IAEA Regional Training Course
on
ASSESSMENT
OF
OCCUPATIONAL EXPOSURE
DUE TO
INTAKES OF RADIONUCLIDES
Measurement of whole body activity in
scanning bed geometry
Introduction
Direct measurement of internal contamination should provide identification of the
radionuclide(s) present in the worker and determine the quantity (activity) of those
radionuclides. However, direct measurements are only useful if the emission from the
radionuclide deposited in the body can be detected with external detectors. This depends on
the abundance and energy of the photons emitted and on the pattern of deposition in the body.
In many cases, the radionuclide is localized to one organ or body tissue, so that the direct
measurement should focus on determination of the activity level in that organ rather than the
whole body. Metabolic models and assumptions about the time of intake are used to
determine the magnitude of the intake. The internal dose assessment will then be based on
this quantity.
Objective
The objective of this exercise is to demonstrate the steps involved in 1) calibration and
2) use of a counting system for identification and measurement of uptake of a radionuclide in
the whole body, and 3) use of the uptake information to determine the intake quantity and the
committed effective dose that results from that intake.
This information will be useful to those concerned with the planning, management and
operation of occupational monitoring programmes including to those responsible for carrying
out individual dosimetry due to intakes of radionuclides based on indirect measurements of
internal contamination.
Scope
The exercise is intended to familiarize the students with the procedures used for whole
body radioactivity measurements, including energy calibration, efficiency calibration,
background measurements, preparation of subjects to be monitored, execution of human
measurements, gamma spectrometric evaluation of measurements and recording the results.
The exercise will focus on the conduct of phantom measurements and data evaluation by
identification of radionuclides incorporated in the phantom and determination of their
activities. Following identification of the radionuclide and determination of the quantity in
the whole body, data on simulated intake will be provided to enable the students to determine
the intake and associated committed effective dose.
Measuring Equipment
Commercial counting systems such as those from Canberra may be used for this
exercise, however the primary components are as follows:
Counting Geometry:
Scanning bed or fixed bed with scanning detectors. A fixed bed with a fixed
detector or detector array could also be used. The bed should be located in a shielded
room, or least be part of a well shielded shadow shield counter.
Detectors:
Sodium iodide or germanium detectors of a suitable size for whole body
measurements. If a single sodium iodide detector is used, it should be of a size to
offer adequate sensitivity – 203 mm diameter  101 mm is recommended. The
sodium iodide should be 76.2 mm x 76.2 mm or less to allow placement near the
thyroid. A germanium detector should have a similar size limitation. Germanium
detectors should also be of a sufficient size to assure adequate sensitivity – 25%
relative efficiency or greater. If a commercial counting system such as that made
by Canberra is used, the germanium detectors supplied with the system should be
adequate.
Electronics:
Multichannel pulse height analyzer (MCA) with sufficient number of channels to
accommodate high resolution spectra from a germanium detector if used. The use
of computer based MCAs facilitates both basic manipulations of spectra and
sophisticated operations such as peak search, evaluation, identification and
calculation procedures.
Ancillary equipment required to extract, amplify and sort electrical signals from the
detector, producing a pulse-height distribution (spectrum), including: pre and main
amplifiers, analogue-digital converter (ADC) and modules providing DC power
supplies (low and high voltage as well).
Phantom
The phantom should be suitable for simulation of distribution of radioactivity in the
whole body, including head, neck and lower limbs. Although BOMAB phantoms probably
are the most commonly used, there are other phantoms that would be suitable. The phantom
may be as simple as a set of 1 L plastic bottles filled with gels containing the radionuclide(s),
with overall dimensions of an adult individual. A few such phantoms are illustrated below.
Outline of the Practical
1. Facility familiarization – The facility staff will provide a brief over view of the
facility to be used for this exercise, including description of the counting equipment,
counting geometry and phantom to be used. If a commercial counting system with
data processing software is used, the familiarization should include a description of
that software that is sufficient to allow the student to understand its use during the
exercise.
2. Energy calibration - The facility staff will demonstrate the energy calibration of the
system to be used for the practical, with student participation if appropriate. The
demonstration should include description of the source(s) used and the major
photons emitted discussion of source positioning, source measurement, and
description of the software used, if relevant.
3. Efficiency calibration – The facility staff will explain how the efficiency calibration
of the counting system has been performed with the phantom to be used in the
practical. If possible, a demonstration should be conducted using a filled phantom.
The explanation and/or demonstration should include description of the
radionuclides used for calibration, the basis for certification of the source activity
(i.e. traceability), and measurement of the source. The demonstration should
include description of how the photopeaks are evaluated, with particular emphasis
on software processing if a commercial system is used. The facility staff should
provide the students with energy dependent efficiency curves for the phantom used
in the practical. These curves should cover the energy of the radionuclide to be
evaluated.
4. Measurement of the “unknown” – The facility staff will demonstrate loading of the
phantom with the unknown radionuclides(s) to be used for the practical. They will
demonstrate positioning of the phantom with emphasis on methods used to achieve
reproducible positioning. Each student will perform a measurement of the phantom.
All measurements should be made with supervision of the facility staff, but the
students should do as much of the “hands-on” work as possible, consistent with
facility rules and safety.
5. Radionuclide identification – Following the measurement, the students will first
evaluate the counting results to determine the energy of the photopeaks and identify
the “unknown” radionuclide(s). This will require determining the energy of the
photopeaks from the spectra and the energy calibration conducted earlier. From this
information and radionuclide decay data to be provided by the facility staff, the
students should identify the radionuclide. If a commercial counting system is used,
the identification should also be done (with guidance from facility staff) through
peak search and built-in radionuclide libraries.
6. Radionuclide quantification – After identification of the radionuclide, the students
will determine the net count rate in the primary photopeak(s) using the same
techniques that were illustrated during the efficiency calibration. Then, using the
efficiency calibration data and the measured phantom-detector separation, the
students will determine the emission rate of photons in the primary photopeak(s).
Finally, the students will determine the activity of the radionuclide using decay data
proved by the facility staff. If a commercial counting system is used, quantification
of the radionuclides should also be done using the system software.
7. Determination of intake – Following identification and quantification of the
radionuclide(s), the facility staff will provide the students with simulated data on
intake, i.e. mode of intake, pattern of intake (acute or chronic) and date. Additional
relevant information that may be needed. Using retention data from ICRP 78 or a
similar reference, the students will determine the estimated intake value (Bq).
8. Determination of Committed Effective Dose (C.E.D.) – The students will use the
intake value determined in step 7, together with dose coefficients from ICRP 78 or
similar reference to determine the committed effective dose value that would be
assigned to the measurement results.
9. Discussion – Following completion of the practical steps 1-8, the facility staff and
lecturers will conduct a discussion to review the experience of the practical,
including problems encountered, questions that the students may have and key
points that the practical has illustrated.
Notes to the Facility Staff and Lecturers
This practical is intended to provide an illustration of the steps involved in system
calibration, subject measurement and data evaluation for a whole body measurement. It is
recommended that more than one unknown radionuclide be used, but no more than three. The
most likely unknown radionuclides for this purpose are fission or activation products. One
should have a gamma ray between 100 and 200 keV.
Emphasis should be given in this practical on the principles involved in routine whole
body counting for occupational protection. Manual evaluation procedures (peak identification
and quantification) have the value that they illustrate the principles behind the measurement
process and help the student better appreciate the advantages (and pitfalls) of more automatic,
software based methods. These points should be stressed. Other issues related to operation of
a direct measurement facility should also be highlighted during the demonstration, including
housekeeping, quality control, etc.
The scenario developed as a basis for intake and committed effective dose
determination from the counting results should be realistic, but not excessively complicated.
It should illustrate the principles involved in making this determination. A scenario based on
a case history would be useful.
References
INTERNATIONAL ATOMIC ENERGY AGENCY, Directory of Whole-Body Radioactivity
Monitors, IAEA, Vienna (1970)
INTERNATIONAL ATOMIC ENERGY AGENCY,
Contamination in Man, IAEA-SM-276, Vienna (1985)
Assessment
of
Radioactive
INTERNATIONAL ATOMIC ENERGY AGENCY, Rapid monitoring of large groups of
internally contaminated people following a radiation accident, IAEA-TECDOC-746, Vienna
(1994)
INTERNATIONAL ATOMIC ENERGY AGENCY, Direct Methods for Measuring
Radionuclides in the Human Body, Safety Series No. 114, IAEA, Vienna (1996)
INTERNATIONAL ATOMIC ENERGY AGENCY, Calibration of Radiation Protection
Monitoring Instruments, Safety Report Series No. 16, IAEA, Vienna (2000).
INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTECTION, Individual
Monitoring for Internal Exposure of Workers, ICRP Publication 78 (1997).
INTERNATIONAL COMMISSION ON RADIATION UNITS AND MEASUREMENTS,
Direct Determination of the Body Content of Radionuclides, ICRU Report 69, Journal of the
ICRU, 3, No.1 2003 (2003).