Radionuclide and Radiation Protection Data Handbook

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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

RADIONUCLIDE AND RADIATION PROTECTION DATA HANDBOOK 2nd Edition (2002)

All rights reserved. No part of this book may be reproduced, stored in a retrieval system or transmitted in any form or by any means, electronic, electrostatic, magnetic, mechanical, photocopying, 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 6 COPYRIGHT  2002 Nuclear Technology Publishing

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

RADIONUCLIDE AND RADIATION PROTECTION DATA HANDBOOK 2nd Edition (2002)

Contents Contents Preface

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INTRODUCTION SCOPE

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CLASSIFICATION OF RADIONUCLIDES

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PHYSICAL CHARACTERISTICS OF RADIONUCLIDES . . . . . . . . . . . . . . . . . . . . . 10 RADIOLOGICAL CONTROL . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 External exposure risks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 Internal exposure risks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 Dose limits for workers . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 EXTERNAL EXPOSURE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 Assumptions . . . . . . . . . . . . . . . . Distant (point source) external exposure. . . . Exposures to a uniformly contaminated surface External contact exposure (receptacles) . . . .

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SHIELDING OF BETA AND GAMMA EMITTERS . . . . . . . . . . . . . . . . . . . . . . . 12 CONTAMINATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 Contact exposure due to external skin contamination. . . . Decay products . . . . . . . . . . . . . . . . . . . . . Decay of heavy elements . . . . . . . . . . . . . . . . Derived surface contamination limits (DSCL) . . . . . . . Determination of removable and fixed contamination values Contamination control . . . . . . . . . . . . . . . . . . Detection probes . . . . . . . . . . . . . . . . . . . . 5

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CONTENTS (continued) INTERNAL EXPOSURE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 Committed effective dose per unit Annual Limits on Intake (ALI) . . The ‘highest dose organ’. . . . . General . . . . . . . . . . . . .

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MAXIMUM RECOMMENDED ACTIVITIES. . . . . . . . . . . . . . . . . . . . . . . . . . . 15 Calculation models . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 Dependence of maximum recommended activities on equipment and working areas . . . . . . . . 15 Limitations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 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 PROTECTION ABSTRACTS, 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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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 published in 1998. That handbook was in turn based on an earlier French laboratory guidebook A Guide on Radionuclides and Radioprotection by D. Delacroix, J.P. Guerre, and P. Leblanc, published in 1994 (1). The earlier publication was very much oriented towards guidance on the handling of radionuclides used in medicine. The present handbook is much more broadly based in its outlook and application and incorporates updated information to take account of the most recent ICRP and IAEA recommendations (1–4). The radionuclides listed in the earlier publication were supplemented by commonly used radionuclides in the nuclear industry and in other areas. Moreover, the main purpose of this handbook (and its 1998 predecessor) is the provision of data sheets and the models or sources used 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 specialists as well as all others who use or transport radionuclides. Its publication should satisfy a major need for all health physics departments and is intended to assist in informing and training personnel in radiation protection. It consists of an explanatory text followed by specific radiation data sheets for selected radionuclides. Because of the disparities in the approaches adopted by different countries, it is essential that users also refer to the relevant national regulations with which they must comply. The present handbook includes 36 additional radionuclides than the previous edition, giving a total of 144 nuclides. Additional data are included covering dose rates above uniformly contaminated surfaces (infinite plane source). Some corrections to data in the previous edition have been incorporated. The list of nuclides is not exhaustive but those included have been selected on the basis of being most commonly used, taking into account the requirements of users in industry, medicine and research. Finally, account is taken of the need to consider decay chain products for several heavy elements. To this end, an appendix is provided giving the decay products and their activities as a function of time (age), together with decay charts for four of the more complex chains. Directeur Central de la Se´curite´ Commissariat a` l’Energie Atomique, France

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RADIONUCLIDE AND RADIATION PROTECTION DATA HANDBOOK 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 provide users of radionuclides in medicine, research and industry with consolidated and appropriate information and data to handle and transport radioactive substances safely. It is mainly intended for users in low and intermediate activity laboratories. Individual data sheets are provided for a wide range of commonly used radionuclides (144 in total). These radionuclides are classified into five different groups as a function of risk level, represented by colours red, orange, yellow, green and blue, in descending order of risk.

INTRODUCTION

in the form of data sheets in an easy to use format for 144 of the most commonly used radionuclides in medicine, research and industry. It is designed as a ready reference source of information, gathering together in one place, in the form of individual data sheets, the most up to date isotopic and radiation protection data. The information includes:

In recent years the need for a compilation of essential data for commonly used radionuclides, for both practice and training, has become increasingly apparent. This handbook contains individual data sheets for a range of radionuclides and is mainly intended for users of unsealed radioactive sources in low and intermediate activity laboratories, industrial applications and transport. Individual data sheets corresponding to the different radionuclides give the physical characteristics, reference transport activities, exemption levels, contamination derived limits and appropriate detection probes, dose rates from skin contamination, half and tenth value shielding thicknesses, external exposure data, dose coefficients for ingestion and inhalation, derived 20 mSv ALI values, and the highest dose organ. The maximum activities recommended to be handled on benches, under fume hoods and in glove boxes are also given. However, this information is most appropriate in laboratories handling low and intermediate levels of radioactive materials. In order to facilitate the user’s understanding, pictorial illustrations have been employed and data is presented in tabular form. It should be noted that the data provided, although drawn from appropriate recent international recommendations, should in no way be taken to supersede existing local or national regulations. The data may, of course, be used to review existing local regulations which may have become out of date. The text preceding the data sheets provides the background, derivation and substantiation for the data given, enabling users to satisfy any quality assurance arrangements that may be in place.

䉬 䉬 䉬 䉬 䉬 䉬 䉬 䉬

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, 䉬 20 mSv ALI values and the highest organ dose, and, 䉬 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 maximum recommended activities are only appropriate to low and intermediate activity laboratories. These laboratories are frequently limited to two to three orders of magnitude higher than the exemption limits by local or national regulations. CLASSIFICATION OF RADIONUCLIDES Radionuclides have been divided in this handbook into five ‘risk groups’. The classification used is based on the BSS ‘quantity’ exemption limit values. Exemption limit values have been chosen for this purpose because they reflect both internal and external exposure risks. Each group is associated with a colour (red, orange, yellow, green and blue successively represent decreasing levels of risk). The data sheets have been appropriately coloured providing the reader with a rapid visual means of appreciating the risk associated with

SCOPE This handbook provides the most up to date internal and external dosimetry information and control criteria 9

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

The nature of the main radiation (5)

particular radionuclides. The five risk groups have been defined as follows: Group Group Group Group Group

1: 2: 3: 4: 5:

exemption exemption exemption exemption exemption

limit limit limit limit limit

␣, ␤, e⫺, X, ␥ radiation and neutron emissions are given, together with the corresponding energies expressed in keV, and corresponding percentage emissions. A maximum of the three most characteristic emissions is given for each type of radiation, the criterion adopted being the relevance of the emissions from the radiation protection standpoint. However, in order to remedy the restrictive effects of this approach (some radionuclides exhibit multiple emissions, e.g. 140La and 152 Eu), the relative percentage emissions not taken into consideration are nevertheless given.

ⱕ 104 Bq (red) = 105 Bq (orange) = 106 Bq (yellow) = 107 Bq (green) ⱖ 108 Bq (blue)

In the rare cases for which exemption levels have not been defined (e.g. 11C, a nuclide sometimes used for medical imaging) a group has been assigned by analogy with other radionuclides of the same type. Attention is drawn to the fact that radioactive substances are also chemical substances, the handling of which may involve hazards of non-radiological origin. Moreover, radioactive substances are often intimately mixed with chemical products having an aggressive action on the human body (toxic, mutagenic and carcinogenic effects). Chemical hazards should therefore never be neglected.

RADIOLOGICAL CONTROL External exposure risks The following modes of exposure are considered. Contact

PHYSICAL CHARACTERISTICS OF RADIONUCLIDES

When handling radioactive materials, the operator may touch the receptacle (e.g. beaker, pipette, syringe) containing the radioactive substance or materials contaminated with this substance. The radioactive substance can also become deposited on the skin.

The following characteristics are considered in these sheets. Half-lives expressed in the most appropriate units (hours, days, years).

Distant source

Specific activities expressed as Bq.g⫺1.

All parts of the operator’s body generally remain distant from the radioactive substances being handled, even though certain parts of the body such as the hands and forearms occasionally approach these substances during handling operations with or without manipulators.

Exemption levels Exempt activity concentrations (Bq.g⫺1) and exempt activities (Bq) are defined in the IAEA Basic Safety Standards (2) and in the L159 Euratom Directive (3). These limits can be used by regulatory authorities to define criteria for exemption against formal registration of the premises and may also be used to develop clearance levels for materials leaving the premises. The exemption levels also apply in the transport regulations. The correct use of these limits by users is of utmost importance. Users should refer to the above mentioned references.

Immersion in a cloud This type of relatively infrequent hazard is encountered when handling gaseous radioactive sources. The corresponding risks must be taken into consideration, for example, in laboratories using cyclotrons to manufacture short half-life radionuclides (e.g. 18F, 11C, 123I). Internal exposure risks

A1 and A2 reference transport activities

The risks arise from three different possible paths into the human body:

A1 and A2 are the maximum activities (expressed in TBq) that can be transported by A type packages. A1 corresponds to radioactive matter in ‘special form’ and A2 to the other cases. In the data sheets, the values correspond to values in the 1996 edition of The IAEA publication Regulations for the Safe Transport of Radioactive Materials (4), with which the international modal transport authorities (RID, ADR, IMO, ICAO, etc.) are harmonised. Users seeking further details are referred to the referenced texts.

䉬 Inhalation after the dispersion of gases, vapours or aerosols in the environment; 䉬 Ingestion by contaminated hands or objects through contact with the mouth; 䉬 Transfer through the skin with or without an associated wound following contamination of parts of the body (which have not been sufficiently rapidly decontaminated) or following simple contact with highly penetrating radionuclides (e.g. tritium in 10

RADIONUCLIDE AND RADIATION PROTECTION DATA HANDBOOK 2nd Edition (2002)

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

EXTERNAL EXPOSURE Assumptions

Dose limits for workers

X ray, ␥, ␤, e⫺ and neutron emissions lead to external exposures depending on the nature and energies of the corresponding radiation as well as the packaging and geometry of the source. Depending on the particular case being studied, superficial exposures (or skin exposures) defined at a depth of 7 mg.cm⫺2 or deep body exposures (7) are taken into consideration. Various handling postures and situations commonly encountered in laboratories are considered. Calculations have been performed for unique values of source or solution activity; in this way a direct comparison between the values obtained under different circumstances can be made. A nominal activity of 1 MBq has been adopted for external exposures, exposures resulting from contacts with receptacles and exposures to distant sources. A nominal value of 1 kBq has been adopted for the case of skin contamination due to droplets and 1 kBq.cm⫺2 for uniform skin contamination. Exposures due to an infinite and uniformly contaminated surface at 1 MBq.cm−2 are also considered. Data are provided for distances of 10 cm and 1 m from the contaminated surface for beta and photon radiation for the skin and deep tissue. The quantity adopted for all geometries for tabulated 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 unsealed sources may not actually represent what is done in practice (e.g. use of a syringe or beaker). However, the user, by analogy, can transform the values indicated into those corresponding to other receptacles of similar dimensions, on the basis that the situations shown are reasonably typical. The following codes have been used for the calculations:

(2,6)

The occupational exposure of any worker should not exceed the following limits, but may also be subject to more restrictive limits in appropriate local or national regulations. 䉬 an effective dose of 100 mSv over a period of five consecutive years (i.e. 20 mSv per year) averaged over this period; 䉬 an effective dose of 50 mSv in any single year; 䉬 an equivalent dose to the lens of the eye of 150 mSv in a year; 䉬 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 for application of the dose limits on effective dose shall be determined by one of the following methods: (a) by comparing the total effective dose with the relevant limit, where the total effective dose ET is calculated 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 from exposure to penetrating radiation during the year. Note that e(g)j,ingestion and e(g)j,inhalation are the committed effective dose per unit intake by ingestion and inhalation for radionuclide j by the group of age g, and Ij,ingestion and Ij,inhalation are the intakes via ingestion or inhalation for radionuclide j during the same period.

䉬 Microshield Version 4.10 has been used for ␥ and X ray calculations (Grove Engineering, 15215 Shady Grove Road, Rockville, USA, 1996) 䉬 Varskin mod 2 for ␤ radiation and Varskin mod 2 modified for monoenergetic electrons (J.S. Durham, Pacific Northwest Laboratory, PO Box 999, Richland, Washington 99352, USA).

(b) by satisfying the following conditions: Hp(d)/DL + ⌺j(Ij,ingestion/Ij,ingestion,L) + ⌺j(Ij,inhalation/Ij,inhalation,L) ⱕ 1

Distant (point source) external exposure

where DL is the relevant dose limit on effective dose, and Ij,ingestion,L and Ij,inhalation,L. are the annual limits on intake (ALI) via ingestion or inhalation of radionuclide j.

The cases of a point source at a distance of 30 cm (average length of forearm) in air and a ‘penicillin’ type vial at a distance of 1 m in air (8) are considered successively. In the latter case, the source is represented by a cylinder with a density of 1 g.cm⫺3, 2.3 cm in diameter and 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 or whole-body dose equivalent due to X and ␥ ray compo-

(c) by any other approved method. Note that the symbols used in the quote are consistent with those used in this handbook but not the original IAEA document. The above quantities are briefly explained in Schedule II of Reference 2. 11

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

nents is distinguished from the superficial or skin dose equivalent due to ␤ and e⫺ components. This is done in order to draw the operator’s attention to the fact that the superficial dose equivalent can be considerably reduced by interposing a screen with a thickness equivalent 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 and the bench or table).

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

Exposures to a uniformly contaminated surface

The range of betas and electrons and the attenuation of X and ␥ rays in shielding materials depends on the energies of the incident radiation and the nature of the shielding material. Total absorption thicknesses have been determined for ␤ and e⫺ radiation in glass and plastic (13); these materials are those most commonly used for this purpose. In the case of ␥ and X rays, the first half-value and tenth-value thicknesses of lead and steel (attenuating the incident radiation by a factor of 2 and 10 respectively) are given. Attention is drawn to the fact that some radionuclides simultaneously emit significant low and high energy components for which first half-value and tenth-value thicknesses may be considerably less than second halfvalue and tenth-value thicknesses. 123I is an example but the data provided here do not cover this situation.

SHIELDING OF BETA AND GAMMA EMITTERS

Exposures due to a uniformly contaminated source (floors) are considered. The contributions due to beta radiation and photons are considered separately for distances of 10 cm and 1 m from an infinitely and uniformly contaminated surface. This separation enables users to make appropriate decisions as to what measurement equipment should be used. Pure beta emitters generally cannot be detected with gamma detectors. Dose rates at 10 cm and 1 m are often similar because the increased distance is compensated by the increased solid angle. Data is quoted for the dose equivalent at depths representing the skin and deep tissue respectively. External contact exposure (receptacles) The cases of a 50 cm3 beaker containing 20 cm3 of solution and a 5 cm3 syringe containing 2.5 cm3 of solution are considered. The beaker is represented by a cylinder with a density of 1 g.cm⫺3, 4 cm in diameter, 1.6 cm high enclosed in a 2 mm thick glass (density 2.7 g.cm⫺3) envelope. The syringe is represented by a cylinder with a density of 1 g.cm⫺3, 1.2 cm in diameter, 2.2 cm high enclosed in a 1 mm thick plastic (density 1 g.cm⫺3) envelope. In both cases dose equivalents have been calculated under 7 mg.cm⫺2 at the level of the solution (an arrow indicates the position taken into consideration) (8). The position indicated corresponds to the maximum dose equivalent to which an operator can be exposed when manipulating these receptacles. Attention is drawn to the fact that these drawings, for pedagogical reasons, show the operator’s hands and fingers as far away from the source as possible. The dose equivalents depicted take both ␤ and ␥ components into consideration. It should, however, be noted that in the case of a beaker (2 mm thick glass), ␤ components (other than those emitted by radionuclides such as 42K, 90Y and 144Pr) can be neglected. In the case of a syringe (1 mm thick plastic), the ␤ contribution can be highly significant (e.g. 24 mSv.h⫺1 for 1 MBq 32P). Note that good agreement is observed between the values quoted here and those found in the literature (9–12). ␤ ray absorption in shielding gives rise to bremsstrah-

CONTAMINATION Contact exposure due to external skin contamination Two situations have been considered to illustrate body contamination. Firstly, exposure due to an extensive uniformly spread out contamination of the skin (9,14,15) is considered and secondly, the projection of a 0.05 cm3 droplet of a radioactive substance. The droplet is represented by a cylinder with a density of 1 g.cm⫺3, a cross sectional area of 1 cm2 and a height of 0.5 mm (8). Dose equivalent values have been calculated for an average 70 ␮m basal layer depth. At this depth the main component of the dose resulting from superficial contamination of the skin is due to ␤ rays and electrons from the radionuclide; the ␥ contribution to the dose is generally just a few per cent. Comparison between the values due to a uniform (infinitely thin) deposit and those due to a droplet highlights the effects of ␤ attenuation in tissue. It is assumed here that penetration of the contamination in the skin can be neglected. Decay products Decaying radionuclides are generally treated as follows: 12

RADIONUCLIDE AND RADIATION PROTECTION DATA HANDBOOK 2nd Edition (2002)

䉬 radionuclides (parent–progeny) couples are assumed to be in equilibrium; 䉬 calculations are performed for a nominal activity of the parent nuclide (e.g. 1 kBq.cm⫺2 for uniform skin contamination); 䉬 parent–progeny radionuclide decay schemes are normally covered in a single sheet, but when physical separation of the constituents is possible (e.g. by elution), progeny products are treated in individual sheets. The 99Mo–99mTc couple is an example of this. The reader will find both 99Mo–99mTc and 99mTc sheets.

䉬 transfer of surface contamination to the skin leading to external extremity exposures 䉬 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 the contaminated zone leading to annual limits due to each of the four modes of exposure (considered separately). Limits of 20 mSv.y⫺1 and 500 mSv.y⫺1 are considered for whole body external exposure and skin exposure, respectively. The most restrictive ALIinhalation and ALIingestion values calculated for 20 mSv.y⫺1 are used in the following derivations of surface activity for limiting internal contamination.

Decay of heavy elements In the case of heavy element decay chains (e.g. lead, radium, uranium, plutonium and americium), where the parent and progeny are not necessarily in equilibrium (e.g. after chemical separation or extraction of an element) the respective activities of each of the progeny in the chain will depend upon the age of the sample. It is not possible to cover all situations likely to be encountered but the Appendix to this handbook includes selected major heavy element decay chains and the activities of respective progeny in the chain are given as a fraction of the parent activity as a function of time (age since separation) for 0.1 day, 1 day, 10 days, 100 days, 1 year, 10 years and 100 years. The parent radionuclides for which this information is provided are marked ‘!!’ in the main data sheets, including an indication of the first and last elements within the chains. A similar approach can be adopted if data on natural uranium or re-constituted ‘natural’ uranium are required by taking into account the isotopic composition. The data in the Appendix are derived using RadDecay version 1.13 (16).

Calculation of Asatm Asatm is the value of the activity per unit surface area of the contaminated zone which leads to an ALIinhalation dose with an atmospheric re-suspension factor, Tatm, of 10-4 m⫺1 and an annual occupational exposure of 2000 h with 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 surface area of the contaminated zone which leads to an ALIingestion dose with an organ transfer factor, Tingestion, of 1 cm2.h⫺1 and an annual occupational exposure of 2000 h. This component may simulate, for example, the risk due to a chronic exposure at a working post. Asingestion (Bq.cm⫺2) is given by the following equation:

Derived surface contamination limits The derived surface contamination limit (DSCL) is a non- regulatory quantity enabling exposure risks due to removable and/or fixed surface contamination to be quantified. Exposure to the DSCL leads to a dose which does not exceed the maximum annual occupational exposure limit (corresponding to 2000 working hours). A model is used to determine the DSCL values corresponding to the different radionuclides dealt with (17). This model takes different modes of exposure and transfer parameters into consideration. These data are given for completeness and the values do not supersede limits imposed by regulations, which are invariably much lower. Occupational exposures may result from:

Asingestion = ALIingestion/(2000 Tingestion) Calculation of Asskin Asskin is the value of the activity per unit surface area of the contaminated zone leading to an annual skin dose of 500 mSv for an occupational exposure of 2000 h and a skin transfer factor, Tskin of 0.1. It is assumed that contamination is eliminated on a daily basis when the user washes on leaving the working zone. Neither radioactive decay nor the renewal of skin cells (which contribute towards eliminating radioactivity) are taken into account. Asskin (Bq.cm⫺2) is given by the following equation:

䉬 transfer of surface contamination to the atmosphere leading to internal exposure through inhalation 䉬 transfer of surface contamination to the organism through ingestion

Asskin = 500/(Tskin × 2000 × Dp) where Dp (mSv.h⫺1) is a conversion coefficient giving 13

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

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

practicable by appropriate maintenance and cleaning operations. When accidental contamination (even of very low level) occurs, decontamination must be carried out as quickly as possible.

Calculation of Assurf Assurf (Bq.cm⫺2) is the value of the activity per unit surface area of the contaminated zone leading to an effective dose equivalent of 20 mSv for an occupation exposure of 2000 h. Assurf (Bq.cm⫺2) is given by the following equation:

Detection probes

The DSCL value corresponding to most radionuclides lies between 1 Bq.cm⫺2 and 105 Bq.cm⫺2, and, in general, is of the order of 100 Bq.cm⫺2. In the case of alpha emitters, the model can lead to very restrictive values; a lower limit of 0.04 Bq.cm⫺2 has been imposed for alpha emitters.

Surface contamination due to many of the radionuclides covered in these sheets can be detected with soft ␤, ␤, ␣, X ray or ␥ probes. Preferences given to particular types of probe or instrument are indicated for each type of radiation. A single + means that such a type of probe or instrument can possibly be used, double ++ means that this type of probe or instrument is recommended, whilst the absence of a + sign means that that type of probe or instrument is not suitable. Absence of a + sign in any box means that direct measurement is inappropriate and a wipe test, in association with a suitable detection system, such as liquid scintillation methods, would be appropriate. This applies only to tritium in this handbook but the same technique can be applied with advantage to low levels of contamination of other radionuclides, especially those with low penetration emissions.

Determination of removable and fixed contamination values

INTERNAL EXPOSURE

Assurf = 20/(2000 Dfloor) where Dfloor (mSv.h⫺1) is a conversion coefficient giving the effective dose equivalent 1 m above a floor infinitely and uniformly contaminated at 1 Bq.cm⫺2(14). Alpha emitters

DSCL values enable derived limits to be determined for removable (or non-fixed) and fixed contamination. The above limits enable the potential risk to users from surface contamination to be assessed together with the most appropriate means for decontamination. In the case of removable contamination, the derived limit is defined as being equivalent to 1/10th of the DSCL value of the radionuclide considered (this allows possible occupational exposures of other origins to be taken into consideration). In the case of fixed contamination, the derived level is defined as the value corresponding to a whole-body exposure from a uniformly contaminated floor equivalent to 1/10th of the annual exposure limits of 20 mSv. The DSCL value for fixed contamination always exceeds the value for removable contamination. For some radionuclides such as ‘pure’ beta or alpha emitters, the ratio of the two values can be highly significant (radiation from the floor being negligible). In this handbook, the DSCL value for fixed contamination is set at a maximum of 100 times the derived limit for removable contamination (e.g. see 14C on page 24). Regulations do not usually include specific limits for fixed contamination since its effect is taken into account in external radiation levels.

Committed effective dose per unit intake Internal exposures resulting from ingestion and inhalation are evaluated with the dose coefficients, e(g)ingestion and e(g)inhalation, expressed in Sv.Bq⫺1, and identically given by the ICRP (18), the IAEA Basic Safety Standards (2) and the L159 Euratom Directive (3). Listed values take the age of exposed individuals into account (⬎1 year, 1–2 years, 2–7 years, 7–12 years, 12– 17 years and ⬎17 years) as well as their status (public or occupational exposure). In this handbook only occupational exposures are considered. The committed effective dose per unit intake via ingestion is given for different gut transfer factor, f1, values. This factor quantifies the proportion of intake transferred to body fluids in the gut and depends on the chemical form of the radionuclide. The committed effective dose per unit intake via inhalation is given for three default lung absorption types (fast, moderate, slow). These coefficients depend on the chemical form and particle size of the aerosol. It is recalled that particle size is quantified by aerodynamic median activity diameter (AMAD) parameters. Inhalation coefficients are given for 1 and 5 ␮m AMAD values. Indication of the different forms for each nuclide that correspond to different gut transfer factors and lung absorption types are given in the data sheets. In some cases, for example phosphorus and sulphur, where the nuclide may form part of the anion, the gut transfer factor or lung absorption types may be predominantly

Contamination control The level of removable contamination of working zones and materials must be kept as low as reasonably 14

RADIONUCLIDE AND RADIATION PROTECTION DATA HANDBOOK 2nd Edition (2002)

determined by the associated cation. These are indicated in the data sheets.

equipment (chemical benches, fume hoods, glove boxes) or in given working areas (controlled or supervised areas). Whilst the purpose of this handbook is to provide data rather than advice on practices, it is felt appropriate to include information on ‘maximum recommended activities’. The data provided are appropriate to low and intermediate level laboratories and not to high level laboratories, where extensive professional radiation protection expertise will be available.

Annual Limits on Intake (ALI) Annual Limits on Intake by ingestion and inhalation (Iingestion,L and Iinhalation,L) are given in order to assist users who are not familiar with the dose per unit intake coefficients discussed above. Iingestion,L and Iinhalation,L are calculated using the following relationship (19): IL = DL/e(g)

Calculation models

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 the most restrictive e(g) values, respectively. They are calculated values (19), not values given by ICRP, and users should be aware of this. The references should be consulted, where appropriate, for further information.

The maximum activity (Ao) corresponding to a particular nuclide and handling situation is characterised by the potential risk associated with handling specific nuclides. External exposure due to ␥ and ␤ radiation components and the re-suspension of the nuclide in the atmosphere are taken into consideration. Ao (Bq) is determined from the following relationship: 1/Ao = 1/Abeta + 1/Agamma + 1/Avolatile

The ‘highest dose organ’ Various metabolic and dosimetric models have enabled retention and excretion functions after intake by inhalation by Reference Man to be established (19–21). These models have also enabled committed whole-body and organ doses to be determined. Some organs preferentially concentrate particular radionuclides. Therefore, the data sheets indicate which is the ‘highest dose organ’ (e.g. the thyroid in the case of iodine nuclides). The ‘highest dose organ’ depends on physicochemical form and whether intake results from ingestion or inhalation. Only inhalation (for an AMAD of 1 ␮m) is considered in this handbook. It should also be noted that these functions depend on the clearance time from the pulmonary region to the remainder of the body (F, M, S). Therefore, the ‘highest dose organ’ may also depend on the elimination class. 90Sr is an example; depending on whether F or S classes are considered either bone surface or lungs can be considered as being the ‘highest dose organ’. However, for some chemical forms the associated anion or cation may be more important in this context.

where Abeta (Bq) and Agamma (Bq) represent the maximum activities determined for a point source at a distance of 30 cm in air. These activities are calculated for 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 contamination risks directly dependent on the volatility of the 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 determined using the most restrictive ALIinhalation values described above; k values, related to the volatility of the product, are defined as follows: 䉬 k = 1: gases, substances with high saturation vapour pressures of about 1 bar at 20°C, substances penetrating the skin, particles smaller than 5 ␮m. 䉬 k = 0.1: substances with saturation vapour pressures of about 0.1 bar at 20°C, particles larger than 5 ␮m. 䉬 k = 0.01: low volatility substances with saturation vapour pressures of about 0.01 bar at 20°C (e.g. water). 䉬 k = 0.001: non-volatile substances with saturation vapour pressures less than 0.01 bar at 20°C.

General Internal exposure values are determined using various established metabolic and dosimetric models and hypotheses, which are subject to review by ICRP. Interpretation of the results of these analyses should be done by specialists. MAXIMUM RECOMMENDED ACTIVITIES

Dependence of maximum recommended activities on equipment and working areas

Operators or competent radiation protection personnel often require order-of-magnitude estimates on the maximum activities that can be handled with available

The following additional rules are used to determine 15

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

ium (3H), tritiated water and 37Ar have been deliberately reduced because of the difficulty of detecting the corresponding surface or atmospheric contamination. This factor can be modified in particular situations, depending on survey and control frequencies. In the case of iodine isotopes (123I, 125I, 131I), the existence of particularly unstable states (with the possibility of I2 release) has been considered. The corresponding limitations are consequently much more stringent.

the recommended limits for benches, fume cupboards and glove boxes: 䉬 It is forbidden to manipulate highly volatile substances (k ⬎ 0.01) or radionuclides with ALIinhalation values lower than 1000 Bq on chemical benches because even a low level contamination could lead to a significant ALIinhalation fraction that would be difficult to detect. 䉬 Manipulating substances under a fume hood is considered to increase protection by a factor of 10 in comparison with an open bench. 䉬 Manipulating substances in a glove box is considered to increase protection by a factor of 100 in comparison with fume hoods. 䉬 Maximum recommended levels of activity in supervised areas are nominally 3/10 of those in controlled zones (this approximately represents the respective exposure limits defined for these areas). However, the installation of glove boxes in supervised zones is generally deprecated; the level of protection afforded by a glove box is considered incompatible with the definition of such zones. It should be noted that the factor of 3/10 may be masked by rounding the data to one significant figure; rounding in any other way may imply an unjustified degree of accuracy in the models used for deriving the recommended levels of activity.

Radiation control Handled or stored activity values must comply with external exposure limits in situations where the external radiation hazard may be more restrictive than the internal radiation hazard. Attention is drawn to this fact in the data sheets by the remark ‘Subject to external exposure requirements which may be more restrictive’. Stored activities may exceed maximum recommended activities by a factor of 10, subject to any limits imposed by local or national regulations. In particular cases, listed values may be varied in local regulations, based upon assessments by professional radiation protection specialists. DESCRIPTION OF DATA SHEETS General

Limitations

Individual data sheets provide data in five main areas, namely basic radionuclide date, external exposure, contamination, internal exposure and maximum recommended activities in laboratory situations. These are divided into physical characteristics, (half life, specific activity and decay scheme, A1 and A2 reference transport activities, exemption levels, data for external exposure for five different geometries, external and internal exposure limits, appropriate contamination detection probes and half-value and tenth-value shielding thicknesses. The maximum activities that can be handled on benches, under fume hoods or in glove boxes are also given. The ‘highest dose organ’ is indicated, even where this is the whole body. As described earlier 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.

Area control The values indicated in these sheets apply mainly to the supervised and controlled areas of low and intermediate activity laboratories. Maximum recommended activities have further been limited to: Group Group Group Group Group

1 2 3 4 5

(red) 0.5 GBq (orange) 5 GBq (yellow) 5 GBq (green) 5 GBq (blue) 50 GBq

These limits have been imposed because such laboratories are not generally equipped with either a means for continuously monitoring the working environment, or for rejecting gases at a sufficient height to ensure that no recycling takes place. It should be noted that values calculated in compliance with these rules are guide values and are only indicative since other non-quantifiable parameters such as working methods and human factors must also be taken into account. Nevertheless, it has been observed that when practices are performed in accordance with these simple rules, only very low levels of atmospheric contamination and exposure occur. The maximum recommended levels for elemental trit-

GLOSSARY Activity The number of nuclear transformations occurring in a given amount of radionuclide per unit time. The SI unit of activity is the reciprocal second, s⫺1, with the special name becquerel (Bq). 16

RADIONUCLIDE AND RADIATION PROTECTION DATA HANDBOOK 2nd Edition (2002)

Annual limit of intake

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

The intake by inhalation or ingestion of a given radionuclide in a year by the Reference Man which would result in a committed dose equal to the relevant dose limit.

Effective dose

Bremsstrahlung

The quantity E, defined as a summation of the tissue equivalent doses, each multiplied by the appropriate tissue weighting factor (E = ⌺TWT HT where HT is the equivalent dose in tissue T and WT is the tissue weighting factor for tissue T). From the definition of equivalent dose it follows that E = ⌺TWT ⌺RWR DT,R.

X rays produced by deceleration of ␤ particles in materials. Committed equivalent dose Following an intake to the body of a radioactive material, the time integral of the equivalent dose rate is called the committed equivalent dose. If the integration time following the intake is not specified, it is implied that the value is 50 years for adults.

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)

Equivalent dose

The thickness of material necessary to reduce the intensity of X or gamma radiation by a factor of 2 and 10, respectively.

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 Radionucle´ ides. 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 De´ bits de Dose ␤ et ␥ en Fonction de la Ge´ ome´ trie 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 Controˆ le Radiopharmaceutique. 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. Dosime´ trie et The´ rapeutique des Contaminations Cutane´ s. CEA Report 5441 (1988). 16. RadDecay Version 1.13. Grove Enginering (Rockville, Maryland, USA) (1996). 17. Delacroix, D., Guerre, J. P. and Leblanc, P. De´ termination des limites de contamination surfacique pour les principaux radionucle´ ides. CEA-Saclay. Report DSCE/SRI, January 1992. 17

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

18

RADIONUCLIDE AND RADIATION PROTECTION DATA (2002) Radiation Protection Dosimetry Vol. 98, No. 1, pp. 19–20 (2002) Nuclear Technology Publishing

TABLE OF RADIONUCLIDES LISTED

Nuclide

Symbol

Tritium Beryllium - 7 Carbon - 11 Carbon - 14 Nitrogen - 13 Oxygen - 15 Fluorine - 18 Sodium - 22 Sodium - 24 Aluminium - 26 Silicon - 31 Phosphorus - 32 Phosphorus - 33 Sulphur - 35 Chlorine - 36 Argon - 37 Argon - 41 Potassium - 40 Potassium - 42 Potassium - 43 Calcium - 45 Calcium - 47 / Scandium - 47 Scandium - 46 Scandium - 47 Vanadium - 48 Chromium - 51 Manganese - 52m Manganese - 52 Manganese - 54 Manganese - 56 Iron - 52 Iron - 55 Iron - 59 Cobalt - 56 Cobalt - 57 Cobalt - 58 Cobalt - 60 Nickel - 63 Nickel - 65 Copper - 64 Copper - 67 Zinc - 65 Gallium - 66 Gallium - 67 Gallium - 68

(3H1) (7Be4) (11C6) (14C6) (13N7) (15O8) (18F9) (22Na11) (24Na11) (26Al13) (31Si14) (32P15) (33P15) (35S16) (36Cl17) (37Ar18) (41Ar18) (40K19) (42K19) (43K19) (45Ca20) (47Ca20 / 47Sc21) (46Sc21) (47Sc21) (48V23) (51Cr24) (52mMn25) (52Mn25) (54Mn25) (56Mn25) (52Fe26) (55Fe26) (59Fe26) (56Co27) (57Co27) (58Co27) (60Co27) (63Ni28) (65Ni28) (64Cu29) (67Cu29) (65Zn30) (66Ga31) (67Ga31) (68Ga31)

Arsenic - 73 Arsenic - 74 Arsenic - 76 Arsenic - 77 Selenium - 75 Bromine - 77 Bromine - 82 Krypton - 81 Krypton - 83m Krypton - 85 Krypton - 85m Rubidium - 86 Strontium - 85 Strontium - 89 Strontium - 90 / Yttrium - 90 Yttrium - 90 Yttrium - 91 Zirconium - 95 / Niobium - 95 Molybdenum - 99 / Technetium - 99m Technetium - 99m Technetium - 99 Ruthenium - 103 / Rhodium - 103m Ruthenium - 106 / Rhodium - 106 Palladium - 103 / Rhodium - 103m Silver - 110m Silver - 111 Cadmium - 109 Indium - 111 Indium - 113m Indium - 115m Tin - 125 Antimony - 122 Antimony - 124 Antimony - 125 / Tellurium - 125m Antimony - 126 Tellurium - 123m Tellurium - 125m Tellurium - 132 / Iodine - 132 Iodine - 123 Iodine - 124 Iodine - 125 Iodine - 129 Iodine - 131 Iodine - 132 Iodine - 133 Xenon - 133 Caesium - 131

Page 21 22 23 24 25 26 27 28 29 30 31 32 33 34 35 36 37 38 39 40 41 42 43 44 45 46 47 48 49 50 51 52 53 54 55 56 57 58 59 60 61 62 63 64 65 19

(73As33) (74As33) (76As33) (77As33) (75Se34) (77Br35) (82Br35) (81Kr36) (83mKr36) (85Kr36) (85mKr36) (86Rb37) (85Sr38) (89Sr38) (90Sr38 / 90Y39) (90Y39) (91Y39) (95Zr40 / 95Nb41) (99Mo42 / 99mTc43) (99mTc43) (99Tc43) (103Ru44 / 103mRh45) (106Ru44 / 106Rh45) (103Pd46 / 103mRh45) (110mAg47) (111Ag47) (109Cd48) (111In49) (113mIn49) (115mIn49) (125Sn50) (122Sb51) (124Sb51) (125Sb51 / 125mTe52) (126Sb51) (123mTe52) (125mTe52) (132Te52/132I53) (123I53) (124I53) (125I53) (129I53) (131I53) (132I53) (133I53) (133Xe54) (131Cs55)

66 67 68 69 70 71 72 73 74 75 76 77 78 79 80 81 82 83 84 85 86 87 88 89 90 91 92 93 94 95 96 97 98 99 100 101 102 103 104 105 106 107 108 109 110 111 112

D. DELACROIX, J. P. GUERRE, P. LEBLANC and C. HICKMAN Caesium - 134 Caesium - 137 / Barium - 137m Barium - 133 Barium - 140 / Lanthanum - 140 Lanthanum - 140 Cerium - 139 Cerium - 141 Cerium - 143 Praseodymium - 143 Praseodymium - 144 Promethium - 147 Samarium - 153 Europium - 152 Europium - 154 Europium - 155 Europium - 156 Erbium - 169 Thulium - 170 Thulium - 171 Ytterbium - 169 Rhenium - 186 Rhenium - 188 Iridium - 192 Gold - 198 Mercury - 197 Mercury - 203

(134Cs55) (137Cs55 / 137mBa56) (133Ba56) (140Ba56 / 140La57) (140La57) (139Ce58) (141Ce58) (143Ce58) (143Pr59) (144Pr59) (147Pm61) (153Sm62) (152Eu63) (154Eu63) (155Eu63) (156Eu63) (169Er68) (170Tm69) (171Tm69) (169Yb70) (186Re75) (188Re75) (192Ir77) (198Au79) (197Hg80) (203Hg80)

113 114 115 116 117 118 119 120 121 122 123 124 125 126 127 128 129 130 131 132 133 134 135 136 137 138

Thallium - 201 Thallium - 204 Lead - 210 Lead - 214 Bismuth - 207 Bismuth - 210 Bismuth - 214 Polonium - 210 Radium - 226 Thorium - 231 Thorium - 234 Protactinium - 234 Protactinium - 234m Uranium - 233 Uranium - 234 Uranium - 235 Uranium - 238 Neptunium - 239 Plutonium - 238 Plutonium - 239 Plutonium - 240 Plutonium - 241 Americium - 241 Americium - 243 Curium - 244 Californium - 252

20

(201Tl81) (204Tl81) (210Pb82) (214Pb82) (207Bi83) (210Bi83) (214Bi83) (210Po84) (226Ra88) (231Th90) (234Th90) (234Pa91) (234mPa91) (233U92) (234U92) (235U92) (238U92) (239Np93) (238Pu94) (239Pu94) (240Pu94) (241Pu94) (241Am95) (243Am95) (244Cm96) (252Cf98)

139 140 141 142 143 144 145 146 147 148 149 150 151 152 153 154 155 156 157 158 159 160 161 162 163 164

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