Lecture 03
Radiation Dosimetry
RCA-ASEANTOM Regional Training Course on
Rapid Radiation Measurement and Individual Dose Assessment
following Nuclear & Radiological Emergency
Office of Atoms for Peace, Bangkok, Thailand
October 7-11, 2019
Lecturer: Osamu KURIHARANational Institutes for Quantum and Radiological Science and Technology (QST)
Responses in Radiation Emergency Medicine (REM)
Tokai-mura Criticality Accident, 1999 Fukushima Nuclear Disaster, 2011
Medical responses
Diagnosis
Decontamination
Treatments
Consultation
Radiological responses
Radiation measurements
Dose assessment /Dose
reconstruction
Radiation protection
A wide variety of situations should be considered.
2
Major experiences in Japan
Classification of radiological accidents
Exposure
SubjectExternal Internal
Worker Criticality accident
Overexposure by
erroneous operation or
improper use
Accidental
contamination at
nuclear facilities or RI
handling facilities
Public Nuclear disaster (e.g., Chernobyl, Fukushima)
NR terrorism
Orphan source (e.g., Goiânia)
3
The dose assessment in radiological accidents are quite different from
that in nuclear medicine or external irradiation therapy.
Types of radiation exposure
External
exposure
Internal
contamination
/exposure
External
contamination
/exposure (skin)
Basically, no need for concerns
about secondary exposure to medical staff 4
External exposure vs. Internal exposure
外部被ばくは,線源を除去(遮へい)すれば,被ばくが…
内部被ばくは,被ばくが…
External exposure terminates when an external source is
removed or shielded.
Internal exposure continues as long as a radionuclide
exists in the body.
External exposure Internal exposure
From nose/mouth
From wound
5
External exposure vs. Internal exposure
External exposure
Internal exposure
α
α β
β
γ
γ
from some organ/tissue
where radionuclides are deposited
from some external source
α:no damage
β:give dose to the skin
γ:give dose to the
tissue or organ
α-emitting nuclides are likely
to give high internal dose
due to relatively high energy,
relatively long half life and
wR.
6
Radiation protection against external exposure
Less time spent near
source: less radiation
received
Greater distance from
source: less radiation
received
Shielding against source:
less radiation received
7
Inverse square law
d
d
d
1d
2d
3d
Intensity ∝1
d2
Applies to the force of Gravity, Light, Electric
Field, Sound, Radiation.
I1
I2=(1d/2d)2I1=(1/4)I1
I3=(1d/3d)2I1=(1/9)I1
8
Inverse square law (one example)
Distance from the precipitation tank (m)
Ne
utr
on
do
se
ra
te (
mS
v/h
)
Building
wall
Cooling tower
Uranium
Solution
Water
jacket
JAEA-Technology 2009-043 (2009)
[only in Japanese]
Relationship between neutron dose rate
and distance
~ 20 mSv/hnear cooling tower
Necessary to remove water in the
cooling water to terminate the criticality
From the scene of the Tokai-mura criticality
accident in 1999
9
Dose assessment (principles)
10
A suitable method (s) for the dose assessment should be used
depending on exposure situations (e.g., external/internal
exposure, radionuclide(s) of concern, magnitude of the dose, the
number of subjects).
As the first response, the methodology by ICRP can be applied to
the dose assessment, understanding that the result is given in the
radiation protection quantities (effective dose [Sv]).
The dose reconstruction in accidents or epidemiological studies
should be performed when radiation-induced health diseases
need to be evaluated (in that case, preferable to assess absorbed
dose [Gy]).
Effective dose
Absorbed Dose (Gy) Energy deposited per unit weight (Joule kg-1)
Tissue Equivalent Dose (Sv)
multiplied by WR
multiplied by WT and sum up
Effective Dose (Sv)
Mathematical phantomA mathematical phantom is used as a mimic for the
human body to perform calculations of doses delivered by external/internal radiations.
In the case of internal exposure, the committed effective dose (CED) as the integrated dose over a certain period of time post intake is calculated.
11
A new concept of effective dose
ICRP Publication 103
Computational phantoms
(ICRP Publication 110)
12
External dose for various irradiation geometries
Figures are taken from ICRP Publication 116. 13
Devices to measure external dose
Personal dosimeters
To measure accumulated doses during operations
Detectors to measure ambient dose rates (e.g., nSv/h, μSv/h, mSv/h)
External dose can be obtained by multiplying the readings with a working period.
14
External dose calculation
Question: A worker is exposed to a 137Cs source (37 TBq) at a
distance of 1.5 m for 10 min. How much dose does he receive?
Γ= 0.0779 (μSv m2 MBq-1 h-1) for 137Cs
E =A Γ t
d2
Activity (MBq) Time (h)
Distance (m)
=3.71070.07791/6
1.52
= 2.1105 (μSv) = 210 (mSv)
Note: For distances shorter than 0.5 m, this equation will give overestimations of the dose
15
Exposure rate constant
Radionuclide Half-lifeEnergy (MeV) and
emission rate
Effective dose rate at 1m
from 1 MBq source
(μSv h-1)
Ambient dose rate at
1m from 1 MBq source
(μSv h-1)
24Na 2.609y 1.275 – 99.9% 0.284 0.333
54Mn 312.1d 0.835 – 100% 0.111 0.13
59Fe 44.5d1.099 – 56.5%
1.292 – 43.3%0.147 0.171
60Co 5.271y1.173 – 100%
1.333 – 100%0.305 0.354
85Sr 64.84d 0.514 – 96.0% 0.0697 0.0826
110mAg 249.8d
0.658 – 94.0%
0.885 – 72.2%
0.937 – 34.1%
1.384 – 24.1%
0.354 0.416
137Cs 30.04y 0.662 – 85.1% 0.0779 0.0927
192Ir 73.83d
0.296 – 28.7%
0.308 – 30.0%
0.317 – 82.7%
0.468 – 47.8%
0.117 0.139
241Am 432.2y 0.0595 – 35.9% 0.00395 0.0529
Reference: Radioisotope pocket data book 10th edition, Radioisotope association (2003),
JAERI Data Code 2000-044 (2000) (in Japanese) 16
External exposure at very small distances
Approximate gamma dose rate to the hand from a 1Ci Sealed Source
Isotope Skin Dose Rate
(R/min)
Dose Rate at
1 cm* (R/min)
Dose Rate at
3 cm* (R/min)
Cs-137 513 28 3.7
Co-60 2075 114 16
Ir-192 813 43 5.5
Ra-226 1310 72 9.7
Approximate gamma dose rate to the hand from a 1GBq Sealed Source
Isotope Skin Dose Rate
(mGy/min)
Dose Rate at
1 cm* (mGy/min)
Dose Rate at
3 cm* (mGy/min)
Cs-137 121 6.6 0.9
Co-60 490 26.9 3.8
Ir-192 192 10.1 1.3
Ra-226 309 17.0 2.3
Note: Original data taken from Table 6 in NCRP Report 40
For converting from R to Gy, 1 R = 8.73E-03 Gy.
* Depth in tissue
* Depth in tissue
17
Is a lead apron effective for shielding?
HVL: Half-value layer, TVL :Tenth-value layer
Lead apron: 0.25mm
(Size: 58cm X 100cm)
Lead thickness (mm) Weight (kg) Effective dose reduction (Cs-137)
0.2 1.3 98%
0.5 3.3 96%
7.0 46 50%
20 132 10%
Thickness of concrete or water (cm)
Thickness of lead or iron (cm)
Eff
ective
do
se
red
uctio
n r
ate
Water
Concrete
Pb
Fe
18
Dose reconstruction for accidental external exposure
In the JCO criticality accident
A. Endo and Y. Yamaguchi, Radiation Res. (2003).
In the overexposure case with gamma radiography
FCA DaSaliva et al., J. Radiat.Prot. (2005).19
Dose reconstruction for A-bomb survivors
Fujita et al. KURRI KR-114 (http://www.rri.kyoto-u.ac.jp/IPA/DS02/KURRIKR114.html)
Individual external dose
= Ambient dose at the place of concern
x Shielding effect x Conversion factor
Whereabouts of A-bomb survivors Calculation of shielding by a house
20
Skin dose calculation
HT(skin)=Cskin CF t
SFEquivalent dose
to the skin(µGy) Shielding factor(-)
Average surface
Concentration(Bq cm-2)
Conversion
factor([nGy h-1]/[Bq cm-2])
Exposure
time(h)
Maximum β-ray energy (MeV)
CF
([nG
yh
-1]/
[Bq
cm
-2])
3000 Calculate the equivalent dose to the
skin for a person who has average
contamination of 40 Bq cm-2 of 131I
for 10 hours
Cskin= 40 (Bq cm-2)
CF = 1419 ([nGy h-1]/[Bq cm-2])
t=10 (h)
HT(skin)=40x1419x10
=567,600 (nGy)
=567 (µGy)
=0.6 (mGy)
Ref. ICRU report 56
depth at 0.07mm
21
External exposure by neutrons in criticality accident
Neutron irradiation
23Na(n,γ)24Na
24Na (radioisotope)
Photons: 1369 keV, 2754 keV
Half-life: 14 hours
Photons
Measurements of 24Na with
a Whole-Body Counter (WBC)
1 Bq of 24Na in the total body
≈ 0.5 ~ 3 μGy(neutron + gamma)
Ref. IAEA TRS-211 (1982)
Momose et al., J. Radiat. Res., 42 (2001)
The stable sodium (23Na) content is 1.4
g/kg weight. ICRP Publ.23
Germanium
semi-conductor
detector
22
Biokinetic model for internal dose calculations
Respiratory
tractGI tract
Transfer compartment (Blood)
Compartment 1 Compartment 2 Compartment j
GI tractBladder
InhalationIngestion
Early feces
Urine Feces
excretion
Flow of radionuclide
ICRP Publ.30 firstly demonstrated the general biokinetic model to calculate the effective dose. Later on, biokinetic models have been updated for some elements as well as the respiratory and alimentary tract models (Publ. 66 and Publ. 100).
Non-recyclic model
23
How to evaluate internal doses
Direct (in-vivo)
monitoring
Bioassay(excreta analysis)
Calculation from
air conc.
Activity in the
body (Bq)
Activity in
excreta (Bq)
Intake (Bq)
Intake (Bq)
Intake (Bq)
Individual
Monitoring Retention rate
Excretion rate
Monitoring
Data
E = Dose Coefficient (Sv Bq-1) Intake (Bq)
24
Individual monitoring for internal contamination
Direct monitoring (e.g., Whole-body, Thyroid, Lungs)
Bioassay (mainly analysis of excreta)
Merit: High sensitivity & Non-invasive
Demerit: Only nuclides emitting photons with detectable
energy and sufficient emission rate
Merit: Applicable for alpha/beta emitters
Demerit: Time consuming, Collecting excreta samples
25
Whole-Body Counter (WBC)
Photon
detector
γ
γ
γ
Photon
detector
Photon
detector
Photon
detector
Detecting photons emitting from the radionuclide inside the body
by the detector placed near the subject
(typical peak efficiency: ~1% for 662 keV of 137Cs)
26
Calibration using a phantom
Phantom Subject
A(Bq) ??? (Bq)
C(cps)
ε = C / ACounting efficiency
(cps Bq-1)C’(cps)
Whole-body content = C’/ ε = (C’/C) A
Attention should be paid for the body size because WBC is based on
relative measurements to the calibration phantom.
Principle of WBC
27
Typical radiochemical procedure for actinides
Bioassay
28
Internal dose calculation
Intake
(Bq)×
Dose Per
Unit Intake
(Sv/Bq)
=Committed
effective dose
(Sv)
Individual monitoring
Direct measurement :
Indirect measurement:
Body content ∕ Retention rate = Intake
Excretion amount ∕ Excretion rate = Intake
Environmental monitoring
Air concentration x Breathing volume = Intake
Concentration in food x Amount of food eaten = Intake
29
Dosimetric quantities needed for internal dose calculations
Biokinetic ModelIntake
1 Bq
InputIRF functions(Bq/Bq Intake)
DPUI
(Sv/Bq)
Output
Output
DPUI:Dose Per Unit Intake
IRF:Intake Retention Function
Dosimetric model
Models for internal dose calculations
Respiratory
tractGI tract
Transfer compartment (Blood)
Compartment 1 Compartment 2 Compartment j
GI tractBladder
InhalationIngestion
Early feces
Urine Feces
excretion
Flow of radionuclide
SEE T Sw E Y AF T S
M
R R R R
TR
30
Example of internal dose calculations
A worker inhaled radioactive dust. He was measured with a whole-
body counter the next day and 60Co of 1E+06 Bq was found.
Evaluate his effective dose.
Intake (Bq) =Measured activity (Bq)
Retention/Excretion rate(-)=
1E+06 (Bq)
0.49
= 2.04E+06 (Bq)
Effective dose (mSv) = Intake (Bq) DPUI (mSv/Bq)
= 2.04E+06 (Bq) 1.7E-05 (mSv/Bq)
= 34.7 (mSv)
From individual monitoring
From ICRP Publication 78
31
Example of internal dose calculations (cont’d)
ICRP Publication 78
Type: The category of an absorption speed in the respiratory tract (F: Fast, M: Moderate, S: Slow)
For Co, oxides, hydroxides, halides and nitrate are Type S and unexpected compounds are Type M.
32
Dose Per Unit Intake (DPUI)
Nuclide Inhalation Ingestion Injection
Type/
form
e(50)inh (Sv Bq-1)f1
e(50)ing
(Sv Bq-1)f1
e(50)inj
(Sv Bq-1)AMAD=1 µm AMAD=5 µm
60Co M
S9.610-9
2.910-8
7.110-9
1.710-8
0.1
0.05
3.410-9
2.510-9
1.910-8
106Ru F
M
S
8.010-9
2.610-8
6.210-8
9.810-9
1.710-8
3.510-8
0.05
7.010-9
3.010-8
131I F
V7.610-9
2.010-8
1.110-8
1.0
2.210-8
2.210-8
134Cs F 6.810-9 9.610-9 1.0 1.910-8 1.910-8
137Cs F 4.810-9 6.710-9 1.0 1.310-8 1.410-8
238U F
M
S
5.110-7
2.810-6
7.710-6
6.010-7
1.810-6
6.110-6
0.02
0.002
4.610-8
8.310-9
2.110-6
239Pu M
S4.710-5
1.510-5
3.210-5
8.310-6
510-4
110-5
110-4
2.510-7
9.010-9
5.310-8
510-4
4.910-4
IAEA Safety Report Series No.37 (2004)DPUI for workers
33
References for internal dose assessment
Publication Items
ICRP Publication 68 Dose coefficients for workers--- Inhalation (1 and 5 micron), Ingestion
ICRP Publication 78 Dose coefficients for workers
--- Inhalation (5 micron), Ingestion
Retention/Excretion rates up to 10 days post intake
ICRP Publication 71, 72 Dose coefficients for the public
--- Inhalation (1 micron), Ingestion
ICRP CD-ROM Dose coefficients for workers and members of the public
--- Inhalation (0.001-10 micron), Ingestion
IAEA Safety Series No.37 Dose coefficients for workers
--- Inhalation (1, 5 micron), Ingestion, injection
Retention/Excretion rates
IAEA EPR-Medical 2005 Dose coefficients for workers and members of the public
--- Inhalation (1, 5 micron), Ingestion
Retention/Excretion rates up to 10 days post intake
ICRP Publication 119 Covering ICRP CD-ROM, External dose coefficients
34
MONDAL3
MONDAL: MOnitoring to Dose CALculation
MONDAL is a software package developed by NIRS and a tool for performing internal dose
calculations on GUI. MONDAL is distributed as requested on a free of charge basis.
If you are interested, please check
http://www.nirs.qst.go.jp/db/anzendb/RPD/mondal3.php35
Retention function
1.E-08
1.E-07
1.E-06
1.E-05
1.E-04
1.E-03
1.E-02
1.E-01
1.E+00
1 10 100 1000
全身
残留
割合
摂取後の日数
3ヶ月 1才 5才 10才 15才 成人
Whole-body (WB) retention rate of 137Cs as a function of time after acute intake
(Inhalation, Type F compounds, AMAD: 1 µm)
Meas.
0.4%on 300th days
for 10y child
Intake is 250 times as large as the WB content
(1/0.004=250)
3 mo. 1y 5y 10y 15y Adult
Time after intake (day)
WB
re
ten
tion
ra
te (
-)
36
Retention function
0
20
40
60
80
100
120
140
160
180
200
0 30 60 90 120 150 180 210 240 270 300 330 360
全身
残留
量(B
q)
摂取期間(日)
3カ月児 1歳児 5歳児 10歳児 15歳児 成人
Cs-137 1日 1 Bq摂取
Annual intake = 365 Bq for all age groups
52Bq
Whole-body (WB) content of 137Cs in the case of chronic ingestion
3 mo. 1y 5y 10y 15y Adult
WB
co
nte
nt (B
q)
Time after intake (day)
Daily intake: 1Bq
37
Exposure pathways in a nuclear accident
IAEA report on Environmental consequences of Chernobyl accident
and their remediation: twenty years of experiences (2006). 38
Thank you for your attention
39