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Thursday, April 4, 2024

Muon-catalyzed fusion

From Wikipedia, the free encyclopedia
https://en.wikipedia.org/wiki/Muon-catalyzed_fusion

Muon-catalyzed fusion (abbreviated as μCF or MCF) is a process allowing nuclear fusion to take place at temperatures significantly lower than the temperatures required for thermonuclear fusion, even at room temperature or lower. It is one of the few known ways of catalyzing nuclear fusion reactions.

Muons are unstable subatomic particles which are similar to electrons but 207 times more massive. If a muon replaces one of the electrons in a hydrogen molecule, the nuclei are consequently drawn 186 times closer than in a normal molecule, due to the reduced mass being 186 times the mass of an electron. When the nuclei move closer together, the fusion probability increases, to the point where a significant number of fusion events can happen at room temperature.

Methods for obtaining muons, however, require far more energy than can be produced by the resulting fusion reactions. Muons have a mean lifetime of 2.2 μs, much longer than many other subatomic particles but nevertheless far too brief to allow their useful storage.

To create useful room-temperature muon-catalyzed fusion, reactors would need a cheap, efficient muon source and/or a way for each individual muon to catalyze many more fusion reactions.

History

Andrei Sakharov and F.C. Frank predicted the phenomenon of muon-catalyzed fusion on theoretical grounds before 1950. Yakov Borisovich Zel'dovich also wrote about the phenomenon of muon-catalyzed fusion in 1954. Luis W. Alvarez et al., when analyzing the outcome of some experiments with muons incident on a hydrogen bubble chamber at Berkeley in 1956, observed muon-catalysis of exothermic p–d, proton and deuteron, nuclear fusion, which results in a helion, a gamma ray, and a release of about 5.5 MeV of energy. The Alvarez experimental results, in particular, spurred John David Jackson to publish one of the first comprehensive theoretical studies of muon-catalyzed fusion in his ground-breaking 1957 paper. This paper contained the first serious speculations on useful energy release from muon-catalyzed fusion. Jackson concluded that it would be impractical as an energy source, unless the "alpha-sticking problem" (see below) could be solved, leading potentially to an energetically cheaper and more efficient way of utilizing the catalyzing muons.

Viability as a power source

Potential benefits

If muon-catalyzed d–t nuclear fusion is realized practically, it will be a much more attractive way of generating power than conventional nuclear fission reactors because muon-catalyzed d–t nuclear fusion (like most other types of nuclear fusion), produces far fewer harmful (and far less long-lived) radioactive wastes.

The large number of neutrons produced in muon-catalyzed d–t nuclear fusions may be used to breed fissile fuels from fertile material – for example, thorium-232 could breed uranium-233 in this way. The fissile fuels that have been bred can then be "burned," either in a conventional critical nuclear fission reactor or in an unconventional subcritical fission reactor, for example, a reactor using nuclear transmutation to process nuclear waste, or a reactor using the energy amplifier concept devised by Carlo Rubbia and others.

Another benefit of muon-catalyzed fusion is that the fusion process can start with pure deuterium gas without tritium. Plasma fusion reactors like ITER or Wendelstein X7 need tritium to initiate and also need a tritium factory. Muon-catalyzed fusion generates tritium under operation and increases operating efficiency up to an optimum point when the deuterium:tritium ratio reaches about 1:1. Muon-catalyzed fusion can operate as a tritium factory and deliver tritium for material and plasma fusion research.

Problems facing practical exploitation

Except for some refinements, little has changed since Jackson's 1957 assessment of the feasibility of muon-catalyzed fusion other than Vesman's 1967 prediction of the hyperfine resonant formation of the muonic (d–μ–t)+ molecular ion which was subsequently experimentally observed. This helped spark renewed interest in the whole field of muon-catalyzed fusion, which remains an active area of research worldwide. However, as Jackson observed in his paper, muon-catalyzed fusion is "unlikely" to provide "useful power production ... unless an energetically cheaper way of producing μ-mesons can be found."

One practical problem with the muon-catalyzed fusion process is that muons are unstable, decaying in 2.2 μs (in their rest frame). Hence, there needs to be some cheap means of producing muons, and the muons must be arranged to catalyze as many nuclear fusion reactions as possible before decaying.

Another, and in many ways more serious, problem is the "alpha-sticking" problem, which was recognized by Jackson in his 1957 paper. The α-sticking problem is the approximately 1% probability of the muon "sticking" to the alpha particle that results from deuteron-triton nuclear fusion, thereby effectively removing the muon from the muon-catalysis process altogether. Even if muons were absolutely stable, each muon could catalyze, on average, only about 100 d-t fusions before sticking to an alpha particle, which is only about one-fifth the number of muon catalyzed d–t fusions needed for break-even, where as much thermal energy is generated as electrical energy is consumed to produce the muons in the first place, according to Jackson's rough estimate.

More recent measurements seem to point to more encouraging values for the α-sticking probability, finding the α-sticking probability to be around 0.3% to 0.5%, which could mean as many as about 200 (even up to 350) muon-catalyzed d–t fusions per muon. Indeed, the team led by Steven E. Jones achieved 150 d–t fusions per muon (average) at the Los Alamos Meson Physics Facility. The results were promising and almost enough to reach theoretical break-even. Unfortunately, these measurements for the number of muon-catalyzed d–t fusions per muon are still not enough to reach industrial break-even. Even with break-even, the conversion efficiency from thermal energy to electrical energy is only about 40% or so, further limiting viability. The best recent estimates of the electrical "energy cost" per muon is about 6 GeV with accelerators that are (coincidentally) about 40% efficient at transforming electrical energy from the power grid into acceleration of the deuterons.

As of 2012, no practical method of producing energy through this means has been published, although some discoveries using the Hall effect show promise.

Alternative estimation of breakeven

According to Gordon Pusch, a physicist at Argonne National Laboratory, various breakeven calculations on muon-catalyzed fusion omit the heat energy the muon beam itself deposits in the target. By taking this factor into account, muon-catalyzed fusion can already exceed breakeven; however, the recirculated power is usually very large compared to power out to the electrical grid (about 3–5 times as large, according to estimates). Despite this rather high recirculated power, the overall cycle efficiency is comparable to conventional fission reactors; however the need for 4–6 MW electrical generating capacity for each megawatt out to the grid probably represents an unacceptably large capital investment. Pusch suggested using Bogdan Maglich's "migma" self-colliding beam concept to significantly increase the muon production efficiency, by eliminating target losses, and using tritium nuclei as the driver beam, to optimize the number of negative muons.

In 2021, Kelly, Hart and Rose produced a μCF model whereby the ratio, Q, of thermal energy produced to the kinetic energy of the accelerated deuterons used to create negative pions (and thus negative muons through pion decay) was optimized. In this model, the heat energy of the incoming deuterons as well as that of the particles produced due to the deuteron beam impacting a tungsten target was recaptured to the extent possible, as suggested by Gordon Pusch in the previous paragraph. Additionally, heat energy due to tritium breeding in a lithium-lead shell was recaptured, as suggested by Jändel, Danos and Rafelski in 1988. The best Q value was found to be about 130% assuming that 50% of the muons produced were actually utilized for fusion catalysis. Furthermore, assuming that the accelerator was 18% efficient at transforming electrical energy into deuteron kinetic energy and conversion efficiency of heat energy into electrical energy of 60%, they estimate that, currently, the amount of electrical energy that could be produced by a μCF reactor would be 14% of the electrical energy consumed. In order for this to improve, they suggest that some combination of a) increasing accelerator efficiency and b) increasing the number of fusion reactions per negative muon above the assumed level of 150 would be needed.

Process

To create this effect, a stream of negative muons, most often created by decaying pions, is sent to a block that may be made up of all three hydrogen isotopes (protium, deuterium, and/or tritium), where the block is usually frozen, and the block may be at temperatures of about 3 kelvin (−270 degrees Celsius) or so. The muon may bump the electron from one of the hydrogen isotopes. The muon, 207 times more massive than the electron, effectively shields and reduces the electromagnetic repulsion between two nuclei and draws them much closer into a covalent bond than an electron can. Because the nuclei are so close, the strong nuclear force is able to kick in and bind both nuclei together. They fuse, release the catalytic muon (most of the time), and part of the original mass of both nuclei is released as energetic particles, as with any other type of nuclear fusion. The release of the catalytic muon is critical to continue the reactions. The majority of the muons continue to bond with other hydrogen isotopes and continue fusing nuclei together. However, not all of the muons are recycled: some bond with other debris emitted following the fusion of the nuclei (such as alpha particles and helions), removing the muons from the catalytic process. This gradually chokes off the reactions, as there are fewer and fewer muons with which the nuclei may bond. The number of reactions achieved in the lab can be as high as 150 d–t fusions per muon (average).

Deuterium–tritium (d–t or dt)

In the muon-catalyzed fusion of most interest, a positively charged deuteron (d), a positively charged triton (t), and a muon essentially form a positively charged muonic molecular heavy hydrogen ion (d–μ–t)+. The muon, with a rest mass 207 times greater than the rest mass of an electron, is able to drag the more massive triton and deuteron 207 times closer together to each other in the muonic (d–μ–t)+ molecular ion than can an electron in the corresponding electronic (d–e–t)+ molecular ion. The average separation between the triton and the deuteron in the electronic molecular ion is about one angstrom (100 pm), so the average separation between the triton and the deuteron in the muonic molecular ion is 207 times smaller than that. Due to the strong nuclear force, whenever the triton and the deuteron in the muonic molecular ion happen to get even closer to each other during their periodic vibrational motions, the probability is very greatly enhanced that the positively charged triton and the positively charged deuteron would undergo quantum tunnelling through the repulsive Coulomb barrier that acts to keep them apart. Indeed, the quantum mechanical tunnelling probability depends roughly exponentially on the average separation between the triton and the deuteron, allowing a single muon to catalyze the d–t nuclear fusion in less than about half a picosecond, once the muonic molecular ion is formed.

The formation time of the muonic molecular ion is one of the "rate-limiting steps" in muon-catalyzed fusion that can easily take up to ten thousand or more picoseconds in a liquid molecular deuterium and tritium mixture (D2, DT, T2), for example. Each catalyzing muon thus spends most of its ephemeral existence of 2.2 microseconds, as measured in its rest frame, wandering around looking for suitable deuterons and tritons with which to bind.

Another way of looking at muon-catalyzed fusion is to try to visualize the ground state orbit of a muon around either a deuteron or a triton. Suppose the muon happens to have fallen into an orbit around a deuteron initially, which it has about a 50% chance of doing if there are approximately equal numbers of deuterons and tritons present, forming an electrically neutral muonic deuterium atom (d–μ)0 that acts somewhat like a "fat, heavy neutron" due both to its relatively small size (again, 207 times smaller than an electrically neutral electronic deuterium atom (d–e)0) and to the very effective "shielding" by the muon of the positive charge of the proton in the deuteron. Even so, the muon still has a much greater chance of being transferred to any triton that comes near enough to the muonic deuterium than it does of forming a muonic molecular ion. The electrically neutral muonic tritium atom (t–μ)0 thus formed will act somewhat like an even "fatter, heavier neutron," but it will most likely hang on to its muon, eventually forming a muonic molecular ion, most likely due to the resonant formation of a hyperfine molecular state within an entire deuterium molecule D2 (d=e2=d), with the muonic molecular ion acting as a "fatter, heavier nucleus" of the "fatter, heavier" neutral "muonic/electronic" deuterium molecule ([d–μ–t]=e2=d), as predicted by Vesman, an Estonian graduate student, in 1967.

Once the muonic molecular ion state is formed, the shielding by the muon of the positive charges of the proton of the triton and the proton of the deuteron from each other allows the triton and the deuteron to tunnel through the Coulomb barrier in time span of order of a nanosecond The muon survives the d–t muon-catalyzed nuclear fusion reaction and remains available (usually) to catalyze further d–t muon-catalyzed nuclear fusions. Each exothermic d–t nuclear fusion releases about 17.6 MeV of energy in the form of a "very fast" neutron having a kinetic energy of about 14.1 MeV and an alpha particle α (a helium-4 nucleus) with a kinetic energy of about 3.5 MeV. An additional 4.8 MeV can be gleaned by having the fast neutrons moderated in a suitable "blanket" surrounding the reaction chamber, with the blanket containing lithium-6, whose nuclei, known by some as "lithions," readily and exothermically absorb thermal neutrons, the lithium-6 being transmuted thereby into an alpha particle and a triton.

Deuterium–deuterium and other types

The first kind of muon–catalyzed fusion to be observed experimentally, by L.W. Alvarez et al., was protium (H or 1H1) and deuterium (D or 1H2) muon-catalyzed fusion. The fusion rate for p–d (or pd) muon-catalyzed fusion has been estimated to be about a million times slower than the fusion rate for d–t muon-catalyzed fusion.

Of more practical interest, deuterium–deuterium muon-catalyzed fusion has been frequently observed and extensively studied experimentally, in large part because deuterium already exists in relative abundance and, like protium, deuterium is not at all radioactive. (Tritium rarely occurs naturally, and is radioactive with a half-life of about 12.5 years.)

The fusion rate for d–d muon-catalyzed fusion has been estimated to be only about 1% of the fusion rate for d–t muon-catalyzed fusion, but this still gives about one d–d nuclear fusion every 10 to 100 picoseconds or so. However, the energy released with every d–d muon-catalyzed fusion reaction is only about 20% or so of the energy released with every d–t muon-catalyzed fusion reaction. Moreover, the catalyzing muon has a probability of sticking to at least one of the d–d muon-catalyzed fusion reaction products that Jackson in this 1957 paper estimated to be at least 10 times greater than the corresponding probability of the catalyzing muon sticking to at least one of the d–t muon-catalyzed fusion reaction products, thereby preventing the muon from catalyzing any more nuclear fusions. Effectively, this means that each muon catalyzing d–d muon-catalyzed fusion reactions in pure deuterium is only able to catalyze about one-tenth of the number of d–t muon-catalyzed fusion reactions that each muon is able to catalyze in a mixture of equal amounts of deuterium and tritium, and each d–d fusion only yields about one-fifth of the yield of each d–t fusion, thereby making the prospects for useful energy release from d–d muon-catalyzed fusion at least 50 times worse than the already dim prospects for useful energy release from d–t muon-catalyzed fusion.

Potential "aneutronic" (or substantially aneutronic) nuclear fusion possibilities, which result in essentially no neutrons among the nuclear fusion products, are almost certainly not very amenable to muon-catalyzed fusion. One such essentially aneutronic nuclear fusion reaction involves a deuteron from deuterium fusing with a helion (He+2) from helium-3, which yields an energetic alpha particle and a much more energetic proton, both positively charged (with a few neutrons coming from inevitable d–d nuclear fusion side reactions). However, one muon with only one negative electric charge is incapable of shielding both positive charges of a helion from the one positive charge of a deuteron. The chances of the requisite two muons being present simultaneously are exceptionally remote.

In culture

The term "cold fusion" was coined to refer to muon-catalyzed fusion in a 1956 New York Times article about Luis W. Alvarez's paper.

In 1957 Theodore Sturgeon wrote a novelette, "The Pod in the Barrier", in which humanity has ubiquitous cold fusion reactors that work with muons. The reaction is "When hydrogen one and hydrogen two are in the presence of Mu mesons, they fuse into helium three, with an energy yield in electron volts of 5.4 times ten to the fifth power". Unlike the thermonuclear bomb contained in the Pod (which is used to destroy the Barrier) they can become temporarily disabled by "concentrated disbelief" that muon fusion works.

In Sir Arthur C. Clarke's third novel in the Space Odyssey series, 2061: Odyssey Three, muon-catalyzed fusion is the technology that allows mankind to achieve easy interplanetary travel. The main character, Heywood Floyd, compares Luis Alvarez to Lord Rutherford for underestimating the future potential of their discoveries.

Nuclear fusion–fission hybrid

From Wikipedia, the free encyclopedia

Hybrid nuclear fusion–fission (hybrid nuclear power) is a proposed means of generating power by use of a combination of nuclear fusion and fission processes.

The basic idea is to use high-energy fast neutrons from a fusion reactor to trigger fission in non-fissile fuels like U-238 or Th-232. Each neutron can trigger several fission events, multiplying the energy released by each fusion reaction hundreds of times. As the fission fuel is not fissile, there is no self-sustaining chain reaction from fission. This would not only make fusion designs more economical in power terms, but also be able to burn fuels that were not suitable for use in conventional fission plants, even their nuclear waste.

In general terms, the hybrid is similar in concept to the fast breeder reactor, which uses a compact high-energy fission core in place of the hybrid's fusion core. Another similar concept is the accelerator-driven subcritical reactor, which uses a particle accelerator to provide the neutrons instead of nuclear reactions.

History

The concept dates to the 1950s, and was strongly advocated by Hans Bethe during the 1970s. At that time the first powerful fusion experiments were being built, but it would still be many years before they could be economically competitive. Hybrids were proposed as a way of greatly accelerating their market introduction, producing energy even before the fusion systems reached break-even. However, detailed studies of the economics of the systems suggested they could not compete with existing fission reactors.

The idea was abandoned and lay dormant until the 2000s, when the continued delays in reaching break-even led to a brief revival around 2009. These studies generally concentrated on the nuclear waste disposal aspects of the design, as opposed to the production of energy. The concept has seen cyclical interest since then, based largely on the success or failure of more conventional solutions like the Yucca Mountain nuclear waste repository

Another major design effort for energy production was started at Lawrence Livermore National Laboratory (LLNL) under their LIFE program. Industry input led to the abandonment of the hybrid approach for LIFE, which was then re-designed as a pure-fusion system. LIFE was cancelled when the underlying technology, from the National Ignition Facility, failed to reach its design performance goals.

Apollo Fusion, a company founded by Google executive Mike Cassidy in 2017, was also reported to be focused on using the subcritical nuclear fusion-fission hybrid method. Their web site is now focussed on their hall effect thrusters, and mentions fusion only in passing.

On 2022, September 9, Professor Peng Xianjue of the Chinese Academy of Engineering Physics announced that the Chinese government had approved the construction of the world's largest pulsed-powerplant - the Z-FFR, namely Z(-pinch)-Fission-Fusion Reactor- in Chengdu, Sichuan province. Neutrons produced in a Z-pinch facility (endowed with cylindrical symmetry and fuelled with deuterium and tritium) will strike a coaxial blanket including both uranium and lithium isotopes. Uranium fission will boost the facility's overall heat output by 10 to 20 times. Interaction of lithium and neutrons will provide tritium for further fueling. Innovative, quasi-spherical geometry near the core of Z-FFR leads to high performance of Z-pinch discharge. According to Prof. Xianjue, this will considerably speed up the use of fusion energy and prepare it for commercial power production by 2035.

Fission basics

Conventional fission power systems rely on a chain reaction of nuclear fission events that release a few neutrons that cause further fission events. By careful arrangement and the use of various absorber materials the system can be set in a balance of released and absorbed neutrons, known as criticality.

Natural uranium is a mix of several isotopes, mainly a trace amount of 235U and over 99% 238U. When they undergo fission, both of these isotopes release fast neutrons with an energy distribution peaking around 1 to 2 MeV. This energy is too low to cause fission in 238U, which means it cannot sustain a chain reaction. 235U will undergo fission when struck by neutrons of this energy, so it is possible for 235U to sustain a chain reaction. There are too few 235U atoms in natural uranium to sustain a chain reaction, the atoms are spread out too far and the chance a neutron will hit one is too small. Chain reactions are accomplished by concentrating, or enriching, the fuel, increasing the amount of 235U to produce enriched uranium, while the leftover, now mostly 238U, is a waste product known as depleted uranium. 235U will sustain a chain reaction if enriched to about 20% of the fuel mass.

235U will undergo fission more easily if the neutrons are of lower energy, the so-called thermal neutrons. Neutrons can be slowed to thermal energies through collisions with a neutron moderator material, the easiest to use are the hydrogen atoms found in water. By placing the fission fuel in water, the probability that the neutrons will cause fission in another 235U is greatly increased, which means the level of enrichment needed to reach criticality is greatly reduced. This leads to the concept of reactor-grade enriched uranium, with the amount of 235U increased from just less than 1% in natural ore to between 3 and 5%, depending on the reactor design. This is in contrast to weapons-grade enrichment, which increases to the 235U to at least 20%, and more commonly, over 90%.

In order to maintain criticality, the fuel has to retain that extra concentration of 235U. A typical fission reactor burns off enough of the 235U to cause the reaction to stop over a period on the order of a few months. A combination of burnup of the 235U along with the creation of neutron absorbers, or poisons, as part of the fission process eventually results in the fuel mass not being able to maintain criticality. This burned up fuel has to be removed and replaced with fresh fuel. The result is nuclear waste that is highly radioactive and filled with long-lived radionuclides that present a safety concern.

The waste contains most of the 235U it started with, only 1% or so of the energy in the fuel is extracted by the time it reaches the point where it is no longer fissile. One solution to this problem is to reprocess the fuel, which uses chemical processes to separate the 235U (and other non-poison elements) from the waste, and then mixes the extracted 235U5 in fresh fuel loads. This reduces the amount of new fuel that needs to be mined and also concentrates the unwanted portions of the waste into a smaller load. Reprocessing is expensive, however, and it has generally been more economical to simply buy fresh fuel from the mine.

Like 235U, 239Pu can maintain a chain reaction, so it is a useful reactor fuel. However, 239Pu is not found in commercially useful amounts in nature. Another possibility is to breed 239Pu from the 238U through neutron capture, or various other means. This process only occurs with higher-energy neutrons than would be found in a moderated reactor, so a conventional reactor only produces small amounts of Pu when the neutron is captured within the fuel mass before it is moderated.

More typically, special reactors are used that are designed specifically for the breeding of 239Pu. The simplest way to achieve this is to further enrich the original 235U fuel well beyond what is needed for use in a moderated reactor, to the point where the 235U maintains criticality even with the fast neutrons. The extra fast neutrons escaping the fuel load can then be used to breed fuel in a 238U assembly surrounding the reactor core, most commonly taken from the stocks of depleted uranium. 239Pu can also be used for the core, which means once the system is up and running, it can be refuelled using the 239Pu it creates, with enough left over to feed into other reactors as well.

Extracting the 239Pu from the 238U feedstock can be achieved with chemical processing, in the same fashion as normal reprocessing. The difference is that the mass will contain far fewer other elements, particularly some of the highly radioactive fission products found in normal nuclear waste. Unfortunately it is a tendency that breeder reactors in the "free world" (like the SNR-300, the Integral fast reactor) that have been built were demolished before operation, as a "symbol" (as Bill Clinton has stated). The Prototype Fast Breeder Reactor passed tests in 2017 and apparently is about to face the same fate, leaving some military reactors and the Russian BN-800 reactor operating, mostly consuming spent nuclear fuel.

Fusion basics

Fusion reactors typically burn a mixture of deuterium (D) and tritium (T). When heated to millions of degrees, the kinetic energy in the fuel begins to overcome the natural electrostatic repulsion between nuclei, the so-called coulomb barrier, and the fuel begins to undergo fusion. This reaction gives off an alpha particle and a high energy neutron of 14 MeV. A key requirement to the economic operation of a fusion reactor is that the alphas deposit their energy back into the fuel mix, heating it so that additional fusion reactions take place. This leads to a condition not unlike the chain reaction in the fission case, known as ignition.

Deuterium can be obtained by the separation of hydrogen isotopes in sea water (see heavy water production). Tritium has a short half life of just over a decade, so only trace amounts are found in nature. To fuel the reactor, the neutrons from the reaction are used to breed more tritium through a reaction in a blanket of lithium surrounding the reaction chamber. Tritium breeding is key to the success of a D-T fusion cycle, and to date this technique has not been demonstrated. Predictions based on computer modeling suggests that the breeding ratios are quite small and a fusion plant would barely be able to cover its own use. Many years would be needed to breed enough surplus to start another reactor.

Hybrid concepts

Fusion–fission designs essentially replace the lithium blanket with a blanket of fission fuel, either natural uranium ore or even nuclear waste. The fusion neutrons have more than enough energy to cause fission in the 238U, as well as many of the other elements in the fuel, including some of the transuranic waste elements. The reaction can continue even when all of the 235U is burned off; the rate is controlled not by the neutrons from the fission events, but the neutrons being supplied by the fusion reactor.

Fission occurs naturally because each event gives off more than one neutron capable of producing additional fission events. Fusion, at least in D-T fuel, gives off only a single neutron, and that neutron is not capable of producing more fusion events. When that neutron strikes fissile material in the blanket, one of two reactions may occur. In many cases, the kinetic energy of the neutron will cause one or two neutrons to be struck out of the nucleus without causing fission. These neutrons still have enough energy to cause other fission events. In other cases the neutron will be captured and cause fission, which will release two or three neutrons. This means that every fusion neutron in the fusion–fission design can result in anywhere between two and four neutrons in the fission fuel.

This is a key concept in the hybrid concept, known as fission multiplication. For every fusion event, several fission events may occur, each of which gives off much more energy than the original fusion, about 11 times. This greatly increases the total power output of the reactor. This has been suggested as a way to produce practical fusion reactors in spite of the fact that no fusion reactor has yet reached break-even, by multiplying the power output using cheap fuel or waste. However, a number of studies have repeatedly demonstrated that this only becomes practical when the overall reactor is very large, 2 to 3 GWt, which makes it expensive to build.

These processes also have the side-effect of breeding 239Pu or 233U, which can be removed and used as fuel in conventional fission reactors. This leads to an alternate design where the primary purpose of the fusion–fission reactor is to reprocess waste into new fuel. Although far less economical than chemical reprocessing, this process also burns off some of the nastier elements instead of simply physically separating them out. This also has advantages for non-proliferation, as enrichment and reprocessing technologies are also associated with nuclear weapons production. However, the cost of the nuclear fuel produced is very high, and is unlikely to be able to compete with conventional sources.

Neutron economy

A key issue for the fusion–fission concept is the number and lifetime of the neutrons in the various processes, the so-called neutron economy.

In a pure fusion design, the neutrons are used for breeding tritium in a lithium blanket. Natural lithium consists of about 92% 7Li and the rest is mostly 6Li. 7Li breeding requires neutron energies even higher than those released by fission, around 5 MeV, well within the range of energies provided by fusion. This reaction produces tritium and helium-4, and another slow neutron. 6Li can react with high or low energy neutrons, including those released by the 7Li reaction. This means that a single fusion reaction can produce several tritiums, which is a requirement if the reactor is going to make up for natural decay and losses in the fusion processes.

When the lithium blanket is replaced, or supplanted, by fission fuel in the hybrid design, neutrons that do react with the fissile material are no longer available for tritium breeding. The new neutrons released from the fission reactions can be used for this purpose, but only in 6Li. One could process the lithium to increase the amount of 6Li in the blanket, making up for these losses, but the downside to this process is that the 6Li reaction only produces one tritium atom. Only the high-energy reaction between the fusion neutron and 7Li can create more than one tritium, and this is essential for keeping the reactor running.

To address this issue, at least some of the fission neutrons must also be used for tritium breeding in 6Li. Every one that does is no longer available for fission, reducing the reactor output. This requires a very careful balance if one wants the reactor to be able to produce enough tritium to keep itself running, while also producing enough fission events to keep the fission side energy positive. If these cannot be accomplished simultaneously, there is no reason to build a hybrid. Even if this balance can be maintained, it might only occur at a level that is economically infeasible.

Overall economy

Through the early development of the hybrid concept the question of overall economics appeared difficult to handle. A series of studies starting in the late 1970s provided a much clearer picture of the hybrid in a complete fuel cycle, and allowed the economics to be better understood. These studies appeared to indicate there was no reason to build a hybrid.

One of the most detailed of these studies was published in 1980 by Los Alamos National Laboratory (LANL). Their study noted that the hybrid would produce most of its energy indirectly, both through the fission events in its own reactor, and much more by providing Pu-239 to fuel conventional fission reactors. In this overall picture, the hybrid is essentially identical to the breeder reactor, which uses fast neutrons from plutonium fission to breed more fuel in a fission blanket in largely the same fashion as the hybrid. Both require chemical processing to remove the bred Pu-239, both presented the same proliferation and safety risks as a result, and both produced about the same amount of fuel. Since that fuel is the primary source of energy in the overall cycle, the two systems were almost identical in the end.

What was not identical, however, was the technical maturity of the two designs. The hybrid would require considerable additional research and development before it would be known if it could even work, and even if that were demonstrated, the result would be a system essentially identical to breeders which were already being built at that time. The report concluded:

The investment of time and money required to commercialize the hybrid cycle could only be justified by a real or perceived advantage of the hybrid over the classical FBR. Our analysis leads us to conclude that no such advantage exists. Therefore, there is not sufficient incentive to demonstrate and commercialize the fusion–fission hybrid.

Rationale

The fusion process alone currently does not achieve sufficient gain (power output over power input) to be viable as a power source. By using the excess neutrons from the fusion reaction to in turn cause a high-yield fission reaction (close to 100%) in the surrounding subcritical fissionable blanket, the net yield from the hybrid fusion–fission process can provide a targeted gain of 100 to 300 times the input energy (an increase by a factor of three or four over fusion alone). Even allowing for high inefficiencies on the input side (i.e. low laser efficiency in ICF and Bremsstrahlung losses in Tokamak designs), this can still yield sufficient heat output for economical electric power generation. This can be seen as a shortcut to viable fusion power until more efficient pure fusion technologies can be developed, or as an end in itself to generate power, and also consume existing stockpiles of nuclear fissionables and waste products.

In the LIFE project at the Lawrence Livermore National Laboratory LLNL, using technology developed at the National Ignition Facility, the goal is to use fuel pellets of deuterium and tritium surrounded by a fissionable blanket to produce energy sufficiently greater than the input (laser) energy for electrical power generation. The principle involved is to induce inertial confinement fusion (ICF) in the fuel pellet which acts as a highly concentrated point source of neutrons which in turn converts and fissions the outer fissionable blanket. In parallel with the ICF approach, the University of Texas at Austin is developing a system based on the tokamak fusion reactor, optimising for nuclear waste disposal versus power generation. The principles behind using either ICF or tokamak reactors as a neutron source are essentially the same (the primary difference being that ICF is essentially a point-source of neutrons while Tokamaks are more diffuse toroidal sources).

Use to dispose of nuclear waste

The surrounding blanket can be a fissile material (enriched uranium or plutonium) or a fertile material (capable of conversion to a fissionable material by neutron bombardment) such as thorium, depleted uranium or spent nuclear fuel. Such subcritical reactors (which also include particle accelerator-driven neutron spallation systems) offer the only currently-known means of active disposal (versus storage) of spent nuclear fuel without reprocessing. Fission by-products produced by the operation of commercial light water nuclear reactors (LWRs) are long-lived and highly radioactive, but they can be consumed using the excess neutrons in the fusion reaction along with the fissionable components in the blanket, essentially destroying them by nuclear transmutation and producing a waste product which is far safer and less of a risk for nuclear proliferation. The waste would contain significantly reduced concentrations of long-lived, weapons-usable actinides per gigawatt-year of electric energy produced compared to the waste from a LWR. In addition, there would be about 20 times less waste per unit of electricity produced. This offers the potential to efficiently use the very large stockpiles of enriched fissile materials, depleted uranium, and spent nuclear fuel.

Safety

In contrast to current commercial fission reactors, hybrid reactors potentially demonstrate what is considered inherently safe behavior because they remain deeply subcritical under all conditions and decay heat removal is possible via passive mechanisms. The fission is driven by neutrons provided by fusion ignition events, and is consequently not self-sustaining. If the fusion process is deliberately shut off or the process is disrupted by a mechanical failure, the fission damps out and stops nearly instantly. This is in contrast to the forced damping in a conventional reactor by means of control rods which absorb neutrons to reduce the neutron flux below the critical, self-sustaining, level. The inherent danger of a conventional fission reactor is any situation leading to a positive feedback, runaway, chain reaction such as occurred during the Chernobyl disaster. In a hybrid configuration the fission and fusion reactions are decoupled, i.e. while the fusion neutron output drives the fission, the fission output has no effect whatsoever on the fusion reaction, eliminating any chance of a positive feedback loop.

Fuel cycle

There are three main components to the hybrid fusion fuel cycle: deuterium, tritium, and fissionable elements. Deuterium can be derived by the separation of hydrogen isotopes in seawater (see heavy water production). Tritium may be generated in the hybrid process itself by absorption of neutrons in lithium bearing compounds. This would entail an additional lithium-bearing blanket and a means of collection. Small amounts of tritium are also produced by neutron activation in nuclear fission reactors, particularly when heavy water is used as a neutron moderator or coolant. The third component is externally derived fissionable materials from demilitarized supplies of fissionables, or commercial nuclear fuel and waste streams. Fusion driven fission also offers the possibility of using thorium as a fuel, which would greatly increase the potential amount of fissionables available. The extremely energetic nature of the fast neutrons emitted during the fusion events (up to 0.17 the speed of light) can allow normally non-fissioning 238U to undergo fission directly (without conversion first to 239Pu), enabling refined natural Uranium to be used with very low enrichment, while still maintaining a deeply subcritical regime.

Engineering considerations

Practical engineering designs must first take into account safety as the primary goal. All designs should incorporate passive cooling in combination with refractory materials to prevent melting and reconfiguration of fissionables into geometries capable of un-intentional criticality. Blanket layers of Lithium bearing compounds will generally be included as part of the design to generate Tritium to allow the system to be self-supporting for one of the key fuel element components. Tritium, because of its relatively short half-life and extremely high radioactivity, is best generated on-site to obviate the necessity of transportation from a remote location. D-T fuel can be manufactured on-site using Deuterium derived from heavy water production and Tritium generated in the hybrid reactor itself. Nuclear spallation to generate additional neutrons can be used to enhance the fission output, with the caveat that this is a tradeoff between the number of neutrons (typically 20-30 neutrons per spallation event) against a reduction of the individual energy of each neutron. This is a consideration if the reactor is to use natural Thorium as a fuel. While high energy (0.17c) neutrons produced from fusion events are capable of directly causing fission in both Thorium and 238U, the lower energy neutrons produced by spallation generally cannot. This is a tradeoff that affects the mixture of fuels against the degree of spallation used in the design.

Wednesday, April 3, 2024

India's three-stage nuclear power programme

Monazite powder, a rare earth and thorium phosphate mineral, is the primary source of the world's thorium

India's three-stage nuclear power programme was formulated by Homi Bhabha, the well-known physicist, in the 1950s to secure the country's long term energy independence, through the use of uranium and thorium reserves found in the monazite sands of coastal regions of South India. The ultimate focus of the programme is on enabling the thorium reserves of India to be utilised in meeting the country's energy requirements. Thorium is particularly attractive for India, as India has only around 1–2% of the global uranium reserves, but one of the largest shares of global thorium reserves at about 25% of the world's known thorium reserves. However, thorium is more difficult to use than uranium as a fuel because it requires breeding, and global uranium prices remain low enough that breeding is not cost effective.

India published about twice the number of papers on thorium as its nearest competitors, during each of the years from 2002 to 2006. The Indian nuclear establishment estimates that the country could produce 500 GWe for at least four centuries using just the country's economically extractable thorium reserves.

The first Prototype Fast Breeder Reactor has been repeatedly delayed – and is currently expected to be commissioned by October 2022  – and India continues to import thousands of tonnes of uranium from Russia, Kazakhstan, France, and Uzbekistan. The 2005 Indo–US Nuclear Deal and the NSG waiver, which ended more than three decades of international isolation of the Indian civil nuclear programme, have created many hitherto unexplored alternatives for the success of the three-stage nuclear power programme.

Origin and rationale

Homi Jehangir Bhabha, the founding Chairman of India's Atomic Energy Commission and the architect of Indian three-stage (thorium) programme

Homi Bhabha conceived of the three-stage nuclear programme as a way to develop nuclear energy by working around India's limited uranium resources. Thorium itself is not a fissile material, and thus cannot undergo fission to produce energy. Instead, it must be transmuted to uranium-233 in a reactor fueled by other fissile materials. The first two stages, natural uranium-fueled heavy water reactors and plutonium-fueled fast breeder reactors, are intended to generate sufficient fissile material from India's limited uranium resources, so that all its vast thorium reserves can be fully utilised in the third stage of thermal breeder reactors.

Bhabha summarised the rationale for the three-stage approach as follows:

The total reserves of thorium in India amount to over 500,000 tons in the readily extractable form, while the known reserves of uranium are less than a tenth of this. The aim of long range atomic power programme in India must therefore be to base the nuclear power generation as soon as possible on thorium rather than uranium… The first generation of atomic power stations based on natural uranium can only be used to start off an atomic power programme… The plutonium produced by the first generation power stations can be used in a second generation of power stations designed to produce electric power and convert thorium into U-233, or depleted uranium into more plutonium with breeding gain… The second generation of power stations may be regarded as an intermediate step for the breeder power stations of the third generation all of which would produce more U-233 than they burn in the course of producing power.

In November 1954, Bhabha presented the three-stage plan for national development, at the conference on "Development of Atomic Energy for Peaceful Purposes" which was also attended by India's first Prime Minister Jawaharlal Nehru. Four years later in 1958, the Indian government formally adopted the three-stage plan. Indian energy resource base was estimated to be capable of yielding a total electric power output of the order shown in the table below. Indian government recognised that thorium was a source that could provide power to the Indian people for the long term.

Energy resource type Amount (tonnes) Power potential (TWe-year)
Coal 54 billion 11
Hydrocarbons 12 billion 6
Uranium (in PHWR) 61,000 0.3–0.42
Uranium (in FBR) 61,000 16–54
Thorium ~300,000 155–168 or 358

Fuel reserves and research capability

According to a report issued by the IAEA, India has limited uranium reserves, consisting of approximately 54,636 tonnes of "reasonably assured resources", 25,245 tonnes of "estimated additional resources", 15,488 tonnes of "undiscovered conventional resources, and 17,000 tonnes of "speculative resources". According to NPCIL, these reserves are only sufficient to generate about 10 GWe for about 40 years. In July 2011, it was reported that a four-year-long mining survey done at Tummalapalle mine in Kadapa district near Hyderabad had yielded confirmed reserve figure of 49,000 tonnes with a potential that it could rise to 150,000 tonnes. This was a rise from an earlier estimate of 15,000 tonnes for that area.

Although India has only around 1–2% of the global uranium reserves, thorium reserves are bigger; around 12–33% of global reserves, according to IAEA and US Geological Survey. Several in-depth independent studies put Indian thorium reserves at 30% of the total world thorium reserves. Indian uranium production is constrained by government investment decisions rather than by any shortage of ore.

As per official estimates shared in the country's Parliament in August 2011, the country can obtain 846,477 tonnes of thorium from 963,000 tonnes of ThO2, which in turn can be obtained from 10.7 million tonnes of monazite occurring in beaches and river sands in association with other heavy metals. Indian monazite contains about 9–10% ThO2. The 846,477 tonne figure compares with the earlier estimates for India, made by IAEA and US Geological Survey of 319,000 tonnes and 290,000 to 650,000 tonnes respectively. The 800,000 tonne figure is given by other sources as well.

It was further clarified in the country's parliament on 21 March 2012 that, "Out of nearly 100 deposits of the heavy minerals, at present only 17 deposits containing about 4 million tonnes of monazite have been identified as exploitable. Mine-able reserves are ~70% of identified exploitable resources. Therefore, about 225,000 tonnes of thorium metal is available for nuclear power program."

India is a leader of thorium based research. It is also by far the most committed nation as far as the use of thorium fuel is concerned, and no other country has done as much neutron physics work on thorium. The country published about twice the number of papers on thorium as its nearest competitors during each of the years from 2002 to 2006. Bhabha Atomic Research Centre (BARC) had the highest number of publications in the thorium area, across all research institutions in the world during the period 1982–2004. During this same period, India ranks an overall second behind the United States in the research output on Thorium. According to Siegfried Hecker, a former director (1986–1997) of the Los Alamos National Laboratory in the United States, "India has the most technically ambitious and innovative nuclear energy programme in the world. The extent and functionality of its nuclear experimental facilities are matched only by those in Russia and are far ahead of what is left in the US."

However, conventional uranium-fueled reactors are much cheaper to operate; so India imports large quantities of uranium from abroad. Also, in March 2011, large deposits of uranium were discovered in the Tummalapalle belt in the southern part of the Kadapa basin in Andhra Pradesh.

Stage I – Pressurised Heavy Water Reactor

The Narora Atomic Power Station has two IPHWR reactors, the first stage of the three stage program

In the first stage of the programme, natural uranium fueled pressurised heavy water reactors (PHWR) produce electricity while generating plutonium-239 as by-product. PHWRs was a natural choice for implementing the first stage because it had the most efficient reactor design in terms of uranium utilisation, and the existing Indian infrastructure in the 1960s allowed for quick adoption of the PHWR technology. India correctly calculated that it would be easier to create heavy water production facilities (required for PHWRs) than uranium enrichment facilities (required for LWRs). Natural uranium contains only 0.7% of the fissile isotope uranium-235. Most of the remaining 99.3% is uranium-238 which is not fissile but can be converted in a reactor to the fissile isotope plutonium-239. Heavy water (deuterium oxide, D2O) is used as moderator and coolant. Since the program began, India has developed a series of sequentially larger PHWR's under the IPHWR series derived from the original Canadian supplied CANDU reactors. The IPHWR series consists of three designs of 220 MWe, 540 MWe and 700 MWe capacity under the designations IPHWR-220, IPHWR-540 and IPHWR-700 respectively.

Indian uranium reserves are capable of generating a total power capacity of 420 GWe-years, but the Indian government limited the number of PHWRs fueled exclusively by indigenous uranium reserves, in an attempt to ensure that existing plants get a lifetime supply of uranium. US analysts calculate this limit as being slightly over 13 GW in capacity. Several other sources estimate that the known reserves of natural uranium in the country permit only about 10 GW of capacity to be built through indigenously fueled PHWRs. The three-stage programme explicitly incorporates this limit as the upper cut off of the first stage, beyond which PHWRs are not planned to be built.

Almost the entire existing base of Indian nuclear power (4780 MW) is composed of first stage PHWRs of the IPHWR series, with the exception of the two Boiling Water Reactor (BWR) units at Tarapur. The installed capacity of Kaiga station is now 880 MW consisting of four 220 MWe IPHWR-220 reactors, making it the third largest after Tarapur (1400 MW) (2 x BWR Mark-1, 2 x IPHWR-540) and Rawatbhata (1180 MW) (2 x CANDU, 2 x IPHWR-220). The remaining three power stations at Kakrapar, Kalpakkam and Narora all have 2 units of 220 MWe, thus contributing 440 MW each to the grid. The 2 units of 700 MWe each (IPHWR-700) that are under construction at both Kakraparand Rawatbhata, and the one planned for Banswara would also come under the first stage of the programme, totalling a further addition of 4200 MW. These additions will bring the total power capacity from the first stage PHWRs to near the total planned capacity of 10 GW called for by the three-stage power programme.

Capital costs of PHWRs is in the range of Rs. 6 to 7 crore ($1.2 to $1.4 million) per MW, coupled with a designed plant life of 40 years. Time required for construction has improved over time and is now at about five years. Tariffs of the operating plants are in the range of Rs. 1.75 to 2.80 per unit, depending on the life of the reactor. In the year 2007–08 the average tariff was Rs. 2.28.

India is also working on the design of reactors based on the more efficient Pressurized Water Reactor technology derived from the work on the Arihant-class submarine program to develop a 900 MWe IPWR-900 reactor platform to supplement the currently deployed PHWR's of the IPHWR series.

Stage II – Fast Breeder Reactor

In the second stage, fast breeder reactors (FBRs) would use a mixed oxide (MOX) fuel made from plutonium-239, recovered by reprocessing spent fuel from the first stage, and natural uranium. In FBRs, plutonium-239 undergoes fission to produce energy, while the uranium-238 present in the mixed oxide fuel transmutes to additional plutonium-239. Thus, the Stage II FBRs are designed to "breed" more fuel than they consume. Once the inventory of plutonium-239 is built up thorium can be introduced as a blanket material in the reactor and transmuted to uranium-233 for use in the third stage.

The surplus plutonium bred in each fast reactor can be used to set up more such reactors, and might thus grow the Indian civil nuclear power capacity till the point where the third stage reactors using thorium as fuel can be brought online, which is forecasted as being possible once 50 GW of nuclear power capacity has been achieved. The uranium in the first stage PHWRs that yield 29 EJ of energy in the once-through fuel cycle, can be made to yield between 65 and 128 times more energy through multiple cycles in fast breeder reactors.

The design of the country's first fast breeder, called Prototype Fast Breeder Reactor (PFBR), was done by Indira Gandhi Centre for Atomic Research (IGCAR). Bharatiya Nabhikiya Vidyut Nigam Ltd (Bhavini), a public sector company under the Department of Atomic Energy (DAE), has been given the responsibility to build the fast breeder reactors in India. The construction of this PFBR at Kalpakkam was due to be completed in 2012. It is not yet complete. The date of commission has been delayed to October 2022 from the previous date in 2019.

Doubling time

Doubling time refers to the time required to extract as output, double the amount of fissile fuel, which was fed as input into the breeder reactors. This metric is critical for understanding the time durations that are unavoidable while transitioning from the second stage to the third stage of Bhabha's plan, because building up a sufficiently large fissile stock is essential to the large deployment of the third stage. In Bhabha's 1958 papers on role of thorium, he pictured a doubling time of 5–6 years for breeding U-233 in the Th–U233 cycle. This estimate has now been revised to 70 years due to technical difficulties that were unforeseen at the time. Despite such setbacks, according to publications done by DAE scientists, the doubling time of fissile material in the fast breeder reactors can be brought down to about 10 years by choosing appropriate technologies with short doubling time.

Fuel Type U238–Pu cycle Th–U233 cycle
oxide 17.8 108
carbide-Lee 10 50
metal 8.5 75.1
carbide 10.2 70

Another report prepared for U.S. Department of Energy suggests a doubling time of 22 years for oxide fuel, 13 years for carbide fuel and 10 years for metal fuel.

Stage III – Thorium Based Reactors

A sample of thorium

A Stage III reactor or an Advanced nuclear power system involves a self-sustaining series of thorium-232uranium-233 fuelled reactors. This would be a thermal breeder reactor, which in principle can be refueled – after its initial fuel charge – using only naturally occurring thorium. According to the three-stage programme, Indian nuclear energy could grow to about 10 GW through PHWRs fueled by domestic uranium, and the growth above that would have to come from FBRs till about 50GW. The third stage is to be deployed only after this capacity has been achieved.

According to replies given in Q&A in the Indian Parliament on two separate occasions, 19 August 2010 and 21 March 2012, large scale thorium deployment is only to be expected "3–4 decades after the commercial operation of fast breeder reactors with short doubling time". Full exploitation of India's domestic thorium reserves will likely not occur until after the year 2050.

Parallel approaches

As there is a long delay before direct thorium utilisation in the three-stage programme, the country is looking at reactor designs that allow more direct use of thorium in parallel with the sequential three-stage programme. Three options under consideration are the Indian Accelerator Driven Systems (IADS), Advanced Heavy Water Reactor (AHWR) and Compact High Temperature Reactor. Molten Salt Reactor may also be under consideration based on some recent reports and is under development.

Advanced Heavy Water Reactor (AHWR)

Of the options, the design for AHWR is ready for deployment. AHWR is a 300 MWe vertical pressure tube type, boiling light water cooled and heavy water moderated reactor, using uranium233–thorium MOX and plutonium–thorium MOX. It is expected to generate 65% of its power from thorium and can also be configured to accept other fuel types in full core including enriched uranium and uranium–plutonium MOX. There was a plan for constructing such an AHWR with a plutonium–thorium core combination in 2007. This AHWR design was sent for an independent pre-licensing design safety review by the Atomic Energy Regulatory Board (AERB), the results of which were deemed satisfactory. AHWR would offer very little growth for the fuel build up that is essential for wide deployment of the third stage, and perhaps the impact on the accumulated fissile material could even be negative.

The AHWR design that will be taken up for construction is to be fueled with 20% low enriched uranium (LEU) and 80% thorium. The low enriched uranium (LEU) for this AHWR design is readily available on the world market. As of November 2011, construction will start after the site is identified in 6 months time. It will take another 18 months to get clearances on regulatory and environmental grounds. Construction is estimated to take six years. If everything goes according to plan, AHWR could be operational in India by 2020. In Aug 2017 the AHWR location was still not announced.

Accelerator Driven System

India's Department of Atomic Energy and US's Fermilab are designing unique first-of-its-kind accelerator driven systems. No country has yet built an Accelerator Driven System for power generation. Dr Anil Kakodkar, former chairman of the Atomic Energy Commission called this a mega science project and a "necessity" for humankind.

Indian Molten Salt Breeder Reactor (IMSBR)

The Indian Molten Salt Breeder Reactor (IMSBR) is under development. Studies on conceptual design of the Indian Molten Salt Breeder Reactors (IMSBR) have been initiated.

Linkages with the Indo–US nuclear deal

U.S. President George W. Bush and India's Prime Minister Manmohan Singh exchange greetings in New Delhi on 2 March 2006

In spite of the overall adequacy of its uranium reserves, Indian power plants could not get the necessary amount of uranium to function at full capacity in the late 2000s, primarily due to inadequate investments made in the uranium mining and milling capacity resulting from fiscal austerity in the early 1990s. One study done for U.S. Congress in that time period reaches the conclusion, "India’s current fuel situation means that New Delhi cannot produce sufficient fuel for both its nuclear weapons programme and its projected civil nuclear programme." An independent study arrives at roughly the same conclusion, "India’s current uranium production of less than 300 tons/year can meet at most, two-thirds of its needs for civil and military nuclear fuel." This uranium shortfall during the deal negotiations was understood by both players to be a temporary aberration that was poised to be resolved with requisite investments in India's uranium milling infrastructure.

Drivers for the deal from the Indian side

It was estimated that after attaining 21 GW from nuclear power by 2020, further growth might require imported uranium. This is problematic because deployment of third stage requires that 50 GW be already established through the first and second stages. If imported uranium was made available, Department of Atomic Energy (DAE) estimated that India could reach 70 GW by 2032 and 275 GW by 2052. In such a scenario, the third stage could be made operational following the fast breeder implementation, and nuclear power capacity could grow to 530 GW. The estimated stagnation of the nuclear power at about 21GW by 2020 is likely due to the fact that even the short "doubling time" of the breeder reactors is quite slow, on the order of 10–15 years. Implementing the three-stage programme using the domestic uranium resources alone is feasible, but requires several decades to come to fruition. Imports of fissile material from outside would considerably speed up the programme.

As per research data, the U238–Pu cycle has the shortest doubling time by a large margin, and that technology's compounded yearly fissile material growth rate has been calculated as follows, after making some basic assumptions about the operating features of the fast breeder reactors.

Type Fissile Material Growth %
oxide 1.73%
carbide-Lee 2.31%
metal 4.08%
carbide 3.15%

Indian power generation capacity has grown at 5.9% per annum in the 25-year period prior to 2006. If Indian economy is to grow at 8–9% for the next 25-year period of 2006 to 2032, total power generation capacity has to increase at 6–7% per annum. As the fissile material growth rate does not meet this objective, it becomes necessary to look at alternative approaches for obtaining the fissile material. This conclusion is mostly independent of future technical breakthroughs, and complementary to the eventual implementation of the three-stage approach. It was realised that the best way to get access to the requisite fissile material would be through uranium imports, which was not possible without ending India's nuclear isolation by U.S. and the NSG.

U.S. analyst Ashley J. Tellis argues that the Indo–US nuclear deal is attractive to India because it gives it access to far more options on its civil nuclear programme than would otherwise be the case, primarily by ending its isolation from the international nuclear community. These options include access to latest technologies, access to higher unit output reactors which are more economical, access to global finance for building reactors, ability to export its indigenous small reactor size PHWRs, better information flow for its research community, etc. Finally, the deal also gives India two options that are relatively independent from the three-stage programme, at least in terms of their dependencies on success or failure. The first option is that, India can opt to stay with the first stage reactors as long as the global supply of uranium lasts. The plus side of this is that it covers any risk from short term delays or failures in implementing the three-stage programme. On the negative side, this is an option that is antithetical to the underlying objective of energy independence through the exploitation of thorium.

The second option, and perhaps the more interesting one, is that India can choose to access the third stage of thorium reactors by skipping the more difficult second stage of the plan through some appropriately selected parallel approach such as the high-temperature gas-cooled reactor, the molten salt reactor, or the various accelerator driven systems.

Stakeholder views on the linkages

United States Secretary of State Condoleezza Rice and Indian External Affairs Minister Pranab Mukherjee, after signing the 123 Agreement in Washington, D.C., on 10 October 2008

Indian commentators welcomed the opportunity simply because they could see that India would be able end its international isolation on the nuclear front and obtain a de facto acknowledgement of it as a nuclear weapon state to some degree, in addition to it being able to obtain the uranium that would increase the success potential of its three-stage programme as well as its efforts to build a "minimum credible nuclear deterrent". It was estimated that the power produced by imported reactors could be 50% more expensive than the country's existing nuclear power cost. However, this was perceived as a minor point in the larger context of the deal. In a U.S. Senate Foreign Relations Committee hearing, Under Secretary for Political Affairs Nicholas Burns' prepared remarks stated that "India had made this the central issue in the new partnership developing between our countries". Indian government proceeded to negotiate and execute the Indo–US Nuclear Deal, which then paved the way for the NSG waiver on international uranium imports to India in 2008.

According to one foreign analyst, the deal could "over time… result in India being weaned away from its… three-phase nuclear program involving FBRs and advanced PHWRs. This would occur should India become confident that it would have assured supplies of relatively cheap natural uranium, including from Australia. Of course, nobody in the Indian nuclear establishment would yet admit to that possibility."

Anil Kakodkar, then Chairman of the Atomic Energy Commission, went to the extent of making public, the milder position of keeping the country's indigenous fast breeder programme out of the ambit of international safeguards, saying "in the long run, the energy that will come out from the nuclear fuel resources available in India (from domestic uranium and thorium mines) should always form the larger share of the nuclear energy programme..." and "our strategy should be such that the integrity and autonomy of our being able to develop the three-stage nuclear power programme, be maintained, we cannot compromise that." The full demand of the Indian scientists, to have the ability to reprocess plutonium from spent fuel of the imported reactors (goes beyond the defensive position of Kakodkar), appears to have been met in the final deal.

According to the Indian government's official position, India's indigenous three-stage nuclear power programme is unaffected by the Indo–US Nuclear Deal; "Its full autonomy has been preserved." Both right and left-wing political parties opposed the deal in the Parliament. The left feared the deal would make the country subservient to U.S. interests, while the right felt it would limit further nuclear testing.

According to one view within the Indian defence establishment, the deal "has for all practical purposes capped Indian ability to field test and proof high yield nuclear weapons till some time in future (about 20 years) when Indian three-stage nuclear fuel cycle based on Thorium fuel matures into mainstream power production, thus eliminating Indian dependence on imported nuclear fuel from NSG countries or if there is a breakout in global nuclear test moratorium."

Indian nuclear energy forecasts

Atomic Power Stations in India
 Active plants
 Under construction
 Planned plants

On the basis of the three-stage plan and assuming optimistic development times, some extravagant predictions about nuclear power have been made over the years:

Bhabha announced that there would be 8,000 MW of nuclear power in the country by 1980. As the years progressed, these predictions were to increase. By 1962, the prediction was that nuclear energy would generate 20,000–25,000 MW by 1987, and by 1969, the AEC predicted that by 2000 there would be 43,500 MW of nuclear generating capacity. All of this was before a single unit of nuclear electricity was produced in the country. Reality was quite different. Installed capacity in 1979–80 was about 600 MW, about 950 MW in 1987, and 2720 MW in 2000.

In 2007, after five decades of sustained and generous government financial support, nuclear power's capacity was just 3,310 MW, less than 3% of India's total power generation capacity.

The Integrated Energy Policy of India estimates the share of nuclear power in the total primary energy mix to be between 4% and 6.4% in various scenarios by the year 2031–32. A study by the DAE, estimates that the nuclear energy share will be about 8.6% by the year 2032 and 16.6% by the year 2052. The possible nuclear power capacity beyond the year 2020 has been estimated by DAE is shown in the table.[115] The 63 GW expected by 2032 will be achieved by setting up 16 indigenous Pressurised Heavy Water Reactors (PHWR), of which ten is to be based on reprocessed uranium. Out of the 63 GW, about 40 GW will be generated through the imported Light Water Reactors (LWR), made possible after the NSG waiver.[116]

Year Pessimistic (GWe) Optimistic (GWe)
2030 48 63
2040 104 131
2050 208 275

Indian Prime Minister Manmohan Singh stated in 2009 that the nation could generate up to 470 GW of power by 2050 if it managed the three-stage program well. "This will sharply reduce our dependence on fossil fuels and will be a major contribution to global efforts to combat climate change", he reportedly said. According to plan, 30% of the Indian electricity in 2050 will be generated from thorium based reactors. Indian nuclear scientists estimate that the country could produce 500 GWe for at least four centuries using just the country's economically extractable thorium reserves.

Thorium energy forecasts

According to the Chairman of India's Atomic Energy Commission, Srikumar Banerjee, without the implementation of fast breeders the presently available uranium reserves of 5.469 million tonnes can support 570 GWe till 2025. If the total identified and undiscovered uranium reserves of 16 million tonnes are brought online, the power availability can be extended till the end of the century. While calling for more research into thorium as an energy source and the country's indigenous three-stage programme, he said, "The world always felt there would be a miracle. Unfortunately, we have not seen any miracle for the last 40 years. Unless we wake up, humans won't be able to exist beyond this century."

Introduction to entropy

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