foia/pa-2015-0050 - resp 1 - partial. group b (records ... · bclwcen 0.2 and 10 pm and wt(h...

LIMITS FOR INTAKES OF RADIONUCLIDES BY WORKERS i

ADDENDUM

ICRP Publication 30, Part I,1979

The following lines of text are amended as shown:

Page 37, line 4 from the bottom: >0.006 (not a)

Page 38, line 3: 0.006< AF(BS (not d )

Page 50, line 11 from the bottom and the last line: (Sv m’ LQ-’ h-‘) (not m-‘)

The figures 5.1 and 5.2 published on pages 24 and 23 of KRP Publication 30 are amended as follows:

2 ‘0 r IO -

5-

2-

I-

15-

)2-

0’1

\ \ -1 \ \ \

$81 Illtli k I IO 20 30 50 70 80 90 35

Percent dcposmon

i

I

4

-J 9 9

Fig. 5. I. Deposition of dust in lhe respiratory system. The percentage or activity or mass of an aerosol which is deposited in Ihe N-P, T-B and P regions is given in relation lo the Activity Median Aerodynamic Diameter (AMAD) of the aerosol distribution. The model is intended for use with aerosol distributions with AMADs bclwcen 0.2 and 10 pm and wt(h geometric standard deviations of less than 4.5. Provisional estimates of deposition further cxlcnding the size range are given by the dashed lines. For an unusual distribution with an AMAD of greater than 20 jam. complete deposition in N-P can be assumed. The model Jas not apply lo aerosols with

AMADs of less than 0.1 pm.

ii REPORT OF COMMITTEE 2

ChSS

D W Y

Cornpart- T T T Region ment day F day F day F

:I, = 0.30) a b 0.01 0.01 0.5 0.5 0.01 0.40 0.9 0.1 0.01 0.40 0.01 0.99

:i,“_.= 0.08) C d 001 0.2 0.95 0.05 0.0) 0.2 0.5 0.5 0.01 0.2 0.01 0.99

e 0.5 0.8 50 0.15 500 0.05 PD, = 0.25) f n.a. n.a. 04 10 04

g h ::f. :I?

5b.O 0’4 500’ 0.4 50 0:os 500 0:15

L i 0.5 I.0 50 I.0 IO00 0.9 j n.a. n.a. n.a. n.a. co 0.1

nodes

Fig. 5.2. Mathematical model used to describe clearance from the respiratory system. The values for the removal half-times. r._, and compartmental fractions, I=._, are given in the tabular portion of the figure for each of the three classec of retained materials. The values given for D 11-). DT_B and 4 Och column) arc the regional depositions for an aerosol with an AMAD of I pm. The schematic drawing identitles the various clearance pathways from compartments a-i in the four respiratory regions, N-P, T-B, P and L.

n.a. - not applicable.

Page 76 of Part 1 and pages 54 and 55 of its Supplement: As a consequence of the Commission’s decision to reduce its recommended dose equivalent limit for the lens of the eye from 0.3 Sv to 0.15 Sv in a year, values of DAC for *’ Kr and 83aKr arc amended as follows.

Derived air concentrations DAC(Bq m-‘) (40 h wk) for isotopes of krypton

Radionuclidc Semi-infinite

cloud 1000m3rootn 500 m’ room 100 m’ room

“Kr

““‘Kr

2 x 10’ 1 x to* lxloy 1 x lo* (5 x 100) (6 x Id) (9 x IO’)

Lens Lens Lens 4x IO’ 4x10” 4x IO’ 4x 10’

(7 x 109) (7 x 109) (7 x 109) (8 x 109) Lens Lens hS Lens

Page 78, The values of ALI and DAC for *OSr are incorrect because no allowance was made for the contribution to committed dose equivalent of *ORb, the dau&ter of *%r. The corrected dosimetric data are given in Supplement B of Part 3. The correct values of ALI and DAC are given below :

. . . LIMITS FOR INTAKES OF RADIONUCLIDES BY WORKERS 111

Annual limits on intake, ALI and derived air concentrations, DAC(Bq m- ‘) (40 h wk) for isotopes of strontium

Inhalation

Oral Class D Class Y

Radionuclide

9r

j,=3x10-’ f,=lx10-2 f,=3x 10-l j-,=1x10-’

AL1 2x lo* 2x 10’ 4x IO8 5x IO8 DAC - - 2 x 10’ 2x 10’

Page 82, The values of ALI and DAC given for ““‘Nb were calculated using the incorrect assumption that the radionuclide would be distributed uniformly to bone surface rather than throughout mineral bone as stated in the metabolic model. The correct values of ALI and DAC are given below. Revised tables of dosimetric data are given in Supplement B to Part 3.

Annual limits on intake, ALI and derived air concentrations, DAC(Bq m-“) (40 h wk) for isotopes of niobium

Inhalation

Oral Class w Class Y

Radionuclide /,=I x10-2 j,=1x10-* /,=1x10-’

““‘Nb AL1

DAC

3xIP (4x l(r) LLI Wall

-

7x10’ 6x IO’

3x 10. 3 x 10’

Page 84, In Section 3 it is incorrect to state that none of the radioisotopes of molybdenum considered has a radioactive half-life greater than 3 days. g3Mo has a half-life of 3.5 x 10’ years and should perhaps be considered to be distributed throughout the volume of mineral bone rather than on its surface following its deposition in the skeleton. Such an assumption will make only a marginal difference to the values of ALI and DAC quoted.

Page 86, In Section 3 it is incorrect to state that none of the radioisotopes of tellerium considered has a radioactive half-life greater than 200 days. ’ 23Te has a radioactive half-life of 10’ ’ years and should perhaps be considered as a volume rather than a surface seeker of mineral bone. This would have the effect of increasing the values of AL1 shown. However, the mass of the AL1 is so great that ‘*‘Te will not present a radiation hazard in any practical circumstance.

Page 87, Some of the values of AL1 and DAC shown for radioisotopes of tellurium are incorrect due to a transcription error. The correct values are as follows:

iv REPORT OF COMMITTEE 2

Annual limits on intake ALI and derived air concentrations, DAC(Bq m-‘) (40 h wk) for isotopes of tdlurium

Radionuclide Oral

Inhalation

Class D Class w

‘“‘Te

INrnTe

“?e

“‘Tc

IJJqe

l”Te

ALI

DAC

ALI

DAC

ALI

DAC

ALI

DAC

ALI

DAC

ALI

DAC

/I=ZxlO-’

I x I@ (2 x I@) Thyroid

I x IO’ (2 x IO’) Thyroid

8x106 (2 x IO’) Thyroid

5xloB (1X104 Thyroid

I x lo8 (2 x I@) Thyroid

6x108 (9X I@? Thyroid

f,=2x IO-’

2x1@ (5 x 108) Thyroid 8x104

2x IO’ (5 x IO’) Thyroid 6x IO’

9x1@ (3 x IO’) Thyroid 4x IO’

8x108 (2x lo9 Thyroid 4x IO5

2x108 (5 x Iti) Thyroid 8x10”

9x108 (2 x I04 Thyroid 4x@

j,=2x IO-’

2x108 (4 x IO) Thyroid 8x10“

I x IO’ (3 x IO’) Thyroid 6x IO’

8x1@ (2 x IO’) Thyroid 3x103

8x108 (2x lo4 Thyroid 4x105

2xloB (5 x I@) Thyroid 8x10”

9x108 (2x IO4 Thyroid 4x I05

Page 88, The second paragraph of Section (c) Distribution and Retention should read as follows:

“Of iodine entering the transfer compartment a fraction, 0.3,1& assumed to be translocated to the thyroid while the remainder is assumed to go directly to excretion. Iodine in the thyroid is assumed to be retained with a biological half-life of 80 days and to be lost from the gland in the form of organic iodine. Organic iodine is assumed to be uniformly distributed among all organs and tissues of the body other than the thyroid and to be retained there with a biological half-life of 12 days. One-tenth of this organic iodine is assumed to go directly to faecal excretion and the rest is assumed to be returned to the transfer compartment as inorjpuric iodine so that the effective half-life of iodine in the thyroid is 120 days.”

Values of AL1 and DAC for radioisotopes of iodine are not changed.

PREFACE

In 1967 the Commission asked ICRP Committee 2 to prepare a new version of its report on Permissible Dose for Internal Radiation, which was issued in 1959 as ICRP Publication 2. It was recognized that the preparation of the report would depend heavily on the advice to be given by the task groups on reference man, on lung dynamics and on the radiosensitivity of the tissues in bone. Completion of the report also had to await a revision of the Commis- sion’s basic recommendations, now available as ICRP Publication 26.

The following were the members of ICRP Committee 2 preparing the final version of the report :

J. Vennart (Chairman); W. J. Bair; G. C. Butler; G. W. Dolphin (Secretary); L. E. Feinendegen; W. Jacobi; J. Lafuma; C. Mays; P. E. Morrow; P. V. Ramzaev; W. S. Snyder (died June 1977) and R. C. Thompson.

In addition, the following were members of the committee during earlier stages of dis- cussions about the report:

B. Chr. Christensen; M. Dousset; M. Izawa; J. Liniecki; L. D. Marinelli; W. G. Marley; K. Z. Morgan; J. Miiller; V. Shamov and C. G. Stewart.

Committee 2 wishes to record its appreciation of the work of N. Adams and M. C. Thorne who materially assisted in the collection of data and in the preparation of this report, and to thank Mrs E. Henry and Mrs J. Rowley for secretarial assistance and great forbearance during the preparation of several drafts.

The dosimetric calculations have been the responsibility of a task group, originally working under the chairmanship of W. S. Snyder and latterly Mrs M. R. Ford, as follows:

W. S. Snyder (died June 1977); S. R. Bernard; L. T. Dillman (since May 1977); Mrs Mary R. Ford; J. W. Poston and Mrs Sarah B. Watson.

Committee 2 wishes to record its indebtedness to the task group and particularly to Mrs Ford for completing this exacting task under the very difficult circumstances following Dr Snyder’s untimely death.

V

APPRECIATION

w. s. SNYDER

The Commission wishes to record its special indebtedness to the late W. S. Snyder for his unsparing work in the preparation of this and many other ICRP reports. The work of Dr Snyder for Committee 2 in particular and for the advancement of radionuclide dosimetry in general was legendary even in his own life-time. He was tireless in his efforts and always willing to impart his great store of knowledge both to his immediate colleagues and to students and specialists throughout the world. With his passing we lost not only an honoured and trusted colleague but also a friend.

vi

Absorbed Fraction (AF) the fraction of energy emitted as a specified radiation type in a specified source tissue which is absorbed in a specified target tissue.

Activity Median Aerodynamic Diameter (AMAD) the diameter of a unit density sphere with the same terminal settling velocity in air as that of the aerosol particle whose activity is the median for the entire aerosol.

Annual Limit on Intake (ALI) The activity of a radionuclide which taken alone would irradiate a person, represented by Reference Man, to the limit set by the ICRP for each year of occupational exposure.

becquerel (Bq) the special name for the SI unit of activity, 1 Bq = 1 s- 1 (~2.7 x IO- 1 ’ Ci).

Cells near bone surfaces (BS) those tissues which lie within 10 pm of endosteal surfaces and bone surfaces lined with epithelium.

Committed Dose Equivalent (Hs,) the total dose equivalent averaged throughout a tissue in the 50 years after intake of a radionuclide into the body.

Cortical Bone equivalent to “Compact Bone” in ICRP Publication 20, i.e. any bone with a surface/ volume ratio less than 60 cm2 cm- 3 ; in Reference Man it has a mass of 4 000 g.

Deposition probability (in lung region) the fraction of the activity or mass of an inhaled aerosol which is deposited in a particular region of the lung.

Derived Air Concentration (DAC) equals the AL1 (of a radionuclide) divided by the volume of air inhaled by Reference Man in a working year (i.e. 2.4 x 10” m’). The unit of DAC is Bq m-j.

Derived Air Concentration for Submersion (DAC(Submersion)) one two-thousandth of the time integral of the concentration of a radionuclide in air which over a working year would alone irradiate a person to the limit specified by the ICRP.

Dose Equivalent (H) the product of the absorbed dose, the quality factor (Q) and the product of any other modifying factors (N). (currently N = 1.)

Dose-Equivalent Limits the maximum values of committed dose equivalent incurred in 1 year by individual organs (non-stochastic limits), and by the uniformly irradiated whole body (stochastic limit) of an occupationally exposed person, which are recommended in ZCRP Publication 26 (1977).

gray (GY) the special name for the SI unit of absorbed dose. 1 Gy = 1 J kg- ’ = 100 rad.

Lung Class (D, W or Y) a classification scheme for inhaled material according to its rate of clearance from the pulmonary region of the lung.

vii

Non-stochastic effects effects such as opacity of the lens of the eye for which the severity of the effect varies with the dose, and for which a threshold may therefore occur.

Quality Factor (Q) theprincipal modifying factor (which depends on the collision stopping power for charged particles) that is employed to derive dose equivalent from absorbed dose.

Red Bone Marrow (active) the component of marrow (assumed in this report to have a mass of 1500 g) which contains the buik of the haematopoietic stem cells.

Reference Man a person with the anatomical and physiological characteristics defined in the report of the ICRP Task Group on Reference Man (ICRP Publication 23).

sievert (Sv) the special name for the SI unit of dose equivalent. 1 Sv = 1 J kg- ’ E 100 rem.

Source Tissue (27) tissue (which may be a body organ) which contains a significant amount of a radionuclide following intake of that radionuclide into the body.

Specific Absorbed Fraction the fraction of energy emitted as a specified radiation type in a source tissue which is absorbed in 1 g of a target tissue.

Specific Effective Energy (SEE (Tt s),) the energy (MeV), suitably modified for quality factor, imparted per gram of a target tissue (T) as a consequence of the emission of a specified radiation (i) from a transformation occurring in source tissue (S).

Stochastic Effects malignant and hereditary disease for which the probability of an effect occurring, rather than its severity, is regarded as a function of dose without threshold.

Target Tissue (2’) tissue (which may be a body organ) in which radiation is absorbed.

Trabecular Bone equivalent to “Cancellous Bone” in ICRP Publication 20, i.e. any bone with a surface/ volume ratio greater than 40 cm2 cm- ’ ; in Reference Man it has a mass of 1000 g.

Transfer Compartment the compartment introduced (for mathematical convenience) into most of the metabolic models used in this report to account for the delay between radioactive material entering body fluids and its deposition in particular tissues.

Weighted Committed Dose Equivalent (wrHsO ,r) the product of the weighting factor and the committed dose equivalent for a specified tissue.

Weighting Factor (wr) the ratio of the stochastic risk arising from tissue T to the total risk when the whole body is irradiated uniformly.

. . . Vlll

1. INTRODUCTION

The International Commission on Radiological Protection issued ICRP Publication 2, the report of Committee II on Permissible Dose for Internal Radiation, in 1959. Some changes and additional recommendations concerning individual radionuclides were subsequently made in ICRP Publication 6 (1964) but no systematic revision of ICRP Publication 2 has been published. The report has served as a satisfactory guide for the control of intakes of radionuclides into the body to meet the basic standards of the Commission, although there have been some misconceptions about its intent and some misuse of its recommendations. New information on the effects of radiation on the body, on the uptake and retention of radioactive materials in body tissues and better data on radioactive decay schemes have accumulated in the intervening period. These factors and changes in the basic recommendations of the Com- mission described in ICRP Publication 26 have made it necessary for the Committee to publish a new report. In 1959, when the last report of Committee 2 was published, the existing recommendations of the Commission for occupational exposure were framed in terms of limits on the dose equivalent received by the organs and tissues of the body in any period of 13 weeks, with an additional restriction on the rate of accumulation of dose with age for the gonads, the blood-forming organs and the lenses of the eyes. Exposure of the individual was limited by the dose equivalent in the critical organ for which the ratio of the dose received to the dose limit was greatest (ICRP Publication 1, 1959 and Addendum, 1960). ICRP Publication 2 recommended values of the maximum permissible concentration in air and in water (MPC) and of maximum permissible body burden (MPBB) for a number of radionuclides. For any radionuclide, continuous exposure of a Standard Man to either of the MPC values during the whole of a working life-time of 50 years would result in his containing the MPBB of the radionuclide at the end of that period. The MPBB was estimated to deliver a dose-equivalent rate in the appropriate critical organ such that the Commission’s recommended limits on dose equivalent for any relevant stated period were not exceeded. Although the Commission emphasized in ICRP Publication 2 that the rate of intake of a radionuclide could be varied, provided that the intake in any quarter was no greater than that resulting from continuous exposure to the appropriate MPC for 13 weeks, the concept of MPC has been misused to imply a maximum concentration in air or water that should never be exceeded under any circumstances. Similarly, MPBB has been misused to imply a maximum for the activity in the body, although it is evident that an intake in excess of MPBB for a radionuclide with a short residence time in the body would not result in the Commission’s limits on dose being exceeded.

For these and other reasons discussed below the concepts of MPC and MPBB are not used in this report.

As discussed in Chapter 4, SI quantities and units are used throughout this report. The special name for the SI unit of absorbed dose is the gray (Gy), of dose equivalent the sievert (Sv) and of activity the becquerel (Bq). Where a limit on dose equivalent recommended by the Commission is given in the text, the value in rem will be shown in parentheses.

Recommendations of the Commission have been developed over the years, first to avoid radiation-induced effects of early onset and later to minimize the possibility of radiation- induced cancer and hereditary disease. There are now sufficient data available about the effects of radiation for the Commission to give values for the risk per unit dose equivalent of radiation-induced fatal cancer in exposed people and of serious disease in their offspring (paras. 36-67, ICRP Publication 26).

2 REF’ORT OF COMMITlEE 2

It is intended that these values will be applied at the doses and dose rates of interest in radiological protection using the hypothesis that risks of these effects are linearly related to dose equivalent without threshold. Therefore, the risks of fatal cancer in an individual, group or population and of hereditary disease in their offspring is determined either by the total dose equivalent received by individuals, independent of dose-equivalent rate and the way dose equivalent is fractionated, or by the collective dose equivalent within the group or population exposed (paras. 22-24 ICRP Publication 26). Accordingly, the Commission proposes to minimize the risks of these diseases by a system of dose limitation (para. 12 ZCRP Publication 26), the main principles of which are as follows:

(a) no practice shall be adopted unless its introduction produces a positive net benefit;

(b) all exposures shall be kept as low as reasonably achievable, economic and social factors being taken into account;

(c) the dose equivalent to individuals shall not exceed the limits recommended for the appropriate circumstances by the Commission.

This report concerns only the derivation of the secondary standards that limit the intake of radionuclides by workers in compliance with clause (c). It is recognized, however, that in many circumstances the cost-benefit analysis and judgment implied in clauses (a) and (b) might result in the average dose to the work-force being far below these limits.

The basic limits to be used for the control of occupational exposure are discussed in Chap ter 2. For occupational exposure to radioactive materials the Commission believes that the time over which the dose equivalent should be integrated is a working life of 50 years .as heretofore. In this report, the total dose equivalent in any tissue over the 50 years after intake of a radionuclide into the body is termed the Committed Dose Equivalent, N,,.

Several organs and tissues will be irradiated following the entry of a radionuclide into the body. In any year of practice, the values of Hse in all the organs of the body must be limited so that the resulting total risk of cancer and hereditary disease is less than or equal to the risk from irradiating the whole body uniformly to the appropriate annual dose-equivalent limit of 50 mSv (5 rem) recommended by the Commission in ZCRP Publication 26. Additional constraints are imposed to prevent the occurrence of those effects which are produced only when some threshold of dose is exceeded. The method used to limit values of Hs, is discussed in Chapter 2.

Secondary and derived limits for use in practice are discussed in Chapter 3. For each radionuclide in each of several chemical forms a secondary standard is given for the Annual Limit on Intake (ALI) either by ingestion or by inhalation. A Reference Man (ZCRP Publication 23) receiving any such intake would be irradiated to the limit set by the Commission for each year of occupational exposure as described in the previous paragraph. Values are also given for the Derived Air Concentration (DAC) of each radioactive material. This is obtained by dividing the AL1 by the volume of air inhaled by Reference Man in a working year. DACs are also derived for submersion of Reference Man in a cloud of a radioactive inert gas and in a cloud of elemental tritium. In such a case, a limit is imposed in any year of practice on the dose equivalents received by the organs and tissues of the body from external radiation.

The values of DAC are analogous to the values of MPC(air) derived in ZCRP Publication 2. It is emphasized that values of DAC must always be used circumspectly. For inhalation of a radionuclide the overriding limit is the AL1 and for submersion it is the time integral of the concentration of the radionuclide in air during any year of practice. No standard is developed here for

LIMITS FOR INTAKES OF RADIONUCLIDES BY WORKERS 3

a derived concentration in water because water is only one source of ingested material. The total activity ingested in any year should be controlled by use of the value of ALI for ingestion.

The general principles used to calculate H,e per unit intake of a radionuclide are described in Chapter 4. In this report, HsO in any target organ includes the dose from photons emitted by the radionuclide in other organs of the body as well as that from radiations arising from the radionuclide in the target organ itself. The total energy absorbed per unit mass in any target organ for each transformation of the radionuclide in any source organ, suitably weighted by the Quality Factor, is known as the Specific Effective Energy (SEE). Hs,-, in any target organ is directly related to SEE and to the number of transformations, U, which occur in source organs during the 50 years after intake of the radionuclide. All the values given for H,e refer to unit intake of the stated radionuclide alone, but include the committed dose equivalent contributed by any daughter radionuclide produced from transformations of the parent radionuclide within the body. The principles of the methods used to calculate U for the parent radionuclide and U’ etc. for daughters are also discussed in Chapter 4. It is acknow- ledged that any model used to describe the kinetics of a radionuclide and its daughters in the body will be an over-simplification of what could happen in practice. The models used in this report are adopted in the interests of providing a unified approach to the dosimetry of radionuclides in the body and are thought to be sufficiently accurate for the present purpose, which is to derive limits for the exposure of workers to radioactive materials.

Specific models for the routes of entry of radioactive materials into the body, namely the Respiratory System and the Gastrointestinal (GI) Tract, are described in Chapters 5 and 6 respectively. The model of the respiratory system is used to calculate the average committed dose equivalent in the lung, considered by the Commission for purposes of radiological protection as a composite tissue comprising the trachea and bronchi, the pulmonary region and its associated regional lymphoid tissues. The model is also used to calculate those fractions of the inhaled activity which are transferred either to body fluids or to the GI tract. Similarly, the model of the GI tract is used firstly to calculate the average committed dose equivalents in anatomically distinct sections of the tract, which are regarded as separate organs in the application of the Commission’s basic limits, and secondly to calculate the fraction of ingested activity transferred to body fluids.

Chapter 7 describes the model used for dosimetry of radionuclides in bone. Committed dose equivalent is estimated for the relevant target tissues, identified as certain cells near bone surfaces and the red bone marrow.

Chapter 8 describes the methods used to derive a DAC for submersion of an individual in a radioactive, chemically inert gas or in elemental tritium. Arguments are presented to show that the limiting factor for tritium is irradiation of the lung, and, for inert gases, external irradiation of the body. It is shown that internal irradiation of the body from absorbed gas can be disregarded.

The relevant metabolic data required to calculate AL1 and DAC for the radioisotopes of a number of elements are set out in pp. 63-l 15 of this report. They give data on daily intake, excretion, body content and distribution for the stable element taken from the report on Reference Man (ZCRP Publication 23) and other sources. A description of the metabolism of compounds of the element is given, together with values for their absorption from the GI tract to the body fluids, their lung clearance classification as used in the dosimetric model of the respiratory system and models describing their retention in the whole body and relevant organs of adults. After each set of metabolic data for an element values are given for AL1 and DAC for radioisotopes with radioactive half-life in excess of 10 min. The relevant dosimetric data for individual isotopes of the element are given in a Supplement to this report in the form of a

4 REPORT OF COMMITI’EE 2

computer print-out. They include information on the radioactive transformations of the radionuclide, values of SEE for a number of target organs from activities in different source organs, the number of transformations over 50 years in several source organs per unit intake by ingestion and inhalation of various chemical forms, corresponding values of H,,, and of ~~~ weighted for sensitivity of the tissues to radiation-induced cancer and hereditary disease, and finally a table of values of ALI and DAC for occupational exposure. Further metabolic data on the elements and dosimetric data on their radioisotopes will be published as they become available.

Chapter 9 is concerned with limitations on the use of the data. It is stressed that the total dose equivalent which an individual has actually received and that to which he might be committed as a result of past exposure to radioactive materials is the primary factor determining his radiation status. The values of Hso, ALI and DAC given here or in the Supplement for radionuclides are criteria that are useful only for interpreting the occupational exposure of adults. Information is not given for other ages. The values of AL1 are based on consideration of radiation dose alone; chemical toxicity has not been considered.

The Commission wishes it to be emphasized that the data given here are to be used only within the framework of its basic recommendations as described in ICRP Publication 26.

Particular attention is drawn to para. 102 of those recommendations which implies that long- continued intakes of radionuclides at or near their ALIs by a considerable proportion of the workers would only be acceptable if a careful cost-benefit analysis had shown that the result- ant risk would be justified. The models used in this report have been chosen, often conserva- tively, to derive values of AL1 and DAC to ensure the protection of workers. The data provided herein should therefore not be used indiscriminately out of context, e.g. to estimate the risk of cancer in individual cases. The various assumptions made should be regarded as such and not as matters of established fact. Further research and data are required to improve the models which have been used.

References ICRP Publication 1, Recommendations of the International Commission on Radiological Protection. (Adopted

9 September 1958.) Pergamon Press, Oxford, 1959. ICRP Report on Decisions at the 1959 Meeting of the International Commission on RadiologicalProtection:

Radiology 74, 116-119 (1960); Am. J. Roentg. 83, 372-375 (1960); Strahlentherapie 112. 481485 (1960); Acta. Radiol. 53, 166170 (1960); Br. J. Radiol. 33. 189-192 (1960).

ICRP Publication 2, Recommendations of the International Commission on Radiological Protection Report of Committee II on Permissible Dose for Internal Radiation (1959). Pergamon Press, Oxford, 1960.

ICRP Publication 6. Recommendations of the International Commission on Radiological Protection (as amen&d 1959 and revised 1962). Pergamon PI-&, Oxford, 1964.

ICRP Publication 9. Recommendations of the International Commission on Radioloaical Protection. (Adovted 17 September 1985.) Pergamon Press, -Oxford. 1966.

. _

ICRP Publication 23, Internutional Commission on Radiological Protection, Task Group Report on Reference Man. Pergamon Press, Oxford, 1975.

ICRP Publication 26. Recommendations of the International Commission on Radiological Protection. Pergamon Press, Oxford, 1977.


2. BASIC LIMITS FOR THE CONTROL OF INTERNAL DOSE

2.1. Introduction

In the decade or so since the Commission’s former recommendations were published, more information has become available on the relation between absorbed doses of ionizing radiation and risks of biological effects. This new information enables the Commission to establish limits for exposure to ionizing radiation on a different basis than it has used heretofore (ICRP Publication 26). Two broad categories of radiation-induced effects are considered, namely,

(i) malignant and hereditary disease for which the probability of an effect occurring, rather than its severity, is regarded as a function of dose without threshold (stochastic effects), and

(ii) effects such as opacity of the lens and cosmetically unacceptable changes in the skin for which a threshold or pseudo-threshold of dose must be exceeded before the effect is induced (non-stochastic effects).

The Commission’s recommendations are intended to prevent non-stochastic effects and to limit the occurrence of stochastic effects to an acceptable level.

2.2. Dose-Equivalent Limits for Occupational Exposure

The Commission believes that non-stochastic effects will be prevented by applying a dose- equivalent limit of 0.5 Sv (50 rem) in a year to all tissues except the lenses of the eyes, for which the Commission recommends a limit of 0.3 Sv (30 rem) in a year. These limits apply irrespec- tive of whether the tissues are exposed singly or together with other organs, and they are intended to constrain exposures that fulfil the limitation on stochastic effects discussed below.

For stochastic effects the Commission’s recommended system for limiting exposure is based on the principle that the limit on risk should be equal whether the whole body is irradiated uniformly or whether there is non-uniform irradiation. This condition will be met if

(2.1)

where wr is a weighting factor representing the ratio of the stochastic risk resulting from tissue (T) to the total risk when the whole body is irradiated uniformly;

H, is the dose equivalent received by tissue (T), and

Hwb is the stochastic dose-equivalent limit for uniform irradiation of the whole body.

For occupational exposure the Commission recommends the value 50 mSv (5 rem) in any year for Hwb and the values of wT shown in Table 2.1. The values are recommended as appropriate for the protection of any worker, independent of age or sex (para. 104-106, ICRP Publication 26).

The value of wT shown for bone surfaces relates to endosteal cells and to the epithelium on bone surfaces (Chapter 7). The value of wr for the remainder of tissues requires further clari- fication. For reasons stated by the Commission in para. 105, ZCRP Publication 26, a value of wr of 0.06 is applicable to each of the five organs or tissues of this remainder receiving the greatest dose equivalents and the exposure of all other tissues in this group is neglected.

REPORT OF COMMITTEE 2

Table 2.1. Weighting factors recommended by the Commission for sto-

chastic risks

Organ or tissue

Gonads Breast Red bone marrow Lung Thyroid Bone surfaces Remainder

0.25 0.15 0.12 0.12 0.03 0.03 0.30

Because of their very low sensitivity to cancer induction, skin and lens (paras. 61-64, ICRP Publication 26) are not considered as part of remainder tissue for limiting stochastic effects. When the gastrointestinal tract is irradiated, the stomach, small intestine, upper large intestine and lower large intestine are considered as four separate organs.

2.3. Limits for the Intake of Radioactive Materials by Workers

The estimates of risk of radiation-induced cancer and hereditary disease on which the Commission’s dose-equivalent limits for stochastic effects are based were made using the hypothesis that risk of an effect is linearly related to dose equivalent. Therefore, it is the total dose equivalent averaged throughout any organ or tissue, independently of the time over which that dose equivalent is delivered, which determines the degree of effect in that tissue. With regard to limits on the intake of a radioactive material into the body, the Commission has reconsidered the question of the time over which this total dose equivalent should be integrated and has concluded that the period of 50 years used heretofore is appropriate for an occupational lifetime. The total dose equivalent in any tissue over the 50 years after intake of a radionuclide into the body is termed the Committed Dose Equivalent, Hso. It is emphasized that this is the dose equivalent which a Reference Man is assumed to receive if he lives for 50 years after his intake of the radioactive material and if no steps are taken to accelerate the removal of the radionuclide from his body.

Therefore, in order to meet the Commission’s basic limits for the exposure of workers, the intakes of radioactive materials in any year must be limited to satisfy the following conditions

7 wTHsO, .,- 5 0.05 Sv (2.2a)

and Hs,,. 5 0.5 Sv (2.2b)

where wz is the weighting factor shown in Table 2.1, and

H 5,,T (in Sv) is the total committed dose equivalent in tissue (T) resulting from intakes of radioactive materials from all sources during the year in question.

Relationship (2.2a) limits stochastic effects and (2.2b) non-stochastic effects, arising from intakes of radioactive materials. With regard to (2.2b), the limit for non-stochastic effects in any tissue is taken as 0.5 Sv (50 rem), since no case is known where lens opacity would be the factor limiting intake of radioactive materials. It might be the limiting factor when the body


is irradiated from the exterior by submersion in a radioactive noble gas and values for the derived air concentration in such cases are discussed in Chapters 3 and 8.

2.4. Restrictions on the Rate of Intake of Radioactive Materials

The Commission believes it is sufficient to limit the intake of radioactive materials in any year to the recommended limits and does not recommend any further restrictions on the rate of intake, except in the case of occupational exposure of women of reproductive capacity and pregnant women as discussed in paras. 35, 115 and 116 of ICRP Publication 26.

Reference ICRP Publication 26, Recommendations of the International Commission on Radiological Protection. F’eraamon

Press. Oxford, 1977.

8 REPORT OF COMMITTEE 2

3. SECONDARY AND DERIVED LIMITS FOR THE CONTROL OF INTERNAL DOSE

3.1. Reference Man

The standards for control of internal dose for workers are derived for an adult person of stated anatomical and physiological features which are defined in the report of the ICRP Task Group on Reference Man (ICRP Publicution 23); see also Table 4.1 (Chapter 4). The transfer of radionuclides to blood following inhalation or ingestion are discussed in Chapters 5 and 6, which set out dosimetric models for the respiratory system and gastrointestinal tract respectively. The metabolic data for the various elements considered in this report include a description of their behaviour in the respiratory system and gastrointestinal tract following inhalation and ingestion and their partition and retention among the various body organs and tissues following entry to body fluids.

3.2. Committed Dose Equivalent (If,,)

The committed dose equivalent (H,,) in a particular organ or tissue is the total dose equivalent to which that organ or tissue would be committed during the 50 years after intake of a radionuclide. Estimates of H,, per unit intake of a radionuclide are made by the methods described in Chapter 4, and values are given for a number of organs and tissues in the Supple- ment which gives dosimetric data for individual radionuclides. These estimates are based on the characteristics of Reference Man.

3.3. Ammal Limit on Intake (ALI)

The annual limit on intake (ALI) of a radionuclide is a secondary limit designed to meet the basic limits for occupational exposure recommended by the Commission and is derived from relationships (2.2a) and (2.2b) in Chapter 2 as follows. AL1 is the greatest value of the annual intake Z which satisfies both of the following inequalities

Z 7 w,-(Zf,,,, per unit intake) 5 0.05 Sv (3. la)

1 (H,,,7 per unit intake) 5 0.5 Sv (3.lb)

where Z (in Bq) is the annual intake of the specified radionuclide either by ingestion or inhalation;

wr is the weighting factor for tissue (T) and has the values shown in Section 2 of Chapter 2, and

H50,T per unit intake (in Sv Bq- ‘) is the committed dose equivalent in tissue (T) from the intake of unit activity of the radionuclide by the specified route.

3.3.1. Application of annual limits on intake

(a) Values of ALI are estimated for exposure, both by ingestion and inhalation, to the specified radionuclide. Any daughter radionuclides produced in the body after intake of the specified radionuclide are taken into account. When the intake consists of a mixture of radionuclides, and/or exposure occurs both by ingestion and inhalation, the values


of AL1 do not directly apply. In all such cases it is recommended that intakes in any one year should be controlled so that the equations set out in Section 2.3 of Chapter 2, relating to the committed dose equivalent from all sources, are satisfied. A more complete statement on the method recommended to control exposure to mixed sources of radioactive materials is given in Chapter 9.

(b) Values of AL1 and of HsO per unit intake are intended for use in the control of occupational exposure and are based on the parameters of a Reference Man. However, when there is reason to believe that a particular individual has taken in more than the ALI, and it is thought necessary to estimate the resulting committed dose equivalent, it is recommended that his age and known biological parameters should be taken into account as far as is practicable.

3.4. Derived Air Concentration (DAC)

Formerly the Commission recommended maximum permissible concentrations in air and water as guides to the control of exposure to radionuclides. Although intended to control exposure over prolonged periods of one-quarter of a year or more the values have often been misused to infer an over-exposure even when they have been exceeded only over a shorter period.

The Commission wishes to emphasize that the limit for inhalation of a radionuclide is the appropriate ALI. The concentration of a radionuclide in air during any year is limited as follows

J C(r) B(t) dt 5 AL1

where at any time 1,

Bq (3.2)

C(r) (in Bq m- ‘) is the concentration of the radionuclide in air;

B(t) (in m3 per unit time) is the volume of air breathed by the worker per unit time, and the limits on integration are over a working year.

The value of B(r) depends on the type of work being performed and for heavy work it could be more than twice the reference value for “light activity” given in ZCRP Publication 23.

For convenience, the Commission recommends values of derived air concentration (DAC). The DAC for any radionuclide is defined as that concentration in air (Bq rnm3) which, if breathed by Reference Man for a working year of 2 000 h (50 weeks at 40 h per week) under conditions of “light activity”, would result in the AL1 by inhalation.

DAC = AL1/(2 000 x 60 x 0.02)

= ALl/2.4 x 10’ Bq m-’ (3.3)

where 0.02 m3 is the volume of air breathed at work by Reference Man per minute under conditions of “light activity” (ZCRP Publication 23).

In the unlikely event of a person being exposed continuously to a DAC, and breathing air at a rate which is on average greater than 0.02 m3 per min, the activity inhaled in a working year would be more than the ALI. In this respect, it is emphasized that the ALI is the overriding limit and the derived limit DAC should always be used circumspectly.

10 REPORT OF COMMIlTEE 2

3.5. Derived Air Concentration for Submersion, DAC(Submersion)

For operations involving exposure to radioactive noble gases and elemental tritium, values are given for the derived air concentration for submersion of a Reference Man in a cloud of the gas, DAC(Submersion), The methods used to derive these values are described below and, in more detail, in Chapter 8.

3.5.1. DAC(Submersion) for elemental tritium

Exposure to elemental tritium in air during any year is limited by consideration of stochastic effects in the lung as follows

w~un~fi~un~ j C(t) dt < 0.05 sv (3.4)

where wLunD is the weighting factor for lung given in Table 2.1, Chapter 2;

fiLun( (in Sv m3 Bq-’ h - ‘) is the dose-equivalent rate to lung from exposure to unit concentration of tritium in air (i.e. 1 Bq m-‘);

C(t) (in Bq m- ‘) is the concentration of elemental tritium in air at any time t and the limits on integration are over a working year.

For convenience, the Commission recommends a value of DAC for elemental tritium which is l/2 000th of the greatest value of jC(t) dt that satisfies relationship (3.4). A worker may be exposed to concentrations of elemental tritium in air greater than the DAC provided that the conditions in relationship (3.4) are satisfied.

3.5.2. DAC(Submersion) for a radioactive noble gas

Exposure to a radioactive noble gas, other than radon and thoron which are the subject of a separate report, during any year is limited by consideration of external exposure of the body as follows

& w&I, j C(t) dt I 0.05 Sv (3Sa)

and

kr j C(t) dt 5 0.5 Sv (3Sb)

and

fiLcnr j C(t) dt < 0.3 Sv (3.5c)

where C(t) (in Bq rnq3) is the concentration of the radioactive noble gas in air at any time t and the limits on integration are over a working year;

wr is the weighting factor for tissue (T) and has the values shown in Section 2 of Chapter 2;

k, (in Sv m3 Bq- l h-r) is th e d ose-equivalent rate in any tissue (T), and

fiLc”S is the corresponding value for the lens of the eye, resulting from submersion of Reference Man in unit concentration of the noble gas in air (i.e. 1 Bq m-‘).

For convenience the Commission recommends a value of DAC for a radioactive noble gas which is l/2 000th of the greatest value of jC(t) dt that satisfies relationship (3.5).


(2 000 is the number of hours in a working year.) A worker may be exposed to concentrations of a noble gas in air greater than the DAC provided that the conditions in relationship (3.5) are satisfied.

Reference ICRP Publication 23, Report of the Task Group on Reference Man. Pcrgamon Press, Oxford, 1975.

12

4. COMMI’ITED DOSE EQUIVALENT (H,,,) AND ANNUAL LIMIT ON INTAKE (ALI)

4.1. Introduction

As explained in previous chapters, the Annual Limit on Intake (ALI) of a radionuclide into the body is determined by values of the committed dose equivalent HsO in all those target tissues which are irradiated to a significant extent. The methods used to calculate H,, per unit of activity taken into the body and AL1 are described below.


4.2.1. Activity

4.2. Quantities and Units

The International Commission on Radiation Units and Measurements (ICRU, 1971) states that the activity, A, of a quantity of a radioactive nuclide is the quotient of dN by dr where dN is the number of spontaneous nuclear transformations which occur in this quantity in the time interval dr.

A dN

=- dt

The special name for the SI unit of activity used in this report is the becquerel (Bq)

1 Bq = 1 s-l (-2.7 x 10-l’ Ci)

4.2.2. Dose Equivalent

Dose equivalent, H (ICRU, 1973), is the product of D, Q and N at the point of interest, where D is the absorbed dose, Q is the quality factor and N is the product of any other modifying factors.

H=DQN

In this report the special name for the SI unit of dose equivalent is the sievert (Sv) and of absorbed dose the gray (Gy).

1 Gy = 1 J kg-; (= 100 rad)

1 sv =lJkg-’ (= 100 rem)

The modifying factors Q and N are dimensionless.

4.3. Committed Dose Equivalent (H,,)

For purposes of planning in radiological protection it is assumed that risk of a given biological effect is linearly related to dose equivalent. In these circumstances, risk of an effect is determined by the total dose equivalent averaged throughout the organ or tissue at risk, independent of the time over which that dose equivalent is delivered. For planning work with radioactive materials the Commission recommends that the appropriate period for integration of dose equivalent is a working life-time of 50 years. The total dose equivalent averaged throughout any tissue over the 50 years after intake of a radionuclide into the body is termed the committed dose equivalent, Hso, which is therefore given by


H50 = c j” Dso,iQiN, dm I s Mdm

(4.1)

where M is the mass of the specified organ or tissue; and, for each type of radiation i;

D 5,,i is the total absorbed dose during a period of 50 years after intake of the radionuclide into the body in the element of mass dm of the specified organ or tissue;

Q, is the quality factor, and

N, is the product of all other modifying factors such as dose rate, fractionation, etc.

The quality factor, Q, is defined as a continuous function of collision stopping power in water (ZCRP Publication 21). Therefore the value of Qi will vary along the track of an ionizing particle and may be different for each element of mass dm in the irradiated tissue concerned. However, in view of the many uncertainties in estimating the dose to a tissue following the intake of a radioactive material, the Commission recommends (para. 20, ZCRP Publication 26) that for internal exposure the value of Q for a given type of radiation may be considered constant and have one of three values as follows

Q = 1 for beta particles, electrons and all electromagnetic radiation including gamma radiation, x rays and bremsstrahlung.

Q = 10 for fission neutrons emitted in spontaneous fission and for protons.

Q = 20 for alpha particles from nuclear transformations, for heavy recoil particles and for fission fragments.

The Commission recommends that the product of all other modifying factors, N, should be taken as 1 for values of dose equivalent less than or equal to the limits specified in Chapter 2. (Para. 18, ZCRP Publication ?6.) It will not be considered further in this report.

Therefore, using the values of Q shown above, which are constant for any type of radiation i, the expression for H,, shown in eqn (4.1) simplifies to

(4.2)

where D,o.i is the total absorbed dose during the 50 years after the intake of the radionuclide into the body averaged throughout the specified organ or tissue for each radiation of type i. A specified organ or tissue is one of those designated in Chapter 2 and in the metabolic data for individual elements.

In this report estimates are made of the committed dose equivalents in a number of target organs from the activity in a given source organ. For each type of radiation i, H50,i in target organ T resulting from radionuclide j in source organ S is the product of two factors

(a) the total number of transformations of radionuclide j in S over a period of 50 years after intake,

(b) the energy absorbed per g in T, suitably modified for quality factor, from radiation of type i per transformation of radionuclide j in S,

i.e. for each radiation of type i from radionuclide j

Hso(T+S)~= QtD,o(T+S)i = U, X 1.6 X 10-13SEE(TcS)i x lo3 sv


where Us is the number of transformations ofj in S over the 50 years following intake of the radionuclide;

1.6 x 10-l ’ is the number of joules in 1 MeV;

SEE (T+ S), (in MeV g- ’ per transformation) is the specific effective energy for radiation type i, suitably modified by quality factor, absorbed in T from each transformation in S (see Section 4.5) and lo3 is the conversion factor from g- 1 to kg-r.

.*. Hso(T+ S’)i = 1.6 X 10-r” UsSEE(Tt S), SV (4.3)

and for all types of radiation emitted by radionuclide j:

H,o(T+ S)j = 1.6 X lo-” [

Us 1 SEE(T+ S), i 1 SV. (4.4)

!

When the radionuclide has a radioactive daughter j’

H5o(T+ S)j+y = 1.6 X lo-” /

+ Us c SEE(T+ S), II sv. i v J

(4.5)

In general, for the intake of any mixture of radionuclides, i.e. parent with daughters and/or other radionuclides, Hso in target T from activity in source S is given by

C H,,(Tc S)j = 1.6 x lo- lo SEE(Tt S), i

lj Sv (4.6)

where the summation in j is over all the radionuclides involved. Finally, target T may be irradiated by radiations arising in several different sources S. The total value of H,, in target T is then given by

H - 1.6 x IO-” 11 5o.r - Sv. s I 1 i (4.7)

4.4. Cellular Distribution of Dose

Values of Hso derived in this report refer to the average committed dose equivalent in a target tissue specified in Section 2, Chapter 2. For parts of the gastrointestinal tract the target tissue is considered to be the mucosal layer (see Chapter 6), for the bone the cells lying within 10 pm of bone surfaces (see Chapter 7), and for the skin the basal layer of the epidermis, taken to be at a depth of 70 pm (see Chapter 8). In all other cases the position of sensitive cells within the target tissue has not yet been specified. It is recognized that there may be circumstances where the effects produced may be different from those expected from considerations of average dose, e.g. for radionuclides that emit radiations of very short range and which concentrate near radiosensitive microvolumes. With regard to radioactive particles, the Commission has expressed the view that, for late stochastic effects, the absorption of a given quantity of radiation is ordinarily likely to be less effective when due to a series of “hot spots” than when uniformly distributed (para. 33, ICRP Publication 26). Consideration also needs to be given to those compounds labelled with radionuclides such as ‘H, ‘*C and “‘1 which are incorporated into the nuclei of cells synthesizing DNA. In such cases, biological effects

LIMITS FOR INTAKES OF RADIONUCLIDBS BY WORKERS 15

may arise from transmutation, i.e. the chemical change of the nuclide together with its sudden change of electric charge and recoil, as well as from the ionization and excitation produced by the emitted radiations.

There seems to be no justification for modifying existing recommendations in radiological protection on the basis of effects from transmutation. With regard to cell sterilization, it is noted that 3H incorporated into thymidine is of similar efficiency to 3H in water, provided that the comparison is made on the basis of absorbed dose in cell nuclei. Although ‘H-thymidine has been demonstrated to produce mutations and cancer, it is thought that irradiation by 3H in this form is not appreciably more effective than external radiation. More work is needed to examine the possibility of mutation effects arising from incorporation of 3H and i4C into certain positions of the pyrimidine ring in DNA and the possible contribution to radiation damage of the Auger effect in 12’I. (See IAEA, 1968; Ertl er al., 1970; Feinendegen, 1975; Warters et al., 1975).

With regard to practical measures in radiological protection, it is worth noting that compounds such as ‘H and i’C labelled thymidine and [ ’ ’ *I]iododeoxyuridine are of particular concern after their direct entry to the blood stream, as might occur by accident at work. Even then their subsequent uptake into DNA is by no means complete (for [‘Hlthymidine, see Feinendegen and Cronkite, 1977 and Bond and Feinendegen, 1966). In the case of [‘251]- iododeoxyuridine, 90-95 % is rapidly catabolized (Hughes ef al., 1964) and the iodine generated behaves in a manner similar to that of iodine administered as iodide, as described in the dosimetric data for that element. When these compounds are ingested or inhaled a much smaller fraction will be available for uptake by DNA due to their catabolism to simpler forms, especially in the gastrointestinal tract (e.g. for [‘Hlthymidine, see Lambert and Clifton, 1968). Finally, it is noted that these compounds are not very volatile under normal circumstances and the probability of their being inhaled at work is small.

This matter will be kept under review. Meanwhile, cases in which the microdistribution of dose is thought to be important are referred to specifically in the metabolic data for the element concerned. For all other cases the committed dose equivalent H,, (Tc S) refers to the mean dose equivalent in a target tissue arising from activity in a source organ (see also para. 32, ZCRP Publication 26). The derivation of the quantities required to evaluate HsO (Tc S), namely SEE(Tt S) and U,, is now discussed.

4.5. Specitic Effective Energy (SEE)

In the dosimetric data for individual radionuclides given in the Supplement to this report, values are given for

SEE(7-c S) = 1 SEE(T+ S), 1

for a number of target and source organs. It is emphasized that the values shown refer only to the radionuclide concerned and do not include any contribution from daughter radionu- elides. Values for daughter radionuclides are given separately. For any radionuclide J SEE (Tt S), for target T and source S is given by

SEE(T+ S), = 7 ‘EiAFFc S)iQf MeV g- ’ per transformation 7

where the summation is over all radiations produced per transformation of radionuclide i in source organ S;

16 REPORT OF COMMI’ITEE 2

I’, is the yield of radiations of type i per transformation of radionuclide j;

E, (in MeV) is the average or unique energy of radiation i as appropriate;

AF(T t s), is the fraction of energy absorbed in target organ T per emission of radiation i in S. For most organs it is assumed that the energies from alpha particles and electrons are completely absorbed within the source organ. Notable exceptions are mineral bone (see Chapter 7) and the contents of the gastrointestinal tract (see Chapter 6). The absorbed fraction of energy from photons is estimated by the use of data on specific absorbed fraction (absorbed fraction per g of target) given in ICRP Publication 23. The absorbed fractions for fission neutrons have been obtained from data given by Dillman and Jones (1975), and Ford et al. (1977);

Q, is the quality factor appropriate for radiation of type i as discussed above in Sec- tion 4.3, and

M, (in g) is the mass of the target organ.

4.5.1. Decay schemes

Only the principal modes of decay are listed in the dosimetric data and these are given to provide some information as a basis for monitoring rather than as an adequate basis for dosimetry. The yields and energies of internal conversion electrons, of Auger electrons and x rays, and of beta and positron emissions have been estimated by the methods of Dillman (1979). Data on the decay schemes used in the calculations are to be published separately.

4.5.2. Masses of organs in the body

The masses of target organs used in this report are taken from ICRP Publication 23 and are listed in Table 4.1. Except for ovaries and uterus they apply to the 70-kg Reference Man.

Table 4.1. Masses of organs and tissues of Reference Man used in this Report

Source organs Mass (8) Target organs Mass (g)

Ovaries Testes Muscle Red marrow Lungs Thyroid ST content SI content ULI content LLI content Kidneys Liver - Pancreas Cortical bone Trabecular bone Skin Spleen Adrenals Bladder content Total body

:: 28 000

1500 1000

20 250 400 220 135 310

1800 100

4000

E 180

4 70000

Ovaries Testes Muscle Red marrow Lungs Thyroid Bone surface ST wall SI wall ULI wall LLI wall Kidneys Liver Pancreas Skin Sdeen Thymus Uterus Adrenals Bladder wall

11 35

28 000 1500 1000

20 120 I50 640 210 160 310

1800 100

2600 180 20 80 14 45


4.6. Number of Transformations in a Source Organ Over 50 Years

The number of transformations of a radionuclide in any organ or tissue of the body during any period of time is the time integral of activity of the radionuclide within that organ or tissue over the stated period of time, and in this report is referred to by the symbol U. The function describing uptake and retention of a radionuclide in a body tissue following its ingestion or inhalation may be very complex and therefore it is convenient to describe the transfer of radionuclides within the body by simple models which facilitate calculation and yet yield estimates of dose sufficiently accurate for the purposes of this report. With certain exceptions, e.g. for alkaline earth radionuclides in bone (see Chapter 7) the models used here are based on the assumption that the body consists of a number of separate compartments. Any organ or tissue may comprise one or a number of compartments. Loss of the radionuclide from any compartment is taken to be governed by first order kinetics. Therefore, the retention of an element in any organ or tissue will usually be described by either a single exponential term or the sum of a number of exponential terms, details of which are given in the metabolic data for individual elements. Exceptions are noted in the metabolic data.

The compartmental models used for the respiratory system and the gastrointestinal system are described in Chapters 5 and 6 respectively. After a radionuclide has been inhaled or ingested it will be translocated to the body fluids at a rate determined by the rate constants for the different compartments in the respiratory and gastrointestinal systems and by the radioactive decay constant of the radionuclide. Its translocation thereafter to the compartments representing the various organs and tissues of the body is shown schematically in Fig. 4.1. The finite time taken for translocation to the organs and tissues of deposition following entry of a radionuclide into the body fluids is represented in the model by transfer compartment a, which is assumed to be cleared by first order kinetics with a half life of 0.25 day, unless otherwise stated in the metabolic data for a particular element. Transformations occurring in the transfer compartment are assumed to be uniformly distributed throughout the whole body of mass 70 000 g. Each organ or tissue of deposition is assumed to consist of one or

From GI tract ond respiratory system

4%

Transfer . comportment

0 ‘\

\

\

r-

-\---1

Tissue Tissue Tissue I Tissue comportment comportment comportment I comportment I

b C d . I-

i I -r-_-I

I I

t

Excretion

Fig. 4.1. Mathematical model usually use.d to describe the kinetics of radionuclides in the body: exceptions to this model are noted in the metabolic data for individual elements


more compartments, and from each of these compartments the radionuclide is translocated at an appropriate rate to the excretion pathways. For simplicity, it is usually assumed that there is no feedback to the transfer compartment either from the routes of excretion or from the organ compartments, although it is recognized that transfer to body fluids happens in practice, and thus no estimate is usually made of dose along the routes of excretion. It should be noted that, because of the above assumption the amount of a radionuclide in transfer compartment a at any time after inhalation or ingestion cannot be used to estimate the amount of the radionuclide present in body fluids at that time.

From the above model the activity q(f) in any compartment at time t is derived using the following equations. In transfer compartment a,

In tissue compartment b,

(4.9b)

and so on, for any number of compartments. where t(r) is the rate of entry of activity of the radionuclide into body fluids at time r after

its inhalation or ingestion, and is calculated as described in Chapters 5 and 6;

A, is the clearance rate of stable isotopes of the element from transfer compartment a;

b, c, etc. are the fractions of stable isotopes of the element transferred from the body fluids to compartments b, c, etc. ;

A,, I,, etc. are the clearance rates of stable isotopes of the element from the compartments b, c, etc., and

A, is the radioactive decay constant of the radionuclide.

Values of b, c, etc.; &, A,, etc. can be derived from the metabolic data for individual elements. Any exceptions to this general method of deriving the activity in a compartment of the

body will be noted in the metabolic data for that element, e.g. iodine.

4.6.1. Build-up of radioactive daughters

In this report, H5v per unit intake, ALI, and DAC refer to the intake of the specified radionuclide alone. However, if the radionuclide has radioactive daughters, an allowance is made for the committed dose equivalent contributed by the build-up of daughters produced in the body from their parent. In general there is little evidence to indicate whether these daughters will remain associated with, and behave as, their parent, or whether, upon being produced, they will assume their own metabolic behaviour. When experimental evidence is available, e.g. concerning the behaviour of the noble gases radon and thoron when produced from their parents 22sRa and 224Ra in the body, it is given in the metabolic data. In all other cases it is assumed that daughters and all subsequent progeny produced in the body (i.e. including the respiratory and gastrointestinal systems), stay with and behave metabolically like the inhaled or ingested parent radionuclide. However, if evidence to the contrary becomes available, it is recommended that this information should be used to calculate revised ALIs for the radionuclide or mixture of radionuclides in question (see Chapter 9).


Using this assumption the activity of a radioactive daughter q’(t) in the transfer compartment a, or in any organ or tissue c0mpartment.b at any time t after intake of the parent radionuclide can be obtained using the following equations;

where f’(r) is the rate of entry of activity of the daughter radionuclide into the transfer compartment at time 1, this activity having resulted from radioactive decay of the parent radionuclide within the respiratory system or the GI tract as described in Chapters 5 and 6;

1; is the radioactive decay constant of the daughter radionuclide;

q.(t) is the activity of the parent radionuclide in the transfer compartment at time t as determined from eqn (4.9a);

d, is the biological rate of clearance from the transfer compartment of a daughter radionuclide produced in the body (assumed to have the same value as that for its parent), and

q:(r) is the activity of the daughter radionuclide in the transfer compartment at time i.

where q;(t) is the activity of the daughter in tissue compartment b at time t;

b is the fraction of the daughter translocated from the transfer compartment to compartment b; it is assumed that this fraction is the same as that for the stable isotope of

the parent radionuclide;

I, and I, are the biological rates of clearance of the daughter radionuclide from the transfer compartment and from compartment b respectively (assumed to be the same as those for the parent radionuclide);

Xa is the radioactive decay constant of the daughter radionuclide;

q:(z) is the activity of the daughter in the transfer compartment at time t, and

qb(t) is the activity of the parent radionuclide in tissue compartment b at time z.

In a similar manner, a system of equations can be derived that describe the activities of a chain of parent and daughter radionuclides, the activity of each daughter being determined by the activity of its predecessor in the chain. The metabolic behaviour of all the radioactive progeny is assumed to be the same as that of the ancestral radionuclide which was taken into the body. These equations, together with those given in Chapters 5 and 6 for the respiratory system and GI tract, completely specify the models usually used in this report to calculate values of the number of transformations, US, in any source organ within the body. Details of the calculations are not given here and the reader is referred to Appendix 1 for a brief discussion of methods which may be employed to estimate these quantities. In the dosimetric data for any radionuclide, values of Us are given for that radionuclide together with values of Vi, U$, etc. for its daughter radionuclides which have built up in the body in the 50 years following ingestion or inhalation of unit activity of their parent.


4.7. Annual Limit on Intake (ALI)

In the dosimetric data for each radionuclide j, values are given for H,, per unit activity ingested or inhaled for a number of organs and tissues irradiated to a significant extent. Corresponding values are given for H,,, per unit intake weighted for the sensitivity of the individual organ or tissue to radiation-induced cancer or hereditary disease using the weighting factors wr given in Chapter 2. Values are also given for ALI by inhalation and by ingestion of a radionuclide in various chemical forms calculated as described in Chapter 3, i.e. ALI, is the greatest value of Ij which satisfies both of the following inequalities

1, I$wAH ),-,r per unit intake) < 0.05 Sv (4.11a)

for stochastic effects and

I,(H,,,, per unit intake) < 0.5 Sv (4.11b)

for non-stochastic effects. In relationship (4.11 a), the summation is for all those tissues specified in Table 2.1, Chapter 2 and the five other tissues in the remainder which receive the greatest committed dose equivalents. A tissue T is deemed to be irradiated to a significant extent if for that tissue the weighted value of H,, per unit intake is greater than 10% of the maximum weighted value of HS, per unit intake in any tissue,

i.e. wTH50,T per unit intake 2 O.l(w,H,,,, per unit intake),,, (4.12)

For any radionuclide only those target tissues that satisfy relationship (4.12) will be listed. Values of SEE(T t S) will be given for every source organ that contributes 1% or more to the committed dose equivalent in a target tissue which satisfies inequality (4.12).

It is noteworthy that intakes of radioactive materials into the body are limited by relationship (4.1 la) unless relationship (4.11 b) overrides because of the possibility of non-stochastic effects in bone surfaces, thyroid or any one of the five tissues in the remainder. This follows as a consequence of the small weighting factors. (wr I 0.06), assigned to these tissues. When the ALI is limited by consideration of non-stochastic effects, the organ or tissue concerned is named below the value of ALI. In all such cases the greatest annual intake that meets the Commission’s recommendation for limiting stochastic effects (relationship (4.11 a)) will be given in parentheses between the ALI and the tissue or organ name. This value is useful when considering the limitation of exposure from several sources (see Chapter 9) and also when application of the Commission’s system of dose limitation (section E, ICRP Publication 26) requires that a worker receives only a fraction of the dose-equivalent limit for stochastic effects. Values of AL1 by ingestion are given for each appropriate value off,, the fraction of an ingested compound of the element which is absorbed into blood (see Chapter 6). Values of AL1 by inhalation are given for each inhalation class D, W or Y (see Chapter 5) appropriate to various chemical compounds of the element concerned.

4.8. Derived Air Concentration (DAC)

As described in Chapter 3, the derived air concentration (DAC) for a radionuclide is related to the appropriate ALI by inhalation as follows

DAC, = ALlj/2.4 X lo3 Bq me3 (4.13)

LIMITS FOR INTAKES OF RADIONUCLIDES BY WORKER8 21

All the values given for DAC and AL1 by inhalation are calculated for an aerosol of Activity Median Aerodynamic Diameter (AMAD) of 1 pm. However, values are given in the dosimetric data for the percentage of the committed dose equivalent arising from deposition in the nasal- passage (NP) the trachea-bronchi (TB) and the pulmonary (P) regions of the respiratory system to enable an appropriate correction to be made for aerosols of different AMAD (see Chapter 5).

4.9. Values in the Dosimetrlc Data

Values of ALI and DAC collected in pp. 63-115 of this report are given to one significant figure; in view of uncertainties concerning the metabolism and effects of radionuclides in the body, greater precision is hardly justified. Values of SEE, Us and H,, per unit intake collected in the Supplement to this report are given to two significant figures to reduce the errors of repeated multiplication.

It is emphasized that all the values of U,, H,,, AL1 and DAC relate to intake by the specified route of entry into the body of the single radionuclide named. They include an appropriate allowance for the activity of any daughter radionuclides produced in the body as a result of intake of the parent. Intakes which include both parent and daughter radionuclides should be treated by the general methods appropriate to mixtures described in Chapter 9.

The values of AL1 do not apply directly when the worker both ingests and inhales the radionuclide nor when he is (in addition) exposed to external radiation (Chapter 9).

All information in the metabolic and dosimetric data refer to Reference Man (ZCRP Publi- cation 23) and is intended to be used for planning occupational exposure to radioactive materials. The data must be used with circumspection for individuals and they will not be valid for younger members of a population (see Chapter 9).

References Bond, V. P. and Feinendegen, L. E. (1966). Intranuclear 3H thymidine; dosimetric, radiobiological and radia-

tion protection aspects. Health Phys. 12, 1007-1020. Dillman, L. T. and Jones, T. D. (1975). Internal dosimetry of spontaneously fissioning nuclides. Health Phys.

29, 11 I-123. Dillman, L. T. (1979). EDISTR-A computer program to obtain a nuclear decay data base for radiation dosi-

metry. ORNL/TM-6689. Ertl, H. H., Feinendegen, L. E. and Heiniger. H. J. (1970). Iodine-125, a tracer in cell biology: physical proper-

ties and biological aspects. Phys. Med. Biof. 15, 447-456. Feinendegen, L. E. (1975). Biological damage from the Auger effect, possible benefits. Radiat. Environ. Biophys.

12, 85-100. Feinendegen, L. E. and Cronkite, E. P. (1977). Effects of microdistribution of radionuclides on recommended

limits in radiation protection, a model. Curr. Top. Radiat. Res. Q. 12, 83-99. Ford, M. R., Snyder, W. S., Dillman, L. T. and Watson, S. B. (1977). Maximum permissible concentration

(MPC) values for spontaneously fissioning radionuclides. Health fhys. 33, 35-43. Hughes, W. L., Commerford, S. L., Gitlin, D., Krueger, R. C., Schultze, B., Shah, V. and Reilly, P. (1964).

Deoxyribonucleic acid metabolism in uiuo: I. Cell proliferation and death as measured by incorporation and elimination of iododeoxyuridine. Fed. Proc. 23, 640448.

International Atomic Energy Agency, Biological Effects of Transmutation and Decay of Incorporated Radio- isotopes. IAEA, Vienna, 1968, pp. 233-241.

ICRP Publication 21, Data for Protection Against Ionizing Radiation from External Sources. Pergamon Press, Oxford, 1973.

ICRP Publication 23, Report of the Task Group on Reference Man. Pergamon Press, Oxford, 1975. ICRP Publication 26, Recommendations of the International Commission on Radiological Protection. Pergamon

Press, Oxford, 1977. ICRU Report 19, Radiation Quantities and Units. ICRU, Washington, 1971.. ICRU Supplement to Report 19, Dose Equivalent. ICRU, Washington, 1973.

22 REPORT OF COMMIlTEE 2

Lamb-t, B. E. and Clifton, R. J. (1968). Radiation doses resulting from the ingestion of tritiated thymidine by the rat. Health Phys. IS, 3-9.

Warters, R. L., Hofer, K. 0.. Harris, C. R. and Smith, J. M. (1977). Radionuclide toxicity in cultured mammalian cells: elucidation of the primary site of radiation damage. Curr. Top. Radiur. Res. Q. 12, 389-407.


5. DOSIMETRIC MODEL FOR THE RESPIRATORY SYSTEM 5.1. Introduction

The dosimetric model developed in this Chapter relates to the inhalation of radioactive aerosols. The special cases of gas and free ion deposition will receive further consideration. Some special cases will be referred to in the metabolic data for individual elements and also Chapter 8.

When radioactive aerosols are inhaled, parts of the respiratory system are irradiated. Other organs and tissues of the body are also irradiated both by radiations originating from the lungs and as a result of translocation of inhaled material to body tissues from the respiratory system.

It is recognized that after the inhalation of radioactive aerosols the doses received by various regions of the respiratory system will differ widely, depending on the size distribution of the inhaled material. In this report the dose received by the nasopharyngeal region has been neglected since for most particle sizes it is usually small in comparison with the doses received by other regions. The dose delivered to some of the pulmonary lymph nodes by insoluble particles cleared from the pulmonary region may be many times greater than that received by the lung tissues. However, from considerations of the possible effects in the whole lymphoid tissue, of which the pulmonary lymph nodes form only a part, and of autopsy data on men who had inhaled particles of plutonium, the Commission has decided that irradiation of the lung is likely to be more limiting than that of lymphoid tissue for inhaled, insoluble, radioactive particles (paras. 52 and 53, ICRP Publication 26). The distribution of dose to cells in the lung from inhaled particles may also be very inhomogeneous. For the induction of malignant disease, the Commission considers that the hazard of radioactive particles in the lung is likely to be less than that of the same amount of material distributed uniformly in the lung (paras. 49 and 50, ZCRP Publication 26). Although in principle the mode1 described here could be used to estimate dose equivalent in the different regions of the respiratory system, and this might be useful for research, in view of other uncertainties, e.g. concerning the precise location of the cells at risk, it is not considered that such estimates are yet warranted for fixing values of ALI. Therefore, for the purposes of radiological protection, the Commission considers that in adults it will be satisfactory to consider the tracheobronchial region, pulmonary region and pulmonary lymph nodes as one composite organ of mass 1000 g (para. 54, ICRP Publicu- tion 26) to which the weighting factor for lung given in Table 2.1 applies.

As explained in previous chapters, the exposure of workers to radioactive materials is limited by consideration of the committed dose equivalents HSO in the tissues of the body. The method used to calculate HS,, has been described in Chapter 4 where it is explained that H,, for any target organ depends on Us, the total number of transformations in a source organ over the 50 years following intake of the radionuclide, and SEE(Tt S’), the specific effective energy absorbed in a target organ from each type of radiation arising in a source organ. The methods used to calculate Us and SEE(Tc S) for parts of the respiratory system, and to calculate HsO for the composite lung tissue defined above, are described below.

5.2. Deposition and Retention Model

Estimates of the distribution and retention of material in the respiratory system are based on a mode1 proposed in the report of the ICRP Task Group on Lung Dynamics (1966). This

24 REPORT OF COMMITTRR 2

model takes account of particle size and also defines three classes of retention which, in pad, reflect the chemical form of the aerosol. Additionally the model provides information on the various routes of elimination from the lungs. Although further consideration by the Task Group and Committee 2 since publication of the report on lung dynamics has resulted in many changes in values of the deposition and clearance parameters, the model is substantially that of the Task Group (see also ZCRP Publication 19).

The respiratory system is divided into three distinct regions-the nasal passage (N-P), the trachea and bronchial tree (T-B) and the pulmonary parenchyma (P). Deposition is assumed to vary with the aerodynamic properties of the aerosol distribution and is described by the three parameters DN_,., D,_,,, and Dp which represent the fractions of inhaled material initially deposited in the N-P, T-B, and P regions, respectively. For a log-normal distribution of diameters, which seems typical of aerosols, the pattern of deposition can be related to the activity median aerodynamic diameter (AMAD) of the aerosol (Fig. 5.1). In this report calculations of committed dose equivalent are for an aerosol with an AMAD of 1 pm. Estimates for other AMADs can be made using the data in Fig. 5.1 as discussed in Section 5.5 below.

To describe the clearance of inhaled radioactive materials from the lung, materials are classified as D, W or Y which refer to their retention in the pulmonary region. This classification applies to a range of half-times for D of less than 10 days, for W from 10 to 100 days and for Y greater than 100 days. The three regions, N-P, T-B and P described above are each divided into two or four compartments as shown in Fig. 5.2. Each of these compartments is

O.‘i ,.,\I ,,,\1,, I I

5 ‘IO 20 30 50 70 80 90 95 99

Percent deposltaon

Fig. 5.1. Deposition of dust in the respiratory system. The percentage of activity or mass of an aerosol which is deposited in the N-P, T-B and P regions is given in relation to the Activity Median Aerodynamic Diameter (AMAD) of the aerosol distribution. The model is intended for use with aerosol distributions with AMADs between 0.2 and 10 pm and wtth geometric standard deviations of less than 4.5. Provisional estimates of deposition further extending the size range are given by the dashed lines. For an unusual distribution with an AMAD of greater than 20 pm. complete deposition in N-P can be assumed. The model does not apply to aerosols with

AMADs of less than 0.1 pm.


Class D N-P

D W Y

Cornpart- T T T Region ment day F day F day F

N-P a 0.01 0.5 0.01 0.1 0.01 0.01 (D- = 0.30) b 0.01 0.5 0.40 0.9 0.40 0.99

T-B C 0.01 0.95 0.01 0.5 0.01 0.01 (&_a= 0.08) d 0.2 0.05 0.2 0.5 0.2 0.99

e 0.5 0.8 50 0.15 500 0.05

;& = 0.25) f n.a. n.a. ; z ::2a*

50 1.0 0.4 0.4 500 1.0 0.4 0.4 50 0.05 500 0.15

L i 0.5 1.0 50 1.0 1000 0.9 j n.a. n.a. n.a. n.a. aJ 0.1 Lymph

nodes

Fig. 5.2. Mathematical model used to describe clearance from the respiratory system. The values for the removal half-times, T.-r and compartmental fractions, F.-I are given in the tabular portion of the figure for each of the three classes of retained materials. The values given for DN+, DT_B and Dp (left column) are the regional depositions for an aerosol with an AMAD of 1 pm. The schematic drawing identifies the various clearance pathways from compartments a-i in the four respiratory regions, N-P, T-B, P and L.

n.a. = not applicable.

associated with a particular pathway of clearance for which the half-time of clearance is T days and the fraction leaving the region by that rate is F. Thus, compartments a, c and e are associated with absorption processes, whereas b, d, f and g are associated with particle transport processes, including mucociliary transport, which translocate material to the gastrointestinal tract. The pulmonary lymphatic system (L) also serves to remove dust from the lungs. It is associated with compartment h in the P region of the lungs from which material is translocated to compartments i and j in the pulmonary lymph nodes. Material in compartment i is translocated to body fluids but that in compartment j is assumed to be retained there indefinitely. This compartment is only considered appropriate for class Y aerosols; for class D and W aerosols the fraction of material entering compartment j from compartment h (F,) is set equal to zero. The clearance of material from each of the compartments described above is assumed to be governed by first order kinetics so that each compartment is associated with a clearance constant 1 and half-time of clearance T = 0.693/86 400 1. The clearance of inhaled material from the lung is, therefore, described by a set of interlinked first order differential equations, as follows

$ CL(~) = &I. DIG,. F, - ~,q,(O - &q.(t) (5.la)

; a,(f) = &>. D,-, . F, - &qdf) - &q,,(t) (5.lb)

$ qc(l) = &>. h-a. F, - bL(O - &L(f) (UC)

; q,i(O = &>. &-a. Fc, + h0) + &(O - &q&) - &q,,(r) (5. Id)

26 REF’ORT OF COMWl7El3 2

(5. le)

(S.lh)

g qX9 = ~JhdO - MO - hdr) (S.li)

where q.(r), q,,(r), etc. are the activities of an inhaled radionuclide in compartment a, b, etc. at time r;

t(r) is the rate of inhalation of activity of the radionuclide;

4 to 1, are the biological clearance rates of compartments a to i;

1, is the radioactive decay constant of the radionuclide, and

F, to Fj are the fractions of material entering the various regions of the lung that are associated with the various compartments contained therein.

Values of I;; to Fj are given in Fig. 5.2 for.the various inhalation classes D, W and Y. Also given are T, to Ti for these various classes. & to 1, are obtained from the values of Ti to T, by use of the formula 1= 0.693/86 4OOT,

where 1 is in s”;

T is in ‘days, and

86400 is the number of seconds in a day.

The classification D, W or Y for different chemical forms of a radionuclide are given in the metabolic data for the individual elements. As stated in Chapter 4 it is assumed that radioactive daughters remain with and behave metabolically like the inhaled parent radionuclide. In fact, there is a paucity of evidence about the behaviour of radioactive daughters produced within the lung. However, it is at least plausible to assume that daughters produced within a radioactive particle stay within the particle. The build-up and clearance of a radioactive daughter produced in the different compartments of the lung is, therefore, governed by a series of first order differential equations similar to those given in eqn (5.1). If q’(r) is the activity of the radioactive daughter in any compartment at time r and q(r) is the activity of its immediate precursor, then

& 4:(t) = G.(t) - &,4X0 - %4:(r) (5.2a)


(5.2b)

; 4x0 = Ghw + MO) + +?;w - bm - Gdlw (5.2d)

g q:to = &4&O - b?:(t) - &q:(r)

$ 4x0 = n;qXO - Mtt) - &q;(t)

$ 4;to = 1xq,to - +?;w - &q;(t)

(5.2~)

(5Jf)

(5.2h)

2 qXt) = nitqi(r) + Finhdt(z) - 4idr) - nitqi(r) (5.2i)

(54)

where XX is the radioactive decay constant of a daughter radionuclide, and

Fi and Fj are the values appropriate to the class of inhaled parent radionuclide.

Similarly, a system of equations can be derived that describe the activities of a chain of parent and daughter radionuclides, the activity of each daughter being determined by its predecessor in the chain. The metabolic behaviour of all the radioactive progeny is assumed to be the same as that of the ancestral radionuclide which was inhaled.

5.3. Transfer of a Radionuclide from the Lungs Directly to Body Fluids or to the Gastrointestinal Tract

From Fig. 5.2 and the above equations (5.la to 5.lj and 5.2a to 5.2j) it can easily be seen that the rate of transfer of a radionuclide directly from the lungs to body fluids(BF(t)), is given by

BFW = k&) + k(r) + &q,(t) + k,(r) (5.3)

whether the radionuclide is inhaled or produced in the lungs. Similarly the rate of transfer of activity of a radionuclide from the lungs to the gastrointestinal tract (G(t)), is given by

G(t) = k,(t) + &q&) (5.4)

whether the radionuclide is inhaled or produced in the lungs. The metabolism of a radionu- elide after reaching either body fluids or the gastrointestinal tract following inhalation of its parent is usually assumed to be governed by the metabolic model for its inhaled parent (see Chapter 4).


The equations above completely specify the lung model used in this report and have been used to calculate the number of transformations (U) in the lung and other organs given in the dosimetric data for the various radionuclides which are collected in the Supplement to this report. Details of these calculations are not given here and the reader is referred to Appendix 1 for a brief discussion of methods that may be employed in estimating these quantities.

5.4. Calculation of Committed Dose Equivalent, Hso, in the Lung

As described in Chapter 4, H50,T for a target organ T from a number of source organs S containing any mixture of radionuclides j is given by

H SO,r = 1.6 x 10-i’ 7 $ us T SEE(T+ S),] Sv (5.5)

where Us is the number of transformations in the source organ S over 50 years following the intake of radionuclide j, and

SEE(Tc S)i (in MeV g- ’ transformation- ’ ) is the specific effective energy absorbed in T from radiations of type i emitted in S.

The methods by which Us can be calculated for the various parts of the respiratory system have been described in Section 5.2 both for the parent radionuclide and for its radioactive daughters. SEE(T+- S)i for any radiation of type i emitted by radionuclide j is given by

SEE(Tt S), = Y,EiAF( T+ S)iQi

M MeV g - ’ per transformation T

(5.6)

where Yi is the yield of radiations of type i per transformation of radionuclide j;

Ei (in MeV) is the average, or unique, energy of radiation i;

AF(Tt S)i is the average fraction of energy absorbed in target organ T per emission of radiation i in source organ S;

Qi is the quality factor appropriate for radiation i, as discussed in Chapter 4, and

Mr (in g) is the mass of the target organ.

As explained in Section 5.1, dose in the N-P region is neglected and the target tissue T in eqn (5.5) is the lung comprising the T-B, P and L regions of total mass 1 000 g. HS O,T has two components-the committed dose equivalent from radioactive material in the lung, (i;e. in target tissue T comprising T-B, P and L) and the committed dose equivalent from photons arising in other organs and tissues of the body, S.

H 50,r = 1.6x 10-i’ T U/T SEE(T+ T), +c s

In each case the summation is for all radionuclides,

7 Vi 7 SEE(T+ S)i 1 Sv (5.7)

including those daughter radionuclides which build up after inhalation of their parent. Values of U{ (= Vi_, + U{ + Vi) and corresponding values for daughters are given in the dosimetric data for the parent radionuclide j. If they contribute significantly to dose, values for Vi and corresponding values for daughters are also given in the dosimetric data for radionuclide j.

Values pf SEE for any radionuclide j are derived from eqn (5.6) in which MT is 1 000 g and AF(T c S), takes the following values:


(a) For SEE@+ T)i, i.e. for radioactive material in the lung itself, AF(Tt T), = 1 for all radiations other than photons and neutrons. For photons and fission neutrons values

(‘4

of AF(T t T)i /M,, the specific absorbed fraction, are either given in ZCRP Publication 23 or derived from data given by Dillman and Jones (1975) and Ford et al. (1977).

For SEE(T+ S)i, i.e. for radioactive materials in organs and tissues S outside the respiratory system, AF(T+ S)i = 0 for all radiations other than photons and neutrons. For photons and fission neutrons values of AF(T + S)JMT, the specific absorbed fraction, are given either in ZCRP Publication 23 or derived from data given by Dillman and Jones (1975) and Ford et al. (1977).

5.5. Particle Size Correction

Values of HSo per unit intake, annual limit on intake (ALI) and derived air concentration (DAC) given in the dosimetric data collected in the Supplement to this report for individual radionuclides are for a radionuclide with an AMAD of 1 pm. Together with values for Hsb, values are given in parentheses for the fraction&, fi_B andf, of the committed dose equivalent in the reference tissue resulting from deposition in the N-P, T-B and P regions respectively. HSo for an aerosol of AMAD other than 1 pm may then be estimated by

&o(A+W H&l pm) =

f D,_p(AMAD) DdAMAD)

N-P DN-P(1 pm) + fT_B D,-*(AMAD)

DT-~(1 elm) +.fp

DPU Irm) (5.8)

where DN_P, DT_B and Dp are the deposition probabilities in the respiratory regions for a given AMAD (Fig. 5.1).

If AMAD is unknown, it is recommended that values of Hso, AL1 and DAC given in the dosimetric data for an aerosol of 1 pm AMAD be used.

References Dillman, L. T. and Jones, T. D. (1975). Internal dosimetry of spontaneously fissioning nuclides. Health Phys.

29, 111-123. Ford, M. R., Snyder, W. S.. Dillman, L. T. and Watson, S. B. (1977). Maximum permissible concentration

(MPC) values for spontaneously fissioning radionuclides. Health Phys. 33, 3543. ICRP Task Group on Lung Dynamics (1966). Deposition and retention models for internal dosimetry of the

human respiratory tract. Health Phys. 12, 173-207. ICRP Publication 19, The Metabolism of Compounds of Plutonium and other Actinides. Pergamon Press, Oxford,

1972. ICRP Publication 23, Tusk Group Report on Reference Mun. Pergamon Press, Oxford, 1975. ICRP Publication 26, Recommendations of the International Commission on Radiological Protection. Pergamon

Press, Oxford, 1977.

30 REPORT OF COMMITTEX 2

6. DOSJMETRIC MODEL ~‘$HE GASTROINTE!3TINAL

6.1. Introduction

As explained in previous chapters of this report, exposure of workers to a radioactive material is limited by consideration of the committed dose equivalent H5, received by organs and tissues of the body following intake of the radionuclide concerned. The general method used to calculate HSO has been discussed in Chapter 4, where it was explained that H,,,, in a target organ T depends on U,, the number of transformations in source organ S over the 50 years following intake of the radionuclide and SEE(T+ s), the specific effective energy absorbed in target organ T from radiations originating in source organ S. The methods used to calculate U,, SEE(T + s) and HSo,s for sections of the gastrointestinal (GI) tract following ingestion of a radionuclide are discussed below. These sections of the GI tract are treated as separate target tissues (para. 105, ICRP Publicaim 26). Organs and tissues of the body other than the GI tract will also be irradiated by that fraction of the radionuclide which enters the body fluids and also, in some cases, by photon irradiation from material retained in the GI tract. The general methods used to calculate H,O,T following entry of a radionuclide into body fluids have been described in Chapter 4. The method used to calculate the activity of an ingested radionuclide or its radioactive daughters, transferred to body fluids from the GI tract, is discussed below.

6.2. DosImetrIc Model

The dosimetric model is based on the biological model developed by Eve (1966). For the purposes of radiological protection the GI tract is taken to consist of the 4 sections shown schematically in Fig. 6.1. (See p. 33).

Each of these sections is considered as a single compartment and translocation from one compartment to the next is taken to be governed by first order kinetics. Thus, if q(t) is the activity of ingested radionuclide in a compartment at time r then the model is completely described by the following equations

(6.lb)

d ;ii q”LI(r) = - ~“L14”LXf) - &&“LIW + &s,(r)

d z %LlW = -&ILLxr) - &lLLlW + ~“Llq”LI(r) (6. Id)

where I, is the radioactive decay constant of the radionuclide in question;

lBqs,(t) is the rate of transfer of activity to body fluids from the small intestine, assumed to be the only site of absorption from the GI tract to body fluids, and

j(t) is the rate of ingestion of activity of the radionuclide at time t.


Similarly, the model for a radioactive daughter produced in the GI tract from its ingested parent radionuclide is completely described by

= -&*qhw - &&lo) + 1XqsdO (6.2a)

(6.2b)

(6.2~)

where & is the radioactive decay constant of the radionuclide in question;

q’(t) is its activity in any compartment and q(r) is the activity of its immediate parent.

The value of 1, can be estimated fromf,, the fraction of a stable element reaching the body fluids following ingestion :

1, - =f, &sl + lB

B

Values off’ are given in the metabolic data for a number of classes of compounds of each individual element. For radioactive daughters, produced by decay of their parents in the GI tract, the value offi used is usually that appropriate to the stable element of which the ingested radionuclide is an isotope (see Chapter 4). Where a valuef, = 1 is given it is assumed that the radionuclide passes directly from the stomach to body fluids and does not pass through other sections of the gastrointestinal tract.

A system of equations similar to (6.1) and (6.2) can be derived that describe the activities of a chain of parent and daughter radionuclides, the activity of each daughter being determined by the activity of its predecessor in the chain. The metabolic behaviour of all the radioactive progeny is assumed to be the same as that of the ancestral radionuclide which was ingested.

6.3. Activity Transferred from the Respiratory System

The activity of an inhaled radionuclide in any section of the GI tract is described by eqns (6.la)-(6.ld) in which the term i(t) in eqn (6.la) is the rate of entry of activity of the radionuclide to the GI tract from the respiratory system (Chapter 5). The activity of a daughter radionuclide in any section of the GI tract is described by eqns (6.2a)-(6.2d) with the addition of a term, +1’(r), to the right hand side of eqn (6.2a). i’(r) is the rate of entry of activity of the daughter radionuclide to the GI tract, the daughter having been produced by decay of its parent in the respiratory system. The value ofj; for any daughter produced in the respiratory system is usually assumed to be the same as that of the parent radionuclide (Chapter 4).


6.4. Activity Transferred to Body Fluids

The activity transferred to body fluids is the integral over 50 years of Aeqs,(i) for the parent and &q;,(r) for its daughter. Their evaluation is described in the Appendix.

6.5. Calculations of Committed Dose Equivalent, H,,,, to Sections of the Gastrointestinal Tract

As described in Chapter 4, H,,,, for any target organ T from a number of source organs S containing a mixture of radionuclides j is given by

H 50.T = 1.6~10-‘~; T W/$SEE(TcS), IJ

Sv (6.4)

where Us is the total number of transformations in a source organ S over 50 years following intake of a radionuclide j, and

SEE(T t s)i (in MeV g- 1 per transformation) is the specific effective energy’from radiations of type i originating in S.

When the source organ is a section of the GI tract, U is the integral over 50 years of appropriate values of q(t) from eqn (6.1) and (6.2). Methods of calculating U for a parent radionuclide and its daughters in a section of the GI tract following ingestion of unit activity of the parent radionuclide are discussed further in the Appendix.

SEE(Tt S),, for any radionuclide j, is given by

SEE(Tc S), = YIEiAF(T+ S),QI

MT MeV g- ’ per transformation

where Y, is the yield of radiations of type i per transformation of radionuclide j;

El (in MeV) is the average, or unique, energy of radiation i;

AF(Tt S), is the average fraction of energy absorbed in T from radiation i arising in

S;

Q, is the quality factor appropriate for radiation i, and

MT (in .g) is the mass of the target organ.

Hso,T is estimated for the walls of each section of the GI tract and it is convenient to consider values of SEE for non-penetrating (np) and penetrating (p) radiations separately. Non- penetrating radiations are, in this context, taken to be heavy recoil particles, fission fragments, a-particles and B-like radiations; penetrating radiations are taken to be x rays, y rays and fission neutrons.

For both p and np radiations HS0.T is calculated for the mucosal layer of each section of the GI tract. For p radiations the average dose to the walls of the tract is used as a measure of the dose to the mucosal layer. Thus, the value of SEE for any target section of the GI tract and for all emissions from radionuclidej within the tract is given by

SEE=z Kp~,pQ,pAW=+ T)np ML

nP MT

'F; YpEpQpAFW'*S)p

MT" MeV g- ’ per transformation


where Y,,r is the yield of a non-penetrating radiation per transformation of the radionuclide;

EnP (in MeV) is the average or unique energy of the non-penetrating radiation;

e np is 20 for recoil atoms, fission fragments and a-particles and 1 for electrons as described in Chapter 4.

-(ML + J-In,, MFL

is the specific absorbed fraction for the mucosal layer of the section. of the GI tract under consideration and is taken to be equal to

where A4; (in g) is the mass of the contents of that section of the GI tract (Fig. 6.1) and v is a factor between 0 and 1 representing the degree to which these radiations penetrate the mucus. The factor + is introduced because the dose at the surface of the contents will be approximately half that within their volume for np radiations. v is taken to be unity for fi particles, zero for recoil atoms and 0.01 for a particles and fission

Ingestion

Stomach (ST)

Upper Large Intestine (ULI)

A ULI

Lower Large

Intestine (LL 1)

A

@ LLI

Excrerlon

Section of GI tract

Mass of Mass of walls* contents*

(9, (f3)

Mean residence time x (day) day-’

Stomach (ST) 150 250 l/24 24 Small Intestine (SI) 640 400 4124 6 Upper Large Intestine (ULI) 210 220 13124 1.8 Lower Large Intestine (LLI) 160 135 24124 1

l From ICRP Publication 23 (1975).

Fig. 6.1. Mathematical model used to describe the kinetics of radionuclides in the gastrointestinal tract

34 REPORTOF'COMMITTEE 2

fragments. Although this latter factor is clearly arbitrary there is experimental evidence that its use is warranted (Sullivan el al., 1960).

MT” (in g) is the mass of the walls of the target section concerned (Fig. 6.1)

YP is the yield of penetrating radiations per transformation of the radionuclide.

Ep (in MeV) is the average or unique energy of the penetrating radiation.

QP is 1 for photons and 10 for fission neutrons;

AF(W t s), is the fraction of the energy of a photon or fission neutron emitted in source organ S which is absorbed in the walls of the target section T concerned.

Values of AF(W + &/My, the specific absorbed fraction for photons, are given in ZCRP Publication 23. Corresponding values for fission neutrons have been derived from data by Dillman and Jones (1975) and Ford et al. (1977).

For the general case of a number of radionuclides, including the case of an ingested parent together with its daughters produced in the tract, HS,, is derived by the summation of U, and SEE over all radionuclides and source organs, as described previously in Chapter 4. In the dosimetric data for each radionuclide collected in the Supplement to this report values of Us are given for the ingested parent and for its radioactive daughters produced in the various sections of the GI tract. SEE values are also given for the various sections of the GI tract as target organs with the contents of these sections and other organs and tissues of the body as source organs.

References Dillman, L. T. and Jones, T. D. (1975). Internal dosimetry of spontaneously fissioning nuclides. Health Phys.

29, 11 l-123. Eve, I. S. (1966). A review of the physiology of the gastrointestinal tract in relation to radiation doses from

radioactive materials. Hea/th Phys. 12, 131461. See also Dolphin, G. W. and Eve, I. S. (1966). Dosimetry of the gastrointestinal tract. Health Phys. 12, 163- 172.

Ford, M. R., Snyder, W. S., Dillman, L. T. and Watson, S. B. (1977). Maximum permissible concentration (MPC) values for spontaneously fissioning radionuclides. Health Phys. 33, 35-43.

ICRP Publication 23. Task Group Report on Reference Man. Rrgamon Press, Oxford, 1975. ZCRP Publication 26, Recommendations of the International Commission on Radiological Protection. Pergamon

Press, Oxford, 1977. Sullivan, M. F., Hackett, P. L., George, L. A. and Thompson, R. C. (1960). Irradiation of the intestine by

radioisotopes. Radiat. Res. 13, 343-355.


7. DOSIMETRIC MODEL FOR BONE

7.1. Introduction

The cells at carcinogenic risk in the skeleton have been identified as the haematopoietic stem cells of marrow, and among the osteogenic cells, particularly those on endosteai surfaces, and certain epithelial cells close to bone surfaces (ICRP Publication II). The haematopoietic stem cells in adults are assumed to be randomly distributed predominantly throughout the haematopoietic marrow within trabecular bone (ICRP Publication 21). Therefore, dose equivalent to those cells is calculated as the average over the tissue which entirely fills the cavities within trabecular bone. For the osteogenic tissue on endosteal surfaces and epithelium on bone surfaces the Commission recommends that dose equivalent should be calculated as an average over tissue up to a distance of 10 pm from the relevant bone surfaces (para. 47, ZCRP Publication 26).

In order to estimate committed dose equivalent in these tissues from an intake of a particular radionuclide, it is necessary to know the distribution and retention of the radionuclide in the various tissues of the skeleton. At present such data are only available for a few radionuclides and it is hoped that further research in this field will be encouraged. Meanwhile, for purposes of radiological protection, the methods described below are considered to be sufficiently accurate to estimate the committed dose equivalent to cells on bone surfaces and to active red bone marrow from intakes of all bone-seeking radionuclides.

7.2. Calculation of Committed Dose Equivalent, Hso, to Cells on Bone Surfaces and Active Red Bone Marrow

As explained in Chapter 4, If,, in a target organ T is given by

H - 1.6~ lo-” C C S0.T - SV S i 1 i (7.1)

where U, is the number of transformations of nuclide j occurring in source organ S, and

SEE(Tt S). = YIEiAF(Tc S)iQi MeV g-’ transformation-’ I MT

(7.2)

where Yi is the yield of radiation of type i per transformation;

E, (in MeV) is the average or unique energy of radiation i as appropriate;

AF(rt S)i is the fraction of energy absorbed in target organ T from radiation i originating in S;

Qi is the quality factor appropriate for radiation of type i, and

MT (in g) is the mass of the target organ.

In the case of bone the two target tissues are the cells near bone surfaces (BS) and the active red bone marrow (RM) as described above. For all radionuclides, except y emitters, the source tissues will normally be cortical and trabecular bone.

Bone dosimetry, therefore, involves estimates of Us for trabecular and cortical bone and estimates of AF(Tt S)i for all radiations, radionuclides, source organs and target tissues.


From metabolic and dosimetric considerations radionuclides and their emissions are divided into the following six broad classes

(a) photons from all radionuclides;

(b) a particles from radionuclides assumed to be uniformly distributed throughout the volume of bone;

(c) a particles from radionuclides assumed to be on bone surfaces;

(d) /3 particles from radionuclides assumed to be uniformly distributed throughout the volume of bone;

(e) fl particles from radionuclides assumed to be on bone surfaces and having an average /3 energy (E,,) greater than 0.2 MeV, and

(f) /? particles from radionuclides assumed to be on bone surfaces for which E,, is less than 0.2 MeV.

It should be noted that a radionuclide can belong to more than one class; for example, it may be a B emitter with a y-decay mode. In such a case the dose from each different type of radiation must be considered separately. This is why AF(Tt S)i is defined for each particular emission from each particular radionuclide. It should also be noted that a clear distinction is made between the dosimetry for radionuclides, such as * * 6Ra from which most transformations occur whilst they are distributed throughout the volumk of bone, and others, such as 23gPu, which tend to be bound to bone surfaces, although they may subsequently be buried by bone deposition (Vaughan, 1973; Marshall and Lloyd, 1973). The decision as to whether a radionuclide is assumed to be distributed throughout bone volume or assumed to be on bone surfaces is dealt with in the metabolic data for the various elements. However, two broad criteria can be listed here.

(1) Isotopes of the alkaline earth elements with radioactive half-lives greater than 15 days are considered to be uniformly distributed throughout the volume of bone (ZCRP Publi- carion 20).

(2) Radionuclides .with radioactive half-lives of less than 15 days are considered to be distributed on bone surfaces, since they are unlikely to move far into the volume of bone before they decay.

Thus 224Ra is assumed to be on bone surfaces and **‘Ra is assumed to be uniformly distributed throughout the volume of bone.

For the purposes of dosimetry a simplified specification of the dimensions and location of the two target tissues in bone has been adopted. Thus, for cells near bone surfaces (BS) the average committed dose equivalent is calculated for a layer of tissue 10 pm thick covering all endosteal surfaces and surfaces lined with epithelium. From the work of Lloyd and Hodges (1971) the total area of the endosteal surfaces in man has been estimated to be 11.2 m*. HOW- ever, more recent work by Beddoe et al. (1976) suggests a value of about 16.0 m* @piers and Vaughan, 1976). By comparison the area of epithelium on bone is small. In this report the total endosteal area of the skeleton has been taken to be 12 m*, half being associated with cortical bone and half with trabecular bone (ZCRP Publications 20 and 23). Therefore, the 10 pm thick layer on bone surfaces over which committed dose equivalent is to be averaged (BS) has a mass of 120 g. The mass of active red bone marrow in cavities within trabecular bone (RM) is taken to be 1 500 g (ZCRP Publication 23).


7.2.1. Estimates of absorbed fractions in skeletal tissues

7.2.1.1. Photon emitters. For photon emitters, S is any organ of the body containing the radionuclide and T is either the bone surface cells or active red bone marrow. Values of AF(T c S)JMr, the specific absorbed fraction, are given in ICRP Publication 23 for the target tissues skeleton and red bone marrow and for all source organs and tissues of interest in this report. The values of specific absorbed fraction given there for skeleton have been taken here as appropriate for bone surface cells.

7.2.1.2. a emitters uniformly distributed throughout the volume of mineral bone. In the following discussion it will be assumed that the radionuclides under consideration are uniformly distributed both in cortical and trabecular bone but it will not be assumed that the concentrations in the two sorts of bone are necessarily equal.

(a) AF(BS c TRABECULAR BONE),. Because of the short range of a particles in tissue, surfaces of trabecular bone can be considered essentially infinite and flat for the purposes of a-particle dosimetry. Under this assumption, Thorne (1977) has calculated the fractional energy absorption in a layer 10 pm thick for various depths of burial of a emitters of various energies. The results of these calculations can be used to estimate values of AF(BS c TRABECULAR BONE)i for a emitters of different energies uniformly distributed throughout the volume of trabecular bone. These values range from 0.018 for radionuclides emitting 3 MeV a particles to 0.032 for radionuclides emitting 8 MeV a particles and are in good agreement with those which can be derived from the work of Mays and Sears (1962). In this report a value of 0.025 has been used for all a-emitting radionuclides assumed to be uniformly distributed throughout the volume of trabecular bone.

(b) AF(BS c CORTICAL BONE),. Part of the 120 g of tissue on bone surfaces over which committed dose equivalent is to be estimated lies on the surfaces of Haversian canals. Because of the small dimensions of these canals (m 50 pm diameter) some “cross-fire” is likely to occur even with a-particles, thus increasing the dose to surface cells. Even so, AF(BS c CORTICAL BONE)i will be less than the corresponding quantity for trabecular bone as source organ because a larger proportion of a-particle emissions will be out of range of the bone surfaces. Since the bone surface areas of the two types of bone may be taken to be equal (ICRP Publication 23) it is readily shown that, in the absence of cross-fire

AF(BS + CORTICAL BONE)i Mx AF(BS .- TRABECULAR BONE)i = z z

where Mx = 1000 g is the mass of trabecular bone and MZ = 4000 g is the mass of cortical bone (ICRP Publication 23);

AF(BS + TRABECULAR BONE)i = 0.025 as estimated in 7.2.1.2.(a).

Therefore, in practice, when the effects of cross-fire are included

AF(BS + CORTICAL BONE),> 0.006.

However, if the radionuclide is uniformly distributed throughout cortical bone, AF(BS + CORTICAL BONE)1 cannot exceed kM,,/M,. Where k is the ratio of the mass stopping power of soft tissue to the mass stopping power of mineral bone for a particles, MEz is the mass of the surface layer associated with cortical bone and Mz is


the mass of cortical bone as before. Thus, taking iUz to be 4 000 g (ICRP Publication 23), MEz to be 60 g and k to be 1.3 (Thorne, 1976) we obtain

0.006 s AF(BS t CORTICAL BONE), 5 0.020

Explicit calculations by Spiers (1974a) for 326Ra lead to a value for AF(BS c CORTI- CAL BONE), of 0.01 and this value has been adopted here for all a-emitting radionu- elides distributed throughout the volume of bone.

(c) AF(RM c TRABECULAR BONE),. From the data given by Thorne (1977), AF(RM t TRABECULAR BONE)I is calculated to be 0.02 for radionuclides emitting 3 MeV a particles and 0.09 for radionuclides emitting 8 MeV a particles, in good agreement with the values that can be derived from the work of Mays and Sears (1962). In this report AF(RM t TRABECULAR BONE)r has been taken to be 0.05 for all a-emitting radionuclides assumed to be uniformly distributed throughout the volume of trabecular bone.

(d) AF(RM c CORTICAL BONE)r. Since active red bone marrow has been defined to lie entirely within the marrow cavities of trabecular bone (ICRP Publication U), AF(RM c CORTICAL BONE), is considerably smaller than AF(RM t TRABECU- LAR BONE)I for a-emitting radionuclides uniformly distributed throughout the volume of bone. The work of Whitwell and Spiers (1976) demonstrates that for radionuclides emitting low energy /I particles AF(RM t CORTICAL BONE), is negligible in comparison with AF(RM t TRABECULAR BONE)I. Since these /I particles have ranges comparable with a particles, this result will also hold true for a-emitting radionuclides. For this reason, AF(RM c CORTICAL BONE)I has been taken to be zero for all a-emitting radionuclides assumed to be uniformly distributed throughout the volume of cortical bone.

7.2.1.3. a emitters assumed to be an bone surfaces. In this report radionuclides assumed to be on bone surfaces are taken to be uniformly spread in an infinitely thin layer over the relevant surfaces of bone. This assumption will result in an over-estimate of the true committed dose equivalents received by bone surface cells and active bone marrow because it disregards burial of radioactive deposits by the deposition of new bone mineral. Although such effects might be of considerable importance for several of the radionuclides considered in this report, it was decided that the problem was too complex and that there were insufficient data available to make any reasonable estimate of the reduction in committed dose equivalents due to this cause. This matter will, however, be kept under review, meanwhile it should be noted that values of AL1 for the relevant radionuclides may be unduly restrictive.

(a) AF(BS t TRABECULAR BONE)i. Thorne (1977) has calculated the fractional energy deposited in a layer 10 pm thick from a-emitting radionuclides distributed as an infinitely thin layer on the surfaces of mineral bone. These calculations show that AF(BS + T RA- BECULAR BONE)i is 0.43 for radionuclides emitting 3 MeV a particles and 0.15 for radionuclides emitting 8 MeV a particles. These values are in good agreement with those which can be derived from the work of Mays and Sears (1962). In this report a nominal value of 0.25 has been adopted for all a-emitting radionuclides assumed to be on bone surfaces.

(b) AF(BS t CORTICAL BONE),. If cross-fire in small bone cavities can be neglected, AF(BS c CORTICAL BONE)I will be equal to AF(BS c TRABECULAR BONE)* for a-emitting radionuclides distributed in a thin layer on bone surfaces. Monte-Carlo

(4

(d)


calculations suggest that such cross-fire effects are only important for cavities having radii of curvature of less than 25 pm (Thorne, 1977) and such cavities are uncommon even in cortical bone (Whitwell and Spiers, 1976). For this reason, AF(BS c COR- TICAL BONE), has been taken to be the same as the corresponding value for trabecular bone, i.e. 0.25 for all a-emitting radionuclides assumed to be distributed on bone surfaces.

AF(RM t TRABECULAR BONE)i. The surfaces of the trabeculae may, to a first approximation, be regarded as infinite flat planes compared with the range of a particles in tissue. If this is assumed and the surfaces are also assumed to be sufficiently far apart to make the effects of cross-fire negligible, then, by simple geometric arguments, AF(RM t TRABECULAR BONE), is equal to 0.5. This value has been adopted in this report.

AF(RM c CORTICAL BONE),. Since active red bone marrow has been defined to lie entirely within the trabecular marrow spaces, AF(RM t CORTICAL BONE)1 will be much less than AF(RM c TRABECULAR BONE)i and may, for the purposes of radiological protection, be taken as zero for all a-emitting radionuclides assumed to be on bone surfaces (see section 7.2.1.2).

7.2.1.4. fl emitters uniformly distributed throughout the volume of bone. Spiers and his co- workers (Spiers, 1968 ; 1969, 1974a, b ; Whitwell and Spiers, 1976) have described how to calculate the average dose rate to cells within 10 pm of bone surfaces and to active red bone marrow from B-emitting radionuclides uniformly distributed throughout the volume of bone. In their nomenclature, D,, is the absorbed dose rate to a small tissue-filled cavity in an infinite extent of mineral bone uniformly contaminated at a level of 1 unit of activity per unit of mass and D,-D4 are the absorbed dose rates for the various source and target organs given in Table 7.1 when mineral bone in an adult man is contaminated at a level of 1 unit of activity per unit of mass.

Table 7.1. Source and target organs of the skeleton

Target

Cells near bone surfaces Active red bone marrow

source Trabecular Cortical

bone bone

4 & 2

Quantities PI-P., are defined such that Pj = DJD,, where j = 1 to 4. These quantities, which are independent of the level to which the bone is contaminated, are given in Table 7.2 for various radionuclides emitting /3 particles of average energy E, MeV.

Since P, is the ratio of the dose rate in the target tissue to the dose rate in a small tissue- filled cavity in an infinite extent of mineral bone contaminated to the same level, it may be shown that

P, = Y,E,AF(T+ ShQiG

I

YIE,Q,W MT MS

M,AW+ 8, i= M&


Table 7.2. Values of the absorbed dose ratio, P,, for bone surfaces and red bone marrow from p-emitting radionuclides uniformly distributed throughout the volume of mineral bone. Estimates from Spiers (1974a) and Whitwell and Spiers

(1976)

Nuclide & MeV P4

l Ta 0.08 0.21 0.42 0.11 0.00 9oSr 0.20 0.21 0.47 0.22 0.02 s9Sr 0.55 0.19 0.47 0.26 0.04 eoys 0.93 0.16 0.45 0.27 0.06

l Assumed to be uniformly distributed throughout the volume. of mineral bone when it is produced in bone from its parent 90Sr.

and, therefore, that

k&J’, AF(‘I’+ S)i = -

MS (7.3)

where Y,, Ei, AF(Tt s), and Q, are as defined previously;

MS is the mass of the source organ;

MT is the mass of the target tissue;

C, is the activity in the source organ, and

k is the ratio of the stopping power of electrons in soft tissue to the stopping power of electrons in bone.

Thus AF(BS t TRABECULAR BONE)I = 1 0.12k p

I

AF(BS + CORTICAL BONE)i 0.12k p

=4 2

AF(RM t TRABECULAR BONE)* = !$k PO

AF(RM +- CORTICAL BONE)I = F P4

where 1.5 kg is the mass of active red bone marrow;

4 kg is the mass of cortical bone;

1 kg is the mass of trabecular bone (ICRP Publication 23);

0.12 kg is the mass of the layer between 0 and 10 pm from the surfaces of cortical and trabecular bone, and

k is taken to be 1.07, a value appropriate for electrons over a wide range of energies (Berger and Seltzer, 1966).

Values of the absorbed fractions for a number of radionuclides are given in Table 7.3. Since the main contribution to dose equivalent in the red bone marrow arises from activity


Table 7.3. Values of absorbed fractions for @ emitters uniformly distributed throughout the volume of mineral bone

Nuclide “Ca YSr s9Sr 9oy*

5 0.08 0.20 0.55 0.93

AF (BS + TRABECULAR BONE), 0.027 0.027 0.024 0.021 AF (BS .- CORTICAL BONE), 0.013 0.015 0.015 0.014 AF (RM + TRABECULAR BONE), 0.18 0.35 0.42 0.43 AF (RM + CORTICAL BONE), 0.000 0.007 0.014 0.021

l Assumed to be uniformly distributed throughout the volume of mineral bone when it is produced in bone from its parent 90Sr.

in trabecular bone, it is reasonable to neglect the contribution to dose equivalent in red bone marrow from fl activity in cortical bone and to take nominal values of absorbed fractions for p-emitting radionuclides uniformly distributed throughout the volume of mineral bone. Hence the values in Table 7.3 may be reduced to the following representative values

AF(BS t TRABECULAR BONE)i = 0.025

AF(BS t CORTICAL BONE)i = 0.015

AF(RM .- TRABECULAR BONE), = 0.35

AF(RM + CORTICAL BONE)r = 0.0.

7.2.1.5. /I emitters assumed to be on bone surfaces.

(a) AF(BS t TRABECULAR BONE)i. If E, is less than 0.05 MeV, then the average range of the /I particles will be similar to the range of a particles in tissue and AF(BS t TRA- BECULAR BONE), will be about 0.25. However, if i$ is about 1 MeV the average range of the /3 particles will be many times the average linear dimensions of a marrow cavity or trabeculum and AF(BS + TRABECULAR BONE), may be taken as 0.025, the same as for B emitters distributed uniformly throughout the volume of bone. There is, therefore, a considerable difference between the absorbed fractions for low and high energy /I emitters on the surface of bone. It is impossible to assign a nominal value to AF(BS +- TRABECULAR BONE)i which would be appropriate over this entire range of energies. Two nominal values have, therefore, been chosen;

(i) AF(BS + TRABECULAR BONE), = 0.25, for all values of E, less than 0.2 MeV, and

(ii) AF(BS t TRABECULAR BONE), = 0.025, for all values of E8 greater than or equal to 0.2 MeV.

The dividing line of 0.2 MeV between the two classes has been chosen from a consideration of the absorbed dose rate to surface cells from activity on the surface below those cells and the absorbed dose rate to these cells from activity on other areas of trabecular bone. The first of these two components falls rapidly as Efl increases, while the second rises rather more slowly with increasing &. Since the value of AF(BS c TRABECULAR BONE), is not much larger for E, = 0.2 MeV than it is for E, = 1 MeV but increases markedly with decreasing energy below 0.2 MeV (Spiers, 1974a) the former energy is chosen to mark the boundary between the two classes.

42

(b)

(c)

(d)


AF(BS t CORTICAL BONE),. By arguments similar to those above, two nominal values of AF(BS c CORTICAL BONE), have also been adopted:

(i) AF(BS + CORTICAL BONE), = 0.25 for E, less than 0.2 MeV;

(ii) AF(BS t CORTICAL BONE) [ = 0.015 for E, greater than or equal to 0.2 MeV.

AF(RM t TRABECULAR BONE),. For E, less than 0.05 MeV the average range of the /? particles will be similar to the range of a particles in tissues. Thus, in this case, AF(RM c TRABECULAR BONE), may be taken as 0.5, the value used for a emitters assumed to be deposited on bone surfaces. If, however, E, is about 1 MeV then the range of the B particles will be considerably greater than the average thickness of trabecular bone. In this case, AF(RM t TRABECULAR BONE), will be about 0.4, the value appropriate to fl emitters of this energy uniformly distributed throughout the volume of trabecular bone (Table 7.3). It is reasonable, therefore, to take AF(RM c TRABECULAR BONE), to be 0.5 for all B emitters mainly distributed on bone surfaces.

AF(RM t CORTICAL BONE),. Since active red bone marrow is assumed to be entirely contained within trabecular bone AF(RM t CORTICAL BONE), is much less than AF(RM c TRABECULAR BONE), for E,s 0.05 MeV (see Section 7.2.1.2). If, however, E, is about 1 MeV then the range of the B particles will be considerably greater than the average thickness of cortical bone. In this case AF(RM t CORTICAL BONE), will be about 0.02, the value appropriate to /? emitters of this energy uniformly distributed throughout the volume of cortical bone, as given in Table 7.3. Thus, again AF(RM t CORTICAL BONE)i is much less than AF(RM t TRABECULAR BONE), and for the purposes of radiological protection may be taken to be zero.

7.2.1.6. Fission fragments and recoil atoms. Fission fragments have ranges in tissue of about 20 pm (Green et al., 1977) which is about the same as that for 3 MeV a particles. Values of absorbed fractions for fission fragments are taken to be the same as for a particles in all cases. The energies of recoil atoms are negligible in comparison with the primary emissions and are disregarded for purposes of bone dosimetry.

7.2.1.7. Summary of values of absorbed fractions.

Table 7.4. Recommended absorbed fractions for dosimetry of radionuclides in bone

Class of radionuclide (see Section 7.2)

Source Target organ organ

OT OI B B B emitter emitter emitter emitter emitter uniform uniform on bone on bone

in bZe in surfaces surfaces volume surfaces volume Ee 2 0.2 MeV Ep < 0.2 MeV

Trabecular bone

Cortical bone

Trabecular bone

Cortical bone

Bone surfaces (B.9 0.025

Bone surfaces

(BS) 0.01 Red bone

marrow (RM) 0.05 Red bone

marrow (RM) 0.0

0.25 0.025 0.025 0.25

0.25 0.015 0.015 0.25

0.5 0.35 0.5 0.5

0.0 0.0 0.0 0.0


7.22. Estimates of the number of transformations in trabecular and cortical bone

7.2.2.1. Parent radionuclides. With the present state of knowledge it is not possible to esti-

mate IJrsAsScuLAs BONE and UCORTKAL BONE independently for most bone-seeking radionuclides. Only for the alkaline earth elements which. are incorporated into mineral bone is there a comprehensive model for the distribution of activity between cortical and trabecular bone (ICRP Publication 20). Although UrRAnxcuLAR BONE and UcoaricAL BoNx cannot be calculated with any precision for most bone-seeking radionuclides, nevertheless estimates of these quantities can be made.

(a) If a radionuclide is deposited uniformly on all bone surfaces and is removed at the same rate from both cortical and trabecular bone, then, since the surface areas of cortical and trabecular bone are assumed to be equal, LJTRABEC-LARBoNE = UcoRTlCAL BONE

= Om5 &dlNERAL BONE*

(b) If a radionuclide redistributes, in a time short compared with both its radioactive half-life and its time of residence in bone, such that it uniformly contaminates all bone mass, then ;

u TRABECULAR BONE MX 2:-

U CORTICAL BONE MZ

where A4x is the mass of trabecular bone and MS is the mass of cortical bone, taken to be 1 and 4 kg, respectively,

from which it fOhOWS that UTRABECULAR BONE N 0.2Uk(lNERAL BONE and that

UCORTICAL BONE = o.guM,NERAL BONE. Therefore, when specific information on the distribution of a radionuclide is not available, the following approximation is used

(i) ~TRABECULAR BONE = UCORTICAL BONE = OJUMINERAL BONE for radionuCli&s assumed to be on bone surfaces,

(ii) UTRABKULAR BONE = 0v2&INERAL BONE and UcORTlcAL BONE = o.8uMINERAL BONE for radionuclides assumed to be uniformly distributed throughout the volume of mineral bone.

For radioisotopes of the alkaline earths, values of U for cortical and trabecular bone can be obtained directly from the retention functions given in ICRP Publication 20.

7.2.2.2. Radioactioe daughters. As explained in Chapter 4 it is usually assumed that all daughter radionuclides produced in the body stay with and behave metabolically like the parent radionuclide that was ingested or inhaled. The implications of this assumption when the metabolic model for the parent radionuclide consists of a linear chain of compartments linked by first-order rate coefficients are set out in Chapter 4 and the Appendix. However, the model used in ZCRP Publication 20 to describe the metabolic behaviour of the alkaline earths is not governed completely by first-order rate constants and the following methods are used to estimate values of U for radioactive daughters that are produced in the body from radioisotopes of these elements. Daughters produced in the respiratory and gastrointestinal systems are considered separately from those produced in the remainder of the body.

In the model used to describe the metabolism of the alkaline earths, material entering the transfer compartment is translocated to three general compartments, namely cortical bone, trabecular bone and soft tissues. Relationships are given for the retained fractions R(t) of

44 REPORT OF COMMITTEE2

stable isotopes of the alkaline earths in these three compartments at any time t after their entry to the transfer compartment. When allowance is made for radioactive decay these values can be used directly to compute values of U in cortical bone, trabecular bone and soft tissue from radioactive daughters of .radioisotopes of the alkaline earths entering the transfer compartment after inhalation or ingestion of the parent radionuclide. For example, “Y produced from “Sr in the respiratory or gastrointestinal systems is assumed to have the same metabolic behaviour as strontium after its entry into the transfer compartment, as described for the general case in Chapter 4.

The fractional rate of biological elimination (or accumulation) of a stable isotope of the alkaline earths in any of the three compartments at time t after its entry to the transfer compartment is given by (dR(t)/df)/R(r) = F(t), i.e. the slope of the retention curve divided by the retained fraction at time t. It is assumed, as for the general case described in Chapter 4, that daughter radionuclides produced in compact bone, trabecular bone and soft tissues will behave metabolically like the stable isotope of the parent radionuclide retained in these compartments, e.g. “Y produced in these compartments from retained “Sr will behave metabolically like Sr. The following equations have been used to describe the retention of a radioactive daughter produced in these three compartments from a radioisotope of one of the alkaline earths

; 4vkoRT1CAL BONE = &@)CORTICAL BONE + F(rkORTK4L BONEq’(fkORT,CA‘ BONE

- &q’(&ORTICAI. BONE (7.4a)

; q’htABEC”LAR BONE = &drhBECULAR BONE

+ F(rhRABECULAR BONEd(rhRABECULAR BONE

- &~‘(fhL4BEC”LAR BONE (7.4b)

d d$(r)SOPT TISSUE = &?(r)SOPTTISSUE + F(r)SOFT TISSUd&OFT TISSUE - %d(&OFT TISSUE

(7.4c)

where q’(r) is the activity of the radioactive daughter in the compartment at time r;

q(r) is the activity of the parent radioisotope of an alkaline earth in that compartment at time r; it can be derived from the retention functions given in ICRP Publicarion 20;

r is the time since introduction of the inhaled or ingested radionuclide into the transfer compartment;

F(r) is the fractional rate of loss (or gain) of the stable isotope of the parent radionuclide and may be either positive or negative depending on the slope of the retention curve, and

Ai is the radioactive decay constant of the daughter radionuclide.

Similarly, a system of equations can be derived describing the activities of a chain of parent and daughter radionuclides, the activity of each daughter being determined by the activity of its predecessor in the chain. In all cases the metabolic behaviour of the daughters is assumed to be the same as that of the ancestral radionuclide which is taken into the body.

LIMITS FOR INTAKES OF RADIONUCLIDES BY WORKERS

7.3. Committed Dose Equivalent, HsO, to Cells near Bone Surfaces and Active Red Bone Marrow

7.3.1. Photon emitters

As discussed previously, H50,T in any target organ for photon emitters is given by

45

H - 1.6x lo-” c c 50.T - Sv s I I

where SEE( T t S)i is given by,

SEE(T+ S)i = Y,E,AF(T+ s),Qi MeV g-i per transformation.

MT

The source organ S may be any organ in the body including the skeleton, while the target organ T is either the bone surface cells or the red bone marrow as appropriate. Values of the specific absorbed fraction AF(Tt S)i/MT for photons are given for the skeleton and for red bone marrow in ICRP Publication 23. The values given there for specific absorbed fraction in the skeleton are here taken as appropriate for cells on bone surfaces.

1.3.2. a and /I emitters

In this case the target organ may either be the bone surface cells of mass 120 g or the active red bone marrow of mass 1500 g. Thus,

H SO.BS = 1.6 x lo- 1 ’ c j [

bL4BEC”LAR BONE c SEE(BS + TRABECULAR BONE),

+ %XflCAL BONE 7 SEE(BS t CORTICAL BONE), 1

Sv (7.5)

Similarly,

H sO,RM = 1.6 x 10-r’

x[ u TRABECIJLAR BONE Sv

j 7 SEE(RM t TRABECULAR BONE), 1 i

(7.6) where SEE(BS + TRABECULAR BONE)i

= YiEiAF(BS t TRABECULAR BONE)iQi

120 MeV g- ’ per transformation ;

SEE(BS + CORTICAL BONE)i

= YiEiAF(BS + CORTICAL BONE)iQi MeV g- I per transformation

120 , and

SEE(RM t TRABECULAR BONE),

= YiEAF(RM + TRABECULAR BONE)iQi MeV g- 1 per transformation

1500

Values of AF(Tc S)i are set out in Table 7.4. In equation (7.6) the committed dose equivalent in red bone marrow from cortical bone,

which is very small, has been disregarded (AF = 0, Table 7.4).


References Reddoe, A. H., Darley, P. J. and Spiers. F. W. (1976). Measurements of trabecular bone structure in man. phys.

Med. Biol. 21, 589-607. Eterger, M. S. and Seltzer, S. M. (1966). Aadilional Slopping Power and Range Tables for Protons, Mesons and

Elecfrons. Washington D.C., Gffi~e of Technology Utilization, National Aeronautics and Space Administra- tion (NASA SF3036).

Green. D., Howells, G. and Thorne, M. (1977). A new method for the accurate localization of aagPu in bone. Phys. Med. Biol. 22, 284-297.

ICRP Publication II, Task Group on Radiosensitivity of Tissues in Bone. Pergamon Press, Oxford, 1968. 1CRP Publication 20, Task Group Report on Alkaline Earth Metabolism in Adult Man. Pergamon Press, Oxford,

1973. ICRP Publication 23, Task Group Report on Reference Man. Pergamon Press, Oxford, 1975. KRP Publication 26, Recommendations of the International Commission on Radiological Protection. Pergamon

Press, Oxford, 1977. Lloyd, E. and Hodges, D. (1971). Quantitative characterization of bone: a computer analysis of microradio-

graphs. Clin. Orrhop. 78, 230-250. Marshall, J. H. and Lloyd, E. (1973). The effect of the remodelling of bone upon the relative toxicities of radium

and plutonium in man and dog. In : Radionuclide Carcinogenesis, Eds. Sanders, C. L., Busch, R. H., Ballou, J. E. and Mahlum, D. D. Oak Ridge., Tenn., USAEC, pp. 421-436. (AEC Symposium Series 29.)

Mays, C. W. and Sears, K. A. (1962). Determination of localised alpha dose. III: From surface and volume deposits of PuZ3’, Th”’ and Razz6. University of Utah, pp. 78-85. (COO-226.)

Spiers, F. W. Radioisotopes in the Human Body. Academic Press, New York and London, (1968). Spiers, F. W. Beta particle dosimetry in trabecular bone. In: Delayed Effects of Bone Seeking Radionuclides,

Eds. Mays, C. W., Jee, W. S. S., Lloyd, R. D., Stover, B. J., Dougherty, J. H. and Taylor, G. N. pp. 95- 108. University of Utah Press, Salt Lake City, (1969).

Spiers, F. W. (1974a). A New Approach to Setting Maximum Permissible Levels of Radionuclides in Bone. Medical Research Council, Committee on Protection against Ionizing Radiations, PIRC 74/3.

Spiers, F. W. (I 974b). Radionuclides and bone-from lzsRa to 90Sr. Br. J. Radiol. 47, 833-844. Spiers, F. W. and Vaughan, J. (1976). Hazards of plutonium with special reference to the skeleton. Nature 259,

53 I-534. Thorne, M. C. (1976). Aspects of the dosimetry of plutonium in bone. Nature 259, 539-541. Thorne, M. C. (1977). Aspects of the dosimetrv of a-emitting radionuclides in bone with particular emphasis on

llsRa and 239Pu. Phys. Med. Biol. 22, 3W6. Vaughan, J. The pattern of la9Pu distribution in the skeleton. In: Uranium, Plutonium, Transplutonium Elements,

Eds. Hodge, H. C., Stannard, J. N. and Hursh, J. B. pp. 397-421. Springer-Verlag, Berlin, (1973). Whitwell, J. R. and Spiers, F. W. The dosimetry of beta-emitters in bone: theoretical methods. In: Contam-

ination par Radionucleides Osteotropes et Radioprotection. pp. 401417. 5th International Meeting of the French Society of Radiation Protection, Grenoble, (1971).

Whitwell, J. R. and Spiers, F. W. (1976). Calculated beta-ray dose factors for trabecular bone. Phys. Med. Biol. 21, 16-38.


8. DOSIMETRIC MODEL FOR SUBMERSION IN A RADIOACTIVE CLOUD

8.1. Introduction

When a person is submerged in a radioactive gas, the skin and other organs of the body may be irradiated both by external irradiation and by internal irradiation from gas absorbed into body tissues. The respiratory system and other organs may also be irradiated by gas contained in the lungs. These sources of irradiation, which are discussed below, limit the exposure of a worker to an inert radioactive gas and to elemental tritium.

8.2. Relative Magnitudes of Dose-Equivalent Rates from External and Internal Radiation

Consider a person submerged in a radioactive cloud of infinite extent and of volume concentration C Bq m- 3. Let the dose-equivalent rate to any tissue from external radiation be h,, from internal irradiation by absorbed gas be fi,, and to the lung from contained gas be h,.

Then tiE, in a small element of tissue in a person submerged in a radioactive cloud of infinite extent, is-given by

k E = c skr,lp, Sv h-’ (8.1)

where pA, the density of air, is about 1 300 g m- ‘;

s (in Sv h- ‘) is the dose-equivalent rate in a small element of any medium of infinite extent uniformly contaminated at a concentration of 1 Bq g- 1 ;

k, which is usually close to unity, is the mass stopping power of radiations in tissue relative to their mass stopping power in air, and

gE is a geometrical factor to allow for shielding by overlying tissues.

The value of gE is idWayS zero for /? emissions from tritium and for all a-particle emissions, since these are unable to reach any of the sensitive tissues of the body, including the lenses of the eyes and the basal layer of the epidermis which, for the purposes of dosimetry, are taken to be at depths of 3 mm and 70 pm respectively (paras. 62 and 64, ICRP Publication 26). For most /I emissions and for low energy photons, gE is about 0.5 for tissues near the surface of the body and tends to zero for deep-iying tissues. For very penetrating photons, gE ap- proaches unity for all tissues of the body.

After prolonged exposure of the person to the cloud an equilibrium is reached between the concentrations of gas in air and tissue. Under these conditions it may be shown that the concentration of gas in tissue CT is given by

Cr = WP, Bq g-’ (8.2)

where pr, the density of tissue, is about lo6 g rnm3, and

6 is the solubility of the gas in tissue expressed as the volume of gas in equilibrium with unit volume of tissue at normal atmospheric pressure.

This solubility coefficient increases with the atomic weight of the gas, e.g. in water at body temperature its value varies from about 0.02 for hydrogen to about 0.1 for xenon (Kaye


and Laby, 1956). These values may be increased by a factor of 3-20 in adipose tissue (Law- rence et al., 1946). Thus the dose-equivalent rate in tissue from absorbed gas, a,,, is given by

II A = s &7*/P* Sv h-r (8.3)

where gA is a geometric factor determined by the dimensions of a person and the range of the radiations concerned. For a and /? emissions and also for low energy photons gA will be approximately unity for tissues at the centre of the body and 0.5 for surface tissues. For more energetic photons, gA is much less than 1 for all tissues and decreases with increasing photon energy.

The dose-equivalent rate in the lung from contained gas, fiL, is given by

fi L=SWLSLIML Sv h-r (8.4)

where V,, the average volume of air contained in the lungs, is about 3 x lo-’ m’;

ML, the mass of the lungs, is taken to be 1000 g (ICRP Publication 23);

gL is a geometrical factor which, for a and B emissions and for low energy photons, is approximately unity. The value of gL decreases with increasing photon energy.

8.2.1. Tritium For this nuclide, k, given by eqn (8.1) is zero for all relevant tissues of the body because

of the short range of the tritium /I emissions in tissue. The ratio of dose-equivalent rate in any tissue from absorbed gas to that in the lung from contained gas is, from eqns (8.3) and (8.4), given by

fi” &AM, y=- HL VLgLPT

(8.5)

Since pr N lo6 g m- 3 and IULJV, = 106/3 g m- 3, while gA and gL are both approximately unity for tritium /I emissions, this expressionreduces to

For tritium 6 is about 0.02 for aqueous tissues and 0.05 for adipose tissues (Lawrence et al., 1946). Thus the dose-equivalent rate in lung from the tritium gas contained within it will be 60 to 150 times that in any tissue from absorbed gas. Therefore, in this report, submersion in tritium gas is limited solely by consideration of dose-equivalent rate in the lung. However, it is emphasized that the limit on exposure to tritiated water is very much less than that for elemental tritium and in most cases in practice exposure to tritiated water will be the limiting factor (see dosimetric data for hydrogen).

8.2.2. Radon and thoron

Limits for exposure to radon and thoron will be the subjects of separate statements by the Commission.

8.2.3. The noble gases All the radioisotopes of the noble gases argon, krypton and xenon considered in this report

emit either photons or B particles of considerable energy. Thus, for tissues near the surface of the body, including the skin, gE will be about 0.5. From eqns (8.1) and (8.4)


8, M&g, z 130 ?=- - HL VLQLPA QL

(8.7)

and since gL cannot be greater than unity, fi, is more than 130 times fiL. Similarly, from eqns (8.1) and (8.3)

fi, p,kg, y=- HA PdQ*

(8.8)

and since 65 2 (Lawrence et al., 1946), pT/pA N 800, and gA I 1, fi, is more than 200 times B,.

Therefore, when applying the system of dose limitation described in Chapter 2, it is clear that, for exposure by submersion in radioisotopes of the noble gases, external irradiation will be of such overriding importance that it alone need be considered. Thus, in this report dose equivalents from absorbed gas and gas contained in the lung have been disregarded. The methods used to calculate dose-equivalent rate from external irradiation are discussed in Section 8.3.

8.2.3.1. Daughter radionuclides. If a daughter radionuclide produced by decay of a radioactive inert gas is itself an inert gas the considerations in 8.2.3 apply and exposure by submersion in the daughter radionuclide will be limited by the dose equivalents in tissues from external radiation. If the daughter radionuclide is not an inert radioactive gas, it can be shown that in practice dose equivalents from daughters produced from their parent absorbed in body tissues will usually be small compared with the external dose from parent and daughter outside the body. In this report dose equivalents from daughters produced from their parent in body tissues have been disregarded. However, consideration must be given to the dose equivalents resulting from the inhalation of any daughter radionuclides which are present in the radioactive cloud in which the person is submerged, as described in Sections 9.4 and 9.7.

8.3. Dose-Equivalent Rates in Body Tissues from Submersion

8.3.1. Photon emitters

The energy spectra from sources of mono-energetic photons in an infinite extent of air have been calculated by Dillman (1971). From these data the dose-equivalent rat& fi Sv h- l to organs and tissues of the body may be calculated for a Reference Man situated within a semi- infinite cloud bounded by the floor on which he stands, using methods described by Poston and Snyder (1974).

The dose-equivalent rate in the lens is assumed to be the same as that in the skin, for the purposes of this calculation taken to extend from O-2 mm depth (ZCRP Publication 23).

No energy spectra are available for photon sources in a cloud of finite dimensions such as that in a room, but an approximate estimate of the dose-equivalent rate in the skin or internal organs is given by

2fiCI - exp( - wW1 (8.9)

where fi is the dose-equivalent rate estimated for a semi-infinite cloud (Poston and Snyder, 1974) and the factor 2 is used because for most room sizes the floor no longer limits irradiation by the cloud to a 2a geometry, at least for the head of a standing worker and most room dimensions;


p,, is the mass energy absorption coefficient in air;

P,, is the density of air, and

r is the effective radius of the room.

When r is the radius of a sphere equal in volume to that of the room the dose-equivalent rate is overestimated by a small factor which depends on the shape of the room.

11 - exp( - pd*r)] is the factor by which the first interaction dose from an infinite cloud is reduced, to allow for the fact that there are no sources of photons outside the room.

In this report correction factors are used for room volumes of 100 m3 (r = 2.9 m), 500 m3 (r = 4.9 m) and 1000 m3 (r = 6.2 m), where such factors are necessary.

8.3.2. Electrons and /I emitters

The dose-equivalent rate in the skin at a depth of 70 pm and in the lens at a depth of 3 mm from a semi-infinite cloud of electron sources is computed by integrating the point kernel of Berger (1971) over the appropriate locus in air where the emitter could contribute to this dose- equivalent rate. The point kernels were corrected for the ratio of mass stopping powers tissue/air taken to be 1.14 for all electron energies (Berger, 1974). In most cases the correction for room size is small and it is not estimated for electrons in this report.

The energies of the bremsstrahlung for electron sources have been calculated by Dillman et al. (1973) and the dose-equivalent rates from these photon sources are estimated as described above. Their contribution to the total dose-equivalent rate is very small.

8.4. Derived Air Concentration (DAC) for Submersion

As discussed in Section 8.2.1 and Chapter 3, exposure to elemental tritium in air in any year is limited by consideration of stochastic effects in the lung as follows,

w‘unlfiLunl j C(r) dz I; 0.05 Sv (a. lo)

where wLunl is the weighting factor for lung given in Table 2.1, Chapter 2;

kLunl (in Sv m- ’ Bq- 1 h- ‘) is the dose-equivalent rate to lung from exposure to unit concentration of tritium in air (i.e. 1 Bq m- ‘), and

C(t) (in Bq m- ‘) is the concentration of elemental tritium in air at any time t and the limits on integration are over a working year.

Exposure to an inert radioactive gas in any year (see 8.2.3 and Chapter 3) is limited by consideration of external radiation of the body as follows

7 w& j C(t) dt < 0.05 Sv (a.lla)

and fi,jC(t)dt50.5 Sv (8.11b)

and kLcnr j C(r) dr ~5 0.3 Sv (a.iic)

where wr is the weighting factor for tissue T and has the values given in Table 2.1, Chapter 2;

fir (in Sv rns3 Bq- ’ h-l) is the dose-equivalent rate in any tissue T, and


B Lcns is the corresponding value for the lens of the eye, resulting from submersion of Reference Man in unit concentration of the inert gas in air (i.e. 1 Bq m- 3).

C(r) (in Bq m- ‘) is the concentration of the inert radioactive gas in air at any time r and the limits on integration are over a working year.

For convenience, the Commission recommends values of derived air concentration (DAC) which are 1/2OOOth of the greatest value of SC(t) dt which satisfies relationship (8.10) for elemental tritium or which satisfies relationship (8.11) for an inert radioactive gas. When the DAC for an inert radioactive gas is determined by consideration of non-stochastic effects, the organ or tissue concerned (usually the skin) is named below the value of DAC. In all such cases a greater value of DAC is given in parentheses. This value is determined by consideration of the Commission’s recommendations for limiting stochastic effects, i.e. relationship (8.1 la) above. This value of DAC determined by consideration of stochastic effects is useful when considering the limitation of exposure from several sources (see Chapter 9), and also when application of the Commission’s system of dose limitation (Section E, ZCRP Pubficotion 26) requires that a worker receives only a fraction of the dose-equivalent limit for stochastic effects.

It is emphasized (see Chapter 3) that DAC must always be used circumspectly. The overriding limit on exposure by submersion to elemental tritium is determined by relationship (8.10) and to an inert radioactive gas by relationship (8.11).

References Berger, M. J. (1971). Distribution of absorbed dose. around point sources of electrons and beta particles in

water and other media. J. Nucl. Med. 12, Suppl. 5, 5-23. Berger, M. J. (1974). Beta-ray dose in tissue-equivalent material immersed in a radioactive cloud. He&h Phys.

26, l-12. Dillman, L. T. Scattered energy spectrum for a monoenergetic gamma emitter uniformly distributed in

an infinite cloud. In : Health Physics Division Annual Progress Report for Period ending 31 Juty 1970. pp. 216 222. Oak Ridge National Laboratory (ORNL 4584), (1971).

Dillman, L. T., Snyder, W. S. and Ford, M. R. Nuclear data compilations of utility in medical and biological applications. In : Nuclear Data in Science ond Technology: Proceedings of a Symposium, Paris, 12-26 March 1973. Vol. 2, pp. 529-539. IAEA, Vienna, (1973).

ICRP Publication 23, ICRP Task Group on Reference Man. Pergamon Press, Oxford, 1975. ICRP Publication 26, Recommendarions of the International Commission on Radiological Protection. Pergamon

Press, Oxford, 1977. Kaye, G. W. C. and Laby, T. H. Tubfes of Physical and Chemical Constunts, 11 th Edition, p. 137.

William Clowes, London, (1956). Lawrence, J. H., Loomis, W. F., Tobias, C. A. and Turpin. F. H. (1946). Preliminary observations on the

narcotic effect of xenon with a review of values for solubilities of gases in water and oils. J. Physiol. 105, 197-204.

Poston, J. W. and Snyder, W. S. (1974). A model for exposure to a semi-infinite cloud of a photon emitter. Health Phys. 26, 287-293.

52 REF'ORTOFCOMMIlTEE2

9. USE AND LIMITATIONS OF THE DOSIMETRIC DATA 9.1. Introduction

It is emphasized that the dose equivalent received by an individual, together with that to which he is committed as a result of an exposure to radioactive materials, is the basis for determining the exposure status of that individual. A value of H,, per unit intake given in the Supplement to this report is a predictive estimate of dose equivalent during a period of 50 years following intake of a radioactive material. A secondary limit, ALI, and the corresponding derived limit, DAC, are useful criteria for interpreting the exposure of workers.

The estimates of Hs,,, AL1 and DAC provided for each radionuclide are based on a great many assumptions which must be appreciated if these quantities are to be properly employed. The general models incorporating these assumptions have been described in preceding chapters. Assumptions specific to each element are discussed in the metabolic data for that element given on pp. 63-l 15 of this report. These assumptions and some of their implications are discussed in the paragraphs that follow.

9.2. Assumptions Concerning Chemical and Physical Form

An effort has been made to consider the effect of chemical and physical form. For example, in the dosimetric model of the respiratory system, discussed in Chapter 5, three general classes of compounds are considered and more than one value for absorption from the gastrointestinal tract (fi) is employed where information is available to justify a choice between different values offr. The values offi listed may not always be applicable in practice and the user should make his own choice from the values given, interpolating between values where appropriate. A means of adjusting for the effect of particle size, to be used when this is known to vary from an AMAD of 1 pm for which Hso, AL1 and DAC are calculated, is described in Chapter 5. The behaviour of any specific material may be expected to vary significantly from that of the model employed, and alterations should be made in the application of the model when specific data are available to justify such alterations.

9.3. Assumptions Concerning Metabolic Models

The specific metabolic models employed in the calculation of the Dosimetric Data are described in the Metabolic Data for the elements on pp. 63-l 15 of this report. References are given concerning the origin or derivation of these models. In some cases the model may be quite detailed and may be supported by much experimental evidence from observations on humans. In other cases, data may be limited to a few observations on a single species of experimental animal. This lack of knowledge about metabolism represents the largest factor of uncertainty in most estimates of committed dose equivalent. If new data become available, it should be possible to make appropriate adjustments in the Dosimetric Data. It is hoped that users of this report will publish data on cases that may occur in practice and that may offer guidance as to metabolic behaviour of specific radionuclides.

9.4. Assumptions Concerning Daughter Radionuclides

It is usually assumed that daughter radionuclides produced from their parent within the body stay with and behave metabolically like their parent. If there is evidence to the contrary


this has been used as stated in the Metabolic Data. When new evidence becomes available this information should be used to revise the values of H5,,, AL1 and DAC.

9.5. Assumptions Concerning the Exposed Individual

The Dosimetric Data are calculated for an occupationally exposed adult. Detailed assumptions concerning this individual are derived from ICRP Publication 23. Some of the more important of these assumptions are discussed and tabulated in Chapters 3 and 4. Assumptions relevant to a particular element are discussed in the Metabolic Data for that element. The Commission does not recommend the use of the data and the models described in this report to estimate committed dose equivalent to members of a population, for example from radionuclides in the environment, by adjustment solely on the basis of differences in mass of organs or magnitude of intake. While some insight into population exposure may be obtained from the data, they were not collected with this purpose in mind. A bibliography, which is by no means exhaustive, concerning the estimation of committed dose equivalent in people of different ages from intakes of radionuclides is given at the end of this Chapter.

9.6. Assumptions Concerning Chemical Toxicity

The values of ALI given in this report and in the collected dosimetric data given in the Supplement relate solely to limitations on the radiation doses received by organs and tissues of the body and not to chemical toxicity. These values, which are stated in units of becquerel, can be converted to masses using appropriate values of specific activity. The Commission is aware that the chemical nature of some materials, e.g. certain isotopes of uranium, may present a greater risk to the worker than their radioactivity. The Commission is also aware that, for some radionuclides of very low specific activity, e.g. “‘In, the mass of material associated with AL1 is greater than it would be reasonable to expect a worker might inhale or ingest in any one year. Accordingly, the values of ALI should always be used with circumspection remembering that the values have been determined by consideration of dose equivalent alone.

9.7. Exposure to a Mixture of Radionuclides by Inhalation, Ingestion and Submersion

Each AL1 has been derived by assuming that a worker takes the radionuclide concerned into his body by only one of the specified routes of ingestion or inhalation. In practice, workers may inhale and ingest a mixture of radionuclides, may also be submerged in a cloud of a radioactive, chemically inert, gas or a cloud of gaseous tritium and may also be exposed to other external sources of radiation. In such circumstances the tabulated values of ALI cannot be used directly and exposures must be limited as described below (see also para. 110, ZCRP Publication 26).

The AL1 for a particular radionuclide is derived from a consideration of both stochastic and non-stochastic effects (see Chapter 2) and this is true also of the DAC for submersion in a radioactive cloud (see Chapter 8). Thus, when considering ingestion and inhalation of a mixture of radionuclides together with submersion in a cloud of a radioactive gas, the total exposure should be limited by ensuring that the inequalities set out in Chapter 2 are satisfied. However, it is recognized that it is often more convenient to use the appropriate ALIs and DACs than to compute the various appropriate values of H,,. For any year of practice the conditions set out in Chapter 2 will be met if the following inequality is satisfied

54 REPORT OF COMMI-IYBB 2

T ( &XNo + T (&),,. + q 2 000 D*C(Zbmersion),. ’ 1

where (I,),,, is the intake of radionuclide j by ingestion in any one year;

(I,),,, is the intake of a radionuclidej by inhalation in any one year;

(ALI,),,, is the Annual Limit on Intake for radionuclide j by ingestion;

(ALI,),uu is the Annual Limit on Intake for radionuclide j by inhalation;

U,.. is jC,,.(f) dt, where t has units of hours, C,..(r) is the concentration of radionuclide j’ in which the man is submerged at time t and the time integral is over 1 year of practice (see Chapters 3 and 8);

DAC(Submersion),.. is the Derived Air Concentration(Submersion) for any relevant radionuclide j’, and

2000 is the number of hours in a working year.

When a worker is exposed to external sources of radiation in addition to unsealed sources of radioactive material, it may be convenient to limit the total exposure in any year of practice as discussed in para. 110 of ICRP Publicurion 26 and in the statement made by the Com- mission after its meeting in Stockholm 1978 (Annals of the ICRP, 1978).

It should be noted however that in certain circumstances the inequalities described above and in para. 110 of ZCRP Publicarion 26 may be more restrictive than the conditions set out in Chapter 2 if one or more of the ALIs or DACs are determined by non-stochastic effects. Therefore, if a proposed regime of exposure based on these inequalities is thought to be unduly restrictive it is recommended that more exact calculations are made to satisfiy the relationship set out in Chapter 2 and para. 103 of ICRP Publication 26. In this respect the dosimetric data set out in the Supplement to this report will be found to contain useful numerical information, e.g. values of HsO,T and wTH5,,r per unit intake of a radionuclide. Values are also given for the intake Z of a radionuclide in any year and of the DAC for a radioactive inert gas that satisfy the Commission’s recommendations for limiting stochastic effects within a working year. Such values are always given in parentheses below the AL1 or DAC Submersion as appropriate whenever these latter are determined by non-stochastic effects.

References

ICRP Publication 26, Recommendafions of the International Commission on Radiological Protection, Pcrgamon Press, Oxford. 1977.

Anna/s ofthe ICRP. Statement after the Stockholm Meeting of the International Commission on Radiological Protection. Ann. ICRP 2(l). 1978.

Bibliography Concerning Estimation of Dose Equivalent from Radionuclides in Peopie of Different Ages

United Nations Scientific Committee on the Effects of Atomic Radiation, Sources and Effects of Ionizing Radiation, United Nations, New York, 1977.

Ibid. 1958: 1962: 1964: 1966: 1969 and 1972. National Academy of Sciences (I 972). The Effects on Populations of Exposure to Low Levels of Ionizing Radia-

tion. A report of the Advisory Committee on the Biological Effects of Ionizing Radiation. National Research Council, Washington D.C.

Medical Research Council (1975). The Toxicity of Plutonium. Her Majesty’s Stationery Office, London.


Medical Research Council (1975). Criteria for Controlling Radiation Doses to the Public after Accidental Escape of Radioactiue Material. Her Majesty’s Stationary @Ike, London.

Papworth, D. G. and Vennart, J. (I 973). Retention of 9oSr in human bones at different ages and the resulting radiation doses. Phys. Med. Biol. 18, 169486.

Vennart, J. and Ash, Patricia J. N. D. (1976). Derived limits for a’S in food and air. Health Phys. 30.291-294. Wash 1400 (1975). Reactor Safety Study: An Assessment of Accident Risks in U.S. Commercial Nuclear Power

Plants. Appendix VI, Appendix D-l to D-39, United States Nuclear Regulatory Commission, October 1975.


APPENDIX: EXACT AND APPROXIMATE SOLUTIONS OF THE COMPARTMENTAL MODELS USED IN THIS REPORT

In this report all the metabolic models used, except that for the alkaline earths (see Chapter 7) can be expressed in the form of systems of first order differential equations with constant coefficients. Exact solutions of such systems of equations under specific boundary conditions have been investigated in ICRP Publication 2 (1959), Jacquez (1972), Rutherford (1913) and Skrable et al. (1974). In particular, most of the models used in this report do not involve feedback and can thus be reduced to the linear chains of compartments discussed by Ruther- ford (1913) and Skrable et al. (1974). Such linear chains are described by systems of differential equations of the form

dqi dt Cf) = ti(t) + &i- l.i)q(i-l)Cr) - A*qi(r) i=2ton (A-1)

where qi(t) is the total quantity of activity in compartment i at time t;

t,(t) is the rate of intake of activity from outside the system into compartment i at time

1;

A(I_-l, Ij is the rate constant for transfer of material from compartment i - 1 to compartment i, and

1, is the rate constant for loss of material from compartment i.

When evaluating the number of transformations U in the various compartments of the chain, calculations are simplified because a quantity of activity is supposed to be instantaneously deposited in the first compartment of the chain and that, subsequent to this, no further intakes of activity from outside the system occur in any of the compartments of the chain. Thus, taking time t to be zero at the time of this single instantaneous intake, j,(t) may be set equal to zero for all i, q,(O) may be set equal to the amount of activity instantaneously deposited in the first compartment and qi(0) may be set equal to zero for all i from 2 to n, i.e. for all other compartments of the chain. Under these boundary conditions Skrable et al. (1974) have shown that

where

fi ai =a,xa,+, x...a,ifn>m i=m

and

,jj!j=lifm>n

(A-2)

LIMITS FOR INTAKES OF RADIONUCLIDBS BY WORKERS 57

The above expression for q,(t) may be integrated between time zero and time T to yield

(A-3)

Thus, if qi(t) is taken to be the activity of a radionuclide in compartment i at time t, and if T is taken to be 50 years, eqn (A.3) is an expression of the integrated activity of that radionuclide in the i’th compartment of the chain in the 50 years following its intake into the first compartment of that chain. Equation (A.3) may also be used to determine the integrated activities of daughter radionuclides in the various compartments of the lung and gastrointestinal tract. This will be made clear by the following examples. It should be noted that eqns (A.2) and (A.3) break down in the case of I, = 1, (p # k). However, in this case it can easily be shown that lim,p_4 [qi(t)] and lim,P_r,[Ui(T)] still exist since these equations are completely antisymmetric in exchange of 1.

A.l. Integrated Activities in the Compartments of a Two Compartment Chain

In this case eqn (A.l) reduces to

dq, dr= - ~,q,(O

de

provided that the boundary‘ conditions discussed above are employed. An example of this might be the distribution of a radionuclide between the transfer compartment and a particular tissue compartment following its instantaneous introduction into the transfer compartment (see Chapter 4). In this case, ql(0) will be equal to Q, the amount of activity initially introduced, ql(t) the activity in the transfer compartment at any time t, q2(t) the activity in the tissue compartment, at time t, II = I, + A,, L(1,2j = bL, and & = & + AR, where 1, is the rate of loss of the stable element from the transfer compartment, AR the radioactive decay constant of the radionuclide, bl, the rate of translocation from the transfer compartment to tissue compartment b and & the rate of loss from that compartment (see Fig. 4.1, Chapter 4).

Solution of these equations can be obtained by substituting into eqn (A.2) to give

(4 - Al) I *

This second expression can be rewritten in the form

ql(t) = Qe-(k+hN

-(A.+Ar)f (Ab+h)r

M) = b& {

Qe tnb _ naj +

Qe-

q2w = b1.Q (A - Lb) {e -(h.+h)t _e-(A.+A~M >

Which emphasizes the build-up and subsequent decline in activity in the second compartment of the chain.


Similarly, substitution into eqn (A.3) gives the time integrals of activity in the two compartments

Uzm = 6n* I Q(l -e -(A.+h)T) Q(1 _ e -(Ab+h)T)

(12, + A,)(& - A,) + (A, + A,) (A, - &)

and this second expression can again be rewritten

bl,Q

u2(T) = _) I

(1 _ e-(“b+h)T) (1 _ e-(h+h)T)

(& +M - (4 + At)

The first term within the bracket represents the integrated activity that would have occurred in the organ, had transfer to it taken place instantaneously. The second term represents a correction for the finite residence time in the transfer compartment.

A.2. Total number of Transformations of an Inhaled Radionuclide in the Tracheobronchial Region of the Lung

Activity in the T-B region of the lung arises from material deposited in the two compartments which comprise that region and material passing through that region from the P region to the GI tract (see Chapter 5). For material deposited in compartment c (see Chapter 5, Fig. 5.2), ql(0) = DT_B I;; per unit activity inhaled and the total number of transformations U, is given by taking i = 1 in eqn (A.3), to yield

u,= DT_-BFe(l - e-“IT)

4

but ,I1 is the rate constant for loss from compartmentcandis,therefore,(& + &)(seeChapter 5). Thus,

u, = &-r&(1 - e+=+*n)) transformations

& + 1,

where q = 1.58 x 10’ is the number of seconds in 50 years and the various rate constants 1 have units of s-r here and throughout the rest of this Appendix.

Similarly, U.. is given by

u* = &Jdl - e-d”d+“R)) transformations

43 + 1, To determine the total number of transformations in the T-B region from material passing from compartment f of the P region to the gastrointestinal tract it is necessary to use eqn (A.3) to calculate U,(T), setting qi(0) = DpFf, At1,1j = &, A, = A, + dR, A, = & + 1, and

41(O) (1 - emA? A& - Al)

+ qd0) (1 - e-9

12(4 - 12)


Which, using the values set out above, yields

W,Fr (lf + AR)e’l(Ad+AR) _ (& + ~&-)l(k+AR)

43 - If transformations

Similarly, the total number of transformations in the T-B region from material passing from compartment g of the P region to the gastrointestinal tract is

Thus, since I,, Ad, L, and I, are all much greater than l/q, the exponential terms in the above expressions can be neglected, and the total number of transformations in the T-B region from all sources following inhalation of unit activity of a radionuclide (U& is given by

u

A.3. Total number of Transformations of a Daughter Radionuclide in the Pulmonary Region

If daughter is produced from its deposited parent in the pulmonary region it is necessary to evaluate U,(T). As before

U,(T) =

Thus, the total number of transformations of daughter in compartment e ((I:) from the inhalation of unit activity of the parent is obtained from the above formula by setting q,(O) = Dd’;, 41,2j = Ai, I, = 1, + AR, 1, = I, + XR and T = q. Thus,

&J&J, uL = (& + &) (1, + 1;) (

, + (A, + &~~-(“‘+“kk _ (1, + qe-(L+“~k

6% - 1,)

and, since I, is very much greater than l/q,

W,F, ui = (a, + A,) (A, + A;)’

Repeating this calculation for compartments f, g and h, the total number of transformations of daughter in the pulmonary region, p, is Vi where,

+ (Ah + AR) (A, + id) I * Similar calculations may be made to

Fr Fs + (A, + &t) (4 + 4s) + (5 + Ajo (AI + A;)

determine the total number of transformations in any


of the compartments of the various models. However, for convenience, approximate values of these total numbers of transformations are set out below for the various compartments of the lungs and gastrointestinal tract.

A.4. Approximate Expressions for Total Numbers of Transformations in the Various Compartments of the Lung and Gastrointestinal Tract

A.4.1. Inhaled and ingested radionuclides

In all compartments of the lungs, except j, clearance half-times are much less than 50 years (see Chapter 5) and approximate expressions for total numbers of transformations in the various compartments may easily be obtained by use of the formulae discussed above. For convenience these expressions are listed in Table A.1.

Table A.1 . Approximate expressions for the number of transformations in the various compartments of the lung following inhalation of one Bq of activity. The symbols are defined in Fig. 5.2, Chapter 5

Compartment Number of

transformations Compartment Number of transformations

C Dr-.F, a + hr

d

Activity enters the gastrointestinal tract from compartments b and d of the lung. Thus, since compartments f and g empty into compartment d, the total activity entering the gastrointestinal tract per unit activity inhaled is approximately

where correction has been made for radioactive decay in the compartments before transfer. Similarly, the activity transported directly to the transfer compartment per unit activity inhaled is approximately,

Since these transfer processes are essentially over in a time short compared with 50 years they can, for the purposes of radiological protection be regarded as occurring instantaneously.

LIMITS FOR INTAKES OF RADJONUCLIDES BY WORKERS 61

For activity entering the gastrointestinal tract by ingestion all residence times are short compared with 50 years and approximate expressions for integrated activity in the various sections of the tract following ingestion of unit activity can be obtained by the methods described previously. These expressions are listed in Table A.2.

Table A.2. Approximate expressions for the number of transformations in the various regions of the gastrointestinal tract following ingestion of 1 Bq of activity.

The symbols are defined in Fig. 6.1, Chapter 6

Region

Stomach

Number of transformations

I

&r + All

Small intestine

Upper large intestine

Lower large intestine

Further, the activity translocated from the gastrointestinal tract to the transfer compartment per unit activity ingested is

and, since all residence times in the gastrointestinal tract are short compared with 50 years, this transfer may, for the purposes of radiological protection, be regarded as occurring instantaneously.

A.4.2. Radioactive daughters of inhaled and ingested radionuclides

As a first approximation it is sufficient to regard all the decay of a parent radionuclide in a compartment of the lung or GI tract as occurring instantaneously at the beginning of the 50 years following intake. The only case in which this assumption is markedly in error is for the inhalation of long lived class Y materials since in this case an appreciable amount of material is retained in the pulmonary lymph nodes over the whole period of 50 years. However, even in this case the committed dose equivalent from radioactive daughter products will not be seriously over-estimated by using this simplifying approximation. Thus, if the number of transformations of a parent in any compartment is A, it will produce in that compartment Ali Bq of its daughter, where nk is the radioactive decay constant of the daughter.

Unless otherwise indicated, in the lung it is assumed that the clearance of a radioactive daughter from a particular compartment is the same as that of its inhaled parent and that there is no mixing of daughter between compartments at the time of its production. In these circumstances the time integrals of daughter activity in the various compartments of the lung are given in Table A3.

Similar expressions can be obtained for the various regions of the gastrointestinal tract and these are given in Table A4.


Table A.3 Approximate expressions for the number of transformations of a radioactive daughter in the various compartments of the lung. A. to Al are the number of transformations of its parent in compartments a to j of

the lung. The other symbols used are defined previously in this appendix and in Fig. 5.2, Chapter 5.

Compartment Number of transformations Compartment Number of transformations

b A& ht. + h;

Table A.4. Approximate expressions for the number of transformations of a radioactive daughter in the various regions of the gastrointestinal tract. Asr, AsIr AvLl, ALLI are the number of transformations of the parent in the various regions of the tract and r\a is the value of the transfer coefficient to body fluids characteristic of the

parent. The other symbols used are defined in Fig. 6.1, Chapter 6

Region Number of transformations

Stomach Al&

References ICRP Publication 2, Report of Committee 2 on Permissible Dose for Internal Radiation. Pergamon Press, Oxford,

1959. Jacquez, J. A. Compartmental Analysis in Biology and Medicine. Elsevier. Amsterdam, (1972). Rutherford, E. Radioactive Substances and their Radiations. pp. 410431. Cambridge University Press, (1913). Skrable, K., French, C., Chabot, G. and Major, A. (1974). A general equation for the kinetics of linear first

or&r phenomena and suggested applications. Health Phys. 27, 155-157.


METABOLIC DATA

PREFACE

In the following pages the relevant metabolic data for each element precedes a table of values of ALI and DAC for radioisotopes of that element having radioactive half-lives greater than 10 min.

The metabolic models described are for compounds of a stable isotope of the element. Retention data given in the literature have, where necessary, been corrected for radioactive decay of the radionuclides concerned. Because of the considerable variation of gastrointestinal absorption from individual to individual, values off,, the fraction of a stable element reaching body fluids after its entry into the gastrointestinal tract, are given to one significant figure only (Section 6.2, Chapter 6). For inhalation, values of AL1 and DAC are given for each different inhalation class (D, W and Y) appropriate for various compounds of the element (Chapter 5).

In the metabolic data, when the symbol I is used for time its unit is always the day unless specified otherwise.

Values of AL1 (Bq) are given for the oral and inhalation routes of entry into the body. It is emphasized that the limit for inhalation is the appropriate AL1 and that the values of DAC (Bq/m3) for a 40-h working week are given only for convenience and should always be used with caution (Section 3.4, Chapter 3). Values of AL1 for inhalation and DAC are for particles with an AMAD of 1 pm. A method of correcting the values for particles of other sizes is described in Section 5.5, Chapter 5 and the required numerical data are given in the Supple- ment to this Volume.

If a value of AL1 is determined by the non-stochastic limit on dose equivalent to a particular organ or tissue, the greatest value of the annual intake that satisfies the Commission’s recommendation for limiting stochastic effects is shown in parentheses beneath the ALI. The organ or tissue to which the non-stochastic limit applies is shown below these two values. When an AL1 is determined by the stochastic limit this value alone is given (Section 4.7, Chapter 4).

All the values of AL1 and DAC given here are for occupationally exposed adults and must be used with circumspection for any other purpose (Chapter 9).


METABOLIC DATA FOR HYDROGEN

1. Metabolism

Data from Reference Man (ICRP, 1975) Hydrogen content of the body 7OOOg

of soft tissue 6300g Daily intake of hydrogen 350 g Water content of the body 42 000 g Daily intake of water, including water of

oxidation 3ooog For a number of soft tissues water comprises about 80% of the mass.

2. Metabolic Models

(a) Elemental tritium

As discussed in Chapter 8 of this report, exposure to elemental tritium in air is limited by consideration of the total dose equivalent received from tritium contained in the lung during any year of practice.

It is emphasized that the limit on exposure to tritiated water is more than four orders of magnitude less than that for elemental tritium and in most cases in practice exposure to tritiated water will be the limiting factor.

(b) Tritiated water

(i) Ingestion. Ingested tritiated water is assumed to be completely and instantaneously absorbed from the gastrointestinal tract and to mix rapidly with the total body water so that, at all times following ingestion, the concentration in sweat, sputum, urine, blood, insensible perspiration and expired water vapour is the same (Pinson and Langham, 1957).

(ii) Inhalation. Exposure to an atmosphere contaminated by tritiated water results in intake of that substance both by inhalation and by absorption through the intact skin.

Osborne (1966) has shown that exposure to an atmosphere contaminated by tritiated water at a concentration of C Bq m- ’ results in the absorption of lo- * C Bq min- ’ through the intact skin. For Reference Man (ICRP, 1975) breathing air at a rate of 0.02 m3 min-’ the rate of inhaling tritiated water is 2 x lo-* C Bq min- ’ and it is assumed that al! of this is absorbed into body fluids. Therefore the total rate of absorption of tritiated water into body fluids is 3 x lo- * C Bq min- 1 or 3.6 x 10’ C Bq in a working year of 2 000 h. It is assumed that this tritiated water is instantaneously distributed uniformly among all the soft tissues of the body.

(iii) Distribution and retention. Data on many humans have indicated that the retention of tritiated water is essentially described by a single exponential over the first months or more (Pinson and Langham, 1957). However, cases have been reported where a second exponential term or even a third have been observed (Snyder et al., 1968; Sanders and Reinig, 1968; Moghissi et al., 1972)

i.e. R(r) = A e-0.693r/Tl +B e_0*69J~/T~ + C e-0.693,/r>

The value of T, is closely related to the turnover of body water. Since this is closely related to fluid intake, considerable variation of T, is to be expected because of variation of personal


habits and ambient temperature (Wylie et al., 1963; Butler and LeRoy, 1965). Values of T1 have been observed in the range 4-18 days; a typical value is 10 days (Butler and LeRoy, 1965) and this is in agreement with the value obtained for Reference Man (ICRP, 1975) who has an intake of 3 000 g water per day and contains total body water of mass 42 000 g.

42000 T, =0.693 x 3ooo N 10 days

The second and third exponential terms suggest the presence of tritium in compartments other than the body water, and indeed, organically bound tritium has been demonstrated in animals chronically exposed to tritiated water (Evans, 1969). However, it may be estimated from the data (Snyder et al., 1968; Sanders and Reinig, 1968; Moghissi er al., 1972) that such pools contribute of the order of 10% of the committed dose equivalent to the whole body deriving from an intake of tritiated water and they have been neglected in this report.

It is concluded that values of the committed dose equivalent to body tissues arising from an intake of tritiated water may be estimated from consideration of the retention of tritiated water alone. This view has been confirmed by several authors, e.g. Snyder et al. (1968) and Lambert and Clifton (1967).

Tritiated water is assumed to be uniformly distributed among all soft tissues at any time following intake and its retention to be described by a single exponential with a half-life of 10 days. Thus, the fraction of tritium, taken into the body as tritiated water, which is retained in the body t days later, is given by

R(t) = exp(-0.693t/lO)

and the concentration in soft tissue, C, for a body content of q Bq is given by

C,, = q/63 000 Bq g-’

where 63 000 g is the mass of soft tissues in the body of Reference Man (ICRP, 1975).

(c) Organic compounds

(i) Ingesrion. When tritium-labelled organic compounds are ingested, a considerable fraction may be broken down in the gastrointestinal tract producing tritiated water. Organic compounds of tritium may also catabolize to tritiated water after they have crossed the gut.

In rodents, more than 90% of tritiated thymidine is broken down in the gastrointestinal tract and only about 2 % of the ingested substance is actually incorporated into DNA (Lambert and Clifton, 1968). The fraction of tritium incorporated into DNA after ingestion of tritiated thymidine is about one-fifth of that after the direct entry of tritiated thymidine into blood (Feinendegen and Cronkite, 1977). The fractional absorption of other nucleic acid precursors and of most other tritiated compounds into the blood is not known; in most instances it is probably greater than that of tritiated thymidine.

(ii) Inhalation. Many organic compounds of tritium are not very volatile under normal circumstances and the probability of their being inhaled as vapours is, therefore, small. In circumstances where they might be inhaled it would be prudent to assume that once they enter the lungs they are instantaneously and completely translocated to blood without changing their chemical form.

(iii) Distribution and retention. Tritiated organic compounds which are metabolic precursors are usually distributed throughout the soft tissues and only rarely specifically concentrate in particular cells. A notable exception is tritiated thymidine which, if not catabolized, is


taken up only by the nuclei of those cells synthesizing DNA. In mice, about 45 % of all DNA- synthesizing cells in the body are located in the lining of the gastrointestinal tract, about 15 % are found in the bone marrow and the rest are distributed mainly among spleen, lymphatic tissue, skin and parenchyma (Hughes er al., 1964). These cells take up approximately 30% of the tritiated thymidine that enters the blood (Feinendegen et al., 1973), and the clearance rate from the blood corresponds to a half-life of less than 1 h. Following ingestion, although the efficiency of incorporation of the compound is reduced by a factor of about 5, the relative distribution among the tissues and cells is considered to be similar to that after uptake from the blood (Feinendegen and Cronkite, 1977).

Some of the cells that incorporate tritiated thymidine into their DNA have a long life span of more than 10 days, renew themselves, and deliver a progeny of differentiated, functionally competent cells. It has been suggested that these are critical ceils for the development of late radiation effects (Cronkite et al., 1973).

Other tritiated compounds that serve as precursors for nucleic acids, are taken up in varying amounts by all nucleated cells in the body and thus become widely distributed throughout soft tissues. In mice these compounds were found to be incorporated from the blood into the tissues that contain large amounts of DNA-synthesizing cells less efficiently than tritiated thymidine (Feinendegen, 1978).

(iv) Absorbed dose to tissues. Average absorbed doses received by organs and tissues in experimental animals have been estimated for a number of organic compounds labelled with tritium; these include folic acid (Lambert and Clifton, 1967), thymidine (Lambert and Clifton, 1968), sex hormones (Vennart, 1969) and corticosteroids (Standeven and Clarke, 1967). Reports of these experiments were reviewed by Vennart (1969) who concluded that, under the radiological protection criteria then accepted, the maximum permissible annual intakes to blood of tritiated thymidine and tritiated folic acid should be about one-third of the maximum permissible annual intake of tritiated water, but that the maximum permissible annual intakes of tritiated sex hormones and tritiated corticosteroids should be about 30 times the maximum permissible annual intake of tritiated water. Consideration of absorbed dose to the cell nucleus (Berry et al., 1966), of estimated biological effectiveness(Lambert, 1969; Bond and Feinendegen, 1966), and of the possibility that the long-lived self-renewing and DNA-synthesizing cells might be the critical cells (Feinendegen and Cronkite, 1977) indicate that the AL1 to blood of tritiated thymidine should be about one-fourth to one-fiftieth of the ALI to blood of tritiated water. Furthermore the AL1 by ingestion of tritiated thymidine might need to be one-tenth of that for tritiated water (Feinendegen and Cronkite, 1977).

In this report, specific values of AL1 are not recommended for organic compounds of tritium but it is noted that they might differ considerably from those for tritiated water and that the value for tritiated thymidine might be as much as 50 times smaller. The exact ratio will depend on the compound and its route of entry into the body. This matter will be kept under review both with regard to any further indications of the relative effectiveness of these compounds as compared with tritiated water, and to their metabolism.

References Berry, R. J.. Oliver, R. and Reiskin. A. B. (1966). Reproductive death of mammalian cells due to beta radia-

tion from incorporated thymidine labelled with ‘H or W. Heulth Phys. 12, 1461-1466. Bond, V. P. and Feinendegen. L. E. (1966). Intranuclear ‘H thymidine: dosimetric, radiological and radiation

protection aspects. Health Phys. 12. 1007-1020. Butler, H. L. and LeRoy, J. H. (1965). Observations of biological half-life of tritium. Health Phys. 11.283485.

68 REPORT OF COMMlTTEE 2

Cronkite, E. P., Robertson, J. S. and Feinendegen, L. E. Somatic and teratogenic effects of tritium. In: Tririm, Eds. Moghissi, A. A. and Carter, M. W. Messenger Graphics Publ., Phoenix, Arizona and Las Vegas, Nevada, 1973.

Evans, A. G. (1969). New dose estimates from chronic tritium exposures. Heulrh P/tys. 16, 57-63. Feinendegen, L. E. (1978). Biological damage from radioactive nuclei incorporated into DNA of cells, impli-

cations for radiation biology and radiation protection. 6th Symposium on Microdosimetry, Brussels, 22-26 May. (CEC, 1979).

Feinendegen, L. E. and Cronkite, E. F. (1977). Effects of micro-distribution of radionuclides on recommended limits in radiation protection, a model. Curr. Top. Rudiar. Res. Q. 12, 83-99.

Feinendegen. L. E.. Heiniger, H. J., Friedrich. G. and Cronkite, E. P. (1973). Differences in reutilization of thymidine in hemopoietic and lymphopoietic tissues of the normal mouse. Cell Tissue Kinef. 6, 573-585.

Hughes, W. L., Commerford. S. L., Gitlin, D.. Krueger, R. C.. Schultze, B., Shah, V. and Reilly, P. (1964). Studies on deoxyribonucleic acid metabolism in oiuo. I. Cell proliferation and death as measured by incorporation and elimination of iododeoxyuridine. Fed. Proc. 23, 640-648.

JCRP Publication 23, Report of the Task Group on Reference Man. Pergamon Press, Oxford, 1975. Lambert, B. E. (1969). Cytological damage produced in the mouse testes by tritiated thymidine, tritiated water

and X-rays. He&h Phys. 17. 547-557. Lambert, B. E. and Clifton, R. J. (1967). Radiation doses resulting from the administration of tritiated folic

acid and tritiated water to the rat. Br. J. Radiol. 40, 56-61. Lambert, B. E. and Clifton, R. J. (1968). Radiation doses resulting from the ingestion of tritiated thymidine by

the rat. Health Phys. 15.3-g. Moghissi, A. A., Carter, M. W. and Bretthauer, E. W. (1972). Further studies on the long term evaluation of the

biological half-life of tritium. He&h Phys. 23, 805-806. Osborne, R. V. (1966). Absorption of tritiated water vapour by people. Health Phys. 12, 1527-1537. Pinson, E. A. and Langham, W. H. (1957). Physiology and toxicology of tritium in man. 1. Appl. Physiol. 10,

108. Sanders, S. M. Jr. and Reinig, W. C. Assessment of tritium in man. In: Diagnosis und Treatment of

Deposited Radionuch’des, Eds. Komberg, H. A. and Nonvood, W. D., pp. 534-542. Excerpta Medica Foun- dation, Amsterdam, 1968.

Snyder, W. S., Fish, B. R., Bernard, S. R., Ford, Mary R. and Muir, 1. R. (1968). Urinary excretion of tritium following exposure of man to HTO-a two exponential model. Phys. Med. Biol. 13,547-559.

Standeven, Margaret and Clarke, D. A. (1967). Estimation of radiation doses to tissues after administration of tritiated corticosteroids to the rat. Br. J. Rudiol. 40, 48-55.

Vennart, J. (1969). Radiotoxicology of tritium and 14C compounds. He&h Phys. 16.429440. Wylie, K. F., Bigler, W. A. and Grove, G. R. (1963). Biological half-life of tritium. He&h Phys. 9, 911-914.

Annual limits on intake, ALI and derived air concentrations, DAC(Bq/m3) (40 h/wk) for isotopes ofhydrogen

Radionuclide Oral Inhalation

3H (Tritiated water)

‘H (Elemental tritium)

ALI 3 x lo9 3 x 109 DAC - 8 x 10’ AL1 - DAC - 2 x10’0


METABOLIC DATA FOR PHOSPHORUS

1. Metabolism

Data from Reference Man (ICRP, 1975).

Phosphorus content of the body of the skeleton of muscle

Daily intake in food and fluids

780 g

7OOg 50 g 1.4 g

(a) Uptake to blood 2. Metabolic Model

Dietary phosphorus is well absorbed from the gastrointestinal tract as are various inorganic compounds of the element (ICRP, 1975; Wiseman, 1964; Honstead and Brady, 1967; Castle et al., 1964). In this reportI, has been taken to be 0.8 for all compounds of the element.

(b) Inhalation classes

The ICRP Task Group on Lung Dynamics (1966) assigned all compounds of phosphorus to inhalation class D except for phosphates of some particular elements which it assigned to inhalation class W. This classification is adopted here and for information concerning the inhalation class appropriate to a phosphate of a particular element the metabolic data for that element, or the task group report, should be consulted.

Inhalation class

D W Y

f 1

8:: -

(c) Distribution and retention

The retention of phosphorus in the body has been reviewed by Jackson and Dolphin (1966). They conclude that the whole-body retention of phosphorus is well described by a ftinction of the form

R(t) = 0.15 e-0.693U0.S + 0.15 e-0.693V2 + 0.40 e-0.693’/19 + 0.30

These four terms have been associated with blood plasma, intracellular fluids, soft tissues and mineral bone respectively (Jackson and Dolphin, 1966; Dyson, 1966).

In this report the model used for the distribution and retention of phosphorus has been based on that proposed by Dyson (1966). Phosphorus entering the transfer compartment is assumed to be retained there with a half-life of 0.5 days. Of this phosphorus 0.15 is assumed to go directly to excretion, 0.15 to intracellular fluids where it is assumed to be retained with a half-life of 2 days, 0.40 to soft tissue where it is assumed to be retained with a half-life of 19 days and 0.30 to mineral bone where it is assumed to be permanently retained. Although permanent retention of phosphorus in mineral bone has been assumed, consideration of the data from Reference Man (ICRP, 1975) indicates a half-life of retention in this tissue of about 1 500 days. However, for dosimetric purposes, permanent retention may be assumed,

70 REPORT OP COMMITTEE 2

since the radioactive half-lives of all isotopes of phosphorus considered in this report are much less than 1500 days.

Phosphorus going either to intracellular fluids or to soft tissues is, for the purposes of dosimetry, assumed to be uniformly distributed throughout all organs and tissues of the body excluding mineral bone.

3. Classification of Isotopes for Bone Dosimetry Stable phosphorus is uniformly distributed in mineral bone. For purposes of dosimetry,

isotopes of phosphorus with radioactive half-lives greater than 15 days are assumed to be uniformly distributed in mineral bone and those with shorter half-lives are assumed to be retained on bone surfaces.

References Castle, J. N., Scott, K. G. and Reilly, W. A. (1964). The skeletal uptake of radiophosphorus (Pss). Am. 1.

Roentgenol. 91, 1128-1131. Dyson, E. D. (1966). A specific activity method for derived working limits of phosphorus-32. Heulrh Phye. 12,

1521-l 526. Honstead, J. F. and Brady, D. N. (1967). The uptake and retention of ‘lP and “Zn from the consumption of

Columbia River tish. Health Phys. 13.455463. ICRP Task Group on Lung Dynamics (1966). Deposition and retention models for internal dosimetry of the

human respiratory tract. Health Phys. 12, 173-207. ICRP Publication 23, Task Group Report on Reference Man. Pergamon Press, Oxford, 1975. Jackson, S. and Dolphin, G. W. (1966). The estimation of internal radiation dose from metabolic and urinary

excretion data for a number of important radionuclides. He&h Phys. 12.481-500. Wiseman, G. Absorption from the Intestine, pp. 241-243. Academic Press, London, 1964.

Annual limits on intake, ALI and derived air concentrations, DAC@q/m3) (40 h/wk) for isotopes of phosphorus

Inhalation

Radionuclide

=P

’ 'P

Oral Class D ClasSW

fl = 8 x 10-r f, = 8 x 10-r f, = 8 x IO-’

AL1 2 x 10’ 3 x 10’ 1 x 10’ DAC - 1 x l(r 6 x 10’ AL1 2 x 10’ 3 x 10’ 1 x 10s DAC - 1 x 10’ 4 x IO’


METABOLIC DATA FOR MANGANESE 1. Metabolism


Manganese content of the body of the skeleton of muscle


12 mg 5.2 mg 1.5 mg 3.7 mg

2. Metabolic Model

(a) Uptake to blood

The evidence suggests that inorganic compounds of manganese are only poorly absorbed from the gastrointestinal tract (Cotzias, 1962; Underwood, 1971; Furchner et al., 1966). Since the fractional absorption of MnCl, from the gastrointestinal tract of man is in the range 0.06 to 0.16 for normal subjects (MacDonald and Figueroa, 1969), fl has been taken to be 0.1 for all compounds of manganese.


The ICRP Task Group on Lung Dynamics (1966) assigned oxides, hydroxides, halides and nitrates of manganese to inhalation class W and all other compounds of the element to inhalation class D. Studies on dogs (Morrow et al., 1964) are in agreement with this classification and it has been adopted here.

Inhalation

ClaSS

D W Y

A

0.1 0.1 -


The metabolism of manganese shows considerable inter-species variation (Furchner et al., 1966; Mahoney and Small, 1968) and is also influenced by the levels of manganese and iron in the diet (Mahoney and Small, 1968; Britton and Cotzias, 1966). Volunteer human studies by Mahoney and Small (1968) have shown that the whole-body retention of manganese in the first 60 days post-injection may be described by

R(r) = 0.3 e-O.693,/4 + 0.7 e-0.693c/3s

Because of the considerable differences in manganese metabolism between species, measurements of tissue distributions of manganese in the rat (Furchner et al., 1966) cannot be directly extrapolated to man. However, these measurementsdo indicate that bone and brain retain manganese considerably more avidly than do most other organs and tissues of the body. In this report, of manganese entering the transfer compartment, 0.35 is assumed to go to bone and be retained there with a half-life of 40 days. Of the remainder, fractions 0.1 and 0.15 are assumed to go to the liver and be retained there with half-lives of 4 and 40 days respectively


while fractions 0.2 and 0.2 are assumed to become uniformly distributed among all other organs

and tissues of the body and to be retained in those organs and tissues with half-lives of 4 days and 40 days respectively.

3. Classification of Isotopes for Bone Dosimetry

There appears to be no information available concerning the microdistribution of manganese

in the skeleton. However, the comparatively rapid clearance of manganese from rat bone (Furchner et of., 1966) suggests that the element is predominantly loosely bound to bone surfaces. Thus, for the purposes of dosimetry, all isotopes of manganese considered in this report are assumed to be uniformly distributed over bone surfaces at all times following their deposition in the skeleton.

References B&on, A. A. and Cotzias, G. C. (1966). Dependence of manganese turnover on intake. Am. 1. Phydol. 211,

203-206. Cotzias, G. C. Manganese. In: Mineral Metabolism, Vol. 2, Part B, Eds. Comar, C. L. and Bronner, F. pp. 403-

442. Academic Press, New York, 1962. Furchner, J. E., Richmond, C. R. and Drake, G. A. (1966). Comparative metabolism of radionuclides in mam-

mals-111. Retention of manganese-54 in the mouse, rat, monkey and dog. Health Phys. 12, 1415-1423. ICRP Publication 23, Report of the Task Group on Reference Man. Pergamon Press, Oxford, 1975. ICRP Task Group on Lung Dynamics (1966). Deposition and retention models for internal dosimetry of the

human respiratory tract. Health Phys. 12. 173-207. MacDonald, N. S. and Figucrua, W. G. Manganese absorption and retention in humans. In: Semi-Annual

Progress Report for the Period Ending 30 June 1969. pp. 51-52. California University, Laboratory of Nuclear Medicine and Radiation Biology, Los Angeles. (UCLA-12-724). 1969.

Mahoney, J. P. and Small. W. J. (1968). Studies on manganese III. The biological half-life of radiomanganese in man and factors which affect this half-life. J. C/in. Inuest. 47, 643-653.

Morrow, P. E., Gibb, F. R. and Johnson, L. (1964). Clearance of insoluble dust from the lower respiratory tract. Health Phys. 10.543-555.

Underwood, E. J. Trace Elements in Human and Animal Nutrition, 3rd cd., pp. 177-207. Academic Press, London. 1971.

Annual limits on intake, ALI( and derived air concentrations, DAC(Bq/m’) (40 h/wk) for isotopes of manganese

Inhalation

Oral Class D Class w

Radionuclide

“Mn

5zMn

)I-Mn

“Mn

‘.Mn

‘6Mn

AL1 DAC AL1 DAC ALI

DAC AL1

DAC AL1 DAC AL1 DAC

f, =l x 10-l

7 x 10”

3 x107 -

1 x 109 (1 x 109) ST Wall

- 2 x 109

- 7 x 10’

- 2 x 10’

-

fi = 1 x 10-l /1-l x10-1

2 x lo9 2 x 109 8 x 10’ 9 x 10’ 4 x 10’ 3 x 10’ 2 x l(r 1 x lV 3 x 109 4 x 109

1 x 106 2 x 10’ 5 x 10’ 4 x 10’

(9 x 10’) Bone surf.

2 x IO’ 2 x 10’ 3 x 10’ 3 x 10’ 1 x lo* 1 x 10’ 6 x 10’ 8 x lV 2 x 10’ 3 x 10’


METABOLIC DATA FOR COBALT

1. Metabolism


Cobalt content of the body of the liver


1.5 mg 0.11 mg 0.30 mg

2. Metabolic Model

(a) Uptake to blood

Data from balance studies with humans indicates a fractional absorption of dietary cobalt from the gastrointestinal tract of between 0.2 and 0.95 (Harp and Scoular, 1952; Engel et al., 1967; Hubbard et al., 1966; Valberg et al., 1969). Absorption of radioactive cobalt, usually administered as CoCl,, is generally rather less and appears to be related to the amount of cobalt administered (Paley and Sussman, 1963; Smith et of., 1972). Evidence obtained from the inhalation of cobalt oxide by man (Sill et al., 1964) suggests that there is only minimal absorption of this compound from the gastrointestinal tract.

Because of the above considerations fl is taken to be 0.3 for organically complexed compounds of cobalt and for all inorganic compounds of the element, except oxides and hydroxides, in the presence of carrier material. fi is taken to be 0.05 for oxides and hydroxides of cobalt and for all other inorganic compounds of the element ingested in tracer quantities.


The ICRP Task Group on Lung Dynamics (1966) assigned oxides, hydroxides, halides and nitrates of cobalt to inhalation class W and all other compounds of the element to inhalation class D. Experiments on mice (ICRP, 1966) and dogs (Barnes, et al. 1976) are in agreement with this classification. However, experience in man (Newton and Rundo, 1971) indicates that insoluble compounds of cobalt may be much more tenaciously retained in the lung than this classification suggests. For this reason, in this report, oxides, hydroxides, halides and nitrates of cobalt are assigned to inhalation class Y and all other compounds of the element are assigned to inhalation class W. Evidence from experiments on man (Paley and Sussman, 1963; Smith et al., 1972) suggests that tracer quantities of cobalt are only poorly absorbed from the gastrointestinal tract. Further, the experimental data suggest that cobalt oxide translocated from the lung to the gastrointestinal tract is only poorly absorbed (Sill et al., 1964). For these reasons fi is taken to be 0.05 for cobalt reaching the gastrointestinal tract following inhalation.

Inhalation ChSS fl

D - W 0.05 Y 0.05

74

(c) Disrribution and retention

REPORT OF COMMlTTFeE 2

Although there have been suggestions (Hollins and McCullough, 1971) that bone might be the largest reservoir for long-term cobalt retention, experiments in which 6oCo was administered daily in drinking water to rats showed that the concentration of the radionuclide in bone was always less than in liver (Thomas et al., 1976). It is assumed here that, except for the liver, systemic cobalt is uniformly distributed throughout the body. The concentration of cobalt in the liver is taken to be four times that in the rest of the body.

The following retention function, derived from the data of Smith et al. (1972) on two human subjects followed for 1000 days, is assumed to apply to all systemically distributed cobalt

R(~) = 0.5 e-0.693r/0.S + 0.3 e-0.693r/6 + 0.1 e-O.693t/6O + 0.1 e-0.693f/600

This function is quite similar to that derived by Letourneau et al. (1972) based on more subjects followed for a shorter time.

Therefore, of cobalt entering the transfer compartment a fraction 0.5 is assumed to go directly to excretion, 0.05 to the liver and 0.45 to all other organs and tissues of the body amongst which it is assumed to be uniformly distributed. Of cobalt translocated from the transfer compartment to any tissue of the body, fractions 0.6, 0.2 and 0.2 are assumed to be retained with biological half-lives of 6, 60 and 800 days respectively. Cobalt is assumed to be retained in the transfer compartment with a half-life of 0.5 days.

3. Classification of Isotopes for the Purpose of Bone Dosimetry

Systemic cobalt is assumed to be uniformly distributed throughout all organs and tissues of the body other than the liver. Therefore, a classification of isotopes of the element for purposes of bone dosimetry is not required.

References

Barnes, J. E., Kanapilly, G. M. and Newton, G. J. (1976). Cobalt-60 oxide aerosols: methods of production and short-term retention and distribution kinetics in the beagle dog. Health Phys. 30, 391-398.

Engel, R. W., Price, N. 0. and Miller, R. F. (1967). Copper, manganese, cobalt and molybdenum balance in pre-adolescent girls. /. Nurr. 92, 197-204.

Harp, M. J. and Scoular. F. I. (1952). Cobalt metabolism of young college women on self-selected diets. J. Nutr. 41, 67-72.

Hollins, J. G. and McCullough, R. S. (1971). Radiation dosimetry of internal contamination by inorganic compounds of cobalt: an analysis of cobalt metabolism in rats. Healrh Phys. 21.233-246.

Hubbard, D. M., Creech, F. M. and Cholak, J. (1966). Determination of cobalt in air and biological material. Arch. Environ. Health 13, 190-194.

ICRP Publicorion 23, Report of the Task Group on Reference Man. Pergamon Press, Oxford, 1975. ICRP Task Group on Lung Dynamics (1966). Deposition and retention models for internal dosimetry of the

human respiratory tract. Health Phys. 12, 173-207. Letoumeau, E. G., Jack, G. C., McCullough. R. S. and Hollins, J. G. (1972). The metabolism of cobalt by the

normal human male: whole body retention and radiation dosimetry. Health Phys. 22.451-459. Newton, D. and Rundo, J. (1971). The long-term retention of inhaled cobalt-60. Health Phys. 21, 377-384. Paley, K. R. and Sussman, E. S. (1963). Absorption of radioactive cobaltous chloride in human subjects.

Metabolism 12, 975-982. Sill, C. W,, Anderson, J. I. and Percival, D. R. Comparison of excretion analysis with whole-body counting

for assessment of internal radioactive contaminants. In: Assessment of Radioactivity in Man. Vol. 1. pp. 217- 229. IAEA, Vienna, 1964.

Smith, T., Edmonds, C. 1. and Barnaby, C. F. (1972). Absorption and retention of cobalt in man by whole- body counting. Health Phys. 22. 359-367.

Thomas, R. G., Furchner, J. E., London, J. E., Drake, G. A., Wilson, J. S. and Richmond, C. R. (1976). Comparative metabolism of radionuclides in mammals- X. Retention of tracer-level cobalt in the. mouse, rat, monkey and dog. Health Phys. 31, 323-333.

Valberg, L. S., Ludwig, J. and Olatunbosun, D. (1969). Alteration in cobalt absorption in patients with disorders of iron metabolism. Gasfroenferology 56, 241-251.


Annual limits on intake, ALI( and derived air concentrations, DAC @q/m’) (40 h/wk) for isotopes of cobalt

Inhalation

Oral Class W Class Y

Radionuclide f, = 5 x 10-s f, = 3 x 10-l /, = 5 x 10-1 f, = 5 x 10-z

=co ALI 4 x 10’ 6 x 10’ 1 x 108 1 x 10’ DAC - - 4 x 104 4 x 104

wo AL1 2 x 10’ 2 x 10’ 1 x 10’ I x 10’ DAC

“CO ALI 3 P-10” - 5 x 10’ 3 x 10’

2 x 10’ 1 x 10’ 2 x 10’ DAC - - 4 x 104 1 x lo4

“Co AL1 6 x 10’ 5 x 10’ 4 x 10’ 3 x 10’ DAC -

xl09 2 x 104 1 x lo4

‘8mC0 ALI 2 x 109 2 3 x 109 2 x lo9 DAC

ZOV - 1 x 106 1 x 106

60co ALI 2 7 x lo6 6 x lo6 1 x lo6 DAC 3 x 10’ 5 x 10’

=?zO ALI 4 XlO~O 4 x1010 1 x lo” 1 x 10” (‘s;$a;;) (y&V

DAC - - 6 x 10’ 4 x 10’ S’CO AL1 7 x 10’ 8 x 10’ 2 x 109 2 x 109

DAC - - 1 x 106 9 x 10’ 6%0 ALI 1 x 109 1 x 109 6 x lo9 6 x lo9

(2 x 109) ST Wall

DAC - 3 x 10’ 2 x IO’


METABOLIC DATA FOR KRYPTON No metabolic model is proposed for krypton. As explained in Chapter 8 of this report,

exposure in a cloud of a radioactive noble gas is mainly determined by external irradiation, since dose-equivalent rates from gas absorbed in tissue or contained in the lungs will be negligible in comparison with the dose-equivalent rates to organs and tissues from external irradiation. The recommended DACs for krypton are, therefore, based on consideration of external irradiation only.

Derived air concentrations DAC (Bq/m3) (40 h/wk) for isotopes of krypton

Radionuclide

‘*Kr

76Kr “Kr

79Kr *‘Kr

OlrnKr

ssmKr

*sKr

O’Kr

O*Kr

Semi-infinite cloud 1000 m3 room 500 m3 room 100 m3 room

1 x 105 1 x IO6 1 x 106 1 x 106 (3 x 10”) (3 x 106)

Skin Skin (6 tk;06)

3 x 105 7 x lo6 9 x 10” 2 x 10’ 1 x 10s 2 x 106 2 x 106 2 x 106

(3 x 106) (4 x 10”) (7 x 106) Skin Skin Skin

6 x 10’ 1 x IO’ 2 x 10’ 3 x 10’ 2 x 10’ 2 x 10’ 2 x 10’ 3 x 10’

(5 ,x,,‘;Y (6 ;e;;“) (9 x IO’) Lens

9 x 108 9 x 108 9 x 10’ 9 x 10’ (7 x 109) (7 x 109) (7 x 109) (8 x 109)

Lens Lens Lens Lens 8 x lo5 5 x 10” 5 x 106 5 x 106

(2 x 10’) (3 x IO’) (4 x IO’) Skin Skin Skin

5 x 106 5 x IO6 5 x 10’ 5 x IO6 (5 x 10’) (1 x 109) (1 x 109)

Skin Skin Skin (2 s”kitpg)

2 x 10’ 8 x 10” 8 x lo5 8 x 10s (5 ;kE6) (6 x 10”) (1 x 10’)

Skin Skin 7 x lo4 2 x lo6 2 x 106 3 x 106

(4 x 106) Skin


METABOLIC DATA FOR STRONTIUM 1. Metabolism


Strontium content of the body of the skeleton of soft tissues


0.32 g 0.32 g 3.3 mg 1.9 mg

(a) Uptake to blood

2. Metabolic Model

Both human and animal data suggest that the fractional uptake of dietary strontium and of soluble salts of the element from the gastrointestinal tract is in the range 0.2 to 0.5 (ICRP, 1975; ICRP, 1973; Snyder et al., 1964; Shimmins, et al. 1967; Taylor, 1967). However, SrTiOb is only poorly absorbed from the gastrointestinal tract (McClellan and Bustad, 1964). In this report, f, has been taken to be 0.3 for soluble salts of strontium and 0.01 for SrTiOJ.


Morrow et al. (1968) have shown that 85SrCl, is rapidly cleared from the lungs. However, SrTiO, is much more tenaciously retained (Fish et al., 1964). In this report soluble compounds of strontium have been assigned to inhalation class D and SrTiO, to inhalation class Y.

Inhalation class fi

D 0.3 W - Y 0.01

(c) Distribution and Retention

A very comprehensive model for the retention of strontium in adults has been developed by the Task Group on Alkaline Earth Metabolism in Adult Man (ICRP, 1973). The total numbers of spontaneous nuclear transformations in soft tissue, cortical bone and trabecular bone during the 50 years following the introduction of 1 Bq of a radioactive isotope of strontium into the transfer compartment can be derived from the retention functions given in the Task Group report (ICRP, 1973).

3. CIassitkation of Isotopes for Bone Dosimetry

As discussed in Chapter 7, isotopes of the alkaline earths with radioactive half-lives of greater than 15 days are assumed to be uniformly distributed throughout the volume of mineral bone, whereas isotopes with radioactive half-lives of less than 15 days are assumed to be distributed in a thin layer over bone surfaces. Thus, “Sr, “Sr and 8gSr are assumed to be distributed throughout the volume of mineral bone and all other radioactive isotopes of stron-


tium are assumed to be distributed over bone surfaces at all times following their deposition in the skeleton.

References Fish, Il. R.. Bernard, S. R.. Royster. G. W. Jr., Farabee, L. B., Brown, P. E. and Patterson, G. R. Jr,

Applied internal dosimetry. In: Health Physics Division Annual Progress Report Period Ending 31 July 1964 pp. 222-226. Oak Ridge National Laboratory (ORNL 3697). 1967.

ICRP Publication 20, Task Group Report on Alkaline Eorrlr Metabolism in Adult Mon. Pergamon Press, Oxford, 1973.

ICRP Publicotion 23, Report of the Task Group on Reference Man. Pergamon Press, Oxford, 1975. McCiellan, R. 0. and Bustad, L. K. Gastrointestinal absorption of Sr” titanate. In Hot&rdBiofogy Research

Annual Report for 1963, pp. 100-101. Washington (HW 80500). 1964. Morrow, P. E., Gibb, F. R., Davies, H. and Fisher, M. (1968). Dust removal from the lung parenchyma: an

investigation of clearana simulants. Toxicol. Appl. Pharmocol. 12. 372-396. Shimmins, J.. Smith, D. A., Nordin, B. E. C. and Burkinshaw, L. A comparison between calcium-45 and

strontium-85 absorption, excretion and skeletal uptake. In: Strontium Metabolism, Eds. Let&n. J. M. A., Loutit, J. F. and Martin, J. H., pp. 149-159. Academic Press, London, 1967.

Snyder, W. S., Cook, M. J. and Ford, M. R. (1964). Estimates of (MPC), for occupational exposure to SrPo, Sra9 and Sr”. Health Phys. 10, 171-182.

Taylor, D. M., The role of oxidative phosphorylation in calcium and strontium absorption from the gastrointestinal tract. In: Strontium Metabolism, Eds. Lenihan. J. M. A., Loutit, 1. F. and Martin. J. H.. pp. 175-180. Academic Press, London, 1967.

Annual limits on intake, ALI and derived air concentrations, DAC(Bq/m)) (40 h/wk) for isotopes of strontium

Inhalation


Radionuclide /, = 3 x 10-L /, = 1 x 10-Z /I = 3 x 10-l f, = 1 x 10-a

‘OSr AL1 4 x 10’0 4 x 10’0 8 x 1O’O 9 x 10’0 DAC 3 x 10’ 4 x 10’

“Sr AL1 9 x10” 9 x10’ 3 x 109 3 x 109 PAC - 1 x lo6 1 x 10’

*‘Sr AL1 1 x 10’ 8 x\07 3 x 10s 1 x 10’ DAC -

;109 1 x 10’ 5 x 104

‘lmSr AL1 8 x lo9 8 2 x 10‘0 3 x 10’0 DAC - 9 x lo6 1 x 10’

a’& AL1 9 x 10’ 1 x10’ 1 x 10’ 6 x 10’ DAC - - 4 x 104 2 x l(r

‘lrSr AL1 2 x 109 1 x 109 5 x 109 6 x IO9 DAC - 2 x 10’ 2 x 10’

s9Sr AL1 2 x 10’ 2 x107 3 x 10’ 5 x 10’ (2 x 10’) LLI Wall

DAC - - 1 x lo4 2 x 10’ 9OSr AL1 1 x 10’ 2 x 10’ 7 x 10’ 1 x IO’

(1 x 106) (8 x IO’) Bone surf. Bone surf.

DAC - 3 x 10’ 6 x 10’ “Sr AL1 8 x 10’ 6 FIO1 2 x 10” 1 x 10’

DAC - 9 x l(r 5 x lV 9?lr AL1 1 x 10’ 1 x10’ 3 x 10’ 2x10’

DAC - - 1 x 10s 1 x 10’


METABOLIC DATA FOR ZIRCONIUM

1. Metabolism

Reference Man (ICRP, 1975) lists a total soft-tissue content of 420 mg of zirconium but gives no information on the skeletal content. Studies on animals suggest that bone contains the bulk of body zirconium (Thomas et ul., 1971; Schiessle et al., 1961).

Daily intake in food and fluids (ICRP, 1975) 4.2 mg

2. Metabolic Model

(a) Uptake to blood

In comparative studies on the gastrointestinal absorption of various zirconium compounds in rats Fletcher (1969) found that the fractional uptake varied from 0.000 3 to 0.002. Shiraishi and Ichikawa (1972) report similar results in adult rats but find gastrointestinal absorption in juvenile animals to be considerably greater. In this report f, is taken to be 0.002 for all compounds of zirconium.


The ICRP Task Group on Lung Dynamics (1966) assigned zirconium carbide to inhalation class Y, oxides, hydroxides, halides and nitrates of zirconium to inhalation class W and all other compounds of the element to inhalation class D. Experiments on mice (Thomas et ul., 1971) and experience in man (Cofield, 1963; Waligora, 1971) are in agreement with this classification and it has been adopted here.

In the absence of any experimental dataf, has been taken to be 0.002 for all compounds of zirconium entering the gastrointestinal tract following inhalation.

Inhalation class J-1

D 0.002 W 0.002 Y 0.002


Furchner et al. (1964) have described the whole-body retention of intraperitoneally injected zirconium citrate in mice over a period of 420 days post-injection by the equation

R(t) = 0.5 e-0.693r/0.5 + 0.2 e-0.693t/7 + 0.3 e-0.693tf7 700

In man there is no evidence of any rapid loss of zirconium from the body following intravenous injection (Mealey, 1957). The short-term component of the retention function given by Furchner et al. (1964) has been dropped and a whole-body retention function appropriate to man has been assumed to be

R(t) = 0.5 e-O.693t/7 + 0.5 e-0.693t/8 000

The long-term component of this function has been assumed to be associated with mineral bone and the,skrt-term component with all other organs and tissues of rhe body. Thus, of


zirconium entering the transfer compartment 0.5 is assumed to be translocated to bone where it is retained with a biological half-life of 8 000 days and 0.5 is assumed to be translocated to all other organs and tissues of the body where it is. retained with a biological half-life of 7 days.

3. Classification of Isotopes for Bone Llosimetry

Zirconium is not a major constituent of mineral bone, nor has it any particular chemical affinity with the alkaline earth elements. For these reasons zirconium is, for the purposes of dosimetry, assumed to be uniformly distributed over bone surfaces at all times following its

deposition in the skeleton.

References Cofield, R. E. (1963). In ciao gamma spectrometry for inhalations of neptunium-237-protactinium-233, cobalt-

60, and zirconium-9%niobium-95 Health Phys. 9, 283-292. Fletcher, C. R (1969) The radiological hazards of zirconium-95 and niobium-95 Health Phys. 16, 209-228. Furchner. J E., Richmond. C. R. and Trafton, G. A. Retention of zirconium 9s after oral and intraperitoneal

administration to mice. In: Biological and Medical Research Group (H4) of the Health Division AnnualReport July 1962 through June 1963. pp. 53-58. Los Alamos Scientific Laboratory (LAMS-3034), 1964.

JCRP Publication 23, Report of the Task Group on Reference Man. Pergamon Press, Oxford, 1975. ICRP Task Group on Lung Dynamics (1966). Deposition and retention models for internal dosimetry of the

human respiratory tract. Health Phys. 12, 173-207. Mealey, J. (1957). Turnover of carrier-free zirconium-89 in man. Nature 179, 673674. Schiessle. W.. Philipp. K., Jung-Langhorst, U. and Schmidtke. I. (1961). Tierexperimentelle inhalationsunter-

suchungen mit radioactivem zirkonium. Sfrahlentherapie 116, X6-584. Shiraishi, Y. and Ichikawa. R. (1972). Absorption and retention of

and adult rats. Health Phys. 22. 373-378. 14*Ce and 9sZr-9sNb in newborn. juvenile

Thomas, R. G.. Walker, S. A. and McClellan, R. 0. (1971). Relative hazards for inhaled 9’Zr and PsNb particles formed under various thermal conditions. Proc. Sot. Exp. Biol. Med. 138,228-234.

Waligora, S. J. (1971). Pulmonary retention of zirconium oxide (“Nb) in man and beagle dogs. Health Phys. 20.89-91.

Annual limits on intake, ALI and derived air concentrations, DAC (Bqlm”) (40 h/wk) for isotopes of zirconium

Inhalation

Oral Class D Class W Class Y

Radionuclide f, = 2 x 10-S f, = 2 x lo-’ fi = 2 x IO-3 f, = 2 x IQ-’

s6Zr

“Zr

09Zr

93Zr

ALI DAC AL1 DAC AL1 DAC ALI

95Zr DAC AL1

9’Zr DAC AL1 DAC

5 x 10’ -

I x 108 -

6 x 10’

5 x107 (1 x 100) Bone surf.

- 5 x 10’

- 2 x 10’

-

1 x 10’ 6 x 10’ 8 x lo6 3 x 10’ I x 10’ 5 x IO4 2 x 10s

(6 x 10’) Bone surf. 1 x 10’ 5 x 10’

(1 x IO’) Bone surf. 2 x 10s 7 x 10’ 3 x 104

1 x 10’ 4 x lo* 2 x 10’ 7 x 10’ 9 x 10’ 4 x 10. 9 x 10s

(2 x 106) Bone surf. 4 x 102 1 x 10’

6 x 10’ 5 x 10’ 2 x 104

9 x IO’ 4 x 104 1 x 10’ 5 x IO’ 9 x 10’ 4 x 10’ 2 x 106

(3 x 106) Bone surf. 9 x 10’ 1 x 10’

4 x 10’ 5 x 10’ 2 x IO’


METABOLIC DATA FOR NIOBIUM

1. Metabolism

Data from Reference Man (ICRP, 1975). Niobium content of soft tissue Daily intake in food and fluids

110 mg 0.62 mg


The observed level of urinary excretion of niobium argues for a fractional absorption of dietary niobium from the gastrointestinal tract of at least 0.6 (Schroeder and Balassa, 1965). The data on intake and soft-tissue content given in Reference Man (ICRP, 1975) also indicate that a large fraction of dietary niobium is absorbed from the gastrointestinal tract. However, studies in small animals with a number of compounds of the element have indicated that the fractional absorption from the gastrointestinal tract is 0.01 or less (Fletcher, 1969; Furchner and Drake, 1971). In this report I, has been taken to be 0.01, since this value is considered appropriate for occupational exposure to compounds of niobium.


Inhaled niobium oxide is tenaciously retained in the lung (Thomas et al., 1971; McClellan, 1968). The oxalate is more rapidly lost from the lung than zirconium oxalate (Thomas et al., 1971), but a substantial fraction is retained by dogs with a biological half-life of greater than 300 days (McClellan, 1968).

Oxides and hydroxides of niobium are assigned to inhalation Class Y and all other compounds of the element to inhalation class W.

Inhalation class J-1

& -

0.01 Y 0.01


Furchner and Drake (1971) have compared the metabolism of niobium in the mouse, rat, monkey and dog and have reviewed the earlier literature. Their studies show preferential retention in mineral bone, with a concentration 10 times the whole-body average, and in kidney, spleen and testis, with concentrations 3 to 5 times the whole-body average.

In monkeys and dogs they found that the retention of niobium following a single intravenous injection was well described by a retention function of the form

R(r) = ae-‘J.69sUT, + (l _ ,.+-0.693Va

The data of Furchner and Drake (1971) indicate a value for a of 0.5, a value for T1 of 6 days and a value for T2 ranging from 100 days in dogs to 460 days in mice. Since all organs and tissues appear to lose niobium at the same rate (Furchner and Drake, 1971) the same retention function is assumed appropriate for each of them. Thus, of niobium entering the transfer compartment, fractions 0.71, 0.018, 0.01, 0.002 and 0.26 are assumed to go to mineral bone,

a2 REPORT OF COMMITTEE 2

kidneys, spleen, testis and all other tissues respectively. Of niobium entering any one of these organs or tissues 0.5 is assumed to be retained with a half-life of 6 days and the other 0.5 is assumed to be retained with a half-life of 200 days.

3. Classification of Isotopes for Bone Lhximetry

Although there have been autoradiographic studies in which the gross distribution of niobium in the skeleton has been measured (Blckstrom et al., 1967) no more detailed work appears to have been performed. In this report, p3mNb and 94Nb, the longest-lived radioactive isotopes of niobium, are assumed to be uniformly distributed throughout the volume of mineral bone at all times following their deposition in that tissue, whereas all other radioactive isotopes of the element are assumed to be distributed over bone surfaces at all times following their deposition in the skeleton.

References Bgckstriim. J., Hammarstrom. L. and Nelson, A. (1967). Distribution of ztrconium and niobium in mice:

autoradiographic study. Acfu. Radiol. 6, 122-128. Fletcher, C. R. (1969). The radiological hazards of zirconium-95 and niobium-95 Health Phys. 16,209-220. Furchner, J. E. and Drake, G. A. (1971). Comparative metabolism of radionuclides in mammals-VI. Reten-

tion of 95Nb in the mouse, rat, monkey and dog. Health Phys. 21. 173-180. ICRP Publication 23, Report of the Task Group on Reference Man. Pergamon Press, Oxford, 1975. McClellan, R. 0. Fission Product Inhalation Program. Annual Report 1967-1968, pp. 115-121. Lovelace

Foundation, Albuquerque (LF 39). 1968. Schroeder, H. A. and Balassa, J. J. (1965). Abnormal trace metals in man: niobium. 1. Chron. Dis. 18.229-241. Thomas, R. G., Walker, S. A. and McClellan. R. 0. (1971). Relative hazards for inhaled 9sZr and 95Nb

particles formed under various thermal conditions. Proc. Sot. Exp. Biol. Med. U&228-234.

Annual limits on intake, ALI( and derived air concentrations, DAC (Bq/m3X40 h/wk) for isotopesofniobium

Inhalation

Radionuclide

Oral

f, = 1 x 10-Z

Class W classy

f, = I x 10-Z fi = 1 x 10-2

**Nb ALI 2 x 109 8 x lo9 8 x IO9

*9Nb

KZin) (I22 min) 90Nb

9’mNb

DAC ALI DAC ALI DAC ALI DAC ALI

DAC AL1 DAC AL1 DAC ALI

DAC AL1 DAC AL1 DAC AL1 DAC

- 4 x 10’

- 2 x 106

- 4 x 10’

3 x10’ (4 x 10’) LLI Wall

- 4 x 10’

- 8 x 10’

- 8 x 10’

(9 x 10’) LLI Wall

- 4 x 10’

8 x:0’ -

5 x l(r -

4 x 106 2 x 109 6 x lo5 7 x 108 3 x 10’ 1 x 10” 4 x 104 5 x 10’

2 x 10’ 7 x 10’ 3 x 10’ 5 x 10’ 2 x lo* 1 x 10’

4 x 10’ 1 x 10’ 4 x lo4 3 x 109 1 x 106 2 x 109 8 x 10’

3 x 10 1 x 109 6 x 10’ 6 x 10’ 2 x 10’ 9 x 10’ 4 x 1t.r 6 x 10’

3 x 10’ 6 x 10’ 2 x 10’ 4 x 10’ 2 x 10’ 8 x 10’

3 x 10’ 9 x 10’ 4 x 104 3 x 109 1 x 106 2 x 109 8 x 10’

LIMITS FOR INTAKES OF RADIONUCUDES BY WORKERS 83

METABOLIC DATA FOR MOLYBDENUM

1. Metabolism


Molybdenum content of the body Daily intake in food and fluids

9.5 mg 0.3 mg

2. Metabolic Model

(a) Uptake to blood

Molybdenum, as salts of molybdic acid, as MOO, or as CaMoO, is readily absorbed from the gastrointestinal tract (Comar, 1948; Underwood, 1971; Anke et al., 1971). However, MoS, is only poorly absorbed (Underwood, 1971). In this report, fi has been taken to be 0.05 for MO!& and 0.8 for all other compounds of the element.


The ICRP Task Group on Lung Dynamics (1966) assigned oxides and hydroxides of molybdenum to inhalation class Y, sulphides, halides and nitrates to inhalation class W and all other compounds of the element to inhalation class D. Experiments on dogs with ammonium molybdate and molbdenum oxide are in agreement with this classification but molybdenum chloride appears to behave more like a class D compound than a class W (Cuddihy er al., 1969).

In this report all compounds of molybdenum have been assigned to inhalation class D except for oxides, hydroxides and MoS, which have been assigned to inhalation class Y.

Inhalation class

D W Y

/I

0.8 -

0.05


Mice given a salt of molybdic acid were killed at 1 h and 1 day, post-injection. The highest concentrations of the radionuclide were found in the livers and kidneys of these animals (Rosoff and Spencer, 1973). Sheep fed molybdenum in their diet had the greatest concentration of the element in the skeleton, all other organs and tissues having roughly equal concentrations (Dick, 1956). In dogs, molybdenum translocated from the lung following inhalation of various compounds of the element is deposited mainly in liver, skeleton, muscle and kidney with liver and kidney containing the highest concentrations (Cuddihy et al., 1969).

In this report, of molybdenum entering the transfer compartment, fractions 0.3, 0.15, 0.05 and 0.5 are assumed to be deposited in liver, mineral bone, kidney and all other tissues respectively. Retention in all of these tissues is assumed to be governed by the whole-body retention of molybdenum in man as measured by Rosoff and Spencer (1964)

R(t) = 0.1 e-0.693V1 + o_ge-0.69WSO


i.e. of molybdenum translocated from the transfer compartment to any organ or tissue of

the body, fractions 0.1 and 0.9 are assumed to be retained with half-lives of 1 and 50 days

respectively.


None of the radiosiotopes of molybdenum considered in this report has a radioactive half- life of greater than 3 days. For the purposes of dosimetry they are, therefore, considered to be uniformly distributed over bone surfaces at all times following their deposition in the skeleton.

References Anke, M.. Hennig. A., Diettrich, M., Hoffmann, G., Wicke. G. and Pflug, D. (1971). Resorption, exkretion

und verteilung von 99 molybdaen nach oraler gabe an laktierende wiederkauer. Arch. Tiererndhr~21, %5- 513.

Comar, C. L. (1948). Radioisotopes in nutritional trace element studies-I. Nucleonics 3, 32-45. Cuddihy, R. G., Boecker. B. B. and Kanapilly. G. M. Tissue distribution of 99Mo in the beagle dog aRet

inhalation of aerosols of various chemical forms. Lovelace Foundation Fission Product Inhalation Program Annual Report 1968-1969. pp. 117-120. Lovelace Foundation. Albuquerque (LF 41). 1969.

Dick, A. T. In: Inorganic Nitrogen Metabolism, Eds. McElroy, W. D., Glass, B., Johns Hopkins Press, Baltimore, 1956.

ICRP Publication 23. Report of the Task Group on Reference Man. Pergamon Press, Oxford, 1975. ICRP Task Group on Lung Dynamics (1966). Deposition and retention models for internal dosimetry of the

human respiratory tract. Health Phys. 12, 173-207. Rosoff, B. and Spencer, H. (1964). Fate of molybdenum-99 in man. Nature 202,410-411. Rosoff, B. and Spencer, H. (1973). The distribution and excretion of molybdenum-99 in mice. Health Phys.

25.173-175. Underwood, E. J. Trace Elements in Human and Animal Nutrition, 3rd edn. Academic Press, London, 1971.

Annual limits on intake, ALI( and derived air concentrations, DAC(Bq/m’) (40 h/wk) for isotopes of molybdenum

, Inhalation


Radionuclide /, = 8 x 10-l fi = 5 x 10-z f, = 8 x 10-l f, = 5 x 10-z

90Mo AL1 DAC

9=Mo AL1 DAC

9’mMo AL1 DAC

99Mo AL1

DAC ‘O’Mo AL1

DAC

2 x 10’ -

1 x 10’ -

4 x 108 -

6 x 10’

- 2 x 109

(2 x 109) ST Wall

-

7 x 10’ -

9 x 10’ -

2 x 100

4 x107 (4 x 10’) LLI Wall

- 2 x 109

(2 x 109) ST Wall

-

3 x 10’ 1 x 105 2 x 108 8 x lo4 7 x lo* 3 x 10’ 1 x 100

4 x 104 5 x 109

2 x 106

2 x 10’ 7 x 1w 7 x 10’ 3 x 10’ 5 x 10’ 2 x 10s 5 x 10’

2 x lo4 6 x lo9

2 x 106


METABOLIC DATA FOR TELLURIUM 1. Metabolism


Tellurium content of soft tissues 8 mg Daily intake in food and fluids 0.6 mg

The data of Schroeder et al. (1967), indicate a total-body content of about 600 mg of tellurium, with about 90% of this in mineral bone. However, the data of Soman et al. (1970) indicate that there is more than 500 mg of tellurium in soft tissues alone. The reason for this wide range of values is unexplained.


Hollins (1969) has measured the fractional absorption of tellurium compounds from the gastrointestinal tract of the rat. His results suggest thatf, should have a value of between 0.1 and 0.2. These results are in broad agreement with those of Scott et al. (1947) who concluded that the fractional absorption of tetravalent tellurium from the gastrointestinal tract was about 0.25. In this report& has been taken to be 0.2.


The ICRP Task Group on Lung Dynamics (1966) assigned oxides, hydroxides and nitratas of tellurium to inhalation class W and all other compounds of the element to inhalation class D. In the absence of any relevant experimental data this classification has been adopted here.

Inhalation class

D 0.2

: 0.2 -


Hollins (1969) has studied the whole-body retention of tellurium in rats and has reviewed earlier data on this subject. His data, which extend to 200 days after administration, indicate that tellurium is much more tenaciously retained in mineral bone than it is in other tissues.

In this report it is assumed that of tellurium entering the transfer compartment half goes directly to excretion. Of the remainder, 0.25 is assumed to be translocated to mineral bone, where it is retained with a biological half-life of 5 000 days, and 0.25 is assumed to be uniformly distributed among all other organs and tissues of the body, where it is assumed to be retained with a biological half-life of 20 days.

The half-life of clearance from the transfer compartment is assumed to be 0.8 days for all compounds of tellurium (Hollins, 1969).

(d) Behaviour of daughters

Some of the isotopes of tellurium considered in this report decay to yield radioactive iso- lopes of iodine. It is assumed that all iodine produced by the decay of tellurium in the body is

86 REPORT OF COMMITTXZE 2

translocated instantaneously to the transfer compartment in inorganic form. The metabolic behaviour of iodine entering the transfer compartment in this way is assumed to be governed by the metabolic model for iodine described in the dosimetric data for that element.


No data have been found concerning the location of isotopes of tellurium in mineral bone. Since many of these isotopes have short radioactive half-lives and none is greater than 200 days, it is assumed, for purposes of dosimetry, that all isotopes of tellurium are retained on bone surfaces, though the resulting values of AL1 may be unduly restrictive.

References Hollins, 3. G. (1969). The metabolism of tellurium in rats. Heulrh Phys. 17,497-505. ICRP Publication 23, Report of the Task Group on Reference Man. Pcrgamon Press, Oxford, 1975. ICRP Task Group Report on Lung Dynamics (1966). Deposition and retention models for internal dosimetry

of the human respiratory tract. Health Phys. 12, 173-207. Schroeder, H. A., Buckman. J. and Balassa, J. J. (1967). Abnormal trace elements in man: tellurium. J. CIvontc

Dis. 20, 147-161. Scott, K. G., Overstreet, R., Jacobson, L., Hamilton, J. G., Fisher, H., Crowley. J., Chaikoff, I. L., Entenman,

C., Fishler, M., Barber, A. J. and Loomis, F. The Metabolism of Carrier-Free Fission Products in the Rat. Oak Ridge, Tenn., USAEC (MDDC-1275). 1947.

Soman, S. D., Joseph, K. T., Raut, S. J., Mulay, C. D., Parameshwaran. M. and Panday, V. K. (1970). Studies on major and trace element content in human tissues. Health Phys. 19, 641-656.

Annual limits on intake, ALI( and derived air concentrations, DAC@q/m3) (40 h/wk) for isotopes of tellurium

Oral

Inhalation

Class D Class w

Radionuclide f, = 2 x 10-I fi = 2 x 10-t fi = 2 x 10-r

‘lsTe

lilTe

“IraTe

‘13”Te

1 J6bllTe

AL1 DAC AL1 DAC ALI

DAC ALI

DAC AL1

DAC ALI

DAC ALI DAC ALI

3 x 10’ -

1 x 10’

2 x101 (3 x 10’) Bone surf.

- 2 x 10’

(4 x IO’) Bone surf.

- 2 x 10’

(4 x 10’) Bone swf.

4 x107 (5 x 10’) Bone surf.

- 3 x 10’

- 2 x 107

8 x 10” 3 x 10’ 2 x 10’ 6 x 104 7 x lo6


3 x 10’ 7 x 106


3 x 10’ 8 x 10’


3 x 10’ 2 x 10’


6 x 10’ a x lo* 3 x 10x 1 x 10’


4 x 10’

1 x 109 5 x 10’ 1 x 10’ 5 x lo* 2 x lo7

6 x 10” 2 x IO7


7 x 10’ 2 x 10’

a x 103 3 x 10’

1 x 104 6 x 10’ 3 x 10’ 9 x 10’

DAC 4 x 10’


Annual limits on intake, ALI( and &rived air concentrations, DAC(Bq/m3) (40 h/wk) for isotopes of tellurium

Inhalation

Oral Class D Class W

fi = 2 x IO-’ fr 32 x 10-l f, = 2 x IO-’

AL1 1 x lo9 2 x IO9 3 x 109 DAC 1 x 106 1 x 106 AL1 2 x 10’ 2 x 10’ 9 x 10’ DAC - 1 x Iti 4 x 10’ AL1 2 x 10’ 2 x 10” 3 x 10’

(2 x 10’) (;h;ty;) (5 x 10’) Thyroid .Thyroid

DAC 8 x lo* 1 x 10’ ALI 2 x 10’ 2 x 10’ 3 x 10’

(g$?J (g$?$ (%:y’

DAC - 1 x lo4 1 x lo* AL1 2 x 106 a x lo6 7 x 106

(5 x 106) Thyroid (gg??) (&$)

DAC 4 x 10’ 3 x 10’ AL1 5 x 100 7 x lo6 1 x 109

(1 x 109) Thyroid (;lg?;)

(Gr:::) DAC - 3 x 10’ 5 x 10’ ALI 1 x 10’ 1 x lo6 2 x 10’

(2 x lO8) (4 x 10’) (5 x 10’) Thyroid Thyroid Thyroid

DAC - 6 x IO’ 1 x 10s ALI 2 x 10’ 1 x 10’ 3 x 10’

(4 x 106) (4 x 10’) Thyroid Thyroid (z:i

DAC - 5 x lo4 1 x 10’

88 REPORT OF COMMlTTEE 2

METABOLIC DATA FOR IODINE

1. Metabolism

Data from Reference Man (ICRP, 1975)

Iodine content of the body 11 mg of the thyroid 10 mg

Daily intake in food and fluids 0.2 mg The size of the thyroid gland and its uptake of iodine from blood are both very dependent

upon the daily intake of stable iodine (Underwood, 1971; Riggs, 1952; Dolphin, 1971). In this report the fractional uptake of iodine by the thyroid has been taken as 0.3. A thyroid mass of 20 g has been used, appropriate for a fractional uptake to the thyroid of 0.3.

2. Metabolic Model

(a) Uptake to blood

Iodine is absorbed rapidly and almost completely from the gastrointestinal tract (Under- wood, 1971), mainly in the small intestine (Riggs, 1952). In this report, fi is taken to be unity for all commonly occurring compounds of iodine.


The ICRP Task Group on Lung Dynamics (1966) assigned all compounds of iodine to inhalation class D. Data from experiments on mice, rats, dogs and sheep (Willard and Bair, 1961; Bair et al., 1963) are in agreement with this classification and it has been adopted here.

Inhalation class

D W Y

fi

1 - -


Riggs (1952), has described a simple three-compartment model for the metabolism of iodine. This is essentially the model which is used in this report.

Of iodine entering the transfer compartment a fraction, 0.3, is assumed to be translocated to the thyroid while the remainder is assumed to go directly to excretion. Iodine in the thyroid is assumed to be retained with a biological half-life of 120 days and to be lost from the gland in the form of organic iodine. Organic iodine is assumed to be uniformly distributed among all organs and tissues of the body other than the thyroid and to be retained there with a biological half-life of 12 days. One-tenth of this organic iodine is assumed to go directly to faecal excretion and the rest is assumed to be returned to the transfer compartment as inorganic iodine.


Some isotopes of iodine decay to yield radioisotopes of xenon. For dosimetric purposes these isotopes of xenon are assumed to escape from the body without decaying to any appreciable extent.



Organic iodine is assumed to be uniformly distributed throughout all organs and tissues of the body other than the thyroid. Therefore a classification of isotopes of the element for the purpose of bone dosimetry is not required.

References

Bair, W. J.. Snyder, M. D., Walters. R. A. and Keough, R. F. (1963). Effect of I”’ on thyroid uptake of inhaled 1”‘. Heulth Phys. 9, 1399-1410.

Dolphin, 0. W. (1971). Dietary intakes of iodine and thyroid dosimetry. He&h Phys. 21, 711-712. ZCRP Publication 23, Report of the Tak Group on Reference Man. Pergamon Press, Oxford, 1975. ICRP Task Group on Lung Dynamics (1966). Deposition and retention models for internal dosimetry of the

human respiratory tract. Health Phys. 12, 173-207. Riggs, D. S. (1952). Quantitative aspects of iodine metabolism in man. Pharmucol. Rev. 4, 284-370. Underwood, E. 1. Trace Elemenls in Human and Animal Nutrition, 3rd ed. Academic Press, London, 1971. Willard, D. H. and Bair, W. J. (1961). Behaviour of I If‘ following its inhalation as a vapour and as a particle.

Acta Radiol. 55,486-496.

Annual limits on intake, ALI and derived air concentrations, DAC(Bq/m’) (40 h/wk) for isotopes of iodine

Radionuclide

Oral

fl = 1

Inhalation

Class D

/I = 1

,301

1 mhq

“‘1

,331

1341

‘331

,361

,331

ALI

DAC ALI

DAC AL1

DAC AL1

DAC ALI

DAC ALI

DAC ALI

DAC AL1

DAC

1 x 100

(t&Z)

4 x108 (5 x 10’) Thyroid

- 4 x 108

(1 x 109) Thyroid

- 1 x 10’

(4 x 10’) Thyroid

- 2 x 106

(6 x 10”) Thyroid

- 1 x 106

(5 x 106) Thyroid

- 8 x 10’

(3 x 106) Thyroid

- 2 x 109

(2 x 109) ST Wall

-

3 x 10’

(g*?;)

1 x 10’ 8 x 10’

3 x 10) 7 x 100

(2 x 109) Thyroid 3 x los 2 x 108

(7 x 10’) Thyroid 9 x 10’ 3 x 106

(1 x 10’) Thyroid 1 x 10’ 2 x 106

(8 x 10”) Thyroid 1 x 103 1 x 106

(4n;r?;)

5 x 10’ 4 x 109

2 x 106


Annual limits on intake, ALI and derived air concentrations, DAC(Rq/m’) (40 h/wk) for isotopes of iodine

Inhalation

Oral Class D

Radionuclide /I = 1 /I - 1 I J9I AL1 2 x 10’ 3 x 10’

(7x& U&)

DAC - I x 101 ,301 AL1 1 x 10’ 3 x 10’

(4 x 10’) (~7xl&) Thyroid

DAC - 1 x IV 1311 ALI 1 x 106 2 x 106

WJ&) (zoy;)

DAC - 7 x 10’ IS1 ALI 1 x 10’ 3 x 10’

(g$) ($yC)

DAC - 1 x 10’ 1321I AL1 1 x 10’ 3 x 10’

(Z$) (gFz;)

DAC - 1 x 10’ 1331 AL1 5 x 106 1 x 10’

Qp;) Wy;)

DAC - 4 x 10’ 1341 AL1 8 x IO’ 2 x 109

(gg?;)

DAC 1rq AL1 3 x10’

7 x 10’ 6 x 10’

v&) (coy;)

DAC - 2 x lo*


METABOLIC DATA FOR CAE!3UM

1. Metabolism


Caesium content of the body of muscle of bone


2. Metabolic Model

(a) Uptake to blood

Evidence indicates that caesium chloride (LeRoy, et al.

1.5 mg 0.57 mg 0.16 mg

10 Fg

1966; Rosoff et al., 1963) and other commonly occurring compounds of caesium @Roy et al., 1966) are rapidly and almost completely absorbed from the gastrointestinal tract. In this reportf, is taken to be unity for all compounds of the element.


The ICRP Task Group on Lung Dynamics (1966) assigned all compounds of caesium to inhalation class D. The experimental evidence (Miller, 1964; Lie, 1964; Boecker, 1969) is in accord with this classification and it has been adopted here.

Inhalation class f1

D 1

7 - -


The available evidence indicates that caesium is distributed uniformly in the body. In no case is the concentration of caesium in an organ or tissue of Reference Man (ICRP, 1975) greater than the concentration in muscle.

The retention of caesium is adequately represented over at least the first 1400 days by a two- exponential expression

R(t) = oe -0.69%/11 + (1 _ a)e-0.693t/Tz

Values of a have been reported in the range 0.06 to 0.15 (Rundo, 1964), values of T, from 1. to 2 days (Rundo, 1964), and values of T, from 50 to 150 days with isolated cases up to 200 days (Rundo, 1964; Lloyd et al., 1970; Cryer and Baverstock, 1972; Richmond et al., 1962).

In this report, it is assumed that, of caesium entering the transfer compartment, a fraction, 0.1, is translocated to one tissue compartment and retained there with a half-life of 2 days, whereas the remainder is transferred to a second tissue compartment and retained there with a half-life of 110 days. It is also assumed that caesium translocated to these compartments is uniformly distributed throughout the body.

92



“‘Cs decays to ’ “Xe and, to a small extent, to r2’“‘Xe. r2 7mXe, which has a radioactive half-life of 75 s, is assumed to decay at its site of production, whereas ’ 2 ‘Xe, with a radioactive half-life of 36 days, is assumed to escape from the body without dtcaying.


Caesium is assumed to be uniformly distributed throughout all organs and tissues of the body. Therefore, a classification of isotopes of the element for the purposes of bone dosimetry is not required.

References Backer, B. B. (1969). Comparison of “‘Cs metabolism in the beagle dog following inhalation and intravenous

injection. Health Phys. 16, 785-188. Cryer, M. A. and Baverstock, K. F. (1972). Biological half-life of ir7Cs in man. Health Phys. 23, 394-395. ICRP Publication 23, Report of the Task Group on Reference Man. Pergamon Press, Oxford, 1975. ICRP Task Group Report on Lung Dynamics (1966). Deposition and retention models for internal dosimetry

of the human respiratory tract. Health Phys. 12, 173-207. LeRoy, G. V., Rust, J. H. and Hasterlik, R. J. (1966). The consequences of ingestion by man of real and simu-

lated fallout. Health Phys. 12,449-473. Lie, R. (1964). Deposition and retention of i3’Cs in the rat following inhalation of the chloride and the nitrate.

Health Phys. 10, 1071-1076. Lloyd, R. D.. Mays, C. W., Church, B. W., Pendleton, R. C. and Mays, S. F. (1970). Does the elimination rate

of cesium in humans change with the seasons? Health Phys. 18,623-629. Miller, C. E. (1964). Retention and distribution of is7Cs after accidental inhalation. Health Phys. 10,1065-1070. Richmond, C. R., Furchner, J. E. and Langham, W. H. (1962). Long-term retention of radiocesium by man.

Health Phys. 8, 201-205. Rosoff, B., Cohn, S. H. and Spencer, H. (1963). Cesium-137 metabolism in man. Radiat. Res. 19, 643-654. Rundo, J. (1964). A survey of the metabolism of cesium in man. Br. J. Radiol. 37. 108-I 14.

Annual limits on intake, ALI and derivtio;i;~;ntrations, DACQ3q/m3) (40 h/wk) for isotopes

Inhalation

Radionuclide

ma

‘s’cs

199fi

13Q

“Iti

‘sscs

13*cs

1 3*T3

Oral Class D

/I = 1 fr = 1 AL1 2 x 109 5 x lo9

(z%? DAC - 2 x 106 AL1 2 x IO9 4 x 109 DAC 1 x 106 AL1 9 x108 1 x 109 DAC - 5 x lo5 AL1 2 x IO9 7 x lo9

(4 x 109) ST Wall

DAC - 3 x lo6 AL1 g x 10’ 1 x lo9 DAC - 5 x 10’ AL1 1 x 10” 1 x IO’ DAC - 6 x 104 AL1 3 x 10’ 4 x 106 DAC - 2 x 10s AL1 4 x 109 5 x 109

(%%) DAC - 2 x 10’


Annual limits on intake, ALI(Bq and derived air concentrations, DAC(Bq/m’) (40 h/wk) for isotopes of caesium

Inhalation

Oral Class D

Radionuclide f, = 1 fi = 1

1’6cs ALI DAC

1311nCs AL1 DAC

“Ts AL1 DAC

13’)<=s ALI DAC

IQ ALI

DAC

3 x IO’ -

4 x 109 -

2 x 10’ -

4 x 10’ -

I x 10’

(&z? -

4 x 10’ 2 x 10’ 7 x 109 3 x lo6 2 x 10’ 1 x 10’ 6 x lo6 2 x 10’ 2 x 109

9 x 10’

REPORT OF COMMITTBE 2

METABOLIC DATA FOR CERIUM

1. Metabolism

No data are given in Reference Man (ICRP, 1975) for cerium.


The fractional absorption of cerium from the gastrointestinal tract of rats has been reported as less than lo- 3 (Durbin et al., 1956). Low values of fractional absorption have also been reported for chickens (Mraz et al., 1964), pigs (McClellan et al., 1965), goats (Ekman and Aberg, 1962) and cattle (Miller et al., 1967). In man, data from a case of accidental inhalation (Sill et al., 1969) also indicate that the fractional absorption of cerium from the. gastrointestinal tract is very small. In this reportf, has been taken to be 3 x 10m4 for all compounds of cerium.


The ICRP Task Group on Lung Dynamics (1966) assigned oxides, hydroxides and fluorides of cerium to inhalation class Y and all other compounds of the element to inhalation class W. This classification is supported by experiments on mice (Gensicke, 1964), rats (Abrams et al., 1946; Hennacy, 1961) and dogs (Tombropoulos et al., 1969) and is adopted here.

(c) Distribution and rerention

Inhalation class f,

E -

3 x 10-d Y 3 x lo-’

Intravenously injected cerium is primarily deposited in the liver, spleen, skeleton, kidney cortex and adrenal glands (Hennacy, 1961; Tombropoulos et al., 1969; McKenney et al., 1961; Pldkova et al., 1971; Moskalev, 1959). In this report it is assumed that, of cerium eoter- ing the transfer compartment, 0.6 is translocated to liver, 0.05 to spleen, 0.2 to bone and 0.15 to all other tissues. Cerium entering these tissues is assumed to be retained in them with a half-life of 3 500 days. This corresponds with the half-life for the whole-body retention of cerium in beagles (Richmond and London, 1966).

3. Claasi5cation of Isotopes for Bone Dosimetry

None of the isotopes of cerium considered in this report has a radioactive half-life of greater than 285 days. Thus, since the chemistry of cerium more closely resembles that of the actinides than it does that of the alkaline earths, all isotopes of cerium considered in this report are assumed to be uniformly distributed over bone surfaces at all times following their deposition in the skeleton.

References Abrams, R.. Seibert, H. C.. Potts, A. M.. Lohr, W. and Pastel, S. Metabolism of Inhaled Fission Proabct

Aerosols. Argonne National Laboratory, Chicago (MDDG248). 1946. Durbin, P. W., Williams, M. H., Gee, M., Newman, R. H. and Hamilton, J. G. (1956). Metabolism of the

lanthanons in the rat. Proc. Sot. Exp. Bioi. Med. 91,784.


Ekman, L. and Atwg, B. (1962). Excretion of niobium-95, yttrium-91, cerium-144 and promethium-147 in goats. Res. Ver. Sci. 2, lt&lOS.

Gensicke, F. (1964). Die wirkung eingeatmeter hexametaphosphataerosole auf den stoffwechsel von inhalier- tern radiocer (*44Ce-‘44Pr) bei ma&n. .S~rcrhlextherapie 124, 242-253.

Hennacy, R. A. Translocation and inhalation of Ce 14*01. In : Hanford Biology Research Annual Report for 1960, pp. 77-80. Washington (HW-69500), 1961.

ICRP Publication 23, Report of the Task Group on Reference Man. Pergamon Press, Oxford, 1975. ICRP Task Group on Lung Dynamics (1966). Deposition and retention models for internal dosimetry of the

human respiratory tract. Healrh Phys. 12, 173-207. McClellan, R. 0.. Bustad, L. IC. and Keough, R. F. (1965). Metabolism of some SNAP radionuclides in minia-

ture swine. Aerosp. Med. 36, 16-u). McKenney, J. R., McClellan, R. 0. and Bustad, L. K. Preliminary observations on Ce144-Pr144 in sheep.

In: Ha&d Biology Research Annual Report for 1960, p. 60. Washington (HW-69500). 1961. Milkr, J. K., Perry, S. C., Chandler, P. T. and Cragle, R. G. (1967). Evaluation of radiocerium as a non-

absorbed reference material for determining gastrointestinal sites of nutrient absorption and excretion in cattle. J. Dairy Sci. SO, 355.

Moskalev, Yu. I. (1959). Experiments on the distribution of Ce 144. Med. Radiol. 4. 52. English translation: JPRS: 2860 (1960).

Mrax, F. R., Wright, P. L., Ferguson, T. M. and Anderson, D. L. (1964). Fission product metabolism in hens and transference to eggs. Health Phys. 10.777-782.

Pl&ovB, A., Tmovec, T., Chorvat, D. and VladBr, M. (1971). The effect of colloidal iron hydroxide on radiocerium metabolism in rats and mice. Strahlen?herapie 142,480+85.

Richmond, C. R. and London, J. E. (1966). Long-term in uiuo retention of cerium-144 by beagles. Nature 211. 1179.

Sill, C. W., Voelz, G. L.. Olson, D. G. and Anderson, J. I. (1969). Two studies of acute internal exposure to man involving cerium and tantalum radioisotopes. Health Phys. 16, 325-332.

Tombropoulos, E. G., Rair, W. J. and Park, J. F. (1969). Removal of inhaled *44Ce-1*4Pr oxide by diethylenetriaminepentaacetic acid (DTPA) treatment. I. 144Ce-‘44Pr oxide prepared by peroxide oxidation. Health Phys. 16, 333-338.

Annual limits on intake, ALI and derived air concentrations, DAC(Bq/ml) (40 h/wk) for isotopes of cerium

Oral

Inhalation

Class W Class Y

Radionuclide f, = 3 x 1o-4 fi = 3 x 10-e f, = 3 x 10-b

‘34& AL1 2 x IO’ 3 x 10’ 2 x 10’

DAC “Q AL1

DAC *rQ AL1

DAC 137P& AL1

DAC 139& AL1

DAC 1.1~ AL1

DAC 141(3 AL1

DAC 144a

ALI

DAC


- 6 x 10’

- 2 x 109

- 9 x 10’


- 2 x 10’

- 6 x 10’

(7 x IO’) LLI Wall

- 4 x 10’


- 8 x IO’

(9 x 106) LLI Wall

-

1 x IO4 1 x 10’ 6 x 10’ 5 x lo9 2 x 106 2 x 10’

7 x 104 3 x 10’ 1 x l(r 3 x 10’

1 x l(r 7 x 10’

3 x lo* 9 x 10’

4 x 10’

1 x 10’ 1 x 100 5 x 104 5 x lo9 2 x 106 1 x IO’

6 x 10. 2 x 10’ 1 x lo* 2 x 10’

9 x 10’ 6 x 10’

2 x 104 5 x 10’

2 x 10’

96 REPORT OF COMIMTTEE 2

METABOLIC DATA FOR POLONIUM

1. Metabolism


Daily intake of polonium 0.1 Bq No data on the polonium content of the body are given in Reference Man (ICRP, 1975).

However, Blanchard (1967) and Holtzman (1963) have reported measurements of the ‘i”Po content of various tissues. Their results indicate that the total 2 “PO content of the body is typically 40 Bq with about 60% of this activity in the mineral bone. This distribution arises because of the production of ‘r ‘PO from “‘Pb in situ and has little bearing on the distribution of inhaled or ingested polonium.

2. Metabolic Model

(a) Uptake to blood

Fink (1950) has reported an absorption of about 20% following oral administration of polonium chloride to a male patient suffering from chronic myeloid leukaemia; 36% absorption by rats has also been reported (Anthony et al., 1956), but no details of the experiments were given. A value forf, of 0.1 has been assumed in the absence of more experimental data.


The ICRP Task Group on Lung Dynamics (1966) assigned oxides, hydroxides and nitrates of polonium to inhalation class W and all other compounds of the element to inhalation class D. Animal experiments with polonium chloride (Fink, 1950; Smith ef al., 1961) are in agreement with this classification and it has been adopted here.

Inhalation class f1

ti 0.1 0.1

Y -


Polonium is deposited in the liver, kidneys and spleen and is associated with the blood, probably on the red cells (Smith et al., 1961). Excretion data on humans have been summarized by Jackson and Dolphin (1966). These data indicate that the whole-body biological retention of polonium is well described by a function of the form

R(t) = e -0.493r130

This retention function has been assumed to apply to individual tissues. Thus, of polonium entering the transfer compartment, fractions 0.1, 0.1, 0.1 and 0.7 are assumed to go to liver, kidney, spleen and all other tissues respectively. Having entered these tissues the polonium is assumed to be retained there with a biological half-life of 50 days.



Polonium is assumed to be uniformly distributed throughout all organs and tissues of the body other than the liver, kidney and spleen. Therefore a classification of isotopes of the element for the purposes of bone dosimetry is not required.

References Anthony, D. S., Davis, R. K., Cowden, R. N. and Jolley, W. P. Experimental data useful in establishing maxi-

mum permissible single and multiple exposures to polonium. In: Proceedings of the International Confirence on the Peaceful Uses of Atomic Energy. Vol. 13, pp. 215-218. United Nations, New York, 1956.

Blanchard, R. L. (1967). Concentrations of 2*oPb and z‘“Po in human soft tissues. He&h Phys. 13, 625-632. Fink, R. M. Biological Studies with Polonium, Radium and Plutonium. McGraw-Hill, London, 1950. Holtzman, R. B. (1963). Measurement of the natural contents of RaD (Pb”O) and RaF (Po’*O) in human

bone-estimates of whole-body burdens. Health Phys. 9, 385-400. ZCRP Publication 23. Report of the Task Group on Reference Man. Pergamon Press, Oxford, 1975. JCRP Task Group on Lung Dynamics (1966). Deposition and retention models for internal dosimetry of the

human respiratory tract. Health Phys. 12, 173-207. Jackson, S. and Dolphin, G. W. (1966). The estimation of internal radiation dose from metabolic and urinary

excretion data in a number of important radionuclides. Health Phys. 12,481-500. Smith, F. A., Morrow. P. E., Gibb, F. R., Della Rosa, R. J., Casarett, L. J., Scott, J. K., Morken, D. A. and

Stannard, J. N. (1961). Distribution and excretion studies in dogs exposed to an aerosol containing polonium- 210. Am. Znd. Hyg. Ass. J.22,201-208.

Annual limits on intake, ALI and derived air concentrations, DAC(Bq/m3) (40 h/wk) for isotopes of polonium

Inhalation

Radionuclide

20’Po

2osPo

2O’Po

2’OPo

Oral Class D Class W

f, = 1 x 10-l fi = 1 x 10-l j-1 = 1 x 10-l

ALI 9 x 10’ 2 x IO9 3 x lo9 DAC - 1 x 106 1 x 106 ALI 8 x 10“ 1 x 109 3 x 109 DAC - 6 x IO’ 1 x lo6 AL1 3 x 10’ 9 x IO’ 1 x lo9 DAC - 4 x 105 4 x 10% AL1 1 x 105 2 x IO4 2 x lo* DAC - 1 x 10’ 1 x 10’


METABOLIC DATA FOR RADIUM

1. Metabolism

Data from Reference Man (ICRP, 1975) Radium content of the body

of the skeleton Daily intake in food and fluids

31 PB 27 pg 2.3 pg


The Task Group on Alkaline Earth Metabolism in Adult Man (ICRP, 1973) cite values for fractional absorption from the gastrointestinal tract in the range 0.15421. These values are derived from studies in which the radium was present in drinking water or incorporated into food. Experiments on the uptake of RaCI, from the gastrointestinal tract have been performed on rats of different ages (Taylor et al., 1962). Uptake was high (79 %) in 14-18 day- old rats but decreased to low values (3.2%) in 60-70 week-old animals.

In this report, fi has been taken to be 0.2 for all compounds of radium.


The ICRP Task Group on Lung Dynamics (1966) interpreted the data of Marinelli et al. (1953) as indicating a 180-day biological half-life for RaSO* in the lung. However, this analysis took no account of extrapulmonary deposits and re-examination of the data suggests a rather more rapid clearance of this compound from the lungs. Since radium sulphate is one of the less soluble compounds of radium, all commonly occurring compounds of radium are assigned to inhalation class W.

Inhalation class fl

D W 0.2 Y -

(c) Distribution and Retention

A very comprehensive model for the retention of radium in adults has been developed by an ICRP task group (ICRP, 1973). The total number of spontaneous nuclear transformations in soft tissue, cortical bone and trabecular bone during the 50 years following the introduction of 1 Bq of a radioactive isotope of radium into the transfer compartment can be derived from the retention functions given in the task group report (ICRP, 1973).


’ 2 6 Ra decays to 222Rn which has a radioactive half-life of 3.8 days. Since radon is a noble gas any 222Rn produced in soft tissue may be assumed to escape from the body without decaying. The fraction of 222Rn retained in mineral bone is a function of the time for which its parent ‘lsRa has been resident in that tissue, varying from 6 y0 at 1 day to 40 y0 in 27 years (Mays et al., 1958). By use of the 226Ra and 222Rn retention functions given by Norris et al. (1955), it can be estimated that the average fraction of 222Rn retained in mineral bone,


evaluated over the 50 years following intake, is about 0.3 and this is the value which has been used in this report. Other, shorter lived, isotopes of radon are assumed to remain totally with their parents.

3. Classi6cation of Isotopes for Bone Dosimetry

As discussed in Chapter 7, isotopes of the alkaline earths with radioactive half-lives greater than 15 days can be assumed to be distributed throughout the volume of mineral bone. Thus, 22sRa and 228Ra are assumed to be distributed throughout the volume of mineral bone, while 223Ra, 224Ra, 22sRa and 2 27 Ra are assumed to be distributed over bone surfaces at all times following their deposition in the skeleton.

References ICRP Publication 20, ICRP Task Group on Alkaline Eorrh Metabolism in Adult Man. Pergamon Press, Oxford,

1973. ICRP Publication 23. Report of the Task Group on Reference Man. Pergamon Press. Oxford 1975. ICRP Task Group on Lung Dynamics (1966). Deposition and retention models for internal dosimetry of the

human respiratory tract. Health Phys. 12 173-207. Marinelli, L. D., Norris. W. P., Gustafson, P. F. and Speckman. T. W. (1953). Transport of radium sulphate

from the lung and its elimination from the human body following single accidental exposures. Radiology 61, 903-91s.

Mays, C. W.. Van Dilla, M. A., Floyd, R. L. and Arnold, J. S. (1958). Radon retention in radium-injected beagles. Rodiat. Res. 8.480-489.

Norris, W. P., Speckman, T. W. and Gustafson. P. F. (1955). Studies on the metabolism of radium in man. Am. J. Roentgen. 73. 785-802.

Taylor, D. M.. Bligh, P. H. and Duggan. M. H. (1962). The absorption of calcium, strontium, barium and radium from the gastrointestinal tract of the rat. Biochem. J. 83, 25-29.

Annual limits on intake, ALI and derived air concentrations. DAC(Bq/ml) (40 h/wk) for isotopes of radium

Radionuclide

Oral

fi = 2 x 10-l

Inhalation

Class w

fi = 2 x 10-i

“‘Ra ALI

DAC nz.Ra AL1

DAC AL1

DAC ALI

‘*‘Ra DAC AL1

‘*ORa DAC ALI

DAC

2 x 10’ (3 x 10’) Bone surf.

- 3 x 10”


- 3 x 10’


- 7 x 10’

(2 x 10”) Bone surf.

- 6 x lOa

(9 x 100) Bone surf.

9 x10’ (1 x 10’) Bone surf.

-

3 x lo*

1 x 10’ 6 x IO4

3 x 10’ 2 x 10’

1 x 10’ 2 x 10’

1 x 10’ 5 x 10’


2 x 10’ 4 x lo4

2 x 10’

100 REPORT OF COMMI’ITEE 2

METABOLIC DATA FOR THORIUM

1. Met&o&an


Daily intake in food and fluids 3 clg No data on the thorium content of organs and tissues are given in Reference Man. The

studies of Lucas et of. (1970) indicate that the total thorium content of mineral bone is 30 pg.

2. Metabolic Model (a) Uptake to blood

Experiments on the absorption of 23zTh(N03)4 from the gastrointestinal tract of the rat (Kendysh, 1966; Traikovich, 1970) indicate a value forfi in the range 5 x lo-’ to 10e2. However, the fractional absorption of 234Th(S04)2 in the form of mock radium dial paint, from the gastrointestinal tract of man, has been found to be in the range low4 to 6 x low4 with an estimated average value of approximately 2 x 10m4 (Maletskos et ol., 1969). In this report, f, has been taken to be 2 x low4 for all compounds of thorium.


The ICRP Task Group on Lung Dynamics (1966) assigned oxides and hydroxides of thorium to inhalation class Y. Experiments on beagles (Ballou and Hursh, 1972) are in agreement with this classification. However, experiments on the inhalation ofThC1, by rats (Boecker ef al., 1963) have indicated that this compound should be assigned to inhalation class W. In this report oxides and hydroxides of thorium have been assigned to inhalation class Y and all other compounds of the element have been assigned to inhalation class W.

Inhalation class f,

D - W 2 x lo-• Y 2 x 10-d


From the results of Stover et al. (1960, 1965), on the retention of thorium in the beagle, the following metabolic model has been derived:

Of thorium entering the transfer compartment, 0.7 is assumed to be translocated to bone, where it is retained with a biological half-life of 8 000 days, 0.04 is assumed to be translocated to the liver, where it is retained with a biological half-life of 700 days, and 0.16 is assumed to be uniformly distributed among all other organs and tissues of the body, being cleared from them with a biological half-life of 700 days. The remaining 0.1 of thorium entering the transfer compartment is assumed to go directly to excretion. For thorium, the transfer compartment is assumed to be cleared with a biological half-life of 0.5 days (Stover et al., 1960, 1965).


Thorium is predominantly deposited upon the endosteal surfaces of mineral bone and only slowly distributed throughout its volume by processes such as resorption and bone remodelling


(Jee et al., 1962). In this report thorium is assumed to be uniformly distributed over bone surfaces at all times following its deposition in the skeleton.

References Ballou, J. E. and Hursh, J. B. (1972). The measurement of thoron in the breath of dogs administered inhaled

or injected ThOl. Her&r Phys. 22. 155-l 59. Boecker, B. B., Thomas, R. G. and Scott, J. K. (1963). Thorium distribution and excretion studies II. General

patterns following inhalation and the effect of the size of the inhaled dose. Health Phys. 9, 165-176. ICRP Publication 23, Report of the Task Group on Reference Man. Pergamon Press, Oxford, 1975. ICRP Task Group on Lung Dynamics (1966). Deposition and retention models for internal dosimetry of the

human respiratory tract. Health Phys. 12, 173-207. Jee, W. S. S., Arnold, J. S., Cochran, T. H., Twente, J. A. and Mica], R. S. Relationship of micro-

distribution of alpha particles to damage In: Some Aspects of Internal Irradiation. Eds. Dougherty, T. F., Jee, W. S. S., Mays, C. W., and Stover, B. J. pp. 27-45. Pergamon Press, Oxford, 1962.

Kendysh, I. N. Distribution and elimination of 2’2Th. In: Distribution and Biological Effects of Radioactive Isotopes, Ed. Moskalev, Yu. I., pp. 22-29. Atomizdat, Moscow. (AEC-tr-6944 (Rev.), pp. 25-34), 1966.

Lucas. H. F., Edgington, D. N. and Markun, F. (I 970). Natural thorium in human bone. Health Phys. 19.739-742. Maletskos. C. J., Keane, A. T., Telles, N. C. and Evans, R. D. Retention and absorption of 224Ra and

23*Th and some dosimetric consequences of 224Ra in human beings. In: Delayed Eflects of Bone Seekiw Radionuclides. Ed. Mays, C. W., pp. 29-49. University of Utah Press. Salt Lake City, 1969.

Stover, B. J., Atherton, D. R., Keller, N. and Buster. D. S. (1960). Metabolism of the Th2’s decay series in adult beagle dogs. 1. Th’” (RdTh). Radiat. Res. 12, 657-671.

Stover, B. J., Atherton, D. R.. Buster. D. S. and Bruenger, F. W. (1965). The Th22* decay series in adult beagles: Ra224, Pb2i2 and BizL2 in selected bones and soft tissues. Radiat. Res. 26, 132-145.

Traikovich, M. Absorption, distribution and excretion of certain soluble compounds of natural thorium. In The Toxicity of Radioactive Substances, Vol. 4: Thorium-232 and Uranium-238, Eds. Letavet, A. A. and Kurlyandskaya, E. B., pp. 20-29. Pergamon Press, London, 1970.

Annual limits on intake, ALI and derived air concentrations, DAC(Bq/m3) (40 h/wk) for isotopes of thorium

Inhalation

Oral Class W Class Y

Radionuclide f, = 2 x 10-4 fi = 2 x 10-b f, = 2 x 10-d

2asTh AL1 2 x 10’ 6 x lo6 5 x lo6 (2 x 10’) ST Wall

DAC AL1 DAC AL1

2 x 10’ 2 x 10’ - 5 x 106

- 2 x 10’


-

1 x lo4 5 x 100

1 x lo* 5 x 100

4 x 10’ (8 x 102)

6 x lo2

Bone surf. 2 x 10-l DAC

AL1 3 x IO-’ 9 x 10’


4 x lo-’ 6 x lo*


2 x 10-l 2 x 10s 1 x 10’

229Th 2 x 10’ (5 x 109

3 x 10’ (9 x 10’) Bone surf. Bone surf.

- 1 x 10s


- 1 x 108

- 3 x 104


- 1 x 10’


-

DAC AL1

1 x 10-Z 2 x 10’


1 x 10-l 2 x IO’ 1 x 105 4 x 101


2 x 10-Z 7 x 106

“OTh

DAC AL1 23’Th

2227-h DAC AL1 1 x 10’

(2 x 102) Bone suri.

4 x 10-x 6 x lo6

DAC AL1 234Th

DAC 3 x 10’ 2 x 10’


METABOLIC DATA FOR URANIUM

1. Metabolism


Uranium content of the body 90 pg of the skeleton 59 pg of the kidneys 7 PI

Daily intake in food and fluids 1.9 pg


Hursh et al. (1969) have examined the uptake of uranyl nitrate hexahydrate from the gastrointestinal tract of man. They conclude that between 0.005 and 0.05 of the compound is likely to be absorbed. A similar result has been reported by Butterworth (1955) for a single human subject. These findings are consistent with the fractional uptake of UOzFs in mongrel dogs (Fish et al., 1960), which has been reported to be 0.0155.

Toxicity studies (Yuile, 1973) have indicated that the gastrointestinal absorption of relatively insoluble compounds such as UOt and UjOg is considerably less than 0.01.

A higher value of absorption, in the region of 0.2, is indicated by dietary data from occupationally unexposed persons (Hursh, and Spoor 1973) but this is for extremely low levels of intake (about 1 pg per day) of uranium which is probably in an organically complexed form.

In this report, fi is taken to be 0.05 for water-soluble inorganic compounds of uranium (hexavalent uranium) and 0.002 for relatively insoluble compounds such as UF4, UOz and UjOs in which the uranium is usually tetravalent.


Uranium compounds such as UF,, U02F, and U01(N0J2 are rapidly absorbed from the lung (Chalabreysse, 1970) and have been assigned to inhalation class D. Less soluble compounds such as UO1, UF, and UCl, have been assigned to inhalation class W (Hursh and Spoor, 1973; Morrow et al., 1972) while highly insoluble uranium oxides, i.e. UO1 and U,O,, are assigned to inhalation class Y (Morrow et al., 1966).

fl is taken to be 0.05 for class D and W materials entering the gastrointestinal tract following inhalation. For class Y materials entering the gastrointestinal tract following inhalation fl is taken to be 0.002.

Inhalation class

D W Y

J-1

0.05 0.05 0.002


Bone and kidney retention functions have been derived (Adams and Spoor, 1974) using post-mortem data from an occupationally exposed man (Donoghue et al., 1972) and data from adults who had ingested uranium at normal dietary levels (Hursh and Spoor, 1973).


ReoNa(r) = 0.2 e-0*683r120 + 0.023 e-“.sssr15 Ooo

103

RKIDNEv(r) = 0.12 e - 0.693r/6 + 0 c,@)52 ,-0.693#/1 500 .

On the basis of the uranium distribution observed when terminal patients were intravenously injected with UOr(NO3)s solution (Bernard and Struxness, 1957), the retention function for all soft tissues excluding the kidneys may be taken to be the same as that for kidney;

I&&f) = 0.12 e-0*693r/6 + 0.00052 e-“*@3r11 ‘O”

Thus, of uranium entering the transfer compartment, fractions 0.2 and 0.023 are assumed to go to mineral bone and be retained there with half-lives of 20 and 5000 days respectively, fractions 0.12 and 0.00052 are assumed to go to the kidneys and to be retained with half- lives of 6 and 1 500 days respectively and fractions 0.12 and 0.00052 are assumed to go to all other tissues of the body and be retained with half-lives of 6 and 1 500 days respectively. Uranium is assumed to be uniformly distributed amongst these other tissues. The remainder of uranium entering the transfer compartment is assumed to go directly to excretion.


Autoradiographic studies of Neuman and Neuman (1948) indicated that uranium was mainly deposited on bone surfaces readily available to the circulation, and that it was only slowly redistributed throughout the volume of mineral bone. However, more recent work by Rowland and Farnham (1969) has shown that redistribution throughout bone volume can take place within a few months of injection. For this reason, 229u, ZJOu, 231u, 235mu, 237u, 239uand

240U are assumed to be distributed uniformly over bone surfaces at all times following their deposition in the skeleton, whereas the long-lived uranium isotopes 232U, 233U, 234U, 235U, 236U and 23sU are assumed to be uniformly distributed throughout the volume of mineral bone at all times following their deposition in the skeleton.

References Adams, N. and Spoor, N. L. (1974). Kidney and bone retention functions in the human metabolism of uranium.

Phys. Med. Biol. 19.460-471. Bernard. S. R. and Struxness. E. Cl. (1957). A study of the distribution and excretion of uranium in man.

USAEC Report ORNL-2304. Butterworth, A. (1955). The significance and value of uranium in urine analysis. Don. Ass. Ixd. Med. Offrs. 5,36. Chalabreysse, J. (1970). Etude et resultats d’examens effectues a la suite d’une inhalation de composes dits

solubles d’uranium naturel. Radioprofecrion 5, l-17 and 305-310. Donoghue, J. K., Dyson, E. D., Hislop, J. S., Leach, A. M. and Spoor, N. L. (1972). Human exposure to

natural uranium. Br. 1. Ind. Med. 29.8149. Fish, B. R., Payne, J. A. and Thompson, J. L. Ingestion of uranium compounds. In: Health Physics Diuision

Annsal Progress Report for Period Ending 31 July 1960, pp. 269-272. Oak Ridge, Tenn., Oak Ridge National Laboratory (ORNG2994), 1960.

Hursh, J. B. and Spoor, N. L. Data on man. In: Uranium, Plutonium, Transplutonium Elements, Eds. Hodge, H. C., Stannard. J. N. and Hursh, J. B.. pp. 197-239. Springer-Verlag, Berhn, 1973.

Hursh, J. B., Neuman, W. R., Toribara, T., Wilson, H. and Waterhouse, C. (1969). Oral ingestion of uranium by man. Health Phys. 17, 619-621.

ICRP Publication 23, Report of the Task Group on Reference Man. Pergamon Press, Oxford, 1975. Morrow, P. E., Gibb, F. R. and Leach, L. J. (1966). The clearance of UO1 dust from the lungs following single

and multiple inhalation exposures. Health Phys. 12, 1217-1223. Morrow, P. E.. Gibb, F. R. and Beiter, H. D. (1972). Inhalation studies of uranium trioxide. Health Phys. 23.

273-280. Neuman, M. W. and Neuman, W. F. (1948). The deposition of uranium in bone. II. Radioautographic studies.

J. Biof. Chem. 175, 71 l-714. Rowland, R. E. and Farnham, J. E. (1969). The deposition of uranium in bone. Health Phys. 17, 139-144. Yuile, C. L. Animal experiments. In: Uranium Plutonium Transplutonium Elements, Eds. Hedge. H. C..

Stannard, J. N. and Hursh, J. B., pp. 165-196. Springer-Verlag, Berlin, 1973.


Annual limits on intake, ALI and derived air concentrations, DAC(Bq/m”) (40 h/wk) for isotopes of uranium

Inhalation

Oral Class D Class W Class Y

Radionuclide f, = 5 x 10-Z /, = 2 x 10-S h = 5 x 10-Z ft = 5 x 10-z fi = 2 x 10-3

23OU ALI

23,” DAC ALI

232,, DAC AL1

233u DAC AL1

234,, DAC ALI

23SU

DAC ALI

236u

DAC AL1

237,, DAC AL1

236u DAC AL1

239U

oco(J

DAC AL1 DAC AL1 DAC

1 x 10’ (2 x 105) Bone surf.

2 x10’

!!LF FE?

8 x10* (1 x 10’) Bone surf.

- 4 x 10s


- 4 x 10’


- 5 x 10’


5 x105 (7 x 10’) Bone surf.

- 6 x 10’

!TLh%

5 XlO” (8 x 10’) Bone surf.

2 x109 -

5 x 10’ -

2 x 106

- 2 x IO’

i2L%? -

2 x 106 (3 x 106) Bone surf.

- 7 x 106

- 7 x 106

-

- 7 x 106

- 8 x 10”

- 6 x 10’


8 x<06

- 2 x lo9

5 x-lo7 -

2 x 10’ (2 x 10’) Bone surf.

6 x loo 3 x IO”

1 x 105 8 x lo3


3 x loo 4 x 10’

(7 x lo*) Bone surf. 2 x 10’ 5 x 10’

(7 x 10’) Bone surf. 2 x 10’ 5 x lo*


2 x 10‘ 5 x 10’

(7 x 10’) Bone surf. 2 x 10’ 1 x 10’

4 x 10L 5 x 10’

(8 x 10’) Bone surf. 2 x 10’ 7 x 109 3 x 106 1 x 10’ 6 x 10’

1 x 10’

5 x 100 2 x 10’

9 x l(r 1 x lo4

6 x loo 3 x 104

1 x 10’ 3 x lo*

1 x 10‘ 3 x lo*

1 x 10’ 3 x l(r

1 x 10’ 6 x 10’

3 x 10’ 3 x lo*

1 x 10’ 6 x lo9 3 x 106 1 x 10’ 4 x 104

1 x 104

4 x loo 2 x IO’

7 x lo* 3 x 10’

1 x 10-l 1 x 10’

6 x lo-’ 1 x 10’

6 x 10-l 2 x IO’

6 x lo-’ 1 x lo3

6 x 10-l 6 x 10’

2 x 104 2 x lo3

7 x 10-l 6 x lo9 2 x 10’ 9 x 10’ 4 x 104


METABOLIC DATA FOR PLUTONIUM

1. Metabolism

No data are given in Reference Man (ICRP, 1975) for plutonium. However, there are measurable quantities of the element in foods and in human tissues as a result of world-wide fallout from nuclear weapons tests. There are also data on the distribution of plutonium in people occupationally exposed to the element. These limited human data, in general, support and confirm the much more extensive metabolic data from studies in experimental animals. The metabolism of plutonium, as understood in 1970, was reviewed by an ICRP task group (ICRP, 1972). More recent data have been reviewed in other publications (Dolphin, 1971; Hodge et al., 1973; Bair and Thompson, 1974), notably in the extensive collection of papers in Volume 36 of the Handbook of Experimental Pharmacology (Hodge et al., 1973).

2. Metabolic Model

(a) Uptake to blood

Absorption of plutonium from the gastrointestinal tract is low. Forf, the task group (ICRP, 1972) recommended a value of 3 x 10 - 5 “for more soluble compounds (e.g. plutonium nitrate)” and a value of lo- 6 for “relatively insoluble material such as PuO, . . .‘*. However, more recent data reviewed by Stather and his co-workers (1979) suggest that higher values of fi may be appropriate. In this report fi is taken as 10m5 for oxides and hydroxides of plutonium and lo- 4 for all other commonly occurring compounds of the element.

It should be noted that data have been reported which indicate a much higher gastrointestinal absorption for certain compounds of plutonium that are unlikely to be encountered in occupational exposures, e.g. hexavalent plutonium compounds, citrates and other organic complexes; absorption is also increased in the very young (ICRP, 1972).


The task group (ICRP, 1972) concluded that no plutonium compounds should be assigned to inhalation class D (except chelates, discussed below), that PuO, should be assigned to inhalation class Y and that all other commonly occurring compounds of the element should be assigned to inhalation class W. Recent data (Bair et al., 1973; Park et al., 1974; Watts, 1975) have emphasized the complexity of the retention patterns of plutonium compounds in the lung, but do not change the inhalation class assignments. Thus, although 23sPu02 is more rapidly translocated from the lung than is 23gPu02 and much larger concentrations of 238Pu accumulate in bone and liver (Park et al., 1974), 238Pu02 is still best described as a class Y compound.

Inhalation class fi

:: -

10-b Y lo-’

(c) Distribution and retention Plutonium absorbed into the blood stream is deposited principally in liver and skeleton.

The relative deposition in these two organs has shown a considerable variability in human


autopsy, and experimental animal data. The task group (ICRP, 1972) has recommended values of 0.45 for the fractional deposition from blood in liver, 0.45 in bone, and the remaining fraction of 0.10 in all other tissues and early excreta. Based on data from human subjects and from several mammalian species, the task group suggested human biological half-lives of 100 years for retention in the skeleton and 40 years for retention in the liver (ICRP, 1972).

In this report, of plutonium entering the transfer compartment, 0.45 is assumed to be translocated to bone and 0.45 to iiver. The data reviewed by Richmond and Thomas (1975) indicate that the fraction of plutonium deposited in the gonads is likely to be about 3 x 10m4 for human males and lo- 4 for human females. In this report it has been assumed that, of plutonium entering the transfer compartment, the fraction translocated to the gonads is 3.5 x lo- 4 for males and 1.1 x lo- 4 for females, these values corresponding to a fractional translocation to the gonads of lo- ’ per g of gonadal tissue in each case. The remainder of plutonium entering the transfer compartment is assumed to go directly to excretion.

Plutonium deposited in gonadal tissue is assumed to be permanently retained there, whereas plutonium deposited in the liver is assumed to be retained with a biological half-life of 40 years and plutonium deposited in the skeleton is assumed to be retained with a biological half-life of 100 years.

(d) Chelated compounds

Chelated forms of plutonium are not considered in this report. It is known that they move to some extent across the gut wall but also that they are more readily excreted than are other compounds of plutonium (Baxter and Sullivan, 1972).

3. Clasdcrtion of Isotopes for Bone Dosimetry

Plutonium mainly deposits upon the endosteal surfaces of mineral bone and is slowly redistributed throughout its volume by processes such as resorption and burial (Vaughan, 1973; Schlenker et al., 1976). For this reason all isotopes of plutonium considered in this report are assumed, for the purposes of dosimetry, to be uniformly distributed over bone surfaces at all times following their deposition in the skeleton.

References

Bair, W. J. and Thompson, R. C. (1974). Plutonium: biomedical research. Science 183, 715-722. Bair, W. J., Ballou, J. E., Park, J. F. and Sanders, C. L. Plutonium in soft tissues with emphasis on the

respiratory tract. In: Uranium, Plutonium, Transplutonium Elements, Eds, Hodge, H. C., Stannard, J. N. and Hursh, J. B., pp. 503-568. Springer-Verlag, Berlin, 1973.

Baxter. D. W. and Sullivan, M. F. (1972). Gastrointestinal absorption and retention of plutonium chelates. Health Phys. 22, 785-786.

Dolphin, G. W. (1971). The biological problems in the radiological protection of workers exposed to s3*Pu. Health Phys. 20, 549-557.

Hodge, H. C., Stannard, J. N. and Hursh, J. B., Eds. Uranium, Plutonium, Transplutonium Elements (Handbook of Experimental Pharmacology, Vol. 36). Springer-Verlag, Berlin, 1973.

ICRP Publication 19, The Metabolism of Compounds of Plutonium and Other Actinides. Pergamon Press, Oxford, 1972.

ICRP Publication 23. Report of the Task Group on Reference Man. Pergamon Press, Oxford, 1975. Park, J. F., Catt, D. L., Craig, K. D., Olson, R. J. and Smith, V. H. Solubility changes of a3sPu oxides

in water suspension and effect on biological behaviour after inhalation by beagle dogs. In: Proc. Third Int. Cong. Int. Radiat. Prot. Assn., Washington DC September 1973, Ed. Snyder, W. S., Vol. 1, pp. 719-724 (CONF-730901-PI). 1974.

Richmond, C. R. and Thomas, R. L. (1975). Plutonium and other actinide elements in gonadal tissue of man and animals. Health Phys. 29, 241-250.

LIMITS FOR INTI)KES OF RADIONUCLIDES BY WORKERS 107

.?&lenker. R. A., Oltman, B. G. and Cummins. H. T. Microscopic distribution of 239Pu deposited in bone from a human injection case. In: The Health Effects of Plutonium atd Radium, Ed. Jee, W. S. S., pp. 321- 328.1. W. Press, Salt Lake City, 1976.

Stather, J. W., Harrison, J. D.. Rodwell, P. and David, A. J. (1979). The gasttointestinal absorption of plutonium and americium in the hamster. To be published in Phys. Med. Biof.

Vaughan. J. The pattern of z’9Pu distributioti in the skeleton. In: Uranium, Plutonium, Transplutonium Elements. Eds. Hodge, H. C., Stannard, J. N. and Hursh, J. B., pp. 397421. Springer-Verlag, Berlin, 1973.

Watts, L. (1975). Clearance rates of insoluble plutonium-239 compounds from the lung. Health Phys. 29.53-59.

Annual limits on intake, ALI and derived air concentrations, DAC(Bq/m’) (40 h/wk) for isotopes of plutonium

Radionuclide

Oral

fr = 1 x 10-b fi = 1 x 10-S

Inhalation

Class w Class Y

f, = 1 x lo-’ /, = 1 x IO-’

AL1 3 x 10’ 3 x IO’ 8 x 106 DAC AL1 DAC AL1

DAC AL1 DAC ALI

DAC AL1

DAC AL1

DAC AL1

DAC AL1

DAC ALI DAC AL1

DAC AL1 DAC

3 x1O’O -

a x IO’ (1 x 106) Bone surf.

5 xros -

3 x 105 (5 x 10’) Bone surf.

- 2 x 103


- 2 x 10’


- 1 x 10’


3 x103 (5 x 105)

Bone surf.

6 x;Os -

3 x 10’ (5 x 10’) Bone surf.

- 8 x 10’

-

3 x1O’O -

6 x lo6

- 5 x 10’

- 3 x 10’


- 2 x 106


2 x106 (3 x IO’) Bone surf.

- 1 x 10’


- 3 x lo6


- 6 x 10’

- 3 x lo6


- a x 10’

-

3 x 103 1 x 10” 5 x 10’ 7 x 10’


3 x 10-l 1 x 10’ 5 x IO’ 2 x 102


9 x 10-a 2 x 10’


8 x 1O-a 2 x 1oa


8 x 1O-a 1 x 104


4 2 x 102


9 x 10-a 1 x 109 5 x 10s 2 x 102


9 x 10-a 2 x IO’ 7 x 10.

7 x IO6 3 x 10’ 9 x 10’0 4 x 10’ 1 x 10”

6 x 10-r 1 x 10’ 5 x I’)* 6 x 10’

(6 x lOa) Bone surf. 3 x IO-’ 5 x 1oa

(6 x IO’) Bone surf. 2 x IO-’ 5 x 10’ (6 x 10’) Bone surf. 2 x 10-l 2 x 10’

(3 x 104) Bone surf. 1 x 10’ 6 x 10’

(6 x lOa) Bone surf. 2 x 10-l 1 x 109 6 x IO5 6 x 10’

(6 x lOa) Bone surf. 2 x 10-I 2 x 10s 6 x 10.

108 REPORT OF ‘COMMITTEE 2

METABOLIC DATA FOR AMERICIUM

1. Metabolism

No data are given in Reference Man (ICRP, 1975) for americium.

Although distributed world-wide in measurable quantities as a result of fallout from nuclear weapons tests, extensive data have not been obtained on the distribution of americium in man. Reliance must, therefore, be placed on animal data, which have been reviewed, in 1970, by an ICRP task group (ICRP, 1972) and again more recently by Durbin (1973).

2. Metabolic Model

(a) Uptake to blood

Absorption of americium compounds from the gastrointestinal tract has been studied only in the rat (Scott et al., 1948; Zalikin et al., 1968; Buldakov and Nifatov, 1972; Moskalev et al., 1973; Sullivan and Crosby, 1975) with reported values ranging from < 10m4 (Scott et al., 1948) to 0.0014 (Buldakov and Nifatov, 1972). In this report,f, has, therefore, been taken to be 5 x 10m4 for all compounds of americium. It should be noted that greater gastrointestinal absorption might be expected for complexed forms of the element and that enhanced absorption has been reported in very young rats (Moskalev et al., 1973; Sullivan and Crosby, 1975).


Experiments on animals (Scott et al., 1948) indicate that compounds of americium are more rapidly cleared from the lung than are compounds of plutonium. This greater mobility, specifically noted for the oxides (Lafuma et al., 1974; Craig et al., 1975), suggests that all compounds of americium should be assigned to inhalation class W.

Inhalation class fl

D - W 5 x lo-. Y -


The considerable data available on the metabolic behaviour of americium in experimental animals are generally in accord with the recommendation of the Task Group on The Meta- bolism of Compounds of Plutonium and Other Actinides (ICRP, 1972) that the metabolic model used for plutonium should also be used for americium. Experimental evidence (Rich- mond and Thomas, 1975) indicates that the fractions of americium and plutonium translocated to the gonads are rather similar.

In this report it has been assumed, that of americium entering the transfer compartment, 0.45 is translocated to bone and 0.45 to the liver. The fraction of americium translocated to the gonads is assumed to be 3.5 x 10s4 for the testes and 1.1 x lo-* for the ovaries, these values corresponding to a fractional translocation to the gonads of lo-’ per g of gonadal tissue. The remainder of americium entering the transfer compartment is assumed to go directly to excretion.

LIMIT8 FOR INTAKES OF RADIONUCLIDES BY WORKERS 109

Americium deposited in gonadal tissue is assumed to be permanently retained there whereas americium deposited in the liver is assumed to be retained with a biological half-life of 40 years and americium deposited in the skeleton is assumed to be retained with a biological half-life of 100 years.


Chelated forms of americium are not considered in this report. It is known that they are more biologically mobile than are other compounds of americium (Durbin, 1973).


Americium mainly deposits upon both periosteal and endosteal bone surfaces, although endosteal surfaces are usually more heavily labelled. The element is slowly redistributed throughout the volume of mineral bone by processes such as resorption and burial (Durbin, 1973). Thus, for the purposes of dosimetry, all isotopes of americium considered in this report are assumed to be uniformly distributed over bone surfaces at all times following their deposition in the skeleton.

References

Buldakov, L. A. and Nifatov, A. P. Biological effectiveness of americium-241 following oral and parenteral administration. In: Biological Eficts of Radiation from External and Internal Sources, Eds. Moskalev. Yu. I., Kalistratova. V. S. pp. 305-331. Meditsina, Moscow, 1972.

Craig, D. K., Cannon, W. C.. Catt. D. L.. Herring, J. P.. Olson, R. J., Powers, G. J. and Watson, C. R. Distribution of srrAm and “**Cm in dogs after inhalation of the oxides. In.: Pacific Northwest Laboratory Annual Report for 1974 to the USAEC Division of Biomedical and Environmental Research. Part I Biomedical Sciences, pp. 17-18 (BNWL-19%Pt. 1). 1975.

Durbin, P. W. Metabolism and biological effects of the transplutonium elements. In: Uranium, Plutonium, Transplutonium Elements, Eds. Hodge, H. C., Stannard. J. N. and Hursh, J. B., pp. 739-896. Springer Verlag, Berlin, 1973.

ICRP Publication 19, The Metabolism of Compounds of Plutonium and Other Actinides. Pergamon Press, Oxford, 1972.

ICRP Publication 23, Report of the Task Group on Reference Man. Pergamon Press. Oxford, 1975. Lafuma, J., N&not, J. C., Morin, M., Masse, Ft., Metivier, H., Nolibe, D. and Skupinski, W. Respiratory

carcinogenesis in rats after inhalation of radioactive aerosols of actinides and lanthanides in various physico- chemical forms. In: Experimental Lung Cancer: Carcinogenesis and Bioassays, Eds. Karbe. E. and Park, J. F., pp. 443-453. Springer-Verlag. Berlin, 1974.

Moskalev, Yu. I., Zalikin. G. A. and Trofimov, V. I. (1973). Influence of age on the resorption of z4’Am from the gastrointestinal tract of rats. Radiobiologiya 13, 155.

Richmond, C. R. and Thomas, R. L. (1975). Plutonium and other actinide elements in gonadal tissue of man and animals. Health Phys. 29,241-250.

Scott, K. G., Copp. D. H., Axelrod, D. J. and Hamilton, J. G. (1948). The metabolism of americium in the rat. J. Biol. Chem. 175,691-703.

Sullivan, M. F. and Crosby, A. L. Absorption of uranium-233, neptunium-237, plutonium-238, americium-241, curium-244 and einsteinium-253 from the gastrointestinal tract of newborn and adult rats. In: Pacific Northwext Laboratory Annual Report for 1974 to the USAEC Division of Biomedical and Environmental Research. Parr 1 Biomedical Sciences, pp. 105-108 (BNWL-1950 Pt. I), 1975.

Zalikin, G. A., Moskalev, Yu. I. and Petrovich, I. K. (1968). Distribution and biological effects of ‘**Am. Radiobiolwiya 8, 65-71. (AEC-tr-6950, pp. 107-l 18.)

110 RRPORT OF COMMITTEE 2

Annual limits on intake, ALI and derived air concentrations, DAC(Bq/ml) (40 h/wk) for isotopes of americium

Radionuclide

Oral

f, = 5 x 10-b

Inhalation

Class W

f, = 5 x 10-d

r37Am ALI

s30Arn

‘42’Am

‘*‘Am

244Am

r4sAm

DAC ALI

DAC ALI DAC ALI DAC AL1

DAC AL1

DAC AL1

DAC AL1

DAC AL1

DAC AL1

DAC AL1 DAC AL1

DAC AL1 DAC

3 x 109 -

1 x 109

- 2 x 10’

- a x 10’

- 5 x 10’


5 x10’ (9 x 104) Bone surf.

- 2 x 10’

5 x IO4 (9 x lo*) Bone surf.

- 2 x lo9

(&%? -

1 x IO’

- 1 x lo9

- 2 x 109

(%El) -

1 x 109

1 x 10‘0 4 x 106 1 x IO’


4 x lo* 5 x 10’ 2 x 10s 1 x 10’ 4 x 10’ 2 x 10s


8 x lo-* 2 x 10’


8 x 1O-z 3 x 10”


1 x 10’ 2 x lo2


8 x lo-* 1 x 10’


6 x 104 6 x lo6


3 x 10’ 3 x lo9 1 x lo6 6 x lo9

3 x IO6 4 x 109 2 x lo6


METABOLIC DATA FOR CURIUM

1. Metabolism

No data are given in Reference Man (ICRP, 1975) for curium.

Reliance must, therefore, be placed on animal data, which have been reviewed by an ICRP task group (ICRP, 1972) and, more recently, by Durbin (1973).

2. Metabolic Model

(a) Uptake to blood

The fractional absorption of curium compounds from the gastrointestinal tract has been studied only in the rat (Semenov, 1971, 1972; Semenov et al., 1973; Sullivan and Crosby, 1975), with reported values ranging from 3 x lo- 5 to 7 x 10m4. By analogy with the chemically similar and more extensively studied americium, fr is taken to be 5 x 10m4. It should be noted that greater gastrointestinal absorption might be expected for complexed forms of curium and that enhanced absorption has been reported in very young rats (Semenov et al., 1973; Sullivan and Crosby, 1975).


As with americium, curium compounds, including the oxide (McClellan et al., 1972; Craig et al., 1975; Sanders, 1975), are more quickly lost from the lung than are the corresponding compounds of plutonium (Durbin, 1973). The limited data suggest that all compounds of curium should be assigned to inhalation class W.

Inhalation class fi

D - W 5 x 1o-4 Y -


The data available on the metabolic behaviour of curium in animals are generally in accord with the task group (ICRP, 1972) recommendation that the metabolic model used for plutonium shotild also be used for the other actinides. Experimental evidence (Richmond and Thomas, 1975) indicates that the fractions of curium and plutonium translocated to the gonads are rather similar.

In this report it has been assumed that, of curium entering the transfer compartment, 0.45 is translocated to bone and 0.45 to the liver. The fraction of curium translocated to the gonads is assumed to be 3.5 x 10s4 for the testes and 1.1 x 10m4 for the ovaries, these values corresponding to a fractional translocation to the gonads of lo- 5 per g of gonadal tissue. The remainder of curium entering the transfer compartment is assumed to go directly to excretion.

Curium translocated to mineral bone is assumed to be retained in that tissue with a biological half-life of 100 years, while curium translocated to the liver is assumed to be retained in that tissue with a biological half-life of 40 years. Curium translocated to the gonads is assumed to be retained indefinitely in that tissue (Richmond and Thomas, 1975).

112


REPORT OF COMMlTTJiE 2

Chelated forms of curium are not considered in this report. It is known that they have greater biological mobility than do other compounds of curium (Durbin, 1973).


The distribution of curium in the skeleton of the rat is indistinguishable from that of americium (Durbin, 1973). The element is deposited mainly on the endosteal surfaces of mineral bone and only slowly redistributed throughout its volume by processes such as resorption and burial. For this reason all isotopes of curium considered in this report are, for the purposes of dosimetry, assumed to be uniformly distributed over bone surfaces at all times following their deposition in the skeleton.

References Craig, D. K., Cannon, W. C.. Catt. D. L., Herring, J. P., Olson, R. J., Powers, G. J. and Watson, C. R.

Distribution of “‘Am and “*Cm in dogs after inhalation of the oxides. In: Pacific Northwest Laboratory Annual Report for 1974 to the USAEC Division of Biomedical and Environmental Research. Part I Biomedical Sciences, pp. 17-18 (BNWL-1950 Pt. 1), 1975.

Durbin, P. W. Metabolism and biological effects of the transplutonium elements. In: Uranium, Plutonium, Transplufonium Elements, Eds. Hodge, H. C., Stannard, J. N. and Hursh, J. B., pp. 739-896. Springer-Verlag, Berlin, 1973.

ICRP Publication 19, The Metabolism of Compounds of Pluronium and Other Actinides, Pergamon Press, Oxford, 1972.

ICRP Publication 23, Report of the Task Group on Reference Man. Pergamon Press, Oxford, 1975. McClellan, R. O., Boyd, H. A., Gallegos, A. F. and Thomas, R. G. (1972). Retention and distribution of ‘“Cm

following inhalation of ‘**CmCls and s44Cm01.rs by beagle dogs. Health Phys. 22, 877-885. Richmond, C. R. and Thomas, R. L. (1975). Plutonium and other actinide elements in gonadal tissue of man

and animals. Health Phys. 29, 241-250. Sanders, C. L. Toxicology of inhaled 24*CmOa in rats. In: Pacific Northwest Laboratory Annual Report for 1974

to the USAEC Division of Biomedical and Environmental Research. Part I Biomedical Sciences, pp..35-37 (BNWL-1950 Pt. 1), 1975.

Semenov, A. I. (1971). Kinetics of the exchange of s4*Cm. Radiobiologiya 11, 155. Semenov, A. I. Distribution of curium 244 in the rat organism. In: Biological Effects of Radiafion from External

and Internal Sources, Eds. Moskalev, Yu. I. and Kalistratova, V. S., pp. 302-305. Meditsina, Moscow, 1972.

fjemenov. A. I., Moskalev, Yu. I. and Zalikin, G. A. (1973). Influence of age on the absorption of ‘**Cm from the rat gastrointestinal tract. Radiobiologiya 13, 155.

Sullivan, M. F. and Crosby, A. L. Absorption of uranium-233, neptunium-237, plutonium-238, americium-241, curium-244 and einsteinium-253 from the gastrointestinal tract of newborn and adult rats. In: Pacific Northwest Laboratory Annual Report for 1974 to the USAEC Division of Biomedical and EitvironmenIal Research. Part I Biomedical Sciences, pp. 105-108 (BNWL-1950 Pt. I), 1975.

LIMITS FOR INTAKES OF RADlONUCLIDES BY WORKERS 113

Annual limits on intake, ALI and derived air concentrations, DAC(Bq/m’) (40 h/wk) for isotopes of curium

Inhalation

Radionuclide

Oral Class W

f, = .5 x 1o-4 f, = 5 x 10-4

ALI DAC ALI

DAC ALI

6 x lo* -

4 x lo6 (5 x 106) Bone surf.

- 5 x 10’

DAC ALI

DAC AL1

DAC AL1

DAC AL1

DAC AL1

DAC AL1

DAC AL1

DAC AL1

- 2 x 10”


- I x 10’


- 9 x 104


- 5 x IO’

(8 x 10.) Bone surf.

- 5 x lo4


- 5 x 104


- 1 x IO4


- 2 x 109

DAC -

4 x 10’ 2 x 10. 2 x lo4


8 9 x 10s


4 x 102 1 x lo4


4 3 x 102


1 x 10-l 4 x 102


2 x 10-l 2 x 10’

(3 s 102) Bone surf.

8 x 1O-3 2 x 102


8 x lo-’ 2 x 102


9 x 10-1 5 x 10’

(9 x 10’) Bone surf. 2 x 10-a 5 x 108


2 x 10s


METABOLIC DATA FOR CALIFORNIUM

1. Metabolism

No data are given in Reference Man (ICRP, 1975) for californium.

Reliance must, therefore, be placed on animal data. These data have been reviewed by an ICRP task group (ICRP, 1972) and, more recently, by Durbin (1973).

(a) Uptake to blood

2. Metabolic Model

Gastrointestinal absorption of 0.001 of intragastrically administered ” *Cf (NO,), has been reported for the rat, although absorption of 0.0001 would be estimated from the reported deposition in liver and bone (Sullivan, 1974). By analogy with the chemically similar and more extensively studied americium, f, is taken to be 5 x 10s4 for all compounds of californium..


Observations on a single case of human exposure (Rundo and Sedlet, 1974) indicate that insoluble compounds of californium may behave as inhalation class Y materials. Therefore, by analogy with plutonium, oxides and hydroxides of californium are assigned to inhalation class Y and all other compounds of the element are assigned to inhalation class W.

Inhalation class I-1

D - W 5 x 10-4 Y 5 x 10-b


Limited data on the metabolic behaviour of californium in mice, rats, hamsters and dogs (Durbin, 1973) are generally in accord with the recommendations of the Task Group on the Metabolism of Plutonium and other Actinides (ICRP, 1972) that the metabolic model used for plutonium be also used for the other actinides. Therefore, in this report, of californium entering the transfer compartment, 0.45 is assumed to be translocated to bone and 0.45 to the liver. The fraction of californium translocated to the gonads is assumed to be 3.5 x low4 for the testes and 1.1 x 10s4 for the ovaries, these values corresponding to a fractional translocation to the gonads of 10” per g of gonadal tissue (Richmond and Thomas, 1975). The remainder of californium entering the transfer compartment is assumed to go directly to excretion. Californium deposited in gonadal tissues is assumed to be permanently retained there, whereas californium deposited in the liver is assumed to be retained with a biological half-life of 40 years and californium deposited in the skeleton is assumed to be retained with a biological half-life of 100 years.


Chelated forms of californium are not considered in this report. They may be assumed to be more mobile than are other compounds of californium (Parker et al., 1962).

LIMITS FOR INTAKES OF RADIONUCLIDJZS BY WORKERS 115


Californium deposits mainly upon the endosteal surfaces of mineral bone and is only slowly redistributed throughout its volume by processes such as resorption and burial (Durbin, 1973). For this reason all isotopes of californium considered in this report are assumed, for the purposes of dosimetry, to be uniformly distributed over bone surfaces at all times following their deposition in the skeleton.

References Durbin, P. W. Metabolism and biological effects of the transplutonium elements. In: Uranium, Plutonium,

Tronsplufonium Efements, Eds. Hodge, H. C., Stannard, J. N. and Hursh, J. B., pp. 739-896. SpringerVerlag, Berlin, 1973.

ICRP Publication 19, TheMetabolism of Compounds of PIulonium and Other Actinides, Pergamon Press, Oxford, 1972.

ICRP Publication 23, Report of the Task Group on Reference Mon. Pergamon Press, Oxford, 1975. Parker, H. G., Low-Beer, A. de G. and Isaac, E. L. (1962). Comparison of retention and organ distribution of

Amz41 and Cf”’ in mice: The effect of in uiuo DTPA chelation. Heolfh Phys. 8, 679-684. Richmond, C. R. and Thomas, R. L. (1975). Plutonium and other actinide elements in gonadal tissue of man

and animals. Health Phys. 29, 241-250. Rundo, J. and Sedlet, J. Retention and elimination of berkelium-249. californium-249 following acute

accidental inhalation. In: Proc. Third Int. Cong. Int. Rodiot. Prot. Assn., Ed. Snyder, W. S., Vol. I, pp. 731- 735 (CONF 730907-PI), 1974.

Sullivan, M. F. Absorption of curium-244 and califomium-252 from the gastrointestinal tract of new born and adult rats. In: Pacific Norfhwest Laboratory Annual Report for 1973 to the USAEC Division of Biomedical and Enuironmenfal Research. Port I Biological Sciences, pp. 15-17 (BNWL-I850 Pt. l), 1974.

116 REPORT OF COMMITTIS 2

Annual limits on intake, ALI and derived air concentrations. DAC@q/m’) (40 h/wk) for isotopes of califomium

Oral

Inhalation

Class W Class Y

Fbdionuclide fi = 5 x lo-* f, = 5 x 10-4 f, = 5 x lo-*

2’4cf AL1

DAC 266Cf ALI

DAC 2q-J AL1

DAC ‘*gCf AL1

DAC zsocf ALI

DAC “‘Cf AL1

DAC tszcf ALI

DAC ‘53(-J AL1

DAC s’*Cf AL1

DAC

9 x lo*

(:T”G) -

1 x IO’

?L%Z -

8 x 105 (1 x 106) Bone surf.

- 4 x IO4

(8 x lo*) Bone surf.

- 1 x 10’


- 4 x Ioc

(8 x 10.) Bone surf.

- 2 x 10s


- 2 x 10’

(3 x IO’) Bone surf.

- 1 x 10”

-

2 x 10’

9 x 10’ 4 x IO’

2 x 102 3 x 10’


1 2 x 102


8 x IO-’ 5 x 10’


2 x 10-I 2 x lo=


8 x 1O-2 1 x 10’


4 x 10-l 1 x 10.

3 x 10’ 8 x 10’ 4 x 10-l

2 x 10’

9 x 10’ 3 x 10’

1 x 102 4 x IO’

2 5 x 10’


2 x 10-i 1 x 10’

4 x 10-I 5 x 10’


2 x 10-l 1 x 10’

6 x 10-l 6 x 104

3 x IO’ 6 x lOa 3 x 10-I

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