The Health Physics Society defines radiation as “energy that comes from a source and travels through space.”¹ The source could be atomic particles such as alpha and beta emissions as well as electromagnetic energy associated with AM/FM radio, radar, visible light, ultraviolet light, x-rays, and gamma rays. Radiation with enough energy to remove an electron from an atom is termed ionizing radiation.² A characteristic of x-rays, gamma rays, alpha and beta particles, radiation having this ability can lead to biological damage when absorbed in human tissue.

This chapter will introduce the common units to describe radiation, sources of radiation exposure, radiation dose limits, and an introduction to radiation biology. The chapter will further focus on radiation safety and protection regulations pertinent to the practice of Nuclear Cardiology. These regulations are governed by the Nuclear Regulatory Commission (NRC) and are found in Title 10, Parts 19, 20, and 35 of the Code of Federal Regulations (CFR). NRC NUREG 1556 Volume 9, Revision 2 provides guidance specific to radioactive materials licensing and offers suggested policies and procedures for radiation safety compliance.

There are several conventional terms used when describing radiation. These include exposure, absorbed dose, and dose equivalent. Named after Wilhelm Roentgen, the scientist who discovered x-rays in 1895, the Roentgen (R) is the unit of radiation exposure in air.³ In comparison to the International System of Units (SI), one Roentgen corresponds to the amount of radiation required to liberate 2.58 × 10^-4 Coulombs per kilogram of air.

The Rad, or Radiation Absorbed Dose, is the measure of the amount of energy absorbed by an object as radiation passes through.⁴ The amount of energy absorbed is dependent on the energy of the incident photon and the composition of the material. The f-factor is a tissue weighting factor used to convert exposure in air (R) to absorbed dose (rad) in tissue taking into account the x-ray or gamma ray energy and effective atomic number of the tissue exposed. For example, a 100-keV gamma photon incident on fat will transfer 91% of its energy, whereas the same photon will deliver 96% of its energy to muscle tissue. Therefore, a source of radiation exposing a point in air to 100 R will deliver a dose of 91 rad to fat tissue and 96 rad to muscle tissue at the same reference point.

Dose equivalent is a term used to quantify the amount of energy deposited in tissue along with the associated biological risk from the type of radiation.⁵ The conventional unit for dose equivalent is rem which is calculated by multiplying the radiation absorbed dose (rad) by a radiation quality factor or QF.⁶ Table 2-1 illustrates that the QF for x-rays, gamma rays, and beta particles is equal to 1 whereas the QF for alpha particles is 20.⁷ This means that an absorbed dose of 10 rads from an x-ray, gamma ray, or beta particle equates to 10 rem (10 rad × 1), whereas the dose equivalent would be 200 rem (10 rad × 20) if originating from alpha particulates.

Although the conventional unit to describe absorbed dose and dose equivalent is the rad and rem, the international community now uses the term Gray (Gy) and Sievert (Sv), respectively. 1 Gy = 100 rad and 1 Sv = 100 rem. Table 2-2 provides a summary of the units of radiation exposure, absorbed dose, and dose equivalent.

Ionizing radiation occurs naturally in the earth’s soil and rock, in the cosmic rays descending from the sun and stars, as well as in our own body.⁸ This ubiquitous background radiation in the United States totals approximately 3.1 mSv/yr across the population.⁹ Exposure to Radon and its decay products accounts for the majority of the natural background radiation.

In addition to natural background, people are exposed to radiation in manufactured products such as smoke detectors, ceramics, and building materials, as well as from medical sources including diagnostic x-rays and nuclear medicine procedures.¹⁰ The effective radiation dose from these products accounts for an additional 3.1 mSv/yr (Fig. 2-1). It is important to note that these estimates are population based and not all individuals are exposed to radiation from manufactured products and radiological examinations.

The term BERT, or Background Equivalent Radiation Time, has recently been introduced to explain levels of radiation dose to patients relative to natural background radiation exposure. Table 2-3 illustrates radiation doses to patients from Nuclear Cardiology procedures and their associated BERT.

The regulations for dose limits in the United States can be found in Title 10, Part 20 of the Code of Federal Regulations (10 CFR 20). Table 2-4 illustrates these limits, which include limiting radiation to occupational workers, the embryo/fetus of an occupational worker, and the public. Both occupational and public dose limits exclude exposure to natural background radiation.

While the dose limit to the public from sources of ionizing radiation is 1 mSv (0.1 rem) per year, this limit may be increased to 5 mSv (0.5 rem) from an infrequent exposure related to another person’s medical procedure providing the authorized user determines the exposure is appropriate.

Because it is impractical to set occupational worker dose limits to that of the public, regulators rely on the Linear Non-Threshold risk model to set limits at a point below which the threshold for radiation effects are known and at a point to minimize the theoretical effects from the stochastic risks from ionizing radiation exposure. The dose limits are established at levels of risk already assumed by occupational workers.¹¹ To this point, radiation dose limits are established at a level where the risk of a fatal cancer from occupational exposure to ionizing radiation is similar to the assumed risk of a fatal work accident, which is 1 in 10,000 annually.¹²

In addition to ensuring dose limits are not exceeded, an institution must adopt investigational levels in order to maintain radiation levels consistent with the As Low As Reasonably Achievable (ALARA) philosophy. These quarterly investigational levels may be established by the institution, or the facility may adopt those recommended by the NRC. As illustrated in Table 2-5, there are two investigational levels associated with various monitoring points on the body. An ALARA Level I is a dose equal to 10% of the annual dose limit whereas an ALARA Level II is triggered at 30% of the annual dose limit. To help ensure annual dose limits are not exceeded, these ALARA investigational levels are routinely checked on a quarterly basis.

The NRC recommends the following actions must be taken when an employee exceeds ALARA investigational levels:

ALARA I Exceeded

The RSO or designee should investigate and review actions that might be taken to reduce a recurrence. No further action is necessary unless determined otherwise by the RSO.

ALARA II Exceeded

The RSO should perform a timely investigation into the cause, take action to reduce the recurrence, and submit a report to the institution’s Radiation Safety Committee.

Patients administered with radioactive material may be released into the public providing they are not likely to expose any individual to a point where their dose equivalent could exceed 5 mSv (0.5 rem).¹³ Furthermore, the patient or patient’s guardian must be provided with a set of instructions to maintain doses to as low as reasonably achievable if the total effective dose equivalent to any other individual is likely to exceed 1 mSv (0.1 rem). Breast-feeding cessation guidelines must also be provided if the dose equivalent to a nursing child could exceed 1 mSv (0.1 rem).

NRC Radiation Guide 8.39, “Release of Patients Administered Radioactive Material”¹⁴ provides guidance for the release of patients, administered activities of radiopharmaceuticals requiring instructions, and breast-feeding cessation recommendations. As illustrated in Table 2-6, patients administered with diagnostic activities of ^99mTc-Sestamibi may be immediately released without radiation safety instructions and do not require stoppage of breast-feeding.