purpose and practice of radiation monitoring

IEE REVIEW

Purpose and practice of radiation monitoringT.F. Johns, B.Sc, F.lnst. P.

Indexing terms: Instrumentation and measuring science, Hazards, Radiation

Abstract: Following a brief description of the biological effects of exposure to ionising radiation, and anexplanation of the general philosophy of those engaged in the practice of radiation protection, a detailedaccount is given of the role which monitoring of various types plays in that practice. Two quite differenttypes of monitoring are involved — for radiation from sources external to the body, and monitoring forradioactive materials which may get inside the body, or are already there. In each'case this involves moni-toring of the working environment, of the worker, and monitoring of the general environment to check forthe effects of any deliberate or accidental releases of radioactive materials which might affect the generalpublic. The special arrangements which may be needed following an accident or emergency are also brieflydiscussed.

1 Introduction

Man has always been exposed to ionising radiation, both tocosmic radiation from space, and to radiation from naturallyoccurring radioactive materials in the earth, the sea, the air,and even in his own body. Until the beginning of this centuryhe was blissfully unaware of these facts. With the increasinguse of ionising radiations and radioactive materials in industry,research and medicine, and with the development of atomicbombs, it has, however, become apparent during this centurythat exposure to ionising radiation is potentially harmful;and so it has been necessary to develop techniques forassessing or measuring, and hence controlling, the radiationexposure of individual workers and of the public at large.This involves a considerable variety of monitoring techniques.The objectives of this review are

(a) to give an elementary description of the biologicaleffects of exposure to ionising radiation

(b) to explain the general philosophy adopted by thoseconcerned with the practice of radiation protection

(c) to discuss, in some detail, the role which monitoringof various types plays in the practice of radiation protection;this will be limited to everyday routine monitoring to measureand control the doses received from man-made radiations,and will not include reference to such matters as, for example,research investigations into the origins, nature and variabilityof cosmic rays.

Although our knowledge and understanding of some of thebiological effects of ionising radiation is not as complete as wewould wish, we do have much more detailed knowledge ofthose effects than we have about the effects of many otherhazardous substances in our environment; and the practicesof radiation protection (including radiation monitoring) aremuch more highly developed and sophisticated than thecorresponding practices currently used in monitoring forother toxic substances. In particular, radiation monitoringtechniques are very much more sensitive than those whichcan be used for assaying any other material.

Many people have come to believe that ionising radiation isuniquely dangerous, partly because we cannot detect it withour ordinary senses. In reality, however, the adoption of thepractices to be described in this review has led to a situationin which the risks of harmful effects from radiation exposureare far smaller than the corresponding risks from many otherphenomena which cause little or no concern.

Paper 1749A, received in final form 14th September 1981. Com-missional IEE ReviewThe author is with the Radiological & Safety Division, Room 105,Building A40, AEE Winfrith, Dorchester, Dorset DT2 8DH, England

2 Different types of irradiation

Before discussing the effects of exposure to radiation, it isdesirable to discuss briefly the different ways in which suchexposure can occur, and to discuss also the various types ofradiation which may be involved.

Some of the materials in our natural environment, includingsome of a primordial nature and others produced by theinteractions of cosmic rays with our atmosphere, are radio-active and emit ionising radiation, i.e. radiation which has thecapacity to produce ions (positively and negatively chargedproducts) in any material through which it passes. If thehuman body is exposed to such radiation, charged particlesare formed inside the cells of the body; and if the number ofcharged particles is sufficiently great, the biochemical pro-cesses in the cells may be upset, and biological damage mayresult.

Nowadays, in addition to naturally occurring radioactivematerials, a large number of artificially produced radioactivematerials are used in industry. Generally speaking, the types ofradiation which they emit are similar to those emitted bynaturally occurring radioactive materials. There are also anumber of machines which produce, and/or emit, ionisingradiation. The most important and well known are X-raymachines, much used in medicine and industry. X-rays areproduced by the absorption and slowing down of energeticelectrons.

There are various types of radiation with which we shallbe concerned in this review:

(a) Alpha particles, which produce very intense localionisation in materials through which they pass, as a result ofwhich they can be stopped very easily, e.g. by a thin piece ofpaper or a few centimetres of air

(b) Beta particles, which produce fairly intense ionisation(although less so than alpha particles), and, as a consequence,also have a fairly short range in tissue or other materials; theyare stopped by a few millimetres of solid or liquid material,or by a few metres of air

(c) X-rays and gamma rays, electromagnetic radiation;these rays produce far less ionisation per unit length of paththan alpha or beta rays, and, as a consequence, they have amuch greater range in tissue or other materials. They canpenetrate well into the human body, and require quite thickshields of most materials to stop them

(d) Neutrons, which only produce ionisation indirectly, asa result of interactions with the nuclei of atoms of materialsthrough which they pass; e.g. a proton (another intenselyionising particle) can be produced if a neutron collides withthe nucleus of a hydrogen atom. Neutrons, like gamma raysand X-rays, can penetrate most materials, including the human

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body, fairly easily; but it should be noted that they can onlydo so because the amount of ionisation produced per unitlength along their path is much smaller than that producedby alpha or beta rays.

The tissues of the body can be exposed to radiation in twodifferent ways, either to radiation from radioactive sourcesor machines outside the body, or to radiation from radioactivematerials which have somehow got inside the body. We referto these two types of irradiation as external and internalirradiation, respectively.

Because alpha particles are stopped by a few centimetresof air, or by the dead layer of skin on the outside of thebody, alpha emitters do not constitute a source of externalirradiation. External beta emitters cause irradiation only ofthe skin and superficial tissues of the body, whereas X-rays,gamma rays and neutrons from sources external to the bodycan also cause irradiation of the deeper-lying and more sensi-tive body tissues.

When radioactive materials get inside the body, in a sensethe situation is reversed. Alpha emitters inside the bodyproduce very intense local irradiation near the point wherethey are located, thus increasing the chance of biologicaleffects. Beta emitters may also cause high local irradiationof one organ or part of the body; but gamma-emitters producemore diffuse irradiation of the whole body.

Internal irradiation is often different from external ir-radiation in a quite different respect. When radioactive mat-erials get inside the body, different parts of the body oftenreceive grossly nonuniform doses, because certain elementsconcentrate in certain parts of the body. For example radio-nuclides of iodine cause very much greater irradiation of thethyroid than of other parts of the body, because they concen-trate in this small organ.

3 Biological effects of ionising radiationIonising radiation can produce two types of injury:

(a) acute effects, which become apparent within a few daysor weeks of the exposure

(b) delayed effects, which may only become apparentseveral decades after the exposure.In the latter case we have to distinguish between somaticeffects, which may occur in the body of the exposed person,and genetic effects, which may occur in a succeeding gener-ation or generations.

Acute somatic effects were first observed soon after thediscovery of X-rays by Rontgen in 1895. In the first fewyears after that discovery, most people were not aware ofthe dangers, and often deliberately exposed their handsand other parts of their bodies to X-ray beams. More recently,since the dangers of radiation have been clearly appreciated,a few people have occasionally been accidentally exposed tolarge doses of radiation, for example by exposing some partof their body (such as a finger) to X-rays, by unwittinglycarrying a radiation source in their pocket, or by going intoan area where, unknown to them, a large radiation sourcewas unshielded. As a result of this experience, we have arather precise knowledge of the acute effects of exposureto large doses of ionising radiation [1].

The severity of these acute effects (as well as the prob-ability of induction of late or delayed effects — see below) isrelated to the amount of energy which the ionising radiationdeposits per unit mass of tissue in the body; this is referred toas the 'dose', or more precisely as the 'absorbed dose'. The SIunit of absorbed dose is the gray (Gy), which corresponds tothe deposition of 1 J/kg of tissue. The unit most commonlyreferred to in the literature of the last 30 years is the rad,which is equal to 0.01 Gy. The mean dose received by eachperson in the UK per annum from cosmic rays, terrestrial

gamma rays and long-lived naturally occurring radioactivematerials in the body is about lOOmrad, or 1 mGy.

If the whole body is irradiated, no obvious effects areobserved for doses of less than about 0.25 Gy. For doses inthe range 0.25 — 1 Gy, received in a short time, a temporaryreduction in the number of white cells in the blood isobserved, the reduction increasing as the dose increases; andthere may be nausea (feeling of sickness) and diarrhoea. Dosesof a few grays in a short time lead to serious effects of twokinds:

(a) There is a marked reduction in the count of white cellsin the blood; as a result the affected individual is markedlyless able to fight infection, and unless protected may die fromsuch infection.

(b) The cells lining the walls of the stomach and intestinesare damaged, and there may be internal bleeding with seriousconsequences.

Doses of 4—6Gy are likely to have serious effects within afew weeks, and for a dose of about 6 Gy, about half of thoseirradiated will die. Larger doses (> 20 Gy) have serious effectson the central nervous system, and death may occur morequickly — within a day or two.

When there is heavy local irradiation of one part of thebody, the effects are rarely lethal but may nevertheless bevery painful and serious. Doses of a few grays lead to red-dening of the skin, and somewhat larger doses cause radiationburns. These are in some ways similar to heat burns, but areoften slower to heal because of damage to deeper tissues. Verylarge doses can lead to necrosis and the need for amputation.

These acute effects occur only if the irradiation takes placein a short time — in minutes or hours. The same doses spreadover months or years will not produce acute effects, since thebody's repair mechanisms can come into play.

Large doses, even if spread over months or years, can,however, occasionally lead to delayed effects of quite adifferent kind, namely to the development of cancer. One ofthe most dramatic indications of this came in the mid-1920sfrom the use of radium in the luminising industry. Radiummixed with zinc sulphide was used for luminising the dialsof instruments, clocks and watches. Many women employedin the luminising industry, especially during the 1914—1918war, used to lick the brushes used for the luminising, in orderto 'point' the brush, and hence get a neat job; and they tookinto their bodies what we would now consider to be grosslyexcessive amounts of radium. Some of that radium remainedin their bodies, much of it in their bones, for many years, andas a consequence some of these women developed bone cancerup to 40 years later. Much of our knowledge of the effects ofradiation comes from very careful and detailed retrospectivestudies of such groups of people [2].

Evidence that large doses of radiation from sources externalto the body can also lead to the induction of cancer manyyears later has been obtained primarily from two epidemio-logical studies:

(a) studies of a group of men who were deliberatelyexposed to large doses of radiation, as part of the treatment ofa spinal condition called ankylosing spondilitis [3]

(b) studies of the survivors of the atomic bombs droppedon Hiroshima and Nagasaki [4] in 1945.

In both cases it was shown that the irradiated groups wererather more likely to develop cancer several years or decadesafter the event, than if they had not been irradiated. Thefollowing points should however be noted:

(a) The types of cancer observed are exactly the same asthose which occur in all populations, either apparently spon-taneously or due to other carcinogenic insults; only theincidence is different.

82 IEEPROC, Vol. 129, Pt. A, No. 2, MARCH 1982

(b) Even in these groups of people who received very largedoses of radiation, only a very small proportion developedcancers which could be ascribed to radiation, and it requiredvery careful and detailed statistical studies on the irradiatedgroups, and on matched reference groups, to demonstratethe effects.

(c) This is an example of a so-called stochastic effect(Reference 5, p. 2). A few persons in the irradiated grouplater developed cancer, but the vast majority were completelyunharmed. In contrast, the acute effects of large doses ofradiation are examples of nonstochastic effects. Each indi-vidual exposed to a sufficiently large dose in a short timesustains an injury of similar severity to that of any otherindividual exposed to the same dose; there is no element of'chance' — real or imagined. For the nonstochastic effects,increasing the dose increases the severity of the injury; whereasfor stochastic effects, the severity remains unchanged, but theprobability of sustaining the injury increases with dose.

It is not absolutely clear whether, and to what extent,exposure to small doses of radiation, similar to those sustainedby the majority of people who work with radiation sources,also causes an increased risk of cancer. A number of retro-spective epidemiologcial studies have been made, similarto those mentioned above on the ankylosing spondolitics andthe Japanese bomb survivors, but on groups of people knownto have been exposed to much smaller doses of radiation.Some of these have claimed to demonstrate definite corre-lations between cause of death and radiation dose [6]; othershave failed to do so [7]. The methodology of some of thestudies has been strongly criticised [8]. These conflictingresults arise, in part, because of the inherent difficultiesof such studies. Even at very high doses, difficulties areencountered because the increased incidence of cancer iscomparable with known natural variations in its incidence.The incidence of 'natural' cancer in populations is both largeand variable; it varies with location, personal habits, socialclass, sex, age, and a host of other variables. It is difficult tofind suitable control groups. At lower doses, where the magni-tude of the effect being sought is obviously smaller, all ofthese difficulties are compounded. Clear evidence on whetheror not low doses do produce an increased incidence of cancerwill probably not become available until the completion of anumber of prospective studies, now taking place, on verylarge groups of people over a long period of time.

The situation regarding genetic effects is rather similar.There is little or no direct evidence that small doses of radia-tion produce genetic effects in man. However, there is abun-dant evidence that radiation causes genetic effects in plantsand animals, and so it is prudent to assume that geneticeffects may be produced, even for small doses in man. Forboth the delayed somatic and genetic effects, it is generallyassumed that the risk of harm is proportional to the radiationdose. The magnitude of these risks is discussed in Section 4.1.

Some types of radiation are more likely than others tocause biological damage, for a given dose. The risk of bio-logical damage is greater for those types of radiation (e.g.alpha particles) which produce a large number of secondarycharged particles per unit length of their track, as a greaternumber of charged particles are then formed in one cell or ina small group of adjacent cells. In order to take account ofthis effect, a 'quality factor' is ascribed to each type of radi-ation (Reference 5. pp. 4-5) , and the 'dose equivalent' corre-sponding to a given dose of a particular type of radiationis calculated as follows:

dose equivalent = dose x quality factor

DE = D x QF

DE is the dose of gamma rays which produces the samebiological effect as a dose D of this other type of radiation;gamma rays are (obviously) assumed to have a quality factorof 1.

It is then to be expected that for a mixture of differenttypes of radiation, the resulting biological damage will besimply related to the total dose equivalent of the mixture, i.e.to

X(dose)i. (QF)i

where (dose)i is the dose due to one particular type of radia-tion, and (QF)i is the corresponding quality factor.

The quality factor is simply a number varying from 1, forgamma rays, X-rays and beta rays, to 20 for alpha rays. If thedoses are in grays, the corresponding dose equivalent is insieverts (Sv). Until recently, dose equivalents were measuredin rems (100 rems = 1 Sv). In this review, the term 'doseequivalent' will be abbreviated to 'dose', except in circum-stances where this could lead to ambiguity.

4 Radiation protection

The dangers of exposure to ionising radiation were appreciatedsoon after the discovery of X-rays, long before the use ofradiation sources and radioactive materials in industry andmedicine became widespread. And, although there was adelay of 30 years in recognising the dangers of taking sub-stances like radium into the body, the practice of radiationprotection, or health physics as it is often called, was devel-oped soon enough to prevent large numbers of people beingkilled or injured in the industry. This is in sharp contrast tothe position in many other industries.

The aims of radiation protection are:(a) to ensure, as far as possible, that people are not exposed

to doses large enough to cause acute radiation injury(b) to control the doses of radiation which workers inevi-

tably receive when working with radioactivity, so that theassociated risks of the induction of delayed effects, i.e. canceror genetic damage, are limited to acceptably small values

(<?) to limit the doses of radiation received by membersof the general public to even lower values (see below).

The first objective is achieved by the use of shielding (oftenvery massive shielding), by use of containment, by the use ofcareful design, interlocks and operating procedures to reducethe risk of accidents, and by ensuring that if and when acci-dents do occur, the effects are ameliorated by the adoption ofappropriate administrative procedures. In this country therehas been a remarkable degree of success in achieving thisobjective, and cases of acute radiation injury are extremelyrare.

However, workers in these industries are inevitablyexposed to some radiation, and those who are engaged in thepractice of radiation protection spend most of their timemeasuring dose rates in the working environment, in the moregeneral environment, and the doses received by individuals,and ensuring that these doses are controlled within appropriatestatutory limits. In almost all countries, the dose limits incor-porated into legislation are based on the recommendations ofthe International Commission on Radiological Protection(ICRP). The ICRP has for many years studied in depth all ofthe available information on the effects of radiation exposure,and has from time to time revised its recommended doselimits to take account of new knowledge.

Current UK legislation in the field of radiation protectionis based on ICRP recommendations made between 1958 and1965. That part concerned with the protection of workersapplies only to those working in factories. New legislation is

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currently being drafted, which will apply to all workers; thisnew legislation will be in line with more recent ICRP rec-ommendations (which we shall shortly discuss in some detail)and a resulting directive from the EEC. The changes intro-duced in the new legislation are unlikely to have much effecton methods of monitoring, although they will, in some cases,affect the treatment of and the interpretation of the results.Almost everything in this review is compatible with thelatest ICRP recommendations, and is likely to be compatiblewith the forthcoming UK legislation.

The main features of the current ICRP system of doselimitation [5] are set out in paragraph 12 of ICRP Publication26, from which the following is taken:

(a) No practice shall be adopted unless its introductionproduces a positive net benefit.

(b) All exposures shall be kept as low as reasonably achiev-able, economic and social factors being taken into account.

(c) The dose equivalent to individuals shall not exceed thelimits recommended for the appropriate circumstances by theCommission.

Recommendation (b) is sometimes referred to as the ALARAprinciple (as low as reasonably achievable). However, it shouldbe noted that the word used is 'reasonably', and that in tryingto reduce doses one should take economic and other con-siderations into account; this is in contrast with recommen-dation (c), which has to be complied with irrespective of cost.In making these recommendations, the ICRP has consideredit prudent to assume that there is a certain risk that anyexposure to radiation, however small, may lead to the delayedeffects already discussed; i.e. it considers that there is nothreshold below which radiation exposure can be consideredcompletely safe.

After careful consideration of the available evidence, theICRP concluded that if 1 000 000 people were each exposedto a dose equivalent of lOmSv (1 rem), this would eventuallylead to the induction of about 125 additional cases of cancer,and 40 additional cases of genetic effects. Its conclusions aresimilar to those of the United Nations Scientific Committeeon the Effects of Atomic Radiation (UNSCEAR) in its 1977report [9]. The assumed risk figures per unit dose at lowdoses are a factor of 2 or 3 smaller than the risks observedfor doses greater than 1 Sv (Reference 9, p. 414, para. 318).

From these risk figures, the ICRP derived values for themaximum permissible doses for individual radiation workersand for members of the general public — see recommendation(c). The limits are 50mSv per year for radiation workers, and5 mSv per year for individual members of the general public.Although the figures are expressed as annual limits, the under-lying purpose is to limit the lifetime risk. The figure forradiation workers (together with the application of theALARA principle) is chosen to ensure that the risks to healthfrom working in the 'radiation industries' are no greater thanthe risks of working in other industries generally consideredto be safe, whereas the limit for members of the generalpublic is chosen to ensure that the associated risk is no greaterthan that associated with, for example, the use of publictransport.

The limits quoted above apply where the whole body isuniformly irradiated. If the irradiation of the body is non-uniform, the ICRP recommends the calculation of an effectivedose equivalent

H = 2wTHT (1)

where HT is the dose equivalent in one organ or part of thebody, and wT is a weighting factor representing the fractionof the total stochastic risk from whole-body irradiation whichis associated with the irradiation of that particular organ or

Table 1: Recommended valuesof ivy

Tissue wj>

Gonads 0.25Breast 0.15**Red bone marrow 0.12Lung 0.12Thyroid 0.03Bone surface 0.03Remainder 0.30*

*A value of 0.06 is applicableto each of the five organs ortissues of the remainder of thebody receiving the highestdose equivalent. Other riskscan be neglected.**The figures given are aver-age figures, for all ages andboth sexes.

part of the body. H is the whole-body dose equivalent whichwould carry the same risk. The recommended values of vvT areshown in Table 1.

Since vvT < 1, it follows that if only one part of the body isexposed, the dose limit for that organ is greater than the doselimit for whole-body irradiation. In order to avoid the possi-bility of nonstochastic effects, however, no part of the bodymay be exposed to more than 500 mSv per year. This require-ment also sets a limit for the dose to those parts of the body,e.g. the skin, or hands and feet, in which the risk of stochasticeffects is negligibly small.

In addition to the above 'basic' limits, the ICRP has calcu-lated a number of 'derived' limits, e.g. for the amount of aparticular radioactive material which a person may inhaleor ingest in a year, without danger of exceeding the basic doselimits. These derived limits are discussed in more detail inSection 5.2. The quantity of a particular radioactive materialwhich may be inhaled or ingested is (inversely) related to whatis often called its radiotoxicity. As we shall see later, someradioactive materials are potentially much more radiotoxicthan others.

If an individual is exposed both to external radiation andto radiation from radioactive materials inside his own body,the total effective dose equivalent is obtained by adding thewhole-body dose equivalent from external radiation to theeffective dose equivalent from the internal emitters calculatedfrom eqn. 1. The total effective dose equivalent is then com-pared with the appropriate limit recommended by the ICRP.

The above description, which is necessarily brief, gives agreatly simplified picture of the biological effects of smalldoses of radiation, and of the basic philosophy of radiationprotection. However, it is hoped that it will be adequate toenable the reader to understand the purpose of the varioustypes of monitoring which will now be described. It is therole of these monitoring procedures with which the review isprimarily concerned.

5 Role and design of radiation monitoring programmes

The practice of radiation protection involves monitoringfor a variety of purposes. The main types of monitoring,and their primary objectives, are shown in Table 2.

Normal monitoring can be divided into three types:(a) monitoring of the working environment(b) monitoring of the individual worker(c) monitoring of the general environment.

Various combinations of these types of monitoring arerequired, depending on the circumstances. In each case wemay have to monitor for radiation from sources external to


Table 2: Types of radiation monitoring

Type of monitoring

Practised forin all externalsituations radiation

forcontami-nation

Additionalrequirements inaccident situations

Working environment

measuring dose ratesin working areas withfixed and portableinstruments

measuring radio-activity in the air ofworkplaces

measuring radio-activity on surfacesin workplaces

use of instrumentsto give warning ofhigh dose rates dueto external radiation.or high levels ofradioactivity in theair

Individual

use of personaldosemeters, e.g.film badges, tomeasure individualdoses

measuring radio-activity in individ-uals by excretionmonitoring orbody monitoring

chromosomedamage studies

special neutrondosemeters

General environment

measuring dose ratesat places near to orremote from workingareas

measuring radio-activity in certaincritical materials(usually foodstuffs) inthe general environ-ment

identification andmeasurement ofreleased material

the body, or for radioactive materials which could get intothe body (or are already inside the body), or both, againdepending on the circumstances.

5.7 Routine monitoring for external radiationMost radiation exposure to workers in industry comes fromradiation sources external to the body, rather than fromradioactive materials which have got inside the body, and somonitoring the working environment for external radiation isvery important. Such monitoring serves several purposes:

(a) It enables one to tell whether or not it is reasonable toallow people to work in the area.

(b) It will indicate places in which dose rates are high;people can be advised to keep away from such areas.

(c) It may indicate the need for additional shielding, or forchanges in operating procedures, in order to reduce dose rates.

(d) It may (and often does) indicate a need to control thelength of time which people should be allowed to spend in thearea, or parts of it.

In such monitoring, one is normally interested in measuringdose rates rather than integrated doses. Only rarely fromsuch monitoring can one obtain a precise estimate of the dosethat a person working in the area will receive. That is becausepeople usually move about in an uncontrolled way in an areawhere dose rates vary widely from place to place, and alsobecause individuals carry out many operations which them-selves grossly alter the radiation dose rate in the vicinity oftheir body, e.g. by moving radiation sources in and out ofshielded containers. However, it is usually possible to predictthe maximum dose a person is likely to receive. As we shall seelater, personal dosemeters are then generally used to measurethe actual doses received. However, if monitoring of theworking area shows that doses are low enough and sufficientlyinvariant, it may be possible to show that the use of personaldosemeters is unnecessary. The ICRP recommends that per-sonal dosemeters are not needed if there is no possibility ofthe doses exceeding 30% of the dose limits recommended bythe ICRP; but in practice personal dosemeters are used muchmore widely than that 'rule' would suggest.

We have already noted that alpha emitters external to thebody do not cause significant irradiation of living tissues.Consequently, most monitoring for external radiation inthe working environment is for gamma and X-rays (and

perhaps beta rays measured with the same instrument); neu-trons, when present, are measured separately.

It is also sometimes necessary to monitor for externalradiation in the general environment, to ensure that adequatecontrol is being exercised over the doses being received bymembers of the general public. We shall see in Section 6.3that this rarely involves the use of personal dosemeters, theadequacy of the control measures being checked entirely bythe measurement of dose rates in the environment.

5.2 Monitoring for contamination and internal irradiationIn addition to the radiation received from sources external tothe body, there is also concern about radioactive substanceswhich may get inside the body and cause internal irradiationof the whole body, or some part of it. When radioactivematerial is not contained in a strong sealed container there isalways a possibility that some of the material will accidentallybe spread into places where it was never intended to get.Such material is referred to as contamination. It is importantbecause it may subsequently be taken into the body byingestion (swallowing), or inhalation (breathing), or in otherways (e.g. via a cut in the skin). In order to control the radi-ation doses to individuals, it is necessary to control the levelsof contamination on surfaces and in the air of places whereradioactive materials are handled. In some cases it is alsonecessary to measure or estimate the amounts of radioactivematerials which people have taken into their bodies. Finally, ifradioactive materials are discharged into the general environ-ment, there are a variety of ways in which, at least in principle,they could cause irradiation of people, e.g. by incorporationinto foodstuffs or water supplies, and, as a consequence,it is necessary to monitor various aspects of the generalenvironment.

There are major differences, not only in technique, butof philosophy, between monitoring for external and internalirradiation. In the former case the quantity being measuredis simply related to the quantity one wishes to control, i.e.dose in appropriate tissue. A personal dosemeter measuresthis quantity directly. The readings of dose-rate meters have,in principle, only to be multiplied by an appropriate timeto give a direct estimate of dose. With contamination moni-toring, however, the situation is different. The results of somekinds of contamination monitoring cannot b& directly inter-

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preted in terms of dose, or even of the resulting or associatedquantity of radioactive material in the body; those resultscannot, therefore, provide a direct comparison with the basicrecommendations of the ICRP. As the Commission itselfhas pointed out in ICRP Publication 12, an adequate systemof interpretation must be provided for measurements ofthis kind, [10], so that, either alone or in combination withother measurements, they can meet one or more of the basicaims of monitoring. Those aims are of course to achievesafety, or to demonstrate that adequate safety has beenachieved, by keeping doses as low as reasonably achievable,and by keeping individual doses within the limits rec-ommended by the Commission.

In this context, the first thing to note is that when we areconcerned with internal irradiation or contamination, thequantity which we attempt to measure or estimate is not dose,but amount of radioactivity. The SI unit used in suchmeasurements is the becquerel; 1 Bq = 1 disintegration persecond (but the unit referred to most often in the literatureof the last half-century is the curie; lCi = 3.7x 1010 Bq,approximately the radioactivity associated with 1 g of radium).

The effective dose equivalent resulting from breathing oringesting a known quantity of a particular radionuclide can be.calculated if one knows how the material becomes distributedin the body, and how quickly it is excreted. By adopting amodel for the ways in which materials behave when inhaledor ingested, and by adopting, for each radionuclide, appro-priate numerical values for the fractions taking various routesin the body, and for the rate constants associated with thetransfers which take place between various compartmentsin the model, the ICRP has derived, for each radionuclide[12], an annual limit of intake (ALI). This is the quantity ofthat radionuclide which, when taken into the body, would, ina period of up to 50 years following the intake, lead to a doseequivalent equal to the limiting value allowed for a radiationworker in a year (where members of the general public areinvolved, the ALI is correspondingly reduced).

The ALI is somewhat less authoritative than the basic doselimits, since its value depends on the validity of the modelused; however the relationship to the basic limits is fairlydirect, and so the ALI is called a secondary limit.

The ALIs of different radionuclides differ from one anotherenormously [12], by up to a factor 106 or more, because ofdifferences not only in their physical properties, but alsobecause of differences in the way they behave in the body, andthe rates at which they are excreted. We shall see that this is ofconsiderable importance when it comes to monitoring forradioactivity in the air or in people. It is not enough tomeasure how much radioactivity is present — it is usuallynecessary also to identify the radionuclide from which theradiation is being emitted, and its chemical form, before onecan assess the radiological significance of the measurement.

From the value of the ALI, the ICRP has also calculatedvalues for what it calls 'derived air concentrations', or DACs,of the radionuclides. The DAC is the maximum air concen-tration which a person could breathe for 40 h each week,50 weeks a year, without exceeding the ALI. The DACsprovide reference standards to which the results of air moni-toring can be related. It is to be noted, however, that a personmay breathe air containing a higher concentration than theDAC, provided he does so for only a short time; that is whythe term 'derived air concentration' has now been adopted[11] by the ICRP in place of what was previously called the'maximum permissible (air) concentration', or MPC.

Interpretation of the significance of the results of someother forms of monitoring is far less easy because it dependson other factors, such as the way in which workers move inor make use of their environment. For example, the relation-

ship between a measurement of surface contamination on thefloor of a workplace and the dose limits is extremely complex.So too is the relationship between the results of excretionanalysis and the corresponding doses to different organs orparts of the body. In these more complex circumstances, twoconcepts are useful in reducing the work of interpretation [10,11]-

The first of these is the concept of an investigation level.Much of the information obtained from monitoring pro-grammes merely confirms that the situation is satisfactoryand that no action is required. It is convenient to use a methodof discarding this information with the minimum of effort,and the concept of an investigation level is of value inachieving this. An investigation level is a numerical value setfor a particular type of measurement above which the resultis sufficiently important to justify further investigation, butbelow which no further action or calculation is required [10].As we shall see later, this concept is particularly useful in urinemonitoring.

The second concept is that of the derived working limit, orDWL, one example of which we shall describe in some detail.The use of a stylised model of exposure pathways can providea quantitative link between the quantity being measured andthe resulting effective dose equivalent in the body. The accu-racy of the link depends on how closely the model has beenchosen to represent the true situation, and this in turn willdepend on how generally applicable the final DWL needs tobe. For surface contamination, it is usual (although in prin-ciple not essential) to adopt a DWL which will apply to anycontamination. In this case the DWL has to apply to even themost highly radiotoxic materials used in the worst possiblecircumstances, and it is consequently rather restrictive forother materials and in other circumstances. There is obviouslythen a great uncertainty about the relationship of themeasured quantity (expressed in terms of the DWL) and theresulting effective dose equivalent. However, the intention isto establish a figure, adherence to which will provide virtualcertainty of compliance with the recommendations of theICRP. By contrast, failure to adhere to the DWL will not

Table 3: Derived working limits for surface contamination[15]

Type ofradiationemitted*

Alpha rays

Beta rays

Type ofsurface

Surfaces inuncontrolled areas:skin

Surfaces incontrolled areat

Surfaces inuncontrolled areas:skin

Surfaces incontrolled areast

Derived working limit(DWL)+§#

MCi/cm2

io-5

irr4

1 0 "

10"3

Bq/nrv

3.7 X

3.7 X

3.7 X

3.7 X

2

103

104

104

10s

*The majority of radionuclides emit either alpha rays orbeta rays, or both, and surface contamination monitoringgenerally aims to detect such radiation,twhen orginally derived, the numbers in the Table wereexpressed in juCi/cm2, and were rounded to 10" s , 10~4, or10-3.t Higher levels are allowed in controlled areas because onlyradiation workers are allowed to work there, and becausespecial precautions are taken, e.g. the banning of eating,drinking and smoking, and the use of special clothing.§The contamination may be averaged over areas varyingfrom 100cm2 for skin to 1000cm2 for inanimate surfaces.#These limits apply only to loose or removable contami-nation. Fixed contamination is merely another sourceof external irradiation.

86 IgEPROC, Vol. 129, Pt. A, No. 2, MARCH 1982

necessarily, or generally, imply a failure to achieve compliancewith those recommendations, and may require only a morecareful study of the circumstances. If contamination is presenton surfaces, there are various ways by which such material cancause irradiation of people, e.g.

(a) direct irradiation as a result of getting the material onthe skin

(b) ingestion, e.g. via the hands(c) inhalation of material which becomes airborne as a

result of disturbance of what was previously on surfaces.The DWL is the level of widespread surface contaminationthat may be present continuously on surfaces, and which issuch that (whatever the nature of the contamination andwhatever the nature of the activities being undertaken inthe area) there is no risk of people receiving doses fromthat source which exceed the ICRP dose limits. It was shownby Dunster [13] that this is the case if levels of surface con-tamination do not exceed the levels shown in Table 3. It willbe noted that the derived limits are lower in uncontrolledareas, since members of the general public may be affected.

If contamination never exceeds the levels shown in Table 3,the resulting doses are generally several orders of magnitudeless than the ICRP limits because:

(a) Any contamination which does exist is normally bothlocalised and transitory.

(b) The fraction becoming airborne is usually far less thanthat assumed in the derivation of the DWL because, forexample, steps are taken to adopt operational and cleaningtechniques which do not create a dust.

(c) The materials being used are generally of much smallerradiotoxicity than the 'worst-case' materials considered in thederivation of the DWL.

These factors are often overlooked. This is, in part, due tothe fact that the numbers quoted were incorporated into UKlegislation [14], whereas much of the associated philosophyspelled out at the time of their orginal derivation [13] wasignored. The way in which the DWLs were intended to beused, and should be used, are:

(a) if monitoring shows that contamination levels are lessthan the DWL, no further action is necessary; but

(b) if levels above the DWL are detected, some thought oraction should be taken, usually initiated by the healthphysicist.

If the contamination can easily be removed, then it should beremoved; but in many cases, where this is not possible, thehealth physicist may be able to show in a variety of ways,that no particular precipitate action is necessary — althoughfrequent detection of such contamination may indicate a needfor improved operating methods. Present legislation [14]requires the immediate cleaning-up of such contamination,irrespective of other considerations (in principle, even if largedoses of external radiation might be received by peoplecleaning up insignificant amounts of contamination). Newlegislation currently being drafted in the UK will probablychange this, and a number of other matters relating to surfacecontamination.

The detection of surface contamination at the DWLrequires the use of instruments of high sensitivity, and so

probes based on the use of Geiger counters or scintillators arenormally employed. The area of the probe can be approxi-mately matched to the averaging areas quoted in Table 3,footnote®.

One of the fundamental strengths of radiation monitoringis that one can detect minute amounts of contamination.Similar limits are not usually derived for other nonradioactivebut toxic materials, because it would be impossible to measurethem. Paradoxically, however, this great inherent advantage of

high sensitivity has in recent years often seemed to thoseworking with radioactive materials to be a handicap ratherthan an advantage. Because of the very high sensitivity avail-able, it is almost always possible to detect some radioactivity,wherever one looks; and having found it, one can almostguarantee that someone will regard that as dangerous, whileremaining blissfully unaware that there may be much greateramounts of other hazardous substances present which cannotbe detected because they are not radioactive.

5.3 Other aspects of design of monitoring programmesReference has already been made to ICRP Publication 12,which discusses DWLs and investigation levels [10]. Thepublication contains much other sensible advice about thedesign of monitoring programmes, e .g.

(i) Monitoring programmes should be designed to meetclearly defined objectives, and these objectives should berecorded.

(ii) Programmes should not be wasteful of time andresources.

(iii) All programmes should be reviewed every few years,and following any major change in the operations involved, orin national or international recommendations relating toradiation protection.

Alas, not enough heed is always paid to these recommen-dations, especially the last one. It is very easy to perpetuatemonitoring programmes; the procedures can easily becomepart of an unthinking routine. Those who are in charge ofmonitoring programmes have to fight very hard to changethem so that they are more in line with current requirements.

6 Monitoring techniques for external radiation

6.1 Monitoring the working environmentMost monitoring instruments for gamma rays and X-raysin the working environment are based on the use of oneof three types of radiation detectors:

(a) ionisation chambers(b) Geiger counters(c) scintillators.

Before discussing each of these, it is worth considering onedesirable property of such instruments. We usually want to beable to measure photon radiation covering a wide range ofenergies, from perhaps 20keV up to a few mega-electron-volts, and obtain a reading which is independent of thatenergy. We may want to measure the rate of exposure inroentgens per hour, i.e. to get a measure of the ionisationwhich the radiation will produce in air; an instrument whichgives a measure of exposure, independent of energy, is said tobe air-equivalent [15]. Or we may want to get a measure ofthe absorbed dose which would be received by tissue placed inthe radiation; an instrument which gives a measure of absorbeddose in tissue, independent of radiation energy, is said to betissue equivalent.

When photons interact with air, or tissue, or othermaterials, what mainly happens is that the photons producesecondary electrons and photons in the medium throughwhich they are passing; these secondary electrons and photonsmay then produce many further secondary particles. At eachstage the energy is reduced. Ultimately the energy is so smallthat no further charged particles are produced, and the remain-ing energy is dissipated as heat. The electrons produced by theprimary radiation result from two processes. Some are photo-electrons, produced when a photon completely ejects extra-nuclear electrons from the atoms of the material throughwhich it passes. Others are produced by Compton scattering,involving bound electrons. The number of photoelectrons


produced depends rather critically on the energy of theradiation and on the atomic number of the absorbing material.If we want a material to have the same response as air (ortissue) it must have approximately the same (average) atomicnumber. If the atomic number is much higher than that of air(or tissue), as it often is, the material will have a much greaterresponse relative to air or tissue for low-energy photons thanit does for higher energy photons, simply because there will bea much higher proportion of photoelectrons. As we shall seelater, this inherent tendency of some radiation detectors toover-respond rather grossly to low-energy photons can becorrected by the use of shielding around the detector.

In ionisation chambers one merely measures the current(or charge) produced when a moderate voltage is appliedbetween two plates in the chamber, collecting the chargedparticles produced in the gas of the chamber by secondaryelectrons resulting from the interaction of the radiationwith the walls of the chamber. It is fairly easy to make thechamber walls of material which is 'air equivalent' or 'tissueequivalent', i.e. having about the same mean atomic numberas air or tissue. The ionisation current is then directly pro-portional to exposure or absorbed dose, irrespective of theenergy of the radiation.

1000

10 100exposure rate, mR/h

1000

•--• Fig. 1 Ionising chamber currents

The currents produced at various dose rates in chambersof various sizes are shown in Fig. 1, corresponding roughlyto 10~13A in a 1 litre chamber due to an exposure rate of10~3 R/s [16]. For the dose rates which are normally ofinterest, these currents are large enough to be measuredwith an electrometer.

However, at the lower end of the dose range, it is necessaryto use rather large chambers, with volumes of several litres,and this is not a very convenient size for portable instruments.As a consequence, in recent years there has been a trendtowards the use of Geiger counters — although ionisationchambers are still widely used in some applications, e.g. forthe measurement of high dose rates, when a small chambercan be used.

In a Geiger counter, each time an ionising particle producesat least one ion pair in the chamber, an avalanche of ionpairs is produced, of such a magnitude that a detectable pulseis produced by each primary ionising event. By countingthe pulses one obtains a measure of the intensity of theradiation. Fundamentally, the count rate of a Geiger counteris more nearly proportional to the flux of the radiation thanto exposure or dose rate. However, by surrounding the Geigercounter with a shield which absorbs some low-energy radiation

(but also permits some to pass through gaps in the shield) it ispossible to arrange that the pulse rate per unit dose is virtuallyindependent of photon energy over a wide range of energies(see Fig. 2). Energy-compensated Geiger counters using suchshields are now being increasingly used in radiation protectionmeasurements.

anode resistor (stainless steel *tin * lead)filter

aluminium case glass envelopetantalum spiral cathode

1000r

500̂ -

200+

100-

2 50-:

20-

100.01 0.1

photon energy, MeV1.0

Fig. 2 Energy response of Geiger counter

In addition to gamma-ray measuring instruments based onthe use of ionisation chambers and Geiger counters, a thirdtype depends on the use of scintillators. The scintillator mostused in radiation protection is sodium iodide doped withthallium (Reference 17, p. 133). The sensitive element canbe as small as a few cubic centimetres, or as large as a fewlitres. The scintillator is always linked to a photomultiplier,often by direct contact, but sometimes by a light guide. Animportant advantage of scintillators in some applications, aswe shall see later, is their ability to perform as spectrometersas well as radiation monitoring devices (Reference 17, pp.149-150). However, although the light output is related tothe energy deposited in the scintillator, most portable surveyinstruments are not designed to measure this directly; theymerely give an indication of count rate or integrated counts,and hence of the fluence of the radiation rather than dose rate.Thus, although such instruments have some attractions (e.g.high sensitivity), they are not ideally suited for the absolutemeasurement of dose rate, and the majority of portable surveyinstruments used for monitoring for gamma rays or X-raysin the working environment or in the more general environ-ment are based on the use of ionisation chambers or Geigercounters.

Typical hand-held portable instruments used for suchmonitoring are shown in Figs. 3 and 4. Where high dose ratesare involved, it is often desirable to use instruments in whichthe detector is remote from the operator. Several types ofmonitor are available in which the detector, in the form of asmall ionisation chamber or a Geiger counter, is mounted


Fig. 3 Portable Geiger instrument being used to monitor gammaradiation level near a flask

Fig. 4 Portable ionisation-chamber based dose rate meter

on a long rod, which may be either fixed or extendible(Fig. 5).

Most monitoring is done with portable instruments likethose shown in Figs. 3 and 4. However, where there is a chancethat very high dose rates could occur unexpectedly, additionalfixed instruments are often installed. The readings of thesemay be displayed in a central control room, but more oftensuch instruments (Fig. 6) are merely used to give an alarm ifthe dose rate exceeds a predetermined value.

Some of the gamma-measuring instruments which makeuse of ionisation chambers as detecting elements can also beused for measuring beta rays, provided there is a thin windowto allow the beta rays to enter the chamber. Some instrumentshave a removeable cap on the chamber (Fig. 4); if measure-ments are made with and without the cap in place, the differ-ence gives a measure of the dose due to beta rays. It is practi-cally impossible to make an absolute or accurate measure ofbeta dose rate in this way. However, fortunately, this is not ofgreat practical importance. In general beta rays cause onlyirradiation of the skin, and, since the dose limit to the skinis 10 times higher than that to the body as a whole, it isusually only necessary to confirm that beta dose rates are notgrossly in excess of photon dose rates.

Neutrons cause little or no direct ionisation, and so cannotbe measured in the same way. There are two main ways inwhich neutrons are monitored. The first involves the use ofhydrogenous materials; collisions between fast neutrons andthe nuclei of hydrogen atoms produce energetic protons;these cause ionisation and can be detected like other directly

Fig. 5 High dose-rate measurements using remote detector

Fig. 6 Detector of high gamma alarm system

ionising rays. The second method involves the use of boron,generally in the form of boron trifloride (BF3) gas in anionisation chamber [18]. Neutrons of low energy interactwith boron nuclei to produce alpha particles inside thechamber, which are then counted in the ordinary way. Orig-inally such counters could be used only for the measurementof low-energy neutrons, but nowadays the most commonlyused neutron monitoring instruments consist of a BF3 counteror other thermal neutron counter surrounded by a sphere orcylinder of polyethylene [19], and partially shielded by athermal neutron absorber. These instruments can be used tomonitor for neutrons over a very wide range of energies,around 0.025 eV to lOMeV. Fast neutrons lose their energy(are moderated) by collisions with hydrogen nuclei in thepolyethylene, to such an extent that they can be detectedby the thermal neutron counter. By careful design of themoderator and thermal neutron absorber the instrument canprovide a reasonable measure of dose equivalent, independent


Fig. 7 Neutron dose-rate monitor capable of measuring neutronsover a very wide range of energies

of neutron energy. A typical instrument of this kind is shownin Fig. 7.

6.2 Monitoring individual workersAs we have already noted, it is usually impossible to obtaina reasonably precise assessment of individual doses from theresults of area monitoring. Consequently, if the dose rates inan area are such that individuals may receive doses approach-ing the ICRP dose limits, personal dosemeters are worn. Theseare normally issued for a period such as a month, at the end ofwhich the dosemeter is withdrawn for assessment, and anotherone is issued for the following month.

Until the last few years, almost all personal dosemetersconsisted of film badges. When photographic film is exposedto ionising radiation, silver is produced from the silver halidein the emulsion in the same way as on exposure to light.On developing, the irradiated film goes black. The amount ofblackening is related to the dose of radiation in a way which isfar from linear; saturation and even reversal occurs at highdoses. Nevertheless, the technique has proved to be a veryimportant one in radiation protection. In the early days ofradiation protection people wore a piece of photographic filmsimply to give a qualitative or crudely quantitative indicationof whether or not they had been exposed to an unduly largedose of radiation. The film, then as now, was wrapped in paperso that it was not blackened by light. Later a filter was intro-duced over one part of the film so that one could tell whetherthe dose came from beta rays or from photons. Still later thisdeveloped into the more sophisticated film badges which areused today. It is possible to make emulsions of differentsensitivities. The film that forms the basis of the film badgeswhich have now been used for many years in the UK containstwo separate emulsions, one of which is about 100 timesmore sensitive than the other. Normally doses are deducedfrom the blackening of the more sensitive one (strictly fromthe combination), but if the dose is so large that the sensitiveemulsion is too black to read, this emulsion can be stripped offand a measurement obtained from the less sensitive emulsion.In this way it is possible to cover a range of doses from about0.1 mSv up to a few Sieverts.

For the reasons already discussed in Section 6.1, a givendose of low-energy photons produces about 30 times as muchblackening (see Fig. 8) as the same dose of higher energyphotons. To overcome this fundamental difficulty, film badgeshave been designed in which different parts of the film arecovered by filters of different thicknesses, which cause varyingattenuation of incident radiations of different energies. Bysuitable choice of filters, by measuring the amount of black-

ening (or apparent dose) which occurs under the variousfilters, and by combining the apparent doses according to asuitable arithmetical formula, it is possible to obtain aneffective response which is practically independent of photonenergy over the range 20 keV up to a few mega-electron-volts.The use of such filters also enables one to distinguish thedoses due to beta rays and photons, respectively. Details of thefilm badge which has been much used in the UK [20] (andelsewhere) are given in Fig. 9, and the energy response of thebadge to radiation incident at 35° to the normal is shownin Fig. 10.

It will be seen that by careful design of the film badge,some of the inherent disadvantages of photographic filmhave largely been overcome, to produce a small, cheap packagewhich can be worn on the trunk of the body to measureexposure or absorbed dose. Large numbers of films can con-veniently be processed together, and semiautomatic or auto-matic densitometers can conveniently be used for measuringthe blackening under the various filters. However, the filmbadge is not without its problems. Because of the use of flatfilters in the badge, the attenuation which occurs in thefilters is fairly critically dependent on the angle of incidenceof the radiation. As a result the directional response is poor for

30r

20

£ 10

0.01 0.1photon energy ,MeV

1.0

Fig. 8 Energy response of photographic film

f e d c b a a b e d e f

Fig. 9 Film badge used in UK

Filter types:a 0.04 dural b 0.028 Cd + 0.012 Pbc 0.028 Sn + 0.012 Pb d windowse50mg/cm2 plastics /300mg/cm 2 plastics#0.012 Pb edge shielding h 0.4 g indium


low-energy photons. For grazing incidence, care is necessary toensure that blackening is not caused by radiation which haspassed under the filter. Blackening of the film can be producedby agents other than radiation. In spite of these problems thefilm badge has been of immense value in radiation protection.

In recent years the dominant position previously occupiedby the film badge has been very seriously eroded as a result ofthe introduction of thermoluminescence dosemeters (TLD).Many crystalline materials exhibit the property of thermo-luminescence. When they are irradiated some of the electronswhich are produced in the crystal by the radiation becomestably trapped at defects or impurity centres which are pre-sent, and they remain there for long periods. If the crystalis then heated, the electrons are removed from the traps, andas a result of their return to a position of lower energy, light isemitted. For a given type of radiation, the amount of lightemitted is proportional to the absorbed dose of radiation, andhence by measuring the light output one can obtain a measureof the radiation dose received [21]. As with other detectors,such as photographic film, the TLD response is stronglyenergy dependent if the crystal contains elements of highatomic number. However, some materials which exhibit theproperty of thermoluminescence, notably lithium fluorideand lithium borate, have mean atomic numbers quite similarto tissue, and so hence they have a rather flat energy responseand are well suited for use as dosemeters.

The first application of such materials in radiation pro-tection occurred in 1964, and was for the measurement ofdoses to people's finger tips [22]. Generally speaking, peopleare strongly discouraged from handling radioactive materialsor sources, because the dose rate near to a source, even avery weak one, can be very high, but if the material in ques-tion is only very weakly radioactive, e.g. because it has a verylong half-life, it may safely be handled. In such cases oneneeds to measure the dose to the most highly irradiated partof the body, namely the palmar surface of the fingers orthumb, since the dose there may greatly exceed the dose tothe trunk of the body, measured by the normal personaldosemeter. It is not feasible to wear something like a filmbadge on the finger tip, but it is possible to wear for examplea small plastic sachet containing a small amount (~ 30 mg)

1.5r

a 1.0

0.5

0.01 0.1 1.0photon energy , MeV

Fig. 10 Dose estimate from film badge

of thermoluminescent powdered lithium flouride (Fig. 11).After use, the lithium flouride is removed from the sachetand placed on the heating tray of a TLD reader (Fig. 12).The light emitted, when the powder is heated to about 300°C,is measured by a photomultiplier and associated electroniccircuits. This gives a measure of the radiation dose received.A schematic diagram of a typical basic TLD reader is shown

Fig. 11 Finger sachet containing powdered lithium flouride

in Fig. 13. It is convenient to use an analogue-to-digital con-vertor to convert the photomultiplier current into pulses,which can be counted using conventional pulse-countingapparatus such as sealers. This avoids the need for rangechanging.

After this technique had, for some years, been universallyadopted as the preferred method of measuring finger doses,people realised that TLD is a particularly simple technique,and as a consequence it has also been increasingly used in

Fig. 12 Simple TLD reader


recent years as a replacement for the film badge, for themeasurement of doses to the trunk of the body. In this appli-cation it is not convenient to use the TLD material in powderform. Instead, elements have been made by pressing or sin-tering, or by incorporation into an inert matrix like poly-

ifying the number and measuring the light output when eachof the discs is heated. A number of predosed test plaques areincluded to check (automatically) that the system is workingproperly. A number of other safeguards are employed withthe same objective. The output of the reader is to a teletype

EHTpower unittype 2124

power unittype 2015

HTpower unittype 2004

• 1 5 0 v

DC/pulseconvertorunit

sealertype2K2

heating andmeasuringunit

heatingcontrolunit

Fig. 13 Schematic diagram of simple TLD reader

i nhib it

thermocoupleindicationand trips

ratemetertype2134

two -channelrecorder

>•«'•"

Fig. 14 Two types of TLD badge (with film badge on left)

tetrafluroenthylent (PTFE). The most commonly used detec-tor elements are 'chips' of LiF or discs of PTFE containingLiF.

The type of badge used in the UK consists of a small plaquecontaining two TL discs, inside an outer holder designed toensure that one disc is essentially bare to measure skin doseand the other is covered by a filter to measure the dose todeeper tissues [24].

Two distinct, but extremely similar, designs are currentlyin use (Fig. 14). The plaque is protected from grease, dirt,and bright sunlight by wrapping it in a thin black plasticbag. Exposure to light is not disastrous as it is with photo-graphic film but exposure to bright sunlight does cause asmall spurious reading. Each plaque contains a unique codednumber.

A variety of automatic and semiautomatic readers havebeen developed (Reference 24, pp. 126—131) for use withthese badges. One semiautomatic machine is shown in Fig. 15.After use, the plaques are removed from the outer cases ofthe badges, and placed in the input magazine of the reader,which will take up to about 200 plaques. On switching on,the reader processes each plaque in turn, sequentially ident-

92

reader and/or punched tape or magnetic tape for transmissionto a computer, which can produce a variety of printouts ofdosimetric information, both individual and statistical.

The type of TLD badge described here has excellent energyand directional responses (Fig. 16). It cannot provide infor-mation about the energy of the radiation, but this is rarely ofmuch consequence; if necessary, such information can, how-ever, be provided by using a multielement badge. The TLDplaque is considerably more expensive than the film pack usedin film badges, but it can be reused many times. The mean costper dosemeter is broadly comparable with that of providingfilm badges. The dose range covered is similar to that of the

Fig. 15 Semi-automatic TLD reader [D.A. Pitman Ltd.]

IEEPROC, Vol. 129, Pt. A, No. 2, MARCH 1982

film badge, i.e. from about 0.1 mSv to a few sieverts. Overthis range the response is almost exactly linear.

All the personal monitoring devices so far described areintended to provide a measure of integrated dose at theend of a certain monitoring period. They all have the obviousdisadvantage that, during the monitoring period, they pro-vide no information at all about either dose or dose rate —although, of course, the dosemeter can be withdrawn and

2.0

i.O

0.01 o.i i.Ophoton energy, MeV

Fig. 16 Energy response of TLD using lithium flouride

read out if there is any particular reason to suspect that a largedose has been received. Such a dosemeter is often all that isrequired to demonstrate that the dose received by an indi-vidual is acceptably small. However, when people are requiredto work in places where dose rates are high, the use of such adevice alone is inadequate. This situation may arise, forexample, if people are required to work for a few hours ordays or weeks in a place where the dose rates are higher thanthose in which a person could be allowed to work continu-ously throughout the year, e.g. if they are required to domaintenance work on a nuclear reactor while the reactor isshut down, in a place which is normally inaccessible when thereactor is operating.

In such situations, the individuals concerned normallywear additional monitoring devices which provide an imme-diate visual indication of the dose they have received inthe current work period. The instrument most commonlyused for this purpose is the quartz-fibre electrometer or QFE(see Fig. 17). This is a simple device, about the same sizeand shape as a fountain pen. A metallised quartz fibre insidethe instrument is repelled by a fixed electrode when theelectrometer is charged up (as in a gold-leaf electroscope).Exposure to radiation causes the electrometer to discharge,allowing the quartz fibre to move closer to the fixed electrode.Its movement in front of a scale can be viewed by a smalltelescope at one end of the instrument.

The electrometer is charged up (to give a zero reading) atthe beginning of a period of work, and read out when theperiod of work in the area is complete. The results of suchmeasurements are often recorded on a dose-record cardwhich provides an hour-to-hour or day-to-day record of dosereceived. The QFE is, however, not very accurate or reliable,e.g. it may go off scale if subjected to mechanical shock, andso it cannot be used alone. Consequently, another dosemetersuch as a film badge or TLD is worn at the same time; i.e.QFEs are used to provide a rough provisional measure ofdose, but the final assessment of dose is almost invariablyderived from the film badge or TLD.

In recent years a number of pocket-sized instrumentsbased on the use of Geiger counters have become available.Some of these measure (gamma) dose rate, and emit either'bleeps' at a rate which is roughly proportional to dose rate, ora sound when the dose rate exceeds a predetermined value.

Other instruments of this type measure integrated dose, whichis usually displayed in digital form; some give an alarm if theintegrated dose exceeds a predetermined value.

The monitoring of neutron doses to personnel is currentlyone of the least satisfactory aspects of radiation protectiondosimetry. Fortunately, only a very small fraction of workersin the radiation industry are exposed to neutrons, and eventhen the dose equivalent due to neutrons usually representsonly a small fraction of the total dose equivalent.

The film badge can provide a measure of the dose fromthermal neutrons. This is achieved by making one of thefilters in the badge from cadmium. When a thermal neutronis absorbed by cadmium, the latter emits a gamma ray;henceif the badge is exposed to thermal neutrons, the area under thecadmium filter becomes blackened, and this can be used tomeasure the fluence of thermal neutrons.

Unfortunately, however, the film badge cannot be used tomeasure the dose from neutrons other than thermal neutrons,arid it is from these other neutrons that significant dosesinvariably arise. Several of the most widely used thermo-luminescence materials are also very sensitive to thermalneutrons. For example those which contain lithium are verysensitive to thermal neutrons if the lithium contains 6 Li and7 Li in the normal proportions. 6Li has a high capture cross-section for thermal neutrons, because of the reaction 1n0 +6Li3->4He2 + 3 H i ; this produces two energetic chargedproducts, both of which deposit their energy in the lithiumflouride or lithium borate.

Some TLD badges, used primarily for measuring beta/gamma doses, also incorporate detectors especially to measurethermal neutrons. Most people consider it more desirable tomeasure beta/gamma doses with TLD materials which areinsensitive to thermal neutrons, e.g. by using LiF in whichthe 6Li is depleted, and then to measure neutron doses, whennecessary, using a separate neutron dosemeter. Lithiumdepleted in 6Li is used in order to avoid the possibility thatexposure to thermal neutrons will lead to spurious results forthe gamma-ray dose.

Fast neutrons (e.g. of energy a few mega-electron-volts)can also be measured, by a quite different technique involvingthe use of nuclear emulsions. The fast neutrons produceknock-on protons in the emulsion, and these in turn producetracks which can be seen under a microscope. Although thecounting of such tracks is tedious, this can provide a satisfac-tory method of measuring fast-neutron fluence. However, it is

eyepiece lens

working range(about 0.7mm)

optical axis

end cap

Fig. 17 Quartz fibre electrometer


Fig. 18 Surface contamination monitor(s)

an unfortunate fact that most of the neutrons to which peopleare exposed in nuclear power stations and similar facilitiesare of intermediate energy ( < 1 MeV), and although theseproduce tracks in nuclear emulsions, the tracks are too shortto be recognised. As a consequence the nuclear track methodis of limited usefulness. The most satisfactory monitoringmethod available at present makes use of the albedo principle(Reference 24, pp. 108—110); fairly energetic neutronsincident on the body are 'thermalised', and scattered byrepeated collisions with hydrogenous substances in the body,and so there is a flux of thermal neutrons 'reflected' fromthe body. This can be measured using a TLD containing(say) 6LiF. The most commonly used albedo dosemetersemploy two 6LiF detectors separated by a thermal neutronabsorber, one to measure the incident flux of thermal neu-trons, and the other the reflected flux. The badge also containsat least one 7LiF detector to measure the gamma dose, andhence to enable one to correct the readings of the 6LiFdetectors for that component. Such dosemeters do not havea 'flat' energy response, and so can only be used if the neutronspectrum remains reasonably constant in both space and time.An appropriate calibration or sensitivity factor has to bedetermined for each facility [26].

6.3 Monitoring the general environmentIn assessing doses received, or likely to be received, by mem-bers of the general public, the whole approach is somewhatdifferent to that adopted in the monitoring of radiationworkers. It is most unusual for personal dosemeters to beissued to members of the general public (except when theyenter controlled areas), although particular groups of peoplehave, for a few months, sometimes been asked to wear per-sonal dosemeters (TLDs) in places where the background ishigh, in order to establish the doses which people in thoseareas are receiving from natural sources. Instead of usingpersonal dosemeters, doses to individuals are assessed fromthe results of environmental monitoring, using dose-ratemeasurements or integrating dosemeters. This assessmentis not as difficult as in the working environment, since doserates are much less variable in both space and time. As pointedout in Section 1, this review is only concerned with the

monitoring of man-made radiation. However, in practice wecan generally only measure the total dose or total dose rate,and, in order to determine the man-made component, wemust subtract the natural component. This is not always easy,as we shall see later.

If the radiation levels we are trying to measure are suf-ficiently high, adequate separation of the natural and artificalcomponents is straightforward, or even unnecessary. Suppose,for example, that we wish to show that members of the gen-eral public are not being exposed to excessive amounts ofexternal radiation emanating from a particular factory. Nor-mally, to achieve this we will merely need to show that thetotal dose rates at all points on the fence surrounding thefactory are less than

5000

24 x 365= 0.57juGy/h

the annual dose received by a person will then be less than5000/uGy = 5mGy (=5mSv) per year: the limiting valuerecommended by the ICRP. Note that in this case we arealmost invariably only concerned with the dose from gammarays. The limiting dose rate of 0.57/iGy/h quoted above isbetween five and ten times higher than the dose rate due tonatural background radiation in most places. It is not difficultto make ionisation chamber instruments or Geiger instrumentswhich will measure such dose rates with adequate precision.The instruments are basically similar to those used for moni-toring the working environment, but they are more sensitiveby a factor of 10 to 100.

Unfortunately, however, the above approach ignores theALARA principle referred to in Section 4.1. The ALARAprinciple is not an easy one to apply, as it involves judgmentsabout what is reasonably achievable when economic andsocial considerations are taken into account, and there is atpresent no generally agreed formula by which this can bedone. However, decisions have been taken in some particularsituations. Notably, government regulations in the USAspecify that the doses resulting from the discharge of radio-active waste from any light-water reactor used for generatingelectricity in the USA shall not exceed about 5% of theaverage exposure from natural background radiation [27]. Asa consequence, it is necessary for the designers and operatorsof such reactors to demonstrate that radiation doses to mem-bers of the general public from emissions of gaseous radio-active materials emitting gamma rays do not exceed about50/iGy per year. Quite apart from any difficulties whichpeople may encounter in designing reactors to satisfy thisrequirement, there are also formidable problems in makingmeasurements which demonstrate compliance. In this case it iscertainly necessary to subtract the natural component of thedose, which far exceeds the dose which may legally be allowedto result from the gaseous discharge. In fact, the allowabledose rate from such discharges is probably less than fluctu-ations in the background radiation dose rate due to naturalcauses, e.g. due to variable snow cover or waterlogging of theground. The changes are also certainly far smaller than fluctu-ations in background levels which have occurred in the pastowing to worldwide fallout from atomic weapon tests.

One might suppose from the above that it would be necess-ary to adopt some form of measurement which could dis-tinguish the radiation dose due to the gaseous emissions fromthe reactor from that due to other causes. One possible way ofdoing this would be to separate the two components byrecognising that rapid fluctuations in the dose rate are to beexpected from gaseous emissions, because of variations in winddirection, whereas the dose rate due to other causes wouldbe expected to fluctuate very much more slowly. It remains


the view of the author of this review that ultimately this willbe the best method to use. However, rather surprisingly,the method which has been adopted almost universally at thepresent time involves the use of TLDs, which measure doserather than dose rate; this has happened, probably, becausethe amount of development required was thought to besmaller. TLDs are placed in the field, at the point wherethe dose is to be measured, for a period of perhaps 1—3months, and then read out at the same time as referencedosemeters treated similarly in the same general locality, butnot exposed to the gaseous emissions [28].

The TLD system normally used in routine personal dosim-etry is not usually capable of measuring doses of less thanabout 100/iGy; but doses about a hundred times smallercan be measured if great care is taken with a large numberof detailed aspects of the dosimetry. There is no difficultyin achieving adequate sensitivity, but there is considerabledifficulty in achieving an adequate signal/noise ratio. Onehas to reduce the 'noise' level — in this case the productionof spurious signals in the reader during readout, due to phen-omena other than radiation-induced thermoluminescence.It is particularly important to exclude oxygen from the reader,and to reduce the number of OFT impurity on the surfaceof thermoluminescence detectors containing LiF. Otherwisethere will be spurious light emission at an unacceptable levelfrom the detector itself. Typically, the best results areobtained by a combination of the following [29]:

(a) use of 'chips' of LiF or other materials in preferenceto powder or Teflon dispersions; light sensitivity of the Teflondevices alone rules them out

(b) use of several chips in each location, taking the meanreading rather than that indicated by one of them

(c) use of scrupulous methods of chemically cleaning thechips before read-out to remove surface impurities

(d) use of carefully controlled annealing treatment of thechips before use

(e) reduction of non-radiation-induced signals by readingout in an atmosphere of nitrogen

(f) use of a photomultiplier with a very low and stabledark current

(g) use of optical filters to differentiate between radiation-and non-radiation-induced signals

(h) control of the temperature range of the read-out cycleto avoid black-body radiation and other spurious emissions atthe top end of the temperature scale.

Some improvement can be achieved by counting single pho-tons rather than by measuring integrated light output (thetechnique normally adopted).

In the laboratory it is possible to measure doses as low as0.6 fiGy, with an error of ±0.2juGy, by adoption of thesemethods. But in field measurements, a number of additionaldifficulties are encountered, and the limit of detection isprobably an order of magnitude higher.

7 Monitoring techniques for contamination and internalirradiation

7.1 Monitoring surface contamination in the workingenvironment

Typical instruments used for monitoring for surface con-tamination are shown in Fig. 18. They generally consistof a probe, of suitable area, connected by a long flexiblelead to a 'box' containing a power supply, ratemeter anddisplay. Because the DWLs for surface contamination by betaand alpha emitters are different, it is necessary to monitorfor the two separately. However, no attempt is made todistinguish between different types of alpha rays or betweendifferent types of beta rays; one simply measures gross counts.

When monitoring for alpha contamination, scintillatorprobes containing zinc sulphide, together with a photomulti-plier, are generally used. Light must of course be excluded byuse of an opaque cover over the front of the probe. Alphaparticles have a maximum range of only a few centimetresin air, and, as they have usually lost some of their energybefore escaping into the air, their normal range is even smaller.As a consequence the probe must be held very close to thesurface being monitored, and the protective surface over thezinc sulphide detector must be very thin. It is very difficultto make probes which are both effective and yet sufficientlyrobust to be used for monitoring a variety of surfaces inindustrial conditions; in practice considerable effort is devotedto repairing probes which have been slightly damaged, andhence become light sensitive.

Similar probes using anthracene or other scintillators arealso used for monitoring for beta contamination. There arealso dual-phosphor probes which contain both types of phos-phor, and so hence can be used for the measurement of bothbeta and alpha contamination. The pulses produced in thephotomultiplier by alpha particles are larger than those pro-duced by beta rays, and so, by use of a suitable discrimi-nator, it is possible to distinguish between the two. By useof a switch, it is possible to measure alpha alone, beta alone,or alpha and beta. In addition to a meter display on con-tamination monitoring instruments, it is normal to providea loudspeaker to provide an aural indication of dose rate,so that the operator can for most of the time concentratehis eyes on the position of the probe, and his ears on thenoise from the loudspeaker. When the dual-phosphor probeis being used, alpha and beta rays produce different noisesfrom the loudspeaker.

Geiger counters are also used for contamination moni-toring, especially for beta rays. End-window Geiger counterswith a very thin window can be used for alpha-monitoring, butare not sufficiently robust for routine use. They usually have avery small area; this is in general a disadvantage when one ismonitoring large areas, but can occasionally be useful forspecial monitoring in places where a probe of normal size willnot go. Geiger counters of the type shown in Fig. 19 are stillmuch used for monitoring for beta contamination, althoughthey are not entirely satisfactory. The area covered is small(20 cm2) and the response to low-energy beta rays is poorbecause of absorption in the rather thick wall of the Geigertube. Fortunately there is an element of 'swings and round-abouts' in this situation — the radionuclides which emitlow-energy beta rays are (ipso facto) of lower radiotoxicity,and so the probe tends to give a truer picture of the potentialhazard than it does of the associated radioactivity. The radi-

Fig. 19 (Old-fashioned) beta probe


ation emitted by some radioactive materials such as tritium iscompletely undetected by instruments of this type because theradiation is completely absorbed in the wall of the countertube. If it is known that such materials may be present, specialmonitoring techniques are required. Fortunately, materialsemitting such low-energy radiation are of extremely lowradiotoxicity.

Direct monitoring for contamination is often difficult inplaces where there are high gamma-ray dose rates due toadjacent radiation sources. Gamma rays are detected, albeitwith low efficiency, by most contamination monitors. Whenone is monitoring for alpha rays, it is possible to discriminatebetween the pulses produced by alpha rays and gamma rays;but effective discrimination is not possible when one is moni-toring for beta rays, and monitoring at the DWL becomesdifficult if the dose rate due to gamma rays is more than a fewtimes normal background levels. The situation can be ameli-orated to some extent by the use of probes like the one shownin Fig. 20, where the Geiger counter in the probe is shieldedfrom external radiation by a tungsten shield. This techniquedoes not work, however, if the gamma radiation is comingin the same direction as the beta rays, e.g. if one is trying tomonitor for contamination on the outside of a shieldedcontainer containing a radioactive source. In such circum-stances, one has to resort to 'smear' monitoring. The surfacein question is smeared or rubbed over with something like apiece of filter paper; the filter paper is then taken to a place

Fig. 20 Shielded beta probe for use in high-background areas

where the level of gamma radiation is low, and the amount ofradioactive material removed on the filter paper is measured.It is normally assumed that smearing removes 10% of theloose radioactive material which was present, although inpractice it is clear that the fraction removed is highly variable,and is often nearer 100%. The imprecision arising from theassumption that 10% is removed is a disadvantage of smearmonitoring; but there are many occasions on which it is theonly method available. Smear monitoring also has the meritthat it distinguishes between fixed and loose contamination —it is obvious that direct monitoring merely gives a measure ofthe total quantity of contamination present, and hence mayoverestimate the amount of the loose contamination, whichalone has the potential to produce internal irradiation.

The same probes which are used for measuring the amountof contamination on floors, benches etc. may be, and oftenare, used for measuring contamination on people's skin, or ontheir clothing. However, if large numbers of people workregularly in a controlled area, where contamination might bepresent, it is usual to provide purpose-built hand-and-clothingmonitors at the exits from the area, to be used for final

Fig. 21 Hand and clothing monitor

checking on leaving the workplace. A typical installation isshown in Fig. 21. Mostly, a check is made for both alphaand beta activity; in older instruments the two were measuredseparately, but in more modern devices the two are measuredsimultaneously. The detectors usually consist of scintilla tors,but gas-filled proportional counters are a satisfactory alter-native. Some hand-and-clothing monitors make a continuousmeasurement of the count rate due to background radiationwhile the machine is not in use; this is then automaticallysubtracted when the machine is used.

In a few places, a completely different type of monitoris used. In this case, instead of using a hand-held probe tomonitor his clothing a bit at a time, the person being moni-tored stands in an enclosure surrounded by a large array ofGeiger counters or other detectors, so that any contami-nation on any part of his body is detected. In some of themore modern (and expensive) instruments of this type thereis a tendency to make the whole monitoring process almostfully automatic. The person being monitored may have topass through a series of turnstiles, which force him to usethe monitor, and will not allow him to leave the area if he iscontaminated. Two designs of monitor of this general typeare shown in Fig. 22.

7.2 Monitoring for airborne radioactivity in the workingenvironment

We have already noted that inhalation is the most likely routeby which radioactive materials get inside the body, and so oneneeds to know, or be able to predict, how much airborneradioactivity is present in working areas. It has sometimesbeen assumed that if no detectable surface contaminationis present, there cannot be any radioactive material in the air.In some circumstances that is wrong; there may be a signifi-cant amount of contamination in the air, even if there is littleor no contamination on surfaces. This is the case when theairborne material is present, not as a result of resuspensionof material previously on surfaces, but as a consequence of


[Nuclear Enterprises Ltd.

[Eberline Instrument Co. Ltd.

Fig. 22 Modern personnel monitors

Fig. 23 Glove box worker (Note: the cylindrical object just abovehis head is an activity-in-air monitor)

direct release in gaseous, vapour or finely-divided particulateform, direct from some containment such as a glove boxused for work on material of high radiotoxicity (Fig. 23).In such circumstances it may be absolutely necessary tomonitor for airborne radioactivity as well as for surfacecontamination; in many other circumstances it is highlydesirable (although perhaps not essential) to do so.

In contrast with what has just been said, air monitoringoften demonstrates that very little airborne radioactivematerial is present, even in places where there are quite highlevels of surface contamination. This happens because theDWL for surface contamination, intended to apply to awide range of conditions, is unduly conservative for situationsinvolving materials of low radiotoxicity, or where surfacesare wet, or only contaminated locally etc.

The type of air monitoring employed varies enormously,depending on the amount and type of radioactive materialwhich is expected, or is normally found, in the area inquestion.

The general object of air monitoring is to detect radioactiveparticulate matter (small dust particles). This can be collectedon a filter paper simply by sucking air through the paper at aknown rate for a known time. Sometimes 'spot' samples aretaken, but often such monitoring goes on continuously. In thesimplest form of monitoring, the filter paper is removedat the end of a predetermined period, or at the end of a shift,or each 24 h, and the gross amount of radioactivity on it ismeasured by putting it in a suitably shielded counting assem-bly. A considerable amount of monitoring of this type is donemerely to confirm, retrospectively, that no significant amountof airborne radioactive material was present.

This type of monitoring is only satisfactory when the riskof airborne radioactivity is small. When there is a greater riskof airborne material being present, or when such radioactivityis known to exist, it is usual to employ rather more sophisti-cated instruments, in which the amount of radioactivitycollected on the filter paper is monitored continuously byplacing a suitable detector near it. One can then arrange for analarm to be given if the amount of activity exceeds a pre-determined level.

Daughter products of naturally-occurring radon and thoronare almost invariably present in the atmosphere at levels which(when expressed in becquerels per cubic metre) are in excessof the DACs of many of the more highly radiotoxic alpha-emitting nuclides [30]. They do not represent a hazardbecause they are short-lived, and hence of low radiotoxicity.But, because they are present, it is not possible to set alarms


at a level corresponding to 1 DAC of plutonium (say). How-ever, fortunately it is not necessary to set alarms as such lowlevels. It is generally possible to set the alarm at a level corre-sponding to (say) 10 DACs for 1 h (or 1 DAC for 10 h etc.). Inthat situation a person may be exposed to (say) 2 DACs forseveral hours without an alarm being given; but it should benoted that:

(a) Even if no alarm is given, and there has been anexposure to a few DACs for a few hours, this will becomeclear if the filter paper is again counted (say) 24 h later whenthe natural short-lived radioactivity has decayed to an insig-nificant level.

(b) The resulting radiation exposure will be a very insig-nificant fraction (1/200) of the permitted exposure for theyear.

(c) Retrospective monitoring of this type is practisedroutinely in places where there is concern about the possiblepresence of materials of high radiotoxicity (usually alphaemitters) in the air. It is not difficult to establish, retrospec-tively, the amount of such materials, even if they are onlypresent at a few percent of the DAC, even when the presenceof natural short-lived radionuclides makes it impossible todetermine their concentration in 'real time'.

The level at which alarms may be set depends (obviously) onthe amount of radon and thoron daughters present. Radonand thoron are continually escaping from radioactive materialsin the ground, in building materials, and so on. The quantitypresent near the earth's surface depends fairly critically onthe state of the weather (strictly the degree of atmosphericstability). If alarm levels are set too low, there is a risk of

Fig. 24 Portable and personal air samplers

of spurious alarms when there is a temperature inversion inthe atmosphere.

There are several ways in which one can reduce the prob-lems associated with these short-lived natural substances. Oneof the most effective is by filtering all air entering the workingarea in question, and by having a rapid ventilation rate. Thefilters remove the radon and thoron daughters, which aremainly deposited on small dust particles in the air (althoughthey do not remove the radon and thoron, which are gaseous).The rapid ventilation ensures that there is not enough time forfurther daughter to build up significantly inside the workingarea; it takes 15—30 min for the levels of the daughters tobuild up to near their equilibrium value.

It is also possible to reduce the reading due to naturalproducts to some extent by energy discrimination when thepulses are counted.

The alpha particles from the natural radioactive materials

are of some what higher energy than those from (say) plu-tonium, and so by counting only pulses corresponding to anenergy of less than (say) 5.5 MeV, one can count virtuallyall of the pulses from the plutonium, but only a small propor-tion of those from the interfering natural radioactivity [31].In some instruments, even more sophisticated techniques areemployed, using second or third channel compensation.

The positioning of air samplers is very important [31].In many situations, the concentration of radioactive materialin the air may vary considerably from place to place. In somesituations, workers may unwittingly create a cloud of radio-active material in their own vicinity. This can happen if theychoose an unwise operating procedure which raises dust fromsurfaces. It can also arise in the case of a man working at aglove box, using a glove in which (unknown to him) there isa slight defect; as he goes about his work, he may be releasingsmall amounts of material of high radiotoxicity very close tohimself. An air sampler placed several metres away may thenunderestimate, by several orders of magnitude, the concen-tration at the man's breathing zone. There are two ways inwhich this difficulty can be overcome:

(a) by using an air monitoring system with a large numberof sampling heads strategically placed, e.g. by having a sam-pling head on the working face of each glove box (Fig. 23)

(b) by the use of personal air samplers (see Fig. 24).

These procedures are only adopted in places where thereis known to be a high risk of localised exposure to materialsof high radiotoxicity. Since different radioactive materialshave very different DACs, it is not only necessary to measurethe concentration of radioactive material in the air; onemust relate the measured value to the appropriate DAC.This presents no problem if only one radioactive materialis in use, or if experience has shown that airborne radioactivityin the area invariably consists predominantly of one material.In other cases, e.g. if a high level of radioactivity is unex-pectedly observed on one occasion in an area where airbornelevels are normally very low, it is necessary to establish thenature of the material in order to determine which value touse for the DAC.

Identification of such materials is usually done mostsimply by gamma spectrometry. Each radionuclide emitsa characteristic spectrum of gamma rays. By the use of asuitable detector with a pulse height analyser, the nature ofthe gamma radiation emitted by a given sample can be deter-mined, and the nature of the radioactive material can bedetermined by comparing the spectrum with that of thevarious possible radionuclides.

Sodium iodide detectors, doped with thallium, can beused for this purpose, but they are of poor resolution, andso are unable to distinguish between certain radionuclideswith similar characteristic gamma spectra. It is generallymuch better to use lithium-drifted germanium detectors.Examples of spectra obtained with these two types of detectorare shown in Fig. 25.

Radioactive materials are sometimes present in the air ingaseous or vapour form; in this case the type of air monitoralready described, based on the use of filter papers, is of nouse. Such materials can be detected, for example, by passingthem through an ionisation chamber or gas-flow counter.

Special techniques are sometimes necessary in particularcases. Difficulties arise, for example, when a small quantity ofa material of high radiotoxicity may be accompanied by amuch larger quantity of some other material of lower radio-toxicity. It may be possible to discriminate by counting thegamma ray emission (rather than alpha or beta radiation)in a narrow energy band. In the case of 131I, for example,substantial discrimination can be acheived against other radio-


nuclides by counting only gamma rays in an energy bandaround 0.36MeV [32].

A quite different form of discrimination is sometimespossible. If, for example, one wishes to determine the amountof tritiated water vapour in air which also contains a sub-stantial concentration of noble gases (a problem encounterednear some nuclear reactors), it is desirable to separate the twomaterials by physical or chemical means before attemptingany count [33]. This can easily be done by sucking the airat a known rate through a bubbler containing water. Thetritiated water vapour in the air is trapped in the water, whilethe noble gases pass through unchanged. By mixing a sampleof the water with a small quantity of liquid scintillant, andthen measuring the light emitted, the quantity of tritiatedwater present, and hence the concentration in the air, caneasily be determined.

7.3 Monitoring for radioactive materials in peopleMonitoring of the working environment often provides ade-quate reassurance that people cannot take, or have taken,significant amounts of radioactive materials into their bodies.There are, however, occasions on which it is desirable to havemore direct confirmation of this. There are other occasions

on which it is known that one or more persons have acciden-tally been exposed to radioactive materials, and one needs todetermine how much radioactivity they have taken into theirbodies. In such cases it is not possible to get a direct measureof the radiation dose; the best that one can do is to determinethe quantity of radioactive material present in the body, orsome part of it, and then to calculate the resulting effectivedose.

There are two methods by which one can determine howmuch radioactive material a person has in his body: by analysisof his excreta, or by measuring the radiation escaping fromhis body. The first method is by determination of the amountof radioactive material which he is excreting in his urine(or occasionally, in his faeces). This is obviously a ratherindirect method. For it to be effective, certain conditions haveto be satisfied. The amount being excreted must be simplyrelated to the amount in the body, and the relationship mustbe repeatable and known. In some cases these conditions arefulfilled, and urine monitoring is a very satisfactory technique.Perhaps the best example is that of tritium. If tritium is takeninto the body, e.g. as tritiated water, it becomes generally, andrather uniformly, dispersed in body tissues, so that the con-centration in the urine is the same as in other body fluids.

Fig. 25 Spectra obtained with Nal and Li(Ge) detectors

NalNB Samples are not identical: isotopic ratios differ

labyrinth entrance

Fig. 26 Whole body monitor

IEEPROC, Vol. 129, Pt. A, No. 2, MARCH 1982

Hence, by measuring the concentration of tritium in the urine,the concentration in the body tissues is determined. Becauseof the facts just stated, tritium taken into the body is rapidlyexcreted (with a half-life of about 10 d), and the concentrationin the body falls rapidly with time; but at any one time theconcentration in the urine provides a good measure of theconcentration in the body. Knowledge of the concentrationat one time, and of the excretion half-time, which can betaken to be 10 d or determined from the rate at which theconcentration in urine falls off, enables one to calculate thedose to the body tissues.

For some other radioactive materials, the rate of excretionis a function, not only of the amount in the body, but of thetime for which the material has been in the body; the fractionexcreted per day falls off quite rapidly with time following an

99

exposure [34]. This presents no great problem if one isattempting to calculate the dose resulting from an accidentalexposure on a known date. However, where urine samples aretaken on a routine basis, say once a month, and a high resultis found on a given sample, there is often no way of knowingwhen the material was taken into the body. It is then usualeither to assume pessimistically that the material was takenin immediately after the last urine sample was provided (thisgives the biggest estimate of the resulting dose to the body)or to assume, rather less pessimistically, that the exposureoccurred half-way through the monitoring period.

Because the excretion rate depends on many other factors,e.g. chemical composition, urine monitoring is not really anideal method for assessing the amount of a radioactive materialin a person. However, it is a relatively cheap method, and ismuch used in routine monitoring in a negative sort of way;provided that the quantity of radioactivity is less than theappropriate investigation level (see Section 5.2) one canassume that no significant exposure has occurred.

In the case of tritium, the concentration in urine canbe measured by mixing a sample of urine with a liquid scin-tillant, and measuring the light output. More usually, a par-ticular element is separated from the urine by means of anappropriate radiochemical procedure; the element in questionis then deposited on a counting tray by evaporation or electro-deposition, and the amount of radioactivity in the sample ismeasured by direct counting of alpha or beta rays.

For most radionuclides, a much more direct and preciseestimate of the amount of radioactive materials in the bodycan h* obtained by measuring the radiation which escapesfrom the body, in what is often (and sometimes erroneously)called a whole-body monitor [35]. The object is generally tomeasure the quantity present of some material which emitsalpha or beta rays. Because of the short range of these radia-tions in tissue, it is not feasible to detect them by monitoringoutside the body. However, such materials frequently emitgamma rays and/or X-rays, and a fraction of these do escapefrom the body. Therefore, by measuring the associated gammaray or X-ray flux from the body, we can assess how muchradioactivity is present in the body. Furthermore, since theradiation emitted is characteristic of the emitter, we can tellnot only the amount, but the nature of the material in thebody by the use of gamma-ray spectrometry.

90-

§ 80.co

% 70

Oin

60di

Q.Si 50

c 30Oo

20

10

K-40

The quantities of radioactive material normally of interestare only of the order of 1000 Bq, and so the number of gammarays emitted from the surface of the body per second persquare centimetre is small. As a consequence the counting hasto be done in a shielded enclosure to reduce the backgrounddue to cosmic rays and other natural radiation sources. Theenclosure normally consists of a shielded room with wallsmade of several inches of steel or lead. Great care has to betaken to use materials with a very low content of radioactivematerial for making the shield walls, and similar care has to betaken in the choice of materials for use inside the shield.The person being measured normally lies on a stretcher orbed, with several radiation detectors above and below hisbody (Fig. 26), so arranged that the combined count ratedepends only on the amount of radioactive material in thebody, and not its distribution. The detectors normally consist oflarge Nal crystals. A typical spectrum obtained from such asystem is shown in Fig. 27. All people contain ^K, about4000Bq on average, and small quantities of 137Cs, fromfall-out from atomic weapon tests.

If it is required to determine not only the amount ofradioactive material present, but also its position in the body,one can make use of a single moveable Nal crystal in thenormal whole-body monitor. Alternatively, one can use aso-called shadow-shield monitor. In this case the subjectis moved slowly, but continuously, underneath a single largeNal crystal; the crystal is itself shielded from above, and thesubject moves in an enclosure which is shielded from belowand from both sides (Fig. 28).

0 0.5 1.0 1.5gamma ray energy , MeV

Fig. 27 Spectrum from whole body monitor of a subject containing

Fig. 28 Shadow-shield body monitor

Although the quantities of radioactive materials to bemeasured are small, adequate sensitivity is available to allowa precise measurement to be made for most gamma emitters.Unfortunately, some of the most radiotoxic subtances, e.g.plutonium, are more difficult to measure because they do notemit a measureable amount of gamma radiation. However,plutonium does emit X-rays in a small proportion of disinte-grations, and these can be measured (with difficulty). If theplutonium is in a form which is soluble in body fluids, andhence collects in the bone, the absorption of the X-rays whichoccurs in the bone is so great that is is virtually impossibleto make a measurement of the X-rays leaving the body; inthis case one has to resort to urine monitoring to determinehow much plutonium is in the body. However, if plutonium isinhaled in an insoluble form (e.g. as the oxide) it is possible tomake meaningful measurements with a counter over the chest[36] (Fig. 29).

Because the flux of X-rays leaving the chest is so small,however, it follows that, even if the measurement is madeinside a heavily shielded enclosure, it is essential to discrimi-


,K ^

Fig. 29 Chest monitor being used

nate against radiation from sources other than plutonium. Thisis done by the use of anticoincidence techniques.

The detector shown in Fig. 30 is a multi-wire gas-filledproportional counter with a large frontal area, speciallydesigned with an in-built anticoincidence system. It has athin front window to enable the X-rays from the subject'schest to reach the counter. The background radiation whichis present is generally of higher energy (and is thus morepenetrating) than the X-rays from plutonium. As a conse-quence , whereas the X-rays are almost all absorbed in the maincounting chamber, and only give pulses on the correspondingcounting wire, the background radiation generally also pro-duces a coincident pulse on one of the surrounding wiresforming part of the anticoincidence shield. The normal objec-tive is to try to disregard, by the use of anticoincidencetechniques, all pulses caused by radiation outside the energyrange (say) 10—25keV. Obviously this does not howeverdiscriminate against other radiation in the energy band 10-25keV leaving the body. Unfortunately there is a significantflux of radiation in this energy band leaving the body, in theform of scattered and degraded radiation originally emanatingfrom 40K and other radionuclides in the body. A correctionhas to be applied for this, the magnitude of the correctionbeing determined from measurements on people of similarbuild who have not been exposed to plutonium.

As an alternative to a proportional counter, a 'phoswich'(abbreviation for phosphor sandwich) may be used [37].This may consist, for example, of a thin (1 mm) wafer ofthallium-activated Nal backed by a thicker (25 mm) layerof thallium-activated Csl; this pair is connected to a photo-multiplier, as with any other scintillator arrangement. Thethin Nal crystal detects both the X-rays from the plutoniumin the subject, and any other background radiation. However,in general the background radiation also produces a pulse inthe Csl, whereas the soft X-rays are totally absorbed in theNal, and so do not produce a pulse in the Csl. The pulsesarising from the two crystals can be distinguished from oneanother because their characteristic delay times are different,about 0.25/xs in the Nal and 1 fxs in the Csl. One determinesthe number of pulses from the Nal where there is no coin-cident event in the Csl. The system is used in the same way asthe proportional counter, and similar corrections are necessaryfor other X-rays leaving the body.

The calibration of such counters presents a number of prob-lems, as a consequence of the absorption of X-rays occurringin the body itself. The amount of absorption depends fairlystrongly on body build, and corrections must also be madefor this.

Fig. 30 Chest monitor construction

7.4 Monitoring following incidentsAs we have already noted, urine monitoring has for manyyears been used primarily as a method of demonstrating, aspart of an ongoing routine, that no significant intake of radio-active materials has occurred. In recent years whole-body andchest monitoring have been increasingly used for the samepurpose.

When there has been a known accidental intake of radio-active material, it is important to obtain as much informationas possible from all available sources, and then to try to obtainthe most precise possible estimate of the amount taken intothe body of its subsequent excretion rate. Very often the firstindication that an incident has occurred is given by an alarmon an air monitor, or a high result on a personal air sampler.This triggers demands for excreta samples from individualswho may have been affected. In this case it can be moreimportant to sample faeces than urine; a high proportion ofmaterial which is inhaled (> 50%) is excreted in faeces in thefollowing week, and so the amount in the faeces can be usedto provide quite a good estimate of the intake.

Once this material has been cleared from the body, bodymonitoring or chest monitoring can provide a measure ofhow much is retained in the body, and subsequent measure-ments will indicate the rate at which it is transferred to otherparts of the body or excreted. In this way a complete pictureof what has happened can be built up, leading to the possi-bility of a fairly precise estimate of the effective dose com-mitment resulting from the accidental intake.

7.5 Monitoring for radioactive materials in the generalenvironment

Naturally occurring radioactive materials are present every-where, in the earth, in the sea, in plants and animals, and inour own bodies. The most important materials present areuranium and its daughters (including radium and radon alreadyreferred to), thorium and its daughters, and potassium, whichcontains about 0.01% of ^ K which is radioactive. All of thesematerials are primordial, i.e. were present when the world wasformed; they have been slowly decaying ever since. Althoughthe concentrations of these materials are generally low, thequantities present in our environment are astronomical com-pared with the quantities of radioactivity which man is likelyto discharge into the environment as waste. For examplethe amount of ^ K in the world's oceans is about a milliontimes greater than the (beta) activity of all the waste dis-charged into the sea in a year from the whole of the UKnuclear power programme.

Nevertheless, very strict controls are exercised over thequantities of radioactive waste which may be discharged intothe atmosphere, or into rivers or the sea, and a considerable


amount of monitoring is performed in order to demonstratethat the resulting doses to man are acceptably small.

We have already noted, in Section 6.3, that the methodsused for assessing the doses received by members of thegeneral public from external radiation sources are quitedifferent in nature from those employed in determining thedoses received by individuals employed in radiation work.Personal dosemeters are hardly ever used to measure thedoses received by members of the general public; instead,reliance is placed on making indirect assessments of suchdoses by measuring dose rates in the environment. A corre-sponding situation exists in relation to doses from internalemitters. Direct, or relatively direct, assessments involvingtechniques such as body monitoring or urine monitoring, arerarely applied to members of the general public. In demon-strating that doses to members of the general public areadequately controlled, reliance is placed on indirect methods,which involve the monitoring of the amount (and nature) ofradioactive material in the environment, or in environmentalmaterials such as crops and foodstuffs.

When radioactive materials are discharged into the atmos-phere, one result may be to increase slightly the externaldose rate in the immediate vicinity of the discharge point,because of the passage overhead of a cloud of radioactive gas.Monitoring for this has already been discussed in Section 6.3.It is unusual for significant amounts of radioactive materialto be discharged into the air in particulate form, since itis usual to remove material from the air by filtration beforedischarge. If particulate matter is discharged, and thusdeposited on the ground, then again the level of externalradiation may be increased; this increase can again be moni-tored by methods already discussed. However, monitoringfor external radiation alone does not always provide ade-quate reassurance. That is because of the fact that someradioactive materials may cause irradiation of man via foodchains. One of the best-known is the grass-cow-milk chain bywhich small releases of 131I can cause significant irradiation ofthe thyroids of humans, and in particular of young children. If1311 is released, some of it falls on grassland. If the grass isgrazed by cows, a particular cow may ingest much of the 131Ideposited on a large area of land; a substantial fraction ofthe 131I which the cow ingests gets into its milk, and if this isdrunk by humans, about 30% or more of the 1311 in the milkgets into the thyroid. This is a very small organ, especially inchildren, and so the radiation dose to it from a given intakeis large. As a result of these factors, the dose to the thyroid ofa person at a certain distance from a given release couldtheoretically be a few hundred times greater than if he wasonly exposed to the 131I by inhalation. In practice the situ-ation is less dramatic because it is rare for a person to get allof his milk from a cow, or herd of cows, grazing permanentlyclose to a release point. Nevertheless there is obviously a needboth to limit fairly drastically the magnitude of such releases,and to monitor their effects.

In such cases it is necessary to take samples of materialssuch as grass or milk, and to measure the very small quantitiesof particular radionuclides (such as 131I) present in them,either by very sensitive gamma spectometry, or by the use ofradiochemical techniques to separate the particular element,the radioactivity of which can then be straighforwardlydetermined in a laboratory with facilities for measuring smallamounts of radioactivity.

The situation is similar if radioactive materials are dis-charged into rivers or the sea. One may need to monitor forenhanced levels of external radiation owing to the depositionof such material of such material on beaches, for example, butin general it is also necessary to monitor materials which maybe involved in food chains. Careful studies are made to estab-

lish so-called critical pathways by which a certain group ofpeople living near a particular installation may get a largerdose of radiation than others, as the result of the release ofone or more radionuclides. For example, in the case of thereprocessing plant at Windscale in Cumbria, which dischargescontrolled amounts of radioactive liquid waste material intothe Irish Sea, special attention was formerly given to oneradionuclide 106Ru, which concentrates to a significant extenton seaweed (porphyra) growing on the Cumbrian coastline,because some of this seaweed was formerly harvested and usedfor the manufacture of laver bread (a 'delicacy' eaten bycertain people in South Wales). Studies of the behaviour of106Ru in the Irish Sea, of the concentration factor in seaweed,of the amount of laver bread eaten by people who were verykeen on it, enabled a limit to be set for the amount of 106Ruwhich could safely be discharged from Windscale. Monitoringfor 106Ru in seaweed on the coast was used to check that allwas well [38]. Note that in these cases it is unusual (althoughnot unknown) for direct monitoring to be done on the peopleconcerned.

In recent years this particular pathway has become lessimportant, partly because seaweed is no longer harvested inthat particular area, and partly because there has been achange in the composition of the discharge from Windscale.That discharge now contains a higher fraction of 137Cs, andthe critical pathway is the consumption of fish caught in theWindscale area and eaten locally. Thus there is particularinterest in monitoring for 137Cs in fish [39]. Fish are caughtin an area immediately south of the discharge point, and thequantity of 137Cs and various other radionuclides present ismeasured.

8 Monitoring in incident and emergency situations

8.1 Accidents involving external radiationFrom time to time people are accidentally exposed to verylarge doses of radiation, e.g. by going into an area where,unknown to them, a very large source is unshielded. Some-times they are wearing a personal dosemeter, sometimes theyare not. Often the dose to different parts of the body is grosslynonuniform. Such situations provide a major challenge tothe health physicist, who has to try to obtain a fairly preciseestimate of the dose received. In a number of cases of thiskind, rather elaborate reconstructions of the incident havebeen made in order to obtain the required information.Each situation tends to be unique and has to be dealt withaccordingly.

In situations of this kind, a biological dosimetry techniqueis available [40]. This can be used when it is known that aperson has been exposed to a large dose of radiation (or whenthere is some suspicion that there may have been such anexposure); it is particularly valuable when a personal dose-meter result is either not available or is thought to be unre-liable. The method depends on the determination of thenumber of defects occurring in the chromosomes of thelymphocytes in a sample of blood taken from the irradiatedperson. The first attempt at such quantitative biologicaldosimetry was made on three men following a criticalityaccident at Hanford in 1962. During the following two dec-ades this method of dosimetry has been systematically devel-oped, and is now a well established routine technique playingan important role complementary to that of physical dosi-metry. About 20 cases of serious (and in some cases fatal)irradiation involving doses of 1 —lOGy have been investigated,in some cases when there were scant data available fromphysical dosemeters. The technique is currently able to detectwhole-body doses down to about 0.1 Gy.

The special case of criticality accident dosimetry is dis-cussed in Section 8.3


8.2 Accidents involving releases of radioactive materialsinto the environment

We have already discussed, in Section 7.4, the monitoringwhich takes place following an accidental release into theworking environment. If, as a result of an accident, radioactivematerial was also released into the general environment, theremight be concern about:

(a) external radiation doses from radioactive materialspassing overhead, or from radioactive materials deposited onthe ground; if doses are likely to be excessive, it might benecessary to evacuate people

(b) releases of radio-iodine isotopes, especially 131I, thiscould cause high doses by direct inhalation, in which caseit would be necessary to issue stable iodine tablets to blockuptake to the thyroid - a harmless but very effective way ofdrastically ameliorating the effects of such uptake; there isalso the possibility of contamination of milk and other food-stuffs, and it might be necessary to ban or control theirconsumption

(c) other releases which could cause irradiation of people,mainly by inhalation; again it might be necessary to evacuatepeople.

In places like nuclear power stations, of course, many pre-cautions are taken to avoid accidents, and further precautionsare taken to avoid releases of radioactive materials even if anaccident were to occur. But arrangements are also made formonitoring teams to be sent out if an accident were to occur,in order to establish as quickly as possible the nature of theradiological situation, so that appropriate remedial actionscould, if necessary, be taken. In general, monitoring wouldbe done using techniques and equipment similar to thoseused in normal operations, as described in earlier paragraphs.

One is, of course, likely to be concerned with the situation inquite a narrow sector downwind of the release point. Air moni-toring in emergency situations is difficult because it is difficultto obtain air samples from appropriate places at appropriatetimes; consequently one is likely to have to use indirect meansof assessing airborne concentrations. Monitoring for 131I maybe made difficult by the presence of other less radiotoxicmaterials in larger quantities. In order to assess the quantityreleased and the amount deposited on grass (which will inturn determine the concentration in milk) it will probablybe necessary to collect grass samples from downwind sitesand subject them to gamma spectrometry in the laboratory,using high resolution lithium-drifted germanium detectors.Crude gamma spectrometry with Nal is possible in the field,but the resolution is likely to be inadequate in many possibleaccident situations.

8.3 Criticality accident monitoringA criticality accident could occur if too much fissile materialwere to be accidentally brought together in one place; thiswould lead to a self-terminating chain reaction with theemission of an intense burst of gamma rays and neutrons.Several such accidents occurred between 1945 and 1960,mainly in the USA. Generally speaking, one person or a fewpersons were exposed to very large doses of radiation in a veryshort time, and some of them died. There was no significantrelease of radioactive material.

Where fissile materials are handled in substantial quantities,elaborate precautions are taken to avoid the accidental assem-bly of a critical quantity of it. But since these precautionsusually depend to some extent on administrative control, thereremains a slight chance of a criticality accident. In this case,automatic criticality alarms are installed. These consist almostinvariably of gamma-ray detectors, similar to those describedin Section 6.1, for detecting any other case of unusually high

gamma dose rates. Because it is vital for them to be workingwhen required, they are generally fitted in a two-out-of-threearrangement, an alarm being given if any two out of threedetectors indicate a high dose rate. The installation is regularlytested by bringing a radioactive source close to each detectorin turn.

If a criticality accident were to occur, it is importantto obtain a reasonably precise estimate of the doses receivedby individuals in the area, especially any who have receiveda large dose, and who may require urgent medical care. Mostof the dose is likely to be due to gamma rays, but neutronsmay account for a significant fraction. The neutron dose isdifficult to measure, as has already been explained in Section6.2. The energy spectrum of the neutrons emitted dependscritically on the amount of moderator present in the imme-diate vicinity of the criticality.

People who work in areas where there are criticality detec-tors are issued with criticality lockets in addition to anyother personal dosemeters [41]. Because it is important toknow the direction of irradiation, they often have equallyspaced lockets on a belt worn round the waist. The type oflocket used in the UK is shown in Fig. 31. Each locket con-tains a number of components intended to provide a measureof the fluence or dose due to neutrons of thermal, inter-mediate and high energy. The locket used in the UK contains:

(a) an indium foil, activated primarily by thermal neutronsbut also to some extent by fast neutrons

(b) a sulphur disc which is activated by fast neutrons(c) a 'sandwich' consisting of a piece of cadmium between

two gold foils. The gold foils become activated, and from theinduced activity it is possible to determine the doses due toboth thermal and intermediate energy neutrons.

The radionuclides produced by irradiation of the variouscomponents are:

(a) in the sulphur, 32P (half-life 14d)(b) in the gold, 198Au (half-life 2.7 d)(c) in the indium, 116mIn (half-life 54min) and 115mIn

(half-life 4.5 h).

Because the half-lives of some of these activation products arequite short, it is necessary to assay the lockets within a shorttime after the accident. The dose from gamma rays is deter-mined from the normal film badge or TLD. It is not advisableto place TLDs in the criticality locket, as there is a danger ofa spurious enhanced reading due to activation of the othercomponents.

People who might be near a place where a criticalityoccurred are issued with a strip of indium in their identifi-cation pass, and/or in their film badge or TLD. Following anaccident, the resulting activation of the foil can be used toidentify people who may have been irradiated, for furtherinvestigation. If they were not wearing criticality lockets,a measurement of their neutron dose can be obtained bymeasuring the amount of 24Na produced by neutron acti-vation of sodium in their bodies (either by putting the personconcerned in a whole-body monitor, or measuring the MNa ina sample of his blood), and by measuring the amount of 32Pproduced by neutron activation of sulphur in the hair. Thesame techniques can of course be used to provide additionalinformation, even when lockets are available.

9 Conclusions

This review has given a fairly detailed account of the currentstate of radiation monitoring. It will perhaps have confirmedthat, as we mentioned earlier, the practices of radiation pro-tection generally, and of radiation monitoring in particular,are much more highly developed and sophisticated than the


corresponding techniques for other toxic substances. Althoughthere have been no dramatic changes in techniques in recentyears, vast numbers of detailed changes of both hardware andprocedures have occurred, as a result of which the nature of

locket cover

serial number disc

card spacer

top gold foil

cadmium foil

bottom gold foil

indium foil

card spacer

polythene spacer

sulpher disc

polythene spacer

locket base

locket retaining button

Fig. 31 (a) Criticality locket (exploded view)(b) Criticality lockets on belt/personal dosemeter.

what is done now is totally different in almost every detailfrom what it was 20 years ago. No doubt this process willcontinue into the future, but it is to be hoped that, beforelong, society may begin to put radiation hazards into betterperspective, and hence to press for improvements in otherspheres to parallel those that have already occurred in radia-tion protection.

In most respects, current practices are reasonably satisfac-tory. There are some areas where improvements are needed,notably in the field of personal neutron dosimetry; it is,however, far from clear at the present time how such improve-ments can be acheived. There is also a need for improvedmethods of monitoring for materials such as plutonium inthe body, but again it is not at all clear how such improve-ments can be made.

A variety of detailed improvements in monitoring tech-niques can be expected. With modern technology, instrumentswill surely become somewhat lighter and more compact, andwe can expect to see a change from analogue to digital dis-plays. We can perhaps expect to see a trend towards moreinstalled instrumentation, but there will continue to be a needfor a lot of manual surveys with portable instruments. It seemscertain that we can expect to see changes in the arrangementsfor monitoring accidental releases, with a considerable increasein the number of instruments installed in the area aroundinstallations from which releases may occur, some, at least,of these instruments feeding back information by radioor direct wires to a central control point.

10 References1 HENSHAW, P.S.: 'Whole-body irradiation syndrome' in CLAUS,

W.D. (Ed.): 'Radiation biology and medicine' (Addison-Wesley,Reading, Massachusetts, 1958), pp. 317-340

2 EVANS, R.D.: 'Radium in man', Health Phys. 1974, 27, pp. 497-510

3 COURT-BROWN, W.M., and DOLL, R.: 'Mortality from cancer andother causes after radiotherapy for ankylosing spondilitis', Br. Med,J. ii. 1965, pp. 1327-1332

4 MORI YAM A, I.M., and KATO, H.: 'Mortality experience ofA-bomb survivors 1950-1972'. JNIH-ABCC Life Span StudyReport 7, Technical report 15-73

5 'Recommendations of the International Commission on Radio-logical Protection'. ICRP Publication 26 (Pergamon Press, Oxford,1977)

6 MANCUSO, T.F., STEWART, A., and KNEALE, G.: 'Radiationexposures of Hanford workers dying from cancer and other causes',Health Phys., 1977, 33, pp. 369-385

7 MARKS, S., GILBERT, E.S., and BREITENSTEIN, B.D.: 'Cancermortality in Hanford workers' in 'Late biological effects of ionisingradiation, Volume 1: Proceedings of the international symposiumheld in Vienna, March 13-17, 1978' (International Atomic EnergyAgency, Vienna, 1978), pp. 369-386

8 REISSLAND, J.A.: 'An assessment of the Mancuso study'. NationalRadiological Protection Board, Harwell, Didcot, 1978, ReportNRPB-R79

9 'Sources and effects of ionising radiation'. United Nations ScientificCommittee on the Effects of Atomic Radiation, Report to theGeneral Assembly, United Nations, New York, 1977

10 'General principles of monitoring for radiation protection of wor-kers'. ICRP Publication 12 (Pergamon Press, Oxford, 1969)

11 DUNSTER, H.J., and JOHNS, T.F.: 'The principles of derivedworking limits and investigation levels in rationalising the designand interpretation of monitoring programmes' in WILLIS, C.A., andHANDLOSER, J.S. (Eds.): 'Health physics operational monitoring'(Gordon and Breach, New York, 1972). pp. 1743-1754

12 'Limits for intakes of radionuclides by workers' ICRP Publication30, Part 1 (Pergamon Press, Oxford, 1979)

13 DUNSTER, H.J.: 'Surface contamination measurements as an indexof control of radioactive materials', Health Phys., 1962, 8, pp. 353-356

14 'The ionising radiations (unsealed radioactive substances) regula-tions, 1968'. Statutory Instruments 1968 No. 780, HMSO, London,1968

15 'Radiation protection instrumentation and its application'. ICRUReport 20, International Commission on Radiation Units andMeasurements, Washington DC, 1971


16 BOAG, J.W.: 'Ionisation chambers' in ATTIX, F.H., and ROESCH,W.C. (Eds.): 'Radiation dosimetry, Second edition, Vol. II: Instru-mentation' (Academic Press, New York, 1966), p. 47

17 RAMM, W.J.: 'Scintillation detectors' in ATTIX, F.H., andROESCH, W.C. (Eds.): 'Radiation dosimetry, Second edition,Vol. II: Instrumentation' (Academic Press, New York, 1966)

18 KUPER, J.B.H., and COWAN, F.P.: 'Dosimetry and instrumentationfor hazard evaluation' in CLAUS, W.D. (Ed.): 'Radiation biologyand medicine' (Addison-Wesley Publishing Company, Reading,Massachusetts, 1958), pp. 438-439

19 ANDERSON, I.O., and BRAUN, J.: 'A neutron rem-counter withuniform sensitivity from 0.025 eV to 10 meV in 'Neutron dosi-metry, Vol. H' (IAEA, Vienna, 1963), p. 87

20 HEARD, M.J., and JONES, B.E.: 'A new film holder for personaldosimetry', in 'Personal dosimetry techniques for external radiation'(European Nuclear Energy Agency, Paris, 1963), pp. 89-106

21 CAMERON, J.R.: Thermoluminescence dosimetry' (University ofWisconsin Press, 1968)

22 JOHNS, T.F.: 'Measurement and assessment of doses from thehandling of reactor fuels' in 'Colloque internationale sur la dosi-metrie des irradiations dues a des sources externes' (Service Centralde Protection contre les Rayonnements Ionisants, Le Vesinet,1967), pp. 80-84

23 MARSHALL, T.O., PATTISON, R.J. TWYMAN, A., PRESTON,H.E , and STEWART, J.C.: The performance of the NRPB thermo-luminescent dosemeter', Nucl. Instrum. & Methods, 1980, 175,pp. 147-149

24 McKINLAY, A.F.: Thermoluminescence dosimetry' (Adam Hilger,Bristol, 1981)

25 'Personnel dosimetry systems for external radiation exposures'IAEA, Technical Report 109 (IAEA, Vienna, 1970), pp. 95-97

26 PIESCH, H., and BURGKHARDT, B.: 'Application of the TLDalbedo technique for monitoring and interpretation of neutron strayradiation fields', Nucl. Instrum. & Methods, 1980, 175, pp. 180-182

27 'Numerical guides for design objectives and limiting conditions foroperation to meet the criterion 'as low as is reasonably achievable'for radioactive material in light-water cooled nuclear power reactoreffluents'. US Federal Register, 40 FR 19439, 1 Aug. 1980

28 PIESCH, E.: 'Application of TLD systems for environmental moni-toring' in 'Proceedings of the 'Applications of TLD' course, Ispra'(Adam Hilger, Bristol, 1980)

Book reviewsModels of high-energy processesJ.C. PolkinghorneCambridge Univeristy Press, 1980, 131 pp., £11.00ISBN: 0-521-22369-5

Field theories of elementary particle interactions have difficultdynamical equations with unmanageable solutions. To come togrips with them, one must first simplify and approximate thefull dynamics in some way; the various resulting prescriptionsare called models. Prof. Polkinghorne has written a compactand elegant introduction to some of these models, intendedfor graduate students.

The first half of the book is about extracting the high-energy behaviour of scattering amplitudes from sets ofFeynman diagrams, leading to Regge singularities and Bjorkenscaling. This is balanced by a long chapter on hybrid models inwhich complete parton-hadron subamplitudes appear, leadingto a fuller discussion of scaling and scale breaking in deepinelastic scattering. A final chapter explains methods forcalculating the discontinuities of scattering amplitudes.

In this area the author is certainly also an authority; astwenty seven of his works, among the listed references, attest.The writing is cool and fluent. The presentation by thepublishers is clear and readable, as we all expect (although aclutch of pages in my copy had eluded the binder's thread).The main trouble with a slim volume is that some things haveto be left out. I would have liked to see more about recentQCD technology, but perhaps it will be covered by a latermonograph in this worthy series.

R.J.N. PHILLIPS

29 SPANNE, P.: Thermoluminesence dosimetry in the jxGy range'Acta Radiologica, 1979, Suppl. 360

30 FRASER, D.C., PERRY, K.E.G., LOOSEMORE, W.R., andSPARKE, W.G.: Techniques for the continuous monitoring ofairborne plutonium activity and experience of their use in a fuel-element fabrication plant' in 'Assessment of airborne radioactivity'(IAEA, Vienna, 1967). pp. 147-160

31 FRASER, D.C.: 'Health physics problems associated with theproduction of experimental reactor fuels containing Pu O2', HealthPhys. 1967, 13, pp. 1133-1143

32 PEIRSON, D.H.: The application of gamma-ray spectrometry inhealth physics' in 'Health physics in nuclear installations' (OECD,Paris, 1959), 147-156

33 McCONNON, D.: The use of water as a sampling medium fortritium oxide'. USEAC Report BNWL-cc-547, 1966

34 'Evaluation of radiation doses to body tissues from internal con-tamination due to occupational exposure'. ICRP Publication 10(Pergamon Press, Oxford, 1968)

35 PEABODY, CO., FRASER, V,M., and SPEIGHT, R.G.: TheAEE Winfrith whole-body monitor'. UKAEA Report AEEW-R.215,1962

36 RAMSDEN, D.: The measurement of plutonium 239 in vivo'.report AEEW-M804, 1968

37 JOHNSON, J.R., and RAMSDEN, D.: 'Investigations into aphoswich detector for plutonium in-vivo monitoring'. UKAEAReport AEEW-M1304, 1975

38 MITCHELL, N.T.: 'Radioactivity in surface and coastal waters ofthe British Isles, 1971', Fisheries Radiobiological LaboratoryTechnical Report FRL9 (Ministry of Agriculture, Fisheries & Food,Lowestoft, 1973)

39 PRESTON, A., JEFFERIES, D.F., and MITCHELL, N.T.: 'Theimpact of I34Cs and I37Cs on the marine environment fromWindscale' in 'Proceedings of a seminar on radioactive effluentsfrom nuclear fuel reprocessing plants, Karlsruhe' (Commission ofthe European Communities, Luxembourg, 1977), pp. 401—420

40 LLOYD, D., and PURROTT, R.J.: 'Chromosome aberration analysisin radiological protection dosimetry', Radia. Prot. Dosimetry, 1981,l,pp. 19-28

41 ADAMS, N.: 'Review of United Kingdom research and experience incriticality dosimetry' in 'Nuclear accident dosimetry systems'(IAEA, Vienna, 1970), pp. 79-101

The rare earths in modern science and technology — Vol. 2G.J. McCarthy, J.J. Rhyne and H.B. Siber (Eds.)Plenum, 1980, 647pp., $59.50ISBN: 0-306-40347-1

This authoritative work presents the proceedings of the 14thrare earth research conference, held in June 1979, at NorthDakota University, Fargo. At this meeting the Frank H.Spedding Award was established in the presence of this pioneerof rare-earth research and development. The first recipient ofthe award, W.E. Wallace, gave a plenary address on thisoccasion, entitled: 'Studies of rare-earth intermetallic com-pounds and rare-earth hydrides'. This talk excellently sum-marises some of the scientific and technological aspects ofthese fascinating materials. Thus rare-earth hydrides, studied inthe speaker's own laboratory, are of scientific interest as theirphysical properties are determined by indirect exchange inter-actions. Rare-earth intermetallics of the type SmCo5 andSm2Co17 have many uses as permanent-magnet materials,LaNi5 in heterogeneous catalysis and the projected applicationsof rare earth; iron systems are also noted, for example astransducer elements. These and other topics are also mentionedin the various papers given elsewhere in the book. Subjectheadings include: spectroscopy (luminescence, fluorescence,laser, Mossbauer, ESR); metallurgy and materials preparation;solution, solvation and analytical chemistry; X-ray and neutrondiffraction; transport and thermal properties; hydrides;magnetism and rare-earth technology. The impression given bythis book is one of immense activity at all levels, coupled withan optimistic view of the promise of these materials for futuretechnological applications.

E.P. WHOLFARTH


purpose and practice of radiation monitoring

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