-d

United Nations Scientific Committee
on the Effects of Atomic Radiation

1977 report to the General Assembly, with annexes

NOTE
The report of the Committee without its annexes appears as Official Records of the General
Assembly, Thirty-second Session, Supplement No. 40 (A/32/40).
In the text of each annex, Arabic numbers in parentheses are references listed at the end.
The designations employed and the presentation of material in this publication do not imply theexpression of any opinion whatsoever on the part of the Secretariat of the United Nations concerning
the legal status of any country, territory, city or area, or of its authorities, or concerning thedelimitation of its frontiers or boundaries.

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Produced by UNIDO, Vienna

ANNEXDRadioactive contamination due to nuclear power production

CONTENTSParagraphs

Paragraphs

INTRODUCTION . . . . . . . . . . . . . . . . . . .

I. MINING AND MILLING
A.
B.

.......... .

Effiuents . . . . . . . . . . . , . . . . . .
Local and regional collective dose
commitments . . . . . . . . . . . . . . .

II. URANIUM FUEL FABRICATION . . . . . .
A.
B.

Effluents . . . . . . . . . . . . . . . . . .
Local collective dose commitment .. .

III. REACTOR OPERATION
A.

B.

.......... .

Effluents • . . . . . . . . . . . . . . . . .
1.
Fission noble gases . . . . . . . . .
2.
Activation gases . . . . . . . . . .
3. Tritium . . . . . . . . . . . . . . .
4.
Carbon-14 . . . . . . . . . . . . . .
5. Iodine . . . . . . . . . . . . . . . .
6.
Particulates in airborne effluents .
7.. Radionuclides in liquid effluents .
.
. .
.
.
Local and regional collective dose
commitment . . . . . . . . . . . . . . .
1.
Fission noble gases . . . . . . . . .
2.
Activation gases . . . . . . . . . .
3.
Tritium . . . . . . . . . . . . . . .
4.
Carbon-14 . . . . . . . . . . . . . .
5.
Iodine . . . . . . . . . . . . . . . .
6.
Particulates in airborne effluents .
7.
Radionuclides in liquid effluents .

.

10-37
13-20
21-37
38-46
41-44
45-46
47-138
49-90
51-60
61-64
65-70
71-81
82-84
85
86-90

~

IV. FUEL REPROCESSING . . . . . . . . . . . .
A.

1-9

Effluents . . . . . . . . . . . . . . . . . .
1.
Krypton-SS . . . . . . . . . . . . .
2.
Tritium . . . . . . . . . . . . . . .
3. Carbon-14 . . . . . . . . . . . . . .
4.
Iodine . . . . . . . . . . . . . . . .
5. Caesium, ruthenium, strontium
6.
Plutonium, americium, curium

91-13892-99
100-102103-105
106-107108-118
119-122123-138
139-108
141-156
144
145-146147
148-149
150-151
152-156

lfltroductiofl
The use of nuclear reactors to generate electricity
has continued to increase since the Committee's firstatte.!l:lpt to assess the radioactive releases from thenuclear fuel cycle and the resulting dose commitment
1.

B.

Local and regional collective dose
commitment . . . . . . . . . . . . . . .
1.
Krypton-85 . . . . . . . . . . . . .
2.
Tritium . . . . . . . . . . . . . . .
3. Carbon-14 . . . . . . . . . . . . . .
4.
Iodine . . . . . . . . . . . . . . . .
5.
Caesium, ruthenium, strontium
6.
Plutonium, americium, curium ..

V. RADIOACTIVE WASTE STORAGE ANDDISPOSAL . . . . • • . . . . . . . . . . . . .
A.
B.

Low- and intermediate-level waste .. .
High-level waste . . . . . . . . . . . . . .

VL COLLECTIVE DOSE COMMITMENT FROM
GL0Br.LLY DISPERSED RADIONUCLIDES . . . . . . . . . . . . . . . . . . . . . . .
A.
B.
C.
D.

Tritium . . . . . . . . . . . . . . . . . .
K.rypton-85 . . . . . . . . . . . . . . . .
Carbon-14 . . . . . . . . . . . . . . . . .
lodine-129 . . . . . . . . . . . . . . . . .

VII. MISCELLANEOUS CONTRIBUTIONS .. .

157-168
158
159
160
161-163
164-166
167-168

169-183
170-178
179-183

184-202
186-190191-195
196-199200-202

Transportation . . . • . . . · . . . . . . .
Nuclear research laboratories . . . . . .

203-212204-207
208-212

VI1L SUMMARY OF COLLECTIVE DOSE COMMITMENTS TO THE PUBLIC FROM NUCLEAR POWER PRODUCTION .•.•..•

213-215

Appendix. Collective dose commitment from release of radioactive materials due to accidents

216-222

A.
B.

Page

References • . . . . • . • . • • . . . . . . . • . . . •215(140). The total installed nuclear electric generating
capacity in the world in 1976 was 79.8 GW from 188
reactors operating in 19 countries (66). The predicted
capacity for 2000 is 2000 GW, about half the value
predicted in the 1972 report of the Committee.
165

2.
The nuclear fuel cycle consists of the process uf
mining and milling of uranium, conversion to fuelmaterial, usually including enrichment in the isotope
2 3 5 U, fabrication of fuel elements, utilization of thefuel in nuclear reactors, reprocessing of spent fuel andrecycled utilization .of recovered fissile materials,
transportation of material between fuel-cycle installations, and disposal of radioactive wastes.
3.
Almost all the radioactivity associated with thefuel cycle is present in stored, spent fuel elements and inwell contained fractions separated from the fuel during
the reprocessing operations. However, at each step of thefuel cycle, releases of small quantities of radioactive
material into the environment may occur. Most of theradionuclides released into the environment are only of
local or regional concern, because their half-lives are shortcompared to the time required for dispersion to greater
distances. Some radionuclides, on the other hand, having
longer half-lives or being more rapidly dispersed, canbecome globally distributed.
4.
The Committee is interested in assessing the dose
commitments due to releases of radioactive material
from all fuel-cycle operations. As the scale of these
operations is related to the nuclear installed capacity
they serve, it seems reasonable to express suchassessments in terms of the dose commitment ( orcollective dose commitment) per unit of electricalenergy generated (MW(e) y).
5.
The total collective dose commitment to the world
population may be assessed by estimating the contribution from four components: the occupationally
exposed group, the local population, the regional orintermediate population, and the world population. This
Annex deals only with the collective doses incurred
through exposure of the public, while the occupational
contributions are analysed in Annex E. The local andregional contributions to the collective dose commitment are discussed separately for each component of thefuel cycle; the global contributions made by nuclides ofworld-wide distribution are discussed for the cycle as awhole. The handling of solid wastes generated in the
nuclear fuel cycle is considered in chapter V of thisAnnex.
6.
Doses and collective dose commitments to local
and regional populations can be estimated satisfactorily
given adequate knowledge of the effluents, the transfer
factors in the routes by which man is exposed and theresults of properly designed environmental monitoring
programmes. However, the data based on measurements
available to the Committee are insufficient for the
purpose of assessing all collective dose commitments,
especially global collective dose commitments, from allfuel cycle operations, and ·some generalizations concerning· effluent release rates and environmental transfer
factors have had to be mad~. Monitoring programmes areusually confined to local areas and are frequently
planned to verify compliance with applicable regulations. More detailed measurements are seldom carriedout and in many situations are not even feasible. In somecases, however, monitoring can contribute to improvements in the assessment of model parameters,monitoring and modelling being in these cases
complementary procedures.
166

7.
The Committee has reviewed information relating
to both approaches. Since the 1972 report, a number ofnational and international symposia have dealt withvarious aspects of environmental dispersion mechanisms
and of population exposures (46, 63, 67. 69. 149, 150).
A wide range of other publications have contributed toincreased knowledge of environmental transport andimproved estimates of radiation doses. In addition,
environmental and safety studies for specific fuel-cycle
installations and a few comprehensive reviews of theentire fuel cycle have been prepared (148, 151, 153,
156, 157, 158, 162).
8.
The discussion of the effluents of nuclear power
installations and the collective dose estimates are, as far
as possible, based on operating experience. Included aredata on releases in all periods of operation, including
those associated with abnormal operation or withperiods in which there is no electricity generated or fuelreprocessed. The averages, therefore, do not represent
typical release rates during normal operation. but insteadreflect the overall- nuclear power experience, normalized
to the total amount of generated electricity. Where
operating experience is insufficiently documented, some
generalizations regarding effluent release rates and theenvironmental transfer parameters are made in order tomake a complete assessment of the collective doses from
nuclear fuel-cycle operations. It must be stressed thatsuch generalizations will always lead to less accurate
estimates of the dose commitments than assessments
based directly on studies of particular releases and theparticular environment into which the releases are made.
Moreover, in the future, improved practices may give
results that are considerably different from those derived
from current and past experience. Therefore, any
application of the assessments given here to more
specific situations or any extrapolation to the future
must recognize these inherent limitations.
9.
In addition to the small releases of radioactive
material during normal operation the possibility existsthat additional amounts of radioactive materials may be
accidentally discharged. A number of studies haveconsidered the potential composition and amounts ofaccidental releases (61). The assessment of population
exposure due to long-term nuclear power programmes
requires the estimation of fre expected cont!ibution ofsuch events. Recent evaluations of their probability
make it possible to assess the expectation of collective
dose from the release of radioactive material due toaccidents in reactors, as discussed in the appendix.

I. MINING AND MILLING
10. Uranium is .obtained from ore mined in several
countries throughout the world. The countries with thelargest production during 1975 were Canada. France,
South Africa and the United States of America. Reservesof varying quantity and grades. and mining operations ofvarying scale are reported in several other countries(110).

11. Uranium ore has been obtained mainly fromunderground mines, although large amounts are now
coming from open-pit mines. In the United States,
70 per cent of the ore production in 1973 came from 33
open-pit mines and the remainder from 122 underground mines (14). In South Africa, uranium is
extracted from gold-mine tailings (117).

those nuclides because the 2 2 6 Ra level has been themost restrictive, making 2 2 6 Ra the most relevant
nuclide of concern. At most mills. tailing ponds are of
sufficient size for waste liquids to evaporate or seep into
the ground. When space is not available, some liquid
effluents are released into rivers. usually after treatment
to reduce the radium content.

12. Since the ore normally contains less than 0.2 percent uranium, processing mills are usually located in themining areas to avoid transportation of large quantitiesof what is ultimately waste material. In some cases,
however, ore is treated locally by heap-leaching, andprecipitated pre-concentrates are then transported to themills. In the mills, the ore is blended and crushed, thenwet ground and transferred as a slurry to leaching tanks.
According to the ore characteristics, either a sulphuricacid or a sodium carbonate alkaline leach process isutilized. The soluble uranium is recovered by solvent
extraction, or, in the case of the alkaline leach process,by precipitation. The uranium concentrate is calcinedand pulverized, forming the final product of the mill. thesemi-refined compound called yellowcake.

16. Releases of radioactive particles and radon gas into
the air during mjlling operations may originate from theore piles, the crushing, grinding and leaching operations,and the tailings piles. The release of radon activity froma dry tailings pile containing 560 pCi 2 2 6 Ra per gram isestimated to be 500 pCi m-2 s- 1 (156). The release rate
at any one location may vary, however, by an order ofmagnitude owing to the effects of such weather variables
as wind speed, barometric pressure, atmosphericstability, rainfall, and snow cover ( 133). A pile saturated
with water, as it may well be at an operating mill. may
have its radon release rate reduced by a factor of 25(156).

A. EFFLUENTS
13. Mine wastes include liquids, solids, and exhaust
ventilation gases. The liquids are mostly water frommine drainage and water used in drilling; it is discharged
to ponds and allowed to evaporate. Leaching of minetailings may be an additional source of liquid waste.
Solid waste is primarily waste rock and low-grade ore.
Ventilation of mines results in the discharge of large
volumes of air containing radon and radon daughters.
14. The radionuclides contained in the ore include
uranium, primarily 238 U. and the uranium daughter
products. such as 2 3 0 Th and 22 6 Ra. In most cases little,
if any, natural thorium {2 32 Th) is contained in the millfeeds. One tonne of ore containing 2 kg U 3 0 8 has anactivity of 566 µCi from each member of the 23 8 U
decay chain, or a total of about 7900 µCi ( 126). About.
14 per cent of the total activity contained in the ore fed
to the mills is recovered in the uranium concentrate(14), while the average recovery of uranium is 91 percent. In addition, the uranium concentrate contains
about 5 per cent of the total thorium and 0.2 per cent of
the radium contained in the ore (126). With little of theparent remaining, the short-lived daughters 2 3 4 Th
(24.1 d), 234 mpa (1.17 min) and 231 Th (25.5 h) arelost by decay from the mill wastes. The remaining 70 per
cent of the activity originally present in the ore is largely
undissolved and is discharged in the solid mill tailings.
15. Most of the radium in the ore is insoluble andremains in the tailing solids; a small portion, about l per
cent or less, is dissolved. Radium in mill liquid effluentsmay vary from 250 to 500 pCil- 1 (14). Up to 50 per
cent of the 2 3 0 Th in the ore may be dissolved in acidsolutions and be contained in liquid effluents until thesolution has been neutralized and the thorium
frecip~tated. The radionuclides 2 2 3 R, 22 7 Th, 2 27 Ac.
10
Pb and 21 0 Po may also be present in the liquid
effluents, but relatively little information is available on

17. From a recent survey of milling practices at active
uranium mills in the United States (126), the annual
airborne radioactive releases from operating mills at two
representative locations have been estimated. Table l
gives t!1e average and normalized (per MW(e) y) releases ofvarious radionuclides from a mill processing 664 OOO t ofore per year (151). The ore is assumed to be dusty andto contain 6 per cent moisture. Dust from ore crushingcontains all radionuclides · in equilibrium. However,
yellowcake dust contributes tp the uranium release andaccounts for the lack of equilibrium in the total release.
The sources of the airborne radioactivity in thisestimate, in addition to the mill itself, include radonfrom the tailings pond and radon and particulates fromthe dry beach area. Since in this particular survey the
maximum value of dry beach area was assumed for an
active mill, namely the value that applies near the end of
the assumed 20-y life of the mill, the releases from thedry beach area have been reduced by a factor of 2 toreflect more nearly average values for an operating mill.
This correction mainly affects the amounts of radon
released.

TABLE 1. ESTIMATED RADIOACTIVE AIRBORNE EFFLU·
ENTS FROM MILLING OPERATIONSRadionuclide
1llu

n•u

134111130111
11•Ra

llOPb
210Bi
21opo
lllRn

Annual activity
release from mill andactive tailings areaO,
(Cl)

0.09
0.09
0.01
0.01
0.009
0.009
0.009
0.009
1400

Normalized releaseb(mCi per MW(e) y)

0.01
O.Ql
0.001
0.001
0.001
0.001
0.001
0.001
170

Source: Reference 126.
a Assuming an ore processing rate of 664 OOO t y- 1 •b Assuming a uranium content of 0.2 per cent and anannual requirement of 160 t of uranium for a I GW(e) light-water
reactor.

167

18. The tailings area remalfllilg after the mill hasceased operation becomes a long-term source ofenvironmental radioactive contamination due to wind or
water erosion and to radon emanation. Stabilization
programmes are generally conducted or planned so thaterosion from tailing areas is alleviated (126). Radon-222
arises from decay of 22 6 Ra in secular equilibrium withlong-lived 230 Tb (80 OOO y). A few per cent of uranium
isotopes in the ore may also remain in the tailings
material, constituting through the decay chain an evenmore long-lived source of 2 2 2 Rn emanation.
19. Emanation from the tailings arises primarily fromradon produced in the surface layer. Of the total. radon
produced, up to about 20 per cent diffuses out of thecrystallized structure of the mineral (28). The relatively
short half-life of 2 22 Rn (3.8 d) results in the decay ofmuch of the radon during gaseous diffusion towards thesurface of the tailings pile. About 90 per cent of theradon released to the air originates in the top 1-2 m(124, 156), and the release rate is not changed byincreasing the tailings depth much beyond about 3 m.
20. Covering the tailings pile with soil reduces theradon emanation rate by a factor of two for each metreof cover. Accordingly, about 6 m of earth cover isrequired to reduce the emanation to values observed
naturally for soils (124).

B. LOCAL AND REGIONAL COLLECTIVE
DOSE COMMITMENTS
21. Kristan et al (78) have reported 22 6 Ra to be themost significant nuclide in water releases from mining
and milling operations. When, as a result of releases,
there is some increased radium concentration in fish,consumption of fish could be the dominant exposure
pathway to individuals in the local population. Anotherpossible pathway is via drinking water.
22. An inactive mill at Durango, Colorado (UnitedStates), has been studied extensively (137, 138). Before
1959, this mill discharged untreated effluents into theAnimas River. Estimated 2 2 6 Ra release rates to the river
were 0.2 mCi d -t in dissolved form and 30 mCi d - 1 insuspended solids (36). The radium concentration in theriver was 12.6 pCi 1- 1 below the mill and 2.9 pCi 1- 1
JOO km downstream, compared to 0.3-0.6 pCi 1- 1
upstream. Concentration factors of 500-1000 for algae
and aquatic insects, 3-4 for fish flesh and 100 for fishbone were reported (137). Because of the environmental
contamination, waste treatment was introduced, which
reduced the concentration of 22 6 Ra in the river toessentially the natural level (137). More recent surveys
of uranium milling areas in the United States have notshown measurable water contamination beyond theplant sites (129, 130, 131).
23. Under the assumption that adequate waste
treatment prevents seepage of process liquid effluents
directly into rivers, the local collective dose commitments from milling operations are expected to dependon the airborne releases (table 1), which contribute toexternal exposure from deposited material and tointernal exposure via the inhalation and ingestion

168

pathways. The mills in the United States are typicallylocated in areas not particularly suited for farming. The
raising of beef cattle in the vicinity of the mills and theconsumption of the beef may represent the only
significant contribution to the intake by ingestion.
Population in the milling areas is generally sparse andlittle growth can be expected. Based on a survey of fivemills in the United States, Sears et al (126) found that a
population of about 5 104 within a radius of 80 km isrepresentative. This corresponds to an average population density of less than 3 km -z , with lower andnon-uniform population densities very close to the mill.
For areas at a greater distance, a population density of
25 krn- 2 would be representative. Variations in thesefigures, particularly the local population density, would
be expected for specific mills in various other countries.
24. The collective dose from direct inhalation ofradioactive materials released during operation can beassessed by considering the dispersion of the airborne
particulates and radon with distance from the mill or
tailings area. Annual average dispersion factors of about
10-6 s m- 3 at 1 km have been found appropriate for themeteorology of typical mill locations (126), while forlarge area sources the value is smaller by a factor of
about 4 ( 156). In this Annex, a value of 5 10-7 s m - 3 at
1 km is assumed to apply for both the mill and thetailings area.
25. Over a limited range, the decrease of theconcentration of activity in air with distance from thesource appea:s to be approximately inversely proportional to the 1.5 power of the distance (I 26), but thisbehaviour cannot be easily verified by measurements
since conce~trations barely exceed natural levels at a few
kilometres, Shear-er and Sill reported measurements ofradon concentrations 5, 1.5 and 0.15 times backgroundat 0.8, 1. 7 and 8.3 km respectively from inactive milltailing areas (127). This corresponds to concentrations
inversely proportional to the 1.5 power of the distance.
Similar results were obtained by Breslin, who analysedvarious measurements of radon and particulate concen•
trations with distance from seven tailings areas (20). The
relationship at much greater distances must be assumed
to be only approximate, but it should serve to giveindications of average dispersion over longer release
periods.
26. Deposition of the airborne particulates reduces thearea over which the integration of collective doses iscarried out. For ground-level releases of particulatematerial, less than 30 per cent of the released material
remains in the plume at 100 km and less than 20 per
cent at 200 km (164). Integrating the collective doses to
500 km, therefore, will include virtually all the collective
dose from airborne particulates released by the mill.
Estimates of the collective doses from direct inhalationof particulates, calculated assuming the population
distribution given in paragraph 23, are summarized intable 2.
27. The collective dose commitment from radon
released during operations from the mill and the tailings
area can be estimated by the same procedure used forthe case of inhalation of particulates. For a radon releaseof 0.17 Ci per MW(e) y (table I), and using a dispersion

TABLE 2. ESTL\fATED COLLECTIVE DOSE COMMITMENT FOR AIRBORNE PARTICULATE RELEASES FROM MILLINGOPERATIONS
Collective dose commitment per unit activity released(man rad Ci- 1)
Ingestion aRadionuclide

External

Lung(whole)

Bone
ma"ow

Bone.
liningcells

Lung(whole/

Bone
ma"ow

Bone
lining
cells

10·•
1o-•
10-•
10·•

0.06
0.06
0.04
0.09

0.1
0.03
0.1
0.3

0.4
0.4
1.0
1.0

1.2
19.0
3.6
1.7

0.06
12.0
0.09
0.7

0.3
180.0
0.6
2.5

13au
230Th
226

Inhalation

Normalized
activity
released(Cl per MW(e) y)Ra

210Pbb

Whole
body

10.0

Normalized incomplete
collective dose
commitment (man rad
perMW(e) y)Lung(whole)

Bone
ma"ow

Bone
liningcells

1 10·•
2 10·•
1 10-•
2 10·•

2 10-•
1 10-•
1 10-•
1 10-•

7 10-•
2 10 4
1 10-•
4 10-•

aThe values are the incomplete collective dose commitments due to deposition on the ground, incurred mainly through thein~estion pathway but also including a contribution from inhalation of resuspended materials. External exposure from deposited activity(' 'Ra including its daughters) is considered separately.
bJncluding the CC''ltribution to dose from 2 1 0 Po daughter.

factor of 5 10- 7 s m - 3 , the normalized time-integrated
concentration in air at 1 km is 0.024 pCi h 1 -i per
MW(e) y. Using a dose factor of 45 µrad per pCi h 1 -I
for radon in equilibrium with its daughters and an
equilibrium factor F = 0.6 (see Annex B), the normalized dose commitment at 1 km is estimated to be
6 10-4 mrad per MW(e) y. At a few kilometres theequilibrium factor outdoors is probably smaller than 0.6
but would increase indoors because of the delay of the
air circulated indoors. Therefore, not much error isintroduced by assuming that the same value (0.6) applies
over tens of kilometres. The local collective dose
commitment within 100 km from the mill can be
estimated to be

29. An approximate estimate of the incomplete
collective dose commitment due to the deposition of thereleased particulates by external exposure and theingestion pathway, including inhalation of resuspended
materials, may be obtained by assuming that the released
nuclides, once deposited on the ground, behave similarly
to their naturally occurring counterparts. Assuming thatthe deposited nuclides are distributed over a depth of
20 cm of soil and that the dose from all relevant
exposure pathways is proportional to the activity
concentration in that layer of soil, the incomplete
collective dose commitment is given by

D, (-t,ru N6 2rcp cip== 2 10-4 man rad per MW(e) ywhere
is the incomplete collective dose commitment
in tissue i due to nuclide j, D;;J is the dose rate in tissue ifrom the natural levels of nuclide j, A/ is the activity perunit surface area in the top 20-cm layer of the naturally
occurring nuclide j, Qi is the activity released of nuclide j(assumed to be equal to the deposited activity), Nr, isthe population density, and r is the assumed duration of
the practice. It is assumed in this model that only decay
removes the deposited nuclides from the soil.

Sc=

f 0°
1

where D 1 is the dose commitment per unit electricity
generated at the reference distance p 1 = 1 km, N 6 is the
population density (3 km - 2 ), and p is the distance fromthe mill.
28. The regional collective dose commitment from100 to 1500 km, obtained from similar integrations, but
adjusted to account for radioactive decay and assuming a
population density of 25 km -:2 (para. 23), is 4.4. 10-3
man rad per MW(e) y. The sum of the local and regional
components of the collective dose commitment is
5 10-3 manradperMW(e)y. Thisvalue refers to the basal
cells of segmented bronchioles; the corresponding
collective dose commitments in the whole lung, gonads,
bone marrow, and bone lining cells are listed in table 3.

D~J

SI.1 = ~~ Q1 N6 t
j

SlJ

30. Representative values for b,~1 and Aj can be foundin Annex B. As most of the released particulate activity
is deposited within a distance of80 km (126). a value of
3 man km-2 is used for N0 • Little is known about theeffective residence time of these nuclides in soil.
However, a conservative ,assessment of the incomplete
collective dose commitment can be obtained byassuming that the availability of the nuclide decreases
only by radioactive decay.

TABLE 3. COLLECTIVE DOSE COMMITMENTS FROM RADON RELEASED DURING ACTIVEMILL OPERATI0Na(man rad per MW(e) y)
TissueGeograpllic region
Local 0·100 km)
Regional (100-1 SOO km)

Segmental
bronchioles

Lung(whole)Gonads,bone marrow,bone lining cells

2 10 4
4 10-3

4 10-•
9 10 4

3 10· 1
7 10-•a Assuming a normalized release of 0.17 Ci per MW(e) y.

169

31. Table 2 presents the calculated incomplete collective dose commitments per unit activity released, theestimated amounts and the incomplete collective dose
commitments per unit electrical energy generated. The
incomplete commitments a,-· integrated over the period
of operation of the mills. and give some indication of the
maximum future annual per caput doses to be expectedfrom that operation, assuming the duration of themilling practice is 100 y. Reasonably assured reserves ofuranium ore correspond to approximately 10 6 t ofuranium (110), a value equivalent to some 6000
light-water reactor-years. At the presently installedcapacity, the mining and milling practice could proceed
for Jess than 100 y, unless further ore reserves become
available.
32. The external exposure due to deposited 2 2 6 Ra
will give an absorbed dose rate in air of the order of
30 rad h- 1 per Ci m - 2 for a plane source and 5 rad h -i
per Ci m - 2 when the material is evenly distributed in theupper 20 cm of the soil, provided the activity per unit
mass is considerably higher than in the soil material.
Assuming a shielding factor of 0.32, the incomplete dose
commitment will be in the range 4-25 man rad Ci- 1 or
(4-25) 10- 6 man rad per MW(e) y. The average values areof the order of 10 man rad Ci- 1 and 10-s man rad per
MW(e)y.
33. Estimating of the contribution to the collective
dose commitment per MW(e) y made by radon releasefrom the tailings pile remaining after the active life of
the mill entails some sophisticated guesswork. Radonemanation could continue for millions of years becauseof the long half-life of the remaining 23 8 U precursor,
but the continued presence of the pile on the surface forsuch a long period is unlikely. This Annex considers only
the simpler but still uncertain case of an uncovered
tailings pile undisturbed for times of the order of thehalf-life of 2 3 0 Th (80 OOO y).
34. The uranium requirement of light-water reactors,
160 t per GW(e) y, is equivalent to 80 OOO t of
0.2-per-cent ore, resulting in just slightly less 'than
80 OOO t of solid tailings material. In the United States,
more than 80 10 6 t of tailings (equivalent to 10 6MW(e) y) occupied1 8 10 6 m 2 in 1970 (156),
corresponding to 8 m 2 per MW(e) y.
35. For a radon emanation rate from the tailings pilesof 500 pCi m - 2 s- 1 (156), and a dispersion factor of
5 10-7 s m - 3 , the normalized concentration in air at
1 km is 2 10-6 pCi 1 -l per MW(e) y, assuming that
1 MW(e) y corresponds to 8 m2 of the tailings pile, as
discussed in the previous paragraph. Over the mean lifeof 230 Th (1.44 X 80 OOO y), the time-integrated radon
concentration is 0.23 pCi y I -l per MW(e) y. Using thedose factors and the dispersion assumptions discussed inprevious paragraphs for the case of radon releases during
operation of the mill, the collective dose commitments
for radon release from the inactive tailings area can beassessed. The dose commitment to segmented bronchioles at 1 km is 50 rnrad per MW(e) y. The local(1-100 km) collective dose commitment is estimated to
be 20 man rad per MW(e) y, the regional (100-1500 km)
value being about 400 man rad per MW(e) y. Both thesevalues correspond to irradiation of the basal cells ofsegmental bronchioles.
170

36. An alternative estimate of the collective dose
commitment due to radon release from the tailings of an
inactive mill can be obtained by comparison with naturalradon emanation from soil and the corresponding radon
concentration in air. With the assumptions used above
and assuming the natural radon exhalation rate to be
0.42 pCi m - 2 s -l and the corresponding radon concentration in air to be 0.1 pCi 1-1 (see Annex B), thecollective dose commitment is calculated to be
1.SC=(SOO pCi m-2 s-1) (0.1 cil- )p. _ _
0.42pC1m 2 s 1X

X (8 m 2 MW(e)- 1 y- 1 )(10-6 km 2 m- 2 ) X
X (25 man km- 2 Xl.44)(80 10 3 yX0.6) X
X (4.5 10-5 rad pCi- 1 h- 1 1)(8760 h y- 1):::::: 650 man rad per MW(e) yin reasonable agreement with the value assessed in theprevious paragraph. The corresponding collective dose
commitment would be 100 man rad per MW(e) y to thewhole lung and 0.8 man rad per MW(e) y to the gonads,
bone marrow and bone lining cells. The incomplete
collective dose commitments (see Annex A), estimatedfor an assumed duration of the mining and milling
practice of 100 y, are smaller than the above values by afactor of 10 3 •37. Table 4 summarizes the estimated values of thecollective dose commitments due to radon release fromthe tailings area of inactive mills. These estimates arequite uncertain, and the true values will depend strongly
on future practice. For example, if the fuel isreprocessed and if uranium is used in breeder reactors,
the corresponding equivalent area of the tailings pile
(8 m 2 per MW(e) y) will be much smaller and, therefore,
the contribution to the collective dose commitment per
MW(e) y made by milling will also be reduced.
TABLE 4. COLLECTIVE DOSE COMMITMENT TO TISSUESFROM RADON RELEASED FROM INACTIVE MILLTAILINGS(man rad per MW(e) y)
Type ofcommitment

Segmental
bronchioles

Lung(whole)Gonads,bone marrow,bone lining cells

Complete
Incomplete

500
0.5

100
0.1

0.8
0.0008

Note: The commitment is for the local plus regional
areas (radius 1-1500 km), assuming no soil cover. The values
should be halved if a soil cover of I m is assumed.

II. URANIUM FUEL FABRICATION
38. The uranium ore concentrates produced at themills are further processed and purified and usuallyenriched in the isotope 2 3 5 U before being converted touranium dioxide or metal and fabricated into fuelelements. Natural uranium, which contains 0.7 per cent235
U, can be utilized in heavy-water or graphitemoderated reactors. Li~ht-water reactors require fuelenriched to 2-4 per cent 3 5 U.

39. Gaseous diffusion is the usual procedure forenrichment. but other methods can also be used. The
U 3 0 8 concentrate extracted from the ore must first be
converted to the volatile compound, uranium hexafluoride (UF 6 ). The hydrofluor process consists ofreduction. hydrofluorination and fluorination of the oreconcentrates to produce crude UF 6 , followed byfractional distillation to obtain the pure product. At thegaseous-diffusion installation, the UF 6 is pumped
through a series of porous barriers which discriminate
against the passage of the heavier isotope of uranium bya factor of 1.0043 at each stage. About 1700 barrierstages are required to produce an enrichment to 4 percent (151 ),.

40. In the final fuel fabrication step, UF 6 is converted
again to an oxide or to the metal for use in fuelelements. For use as fuel in light-water reactors, thedioxide powder is pelletized, sintered and loaded into

stainless steel or zircaloy tubing which is then capped
and welded. Uranium-metal fuel for use in somegas-cooled reactors is clad in magnesium alloy (magnox).
The fuel rods are assembled in arrays as fuel assemblies
for installation in the reactor core.

A. EFFLUENTS
41. The radioactive materials discharged to theenvironment from fuel fabrication, conversion andenrichment installations are limited. Most of theuranium compounds are solid, and conventional
air-cleaning equipment is used to remove particulates
from airborne wastes. Liquid wastes are collected insettling tanks or ponds. Estimates of release amounts
from typical installations, based on limited monitoring
information, have been published in the United States(151). A summary of these estimates is given in table 5.

TABLE 5. ESTIMATED RADIOACTIVE RELEASES FROM FUEL FABRICATION OPERATIONSIN THE UNITED ST ATES
Installation release
rate (Cly·')u

Airborne

LiquidEquivalentinstallation
size (GW(e))

0.04
0.18
0.005

1.2
1.8
0.5
0.09
0.04
0.26
0.26

27.5
90.0
26.0
27.5
27.5
26.0
26.0

ll6Ra
l~

l3'4Th2:i4paNormalized release(lo-• Cl per MW(e) y)
Airborne

LiquidOperation

1.5
2.0
0.2

44.0
20.0
20.0
3.4
1.5
10.0
10.0

Conversion
Enrichment
Fabrication
Conversion
Conversion
Fabrication
Fabrication

Source: Reference 151.

42. Residual amounts of 226 Ra and 230 Th areremoved from the uranium material in the conversion
operation. Small amounts of these isotopes may be
included in the liquid effluents from the conversion
installations. The thorium daughters of uranium areseparated at the mill, but 234 Th, with a half-life of24 d,grows back to over 99 per cent secular equilibrium in
170 d; small amounts of 234 Th and its short-lived
daughter 2 3 4 Pa are thus assumed present with uranium
in liquid effluents from the fuel fabrication plant.
43. Estimates of the release of radioactive materials
into the environment from enrichment and fuelfabrication installations in the United Kingdom are alsoavailable (106). The release of uranium to theatmcsphere from the Capenhurst enrichment plant in
the period 1972-1975 averaged 5 kg y- 1 • Assuming
280 µCi of 238 U and of 234 U per kilogram of U 3 0 8 ,the annual release is equivalent to 0.0028 Ci.

44. The alpha activity released annually in liquid
effluents from the Capenhurst plant for the same period
is estimated to be 0.09 Ci (106). The Springfields fuelfabrication plant in the United Kingdom of Great Britain
and Northern Ireland discharged small amounts of liquidradioactive waste into the tidal river Ribble. Averagealpha activity released annually in liquid effluents during
1972-1975 was 34 Ci {106).

B. LOCAL COLLECTIVE DOSE COMMITMENT
45. The liquid discharges from the Springfields fuelfabrication plant result in external exposure ofindividuals who frequent the river bank near the outfall.
The annual collective dose due to the local population isestimated to be less than 0.1 man rad (100).
46. Estimates of collective dose commitments fromairborne releases can be obtained in a manner similar tothat used for releases from milling operations. The
combined airborne release of uranium from all fuelfabrication operations, on the assumption that the values
given in table 5 are representative. is about 4 10-6 Ci per
MW(e) y. A population density of 25 km- 2 has been
assumed to apply at all relevant distances from the fuelfabrication installations. The estimated collective dose
commitment assumes that the released uranium, once
deposited on the ground, will remain available to enter
the ingestion or the resuspension-inhalation pathway to
man for the entire mean-life of 238 U. This assumption
obviously leads to a considerable overestimation of thecomplete collective dose commitments. The values forthe incomplete collective dose commitments (seeAnnex A) given in table 6 assume that fuel fabrication
operations to serve reactors will be continued for alimited time, which for purposes of calculation is taken
to be 500 y.
171TABLE 6. ESTIMATED INCOMPLETE COLLECTIVE DOSE COMMITMENTS FOR URANIUM IN
AIRBORNE EFFLUENTS FROM FUEL FABRICATION OPERATIONSa

Per unit electricalenergy generated(man rad per MW(e) y)
Total

Pfa" unit activity released(man rad per Ci)
Inhalation

Ingestionb

Lung

Bone
marrowBonelining
cells

0.5

0.8

4.0

Lung

Bone
marrowBonelining
cells

2

0.09

0.4

Lung

Bone
ma"ow

Bone
liningcells

1 10-•

4 10-•

2 10-•

0
1ncluding conversion, enrichment and fabrication and assuming a normalized 2 3 • U release of
410-• CiperMW(e)y.
bThe values are collective doses due to deposition on the ground. They are mainly incurred
through the ingestion pathway, but include also the contribution from inhalation of resuspended
materials.

Ill. REACTOR OPERATION
47. The common reactor types include thepressurized-water moderated and cooled reactor (PWR),
the boiling-water moderated and cooled reactor (BWR),
the gas-cooled, graphite-moderated reactor (GCR), thewater-cooled, graphite-moderated reactor (LWGR), theheavy-water moderated reactor (HWR). which may becooled by heavy water, light water or gas, and the fast
breeder reactor (FBR).
48. Table 7 gives the number of operational reactorsof each type and the electric generating capacity foreach country at midyear in 1976(74). The total installed
nuclear generating capacity in the world in 1976 was
79.8 GW(e) from 188 reactors in 19 countries.

A. EFFLUENTS
49. During the operation of a nuclear reactor,
radioactive fission and activation products are produced.
These radioactive materials are for the most part
retained within the fuel elements. Most of theradionuclides which diffuse into or are formed within
the coolant are removed by the gaseous and liquid waste
processing systems. Low-level releases, which occur
during normal operation, are controlled and monitored.
50. The type and quantity of radioactive materials
released from reactors depends on the reactor type andon the specific waste processing systems utilized.
Radionuclides may reach the environment througheither the gaseous or liquid effluent streams. In the
airborne effluents are found fission noble gases (krypton
and xenon isotopes), activation gases ( 41 Ar, 14 C, 16 N,
35S), tritium, radioactive halogens and particulates. In
the liquid effluents are tritium, fission products andactivated corrosion products.

NTP per thermal megawatt-day. The fraction thatescapes from the fuel pellets during normal operation
generates gas pressure in the fuel can. The appearance of
noble gas activity in effluent streams results mainly fromfuel-cladding failure. Only insignificant amounts areformed in the coolant by fission of trace uranium on
external fuel-element surfaces.
52. In PWRs the primary coolant is maintained in asealed loop, which is opened only infrequently.
Therefore, much of the short-lived noble gas activity
decays before venting. Short-lived noble gases originate
mainly from water that has leaked from the primarycoolant system.
53. A side stream of the primary coolant in PWRs is
continually diverted into a subsystem (primary coolant
letdown) for control of chemical content and for
coolant purification. The gaseous waste effluents fromthis system are compressed into tanks and held for aperiod of 30-120 d for decay of short-lived isotopes.
Other gaseous effluent streams in PWRs originate fromthe main condenser exhaust, secondary coolant blowdown, reactor building ventilation including containment purg~s (-4 times a year), and turbine andauxiliary building ventilation (166).
54. In table 8, which lists the reported noble gas
releases from PWRs, it can be seen that the normalized
release amounts span a considerable range. Leakage is an
important and variable source term. The total activity
released divided by the net electrical energy generated
for all PWRs for which effluent data are available varied
from 10 to 20 Ci per MW(e) y during 1970-1974. The
composition of the release from PWRs is primarily
longer-lived 13 3 Xe (5.3 d), but some shorter-lived noble
gas isotopes are also present. Krypton-85 (10.7 y) may
provide a few per cent to the PWR release activity
composition. The activity of specific noble gas isotopes
in the releases reported for PWRs in the Federal
Republic of Germany and in the United States during
1974 are given in table 9.

1. Fission noble gases

51. Several isotopes of the noble gases krypton andxenon occur by fission in the nuclear fuel. The total
volume of xenon and krypton generated is about 25 cm3172

55. In BWRs, continuous removal of non-condensable
gases in the steam flow occurs via the main condenser
air-ejector system. The noble gases enter the gaseous
waste stream primarily at this point. Secondary

TABLE 7. WORLD :NUCLEAR ELECTRIC GENERATING CAPACITY IN 1976
Net values in MW(e) and number of reactors in parentheses
Reactor rype
Country

PWR

BWR

GCR

United States

24 398(34)

15 208
(23)

330

2 879
(5)

Japan

L WGR

FBR

Tora/

39 936
(58)

(1)

6 660
(31)

United Kingdom

HWR

92

244

(1)

(2)

6 996
(34)
6 297

3 264

154

(7)

(1)

l 148
(4)

14

52

5 247

(l)

(1)

(11)

(13)

Germany, Federal
Republic of

4 033

Union of Soviet
Socialist Republics

2 094

94

2 882

146

(6)

(5)

(11)

(2)

822

2 362

3 184

(1)

(4)

(5)

(5)

SwedenFrance

2 133
(7)

270
(l)

250
(1)

l 663

(4)

Switzerland

(10)
(7)

l 663

Spain

2 723

2 540

2 540
(7)

CanadaBelgium

70
(1)

5 216
(24)

(4)

153

440

480

1 073

(1)

(1)

(l)

(3)

700

306

(2)

(l)

1 006
(3)

German Democratic
Republic

879

879

(3)

(3)

Bulgaria

837

837
(2)

(2)

lndia
Italy
Netherlands

396

207

(2)

(1)

603
(3)

542

242

150

150

(1)

(1)

(1)

447

52

499

(1)

(1)

(2)

Argentina

(3)

319

319

(1)

(1)

Pakistan

125

125

(1)

(1)

Czechoslovakia

110

110

(l)

(1)

Total

39 416(66)

23 420
(49)

9 921
(43)

3 423

2 882

640

79 794

(13)

(11)

(5)

(188)

Source: Reference 66.

pathways include the turbine gland-seal system, thecondenser mechanical vacuum pump, and any leakage ofprocess fluids to ventilated building spaces ( 165).
56. Table 8 also lists the reported noble gas released
activities from BWRs. The normalized releases vary fromless than 100 to more than 10 OOO Ci per MW(e) y. The
value representing the overall operating experience has
declined from about 1500 Ci per MW(e) y in 1970 to
IOOOCiperMW(e)y in 1974. The release composition ofnoble gas isotopes from BWRs varies somewhat,
dependent on the specific hold-up time. The hold-up
times range from 20 to 30 min for older BWRs to 2 h or
more for more recently designed hold-up systems. The
reported release compositions for BWRs in the Federal
Republic of Germany and in the United Sates during

1974 are given in table 9. Important components of thereleases are 135 Xe (9.2 h), 133 Xe (5.3 d), 138 Xe(17 m), 88 Kr (2.8 h), 87 Kr (76 m) and ssmKr (4.4 h).
The specific release activities are not necessarily typicalfor any one reactor, but the total normalized release issimilar to the overall average from table 8.
57. In GCRs. fission noble gas release is insignificant.
Gas-cooled reactors include the Magnox type, theadvanced gas-cooled reactor (AGR) and the hightemperature gas-cooled reactor (HTGR). The Magnox
type utilizes metallic uranium fuel and C0 2 coolant.
Because of the magnesium-alloy cladding fuel temperatures must be kept at less than 650°C, at which
temperature release of noble gases from the fuel material
is negligible.
173

T ADLE 8. NOBLE GASES DISCHARGED IN AIRBORNE EFFLUENTS FROM REACTORS IN VARIOUS COUNTRIES, 1970-1974....;i.i,.

ReactorNetelectricalpower(MW(e))Start-up
year

Release (hCiJ
1970

1971

1972

Normalized release (Ci per MW(e) y J
1973

1974

1970

1971

1972

1973

1974

PWR

Belgium
BR-3, Mo!

1962

10

France
SENA, Chooz

1967

270

0.003

4.5

31.3

19.9

Germany, Fed. Rep. of
Obrigheim
Stade
Oiblis

1968
1972
1974

328
630
1 147

7.7

1.46

3.20
2.45

2.93
2.61

Italy
Trino, Verccllese

1964

247

0.019

0.59

1.03

Japan
Milrnma I
Mihama 2
Takahama l

1970
1972
1974

340
500
826

0.9

1.4

0.62
0.26

Netherlands
llorsscle

1973

447

United States
Yankee Rowe
Indian Pt. l
San Onofre
Connecticut Yankee
R. E. Ginna
Point Beach, 1,2
11. B. Robinson
Palisades
Maine Yankee
Surry, 1,2
Turkey Pt. 3,4
Indian Pt.. 2
Ft. Calhoun
Prairie Island, 1,2
Oconce 1,2,3Zion 1.2
Arkansas 1
KewauneeThree Mile Island

1961
1962
1968
1968
1970
1970/72
1971
1971
1972
1972/73
1972/73
1973
1973
1973/74
1973/74
1973
1974
1974
1974

175
265
430
575
420
2 X 497700
700
790
2 X 788
2 X 693873
457
2 X 530
3 X 886
2 X 1 050
820
520
810
Total

Electrical energy generated (MW(e) y)
Overall normalized release (Ci per MW(e) y)
11 IV

Germnny, Fed. Rep. of
Kahl,n,.:,

190

0.25

0.51
0.34

0.07
0.34
0.07

0.31

5.83
0.040
0.61
1.78
0.0074
0.76
9.74
2.31
0.00003
6.36
55.0
4.66
5.58
0.30
0.36
19.4
2.99
0.20
3.35
0.92

0.017
1.7
4.2
0.7
10.0

0.013
0.36
7.67
3.25
31.8
0.84
0.018

0.018
0.54
19.1
0.65
11.8
2.81
0.26
0.51
0.002
0.00001

0.035
0.12
11.0
0.032
0.58
5.75
3.1
0.45
0.16
0.87
0.53
0.015
0.066
0.008
9.3
0.004

-25.2

51.9

--74.8

--

1 906
13.2

3 124
16.6

3 960
18.9

13.5
0.89
0.06

58.6

-135.1

6 083

9 045

9.6

14.9

0.02

20

130

27

5.7

12

82

6.5

0.13

3.8

4.5

11

5.6

4.9

5.4
1.2

2.8
1.1
0.27

4.0

18

0.25
3.8

0.29

59

42
0.12
1.5
9.9
7.2
1.7
0.41
1.0
0.99
0.38
0.96
3.6
35
0.052

0.38
4.5
5.0
0.015
3.2
13
4.2
0.0038
16
!12
5.2
15
I. I
2.2
3[
5.6
3.0
18
3.8

1.1

0.12
39
12
1.7
38

0.08
4.1
21
6.8
100
2.1
0.061

1.3
43
7.9
0.51
2.3
0.04
0.0003--~

1.3.J.

Overall normalized release (Ci per MW(c) Y)

Germany, Fed. Rep. of
Kahl
Gundremmingen
LingenWiirgassen

1961
1966
1968
1972

15
237
174
640

2.90
7.35
110

Italy
Garigliuno

1964

152

275

640

290

Japan
Tsuruga
Fukushima 1,2

1969
1970/74331
460/784

130
1.2

42
70

4.9
97

5.2
4.9

Netherlands
Dodewaard

1968

52

2.41

2.76

8.32

United States
Dresden l
Big Rock Pt.
Humboldt Bay
Lacrosse
Oyster Creek
Nine Mile Pt.
Dresden 2,3
Millstone Pt. 1
MonticelloQuad Cities, 1,2
Pilgrim 1
Vermont Yankee
Peach Bottom 2,3

1960
1962
1963
1969
1969
1969
1970/71
1970
1971
1972
1972
1972
1973/74

200
70
69
53
640
625
2 X 800652
545
2 X 800655
514
2 X 1 OOO

900
280
540
0.95
110
9.5
250

753
284
514
0.53
516
253
580
276
75.8

Browns Ferry

1973

I 050

Cooper Station

1974

760
Total

Electrical energy generated (MW(c) y)
Overall normalized release (Ci per MW(e) y)

2.10
6.50
8.70

220
35
960

160
29
75

3 300

4 800

6 OOO

5.6
14

470
160

170
230

6.70

4.16

57

877
338
430
30.6
866
517
429
726
751
132
18.1
55.2

840
230
350
91
810
872
880
79
870
900
230
180
0.004

98.4
188
572
49.0
279
558
627
912
1 490
950
546
63.9
0.2
64
1.6

5 700
6 100
11 OOO
64
270
44
1 800

11.0
5.1
0.59

0.48
23.7
2.6
0.56

0.99
4.15
9.5
0.05

2 619

4 024

5 587

6 376

6 438

I 790
I 460

3 121

4 369
1 280

5 552
l 150

6 166
1 040

1 290

37
120
18
2.4

95
19
180
0.9

19
330

18
22

33
29

60

220

160

130

9 500
7 200
14 OOO
21
I 200
770
1 300
650
450

7 OOO
5 900
10 OOO
I 100
I 700
1 400
460
2 OOO
1 900
300
200
I 100

13 OOO
4 800
7 300
4 OOO
2 OOO
2 200
890
370
2 300
780
490
880

2 500
4 900
13 OOO
1 300
660
I 500
830
2 200
4 500
1 OOO
2 400
220
0.4
150
7.6

53
84
9

GCR

1963/}
1965/
1966
1969/}
1970
1972

70
200
480
480
505
540

Germany, Fed. Rep. of
AVR-Jlllich

1968

Italy
l~1lina
Japan
Takai

FranceEDF 1,2,3 Chinon
SL 1,2 St. Laurent
Bugey

8.09

4.23

11.52

2.8

19.6

10.9

23.2

8.7

0.31

3.43

3.86
0.84

5.0
3.1

19.3

9.5

5.9
6.5

7.0
10.6

13

0.031

0.027

0.028

0.021

3

2.4

2.7

1963

153

2.50

2.47

3.66

18.4

11.7

26.6

1965

166

4.9

4.8

5.9

6.0

6.3

46.7

42.1

52.7

Total

15.8

14.9

25.8

16.9

6.3

0.010

-..I
U'I

Electrical energy generated (MW(e) y)
Overall normalized release (Ci per MW(e) y)

880
18.0

1 085
13.7

I 546
16.7

1 444
11.7

113
55.8

52.6

55.8

'

-....,TABLE 8 (conti1111ed)

0\Netelectricalpower(MW(e))Start-up
year

ReactorNormalized release (Ci per MW(e) y)

Release (Ii Ci)
1970

1971

1972

1973

1974

1970

1971

1973

1972

1974

I/WR

Argentina
Atucha1974

CanadaDouglas Pt.
Pickering 1,2,3,4

1966
220
1971/72/73 4 X 508

FranceEL-4-Monts d'Arree

1967

70

Germany, Fed. Rep. of
MZFR-Karlsruhe

1965

51

17

1.8

320

l 500

161

0.072

Total
Electrical energy generated (MW(e) y)
Overall normalized release (Ci per MW(e) y)

58.l

40.0

31.0

53.8

144.5

130

0.53

0.96

0.22

3.1

0.95

161

112

186

161

107

402

808

1 694

1 770

1 500

280

230

95

3

170

57

19

2 700

2 500

2 500

14

19

19

2

23

5.9

Sources: References 1, 26, 27, 43, 44, '.17-99, 120, 141, 143, 144, 161, 163.

TABLE 9. COMPOSITION OF NOBLE GASES DISCHARGED IN AIRBORNE EFFLUENTS FROM REACTORS

ReactorElectricalenergygeneratedin 1974(MW(e) y)

Amount released (Cl)
133

Xe

"sxe

293.4
608.2

12100
711

I 320
156

103.9
353.9

16.4
l 420

1.2
9.7

502.3
239.7
764.8
547.9
8.9
411.0
673.5
901.8
274.0

2.3
630
6 040
2 110
0.02
5 460
54 600
3 980

0.4
110
I 300
43.1
0.01
23.7
223
172
1.2

138

Xe

asKr

131

mxe

.. Kr

llsmxeumK,u,m Xe

97

Kr

41

Ar

37

Ar

l' WR

Germany,
Fed. Rep. of
Obrigheim
Stade
United States
Yankee RoweSan Onofre
Connecticut
YankeeR. E. Gi111m
Pt. Beach 1.2
H. B. Robinson
Palisades
Maine Yankee
Surry 1,2
Turkey Pt. 3,4
Ft. Calhoun

O.Q3

35
23
0.0005

321

15.8

1.5
63.9
4.0
8.9
23.l
532
8.1
291
0.5

0.02
105

394
0.8
0.008
88.6
38.0

0.03

0.0006

0.3

0.3

542

19.3
14.5

322

51.1

0.06
0.9

0.02
177

0.02

0.9

0.2
7.1
374
2.8
0.0002
19.8
17.5

133

251

I. I
32.0
3.2

0.00003
3.3
2.6
22.2

14.9
4.7

15.0
52.1

0.5

Total

,

..
Prairie Island
Oconee 1.2,3Zion 1,2
Arkansas I
KawauneeThree Mile
Island

160.0
632.4
536.5
65.l
182.6

357
18 600
2 980
0.02
2 690

0.1
527
6.3

239.7

891

12.6

Total

112 600

4 035

7 500

1435

Corrcspondiwi electricalenergy ("'.{ (e) y)
Normnlized release(Ci per MW(e) y)
Proportion of total (%)

0.0005
0.7

16.2

0.06

0.02

0.03
37.5

128
0.7
336.8

1 771

956.8

5 690

643.3

4 029

0.003
613.7

3 978

0.03

0.001
74.5

0.2
37.2

0.00001

0.03
2.2

2.5
11.I

0.0002

0.5

6.7

499.4

395.8

373.4

2 855

4 874

4 498

1.1

1.6

87.4

0.00010.0001
2.7
5.8

272.4

3 644

236.3

4 804

0.5

104

15.0

0.54

0.19

0.17

0.16

0.15

0.13

0.10

0.09

0.07

0.05

<0.01

17

90

3.2

1.1

1.0

1.0

0.9

0.8

0.6

0.5

0.4

0.03

<0.1

100

"'Xe

.. K,

33

Other

Amount released (kCi)
"' Xe

"'Xe

"Kr

umKr

umKr

usmxe

"Kr

131

0.04
0.2

0.2
0.5
0.4
0.005

0.1
0.6
0.001

0.04
0.04

0.3
1.0

0.03

0.04
0.3
0.7
0.02

0.4

3.8
4.7
3.3
0.03

0.7
1.2
2.9
0.2

3.7
0.8
0.2
0.006

Xe

'

mxe

BWR
Germany,
Fed. Rep. of
Kahl
10.4
Gundremmingen 219.2
Lingen52.9
Wiirgassen
55.7
United States
Dresden 1
40.0
Dig Rock Pt.
38.8
llumboldt Day
43.4
Lacrosse
37.7
Oyster Creek
422.4
Nine Mile Pt.
376.7
Dresden 2,3
753.4
Millstone Pt.
411.0
Monticello331.1
Quad Cities 1.2 924.7
Pilgrim I
228.3
Vermont Yankee 285.4
Drowns Ferry
593.6
Cooper Station
205.5
Total
Correspon~ electricalenergy ((c) y)

0.6
0.2
1.7
0.02
10

1.9
0.6
0.001
2.1

107
6.6
73.4
182
185
257
533
270
153
34.8
0.5
0.1

5.7
17.5
76.4
3.8
50.7
133
95.1
142
200
175
90.8
0.8
0.8
0.001

36.2
1.7
22.6
59.3
83.8
255
216
193
81.4
22.2
4.5
0.01

1 814.9

993.4

980.3

0.02
0.4
0.7
0.002
48
54.8
169
18.2
55.1
61.5
67.6
66.5
223
61
63
3.2
5.8
897.8

0.04
0.1
1.0
0.007
11
30.2
75
4.8
47.3
78.7
111
89.9
210
127
71.2
1.2
1.4
0.2

0.02

1.6
10.2
21.9
1.5
16.l
43
38.3
51.6
98.4
61
36.3
1.7
0.4
0.04

860.0

139.6

14.4

125

0.02

20

8.2
65.6
8.9
13.9

0.005

46.5
25.2

9.8
6.6

6.3
0.2

45.6

383.1

195.5

62.5

13.2

5.0

6.4

1.5
0.05

4 991

5 030

4 981

4 825

·5 030

1 446

5 030

3 388

2 403

I 020

606

I 474

Normalized release(Ci per MW(e) y)

360

200

200

190

170

100

80

60

30

10

8

4

I 400

Proportion of total (%)

26

14

14

13

12

7

5

4

2

0.9

0.6

0.3

100

.....

-..l
-..l

0.06
0.3
1.4
0.004

Sources: References SS, 163.

an AGR due to a coolant leakage of 1 per cent per day
are about 3 Ci per MW(e) y for 41 Ar and 2 Ci per
MW(e) y for 16 N (I 13).

58. The AGR utilizes slightly enriched U0 2 as fueland C0 2 as coolant. The cladding is stainless steel.
Somewhat higher operating temperatures are reached,about 1300°C, but noble gas release is still expected tobe quite low. Commercial AGRs have come into
operation only recently.

63. Another airborne effluent from GCRs is 3 5 S;
arising from (n,-y) reactions with 34 S and (n.p)
reactions with 3 5 CL The impurity levels of sulphur andchlorine in moderator graphite are approximately 50 and
2 ppm, respectively (113). Release of 35 S from the
Oldbury and Wylfa (Magnox) reactors in the United
Kingdom averaged 0.002 Ci per MW(e) y during
1972-1974 (106).

59. The helium-cooled HTGR utilizes graphite-coated,highly enriched U0 2 or UC 2 as fuel. The prototype
HTG~s thus far constructed, AVR (Federal Republic of
Germany), Dragon (United Kingdom) and Peach Bottom(United States), have utilized charcoal filters to retain
the noble gases which reach the primary coolant for long
periods. The release composition is thus primarily 1 3 3 Xeand 8 5 Kr. Noble gas release reported from Peach
Bottom averaged about 5 Ci per MW(e) y for 1968-1973(141, 161). For the AVR reactor, the noble gas releasehas been 2-3 Ci per MW( e) y for 1969-1972 (27). Larger
HTGRs are expected to release noble gases at the rate of
1-3 Ci per MW(e) y (49).

64. Nitrogen-16 (7.3 s) is also formed in the coolant inlight-water reactor cores. In BWRs a portion of the 1 6 N
is transported with the steam before decay to theless-shielded turbines. The photons (6.1 and 7.1 MeV)
from 1 6 N decay in the turbines are a potentiallysignificant component of the gamma radiation field near
the turbine building (86, 87).

60. Table 8 also gives a limited amount of data on the
total activity of noble gases released from GCRs and
HWRs. In some cases, measurements at GCRs did not
distinguish fission noble gases from 4 1 Ar. Therefore, inreports of release data from these reactors, the amountsof fission noble gases are upper limits.

3. Tritium
65. Tritium occurs by ternary fission in nuclear fueland also by neutron activation reactions with lithiumand boron isotopes dissolved in, or in contact with, theprimary coolant and with naturally occurring deuteriumin the primary coolant. Erdman (31) has estimated thenormalized tritium generation rate in LWRs fromternary fission to be about 15 Ci per MW( e) y, based on a
recent measurement of 0.85 10-4 tritons per thermal
fission in 235 U (34), and assuming 2 10-4 tritons per
thermal fission in 239 Pu and 238 U (31). The estimate
assumes 55 per cent fissions in 235 U, 41 per cent in
239
Pu and 4 per cent in 238 U. Other estimates indicate
about 23 Ci per MW(e) y ( 107). An average assumed foruse in this Annex is 20 Ci per MW(e) y. The generation
rate estimate for a fast breeder reactor is also 20 Ci per
MW(e) y (31).

2. Activation gases

61. In contrast to the noble gas fission products
released in BWR and PWR operation, several activation
gases are formed in GCR operation. Direct activation of
the C0 2 coolant gives rise to 16 N by the (n, p) reaction
with 16 0. Argon-41 arises from the (n, 'Y) reaction with40
Ar in the coolant and also in shield-cooling air, where
this is used.
62. The amounts of 4 1 Ar released depend on thereactor characteristics. Release rates of 41 Ar from GCRs
in the United Kingdom are not measured routinely, but
for a Magnox reactor, the rate has been estimated to beapproximately 20 Ci h- 1 from shield cooling and an
additional 10 Ci d- 1 from C0 2 leakage (38). A numberof recent measurements at three reactors in the United
Kingdom indicate 4 1 Ar releases on the order of
100-300 Ci per MW(e) y (115). Estimated discharges for

66. Estimated generation rates and appearance oftritium in effluent streams of reactors are summarized intable 10. Assessments of tritium activity in the coolant
have generally assumed that 1 per cent of the tritium
generated in the fuel is transferred to the coolant throughpinhole defects in the fuel element cladding. In earlier
PWRs, which utilized stainless steel cladding, as much as

TABLE 10. ESTIMATED RATES OF GENERATION OF TRITIUM AND OF ITS APPEARANCE IN EFFLUENT STREAMS OF
DIFFERENT TYPES OF REACTORS(Ci per MW(e) y)
B WR

Generation

Effluent
streamFissionSoluble boron shim
Deuterium activation
Lithium activation
Boron activation(control rods)

20

0.2

0.01

0.01

Total

29

SourceGCR

PWR

Generation

Effluent
stream

20
0.5
0.001
0.02

21

Generation

Effluent
streamGeneration

Effluent
stream

0.2
0.5
0.001
0.02

20

0.2

20

0.2

600

zoa

0.5

0.1

l

21

0.3

620

20

9
0.2

Sources: References 39, 76, 128, 136.
aoepending on the irradiation time and on the net leakage oi heavy water.

178

HWR

--MW(e) y in airborne effluents. For BWRs. ti].e averages
are 0.1 and 0.05 Ci per MW(e) yin liquid and airborne
effluents respectively.

80 per cent of the fission-produced tritium could
permeate the cladding. With the zircaloy cladding incurrent usage, 1 per cent or less of the fission-produced
tritium is expected to appear in the effluent streams(85). Although tritium is formed within the control rodsin BWRs, it has not been shown to diffuse through the
B4 C matrix of the control rods (136).

68. In GCRs, tritium is produced by activation oflithium impurities in the graphite moderator as well as
by ternary fission. Use of helium as coolant in HTGRs
results in tritium generation by the (n, p) reaction with3
He. Estimates of the normalized generation rates are
0.5 Ci per MW(e) y from lithium activation (39) and
3·6 Ci per MW(e) y from helium activation (136).

67. Reports on PWR operation (tables 11 and 12)
show the average normalized release of tritium to beabout 2 Ci per MW(e) yin liquid effluents and 0.2 Ci per

TABLE 11. TRITIUM DISCHARGED IN AIRBORNE EFFLUENTS FROM REACTORS IN
VARIOUS COUNTRIES, 1973-1974

ReactorReleaseNormalized release(Ci)(Ci perMW(e) y)

1973

1974

1973

1974

20.2

11.5
11.1

0.067

0.039
0.018

8.4
25.4
268.9
50.6
1.1
25.5
2.5
0.18
1.89
42.4
4.1
2.0
0.33

3.8
0.32
91.4
0.012
0.37
42.8
51.5
0
7.2
60.4
9.2
19.9
0.75
3.9
878
180
0.030
109
12.7

0.070

0.037
0.00240.26
0.000023
0.0015
0.056
0.094
0
0.017
0.090
0.010
0.053
0.0027
0.024
1.39
0.34
0.00046
0.60
0.053

PWR

Germany, Fed. Rep. of
Obrigheim
Stade
United States
Yankee Rowe
Indian Pt. 1
San Onofre
Connecticut Yankee
R. E. Ginna
Pt. Beach 1,2
H. B. Robinson
Palisades
Maine Yankee
Surry 1,2
Turkey Pt. 2,3
Indian Pt .2
Ft. Calhoun
Prairie Island
Oconee 1.2
Zion 1,2
Arkansas 1
KewauneeThree Mile Island

13.1

Total

466

1494

Electrical energy (MW(e) y)

5 393

8 014

Overall normalized
release (Ci per MW(e) y)

1.02
0.18
0.00280.039
0.00580.00066
0.00490.051 ·
0.00760.050
0.0048
0.049

0.09

0.19

77.1
1.9
50.6

38.7
1.7
18.3
0.42
15.8
114
7.9

1.61
0.040
2.24
0.00078
0.067
0.010
0.0079

1.00
0.040
0.49
0.00098
0.042
0.15
0.019

34.0
14.0
1.0

29.0
8.0
2.2
5.6
0.65
0.016

0.030
0:030
0.00490.031
O.D35
0.00780.00940.00150.000077

Total

217

242

Electrical energy (MW(e) y)

4 340

5 094

BWR

United States
Big Rock Pt.
Humboldt Bay
Lacrosse
Oyster Creek
Nine Mile Pt.
Dresden 2,3
Millstone Pt.
MonticelloQuad Cities 1,2
Pilgrim 1'l{ermont Yankee
Peach Bottom 2,3
Browns Ferry
Cooper Station

Overall normalized
release (Ci per MW(e) y)

0.32
26.8
10.0
1.7

0.050

0.048

179

TABLE 11 (continued)
Nonnalir.ed release(Ci per MW(e) Y)

Release
(Ci)

1974

1973

Reactor

1973

1974

HWR

Argentina
AtuchaCanadaPickering217
36 50024 800Total

36 50025 017

Electrical energy (MW(e) y)

1 630

1 723

Overall normalized
release (Ci per MW(e) y)

2.0

22.4

14.S

30

12

22.4

15.4

0.10
0.67

0.037

GCR

United Kingdom
Oldbury
Wylfa

200

Total

230

12

Electrical energy (MW(e) y)

588

321

Overall normalized
release (Ci per MW(e) y)

0.39

0.04

Sources: References 26, 44, I 06, 161, 163, 171.

69. The release rate of tritium from GCRs varies withreactor type. Penetration of the Magnox cladding bytritium produced in fuel can be assumed to be about
1 per cent. For HTGRs an estimated 30-90 per cent of
the tritium originating in the fuel and graphite
moderator may reach the coolant (39, 136). The tritiumis removed in coolant-gas purification, but a 4 per centdiffusion loss through the condenser walls is anticipated,or about 3 Ci per MW(e) y (39). Based on the experience
with GCRs, primarily for the Magnox reactors in the
United Kingdom, tritium release in liquid effluents(table 12) is about 0.2 Ci per MW(e) y, and in airborne
effluents it ranges from 0.04 to 0.4 Ci per MW(e) y, on
the basis of the limited information available (table 11).
70. The amount of tritium generated in HWR fuel by
ternary fission is approximately the same as in LWRs, but
activation of deuterium in the D 2 0 coolant andmoderator provides much more. Kouts estimates thetritium production rate in the coolant to be 600 Ci per
MW(e) y (76). Tritium concentration in the primarycoolant and moderator of a HWR depends upon theirradiation time of the heavy water and the reactor size.
The D2 0 leakage from HWRs is kept as small as possible
for economical and radiological reasons. Leakage from
the MZFR reactor in the Federal Republic of Germany
averages about 1.5 I d- 1 , about -0.5 per cent of the total
D 2 0 amount per year (39). For larger reactors, annual
losses of from l per cent (Atucha reactor in Argentina
and Pickering reactors in Canada) to 2-3 per cent can be
anticipated. For the optimal loss of 0.5 per cent per
year, the normalized tritium release rate from a HWR
ranges from 3 Ci per MW( e) y in the first year ofoperation to about 20 Ci per MW(e) y in the tenth.
Based on the latter value as representative of the reactor
180

life, the normalized tritium release is estimated to be
16 Ci per MW(e) y in airborne effluents and 4 Ci per
MW(e) yin Uquid effluents (39). For larger leakage rates,on the order of l per cent per year, these values would
be attained after a few years of reactor operation.
Information on actual releases of radioactivity from
HWRs is limited; however, the data reported for the
Pickering reactors in Canada ( tables 11 and 12) roughly
agree with these estimates ( 171 ). For the Atucha reactor
in Argentina, normalized tritium release in liquid
effluents was 0.8 Ci per MW(e) y during its initial year ofoperation (1974) and 3 Ci per MW(e) y in 1975 (26).
Tritium in airborne effluents from Atucha was 2.0 and
3.8 Ci per MW(e) yin 1974 and 1975, respectively (26).

4. Carbon-14
71. Very little attention has been given to 14 C
production in and release from reactors. The dose rates
associated with the releases are very low; however, with
its long half-life (5730 y), 14 C can make a significant
contribution to the collective dose commitment.
72. Carbon-14 is produced in LWRs and HWRs by(n, a) reactions with 1 7 0 present in the oxide fuel andin the moderator, by (n. p) reactions with 14 N present
in the fuel as impurities, and by ternary fission. Ternary
fission production per unit electrical energy generated ispractically independent of reactor design. while thenormalized production of 14 C 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.

-73. In 1974, Bonka estimated the normalized 14 C
production in light water reactors to be about 0.03 Ci
per MW(e) y (18). Magno, also in 1974, estimated the
production to be about 0.05 Ci per MW(e) y (90). Hayes
et al (45) estimated the production to be 0.02 Ci per
MW(e) y for PWRs and 0.04 Ci per MW(e) y for BWRs.
Recent estimates by Davis (29) gives values of-0.02 Ci
per MW(e) yin PWRs and BWRs. A study carried out for
Swedish power reactors (84) estimated the values to be
0.014 Ci per MW(e) y for PWRs and 0.024 Ci per
MW(e) y for BWRs.
74. Assuming, as in the assessments mentioned above,that the average thermal neutron flux density is the samein the fuel and moderator, that the relevant (n, p) and(n, a) reactions follow the 1/v law, as several studies
indicate (42), and that there are 10 ppm nitrogen
impurities in the fuel, production rates of 1 4 C have been
estimated from reactor design information compiled by
IAEA (57). For recent LWRs, the values estimated are,for PWRs, 0.012 Ci per MW(e) yin the fuel and 0.005 Ci
per MW(e) yin the moderator and, for BWRs, 0.016 Ci
per MW(e) y in the fuel and 0.008 Ci per MW(e) yin themoderator.
75. These results indicate the normalized 14 C
production to be about 0.02 Ci per MW(e) y. The amountwhich can be assumed to be released at the reactor,
30. per cent of the total production, is the amountproduced in the moderator, which will vary dependingon the moderator size. The 14 C produced in the fuel will
presumably be released, at least partially, at the
reprocessing plant.
76. From measurements of decay-tank gas andcontainment air at three PWRs, Kunz estimated the 1 4 C
release in airborne effluents to be approximately
0.006 Ci per MW(e) y (81). In subsequent analyses ofoff-gas from the main condenser air ejector of a BWR,
Kunz estimated the 14 C release in airb_orne effluents tobe 0.016 Ci per MW(e) y (82). This result for the
R. E. Ginna reactor in the United States corresponds toa somewhat larger moderator size. From detailed
measurements at the Oyster Creek BWR in the UnitedStates. Blanchard et al ( 15) estimate the 14 C release tobe 0.006 Ci per MW(e) y from the main condenser air
ejector and 0.002 Ci per MW(e) y from the building
ventilation air. Occasional stack monitoring, however,would indicate, if considered representative, a larger
airborne release of 0.018 Ci per MW(e) y. Additional
measurements at Oyster Creek indicated 14 C discharged
in liquid effluents to be 2 10-5 Ci per MW( e) y (15).
77. A 1 4 C airborne normalized release rate of 0.013 Ci
per MW(e) y was reported for the Yankee Rowe PWR in
the United States for 1972 (146). This reactor reported
0.005 Ci per MW(e) y in airborne effluents in 1974 and3 10- 5 Ci per MW(e) y in liquid effluents (163). Someinitial measurements of 14 C airborne release fromreactors operating in the USSR, however, indicate larger
airborne releases, namely 0.2, 0.46 and 0.9 Ci per
MW(e) y for a PWR, BWR and LWGR, respectively(121). More recent results are somewhat lower. These
measurements presumably reflect different reactor
design. moderator volume and nitrogen impurity levels.
78. Of the 14 C in the BWR off-gas sample measured
by Kunz, 95 per cent was in the form of C0 2 , 2.5 per

cent, CO and 2.5 per cent, hydrocarbons. Due to the
reducing atmosphere of the hydrogen cover gas of the
PWR, however, over 80 per cent of the 14 C activity
released from this reactor type was in organic forms CH 4
and C2 H 6 (81, 82).
79. In gas-cooled, graphite moderated reactors the
major source of 14 C production is the graphite
moderator, due to the 13 C (n, r) 14 C reaction and also
to the 14 N (n p) 14 C reaction based on the incorporated
nitrogen impurity. Normalized production rates havebeen estimated to be about 0.18 Ci per MW(e) y in
Magnox reactors and 0.3 Ci per MW(e) y in AGRs (74).
Production of 14 C in the carbon dioxide coolant, mainly
from activation of nitrogen impurities and from the 1 7 0( n, a) 1 4 C reaction, is a smaller source estimated to beabout 3 10-3 Ci per MW(e) y for Magnox reactors and10-2 Ci per MW(e) y for AGRs (74). Carbon-14
discharges from Magnox reactors and AGRs result from
coolant leakage and include 14 C released in the coolant
from corrosion of the moderator; estimated releases are
0.006 Ci per MW(e) y for Magnox reactors and 0.017 Ci
per MW(e) y for AGRs. Assessments of 14 C production
by HTGRs range from 0.012 Ci per MW(e) y (29) to0.25 Ci per MW(e) y (90) and originates primarily fromthe 13 C (n, -y) 14 C reaction. Insignificant releases of 14 Cfrom these reactors are anticipated. Rather, the 14 C
would be released at the reprocessing plant ( 17).
80. The production of 14 C in HWRs is due to thesame type of nuclear reactions described for LWRs
(para. 72). However, due mainly to the large moderator
mass. the production rate of 14 C in HWRs is expected tobe considerably larger than in LWRs. Using theassumptions described in paragraph 74, the production
rate in the Atucha reactor in Argentina is estimated tobe 0.46 Ci per MW(e) y, with 90 per cent generated in
the moderator. The production of 14 C in CANDU
reactors is estimated to be 0.42 Ci per MW(e) y, 95 percent being produced in the moderator (120).
81. Much more data, particularly from continuous
monitoring, must be forthcoming before a reliable
assessment of the 14 C generation rate in reactors and its
release characteristics can be made.

5. Iodine
82. A number of iodine isotopes are produced inreactors by fission and by decay of other fission
products. Release estimates have been published for131
I(8.05d), 132 I(2.3h), 133 I(2Ih), 134 1(53m),
135 I(6.7h) and 129 I(I.610 7 y) (16). Equilibrium
amounts of all but 1 2 9 I are rapidly established in thereactor core, and releases are more a function of theintegrity of the coolant system than of the amount ofelectricity generated. Iodine-131, with a half-life of 8 d,is the main concern from an environmental point ofview. Its activity per unit of thermal power at
equilibrium in uranium fuel varies from about
25 kCi MW- 1 at the beginning of the fuel irradiation toabout 30 kCi Mw- 1 at the end, as a result of the larger
fission yields of plutonium, which contributes increasingly to power production as the bum-up proceeds.
Long-lived 1 2 9 I has not been identified in power reactor
181

TABLE 12. TRITIUM DISCHARGED IN LIQUID EFFLUENTS FROM REACTORS IN VARIOUS COUNTRIES, 1970-1974

....
00

N

ReactorStart-up
year

Net
electricalpower(MW(e))

1970

1971

1972

1973

1967

270

340

706

1 762

1 849

1968
1972
1974

328
630
1 147

1964

247

135

1 117

1 078

1970
1972
1974

340
500
826

32

140

85
150

1973

447

1974

820

Normalized release (Ci per MW(e) y)

Release (Ci)
1974

1974

1970

1971

1972

1973

2.3

3.2

7.2

5.7

1.2

0.89
0.26

1.2
0.24

0.51
0.05
0.08

0.95

7.2

4.8

0.40

0.56

0.67
0.63

1.3

0.69

1.9
0.72
0.49

0.02

0.53

PWR
France
SENA, Chooz
Germany, Fed. Rep. of
Obrigheim
Stade
BiblisItaly
Trina, Vercellese
Japan
Mihama 1
Miliama 2
Takahama 1
Netherlands
Borssele
SwedenRinghals 2
United States
Yankee Rowe
Indian Pt. l
San Onofre
Connecticut Yankee
R. E. Ginna

Point Beach, 1,2
H. B. Robinson
Palisades
Maine Yankee
Surry. 1,2
Turkey Pt. 3 ,4Indian Pt. 2
Ft. Calhoun
Prairie Island, 1,2Oconee 1,2,3Zion 1,2
Arkansas I
KewauneeThree Mile Island

1961
1962
1968
1968
1970
1970/72
1971
1971
1972
1972/73
1972/73
1973
1973
1973/74
1973/74
1973
1974
1974
1974

311

120
190

149
32.2
8.3

47
230
130
171

0.27

6.3
I 500
410
4 800
7 400
110

Total

14 72715 617

I 616

3 124

Electrical energy generated (MW(e) y)

326
112

1.8

175
265
430
575
420
2 X 497700
700
790
2 X 788
2 X 693873
457
2 X 530
3 X 886
2 X I 050
820
520
810

Overall normalized release (Ci per MW(e) y)

243
96.7

9.1

I 680
725
4 570
5 830
154
266
118

5.0

803
574
3 480
5 890
119
563
405
208
9.2
5.0

694
138
4 070
3900
286
556
432
185
154
448
329
27.5
15.8
10·•
70.7

314
684
3 810
2 240
195
833
449
8.1
219
245
580
47.9
124
142
350
2.3
25.6
92.4
130

15 47113 905

11 265

3 960
3.9

6 162

9 069

--2.3

1.2

I. 7
148
14.6
5.9

7.9
215
9.0
3.6

10.3
9.4
13.1
17.4
0.42

9.7
4.7
11.5
11.6
0.47
0.68
0.40

10.9
4.0
10.8
12.0
0.43
1.1

0.81
0.96
0.17
0.12

5.8
15.5
14.1
0.74
0.86
1.0
0.68
0.40
0.54
0.61
0.69
0.23
3.3 10·•
0.26

3.0
5.0
10.8
4.5
0.81
1.1
0.82
0.91
0.53
0.36
0.64
0.13
0.45
0.89
0.55
0.0043
0.39
0.51
0.54

BWR
Germany, Fed. Rep. of
Kahl
Gundremmingen
Ungcn
Wilrgassen
Italy
Garigliano
Japan
Tsuruga

1961
1966
1968
1972

15
237
174
640

1964

152

1969

331

1.4
45.6
31.7
5.0

1.2
78.3
24.0
3.2

5.0

3.0

6.2

5.3

0.11
0.20
0.28
0.059

8.2

21.0

0.38
0.40
0.05

0.038

0.062

0.025

0.020

0.30
0.75
0.13
0.03

0.76
0.98
0.17
0.06

0.029

0.12

Netherlands
Dodewaard
SwedenOskarshamn 1
Ringhals 1
United States
Dresden I
Big Rock Pt.
Humboldt Bay
Lacrosse
Oyster Creek
Nine Mile Pt.
Dresden 2,3
Millstone Pt. I
MonticelloQuad Cities, 1,2
Pilgrim l
Vermont Yankee
Peach Bottom 2,3
Browns Ferry
Cooper Station

1968

52

1971
1974

440
760

1960
1962
1963
1969
1969
1969
1970/71
1970
1971
1972
1972
1972
1973/74
1973
1974

2.4

1.3

3.2

4.9

4.9

5.0
54.0
7.0
20.0
22.0
20.0
31.0

8.7
10.3
1.5
91.4
21.5
12.4
38.5
12.7
0.6

43.3
10.4
13.0
120.0
61.6
27.8
25.9
20.9
0.0001
4.7
4.2
0

18.5
19.7
51.3
103.0
35.9
46.5
25.8
3.7
0
24.5
0.4
0.2

18.8
50.1
31.7
115.0
14.1
18.7
22.6
24.l
0
34.0
10.5
0
10.0
2.8
1.7

Total

198

263

450

513

629

l 284

2 698

4 072

5 388

5 838

Electrical energy generated (MW(e) y)

0.028

0.086

0.12

0.15

0.097

0.11

0.095

0.15
0.071
0.23

10.9
2.0

200
70
69
53
640
625
2 X 800652
545
2 X 800655
514
2 X l OOO
1 050
760

Overall normalized release (Ci per MW(e) y)

0.056

0.031
1.24
0.14
1.31
0.054
0.091
0.22

0.12
0.27
0.19
3.81
0.047
0.036
0.084
0.030
0.0036

0.34
0.24
0.31
4.38
0.12
0.075
0.028
0.058
2 10- 1
0.011
0.045

0.12

0.062

0.12

0.20
0.45
0.11
0.18
0.051
0.049
0.054
0.053
0.60

0.15
0.49
0.28
0.11
0.090
0.17
0.18
0.087
0.63

0.17
1.04
0.096
0.14
0.076
0.14
0.048
0.25
0.17

0.29
0.41
1.07
4.56
0.087
0.12
0.026
0.017
0
0.021
0.000920.00097

0.47
1.31
0.73
3.05
0.033
0.50
0.030
0.059
0
0.037
0.046
0
0.017
0.00640.0083

0.84
1.08
0.11
0.50
0.083
0.63
0.047
0.046
0.39

0.25
0.51
Cl.II
0.16
0.051
0.70
0.12
0.32
0.28

0.11

GCR

Italy
Latina

1963

153

16.7

13.0

16.9

United Kingdom
Berkeley
Bradwell
Hinkley Pt.
Trawsfynydd
DungcnessSizewell
Oldbury
Wylfa
Huntcrston

1962
1962
1965
1965
1965
1966
1967
1971
1964

2 X 138
2 X 125
2 X 230
2 X 195
2 X 205
2 X 2102 X 200
2 X 3222 X 160

60.1
95.3
18.6
67.7
18.6
20.9
17.3
0.55
159.0

43.1
102.0
24.9
41.9
35.5
7.0
64.4
30.2
162.0

44.2
251.0
38.6
46.0
28.9
53.2
15.0
82.7
37.5

Total

475

594

614

l 158

784

2 580
0.18

3 004

2 659

2430
0.48

2 887
0.27

Electrical energy generated (MW(e) y)
Overall normalized release (Ci per MW(e) y)
HWR
Argentina
AtuchaCanadaPickering 1,2,3,4

,_.
00
w

1974

Electrical energy generated (MW(e) y)

0.23

56.7
117.0
39.0
60.0
20.0
253
37.4
134
67.0

0.83

90.3

320

1971/72/73 4 X 508
Total

Overall normalized release (Ci per MW(e) y)

0.20

200
198
30
116
30.5
208
13.6
275
86.7

402
402

986
986

5 130
5 130

14 400--14 490

344

658

l 630

l 723

3.1

8.4

1.2

Sources: References 1, 3, 26, 27, 97, 99, 1OS, 14.J, 143, 146, 161, 163, 171.

1.5

1.5

3.1

9.0

,

environs, and its significance, in terms of possible releaseactivities, is much Jess than that of the other iodine
isotopes. However, it is of greater significance in thereleases from fuel reprocessing plants and because of its
long-term contribution to the collective dose commitment.
83. Iodine measurements at six United States LWRs
indicated that on the average 73 _per cent of the iodine inreactor off-gas is organic, 22 per cent hypoiodous acid,and 5 per cent elemental (I 14). All forms were alsopresent in ventilation exhaust air, though the release ratefrom this source is usually much less than for the reactor
off-gas system. The release of iodine nuclides depends
strongly on the filter systems used in the installation.

84. Table 13 gives the reported iodine isotope releases
in airborne effluents from various reactors. The overall
1 3 1 I effluent average is (2-5) 10-3 Ci per MW(e) y from
BWRs and (5·50) 10- 5 Ci per MW(e) y from PWRs.
These values are not necessarily typical for any one
reactor. The normalized iodine release amounts for theindividual reactors given in table 13 show extremely
wide variability. Unusually high results from one or two
reactors can easily increase the overall effluent average.
The limited data from GCRs and HWRs indicate that the
average airborne releases are comparable to that of
PWRs.

6. Particulates in airborne effluents
85. Most of the fission and activation nuclides canform aerosols, which may be released with the airborne
effluents. Particulates may arise either directly or as
decay products of fission noble gases. Aerosols originate
particularly from primary coolant leaks, but they canalso be generated by working on contaminatedcomponents or cleaning contaminated surfaces. Highefficiency particulate air (HEPA) filters retain all but thevery small fraction of the smallest particles. Released
particulate activities are thus very low. Table 14 lists thereported activities and composition of particulates
released in airborne effluents from reactors in the UnitedStates during 1974 (163). A few additional radionuclides
were less frequently identified (163). The average total
activities per unit of electrical energy generated are
0.5 mCi per MW(e) y from PWRs and 1.1 mCi per
MW(e) y from BWRs. Over 80 per cent of the average
PWR release activity is 88 Rb, daughter of the noble gas
isotope 8 8 Kr. Other small contributors are 1 34 Cs.
137
Cs, 58 Co, 6 °Co and 54 Mn. For the BWRs,
140.Ba + 140 La and 89 Sr each account for about 30 per
cent of the total activity, with 131 I, 5 8 Co, 6 ° Co, 134 Cs
and 137 Cs each contributing 10 per cent or less.

7. Radionuclides in liquid effluents
86. Radionuclides other than tritium in liquid
effluents include a number of fission and activatedcorrosion products. The amount and composition of thedischarge depend strongly on the type of reactor. The
184

available data on discharges during 1974 are listed intable 15. The average normalized releases. based on
operating experience, are approximately 0.008, 0.06 and0.15 Ci per MW(e) y for PWRs, BWRs and GCRs,
respectively. The averages for earlier data (1972 liquid
effluents) are similar for BWRs and GCRs but somewhat
higher (0.03 Ci per MW(e) y) for PWRs (27. 146). The
radionuclides 1 37 Cs and 13 4 Cs account for 30-50 per
cent of the total activity discharged by PWRs and GCRs
and 70 per cent of that from BWRs: 1 31 I and 1 3 3 I makeup 10-40 per cent and 58 Co and 6 °Co about 15 per centof the activity in liquid effluents from LWRs. Dissolved
noble gases are also present in liquid effluents from
LWRs,' and in addition several less commonly identified
radionuclides have been reported (163).
87. The activity of 1 3 7 Cs reported as discharged isslightly greater than that of 134 Cs. For reactors
reporting the release of both caesium isotopes, the134
Cs/ 13 7 Cs activity ratio is 0.6 for the PWRs and 0.7
for the BWRs. For the GCRs in the United Kingdom theratio is 0.2. On the basis of an analysis of several years ofliquid effluent release data, primarily from reactors in
the Federal Republic of Germany, Schiesser estimatesthat typical normalized releases of 1 3 7 Cs and 1 3 4 Csfrom large LWRs and 5.0 ·and 3.1 mCi per MW(e) y,
respectively. and of 8 9 Sr and 9 0 Sr, 0.63 and 0.063 rnCi
per MW(e) y, respectively (125).
88. The iodine isotopes 13 1 I and 1 3 3 I are released inliquid effluents in amounts comparable to the airborne
release from BWRs. Iodine is readily soluble and withlittle hold-up of liquid releases would be expected to beprevalent in liquid effluents. The iodine isotopes are notreported in liquid discharges from GCRs.
89. The activated corrosion products present in liquid
effluents can be expected to vary widely, even among
reactors of the same type. In addition to 5 8 Co and
6°Co, the isotopes 51 Cr and 54 Mn are quite prevalent inliquid effluents from LWRs. The reported activities of99
Mo, 65 Zn and 59 Fe are somewhat less. The isotopes
55
Fe and 63 Ni are not routinely reported in LWR
releases, but Hutchinson et al (56) measured activities
comparable to other corrosion products in liquid-waste
samples. Blanchard et al (15, 72), however, did notdetect 6 3 Ni in liquid effluents from PWRs and BWRs.
The Palisades PWR in the United States reported
relatively high releases of activated corrosion products.
including 6 3 Ni, during 1974, associated with a relatively
long maintenance period (163).
90. Most of the activity in liquid effluents from GCRs
arise from spent-fuel storage pools. Additional sources
are the reactor-vessel cooling system, the coolant·drying
installation and various other low-level sources. Liquid
streams are collected, treated and released with thecondenser coolant water. In GCR liquid effluents, theactivation products include 35 S, 4 5 Ca and 32 P. Activity
released in liquid effluents from gas-cooled Magnox
reactors (table 15) is somewhat higher than from LWRs.
In the United Kingdom all releases are to the sea, except
for Trawsfynydd, from which releases are to a
fresh-water lake. Liquid wastes are filtered and treatedadditionally by filtration in non-regenerable ionexchange beds (60).

---------------------- ·--TABLE 13. IODINE DI~CHARGED IN AIRBORNE EFFLUENTS FROM REACTORS IN VARIOUS COUNTRIES, 1970·1974/odlne 133/odlne-131
Normalized release(10-• Ci per MW{e) y)

Release(Cl)
Reactor

1970

1971

1972

0.063

2 10-•

0.001

1973

1974

1970

lodl11e-135

0

1971

1972

Rt!lease

(Ci)

1973

1974

1.8
9.1

1.7
1.8
0.06

10.4

4.5

1972

1974

Release(Ci)

Normalized release(10-• Cl per MW(e) y)
1972

1974

1972

Nonnallzed release(10·• Ci per MW(e) y)

1974

1972

1974

PWR
Belgium
BR·3, Mo!
France
SENA, Chooz

0.0006

0.015

0.006
0.047

0.001

l 10-•

Netherlands
Borsscle

u,

0.005
0.043

0.005
0.011
6 10-•

0.034

0.00023 0.0027 0.00071
0.058
0.012
0.09
4.4 I o-• 0.42
0.000190.01
0.0013 4.6 10-•0.034
0.00053 0.00028
0.0097 0.011
0.088
0.30
0.046
0.0087 0.29
0.0099
1. 7 10-• 0.0016 0.054
1.2 10-• 0.042
0.10
3.41
O.OS5
0.00029 0.37
4.3 10-• 0.00042
1.3 I o-• 0.00045
0.012
0.032
6.7 10- 1 0.014
0.053
0.024
0.028

15.6

5.8

2.2
12.7

0.42

0.65

0.00040.68
0.00029
0.32
2.25
65.7
0.004540.8
1.2 10-•0.053
0.013 160
9. t 1o-• 6.3 10-•
0.013
2.04
0.47
0.14
0.0015 0.00017 0.54
12.1
0.12
11.5
0.038
1.82
1.69
0.0034
68.2
8.47
0.010
4.0
104
112
13.2
0.0042 0.0019 7.9
0.0032 0.41
0.11
5.04
4.5 10-• 0.020
0.029
15.0
0.034
10.3
378
0.056
0.73
98.0
5.0 10-•0.15
0.063
0.00014
0.28
0.58
4.47
5.01
0.000370.00110.00083 2.65
81.4
13.1
0.00270.00013
1.18

0.28
3.25
0.0034
0.069
5.02
0.63
112
0.46
2.90
3.74
14.9
0.018
0.086
0.059
0.21

0.000390.37

3.1 10-s

0.00880.00095
0.00190.17
21.8

0.0011

1.46
0.052

2.1
0.00420.00520.63
0.57

4.2 10-•1.6 10-•8.1 10-•

0.015
0.010
0.00130.0023
2.0 10- 1

1.25
8.2 10-•

0.02

0.20

1.20

4.37

0.058

0.17

0.011

0.015

Electrical energy
generated (MW(c) y) 438

412

3 133

5 471

8436

863

6 544

53

4 079

Overall normalized
releaseCi perMW(c) y)

3.9

6.3

22

52

0.67

2.6

2.1

0.37

Total

00

9.4

0.0025

United States
Yankee Rowe
Indian Pt. I
San Onofre
Connecticut Yankee
R.E.Ghma
Point Beach 1,2
II. B. Robinson
Palisades
Maine Yankee
Surry 1,2
Turkey Pt. 2,3
Indian Pt. 2
Ft. Calhoun
Prairie Island
Oconee 1,2Zion 1,2
Arkansas I
KewauneeThree Mile Island

-

80

0.023

Germany, Fed. Rep. of
Obrigheim
0.045
Stade
BiblisItaly
Trino, Vercellese

970

0.11

oo-•

25

.....

TABLE 13 (co11ti1med)

00°'

lodine-133

lodine-131
Normalized release(10- 3 Ci per MW(e) y)

Release(Cl)

Rellctor

1970

1971

Germany, Fed. Rep. of
Kahl
0.005
Gundremmingen
0.20
Lingen0.26
Wiirgasscn

0.0025
0.34
0.38

1972

1973

1974

1970

1971

1972

0.20
0.15
< 10-4

0.002
1.96
0.016
< 10- 4

0.003
0.12
0.002
< 0.0007

0.38
0.94
2.3

0.19
1.5
3.3

0.71

0.98

1.2

1973

Release(Ci)
1974

lodine-135
Release(Cl)

Normalized release(10- 3 Ci per MW(e) YI

1972

1974

1972

1974

0.017
1. 7
0.47
7.04
0.55

4.41
0.26
1.19
0.019
10.5
1.66
14.7

0.39
41.1
17.2
14.1
1.5

110
6.8
27.4
0.49
24.9
4.4
19.5

1.16
0.60
0.92
0.16

11.0
16.0
1.12
0.12

2.9
1.4
9.9
3.3

33.2
17.3
4.9
0.42

1972

Normalized release(JO"' Ci per MW(e) y)

1974

1972

1974

B WR

Italy
Garigliano
Japan
Tsuruga
Fukushima 1,2

0.94
2.5
< 0.002

0.32
0.29
9.9
0.53
0.15
0.038
< 0.0004 < 0.01

0.06

0.13

0.06

0.38

1.06
0.06

0.24
0.25

0.20
0.79

0.28
0.34

1.4

4.2
0.19

0.92
0.84

0.70
3.5

1.6
0.69

0.00050.009

0.00150.001

0.0095

0.012

0.020

0.040

0.024

0.03

0.04
4.6
0.17
0.18
6.7
1.96
0.49
0.15
1.2
5.5
0.46
0.06

0.41
0.09
0.50
0.033
2.31
0.72
3.98
3.18
5.63
8.82
1.44
0.35
0.00660.38
0.13

Netherlamls

Dodewaard
United Slates
Dresden I
Big Rock Pt.
Humboldt Bay
Lacrosse
Oyster Creek
Nine Mile Pt.
Dresden 2,3
Millstone IMonticelloQuad Cities 1,2
Pilgrim I
Vermont Yankee
Peach Bottom 2,3
Browns Ferry 1
Cooper Station
Total

0.12
0.40
0.69
6.26
0.89
1.23
0.58
0.73
0.03
0.17

0.91

Electrical energy
generated (MW(e) y) 747
Overall normalized
release(mCi per MW(c) y)
1.2

2.8
9.7
25.3
12.6
2.4
0.99
3.4
1.4
1.6
0.3
3.5

1.1
96.0
3.5
8.0
16.0
4.9

0.70
3.2
4.7
0.97
0.29

10.3
2.3
11.6
0.87
7.8
1.9
5.3
7.7
17.2
9.5
6.3
1.2
0.11
0.086
0.062

0.045
0.086

220
0.016
0.28

8.66
0.00063
0.012
0.012
5.82
0.28
2.26
0.070

0.10
0.42

9.54
0.17
19.3
12.0
11.8
1.87
0.41
3.44

1.98

12.0

24.5

29.7

12.6

61.1

8.4

67.2

1 094

3 312

5 509

6 134

1 972

4 129

1 746

3 806

1.8

3.6

4.4

4.8

6.4

15.0

4.8

18.0

0.44
11.7
0.76
5.6
0.16

22.6
0.45
25.6
36.2
12.8
8.2
0.92
16.7

---

- • -

•~I---•- - - ·... - - - - - . ~ - - - - · - - - · - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -lodine•l 33

lodine•l31(Ci)

Reactor

1970

1971

1972

Release(Cl)

Normalized release(10-• Ci per MW(e) y)

Release
1973

1974

1970

1971

1972

1973

1974

GCR
FranceChinonSt. Laurent
Bugey
Total

Electrical energy
generated (MW(e) y)
Overall normalized
releaseoo-• Ci per MW(e) y)

0.027
0.065
0.0001

5.4
9.9
0.08

0.09

I 285
7.2

HWR
Argentina
Atucha9.1

0.0098CanadaPickering

0.13

0.00550.0040

Total

0.13

0.00550.0138

Electrical energyb,encrated (MW(e) y)

658

I 630

I 723

Overall .normalized
release(I o-• Cl per MW(e) y)

20.0

0.34

0.80

Sources: References 1, 26, 27, 43, 44, 99,146,161,163,171.

20.0

0.34

0.25

1972

1974

Jodine-135

Normalized release(UT 5 Cl per MW(e) y)
1972

1974

Release
{Cl)
1972

1974

Normalized release(lo-• Ci per MW{e) y)
1972

1974

TABLE 14. PARTICULATES IN AIRBORNE EFFLUENTS FROM UNITED STATES REACTORS

00
00

Electrical e11ergy
ge11erated
ill 1974(MW(e) y)

ReactorAmou11t released (met) during 1974
11

Rb

PWR
Yankee Rowe
Indian Pt. 1
Connecticut Yankee
R. E. Ginna
Pt. Beach 1,2
H. B. Robinson
Palisades
Maine Yankee
Surry 1,2
Turkey Pt. 3,4
Indian Pt. 2
Ft. Calhoun
Oconee 1,2,3Zion 1,2
Arkansas l
KawauneeThree Mile Island

103.9
137.0
502.3
239.7
764.8
547.9
8.9
411.0
673.S
901.8
376.7
274.0
632.4
536.S
65.1
182.6
239.7

542
0.16
0.92

170

Cs

"'M11

nco

•

co

,,,.Cs

"Fe

"'Sr

0.016
7.26

0.41
0.84

0.95
0.85

2.18
7.05

0.0081
2.74

0.047

0.74
0.052

0.020
0.23

6.3 I o-•0.0023
28.0
0.70
2.0
0.41
1.16
17.7
41.6
0.023
0.070
0.27

0.012
37.2
0.36
0.58
0.19
1.16
15.8
7.06

0.00492.5 I0-4

0.066
0.042
9.3 10-•0.0001
52
0.43

0.010

s.s 10-•

'

37

0.12
0.016
1.71
88
0.53

0.024
0.60
37.4
2.91
7.67

0.032
0.040
136

0.37
0.0023849--98.049.9

Corresponding electrical energy (MW(e) y)

I 963

4 786

Normalized release (µCi per MW(e) y)

430

21.0

140

80

74.4

Total

BWR
Dresden I
Big Rock Pt.
Humboldt Bay
Lacrosse
Oyster Creek
Nine Mile Pt.
Dresden 2,3
Millstone Pt.
Quad Cities 1,2
Pilgrim I
Vermont Yankee
Browns Ferry
Cooper Station

40.0
38.8
43.4
37.7
422.4
376.7
753.4
411.0
924.7
228.3
285.4
593.6
205.S

Total

---

126
1.2
58.1
886
6.9
4,7
1.63
35.9
195
I 390

0

55.3

2 989

6 030

4 798

4 033

17.0

16.0

15.0

14.0

••sr

"' I

••co

134

78.1

32.9

45.0
5.23
0.35

33.4
334
29.6
34.0

0.16
0.023

8.4
7.54
267
12.9
3.5
1.38
0.031

---

I 337

464

351

uoBa

••sr

0.5

0.52

0.11
0.023

7.8 10-•0.091
0.0046

0.018
3.08

0.092
I.OS
3. 7 10-•0.012
0.00050.00250.11
0.40

0.0039

0.011
0.042

0.012
0.27
3.4 10-•1.2 10-•
0.034

0.14

94.1

202
4.7
47.4
3.65
991
3.35
5.83
0.89

"er

0.14

--71.6---''"c:s

Cs

"er

1.59
0.27
1.2
1.7
3.66
15.6
2.76
3.44
0.15
l.l
38.4
28.8

7.69
1.S

4.58
49.2
2.68
6.1
0.014

3.4 10-•

--

1.9

0.68

0.51

I 655

3 267

1 580

1.9

0.60

0.43

'"C:s

••co

3.6
7.79
4.47
1.2
2.6
7.41
33.2
3.67
4.9
0.44
0.78
9.15

s.s

--3.1

54

M11

0.78

0.46

l 387

2 460

3 202

0.37

0.32

0.14

0

59

'

Sr

Fe

3.3
1.7
9.3
6.13
0.71
0.99
0.0077

0.17
l.S
1.2
4.61
11.0
0.41
0.095
0.0055

••z,.

0.030
0.012

0.040
0.14
0.20
0.18

0.056

0.63
0.0056
0.098

0.046
0.21
0.052
0.089

--

--

---

--

--

---

---

--

71.8

79.2

56.4

23.9

20.0

I. I

98.7

7.1 10-•

--

1.0
1.29
0.47

1.6
0.16
47.4
0.79
0.65
0.31

--

0.63

Corresponding electrical energy (MW(c) y)

3 534

4 037

2 552

3 524

4 321

3 155

4 155

3 157

3 484

4 037

2 363

2 308

Normalized release (µCi per MW(e) y)

390

330

180

100.0

23.0

23.0

19.0

18.0

6.9

s.o

0.47

0.27

Source: Reference 163.

TABLE IS. RADIONlJCLIDES OTIIER TIIAN TRITIUM DISCHARGED IN LIQUID EFFLUENTS FROM VARIOUS REACTORS DURING 1974
Electricalenergygmuated(MW(e}y)ReactorAmowrt di.fcharged (Cl)

'" /

'"Cs

0.083
0.00070.00110.61
0.041

0.00060.40
0.047
9 10·•
0.00010.018
0.82
0.00560.0032
0.45

IJI /

,,.Cs

11

Co

••co

"er

••M,r

124

Sb

iuce

ao,Ru

11omAg

0.40
0.013

0.64
0.013
0.040

0.27
0.015
0.00320.019
0.00460.073
0.00430.0034

0.018
0.020
0.27

0.00250.00100.027

0.00200.20

0.0003
0.16
0.070
0.68
0.032
0.00020.16
0.43

5 10-•

0.00030.0011
3 10·•

0.0001
0.12
0.35
0.42
0.014
0.0013
0.22
0.70
0.27
1.93
0.16
1.34
0.14
0.00030.47
0.0008
0.033
0.29
0.018

19

Sr

'" I

59

Fe

"Nb

"zr

0.00290.001 l

,.Mo

90

Sr

PWR
Germany, Fed. Rep. of
Obrighcim
293.5
Stade
608.2
Diblis100.8
United States
Yankee Rowe
Indian Pt. I
San Onofre
Conn. Yankee
Ginna
Pt. Beach 1,2
H. B. Robinson
Palisades
Maine Yankee
Surry 1,2
Turkey Pt. 3,4
Indian Pt. 2
Ft. Calhoun
Prairie Island
Oconee 1,2,3Zion 1,2
Arkansas I
KewauneeThree Mile Island

103.9
137.0
353.9
502.3
239.7
764'.8
547.9
8.9
411.0
673.S
901.8
376.7
274.0
160.0
632.4
536.5
65.1
182.6
239.7

Total
Corresponding electricalenergy (MW(e) y)
Normalized release(mCi per MW(e) y)

I 1.0
1.07
2.18

0.27
0.039
0.098
0.15
1.48
0.99
6.83
0.068
0.24

5 10·•

0.029
0.0005
0.14
4.67
0.099

0.0030
0.013
0.038
0.14
0.70
1.49
3.73
0.036
0.056

0.017

0.011
0.00450.0034
2 10-•
0.022
0.29
0.00250.36
0.0082
0.062
0.011

0.013

0.019

0.00770.017

0.00770.00020.00050.00060.017
0.00010.00910.00450.00530.00210.00890.00020.0090

1.00
0.029
0.14
0.00294 10·•
0.028

0.0081
0.014
0.021

0.017
0.28
0.015
8 10·•

0.048
0.00400.00050.00020.00030.00620.00200.032

0.00030.020
0.00940.00194 10-•

0.00120.0001
2 10·•
0.00120.00070.00260.012

0.00020.00110.00280.0051

I 10·•
0.00150.00020.0003
7 10·•

0.00210.00160.00303 10··

0.00020.00010.00290.00200.00130.017

0.0001
6 10"'

0.00180.00010.00030.00240.00020.00170.0003

2 10·•
0.0007
4 10·•

0.00250.01350.01650.00880.0144

3 206

2 262

4 054

2 570

6 484

0.013

0.013

0.00600.00410.00340.0022

tos,

s•Fe

9Sz,

141

Ce

,,omAg0.00100.052
3 10· 1
2 10·•

0.0015I I o·•
0.00040.00060.00300.011

0.00410.00040.00090.00770.029
0.00230.00410.00800.019
0.021
0.00920.00020.00720.00350.00030.0027

15.78

11.34

5.05

6.85

7.17

3.03

0.803

0.874

0.354

0.304

0.05310.04580.08350.03700.0437

6 733

6 689

4 018

6 624

8 114

7 167

4 440

7 281

3 621

3 714

2 180

2 551

6 073

2 836

2.3

I. 7

1.3

1.0

0.88

0.42

0.18

0.12

0.098

0.082

0.024

0.018

0.014

,11c.r,,.C.t

60

,,. ,

••M,r

"Co

•• Sr

"Cr

.... Ce

Uzn

14

99

I JJ/

0.00170.12
0.00440.00050.00020.059
0.000120.00050.00280.040
0.00570.36

0.00020.11

0.00980.057

0.0007

6 10·•

0.014

0.16

2 10·•
0.53
0.00210.00280.72

0.034
0.00520.034

0.0060.

0.00980.00316.81
0.0092
0.015
0.23
0.50

0.074
0.013
0.012
0.85
0.62
1.62

0.061
0.11
0.27
0.059

0.43

0.24

1.10

2 10·•

Na

8 10-•

0.016

0.092

6 10"'

0.00760.0064
8 10-•

I 10·•

0.060

0.0015

7 10·•
0.00080.00440.00020.00090.00320.001
0.00049 JO"

0.0077
0.018

0.11

Co

0.040

0.19
0.51

4 10··

0.00560.001 I
0.0001Mo

B WR
Germany, Fed. Rep. of
Kahl
10.4
Gundremmingen
219.1
Lingcn54.9
Wiirgassen
55.1
United States
Dresden I
Big Rock Pt.
llumboldt Bay00\0

Lacrosse
Oyster Creek
Nine Mile Pt.
Dresden 2,3
Millstone Pt.
MonticelloQuad Cities I , 2

40.0
38.8
43.3
37.7
422.4
376.7
753.4
411.0
331.1
924.7

1.37
0.67
0.18
0.075
2.10
1.37
2.72
1.51
0.026
0.015
9.78
4.64
6.79
2.77
95.2
72.1
no liquid releases
6.00
2.95

0.28
0.53
0.074
5.16
12.3
9.67

0.014
1.01
0.55
13.4

0.036
0.049
0.23
0.079
0.044
2.46
0.72
6.18

3.01

2.09

0.3R

0.10

0.33

0.12
0.00310.023
0.00020.0015

0.021
0.14
0.096
0.0020
0.17

0.65

0.1 I
0.36
0.00160.016
0.48
0.00320.17
0.12

0.00870.00190.00220.010

0.018
0.00120.0062
0.12
0.028
0.21

0.012
0.010

0.045

0.00020,130.00088 10·•

2 10·•TAIILE 15 (co11tim1ed/

.....

'°
0

Electricalenergygenerated
Rea,:tor(MW(e) y)

A11101111t discharged (Ci)il'C.r

,,.c,

••co

131 /

54

1.58

0.50

0.96

0.17
O.IO

0.22

0.18
0.00750.077

125.8786.90

Corresponding electricalenergy (MW(e) y)

5 032

Normalized release(mCi per MW(c) y)

51

Co

"Sr

"Cr

144Ce

6SZtr

0.28

0.30

0.016

0.12

0.0008

0.056

0.014
0.045
0.57

0.00520.00280.014
0.00520.00240.0098

0.091
0.069
0.72

33.21

17.44

10.50

9.13

4.00

2.70

5 032

5 435

4 884

5 409

5 370

4 993

25.0

17.3

6.1

3.6

1.9

1.7

1ncs

"s

034

go Sr

••y

224.9
204.3
350.5
369.9
388.l
360.7
320.8
424.7
243.2

11.3
27.5
61.8
2.20
43.0
6.49
12.6
0.036
32.5

3.26
7.11
4.13
11.1
6.07
4.69
6.68
0.26
8.58

2.19
3.33
5.63
0.32
13.0
1.10
2.28
0.008
11.4

1.99
16.6
6.25
0.17
1.45
0.70
4.24
0.024
1.94

Total

197.3

51.84

39.24

2 887

2 887

68.3

Mn

,.,Na

t9Mo

t'l3/

90Sr

59

0.010

0.00110.00270.0051

9Szr

l"lce

110mAg0.023

0.0010

7 10-•

0.00020.00090.016
0.00060.00870.00340.00200.0038Fc

United States(co11ti1111ed)

Pilgrim I
Vermont Yankee
llrowns Ferry I
Peach Bottom 2,3
Cooper Station

228.3
285.4
593.6
441.8
205.5

no liquid releases

0.30
0.0030
0.020

0.017
0.21

0.775

0.994

0.399

0.497

0.528

0.490

0.212

0.05100.01940.0121

4 656

2 283

4 239

2029

3 04S

3 444

4 993

4 629

4 244

2 124

2044

0.80

0.58

0.34

0.23

0.20

0.16

0.15

0.098

0.046

0.012

0.00910.0059'"Sb

141P,n55

106Ru•••n,,

12smTe

"ea

J44Ce

1.99
16.6
6.25
0.17
1.45
0.70
4.24
0.024
1.94

0.092
0.36
16.0
3.14
0.069
0.048
0.16
0.001

0.25
4.50
4.75
0.095
0.14
0.095
0.098
0.008

0.12
0.092
0.092
0.023
4.86
I. 71
1.71
O.D90
4.00
4.00
4.13
2.13
0.30
0.019
0.30
0.76
l.86
< 0.035 < 0.035 < 0.035
0.048
0.59
0.048
0.016
0.20
0.033
0.033
0.033
0.004
0.056
0.004
0.001

0.85
0.63
0.25
0.019
0.62
0.52
0.52
0.001
0.41

0.069
1.08
1.25
0.019
0.069
0.016
0.033
0.004

33.32

33.32

19.87

9.94

9.82

6.19

6.19

5.05

3.83

2.54

2 887

2 887

2 887

2 644

2 644

2644

2 664

2 664

2664

2 887

2 644

18.0

13.6

11.5

11.5

7.5

3.8

3.7

2.3

2.3

1.9

1.3

0.96

'""Pr

'llp

••co

••sr

••o-

,,.. Sb"Y""'E11

95

•• C

uz"

s<Mn

..Fe

224.9
204.3
350.5
369.9
388.l
360.7
320.8
424.7
243.2

0.069
1.08
1.25
0.019
0.069
0.016
0.033
0.004

0.069
0.72
0.25
0.25
0.21
0.43
0.13
< 0.0003
0.1B

Total

2.54

Corresponding clectricnl
energy (MW(c) y)
Normalized release(mCi per MW(e) y)

Total

0.00110.00190.0081Cs

Fe

0.00180.010

GCR
United Kingdom
Berkeley
Bradwell
Hinkley rt.
Trawsfynydd
DuneenessSizewell
Oldbury
Wylfa
l luntcrston

Corresponding electricalenergy (MW(c) y)
Normalized release(mCi per MW(c) y)

Berkeley
Bradwellllinkley Pt.
Trawsfynydd
Duni:encss
Sizewell
Oldbury
Wylfa
Huntcrston

N/J

0.046 < 0.012< 0.0450.18
0.13< 0.063< 0.0095 0.019< 0.035 < 0.035< 0.0080 0.016
0.065 <0.016
0.001
0.004

<0.012 < 0.012<0.0450.09< 0.063 <0.063< 0.0095 0.019< 0.035 < 0.035
0.016
0.016
<0.016
0.033
0.004
0.005

0.28
0.023
0.16
<0.045 <0.045
1.53
0.13
0.38
1.00
0.038 < 0.0095 < 0.0095
0.069
0.21
0.35
0.048
0.080
0.11
0.033
0.13
0.85
0.008
0.016
0.001

0.069
0.046 < 0.012
0.023<0.045< 0.045
0.18
0.27
0.13
0.25
0.88
0.13
0.057 < 0.0095 < 0.0095 0.019
0.069
0.35
< 0.035 <0.035
0.064
0.032 < 0.008
0.016
0.13
0.033 <0.016
0.033
0.004
0.001 < 0.0003 0.004

2.23

1.87

1.79

1.76

1.10

0.987

0.395

0.289

0.237

0.219

0.110

0.725

2 644

2 887

2 644

2 644

2 644

2 644

2 644

2 644

2 644

2 644

2 644

2 644

2 644

0.96

0.77

0.71

0.68

0.67

0.42

0.37

0.15

0.11

0.090

0.083

0.042

0.027

Sourcr.s: References 3, 106, 163.

rI

effective stack height and average weather conditions.
Assuming a population density of 100 km -l in theregion from 1 to 100 km and 25 km- 2 at greater
distances, the corresponding collective doses per unit
activity discharge have been calculated (table 16). Whenthese results are applied to actual or generalized released
amounts. the total dose and the relative contributions of
the various nuclides can be obtained. The specific noble
gas release composition can vary substantially, depending on how the waste gas is processed.

B. LOCAL AND REGIONAL COLLECTIVE
DOSE COMMITMENT

I
I

91. Routine environmental monitoring is conducted toa greater or lesser extent around all reactor installations.
More detailed, investigative studies to characterize
reactor environmental contamination caused by releases
have been conducted around specific types of reactors( 15, 70, 71. 72). Except in the immediate vicinity of thereactor, the levels of radioactive contamination are notusually readily detectable. Dose assessments must
therefore rely primarily on calculations of theenvironmental dispersion of the released radioactivity.
Generalized calculations can be expected to indicate
roughly the relative significance of the release amounts
and of the exposure pathways. More specific dose
assessments at a given site may require a different
selection of model parameters to fit the local
conditions.

94. Using the average release composition based on
operating experience (table 9), the examples of table 17
have been calculated. In the computation for the BWR
given in table 17, the gonad doses at 1 and 10 km are
due primarily to 88 Kr, 87 Kr, 138 Xe and 135 Xe. The
88 .Kr daughter, 88 Rb, although largely prevented frombeing released by the particle filters. builds up as theparent decays and is an important dose contributor at
10 km. The same holds for the 138 Xe daughter, 138 Cs.
At I 00 km, 13 5 Xe (9.2 h) becomes the dominant
contributor. The fractional contribution from therelatively long-lived 1 3 3 Xe ( 5.3 d) increases with
distance. At 1000 km, only 13 5 Xe and 133 Xe makesignificant dose contributions. Estimates of the corresponding collective doses per unit of noble gas activity
released are 4.5 10-4 man rad Ci- 1 within 100 km. andan additional 0.4 10-4 man rad Ci- 1 in the region from
100-1000 km. For the average BWR total normalized
release based on operating experience. 1100 Ci per
MW(e) y (table 8), the local collective dose would be
0.5 man rad per MW( e) y ( to I 00 km), and the
additional regional contribution would be 0.05 man rad
per MW\e) y (100-1500 km), including the contribution
from 1 3 Xe beyond 1OOO km.

I. Fission noble gases
92. The noble gases released from reactors usually
make the largest contribution to the dose to the local
population. Considerable effort has been devoted to thedevelopment and application of calculational methods toestimate the airborne dispersion (23, 96), and in a few
cases careful attempts to measure the actual exposure
from the noble gas plume have been successful (9, 37,98).
93. Bryant has calculated the gonad dose per unit
activity discharged that would be received by the
population at distances from 1 to 1000 km for each ofnoble gas nuclides (23). The calculations assume a 30 m

TABLE 16. GONADAL DOSE FACTORS FOR NOBLE GASES RELEASED IN AIRBORNE EFFLUENTSa
Collective dose per unit
release (10" 3 man rad Ci" 1)b

Dott per unit releaseoo·• rad cr 1
1Radionuclide

Half.Jife

1 km

10km

umKr
asmKr

1.86 h4.4 h10.76 y76m
2.8h
17.8 m3.2 m
15.4 m11.8 d2.26 d5.27 d5.27 d
15.6 m
9.2h
3.9m
17m
32.2m
1.83 h

3.2
18.4
0.18
66.6
101.0
8.2
76.l
23.8
13.3
22.5
0.0014
2.3
38.1
25.4
4.8
72.9
12.0
85.6

0.16
0.92
0.012
2.2
5.4
1.7
0.0032
0.27
0.76
1.3
0.00079
0.14°
0.24

Ul(r
87

Kr
aaKr
aaRbc
89Kr
89 Rbd131 mxe
133mxe133xee

u3Xe
usmxe
usxe
'"Xe
lllxe
u•csK
41 Ar

100 km0.010
0.000760.00160.025
0.00791000 km

0.000048

0.044
0.067
0.000430.00650.00230.00110.000120.00026

1.3

0.032

0.000290.00041
0.70
1.3
3.2

0.000070
0.0057

1-10 km10-100 km

0.014
0.069
0.00080
0.20
OAO
0.095

0.015
0.18
0.0050
0.20
0.78
0.29

0.055
0.092
0.000028
0.0094
0.052
0.097

0.31
0.47
0.00160.050
0.0051
0.34

0.16
0.098
0.26

0.018
0.066
0.22

100-1000 km

0.00780.0080
0.012

0.41
0.46
0.014
0.0541
0.11

0.0023

Source: Reference 23.
aEffective stack height of 30 m; average weather conditions.
b Assumed 100 man km-• in the region 1-100 km and 25 man km -l at distances> 100 km.
Cfer unit activity of • 8 Kr released.
dfer unit activity of•• Kr released.
eper unit activity of 1 33 mxe released.
!The extrapolated collective dose in the region 1000-1500 km, per unit activity of 1 3 3 Xe released, is 0.017 10 _, man rad Ci " 1 •6Per unit activity of 1 3 8 Xe released.

191

TABLE 17. GONADAL DOSE FROM NOBLE GASES RELEASED IN AIRBORNE EFFLUENTSFROM REACTORSa
Collective dose per unit release(lo-• man rad a- 1)

Dose per unit release(JO-' rad a- 1)
Type of reactor

J km

10km

100km

1 OOO km

BWR
PWR

44
5.7
86

2.0
0.26
3.2

0.015
0.00790.00570.000120.00027

GCR

Relative contribution to doseRadionuclide
BWR
135 XeuKr
IBRbb
133 Xecuaxe
1:,acsb
B7Kr
••mKr
umKr
l35mXel5Kr
131 Xe131mxe133xebFractionof totalrelease
(%)

26
14
14
13
12
7
5

4
2
0.9
0.6

(%)

JOO km

0.16
0.020
0.26

0.041
0.059
0.0023

0.29
O.Q70
0.22

Relative contribution tocollective dose (%)
10100km100l 000km

30
37
14
2.4
0.8
2.9
8.2
0.4
3.1
0.07
0.03

70
4.1

0.8

16
35
8.3
0.8
13
7.9
15
0.6
2.2
1.3
0.01

5.5

0.3

LO
0.003

6.7
0.2

42
15
8.0
0.04
2.8
18
2.1
2.1
2.3

64
14
0.3
0.07
4.4
10
0.06
1.5
3.3
0.01
1.1
0.9

82
5.6

100

100

10km

15
32
2.6
0.7
21
3.5

17
55
39
23
12
7.4
6.1
1
4.7
0.06
8.7
14
1.3
0.6
3.3
2.4
0.5
0.01
0.1
0.000227
0.4
0.0003 0.02

63

48
15
2.7
0.05
2.9
19
0.7
2.1
2.5
0.002
3.4
3.7

74
12

86
3.2

0.1
5.6
2.8

0.2
8.5

100

100

0.5
2.1
3.5
0.008
0.1
0.3

1001 OOO km

1-10 km

1 000kmJ km

18101-10 km 100km

30

0.6

18

1.0
0.4

PWR
133 XeC
nsxe
""Xe
··Kr
"'mXe

.. Kr
nsmxe
asmKr
133mxe'''Xeb
17Kr

41Ar

90
3
1
1
1
0.9
0.8
0.6
0.5

37
13
13
0.03
2.4
16
5.4
2.0
2.0

0.4
0.3

4.7
4.5

0.8
4.2
0.03
0.08
0.2

2.0
0.2

4.0
3.9

0.1
6.9
0.2
0.08
3.9
0.1
0.01

GCR
41 ...\r

100

100

100

100

aThe dose factors of table 16 were used; therefore, the results apply to a 30-m effective stackheight and average weather conditions.
bFormed after release and decay of parent radionuclide.
CThe extrapolated collective dose in the region 1 000-1 500 km due to 133 Xe is 0.0024 10-•and 0.015 10- 3 man rad Ci" 1 for the BWR and PWR, respectively.

95. Using reported noble gas release data and actual
average local meteorology and population factors, Martin
has computed the collective dose to the population within
80 km of several BWRs in the United States (96). The
results for 1971 ranged from 0.04 to 4 .0 man rad per
MW(e) y for eight reactor sites. The average noble gas
release was 1718 Ci perMW(e) y. The collective dose was
900 man rad from 1.85 GW(e) y of electrical energy
generated. corresponding to 0.5 man rad per MW( e) y.
Approximately similar results were obtained for 11 BWR
sites in 1972 (94). Preliminary results for 1973 indicated aprobable slight reduction in the collective dose due to thelonger hold-up times of the newer BWRs (95). The results
192

of Martin, which are based on actual population andmeteorology data, are in good agreement with theestimates of the previous paragraph.
96. Attempts to evaluate the absorbed dose rates in air
from released noble gas by actual measurements within a
few kilometres of BWRs have met with success (9, 98).
Sensitive ionization chambers clearly register theoverhead passage of the plume; however. some care isneeded to distinguish similar abrupt changes in the
background exposure rate due to changes in theradon-daughter contribution during and shortly after arainfall.

97.. Gogolak has used the results of a monitoring
network of continuously operating ionization chambers
to evaluate model dispersion calculations (37). Seven
ionization chambers were placed at distances of
1.3-8 km from a BWR located at a sea-coast site. In themost prominent wind direction, the absorbed dose in air
per unit of activity released, measured over a 200-d
period, averaged 4.2 10-9 rad Ci- 1 at 2.6 km and1.5 10-9 rad Ci- 1 at 8 km. The computed absorbed
doses in air, with careful correlation of stack-effluent
release monitoring and local meteorology, agreed very
well with the measurements. For the actual population
around the reactor, Gogolak determined the whole-bodycollective dose per unit of electrical energy generated tobe 0.02 man rad per MW(e) y from O to 16 km andestimated it to be 0.05 from O to 80 km and 0.09 man
rad per MW(e) y overall. The estimates include a shielding
and occupancy factor of 0.5 and a tissue screening andabsorption factor of 0.8. The collective dose estimates ofMartin and the example in table 17 have not beencorrected for shielding by buildings.
98. The main contributor to the collective dose fromnoble gases released from a PWR (table 17) is expected
to be primarily 133 Xe, due to the long hold-up of the
primary gaseous-waste processing system, but leakages
also result in some release of shorter-lived noble gas
nuclides. The collective dose appears to be primarily due
to 13 3 Xe, 135 Xe and 88 Kr, with 13 3 Xe becoming
increasingly a more important contributor to the dose at
greater distances. Small releases of the longer-lived
133 mxe (2.3 d) and 133 mxe (12.0 d) occur, but theircontributions to the collective dose remain small out to1000 km. The contribution from 8 5 Kr ( 10.7 6 y) to thecollective dose in the local and regional areas isinsignificant.
99. The collective dose per unit of noble gas activity
released by PWRs is 0.9 10-4 man rad Ci- 1 within
100 km and 0.6 10-4 man rad Ci- 1 from 100 to1000 km (table 17). For the average PWR normalized
release based on operating experience, about 15 Ci per
MW(e) y (table 8), the local collective dose would be
0.0014 man rad per MW(e) y (to 100 km). with anadditional regional contribution of 0.0011 man rad per
MW(e) y (100-1500 km), including the extrapolated
contribution from 1 3 3 Xe beyond 1OOO km. The total ofthese two contributions is 0.0025 man rad per MW(e) y.
Using actual population data and reported releases for
two United States PWRs for 1972, Martin computed thecollective dose to the population within 80 km of thereactors to be 0.005 man rad per MW(e) y (94).

2. Activation gasesiII100. Due to the short half-life of 41 Ar ( 1.83 h), thedose decreases quite rapidly with distance. A calculationof the individual gonad dose and collective dose is given
in table 17 for the GCR. assuming the noble gas release
to be lOO per cent 41 Ar. For a population density of
100 km - 2 , the collective dose per unit activity released
to 100 km is 4.8 10-4 man rad Ci- 1 . For an 41 Ar
release from GCRs of 200 Ci per MW( e) y, the collective
dose is estimated to be 0.1 man rad per MW(e) y. Clarke
has calculated collective doses from 4 1 Ar released from

Magnox reactors in the United Kingdom (24). His resultsare (4.8-30) 10·4 man rad Ci- 1 , depending on thereactor site, corresponding on the average to 0.13 man
rad per MW(e) y. Reductions by factors of two to threewould be appropriate to account for indoor shielding.
Nagel has discussed initial attempts to compare
calculations with measurements of the gamma dose from41
Ar (103).
101. The behaviour in the environment of 3 5 S released
from GCRs is not well understood. Ingestion of milk isthought to be the significant pathway to man; there are.
however, uncertainties in estimating the deposition
velocity and the transfer to milk. Passant has given a
value of 0.023 rad per curie released as a rough estimateof the dose to bone marrow per unit activity released at
the reactor boundary (113). Vennart and Ash haverecently reviewed the dosimetry of inhaled and ingested
35
S(l68).
102. The absorbed dose rate at the site boundary fromreleased from GCRs is given as 5 10-s rad y- 1
(whole-body) for a 2000 Ci y· 1 release (113). Absorbed
dose rates in air from 1 6 N radiation originating from theturbine buildings of BWRs (about 5 rnrad y- 1 at 0.5 km)
are quite dependent on source and field geometry (86,
87).
16 N

3. Tritium

103. Tritium present in airborne effluents from reactors
can contribute to the local collective dose primarily
through the inhalation pathway and secondarily throughthe ingestion pathway. The local collective dose
contribution after deposition through the ingestion
pathway may be quite variable from site to site owing todifferences in local hydrology and water usage. This
pathway is accounted for in the general assessment of
the global collective dose commitment in chapter VI.
The intake of airborne tritium occurs by inhalation(20 m 3 air inhaled per day) and by passage through skin.
Intake through the skin is assumed to be equal to thatthrough inhalation. The dose to soft tissue per unit
activity intake in the form of tritiated water is
84 rad Ci- 1 (12). Assuming an atmospheric dispersion
factor of 5 10- 7 s m - 3 at 1 km from the release and areduction in concentration due to further dispersion
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 and including the above
factors, we obtain the estimated local collective dosefrom airborne tritium per unit activity released:
1
S = (5 10- 7 s m- 3 ){40m 3 d- 1 ) ( -- d s- 1 )(84 rad Ci- 1 )(100 man 1cm- 2 )

f

86400

00

,.-u 2 n rdr

the local collective dose from airborne tritium per unit
activity released is estimated to be 2 10-4 man rad perCi.
104. The average discharges of airborne tritium basedon operating experience (table 11) are (Ci per MW(e) y),for PWRs and GCRs, 0.2; for BWRs, 0.05; for HWRs. 18.
Using the above value of collective dose per unit release.
193the estimated local collective doses, for the assumed
local situation, are 4 10-s, 1 10- 5 , and 4 10- 3 man rad
per MW(e) y for the PWR and GCR, the BWR, and the
HWR, respectively.
105. Tritium in liquid effluents from reactors cancontribute most directly to the local collective dose if
the discharge is to a river that provides a drinking watersupply. The collective dose per unit activity released,
estimated from the drinking water intakes given intable 22 (discussed in paragraph 126) is 0.007 man rad
CC 1 • The tritium releases reported from the most recent
operating experience (table 12) are about 1 and 0.1 Ci
per MW(e) y from PWRs and BWRs, respectively. From
HWRs the release is assumed to be 4 Ci per MW(e) y. The
local collective dose estimates would thus be 0.007,
0.0007, and 0.03 man rad per MW(e) y for the PWR,
BWR, and HWR, respectively.

behaviour of hypoiodous acid is uncertain; it may simply
decompose to the elemental and organic forms (114).
Physico-chemical transformations occurring during at·mospheric transport may also affect the amount of
iodine in its various organic forms, since these are n.ot
stable in sunlight. The elemental form would be
expected to become rapidly associated with airborne
aerosols, so that deposition at distances beyond theimmediate vicinity of the release would be largely
governed by the particulate behaviour. This is also
shown by fallout measurements (116).

4. Carbon-14

109. A recent reassessment by Hoffman (53) of iodine
deposition factors fo.· predicting environment transfer
has shown that the experimentally derived values of thedeposition velocity have frequently been misinterpreted.
As the measurements relate directly the actual
deposition rate on vegetation to the concentration in air,the use of an additional factor to account for theintercepted fraction of the total deposition is therefore
inappropriate.

106. Carbon-14 makes a very small contribution to the
local collective dose. It is of more significance in termsof entering the carbon cycle and contributing to theglobal collective dose, as discussed in chapter VI. The
local collective dose due to 14 C airborne releases fromreactors can be estimated by comparing the resulting
specific activity in air (which contains 0.16 g of C per
cubic metre) to the natural specific activity (and thus
the natural 14 C dose rate). Using the same atmosphericdispersion model and the same population assumptions
as in the case of airborne tritium (para. 103), the local
collective dose commitments per unit activity released
are estimated to be 0.12, 0.1, 0.37, and 0.40 man rad Ci- 1
for the lung, gonads, bone lining cells and bone marrow,respectively.

110. Recent determinations of deposition velocities as
the ratio of the total deposition on vegetation per unit
ground surface area to the time-integrated concentrationin air give 2 cm s -I for the deposition of elemental
iodine on grass and 0.01 cm ·s-1 or less for methyl
iodide (47). In the case of particulates (mean aerosol
particle diameter 4 µm), a value ofO.l cm s- 1 was foundrepresentative for grass, about 0.2 cm s- 1 for clover andabout 0.03 cm s- 1 for vegetation-free soil (54). Since
the deposition velocity varies with the vegetation
density, less variability is encountered by normalizing
the values by the mass of dry vegetation per unit area(52). For the conditions under which the particulate
deposition velocities were measured, the normalized
value for grass is 0.005 m3 kg- 1 s- 1 (54).

107. For a normalized 14 C release from a light-water
reactor of 6 10- 5 Ci per MW(e) y (para. 76), the local
collective dose commitments are estimated to be 7 10-4 •6 10- 4 , 2.2 10-3 and 2.4 10- 3 man rad per MW(e) y forthe lungs, gonads, bone lining cells and bone marrow,respectively. As the inhalation intake of carbon is about
3 g d- 1 and only 1 per cent of this intake is retained in
the body (13), whereas the ingestion intake is 300 g d- 1
with 100 per cent retention (68), the inhalation pathway
can account for only 10-4 of the local collective dose.
Regional contributions to the collective dose have beenassessed by Kelly et al. (74), considering the first passageof the released material over a country or region. These
contributions are also very small compared to the global
collective dose commitment from 1 4 C.

111. The assessment of the transfer of iodine to milk,
which is the main pathway to man, requires theknowledge of the value of the following parameters in
addition to the deposition velocity: the residence
half-time of particulates on vegetation (14 d), corresponding to an effective mean-life of 20.2 d for 1 2 9 I and7.36 d for 13 1 I; the average mass of grass consumed per
cow and per day under average agricultural conditions
during the grazing season, taken to be 11 kg (dry) perday (51 ); the fractional transfer of the daily ingested
activity by the cow per unit volume of produced milk,
0.01 d 1 -I; and the fractional pasture-grazing time,
assumed to be 0.5 for 131 I and 1.0 for 1 2 9 I. Taking
account of all these parameters, an integrated air
concentration of I pCi d m -.3 of 1 3 1 I or 1 2 9 I would
thus result in an integrated milk concentration for 131 I of
175 pCi d 1- 1 and for 129 1 of 960 pCi d I -i. Transfer
via wet deposition is usually insignificant over the courseof a year. For specific times or locations, however, it
may indeed be significant.

5. Iodine108. Assessment of local dose from 1 3 1 I released fromreactors is complicated by the various forms that iodine
may take-particulate, elemental, organic or as hypoiodous acid. Elemental iodine readily deposits on forageand enters the cow-milk-man pathway. Organic iodine,
however, is retained much less efficiently by vegetation,and its deposition velocity is 200-1000 times smaller
than that of the elemental form (2, 47). Particulateassociated iodine will be deposited at rates intermediate
between those for the elemental and organic forms. The
194

112. To assess the transfer of iodine from air to fresh
leafy vegetables, the required parameters are: thedeposition velocity and the residence time on thevegetation (same values as discussed above); thefractional removal by washing, 0.4 (157); the decay due
to an average 7-d marketing delay, if applicable, 0.55 for
131
I (I 57); and a dry-to-wet vegetation weight ratio of
0.5. Time-integrated concentrations of 350 pCi d kg- 1

(fresh weight) for 131 1 and 1740pCidkg- 1 (fresh
weight) for I 29 l are calculated for the case of atime-integrated air concentration of I pCi d m - 3 •113. An estimate of the transfer of 1 29 I to beef canalso be made. The parameters required are: thedeposition velocity, 0.005 m3 kg- 1 s- 1 ; residence meantime on grass, 20.2 d; grass consumption rate, 11 kg(dry) per day; fractional transfer of daily ingested
activity per unit mass of meat, 0.003 d kg- 1 (111); andthe fractiorial pasture time, 1.0. The resulting transfer
factor is 290 pCi d kg- 1 per 1 pCi d m - 3 of 129 1 in air.
114. The estimated transfer factors from air to milk,
leafy vegetables or meat have not been verified in thecase of actual radioiodine releases from reactors. A
careful attempt to measure 131 I in the environment
around a reactor during normal operation has beenreported {170). Low levels of 131 I in air were detected
near three BWRs. but only a few samples of vegetationand milk showed 131 I activity above the detection limit.
The transfer of 1 3 1 I from deposition on grass to milk
has also been estimated from fallout field measurements(see Annex C). These estimates of the transfer factor are
somewhat lower than the value given in paragraph 11,
but it has been suggested that the availability of fallout
131
I might be reduced because of its physical form
(140). Because of the variability of all of the intervening
factors, the assessment of transfer factor given in thepreceding paragraphs can be expected to give only ageneral indication of the doses from airborne releases ofiodine.
115. In a review of the iodine metabolic parameters
certain values have been suggested for use in computing
the age-dependent doses to the thyroid gland following
an ingestion or inhalation intake (51). The values for
131
I and 12 9 I are given in table 18.

116. The average consumption rate of milk (68), leafy
vegetables (157) and beef (one third of maximum meat·consumption) (132) and the inhalation rate (68) for
representative ages are given in table 19. These intake
rates and the transfer factors estimated for the various
pathways were used to compute the thyroid doses perunit time-integrated activity concentration in air, also
listed in table 19. The highest dose is to the infant and isdue almost entirely to milk consumption.
TABLE 19. DATA RELATING TO THE ENVIRONMENTAL
PATHWAYS AND THYROID DOSES FROM' 3 i I and 1 2 • I

Age
6 months

Inhalation pachway
Breathing rate (m 3 y- 1)

1 150

3 530

6 440

8 030

Ingestion pathway
consumption rate~filkO/d)
Leafy vegetables (kg/y)Beef (kg/y)

0.9
0
0

0.5
13
8

0.4
20
15

0.25
30
27

Inhalation
Ingestion°

14

19

12

11

l,lilkLeafy vegetables

920
0

230
33

61
17

30
20

Totalb

930

280

90

61

Dose per unit integrated
concentration in air(mrad per pCiy m·'J

'" I

u•r
Inhalation
Ingestion a
Milk

Age
6 months
2.0

4.0

14

20

Fractional uptake
by the thyroid
Ingestion
Inhalation

0.40
0.30

0.35
0.26

0.35
0.26

0.35
0.2§

0.18

0.18

0.19

0.19

6.0

6.3

6.9

7.6

IJII

Source: Reference 51.

29

27

67

6 300
0

1 900
250
26

780
190
24

960
570
86

6 300

2 200

1 OOO

1 700

0

aParticulate form assumed. Environmental transfer andtherefore dose - 20 times higher for the elemental form and- 20 times less for the organic form.
bRounded.

4 years 14 years Adult

Thyroid mass (g)Energy per
disintegration (MEV)
Effective half-timein the thyroid (d)Dose per unit intake(rad µCi" 1 )Ingestion
Inhalation
12•1
Energy per
disintegration (MEV)
Effective half-timein the thyroid (d)Dose per unit intake(rad µCi "1)
Ingestion
Inhalation

17

Leafy vegetables
Beef
TotalbTABLE 18. DATA RELATING TO THE AGE-DEPENDENTDOSE TQ THE THYROID GLAND FROM 1 3 I! AND
29
I! INTAKES

4 years 14 years Adult

16
12

7.3
5.5

2.4
1.8

1.9
1.4

0.060

0.061

0.063

0.064

23

28

48

136

20
15

11
8.3

5.6
4.2

1.1

117. A population-weighted dose factor can beobtained by assuming that the doses to the three groupsof children are representative of the age groups 0-1, 1-9,
and 10-19 y, respectively, and that these groups containrespectively 2J6 and 20 per cent of the ·population (157),
the population-weighted thyroid dose factor is 120 mrad
per pCi y m - 3 for 1 31 I and 1700 mrad per pCi y m - 3
for 129 I.
118. The local throid collective dose commitment perunit activity of 1 1 I released from the reactor can becalculated using the same dispersion assumptions andpopulation density values discussed previously:S = (5 10-1sm-3)(3.ls\01 ys-1)(120 mrad pCi- 1 y- 1 m3 )(100 mankm-2 )
100

8.4

f

,-1.s 2nrdr

1= 22 man rad per Ci

195

On the basis of average operating exf':rience {table 13),the normalized airborne release of 3 1 I from BWRs is
5 10- 3 Ci per MW(e) y. The thyroid collective dose
commitment per unit of electrical energy generated istherefore estimated to be 0.11 man rad per MW(e) y. The
normalized collective dose commitments for PWRs,
GCRs and HWRs are smaller by two orders ofmagnitude.
6. Particulates in airborne effluents

119. The activities of particulates released in airborne
effluents may be quite variable but are relatively low.
From releases based on average operating experience(table 14). estimates may be made of the local collective
dose commitments from immersion and from inhalation

intake while the radioactive material remains airborne
and from the ingestion pathway and the exposure tocontaminated ground following deposition.
120. The highest particulate activity discharge from
PWRs was 0.0004 Ci per MW{e) y of 88 Rb. This would
correspond to 0.004 Ci per MW(e) y of the precursor
88
Kr. Compared to 8 8 Kr activity releases of 280 Ci per
MW(e) y from BWRs and 0.09 Ci per MW(e) y at PWRs,
the release amounts and therefore the collective doses
from 8 8 Rb particulate discharge are one or more ordersof magnitude lower than those from 8 8 Rb that arise
from discharge of 8 8 Kr.
121. Estimates of the collective dose commitments by
the various pathways following unit release of activity inairborne effluents, are listed in table 20. The external

TABLE 20. LOCAL COLLECTIVE DOSE COMMITMENT FACTORS FOR RADIONUCLIDES IN AIRBORNE PARTICULATEEFFLUENTS
Collective dose commitment per unit activity released(man rad cr 1J
Ingestion

Inhalation

Radionuclide
seeo
60Co
e9sr
9osr
t06Ru
t34Cs
137Cs

1•0Ba

immersion
total bodyLung(whole)Bonemarrow

0.001
0.003

0.09
1.9
0.01
0.03
5.6
0.07
0.06
0.009

0.005
0.08
0.1)2
1.0
0.009
0.07
0.05
0.005

0.00020.002
0.00070.0003

gamma dose from the passing cloud (immersion
exposure) makes a minor contribution to the collective
dose commitment. The dose assessments from immersion and from inhalation assume decreasing concentrations with distance integrated to 100 km and a
population density of 100 man km-2 in the local region. The
dose estimates from the ingestion pathway and fromexposure to contaminated ground do not require thedeposition regions to be specified, provided that the
population density can be assumed to be constant in therelevant area (JOO km-2 is assumed). The transfer
factors from deposition to diet and to man for strontium
and caesium nuclides are derived from fallout fieldstudies (see Annex C). The assessment of the exposure
to contaminated ground is also discussed in Annex C for
most of the radionuclides in table 20. The additionaldose factors for 6 °Co and 134 Cs assume a plane-source
distribution maintained in time (10) and may thus beslight overestimates, as some downward movement insoil is expected over the mean-lives of these radionuclides.
122. Estimate of the collective dose commitments due
to radioactive particulates released in airborne effluentsfrom reactors per unit electrical energy generated are
given in table 21, obtained by applying the dose factorsof table 20 to the releases representing average operating
experience. The collective dose commitments are
somewhat higher for BWRs than PWRs, ranging from
3 10- 5 to 2 10-4 man rad per MW(e)y to the lung. due
primarily to 6 °Co, and 3 10- 3 to 1 10-2 man rad per
MW(e) y to total body. due to 6 °Co from external

196

Total
body

Exposure tocontaminatedgroundTotal bodyBonelining
cells

Air

Bone
marrow

1.2
80.0

0.06
100

140
1.9
21.0
46.0
0.5

30
20

exposure and to t 34Cs and t37cs from external
exposure and through the ingestion pathway. The dose
commitments to bone marrow and bo:.e lining cells aredue almost entirely to the smllll amounts of 90 Sr
released.
TABLE 21. LOCAL COLLECTIVE DOSE COMMITMENTS
FROM PARTICULATES IN AIRBORNE EFFLUENTSFROM REACTORS
Normali=ed
activity
released(Ci per
MW(e)y)Collective dose per unit electricalenergy generated (man rad per MW(e) y)
Total
body

Lung
(whole)Bone
Bonemarrow liningcells

PWR

"'Cs
•oeo
134Cs
seeo
14oBa
90Sr
89
Sr

2.1 10-s
1.5 10-s
1.4 10-s
1.6 10-•3.2 10- 1
1.4 10- 1
6.0 10·'Total

1.4 10-> 0.1 10-s
1.2 10-> 2.9 10·•
0.7 10-> 0.1 10-•0.02 10·3 0.1 10-sO.o2 10-• 0.03 10-•0.04 10· 1
0.06 10·'
3

10-3

3

0.1 10·•0.1 10-s
0.1 10·•0.08 10-•0.02 10· 1
1.4 10-• 2.0 10·•0.05 10·•10·• 210-• 2

10-s

B WR
6

°Co 1.0 10-•"''Cs 2.3 10·•137Cs
1. 7 10-•uoBa 3.9 10-•58
Co 1.8 10-s0
5.0 10-•' Sr
19
Sr 3.3 10-•Total

8.0 10-• 1.9 10""'
1.2 10-3 0.02 10-•I. I 10·• 0.01 10-•0.2 10-> 0.04 10·•0.02 10-> 0.02 10-•0.02 10·•0.03 10-•10·•

20.08 10-•0.02 10-4
0.09 10·•0.02 10-•o.o9 1o-•
5.1 10-• 7.0 10-•
0.3 10-•10·• 510·· 710-•

7. Radionuclides in liquid effluents
123. Radionuclides discharged in liquid effluents may
result in doses to man through the pathways of drinking
water and fish consumption for releases to fresh water,and of ocean fish and shellfish consumption for releases
to salt water. A portion of the population may also be
exposed on shorelines to external irradiation fromsediments. In areas where irrigation of crops is practised,especially when sprinkling methods are utilized, asignificant contribution to the collective dose could
result. This pathway should then also be considered.
Ingestion of water fowl and immersion exposure fromswimming are other pathways, but these have generally
been found to be insignificant contributors to thecollective dose (19).
124. Assessments for specific reactors can be performedby means of rather specific transfer models and using theinformation provided by environmental monitoring. This
monitoring is usually carried out along the criticalpathways leading to human exposures. Site-specificassessments for the liquid discharges of nine GCRs in the
United Kingdom {106) show ·that the local annual
collective dose to the gonads in the vicinity of thesereactors is in the range of 10- 1 to I man rad. Combining
these values with the electrical energy generated in eachcase (106), the normalized gonad collective dose isestimated to be in the range of 5 10-4 to 4 10-3 man
rad per MW(e) y.
125. In the general case, the dose assessments require
information on the discharged activity of eachradionuclide, the volume of the receiving water into
which the radioactivity is diluted, the concentrationlevels which are reached in fish and shellfish, the factors
regarding removal to sediments and exchange rates ofwater bodies, and the number of individuals in the
populations who use the water for drinking purposes andwho consume fish.
126. The local and regional collective dose commitments from J1!dionuclides in liquid effluents can be
estimated using the expression

where= individual

consumption
rate of pathway item k= tissue dose per unit activity in~ested.
By using the normalized release rate for Ai (Ci per
MW(e) y), the collective dose will be expressed in unitsof man rads per MW(e) y. The total collective dose
commitment per unit of electrical energy generated sc is
obtained by a further summation over all radionuclides.
127. The quantity ~ (pCi y 1-1 ) is the integral
V(T+ /\.)

water concentration for release of an activity A (pCi) or,
alternatively, the equilibrium concentration in water
Cw.i (pCi 1-1 ), for a constant continuing release rate
(Ci y- 1 ). The equilibrium concentration in fish or shellfish is Cw,i lk,i (pCi/kg fish), where fk,i is theappropriate fresh water or salt water concentration
factor.
128. For radionuclide releases to small volumes ofwater, the concentrations in water or fish may be high,but the population which can be served with drinking
water or by fish production will be limited. For releases
to larger water volumes, the concentrations will be less,but the populations involved will be correspondingly
larger. In fact, the Nk/V relationship could be taken, ina crude approximation, as relatively constant. the inverseof which indicates the water utilization with regard to
the specific pathway k per individual in the population.
129. For the drinking water pathway, a fresh water
utilization ratio of 3 107 1 man -t will be assumed inestimating collective dose commitments from generalized liquid releases. 'This value of V/Nk is assumed to bea global average, obtained from an estimated global totalof 1.3 I 0 1 7 l of fresh water in lakes (1 101 7 I) and rivers(annual flow 0.3 I 0 1 7 1) (5), serving the world
population of 3.8 109 • Similar results are obtained forregional areas. Martin (93) estimates that a population of
3.1 106 derives drinking water from Lake Ontario and
the St. Lawrence River, comprising an available volumeof 2.2 10 14 1. This gives a water utilization factor of
7 10 7 1 man- 1 • In the Danube River studies, Ruf(l22)quotes annual flows of (3-7) 10 12 1, from which several
million derive their drinking water. This gives a factor of
about 2 10 6 1 man -I. For the large flow of theMississippi River. the estimated factor is 3 10 8 1 man -I(159). Even more widely varying results would bepossible for more localized situations.

= activity ofnuclide i released

Ai(pCi)to water= volume of receiving water= turnover time of receiving

V(l)

water

= radioactive decay constantfor nuclide i= number of individuals forNk (man)pathway k

fk.i (pCi kg- per pCi 1- ) = concentration factor for
item in pathway k for
nuclide i11130. Average fish plus sea-food consumption per
individual is about 8 kg y- 1 . ranging from 4 to 6 kg y- 1
in the Near East and Africa to 10-14 y-1 in the Far
East and Europe (68). It may be assumed that theannual consumption is 6kgy- 1 ocean fish, 1. kgy- 1
fresh water fish and 1 kg y- 1 shellfish. Total fresh water
fish consumption by the world population is thus
3.8 109 kg y- 1 , which, when a correction is made foredible weight, agrees well with the estimated annual
global harvest cif 10 1 0 kg landed weight (5). Dividing by
the global fresh water volume given in the above
paragraph, we obtain 3 10-s man kg y- 1 1-1 , which will
be assumed to be the factor Nklk/V needed in estimatingcollective doses from fresh water fish consumption.

kg

197

131. Similarly, ocean-fish consumption by the world
population is 22.8 109 kg y- 1 : The annual global ocean
fish harvest is estimated to be 60 109 kg landed weight(5). The ocean surface area is 3.6 108 km 2 , andassuming the depth for fish production extends to 30 m,
the water volume is 10 19 I. Of course, most of the fish
are taken from coastal regions, comprising only about
10 per cent of the ocean's surface. However, most of thelarger .water volume can by expected to circulate
through the fish er<:>duc~on regions. Thus, the factor
Nklk/V which is required for the salt water fish pathway
is 210- 9 man kgy- 1 r 1 • For the shellfish pathway,
with consumption one sixth that of ocean fish, a valueof the factor NkI,iV of 3 10- 1 0 man kg y- 1 1- 1 will be
used in this Annex.
132. For estimating collective doses from exposure to
sediments, rather more specific assumptions must be
made regarding the receiving waters and the potential
population involved. Booth et al. (19) have calculated
sediment exposure rates based on releases to a water
volume of 10 1 2 1 with a water turnover rate of
0.1 y- 1 • This corresponds to an annual dilution volumeof 1011 I. Dividing by the fresh water utilization factor
derived above of 3 107 1 man -i, it can be assumed thatthe exposures apply to a population of 3000. These
values will be used to obtain collective dose estimates for
sediment exposure in the general release situation.
133. For assessments involving releases to the sea, thedilution volume will be assumed to be 101 5 I, with aturnover of 1 y- 1 • The salt water utilization factor fromglobal values of ocean volume and population is 10 1 9 1
per 3.8 109 man, or 3 109 l man -l. Thus, for a dilution
volume of 10 1 5 1, the corresponding population is
300 OOO. For releases to salt water, this value will be
used to obtain collective doses from the individual dose
rates from sediment exposure calculated by Booth et al(19). The estimates of populations exposed to sediments
are admittedly rather arbitrarily selected and should
certainly be adjusted for estimating the collective dose at
specific sites.
134. Correction factors to account for removal ofradionuclides from water to sediments depend on the·turnover rate, that is, the flow-through oi: replacement
rate of the receiving water. The correction becomes lessimportant for fast moving water ( r = 10 or 100 y- 1 ).
For releases to salt water, in which depths greater than
I Om can be assumed, the correction is also negligible(19). This, however, will depend on how the effluentsare admitted to the water. For near-shore releases not far
from the bottom, sediment removal is considerable formany radionuclides. Even for surface releases, sediment·
ation may in some cases be an important process,ultimately removing the activity from the water.
Whether sediments once deposited might eventually
become resuspended by storms or by winds from shore
areas is a question that is being studied.
135. In the case of releases to fresh water, a turnoverrate of 0.1 y- 1 will be assumed for the generalized
assessments. For 13 7 Cs, the correction for sedimentremoval is taken to be 0.29 (19); therefore, the198

time-integrated activity for an activity release of 1 Ci is1
0.29(r += 2.4 Ci y. For lake Huron, which wasanalysed by Barry ( 4), the integrated value for a unit
activity release of 1 7 Cs was 1.4 Ci y, in reasonableagreement with the assumed generalized case. Mitchelland Jeffries (IOI), however, estimate a more rapid
turnover rate for Lake Trawsfynydd, the half-time being
50 d, corresponding to r= 5 y- 1 • For releases to salt
water, a turnover rate of 1 y- 1 will be assumed with noadditional removal of activity to sediments. Kautsky(73) estimates the turnover of water in the North Sea tobe of this order, and Hetherington (48) reports theresidence half-time of water in the Irish Sea to beapproximately 1 y.

)lr136. The factors which are required to estimate
collective doses ( using the formula in paragraph 126) fora number of radionuclides that may be present in liquid
effluents released to fresh and salt water are listed intable 22. The corrections to account for removal tosediments are also included, and in addition, factors are
given for the removal of activity during treatment ofdrinking water (30, 122). The concentration factors for
fresh-water fish are those compiled by Thompson (135)and for marine fish and shellfish by Freke (35). The
factors relating dose to activity intake included intable 22 have the follow characteristics: (a) values forthe total body (TB) from 3 H, 14 C, 65 Zn, 134 Cs, and
13 7
Cs apply approximately equally to most single-organ
exposures; (b) the dose factors for thyroid (Thy) aretaken from table 18; (c) for 90 Sr the dose factor usedrefers to exposure of bone lining cells; the dose to bonemarrow would be 30 per cent smaller; ( d) the doses from32
P, 89 Sr, and 23 9 Pu are in total bone; (e) for the otherradionuclides the highest doses are experienced in thelower large intestine (LLI).
137. Collective dose estimates for the composition andamounts of radionuclides other than 3 H in liquid
effluents from reactors reported from operating
experience (table 15) are given in table 23. The
collective doses to the total body include contributions
from ingestion of 137 Cs, 134 Cs and 65 Zn, and fromexternal exposure of 137 Cs, 134 Cs, 60 co, 54 Mn, 144 Ce,
51
Cr, 1 2 5 Sb, and 1 0 6 Ru in sediments. The dose to thelower large intestine is due primarily to 6 °Co and 54 Mn
for PWRs and BWRs and to 1 2 5 Sb for GCRs. The doseto bone from 3 2 P in releases to salt water from GCRs is
estimated to be comparable to the 90 Sr dose, but ismuch less for release to fresh water. Since the averageradionuclide release from GCRs applies chiefly toreactors which discharge to the sea, a separate example isincluded in table 23 for discharges to fresh water fromthe GCR at Lake Trawsfynydd.
138. The estimates of collective dose in table 23
correspond to average release amounts from reactors andfor generalized release situations. The collective doses
for specific reactor situations may differ strongly fromthem. There may also be particular radionuclides or
special pathways which might be overlooked. In general.
however. the results seem to be consistent with the dose
estimates from environmental sampling (15, 70, 71,
100).

TABLE 22. COLLECTIVE DOSE FACTORS FOR RADIONUCLIDES IN LIQUID EFFLUENTS RELEASED TO FRESH AND SALT WATER

Radionuclide

Radio·
activehalf·life
Ty,

Sediment
removalcorrection
factors;

1Yme
integralof,mitactivity
in waterWj (Cly)

Drinking•
water
treatment
removalfactor

1.0
0.78
1.0
0.97
0.83
0.33
0.96
1.0
0.78
0.81
1.0
0.92
0.59
1.0
0.94
1.0
1.0
0.51
0.29
0.63

6.39
7.79
0.056
0.105
0.914
1.43
0.845
0.196
6.27
0.197
0.136
0.147
0.743
0.964
2.63
10.0
0.032
1.19
2.36
0.648

1.0

10.0

1.0
1.0
0.4
0.9
0.5
0.5
0.4
0.5
0.5
0.1
0.1
0.1
0.1
0.5
0.5
0.8
0.8
0.2
0.2
0.1
0.1

fjj

Concentration Dosefactor, fish
per unit
activity
f2i(pCi kg-• fish ingested
per pCi r
Dj(10" 6 radCr'J
water)

Extimal dosefrom sedimentper unit
activity
released(10- 6 rad er')Collective dose per unit activity released(man rad er')
Drinking
water

Fish

SJ;

S21

0.00710.057
0.011
0.0012
0.11
0.37
0.029
0.080
29.0
0.00830.00340.0046
0.19
0.41
0.33
I 050
1.0
0.23
0.27
0.16
0.25

1.4 10-•0.60
6.4
0.00012
0.052
0.033
0.16
0.00180.66
0.000622.3
0.001 I
0.042
0.00430.0015
45
0.045

Total
(internal)

Sedime11t
exposure(external)

Fresh water

•11

uc
"P
.,er
54

Mn

60Co
65

Zn
.. Sr
9osr
9•zr
95
Nb
'"'Ru
1o•Ru
IIOlllAg
125

Sb

129)

"' 1
""Cs
37
'
Cs
'44Ce
239p,,

12.26 y5 730 y14.3 d27.7 d313 d5.26 y244 d50.5 d28.5 y
63 d
35 d
41 d
1.0 y
270 d
2.7 y
1.6 10 7 y
8.05 d
2.1 y
30.0y
290 d
24 400 y

0.9
4 600
100 OOO
40
100
20
I OOO
5
5
3.3
30 OOO
10
10
2.3
1
15
15
400
400
25
10

0.000084
0.000560.038
0.00097
0.019
0.039
0.0065
0.062
0.7
0.032
0.019
0.024
0.19
0.065
0.019
10.0
3.1
0.075
0.043
0.19
0.019

0.0023
2.76
162
0.30
1.13
0.0090
0.11
1.73
0.84
2.15
6.7 10-•
36
235
0.33

1.1

1.2
0.092
0.06

0.00710.66
6.4
0.0013
0.16
0.40
0.19
0.082
30.0
0.00892.3
0.0057
0.23
0.41
0.33
1 100
1.1

1.3
1.5
0.25
0.31

TB
TB
Bone
LU
LU
LU
TB
Bone
Bone
LLI
LLI
LLI
LLI
LLI
LLI
Thy
Thy
TB
TB
LU
Bone

6.8 10-•0.00830.48
0.000900.0034
2.7 10-•0.00032
0.00520.00250.00652.0 10· 7
0.11
0.70
0.00099

Salt water
Concentration factor
Fish
Shellfish
S3;
S4i

Concentration factor
Fish
Shellfish
C3;

'H

'4c
32p
51

Cr
Mn
6
°Co
••zn
19
Sr
90
Sr
95
Zr
95
Nb
'"'Ru
0
' 'Ru
54

110l11Ag
12
'Sb
"'l

"' l

-

\0
\0

'11

:I

u•cs
•J?cs
•••ce

219Pu

12.26 y5 730 y14.3 d27.7 d
313 d
5.26y
244 d
50.5 d28.5 y
63 d
35 d
41 d
1.0 y
270 d
2.7 y
1.6 107 y
8.05 d2.1 y30.0 y290 d
24 400 y

1.0

0.946
1.0
0.056
0.099
0.553
0.884
0.491
0.166
0.976
0.199
0.122
0.139
0.591
0.516
0.796
1.0
0.031
0.752
0.977
O.S34
1.0

C41

1
1
10 OOO
100
3 OOO
100
5 OOO
1

1
30
100
3

3
1 OOO
l OOO
20
20
30
30
30
3

1
1
10 OOO
1 OOO
50 OOO
10 OOO
50 OOO
I
l

100
200
100
100
5 OOO
l OOO
100
100

so
so

100
300

0.000084
0.000560.038
0.00097
0.019
0.039
0.065
0.062
0.7
0.032
0.019
0.024
0.19
0.065
0.019
10.0
3.1
0.07S
0.043
0.19
0.019

1.1 10-•0.00060.00520.000160.00420.00004
0.000012
0.000076
0.000740.00025
3.6 10- 7
0.00160.00820.0028

1.6 10- 7
1.1 10-6
0.043
1.9 10-•0.063
0.00690.032
2.1 10-•0.00140.000380.000460.000020.00067
0.067
0.030
0.40
0.00380.00340.002S
0.00610.00011

2.4 10-•I. 7 10" 7
0.0064
2.9 10-•
0.16
0.10
0.048
3.1 10-•0.000200.000190.000140.00010.00340.050
0.00450.30
0.00290.0008S
0.00063
0.00300.0017

1.8 10- 7
1.3 t0·•0.049
4.8 1o·•
0.22
0.11
0.082
2.4 10-•0.00160.00057
0.00060
0.000120.0041
0.12
0.035
0.70
0.00670.00430.00310.00910.0018

TB
TB
Bone
LLI
LU
LLI
TB
Bone
Done
LLI
LLI
LLI
LLI
LU
LLI
Thy
Thy
TB
TB
LLI

3.3 10- 7
0.000180.0016
4.8 I o-•0.0013
1.2 10-•
3.6 10-•2.3 10-•0.00022
7.5 10-•1.1 10-•0.00049
0.00250.00084

Bone

Notes: i =radionuclide; k =pathway= I, 2, 3, 4 for drinking water, fresh-water fish, salt-water fish ond shellfish, respectively; hi= decay constant= 0.693/Ty,; r= water turnover rate= 0.1 y·• (fresh
water), 1.0 y·' ~salt water); S~ i = llifk ;NklkD//Vi;where Wi=·tAiSf(T+ "A.if';/1 =4381 y·',Nt/V= 3 10·1 man 1· 1 ; / 2 = I kg y· 1 , N 2 / 2 /V = 3 10·• man kg) y· 1 1· 1 ; / 3 =6 kg y·•, N 3 / 3 /V= 2 10· 9 man kg y·• 1· 1 ;•N 4 ! 4 /V = 3 10· 0 man kg y· 1· 1 •TABLE 23. COLLECTIVE DOSE FROM RADIONUCLIDES OTHER THAN TRITIUM IN LIQUID EFFLUENTS FROM REACTORS
Normalized collective dose (man rad per MW(e) y)
Release ro fresh water

Release ro salr water
Normalized
acrivity released{Ci per MW(e) Y)

Total
body

Lower

Lower
large
intestine

Thyroid

Bone

PWR
31
' Cs
,,.Cs
60
Co

0.002
0.001
0.00040.00010.00008
0.00020.000002
0.002

.. Mn

... Cc

uer

90Sr

"'I

Total

1.1 10-s
0,5 10-•0.06 10-• 4.4 10-•0.02 10-• 2.2 10-•0.07 10-, 0.07 10-s
0.01 10-1 0.01 10-'

3.2
2

10-s

7

10-s 3Total
body

4.4 10- 3
1.4 10-3
0.2 10·3
0.08 10-•0.08 10-,0.01 10-•

10-9

large

Bone

intestine

1.6 10-•1.6 10-s
0.2 10 ...
0.03 10·•
6.0

10-s

6

10-•

2.2 10-•1.3 10·•

10-9

10·•

6

10-3 2

10-•

Thyroid

2

10-3

BWR

0.025
0.017
0.006
0.002
0.00060.00010.0008
0.004

'"Cs
134Cs

•oeo
.. Mn

51 Cr
,osr
89
Sr
"'I

Total
GCR

137Cs

•34cs
65

Zn
'°Co
, .. Sb

•••ce

1o•Ru
32p
tosr
89

Sr

1.4 10 ...
0.8 10 ...
0.1 10 ... 6.6 10 ...
0.04 10-• 4.4 10 ...
0.02 10-• 0.03 10-'2

10-• 1

Salt
water

Fresh
water

0.068
0.014
0.00008
0.0007
0.008
0.001
0.002
0.00080.01
0.00070.006
0.00090.00005
0.00010.008
0.00005
0.00080.00070.0005

3.8 10 ...
0.7 10-•0.06 10 ...
0.01 10 ...
0.08 10-s
0.08 10-•0.05 10-•

Total

5

10 ...

1.6
0.2

10-•
10-•

10-• 210-•

0.8 10-•2.8 10 ...
0.09 10 ...
0.08 10 ...

4

IV. FUEL REPROCESSING
139. At the fuel reprocessing stage of a completely
integrated nuclear fuel cycle, the uranium andplutonium in the irradiated fuel are recovered for reusefrom the fission products that have accumulated during
fuel burn-up in the reactor. The spent fuel received at
the reprocessing plant is stored under water, which
serves both as a transparent radiation shield and as acoolant. The fuel elements to be processed aretransferred to a head-end cell, where they aredisassembled, and the individual rods are treatedmechanically to expose the fuel. The fuel is then leachedwith nitric acid140. A solvent extraction process is then used toseparate the uranium and plutonium from the fission
and activation products. Nearly all fuel reprocessing
facilities employ some form of the Purex process. whichuses the organic complexing compound tributyl
phosphate to extract both the uranium and theplutonium into the organic phase. The uranium andplutonium can be separately back-extracted from the200

3.0 10-•0.07 10-•
4.4 10-•2.7 10-•3.9 10-•1.6 10-s
0.02 10-'10 ... 6

5.5 10-2
2.4 10-2
0.3 10-2 2.4 10-•0.02 10- 3 0.3 10-•0.04 10-• 0.08 10·•

10-•

3

10-•

8

10-2

1.3
0.1

10-2
10-2

3

3

10-•

0.4
1.5

10-2
10-2

2

10- 2

10-•

0.05 10-3 2.6 10-•0.05 10-6 0.01 10-•0.04 10 ... 0.2 10-•10-2 310-•

4· 10-•

organic solution with nitric acid. The product nitrates
are further purified and converted to oxides suitable forrefabrication into fuel elements.

A. EFFLUENTS
141. Waste processing to prevent large releases ofradioactivity to the environment is a complex task at
reprocessing plants. The gaseous and volatile fission
products (I, 3 H, Kr, Xe, Ru, Te and Cs) are largely
separated from the fuel during the fuel dissolution step.
The dissolver off-gas is treated for nitric acid recovery
and for iodine removal before being blended wi~ theoff-gas from other vessels in the plant. The vessel off-gas
is usually treated by caustic scrubbing, drying andfiltering through high efficiency filters before discharge
from a stack. The aqueous wastes. containing almost allof the fission products, are generally concentrated byevaporation and stored in underground tanks and will
eventually be solidified. The evaporator overheads areusually sufficiently decontaminated of radioisotopes topermit their discharge to the environment under existing
regulations.

TABLE 24. RADIONUCLIDES DISCHARGED FROM VARIOUS FUEL REPROCESSING PLANTS, 1960.1974

Year

Electricalenergyequivalentof fuelprocessed(MW(e) y)

Activity released (Ci)(a) In airborne effluents

••x,

3H

90

100 Ru

Sr

a•s,

'"I

131/

137

Cs

134

Cs

Windscale (United Kingdom)
1972
1973
1974

- 2 500 I OOO OOO

Total (Ci)

1 OOO OOO

Normalized release1972(Ci per MW (e) y)

400

0.097
0.75
0.15

23.6

0.004

1.0

24.8

0.004

4 10-s

9 10-•

2 10-•

1.2

0.01

Nuclear Fuel Services (United States)
1966
1967
1968
1969
1970
1971
1972

210
900
530
830
660
720
0

77 OOO
328 OOO
193 OOO
300 OOO
180 OOO
221 OOO
0

Total (Ci)

1 299 OOO

340

Normalized
release 19661972 (Ci per
MW(e) y)-soo
0.000360.041

0.00130.029<0.06
0.0066<0.03<0.03
0.00240.00096-3 075

0.041

0.030

0.84

6 10-s

4 10-s

- 500
- 500

-soo

I 070
5.5

0.00260.00027
0.00210.00090.00060.01

0.00260.00240.00090.0006

8 10-•

4 10-•

3 10-•

1 10-•

8 10- 7

103Ru

s•s,

11 600
3 200
1 800
9 600
1200
1 800
2 500
2 200
1 800
1400
890
830
1160

980
1 370
500
170
190
170
90
140
40
230
470
390
l 080

(b) In liquid effluents
La Hague (France)

1971
1972
1973
Total (Ci)11omAgRu

137C,y

144

7 900
7 600
7 OOO

6 700
1 200
2 OOO

3900
1 400
I 700

1450

90
340
70

22500

9 900

7 OOO

1450

500

106

Ce

95

Zr-Nb

Windseale (United Kingdom)
13'Cs

1960
1961
1962
1963
1964
1965
1966
1967
1968
1969
1970
1971
1972
1973
1974

3H

Ru

tsz,

39 60025 30023 OOO33 40023 10021 OOO
24 90017 20024 20022 90027 60036 40030 SOO
37 80029 200

2400
1 700
940
560
21600
17 70014 10018 80028 10031 6009100
18 OOO
25 600

'

06

90Sr

, .. Ce

6 300
7 900
4 300
3300
20 80033 70023 40025 70037 10030 OOO
9 900
17 30023 SOO

520
490
1 020
550
970
1 160
910
1 390
1 370
2 950
6 OOO
12 30015 200
7 450
10 600

890
2 200
2400
1 400
3 200
3 500
6 900
13 70010 OOO
13 50012 40017 20013 70014 600
6 530

95

Nb

""Cs

"·Pu

""2 500

910
1100
1 100
370
1 300
1 200
1 200
1 600
1500
12 OOO
30 10035 80034 800Total (kCi)

123.0

158.3

416.1

190.2

243.2

62.9

122.0

19.8

8.8

40.0

5.8

Normalized release1972 (Ci per MW(e) y) , 13.9

13.4

12.2

10.2

9.4

6.1

5.5

2.3

0.62

0.46

0.43

16 10020 30024 80032 OOO

31 50033 600

630
7 010
6 370
5 820

492
828
816
936
1 128
l 558
1 776
l 248

201

TABLE 24 (continued)

Year

Electricalenergyequivalentof fuelprocessed(MW(e) y)

Activity released (Ci)

Nuclear Fuel Services (United States)
3H
1966
1967
1968
1969
1970
1971
1972

106Ru

•ner

90 Sr

'""Cs

120 I

""Pu

231Pu

2.27
0.08

1.83
0.19

0.07
0.03
0.2
0.34
0.21
0.06

0.00480.00050.00420.0007

"'Sr

210
900
530
830
660
720
0

290
4 200
2 650
5 980
4 520
3 850
600

48.2
6.1

10.0
1.1

9.9
5.0
10.1
14.2
6.6
0.7

Total (Ci)

22 090

54.3

11.1

46.5

3.35

2.02

0.91

0.00530.0049

5.7

0.075

0.015

0.013

0.00330.00282.5 10-4

7.4 10-'

6.8 10- 6

Normalized
release 1966·1972(Ci per MW(e) y)

Sources: References 21, 48, SO, 93,100,106,123,134,147,172.

142. The radionuclides of concern in reprocessing plant
effluents include primarily 3 H, 14 C, 85 Kr, 129 1, 131 1,
106
Ru, 1 34 Cs, 1 37 Cs, and the actinides. General
discussions of the occurrence and release pathways from
reprocessing plants have appeared in several reports (8,83, 107, 109, 151, 158), and more detailed accounts
have been published for tritium (39, 80), noble gases(49, 80), and iodine (16, 118). Specific information
available on radionuclide discharges from reprocessing
plant operating experience is given in table 24. The
release data from the NFS plant in the United States
represent a completed experience, as this plant wasclosed for modifications in 1972.

143. The release of radioactive materials from the fuelreprocessing plant depends upon the type of fuel,irradiation history, cooling time and the specific waste
processing systems of the reprocessing installation. The
typical radionuclide composition of LWR fuel following
an average bum-up of 33 OOO MW(t) d per tonne and a.
cooling time of 150 dis given in table 25 (107). Bum-up
differs for the various light-water reactor types. The
bum-up levels achieved in presently operating reactors,
and therefore the amounts of longer-Jived radionuclides
generated, have been somewhat less than the example
given in table 25. The cooling time has also been on the
average much longer than 150 d (167).

TABLE 25. RADIONUCLIDE CONTENT OF LWR FUEL
Burn-up, 33 OOO MW(t) d 1-•; cooling time, 150 d\.
Radionuclide

3H··KrSr

89

90 Sr

••y
••y

•s2r
95Nb
tsmNb
103Ru
"''Ru
10,mRh106Rh
11omAg"sSb
12,mTe12,mTe1"Te
12•mTe
u•Te

'"I
"' I

Per unit
mass of fuel(Ci ,-·,

692
11 20096 OOO
76 60076 600159 OOO
276 OOO
5180005 860
89 100410 OOO
89 100410 OOO
261
8 130
3 280
6 180
6 110
6 690
4 290
0.038
2.17

Per unit
electricity
generateda(Ci per MW(e) y)

23
375
3 220
2 570
2 570
5 330
9 250
17 360196
2 986
13 7402 990
13 740
8.7
272
110
207
205
224
144
0.0013
0.073

Radio·
nuclide.... Cs

•nc~

,nmBa
i4DBa
140La

•••ee

144Ce
143Pr
144Pr
14

'Pm

u•sm

154 Eu
•ssEu
160Tb
2••Np
238Pu
"'Pu
240Pu
241Pu
, •• Am
242cm
244Cm

Source: Reference I 07.
a Assuming a reactor thermal efficiency of 0.33.

202

Per unit
mass of fuel(Cit- 1)213 OOO
106 OOO
99 600430
495
56 700770 OOO
694
770 OOO
99 400
1 150
6 820
6 370
300
17.4
2 810
330
478
115 OOO
200
15 OOO
2 490

Per unit
electricity
generateda(Ci per MW(e) y)

7 140
3 550
3 340
14.4
16.6
1 900
25 81023.3
25 810
3 330
38.S
229
214
10.1
0.6
94
11

16
3 850
6.7
503
84

I. Krypton-85
144. The thermal fission yield for 8 5Kr is 0.285 percent for 235 U and 0.144 per cent for 239 Pu (31),corresponding to activities per unit of thermal energy of
0.415 and 0.210 Ci per MW(t) d or per unit electricity
generated of 505 and 255 Ci per MW(e) y for 23 5 U and2 3 9Pu. respectively. Assuming that 41 per cent of thefissions are fissions of 2 3 9Pu in LWR fuel with· a burn-up
of 33 OOO MW( t) d per tonne (31 ), the 8 5Kr activity
generated per unit electricity generated is about 400 Ci
per MW(e) y. A more detailed calculation estimates thegeneration per unit mass to be 11 200 Ci/C 1 ( 107),
corresponding to about 375 Ci per MW(e) y, assuming athermal efficiency of 33 per cent. Almost all (>99 percent) of the 8 5Kr generated by fission in zircaloy-clad
LWR fuels is expected to be released at the reprocessing
plant (80). The operating experience shown in table 24
confirms this expectation.

2. Tritium
145. Most of the tritium separated from fuel during
dissolution appears in the liquid-waste stream. Some iscarried into the dissolver off-gas stream. For the NFS
plant in the United States only about 7 per cent of thetritium in the fuel was released through the stack (25).
More recent reprocessing-plant designs call for about
75 per cent of the tritium present in fuel to be released
to the air and liquid waste discharges to be prevented
entirely (80).
146. A portion of the tritium in fuel elements is
immobilized as a solid zirconium compound in thezircaloy cladding. The fraction thus retained with thehulls is a function of the linear power rating of the fuel.
For present LWR fuel this amounts to about 15 per cent(80). Of an estimated 19 Ci per MW( e) y generation rate,
Grathwohl (39) estimates that 1.2 Ci per MW(e) y are
released at the reactor (PWR), 16 Ci per MW(e) y are
released at the reprocessing plant and 1.8 Ci per MW(e) yare retained with solid wastes. Similar released amounts
are expected in reprocessing HWR fuel, while about one
half of this amount is expected for AGR fuel and one
tenth for HTGR fuel (39). Operating experience(table 24) shows 3H release in liquid effluents to beabout 13 Ci per MW(e) y for Windscale and about 6 Ci
per MW( e) y for the NFS plant. An additional 1 Ci per
MW(e) y was estimated released in airborne effluentsfrom NFS.
3. Carbon-14
147. There are no data available on actual 14 C
discharges from fuel reprocessing plants. In this Annex itis assumed that all of the 14 C present in spent fuel isdischarged in the airborne effluents during reprocessing,probably in the fonn of C0 2 •4. Iodine148. A tonne of reference LWR fuel (table 25) contains
2.17 Ci of 13 1 I and 0.038 Ci of 129 L corresponding tonormalized contents of approximately 0.073 and0.0013 Ci per MW( e) y for 131 I and 12 9 I, respectively.
1
The 3 1 I content depends very much on the coolingtime. In current practice, the fuel is still held for at least
150 d before reprocessing, and in many cases the spent
fuel is stored for much longer periods.
149. Airborne release of 12 9I at the Windscale and NFS
reprocessing plants has been reported to be (2-4)10-6 Ci
per MW(e) y (table 24). The normalized liquid release of
129 1 at NFS, 2.5 10-4 Ci per MW(e) y, corresponds toabout 25 per cent of the total in the fuel. Release ofiodine to the sea from Windscale can be as much as
75 per cent of the throughput (22). The global collective
dose estimates for 12 9I are thus strongly dependent on
the liquid releases. The chemical form of the iodine is ofimportance, particularly for the airborne releases. Haller( 41) has reported a wide variation in chemical form of
131 I discharged from a fuel reprocesssing plant.
5. Caesium, ruthenium, strontium
150. Other radionuclides released from fuel reprocessing plants include 137 Cs, 134 Cs, 106 Ru and 90 Sr.
For the airborne effluents, the decontamination factors(ratio of activity processed to activity released) are of
the order of 10 7-10 10 , based on the NFS experience(90Sr, 410'; io6Ru. 310s; 137Cs, 4109; I34Cs,
9 109). In liquid effluents, somewhat greater fractions of
the processed amounts of these fission products werereleased. For release to fresh-water streams at NFS, thedecontamination factors were 105 -10 6 . Liquid wastes
from NFS were passed through a series of three smalllagoons before discharge to the stream. This systemprovided a 30-60 d hold-up of the liquids and, in
addition, removed considerable amounts of activity by
sedimentation: >90 per cent of 144 Ce. 147 Pm, 95 Zr
and 239 Pu, and 60-75 per cent of 134 Cs, 137 Cs and6
Co (89). Releases of 1 0 6 Ru, 1 2 5Sb and 9 0 Sr were lessaffected by the lagoon hold-up.

°

151. For the Windscale situation, in which release is tothe sea, the achievement of decontamination factors of
102-10 3 is practised. Ratios of 8 9 Sr, to 9 0 Sr and 1 0 3 Ru
to 1 0 6 Ru in liquid discharges indicate fuel-cooling timesof the order of I y (125). After 1969/70, releases oflong-lived radionuclides, particularly 137 Cs, fromWindscale increased, due primarily to leakage from fuelelements in the storage pools.

6. Plutonium. americium. curium
152. Considerable attention has recently been focusedon the transuranic elements, particularly plutonium:
large amounts would become available as a result of
increased reprocessing. It could be used in the operationof breeder reactors or in thennal reactors as recycle fuel.
Included in table 25 are the activities of transuranic
isotopes in the fuel elements of LWRs.
153. During reactor operation, 239 Pu is formed in theuranium fuel by neutron capture by 2 3 8 U, followed byrelatively rapid beta decays of 2 3 9 U and 2 3 9 Np.
Generally more than half of the 2 3 9 Pu undergoes
fission, thus contributing to the ~nergy produced by thereactor. A small amount is transformed by successive
neutron capture to 2 4 0 Pu, 24 1 Pu, 2 4 2Pu and 2 4 3 Pu.
Beta decay of 2 4 1 Pu and 2 4 3Pu produce 241 Arn and
243 Arn which, following subsequent neutron capture
and beta decay, give rise to 242 Cm and 244 Cm. These
interactions are illustrated in figure I.
203

2

"cm

I

tl!T" =

26 mini

243 Am ~ 244 Arn(n, ')'Ilrjl(T,. = 14.Jvl(n,')')/l(T" s 4.98 h)rjl(T,. • 22hl
236Np ~ 237Np ~ 238Np

1

rli(T ,. :nsu ~ n&u ~(n, 7)(n, ')')

:\s

239Np

1/l(T" •d)23.5 min)n,u ~ 238u ~ 239u(n, 2n)(n, 7)

Figure I. Formation scheme for important transuranic radionuclides (31)154. The isotopic composition of the plutonium in aspent fuel of high bum-up is as follows (weight percent): 239 Pu, 60; 240 Pu, 22; 241 Pu. 12; 242 Pu,4.5;
238
Pu, 1.5 (31 ). The composition changes with time
however. Plutonium-238 activity increases as 242 Cm
decays, 2 4 0 Pu increases as 2 4 4 Cm decays, and 2 3 9 Pu
increases slightly as 243 Am decays to 239 Np, which
subsequently decays to 2 3 9 Pu. The amounts of 2 4 1 Am
increase significantly as 241 Pu decays.
155. Most of the half-lives of the transuranic isotopes
are quite long: 239 Pu (24400y), 240 Pu (6580y),238
Pu (86y). 241 Pu (13.2y), 241 Am (458y) and244
Cm (I 7 .6 y ). Relatively short-lived is 2 4 2 Cm
(I 63 d). which forms a significant fraction of the alphaactivity in spent fuel.
156. Measurements at the NFS fuel reprocessing plantin the United States showed that small amounts ofplutonium were discharged in the liquid effluent stream(table 24). In a paper published in 1974, it was suggested
that 0.1-1 per cent of the plutonium throughput was
being released in liquid effluents from the Windscale
reprocessing plant ( 102).

B. LOCAL AND REGIONAL COLLECTIVE
DOSE COMMITMENT
157. The evaluation of the total collective dose fromradionuclides released from fuel reprocessing plants
requires consideration of the exposure of the local
population surrounding the plant and the long-term
global exposure. The global dose estimates are presented
in a subsequent section. The local environment aroundthe NFS plant, being the first commercial fuelreprocessing plant in the United States, has been studiedin considerable detail. A comprehensive review of the
population doses from radioactivity discharged from theplant has recently been presented by Martin (93). Moredetailed surveys to evaluate the doses incurred throughthe ingestion pathway were made by Magno et al. (91 ).
The releases of radioactivity into the Irish Sea from theWindscale plant and the corresponding exposures are still
being closely studied (106). Considerable detail in
obtaining inventories of activity in the sea and sediments
and in tracing the activity in its passage into the North204

Sea are being aclueved (I 00). Some results ofenvironmental surveys Mound the reprocessing plant at
La Hague have also been published (123).
I. Krypton-85
158. The collective dose commitment from 85 Kr
released from reprocessing plants is most significant
from the global standpoint. As discussed in chapter VI,
it is estimated that the collective dose to the gonads dueto the release of 375 Ci per MW(e) y of 8 5 Kr is 0.01 man
rad per MW(e) y during the first two years, until uniform
atmospheric concentrations are achieved. The collective
dose received as the 8 5 Kr initially disperses through thelocal and regional area surrounding the reprocessing
plant is much smaller. Using the same local and regional
dispersion assumptions as in previous sections of thisAnnex, namely a dispersion factor of 5 10- 7 s m- 3 at
1 km and a concentration decreasing as the inverse 1.5
power of the distance, a population density of
100 km - 2 , and a gonadal dose factor (average for testes
and ovaries) of 1.1 104 rad m3 Ci- 1 y- 1 (104), thegonadal collective dose in the local region ( 1-100 km
from the plant) per unit 8 5 Kr activity released isestimated to be 2 10-6 man rad Ci- 1 • If all the 8 5 Kr
present in the spent fuel were released, the estimatedlocal normalized collective dose to the gonads would be
0.0007 man rad per MW(e) y. For other tissues the dose
factors are 1.8104 . 3.1104 , 1.8 10 6 rad m 3 Ci- 1 y- 1
for red bone marrow, whole lung, and skin, respectively(104). The local normalized collective doses are O.OOL
0.002. and 0.1 man rad per MW(e) y for bone marrow.
lung, and skin, respectively.
2. Tritium
159. An estimated 16 Ci of 3 H per MW(e) y could be
released at fuel reprocessing plants (para. 146). primarily
in liquid effluents. For a release of 1 Ci per MW(e) y inairborne effluents, as was the NFS experience (table 24).
and assuming a collective dose factor of 2 10-4 man rad
per curie released (para. 103), the local collective dosewould be 2 10-4 man rad per MW(e) y. For liquid
releases the factors for 3 H in table 22 can be applied.
For the release to salt water (Windscale) of 13 Ci per
MW(e) y, the collective dose is 2 10- 6 man rad per

r
I

I

MW(e) y, and to fresh water (NFS), of 6 Ci per MW(e) y,
the collective dose estimate is 0.04 man rad per

MW(e)y.
3. Carbon-14
160. In paragraph 106 the local collective dose
commitments per unit activity released following
airborne releases of 1 4 C were estimated to be 0.1, 0.12,
0.37 and 0.40 man rad Ci- 1 for gonads,, lung. bonelining cells, and bone marrow, respectively. For an
anticipated release of 0.014 Ci per MW(e) y from thereprocessing plant (paras. 73 and 147) the local
collective dose commitments are estimated to be as
follows (man rad per MW(e) y): gonads, I 10- 3 ; lungs,
2 10- 3 : bone lining cells, 5 10-3 ; bone marrow, 6 10-3 •4. Iodine161. Members of the public can incur thyroid doses
from 1 31 I discharges in airborne effluents from fuelreprocessing plants. A spent-fuel cooling time of 150 d
and a fractional release of the order of 10-3 for airborne
releases of iodine would imply a release of 7 10-5 Ci per
MW(e) y, of the same order as releases by PWRs and
GCRs and much smaller than BWR releases. The NFS
reprocessing plant reported 1 3 1 I releases of I I O-s Ci
per MW(e) y in 1968 and 0.3 10-s Ci per MW(e) y in1972. From the Windscale reprocessing plant the 13 1 I
release was 9 10- 3 Ci per MW(e) y in 1972 (table 24),
but was lower by factors from 20 to 2000 in subsequent
years (106). The local collective dose to the thyroid.
using a dose factor averaged for the age distribution
given in paragraph 117, is estimated to be 22 man rad
Ci- 1 for 13 1 I released in airborne effluents. The local
collective dose per unit of electricity generated would
therefore be expected to be of the order of 0.002 man
rad per MW(e) y for spent fuel held 150 d, and about
100 times more for the highest reported releases in
reprocessing plants. Since large amounts of fuel are being
held in storage for much longer periods than 150 d, local
doses from 1 3 1 I should generally be insignificant when
this fuel is reprocessed.
162. The dosimetry of 129 I was considered in table 18
and the thyroid doses based on representative foodconsumption estimates were given in table 19. The
thyroid dose per unit integrated 1 2 9 I concentration in
air through the inhalation and ingestion pathways,
weighted for the assumed population age distribution(para. 117), is 1700 mrad per pCi y m - 3 • which is thesame as for the adult thyroid. Using the same dispersion
assumptions and population density as in paragraph 103,
it is possible to estimate the local collective dose from
129
I in airborne effluents to be 300 man rad Ci- 1 • Thisis approximately equal to the local collective dose
commitment provided that soil accumulation does not
greatly increase the levels of 1 2 9 I in plants. Soldat
estimates that soil accumulation of 129 I adds 1 per centper year to the leafy vegetable pathway and 1.3 per centper year for milk consumption, assuming that all of theiodine in soil is retained in the root zone (132). Killin etal. (79), however, report that iodine penetrates into soil
. more rapidly and more deeply than other fission
products. The global contribution to the collective dose
commitment from 129 I released to the environment is
considered in chapter VI.

163. For an 129 I airborne release (table 24) of 4 10- 6
Ci per MW(e) y, the local collective dose commitment tothyroid is 0.001 man rad per MW(e) y. The local andregional collective dose per unit release of 1 2 9 I in liquid
effluents is 1100 man rad Ci- 1 for fresh water and
0.7 man rad Ci- 1 for salt water (table 20). For the
reprocessing plant release experience of 2.5 10-4 Ci per
MW(e) y to fresh water (table 24), and 9.8 10-4 Ci per
MW(e) y to salt water (para. 149). the local and regional
collective dose is estimated to be 0.3 and 0.0007 man
rad per MW(e) y for fresh and salt water releases,
respectively.

5. Caesium, ruthenium, strontium
134

106

89

164. The fission products 137 Cs.
Cs.
Ru,
Sr
and 90 Sr are important components of fuel reprocessing
plant releases and contributors to the collective dose.
Caesium-137 and 134 Cs may concentrate in fish andresult in total body doses to the consumers of the fish.
In the past few years, 1 3 7 Cs has contributed the largest
component of the collective dose from radioactive
releases from the Windscale reprocessing plant (100,106). Ruthenium-106 is known to accumulate in theseaweed Porphyra growing near the Windscale discharge
area in the North Irish Sea, from which the foodstuff
laverbread is produced. However, recent changes in theharvesting practice of the seaweed have considerably
reduced the 1 0 6 Ru contribution to the collective dose
144(100). Caesium, 106 Ru, 95 Zr-95 Nb andCe insediments cause external exposures. All of these nuclides
may contribute small collective doses from inhalation
following airborne release from reprocessing plants.
165. Approximate estimates of local collective doses
due to the release of airborne particulates can beobtained by using dose factors derived as for reactor
releases (table 20). These dose factors account for direct
exposure to the radioactive cloud, exposure by theinhalation and ingestion pathways, and by external
exposure to deposited radioactivity. Data on typicalreleases of radioactive particulates in airborne effluentsfrom reprocessing plants are available from the NFS
experience (table 24). These average release amounts areused in the example of local collective dose determination given in table 26. The local collective dose
commitments are of the order of 10-4 man rad per
MW(e) y. Ruthenium-106 is an important contributor tothe collective dose commitment to the lung via
inhalation and to the total body from exposure tocontaminated ground. Caesium-137 and 134 Cs also
contribute to the total body dose through the ingestion
pathway and by external exposure. Strontium-90 is
responsible for the dose to bone marrow and bone lining
cells through the ingestion pathway.
166. Approximate estimates of the collective doses
from radionuclides in liquid effluents can be obtainedusing the dose factors of table 20 and the reported
releases from fuel reprocessing plants (table 24). The
results are given in table 27. The collective doses from3
H and 1 2 9 I in liquid effluents. which were considered95
separately above. are also listed, as are those fromZr.
95
Nb, 144 Ce, and 239 Pu. The estimates are based on
the operating experience of the Windscale plant for
releases to salt water and of the NFS plant for releases tofresh water. For most radionuclides, they are somewhat
205

TABLE 26. WCAL COLLECTIVE DOSE COMMITMENTS FROM PARTICULATES IN AIR·
BORNE EFFLUENTS FROM A FUEL REPROCESSING PLANT'Radio·
nuclide4 10-s
1 10- 6
8 10· 7
6 10-s
3 10- 6

t06Ru

u,cs

134Cs

9osr
B9Sr

Collective dose (man rad per MW(e) y)

Normalized
activity
released(Ci per MW(e) y)

Total

Total
body

Lung
(whole)Bonemarrow

7.6 10-s
6.6 10-s
4.1 10-s

2.2 10-•0.06 10-6
0.06 10- 6
0.02 10-•0.03 10-6

0.04 10-s
0.05 10- 6
0.06 10- 6
6.1 10· 3
0.02 10-s

10-•

2

2

10-•

10- 3

6

Bone lining
cells

8.4 10- 3
8

10· 3

aExample based on NFS data, 1971-1972 (table 24).

TABLE 27. COLLECTIVE DOSE FROM RADIONUCLIDES IN LIQUID EFFLUENTS FROMFUEL REPROCESSING PLANTS

Radio•

nuclideNormalized
activity
released(Ci per MW(e) y)Collective dose (man rad per MW(e) y)
Total
body

Large lower
intestine

Bone

ThyroidRelease to salt Water

aH

13

13'Cs

14
10
2.3

9szr
134Cs
144Ce

5.5

'06Ru

12
9.4
0.5
6.1
0.4
0.6
- 0.001

9sNb

1oaRu
90Sr

s9sr
239Pu
129I

0.000002
0.08
0.01
0.01
0.005
0.00030.00010.000002

0.006
0.05
0.05
0.006
0.00006
0.01
0.00001
0.001
0.0007

Total

0.1

0.1

0.01

0.0007

Release to fresh water

'H

137Cs
134
Cs
106

Ru

90

Sr

89

Sr

239 Pu
129

1

6
0.015
0.00280.075
0.013
0.0033
7.4 10- 6
2.5 10°"
Total

0.04
0.03
0.004
0.00040.4
0.00030.000002
0.3
0.07

larger than the previously published estimates (91, 100,140), primarily because only small, identifiable. local
populations had been taken into account in making theearlier estimates. Generalized environmental dispersion
has been assumed in the assessments of table 27,
however, and differences from more realistic site-specificanalyses are expected. For example. the 1 37 Cs releases
in liquid effluents from the Windscale plant during
1972/73 have resulted in an estimated collective dose to
the entire fish-consuming population in Western Europe(106) of 3000 man rad (half of which is to the United
Kingdom population), corresponding to about 1.2 man
rad per MW(e) y.
6. Plutonium, americium. curium
167. Both the inhalation and the ingestion pathways to man of the transuranic elements have been
investigated, with important data on widespreadenvironmental effects coming from fallout studies
206

0.02

0.02

0.4

0.3

(Annex C). Extremely low concentrations were found infoods; this fact, combined with a very low uptake fromthe gut, indicates that ingestion contributes very little tobody content. Inhalation intake could conceivajly be ofmore significance. Annex C discusses the inhalation
pathway and the problems of resuspension of plutonium
deposited on the soil. Plutonium in airborne effluentshas not been reported for fuel reprocessing plants.
However, there are data available on total alpha activity
released ( 106).
168. For releases in liquid effluents, no unusual
concentrating mechanisms seem to exist, although somehigh concentration factors have occasionally beenreported (154). Plutonium is largely removed tosediments. Hetherington et al. (48) have discussed someof the detailed studies of plutonium behaviour in themarine environment following discharge from theWindscale reprocessing plant. Approximate estimates ofthe local and regional collective dose commitments areincluded in table 27.

V. RADIOACTIVE WASTE STORAGE
AND DISPOSAL
169. Only a small fraction of the radioactivity
generated in the nuclear fuel cycle is released to theenvironment during nonnal operations. Except for
uranium, plutonium, and certain other useful nuclides
present in spent fuel, the radioactivity associated withthe fuel cycle is considered waste, which must be
subjected to suitable treatment and storage or disposal.
Storage is taken to mean any arrangement intended toallow the waste to be retrieved if necessary at any future
time; the waste may be temporarily inaccessible, but its
position and other details have been documented andthe means are available to retrieve it. Disposal implies arelinquishing of control over the waste.

A. LOW· AND INTERMEDIATE-LEVEL WASTE
170. The tailings from uranium mills are one example
of low-activity solid waste. Common fonns of solid
waste arising in most types of reactors include spention-exchange resins, air and liquid filters, reactor
components (control rods, replaced pressure tubes),
reactor instruments, protective clothing and miscellaneous paper, plastic and metal wastes. In addition,
miscellaneous •.wastes (e.g., obsolete buildings andequipment) may arise from any installation.
171. Solid wastes can often be reduced in volume,
either by compression or incineration. They are thenplaced in storage or disposed of at surface- or deep-burialsites. Most nuclear sites have some provision for storageof solid wastes. Some reactors provide for storage ofsolid wastes throughout the reactor lifetime. The final
disposition of these wastes will depend on the decision
as to the final method of disposal of the reactor itself.
172. It is common practice to dump low-activity solid
waste untreated into trenches and cover with earth. At
less suitable sites, or if more hazardous materials are to
be buried, additional barriers are needed to retard thetransfer of actjvity from the waste to water supplies.
These may be provided by constructing concrete wells or
pits, or by incorporating the waste in concrete orbitumen.
173. The limitations of shallow burial have been noted.
In particular, if significant quantities of long-lived
radioisotopes such as 9 0 Sr and 1 37 Cs are included,
satisfactory containment is required for a period
sufficient for decay. Surface burial is generally
considered unsuitable for waste significantly contaminated with transuranium elements, except as temporarystorage (108). Since 1970 it has been a requirement in
the United States that all solid wastes with significant
contamination of transuranium nuclides be segregatedand stored retrievably (142).
174. Suitable conditions for a surface-burial site cannot
always be found owing to dense population or
unfavourable geological or hydrological conditions. The
possibility of using abandoned mine workings andnatural geological formations as possible sites has
therefore been investigated.

175. One of the most thoroughly investigated repositories is the Asse II salt mine in the Federal Republic of
Germany. After acquisition of the mine in 1965, studies
were made of the stability of the mine room, pillars andsafety roofs, the safety of the shaft, and thehydrogeology of the environs. The completion of thestudies and a radiological safety assessment enabled thefirst batch of low-level wastes to be deposited in 1967.
The depth of storage of these wastes is about
750 metres. The mine has been used for both low- andintennediate-level radioactive wastes, which generate
negligible heat. It is estimated that this one mine will
suffice for the low- and intermediate-level wastes
generated in the Federal Republic of Germany until
2000 (109).
176. Out of many sites explored in Austria, a gypsum
mine and an artificial cavern of 20 OOO m 3 were foundsuitable for pennanent storage of low- and intermediate-level radioactive waste, provided that certain
additional construction work was undertaken (108).
177. In the south of Spain, several old mines (inparticular the Beta mine north of Cordoba) were also
suitable for the permanent storage of low- andintennediate-level wastes.
178. The possibility of using the sea to receive anddilute low- and intermediate-level radioactive wastes has
been recognized for several decades. The first deliberate
introductions of packaged radioactive material into thesea were probably by the United States into the Pacificin 1946. Since about 1949, the United Kingdom andother European countries have made similar disposals inthe deep waters of the north-eastern Atlantic Ocean. As
a result of the United Nations Conference on the Law of
the Sea in 1958, IAEA has produced recommendations
that form the basis for deep-sea disposal operations.

B. HIGH-LEVEL WASTE
179. The high-level liquid waste, generated if fuel isreprocessed, contains more than 99 per cent of thefission product and actinide activity in the fuel, except
for plutonium and uranium, which are present to theextent of less than 1 per cent. Although some may bedisposed of in the liquid state, it appears likely that therest will be solidified in some manner to render its
handling, transport and storage safer and to facilitate its
eventual disposal. By the end of the century, the total
high-level liquid waste produced throughout the worldconcomitantly with about 10 7 MW(e) y of nuclear
electric power will be equivalent to about 104 m 3 ofsolidified material.
180. The view has been expressed that present storage
methods are satisfactory, both on economic and on
safety grounds, as an interim measure for the present
and near future.
181. Various disposal strategies for these high-levelliquid wastes are discussed in detail in several
publications (6, 7, 75, 152, 167). Those alternative
strategies which have been subjected to the most207

'fl"""'"

I
intensive scientific and engineering study on the basis of
their promise of providing satisfactory isolation of thewastes from man's environment and of being feasible
economically include:
Solidification, followed by eventual disposal
into g~ological formations
Solidification, followed by eventual disposal
under or on to the sea bed
Storage, followed by disposal of the liquid
wastes directly into geological formations
These three alternatives remain the subject of intensive
scientific and engineering study. In addition, some othermethods have been examined for their scientific andeconomic feasibility including disposal of solidified
waste in the polar ice caps; extra-terrestrial disposal; andseparation, and nuclear transmutation, of the actinide
content of the waste.

185. The very long-lived nuclides, 14 C (5730 y) and
12 9
I (1.6 10 7 y), pose a special problem. Rather
extensive extrapolation into the future is necessary tocover their mean lifetime. In this regard the concept of
the incomplete dose commitment, introduced in
Annex A, is of special utility for assessing the future
maximum per caput annual doses resulting from acontinuing but finite practice. Thus, if one can foreseethat power production by nuclear fission will have afinite duration, which could be of the order of several
hundred years, the incomplete dose commitment can be. used to estimate the ma..x.imum annual per caput dosethat will be experienced in the future due to thatpractice. Although the exercise of computing thecomplete collective doses can be carried out, assuming
an ultimate size for the world population, thesignificance of the results cannot be readily interpreted.

A. TRITIUM
182. To date, no national authority has reached a final
decision on the precise future strategy that it will follow
on storage and disposal of high-level waste. Indeed, nomajor disposal operation involving high-level waste from
nuclear power production has yet been undertaken by
any country, and in all countries generating this waste itremains in storage.

183. Doses resulting from occupational exposure in
waste-storage operations are covered in Annex E.
Because high-level wastes from nuclear power production continue to be maintained in storage by national
authorities, with the precise method of eventual disposal
undecided, the Committee feels that it is unable at thisstage to make an adequate assessment of collective dose
commitment for the world population from disposal ofthese wastes. However, the Committee is satisfied thatthe collective dose contribution from present wastestorage practices is very small compared to thecontributions from other parts of the nuclear fuel cycle.

VI.

COLLECTIVE DOSE COMMITMENTFROM GLOBALLY DISPERSED RADIONUCLIDES

184. The radionuclides that contribute to the global
collective dose are those which are sufficiently long-lived
to achieve widespread distribution in the environment.
These include 3 H, 1 4 C, 8 5 Kr and 129 1. Other
longer-lived radionuclides, such as 1 3 7 Cs and 2 3 9 Pu,
probably become less widely dispersed after deposition
in soil or sediments following release in a local region.
Consideration of the global dispersion and the potential
dose contributions has appeared in several reports (5, 11.
22, 74, 88, 92). Estimates of the individual andcollective dose commitments from 3 H and 8 5 Kr are
becoming fairly well established, as experience has led toa good understanding of their environmental behaviour.
For radionuclides discharged in liquid effluents, themethods discussed in chapter IV may be applied as afirst approximation.
208

186. The collective dose to the global population fromreactor and reprocessing plant releases of tritium can beassessed from the relationships determined for the 3 H
released into the environment by nuclear explosions.
The dose commitment in that case is estimated to be
2 mrad from 2500 MCi released, or 0.8 10- 12 rad Ci- 1
for the northern hemisphere (Annex C). This estimate
was derived from measurements of the environmental
persistence of fallout 3 Hin surface waters.
187. A rough estimate of the dose commitment for 3 H
can be obtained by noting that fallout tritium reachedlevels of around 4000 pCi 1-1 in surface waters and that
subsequent decline of concentration was exponential
with a half-time of 3.2 y (12). This corresponds to an
integrated concentration of 18 400 pCi y 1- 1 , or about
20 OOO pCi y 1- 1 including levels occurring prior to the
maximum concentration. Assuming that the total water
intake by man in food, air and drinking water is about
3.1 I d- 1 , and that the tissue dose per unit activity oftritium intake is 84 rad Ci- 1 (12), a transfer factor of
9.5 104 rad per Ci y 1- 1 is estimated. The dose
commitment from the integrated fallout tritium levels isthus estimated to be about 2 mrad, in agreement withthe estimate of the previous paragraph.
188. An alternate approach to estimate the dose
commitment (1 I, 22) consists of assuming that all
significant discharges of 3 H as tritiated water in the near
future will occur in the northern hemisphere, and thatthe tritiated water will be dispersed in the circulatingwaters of that hemisphere ( about 10 2 2 g). It is also
assumed that the time of exchange with the southernhemisphere and with waters below the thermocline issubstantially larger than the half-life of tritium. Thisapproach leads to a lower estimate for the dose
commitment, because the higher concentrations of 3 H in
surface waters before complete mixing occurs are
neglected.
189. The collective dose commitment per unit activityof 3 H released can be assessed by the procedures
outlined in Annex A. In the time scale of interest, thefunction describing the population growth can beapproximated by an exponential with a fractional

increase rate a about 2 per cent per year. The present
northern hemisphere population N 0 is estimated to beabout 3.2 10 9 (100). The collective dose commitment
sc is related in this case to the dose commitment Dc by
the equation

where AE is the effective decrease constant of 3 H in
surface waters. For assessments based on the model
presented in paragraphs 186 and 187, XE is equal to
1/4.6 y {the effective mean life). With the value of Dc
given in paragraph 186, the collective dose commitment
is estimated to be about 3 10- 3 man rad per curie oftritium released. For assessments based on the model
outlined in paragraph 188, AE is the decay constant of3
H. The collective dose commitment for this model isabout 1 10- 3 man rad Ci- 1 , probably an underestimate
for the reasons given in paragraph 188.
190. Using the value of 3 10- 3 man rad Ci- 1 for theestimated reprocessing plant release of 16 Ci per
MW(e) y (para. 146), the collective dose commitment isassessed to be about 5 10-2 man rad per MW(e) y.
Releases of tritium from reactors (tables 11 and 12)
make additional contributions of the order of I per centof this value in the case of BWRs and GCRs, 20 per centin the case of PWRs, and on the average 130 per cent in
the case of HWRs.

193. Machta has developed a model which takes
account of dispersion in the first 2 y after release (88).
The local wind-rose is used for the fust 6 h (about
100 km of travel), air trajectory climatology from 6 h toabout 3 d, mean low-level airflow from 3 to 30 d,constituting the first hemispheric pass, and finallynorth-south and vertical mixing until uniform airconcentrations are achieved. For the reprocessing plantrelease of 37 5 Ci per MW(e) y, the combined collective
dose contribution of these steps in the first two years isabout 0.01 man rad per MW(e)y, using an average gonadal
dose factor of 1.1 104 rad m 3 Ci- 1 y- 1 • Variations in
the local population around the release point can
introduce variations of about 10-20 per cent in theestimated collective dose contribution during the firsttwo years following release, but there is little difference{l-2 per cent) in the subsequent collective dose
contribution.
194. From the third year after release on. the airconcentration is subject only to radioactive decay. The
collective dose commitment can be calculated by theprocedures given in Annex A, assuming an exponential
population growth in the time scale of interest, with afractional increase rate a of about 2 per cent per year
and a population at the beginning of the third year N 0
of about 3.9 10 9 • As 1 Ci of 8 5 Kr after uniform mixing
in the atmosphere ( 4 10 1 8 m 3 ) gives an average gonadal
dose rate R = 2.8 10- 15 rad y- 1 • the collective dose
commitment per unit activity released is

sc =f"' b (t) N (t) dt= ~o No =2.410- 4 man rad Ci- 1
B. KRYPTON-85
191. Krypton-85 {10.7 y half-time) decays 99.57 percent of the time by beta-ray emission (mean energy
0.251 MeV). In 0.43 per cent of the disintegrations alower-energy beta ray (mean energy 0.0475 MeV) and aphoton (0.514 MeV) are emitted. The dose to the
gonads and bone marrow is primarily due to photons
from 85 Kr decay in air. When a person is immersed in alarge volume of air containing 8 5 Kr, the dose rates perunit concentration of 8 5 Kr in air are 0.62 I0 4 , 1.6 104 ,and 1.8 104 rad y- 1 per Ci m- 3 in testes, ovaries, andred bone marrow, respectively (104). Beta decay of
85
Kr in air contributes almost 100 per cent of the dosein skin, and beta rays originating in the airways of Jungaccount for 58 per cent of the lung dose. The doses perunit concentration of 8 5 Kr in air are 1.8 10 6 rad m 3
Ci- 1 y- 1 for skin and 3.1 104 rad m 3 Ci- 1 y- 1 for lung(104).
192. The collective dose commitment from 8 5 Kr
generated during nuclear power production is due almost
entirely to the release from reprocessing plants, therelease from reactors being quite small in comparison. As
the solubility of krypton in water is negligible, there isno sink for the krypton released into the atmosphere,
and rather uniform concentrations are achieved in aperiod of about two years. In the 1972 report of theCommittee. the distribution of krypton was taken to bealmost homogeneous over the surface of the globe andthroughout the troposphere. This assumption issubstantiated by results of measurements of 8 5 Kr in
surface air at different latitudes (33, 112).

oA-a195. For the reprocessing plant release of 375 Ci per
MW(e) y, the activity remaining after two years is about
330 Ci per MW(e) y and the resulting collective dose
commitment is about 0.08 man rad per MW{e) y. The
total collective gonadal dose commitment from 8 5 Kr,
including the contribution of the first two years. istherefore estimated to be 0.09 man rad per MW(e) y.
The corresponding values for bone marrow and lung are
0.15 and 0.25 man rad per MW(e) y.

C. CARBON-14
196. The dose commitment from 14 C releases from
nuclear power installations can be assessed by assumingthat the environmental distribution and behaviour will
follow that of natural 14 C. The dose commitment perunit activity released has been estimated by thisprocedure to be approximately 4.6 10-s rad Ci- 1 forthe whole body, 2 10-s rad Ci- 1 for gonads and Jung,and about 7 10-s rad Ci- 1 for endosteal cells and bonemarrow {Annexes B and C).
197. As this dose is delivered over a very long time,
much longer than the time required by the world
population to reach an equilibrium value. it can beshown (Annex A) that the collective dose commitment
is, very approximately, given by sc ""'DcNoo, where Noo
is the assumed upper bound of the population size.
Assuming a value of 10 1 0 for Noo. the collective dose
commitment per unit activity release is 460 man rad209

Ci- 1 for the whole body, 200 man rad Ci- 1 for gonads
and Jung, and 700 man rad Ci- 1 for endosteal cells andbone marrow.
198. The combined 14 C release from light-water
reactors and reprocessing plants (about 0.02 Ci per
MW(e) y) gives therefore a whole-body collective dose
commitment of 9 man rad MW(e) y, while the value forgonads and lung is about 4 man rad per MW( e) y and thevalue for endosteal cells and bone marrow is about
14 man rad per MW(e) y. As is discussed in Annex A, it
would not be realistic to use these collective dose
commitments to assess the maximum per caput annual
dose in the future, because it cannot be assumed that thepractice will continue for the long time required toachieve equilibrium. The incomplete collective dose
commitment is the proper quantity for this purpose.
199. Assuming that power production by nuclear
fission will last for a few hundred years (for example
500 y) it is possible to calculate the incomplete
collective dose commitment by an appropriate model forthe carbon cycle . in the environment. Using theenvironmental model that incorporates thick diffusive
layer between the surface and deep oceans, discussed in
Annex C, the incomplete whole-body collective dose
commitment is estimated to be about 2 man rad per
MW(e) y. For gonads and lung, the corresponding value
is 0.9 man rad per MW( e) y and for endosteal cells andbone marrow 3 man rad per MW(e) y.

a

For computing the dose commitment, therefore, therelease of 1 Ci of 1 2 9 I can be considered to result in a
specific activity of 2.6 10- 17 Ci g- 1 , giving an averageabsorbed dose rate to thyroid of 1.4 10- 11 rad y- 1 .
202. Assuming that about half of the 1 2 9 I present inspent fuel is discharged in liquid releases (para. 149),
that 10 10 is the upper bound of the world population,and that only decay decreases the availability of 1 2 9 L
the collective dose commitment to the thyroid isestimated to besc= (6.s 10-4 __g_.) (1.4 10-11 rad_) ( 1.6 10MW(e).y

y CI

7

0.693

y)

(10 1 0 man)= 2100 man rad per MW(e) yAs discussed in the case of 14 C, a dose commitment
delivered over such an enormous amount of time is of
doubtful meaning. In particular, it cannot be used forassessing future maximum per caput annual doses
resulting from a continuing practice. A more usefulquantity in this case, the incomplete collective dose
commitment, can be calculated with the assumption thatpower generation by nuclear fission will last for a few
hundred years. Using 500 y as a representative value, theincomplete collective dose commitment to the thyroid isestimated to be 0.05 man rad per MW(e) y.

VII. MISCELLANEOUS CONTRIBUTIONS
D. IODINE-129
200. The extent to which locally released 1 2 9 I
participates in the global iodine cycle is not certain, but
because of its very long half-life (1.6 107 y) it may beconsidered to become widely distributed. The specificactivity approach has been the usual method to assess
the thyroid dose commitment from 12 9 I discharges. A
specific activity of 1 pCi g- 1 stable iodine in thyroid
tissue results in an absorbed dose rate of 0.9 10-7 rad
y- 1 in the thyroid of a 6-month-old child, 2.1 10- 7
rad y- 1 in a 4-y-old, 3.5 10- 7 rad y- 1 in a 14-year-old,and 7.2 10- 7 rad y- 1 in the adult thyroid. The absorbed
dose rate increases with age because the iodine
concentration increases, being 80, 180 and 300 ppm atages 6 months, 4 y and 14 y respectively (119), and
600 ppm in the adult thyroid (68). An average absorbed
dose rate, weighted for an assumed population age
distribution (para. 117), is 5.5 10- 7 rad y- 1 per
pCi g- 1 •201. Stable iodine is non-uniformly distributed innature. The largest fraction, however, is in sea water.
Evaporation of sea water contributes most of theexchange between hydrosphere and atmosphere. Atmospheric water is exchanged within about 10 d. and it
follows therefore that there is a rapid exchange of
iodine. Assuming 102 2 g to be the mass of circulatingwater with an iodine concentration of 0.064 ppm (139),
there is an initial mixing capacity for 129 I of 6.4 10 14 gof stable iodine. Over the mean life of 1 2 9 I, exchangewith the deep ocean must also be considered. Uniform
mixing is achieved within 100-200 y. The mixing
capacity for 129 I is thereby increased by a factor of 60.
210203. Several operations conducted in support of fuelcycle installations may make additional contributions tothe collective dose commitment from nuclear power
production. Examples of such operations are thetransportation of materials between fuel cycle installations and some research and development activities.

A. TRANSPORTATION
204. Part of the operating requirement of the nuclear
fuel cycle is the need to transport radioactive materials:
unirradiated fuel from the fabrication plant to thereactor, spent fuel from the reactor to the reprocessing
plant, and solid wastes from the reactor and reprocessing
plant to waste storage or burial grounds. During thistransportation, two groups of people will receive
radiation doses: the transport workers and the generalpublic. The contribution made by the exposure oftransport workers to the total collective dose resulting
from the generation of nuclear power is discussed in
Annex E.
205. Nuclear fuel and radioactive waste materials areusually shipped by truck, rail or barge. The transport ofradioactive material is subject to regulations, which invirtually all countries are based on the IAEA model
regulations (58). The number of shipments and theshipping distances may vary greatly from installation to
installation and from country to country. Typicaldistances and numbers of shipments to support a large
reactor have been published in the United Kingdom(169) and the United States (145).

TABLE 28. ESTIMATED COLLECTIVE DOSE TO THE PUBLIC DUE TO SHIPMENTS TO AND FROM REACTORS(man rad per MW(e) y)

ReactorCountry

UnitedKingdom
UnitedStates

type

GCR

LWR

PopulationgroupaIrradiated fuelUni"adiated fuel

A
B

2
l

10-s
10-s

A
B

4

10-1

1.3 10-•(l 10 6)

1
3

10~
10~

(10 3 )
(3 10 6)(60)
(3 105)

1.2 10-1
1.6 10- 3(700)(7 105)(500)

Solid waste

Total} 4.3 10~
9.1 10-•

6.5 10~(600)(3 105)} 4.4 10- 3

Note: The number in parentheses is the number of individuals involved.
a A= individuals in close proximity; B = individuals along the route.

206. For purposes of estimating the collective dose, thepublic may be divided into two groups: those individuals
along the route who may come into relatively close
proximity to the shipment and those at some distancefrom the shipment as it passes. The collective doses tothe first group are estimated by the same procedure asthat described in Annex E. The collective doses to thesecond group are estimated by calculating the dose rate
as a function of distance from the package out to about
1 km, multiplying the average dose rate by the
population density along the route and integrating theresult over the distance travelled. A summary of thecalculated doses to the public is given in table 28.
207. The estimated total collective dose to the publicfrom nonnal transportation associated with reactor
operations is in the range of 4 10-4 to 4 10-3 man rad
per MW(e) y. The risk of additional radiation exposure
from accidents is extremely small. The probability of atransportation accident, of which less than 1 per centinvolve a severe impact or fire, is about 10- 6 pervehicle-mile (145). Combining the low probability ofaccident occurrence with the low probability ofradioactivity release in the event of an accident, it has
been concluded that collective doses from transport
accidents are an insignificant addition to the collective
doses from nonnal transportation (145).

B. NUCLEAR RESEARCH LABORATORIES
208. A portion of the radioactivity released to theenvironment from nuclear research laboratories may be
attributed to support for continued operation or future
development of the nuclear fuel cycle. However. otheractivities at nuclear centres, such as for example
radioisotope production and processing, and other typesof research, are usually responsible for a large portion of
the radioactive releases.
209. Reported estimates of collective dose to the local
population during 1973 from 28 research andoperational installations in the United States range from
10-6 to 200 man rad {160). The highest values are forthe largest research and production centres, SavannahRiver, Argonne, Mound. and Hanford. The totalcollective dose from all installations (about 450 manrad) divided by the electrical energy generated by
nuclear reactors in the country in the same year gives

0.05 man rad per MW(e) y. A much smaller, but
indeterminate value represents the portion of thiscollective dose incurred only in support of nuclear fuelcycle operations.
210. The collective dose to gonads from radionuclides
in liquid effluents from the research reactor installations
Winfrith and Dounreay in the United Kingdom havebeen estimated not to exceed 0.1 and 1 man rad,respectively, during 1972-1973 _by the shellfish pathway{100). From the Harwell Laboratories the estimatedcollective dose is 100 man rad by the drinking water
pathway from radioactivity released to the River Thames(100). Contributing to this value is the radioactivity,
primarily tritium, released from the radioisotope
production plant at Amersham (106).
211. Data on radioactive materials released from thenuclear research centres Jiilich and Karlsruhe and fromresearch reactor sites in the Federal Republic of
Gennany have been reported (55). Only for the
Karlsruhe laboratories are the releases of any significance. Individuals in the vicinity of the centre may
receive an estimated annual dose of 0.3 mrad due
primarily to fish consumption (55). The fraction
attributable to support of the nuclear fuel cycle cannot
be readily estimated.
212. In general, it can be concluded that thiscomponent of the collective dose commitment is verysmall compared to the contribution of other phases ofnuclear power production. The collective dose contribution of research and development through occupational
exposure is discussed in Annex E.

VIII. SUMMARY OF COLLECTIVE DOSE
CO!\lMIT.l\IENTS TO THE PUBLIC
FROM NUCLEAR POWER PRODUCTION
213. The preceding sections of this Annex contain the
results of the Committee's assessments of the collective
dose commitments, excluding the contribution ofoccupational exposure, associated with nuclear power
production. These collective dose commitments, per unit
electrical energy generated, are summarized in table 29.
211

TABLE 29. SUMMARY OF NORMALIZED COLLECTIVE DOSE COMMITMENTS TO THE
PUBLIC DUE TO NUCLEAR POWER PRODUCTION(man rad per MW(e) y)

GonadsWholelung

0.00001
0.00080.00004
0.1

ThyroidBonemarrowBoneliningcells

0.00001
0.00080.000020.00080.00020.00080.000004

0.00002

0.2
0.004
0.002

0.2
0.004
0.002

0.006

0.006

Local and regional contribution
Mining and milling
U, Ra, Tha

Rna
Fuel fabricationua

0.00001

Reactor operation b
Atmospheric pathways
Kr, Xe,

41

AI

'H
•4c

0.2
0.004
0.00060.2
0.004
0.0007

0.006

0.006

0.2
0.004
0.0006
0.1
0.006

0.03
0.01

0.03
0.01

0.03
0.02

0.03
0.01

0.03
0.01

0.00070.00020.001

0.002
0.00020.002

0.001
0.00020.006

0.001
0.00020.005

0.00020.00040.00070.00020.001
0.2
0.0002

0.006

0.008

0.04

0.04

0.04

0.04

0.09

0.09

0.04
0.3
0.09

0.2

0.2

0.003

0.003

0.003

0.003

0.003

0.1
0.09
0.9

0.1
0.25
0.9

0.1
0.09
0.9
0.5

0.1
0.15
3.0

0.1
0.15
3.0

1.5

1.7

2.0

3.8

3.8

"' I

'cs, Sr, Co, Ru
Water pathways'H
Cs, Co, Mn, I
Fuel reprocessing
Atmospheric pathways
85Kr

3H
14c

"'I, i2111
Cs, Ru, Sr
Water pathways

3H

1291

Cs, Ru, Sr
Transportation
External
Global contribution
Fuel reprocessing and reactor
operation

•H

85Kr

14ca
12, a

1

Total

0
The values are the incomplete collective dose commitments. The complete dose commitments
delivered over thousands to millions of years are as follows (man rad per MW(e) y):
Mining and milling:
U lung, 37; bone marrow, 6S; bone lining cells, 2l!O
Rn lung, 100
Fuel fabrication
U lung, 120. bone marrow, 220; bone lining cells, 920
Global14
C gonads and lung, 4; bone marrow and bone lining cells, 14
129
I thyroid, 2 100
hvaJues are averages based on the reactor-type distribution given in table 7.

For the contributions of the very long-lived nuclides, thevalues in the table are the incomplete collective dose
commitments, the complete commitments being given in
the footnotes. The use of these quantities is discussed in
Annex A, the incomplete commitments being ofparticular interest for assessing the future maximum
annual per caput doses in the case of continued
practices. The other required inputs for these assessments are the nuclear installed capacity and the world
population size. Considerable caution, however. is
required for ·extrapolation into the future. The activity
discharged per unit electricity generated has generally
decreased during the past several years because of212

improvements in technology and increases in regulatory
control. Since it is very difficult to predict what the
future course of technology and control will be, theextrapolation of the dose commitments to future power
production may not necessarily be valid.
214. The data summarized in table 29 indicate that
more than 50 per cent of the collective dose
commitment to the world population from nuclear
power production is due to the global dispersion of thelong-lived radionuclides 14 C, 8 5 K.r and 3 H. In somecountries. technologies are under development for
restricting the release of these radionuclides and also

12 9

I. While collection and retention of these radio·
nuclides will reduce the projected global collective dose
commitments, it can also present additional radiation
sources for potential local and regional exposures fromprocessing operations, handling, transportation, storage
and disposal of the collected concentrated radionuclides.
215. The collective dose commitment can be expressedas the period of time during which natural radiation
would have to be doubled to give an equivalent dose
increase. The collective dose commitments to variousorgans shown in table 29 for the production of
1 MW(e) y range from about 2 to about 4 man rad. For
the present installed capacity of80 GW(e) (table 7), andassuming a utilization factor of 70 per cent, the wholebody collective dose commitment per year of electricenergy production is about 1.3 10 5 man rad, corresponding to about 3.5 h of natural radiation exposure.

Appendix

COLLECTIVEDOSE
COMMITMENTFROMRELEASE OF RADIOACTIVE MATERIALSDUE TO ACCIDENTS

216. Although accidents involving the release ofradioactive material or unplanned personal exposure canoccur at any stage of the fuel cycle, most attention has

been focused on reactor accidents. This is not because
reactor accidents are more likely, but because thepotential consequences for the general public of somereactor accidents are much greater than from accidents
at other stages in the fuel cycle. In view of the low
accident frequency and limited consequences of
non-reactor accidents, only reactor accidents will becovered in this appendix.
217. So far, in several hundred reactor-years ofoperation, there has not been an accident in acommercial power reactor which has led to any exposureof the public. The frequency of accidents involving
radioactivity in the entire nuclear industry, including
research and development, has been very low and mostof those which have occurred have been at experimental
or pilot plants. Of these, only a few have involved
exposures of members of the public.
218. Although no power reactor accidents involving
exposure of the public have occurred, the commitment
to a large nuclear programme implies a commitment to acertain accident probability. In this sense. therefore,
there exists an "accident dose commitment" which isincurred by the operation of power reactors. Since thepast experience does not provide estimates of thefrequency of extremely unlikely events. any accident
analysis must be based on a probability approach whichtakes into account reliability considerations.

"'0w 10-•
w

u
X
w
..J
..J

3:

w

"'00
w

10-•

>

~

w

..J
..J

0
u
w

~
~
~
a:
<
w
>

10·•

.
a:
0
u

<
w
a:

10-,

a:

w
a.

.
>

:::;
iii

<

"'

0
a:
a.

,a-•

WHOLE-BOOY COLLECTIVE DOSES

Figure II. Probability distribution (or the whole-body collective dose per reactor year. Reactorof 1000 MW(e). (Uncertainties can be accommodated by allowing variations in the average curve offactors of 1/3 and 6 in probability and 1/3 and 3 in consequence.) (155)213219. The most comprehensive analysis to date (155),
based on a detailed study of the sequence of events or
faults that might lead to release. expressed its main
conclusions in terms of the probabilities of different
acci~ent releases per reactor-year of operation of large
light-water reactors. The doses and collective doses thatwould result from a given release were assessed by theuse of a consequence model, also stochastic in nature,
which took account of different weather conditions andpopulation distributions.
220. The probability of incurring an accidental
collective dose of a given value ( or within a given range)
can be obtained from the convolution of the functions
representing the release model and the consequencemodel. An example of such a combined function isrepresented in figure II, which gives the probability ofexceeding selected collective dose values.
221. In the example quoted above, the probability
density function (the probability increment per unit

214

collective dose) f(S} can be obtained by differentiationof the function of figure II. The accident collective dose
commitment ~cc, defined as the expectation ofcollective dose due to accidents, is the integral

where, in practice, the integration is carried out over the
possible range of collective doses. For large light-water
reactors (1000 MW(e)). this integration gives a value ofabout 250 man rad per reactor-year, implying anormalized accident collective dose coIJJ.mitment of0.25 man rad per MW(e) y.
222. It should be pointed out that the value quoted
above includes individual doses for which proportionality of dose and risk does not apply, e.g., in the case of
acute effects. Therefore, it is not possible to assess therelative detriment by comparing this accident collective
dose to the collective dose from normal operations.

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