The International Panel on Fissile Materials (IPFM) was founded in January 2006. It is an independent group of arms-control and nonproliferation experts from seventeen countries, including both nuclear weapon and non-nuclear weapon states.

The mission of the IPFM is to analyze the technical basis for practical and achievable policy initiatives to secure, consolidate, and reduce stockpiles of highly enriched uranium and plutonium. These fissile materials are the key ingredients in nuclear weapons, and their control is critical to nuclear disarmament, halting the proliferation of nuclear weapons, and ensuring that terrorists do not acquire nuclear weapons.

Both military and civilian stocks of fissile materials have to be addressed. The nuclear weapon states still have enough fissile materials in their weapon stockpiles for tens of thousands of nuclear weapons. On the civilian side, enough plutonium has been separated to make a similarly large number of weapons. Highly enriched uranium is used in civilian reactor fuel in more than one hundred locations. The total amount used for this purpose is sufficient to make about one thousand Hiroshima-type bombs, a design potentially within the capabilities of terrorist groups.

The Panel is co-chaired by Professor R. Rajaraman of Jawaharlal Nehru University in New Delhi and Professor Frank von Hippel of Princeton University. Its members include nuclear experts from Brazil, China, France, Germany, India, Ireland, Japan, South Korea, Mexico, the Netherlands, Norway, Pakistan, Russia, South Africa, Sweden, the United Kingdom and the United States. Professor Jose Goldemberg of Brazil stepped down as co-chair of IPFM on July 1, 2007. He continues as a member of IPFM. Short biographies of the panel members can be found at the end of this report. IPFM research and reports are shared with international organizations, national governments and nongovernmental groups. It has full panel meetings twice a year in capitals around the world in addition to specialist workshops. These meetings and workshops are often in conjunction with international conferences at which IPFM panels and experts are invited to make presentations.

Princeton University’s Program on Science and Global Security provides administrative and research support for the IPFM.

IPFM’s initial support is provided by a five-year grant to Princeton University from the John D. and Catherine T. MacArthur Foundation of Chicago.

Disclaimer

Not all of the authors of the chapters in this report agree with
all of the conclusions of the Overview chapter.

Specifically, Gennadi Pshakin, who authored chapter 5, does not share the Overview chapter’s
pessimistic conclusions about the future of fast breeder reactors.

M. V. Ramana

India is one of only two countries that are currently constructing commercial scale breeder reactors. (The other is Russia.) Both the history of the program and the economic and safety features of the reactor suggest, however, that the program will not fulfill the promises with which it was begun and is being pursued.

History

Breeder reactors in India were originally proposed in the 1950s as part of a three- stage nuclear program as a way to develop a large autonomous nuclear power program despite India’s relatively small known resource of uranium ore.1

The first stage of the three-phase strategy involves the use of uranium fuel in heavy water-reactors, followed by reprocessing the irradiated spent fuel to extract the plutonium.

In the second stage, the plutonium is used to provide startup cores of fast breeder reactors. These cores would be surrounded by blankets of either depleted or natural uranium, to produce more plutonium. If the blanket were thorium, it would produce chain-reacting uranium-233. So as to ensure that there is adequate plutonium to construct follow-on breeder reactors, however, breeder reactors would have to be equipped with uranium blankets until the desired nuclear capacity was achieved.

The third stage would involve breeder reactors using uranium-233 in their cores and thorium in their blankets. Though the thorium-uranium-233 cycle would result in slow growth of nuclear power, presumably the rationale for going to this stage was to completely eliminate the requirement for uranium.

The three-stage program remains the official justification for pursuing breeders, despite their slow and disappointing progress.

A version of this chapter has been published in Science and Global Security 17 (2008): 54-67.

Though India’s Department of Atomic Energy (DAE) has been talking about breeder reactors since its inception, work on even conceptual studies on breeders began only in the early 1960s. In 1965, a fast reactor section was formed at the Bhabba Atomic Research Center (BARC) and design work on a 10 MWe experimental fast reactor was initiated.2 This seems to have been abandoned and, in 1969, the DAE entered a collaboration agreement with the French Atomic Energy Commission (CEA) and obtained the design of the Rapsodie test reactor and the steam generator design of the Phenix reactor.3 This was to be the Fast Breeder Test Reactor (FBTR), India’s first breeder reactor.

As part of the agreement with the CEA, a team of approximately thirty Indian engineers and scientists were trained at Cadarache, France. Once they returned, they formed the nucleus of the Reactor Research Centre (RRC) that was set up in 1971 at Kalpakkam to lead the breeder effort. In 1985, this was renamed the Indira Gandhi Centre for Atomic Research (IGCAR). Over the years, the center has emerged as the main hub of activities related to India’s breeder program.

Frank von Hippel

The possibility of a plutonium-fueled nuclear reactor that could produce more fuel than it consumed (a "breeder reactor") was first raised during World War II in the United States by scientists in the atomic bomb program. In the following two decades, the Soviet Union, the United Kingdom, France, Germany, Japan and India followed the United States in establishing national plutonium breeder reactor programs, while Belgium, Italy and the Netherlands joined the French and German programs as partners. In all of these programs, the main driver was the hope of solving the long-term energy supply problem using the large scale deployment of nuclear energy for electric power. Plutonium-fueled breeder reactors appeared to offer a way to avoid a potential shortage of the low-cost uranium required to support such an ambitious vision using other kinds of reactors.

Uranium proved to be much more abundant than originally imagined and, after a fast start, nuclear power growth slowed dramatically in the late 1980s and global nuclear capacity is today about one-tenth the level that had been projected in the early 1970s. The urgency of deploying fast-neutron reactors for plutonium breeding therefore abated — at least in the western Organization for Economic Co-operation and Development (OECD) countries. In India and Russia, however, concerns about potential near-term uranium shortages persist, and new demonstration breeder reactors are being built. China, which currently is building up its nuclear capacity at an enormous rate, is considering the possibility of building two Russian-designed breeder reactors. Because of the high costs and reliability and safety issues that are detailed below, however, no commercial breeder reactors have been deployed.

Interest in fast-neutron reactors persists in the OECD countries for a new reason, political difficulties with storing or disposing of spent fuel. "Reprocessing" spent fuel does not eliminate the problem of siting a geological repository but a reprocessing plant does provide an interim destination that has proved a path forward with regard to the spent fuel problem in a number of nations.

Spent-fuel reprocessing was originally launched in countries that planned to deploy breeder reactors. They wanted separated plutonium for manufacturing startup fuel for their first breeder reactors. Standard light-water-reactor spent fuel contains about one percent plutonium. In the absence of breeder reactors, the separated plutonium has become a disposal problem and some countries have decided to recycle it into fuel for the same reactors that produced it. Slow-neutron reactors are relatively ineffective, however, in fissioning some of the plutonium isotopes, which therefore build up in recycled fuel.

Fast-neutron-reactor advocates argue that, if the plutonium and other long — lived transuranics in spent fuel could be fissioned almost entirely, the political problem of finding a geological disposal site for radioactive waste consisting of mostly shorter-lived fission products would become much easier. Fast neutron reactors would be more effective in fissioning all the transuranic isotopes. Fast — neutron breeder reactors could be converted into transuranic "burner" reactors by removing the plutonium-breeding uranium blankets around their cores and flattening the cores into more of a "pancake" shape so that more neutrons would leak out of them.

Table 1.1 Major experimental, pilot and demonstration fast breeder reactors.1

This report looks at the experience and status of breeder reactor programs in France, India, Japan, the Soviet Union/Russia, the United Kingdom and the United States. The major breeder reactors built in these countries are listed in table 1.1. Germany also built two breeder reactors. All were sodium cooled.

The problems described in the country case studies in the following chapters make it hard to dispute Admiral Hyman Rickover’s summation in 1956, based on his experience with a sodium-cooled reactor developed to power an early U. S. nuclear submarine, that such reactors are "expensive to build, complex to operate, susceptible to prolonged shutdown as a result of even minor malfunctions, and difficult and time-consuming to repair."2

Fast-neutron breeder reactors

Fissile isotopes are the essential nuclear materials in both nuclear reactors and nuclear weapons. They undergo fission when they absorb neutrons and, on average, release more neutrons than they absorb. This makes a sustained chain — reaction possible in a "supercritical mass." This supercritical mass must contain a significant concentration of fissile isotopes and must be large enough so that only a small fraction of the neutrons escape without interacting.

The most important fissile materials are uranium-235 and plutonium-239. Uranium-235 is found in nature, constituting 0.7 percent of natural uranium. Plutonium-239 is created when uranium-238 (99.3 percent of natural uranium) absorbs a neutron (figure 1.1).

The vast majority of deployed power reactors around the world are fueled with low-enriched uranium and use a neutron "moderator" — in most cases, ordinary water, which also serves as the reactor coolant — that slows the neutrons and increases the likelihood that they will be captured by uranium-235 and cause it to fission. Such reactors are called "light-water reactors" and are fueled by uranium typically enriched to four-five percent in uranium-235. Light-water reactors are so named to distinguish them from the "heavy-water reactors" developed by Canada, which are fueled by natural uranium. In both types of reactors, the neutrons lose most of their energy in collisions with hydrogen, ordinary hydrogen in light — water reactors and heavy-hydrogen or deuterium in heavy-water reactors. In both types of reactors, some of the extra neutrons from uranium-235 fissions are also captured by uranium-238, converting it into chain-reacting plutonium-239 — but not enough to replace the fissioned uranium-235.

Virtually all breeder reactor programs have focused on reactors that do not use water as a coolant, so that the neutrons propagating the chain-reaction remain energetic (fast).

In order to be supercritical with fast neutrons, the "cores" of breeder reactors contain over 20 percent of fissile material — usually plutonium-239 — mixed with natural or "depleted uranium" (the residue after uranium-235 has been extracted from natural uranium by uranium-enrichment plants). Surrounding this core on all sides is a "blanket" — usually also consisting of natural or depleted uranium. Neutrons that leak out of the core are absorbed by the uranium-238 in the blanket and convert it into plutonium. Because such a reactor produces more plutonium than it consumes, its ultimate fuel is uranium-238, which is 140 times more abundant than uranium-235.

Plutonium breeder reactor programs have focused on fast-neutron reactors because, when a fast neutron fissions a plutonium-239 nucleus, more secondary neutrons are produced per fission than with any other combination of neutron speed and fissile isotope.3 Fast-neutron plutonium-fueled reactors can therefore breed extra fissile material more rapidly than any other reactor system. Despite the safety, cost and reliability issues of fast-neutron reactors, this fact determined their choice as the preferred technology at a time when the global population of nuclear power reactors was expected to double every decade indefinitely. The extra plutonium produced by fast-neutron reactors could be used to provide startup fuel for additional plutonium breeder reactors, allowing the number of breeder reactors to grow at a high rate.

In Russia, given the cost and safety problems associated with plutonium fuel thus far, demonstration fast-neutron reactors have been fueled with highly-enriched uranium, enriched to between 20 and 30 percent uranium-235.

Why commercialization of breeder reactors failed

The rationale for pursuing breeder reactors — sometimes explicit and sometimes implicit — was based on the following key assumptions:

1. Uranium is scarce and high-grade deposits would quickly become depleted if fission power were deployed on a large scale;

2. Breeder reactors would quickly become economically competitive with the light-water reactors that dominate nuclear power today;

3. Breeder reactors could be as safe and reliable as light-water reactors; and,

4. The proliferation risks posed by breeders and their "closed" fuel cycle, in which plutonium would be recycled, could be managed.

Each of these assumptions has proven to be wrong.

Uranium is cheap and abundant. Breeder reactors were seen as a solution for the uranium scarcity problem because, by converting uranium-238 into chain­reacting plutonium, they can potentially increase one-hundred-fold the amount of fission energy that can be extracted from a kilogram (kg) of uranium and make it economically feasible to mine much lower grades of uranium ore.4

In 2007, uranium requirements for the global fleet of nuclear power reactors were 67,000 metric tons — approximately 180 tons per gigawatt of generating capacity per year. The International Atomic Energy Agency (IAEA) projects that global nuclear capacity will increase and that uranium requirements will increase correspondingly to between 94,000 and 122,000 tons a year in 2030.5

In 2008, the biennial report put out by the OECD Nuclear Energy Agency, Uranium 2007: Resources, Production and Demand — also known as "the Red Book" — found that, despite inflation, global known conventional resources of uranium recoverable for less than $130/kg had increased from about 4.7 to about 5.5 million tons. The Red Book also reported estimates from 27 countries that, with further exploration, an additional 7.6 million tons of uranium would be discovered in the same cost range.6 At $130/kg, the cost of uranium would contribute 0.3 U. S. cents to the cost of a kilowatt-hour of nuclear electricity.

In the long run, worldwide, the amount of uranium recoverable at low cost is virtually certain to be far greater than the numbers reported in the Red Book. If plausible estimates of geological abundance are used, the amount of uranium still to be discovered at recovery costs up to $130/kg would be 50-126 million tons.7 This corresponds to 500 to 1000 times the projected demand in 2030.

It will be seen from figure 1.2 that the price of uranium on the spot market went significantly above $130/kg during the late 1970s and then again after 2005. Except for these two periods when there was disequilibrium between supply and demand, prices have been less than $50 per kg. The 1970s price peak was due to the expectation of an enormous expansion in nuclear power capacity. This expectation was not realized but large stockpiles of uranium were built up and then sold off during the subsequent decades resulting in the closure of many uranium mines. The sale by Russia to the U. S. of low-enriched uranium blended down from 500 tons of weapon-grade uranium from excess Cold War weapons at a rate sufficient to fuel half of the U. S. nuclear capacity extended the period of low demand for freshly mined uranium.8 The stockpiles of natural uranium have been largely used up, however, and the blend-down of the Russian weapon-grade uranium will be completed in 2013. The most recent uranium price peak therefore reflected, at least in part, the expectation, compounded by speculation, that there might be uranium shortages before uranium-mining capacity increases again to the level required to support growing demand.

In any case, unlike the situation with oil or gas-fueled power plants, the cost of uranium fuel can double without having a significant impact on the cost of nuclear power. As noted above, at $130/kg, the cost of uranium contributes only 0.3 cents to the cost of a kilowatt-hour (kWh), which is about 5 percent of the cost of electricity produced by a new light-water reactor.9

Breeder reactors are costly to build and operate. Governments of countries in the OECD have together reported that they have spent about $50 billion (2007$) on fission and breeder reactor research and development (figure 1.3). Of this total, the United States reported that it had spent $15 billion, Japan $12 billion, the United Kingdom $8 billion, Germany $6 billion and Italy $5 billion. France

reported only $1 billion in expenditures but this was obviously an incomplete report, given that the total cost of the Superphenix demonstration project alone is estimated at FRF 65 billion (1998FRF) or $14 billion (2007$) (see chapter 2).

Russia and India, which are both outside the OECD, have spent large amounts on breeder research, development and demonstration. The Soviet Union and Russia alone have spent an estimated $12 billion (see chapter 5). Yet none of these efforts has produced a reactor that is anywhere near economically competitive with light-water reactors.

The individual country studies make clear that, without astronomically high uranium prices, breeder reactors are unlikely to be economically competitive with light-water reactors. For "demonstration" liquid-sodium-cooled reactors the capital costs per kilowatt (KW) generating capacity have typically been more than twice those of water-cooled reactors of comparable capacity. Since breeder reactors were never built in quantity, it could be expected that, in production, this cost ratio would decline. Few if any argue today, however, that the capital costs for breeder reactors could be less than 25 percent higher than for water-cooled reactors of similar generating capacities. This would be a capital cost difference on the order of $1000 per kilowatt of generating capacity. At a 10 percent capital charge and a 90 percent average capacity factor, this would translate to a cost difference of about 1.3 cents per kilowatt hour.

Detailed economic comparisons of light-water reactors and breeder reactors using different breeding ratios, fuel reprocessing and fabrication costs, and capital costs show that direct disposal of spent light-water-reactor fuel would be far less expensive than reprocessing and plutonium recycle in breeder reactors under a wide range of assumptions.12

Fast-neutron reactors have special safety problems. As already noted, fast-neutron reactors cannot use water as a coolant because collisions with the hydrogen nuclei in water quickly remove most of the kinetic energy from the neutrons. Also, in order to sustain a chain-reaction with fast neutrons, the fissile material in a reactor core must be more concentrated. As a result, fast-neutron — reactor cores are smaller than those of light-water reactors with the same power. This necessitates the use of a coolant that can efficiently carry away the heat. The coolant that has been used in all demonstration breeder reactors to date is a liquid metal that melts at relatively low temperatures — sodium.

Sodium has both safety advantages and disadvantages compared to water. Its primary safety advantage is that the reactor operates below the boiling point of liquid sodium (883 °С) and therefore at low pressure. By contrast, water-cooled reactors operate at high pressures — over 150 atmospheres for pressurized water reactors. Therefore, if there is a large break in a pipe of a water-cooled reactor, the water flashes into steam, leaving the reactor’s intensely hot fuel without coolant unless the core is flooded with emergency cooling water. In the case of a sodium — cooled reactor, however, unless the break is below the top of the core, the sodium will continue to cover the core and absorb heat.

Sodium’s major disadvantage is that it reacts violently with water and burns if exposed to air. The steam generators, in which molten-sodium and high-pressure water are separated by thin metal, have proved to be one of the most troublesome features of breeder reactors. Any leak results in a reaction that can rupture the tubes and lead to a major sodium-water fire.

As the country studies detail, a large fraction of the liquid-sodium-cooled reactors that have been built have been shut down for long periods by sodium fires. Russia’s BN-350 had a huge sodium fire. The follow-on BN-600 reactor was designed with its steam generators in separate bunkers to contain sodium-water fires and with an extra steam generator so a fire-damaged steam generator can be repaired while the reactor continues to operate using the extra steam generator. Between 1980 and 1997, the BN-600 had 27 sodium leaks, 14 of which resulted in sodium fires (see chapter 5).

Leaks from pipes into the air have also resulted in serious fires. In 1995, Japan’s prototype fast reactor, Monju, experienced a major sodium-air fire. Restart has been repeatedly delayed, and, as of the end of 2009, the reactor was still shut down. France’s Rapsodie, Phinix and Superphinix breeder reactors and the UK’s Dounreay Fast Reactor (DFR) and Prototype Fast Reactor (PFR) all suffered significant sodium leaks, some of which resulted in serious fires.

Sodium also creates radiation problems. When it absorbs a neutron, ordinary sodium-23 becomes sodium-24, a gamma-emitting isotope with a 15-hour half­life. The sodium that cools the core therefore becomes intensely radioactive. To ensure that a steam-generator fire does not disperse radioactive sodium, reactor designers have inserted an intermediate sodium loop. The heat generated from the reactor is transferred to non-radioactive sodium through a sodium-sodium heat exchanger. The non-radioactive sodium delivers the heat to the steam generators. The extra sodium loops and associated pumps contribute to the high capital costs of breeder reactors.

Finally, light-water-cooled reactors have the critical safety characteristic that, if the water moderator is lost, the chain-reaction stops. It is impossible to sustain a chain-reaction in 4 to 5 percent enriched uranium without slowing the neutrons so that they are captured preferentially by uranium-235. In the absence of the water, the fast neutrons will be absorbed mostly in uranium-238 and the chain — reaction ends.

By contrast, in a fast-neutron reactor, the concentration of plutonium is high enough that it can sustain a chain-reaction even in the event of a coolant loss. Indeed, except for special core configurations, the reactivity will increase if the coolant is lost.13 Furthermore, if the core heats up to the point of collapse, it can assume a more critical configuration and blow itself apart in a small nuclear explosion.14 Whether such an explosive core disassembly could release enough energy to rupture a reactor containment and cause a Chernobyl-scale release of radioactivity into the environment is a major concern and subject of debate. (See chapter 3 for a discussion of this debate in India.)

Sodium-cooled reactors have severe reliability problems. The reliability of light-water reactors has increased to the point where, on average, they operate at about 80 percent of their generating capacity. By contrast, a large fraction of sodium-cooled demonstration reactors have been shut down most of the time that they should have been generating electric power. A significant part of the problem has been the difficulty of maintaining and repairing the reactor hardware that is immersed in sodium. The requirement to keep air from coming into contact with sodium makes refueling and repairs inside the reactor vessel more complicated and lengthy than for water-cooled reactors. During repairs, the fuel has to be removed, the sodium drained and the entire system flushed carefully to remove residual sodium without causing an explosion. Such preparations can take months or years.

In contrast, when a water-cooled reactor is shut down, the top of the pressure vessel can be removed and the reactor cavity that holds the pressure vessel can be flooded with water to provide shielding against the radioactivity of the fuel and the irradiated steel. Repairs can take place guided by underwater periscopes and video cameras.

The history of the world’s only commercial-sized breeder reactor, France’s Superphenix, is dominated by lengthy shutdowns for repairs (see chapter 2). Superphenix went critical and was connected to the grid in January 1986 but was shut down more than half of the time until operations ceased in December 1996. Its lifetime capacity factor — the ratio of the number of kilowatt-hours that it generated to the number it could have generated had it operated continually at full capacity — was less than 7 percent. The histories of Japan’s Monju and the U. K.’s Dounreay and Prototype Fast Reactors and the U. S. Enrico Fermi 1 demonstration breeder reactor power plants were similarly characterized by prolonged shutdowns (see chapters 4, 6 and 7). Russia’s BN-600 has experienced a respectable capacity factor but only because of the willingness of its operators to continue to operate it despite multiple sodium fires.

The fast-neutron reactor fuel cycle provides easy access to plutonium for weapons. All reactors produce plutonium in their fuel but breeder reactors require plutonium recycle, the separation of plutonium from the ferociously radioactive fission products in the spent fuel. This makes the plutonium more accessible to would-be nuclear-weapon makers. Breeder reactors — and separation of plutonium from the spent fuel of ordinary reactors to provide startup fuel for breeder reactors — therefore create proliferation problems.

This fact became dramatically clear in 1974, when India used the first plutonium separated for its breeder reactor program to make a "peaceful nuclear explosion." Breeders themselves have also been used to produce plutonium for weapons. France used its Phenix breeder reactor to make weapon-grade plutonium in its blanket. India, by refusing to place its breeder reactors under international safeguards as part of the U. S.-India nuclear deal, has raised concerns that it might do the same.

India’s Prototype Fast Breeder Reactor (PFBR), expected to be completed in 2010, will have the capacity to make 90 kg of weapon-grade plutonium per year, if only the radial blanket is reprocessed separately and 140 kg per year if both radial and axial blankets are reprocessed.15 The Nagasaki bomb contained 6 kg of weapon-grade plutonium and modern weapons designs contain less. At 5 kg per warhead, the PFBR would produce enough weapon-grade plutonium for 20-30 nuclear weapons a year, a huge increase in production capacity in the context of the South Asian nuclear arms race.

The G. W. Bush Administration proposed to make reprocessing more "proliferation resistant" by leaving some of the other transuranic elements (neptunium, americium and curium) mixed with the plutonium.16 Even if all the transuranics

were left mixed with the plutonium, however — a project that the U. S. Department of Energy abandoned when it learned that the technology was not in hand — the gamma radiation field surrounding the mix would still be less than one-hundredth the level the IAEA considers self-protecting against theft and thousands of times less than the radiation field surrounding plutonium when it is in spent fuel (figure 1.4).

The budget for the FBTR was approved by DAE as early as September 1971 and it was anticipated that the FBTR would be commissioned by 1976.4 The reactor finally attained criticality only in October 1985 and the steam generator began operating in 1993.5

Much of the first one and a half decades of the FBTR’s operations were marred by several accidents of varying intensity. Two of these are described below in some detail to illustrate the complexities of dealing with even relatively minor accidents and the associated delays, as well as the hazards posed to workers. When viewed in combination with similar experiences elsewhere, these circumstances suggest that it is unlikely that sodium-cooled breeder reactors will ever perform with the reliability that water-cooled reactors have demonstrated over the past two decades.

In May 1987 there was a major incident that took two years to rectify.6 This occurred as a fuel subassembly was being transferred from the core to the periphery.7 The problem began with the failure of a protective circuit involved in the rotation of the plug to move the selected fuel assemblies. For some reason, this protective circuit was bypassed and the plugs were rotated with a foot long section of one fuel subassembly protruding into the reactor core. This resulted in the bending of that specific subassembly as well as the heads of 28 reflector subassemblies on the path of its rotation. Various maneuvers to rectify the situation did not help and only resulted in one reflector subassembly at the periphery getting ejected as well as the bending of a sturdy guide tube by 32 cm. The last event has been described as the result of "a complex mechanical interaction" which seems to suggest that how it happened was never really understood.

Extensive repairs were required before the reactor could be restarted. First, the guide tube had to be cut into two parts using a specially designed remote cutting machine while ensuring that none of the chips produced during the cutting process fell into the core.8 Then the damaged reflector subassemblies had to be identified using a periscope. Finally part of the sodium had to be drained out and the damaged subassemblies removed using specially designed grippers. As might be expected, all of this took time and reactor operations commenced only in May 1989.9

The second accident described here is one that is common in fast breeder reactors — a sodium leak. That this occurred seventeen years after the reactor was commissioned underscores the generic nature of such accidents. The leak occurred in September 2002 inside the purification cabin, which houses the pipelines of the primary sodium purification circuit.10 The cause of the leak is said to have been "the defective manufacturing process adopted in the manufacture of the bellows sealed sodium service valves". By the time the leak could be confirmed and controlled, approximately 75 kilograms of sodium had spilled over and solidified on the cabin floor and various components in that cabin.

Removing this radioactive sodium was a major effort. To begin with, even to approach the cabin, the workers had to wait ten days to allow for a reduction in the radioactivity from the sodium, some of which had absorbed a neutron to become Na-24, a gamma emitter (15-hour half-life). Even then, in areas near the spilled sodium, the dose rate was as high as 900 millisieverts per hour (mSv/h).11 Another problem resulted from the whole cabin normally being surrounded by a layer of nitrogen so as to avoid sodium burning. At first, IGCAR tried to simply replace the nitrogen with regular air so that cleanup workers could breathe. But this led to sparks and fires involving the spilled sodium. These had to be put out with dry chemical powders — but then this led to lots of dust being suspended in the atmosphere and made visibility poor. Once again nitrogen had to be reintroduced. Workers were then sent in with masks that had tubes feeding them with breathing air. Much of the work had to be done remotely, which, while lowering radiation exposure, made it a very slow operation. In all, removing the 75 kg of sodium and bringing the cabin back to normal conditions took approximately three months.12

The FBTR has also seen several other accidents and unusual occurrences, such as unexplained reactivity transients.13 Overall, the reactor’s performance has been mediocre: it took fifteen years before the FBTR even managed fifty plus days of continuous operation at full power.14 In the first twenty years of its life, the reactor has operated for only 36,000 hours, i. e. an availability factor of approximately 20 percent.15 Despite this checkered history, IGCAR claims to have "successfully demonstrated the design, construction and operation" of a fast breeder reactor.16

The Prototype Fast Breeder Reactor

Even before the FBTR came on line, the DAE started making plans for a larger Prototype Fast Breeder Reactor (PFBR). In 1983, the DAE requested budgetary support from the Government.17 The first expenditures on the PFBR started in 1987-88.18 In 1990, it was reported that the Government had "recently approved the reactor’s preliminary design and has awarded construction permits" and that the reactor would be on line by 2000.19 Construction of the reactor finally began in October 2004 and was projected to be commissioned in 20 1 0.20 The PFBR will likely suffer from the two problems that have plagued breeder reactors elsewhere: the risk of a catastrophic accident and poor economics. It will also be a source of weapon-grade plutonium that might be used for the strategic program. See the discussion in chapter 2 of France’s use of its first demonstration breeder reactor Phenix to produce weapon-grade plutonium for France’s weapons program.

After six decades and the expenditure of the equivalent of tens of billions of dollars, the promise of breeder reactors remains largely unfulfilled and efforts to commercialize them have been steadily cut back in most countries.

Germany, the United Kingdom and the United States have abandoned their breeder reactor development programs. Despite the arguments by France’s nuclear conglomerate Areva, that fast-neutron reactors will ultimately fission all the plutonium building up in France’s light-water reactor spent fuel,18 France’s only operating fast-neutron reactor, Phenix, was disconnected from the grid in March 2009 and scheduled for permanent shutdown by the end of that year.19 The Superphenix, the world’s first commercial-sized breeder reactor, was abandoned in 1998 and is being decommissioned. There is no follow-on breeder reactor planned in France for at least a decade.

Japan’s Monju reactor operated for only a year before it was shut down by an accident in 1995 and it had not resumed operation as of the end of 2009. There are plans for a new demonstration reactor by 2025 and commercialization of breeder reactors by 2050 but there is reason to doubt these projections. Japan’s Government is not willing to kill its breeder program entirely, because, as in France, the breeder is still the ultimate justification for Japan’s spent fuel reprocessing program. For decades, however, the Japanese Government has been reducing funding for its breeder program and shifting commercialization further and further into the future (see chapter 4).

Russia and India are building demonstration breeder reactors. In both cases, however, their breeder (and spent fuel reprocessing) programs leave much to be desired regarding the availability of data on reliability, safety and economics. In the case of India, there is also the potential for use of breeder reactors to produce plutonium for weapons. The high costs of commercial breeder reactors and an international Fissile Material Cutoff Treaty that bans production of fissile materials for weapons will force some of these issues into the open and foster new debates about the value of these breeder programs.

In the United States, during the G. W. Bush Administration, fast reactors returned to the agenda as "burner" reactors. In an initiative started in 2006 labeled "The Global Nuclear Energy Partnership (GNEP)," the U. S. Department of Energy proposed that sodium-cooled fast-neutron reactors be used to make the radioactive waste in spent reactor fuel more manageable. With the removal of the uranium blankets around their cores, fast-neutron reactors would, like light-water reactors, breed less fissile material than they burned. The high-energy neutron spectrum of the sodium-cooled reactors would be more effective, however, in fissioning the non-chain-reacting isotopes of plutonium and minor transuranic elements. Already in 1996, however, a National Academy of Sciences assessment commissioned by the U. S. Department of Energy, had concluded that such an effort would have very high costs and marginal benefits and would take hundreds of years of recycling to reduce the global inventory of transuranic isotopes by 99 percent.20 The Obama Administration and the U. S. Congress share this skepticism and propose a new research and development program to investigate alternative strategies for managing U. S. spent fuel.21

The breeder reactor dream is not dead but it has receded far into the future. In the 1970s, breeder advocates were predicting that the world would have thousands of breeder reactors operating by now. Today, they are predicting commercialization by approximately 2050. In the meantime, the world has to deal with the legacy of the dream; approximately 250 tons of separated weapon-usable plutonium and ongoing — although, in some cases struggling — reprocessing programs in France, India, Japan, Russia and the United Kingdom.

kg corresponding to less than 0.1 cents per kWh. Operations and maintenance would add approximately 1.5 cents/kWh. Finally, for a light-water reactor costing $4000/KWe operating at a 90 percent capacity factor the capital charge would be 5 cents/kWh, assuming a 10 percent capital charge; Massachusetts Institute of Technology, The Future of Nuclear Power, An MIT Interdisciplinary Study (Cambridge: MIT Press, 2003), Appendix 5A.

This figure is based on figure 5 of Matthew Bunn, Steve Fetter, John Holdren and Bob van der Zwaan, "The Economics of Reprocessing Versus Direct Disposal of Spent Nuclear Fuel," Nuclear Technology 150 (June 2005): 209. It has been updated by Steve Fetter through 2006 and the author through 2007 (average U. S. price) and 2008 and early 2009 (spot price) based on U. S. Energy Information Administration, "Average Price and Quantity for Uranium Purchased by Owners and Operators of U. S. Civilian Nuclear Power Reactors by Pricing Mechanisms and Delivery Year," <http://www. eia. doe. gov/cneaf/nuclear/umar/table5.html> (accessed 15 September 2009); and Uranium Intelligence Weekly respectively.

From International Energy Agency R&D Statistics Database, <http://www. iea. org/Textbase/stats/rd. asp> (accessed 15 September 2009). Unavailable values have been replaced with zeros.

Matthew Bunn et al., op. cit., 209. For example, for utility financing at a 10 percent discount rate, central values for reprocessing and breeder core fabrication of $1000 and $1500 per kg of heavy metal respectively, and only a small capital cost difference between light-water reactors and breeders of $200/KWe (5 percent of current light-water reactor capital costs), the breakeven uranium price would be $340 per kg — far greater than projected prices even if nuclear power grows substantially in the coming decades.

Since the sodium slows the neutrons somewhat, its removal increases reactivity since both the fission probability of plutonium-239 and the number of neutrons released per fission increase with neutron energy. The only way to offset this positive reactivity feedback from loss of coolant is to design the core geometry so that leakage of neutrons out of the fuel region of the core increases as the sodium is lost. This requires either that the core be pancake shaped or that neutron absorbing blanket fuel assemblies be dispersed among the fuel assemblies.

H. A. Bethe and J. H. Tait, "An Estimate of the Order of Magnitude of the Explosion When the Core of a Fast Reactor Collapses," UKAEA-RHM 56 (1956).

Alexander Glaser and M. V. Ramana, "Weapon-grade Plutonium Production Potential in the Indian Prototype Fast Breeder Reactor," Science and Global Security 15 (2007): 85-105.

The National Energy Policy Development Group, "Report of Vice President Cheney’s 2001 Energy Task force," 5-21. See also U. S. Department of Energy, Office of Nuclear Energy, Science, and Technology, "Report to Congress on Advanced Fuel Cycle Initiative: The Future Path for Advanced Spent Fuel Treatment and Transmutation Research", (2003).

Robert Hill, Argonne National Laboratory, "Advanced Fuel Cycle Systems: Recycle/Refabrication Technology Status," September 2005. See also, E. D. Collins, Oak Ridge, "Closing the Fuel Cycle can Extend the Lifetime of the High-Level-Waste Repository," American Nuclear Society, 2005 Winter Meeting, Washington, DC; and, Jungmin Kang and Frank von Hippel, "Limited Proliferation Resistance Benefits from Recycling Unseparated Transuranics and Lanthanides from Light-water Reactor Spent Fuel," Science and Global Security 13 (2005): 169.

Areva is reprocessing France’s low-enriched uranium fuel and recycling the plutonium into mixed-oxide (MOX, uranium-plutonium) fuel for light — water reactors. The spent MOX fuel contains approximately two thirds as much plutonium as the fresh MOX fuel but the plutonium contains an increased fraction of the even isotopes, plutonium-238, plutonium-240 and plutonium-242, that are difficult to fission in thermal reactors. The spent MOX fuel therefore is being stored in the hopes that fast-neutron reactors will eventually be built that can fission all the plutonium isotopes efficiently.

Mycle Schneider, Steve Thomas, Antony Froggatt and Doug Koplow, "World Nuclear Industry Status Report 2009," (August 2009), 101.

Committee on Separations Technology and Transmutation Systems, National Research Council, Nuclear Wastes: Technologies for Separation and Transmutation (Washington, D. C: National Academies Press, 1996).

In June 2009, the U. S. Department of Energy announced that it was cancelling development of the Global Nuclear Energy Partnership Programmatic Environmental Impact Statement because it was no longer pursuing domestic commercial reprocessing.

There are a number of reasons to doubt the safety of the PFBR design.21 As with other breeder reactors, the PFBR design is susceptible to catastrophic accidents involving large and explosive energy releases and dispersal of radioactivity following a core meltdown. The potential for such Core Disruptive Accidents (CDA) comes from the reactor core not being in its most reactive configuration. If conditions during an accident cause the fuel bundles to melt and rearrange, the reactivity could increase leading to further core rearrangement and a potential positive feedback loop. Another unsafe feedback effect that is present in the PFBR design is its positive sodium void coefficient. This means that if the coolant heats up and becomes less dense, forms bubbles, or is expelled from the core, the reactivity increases. The magnitude of the void coefficient is a measure of the feedback and tends to increase with core size.22 For the core design that has been adopted for the PFBR, it has a value of $4.3.23

Compounding the safety risks that come with this large and positive sodium void coefficient, the PFBR design also has a relatively weak containment, which is designed to withstand only 25 kilopascals (kPa) or one quarter of an atmosphere of overpressure.24 This maximum overpressure that the PFBR containment is designed for is lower than some other demonstration reactors (table 3.1). If one considers the ratio of the containment volume times its design overpressure divided by the reactor power, V*P/E, the PFBR containment is weaker than those of all other demonstration breeder reactors except the Prototype Fast Reactor (PFR).25 The difference appears more acute when the higher positive sodium void coefficient of the PFBR in comparison to other breeder reactors is taken into account

It is of course possible to design containments to withstand much higher pressures. Containments for light-water reactors routinely have design pressures above 200 kPa.26 The DAE justifies this choice of containment design by arguing that its safety studies demonstrate that the maximum overpressure expected in a CDA involving the PFBR is smaller than this overpressure. But these results are based on favourable assumptions, in particular, that only limited parts of the reactor core would participate in the CDA and that approximately 1 percent of the thermal energy released would be converted into mechanical energy. Based on such assumptions, the DAE estimates that the maximum credible energy release in a CDA is 100 megajoules (MJ).27 It then calculates that such a CDA leading to sodium leakage into the containment would result in a containment overpressure of 20 kPa.

There are, however, good reasons to consider much larger energy releases from a worst-case CDA to the extent of several hundreds of megajoules in the evaluation of the safety of a reactor design, especially one as large as the PFBR. Table 3.2 shows that the calculated CDA energy releases for a number of breeder reactors are much higher than that of the PFBR, both absolutely and when scaled by reactor power.

The energy releases from core collapse depend sensitively on the reactivity insertion rate, which is the rate at which the fuel rearrangement increases (inserts) reactivity.28 The DAE’s calculation of the maximum CDA energy release is based on a reactivity insertion rate of $65/s, which itself is the result of assuming only limited core disassembly.29 There is ample reason and precedent to use an insertion rate of $100/s as a benchmark for disassembly calculations, with the caveat that it still is not quite an upper bound.30 Likewise, the efficiency of conversion could be much larger than the 1 percent assumed by the DAE. Tests at the UK’s Winfrith facility with core melt amounts of up to 25 kg suggest energy-conversion efficiencies of approximately 4 percent.31 For a reactivity insertion rate of 100 $/s, and an energy conversion efficiency of 1 percent, the energy release from a CDA is 650 MJ.32 It has been estimated that a 650 MJ CDA could lead to an overpressure of approximately 40 kPa on the containment, clearly much higher than the design limit of the containment building.33 Higher conversion factors would imply higher mechanical energy releases and thus higher overpressures and higher likelihoods of containment failure.

To summarize, there are good reasons to believe that the containment of the PFBR does not offer adequate protection against a severe CDA, especially given the many uncertainties inherent in calculations of CDA release energies.

Mycle Schneider

France’s program to produce and separate plutonium began immediately after the Second World War. While the initial purpose was to obtain plutonium for the nuclear weapons program, very early on the fast breeder reactor became a second strategic goal. European cooperation was another goal and the EUROCHEMIC consortium was created in 1957 with the participation of 10 countries; France and Germany held the largest shares with 17 percent each.1

The first reprocessing plant, the "plutonium factory" (usine de plutonium, UP1), began operating in Marcoule in 1958 and the first proposal for the experimental fast reactor Rapsodie was drawn up that year. Preliminary studies for a 1000 megawatt electric (MWe) reactor were conducted as early as 1964.

The behavior of materials was tested under neutron irradiation in Harmonie starting in 1965 and breeder core configurations were studied in the critical facility, Masurca, starting in 1966. These research facilities were located at the Cadarache site in southern France. Much later, in 1982, the Esmeralda facility, also at Cadarache, was designed to study sodium fires. While most of the research was financed by the French Atomic Energy Commission (CEA), up to 35 percent of some research projects were funded by EURATOM.

In 1966, the second commercial reprocessing plant UP2, financed entirely by the CEA (with the civil and military budgets paying equal shares), started operations at La Hague by separating plutonium from gas-graphite reactor fuel. In Belgium, the EUROCHEMIC plant began operating in 1967. It operated until 1974 and reprocessed 181.3 tons of spent fuel of various types and origins. Two years later the CEA started up a light-water reactor head-end at La Hague (UP2-400) and launched the 100 percent daughter company COGEMA under private law. Foreign (German) light-water reactor fuel was sent to La Hague as early as 1973. There had been no experience with reprocessing light-water reactor fuel with much higher burn-ups than gas-graphite reactor fuel and it took COGEMA eleven years, until 1987, to operate at a nominal capacity of 400 tons per year.

A version of this chapter has been published in Science and Global Security 17 (2008): 36-53.

The main argument offered for the DAE’s pursuit of breeder reactors is that India has only "modest uranium reserves" of approximately 60,000 tons, "which can support 10 000 MWe (megawatt electric) of PHWR (pressurized heavy-water reactor) capacities".34 While widely repeated, this formulation is misleading. India’s uranium resource base cannot be represented by a single number. As with any other mineral, at higher prices it becomes economic to mine lower grade and less accessible ores. Exploiting these would increase the amount of uranium available. Therefore, uranium resources can only be specified as a function of price.

As a way of evaluating the economics of breeder reactors, the cost of generating electricity at the 500 MWe PFBR can be compared with that at the PHWR,35 the mainstay technology of the Indian nuclear program.36 In order to address the argument about India’s limited uranium reserves, it must be understood that the reserves are a function of uranium price, which allows calculation of the crossover price when the two technologies generate electricity at the same cost.

The total construction cost of the PFBR is estimated as Rs. 34.92 billion (mixed year Rupees; overnight construction cost of $646 millions in 2004 dollars). The overnight unit cost is $1292/kilowatt (KW) and is lower than the corresponding figure for recent Indian PHWRs of $1371/KW. This is quite in contrast to experiences around the world that suggest that breeder reactors are much more expensive than water moderated reactors; for light-water reactors, a typical estimate of the minimum cost difference is $200/kilowatt electric (KWe).37 The PFBR’s estimated construction cost is also much lower than estimates of breeder reactor construction costs elsewhere; the Nuclear Energy Agency (NEA) gives a range of $1850-2600/KWe ($2000) or $2000-2800 (2004 dollars) for mixed-oxide fueled (MOX) fast reactors.38 Actually constructed breeder reactors in other parts of the world also bear out the expectation of higher costs. Construction costs for the French Phenix reactor with a capacity of 250 MWe totalled $800 million in 1974 French FRF ($800 million in 2004 dollars) or $3200/KW. However, a further €600 million ($870 million in 2004 dollars) were spent on Phenix upgrades between 1997 and 2003. The 1240 MWe Superphenix was even more expensive. For these technical reasons, and the DAE’s history of cost overruns at all the reactors it has constructed, it is likely that the PFBR capital cost will be higher than this projected value.39

In economic terms, the primary material requirement for the PFBR is plutonium. The PFBR design requires an initial inventory of 1.9 tons of plutonium in its core.40 Based on a detailed model of the reactor, it has been estimated that at a 75 percent capacity factor, the PFBR requires 1012 kg of plutonium every year for refueling during equilibrium conditions.41 The plutonium for the initial core and the first few reloads will have to come from reprocessing of PHWR spent fuel. At a real discount rate of 6 percent, reprocessing costs approximately $659 per kg of uranium in the fuel, which corresponds to a plutonium cost of $178/g.42 Because of the higher plutonium content of the PFBR spent fuel, the unit cost of subsequent plutonium requirements would be lower, approximately $43/g.43

Following the Nuclear Energy Agency, the costs of fabricating breeder reactor core fuel and (radial) blanket uranium fuel have been assumed to be $1512/kg and $540/kg.44 The base case assumes costs of $200/kg for natural uranium and $200/ kgU for fabrication of uranium fuel for heavy-water reactors. The high base costs of uranium reflects the higher mining costs of poor quality uranium ore in India

Table 3.3 shows the difference in the levelised cost, at a real discount rate of 6 percent, of producing electricity at the PFBR and at the proposed 2 x 700 MW twin unit PHWRs.45 The economics of the PFBR will be key to the future of breeder reactors in India. The DAE has argued that the "primary objective of the PFBR is to demonstrate techno-economic viability of fast breeder reactors on an industrial scale".46 The results presented here show that the PFBR will not be viable, even at the projected costs and for optimistic assumptions about capacity factors. As table 3.4 shows, breeder reactors across the world have operated with relatively low cumulative load factors. There is no reason to expect that the PFBR experience

would not be similar, and a capacity factor of 50 percent might well be more plausible. This would result in a levelised cost of 8.35 cents/kilowatt hours (kWh), 139 percent more expensive than PHWRs.

As mentioned earlier, the main rationale offered for the pursuit of expensive breeders is the shortage of uranium. The validity of this rationale has been examined by increasing the price of uranium from $200/kg to the crossover value where breeders become competitive. For the optimistic base case, with a PFBR capacity factor of 80 percent, the levelised costs of electricity from the PFBR and PHWR are equal at a uranium price of $1375/kg. At a PFBR capacity factor of 50 percent, the crossover price is $2235/kg

These prices are much higher than current values and significantly larger quantities of uranium will be available at these prices. The distribution of uranium among the major geological reservoirs in the earth’s crust corresponds to a roughly three hundred fold increase in the estimated amount of recoverable uranium for every ten fold decrease in the ore grade.47 Based on this, and assuming that mining cost is inversely proportional to ore grade, one can surmise that the available uranium at costs less than $1375/kg and $2235/kg are approximately 124 and 417 times current reserves respectively. This is an underestimate because it ignores the general trends of reduced mining costs due to learning and improved technology.48 In any case, India should have sufficient uranium for a nuclear energy sector based on PWHRs for many decades, with no reprocessing and breeder reactors.

Plutonium for weapons?

There may be another reason for the DAE’s attraction to breeder reactors. This stems from the source of DAE’s institutional clout: its unique ability to offer both electricity for development and nuclear weapons for security. This was revealed quite clearly during the course of negotiations over the U. S.-India nuclear deal, where in an ostensibly civilian agreement, much of the DAE’s efforts were aimed at optimizing its ability to make fissile material for the nuclear arsenal within various constraints, especially the shortage of domestic low-cost uranium.49 Most prominently, the DAE focused a lot of attention on keeping the fast breeder program outside of safeguards. In a prominent interview to a national newspaper, the head of the DAE said:

Both, from the point of view of maintaining long-term energy security and for maintaining the minimum credible deterrent, the fast breeder programme just cannot be put on the civilian list. This would amount to getting shackled and India certainly cannot compromise one (security) for the other.50

In parallel, the DAE did not classify its reprocessing plants or its stockpile of reactor-grade plutonium as civilian. This allows for the possibility that breeder reactors like the PFBR could be used as a way to launder unsafeguarded reactor-grade plutonium, both in the historical stockpile as well as from future production at unsafeguarded reprocessing plants, into weapon-grade plutonium. While reactor — grade plutonium is consumed in the core of the PFBR, weapon-grade plutonium is produced in the radial and axial blankets. Based on neutronics calculations for a detailed three-dimensional model of the reactor, it has been estimated that 92.4 kg and 52 kg of weapon-grade plutonium will be generated in the radial and axial blankets (93.7 percent and 96.5 percent plutonium-239) respectively in the PFBR each year at a 75 percent capacity factor.51

If the blanket fuel elements are reprocessed separately rather than jointly with the core fuel elements, then the plutonium contained in them can be used for weapons. To make up for this, approximately 346 kg of reactor-grade plutonium derived from reprocessing spent fuel from India’s PHWRs would have to be used in the PFBR annually. The existing stockpile of reactor-grade plutonium and PHWR spent fuel is adequate to meet this need for decades. Such a strategy would increase the DAE’s weapon-grade fissile material production capacity several-fold.

Future projections

The PFBR is to be the first of the many breeder reactors that the DAE envisions building. The DAE’s current projections are that nuclear power would grow to 20 gigawatt electric (GWe) by 2020 and to 275 GWe by 2052, including 260 GWe in metallic fueled breeders.52 More recent media statements following the nuclear suppliers group lifting of its ban on nuclear trade with India project even larger rates of growth of India’s breeder capacity. These projections seem to assume that spent fuel from imported light-water reactors fueled with imported uranium will be reprocessed and the plutonium extracted will also be used to provide startup fuel for breeder reactors.

These projections are primarily based on assumptions about the doubling time, the time it would take a breeder reactor to produce enough plutonium to fuel a new breeder reactor core. Since MOX fueled reactors have lower breeding ratios, by 2020 the DAE plans to switch to constructing breeders that use metallic fuel, which could have a much higher breeding ratio.53 A higher breeder ratio will result in a shorter doubling time. The rate of growth also depends sensitively on the out-of-pile time, the time period taken for the spent fuel to be cooled, reprocessed, and fabricated into fresh fuel. The DAE optimistically assumes that all of this can be accomplished within one year.54

The DAE’s methodology is flawed, however, and does not account correctly for plutonium flows.55 To start with, the base capacity of metallic fueled breeder reactors (MFBRs) assumed for 2022 of 6 GWe, which is necessary for the 2052 projection, would require approximately 22 tons of fissile plutonium for startup fuel. The DAE does not have enough reprocessing capacity currently to handle all the spent fuel produced by the heavy water reactors that are operating and under construction. Even if the DAE does manage to inexplicably obtain the necessary plutonium to construct a MFBR capacity of 6 GWe with some to spare, under the DAE’s assumed rate of growth, the plutonium stockpile would decline by approximately 40 tons just in the first ten years, even with an optimistic one year out of pile time. This is due to a three year lag between the time a certain amount of plutonium is committed to a breeder reactor and additional plutonium, which could be used as startup fuel for a new breeder reactor, is produced by reprocessing the irradiated spent fuel containing the initial plutonium.

A more careful calculation that takes into account the plutonium flow constraints shows that the capacity for MFBRs based on plutonium from the DAE’s heavy water reactor fleet will drop from the projected 199 GWe to 78 GWe by 2052.56 If the out-of-pile time were projected to be a more realistic three years, the MFBR capacity in 2052 based on plutonium from PHWRs will drop to 34 GWe.

While these figures may seem large compared to India’s current nuclear capacity of only 4.1 GWe, they should be viewed in relation to the projected requirements, under business-as-usual conditions, of approximately 1300 GWe total generating capacity by mid-century. Further, the only constraint assumed here is fissile material availability. It assumes that there will be no delays due to infrastructure and manufacturing problems, economic disincentives due to the high cost of breeder electricity, or accidents. All of these are realistic constraints and render even the lower end of the 2052 projections quite unrealistic.

Conclusion

Breeder reactors have always underpinned the DAE’s claims about generating large quantities of cheap electricity necessary for development. Today, more than five decades after those plans were announced, that promise is yet to be fulfilled. As elsewhere, breeder reactors are likely to be unsafe and costly, and their contribution to overall electricity generation will be modest at best.

Construction of France’s first experimental sodium-cooled reactor, Rapsodie, started in 1962 and it went critical on 28 January 1967 with a nominal capacity of 20 megawatts thermal (MWt). At the end of 1967, its power was increased to 24 MWt, and in 1970, after core redesign, to 40 MWt. Its operating power was reduced to 22 MWt in June 1980 to minimize the thermal stresses thought to be the source of cracks in the reactor vessel. The reactor operated until April 1983, when it was shut down permanently.

Rapsodie was a loop-type reactor, with the heat exchanger between the primary and secondary sodium loops outside the reactor vessel. It was as close as possible to the basic design imagined for commercial applications (molten-sodium coolant, reactor material, power density, etc). The core contained 31.5 kilograms (kg) of plutonium-239 and 79.5 kg of uranium-235. The mean duration of reactor runs was 80 days and the fuel reached burn-ups of 102,000 MWd/t.2

Tatsujiro Suzuki

This paper reviews the history, status, and probable future of fast reactor and associated fuel cycle development in Japan. The fast breeder reactor and its closed fuel cycle have been the cornerstone of Japan’s nuclear-energy development program since the 1950s. For economic, technological, and political reasons, Japan’s development and implementation of these technologies are significantly delayed. The budget for fast breeder reactor development has steadily declined since the mid-1990s, and its commercialization target has slipped from the 1980s to the 2050s. An accident at the Monju prototype reactor contributed to these delays and triggered a fundamental shift from research and development (R&D) and early commercialization to an emphasis on advanced fuel cycles.

Nevertheless, Japan is still committed to fast-reactor development. This paper examines the motivation for its continued commitment to a fast reactor program and concludes that several non-technological factors, such as bureaucratic inertia, commitments to local communities, and an absence of R&D oversight have contributed to this entrenched position. Japan is currently reorganizing its R&D programs with the goal of operating a demonstration breeder reactor by about 2025. This effort is in response to the government sponsored Nuclear Power National Plan and the Bush Administration’s Global Nuclear Energy Partnership Program (GNEP). Breeder R&D programs face significant obstacles, such as plutonium-stockpile management, spent fuel management, fuel cycle technologies, and arrangements for cost and risk sharing between industry, national and local governments. As a result, it is likely that fast breeder reactor programs will continue to slip.