The total radiation exposure to an individual is directly proportional to the time of exposure to the radiation source . The longer the exposure, the higher the radiation dose. Therefore, it is wise to spend no more time than necessary near radiation sources.

The intensity of a radiation source, and hence the radiation exposure, varies inversely as the square of the distance from the source to the point of exposure . It is recommended that an individual should keep as far away as practically possible from the radiation source. Procedures and radiation areas should be designed so that individuals conducting the procedures or staying in or near the radiation areas receive only minimum exposure.

The radiation exposure from γ-ray and x-ray emitting radionuclides can be estimated from the exposure rate constant , Γ, which is defined as the exposure from γ-rays and x-rays in R/h from 1 mCi (37 MBq) of a radionuclide at a distance of 1 cm. Each γ– and x-ray emitter has a specific value of Γ, which has the unit of R · cm²/mCi · h at 1 cm or, in System Internationale (SI) units, µGy · m²/GBq · h at 1 m. The Γ values are derived from the number of γ-ray and x-ray emissions from the radionuclide, their energies, and their mass absorption coefficients in air.¹ Because γ-rays or x-rays below some 10 or 20 keV are absorbed by the container and thus do not contribute significantly to radiation exposure, often γ-rays and x-rays above these energies only are included in the calculation of Γ. In these instances, they are denoted by Γ ₁₀ or Γ ₂₀. The values of Γ ₂₀ for different radionuclides are given in Table 16.1.

The exposure rate X from an n-mCi radionuclide source at a distance d cm is given by

where Γ is the exposure rate constant of the radionuclide.

It should be pointed out that because the patient is not a point source, the exposure rate does not vary exactly as the inverse square of the distance.

Various high atomic number (Z) materials that absorb radiations can be used to provide radiation protection . Because the ranges of α– and β-particles are short in matter, the containers themselves act as shields for these radiations. γ-Radiations, however, are highly penetrating. Therefore, highly absorbing material should be used for shielding of γ-emitting sources, although for economic reasons, lead is most commonly used for this purpose. The half-value layer (HVL) of absorbent material for different radiations is an important parameter in radiation protection and is related to linear attenuation coefficient of the photons in the absorbing material. This has been discussed in detail in Chap. 6.

Obviously, shielding is an important means of protection from radiation. Radionuclides should be stored in a shielded area. The radiopharmaceutical dosages for patients should be carried in shielded syringes. Radionuclides emitting β-particles should be stored in containers of low-Z material such as aluminum and plastic because in high-Z material, such as lead, they produce highly penetrating bremsstrahlung radiations. For example, ³²P is a β ⁻ emitter and should be stored in plastic containers instead of lead containers.

It should be obvious that the radiation exposure increases with the intensity of the radioactive source . The greater the source strength, the more the radiation exposure. Therefore, one should not work unnecessarily with large quantities of radioactivity.

The film badge is most popular and cost-effective for personnel monitoring and gives reasonably accurate readings of exposures from β-, γ– and x-radiations. The film badge consists of a radiation-sensitive film held in a plastic holder (Fig. 16.2a, b). Filters of different metals (aluminum, copper, and cadmium) are attached to the holder in front of the film to differentiate exposure from radiations of different types and energies. Filters of metals of different densities stop different energy radiations, thus discriminating exposures from them. After exposure the optical density of the developed film is measured by a densitometer and compared with that of a calibrated film exposed to known radiation. Film badges are usually changed monthly for radiation workers in most institutions. Film badges provide an integral dose and a permanent record. The main disadvantage of the film badge is the long waiting period (a month) before the exposed personnel know about their exposure. The film badge also tends to develop fog resulting from heat and humidity, particularly when in storage for a long time, and this may obscure the actual exposure reading. The film badges of all workers are normally sent to a commercial firm that develops and reads the density of the films and sends back the report of exposure to the institution. The commercial firm must be accredited by the National Voluntary Laboratory Accreditation Program (NVLAP) of the National Institute of Standards and Technology.

A thermoluminescent dosimeter (TLD) consists of inorganic crystals (chips) such as lithium fluoride (LiF) and manganese-activated calcium fluoride (CaF₂:Mn) held in holders like the film badges and plastic rings (Fig. 16.2c) . When these crystals are exposed to radiation, electrons from the valence band are excited and trapped by the impurities in the forbidden band. If the radiation-exposed crystal is heated to 300 to 400°C, the trapped electrons are raised to the conduction band; they then fall back into the valence band, emitting light. The amount of light emitted is proportional to the amount of radiation absorbed in the TLD. The amount of light is measured and read as the amount of radiation exposure by a TLD reader, a unit that heats the crystal and reads the exposure as well. The TLD gives an accurate exposure reading and can be reused after proper heating (annealing).

It should be noted that exposure resulting from medical procedures and background radiations are not included in occupational dose limits . Therefore, radiation workers should wear film badges or dosimeters only at work. These devices should be taken off during any medical procedures involving radiation such as radiographic procedures and dental examinations, and also when leaving after the day’s work. Also radiation workers should not wear these badges for certain period of time after undergoing a diagnostic or therapeutic nuclear medicine procedure or radiation therapy permanent implant procedure.

Although 10CFR20 does not spell out the conditions of the decay-in-storage method , 10CFR35.92 describes this method in detail. Radionuclides with half-lives less than 120 days usually are disposed of by this method. These radionuclides are allowed to decay in storage and monitored before disposal. If the radioactivity reading of the waste is indistinguishable from background, it can be disposed of in the normal trash after removal or defacing of all radiation labels. This method is most appropriate for short-lived radionuclides such as ^99mTc, ¹²³I , ²⁰¹Tl, ¹¹¹In, ⁶⁷Ga and ¹³¹I commonly used in nuclear medicine. Radioactivities should be stored separately according to half-lives for convenience of timely disposal of each radionuclide.

The NRC permits radioactive waste disposal into the sewerage system provided the radioactive material is soluble or dispersible in water and the quantity disposed of monthly does not exceed the maximum permissible limits set in 10CFR20 . Disposal depends on the total volume and flow rate of water used but is limited to 1 Ci (37 GBq) of ¹⁴C, 5 Ci (185 GBq) of ³H , and 1 Ci (37 GBq) of all other radionuclides annually. Excreta from humans undergoing medical diagnosis or treatment with radioactive material are exempted from these limitations. However, items contaminated with radioactive excreta (e.g., linen, diapers, etc., contaminated with urine or feces) are not exempted from these limitations. To adopt this method of radioactive disposal, one must determine the total volume and the flow of sewer water in the institution and the number of users of a specific radionuclide so that for each individual user, a limit can be set for sewer disposal of the radionuclide in question.

This method of transfer to an authorized recipient is adopted for long-lived radionuclides and usually involves transfer of radioactive wastes to authorized commercial firms that bury or incinerate at approved sites or facilities .

Although the columns of the ⁹⁹Mo–^99mTc generators may be decayed to background for disposal to normal trash, a convenient method of disposing of this generator is to return them to the vendors, who let them decay and later dispose of them. Normally, the used generator is picked up by the authorized carrier when a new one is delivered.

A licensee may adopt methods of radioactive waste disposal different from those mentioned here, provided regulatory agency approval is obtained . Impact of such disposal methods on environment, nearby facilities, and population is heavily weighed before approval. Incineration of solid radioactive waste and carcasses of research animals containing radioactive materials is allowed by this method . Radioactive gases such as ¹³³Xe and ¹²⁷Xe are released by venting through the fumehood, provided their maximum permissible concentration at the effluent side of the exhaust to the atmosphere does not exceed the NRC limits. Radioactive waste containing 0.05 µCi (1.85 kBq) or less of ³H or ¹⁴C/g of medium used for liquid scintillation counting or animal tissue may be disposed of in the regular nonradioactive trash.

Records must be maintained as to the date of storage and the amount and kind of activity stored in a waste disposal log book. The stored packages must be labeled with pertinent information. The date of disposal and the amount of disposed activity must also be recorded in the log book, along with the initials of the individual disposing of the waste.

As already mentioned, applications for a license and its renewals must be made by the licensee’s management for the medical uses of by-product materials. Amendments to the license must be made by the licensee’s management for the following:

Notification of the above must be made within 30 days of occurrence. Change or addition of areas of use for uptake and dilution (10CFR35.100) and for localization and imaging (10CFR35.200) need not be amended. Licenses with Type A specific license of broad scope are exempt from these requirements.

According to 10CFR35.24, the licensee’s management is responsible for the overall implementation of the radiation protection program in the medical uses of by-product material. The licensee’s management shall approve in writing all new authorized users , radiation safety officer , or nuclear pharmacist, and ministerial changes in the radiation safety program that do not require license amendment (10CFR35.26).

The licensee’s management shall appoint a Radiation Safety Officer (RSO) , who accepts in writing responsibilities to implement a radiation protection program. It may appoint one or more temporary RSOs for 60 days in a year, if all conditions of an RSO are met.

The licensee’s management also must appoint a Radiation Safety Committee (RSC) , if the licensee is authorized for two or more different types of uses of by-product material. Examples are the use of therapeutic quantities of unsealed by-product material (10CFR35.300) and manual brachytherapy (10CFR35.400), or manual brachytherapy and low-dose-rate therapy units (10CFR35.600), or teletherapy units (10CFR35.600) and gamma knife units (10CFR35.600). Use of by-product materials for both uptake and dilution (10CFR35.100) and imaging and localization (10CFR35.200) does not require an RSC. The RSC must include as a minimum an authorized user of each type of use permitted in the license, the RSO, a representative of the nursing service, and a representative of management, and in addition, other members, if appropriate. The NRC does not prescribe any definite frequencies of the RSC meetings nor record-keeping of the minutes.

According to 10CFR35.27, a licensee that permits an individual to work under an authorized user or authorized nuclear pharmacist using by-product material must instruct the supervised individual to follow strictly all regulations and conditions of the license and all procedures involving by-product material. There is no requirement for periodic review of the supervised individual’s work and records. The licensee is responsible for the acts and omissions of the supervised individuals.