international atomic energy agency individual dose assessment assessment of occupational exposure...

International Atomic Energy Agency

Individual Dose Assessment

ASSESSMENT OF OCCUPATIONAL EXPOSURE DUE TO INTAKE OF

RADIONUCLIDES


Individual Dose Assessment – Unit Objectives

The objective of this unit is to provide an overview of the use monitoring measurement to assess the exposure from internally deposited radionuclides. It includes a discussion of the use of material and individual specific data to improve dose estimates, and the role of task and special monitoring in assessment of internal exposure.At the completion of the unit, the student should understand the principles involved in dose assessment, and how to apply these principles.


Individual Dose Assessment – Unit Outline

Introduction Need for Monitoring Routine Monitoring Programme Design Methods of Measurement Monitoring Frequency Reference Levels Use of Material & Individual Specific Data Task Related Monitoring Special Monitoring


IntroductionIntroduction


Monitoring objective

The general objective of operational monitoring programmes is the assessment of workplace conditions and individual exposures

The assessment of doses to workers routinely or potentially exposed to radiation through intakes of radioactive material constitutes an integral part of any radiation protection programme and helps to ensure acceptably safe and satisfactory radiological conditions in the workplace


Individual monitoring methods

Individual monitoring for intakes is done by: Direct methods

Whole body counting Organ counting (e.g. thyroid or lung

monitoring) Indirect methods

Analysis of samples of excreta Analysis of selected body fluids or tissues Personal air samplers is also used


Workplace monitoring

Workplace monitoring is used in many situations involving radionuclide exposure

May be used to demonstrate satisfactory working conditions or where individual monitoring may not be sufficient

May be appropriate when contamination levels are low, for example in a research laboratory using small quantities of radioactive tracers


Monitoring techniques for internal dose estimation

Monitoring for radionuclide intake dose estimation may include one or more techniques:

Sequential measurement of radionuclides in the whole body or in specific organs;

Measurement of radionuclides in biological samples such as excreta or breath;

Measurement of radionuclides in physical samples such as filters from personal or fixed air samplers, or surface smears.


Determination of committed effective dose

Measurements are used to determine intake The intake, multiplied by the dose coefficient,

gives an estimate of committed effective dose Dose coefficients have been calculated by the

ICRP and are given in the BSS In some situations, direct measurements may

be used to determine whole body or individual organ dose rates directly


AMAD (activity median aerodynamic diameter)

The aerodynamic diameter of an airborne particle is the diameter that a sphere of unit density would need to have in order to have the same terminal velocity when settling in air as the particle of interest.

The thermodynamic diameter is the diameter that a sphere of unit density would need to have in order to have the same diffusion coefficient in air as the particle of interest.


DOSE COEFFICIENTS FOR SELECTED RADIONUCLIDES

Radionuclide

Inhalation Ingestion

Type /form (a)e(g)inh (Sv/Bq)

f1 e(g)ing (Sv/Bq)AMAD = 1μm AMAD = 5μm

H-3 HTO (c) 1.8 E-11(b) 1 1.8 E-11

OBT 4.1 E-11(b) 1 4.2 E-11

Gas 1.8 E-15(b)

C-14 Vapour 5.8 E-10(b) 1 5.8 E-10

CO2 6.2 E-12(b)

CO 8.0 E-13(b)

P-32 F 8.0 E-10 1.1 E-09 0.8 2.3 E-10

M 3.2 E-09 2.9 E-09

Fe-55 F 7.7 E-10 9.2 E-10 0.1 3.3 E-10

M 3.7 E-10 3.3 E-10

Fe-59 F 2.2 E-09 3.0 E-09 0.1 1.8 E-09

M 3.5 E-09 3.2 E-09

Co-60 M 9.6 E-09 7.1 E-09 0.1 3.4 E-09

S 2.9 E-08 1.7 E-08 0.05 2.5 E-09

Sr-85 F 3.9 E-10 5.6 E-10 0.3 5.6 E-10

S 7.7 E-10 6.4 E-10 0.01 3.3 E-10


Need for Monitoring


Designation of workplace areas

Determination of the need for monitoring begins with designation of workplace areas

Supervised areas

Controlled areas

Area designation is based on knowledge of workplace conditions and the potential for worker exposure


Designation of workplace areas

A worker should be enrolled in an internal exposure monitoring programme when there is a likelihood of an intake that exceeds a predetermined level

Guidance on the designation of areas is given in the Guide on Occupational Exposure

If operational procedures are set up to prevent or reduce the possibility of intake, a controlled area will, in general, need to be established


Establishing the need for monitoring

Individual or area monitoring need depends on: Amount of radioactive material present Radionuclide(s) involved Physical and chemical form Type of containment used Operations performed and General working conditions


Establishing the need for monitoring

Examples:

Workers handling sealed sources, or unsealed sources in reliable containment, may need to be monitored for external exposure, but not necessarily for internal exposure

Workers handling radionuclides such as tritium, I-125 or Pu-239 may need monitoring for internal exposure, but not for external exposure


To monitor or not to monitor?

The decision to conduct intake monitoring may not be simple

Routine monitoring only for: Workers in controlled areas Contamination control and When significant intakes can be expected

From experience, if a C.E.D. > 1 mSv is unlikely, Individual monitoring may be unnecessary Workplace monitoring may be in order


Situations that may call for monitoringSituations that may call for monitoring

Some situations where routine individual monitoring may be appropriate include:

Handling of large quantities of gaseous or volatile materials, e.g. 3H and its compounds in;

Large scale production processes Heavy water reactors and Luminizing;

Processing of plutonium and other transuranic elements;



Mining, milling and processing of thorium ores Use of thorium and its compounds – can lead

to exposure from radioactive dusts, and thoron (Rn-220) and its progeny);

Mining, milling and refining of high grade uranium ores;

Processing of natural and slightly enriched uranium, and reactor fuel fabrication;



Bulk production of radioisotopes; Working in mines and other workplaces where

radon levels exceed a specified action level; Handling radiopharmaceuticals, such as I-131

for therapy, in large quantities; Reactor maintenance exposure due to

fission and activation products


Individual vs. Workplace monitoring

Individual monitoring may not be feasible for some radionuclides because of: Radiation type(s) emitted and Detection sensitivity of monitoring methods

In such situations, reliance must be placed on workplace monitoring

However, for some radionuclides, e.g. 3H, individual monitoring may be more sensitive than workplace monitoring


Monitoring for new operations

Individual monitoring is likely to be needed for new operations

As experience in the workplace is accumulated, the need for routine individual monitoring should be kept under review

Workplace monitoring may be found to be sufficient for radiological protection purposes


Routine Monitoring Programme Design


Consider monitoring limitationsConsider monitoring limitations

Monitoring conducted on a fixed schedule for selected workers is routine monitoring

Internal exposure monitoring has several limitations

These limitations should be considered in the design of an adequate monitoring programme

Monitoring does not measure directly the committed effective dose to the individual


Internal exposure monitoring limitations

Monitoring does not measure directly the committed effective dose to the individual

Biokinetic models are needed to: determine activity in the body from excreta

sample activity levels, determine intake from body content, calculate the committed effective dose

from the estimated intake


Further internal monitoring limitations

Measurements may be subject to interference from other radionuclides present in the body:

Natural 40K present naturally

Cs-137 from global fallout

Uranium naturally present in the diet

Radiopharmaceuticals administered for diagnostic or therapeutic purposes


Interference from “background” radionuclides

Establish the radionuclide body content from previous intakes

Particularly important when the non-occupational intakes are elevated, e.g. in mining areas high domestic radon exposure

Workers should have bioassay measurements before working with radioactive materials to establish a ‘background’ level.


Interference from Radiopharmaceuticals

Radiopharmaceuticals can interfere with bioassays for some time after administration

Duration of interference depends on: Properties of the agent administered and Radionuclides present at the workplace

Request workers to report administration of radiopharmaceuticals

It can then be determined if adequate internal monitoring can be performed



The results of an individual monitoring programme for the estimation of chronic intakes might depend on the time at which the monitoring is performed

For certain radionuclides with a significant early clearance component of excretion, there may be a significant difference between measurements taken before and after the weekend



For nuclides with long effective half-lives, the amount present in the body and the amount excreted depend on, and will increase with, the number of years for which the worker has been exposed

In general, the retained activity from previous years’ intakes should be taken to be part of the background for the current year


Timing of measurements is important

Results for the estimation of chronic intakes can depend on when the monitoring is done

If radionuclides have a significant early clearance, difference between pre- and post-weekend measurements may be significant

These should be reviewed individually if chronic exposure is possible


Timing of measurements is important

If nuclides have long effective half-lives, Amount present in the body and Amount excreted

depend on the number of years for which the worker has been exposed

These amounts may increase with exposure Retained activity from previous years’ intakes

should generally be taken to be part of the background for the current year



The analytical methods used for individual monitoring sometimes do not have adequate sensitivity to detect the activity levels of interest


Air and surface monitoring

Analytical methods may not have adequate sensitivity

A system of workplace and personnel monitoring may be needed to determine radionuclide intake quantities

Fixed or personal air samplers (PASs) may be used to determine the airborne concentrations of radioactive material


Air and surface monitoring

Air sampling results, together with standard or site specific assumptions: Physical and chemical form of the material Breathing rate and Worker exposure time

to estimate inhalation intakes Surface monitoring may also indicate intake

potential or need for detailed area monitoring But, models for estimating intake from surface

contamination are particularly uncertain


Methods of Measurement


Direct vs. Indirect measurements

Radionuclide intake can be determined by either direct or indirect measurement methods

Direct measurement of photons is also referred to as body activity measurements, whole body monitoring or whole body counting

Indirect measurements include activity in either biological or physical samples

Each type has advantages and disadvantages The selection of over than another depends on

the nature of the radiation to be measured


Direct measurements

Direct methods are useful only for those radionuclides which emit photons: Of sufficient energy, and In sufficient numbers, To escape from the body and Be measured by an external detector

Direct measurements are particularly useful for fission and activation products


Direct measurements Radionuclides which do not emit energetic

photons (e.g. 3H, 14C, 90Sr-90Y) can usually be measured only by indirect methods

Pu-239 emits weak x-rays and may be measured by either method

Some higher energy beta emitters, e.g. 32P or 90Sr-90Y, can sometimes be measured ‘directly’ via the bremsstrahlung produced

These measurements have a relatively high minimum detectable activities and are not usually employed for routine monitoring


Direct measurements

Direct measurements: Rapid Convenient Can estimate activity in the whole body or

a defined part of the body Less dependent on biokinetic models than

indirect monitoring


Direct measurements

May have greater calibration uncertainties, especially for low energy photon emitters

May require the worker to be removed from work involving radiation exposure for the period over which the retention characteristics are measured

Often need special, well shielded, and expensive facilities and equipment.


Direct measurements – Qualitative applications

Useful in qualitative and quantitative determinations of radionuclides

Can assist in identifying the mode of intake by determining the distribution of activity

Sequential measurements can reveal activity redistribution and give information about the total body retention and biokinetics


Indirect measurements

Generally interfere less with workers duties However, require access to a radiochemical

analytical laboratory Analytical laboratory may also be used for

measuring environmental samples Perform high level (e.g. reactor water

chemistry) and low level (e.g. bioassay or environmental samples) work in separate laboratories


Indirect measurements - Excreta

Excreta measurements determine the rate of loss of radioactive materials from the body by a particular route

Must be related to body content and intake by a biokinetic model

Radiochemical analyses low detection levels sensitive detection of body activity


Indirect measurements – Air samples

Can be difficult to interpret - air concentration may not represent breathing zone

Personal air sampler (PAS) placed on the worker’s lapel or protective headgear can collect more representative samples

Sample comprising only a few particles still a problem

Air concentrations + breathing rates + measured exposure times = estimated intake


Indirect measurements – Air samples

Use of PASs only estimates intake Cannot be used to refine a dose estimate based

on individual retention characteristics PAS measurements cannot be repeated Can provide intake estimates for nuclides such

as 14C (particulate), 239Pu, 232Th and 235U, when other methods may have sufficient sensitivity

Interpretation depends on the dose coefficients and the derived air concentration (DACs)


Particles size is important

Particle size influences deposition of inhaledparticulates in the respiratory tract

Correct interpretation of bioassay and dose assessment depends on particle size data

Determine airborne particle size distribution using cascade impactors or other methods

BB

bb

Al

ET2

0.1 1 10 100AMAD (m)

100

10

1

0.1

0.01

Reg

iona

l dep

ositi

on (%

) ET1


Particles size is important

Measurements should, at least, include the concentration of the respirable fraction

Some models for interpreting PAS results discriminate against non-respirable particles

Dose assessment improves with more site and material specific information


Measurement detection limits

Measurement methods have limits of detection arising from: Naturally occurring radioactive materials Statistical fluctuations in counting rates,

and Factors related to sample preparation and

analysis Minimum significant activity (MSA) and

minimum detectable activity (MDA)will be discussed in another unit


Monitoring Frequency


Individual monitoring frequency

BSS:

“The nature, frequency and precision of individual monitoring shall be determined with consideration of the magnitude and possible fluctuations of exposure levels and the likelihood and magnitude of potential exposures.”

Characterize the workplace to determine the appropriate frequency and type of monitoring!



Identify radionuclides in use and determine their chemical and physical forms

Consider possible changes of these forms under accident conditions;

e.g. the release of uranium hexafluoride into the atmosphere results in the production of HF and uranyl fluoride



Chemical and physical forms (e.g. particle size) determine material behaviour on intake and biokinetics in the body

These in turn determine the excretion routes and rates, and hence the type of excreta samples to be collected and their frequency


Proper frequency minimizes intake uncertainty

Set bioassay sampling schedules to minimize intake estimate uncertainties due to the unknown time of an intake, i.e. If acute intake occurs immediately after

previous assay, Assuming intake at the monitoring period

midpoint underestimates the intake Monitoring period should be short enough that

the underestimate factor of 3


Determining the monitoring frequency

Monitoring period, ΔT, depends on: Radionuclide retention, R(t) Radionuclide clearance, E(t) Sensitivity of the measurement process, i.e

measurement MDA Acceptable uncertainty Committed effective dose, e(50)


Determining the monitoring frequency

For in vivo measurementse(50) MDA/R(ΔT) 365/ΔT ≤ 1 mSv/year

For in vitro measurementse(50) MDA/E(ΔT) 365/ΔT ≤ 1 mSv/year

Maximum overestimation shouldn’t exceed 3 If exposure occurs at ΔT/2, this means;

R(1)/R(ΔT/2) ≤ 3E(1)/E(ΔT/2) ≤ 3


Recommended maximum time intervals for routine monitoring

In vitro measurements In vivo measurements Isotope Type Urine (days) Whole Body (days) Thyroid (days) 3H HTO 30 - - 14C Organic 30 - - Dioxide 180 32P F 30 - - 35S F 15 - - 36Cl F 30 - - 51Cr F (15) 15 - 54Mn M - 90 - 59Fe M - 90 - 57Co S (180) 180 - 58Co S (180) 180 - 60Co S (180) 180 - 89Sr F, S 60 - 90Sr F, S 180 - 110mAg S - 180 - 125I F (90) - 90 131I F (15) - 15 137Cs F (180) 180 - 147Pm S 180 - 226Ra M 180


Suggested maximum time intervals for routine monitoring for uranium compounds

Material Type* Urine (days)

Faeces (days)

Lungs (days)

Natural / Depleted U F and M 90 Uranium hexafluoride F 90 Uranium peroxide F 30 Uranium nitrate F 30 Ammonium diuranate F 30 - - Uranium tetrafluoride M 90 180 180 Uranium trioxide M 90 180 180 Uranium octoxide S 90 180 180 Uranium dioxide S 90 180 180


Suggested maximum time intervals for routine monitoring for actinide compounds

Isotope Type Urine (days)

Faeces (days)

Lungs (days)

228Th S 180 180 - 232Th S 180 180 - 237Np M 180 180 - 238Pu S 180 365 - 239Pu S 180 365 - 241Am M 180 365 180 244Cm M 180 365 -


Recommended monitoring interval tolerances

Unreasonable to expect bioassay measurements to be preformed on exact schedule

Monitoring interval - Days Tolerance - Days15 230 460 790 14

180 30365 30


Schedule to avoid missing an intake

Schedule monitoring to ensure an intake above a predetermined level is not ‘missed’

Intake could be missed if, As a result of clearance, Body content or daily excretion Declines to a level below the minimum

significant activity of the measurement During the time between intake and

measurement



m(t) - Fraction of an intake in the body (direct measurement) or being excreted from the body for indirect measurement, depends on:

Physical half-life

Biokinetics of the radionuclide, and

Is a function of the time since intake



An intake I and the resulting committed effective dose E(50) would be missed if,

I m(t) is less than the MSA

Monitoring frequency should be set so that intakes corresponding to more than 5% of the annual dose limit are not missed.


Monitoring frequency depends on sensitivity

Monitoring frequency is largely driven by the sensitivity of the measurement technique

Measurement techniques should be as sensitive as possible

However, associated costs - Most sensitive techniques Frequent monitoring measurements

should be balanced against risk doses are underestimated or missed


Additional methods for better sensitivity

Measurement method and frequency should detect intakes a specified dose limit fraction

Goal cannot be realized because: Lack of analytical sensitivity Unacceptably long counting times Short sampling intervals required for

excreta collection Additional methods – e.g. improved workplace

monitoring and personal air sampling - should be used for adequate worker protection


Use of Reference Levels


Reference levels

Reference levels are helpful in management of operations

Expressed in terms of measured quantities or other quantities to which measured quantities can be related

If exceeded, take specified action or decision Reference levels usually based on committed

effective dose E(50) for radionuclide intake


Reference levels

Appropriate dose limit fraction corresponding to each reference level should be established

Take other sources of exposure into account

Recording Levels and Investigation Levels relevant to internal contamination monitoring for occupational exposures.


Recording level

Defined as “a level of dose, exposure or intake specified by the regulatory authority at or above which values of dose, exposure or intake received by workers are to be entered in their individual exposure records”

Example - RL for a radionuclide intake set to correspond to a committed effective dose of 1 mSv (0.001 Sv) from a year’s intakes


Recording level

For N monitoring periods per year, the recording level for intake of radionuclide j in a monitoring period would be given by:

jj )g(Ne

.RL 0010


Investigation level

Is “the value of a quantity such as effective dose, intake or contamination per unit area or volume at or above which an investigation should be conducted”

Investigation level for radionuclide intake - A value of committed effective dose above which monitoring results justify further investigation

Set by management, depends on programme objectives and type of investigation to be done


Investigation level

For routine monitoring, the investigation level for a radionuclide intake is set in relation to: Type and frequency of monitoring Expected level and variability of intakes

Numerical value of the investigation level depends on conditions in the workplace

Investigation level may be set for; Individuals in a particular operation, or Individuals within a workplace without

reference to a particular operation


Investigation level – An example

Routine operation with routine monitoring IL set at a committed effective dose of 5

mSv (0.005 Sv) from a year’s intakes For N monitoring periods per year, the IL (in

Bq) for the intake of any radionuclide j in any monitoring period is:

where e(g)j is the dose coefficient for inhalation or ingestion

jj )g(Ne

.IL 0050


Derived levels

Measured quantities are radionuclide activities in the body or excreta samples

It is convenient to establish reference levels for the measurement results themselves

These are termed derived investigation levels (DILs) and derived recording levels (DRLs)

Measurement results that imply radionuclide intakes or committed effective doses at the corresponding reference levels


Derived levels

Derived investigation and recording levels are calculated separately for each radionuclide

Specific to the radiochemical form in the workplace Are a function of time since intake For the previous examples,

)t(m

)g(Ne.DIL

jj 0

0050


Derived recording level

t0 (time elapsed between intake and bioassay) is usually set as 365/2N days - assumes that intake occurs at the mid-point of the monitoring period, and

)t(m)g(Ne

.DRLj

j 00010


Derived levels

Measurement result should always be maintained in the radiation monitoring records for the workplace and for the individual

For worker exposure to external radiation or to multiple radionuclides, management may need to reduce the derived levels for individual radionuclides appropriately.


Use of Material and Individual Specific Data


Biokinetic models

Biokinetic models for most radionuclides Developed by the ICRP Use reference parameter values Are based on Reference Man data, and Observed radionuclide behaviour in

humans and animals Have been developed for defined chemical

forms of radionuclides, and Are generally used for planning purposes


Biokinetic models

Characterize particular workplace conditions to determine forms actually present

In some circumstances, the chemical or physical forms of the radionuclides will not correspond to the reference biokinetic models

Then, material specific models may need to be developed


Specific biokinetic models

For small intakes are small, i.e. a few per cent of the dose limit, reference models are probably good enough

If the intake estimate 1/4 dose limit, model parameters for; Specific material(s), and Individual(s)

may be needed for better estimate of the committed effective dose


Specific biokinetic models

Specific models can be developed from sequential direct and indirect measurements of the exposed workers

Analysis of workplace air and surface contamination samples can also assist in the interpretation of bioassay measurements

Example - Measure 241Am/ 239,240Pu from direct lung measurement of 241Am to assess plutonium intakes or inhaled particle solubility


Need for specific information

Common example – aerosol particle size a worker would likely inhale differs significantly from ICRP 5 μm AMAD default value

Fractions of inhaled materials deposited in various regions of the respiratory tract would have to be determined from the ICRP respiratory tract model, and

An appropriate dose coefficient calculated


Need for specific information

More specific information may also be needed on the material solubility characteristics

Can be obtained from experimental studies in animals or by in vitro solubility studies

Retrospective determination of particle characteristics may be difficult

Consideration should be given to obtaining material specific information when setting up worker monitoring programmes


Individual variability

There are differences between individuals in excretion rates and other biokinetic parameters for the same intake

Individual variability may be more significant than the differences between generic and individual specific biokinetic models

Excreta sample collection periods should be sufficiently long to reduce this variability, e.g. 24 hours for urine and 72 hours for faeces

Use of individual specific model parameters should be rare under routine circumstances.


Task Related Monitoring


Task related monitoring is,

Not routine, i.e. it is not regularly scheduled Conducted to provide information about a

particular operation, and give a basis for decisions on the conduct of the operation

Useful when short term procedure conditions would be unsatisfactory for long term use

Usually conducted the same as routine monitoring, unless the circumstances of the operation dictate otherwise


Special Monitoring


Special monitoring

May be necessary as a result of; Known or suspected exposures An unusual incident,

e.g. loss of containment of radioactive materials as indicated by an air or surface sample, or

Following an accident


Special monitoring

Usually prompted by a result of a routine bioassay measurement that exceeds the derived investigation level

It may also result from occasional samples such as nose blows, swipes or other monitoring.


Special monitoring

Measurement techniques for special monitoring usually the same as routine measurement

However, improved sensitivity or a faster processing time may be needed

Advise the laboratory that the sample analysis or the direct measurement has priority over routine measurements, and


Recommended methods for special monitoring after inhalation

In vitro measurements In vivo measurements nasal Urine Faeces Organ

Isotope NB EA Spot sample 24 h 72 h WB Th

3H ** ** 14C ** ** * 32P ** * 35S ** * 51Cr ** ** ** 54Mn ** ** ** ** 59Fe ** ** ** 58, 60Co ** ** ** ** 90Sr ** ** 110mAg ** ** ** ** 125, 131I ** ** ** 137Cs ** ** * ** 147Pm ** ** 226Ra ** ** Legend ** Recommended * Supplementary

NB: Nose blow EA: Expired air WB: Whole body Th: Thyroid



In vitro measurements In vivo measurements

nasal Urine Faeces Isotope Nose

blow Spot

sample 24 h 72 h Lung Natural / Depleted U ** ** ** * Uranium hexafluoride ** ** ** Uranium peroxide ** ** ** Uranium nitrate ** ** ** Ammonium diuranate ** ** ** Uranium tetrafluoride ** ** ** * * Uranium trioxide ** ** ** * * Uranium octoxide ** ** ** ** Uranium dioxide ** ** ** **



Legend NB: Nose blow EA: Expired air

In vitro measurements

nasal Urine Faeces In vivo

measurements

Isotope NB EA 24 h 72 h Lung 228Th ** ** ** ** 232Th ** * ** ** 237Np ** ** ** 238Pu ** ** ** 239Pu ** ** ** 241Am ** ** ** ** 244cm ** ** **


Special monitoring

The frequency of follow-up monitoring may be changed

Inform the laboratory that samples may have a higher than normal level of activity

The measurement technique can be tailored to the special monitoring situation, and

Necessary precautions may be taken to prevent contamination of other samples


ReferencesFOOD AND AGRICULTURE ORGANIZATION OF THE UNITED NATIONS, INTERNATIONAL ATOMIC ENERGY AGENCY, INTERNATIONAL LABOUR ORGANISATION, OECD NUCLEAR ENERGY AGENCY, PAN AMERICAN HEALTH ORGANIZATION, WORLD HEALTH ORGANIZATION, International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources, Safety Series No. 115, IAEA, Vienna (1996).

INTERNATIONAL ATOMIC ENERGY AGENCY, Occupational Radiation Protection, Safety Guide No. RS-G-1.1, ISBN 92-0-102299-9 (1999).

INTERNATIONAL ATOMIC ENERGY AGENCY, Assessment of Occupational Exposure Due to Intakes of Radionuclides, Safety Guide No. RS-G-1.2, ISBN 92-0-101999-8 (1999).

INTERNATIONAL ATOMIC ENERGY AGENCY, Direct Methods for Measuring Radionuclides in the Human Body, Safety Series No. 114, IAEA, Vienna (1996).

INTERNATIONAL ATOMIC ENERGY AGENCY, Indirect Methods for Assessing Intakes of Radionuclides Causing Occupational Exposure, Safety Guide, Safety Reports Series No. 18, ISBN 92-0-100600-4 (2000).

INTERNATIONAL COMMISSION ON RADIATION UNITS AND MEASUREMENTS, Direct Determination of the Body Content Of Radionuclides, ICRU Report 69, Journal of the ICRU Volume 3, No 1, (2003).

INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTECTION, Individual Monitoring for Internal Exposure of Workers: Replacement of ICRP Publication 54, ICRP Publication 78, Annals of the ICRP 27(3-4), Pergamon Press, Oxford (1997).

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