unscear 2001 exposicion ocupacional

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ANNEX E

Occupational radiation exposures

CONTENTS

Page

INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 499

I. DOSE MONITORING AND RECORDING PRACTICES . . . . . . . . . . . . . . . . 500A. QUANTITIES MEASURED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 500

1. Protection quantities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5002. Quantities for external radiation exposure . . . . . . . . . . . . . . . . . . . . 5003. Quantities for internal radiation exposure . . . . . . . . . . . . . . . . . . . . 5024. Total effective dose . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5025. Special quantities for radon . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 503

B. MONITORING PRACTICES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5031. External radiation exposure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5042. Internal radiation exposure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 505

C. DOSE RECORDING AND REPORTING PRACTICES . . . . . . . . . . . . . . 506D. CHARACTERISTICS OF DOSE DISTRIBUTION . . . . . . . . . . . . . . . . . 508E. ESTIMATION OF WORLDWIDE EXPOSURES . . . . . . . . . . . . . . . . . . 509

II. THE NUCLEAR FUEL CYCLE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 510A. URANIUM MINING AND MILLING . . . . . . . . . . . . . . . . . . . . . . . . . . . 511B. URANIUM ENRICHMENT AND CONVERSION . . . . . . . . . . . . . . . . . 513C. FUEL FABRICATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 513D. REACTOR OPERATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 514

1. Light-water reactors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5162. Heavy-water reactors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5173. Gas-cooled reactors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5184. Light-water-cooled graphite-moderated reactors . . . . . . . . . . . . . . . 5185. Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 519

E. FUEL REPROCESSING . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 519F. WASTE MANAGEMENT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 520G. RESEARCH IN THE NUCLEAR FUEL CYCLE . . . . . . . . . . . . . . . . . . . 521H. SUMMARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 522

III. MEDICAL USES OF RADIATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 524A. DIAGNOSTIC RADIOLOGY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 525B. DENTAL PRACTICE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 526C. NUCLEAR MEDICINE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 527D. RADIOTHERAPY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 528


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ANNEX E: OCCUPATIONAL RADIATION EXPOSURES

Page

E. ALL OTHER MEDICAL USES OF RADIATION . . . . . . . . . . . . . . . . . . 529F. SUMMARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 529

IV. INDUSTRIAL USES OF RADIATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 531A. INDUSTRIAL IRRADIATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 532B. INDUSTRIAL RADIOGRAPHY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 533C. LUMINIZING . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 534D. RADIOISOTOPE PRODUCTION AND DISTRIBUTION . . . . . . . . . . . . 534E. WELL LOGGING . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 535F. ACCELERATOR OPERATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 535G. ALL OTHER INDUSTRIAL USES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 536H. SUMMARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 536

V. NATURAL SOURCES OF RADIATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 537A. COSMIC-RAY EXPOSURES TO AIRCREW . . . . . . . . . . . . . . . . . . . . . 537B. RADON EXPOSURES IN WORKPLACES . . . . . . . . . . . . . . . . . . . . . . . 539

1. Underground mining . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5402. Exposures above ground . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 541

C. EXPOSURES IN MINERAL PROCESSING INDUSTRIES . . . . . . . . . . . 542D. SUMMARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 545

VI. DEFENCE ACTIVITIES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 545A. NUCLEAR WEAPONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 545B. NUCLEAR-POWERED SHIPS AND THEIR SUPPORT FACILITIES . . 546C. SUMMARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 546

VII. MISCELLANEOUS OCCUPATIONAL CATEGORIES . . . . . . . . . . . . . . . . . . 547A. EDUCATIONAL ESTABLISHMENTS . . . . . . . . . . . . . . . . . . . . . . . . . . 547B. VETERINARY MEDICINE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 547C. OTHER OCCUPATIONAL GROUPS . . . . . . . . . . . . . . . . . . . . . . . . . . . 548

D. ACCIDENTS WITH SERIOUS EFFECTS . . . . . . . . . . . . . . . . . . . . . . . . 548E. SUMMARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 551

CONCLUSIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 551

Tables . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 556 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 649

498


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INTRODUCTION

1. There is a wide variety of situations in which people atwork areexposedtoionizingradiation.Thesesituations rangefrom handling small amountsofradioactivematerial, such asfor tracer studies, to operating radiation-generating or -gaug-ing equipment, to working in installations of the nuclear fuelcycle. Therearealsosituations where the exposureofworkersto natural sources of radiation is sufficiently high to warrantits management and control as an occupational hazard.

2. Theconventionaldefinitionofoccupational exposure toany hazardous agent includes all exposures incurred at work,regardless of source [I18]. However, to distinguish theexposures that should be subject to control by the operatingmanagement from the exposures arising from the generalradiation environment in which all must live, the termoccupational radiation exposure is usually taken to mean

thoseexposures that are received at work that can reasonablyberegarded as the responsibilityofthe operatingmanagement[I5, I12]. Such exposures are normally also subject toregulatory control, with the requirements for practices asdefinedbyICRP in itsPublication60 [I12]being applied. Theexposures are usually determined by individual monitoring,but sometimes by other methods. An important objective of such determinations is to provide information on theadequacy of protection measures, and they are a key input tooperational decisions related to the optimization principle. Inaddition, they demonstrate compliance with relevant doselimits.

3. The Committee is interested in reviewing thedistributions of individual annual effective doses and annualcollective effective doses from occupational radiationexposures in various sectors of industryor from various typesof source. It is of particular interest to examine the changesthat have taken place over time with the introduction of improved practices, new technology, or revised regulations.

4. Data onoccupational radiation exposureswere given intheUNSCEAR1977, 1982, 1988, and1993 Reports [U3, U4,U6, U7]. Differences existed, and indeed stilldo exist, amongcountries in the procedures for monitoring and reportingoccupationalexposures;thesedifferences reflect, amongotherthings, differences in regulatory requirements. As a result,comparisons of data on doses are not always straightforwardand may be somewhat limited in scope. Over the years, suchcomparisons have shed light on these differences, and anumber of recommendations have been made. Particularattention wasdrawn to theneed fordataon thepattern of doseaccumulation over a working lifetime, especially for thoseoccupations in which higher levelsof individualexposure areencountered, and to the value of reporting doses in narrowerbands of individual dose. Such data are not readily available,however.

5. The main objectives of the analysis of occupationalradiation exposuresremain, as in theprevious assessmentsof the Committee, as follows:

(a) to assess annual external and committed internaldoses and cumulative doses to workers (both theaverage dose and the distribution of doses within theworkforce) for each major practice involving the useof ionizing radiation. This provides a basis forestimating the average individual risks in aworkforce and within its subgroups;

(b) to assess the annual collective doses to workers foreach of the major practices involving the use of ionizing radiation. This provides a measure of thecontribution made by occupational exposures to theoverall impact of that use and the impact per unitpractice;

(c) to analyse temporal trends in occupational exposuresin order to evaluate the effects of changes inregulatorystandardsorrequirements (e.g. changes in

dose limits and increased attention to making dosesas low as reasonably achievable), new technologicaldevelopments, modified work practices, and, moregenerally, radiation protection programmes;

(d) to compare exposures of workers in differentcountries and to estimate the worldwide levels of exposure for each significant use of ionizingradiation; and

(e) to evaluate data on accidents involving the exposureof workers to levels of radiation that have causedclinical effects.

6. The Committee has evaluated five-year averageexposures beginning in 1975. The detailed data presentedin this Annex are for 1990 1994, but data for previousperiods are provided for comparison. Occupationalexposures in each major practice or work activity arereported, indicating trends with respect to the data in theearlier assessments and identifying the main contributors.Exposures from different countries are compared, andworldwide exposures are determined for each category of work in which radiation exposures occur.

7. The data in thisAnnex wereobtained in much the samewayasthedata for theUNSCEAR 1993 Report [U3]. Data onoccupational exposures from man-made sources of radiation(nuclear power, defence activities, and industrial and medicaluses of radiation) are systematically collected by manynational authorities. The Committee obtained these data bymeans of a questionnaire, the UNSCEAR Survey of Occupational Radiation Exposures, which it distributed tocountries throughout the world. The data have beensupplemented by other (usually published) sources of information; for the nuclear power industry, for example, thesource is the databank of the Organization for EconomicCooperation and Development/Nuclear Energy Agency(OECD/NEA) [O2, O5].However, thedata set isbynomeans

complete, and procedures have been developed by theCommittee to deriveworldwide doses from the data availablefor particular occupational categories (see Section I.E).


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8. The data on doses arising in the commercial nuclearfuelcycleare reasonablycomplete.Where dataare missingor incomplete, doses can be calculated from worldwidestatistics on capacity and production in the various stagesof the fuel cycle. Thus the worldwide annual collectiveeffective dose from a given part of the nuclear fuel cycle isestimated to be the total of the annual collective effectivedoses from the reported data scaled according to the totalworldwidestatistic(uraniummined,fuelfabricated,energygenerated, etc.).

9. For exposures to radiation in other operations, thecalculations are scaled according to the gross domesticproduct (GDP) of countries. The GDP is reasonablycorrelated with the level of both industrial activity andmedical care in a country. To make the calculations morereliable, the values of GDP are applied to regional data,and the results are summed over all regions. For thispurpose, the world was divided into seven regions: the

OECD excluding the United States; the United States;eastern Europe and the countries of the former USSR;Latin America; the Indian subcontinent; east and south-west Asia; and the remaining countries.

10. Exposures from natural sources of radiation, with afewexceptions, havegenerallynot been subject to thesamedegree of control as exposures from man-made sources.The few exceptions are exposures in uranium mines andmills and in practices where purified forms of naturallyoccurring radioactive substances, such as 226Ra andthorium, are handled.

11. The principal natural sources of radiation exposure of interest other than those that have traditionally been directlyrelated to the work (e.g. those in the mining and milling of uranium ores) areradonin buildings,non-uraniumminesandother underground workplaces; cosmic rays at aircraftaltitudes; and materials other than uranium or thorium oresthat contain significant traces of natural radionuclides. Theexposures of individuals in the first two situations are oftencomparable to, if not in excess of, the exposures currentlyreceived from man-madesources. Furthermore, there is somescope for the reduction of these exposures, particularly those

from radon. The large number of workers involved,particularlyin theminingindustry, resultsin annualcollectiveeffective doses that are substantially higher than those fromman-made sources of radiation.

I. DOSE MONITORING AND RECORDING PRACTICES

12. A number of difficulties are encountered indetermining occupational exposures. External radiationfields may be non-uniform in space and time and may beof various types and a wide range of energies. Internalexposures may also occur. Workers may be frequentlyexposed, seldom exposed, or hardly exposed at all. Thedifficulties may be addressed in various ways, as reflectedin thevarietyofmonitoring procedures and dose recordingpractices adopted in countries throughout the world. Thistopicwasaddressed in some detail in the UNSCEAR 1993Report [U3]. However, to the extent that attention stillneeds to be drawn to it or that changes have occurred thatmay affect the interpretation of results, the topic isdiscussed further in this Chapter.

A. QUANTITIES MEASURED

1. Protection quantities

13. The basic physical quantity used in radiologicalprotection is theabsorbeddose,DT, averaged over an organor defined tissue. The absorbed dose is expressed in theunit gray(Gy), with 1 Gy equal to 1 joule per kilogramme.To account for the type of the radiation and the differencesin ionization density, a further quantity has beenintroduced, the equivalent dose, HT, which is the average

absorbed dose in an organ or tissue multiplied by adimensionless factor called the radiation weighting factor,wR. Equivalent dose is expressed in the unit sievert (Sv).

14. The effective dose, E, also expressed in Sv, has beendefined to take account of the fact that the probability of stochastic effects for a given equivalent dose varies withthe organ or tissue irradiated. The factor by which theequivalent dose in a tissue or organ is weighted is calledthe tissue weighting factor, wT, the values being chosensuch that theeffective dosegivesa measureofthe radiationdetriment irrespective of how that dose was received. Inparticular, this approach allows effective doses fromexternal and internal exposures to be aggregated.

15. Effective dose and equivalent dose are the basicquantities for radiological protection purposes in which, forexample, dose limits are expressed [I12]. The effective doselimit is intended to limit the total health detriment fromradiation exposure due to stochastic effects. Limits onequivalent doseare requiredfor skin andthelens of theeye toensure that deterministic effects are avoided in these tissues.These protection quantities relate, as appropriate, to the sumof the effective or equivalent doses from external sources andthe committed effective or equivalentdosesfrom the intakeof radionuclides. Dose quantities are discussed in detail inAnnex A, Dose assessment methodologies .

2. Quantities for external radiation exposure

16. The basic quantities for physical measurementinclude particle fluence, kerma, and absorbed dose. Theyare the quantities used by national standards laboratories.


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PHYSICAL QUANTITIES

Compared by measurement and calculations

OPERATIONAL QUANTITIES PROTECTION QUANTITIES

Fluence,

Ambient dose equivalent, H*(d) Organ absorbed dose, DT

T

Kerma, K

Directional dose equivalent, H'(d, ) Organ equivalent dose, H

Absorbed dose, D

Monitored quantitiesand

instrument responses

Calculated using Q(L) and sample phantoms (sphere or slab)validated by measurementsand c alculations

Related by calibration

and calculation

Calculated using w , wand anthropomorphic phantoms

R T

(using w , wand anthropomorphic phantoms)

R T

pPersonal dose equivalent, H (d) Effective dose, E

However, the need for measurable quantities for externalradiation exposure that can be related to the protectionquantities has led to the development of operationalquantities, which provide an estimate of effective orequivalent dose that avoids underestimation and excessiveoverestimation in most radiation fields encountered inpractice.

17. There are three operational quantities of particularinterest in the measurement of radiation fields forprotection purposes: the ambient dose equivalent, H*(d);the directional dose equivalent, H'(d, ); and the personaldose equivalent, Hp(d). All these quantities are based onthe dose equivalent at a point and not on the concept of equivalent dose. The ambient dose equivalent and thedirectional dose equivalent are appropriate forenvironmentalandareamonitoring,theformer for stronglypenetrating radiation and the latter for weaklypenetratingradiation. The ambient dose equivalent at a point in a

radiation field is the dose equivalent that would beproduced bythe corresponding aligned and expanded fieldin the ICRUsphere at a depth d on the radius opposing thedirection of the aligned field. The directional doseequivalent at a point is the dose equivalent that would beproducedbythe corresponding expanded field in the ICRUsphere at a depth d on a radius in a specifieddirection. Theconcepts of expanded and aligned fields are given inICRU Report 39 [I19] to characterize fields that arederived from the actual radiation fields. In the expandedfield, the fluence and its angular and energy distributionhave the same values throughout the volume of interest asat the actual field at the point of reference. In the alignedand expanded field, the fluence and its energy distributionare the same as in the expanded field, but the fluence isunidirectional.

18. The personal dose equivalent, Hp(d), is the doseequivalent in soft tissue below a specified point on thebodyatan appropriate depth d. This quantity can be used formeasurements ofsuperficial and deep organ doses, dependingon the chosen value of the depth in tissue. The depth d isexpressed in millimetres, and ICRU recommends that anystatement of personal dose equivalent should specify thisdepth. For superficial organs, depths of 0.07 mm for skin and3 mm for the lens of the eye are employed, and the personaldoseequivalentsfor thosedepths aredenoted byHp(0.07) andHp(3), respectively. For deep organs and the control of effective dose, a depth of 10 mm is frequently used, with thenotation Hp(10).

19. Personaldoseequivalent quantities aredefined in thebody and are therefore not directlymeasurable. They varyfrom person to person and from location to location on aperson, because of scattering and attenuation. However,Hp(d) can be assessed indirectly with a thin, tissue-

equivalent detector that is worn at the surface of the bodyand covered with an appropriate thickness of tissueequivalent material. ICRUrecommends that dosimetersbecalibrated under simplified conditions on an appropriatephantom [I20].

20. The relationship between the physical, protection,and operational quantities is illustrated in Figure I. Theyare discussed more fully in ICRP Publication 74 [I16],which provides conversion coefficients for use inradiological protection against external radiations. It wasconcluded that there is an acceptable agreement betweenthe operational and protection quantities for radiationfields of practical significance when the operationalquantities are based on the Q/LET relationship given inICRP Publication 60 [I12].

Figure I. Relationship of quantities for radiological protection monitoring purposes [I16].

21. In most practical situations, dosimeters providereasonableapproximationstothepersonaldoseequivalent,Hp(d), at least at the location of the dosimeter. When theexposure of the body is relatively low and uniform, it is

common practice to enter the dosimeter reading, suitablycalibrated, directly into the dose records as a surrogate foreffective dose. However, because the personal doseequivalent generally overestimates the effective dose, this


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E(t) Hp(d) j

e j,inh(50) I j,inh j

e j,ing(50) I j,ing

practice results in overestimated recorded and reporteddoses, with the degree of overestimation depending on theenergyofthe radiation and thenature ofthe radiation field.For many practical situations involving relativelyuniformexposuretofairlyhigh-energygammaradiation, thedegreeof overestimation is modest; for exposure to low-energygamma or x radiation, the overestimation can be substan-tial. For photon energies below ~50 keV, theeffective dosecan be overestimated by a factor of 2, depending on theorientation of the body.

22. For exposure to spatially variable radiation fields orwhere there is partial shielding of the body or extremevariations in the distances of parts of the body from thesource, the relationships between the dosimeter measure-mentandtheeffectivedosearemorevariableandcomplex.Where the circumstances so justify, additional measure-ments or theoretical analysis have been used to establishreliable relationships on a case-by-case basis for the

exposure conditions of interest. The direct entry of dosimeter measurements into dose records in these morecomplex situations (or the use of very simple anddeliberatelycautious assumptions to establish therelation-ships between the two quantities) leads, in general, tooverestimates in the recorded exposures. Where suchpractice has been adopted in the recording ofdoses, care isneeded in their interpretation, in particular when they arebeing compared with doses arising elsewhere. Theinformation available to the Committee is generally notsufficient to allow the exercise of such care in interpretingrecorded values.

23. For itspreviousassessments, theCommitteeadopted theconventionthatall quantitative resultsreportedbymonitoringservices represent the average absorbed dose in the wholebody(or theeffective dose). It is further assumed that the dosefromnormalnaturalbackgroundradiation hasbeensubtractedfrom the reported results, although this was not always clearfromtheresponsesto thequestionnaire. It is alsoassumed thatmedical radiation exposures have not been included. TheCommittee recognized that it is almost always the readingfrom the dosimeter, suitably modified by calibration factors,that is reported, without considering its relationship to theabsorbed doses in the various organs and tissues of the body

or to the effective dose. This is still regarded as a reasonableconvention, in particular as most data are for externalexposure of the whole body to relatively uniform photonradiation of moderately high energy. Where exposure of thebody is very non-uniform (especially in medical practice) orwhere exposure is mainly to low-energyradiation, the use of this convention may result in an overestimate of effectivedoses,which thenneedsappropriatequalification.Becausetherelationship between the reported dosimeter reading and theaverage absorbed dose in the whole body (or the effectivedose) varies with the circumstances of the exposure, cautionneeds tobeexercisedwhen aggregatingor directlycomparingdata fromverydissimilar types ofwork. Thereported data areappropriately qualified where the adoption of the aboveconventioncould leadtoa significantmisrepresentation oftheactual doses.

3. Quantities for internal radiation exposure

24. Radionuclides taken into the body will continue toirradiate tissue until they have been fully excreted or havefully decayed. The committed effective dose for occupa-tional exposure, E(50), is formally defined as the sum of the products of the committed organ or tissue equivalentdosesandtheappropriateorganortissueweightingfactors,where 50 is the integration time in years following intake.Thecommittedequivalentdose,HT(50), is formallydefinedas the time integral of the equivalent dose rate in aparticular tissue or organ that will be received by anindividual following intake ofradioactivematerial into thebody, where 50 is, again, the integration time in yearsfollowing intake.

25. In thecalculation ofE(50) and, where appropriate, of HT(50), the dose coefficient is frequently used. Foroccupational exposure, this is the committed effective dose

per unit acute intake, e(50), or committed tissue equivalentdose per unit acute intake, hT(50), where 50 is the timeperiod in years over which the dose is calculated. The unitis sievert per becquerel.

26. ICRP has recommended that the annual limit onintake (ALI) shouldbebased on a committedeffective doseof 20 mSv [I12]. The annual limit on intake (Bq) can thenbe obtained by dividing the annual average effective doselimit (0.02 Sv) bythe dose coefficient, e(50)(Sv Bq 1). Thedose coefficients for occupational exposure for inhalationand ingestion of radionuclides based on the radiation andtissue weighting factors in ICRP Publication 60 [I12] andthe new Human Respiratory Tract Model for RadiologicalProtection [I14] are given in ICRP Publication 68 [I15].

4. Total effective dose

27. The total effective dose, E(t), during any timeperiod,t, can be estimated from the following expression:

where Hp(d) is the personal dose equivalent during time

period t at a depth d in the body, normally 10 mm forpenetrating radiation; e j,inh(50) is the committed effectivedose per unit activity intake by inhalation fromradionuclide j, integrated over 50 years; I j,inh is the intakeof radionuclide j by inhalation during time period t;e j,ing(50) is the committed effective dose per unit activityintake by ingestion from radionuclide j, integrated over 50years; I j,ing is the intake of radionuclide j by ingestionduring time period t.

28. The conversion coefficients for use in radiologicalprotection against external radiation are given in ICRPPublication 74 [I16]. Except for radon progeny, values of thecommitted effective dose per unit intake for inhalation,e j,inh(50), and ingestion, e j,ing(50), are found in ICRPPublication 68 [I15], which takes account of the tissue


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weighting factors in ICRP Publication 60 [I12] and the newlung model in ICRP Publication 66 [I14]. It is assumed thatthe data provided to the Committee will have been based onthese conversion coefficients. The parameters for radon aregiven below.

5. Special quantities for radon

29. Special quantities and units are used to characterizethe concentration of the short-lived progeny of both 220Rn(commonlyknown as thoron) and222Rn (commonlyknownas radon) in air and the resulting inhalation exposure (seeICRP Publication 65 [I13]).

30. The potential alpha energy, p, of an atom in thedecay chain of radon or thoron is the total alpha energyemitted during the decay of this atom to 206Pb or 208Pb,respectively. The SI unit is joule, J; MeV is also used. Thepotential alpha energyconcentration, cp, ofanymixture of short-lived radon or thoron decayproducts in air is the sumof the potential alpha energy of these atoms present perunit volume of air, and the SI unit is J m 3. The potentialalpha energy concentration can also be expressed in termsof the unit working level (WL), which is still used in somecountries. One WL is defined as a concentration of potential alpha energyof 1.30 108 MeV m 3. The potentialalpha energy concentration can also be expressed in termsof the equilibrium equivalent concentration, ceq, of theparent nuclide, radon. The equilibrium equivalentconcentration for a non-equilibrium mixture of radon

progeny in air is that activity concentration of radon inradioactive equilibrium with its short-lived progeny thathas the same potential alpha energy concentration, cp, asthe non-equilibrium mixture. The SI unit of theequilibrium equivalent concentration is Bq m 3.

31. The exposure of an individual to radon or thoronprogeny is determined by the time integral of thepotential alpha energy concentration in air or of thecorresponding equilibrium equivalent concentration. Inthe former case, it is expressed in the unit J h m 3 and inthe latter, in the unit Bq h m 3. The potential alpha

energy exposure is also often expressed in the historicalunit working level month (WLM). Since this quantitywas introduced for specifying occupational exposure,one month was taken to be 170 hours. Since 1 MeV =1.60210 13J, the relationship between the historical andthe SI unit is 1 WLM = 3.54 10 3 J h m 3. The factor forconverting from WLM to effective dose has been thesubject of some debate. The Committee has adopted aradon dose coefficient of 9 nSv (Bq h m 3) 1. However,the ICRP derived a conversion convention of 5 mSv(WLM) 1 or 6 nSv (Bq h m 3) 1, which was used in thequestionnaire sent to national authorities in gatheringinformation for theAnnex. As a result of this difference,the data in this Annex for radon exposure situationsunderestimate the doses by about 30%.

B. MONITORING PRACTICES

32. For many reasons, worker monitoring practices differfrom country to country, from industry to industry, andsometimes even from site to site within a given industry.Someofthese differences stem from historical, technological,cost, or convenience considerations. In general, monitoringpractice is such that more workers are individuallymonitoredthan isstrictlynecessarytomeet regulatoryrequirements, withthe consequence that only a fraction of those monitoredreceive measurabledoses. Although these differencesmaynotseriouslyaffect thequalityofthe data, theycould lead to somedifficulties in making valid comparisons of results.

33. It is convenient tosubdividemonitoring programmesinto a number of categories. Routine monitoring isassociated with continuing operations and is intended todemonstrate that the working conditions, including thelevels of individual dose, remain satisfactory and meet

regulatory requirements. This sort of monitoring is largelyconfirmatory in nature, but it underpins the overallmonitoring programmes that should be undertaken tocontrol occupational exposure. The most common type of routine monitoring is that undertaken using passivedevices, such as film badges or TLDs. Such dosimeters aregenerallyworn bypersonnel for a set period, and at the endof this period they are read and the doses recorded. In themain, the information used in this Annex comesfrom suchmonitoring programmes,although the approaches adoptedand the degree of quality control exercised over themeasurements vary from country to country.

34. To obtain a more up-to-date understanding of workerexposures, additional task-related monitoring is oftenundertaken. The intention of such monitoring is to providedata to support immediate decisions on the management of operations and optimization of protection. Task-relatedmonitoring is usually based on some type of direct-readingdosimeter, such as a digital electronic dosimeter or a quartz-fibre electroscope, although multi-element TLD systems arealso used. Some examples are given in this Annex.

35. Special monitoring may also be conducted whendeemed necessary. It is investigative in natureand typically

covers a situation in the workplace where insufficientinformation is available to demonstrate adequate control.It is intended to provide detailed information that willelucidate any problems and define future procedures.

36. ICRP indicates [I12] that three important factorsshould influence the decision to undertake individualmonitoring: the expected level ofdose or intake in relationto the relevant limits, the likely variations in the dose andintakes, and the complexity of the measurement andinterpretation procedures that make up the monitoringprogramme. In practice, it is usual for all those who are

occupationally exposed to external radiation to beindividually monitored (i.e. to wear personal dosimeters).When doses are consistently low or predictable, other


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methods of monitoring are sometimes used, as in the caseof aircrew where doses can be calculated from flightrosters. The third factor results in an approach to themonitoring for external radiation that is different from thatfor intakes and the resulting committed effective dose.

1. External radiation exposure

37. The approach followed in many countries is tomonitor the external radiation exposures of all individualswho work routinely in designated areas. However, on thebasis of the recommendations of ICRP [I10], a distinctionhas often been made in monitoring programmes betweenthose who can exceed 3/10 of the relevant dose limit andthose who are most likely not to exceed. While individualmonitoring may well have been carried out for thosein thesecond category, the difference in monitoring lies largelyin the degree of quality control that is exercised over themeasurement. For the Committee, it is important to know

whether doses to both groups of workers have beenreported to it.

38. Monitoring programmes usually specify how andwhere personal dosimeters are to beworn to obtain the bestestimate of effective dose or equivalent dose, asappropriate. In general, a dosimeter is placed on the frontof the body. This is satisfactory provided that thedosimeters have been designed to measure Hp(10).

39. Where lead aprons are used in medical radiology,different approaches havebeen adopted. In somecases, theassessment of effective doses to workers is carried out bymeans of a dosimeter worn on the trunk, under the apron.Where doses are likely to be significant, such as ininterventional radiology, two dosimeters are sometimesused, one worn under the lead apron and the other wornoutside. The purpose of the second dosimeter is to assessthe contribution to the effective dose of irradiation of unshielded parts of the body [N6]. Where doses are lowand individual monitoring is only intended to give anupper estimate of exposure, single dosimeters may havebeen worn outside the apron. Measurements made onphantoms using x-ray beams of 76 and 104 kVp haveshown that estimates of the effective dose without the lead

apron were within 20% of expected values; estimates withthe dosimeter worn on the waist underneath the lead apronwere lower than the expected values [M1]. The resultssuggest that accurate estimation of the effective dose frompersonal dosimeters under conditions of partial bodyexposure remains problematic and is likely to require theuse of multiple monitors, which is not often done.Differing monitoring practices in medical radiology maytherefore affect the validity of any comparisons of dataacquired.

40. Thechoiceof dosimeter will dependon the objectivesof the monitoring programme and on the method of interpreting the data to be used. In practice, the basicchoice for penetrating radiation has usually been betweena dosimeter giving information on the personal dose

equivalent at 10 mm depth and a discriminating devicegiving some indication of the types of radiation and theireffectiveenergies. For a widerangeofenergies, TLDswithdetectors that exhibit little energy dependence of tissuedoseresponseand arecoveredwith tissue-equivalent filtersof appropriate th icknesses are an example of the former.Multi-element dosimeters using either photographic filmor thermoluminescent material, with filters of differentatomic numbers and thicknesses, are an example of thesecond type.

41. The quality and accuracy of personal electronicdosemeters is improving rapidly, and in a few countriesthey have already been approved for formal doseassessment for some types of radiation to meet regulatoryrequirements. The approvals have tended to be limited tospecific groups of workers [C2], but the pace of development is such that they are being considered asalternatives to photographic film and TLDs. They offer a

low threshold limit of detection and a digital read-out.42. Personaldosimeters that respondto neutronsover thecomplete energy range of interest are not available, andsome of the current methods of assessment may berelatively expensive and time-consuming. Where thecontribution to effective dose from neutrons is smallcompared with that from photons, the dose is sometimesdeterminedbyreferenceto thephoton doseand an assumedratio of the two components. Alternatively, use is made of measurements in the workplace environment and anassumed occupancy.

43. Monitoring for incident thermal and epithermalneutrons is performed using detectors with high intrinsicsensitivity to thermal neutrons (e.g. some TLDs) or detectorssensitive to other types of radiation (photons and chargedparticles) and a converter. Neutron interactions in the con-verter produce secondary radiations that are detectable bythedosimeter. Themost common example of the latter techniqueis the film badge used with a cadmium filter. Some dosi-meters have been designed such that they respond, in themain, to thermal and epithermal neutrons produced in thewearers body by moderation and scatter of higher energyneutrons incident on the body. These albedo neutron

dosimeters have good response characteristics up to 10 keVneutron energy and, by normalization appropriate to theworkplace field, are used where the neutron personal doseequivalentisdominated byneutronsoutsidethisenergyrange.The normalization process is critically dependent on theneutron spectrum, and if this is not well known or isvariable,significant errors mayresult.

44. The assessment of personal dose equivalents fromfast neutrons is carried out by means of nuclear emulsiondetectors, bubble detectors, or track-etch detectors (e.g.poly-allyl diglycol carbonate, PADC). Nuclear emulsiondosimeters can measure neutrons at thermal energies andat energies above700 keV. Theyhavethe disadvantages of being relatively insensitive to neutrons with intermediateenergies and being sensitive to photons, and they suffer


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from fading. Bubble detectors respond to fast neutronsfrom 100 keV upwards and have the advantage that theyare direct-reading, non-sensitive to photons, and reusable,but they have the disadvantage of being temperature- andshock-sensitive. Track-etch detectors based on PADCrespond to fast neutrons from about 100 keV upwards.

45. There is a highly complex relationship between theexposure to radiation and the effective dose. Models arerequired that are intended to give results that are not likelyto underestimate the consequences of exposure, thoughwithout overestimating them excessively. This is theobjective of the operational quantities.

46. In the workplace, the dose rate in air varies as afunction of position and time. In the body, the equivalentdose in an organ or tissue is related to the dose equivalentat the surface by factors such as the type and quality of theradiation, thenon-uniformityof the field, the orientation of

the worker relative to the field, and the position andcomposition of the organs and tissues within the body.Several of these factors will be functions of both time andposition in the workplace.

47. A dosimeter worn on the surface of the body is bestregarded as a sampling device. It provides a measure of thedose equivalent to the skin and underlying tissue in theimmediatevicinityof the dosimeter. A personal dosimeter ona phantom can be calibrated in terms of the measured orcalculated valuesofthe personaldose equivalentHp(d). Whenworn on the body of a person facing a unidirectional field of radiation, it will indicate thepersonaldose equivalent. Wherea worker moves about the workplace, resulting effectively ina multidirectional field, a personal dosimeter will provide anadequate measure of the personal dose equivalent. Further-more, the personal dose equivalents will, for most combina-tions of exposure, overestimate the effective dose. In somecases, the overestimation may be substantial.

48. There are three main areas of uncertainty inindividual monitoring for external radiation:(a) that which is inherent in dose calibrations;(b) that due to the measurement of the operational

quantity Hp(10) as compared with the reading of anideal dosimeter for the measurement of the quantitywhen worn on the same point on the body; and

(c) that which occurs if the dosimeter is not worn at theappropriate point on the body.

These uncertainties and how they are dealt with by thedosimetry services could also have an impact on thecomparisons made in this Annex.

49. Many countries appear to follow the guidance givenin ICRP Publication 35 [I10]. This defines acceptableuncertainties in routine monitoring for external radiation.Near the dose limits, the recommendation is that theuncertainty should be within a factor of 1.5 in eitherdirection. Some relaxation is allowed at lower doses. It hasbeen shown that these recommendations can be met by the

majority of personal dosimeters currently in use, as far asthe measurement of Hp(10) is concerned [M2]. It must beappreciated, however, that the relationship between Hp(10)and E introduces further errors, for example for photons.These are relatively small at higher photon energies (e.g.>0.5 MeV), but large overestimates can occur at lowerenergies, up to a factor of 5 at 10 keV.

2. Internal radiation exposure

50. There are three approaches to the determination of intake and internal dose:(a) byquantification ofexposureto radioactivematerials

in termsoftheir time-integrated air concentration viaair sampling techniques;

(b) by the determination of internal contamination viadirectin vivo measurements (in vivo methods includedirect measurements used for assessing gamma and

x-rayemitters and measurementsofbremsstrahlung,bymethodssuch aswhole-body, thorax, skeleton,andthyroid counting); and

(c) by the measurement of activity in in vitro biologicalsamples (in vitro methods are usually based onanalysis of urine or faecal samples).

In practice, the approach adopted for a situation willdepend on the abilities of the various options to indicatedoses in that particular situation.

51. The choice between the three approaches isdetermined by the radiation emitted by the radionuclide;

the biokinetic behaviour of the contaminant; its retentionin the body, taking into account both biological clearanceand radioactive decay; the required frequency of measure-ments; and the sensitivity, availability, and convenience of the appropriate measurement facilities. The most accuratemethod in the case of radionuclides emitting penetratingphoton radiation is usually in vivo measurements.However, even when this method can provide informationon the long-term accumulation of internal contamination,it may not be sufficient for assessing committed dose dueto a single years intake. The assessment may also needdata from air monitoring. In many situations, therefore, acombination of methods is used. For radon doseassessments, however, air monitoring (individual or area)is the only available routine method.

52. There are two methods for the determination of exposure to airborne contamination:(a) the use of representative/area air monitoring data,

combined with a knowledge of occupancy of indivi-dual workers within each sampling area and anassumed breathing rate. This method is often used insituations where the more significant intakes areassociated with well defined work activities; and

(b) the routine useofpersonal air samplers. This is oftenused where significant contributions to internalexposure are not linked to identifiable fixedlocations.


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53. Intakesof radioactivematerial are normallyassessedroutinely for workers who are employed in areas that aredesignated as controlled, specifically in relation to thecontrol of contamination, and in which there are groundsfor expecting significant intakes. However, there aredifficulties in comparing dataon internal doses in differentcountries because of the different approaches that are usedto monitor and interpret the results. Measurements in aroutine monitoring programme are often made atpredetermined times not necessarily related to a particularintake event, and it is therefore necessary to make someassumptions about the pattern of intakes. Guidance oninterpreting the results of measurements of intakes of radionuclidesbyworkerswasgiven in ICRP Publication 54[I11]. This publication has been replaced, however, by anewdocument [I1] that uses current biokinetic modelsandis consistent with ICRP Publication 68 [I15]. In keepingwith the ICRP advice, it is usual for the results of in vivoand in vitro monitoring measurements to be interpreted

using the assumption that the intake took place at themid-pointofthe intervalbetweenmonitoringtimes.Assessmentof doses from air sampling data requires knowledge of thephysical and chemical properties of the radioactivematerials, including the particle size and solubility inbiological fluids. The current recommendation of ICRP[I15] is that a default value of 5 m should be used for theparticle size; previously, a value of 1 m was recom-mended and maystill be in use. A major difficulty in usingarea air sampling data to assess dose is whether themeasurement data can be related to the activityconcentration in the breathing zone. There is also theparticular difficulty in interpreting area air sampling datawhen the contamination is due to localized sources orwhere only a few particles of radioactive material canrepresent a significant intake.

54. With the techniques currently available, it isgenerally not possible to obtain the same degree of precision in routine assessments of dose from intakes of radioactive material as is possible with external radiation.The dose assessment falls into three stages:(a) individual monitoring measurements;(b) assessment of intake from the measurements; and(c) assessment of doses from the intake.

The overall uncertainty in the assessed dose will be acombination of the uncertainties in these three stages. Agoodexampleof the uncertaintiesinvolvedand the relativemeritsofvariousdoseassessment techniques isprovidedbya studyofchronic low-level exposure ofworkers in nuclearfuel reprocessing [B3]. The study was able to compareassessments of intakes from static air sampling (SAS) andpersonal air sampling (PAS) and to then compare doseassessments from personal air sampling and biological invitro samples. In the first of these comparisons, the doseassessed by personal air sampling was about an order of magnitude larger than that implied by static air sampling.For the group as a whole, over a seven-year period therewas reasonable agreement between the geometric meancumulative doses (23 mSv for biological sampling and

30 mSv for personal air sampling). However, there was alack of correlation when viewed at any individual level,with nosingle identifiable factor to explain the difference.This must cast some doubt on the adequacyof personal airsamplers for estimating annual intakes of individualworkers at the levels of exposure encountered inoperational environments.

55. In practice, there are relatively few occupationalsituations in which internal exposures to man-madesources of radiation are significant, and significantexposures have generallybeen decreasing. Exposures maystill be significant in a number of situations, however: thehandling of large quantities of gaseous and volatilematerials such as tritium (e.g. in the operation of heavy-water reactors and in luminizing); reactor fuel fabrication;the handling of plutonium and other transuranic elements(e.g. in the reprocessing of irradiated fuel and in nuclearweapons production); and some nuclear medicine

situations. Significant internal exposures to naturalradionuclides can occur in the mining and processing of radioactive ores, particularly uranium ores but also someother materials with elevated levels of naturalradionuclides (e.g. mineral sands). Significant exposure toradon can also occur in other mines, underground areassuch as show caves (e.g. those that are open to tourists),andsomeabovegroundworkplacesnotnormallyassociatedwith radiation exposure.

C. DOSE RECORDING AND REPORTINGPRACTICES

56. In most countries dose recording and reportingpractices are governed by regulations and can be differentfor various categories of workers depending on theiranticipated levels of exposure. Like monitoring practices,they vary from country to country and may significantlyaffect the reported collective doses. The most importantdifferences arise from the following:

(a) the recording of doses less than the minimumdetectable level (MDL);

(b) the measurement technique used, for example, TLD,film, or electronic dosimeter in the case of externalradiation exposure;

(c) the assignment of doses to fill missing recordperiods;

(d) the treatment of unexpectedly high doses;(e) the subtraction of background radiation doses;(f) the protocol for determining who in the workforce

should be monitored and for whom doses should berecorded in particular categories; and

(g) whether or not internal exposures are included ortreated separately.

57. The recording level is the level above which a resultis considered tobesignificantenough toberecorded, lowervalues being ignored [I12]. Recent advice from ICRP is


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that therecording level for individual monitoringshouldbebased on the duration of the monitoring period and anannual effective dose no lower than 1 mSv [I17]. Inpractice, littleuse is made ofrecording levels in individualmonitoring for external radiation exposure, and manycountries adopt the practice of recording all measureddoses above the MDL for the technique used. When dosesaredeterminedto be less than the MDL, the valuerecordedmaybezero, some pre-designated level, or the MDL itself.These differences affect the comparability of results.Furthermore, the MDLwill varywith the device used. Forexample, theMDL associated with electronicdosimeters isgenerally much lower than that for film badges or TLDs.Electronic dosimeters have not been extensively used forthe assessment of individual dose for record keepingpurposes, but this situation is changing. This could lead tosignificant differences in the recording of low levels of external exposure. For instance, during the first fourmonths of operation of an electronic dosimetry system at

SizewellBnuclearpower plant in the UnitedKingdom, themonthly collective dose measured by film badges washigher by a factor of 20 than that measured by electronicdosimeters [R1]. It is therefore important tounderstand theimplications ofrecording levelsand differentMDLson theaverage individual dose and collective dose.

58. When dosimeters are lost or readings are otherwisenot available, administrative procedures are then used inassigning doses to individual dose records. These areassumed doses to the workers for the appropriateperiodforwhich measurements are not available. A variety of procedures are used in determining the assigned dose.These includetheassignment ofthe appropriateproportionof the annual limit for the period for which the dosimeterwas lost; the assignment of the average dose received bythe worker in the previous 12 months; and the assignmentof the average dose received by co-workers in the sameperiod. Some of these procedures can distort recordssignificantly, particularly if large numbers of dosimetersare lost within a particular occupational group. Wherethisis the case, direct comparisons with other data may beinvalid or, at least, need qualification. A similar situationmay arise in the treatment of unexpectedly high measureddoses that are considered not to be a true reflection of the

actual doses received.59. The background signal of a dosimeter involvescontributions from both the non-radiation-induced signalsfrom the dosimeter and the response of the dosimeter tonatural background radiation. This signal is oftensubtracted from the actual dosimeter reading beforerecording. In manycountries, the practiceis to usea singlevalue that takes account of the contributions to thebackgroundsignal, that from natural background radiationbeing the average for the country as a whole. Where thereare significant variations in the gamma-ray contributionfrom natural sources, this practice may have someinfluence on the individual doses that are recorded,particularly where the occupational exposures are similarin magnitude to those from the natural environment.

60. In the past, internal and external exposures weregenerally recorded separately. Furthermore, there weresignificant variations in the reporting levels for internalcontamination, and this added to the difficulty of compilingmeaningful statistical information. There isnowincreased emphasis on recording the sum of the annualeffective dose fromexternal irradiation and the committedeffective dose from internal irradiation. Such data willenable more valid comparisons to be made of the radio-logical impactofdifferent practices.However,comparisonsof the more recent data with data for earlier periods willneed to be treated with caution. For example, internalexposures in someoccupations and industries (fuelfabrica-tion and fuel reprocessing) may have been significantduring the periods covered in previous assessments by theCommittee but may not have been included in the data.Furthermore, inclusion of internal doses may result in anapparent step increase in the level of exposure received byworkers in industries where internal exposure contributes

significantly.

61. A major cause of difficulty in comparisons,particularly of average individual and collective doses, isthe protocol used for determining who in the workforce isto be monitored and to have data recorded within anyparticular category. For instance, it is important to knowwhether the data for nuclear power operations includedoses to visitors, administrative staff, and contract workersin addition to the companys employees.

62. In the UNSCEAR 1993 Report [U3], the advantagewas noted of reporting data according to an agreedcategorization scheme of work and also the difficulty of doing so, particularly in view of the differences in long-established national practices. The categories used by theCommittee in this Annex are given in Table 1; there aresome differences between this categorization and that usedin the UNSCEAR 1993 Report. The main differences arethat veterinarypracticeand educational establishmentsarenow placed in a miscellaneous category, and there is somedevelopment of the section on natural radiation. Howeverthe approach adopted should still permit broad com-parisons to be made with the data in the UNSCEAR 1993

Report. The dose monitoring and recording procedures foroccupational exposure obtained from the UNSCEARSurvey of Occupational Radiation Exposures are given inTable 2. The data are not comprehensive for some of theattributes.

63. Any harmonization of the way data are recorded invarious countries would help in future surveys. TheEuropean Union has an ongoing project, European StudyofOccupationalExposure (ESOREX) [F3], to compare theadministrative systems of the member states that are usedfor registering individualoccupational exposure,to identifydifferences, and to analyse the possibilityof harmonizationwithin Europe. The project has alsobeen extended to covercentral and east European countries [F4].


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S

N

i 1Ei

S

r

j 1N j E j

NRE N(>E)

N

SRE S(>E)

S

D. CHARACTERISTICS OF DOSEDISTRIBUTION

64. Dose distributions are the result of many constraintsimposed bythe natureof the work, bymanagement, bytheworkers, and by legislation. In some job categories it maybe unnecessary for workers ever to receive more than verylow doses, whereas in other jobs workers may have to beexposed to high doses fairly routinely. Managementcontrols act as feedback mechanisms, especially whenindividual doses approach the annual dose limit, or someproportion of it, in a shorter period of time.

65. TheCommitteeis principallyinterested in comparingdose distributions and in evaluating trends. For thesepurposes, it identified three characteristics of dosedistributions as being particularly useful:(a) the average annual effective dose (i.e. the sum of the

annual dose from external irradiation plus thecommitted dose from intakes in that year), E ;

(b) the annual collective effective dose, S (referred to asM in some earlier UNSCEAR reports), which isrelated to the impact of the practice; and

(c) the ratio, SRE, of the annual collective effective dosedelivered at annual individual doses exceedingE mSv to the total collective dose. SR (referred to asMR in some earlier UNSCEAR reports) provides anindication of the fraction of the collective dosereceived by workers exposed to higher levels of individual dose. This ratio is termed the collectivedose distribution ratio.

66. Another ratio, NRE, ofthe number of workersreceivingannual individual doses exceeding E mSv to the totalmonitored or exposed workforce, is reported in manyoccupational exposure statistics, often when the ratio SRE isnot provided. The more frequent reporting ofthe ratio NRE isprobablydueto theeasewith which it can beestimated. In thepast, the Committee was somewhat concerned because of theratios potential sensitivity to howthe size ofthe workforce isdefined (those monitored, those measurably exposed, etc.);comparisons of values of this ratio for different occupationsand in different countries would, in general, require somequalification. The ratio SRE, on the other hand, is relativelyinsensitiveto thisparameter and is thereforea better meansof affording fair comparisons between exposures arising indifferent industries or practices. Notwithstanding thelimitations of the ratio NRE, it is included in thecharacteristics reported by the Committee. This reflects itspotential for use in more limited circumstances (e.g. whenanalysing trends with time in a given workforce or makingcomparisons between workforces that have been defined incomparableways). The ratioSRE, however, remains the mostappropriate basis for comparing data generally.

67. The annual collective effective dose, S, is given by

where Ei is the annual effective dose received by the ithworker and N is the total number of workers. In practice,S is often calculated from collated dosimetry results usingthe alternative definition

where r is the number of effective dose ranges into whichthe dosimetry results have been collated and N j is thenumber of individuals in the effective dose ranges forwhich E j is the mean annual effective dose. The averageannual effective dose, E , is equal to S/N. The numberdistribution ratio, NR, is given by

where N(>E) is the number of workers receiving annual

doses exceeding E mSv. The annual collective dosedistribution ratio, SR, is given by

where S(>E) is the annual collective effective dosedelivered at annual individual doses exceeding E mSv.

68. The total number of workers, N, warrants furthercomment, as it has implications for the various quantitiesestimated. Depending on the nature of the data reported andsubject to the evaluation (or the topic of interest), the numberof workers may be those monitored, those classified, thosemeasurably exposed, the total workforce, or some subsetthereof. These quantities, therefore, will always be specific tothe nature and composition of the workforce included in theestimation; when making comparisons, caution should beexercised to ensure that like is being compared with like.These aspects were discussed in Section I.C, where theimplications of different monitoring and reporting practicesfor the assessed average individual and collective doses wereidentified. In this Annex, consideration is, to the extentpracticable, limited to the estimation of the above quantitiesfor the monitored and measurably exposed workforces;

however, lack ofuniformitybetween employers and countriesin determining who should be monitored and/or whatconstitutes measurably exposed means that even thesecomparisons between ostensibly the same quantities are lessrigorous than might appear. Where necessary, quantitiesestimated for a subset of the workforce (e.g. those measurablyexposed)can be transformed to applyto the whole workforce;methodsofachieving this, based oncharacteristics ofthe dosedistributions, are discussed below.

69. In summary, the following characteristics of dosedistributions will be considered by the Committee in thisassessment of occupational exposures:(a) the average annual effective dose (i.e. the sum of the

annual dose from external radiation and thecommitted dose from intakes in that year), E ;


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Sw

1f

n

c 1S

c

f

n

c 1Pc / Pw

Sw m

r 1Sr

Sr 1gr

nr

c 1Sc

gr nr

c 1Gc / Gr

(b) the annual collective effective dose (i.e. the sum of the annual collective dose from external irradiationand the committed collective dose from intakes inthat year), S;

(c) the collective dose distribution ratio, SRE, for valuesof E of 15, 10, 5, and 1 mSv; and

(d) the individual dose distribution ratio, NRE, for valuesof E of 15, 10, 5, and 1 mSv.

E. ESTIMATION OF WORLDWIDEEXPOSURES

70. Inevitably, the data provided in response to theUNSCEAR Survey of Occupational Radiation Exposureswere insufficient for estimating worldwide levels of dose.Procedures were therefore developed by the Committee toderive worldwide doses from the data available forparticular occupational categories. Two procedures were

developed, one for application to occupational exposuresarising at most stages in the commercial nuclear fuel cycleand the other for general application to other occupationalcategories.

71. In general, the reporting of exposures arising in thecommercial nuclear fuel cycle is more complete than thatof exposures arising from other uses of radiation. Thedegree of extrapolation from reported to worldwide dosesis, therefore, less, and this extrapolation can be carried outwith greater reliability than for other occupationalcategories. Moreover, worldwide statistics are generallyavailable on capacity and production in various stages of the commercial nuclear fuel cycle. Such data provide aconvenient and reliable basis for extrapolating toworldwide levels ofexposure. Thus, the worldwide annualcollective effective dose, Sw, from a given stage of thenuclear fuel cycle (e.g. uranium mining, fuel fabrication,or reactor operation) is estimated to be the total of annualcollectiveeffectivedosesfromreportingcountriestimesthereciprocal of the fraction, f, of world production (uraniummined, fuel fabricated, energy generated, etc.) accountedfor by these countries, namely,

where Sc is the annual collective dose from country c andn is the number of countries for which occupationalexposure data have been reported. The fraction of totalproduction can be expressed as

where Pc and Pw are the production in country c and in theworld, w, respectively.

72. The annual number of monitored workers worldwide,Nw, is estimated by a similar extrapolation. Because the data

aremore limited, the worldwide distributionratios, NRE,w andSRE,w, are simply estimated as weighted averages of thereported data. The extrapolations to worldwide collectiveeffective doses and numbers of monitored workers and theestimation of worldwide average distribution ratios areperformed annually. Values of these quantities have beenaveraged overfive-yearperiods, andtheaverageannual values

are reported in this Annex.73. For occupational exposures to radiation from practicesother than operations of the nuclear fuel cycle, statistics arenotsoreadilyavailableon theworldwidelevel ofthe practicesor their distribution among countries. In thesecasesa simplerand, inevitably, less reliable method ofextrapolation has to beused. A varietyofapproachesarepossible (e.g. scaling bysizeof population, by employment in industrial or medicalprofessions, or by some measure of industrial output). In theend, it seemed to be most practical and reasonable to extra-polate on the basis of GDP [U14]. Several considerationsinfluence the choice of this quantity in preference to others,

notably the availability of reliable worldwide statistics onGDPs and their potential for general application; the latter isa consequence of the expectation that GDP is reasonablycorrelatedwith both thelevelofindustrialactivityandmedicalcare in a country, characteristicsunlikelyto bereflectedin anyother singlequantity.Tomakethe extrapolation morereliable,it is applied not globally but separately over particulargeographic or economicregions, followed bysummation overthese regions. This results in extrapolations of available datawithin groups of countries with broadly similar levels of economic activity and allows for general geographicalcomparisons.

74. The worldwide annual collective effective dose forother uses of radiation, is estimated as

where

where Sr is the annual collective effective dose ingeographic or economic region r, nr is the number of countries in region r for which occupational exposure datahave been reported, m is the number of regions, and gr isthe fraction of GDP of region r, represented by thosecountries for which occupational exposure data areavailable and is given by

where Gc and Gr are the GDPs of country c and region r,respectively.

75. Theaboveequationsare appliedto estimatecollectivedoses for those regions for which occupational exposure


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Sr Gr n

c 1Sc /

n

c 1Gc

data are available for at least one country within theregion. For those regions for which nodata for anycountrywere reported, a modifiedapproachfor estimating regionalcollective dose is adopted:

76. Theannualnumber ofmonitoredworkersworldwide,Nw, is estimated by the same procedure. The worldwidedistribution ratios are estimated as for operations of thenuclear fuel cycle, but the averaging is performed on aregional basis before summing over all regions. Thenumber of measurably exposed workers worldwide, Mw, isestimated in a similar manner.

II. THE NUCLEAR FUEL CYCLE

77. A significant source of occupational exposure is theoperation of nuclear reactors to generate electrical energy.This involves a complex cycle of activities, including themining and milling of uranium, uranium enrichment, fuelfabrication, reactor operation, fuel reprocessing, wastehandling and disposal, and research and development

activities. Exposures arising from thispracticewere discussedandquantifiedin theUNSCEAR 1972 [U8], 1977 [U7], 1982[U6],1988[U4],and 1993[U3] Reports,with comprehensivetreatment in the UNSCEAR 1977 and 1982 Reports. Incomparison with manyothersourcesofexposure, thispracticeis well documented, and considerable quantities of data onoccupational dose distributions are available, in particular forreactor operation. This Annex considers occupationalexposure arising at each main stage of the fuel cycle. As thefinal stage, treatmentand disposal ofthemain solid wastes, isnot yet sufficiently developed to warrant a detailedexamination of potential exposures, it is given only verylimited consideration. However, for the period underconsideration,occupationalexposuresfromwastedisposalarenot expected to significantly increase the sum of the dosesfrom the other stages in thefuel cycle. For similar reasons, noattempt ismadeto estimate occupational exposuresduringthedecommissioning of nuclear installations, although this willbecome an increasingly important stage.

78. Each stage in the fuel cycle involves different types of workers and work activities. In some cases, e.g. for reactoroperation, the data are well segregated, while in others theavailabledataspan severalactivities, e.g.uranium mininganduranium milling. Where the data span a number of activities,

this is noted in footnotes to the tables. The data onoccupational exposures for each of the activities are derivedprimarily from the UNSCEAR Survey of OccupationalRadiation Exposures but also fromother sources, particularlythe Information System on Occupational Exposure of theOECD/NEA [O4, O5].

79. For each stage of the fuel cycle estimates are made of the magnitude and temporal trends in the annual collectiveand average individual effective doses, the numbers of monitored workers, and the distribution ratios. The collectivedoses are alsoexpressedin normalized terms, that is, per unitpractice relevant to the particular stage of the cycle. Foruranium mining and milling, fuel enrichment, fuelfabrication, and fuel reprocessing, the normalization isinitially presented in terms of unit mass of uranium or fuel

produced or processed; an alternative way to normalize is interms of the equivalent amount of energy that can be (or hasbeen)generated bythe fabricated(or enriched)fuel.Thebasesfor the normalizations, namely, the amounts of mineduranium, the separative work during enrichment, and theamount offuel required to generate a unit ofelectrical energy

in variousreactor types, are given in Annex C, Exposures tothepublic fromman-made sourcesof radiation . For reactors,the data may be normalized in several ways, depending onhowtheyare to be used. In this Annex, normalized collectivedoses are given per reactor and per unit electrical energygenerated.

80. To allow proper comparison between the doses arisingat different stages of the fuel cycle, all the data are ultimatelypresented in the same normalized form, in terms of theelectrical energy generated(or the amount of uranium minedor fuel fabricated or reprocessed, corresponding to a unit of energy subsequently generated in the reactor), which is theoutput of the nuclear power industry. This form of normalization is both valid and useful when treating dataaccumulated over a large number of facilities or over a longtime. It can, however, be misleading when applied to data fora single facilityfor a short time period; this is because a largefraction of the total occupational exposure at a facility arisesduring periodic maintenance operations, when the plant isshut down and not in production. Such difficulties are,however, largely circumvented in this Annex, since the dataare presented in an aggregated form for individual countriesand averaged over five-year periods.

81. Various national authorities or institutions have useddifferent methods to measure, record, and report theoccupational data included in this Annex. The main featuresof the method used by each country that responded to theUNSCEAR Survey of Occupational Radiation Exposures aresummarized in Table 2. The potential for such differences tocompromise or invalidate comparisons between data isdiscussed in Section I.A.3. The reported collective doses andthe collective dose distribution ratios are largely insensitive tothe differences identified in Table 2, so these quantities cangenerally be compared without further qualification. Theaverage doses to monitored workers and the numberdistribution ratios are, however, sensitive to decisions andpractice onwhoin a workforceistobemonitored. Differencesin these areas could not be discerned from responses to theUNSCEAR Surveyof Occupational Radiation Exposures, so


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theycannot be discerned from Table2. However, becausethemonitoring of workers in the nuclear power industry is ingeneral fairly comprehensive, comparisons of the averageindividual doses (and number distribution ratios) reportedhere are judged to be broadly valid. Nonetheless, it must berecognized that differences in monitoring and reportingpractices do exist, and theymay, in particular cases, affect thevalidity of comparisons between reported data; to the extentpracticable, where such differences are likely to be importantthey are identified.

A. URANIUM MINING AND MILLING

82. Uranium is used for military, commercial, and researchpurposes. It is widely distributed in the earths crust, andmining is undertaken in over 30 countries [O3]. Commercialuraniumuse is primarilydeterminedbythe fuel consumptionin nuclearpowerreactorsandnuclear research reactors andbythe inventory requirements of the fuel cycle. Uraniumrequirementsfor power reactorscontinue to increasesteadily,while therequirementsfor research reactorsremainmodest bycomparison. The annual production of uranium in variouscountries in the years 1990 1997 is given in Annex C, Exposures to the public from man-made sources of radiation , and more detailed information can be found in anOECD/NEA publication [O3].

83. The mining of uranium is similar to that of any othermaterial. It mainly involves underground or open-pittechniques to remove uranium ore from the ground, followedby ore processing, usually at a location relatively near themine.Themillingprocessinvolves thecrushing andgrindingof raw ores, followed by chemical leaching, separation of uranium from the leachate, precipitation of yellowcake [K4],and drying and packaging of the final product for shipment.In response to thedeclining priceofuranium, theemphasisinrecent years has been on lower-cost methods for extractinguranium [O3]. The percentage of conventional undergroundminingwasreducedfromabout 55%toabout 45%from1990to 1992. The lower-cost methods are open-pit mining, in situleaching, and by-product production (e.g. from the mining of other minerals such as gold). The percentage fromconventional open-pit mining increased during this period,

from 38% to 44%; that from in situ leaching from 5.7% to9.1%; and that from by-product production from 1.1% to2.2%. In 1992, there were55 operatinguranium mines in theworld in over 21 countries, with 32% of the productioncoming from Canada alone. About 84% of the world'sproduction came from only 12 countries: Australia, Canada,France,Kazakhstan,Kyrgyzstan, Namibia,Niger,theRussianFederation, South Africa, Tadjikistan, Uzbekistan, and theUnited States [G2] (see Table 28 of Annex C, Exposures tothe public from man-made sources of radiation , for annualproduction of uranium in other years between 1990 and1997).

84. The mining and milling of uranium ores can lead toboth internal and external exposures of workers. Internalexposure may arise from the inhalation of radon gas and its

decay products and radionuclides in ore dust. The extent of internal exposure will depend on many things, including theoregrade, the airborne concentrations of radioactiveparticles(which vary depending on the type of mining operation andthequalityofventilation), and theparticlesize distribution.Inunderground mines, the main source of internal exposure islikely to be radon and its decay products. Because of theconfined space underground and practical limitations to thedegree of ventilation that can be achieved, the total internalexposure is ofgreater importance in underground mines thanin open-pit mines. In open-pit mines, the inhalation of radioactive ore dusts isgenerally the largest source ofinternalexposure, although the doses tend to be low. Higher dosesfrom this source would be expected in the milling of the oresand production of yellowcake.

85. With the emphasis on low-cost uranium production,new projects are expected to focus on high-grade un-conformity and sandstone-type deposits. These may be

amenable to in situ leaching techniques, but where under-ground mining is used, exposures of workers are likely tocontinue to be of concern. In future surveys there will be aneed toconsider theexposuresthat ariseduring therehabilita-tion of old mining operations. For example in Germany,where uranium mining is no longer undertaken, annualexposures to workers due to the removal of uranium miningresidues are estimated for 1995 to be distributed as follows:1 6 mSv, 1,250 workers; 6 20 mSv, 230 workers; and>20 mSv, no workers [S2]. The exposures result fromexternalradiation,inhalationofradioactivedust particles, andinhalation of radon progeny.

86. Exposure data for mining and milling of uraniumores from the UNSCEAR Survey of OccupationalRadiation Exposures for 1990 1994 are given in Tables 3and 4, respectively; and trends for the four periods from1975 are given in Figure II. The questionnaire askedrespondents to use a conversion factor for exposure toradon decay products of 5 mSv per WLM, the valuerecommended by ICRP [I12].

87. Over the threeprevious five-year periods the averageannual amountsofuranium mined worldwide were 52, 64,and 59 kt, a reasonably constant level of production, with

by far the largest part mined underground. As has alreadybeen mentioned, therehas more recently been a moveawayfrom underground mining and a reduction in the amountmined. For the 1990 1994 period, the average annualamount mined was 39 kt, a reduction of about one third.Theyear-on-year figures showed a steadydownward trend,from 49.5 kt in 1990 to 31.6 kt in 1994. During this perioda number of countries, including Bulgaria, Germany, andSlovenia, reported that mining operations had ceased,although some exposures continued from measures to treatthe closed-down mining operations. These trendswould beexpected to affect both the magnitude of the collectivedoses and the dose profiles, and indeed they do so.

88. The data set for 1990 1994 is smaller than for thepreceding period, 1985 1989, with data from10 countries as


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Figure II. Trends in numbers of monitored workers,doses to workers, and collective doses for mining,milling, enrichment and fuel fabrication.

opposed to 14 countries, respectively. The 1985

1989 datawere dominated by underground mining data from SouthAfrica, which accounted for some 70% (82,000) of the totalreportedmonitored workers(114,000)and55%(278manSv)ofthe reported collective dose(507manSv). China also madean important contribution to the 1985 1989 data, with areported collective doseof114man Sv, some 22% ofthe totalreported. The lack of data for 1990 1994 from South Africaand China (and, to a lesser extent, from India and the UnitedStates) distorts any extrapolation to arrive at a world figure.For the earlier periods the extrapolation for the number of monitored workers and collective dose worldwide was basedon the ratio between the total amount of ore produced by thereportingcountriesand total worldproduction.Employing thesame approach to the 1990 1994 period would give aworldwide monitored population of 28,000 and an average

annual collective effective dose of 140 man Sv. Both of theseestimatesare an order of magnitude less than for 1985 1989.The Committee regarded this as a significant underestimateand has instead chosen to make estimates for thosecountriesthat had not reported for 1990 1994 but that did report for1985 1989, before extrapolating on the basis of worldwideproduction of uranium ore. This approach has the benefit of ensuring that major contributors such as South Africa andChina are more adequately accounted for. The estimates forthese countries (shown in square brackets in Table 3) arebased on the average trends for countries reporting for both1985 1989 and 1990 1994 and take into account the bestestimates of uranium ore production. On this basis, theaverage annual number of monitored workers worldwide fellfrom260,000 in 1985 1989 to 69,000 in 1990 1994. For theprevious two periods the numbers had been 240,000 and310,000. This reduction by a factor of3 or4 isalsoseen in thevalues for average annual collective effective doses. For thethree previous periods the worldwide estimates were 1,300,

1,600 and 1,100 man Sv, but for 1990

1994 the value was310 man Sv. Similarly, the averagecollectivedose per unit of uranium extracted had been 26, 23, and 20 man Sv per kt forthe three previous periods and wasdown to 7.9 man Svper ktfor 1990 1994; the corresponding values for averagecollectivedose per unit energy were 5.7, 5.5, and 4.3 man Svper GWa, falling to 1.7 man Svper GWa for 1990 1994 (seeFigure III). However, the estimated average annual effectivedose, 4.5 mSv, was marginally higher than for theimmediatelyprecedingperiod, when it was4.4mSv.With thedoses from underground mining dominating the collectivedose and the known difficulties in reducing individual doses,the data would be consistent with a worldwide reduction inunderground mining activity coupled with more efficientmining operations.

Figure III. Normalized collective effective dose perunitenergyproduction for mining, milling, enrichment andfuel fabrication.

89. Data on exposure to workers from uranium millingwere provided from only two countries, Australia andCanada, and are given in Table 4. In line with theirreductions in mining, both countries show significantreductions in the number of monitored workers and thecollective dose. It is difficult toextrapolateworldwide fromthese data, but crude estimates can be made. Asin previous


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UNSCEAR reports it is assumed that the amount of uranium milled is equal to the amount mined. Thecombined data for the two countries reporting show areduction by a factor of about 4 in the average annualcollective dose and about a factor of 2 in the number of monitored workers relative to 1985 1989. These factorsare in line with the trends for uranium mining, and itwould seem appropriate to apply themto deriveworldwideestimates for 1990 1994. Doing so leads to worldwideestimates for average annual monitored workers of 6,000compared with 12,000, 23,000, and 18,000 in each of thethree previous periods; to an average annual collectiveeffective dose of 20 man Sv compared with 124, 117, and116 man Svin eachofthe three previous periods; and to anaverage annual effective dose of 3.3 mSv compared with10.1, 5.1, and 6.3 mSv in each of the three previousperiods.

B. URANIUM ENRICHMENT ANDCONVERSION

90. Uranium conversion is the process by which UO2,which is the chemical form of uranium used in mostcommercial reactors, isproduced for the fabricationofreactorfuel. In reactors that use fuel slightly enriched in 235U(generally about 3%; natural uranium contains about 0.7%235U), uranium from the milling process must be enrichedbefore fuel fabrication. Thus, the U3O8 from the millingprocess is converted to UO2 by a reduction reaction with H2.The UO2 is then converted to UF4 by the addition of hydrofluoric acid (HF), and then to UF6 using fluorine (F2).This gaseous product, UF6, is then enriched in 235U. Most of this was done by the gaseous diffusion process, butincreasingly, gaseous centrifuge techniques are being used.Oncetheenrichment processhasbeencompleted, theUF6 gasis reconverted into UO2 for fuel fabrication. Occupationalexposures occur during both the conversion and enrichmentstages, with, in general, external radiation exposure beingmore important than internal radiation exposure. Workersmay, however, be exposed to internal radiation, particularlyduring maintenance work or in the event of leaks.

91. During 1990 1994 most enrichment services came

from fivesuppliers:Department ofEnergy(UnitedStates),Eurodif (France), Techsnabexport (Russian Federation),Urenco (Germany, Netherlands and United Kingdom) andChina. (Entities in those same countries, plus Canada,offered services for the conversion process that precedesenrichment.) The enrichment capacity of these and a fewother small producers has been estimated at between 32and 35 million separative work units (MSWu) per annumduring 1990 1994 compared with demand of between 23and 27 MSWu [O8, O9]. Exposure data for 1990 1994 aregiven for Canada, France, Japan, the Netherlands, SouthAfrica, the United Kingdom, and the United States inTable 5. With three exceptions thedataare for enrichmentby the diff

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