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