Air cooling was used in the vety first generation of nuclear reactors, namely, graphite-moderated natural uranium “piles,” which were built in both the

United Kingdom and the United States in the 1940s. The largest air-cooled reac­tors were at the U. K. Atomic Energy Authority’s Windscale establishment and were designed for plutonium production. The main problem with air as a coolant is that it is an oxidant, i. e., it supports combustion. In the case of the Windscale graphite-moderated piles, there was something of a dilemma: if the pile temperature was too high, the graphite oxidized, but if the pile temperature was too low, the graphite atoms could become permanently displaced from their natural positions by neutron bombardment. At higher temperatures, the atomic vibrations are sufficient to shake them back to their normal positions. Displacement of the atoms results in energy being stored, with possible acci­dent connotations, which we will discuss in the context of the Windscale acci­dent in Chapter 5.

Despite the ready availability of air, its oxidizing properties rule it out as a vi­able coolant in modern high-temperature reactors.

The Browns Ferry nuclear power plant in Alabama consists of three 1065- MW(e) boiling-water reactors. On March 22, 1975, a workman who was lying on his side used a lighted candle to test for leakage of air around cable pene­trations through a concrete wall at the Unit I plant. A hole was found, and the workmen stuffed some polyurethane sheet into it and tested again for leaks. The leak persisted, and the candle flame ignited the polyurethane sheet. The air rnshing into the hole spread the fire into the hole and away from the workmen, so that they could not extinguish it with fire extinguishers. The fire burned for 7 h before it was put out. Units I and 2 were both at full power when the fire started. (Unit 3 was under construction and was not affected by the accident.) The fire spread horizontally and vertically, affected about 2000 cables, and caused damage that cost about $10 million to repair. There was a reluctance to use water on the fire until both reactors were in a stable shutdown condition because of the possibility of short-circuiting. Once water was used, the fire was rapidly put out.

Both reactors were shut down. However, because of the fire, both the shut­down cooling system and the emergency core cooling system for Unit 1 were in­operable for several hours. The operators had to use alternative means of injecting water into the reactor, which included a pump used in connection with the control rod drive system and pumps used for returning condensate to the sys­tem. The use of these alternative water supplies required depressurization of the reactor, and during this maneuver, the water level over the core dropped to 1.2 m above the top of the fuel. However, sufficient cooling was provided throughout the incident to prevent the core from overheating. No significant problems were encountered with the cooling of Unit 2, and the high-pressure cooling system (HPIS) was successfully initiated. There was no release of radioactivity off-site, and no one on the site was seriously injured. Both units were, however, out of operation for over 1 year while the damage was repaired.

The main lesson from the Browns Feny incident was related to what is called common mode failure. All the cables related to the safety systems passed through a single duct and failed in a common mode (despite the diversity in­troduced as discussed in Chapter 4), and all the systems failed when there was a fire. The moral is that the designer should ensure that each of the indepen­dent systems is truly independent and that supplies and controls to the instm — mentation and actuation devices should not pass along common ductwork. The technical term for this is segregation, and after the Browns Feny incident the provisions for segregation were significantly improved. For example, 3-h fire-re­sistant physical barriers are now placed between components, and when this is not possible the cables are separated by significant distances (typically 7 m) and protected by active fire-fighting equipment so that the possibility of a fire spreading from one to another is remote.

Figure 7.5 shows the refueling route for a large pool-type fast breeder reactor. The objective in this refueling process is to keep the used fuel permanently under sodium, which acts as a heat sink for the decay heat. The fuel is extracted by a grab attached to a rotating plate above the reactor. It is extracted from the core and, still under sodium, is passed into an intermediate buffer store, from which it is transferred through a sloping transfer line (also sodium-filled) to a sodium-cooled spent fuel store, where it is kept for 100-200 days before being transferred to the reprocessing plant.

What is energy? There is general awareness of the problem of depletion of the world’s energy resources. People understand energy in terms of those re­sources, namely, the supplies of oil, gas, and coal and the electricity derived from them. All of these items have made an increasingly large demand on na­tional and personal budgets.

The engineer has, by training, a somewhat different concept of energy. This de­rives from his or her undergraduate training in the field of thermod^ynamics, which is the science of energy and energy conversion. We do not intend to try to provide a basic course in thermodynamics; however, for the rest of this ^юк to be rea­sonably intelligible, it is important that some of the basic concepts be stated.

The concept of doing work to lift objects or to move an object such as a bi­cycle along is a commonly accepted one. Thus, it is relatively easy to under­stand the concept of energy as a measure of the ability to do work. Energy can appear in different forms as follows:

1. Kinetic Energy. This is energy associated with movement, for example, that of a flywheel or a moving locomotive.

2. Potential Energy. This is energy possessed by virtue of position, typically in the earth’s gravitational field. For instance, a child sitting on the higher end of a seesaw has greater potential energy than a child sitting on the lower end. Likewise, water in a mountain lake has greater potential energy than water at sea level.

3. Chemical Energy. Matter consists of atoms that are combined together in molecules. Molecules of different substances can react to release energy,

and this releasable energy is often termed chemical energy. For example, chemical energy is released when gasoline combines with air in the cylin­ders of a car’s engine.

4. Electrical Energy. Atoms consist of a central mass, known as the nucleus, around which a cloud of electrons circulates (see Figure 1.1). If there is an excess or deficit of electrons in one part of a body, the body is said to have an electrical charge and, by virtue of this, to have electrical energy. An ex­ample of this is a thunderstorm, where the clouds are charged electrically with respect to the ground.

5. Nuclear Energy. Normally, the nucleus of an atom is stable and will re­main indefinitely in its present state. An example is the nucleus of an atom of iron; no matter how much we would like it to happen, iron will never change into another element, such as gold. However, the atoms of some el­ements are unstable and can change into another form spontaneously, by the emission of radiation. We shall discuss the forms of radiation emitted further in Section 1.2; it is sufficient here to note that the radiation emitted has kinetic energy and the disintegration process results in the release of energy associated with the nucleus, namely, the nuclear energy. If the nu­cleus could be weighed before the disintegration, and the resulting nucleus and all particulate components of the radiation weighed afterward, it would be observed that a small change in mass had occurred due to the conver-

sion of mass into energy. The relationship between the loss of mass m and the energy released E is given by Einstein’s famous equation:

E = mc2

where c is the velocity of light, namely 300,000 kilometers per second (186,000 miles per second). The amount of energy deriving from a mass loss is enormous; for example, 100 kilograms of mass fully converted into energy would supply all the energy needs of the United Kingdom (at the present rate of usage) for a year. Each kilogram of mass, fully converted, is equivalent to the energy available by burning 3 million tons of coal. In a typical nuclear re­action, however, only a tiny fraction of the mass is converted into energy, typ­ically -0. 1 %. The disintegration of an unstable nucleus, and the consequent release of nuclear energy, can be stimulated by exciting the nucleus by bom­barding it with radiation. This is at the heart of the fission reaction process, which we shall discuss further below. Nuclear energy can also be released, as we shall see, by the fusion of very light atoms into heavier ones.

6. П^^ші Entergy. The atoms of all substances are in constant motion. In a solid the atoms are held in an approximately fixed position with respect to one another. However, they all vibrate to an extent that increases with in­creasing temperature. The energy associated with this vibration is called thermal energy. In fluids (namely, liquids and gases), two or more atoms may be combined with each other chemically in the form of molecules. These molecules have vibrational energy, but in the fluid state they may also have translational energy arising from their motion in space and rota­tional energy arising from their rotation. All of these components of energy add up to the thermal energy of the fluid. It will be seen from this descrip­tion that thermal energy is of a special type. It is associated with atomic or molecular movements that are randomly directed. This makes it very much more difficult to convert thermal energy into other forms of energy, as we shall see below.

The intensity of atomic or molecular movement is a measure of the energy content of a piece of matter. A body that has a high intensity of atomic or mol­ecular movement will transfer energy to an adjacent body with a lower intensity of movement. This process of transfer of thermal energy is known as conduc­tion, and we define a quantity known as temperature as a measure of the abil­

ity of a body to transfer thermal energy to adjacent bodies by the conduction process. If the temperature of a body is higher than that of adjacent bodies, heat will be conducted from it; if it is lower, the reverse is true. We conveniently choose a scale of temperature in terms of certain transitions that occur in na­ture. Specifically, we define the melting point of ice as zero degrees centigrade (0°C) and the boiling point of water as 100 degrees centigrade. In energy con­version processes involving thermal energy, it is convenient to define an alter­native temperature scale, commonly referred to as the scale of absolute temperature. Here, the measure of temperature is the kelvin (K) rather than the degree centigrade. Zero kelvin corresponds to -273.17°C and is the condition in which all atomic and molecular motions have effectively ceased.

In a system that does not receive energy from or emit energy to the outside, the total amount of energy can be increased only by converting mass into en­ergy via nuclear processes. In the absence of these processes, the total amount of energy remains constant (this is the basis of the first law of thermodynamics). However, within the given system, the form of energy may change (e. g., chem­ical energy may be converted into thermal energy or thermal energy may be converted into mechanical energy). Before discussing these conversion processes, we shall digress briefly to discuss and explain the units by which en­ergy is measured, since these are vital in what follows in this book.

The prime examples of this type of circuit are those used in the Magnox, pres- surized-water, and CANDU reactors. The circuits for these reactors are illus­trated in Figures 2.4, 2.8, and 2.11). In normal reactor operation there is a multiplicity of loops, as illustrated in Figures 3.5 and 3.6, which show the posi­tions of the individual loops in the P^TC. and C. ANDU systems, respectively. Note that in the P^TC. the loops come together in the reactor core, whereas in the CANDU reactor they are always totally separate. This has important impli­cations for safety considerations with these reactors, as we shall see in Chapter

4. A typical modern large P’^TC has three or four loops, depending on the size, each loop handling typically 300 ^MW of electric power production (corre­sponding to generation in the reactor core of 900 ^MW of thermal energy for each loop). Smaller reactors have two loops, with the size of the steam genera­tors and other components within a loop kept approximately the same. Some

of the earlier P^WRs had a four-loop design, notwithstanding their smaller over­all size and much smaller output per loop (e. g., the Shippingport and Yankee Rowe reactors). The move toward standardization in the mid-1960s led to much larger reactors and much larger powers per loop.

The Enrico Fermi reactor was a sodium-cooled fast breeder demonstration re­actor, producing 200 MW(t) [61 MW(e)]. The plant was located near Lagoona Beach, Michigan, and started operation in 1963. After extended low-power op­eration, power raising took place during 1966. When this was being done, it was noted that the coolant temperatures above 2 of the 155 fuel assemblies (clusters of fuel rods) were higher than normal and the temperatures above an­other assembly were lower than normal.

The reactor was shut down, and the fuel assemblies were rearranged in the core to determine whether these abnormal temperatures were dependent on location in the core or were characteristic of the fuel assemblies themselves.

On October 5, 1966, the rise to the selected power level [67 MW(t)] for these tests on the rearranged fuel elements was begun. At about 3 p. m., with the re­actor at a power level of 20 MW(t), the reactor operator observed a control sig­nal, indicating that the rate of change of neutron population was erratic. The problem had been experienced before and was thought to be due to random electrical fluctuations in the control system. The reactor was placed on manual control, and when the instability disappeared, automatic control was again se­lected and the increase in power resumed.

At 3:05 p. m., with the reactor power at 27 MW(t), the erratic signal was again oh — selved. Shortly after that it was noted that the control rods were withdrawn farther than normal. A check of the core exit temperatures showed that the outlet tem­peratures from two subassemblies were abnormally high at 380 and 370°C (715 and 695°F), compared with a mean bulk outlet temperature of 315°C (600°F).

At 3:09 p. m., alarms occurred from the ventilation monitors in the upper building ventilation exhaust ducts. The building was automatically isolated—no one was inside at the time—and a radiation emergency was announced. The re­actor power increase was stopped at 31 ^W(t), and a power reduction was started. By 3:20 p. m., the power had decreased to 26 ^W(t) and the reactor was manually tripped and shut down indefinitely. Over the next year, many of the assemblies were removed and examined, and it was found that the bulk of the fuel in two of the fuel assemblies had melted. It was not until the end of the ex­amination period that the cause of the accident was discovered. The cause was relatively trivial. Below the core, six small Zircaloy plates had been installed to guide the flow of sodium into the upward direction. One of these Zircaloy plates had broken loose and blanked off the entry to a few subassemblies, causing almost total flow starvation.

The damage to the reactor was repaired with a specially designed remote handling tool, and the reactor reached full power output again in October 1970, 4 years after the accident.

Although the Enrico Fermi accident led to no injury or release of activity out­side the containment shell, 10,000 curies of fission products were released to the circulating sodium coolant. The accident focused attention on the potential problems of flow blockages caused by foreign bodies within the circulating sodium. In particular, any part of the reactor that may be susceptible to vibration damage, causing the release of foreign material, must be carefully evaluated. In the design of modern reactors, very thorough flow testing of the various com­ponents is carried out. It is noteworthy that the zirconium plates were added at a very late stage in the design and may not have had the same level of quality assurance as the other components in the Enrico Fermi reactor. Late “fix-ups" of this kind and of the kind that occurred at Hunterston must be avoided.

The damage to the fuel assemblies did not propagate to adjacent fuel assem­blies, and the evidence from this incident that the accident did not escalate was encouraging.

As we saw in Section 8.1, a large variety of low-level wastes also arises from the nuclear program. Waste consisting of miscellaneous rubbish (such as rubber gloves and tissues contaminated with traces of radioactive material) is typically contained in steel drums, compacted to reduce bulk and then placed in steel containers and disposed of in a shallow trench area covered with at least 1 m of soil. Measurements on such an engineered disposal facility have indicated that the radiological significance of the disposal is negligible.

Wastes with medium levels of radioactivity from reprocessing, power reactor operations, and decommissioning, as well as plutonium-contaminated wastes, are usually contained within a concrete or bitumen matrix within stainless steel drums. There is widespread agreement that geological disposal is the best solu­tion for the management of such wastes. A geological repository of the type il­lustrated in Figure 8.6 would be suitable. Sweden already has an operational repository under the seabed for medium — and low-activity wastes at Forsmark; Finland has a similar repository at it Olkiluoto power station site. These facilities are about 100 m below the seabed-ground level. In Britain, UK Nirex Ltd. was set up in 1985 with the responsibility of providing radioactive waste disposal fa­cilities. Currently, a deep underground site near Sellafield, northwest England, is being investigated for a geological repository. An underground rock laboratory is planned as the first stage prior to the construction of the facility, expected to be brought into operation around 2010.

Liquid wastes at low activities arise from all nuclear sites, particularly from reprocessing plants, and are discharged under the regulations laid down by the licensing authority. Obviously, great care must be taken to avoid any public

danopr frnm such disrhar. es

Gaseous wastes, typically noble gas isotopes, are also produced from reac­tors and reprocessing plants. These are normally discharged to the atmosphere under carefully controlled conditions.

A final point on disposal concerns the decommissioning of a nuclear plant. Decommissioning is done in stages; stage 1 is concerned with the removal of spent fuel—defuelinrr—from the reactor. This starts at shutdown and can take up to 3 years for a large gas-cooled reactor. The spent fuel that is discharged is then managed in the same way as “operational” spent fuel. This reduces the total amount of radioactivity at the reactor site to less than one-seventh that at shutdown. The second stage involves all the dismantling of all nonradioactive plant and buildings other than the reactor and its concrete biological shield. This stage follows on from stage 1 and takes 5 to 10 years. The reactor building itself is then sealed for a period of surveillance. Finally, stage 3 involves the complete dismantling of the reactor and returning the site to a “greenfield” sta­tus. This stage occurs about 100 years after shutdown and takes about 10 years to complete.

A variant on this strategy involves the construction of a high-integrity in­truder-proof containment around the reactor building—Safestore-—that can be left for periods of up to 100 years before the final dismantling of the reactor. This strategy allows the maximum time for the radioactivity in the reactor build­ing to decay, thus minimizing the hazard when actual dismantling takes place. Modern P’^TC stations are designed for the replacement of all components with the exception of the reactor pressure vessel, and are therefore relatively straightfoiward to decommission.

So far about 80 nuclear reactors have been shut down worldwide and sev­eral sites have been cleared completely—the world’s first civil P’^TC station, Shippingport, for example. In the United Kingdom, decommissioning has started at three of the older Magnox station sites, Berkeley, Hunterston, and Trawsfynydd. Handling and disposal of radioactive waste from decommission­ing follow similar routes to reprocessing and reactor operational wastes. De­commissioning represents only a small fraction (approximately 5% maximum) of nuclear generating costs.

REFERENCES

Cooper, J. R., and J. W. Rose (1977). Technical Data on Fuel, p. 53. Scottish Academic Press.

Ealing, C. J. (1994). "Experience and Application of the GEC Alsthom Modular Vault Dry Store.” Nuclear Engineer 35 (March-April): 48-54.

Janbury, K. 0994). “Transport, Storage and Final Disposal of Spent Fuel in the Federal Republic of Germany.” Nuclear Engineer 35 (May-June): 78-83.

OECD 0988). Environmental Impacts of Renewable Energy. Report by the Organization for Economic Cooperation and Development.

Passant, F. H. 0994). “Waste Management and Decommissioning.” Nuclear Energy 33 (4): 223-229.

Stevens-Guille, P. D., and F. E. Pave 0994). “Development and Prospects of Canadian Technology for Dry Storage of Used Nuclear Fuel.” Nuclear Engineer 35 (March-April): 64-71.

The low volumetric power density and low operating temperatures and pres­sures of the Magnox stations led to a search in the United Kingdom for an im­proved design. The resulting advanced gas-cooled reactor (AGR) is illustrated in Figure 2.5. In common with the Magnox reactor, the AGR uses carbon dioxide as a coolant, but the coolant pressure in the AGR is 40 bars (600 psia) and the coolant outlet temperature is 650°C. To achieve these higher temperature and pressure conditions, it was necessary to make a radical change in the design of the fuel. The fuel was changed to uranium oxide, mounted in the form of pel­lets inside thin-walled stainless steel tubes, which had small transverse ribs ma­chined on the outside (Figure 2.6). These tubes (sealed at each end) were grouped in bundles of 36 (see Figure 2.6). Since the high temperatures require the use of a stainless steel can, the can material is a significant absorber of neu­trons, unlike that in the Magnox reactor, and it is necessary to enrich the ura­nium in the fuel to about 2.3% 235U (about three times the natural 235U content). The AGR design benefited from the Magnox developments, particularly the de­sign of the gas circulation system. The steam generators were mounted inside the prestressed concrete vessel, as illustrated in Figure 2.5. Since the C02 reac­tor coolant is now at a high temperature, the steam generators can be designed to provide steam under conditions similar to those found in the most efficient fossil-fuel power plant, i. e., steam at 170 bars and 560°C. This gives the AGR a considerable advantage. Its steam cycle efficiencies are around 40%, the highest of any nuclear reactor operational at present.

Referring to Table 2.3, we see that the average volumetric power density of an AGR is around three times that of the highest-rated Magnox station. The av­erage fuel rating is also higher, by a factor of approximately 4. This leads to a more compact, capital-effective design. Nevertheless, a number of technical problems in the AGR design had to be solved. One was that the carbon dioxide coolant might react with the graphite moderator under the high temperatures and radiation fields in the reactor to produce carbon monoxide by the reaction:

C02 + C -> 2CO

which would corrode the graphite and reduce its strength. It was found that precise control of the carbon monoxide and water vapor content, together with the addition of methane in small concentrations, inhibited this reaction and minimized the rate of attack on the graphite. However, too high concentrations of methane and carbon monoxide could lead to carbon formation on the fuel elements, which would impair the heat transfer by reducing the turbulence

Double Sktared Grapeite Stove

• Improved graphite to withstand longer reactor dwell

• Modified design of graphite sleeve to improve strength

Brace

• Streamlined grids and braces to reduce pressure drop

Fuel Pins

• Strong cladding material to withstand longer reactor dwell

• Coating on pins to reduce oxidation

•

Large grained UO, fuel pellets for improved fission product retention

caused by the ribs. Fortunately, there is a range of methane and carbon monox­ide concentrations (called the coolant “window") in which the satisfactory op­eration is possible without excessive corrosion or deposition.

An alte^tive design is the so-called high-temperature gas-cooled reactor (HTGR). The use of helium rather than carbon dioxide overcomes the graphite oxidation problem. Helium is inert and consequently allows higher coolant tem­peratures. The uranium fuel is in the form of coated particles. A kernel of low-en­riched uranium carbide is coated with successive layers of pyrolytically deposited carbon and impervious silicon carbide (to retain the fission products). Two dis­tinct lines of reactor development have been pursued. One line is the so-called pebble-bed reactor, developed in Germany, whose core consists simply of a stack of graphite spheres in which the coated fuel particles are embedded. A second line of development, initiated in Europe but carried forward in the United States, is the prismatic core in which vertical replaceable graphite prisms containing graphite fuel rods (in which the coated particles are embedded) and coolant pas­sages make up the core. Typically, core power densities range between 5 and 10 ^W/m3 with helium coolant outlet temperatures up to 1000°C. A number of pro­totype HTGR plants have been built to demonstrate both the pebble-bed and the prismatic-core concepts, although no commercial power plant is currently in operation.

The boiling-water reactor, like the P^^, has multiple provisions for cooling the core in the event of an unplanned depressurization or loss of coolant within the reactor. A typical B^^ emergency core cooling system is illustrated in Figure 4.27 It is composed of four separate subsystems, namely the high-pressure corespray (HPCS) system, the automatic depressurization system (ADS), the low-pressure corespray (LPCS) system, and the low-pressure coolant injection (LPCI) system.

The HPCS pump takes water from the condensate storage tank and/or the pressure suppression pool as shown in Figure 4.27 The water in the system is piped into the vessel and feeds semicircular perforated rings that are designed to spray water regularly over the core and onto the fuel assemblies. This system operates over the full range of reactor pressure and is activated when the water level in the reactor drops below a preset level or the pressure in the contain­ment vessel reaches a high value.

If the HPCS cannot maintain the water level or if it fails to operate, the reac­tor pressure is reduced automatically by operation of the ADS, which dis­charges fluid from the vessel into the pressure suppression pool. The depressurization allows the LPCI and LPCS systems to come into operation, and these provide sufficient cooling. The LOCS pump takes its water from the sup­pression pool and discharges from a circular perforated pipe in the top of the reactor vessel above the core; it is actuated in much the same way as the HPCS. The LPCI system is used for residual heat removal on a long-term basis.

Hydrogen can be formed at various stages of a severe accident as a result of chemical accidents between steam and various metals. This hydrogen can burn or detonate, hazarding the containment systems.

The most important contributor to the hydrogen formation process is the ox­idation of the zirconium cladding of the fuel:

Zr + 2H20 = Zr 02 + 2H2

The reaction is exothermic adding to the decay heat. The extent of the chemi­cal reaction is determined by a number of factors, including the access of steam to unreacted metal and the geometiy of the core debris. Other materials that react include chromium and iron and even uranium dioxide.

Hydrogen may he formed at various phases of the accident:

1. When the initial heat-up occurs; perhaps 20-40% of the cladding may react in the first 10 or 20 minutes

2. When further water from the ECCS system or reactor coolant pumps contacts the hot debris

3. When molten debris jets or falls into the vessel lower head and vaporizes water to steam, which then has access to relatively undamaged fuel in the core above

4. When the pressure lower head fails and the molten debris attacks the con­crete of the vessel cavity and containment

In the case of a large loss-of-coolant accident (LOCA), the hydrogen may be released to the containment as it is formed. Conversely, where the primary circuit remains intact, the hydrogen release may occur at the time of lower head failure.

Hydrogen can react with the oxygen within the containment in one of two ways. The first way is by deflagration, or a diffusion flame in which the unburned gas is heated by conduction to a temperature sufficiently high for a chemical re­action. Whether a combustion reaction takes place depends on reaching the min­imum concentration of the hydrogen, i. e., 4-9% by volume. While diffusion flames and slow deflagrations add to the heating load and therefore the pressur­ization of the containment, they do not represent a serious threat to the integrity of most designs. Such a deflagration occurred during the TMl-2 accident.

In the second way, in a detonation, the unburned gas is heated by compres­sion in a shock wave. Initiation can come from a spark or other high-energy source. The consequences of a detonation depend on the concentration of the hydrogen (the higher the concentration the higher the combustion pressure) and the geometry of the containment internals.

One means of controlling hydrogen is to have an inert atmosphere (nitro­gen) in the containment. This is used on some reactor designs, particularly those of BWRs, but has operational disadvantages. Other techniques include catalytic recombiners (which react hydrogen and oxygen to form steam) and ig­niters (which deliberately ignite the hydrogen at the lower mixture concentra­tion) installed at various locations within the containment.