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Nuclear Energy

Fission nuclear energy is released from the splitting of a fissionable nucleus struck by a neutron. In nuclear reactor applications, relevant neutron energies lie around or below 10 MeV (1eV=1.60210 × 10−19 Joule) and thus only certain isotopes of thorium, uranium and plutonium are fissionable in the practical sense. Nuclei which can undergo fission when absorbing an arbitrarily slow, or low-energy neutron are called fissile. Nuclear fuel is always designed to contain at least one of the fissile isotopes U233, U235, Pu239 and Pu241. Of these, U235 is the only one that occurs in any separable quantity in nature (Zweifel, 1973).

Besides yielding fission product nuclei, emitting gamma radiation and releasing energy, additional neutrons are emitted from fission. These emitted secondary neutrons can be utilized to induce further fission reactions. Hence a chain reaction can be perpetuated. When nuclear reactors operate at a steady fission rate they are referred to as being critical.

The neutron was discovered in 1932 by Chadwick and fission was discovered in 1938 by Hahn and Strassmann, while the first self-sustaining chain reaction experiment was carried out by a team led by Fermi in 1942 (ANS, 1992).

Fusion

Nuclear fusion refers to the merging of two light nuclei into a heavier one with a simultaneous release of energy. Solar energy is generated by fusion. Research and design work are under way to develop a fusion reactor in which a stable, confined fusion reaction chain could produce a net amount of energy.

Reactor physics

In addition to inducing fission, neutrons can interact with fuel and reactor materials in a number of other ways. Neutrons can be lost through radiative capture or can interact by scattering from these nuclei. Neutron scattering collisions with lighter nuclei are elastic and reduce neutron energy very effectively. The probability of these interactions is measured in terms of cross-sections specified for each combination of target nuclei and colliding particles, among which the neutron is the primary one.

The probability of a neutron inducing a fission reaction-i.e., the fission cross-section-of a fissionable nucleus depends strongly on the kinetic energy of the impinging neutron. Taking into account all parasitic absorptions, there are two energy ranges of neutron energies which appear most favorable for sustaining the chain reaction. Depending on which one of these two neutron energy domains a given reactor operates in, it is referred to as a thermal or, alternatively, a fast reactor (Zweifel, 1973).

In thermal reactors the bulk of the neutron collisions with the fuel that lead to fission occur with neutrons of energies of between 0.01 and 1 eV. These energies correspond to the energy of thermal motion of the target nuclei. Neutrons released from fission have energies of the order of 1 MeV. In thermal reactors, the neutron energies are reduced, or neutrons slow down through elastic collisions in the moderator. In collisions, light elements absorb neutron energy more effectively than heavy ones. Therefore, typical moderators are H2O, D2O, graphite and beryllium (Rahn et al., 1984; Glasstone and Sesonske, 1981).

Fast reactors, on the other hand, are designed so that the bulk of neutrons lie at energies above 1 keV. Rather than involving a moderator, these cores are designed to avoid collisions with light nuclei. The heat generated in the reactor core is removed using a liquid or gaseous coolant circulating from the core to a heat exchanger. The most commonly used coolant is ordinary water. Light-water reactor (LWR) refers to a reactor type in which ordinary water has a dual function, serving both as the moderator and as the coolant. Other coolants used in thermal reactors include heavy water (D2O) and He as well as CO2 gas. A fast reactor core is by nature very compact and the coolant is required to have a large thermal capacity. Therefore, liquid sodium metal is used in the cooling circuit of fast reactors.

The triad consisting of fuel, moderator and coolant is used in Table N4 to classify the most important power reactors currently in use. About 75 per cent of the installed nuclear capacity in 1993 was based on light-water reactors (IAEA, 1992).
Table N4 Most common types of nuclear power reactors

Moderator

Coolant

Fuel

Cladding

Type designation

Water

Water

UO2, enriched to about 3%

Zirconium alloy

LWR, BWR, PWR, VVER

Graphite

CO2

U metal, natural U, UO2 enriched to 2%

Magnesium, steel

Magnox, UNGG, AGR

He

UO2, UC, enrichment can vary from 3% to 93%

Graphite

HTR

Water

Enriched UO2

Zirconium, steel

RBMK

Heavy water

Heavy water

UO2, natural U

Zirconium

CANDU

Water

Enriched UO2

Zirconium

SGHWR, ATR

None

Liquid sodium

U-PuO2

Steel

LMFBR

There are two major types of LWRs as illustrated in Figure N9. In a pressurized water reactor (PWR) the primary coolant circuit is kept under pressure at around 150 bar. The heat is transformed into steam in a steam generator, from the secondary side of which the steam flows into a turbine. In a boiling water reactor (BWR) the coolant pressure is around 70 bar and the coolant water is allowed to boil within the reactor core. Steam is separated within the reactor vessel and proceeds then directly to the turbine-generator part of the plant (Weisman, 1983).
Figure N9 Light-water nuclear reactors (LWRs):
(a) pressurized water reactor (PWR), (b) boiling water reactor (BWR)

Nuclear fuel

The relative abundance of the isotope U235 is 0.71 per cent in natural uranium. As indicated in Table N4, a multiplicative system cannot be based on natural uranium except in the cases where heavy water or graphite is used as the moderator. In most reactor types uranium is enriched, or its relative content of U235 is increased.

Uranium is extracted as U3O8 from the ore. The mill tailings slurry flows into a tailings pond. As the tailings contain natural radioactive decay products, the operation must be managed with care. Prior to enrichment uranium is converted into uranium hexafluoride (UF6). In a LWR a typical enrichment is 3.0 to 3.5 per cent in U235. Enrichment technologies include gaseous diffusion, centrifuge and laser separation. Enriched uranium is then converted into UO, for fabrication. In LWRs uranium dioxide is manufactured into pellets that have a diameter of the order of 7.5 to 9 mm and a height of about 10 mm. These UO2 pellets are assembled into a cladding tube made of a zirconium alloy to form fuel rods. A commercial LWR core contains some 50,000 or more fuel rods. The height of the core is about 4 m. New fuel is loaded into the reactor core during a refueling shutdown every 12 to 18 months. Typically, a reload batch is a third of a PWR core and a fifth of a BWR core and, consequently, fuel residence time varies between 3 and 5 years.

A 1,000 MW (electric) LWR operating at a 75 per cent load factor produces over 6.5 TWh of electricity per year. The amount of fresh enriched uranium fuel required annually is about 20 tonnes. Depending on the tails assay, which usually varies from 0.2 to 0.3 per cent of U235, an amount of some 7 tonnes of natural uranium is consumed for each ton of uranium enriched to 3.5 per cent (Silvennoinen, 1982). The materials balance for the front end of the fuel cycle, or the pre-reactor stages, is summarized in Table N5.
Table N5 Materials balance for a 1,000 MW LWR fuel loading

Output

Stage

Input: annual amount, tonnes U

Chemical compound

Annual amount, tonnes U

Mining and milling

30,000-50,000

U3O8

140

Conversion

140

UF6

140

Enrichment

140

UF6

20

Fueld fabrication

20

UO6

20

Reactor core

20

UO2

20 spent fuel

When spent fuel is discharged from the reactor there are two options to follow. After a period of interim storage, spent fuel can be classified as waste and disposed of (see below in this entry and also entry on Underground Storage and Disposal of Nuclear Wastes). Alternatively, spent fuel can be reprocessed. The separated uranium and plutonium can be recycled. For this reason, the flow of nuclear materials is referred to as a fuel cycle.

A simplified chart of the LWR fuel cycle is show in Figure N10. The 20 tonnes of spent fuel discharged annually from a 1,000 MW LWR contain some 200 kg of fissile plutonium which can be used as mixed U-PuO2, or mixed oxide for fueling a LWR or a fast reactor.
Figure N10 Light water reactor fuel cycle

Development of power reactors

The experimental fast breeder reactor EBR-1 generated the first nuclear electricity in the USA in 1951. The first nuclear reactor to generate an appreciable amount of electricity was the 5 MW APS-1 in Russia in 1954. APS-1 was the precursor of the notorious Chernobyl RBMK type (boiling-water cooled, graphite moderated) reactors. The first UK power reactors at the Calder Hall plant were gas-cooled, natural uranium fueled, graphite-moderated Magnox reactors. France and Canada also developed very early their own types of power reactors using natural uranium fuel. In Table N4 they are referred to as UNGG and CANDU, respectively (Rahn et al., 1984; Leclerq, 1986). The first nuclear power station to produce commercial nuclear electricity in the USA was the 60 MW Shippingport PWR, which was in operation from 1957 to 1982 (ANS, 1992). In the 1970s the French went over to LWRs, which is now the dominating reactor type also in Japan and Germany. The first commercial British LWR came onstream in 1994.

The unit size of LWRs has been gradually increased to 1,200-1,400 MW. In most cases there are several reactor units, sometimes up to eight reactors, on a given power plant site.

Nuclear electricity generation

In 1993 there were about 430 power reactors in operation in 29 countries, representing a total installed capacity of 343,000 MW. Some 70 plants were under construction, with a total generating capacity of 58,000 MW. In the early 1990s only a few orders for new nuclear power plants have been placed, mainly in Far Eastern countries with high economic growth rates (IAEA, 1992).

Worldwide the total nuclear electricity supplied exceeds 2,000 TWh/a; 17 per cent of the total electricity generated and about 6 per cent of all commercial primary energy use is nuclear. Other major sources of electricity are coal (42 per cent), hydropower (19 per cent), natural gas (12 per cent) and oil (10 per cent) in the 1990 statistics. LWR has taken by far the largest market share, PWR 25 per cent and BWR 20 per cent. Gas-cooled reactors represent 9 per cent, followed by CANDUs with 7 per cent and RBMKs with 5 per cent. The world's largest producer of nuclear electricity is the USA, with 613 TWh from 113 reactors. In 1991, 12 countries produced more than 25 per cent of their electricity in nuclear plants. France headed the list with about 75 per cent.

The technical design lifetime of nuclear power plants is 30 to 40 years. Large-scale retirement of existing plants will start during the decade 2000 to 2010. Major research and development and demonstration programs investigate how to alleviate materials aging phenomena and provide a basis for extension of the life of power plants.

Cost of nuclear electricity

Nuclear power is a capital-intensive mode of electricity production. An illustrative cost breakdown is given in Table N6. Fuel costs typically amount to less than 20 per cent of the total cost of nuclear electricity. The three major components of the front-end fuel costs relating to purchase of natural uranium, enrichment and fuel fabrication are roughly of equal size.
Table N6 Breakdown of the nuclear costs for a 1,000 MW LWR

Cost component

Estimate in US$, excluding taxes

Approximate percentage contribution to generating cost

Construction costs

1,750 million

63

Fixed operating costs

20 million/year

7

Variable operating costs

3 per MWh

9

Fuel costs

5 per MWh

17

Nuclear waste management and decommissioning

900 million

4

The illustrative cost estimates of Table N6 can be converted into explicit kWh costs as soon as the costing basis, discount and interest rates and plant performance parameters are fixed. The relative contributions shown in Table N6 reflect typical values for OECD countries. In 1991 US dollars the bus-bar production costs vary typically from 35 to 50 million per kWh in OECD countries (OECD Nuclear Energy Agency, 1993). In calculating the relative share of waste management and decommissioning costs for Table N6, these costs are discounted to their present value. As a major part of the cost is incurred in the future, the present value is small compared to the nominal one.

Nuclear safety

Safety objectives and principles

The International Atomic Energy Agency (IAEA) has assembled a comprehensive set of safety objectives and principles (International Nuclear Safety Advisory Group, 1988). The document reflects the latest practices and requirements that have evolved in the leading countries of nuclear power technology. Detailed safety requirements can vary considerably between different countries.

Release of radioactive materials during normal operation

Nuclear power plants have technical systems for concentrating the radioactive materials collected from the plant process systems and for storing and preparing them for appropriate disposal (see below in this entry and also entry on Underground Storage and Disposal of Nuclear Wastes). Small amounts escape or are released in a controlled manner into the environment. Releases and concentrations in the environment are monitored continuously by instruments and by collecting samples in the vicinity of the plant (Rahn et al., 1984). Gaseous releases into the air consist primarily of noble gases (krypton and xenon) and iodine. They are formed as fission products in the fuel. Minute amounts of these nuclides occasionally escape into the reactor coolant circuit from leaking fuel rods. On average, less than 1 in 10,000 fuel rods develops a leak during the 3- to 5-year residence time in the reactor. Liquid releases include activated corrosion products (cobalt, manganese, iron and zinc), and tritium. The long half-life (12 years) of tritium and its propensity to become part of water molecules makes it difficult to remove it from the process waters.

Regulatory authorities set limits on permissible releases from the plant. These limits imply that the radiation dose from the releases to the most exposed group of the population in the plant vicinity is only a small fraction of the dose received from other natural and man-made sources. A representative dose limit for normal operation is 0.1 mSv/yr. This can be compared to radiation from natural sources from which an average person can receive an annual dose of 4 to 6 mSv, depending on location, mostly from indoor radon. Actual releases from operating plants are a small fraction of the permissible levels.

Successive safety barriers

The risk due to reactor accidents is managed by preventing the release of significant amounts of radioactive substances into the environment by means of successive barriers. The strategy of defense-in-depth is adopted to maintain these barriers under all conceivable conditions (Glasstone and Sesonske, 1981).

In LWRs the following four barriers exist:
  • the ceramic fuel pellets effectively bind the radioactive fission products;
  • the fuel cladding tubes are gas-tight metal tubes;
  • the reactor core is housed in a steel pressure vessel with a wall thickness up to 150 mm (PWR); and
  • the reactor vessel and most components of the reactor cooling system are located within a heavy leak-proof containment building, made of steel plates or reinforced concrete.

There are further structures outside the primary containment that form a secondary containment. Together the containment buildings also provide protection against external impacts, such as crashing airplanes. Only if all the barriers fail can an accident with considerable environmental consequences develop.

Prevention of accidents is the first safety priority. The occurence of events which might initiate accidents is also minimized by relying on qualified staff, applying systematic quality assurance, requiring wide safety margins and incorporating fault tolerance in the design, and specifying limits for acceptable operating conditions.

Nevertheless, designers must assume that equipment and system failures and human errors are possible and incorporate engineered safety systems to detect and remedy them and to prevent the escalation of abnormal conditions into accidents. There are multiple automatic systems that shut the reactor down, provide emergency cooling of the core and the containment and supply independent power for essential functions. Should a severe accident happen, management strategies and on-site and off-site emergency plans are developed to mitigate the consequences of even very unlikely accidents.

Quantitative safety goals

Probabilistic safety assessment (PSA) provides a methodology for assessing the probability of consequences of different faults and errors that may take place at the plant. Within the PSA framework, quantitative safety goals for plant design and operational safety performance are defined. A widely accepted safety target for existing nuclear power plants implies that the likelihood of occurrence of severe core damage is less than once per 10,000 operating years. Accident management and mitigation measures should further reduce the probability of a large release of radioactivity into the environment necessitating short-term off-site protective measures or resulting in extensive restrictions in land-use, to less than once in 100,000 years (International Nuclear Safety Advisory Group, 1992).

Safety experience

Experience from operating nuclear power plants exceeded 6,500 reactor-years at the end of 1992 and accumulates at a rate of more than 400 reactor-years annually.

As far as commercial nuclear power plants using light- or heavy-water reactors are concerned there has been one severe core damage accident, that at the Three-Mile Island plant in 1979 (Rahn et al., 1984). The successive barriers proved effective and there was a rather insignificant release of fission products from the plant. The accident resulted in a collective dose of some 30 to 40 man Sv distributed over a population of about 1 million people. The inhabitants within a five mile radius from the plant received an average dose of 0.08 mSv. Largely due to the lessons learned from the Three Mile Island accident, the current probability of severe core damage to US reactors is now lower by a factor of more than 10 in comparison with the situation before 1979.

The Chernobyl accident in 1986 resulted in a considerable release of radioactive materials and major off-site protective measures were needed to mitigate the consequences. It is the most serious accident ever experienced at a commercial nuclear power plant. The accident was made possible by, and its severe consequences were a result of, a combination of human failures, negligent safety culture and major design deficiencies. The equivalent collective dose commitment due to the Chernobyl catastrophe is estimated at 600,000 man Sv. Some 40 per cent of the exposure was experienced in Ukraine and the neighboring parts of the former Soviet Union and 57 per cent in the rest of Europe, with the remaining 3 per cent being distributed in other countries in the northern hemisphere (UNSCEAR, 1988).

The International Nuclear Event Scale (INES), is an internationally agreed system operated by the IAEA for consistent classification of incidents and accidents at nuclear power plants. It aims at standardizing the reporting of nuclear events worldwide and facilitating communication between the nuclear community, the media and the public. The criteria for the classification of less severe events are in broad terms as follows:
  • Level 1, Anomaly. Essential deviation from the authorized operating regime.
  • Level 2, Incident. Spread of contamination on-site; or, overexposure of a worker; or, significant failures in safety provisions.
  • Level 3, Serious incident. Very small release, public exposure at a fraction of prescribed limits; or, acute health effects to a worker; or, near accident, no safety layers remaining.

Events of classes 4 to 7 represent actual accidents of increasing seventy. The Three Mile Island accident is classified as Level 5, based on the impact on-site. The Chernobyl accident with its widespread health and environmental effects is classified as Level 7.

National bodies, such as the US Nuclear Regulatory Commission, and international organizations, such as the World Association of Nuclear Operators, have focused their attention on the development of operational and safety performance indicator systems. The main purpose of these indicators is to identify early signals of deteriorating performance and provide a warning of impending problems before an actual incident or accident occurs. In addition to their preventive function, the indicators also monitor the effectiveness of corrective actions taken at the plant. Some representative indicators are listed in Table N7, together with typical actual values reached at well-performing plants.
Table N7 Performance indicators for nuclear power plants

Parameter

Actual values reached at well-performing plants

Plant capacity factor

80%

Number of reactor safety shutdowns per year

<1

Collective annual radiation dose to plant workers

<1 man Sv/unit

Radiation dose to the most exposed group of population in plant vicinity as inferred from releases

1 μSv/yr

Volume of plant operational waste

<100 m3/yr/unit

Number of operational events with safetysignificance:

<1/yr

INES severity level 1

INES severity level 2

A few during the plant life

INES severity level 3 or higher

Very rare

Nuclear wastes

Waste categories

Radioactive wastes fall into two distinct categories: low- and high-level wastes. Low-level wastes contain small amounts of radioactive nuclides with short half-lives. These wastes arise from the plant operation and range from used filters and ion-exchange resins of reactor water purification systems to protective clothes and scrap from maintenance work. Sometimes the low-level wastes contain higher amounts of radioactivity and are referred to as intermediate-level waste.

The wastes of the lowest level of radioactivity are usually compacted at high pressure or incinerated with retrieval and subsequent compaction of the ashes. Wet wastes are immobilized in concrete or bitumen. Decommissioning of reactors yields low-level short-lived wastes similar to those from the operation of the reactor. The volume of low-level waste generated annually in operating a 1,000 MW nuclear power plant varies between 300 and 1,000 m3 depending on reactor type, volume reduction and immobilization techniques and regulatory limits applied. The volume of wastes arising from decommissioning of a large LWR is about 7,000 to 20,000 m3 (OECD Nuclear Energy Agency, 1986).

The fission and activation products generated in nuclear reactors are almost exclusively contained in the used fuel elements. Consequently, it is a major undertaking in nuclear waste management to provide safe and cost-effective techniques for the storage, transport and processing of the high-level wastes originating from spent fuel. Some decommissioning wastes contain a relatively high amount of radioactivity and may require similar treatment as spent fuel.

Spent fuel

The activity inventory of spent fuel as a function of time after reactor shutdown is presented in Table N8. Only the most important nuclides are identified specifically, but the total and subtotal figures cover the whole spectrum of radioactive nuclides present in the spent fuel. Up to about 100 years' cooling time, the total radioactivity content is dominated by fission products and thereafter by actinides and their daughter nuclides. Although the reduction rate by radioactive decay is initially quite fast, the spent fuel remains highly radioactive for long time periods, requiring careful long-term isolation from the biosphere.
Table N8 Radionuclide inventories of spent BWR fuel with an initial enrichment of 3.6 per cent irradiated to an average discharge burn-up of 36 MWd/tU after different cooling times, GBq/tU

Cooling time (yrs)

Nuclide

Half-life (yrs)

0

100

10,000

million

Fission and activation products

C-14

5.7 × 103

1.0 × 101

1.0 × 101

3.0 × 100

Se-79

6.4 × 104

1.6 × 101

1.6 × 101

1.5 × 101

Sr-90

2.9 × 101

2.8 × 106

2.6 × 105

Zr-93

1.5 × 106

7.1 × 101

7.1 × 101

7.1 × 101

4.5 × 101

Tc-99

2.1 × 105

5.2 × 102

5.2 × 102

5.0 × 102

2.0 × 101

Pd-107

6.5 × 106

4.9 ×10×0

4.9 × 100

4.9 × 100

4.4 × 100

Sn-126

1.0 × 105

3.2 × 101

3.2 × 101

3.0 times; 101

3.1 × 10-2

I-129

1.6 × 107

1.3 × 100

1.3 × 100

1.3 × 100

1.2 × 100

Cs-135

2.3 × 106

2.4 × 101

2.4 × 101

2.3 × 101

1.7 × 101

Cs-137

3.0 × 101

4.1 × 106

4.1 × 105

Subtotal

4.2 × 109

1.3 × 106

7.5 × 102

1.3 & times; 102

Actinides and their daughters

Ra-226

1.6 × 103

1.0 × 10-3

6.4 × 100

1.9 × 101

Th-229

7.3 × 103

8.1 × 10-1

4.4 × 101

Th-230

7.7 × 104

5.1 × 10-2

8.2 × 100

1.9 ×101

Pa-231

3.2 × 104

2.3 × 10−3

1.4 × 10−1

1.1 × 100

U-233

1.6 × 105

8.5 × 10−3

2.4 × 100

4.4 × 101

U-234

2.5 × 105

3.8 × 101

7.0 × 101

9.5 × 101

1.7 × 101

U-235

7.0 × 108

6.7 × 10−1

6.7 × 10−1

7.8 × 10−1

1.1 × 100

U-236

2.3 × 107

1.0 × 101

1.0 × 101

1.4 × 101

1.5 × 101

U-238

4.5 × 109

1.2 × 101

1.2 × 101

1.2 × 101

1.2 × 101

Np-237

2.1 × 106

1.6 × 101

2.1 × 101

5.7 × 101

4.2 × 101

Pu-238

8.8 ×101

1.5 × 105

7.5 × 104

Pu-239

2.4 × 10 4

1.3 × 104

1.3 × 104

1.0 × 104

Pu-240

6.5 × 103

1.9 × 104

1.9 × 104

6.8 × 103

Pu-242

3.8 × 105

8.1 × 101

8.1 × 101

8.0 × 100

1.4 × 101

Am-241

4.3 × 102

8.8 × 103

1.8 × 105

8.4 × 100

Am-243

7.4 × 103

1.1 × 103

1.1 × 103

4.3 × 102

Subtotal

1.2 × 109

3.5 × 104

1.8 × 104

7.7 × 102

Fuel cladding and other structural materials

C-14

5.7 × 103

1.3 × 101

1.3 × 101

3.9 × 100

CI-36

3.0 × 105

4.2 × 10−1

4.2 × 10−1

4.1 × 10−1

4.2 × 10−1

Ni-59

8.0 × 104

9.6 × 101

9.6 × 101

8.8 × 101

1.7 × 10−2

Zr-93

1.5 × 106

1.6 × 101

1.6 × 101

1.6 × 101

1.0 × 101

Nb-94

2.0 × 104

4.9 × 100

4.9 × 100

3.4 × 100

Subtotal

1.6 × 107

7.7 × 103

1.3 × 102

2.0 × 101

Grand total

5.5 × 10 9

1.7 × 10 6

1.9 × 10 4

9.2 × 10 2

Spent fuel management strategies

Once removed from the reactor core, spent fuel is stored on-site in water pools before it is either placed in interim storage away from the reactor for a period which may last decades, conditioned after the decay period, and stored before its final disposal in a geologic repository; or reprocessed after additional storage away from the reactor. The resulting liquid high-level waste, containing mostly fission products and a small proportion of the actinides, is then immobilized in a stable borosilicate glass matrix for disposal in a geologic repository. The direct disposal policy has been adopted in the USA, Spain and Sweden, whereas reprocessing policy is followed in France, Japan and Belgium.

The first one of these two alternatives corresponds to the so-called once-through fuel cycle and implies that spent fuel is considered as waste, even if it still contains valuable fissile material. In the other option the spent fuel is chemically processed to separate waste products and useful uranium and the plutonium produced from uranium by neutron capture in the fuel during irradiation in the reactor core (see section on Nuclear fuel in this entry). The separated uranium and plutonium can be used for fabricating new fuel elements.

The radionuclides contained in the remaining high-level waste stream decay considerably faster in the long term than those in the unreprocessed spent fuel. These economic and safety benefits of the reprocessing option are to be balanced against the additional costs involved in the chemical processing.

The radionuclides of principal concern in the atmospheric and liquid effluents from a reprocessing plant are the long-lived nuclides: 3H, 14C, 85Kr, 90Sr, 106Ru, 129I, 134Cs, 137Cs and isotopes of transuranium elements. The activity of the effluents depends upon the specific treatment, processing and effluent control design of the plant. According to assessments published by the UN, a normalized local and regional collective dose equivalent commitment brought about by atmospheric effluents is estimated to be about 0.3 man Sv/GWa, or 0.7 man Sv/GWa from aquatic discharges (UNSCEAR, 1982).

Disposal facilities for spent fuel or high-level waste

The preferred option for long-term disposal of either spent fuel as such or the high-level waste from reprocessing is to place it in a deep geologic repository. In generic feasibility studies, several geologic host media have been shown to be suitable for safe long-term isolation of long-lived radioactive wastes. Final disposal of spent fuel in crystalline rock is illustrated in Figure Nlla (Salo, 1992). The tunnel system is located at a depth of about 500 m. In each tunnel several individual disposal holes are drilled into the floor. Each hole houses one disposal canister and is surrounded by the clay-like material bentonite. After contact with groundwater, ben-tonite swells considerably and its water conductivity is greatly reduced. The encapsulation of fuel bundles is also shown in Figures Nllb and Nllc. The disposal canister has a two-layer structure; the outer copper canister provides a long-lasting barrier against corrosion, while the inner steel canister provides the necessary strength against mechanical stresses.
Figure N11 Encapsulation and final disposal facilities for spent nuclear fuel: (a) final disposal of spent fuel in crystalline rock; (b) canister for spent nuclear fuel; (c) canister for spent nuclear fuel in disposal tunnel
Figure N11

Assessment of long-term safety of waste repositories

The long-term safety of final disposal in a geologic repository must be demonstrated by experiments at a sub-system level. Moreover, the overall long-term performance of the total system has to be assessed by employing a set of mathematical models describing the behavior of the various barriers. A considerable amount of research and development has been carried out in this area, including the development of underground rock laboratories and other near- and far-field research facilities (OECD Nuclear Energy Agency, 1991).

Although the disposal of high-level waste has yet to be demonstrated in practice, a widely accepted view is that deep geologic disposal of spent fuel and high-level waste, using the multiple barrier concept, is the most technically sound, feasible and safest option available. An appropriate use of safety assessment methods (OECD Nuclear Energy Agency, 1988), coupled with sufficient information from disposal sites, can provide the technical basis to assess whether any given disposal system would offer a satisfactory level of safety for both the present and future generations.

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  • Cross-references
  • Hazardous Materials Transportation and Accidents; Hazardous Waste; Nonrenewable Resources; Nuclear Winter, Possible Environmental Effects; Risk Assessment; Underground Storage and Disposal of Nuclear Wastes