Background

7.2.1 How did we get here? A brief history of nuclear energy

Several descriptions of the history of nuclear energy clearly overlap the history of nuclear weapons (Goldschmidt, 1982). In Appendix A we show a chronology of important events related to the development of nuclear fission, starting with nuclear weaponry and proceeding to civilian nuclear energy. In addition, general books about nuclear energy discuss the history and the basic issues very clearly (Bodansky, 1997).

7.2.2 Basic physics of fission

An atomic nucleus of atomic mass A and charge Z (atomic number) can be thought of as composed of Z protons and (A —Z) neutrons. These constituents are bound into the nucleus by the attractive nuclear forces. In accordance with Einstein's famous equation (E=mc2) this leads to a mass defect (or binding energy/c2), AM, according to the equation:

M(A,Z) = Z Mp + (A —Z) Mn — AM, where Mn and Mp are the masses of the neutron and proton, respectively.

The mass defect per nucleon, AM/A, is plotted in Fig. 7.2. The initial increase in AM/A as A increases is a result of nuclear forces, and the decreases of the per-nucleon mass defect as A becomes larger is a result of the repulsion of the Coulomb forces between the protons. From this curve it may be seen that if two light nuclei combine to form a heavier one (i.e., nuclear fusion), energy is released, but above the common element iron (A = 56) energy is released if a heavy nucleus is split into two lighter fractions (e.g., nuclear fission). It is also important that light nuclei usually have equal numbers of protons and neutrons, whereas heavy nuclei have a considerable neutron excess (uranium 238 has 92 protons and 146 neutrons). In this chapter we are concerned with fission. Heavy elements can be stable because the attractive nuclear forces are strong at small distances (10—13cm), whereas the repulsive Coulomb forces act over much greater distances.

These basic features of nuclear masses were known at least as far back as the time of the early mass measurements (Aston, 1919), but it remained a puzzle how to unlock the energy available in the atom. The secret was discovered by Hahn and Strassman (1939), deliberately brought out of Nazi Germany by Lise Meitner, and theoretically described by Bohr and Wheeler (1939). When uranium is bombarded by slow neutrons (e.g., neutrons with energies close to c o

o hJ

92Kr

£Ba

—> • -

4He* /

Release Fission

235U

v f

through

LiT

/

.Energy through

Release Fusion

J3 H

|2h

20 40 60 80 100 120 140 160 180 200 220 240

Number of Nucléons per Nucleus, A

Figure 7.2 Binding energy (mass deficit) versus atomic mass showing fusion (2H + 3H —>4He + n + 17.6 MeV) and fission (235U + n—>141Ba + 92Kr + n + 205 MeV) approaches to increased nuclear stability and attendant energy releases.

ambient thermal energy), this "thermal" neutron is absorbed into the nucleus and a compound nucleus is formed (in an excited state). If the compound nucleus has an atomic number divisible by four (an "even-even" nucleus), it will immediately split into two fragments of unequal mass. In the process, as first shown by Joliot, von Halban and Kowarski (1939), additional prompt free neutrons are produced (on average about 2.5 neutrons per fission). This release of extra (energetic) neutrons leads to the possibility of a chain reaction where at least one of the produced neutrons goes on to produce another fission, which in turn produces more neutrons. Approximately 81% of the energy produced -three million times as much per unit mass as in the burning of carbon - is in the kinetic energy of the fission fragments1.

Of the naturally occurring elements, uranium of atomic mass 235 is unique in satisfying the requirement of Bohr and Wheeler (e.g., 236 is divisible by 4). Uranium 235 is only present to an extent of 0.71% in normal ores (99.3% is uranium 238 which is not fissionable by slow neutrons), so not all the nuclear energy in the uranium can be released. But other elements can be produced by capture of a neutron by the nucleus. In particular, when uranium 23 8 (238U) captures a neutron it becomes uranium 239; successive radioactive (beta) decay leads to neptunium 239 (239Np) and plutonium 239 (239Pu, discovered by Seaborg and collaborators in 1940). A further neutron capture (by plutonium 239) leads to the excited nucleus plutonium 240 which, having an atomic number divisible by four, immediately splits. This sequence describes the process of breeding a "fissile" (meaning fissionable by slow neutrons) element plutonium 239 from the "fertile" element uranium 238, which is 141 times more plentiful than uranium 235. Further neutron capture leads to heavier nuclei which are unstable in nature and more easily fissioned, as is shown in Fig. 7.3a. In addition to uranium 235 being fissile, uranium 233 (also discovered by Seaborg) is fissile. While not present in nature, it can be bred from neutron capture on thorium 232 (232Th), which is four times more plentiful than 238U, as is shown in Fig. 7.3b.2

The important fissile elements for nuclear energy are 233U, 235U, 239Pu and to a lesser extent the heavier transuranic elements (most plutonium isotopes, americium isotopes, curium isotopes, etc.). Fission can occur with either slow or fast neutrons. The probability of capture of a neutron by a nucleus, however,

1 Typically, of the 210 MeV (3.2 X10"11 J/fission, 19.3 TJ/mole, or 82.1 TJ/kg 235U) released per fission, 81.0% appears as fission-product kinetic energy, 6.8% as fission-product decay, 3.9% as prompt gamma radiation, and the remaining 8.3% as neutrinos.

2 232U is formed from the beta decay of 232Pa, which in turn derives from (n, 2n) reactions on the fertile material 232Th followed by a beta decay of the resulting 231Th; the 232U contributes significant alpha activity (heating), and from the 1.91-yr 228Th daughter product, which leads to a number of short-lived daughters, some (212Bi and 208Tl) emitting very energetic gamma radiations.

Figure 7.3a Principal nuclear reactions and transitions in uranium, showing reaction cross-sections (1 barn = 1028m2) and transition half-lives.

Figure 7.3b Principal nuclear reactions and transitions in thorium, showing reaction cross-sections (1 barn = 1028m2) and transition half-lives.

is proportional to the time the neutron is within the range of the short-range nuclear forces (e.g., proportional to the inverse of the neutron speed, 1/v) and, therefore, is greater for slow neutrons. It is, therefore, easier to make a chain reaction with slow neutrons, although it is well known that a chain reaction is possible with fast neutrons (as in a nuclear bomb). Nuclear reactors based on fissions from these slow or thermal neutrons require a "moderator" to slow down the energetic neutrons released from fission through collisions with the nuclei of the moderator. A desirable moderator will have light nuclei for slowing down (more energy lost by the neutron on average per collision), as well as a low neutron capture cross section (e.g., reduced "parasitic" loss of the valuable neutron to the moderator nuclei). This choice of one of the three fissile elements (233U, 235U, 239Pu) combined with the choice of moderator gives a matrix of possible nuclear-reactor configurations discussed in Section 7.2.3 and elaborated in Table 7.3.

7.2.3 Taxonomy of nuclear reactors

The balance of the neutron production (through fission) and consumption (e.g., all absorptions plus leakage from the reactor core) is crucial to the sus-tainment of a controlled chain reaction for times sufficient to allow the safe extraction of an economic amount of energy (measured usually in units of GWtd) from a given mass of fuel (heavy metal, HM). For the U-Pu fission cycle depicted in Fig. 7.3a, 235U is consumed, plutonium isotopes increase, and fission products build up. Figure 7.4 illustrates these changing nuclide concentrations in a large pressurized-water-reactor (PWR, a kind of LWR; Benedict, Pigford, and Levi, 1981), with time expressed in terms of integrated thermal energy per unit of initial fuel, GWd/tonneHM. As time progresses, fissions in plutonium isotopes increase, and neutron absorptions in the increasing concentrations of fission products increase. Typically (Marshall, 1983), each fission produces 2.6 neutrons, 63% of which come from 235U and 32% are created from the fissioning of 239Pu that is building into the fuel (Fig. 7.4). Roughly 39% of these 2.6 neutrons go on to produce fissions in the fuel, and another 39% are absorbed in the fuel without causing fission; of these 58% are absorbed in 238U, 15% are captured in 239Pu, and the remaining 15% are captured in 235U. The remaining 22% of "lost" neutrons are absorbed in structure, coolant, fuel cladding, and control rods (76% of the non-fuel absorptions, or 17% of the original 2.6 fission neutrons), with fission-product absorptions accounting for 24% of the remaining "lost" neutrons [5% of the original (235U + 239Pu) fission neutrons]. Typically, leakage out of the reactor core per se is much less than a percent of the fission neutrons produced.

Achieving the correct mixture of fuel and moderator to assure an exact balance between neutron production and consumption over reasonable times of steady-state fission power production would be very difficult. Control of this desired steady state would be impossible for either thermal- or fast-spectrum reactors were it not for the delay in a small fraction of the fission-neutron emissions from certain fission products having to undergo beta decay before shedding an excess neutron. This delay ranges from a fraction of a second to nearly a minute, with an average delay being in the range 10-20 seconds. While the percentage of all fission neutrons that are in this delayed-neutron class vary with fissile isotope (0.28% for 233U, 0.64% for 235U, 0.21% for 239Pu), the time scale for control of the neutron population in a critical reactor is increased from microseconds to milliseconds, which is in the regime of most mechanical control systems (e.g., neutron-absorbing control rods moved into and out of the reactor core).

On the basis of the foregoing description, a nuclear fission reactor requires four primary systems to provide safe and economic thermal-power generation while simultaneously satisfying the constraints imposed by the above-described neutron balance:

• Moderator (low atomic mass) material that efficiently (few collisions required) slows down energetic fission neutrons;

• Structure within/surrounding the core (including fuel cladding); and

• Heat transfer/transport medium (coolant).

Additionally, while not directly impacting the neutron balance and the physics of the core, the system that converts the thermal energy released (mainly as fission-product kinetic energy) from particle kinetic energy to electrical energy on the grid is an essential element of the overall reactor system. Integral to the nuclear power plant, but not considered here, are all operational and offnormal safety and control systems. Within the neutronic constraints and limitations described above can be fitted a wide range of material options for the fuel, moderator (if needed), structure, and coolant. Material systems that have been considered and implemented in the past are listed in Table 7.3. Included in this table are typical materials with high (thermal) neutron absorption cross section used to control the neutron population within the reactor core by insertion into the coolant, into the fuel, or into separately manipulated control rods. Hence, a rich array of material combinations is available to define the economics, safety, and sustainability of nuclear energy (Todreas, 1993).

By far the most common reactor system in use today around the world is based on a fuel composed of fissile uranium (or plutonium) oxide, with both the

Table 7.3 Materials options matrix for nuclear fission energy

Fuel"

Fuel forms''

Control rods Moderator®

Structure^

Coolant'1

Conversion'

Fissile

Fertile"

Metal alloy

Oxides

Carbides

Nitrides

Cermets

Dispersions

Liquids

• Aqueous slurry

Gaseous (UF6)

Boron (10B) Cadmium Europium Gadolinium

Beryllium Light water Heavy water Graphite none^

Steels

• Austenitic

Silicon carbide Concrete

Fuel cladding

• Zirconium alloys

• Stainless steels

• Graphite/carbides

Helium

Carbon dioxide Light water Heavy water Sodium Lead eutectic Other liquid metals

• Lead Molten salts

• KCl/LiCl Organic liquids

Steam turbine^ Gas turbine' MHD™ NPL" Chemical0

Notes:

a A small fraction of the fertile isotopes 238U and 232Th undergo fission upon absorption of an energetic neutron. b Plus 241Pu in a thermal neutron spectrum; all plutonium isotopes fission in an energetic neutron spectrum; 239Pu formed upon neutron absorption in 238U according to Fig. 7.3. c Formed upon neutron absorption in 232Th according to Fig. 7.3. d Solid fuel metrices clad in zirconium-based alloy or stainless steel. e For decreasing neutron energy from few MeV at birth to near "room temperature".

For fast or epithermal spectrum reactors, or for thermal systems with 235U enrichments about 20%. Cladded fuel-pin grids, in-core support structure, pressure vessel, coolant piping.

Coolant and moderator functions can be combined in some cases (e.g., light water); other coolants that have been considered include heavy water (D20), liquid metals (Na, NaK, Pb), and pressurized gases (C02, He).

Refers primarily to thermal-to-electric conversion; non-electric applications envisage the delivery of high-temperature (800-900 °C) process heat by means of a liquid-metal (sodium, sodium-potassium eutectic, lead or lead-bismuth eutectic); the use of chemical storage and transfer of nuclear energy has also been suggested.

Largely the Rankine thermodynamic cycle; steam generation either directly within the reactor core or through a secondary coolant represent further sub-options.

Largely the Brayton thermodynamic cycle; as with the steam-based conversion, either direct drive or the use of a secondary working fluid define sub-options. Magnetohydrodynamic conversion.

Nuclear pumped laser based on direct fission-product energy conversion.

For example, (endothermic) methane reforming at reactor site; transport of the C0/H2/C02 synthesis gas mixture to a utility/user site; followed by (exoergic) re-menthanation and use of released chemical energy, and either reinjection or recirculation of the substitute natural gas.

Neutron Count Rate Relative With Burnup

FUEL BURNUP, GWtd/tonne HM

Figure 7.4 Change in fractional nuclide concentrations with burn-up for a 1060-MWe PWR, showing the depletion of the 235U driver fuel and the buildup of both fission products and plutonium isotopes (Benedict et al., 1981).

FUEL BURNUP, GWtd/tonne HM

Figure 7.4 Change in fractional nuclide concentrations with burn-up for a 1060-MWe PWR, showing the depletion of the 235U driver fuel and the buildup of both fission products and plutonium isotopes (Benedict et al., 1981).

neutron-moderating and the cooling functions provided by ordinary (light) water, and with electricity generated by high-pressure steam driving a turbinegenerator system. Approximately 76% of all commercial nuclear power plants (NPPs) are of this light-water reactor (LWR) kind [55% pressurized-water reactors (PWRs) and 21% boiling-water reactors (BWRs)]. Heavy-water-moder-ated/ light-water-cooled reactors operated on natural (unenriched in the 235U isotope) uranium, and gas-cooled (helium)/graphite-moderated reactors represent important, but minor, contributors to the present mix of world nuclear power plants. Finally, a number of small reactors operated without strongly neutron-moderating (neutron thermalizing) materials and, therefore, sustained by fast or energetic neutrons, using liquid-metal (sodium) coolants, are being operated to investigate the physics and technology of such systems that might ultimately be needed to exploit the much larger 238U fertile-fuel resource via the breeding of plutonium fissile fuel (Fig. 7.3a). Recently (Galperin et al., 1997; Murogov et al., 1995), interest is developing in exploiting the 232Th-233U breeding fuel cycle (Fig. 7.3b) in thermal-neutron LWRs; this interest is driven by desires to address both proliferation, resource, and waste concerns by building on the well-developed thermal-spectrum LWR technology. It has been claimed (West, 2000) that a fast-neutron reactor can also be used to produce 233U uranium fuel that is as proliferation resistant as any LWR fuel.

As is seen from the chronology given in Appendix 7.A, the evolution of the somewhat hegemonic spectrum of commercial NPPs described above resulted largely from the needs and economics associated with a dual-use approach to military and peaceful applications of nuclear energy. Furthermore, the development and deployment of both nuclear power plants and the all-important "back-end" [e.g., used-fuel storage, reprocessing (if at all), and waste disposal] has been and continues to be subject to concerns and constraints related to the proliferation of nuclear-weapon materials to presently non-nuclear-weapon states. These concerns and constraints prevail even though nuclear-explosive materials have yet to be obtained for that dark purpose from the civilian nuclear fuel cycle (Meyer, 1984).

Appendix 7.B (Nuclear News, 1992) summarizes in table form contemporary reactor design and performance parameters for: pressurized-water reactors (PWR), boiling-water reactors (BWR), modular high-temperature gas-cooled reactors (MHTGR), Canadian heavy-water-moderated (D2O) natural-uranium reactor (CANDU), and the advanced liquid-metal-cooled fast-spectrum reactor (ALMR). Prototypical diagrams for each of these four NPPs are given in Figs. 7.5-7.9, with narrative descriptions (Benedict et al., 1981; Marshall, 1983; Cochran and Tsoufanidis, 1990) being given in the respective captions and associated endnotes. Similarly, Figs. 7.10-7.12 give "top-level" material flows fuel cycles for LWRs, MHTGRs, and LMFBRs,

High Pressure Turbine

Low Pressure Turbine

Generator

Transmission Lines

Generator

Condensate Pump

NOTE: 3 or 4 Steam Generator Loops in a 1000 MVIfePlant used

Figure 7.5 Schematic diagram of a pressurized-water reactor (PWR).3

3 The PWR core is fueled with low-enriched (LEU, 3.5%253U) uranium-oxide (U02) pellets that are contained (clad) in zirconium-alloy tubes, through which the majority of the fission energy is conducted to the flowing pressurized-water coolant/moderator. The water is thereby heated as it passes through the reactor core and flows out of the core to a steam generator, where heat is exchanged to create steam in this secondary coolant loop; the steam is used to drive a turbine-generator unit to make electricity. The primary-coolant system (core, pressure vessel, primary-coolant pump, steam generators, etc.) are housed in a primary containment building; the fuel, cladding, pressure vessel, and containment building present sequential barriers to loss of radioactive materials to the environment.

To DW-Sump

Steam Dryers & Separators

Generator Transmission Lines

Steam Dryers & Separators

Generator Transmission Lines

Feed Water Pump

Feed Water Heater

Figure 7.6 Schematic diagram of a boiling-water reactor (BWR).4

Condensate Pump

Recirculation Pump mmm primary cycle (Radioactive Steam) primary cycle (Reactor Grade Water)

Feed Water Pump

Feed Water Heater sssa COOLING CYCLE (Non-Demineralized Water)

Figure 7.6 Schematic diagram of a boiling-water reactor (BWR).4

Condensate Pump

4 Like the PWR, the BWR core is fueled with low-enriched (LEU, 3.5% 235U) uranium oxide (U02) pellets that are contained (clad) in zirconium-alloy tubes, through which the majority of the fission energy is conducted to the flowing pressurized-water coolant/ moderator. The water is thereby heated as it passes through the reactor core. Unlike the PWR, water boils as it passes upwards through the core and turns into steam within the primary pressure vessel. This steam is passed directly to the turbine-generator, condensed, and sent back to the reactor core by means of feedwater pumps. The primary-coolant system (core, pressure vessel) is housed in a primary containment building; the fuel, cladding, pressure vessel, and containment building present sequential barriers to loss of radioactive materials to the environment.

rw actor vvsmI cooled by natural circulation ot air cual encapsulatsd by high-maltinQ-PuBiw-v-K point ceramic

Reactivity decrease* a« tempera-lure incraatOT

rw actor vvsmI cooled by natural circulation ot air cual encapsulatsd by high-maltinQ-PuBiw-v-K point ceramic

Reactivity decrease* a« tempera-lure incraatOT

Steam to turbine*

Buried in underground vault

Figure 7.7 Schematic diagram of a modular high-temperature gas-cooled reactor (MHTGR).5

Steam to turbine*

Buried in underground vault

Figure 7.7 Schematic diagram of a modular high-temperature gas-cooled reactor (MHTGR).5

5 Although the HTGR has not been widely exploited commercially, its high level of inherent safety to coolant loss [large thermal mass, high-temperature (refractory) materials, and high-temperature helium, capable of direct cycle to a gas turbine], its high fuel efficiency when operated on a thorium fuel cycle, and its ability to produce process heat for non-electric application all combine to make the MHTGR an important concept in the long-term. Some of these benefits accrue at low power density and, hence, higher capital cost, which must be traded carefully with the very high fuel burn-up capability (>100 GWtd/tonneHM) and higher thermal-to-electric conversion efficiencies offered by this concept.

Harley Carburetor Cutaway View

Figure 7.8 Schematic diagram of the Canadian deuterium natural uranium reactor

Figure 7.8 Schematic diagram of the Canadian deuterium natural uranium reactor

6 The CANDU uses natural (unenriched) uranium in oxide form (UO2) as a fuel, and heavy water (D2O) as a coolant. This combination of moderator (deuterium is a more efficient moderator than hydrogen when neutron-absorption properties are taken into account) and coolant improves the overall neutron balance to the extent that unenriched uranium can be used. The pressurized-water coolant and associated steam generators are essentially the same as that used in a PWR. The relatively low power density in the CANDU reactor core allows for refueling at full power ("on-line" refueling), thereby giving very high plant availability factors (>90%), and compensating for the (capital) cost penalties of operating a lower power-density system.

Liquid Metal Fast Breeder Reactor
Figure 7.9 Schematic diagram of the liquid-metal fast breeder reactor (LMFBR)7

7 Both a loop and a pool design for the LMFBR are shown. A moderator per se is not required for this fastneutron sodium-cooled reactor; metallic, oxide or carbide fuels have been considered. The reactor is comprised of three main regions containing either stainless-steel-clad fissile (initially 235U, later 239Pu) or fertile (238U) fuels - the core, the axial blanket, and the radial blanket. The uranium matrix in all three regions is comprised of depleted uranium from enrichment-plant tailings (i.e., 235U content is below the 0.71% value found in natural uranium). The liquid-sodium coolant heats (and is activated by neutrons) as it passes through the core, and transfers this heat to a secondary coolant, which is not radioactive. It is then directed to a steam generator, where the steam so created is used to drive a turbine-generator unit.

Phase Diagram For Uf6

Figure 7.10 Schematic fuel-cycle diagram for LWRs with and without plutonium recycle. Parts of the fuel cycle where significant quantities of weapons usable nuclear materials may reside are indicated by the heavy lines (Willrich and Taylor, 1974), whereas dashed lines indicate process flows that are generally not yet implemented in the US.

Figure 7.10 Schematic fuel-cycle diagram for LWRs with and without plutonium recycle. Parts of the fuel cycle where significant quantities of weapons usable nuclear materials may reside are indicated by the heavy lines (Willrich and Taylor, 1974), whereas dashed lines indicate process flows that are generally not yet implemented in the US.

along with indications of where in the fuel cycle elevated proliferation concerns may arise (Willrich and Taylor, 1974). For each of these three figures, operations conducted at distinct locations are designated by boxes, with the solid lines/arrows indicating inter-facility material flows; for situations where plutonium recycle is not occurring, dashed lines/arrows are used. Both the LWR and the MHTGR fuel cycles start with mining/milling/conversion (of oxide to

uranium mining

—U(N) ore—»

uranium milling

—U(N)3

thorium

—Th ore-*

thorium

mining

milling

U(N)F6 production

enrichment

U(mixed)F6

U(mixed)F6

irradiated fuel assemblies [U(H)C2 +U233C2] + fission products fuel reprocessing

U(H)C2 +ThC2 + graphite fuel assemblies fuel assemblies w/U233C2 added irradiated fuel assemblies [U(H)C2 +U233C2] + fission products

U(H)C2 +ThC2 + graphite fuel assemblies fuel assemblies w/U233C2 added

ThO,

Depleted Uranium

ThO, fuel reprocessing

■U233C, or U2330, small quantities of residual U235 and U236

U233C, or U2330, y233

storage waste storage

U(N) = natural uranium U(H) = highly enriched uranium (U235/U 90-95%)

Figure 7.11 Schematic fuel-cycle diagram for MHTGR Th-U fuel cycle with and without 233U recycle. Parts of the fuel cycle where significant quantities of weapons usable nuclear materials may reside are indicated by the heavy lines (Willrich and Taylor, 1974), whereas dashed lines indicate process flows that are not yet implemented.

Lmfbr Flow Diagram

Figure 7.12 Schematic fuel-cycle diagram for LMFBR U-Pu fuel cycle based solely on the use of depleted uranium. Parts of the fuel cycle where significant quantities of weapons usable nuclear materials reside are indicated by the heavy lines (Willrich and Taylor, 1974). Generally, the pyro-chemical/electro-chemical (non-aqueous) processing of LMFBR irradiated fuel and blanket materials would occur integrally with each

LMFBR power-plant operation.

Figure 7.12 Schematic fuel-cycle diagram for LMFBR U-Pu fuel cycle based solely on the use of depleted uranium. Parts of the fuel cycle where significant quantities of weapons usable nuclear materials reside are indicated by the heavy lines (Willrich and Taylor, 1974). Generally, the pyro-chemical/electro-chemical (non-aqueous) processing of LMFBR irradiated fuel and blanket materials would occur integrally with each

LMFBR power-plant operation.

gaseous UF6)/enrichment operations, whereas the LMFBR fuel cycle depicted in Fig. 7.12 is assumed to be sustained on depleted uranium from the enrichment operations. A fuel cycle based on the use of thorium in LWRs would look similar to that for the MHTGR (Fig. 7.13), except that the 235U enrichment required during the early phases of this Th-U/LWR fuel cycle would be below 20%.

7.2.4 Decay heat and fission-product radioactivity

The splitting of the nucleus leads to fission products that, according to both experiment and the Bohr-Wheeler theory (Bohr and Wheeler, 1939), are unequal in mass. These fission-product nuclei have neutron excesses (as the parent nuclei have) and are, therefore, unstable against beta decay. In general they will decay by a succession of beta-ray (i.e., energetic electrons) emissions until a more stable nucleus results. These decays result in two problems that are unique to nuclear fission energy. The radioactive decay produces heat (8% of the kinetic energy immediately after cessation of the nuclear chain reaction,

Fission Products Radioactivity
Figure 7.13 Fission-product radioactivity (solid lines) and nuclear decay heat per unit fission power (dashed lines) as a function of time after shutdown of the fission process for a range of full-power operating times.

~ 1% after one day, and ~0.2% after one month) and also various other radioactive emissions. The decay heat can produce an accident out of a malfunction of the coolant system in a non-passively cooled design, and the radioactivity, if released, can cause cancer and kill people. These two unique features cause considerable public concern. Figure 7.13 shows the nuclear decay heat as a percentage of full power as a function of time after the fission process has stopped, as well as the radioactivity as a function of time after cessation of the chain reaction. (Cohen, 1977; Benedict et al., 1981; Bodansky, 1997). Both nuclear after-heat and radioactivity rates after cessation of fission fall roughly as 1/t12, as noted by Way and Wigner (1948), rather than exponentially because it is a sum of exponential decays.8

8 Benedict, Pigford and Levi (1981, pp. 55-9) parametrize afterheat decays more precisely as eV/s = 3.0/t19 + 11.7/t14, with this energy carried equally by gamma and beta rays at 25% each, and the rest carried off by neutrinos; also Ci/W(fission) = 1.9/102 + 1/(T+t)02, where T is the time at full power, and t is the time of cooling after reactor shutdown; decay heat as a fraction of full fission power is 0.0042(1/t02 + 1/(T+ t)02) + 0.0063(1/t04 + 1/(T +t)04).

3 106

3 106

135Ç

;sm n

rV29i

ä1 + 93

107Pd

99T

93Zr

126Snt

h

79Sen

137Cs

90Sr

!151Sm

hrfiDL

à

4 +

1

U-oyole Th-cycle

C3 i L|HH r

S □

h

+

fflS

ftn

EB

+

10-2

Half Life of Nuclide (year)

Figure 7.14 Distribution of specific fission yields and associated fission-product halflife at equilibrium for both the uranium and thorium fuel cycles (Takagi and Sekimoto,

1992).9

Figure 7.14 depicts key fission-product radionuclides in terms of masses in equilibrium per unit of nuclear power as a function of fission-product half-life. Only a handful of isotopes have a lifetime that is sufficiently long to present a long-term waste-disposal concern. It is important to realize that these calculations have no ambiguity. The only legitimate ambiguity is what are the consequences on the reactor core of this after-heating rate under conditions where the fission chain reaction has ceased, but the coolant flow is reduced or stopped. At small times after shut down (hours to few days), the important radionuclides from a public point of view are isotopes of radioactive iodine, because iodine concentrates in cows' milk and, after ingestion of the milk, in the human thyroid. In the time period from a year to 100 years the most important radionuclides are strontium 90 (a beta emitter that concentrates in the bone and can cause leukemia) and cesium 137, with half-lives of 28.1 and 30.0 yr, respectively. Both of these isotopes produce long-term consequences to health in the event of severe reactor accidents.

After 300 years the fission products have decayed to a level where they are overshadowed by the transuranic elements produced by (non-fissioning)

9 These specific yields correspond to repository inventories that are in decay equilibrium with the corresponding reactor production rate. The main constituents of waste in the far future are the seven isotopes indicated in the upper right of this chart, which would be prime candidates for collection and transmutation (using either reactor or accelerator neutrons) to shorter-lived products to alleviate repository requirements.

neutron capture on the uranium/plutonium fuel. If these transuranics remain with the waste stream, the radioactivity level is about 10 times that of the original uranium ore. If plutonium and other transuranic elements are removed, the radioactivity of the waste stream is then less than the radioactivity of the original ore. Removal of plutonium in particular has the advantage that the waste disposal facility does not become a future "plutonium mine" for easily available bomb fuel after a few hundred years (Section 7.3.2.3). Removal of plutonium from the (still useful) uranium and the fission products (real waste) represents an inter-temporal trade-off related to human radiation exposure and proliferation risk about which clear resolution remains elusive. Additionally, the economics of plutonium recovery and recycle (to LWRs) is dictated largely by the price of uranium fuel (both for ore recovery and enrichment) and processing costs; generally, the economics based on present-day costs and technology are not favorable for plutonium recycle (i.e., reprocessing costs must be below —1000 $/kgHM for breakeven when uranium costs are 100 $/kgU).

7.2.5 Present status of nuclear power

Nuclear energy (NE) was introduced over four decades ago into the commercial electricity market, and presently provides 18% of the global electrical energy supply and contributes —8% of primary energy. A total generation capacity of 350 GWe is available in 442 nuclear power plants (NPPs) operated in 32 countries around the world (primarily in OECD countries, with a shift to the developing countries expected in the next century; Section 7.5). The mix of these reactors in 1993 was PWRs (55%, Fig. 7.5), BWRs (21%, Fig. 7.6, PWR/BWR = 2.7), GCRs (9%, Fig.7.7), PHWRs (7%, Fig. 7.8), water-cooled/graphite-moderated reactors (4.5%, all in the FSU), and other kinds of reactors (3.5%). Nuclear energy provides over 40% of the electricity in nine countries and 20-40% of the electricity supply in ten countries. If the 2130 TWeh of nuclear electricity generated in 1995 had been produced by fossil fuels, global carbon emissions would have been greater by 8%.

Nuclear energy was introduced rapidly in the 1960s and 1970s, with worldwide deployment rates of 13.5 GWe/yr in the 1970s and 14.6 GWe/yr in the 1980s; the average deployment rate in the US during this period was 5 GWe/yr (Diaz, 1998). No new orders for NPPs have been placed in the United States since 1978, leading to the capacity being pegged at 98.8 GWe. If nuclear energy is phased out in the US with no new plants being built, but with existing systems being operated for their presently licensed period, no nuclear power plant would be operating by the year 2030. On the other hand, if a 20-year license extension is approved, nuclear power generation would be extended to

Figure 7.15 New NPP capacity additions required according to the medium variant case considered in the IAEA study (Wagner, 1997).

the year 2050. Figure 7.15 illustrates this phase-out scenario, as well as possible levels of nuclear energy capacity additions out to the year 2050 (Meneley, 1997).

After an increase in operating costs during the late 1970s, and a very rapid increase in costs following the Three Mile Island accident in 1978, the overall operational and economic performances of NPPs in the USA have improved since 1985, as is shown in Fig. 7.16 (Diaz, 1998). This economic performance, however, still has not returned to 1973 levels. The economic performance of nuclear power varies somewhat from country to country. A recent survey (OECD, 1998a) of 15 OECD countries (Belgium, Canada, Denmark, Finland, France, Hungary, Italy, Japan, the Republic of Korea, the Netherlands, Portugal, Spain, Turkey, the United Kingdom, and the United States) and five non-OECD countries (Brazil, China, India, Romania, and Russia) on the cost of electricity generation from nuclear, coal, gas, biomass, solar (photovoltaic), and wind is summarized in Table 7.4. Production-cost comparisons with fossil fuels are also included on Fig. 7.16, with nuclear energy production costs (unburdened of capital costs) being competitive. (Section 7.4 addresses country-specific technological responses to future NE developments.)

A decrease in collective radiation dose per unit of generated electrical energy has occurred (Lochard, 1997), and a dramatic decrease in radioactive liquid releases (without tritium) per unit of generated energy for NPPs is also reported. The number of accidents with significant radiological consequences

Figure 7.16 US NPP operations and maintenance costs (a) and electricity production costs (b) over the past two decades (Diaz, 1998).

per unit of generated energy has also shown a dramatic decrease (Lochard, 1997).

7.2.6 Fuel supply

Following from the brief words in the overview (Section 7.1.4) we will show that fuel resources do not place a limitation on development of nuclear energy. The world uranium resources are summarized in Table 7.2 and Fig. 7.1 (Benedict, 1971; OECD, 1996, 1998a). Based on OECD estimates, Table 7.5 gives the energy equivalent of this resource. Proven reserves correspond to 1.5 MtonneU, reasonably assured reserves are 4.0 MtonneU, and total resources and reserves equal 18.5 MtonneU (excluding recoverable low-concentration

Table 7.4a Economic data for electricity generation condensed from a range of sources taken from a recent NEI survey (1998)a

($/kWey)

Fuel cycle (0/kWh)

Total bussbar generation cost (0/kWeh)

Nuclear

Coal

Gas

2110 1460 810

49 57 24

0.65

Note:

a All costs are expressed in 1 July 1996 dollars ($) or cents (0) of the United States. Capital costs include contingency, interest during construction at 5% discount rate, and refurbishment and decommissioning (also discounted at 5%) when applicable. Fuel cycle and total generation costs are calculated at 5% discount rate for 40 years lifetime and 75% load factor.

Table 7.4b Total bussbar generation cost /kWeh)

Year

1998

2030

Biomass

8

3.5

Solar PV

~30

5.2(3.5 + 2)b to 7(5 + 2)b

Wind

8(6 + 2)b

b The second number in the bracket is for the cost of backup electricity for these stochastic energy sources.

Note:

b The second number in the bracket is for the cost of backup electricity for these stochastic energy sources.

Table 7.5 Global occurrences of natural uranium (Mtonne)a

Category Mtonne U 100 EJc

Reasonably assured conventional reserves (<80 $/kgU) 4.0b 324.1

Undiscovered conventional resources (80-130$/kgU) 2.8 226.9

Speculative conventional resources (>130$/kgU) 2.0 162.1

Speculative unconventional resources (130-600$/kgU) 9.7 786.0

Total reserves and resources 18.5 1500.0

Notes:

b 1.5 Mtonne (122000 EJ) proven reserves.

c Based on complete fissioning of all uranium, of which 0.7205% is 235U.

uranium in granite and seawater). Complete fissioning of all 235U in the total reserves and resources (18.4 Mtonne) represents an energy resource of —10766 EJ (1.5 X 106 EJ if all uranium were fissioned). Including the less-known thorium (233U) resource would increase the 1.5 X 106 EJ resource by a factor of two to four. Compared to the fossil-energy resource listed in Table 7.6, the total uranium and fossil (including oil shale and clatherated methane) resources are comparable (e.g., 1.5 X 106 versus 1.0 X 106 EJ for total uranium versus total fossil), but the ratio of uranium fuel resource to fossil fuel resource without oil shale and clatherates (97892 EJ) amounts to —15.3.

The present world demands for total primary energy and for nuclear energy (350 GWe, —70% plant availability, 35% thermal-to-electric conversion efficiency) amounts to 360 EJ/yr and 22 EJ/yr, respectively. At the present nuclear energy demand rate, most (90.7%) of the 235U in the reasonably assured reserve category (4.0 MtonneU) would be consumed in a once-through (no plutonium recycle) fuel cycle by the year 2100. A linear increase of nuclear energy capacity to 2000 GWe would increase this 100-year uranium requirement by a factor of 2.7 (9.8 MtonneU), which amounts to approximately half of the 235U in the total reserves and resource category (Table 7.2). These estimates are based on the total consumption of the 235U resource; in actuality, only the fraction (xp/xf)(xf - xt)/(xp - xt) is extracted in the enrichment process, which for xp(product) = 0.035, xt(tailings) = 0.002, and xf(feed) = 0.007205 amounts to 77%. Advances that increase the efficiencies and decrease the cost of uranium enrichment (e.g., gaseous diffusion ^ centrifugation ^ laser isotope separation) can improve this situation first by extracting more of the 235U isotope and second by removing 236U buildup from recycled uranium fuel (Meneley, 1997). Generally, the long-term uranium requirement for a given demand scenario depends strongly on the efficiency with which the energy content in uranium (and thorium) is utilized, and the degree to which the world uranium resource is developed in the long run is dependent on the extent to which the four cardinal issues for nuclear energy are resolved (i.e., safety, waste, proliferation, and cost). Lastly, advances in removing economically even a small fraction of —4000 MtonneU from seawater at cost 200-300 $/kgU or less would indefinitely extend the resource limit for fission while possibly obviating the need for the fast-neutron, plutonium breeder.

Central to the sustainability of nuclear energy in the longer term (>100 years) is the efficient use of the uranium (and thorium) resource (Section 7.2.6). To illustrate quantitatively the present situation and the potential for waste, a simplified once-through uranium fuel-cycle for an LWR (OT/LWR) is considered, wherein uranium ore is mined at a rate E(tonneU/yr), enriched in 235U content to a weight fraction xp, and fissioned in an LWR of capacity PE,

Table 7.6 Summary of fossil fuel reserves and resources (EJ)

Oila

• Identified reserves

• Undiscovered reserves

• Ultimate reserves

6299 2626

8925

Heavy and extra-heavy crude oil6

• Identified resources

• Undiscovered resources

• Total resources

2719 541

3261

Bitumen,6 original identified recoverable resources

2490

• Measured resources

• Indicated resources

• Total resources

18741 60559

79300

Conventional gasa

• Identified reserves

• Undiscovered reserves

• Total reserves

5239 4652

9890

Unconventional gasc

• Coal-bed methane

• Fractured shale

• Tight-formation gas

• Clatherates

Remaining in situ after production Total less clatherates

9702 17262 7938 783216 494

• Proven recoverable reserves

• Anthracited

• Sub-bituminous

Additional recoverable reserves Total reserves

22941 15217 3942 3782 14989

37930

Total fossil fuel

• Less clatherates

• Less clatherates and oil shale

960408 177192 97892

a Masters et al. (1994). 6 Masters, et al. (1987). c Rogner (1996, 1997). d WEC (1995).

thermal-to-electric conversion efficiency %H, and a once-through (no plutonium recovery/recycle) exposure of BU (GWtd/tonneHM). The rate of used-fuel generation is nominally 26 tonneHM/GWeyr. This spent material for a typical LWR contains xPu ~ 0.009 weight fraction plutonium, of which ~73% is fissionable; for the present world nuclear energy capacity of 350 GWe, the plutonium generation rate amounts to ~80 tonnePu/yr, without recycle as MOX (mixed plutonium-uranium oxide) back to the LWRs. The rate of potential energy loss relative to PE to both the repository and the depleted-uranium enrichment-plant tailings stream can also be estimated: ~0.5% of the (electric) energy stored in the uranium resource is extracted by the OT/LWR fuel cycle. For the tailing and enrichment fractions xt = 0.002 and xp = 0.035, 6.4 units of natural uranium are needed to produce one unit of low-enriched uranium (LEU) feed to the reactor. The rate of uranium resource depletion amounts to 165.6 tonneU/GWeyr for these conditions, and at the present world nuclear energy capacity of PEo = 350 GWe, the cumulative uranium demand would be 5.9 MtonneU by the year 2100; if that capacity were to triple linearly to 1050 GWe by the year 2100, the uranium resource requirement would double in 2100 to 11.8 MtonneU. These uranium demands compare to 2.12 Mtonne in reserves, 4.51 MtonneU in known resources, and 15.5 MtonneU in total conventional resources (OECD, 1996) (Table 7.5; OECD, 1998a,b). Generally, fuel resources do not present an important limitation to the use of nuclear energy for most of the 21st century; public acceptance of this large, hazardous technology does, however.

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