Measurement of Tritium by Liquid Scintillation Counting

IAEA Regional Training Course
Laboratory Practices Module VI.2/1
Determination of tritium in urine by liquid
scintillation counting
Tritium is encountered in several occupational environments. Tritium decays to helium
by emitting beta particles with a maximum energy (Emax) of 18.6 keV. Its half-life is 12.35
years. Due to the very low energy of the beta radiation of tritium, the liquid scintillation
counting is an ideal method to determine its activity concentration in a given sample. The
preferred sample for bioassay is urine.
Liquid scintillation counting is based on the conversion of the kinetic energy of beta
particles into light photons. In the sample vial, the scintillation cocktail converts the radiation
emitted from the radionuclide into light pulses. The liquid scintillation analyser detects the
light photons when they enter the photomultiplier tubes. Samples often contain materials that
interfere with the radiation detection process – quenching agents. Two types of quenching
can occur. First, a chemical quenching agent interferes with the production of photons in the
scintillator. Second, a colour-quenching agent hinders the detection of the photons. Both types
of quenching will decrease the counting efficiency. In general, for quantitative measurements
corrections must be made for quenching. However, when counting tritium in the form of
HTO, the sample can be distilled. The tritiated water follows the H2O and the resultant
purified sample does not require quench correction.
FIG. 1. A typical quench curve
When counting low energy betas, it is necessary to use high amplification and voltage to
make the pulses detectable. As a result, the photomultiplier noise will be high. For low energy
beta-emitters, the amplitude of signals is in the range of the noise impulses. Two or more
photomultiplier are operated in coincidence to decrease the noise. The noise in the different
photomultipliers is independent, so noise counts will only occur from random coincidence,
significantly reducing the noise contribution.
The purpose of this exercise is to demonstrate the steps and procedures involved in
performing a urine bioassay by liquid scintillation counting. This includes determination of
counting efficiency, sample preparation, background measurement, sample counting,
performing quenching correction, and determination of the tritium concentration in the
sample. It will also demonstrate the use of the sample activity concentration, together with a
worker intake scenario to determine tritium intake and committed effective dose from the
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.
The exercise intends to familiarize the students with the procedure applied in excretion
analysis of tritium. The exercise will focus on determination of tritium in urine by liquid
scintillation analysis. The exercise will address sample collection, sample preparation
(physical treatment, chemical separation and preparation for counting), radioactivity
measurements, data evaluation and recording the results. The exercise will not deal with other
biological (faeces, blood, etc.) and physical (air, swipe, etc.) samples, but the measurement
procedure can be also applied to these cases following the necessary modification.
Outline of the Practical
1. Facility familiarization – The counting facility staff will provide an overview of the
use of the liquid scintillation counter, specific calibration details and sample
preparation information.
2. Calibration – The calibration is performed by preparing one or more standard
samples. Each student will participate in standard sample preparation. The
standards will counted together with background samples and the bioassay samples.
3. Sample preparation– The following is a suggested procedure for sample preparation.
It may be modified to be consistent with counting facility procedures, if necessary.
Each student should prepare at least one sample and one standard
a. Sampling - Collect and homogenise about 1 L of urine (it is about the nominal
daily output of urine from human body) into cleaned and dried glass or
polyethylene flask. Close the flask well and store it in refrigerator, if it is
b. Sample preparation - Since the urine samples could contain other
radionuclides and inorganic and organic chemicals that cause quenching, it is
necessary to introduce a distillation step to remove contaminants. Several
different techniques are used to recover a pure distilled fraction: (i) normal
distillation, (ii) azeotropic distillation, (iii) vacuum distillation.
c. Background samples – Background samples are prepared in adding distilled
water to the scintillator in volumes equal to that used for the bioassay samples.
4. Measurement - After distillation, an aliquot is mixed with liquid scintillation
cocktail and counted in a liquid scintillation spectrometer. Standards and
background samples are prepared and counted with each group of samples. The
tritium activity concentration in the sample is determined from
 R  R0  t
c= 
   V 
c is the tritium activity concentration of the sample, in becquerels per L
R is the counting rate of the sample in counts per second,
R0 is the counting rate of the blank in counts per second,
 is the counting efficiency,
V is the volume of sample in the counting vial in L,
 is the decay constant of tritium in reciprocal years ( = 0.05576)
t is the interval between sampling and counting in years.
If t <0.5 year, as in the normal practice of occupational monitoring, the last factor
of the equation can be ignored.
5. Determination of intake – Following quantification of the tritium, 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).
6. Determination of Committed Effective Dose (C.E.D.) – The students will use the
intake value determined in step 5, 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.
7. Discussion – Following completion of the practical steps 1-6, 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 sample
preparation and counting by liquid scintillation. Emphasis should be given in this practical on
the principles involved in routine tritium bioassay for occupational protection. Other issues
related to operation of an analytical laboratory 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.
Contamination in Man, IAEA-SM-276, Vienna (1985)
OFFICE, Occupational Radiation Protection, Safety Standards Series No. RS-G-1.1, IAEA,
Vienna (1999)
OFFICE, Assessment of Occupational Exposure due to Intakes of Radionuclides, Safety
Standards Series No. RS-G-1.2, IAEA, Vienna (1999)
INTERNATIONAL ATOMIC ENERGY AGENCY, Indirect Methods for Assessing Intakes
of Radionuclides Causing Occupational Exposure, Safety Reports Series No. 18, IAEA,
Vienna (2000)
INTERNATIONAL ATOMIC ENERGY AGENCY, Calibration of Radiation Protection
Monitoring Instruments, Safety Report Series No. 16, IAEA, Vienna (2000)
Monitoring for Intakes of Radionuclides by Workers: Design and Interpretation, Publication
54, Ann. ICRP 19 1-3, Pergamon Press (1988)
Monitoring for Internal Exposure of Workers (Replacement of ICRP Publication 54),
Publication 78, Ann. ICRP 27 3-4, Pergamon (1997)
of Bioassay Procedures for Assessment of Internal Radionuclide Deposition, Rep. 87, NCRP,
Bethesda, MD (1987)
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