Chapter 21. Radiation Protection

Protection of biological entities from the hazard of radiation exposure is a fundamental requirement in the application of nuclear energy. Safety is provided by the use of one or more general methods that involve control of the source of radiation or its ability to affect living organisms. We will identify these methods and describe the role of calculations in the field of radiation protection. Thanks are due to Dr. James E. Watson, Jr. for his excellent suggestions on this chapter.

21.1. Protective Measures

Radiation and radioactive materials are the link between a device or process as a source and the living being to be protected from hazard. We can try to eliminate the source, remove the individual, or insert some barrier between the two. Several means are thus available to help ensure safety.

The first is to avoid the generation of radiation or isotopes that emit radiation. For example, the production of undesirable emitters from reactor operation can be minimized by the control of impurities in materials of construction and in the cooling agent. The second is to be sure that any radioactive substances are kept within containers or multiple barriers to prevent dispersal. Isotope sources and waste products are frequently sealed within one or more independent layers of metal or other impermeable substance, and nuclear reactors and chemical processing equipment are housed within leak-tight buildings. The third is to provide layers of shielding material between the source of radiation and the individual and to select favorable characteristics of geological media in which radioactive wastes are buried. The fourth is to restrict access to the region where the radiation level is hazardous and take advantage of the reduction of intensity with distance. The fifth is to dilute a radioactive substance with very large volumes of air or water on release to lower the concentration of harmful material. The sixth is to limit the time that a person remains within a radiation zone to reduce the dose received. We thus see that radioactive materials may be treated in several different ways: retention, isolation, and dispersal; whereas exposure to radiation can be avoided by methods involving distance, shielding, and time.

The analysis of radiation hazard and protection and the establishment of safe practices is part of the function of the science of radiological protection or health physics. Every user of radiation must follow accepted procedures, and health physicists should provide specialized technical advice and monitor the user's methods. In the planning of research involving radiation or in the design and operation of a process, calculations must be made that relate the radiation source to the biological entity by use of exposure limits provided by regulatory bodies. Included in the evaluation are necessary protective measures for known sources, or limits that must be imposed on the radiation source, the rate of release of radioactive substances, or the concentration of radioisotopes in air, water, and other materials.

The detailed calculations of radiological protection are very involved for several reasons. A great variety of situations should be considered, including reactor operations and uses of isotopes. Many scientific and engineering disciplines are needed—physics, chemistry, biology, geology, meteorology, and several engineering fields. Increased use of computers favors the development of more sophisticated calculation methods while providing increasing convenience. The collection of new experimental data on the interaction of radiation and matter and the relationship of dose and effect results in evolving recommendations and regulations. Finally, the enhanced awareness of radiation and concern for safety on the part of the public have prompted increased conservatism, which entails refinement in methods and a requirement for fuller justification of methods and results.

In the operation of nuclear power plants and uses of radioisotopes, adherence to government regulations is mandatory to maintain a license. The principal document of the United States Nuclear Regulatory Commission (NRC) is Code of Federal Regulations: 10 Energy (see References). Part 20 “Standards for Protection Against Radiation,” has an abbreviated designation 10CFR20.

The establishment of regulations is a slow process, starting with the study of research information by advisory bodies such as International Committee on Radiological Protection (ICRP) and National Council on Radiation Protection and Measurements (NCRP), recommendations for dose limits and protection policies, review by the regulatory body with input from the public, institutions, and industry, with final issuance of mandatory requirements, along with guidance documents. As a consequence, the limits and method for different situations may be inconsistent but fundamentally safe. A case in point is the older use of a “critical organ” and maximum permissible concentrations of radionuclides and the newer use of “committed effective dose equivalent” referring to the summation of all effects on the body. The old and the new are contrasted in the NRC's discussion of regulation 10CFR20 in the Federal Register of January 9, 1986. Some of the earlier regulations are still applicable. We will present examples of both methods for two reasons: (a) to help the reader make use of all pertinent literature of radiological protection, and (b) to illustrate the trend toward greater precision and realism in radiation protection.

We now discuss the relationship of dose to flux, the effect of distance and shielding materials, internal exposure, environmental assessment, and dose limits for workers and the public.

21.2. Calculation of Dose

Some simple idealized situations will help the reader understand concepts without becoming involved in intricate calculations. The estimation of radiation dose or dose rate is central to radiation protection. The dose is an energy absorbed per unit mass as discussed in Section 16.2. It depends on the type, energy, and intensity of the radiation, as well as on the physical features of the target. Let us imagine a situation in which the radiation field consists of a stream of gamma rays of a single energy. The beam of photons might be coming from a piece of radioactive equipment in a nuclear plant. The stream passes through a substance such as tissue with negligible attenuation. We use the principles of Chapter 4 to calculate the energy deposition. Flux and current are the same for this beam (i.e., j and φ are both equal to nυ). With a flux φ cm⁻²−s⁻¹, and cross section Σ cm⁻¹, the reaction rate is φΣ cm⁻³−s⁻¹. If the gamma ray energy is E joules, then the energy deposition rate per unit volume is φΣE J cm⁻³−s⁻¹. If the target density is ρ g-cm⁻³, the dose in joules per gram with exposure for a time t seconds is

This relationship can be used to calculate a dose for given conditions or to find limits on flux or on time.

For example, let us find the gamma ray flux that yields an external dose of 0.1 rem in 1 y with continuous exposure. This is the dose limit to members of the public according to 10CFR20 (Section 16.2). Suppose that the gamma rays have an energy of 1 MeV and that the cross section for energy absorption with soft tissue of density 1.0 g/cm³ is 0.0308 cm⁻¹. With a quality factor of 1 for this radiation, the numerical values of the dose and the dose equivalent are the same, so

Also E = 1 MeV = 1.60 × 10⁻¹³ J. Solving for the flux, or

This value of the gamma ray flux may be scaled up or down if another dose limit is specified. The fluxes of various particles corresponding to 0.1 rem/y are shown in Table 21.1.

Table 21.1. Radiation Fluxes (0.1 rem/y)

Radiation Type	Flux (cm−2 − s−1)
X- or gamma rays	6.4
Beta particles	0.10
Thermal neutrons	3.1
Fast neutrons	0.085
Alpha particles	10−5

Another situation is the exposure of a person to air containing a radioactive contaminant, for example the noble gas krypton-85, half-life 10.73 y, an emitter of beta particles of average energy 0.251 MeV. Let us derive and apply a formula for the case of continuous exposure during working hours. We wish to relate dose H in rems to activity A in μCi, with an exposure time of t seconds. A rough estimate comes from a simple assumption—that the person is immersed in a large radioactive cloud, and that the energy absorption in air, E^a, is the same as in the human body and the same as that released by decay of the radionuclide, E^r. Write expressions for each of these,

Equate these and solve for the dose, but reducing the figure by a factor of 2 if the person is on the ground and the cloud occupies only half of space. The result is

Assume continuous exposure for 40 h/w, 50 w/y, 3600 s/h, so that t = 7.2 × 10⁶ s. Insert E = 0.251 MeV and H = 5 rems, the annual dose limit for plant workers. Solve for the activity This agrees fairly well with the figure of 1 × 10⁻⁵ listed in the 1993 edition of the old NRC 10CFR20. We will see in Section 21.7 that the latest method yields a larger dose limit.

21.3. Effects of Distance and Shielding

For protection, advantage can be taken of the fact that radiation intensities decrease with distance from the source, varying as the inverse square of the distance. Let us illustrate by an idealized case of a small source, regarded as a mathematical point, emitting S particles per second, the source “strength.” As in Figure 21.1, let the rate of flow through each unit of area of a sphere of radius R about the point be labeled φ (cm⁻² − s⁻¹). The flow through the whole sphere surface of area 4π R² is then φ 4π R², and if there is no intervening material, it can be equated to the source strength S. Then This relation expresses the inverse square spreading effect. If we have a surface covered with radioactive material or an object that emits radiation throughout its volume, the flux at a point of measurement can be found by addition of elementary contributions.

Figure 21.1. Inverse square spreading of radiation.

Let us consider the neutron radiation at a large distance from an unreflected and unshielded reactor operating at a power level of 1 MW. Because 1 W gives 3.3 × 10¹⁰ fissions per second (Section 6.4) and the number of neutrons per fission is 2.42 (Section 6.3), the reactor produces 8.0 × 10¹⁶ neutrons per second. Suppose that 20% of these escape the core as fast neutrons, so that S is 1.6 × 10¹⁶ s⁻¹. Apply the inverse square relation, neglecting attenuation in air, an assumption that would only be correct if the reactor were in a spacecraft. Let us find the closest distance of approach to the reactor surface to keep the dose below 100 mrems/y as in Table 21.1. The limiting fast flux is 0.085 cm⁻²−s⁻¹. Solving the inverse-square formula, we obtain

This is a surprisingly large distance, approximately 760 miles. If the same reactor were on the Earth, neutron attenuation in air would reduce this figure greatly, but the necessity for shielding by solid or liquid materials is clearly revealed by this calculation.

As another example, let us find how much radiation is received at a distance of 1 mile from a nuclear power plant, if the dose rate at the plant boundary, -mile radius, is 5 mrems/y. Neglecting attenuation in air, the inverse-square reduction factor is giving 0.31 mrems/y. Attenuation would reduce the dose to a negligible value.

The evaluation of necessary protective shielding from radiation makes use of the basic concepts and facts of radiation interaction with matter described in Chapters 4 and 5. Let us consider the particles with which we must deal. Because charged particles—electrons, alpha particles, protons, etc.—have a very short range in matter, attention needs to be given only to the penetrating radiation—gamma rays (or X-rays) and neutrons. The attenuation factor with distance of penetration for photons and neutrons may be expressed in exponential form exp(−Σr), where r is the distance from source to observer and Σ is an appropriate macroscopic cross section. In shielding analysis; this is called the linear attenuation coefficient, μ, with units cm⁻¹. Now Σ or μ depends on the number of target atoms, and through the microscopic cross section Σ also depends on the type of radiation, its energy, and the chemical and nuclear properties of the target.

For fast neutron shielding, a light element is preferred because of the large neutron energy loss per collision. Thus hydrogenous materials such as water, concrete, or earth are effective shields. The objective is to slow neutrons within a small distance from their origin and to allow them to be absorbed at thermal energy. Thermal neutrons are readily captured by many materials, but boron is preferred, because accompanying gamma rays are very weak.

Let us compute the effect of a water shield on the fast neutrons from the example reactor used earlier. The macroscopic cross section appearing in the exponential formula exp(−Σr) is now called a “removal cross section,” because many fast neutrons are removed from the high-energy region by one collision with hydrogen and eventually are absorbed as thermal neutrons. Its value for fission neutrons in water is approximately 0.10 cm⁻¹. A shield of thickness 2.5 m = 250 cm would provide an attenuation factor of exp(−25) = 10^−10.9 = 1.39 × 10⁻¹¹. The inverse-square reduction with distance is

The combined reduction factor is 1.77 × 10⁻¹⁷; and with a source of 1.6 × 10¹⁶ neutrons/s, the flux is down to 0.28 neutrons/cm² − s, which is somewhat higher than the safe level of 0.085 as in Table 21.1. The addition of a few centimeters of water shield would provide adequate protection, for steady reactor operation at least. Computer Exercise 21.B describes a program NEUTSHLD that finds fast flux from a fission source as a point or a plane.

For gamma ray shielding, in which the main interaction takes place with atomic electrons, a substance of high atomic number is desired. Compton scattering varies as Z, pair production as Z², and the photoelectric effect as Z⁵. Elements such as iron and lead are particularly useful for gamma shielding. The amount of attenuation depends on the material of the shield, its thickness, and the photon energy. The literature gives values of the mass attenuation coefficient μ/ρ, which is the ratio of the linear attenuation coefficient μ (macroscopic cross section Σ) and the material density ρ, thus it has units cm²/g. Typical values for a few elements at different energies are shown in Table 21.2. For 1 MeV gammas in iron, for example, density 7.86 g/cm³, we calculate Σ = (0.0596) (7.86) = 0.468 cm⁻¹. In contrast, for water H²O, molecular weight 2(1.008) + 16.00 = 18.016, the average value of μ with numbers from Table 21.2 with weight fractions is

Table 21.2. Mass Attenuation Coefficients (cm²/g). (NUREG/CR-5740, 1991)

Energy (MeV)	H	O	A1	Fe	Pb	U
0.01	0.385	5.76	2.58	169.6	125.7	173.7
0.1	0.294	0.151	0.161	0.342	5.35	1.72
1	0.126	0.0636	0.0613	0.0596	0.0684	0.0757
2	0.0876	0.0445	0.0432	0.0425	0.0454	0.0479
10	0.0324	0.0208	0.0231	0.0299	0.0496	0.0519

This is also the value of Σ because ρ = 1. Thus to achieve the same reduction in gamma flux in iron as in water, the thickness only need be 15% as much.

As an example of gamma shielding calculations, let us find the flux of 1-MeV gamma rays that have made no collision in arriving from a point source. This uncollided flux is a product of a source strength S, an exponential attenuation factor exp(−Σr), and an inverse square spreading factor 1/(4πr²), i.e.,

For example, find the uncollided flux at 10 cm from a 1 millicurie source (S = 3.7 × 10⁷/s). We readily calculate Σ for lead, μ/ρ = 0.0684 cm²/g, density 11.3 g/cm³, to be 0.773 cm⁻¹, and Σr = 7.73. Inserting numbers,

This is not the complete flux that strikes a receptor at the point, because those scattered by the Compton effect can return to the stream and contribute as sketched in Figure 21.2. To account for this “buildup” of radiation a multiplying buildup factor B depending on Σr is introduced. Figure 21.3 shows B for 1 MeV gammas in the most common shielding materials—lead, iron, and water. The total flux is then which shows that the buildup factor is the ratio of the actual flux to the uncollided flux. It remains to find B from the graph or tables, as 3.04, so that the flux is

Figure 21.2. Buildup effect.

Figure 21.3. Buildup factors, 1 MeV gammas. B for concrete and aluminum are about the same.

This calculation was rather straightforward, but it is more difficult if the flux is known and one wants to find the distance. Note that r appears in three places in the flux formula, so trial-and-error methods are needed. This tedious process is greatly assisted by use of the computer program EXPOSO, see Computer Exercise 21.A. To bring the exposure down to 5 mrems/y, the value of r is approximately 15 cm.

Although calculations are performed in the design of equipment or experiments involving radiation, protection is ultimately assured by the measurement of radiation. Portable detectors used as “survey meters” are available commercially. They use the various detector principles described in Chapter 10, with the Geiger-Mueller counter having the greatest general usefulness. Special detectors are installed to monitor general radiation levels or the amount of radioactivity in effluents.

The possibility of accidental exposure to radiation always exists in a laboratory or plant despite all precautions. To have information immediately, personnel wear dosimeters, which are pen-size self-reading ionization chambers that detect and measure dose. For a more permanent record, film badges are worn. These consist of several photographic films of different sensitivity, with shields to select radiation types. They are developed periodically, and if significant exposure is noted, individuals are relieved of future work in areas with potential radiation hazards for a suitable length of time.

Newer devices include the thermoluminescent detector reader (TLD), discussed in Section 10.3, which when heated releases photons in proportion to dose received. Solid-state meters have liquid crystal display (LCD) that involves transparent polarizing sheets with electric potential between them. Audible warning signals are available on some instruments.

Operation, maintenance, and repair of nuclear equipment involve some possible exposure to radiation. Even though it is assumed that any radiation is undesirable, it is necessary on practical grounds to allow a certain amount of exposure. It would be prohibitively expensive to reduce the level to zero. A basis for what action to take is the philosophy expressed in the phrase, “as low as is reasonably achievable,” with the acronym ALARA. Planning, design, and operation are done with the ALARA principle in mind. For example, a repair job on contaminated equipment is planned after making careful surveys of radiation levels. The repair is to be carried out by a small crew of well-trained people who will do the work quickly and with minimum contact with the radiation sources. Temporary shielding, special clothing, and respirators are used as needed to minimize doses. Factors considered are: (a) the maximum exposure both to individuals and to the group of workers as a whole, (b) other nonradiological risks, (c) the state of technology, and (d) the economic importance of the operation being performed. If the expected total dosage to the group is more than a fraction of the allowed quarterly dose, a formal ALARA evaluation is made, accounting for both the dollar costs and the dose costs. For a complete discussion by the NRC of the regulatory implications of ALARA, see References.

21.4. Internal Exposure

We now turn to the exposure of internal parts of an organism as a result of having taken in radioactive substances. Special attention will be given to the human body, but similar methods will apply to other animals and even to plants. Radioactive materials can enter the body by drinking, breathing, or eating, and to a certain extent can be absorbed through pores or wounds. The resulting dosage depends on many factors: (a) the amount that enters, which in turn depends on the rate of intake and elapsed time; (b) the chemical nature of the substance, which affects affinity with molecules of particular types of body tissue and which determines the rate of elimination (the term biological half-life is used in this connection, being the time for half of an initial amount to be removed); (c) the particle size, which relates to progress of the material through the body; (d) the radioactive half-life, the energy, and kind of radiation, which determine the activity and energy deposition rate, and the length of time the radiation exposure persists; and (e) the radiosensitivity of the tissue, with the gastrointestinal tract, reproductive organs, and bone marrow as the most important.

In the older regulatory framework, limiting concentrations of radionuclides in air or water are calculated with the concept of “critical organ,” the one receiving the greatest effective dose from a certain ingested radionuclide. The organ selected thus dominates the hazard to the body, and effects on other organs are neglected. We apply the method to calculate the maximum permissible concentration (MPC) in units μCi/cm³ of iodine-131 in water consumed by plant workers. I-131 has a half-life of 8.0 d and releases 0.23 MeV of beta-gamma energy per decay. The thyroid gland, of mass 20 g, will be taken as the critical organ because of the affinity of the thyroid for iodine. According to ICRP 2 (see References), the allowed annual dose is 30 rad. We first find the activity A that will yield that dose. The method of Section 21.2 is applied again. The energy absorbed is

The energy released is

Equate and solve for A = 0.139 μCi.

Now we find the rates of supply and elimination of I-131 to the organ, assumed to be in balance in steady state. With the formula of Section 16.1, with biological half-life of 138 days, the effective half-life t^E is 7.56 days and the decay constant λ^E is 0.0917 d⁻¹. Thus the elimination rate is proportional to λ^EA = (0.0917)(0.139) = 0.0127. The consumption rate of water for the standard man is 2200 cm³ per day, but it is assumed that workers drink 1.5 times the average during their 8-h day, and they work only 50 wk at 40 h/wk. The rate of intake of contaminated water is thus 755 cm³/d, and if 30% of the iodine goes to the thyroid, the supply rate of I-131 is (755)(0.3)(MPC). Equate rates and solve for MPC = 5.65 × 10⁻⁵ μCi/cm³, which rounds off to 6 × 10⁻⁵ μCi/cm³, the figure appearing in the older (1993) version of 10CFR20.

When there is more than one radioisotope present, the allowed concentrations must be limited. The criterion used is where i is an index of the isotope. This equation says the sum of quotients of actual concentrations and maximum permissible concentrations must be no greater than 1.

21.5. The Radon Problem

The hazard of breathing air in a poorly ventilated uranium mine has long been recognized. The death rate of miners has historically been higher than that for the general population. The suspected source is the radiation from radioactive isotopes in the decay chain of uranium-238, which by emission of a series of α particles eventually becomes lead-206. The data are clouded by the fact that uranium miners tend to be heavy smokers.

Well down the chain is radium-226, half-life 1599 y. It decays into radon-222, half-life 3.82 d. Although radon-222 is an alpha emitter, its shorter-lived daughters provide most of the dosage. Radon as a noble gas along with its suspended particulate decay products is breathed in with air. Some radioactive particles deposit on the lung surfaces. Decay of the radon and its daughters releases ionizing radiation.

The problem of radon near piles of residue from uranium mining, the mill tailings, has been known, and rules adopted about earth covers to inhibit radon release and about use of the tailings for fill or construction. More recently it has been discovered that a large number of United States homes have higher than normal concentrations of radon. Such excessive levels are due to the particular type of rock on which houses are built. Many homes have a concentration of 20 picocuries per liter, in contrast with the average of approximately 1.5 pCi/L and in excess of the EPA limit of 4 pCi/L. In recent years, EPA has given the subject a great deal of attention.

Application of dose-effect relationships yields estimates of a large number of cancer deaths from the radon effect, as high as 20,000 per year in the United States. Such numbers depend on the validity of the linear relationship of dose and effect discussed in Section 16.3. If there were a threshold or if there were a hormesis effect, the hazard would be very much smaller and mitigation costs greatly reduced. See References for the history of radon in mines, spas, and homes.

It was originally believed that the radon concentrations in houses were high because of conservation measures that reduced ventilation. Investigations revealed that the radon comes out of the ground and is brought into the home by drafts, similar to chimney action. Temperature differences between the air in the house and in the ground beneath cause pressure differences that cause the flow. One might think that covering the earth under a house with plastic would solve the problem, but even slight leaks let the radon through. In areas known to have significant radon levels, it is considered wise for homeowners to obtain radon test kits, which are rather inexpensive. If levels well above 4 pCi/L are found, action is recommended. The best solution is to ventilate a crawl space or to provide a basement with a small blower that raises the pressure and prevents radon from entering.

The dimensions of the problem are yet not fully appreciated nationally; continued study is required to determine the proper course of action at the national level.

21.6. Environmental Radiological Assessment^[†]

The NRC requires that the ALARA principle, discussed in Section 21.3, be applied to the releases of radioactive materials from a nuclear power plant. A deliberate effort is to be made to stay below the specified limits. These refer to any person in the unrestricted area outside the plant. According to 10CFR50, Appendix I, the annual dose resulting from a liquid effluent must be less than 3 millirems to the individual's total body or 10 millirems to any organ. The dose from air release must be less than 10 millirems from gamma rays and 20 millirems from beta particles. To comply with ALARA, it is necessary for the plant to correlate a release of contaminated water or air to the maximum effect on the most sensitive person. An acceptable method to calculate releases and doses is found in NRC's Regulatory Guide 1.109, October 1977 (see References). This “Reg. Guide” discusses the factors to be considered, gives useful formulas, and provides basic data. Older health physics methods are used, but because the dose limit sought is very small, the results are conservative. Among the important factors are:

1. The amounts of each radioisotope in the effluent, with special attention to cesium-137, carbon-14, tritium, iodine, and noble gases.
2. The mode of transfer of material. The medium by which radioactivity is received may be drinking water, aquatic food, shoreline deposits, or irrigated food. For the latter, pathways include meat and milk. If the medium is air, human beings may be immersed in a contaminated cloud or breathe the air, or material may be deposited on vegetables.
3. The distance between the source of radioactivity and person affected and how much dilution by spreading takes place.
4. The time of transport, to account for decay during flow through air or by streams, or in the case of foodstuffs, during harvesting, processing, and shipment.
5. The age group at risk: infant (0 to 1 y), child (1 to 11 y), teenager (11 to 17 y), and adult (17 and older). Sensitivities to radiation vary considerably with age.
6. The dose factor, which relates dose in millirems to the activity in picocuries. These numbers are tabulated according to isotope, age group, inhalation or ingestion, and organ (bone, liver, total body, thyroid, kidney, lung, and GI tract).

^† Appreciation is extended to Mary Birch for helpful discussions.

As an example, let us make an approximate calculation of the dose resulting from a continuous release of radioactive water from a nuclear power station into a nearby river. Assume that each day there is a release of 1000 gallons of water contaminated with a single radioisotope cesium-137, half-life 30.2 y. Also assume an activity in the water of 10⁵ pCi/L. The activity in the discharge of (1000 gal/d)(1440 min/d) = 0.694 gal/min is diluted by a stream flow of 2 × 10⁴ gal/min, down to (10⁵)(0.694)/(2 × 10⁴) = 3.47 pCi/L. The potential radiation hazard to the population downstream is by two types of ingestion: drinking the water or eating fish that live in the water. The age groups at risk are infants (I), children (C), teenagers (T), and adults (A). Consumption data are as shown in Table 21.3.

Table 21.3. Consumption by Age Group (Table E-5, Reg. Guide 1.109)

	I	C	T	A
Water (liters/y)	330	510	510	730
Fish (kg/y)	0	6.9	16	21

The row in the table that refers to fish must be multiplied by a bioaccumulation factor of 2,000 (its units are pCi/liter per pCi/kg). Consider the dose to an adult. To the consumption rate of water of 730 liters/y must be added the effect of eating fish, (2,000) (21) = 42,000, giving a total of 4.27 × 10⁴ liter/y. Now apply a dose conversion in mrems per pCi for cesium-137 as in Table 21.4. Each number should be multiplied by 10⁻⁵. The adult total body dose conversion factor is 7.14 × 10⁻⁵ mrems/pCi. Thus, the yearly dose is Because this is well above the limit of 3 mrems, a reduction in rate of release will be required.

Table 21.4. Ingestion Dose Conversion Factors in units of 10⁻⁵ (Table E-12, Reg. Guide 1.109)

Group	Bone	Liver	Total body	Kidney	Lungs	GI tract
I	52.2	61.1	4.33	16.4	6.64	0.191
C	32.7	31.3	4.62	10.2	3.67	0.196
T	11.2	14.9	5.19	5.07	1.97	0.212
A	7.97	10.9	7.14	3.70	1.23	0.211

The general environmental effect of supporting parts of the nuclear fuel cycle must be described in an application for a construction permit for a power reactor. Data acceptable to the NRC for that purpose appear in the Code of Federal Regulations, Part 51.51, as “Table of Uranium Fuel Cycle Environmental Data.”

21.7. Newer Radiation Standards

A major revision of regulations on radiation exposure was proposed by the NRC in 1986, published as a Final Rule in 1991, and required for use from January 1, 1994. The newer version of the rule 10CFR20,^[†] intended to provide greater protection for both workers and the public, was based on recommendations of the International Committee on Radiological Protection (ICRP).

^† Federal Register, Vol. 56, No. 98, Tuesday, May 21, 1991, p. 23360 ff. The Introduction contains useful reading on the history of dose regulations in the United States.

The improved regulations are more realistic in terms of hazards and bring to bear accumulated knowledge about radiation risk. The complicated task of deducing doses is accomplished by computer methods. Whereas the traditional limits on dosage are based on the critical organ, the new 10CFR20 considers the dosage to the whole body from whatever sources of radiation are affecting organs and tissues. Radiations from external and internal sources are summed to obtain the total dose. Also, long-term effects of radionuclides fixed in the body are added to any short-term irradiation effects. The bases for the limits selected are the risk of cancer in the case of most organs and tissues and the risk of hereditary diseases in offspring in the case of the gonads.

A new concept called “committed effective dose equivalent” is introduced. Recall from Section 16.2 that dose equivalent is the product of absorbed dose and the quality factor. The word “committed” implies taking account of future exposure after ingestion of radioactive material. The time span is taken to be a typical working life of 50 y (e.g., between ages 20 and 70). Suppose that a certain radionuclide is deposited in an organ of the human body. Over time thereafter the nuclide decays and is eliminated but provides a dose to that organ. The total dose, labeled H⁵⁰, is called a committed dose equivalent. It is assumed that the dose is experienced within the year the nuclide is deposited, which will be more nearly true the shorter the effective life in the body.

To calculate H⁵⁰, suppose that N⁰ atoms are deposited in a gram of an organ or tissue. The number left after a time t is where t^e is the effective half-life, as discussed in Section 16.1. The number that have been lost is N^L = N⁰ − N, and the fraction of these that decay is t^e/t^H, as shown in Exercise 16.6. Thus the number that decay is

As each nucleus decays, it delivers energy E, and thus the committed dose equivalent is

Let us apply these relations to some radionuclides. The half-life of tritium of 12.3 y is a fairly large fraction of 50 y but the biological half-life is only t^b = 10 d, so t^e is also approximately 10 d. The fraction that decays within the organ is 10/(4.5 × 10³) and the fraction lost is almost exactly 1. In contrast, for plutonium-239, t^H = 2.4 × 10⁴ y, t^b = 100 y for bone, and t^e = 99.6 y. The fraction left after 50 y is (1/2)^50/100 = 0.707, whereas the fraction lost is 0.293. Of these, decay accounts for only 99.6/(2.4 × 10⁴) = 0.0042.

Finally, the word “effective” takes account of the relative risk associated with different organs and tissues by forming a weighted sum by use of weighting factors w^T as listed in Table 21.5. If (H⁵⁰)^T represents the committed dose to organ or tissue T, the effective dose is a sum over T,

Table 21.5. Organ and Tissue Radiation Weighting Factors (10CFR20)

Organ or Tissue	Weighting Factor
Gonads	0.25
Breast	0.15
Red bone marrow	0.12
Lung	0.12
Thyroid	0.03
Bone surfaces	0.03
Remainder[†]	0.30
Whole body	1.00

^† 0.06 each for five organs.

If only one organ were important, as in the case of iodine-131 in the thyroid, the effective dose to the whole body would only be 3% of what it would be if the same dose were delivered throughout the body.

From the factors in Table 21.5 and from the knowledge of chemical properties, half-life, radiations, and organ and tissue data, the NRC has deduced the limits on concentration of specific radionuclides. Dose restrictions are for an annual limit of intake (ALI) by inhalation or ingestion of 5 rems/y (or a 50-year dose of 50 rems) for a plant worker. The derived air concentration (DAC) would give one ALI in a working year through breathing contaminated air. Extensive tables of ALI and DAC for hundreds of radioisotopes are provided in the new 10CFR20. They allow the calculation of exposure to mixtures of isotopes.

The two quantities are related by where the numerical factor is a product of four things: 50 wk/y; 40 h/wk; 60 min/h; and 2 × 10⁴ ml (air breathed per minute).

A distinction is made between two types of dose: The first is “stochastic,” which is the same as “probabilistic,” defined as dosages related to the chance of cancer or hereditary effect, with the number of health effects proportional to the dose. The worker dose limit for stochastic effects is 5 rems/y. The second is “nonstochastic” or “deterministic,” which are doses to tissues for which there is a threshold dose for an effect, so that a definite limit can be set on an annual dose (e.g., 50 rems). The skin and the eye lens are examples.

We can revisit the situation of a cloud of radioactive krypton-85 as in Section 21.2. Detailed calculation on all organs lead to the conclusion that only the skin will be significantly affected and thus the nonstochastic limit applies. The ALI and DAC values are correspondingly higher, the latter being 1 × 10⁻⁴ μCi/cm³, 10 times the value in the old 10CFR20. For other radionuclides and modes of exposure, the new calculated concentrations can be smaller, the same, or larger than the old.

An example adapted from NRC material will be helpful in understanding the new rule. Suppose that a worker in a nuclear plant receives 1 rem of external radiation and also is exposed over 10 working days to concentrations in air of iodine-131 of 9 × 10⁻⁹ μCi/ml and of cesium-137 of 6 × 10⁻⁸ μCi/ml (these correspond to the older MPCs). What is the fraction (or multiple) of the annual effective dose equivalent limit? We sum the fractions that each exposure is of the annual limit of 5 rems. The external exposure contributes . The ALI figures, taking account of the ICRP weighting factors for the various organs for the two isotopes, are 50 μCi for I-131 and 200 μCi for Cs-137. We need to find the actual activities taken in. With the standard breathing rate of 1.2 m³/h, in 80 h the air intake is 96 m³. The activities received are thus 0.86 μCi for I and 5.8 μCi for Cs. The corresponding fractions are 0.86/50 = 0.017 and 5.8/200 = 0.029, giving a total of external and internal fractions of or approximately of the limit. In this particular case, the expected hazard is lower than by the older method.

Other features of the new rule are separate limits on exposures (a) of body extremities—hands, forearms, feet, lower legs; (b) of the lens of the eye; and (c) of an embryo and fetus. The risk to the whole body per rem of dosage is 1 in 6000. For the limit of 5 rems the annual risk is 8 × 10⁻⁴, which is approximately eight times acceptable rates in “safe” industries. The figure is to be compared with the lifetime risk of cancer from all causes of approximately 1 in 6.

Dose limits for individual members of the public (0.1 rem/y) are quite a bit lower than those working with radionuclides (5 rems/y). In calculating concentrations of radionuclides in air released to an unrestricted area, differences in time of exposure, breathing rate, and average age are accounted for by dividing worker DAC values by 300 for inhalation or 219 for submersion. Examples (in μCi/ml) are Cs-137 (6 × 10⁻⁸)/300 = 2 × 10⁻¹⁰ and gaseous Xe-133 (1 × 10⁻⁴)/219 = 5 × 10⁻⁷.

21.8. Summary

Radiation protection of living organisms requires control of sources, barriers between source and living being, or removal of the target entity. Calculations required to evaluate external hazard include the dose as it depends on flux and energy, material, and time; the inverse square geometric spreading effect; and the exponential attenuation in shielding materials. Internal hazard depends on many physical and biological factors. Maximum permissible concentrations of radioisotopes in air and water can be deduced from the properties of the emitter and the dose limits. Application of the principle of ALARA is designed to reduce exposure to levels that are as low as reasonably achievable. There are many biological pathways that transport radioactive materials. New dose limit rules are based on the total effects of radiation—external and internal—on all parts of the body.

21.9. Exercises

1. What is the rate of exposure in mrems/y corresponding to a continuous 1-MeV gamma ray flux of 100 cm⁻² − s⁻¹? What dose equivalent would be received by a person who worked 40 h/w throughout the year in such a flux?
2. A Co-60 source is to be selected to test radiation detectors for operability. Assuming that the source can be kept at least 1 m from the body, what is the largest strength acceptable (in μCi) to assure an exposure rate of less than 500 mrems/y? (Note that two gammas of energy 1.17 and 1.33 MeV are emitted.)
3. By comparison with the Kr-85 analysis, estimate the MPC in air for tritium, average beta particle energy 0.006 MeV.
4. The nuclear reactions resulting from thermal neutron absorption in boron and cadmium are Which material would you select for a radiation shield? Explain.
5. Find the uncollided gamma ray flux at the surface of a spherical lead shield of radius 12 cm surrounding a very small source of 200 mCi of 1 MeV gammas.
6. Concentration limits of some radionuclides in water released to the public, according to 10CFR20 in the old and new versions are listed:

    

   Radionuclide	Concentration Limits (μCi/ml)
   	Old	New
   Tritium	3 × 10−3	1E-3
   Cobalt-60	3 × 10−5	3E-6
   Strontium	3 × 10−7	5E-7
   Iodine-131	3 × 10−7	1E-6
   Cesium-137	2 × 10−5	1E-6

   Calculate the ratio new/old for each radionuclide.
7. Water discharged from a nuclear plant contains in solution traces of strontium-90, cerium-144, and cesium-137. Assuming that the concentrations of each isotope are proportional to their fission yields, find the allowed activities per ml of each. Note the following data:

    

   Isotope	Half-life	Yield	Limit (μCi/ml)[†]
   90Sr	29.1 y	0.0575	5E-7
   144Ce	284.6 d	0.0545	8E-6
   137Cs	30.2 y	0.0611	1E-6

   
   ^† According to 10CFR20 (1993 version).
8. A 50-year exposure time is assumed in deriving the dose factors listed in Section 21.6. These take account of the radioisotope's physical half-life t^p and also its biological half-life t^b, which is the time it takes the chemical to be eliminated from the body. The effective half-life t^e can be calculated from the formula Find t^e for these three cases cited by Eichholz (see References):

    

   Radionuclide	tp	tb
   Iodine-131	8.04 d	138 d
   Cobalt-60	5.27 y	99.5 d
   Cesium-137	30.2 y	70 d

   If t^p and t^b are greatly different from each other, what can be said about the size of t^e?
9. The activities of U-238, Ra-226, and Rn-222 in a closed system are approximately equal, in accord with the principle of secular equilibrium. Assuming that the natural uranium content of soil is 10 ppm, calculate the specific activities of the isotopes in microcuries per gram of soil (Table 3.1 gives half-lives needed to calculate).

Computer Exercises

1. Program EXPOSO looks up gamma ray attenuation coefficients and buildup factors on data tables and finds the radiation exposure at a distance from a point source.
   • Run the program and explore its menus.
   • Verify that the flux at 10 cm from a point millicurie 1 MeV gamma ray source in lead is 39.2/cm²-s.
   • Use the program to find the lead distance from a millicurie 1 MeV source that yields 5 mrems/y, to within one millimeter.
   • Check the figures for a reactor in space (Section 21.3) with the shield option 7 (none).
2. A small research reactor core is located near the bottom of a deep pool of water. The water serves as moderator, coolant, and shield. (a) With a power of 10 MW and a fission neutron leakage fraction of 0.3, estimate, with the point source version of the computer program NEUTSHLD, the uncollided flux of fast neutrons at a distance of 20 ft from the core, treated as a point source. (b) Samples to be irradiated are placed near the core, the dimensions of which are 30 cm × 30 cm × 60 cm high. Assuming that the neutron source strength per unit area is uniform, calculate, with the plane version of NEUTSHLD, the fast neutron flux at 10 cm from the center of a large face of the core.
3. A study is made of leukemia incidence over a 100-km² area in the vicinity of a nuclear power plant. Some apparent clustering of cases is observed that might be attributed to proximity or wind direction. Run computer program CLUSTER to see how small samples of completely random statistical data normally are clustered. Then edit line 410 of the program from 100 to 1,000 and then to 10,000 to see the population becoming more uniform.
4. To improve the uniformity of irradiation of large objects in a water pool, a set of five “point” cobalt-60 sources (average gamma ray energy 1.25 MeV) are arranged in a plane at coordinates in centimeters (0, 0), (20, 20), (20, −20), (−20, −20), and (−20, 20). Explore the variation of total gamma flux over a parallel plane 10 cm away with computer program EXPOSO to calculate contributions of each source. Compare with results in a case where all five sources are concentrated near the point (0,0).

21.10 References

Radiation Information Network Radiation Information Network

http://www.physics.isu.edu/radinf/index1.html http://www.physics.isu.edu/radinf/index1.html

Numerous links to sources Numerous links to sources. Created by Bruce Busby, Idaho State University.

Schleien, et al 1998 Bernard Schleien, Lester A. Slayback Jr, Brian Kent Birky, Handbook of Health Physics and Radiological Health 3rd Ed. 1998 Williams & Wilkins Baltimore A greatly expanded version of a classical document of 1970.

Cember and Johnson 2008 Herman Cember, Thomas E. Johnson, Introduction to Health Physics 4th Ed. 2008 McGraw-Hill New York A thorough and easily understood textbook

Cember and Johnson 1999 Herman Cember, Thomas E. Johnson, The Health Physics Solutions Manual 1999 PS&E Publications Silver Spring, MD Recommended by the second author

Bevelacqua, 1995 Joseph John Bevelacqua, Contemporary Health Physics: Problems and Solutions 1995 John Wiley & Sons New York

Bevelacqua, 1999 Joseph John Bevelacqua, Basic Health Physics: Problems and Solutions 1999 John Wiley & Sons New York Helpful in preparing for CHP exam

Turner, 1995 James E. Turner, Atoms, Radiation, and Radiation Protection 1995 John Wiley & Sons New York

Dowd and Tilson 1999 Steven B Dowd, Elwin R. Tilson, Practical Radiation Protection and Applied Radiobiology 1999 W. B. Saunders Philadelphia All about radiation, its effects, and protection, from a nuclear medicine viewpoint. An appendix of Web sites maintained also at http://www.radscice.com/dowd.html

Shultis and Faw 1996 J. Kenneth Shultis, Richard E. Faw, Radiation Shielding 1996 Prentice-Hall Upper Saddle River, NJ Includes transport theory and Monte Carlo methods

Wood, 1982 James Wood, Computational Methods in Reactor Shielding 1982 Pergamon Press Oxford

Rockwell, 1956 Theodore Rockwell III, Reactor Shielding Design Manual 1956 McGraw-Hill New York A classic book on shielding calculations that remains a valuable reference

Eichholz, 1985 Geoffrey G Eichholz, Environmental Aspects of Nuclear Power 1985 Lewis Publishers Chelsea, MI

Faw and Shultis 1999 Richard E. Faw, J. Kenneth Shultis, Radiological Assessment: Sources and Doses 1999 American Nuclear Society La Grange Park, IL Fundamentals and extensive data. A reprint with a few changes of a 1991 book published by Prentice-Hall

Till and Meyer, 1983 John E. Till, H.Robert Meyer, Radiological Assessment, A Textbook on Environmental Dose Analysis, Nuclear Regulatory Commission 1983 Washington, DC, NUREG/CR-3332

Miller and Weidner 1986 Kenneth L. Miller, William A Weidner, CRC Handbook of Management of Radiation Protection Programs 1986 CRC Press Boca Raton An assortment of material not found conveniently elsewhere, including radiation lawsuit history, the responsibilities of health physics professionals, information about state radiological protection agencies, and emergency planning. More than half of the book is a copy of regulations of the Department of Transportation.

Radon Exposure of the United States Population—Status of the Problem, 1991 Radon Exposure of the United States Population—Status of the Problem 1991 NCRP Commentary No. 6, National Council of Radiation Protection and Measurements Bethesda, MD

International Basic Safety Standards for Protection Against Ionizing Radiation and for the Safety of Radiation Sources, 2004 International Basic Safety Standards for Protection Against Ionizing Radiation and for the Safety of Radiation Sources International Atomic Energy Agency, Vienna, Safety Series No. 115, CD-ROM 2004. Document can be downloaded from IAEA website.http://www-ns.iaea.org/standards/documents/default.asp?sub=160

Code of Federal Regulations Code of Federal Regulations, Energy 10, Office of the Federal Register, National Archives and Records Administration, United States Government Printing Office, Washington, DC (annual issuance).

10CFR Part 20—Standards for Protection Against Radiation 10CFR Part 20—Standards for Protection Against Radiation

http://www.nrc.gov/reading-rm/doc-collections/cfr/part020 http://www.nrc.gov/reading-rm/doc-collections/cfr/part020

Links to each subpart Links to each subpart (e.g., C: Occupational Dose Limits and D: Radiation Dose Limits for Individual Members of the Public).

NRC Update: New Reactor Licensing NRC Update: New Reactor Licensing

http://hps.ne.uiuc.edu/numug/archive/2006/presentations/harvey_ppt.pdf http://hps.ne.uiuc.edu/numug/archive/2006/presentations/harvey_ppt.pdf

PowerPoint presentation by R. Brad Harvey of NRC Office of Nuclear Reactor Regulation PowerPoint presentation by R. Brad Harvey of NRC Office of Nuclear Reactor Regulation, including present activities and future plans, 2006.

NRC Regulatory Guides NRC Regulatory Guides

http://www.nrc.gov/reading-rm/doc-collections/reg-guides http://www.nrc.gov/reading-rm/doc-collections/reg-guides

Issued in ten broad divisions Issued in ten broad divisions.

Select division 1… Select division 1…

1.109 Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR Part 50 Appendix I 1.109 Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR Part 50 Appendix I October 1977 (pdf 2.71 MB).

1.111 Methods for Estimation of Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors 1.111 Methods for Estimation of Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors July 1977 (pdf 1.25 MB).

Select division 8 Select division 8. Occupational Health.

8.8 Information Relevant to Ensuring that Occupational Radiation Exposures 8.8 Information Relevant to Ensuring that Occupational Radiation Exposures at Nuclear Power Stations Will be as low as Reasonably Achievable June 1978 (pdf 1.19 MB).

Committee on the Biological Effects of Ionizing Radiations Committee on the Biological Effects of Ionizing Radiations, National Research Council, Health Risks of Radon and Other Internally Deposited Alpha-Emitters, BEIR IV 1988 National Academy Press Washington DC http://books.nap.edu/openbook.php?isbn=0309037972

Select sections from Web site Select sections from Web site. Emphasizes lung cancer and the relationship of smoking and radon.

Committee on Health Effects of Exposure to Radon Committee on Health Effects of Exposure to Radon (BEIR VI), National Research Council, Health Effects of Exposure to Radon 1999 National Academy Press Washington DC http://books.nap.edu/openbook.php?isbn=0309056454 Select sections from Web site. Conclusion: new information needs to be considered and improved models developed

Radon Update Radon Update

http://www.physics.isu.edu/radinf/radon.htm#top http://www.physics.isu.edu/radinf/radon.htm#top

Based on an article by Dr. A. B Based on an article by Dr. A. B. Brill in Journal of Nuclear Medicine February 1994 Physics Department, Idaho State University provided by

Radon and hormesis Radon and hormesis

http://www.belleonline.com/newsletters/volume7/vol7-1/riskmanagement.html http://www.belleonline.com/newsletters/volume7/vol7-1/riskmanagement.html

Article by K. T. Bogen and D. W. Layton Article by K. T. Bogen and D. W. Layton.

Radiation, Science, and Health Radiation, Science, and Health

http://www.radscihealth.org/rsh http://www.radscihealth.org/rsh

Select Documents Select Documents. Organization criticizes conservatism of standards advisory bodies and government regulators.

NIST Physics Laboratory NIST Physics Laboratory

http://physics.nist.gov http://physics.nist.gov

Select Physical Reference Data Select Physical Reference Data, X-ray and Gamma-Ray Data. Figures differ slightly from those in NUREG/CR-5740.

ICRP Publication 2 ICRP Publication 2Recommendations of the International Commission on Radiological Protection Health Physics 3 1960 Pergamon Press Oxford