12.160. In calculating the expected radiation doses received from any airborne radioactive material that might escape from the reactor contain­ment structure, it is assumed that the material forms a plume similar to that from a smokestack but closer to the ground. The plume is carried forward by the wind while diffusion causes it to spread in two perpendicular directions, i. e., laterally (crosswind) and vertically. Observations made on small plumes indicate that diffusion would result in a Gaussian distribution (§9.165) of the radioactivity concentration about a centerline. According to the Gauss plume model, assuming that the containment vessel constitutes a continuous point source, the distribution of the radioactivity on the ground, from a specific radionuclide, would be represented by

(12.3)

where x(*, y) Bq/m3 is the ground-level concentration of radioactivity at a point x, y; Q Bq/s is the source strength of the given nuclide, и m/s is the wind speed, assumed to be uniform in the x direction, h m is the effective release height of the radioactivity, and у m is the lateral distance of the receptor from the plume centerline; ay and az are the standard deviations of the distribution in the plume in the у (lateral) and z (vertical) directions, respectively.

12.161. The values of ay and a2, which are functions of the distance x of the receptor from the source, have been determined from experimental studies; they are presented in graphical form in Figs. 12.12 and 12.13, for various atmospheric conditions, designated Pasquill categories A through F. Pasquill A refers to an extremely unstable atmosphere whereas Pasquill F is a moderately stable condition. Thus, if Q, w, and h are known, x(*, y) can be determined for various atmospheric conditions from equation

12.162.

The source strength Q for a particular radionuclide is deter­mined in the following manner. The saturation (or equilibrium) activity in a reactor which has been operating at a specified thermal power can be calculated along the lines indicated in §2.122. Values for the radioiodines obtained in this manner are given in Table 12.2. From the equilibrium activity of a particular radionuclide, the assumed release fraction, the di­mensions of the containment structure, and the maximum design leakage rate from the containment vessel, the source strength Q in Bq/s can be readily evaluated.

12.163. In making radiation dose calculations, у in equation (12.3) is

set equal to zero; the result

then applies to the radioactivity concentration on the ground along the plume centerline where the dose would be a maximum for a given value of x. This expression is assumed to hold for the first 8 hours after a presumed release of radioactive material, e. g., after an LOCA. For times greater than 8 hours, the wind is supposed to vary somewhat in direction so that the plume is spread uniformly over a 22.5° sector; the resulting equation is then

12.164. For radioactivity escaping from a containment structure, the release is fairly close to ground level; hence h in equations (12.4) and (12.5) is taken as zero and the exponential terms are then equal to unity. The atmospheric conditions for the first 24 hours are assumed to correspond

to Pasquill F with a wind speed of 2 m/s (about 4.5 mph); under these conservative conditions the calculated doses will be much larger than would be generally expected. More realistic atmospheric conditions are postulated for later times.

12.165. In the case of a release near ground level, allowance must be made of the turbulent wake resulting from the presence of the reactor building. The wake will cause additional dispersion of the radioactivity, thereby reducing the concentration on the ground. For the first 8 hours after a release, a dispersion factor, not exceeding 3, is introduced into the denominator of equation (12.4) with h = 0; the factor decreases with the (minimum) cross-sectional area of the building (at a given distance from the source) and with the distance (for a given area). After 8 hours, the effect of the building wake is ignored, i. e., the factor is unity.

12.166. Examination of equations (12.4) and (12.5) shows that, for a given (or zero) release height h, the quantity x(x)u/Q is a function only of the distance in the x direction from the release point and of the Pasquill criterion. Curves of x(x)u/Q versus x for various release heights and at­mospheric conditions have been published. These curves greatly facilitate the evaluation of x([24]) from a known source.

Whole-body dose

12.167. For a radioactive cloud (or plume) with dimensions that are small relative to the effective range of the gamma rays emitted from the cloud, the dose at a given point will include radiations received from various parts of the cloud. Calculations of the dose rate (or dose) are then very complex. If, however, the cloud is large and uniform, an equilibrium con­dition exists, and the rate of energy absorption per unit volume of air will be the same as the rate of energy release. Suppose that the concentration of a given radionuclide in a large uniform cloud is x Bq/m3,_i. e., x dis/s • m3, and the average gamma-ray energy per disintegration is Ey MeV, i. e., 1.60 x 10~13 Ey J/dis, then

Rate of energy release (or absorption) in air = 1.6 x 10~13

where p is the density of air in kg/m3. In soft-body tissue, the energy absorption would be 1.1 times that in air;* hence, upon taking p as about 1.3 kg/m3 and recalling that 1 rad represents the absorption of 10~2 J/kg,

it follows that

Absorbed dose rate in body tissue ~ 1.2 x 10-11 Ey rad/s. (12.6)

12.168. The value of x obtained from equation (12.4) or (12.5) is for a receptor at ground level. The gamma-ray dose received by an individual on the ground at the center of an “infinite” cloud would be roughly half that given by equation (12.6) owing to the presence of the ground which limits the source to a 2tt solid angle. The gamma-ray dose rate in body tissue (or the whole body) would then be

Whole-body dose rate ~ 0.6 x 10~n Ey rem/s,

where rems have been used in place of rads since they are essentially equivalent for gamma rays (§6.37). If the radionuclide concentration is assumed to remain constant over the exposure period of t s, the gamma — ray dose received in that time is

Dy ~ 0.6 x 10-11 XEy t rem.

In view of the various assumptions and approximations made, this expres­sion gives doses that are substantially larger than would realistically be expected.

Thyroid dose

12.169. The dose to the thyroid would arise mainly from breathing air containing radioiodines (see Table 12.2). If x, Bq/m3 is the average con­centration of radioiodine in the air, В m3/s is the breathing rate, and t s is the time during which the iodine-laden air is breathed, the initial amount C0(l) Bq of radioiodine taken into the lung is given by

Q)(0 = XiBt Bq.

For a conservative calculation, no allowance would be made for depletion of radioactivity from the plume as a result of deposition on the ground or for radioactive decay in transit from the reactor site to the point at which C0(z) is to be determined. In a realistic calculation, however, such factors would be taken into account.

12.170. Hence, in a conservative calculation, x,- is obtained from equa­tion (12.3), etc., with Qt Bq/s equal to the rate of escape of radioiodine from the containment vessel (cf. §12.159). For the first 8 hours after a postulated accident, В is assumed to be 3.47 x 10“4 m3/s, and from 8 to 24 hours it is 1.75 x 10"4 m3/s. Subsequently, until the plume has passed, the normal breathing rate of 2.32 x 10-4 m3/s is assumed. From these data, the value of C0(/) for a particular radioiodine can be determined. For iodine-131, this is related to the thyroid dose commitment by equation (6.8) in the form

Doo (in rem) = C0(i) (in Bq) x 4.1 x 10~5,

where rads are replaced by rems because they are essentially equivalent in this instance. Similar evaluations may be made for the other radioiodines or, more simply, the dose from iodine-131 may be multiplied by 1.9 to obtain the total iodine dose. The factor 1.9 takes into account the equilib­rium amounts (Table 12.2) and the dose equivalents per becquerel of the various isotopes.

9.56. For a reactor operating in the steady state, the spatial distribution of temperature in any component, e. g., a fuel rod, is given by equation (9.16), assuming the medium to be isotropic and the thermal conductivity independent of temperature. In a transient situation, however, such as when the reactor is being started up or shut down, the steady-state condition is not applicable. In deriving the steady-state equation (9.15) for one­dimensional heat flow, of which equation (9.16) is the general form, a heat (or energy) balance was obtained by equating the difference between the rates of heat flow into and out of a volume element to the rate of internal heat generation.

9.57.

For the transient situation, the energy balance (or conservation) must include the time rate of change of internal energy in the volume element. This is given by cpp{A dx){dt! d%), where cp is the specific heat of the material and p is its density ; A dx is the volume of the element under consideration (§9.42) and dtid% is the rate of increase of temperature t with time 0. Hence, equation (9.15) becomes

or

d2t _ Q(x) 1 dt

dx2 к a dO ’

where a, called the thermal diffusivity of the material, is defined by

The general form of equation (9.26) for the spatial distribution of tem­perature in a conducting medium with internal heat generation under tran­sient conditions is

which may be compared with the steady-state equation (9.16).

9.58. In the analysis of transient behavior, the appropriate form of equation (9.27) is generally solved by numerical methods using a computer; for this purpose, the differential equation must be expressed in finite — difference form. To illustrate the general principles used in these (and many other) calculations, the simple case will be considered of conduction in one dimension in a component (or components) which may be regarded as being made up of several adjacent volume elements (or cells).

9.59. The system can be treated as a series of grid points or nodes, with the characteristic properties, e. g., the temperature, of each cell being rep­resented by the value at its nodal point. Three such cells, designated і — 1, і, and i + l, respectively, with the corresponding nodal points, assumed to be at the cell centers, are shown in Fig. 9.9; the nodal temperatures are

th and ti+1. The dimension of each cell in the x direction is Ax, and the volume is A Ax, where A is the cross-sectional area of the cell.

9.60.

Suppose, for the present, that there is no internal heat generation. Conservation of energy then requires that the difference between the rates of heat flow into and out of the cell і be equal to the rate of increase of internal energy cpp(A Ax^dt^dQ), If qt-u is the rate of heat flow in the x direction from cell і — 1 into cell /, and qii+г is the rate of heat flow out

qu+1 = кA

assuming к to be independent of temperature. If these values are inserted into equation (9.28), it is found that

ti_l + ti+l — 2tj _ 1

(Ax)2 add

It can be readily shown that the left side is the linear-difference form of + cPt/dx2 so that equation (9.29) is the equivalent of equation (9.26) with Q(x) equal to zero.

9.61.

In order to express dtt/dQ in linear-difference form, the nodal approach is extended into a time coordinate 0 as in Fig. 9.10, where n — 1, n, and n + 1 indicate successive times. For a time step A0 from n

to n + 1, the temperature change Att in the cell і is tin+1 — ti n, so that

dti^ M = tj, n+i ~ Kn d% A0 A0

Upon insertion of this result into equation (9.29) for the time 0„ and rearranging, it is found that

It is thus seen that if the temperatures tn at the various nodal points are known at the time 0„, the temperature tn + 1 at the time 0„ + 1, i. e., after a specified time increment Д0, can be calculated directly from equation (9.31). When the values of t at all the nodal points at the time 0„ + 1 are known, the calculation can proceed to the next time step, i. e., to time 0n+2, and so on. This procedure is referred to as the explicit method for solving the differential equation (9.26) with Q(x) equal to zero. Internal heat gener­ation may be included, however, without affecting the general principles.

9.62. A simplification of equation (9.31) is possible if the time and distance increments Д0 and Ax are chosen so that the dimensionless quantity

aA0

(АхГ

It is apparent from equation (6.31) that, in these circumstances, the tem­perature at a node і after a time step is equal to the mean of the temper­atures at the two adjacent nodes before the time step, i. e., fl> + 1 =

9.63. In any event, it has been shown that for a numerical solution of equation (9.31) to be possible, аД0/(Д*)2 must be positive and less than

0. 5. This limitation on аД0/(Д*)2 places a restriction on the magnitude A0 of the time step, relative to (Дх)2/а, that can be used in the computations. Consequently, when the calculations must be made for an extended time period, a large number of time intervals may have to be used. Furthermore, it is found that if the time intervals are small, instabilities arise in applying the numerical techniques.

9.64. When, for one reason or another, longer time intervals are re­quired than are permitted by the explicit method, the implicit form of the difference equation is used. The spatial finite-difference temperature terms are expressed at the advanced time point n + 1 instead of at the point n as in the explicit procedure. That is to say, equation (9.30) is inserted into equation (9.29) in which the temperatures are now tt_l n + 1, ti+l n+1, and

ti n +1- It is then found that

9.65. In order to solve this equation for the temperatures at the time 0n+1 from the known value ti n at the preceding time 0n, it is necessary to write a set of simultaneous equations for all the space points at each time step. These equations can be represented by a tridiagonal matrix, i. e., a matrix in which only the main diagonal and the two adjacent diagonals are nonzero. Such a matrix is readily solved on a computer by the Gauss elimination procedure [4]. The implicit method requires a greater calcu — lational effort at each time step than does the explicit method, but larger time steps can be taken and stability problems are eased; as a result, the overall computer time is often less than for the explicit method.

9.66. In the foregoing discussion a simple situation has been treated to provide some insight into the general principles used in solving transient problems in heat transmission. In practice, boundary conditions, e. g., a convection boundary at the fuel cladding-coolant interface, must be in­cluded in the calculations. Allowance may also be required for internal heat generation, which is usually time dependent. In addition, in some cases it may be necessary to determine the spatial temperature distributions in more than one dimension. It should be noted, too, that a, which is a characteristic property of the material, may change from one nodal point to another. The calculations are thus often quite complex and are per­formed by means of appropriate computer codes.

9.182. Gas-cooled reactors such as the HTGR have the advantage of not having the boiling crisis design limit characteristic of water-cooled reactors. The core thermal design is based on classical methods describing conduction and convection transport to a gas with the aid of suitable com­puter codes. A design objective is to achieve a core power distribution so that the coolant exit temperature will be radially uniform. Power peaking factors developed by computer codes are used primarily as a design aid in modeling the core power distribution.

10.94. The capital costs (or charges) are those associated with the con­struction of the plant. For engineering economic estimates, as opposed to corporate accounting, it is usually adequate to consider charges on an annual basis during the plant lifetime. These charges are primarily the cost of the required investment needed to construct the plant, allowance for depreciation, and some taxes. A charge for decommissioning may be in­cluded in the depreciation allowance or carried as a separate contribution to the generating cost. We have followed the latter approach in Table 10.2.

10.95. Construction costs consist of direct and indirect charges. The direct costs are those for equipment, materials, and labor required to build the plant. An allowance for contingencies is often included with the direct costs. Indirect costs are essentially all other costs incurred in building and testing the plant, so that it can be turned over by the manufacturer to an electric utility ready for continuous operation at its rated capacity.

10.96. Direct construction costs include costs of such items as land, structure and site facilities, reactor plant equipment (exclusive of fuel), turbine plant (generating) equipment, electric (power transmission) plant equipment, miscellaneous equipment, and contingency allowance. (The latter is sometimes included with the indirect costs.) The direct costs also include the cost of materials and of labor; in some categories, e. g., struc­tures and site facilities and electric plant equipment, labor costs constitute more than half the total.

10.97. Indirect construction costs cover several broad areas. One is professional services for engineering, design, licensing activities (see Chap­ter 12), and supervision of construction. A second area includes temporary facilities, equipment, and services during construction, the plant owner’s general and administrative expenses, and reactor operator training and plant startup. A major indirect cost is the interest on funds expended during the design and construction of the plant. The interest charge on each type of expenditure is based on the amount spent, on the elapsed time between the expenditure and the conclusion of the construction period, and the interest rate. The construction period, i. e., from the award of initial con­tracts for reactor plant construction to the time when the plant is ready for commercial operation, may be as long as ten years or more. But the first few years are devoted primarily to design and licensing efforts, and expenditures are relatively small. Subsequently, after a construction permit is issued by the U. S. Nuclear Regulatory Commission, expenditures for equipment and labor rise markedly. Hence, most of the interest charges are incurred during the later stages of the construction period.

10.98. An escalation cost, which is applicable to both direct and indirect construction costs, arises from increases in cost due to inflation between the time the costs are first estimated and the time a firm purchasing com­mitment is made. The escalation allowance depends on the estimated cost of the item, the time when the item is purchased after the estimate is made, and an annual percentage to reflect the probable consequences of inflation. Since the inflation rate varies from year to year and is difficult to predict, there is a large uncertainty associated with the escalation allowance.

10.99. Now that we have determined the total construction cost, we need to develop an annual fixed-charge contribution to the generating cost that is based on it. After the plant is placed in full-power operation, div­idends on stocks and interest on bonds must be paid each year. Such stocks and bonds were issued to obtain the funds to build the plant. As mentioned in §10.87, other annual “fixed” charges include some taxes and an allow­ance for depreciation. Although depreciation of present plants is based on an estimated 40-year lifetime, it should be noted that advanced plants are designed for a 60-year lifetime (Chapter 15). All of these fixed annual charges may be estimated by applying a percentage rate to the total con­struction cost. Since this percentage rate depends on the interest rates prevailing during the years of construction and other parameters, any value cited here is likely to become dated quickly. However, an annual rate of the order of 14 percent for an investor-owned utility may be used for orientation purposes.

Sources of Radioactivity

11.90. For reactors cooled (and moderated) by light water, essentially all the radioactivity present in the effluent from a nuclear power plant originates in the reactor vessel from two sources: (1) the escape of fission products from the fuel into the coolant water and (2) activation by neutrons of corrosion and erosion products, e. g., iron, chromium, nickel, and cobalt, and other substances, e. g., boron (in boric acid), dissolved or suspended in the water. Gaseous radioactive isotopes of oxygen, nitrogen, and argon, in particular, are also formed by neutron activation of these elements in the water molecule and in dissolved air. Most of these isotopes have fairly short half-lives, and they will have largely decayed to stable nuclides before being discharged to the atmosphere. However, although these radioactive species are a negligible environmental hazard, they must not be overlooked in connection with the protection of plant workers.

11.91. Fuel rods are clad with zircaloy to prevent escape of fission products, but in a small fraction of the rods, typically below 0.1 percent, toward the end of their operating life small holes and cracks may develop in the cladding as a result of welding defects or localized corrosion. Con­sequently, some fission products will escape into the coolant. These consist mainly of the gaseous radioisotopes of krypton and xenon and of the readily volatile element iodine. In addition, solid fission products are extracted to a small extent by the high-temperature water. Hence, radioactive fission products can be found in both gaseous and liquid effluents from a reactor installation. Corrosion and erosion products from the reactor vessel, cool­ant pumps, steam generator, and piping are inevitably present in the water and these are activated by neutrons from the reactor core (§7.78).

11.92. Another important source of radioactivity in the water coolant is tritium, the radioactive isotope of hydrogen. About one fission in 104 of uranium-235 is a ternary fission in which tritium is one of the products; the other two products are heavier nuclides similar to ordinary (binary) fission products. Some tritium is also formed by the interaction of neutrons with boron in the control elements of BWRs. (The control rods of PWRs do not contain boron.) Thus, tritium can escape into the coolant through cladding defects in both fuel rods and control elements. Tritium, is also formed directly in the coolant by the capture of neutrons by deuterium, i. e., heavy hydrogen, nuclei that are normally present in water. In PWRs, the major sources of tritium are the reactions of neutrons with boron (in the boric acid) used as a chemical shim (§5.187) and with lithium (in lithium hydroxide) used to control the pH of the water. Regardless of its origin, most of the tritium in the reactor coolant is found in the form of НТО (§6.14); a small proportion occurs as HT gas but this is of little significance.

Transient with failure of the reactor protection system (TC)

12.102. Should there be a malfunction combined with failure of the reactor protection system, the reactor will continue to generate power. However, it is assumed that the main steam valves will isolate the reactor from the turbine generator and the recirculating pumps will trip. As a result of rising steam pressure, safety relief valves will open, releasing steam to the suppression pool, with the resulting core voiding reducing the reactor power. Subsequent events would depend on the ability of water makeup and coolant injection systems to keep the core covered. Insufficient suppression pool heat removal capacity could result in excessive pressur­ization of the containment and failure, while core uncovering would result in fuel failure and fission product release.

Transient with failure to remove residual core heat (TW)

12.103. In this type of sequence, the reactor would trip, but the suppres­sion pool would fail to remove the continuing production of residual heat. Containment failure could occur as a result of excessive pressure and the ECC systems, which use the suppression pool as a water source, becoming ineffective because the water is now saturated and cannot be pumped. Core uncovering is then likely.

8.37. Familiarity with information resources is an important decision tool for the nuclear reactor engineer. Traditionally, the published literature has been the first step in developing background to solve a given problem. Although an understanding of relevant principles is still essential, conven­ient access to a variety of data bases and other information can expedite problem solving. Our purpose here is to draw attention to some of these resources. A useful guide to nuclear engineering resources is given by Jedruch [8].

8.38. Most practicing engineers recognize the need to update their ex­pertise continually through routine scanning of professional society news magazines and journals devoted to their speciality. We assume that they have access to a technical library and know how to use it to find articles in the archival literature and books of interest. Computerized catalogs are available in most major libraries. Therefore, extensive searches can be carried out quickly. Our objective here is to remind the nuclear reactor engineer of several other important information sources.

Abstracts

8.39. Atomindex, published by the International Atomic Energy Agency, provides comprehensive abstract coverage of the entire field of nuclear engineering and science. Energy Research Abstracts published by the U. S. Department of Energy provides coverage of nuclear and other energy — related activities with particular attention to those is the United States. Research reports, conference proceedings, and various safety-related reg­ulations are covered. It is issued semimonthly and is indexed annually. Therefore, it represents a convenient first step for a literature search. Since it is arranged by subject matter, with detailed subcategories, it is easy to browse through contributions in areas of interest. Sources for the literature cited are given. Another resource is the Engineering Index, which covers all areas of engineering.

9.102. The critical heat flux, at which burnout is expected to occur, is an important design consideration in water-cooled (and moderated) re­actors. A knowledge of burnout conditions is also necessary for possible accidental situations, such as might arise from loss of coolant flow or excessive fuel temperatures due to a power excursion. Consequently, meth­ods for estimating the critical heat flux in a reactor are required. Much experimental work has been done to develop the many correlations reported in the literature. In addition, much effort has been devoted to the development of theoretical models which would support empirical correlations.

9.103. For PWRs, the so-called W-3 correlation for the DNB condition has been widely used.[7] This is a many-termed complex expression for the critical heat flux with pressure, mass velocity, steam quality, enthalpy, and system dimensions as parameters [11]. In practice, the equation is readily solved as part of a digital computer calculation, and some general conclu­sions are of interest. The DNB flux decreases as the quality increases and hence reaches a minimum at the coolant channel exit. An increase in coolant-mass velocity tends to increase the DNB flux. The pressure effect is complex but an increase in critical heat flux can occur under PWR conditions as the system pressure is increased. As part of the core design effort, the DNB condition is calculated at various axial locations to deter­mine the closest approach to the design flux (see Fig. 9.22). Since the DNB flux and heat flux both decrease toward the exit, the closest approach is normally in the upper half of the channel.

9.104. The conditions in a BWR differ from those in a pressurized — water system largely because of the higher quality of the fluid in the former case. The approach generally used for determining the critical heat flux is based on numerous experiments with electrically heated rod bundles with fluid flow resembling that in an actual reactor. The results are incorporated into a thermal-hydraulic computational model, together with correlations that take into account the steam quality and coolant flow rate [12].

10.29. The use of burnable absorber rods or lumped burnable poison (LBP) plays an important role in PWR core management. In Westinghouse cores, these may take the form of aluminum oxide-boron carbide rods placed in those positions in a fuel assembly normally left open for the insertion of rod cluster control elements (§5.185). This means that LBP rods may only be used in assemblies not assigned to rod cluster control positions.

10.30. Since the first core initially contains no fission-product poisons, it is necessary to use burnable absorber rods in addition to soluble boron to compensate for the excess reactivity. It is not possible to use additional soluble boron as an alternative since the need to maintain a negative mod­erator temperature coefficient limits the boron concentration in the cool­ant. However, since first core use is limited to new reactors, most LBP use is for tailoring the individual assembly reactivity to avoid local power peaking in designing reload core patterns (§10.46). However, it is empha­sized that neutrons are lost unproductively in LBP rods. Therefore, only sufficient rods should be used to control the power distribution so that it is within licensing peaking factor limits. These peaking factors are normally determined by loss-of-coolant accident criteria (§12.90).

10.31. During the burnup cycle, burnable poison rods are depleted, fission product poisons are formed, and the fissile atom concentration is reduced. Although this generally results in a loss of fuel assembly reactivity, attention must be given to the relative reaction rates so that the fixed absorber is not depleted so rapidly that an unacceptable power peak would appear in midcycle. Another concern with some absorbers is the residual poisoning effect remaining after the initial burnup cycle. This leads to a reactivity penalty when the fuel assembly is reinserted in subsequent cycles (see Fig. 13.3). Therefore, a general design objective is to minimize the use of lumped burnable poison rods.

11.10. The environmental effects associated with the generation of nu­clear power are the result of a number of processes and operations. First, we have the various steps, which start with mining the ore required to manufacture the fuel assemblies loaded into the reactor (§10.4 et seq.). Although there are low-level radioactive wastes associated with milling and fuel manufacturing operations, their treatment is relatively routine and has not attracted public concern.

11.11. Nuclear power plants during normal operation release various gaseous and liquid effluents that affect the environment. The extent of the impact is determined by the transport of radionuclides through so-called “pathways” from their point of release to where people would be affected.

Releases must be “as low as is reasonably achievable” as a licensing re­quirement (§6.63). We will treat this topic in the next section but will defer a discussion of plant waste treatment systems and the management of low — level wastes until later in this chapter. The impact of accidents on the environment is treated in Chapter 12.

11.12. The management of spent fuel and high-level radioactive wastes has attracted considerable controversy. This is the third category to be discussed. Here again, the pathway concept is useful since harm is done only if the radioisotopes of concern manage to migrate to people in amounts above permissible levels.