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Radiation Safety and Dosimetry
https://t.me/med1917
Michael Lassmann and Uta Eberlein
6
Contents
6.1 Introduction 103
6.2
Basic Quantities and Definitions 104
6.2.1 Absorbed Dose (D) 104
6.2.2 Dosimetry in Nuclear Medicine 104 Effective Dose (E) 105
6.2.3
6.2.4 External Exposure 105
6.3
Patient Dosimetry for SLN
Diagnostics with Tc-99m 106
6.3.1 Patients 106
6.3.2 Pregnancy 107
6.3.3 Lactating Women 107
Patient Dosimetry for Other Isotopes 107
6.4
6.5
Staff Exposure for SLN Diagnostics
with Tc-99m 107
6.5.1 General Rules 107
6.5.2 Staff in Nuclear Medicine Department 108
6.5.3 Staff in Operating Room 108
6.5.4 Pregnant Staff in Operating Room 108 Staff in Pathology Department 108
6.5.5
6.5.6 Radiation Safety Precautions 109
6.5.7 Radioactive Clinical Waste 109
6.6 Staff Exposure for Other Isotopes 109
6.7
Discussion and Conclusion 109
References 109
Abstract
The use of radioactive substances for senti­nel lymph node (SLN) biopsy needs consid­eration on how to optimize radiation safety issues. In this chapter an overview on basic radiation-related quantities, definitions, and on patient dosimetry is given. Values of absorbed and effective doses are provided for patients, the staff in the operation theater and in the pathology department, and for waste disposal. For radiolocalization of SLN with Tc-99m, the dose to the patients and the exposure of the staff are low; for radiopharmaceuticals with longer half-lives and/or positron emitters, individual moni­toring of staff exposure and contamination should be considered for larger patient numbers.
6.1 Introduction
M. Lassmann (*) • U. Eberlein Department of Nuclear Medicine,
University Würzburg, Würzburg, Germany
e-mail: lassmann_m@ukw.de
© Springer International Publishing Switzerland 2016 K. Herrmann et al. (eds.), Radioguided Surgery: Current Applications and Innovative Directions in Clinical Practice, DOI 10.1007/978-3-319-26051-8_6
The use of radioactive substances for sentinel lymph node (SLN) biopsy needs the optimization of radiation safety issues for patients, for the staff in nuclear medicine departments, in the operating theater, pathology laboratories, and also the dis­posal of radioactive waste.
103
104
dm
rr
kh
=←
()
kh
()
kh
()
https://t.me/med1917
M. Lassmann and U. Eberlein
6.2 Basic Quantities
Definitions
and
6.2.1 Absorbed Dose (D)
In a lot of publications in nuclear medicine, the term “dose” is used for describing the adminis­tered activity as well as the “absorbed dose.” It is, however, mandatory to distinguish between these two terms. The appropriate term for the quantity of interest in radiation protection and dosimetry, however, is “absorbed dose” (D), expressed in units of gray (Gy) [1].
According to ICRP103 [1], the absorbed dose is defined as the mean energy dε imparted to mat­ter of mass dm by ionizing radiation:
e
d
D
=
(6.1)
The SI unit for absorbed dose is joule per kilo­gram (J kg−1), and its special name is gray (Gy) [1]. In nuclear medicine, dε represents the num­ber of radionuclide disintegrations in a particular volume multiplied by the energy emitted per dis­integration of the radionuclide and multiplied by the fraction of emitted energy that is absorbed by a particular (target) mass.
The concept of absorbed dose is only applica­ble in a macroscopic scale as it is generally done in nuclear medicine dosimetry. In “macrodosim­etry”, one considers mean parameters (mean doses). It should be noted that the mean absorbed dose can be calculated for a large (i.e., organ) volume or at the microscopic level, as long as the criteria applicable to macrodosimetry are met. The energy deposition, however, is a stochastic process and shows inherent statistical fluctua­tions. If particle flux-and energy deposition-is large enough, then the mean absorbed dose is rel­evant as the standard deviation due to stochastic fluctuations is small.
6.2.2 Dosimetry in Nuclear
Medicine
Methods for calculating the absorbed dose from administration of a radiopharmaceutical were
first standardized in the 1960s by the Medical Internal Radiation Dosimetry (MIRD) commit­tee, with the initial aim of estimating average doses to critical organs resulting from diagnostic procedures [
2]. Essentially this methodology
allows the calculation of absorbed dose using the simplified version of the basic equation [3]:
∼
A
h
(6.2)
Dr r
¬
Dr rS
←
()
kh
: the mean absorbed dose to a target region rk from the cumulated activity in source region rh.
Ãh: the cumulated activity (i.e., the integral of the
activity-time curve from zero to infinity) in a given target region rh.
Sr r
¬
: the radionuclide-specific S factor for target region rk and source region rh per unit cumulated activity in source region rh.
à denotes the total number of radioactive
decays occurring within an organ in which a radiopharmaceutical accumulates (the “source organ”). The MIRD S factor accounts for the energy released from each radioactive decay and the relative geometry of the source organ and the organ for which the absorbed dose is to be calcu­lated. Thus, the cumulated activity is dependent on biological parameters, while the S factor deals with the physical components of the absorbed dose.
There is no assumption made concerning the
source or target, other than that the radioactive distribution is homogeneous in the source h: the source and target can be of any size or composi­tion. Theoretically, if the activity in the source has a heterogeneous distribution, it is possible to subdivide the source into smaller volumes in which the activity can be considered to be homogeneous.
As the cumulated activity constitutes the sum
of all radioactive emissions during the time con­sidered, the energy deposition rate with time is not taken into account, although the absorbed dose rate (Gy s
−1
) is known to impact on the bio­logical consequences of the irradiation. The rele­vance of this is indicated by the fact that the aim
EH
TT
å
HH
At
r
6 Radiation Safety and Dosimetry
https://t.me/med1917
105
of most dosimetric studies is to relate a physical parameter (i.e., energy absorbed per unit mass) to the observed biological effect: the relation between these two parameters may not be simple, and scientifically sound dosimetric protocols are simply a prerequisite for subsequent radiobio­logical studies.
6.2.3 Effective Dose (E)
For a risk assessment of medical diagnostic pro­cedures involving ionizing radiation and radia­tion protection, the concept of the effective dose (E) has been widely adapted [1, 4]. The risk asso­ciated with the effective dose is based on assump­tions such as the concept of considering the risk to the general public or to workers. This does not necessarily reflect the situation for patients in nuclear medicine as, particularly in patients with cancer, the life expectancy is likely to be different of that of the general public or of workers. Another aspect is the strong age and sex depen­dency of the radiation risk, which is not included in the effective dose. Therefore, the effective dose should not be used for individual risk­benefit assessments in patients; instead, the rele­vant quantity is the equivalent dose or the absorbed dose to irradiated organs. However, for comparing different medical procedures, effec­tive dose is a useful quantity [1]. It is defined as
=
w
T
(6.3)
F
+
E
=×
å
6.1 summarizes the tissue weighting fac-
Table
T
TMT
w
T
2
(6.4)
tors according to ICRP103.
According to ICRP103 [1], the occupational exposure in planned exposure situations should be limited to an effective dose of 20 mSv per year, averaged over defined 5-year periods (100 mSv in 5 years), with the further provision that the effective dose should not exceed 50 mSv in any single year. For public exposure in planned exposure situations, ICRP recommends that the limit should be expressed as an effective dose of 1
mSv in a year. The annual exposure of the hands and feet should not exceed 500 mSv (organ dose) in occupational exposure; no values are provided by the ICRP for the general public.
6.2.4 External Exposure
The absorbed dose D imparted by the external exposure of a point source and by penetrating radiation, an activity A at a distance r, an expo­sure rate constant Γ (dependent on the isotope), and an exposure duration Δt is described by Eq. 6.5:
Γ⋅ ⋅∆
D
=
2
An overview of the most common exposure rates, exposure durations, distances, activities to be
(6.5)
where ωT is the tissue weighting factor for the tis­sue T and represents the radiation sensitivity for stochastic radiation damage of each tissue con­sidered. The sum over all tissue weighting factors is 1: ΣT ωT = 1. HT is the organ dose (unit: Sv) caused by internal or external exposure. For nuclear medicine applications using beta and gamma emitters, the organ doses are calculated by multiplying the absorbed dose (unit: Gy) by the radiation weighting factor for photons and electrons (=1 Sv/Gy). According to ICRP103 [1], the effective dose E is then the arithmetic mean of the gender-specific organ doses H
F
and H
(female):
T
M
(male)
T
Table 6.1 Tissue weighting factors according to ICRP103 [1]
Tissue ω Bone marrow (red), colon, lung,
stomach, breast, remainder tissues Gonads 0.08 0.08 Bladder, esophagus, liver, thyroid 0.04 0.16 Bone surface, brain, salivary glands,
skin
a
Remainder tissues: adrenals, extrathoracic (ET) region, gall bladder, heart, kidneys, lymphatic nodes, muscle, oral mucosa, pancreas, prostate (male), small intestine, spleen, thymus, uterus/cervix (female)
a
0.12 0.72
0.01 0.04
Total 1.00
ΣT ω
T
T
106
M. Lassmann and U. Eberlein
https://t.me/med1917
Table 6.2 External exposure rates and exposures for typical SLN applications
Isotope Tc-99m In-111 F-18 Co-57 Exposure rate 22 Typical activity 10 Exposure for 2
at a distance of
0.3 m
Data taken from the radionuclide and radiation protection handbook [ *Similar exposure rate constants for many other PET nuclides **4 h after administration of 300 MBq F-18-FDG
h
0.005
2
/(GBq h) 87 μSv m2/(GBq h) 37 μSv m2/(GBq h)* 15 μSv m2/(GBq h)
μSv m MBq 150 MBq 75 MBq <1 MBq
mSv 0.03 mSv 0.1 mSv** <0.0003 mSv
5]
expected, and potential exposure scenarios and exposures in SLN procedures is provided in Table 6.2 (data taken from Delacroix et al. [5]). With the exception of the application of PET nuclides and In-111, the exposure scenarios do not exceed 0.06 mSv for the hands and 0.005 mSv for body exposure. For comparison, the total effective dose to individual members of the pub­lic should not exceed 1 mSv per year [1]. Concerning contamination of the hands, the dose rate after contamination with a droplet of 0.05 ml and an activity of 1 kBq leads to an exposure dose rate of 9 μSv/h, 0.06 mSv/h, and 0.8 mSv/h for Tc-99m, In-111, and F-18, respectively [5].
Reported radiation exposures for Tc-99m­based procedures were less than 0.4 mSv for the surgeon hands and 0.02 mSv for the surgeon body [6]. In procedures involving Tc-99m and in typical quality assurance procedures, therefore, the exposure by external irradiation is low. Special precaution measures, however, might need to be considered when applying SLN diag­nostics after the use of PET-tracers such as F-18.
6.3 Patient Dosimetry for SLN
Diagnostics with Tc-99m
6.3.1 Patients
As has been stated in the EANM-EORTC general recommendations for sentinel lymph node diag­nostics in melanoma [7], lymphoscintigraphy is a procedure involving low activities. The estimated local radiation dose varies depending on the administered activity, injection site, volume of tracer, the application of multiple injections, and
retention time [
8]. The different radiopharmaceu-
ticals used for SLN imaging are associated with minor differences in dosimetry. The local absorbed dose at the injection site with respect to the most common radiocolloids is less than 50
mGy/MBq [8–10].
For breast cancer, Waddington et al. estimated the mean absorbed dose to the tissue of the affected breast to be 0.72 mGy/MBq resulting in an absorbed dose of 11 mGy for a tracer administra­tion of 15 MBq [8]. The sentinel lymph node itself will receive an absorbed dose between 0.44 and
2.5 mGy/MBq depending on the size of the lymph node, assuming 1.0 % uptake of tracer and physi­cal half-life. Taking all other organs that are also irradiated into account, Waddington et al. estimate the effective dose for SLN diagnostics in breast cancer to 0.3 mSv for an injection of 15 MBq Tc-99m-colloids [8]. For this procedure, dosime­try data were also published by Law et al. [11]. These authors report an effective dose of 5.1 μSv/ MBq, a value which is significantly lower than the
μSv/MBq reported by Waddington et al. [8].
21 Values from Waddington et al. are also systemati­cally higher with respect to the upper limit of breast absorbed dose (720 vs 35 μGy for the injected breast). Cremonesi et al. [12] estimated a mean of 0.8 mGy/MBq at the injected breast and
0.05 mGy/MBq at the sentinel lymph node. Extensive calculations performed at the Memorial Sloan Kettering Cancer Center have confirmed the safety of the procedure for breast cancer by report­ing an effective dose around 0.2 mSv [13].
Melanoma originates from skin tissue that is relatively less radiosensitive than many other tissues; the tissue weighting factor defined by the International Committee of Radiation Protection
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