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RADIONUCLIDE AND RADIATION PROTECTION DATA HANDBOOK 2002 D. Delacroix* J. P. Guerre** P. Leblanc** C. Hickman * Commissariat a ` l’Energie Atomique, CEA/DAM - Ile de France, France **Commissariat a ` l’Energie Atomique, CEA/Saclay, France ISBN 1 870965 87 6 RADIATION PROTECTION DOSIMETRY Vol. 98 No 1, 2002 Published by Nuclear Technology Publishing

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Page 1: RADIONUCLIDE AND RADIATION PROTECTION DATA HANDBOOK · PDF file5 Radiation Protection Dosimetry Vol. 98, No. 1, pp. 5–6 (2002) Nuclear Technology Publishing RADIONUCLIDE AND RADIATION

RADIONUCLIDEAND

RADIATION PROTECTIONDATA HANDBOOK 2002

D. Delacroix*J. P. Guerre**P. Leblanc**C. Hickman

* Commissariat a l’Energie Atomique, CEA/DAM - Ile de France, France**Commissariat a l’Energie Atomique, CEA/Saclay, France

ISBN 1 870965 87 6RADIATION PROTECTION DOSIMETRY Vol. 98 No 1, 2002Published by Nuclear Technology Publishing

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RADIONUCLIDE AND RADIATION PROTECTION DATA HANDBOOK2nd Edition (2002)

All rights reserved. No part of this book may be reproduced, stored in a retrieval system ortransmitted in any form or by any means, electronic, electrostatic, magnetic, mechanical, photo-copying, recording or otherwise, without permission in writing from the publishers.

British Library Cataloguing in Publication Data

A catalogue record of this book is available at the British Library

ISBN 1 870965 87 6COPYRIGHT 2002 Nuclear Technology Publishing

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Radiation Protection DosimetryVol. 98, No. 1, pp. 5–6 (2002)Nuclear Technology Publishing

RADIONUCLIDE AND RADIATION PROTECTION DATA HANDBOOK2nd Edition (2002)

ContentsContents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5

Preface . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7

INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9

SCOPE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9

CLASSIFICATION OF RADIONUCLIDES . . . . . . . . . . . . . . . . . . . . . . . . . . . 10

PHYSICAL CHARACTERISTICS OF RADIONUCLIDES. . . . . . . . . . . . . . . . . . . . . 10

RADIOLOGICAL CONTROL . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10

External exposure risks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10Internal exposure risks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10Dose limits for workers . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11

EXTERNAL EXPOSURE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11

Assumptions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11Distant (point source) external exposure. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11Exposures to a uniformly contaminated surface . . . . . . . . . . . . . . . . . . . . . . . . . 12External contact exposure (receptacles) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12

SHIELDING OF BETA AND GAMMA EMITTERS . . . . . . . . . . . . . . . . . . . . . . . 12

CONTAMINATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12

Contact exposure due to external skin contamination. . . . . . . . . . . . . . . . . . . . . . . 12Decay products . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12Decay of heavy elements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13Derived surface contamination limits (DSCL) . . . . . . . . . . . . . . . . . . . . . . . . . . 13Determination of removable and fixed contamination values . . . . . . . . . . . . . . . . . . . 14Contamination control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14Detection probes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14

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CONTENTS (continued)

INTERNAL EXPOSURE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14

Committed effective dose per unit intake . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14Annual Limits on Intake (ALI) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14The ‘highest dose organ’. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15General . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15

MAXIMUM RECOMMENDED ACTIVITIES. . . . . . . . . . . . . . . . . . . . . . . . . . . 15

Calculation models . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15Dependence of maximum recommended activities on equipment and working areas . . . . . . . . 15Limitations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16

DESCRIPTION OF DATA SHEETS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16

GLOSSARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16

REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17

TABLE OF RADIONUCLIDES LISTED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19

DATA SHEETS (144) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21

APPENDIX – HEAVY ELEMENT DECAY . . . . . . . . . . . . . . . . . . . . . . . . . . . 165

is abstracted or indexed in RADIATION PROTECTIONABSTRACTS, Chemical Abstracts, CURRENT CONTENTS, Energy Information Abstracts (Cambridge),EXCERPTA MEDICA (EMBASE), Health and Safety Science Abstracts (Cambridge), INIS ATOMINDEX(hard copy and CD ROM), INSPEC, Nuclear Energy (Czech Republic), QUEST and Referativmaja Zhurnal.

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Radiation Protection DosimetryVol. 98, No. 1, p. 7 (2002)Nuclear Technology Publishing

Preface

RADIONUCLIDE AND RADIATION PROTECTION DATA HANDBOOK (2002)

This handbook is an updated and expanded (2nd edition) version of the handbook with the same title publishedin 1998. That handbook was in turn based on an earlier French laboratory guidebook A Guide on Radionuclidesand Radioprotection by D. Delacroix, J.P. Guerre, and P. Leblanc, published in 1994(1). The earlier publicationwas very much oriented towards guidance on the handling of radionuclides used in medicine. The present handbookis much more broadly based in its outlook and application and incorporates updated information to take accountof the most recent ICRP and IAEA recommendations(1–4). The radionuclides listed in the earlier publication weresupplemented by commonly used radionuclides in the nuclear industry and in other areas. Moreover, the mainpurpose of this handbook (and its 1998 predecessor) is the provision of data sheets and the models or sourcesused to assemble the data, rather than as a radiation protection guide for laboratory users.

This practical handbook of data for handling radioactive materials is intended for radiation protection specialistsas well as all others who use or transport radionuclides. Its publication should satisfy a major need for all healthphysics departments and is intended to assist in informing and training personnel in radiation protection. It consistsof an explanatory text followed by specific radiation data sheets for selected radionuclides. Because of the dispari-ties in the approaches adopted by different countries, it is essential that users also refer to the relevant nationalregulations with which they must comply.

The present handbook includes 36 additional radionuclides than the previous edition, giving a total of 144nuclides. Additional data are included covering dose rates above uniformly contaminated surfaces (infinite planesource). Some corrections to data in the previous edition have been incorporated. The list of nuclides is notexhaustive but those included have been selected on the basis of being most commonly used, taking into accountthe requirements of users in industry, medicine and research. Finally, account is taken of the need to considerdecay chain products for several heavy elements. To this end, an appendix is provided giving the decay productsand their activities as a function of time (age), together with decay charts for four of the more complex chains.

Directeur Central de la SecuriteCommissariat a l’Energie Atomique, France

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Radiation Protection DosimetryVol. 98, No. 1, pp. 9–18 (2002)Nuclear Technology Publishing

RADIONUCLIDE AND RADIATION PROTECTION DATAHANDBOOK 2nd Edition (2002)D. Delacroix*, J. P. Guerre**, P. Leblanc** and C. Hickman*Commissariat a l’Energie Atomique, CEA/DAM - Ile de France, France**Commissariat a l’Energie Atomique, CEA/SACLAY, France

Abstract — This handbook is a reference source of radionuclide and radiation protection information. Its purpose is to provideusers of radionuclides in medicine, research and industry with consolidated and appropriate information and data to handle andtransport radioactive substances safely. It is mainly intended for users in low and intermediate activity laboratories. Individualdata sheets are provided for a wide range of commonly used radionuclides (144 in total). These radionuclides are classified intofive different groups as a function of risk level, represented by colours red, orange, yellow, green and blue, in descending orderof risk.

INTRODUCTION

In recent years the need for a compilation of essentialdata for commonly used radionuclides, for both practiceand training, has become increasingly apparent.

This handbook contains individual data sheets for arange of radionuclides and is mainly intended for usersof unsealed radioactive sources in low and intermediateactivity laboratories, industrial applications and trans-port.

Individual data sheets corresponding to the differentradionuclides give the physical characteristics, referencetransport activities, exemption levels, contaminationderived limits and appropriate detection probes, doserates from skin contamination, half and tenth valueshielding thicknesses, external exposure data, dose coef-ficients for ingestion and inhalation, derived 20 mSvALI values, and the highest dose organ. The maximumactivities recommended to be handled on benches, underfume hoods and in glove boxes are also given. However,this information is most appropriate in laboratorieshandling low and intermediate levels of radioactivematerials. In order to facilitate the user’s understanding,pictorial illustrations have been employed and data ispresented in tabular form.

It should be noted that the data provided, althoughdrawn from appropriate recent international recommen-dations, should in no way be taken to supersede existinglocal or national regulations. The data may, of course,be used to review existing local regulations which mayhave become out of date.

The text preceding the data sheets provides the back-ground, derivation and substantiation for the data given,enabling users to satisfy any quality assurance arrange-ments that may be in place.

SCOPE

This handbook provides the most up to date internaland external dosimetry information and control criteria

in the form of data sheets in an easy to use format for144 of the most commonly used radionuclides in medi-cine, research and industry. It is designed as a readyreference source of information, gathering together inone place, in the form of individual data sheets, the mostup to date isotopic and radiation protection data. Theinformation includes:

� half-life and specific activity,� main emissions,� IAEA Basic Safety Standards exemption levels,� IAEA A1 and A2 transport values,� external exposure data for a range of geometries,� surface contamination data, detection and limits,� shielding information,� ICRP dose per unit intake data by ingestion and

inhalation,� 20 mSv ALI values and the highest organ dose, and,� maximum recommended activities in Controlled and

Supervised Areas.

Most of the data provided are relevant to all users.However, the section and data dealing with maximumrecommended activities are only appropriate to low andintermediate activity laboratories. These laboratories arefrequently limited to two to three orders of magnitudehigher than the exemption limits by local or nationalregulations.

CLASSIFICATION OF RADIONUCLIDES

Radionuclides have been divided in this handbookinto five ‘risk groups’. The classification used is basedon the BSS ‘quantity’ exemption limit values. Exemp-tion limit values have been chosen for this purposebecause they reflect both internal and external exposurerisks. Each group is associated with a colour (red,orange, yellow, green and blue successively representdecreasing levels of risk). The data sheets have beenappropriately coloured providing the reader with a rapidvisual means of appreciating the risk associated with

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particular radionuclides. The five risk groups have beendefined as follows:

Group 1: exemption limit � 104 Bq (red)Group 2: exemption limit = 105 Bq (orange)Group 3: exemption limit = 106 Bq (yellow)Group 4: exemption limit = 107 Bq (green)Group 5: exemption limit � 108 Bq (blue)

In the rare cases for which exemption levels have notbeen defined (e.g. 11C, a nuclide sometimes used formedical imaging) a group has been assigned by analogywith other radionuclides of the same type.

Attention is drawn to the fact that radioactive sub-stances are also chemical substances, the handling ofwhich may involve hazards of non-radiological origin.Moreover, radioactive substances are often intimatelymixed with chemical products having an aggressiveaction on the human body (toxic, mutagenic and car-cinogenic effects). Chemical hazards should thereforenever be neglected.

PHYSICAL CHARACTERISTICS OFRADIONUCLIDES

The following characteristics are considered inthese sheets.

Half-lives expressed in the most appropriate units(hours, days, years).

Specific activities expressed as Bq.g�1.

Exemption levels

Exempt activity concentrations (Bq.g�1) and exemptactivities (Bq) are defined in the IAEA Basic SafetyStandards(2) and in the L159 Euratom Directive(3).These limits can be used by regulatory authorities todefine criteria for exemption against formal registrationof the premises and may also be used to develop clear-ance levels for materials leaving the premises. Theexemption levels also apply in the transport regulations.The correct use of these limits by users is of utmostimportance. Users should refer to the above men-tioned references.

A1 and A2 reference transport activities

A1 and A2 are the maximum activities (expressed inTBq) that can be transported by A type packages. A1

corresponds to radioactive matter in ‘special form’ andA2 to the other cases.

In the data sheets, the values correspond to values inthe 1996 edition of The IAEA publication Regulationsfor the Safe Transport of Radioactive Materials(4), withwhich the international modal transport authorities(RID, ADR, IMO, ICAO, etc.) are harmonised. Usersseeking further details are referred to the referencedtexts.

The nature of the main radiation(5)

�, �, e�, X, � radiation and neutron emissions aregiven, together with the corresponding energiesexpressed in keV, and corresponding percentage emis-sions. A maximum of the three most characteristic emis-sions is given for each type of radiation, the criterionadopted being the relevance of the emissions from theradiation protection standpoint. However, in order toremedy the restrictive effects of this approach (someradionuclides exhibit multiple emissions, e.g. 140La and152Eu), the relative percentage emissions not taken intoconsideration are nevertheless given.

RADIOLOGICAL CONTROL

External exposure risks

The following modes of exposure are considered.

Contact

When handling radioactive materials, the operatormay touch the receptacle (e.g. beaker, pipette, syringe)containing the radioactive substance or materials con-taminated with this substance. The radioactive substancecan also become deposited on the skin.

Distant source

All parts of the operator’s body generally remain dis-tant from the radioactive substances being handled, eventhough certain parts of the body such as the hands andforearms occasionally approach these substances duringhandling operations with or without manipulators.

Immersion in a cloud

This type of relatively infrequent hazard is encoun-tered when handling gaseous radioactive sources. Thecorresponding risks must be taken into consideration,for example, in laboratories using cyclotrons to manu-facture short half-life radionuclides (e.g. 18F, 11C, 123I).

Internal exposure risks

The risks arise from three different possible paths intothe human body:

� Inhalation after the dispersion of gases, vapours oraerosols in the environment;

� Ingestion by contaminated hands or objects throughcontact with the mouth;

� Transfer through the skin with or without an associa-ted wound following contamination of parts of thebody (which have not been sufficiently rapidlydecontaminated) or following simple contact withhighly penetrating radionuclides (e.g. tritium in

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tritiated water) or chemically aggressive substances(acidic solutions, solvents, etc.).

Dose limits for workers(2,6)

The occupational exposure of any worker should notexceed the following limits, but may also be subject tomore restrictive limits in appropriate local or nationalregulations.

� an effective dose of 100 mSv over a period of fiveconsecutive years (i.e. 20 mSv per year) averagedover this period;

� an effective dose of 50 mSv in any single year;� an equivalent dose to the lens of the eye of 150 mSv

in a year;� an equivalent dose to the extremities (hands and feet)

or the skin of 500 mSv in a year.

It is noted that the IAEA(2) states that

Compliance with the foregoing requirements forapplication of the dose limits on effective doseshall be determined by one of the followingmethods:

(a) by comparing the total effective dose with therelevant limit, where the total effective dose ET iscalculated according to the following formula:

ET = Hp(d) + �je(g)j,ingestion.Ij,ingestion

+ �je(g)j,inhalation.Ij,inhalation,

where Hp(d) is the personal dose equivalent fromexposure to penetrating radiation during the year.Note that e(g)j,ingestion and e(g)j,inhalation are the com-mitted effective dose per unit intake by ingestionand inhalation for radionuclide j by the group ofage g, and Ij,ingestion and Ij,inhalation are the intakes viaingestion or inhalation for radionuclide j during thesame period.

(b) by satisfying the following conditions:

Hp(d)/DL + �j(Ij,ingestion/Ij,ingestion,L)

+ �j(Ij,inhalation/Ij,inhalation,L) � 1

where DL is the relevant dose limit on effectivedose, and Ij,ingestion,L and Ij,inhalation,L. are the annuallimits on intake (ALI) via ingestion or inhalationof radionuclide j.

(c) by any other approved method.

Note that the symbols used in the quote are consistentwith those used in this handbook but not the originalIAEA document. The above quantities are brieflyexplained in Schedule II of Reference 2.

EXTERNAL EXPOSURE

Assumptions

X ray, �, �, e� and neutron emissions lead to externalexposures depending on the nature and energies of thecorresponding radiation as well as the packaging andgeometry of the source. Depending on the particularcase being studied, superficial exposures (or skinexposures) defined at a depth of 7 mg.cm�2 or deepbody exposures(7) are taken into consideration.

Various handling postures and situations commonlyencountered in laboratories are considered. Calculationshave been performed for unique values of source or sol-ution activity; in this way a direct comparison betweenthe values obtained under different circumstances canbe made. A nominal activity of 1 MBq has been adoptedfor external exposures, exposures resulting from con-tacts with receptacles and exposures to distant sources.A nominal value of 1 kBq has been adopted for the caseof skin contamination due to droplets and 1 kBq.cm�2

for uniform skin contamination. Exposures due toan infinite and uniformly contaminated surface at1 MBq.cm−2 are also considered. Data are provided fordistances of 10 cm and 1 m from the contaminated sur-face for beta and photon radiation for the skin and deeptissue. The quantity adopted for all geometries for tabu-lated data is dose equivalent, expressed in mSv.h�1

(millisievert per hour).The handling postures adopted in this handbook for

some operations with radionuclides in the form of un-sealed sources may not actually represent what is donein practice (e.g. use of a syringe or beaker). However,the user, by analogy, can transform the values indicatedinto those corresponding to other receptacles of similardimensions, on the basis that the situations shown arereasonably typical.

The following codes have been used for the calcu-lations:

� Microshield Version 4.10 has been used for � andX ray calculations (Grove Engineering, 15215 ShadyGrove Road, Rockville, USA, 1996)

� Varskin mod 2 for � radiation and Varskin mod 2modified for monoenergetic electrons (J.S. Durham,Pacific Northwest Laboratory, PO Box 999, Rich-land, Washington 99352, USA).

Distant (point source) external exposure

The cases of a point source at a distance of 30 cm(average length of forearm) in air and a ‘penicillin’ typevial at a distance of 1 m in air(8) are considered success-ively. In the latter case, the source is represented by acylinder with a density of 1 g.cm�3, 2.3 cm in diameterand 2.5 cm high enclosed in a 1.5 mm thick glass(density 2.7 g.cm�3) envelope.

In the case of a point source at 30 cm, the deep orwhole-body dose equivalent due to X and � ray compo-

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nents is distinguished from the superficial or skin doseequivalent due to � and e� components. This is donein order to draw the operator’s attention to the fact thatthe superficial dose equivalent can be considerablyreduced by interposing a screen with a thicknessequivalent to the maximum range of the � and e�

components.A deep dose equivalent is given for a vial at 1 m (this

corresponds to the distance between the operator andthe bench or table).

Exposures to a uniformly contaminated surface

Exposures due to a uniformly contaminated source(floors) are considered. The contributions due to betaradiation and photons are considered separately for dis-tances of 10 cm and 1 m from an infinitely and uni-formly contaminated surface. This separation enablesusers to make appropriate decisions as to what measure-ment equipment should be used. Pure beta emittersgenerally cannot be detected with gamma detectors.Dose rates at 10 cm and 1 m are often similar becausethe increased distance is compensated by theincreased solid angle. Data is quoted for the doseequivalent at depths representing the skin and deeptissue respectively.

External contact exposure (receptacles)

The cases of a 50 cm3 beaker containing 20 cm3 ofsolution and a 5 cm3 syringe containing 2.5 cm3 of sol-ution are considered. The beaker is represented by a cyl-inder with a density of 1 g.cm�3, 4 cm in diameter,1.6 cm high enclosed in a 2 mm thick glass (density2.7 g.cm�3) envelope.

The syringe is represented by a cylinder with a den-sity of 1 g.cm�3, 1.2 cm in diameter, 2.2 cm highenclosed in a 1 mm thick plastic (density 1 g.cm�3)envelope.

In both cases dose equivalents have been calculatedunder 7 mg.cm�2 at the level of the solution (an arrowindicates the position taken into consideration)(8).

The position indicated corresponds to the maximumdose equivalent to which an operator can be exposedwhen manipulating these receptacles. Attention is drawnto the fact that these drawings, for pedagogical reasons,show the operator’s hands and fingers as far away fromthe source as possible.

The dose equivalents depicted take both � and �components into consideration. It should, however, benoted that in the case of a beaker (2 mm thick glass), �components (other than those emitted by radionuclidessuch as 42K, 90Y and 144Pr) can be neglected. In the caseof a syringe (1 mm thick plastic), the � contribution canbe highly significant (e.g. 24 mSv.h�1 for 1 MBq 32P).

Note that good agreement is observed between thevalues quoted here and those found in the literature(9–12).

� ray absorption in shielding gives rise to bremsstrah-

lung (X rays). With high activity sources bremsstrah-lung radiation can make a significant contribution todose. Attention is drawn to this in the data sheets. It isfor this reason that ’Brem. Rad.’ is indicated on datasheets corresponding to pure � emitters (14C, 35S, 33P,36Cl and 45Ca) for which the calculated dose ratesbehind shielding are very low or even zero. These radio-nuclides are very much used in laboratories.

SHIELDING OF BETA AND GAMMA EMITTERS

The range of betas and electrons and the attenuationof X and � rays in shielding materials depends on theenergies of the incident radiation and the nature of theshielding material.

Total absorption thicknesses have been determinedfor � and e� radiation in glass and plastic(13); thesematerials are those most commonly used for this pur-pose. In the case of � and X rays, the first half-value andtenth-value thicknesses of lead and steel (attenuating theincident radiation by a factor of 2 and 10 respectively)are given.

Attention is drawn to the fact that some radionuclidessimultaneously emit significant low and high energycomponents for which first half-value and tenth-valuethicknesses may be considerably less than second half-value and tenth-value thicknesses. 123I is an example butthe data provided here do not cover this situation.

CONTAMINATION

Contact exposure due to external skincontamination

Two situations have been considered to illustratebody contamination. Firstly, exposure due to an exten-sive uniformly spread out contamination of theskin(9,14,15) is considered and secondly, the projection ofa 0.05 cm3 droplet of a radioactive substance. The drop-let is represented by a cylinder with a density of1 g.cm�3, a cross sectional area of 1 cm2 and a heightof 0.5 mm(8).

Dose equivalent values have been calculated for anaverage 70 �m basal layer depth. At this depth the maincomponent of the dose resulting from superficial con-tamination of the skin is due to � rays and electronsfrom the radionuclide; the � contribution to the dose isgenerally just a few per cent. Comparison between thevalues due to a uniform (infinitely thin) deposit andthose due to a droplet highlights the effects of � attenu-ation in tissue. It is assumed here that penetration of thecontamination in the skin can be neglected.

Decay products

Decaying radionuclides are generally treated asfollows:

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� radionuclides (parent–progeny) couples are assumedto be in equilibrium;

� calculations are performed for a nominal activity ofthe parent nuclide (e.g. 1 kBq.cm�2 for uniformskin contamination);

� parent–progeny radionuclide decay schemes arenormally covered in a single sheet, but when physi-cal separation of the constituents is possible (e.g. byelution), progeny products are treated in individualsheets. The 99Mo–99mTc couple is an example of this.The reader will find both 99Mo–99mTc and 99mTcsheets.

Decay of heavy elements

In the case of heavy element decay chains (e.g. lead,radium, uranium, plutonium and americium), where theparent and progeny are not necessarily in equilibrium(e.g. after chemical separation or extraction of anelement) the respective activities of each of the progenyin the chain will depend upon the age of the sample. Itis not possible to cover all situations likely to beencountered but the Appendix to this handbook includesselected major heavy element decay chains and theactivities of respective progeny in the chain are givenas a fraction of the parent activity as a function of time(age since separation) for 0.1 day, 1 day, 10 days, 100days, 1 year, 10 years and 100 years. The parent radio-nuclides for which this information is provided aremarked ‘!!’ in the main data sheets, including an indi-cation of the first and last elements within the chains.

A similar approach can be adopted if data on naturaluranium or re-constituted ‘natural’ uranium are requiredby taking into account the isotopic composition.

The data in the Appendix are derived using RadDecayversion 1.13(16).

Derived surface contamination limits

The derived surface contamination limit (DSCL) is anon- regulatory quantity enabling exposure risks due toremovable and/or fixed surface contamination to bequantified. Exposure to the DSCL leads to a dose whichdoes not exceed the maximum annual occupationalexposure limit (corresponding to 2000 working hours).A model is used to determine the DSCL values corre-sponding to the different radionuclides dealt with(17).This model takes different modes of exposure and trans-fer parameters into consideration. These data are givenfor completeness and the values do not supersede limitsimposed by regulations, which are invariably muchlower.

Occupational exposures may result from:

� transfer of surface contamination to the atmosphereleading to internal exposure through inhalation

� transfer of surface contamination to the organismthrough ingestion

� transfer of surface contamination to the skin leadingto external extremity exposures

� whole-body exposure due to surface contamination.

The DSCL is defined by the following equation:

1/DSCL = 1/Asatm + 1/Asingestion

+ 1/Asskin + 1/Assurf

where Asj is the activity per unit surface area of thecontaminated zone leading to annual limits due to eachof the four modes of exposure (considered separately).Limits of 20 mSv.y�1 and 500 mSv.y�1 are consideredfor whole body external exposure and skin exposure,respectively. The most restrictive ALIinhalation andALIingestion values calculated for 20 mSv.y�1 are used inthe following derivations of surface activity for limitinginternal contamination.

Calculation of Asatm

Asatm is the value of the activity per unit surface areaof the contaminated zone which leads to an ALIinhalation

dose with an atmospheric re-suspension factor, Tatm, of10-4 m�1 and an annual occupational exposure of 2000 hwith a respiratory volume rate, R, of 1.2 m3.h�1.

Asatm (Bq.cm�2) is given by the following equation:

Asatm = ALIinhalation/(Tatm × 2000 R × 104)

Calculation of Asingestion

Asingestion is the value of the activity per unit surfacearea of the contaminated zone which leads to anALIingestion dose with an organ transfer factor, Tingestion,of 1 cm2.h�1 and an annual occupational exposure of2000 h. This component may simulate, for example, therisk due to a chronic exposure at a working post.

Asingestion (Bq.cm�2) is given by the followingequation:

Asingestion = ALIingestion/(2000 Tingestion)

Calculation of Asskin

Asskin is the value of the activity per unit surface areaof the contaminated zone leading to an annual skin doseof 500 mSv for an occupational exposure of 2000 h anda skin transfer factor, Tskin of 0.1. It is assumed thatcontamination is eliminated on a daily basis when theuser washes on leaving the working zone. Neither radio-active decay nor the renewal of skin cells (which con-tribute towards eliminating radioactivity) are takeninto account.

Asskin (Bq.cm�2) is given by the following equation:

Asskin = 500/(Tskin × 2000 × Dp)

where Dp (mSv.h�1) is a conversion coefficient giving

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the dose rate in tissue beneath 7 mg.cm�2 for a skincontamination of 1 Bq.cm�2(14,15).

Calculation of Assurf

Assurf (Bq.cm�2) is the value of the activity per unitsurface area of the contaminated zone leading to aneffective dose equivalent of 20 mSv for an occupationexposure of 2000 h.

Assurf (Bq.cm�2) is given by the following equation:

Assurf = 20/(2000 Dfloor)

where Dfloor (mSv.h�1) is a conversion coefficient givingthe effective dose equivalent 1 m above a floor infinitelyand uniformly contaminated at 1 Bq.cm�2(14).

Alpha emitters

The DSCL value corresponding to most radionuclideslies between 1 Bq.cm�2 and 105 Bq.cm�2, and, in gen-eral, is of the order of 100 Bq.cm�2. In the case of alphaemitters, the model can lead to very restrictive values;a lower limit of 0.04 Bq.cm�2 has been imposed foralpha emitters.

Determination of removable and fixedcontamination values

DSCL values enable derived limits to be determinedfor removable (or non-fixed) and fixed contamination.The above limits enable the potential risk to users fromsurface contamination to be assessed together with themost appropriate means for decontamination.

In the case of removable contamination, the derivedlimit is defined as being equivalent to 1/10th of theDSCL value of the radionuclide considered (this allowspossible occupational exposures of other origins to betaken into consideration).

In the case of fixed contamination, the derived levelis defined as the value corresponding to a whole-bodyexposure from a uniformly contaminated floor equival-ent to 1/10th of the annual exposure limits of 20 mSv.The DSCL value for fixed contamination alwaysexceeds the value for removable contamination. Forsome radionuclides such as ‘pure’ beta or alpha emit-ters, the ratio of the two values can be highly significant(radiation from the floor being negligible). In this hand-book, the DSCL value for fixed contamination is set at amaximum of 100 times the derived limit for removablecontamination (e.g. see 14C on page 24). Regulations donot usually include specific limits for fixed contami-nation since its effect is taken into account in externalradiation levels.

Contamination control

The level of removable contamination of workingzones and materials must be kept as low as reasonably

practicable by appropriate maintenance and cleaningoperations. When accidental contamination (even ofvery low level) occurs, decontamination must be carriedout as quickly as possible.

Detection probes

Surface contamination due to many of the radio-nuclides covered in these sheets can be detected withsoft �, �, �, X ray or � probes.

Preferences given to particular types of probe orinstrument are indicated for each type of radiation. Asingle + means that such a type of probe or instrumentcan possibly be used, double ++ means that this type ofprobe or instrument is recommended, whilst the absenceof a + sign means that that type of probe or instrumentis not suitable. Absence of a + sign in any box meansthat direct measurement is inappropriate and a wipe test,in association with a suitable detection system, such asliquid scintillation methods, would be appropriate. Thisapplies only to tritium in this handbook but the sametechnique can be applied with advantage to low levelsof contamination of other radionuclides, especially thosewith low penetration emissions.

INTERNAL EXPOSURE

Committed effective dose per unit intake

Internal exposures resulting from ingestion andinhalation are evaluated with the dose coefficients,e(g)ingestion and e(g)inhalation, expressed in Sv.Bq�1, andidentically given by the ICRP(18), the IAEA BasicSafety Standards(2) and the L159 Euratom Directive(3).Listed values take the age of exposed individuals intoaccount (1 year, 1–2 years, 2–7 years, 7–12 years, 12–17 years and 17 years) as well as their status (publicor occupational exposure). In this handbook only occu-pational exposures are considered.

The committed effective dose per unit intake viaingestion is given for different gut transfer factor, f1,

values. This factor quantifies the proportion of intaketransferred to body fluids in the gut and depends on thechemical form of the radionuclide.

The committed effective dose per unit intake via inha-lation is given for three default lung absorption types(fast, moderate, slow). These coefficients depend on thechemical form and particle size of the aerosol. It isrecalled that particle size is quantified by aerodynamicmedian activity diameter (AMAD) parameters. Inha-lation coefficients are given for 1 and 5 �m AMADvalues. Indication of the different forms for each nuclidethat correspond to different gut transfer factors and lungabsorption types are given in the data sheets. In somecases, for example phosphorus and sulphur, where thenuclide may form part of the anion, the gut transferfactor or lung absorption types may be predominantly

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determined by the associated cation. These are indicatedin the data sheets.

Annual Limits on Intake (ALI)

Annual Limits on Intake by ingestion and inhalation(Iingestion,L and Iinhalation,L) are given in order to assistusers who are not familiar with the dose per unit intakecoefficients discussed above. Iingestion,L and Iinhalation,L arecalculated using the following relationship(19):

IL = DL/e(g)

where DL is the average annual limit of 20 mSv. IL

values can be considered to be ALI values.ALIingestion and ALIinhalation values are given in the

data sheets for an average annual 20 mSv limit and themost restrictive e(g) values, respectively. They are cal-culated values(19), not values given by ICRP, and usersshould be aware of this. The references should be con-sulted, where appropriate, for further information.

The ‘highest dose organ’

Various metabolic and dosimetric models haveenabled retention and excretion functions after intakeby inhalation by Reference Man to be established(19–21).These models have also enabled committed whole-bodyand organ doses to be determined. Some organs prefer-entially concentrate particular radionuclides. Therefore,the data sheets indicate which is the ‘highest doseorgan’ (e.g. the thyroid in the case of iodine nuclides).The ‘highest dose organ’ depends on physicochemicalform and whether intake results from ingestion or inha-lation. Only inhalation (for an AMAD of 1 �m) is con-sidered in this handbook. It should also be noted thatthese functions depend on the clearance time from thepulmonary region to the remainder of the body (F, M,S). Therefore, the ‘highest dose organ’ may also dependon the elimination class. 90Sr is an example; dependingon whether F or S classes are considered either bonesurface or lungs can be considered as being the ‘highestdose organ’. However, for some chemical forms theassociated anion or cation may be more important inthis context.

General

Internal exposure values are determined using variousestablished metabolic and dosimetric models andhypotheses, which are subject to review by ICRP.Interpretation of the results of these analyses should bedone by specialists.

MAXIMUM RECOMMENDED ACTIVITIES

Operators or competent radiation protection personneloften require order-of-magnitude estimates on themaximum activities that can be handled with available

equipment (chemical benches, fume hoods, gloveboxes) or in given working areas (controlled or super-vised areas). Whilst the purpose of this handbook is toprovide data rather than advice on practices, it is feltappropriate to include information on ‘maximumrecommended activities’.

The data provided are appropriate to low and inter-mediate level laboratories and not to high level labora-tories, where extensive professional radiation protectionexpertise will be available.

Calculation models

The maximum activity (Ao) corresponding to a parti-cular nuclide and handling situation is characterised bythe potential risk associated with handling specificnuclides. External exposure due to � and � radiationcomponents and the re-suspension of the nuclide in theatmosphere are taken into consideration.

Ao (Bq) is determined from the following relation-ship:

1/Ao = 1/Abeta + 1/Agamma + 1/Avolatile

where Abeta (Bq) and Agamma (Bq) represent themaximum activities determined for a point source at adistance of 30 cm in air. These activities are calculatedfor the following dose rate limits:

250 �Sv.h�1 under 7 mg.cm�2 for beta components,10 �Sv.h�1 for gamma components.

Avolatile (Bq) is related to potential atmospheric con-tamination risks directly dependent on the volatility ofthe radioactive product.

In free air, Avolatile is given by the expression:

Avolatile = 10�2 ALIinhalation/k,

where k is the volatility factor.It should be noted that the maximum recommended

activities given in these sheets have been determinedusing the most restrictive ALIinhalation values describedabove; k values, related to the volatility of the product,are defined as follows:

� k = 1: gases, substances with high saturation vapourpressures of about 1 bar at 20°C, substances pen-etrating the skin, particles smaller than 5 �m.

� k = 0.1: substances with saturation vapour pressuresof about 0.1 bar at 20°C, particles larger than 5 �m.

� k = 0.01: low volatility substances with saturationvapour pressures of about 0.01 bar at 20°C (e.g.water).

� k = 0.001: non-volatile substances with saturationvapour pressures less than 0.01 bar at 20°C.

Dependence of maximum recommended activitieson equipment and working areas

The following additional rules are used to determine

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the recommended limits for benches, fume cupboardsand glove boxes:

� It is forbidden to manipulate highly volatile sub-stances (k 0.01) or radionuclides with ALIinhalation

values lower than 1000 Bq on chemical benchesbecause even a low level contamination could leadto a significant ALIinhalation fraction that would be dif-ficult to detect.

� Manipulating substances under a fume hood is con-sidered to increase protection by a factor of 10 incomparison with an open bench.

� Manipulating substances in a glove box is consideredto increase protection by a factor of 100 in compari-son with fume hoods.

� Maximum recommended levels of activity in super-vised areas are nominally 3/10 of those in controlledzones (this approximately represents the respectiveexposure limits defined for these areas). However,the installation of glove boxes in supervised zonesis generally deprecated; the level of protectionafforded by a glove box is considered incompatiblewith the definition of such zones. It should be notedthat the factor of 3/10 may be masked by roundingthe data to one significant figure; rounding in anyother way may imply an unjustified degree of accu-racy in the models used for deriving the rec-ommended levels of activity.

Limitations

Area control

The values indicated in these sheets apply mainly tothe supervised and controlled areas of low and inter-mediate activity laboratories. Maximum recommendedactivities have further been limited to:

Group 1 (red) 0.5 GBqGroup 2 (orange) 5 GBqGroup 3 (yellow) 5 GBqGroup 4 (green) 5 GBqGroup 5 (blue) 50 GBq

These limits have been imposed because such labora-tories are not generally equipped with either a meansfor continuously monitoring the working environment,or for rejecting gases at a sufficient height to ensure thatno recycling takes place.

It should be noted that values calculated in com-pliance with these rules are guide values and are onlyindicative since other non-quantifiable parameters suchas working methods and human factors must also betaken into account. Nevertheless, it has been observedthat when practices are performed in accordance withthese simple rules, only very low levels of atmosphericcontamination and exposure occur.

The maximum recommended levels for elemental trit-

ium (3H), tritiated water and 37Ar have been deliberatelyreduced because of the difficulty of detecting the corre-sponding surface or atmospheric contamination. Thisfactor can be modified in particular situations,depending on survey and control frequencies. In thecase of iodine isotopes (123I, 125I, 131I), the existence ofparticularly unstable states (with the possibility of I2

release) has been considered. The corresponding limi-tations are consequently much more stringent.

Radiation control

Handled or stored activity values must comply withexternal exposure limits in situations where the externalradiation hazard may be more restrictive than theinternal radiation hazard. Attention is drawn to this factin the data sheets by the remark ‘Subject to externalexposure requirements which may be more restrictive’.Stored activities may exceed maximum recommendedactivities by a factor of 10, subject to any limits imposedby local or national regulations.

In particular cases, listed values may be varied inlocal regulations, based upon assessments by pro-fessional radiation protection specialists.

DESCRIPTION OF DATA SHEETS

General

Individual data sheets provide data in five main areas,namely basic radionuclide date, external exposure, con-tamination, internal exposure and maximum rec-ommended activities in laboratory situations. These aredivided into physical characteristics, (half life, specificactivity and decay scheme, A1 and A2 reference trans-port activities, exemption levels, data for externalexposure for five different geometries, external andinternal exposure limits, appropriate contaminationdetection probes and half-value and tenth-value shield-ing thicknesses. The maximum activities that can behandled on benches, under fume hoods or in gloveboxes are also given. The ‘highest dose organ’ is indi-cated, even where this is the whole body. As describedearlier the data sheets are coloured red, orange, yellow,green or blue, indicating, in descending order of risk,the level of risk posed by the particular radionuclide.

GLOSSARY

Activity

The number of nuclear transformations occurring ina given amount of radionuclide per unit time. The SIunit of activity is the reciprocal second, s�1, with thespecial name becquerel (Bq).

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Annual limit of intake

The intake by inhalation or ingestion of a given radio-nuclide in a year by the Reference Man which wouldresult in a committed dose equal to the relevant doselimit.

Bremsstrahlung

X rays produced by deceleration of � particles inmaterials.

Committed equivalent dose

Following an intake to the body of a radioactivematerial, the time integral of the equivalent dose rate iscalled the committed equivalent dose. If the integrationtime following the intake is not specified, it is impliedthat the value is 50 years for adults.

Equivalent dose

The equivalent dose, HT,R is defined as

HT,R = WR DT,R

REFERENCES

1. Delacroix, D., Guerre, J. P. and Leblanc, P. Radionuclides and Radioprotection. (French Atomic Energy Commission),(June 1994).

2. IAEA. International Basic Safety Standards for Protection against Ionising Radiation and for the Safety of Radiation Sources.Safety Series No. 115, (Vienna: IAEA) (1996).

3. EURATOM DIRECTIVE; L159, 13 May 1996.4. IAEA. Regulations for the Safe Transport of Radioactive Material. 1996 Edition. Safety Standards Series No.

ST1/Requirements (Vienna: IAEA) (1996).5. Lagoutine, F., Coursol, N. and Legrand, J. Tables des Radionucleides. CEA-ORIS, 4 vol, (1983–1987).6. ICRP. 1990 Recommendations of the International Commission on Radiological Protection. Publication 60, (Oxford:

Pergamon) (1990).7. ICRP. Data for Use in Radioprotection against External Radiation. Publication 51 (Oxford: Pergamon) (1987).8. Delacroix, D., Chazot, C. and Guerre, J. P. Calcul des Debits de Dose � et � en Fonction de la Geometrie de la Source.

CE. Report DSCE/SRI-A/93-362, (April 1993).9. Takaku, Y. and Kida, T. Radiation Dose to the Skin and Bone of Fingers from Handling Radioisotopes in a Syringe. Health

Phys 22, 295–297 (1971).10. Henson, P. W. Radiation Dose to the Skin in Contact with Unshielded Syringes Containing Radioactive Substances. Br. J.

Radiol. 46, 972–977 (1973).11. Perotin, J. P. and Goubert, J. Evaluation des Risques d’Irradiation des Mains dans un Laboratoire de Controle Radiopharma-

ceutique. D. CEA-Saclay. Report SPR/SRI (1979).12. Schmidt, W., Nowotny, R., Kletter, P. and Frisschauf, H. Radiation Exposure due to 99mTc and 131I Manipulated in Syringes.

J. Nucl. Med. 4, 389–391 (1979).13. Moreau, A. La Radioprotection dans les Laboratoires de Faible et Moyenne Radioactivite. CEA-Saclay. Report SPR/SRI,

1981.14. Kocher, D. C. and Eckerman, K. F. Electron Dose Rate Conversion Factors for External Exposure of the Skin from Uniformly

Deposited Activity on the Body Surface. Health Phys. 53(2) 135–141 (1987).15. Piechowski, J., Menoux, B., Chaptinel, Y. and Durand, F. Dosimetrie et Therapeutique des Contaminations Cutanes. CEA

Report 5441 (1988).16. RadDecay Version 1.13. Grove Enginering (Rockville, Maryland, USA) (1996).17. Delacroix, D., Guerre, J. P. and Leblanc, P. Determination des limites de contamination surfacique pour les principaux

radionucleides. CEA-Saclay. Report DSCE/SRI, January 1992.

where DT,R is the absorbed dose delivered by radiationtype R averaged over a tissue or organ T and WR is theradiation weighting factor for type R radiation. Whenthe radiation field is composed of different radiationtypes, the equivalent dose is HT = �RWR DT,R.

Effective dose

The quantity E, defined as a summation of the tissueequivalent doses, each multiplied by the appropriatetissue weighting factor (E = �TWT HT where HT is theequivalent dose in tissue T and WT is the tissue weight-ing factor for tissue T). From the definition of equivalentdose it follows that E = �TWT �RWR DT,R.

Radioactive half-life

Time required for a radioactive substance to lose 50%of its activity by radioactive decay.

Half-value and tenth-value thickness (or layer)

The thickness of material necessary to reduce theintensity of X or gamma radiation by a factor of 2 and10, respectively.

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18. ICRP. Dose Coefficients for Intakes of Radionuclides by Workers. Publication 68 (Oxford: Pergamon) (1994).19. ICRP. Individual Monitoring for Internal Exposure of Workers, Replacement of ICRP Publication 54. Publication 78 (Oxford:

Pergamon) (1997).20. ICRP. Limits for Intakes of Radionuclides by Workers. Publication 30 (Parts 1–4 and Supplements) (Oxford: Pergamon)

1979–1988).21. ICRP. Individual Monitoring for Intakes of Radionuclides by Workers: Design and Interpretation. Publication 78 (Oxford:

Pergamon) (1998).

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TABLE OF RADIONUCLIDES LISTED

Nuclide Symbol Page

Tritium (3H1) 21Beryllium - 7 (7Be4) 22Carbon - 11 (11C6) 23Carbon - 14 (14C6) 24Nitrogen - 13 (13N7) 25Oxygen - 15 (15O8) 26Fluorine - 18 (18F9) 27Sodium - 22 (22Na11) 28Sodium - 24 (24Na11) 29Aluminium - 26 (26Al13) 30Silicon - 31 (31Si14) 31Phosphorus - 32 (32P15) 32Phosphorus - 33 (33P15) 33Sulphur - 35 (35S16) 34Chlorine - 36 (36Cl17) 35Argon - 37 (37Ar18) 36Argon - 41 (41Ar18) 37Potassium - 40 (40K19) 38Potassium - 42 (42K19) 39Potassium - 43 (43K19) 40Calcium - 45 (45Ca20) 41Calcium - 47 / Scandium - 47 (47Ca20 / 47Sc21) 42Scandium - 46 (46Sc21) 43Scandium - 47 (47Sc21) 44Vanadium - 48 (48V23) 45Chromium - 51 (51Cr24) 46Manganese - 52m (52mMn25) 47Manganese - 52 (52Mn25) 48Manganese - 54 (54Mn25) 49Manganese - 56 (56Mn25) 50Iron - 52 (52Fe26) 51Iron - 55 (55Fe26) 52Iron - 59 (59Fe26) 53Cobalt - 56 (56Co27) 54Cobalt - 57 (57Co27) 55Cobalt - 58 (58Co27) 56Cobalt - 60 (60Co27) 57Nickel - 63 (63Ni28) 58Nickel - 65 (65Ni28) 59Copper - 64 (64Cu29) 60Copper - 67 (67Cu29) 61Zinc - 65 (65Zn30) 62Gallium - 66 (66Ga31) 63Gallium - 67 (67Ga31) 64Gallium - 68 (68Ga31) 65

Arsenic - 73 (73As33) 66Arsenic - 74 (74As33) 67Arsenic - 76 (76As33) 68Arsenic - 77 (77As33) 69Selenium - 75 (75Se34) 70Bromine - 77 (77Br35) 71Bromine - 82 (82Br35) 72Krypton - 81 (81Kr36) 73Krypton - 83m (83mKr36) 74Krypton - 85 (85Kr36) 75Krypton - 85m (85mKr36) 76Rubidium - 86 (86Rb37) 77Strontium - 85 (85Sr38) 78Strontium - 89 (89Sr38) 79Strontium - 90 / Yttrium - 90 (90Sr38 / 90Y39) 80Yttrium - 90 (90Y39) 81Yttrium - 91 (91Y39) 82Zirconium - 95 / Niobium - 95 (95Zr40 / 95Nb41) 83Molybdenum - 99 / Technetium - 99m (99Mo42 / 99mTc43) 84Technetium - 99m (99mTc43) 85Technetium - 99 (99Tc43) 86Ruthenium - 103 / Rhodium - 103m (103Ru44 / 103mRh45) 87Ruthenium - 106 / Rhodium - 106 (106Ru44 / 106Rh45) 88Palladium - 103 / Rhodium - 103m (103Pd46 / 103mRh45) 89Silver - 110m (110mAg47) 90Silver - 111 (111Ag47) 91Cadmium - 109 (109Cd48) 92Indium - 111 (111In49) 93Indium - 113m (113mIn49) 94Indium - 115m (115mIn49) 95Tin - 125 (125Sn50) 96Antimony - 122 (122Sb51) 97Antimony - 124 (124Sb51) 98Antimony - 125 / Tellurium - 125m (125Sb51 / 125mTe52) 99Antimony - 126 (126Sb51) 100Tellurium - 123m (123mTe52) 101Tellurium - 125m (125mTe52) 102Tellurium - 132 / Iodine - 132 (132Te52/132I53) 103Iodine - 123 (123I53) 104Iodine - 124 (124I53) 105Iodine - 125 (125I53) 106Iodine - 129 (129I53) 107Iodine - 131 (131I53) 108Iodine - 132 (132I53) 109Iodine - 133 (133I53) 110Xenon - 133 (133Xe54) 111Caesium - 131 (131Cs55) 112

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Caesium - 134 (134Cs55) 113Caesium - 137 / Barium - 137m (137Cs55 / 137mBa56) 114Barium - 133 (133Ba56) 115Barium - 140 / Lanthanum - 140 (140Ba56 / 140La57) 116Lanthanum - 140 (140La57) 117Cerium - 139 (139Ce58) 118Cerium - 141 (141Ce58) 119Cerium - 143 (143Ce58) 120Praseodymium - 143 (143Pr59) 121Praseodymium - 144 (144Pr59) 122Promethium - 147 (147Pm61) 123Samarium - 153 (153Sm62) 124Europium - 152 (152Eu63) 125Europium - 154 (154Eu63) 126Europium - 155 (155Eu63) 127Europium - 156 (156Eu63) 128Erbium - 169 (169Er68) 129Thulium - 170 (170Tm69) 130Thulium - 171 (171Tm69) 131Ytterbium - 169 (169Yb70) 132Rhenium - 186 (186Re75) 133Rhenium - 188 (188Re75) 134Iridium - 192 (192Ir77) 135Gold - 198 (198Au79) 136Mercury - 197 (197Hg80) 137Mercury - 203 (203Hg80) 138

Thallium - 201 (201Tl81) 139Thallium - 204 (204Tl81) 140Lead - 210 (210Pb82) 141Lead - 214 (214Pb82) 142Bismuth - 207 (207Bi83) 143Bismuth - 210 (210Bi83) 144Bismuth - 214 (214Bi83) 145Polonium - 210 (210Po84) 146Radium - 226 (226Ra88) 147Thorium - 231 (231Th90) 148Thorium - 234 (234Th90) 149Protactinium - 234 (234Pa91) 150Protactinium - 234m (234mPa91) 151Uranium - 233 (233U92) 152Uranium - 234 (234U92) 153Uranium - 235 (235U92) 154Uranium - 238 (238U92) 155Neptunium - 239 (239Np93) 156Plutonium - 238 (238Pu94) 157Plutonium - 239 (239Pu94) 158Plutonium - 240 (240Pu94) 159Plutonium - 241 (241Pu94) 160Americium - 241 (241Am95) 161Americium - 243 (243Am95) 162Curium - 244 (244Cm96) 163Californium - 252 (252Cf98) 164

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Radiation Protection DosimetryVol. 98, No. 1, pp. 165–168 (2002)Nuclear Technology Publishing

APPENDIX – HEAVY ELEMENT DECAYPb-210 decay Pb-210 to Po-210 (see pages 13 and 141)

Nuclide Half-life 0.1 day 1 day 10 days 100 days 1 year 10 years 100 years

Pb-210 2.22E+1 yr 1.00E+00 1.00E+00 1.00E+00 1.00E+00 1.00E+00 1.00E+00 1.00E+00Bi-210 5.01E+0 d 1.37E-02 1.29E-01 7.48E-01 1.00E+00 1.00E+00 1.00E+00 1.00E+00Po-210 1.38E+2 d 3.44E-06 3.30E-04 2.52E-02 3.69E-01 8.43E-01 1.02E+00 1.02E+00

Ra-226 decay Ra-226 to Po-210 (see pages 13 and 147)

Nuclide Half-life 0.1 day 1 day 10 days 100 days 1 year 10 years 100 years

Ra-226 1.60E+3 yr 1.00E+00 1.00E+00 1.00E+00 1.00E+00 1.00E+00 1.00E+00 1.00E+00Rn-222 3.82E+0 d 1.80E-03 1.65E-01 8.40E-01 1.00E+00 1.00E+00 1.00E+00 1.00E+00Po-218 3.05E+0 min 1.74E-02 1.65E-01 8.40E-01 1.00E+00 1.00E+00 1.00E+00 1.00E+00Pb-214 2.68E+1 min 1.27E-02 1.61E-01 8.30E-01 1.00E+00 1.00E+00 1.00E+00 1.00E+00Bi-214 1.99E+1 min 9.46E-03 1.58E-01 8.30E-01 1.00E+00 1.00E+00 1.00E+00 1.00E+00Po-214 1.64E-4 sec 9.46E-03 1.58E-01 8.30E-01 1.00E+00 1.00E+00 1.00E+00 1.00E+00Pb-210 2.22E+1 yr 2.70E-08 6.60E-06 4.55E-04 8.01E-03 3.01E-02 2.66E-01 9.73E-01Bi-210 5.01E+0 d 8.90E-11 2.84E-07 1.69E-04 7.40E-03 2.95E-02 2.66E-01 9.73E-01Po-210 1.38E+2 d 8.40E-15 3.44E-10 2.40E-06 1.41E-03 1.57E-02 2.53E-01 9.73E-01

U-235 decay U-235 to Po-211 (see pages 13 and 154)

Nuclide Half-life 0.1 day 1 day 10 days 100 days 1 year 10 years 100 years

U-235 7.04E+8 yr 1.00E+00 1.00E+00 1.00E+00 1.00E+00 1.00E+00 1.00E+00 1.00E+00Th-231 2.55E+1 hr 6.31E-02 4.78E-01 1.00E+00 1.00E+00 1.00E+00 1.00E+00 1.00E+00Pa-231 3.28E+4 yr 1.84E-10 1.53E-08 4.90E-07 5.70E-06 2.10E-05 2.11E-04 2.11E-03Ac-227 2.18E+1 yr �E-10 �E-10 1.90E-10 2.44E-08 3.30E-07 3.03E-05 1.48E-03Fr-223 2.18E+1 min �E-10 �E-10 �E-10 3.36E-10 4.55E-09 4.18E-07 2.04E-05Ra-223 1.14E+1 d �E-10 �E-10 �E-10 2.42E-10 4.17E-09 4.15E-07 2.04E-05Rn-219 3.96E+0 sec �E-10 �E-10 �E-10 2.42E-10 4.17E-09 4.15E-07 2.04E-05Po-215 1.78E-3 sec �E-10 �E-10 �E-10 2.42E-10 4.17E-09 4.15E-07 2.04E-05Pb-211 3.61E+1 min �E-10 �E-10 �E-10 2.42E-10 4.17E-09 4.15E-07 2.04E-05Bi-211 2.13E+0 min �E-10 �E-10 �E-10 2.42E-10 4.17E-09 4.15E-07 2.04E-05Tl-207 4.77E+0 min �E-10 �E-10 �E-10 2.42E-10 4.15E-09 4.14E-07 2.03E-08

Po-211 5.16E-1 sec �E-10 �E-10 �E-10 �E-10 �E-10 1.13E-09 5.6E-08

Th-227 1.87E+1 d �E-10 �E-10 �E-10 1.44E-08 2.81E-07 2.95E-05 1.45E-03Ra-223 1.14E+1 d �E-10 �E-10 �E-10 9.76E-09 2.55E-07 2.93E-05 1.45E-03Rn-219 3.96E+0 sec �E-10 �E-10 �E-10 9.76E-09 2.55E-07 2.93E-05 1.45E-03Po-215 1.78E-3 sec �E-10 �E-10 �E-10 9.76E-09 2.55E-07 2.93E-05 1.45E-03Pb-211 3.61E+1 min �E-10 �E-10 �E-10 9.75E-09 2.55E-07 2.93E-05 1.45E-03Bi-211 2.13E+0 min �E-10 �E-10 �E-10 9.75E-09 2.55E-07 2.93E-05 1.45E-03Tl-207 4.77E+0 min �E-10 �E-10 �E-10 9.72E-09 2.55E-07 2.92E-05 1.45E-03

Po-211 5.16E-1 sec �E-10 �E-10 �E-10 �E-10 6.97E-10 7.99E-08 3.97E-06

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D. DELACROIX, J. P. GUERRE, P. LEBLANC and C. HICKMAN

166

U-238 decay U-238 to Po-210 (see pages 13 and 155)

Nuclide Half-life 0.1 day 1 day 10 days 100 days 1 year 10 years 100 years

U-238 4.47E+9 yr 1.00E+00 1.00E+00 1.00E+00 1.00E+00 1.00E+00 1.00E+00 1.00E+00Th-234 2.41E+1 d 2.87E-03 2.83E-02 2.50E-01 9.44E-01 1.00E+00 1.00E+00 1.00E+00Pa-234m 1.17E+0 min 2.83E-03 2.83E-02 2.50E-01 9.44E-01 1.00E+00 1.00E+00 1.00E+00Pa-234 6.70E+0 hr 5.14E-07 2.87E-05 3.86E-04 1.50E-03 1.60E-03 1.60E-03 1.60E-03U-234 2.44E+5 yr �E-10 �E-10 �E-10 8.30E-10 4.10E-09 4.49E-08 4.3E-07Th-230 7.7E+4 yr �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 2.0E-10Ra-226 1.60E+3 yr �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10Rn-222 3.82E+0 d �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10Po-218 3.05E+0 min �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10Pb-214 2.68E+1 min �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10Bi-214 1.99E+1 min �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10Po-214 1.64E-4 sec �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10Pb-210 2.23E+1 yr �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10Bi-210 5.0E+0 d �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10Po-210 1.38E+2 d �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10

U-234 2.44E+5 yr �E-10 1.01E-08 1.01E-18 5.21E-07 2.56E-06 2.80E-05 2.83E-04Th-230 7.7E+4 yr �E-10 �E-10 �E-10 �E-10 �E-10 1.25E-09 1.27E-07Ra-226 1.60E+3 yr �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 2.27E-10Rn-222 3.82E+0 d �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 1.81E-09Po-218 3.05E+0 min �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 1.81E-09Pb-214 2.68E+1 min �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 1.81E-09Bi-214 1.99E+1 min �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 1.81E-09Po-214 1.64E-4 sec �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 1.81E-09Pb-210 2.23E+1 yr �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 8.44E-10Bi-210 5.01E+0 d �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 8.44E-10Po-210 1.38E+2 d �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 8.28E-10

Pu-241 decay Pu-241 to Pb-209 (see pages 13 and 160)

Nuclide Half-life 0.1 day 1 day 10 days 100 days 1 year 10 years 100 years

Pu-241 1.44E+1 yr 1.00E+00 1.00E+00 1.00E+00 1.00E+00 1.00E+00 1.00E+00 1.00E+00Am-241 4.32E+2 yr 4.39E-07 4.39E-06 4.39E-05 4.48E-04 1.64E-03 2.04E-02 3.58E+00Np-237 2.14E+6 yr �E-10 �E-10 �E-10 �E-10 2.69E-10 3.58E-08 9.85E-05Pa-233 2.70E+1 d �E-10 �E-10 �E-10 �E-10 2.17E-10 3.51E-08 9.85E-05U-233 1.59E+5 yr �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 1.85E-08Th-229 7.34E+3 yr �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10Ra-225 1.48E+1 d �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10Ac-225 1.00E+1 d �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10Fr-221 4.8E+0 min �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10At-217 3.23E-2 sec �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10Bi-213 4.56E+1 min �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10Tl-209 2.20 E+0 min �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10Pb-209 3.24E+0 hr �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10

Po-213 4.20E-6 sec �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10Pb-209 3.25E+0 hr �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10

U-237 6.75E+6 2.50E-07 2.39E-06 1.57E-05 2.48E-05 2.46E-05 2.45E-05 2.45E-05Np-237 2.14E+6 yr �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 2.00E-08Pa-233 2.70E+1 d �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 2.00E-08U-233 1.59E+5 yr �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10Th-229 7.34E+3 yr �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10Ra-225 1.48E+1 d �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10

Continued

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Pu-241 decay (Continued)

Ac-225 1.00E+1 d �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10Fr-221 4.8E+0 min �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10At-217 3.23E-2 sec �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10Bi-213 4.56E+1 min �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10Tl-209 2.20 E+0 min �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10Pb-209 3.24E+0 hr �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10

Po-213 4.20E-6 sec �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10Pb-209 3.25E+0 hr �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10

Am-243 decay Am-243 to Po-211 (see pages 13 and 162)

Nuclide Half-life 0.1 day 1 day 10 days 100 days 1 year 10 years 100 years

Am-243 7.38E+3 yr 1.00E+00 1.00E+00 1.00E+00 1.00E+00 1.00E+00 1.00E+00 1.00E+00Np-239 2.35E+0 d 2.90E-02 2.55E-01 9.47E-01 1.00E+00 1.00E+00 1.00E+00 1.00E+00Pu-239 2.41E+4 yr 1.14E-10 1.05E-08 5.33E-07 7.60E-06 2.84E-05 2.87E-04 2.88E-03U-235 7.04E+8 yr �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 1.42E-10Th-231 2.55E+1 hr �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 1.42E-10Pa-231 3.28E+4 yr �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10Ac-227 2.18E+1 yr �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10Fr-223 2.18E+1 min �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10Ra-223 1.14E+1 d �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10Rn-219 3.96E+0 sec �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10Po-215 1.78E-3 sec �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10Pb-211 3.61E+1 min �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10Bi-211 2.13E+0 min �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10Tl-207 4.77E+0 min �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10

Po-211 5.16E-1 sec �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10

Th-227 1.87E+1 d �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10Ra-223 1.14E+1 d �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10Rn-219 3.96E+0 sec �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10Po-215 1.78E-3 sec �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10Pb-211 3.61E+1 min �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10Bi-211 2.13E+0 min �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10Tl-207 4.77E+0 min �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10

Po-211 5.16E-1 sec �E-10 �E-10 �E-10 �E-10 �E-10 �E-10 �E-10

Note:

The above tables for U-235, U-238, Pu-241 and Am-243 should be read in conjunction with the corresponding decay chainsshown overleaf. These chains clearly show the branching in the decay chains, clarifying the lines apparently repeated in the tables.

REFERENCE

1. RadDecay for Windows Version 1.13. Grove Engineering (Maryland) 1996.

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