Nuclear fuels (CASE STUDY)

Nuclear fuels 

Introduction

Nuclear fuel is a material that can be 'consumed' by nuclear fission or fusion to derive nuclear energy. Nuclear fuel can refer to the fuel itself, or to physical objects (for example bundles composed of fuel rods) composed of the fuel material, mixed with structural, neutron moderating, or neutron reflecting materials.

Most nuclear fuels contain heavy fissile elements that are capable of nuclear fission. When these fuels are struck by neutrons, they are in turn capable of emitting neutrons when they break apart. This makes possible a self-sustaining chain reaction that releases energy with a controlled rate in a nuclear reactor or with a very rapid uncontrolled rate in a nuclear weapon.

The most common fissile nuclear fuels are

1-      Uranium-235 (²³⁵U)

2-       Plutonium-239 (²³⁹Pu).

The actions of mining, refining, purifying, using, and ultimately disposing of nuclear fuel together make up the nuclear fuel cycle.

Not all types of nuclear fuels create power from nuclear fission. Plutonium-238 and some other elements are used to produce small amounts of nuclear power by radioactive decay in radioisotope thermoelectric generators and other types of atomic batteries.

 Also, light nuclides such as tritium (³H) can be used as fuel for nuclear fusion.

Nuclear fuel has the highest energy density of all practical fuel sources.

• Uranium is a very heavy metal which can be used as an abundant source of concentrated energy. 
• Uranium occurs in most rocks in concentrations of 2 to 4 parts per million and is as common in the Earth's crust as tin, tungsten and molybdenum. Uranium occurs in seawater, and can be recovered from the oceans. 
• Uranium was discovered in 1789 by Martin Klaproth, a German chemist, in the mineral called pitchblende. It was named after the planet Uranus, which had been discovered eight years earlier.
• Uranium was apparently formed in supernova about 6.6 billion years ago. While it is not common in the solar system, today its slow radioactive decay provides the main source of heat inside the Earth, causing convection and continental drift. 
• The high density of uranium means that it also finds uses in the keels of yachts and as counterweights for aircraft control surfaces, as well as for radiation shielding.
• Uranium has a melting point is 1132°C. The chemical symbol for uranium is U.

The Uranium Atom

On a scale arranged according to the increasing mass of their nuclei, uranium is one of the heaviest of all the naturally-occurring elements (Hydrogen is the lightest). Uranium is 18.7 times as dense as water.

Like other elements, uranium occurs in several slightly differing forms known as 'isotopes'. These isotopes differ from each other in the number of uncharged particles (neutrons) in the nucleus. Natural uranium as found in the Earth's crust is a mixture largely of two isotopes: uranium-238 (U-238), accounting for 99.3% and uranium-235 (U-235) about 0.7%. 

Figure (2)

(Uranium metal highly enriched in uranium-235)

The isotope U-235 is important because under certain conditions it can readily be split, yielding a lot of energy. It is therefore said to be 'fissile' and we use the expression 'nuclear fission'.

Meanwhile, like all radioactive isotopes, they decay. U-238 decays very slowly, its half-life being about the same as the age of the Earth (4500 million years). This means that it is barely radioactive, less so than many other isotopes in rocks and sand. Nevertheless it generates 0.1 watts/tonne as decay heat and this is enough to warm the Earth's core. U-235 decays slightly faster.

Energy from the uranium atom

The nucleus of the U-235 atom comprises 92 protons and 143 neutrons (92 + 143 = 235). When the nucleus of a U-235 atom captures a moving neutron it splits in two (fissions) and releases some energy in the form of heat, also two or three additional neutrons are thrown off. If enough of these expelled neutrons cause the nuclei of other U-235 atoms to split, releasing further neutrons, a fission 'chain reaction' can be achieved. When this happens over and over again, many millions of times, a very large amount of heat is produced from a relatively small amount of uranium.

It is this process, in effect "burning" uranium, which occurs in a nuclear reactor. The heat is used to make steam to produce electricity.

 Figure (3)

Nuclear power stations and fossil-fuelled power stations of similar capacity have many features in common. Both require heat to produce steam to drive turbines and generators. In a nuclear power station, however, the fissioning of uranium atoms replaces the burning of coal or gas.

Types of Nuclear fuels

1-Oxide fuel

For fission reactors, the fuel (typically based on uranium) is usually based on the metal oxide; the oxides are used rather than the metals themselves because the oxide melting point is much higher than that of the metal and because it cannot burn, being already in the oxidized state.

1.1 UOX

Uranium dioxide is a black semiconductor solid. It can be made by reacting uranyl nitrate with a base (ammonia) to form a solid (ammonium uranate). It is heated (calcined) to form U₃O₈ that can then be converted by heating in an argon / hydrogen mixture (700 °C) to form UO₂. The UO₂ is then mixed with an organic binder and pressed into pellets, these pellets are then fired at a much higher temperature (in H₂/Ar) to sinter the solid. The aim is to form a dense solid which has few pores.

The thermal conductivity of uranium dioxide is very low compared with that of zirconium metal, and it goes down as the temperature goes up.sIt is important to note that the corrosion of uranium dioxide in an aqueous environment is controlled by similar electrochemical processes to the galvanic corrosion of a metal surface.

Figure (4 )

1.2 MOX

Mixed oxide, or MOX fuel, is a blend of plutonium and natural or depleted uranium which behaves similarly (though not identically) to the enriched uranium feed for which most nuclear reactors were designed. MOX fuel is an alternative to low enriched uranium (LEU) fuel used in the light water reactors which predominate nuclear power generation.

Some concern has been expressed that used MOX cores will introduce new disposal challenges, though MOX is itself a means to dispose of surplus plutonium by transmutation.

Currently (March, 2005) reprocessing of commercial nuclear fuel to make MOX is done in England and France, and to a lesser extent in Russia, India and Japan. China plans to develop fast breeder reactors and reprocessing.

Figure (5)

2-Metal fuel

Metal fuels have the advantage of much higher heat conductivity than oxide fuels but cannot survive equally high temperatures. Metal fuels have a long history of use, stretching from the Clementine reactor in 1946 to many test and research reactors. Metal fuels have the potential for the highest fissile atom density. Metal fuels are normally alloyed, but some metal fuels have been made with pure uranium metal. Uranium alloys that have been used include

1-      Uranium aluminum

2-      Uranium zirconium

3-      Uranium silicon

4-      Uranium molybdenum

5-      Uranium zirconium hydride

 Any of the aforementioned fuels can be made with plutonium and other actinides as part of a closed nuclear fuel cycle. Metal fuels have been used in water reactors and liquid metal fast breeder reactors, such as EBR-II.

2.1 TRIGA fuel

TRIGA fuel is used in TRIGA (Training, Research, Isotopes, General Atomics) reactors. The TRIGA reactor uses uranium-zirconium-hydride (UZrH) fuel, which has a prompt negative temperature coefficient, meaning that as the temperature of the core increases, the reactivity decreases—so it is highly unlikely for a meltdown to occur. Most cores that use this fuel are "high leakage" cores where the excess leaked neutrons can be utilized for research. TRIGA fuel was originally designed to use highly enriched uranssium, however in 1978 the U.S. Department of Energy launched its Reduced Enrichment for Research Test Reactors program, which promoted reactor conversion to low-enriched uranium fuel. A total of 35 TRIGA reactors have been installed at locations across the USA. A further 35 reactors have been installed in other countries.

2.2Actinide fuel

In a fast neutron reactor, the minor actinides produced by neutron capture of uranium and plutonium can be used as fuel. Metal actinide fuel is typically an alloy of zirconium, uranium, plutonium and the minor actinides. It can be made inherently safe as thermal expansion of the metal alloy will increase neutron leakage.

3-Ceramic fuels

Ceramic fuels other than oxides have the advantage of high heat conductivities and melting points, but they are more prone to swelling than oxide fuels and are not understood as well.

3.1Uranium nitride

This is often the fuel of choice for reactor designs that NASA produces, one advantage is that UN has a better thermal conductivity than UO₂. Uranium nitride has a very high melting point. This fuel has the disadvantage that unless ¹⁵N was used (in place of the more common ¹⁴N) that a large amount of ¹⁴C would be generated from the nitrogen by the (n,ps) reaction. As the nitrogen required for such a fuel would be so expensive it is likely that the fuel would have to be reprocessed by a pyro method to enable to the ¹⁵N to be recovered. It is likely that if the fuel was processed and dissolved in nitric acid that the nitrogen enriched with ¹⁵N would be diluted with the common ¹⁴N.

 

3.2Uranium carbide

Much of what is known about uranium carbide is in the form of pin-type fuel elements for liquid metal fast breeder reactors during their intense study during the '60s and '70s. However, recently there has been a revived interest in uranium carbide in the form of plate fuel and most notably, micro fuel particles (such as TRISO particles).

The high thermal conductivity and high melting point makes uranium carbide an attractive fuel. In addition, because of the absence of oxygen in this fuel (during the course of irradiation, excess gas pressure can build from the formation of O₂ or other gases) as well as the ability to complement a ceramic coating (a ceramic-ceramic interface has structural and chemical advantages), uranium carbide could be the ideal fuel candidate for certain Generation IV reactors such as the gas-cooled fast reactor.

4-Liquid fuels

Liquid fuels are liquids containing dissolved nuclear fuel. Liquid-fueled reactors generally have large negative feedback mechanisms and therefore are particularly stable designs; however the liquid fuel form also has the disadvantage of being easily dispersible in the event of an accident, such as a leak in the primary system.

4.1Molten salts

Molten salt fuels have nuclear fuel dissolved directly in the molten salt coolant. Molten salt-fueled reactors, such as the liquid fluoride thorium reactor (LFTR), are different than molten salt-cooled reactors that do not dissolve nuclear fuel in the coolant.s

Molten salt fuels were used in the LFTR known as the Molten Salt Reactor Experiment, as well as other liquid core reactor experiments. The liquid fuel for the molten salt reactor was a mixture of lithium, beryllium, thorium and uranium fluorides: LiF-BeF₂-ThF₄-UF₄ (72-16-12-0.4 mol%). It had a peak operating temperature of 705°C in the experiment, but could have operated at much higher temperatures, since the boiling point of the molten salt was in excess of 1400°C.

4.2Aqueous solutions of uranyl salts

The aqueous homogeneous reactors (AHRs) use a solution of uranyl sulfate or other uranium salt in water. Historically, AHRs have all been small research reactors, not large power reactors. An AHR known as the Medical Isotope Production System is being considered for production of medical isotopes.^s

5-Common physical forms of nuclear fuel

Uranium dioxide (UO₂) powder is compacted to cylindrical pellets and sintered at high temperatures to produce ceramic nuclear fuel pellets with a high density and well defined physical properties and chemical composition. A grinding process is used to achieve a uniform cylindrical geometry with narrow tolerances. Such fuel pellets are then stacked and filled into the metallic tubes. The metal used for the tubes depends on the design of the reactor. Stainless steel was used in the past, but most reactors now use a zirconium alloy which, in addition to being highly corrosion-resistant, has low neutron absorption. The tubes containing the fuel pellets are sealed: these tubes are called fuel rods. The finished fuel rods are grouped into fuel assemblies that are used to build up the core of a power reactor.

Cladding is the outer layer of the fuel rods, standing between the coolant and the nuclear fuel. It is made of a corrosion-resistant material with low absorption cross section for thermal neutrons, usually Zircaloy or steel in modern constructions, or magnesium with small amount of aluminium and other metals for the now-obsolete Magssnox reactors. Cladding prevents radioactive fission fragments from escaping the fuel into the coolant and contaminating it.

• 

Figure (6)  NRC Image of fresh fuel pellets ready for assembly.

5.1PWR fuel

PWR fuel assembly (also known as a fuel bundle) This fuel assembly is from a pressurized water reactor of the nuclear-powered passenger and cargo ship NS Savannah. Designed and built by the Babcock and Wilcox Company.

Figure (7)

Pressurized water reactor consists of cylindrical rods put into bundles. A uranium oxide ceramic is formed into pellets and inserted into Zircaloy tubes that are bundled together. The Zircaloy tubes are about 1 cm in diameter, and the fuel cladding gap is filled with helium gas to improve the conduction of heat from the fuel to the cladding. There are about 179-264 fuel rods per fuel bundle and about 121 to 193 fuel bundles are loaded into a reactor core. Generally, the fuel bundles consist of fuel rods bundled 14×14 to 17×17. PWR fuel bundles are about 4 meters long. In PWR fuel bundles, control rods are inserted through the top directly into the fuel bundle. The fuel bundles usually are enriched several percent in ²³⁵U. The uranium oxide is dried before inserting into the tubes to try to eliminate moisture in the ceramic fuel that can lead to corrosion and hydrogen embrittlement. The Zircaloy tubes are pressurized with helium to try to minimize pellet-cladding interaction which can lead to fuel rod failure over long periods.

Fuel Assembly

PWR fuel assemblies differ from BWR assemblies in that the control rods (called Rod Control Cluster Assemblies) have 16 to 20 rods and enter tubes in the assembly.The PWR fuel assembly usually has a fuel rod arrangement of 14 x 14 up to 17 x 17. Click for a 275K detailed illustration of a Westinghouse assembly.

5.2BWR fuel

In boiling water reactors (BWR), the fuel is similar to PWR fuel except that the bundles are "canned"; that is, there is a thin tube surrounding each bundle. This is primarily done to prevent local density variations from affecting neutronics and thermal hydraulics of the reactor core. In modern BWR fuel bundles, there are either 91, 92, or 96 fuel rods per assembly depending on the manufacturer. A range between 368 assemblies for the smallest and 800 assemblies for the largest U.S. BWR forms the reactor core. Each BWR fuel rod is back filled with helium to a pressure of about three atmospheres (300 kPa).

  Figure (8)

5.3CANDU fuels

  Figure (9)

CANDU fuel bundles Two CANDU ("CANada Deuterium Uranium") fuel bundles, each about 50 cm long, 10 cm in diameter. Photo courtesy of Atomic Energy of Canada Ltd.

CANDU fuel bundles are about a half meter long and 10 cm in diameter. They consist of sintered (UO₂) pellets in zirconium alloy tubes, welded to zirconium alloy end plates. Each bundle is roughly 20 kg, and a typical core loading is on the order of 4500-6500 bundles, depending on the design. Modern types typically have 37 identical fuel pins radially arranged about the long axis of the bundle, but in the past several different configurations and numbers of pins have been used. The CANFLEX bundle has 43 fuel elements, with two element sizes. It is also about 10 cm (4 inches) in diameter, 0.5 m (20 in) long and weighs about 20 kg (44 lb) and replaces the 37-pin standard bundle. It has been designed specifically to increase fuel performance by utilizing two different pin diameters. Current CANDU designs do not need enriched uranium to achieve criticality (due to their more efficient heavy water moderator), however, some newer concepts call for low enrichment to help reduce the size of the reactors.

6-Less common fuel forms

Various other nuclear fuel forms find use in specific applications, but lack the widespread use of those found in BWRs, PWRs, and CANDU power plants. Many of these fuel forms are only found in research reactors, or have military applications.

6.1 Magnox fuel

Figure (10) A magnox fuel rod

Magnox reactors are pressurised, carbon dioxide cooled, graphite moderated reactors using natural uranium (i.e. unenriched) as fuel and magnoxs alloy as fuel cladding. Working pressure varies from 6.9 to 19.35 bar for the steel pressure vessels, and the two reinforced concrete designs operated at 24.8 and 27 bar. Magnox is also the name of an alloy—mainly of magnesium with small amounts of aluminium and other metals—used in cladding unenriched uranium metal fuel with a non-oxidising covering to contain fission products. Magnox is short for Magnesium non-oxidising. This material has the advantage of a low neutron capture cross-section, but has two major disadvantages:

• It limits the maximum temperature, and hence the thermal efficiency, of the plant.
• It reacts with water, preventing long-term storage of spent fuel under water.

Magnox fuel incorporated cooling fins to provide maximum heat transfer despite low operating temperatures, making it expensive to produce. While the use of uranium metal rather than oxide made reprocessing more straightforward and therefore cheaper, the need to reprocess fuel a short time after removal from the reactor meant that the fission product hazard was severe. Expensive remote handling facilities were required to address this danger.

                Figure (11)

particle which has been cracked, showing the multiple coating layers

6.2 TRISO Fuels

Tristructural-isotropic (TRISO) fuel is a type of micro fuel particle. It consists of a fuel kernel composed of UO_X (sometimes UC or UCO) in the center, coated with four layers of three isotropic materials. The four layers are a porous buffer layer made of carbon, followed by a dense inner layer of pyrolytic carbon (PyC), followed by a ceramic layer of SiC to retain fission products at elevated temperatures and to give the TRISO particle more structural integrity, followed by a dense outer layer of PyC. TRISO fuel particles are designed not to crack due to the stresses from processes (such as differential thermal expansion or fission gas pressure) at temperatures up to and beyond 1600°C, and therefore can contain the fuel in the worst of accident scenarios in a properly designed reactor. Two such reactor designs are the pebble bed reactor (PBR), in which thousands of TRISO fuel particles are dispersed into graphite pebbles, and the prismatic-block gas-cooled reactor (such as the GT-MHR), in which the TRISO fuel particles are fabricated into compacts and placed in a graphite block matrix. Both of these reactor designs are very high temperature reactors (VHTR) [formally known as the high-temperature gas-cooled reactors (HTGR)], one of the six classes of reactor designs in the Generation IV interactive.

Figure (12)

TRISO fuel particles were originally developed in Germany for high-temperature gas-cooled reactors. The first nuclear reactor to use TRISO fuels was the AVR and the first powerplant was the THTR-300. Currently, TRISO fuel compacts are being used in the experimental reactors, the HTR-10 in China, and the HTTR in Japan.

6.3QUADRISO fuel

In QUADRISO particles a burnable neutron poison (europium oxide or erbium oxide or carbide) layer surrounds the fuel kernel of ordinary TRISO particles to better manage the excess of reactivity. If the core is equipped both with TRISO and QUADRISO fuels, at beginning of life neutrons do not reach the fuel of the QUADRISO particles because they are stopped by the burnable poison. After irradiation, the poison depletes and neutrons stream into the fuel kernel of QUADRISO particles inducing fission reactions. This mechanism compensates fuel depletion of ordinary TRISO fuel. In the generalized QUADRISO fuel concept the poison can eventually be mixed with the fuel kernel or the outer pyrocarbon. The QUADRISO [1] concept has been conceived at Argonne National Laboratory.

Figure (13)

QUADRISO Particle

6.4 RBMK fuel

RBMK reactor fuel was used in Soviet designed and built RBMK type reactors. This is a low enriched uranium oxide fuel. The fuel elements in an RBMK are 3 m long each, and two of these sit back-to-back on each fuel channel, pressure tube. Reprocessed uranium from Russian VVER reactor spent fuel is used to fabricate RBMK fuel. Following the Chernobyl accident, the enrichment of fuel was changed from 2.0% to 2.4%, to compensate for control rod modifications and the introduction of additional absorbers.

6.5CerMet fuel

CerMet fuel consists of ceramic fuel particles (usually uranium oxide) embedded in a metal matrix. It is hypothesized^{[by whom?]} that this type of fuel is what is used in United States Navy reactors. This fuel has high heat transport characteristics and can withstand a large amount of expansion.

7 Radioisotope decay fuels

7.1 Atomic battery

The terms atomic battery, nuclear battery and radioisotope battery are used interchangely to describe a device which uses the radioactive decay to generate electricity. These systems use radioisotopes that produce low energy beta particles or sometimes alpha particles of varying energies. Low energy beta particles are needed to prevent the production of high energy penetrating Bremsstrahlung radiation that would require heavy shielding. Radioisotopes such as tritium, nickel-63, promethium-147, and technetium-99 have been tested. Plutonium-238, curium-242, curium-244 and strontium-90 have been used.

There are two main categories of atomic batteries: thermal and non-thermal. The non-thermal atomic batteries, which have many different designs, exploit charged alpha and beta particles. These designs include the direct charging generators, betavoltaics, the optoelectric nuclear battery, and the radioisotope piezoelectric generator. The thermal atomic batteries on the other hand, convert the heat from the radioactive decay to electricity. These designs include thermionic converter, thermophotovoltaic cells, alkali-metal thermal to electric converter, and the most common design, the radioisotope

8-Fusion fuels

Fusion fuels include tritium (³H) and deuterium (²H) as well as helium-3 (³He). Many other elements can be fused together, but the larger electrical charge of their nuclei means that much higher temperatures are required. Only the fusion of the lightest elements is seriously considered as a future energy source. Although the energy density of fusion fuel is even higher than fission fuel, and fusion reactions sustained for a few minutes have been achieved, utilizing fusion fuel as a net energy source remains a theoretical possibility.

First generation fusion fuel

Deuterium and tritium are both considered first-generation fusion fuels; they are the easiest to fuse, because the electrical charge on their nuclei is the lowest of all elements. The three most commonly cited nuclear reactions that could be used to generate energy are:

²H + ³H n (14.07 MeV) + ⁴He (3.52 MeV)

²H + ²H n (2.45 MeV) + ³He (0.82 MeV)

²H + ²H p (3.02 MeV) + ³H (1.01 MeV)

Second generation fusion fuel

Second generation fuels require either higher confinement temperatures or longer confinement time than those required of first generation fusion fuels, but generate fewer neutrons. Neutrons are an unwanted byproduct of fusion reactions in an energy generation context, because they are absorbed by the walls of a fusion chamber, making them radioactive. They cannot be confined by magnetic fields, because they are not electrically charged. This group consists of deuterium and helium-3. The products are all charged particles, but there may be significant side reactions leading to the production of neutrons.

²H + ³He p (14.68 MeV) + ⁴He (3.67 MeV)

Third generation fusion fuel

 Aneutronic fusion

Third generation fusion fuels produce only charged particles in the primary reactions, and side reactions are relatively unimportant. Since a very small amount of neutrons is produced, there would be little induced radioactivity in the walls of the fusion chamber. This is often seen as the end goal of fusion research. ³He has the highest Maxwellian reactivity of any 3rd generation fusion fuel. However, there are no significant natural sources of this substance on Earth.

³He + ³He 2p + ⁴He (12.86 MeV)

Another potential aneutronic fusion reaction is the proton-boron reaction:

p + ¹¹B → 3⁴He

Under reasonable assumptions, side reactions will result in about 0.1% of the fusion power being carried by neutrons. With 123 keV, the optimum temperature for this reaction is nearly ten times higher than that for the pure hydrogen reactions, the energy confinement must be 500 times better than that required for the D-T reaction, and the power density will be 2500 times lower than for

 9- Nuclear Fuel Cycle

Types of Fuel Cycle

The nuclear fuel cycle is the progression of steps in the utilization of fissile materials, from the initial mining of the uranium (or thorium) through the final disposition of the material removed from the reactor. It is called a cycle because in the general case, some of the material taken from a reactor may be used again, or “recycled.”

Fuel cycles differ in the nature of the fuel used, the fuel’s history in the reactor, and the manner of handling the fuel that is removed from the reactor at the end of the fuel’s useful life (known as the spent fuel). For uranium-fuels reactors—which means virtually all commercial reactors—a key difference is in the disposition of the plutonium and other actinides that are produced in a chain of neutron captures and beta decays that starts with neutron capture in 238U to produce 239Pu (see Section 7.4).1 The actinides are important because (1) some, especially 239Pu, are fissile and can be used as nuclear fuel complicating the problems of nuclear waste disposal. The three broad fuel cycle categories are as follows:

1 The actinides are the elements with atomic numbers Z greater than or equal to that of actinium (Z = 89). (The terminology is not uniform and, sometimes, actinium is not included among the “actinides.”) Neptunium (Z = 93), americium

(Z = 95), and curium (Z = 96) are referred to as minor actinides in view of their low abundance in spent fuel compared to uranium (Z = 92) and plutonium

(Z = 94). Elements with atomic numbers greater than 92 are termed transuranics

elements.

Figure (14)

Once-through fuel cycle. This is sometimes called an open fuel cycle or a “throw-away” cycle. It is not really a cycle, in that the spent fuel is treated as waste when it is removed from the reactor and is not used further. The

239Pu and other actinides are part of these wastes.

 Reprocessing fuel cycle. In the present standard reprocessing fuel cycle, plutonium and uranium are chemically extracted from the spent fuel. The plutonium is used to make additional fuel, often by mixing it with uranium oxides to produce mixed-oxide fuel (MOX) for use in thermal reactors.

This provides additional energy and changes the nature of the wastes. In potential variants of the reprocessing fuel cycle, the minor actinides would also be extracted, and they and the plutonium would be incorporated in fresh fuel for fast reactors .

◆ Breeding cycle. For this cycle, the reactor is designed so that there is more fissile material (mostly 239Pu) in the spent fuel than there was in the fuel put into the reactor (see Section 8.3). As in the reprocessing fuel cycle, the plutonium can be removed and be used in another reactor. With a sequence of such steps, fission energy is in effect extracted from a substantial fraction of the 238U in uranium, not just from the small 235U component, increasing the energy output from a given amount of uranium by a factor that could, in principle, approach 100.

It may be noted that uranium accounts for most of the mass of the nuclear wastes in the once-through cycle. It is separated out in the reprocessing and breeding cycles for possible reuse in reactor fuel.

At present, all U.S. commercial reactors and the majority of reactors worldwide are operating with a once-through fuel cycle, although some countries, particularly France, have large-scale reprocessing programs with use of plutonium in the form of MOX fuel. It should be noted, of course, that even in the once-through fuel cycle, the potential for eventually using the fuel in a reprocessing cycle remains until the fuel is disposed of irretrievably. No country is employing a breeder cycle at this time, although France appeared on the verge of attempting such a program with its Phenix and Superphenix reactors—but this effort has been abandoned, at least for the time being (see Section 8.3.3).

Although virtually all of the world’s commercial reactors have used uranium fuel, there is continuing interest in the use of thorium fuel.2 In a thorium fuel cycle, the thorium (all 232Th in nature) serves as the fertile fuel. Neutron capture and beta decay result in the production of 233U, which has favorable properties as a fissile fuel. To start the thorium cycle, a fissile material such as

235U or 239Pu is needed, but once begun, it can be sustained if enough 233U is produced to at least replace the initial fissile material. It is often argued that a thorium cycle is preferable to a uranium cycle, because if 233U is ex-

2 The Fort St. Vrain high-temperature, gas-cooled, graphite-moderated reactor in

Colorado, which was shut down in 1989, is one of several exceptions to the exclusive

Front End of the Fuel Cycle

tracted from the spent fuel, it can be “denatured” by mixing it with natural uranium to make a fuel that cannot be used in a bomb. Bomb material could be obtained only after the isotopic separation of 233U. In contrast, bomb material can be obtained from a uranium-fueled reactor by chemical separation of the plutonium (see Chapter 17). Isotopic separation is technically more difficult than chemical separation; thus, a thorium fuel cycle could be more proliferation resistant than a uranium fuel cycle unless, in the latter case, the plutonium is well protected from diversion or theft.

 Steps in the Nuclear Fuel Cycle

The steps in the fuelcycle that precede the introduction of the fuel into the reactor are referred to as the front end of the fuel cycle. Those that follow the removal of the fuel from the reactor comprise the back end of the fuel cycle. At present, there is only a truncated back end to the fuel cycle in the United States, as virtually all commercial spent fuel is accumulating in cooling pools or storage casks at the reactor sites.

Implementation of a spent fuel disposal plan, or of a reprocessing and waste disposal plan, would represent the “closing” of the fuel cycle. This closing is viewed by many to be an essential condition for the increased use of nuclear power in the United States—and perhaps even for its continued use beyond the next several decades.

Key aspects of the fuel cycle will be surveyed in the remainder of this chapter. The fuel cycle will be discussed particularly in the context of light water reactors, in view of their dominance among world nuclear reactors. The main aspects are relevant to other types of reactor as well. A more extensive   treatment of the crucial step of waste disposal will be given in Chapters 10–13.

                                      

10 Uranium Mining and Milling

Uranium Deposits in the Earth’s Crust

The concentration of uranium varies greatly among geological formations.

The average concentration in the Earth’s crust is about 3 parts per million

(ppm) by weight, but extremes extend from under 1 ppm to something in the neighborhood of 500,000 ppm.3

Uranium resources are widely distributed, with substantial uranium production in many countries, including Australia, Canada, Kazakhstan, Namibia,

3 For example, one deposit in Canada is identified as having zones of “over 50%

U3O8,” which translates to over 42% uranium [3].

Niger, the Russian Federation, South Africa, the United States, and Uzbekistan

[4, p. 36]. Most of the uranium now used in the United States is imported, with Canada being the largest supplier. Through 2000, the United States had been the world’s leader in cumulative production of uranium, with Canada a close second. However, the U.S. share of production declined by the 1990s, and now Canada and Australia are the world’s main suppliers of uranium.

Together, they accounted for about 50% of world production in 2000 [4].

Uranium in rocks is mostly in the form of a uranium oxide, U3O8. In “conventional” mining, the rock is extracted from open pit or underground mines, the U3O8 is then extracted in the milling process by crushing the rock and leaching with acid, and the U3O8 is then recovered from the liquid and dried. The concentrated U3O8 is known as yellowcake.4 In an “unconventional” method, appropriate for only certain types of uranium deposits, U3O8 is extracted by in situ leaching (i.e., by pumping a leaching agent through the ore without physical removal of the rock).5

Figure (15)

At low concentrations, the uranium content is expressed in terms of the uranium grade, given in percent by weight of either uranium or U3O8. Thus, ore which is 1000 ppm of uranium corresponds to grades of 0.100% U or 0.118%

U3O8.6 In an extensive 1983 study of U.S. uranium ores, deposits were listed with U3O8 grades ranging from under 0.01% to over 1.8%, with a median of about 0.1% [6, p. 39]. The higher the grade, the less the amount of ore that must be extracted, which, in general, leads to lower costs. Ores below a grade of 0.05% are considered low-grade ores and have not been widely needed. Of course, the ultimate criterion is overall cost, not grade per se, and at one time open-pit mining utilized ores down to 0.04% [7, p. 411].

During 2001, most of the uranium extraction in the United States was done by in situ leaching, using ores ranging in grade from 0.09% to about 0.20% U3O8 [8]. Worldwide, conventional mining dominates but is now economically practical only at higher uranium grades (“above a few tenths of a percent”)

Radon Exposures from Uranium Mining and Mill Tailings

In the early days of uranium mining, little attention was paid to radiation safety. In the Middle Ages, long before uranium had been identified as an element, metal miners in southern Germany and Czechoslovakia contracted lung ailments, called Bergkrankheit (“mountain sickness”). Modern scientists have attributed the ailment to lung cancer caused by a high uranium concentration

4 U3O8 is not yellow in its pure form. Yellowcake is about 85% U3O8 [5, p. 241], and the yellow color results from another uranium compound in the ore.

5 The designations “conventional” and “unconventional” correspond to those,

6 In international usage, “grade” usually refers to U content, whereas in U.S. DOE documents it refers to U3O8 content. Note: U3O8 is 84.8% uranium and 15.2% oxygen, by weight. that, by chance, was in the rock formations being mined. The decay of the radionuclides in the uranium series proceeds from 238U through several steps to 226Ra and then to radon gas (222Rn) and its radioactive progeny. Inhalation of these “radon daughters” can lead to lung cancer (see Section 3.5.1).

As one would expect, the problem of radon exposure is more extreme in uranium mines than in other sorts of mine. It became a particularly serious problem in a number of countries—for example, in the United States,

Czechoslovakia, and Canada—when large-scale uranium mining was begun in the 1940s to meet the demands of nuclear weapon and nuclear power programs.

By the late 1950s, steps were initiated in the United States to reduce radon exposures, mainly through better ventilation, and by the 1970s, the average exposures of uranium miners had become quite low (lower than that from indoor radon in many homes). However, a good deal of damage had already been done, and there is unambiguous evidence of increased lung cancer fatalities among uranium miners.

The residues of the milling operation, representing the remainder of the ore after extraction of the U3O8, are the mill tailings. All of the uranium progeny, starting with 230Th, are present in the tailings.7 The radionuclide 230Th has a half-life of 75,400 years and thus sustains the remainder of the uranium series for a long period of time. This results in the continuous production of radon, some escaping to the atmosphere. Of course, these steps do not increase the rate of radon production above what it would have been without mining, but the radon in the tailings can more readily reach the atmosphere than can radon in underground ore. At one time, this was viewed by some as constituting an important environmental hazard, and it is still deemed necessary to take remedial measures to limit radon emissions from the tailings (using overlying layers of material to impede radon escape). However, interest in the issue has diminished as it has become obvious that exposures from “normal” indoor radon pose a much more serious problem, in terms both of the number of people impacted and the magnitudes of the radon concentrations to which they are exposed.8

11 Enrichment of Uranium

Preparation for Enrichment: Conversion

There are a variety of approaches to the enrichment of uranium, each taking advantage of the small mass difference between 235U and 238U. In the most used of these processes, it is necessary to have the uranium in gaseous form. For that purpose, the U3O8 is chemically converted to gaseous uranium hexafluoride,

UF6. This is the compound of choice, because UF6 is a gas at lower

7 The 234U remains in the yellowcake and the radionuclides between 238U and 234U in the uranium series are short-lived.

8 For a comparison of the hazards from mill tailings and indoor radon,

Degrees of Enrichment

Natural uranium has an isotopic abundance by number of atoms of 0.0055%

234U, 0.720% 235U and 99.275% 238U.9 In the remainder of the discussion of uranium isotopic enrichment, we will follow the standard practice of describing the 235U fraction in terms of mass rather than, as is common in many other scientific applications, of number of atoms.10 For natural uranium, the 235U abundance by mass is 0.711%. The presence of the small amount of 234U is often ignored, because corrections on the order of 10−4 or less are irrelevant.

The fissile nuclide in thermal reactors is 235U. For reactors that require uranium with a higher fraction of 235U than is found in natural uranium, enrichment is necessary. This is, of course, the case for light water reactors

(LWRs). Fuel used in LWRs in past years has been enriched to 235U concentrations ranging from under 2% to over 4%. The anticipated average for the United States, for cumulative production up until about 2010, is 3.0% for

BWR fuel and 3.75% for PWR fuel.11

The material used in LWRs is known as slightly enriched uranium, in contrast to the highly enriched uranium used for nuclear weapons and submarine reactors. Within the core of a given reactor, enrichments vary with the location of the fuel assemblies. As discussed later in the context of the burnup of fuel, there is a general trend toward using fuel with higher initial enrichments.

The products of the enrichment process are the enriched material itself and the depleted uranium, sometimes called enrichment tails. Typically, enrichment tails have in the neighborhood of 0.2% to 0.35% 235U remaining [12,

p. 7]. As one goes to lower concentrations of 235U in the tails, the consumption of uranium ore is reduced, but the cost of enrichment is increased. Thus, there is a trade-off.

The depleted uranium is sometimes used in special applications. Its use in armor-piercing shells, where the high density of uranium is advantageous

(ρ ≈ 19 g/cm3), has led to some public concern about the resulting environmental risks. However, depleted uranium has a lower specific activity than

9 The 234Uarises as a member of the 238U series, with an abundance relative to 238U that is inversely proportional to the half-lives of the two isotopes (2.45 ?105 yr and 4.468 ?109 yr, respectively).

10 These descriptions of isotopic abundance are related by the expression w = [(1−

δ)/(1 − xδ)] x, where, specialized to the case of uranium, w is the ratio of 235U mass to total uranium mass, x is the ratio of the number of 235U atoms to the total number of uranium atoms, and δ is the ratio of the difference between the 238U and 235U atomic masses to the 238U atomic mass. For low enrichments (with

δ = 0.0126 for uranium), w

=. 0.987x, and there is little difference between the two formulations. For natural uranium, x = 0.00720 and w = 0.00711.

11 This is the planning basis for the Yucca Mountain nuclear waste repository  does natural uranium, and there is no evidence of appreciable radiation hazards except for occupants of a closed vehicle that has been struck by a shell that partially vaporizes within it.12s

Methods for Enrichment

The leading enrichment methods in terms of past or anticipated future use are as follows:13

Gaseous diffusion. The average kinetic energy of the molecules in a gas is independent of the molecular weight M of the gas and depends only on the temperature. At the same temperature, the average velocities are therefore

 nversely proportional to

M. For uranium in the form of UF6, the ratio of the velocities of the two isotopic species is 1.0043.14 If a gas sample streams past a barrier with small apertures, a few more 235U molecules than 238U molecules pass through the barrier, slightly increasing the 235U fraction in the gas. The ratio of 235U/238U before and after passing the barrier is the enrichment ratio α. Its ideal or maximum value is given by the velocity ratio α = 1.0043. However, one cannot calculate the number of stages of diffusion needed to achieve a given enrichment merely in terms of powers of α, because the ideal value is not achieved in practice and because it is necessary to continually recycle the less enriched part of the stream. Typically, if one starts with natural uranium (0.71%) and with tails depleted to 0.3%, it is found that about 1200 enrichment stages are required to achieve an enrichment of 4%

Centrifuge separation.

Any fluid—liquid or gaseous—can be separated in a high-speed centrifuge. The centrifugal action causes the heavier component to become more highly concentrated at large radii. As in gaseous diffusion, only a small gain is made in any one stage, and high enrichments of the UF6 are reached using multiple centrifuge stages, with the slightly enriched output of one stage serving as the input to the next one. The centrifuges used for uranium enrichment are rotating cylinders. Uranium that is slightly enriched in 238U (and depleted in 235U) can be extracted from the outer region of the cylinder and returned to an earlier stage in the centrifuge cascade. Uranium slightly enriched in 235U can be extracted

 regions near the center and used as input to the next higher stage in the array of centrifuge units. High enrichments of the UF6 are reached using multiple centrifuge stages. The power requirement for a given degree

of enrichment is much less for centrifuge separation than for diffusion separation.

◆

 Aerodynamic processes. These processes exploit the effects of centrifugal forces, but without a rotating centrifuge. Gas—typically UF6 mixed with

12 Inthis case, direct damage from the shell is a still greater concern.

13 Detailed discussions of these methods are given in, for example, Refs. [13] and

14 The atomic mass of fluorine (F) is 19.00 u. hydrogen—expands through an aperture, and the flow of the resulting gas  stream is diverted by a barrier, causing it to move in a curved path. The more massive molecules on average have a higher radius of curvature than

do the lighter molecules, and a component enriched in 235U is preferentially selected by a physical partition. The process is repeated to obtain successively greater enrichments. The gas nozzle process was developed in Germany as the Becker or jet nozzle process. A variant with a different

geometry for the motion of the gas stream, the so-called Helikon process, has been developed and used in South Africa.

◆Electromagnetic separation.

 When ions in the same charge state are accelerated through the same potential difference, the energy is the same and the radius of curvature in a magnetic field is proportional to

M. Thus, it is possible to separate the different species magnetically. This separation

 can be done with ions of uranium and, so, conversion to UF6 is not, in principle, necessary. Overall, this approach gives a low yield at a high cost in energy, but it has the advantage of employing a relatively straightforward technology.

Laser enrichment. The atomic energy levels of different isotopes differ slightly.15 This effect can be exploited to separate 235U from 238U, starting with uranium in either atomic or molecular form. For example, in the atomic vapor laser isotope separation (AVLIS) method, the uranium is in the form of a hot vapor. Lasers precisely tuned to the appropriate wavelength are used to excite 235U atoms, but not 238U atoms, to energy levels that lie several electron volts above the ground state. An additional laser is used to ionize the excited 235U atoms.16 The ionized 235U atoms can be separated from the un-ionized 238U atoms by electric and magnetic fields. An alternative to the AVLIS method is the SILEX process (separation of isotopes by laser excitation). It is based on the selective dissociation of UF6 (a gas) into UF5 (a solid) [16]. The costs in energy ofs laser enrichment are lower than those of other enrichment methods, but a sophisticated laser technology is required, and, to date, there are no commercial facilities for laser enrichment of uranium. Once mastered, the laser technique is expected to be relatively inexpensive. On the negative

 , there have been fears that if the technique develops sufficiently, laser separation may make it easy for small countries or well-organized terrorist

 to enrich uranium for nuclear weapons.

15 This “isotope effect” was responsible for the discovery of 2H. It arises for two reasons: (1) the atomic energy levels depend on the reduced mass of the electrons, which differs from the mass of a free electron by an amount proportional to me/M, where me is the electron mass and M the atomic mass, and (2) the energy levels of heavy atoms depend in a small measure on the overlap between the wave functions of the innermost electrons and the nucleus, with differences between isotopes due to differences in their nuclear radii.

16 The ionization energy to remove an electron from uranium in its unexcited (ground) state is 6.2 eV.

 Fuel Utilization

 Burnup as a Measure of Fuel Utilization

Thermal Efficiency of U.S. Reactors

The thermal efficiency of a reactor is the ratio of the electrical energy produced to the total heat energy produced. Since 1973, the average thermal efficiency

of U.S. reactors has ranged between 30.6% and 32.1%, according to DOE compilations [23, Table A6]. There has been a gradual improvement with time, and since 1985 it has been above 31.5%, reaching 32.1% for the years

 –2002. We will use the approximate figure of 32% as the nominal average efficiency of LWRs.

19 For a discussion of the details of these processes, see Ref. [22, Section 7.5].

Basic Unit for Burnup: GWDT per MTHM

A useful measure of the performance in the nuclear fuel cycle is the energy obtained per unit mass of fuel, known as the fuel’s burnup. The burnup is commonly specified in megawatt-days or gigawatt-days of thermal output per metric tonne of heavy metal (MWDT/MTHM or GWDT/MTHM). This is a cumbersome notation for repeated use, and we represent GWDT/MTHM in a more compact form as GWd/t (gigawatt-days per tonne). In standard energy units, 1 GWd = 8.64 ?1013 joules (J).

For U.S. reactors, as well as most reactors elsewhere, the “heavy metal” in the original fuel is uranium.20 The fuel is in the form of uranium oxide

(UO2). About 12% of the mass of the fuel is oxygen and, therefore, there is a distinction between the mass of heavy metal and the mass of the fuel.

The heavy metal in the spent fuel removed from the reactor is still primarily uranium, but it also includes isotopes of plutonium and—to a small extent— other transuranic elements. Typically, the mass of heavy metal is about 3%  fission of plutonium isotopes).21

A 1000-MWe reactor, operating at a typical thermal efficiency of 32%, produces energy at the rate of 3125 MWt (where it is explicitly indicated that this is the thermal output). One gigawatt-year of electric power therefore represents a thermal output of 1141 GWd(t). If, for example, the average

burnup in a reactor is 40 GWd/t, the fuel consumption is 28.5 tonnes of enriched uranium per gigawatt-year. Trends in Burnup of LWR Fuel Average burnupvalues for past years are shown in Table 9.1 along with the average projected for the fuel to be deposited at the Yucca Mountain waste repository. Overall, there has been a trend with time toward higher burnup, on average roughly doubling in the 25 years from 1973 to 1998 and projected to continue to rise. Thus, in a 1993 DOE projection, it was expected that the median PWR fuel burnup for standard assemblies would be about 43

GWd/t in the year 2000—a value actually achieved in 1998—and 51 GWd/t

20 The main exception is for reactors that use a mixture of uranium and plutonium

 (21 The designation “metric tons of heavy metal” (MTHM) commonly appears in

of the utilization and disposal of nuclear fuel. Here, MTHM refers to the heavy metal mass of the initial fuel. Alternatively, this can be made explicit

by using the designation “metric tonnes of initial heavy metal” (MTIHM). In

effect, “MTHM” and “MTIHM” are used interchangeably and mean the same

thing. Thus, the U.S. spent fuel inventory at the end of 1995 is given as 31926

MTHM in the 2002 Yucca Mountain EIS [24, Table A-7] and as 31952 MTIHM in

a 1996 report [25, Table 1.2]. (The 0.08% difference is insignificant compared to

the difference of several percent in the actual heavy metal contents of the initial

and spent fuel.)

                               Back End of Fuel Cycle

Handling of Spent Fuel

Initial Handling of Reactor Fuel

Periodically, a portion of the fuel in the reactor is removed and replaced by fresh fuel. In typical past practice, an average sample of fuel remained in the reactor for 3 years, and approximately one-third of the fuel was removed each year, with a shutdown time for refueling and maintenance of up to about 2 months. The trend is now to extend the interval between refueling operations and to reduce the time for refueling.25 Currently, time intervals of 18 months and shutdowns of 1 month are typical.

When the spent fuel is first removed from the reactor, the level of radioactivity is very high, due to the accumulation of radioactive fission products and radioactive nuclei formed by neutron capture. Each radioactive decay involves the release of energy, which immediately appears as heat, so the fuel is thermally hot as well as radioactively “hot.” Independent of the reprocessing question, the first stage is the same, namely allowing the fuel to cool both thermally and radioactively. The cooling of the fuel normally takes place in water-filled cooling pools at the reactor site.

Originally, it was planned to keep the spent fuel at the reactor for roughly 150 days and then to transfer it to handling facilities at other locations. The nature of the next step, in principle, depends on whether the fuel is to be disposed of as waste or reprocessed. However, as yet, this “next step” has been much delayed in the United States because no off-site facilities have been developed. Instead, almost all of the fuel has remained at the reactor sites—in many cases for more than 20 years.

In the absence of alternatives, some U.S. utilities are transferring older fuel rods from cooling pools to air-cooled (dry storage) casks at the reactor site. This may provide a workable temporary solution to the long delay in implementing a national waste disposal program. However, it is only a stopgap because the reactor operator cannot be counted on to be willing and able to supervise the spent fuel for prolonged periods of time (see Section 11.1.3).

Disposal or Storage of Spent Fuel

For many years, it had been assumed that all U.S. civilian nuclear waste would be reprocessed, but U.S. reprocessing plans have been abandoned. Instead,

25 An annual refueling shutdown of 2 months would mean a maximum capacity factor of 83%, which is well below the present U.S. average. official plans now call for disposing of the spent fuel directly, while retaining for many decades the option of retrieving it. The fuel is to remain in solid form and the fuel assemblies eventually placed in protective containers and ultimately moved in secure casks to either a permanent or an interim repository site.

In the latter case, the waste would be moved to a permanent repository at a later time.

A distinction is sometimes made been “disposal” and “storage.” The former suggest permanence, whereas the latter suggests the possibility that the spent fuel might be later retrieved. This possibility is made explicit in retrievablestorage systems, where the permanent sealing of the repository is deferred, allowing the spent fuel to be recovered should this be desired at a later time.26 In this case, the reprocessing option is not foreclosed, and the spent fuel may ultimately not be a “waste.”

There are several motivations for maintaining retrievability: (a) It allows for remedial action in case surprises are encountered in the first decades of waste storage that require modifying the fuel package or the repository; (b) it keeps open the option of recovering plutonium from the fuel; and (c) it allows the recovery of other materials deemed useful—for example, fission products for use in medical diagnosis and therapy or in the irradiation of food or sewage sludge. When the placement becomes irreversible, with no prospect of retrieving the fuel, this becomes final disposal.

Reprocessing

Extraction of Plutonium and Uranium

The alternative to disposing of the spent fuel is to reprocess it and extract at least the uranium and plutonium. In reprocessing, the spent fuel is dissolved in acid and the plutonium and uranium are chemically extracted into separate

streams, for use in new fuel. The most widely used method for this is the

suggestively named PUREX process.

Most early U.S. plans for reprocessing assumed that 99.5% of the U and

Pu would be removed. The remainder constitutes the high-level waste. In

the traditional plans, the wastes include almost all of the nonvolatile fission products, 0.5% of the uranium and plutonium, and almost all of the minor

actinides [i.e., neptunium (Z = 93), americium (Z = 95), and curium (Z =

96)]. The uranium represents most of the mass of the spent fuel, but the fission

products contain most of the radioactivity.

Extraction can be more complete than contemplated in the original U.S.

thinking. The French program has exceeded the 99.5% goal, separating out

more than 99.9% of the uranium and 99.8% of the plutonium [33, p. 28].

There is no essential reason to limit extraction to plutonium and uranium,

26 Plans for the Yucca Mountain repository call for it to remain open for perhaps as much as several hundred years, but the preservation of the reprocessing option does not now appear to be the major motivating factor although the former represents the valuable fuel and the latter represents

the bulk of the mass. It is possible to extract other radioisotopes as well,

either because they are deemed pernicious as components of the waste or

because they are useful in other applications. The minor actinides have been

of particular interest. They include long-lived products whose removal would

decrease the long-term activity in the waste. One option is to separate them and return them to a reactor where they would be transmuted in neutron reactions.

Of course, if the chief goal is safety, it is necessary to balance the benefits from decreased activity in the wastes against the increased hazards of handling and processing them when they are still very hot. At present, this further separation option has not been adopted in the major reprocessing programs in France and the United Kingdom, and the minor actinides remain with the  products [34, p. 149].

The residue of reprocessing constitutes the wastes. They are to be put in

solid form for eventual disposal. The standard method is to mix the high-level waste with molten borosilicate glass and contain the solidified glass in metal canisters. Although other solid waste forms have been suggested, borosilicate glass has been used in the French nuclear program and had figured prominently in the original U.S. plans for reprocessing commercial wastes. It is being used for the sequestering of already reprocessed U.S military wastes at the Savannah River site in South Carolina and is planned for the wastes at the Hanford reservation in Washington state.

Status of Reprocessing Programs

Until the late 1970s, reprocessing had been planned as part of the U.S. nuclear power program. A reprocessing facility at West Valley, New York was

in operation from 1966 to 1972, with a capacity of 300 MTHM/yr. This is enough, roughly speaking, for the output of 10 large reactors. There were plans for further facilities at Morris (Illinois) and Barnwell (South Carolina) which would have substantially increased the reprocessing capacity. However, all these plans have been abandoned.27

In part, the abandonment was impelled by technical difficulties. There had been high radiation exposures of workers at West Valley and the plant was shut down in 1972; plans to remodel and expand it were later aborted. When the Morris plant was first tested with nonradioactive materials, it did not perform reliably, and the General Electric Co., which was building the plant, decided there were serious difficulties. The Barnwell plant moved ahead until the early 1980s, but it faced problems of meeting increasingly strict standards on permissible radioactive releases.

In uranium supply, and uranium prices were low enough to remove the economic incentive for reprocessing. Further, an important body of opinion had developed in the United States against reprocessing, on the grounds that it might make plutonium too readily available for diversion into destructive devices.

This view was expressed in the 1977 report Nuclear Power Issues and Choices, sponsored by the Ford Foundation and authored by an influential group of national science policy leaders. In its conclusions on reprocessing, the report stated:

[T]he most severe risks from reprocessing and recycle are the increased opportunities for the proliferation of national weapons capabilities and the terrorist danger associated with plutonium in the fuel cycle.

In these circumstances, we believe that reprocessing should be deferred indefinitely by the United States and no effort should be made to subsidize the completion or operation of existing facilities. The United

States should work to reduce the cost and improve the availability of alternatives to reprocessing worldwide and seek to restrain separation and use of plutonium. [36, p. 333]

Consistent with this thinking, the Carter administration decided in 1977 to “defer indefinitely the commercial reprocessing and recycling of the plutonium produced in U.S. nuclear power programs” [37, p. 54]. Work on U.S. reprocessing plants for commercial fuel was phased out, culminating in the closing

of Barnwell at the end of 1983

Nonetheless, reprocessing has been pursued in other countries. The largest reprocessing programs are in France and the United Kingdom, both of which completed major expansions of reprocessing capacity in 1994 to handle both domestic and foreign fuel. In addition, a large facility is being built in Japan.

France has the most fully developed fuel cycle. Although much of its present reprocessing capacity is devoted to foreign orders, it also has a program of reprocessing and plutonium recycle of domestic fuel.

Year of Capacity

Country Location Start-up (MTHM/yr)a

In operation

France La Hague (UP2)a 1976 800

France La Hague (UP3) 1989 800

India Tarapur 1974 100

India Kalpakkam 1998 100

Japan Tokai-mura 1977 90

Russia Chelyabinsk 1984 400

United Kingdom Sellafield (B205) 1964 1500

United Kingdom Sellafield (Thorp) 1994 1200

Under construction

China Diwopu 2002 (?) 25–50

Japan Rokkasho-Mura 2005 800 aUP2 was upgraded and redesignated as UP2-800, with full capacity reached in 1994 sin the reactor core, with the remainder ordinary uranium-oxide fuel [41, 42].28

Some LWRs, however, have been designed to accommodate a full load of MOX

fuel.29

By 2001, about 20 PWRs in France (out of 58) were using MOX for onethird of their fuel [34, p. 138]. In the United States, the interest in MOX fuel has been motivated by the need to dispose of plutonium from dismantled nuclear weapons (see Section 18.3.3). Toward this end, the DOE is planning to build facilities for conversion of plutonium into MOX fuel at its Savannah River site. At least one nuclear plant operator (Duke Energy) has made a

28 It is more difficult to control a thermal reactor using plutonium than one using uranium. Contributing reasons include (a) the delayed neutron fraction, β, I smaller for 239Pu than for 235U and (b) the fission cross section resonance in

239Pu near 0.3 eV (see Figure 6.1) leads to a positive feedback if the reactor

temperature rises. In addition, with 239Pu, the neutron and gamma-ray spectra

are more energetic than with 235U, causing more radiation damage [42, p. 119].

As a result, it is necessary to have design changes, including more control rods, if a full load of MOX fuel is used in place of uranium fuel. This cannot be readily

accomplished in most LWRs. However, it is possible in the so-called System-80

PWRs. Three such reactors are in operation at the Palo Verde nuclear plant in

Arizona, but, at present, no U.S. LWR is licensed by the NRC to operate with MOX fuel.

29 See Section 18.3 for a further discussion of MOX fuel, in the context of the burning

of plutonium from dismantled nuclear weapons. commitment to use MOX fuel in some of its reactors, beginning in 2007 if plans proceed according to the initial schedule [43].

Advanced Aqueous Process

In the widely used PUREX process, the plutonium and uranium are extracted and the fission products and minor actinides constitute the wastes. The advanced aqueous process is a modification of the PUREX process in which the

 stage to reduce the bulk of the material that must be dealt with in the further chemical processing [44, p. 60]. The two product streams, one of uranium and

the other of plutonium and the minor actinides, are used to fabricate fuel for use in either thermal or fast reactors.

The UREX Process

An alternative to the advanced aqueous process is the uranium extraction

process (UREX and UREX+). It differs in the means of separating out the

uranium. Several output streams are specifically identified in this process [45,

p. II-3]:

1. Uranium. The uranium is extracted in very pure form (“at purity levels

of 99.999 percent”). The leaves it free of highly radioactive contaminants

and makes it easy to handle for disposal or reuse in a reactor.

2. Plutonium and minor actinides. Neptunium, americium, and curium are

retained with the plutonium. These elements can be incorporated into the

reactor fuel.

3. Long-lived fission products. Long-lived fission products (in particular,

iodine-129 and technicium-99) are separately extracted, for destruction

in a reactor (see Section 11.3.3).

4. Other fission products. These become the wastes. The waste disposal problem is simplified because the long-lived radionuclides have, for the most

part, been removed.

Pyroprocessing

The above-discussed chemical reprocessing processes are known as aqueous processes. An alternative approach, under active exploration for use in conjunction with future reactors, is the pyroprocess or electrorefining process. In

this method, the spent fuel is dissolved at very high temperatures in molten

cadmium, creating an “electrolytic bath.” Groups of chemical elements are

separately extracted on the basis of differences in the potentials at which they

dissolve and ionize. In particular, ions of the actinides, including uranium, plutonium, and the minor actinides, are attracted to cathodes and are extracted.

The actinides are then incorporated in the fabrication of new fuel

elements.

Full Actinide Recycle

A fuel cycle based on the nearly complete extraction of plutonium and minor

actinides (collectively, the transuranic elements) and their consumption by

fission in fast reactors has been sketched in the MIT report The Future of

Nuclear Power [46]. This cycle envisages a global nuclear economy in 2050

with a capacity of 1500 GWe based on a balanced combination of thermal

and fast reactors. The thermal reactors are assumed to be LWRs, fueled by

enriched uranium oxide. The fast reactors, undefined as to type, are fueled

by transuranics obtained from the LWR spent fuel. Pyroprocessing is used to

extract the transuranics from both the thermal and fast reactor spent fuel.

For each load of fresh fuel in the fast reactors, 20% of the transuranics are

consumed in fission and the remaining 80% are available for recycle.

A balanced system is one in which the spent fuel from the thermal reactors

provides the transuranics needed to make up for those consumed in the fast

reactors. With the assumptions made in the MIT analysis, this is achieved

by having slightly more capacity in the thermal reactors than in the fast

reactors (815 GWe and 685 GWe, respectively). A variety of choices exist for

the fast reactors, including some of the Generation IV reactors discussed in

Section 16.6.

The uranium requirement for the entire fuel cycle is the amount needed

to provide for the 815 GWe of LWRs, which is 54% of the amount needed

if LWRs accounted for the full 1500-GWe capacity. A further, and perhaps

even more important, benefit is the almost complete elimination of plutonium

and minor actinides from the stream of wastes that require permanent

disposal.

General Features of Reprocessing Options

Any fuel cycle that recycles the fissile components of the spent fuel (mainly the

remaining 235U and the plutonium isotopes 239Pu and 241Pu), increases the

energy obtained from the existing uranium resources. If the minor actinides

are included with the uranium and plutonium in the new fuel, the wastes will

have much less long-term radioactivity than wastes in the once-through fuel

cycle. The mass of the spent fuel is greatly reduced if the uranium is either

returned to the reactor or is separated from other radionuclides to become

low-activity depleted uranium. The fission products then constitute the waste

product that requires long-term disposal. This greatly reduces the mass of the

waste product and the period during which it must be kept isolated from the

environment. All of the reprocessing fuel cycles that have been described in this section achieve these resource extension and waste reduction benefits. The PUREX

process accomplishes much of this, but in its standard form the minor actinides are not removed.

A long-standing objection to reprocessing is based on the increased proliferation risks if 239Pu is in wide circulation. The above-described methods lessen the risks in two ways: (1) The presence of the minor actinides increases

the activity of the fuel and makes it more difficult to handle, and (2) the reprocessing and fuel fabrication facilities can be located adjacent to the reactor,

making theft or diversion of the fuel very difficult without the collaboration

of the plant operators. The collocation aspect was particularly stressed in

planning documents for the Integral Fast Reactor (see Section 16.5.1) and

combined facilities were tested on a small scale using fuel from the experimental

breeder reactor in Idaho (EBR-II).

A further objection is based on economics. Given present demand and

prices, it is more expensive to reprocess spent fuel than to obtain fuel from

newly mined uranium. This is a cogent objection at the present scale of nuclear

power use. The advantages of these reprocessing approaches become more

relevant in the context of a possible major expansion of nuclear power.

At present, the reprocessing approaches discussed here are in the development and study stage, except for the long-used PUREX process. In general,

the pyroprocessing technique is more suitable for use with fuel in metallic form,

while the aqueous processes are more suitable for oxide fuels. Thus spyroprocessing was originally studied for use with metallic fuel from a sodium-cooled fast reactor. However, either class of process could be used with a wide variety of fuel forms, given appropriate pretreatment stages.

 Waste Disposal

All countries with announced plans for disposing of high-level radioactive

wastes are planning on eventual disposal in deep geologic repositories, typically

made by excavating caverns or holes in favorable environments. Many of

the plans for these permanent disposal facilities include a period during which

the waste could still be retrieved.

Deep geologic disposal has been the favored course in U.S. thinking since

the first attempts to formulate plans. There have been continuing efforts to

locate and design a suitable facility. A site at Yucca Mountain in Nevada was

selected in 1987 as the candidate for a U.S. repository and it has been under

intense study since. The DOE in 2002, with the subsequent concurrence

of the president and Congress, recommended going ahead with the Yucca

Mountain project. The announced goal is to have a facility ready to receive

wastes by 2010, subject to approval by the Nuclear Regulatory Commission.

 Price of Uranium

Conventional Units for Amounts of Uranium

The magnitude of uranium resources can be specified in terms of the amount

of uranium oxide (U3O8) or the amount of natural uranium (U). Commonly,

U.S. organizations expressed the resources in short tons of U3O8, whereas

international organizations, such as the OECD, use tonnes of uranium. The

units are related by the equivalence:30

1 ton of U3O8 = 0.769 tonnes of U.

In terms of the units in which uranium prices are usually couched, 1 kg of U is equivalent to 2.60 lbs of U3O8; therefore, a price of $100/lb of U3O8 is equivalent to $260/kg of U.

Uranium Prices and Electricity Costs

Uranium prices are now very low compared to those projected several decades

ago. They dropped markedly in recent years, in large measure due to the lag

in the expansion of nuclear power. U.S. prices peaked in 1978 at an average

of $43/lb of U3O8 [47]. In 2001, the average price paid by U.S. utilities was

$10/lb U3O8 ($26/kg U) [17, p. 11].

The relationship between uranium price and the contribution of uranium

fuel costs to electricity costs depends on the effectiveness of fuel utilization. As

was discussed in Section 9.3.4, the approximate requirement for LWRs is 200

tonnes of uranium per gigawatt-year. Then, for example, uranium at $100/kg

corresponds to a cost of $20 million per gigawatt-year (8.76 ?109 kWh), or

0.23/c/kWh.

A useful overall equivalence for LWR uranium costs is

$100/lb of U3O8 = $260/kg of U ≈ 0.59/c/kWh.

The 2001 price ($26/kg of U) corresponds to a contribution to the cost of

electricity of about 0.06/c/kWh. This is roughly 1% of the total cost of electricity

from nuclear power plants. Thus, uranium would remain “affordable”

even with a large increase in uranium prices. For example, uranium would

cost 1/c/kWh at a price of $440/kg of U.

References

1. Organization for Economic Co-operation and Development, Nuclear Energy

Agency, Trends in the Nuclear Fuel Cycle: Economic, Environmental and Social

Aspects (Paris: OECD, 2001).

2. U.S. Department of Energy, Nuclear Power Generation and Fuel Cycle Report

1997, Energy Information Administration Report DOE/EIA-0436(97) (Washington,

DC: U.S. DOE, 1997).

3. Uranium Information Center, Geology of Uranium Deposits, Nuclear Issues

Briefing Report No. 34 (Melbourne, Australia: UIC, 2001) [From:

http://www.uic.com.au/nip34.htm]

4. Organization for Economic Co-operation and Development, Nuclear Energy

Agency, Uranium 2001: Resources, Production and Demand, Joint Report of

OECD Nuclear Energy Agency and IAEA (Paris: OECD, 2002).

5. Manson Benedict, Thomas H. Pigford, and Hans Wolfgang Levi, Nuclear Chemical

Engineering, 2nd edition (New York: McGraw-Hill, 1981).

6. U.S. Department of Energy, Statistical Data of the Uranium Industry, Report

GJO-100(83) (Grand Junction, CO: U.S. DOE, 1983).

7. DeVerle P. Harris, “World Uranium Resources,” Annual Review of Energy 4,

1979: 403–432.

8. Luther Smith, U.S. Department of Energy, private communication (August

2002).

9. Ahmad E. Nevissi and David Bodansky, “Radon Sources and Levels in the Outside

Environment,” in Indoor Radon and Its Hazards, David Bodansky, Maurice

A. Robkin, and David R. Stadler, eds. (Seattle: University of Washington Press,

1987): 42–50.

10. James J. Duderstadt and Louis J. Hamilton, Nuclear Reactor Analysis (New

York: Wiley, 1976.)

11. U.S. Department of Energy, Office of Civilian Radioactive Waste Management,

Yucca Mountain Science and Engineering Report, Report DOE/RW-0539

(North Las Vegas, NV: U.S. DOE, 2001).

12. Ronald Allen Knief, Nuclear Engineering: Theory and Technology of Commercial

Nuclear Power, 2nd edition (Washington, DC: Hemisphere Publishing Company,

1992).

13. Allan S. Krass, Peter Boksma, Boelie Elzen, and Wim A. Smit, Uranium Enrichment

and Nuclear Weapon Proliferation (London: Taylor & Francis, 1983).

Cycle, W. Marshall, ed. (Oxford: Clarendon Press, 1983): 104–158.

15. R.E. Leuze, “An Overview of the Light Water Reactor Fuel Cycle in the U.S,”

in Light Water Reactor Nuclear Fuel Cycle, R.G. Wyneer and B.L. Vondra, eds.

(Boca Raton, FL: CRC Press, 1981).

16. Uranium Information Center Ltd, Uranium Enrichment, Nuclear Issues Briefing