INTERNATIONAL PROGRAMME ON CHEMICAL SAFETY
ENVIRONMENTAL HEALTH CRITERIA 25
SELECTED RADIONUCLIDES
TRITIUM
CARBON-14
KRYPTON-85
STRONTIUM-90
IODINE
CAESIUM-137
RADON
PLUTONIUM
This report contains the collective views of an international group of
experts and does not necessarily represent the decisions or the stated
policy of the United Nations Environment Programme, the International
Labour Organisation, or the World Health Organization.
Published under the joint sponsorship of
the United Nations Environment Programme,
the International Labour Organisation,
and the World Health Organization
World Health Orgnization
Geneva, 1983
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CONTENTS
Paragraphs
ENVIRONMENTAL HEALTH CRITERIA FOR SELECTED RADIONUCLIDES
PREFACE . . . . . . . . . . . . . . . . . . . . . . 1 - 6
I. INTRODUCTION . . . . . . . . . . . . . . . . . . . . 7 - 22
II. TRITIUM . . . . . . . . . . . . . . . . . . . . . . 23 - 77
A. INTRODUCTION . . . . . . . . . . . . . . . . . . 23 - 25
B. SOURCES . . . . . . . . . . . . . . . . . . . . 26 - 57
1. Natural tritium . . . . . . . . . . . . . . 26 - 29
2. Nuclear explosions . . . . . . . . . . . . . 30 - 33
3. Nuclear fuel cycle . . . . . . . . . . . . . 34 - 51
4. Tritium production plants . . . . . . . . . 52 - 54
5. Consumer products . . . . . . . . . . . . . 55 - 56
6. Controlled thermonuclear reactors . . . . . 57
C. BEHAVIOUR IN THE ENVIRONMENT . . . . . . . . . . 58 - 61
1. Natural and fallout tritium . . . . . . . . 58 - 59
2. Industrial releases . . . . . . . . . . . . 60 - 61
D. TRANSFER TO MAN . . . . . . . . . . . . . . . . 62 - 63
E. DOSIMETRY . . . . . . . . . . . . . . . . . . . 64 - 77
1. Dose per unit intake . . . . . . . . . . . . 64 - 66
2. Dose per unit release . . . . . . . . . . . 67 - 77
F. REFERENCES
III. CARBON-14 . . . . . . . . . . . . . . . . . . . . . 78 - 112
A. INTRODUCTION . . . . . . . . . . . . . . . . . . 78 - 80
B. SOURCES . . . . . . . . . . . . . . . . . . . . 81 - 98
1. Natural carbon-14 . . . . . . . . . . . . . 81
2. Nuclear explosions . . . . . . . . . . . . . 82 - 84
3. Nuclear fuel cycle . . . . . . . . . . . . . 85 - 98
C. BEHAVIOUR IN THE ENVIRONMENT . . . . . . . . . . 99 - 102
D. TRANSFER TO MAN . . . . . . . . . . . . . . . . 103 - 105
E. DOSIMETRY . . . . . . . . . . . . . . . . . . . 106 - 112
1. Dose per unit intake . . . . . . . . . . . . 106 - 107
2. Dose per unit release . . . . . . . . . . . 108 - 112
F. REFERENCES
IV. KRYPTON-85 . . . . . . . . . . . . . . . . . . . . . 113 - 150
A. INTRODUCTION . . . . . . . . . . . . . . . . . . 113 - 117
B. SOURCES . . . . . . . . . . . . . . . . . . . . 118 - 128
1. Natural krypton-85 . . . . . . . . . . . . . 121
2. Nuclear explosions . . . . . . . . . . . . . 122 - 123
3. Nuclear fuel cycle . . . . . . . . . . . . . 124 - 128
C. BEHAVIOUR IN THE ENVIRONMENT . . . . . . . . . . 129 - 137
1. Dispersion in the atmosphere . . . . . . . . 130 - 133
2. Removal from the atmosphere . . . . . . . . 134 - 137
D. TRANSFER TO MAN . . . . . . . . . . . . . . . . 138 - 141
E. DOSIMETRY . . . . . . . . . . . . . . . . . . . 142 - 150
1. Dose per unit exposure . . . . . . . . . . . 143 - 144
2. Dose per unit release . . . . . . . . . . . 145 - 150
F. REFERENCES
V. STRONTIUM-90 . . . . . . . . . . . . . . . . . . . . 151 - 211
A. INTRODUCTION . . . . . . . . . . . . . . . . . . 151 - 154
B. SOURCES . . . . . . . . . . . . . . . . . . . . 155 - 165
1. Nuclear explosions . . . . . . . . . . . . . 155 - 156
2. Nuclear fuel cycle . . . . . . . . . . . . . 157 - 165
C. BEHAVIOUR IN THE ENVIRONMENT . . . . . . . . . . 166 - 185
1. Movement in soil . . . . . . . . . . . . . . 166
2. Transfer to plants . . . . . . . . . . . . . 167 - 171
3. Transfer to milk . . . . . . . . . . . . . . 172
4. Transfer to diet . . . . . . . . . . . . . . 173 - 181
5. Aquatic behaviour . . . . . . . . . . . . . 182 - 185
D. TRANSFER TO MAN . . . . . . . . . . . . . . . . 189 - 192
E. DOSIMETRY . . . . . . . . . . . . . . . . . . . 193 - 211
1. Dose per unit intake . . . . . . . . . . . . 193 - 197
2. Dose per unit release . . . . . . . . . . . 198 - 211
F. REFERENCES
VI. IODINE . . . . . . . . . . . . . . . . . . . . . . . 212 - 269
A. INTRODUCTION . . . . . . . . . . . . . . . . . . 212 - 214
B. SOURCES . . . . . . . . . . . . . . . . . . . . 215 - 234
1. Natural production . . . . . . . . . . . . . 215 - 216
2. Nuclear explosions . . . . . . . . . . . . . 217 - 220
3. Nuclear fuel cycle . . . . . . . . . . . . . 221 - 234
C. BEHAVIOUR IN THE ENVIRONMENT . . . . . . . . . . 235 - 255
1. Nuclear explosions . . . . . . . . . . . . . 235 - 241
2. Industrial releases . . . . . . . . . . . . 242 - 255
D. TRANSFER TO MAN . . . . . . . . . . . . . . . . 256 - 259
E. DOSIMETRY . . . . . . . . . . . . . . . . . . . 260 - 270
1. Dose per unit intake . . . . . . . . . . . . 260 - 261
2. Dose per unit release . . . . . . . . . . . 262 - 270
F. REFERENCES
VII. CAESIUM-137 . . . . . . . . . . . . . . . . . . . . 271 - 336
A. INTRODUCTION . . . . . . . . . . . . . . . . . . 271 - 274
B. SOURCES . . . . . . . . . . . . . . . . . . . . 275 - 282
1. Nuclear explosions . . . . . . . . . . . . . 275 - 276
2. Nuclear fuel cycle . . . . . . . . . . . . . 277 - 282
C. BEHAVIOUR IN THE ENVIRONMENT . . . . . . . . . . 283 - 309
1. Fixation in soil . . . . . . . . . . . . . . 283 - 286
2. Transfer to plants . . . . . . . . . . . . . 287 - 290
3. Transfer to milk . . . . . . . . . . . . . . 291
4. Transfer to meat . . . . . . . . . . . . . . 292
5. Transfer to diet . . . . . . . . . . . . . . 293 - 301
6. The lichen-caribou-man foodchain . . . . . . 302 - 303
7. Aquatic behaviour . . . . . . . . . . . . . 304 - 309
D. TRANSFER TO MAN . . . . . . . . . . . . . . . . 310 - 319
1. Absorption and distribution in tissues . . . 310 - 314
2. Retention half-time . . . . . . . . . . . . 315 - 317
3. Transfer factor . . . . . . . . . . . . . . 318 - 319
E. DOSIMETRY . . . . . . . . . . . . . . . . . . . 320 - 336
1. Dose per unit intake . . . . . . . . . . . . 320 - 324
2. Dose per unit release . . . . . . . . . . . 325 - 336
F. REFERENCES
VIII. RADON . . . . . . . . . . . . . . . . . . . . . . . 337 - 395
A. INTRODUCTION . . . . . . . . . . . . . . . . . . 337 - 340
B. SOURCES . . . . . . . . . . . . . . . . . . . . 341 - 351
1. Outdoors . . . . . . . . . . . . . . . . . . 341 - 344
2. Indoors . . . . . . . . . . . . . . . . . . 345 - 351
C. BEHAVIOUR IN THE ENVIRONMENT . . . . . . . . . . 352 - 375
1. Release from soil . . . . . . . . . . . . . 352 - 355
2. Dispersion in air . . . . . . . . . . . . . 356 - 361
3. Indoor behaviour . . . . . . . . . . . . . . 362 - 365
4. Radon daughter concentrations . . . . . . . 366 - 375
D. TRANSFER TO MAN . . . . . . . . . . . . . . . . 376 - 380
E. DOSIMETRY . . . . . . . . . . . . . . . . . . . 381 - 395
1. Dose per unit exposure . . . . . . . . . . . 381 - 393
2. Dose per unit release . . . . . . . . . . . 394 - 395
F. REFERENCES
IX. PLUTONIUM . . . . . . . . . . . . . . . . . . . . . 396 - 456
A. INTRODUCTION . . . . . . . . . . . . . . . . . . 396 - 401
B. SOURCES . . . . . . . . . . . . . . . . . . . . 402 - 411
1. Nuclear explosions . . . . . . . . . . . . . 402 - 404
2. Nuclear fuel cycle . . . . . . . . . . . . . 405 - 406
3. Other sources . . . . . . . . . . . . . . . 407 - 411
C. BEHAVIOUR IN THE ENVIRONMENT . . . . . . . . . . 412 - 434
1. Movement in soil . . . . . . . . . . . . . . 412 - 416
2. Transfer to plants . . . . . . . . . . . . . 417 - 418
3. Transfer to animals . . . . . . . . . . . . 419 - 420
4. Transfer to diet . . . . . . . . . . . . . . 421 - 425
5. Aquatic behaviour . . . . . . . . . . . . . 426 - 434
D. TRANSFER TO MAN . . . . . . . . . . . . . . . . 435 - 443
E. DOSIMETRY . . . . . . . . . . . . . . . . . . . 444 - 456
1. Dose per unit intake . . . . . . . . . . . . 444 - 448
2. Dose per unit release . . . . . . . . . . . 449 - 456
F. REFERENCES
X. RADIATION EFFECTS . . . . . . . . . . . . . . . . . 457 - 476
A. SOMATIC EFFECTS . . . . . . . . . . . . . . . . 459 - 463
1. Early somatic effects . . . . . . . . . . . 459 - 461
2. Late somatic effects . . . . . . . . . . . . 462 - 463
B. GENETIC EFFECTS . . . . . . . . . . . . . . . . 464 - 465
C. DOSE-RESPONSE RELATIONSHIPS . . . . . . . . . . 466 - 469
D. RISK ESTIMATES . . . . . . . . . . . . . . . . . 470 - 476
XI. CONCLUSIONS . . . . . . . . . . . . . . . . . . . . 477 - 491
A. RADIONUCLIDES AND THE ENVIRONMENT . . . . . . . 477 - 481
B. DOSE ASSESSMENTS . . . . . . . . . . . . . . . . 482 - 487
C. EFFECTS EVALUATION . . . . . . . . . . . . . . . 488 - 491
XII. ANNEX
EXCERPTS FROM "BASIC SAFETY STANDARDS FOR RADIATION
PROTECTION 1982 EDITION"
NOTE TO READERS OF THE CRITERIA DOCUMENTS
While every effort has been made to present information in the
criteria documents as accurately as possible without unduly
delaying their publication, mistakes might have occurred and are
likely to occur in the future. In the interest of all users of the
environmental health criteria documents, readers are kindly
requested to communicate any errors found to the Division of
Environmental Health, World Health Organization, Geneva,
Switzerland, in order that they may be included in corrigenda which
will appear in subsequent volumes.
In addition, experts in any particular field dealt with in the
criteria documents are kindly requested to make available to the
WHO Secretariat any important published information that may have
inadvertently been omitted and which may change the evaluation of
health risks from exposure to the environmental agent under
examination, so that the information may be considered in the event
of updating and re-evaluation of the conclusions contained in the
criteria documents.
ENVIRONMENTAL HEALTH CRITERIA FOR SELECTED RADIONUCLIDES
At the request of the United Nations Environment Programme
(UNEP), the United Nations Scientific Committee on the Effects of
Atomic Radiation (UNSCEAR) prepared a paper on the Environmental
Behaviour and Dosimetry of Radionuclides. In accordance with the
UNEP proposal, the paper, which was prepared during the 27th - 29th
sessions of the Committee and was completed and approved at the
30th session in 1981, is now being published in the WHO/UNEP
Environmental Health Criteria series. The EHC document, which is
entitled "Selected Radionuclides", comprises the integral report
prepared and edited by UNSCEAR, together with an annex consisting
of excerpts taken from "Basic Safety Standards for Radiation
Protection 1982 Edition", Safety Series No 9, a document prepared
jointly by IAEA/ILO/NEA(OECD)/WHO, and published by the
International Atomic Energy Agency, to give guidance to the
appropriate national authorities on the establishment of limits for
radionuclides. The selected radionuclides discussed in the
Environmental Health Criteria document are those of environmental
importance for the general population and radiation workers.
Dr E. Komarov, Environmental Health Division, World Health
Organization, was responsible for the final layout of the
Environmental Health Criteria document.
The assistance of Dr B.G. Bennett (Monitoring and Assessment
Research Centre, MARC) in the scientific editing of the
Environmental Health Criteria document is gratefully acknowledged.
The contents of the 1982 UNSCEAR report to the General Assembly
of the United Nations were taken into account during the
preparation of the paper on the Environmental Behaviour and
Dosimetry of Radionuclides, but the report was not quoted as it had
not been issued at that time.
ENVIRONMENTAL BEHAVIOUR AND DOSIMETRY OF RADIONUCLIDES
1. PREFACE
1. The release of radioactive materials to the environment
potentially exposes populations to ionizing radiation and increases
the risk of incurring deleterious health effects. The associations
of released amounts to effects establish the health criteria for
radionuclides, i.e., the quantitative relationships that would be
required to establish release limits governing the management of
radioactive materials used by man.
2. This report has been prepared by the United Nations Scientific
Committee on the Effects of Atomic Radiation (UNSCEAR) for the
United Nations Environment Programme (UNEP) to provide background
information in establishing such health criteria. In this report
the more general considerations of environmental behaviour of
several radionuclides are discussed, including sources, transport
to man and dosimetry. The radionuclides discussed are those most
frequently released from natural and man-made sources and the
greatest contributors to population radiation exposure under normal
circumstances.
3. The compilation of the relevant information is based largely on
the detailed presentations and evaluations of the sources of
ionizing radiation by UNSCEAR in its reports to the United Nations
General Assembly. The reader is referred to these reports for
general concepts and for assessments of the dose commitments to man
from exposures to sources such as natural radioactivity, fallout
from atmospheric nuclear testing, releases from nuclear power
production, occupational and medical irradiations.
4. Further information to be considered in establishing health
criteria for radionuclides is that on health effects of
irradiations. The relationships between radiation dose and risks
of health effects in man have recently been re-evaluated based on
the available data. This information can be found in the 1977
report of UNSCEAR. Only a brief summary of the general aspects of
radiation effects and of radiation protection considerations is
presented here.
5. The establishment of release limits for radionuclides in
particular situations cannot be accomplished without rather more
detailed considerations of the local and regional environment and
the special pathways of transfer to man. With this in mind, it is
recognized that the material given here can only serve as
background guidance.
6. The following scientists have contributed in the preparation of
this report: Dr. W.J. Bair, Dr. D. Beninson, Dr. B.G. Bennett, Dr.
A. Bouville, Dr. P. Patek, Dr. G. Silini and Dr. J.O. Snihs.
I. INTRODUCTION
7. Radionuclides are a special class of environmental substances.
They are the unstable configurations of chemical elements which
undergo radioactive decay, emitting radiation in the form of alpha
or beta particles and x or gamma rays. The interaction of radiation
with biological materials causes energy to be released to these
materials which may result in a variety of harmful effects.
Radiation is thus a potential hazard to man, although it may also
be used in many beneficial ways, as in medical diagnosis and
treatment, in industrial and consumer products and in the
generation of electricity with nuclear reactors.
8. The realization of the harmful potential of ionizing radiation,
which was dramatically brought to the attention of the public by
the atomic bombing of Hiroshima and Nagasaki in 1945, was the cause
of considerable attention that has been paid throughout the years
to the effects of radiation. As a result of these studies, a great
deal is now known about radionuclide behaviour in the environment
and in man and about the somatic and genetic consequences of
irradiation. This information surpasses by far that relating to
any other class of environmental pollutants.
9. Considerable experience has been gained in environmental
radiation measurements, particularly in tracing the movement of
fallout radionuclides produced in atmospheric testing of nuclear
weapons. Much of this information has in turn contributed to the
general knowledge of atmospheric and oceanic transport processes
and of bio-geochemical cycles of elements. Extensive studies of
radiation effects in animals and numerous epidemiological surveys
of exposed population groups have by now been conducted. They have
considerably enlarged our understanding of the biological effects
of radiation on man and the environment, although uncertainties
still remain, particularly regarding the basic mechanisms of action
and the risk evaluations at low doses and dose rates [U1-U7].
10. A few definitions and general concepts should be introduced
before the detailed presentation of radionuclide assessments. The
basic unit of radioactivity is the becquerel (Bq), corresponding to
one disintegration per second. The previously used unit was the
curie (Ci), one Ci corresponding to 3.7 1010 Bq.
11. The basic measure of radiation interaction in irradiated
materials is the absorbed dose (D). This quantity is also the
basis of health risk estimates, under the assumption of a linear
relationship between dose and risk. The absorbed dose is defined
as the mean energy (joules) imparted to the irradiated material per
unit mass (kg) at the point of interest. The unit of absorbed dose
is ca11ed the gray (Gy) which corresponds to 1 J/kg. The unit of
absorbed dose previously in use, the rad, is one hundred times
smaller than the Gy.
12. Radiations of different types and energies have different
effectiveness for producing effects, depending on the amount of
energy transferred per unit length (LET) along the path of the
charged particles. In order to quantify this differing
effectiveness, use is made of a normalizing quantity called the
quality factor (Q). For general purposes of radiological
protection the assumed values of Q are: 1 for x and gamma rays and
for electrons; 10 for neutrons and protons; 20 for alpha and
multiply charged particles.
13. The product of the absorbed dose, D, and the quality factor,
Q, is termed the dose equivalent (H). The unit of dose equivalent
is the sievert (Sv). The previously used unit was the rem (1 rem =
0.01 Sv). Use of the dose equivalent allows the summation of doses
from all types of radiation of different biological effectiveness.
14. The exposure of an individual to a source of radiation may be
expressed in terms of the absorbed dose or dose equivalent during
the period of exposure. In the natural radiation environment the
exposure is continuous and it is sufficient to give the annual
average dose or dose rate. There are important spatial variations
to be considered, for example, as a function of the altitude in
case of exposure to cosmic radiation or as a function of the
geographical location due to the different radionuclides present in
soil.
15. For specific releases of radioactive materials into the
environment (atmospheric nuclear tests, operation of nuclear
reactors) there are also important temporal variations in the
exposure. In order to account for the exposures which will occur
in the future from specific sources, use is made of the dose
commitment (Dc). This quantity is the infinite time integral of
the average individual dose rate. Dose commitments may not
represent doses to specific individuals. For example, if the
radionuclide released has a very long half-life, the dose
commitment is derived from the doses to successive generations in
the population.
16. The collective harm to a population resulting from the
exposure of all individuals is related to the collective dose in
the population, particularly if the linearity of the relationships
between dose and effects may be assumed for the exposures involved.
The collective dose (S) in a given population is the summation of
products of the average individual doses and the number of
individuals in each range of doses. The summation may become an
integral for continuous variations over the entire range of doses.
The unit of the collective dose is man Gy and the corresponding
unit of collective dose equivalent is man Sv.
17. The measure of the total exposure of a population from a
specified source or release practice is the collective dose
commitment (Sc), defined as the infinite time integral of the
collective dose rate. The relevant units are man Gy, or man Sv in
case of the collective dose equivalent commitment.
18. In radiation exposure assessments, it is often necessary to
account for the different sensitivity of individual organs of the
body with respect to each other or to irradiation of the whole
body, particularly in the case of internally deposited
radionuclides. Weighting factors for the relevant organs may be
derived for this purpose from relative risk estimates. These
factors will be listed in the section on radiation effects with
some additional discussion.
19. The summation of the products of the weighting factors and the
dose equivalents for individual organs gives a single measure to be
used as an index of health detriment, called the effective dose
equivalent (HE). The concepts of collective and committed doses
may also be used with this quantity. Thus a final quantity for
health assessments may be the collective effective dose equivalent
commitment, (ScE) which is a collective dose, weighted for the
effects of doses within the body and dose distributions within the
population.
20. The chain of events leading from the release of radioactive
materials into the environment to the irradiation of human tissues
may be expressed schematically as a series of compartments
connected by transfer pathways. Such models are necessarily
simplifications of the actual transfer pathways. The following
diagram illustrates the transfer stages most usually considered in
assessments by UNSCEAR.
21. The basic task in the assessment process is to evaluate the
transfer factors (Pi,j) which relate the appropriate quantity of
radioactivity amount or dose in step i of the sequence to the
appropriate quantity in the subsequent step j. Since the desired
quantity in the final step is the time integrated dose rate, the
dose commitment from a specific source, the quantities in the other
steps are the time integrated activity concentrations. The
transfer factor is the quotient of time integrated quantities in
successive compartments. The total transfer factor for steps in
series is the product of the transfer factors involved. The total
transfer factor of several parallel pathways is the sum of the
transfer factors of the individual pathways.
22. There are many common features of the behaviour of different
radionuclides in the environment and their transfer to man. For
example, the physical dispersion of radionuclides in the
environment following release from a source is largely the same for
broad classes of material, such as particulates and gases. Several
models used to describe the transfer of radioactive material within
an environmental medium or from one medium to the next have general
applicability. A review of such general behaviour and modelling
procedures can be found in the 1982 report of UNSCEAR [U8].
Therefore, in the presentations which follow only the rather more
specific aspects of environmental behaviour and dosimetry of the
radionuclides are considered.
REFERENCES
U1 United Nations. Report of the United Nations Scientific
Committee on the Effects of Atomic Radiation. Official
Records of the General Assembly, Thirteenth Session,
Supplement No. 17 (A/3838). New York, 1958.
U2 United Nations. Report of the United Nations Scientific
Committee on the Effects of Atomic Radiation. Official
Records of the General Assembly, Seventeenth Session,
Supplement No. 16 (A/5216). New York, 1962.
U3 United Nations. Report of the United Nations Scientific
Committee on the Effects of Atomic Radiation. Official
Records of the General Assembly, Nineteenth Session,
Supplement No. 14 (A/5814). New York, 1964.
U4 United Nations. Report of the United Nations Scientific
Committee on the Effects of Atomic Radiation. Official
Records of the General Assembly, Twenty-first Session,
Supplement No. 14 (A/6314). New York, 1966.
U5 United Nations. Report of the United Nations Scientific
Committee on the Effects of Atomic Radiation. Official
Records of the General Assembly, Twenty-fourth Session,
Supplement No. 13 (A/7613). New York, 1969.
U6 United Nations. Ionizing Radiation: Levels and Effects.
A report of the United Nations Scientific Committee on the
Effects of Atomic Radiation to the General Assembly, with
annexes. United Nations sales publication, No. E.72.IX.17
and 18. New York, 1972.
U7 United Nations. Sources and Effects of Ionizing
Radiation. United Nations Scientific Committee on the
Effects of Atomic Radiation 1977 report to the General
Assembly, with annexes. United Nations sales publication
No. E.77.IX.I. New York, 1977.
U8 United Nations. Ionizing Radiation: Sources and
Biological Effects. United Nations Scientific Committee
on the Effects of Atomic Radiation 1982 report to the
General Assembly, with annexes. United Nations sales
publication No. E.82.IX.8. New York, 1982.
II. TRITIUM
A. INTRODUCTION
23. Tritium, 3H, is a radioactive isotope of hydrogen which decays
into the stable nuclide 3He. Tritium is a pure beta-emitter with a
half-life of 12.3 a, a maximum energy of 18 keV and an average
energy of 5.7 keV. Tritium is produced naturally in the
atmosphere, where it results from the interaction of cosmic ray
protons and neutrons with nitrogen, oxygen, and argon. Man-made
tritium, in amounts substantially larger than the natural
inventory, has been injected into the stratosphere by thermonuclear
explosions. In addition, tritium is produced during the operation
of nuclear reactors.
24. There are many applications of tritium in industry. It is
widely used in consumer products, such as radioluminous timepieces
and also as a tracer in biomedical research. Environmental tritium
is mainly found as tritiated water. As such, it follows the
hydrological cycle and penetrates into all components of the
biosphere, including man.
25. This document is mainly based on the 1977 UNSCEAR report [U1],
but makes also extensive use of the contents of recent reviews or
symposia on tritium [I4, J1, M7, N1, N2].
B. SOURCES
1. Natural tritium
26. Natural tritium is produced by nuclear reactions in the
atmosphere and, to a much smaller extent, in the hydrosphere and in
the lithosphere. In addition, some tritium may be created in the
extra-terrestrial environment and enter the atmosphere along with
cosmic rays. Most of the natural tritium is found in the
environment as tritiated water, generally designated as HTO.
27. In the atmosphere, natural tritium is produced by the
interaction of high energy cosmic rays with atmospheric nitrogen
and oxygen. The estimates of the number of atoms of tritium
produced per unit time and per unit area of the earth's surface
range from 0.1 to 1.3 cm-2 s-1 [U1]. In the UNSCEAR 1977 report
[U1], a production rate of 0.25 cm-2 s-1 was adopted; this
corresponds to a production rate of 3.6 1016 Bq a-1 in each
hemisphere and to a global inventory of 1.3 1018 Bq at equilibrium.
28. It has been suggested that tritium might be ejected from the
sun during solar flares [L1] and from stars [F1]. Flamm et al.
[F2] estimated that the solar flares could account for an
additional production rate, averaged over the solar cycle, of 0.1
cm-2 s-1.
29. In the lithosphere and in the hydrosphere, tritium is produced
by interaction of neutrons with 6Li nuclides. The production rates
have been assessed at 10-3 cm-2 s-1 in the lithosphere and at 10-6
cm-2 s-1 in the hydrosphere [F1, K1].
2. Nuclear explosions
30. Nuclear tests have been conducted in the atmosphere since 1945
and have produced tritium in amounts that greatly exceed the global
natural activity. The tritium activity arising from atmospheric
nuclear tests can be estimated from the fission and fusion yields
or from environmental measurements.
31. Bennett [B1] has published an estimate of the total and
fission yields for each reported atmospheric test from 1945 to
1978; according to that compilation, 422 nuclear tests were
conducted in the atmosphere up to 1979, with cumulative yields of
217 Mt and 328 Mt for fission and fusion, respectively. The tritium
activity produced per unit yield depends on the characteristics of
the device, as well as on those of the explosion site, but is in
any case much greater for fusion than for fission [N1]. Miskel
[M1] estimated the yield for fission explosion to be 2.6 1013 Bq
Mt-1 and that for fusion to be typically 7.4 1017 Bq Mt-1. The
total tritium activity produced by atmospheric tests is thus
assessed at
328 Mt (fusion) x 7.4 1017 Bq Mt-1 = 2.4 1020 Bq
Most of this activity was produced during the large yield test
series which took place during 1954-1958 and 1961-1962; the
contribution of the nuclear tests carried out since 1964 is less
than 5% of the total.
32. Almost all the tritium produced by fallout occurs as HTO in
the atmosphere and the hydrosphere, and thus follows the
hydrological cycle. The total activity injected can therefore be
conceivably derived from measured concentrations in water samples.
From the study of Schell et al. [S1] on the tritium concentrations
in precipitation at stations in the IAEA network, it can be
estimated [U1] that the total production was about 1.7 1020 Bq.
Other estimates, using vertical profiles of 3H in the oceans as a
basis, lead to injections of 1.2 1020 Bq (in the oceans only) [O1],
1.3 1020 Bq [B2, U1], and 2.0 1020 Bq [M2].
33. All the estimates presented above are in fairly good
agreement, as they lie in the limited range from 1.2 1020 to 2.4
1020 Bq. In its 1977 report UNSCEAR adopted a value of 1.7 1020 Bq
for the total globally dispersed activity of tritium produced in
atmospheric tests up to 1976 [U1].
3. Nuclear fuel cycle
34. Tritium occurs in nuclear reactors by ternary fission in the
fuel and also by neutron activation reactions with lithium and
boron isotopes dissolved in, or in contact with, the primary
coolant as well as with naturally-occurring deuterium in the
primary coolant (Figure II.I).
35. Most of the fission product tritium produced in the fuel rods
is usually retained within the fuel and is not released into the
environment at the reactor site; it is instead released during fuel
reprocessing, if that practice is carried out. The activity
produced in the coolant is partly or entirely released in the
effluent streams according to the waste management practices at the
plant.
36. Releases into the environment are mainly in the form of HTO in
reactors that use water as primary coolant, as well as in fuel
reprocessing plants.
(a) Nuclear reactors
37. Four types of reactors have been considered (PWR, BWR, HWR,
GCR), the emphasis being on PWRs and BWRs which currently represent
the largest share of nuclear capacity. Estimated generation rates
and appearance of tritium in effluent streams of reactors are
summarized in Table II.1.
38. The annual production of fission product tritium in the fuel
rods of a pressurized water reactor (PWR) is in the range of 6 to 9
1011 Bq per MW(e)a [N1]. A small percentage, 1% or less, is
expected to be released into the coolant through defects in the
cladding, currently made of zirconium alloy. In contrast, the use
of stainless steel cladding in earlier PWRs resulted in the release
to the coolant of most of the tritium produced in the fuel.
39. Tritium generation in the primary coolant (water) of a
PWR is mainly due to reactions with boron (2.6 1010 Bq per
MW(e)a) which is dissolved as boric acid to control
reactivity; in addition, the maintenance of 2 ppm lithium
hydroxide for pH control [L2] results in the formation of
about 7 108 Bq per MW(e)a.
40. Environmental tritium discharges from PWRs depend on waste
management practices as well as on the materials used in the
reactor. Average normalized releases of tritium were shown in the
UNSCEAR 1977 report [U1] to be about 7 1010 Bq per MW(e)a in liquid
effluents and 7 109 Bq per MW(e)a in airborne effluents for the
reactors in operation in 1973-1974. However, large differences
between PWRs are due to the type of fuel cladding. For an old
reactor using stainless steel Kahn et al. [K3] measured 3H releases
of about 4 1011 Bq per MW(e)a in liquid effluents and 4 1010 Bq per
MW(e)a in airborne effluents, whereas the combined releases of 9
PWRs with zirconium alloy clad fuel (current practice) were
reported by NCRP [N1] to be about 3 1010 and 109 Bq per MW(e)a in
liquid and airborne effluents, respectively.
41. In boiling water reactors (BWRs) tritium is produced by
ternary fission in the fuel at about the same rate as in PWRs (6 to
9 1011 Bq per MW(e)a). The generalized use of zirconium alloy
cladding limits the tritium release into the coolant to less than 7
109 Bq per MW(e)a.
42. Tritium can be generated by neutron activation in the coolant
and in the control rods. Prior to 1971, control rods of boron
carbide were used in BWRs [S2]; the production of tritium by
activation of these control rods has been estimated to be about 3
1011 Bq per MW(e)a. However, tritium has not been shown to diffuse
through the boron carbide matrix [T1]. In the coolant itself,
tritium is generated by activation of deuterium at a rate of about
4 108 Bq per MW(e)a.
43. Tritium activities discharged from BWRs into the environment
are lower than those of PWRs because less tritium is produced in or
diffuses into the primary coolant. UNSCEAR [U1] reported the
average discharge rates to be 4 109 and 2 109 Bq MW(e)a in liquid
and airborne effluents, respectively.
44. The amount of tritium generated in fuel of heavy water
reactors (HWR) by ternary fission is approximately the same as in
light water reactors, but it is largely exceeded by the production
in the D2O coolant and moderator by neutron activation, which has
been estimated to be about 2 1013 Bq per MW(e)a [K2].
Table II.1 Estimated rates of generation of tritium and of its release in effluent streams of
different types of reactors (1010 Bq per MW(e)a) [G1, K2, S2, T1, U1]
---------------------------------------------------------------------------------------------------------
PWR BWR HWR GCR
Source ------------------------------------------------------------------------------------------
Generation Effluent Generation Effluent Generation Effluent Generation Effluent
stream stream stream stream
---------------------------------------------------------------------------------------------------------
Fission 75 < 0.7 75 < 0.7 55 < 0.6 75 < 0.7
Activation
Deuterium 0.004 0.004 0.04 0.04 2000 75a/
Lithium 0.07 0.07 2 0.4
Boron 2.6 2.6 30 0
Rounded total 80 3 110 0.5 2000 75 80 1
---------------------------------------------------------------------------------------------------------
a/ Depending on the irradiation time and on the net leakage of heavy water.
45. Environmental discharges depend upon the D2O leakage which is
kept as small as possible for economical and radiological reasons,
and upon the tritium activity in the coolant and moderator, which
builds up with the irradiation time. Annual losses of from 0.5% to
3% are anticipated in HWRs [U1]. For the optimal loss of 0.5% per
year, the normalized tritium release rate ranges from 1011 Bq per
MW(e)a in the first year of operation to about 7 1011 Bq per MW(e)a
in the tenth year. Based on the latter value as representative of
the reactor life, the normalized 3H release rates are estimated to
be 6 1011 and 1.5 1011 Bq per MW(e)a in airborne and liquid
effluents, respectively [G1]. Reported releases roughly agree with
these estimates: they are 6.3 1011 and 2.6 1011 Bq per MW(e)a for
the Pickering A station in Canada, in airborne and liquid
effluents, respectively whereas the Atucha reactor in Argentina
releases about 8 1011 Bq per MW(e)a both in airborne and in liquid
effluents.
46. In gas-cooled reactors (GCR), tritium is produced by ternary
fission (about 7 1011 Bq per MW(e)a) and by activation of lithium
in the graphite moderator. Based on the experience with UK
reactors (mainly Magnox reactors), the tritium release is about 7
109 Bq per MW(e)a in liquid effluents and ranges from 109 to 1010
Bq per MW(e)a in airborne effluents [U1].
(b) Fuel reprocessing plants
47. At the fuel reprocessing stage of the nuclear fuel cycle (if
it is undertaken) the elements uranium and plutonium in the
irradiated nuclear fuel are recovered for reuse in fission
reactors. When the fuel elements are reprocessed, the uranium is
first taken out of its cladding material and then dissolved in
nitric acid. Most of the tritium released from fuel during
dissolution appears in the liquid waste stream while some is
carried out in the dissolver off-gas stream and a portion is
immobilized as a solid zirconium compound in the cladding.
48. In 1980, the only reprocessing plants operating commercially
in the world were at Windscale (U.K.) and La Hague and Marcoule
(France); their combined capacity was much lower than the amount
of fuel discharged from reactors worldwide. Luykx and Fraser [L3]
have expressed the reported releases from the three reprocessing
plants during the 1974-1978 time period in terms of activity
discharged per unit of electricity generated. The average figures
for each plant are given in Table II.2.
Table II.2 Average normalized tritium activities
discharged into the environment by fuel
reprocessing plants (1010 Bq per MW(e)a) [L3]
--------------------------------------------------
Plant Airborne Liquid Total
location effluents effluents
--------------------------------------------------
Windscale 17 55 72
La Hague 0.4 28.5 29
Marcoule 5.2 41 46
--------------------------------------------------
49. As there is no retention system for tritium in the currently
operating reprocessing plants, the activity released corresponds to
that which is contained in the fuel elements (with the exclusion of
cladding) at the time of reprocessing. The production rate of
tritium in reactors being about 75 1010 Bq per MW(e)a (Table II.1),
approximately half of the theoretical fuel content seems to be
unaccounted for at the La Hague and Marcoule plants.
(c) Summary
50. In 1980, the installed nuclear capacity was 1.25 105 MW(e) on
a worldwide scale [I1]. Assuming an average load factor of 0.6,
the energy produced was 7.5 104 MW(e)a. Using the average figures
given previously for production and release in the types of
reactors considered, the global production and release of tritium
at the reactor sites in 1980 are estimated to be about 1.5 1017 Bq
and 4 1015 Bq, respectively. Table II.3 provides a breakdown of
the environmental discharges from reactors according to reactor
type.
Table II.3 Estimated global discharge of tritium from nuclear
power stations in 1980
----------------------------------------------------------------
Estimated tritium discharges
Reactor Number Capacity in 1980 (Bq)
type [MW(e)] Airborne Liquid Total
effluents effluents
----------------------------------------------------------------
PWR 96 64239 3.9 1013 1.2 1015 1.2 1015
BWR 62 35170 4.2 1013 8.4 1013 1.3 1014
HWR 14 5963 5.4 1014 2.1 1015 2.6 1015
GCR 36 7086 1.3 1013 3.0 1013 4.3 1013
Other 33 12527 - - -
----------------------------------------------------------------
Total 241 124985 6.3 1014 3.4 1015 4.0 1015
----------------------------------------------------------------
51. In comparison, the tritium releases reported for the three
currently operating commercial fuel reprocessing plants were about
2 1015 Bq in 1978. All together, the current tritium production
rate in the nuclear fuel cycle is comparable to the natural
production rate, whereas the release rate is about 20 times less.
4. Tritium production plants
52. Artificial production of tritium on an industrial scale is
necessary to provide an essential component of thermonuclear
weapons. In addition, relatively small amounts of tritium are used
for other industrial and scientific applications. The most
economical way to produce tritium is the irradiation of lithium
metal, alloys or salts in a nuclear reactor [J1]. The tritium is
isotopically separated from other hydrogen isotopes and is
processed in tritium-handling plants [C1].
53. Tritium airborne release rates from Savannah River Plant,
which is the primary production source of tritium in the U.S.A.,
have ranged from 1.4 1016 Bq a-1 to 9.9 1016 Bq a-1 from 1974 to
1977 with an average of 4.1 1016 Bq a-1 [M4]. Under normal
operating conditions, the releases are about 20% HT and 80% HTO.
However, accidental airborne releases, which seem to be essentially
in the gaseous HT form, have raised the contribution of HT to the
total activity released to 60% in 1974 and 57% in 1975 [M4]. The
activity of tritium released in the liquid effluents appears to be
about 10% of that in the airborne effluents [N1].
54. Data on releases from other tritium production plants have not
been found in the literature. However, an indirect estimate of 7
1016 Bq for the worldwide release of HT in 1977 has been made by
Mason and Ostlund [M3] on the basis of their measurements of the
atmospheric HT content.
5. Consumer products
55. Tritium has been used extensively in the dial-painting
industry for the illumination of timepieces, the radiation emitted
by 3H being converted into light by a scintillator which is usually
zinc sulfide containing small amounts of copper or silver. In
recent years, this illumination system has been in competition with
the tritium gas-filled glass tubes, coated internally with
phosphor, which are used to illuminate some types of LCD (liquid
crystal display) watches. Exit signs and electronic tubes are
other types of consumer products containing tritium [C2, K4, U1,
W1].
56. In luminous compounds, the fractional release rate of tritium,
in the form of HTO, HT and short-chain organic radicals of the
styrene type, is about 5% annually [K4, K5] while it is negligible
from gas-filled glass tubes. It has been estimated that about 7
1016 Bq was processed in 1978 in the worldwide production of
timepieces and that the activity released is probably under 1014 Bq
a-1 for luminous compounds and 2 1012 Bq a-1 for gas-filled glass
tubes [K5]. Environmental releases due to breakage through accident
or disposal could be more important [C2, W1].
6. Controlled thermonuclear reactors
57. Large-scale use of controlled thermonuclear reactors for heat
or power generation seems quite unlikely in the next 25 years.
However, if thermonuclear reactors come into use, they will contain
substantial inventories of tritium and will pose considerable
tritium management problems [N1]. The production of tritium in a
nominal 1000 MW(e) controlled thermonuclear reactor is anticipated
to be about 5 1017 Bq d-1 and the inventory of the order of 1019 Bq
[C1, C3, H1]. In order to prevent massive releases of tritium into
the environment, an extraordinary degree of control will be
required. However, conceptual designs for fusion power plants show
that the effluent release rate can be limited to 4 1013 Bq a-1 by
applying present-day tritium technology [C3].
C. BEHAVIOUR IN THE ENVIRONMENT
1. Natural and fallout tritium
58. Natural and fallout tritium are mainly produced in the
stratosphere where they are essentially found in the HTO form.
Tritiated water vapour is transferred from the stratosphere to the
troposphere with a half-time of about one year, then from
troposphere to the earth's surface through rainfall and molecular
exchange with a half-time of about ten days. Tritiated water then
follows the hydrological cycle. Water deposited on the ocean
surfaces is diluted in the mixed layer. Part of it evaporates back
to the atmosphere, with a much lower concentration, while a smaller
fraction is transferred to the deep ocean. Tritiated water
deposited on land surfaces is partitioned partly to surface run-off
(leading to a pond, a lake, a stream, or an ocean) and partly to
infiltration in the soil from where it can be absorbed by plants,
evaporate, or move with groundwater to a surface stream or to an
ocean.
59. Part of the tritiated water deposited on soils finds its way
into vegetable and animal products and thus contaminates dietary
foodstuffs. Tritium incorporated into those biological materials,
and in soil and sediments as well, is found to be present in at
least two separable fractions, one easily exchangeable, that is
available by freeze-drying (free water tritium fraction) and one
less easily exchangeable, available by combustion ("organically
bound" fraction) [B6]. The analysis of soil, water, and various
components of the diet in the New York area in 1978 [B6] revealed
that water, soil and diet were in equilibrium with respect to free
water tritium; however, the specific activities (activity
concentration per unit mass of hydrogen) of the "organically bound"
tritium in various foodstuffs were higher by a factor of 2 to 4
than those of the water tritium. It is suggested that tritium was
incorporated uniformly into biological materials during the period
of highest deposition rates in the early 1960s and that differences
in specific activities developed due to longer biological residence
half-time of the "organically bound" fraction compared to the free
water tritium fraction [B6].
2. Industrial releases
60. Industrial releases consist mainly of HTO and HT, and probably
tritiated methane, CH3T [B7]. The residence times of HT and CH3T
in the atmosphere are not known with certainty but the estimates
point to average values of 5 to 10 years [B7]. The main removal
processes are bacterial action and photochemical oxidation for HT
and photochemical oxidation alone for CH3T [B7]. In both cases,
the resulting product is presumably HTO. As HTO is much more
biologically active than HT and CH3T, it is this tritium compound
that is of most concern in the case of industrial releases.
61. Industrial releases may be to the atmosphere or to water
(river or sea). In addition, releases to ground water have taken
place but they are of little consequence as the movement of water
in suitable aquifers is very slow. The environmental behaviour of
HTO released by industry is not different from that from natural or
fallout sources.
D. TRANSFER TO MAN
62. Transfer to man of environmental HTO takes place via
inhalation, diffusion through skin and ingestion of beverages and
foodstuffs; in the case of HT, inhalation is the only meaningful
pathway to man. Exposure to an atmosphere contaminated with
tritiated water vapour results in total absorption of the inhaled
activity through the lungs and absorption of about 50% of that
amount through the intact skin [I2]. Ingested tritiated water is
completely absorbed from the gastro-intestinal tract.
63. Absorbed tritiated water is rapidly distributed throughout the
body via the blood. Tritiated water in blood equilibrates with
extracellular fluid in about 12 minutes. However, in poorly
vascularized tissues, such as bone and fat, equilibrium with plasma
water may take days to weeks [N2, W2]. The biological half-life of
tritium in the body following intake of tritiated water has been
found to range from 2.4 to 18 days among 300 individuals [B3, W3].
The experience from observations of human cases of accidental
tritium exposures with intakes large enough to allow relatively
long-term monitoring shows that the excretion rate can be
represented as the sum of three exponentials with half-times of
residence of the order of 10 days, one month, and one year [L4, M5,
M6, S3]. The first component is believed to reflect the turnover
of body water while the second and the third components are likely
to represent the turnover of tritium incorporated into organic
compounds.
E. DOSIMETRY
1. Dose per unit intake
(a) Tritiated water
64. External irradiation from tritium does not need to be
considered as the range of the electrons emitted by decay (at most
6 µm in soft tissue) is shorter than the depth of the basal cells
in the epidermis. Following a chronic intake of 1 Bq 1-1 of
tritium (as HTO) in air, water and food the equilibrium dose rate
in active wet tissue (the totality of soft tissues with the
exclusion of fat) is 2.6 10-8 Gy a-1. Of that dose, 16% is
calculated to be due to tritium contained in organic pools of the
body. These results were derived by Bennett [B4] based on human
retention data.
65. When all the sources of intake (air, water and food) are
assumed to be contaminated at the same level, use can be made of
the specific activity model which consists in assuming that the
specific activity of tritium (activity concentration per unit mass
of hydrogen) in the body is the same as that in the intake. A
chronic intake of tritium at a concentration of 1 Bq per litre of
water would thus give rise to an absorbed dose averaged over the
whole body of
10-3
1 Bq 1H2O gH2O gH
----- x ----------- x 18 ---- x 0.1 -----
1H2O gH2O gH gbody
MeV s Gy gbody
x 5.7 10-3 ------ x 3.16 107 --- x 1.6 10-10 ----------
Bq s a MeV
= 2.6 10-8 Gy a-1 per Bq 1-1
This result is numerically equal to that of Bennett [B4]. The
doses in individual tissues depend on their hydrogen
concentrations. According to the values adopted for the Reference
Man of ICRP [I2], the hydrogen concentration per unit mass is the
same (10%) in total body and in total soft tissues and is, as a
first approximation, uniform in the soft tissues. Hydrogen content
is lowest in mineral bone (about 4%) and highest in adipose tissue
(12%). Since the range of the beta-particles emitted by tritium
decay is very small, it can be assumed that all the energy emitted
in a given tissue is absorbed in the same tissue. The effective
dose equivalent is therefore numerically equal to the absorbed dose
averaged over the whole body and is 2.6 10-8 Sv a-1 per Bq 1-1.
Assuming a rate of intake of 3 litres of water (in beverages and in
food) per day and a water vapour atmosphere concentration of 8 g m-3,
the effective dose equivalent per unit intake is found to be 2.2
10-11 Sv Bq-1 while the effective dose equivalent rate per unit
atmospheric concentration would be 2.1 10-9 Sv a-1 per Bq m-3.
(b) Tritiated hydrogen
66. The doses from inhalation of HT are much lower than those from
HTO for a given atmospheric concentration of tritium. The dose rate
to the lungs per unit concentration of HT in air is about 10-14 Gy
h-1 per Bq m-3 [I3], while the doses in tissues from the absorbed
gas are 60 to 150 times smaller [I3]. The corresponding effective
dose equivalent rate per unit concentration in air is therefore 1.1
10-11 Sv a-1 per Bq m-3.
2. Dose per unit release
(a) Natural tritium
67. Doses from natural tritium can be estimated from the few
tritium measurements in environmental materials that were carried
out before the contamination with fallout (or that had been
preserved from contamination). Activity concentrations of
continental surface waters were then found to be in the range from
0.2 to 0.9 Bq 1-1 [K1]. The production rate of natural tritium
being constant in time and relatively uniform on the global scale,
the concentrations in all the components of human intake (air,
water and food) of natural tritium are in steady-state equilibrium
with the concentrations in continental surface waters. Using the
specific activity approach, it is assumed that the specific
activity of natural tritium is the same in the continental surface
waters, in all the components of human intake and in the body. The
effective dose equivalent rate is thus found to range from
0.2 Bq 1-1 x 2.6 10-8 Sv a-1 per Bq 1-1 = 5.2 109 Sv a-1 to
0.9 Bq 1-1 x 2.6 10-8 Sv a-1 per Bq 1-1 = 2.3 10-8 Sv a-1, being
therefore of the order of 10-8 Sv a-1. The effective dose equivalent
commitment per unit release would then be
10-8 Sv a-1
--------------- ca. 1.4 10-25 Sv per Bq
7.2 1016 Bq a-1
Taking the world's population to be 4 109 people, the global
collective effective dose equivalent commitment per unit of
activity produced is about 5 10-16 man Sv per Bq.
(b) Nuclear explosions
68. The doses from fallout tritium can be estimated in the same
way as those from natural tritium. On the basis of the variation
with time of the tritium activity concentration in surface waters
[B5] and of the latitudinal distribution of the fallout deposition
[S1], UNSCEAR [U1] estimated the effective dose equivalent
commitments to the populations of the northern and southern
hemispheres to be 2 10-5 and 2 10-6 Sv respectively.
69. The effective dose equivalent commitment from fallout tritium
was also estimated indirectly, using the relationship obtained for
natural tritium between the production rate and the dose rate
W
Hc = gammao -
B
where Hc is the effective dose equivalent commitment (Sv) from
production of fallout tritium in a given hemisphere; gammao is the
effective dose equivalent rate from natural tritium (gammao = 10-8
Sv a-1); W is the activity of tritium released by nuclear
explosions (1.5 1020 Bq in the northern hemisphere and 0.2 1020 Bq
in the southern hemisphere); and B is the natural rate of
production (3.6 1016 Bq a-1 in each hemisphere). The effective
dose equivalent commitments thus derived are 4.2 10-5 Sv for the
population of the northern hemisphere and 5.6 10-6 Sv for the
population of the southern hemisphere. These results are higher
than the direct estimates by a factor of 2 to 3. The global
collective effective dose equivalent commitments per unit activity
released are estimated to be 9 10-16 and 4 10-16 man Sv Bq-1 using
the latter and the former method, respectively. UNSCEAR [U1] used
an intermediate value of 8 10-16 man Sv per Bq.
(c) Nuclear installations
70. While the production of natural and fallout tritium brings
about a relatively uniform contamination of the whole biosphere,
the releases from nuclear installations occur at discrete points on
the earth's surface giving a highly heterogenous spatial
distribution of concentrations.
71. UNSCEAR'S practice is to divide the collective doses into two
components: the local and regional collective doses, which are due
to the first passage of the plume, over distances of 100 to 1000 km
from the point of release, and the global collective doses, which
arise from the mixing of tritium in the whole biosphere. As the
doses per unit concentration of tritium in air are much higher for
HTO than for HT, tritiated water will be the only compound
considered in the estimate of the local and regional collective
doses.
(i) Local and regional collective dose
72. A distinction is made between airborne and liquid effluents.
Tritium present in airborne effluents can contribute to the local
and regional collective doses through inhalation, absorption
through skin and ingestion. As the contribution from the ingestion
pathway is quite variable from site to site owing to differences in
local hydrology and water usage, UNSCEAR [U1] has not taken this
pathway into account in its assessment of the local and regional
collective doses. Assuming an atmospheric dispersion factor of 5
10-7 s m-3 at 1 km from the release and a reduction in
concentration inversely proportional to the 1.5 power of the
distance expressed in kilometres, the local collective dose per
unit activity released can be assessed by integration over the
local area. Integrating from 1 to 100 km for a population density
of 100 km-2, UNSCEAR [U1] estimated the local collective dose from
airborne tritium per unit activity released to be about 5 10-17 man
Sv per Bq.
73. The collective dose commitment from the input of 3H to water
bodies, normalized per unit activity released, can be estimated
[U1], using the expression
c sigmak Nk Ik fk phi
S = -------------------
1 V(lambda + 1/tau)
where V is the volume of the receiving waters, tau is the turnover
time of receiving waters, lambda is the decay constant of 3H, Nk is
the number of individuals exposed by pathway k, Ik is the
individual consumption rate of pathway item k, fk is the
concentration factor for the consumed item in pathway k, and phi is
the collective dose per unit activity ingested collectively by the
exposed group.
1
74. The quantity V(lambda + 1/tau) is the infinite time integral
of the water concentration per unit of activity released, while the
quantity multiplied by fk is the infinite time integral of the
concentration in the consumed item k. For radionuclide inputs into
small volumes of water, the concentrations in water and in fish
will be high, but the population which can be served with drinking
water or by fish consumption will limited. For inputs into larger
volumes of water, the concentrations will be smaller, but the
populations involved will be correspondingly larger. It is
reasonable, therefore, to assume as a first approximation that the
quantities Nk/V are relatively constant, independent of V. The
values for these quantities as well as values for the other
parameters of the above expression have been extensively discussed
[U1].
75. A summary of the values used in the assessment, based on
UNSCEAR [U1], and the evaluation of the collective dose commitments
for a release of 1 Bq of 3H in liquid effluents are given in Table
II.4.
(ii) Global collective dose
76. For HTO releases, the global collective effective dose
equivalent commitment established for fallout tritium (8 10-16 man
Sv per Bq) can be applied without change. With respect to HT
releases, if it is assumed that the conversion to HTO takes place
on average 5 years after the discharges, the global collective
effective dose equivalent commitment is estimated to be
-0.693
8 10-16 e x 5/12.3 = 6 10-16 man Sv per Bq.
Table II.4 Collective dose factors for 3H in liquid effluents
---------------------------------------------------------------------------
Fresh water Salt water
---------------------------------------------------------------------------
Activity released, A 1 Bq 1 Bq
Turnover time of receiving water, 10 a 1.0 a
Sediment removal correction factor, s 1.0 1.0
Time integral of activity in water,
As
W = ------------ 6.36 Bq a 0.946 Bq a
1/tau+lambda
Water utilization, V/N 3 107 1/man 3 109 1/man
---------------------------------------------------------------------------
FRESHWATER PATHWAYS
1 Drinking water
Treatment removal factor, f1 1.0
Consumption, I1 438 1 a-1
Collective dose commitment
c NI
S1 = W f1 (--)1D 2 10-15 man Sv
V
Table II.4 (contd.)
---------------------------------------------------------------------------
Fresh water Salt water
---------------------------------------------------------------------------
2. Fish
Concentration factor, f2 1.0
Consumption, I2 1 kg a-1
Collective dose commitment
c NI
S2 = W f2 (--)2D 5 10-18 man Sv
V
SALT WATER PATHWAYS
3. Fish
Concentration factor, f3 1.0
Consumption, I3 6 kg a-1
Collective dose commitment
c NI
S3 = W f3 (--)3D 4 10-20 man Sv
V
4. Shellfish
Concentration factor, f4 1.0
Consumption, I4 1 kg a-1
Collective dose commitment
c NI
S4 = W f4 (--)4D 7 10-21 man Sv
V
---------------------------------------------------------------------------
(iii) Summary of collective dose commitments per unit activity released
77. Table II.5 summarizes the values obtained above for the
collective effective dose equivalent commitments per unit of 3H
activity released. With respect to the local and regional
component due to industrial releases, the largest collective
effective dose equivalent commitment per unit activity released is
obtained for a river discharge and the smallest for a sea discharge
while an intermediate value is found for the airborne discharge.
Table II.5 Summary of collective effective dose
equivalent commitments per unit tritium activity released
(man Sv per Bq)
---------------------------------------------------------
Origin Local and regional Global
component component
---------------------------------------------------------
Natural 5 10-16
Nuclear tests 8 10-16
Industry
Airborne discharge 5 10-17 (HTO) ) 8 10-16 (HTO)
River discharge 2 10-15 )
Sea charge 5 10-20 ) 6 10-16 (HT)
---------------------------------------------------------
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implications. Nature 263: 103-106 (1976).
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studies on the long-term evaluation of the biological half-life
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incorporated in genetic material. NCRP No. 63 (1979).
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the Environment. IAEA, Vienna, 1979.
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environmental tritium. p. 374-385 in Physical Behaviour of
Radioactive Contaminants in the Atmosphere. IAEA, Vienna,
1974.
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Medica Foundation, Amsterdam, 1968.
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noble gas fission products in the nuclear fuel cycle. I.
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Radiation 1977 report to the General Assembly, with annexes.
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unrestricted use of consumer products containing this
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life of tritium. Health Phys. 9: 911-914 (1963).
III. CARBON-14
A. INTRODUCTION
78. Carbon-14 has always been present on the earth. It is
produced by cosmic ray interactions in the atmosphere. This
nuclide is a pure beta-emitter, with a half-life of 5730 years, a
maximum energy of 185 keV and an average energy of 49.47 keV [N1].
79. Carbon is one of the elements that are essential to all forms
of life and is involved in most biological and geochemical
processes on the earth. Associated with the stable isotopes of
carbon (12C and about 1.1%13C), there is a very small amount of 14C
formed in the atmosphere and which has subsequently entered in the
carbon cycle. The specific activity of biological carbon, as
measured in wood samples grown in the nineteenth century, was 0.227
± 0.001 Bq per gram of carbon [T1], corresponding to an atmospheric
inventory of 1.4 1017 Bq. During the present century the specific
activity of 14C has decreased due to the diluting effect of
releases into the atmosphere of carbon dioxide from the combustion
of fossil fuels. This effect (the Suess effect) accounts for a
reduction of a few percent.
80. In addition to its natural production, carbon-14 is also
produced by the detonation of nuclear explosives and by the
operation of nuclear reactors. The assessment of the collective
dose commitments from the releases of man-made carbon-14 is
facilitated by knowledge of the production rate of natural
carbon-14.
B. SOURCES
1. Natural carbon-14
81. Carbon-14 is produced by the action of cosmic ray neutrons on
nitrogen atoms, both in the stratosphere and in the upper
troposphere. UNSCEAR [U3] has estimated the natural production
rate to be about 1015 Bq per year, a value which has been derived
from assessments of the natural 14C inventory. The production rate
has also been estimated directly from assessments of cosmic ray
neutrons and the values obtained by different authors range from 1
to 1.4 1015 Bq per year [U3]. Considering the uncertainties
involved in determining both the direct production rate and also
the total 14C inventory of the earth, the estimates are in
reasonable agreement.
2. Nuclear explosions
82. Carbon-14 is formed in nuclear explosions through the capture
of excess neutrons by atmospheric nitrogen. After large
atmospheric nuclear explosions, most of the 14C is transported into
the stratosphere, from where it equilibrates with the troposphere
with a half-time of 1 to 2 years [U3].
83. The inventory of 14C from nuclear explosions has been
estimated from measurements of excess specific activity in the
troposphere and in the surface ocean waters, and models
representing the exchange of 14C between the atmosphere, the
biosphere and the ocean. UNSCEAR [U3] has estimated that nuclear
explosions up to 1972 have injected into the atmosphere 2.1 1017 Bq,
while subsequent injections have increased this amount by about 1%.
84. For the past pattern of atmospheric nuclear explosions, the
production mentioned above corresponds to about 3.7 1014 Bq per
megaton. This value, however, is not representative of any given
nuclear explosion, because the production of 14C will depend on the
type of nuclear device exploded and also on whether the explosion
took place on the surface of the earth or high in the atmosphere.
A "surface" test will produce approximately 50% of the 14C that
would be produced by the same device in an "air" test, because
about one half of the escaping neutrons will be captured in the
soil or water rather than in the atmosphere.
3. Nuclear fuel cycle
85. Carbon-14 is produced in nuclear reactors and is released to
the environment at the reactor itself or at reprocessing plants
where spent fuel is reprocessed. Only recently has attention been
given to the production and release of this radionuclide at nuclear
fuel cycle installations.
(a) Nuclear reactors
86. The production of carbon-14 in nuclear power reactors is due
to several nuclear reactions in the fuel, core construction
materials and moderator. Figure III.I summarizes the relevant
reactions.
87. Production rates depend upon the neutron flux, the shape of
the respective neutron spectra and the resulting effective cross
sections, on the amount of the target elements present in different
reactor components and on the abundance of the target isotopes in
the target elements. The target elements are uranium, nitrogen,
oxygen, and also carbon in the case of graphite moderated reactors.
Nitrogen is present as an impurity in the fuel, as dissolved gas in
the coolant, as nitrogen compounds sometimes used for pH control in
the coolant, and as an impurity in structural materials. Oxygen is
present in water moderators and coolants, in CO2 coolants, and in
oxide fuels (e.g., UO2).
88. The place of origin of 14C within a nuclear reactor has a
strong influence on the discharge pathway. One can basically
distinguish between three locations of 14C generation, namely, 14C
in the fuel, 14C in structural materials of the core (and solid
moderator, if applicable) and 14C in the reactor coolant (and
liquid moderator, if applicable).
89. The 14C produced in liquid or gaseous coolants will be present
in different chemical compounds (CO2, CO, methane), depending on
the chemical environment. Under the influence of intensive
radiation fields several chemical reactions may occur, influencing
the chemical form of carbon-14. The compounds in the coolant are
released mainly together with off-gas and waste water from the
coolant purification and treatment system. Part of the carbon-14
also leaks from the primary coolant circuit into the plant
ventilation system and is released with ventilation air.
90. Significant reactions for the production of 14C in light water
reactors (LWR) are: (n) reactions with 17O present in the oxide
fuel and in the moderator; (n, p) reactions with 14N present in the
fuel as impurities; and ternary fissions. Ternary fission
production per unit electrical energy generated is practically
independent of reactor design, while the normalized production of
14C by the other reactions depends on the enrichment of the fuel,
the relative masses of the fuel and moderator, the concentration of
nitrogen impurities in the fuel and the temperature of the fuel and
moderator.
91. In boiling water reactors (BWR), the gaseous 14C is
transported with the steam until it arrives at the turbine
condenser. There the gases are continuously withdrawn over a
catalytic recombiner to burn the hydrogen and oxygen produced by
radiolysis of the primary water. Measurements have shown that one
half or more of the total 14C produced in the coolant will be
discharged in the form of CO2 together with the filtered gases from
the turbine condenser. There are other pathways of release of 14C,
mainly caused by leakage from the primary circuit into the reactor
building and the turbine hall. These releases are also mainly in
the form of CO2. A part of the 14C remains dissolved in the
primary water purification and treatment systems, causing smaller
sources of release, for example in the auxiliary building and
finally in the waste water system.
92. The primary circuit water of a pressurized water reactor (PWR)
contains hydrogen in excess to recombine the oxygen produced by
radiolysis. Under such reducing conditions compounds like methane
will be formed. Therefore, contrary to the BWR, a PWR will release
most of the 14C bound in hydrocarbons. The main release pathways
for gaseous compounds of 14C in PWRs are leakages of the primary
water circuit into the containment air and the degasification of
the primary water. The escaping or withdrawn gases may be stored
in decay tanks prior to release, and the gaseous 14C compounds can
be oxidized to CO2 or released through charcoal beds. Leakages may
also arise in the auxiliary building from the primary water
purification and treatment systems by way of degasing. Also, a part
of the 14C compounds stays dissolved in the water and is released
at the different steps of the waste water treatment.
93. The total environmental release of carbon-14 at the reactor,
expressed as a fraction of the production rate, is on the average
about 50% in BWRs and 30% in PWRs, but the value is quite variable,
as has been shown by several recent monitoring programmes [R1, L1].
UNSCEAR summarized the estimates of production in LWRs from several
authors, the values being in the range 0.5 to 1.9 109 Bq per
MW(e)a, and also derived an independent value of about 0.7 109 Bq
per MW(e)a [U3].
94. Carbon-14 is generated in heavy water reactors (HWR) through
reactions similar to those described for LWRs. Owing mainly to the
large moderator mass, the production rate of 14C in HWRs is
expected to be considerably larger than in LWRs [U3]. The
production rate in pressure vessel reactors is estimated to be 1.7
1010 Bq per MW(e)a, with 90% generated in the moderator. The
production of 14C in CANDU reactors is estimated to be 1.6 1010 Bq
per MW(e)a, 95% being produced in the moderator.
95. In gas-cooled graphite-moderated reactors (GCR), the major
source of 14C production is the graphite moderator, due to 13C(n,
gamma)14C reaction and also to the 14N(n, p)14C reaction based
on the incorporated nitrogen impurity. Production rates have been
estimated to be about 0.7 1010 Bq per MW(e)a in Magnox reactors and
1.1 1010 Bq per MW(e)a in advanced gas-cooled reactors (AGR) [U3].
Production of 14C in the carbon dioxide coolant, mainly from
activation of nitrogen impurities and from the 17O(n, alpha)14C
reaction, is a smaller source estimated to be about 108 Bq per
MW(e)a for Magnox reactors and 4 108 Bq per MW(e)a for AGRs.
96. Carbon-14 discharges from Magnox reactors and AGRs result from
coolant leakage and include 14C released to the coolant from
corrosion of the moderator. The fraction released at the reactor
is about 3% in Magnox reactors and about 6% in AGRs, of the total
production rate of 14C in these reactors [U3].
(b) Fuel reprocessing plants
97. While the 14C produced in the reactor coolant and moderator
has a potential for immediate release at the nuclear reactor, the
14C produced in the fuel will be released only later during nuclear
fuel reprocessing. Depending on reprocessing plant operation
characteristics the release may be continuous or discontinuous.
There are few measurements of 14C releases from reprocessing
installations [S1], but it seems reasonable to assume that almost
all the inventory of the fuel elements is released during the
chemical dissolution of the fuel. In the case of the Purex process
the 14C is released in the form of CO2.
(c) Summary
98. A very rough estimate can be made of the total production and
release of 14C from nuclear fuel cycle installations, based on the
average values given above. Installed nuclear capacity worldwide
in 1980 was 1.25 105 GW(e) [I2]. Assuming an average load factor
for reactor operation of 0.6, the energy produced was 7.5 104
GW(e)a. Global production and release of 14C from reactor sites
are thus estimated to be about 1.4 1014 Bq and 6 1013 Bq,
respectively. The estimated discharges by reactor types are given
in Table III.1. There are no estimates of production and release
from other reactor types representing 10% of the total installed
capacity. The difference between production and reactor discharge
estimates will largely represent the release from reprocessing
plants, to the extent that the fuel is eventually reprocessed.
Table III.1 Estimated global discharge of carbon-14
from nuclear power stations in 1980
------------------------------------------------------------------------
Reactor Reactor Capacity Production rate Release Estimated
type number [MW(e)] [Bq per MW(e)] fraction carbon-14
(%) discharge (Bq)
------------------------------------------------------------------------
PWR 96 64239 7 108 30 8 1012
BWR 62 35170 7 108 50 7 1012
HWR 14 5963 1.6 1010 70 4 1013
GCR 36 7086 9 109 5 2 1012
Other 33 12527 - - -
------------------------------------------------------------------------
Total 241 124985 6 1013
------------------------------------------------------------------------
C. BEHAVIOUR IN THE ENVIRONMENT
99. Carbon-14 is present in atmospheric carbon dioxide, in the
biosphere, and in the bicarbonates dissolved in the ocean. The
specific activity of natural 14C in the terrestrial biosphere, as
measured in wood grown in the nineteenth century, was 0.227 ± 0.001
Bq per gram of carbon. The Suess effect, accounting for a few
percent decrease of specific activity at present, could reach a
figure of the order of 20% in the year 2000 [U2], but is of little
importance in the long range, when fossil fuel resources are
exhausted.
100. Leaving aside the Suess effect, it has been suggested,
however, that the present-day inventory does not correspond to the
equilibrium value, but is increasing. In fact, measurements of
wood samples of known age show that only cyclic variations of
atmospheric 14C, amounting to a few percent, have occurred in the
past 6000 years [U2]. Two types of variations have been
recognized: one, with a time scale of the order of 100 years, has
been explained by the solar wind modulation of the cosmic-ray flux
density; the other, with a time constant of more than 1000 years,
may largely be due to a variation of the geomagnetic shielding of
the earth.
101. Contrary to the case of natural carbon-14, the levels of man-
made carbon-14 are not at steady state in the different
compartments of the environment. Due to the very long mean life of
carbon-14, continuing practices are not expected to last long
enough to allow the environmental levels to reach the steady state.
The predictions of the time-evolution of 14C levels in the
atmosphere, biosphere and ocean after a release into the
environment require, therefore, the use of compartment models.
102. Many models describing the dispersion of released 14C, and
the subsequent exchange between the different compartments involved
in the carbon cycle, have been proposed [C1, P1, N2, Y1, N3].
UNSCEAR [U3] also developed a dynamic model for the assessment of
doses from 14C released by nuclear explosions. This model includes
compartments for the atmosphere and short-term biosphere, the
terrestrial biosphere, the surface ocean and the deep ocean, and
represents the thermocline layer in the ocean as a thick diffusion
barrier.
D. TRANSFER TO MAN
103. Carbon-14 released to the environment enters the carbon
cycle, giving rise eventually to increased levels in man. From
measurements of fallout carbon-14, it was noted that the specific
activity in human tissue comes into equilibrium with that of
atmospheric CO2 with a delay time of about 1.4 years [N5].
104. Intake of carbon by man is primarily from diet. Ingestion
intake is of the order of 300 g d-1 with nearly complete
absorption, whereas inhalation intake is about 3 g d-1 with only 1%
retained in the body [U3]. The total carbon content of the body is
1.6 104 [I1]. The quotient of this with the intake rate gives an
estimated mean residence time of carbon in the human body of 53
days.
105. Man comes, therefore, into fairly rapid equilibrium with
carbon-14 in his immediate environment. It is generally sufficient
in carbon-14 dose calculations to adopt a steady-state model which
assumes that the specific activity of carbon in tissues is in
equilibrium with that in air and in the diet.
E. DOSIMETRY
1. Dose per unit intake
106. An intake of carbon-14 at a specific concentration of 0.23 Bq
per gram of carbon, corresponding to the present value for natural
carbon-14, gives rise to the following absorbed dose rate averaged
over the whole body
Bq Gy g 0.049 MeV/Bq s
0.23 -- 1.6 10-10 ---- ----------------
gc MeV 7 104g
3.15 107 s/a 1.6 104 gc = 13 µGy a-1
The dose rates in individual tissues depend on their carbon
concentrations. The carbon content per unit mass averages 23% for
the whole body, but ranges from 9% in gonads and 10% in lungs to
41% in red bone marrow and 67% in adipose tissue [I1]. The annual
absorbed doses are 5 µGy in gonads, 6 µGy in lungs, 20 µGy in bone-
lining cells and 22 µGy in red bone marrow [U3]. The tissue-
weighted annual effective dose equivalent from natural carbon-14 is
12 µSv.
107. This dose is due almost entirely to ingestion intake of
carbon-14. If the carbon intake rate is 300 g d-1 at the specific
activity of 0.23 Bq g-1, the intake rate of 14C is 69 Bq d-1. The
effective dose equivalent per unit ingestion intake of 14C is
12 10-6 Sv/a 1 a
------------ ------- = 5.2 10-10 Gy Bq-1
69 Bq/d 365 d
The dose factor for inhalation intake is less by a factor of 10-2,
since absorption into the body is that much less by this pathway.
2. Dose per unit release
108. The doses given above for natural carbon-14 correspond to
the annual global production of 1015 Bq. This production is
essentially constant in time and uniform over the world. Therefore,
equilibrium has become established. The effective dose equivalent
commitment per unit release is
12 10-6 Sv/a
------------ = 1.2 10-20 Sv Bq-1
1015 Bq/a
The collective dose equivalent rate from natural carbon-14 to the
world population of 4 109 people is 4.8 104 man Sv a-1.
109. The assessment of the dose commitment from a given release of
man-made carbon-14 is carried out by direct analogy with natural
carbon-14. Once the released carbon-14 enters the global carbon
cycle, the effective dose equivalent commitment per unit release is
1.2 10-20 Sv Bq-1.
110. It is difficult to assess with precision the collective dose
commitment per unit release of carbon-14, because the projected
increase in the world population is very uncertain. Assuming that
it will attain an equilibrium value of 1010 persons, in a time
short compared with the mean effective life of 14C [U3], the
collective effective dose equivalent commitment per unit released
is approximately 1.2 10-10 man Sv per Bq.
111. In order to calculate the complete collective dose commitment
[U3] required for assessments of maximum future mean annual doses
from a continuing but finite practice releasing 14C, it is
necessary to use dynamic models predicting the time evolution of
environmental levels. Assuming that power production by nuclear
fission will last for a few hundred years (for example, 500 years),
the incomplete collective dose commitment can be calculated using
the model with diffusion barrier already mentioned. The incomplete
collective dose commitment, integrated over 500 years, is about 2.3
10-11 man Sv per Bq released. This value is somewhat higher than a
value of about 1.4 10-11 man Sv per Bq which can be deduced from a
recent assessment of the environmental significance of 14C [N3],
but in view of the uncertainties involved, the difference is
probably insignificant.
112. The contribution of local and regional exposures to the
collective dose commitment is very small, of the order of a
percent, and can be neglected [N3]. The assessment of individual
doses at some selected locations, however, is necessary for
radiation protection purposes. Its calculations can be carried out
by the use of specific activity methods. One simple model assumes
that the specific activity of 14C in air is equal to that in the
body. A more sophisticated calculation assumes that the specific
activity in the vegetation at the location of interest is equal to
that of air. The dose can then be assessed from knowledge of the
relative proportion of contaminated food in the diet. Both methods
require the use of micrometeorological models to assess
quantitatively the dispersion of 14C from the release point to the
locations of interest. Some publications [U4, N4, C2], present
improvements to the classical formulations describing the local
atmospheric dispersion.
F. REFERENCES
C1 Craig, H. The natural distribution of radiocarbon and the
exchange time of carbon dioxide between atmosphere and sea.
Tellus 9: 1-17 (1957).
C2 Clarke, R. A model for short and medium range dispersion of
radionuclides released to the atmosphere. A first report of a
working group on atmospheric dispersion. NRPB-R91 (1979).
I1 International Commission on Radiological Protection. Report of
the task group on reference man. International Commission on
Radiological Protection publication 23 (1975).
I2 International Atomic Energy Agency. Power reactors in member
states. IAEA, Vienna, 1980.
L1 Luykx, F. and G. Fraser. Radioactive effluents from nuclear
power stations and nuclear fuel reprocessing plants in the
European community: discharge data 1962-76. Radiological
aspects. Commission of the European Communities. V/4604/78-EN
(1978).
N1 National Council on Radiation Protection and Measurements. A
handbook of radioactivity measurements procedures. National
Council on Radiation Protection report No. 58 (1978).
N2 Nydal, R. Further investigation on the transfer of radiocarbon
in nature. J. Geophys. Res. 73: 3617-3635 (1968).
N3 Nuclear Energy Agency, OECD. Radiological significance and
management of H-3, C-14, Kr-85 and I-129 arising from the
nuclear fuel cycle. Report by an NEA group of experts.
OECD/NEA (1980).
N4 NRPB and CEA. Methodology for evaluation of radiological
consequences of radioactive effluents released in normal
operations. Commission of European Communities. V/3865/79
(1979).
N5 Nydal, R., K. Lovseth and O. Syrstad. Bomb 14-C in the human
population. Nature 232: 418-421 (1971).
P1 Plesset, M. and A. Latter. Transient effects in the
distribution of carbon-14 in nature. Proceeding of the
National Academy of Sciences 46: 232-241 (1960).
R1 Riedel, H. and P. Gesewsky. Zweiter Bericht über Messungen zur
Emission von Kohlenstoff-14 mit der Abluft aus Kernkraftwerken
mit Leichtwasserreaktor in der Bundesrepublik Deutschland.
Bundesgesundheitsamt report STH-13/77 (1978).
S1 Schuettelkopf, H. and G. Herrman. 14-CO2 Emissionen aus wer
Wiederaufarbeitungsanlage Karlsruhe. p. 189 in Report for the
Commission of the European Communities. V/2266/78-D (1978).
T1 Telegadas, K. The seasonal atmospheric distribution and
inventories of excess carbon-14 from March 1955 to July 1969.
HASL-243 (1971).
U2 United Nations. Report of the United Nations Scientific
Committee on the Effects of Atomic Radiaton to the General
Assembly, with annexes. Volume I: Levels, Volume II: Effects.
United Nations sales publication No. E.72.IX.17 and 18. New
York, 1972.
U3 United Nations. Sources and Effects of Ionizing Radiation.
United Nations Scientific Committee on the Effects of Atomic
Radiation 1977 report to the General Assembly, with annexes.
United Nations sales publication No. E.77.IX.I. New York,
1977.
U4 U.S. Nuclear Regulatory Commission. Regulatory Guide 1.111
(1977).
Y1 Young, J. and A. Fairhall. Radiocarbon from nuclear weapons
test. J. Geophys. Res. 73: 1185-1200 (1968).
IV. KRYPTON-85
A. INTRODUCTION
113. Krypton is element number 36 in the periodic table. It
belongs to the group of inert gases together with helium, neon,
argon, xenon and radon. It occurs naturally in the atmosphere to
an estimated extent of 1 to 2 10-6 by volume.
114. The naturally occurring stable krypton isotopes and their
atom percentage abundances are: 78Kr (0.35%), 80Kr (2.27%), 82Kr
(11.56%), 83Kr (11.55%), 84Kr (56.9%), 86Kr (17.37%) [N1]. The
radioactive isotopes of krypton include mass numbers of 74-77, 79,
79m, 81, 81m, 85, 85m, 87-95 and 97. Some of these occur naturally
in low trace amounts as a result of cosmic ray induced reactions
with stable krypton isotopes and by spontaneous fission of natural
uranium.
115. The radioactive isotope 85Kr is produced in nuclear fission.
With a half-life of 10.7 years, it can become widely dispersed in
the atmosphere following release. The average fission yields
differ by about a factor of 2 for 239Pu and 235U, being about 0.6
and 1.3 atoms per 100 fissions, respectively (Table IV.1).
Table IV.1 Fission yields of
krypton-85 [C2]
-----------------------------
Fission yield (%)
Nuclide thermal fast
-----------------------------
232Th 4.14
233U 2.28 2.12
235U 1.32 1.33
238U 0.74
239Pu 0.558 0.62
-----------------------------
116. The decay scheme of 85Kr is presented in Figure IV.I. Two
beta particles and a single gamma photon are emitted, along with
several low-energy conversion electrons and x-rays.
117. Being chemically inert, krypton and other inert gases are not
usually involved in biological processes. They are, however,
dissolved in body fluids and tissues following inhalation. Krypton
is characterized by low blood solubility, high lipid solubility and
rapid diffusion in tissue [K1]. The biological involvement of
inert gases has been noted in numerous studies [K1].
B. SOURCES
118. Krypton-85 is produced by cosmic ray interactions in the
atmosphere, in nuclear power reactors, and nuclear explosions. The
main release source is the dissolution step in the reprocessing of
nuclear fuel.
119. Concentrations of 85Kr in the atmosphere increased sharply
after 1955 due to the production and testing of nuclear weapons and
the development of the nuclear power industry. More recently the
input rates of 85Kr into air have decreased [H2]. There have been
reductions in plutonium production for military purposes and in
nuclear fuel reprocessing.
120. A review of 85Kr measurement data for 1950-77 has been
prepared by Rozanski [R1]. The most recent data indicate that
concentrations in air have stablized at about 0.6 Bq/m3 in the
northern hemisphere and 0.4 Bq/m3 in the southern hemisphere [R1].
The major sources are in the northern hemisphere, accounting for
the higher levels in that hemisphere.
1. Natural krypton-85
121. Krypton-85 is present in small amounts in the environment as
a result of spontaneous fission of natural uranium and interactions
of cosmic ray neutrons with atmospheric 84Kr. The steady state
environmental inventories of 85Kr from these sources have been
calculated: 7.4 1010 Bq in the upper 3 m of the total land and
water surface due to spontaneous fission of natural uranium, 3.7
1011 Bq in the atmosphere from cosmic ray production and 3.7 105 Bq
in the oceans from the atmospheric source [D1]. These estimates,
in comparison with the estimates of man-made sources of 85Kr to
follow, are negligible in contributing to the world's total 85Kr
inventory.
2. Nuclear explosions
122. Since 85Kr is produced during fission, it has been generated
by nuclear weapon tests. The total amount of 85Kr produced in
nuclear testing can be calculated from the ratio of 85Kr/90Sr
fission yield of 0.08, giving an activity ratio of 0.22 [C2].
Measurements of 90Sr activity have been reported and discussed in
the reports of UNSCEAR [U1-U7]. There have been 6 1017 Bq of 90Sr
produced in weapon testing through 1976 [U7], corresponding to
about 1.3 1017 Bq of 85Kr.
123. Another source of 85Kr associated with nuclear weapons is in
the production of plutonium in military reactors. The amount of
85Kr released from this source is estimated to be two times higher
than that from the weapon tests [D1]. Naval propulsion reactors
also contribute to the 85Kr inventory with an annual production in
the region of 1.1 to 1.9 1016 Bq [B1]. Including all sources, the
total amount of 85Kr produced in operations for military purposes
is still rather small in comparison to the prospective generation
of 85Kr by the nuclear power industry.
3. Nuclear fuel cycle
124. Krypton-85 is produced by fission in the fuel of nuclear
reactors and in very low trace amounts in the moderator or coolant,
due to contamination with fissile material. The rates of 85Kr
production are related to the type of fuel and degree of burn-up.
Production and emission rates may be conveniently normalized to
unit electrical energy generated (for power reactors) or to the
electrical energy generated by the reactors serviced (for fuel
reprocessing plants).
125. The amounts of 85Kr produced vary according to reactor type.
For thermal reactors, the range of estimated production is about
1.1 to 1.5 1013 Bq/MW(e)a. For FBRs the values are about 25%
smaller [E1, M1], for HTGRs 50% higher [B3]. A production rate of
1.4 1013 Bq/MW(e)a has been correlated with some measurements from
reprocessing plants [U7] and this value can be taken for general
evaluations.
126. An estimate of 85Kr annual generation from reactor operation
can be obtained from the installed capacity of nuclear reactors of
1.25 105 MW(e) worldwide in 1980 [I1], with the assumptions of 60%
utilization and average 85Kr generation rate of 1.4 1013 Bq/MW(e)a:
1.25 105 MW(e)a x 0.6 x 1.4 1013 Bq/MW(e)a = 1 1018 Bq/a
The actual release rate is less, since delays occur before
reprocessing and not all fuel is reprocessed.
127. Reported releases of 85Kr and other fission noble gases were
listed in the 1977 report of UNSCEAR [U7]. There are large
differences in the release values of the various plants. Although
the relevant data are not very extensive, there are indications of
improved retention of 85Kr at reactors in recent years due to the
installation of additional hold-up tanks or adsorption columns.
128. In the reprocessing plant the spent fuel elements are
dismantled and the nuclear material dissolved. Procedures to
separate 85Kr from gaseous effluents and to provide long-term
retention are under study, but current practice is to allow
controlled release to the atmosphere.
C. BEHAVIOUR IN THE ENVIRONMENT
129. Krypton-85 discharged to the environment disperses in the
atmosphere and largely remains there until decay. It can become
washed out by rain and diffuse into surface layers of soil and
oceans, but these processes account for very little transfer of
85Kr from the atmosphere.
1. Dispersion in the atmosphere
130. Materials released to the atmosphere are transported downwind
and dispersed according to atmospheric mixing processes. The
estimation of this dispersion is commonly approached by using a
diffusion-transport equation. Several models have been developed
for this purpose, using a variety of boundary conditions and
simplifying assumptions. Most of them are based on the Gaussian
plume diffusion model [S1, I2], which has been shown to be adequate
in many practical situations. The krypton concentrations in air at
various distances for a release from a 30 m high stack are shown in
Table IV.2 [C5].
Table IV.2 Krypton-85
concentration in air for a
release of 1 Bq/s (stack height
30 m, Pasquill category D)
[C5]
-------------------------------
Distance Concentration
(km) (Bq/m3)
-------------------------------
1 4.8 10-7
10 1.3 10-8
100 4.4 10-10
1000 3.2 10-11
-------------------------------
131. For estimation of dispersion at greater distances, some
shortcomings in the Gaussian model are evident in the assumptions
that the meteorologic conditions and the direction of the wind
remain constant throughout the transit of the plume. To overcome
these difficulties, long-range models have been developed [A1, D3,
M2], which follow the trajectories of masses of air passing over
the release point and take into account the changing meteorologic
conditions with time. A survey of several diffusion models and of
their applications is given in [C5].
132. The global circulation of 85Kr can be approximated by a
simple compartment model, consisting of single compartments
representing the atmosphere in the northern and in the southern
hemispheres. Following a single release, equilibrium
concentrations in the atmosphere are achieved after about two
years. Further decrease in concentrations is due to radioactive
decay. In applying this model, Kelly et al. [K3] determined that
the integral concentration in air would be 5.3 10-18 Bq a m-3 per
Bq released. The atmospheric mass was assumed to be 3.8 1021 g,
equivalent to 3.1 1018 m3 at STP.
133. The dispersion calculations of Machta et al. [M2] are based
on detailed meteorological considerations and allow population-
weighted exposures to be determined. Table IV.3 lists the average
surface air concentrations of 85Kr in latitude bands following
release of 1 Bq in the 30-50° N latitude band. Uniform
concentrations are achieved after two years, after which the
integral concentration until complete decay is
10.73 a
22 10-20 Bq/m3 ------- = 3.4 10-18 Bq a/m3
1n 2
Adding the contributions from the first two years gives
3.9 10-18 Bq a/m3 for the population weighted integral
concentration of 85Kr in air from a release of 1 Bq.
2. Removal from the atmosphere
134. There is very little removal of 85Kr from the atmosphere,
except by radio-active decay. The low solubility of krypton in
water limits the accumulation of 85Kr in rainwater. Adsorption of
85Kr on particulate matter in air and subsequent deposition of the
particles provides a removal means of very low efficiency [N1].
135. The transfer of 85Kr to soil can occur by diffusion
processes; however, estimates of this transfer can account for
only about 0.05% of the total krypton in the atmosphere [N1].
Therefore, soil in general is not an important removal sink for
85Kr.
136. The efficiency of the oceans as a sink for 85Kr can be
determined from the natural krypton content of the atmosphere and
of the mixed layer of the ocean. From estimates of the krypton
concentration in air, the atmospheric volume and the density
krypton (STP), a total mass of about 1.64 1016 g of krypton in the
atmosphere is calculated [N1]. Assuming that the mixed layer of
the ocean extends to 100 m depth and an area of 3.6 1018 cm2, and
using the measured average krypton concentration in this layer of
seawater of 5 10-8 by volume [B4], a total mass of 6.7 1012 g of
krypton in the mixed layer of the ocean is obtained. This
corresponds approximately to 0.04% of the atmospheric mass of
krypton.
Table IV.3 Average surface air concentration of krypton-85
(1 Bq emitted uniformly over one year in 30-50° N latitude band)
[M2]
---------------------------------------------------------------
Krypton-85 concentration Population
Latitude band (10-20 Bq/m3) distribution %
Year 1 Year 2 Year 3
---------------------------------------------------------------
70 - 90° N 23 32 22 -
50 - 70° N 25 31 22 12.6
30 - 50° N 23 30 22 32.0
10 - 30° N 19 27 22 39.0
10° N - 10° S 11 22 22 11.5
10 - 30° S 6.3 22 22 3.4
30 - 50° S 5.1 20 22 1.5
50 - 70° S 4.3 19 22 0.05
70 - 90° S 3.8 19 22 -
Population weighted
integral
concentration
(10-20Bq a/m3) 19.5 27.6 22.0
---------------------------------------------------------------
137. An estimate of the total mass of krypton in the oceans as a
whole is obtained using an average concentration by volume of
krypton in the oceans of 9 10-8 [B4], a total ocean volume of 1.4
1024 cm3, and a krypton density of 3.73 10-3 g/cm3 at STP. This
calculation results in a total ocean inventory of about 4.7 1014 g
of krypton, or approximately 3% of the total atmospheric krypton
[N1]. These figures clearly indicate that the oceans can serve
only as a minor sink for 85Kr discharged into the atmosphere.
D. TRANSFER TO MAN
138. Following release to the atmosphere 85Kr becomes widely
dispersed. Exposure of man occurs by external irradiation from the
passing cloud or the dispersed gas and by internal irradiation
following inhalation of 85Kr and absorption in tissues.
139. After intake, 85Kr is distributed in the body by blood and
lymph fluids and is absorbed in the various tissues. A person
immersed in an atmosphere of 85Kr at low concentration would rather
quickly come into equilibrium with it. The concentrations in body
tissues are determined by multiplying the concentration in air by a
partitioning factor, called the Ostwald's coefficient. The
relevant values reflect the rate at which tissues are perfused with
blood, the solubility of the gas in the several tissues and the
velocity of diffusion of krypton across anatomical boundaries. The
concentration of 85Kr in the body is not uniform, the concentration
in the adipose tissue being nearly 50 times higher than that in
other parts of the body.
140. As a first approximation, one may only account for a
difference in the absorption behaviour of krypton in fat and non-
fat tissues, with values of the Ostwald coefficient of 0.45 for fat
and 0.07 for non-fat tissue [N1]. Other more elaborate models use
weight-related coefficients, where the density of the absorbing
tissue is taken into account [S2].
141. The total body retention of 85Kr has been subjected to
exponential analysis. Several clearance rates have been
recognized. Recent work has suggested a model for krypton in the
body consisting of five compartments [C6]. The fastest component
probably represents the clearance from circulating blood,
particularly blood plasma (T´ = 21.5 ± 5.7 s). The second
component (4.74 ± 2 min) appears to be representative of
haemoglobin clearance. The next slower component (19.8 ± 6.6 min)
is most likely related to clearance of krypton from muscle. The
two components with the slowest clearance rates can be related to
body fat compartments. A half-time of about 2.4 h is attributed to
a fat compartment not located in adipose tissue. The retention
half-time of krypton in adipose tissue is the slowest component and
is correlated significantly with the total body fat content. The
relationship is T´(h) = 0.17 (percentage fat) + 0.75 [C6].
E. DOSIMETRY
142. Krypton-85 released to the environment causes a radiation
dose to man through external irradiation from amounts in air and
through internal irradiation from amounts within the body. Tissues
are irradiated both from the activity in the organ itself and from
the activity present in the surrounding organs.
1. Dose per unit exposure
143. The equilibrium absorbed dose rates to body organs per unit
concentration of krypton-85 in air are summarized in Table IV.4
[N1]. For comparison, the recently published values of the ICRP
are also listed [I3]. The ICRP values represent minor adjustments,
except for the lungs, for which the beta dose due to 85Kr in the
airways of the lungs has been disregarded.
144. The dose equivalent rates in various organs are listed in
Table IV.5. These are the ICRP values [I3]. The quality factor
for 85Kr radiation is one. Therefore the dose equivalent rates are
numerically equal to the absorbed dose values. When combined with
the tissue weighting factors suggested by the ICRP to account for
varying incidence of h