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358
199 NE
iii
n
d
18 Radiation Protection andRegulations
The radiation exposure from γ-ray and x-ray emitting radionuclides can be estimated from the exposure rate constant, Γ, which is dened as the exposure from
γ-rays and x-rays in R/h from 1mCi (37MBq) of a radionuclide at a distance of
1cm. Each γ- and x-ray emitter has a specic value of Γ, which has the unit of
R·cm2/mCi·h at 1cm or, in System Internationale (SI) units, μGy·m2/GBq·h at
1m. The Γ values can be calculated using the formula
where Ni is the fractional abundance of photons of energy Ei in MeV, and μi is the
mass absorption coefcient (cm2/g) of photons of energy Ei in air. Because γ-rays or
x-rays below some 10 or 20keV are absorbed by the container and thus do not contribute signicantly 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
Γ10 or Γ20. The values of Γ20 for different radionuclides are given in Table18.1.
The exposure rate X from an n-mCi radionuclide source at a distance d cm is
given by
x
2
(18.1)
where Γ is the exposure rate constant of the radionuclide.
Table 18.1 Exposure rate constants in lead for commonly used radionuclides in nuclear medicine
2
Г (C ∙ m
Radionuclides
37
Cs 6.64× 10
18
F 1.10× 10
131
I 4.26× 10
99m
Tc 1.54× 10
111
In 6.70× 10
67
Ga 1.55× 10
68
Ga 1.05× 10
57
Co 1.09× 10
60
Co 2.50× 10
201
Tl 8.72× 10
99
Mo 1.78× 10
177
Lu 3.52× 10
a
Adapted with permission from Smith DS and Stabin MG.Exposure rate constants and lead shield-
1cm)
/kg ∙ MBq ∙ s at
‐13
‐12
‐13
‐13
‐13
‐13
‐12
‐13
‐12
‐14
‐13
‐14
ing values for over 1100 radionuclides. Health Physics (2012); 102(3): 271–291. DOI: 10.1097/
HP.0b013e318235153a. PMID: PMID: 22420019
b
These values are obtained by multiplying the values in column 3 by 27.027, since the R ∙ cm2/mCi
∙ h at 1cm=27.027 μGy ∙ m
2
/GBq ∙ h at 1m
Г (µGy ∙ m2/GBq ∙ h at
b
1m)
Г (R ∙ cm2/mCi ∙ h at
1cm)
92.70 3.43
153.51 5.68
59.46 2.20
21.08 0.78
93.51 3.46
2.16 0.08
146.76 5.43
25.95 0.96
348.65 12.9
12.16 0.45
24.86 0.92
4.89 0.181
a
./
Rh
..
./
18.2 Radiation Protection
Problem 18.1
Calculate the radiation exposure at 25 cm from a vial containing 30 mCi
(1.11GBq) of
201
Tl.
Answer
The exposure rate constant Γ20 of
201
Tl is 0.45 R·cm2/mCi·h at 1cm from
Table18.1. Therefore, using Eq. (18.1), at 25cm
.
30 045
X
25
2
21 6
m
359
Because Γ20 of
1.11GBq of
201
201
Tl in SI units is 12.16 μGy·m2/GBq·h at 1m, X for
Tl at 25cm is
1111216
X
215 96
025
.
2
Gy h
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.
18.2.5.3 Shielding
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 the linear attenuation coefcient 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.
360
Problem 18.2
18 Radiation Protection andRegulations
Calculate the number of HVLs and the amount of lead necessary to reduce the
exposure rate from 100mCi (3.7GBq) of
131
I to less than 10 mR/h at 10cm
from the source. (Γ=2.17R·cm2/mCi·h at 1cm and 1 HVL=3mm of lead).
Answer
x
Exposureat10cm=
2170 100
2
10
= 2170mR /h.
A factor of 2170/10=217 or more would be needed to reduce the exposure
to less than 10 mR/h. In terms of HVL, 28=256, that is, 8 HVLs would be
needed. Since 1 HVL=3mm of lead, 8 HVLs would be equal to 24mm.
Therefore, 8 HVLs or 24mm of lead would be necessary.
18.2.5.4 Activity
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.
18.2.6 Personnel Monitoring
According to 10CFR20.1502, personnel monitoring is required under the following
conditions:
1. Adults likely to receive in 1year a dose in excess of 10% of the annual limit of
exposure from the external radiation source.
2. Minors likely to receive, in 1year, from external radiation sources external to the
body, a deep dose equivalent (Hd) in excess of 0.1rem (1mSv), a an eye lens
dose equivalent (Hd) in excess of 0.15rem (1.5mSv), or a shallow dose equivalent to the skin or to the extremities in excess of 0.5rem (5mSv).
3. Declared pregnant women likely to receive during entire pregnancy a deep dose
equivalent (H
4. Individual entering a high radiation or very high radiation area.
Monitoring for occupational intake of radioactive material is also required if the
annual intake by an individual is likely to exceed 10% of the ALIs in 10CFR20,
Appendix B, and if minors and declared pregnant women during the entire period of
pregnancy are likely to receive a committed effective dose equivalent in excess of
0.1rem (1mSv) in 1year.
Four devices are used to measure the exposure of ionizing radiations received by
an individual: the pocket dosimeter, the lm badge, the thermoluminescent
) in excess of 0.1rem (1mSv).
E, 50
18.2 Radiation Protection
361
Fig. 18.2 (a) Film badge.
(b) Film badge holder. (c)
TLD ring badge. (d)
Pocket dosimeter
ab
cd
dosimeter and the optically stimulated luminescent dosimeter. The pocket dosimeter (Fig. 18.2d) has been described in Chap. 7 and other three devices are
described below.
18.2.6.1 Film Badge
The lm badge is most popular and cost-effective for personnel monitoring and
gives reasonably accurate readings of exposures from β-, γ-, and x-radiations. The
lm badge consists of a radiation-sensitive lm held in a plastic holder (Fig.18.2a,
b). Filters of different metals (aluminum, copper, and cadmium) are attached to the
holder in front of the lm 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 lm is measured by a densitometer and compared with that of a calibrated lm 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 lm badge is the long waiting
period (a month) before the exposed personnel know about their exposure. The lm
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
lm badges of all workers are normally sent to a commercial rm that develops and
reads the density of the lms and sends back the report of exposure to the institution. The commercial rm must be accredited by the National Voluntary Laboratory
Accreditation Program (NVLAP) of the National Institute of Standards and
Technology.
18.2.6.2 Thermoluminescent Dosimeter
A thermoluminescent dosimeter (TLD) consists of inorganic crystals (chips) such
as lithium uoride (LiF) and manganese-activated calcium uoride (CaF2: Mn) held
in holders like the lm badges and plastic rings (Fig.18.2c). 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
362
18 Radiation Protection andRegulations
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).
18.2.6.3 Optically Stimulated Luminescence Dosimeter
Current dosimeter of choice in radiation dosimetry is the optically stimulated luminescence (OSL) dosimeter, which works like TLD except that during stimulation of
the radiation-exposed crystals, light is used instead of heat. The detector materials
used in the dosimeters are quartz or aluminum oxide, which are made imperfect by
adding some impurities such as traces of carbon (C). Electron-hole pairs are created
by the interaction of radiation with the crystals, which are trapped in the impurity
lattices located in the forbidden band between the valence band and the conduction
band. When laser light or an LED is applied to the crystals, the trapped electrons are
stimulated and move to the conduction band. The electrons subsequently combine
with the positive holes creating visible light (luminescence), which is fed into a PM
tube to produce a signal.
Landauer Inc. and Mirion Technologies (Capintec) Inc are the primary vendors
for OSL dosimeters. These dosimeters are designed in different shapes with different lters to differentiate between β‐ and γ radiations. Ring-shaped dosimeters are
also available for ngers too. The range of readings of the dosimeter is 1mR to 1000
mR for γ rays and 10 mR to 1000 mR for β‐ radiations. These dosimeters are useful
for personnel monitoring as well as for invivo dosimetry in radiation therapy. OSL
dosimeters are not affected by heat, light, and humidity.
Radiation workers normally wear the dosimeters on the chest or waist, and ringshaped dosimeters on ngers for a month, after which they are sent to a commercial
rm accredited by the National Voluntary Laboratory Accreditation Program
(NVLAP) of the National Institute of Standards and Technology (NIST) for processing, after which the reports of exposure readings are sent back to the institution.
Landauer Inc. and Mirion Technologies (Capintec) Inc. are two NVLAP-accredited
commercial rms. Radiation workers are informed of their monthly exposures, and
the permanent records of exposure are kept by the Radiation Safety Ofce under the
authority of the licensee.
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 lm 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.
18.3 Radiation Regulations
363
18.2.7 Dos andDon’ts inRadiation Protection Practice
Do wear laboratory coats and gloves when working with radioactive materials.
Do work in a ventilated fume hood while working with volatile material.
Do cover the trays and workbench with absorbent paper.
Do store and transport radioactive material in lead containers.
Do wear a lm badge while working in the radiation laboratory.
Do identify all radionuclides and dates of assay on the containers.
Do survey work areas for contamination as frequently as possible.
Do clean up spills promptly and survey the area after cleaning.
Do not eat, drink, or smoke in the radiation laboratory.
Do not pipette any radioactive material by mouth.
Do monitor hands and feet after the day’s work.
Do notify the radiation safety ofcer (RSO) in the case of any major spill or other
emergencies related to radiation.
18.3 Radiation Regulations
18.3.1 License
Licenses are issued by the NRC or the Agreement State to various facilities, institutions, or individuals for the use of by-product materials and fall into several categories depending on the specic use as described below:
18.3.1.1 General License
A general license (10CFR31) is given to any physician, veterinarian, clinical laboratory, or hospital to acquire, possess, transfer, or use of the following by- product
materials in prepackaged units containing limited activities (given in parenthesis)
for each specic use:
125
I (10μCi or 370kBq),
131
I (10μCi or 370kBq), 14C (10μCi
or 370kBq)), 3H (50μCi or 1.85MBq)), 59Fe (20μCi or 740kBq), 75Se (10μCi or
370kBq), and 57Co (10μCi or 370kBq). The licensee must be an authorized user
according to 10CFR35 and can possess a maximum of 200μCi (7.4MBq) of the
approved radionuclide at any one time and at any one location.
18.3.1.2 Specific License of Limited Scope
This license is granted to private or group practices and medical institutions for
medical use of by-product materials in humans. The authorized users are specically listed on the specic license of limited scope, but only limited quantities of
specic radionuclides for intended uses are granted. For patients requiring hospitalization under 10CFR35.75, only hospitals having inpatient facilities are authorized
to treat such patients. The specic license of limited scope can also be issued to
mobile services.
364
18 Radiation Protection andRegulations
18.3.1.3 Specific Licenses of Broad Scope
The specic licenses are given in two categories: one to manufacture or transfer for
commercial distribution certain items containing by-product material (10CFR32)
and the other to possess, use, and transfer by-product material in any chemical or
physical form with the limitations of the maximum activity specied (10CFR33).
The former types of specic licenses are typically given to commercial manufacturers. The latter type is called the specic license of broad scope or “broad license”
and has three categories based on the maximum activity allowed for the receipt,
acquisition, ownership, possession, use, and transfer of any chemical or physical
form of by-product material (10CFR33.11). The Type A broad license allows specied quantities of activities usually in multicuries; the Type B broad license allows
maximum activities of by-product material specied in 10CFR33.100, Schedule A,
Column I; and the Type C license permits maximum activities of byproduct material
specied in 10CFR33.100, Schedule A, Column II, which are an order of magnitude
less than those in the Type B license. In type B and Type C broad scope licenses, if
two or more radionuclides are possessed at any time, then the possession limit of
each is determined by calculating the ratio of the radionuclide in possession to the
applicable quantity in part 33.100, schedule A, column I for Type B and those in
column II for type C, and summing them up, which should not exceed 1.
In the Type A license, a radiation safety committee and a radiation safety ofcer
are required to implement and monitor all aspects of radiation safety in the use and
disposal of by-product material. Such licenses are mainly offered to large medical
institutions with previous experience that are engaged in medical research, and in
diagnostic and therapeutic uses of by-product material. Individual users are authorized by the radiation safety committee to conduct specic protocols using byproduct materials.
The Type B specic license requires a radiation safety ofcer, but no radiation
safety committee, to implement and monitor all radiation safety regulations. The
Type C specic license requires neither the radiation safety ofcer nor the committee, but a denite statement that the by-product material will be used by the licensee
or by persons under his direct supervision who has the training specied in
10CFR33.15.
In all cases of specic licenses, an application must be led to the NRC using the
NRC Form 313 with all information related to the possession, use, and disposal of
by-product materials.
18.3.2 Radiation Safety Committee
The management of an institution is required to establish a Radiation Safety Committee
(RSC) for certain aspects of the use of radioactive materials such as in therapeutic use.
For uptake ane dilution (10CFR35.100) and imaging and localization (10CFR35.200),
an RSC is not required. An RSC is composed of an authorized user for each type of
use, the RSO, a nurse representative, a management representative and members of
18.3 Radiation Regulations
365
other disciplines as appropriate. One of the members is designated by the management as the chairman of the committee. The committee meets as frequently (minimum quarterly) as needed. The function of the committee is to review the proposals
submitted by institutional members, and based on the review regarding the safe use of
radiation, the committee approves or disapproves the proposal.
18.3.3 Radiation Safety Officer
A Radiation Safety Ofcer (RSO) is a qualied person with training and experience
in the use of byproduct materials and management of a radiation protection program. Of many responsibities, the following are the prime duties:
Implementation of all radiation protection programs mandated by the license.
Preventing or even stopping unsafe activities.
Monitoring personnel exposure record and advising how to reduce exposure.
Acting as a liaison between the facility and the NRC or Agreement State.
Keeping inventories and sealed sources.
Control of waste disposal.
Training of Personnel.
Implementation Department of Transportation regulations.
Investigation of the radiation accident and spill.
Annual audit of the program.
The management can appoint an Associate Radiation Safety Ofcer (ARSO) to
support the RSO in his activities.
18.3.4 Medical Uses ofRadioactive Materials
The NRC and Agreement States regulate the medical uses of by-product materials
by implementing 10CFR35. There are six categories of medical uses of radioactive
materials according to 10CFR Part 35. They are: (1) radiopharmaceuticals for
uptake, dilution, and excretion (10CFR35.100); (2) radiopharmaceuticals for imaging and localization including generators and kits (10CFR35.200); (3) radiopharmaceuticals for therapy (10CFR35.300); (4) sealed sources for brachytherapy
(10CFR35.400); (5) sealed sources for diagnosis such as sources of
for bone mineral analysis (10CFR35.500); and (6) sealed sources for teletherapy,
such as sources of 60Co and
137
Cs in teletherapy units or gamma stereotactic radio-
surgery units (10CFR35.600).
The regulations for the medical use of all radioactive materials are given in
10CFR35, but radiopharmaceuticals under categories 1, 2, and 3 only are relevant in
nuclear medicine. These radiopharmaceuticals must be approved for human clinical
use by the FDA under an IND or NDA. The
99
99m
Mo‐
Tc generator and reagent kits are used to prepare
99m
Tc activity is eluted from the
99m
Tc-labeled radiophar-
maceuticals according to instructions given by the manufacturer in the package
125
I and
153
Gd
366
18 Radiation Protection andRegulations
inserts. Only reagent kits that are approved by the FDA under an IND or NDA may
be used for radiopharmaceutical preparation. Many other radiopharmaceuticals are
prepared by the manufacturers using appropriate labeling methods. The following is
a brief description of the pertinent rules of 10CFR35.
18.3.4.1 Applications, Amendments, andNotifications
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:
(a) Appointment or discontinuation of an authorized user, radiation safety ofcer,
authorized medical physicist, or authorized nuclear pharmacist
(b) Change of name or address of the licensee
(c) Change or addition of the use areas
(d) Use of excess or new by-product materials not permitted before in the license
Notication of the above must be made within 30days 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 specic
license of broad scope are exempt from these requirements which are managed by
the RSC of the institution.
18.3.4.2 Authority andResponsibilities oftheLicensee
According to 10CFR35.24, the licensee’s management is responsible for the overall
implementation of the radiation protection program in the medical uses of byproduct material. The licensee’s management shall approve in writing all new
authorized users, radiation safety ofcer, 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 Ofcer (RSO), who
accepts in writing responsibilities to implement a radiation protection program. It
may appoint one or more temporary RSOs for 60days 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 byproduct material. Examples are the use of therapeutic quantities of unsealed byproduct 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 byproduct 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 denite
frequencies of the RSC meetings nor record-keeping of the minutes.
18.3 Radiation Regulations
367
18.3.4.3 Supervision
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.
18.3.4.4 Mobile Nuclear Medicine Service
According to 10CFR35.80, a licensee providing mobile nuclear medicine service to
a client must
(a) Have a letter, or memorandum of understanding (MOU), signed by the licensee
and the management of each client spelling out the details of the responsibility
and authority of the client and the licensee
(b) Calibrate and check the instruments for measuring dosages and surveying
(c) Measure dosages and perform surveys of the area of uses at the client
address, and
(d) The client must have a license for receiving and using by-product material.
18.3.4.5 Written Directives
According to 10CFR35.40, a written directive is required when a dosage greater
than 30μCi (1.11MBq) of
uct material other than
131
I-NaI or a therapeutic dosage of an unsealed byprod-
131
I-NaI is administered to a patient or human research subject. The written directive must be dated and signed by an authorized user and must
contain the patient’s name, the dosage, the name of the drug, and route of administration. A revision of the written directive can be made, if necessary, provided it is
signed and dated by the authorized user before administration. In case of an emergency, an oral revision to an existing written directive is acceptable, which must be
followed by a written directive within 48h.
According to 10CFR35.41, the licensee shall develop and maintain a copy of the
written procedures for the written directive that include specic verications of the
identity of the patient before each administration, and that the administration is in
accordance with the written directive. The identity of the patient may be veried by
the name, driver’s license, birthday, any hospital’s I.D. number, and so on.
18.3.4.6 Measurement ofDosages
According to 10CFR35.63, all dosages for patient administration must be measured
in an instrument (dose calibrator) that is calibrated with nationally recognized standards or the manufacturer’s instructions (10CFR35.60). Although the methods of
calibration are not specically prescribed in 10CFR35, the constancy, accuracy,
linearity, and geometry of the dose calibrator must be checked as described in
Chap. 7.