IONIZING RADIATION:
SOURCES AND BIOLOGICALEFFECTS
United Nations Scientific Committee
on the Effects of Atomic Radiation
1982 Report to the General Assembly, with annexes

UNITED NATIONSNew York, 1982

NOTE
The report of the Committee without its annexes appears as Official Records of the General Assembly, Thirtyseventh Session, Supplement No. 45 (A/37/45).
The designations employed and the presentation of material in this publication do not imply the expression of
any opinion whatsoever on the part oft he Secretariat of the United Nations concerning the legal status of any country,
territory. city or area, or of its authorities, or concerning the delimitation of its frontiers or boundaries.

UNITED NATIONS PUBLICATION
Sales No. E.82.IX.806300PANNEXF
Exposures resulting from nuclear power production

CONTENTSParagraphs

Paragraphs

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

1-11

I. MINING AND MILLING ..••.......

12-32

A.
B.

15-24

Effluents ...................... .
Local and regional collective dose
commitments ...........•.......

25-32

II. URANIUM FUEL FABRICATION .. .

33-43

Effluents ...................... .
Local and regional collective dose
commitments .................. .

36-39

III. REACTOR OPERATION ........... .

44-141

A.

B.

A.

B.

C.

Effluents ...................... .
1.
Fission noble gases ........ .
2.
A<;ti_vation gases ........... .
3.
Tntmm .................. .
Carbon-14 .....•.•.........
4.
5.
Iodine ................... .
6.
Particulates in airborne effluents ..................... .
7.
Liquid effluents ........... .
Local. and regional colle.ctive dose
commitments .................. .
1.
Fission noble gases ........ .
2. Activation gases ........... .
Tritium .................. .
3.
4.
Carbon-14 .....•...........
5.
Iodine ................... .
6.
Particulates in airborne effluents ..................... .
7.
Liquid effluents ........... .
Reactor accidents ............... .

40-43

47-83
49.55
56-59
60-64
65-70
71-73
74-76
77-83
84-133
87-95
96-97
98-103104-105
106-112113-116
117-133
134-141

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

142-168

Effluents ...................... .
1.
Krypton-SS ............... .
2.
Tritium .................. .
3.
Carbon-14 ................ .
4.
Iodine ................... .
S.
Radioactive aerosols ....... .
6.
Liquid effluents ........... .

145-159149
150-151
152-153154-155
156-157
158-159

A.

B.

Local and regional collective dose
commitments .................. .
1.
Krypton-SS ............... .
2.
Tritium and carbon-14 ..... .
3.
Other atmospheric releases .•.
4.
Liquid effluents ...•..•.....
5.
Atmospheric and liquid effluents from a notional plant ...

V. COLLECTIVEDOSE
COMMITMENTS FROM THE GLOBAL DISPERSION OF RADIONUCLIDES ... .

A.
B.
C.
D.

160-168161
162-163164
165-166
167-168

169-188

Krypton-SS .................... .
Tritium ....................... .
Carbon-14 .................... .
lodine-129 .................... .
Summary ..................... .

172-174
175-178
179-183
184-187
188

VI. RADIOACTIVE WASTE STORAGE
AND DISPOSAL .................. .

189-200

E.

A.
B

Low and intermediate level wastes ..
High-level wastes ............... .

CONTRIBUVII. MISCELLANEOUSTIONS ..............•.............

190-193194-200

201-205

Transportation ................. .
Nuclear research Installations .... .

202-203204-205

VIII. SUMMARY OF NORMALIZED COLLECTIVE EFFECTIVE DOSE EQUIVALENT COMMITMENTS TO THE
PUBLIC FROM NUCLEAR POWER
PRODUCTION .................... .

206-211

A.
B.

Pagt

References .............................. .

327

249

Introduction
I. The use of nuclear reactors to generate electric
power has increased since the Committee's previous
assessment of the releases of radioactive materials fromthe nuclear fuel cycle and their resulting dose commitments in Annex D of the 1977 report [Ul). The total
installed nuclear generating capacity in the world in
1979 was about 120 GW(e) from some 235 powerreactors operating in 22 countries [11, K 12). This represents an approximate doubling of nuclear generation
during the years 1975 to 1979 covered by this Annex. In
1981 the installed electric generating capacity was 144.4
GW(e) from 261 reactors with 209.8 GW(e) (227
reactors) under construction [112). Projections of theworld nuclear generating capacity in the year 2000 are
somewhat speculative, but the figure seems likely to beabout 1300 GW(e), or a little over half of the estimate
reported by the Committee in Annex D of the 1977
report [Ul]. This expectation is based on revised
estimates of about 400 GW(e) nuclear generating plantin North America by the year 2000 [19), towards 350
GW(e) for the Soviet Union and Eastern Europe [U2,19), and the remainder being made up of perhaps 300
GW(e) in Western Europe [Wt, 19], 100 GW(e) forJapan [M8] and 150 GW(e) in developing countries. In
the International Nuclear Fuel Cycle Evaluation(INFCE) [19) a range of 1030-1650 GW(e) was
predicted for world installed nuclear capacity in theyear 2000.

2. The nuclear fuel cycle includes the mmmg andmilling of uranium ores, conversion to nuclear fuelmaterial, which usually includes enrichment of theisotopic content of mu, fabrication of fuel elements,
production of power in the nuclear reactor, reprocessing of irradiated fuel and recycling of fissile andfertile nuclides recovered, and disposal of radioactive
wastes. In addition, nuclear fuel materials are transported between installations at various stages in thecycle. In recent years several reports by the United
Nations Environment Programme have assessed theenvironmental impact of the nuclear fuel cycle [E7. US]
and in the INFCE studies radiation risks were also
considered [19).
3. Almost all of the artificial radionuclides associatedwith the nuclear fuel cycle are present in the irradiatednuclear fuel, although some neutron activation of structural and cladding materials takes place, and thenaturally-occurring radionuclides are present at theuranium mining and milling stage. The majority ofirradiated fuel elements are currently stored. However,where reprocessing takes place, the radioactive fission
products and transuranium elements are stored ashighly active liquors in tanks isolated from theenvironment. At each step of the fuel cycle releases ofsmall quantities of radioactive material to theenvironment may occur. Most of the radionuclides
released to the environment are only of local orregional concern because of their short radioactive halflives or limited environmental mobility. Some radionuclides, because of a combination of long radioactive
half-lives and rapid dispersal in the environment,
become globally distributed.
4. The interest of the Committee is in assessing theoverall health detriment and the doses to individual
members of the public due to the releases of radioactive
materials at each stage of the nuclear fuel cycle.
Because of the system of control applied to environmental releases from nuclear power installations, doses
250

to individual members of the public are generally well
below the relevant dose limits and correspond to lowlevels of individual risk. Individual dose levels decrease
rapidly with distance from a given source and the doses
to most exposed individuals vary widely from installation lo installation and between one location andanother. In this Annex an indication of the range ofdoses lo the most exposed individuals at each stage ofthe nuclear fuel cycle is given. To evaluate the total
impact of nuclides released at each stage of the fuelcycle, calculated results are first presented in terms ofcollective absorbed dose commitments to various bodyorgans or tissues, normalized to the production of unit
quantity of electric energy generation, expressed as
man Gy per GW(e) a. Annex A describes the way inwhich the absorbed doses may be combined to give thequantity "collective effective dose equivalent" which isassumed to be proportional to health detriment. Valuesof coll,ective effective dose equivalent commitment perunit of electrical energy actually generated are thus
calculated for each stage of the nuclear fuel cycle.
5. The collective dose commitment from nuclear
power arises in four population groups: the occupationally exposed, the population within a few hundred
km from the site, the population within a few thousand
km from the site, and the total wortd population. This
Annex deals only with exposure of the public, as
occupational contributions are dealt with in Annex H.
Each stage of the nuclear fuel cycle is considered separately and the local and regional dose commitments are
given for atmospheric and aquatic discharges. The
global contributions from those nuclides which
irradiate the world population are then discussed forthe fuel cycle as a whole. Finally, the disposal of solid
wastes from the nuclear fuel cycle is reviewed.
6. Collective dose commitments to local and regional
populations must be estimated by environmental
modelling, as the activity concentrations from effluentsfrom the nuclear fuel cycle are very low, both in thegeneral population and in environmental materials.
Monitoring of activity concentrations due to effluent
releases has concentrated on areas surrounding nuclear
facilities to ensure compliance with applicable regulations. In recent years computer models have beendeveloped which enable estimates to be made of doses
to large populations over long periods of time [ClO, El,
Ml]. The values of parameters for the transfer ofradionuclides in these models are taken from environmental monitoring results or from experimental observations.
7. The source terms for the releases of radioactive
effluents from nuclear installations are usually readily
available to the Committee and reflect the operating
histories, including abnormal periods of operation. The
Committee has reviewed reported discharge data andhas produced average releases per GW(e) a generated.
These normalized releases do not apply, therefore, toany one plant, but are deemed to be representative of
current nuclear power generation. They, therefore,
reflect differences in reactor design and changes inrelease rates between newer and older installations.
Future practices may give rise to results considerably
different from those derived from past and current
experience and any extrapolation to the future must be
undertaken with caution. To estimate the collective
dose commitments corresponding to these averaged
releases, the Committee has decided that a model
facility at a representative site be established for eachstage of the fuel cycle: mining and milling, fuel fabrication, reactor operation and reprocessing. The
environment receiving the typical releases from eachmodel facility was chosen to represent broad averages
containing typical features of existing sites andrenecting the most common environmental pathwa)'.s.
Such generalizations are intended to give dose commitments renecting the overall impact of the nuclear power
programme and will not be applicable to any one gi~ensite without due consideration of its own specificenvironmental pathways and of the particularradioactive release.
8. The calculation or the collective absorbed dose
commitment to any organ or tissue requires theintegration over infinite time of the collective absorbed
dose rate in that organ or tissue. As described in Annex
A it is convenient in evaluating the collective dose
c~mmitmenL and other similar quantities, to distinguish between external .a':d internal irra?iat!on o~ thebody. For external irrad1at1on the calculation 1s straightforward. but for internal emitters the integration_ iscomplex. particularly for nuclides of long retention.J times, when detailed knowledge is required of the time
variation of dose equivalent rates in body organs andtissues following intake. The Committee, therefore,
decided to represent the collective dose commitm:nt
for internal radiation as the integral of the collective
intake multiplied by the committed dose per unit
intake. The collective intake of a nuclide is a quantity
readily calculable and tabulations of dose per unit
intake are available. The Committee decided to use thecommitted dose, defined as the 50 year integral of dose
rate in any organ or tissue following intake, for the
average dose per unit intake in a normal population.
This may lead to a slight overestimate of average dose
for nuclides of long retention times in the body, since
the mean life expectancy of a population is a little lessthan 50 years. This formulation of collective dose
commitment can also lead to slight overestimates of its
truncated value for time periods comparable with themean life expectancy of the population. but the error issmall and compensated by the ready availability of
tabulations from which committed absorbed dose to theparticular body organs of interest may be obtained [12,Al. ClO].

9. In the estimation of collective dose commitments, it
is clear that assumptions have to be made about the sizeand habits of the exposed future populations. For thisstudy it is assumed that the magnitude of the world
population remains stable and that no major changes inage structure occur. It is further assum~d that t~edietary and other habits of the population remam
constant, which is a reasonable assumption for shorttimes, although the uncertainties must increase as
longer times are considered. Finally. it is assumed thatthe whole population is represented by adults for the
purposes of evaluating doses from inhalation andingestion. This assumption is valid when estimatingcollective dose commitments for populations, because
children only comprise a fraction of the total, andthe difference in their dose per unit level of activity
in environmental materials is small. except for a very
few nuclides such as those of transuranium elements
and 90Sr.
10. Very long-lived nuclides pose a special problem.
One example is 1291 (1.6 101 a). Another example isradon from mill-tailings containing 230'fh (8 104 a) and
238LJ (4.5 J09 a). Assessments of human exposures oversuch periods of time are obviously hypothetical andtheir relevance is doubtful. Dose commitments assessed

for the purpose of calculating the maximum dose ratein the future, however. should only include integration
over periods of time equal to the expected length of thepractice that causes environmental pollution. Assuming
this period to be 500 years for nuclear power
production. the long life-time of some nuclides will not
influence the assessment of such truncated dose
commitments. The problem therefore only arises when
the complete dose commitment is assessed. That
assessment will be extremely uncertain for the longlived nuclides and should therefore not be presented as
single figures but preferably in tables or diagrams
which indicate when the late dose contributions are
expected to occur. One long-lived nuclide, 14 c (5730 a),does not pose as great a problem as indicated by its
long half-life. The early disappearance of this nuclide
from the human environment causes as much as about
10% of the complete dose commitment to be delivered
within 7? years.
11. In addition to the small releases of radioactive
materials which take place during normal operation,
unplanned releases may occur. This Annex also reviews
reactor accidents which have led to unplanned releasesof activity into the environment, together with estimatesof the resulting collective doses.

I.

MINING AND MILLING

12. Uranium is obtained from ore mined in several
countries of the world, the major producers being
Canada, France, South Africa, United States. with
Australia soon to join them. Major efforts in uranium
exploration and production are also being made incountries such as Algeria, Argentina, Brazil. Gabon,
India, Iran and Niger, while many other developing
countries are making moderate efforts at uranium
exploitation [B25]. World uranium production was
38 OOO tin 1979 [H13] and the planned capacity for 1990
is 120 OOO t. However, the growth of nuclear power needs
to be more predictable in order to provide the incentive
and lead times to establish the necessary mining andmilling operations. Table 1 shows the 1979 uranium
production capability from the major producers andgives for guidance the estimated ''reasonably assured"
resources in those countries, assuming a price ofUSS 130 per kilogram of uranium. The estimates ofuranium resources vary widely depending upon theprice of uranium assumed; unconventional so~rces,
e.g., coals and lignites, are not included.
13. Uranium mining operations involve the removalfrom underground of large quantities of ore containing
uranium and its daughter products at concentrations up
to several thousand times the concentrations of these
nuclides in the natural terrestrial environment. The
concentration of uranium in mined ores is between 0.1
and about 3% U30s. The mining is carried out either by
underground or open pit methods and these techniques
accounted for nearly all new uranium production in
1979. Each method produced about the same amount of
uranium, although more ore was produced by open pit
mining, as it is generally of lower grade. Some underground mines are less than 30 m below the surface and,conversely, some open pit mines operate to depths of
150 m. In recent years in situ solution mining has been
carried out, although that, together with heap leaching
techniques, only accounted for a few percent of world
production. The main radioactive release from underground mining is 222Rn in the mine ventilation air,25]

while for open pit mining there are also radioactive
dust emissions.
14. Uranium
milling
operations
involvetheprocessing of large quantities of ore to extract thepartially refined uranium. The uranium concentrate iscalled yellowcake. which is then used as feed at the fuelfabrication plant, where it is further refined, converted
and enriched, if necessary, for use in reactors. Becauseof the large quantities of ore lo be processed. mills tend
to be near the uranium mines to minimize transportation. In some cases, where ore is treated locally byheap leaching, precipitated preconcentrates are transported to the mills. In 1979 more than 50 uranium mills
were in operation [G1, F5, Sl6] which processed over
65 million tons of ore. Approximately half of the ore
was new and the remainder was tailings from existing
operations such as the gold mining industry in South
Africa. The process of uranium extraction involves thefollowing steps: crushing, grinding, chemical leaching,separation of the uranium from the leach solution,
precipitation, drying and packing of the yellowcake.
The mill processes fall into three general types, acidleach solvent extraction, acid leach ion exchange, andalkaline leach; most mills use the acid leach solvent
extraction process. The steps in the milling process
which lead to the major emissions of radioactive
materials are the front end crushing operations and thedrying and packaging of yellowcake.

A.

EFFLUENTS

15. Gaseous radioactive effluents from mines arealmost entirely composed of 222Rn in the ventilation air
which is discharged in large quantities. Liquid wastes
result from mine drainage and process feed water; they
are discharged to ponds for settling of solids and thewater is either allowed to evaporate or is released to theenvironment. Mine drainage water can also be used asprocess feed for the mill. or may be diluted. treated anddischarged [C7]. Leaching of mine tailings may also bea source of liquid waste but generally wastes from themilling process are of more importance. Solid waste iscomposed of rock and very low grade ore. Only limited
information is available on radionuclide emissions
from underground uranium mines, and in general only
222Rn emissions are reported, as this gas represents the
major airborne radioactive component of the effluent[N21, L8]. Particulate emissions are believed to be farless significant. In a 1978 survey of underground mines
in New Mexico. l22Rn emissions per unit mass of ore
mined in the range of 0.4 to 8 GBq t-I were reported[Jl]. The variation is mainly dependent upon the gradeof ore mined, with ores containing up to 1 or 2%
uranium at Nabarlek in Australia giving 1-2 GBq t- 1
[L9) and 0.1% ores at underground Canadian mines
giving results 10 times lower [W6].
16. The major airborne radioactive component in
effluents from open pit mining is also believed to be
222Rn [El, C7, W6]. As large areas are involved, it is notpossible to directly measure the radon emissions, butthey may be inferred from emission rates from particular surfaces of the mine or from 222Rn concentrations
downwind of the mine. The results at the Ranger Minein Australia for radon release per unit mass of ore
produced during mining were 0.1 GBq t-1 [D7] and theemanation rate per unit ore grade is the same as
Nabarlek underground mine. Results for several open
pit mines in Canada average 0.2 GBq 1-1 for ore at252

about 0.25% uranium [E8]. Jn a recent study of eight
open pit uranium mines in Wyoming, an estimate of the
222Rn emission per unit mass of ore produced was 0.2
GBq t-1 [N3]. The grade of ore obtained from open pit
mines are typically 0.1-0.2% uranium and thus theradon emission normalized for uranium content seems
similar for underground and open pit mines at l GBq
per tonne and per cent of uranium.
17. Uranium ore requirements per unit of electricalenergy generated vary somewhat between thecommercial nuclear reactors currently in use. In the
recent I NFCE studies [19] the light water reactors(L WRs) are assumed to require 205.4 t of uranium
heavy metal to be extracted for the production of 1
GW(e) a of electrical energy, while heavy water reactors(HWRs) operating on the uranium cycle require 178.8 t[GW(e) a]-1 of heavy metal. It is estimated by theCommittee that gas cooled reactors (GCRs) using
natural uranium metal require 270 t [GW(e) a]-1 ofuranium metal, assuming a mean fuel burnup of 4.5
GW d t-1 and a thermal efficiency of 30%. whereas
advanced gas 1-eooled reactors require 219 t [GW(e) aJ-1of heavy metal. These uranium requirements are calculated on a "once through" basis in that no recycling of
uranium or plutonium is assumed. Were plutonium tobe recycled in LWRs or HWRs, the heavy uranium
metal requirements would be reduced to 120 and 75 t
[GW(e) aJ-1, respectively. The introduction of fast
breeder reactors (FBRs) would reduce the heavy metal
input to 1 t [GW(e) aJ-1 of uranium for plutonium
recycle, an improvement of a factor of about 200 in
uranium utilization. A heavy water reactor operating ona uranium-thorium cycle breeding mu would require
7 t [GW(e) aJ-1 of thorium metal.
18. The grade of ore mined is variable but is usually
between 0.1 and about 3.0% U 30g. In the United States
it is higher from underground mines at an average 0.2%
U30s, than from open pit mines with 0.11% U30s [El].
Thus the L WR uranium heavy metal requirement of
205 t [GW(e) aJ-1 corresponds to a mining rate of
1.2 1os t of ore from a United States underground mineor 2.2 10s t from a United States open pit mine. The
Nabarlek mine in Australia produced 2% U30s ore, sothat the LWR requirement is met by an ore productionof just over 1.2 104 t {L9]. However, since the corresponding normalized 222Rn releases appear to be
similar for underground and open pit mines per unit
concentration of uranium in ore, the Committee
estimates radon releases to be 20 TBq [GW(e) a)-I. This
compares with the Committee's estimate in Annex D of
the 1977 report [Ul] of 6.3 TBq [GW(e) aJ-1 from mining
and milling.
19. The activity content of the mined ore is predominantly due to 238U and its daughter products; there isvery little, if any, natural thorium in most ores.
However, ore from the Elliot Lake region in Canada
averages about 0.2% natural thorium. One tonne of ore
containing, say, 2 kg of U30g has an activity of 21 MBq
from each of the 14 principal members of the 23RU
decay chain, a total of about 0.29 GBq [Sl]. The
operation of uranium mills has thus resulted in theaccumulation of large quantities of waste tailings
containing significant quantities of uranium daughter
nuclides. There are currently known to be some 120
million tonnes of tailings stored at active mill sites
mainly in the United States and Canada [N21] andcurrent uranium demand trends would increase thisfigure to about 5 108 t by the year 2000. About 14% of
the total activity in the ore feed appears in the uranium
concentrate which achieves better than 90% uranium

extraction. In the resulting solid wastes, with the parent
nuclides removed and short-lived daughters 234Th (T112
24.1 d), 234mPa (T112 1.17 min) and 231Th (T112 25.5 h)quickly decaying. some 70% of the original activity
remains and is essentially due to 2JOTh (T112 8 104 a) and
its daughters.
20. Uranium tailings are discharged from the millusually in slurry form at about 50% solids, to an
impoundment area. Tailings piles typically cover areasof between 30 and 60 hectares. The tailings compriseabout 70% sand and 30% slimes and about 85% of theactivity is contained in the slime fraction [SI. N21].
Although in the early days of uranium mining andmilling liquid effluents were discharged to local watercourses [02). the present practice is to minimize liquid
efnuent by recycling or evaporation of the water.
21. In dry climates there are essentially no liquid
discharges which lead to radiation exposure of the
public. This has been confirmed by environmental
monitoring which has not detected measurable watercontamination beyond plant sites [S2, S3, S4]. Studies
on tailings piles in Illinois indicate that thorium andradium concentrations fall off to background levels
within 1 m below the bottom of the tailings and that theactivity does not migrate into groundwater [N2]. In .
contrast, Canadian mills operate in a wet environment
and the usual practice is to use small lakes and depressions for tailings disposal [W6]. Tailings are containedby dams and the overflows treated with barium chloride
to co-precipitate the radium as Ba(Ra)S04. Further
barium chloride addition and settling may be allowed
before discharge of effluents to waters available to the
public. After treatment, the dissolved 226Ra concentrations are usually less than 0.4 Bq 1-1 with 0.2-7 Bq J-1
suspended 226Ra [MlOJ. Typical annual releases of226Ra from tailings into the watershed are of the orderof 40 GBq, corresponding to less than 1 GBq [GW(e) a]-1.
22. Sources for atmospheric emissions of radionuclides are the ore crushing and grinding circuits. theyellowcake drying and packaging operations, and. thetailings. Emission rates from different plants vary
widely, owing in part to different process and control
technologies. For a typical mill processing about 2000 tof ore per day, the major source of atmospheric dust
emissions is from the yellowcake drying and packaging
processes; the reported ranges of emissions are 1-4
GBq a-1 for 238U; 0.2-2 GBq a-1 for 230'fh, 226Ra and
2!0Pb; and 1-7 TBq a-1 for 222Rn [N4, N5. N21. E8. D7,
N6, N7). More recent mills may achieve releases fromthe crushing and storage processes of about 0.04-0.16
GBq a-1 particulate emissions [N2, N8]. Atmosphericemissions from the tailings area are in the ranges of
7-500 MBq a-t for 238U and 2J4U; 0.1-8 GBq a-1 for
230TH, 226Ra and 21ops; and 500 GBq to 300 TBq a-1
for 222Rn [N2. N3, N4, N5, N6, N7. N8, N21, D7]. The
amount of particulate airborne emissions from tailingareas depends upon the size of dry tailings beach areas
which are subject to wind and weather erosion, while
the radon emission depends on diffusion from theground. The radon exhalation rate appears to be about
1 Bq m-2 s-1 per Bq g-t of 226Ra in the tailings [N2].
allhough the figure can vary by an order of magnitudedepending on meteorological conditions such as windspeed, atmospheric stability and rainfall [Tl]. The
release of radon from an uncovered tailings pilecontaining about 21 Bq g-t of 226Ra wilJ thus be about
21 Bq m-2 s-1. Tailings impoundment areas almost
completely covered by water wilJ have very low radionuclide emissions {N8).

23. To estimate the environmental releases, a model
mill facility has been established based on the data
given for the United States [El, N21. St]. The model
mill processes 600 OOO t per year of 0.2% grade ore.
Dust from ore crushing is assumed to contain allradionuclides in equilibrium but yelJowcake drying andpackaging contributes to the uranium release andaccounts for the lack of equilibrium in the release. The
milJ is assumed to produce uranium at a rate equivalent
to 5 GW(e) a per year and to operate for 20 years. The
normalized atmospheric releases from the milling
operations and from the tailings piles arc shown in
Table 2 together with the mining contribution from
222Rn. The typical tailing impoundment area assumed is
60 hectares with a radon emission rate of 6.7 TBq ha-I
a-1. The mill processes ore equivalent to 100 GW(e) aand thus tailings areas amount to 0.6 ha [GW(e) a]-1. It
is assumed that there is a 2 m covering of earth which
reduces the radioactive emissions by a factor of 4. The
dust emi!ision data for the tailings disposal area werebased on assumptions that average wind speeds
produced 0.017 g ha-I s-t of particles smaller than 10µm; 10-80 µm particles accounted for 0.04 g ha-I s-1
with assumed nuclide compositions of 3 Bq g-1 for nsu
and 60 Bq g-t of all the daughter radionuclides [El, St].
Activities of 235U and daughters are some two orders ofmagnitude below the values for 238U.
24. The tailings remain after the mill has ceased
operation and can become a long-term source ofradioactive contamination due to wind and watererosion, leaching and radon emanation. Hence thenormalized releases quoted in Table 2 for tailings arequoted per year per unit electric generation. Stabilization programmes are generally conducted or planned
so that erosion is alleviated using materials which may
be native soils, gravel or so-calJed Rip-rap cover, clays,
or even artificial or synthetic covers or sealants such asasphalt or polyvinyl chloride [N2, N21, C7]. Although
up to a few per cent of the original uranium isotopes
remain in the tailings, the major source of long-lived
activity for about 5 105 a is 2JOTh (T112 8 104 a) which
continues to produce 226Ra and corresponding radon
releases. The diffusion of radon in the top few metres oftailings is responsible for most of the release and therelease rate is independent of the depth of the tailings
beyond about 3 m [S5]. The reduction in radon
emanation by soil covering depends on the type anddepth of cover. In part because of its smaJJ particle size,
clay can hold moisture and present a more effective
barrier to radon diffusion than more common soils.
Hence, incorporation of clay can reduce the depth ofsoil cover required to achieve radon exhalation rates
similar to those of normal soils [N2].
B.

LOCAL AND REGIONAL COLLECTIVE DOSECOMMITMENTS

25. The uranium mining and milJing sites tend to be
in areas of low population density. In many areas theconditions are arid and not suitable for farming while
those mines in areas of high precipitation are usually
remote with again little local farming. In arid areas
there is negligible release of radionuclides to theaquatic environment and the radiation doses to the
public arise predominantly from airborne effluents. In
areas of high precipitation, it is established that theliquid effluent population doses are dominated by
226Ra in drinking water and aquatic foodstuffs [EB,KlO, W6]. In a detailed study of Canadian uranium
mines and mills which discharged radionuclides to253

lakes and water courses, the collective doses resulting
from drinking water, fish consumption and externalradiation from sediments were in general significantly
less than the collective doses due to the atmosphericreleases from the mine, mill and tailings areas [E8]. The
226Ra concentrations in surface waters were found to be
no higher than 10-1 Bq 1-1. The levels of maximum
individual dose reported were less than 80 µSv a-1 [E8].
26. Under the assumption that adequate waste
treatment prevents seepage of process liquid effluents
directly into rivers, the local and regional collective
dose commitments are assumed by the Committee todepend upon the airborne releases which contribute toexternal exposure from deposited material and tointernal irradiation via the inhalation and ingestion
pathways. In Annex D of the 1977 report [U1], theCommittee has previously used a population density of
3 km-2 within a few hundred km of the mine and milland a uniform population density of 25 km-2 from thisto 2000 km. These figures still remain suitable on thebasis of reported population densities in areas ofmining and milling [Sl, El, E2, E8, W6]. The collective
dose commitments to the local and regional populationfrom the particulate releases from the model mill
(Table 2) were derived using the values presented
for the naturally-occurring radionuclides in Annex C,
modifying the population density to 25 km-2 from thevalue of 100 km-2 used for estimating the impact of coal
fired power stations. A deposition velocity of 10-2 m s-1
was chosen for particulate releases and the majority of
the collective dose arises within the first few hundredkm.
27. For radon releases an atmospheric dispersion
calculation was undertaken to estimate collective doses.
The meteorological dispersion characteristics of therepresentative site are shown in Table 3. They are
typical of the local environment in arid areas and are
the values chosen in the Rasmussen study as applying
to a semi-arid area [R3]. Rainfall occurs for only 0.3%of the time and only in neutral conditions (Pasquill
category D); the effective height of release is 10 m. The
high incidence of category F conditions is a phenomenon local to the mine environment and the conditions are assumed to persist only up to 100 km. Beyond
that distance the meteorological conditions are assumed
to become neutral out to 2000 km. The assumption ofan equal frequency of all wind directions is consideredreasonable for this study since uniform population
distributions are assumed. The atmospheric dispersion
modelling has been fully described in Annex A and has
been utilized here in computer calculations using themethodology developed for the Commission of theEuropean Communities for the assessment of effluent
releases [CIO]. The dosimetry for radon inhalation wasthat assumed in Annex D. which gives an average
effective dose equivalent per unit inhaled activity ofradon daughters of 1.310-S Sv Bq-1. Only effective dose
equivalent calculations are presented for radon in thisAnnex. The equilibrium factor for the short-lived radon
daughters is taken as 0.6.
28. The resulting normalized local and regional
collective dose commitments from mining and millingare shown in Tables 4 and 5. For the particulate releases
inhalation, ingestion and external irradiation allcontribute to the dose commitment and a further
description of the analysis can be found in Annex C.
The resulting collective effective dose equivalentcommitment is approximately half that estimated forthe radon releases from milling operations. The total
254

normalized collective effective dose equivalentcommitment from mining and milling is 0.54 man Sv[GW(e) a]-1, of which 93% is from the mining operation.
The results are generally similar to those estimated by
the Committee in Annex D of the 1977 report [UI]. ll isassumed in all these calculations Lhat the indoor air
concentrations arc the same as outdoors.
29. The estimated doses to most exposed members of
the public are highly dependent upon the characteristicsof the particular location of the mine and mill. Annual
effective dose equivalents between a few hundred µSv
to several mSv have been estimated for typicalemissions from mines and mills [El, E8, N21. MIO]. For
the model mine and mill used by the Committee theannual effective dose equivalent at 500 m from thesource is 900 µSv from radon releases, assuming an
atmospheric dilution factor of 5 lo-6 Bq s m-3 per Bq
released, and doses from the particulate releases are 50
times rower.
30. As discussed in section I.A., 230Th in tailings piles
will provide a long-term source of radon emission. If
the release rate were to continue throughout the mean
lifetime of 1.1 ]05 a, then the emanation rate given in
Table 2 for 222Rn would give a normalized collective
effective dose equivalent commitment of about 2800
man Sv [GW(e) a]-1. The corresponding particulate
releases are estimated to give an additional 50 man Sv[GW(e) a]-1. A small amount of 230'fh would still remain
supported by the residual uranium in the tailings. These
results must be considered highly speculative becauseof the assumptions of the duration of the constantrelease and of the fixed population density and habits.
Table 6(a) shows how the normalized collective
effective dose equivalent commitment from the tailings
varies with the period of time over which the radon andparticulate releases are assumed to be released. The
results must be extremely uncertain over such geologic
time scales and indeed present day tailings
management may lead to radon emanation rates nogreater than the ambient levels for soils in the millvicinity, so that almost no long term dose commitment
arises.
31. In the INFCE studies (19] the dose commitment
from tailings was estimated on the assumption that theradon emanation continued for 103 a, by which time thetailings are assumed to have been eroded and lead to afurther dose commitment from the aquatic environment(freshwater and marine). The dose commitment fromthe aquatic environment is enhanced by the small
percentage of uranium left in the tailings, here assumed
to be 10% The regional marine model described in
Annex A has been used to estimate the collective dose
commitment from the tailings, assuming they are
eroded into coastal waters and disperse throughout theworld's oceans. The resulting normalized collective
effective dose equivalent commitments are shown in
Table 6(b) and, because of the nuclide content, thesedose commitments are almost independent of the timeat which they are released into the marine environment.
The normalized radon collective effective dose equivalent commitment for 103 a release is about 25 man Sv[GW(e) a]-1 and the uranium figure is 460 man Sv[GW(e) a]-1. These figures compare with the INFCE
estimates of 10 and 360 man Sv [GW(e) a]-1, respectively.
32. Previously the Committee has also used a
simplified model for estimating the normalized
collective effective dose equivalent commitment from

tailings by comparing with natural radon emanationfrom soil and the corresponding radon concentration in
air. With the model mine and mill data from Table 2
and the normal radon exhalation rate of 20 mBq m-2
s-1, leading to a standing equilibrium equivalent radon
concentration of 1.8 Bq m-3 (Annex D), and an effective
dose equivalent per unit of integrated air concentrationof 9.2 10-5 Sv per Bq m-3 a (Annex D), the normalized
collective effective dose equivalent commitment fromtailings over the mean life of 23DTh is calculated to be
S' = (6.7 TBq ha- 1 a- 1) (0.6 ha [GW (e) aJ- 1 (0.25)

1.8 Bq m- 3 )x (25 10-6 man m- 2)( 0.02 Bq m- 2 s- 1

( 1.44 x 80 103 a) (9.2 10- 5 Sv Bq- 1 m3 a- 1) xx ·(3.17 10- 8 a s- 1) = 760 man Sv [GW (e) aJ- 1•

This is in agreement with the value of 2800 man
Sv [GW(e) aJ-1 calculated using the atmosphericdispersion model (Table 6). The previous estimate by
the Committee, in Annex D of the 1977 report [Ul],based on the tailings release rate assumptions andradon dosimetry, was equivalent to a collective
effective dose equivalent commitment of 2300 man Sv[GW(e) a]-1. These values must be extremely uncertainand depend upon future practice, both in the choice offuel cycle, since plutonium recycle in fast reactors could
reduce the uranium ore requirement per unit energy
generated by a factor of about 200 with the resulting
same reduction in collective dose, and in the
management practices of the tailings themselves. Also
the downwards migration.-of radionuclides in soils at
rates of 2 10-3 m a-t could reduce the collective dose
commitment from radon releases by up to two orders ofmagnitude.

II.

URANIUM FUEL FABRICATION

33. The uranium ore concentrate produced at themills is further processed and purified and often
enriched in the isotope mu before being converted into
uranium oxide or metal and fabricated into fuelelements. Natural uranium which contains 0.7% 235U
can be utilized in graphite or heavy-water moderated
reactors (HWRs). Light water reactors (LWRs) andadvanced gas-cooled reactors (AGRs) require enrichedfuel of between about 1 and 4% mu. Before uranium
can be enriched it must be converted from the oxide
form U30s to uranium hexafluoride (UF6), the gaseous
form for use in enrichment plants.
34. Conversion takes place, for example, at two UF6
production facilities in the United States at Sequoyah,
Oklahoma and Metropolis, Illinois and in the United
Kingdom at Springfields, although other conversion
facilities exist. Two industrial processes are used forUF6 production, dry hydrofluor and solvent extractionand each is responsible for about half the UF6
production in the United States. The hydrofluor process
consists of reduction, hydrofluorination and fluorination of the ore concentrates to produce crude UF6,
followed by fractional distillation to obtain a pureproduct. The solvent extraction process employs a wet
chemical solvent extraction step at the start of theprocess to produce high purity uranium for the subsequent reduction, hydrofluorination and fluorination
steps.

mu

35. Enrichment of the isotopic content ofusually
takes place at a gaseous diffusion plant, although
increasing use is made of centrifuge techniques. At agaseous diffusion installation the UF6 is pumped
through a series of porous membranes which discriminate against the passage of the heavier isotope ofuranium by a factor of 1.0043 at each stage. Some
1700 stages arc required to produce an enrichment of
4% [U3]. Centrifuge technology utilizes rapidly rotating
cylindrical vessels to separate the isotopes of uranium
and utilizes only ten stages to reduce to the normal
tailings levels using only 10% of the electricity used by adiffusion plant [C7]. In the final fuel fabrication step
the UF6 is chemically converted to U02 or to uranium
metal for use in fuel elements. For use in LWRs or
AGRs the dioxide powder is sintered into pellets andloaded into zircaloy or stainless steel cladding to
produce fuel pins which are filled with helium andwelded with end caps. Uranium metal without
enrichment is used in Magnox reactors clad in amagnesium alloy (Magnox) can. For HWRs unenricheduranium dioxide is normally used. After the enrichment
process large quantities of depleted uranium remain inwhich the 235U content is 0.3% or more. This uranium
may become a source of public exposure if it isdisposed of: at present it is stored for possible use inbreeder reactors and for other purposes.
A.

EFFLUENTS

36. Emissions of radionuclides from the conversion.
enrichment and fuel fabrication processes are small.
Most of the uranium· compounds are solid and conventional equipment may be used to remove particulates
from airborne effluents. Liquid wastes are collected insettling tanks or ponds .. Estimates of releases fromtypical installations have been published previously in
the United States [U3]. More recently the United States
Nuclear Regulatory Commission has reported environmental releases from fuel fabrication plants at sixmonth intervals [N18].
37. Residual amounts of230Th and 226Ra are removed
from the uranium ore concentrate in the conversion
process and small amounts of these nuclides anduranium appear in effluent streams. Atmosphericannual discharges from typical conversion plants in the
United States are reported at about 3 GBq for 238U,
234U, 234Th: 33 MBq for 230Th; 3 MBq for 226Ra: and 74
MBq for 235U [El, N18]. In the United Kingdom.
British Nuclear Fuels Ltd. published annual reports onradioactive discharges and monitoring [B2, B22]. The
annual atmospheric discharge from Springfields has
been about 0.02 TBq of natural uranium from 1977 to
1979. The release of uranium from the Capenhurst
enrichment plant in the United Kingdom was reported
at 0.2 GBq in 1977 and 1979 and at 0.15 GBq in 1978
[82, B22]. The Swedish fuel fabrication facility released
between 13 and 60 MBq of 2 or 2.3% enriched uranium
per year to the atmosphere between 1976 and 1979 [G3].
38. Annual liquid discharges from Capenhurst
contain about 1 GBq of uranium and 0.5 GBq of beta
activity, and Springfields releases about 1 TBq a-1 ofuranium. The effluent from Springfields is released into
the tidal waters of the river Ribble, while waste from
Capcnhurst is now discharged through the sewage
system. The total beta activity in Springfields liquid
effluent discharges greatly exceeds the alpha activity
because of the removal and discharge of the short-lived
isotopes 234Th (24.1 d) and 234mpa (l.17 min). Annual255liquid eflluent discharges of mu at United States sites
are about 40 G8q for conversion, 67 GBq forenrichment and 17 G8q for fuel fabrication [U3].

near neutral conditions (categories C and D) andaccounts for some 10% of the annual frequency distribution (Table 8).

39. Table 7 shows discharges from the model fuelconversion. enrichment and fabrication facilities. The
atmospheric emissions of the model conversion facility
were based on operating data taken from Sears et aL[S6) for plant with low impurity feed. The atmosphericdischarges from the model fuel fabrication facility werebased on the work of Pechin et al. [Pl]. The model
enrichment facility is based on the Capenhurst
discharges to atmosphere [82). The throughput of fuel isassumed to be 104 t a-1 of uranium at the conversion
facility and enrichment plant, which is reduced to 1500
t a-1 of enriched fuel for the fabrication plant. The
liquid discharges from the model facilities were derived
from Capenhurst [82. 822] and United States data [U3].
The model plant is assumed to be at a site discharging
into fresh water. The discharge data for the model
facility are about half those used in Annex D of the
1977 report [Ul] for airborne releases and the same forliquid effluents.

42. The normalized collective absorbed dose equivalent commitments for the particulate releases areshown in Table 9, where it can be seen that inhalationof uranium isotopes is the main route of exposure.
These collective dose commitments are estimated on theassumption that the radionuclides deposited onto theground migrate downwards fairly rapidly and thus
become unavailable. In fact, owing to the long halflives of the nuclides of the uranium decay chain, they
may in the far future return to man by various
pathways, although the corresponding collective dose
commitments are probably small. The estimatedcollective absorbed dose commitments from ingestionof foodstuffs are estimated to be a factor of about 10
below those due to inhalation.

B.

LOCAL AND REGIONAL COLLECTIVE DOSECOMMITMENTS

40. The exposure as a result of liquid effluents from
United Kingdom fuel fabrication operations has beenreported as mainly due to external irradiation by the
sediments on the river bank near the Springfields plant.
The annual collective effective dose equivalent has
been estimated at less than 10-3 man Sv [T4]. For liquid
releases into a freshwater environment Pechin has
shown that the most exposed individual receives a doseof less than 10% of that received from the gaseous
release and has estimated a representative normalized
collective effective dose equivalent commitment of
about 2 104 man Sv [GW(e) aJ-1 to the population
exposed to aquatic pathways from the river [PI]. Thereis great variability in the behaviour of effluents released
to the aquatic environments and the collective dose
commitment can only be regarded as indicative of theorder of magnitude. It seems probable, however, thatthe major exposure of the population from fuelconversion and fabrication processes arises as a resultof the discharges to the atmospheric environment.
41. The location of the model facility has been chosen
to be representative of the northern United States andnorthern Europe. The most significant pathway thatemerges, which leads to exposure of the population, isthe direct inhalation route, as will be shown below. The
population distribution around the model facility is
assumed to be constant at 25 km-2 and the collective
doses for the particulate releases are derived using theresults given in Annex C for technologically enhanced
releases of radionuclides. The results are modified forthe different isotopic composition and population
density. The particulate releases are assessed for inhalation from the plume, ingestion of foodstuffs contaminated by activity deposited from the plume and by
external irradiation from the ground deposited activity.
For the radon releases a meteorological dispersion
calculation was performed using an effective dose
equivalent per unit inhaled activity of radon daughtersof 1.3 lQ-8 Sv Bq-1 and an equilibrium factor of 0.6
between radon and its short-lived daughters fromAnnex D. For the calculation of radon releases the
distribution of Pasquill weather categories is typical ofa moderate climate in which rain is assumed to occur in
256

43. In summary, the normalized collective effective
dose equivalent commitment due to uranium fuel fabrication is estimated to be 2 10-3 man Sv [GW(e) a]-1, asshown in Table 10. The main contribution is inhalationof the isotopes of uranium. Radon releases contributeabout 20% of the total. This figure is higher than that
estimated in Annex D of the 1977 report [Ul] for asimilar discharge, but is small compared with the doses
from mining and milling. It has been reported that thecritical group for uranium fuel fabrication plants arethose exposed to sediments on banks of waterways near
the site [B22]. For the Springfields site in the United
Kingdom this critical group may be exposed at 10-9 Sv
a-1, but the pathway contributes little to the totalcollective effective dose equivalent.

III.

REACTOR OPERATION

44. Most of the electrical energy generated by nuclear
power is produced by thermal reactors in which the fastneutrons produced by the fission process are slowed
down to thermal energies by use of a moderator. The
smaller the atomic mass of the moderator, the more
efficiently it removes energy from the neutrons. The
most common materials which have been employed asmoderators in thermal power reactors are light water,
heavy water and graphite. The choice of moderator
greatly affects the design of the reactor, its size and heatremoval system.
45. The uranium fuel is contained in discrete pins,
both to prevent leakage of produced radioactive fission
products into the coolant circuit and also to improve
the neutron economy by reducing the parasite neutron
captures in the resonance neutron energy region of
2JBU. The heat generated in the fuel pins by the slowing
down of the fission fragments is removed by forced
convection. The most usual coolants are light or heavy
water and carbon-dioxide gas. In the case of fastbreeder reactors, the neutrons are not mo~erated andinduce fissions with their energies close to those atwhich they were produced. The usual heat removal
system from the core is by means of liquid sodium
metal, which is a better heat transfer medium and does
not significantly moderate the neutrons.
46. The number of operational reactors of each type
and the generating capacity for each country or area
utilizing commercial reactors together with the installedcapacity per caput are shown in Table 11. There were
235 reactors in 22 countries with an installed capacity

of about 120 GW(e) in 1979 [11, K 12]. The reactor types
include the pressurized water moderated and cooled
reactor (PWR), the boiling water moderated and cooled
reactor (BWR), the Magnox and advanced gas-cooledgraphite moderated reactors (GCR), the light-watercooled graphite moderated reactor (LWGR), the heavywater moderated and cooled reactor (HWR) and thefast breeder reactor (FBR). The average installedcapacity per caput was 0.07 kW(e) with the highest
value being 0.46 in Sweden and the other developed
countries averaging between 0.1 and 0.2 kW(e).
A.

EFFLUENTS

47. During the production of power by a nuclear
reactor radioactive fission products are formed within
the fuel and neutron activation produces radioactive
components in structural and cladding materials.
Radionuclides are found in the coolant both because
the coolant becomes activated, because of diffusion offission product elements with radioactive nuclides fromthe small fraction of the fuel with defective cladding,and because of corrosion of the structural and cladding
materials. All reactors have treatment systems for theremoval of radionuclides from gaseous and liquid
wastes which arise from leakage out of the core or fromclean-up of the coolant. Low level releases which occur
are controlled and monitored.
48. The quantities of different types of radioactive
materials released from reactors depend on the particular design and on the specific waste treatment plantinstalled. Radionuclides released to the atmosphericenvironment include noble gases from fission (krypton
and xenon), activation gases (I4C. I6N, 35S, 41Ar),
tritium. iodine and particulates. Radionuclides
discharged to the aquatic environment in liquid
effluents include tritium. fission products and activatedcorrosion products. Results are presented for annual
normalized release, i.e., per unit of electrical energy
generated in that year, averaged over all reactors of agiven type (PWR, BWR. etc.). Normalized results are
not presented for individual sites because releases inany one year often reflect the need for maintenance orirregular procedures which are the result of a number ofprevious years operation. The total releases of radionuclides between 1975 and 1979 have been divided by
the total production of electrical energy over the same
years in order to obtain a representative normalized
release over the period covered by this Annex. These
results are subsequently used to assess collective dose
commitments. In the tables which follow open entries
mean that no data were available, entries characterized
by a dash mean that no releases were reported.

1.

Fission noble gases

49. There are at least nine radioactive isotopes ofkrypton and eleven radioactive isotopes of xenon
formed by the fission process. Most of them have very
short half-lives (seconds to minutes) and decay before
they migrate significantly within the fuel. A fraction of
the noble gas inventory of the fuel pins diffuses to thefree space between the fuel and the cladding, leading toa build-up of gas pressure. The presence of noble gasesin the coolant is generally due to fuel cladding failure.
50. In PWRs the primary coolant is in a sealed loop
and off-load refuelling is employed. Short-lived
radioactive noble gases, therefore, only appear becauseof leakages of the primary circuit water. The primarywater coolant in a PWR is continually purged forcontrol of chemical composition and purification.
Gaseous wastes are released in the process and are held
under pressure in tanks for between 30 and 120 days fordecay of short-lived nuclides. Other gaseous effluent
streams in PWRs originate from the condenser exhauston the steam circuit, secondary coolant blowdown,
reactor building ventilation, including containment
purges (about 4 a-1), and turbine plus ancillary building
ventilation [N9].
51. Table 12 lists the reported discharges of noble
gases from PWRs. The information is taken from
Phillips and Gruhlke [P2]. Beebe [823), Luykx andFraser [L 1, L6], Decker [D6]. van Daatsellar [D8].
Kumatori [K4], Errera [E9], Godas [G3], Norlinder[N19] and Salo [S14]. The releases span more than threeorders of magnitude partly because of variation indesign and partly because of the need for irregular
operations or maintenance. For this reason, the averagenoble gas release from PWRs was obtained by dividing
the total release by the actual total amount of electricalenergy generated. The normalized annual release has
been similar over the five years 1975 to 1979. The
average normalized release over that period was 430
TBq [GW(e) aJ-I. The composition of the release from
PWRs is predominantly from 133Xe (T112 5.3 d) but
some shorter-lived nuclides are present. particularlymxe (T112 9.2 h). In Table 13 the nuclide compositionof noble gases is presented for United States PWRs in
the year 1979 [B23] and the normalized release per unitof electrical energy generated is given. It can be seenthat 85K.r (T112 10.7 a) represents 5% of the normalized
release. Figures for the nuclide composition of noble
gases in European PWRs are essentially similar [Ll,L6].
52. In BWRs non-condensable gases in the steam flow
are continuously removed by the main condenser airejector system. This is the main source of noble gas
release into the gaseous waste stream. Secondary
pathways include the purging system for the turbine
gland seals, the condenser mechanical vacuum pump
and any leakage of process fluids to ventilated building
spaces [NlO]. Table 14 lists the discharges reported
from BWRs in consecutive years derived from references [823, D6, D8, E9. G3, K4, L1, L6, N19, P2, S14].
The releases vary by more than six orders of magnitude
per unit of installed capacity. The value averaged over
all operating experience has been falling year by year
from 56 PBq [GW(e) a]-1 in 1970, to 37 PBq [GW(e) aJ-1
in 1974 and to 4.4 PBq [GW(e) aJ-1 for BWRs in 1979.
This consistent reduction in discharges of noble gases
from BWRs is partly a result of the commissioning of
new plants with lower release rates, but is also due toreductions in discharges from many existing plants forthe same annual electrical output. About 85% of thenoble gas discharges from BWRs in 1979 came fromthree plants. The average normalized release from 1975
to 1979 was 8800 TBq [GW(e) a]-1.
53. Table 15 gives the nuclide composition of noble
gas releases from BWRs in the United States during
1979 [B23]. The release composition varies greatly.
depending on the waste treatment hold-up time which
varies from less than half an hour on older plants toseveral hours on newer BWRs. The normalized releases
per unit of electricity generated by all United States
BWRs is also given and it can be seen that contributions to the total activity released are made by a numberof nuclides including 138Xe (Tu2 17 min), 135mXe (T 112
257

15 min), rnxe (T 112 9.2 h), mxe (T 112 5.3 d), HHKr (T112
2.8 h), 87 Kr (T112 76 min), 8SmKr (T112 4.4 h) and 85Kr
(T112 10.7 a).
54. In gas-cooled reactors noble gas releases are
insignificant. For reactors of the Magnox type utilizing
uranium metal fuel, because of the magnesium alloy
cladding, temperatures must be kept below 650°C andthe reactors arc equipped with detection systems for
failed fuel pins which are quickly removed from theoperating core. The AGR utilizes enriched uraniumoxide fuel in stainless steel cladding with C02 coolant.
Again, early operating experience does not indicate that
noble gas releases occur. As with Magnox reactors,
AGRs employ on-load refuelling and failed fuel pins
are easily removed.

55.

Releases of noble gases for heavy water reactors
are presented in Table 16 using data from Argentina[B21] and Canada [MIO]. The normalized release
averaged between 1975 and 1979 is 460 TBq[GW(e) a}-1. There is little release reported for fast
breeder reactors in France and the United Kingdom[Fl]. There has been little exploitation of high temperature reactors in recent years, although one or two
experimental facilities remain.

2.

Activation gases

56. Although gas-cooled reactors do not generally
release noble gas fission products as do BWRs and
PWRs, several activation gases are formed in GCR
operation. Direct activation of the oxygen in the C02
coolant gives rise to 16N by the (n,p) reaction on 160.
Argon-41 arises from (n,y) reactions in the stable 40Ar
content of air, either present as an impurity in the
coolant circuit or used for shield cooling in early GCRs
with steel pressure vessels.
57. The amount of 41Ar (T112 1.8 h) released depends
upon the detailed design of the reactor. Release rates of
41Ar are not measured routinely in the United
Kingdom, but measurements have been reported by
Clarke and Wilson [Cl] and give averages of between
about 2 and 15 PBq [GW(e) aJ-1 or between about 20
and 150 MBq s-1. For AGRs the primary source of41Ar
releases is from the leakage of the coolant to atmosphere and releases of the order of about 0.1 PBq[GW(e) a}-1 are reported [L6]. The activation gas
discharges from GCRs are given in Table 17 for theyears 1975-1979. The average normalized releasebetween 1975 and 1979 was 3.2 PBq [GW(e) aJ-1.
58. Another airborne gaseous efnuent of GCRs is 35S(T 112 87.5 d) arising from (n, y) reactions on 34S present
as an impurity in the graphite core, and from (n,p)
reactions on JSCJ also present as an impurity. Discharge
rates have been reported in the range of about 1 to 4
GBq d-1 [G4] but may reduce as the reactors come tofull power. The impurity levels of sulphur and chlorine
in the moderator graphite are reported as approximately 50 and 2 ppm, respectively [P3]. Measured
releases of JSS from Oldbury and Wylfa (Magnox) andfrom Hinkley B (AGR) in 1976 were 52, 89 and 81
GBq, while in 1977 they were 17, 200 and 211 GBq,
respectively [G4]. In 1978 ·the values were 28, 174 and
204 GBq, respectively. The average normalizeddischarge was 170 [GW(e) a}-l and the chemical form
was carbonyl sulphide.
258

59. Nitrogen-16 (T 112 7 s) provides essentially only
direct external irradiation at nuclear power plants. The
photons produced in its decay have energies of 6.1 and
7.1 McV. In BWRs the 16N generated is in part transported with the steam to the turbine buildings
producing an external gamma field. External .gamma
dose rates at the perimeter fence of Central Electricity
Generating Board nuclear power stations in the United
Kingdom have been reported as being mainly due tol6N [G4].

3.

Tritium

60. In LWRs tritium arises from ternary fission in the
nuclear fuel and from neutron activation reactions withlithium and boron isotopes dissolved in or in contact
with the primary coolant. The normalized tritium
production rate has been estimated for LWRs as 0.56
PBq [GW(e) a]-1 from ternary fission [E3], assuming ayield of 0.85 lQ-4 per thermal fission in mu [F2] and2 lQ-4 per thermal fission in 239Pu and mu [E3]. The
estimate assumes that 55% of the integrated number offissions have occurred in 23SU, 41% in 239Pu and 4% in
238U. Other estimates have given production rates of0.85 PBq [GW(e) a]-1 [01, K2]. An average value ofabout 0.75 PBq [GW(e) aJ-1 is assumed here.
61. Estimated generation rates of tritium from
activation reactions depend on the assumed concentrations of the parent nuclide. In PWRs it is thought to bemainly due to reactions with the boron in the coolant
water which is used for reactivity control. In BWRs it ismainly from boron in control rods. In GCRs it is due tolithium impurities in the graphite and to the presence ofwater vapour in the core. For HWRs it is principally
due to activation of the deuterium moderator andcoolant [K 1, S7, T2]. The generation rate from
activation only exceeds the ternary fission source for
HWRs, where the activation rate is some 30 times
higher at about 25 PBq [GW(e) aJ-I. About 1% of thetritium formed in the fuel elements is usually assumed
to appear in the coolant circuit and to find its way toefnuent streams.
62. Table 18 gives the reported tritium discharges inairborne effluents for LWRs and HWRs. For BWRs theannual normalized release of tritium to atmosphere isfairly constant and averages 3.4 TBq [GW(e) aJ-1 for
1975-1979, while for PWRs the value varies between 4
and 15 TBq [GW(e) aJ-1 and the 1975-1979 average is
7.8 TBq [GW(c) a]-1. The atmospheric discharges oftritium are seen to average 540 TBq [GW(e) aJ-1 for
HWRs between 1975 and 1979. The discharges oftritium in liquid efnuents are shown in Table 19 and thenormalized average release derived for 1975-1979 is 1.4
TBq [GW(e) aJ-1 for BWRs and 38 TBq [GW(e) a]-1 for
PWRs. For European PWRs with stainless steel fuelclad, the tritium figure in liquid discharges is 300 TBq[GW(e) a]-1. Gas-cooled reactors are seen to release 25
TBq [GW(c) aJ-1 of tritium in liquid discharges
averaged between 1975 and 1979. The highest releasesof tritium in liquid effluents are from HWRs with anaverage normalized release of 350 TBq [GW(e) a]-1,
63. Bonka [B5] has recently quoted discharges per unitof electrical energy generated of 4 TBq [GW(e) a}-' toair and 40 TBq [GW(e) a}-' to water for PWRs andnormalized rates of 2 TBq [GW(e) aJ-1 (atmospheric)and 6 TBq [GW(e) a]-1 (aquatic) for BWRs. For AGRs
the atmospheric discharge rate of tritium amounts toabout 10 TBq [GW(e) a}-1, about the same as the level inliquid discharges [G6]. In Magnox reactors, the coolant
circuit is continually dried to remove water vapour andtritium produced in the circuit appears primarily in theliquid effluent removed by the humidryers.
64. Gorman and Wong [G7] have estimated 3H
normalized production in Canadian HWRs to be up to
89 PBq [GW(c) aJ-1 in the moderator, compared with
0.74 PBq [GW(e) aJ-1 in the fuel and the normalized
emission is given as 630 TBq [GW(c) a]-1 for airborne
efOuents and 260 TBq [GW(e) aJ-1 in liquid discharges.
These are comparable with the normalized atmosphericand liquid discharges reported here averaged over theyears 1975-1979. In HWRs the build-up of tritium inthe moderator depends upon the irradiation history andon the leakage rate of deuterium from the core, which
will be at least 0.5% and perhaps a few per cent of theinventory per year. These conditions affect the environmental releases, as can be seen from Table 19. where
there is a general upward trend in tritium in liquid
effluents. However, tritium control and removal
systems are being developed and releases could be
expected to reduce. Averaged over the years 1975-1979
the reported HWR releases to atmosphere are similar to
those to the aquatic environment.

4.

Carbon-14

65. Discharges of 14C have been of increasing interestin recent years because of the long half-life (5730 a) of
the isotope and its significant contribution to thecollective dose commitments. Carbon-14 is produced in
LWRs and HWRs by (n,a) reactions with 170 present in
the oxide fuel and moderator; by (n,p) reactions with
14N present in the fuel as impurities; and by ternary
fission. The production rate by ternary fission isvirtually independent of reactor design, while thenormalized production of 14C by the other routes
depends on the fuel enrichment, the relative masses offuel and moderator, the concentration of nitrogen in
the fuel, and the fuel and moderator temperatures. In
gas-cooled graphite moderated reactors, the graphite
moderator is a major source of 14C production due tothe BC (n, y) 14C reaction and the 14N (n,p) 14C reaction
on nitrogen impurities. Production of 14C from the C02
coolant has been estimated to give only a few per centof the total contribution from all sources. Because of
the large moderator mass, 14C is produced mainly from170 reactions in the moderator in HWRs.
66. The 14C content of LWR fuels per unit energy
generated has been estimated at 0.22 TBq [GW(th) aJ-1
[K2], i.e., 0.67 TBq [GW(e) a]-1, assuming a reactor
thermal efficiency of 33%. As noted above, such
estimates are dependent upon the assumptions on
nitrogen impurity levels in the fuel and values up to 1.9
TBq [GW(e) a]-1 have been reported [M2]. More recent
estimates by Davis [D2] gave values of 0.74 TBq[GW(e) aJ-1 for both PWRs and BWRs and a study
carried out for the Swedish power reactors [L2]
estimates 0.52 TBq [GW(e) aJ-1 for PWRs. Hayes and
MacMurdo have estimated 14C normalized production
to be 0.22 and 0.55 TBq [GW(e) a]-1 for PWRs andBWRs, respectively [H2]. Bonka et al. [84] gave valuesof 0.9 TBq [GW(e) aJ-1 for PWRs and 1.1 TBq[GW(e) a]-1 for BWRs. It was estimated in Annex D of
the 1977 report [Ul] that approximately 30% of the 14C
total production is in the moderator for both PWRs andBWRs. This is the likely source of release to theenvironment from the reactor while, based on experi-mental studies, most of the content of the fuel appears
to be released during reprocessing [S8].
67. In gas-cooled reactors a normalized production
within the fuel of 0.96 TBq [GW(th) a)-I has been
estimated [K2] which, assuming a thermal efficiency
from Magnox reactors of 30%, gives a normalized
production of 3.2 TBq [GW(e) a]-I. The primary sourceof 14C is the graphite moderator in which a normalized
production of 9.3 TDq [GW(e) aJ-1 has been estimated.
The normalized production in the carbon dioxide
coolant by the 170 (n,a) 14C has been estimated at 0.11
TBq [GW(e) a]-1 for Magnox and 0.37 TBq [GW(e) a]-1
for AGRs [K2]. For HWRs the normalized 14C
production is estimated at 0.74 TBq [GW(e) a)-I in thefuel and at 21 TBq [GW(e) aJ-1 in the moderator [W3].
68. A programme of measurements have been madein the ~ederal Republic of Germany by the Bundesgesundheitsamt [Rl, S15]. the results of which are given in
Table 20. In European PWRs only a small fraction of
the 14C is in the form of C02; it is mostly present inmethane or other hydrocarbons. The normalizeddischarge is about 220 GBq [GW(e) aJ-1 [Ll, S15, Rl].
For European BWRs the release of 14C appears to be
more than 95% as C02 and the normalized release rate
is about 520 GBq [GW(e) a}-1. In a detailed study of the
Oyster Creek (United States) BWR Blanchard has
measured 14C normalized releases of 220 G Bq
[GW(e) a]-1 from the condenser air ejector and 74 GBq[GW(e) aJ-1 from building ventilation air [B3]. The same
study revealed liquid discharges of l4C at 0.74 GBq[GW(e) aJ-1. For a PWR, Kahn et al. have reported l4C
measured atmospheric discharges of 37 GBq[GW(e) aJ-1 [K3].
69. Carbon-14 discharges from gas-cooled reactors
result from the leakage of the primary coolant (typicallya few per cent per day) which contains radionuclides
released to the coolant by corrosion of the graphite
moderator; estimated normalized releases are 0.22 TBq[GW(e) a]-1 for Magnox and 0.63 TBq [GW(c) aJ-1 for
AGRs [P3]. Groome has reported [G6] that the total 14C
discharge from Magnox reactors is 3.7 TBq, corresponding to 1.1 TBq [GW(e) a}-1, and that AGRs are
expected to give similar normalized release.
70. For HWRs it is reported that about half the 14C
moderator production is released to atmosphere, givinga normalized release of 10 TBq [GW(e) a]-1 [W3], whileother reported normalized releases are about 17 TBq[GW(e) a]-1 [821].
5.

Iodine71. The volatile element iodine is produced in thefission process. Its yield is almost independent of
whether uranium or plutonium isotopes are undergoingfission. The isotopes of iodine of interest in radiologicalassessments are 129J (T112 1.6 10' a), 131J (T112 8.04 d),1321 (T 112 2.3 h), 1331 (T112 21 h), I34J (T 112 53 min) andBSJ (T112 6.6 h). Owing to the short half-lives of all theisotopes except 129], equilibrium activity concentrations
are achieved quickly and releases depend on thenumber of fuel cladding failures and coolant leakage
rate. I odine-131 has been studied for many years inview of its mobility in the environment and selective
human organ irradiations. In recent years 129] hasreceived more attention, although its release rate isextremely low, because of its contribution to thecollective dose commitment through ils long availa259

hility in the environment. or particular interest in fuelreprocessing, t29J is not generally reported in routinedischarges from nuclear power plants.
72. Table 21 gives the year by year reported atmospheric discharges of iodines from power reactors invarious countries. There are wide differences both in
the quantities and in the nuclide composition of thereleases due to different waste treatment systems. Table
22 gives the isotopic composition of atmosphericreleases of iodine from United States PWRs and BWRs
in 1979, together with the normalized release rates. The
annual iodine normalized discharge for PWRs has been
fairly constant and averages 5.0 G8q [GW(e) a}- 1
between 1975 and 1979. For BWRs the 1975-1979
averaged normalized discharge is 410 GBq [GW(e) aJ-1.
The comparable figures for HWRs are between 0.08
and 3.1 GBq [GW(e) aJ-1 [MIO]. It is clear from Table 22that while for PWRs 131J contributes about one third of
the total iodine discharge, for BWRs it represents lessthan 10% of the iodine release. The normalized releases
are not typical of any one reactor but reflect the
average for the nuclear power industry. Few data are
available on the proportions of organic and inorganic
forms of the iodine released to the atmosphere, but
analysis for power stations in the Federal Republic of
Germany shows that usually less than 1o/o of the iodine
released in gaseous effluent is in particulate form [W2].
73. Measurements at six power reactors in the UnitedStates indicated that on average 73% of the iodine in thereactor off-gases was in the organic form, 22% was
hypoiodous acid and 5% elemental [P7]. All forms werealso present in ventilation exhaust air, though therelease rate from this source is usually much less thanfrom the reactor off-gas system. The release of iodine
isotopes depends strongly on the filtration system usedat the plant.
6.

Particulates in airborne effluents

74. Radionuclides in particulate form can arise
directly or as decay products of fission noble gases or
may arise from corrosion of materials in the primarycoolant circuit. Aerosols are generated because of
primary coolant leaks or because of maintenance workon active components removed from the primarycircuit. The air in all areas where aerosols might arise iscontinually purged and the plenum activity is filteredthrough high efficiency particulate (HEPA) filters
which retain all but the finest aerosols. Releases ofparticulate activity are very low and the nuclide composition is essentially unique to each operating plant; it
depends on the particular impurities in cladding andstructural materials, coolant chemistry and fuel failuremodes. The releases also vary from time to time becauseof different operational and maintenance needs andpractices.
75. As a consequence, the range of reported nuclides
in parliculate atmospheric discharges is very large, up
to several tens of nuclides at any one plant [D3, 823,
824). Radionuclides identified and reported at various
plants include: 'Be, 22Na. 24Na, 51Cr, 54Mn, 56Mn, 59Fe,
s1c 0 , S8Co. 60Co, 63Ni, 65Zn, 76As, 88Rb. 89Sr, 90Sr. 91Sr,
95Zr, 97Zr, 95Nb, 99Mo, 99mTc, t03Ru, IOSRu, 106Ru,
IOl!mAg, llOmAg, 1IJSn, 1t5Cd, t22Sb, t24Sb. t25Sb, t23mSn,
t23mTe, 134Cs, 136Cs, mes, IJ9Ce, 1408a, l40La, t41Ce,
144Ce, t82Ta.
76. It is thus clear that averaging the nuclide composition of particulate releases by reactors is difficult and260

may mean little. There is no indication in current
discharge data of any one nuclide dominating therelease for any given reactor type. In Table 23, theaverage particulate discharges from reactors year by
year are given as the reported sum of activity. Over theperiod 1975-1979 normalized releases of particulates
from PWRs averaged 2.2 GJ3q [GW(e) aJ-1, while the
BWR nonnalizecl releases to atmosphere were 53 GBq[GW(e) aJ-1. The normalized release from GCRs was 1.0
GBq [GW(e) aJ-1, from the Argentinan HWR was
0.044 GBq [GW(e) a]-1, from the Canadian HWRs 0.9
GBq [GW(e) aJ-t [MIO] and from the fast reactor Phenix
was 4.8 MBq [GW(e) aJ-1 (L6].

7.

Liquid effluents

77. The sources of radionuclides other than tritium inliquid efnuents are essentially the same as thosedescribed for particulate releases to the atmosphere andthe discharges reported are equally as varied. The
amount and composition of the discharge depends
upon the design and operating practice of the reactor,
impurity levels and trace quantities of materials in
structural and cladding components. The available
discharge data for 1975-1979 are given in Table 24 for
operating reactors. Table 25 shows the isotopic composition of liquid effluent discharges for reactors in the
United States in 1979. The isotopic composition ofliquid effluents from GCRs in the United Kingdom for1979 is shown in Table 26.
78. The normalized release levels based on thereported discharges for each reactor type using theactual reported generation of electrical energy averaged
between 1975 and 1979 are approximately (Table 24):
PWR: 180 GBq [GW (e) aJ-t
BWR: 290 GBq [GW (e) aJ-t
GCR: 4800 GBq [GW (e) aJ-t
HWR: 470 GBq [GW (e) a]-1.
Canadian HWR averaged results are some 10 times
lower [M 10). The elevated levels for GCRs in the
United Kingdom reflect the fact that discharges are
made (with the exception of Trawsfynydd) to themarine environment. Swedish reactors also discharge tothe sea and have slightly higher normalized liquid
effluent release rates than LWRs of the United States orEurope. In Annex D of its 1977 report [U1], theCommittee quoted normalized releases for PWRs,
BWRs and GCRs of 296, 2220 and 5550 GBq[GW(e) aJ-t. The average normalized aquatic releases
for BWRs over the 5-year period 1975-1979 havebeen reduced by about an order of magnitude andnormalized discharges in the last few years have beeneven lower. This reduction does not appear to havebeen due to removal of specific nuclides from the liquid
waste streams, but appears to be applicable to all the
nuclides constituting the release.
'-

79. In 1979 the isotopes 134Cs and 137Cs represented
10% and 35% of the activity concentration in liquid
discharges from the United States PWR and 8WR
systems, respectively. In Annex D of the 1977 report[Ul], the Committee found that caesium isotopes contributed 70% of the discharged activity concentration for
BWRs and GCRs, and 30-50% for PWRs. For L WRs
the remaining contributions to activity in aqueous
discharges arise from a number of nuclides; cobalt
isotopes ssco and 6DCo contribute about 65% of the,acuv1ty in PWRs and iodine isotopes some 6%. For
BWRs cobalt isotopes contributed 10% of the liquid
effluent activity concentration in 1979 and iodines 5%.
80. The isotopic ratio of 134Cs to 137Cs from GCRs
was 0.28 in 1977, compared with 0.2 found by theCommittee in Annex D of the 1977 report [Ul]. The
increase in I34Cs level probably reflects the increasingfuel irradiation time achieved in Magnox reactors (nowapproximately 4.5 GW d t-1). As fuel burnup increases,
the amount of the long half-life isotope 137Cs increases
almost linearly, whereas 134Cs is produced primarily
from neutron captures in the stable fission product
133Cs, and, since the amount of 133Cs increases linearlywith burnup, the production of 134Cs increases nearly as
the square of the fuel burnup. Its radioactive half-life(2.1 a) means that some radioactive decay takes place,but the t34Cs/l37Cs ratio should increase nearly linearlywith burnup. For BWRs the 134Cs/137Cs ratio is 0.77
and for PWRs it is 0.57.
81. The release of 1311 into liquid effluents contributed
about 4.7 GBq [GW(e) a]-1 for BWRs in the UnitedStates in 1979. In the same year, 1311 normalized
releases from PWRs average out at 4.6 GBq [GW(e) aJ-1
in the United States, 4.8 GBq [GW(e) a]-1 in Sweden
and 6.7 GBq [GW(e) a]-1 in western Europe. The liquid
discharges of 1311 are therefore similar for BWRs and
PWRs and compare with the atmospheric normalized
releases of 1311 of 42 GBq [GW(e) aJ-1 from BWRs, and
1.9 GBq [GW(e) aJ-1 from PWRs. The normalized
releases of other iodine isotopes in liquid effluentsamount to about 3 GBq [GW(e) aJ-1 from 1331 and 1351
in PWRs, and 2 GBq [GW(e) aJ-1 in BWRs.
82. There is a wide range of activation products andfission products reported in liquid discharges andresults vary widely from reactor to reactor. It isapparent that one or two nuclides consistently
contribute to the discharge from LWRs. Amongst these.
24Na is widely reported and contributes to radionuclide
discharges from both BWRs and PWRs. Also ssco
discharges are as high or higher than 60Co discharges
and 89Sr contributes to BWR liquid effluents.
83. For GCRs the next most significant contributions
to aqueous releases after caesium are from 3SS and 90Sr.
The 35S is collected in the humidryers which remove
water from the gas circuit and appears in the liquid
effluent stream at a normalized release of 547 GBq[GW(e) aJ-1, compared with 170 GBq [GW(e) aJ-1 ingaseous releases (paragraph 58).
B.

,.;

1

LOCAL AND REGIONAL COLLECTIVE DOSECOMMITMENTS

84. National authorities usually require an environmental monitoring programme in the vicinity of a
nuclear power plant to be carried out either by theoperator, or by another competent agency, or both.
Detailed investigative studies have been reported [K3,E4, B3) and, in general, levels of radioactive contamination are not readily detectable except in theimmediate vicinity of the plant. Dose assessments forthe population. therefore, rely on modelling theenvironmental transport and transfer of radioactive
materials. In recent years trajectory modelling has beendeveloped for long-range atmospheric dispersion calculations in which historical meteorological data is used
to calculate the paths of discrete masses of air [Zl]. The
Committee has felt these models are not yet sufficiently

developed to offer advantages
described in Annex A.

over the models

85. The object of the Committee in making its present
assessment is to give a representative value of thecollective dose commitments per unit electric energy
generated by nuclear power stations and to reflect thelevels of dose received by most exposed individuals.
The results will not apply to any one reactor or locationand the collective dose commitments should not be_ applied to the known discharge rate of a given reactor
to obtain estimates of total health detriment. To
undertake such a study the values of parameters in themodels used would need to be specific to that site (forexample, for the meteorological dispersion) and to theparticular local terrestrial pathways.
86. In the following sections, the normalizeddischarges of radionuclides found in the previous
sections are assessed at a model reactor site using themethodologies outlined in Annex A. The site is mostrepresentative of the areas of Europe and the NorthEastern United States. as those areas contain the
majority of the power producing reactors. Agricultural
production patterns and population distribution are
most typical of those areas. The parameter values in themodels necessary for the assessment of each of thesource terms are described in tum below.
1.

Fission noble gases

87. The exposure of the population from noble gas
discharges to the atmosphere is by external ~-- andy-radiation. For ~--irradiation a semi-infinite cloudmodel is adequate, but for y-exposure, because of thelong mean free path of y rays in air, a finite cloudmodel must be used. The models are described in
Annex A. The finite cloud-y calculation integrates
contributions to the photon tluence from radioactive
source terms throughout the volume of the plume. For
the present calculations the computer code ESCLOUD
has been used [J2] which is essentially similar to othercodes developed in the Federal Republic of Germany
by Vogt [Vl] and Rohloff [R2] and in the United Statesat the Oak Ridge National Laboratory [Ml].
88. The energy deposition build-up factors for they-exposure at each downwind distance of interest weretaken from Chilton [C2]. The absorbed dose in air iscalculated using values of absorbed dose in air per unit
tluence [Hl] and the ratios of the absorbed doses inhuman body organs and tissues per unit absorbed dosein air have been obtained from Poston and Snyder [P4].
The effective dose equivalent is calculated using theprocedure outlined in Annex A for presentation ofsummary results. The absorbed doses in skin have beentaken into account using a weighting factor of 0.01. The
nuclear decay schemes have been taken from Nichols[Nll, N12]. Nair [N13] and Despres et al. [D9). Thesesame data sources were used for the ~--energies for skindose equivalent calculations.
89. The meteorological data applicable to the modelsite are given in Table 27. The frequency distribution of
Pasquill meteorological categories is assumed to be thesame in each sector, which is an approximation toobservations. A representative frequency distribution ofwind directions has been taken. Rain is only assumed
to occur for 3.5% of the time and only in near neutral
conditions. A typical population distribution for
Northern Europe and North-Eastern United States has
261been taken and the summary of individuals in distancebands and cumulative population all round the site arealso given in Table 27. The cumulative population
within a radius of 2000 km from the site is 260 million
people, giving an average population density of about
20 km-2. The population density within 50 km of thesite is an average of about 400 km-2. An effective heightof release of 30 m has been chosen for collective dose
calculations. although the choice of height has only asecondary inOuence on the results. The individual doseto a member of a critical group is of course highly
dependent upon stack height.
90. Using the normalized releases for PWRs from
Table 12 for noble gas atmospheric releases and thenuclide composition from Table 13. the normalized
collective absorbed dose commitments averaged
between 1975 and 1979 from the model PWR facility
have been calculated and are shown in Table 28. The
normalized release term is 430 TBq [GW(e) aJ-1 andonly those isotopes which contribute significantly to thecollective absorbed doses are presented. The growth ofdaughter nuclides, e.g .• BBKr - BBRb has been includedin the calculations.
9 I. The normalized collective effective dose equivalent commitment totals 4.2 10-2 man Sv [GW(e) a]-1,
compared with the Committee's previous assessment in
Annex D of the 1977 report [Ul] of 2.5 10-2 man Sv[GW(e) a]-1, even though releases have been reduced byabout a factor of 2. About 80% of the collective dose
commitment is given by the single isotope mxe.
Xenon-135 is responsible for about 11% of thecollective dose commitment and 88Kr for a further 4%of the total. The difference in the assessment is mainly
due to the somewhat higher population density within
the first few hundred km of the site (towards 300 km-2)
as compared with a value of 100 km-2 used previously.
This reflects the closer siting of reactors to centres of
population in the last decade. In Table 29 the spatial
distribution of collective dose for the PWR is shown; ofthe collective dose, 90% is accumulated within 500 km,
and the bulk of the contribution (nearly 60%) arises
between 100 and 500 km. There is little contribution
from inhalation of radioactive daughter products. The
estimates made here include an allowance for theshielding from buildings and fraction of time spent
outdoors.
92. For BWR normalized releases for the years
1975-1979. a value of 8800 TBq [GW(e) a]-1 was takenfrom Table 14. together with the isotopic composition
shown in Table 15. and the resulting collective dose
commitments are shown in Table 30. The normalized
collective effec1ive dose equivalent commitment is 1.9
man Sv [GW(e) aJ-1, compared with the Committee's
previous estimate in Annex D of the 1977 report [Ul) of
5.5 man Sv [GW(e) a)-I. The main isotope contributing
to the normalized collective effective dose equivalentcommitment is SBKr (T112 2.8 h) giving about 50% of thetotal with its daughter isotope 811Rb (T112 15.4 min). This
RSRb component is only due to the decay of the SBKr in
the atmosphere. Most of the remainder of thenormalized collective effective dose equivalentcommitment arises from: t35Xe (T 112 9.2 h), 21%; 13BXe(T112 17 min). 14%; B7 Kr (T112 1.3 h), 7% and mxe (T112
5.27 d), 5%. The growth of 138Cs (T112 32.2 min) fromdecays of usxe is included in the dose calculations inTable 30. The collective dose commitments due tonoble gases include a contribution from inhalation of
the BBRb isotope. -only those isotopes contributing
262

significantly to the collective dose commitments areincluded in Table 30.
93. The spatial distribution of the normalized
collective effective dose equivalent commitment overdistance for the model BWR is shown in Table 31.
More than 80% of the collective dose is accumulated
within 50 km of the site, and 40% within 10 km. Thisbehaviour is caused by the dominant contribution of
BBKr which decays by a half-life about every 40 km oftravel distance. and by the fact that all the shorter-lived
nuclides decay within the first tens of kilometres.
94. Martin and Nelson computed the collective doses
to populations within 80 km of eight BWRs in the
United States [M3). The collective absorbed whole-bodydose was 9 man Gy from 1.85 GW(e) a of power
generation, corresponding to 5 man Gy [GW(e) a)-I.
There h,as been a very marked tendency for releases tobe reduced partly as newer plants come into operation.
and Martin's estimate is consistent with the assessment
made here when allowance is made for the reduction inreleases.
95. In summary, the normalized collective effective
dose equivalent commitment from noble gas releases is0.63 man Sv [GW(e) a]-1, based on the weighted
electricity production by PWRs and BWRs as a fractionof total nuclear generation over the years 1975-1979.
The annual effective dose equivalents to most exposed
individuals of hypothetical critical groups have been
calculated at 40 µSv for the model BWR and more than
100 times lower for PWRs, taking an average over the
first 2 km from the site. Reported levels of annual dose
equivalent rates to most exposed individuals are
generally a few tens of µSv, although one or two plants
can give figures of a few hundred µSv [Ll, L6].

2.

96.

Activation gases

The primary interest has been in the release of~ 41Ar which, because of its short half-life (1.83 h), contri-

butes most of its dose within a few tens of kilometres ofthe site. although the exact result is clearly highly
dependent upon the close-in population density. The
normalized release of 41Ar from GCRs between 1975
and 1979 is taken from Table 17 as 3240
TBq [GW(e) a]-1 and the associated normalized
collective effective dose equivalent commitment fromthe model site is 0.95 man Sv [GW(e) a)-I. Clarke andWilson [C1] have reported collective doses in the
United Kingdom which correspond to values between
0.3 and 1.0 man Gy [GW(e) a]-1 and individual dose
equivalents of up to a few hundred µSv for early GCRs,
depending on the site. all of which are fairly remote.
The weighted normalized dose commitment. allowing
for GCR nuclear generation as a fraction of allelectricity produced over the years 1975-1979. is about
0.1 man Sv [GW(e) aJ-t. For reasons of convenience, thedoses arising from releases of 41 Ar from LWRs were
included with the doses due to the fission noble gases(paragraphs 87 to 95).
97. The consequences of the release of35S from GCRs
have been studied in some detail by environmental
monitoring. The isotope is released in the form ofcarbonyl sulphide (COS) which has a very low
deposition velocity and a slow reaction rate in air. The
major route of exposure of the population is throughthe ingestion of milk: estimates of collective
whole-body dose per unit activity released have been

made by Linsley [L7] which yield 2.2 1~ man Gy(GBq)-1 for typical United Kingdom conditions. Taking
a normalized release rate of 0.17 TBq [GW(e) a]-1 thus
gives rise to 3.7 10-2 man Gy [GW(e) a]-1 and thecontribution to collective effective dose equivalentcommitment weighted by the fraction of nuclear
electricity from GCRs is 3.8 10-3 man Sv [GW(e) a]-1.
3.

Tritium

98. The transfer of tritium between the atmosphereand the terrestrial environment is particularly complex
because of the hydrogen cycle in biological systems.
Tritium released to the environment will make acontribution to the collective close commitment by
becoming globally dispersed and this is considered inchapter V. The assessment of the local and the regional
collective doses from atmospheric discharges of tritiumis performed slightly differently from the method
outlined in the introduction to this Annex.
99. To assess the collective dose, a specific activity
model has been assumed for transfer through the terrestrial environment. It is assumed that tritium in man andin the terrestrial environment rapidly achieves
equilibrium with the tritium in the atmosphere. The
specific activity of tritium taken into the body is equal
to that in atmospheric water vapour at the point ofinterest. The specific activity of tritium in atmosphericwater vapour is determined by the atmosphericdispersion to the point of interest and the concentrationof water vapour in the atmosphere (8 g m-3 annual
average value).
100. This specific activity approach assumes that allwater taken in by humans, whether by inhalation andabsorption through the skin or by ingestion of water orfoodstuffs in normal diet, is contaminated at the
specific activity appropriate to the point of interest.
This is a conservative assumption. It also fails to distinguish any temporal distribution in the dose which may
be extended over some considerable time. The dose is,therefore, only assessed on a specific activity model forthe period of discharge. The annual intake of water has
been obtained from the data given by ICRP in its publication 23 (15). The total water intake rate by all routes isassumed to be 3 kg d-1 for men and 2.1 kg d-1 for
women; the respective inhalation and ingestion intake
rates have been assessed assuming an annual average
concentration of water vapour in air of 8.1 g m-3 and amean adult inhalation rate of 20 m3 d-1. The intake byinhalation is assumed to be accompanied by an equal
intake by skin absorption so that the net annual intakesof H20 become 130 kg inhaled and 800 kg ingested.
101. The collective whole-body absorbed dose
commitment to the local and regional population on
this basis is evaluated, assuming normalized atmospheric discharges taken from Table 18 of 3.4 TBq[GW(e) a]-1 for BWRs, 7.8 TBq [GW(e) a]-1 for PWRs,
11 TBq [GW(e) a)-I for GCRs and 540 TBq [GW(e) aJ-1
for HWRs. The local and regional collective doses perunit electrical energy generated are shown in Table 32.
The ingestion pathway appears to be more importantby a factor of about 6 than the inhalation pathway andthe collective absorbed doses to all body organs may be
regarded to be the same as the collective effective dose
equivalent because of the assumption of the models.
102. The collective effective dose commitment varies
between 0.04 and 5.6 man Sv [GW(e) a]-1. The previous

estimation in Annex D of the 1977 report [Ul] gave anormalized collective dose commitment of 4 J()--4 man
Gy [GW(e) a]-1 for PWRs and a normalized dischargeof 7.4 TBq [GW(e) a]-1. The reason for the difference incollective dose commitment per unit discharge of afactor of about 200 higher in the present estimate is dueto the previous estimate being given only for the local
population within JOO km (about 25% of the local andregional collective dose); to the fact that the previous
estimation only considered inhalation (about 15% of
total intake); and to the greater by a factor of about 4
population density. In summary. the normalized
collective effective dose equivalent commitment foratmospheric releases of tritium, weighted by the
proportion of electricity generated. is 0.46 man Sv[GW(e) aJ-1. For the model site taken by the Committee.
individual annual effective dose equivalents from
LWRs and GCRs are about lQ-6 Sv, while the HWR
model r.esults are about 1~ Sv.
103. For tritium in liquid effluents the model river site
(Table 38) gives a collective whole-body dose
commitment per unit activity discharged of 8.1 1~
man Gy TBq-t on the assumption that the river is used
as a source of drinking water. Using the normalizeddischarges for 1975-1979 in Table 19 (1.4 TBq[GW(e) a]-1 for BWRs, 38 TBq [GW(e) a]-1 for PWRs
and 350 TBq [GW(e) a]-1 for HWRs) leads to collective
effective dose equivalent commitments of 1.1 10-3 man
Gy [GW(e) a]-1 for BWRs, 3.1 10-2 man Gy [GW(e) a]-1
for PWRs and 0.28 man Gy [GW(e) a]-1 for HWRs. The
models indicate dose commitments 10 times lower foraquatic effluents than for atmospheric effluents per unit
release, while the reported release data indicate theatmospheric pathways as the more significant. The
contribution to the normalized collective effective dose
equivalent commitment from liquid releases of tritium
weighted by generation of each reactor type is 0.04 man
Sv [GW(e) a]-1.
4.

Carbon-14

104. The local and regional collective doses due to l4C
releases from reactors only represent a small proportionof the total dose commitments. The main significanceof 14C is due to its entry into the carbon cycle andresulting global dispersion, leading to long-term irradiation which is considered in chapter V. The assessmentof the first pass regional collective dose commitments
may be made using the same specific activity approachthat was used for tritium in the preceding subsection.
For a release to atmosphere the specific activity will be
determined by the atmospheric dispersion and thecarbon concentration in the atmosphere, taken as 0.16 g
m-3. The intake of carbon is assumed to be that given
by ICRP in publication 23 [IS]. that is. 93 kg a-1 byingestion and 1.2 kg a-1 by inhalation, based on the
average level of carbon in the atmosphere andassuming a breathing rate of 20 m3 d-1. As with tritium,
it is assumed that all components of the diet arecontaminated at the specific actiYity applicable to thedownwind distance of interest from the source.
105. The form of release of carbon-14 is taken to be as
C02. The collective whole-body absorbed dose
commitment, corresponding to the production of
1 GW(e) a, for the model site and discharges of 518 GBq
for the model BWR, 222 GBq (PWR), 1100 GBq (GCR)
and 17 TBq (HWR) are given in Table 33. The results
range from 0.9 to 30 man Gy, and it is clear that theinhalation pathway accounts for little of the dose. The
263

study of the Nuclear Energy Agency of the OECD
[N 14) has given a regional collective effective dose
equivalent commitment per unit activity associated withthe release or 14C of 0.68 man Sv TBq-1. However, themajority of the collective effective dose equivalentcommitment from the release will arise from the global
contribution. The present estimates must be qualified
for two reasons. Firstly. the figures will be overestimates because of the assumption of a specific activity
model: and, secondly, the assumption or all components of diet being contaminated at the specific activity
corresponding to that point means that the time distribution in the delivery of the dose is ignored. The
normalized collective effective dose equivalentcommitment, weighted by generation of electric energy,is 2.8 man Sv [GW(e) aJ-1, For the model site. theannual effective dose equivalents to most exposed
individuals would be 2 µSv for the PWR, 5 µSv for the
BWR, 10 µSv for the GCR and 200 µSv for the HWR.

5.

Iodine106. Releases of radioactive iodine from nuclear
power plants are small and there is only a small contribution to the total local and regional collective dose
commitments from reactor discharges. lodine-129
because of its long half-life enters the global cycle foriodine and potentially irradiates the global population
for many millions of years. The release of IJIJ contributes only to the local and regional collective doses but
its assessment is complicated by the chemical form inwhich the iodine is released, i.e., elemental, organic orparticulate. Elemental iodine readily deposits onvegetation and enters the terrestrial foodchains. The
deposition rate of organic iodine is between 200 and
100 times less per unit air concentration than that of theelemental form [H3, S9]. The exact value of deposition
velocity for a particular circumstance depends upon thesize of particles, the reactivity of the vapour, the natureof the underlying surface and the meteorological conditions. In this assessment 75% of the iodine released isassumed to be in organic form and 25% elemental(subsection III.A.5).
107. A representative deposition velocity of 5 10-3 m
s-1 is used here for elemental iodine and is also applicable for particulates with mean aerodynamic
diameters of a few micrometres depositing on a wide
variety of surfaces. In particular circumstances withdefinedphysico-chemical
forms
and
specificvegetation, more appropriate values would be needed.
In fact, elemental iodine becomes absorbed onto
aerosols in the atmosphere and its behaviour is thengoverned by that of the aerosol. For organic iodine adeposition velocity of 5 1o-s m s-1 has been chosen as
typical. These deposition velocities refer to removalfrom the plume and once the activity is removed. only acertain fraction is found on the surface of thevegetation.
108. Hoffman has shown that care must be taken to
distinguish between the total removal from the air, andthe fraction intercepted by the vegetation [H4] and thatmany experiments have been misinterpreted. Considerable variation has been observed in the measured
values of interception factors and removal rates fromplant surfaces [86, G8. H4]. An interception fraction of
0.2 is used here as a compromise between the somewhat
higher fraction for dry deposition and the lower one forwet deposition. The remaining fraction of the deposit.
264

0.8, is transferred directly to the ground surface. The
removal of the radionuclides from plant surfaces due tothe actions of wind and rain and plant growth is taken
to occur with a half-life or 14 days for pasture grass [B6]and 30 days for all other plants [G8].
109. Calculations have been performed for iodine
releases from the model PWR and BWR reactors into
the model environment, using the computer codesdeveloped by Simmonds et al. [SlO] which provide amethodology for dealing with the assessment of theintakes of radionuclides by grazing animals and thetransfer to human diet. The concentration factor for
iodine in root vegetables, pasture and grain is assumed
to be 0.02 and the translocation fractions to the edible
portions are 0.1 for vegetables and grain [Nl5, F3].
110. T,he consumption rate of grass (dry weight) bycows is taken as 5.1 103 kg a-1, the half-time in thebovine GI tract is 15 h, the mean-life of milk cows istaken as 6 a, the weight of meat 230 kg and the milk
production rate 10 1 d-1 [Sll]. The time integrals ofactivity in animal products derived from cows grazingon contaminated pasture have been given [ClO].
Assuming a unit deposition rate (1 Bq m-2 s-'), the timeintegrated concentrations of 131 I in beer. liver and milk
are 4.4 104, 4.4 104 and 7.5 104 Bq a kg-I, respectively.
111. The dosimetric and metabolic data for 131 I havebeen derived from Adams [Al] and ICRP publication
30 [12]. The model site has production rates offoodstuffs typical of Europe and North-Eastern UnitedStates at the typical densities of production. The densityof cattle in grazing areas is over 200 km-2 and averages
out at about 15 km-2 over an area of radius 2000 km,
similar to the average of values given by Eckerman etal. for the states of New York, Pennsylvania,
Washington, Virginia, Illinois and Ohio [E2). Vegetableproduction is given as a fraction of land, leading to 2
10-2 km2 per km2 averaged over the same area. The
collective absorbed dose commitments are shown in
Tables 34 and 35 for 1975-1979 normalized emissions
to atmosphere of iodine: 410 GBq [GW(e) a]-1 for
BWRs and 5.0 GBq [GW(e) aJ-' for PWRs, respectively(Table 21). HWR normalized releases appear to be
similar to those for PWRs. The isotopic composition of
the iodine released was assumed to be that averaged
over United States PWRs and BWRs (Table 22).
112. For PWRs the normalized collective effective
dose equivalent commitment is assessed at 6.6 lQ-4 man
Sv [GW(e) aJ-1 and the collective thyroid dose at 2.1
10-2 man Gy. The largest component of the collective
dose comes from inhalation, with ingestion contributing almost equally; most of the ingestion dose arises
from milk. In the case of BWRs the normalized
collective effective dose equivalent commitment is 1.9
10-2 man Sv [GW(e) aJ-1 and the normalized collective
thyroid dose is 5.9 10-1 man Gy [GW(e) aJ-1, The
fractional contributions by the ingestion pathways are
smaller than for the PWR. The main route of exposurefor 133] and t35J is, as expected, inhalation althoughthese isotopes contribute about 22% to the collective
thyroid dose for BWRs and 4% for PWRs. The
normalized collective effective dose commitment
averaged over the proportions of electricity generated is
6.5 10-3 man Sv [GW(e) aJ-1. Representative annual
effective dose equivalents for individuals at about 1 km
from the model site are 10 µSv for BWRs and some 30
times lower for PWRs.

6.

Particulates in airborne efflumts

113. As reported in subsection 111.A.6, the quantities
of radionuclides in particulate releases are highlyvariable between reactors of the same type and indeed
for the same reactor from year lo year. Also, there areseveral tens of nuclides identified which contributesignificantly to the releases. The approach adopted here
is to release the normalized discharge rate from PWRs
and BWRs and assume that the rate is composed ofequal amounts of activity concentration from a range ofthe nuclides most frequently reported to he present in
atmospheric discharges. Again the releases take placefrom the model facility and the collective dose commitments evaluated via all the pathways. The transfer data
for the nuclides in environmental materials have beentaken from the work of Ng et al. [N15, N16], Fletcher
and Dotson [F3] and Linsley et al. [L3).
114. Much of the external contamination on plants
when harvested is removed before consumption by man(e.g., washing or milling of grain). There is recent
evidence that 90% of the external contaminant isremoved during preparation and processing before
consumption by man. based on the observations onvalues for the transfer of plutonium from the outside
surfaces of grain to flour [A2]. Other pathways which
need to be considered are the root uptake of radionuclides migrating downwards through the soil, theirresuspension from the ground onto plant surfaces, andtheir translocation from the surfaces of plants to theinternal tissues. In addition, account must be taken of
the regular removal of radionuclides by harvesting
crops. A mean growing period of 100 days is assumed
here.
115.

The nuclides considered were: S1Cr, S4Mn. S9Fe,

ssc 0 , 60Co, 65Zn, 89Sr, 90Sr, 9oy, 95Zr, 95Nb, 124Sb, 134Cs,
136Cs, 137Cs, 140Ba, 140La, 141Ce, and t44Ce. The
normalized total particulate release for PWRs was
2.2 GBq [GW(e) a]-1, and that for BWRs was taken as
53 GBq [GW(e) a)-I corresponding to the 1975-1979
averaged discharges for release (Table 23). The
resulting collective absorbed dose commitments perunit electrical energy generated are shown in Table 36.
For both BWR and PWR the highest collective organdose commitments are to bone lining cells, althoughmost organs and tissues receive similar doses. In thecase of GCRs. the normalized releases and resulting
doses are similar to those for PWRs, while those for
HWRs are about an order of magnitude lower.
116. For PWR releases the sum of the collective
effective dose equivalent commitments by all routes is0.012 man Sv [GW(e) a]-1. most of which arises from theexternal dose and ingestion from ground deposits ofactivity. Results for GCR releases are essentially thesame. The major contributions to this ground-y doseare mes and 60Co (80%) most of the remaining dosecoming from 134Cs and 54Mn. The dose distribution
through all the body organs is similar because it is from
penetrating external radiation. Some 95% of thecollective effective dose equivalent commitment isreceived in the first 50 years following grounddeposition. The nuclides contributing to the collective
effective dose equivalent commitment via grain are 90Sr
(30%), t34Cs (30%), 137Cs (30%) and 60Co (5%). For thegreen and root vegetable routes, 80% of the collective
effective dose equivalent commitment arises from 90Sr
and the remainder from 137Cs. For beef, the two
caesium isotopes and 90Sr contribute equally to thecollective effective dose equivalent commitment. In thecase of the BWR releases the time distribution of thecollective effective dose equivalent commitment is thesame as for the PWR and the same isotopes and routes
are important: the collective effective dose equivalentcommitment is 0.29 man Sv [GW(e} aJ-1. The
normalized result for all reactors thus becomes 0.1 man
Sv [GW(c) aJ-1. Individual annual effective dose equivalents amount to 1 nSv for BWRs at typically 1 km;
corresponding values for exposures to PWR releasesarc 10 times lower.

7.

Liquid effluents

117. For releases of radionuclides into fresh water thereceiving medium is usually a river or a lake, and thepathways leading to human exposures are drinking
water, ingestion of fish, irrigation leading to contamination ,of foodstuffs. and external irradiation fromsediments. For discharges to the marine environment it
is usually sufficient to consider the ingestion offoodstuffs, including ocean fish and crustacea. Other
pathways exist, such as swimming in contaminatedwaters and the consumption of unusual foodstuffs. butthese contribute little to the collective dose commitments [P1, B7].
118. The collective doses resulting from discharges tothe aquatic environment are more difficult to estimate
using generalized models than are those from atmospheric releases. This is because the local dispersion
from the discharge point is very dependent upon particular site characteristics, such as sedimentation andwater volume flow patterns. The collective dose from areactor discharge is, therefore, highly variable
depending upon how much activity is transferred to theareas where foodstuffs are significantly harvested. Sitespecific assessments have identified the criticalpathways to man and estimated doses to critical groups
for reactor liquid discharges_ The reported annual doses
to the critical group are estimated to be in general lessthan 5 µGy [H5, M4. B7, M10, W3].
119. Mitchell suggests that the annual collective doseto the United Kingdom from radioactive liquid
effluents into the sea had come into equilibrium in 1976
[M5]. On this basis the collective whole-body absorbed
dose to the United Kingdom population in 1976
resulting from the discharge of about 8 TBq of 137Cs
each year from all the coastal nuclear stations during
the period 1974-1976 can be estimated at about 0.2 man
Gy. Hetherington has estimated [H5] that the collective
whole-body dose per unit discharge to the populationof Europe is 0.16 man Gy TBq-1 at equilibrium for acontinuous discharge of mes. This leads to an estimateof 1.0 man Gy as the collective absorbed whole-bodydose to the United Kingdom and Europe for 1979 total
mes discharges from United Kingdom reactor liquid
effluents and corresponds to 0.24 man Gy [GW(e) aJ-1.
120. The Committee. in Annex D of the 1977 report[U1], discussed the difficulty of assigning values toparameters in assessing liquid effluents, in particular,
the water utilization rate and flow rates for rivers, thefish production rates and sedimentation rates. Any
assessment of a given site must, therefore. utilize sitedependent values of parameters and even models
developed specifically to model the movements ofactivity in the local aquatic environment. Generalized
models for a river and a typical marine environment
have been described in Annex A and these models are265now applied to a model site either discharging into ariver or into local coastal waters.
121. For the present assessment, the site environmental characteristics are chosen to be representative of
the receiving media in northern Europe and thenorthern states of the United States and do not refer toany one particular site. The results arc intended to be
indicative of the collective dose commitments per unitof electricity generated by nuclear power stations and,
therefore, the collective dose per unit discharge should
not be combined with the discharges reported for a
specific site in an attempt to estimate the collective dose
commitment from that site.
122. The 1975-1979 normalized releases for PWRs,
BWRs and GCRs given in Table 24 with the isotopic
composition shown in Tables 25 and 26 are used assource terms for the model sites and shown in Table 37.
Only those nuclides which contribute significantly tothe collective dose from each installation are shown.
Also shown in Table 37 are the concentration factors
assumed for the freshwater and marine environments.
The concentration factor is the quotient of the activity
per unit weight of the animal considered (fish,
Crustacea. etc.) and the activity per unit volume offiltered water [Bq t-1/(Bq m-3)]. The factors are basedon dry weight of sediments and wet weight of the edible
fraction of other materials. The marine concentration
factors are taken from Ancellin et al. [A3) and the International Atomic Energy Agency work on models foruse in sea dumping of radioactive wastes (16].
123. The freshwater concentration factors for fish
were taken from Thompson et al. [T3J and those forsediments from Booth's model [B8]. The quantities offresh water extracted from the model river site fordrinking purposes have been estimated assuming that
populations bordering the river take their water at a
rate per individual of 1.5 10-S m3 s-1, given for a studyof the Rhine-Meuse region by Bayer [B9]. The model
river has three sections below the point of discharge
and the relevant model data are given in Table 38.
124. Values in Table 38 are selected to be representative values and many vary by several orders ofmagnitude in both directions for any given site. Nevertheless they are deemed to reflect the situation averaged
over several countries [B9]. The population along thesides of the river is assumed to be 20 km-1 for the
purpose of calculating external exposure fromsediments. This population is assumed to spend 200 h
a-1 on average on the river bank [B9]. The doses per unit
activity ingested were again consistent with Adams [Al]and ICRP publication 30 (12].
125. The collective absorbed dose commitments fornormalized releases to the freshwater rece1v111g
medium, the model river, are shown in Table 39 for
PWR and BWR systems. The total normalized
collective effective dose equivalent commitment for the
BWR is estimated at 2.8 10-3 man Sv [GW(e) a}-1; twothirds of this dose comes from drinking water and onethird from fish consumption. For drinking water 20% of
the collective effective dose equivalent commitment isfrom 131J and nearly all the dose from ingestion arises
from caesium isotopes. For the PWR the total
normalized collective effective dose equivalentcommitment is 1.0 10-3 man Sv [GW(e) aJ-1 withdrinking water giving 78% of the total. The contribution
from drinking water is given almost equally by 1311,
60Co, t34Cs and 137Cs. For ingestion, the caesium
266

isotopes give more than two-thirds of the total
normalized collective effective dose equivalentcommitment. The normalized collective effective dose
equivalent commitment, weighted for electricity
generated, is 1.4 10-3 man Sv [GW(e) aJ-1,
126. For releases to salt water it is necessary toconsider the dispersion close to the site where theeffects of sedimentation may lead to a reduction in theactivities which become available for dispersion on aregional scale. Using the local marine model describedin Annex A, the values of parameters adopted for themodel site are as follows: volume, 2 km3; water depth,
15 m; rate of sedimentation, I 00 g m-2 a-t; suspended
sediment load, 5 g m-3; rate of renewal of water in thecompartment, 50 a-t. These values have been taken as
typical of a coastal marine site from the data detailed
by Booth et al. [B7] and the study for the Commissionof the European Communities by the National Radiological Protection Board of the United Kingdom jointlywith the Commissariat a l'Energie Atomique of France[ClOJ. In general, from 50% to 90% of the activity
released to the local marine environment becomes
available for the regional dispersal.
127. In the regional dispersal model the sedimentation rates in coastal waters on the continental shelfvary in the range of about 10 to 100 g m-2 a-t [A3, 16]
and the suspended sediment loads are of the order of
0.1 to 6 g m-3 [A3, 16, Cl OJ. with the lower values
applying to the ocean compartments. At any given time
the activity in the water column is partitioned between
the water phase and the suspended sediment material.
The fraction of activity remaining in solution will behigh for elements such as caesium which do not concentrate on sediments, but for other nuclides such as
ruthenium or plutonium the fraction remaining in thewater phase can be reduced by a factor of 10 for
suspended sediment loads approaching 100 g m-3 [H6].
128. Fish-catch data and production of other marine
foodstuffs vary greatly from location to location.
Reported landings of fish have been obtained from theConseil International pour l'Exploitation de la Mer[C3]. Typical values from the Baltic Sea are 4.6 1os t a-1
and the North Sea catch is 2 106 t a-1, while some 3 105 t
a-l are taken from the Arctic ocean. Catches ofmolluscs and crustacea vary in the range of 8 ]03 t a-t
for the Baltic and perhaps 2 1os t a-t for the Bay of
Biscay and the southern North Sea.
129. Using the normalized discharge rates given in
Table 37. the collective effective dose equivalentcommitments from the notional BWR, PWR and GCR
on the model coastal site have been computed using theregional marine model prepared for the Commission of
the European Communities [ClO, C8]. The notional
release was into the receiving waters of the eastern
English Channel. This area is representative of manysites both in Europe and North America.
130. The resulting collective absorbed dose commitments are shown in Table 40. The dose per unit intake
was again derived from the lCRP 30 models andmetabolic data [12] and the results presented by Adams
which uses the same basic data [Al]. The collective
doses were estimated by taking the time integrals of theactivity in the filtrate fraction of the water in eachcompartment and multiplying by the concentration
factors already listed for the marine environment inTable 37. The edible fraction of the marine foodstuffs
has been taken as 50% for fish, Crustacea and molluscs[Cl OJ.

131. A large proportion of the collective dose
commitment for GCRs arises from isotopes of caesium.
The radionuclide releases from the gas-cooled reactors
are higher because they are essentially all coastal sited,
whereas the normalized discharge figures for LWRs are
based largely on inland siting experience. The collective
effective dose equivalent commitment for gas-cooled
reactors is 0.18 man Sv [GW(e) a]-1 which can becompared with the estimate by Hetherington [H5] of
0.24 man Sv [GW(e) a]-1. For BWRs the corresponding
collective effective dose equivalent commitment is 4.2
10-2 man Sv [GW(e) a]-1, mainly coming from 65Zn inmolluscs, and the remainder from fish, with the largest
contributions from caesium and 65Zn. The collective
effective dose equival~nt commitment for PWRs is
6.0 10-3 man Sv [GW(e) a]-1, about one-third comingfrom caesium in fish and many of the isotopes contributing to the mollusc component. The normalized
collective effective dose equivalent commitment,
weighted by electricity production of each reactor typeis 3.5 10-2 man Sv [GW(e) a]-1 for marine discharges.
132. The collective effective dose equivalentcommitment from caesium isotopes is not very
dependent upon the location at which they enter
northern European waters, the variation for unit
discharge being between 150 Bq (integrated intake) for
releases into the eastern Irish Sea, 130 Bq for the
English Channel (east), 120 Bq for the southern North
Sea and 27 Bq for a release into the Bay of Biscay.
These results are per unit release into the relevant
regional compartment from the local model. The results
obtained may certainly be taken as typical of Europe
and the collective effective dose equivalentcommitment is essentially all delivered over the firstfew years after discharge.
133. Again it must be emphasized that the figures
given in Tables 39 and 40 are representative of thegeneration of unit quantity of electricity and should not
be applied to a specific site where particular environmental pathways exist which have not been consideredhere and might lead to significant changes of thecollective dose contributions. The normalized collective
effective dose equivalent commitment due to aquatic
discharges has been estimated on the assumption that
half the discharges are to freshwater and half to marine
environments. The result amounts to about 0.02 man Sv[GW(e) aJ-1.

useful for other purposes. should not be used as a basis
for estimating any speculative future component ofcollective dose commitment. Contributions from
accidents to occupational exposures are dealt with inAnnex H.
136. The 3Ccident at Three Mile Island has been thesubject of many reports, particularly from the UnitedStates Nuclear Regulatory Commission and thePresident's Commission [K9]. The basic form of theaccident was very simple, but the details were extremelycomplicated. The pumps providing feed water to theboilers stopped and the safety system shut down theturbine. A relief valve in the reactor primary cooling
circuit opened correctly and the reactor was shut down
automatically. A second fault was the failure of therelief valve to close when the reactor primary circuit
pressure fell, while the control room instrumentation
indicated closure. As the primary circuit pressure fell,high pressure emergency core cooling was automatically injected into the circuit but this now was stopped
by the operators in the mistaken belief that the reactor
was too full of water. This action was caused byinstrument readings correctly showing high water levels
in part of the system, although it was caused by water
being forced out of the core by the generation of steam.
At this point severe damage to the fuel elements
occurred.
137. Steam and fission products released from thedamaged fuel are now thought to have left the reactor
by a relief valve in the primary water makeup system
which was kept running to maintain the boron concentration. The condensed water was retained in thecontainment building. This building is designed to seal
automatically in the event of a pressure rise, but none
occurred and it took four hours before the building wassealed. During this time fission product gases escapedto the atmosphere, and gave rise to irradiation of the
public. Further releases took place later when contaminated waste water was discharged from thecontainment building. Most fission products wereretained in the water but a release of about 370 PBq ofnoble gases, mainly mxe. and some 550 GBq of 1311
were released to the atmosphere.

134. There have only been two reactor accidents
which are known to have caused measurable irradiationof the public: Three Mile Island in March 1979 and theWindscale reactor accident in October 1957. The latter
accident was at a military reactor, but the collective
dose commitments have been included here on the
grounds that the reactor contributed partly to the development of gas-cooled civil reactors.

138. The accident released large amounts of activity
from failed fuel in the core but the environmental
releases and the resulting exposure of the public wassmall. The collective whole-body dose commitment was
mainly due to the 133Xe release and has been estimatedat between 16 and 35 man Gy within 50 miles [B19]. The
corresponding mean value and the value given in the
Kemeney report [K9] is 20 man Gy. Individual levels ofdose averaged 1.5 10-5 Gy within 50 miles of the plant
and the maximum absorbed dose which any number ofthe public could have received has been estimated at
85 10-5 Gy [K9] from external gamma-irradiation. The
contribution to the collective whole-body dose
commitment due to mxe dispersion beyond 50 miles
can be estimated roughly from the results in Table 29
for the model PWR, where 133Xe dominates the release.
Over all distances the whole-body dose commitment
might be expected to be twice that within 50 miles.

135. It is impossible, on the basis of these two
accidents, to retrospectively calculate a component of
the collective dose commitments due to accidents
involving public exposure from nuclear power. The
Committee has decided that the probabilistic
approaches, which predict the risk of reactor
programmes by extrapolating into the future, while139. In the 1957 Windscale reactor fire, the accident
began during a routine release of the Wigner Energy
stored in the graphite (a phenomenon caused in low
temperature graphite by neutron bombardment, leading
to the creation of interstitials and the stressing of themoderator). Due to errors in operation, the fuel became
overheated and, in the once-through air cooled system,C.

REACTOR ACCIDENTS

Collective dose commitments due to releases ofradioactive materials in accidents267

caught fire. The fire lasted for about three days andmajor releases of iodine occurred on two occasions;once when air now was started through to core in an
attempt to cool it, and secondly when water was
pumped into the reactor which finally extinguished thefire. The Windscale reactor has never been utilized
since.
140. A theoretical re-analysis of the Windscale
accident has been given by Clarke [C6] who estimatedthe activities released and dosirnetric consequences out
to 50-100 km. Extensive environmental measurements
were undertaken at the time of the accident, which canbe used to evaluate the collective doses. Individual
doses were re-examined by Baverstock and Vennari[B20]. The release of 131} has been estimated at some
740 TBq accompanied by 44 TBq 137Cs, 12 TBq 106Ru
and 1.2 PBq of 133Xe. Collective dose estimates were
not made at the time, but measurements of activity inthe thyroids of adult individuals in Leeds and London
indicated thyroid dose commitments of 10-3 and 4 1(}-1
Gy, respectively [820], with young children receiving
doses of twice this value. Maximum doses to local
individuals close to the site were estimated to be of theorder of 10-2 Gy to the thyroid of adults and perhaps
10-1 Gy to children's thyroids [C6, LIO].
141. The contamination of pasture land was widespread, the majority of the released activity passing
south-south-east from Windscale, passing directly
towards London, and eventually passing over Belgium
before turning northwards to Norway. In a new evaluation of the accident to assess the collective doses theprincipal route for irradiation was shown to be iodine
in milk [C9]. Although other nuclides did not deposit asreadily on pasture they have been shown to contributesignificantly to the collective dose. The estimate given
by Crick et al. [C9] for the collective thyroid dose
commitment is 1.8 104 man Gy of which about 25% wasderived from inhalation and 75% from ingestion. The
corresponding collective effective dose equivalentcommitment is 6 102 man Sv. The collective effective
dose equivalent commitment from all isotopes andpathways was estimated to have been 1.3 103 man Sv, of
which somewhat less than 50% was due to iodine
isotopes and thyroid irradiation. External irradiationfrom ground deposits of activity was estimated tocontribute another 40%, while the remaining 10% arose
mainly from ingestion of foodstuffs contaminated by
nuclides other than iodine [C8].

IV.

FUEL REPROCESSING

142. At the fuel reprocessing stage of the nuclear fuelcycle the elements uranium and plutonium in theirradiated nuclear fuel are recovered for use again infission reactors. The spent fuel elements are stored
under water (which serves both for radiation shielding
and for cooling) while waiting for reprocessing. Fuelelements are usually left until all the short-lived isotope
131J has decayed to insignificant amounts (usually atleast 120 days). One reprocessing plant can serve awhole nuclear reactor programme. so that the quantitiesof the nuclides significant from the health point of viewwhich pass through the plant will be rather high inabsolute terms, but may be small per unit of electricalenergy generated. When the fuel elements are reprocessed the irradiated fuel is first taken out from itscanning material and then dissolved in nitric acid. Thisis known as the head-end process.
268

143. A solvent extraction process is next used for the
separation of uranium and plutonium from the fission
products, and the remaining transuranic elements.
Nearly all reprocessing facilities employ the PUREX
process which uses the organic complexing compound
tributyl phosphate to extract both the uranium and theplutonium into the organic phase. The uranium andplutonium can be separately recovered from theorganic phase using nitric acid. The resulting nitrates
are rurthcr purified and then converted to oxides
suitable for storage until they are refabricatcd into fuelelements.
144. The only reprocessing plants operating commercially in the world are at Windscale (United Kingdom).
La Hague and Marcoule (France). In addition, there
are several small experimental reprocessing facilities,
such as the one at Karlsruhe in the Federal Republic of
Germany. The capacity of the Windscale plant is 2 IQ3 t
a-1 and that of La Hague is 900 t a-1 for GCR fuel and
400 t a-1 LWR fuel. The capacity of the La Hague plantfor LWR is thought likely to increase to 800 t a-1 in
1983/84, when essentially all the French GCR fuel will
be reprocessed at Marcoulc. except for some 150 t a-1
which will be used at La Hague to dilute the fast reactor
fuel from Phenix when it is reprocessed at La Hague[LI]. The assessment performed here considers first thecommercial reprocessing plants and then establishes amodel facility for representative calculations ofcollective and individual dose assessment.
A.

EFFLUENTS

145. The design and operation of reprocessing plants
to avoid releases of large amounts of radionuclides iscomplex. The gaseous and volatile fission products (I,3H, C, Kr. Xe, Ru, Tc and Cs) are largely separated
from the fuel solution at the dissolution stage. The
dissolver off-gas is treated for nitric acid recovery andfor iodine removal before being blended with the otheroff-gases from other vessels in the process. The vessel
off-gas is usually treated by caustic scrubbing, drying
and filtering through high efficiency filters before
discharge from a tall stack. The aqueous wastes.
containing almost all the fission products and transuranium elements are concentrated by evaporation andstored in double-containment stainless steel tanks to
await further treatment before disposal.
146. The radionuclides of principal concern in reprocessing plants efnucnts arc primarily the long-lived
nuclides, 3H, 14C, 85Kr. 90Sr, tut>Ru, 1291. 134Cs, mes andisotopes of transuranium elements. There have been
many publications on the releases. environmental
pathway analysis and dosimetric consequences ofreprocessing plant cfnucnts [82, B10, B 11, K2, N17, PS,
P6, V2]. Many other papers are to be found dealing
with the specific nuclides of interest. particularly 3H,
14C, 85Kr and 129J. which have also been the subject or a
special study group sponsored by the Nuclear Energy
Agency of the OECD [N14].
147. In Tables 41 and 42 the annual discharges forWindscalc, La Hague and Marcoule are presented foratmospheric and liquid cfnuent discharges, respectively. The amount of activity in effluents depends not
only upon the specific waste treatment and processing
systems of the reprocessing plant, but also on the typeof fuel reprocessed, its irradiation history and storage(cooling) time. The discharges of liquid effluents from
Marcoule are controlled to lower amounts of radio-

nuclides than rrom the other two sites because Marcou ledischarges into a river, the Rhone, which nows into theMediterranean Sea. The other two sites are coastal anddischarge to the sea. The sources or data for thedischarges were Luykx [Ll. L6] and Dritish Nuclear
Fuels [D2, 1322]. For the liquid discharges fromWindscale an isotopic breakdown or the discharge insuccessive years is given in Table 43.
148. The throughput of fuel at each reprocessing plant
has been calculated on the basis of its reported 85 Kr
discharges and the Committee's estimate of the 85Kr
content of fuels from different reactor types as given in
the following paragraphs. These figures imply that at
Windscale the throughput of rue! in 1978 corresponded
to less than 2 GW(e) a and the 85Kr discharges havefallen from the levels in previous years. The figures for1979 imply an increase in fuel throughput; however thecorresponding amounts of electricity generated by
nuclear power stations do not agree with reported
electrical energy production in the United Kingdom[C4]. The implication is that there is a time lag in reprocessing fuel and that there is a growing backlog of fuelcommitted for reprocessing.
1.

Krypton-SS

149. Krypton-85 is the only noble gas of interest
released from reprocessing plants. It is essentiallyalways totally released from the fuel at the dissolution
stage and, consequently, knowledge of the 85Kr
discharge from a plant can be used to assess thethroughput of fuel. The production of 85K.r per unit
energy generated in typical reactor fuels have beentaken from White [W4] as 4.1 PBq [GW(th) aJ-1 for
GCRs and 3.6 PBq [GW(th) aJ-1 for LWRs, on the basisof 3.5 GW d t-1 and 33 GW d t-1 fuel burnups.
Assuming the average thermal efficiency of a GCR is
30% and that of the LWRs is 31% leads to an inventoryof 14 PBq [GW(e) aJ-1 for GCRs and 11.5 PBq[GW(e) aJ-1 for PWRs. Bernero et al. [BIO] haveestimated for reprocessing fuel from LWRs in the
United States that 45 GW(e) a generates 520 PBq of
85Kr to be released to the atmosphere, corresponding to11.5 PBq [GW(e) a)-I which is similar to the value
derived here.
2.

Tritium

150. It was estimated in paragraph 60 that thenormalized production of tritium in LWR fuel totalledsome 0.75 PBq [GW(e) aJ-1. Using these figures for
GCR fuel, the total tritium inventory passing throughWindscale in 1979 was about 1.9 PBq. In Table 42 it isseen that for the same year liquid efnuent discharges
totalled 1.2 PBq. while Table 1 indicates that 0.29 Pllq
were released to atmosphere. Thus, most of the tritiumin the fuel appears in the effluent streams during reprocessing and perhaps about 20% is emitted to atmosphere. A recent retrospective study on the W AK reprocessing plant in the Federal Republic of Germany
reveals that 2% of the tritium in LWR fuel is released toatmosphere [H7]. Releases of tritium to atmosphere perunit of electrical energy generated. averaged between
1975 and 1979. are 3.2 TBq [GW(e) aJ-1 (La Hague) 46
TBq [GW(e) a]-1 (Marcoule) and 133 TBq [GW(e) a]-1
(Windscale).
151. A portion of the tritium in LWR zircaloy clad
fuel elements is immobilized as a solid zirconium

compound. This has been estimated to account for 15%
or tritium production [K5]. An analysis by Luykx andFraser for European reprocessing plants [LI] indicatedthat for 1973-1974 at Eurochemic (Belgium. no reprocessing since 1974) tritium discharges were 0.48 PBq[GW(e) aJ-1, which corresponds to about 65% of theCommittee's estimated throughput; 26% of this was toatmosphere. For the WAK Karlsruhe plant in theFederal Republic of Germany (1973-1976) the
discharges were 0.44 PBq [GW(e) aJ-1, 60% of
throughput. with 3% being discharged to atmosphere.
3.

Carbon-14

152. Routine measurements of discharges of t4C fromthe Windscale reprocessing plants are now available
and reported in Table 41 [822). Schuettelkopf andHerrm~nn [S8] have reported that for a 2-year period.
when PWR and BWR fuel was being reprocessed at
Karlsruhe. normalized atmospheric discharges of 14C in
the form 14C02 were at a level of between 0.46 and0.51 TBq [GW(e) aJ-1. The t4C content of such fuels istaken as 0.66 TBq [GW(e) aJ-1 based on the results of
Kelly [K2], assuming a thermal efficiency of 31% at thereactor, so that about 75% of the inventory was released
to atmosphere. The t4C levels in fuel are highly
sensitive to nitrogen impurity levels in fuel.
153. For GCRs a normalized production rate within
the fuel of 0.96 TBq [GW(th) aJ-1 is assumed [K2]. andtaking a reactor efficiency of 30% leads to a normalized
production rate of 3.2 TBq [GW(e) a)-t. The potential
discharges from GCRs fuel are greater than those forthe quantity of LWR fuel for the same electricity
production. The throughput of 14C at Windscale in1979 can be assumed to be about 8 TBq on this basis
and from Table 41 it can be seen that the reported
atmospheric release was 3.5 TBq, i.e., 44% of
throughput.
4.

Iodine154. The 131) content of irradiated nuclear fuel varies
depending upon the cooling time and the final power
rating of the fuel before discharge. The 131J normalized
content of LWR fuel cooled to 180 days is estimated tobe 2.8 TBq [GW(e) aJ-1, falling to 1.4 GBq [GW(e) aJ-1
at 270 days and to 550 kBq [GW(e) aJ-1 at 1 year.
For 1291 the arisings depend on burnup and are assessed
to be between 37 GBq and 74 GBq [GW(e) a]-1. Sincefuel is generally cooled to about a year before reprocessing. t31J discharges are generally extremely small.
For 1975-1978 at Windscale atmospheric releases of
131J were averaging 1.7 GBq [GW(e) a)-I, while La
Hague averaged 32 GBq [GW(e) aJ-1 and Marcoule 144
Gllq [GW(e) aJ-1.
155. Atmospheric discharges of 129I are now reported
for Windscale and average 2.2 GBq [GW(e) aJ-1.
Discharge data for the WAK reprocessing plant in theFederal Republic of Germany indicate that in
1975-1976 the 129 1 normalized releases averaged 11 and
0.4 GBq [GW(e) a]-1, representing 25% and 1% of the1291 throughput in fuel [B 12]. The reduction in 1976
resulted from the installation of a new filtration system
for the dissolver off-gases. In a series of measurementsfrom November 1975 to August 1977 the average value
for the components of 129{ discharges were reported as
74% inorganic, 23% organic, 2% aerosol (812]. Theseaverages conceal a wide variation composition, the269

inorganic form having a range of 21-97%, organic
2-54% and elemental 0.04--14%. Annual liquid
discharges of 1291 from Windscale are about 0.1 TBq,
corresponding to about 40 GBq [GW(e) a}-'.
5.

Radioactive aerosols

156. In Table 41 figures are given for total a activity
discharged to atmosphere. Although a wide variety of
CL-emitting nuclidcs are present, United States
experience of a particular plant there suggests that the
majority of the a activity will be from plutonium
isotopes [F4]. Windscale reports (B2) that in 1978 of theactivity associated with the CL-emitting radionuclides
released, 71% was from plutonium isotopes and theremainder was from 241Am and 242Cm. The percentage
isotopic composition of the plutonium a emitters willroughly be as follows:
240Pu

GCR
LWR

16
78

44

9

40
13

0.1
0.04

The normalized release rate for a aerosols fromWindscale in 1979 was 0.4 GBq [GW(e) aJ-1.
157. The beta aerosol results are also given in
Table 41, the largest component at Windscale being
mes discharges in aerosols. In addition, 90Sr is separately identified, the remaining activity comprising thefollowing nuclides: 95Zr/Nb, I06Ru, t25Sb, t34Cs, 144Ce.
The discharges of 137Cs and 9DSr have been substantially higher since 1974 compared with previous years,
reflecting higher discharges from the solid waste siteused primarily for the storage of Magnox fuel cladding
and these discharges are expected to be reduced in thenear future [L4). Of these discharges about half arereleased from high stacks. Approximately 50% of themes and 90Sr discharges took place from stacks ofheight less than 46 m [H8) and are associated with highactive silos and cooling pond water. Site discharges of aactivity to atmosphere mainly arise from the researchand development laboratories and from the plutonium
recovery plant [L4).
6.

B.

LOCAL AND REGIONAL COLLECTIVE DOSECOMMITMENTS.

160. The evaluation of the collective dose commitments from reprocessing nuclear fuel requires a study
both of the local and regional effects and of the global
consequences of the releases. Estimates of the local andregional contribution to collective dose commitments
are given here and the global contribution in chapter V.
Monitoring of the local environment and regional
distribution of activity from Windscale is extensive andwidely reported in annual reports [82, 822, H5, M4,
M5) and in the open literature [H8, H9, P5, P6, WSJ.
Studies on the fate of liquid effluents from La Hague
have also recently been published [Y2]. In this section
the collective absorbed dose commitments areevaluated for normalized discharges from Windscaleand La Hague. However, since there are only threeknown operating commercial reprocessing plants, theircontributions of collective dose per unit of electricalenergy generated may not be representative of thatwhich will be made by other installations when theycommence processing irradiated fuel. Therefore asummary of collective dose commitments is given for
typical discharge figures quoted for new designs ofreprocessing plant. The characteristics of the
population data and meteorological conditions used forthe model plant are those applicable to the Windscalesite in the United Kingdom and are shown in Table 44.

Liquid effluents

158. The liquid effluents discharged from Windscale,
La Hague and Marcoule are shown in Table 42 for total
a, total ~ and the releases of tritium, 90Sr and 106Ru. A
wide range of nuclides appears in waste streams and theisotopic composition of Windscale liquid effluent
discharges for 1977. 1978 and 1979 are shown in Table
43. The a-isotopic composition in Windscale liquid
effluents largely reflects the isotopic composition of
GCR fuel [H9). Americium-241 has also contributed
significantly to Windscale releases, although thepercentage has been decreasing from over 50% of thetotal a-discharges in 1972 to about 10% of this value
since 1977: this reduction is due to more efficient
removal of 241Am from waste streams prior to discharge[W5]. The normalized release rate of a activity fromWindscaleintoliquid effluentsiscurrently
25 TBq [GW(e)aJ-1, and the figures for Marcoule and
La Hague are 0.016 and 0.24 TBq [GW(e) aJ-1, respectively. The Marcou le discharges are to the river Rhone.
159. The liquid discharges of ~ activity shown in
Table 42 give normalized release rates of 3.7, 0.52 and0.04 PBq [GW(e) a)-I for Windscale, La Hague andMarcoule, respectively, in 1978. The isotopic composition of the release varies between La Hague and270

Windscale. In liquid effluents from La Hague up to
1976. 65% of the total fl activity concentration was from106 Ru with 13% from 137Cs. 12% from 144Ce and 7.6%
from 9 DSr. The isotopic composition for Windscale is
63% mes and 134Cs, 11% JU{,Ru, 8.3% 90Sr. theremainder being composed of a number of nuclides.
The Windscale caesium discharges which began to risein 1974 are not due to reprocessing as such but are theresult of the corrosion of Magnox fuel cans by water inthe storage pond. Corrective action has now been
undertaken to reduce the discharge levels hy the use ofion exchange resins in the effluent treatment plant. The
result has been that in 1979 the liquid efnuent discharge
was reduced to 1.6 PI3q [GW(e) aJ-1.

t.

Krypton-85

161. The assessment of the averaged 85Kr discharges
from Windscale between 1975 and 1979 has been
performed using the methodology provided for theCommission of the European Communities [ClOJ. The
average annual release of 8SKr was 35 PBq and theresulting collective absorbed dose commitments areshown in the summary Table 46. The collective effective
dose equivalent commitment is estimated to be 0.074
man Gy from the cloud y-irradiation, using the doseconversion factors of Poston and Snyder [P4). A further
contribution from the cloud-P irradiation amounts to 19
man Sv collective absorbed dose in skin, which contributes 0.19 man Sv to the collective effective dose equivalent commitment. The normalized local and regional
collective effective dose equivalent commitment for the
85Kr annual discharge from Windscale averaged from1975 to 1979 is 0.1 man Sv [GW(e) aJ-1.
2.

Tritium and carbon-14

162. In this case a specific activity model is used as
described in paragraphs 99 and 100. The averageannual release of tritium to atmosphere from Windscale
between 1975 and 1979 was 0.33 PBq, or 0.13 PBq[GW(e) a]-•. The associated normalized collective dose
commitments are shown in Table 45. The specilic
activity model chosen with the dosimetric results ofAdams [A 1] suggests that the individual organ doses(thyroid. gonad, etc.) expressed in gray are essentiallythe same as the effective dose equivalent expressed insievert. The normalized release of 3H to atmosphereleads to a local and regional collective effective dose
equivalent commitment of 0.35 man Sv [GW(e) a]-•.
Releases lo the regional marine environment lead tolower dose commitments: the normalized tritium
release from Windscale in the period 1975-1979 was
0.45 PBq [GW(e) a]-1 (Table 42) leading to a collective
effective dose equivalent commitment of about 8 IQ-4
man Sv [GW(e) a]-1, using the distributions of marine
foodstuffs, water movements and sedimentation rates
taken from chapter I I I.
163. For l4C atmospheric releases the normalized
collective effective dose equivalent commitment in thelocal and regional population is 0.69 man Sv[GW(e) a]-1 for the population distribution within 2000
km of the Windscale site. The average annual
throughput of fuel at Windscale from 1975 to 1979
corresponds to 2.6 GW{e) a, leading to a collective
effective dose equivalent commitment per year of
release of 1.8 man Sv. If the same discharge took place
to the marine environment the regional model suggests
the integrated activity found in foodstuffs would be 2.5
lQ-4 Bq a per Bq discharged. Thus the same annual
release in the marine environment would lead to 2.7
man Sv, essentially the same ligure as for the release toatmosphere.
3.

marine foodstuffs from the areas of northern European
waters. as described in Annex A and paragraphs
126-128. The releases occur to a local marine modeland then subsequently into the regional model. The
experimental evidence strongly suggests that a large
proportion of the plutonium discharges into the eastern
Irish Sea is rapidly fixed on to sediments in the local
dispersion and only a small fraction is available fortransfer into the regional waters. The large suspended
sediment load in the Irish Sea accounts for the considerable difference in plutonium dose commitments perunit discharge between the Channel site and the eastern
Irish Sea site.
166. In Table 47 the annual average releases fromWindscale and La Hague from 1975 to 1979 have also
been used with the Committee's models to obtaincollective effective dose equivalent commitments. For
Windscale the result is 311 man Sv of which 90% is due
to mes' discharges and the remainder nearly all due to
106Ru. The normalized collective dose commitment isthus 124 man Sv [GW(e) a)-I. For the La Hague
discharges the collective effective dose equivalentcommitment is 84 man Sv with 86% due to 106Ru andthe remainder due to 137Cs discharges. The normalized
collective dose commitment for La Hague is 53 man Sv[GW(e) aJ-1, although this is probably an overestimate
because of the assumptions in the models used. The
normalized results for La Hague and Windscale areseen to be similar. From Table 43 it is seen that in 1979
the releases of caesium and ruthenium from Windscale
to the marine environment were half those in 1978, sothat thecollective effective dose
equivalentcommitment will be reduced to 62 man Sv [GW(e) a]-I,
Annual effective dose equivalents to most exposed
individuals near Windscale were probably less than
0.5 mSv from 1979 discharges of 137Cs.

Other atmospheric releases

164. Of the other nuclides released to the atmosphere,
apart from 129J which makes a significant contribution
to the global collective dose commitment because of its
Jong half-life (1.6 107 a) and is considered in detail inchapter V, the remainder contribute only to the local
and regional dose commitment. Again, the spatial
patterns of population distribution and of agricultural
production applicable to the United Kingdom andEurope were taken from Simmonds and Linsley [S12]and the study undertaken for the Commission of theEuropean Communities [ClO]. The release rates of thevarious isotopes discharged to atmosphere fromWindscale given in Table 41 were used and theresulting collective dose commitments for those
nuclides contributing are shown in Table 46. The totalcollective effective dose equivalent commitment comesto 6.2 man Sv as a result of one year of average atmospheric discharges between 1975 and 1979. The result isdominated by the collective effective dose equivalentcommitment from ingestion (4.4 man Sv). Of theremainder, 1 man Sv arises from ground deposition(137Cs), 0.48 man Sv from inhalation (3H, 239Pu, 240Pu.
241Am) and 0.26 man Sv from BSKr doses in the local
and regional population.
4.

Liquid effluents

165. Jn Table 47 the collective absorbed dose commitments are presented for releases of the nuclides inliquid effluents from both the Windscale and La Hague
sites averaged from 1975 to 1979. The dose commitments include doses from the consumption of all

5.

Atmospheric and liquid effluents from a notional
plant

167. There are significant uncertainties in predicting
dose commitments over the long periods of time during
which the long-lived isotopes may be available in theenvironment. Furthermore, the collective dose commitments derived from present day releases fromWindscale and La Hague are probably not representative of future reprocessing practices if all L WR fuelcurrently stored were reprocessed. Notional reprocessing plant designs exist, some of which are sitedinland. with very low discharges to the aquatic
environment. Consideration is also being given to
retention of SSKr, 3H. 14C and 129). nuclides which are
currently released into the atmosphere. In order toobtain a more representative set of collective dose
commitments. a model oxide fuel reprocessing facility
is assumed which discharges into a marine
environment. The discharges are based on the implieddischarges of the British Nuclear Fuels Ltd. thermaloxide reprocessing plant (THORP) as reported at theWindscale Inquiry [PS]. Table 48 gives a summary of
the notional discharges per unit energy generated forthe model facility assuming a fuel throughput equivalent to 40 GW(e) power and the associated collective
effective dose equivalent commitments. It has been
assumed that there is no retention of 14C or 85Kr in
atmospheric discharges and in this case 14C is the main
contributor to the collective effective dose equivalentcommitment from atmospheric releases, although the
14C content of LWR fuels is much less than for Magnox(paragraphs 66-67). On the basis of the very notional271

discharges assumed here, the liquid effluents are as
significant as the atmospheric releases with 137Cs and
106Ru being the main contributors. Jodine-129 releases
contribute to the global collective dose and this isconsidered in chapter V. The representative local andregional normalized collective effective dose equivalentcommitment from reprocessing is thus 1 man Sv[GW(e) aJ-t.
168. The representative values of doses to criticalgroups from the discharges at the model facility arepresented in Table 49. For atmospheric releases thelargest single contribution to individual effective dose
equivalent is from 14C and a representative annual
effective dose equivalent to the critical group is 25 µSv.
The liquid effluents contribute more to the total
effective dose equivalent, with mes being the mostimportant nuclide. The results must only be used toindicate the order of magnitude of doses foreseen indesigns of reprocessing plants. The annual effective
dose equivalent to the most exposed group amounts to
200 µSv for marine discharges. These doses for criticalgroup exposures are similar to those from the proposed
Barnwell Nuclear Fuel Plant in South Carolina. UnitedStates [P9].

V.

COLLECTIVE DOSE COMMITMENTS
FROM THE GLOBAL DISPERSION OF
RADIONUCLIDES

169. The radionuclides that contribute to the global
collective dose commitment are those which are sufficiently long-lived and readily migrate through theenvironment, thereby achieving widespread distribution. Those of primary interest are 3H, 14C, 85Kr
and 129) and assessments of the global collective doses
from the release of these nuclides have appeared inseveral reports [ClO, K2, N14]. The environmental
transfer of 3H and 85Kr is becoming fairly wellestablished and more reliable estimates of collective
dose commitments can be made. Long-lived nuclides
such as 137Cs and 239Pu are less mobile in theenvironment and become far less dispersed after
deposition onto soil or onto sediments, following
release into the local region. Some other nuclides are
considered which are released in the nuclear fuel cycle
in a mobile form and have sufficiently long half-lives toirradiate populations beyond the regional area.
170. The very long-lived nuclides, for example, 129J
(1.6 101 a) pose a special problem because extrapolation
far into the future is required to estimate the dose
commitment. This introduces uncertainties because of
the unknown population size, its dietary habits and,amongst other things, environmental changes. Thisimplies that little weight can be placed on dose commitments for decision-making purposes. Estimates of theincomplete collective dose commitment from these
nuclides is useful firstly in demonstrating the time
distribution of the dose commitment, and secondly to
calculate an estimate of the per caput annual doses
resulting from a finite continuing practice. Thus, if
electrical energy production by nuclear fission has afinite duration, of the order of a few hundred years. theincomplete collective dose commitment to that time
divided by the mean population size will indicate themaximum per caput dose rate that may be experienced
in the future from that practice. Collective dose
commitments must be treated with caution since thesignificance of the results must be carefully interpreted.
This topic is fully ·discussed in Annex A.
272

171. ln the following paragraphs collective dose
commitments (and incomplete values where of interest)
have been estimated for those nuclides which become
globally dispersed and irradiate the world population.
This dose is received in addition to that received during
the local and regional dispersion, the so-called first passdose, from the point of discharge. Recently a report has
been published by the Nuclear Energy Agency of theOECD which discusses the collective dose commitments from releases of 3H, t4C, 85Kr [N14]. In thischapter extensive reference is made to the results
obtained by the study group who produced the reportfor the Nuclear Energy Agency.

A.

KRYPTON-85

172. The collective dose commitments from B5Kr
generated during nuclear power production arise
almost entirely from the release at reprocessing plants,
as reactor releases are small in comparison. Since
krypton is an inert gas it disperses through the atmosphere and achieves a uniform concentration in a period
of about two years. Although the assumption of instantaneous dilution in the world's atmosphere, as used in
Annex E. gives a reasonable estimate of time integral ofconcentration, a two-compartment model similar tothat proposed by Kelly et al. [K2] is used in which thereleased krypton is assumed to be instantaneously
dispersed throughout the troposphere of the northernhemisphere which is assumed to have a height of 10 km
and a mass of 1.9 1021 g. Exchanges take place between
the troposhere of the two hemispheres with a half timeof about 2 years. Within a few years the 85Kr becomes
uniformly dispersed and the sole removal mechanism isradioactive decay.

173. The concentration in air and the time integral ofactivity concentration of 85Kr for a discharge of 1 Bq
s-1 for a year is shown in Figure I. The whole-bodyabsorbed dose commitment per unit time integral of air
concentration of 85Kr has been taken as 4.3 1Q-9 Gy (Bqa kg-1)-1 [Cl OJ, and the collective skin dose commitment
due to the ~--irradiation as 5.4 10-1 Gy (Bq a kg-1)-t.
These values are in good agreement with the dose rates
per unit air concentration of 85Kr as given by ICRP
publication 30 [12]. The collective absorbed dose
commitment from the gamma contribution is 5.5 10-19
man Gy for a release of 1 Bq s-1 for one year, while thecollective absorbed dose in skin for the same release is
7.4 10-11 man Gy.
174. The normalized 85Kr discharges from reprocessing of nuclear fuel have been given in paragraph
149 as 11 PBq [GW(e) aJ-1 for a typical LWR and
14 PBq [GW(e) a]-1 for a GCR. The corresponding
collective effective dose equivalent commitments arecalculated, assuming a world population of 4 109 during
the irradiation period,'to be 0.81 man Sv [GW(e) a]-1 ofreprocessed LWR fuel and 0.98 man Sv [GW(e) a)-I for
GCR fuel for the external gamma irradiation. The
collective absorbed dose commitment to skin is 110
man Gy [GW(e) aJ-1 for LWR and 130 man Gy[GW(e) aJ-1 for GCR fuel, which adds 1.1 and 1.3 man
Sv [GW(e) aJ-1 to the collective effective dose equivalentcommitments, respectively. The total collective effective
dose equivalent commitment weighted for energy
produced is 1.9 man Sv [GW(e) a]-1. All of the dose
commitment is delivered within the first 50 years
following the year of release. These results comparewith the Committee's previous estimate in Annex D of
the 1977 report [Ul] of a gonadal dose commitment of

0.9 man Sv [GW(c) aJ-1. The results of the Nuclear
Energy Agency expert group report give a collective
effective dose equivalent commitment for 85Kr global
dispersion of 0.81 man Sv [GW(e) aJ-1 for LWR fuel, if
the same world population is assumed as used here.
ll.

whether to atmosphere or hydrosphere, is immediately
dispersed and exchanged with the hydrogen content of
the circulating waters of the hemisphere into which thedischarge is made.
177. The subsequent circulation of tritium is determined, however. by the exchange of waters between thetwo hemispheres and the deep oceans. The model is
described in Annex A and the exchange rates taken are
shown below. The mass transfer rates are obtained bymultiplying the exchange rates by the relevant mass ofeach component.

TRITIUM175. The collective dose to the global population fromreactor and reprocessing plant releases of tritium canbe estimated from the evidence concluded from fallout
measurements following atmospheric weapons testing.
The absorbed dose commitment has been estimated tobe 5 1()-4 man Sv per TBq released (see Annex E).

Circulating waters (N. hemisphere) circulating waters (S. hemisphere)
Circulating waters (S. hemisphere) circulating waters (N. hemisphere)
Exchange between deep oceans (N. hemisphere~ S. hemisphere)
Exchange between circulating surface
waters and deep oceans
Exchange between deep oceans andcirculating waters

176. In recent years a number of models have been
proposed to describe the global circulation of tritiumfor predicting dose commitments [B13, 814, 815, ClO,
K2, N20]. In a very simple model emissions to atmosphere or to hydrosphere arc not distinguished. since theexchange of water between the atmosphere and theremaining circulating waters of the globe is rapid. Insuch models it has been usual to assume that thereleased tritium is immediately and uniformly distributed in the circulating waters of the northern hemisphere to a depth of 75 m [B14, B15. Ul). More recent
models [ClO, K2] still assume that discharged tritium,

0.1

a- 1

0.05

a-1

0.005 a-1
0.1

a-I

0.0014 a-I

Using this model, the time integral of activity concentration in the circulating waters of the Northern hemisphere is shown in Figure I to be 1.4 10-11 Bq a kg-I fora release of 1 Bq s-t for a year. Assuming an individual

10-11 ~ - - - - - - 1-- - - I'...................... ,

''
·;;;
C

~ ,o·n~,'"'"'Troposhere, northern hemisph&re

co

''

.E
~

'\

~

10-10?

-~
E

\

C

.g

~

\

~

8

\

~

0

\
\

8

Il
.
11

I
\

10-"L _ _ _

er

·.;

\

0

.

85Kr

--

1

~-

10

10· ~
'-~---------100

TIME (alII

10-11 , - - - - - - - - ~3HCirculating water. 11orthern hemisphere ,

"'"'
C'

co

'
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0
C

.2

;

''

\
\

\
\

\

\

j ,o-" }-- - - - - - ·------io\ - - - -1()()

TIME (alFigure I. Time variation and time Integrals of environmental concentrations of 3H
and 85Kr for releases ol 1 Bq , - 1 during one year [C10]273intake of water of930 kg a-1, as given by ICRP in publication 23 [15], a population of 4 109, and a dose per unit
intake of 1.7 I 0-11 Gy Bq-1 (A I, 12]. the collective
effective dose equivalent commitment per unit releaseis estimated to be 2.8 I0-5 man Gy TBq-1. Assuming thenormalized release of tritium in reprocessing to be
0.6 Pllq [GW(e) aJ-1 plus about 40 TRq [GW(e) aJ-'for the average of reactor discharges, gives a total of
0.64 PBq [GW(e) aJ-1 and a normalized collective
effective dose equivalent commitment of 1.8 !(}-2 man
Sv [GW(e) aJ-1. The result is directly proportional to theassumed population size.
178. The Nuclear Energy Agency expert groupassumed a population of 1010 so the present model
would predict 6.8 J(}-5 man Gy TBq-1. The expert group
derived a value of 4.6 1(}-4 man Gy TBq-1 and theCommittee, in Annex D of its 1977 report [UI), gave avalue of 8.1 1(}-4 man Gy TBq-1. The figure derived
from the comparison of the natural dose rate andproduction of tritium (see Annexes B and E) would be
about 5 J(}-4 man Gy TBq-1 for a uniform production
over the world, assuming a global population of 4 1Q9.
One reason for the smaller value obtained in this Annex
is the inclusion of the removal rates of tritium to deep
oceans which leads to a reduction of a factor of 4 to 5 inintegrated tritium intakes. The local and regional
effective dose equivalent commitment per unit energy
generated for atmospheric releases of tritium amount toabout 0.5 man Sv [GW(e) aJ-1, so that the global
contribution is found to be small in comparison(paragraph 102).

compartment to which it is released. The subsequentcirculation between compartments governs the specificactivity of 14C in all carbon inhaled or ingested by man,
which is assumed to be taken from the circulating
carbon in the northern hemisphere. The mass of circulating carbon in the northern hemisphere has been
estimated at 3.6 1014 kg by Ekdahl et al. [E5] and thisvalue is assumed to remain constant.
181. In Figure II the airborne concentration and thetime integrals of activity concentration in carbon are
shown for a release of I Bq s-1 for a year to either theatmospheric or aquatic environments. The doses in all

T
14c
ChtlJl,1tlno csirbon,
oorlh1,1n hr!m1sphgre§

8. 10· 11

<J

!!

0
10- 9

__g 1o·•L

~

g
1o'

10'

10"

1o'

1o>

,=

TIME f•I

CARBON-14

179. The global dose commitment from ' 4 C releases
from the nuclear industry is estimated by constructing
compartment models which renect the environmental
distribution and behaviour of naturally-produced t4C.
Models of varying degrees of complexity have been
produced using between 2 and 20 compartments [H2,K2, K6, l'v12, S13, Ul]. In the two-compartment model
the stratosphere. troposphere and ocean surface areconsidered as one compartment which exchanges withthe deep ocean [H2]. In the 20-compartment model
there are four latitude bands each with compartmentsin the stratosphere, troposphere, terrestrial biosphere,
mixed ocean and deep ocean [M6]. An additional input
to some models is the increased stable carbon in theatmosphere due to the combustion of fossil fuels. whichtends to decrease the specific activity of 14C. It is noteasy, therefore, to evaluate the various estimates of dose
commitment. However, there is evidence [K6] that,
owing to the long half-life of 14C compared with therate of environmental transport of carbon, estimates ofdose commitment are fairly insensitive to the details of
the environmental models except for the rate of transfer
to the deep ocean which affects the result.

'I

'

§

'\

8. 10·' -

t

c
C
0

f ,o~"
8

£

e

10··

L-----~----~_..___, ,o·· tr
10•180. The model chosen for the present estimates has
been described in Annex A and is an 8-compartment
model allowing for two hemispheres comprising
humus, circulating carbon, surface ocean and deep
ocean. The circulating carbon compartment represents
carbon in the troposphere and those parts of the terrestrial biosphere which are subject to rapid cycles ofgrowth and decomposition. The humus compartment
represents the carbon content of the terrestrial
biosphere which circulates more slowly. Carbon-14
discharged to atmosphere or hydrosphere is assumed to
be instantaneously mixed with the carbon in the
274

ie
0

8 10-J.) ~ - ~ - - - . - - - - , - - - - . - ~ ' , - - - ~ 10·•
10'

C.

~
8

1o'

HY

1a'

TIMEC,J

10',=

Figure ll. Time variation and time Integrals of environmental concentrations of 14C and 129) for releases of
1 Bq s-1 during one year [C10]organs and tissues are assumed to be equal and theeffective dose equivalent per unit intake is taken as
5.7 10-10 Sv Bq-1 for ingestion and 6.4 J(}-12 Sv Bq-1 forinhaled C02. The Reference Man (15] annual intakes ofcarbon are taken as 93 kg a-1 ingested and 1.2 kg a-1
inhaled. A population of 1010 is assumed for this calculation and is taken to be constant during the integration

period. This figure represents an estimated equilibrium
population figure during the irradiation.
182. The collective effective dose equivalentcommitment per unit release is estimated at 67 man Sv
TBq-1 as an average for both atmospheric and aquatic
releases. This compares with an estimate of the incomplete effective dose equivalent commitment per unit
release given by the Nuclear Energy Agency expert
group of 70 man Sv Tllq-1 for the same population
figure used here. The expert group, incomplete
collective effective dose equivalent commitment was to104 years, during which time they expected 67% of thedose commitment to have been delivered [N14); thus
their estimate of the complete collective effective dose
equivalent commitment per unit release is about 105
man Sv TBq-1. The answers are thus in good agreement
given the assumptions made and the associated uncertainties over long time periods. Based on the dose rates
due to natural 14C production (Annex 13), the collective
effective dose equivalent commitment per unit releaseis 120 man Sv TBq-1, confirming the results of themodels.
183. The estimated local and regional collective
effective dose equivalent commitments given in Tables
33 and 45 correspond to 2.0 or 0.5 man Sv TBq-1,
depending upon site, and thus represents a few per centof the total collective effective dose equivalentcommitment per unit release for 14C. The normalized
release of 14C is given as 0.4 TBq [GW(e) aJ-1 from themodel reprocessing plant handling L WR fuel (Table48), and LWR reactor releases can be expected tocontribute about the same amount (Table 33). The
normalized collective effective dose equivalentcommitment is thus 54 man Sv [GW(e) a)-I for LWRs,
only half of which arises from atmospheric discharges
at the reprocessing plant. HWR reactor operation leads
to 17 TBq [GW(e) a]-1 (Table 33) and reprocessing
would give little addition; the global collective effective
dose equivalent commitment per unit energy generated
in HWRs is thus 1.1 103 man Sv [GW(e) a}-1. The
1000-year incomplete values are about 20% of theinfinite values. 11 man Sv [GW(e) aJ-1 for LWR releases
and 200 man Sv [GW(e) a)-I for HWR operation. The
1000-year incomplete collective effective dose equivalent commitment is 21 man Sv [GW(e) aJ-1 averaged
over electricity production by each reactor type, andthe infinite value is 110 man Sv [GW(e) a)-I.

D.

lODINE-129

184. Iodine-129 emitted to the atmosphere, because ofits mobility in the environment, becomes rapidly incorporated in foodstuffs ingested by individuals. Iodine in
the environment is not uniformly distributed and thehighest concentrations are found in sea water. Models
proposed for the environmental transport of 1291 havebeen discussed in Annex A. and range from simple
assumptions of instantaneous mixing in the oceans withstable iodine. as in Annex D of the 1977 report [Ul ), tomulticompartment models taking account of transfer
between surface waters and the deep ocean [K7] andbetween these waters and the land and atmosphere[B16]. The most important feature to be included
appears to be the transfer to the deep oceans [K7]. The
model utilized here probably leads to an overestimateof collective dose commitment since effects such as
sedimentation are not taken into account. It isextremely difficult to estimate dose commitments forperiods of' tens of millions of years. The increasing
uncertainty with which doses can be predicted in the far
future means that less weight can be placed on them fordecision making.
185. In Figure II the integrated act1v1ty concentrations of 1291 are given for a 1 Bq s-1 for a year's
discharge whether to atmospheric or aquatic environments. Once more an equilibrium population of 1010 isassumed and held constant throughout the irradiation
time. These integrated concentrations are per kg of
stable iodine so that collective dose commitments arefound by multiplying the integral of activity concentration by the population size, the individual rate ofintake of iodine and the committed dose per unit
intake. The thyroid dose equivalent is 3.3 lfr-6 Sv Bq-1
ingested, giving an effective dose equivalent per unit
intake of 9.9 1~ Sv Bq-1 ingested [Al, I2]. The intake
of iodin,c is 7 J0-5 kg a-I [15]. The collective effective
dose equivalent commitment per unit release is thus
found to be 1.4 104 man Sv TBq-1, while the value
truncated to 104 years would be 4.1 man Sv TBq-1. The
answers must clearly be very uncertain for this nuclide
because most of the dose commitment is delivered
beyond 1Q6 years into the future.
186. The incomplete collective effective dose equivalent commitment per unit 129I release for the same
population as used here was estimated by the Nuclear
Energy Agency expert group for an integration time of
104 years as 8.1 man Sv TBq-1 [Nl4). The NEA groupcalculated 104 year incomplete collective dose equivalent commitments because they assumed this to be afeasible containment period and thus they were interested in detriment averted. The value here is just overhalf that value at 104 years, but the collective dose isincreasing steeply with time at this point and the results
are very sensitive to parameter values. The NEA groupcomment that the complete collective effective dose
equivalent commitment per unit 129J release is some3500 times the 104 year value, i.e., 2.8 104 man Sv TBq-1.
compared with the Committee's current estimate of 1.4
104 man Sv TBq-1. A factor of about 3400 has beenfound here between the 104 year incomplete value andthe complete dose commitment. The normalized release
for the model reprocessing plant discharge was 0.2 G Bq
[GW(e) aJ-1 (Table 48) to atmosphere and 40 Gl3q
[GW(e) aJ-1 in liquid effluents, giving a total of 40 GBq[GW(e) a)-I.
187. The resulting normalized global collective
effective dose equivalent commitment for 129I releases
is thus 560 man Sv [GW(e) a)-I from reprocessing plantdischarges, reactor releases being insignificant bycomparison. The collective thyroid dose equivalentcommitment is about 19 OOO man Sv [GW(e) aJ-1. The
local and regional collective effective dose equivalentcommitment has been estimated as 4 10-3 man Sv[GW(e) a]-1 for atmospheric releases and 8 10-3 man Sv[GW(e) a)-I for liquid discharges (Table 48). The local
and regional collective dose commitments are delivered
over a relatively short time scale due to environmental
removal processes (less than 500 a) and over the sametime period the normalized global collective effective
dose equivalent commitment is about 3 10-2 man Sv[GW(e) a)-I.
E.

SUMMARY

188. A summary of the normalized releases of thenuclides of global significance, together with their275

collective effective dose equivalent commitments. areshown in Table 50. Incomplete values of collective dose
commitment are shown to illustrate their time dependence. It should he noted that, in addition to 3H, t4C,
85 Kr and 129J, a fraction of those radionuclides released
into the atmosphere with hair-lives greater than a few
days and deposition velocities of less than about 10-3 m
s- 1 will leave the regional area rind potentially become
globally dispersed. Their contribution to the collective
dose commitment beyond the regional area has been
estimated using the 85Kr model and results. It may be
deduced that the global dose commitment increases
with increasing half-life and equals that from theregional area when the half-life is about- 0.5 a. Shortlived nuclides such as mxe (T112 5.3 d) and DJ I (8.05 d)in organic form can thus only be expected to add a few
per cent to the regional dose commitment by becoming
globally dispersed. The release of 35S (T 112 87 .5 d)
however would contribute an extra 50% of the local andregional dose commitment. These three nuclides
contribute 2 10-3 man Sv [GW(e) a]-1 from global
dispersion which is 0.05% of the collective dose
commitment from reactor operation.

VI.

RADIOACTIVE WASTE STORAGE ANDDISPOSAL

189. Only a small fraction of the artificial activity
generated in the nuclear fuel cycle is released to theenvironment during normal operations. Except for theuranium. plutonium, and certain other nuclides present
in spent fuel, the radionuclides generated by the
nuclear fuel cycle are generally considered as waste
which must be subjected to suitable treatment followed
by storage or disposal. Storage is taken to mean any
arrangement intended to enable retrieval of the waste atsome future time; the waste may be temporarily
inaccessible, but there is an intent, by surveillance anddocumentation, to retrieve. Disposal implies that
control over the waste has been relinquished. It is useful
to distinguish between high-level wastes, which may
consist of unreprocessed fuel or may arise in liquid
form following reprocessing of nuclear fuel and containmore than 99% of the fission product and actinide
radionuclides in the fuel, and the low and intermediatelevel wastes which arise both in reprocessing and inreactor operation. The latter are, for example, used ion
exchange resins, air and liquid filters, in-reactor components (control rods. instrumentation). contaminatedclothing and equipment. A further category of waste is
known as plutonium contaminated material, which is a
low-level waste but because of the long half-lives of the
a-emitting nuclides covered in this category, needs tobe treated similarly to intermediate level wastes.

A.

LOW AND INTERMEDIATE LEVEL WASTES

190. The tailings from uranium milling are one
example of low-level solid waste, the radiological
exposures from which have been considered in chapter
1. Solid wastes may be treated by compaction or incineration to reduce their volume before being placed instorage or disposed of in shallow land or deep burialsites. Most nuclear sites have provision for the storageof solid wastes generated through the reactor lifetimeand the final disposition of these wastes probably will
depend upon the decision as to the final method of
disposal of the reactor itself.
276

191. It is common practice to dump low-activity solid
wastes untreated into trenches and to cover with earth.
Jn the United States there were six commercial lowlevel waste burial facilities and five major active sites
operated by the Department of Energy in 1978 [A4].
Three of the commercial facilities are currentlv
inoperative, three of the sites are closed indefinitely andresponsibility for perpetual care is expected to beplaced upon the respective States. In general the wastes
arc placed as received into trenches excavated in theexisting soil and the removed material used to cover thewastes once the trench is filled to capacity. The
overburden is sometimes compacted and usually
mounded to promote water run-off. The capacity of thecurrently open commercial sites is between 105 and 106
m3 and typical areas would be of the order of 1.5 km2,
in low population density often semi-arid areas. A
recent review of the radioactive waste inventory at theMaxey. Flats site [G10] revealed the following radionuclides which are attributable to the nuclear fuel cycle:
20 PBq
1.3 PBq
3.0 PBq
0.56 PBq

mes

216Ra
2l8Pu

239Pu

0.93 PBq
0.18 PBq
1.6 PBq
0.14 PBq

192. Emissions from waste handling operations havebeen studied at Maxey Flats [A4] and also at otherwaste disposal sites [D4, LS, M7, S17]. Atmosphericreleases from normal waste handling and from Ieachate
evaporators have been studied by Blanchard et al. [B 17]
and assessments made by the United States Environmental Protection Agency [El] give estimates ofcollective effective dose equivalent commitment arisingfrom one year of release of 4.2 10-2 man Sv per facility,
leading to normalized collective effective dose equivalent commitments of less than 4 J0-5 man Sv[GW(e) a]-1.
193. Disposal of low-level packaged radioactive waste
in the deep ocean is governed by the International
Convention on the Prevention of Marine Pollution by
Dumping of Wastes and Other Matter (the London
Convention). The special permits granted by theappropriate national authorities for radioactive waste
should take into account the relevant IAEA definitionof high-level waste unsuitable for disposal at sea andrecommendations for the purpose of the London
Convention. In addition, sea disposal operations undertaken by NEA member countries are required toconform to the terms of the Multilateral Consultationand Surveillance Mechanism which was established in1979 by a decision of the OECD Council. Thismechanism was set up to further the objectives of the
London Convention and provides a framework forcooperation between participating countries. It
provides for the development of standards, guidelines.
practices and procedures for sea disposal operations
and for international consultation on and surveillanceof operations to verify that these are carried out maccordance with internationally established rules.
B.

HIGH-LEVEL WASTES

194. The majority of irradiated nuclear fuel which has
been removed from reactors is currently stored awaiting
national decisions on whether to dispose of the fueldirectly or to reprocess and recycle fissile nuclides.
When reprocessing takes place, high-level wastes are
currently stored as liquids. The intent is to solidify insome manner to facilitate further handling, storage andIeventually disposal. In France a decision has been
made about the process for solidifying high-level wastes
and a vitrification plant was commissioned at Marcoulein \978. After about a year's operation during 1979
some 250 glass blocks had been produced and stored in
air-cooled facilities. Other countries are pursuingresearch into the most appropriate form for solidifying
high-level wastes, for containing the wastes, and to findthe most suitable location for disposal. Assuming aworld nuclear generating capacity of 1300 GW(e) by
the year 2000. it might be expected that some 5000
GW(e) a will have been generated and assuming thewaste is vitrified in a cylindrical form similar to thosedeveloped in the French AVM process several tens of
thousands of cylinders will be required.
195. The procedures for disposal being studied widelyare either disposal in deep geological formations, ordisposal on or under the ocean bed. International
studies are being undertaken on research needs toassess ocean disposal and many countries are actively
investigating geologic disposal in either salt or hardrock (granite, gneiss or basalt). In addition, there is thepossibility of using argillaceous rock as a repository.
The main barriers which can prevent the return ofactivity to man's environment from a geologic repository or can influence the rate at which activity returns
are: the waste form and its container; geologic
containment of the waste within the rock formation;
retardation of activity during transport throughgeologic media; and dispersion and dilution of activity
in the biosphere.
196. Assessments of geologic disposal have been
undertaken in many countries [E6, K8, K 11]. In the
United Kingdom [H11, H12], assessments estimate not
only public health consequences but are performed
with the intention of using the theoretical models toperform sensitivity analysis so that those parameters
whose uncertainties have a significant effect on theoverall result can be identified for further research. In
the United States a number of studies have been undertaken on a range of geologic media and the present
emphasis is also on the identification of areas of uncertainty. so that research needs and priorities can beproperly co-ordinated in the development of the overall
strategy for high-level waste disposal [E6]. In Sweden acomprehensive study of the radiological consequencesof geological disposal of unreprocessed nuclear fueland vitrified high-level waste has been undertaken [K8].
197. Disposal of high-level wastes from nuclear power
production has not yet taken place. There has only beenstorage under surveillance by national authorities,
awaiting a final decision on processing. Therefore theCommittee has to make its estimate of the potential
collective dose commitments from high-level waste
disposal on tht basis of theoretical studies. For normal
operations in the preceding steps in the nuclear power
cycle, truncated dose commitments per unit of electricalenergy generated have been used as a tool for theassessment of the maximum dose rate in the future, if
the practice were continued at constant rate. The
integration period was taken to be equal to the expected
length of the practice. For high-level waste contained
over periods which are longer than the assumed periodof practice. such truncated dose commitments are notneeded, since the dose rate at the time when theradioactive material may reach the biosphere can becalculated directly and there would be no continued
practice that would add to it in later years. The
potential collective dose commitments given below

have been based on the INFCE analysis of waste
disposal (19]. as this was an exercise involving many
countries active in the development of nuclear power.
No assessment is made by the Committee of dose
commitments. but rather the INFCE values are quoted.
The INFCE analysis of geologic repositories assumes
that the lifetime of stainless steel containers is 103 a andthe retention time of wastes in a vitreous matrix is 104 a,
although these figures have little influence on theoverall result as the time scale of the subsequentgeologic transfer is much greater. For a salt repository
the waste is only exposed to groundwater due to a
disruptive event which is assumed to have a probabilityof occurrence approaching unity in JOLI06 a. A
detailed analysis of this scenario showed that the mean
arrival times of the nuclides of significance, 99Tc, 129),
135Cs and the actinides, in fresh water was of the orderof several million years. In the repository in nominal
hard rock, the canisters would be in contact with water
which has a low now velocity. Transport of dissolved
species through the buffer material surrounding thecanisters will be governed by diffusion which isextremely slow. Further retardation of many nuclides is
expected by chemical interactions with the buffermaterial and the host rock. The INFCE study assumes
that because of these processes it will be between 105
and 106 a before radionuclides from the waste arrive infresh surface waters and the radiological consequences
are calculated following the different pathways tohumans [I I OJ.
198. Dose commitments are first evaluated assuming
the radionuclides are released to fresh water for whichsimple assumptions are made about turnover time
(10 a), fractions of water drunk (2.7 10--5 of the available volume per year) and fractions of fish content
consumed (3 10--s a-1). These figures are global
averages. It is assumed that all the long-lived nuclides
released from a repository eventually reach the ocean
where sedimentation processes become an important
removal mechanism for radionuclides from thebiosphere. In the consideration of population exposure,in addition to consumption of fish and other sea foods.
resuspension of sediments provides inhalation doses
and external exposure to sediments occurs along theshoreline.
199. The dose commitments due to waste disposal
depend upon whether unreprocessed fuel or high-level
reprocessing wastes are considered and from which fuelcycle they arose. The doses would not be received until
after 105 or 106 a, so they will be subject to considerable
uncertainty and the absolute numerical values should
be used with caution. Their main virtue is in providing
the ability to compare the cycles. Table 51 shows thecollective effective dose equivalent commitments forthe major fuel cycles under consideration for the future
where the contribution from the uranium isotopes andtheir radioactive daughters are separately identified
from other elements in the waste. Results are presented
assuming both 105 and 106 a migration times throughthe geosphere to show the range to be expected for dose
commitments. The once-through fuel cycles give higher
values of collective dose commitment and the fast
breeder reactor gives lower values because of its more
efficient use of uranium. Assuming LWR fuel is reprocessed, then whether the plutonium is recycled through
LWRs or in FBRs, the normalized collective effective
dose equivalent commitment is between 20 and 50 man
Sv [GW(e) aJ-1, regardless of the actual time chosen for
migration through geologic media.
277

200. As regards the levels of individual dose
associated with a geologic repository, studies on a salt
medium [US] and on hard rock [I I O. K8] have been
undertaken, assuming wastes from 100 GW(e) a ofelectrical energy production are disposed of. For thesalt repository the annual whole-body dose equivalent
was between 5 J0-8 and 3 10-s Sv, some 3.5 ]06 a after
the event initiating ground water ingress. For hard rock.
maximum individual annual dose equivalents werepredicted some 4 l os a after disposal and ranged from
10-S to 10"6 Sv. These doses should be regarded asindicative rather than definitive and the conclusion to
be drawn is that either disposal option for any fuelcycle will probably lead to very low levels of individual
annual dose in the far distant future.

VII.

MISCELLANEOUS CONTRIBUTIONS

201. Other aspects of the nuclear fuel cycle which
contribute to the collective dose commitment from theuse of nuclear power are transportation of irradiatednuclear fuel and the operation of nuclear researchfacilities.
A.

TRANSPORTATION

202. Unirradiated nuclear fuel is transported toreactor sites from fuel fabrication facilities andirradiated nuclear fuel is transported from the reactor
sites to reprocessing or fuel storage facilities. The
transport of radioactive materials is subject to national
regulations generally based on the IAEA regulations[17]. Shipments occur by road, rail and sea, and thenumber of shipments and distances travelled vary
widely from country to country. Typical studies havebeen carried out in the United Kingdom [818] and in
the United States [U4] and for the Commission of theEuropean Communities [CS].
203. Calculations can be performed knowing the dose
rate distribution as a function of distance from theirradiated fuel flask and the population density along
the routes used for transportation and up to a distanceof perhaps I km away from the pathway. Estimates of
the normalized collective effective dose equivalentcommitment associated with irradiated fuel transport isin the range 10-3 to 10-2 man Sv [GW(e) a)-I. This does
not include any contribution from nask accidents
which could lead to radiation exposures of the public.
It is difficult to add any meaningful component to total
health detriment from accidents since there is a probability distribution of a whole range of accidents leading
to different releases, for each of which there are probability distribution functions representing the chance ofdifferent meteorological conditions (weather category
and wind direction), and widely different population
distributions depending upon the location of theincident and meteorology.
B.

NUCLEAR RESEARCH INSTALLATIONS

204. A proportion of the radioactive materials
released to the environment from nuclear research facilities may be attributed to support for continued
operation or future development of nuclear power.
However, other activities at nuclear facilities such asradioisotope production and processing and other typesof research are usually responsible for a large fractionof the environmental releases. Since the fraction of the278

releases to the environment which may be attributed to
the generation of electricity by nuclear power is
unknown, it is not possible for the Committee to assess
their contributions to overall dose.
205. Reported estimates of discharges have been used
to calculate the associated collective doses in 1977-1978
from 27 nuclear research and operational institutions in
the United States [E 1]. The annual collective dose
equivalents ranged from 10-10 man Sv up to 2 man Sv
and the total value was about 6 man Sv from all installations. over half of this being due to 3H releases fromtwo sites (Savannah River and Ames Laboratory). Onethird of the total was due to 41Ar discharges from the
Argonne National Laboratory. The collective doses
from nuclear research facilities in the United Kingdom
have been estimated [T4] based on reported discharge
data. The only significant contribution is from thetritium discharged to the river Thames by the United
Kingdom Atomic Energy Authorities' Research
Establishment at Harwell, for which the annual
collective dose equivalent via the drinking water
pathway to the population of Greater London isestimated to be 1 man Sv. Data on radioactive materials
released from the nuclear research centres Jiilich andKarlsruhe and from research reactor sites in the Federal
Republic of Germany have been reported and thecollective doses associated with nuclear power
production are negligible.

VIII. SUMMARY OF NORMALIZED COLLECTIVE EFFECTIVE DOSE EQUIVALENT COMMITMENTS TO THE
PUBLIC FROM NUCLEAR POWER PRODUCTION
206. The preceding chapters contain the results of theCommittee's assessments of collective dose commitments and collective effective dose equivalent commitments. excluding the contributions from occupational
exposure. associated with nuclear power production.
The collective effective dose equivalent commitments,
which have been averaged over reactor types according
to their total contribution to electrical power
production, are summarized in normalized form, i.e.,
per unit of electrical energy generated. in Table 52.
Normalized collective effective dose equivalentcommitments are presented. so that different steps inthe cycle may be compared and appropriate estimatesof the health detriment may be estimated using theprocedures outlined in Annex A. For releases during
the operational phase in the nuclear fuel cycle, i.e.,
excluding waste disposal, the normalized collective
effective dose equivalent commitment is 5.7 man Sv[GW(e) a]-1, of which some 90% is delivered in the year
the wastes are discharged and some 98% within 5 years.
This estimate is made using the notional reprocessing
plant efnuents. Present day reprocessing may givenormalized collective effective dose equivalent commitments 10-20 times higher; however, since less than 10%
of fuel is reprocessed, the weighted contribution would
be similar to that here.
207. For those long-lived nuclides which become
globally dispersed incomplete normalized collective
effective dose equivalent commitments are also given in
Table 52. The incomplete dose commitments give an
indication of the time distribution of the dose
commitment and can also be used to derive themaximum annual per caput effective dose equivalents

by dividing by the mean population size, assuming the
production of electricity by nuclear power continues
throughout the period chosen for the integration time,
and that the release rates remain constant. The
normalized collective effective dose equivalentcommitment due to global dispersion of nuclides
released during fuel cycle operations is 670 man Sv[GW(e) a)- 1, 90% of which is delivered in the time
period 104-1 QR a after release. The incomplete value to500 a, which might be taken as the duration of thepractice of generation of electricity by fission power is
18 man Sv [GW(e) aJ-'. If this figure is used to derive anannual per caput effective dose equivalent, it iscomparable with that obtained from the local andregional normalized collective effective dose equivalentcommitment.
208. Considerable caution must be exercised in
extrapolation of doses into the future because the trend
over the last five years has been for reactor releases tothe environment to decrease despite the increasing
electrical output of plants. This has been due partly to
improvements in control technology and partly to thedevelopment of new concepts in radiation protection.
Also, the collective doses due to global dispersion ofradionuclides may not be typical of future practices;for example, in the model reactor and reprocessing
facilities assumed here for LWR assessments, it was
taken that long-lived nuclides such as 85Kr and 14C
were emitted untreated to the atmosphere, whereas they
may be contained and immobilized. Finally, when
extrapolating thousands or millions of years into the
future it is difficult to predict the population size, its
dietary and other habits as well as medical services sothat the estimation of global collective dose commitments must be regarded as highly speculative.
209. The results in Table 52 indicate that the largest
contributions to the normalized collective effective dose
equivalent commitments arise from mining and milling
tailings. The incomplete dose commitments are proportionally less and depend entirely on the assumptions
made about the amount and type of covering on
tailings and the time before the activity either migrates
downwards into the ground and becomes unavailable,
or is eroded into the aquatic environment. If the tailings
are immobilized by use of polyvinylchloride or asphalt
coverings, the radon emanation rate can be reduced toambient levels and the contribution to dose
commitment is essentially zero; similarly downwards
migration of uranium and thorium with a rate of 2 10-3
m a-1 would lead to a normalized collective effective
dose equivalent commitment of the order of 30 man Sv[GW(e) a]-1. Assuming that the tailings are eroded into
the aquatic environment. leads to a normalized
collective effective dose equivalent commitment of 460
man Sv [GW(e) a)-I, of which 76% is contributed by
2l0Po. The results will be highly dependent upon thelocal site characteristics and the answer given here must
be uncertain by at least an order of magnitude. Finally,
it must be recognized that any decision to use fast-

breeder reactors, by more efficient utilization of
uranium resources, could reduce the normalized
collective effective dose equivalent commitment fromtailings by two orders of magnitude.
210. In order to estimate the maximum per caput
annual effective dose equivalent in the future as a resultof nuclear power production, an incomplete collective
effective dose equivalent commitment truncated overthe expected duration of the practice of generation ofelectricity by fission power, taken here as 500 a. must be
used. The releases during the operational stage of thenuclear fuel cycle lead to a local and regional
normalized collective effective dose equivalentcommitment of 5_7 man Sv [GW(e) aJ-1 of which 98% isreceived in the first few years after discharge. For those
nuclides which become globally dispersed, the incomplete normalized collective effective dose equivalentcommitment to 500 a is about 18 man Sv [GW(e) a]-1.
The choice of 500 a as a mean duration of the practiceof producing power by nuclear fission implies the useof breeder reactors and the rate of mining would
decrease. The normalized incomplete collective
effective dose equivalent commitment from mining andmilling, based on the present fuel cycle. is therefore to
100 a and is likely to be due mainly to radon releases
giving 2.5 man Sv [GW(e) a]-1. Thus. on the pessimistic
assumption that no technological improvements are
made and current levels of discharge continue for 500 a,
the maximum annual collective effective dose equivalent would be about 25 man Sv [GW(e) a)-1. The
annual collective and per caput effective dose equivalents for a notional nuclear programme to the year 2500
are shown in Table 53, again assuming that present
release levels are not reduced and that the annual~eneration of electric energy reaches 1 kW a per caput,
1.e., some 10 4 GW(e) a in 2500. It can be seen that evenwith the maximizing assumptions made here. the levelof annual per caput effective dose equivalent rises tothe equivalent of 1% of natural background radiation.
The annual per caput effective dose equivalent would
reduce after the end of the practice, to about 1% of the
final values after 100 a.
211. Normalized collective effective dose equivalentcommitments from mill tailings, waste disposal and theglobal dispersion of the long-lived nuclides 14C and 1191
must be uncertain because of difficulties in predicting
future practices, population sizes and habits, environmental pathways and human metabolism. At present
the estimate is of the order of 4 103 man Sv [GW(e) aJ-1,
mainly delivered in the time period starting 105 years
from now. Such figures should not be relied upon for
decision-making purposes and the Committee therefore
recommends that little significance be attached to thesecompletecollective
effectivedose
equivalentcommitment estimates. The Committee however recommends that more research be undertaken in order tobetter quantify the incomplete collective effective dose
equivalent commitments.

279

T a b 1 e

Attainableroduction ea abilit of uraniuM in 1979an es r.iate reserves'lilCountry
or area3:

Production
capability
in 1979(t)

Estimated
reserves
a/

135
600
103
6900
2950
1000
200
3700
3350
s2qo
339
0
19000

37000
3540001640001,029000
10000035000
13000
17500021000052500019000'304000
1,831000Argentina
Austra 1ia
BrazilCanadaFranceGabon
Indiailamibi a
lli gerSouth Africa
Spain
SwedenUnited States{t)a/ Reasonably assured reserves, assuming a
price of USS 130 per kilogram of uranium.

T a b l e2

Estimated radioactive airborne effluentsfrom the model m1ne and operations at the model millRadionuclide

llormalized releasefrom mine[GBq (GW[e] a,- 1}

Normalized releasefrom mi 11 i ng
[GBq (Gll[e] a)- 1]

Normalized annual
release fromtailings area
'G6q (GU[e] a)" 1J

0.66
0.66
0.074
0.04

0.00070.0007
0.015
0.015
0.015
0.015
1000

~~u

230Th
226Ra
210Pb
210Po
222Rn

0.04

0.04
880

20000

Table

3

Po ulation and atmos heric dis ersion arameters
at t e mo e mine an m1
site!
Pasqui 11 ea tegory

Frequency ('Jo)
Height of atmosphericmixing depth (m)Wind speed (rn s" 1)

A

ll

C

3.8

5.1

6.5 25.5 22.6

02000 2000 1000 1000
2

5

5

E

200
3

F

-+

rain

36.2 0.3
200 10002
5

a/ The effective height of release is 10 m. The population density
is 3 km" 2 for 0-100 km and 25 kr.1· 2 for 100-200 kr.1.

280

T a b l e

4

Estimated normalized collective absorbed dose co11'111it.mentsfrom earticulale a1r6orne re1eases rrom uranium milling

florma 1 i z1ed co 11 ecti ve absorbed dose commitment '.10-S man Gy (Gl>'[e] af 1 J
GonadsBreastCloud passage:
Inhalation
23Bu

Ont!
liningcells

Red bonemarrow

Lungs

Thyroid

Kidneys

Liver

210

0.2

2

0.9

0.006

234u

2SO

0.2

3

1

0.006

230Th

26

15

200

0.04

0.3

226Ra

0.9

0.03

0.4

0.006

0.006

210Pb

0.5

0.2

2

0.4

0.8

210p 0

1.2

0.02

Q.02

0.6

0.2

Total (rounded) (a)Activity deposited:
Interna 1 irradiation
238u

490

16

210

Remainder
tissues

3

2

2

2

5

29

2

17

2

234u

2

2

2

3

26

2

16

2

2

230Th

0.01

0.01

0.7

0.8

11

0.01

0.4

0.01

0.01

2

226Ra

0 .1

0.1

0.1

0.4

4

0.1

0.1

0.1

0.1

2!0Pb-210Po

4

4

2

4

26

4

4

4

4

Tota 1 (rounded) (a)

8

8

7

13

96

9

37

8

8

Activity deposited:
External irradiation (y)23 to all organs and tissues

Table

5

Normalized collective effective dose e uivalent commitmentsfrom particulate and gaseous a,r crne re eases fror.i mining an m, llingCollective effective dose equivalentcommitn1ent (10- 2 man Sv [G\/(e) aj-i)lnha l ati on
during thepassage ofthe cloudMilling
Particulates
Gaseous radon-222
and daughters

2.3

MiningGaseous radon-222
and daughterssoTota 11.3

lnterna lExterna 1
irradiation i rradi a ti on
due to the due to theactivity
activity
depositeddeposited

1.5

0.2
]
0.1

Total

0.02
2.4

50

54

281

T a b 1 e

6Norma 1i zed(a)Atmospheric releases of radon and airborne
particulates as a function of the time forwhich the erosion continues
llorma li zed col lecli ve effective
dose equivalent commitment
1)(man Sv [Gll(e)

ar

2JBu
2J4u
230Th
226Ra
210Pb
210p 00.000005 0.005
0. 000006 0.006
0. 0004
0.44
0.000009 0.009
0.00002
0.02
0.00002
0.02
0.025
25

0.5
0.6
44
0.9
2
2
2500

0.026

2600

.

222Rn
Total (rounded)
(b)

10 5

103

Release duration (a)26

Re leases of tailings into the marine environr.ient
Normalized collective effective
dose equivalent commitment(man Sv [GW(e) a]- 1)
238U
234Th
234u
230Th
226
Ra
210Pb
21 0Po

0.15
8.8
o.17
0.54
18.6
79.7
350

Total

460

Table

7

Norma 1i zed effluent di scha 1es from the model fue 1 conversion,
enr1chment and abrication fac1l1ties(MDq [GW(e) a]- 1)[ E1 , B2, 822, Pl , 56]

238u
235u
234u
234Th
230Th
226Ra
222Rn

282

Aquatic

Atmos?hericlsoto?eConversion

Enrichment

Fabricaticn

74
2.0
74
74
0.74
0.07
8140

3.7
0.07
3.7
3,7

0.74
0.22
7.4
0. 74

versionCon-richment

En-

Fabrication

814
20
814

370
7.4
370

370

56
126

7.4

370
370

Table8

Meteoro lo i ea l
aroun
used
Pasquill weather category
Quantity

Dr yFrequency (%) _~ind speed (ms 1 )Depth of mixing
layer {m)Ra i n

A

B

C

D

E

F

C

D

0.3

4.5
2

12. I

63
5

5

5

4.6
2

s

1. 7

8.4

10DO

1000

200

I

2000

2000

3

200

1000

5

!OOO

a/ The assumed stack height is 60 m.
The assumed population distribution is uniform at 25 km- 2 .

Table

9

Normalized collective absorbed dose co11J11itments
by inhalation for a1rborne r~leases from fuel fabr1cation
Nonnalized collective absorbed dose commitment fl0- 6 man Gy (GH~e] a)- 1 ]onebone
Thyroid Kidneys Liver Remainder
GonadsBreast Lungs Red
marrow lining
tissuescells
Cloud passage:
Inhalation
2380
234u
230Th
226Ra

250
320

0.2
0.2

3

I

3

2

20
0.007

0.004

Tota 1 (rounded) (a)

570

2

30

2

35
34

0.02

Activity deposited:
Internal irradiation
238 0
234u
230Th
226Ra

2
2
0.001
0.002

2
2
0.001
0.002

2

6

2
0.07
0.002

4
0.08
0.008

Tota 1 (rounded) (a)

4

4

4

10

Activity deposited:
External irradiation (y)

3

0.08

2
2
0.001
0.002

20
20
4
0.002

2
2
0.001
0.002

2
2
0.001
0.002

70

4

40

4

4

1

0.4 to all organs and tissues283

Ta b l e

10

Normalized collective effective dose
equivalent commitment[10- 4 n\:ln Sv (GW[e) a)-I)

Radionuclide

Cloudpassage:
Inhalation

Internal
i rradi ati on

238u
234u
230Th
226Ra
222Rn

6
7
0.2
0.0004
4

0.7
0.9
0.01
0.001
0.02 ,

Total {rounded)20

2

Activity deposited

Ta b l e

External

i rradi at ion

I

0.004

0.004

11

World nuclear eneratin ea acit in 1979
rNet values in We an nurn ers o un1 ts 1n parentheses][11, K12]

Reactor typeCountry
or area

PWR

BWR

Total
capa·
city

kW(e)
per caput

0.34

0.015

1.66

0.17

0.82

0.10

5.49

0.25

0.38

0.024

1.08

0.23

8.16

0.15

1.29

0.074

8.61

0.14

0.60

0.001

0.54

0.01

14.47

0.16

0.56

0.17

0.50

0.038

0.13

0.002

1.07

0.031

3.70

0.46

1.94

0.31

10.62

0.043

6.98

0.13

49.99

0.24

1.21

0.084

10.01 6.43 7.89 0.65 120.14(44) ( 17) ( 16) (6)

0.07

GCR

Argentina
Belgium
Bulgaria
CanadaCzechoslovakia
FinlandFranceGerman Dern. Rep.
Germany, Fed.Rep.of
India
Italy
Japan
Republic of Korea
Netherlands
Pakistan
Spain
Sweden

USSR

0.38
{1)
0.42
{1)
5.71
(7}
1.29
(4)
5.37
(6)
0.24
( 1)
6.44
(9)
0.56
{1}
0.45
{1)

0.66
( 1}

3.16
{5}
0.40
(2)
0.15
(1)
7. 72
(12)

0.44

{I)

(1)

0.80

2.90
(5)
0.32

1.62
(3)
2.49

0.01 0.05
( 1)
( 1)
0.20
( 1)
0.15
( 1)
0.16 0.15
( 1)
( 1)

0.25
( 1)
0.02
{1)

o.4e
( 1)

0.13
{1)

(I)

0.09
(5}

Other Asia

31.88 17.78
(42}
(26)
1.21
{2}

Total

61.49
(92)

2.13 0.07
{7}
( 1}

0.05
{1)

0.15

United Kingdom

284

FBR

5.49
{11)

(7)

United States

LWGR

0.34
( 1)

1.66
(3)
0.82
(2)

( 1)

Switzerland

HWR

33.67
(60}

6. 75
(32)
0.33
{1}

7.89 0.15( 16) (2)
0. 23
{1)

Ta b 1e12

ases dischar ed in airborne effluents from PWRs
1n various countries[B23, 06, OS. rg, G:l, I<~. [l, Li>, fM, P2, S14]

Noble

Country
andreactorStartupyear

Electrical
capacity[Gli(e)]Belgium
IJOell,2
Tihange

1974/75
1975

0.79
0.87

Finland
Loviisa

1977

0.42

FranceFessenheim
Chooz
Bugey 2,3

1977
1967
1978

1.78
0.30
J.85

1968
1972
1974/76
1976
1978

~
rno

Release (TBq)
1976

1977

1978

1979

7. 7 30
170
17

28
55

17
58

44
14

2.2

1.4

1.6

1975

100

183

70
103

73
120
4.1

0.33
0.63
2.39
0.81
1.23

296
47
62

12
389
54
19

14
132
156
70

17
18
79
4.3
1.9

1964

0.25

17

6.6

2.2

20

, a ia 1~
Mihama 2
Mihama 3
Takahama l
Takahama 2
Ohi 1
Ohi 2
Ikata 1Genkai 1

1970
1972
1976
1974
1975
1977
1978
1977
1975

0.34
0.50
0.83
0.83
0.83
!.17
i.17
0.57
0.57

2.7
7.0
4.8
3.0

2.1
30
10
3.6
3.3

2.7
14
6.3
4.4
2.3
0.4

1.5

1.8

2.2
2.0

1.7 2727
6.7
0.59
0.9
0.26
3.1
3.0
2. l
2.4
9.3
3.0
0.33 2.3
3.0
3.4
1.6
1.0

Netherlands
Borssele

1973

0.45

97

37

15

8

Swedenlmiglials 2

1975

0.80

11

5.4

22

1974
1979
1976
1974
1975
1977
1977
1978
1973
1969
1968
1973/76
1974
1972
1975
1978
1973/741971
1970/72
1973/74
1974
1970
1976
1976
1967
1972 /73
1974
1975

0.84
0.95
0.80
1.62
2.14
0.80
0.92
0.83
0.46
0.47
0.55
1. 77
0.52
0.77
0.80
0.90
2.64
0.64
0.99
1.01
0.87
0.66
0.78
l.08
0.44
0.55
0.79
l.08

Germani, Fed. Ree. of
Obrigheim
Stade
Biblis A,B
Neckarwestheim
Unterweser

United States
Arkansas 1
Arkansas 2
Beaver Valley
Calvert Cliffs 1,2
Cook 1,2
Crystal River
Davis Besse
FarleyFort Calhoun
R.E. Ginna
Haddam Neck
Indian Point
Kewaunee
Maine Yankee
Millstone Point 2
North Anna
OconeePalisades
Point Beach l , 2
Prairie Island
Rancho Seco
H.B. Robinson
St. Lucie
Salem 1San Onofre
Surry I ,2Three Mile Island 1
Trojan

144
8.5

38

210

514

277

3.2

286
36

2
825
141
112
47

14
1020
1795
254
77

3.8
8.5
43
12
38

315
168
65
378
403
2802
62
118
26
28
204
343
5.6
73
13
232
1772
2.5

131
16
80
141
50
389
204
118
36
18
115
79
18
392
303
592
551
91
59
90
16
48
152
11
53
57
84
28
559
562
1628 1317 1612
97
1.1
2.2 12
1676
74
42
19
36
64
81
25
47
26
4.4
4.7
74
263
326
43
29
18
33
56
940 1084570
7.3
0.38 9.2
5.6 67
66
15
24
298
710
703
161
66
134
102
614
581
83
113
11
34

285

Table 12, continued
Country
andreactorStartupyear

Turkey Point.
Yankee RoweZion 1,2

1972/731960
1973/74

Electri ea 1
capacity
1975[GW(e)]

1976

1977

1978

1979

496

577

862

I

l

1676

5254

1191

870
24
2505

392
6.7
1262

16.9
402

19.2
570

23.B
388

29.4
432

27.5
370

I. 32
0.17
2.08

Total annual energy
generated [GW(e) a],lormalized release [TBq (GW[e] af 1]

Re 1ease (Tllq)

430 TBq [GW(e) a]-lAverage 1975-1979

Ta b l e

4 ,613

lsotoeic comeosition of noble 9ases discharged from PWRs
,n tne On1teo States our1n9 1979
[B2j]

ReactorArkansas 1
Arkansas 2
Beaver Valley
Calvert Cliffs 1,2
Cook 1,2
Crystal River
Davis Besse
Joseph M. Farley
Fort Calhoun
R.E. Ginna
Haddam NeckIn di an Point
Kewaunee
Maine Yankee
Millstone Point 2
North Anna
Oconee 1,2,3
Palisades
Point Beach 1,2
Prairie Island 1,2
Rancho Seco
H.B. Robinson
Salem
St. Lucie
San Onofre
Surry 1 ,2Three Mile Island 1
TrojanTurkey Point
Yankee Rowe
Zion

EnergyStartup generated
year[GW(e) a]
1974
1979
1976
1975/77
1975/78
1977
1977
1978
1973
1969
1967
1973
1974
1972
1975
1978
1973/741971
1970/72
1973/74
1974
1970
1976
1976
1967
1972/73
1974
1975
1972/73
1960
1973

o. 397

0.916
0.221
1.161
1. 373
0.453
0.381
0.211
0.440
0.355
0.493
1.142
0.412
0.537
0.520
0.507
1.708
0.415
0.808
0.865
0.687
0.482
0.250
0. 592
0.401
0.343
0.266
0.631
0.811
0.149
1.238

Total annual energy
generated [GW(e) a]
19.165
Normalized release [TBq (GW[e] a)- 1]

286

Release (Tllq)
41Ar 85mKr 85Kr
0.30
5.3
0.005
0.019
17
0.02
11
0.078
0.033
0. 31
0.068
0.007
1.0
2.1
0.1
0.06
0.05
0.011
0.92
0.40
0.083
0.67
0.016
2.0
0.062
0.28

2 .1

87Kr

'88

Kr

131mxe 133mxe 133Xe 135mxe 135Xe

0.11 2.8
0.001
0.52
0
18
1.6
0.036
0.045 0.048 0.14
1.3
0.007 326
0.88
25
7.8
19
36
49
0.61 0.43 0.019 0.023
0.26 560.001 0.78 0.062 0.007 3.5
1.4
0.01 0.48 0.004 0.016
0.19
0.036 0.018
0.11 0. 74
0.051 37
0.13 0.12
1.5
0.29
0.29
3.8
0.92 241.0
3.6
0.016 1.0
0.013
0.007
2.4
1.0
0. 36
0.67 0.05 0.01 0.02 0.12 0.58
40
0.26
1.1
14
27
3.4
0.090
0.021 0.002
2.3
1.2
1.2
3.6
0.11
0.068
0.03 0.048 0.09
0.52 0.27 0.007 0.041 0.085 0.92
3.2

0.043
4.9
0.08
0.029

0.20

0.49

1. 7

0.042
0.050
0.060
0.062

0.050
2.9
2.3
0.013
0.007
0.038
0.035
0.060
0.097

0.04
5.0
10
0.019 0.084
0.041 0.02
0.007
0.034 0.058
0.0099 0.036
0.050 0.093
1.5

0.12
4.9
0.073
0.36
0.093 0.26
0.092 0.28
0.075 0.25
0.012 0.069
8.8
0.63

2.2

25

0.84

6.3

2.0

31
0.10

5.0

289
24
143
19
45
0.027
363
13
70
7.8
2290 21212
3.3 0.51 0.24
95
5.1
25
0.24
1.9
25
155 0.027 1.1
297 0.38 11
1.1 0.41
67
1.4
648
0.64
224 0.02 7.3
1646 1.429
2.4 0.007
11. 3 0. 78 12
25
0.71
300
24
56
0.87
8.4 0.56
492 0.41 9.4
0.53
19
63
2.5
0.95
81
32 0.13 1.4
389 0.068 1.1
4.3 0.82 0.94
1132
118

469

1.3

26

138xe

125
1.2
0.01
0.13
0.84

0.01
0.15
1.2

4.7

0.016
0.026
0.028

6.9

Ta b 1e14

Noble gases discharged in airborne effluents from BWRs
in various countries 1g?5-1§7g[B23, D6, DS, t9, G~. RZ, [1, [6, N19, P2, S14 J
Country
andreactorStartupyear

Electrical
capacity[GW(e)]

Release (TBq)
1975

1976

1977

1978

1979

0.25

0.014

FinlandllTITTuoto

1978

0,660

German.l:'., Fed. Ree.of
Gundremmingen
LingenWiirgassen
llrunsbiitte l
Isar

1966
1968
1971
1976
1977

0.237
0.182
0.640
0.?70
0.870

274
1295
4.5

195
237
18
36

9.2
8.5
4.9
29
121
116
280
0.9 40

1964
1978

0.152
0.548

8456'

8861

3328

2886

Caorso~u gaTokai 2
Fukushima l
Fukushima 2
Fukushima 3
Fukushima 4
Fukushima 5
Hamaoka 1
Hamaoka 2
Shimane

1970
1978
1971
1974
1976
1978
1978
1976
1978
1974

0. 3571.100
0.460]
0. 784o. 7841
0,784
0.784
0.540
0.840
0.460

85

70

23

16

8.1

196

1070

17

740

133

0.63

0.25

1.1

0.77 0.67
0.024 0.0025

Netherlands
Dodewaard

1968

0. 05278

SwedenU's'l<arshamn 1
Oskarshamn 2
Ringhals 1
Ba rseback 1
Barseback 2

1970
1974
1974
1975
1976

0.460
0.590
0. 780
0.590
0.590

~
ar gliano

United States
819 Rock Point
Browns Ferry
Brunswick
CooperDresden l
Dresden 2.3
Duane Arnold
J.A. Fitzpatrick
Edwin I. Hatch
Humboldt Bay
Lacrosse
Mi 11 stone PointMonticelloNine Mile Point
Oyster Creek
Peach Bot tom
Pilgrim
Quad Cities
Vermont Yankee

1963
1973/77
1975 /76
1974
1960
1971 /72
1975
1975
1975
1963
1969
1971
1971
1969
1969
1974
1972
1973
1972

- 0.0026

159
46
85

230

481

159

119

5402
1.4
1712
0.78

5217
31
78
3
0. 37

4613 4700
442902900 19300
2.9 0.97
2.0
3.3

496
699
323
0.075
1872
562
3.195
932 2974 6142 5809 10027
6.8
1.580
685 9102 3381 3184
1420
47
0. 764729
151 11250.200 19240 17020 19240 31450
6.8
1.545 13653 1199 11581
1502 2557
0.515
57
195
143
342
58
862
0.300
151
1709
218
125
0.717
57
115
70
60
142
0. J6510952 3441
0.)48
2109 4588 1573
312
385
762
0.652 109890 18759 22940 20942
254
0.536
5735
422
238
149
0.510 48100 6512
131
38
112
7622 6142 6549 36926 37470
0.520
2.)86
480 7730 2631
1425 4592
0.564
3885 6771 15281
1210
514
1.538
947
4070 1166
1199 1644
0.514
124
131
183
299
106

Total annual energy
generated (GW(e) a]
Normalized release (TBq (GWfe] a)- 1]

9.53
25447

Average 1975-1979

13.04
7104

13.24
8103

18.1 20.06500 4407

8800 TBq (GW(e) a]-l287

Ta b l e

15

Isotoeic comeosition of noble gases discharged from BWRs
1n tne Onitea States aur1n9 197[B23)

Reactor

Big Rock Point 1
Browns Ferry 1,2 ,3Brunswick 1,2
CooperDresden 1
Dresden 2 ,3Duane Amold
J .A. Fitzpatrick
Edwin I Hatch
Lacrosse
Millstone Point lMonticelloNine Mile Point
Oyster Creek
Peach Bottom 2.3
Pilgrim
Quad Cities 1,2
Vermont Yankee

EnergyStartup generated
year [GW(e) a]
1962
1973/77
1975/771974
1959
1970/71
1974
1974
1974
1967
1970
1970
1969
1969
1973/74
1972
1971 /72
1972

0.013
2.393
0.810
0.591
1.013
0.352
0. 349
0.401
0.024
0.505
0.522
0.354
0.541
1. 740
0.574
1.075
0.414

Total annual energy
generated [GW(e) a]
11.670Normalized release [TBq (GW[e] a)-l]

Reactor

Big Rock Point 1
Browns Ferry 1,2,3
Brunswick 1,2
CooperDresden 1
Dresden 2 ,3Duane Arnold
J .A. Fitzpatrick
Edwin I Hatch
Lacrosse
Millstone PointMonticelloNine Mile Point
Oyster Creek
Peach Bottom 2,3
Pilgrim
Quad Cities 1 ,2Vermont Yankee

EnergyStartup generated
year [GW(e) a]
1962
1973/77
1975/771974
1959
1970/i 1
1974
1974
1974
1967
1970
1970
1969
1969
1973/741972
1971/72
1972

0.013
2.393
0.810
0.591
1.013
0.352
0. 349
0.401
0.024
0.505
0.522
o. 3540.541
1. 740
0.574
1.075
0.414

Total annual energy
generated (GW(e) a]
11.670Normalized release (TBq {GW[e] a- 1]

288

Release (TBq)
41Ar
44
41

83mKr

85mKr

85Kr

87Kr

88Kr

89 Kr

90Kr

9.9

12
50
88
74

0.037
8140

31
73
316
100

17
101
281
188

24

26

32
4.7
5.8
6.8
14,
0. 31

13
0.13

69
3.2
2.4
0.37
41
8.4
1.9

136
21
4 .1
2.8
25
2.2
1.0
16
5032
3.1
204
190
3.2

530

5.2

2.3

25

3.9

54
3.8
1.5
2.0
0. 37

7.9

7.6

0.36
0.74

1613
16
96
121
1. 3

0.002

5217
4.1
83
488
3.8

180

690

550

0.06
0.49
36

1.2

1.1

Release (TBq)
131mxe133mxeI33xe

135mxe

135Xe

137xe

138xe

139Xe

0.11

0.32

20
38
1217
358

64
610
240
45

35

29
11

40
121
888
9.4

38

1. 7

5.8
319
84
310
6.8
219
212
70
110
27
12
10
1298
3922
71
481
164

250
9.4
0.6
0.74
33
119
2.8
6.0
3648
3.2
7.6
33
36

466
42
29
2.1
120
74
2.1
5.1
8362
566
19
97
8.1

630

450

980

4.6

1.2
1.2
3.0

14
0.35
81
0.07

0.02

5.0

72

9.1

10

0.23

2.9
269
47
23

33

1365
5.1
3.3
2.4
120
182
41
12
12284
0.97
32
235
81

3.7

1300

3.3

Ta b l e

16

Noble sases discharsed in airborne effluentsfrom R\;/Rs in various countries 1975-1979[ B21, fi!O]

Country
andreactorStartupyear

Argentina
Atucha1974
CanadaBruce A
1976/79
Pickering A 1971/73

Electrical
capacity[GW(e)]

Release (TBq)
1975

1976

1977

1978

1979

0.320

9.3

160

71

311

252

4 X 0.755
4 X 0.514

162

104

1254
159

1550
152

3084
218

2.2
86

2.6
102

3.2
464

4.1
491

4.3
827

Total annual energy
generated [GW(e) a]
Normalized release [TBq (GW'.e) a)·l;

460 TBq [GW(e) a]-lAverage 1975-1979

Ta b1e17

Activation 9ases released from GCRs in various countries 1975-1979
Country
andreactorFrancetFi1iion 1
Chinen 2
Chinen 3
St. Laurent des Eaux 1
St. Laurent des Eaux 2
Bugey 1Italy
LatinaJapan
TokaT
United Kingdom
CalderChapel crossBerkeley
Bradwell
Hunterston A
Hunters ton B
Trawsfynydd
Hinkley Point A
Hinkley Point B
Dungeness A
Sizewell
Oldbury
Wylfa

Startupyear

Electrical
capacity[GW(e))

Release (TBq)
1975

1976

1977

1978

182

145

94

1979

1963
1965
1966
1969
1971
1972

0.070
0.210
0.400
0.460
0.515
0.540

] 224
] 129
195

107
114

155
89

250
107

1963

0.153

96

92

89

97

1966

0.166

207

229

289

307

333

1956
1959
1962
1962
1964
1976
1964
1965
1976
1965
1965
1967
1971

0.200
0.192
0.276
0.250
0.300
1.240
0.390
0.460
1.240
0.410
0.420
0.416
0.840

548
1184
592
555

962
1184
555
666
555

1073
1184
444
592
740

1100
1200
444
555

74
5550
2960

5550
2960

4810
2960

5180
2590

1110
2220

1110
2220

1110
2220

444
2220

4.59
3381

5 .10
3241

5.07
3153

4.4
3194

Total annual energy
generated (GW(e) a)
Normalized release [TBq (GW[e] a)- 1]
Average 1976-1979

4.42

3240 TBq [G~(e) a)-l289

Tab 1 e

Country
andreactorP II R
FranceOugey 2, 3
Chooz
Fessenheim
Gennany, Fed.Rep.of
Obngheim
Stade
Biblis, A,B
Neckarswestheim
Unten,ieser~
rrno
Netherlands
Borssel eUnited States
Arkansas 1
Arkansas 2
Beaver Valley
Calvert Cliffs
Cook 1
Crysta 1 River
Davis Besse
Joseph M. Farley
Fort Calhoun
R.E. Ginna
Haddam Neck
Indian Point
Kewaunee
Maine Yankee
Mills tone Point II
North Anna
OconeePoint Beach
Prairie Island
Rancho Seco
H.B. Robinson
Salem
St. Lucia
San Onofre
Surry
Three Mile Is land
TrojanTurkey Point
Yankee Rowelotal annual energy
generated [GW(e) a]
Nonnalized relea~y[T8q (GW[e] a) ]18

Release (TBq)
1975

]976

)978

1979

44

0.99
0.56
0.48

2.3
0.78
0.48
0.074

0.85
0.59
1.26
0.92

0.96
0.99
2.29
1.04

0.63
1.2
2.4
I. I

0.21

0.14

0.60

0.44

0.33

0.37

0.93

0.9

0.019

0.25

7.0

0.22

138

7.9
4.3
0.007
4.5
0.0014

0.47
0.15
0.82
0.19
0.5P
0.4
0.54
0.054
3.6
6.4
0.48
0.26
0.11
3.8
0.070
0.98
29.7
5.8
5,3
0.38
108
12.6
1.0
3.0
2.4
0. 39
0.035
0.13

0.088
0.21
2.59
0.48
1.37
0.17
0.063

0.093
0.89
27
0.89
0.026
0.14
0.56

0.11
1.90
0.0004
0.44
0.14
0.078
0.070

62.9
15.5
0.37

18.5
14.8
1.22
0. 34
5.85

0.13
0.074

0.074
1. 74
13.7
26.6
0.056
0.19
0.074

2. 31
7.22
3.24
o. 78
2.26
1.9
11.8
2.81
32 .5
4.77
0.11
0.14
0.12

14
0.06
0.72
0.96
1.2
2.9
0.14
1.6
2.5
0.38
0.46
0.041
1.5
0.66
2.5
6,3
5.3
8.9
1.6
5.3
19.8
1.1
8.1
7.9
0.15
0.13
0.11

15.7

16.8

21.3

24.4

22.8

6.2

15.2

4.5

5.9

8.5

0.046
0.0007

7.14
1.26
1.18
1.48

0.004

Average 1975-1979

290

1977

B WR
Germany, Fed. Rep.of
Gundremm1 ngen
LingenWiirgassen
BrunsbUttel
lsar

3.7
1.1
0.07

1.0
0.22

M*
r gliano

0.026

0.55

Netherlands
Dosewaard
United States
B1 g Rock Point
Browns Ferry
Brunswick
CooperDresden 1
Dresden 2 ,3

0.27
0.19
0.07
1.6
1. 3
8.1

0. 31
0.021
0.81
2.5
2.3
5.2

0.0074

I. I

7.8 TBq [GW(e) arj

0.26
0.20
0.89
0.18

0.037
0.078
0.28
0.24
0.056

0.40
0.87
0.052
1.9
0.56
18

o. 31

1.1
0.37
0.21
3.3
12

0.014

o. 33
0.34
0.25

0.12
I. 3
0.88
0.92
0.25
6.8

Table 18, continued
Country
andreactorDuane Arnold
J.A. Fitzpatrick
Edwin I. Hatch
Humboldt Oay
Lacrosse
Millstone PointMonticelloNine Mile Point
Oyster Creek
Peach Bottom
Pilgrim
Quad Cities
Vennont Yankee
Total annual energy
generated [GW(e) a]
Normalized releas~ 1(TBq (GW[e) a) ]Release (TBq)
1975

1976

1977

1978

1979

0.70
0.011
0.066
0.093
0.63
0.63
3.4
0.10
0.011
2.7
1.4
0.26

0.59
0.56
0.052
0.048
0.48
I. I
2.8
0.70
0.041
1.0
1.4
11.0
0.52

0.56
0.26
0.085
0;0024
0.32
2.4
5.1
1. 7
0.044
3.7
0.023
1.5
0.85

0.22
0.28
0.22
0.0015
0.29
1.8
8.7
3.2
1.4
0. 74
J.5
3.7
0.45

0.33
0.24
0.54
0.002
0.17
2.2
8.1
1.5
1.4
1.0
5.8
5.4
0.61

7. I

9.6

10.6

13.1

13.2

3.8

3.5

3.8

3.2

2.9

3.4 TBq [GW(e) a]·lAvera,~e 1975-1979
H WR
Argentina
AtuchaCanadaBruce
Pi eked ng
Total annual energy
generated (GW(e) a]
Nonnalized releas~ 1(TBq (GW[e] a) ]

38

220

224

222

237

891

315
1628

481

918

962

1399
1147

2.0

2.6

3.2

4.1

4.3

478

427

677

406

647

540 TBq (GW(e) a]-lAverage 1975-1979

Ta b l e

Country
andreactor
P WR
Finland
Lov11sa
FranceBugey 2 ,3Chooz
Fessenheim
German.z:, Fed. Rep.of
Obr1 ghe1m
Stade
Biblis A,B
NeckaNestheim
UnteNeser~ino~
a a 1,2

Mi hama 3
Takahama 1,2

19

Release (TBq)
1975

1976

1978

1.2

5.0

92

71

96
3.1

3.3
65
31

6.2
3.9
4.1

4.7
1.6
13
0.19

5.5
4.8
13
J .1

4.8
4.9
28
5.0
0.0081

44

27

64

77

2.4

5.2
I. 7
13

4.1
3.7

8.1
5.2
17
4,4
II
6.7

13

11

0.014
3.3

Ohi 1,2

Ikata 1Genkai

1977

2.3

1.1

11

1979

5.0
5.4
18
3.8
4.0

7.4
3.6
8.5
16
4.1
11

291

Table 19, continued
Country
andreactor
Netherlands
Borsse leSwedenRinglials 2
United States
Arkansas 1
Arkansas 2
Beaver Valley
Calvert Cliffs
Cook 1
Crystal River
Davis Besse
Joseph M. Farley
Fort Calhoun
R.E. Ginna
Haddam Neck
Indian Point
Kewaunee
Maine Yankee
Millstone Point 2
North Anna
OconeePalisades
Point Beach
Prairie Island
Rancho Seco
H.B. Robinson
Salem
St. Lucie
San Onofre
Surry
Three Mile Island
TrojanTurkey Point
Yankee Rowe
Zion
Total annual energy
generated [GW(e) a)]
Normalized releas1£ 1[TBq (GW[e] a) ]Release (TBq)
1975

1976

1977

1978

1979

2 .1

1.5

1 .5

8.4

8.3

10

8.8

11

8.9

17

7.8

9.1

10.8

9.7
2.1

0.32
10
7.1

4.0
21
11
6.1
0.33

4.5
8.9
179
12
6.7
14
10

5.8
4.4
247
14
11
5.7
7.8

12.9
16.8
23
5.7
8.0
2.2
5.6
8.9
146
28,4
10,9
11. 7
7,4
10.4
43
3.7
48
20

6.2
1.95
3.5
19
45
6 .1
9.1
3.5
9.5
8.9
131
18.2
9.2
7.5
9.4
11.6
33
4.7
33
23.3

4.1
9.6
210
2.9
10

6.5
0.28
131
1.5
33
0.017
4,9
23

29
9.1
38

81
71
0.36
2.1
37
27
0.0037 50
0.0031036
250.0015 11
9,0
0.49
125
66
29
15
7.0
7 .1
1. 3
12
34
29
5.8
5.1
27
28

16.9

19.2

23.8

29,4

25.8

52.9

38.7

40.0

33.0

30.0

148
16
17

Average !975-1979
B WR
Finland
ITTCTTuoto
Germany, Fed.Rep.of
Gundre111TI1ngen
LingenWiirgassen
Brunsbiitte l
Isar~gliano~uga

Tokai 2
Fukushima 1
Fukushima 2
Fukushima 3
Fukushima 4
Fukushima 5
Hamaoka 1
Hamaoka 2
Shi mane

Swedenllsl<iirsh amn
Ringhals 1
BarsebackUnited States
Big Rock Point
Browns Ferry
Brunswick

292

17.5
16.5
4.7
92,5
27 .6
5.7
5.9
43
7.2
27

15.9
86
4.7
85.8
13.2
2 .1
2.5
35

6.5
22

38 TBq [GW(e) ar 1

0.0086

0.46
0.026
0.028
1.5
0.085
0.80

4.7
0.63
0.15

1.9
0.56
1.1

0.81
0.092
1. 3
0.34
0.0015

0.093
0.070
2.0
0.70
0.17

0.19

0.67

0.56

0.37

1.5

1.9

1.0

o. 78

1.4
0.31
0.13
0.16
0.00520.000027
0.078
0.63
0.16

0.24

0.18

0.14
0.067
0.13
0.19
0.034
0.10
0.037 ]
0.085
0.19

1.3
0.85
0.52

1. 3
0.56
0.74

I .6
0.32
1.3

2.7
1. 7
1.1

0.074
0.15
0.22

0.33
0.89
0.33

0.15
1.1
0.52

0.20
0.49
1.1

0.14
0.00170.13
0.0073
0.24
0.14

0.22
0.37
0.11

Table 19, continued
Country
andreactorCooperDresden 1
Dresden 2. 3
Duane Arnold
J.A. Fitzpatrick
Edwin I . Hatch
Humboldt Bay
Lacrosse
Mi 11 stone PointMonti ce 11 oNine Mile Point
Oyster Creek
Peach Bottom
Pilgrim
Quad Cities
Vermont Yankee
Total annual energy
generated [GW(e) a]
Normalized releas~ 1[TBq (GW[e) a) )Release (TBq)
1975

1976

1977

197B

1979

0.30
0.01
2.0
0.012
0.18
0.22
o. 74
4.7
3.0

0.30
0.0007
0.74
0.013
0.16
0.33
0.48
1.5
0.74

0.33
0.00330.19
0.0078
0.12
0.44
0.020
1.8
0.16

0.28
0.49
0.71
4.4
0.070
0.33
0.001
1. 7
0.12

0.25
0.056
0.71
0.010
0.056
0.%
0.002
1.3
0.29

1.0
0.67
1.1
0.67
2.0

0.093
1.4
2.7
1. 7
1.8
0.059

0.092
0.70
2.6
1.2
0.98
0.031

0.73
1.2
0.11
0.64

0.25
0.052
1.6
0.50
0.63
0.15

9.53

13.04

13.24

18.1

20.0

2.6

1.6

1. 3

1.1

1.0

Average 1975-1979
GCR
Franceillnciii
St. Laurent
Bugey 1

9.0

3.9
19
7.2

1.4 TBq [GW(e) a]-l

4.1
13
8.9

~
o a

4.8
25
14
0.0006

0.033

2.8
4.4
1.6
2.1
3.6
160
0.78
1.6
0.22
4.6

United Kingdom

CalderChapelcrossBradwellBerkeley
Hunterston A
Hunterston B
Traws fynydd
Hinkley Point A
Hinkley Point B
Dungeness A
Sizewel 1
Oldbury
Wylfa
Total annual energy
generated [GW(e) a]
Normalized releas~ 1[TBq (GW[e] a) ]

0.26
3.3
2.6
2.0

0.93
1.8
0.52
4.8

0.33
11
1.1
2.4
1.6
0.59
0.89
0.11
1.3
2.3
0.70
7.3

0.074
7.4
1.9
2.0
2.0
0.48
1.2
27
0.96
1.6
0.56
11

1.2
3.8
0.59
2.0
85
0.56
2.0
59
1.2
1.1
0.27
39

4.42

4.59

5.10

5.07

4.4

6.91

13.0

16.0

47.2

41.0

3.3
2.0

25 TBq fGW(e) a]-lAverage 1975-1979
H WR
Argentina
AtuchaCanadaBruce
Pickering
Total annual energy
generated [GW(e) a]
Normalized releas~ 1[TBq (GW(e] a) ]

31

76

218

229

259

540

224

36
703

152
1188

546
1421

2.0

2.6

3.2

4. 1

4.3

286

115

299

383

518

Average 1975-1979

350 TBq [GW(e) a)-l293Tab 1 e

20

Carbon-14 discharges from U:Rs
in the Federal Refublic of Germany 1976-1978

Rl, S15J

Release (Gllq)

Reactor

1977

1976

1978

!....!i..!!. 2_/
111 ( 15) 33 (19)111(111) 55 (55)78 (22)178 (15)181 (4.4) 168 ( 15)
148(4.1) 144 (6)Obrigheim
Stade
Bibl is A
Bibl is B
Neckarwestheim
llormal ized releas~ 1[GBq (G\J[e] a) ]222!.!LI!. £_/Gundrenrni ngen
Lingenl{urgassen
Brunsbiittel

137
42

37
7.4
229
159

3.7
229
167

Normalized releas~ 1[GBq (GW[e] a) ]
a/

~

518

co bound values in parentheses.
Co 2
2 values only.

21

T a b 1 e

Iodine releases to atmos~here from reactors 1975-1979
re21. B23, 06, 08, £9, G , K4, Ll, (6, Nl9, P2, S14]Country
andreactorRelease (GBq)
1975

1976

1977

1978

1979

0.089
0.021

0.19
0.74

0.089
0.081

0.37

0.33
0.11

P WR

~e;f mTihange
Finland
Lov11sa
Germany, Fed. Rep. of
Obrighe1m
Stade
Biblis A,B
Unterweser
Neckan,iestheim
~

Trino

0.04
0.41
0.37
0.19

0.074
0.74
0.37

0.022
0.96
0.15

0.074

I. 7

0.0021
0.026
0.14
0.59
0.00074
0.26

0.03
0.086
0.68
0.045
0.032

0.00480.037
0.0078
0.014
0.041
0.089
0.028

0.00170.00003 0.0033

0.048

0.012
0.18
0.00520.00740.014
0.078
0.00820.070
0.00067

0.0089
0.093
0.010
0.010
0.078
0.00220.00056

0.52

0.31

0.13

0.0078

0.27

0.81

0.69

0.27

1.5

0.41

0.11

0.16
0.17
0.015
15.5
0.55
0.47
58.0
0.20

J_!_2!11_f.lTiiama 1

Mihama 2
Mihama 3
Takahama I ,2
Ohoi 1
Ohoi 2lkata I
Netherlands
Borsse leSwedenRTrigTia 1s 2
United States
Arkansas I
Arkansas 2
Beaver Valley
Ca 1vert C1iffs
Cook lFort Calhoun
Crystal River
Davis Besse

294

0.81

0.008118.0
0.0063 0.052
27.0
0.26
1. 3

0.0056 0.63
14.0
7.4
3.7
0.48
4.8
0.30
0.058 31.0
0.0041 0.11

Table 21, continued
Country
andreactor
Joseph M. Farley
R.E. Ginna
lladdam Neck
Jndi an Point
Kewaunee
Maine Yankee
Millstone Point 2
North Anna
OconeePalisades

Point [leach
Prairie Island
Rancho Seco
H.B. Robinson
Salem
St. Lucie
San Onofre
Surry
Three Mi le Island
TrojanTurkey Point
Yankee Rowe
Zion
Total annual energy
generated [GW(e) a]
Normalized releas~ 1[GBq (GW[e) a) ]Release (GDq)
1975

1976

1977

1978

2.4
0.033
15.0
o. 74
0.22
0.0017

1.3
0.027
10.0
0.12
0.059
1.8

0.41
16.0
7.0
0.78
0.007
0.85

5.9
1.1
0.23
0.93
0.04
9.3

9.3
4.8
0.035

0.074
0.17
20.0
0. 32

17.0
0.1

19.0
0.048
3.3

0.0017
0.32
0.11
3.0
0.15
0.074
0.53
1.1
6.3
14.0
0.89
1.6
0.56
0.23
0.30
0.74
0.096
18
0.44
0.070
0.20
38
3.0
• 0.0066
0.0078
18.0
2.2
0.78
0.96
0.97
0.89
52.0
18.0
0.0085 0.014
1.6
2.6

16.9

19.2

23.8

29.4

27.5

4.70

6.41

5, 32

4.68

4.18

0.96
0.030
5.2
o. 76
0.21
0.22

Average 1975-1979 5.0 GBq [GW(e)B WR
FinlandlllTiluoto
Germanl, Fed.Ree. ofGundremm,ngen
LingenWUrgassen
Brunsbi.ittel
lsar

0.017
9.3
48
0.052

13
0.19
1.9
0.052
1. 7
1.0
0.00074 0.56

~gli ano

0.59

1.3

0.56

~
s ga

1.2

1.0

0.37

0.48

8.9
0.0037

1.0
0.18

Takai 2
Fukushima 1.2
Fukushima 3,4
Fukushima 5
Hamaoka 1,2
Shimane1979
0.0005
0.71
1.8
6.1
0.043
4.8

0.62
1.4
7.4
1.0
0.78
0.16
0.20
0.006
0.30
7.4
0.0045
0.23
0.41
1. 7
1.7
0.014
0.45

ar 1
0.023

0.00280.00074
4.1
2.5
1.8
0.018
0.003

0.19
0.012
32.0
0.048

0.17
0.028
5.9
0.096
0.0052

SwedenJITiignal s 1
Oskarshamn
BarsebackUnited States
B1 g Point RockBrowns Ferry
Brunswick
CooperDresden 1
Dresden 2,3Duane Arno 1d
J.A. Fitzpatrick
Edwin 1. Hatch
Humbo 1dt Bay
Lacrosse
Mi 11 stone PointMonticelloNine Mile Point
Oyster Creek
Peach Bottom

0.67
2.7
0.070
10.0
22.0
0.10
16.0
211.0
4.1
15.0
0.67
0.24
41.0
4.8
2331.0555.0
222.0
1517.0
1.3

0.32
8.9
0.031

1.2
5.5
0.089

23
5.0
0.10

5.4
5.9
7.4
0.24
14.4
22 .0
5.9
5.8
78.0
38.0
17.0
28.0
5.4
3.3
1. 7
8.0
997.0
85.0 2020.0
0.12
4.8
4.8 2110.0 1298.0
1583.0
4.1
4 .1
569.0
9.4
215.0
7.0
0.86
0.084
0.13
0.27
0.92
0.00044
14.0
9.3
1.5
0.82
3.7
1332 .0 2257 .o 1417 .0
141.0
37.0
20.0
15.0
10.3
44.0
52.0
318.0
5.8
1702.0 1517.0 2294.0 2868.0
52.0
49.0
70.0
63.0

295

Table 21, continued
Country
andreactor

1975

1976

1977

1978

1979

Pilgrim
Quad Cities
Vermont Yankee

229.0
107.0
15.0

74 .0
159. 0
2.6

126.0
222.0
0.81

61.0
407.0
1l. 0

52.0
523.0
21.0

Release (GBq)

Total annual energy
generated (GW(e) a]
Normalized releas~ 1(GBq (GW[e] a) )

9.53

13.04

13.24

18.1

20.0

561

314

484

501

280

Average 1975-1979 410 GBq [GW(e) ar 1
GC R
Francerliiiiiin

0.081
0.070
0.41

0.78
0.81
0.59

Bugey
St. Laurent

~akan
o a1
H \.I R
Argentina_~tucna [GBq (GW[e] a) 1 ]

0.24

1.2

0.16

0.00410.0030

5.4

8.7

Average 1975-1979 3.1 Gtlq [GW(e) ar 1

Ta b 1e22

lsotoeic comeosition of iodine releases
from reactors 10 tne On1teo ~tates 111 1979
[B23]
Release (GBq)
Reactor
1311

P WR
Arkansas I
Arkansas 2
Beaver Valley
Calvert Cliffs
Cook lCrystal River
Davis Besse
Fort Calhoun
Joseph M. Farley
R.E. Ginna
Haddam Neck
Indian Point
Kewaunee
Maine Yankee
Millstone Point 2
North Anna
OconeePalisades
Point Beach
Prairie Island
Rancho Seco
H.tl. Robinson
St. Lucie
Salem
San Onofre
Surry 1,2
Three Mi le Island 1
Trojan

296

0.16
0.17
0.015
11.1
0.53
0.69

1321

134 I

1351

0. 0067
0.28

O. lB
0.048
0.0005
0.35
1.7
4.3
0.020
2.0
0.53
1.3
5.1
0. 70
0.16
0.14
0.17
0.0014
3.7
0.20
0.045
0.23
0.41
0.53

1331

0.044
0. 00780.10
0.36

0.00030.0015
4.0
0.026
0.15
0.023
0.067
0.32
0.093
0.46
0.011
2.6
0.036
0.052
1.8
0. 31
0.26
0.023
0.030
0. 00213.7
0.10

0.00440.0029

0.001

0.47

0.00180.14

00.00035

0.033
0.008

58.0
0.0041
0.35
0.14
1.3
0.0039
0.23
0.059
0.37
0,083
0.0029
0.004
0.004

0.70

Table 22. continued
Release (GBq)
Reactor1311

Turkey Point
Yankee Row
Zion 1,2

132 I

1341

0.58
0.00280.22

1.2
0.00670.23

Total annual energy
generated [GW(e) a] 19 .165
Nonnalized releas~ 1
1.9[GBq (GW(e] a) ]B WR
Big Rock Point
Browns Ferry 1,2 ,3Brunswick 1 ,2
CooperDresden 1
Dresden 2,3Duane Arno 1d
J.A. Fitzpatrick
Edwin I. Hatch
Lacrosse
Mi 11 stone PointMonticelloNine Mile Point
Oyster Creek
Peach Bottom 2,3
Pilgrim
Quad Cities 1,2
Vermont Yankee

1331

1351
0.065
0.00480.0017

0.042

0.78

1.3

0.11
1.1
17,.8
0. 76

0.13
3.7
4.2
4.7

390
2.7
0.44
0.043
0.28
59
3.4
1.5
1095
23
5.7
210
4.4

834
566
0.21

0.011
1.0
4.7
2.5
0.12
75
0.35
0.21
0.88
0.41
15
0.89
1.0
326
9.5
3.7
35
15

0.011

Total annual energy
generated (GW(e) a] 11.670
Normalized releas~ 1[GBq (GW[e] a) ] 42

155

0.11

T a b 1 e

3.2

0.030

0.088
67
6.0
3.3
1447
16
43
278
1.8

0.0026281

23Particulate releases from reactors in various countries 1975-1979(823, D6, DB, £9, GJ, K4, LI. L6, Nl9, P2, s14)
Country
andreactorRelease (GBq)
1975

1976

1977

1978

7 .8

6. 3

0.0018

1.1

6.7
3.2

1979

P WR

~

6.7

Tihange
Finland0.042

LOVll sa

FrancerfioozFessenheim
Bugey 2 ,3Germany, Fed. Rep.of
06righe1m
Stade
Biblis A,B
Neckarwes theim

0.089

0. 063

3,4

0.44
0.67
0.059

0.93

0.30
0.26
0.11
0.019

0.24
0.33
0.20
0.45

0.16
0.37
0.0013
0.19

0.075
0.025
0.016
0.025

l. l

0.22

~

irTno
Netherlands
Borsse leSwedenITTrigli'a 1s 2
United States
Arkansas 1
Arkansas 2
Beaver Valley
Ca 1vert C1iffs

0.037

0.0037
0.052

0.025

0.00052

0.004

0.30

0.26

0.083

0.009

0.41

0.63

0.0044

0.044

0.41

0.00005 0.0002
0.59
5.2

0.67

0.00074
0.00740.00074 0.00059
0.50
0.18

297

Table, ?l, continued
Country
andreactorCook 1,2
Crystal River
Davis Besse
Joseph M. Farley
Fort Calhoun
R.E. Ginna
Haddam Neck
Indian Point
Kewaunee
Maine Yankee
Millstone Point 2
North Anna
Oconee
Pa 1i sades
Point Beach
Prairie Island
Rancho Seco
H.B. Robinson
St. Lucie
Salem
San Onofre
Surry
Three Mile Island
TrojanTurkey Point
Yankee Rowe
Total annual energy
generated [GW(e) a]
Normalized releas~ 1(GBq (GWfe] a) ]Release (GBq)
1976

1977

1978

1979

0.00020.52
0.037
0.00880.00410.00160.078
46
24
0.00930.00040.01
0.001
o. 0063
1.0
0.012
0.078
0.11

0.00270.0026
0.036
0.56
0.021
0.012
0.014

0.010
0.022
3.5
0.13
0.00290.050

3.6
0.00060.00831.5
0.00450.0048
0.081
6.9
0.059
0.0081
0.029
0.036
1.9
0 .15
0.37
0.0015
0.52
0.013

2.2
0.30

3.7
7.8
0.56
0.041
0.59
0.067
0.0089 •0.041
0.002
0.093
0.093
0.021
0.00056 2.8
100.00001 1.3
0.00067 0.00018 0.093
0.56
3.0
o. 13
0.16
0.44
3.7
0.95
0.37
0.24
2.8
1. 9
2.9
0.00081 0.0012
0.0002

2.2
0.16
0.033
0.81
0. 0059
0. 34
0.11
12
0.003
0.037
3.0
0.81
3.7
0.059
0. 35
0.003
0.048
0.015
3.7
0.081
0.00074
0.052
0.04
0.74
l. 7
0.003

15.7

17.3

21.8

24.8

23.6

5. 77

1.27

1.54

1.85

1.28

1975

l. 3
2.9
0.0037

Average 1975-1979 2.15 GBq [GW(e} a]-l
B WR
Finland
ITTTITiiot 0
Germany, Fed.Ree.of
Gundremmingen
LingenWUrgassen
BrunsbUtte l
Isar
Italy
Gangl iano
Nether 1ands
Dodewaard
S11eden
IYsk'arsh arnrn
Ringhals l
Barsebiick
United States
B1g Rock Point
Browns Ferry
Brunswick
CooperDresden l
Dresden 2,3Duane Arnold
J.A. Fitzpatrick
Edwin I. Hatch
Humboldt Bay
Lacrosse
Millstone PointMonticelloNine Mile Point
Oyster Creek
Peach Bottom
Pilgrim

298

0.30
3.7
4.0

0.22

3.6
2.2
0.14
1.1
13.0
153.0
0.33
0.67
0.01
31.0
0.48
7.0
24.0
16.0
6.7
0.14
24.0

0.0059

0.33

0.15
0.056
2.1
1.4
0. 16

0.004
0.0013
1.4
7.4
0.79

0.19
0.019
0.63
0.26

0.27
0.070
1.4
2.7
0.0011

l .2

0.74

0.12

0.11

0.035

15
4.9
0.096

8.3
20
0. 72

l!

0. 32
1.5
2 .0
2.4
5.2
0.70
0.41
0.17
17.0
22.0
167.0
1561.0
0.27
0.25
0.70
1.1
0.031
0.017
0.015
1.0
0.41
0.85
7.4
5.6
0.74
1.9
3.7
1.8
a.1
48.0
1.4
I. 0
7.8
12.0

1.2
1.6
0.22
3.7
4.1
0.081
41
89.0
0.85
2.0
0.085
0.026
0.12
50
0.50
1.0

296.0
0. 46
2.1

1.6
94
0. 30
0.059
4.8
25
0 .11
O. 77

185,0
0.89
0.32
0.078
0.004
0.48
7.0
0. 37
0.74
19.0
0.30
2.0

Table 23, continued
Release {Gllq)

Country
andreactor

1975

1976

1977

1978

1979

Quad Cities
Vermont Yankee

15.0
0.074

20.0
0.21

27 .0
0.28

41.0
0.20

23.0
1.1

7. 1

10.9

11.9

15.5

15.5

43.0

24.4

144 .o

35. 5

24.2

Total annual energy
generated [GW(e) a}
Norma 1i zed re leas~ 1[GBq (G~Jfe) a) ]Average 1975-1979 52.7 GBq fGW(e) ar 1
GC R
Francet1iiiiori
Bugey
St. Laurent

0. 37
0.063
0.048

0.67
0.30
0.070

,0.48
0.28
0.23

0.48
0.67
0.16

0 .11
0.14

0.096
0.11

2.0
0.37
1. 3
0.15

0.15
0.074
0.37
0.019
0.37
0.41
1.1
0.52
0.48
0.074
0.37

0.15
0.15

0.27
0.48
0.48
0.70
0.70
0.41
0.067
0. 31

0.11
0.11
0.37
0.26
0.67
0.37
0.67
0.41
0.56
0.11
0.37

4.42

4.59

5.10

5.07

4.3

1.19

1.02

0.98

1.03

0.86

~1
United Kingdom
BradwellBerkeley
Hunters ton A
Hunters ton B
Traws fynydd
Hinkley Point A
Hinkley Point B
Dungeness A
Sizewell
Oldbury
Wylfa
Total annual energy
generated rGW(e) a}
Normalized releas~ 1(GBq (GW!e} a) }

0.41
0.41

Average 1975-1979
H WR
Argentina
lltucha [i'.;Bq (GW[e J af 1]

0.018

0.037

0.41
0. 74
1.1
0.33
0.45
0.10
0.26

1.0 GBq [GW(e) ar 1

0.028

0.060

0.078

Average 1975-1979 0.044 GBq [GW(e) a]-lCountry
andreactorRelease (GBq)
1975

1976

1977

1978

381
14

1840
31

625
132

696

4

18

1979

P WR
Belgium~

Tihange
Finland
Lov11 sa
FranceLfioozFessenheim
Bugey 2 ,3Germanx, Fed. Rep.of
Obr, ghe1m
Stade
Biblis A

318

64
10
27

95

36
12
B.l

4. l
256
9.6
14
3.8

72

15

174
364
6.3
3.7
4.7

6.3
9.7
1.9

299

Table 24, continued
Country
andreactorRe 1ease (G[lq)1975Bibl is B
Neckarswestheirn
Unterweser

~
nno
~ahan
1 arna 1
Mi hama 2
Mihama 3
Takaharna 1
Takaharna 2
Ohi 1
Ohi 2
lkata 1Genkai I
Netherlands
Borssele
SwedenIITiigfial sUnited States
Arkansas I
Arkansas 2
Beaver Valley 1
Calvert Cliffs
Cook I
Crystal River
Davis Besse
Joseph M. Farley
Fort Calhoun
R.E. Ginna
Haddam Neck
Indian Point
Kawaunee
Maine Yankee
Millstone Point 2
North Arma
Oconee 1.2,3
Palisades
Point Beach 1, 2
Prairie Island 1,2
Rancho Seco
H.B. Robinson
St. Lucie
Salem 1San Onofre
Surry 1,2
Three Mi le Island
TrojanTurkey Pt. 3,4
Yankee RoweZion 1,2
Total annual energy
generated [GW(e) a]
Norma 1i zed re 1eas~ 1[GBq (GWfe] a) ]

54

1976

~
r gliano~surugaTokai 2
Fukushima

300

1978

11
8.9

I. I
5.6

4.4
1.1

100

52

48

1979
7.8
2. I
0.07

0. 74

0.22

0.28

0.32

0.34

0.16

0.01
0.044
0.14
0.31
0.0014 ] 0.041

0.048
0.074

0.078
0.063

0.037

0.048

0.06
0.007

31

16

II

11

225

427

209

115

115

485

167

223

53
9.6

6.3
44
69

24
129
56
0.57
0.96

20
26
4.8
184
105
105
9.6

13
2.4
63
112
47
16
57

9.7
227
55
1.1
3.3
3.7
22
2.2

293
16
120
3.7

1340
3.4
56
0.49

111
26
3.8
103
9.9
241
3.6
25
0.18

114
48
4.5
289
95
15
1.6
2.2
9.0
3.2
32
87
33
17
180
22
34
4.7
27
0.33

14
12
3.0
215
0.37
107
45
215
364
343
12~6
2424
2.6
3.7
7.2
102
155
114
320
329
0.74
0.37
0.67
0.37
35
5.9

6.6
104
148
437
89
23
26
123
3.0
35

99
147
407
94
18
21
15
0.43
26

16.9

19.2

23.8

29.4

27.5

145

310

306

126

74

0.033
60

13
16
44
182

27
119
0.74
187
128
87
17
0.37
17

Average 1975-1979
B WR
Finland
ITTTiiuoto
Germany, Fed. Rep.of
Gundrernm1 ngen
LingenWUrgassen
BrunsbUt te l
lsar

1977

JS

184 GIJq [GW(e) ar 1

1.2

7.7
B.3
0.93
16
17
9.7

47
1.5
69

43
9.6
41
83

64
0.37
58
61
1.5

20
0.37
20
52
7.8

116

139

152

111

1.5

1.9

11

1.0
0.14

0.78
0.14
0.067

1.4
0.31
0.13

Table 24. continued
Country
andreactorfukushinl<'I 2
Fukushima 3
Fukushima 4
Fukushima 5
Hamaoka 1
flaniaoka 2
ShilUilne
Sweden
IJ'sta'rsh amn 1,2
Ringhals 1
Barseback 1,2
United States
81 g Rock Pornt lBrowns ferry 1,2,3
Brunswick 1 .2
CooperDresden 1
Dresden 1,2Duane Arnold
J.A. Fitzpatrick
Edwin I. Hatch
Humboldt Bay
lacrosse
Mi 11 stone PointMonticelloNine Mi le Point
Oyster Creek
Peach Bottom 2,3
Pilgrim
Quad Cities
Vermont Yankee
Total annual energy
generated (GW(e) a]
Normalized releas!l 1[GBq {GW[e] a) ]Release (GBq)
1975

1976

1977
0.017
0.13
0.0073
0.24

0.14

0.24

0.18

1978

1979

0.13
0.16
0.19
0.00520.034 0.000027
0.10
0.078
0.037
]
0.63
0.085
0.19
0.16

903
232
64

485
249
68

158
148
140

155
223
58

74
100
70
64
31
30
3.7
197
2.2
140
525
7360

28
146
122
2.6
13
45
3.7
222
1.5
37
214
357

14
44
230
• 28
22
16
0.085
33
925
34
788
19

10
488
129
113
12
15
10
58
1.5
7.2
328
6.5

33
66
189
92
0.98
9.8
0.02
24
1.8
3.5

780
15
34
296
633
0.37

79
8.1
125
86
258
0.37

11
3.6
83
126
50
5.7

0.55
189
65
83

70
0.25
722
19
48
0.01

9.53

13.04

13.24

18.1

20.0

1100

251

273

121

96

124.7

7.8

Average 1975-1979 290 GBq (GW(e)a ]-l
GCRFranceCr11nonBugey 1
St. Laurent
~akan
o a1
United Kingdom
Chape lcrossBradwellBerkeley
Hunterston A
Hunterston B
Trawsfynydd
Hinkley Point A
Hinkley Point B
Dungeness
Si zewell
Oldbury
Wylfa
Total annual energy
generated (GW{e) a]
Normalized releas~ 1(GBq (GH[eJ a) ]

24
511
174

21
133
110

8.5
144
181

13
401
300

640
4400
2000
4260

2940
740
1010
125

1200
2420
4140
5880
22
740
5110
41
1710
1090
1860
241

337
2450
5480
5440
44
500
4440
44
1680
1590
2440
688

2830
2000
1180
2220
220
648
4070
189
1410
910
1120
977

289
2812
600
1110
1420
706
600

4.42

4.59

5.10

5.07

4.3

5280

5390

4990

3650

4628

629
5880

Average 1975-1979
H WR
Ar9entina~tucha [GBq (GW[e] a)- 1 )

520

750

Average 1975-1979

9100
1576
1687

4767 GBq [GW(e) a]-1

530

230

390

473 GBq [GW(e) a]-1

301

Ta b 1e25

Radionuclide comeosition of liguid dischar~es other than tritium from reactors in the United States in 1979
Release (GDq)

Re;ictor
131 I
P WR
Arkansas 1
Arkansas 2
Beaver Va 1ley 1
Calvert Cliffs 1
Cook 1
Crysta 1 River
Davis Besse
Joseph M. Farley
Fort Calhoun
R.E. Ginna
Haddam lleck
Indian Point
Kawaunee
Maine Yankee
Millstone Point 2
North Anna
Oconee 1 ,2 ,3Palisades
Point Beach 1 ,2
Prairie Island 1,2
Rancho Seco
H.B. Robinson
St. Lucie
SalemiSan Onofre
Surry 1,2
Three Mi le Is 1and
TrojanTurkey Pt. 3 ,4Yankee RoweZion 1,2
Total annual energy
generated \GW(e) a}
Nonnalized releas~ 1(GBq (GW[e l a) ]

302

10

8.9
0.032
24
0.44
2.2
0.13
0.049
0. 78
0.34
2.5
5.1
0.022
15
4.4
5.9
5.3
0.014
0.33
0.028
0.14
1.8
0.70
0.93
2.4
5.3
0.47
0.74
0.15
0.43

1321

1331

0.00040.0033

0.012
1. 7
0.030
14
0.024
0.070
0.00740.00370.10

0.46
0.00630.0048

1341

1351

24Na

51cr

54Mn

0.0037

0.19

0.002
2.5

11
0.67
0.048
24
5.2
0.63
0.034
0.093
0.57
0.13
o. 31
2.9
1.9
0.0023
8.2
0.033
0.36
0.47
0.020
0.016

1. 5
1.4
0.27
4.1
2.9
0.20
0.24
0.11
0.13
0.048
o. 093
1. 3
0.63
0.014
5.9
0.048
0.27
0.46
0. 27
0.0013
2.1
15
0.28
1.5
0.054

0. 39

5.2
3.5
8.3
7.4
0.25
2.7
0.18
0.015
2.2

1.1

4.6

2.2

1.3
0.00560.00026

0.52
0.00330.00370.056

0.000960.00370.0059
0.019

0.21
0.19
0.89
0.36

0.074

0.096

0.36

6.3

0.78

3.1

0.00570.35
0.11

0.0037o. 0910.0032
0.16

5.2

4.9

4.3

7.1

8.9

0.49

0.071
0.0037
0.11
2.1
0.015
0.00057

0.026
1.2

0.0067
0.19

0.052
3.0
0.012
0.009656Mn

57Co

0.0037

0.11
0.12

0.024

0.24
0.085
0.00360.089
0.000150.00020.00180. 00230.00160.00170.74
0.018
0.000190. 0059

1.6

0.0042
0.054

1.1

0.074
0.048
0. 005

l. 3

0.041
0.004
0.61

0.024

19.165
4.6

0.31

l. 7

0.24

0.68

0.085

0.084

Table 25, continued
Release (GDq)
Reactor
58co
P WR
Arkansas 1
58
Arkansas 2
30
Beaver Valley 1
1.6
Ca 1vert Cliffs 1
141
Cook 1
33
Crysta 1 River
4.3
Davis Besse
0.85
Joseph M. Farley
0.96
Fort Calhoun
0.41
R.E. Ginna
0.081
Haddam Neck
0.52
lndi an Point
11
Kawaunee
11
Maine Yankee
0.67
Millstone Point 2
60
North Anna
4.8
Oconee 1,2,3
6.7
Palisades
2.4
Point Beach 1,2
1.8
Prairie Island 1,2
0.11
Rancho Seco
H.B. Robinson
3.7
St. Lucie
37
Salem 152
San Onofre 1
343
67
Surry I .2
Three Mile Island 1 2.3
Trojan5.2
Turkey Pt. 3,4
1.8
Yankee Rowe
0.015
Zion 1,2
5.2

59Fe

60co

0.74
17
0.56
0.81
0.033 2.3
1.1
12
0.70
20
0.093 1.3
0.0093 0.093
0.019 0.81
0.074 0.15
0.0074 1. s
0.00015 3.6
0.56
S.9
0.085 6.7
0.052
58
0.67
0.0033 0.47
0.019 1.8
0.047 0.46
0.0033 0.59
0.015 0.03S
0.085
0.37
0.98
7.3
0.00600.00490.15
0.00630.041

Total annual energy
generated !GW(e) a] 19 .165
Normalized releas~ 1
0.71(GBq (GW(e] a) ] 46

97zr

95Nb

97Nb

99Mo

0.01
1.1
0.031
0.0016 0.0033
0.019
0.096 5.6
0.11
0.0044 3.0
0.26
0.26
0.0017 0.012
0.00074 0.0074
0.0048 0.079
0.033
5.4
0.00015 0.25
0.13
0.18
0.33
0.52
1.8
0.030 0.0063 0.11
0.00170.00150.0044 0.048 0.0063 1.1
0.18
0.033
0.13
3.0
0.27
0.37
0.011
0.021
0.0026 0.00063 0.081
0.00034 -

2.0
0.047

0.16
0.022

0.11
0.031

6Szn

89sr

0.013 0.11
0.0013 0.24
0.00070.085
0.00960.21
0.034 1.8
0.00140.01
0.00110.089 0.018

90sr

95zr

0.015

0.59
3.0

0.45

0.023
0.046
0.024

0.52
0.036
0.00130.032
0.041
0.00015
6.3

0.13
2.6

3.3

0.0024
0.037

0.15
0.054
0.046

l.O

O.OS9

2.9
0.12
7.0
70
25
61
0.54
0.015
6.4
o. 0097
3.1
0.0069 0.0036
7.1

0.00004 0.67
0.0095
0.37
0.60
0.025
0.59
0.052 0.016
0.71
3.4
0.27
0.00110.34
0.065 0.0037 0.48
0.78
0.29
0.048 0.029
0.012
0.0006 0.0021
0.021 0.016 0.14
0.50

17

0.54

0.12

0.070
0.002S

0.028

0.086

1.0

0.0084 0.60

0.00740.00032
0.000780.00440.0083

0.23

0.38

303

Table 25, continued
Release (GBq)

Reactor
99mTc
P WR
Arkansas
Arkansas 2
Beaver Valley 1
Calvert Cliffs
Cook 1
Crystal River
Davis Besse
Joseph M. Farley
Fort Calhoun
R.E. Ginna
Haddam Neck
Indian Point
Kawaunee
Maine Yankee
Millstone Point 2
North Anna
Oconee 1,2,3
Palisades
Point Beach I ,2
Prairie Island 1,2
Rancho Seco
H.B. Robinson
St. Lucie
Salem 1San Onofre
Surry I ,2Three Mile Island 1
TrojanTurkey Pt. 3,4
Yankee RoweZion 1,2

103Ru

J06Ru

Jl()nAg

0.12
0.0051

0.040

1.1
0.061

124Sb

125Sb

134Cs

136cs

137Cs

3.6
5.6
0.057 0.0081 0.95
0.023
0.10
14
0.47
0.93
3.5
1.0
10
21
0.12
0.048
0.15
I. 7
0.014
0.00742.5
0.00330.048
0.12
0.0013 0.0063
0.0052 0.00019 0.016
0.00004 0.016 0.059
2.3
0,070 0.37
0.070
0.011 0.05
0.12
0.028
0.049 0.48
0.00015 2.4
1.4
0.00015 1.5
0.002
1.8
0.00040.0074 0.011
5.2
16
0.22
2.4
2.9
1.2
2.1
3.2
' 0.12
0.78
12
1.6
14
2.4
3.3
0.0028 0
0
0
4.6
0
0.061
4.6
0.37
8.9
0.036 0.0020 0.25
0.37
0.074
0.29
0.00048 2.0
0.081
0.0085 0.019 0.56
0.0034
0.027
0.036

0.027

1.2

1.2

0.50

1.1

0.011
0.22
0.010

0.057
0.63
0.12
0.0011
0.24

0.63

0.008
0.00440.00130.029

Total annual energy
generated (GW(e) a] 19 .165
Normalized releas~ 1(GBq (GW(e] a) ] 0.087 0.23

0.12

0.031
1.5
0.45

0.041
1.0

I. 7
0. 12
0.019

0.016 0.098
0.16
0.37
0.58
0.63
0.0016
0.80

0.13
10
2.1
0.87
16
0.86
0.078
I. I
0.075
4.4

0.35

4.4

0.21

0. 76

I. I
14
1.6
l. 7
0.0042 36
0.058 3.7
0.0032 0.19
1.8
0.098
5.3

7.7

140Ba/La 14lce

144ce

0.25
0.018

0.037
0.0090
0.21

0.70
0.016
0.013
0.016
0.0037
0.14
0.013
0.017
I. 7
0.059

0.022 0.17
0.00740.00019 0.0033
0.12
0.0025 0.0070
2.0
10
0.50
0.0096

0.035

0.018

0.13

0.12

0.17

1.1

0.074

0.013

0.00070 0.019
0.030 0.11

0.015
0.84
0.54

0.0060 0.12
0.0019 0.0013
0.074 0.14

0.00610.070

0.00210.0083

0.29

0.16

0.57

Release (GBq)
Reactor
131 l

1321

1331

0

0.010
0.78
0.65

1341

1351

24Na

Sier

54Mn

0.0014
10
60
0.046

0.21
17
13
2.1

2.2
3.7
18
5.5
0.037
l. 1
Q.0037
5.3
0.035
0.18
2.4
1.0

56Hn

57Co

B WR
Big Rock Point
Browns Ferry 1,2 ,3
8runswi ck 1 ,2
CooperDresden I
Dresden 2, 3Duane Arno 1d
J.A. Fitzpatrick
Edwin I. Hatch
Humboldt Bay
Lacrosse
Mi 11 stone PointMonticelloNine Mile Point
Oyster Creek
Peach Bottom 2,3
Pi }grim
Quad Cities
Ve nnont Yankee

1.4
8.0
3.4

0.012

0.18
0.023
0.18

0.00020.010
0.0019

1.8
0.20

0.17
0.0004

0.94
0.028

0.12

0.027
17
0.005

0.052
36
0.00070.00070.009
0.0008

1.8
0.11

0.023
4.4

0.0009
343

4.2

1. 7

0.0008

0.41

0.14
0.12
0.16
0.29

0.33
0.014

0.0002

Total annual energy
generated (GW(e) ~ 11.670
Normalized releas~ 1[GBq (MW[e] a) ) 4.7
0.025

304

o. 002336

0.00190.81
0.0024
0.27
0.48
0.26
0.0022

3. I

3.5

0.096
2.4
0.25

0.10
0.024
0.00530.00037
0.00370.00062.5

0.00015

0.011

0.22

Table 2S, continued
Release (GBq)

ReactorB WR
Big Rock Point 1
Browns Ferry 1,2, 3
Brunswick 1,2
CooperDresden I
Dresden 1,2Duane Arnold
J.A. Fitzpatrick
Edwin I. Hatch
Humboldt Bay 3
Lacrosse
Millstone Point 1
MonticelloNine Mile Point
Dys te r Creek
Peach Bottom 2,3
Pilgrim
Quad Cities
Vermont Yankee

58co

S9Fe

60co

65zn

89Sr

90Sr

0.063
0.85
3.6
1.3
0.051
0.24

0.19
4.8
0.47
7.9
1.1
18
0.37
11
0.0037 0.46
0.012 5.7
0.00968.9
0.061
0.0052 0.29
0.44
0.061 33
0.070 3.6

0.034
11
0.68
6.5

0.25
0.30

0.016
0. 32
0.014
3.6
0.072
0.30
0.00210.0074
0.012

1.1
0.006

0.032

0.042
0.14
0.008
0.27
0.00740.07
0.00210.00560.00310.078
0.43
0.0067

17
0.012
17
6.0
3.1
0.025
0.17
7.o
0.00056 0.0011

0.037
0.0011
o. 74
0.078
19
0.00020.!20.00012 0.030
0.019 0.0052
0.20
0.0001 0.00085

-

3.0

0.12

0.015

2.1
0.089
18
0.15
0.22

2.9
0.059 0.013
0.089
0.0004 0.001

Total annual energy
generated [GW(e) a] 11.670
Normalized release[G8q (GW[e] a)-1) 2.5
0.20

Jl

3.2

11

95zr

97zr

9SNb

97Nb

99Mo

0.035

0.35
0.11
0.21

4.9
0.17
0.21
0.011
0.07

0.18
0.013
0.17

0.00240.0013 0.00015 0.0033 0.0011

0.019
0.0025

3.7
0.016

0.062
0.00090.76

6.0
0.0003 0.067

0.19
0.002

0.017

0.00590.00740.00011

0.023

0.019

0.066

137Cs

140 Ba/La

144ce

239Np

18
2.8
21
14
0.21
1.1
0.0003
3.4
0.22
2.2
18
0.17

0.059
0.19

37
0.026
121
1.1
12
0.00041

0.023
0.89
0.30
6.1
0.0012

0.023

22

1.0

1.0

Release. (GBq)
Reactor
9brc
B WR
Big Rock Point 1
Browns Ferry 1,2,3
Brunswick 1,2
CooperDresden 1
Dresden 2 ,3Duane Arnold
J.A. Fitzpatrick
ECMi n I. Hatch
Humboldt Bay 3
Lacrosse
Millstone Point 1
Monti celloNine Mile Point
Oyster Creek
Peach Bottom 2,3
Pilgrim
Quad Cities
Vermont Yankee

10\u

0.35
0.41
0.19
0.00370.069
0.00130.0021
1.3

0.014
3.7
0.00741.1

0.0067

llOmAg124:.,J

134cs

0.052·
0.60

0.16
0.41
0.003
0.11

1.9
1.9
12
11
0.076
0.41
0.0009
1.6
0.20
0.27
4.5
0.037

0.00150.23
0.00030.00220.000520.00130.00370.00480.0048

0.01
0.026

0.020

0.060

17

0.027

0.023

0.43
0.014
0.042

13
0.0001
145
0.22
7.4
0.37
0.00032

0.00011

Total annual energy
generated [GW(e) a] 11.670
Normalized releas~ 1
0.10[GBq (GW[e] a) ) 0.51

136cs

0.076

0.047
0.41

0.70
0.037
0.033
0.00059 0.0026
0.001
0.0005
0.074
3.7
12
0.12
0.026

0.0022
3.1
0.059
0.007
1.1

0.40

305

Tab 1 e

26

Lifrom GCRs

Release (GBq)

Reactorllradwe 11
Berkeley
Traws fynydd
Hinkley A
DungenessSizewell
Oldbury
1/ylfa

355

55re

60Co

250
350
120
210
46
140
260
12

79
2
3

14
2

89Sr

8

l

4
4

0.6
6
3
0.6

14
2
0.1
11
3
4
7
0.6

6

10

0.6

18
4

Total annual energy
generated
4,3 GW(e) aNonnalized releas~ 1[GBq (GW[e] a) ]320
28

106Ru 125Sb

90Sr

134Cs

I 37cs

144ce

150
78
27
350
95
91
52
23

2
2
6
67

2
30
10
140
I
2
o. 7 4
7
4
I
2

120
190
9
150
120
170
18
37

670
860
49
1160
730
870
200
480

17
15
3
67
2
7
8

200

19

190

1170

28

Tab I e27

46l

Po ulation distribution around the model reactor site
an meteoro 091ca c aracier1st cs o its ocat1on
METEOROLOG !CAL CHI\RI\CTER I ST! CS
Pasqui 11 weather ea tegory
Dr yQuantityA

Frequency (%) _\:ind speed (m s 1)
Depth of mixing
layer (m)2000

Ra i n

e

C

D

E

F

C

D

11
2

19
5

32
5

16
3

18
1

l

25

5

5

2000 1000

1000

200

200

1000

1000

. l ' 0.16, 0.13, 0.074

Stack height: 30 m
POPULATION DISTRIBUTION

306

Distance (km)

0 - 1

1 - 2

2 - 5

5 - 10

10 - 20

20 - 50

Population
Cumulative

1300
1300

3000
4300

26000
30000

90000
1200004300005500002800000
3300000

Di stance (km)50-100100-200

200-500

500-1000

1000-2000

Population
Cumulative

6100000
9400000

20000000
29000000

70000000 140000000 17000000
99000000 240000000 260000000

Ta b l e28

Normalized local and re ional collective dose conunitments
or no e 2ases ram t e mo e

Normalized collective absorbed dose conrni tment (10-4man Gy [GH(e) a]-1)

Radionuclide
Gonads41Ar
85mt:r
85Kr
87Kr
88Kr
13lmxe133mxe133Xe
135mxe
135Xe
138Xe

4.3
1.2
1.2
0.7
12
0.8
2.4
220
0.2
40
1.8

Total
(rounded) 280

bone
Breast Red
marrow

5.3
1.3

5.7
2.2
2.0
1.0
16

Lungs

Thyroid

5.0

4.8
1.5
1.4
0.9
15
0.7
2.7
350
0.2
42
2.1
420

1. 3

0.9
15
0.8
2.6
270
0.2
43
2.2

4.2
560
0.3
64
2.4

1.6
0.9
22
0.6
2.4
250
0.2
42
2.1

350

660

330

1. 7

1.2

Tab 1 e£lone
lining
cells

Liver

Skin

Remainder
tissues

4.2
560
0.3
64
2.4

5.0
1. 3
l.6
0.9
' 15
0.6
2.4
250
0.2
42
2.1

9.8
6.4
410
3.9
41
47
44
2700
0.5
180
4.8

5.3
1.4
1. 7
0.9
15
0.9
2.8
310
0.2
46

660

320

3450

380

5.7
2.2
2.0
1.0
16
1.2

2.2

llormalized
collective
effectivedose
equivalentcolllTiitment-4(10 man rv
fGW(e) ar )
5.1
1.5
5 .7
0.9
16
1.3
3.2
340
0.2
48
2.2
420

29

Normalized collective effective dose equivalent commitment
for noble gases from the model PUR
as a function of distance

Distance(km)

llorma 1i zed co 11 ecti veeffective dose
equivalent commitment
r 10- 4 man rv(Gl/fe] a)- ]Percentage
distribution

0-2
2-5
5-10
10-20
20-50
50-100100-200
200-500
500-1000
1000-2000

5
12
18
27
40
55
114
122
20
6.3

1.2
2.9
4.2
6.5
9.6
13.2
27.2
29. '.)
4.8
1.5

Total
( rounded)

420

100

307

Ta b 1e30

Normalized local and regional collective dose commitments
For no61e gases from tfie moae1 Bim:
Norma 1i zed
collective
effectivedose
equivalent
Remainder conrnitment(man Svti ssucs
1 Gl/(e) ar 1 )Uonnalized collective absorbed dose commitment (mun Gy [GW(e) af 1)

Radionuclide
Gonadsbone
Breast Red
marrow

Lungs

Done
Thyroid liningcells

Li vcr

Skin

41Ar
85mKr
85Kr
87Kr
88Kr
13llllxe
133Xe
135ffixe
135xe
138Xe

0.003
0.02
0.007
0.10
0.69
0.001
0.06
0.13
0.32
0.21

0.004
0.02
0.01
0.13
0.87
0.001
0.08
0.02
0.35
0.26

0.005
0.04
0.01
0.14
0.92
0.001
0.16
0.02
0.52
0.28

0.004
0.02
0.009
0.12
1.2
0.001
0.07
0.02
0. 34
0.35

0.004
0.03
0.008
0.13
0.88
0.001
0.10
0.01
0. 34
0.25

0.005
0.04
0.01
0.14
0.92,
0.002
0.16
0.02
0.52
0.28

0.004
0.02
(J.009
0.12
0.83
0.001
0.07
0.02
0.34
0.25

0.008
0.11
2.5
0.56
2.3
0.02
0. 79
0.04
1. 5
0.55

0.004
0.03
0.01
0.13
0.87
0.001
0.09
0.02
0.37
0.26

0.004
0.03
0.034
0.13
0.90
0.001
0.10
0.02
0.39
0.26

Total
(rounded)

1.4

l. 7

2.2

2.2

1.8

2.1

1. 7

9.1

1.8

1.9

T a b 1e31

Normalized collective effective dose equivalent comrnitrnent
for noble ases from the model BWR
as a unction of distance1

Percentage
distribution(km)

Normalized collective
effective dose
equivalent commitment[man Sv (GW[e) a)- 1)

0-2
2-5
5-10
10-20
20-50
50-100100-200
200-500
500-1000
1000-2000

0.19
0.27
0.30
0.34
0.46
0.18
0.10
0.04
0.01
0.01

10
14
16
18
24
9.3
5.4
2.2
0.5
0.5

Total

1.9

Distance

Table

100

32

tiorrna l i zed loca 1 and re i ona 1 co 11 ecti ve absorbed dose commi trnents
rorn tr1 ,urn 1n air orne 1sc arges from the rno e reactor

Reactor

llWR
P\JR
GCR
HWR

308

Normalizeddischarge[TBq
cm,reJ a)- 1]
3.4

7.8
11

540

Normalized collective whole-bodyabsorbed dose commitment-I[man Gy (GU[e] a) J
Ingestion

Inhalation

Total

0.03
0.07
0.1
4.8

0.005
0.01
0.02
0.8

0.04
0.08
0.1
5.6

Tab l e 33-----Normalized local and re~onal collective absorbed dose commitmentsfrom ea on-14 releases to atmosphere

Reactorllormalized collective whole-bodyabsorbed dose commitment[man Gy (G\/[e] a)- 1 ]Hanna 1i zed
discharge[TBq(G\/[e] a)- 1]

Ingestion

Inhalation

Total

0.5
0.2
1.1
17

0.92
o. 39
2.2
30

0.001
0.00006
0.0003
0.005

0.92
0.39
2.2
30

B~JR
PWR
GCR
Hl/R

Ta b 1e34

Normalized local and regi ona 1 co 11 ecti ve dose commitments
for 1oiline releases from tfie model P1IR

Normalized collective absorbed dose commitment (10- 4 man Gy '.GU(e) ar 1 )Pathway
GonadsDirect cloudi rradi ati on
Inhalation
Ingestion ~/Externa 1 dosefrom grounddeposits
Resuspension
Total( rounded)bone
Breast Red
marrow

Lungs

Bone
Thyroid 1i ni ng
cel 1s

Liver

Skin

Remainder
tissues

Normalized
collective
effectivedose
equivalent
comrni tment(10- 4 man !V[GIJ(e) ar )
a/

0.04
0.01
0.006

0.05
0.04
0.02

0.06
0.03
0.02

0.05
0.42
0.02

0.04
120
84

0.06
0.03
0.02

0.05
0.02
0.009

0.11
0.03
0.01

0.02
0.04
0.02

0.05
3.7
2.6

0.18

0.22

0.32

0.13

0.19

0.35

0 .11

0.27

0.22

0.22

0.00020.07

0.00001

0.000009 0.00001

0.00002

0.002

0.45

0.19

0.32

6.6

0.000005 0.00002 0.00001
0.24

0.33

0.42

0.62

210

0.42

~/

Percentage contribution by iodine isotopes to collective effective dose equivalent commitment:
1311 96 %; 133! 2 %; 135! 2 %.
b/ Percentage contribution of each pathway to ingestion doses: milk 90 %; beef 5 %, green vegetables 5 %.

309

Ta b l e

35

Normalized local and re ional co 11 ecti ve
or 10Norma 1i zed
collective
effectivedose
equivalent
tment
Remainder conrni-4tissues(10 man ~v{Gll(e) ar )!I

llorma 1i zed co 11 ecti ve absorbed dose conrnitment {10- 4 man Gy [GW{e) aJ- 1)
Pathway
GonadsDirect cloud
irradiation
Inhalation
Ingestion~/
txtema l dosefrom grounddeposits
Resuspension
Total{rounded)bone
Breast Red
marrow Lungs

Thyroid

3.5
0.91
0.16

4.4
1. 7
0.55

5.1
1.5
0.38

3,8
4.2
32
3900
0.44 2000

7.2

9.0

12

6.8

7.8

0.00020.00050.0003

0.006

1.6

12

16

19

43

5900

Bone
lining
cells

Liver

5.1

4.2

1.4

I. 4

o. 36

Skin

0.22

9.8
1.2
0. 35

0.50

4.2
120
62

13

5.9

11

8. 7

8,7

0.00030.00030.00030.0005

0.05

20

12

22

15

190

4.4
1. 7

!/ Percentage contribution by iodine isotopes to collective effective dose equivalent conrnitment:
1311 78 %; 1321 0.001 %; 1331 17 %; 1351 5 %.
b/ Percentage contribution of each pathway to ingestion doses: milk 90 %; beef 5 %, green vegetables 5 %.

310

)

Ta t 1e36

Nonnalized local and regional collective dose commitmentsfor ~artlcula£e releases from t~e moael Plllt ana D~

Normalized co 11 ect i vc absorbed dose commitment (I0- 3 man Gy [GW(e) a]- 1}
Pathway
GonadsP WR
Direct cloud
irradiation
Inhalation
Ingestion~/
Externa 1 dosefrom grounddeposits
Resuspension
Total(rounded)

Lungs

ThyroidBonelining
cells

Liver

Skin

Remainder
tissues

0.001
0.04
2.1

0.002
0.06
1.9

0.002
0.41
14

0.002
1.4
1.9

0.002
0.05
!.9

0.002
0.82
28

0.002
0.11
2

0.003
0.04
1.9

0.002
0.06
1.9

0.002
0.27
4.4

5.9

7.7

8.3

7.2

6.4

8.6

6.6

9,6

7.3

7.3

0.001

0.002

0.02

0.04

0.002

0.04

0.004

0.001

0.002

0.009

8.1

9.7

23

11

8.4

37

8.9

11

9.2

12

0.04
1.4
47

0.05
10
350

0.04
35
45

0.04
l.3
48

0.05
20
680

0.04
2.8
52

0.07
0.91
36

0.04
46

0.04
6.6
110

190

200

180

160

210

160

240

180

180

B WR
Direct cloud
irradiation
0.03
Inhalation
0.93
Ingestion 2_/ 52
External dosefrom ground 150
deposits
Resuspension
0.03
Total
(rounded}bone
Breast Red
marrow

Norma 1i zed
collective
effectivedose
equivalentcommitment(I0- 3 man ~v[GW(e) af )

200

1.4

0.05

0.42

1.1

0.04

0.85

0.09

0.03

0.05

0.22

240

560

260

210

910

220

270

230

290

a/ The pathways contributing to the ingestion dose are: grain 30 %; green vegetables 25%; milk 15 %;milk products 10 %; root vegetables 10 ~: beef 5 %; others 5 %.

311

Ta b 1e37

Radionuclides considered for li uid effluent calculations
rom t e mo e reactor s 1 te, nonna 1 ze re eases anconcen tra tfon rac tors for mad ne and fresfiwa ter meihaNormalized
release[GBq/(GH[e] a);Radionuclide

P\JR

Bl~R

54Mn
58Co
60co
65zn
89Sr
90Sr
106Ru
llOmAg
125Sb
1311
134cs
137Cs
144CP

Moll- Sedi- Seauses
n-en ts weed

Sediments

Fish Crustacea

GCR

500 10000 10000
100 1000 1000JOO
!OOO 1000
2000 5000 100000
10
JO
lO
10
1
500 2000
1000 5000 50000
500
300
100
10
100
100
50
30
30
so
30
30
10 1000 1000

10
4.0
6.9 90
34
30
25
8.8
8.2
1.0
0.33 0.2 420
34

1.2
80
9
9 392
16 2600

13
46
60(m3 t-1)

Freshwater
concentration
factor(rn3 t -1)

Marine concentration factor

so

Table

Fish

10000 10000 10000 300
10000 !OOO 30000 300
10000 !OOO 30000 300
10000 1000 1000 1000
500
10 2000
30
500
JO 2000
30
10000 2000 40000
1010000 !OOO
3
200
10000
JOO
300 1000
100 1000
200
30
500
30 30000 1000
SOO
30 30000 1000
10000 1000 30000 1000

38

Parameters for the river model used to assess reactor releases

SectionWater
velocity(m s -l)Sediment
velocity-1(m s )

Suspended
sedimentload-3(g m )

Drinking
water
extraction(m 3 a -1 )Fishproduction(t a- 1 )

I
2
3

0.8
0.75
0.7

0.00020.000150.000140
45

100000140000
31000

JOO
66
600

so

Ta b 1e39

Normalized collective dose commitments from li9uid reactor effluents
aiscnargeo into tne mooe1 river

Normalized collective absorbed dose commitment(10- 4 man Gy [GW(e)ar 1)

Reactor type
and pathway
Gonadsbone
Breast Red
marrow

Lungs

ThyroidBonelining Remainder
tissuescells

Norma 1ized
co 11 ecti veeffectivedose
equivalentco1T111i tment(I0- 4 man ~v
fG~/(e) af )PW R

Drinking water
1311
58CD
6Dco
90sr
134cs
J 37cs
Total
312

1.4
0.009
1.2
J.6

0.02
0.33
0.70
0.009
1.1
1.5

0.02
0.39
0.82
1.2
1.2
1.5

0.02
0.29
0.56
0.009
1.1
1.5

94
0.27
0.49
0.009
I. 2

0.02
0.29
0.59
2.6

0.79
3.1

I. I

I. 3

s

1.5

1.4

2.8
0.53
1.6
0.21
I. 2
1.5

4.8

3.7

5.1

3.5

97

6. I

6.5

7.8

0.008

o.sa

I.

Table 39, continued

Normalized collective absorbed dose commitment(J0- 4 man Gy [GW(e)ar 1 )Reactor type
and pathway
Gonadsbone
Breast Red
marrow

Lungs

ThyroidBoneliningeel ls

58Co
60co
90sr
134cs
137Cs

0.00060.11
0.26
0.0003
o. 76
1.0

0.002
0.06
0.13
0.0003
0.68
0.93

0.001
0.07
0.15
0.05
0.76
0.93

0.002
0.06
0.11
0.0003
0.68
0. 93

6.7
0.05
0.09
0.0003
0.72
0.93'

0.001
0.06
0.11
0.10
0.68
0.93

Total

2.1

1.8

2.0

1.8

8.5

1.9

Fish
1311

Remainder
tissues

Normalized
collective
effectivedose
equivalentconrnitment-4(10 man_~v[GW(e) a) )

0.81
0.91

0.20
0.10
0.30
0.008
0.75
0.94

2.4

2.3

0.14
0.59

External
58co
60co
134cs
137cs

0.00140.033
0.00230.0079to all organs and tissues
to all organs and tissues
to al 1 organs and tissues
to a11 organs and tissues

0.00140.033
0.00230.0079

Total

0.045

to all organs and tissues

0.045

B WR

Drinking water
131 I
60Co
65zn
89Sr
90sr
134cs
137cs

5.2
5.2

130
0.46
0.88
0.06
1. 3
5.4
5.2

0.02
0.55
1.4
1.1
3.9
5.2
5.2

5.3
5.2

4.0
1.6
1.1
0.49
0.34
5.4
5.6

16

12

140

18

18

19

0.002
0.12
1.2
0.002
0.0005
3.2
3.3

0.002
0.14
1. 7
0.03
0.07
3.6
3.3

o.~02
0.10
1.2
0.002
0.0005
3.2
3.3

10
0.10
1.1
0.002
0.0005
3.4
3.3

0.002
0.10
1. 7
0.04
0.15
3.2
3.2

7.5

8.8

7.5

18

8.1

0.01
1.3
1.0
0.06
0.01
5.8
5.6

0.03
0.63
1.0
0.06
0.01
5.2
5.2

0.02
0.75
1.4
0.75
1.9
5.8
5.2

0.03
0. 53
0.88
0.06

14

12

Fish
1311
60co
65zn
89Sr
90Sr
134cs
137cs

0.08
0.24
1.2
0.002
0.0005
3.6
3.5

Total

8.8

External
60co
134cs
137cs

0.033
0.012
0.030

to all organs and tissues
to a11 organs and tissues
to all organs and tissues

0.033
0.012
0.030

Tota I

0.075

to al 1 organs and tissues

0.075

Total

1. 3

3.3
1.2
1.1

0.62
1.6
0.05

0.29
0.30
1.4

4.7
3.8

0.02
0.01
3.8
3.5

11

9.3

313

Ta b 1e40

Norma 1i zed co 11 ect i ve dose corrrnitments for the mode 1 PWR, BWR and GCR
on a notional slte in t~e eastern Englls~ C~annel

Norma 1i zed co 11 ec t i ve absorbed dose commitment(10- 4 man Gy (GW(e) a]- 1)

Reactor type
and pathway

Lungs

Bone
Thyroid lining
cells

Upper
Lower
large
large
intestine intestine

0.06
0.05
1.4
0.05
3.6
7.6

0.30
0.10
I, 7
0.13
3,4
7.6

o. 72

0.25
3.8
7.6

0.11
0.071
1.6
0.22
3.4
7.6

13

15

13

13

134cs
137cs

0.29
0.28
1.5
0.11
0.09
0.17

0.14
8.6
0.73
0.05
0.08
0.16

0.25
0.12
0.87
0.06
0.09
0.16

0.11
0.04
0.59
0.05
0.08
0.16

Total

2.4

1.2

1.5

Mo 11 uses
54Mn
58co
60co
llOmAg
!34cs
137cs

2.5
2.8
11
9.4
0.70
1.1

1,2
0.89
5.3
4.2
0.63
1.0

Total

28

Grand total

47

Gonadsbone
Breast Red
marrow

0.14
0.15
2.0
0.20
3.4
7.6

0.25
0.21

134cs
137cs

0.29
0.48
4.0
0.45
3.8
8.2

Total

17

P WR
Fish
54Mn
58co
60co
l lOmAg

Crustacea
54Mn
58co
60co
l lOmAg

314

Norma 1i zed
co1'ective
effectivedose
equivalentconrni tmen t(10- 4 man ~~!(lW(e) ar )

1.6
4.2
8.8

1.1
3.3
20
2.9
4.4
8.8

13

28

41

19

0.06
0.03
0.51
0.01
0.09
0.15

0.30
0.06
0.62
0.03
0.08
0.16

0.72
0.96
3.9
o. 37
0.10
0.18

1.1
1.9
7.3
0.68
0.10
0.18

0.37
0.37
1.8
0.18
9.5
0.17

1.0

0.86

1.2

6;2

11

3.0

2 .1
1.2
6.3
5.1
0.70
1.0

0.98
0.42
4.3
4.6
0.63
1.0

0.49
0.32
3. 7
1.0
0.67
1.0

2.6
0.59
4.5
2.7
0.63
1.0

6.2
9,9
28
33
0.78
1.2

9.8
23
53
61
0.81
1.2

3.2
3,8
13
16
0.74
1.1

13

16

12

7.2

12

79

150

38

28

32

26

21

26

110

200

60

2.4

I. 7
II

0.37
0.64
4.9

0.77
4.0

8.2

Table 40, continued

Nonnalized collective absorbed dose commitment(10- 4 man Gy [GW(e) a]- 1)

Reactor type
and pathway

ThyroidBoneliningcel 1s

Upper
Lower
large
large
intestine intestine

0.25
1.5
12
17
29

0. 12
1.3
12
18
29

0.65
1.5
18
17
29,

1.6
9.7
16

69

60

60

0.30
0.65
1.2
0.42
0.59

0.54
0. 77
1.6
0.47
0.59

0.25
0.53
1.0
0.42
0.59

4 .1

3. l

4.0

Moll uses
54Mn
60co
65zn
l 34cs
137Cs

5.4
9.8
280
3.6
4.1

2.6
4.9
270
3.2
3.8

Total

300

Grand tota 1 370

Gonadsbone
Breast Red
marrow

0.62
3.7
13
19
31

0.30
1.8
13
17
29

0.54
2.2
18
19
29

67

61

Crustacea
54Mn
60co
65zn
134cs
l 37Cs

0.62
1.3
1.1
0.47
0.64

Total

B WR
Fish
54Mn
60co
65zn
134cs
l 37Cs
Total

Lungs

2.5

Nonnalized
collective
effectivedose
equivalentcorrmi tment(I0- 4 man ~v[GW(e) ar )

33

19
22
33

0.81
4.5
15
20
31

66

81

95

71

0.12
0.46
1.0
0.44
0.59

0.65
0.55
1.6
0.42
0.59

l .6
3.4
1.4
0.51
0.69

2.5

6.5
1.6
0.54
0.69

0.81
1.6
1.3

2.8

2.6

3.8

7.6

12

4.8

4.7
5.8
390
3.6
3.8

2. l
4.0
250
3.2
3.8

1.1
3.4
250
3.4
3.8

5.6
4.1
390
3.2
3.8

13
26
340
4.0
4.4

21
49
400
4.2
4.4

7.0
12
320
3.8
4.1

280

410

260

260

410

390

480

350

350

480

330

320

480

480

580

420

21

18

0.49

0.64

315

Table 40, continued

Nonnalized collective absorbed dose commitment(10- 4 man Gy fGW(e) a]- 1)

Reactor type
and pathway
Gonads
GC R
Fish
60co
90Sr
106Ru
125Sb
134cs
137Cs
144ce
Total

Lungs

ThyroidBone
1iningeel ls

1.3
l. 1
3.0
1.5
1.8
1.2
0.39
0.39
55
0.39
0.39
120
0.01
0.01
0.01
0.01
0.01
0.02
2.1
4.6
1.3
0.96
20
6 .1
140
160
150
160
140
140
1200
1200
1200
1200
1200
1300
0.0006 0.0005 0.13
0.004
0.001
0.009
1500

1300

1400

1300

1400

1500

Upper
Lower
large
large
intestine i nte5ti ne

7.9
1. 7
0.23
48
180
1400
2.2

15
6 .1
0.66
130
190
1400
6.4

3,7
10
5.4
14
170
1300
0.51

1600

1700

1500

Crustacea
60co
90Sr
106Ru
1255b
134cs
137Cs
144ce

1.1
0.14
0.33
0.15
4.0
27
0.02

5.3
0.14
0.29
0.51
3.6
25
0.006

0.63
20
0.31
8.11
4.0
25
0.04

0.43
0.14
0.29
0.03
3.6
25
0.003

0.37
0.14
0.29
0.02
3.8
25
0.002

0.45
42
0.29
0.49
3.6
25
0.06

2.8
0.62
5.2
1.2
4.4
29
11

5.3
2.2
15
3.0
4.6
29
31

1.3
3.6
1.2
0.34
4.2
27
2.5

Total

33

34

50

29

30

72

54

90

39

Molluscs
60co
90Sr
106Ru
125Sb
l34cs
137Cs
144Ce

8.1
0.91
12
0.37
30
180
0.20

4.1
0.91
10
0.13
27
170
0.06

4.8
130
0.28
30
170
0.45

3.3
0.91
10
0.08
27
170
0.03

2.9
0.91
10
0.06
29
170
0.03

3.4
260
10
1.2
27
170
0.64

0.00002 41
3.9
14
190
530
3.0
7.7
34
35
190
190
110
320

230

210

350

210

210

470

530

·1100

310

Grand tota 1 1700

1600

1800

1600

1600

2000

2200

3000

1800

Total

316

bone
Breast Red
marrow

Nonna 1i zed
co 11 ect i ve
effectivedose
equivalentcommitment(10- 4 man fV[GW(e) ar )

11

10
23
43
0.86
32
180
26

Table41

Activity released (TBq)
Year

Particulate
releaseEnergy(G~l(e) il)

311

14c

05Kr

Total a0.00280.00190.00100.000890.0010

Isotopic composition of particulate release(total 13)
131 I

134cs

~/lNDSCALE (United Kingdom)
0.070
0.13
0.041
0.28
0.0059 0.0026
0.048 0.0078 0.0026 0.015 0.027
0.34
0.009 0.003 0.006 0.03
0.31
0.025

0.23
0.24
0.24

90Sr

1975
3.2
1976
3.2
1977
2 .1
1978
1.8
1979
2.5
1975-1979
Normalized rele,sefTBq (GWfe) a)· J

444
444
296
222
290

4. l
3.5

44000
44000
33000
26000
35000

133

1.8

14000

0.0006

1975
1.6
1976
0.84
1977
1.8
1978
2 .1
1975-1979
Nonnalized rel~fse[TBq (GW[e] a) ]

3.3
1.8
11
4.1

24000
13000
25000
29000

LA HAGUE (France)-10
5.9 10_10 0.0006
7.4 10_7
0.00030.0001
1.1 10_6
0.0001
1.1 103.2

14000 1.9 10- 1

1975
0.24
1976
0.22
1977
0.32
1978
0.83
1975-1978
Normalized rel~fse[TBq (GW[e] a) ]

4.4
4.4
2.8
63

3700
3500
4400
11000

46

14000 4.0 10- 6

106Ru

1291

!34cs

Total 130.088

0.0002MARCOULE (France)-7
8.1 10_7
0.001
0.001
6.7 10_6
3.2 10 6
0.0002o.ooD3
1.1 100.002

317

Ta b l e

42

Radi onuc l ides released to the a uatic environmentrom ue re rocess ng
2, 822, L1 , L6,
Activity released (TlJq)
Year

Tota 1 ~Total a(othe~
than 11)

311

90Sr

106Ru

llltlDSCALE (United Kingdom)
1400
466
9065
762
6771
1200
381
766
427
7129
910
816
7124
1050
597
810
4100
1200
250
390

1975
85
1976
60
46
1977
1978
68
1979
62
1975-1979
Uonnalized rel~pe[TBq (G~[e] a) l
25
2671

450

166

LA HAGUE (France)
75
1975
0.49
1180
40
1976
0.37
714
73
331
1977
0.67
765
140
1978
0.51
1092
728
1975-1978llonnalized releise[TBq (Gl/[e] a)- ]
52
0.32
592
272
IIARCOULE (France)
1975
0.019
42
1976
23
0.011
30
117
1977
0.013
35
1978
270
0.013
1975-1978
llormal ized rel~ise[TBq (GW[e] a) )
337
0.035
81

318

277
829
555
540
801

430

0.93
0.41
0.37
0.88

33
20
27
25

1.6

65

1 a b 1e

43

Isotopic corn osition of 11 uid effluentsfrom t e 10 sea e reerocess1n9 e ant
rn2, i322]
Annual di!.charge (TBq)
Radionuclide
1977
35 5
s4r-,n
55Fe
60co
63fli
65In
89Sr
90Sr
95zr
951lb
99Tc
103Ru
106Ru
1lOmAg
125Sb
129 r
134cs
!37Cs
l44ce
152Eu
154Eu
155Eu
Uranium
237(1p238Pu
239+240Pu
241Pu
24lf11;i
242cm
243+244Cr.,

427
92
203

816

0.11
594
4480
152

36
981
3.6

Electricity production
fror., fuel reprocessed
2 .1[GU(e) a]

1978

1979

1.8
0.22
2.3
1.0
1.5
0.037
9.9
597
82

1.4
0.037
1.8
0.52
0.18
0.022
7.5
250
60
98
43
5.8
390
0.033
14
0.12
240
2600
83
3.7
1.9
4.0
6000 (kg)
0.33
12
37
1500
7.9
0.37
0.15

148

179
8.5
810
0.33
29
0.074
404

4090
104
10
38
7.8
10936 (kg)
0.60
12
46
1773
7.9
0.55
0.33

1.8

2.5

319

Po ulation distribution around the l·Jindscale re rocessing
an meteoro og ea c aracter sties o its ocat1onlant

METEOROLOG !CAL CHARACTER !STICS

Pasqui 11 weather ea tegory
Quantity0 r yFrequency (%) _liind speed (m s 1 )Depth of mixing
1ayer (m)

R a i

A

0

C

D

E

F

0.3
1

4.5
2

12.l
5

63

s

5
3

4.6
I

s

2000

2000

1000

1000

200

200

C

D

1. 7

8.4

1000

1000

Uistribution of wind direction frequencies (30° sectors):
0.091, 0.089, 0.096, 0.11, 0.098, 0.069, 0.049, 0.034, 0.056,
0.10, 0.11, 0.094
Stack height: 100 m (effective)
PO PU LAT JOU DI STR IBUT I ON

Distance (km)

O- l

- 2.5

2.5 - 6

6 - 12,5

12.5-17.S

Population
Cumulative

36
35

230
220

1100
1400

12000
14000

17000
31000

Distance (km}17.5-60

60-150150-375375-900

900-2700

Population
Cumulative

71000
74000

20000000
21000000

36000000
57000000

Ta b l e

110000000 82000000
170000000 260000000

45

Averaged local and regional normalized
whole-body absorbed dose coll'fllitments from tritium and carbon-14
discharges to the atmosphere from llindscale (United Kingdom)Radionuclide320

tlorma 1i zed
collective absorbed whole-bodydose commitment
rman Gy (Gll'e] a)- 1 ]Ingestion

Inhalation

Total

0. 30
0.69

0.049
0.0001

0. 35

0.69

n

s

46

Ta b l e

from Windscale United Kin dom

Collective dose commitmentsCollecthe absorbed dose co1T111itment (man Gy}

Collective
effectivedose
equivalentco1T111itment
man Sv

GonadsBreastRed bonemarrow

Lungs

Thyroid!lone
lining
cells

Liver

Skin

CLOUD
85Kr

0.26

0.058

0.078

0.095

0.074

0.065

0.095

0.074

19

0.078

DEPOSIT
137Cs

1.1

0.87

1.2

1.4

1.1

0.97

1.4

1.0

1.4

1.1

0.3
0.13
0.000740.000680.000095
0.000220.13
0.00430.00440.00560.0015

0.13
0.02
0.018
0.00052
0.0071

0.13

0.13
0.059
0.054
0.075
0.019

0.13
0.012
0.011
0.016
0.0041

0.13

0.13

137cs

0.3
0.11
0.10
0.08
0.04
0.007
0.006

0.005

0.00540.0041

0.005

Total

0.47

INGESTION
3H
134Cs
137cs
14c
90Sr
1291

0.78
0.13
1.1
1.8
0.24
o. 34

0.78
0.11
0.98
1.8
2.8

0.78
0.13
1.1
1.8
0.0096

0.78
0. 11
0.98
1.8
0.0096

0.78
0.086
0.76
1.8
0.0096

Total

4.4

GRAND TOTAL

6.2

Pathway

INHALATION
3H
239Pu
240Pu
241Am
238Pu
129 1

0.006

0.005

0.006

0.0056

0.24
0.005

0.78
0.13
1.1
1.8

0.78
0.11
0.98
1.8

0.78
0.13
0.98
1.8
1.3

0.78
0.11
0.98
1.8

0.78
0.11
0.98
1.8

Remainder
tissues11321

T a b l e

47

Collective absorbed
into t e
an into t e

and 1979

Collective absorbed dose commitment (man Gy)
Pathway

GonadsBreastRed bonemarrow

Lungs

ThyroidBoneliningeel lsLiver

Lower large
intestine

250

270

WINDSCALE (United Kingdom)
Fish
137Cs

270

Mollusc and crustacea
106Ru
Fish, crustacea, mollusc
90Sr
Mollusc
239 Pu and 240 Pu
241Pu
Total

av
a

8.7

250
7.4

250
8.1

25u

250

7.4

250

7.4

380

7.4

25

54

0.16
2.7

2
36

0.42
5.9

0.50
2.3

260

350
2

260
0.42

650
0.50

Collective effective dose equivalent commitment:

311 man Sv

8.8

9.5

0.024
0.46
280
0.024

260

290
0.16

260

LA HAGUE (France)
Fish
!37Cs
Mollusc and crustacea
106Ru
Mollusc
239Pu and 240Pu
241Pu
Tota 1

6Y
a

9.5

8.8

8.8

8.8

8.8

8.8

20

17

19

17

17

17

0.005
30
0.005

0.036
26

28
0.036

26

26

Collective effective dose equivalent commitment:

322

880

0.42

0.09

0.11

26
0.42

11
0.09

890
0.11

84 man Sv

Ta b l e

48

Local and reor nonna

Normalized
releaseRadionuclide[TBq (GW!e; a)-ll
ATMOSPHERIC
JH
14c
85Kr
90Sr
1291
238,239 ,240Pu

Normalized collective
effective dose
equivalent co11111itrnent[man Sv (GWfe) a)- 1]

60
0.4
10000
0.00070.00020.00001

0.1
0.2
0.03
0.006
0.004
0.002
Total

AQUATIC
l37Cs
l06Ru
90sr
1291
238.239 .240Pu

7
10

0.3

0.4
0.3
0.02
0.008
0.0006

2

0.04
0.5
Total

Ta b l e

0.749

Representative annual effective dose equivalents
to most exposed ind1v1duals
from reprocessing LIJR fuel at the model facility

Radionuclide

Annual effective
dose equivalent(uSv)

ATI-10SPHERIC
311
t4c
85Kr
90Sr
1291

5
10

s
2

2

a emitters

Total

AQUATIC
90sr
106Ru
137cs
1291
a ernitters

25
2
50
150
I

0.5

Total (rounded) 200

323

Tab l e50
s i nifi canceequ va en co111111 ents

Radionuclide

3H
85Kr
14c
1291

florma l i zed
collective effective dose equivilent
cOOlllitment 'man Sv (GIi'. e] a f ]ilorma l i zed
re leaserrnq(Gl:' e: a) - l,Integration t i1;1e (a}
104
106
10 2

101

640
11000
17
0.04

0.015
0.9
3

0.02
1.9
10
0.02

Tab l e

0.02
1.9
70
0.2

0.02
1.9
110
23

108
0.02
1.9
110
560

51

Collective effective dose e9uivalent commitmentsfrom high-level waste disposal per unit electrical energy generated
for different fuel cycles! I11 ~
Normalized collective effective dos~ 1equivalentcorrmitment !man Sv (GWe) a) J
LUR
DISPOSAL OF UNREPROCESSED FUEL
Uranium content of fueltligration time 106 a
62
Migration time 10 5 a
71
All other elements
Migration time 106 a
27
Migration time 105 a
200
89
Total {rounded) {106 a)
270( 105 a)
ll/R

h'WR

300
170
10
280
310
450

DISPOSAL OF REPROCESSING HASTESUranium content of fuelMigration time 106 a
22
Migration time 105 a
23
All other elements
Migration time 106 a
9Migration time 105 a
27
Total {rounded) {106 a)
31{105 a)
so

324

HUR

IITR

Uranium
Uranium
Plutonium Thorium

UraniumThorium

FBR(Plutonium
recycle)

0.8
1.1

130
72

22
120

18
51

6
45

14
140

24
110

19
52

140
120

36
260

25
120

1.1

6.0

Ta b l e

52

Summar of normalized collective effective dose e uivalent corn:nitments
tote pu 1c rom nuc ear power pro uct1on
Normalized collective effective
dose equivalent COl111!1itment[man Sv (GW[e] a) ]LOCAL AND REGIONAL CONTRIBUTION
MiningRadon
Milling (excluding releases from tailings)
Uranium, thorium, radium
Radon
Fuel fabrication
Uranium
Reactor releases
Atmospheric- Noble gases- Tritium- Carbon-14- lodi nes- Particulates (caesium, ruthenium,
cobalt)

0.5
0.015
0.02
0.002

Total

0.5

Total

0.04

Total

0.002

Total

4.1

Total

0.06

Total

0.3

Total
Total

0.7
0.003

0.6 •
0.5
2.8
0. 06
0.1

Aquatic- Tritium- Other (caesium, ruthenium, cobalt)
Fuel reprocessing
Atmospheric- Tritium- Krypton-85- Carbon-14- a emitters

0.04
0.02

0.1
0.03
0.2
0.002

Aquatic- Caesium 134, 137- Ruthenium-106- Strontium-go- a emitters- lodine-129

0.4
0.3
0.02
0.0006
0.008

Transportation

0.003

5. 7 man Sv (GU(e) ar 1

TOTAL for operations in the nuclear fuel cycle

Complete and incomplete normalized
collective effective dose equivalentcommitment[man Sv (GW(e] a)- 1]
10 1
GLOBAL CONTRIBUTION
FROM OPERATIONS IN THE
NUCLEAR FUEL CYCLE
Tritium
Krypton-SS
Carbon-14
Iodine-129
TOTAL (rounded)
WASTE DISPOSAL(assuming LWR fuelreprocessed and plutonium
utilization in LWRs and FBRs)
Mill tailings (radon) a/(uranium) o/
High-level wastes
a/
b/

0.015
0.9

Integration period (a)
2
4
10
10
106

10 8

0.02
1.9
10
0.02

0.02
1.9
70
0.2

0.02
l. 9
110
28

110
560

3.9

12

72

140

670

0.25

0.25

250
460

2800
460
30

2800
460
30

3

0.02
1.9

Assuming radon emanation continued at the same rate as at disposal
with 2 m earth cove ing.
Assuming that at 10 3 a the tailings are eroded into fresh water,then marine environments.

325

T a b l e

53

Annual per caput doses frOG'I the continued generationof nuclear electric power to the year 2500
Year

Annual projected nuclear generation (GW(c) a]
Annual collective effective
dose equivalent {man Sv)
World population
Annual per caput
effective dose equivalent (µSv)
Percentage of average exposure
to natural sources of radiation (%)

326

1980

2000

2100

80

1000

10000 10000

500
4 10

9

10000 200000 250000
1010 1010
1010

0.1
0.005

2500

20
0.05

25

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NJO
N 11
N 12
N 13
N 14

NIS

Nl6

N17

N18

N19
N20
N21
01
PI

P2
P3

P4
PS

P6

P7

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