SOURCES, EFFECTS
AND RISKSOF IONIZING RADIATION
United Nations Scientific Committee on the Effectsof Atomic Radiation
1988 Report to the General Assembly, with annexes

UNITED NATIONSNew York, 1988

NOTE
The report of the Committee without its annexes appears as Official Records of theGeneral Assembly, Forty-third Session, Supplement No. 45 (A/43/45).
The designations employed and the presentation of material in this publication d o
not imply the expression of any opinion whatsoever on the part of the Secretariat of theUnited Nations concerning the legal status of any country, territory, city or area. or of its
authorities, or concerning the delimitation of its frontiers o r boundaries.

UNITED NATIONS PUBLICATION
Sales No. E.88.IX.7
ISBN 92-1- 142143-8
09000PANNEX B
Exposures from nuclear power production

CONTENTSParagraphs

Paragraphs

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

1-20

1. .MINING AND MILLING ........

. Effluents ....................

21-37
22-30

B . Local and regional collective dose
commitments ................

3 1-32

C . Occupational exposures .......

33-37

I1. URANILhI FUEL FABRICATION

38-49

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

42-44

C. Occupational exposures .......

47-49

A

A.

Ill . REACTOR OPERATION .......
.A . Effluents ...................
1 . Fission noble gases .......
2 . ?ictivation gases ..........
.;. Tritium .................
4 . Carbon-I4 ..............
5 . Iodine ..................
6 . Particulates in airborne
effluents ................
. Liquid effluents ..........

-

B . Local and regional collective dose
commitments ................
I . Fission noble gases . . . . . . .
2 . .Activation gases . . . . . . . . . .
3 . Tritium .................
4. Carbon-I4 . . . . . . . . . . . . . .
5. Iodine ..................
6. Particulates in airborne
effluents ................
7 . Liquid effluents ..........
C . Occupational exposures .......

45-16

I

...........
D . Solid ~ a s t disposale
1 . Solid waste production ....
2. Solid waste disposal facilities .....................
3. Collective dose commitments ...................

122- 145
122- 128

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

146-178
148-161129-135
136-145

B . Local and regional collective dose
commitments ................ 162- 170
I . Krypton-85 ..............
163
2. Tritium and carbon-I? .... 164-165
3. Other atmospheric releases
166
4 . Liquid effluents .......... 167- 170
C . Occupational exposures ....... 17 1- 172
D . Solid waste disposal ........... 173-178
V . COLLECTIVE DOSE COMMITSIENTS FROM GLOBALLY DISPERSED RADIONUCLIDES .....

179-190.\ . Krypton-85 ..................

182

B . Tritium .....................

183-184

C . Carbon-14 ..................

185-187

D . Iodine-129..................

188-190

VI . TRANSPORT ...................

191-193

VII . SUMMARY ....................

194-198
Pages

Tables ...............................

163

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

234

Introduction
1. The generation of electric energy by nuclear
reactors has increased since the Committee's assessment of doses from radioactive materials released
during nuclear fuel cycle operations, as reported in
Annex F of the UNSCEAR 1982 Report [UI]. The
total world installed nuclear electricity generating
capacity at the end of 1987 was 298 G W from 417 units
in 26 countries [Ill. This represents an approximate
doubling of nuclear capacity since the UNSCEAR
1982 Report. as rnay be seen from Figure I. Kuclear
power was responsible for some 16% of the tvorld's
electricity generated in 1987, and currently some 120
reactors are under construction with an electrical
capacity of 101 G W [Ill. Projections for world
nuclear generating capacity for the year 2000 are still
somewhat speculative, but the figure seems likely to bein the range of 400-500 G W [12], somewhat less thanearlier expectations but still representing a further
expansion of 30-60% from currently installed capacity.

2. The number of power reactors operating at theend of 1987, their type and generating capacities foreach country of the tvorid is shown in Table 1. The
reactor types include the pressurized water moderated
and cooled reactor (PWR). the boiling tvater moderated
and cooled reactor (BWR), the gas cooled reactors( G C R ) of the Magnox a n d advanced gas cooled(AGR), graphite moderated type, the light tiratercooled graphite moderated reactor (LWGR), theheavy water moderated and cooled reactor (HWR),

and the fast breeder reactor (FBR). The installedcapacity per caput is also given in Table 1: it is highest
in Sweden at 1.14 kW per caput and ranges fromabout 0. I to over 0.8 kW per caput in other developed
countries. The average installed capacity per person atabout 0.14 kW represents an increase of 100Ci; overthe equivalent figure (0.07 kW) reported in the
UNSCEAR 1982 Report. Table 3 shows the amounts
and percentage of electricity generated in countries by
nuclear power in 1987 [I2]. The highest use of nuclear
reactors for electricity generation was in France (70Si )and Belgium (66%).
3. The nuclear fuel cycle includes the mining andmilling of uraniunl ores. conversion to nuclear ft~elmaterial, which usually includes the enrichment of theisotopic content of 235U and fabrication of fuelelements; the production of energy in the nuclear
reactor; the storage of irradiated fuel, or its reproccssing with the recycling of the fissile and fertile materials
recovered, and the storage and disposal of radioactive
wastes. Almost all of the artificial radionuclides
associated with the nuclear fuel cycle are present in
the irradiated nuclear fuel, although some neutron
activation of structural and cladding materials takes
place. The majority of irradiated fuel elements are
currently stored; when reprocessing takes place, thehighly active liquid wastes containing fission products
and transuranium elements are stored in tanks isolated
from the environment until the) can be solidified.
Solid wastes. arising at each stage of the fuel cycle. aremainly stored, although some wastes are disposed of.

YEAR
Flgure I.

The installed nuclear electric energy capaclty on 31 December
between 1979and 1987.
[Il, 12, 13, 14, 15, 16, U l ]In routine operation of nuclear installations, small
quantities of radioactive materials are released in
effluents. which disperse in the environment and resultin low-level exposures of the public.
The interest of the Committee is in assessing theradiation doses t o individual members of the publicfrom releases of radioactive materials and also thedoses to workers from normal operation of thenuclear fuel cycle. Exposures of the public from highlevel wastes, which arise in fuel reprocessing, have notbeen assessed by the Committee. as these wastes arestill in storage. The majority of irradiated fuel is not
being reprocessed. Preliminary estimates are made ofthe exposures in the future resulting from current
disposals of radioactive solid wastes. The significant
release of radioactive materials and the exposures toworkers and the public that resulted from the accident
at the Chernobyl nuclear power reactor are discussed
in detail in Annex D, "Exposures from the Chernobyl
accident", and Annex G "Early effects in man of highdoses of radiation".
4.

The quantities of radionuclides in effluents from
nuclear facilities are usually reported and available to
the Committee. reflecting the operational history ofeach plant, including periods of abnormal operationand maintenance shut-down. In this Annex the Committee reviews discharge data for the six-year period
1980-1985 and estimates average releases per unit ofelectric energy generated for each major power reactor
type. Because the data for 1985 are incomplete.
normalized releases are presented for the quinquennium
1980-1984. These normalized releases d o not apply. ofcourse, to any one plant but are deemed to be
representative of current nuclear power generation.
Future practices may lead to discharge levels considerably different from the normalized values presented
here, which include new and old plants: therefore, any
extrapolation to the future must be undertaken withcaution.
5.

15. Because of the system of controls applied toenvironmental releases from nuclear power installations. doses to individual members of the public
correspond to low levels of individual risk. The doses
to the most exposed individuals vary widely frominstallation to installation and from one location toanother, and the level of individual dose generally
decreases rapidly with distance from a given source. In
this Annex a n indication is given of the range ofindividual doses associated with each type of installation. T o evaluate the total impact of radionuclides
released at each stage of the fuel cycle, results arepresented in terms of the collective effective dose
equivalent commitment per unit quantity of electricenergy produced, expressed as man Sv per G W a.

7. The collective dose commitment from nuclear
power production is considered in four populationgroups: the occupationally exposed; the local population, being those within about 100 kilometres of thesite; the regional population, those within about a
1,000 kilometres of the site; a n d the remaining world
population. Each stage of the nuclear fuel cycle istreated separately. and the occupational, local andregional dose commitments are evaluated. The contributions from nuclides that, because of a combinationof long radioactive half-lives and rapid dispersal in theenvironment, become globally dispersed and irradiate
the world population are then discussed for the fuelcycle as a whole.
8. Collective dose commitments to local and regional
populations must be estimated by environmental
n~odelling, as the activity concentrations resulting
from effluents from nuclear fuel cycle operations arevery low both in environmental samples and in thegeneral population. Monitoring of activity concentrations due to effluent releases has concentrated on
areas immediately surrounding nuclear facilities toensure compliance with relevant regulations. T o estimate collective dose commitments it was decided in
the UNSCEAR 1987- Report to establish a model
facility at a representative site for each stage of thefuel cycle; mining and milling, fuel fabrication. reactor
operation and reprocessing. The envirorlnlent receiving
the normalized releases from each model facility waschosen t o represent broad averages containing typicalfeatures of existing sites and reflecting the mostcommon environn~ental pathways. Such generalizations gave dose commitments indicative of the impactof the overall nuclear power programme though notapplicable to any one site. In the UNSCEAR 1982
Report, the collective doses were evaluated for reported
discharges at the three operating commercial reprocessing plants at Sellafield in the United Kingdom andCap de la Hague and blarcoule in France.
9. The methods used by the Committee for estimating
the disoersion of radionuclides released to the atmosphere o r hydrosphere and the resulting doses to individuals were described in Annex A of the UNSCEAR
1982 Report. The Committee considers that, in general.
these methods and the model facilities and representative sites used in the UNSCEAR 1982 Report are
still valid for assessing the current impact of discharges
from the fuel cycle. Therefore, in this Annex, thecollective effective dose equivalent commitments are
obtained by scaling the dosimetric results from the
UN SCEAR 1982 Report, allowing for different releasesof the various radionuclides involved. The Committee
has decided to treat the reprocessing contributiondifferently in this Report. The hypothetical model
facility is not used, but rather, in order to reflect theactual dose contributions made. the normalized dose
commitments from the fraction of fuel reprocessed is
added to the contributions from the rest of the fuelcycle.
10. Very long-lived nuclides pose a special problem.
One example is(half-life: 1.6 10' a), while anotheris radon gas, which emanates from mill tailings
containing I3OTh (half-life: 8 10' a) and '.''U (half-life:
4.5 lo9 a). Assessments of human exposures over such
periods of time are clearly hypothetical and therelevance of the results is doubtful. Dose commitments
assessed for the purpose of calculating maximum dose
rates in the future involve integration over the periodof practice leading to the release of the radioactive
material. This approach is taken in this Annex for
effluents. For the solid waste disposal assessment, it isin general only possible to assess the collective
effective dose equivalent commitment.

exposures throughout the nuclear fuel cycle arebrought up t o date.

11. There have been a number of attempts togenerate rigorous definitions of the waste categories
generally referred to as low-, intermediate- and highlevel wastes [I 131. Although precise definitions havebeen agreed for particular purposes, the schemes proposed have not been universally satisfactory. None theless, the general characteristics of the three waste types
are reasonably well established.

18. With regard to assessing occuparional exposures,the relationship between measurements of external
irradiation made in radiation fields by film, thermoluminescent or other personal dosimeters and theabsorbed doses in the tissues and organs of the bodywas discussed in the UNSCEAR 1982 Report. The
Committee adopted the convention that all numericalresults reported by monitoring services represent the
average absorbed dose in the whole body, recognizingthat these are almost always readings from thedosimeters that are r e ~ o r t e d without.
consideration of
the relationships to the absorbed doses in organs andtissues of the body. In this Annex the Committee
adopts a similar convention; but to simplify comparisons, and because most exposures are to penetrating gamma-radiation, the numerical result is taken to
represent the effective dose equivalent, Exposures ofuranium miners to radon and its daurrhters are also
expressed in terms of effective dose equivalent.

12. High-level wastes (HLW) are primarily the spent
fuel elements or the solidified Lvaste products from
reprocessing. They have high activity concentrationsof both actinides and fission products and are
significantly heat-generating. As file1 elements are asignificant potential source of fissile material, they willusually be stored in the short-to-medium term rather
than disposed of. Occasionally, other waste streams
with high activity concentrations are also regarded as
H L W , but the quantities of activity in them are
relatively small.
13. Intermediate-level wastes (ILW) are defined tosome extent by exclusion from the other two categories: t h e contain either actinides or long-lived
beta/gamma emitters in quantities that are not negligible o r substantial activity concentrations of shorterlived beta/gamma enlitters and are not significantly
heat-generating.
14. Low-level wastes (LLW)contain primarily reasonably short-lived beta/gamnia emitters in low-tomoderate activity concentrations. They may containactinides or long-lived beta/gamma emitters but onlyin very small quantities.

15. There will be other categories of materials thatare uncontaminated, even though they were generated
a t a nuclear site or are of such a low level of activity
concentration that they can be exempted from therequirements for storage and disposal as radioactive
waste. The rationale for such exemption is that theradiological impact of uncontrolled disposal of thesematerials is insignificant [I14. N7]. These wastes are
not considered part of this study. as their potential forradiological impact is by definition very lo\\ incomparison with that from the other waste categories.
16. In this preliminary assessment of doses fro111
disposed wastes. only LLW and some categories of
ILW are considered to be disposed of by shallow land
burial. All other Ivastes are stored under conditions
such that the doses to members of the public are
essentially zero. and doses to occupational workers
are included in those assessed for other operations at
the same sites.

17. The Committee presented detailed comprehensive
revie~vs of occupational exposures. including thosefrom the nuclear fuel cycle, in both the UNSCEAR
1977 Report [U2] and the UNSCEAR 1982 [GI]Report. In this Annes the data on occupational-

19. The characteristics of occupational dose distributions identified by the Committee as of interest
were: ( a ) the annual average effective dose equivalent
Hefh which is related to the average level of individual
risk: this average has generally been calculated for all
individuals monitored in a given occupational group;(b) the annual collective effective dose equii.alent, Serf,
which is related to the impact of the practice: (c) thecollective effective dose equivalent distriburion ratio,
defined as the ratio of' the annual collective effective
dose equivalent delivered at annual effective dose
equivalents exceeding 15 mSv to the total collective
effective dose equivalent. This is related to the
proportion of workers exposed to higher levels of
individual risk. These characteristics may be obtained
for any form of the dose distribution, ~vhetheror not
it exhibits a log-normal o r other defined response overa n part of thc effective dose equivalent range. The
collcctivc cffectivr dose equivalent i h usually calculated
from collated dosimetry results using the definition

where N , is the number of individuals in the effective
dose equivalent range i for tvhich HCil.,is the meanannual effective dose equivalent. The annual average
effective dose equivalent, R,,,-,is giiten by

where N is the total number of ivorkers monitored.
20. The normalized measure of the impact of thevarious components of the nuclear fuel cycle is thecollective effec~ive dose equivalent per unit electricenergy generated. This is calculated as an average o \ e r
a complete power programme o r over several years to
avoid anomalies such as those connected with the
shut-down of reactors for maintenance. The results fordoses from occupational exposures and to the local.
regional, and global populations exposed as a result ofeffluent discharges to the environment may be taken
to be a relative measure of the health impact of
nuclear power production.

I. MINING AND MILLING
21. Uranium mining operations involve the removalfrom the ground of large quantities of ore containing
uranium a n d its daughter products at concentrations
between a tenth and a few per cent U,O,. These
concentrations are several thousand times the concentration of these nuclides in the rest of the natural
terrestrial environment. Uranium is mainly minedusing underground or open-pit techniques, othermethods such as heap leaching accounting for only afew per cent of the world production. The quantities
produced during the period 1980-1984 are given in
Table 3. Milling operations involve the processing ofthese large quantities of ore to extract the uranium in apartially refined form, often known as yellow-cake. Thisis further refined. converted and enriched, iS necessary.
before fabrication into fuel elements. Uranium mills
tend to be located near mines to minimize transportation. The number of mills operating is related touranium demand.
A.

EFFLUENTS

22. The predominant gaseous effluent from active
uranium mines is ?::Rn in the ventilation air fromunderground mines or released into the pit fromsurface mines. In a study covering 27 mines [J2] thisaccounted for 9 7 4 of the radon released. A recent
study [N5] has also shown that for some surface
mines. especially where a large volume of overburden
has to be removed to expose the ore, waste rock piles
formed a source of radon of a magnitude con~parable
to that of the pit. Release rates per unit inass of ore
were estimated in the UNSCEAR 1982 Report at
about 1 GBq I - I from underground mines and about
0. I GBq t-' from surface mines. In general, however,
the ore from underground mines was estimated tohave about 10 times the uranium concentration of thatfrom surface mines; the normalized radon emission
was thus taken for both types to be 1 GHq t - I of ore
for I%, uranium oxide in the ore. Particulates inand its d a u g h ~ e r sandairborne dust contain "1
sometimes "'Th and its daughters.
23. The results of measurements o r estimates of
either total radon emission rates o r normalized radon
emission from a number of mines are given in Table 4.
The data for underground mines relate to the ventilation air from the shaft, those for surface mines. to themine pit. The results support retention of an overall
normalized radon emission of I GBq t-' of ore for Iq
uranium oxide in the ore.
23. The uranium requlrements per unit elcctrlc energ)
generated vary someu.hat between current designs ofthermal reactors: but the heavy metal requlrements
are generall) in the range of 150-250 t (GW'a ) I . The
grade of ore mined at present is usually betueen 0.1
a n d 1% U:O,. Taking a typical value for undergroundmines from the United States of 0.2% [El], thenormalized radon releases are about 70 TBq (GU' a)-'.
This is the same value that was estin~ated in the
UNSCEAR 1982 Report.

25. The processing of uranium at the mill was
described in the UNSCEAR 1982 Report, as were thebroad characteristics of the tailings piles, where mostof the activity not extracted as usable uranium resides.
This activity is predominantly ?j0T'hand its daughters.
There are airborne emissions during operation of amill. mainly of :?:Rn together with :3wU.
:jOTh. 22bRa
and ?I0Pb. The ranges of airborne release rates for atypical mill estimated in the UNSCEAR 1982 Report
are shown in Table 5.
26. During operation of a mine. there are stockpilesof ore and piles of sub-ore. overburden and waste
rock. After closure there will typically be a pile ofoverburden, possibly covered by sub-ore, in caseprocessing of this becomes economically viable in the
future. These also act as sources of airborne emissions.
principally of Z22Rn.An estimate of the radon emanation rate from waste rock per 1% ore grade in the
United States is 100 Bq m-Z s-' [N4]. The number ofinactive mines in the United States was estimated to
be about 1,250 surface and 2,000 underground in 1980
[H9]. Some useful measurements have been made ofradon emanation rates under d n conditions over awide range of ore grades in the Northern Territory ofAustralia [L2. M5]. These suggest that a radon
exhalation rate of 50 Bq m-I s-I per 1% ore grade
is widely applicable: this figure is equivalent to
0.5 Bq m-' s-I per Bq g-I.
27. Extraction of uranium during milling is clearlymade as con~pletea s possible but cannot reach 100%.
Typically, the residual tailings from the mill willcontain from 0.001 to 0.01% U,O,. depending on thegrade of ore and the extraction process. Tailings aredischarged from mills into impoundmenrs, the characteristics of which depend on the local climate andgeology [TI]. From the point of view of estimating
effluents, the major differences are whether the tailings
pile is wet or d r j and whether it has been covered. All
tailings piles act as sources of airborne releases,
although if they are completely covered by water, therates can be e\tremely I O N . Estimates of radon
emanation for a number of typical mill tailings areasand impoundments are shown in Table 6. Most ofthese are taken from an extensive study by the Nuclear
Energy Agency (NEA) [NS]. The radon exhalation
rate per unit area and specific activity of ?IbRa wasestimated in the UNSCEAR 1982 Report at about
1 Bq m-: s-I per Bq g-I of ??"Ra in the tailings, although
it Has noted that the rate could vary from effectively
zero to an order of magnitude higher than the above
figure. I t has been suggested that a more realistic
figure \vould be 0.2-0.5 Bq m-' s-I per Bq g-' [S12].
For comparison, 0.015 U,O, ore contains approximately I Bq g-' of "bRa. Detailed measurements havebeen carried out o n seven tailings dams in South
Africa [A7]. giving a mean radon exhalation rate of
0.4 Bq n ~ - :s-' per Bq g-I for a radium concentration
ranging from 0.2 to 0.7 Bq g-I. Measurements on
tailings in the Elliot Lake area of Canada [B26]
sho\ved a range from 0.2 to 7.6 Bq m-' s-I per Bq g-I.
Experimental investigations on t u o types of bare dry
tailings in Australia [S13] showed exhalation rates
from 0.3 to 0.7 Bq m-2 s-I per Bq g-I; these were reduced
by a factor of 3 for 1 m dry cover and by more than afactor of I O for 1 m moist cover.

28. In considering the longer-term impact of effluentsfrom tailings piles. i t must be assumed that activity
concentrations from uranium nuclides remain practically constant indefinitely, due to their long halflives. The rest of the activity in the tailings isdominated by :'OTh, which has a half-life of 80,000 a.
The radionuclides in the decay chain from Z30Thwith
the greatest radiological significance are '26Ra, which
can be leached out by water access, 2'0Pb and 222Rn,
which can escape into the air.
29. At present, tailings have tended to be kept in
open. uncontained piles or behind engineered dams or
dikes with solid or water cover. I t is- likely, however.
that some further engineering will be carried out tominimize the release of radionuclides from the abandoned piles. Such techniques were analysed in the
NEA study [N5] for a number of sites. The radon flux
density varied by factors of more than lo6. dependento n the treatment assumed, showing that this is clearlya crucial parameter in the assessment of the impact oftailings piles. The options assumed for one typical site
in an arid region and the relative radon flux densities
assumed to result are shown in Table 7. Similar
reductions in radon emission have been found using
covers of various types [H 101. Assuming some reasonably impermeable cover is used, the radon exhalation
rate from a typical tailings pile is taken to be
10"qm-' a-I. This is less than the figure assumed for
emanation from the unstabilized material stockpiled
around working mines and comparable with the value
expected to be achieved in the United States [E4]. The
cover is assumed to provide some protection against
erosion. s o that the radon exhalation rate remains
essentially constant with time. Otherwise. an increaseof u p to double the initial rate of emanation from abare pile could have been expected over a period of
about lo4 years [N5]. As can be seen from the resultsof the UNSCEAR 1977 and UNSCEAR 1982 Reports,
these are critical assumptions in determining theoverall impact of the fuel cycle.
30. Mine and mill sites in dry areas give rise toeffectively no liquid effluents. For those in wet
climates. however, run-off water will contain radionuclides and may need trealrnent before release into
watercourses. The most important radionuclide inliquid effluents is 226Ra. and typical releases at wet
sites were estimated in the UNSCEAR 1982 Report tobe 1 GBq (GW a)-'. A review by Kaufmann [K5]suggests values of the order of 0.1 GBq ( G W a)-',given normal procedures for water treatment.
B.

LOCAL AND REGIONAL COLLECTIVE
DOSE COMMITMENTS

31. In the dose estimation procedure used in the
UNSCEAR 1982 Report, the typical characteristics ofa mine and mill site in terms of population density,
rainfall, farming, etc. were first established. The
population densities used were 3 k m 2 for 0-100 km and
25 km-' for 100-2.000 km. A deposition velocity ofrn s-I was taken for particulate releases. The
collective dose for radon release was then calculatedusing an atmospheric dispersion model with charac-

trristics typical of a semi-arid area and an effective
release height of 10 m. The atmospheric dispersion
model was described in the UNSCEAR 1982 Reportand in the original reference [CI]. The resultant
collective effective dose equivalent commitments perunit activity released are shown in Table 8. with the
exception of the figure for radon. This has been
reduced for the reasons discussed in Annex A whichhave led to a reduction in the dosimetric coefficientfor outdoor air from 17 to 9 nSv h-' per Bq m-'.
These figures have been used in this Annex t o estimate
the normalized collective effective dose equivalentconlmitnlents from current atmospheric releases which
is about 0.3 man Sv ( G W a)-'. The doses from liquid
effluents are negligible by comparison.
32. Using the figure estimated for the ini~ialrate ofexhalation of radon from a typical tailing pile leads toan annual release of about 1 TBq ha-'. The propuction of a mine generates about 1 ha ( G W a)-' ' oftailings, so the releases during a period of five years,
corresponding to the duration taken for the current
discharge, would add a normalized collective effective
dose equivalent commitment of 0. I man Sv ( G W a)-'.
The rate of release as a function of' time is assumed tobe constant, and given the very long duration of thesource. the normalized collective effective dose equivalent commitment is proportional to the durationconsidered reasonable for assuming the release. Taking
this period to be lo4 years for the sake of illustration,
the result is an estimated 150 man Sv ( G W a)-'. An
alternative perspective on this component can beobtained by assessing the truncated collective effective
dose equivalent commitments up to different times.
Some examples of the results of such calculations forthe various coverings described in Table 7 are shownin Table 9, taken from the same study [N5].

C.

OCCUPATIONAL EXPOSURES

33. The main source of radiation exposure of underground uranium miners is radon and its daughters.
The annual average exposure of underground miners
was taken to be 1.5 WLM in the UNSCEAR 1982
Report; this was converted to an annual effective dose
equivalent of about 13 mSv. Surface miners have alower exposure to radon and daughters, with annual
doses estimated to be about 3-4 mSv, but they andunderground miners are exposed through inhalationof dust containing uranium and its daughters. Bothunderground and surface miners are also exposed tosome external gamma radiation. The estimate ofannual doses for underground miners was rather
broad in the UNSCEAR 1982 Report, 1-10 mSv; thatfor surface miners was taken to be 1-2 mSv. Where theauthors have not carried out their own conversions,use has been made of the conversion coefficients givenby the International Commission on RadiologicalProtection [I121 between committed effective dose
equivalent and time integrated equilibrium equivalentradon daughter concentration in air of 17 nSv h-' per
Bq m-' o r 10 mSv WLM-I, where 1 WLM is one
working month (170 ha) of exposure to a potentialalpha-energy concentration of 2.08 lo-' J m-' [IIO].

34. Exposure of uranium miners to radon anddaughters has been monitored by a combination of
measurement of levels in air at a variety of places
through the mine and estimates of the time spent byminers in those places. In recent years, however, there
has been considerable development work on dosimeters suitable for monitoring of radon daughter
exposures for individual underground uranium miners.
Some recent results for underground uranium miners
are shown in Table 10. The United States data for
1980, assumed to be primarily for undergroundminers. are from a very general summary prepared by
the Environmental Protection Agency [E3]; those for
198 1 and 1982 relate only to the mines in New Mexico[SS]. The data for Canada (A41 include exposures at
the uranium mills associated with the mines. Data can
be clearly separated into underground and a surface
mine for the Canadian mines, and the results for thesurface mine at Key Lake [A41 are shown in Table 11.
A comparison between mine company records andexposures based on measurements by inspectors for1979 and 1980 in the United States showed reasonableagreement [C7]. In this study the annual average
effective dose equivalent to underground workers in
61 mines from exposure to radon and daughters wasestimated to be in the range of 18-29 mSv, dependingon the assumptions made in deducing the personnel
exposures from the measurements in working areas.
This is somewhat higher than the estimates given in
Table 10.
35. Information on gamma exposures to workers inboth underground and open pit mines in Canada [A41shows annual average effective dose equivalents ranging
from 0.1 to 3.4 mSv for the years 198 1-1983. Someunderground mines showed average gamma doses as
low as for surface mines, but the major undergroundmines employing more than 80% of the work-force
had an annual average effective dose equivalent of
3 mSv. An estimate of 3 mSv as the annual average
effective dose equivalent from inhalation of dust has
also been made for the Ranger surface mine inAustralia [A8].
36. Taking all the above information into account,
the average annual effective dose equivalent to underground uranium miners from both external exposureand radon daughter exposure is 10-12 mSv; that forsurface miners is lower, possibly around 5 mSv. Given
the predominance of underground miners. an overall
annual average of 10 mSv seems a reasonable estimate
for the early 1980s. Taking the productivity to be 3 t a-!
of natural uranium per miner and a natural uranium
requirement of about 200 t (GW a)-'. the normalized
collective effective dose equivalent would be 0.7 man
Sv ( G W a)-'. This is comparable t o the estimate in the
UNSCEAR 1982 Report of 0.9 man Sv (GW a)-',which was rounded up to I man Sv (GW a)-'.
37. Recent data on doses received by 3.000 workers at
uranium mills in the United States show a n annual
average effective dose equivalent of 2.7 mSv [E3]. The
external average effective dose equivalent to 131 workers
a t the Nabarlek mill in Australia during the period
198 1-1982 was 1.5 mSv [M9] and at the Ranger millduring the period 1985-1986 as low as 0.9 mSv [A8].

The contribution from workers at mills to the collective effective dose equivalent per unit electric energy
generated is so small that in the UNSCEAR 1982
Report it was not included as a separate item. Thissituation does not appear to have changed.

11.

URANIUM FUEL FABRICATION

38. The uranium ore concentrate produced at themills is further processed and purified and converted
to uranium tetrafluoride (UF,), and then to uranium
hexafluoride (UF,), if it is to be enriched in theisotope 215U. before being converted into uraniumoxide or metal and fabricated into fuel elements.
Natural uranium, containing 0.7% :3SU, is used ingraphite o r heavy water moderated reactors. Enrichments of 2-5% are required for light ivater reactors(LWRs) and advanced gas cooled reactors (AGRs).
39. T o produce natural uranium metal fuel, theuranium tetrafluoride is compressed with shredded
magnesium and heated. and the resulting reduced
uranium is cast into rods that are machined andinserted into cans. Natural uranium oxide is sinteredinto pellets and clad in zirconium alloy for HWR fuelpins. For LWR and AGR fuel, the UF, is converted
into the gaseous form UF,. The first type of enrichment piant to be developed con~n~erciallyemployed
the gaseous diffusion process. In this, the UF, diffuses
through a porous membrane, the lighter compoundcontaining 235Uand 'j4U diffusing more rapidly thanthe heavier compound containing 238U.Partial separation occurs, but in practice many stages of suchmembranes are required in series t o provide a cascade.
40. The pumping power required t o mo\-e the UF,
through the cascade requires a large amount ofelectric energy. The alternative gas centrifuge process
consumes only about 5% of the electric energy
demanded by the diffusion process. The gas centrifuge
process is based on the separation effect on a mixtureof UF, isotopes in a strong centrifugal field in arotating cylinder, suitably combined with the cascading
effect of counter-current circulation. More separation
is attained in one centrifuge stage than one diffusion
stage but. as the mass flow is less, a series-parallel
configuration is required.
41. T o fabricate LWR fuel the enriched UF, isconverted to the oxide (UO,) powder, which is
granulated, sintered and pressed into pellets. These areinserted into tubes (cladding) that are sealed after
being filled with pellets. For LWR fuel cans zirconium
alloy is used, while for AGRs stainless steel cans are
adopted. After the enrichment process, large quantities of depleted uranium remain, containing about
This uranium would become a
0.3% o r more 'I".
source of public exposure were it to be disposed of,but currently it is stored for possible use in breeder
reactors and for other purposes. The solid wastes
arising during operation of the uranium fuel fabrication facilities will contain the same radionuclides as
those at uranium mines and mills, but will be trivial inquantity by comparison. It does not, therefore, seem
worth while to assess their impact separately.

A.

EFFLUENTS

42. Emissions of radionuclides from the conversion,enrichment and fuel fabrication processes are generally
small. Most of the uranium compounds are solid andare easily removed from airborne effluent streams,
while settling tanks are used to reduce liquid effluent
discharges. Few data published in the United States o r
Europe give discharge rates of radionuclides fromthese fuel cycle facilities. The Committee concluded in
the UNSCEAR 1982 Report that discharges lipere
small and estimated releases from model facilities
producing LU'R fuel. In the United Kingdom, reported
discharges are given in terms of total alpha, total beta
activity and masses of uranium (Table 12). and someisotopic breakdown can be obtained [G2] for centrifuge enrichment plant effluents. Most of the betadischarges are from the short-lived :'"Pa(half-life:
1.17 min) which is separated with "'Th (half-life:
24.1 dl. Canadian data-are also available for effluentsfrom a conversion plant [A4, LI] with their isotopic
composition [XII]. There are small releases of "Tcreported from the British enrichment plant. indicatingsome recycling of reprocessed uranium, but these
releases are atypical and no dose assessment has beenmade.
43. The data presented in Table I? have been used toobtain the effluent releases which are applied to thesame model facility sited on a river as tias used in the
UNSCEAR 1982 Report. The normalized releases are
based on an LWR cycle uranium requirement of 150 t(GW a)-' and an HLVR cycle requiren~entof 170 I( G W a)-'. The results are given in Table 13 foratmospheric and aquatic effluents. The conversion
plant figures are based on data from Canada [MZ]. as
are those for fabrication. since these relate to freshwater discharges in contrast to the British figures.
which relate to marine discharges. The values quoted
in Table 13 are typical figures taken from those
calculated for the five Canadian fabrication plants,
based on a fuel cycle requirement of 170 t IGW a)-'.
The discharges of L34Thare obtained by assuming that
this radionuclide is in equilibrium with ?"U.

B.

LOCAL A N D REGIONAL COLLECTIVE
DOSE COMMITMENTS

45. The Committee concluded in the UNSCEAR
1982 Report that releases to the atmosphere provided
the major exposure to the population (over 90%) fromfuel conversion, enrichment and fabrication processes.
To obtain an order of magnitude assessment of thecollective dose commitments, the Committee specified
a model facility with a constant population density of
25 km-2 out t o 2.000 km. This was chosen to be
representative of North America and Europe, andcollective dose commitments were derived for inhalation from the plume, ingestion of foodstuffs contaminated by activity deposited from the plume and by
external irradiation from the activity deposited on the
ground. The same results have been used here. but thecollective effective dose equivalent commitments havebeen scaled for the normalized releases derived inTable 13; the resultant doses are given in Table 14.
The most significant pathway of exposure coniinues tobe inhalation of particulate activity, with radon
daughters contributing about 15% of the dose.
46. In summary, the normalized collective effective
dose equivalent commitment due to uranium fuelfabrication is estimated to be 2.8 10-j man Sv ( G W a)-'.
The main contribution arises froni inhalation of theisotopes of uranium. The figure is similar to thatderived in the UNSCEAR 1982 Report [2.0 10-'man Sv( G W a)-']. Individual doses in the vicinity of fuelfabrication facilities are estimated to be less than
50 mSv per year for members of the public [B I. B2,
B3. BX. B 16, 13291.
C.

OCCUPATIONAL EXPOSURES

37. The annual average effective dose equivalents toworkers in fuel fabrication plants were found in the
UNSCEAR 1982 Report to be generally low. ranging
from 0.3 to 3 mSv. The annual collective effective dose
equi\~alcntdistribution ratio (see paragraph 19) \vasalso in general small. often approaching zero. Data on
the number of workers employed and the corresponding annual average individual and collective effective
dose equivalents are given for some countries in
44. The results in Table 13. which were derived from
Table 15. These are not always complete for a c o u n t r reported discharges, can be compared H ith the effluentsfor any particular year and could include workers notfrom the model facilities quoted in the UNSCE.AR
strictly employed in fuel fabrication. For example. the1982 Report. which were based mainly on the notional
data from the United States [N2] are quoted as
results produced by the Environmental Protection
corresponding to fabrication and reprocessing. but i t
Agency [E2]. The results using present data suggest
has been assumed that the contribution from reprothat for conversion, atmospheric releases are generally
cessing in the years 1980 and 1981 tvas negligible: theabout twice those quoted prc\ iously for uranium anddata for the United Kingdom [H8, RI?] include
thorium isotopes, white aq~laticreleases as reported
exposures during enrichment. Annual average doses toare about 10% of those assumed previously, and in thefuel fabrication workers have remained lour, in the' of that in the Environmental
case of IZbRaare only 13
range of 1-2 mSv. The collective effective dose equiProtection Agency model facility assumed pre~~iously. valent distribution ratio t i ) r Uni~cd Stales u.orkers.
Atmospheric releases from enrichment are about one\vhich was 0.12 in 1980, decreased to 0.09 in 1981
half of those quoted in the UNSCEAR 1982 Report:[N2]: that for British \vorkers was 0 in 1982 and 0.02
liquid effluents are only a few per cent of thein 1983 [ B 121; that for Japanese workers \\'as 0 in theCommittee's previous estimates. For fuel fabrication.
period 198 1 - 1984 [TI 21.
based on a weighted average of natural and enrichedfuels, the atmospheric and aquatic releases are again
48. Some data on the external doses from theabout one half the previously assumed values.
fabrication of plutonium fuel at the PNC works inJapan have been published [A5]. These are shown in
Table 16. During the period 1977-1982 the total
amount of fuel fabricated was 37.6 t for an advanced
H W R and 1.2 t for an FBR. From 1980 to 1982. it
was necessary to process reactor grade plutonium
recovered from high burn-up fuel, and this led to anincrease in both average and collective doses to thework-force.
49. The estimates of normalized collective effective
dose equivalent in the UNSCEAR 1982 Report wereconsiderably reduced from previous estimates; theoverall figure estimated to be 1 man Sv (G\i: a)-'.
More recent estimates are shown in Table 15. The
normalized collective effective dose equivalents for
Canada and the United Kingdom were obtained bydirectly relating the collective dose in a year to theelectric energy generated in the year [I3. 14. 15. 161. asseems appropriate for nuclear power programmes in
an approximatel> equilibrium situation. For the United
Stares the same assumption is made a s in the
UNSCEAR 1982 Report; 604; of the fuel Fabricated isfor United States nuclear power stations. For Japan.
figures for 1981-1984 are used [J3], and these havebeen tentatively related to the total electric energy
generated in the corresponding years by nuclear
power. Giving appropriate weight to more recent
data, a n overall average of 0.5 man Sv ( G W a)-I now
seems more appropriate.

111.

REACTOR OPERATION

50. Nearl! all the electric energy generated by
nuclear power is produced in thermal reactors inwhich the fast neutrons produced by the fission
process are slowed down to thermal energies by use ofa moderator. The most common materials still used
for moderators are light ivater. heavy water andgraphite. The choice of moderator and coolant. lighto r heavy water or carbon dioxide gas. greatly affects
the design. size and heat reniotal system of thereactor.
51. The uranium fuel is contained in discrete pins,
both t o prevent leakage of the radioactive fission
products into the coolant circuit and to improve neutron
economy by reducing parasitic neutron captures in theresonance neutron energy region of '-'WU.
The heat
generated in the fuel pins by the slowing down of thefission fragments is removed by forced con\.ection. themost usual coolants being light o r heal! \yarer or
carbon dioxide gas. In the case of fast reactors. theneutrons are not modcrated and induce fissions \vithenergies close to those at which they are produced.
The usual heat removal system is liquid sodium. which
is a good heat transfer medium and does not greatly
moderate the neutrons.
.I\.

EFFLUENTS

52. 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 formed in the coolant circuit
because the coolant becomes activated, because of the
diffusion of fission product elements with radioactive
isotopes from the small fraction of the fuel withdefective cladding, and because of the corrosion of
structural and cladding materials anywhere in the
coolant circuit which leads to particles being carriedthrough the core and becoming activated. All reactors
have treatment systems for the removal of radionuclides from gaseous and liquid wastes. which arise
from leakage out of the core or from clean-up of thecoolant.
53. The quantities of different radioactive materials
discharged~fromreactors depend on the reactor type.
its design and the specific waste treatment plantinstalled. Radionuclides discharged to atmosphereinclude fission noble gases (krypton and xenon),activation gases ( I T , IbN, '%, 41Ar).tritium. iodine
and particulates. Radionuclides released into theaquatic environment in liquid effluents usually include
tritium, fission products and activated corrosion
products. The discharge data for the years 1980-1985
are presented in this section, and the annual normalized
releases are evaluated for each reactor type andaveraged over all reactors of each type as TBq( G W a)-'. Normalized results are not presented forindividual sites because releases in any one year may
reflect a need for maintenance or irregular procedures
which are the culmination of a number of vears ofprevious operation. The total releases of radionuclides
between 1980 and 1984 have been normalized by
dividing by the total electric energy generated ( G W a )over the same period. These normalized releases areused to assess collective dose commitments because
the 1985 data were incomplete. Generally, the normalized releases for 1985 from the ~ a r t i a datallead tolower values than for the previous five years. although
the 1980-1985 averages are mostly within 10% of the
1980- 1984 averages.
1.

Fission noble gases

At least nine identified radioactive isotopes ofkrypton and 1 1 of xenon are formed during fission.
htosr have half-lives of minutes or seconds and decay
before they migrate significantly in the fuel. A fractionof the noble gas inventory of the fuel pins diffuses tothe fret: space between the fuel and the cladding,leading to a build-up of gas pressure. The presence of
noble gases in the coolant circuit is generally anindication of fuel cladding failure.
54.

55. Table 17 lists tlie reported discharges of noble
gases irom PWRs. The releases span many orders ofmagnitude partly because of the design of newer
plants and partly because of the need for irregular
operations and maintenance. Thus, the normalized
releases presented are averaged over all PWR electricenergy production from 1980 to 1984. Short-lived
noble gases only appear in PWR effluents because ofleakages in the primary water pressure circuit. Gaseous
wastes can also arise from the condenser exhaust on
the steam circuit and from blow-downs or contain-

ment purges. These wastes are usually held underpressure in delay tanks to allow decay of short-lived
isotopes before release. The isotopic composition ofnoble gases released from PWRs in the United Statesin 1982 is shown in Table 18. Comprehensive data are
available for each year from United States reactors;
a n d data available from other countries are similar to
those from the United States. The data for the UnitedStates for 1982 are therefore assumed to be representative of the isotopic con~positionof releases between
1980 and 1984 and are used for dose estimation.

be replaced. Releases of noble gases from HWRs a n d
LWGRs are given in Table 21. Normalized release for
HWRs has been 212 48 TBq ( G W a)-'. similar tothat for PWRs. The highest figures are for LWGRs at
5,470 + 1.370 TBq (GW a)-I, about three times thefigures for BWRs. The available discharge data
indicate that FBRs have lower releases of noble gases:
measurements at BN-350. an FBR in the USSR.
indicate 65-130 TBq ( G W a)-' [P6].

+2. Activation gases
56. Data for 1985 are incomplere and the releases are
not inclgded in the nornlalized set. The normalized
releases seem to have remained fairly steady overthe five-year period 1980-1984, but the average of
218 -t 40 TBq (GW a)-' appears to be about half ofthe value reported previously by the Committee[430 TBq ( G W a)-']. Xenon-1 33 accounted for 75% of
the discharge and Ij5Xe for 13%. In the UNSCEAR
1982 Report the comparable figures were 85% and
5%. respectively. Some of the reduction in discharges
is thought to be due to better fuel can performance.
which would account for the lower releases t o coolingwater. The other feature is the inclusion of' neuer
reactors wilh lower levels of discharge.
57. Reported discharges of noble gases from BUrRs
are shown in Table 19. The releases vary by six ordersof magnitude, although the average releases continue
to have been reduced from previous years. The
normalized releases are shown in Table 19 for all
BWRs from 1980 to 1985. The main source of noble
gas release from BWRs is gases in the steam circuitthat are continuously removed by the main condenser
air-ejector system. The isotopic composition depends
on the hold-up time, which is usually less than for
PWRs, thus allowing more short-lived isotopes to bereleased. Table 20 gives the radionuclide compositionof noble gas releases from. United States BWRs in
1982, which is similar to that of reactors in othercountries. These figures again are taken to be representative of BWR releases in all countries during theperiod 1980- 1983 and are used for dose assessment.
58. F o r BWRs the average discharge rate for noble
gases during the five-year period 1980-1984 is
2,150 f 520 TBq (GW a)-! compared with 8.800 TBq( G W a)-' reported in the UNSCEAR 1977 Report for
1975-1979. This reduction seems to have been achieved
because of significant reductions in releases fromthose reactors that previously had the highest discharge rates. The normalized release for 1985 issignificantly lower (460 TBq ( G W a)-!), partly becauseof the missing data but mainly because of very large
reductions in discharges from the largest previous
sources (Browns Ferry and Brunswick). The isotopic
composition shown in Table 20 reveals that most of
the activity consists of 68Kr (half-life: 2.8 h), IJ3Xe(half-life: 5.3 d), I3'Xe (half-life: 9.2 h) and 'j6Xe (halflife: 17 min) in almost equal quantities.
59. In GCRs, noble gas releases are insignificant
compared with activation gases. Magnox reactors,
AGRs, LWGRs and HWRs utilize on-load refuellingand, in the event of fuel element failure, fuel rods can

60. Although GCRs d o not generally release fission
noble gases, several gases are formed during gas
cooled reactor operation. These are primarily "Ar.
formed by activation of the stable argon in air. and 'S
produced from sulphur and chlorine impurities in thegraphite core. The discharge data for "Ar are reported
in Table 22. For " S . measurements were made in the
United Kingdom at Hinkley B, Oldbury and Wyifa[HI, H2], and discharges have consistently averaged
0.2 TBq ( G W a)-I with only about 20% variation
around the mean.
61. The quantity of "Ar (half-life: 1.8 h ) released
depends upon the detailed design of the reactor. F o rearly Magnox reactors having steel pressure vessels,
the principal source of 41Aris the activation of stableargon in the air used as cooling air around the outsideof the pressure vessel. For advanced reactors withprestressed concrete pressure vessels, the principal
source of "Ar is leakage of the coolant CO,, whichcontains small amounts of air, to the atmosphere. The
normalized releases from AGRs are 5-15'2 of thevalues for Magnox reactors. The average normalized
release from GCRs is 2,320 220 TBq ( G W a)-I
compared to 3,240 TBq ( G W a)-I in the UNSCEAR
1982 Report. The reduction is due to the proportionof power now generated by AGRs and addition of
French data. For BN-350, the Soviet FBR, normalized'IAr releases average 470 TBq ( G W a)-! [K4. P6].

+

62. Nitrogen-16 (half-life: 7 s) causes direct externalradiation at nuclear power plants. The photons produced in its decay have energies of 6.1 and 7.1 MeV.
In BWRs, the I6N generated in the coolant water istransferred in the steam t o the turbine buildings.
Direct radiation from gas ducts in steel pressure vesselgas cooled reactors produces the major dose to
individuals close to those sites.
3.

Tritium

63. In LWRs tritium arises from ternary fission in
the nuclear fuel and from the neutron activation oflithium and boron isotopes dissolved in, or in contact
with. the primary coolant. The Committee assessed in
the UNSCEAR 1982 Report the tritium production
rate from ternary fission as 0.75 PBq ( G W a)-'.
Tritium generation from activation reactions in PWRs
seems to result mainly from boron, which is used forreactivity control, in the coolant, whereas in BWRs it
results mainly from boron in control rods. In GCRs it
is the result of lithium impurities in the graphite a n dthe presence of water vapour in the core. For HWRs it
is principally the result of the activation of thedeuterium moderator and coolant. The activation
production rate only exceeds that from ternary fissionin HWRs, where the activation rate was previously
estimated by the Committee to be 30 times higher at
about 25 PBq ( G W a)-'.
64. Table 23 presents the tritium releases to theatmosphere for 1980-1985 for PWRs. BWRs and
HWRs. For PWRs the normalized release over the
five-year period 1980- 1984 is 5.9 f 2.4 TBq (GM' a)-'a n d no particular trend is apparent over this period.
The corresponding figure was 7.8 TBq ( G W a)-' for
1975- 1979. The BWR releases normalized for the same
period average 3.4 1.6 TBq ( G W a)-', comparedwith 3.4 TBq (GM' a)-' for 1975-1979. The decrease inannual BWR normalized releases from 1980 to 1984
seems primarily attributable t o reductions from theDresden nuclear plant alone. while the higher figurefor 1985 is due to Hatch 1. These figures indicate thatabout 1% of the tritium produced in the fuel of LWRs
finds its way into the coolant and then enters airborne
effluent streams. For HWRs the production of tritiumin the moderator is the most probable source of tritium
releases, which averaged 670 f 190 TBq ( G W a)-' for
1980-1984, as compared with 540 TBq ( G W a)-' for
1975-1979. For some HWRs. however, the coolant
may be the main source of tritium production. Thereis little release of tritium to the atmosphere from
Magnox gas cooled reactors mainly because humidriers
remove water vapour from the gas circuit. There issome release of tritium to the atmosphere from AGRs,
and the normalized release is 5.4 f 0.9 TBq (GW a)-',similar to LWR releases.

+

65. From Table 24 it can be seen that releases oftritium to the hydrosphere from PWRs have been
fairly constant over the past five years, and the 19801984 normalized average is 27 2 1.8 TBq (GW a)-',with the figure for 1985 similar. This compares to the
38 TBq (GW a)-' obtained for 1975- 1979. The comparable figures for BWRs are 2.1 f 0.5 TBq ( G W a)-' for
1980-1984, which is 50% higher than the 1.4TBq( G W a)-' for 1975-1979 and no trend is apparent. For
G C R s the normalized release to surface waters is
9 6 f 13 TBq ( G W a)-', which contrasts with 25 TBq( G W a)-' for 1975-1979. There appears to have beena significant increase in tritium releases from GCRs
over the past five years. HWR releases in liquid
effluent streams averaged 290 2 68 TBq ( G W a)-' for
1980-1984, compared with 35OTBq ( G W a)-I for
1975-1979. LWGRs have low liquid releases at 1.7 TBq(G W a)-'.

[B6, V4]. similar to other PWRs and AGRs. The
liquid discharges amount to about 5 TBq ( G W a)-'and 1 TBq ( G W a)-' for PWRs and LWGRs, respectively [B6, P6. V4]. Measurements indicated that on
average 90% of the atmospheric releases of tritium
was in oxide form [B6]. Practical experience a t the
Novovoronezh APS (PWR) showed that it is possible
to reduce the tritium concentration in the coolant
water by 50% [B7].

68. Discharges of ''C are of interest because of its
long half-life (5.730 a) and contribution to collective
dose commitments. Estimates of "C production in
fuels depend o n the nitrogen level in the fuel can,although some is produced from reactions o n oxygenin oxide fuels. The Committee concluded in the
UNSCEAR 1982 Report that the normalized production rate within fuel for PWRs. BWRs. G C R s a n d
HWRs was close to I TBq ( G W a)-'. Little of this isreleased into the reactor coolant circuits, it appears to
be released during reprocessing [BI, B2, B3, 88. B16,
B291. Carbon-14 is produced in the moderators of all
reactors, production in H WRs being perhaps 100 times
greater than in LWRs or GCRs, because of the ''0(n, a) 14C reaction in the greater mass of oxygen in themoderator, and there is a consequential release.
69. The National Council on Radiation Protection
and Measurements (NCRP) [NI] has estimated the
production rate of I4C in PWRs to be between 2 a n d
3 TBq ( G W a)-', and for BWRs 3-4 TBq ( G W a)-'.
arising in both cases mainly in stainless steel a n dzirconium alloy. For the estimation of release rates tothe environment, NCRP assumes that the "C formedin the hardware remains there. but that the fraction
formed in dissolved nitrogen in the cooling water istotally released. The NCRP estimate for PM'Rs is
370 GBq ( G W a)-' and for BWRs 220 GBq ( G W a)-L.
The NCRP estimate of the release of "C t o theenvironment for FBRs is essentially zero a t thereactor.

66. Thus, about 0.3% of BWR tritium production
appears in liquid effluents, with a similar amountgoing to the atmosphere. For PWRs about 3% of thetritium produced is in liquid effluents. about five times
more than that going to the atmosphere. For HWRs
liquid effluents are about one half those discharged tothe atmosphere.

70. Environmental discharges of I4C are not routinely
reported for all reactors. The data summarized inTable 25 are from a series of measurements made inArgentina, the Federal Republic of Germany, Finland,and USSR. For BWRs it appears that essentially all the"C appears a s carbon dioxide, and the normalized
release rate for 1980- 1984 is 330 f 1 10 GBq ( G W a)-'.
significantly less than the Committee's estimate in the
UNSCEAR 1982 Report of 520 GBq ( G W a)-'. F o r
PWRs in the Federal Republic of Germany [Wl],Finland [B17] a n d the USSR [RI], the data indicate arelease rate of 345 f 80 GBq (GW a)-', which is significantly higher than the figure of 220 GBq ( G W a)-'given in the UNSCEAR 1982 Report. For PWRs only
about 5 5 0 % of the emission appears as CO,. It now
appears that normalized I4C releases from PWRs a n d
BWRs are similar.

67. For PWRs and LWGRs in the USSR theatmospheric releases of tritium are reported t o average
7.4 TBq ( G W a)-' and 1.9 TBq ( G W a)-I, respectively,

71. In recent measurements at three LWRs in the
United States [K2], two PWRs emitted an average of390 GBq ( G W a)-'. The source of I4C was different a tthe two sites: the first had 42% arising from venting ofgas decay tanks, 35% from auxiliary building ventilation and 32% from containment venting: the second
had emissions resulting primarily from pressure relief
venting and purging of the containment air, with only
7% from venting of gas decay tanks. For the BWR,
the discharge rate was 460 GBq (GW a)-' with 97% of
the release via the off-gas discharge, which was 95%
I4CO,. For the PWR 94% of the discharge was "CH,.
The I4C content of liquid and solid wastes was lessthan 5% of the aerial discharge for all reactors.
72. Measurements at LWGRs in the USSR gave
average releases of 1.3 TBq (GW a)-I [RI]. In the
United Kingdom, reported releases were 0.74 TBq(G W a)-' from Magnox reactors and 1.9 TBq (G W a)-'from AGRs. Weighted by energy production, thenormalized IJC release for GCRs is 1.1 TBq (GW a)-I
[H8]. The main source of IT releases from GCRs is
the leakage of the primary coolant, at a rate typicallyof a few per cent per day, which contains radionuclides
released to the coolant by corrosion of the graphite
moderator.
73. For HWRs it has been reported that a significant
fraction of the inventory formed in the moderator canbe released to atmosphere. Measurements at Atucha I[B 18, 031, however, for 1983-1985 have indicated that
releases are significantly lower than previously calculated for 1980-1982. The five-year normalized releaseis 6.3 3.3 TBq (GW a)-', whereas the Committee
had estimated 17 TBq (GW a)-' in the UNSCEAR
1982 Report. The form is again variable, between 40
and 80% being reported as CO,. In Argentina, regularmonitoring of discharges of lJC has continued forseveral years so that more reliable estimates can be
made.

+

5.

Iodine74. The volatile element iodine is produced by thefission process, the isotopes of radiological interest
being "'1 (half-life: 1.6 10' a), "'1 (half-life: 8.04 d),l3'I (half-life: 2.3 h). 'j31 (half-life: 21 h). :"I (half-life:53 m) and 'j51 (half-life: 6.6 h). Because. apart from
Iz9I, the iodine isotopes have such short half-lives,
equilibrium activity concentrations in the fuel arereached quickly and releases depend on the number offuel cladding failures and the coolant leakage rate.
Iodine has been studied for many years in view of its
mobility in the environment and selective thyroid
irradiation. Because of its long half-life, I z Y I is ofinterest in evaluating collective dose commitments;
however, its release from reactors is very small andoften not reported. Most of Iz9I in fuel is released
during reprocessing, from which i t makes a greater
contribution than from reactor operation.
75. Table 26 gives the reported atmospheric discharges of I3'I from operating reactors in various
countries for 1980-1985. There are considerable differences in the absolute quantities; these appear to be
attributable to differences in the ages of the plants andin the waste treatment designs. There does not appear
to be any trend in PWR releases, but BWR normalized
data show a sharp downward trend.

76. The annual normalized discharges of 13'1 from
PWRs were 1.75 0.33 GBq (GW a)-' for 1980-1984,
not significantly changed when compared with 1.9 GBq(GW a)-I for 1975-1979. The I3'I releases from BWRs
for 1980-1984 have averaged 9.3 F 4.9 GBq (GW a)-I
compared with l 3 ' I releases of 40 GBq (GW a)-' for
1975-1979. This reduction was because the few reactors that had large releases are currently releasing far
less. The results for HWRs indicate releases of0.23 f 0.08 GBq (GW a)-I. From early GCRs, which
utilize on-load refuelling, releases were negligible, andreleases from AGRs were 1.4 2 1.1 GBq (GW a)-L.
LWGRs released 8 0 2 40 GBq (GW a)-I [All, andmeasurements indicated that 60% of the iodine in thereactor off-gases was in organic form, 40% inorganic
and about 1% particulate [B9, D 1, S6].

+

77. The isotopic composition of iodine releases from
LWRs in the United States in 1982 is shown in Table 27[T5]. The isotopic conlposition was taken to be
representative of reactor operations in all countries
and was used as the basis for dose calculations. For
PWRs about 25% of the discharge is accounted for by" I 1 and 75% by ')'I, compared with the figures
reported by the Committee in the UNSCEAR 1982
Report of 30% accounted for by 13'I. For BWRs, 1 3 ' 1
releases represented about 7% of the discharges. with
I3jI and 'jSI contributing 28% and 65%, respectively.
This compares with less than 10% previously reportedforand 30% and 60% for l3jI and 'j51, respectively.
For LWGRs, 24% is accounted for by I3'I, 43% by I3'Iand 33% by l J S I [B21]. It might be concluded that
there was little change in the isotopic composition in
the periods 1975-1979 and 1980-1984.
6.

Particulates in airborne effluents

78. 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 circuit leaks or because of maintenance workon active components removed from the priman
circuit. The air in all areas where aerosols might arise
is continually purged and the plenum activity isfiltered by high efficiency particulate (HEPA) filters.
Results of recent measurements on particle size
distributions indicate a mean aerodynamic diameter ofl pm for fission products and IOpm for activation
products [B4]. Measurements at LWGRs in the
USSR have indicated mean aerodynamic diameters of
0.1-0.4 pm for particulates: for I 3 ' I and "Cr, 30-30q
were particulates with a mean aerodynamic diameterof less than 0.1 { r r n [BIO. C?].
79. Releases of particulate activity to the atmosphere
are s~lmmarizedin Table 28 for reactors around rhe
world. The quantities are extremely low. and thenuclide composition appears to be uniqlre to each
operating plant; it depends on the particular impurities
in cladding and structural materials. coolant chemistry
and fuel failure modes. The isotopic conlposition of
the release from a plant can vary from year to year,
because of different operational and maintenanceneeds. Consequently, the range of nuclides reported inatmospheric discharges is extremely large, several tens
PWR releases had been reduced by a factor of about 2of nuclides often being reported from one plant. No
and BWR releases by a factor of 10 from the figures
single nuclide can be identified as contributing the
given in its UNSCEAR 1977 Report. These reductions
majority of the activity released for any one type of
do not seem to have been attributable to the removalof any single nuclide but are applicable to all nuclides
reactor. Radionuclides identified include 'Be, "Na.
slCr, '4Mn, '9Fe, "Co, 5 a C ~6,0 C ~63Ni,,
65Zn, 7 6 A ~ , in the release.
"Rb, gqSr, 90Sr, 9LSr.9'Zr, 97Zr. 95Nb. 99Mo, 9 9 m T ~ ,1 0 3 ~ 1~ 0. 5 ~ 1~ .0 6 ~l O~ 8 m ,~IlomAg,~'"Sn. 'I5Cd. '22Sb,
83. The isotopic composition of liquid effluents froml24Sb, l25Sb l:3mSn 1Z3nlTe, 134Cs, 137CS, 13YCe, LJOBa,
United States reactors in 1982 is shown in Table 30.
About 20% of the normalized PWR discharge is due
IJ0La. 14'Ce, I4'Ce and lR2Ta.
to 5BCoand almost 20% to '"I, while "'Cs accounts
for about 11%. In the BWRs about 30% of the release
80. For PWRs the normalized release was 4.5 f 2.9
is due to 60Co and about 13% to I3'Cs. the other
GBq (GW a)-' for 1980-1984, compared with 2.2 GBq(GW a)-' for 1974-1979. For BWRs, the average
nuclides with significant contributions being "Na andrelease was 43 f 24 GBq (GW a)-', compared withb'Zn; "'I contributed about 3%. These figures repre53 GBq (GM7a)-I for 1974-1979. For HWRs thesent small changes from those in the UNSCEAR 1982
data yield 0.04 0.016 GBq (GW a)-'. similar to theReport. with some reduction in the percentages of
0.044 GBq (GW a)-! normalized release reported
caesium.
previously, while for LWGRs the average releaseappears to have been 15.7+ 16.2 GBq (GW a)-'.
84. For GCRs. 409% of the discharges to the aquatic
There were no figures previously for LWGRs. nor wereenvironment are due to I3'Cs and the ratio of "'Cs tothere any for GCRs. which now average 1.4 f 0.8 GBq"'Cs is 0.22. compared with 0.6 for PWRs and 0.5 for(GW a)-'.
BWRs. which reflects differences in fuel burn-up.
About 16% of GCR releases is due to 3SS. and POSr
accounts for about 6%.
7. Liquid effluents*

81. The sources of radionuclides other than tritiumin liquid effluents are essentially the same as thosedescribed for particulate releases to the atmosphere.
The reported levels of discharge are equally variable.
the magnitude and isotopic composition depending
upon the design and operating practice of the reactor.
impurity levels and trace quantities of material instructural and cladding components. Table 29 summarizes reported liquid effluent discharges from reactors around the world. In Table 30 the isotopic
composition of liquid discharges from power reactors
in the United States in 1982 is presented. and in
Table 31. that for GCRs in the United Kingdom isgiven. also in 1982.
82. The normalized release levels based on thereported discharges for each reactor type using reported
figures for electric energy generated between 1980 and1984 can be summarized from Table 29 and contrasted
with the figures given in the UNSCEAR 1982 Report.
PWR: 132?49GBq(GWa)-I.
compared with 180 GBq (GW a)-!
BWR: 115 i- 47 GBq (GW a)-',compared with 290 GBq (GQ' a)-'
GCR: 4.520 1.790 GBq (G W a)-'.
compared with 4.800 GBq (GU' a)-I
HIVR: 25.7 I 8.7 GBq (GW a)-',compared with 470 GBq (GW a)-'+The normalized releases for PWRs betitreen 1980 and1984 are similar to previous years although there has
been an increasing trend. while BWR releases are less.
Canadian HWRs were previously reported as giving
discharges of about 50 GBq (GW a)-l. while thehigher figures for the GCRs reflect the fact thatdischarges are made, with the exception of Trawsfynydd. to the marine environment. I t appears fromthe above results that aquatic discharges from BWRs
have been reduced by a factor of 2.5. In the
UNSCEAR 1982 Report, the Committee found that

85. There is a wide range of activation products andfission products reported in liquid effluent discharges,
and the isotopic composition varies even between
reactors of the same type. The normalized figures areused. however, to make an estimate of the collective
doses due to liquid effluent discharges.
B.

LOCAL AND REGIONAL COLLECTIVE
DOSE COMMITMENTS

86. National authorities usually require environmental
monitoring programmes in the vicinity of a nuclear
power plant to be carried out by the operator, anothercompetent agency or both. In general, activity concentrations of radioactive materials from effluent discharges are loo low to be measurable except close to
the immediate point of discharge. Dose estimates forthe population, therefore. rely on modelling theenvironmental transfer and transport of radioactive
materials.
87. In the UNSCEAR 1982 Report, rhe Committee
established a model site that was most representativeof areas of northern Europe and north-eastern UnitedStates, since those areas contain a large proportion of
the power-producing reactors. Agricultural production patterns and population distributions typical of
those areas were also established. The cumulati\~epopulation within 2.000 km of the site is about 2.5 lo8,
giving an average population density of 20 km-2.
Within 50 km of the site, the population density was
taken to be 400 km-2 in order to to reflect current
siting practice. The objective of the Committee remains
unchanged to give a representative value of thecollective dose commitments per unit of electricenergy generated by nuclear power stations and toreflect the levels of dose received by the most exposed
individuals. The results do not apply to any one
reactor or any one location, and the collective dosecommitments should not be applied t o a given reactor
with known discharge data to obtain estimates ofhealth detriment.
1.

Fission noble gases

88. Using the normalized releases for PWRs from
Table 17 for noble gas atmospheric releases and theradionuclide composition from Table 18, the normalized collective effective dose equivalent commitments
averaged between 1980 and 1984 were calculated for themodel PWR facility and are shown in Table 32. The
normalized release term from Table 17 is 218 TBq( G W a)-', and the radionuclides that contribute significantly to the collective effective dose are 'j3Xe and
13'Xe. The in-growth of daughter products. e.g., &%bfrom %r, has been included in the dose calculations,which are those presented in the UNSCEAR 1982
Report, but scaled for the different normalized releaseand isotopic composition.
89. The normalized collective effective dose equivalent commitment amounts t o 2.6 10-' man Sv( G W a)-' compared with the Committee's assessmentof 3.2man Sv (GW a)-', which was given in
Annex F of the UNSCEAR 1982 Report. This reflects
the reduction in discharges with little difference in the
distribution of radionuclide composition. About 64%
of the total collective dose is now due to '33Xe(80% in1982) a n d 28% to IsSXe ( I 1% in 1982). As in the
UNSCEAR 1982 Report, some 90% of the collective
dose commitment is accumulated within 500 km.
There is little contribution from the inhalation ofradioactive daughter products, and the dose estimates.
as before, include an allowance for the shielding frombuildings and the fraction of time spent indoors.
90. For the quinquennium 1980-1984. Table 19 shows
the normalized releases from B WRs to be 2.150 TBq( G W a)-', compared with the value of 8,800 TBq(GW a)-' given in the UNSCEAR 1982 Report. Taking
the relative isotopic composition from Table 20. thenormalized collective effective dose equivalent commitment is given in Table 33 as 0.56 man Sv ( G W a)-'.
compared with the Committee's estimate in the
UNSCEAR 1982 Report of 1.9 man Sv ( G W a)-'. The
main isotope contributing t o the collective dose is "Kr(half-life: 2.8 h) accounting for about 57%. somewhat
more than in 1982. Most of the remainder of thecollective dose arises from '3SXe (16%), 138Xe(9%)
a n d 13>Xe(8%), in somewhat smaller proportions thanin the UNSCEAR 1982 Report.
91. The in-growth of "'b(half-life: 15.4 min) from"Kr and '38Cs (half-life: 32.2 min) fromdecaysare included in the dose estimation, and the collective
doses include a contribution from the inhalation of
the a8Rb and '3sCs radioisotopes. The spatial distribution of the normalized collective effective dose
equivalent commitment is biased towards the source,with more than 80% of the dose accumulated within
50 km and nearly 50% within 10 km. This behaviour
is caused by the dominance of 88Kr,which decays witha half-life corresponding to about 40 km distancetravelled.

92. The normalized release of noble gases from
HWRs is 212 TBq ( G W a)" (Table 21), and assuming
the same relative isotopic composition as PWRs, thenormalized collective dose commitment is 0.024 m a n Sv
( G W a)-', while for LWGRs a normalized release of
5.470 TBq ( G W a)-' (Table 21) and the assumption of
an isotopic composition similar to that of BWRs yielda normalized collective effective dose equivalent commitment of 0.72 man Sv (GW a)-'.
93. In summary, the normalized collective effective
dose equivalent commitment from noble gas releases
is 0.20 man Sv (GM1 a)-', based on the five-year(1980-1984) weighting of electricity generated by
PWRs, BWRs, HWRs and LWGRs. The Committee
gave a figure of 0.63 man Sv (GW a)-' in the
UNSCEAR 1982 Report, so that an average reductionof dose from noble gas effluents of about a factor of 3
has been found owing to reductions in reported
discharge levels, mainly from BWRs. The annual
average effective dose equivalent to the most exposed
individuals in hypothetical critical groups has been
calculated at IOjlSv for the model BWR and morethan 10 times lower for rhe PWR site, taking a n
average distance of about 1 km from the site. Manyreactors give lower doses, although for some early
BWRs, the doses could be about 10 times higher.

2.

Activation gases

94. The primary activation product of interest forgaseous releases is 4'Ar. Because of its short half-lifc(1.83 h), it contributes most of its collective dose
within a few tens of kilometres of the release point,
although the exact result depends on the close-in
population density. The normalized release of "Ar
from GCRs (Table 22) between 1980 and 1984 is
2,320 TBq ( G W a)-', and the associated collective
effective dose equivalent commitment is 0.78 man Sv( G W a ) - ' , compared with the estimate in the
UNSCEAR 1982 Report of 0.95 man Sv (GW a)-'.
The reduction is due to the fact that new AGRs areproducing electricity with much lower 4'Ar discharges
than GCRs. The weighted collective effecrive dose
equivalent commitment, allowing for the fraction ofelectricity generated by GCRs, is 0.039 man Sv(GW a)-', significantly lower than the value given in
the UNSCEAR 1982 Report of 0.1 man Sv ( G W a)-'.
Because of reporting procedures, "Ar releases for
LWRs are included in noble gas data as shown in
Tables 18 and 20.
95. The consequences of the release of j5S from
GCRs have been studied in detail. The isotope isreleased in the form of carbonyl sulphide (COS),
which has a low deposition velocity and a slow
reaction rate in air. The major route of human
exposure is via milk, and the Committee estimated
2.2 lo-' man Sv GBq-' in the UNSCEAR 1982
Report. so that, using the nornlalized release of
200 GBq ( G W a)-', the normalized collective effective
dose equivalent commitment is 0.044 man Sv ( G W a)-'and the contribution weighted for GCR electricity
production is 2.4 10-3 man Sv (GW a)-', comparedwith 3.8
man Sv (GW a)-' in the UNSCEAR 1982
Report.

3. Tritium
96. The collective effective dose equivalent commitment to the local and regional population was
evaluated in the UNSCEAR 1982 Report on the basisof a specific activity model. For atmospheric releases
the Committee obtained 1.5 lo-' man Sv TBq-' byinhalation and 9 lo-' man Sv TBq-' by ingestion to
give a total of 0.0 11 man Sv TBq-' released.
97. Normalized tritium atmospheric releases for thequinquennium 1980- 1984 from PWRs are 5.9 TBq(GW a)-' (paragraph 64). giving 0.065 man Sv (GW a)-';BWR releases of 3.4 TBq (GW a)-' give 0.037 man Sv( G W a)-': HWR releases of 670 TBq ( G W a)-' give
7.4 man Sv (GW a)-' for atmospheric releases. Releases
from GCRs and LWGRs are comparable with PWRs
a n d give similar dose contributions. In summary,weighted by the amount of electricity generated byreactor type, the normalized collective effective dose
equivalent commitment for atmospheric releases oftritium is 0.53 man Sv (GW a)-', compared with theCommittee's estimate given in the UNSCE.4R 1982
Report of 0.46 man Sv (GW a)-'. For the model siteused by the Committee, the annual individual effective
dose equivalent for critical groups would be less than
1pSv from LWR atmospheric 'H releases, while the
H W R dose would be 50 pSv per year.
98. For tritium in liquid effluents, the river model
used by the Committee gave a collective effective dose
equivalent commitment of 8.1 lo-' man Sv TBq-', on
the assumption that the river is used as a source ofdrinking water. Using the normalized discharges for
1980-1984 for PWRs, BWRs, HWRs, GCRs and
LWGRs from paragraph 65. the normalized collective
effective dose equivalent commitments were calculated
and are shown in Table 34. The normalized doseweighted by the proportion of electricity generated byeach reactor type is 0.033 man Sv (GW a)-', which
compares with the estcmate of 0.04 man Sv ( G W a)-'given in the LrNSCEAR 1982 Report. The doses fromaquatic discharges are therefore about 16 times lower
than for atmospheric effluents per unit electric energy
generated, similar to the difference of a factor of 10
reported in the UNSCEAR 1982 Report.

99. The local and regional collective doses attributable to l4C releases from reactors represent only asmall proportion of the total dose commitments. The
main significance of "C stems from its entry into thecarbon cycle and the resulting global dispersion.
leading to long-term irradiation, which is consideredin chapter V. The first pass local and regional
collective dose commitment was previously assessedby the Committee using the specific activity approach
which was also used for tritium. The Committee also
assumed in the UNSCEAR 1982 Report that the formof release of 'Twas COz. The normalized local andregional collective effective dose equivalent commitment per unit release previously determined by theCommittee was 1.8 man Sv TBq-I for ingestion and
0.0003 man Sv TBq-' for inhalation following release to
the atmosphere. The normalized doses per unit electricenergy generatedare shown in Table 35 and are based on
the normalized releases taken from Table 25.
100. The normalized collective doses ranged from0.59 man Sv (GW a)-' for BWRs t o over 11 man Sv(GW a)-' for HWRs. The weighted average, allowing
for the proporrion of electricity generated by eachreactor type, was 1.6 man Sv (GW a)-' to the local
and regional population. This is about one half of theestimate of 2.8 man Sv (GW a ) - ] given in the
UNSCEAR 1982 Report, largely because of lower
reported HWR releases. For the model site the annual
effective dose equivalents to most exposed individuals
was 3 pSv for PWRs and BWRs, 10 /IS\' for GCRs,
about 70 pSv for H WRs and about 13 1tSv for LWGRs.

5. Iodine101. Releases of radioactive iodine from nuclear
power plants are small, and there is little contribution
to the local and regional collective effective dose
equivalent commitment. Because of its long radioactive half-life. 1291 enters the global cycle for iodine
and has the potential to irradiate the global population for tens of millions of years. The release of '"Icontributes only t o the local and regional collective
doses, but its assessment is complicated by thechemical form in which the iodine is released, i.e.,elemental, organic o r particulate. Elemental iodine
readily deposits on vegetation and enters terrestrial
food chains. The deposition of organic iodine isusually less than 1% of that for elemental iodine perunit time integrated air concentration. In this Annex.
as in the UNSCEAR 1982 Reporl, the Committee
assumes that 75% of the iodine released is in organic
form and 25% in elemental form.
102. In the dose evaluation used in the UNSCEAR
1982 Report, the collective effective dose equivalent comman Sv
mitment per unit release of '''1 was 4.0
GBq-'. Taking the releases of ''II from Table 26, thenormalized collective doses for "'I per unit of electricenergy generated were calculated and are shown in
Table 36. Results for the other iodine isotopes arefound by scaling from the results in the UNSCEAR
I982 Report, allowing for the change in isotopic
composition. The PWR figures are about 10% lower
than in the UNSCEAR 1982 Report, while the BWR
results are about 25% of those found previously. For
both BWR and LWGR reactors the short-lived '"I and'j51 make significant additions to the dose. The weighted
average, taking into account the proportion of electricity generated by each reactor type, is 3.3
man Sv(GW a)-'. Representative effective doses to individuals
about 1 km from the model site are about 0.5 pSv per
year for PWR releases and 4 pSv per year for the B WR
releases. As in the UNSCEAR 1982 Report, 90% of
the collective dose contribution is estimated to comefrom the milk pathway.
6. Particulates in airborne effluents
103. As noted in paragraph 79, the quantities ofradionuclides in particulate releases to the atmospheremay vary greatly, even if releases from reactors of thesame type o r those from the same reactor from year toyear are compared. Furthermore, there are several
tens of radionuclides identified in the releases. The
solution previously adopted by the Committee for
estimating doses was to assume that the normalized
releases are composed of equal amounts of activity
concentration from a range of nuclides most frequently reported in atmospheric discharges.
104. Dosimetric calculations allowed for transfer
through foodchains to man as well as external
irradiation from deposited radionuclides and inhalation from the dispersing plume of activity. Allowance
was made for uptake by roots of growing vegetation.
The full environmental modelling and resultant doses
were described in the UNSCEAR 1982 Report. The
nuclides considered were "Cr, 54Mn,59Fe, ' T o , "OCo,h'Zn. a9Sr, 9%r,
95Zr,95Nb, IZ4Sb,l3'Cs, 136"6Cs,]3713?Cs, 140Ba, !"La, :4!Ce and I4'Ce.
105. The collective effective dose equivalent commitments per GBq release of the isotopic mixture is takenfrom the UNSCEAR 1982 Report to be 5.4 lo-' man Sv(GBq)-I. with nearly two thirds coming from externalradiation from deposited activity and one third fromingestion. The collective doses per unit energy generated
have been calculated using the normalized releases
from Table 28 and arc shown in Table 37, from tvhich
it can be seen that the most important pathway is theexternal dose received from activity deposited on the
ground, followed b) the dose from ingested foodstuffs. The normalized doses cover three orders ofmagnitude. with HWRs giving the louest figure of
0.00022 man Sv (G\V a)-I and RWRs the highest valueof 0.23 man Sv (GW a)-I. The doses can be comparedwith the previous figure of 0.012 man Sv (Gki' a)-! for
PN'Rs, about one half of the current estimate. For
BM'Rs the previous figure of 0.29 man Sv (GiV a)-!
was slightly higher than the present value. GCR
figures are significantly lower than before (0.007 man Sv(GW a)-' compared with 0.012 man Sv (GU' a)-').
106. Some 9 5 5 of the collective effective dose equivalent commitment from ground deposits is delivered
within 50 years of the deposition and the major
nuclides contributing are '"Cs and 60Co. For ingestion,'OSr, 'j4Cs and I3'Cs all contribute equally by three pathways: grain, vegetables and meat. The normalized collective effective dose equivalent commitment, weighted
for the proportion of electricity produced by eachreactor type. is 0.08 man Sv ( G W a)-', essentially thesame as the estimate of 0.1 man Sv (GM' a ) - [ In the
UNSCEAR 1981 Report. Individual effective dose
equivalents from the normalized releases at the end of
a plant's operating lifetime range from about 0.01 jtSv
a t 1 km from the model BWR to 1.000 times less for
H WRs.
7.

external radiation from sediments. For discharges tomarine environments it is usually sufficient to consider
the ingestion of ocean fish and crustacea. In the
UNSCEAR 1982 Report pathways such as swimmingin contaminated waters o r consumption of unusual
food items were considered to contribute little to thecollective dose commitment.
108. The Committee has recognized the difficulty inassigning values to parameters in assessing the consequences of liquid effluents. in particular. water utilization and flow rates for rivers. fish production rates
and sedimentation rates. The assessments based on themodel used in the UNSCEAR 1982 Report, therefore,
must be regarded as merely giving a representativevalue of nuclear power impact and should not beapplied to discharges from a specific site to estimate
collective doses from that site.
109. The normalized releases for PWRs, BWRs.
GCRs. HWRs and LWGRs for 1980-1984 weresummarized in paragraph 82, and the isotopic composition for these discharges were assumed to be thoseof United States reactors shown in Table 30 and thoseof United Kingdom reactors in Table 31. Collective
effective dose equivalent commitments were evaluatedassuming the discharges took place to freshwater a n d
to marine environments. The results are shown in
Tables 38 and 39.
110. The normalized collective effective dose equivalent con~nlitment for releases lrom the PWR tofresh water was 1.6
man Sv (GW a)-', comparedwith the finding of 1 lo-' man Sv (GW a)-! in the
UNSCEAR 1982 Report. Drinking water accounted
for about 80% of the total and l"I was the major
contributing nuclide (70% of the dose). For BWRs thenormalized collec~ive effective dose equivalent is
6.6 lo-' man Sv (GW a)-', compared with the assessment of 2.8
man Sv ( G W a)-! in the UNSCEAR
1982 Report. The reduction is the result of an overall
decrease of discharge, as well as a greater reduction in
the more significant radiological nuclides. About 75%of the dose was from drinking water, and 60Co. I3'I,
1 i - 1 ~ and~I3'Cs contributed almost equally to thisdose. No results were provided for GCRs since thesewere coastal-sited. The collective dose from theseradionuclides normalized for the amount of electricity
generated were 1 . 1 10-3 man Sv (GW a)-! for PWR
and BWR releases to the model river.
1 1 1 . For PWR releases to salt water the normalized
results are shown in Table 39. The collective effective
dose equivalent commitment was 3.6 lo-' man Sv( G W a)-', about one half the value found in the
UNSCEAR 1982 Report. The fish and mollusc path\+lays were both equally important, although the mostimportant nuclide was different for each pathway:"'Cs for fish and 60Co for m o l l u ~ c ~ .

Liquid effluents

107. Aquatic releases are made into freshu.ater o rmarine environments. For releases of radionuclides
into rivers or lakes. the pathways of exposure werepreviously taken by the Committee to be drinking
water, irrigation leading to transfer to foodstuffs. and112. For RWR releases to the marine environment.
the normalized collective effective dose equivalentcommitment was 3.8 lo-' man Sv ( G W a)-L.comparedwith the figure of 4.2 lo-* man Sv (GW a)-' given in
the UNSCEAR 1982 Report. The major contribution.
was, as before, from "Zn, which concentrates inmolluscs, and therefore the marine results differedmarkedly from those for fresh water. For GCR
releases the normalized collective effective dose equivalent commitment was 0.19 man Sv ( G W a)-',essentially the same value found in the UNSCEAR
1982 Report. The majority of the dose arose fromdischarges of I3'Cs. The weighted normalized collective
effective dose equivalent commitment, allowing f o r therespective electricity generation was 0.025 man Sv( G W a)-'.
113. Again. it should be emphasized that the figures
given in Tables 38 and 39 are representative of thegeneration of unit quantity of electric energy andshould not be applied to a specific site whereparticular releases and specific environmental pathways exist that have not been considered here andmight lead to significant differences in collective dose
contributions. The normalized collective effective dose
equivalent commitment due to aquatic discharges has
been estimated to be 0.013 man Sv ( G W a)-', assumingthat one half the discharges are made to fresh waterand one half to the marine environment.

C.

OCCUPATIONAL EXPOSURES

114. As was noted in the UNSCEAR 1982 Report.
more data on occupational exposure to radiation arereported for reactor operation than for any other area.
There are difficulties in normalizing data on occupational exposure to the electric energy generated,particularly for water reactors, as most of the doses
are incurred during maintenance when no energy is
produced. Normalized results are therefore only usefully derived over several years for a number ofreactors. Average annual effective dose equivalents toreactor workers were estimated t o be similar in the
UNSCEAR 1977 and 1982 Reports and ranged from 3
to 8 mSv. During the same period. however. there wasa large increase in the number of workers per reactor
in the United States. The trend in the normalized
collective effective dose equivalent was downwards,
but oilerall the best estimate for LWRs was taken in
the UKSCE.AR 1982 Report t o be 10 man Sv( G W a)-', the same as in the UNSCEAR 1977 Report.
115. .AS can be seen from Table 1, PWRs and BWRs
have been installed in many countries, althoughinstalled capacity in 1987 was still dominated by rhe
United States. Recent data on occupational exposureand normalized collective effective dose equivalents
are given in Table 40 for PWRs a n d BWRs. For somecountries, the data are comprehensive and publishedregularly by the appropriate authorities. For othercountries, data are not available for all years o r all theunits installed. h o n e the less, the data are sufficient togive a reasonably comprehensive indication of thesituation world wide, as substantial numbers of LWRs
enter the middle phase of their predicted operating
lifetimes. In general. the data on electricity generated
were taken from the summaries produced by theInternational Atomic Energy Agency (IAEA) [I3. 14.
15, 161, if not otherwise given in the references for aparticular country.

116. A comprehensive survey of data on LWRs in
Western Europe has been carried out [B33]. The datacannot be added to Tables 40 o r 41, as the reactors
were not identified specifically by country. The normalized collective effective dose equivalent for 23 PWRs
dropped from about 6 man Sv ( G W a)-' in 1980-1981
to 4 man Sv ( G W a)-' in 1984. The comparable figuresfor 17 BWRs were more variable but were in the rangeof 3-6 man Sv ( G W a)-'. A particular study o n PWRs
has also been carried out by Lochord and Benedittini[L5]. A distinct difference is emerging between PWRs
and BWRs in the annual collective dose per reactor
and per unit electric energy generated for reactors ofsimilar electrical capacity. This trend, which is illustrated in Table 41 for PWRs and BWRs from theUnited States and Japan, becomes more apparent as
the reactors enter the second decade of their operating
life and has been reported in the Federal Republic of
Germany [E7]. Japan [19. T121, Sweden [P7] and theUnited States [N3]. The collective western Europeandata, however, d o not support the conclusion [B33].
The collective dose can be higher in BWRs than in
PWRs by up to a factor of 2, possibly because more
maintenance work has to be performed in radiation
areas on BWRs, especially around the turbines.
117. It appears that the significant trend towards
increasing numbers of workers per reactor, which wasnoted in the UNSCEAR 1982 Report, especially in the
United States, levelled off in the early 1980s. This
aspect has been studied in detail by the Nuclear
Regulatory Commission [N3]. who showed that
although the number of workers per reactor doubled
from 600 to 1,200 over the period 1975-1980, it
remained constant at the higher figure during theperiod 1980-1983. The collective effective dose equivalent distribution ratio (see paragraph 19) was
assessed separately for PWRs and BWRs [N3]. For
both types of reactors the average values of the ratios
for the years 1981-1983 were in the range of 0.4-0.6.
Annual average doses have been quoted for the years
1980-1982 at three PWRs in the USSR [V2, V3]. The
values range from 4-8 mSv and provide detailed
support for the overall average figure of 5.6 mSv given
by Varobyov [Vl] and used in Table 40. F o r the
Novovoronezh PWR [V5], however. the normalized
collective effective dose equivalent in 1980 of 3.1 man Sv( G W a)-', based o n an annual collective effective dose
equivalent of 4.0 man Sv and an electric energy
generated of 1.3 G W a [All was somewhat lower thanthe overall figure of I I man Sv (GW a)-' used inTable 40 [Vl]. Data on doses to personnel at LWGRs
in the USSR have been reported at some reactors [P4,B151. For the two reactor units at Kolskaya. thecollective dose in 1980 of 2.3 man Sv was typicalof earlier years: the normalized collective dose of
1.9 man Sv ( G W a)-' for 1980 was some\vhat lower
than the average value of 2.5 man Sv ( G W a)-' for
1977-1980. The annual average dose to personnel was
about 5 mSv over that period. The data for Japan inTables 40 and 41 were compiled mainly from detailed
statistics supplied by Kumatori [K 11 and Terasima[TI 21.
118. Recent data for HWRs in Canada are shown in
Table 42 [A4]. The values include internal doses from

exposure to tritium. The electric energy generated was
obtained from IAE.4 tabulations [I3, 14, 15, 161. The
results for Canada show considerably higher collective
effective dose equivalent per unit energy generated
than would be obtained by considering only the twolargest power plants that produce the bulk of theenergy. The lifetime collective effective dose equivalent
generated by the Atucha 1 HWR in Argentina has
been estimated at 27 man Sv (GW a)-' [P8].
119. Most of the GCRs in the world are in the
United Kingdom. Comprehensive data on them [HI,H2, H3. H4. P5, W2] are shown in Table 42. The stepchange in the number of workers after 1980 is because
the figures prior to 1981 did not include workers notdirectly employed by the Central Electricity Generating
Board. The collective effective dose equivalent distribution ratio has been low: less than 0.01 in 1984 [P5].
The GCR in Japan is of a similar design to the early
GCRs in the United Kingdom [Kl]; that in the UnitedStates is a small high temperature gas cooled reactor[B27, N3], so the doses from it are not directlycomparable to the others.
120. Collective doses to personnel at the Dounreay
establishment in the United Kingdom concerned with
operation of the prototype FBR were 0.15 man Sv in1984 and 0.29 man Sv in 1985 [U3]. Most of thiscollective dose resulted from charge machine refurbishing and irradiated fuel cell entries. Although many
aspects of the design and operation of FBRs werereviewed at a recent symposium [I 1 I], no data on
occupational exposures were reported. The Committee
would welcome more information on this aspect,
especially from prototype and nearly commercial-scale
reactors.
121. Despite the lack of data from some countries
and for some years. there is enough information fromthe countries for which each reactor type is installed inlarge numbers to make a reasonable estimate of thenormalized occupation dose for the quinquennium
1980- 1984 for the major reactors. These estimates,
which are based on the data in Tables 40 and 42, aregiven in Table 43. Bearing in mind the world-wide
predominance of LWRs, the overall estimate must be
heavily weighted by the estimate for this reactor type,but a figure of 10 man Sv (GW a)-' does not appear
unreasonable.

short-term storage at reactors for initial decay heatremoval. In addition, waste may include some structural components from the core or fuel, such as theouter fuel element structures from GCR and AGR
fuel elements.
123. The radionuclides present in the above wastes
are fission products, activation products and actinides,
the particular radionuclides, quantities and relative
activities being dependent on the reactor type, thestate of the fuel cladding, the levels of corrosion. etc.
In most cases these wastes will be the bulk of what isnormally classified as intermediate-level wastes (ILW),
i.e., wastes containing substantial activity concentrations but not significantly heat generating.
124. The other main cause of solid wastes during
operation is the protective material of various kinds
used around the station. Much of this is burnable andconsiderable volume reduction can be achieved byincineration and compaction. The radionuclide composition is even more variable than for the wet wastes
and the activity concentrations are small to zero.
These wastes are generally classified as low-level\vastes (LLW).
125. In order to characterize the wastes for analysisof the impact of disposal, it is necessary to determine
the volumes and the activity concentrations withidentification of the relative quantities of importantradionuclides, especially long-lived radionuclides andany actinides.
126. There were a number of studies in the mid1970s on the quantities of wastes produced at LWRs
[B30, M7, M8]. The results, summarized in Table 44,
have been extracted from a review carried out by theEnvironmental Protection Agency of the United Stares[E5]. More recent reviews have been carried out in the
United States by the Nuclear Regulatory Conlmission[N8, N9] and by the United States Department ofEnergy [D4]. as well as in other countries using LWRs
or planning to do so [NIO]. On the basis of thesestudies, the quantities of conditioned wastes arisingfrom LWRs per unit energy generated are assumed to
be as shown in Table 45. These are only approximate;variations of up to an order of magnitude are possible
in particular circumstances, depending on the type oftreatment or conditioning used.

1. Solid waste production

127. The assumed radionuclide compositions for thewastes in Table 45 are shown in Table 46. These arebased primarily on the analyses reported in the UnitedStates [N8, N9, E6] and the United Kingdom [PIO].

122. During operation of a power station, solid
wastes are generated in a number of ways. In LWRs
the main cause is treatment of the circulating water,
giving rise to spent ion exchange resins, filter sludges
and evaporator concentrates. Although these are
originally wet wastes and may be stored in this form
at the site, they are generally solidified before disposal.
A similar type of waste arises from purification of thewater in spent fuel storage ponds at reactors. Even
though fuel elements are eventually removed for longterm storage or reprocessing, provision is made for128. The quantities and activity concentrations ofoperating wastes of HWRs derived from data given
for Canadian CANDU reactors [B31] are assumed to
be as shown in Table 47. The quantities of wastes
from operational GCRs have recently been reviewed[F3], and the results are also summarized in Table 47.
The main differences in radionuclide composition
from the LWR wastes are the higher alpha activity of
the Magnox reactor sludges and the graphite debris
containing 14C. Although there are significant differences between reported inventories for Magnox

D. SOLID WASTE DISPOSAL

reactors and AGRs [PI21 and LWRs, the compositionin Table 46 is taken for this preliminary study, given
the predominance of LWRs.

2.

Solid Haste disposal facilities

129. A large proportion of the LLW produced at all
facilities during operation can be disposed of by burial
at a shallow depth. Burial facilities range from simple
trenches or pits containing untreated wastes andcapped with soil. to concrete structures containing
conditioned wastes and capped with weather-resistantmaterials. These will be referred to as trenches andengineered disposal facilities.
130. Considerable quantities of LLW have been
disposed of in such facilities throughout the past few
decades. Many of the earlier disposal sites were notused for disposal of wastes from the generation of
nuclear power. except perhaps for some research anddevelopment aspects. For example, there are 14 sites
in the United States operated by the United States
Department of Energy for the disposal of wastes
generated from certain defence research activities.
Some major closed and currently operating LLW
burial sites are shown in Table 48 [C8, C9, N9]. These
have accepted wastes from a range of operations[H 15, M 101.
131. Typical simple trenchs are about 10 m deep and
25 m wide and could be from 100 to 200 m in length,
depending on the site. They are covered by about 1 mof compacted soil. The waste is not conditioned except
to render the material non-combustible where necessary.
This is similar to the minimum engineered trench
specified by Pinner [Pll] and the base case of the
Nuclear Regulatory Commission of the United States[N9]. Some major routes by which radionuclides will
be released from such a trench will be into rainwaterpercolating through the trench and into ground water.
There will be considerable differences between the
behaviour of elements that form easily soluble compounds, such as iodine, and those that do not, such asuranium. as well as a marked dependence on theenvironmental and hydrological conditions of the site.

132. On the basis of knowledge of the chemical
behaviour of the elements of which there are important radionuclides, the release behaviour of the wastes
can be classified into three groups [PI I]. It is also
assumed that. since LLW is usually disposed of intrenches without packaging, radionuclides begin to bereleased into water as soon as the site is closed. The
reference site is assumed to be above the water table inreasonably permeable, weathered material that has anunderlying less permeable rock. In a site with thesecharacteristics. water filtering through the waste willtend to move down through the unsaturated zone
until it reaches the water table and the impermeableboundary where i t moves downslope. I t is assumed toreach a stream at a distance of 2,000 m from the site.
133. Some categories of ILW containing radionuclides
with longer half-lives or at activity concentrations too
great for disposal in simple trenches have been

disposed of in engineered shallow disposal facilities. A
typical facility is an excavation about 20 m deep and
25 m wide lined with 1 m of concrete. Such facilities
are filled with concreted wastes to about one half theirdepth. the interstices being filled with concrete andfinished with layers of concrete and clay to form animpervious cap. The canisters and concrete around thewastes will prevent rain or ground-water access for aconsiderable time, which is taken to be 100 years.
After this time it is assumed that all radionuclides are
released into percolating water at a constant fractional
rate of lo-: a-l.
134. As a result of the greater depth of emplacement.
it is likely that engineered facilities would be positioned
below the water table. I t is also sensible to locate thefacilities in materials with good sorption properties, so
the reference site is assumed to bc in clay. M a n clay
outcrops are associated with harder, more permeable
rocks leading to artesian conditions, i.e.. rising groundwater. The trench would interfere with this locally sothat the eventual flow pattern assumed for thereference site is that water infiltrating the trench fromabove will tend to move downward, then upwards andoutwards, eventually entering streams at a distance ofabout 1.000 m on either side of the site [PIO].
135. An alternative method of disposal for packaged
solid wastes is to dump them into a sea-bed atconsiderable depth. Although such disposals werecarried out for many years, they ceased in 1982 undera temporary moratorium. The amounts of wastes
disposed of to this date in the north-east Atlantic havebeen summarized by NEA [N6] and are given in
Table 49. I t is not possible to assign the wastes to aparticular power programme. and it is known thatsome of the major radionuclides, such as "C. arose aswastes in the form for sea dumping largely from the
preparation of radiopharmaceuticals.

3.

Collective dose commitments

136. After closure of a burial facility. there will be aperiod during which control over the site is maintained.
This does not necessarily preclude the transport ofradionuclides released from the wastes into percolating
ground water but could reasonably be relied upon toprevent major human intrusion into the site, such as
for building purposes. Thus, during the controlled
period, taken to be 100 years. only release by watercontact is considered; other pathways are assessed
after this period. The major pathways possibly leading
to exposure are shown in Figure 11.
137. The actual transport of radionuclides withground water, after release from the waste, thecontainer and any surrounding engineered structures,will be very dependent on the hydrogeologic characteristics of the site. Considerable effort is being
devoted to the development of calculational techniques
capable of handling detailed knowledge of particularsites. For this study a more general approach isappropriate, such as the one adopted in other genericappraisals of shallow land burial [PIO, N9. N 101.

External
irradiation

I

Inhalation

i
Crops

Farming

Ingestion

Animalsr

GroundwaterContactby water

I

Soi 1

I

Surfacewater

I [Figure 11.

'

Seas

1
4

Marinepathways,

1

General release and transport mechanism and pathways from shallow burial facilities.

138. The simplest representation of ground waterflow velocity. caused only by the natural hydraulicgradient. is that given by Darcy's law:where V, is the ground water flow velocity, K, is thehydraulic conductiiity. k, is the hydraulic gradientand i L is the kinetic porosity. This is the basis forseveral transport codes, such as FEFLON'. which isused in the Nordic study [N 101, GEOS. which is usedby the Yational Radiological Protection Hoard of theUnited Kingdom [H12]. and that used by the Nuclear
Regulatory Commission of the United States [N9]. I t
is adapted to radionuclide transport through a porous
medium by including a retardation factor or distribution coefficient.
139. Generic assessments have also been carried out
using a somewhat more realistic two-dimensionalmodel, NAMMU [R4]. to calculate the pressure head
distributions and, hence. flowpaths and velocities insaturated porous media. This has been applied tomigration through undisturbed clay and to movementin the surface soil layer [PIO]. Whatever calculational
method is used, the general result is that those
nuclides with small retardation factors o r distribution
coefficients, such as tritium, I4C, 99Tc and '291, moveat a velocity close to that of the ground water.
whereas nuclides with large retardation factors o r
distribution coefficients. such as 23sU. '36U and 2'9Pu.
move very slowly. The values for some importantradionuclides are intermediate. Values adopted inthree major studies [H12, N9. NlO] are reasonably
consistent.

140. The output from the radionuclide transport
calculations is the rate of input of activity into either
the nearest stream, as described for the generic site, o r
via ground water into soil that could be used forfarming. b7ater could also be abstracted via a well. In
calculating doses i t is assumed that the water forms asource of drinking water for humans and animals. It isalso assumed that fish from the stream are caught a n deaten. The river model is a compartment type. witheach compartment rspresenting a homogeneous freshwater body and incorporating adsorption on to andresuspension of sediment particles [I 151. The flow rateof the river is taken to be 6 lo6 m' a-'. The eventual
transfer from rivers via estuaries to the sea is also
included, In assessing doses from drinking water. it isassumed that suspended sediments are removed byfiltration. Collective doses from streams and wells areassessed on the assumption that 0.2% of the flow rateand 1% of the abstraction rate are actually ingested.
1-11. If the land is used for farming. this will give rise
to a large number of exposure pathu.ays. The contamination can result from transfer directly upbvards
through the soil from ground water or via streams andrivers through irrigation. The calculation of collective
doses requires an esrimate of the total quantities ofeach foodstuff consumed. shown in Table 50. together
with average values for activity concentrations obtained
from the radionuclide transport models.
142. The collective dose equivalent rates per unit
activity as a function of time after release from anengineered facility via all the pathways are shown fora number of important radionuclides in Tables 51, 52.

and 53. for three major time periods of interest. In
general, farming and water consumption pathways
both contribute significantly to the collective dose.
143. The results of applying the specified models areshown in Tables 54 and 55 for the shallow earth
trench and the engineered trench. respectively. The
results are presented per unit activity in the trench andshow the collective effective dose equivalent commitment and the maximum collective effective dose
equivalent rate. Also shown is the time at which thespecified percentage of the maximum collective effective dose equivalent rate is reached [S 151.
144. Using the estimated volume for LLW from
Table 45 of 200 m3 (GW a)-I at an activity concentration of 1 GBq m-' as appropriate for PWRs and assuming the radionuclide composition given in Table 46. it
can be seen from Table 54 that the normalized
collective effective dose equivalent commitment fromburial of these relatively short-lived wastes is less than
10-lo man St' (GW a)-'. Only if long-lived radionuclides were present could there be a collective doseof any significance; if this proportion were taken to be
one thousandth of the quantity present in ILW, as
shown in Table 46, the normalized collective effective
dose equivalent commitment would be aboutman Sv (GU' a)-'.
135. Taking the estimated volume for ILL\' from
Table 35 of 50 m' (GLV a)-' at a n activity concentration of 100 GBq m-'. again. as appropriate for PWRs,
and combining this with the data from Table 55 forthe radionuclides specified to be present in Table 46,
the normalized collective effective dose equivalentcommitment from disposal of I L W in a model
engineered trench is 0.5 man Sv (GI\' a)-'. The main
contribution is from the 0.1% by activity of lJC
assumed to be present in the waste. The contribution
by radionuclide is shown in Table 56.

I .

FUEL REPROCESSING

146. At the fuel reprocessing stage of the nuclear fuelcycle, the elements uranium and plutonium in theirradiated nuclear fuel are recovered to be used again
in fission reactors. Spent fuel elements are stored
under water. which provides both biological shielding
and cooling. while waiting to be reprocessed. Fuelelements are usually left until the short-lived "'I has
decayed to a low level. normally a minimum of fouro r five months. Since one reprocessing plant can serve
large numbers of nuclear reactors. the quantities of
nuclides passing through the plant that are significant
from the point of view of health will be high inabsolute terms. Careful design limits discharges, however, so that releases per unit of electricity generated
by the fuel passing through the plant, i.e.. (GW a)-],may be relatively small.
147. The only commercial operating reprocessing
plants are at Sellafield (formerly Windscale) in the
United Kingdom and Cap de la Hague and Marcoulein France. The capacity at Sellafield is 2,000 t a-I(heavy metal) and that of Cap de la Hague is 900 t a-'oxide fuel. while the plant at Marcoule processes u p to
400 t a-I of GCR metal fuel. The annual throughput
of irradiated fuel from civilian power programmes inthese three reprocessing plants is currently equivalent
to about 8 G W a of electric energy, representing about
5% of the reported annual nuclear electric energy
production (189 G W a. Table 2). Thus, the majorityof irradiated fuel, which arises from LWRs, is not
reprocessed but is stored pending future policy decisions as to whether to dispose or reprocess. A
summary of the attitudes of countries with powerreactors towards reprocessing is given in Table 57.
while in Table 58 national programmes for commercial
reprocessing are given [C3].
A.

EFFLUENTS

148. The design and operation of reprocessing plants
to avoid releases of large amounts of radioactive
material is complex. The gaseous and volatile fission
product elements (iodine. tritium, carbon, krypton.
ruthenium, technetium. xenon and caesium) are largely
separated from the fuel when it is dissolved in nitricacid. The dissolver off-gas is treated for nitric acidrecovery and iodine removal before being mixed withthe off-gases from other stages in the process. The\lessel off-gases are treated by caustic scrubbing,
drying and filtering through high efficiency filtration
systems before being discharged to the atmosphere.
The aqueous wastes containing almost all the fission
products and transuranic elements are concentrated
by evaporation and stored in double containment
stainless steel tanks before they are treated further.
149. The radionuclides of principal concern in reprocessing effluents are the long-lived nuclides: 'H.
14C. 8 5 ~ ~1291. , IJJCs, "'CS
and isotopes of transuranium elements. Table 59 lists the reported discharges to the atmosphere. and Table 60 those inaquatic releases, from Sellafield, Cap de la Hague andMarcoule for 1980-1985. The amount of activity in theeffluents depends upon the specific waste treatment
and processing design of the plant. as well as the typeof fuel processed, its irradiation history and storage
time prior to reprocessing. Table 61 gives the isotopic
composition of liquid effluent discharges from the
Sellafield and Cap de la Hague plants in greater detailfor the years 1980-1985. Atmospheric release data andliquid discharge data for Marcoule are not available
beyond 1980.
150. The throughput of fuel at both Sellafield and
Cap de la Hague has been calculated on the basis of"Kr reported discharges and on the assessment of the
L5Krgeneration in different reactor fuel cycles made in
the UNSCEAR 1982 Report. In that Report, theCommittee has used production rates of 14 PBq(GWr a)-I for GCRs and 11.5 PBq (GW a)-' for PWRs.
These figures make assumptions about fuel bum-up
and reactor thermal efficiency that are not likely t ohave changed significantly since the UNSCEAR 1982
Report. On this basis. the electric energy productionof the annual throughput of fuel at Sellafield hasvaried between 1.7 G W a and 3.7 G W a, while forCap de la Hague the range has been 2.4 to 6.1 GW a.
For Marcoule, there are little data on atmospheric discharges. although the electric energy of annual fuelthroughput for 1980 has been estimated at 1.4 G b ' a.
151. For tritium, the Committee used in the
UNSCEAR 1982 Report a production rate in LWR
fuel of 0.75 PBq (GW a)-I and, assuming this applies
to GCR fuel, the inventory passing through Sellafieldhas varied between 1.3 PBq ( 1985) and 2.8 PBq (1981).
In 1981, atmospheric discharges of 3H were 0.46 PBq
and liquid discharges 2 PBq, giving 2.46 PBq, comparedwith the estimated throughput of 2.8 PBq. Thus, itappears that nearly all the tritium in irradiated fuel isreleased in reprocessing and about 20% is released tothe atmosphere. This is identical to the percentage
estima~edin the UNSCEAR 1982 Report. The reniaining tritium may be immobilized in cladding wastes.
For Cap de la Hague the normalized releases oftritium have been 0.26 PBq (GW a)-' of which only
about 1% is in reported atmospheric releases and thus
it seems that only about one third of the throughput isreleased.
152. The results of routine measurements of I4C
atmospheric discharges from the Sellafield reprocessing
plant are given in Table 59. The normalized production
rate of I'C in GCR fuel was estimated in the
UNSCEAR 1982 Report at 3.2 TBq (GWr a)-'. Atmospheric discharges from Sellafield therefore seem toaccount for essentially the whole of the estimatedthroughput of I4C between 1980 and 1985. For the
French reprocessing plants, 14C discharges are notreported. In the UNSCEAR 1982 Report, the Committee estimated the I4C content of LWR fuel to be
0.66 TBq (GW a)-'. of which about 75% was assumed
emitted to the atmosphere, but in view of the
Sellafield data. all I4C can be considered released tothe atmosphere for the dose assessment.
153. The 13'1 content of irradiated nuclear fuelvaries, depending upon the cooling time and the final
power level of the fuel discharge. The "'I normalized
content of LWR fuel cooled for six months isestimated at 2.8 TBq (GW a)-I, falling by a factor of
2.000 for a cooling period of nine months. Since
irradiated fuel is generally cooled for at least a year
prior to reprocessing, "'I discharges are very small.
For 1980-1985. Sellafield atmospheric releases of l 3 I l(Table 59) gave normalized values of 19 GBq (GW a)-I;
the corresponding figure for 1975- 1979 was 1.7 GBq(GW a)-'. the increase being due to a high releasefigure in 1981.
154. The quantity of "'1 in fuel depends upon burnup and is assessed at 37-74 GBq (GW a)-I. Atmospheric discharges of "'I, as well as liquid effluent
amounts, have been reported for Sellafield and Cap dela Hague but not for Marcoule. The normalized
atmospheric release is 3.7 GBq (GW a)-I for Sellafield
and 4.9 GBq (GW a)-' for Cap de la Hague for theperiod 1980-1985, which is about twice the value givenin the UNSCEAR 1982 Report. Liquid effluents
averaged about 30 to 60 GBq (GW a)-! at each plantover the same period, compared with 40 GBq (GW a)-'previously reported. I t seems likely that all the 12q1 in
the fuel is released with a few per cent going to theatmosphere.
155. Atmospheric releases of aerosols are summarizedin Table 59. The normalized alpha releases from
Sellafield are 0.2 GBq (GW a)-I, of which more than
75% is due to plutonium isotopes [BI, B2. B3. B8.
B16], the remainder being accounted for by 24'.4mand
242Cm.This figure is one half that reported in the
UNSCEAR 1982 Report. The alpha-aerosol results areavailable from France for Cap de la Hague and aresome 30 times lower. For atmospheric beta releases.
the largest component from Sellafield is I3'Cs, although
since 198 1 the levels have been reduced. The normalized
release is 63 GBq (GW a)-', compared with 88 GBq(GW a)-I for 1975-1979. For Cap de la Hague thenormalized release is 0.04 GBq (GW a)-' for betaaerosols. and no isotopic breakdown is available. The
reduction in aerosol releases in recent years from
Sellafield is the result of improvements in the Magnox
cladding silo stores, including the installation of inert
gas blankets and filtration systems.
156. The liquid effluents discharged from Sellafield.
Cap de la Hague and Marcoule are given in Table 60
for total alpha, total beta, 'H. '"Sr, Io6Ru and I3'Cs.
There is a yearly isotopic breakdown for the French
plant at Cap de la Hague but not for Marcoule. The
isotopic compositions of Sellafield and Cap de laHague discharges are given in Table 61 for 1980-1985.
157. The normalized alpha release from Sellafield tothe sea is 8.0 5.2 TBq (GU' a)-', compared with anaverage of 25 TBq (GW a)-' between 1975 and 1979.
For Marcoule and Cap de la Hague the figures are
0.063 and 0.16 TBq ( G b l a)-!. while for the previous
period they were 0.016 and 0.24 TBq (GU' a)-I. Mostof the Sellafield alpha activity was 239"2'0P~,
and thelevel of alpha discharge has been reduced by a factorof 6 over the reporting period.

+

158. For liquid discharges of beta activity thenormalized releases from Sellafield, Cap de la Hagueand Marcoule are 0.97, 0.24 and 0.027 PBq (GW a)-!,respectively. compared with 3.7, 0.52 and 0.04 PBq(GW a)-' for 1975-1979. The isotopic composition of
the effluents varies between the sites: 55-70% of the
Sellafield discharge is attributable to 13'Cs, whereas
40% of the Cap de la Hague discharge is attributable
to Io6Ru.Tile "'Cs levels from Sellafield were reduced
by a factor of 9 over the review period. although the'06Ru levels remained constant at about 400 TBq(GW a)-' until 1985. After 'H. Io6Ru is the mainisotope released from Cap de la Hague; the lWRu
discharges are comparable to those of the Sellafield
plant.
9
Monitoring of the marine environment is undertaken by regulatory authorities to ensure compliancewith authorized discharges and to ensure that doses toexposed populations are at the levels predicted. The
results of monitoring around the United Kingdom inthe vicinity of all operating nuclear plants have been
published by Hunt [H5, H6, H7]. The most significant

results arise from discharges of the Sellafield plant.
Measurements of activity in fish and shellfish in 1983
are shown in Tables 62 and 63 for various locations
around the United Kingdom. In order to interpretthese results, consumption data are required to assess
intakes of radionuclides.
160. Aarkrog [A21 has summarized bio-indicatorstudies in Nordic waters to identify levels of radioactive contamination. The marine bio-indicators are
the blue mussel (Mytilus edulis) and bladder wrack(Fucus vesiculosus). which are sensitive to contamination from nuclear fallout and from Sellafield discharges and nuclear power plants in Sweden. Finland
and the rest of coastal northern Europe. Discharges
from Sellafield have been traced from the lrish Sea.
along the western Norwegian coast, down along
eastern Greenland and then western Greenland. The
transit time from the lrish Sea is measured to be four
years and the activity concentration is diluted by afactor of 100.
161. The measured concentrations of "'Cs in sea
water decrease from the highest levels of 1.000 Bq kg-'near Sellafield to 8-10 Bq kg-' in the Baltic Sea,
1-2 Bq kg-' near Greenland and less than I Bq kg-' near
Iceland. Levels of 99Tcfrom Sellafield discharges closely
follow those of I3'Cs. Measurements of plutonium showenhanced levels primarily in British and lrish coastal
waters. although very low levels have recently been
detected in areas further from the coast.

B.

LOCAL AND REGIONAL COLLECTIVE
DOSE COMMITMENTS

162. The evaluation of the collective dose commitments from reprocessing nuclear fuel requires a studyof the local and regional effects and of the global
consequences of the releases. Estimates of the local
and regional collective dose commitments are given inthis section and the global contribution is provided inchapter V. The collective dose commitments areevaluated for the normalized discharges from Sellafield
and Cap de la Hague by scaling the normalized results
given in the UNSCEAR 1982 Report. As there are
only three reprocessing plants operating with significant commercial throughput of fuel, the collective
effective dose equivalent commitments per unit ofelectric energy generated are weighted by the fractionof fuel reprocessed to provide the current contribution
from all operating reactors. In the UNSCEAR 1982
Report. the Committee gave typical discharge figures
for notionzl new designs of reprocessing plant. This
has not been repeated in this Report since all fuel may
not be reprocessed. The weighted average. therefore.
reflects actual exposures from the nuclear fuel cycle ascurrently operated.

163. The averaged 8'Kr normalized discharge from
Sellafield between 1980 and 1985 was 14 PBq (GW a)-'(Table 59), and the collective effective dose equi-valent commitment obtained by the Committee in the
UNSCEAR 1982 Report was 0.0074 man Sv (PBq)-I.
Thus. the normalized local and regional collective
effective dose equivalent commitment is 0.1 man Sv(GW a)-'. The normalized discharge from Cap de laHague is 11 PBq (GW a)-'. giving 0.08 man Sv(GW a)-'. The average annual electric energy generated
in recent years has been over 160 GW a, and anannual amount of fuel equivalent to 8 GW a was
reprocessed. Thus, the normalized collective effective
dose equivalent commitment fromis 0.005 man
Sv (GW a)-' electric energy generated.

2.

Tritium and carbon-14

164. The Committee used in the UNSCEAR 1982
Report specific activity models to estimate collective
doses from 'H and 14Cdischarges. The dose resulting
from the release of tritium to the atmosphere wasestimated in that Report at 0.0027 man Sv TBq-I for
the Sellafield site. some four times lower than thevalue for the reactor site, due to differences in site,
population density and meteorological conditions.
Releases from Sellafield to the atmosphere averaged
I20 TBq (GW a)-' (Table 59), giving a collective
effective dose equivalent commitment of 0.32 man Sv(GM'a)-', 85% of which was from ingestion. For Cap
de la Hague the normalized release of 3.5 TBq(GW a)-' gives a collective effective dose equivalentcommitment of 0.01 man Sv. Releases to the regional
marine environment were estimated in 1982 to lead tolower dose commitments. Using the value derived in
the UNSCEAR 1982 Report of 1.8 lo-' man Sv PBq-I
released to oceans and the average release to the seaof 579 TBq (G W a)-' for 1980-1985 from Sellafield(Table 60) leads to a collective dose commitment of 1
lo-' man Sv (Gw a)-'. The total normalized collective
effective dose equivalent commitment weighted by the
relative energy of fuel reprocessed at Sellafield and
Cap de la Hague is 0.15 man Sv (G W a)-', essentiallythe same figure that was given in the UNSCEAR 1982
Report. Again, allowing for the fraction of fuelreprocessed. the weighted normalized collective effective dose equivalent commitment is 0.007 man Sv(GW a)-'.
165. For "C releases to the atmosphere, the Committee estimated the collective effective dose equivalentcommitment at 0.4 man Sv TBq-I, with essentially thesame value per TBq released to the aquatic environment. Averaged atmospheric releases from Sellafield arereported to be 3.5 TBq (GW a)-', giving a normalized
collective effective dose equivalent commitment of
1.4 man Sv (GW a)-', compared with 0.69 man Sv(GW a)-' quoted in 1982. The difference is accounted
for by the reported discharges to the atmosphere being
double the values reported in 1975-1979. It would
appear that the total throughput of I4C at Sellafield is
now accounted for in atmospheric releases. For Cap
de la Hague. the assumed release of 0.66 TBq (GW a)-'gives a normalized collective effective dose equivalentcommitment of 0.3 man Sv (GW a)-'. Weighted forthe fraction of fuel reprocessed. the contribution is
0.04 man Sv (GW a)-'.

3.

Other atmospheric releases

166. Of the other nuclides released to the atmosphere.
Iz9I becomes globally dispersed and makes a contribution to the collective dose commitment over a
prolonged period, while the remainder contribute only
to the local and regional collective dose commitment.
A summary is given in Table 64. The total, averaging
over Sellafield and Cap de la Hague, amounts to
1.3 man Sv (GW a)-', compared with the assessmentof 3 man Sv (GW a)-' for atmospheric releases madein the UNSCEAR 1982 Report. Some 65% of the doseis now due to ''C discharges, which are reported to be
twice the previous levels and which counteract reductions in discharges of other nuclides, particularly
actinides. Weighted by the proportion of electricenergy generated, the normalized collective effective
dose equitralent commitment from atmospheric releases
during reprocessing is 0.07 man Sv ( G W a)-'.

1. Liquid effluents

167. The results are presented in Table 65 for thenormalized collecti\~eeffective dose equivalent commitments for marine discharges from reprocessing at
Sellafield and Cap de la Hague. The environmental
dosimetric models are appropriate for the specificcoastal waters of northern Europe and were fully
described in the UNSCEAR 1982 Report. They
assume that consumed fish, molluscs and crustacea
are the important food pathways to man.
168. For Sellafield, normalized liquid discharges for
1980-1985 were reduced by a factor of 3 since theperiod 1975-1979. The collective effective dose equivalent commitment per TBq for marine discharges
from Sellafield found by the Committee in the
UNSCEAR 1982 Report was 0.068 man Sv for "'Cs.
0.034 man Sv for 'ObRu and 0.025 man Sv for alphaemitters. The principal route of exposure is I3'Cs in
consumed fish, as before. and for 1980-1985 thecaesium contribution is some 85% of the totalcollective dose. The normalized collective effective
dose equivalent commitment for 1980-1985 fromliquid discharge from Sellafield is 44 man Sv (GM' a)-'.
the estimate made in the UNSCEAR 1983 Reportbeing 124 man Sv ( G W a)-', If data for 1985 alone aretaken, the normalized release gives a collective effective
dose equivalent commitment of 25 man Sv (GU' a)-'.
reflecting the lower discharges after the installation ofa new plant to remove radioactive substances from
effluent streams.
169. In the case of C a p de la Hague, discharges also
seem t o have been reduced. The Committee's models
used in the UNSCEAR 1982 Report gave doseconversion factors per TBq released from Cap de laHague of 0. I man Sv for '06Ru. 0.09 man Sv for "'Cs
and 0.4 man Sv for alpha-emitters. The normalized
collective effective dose equivalent commitment shownin Table 65 is 1 1 man Sv ( G W a)-', which compareswith the figure of 53 man Sv ( G W a)-' given in the
UNSCEAR 1982 Report. The majority of the dose
arises from the discharge of lobRu.

170. The collective effective dose equivalent commitment weighted for the relative amount of electricity
produced by the fuel reprocessed at each plant is thus
25 man Sv ( G W a)-' and, after allowing for the
proportion of fuel reprocessed commercially. thenormalized contribution is 1.2 man Sv ( G W a)-'.
Annual committed effective dose equivalents to thecritical group of winkle eaters close to the Sellafieldsite were reported to be 0.5 mSv in 1985 [B29]. The
doses are reduced as discharge levels fall.
C.

OCCUPATIONAL EXPOSURES

171. I t was noted in the UNSCEAR 1982 Reportthat experience of fuel reprocessing is limited to a few
countries and that plant design and historical operating
conditions may not represent the best current potential
for new plants. This view is supported by a recent
review of the trends in the annual collective a n d the
maximum individual occupational doses in a numberof reprocessing plants [B22]. The review covered not
only the large operating reprocessing plants at C a p dela Hague and Sellafield. but also the pilot plants
WAK at Karlsruhe, Federal Republic of Germany. theEurochemie plant at Mol. Belgium, the PNC plant at
Tokai Mura in Japan, and the Idaho and SavannahRiver plants in the United States. Although recognizing
the differences in sizes and design age of the various
plants and that some of them reprocess fuel for
military as well as civilian purposes. a downward
trend in average doses was observed starting during
the period 197 1 - 1973 and ending during the period
1980- 1982. The annual average effective dose equivalent dropped from 4-15 mSv in the early 1970s to
2-4 mSv early in the 1980s. Data since the UNSCEAR
1982 Report are summarized in Table 66 for Japanand the United Kingdom [AS. 813. B23. 828. H8].
The data for Japan refer only to the PNC plant at
Tokai Mura. An estimate of 0.5 man Sv has been
made of the neutron collective dose equivalent at
Sellafield in 1982 [B24]. Data for Cap de la Hagueand Marcoule from 1973 to 1985 are given in Table 67.
taken mainly from the recent comprehensive re\,ie\v by
Henry [H 131. This also shou's annual average effective
dose equivalents of about 2 mSv in the period 1982- 1985
at both establishments.
172. In the UNSCEAR 1982 Report the normalized
collective effective dose equivalents for the plants at\tlindscale (now Sellafield). United Kingdom. and C a p
de la Hague, France. were estimated to be 18 and
6 man Sv ( G W a)-', respectively. Some revised estimates
for the United Kingdom are given in Table 66. based on' X r discharges related to energy throughput and a fuelcontent of 14 PBq (GU' a)-'. The normalized value for
Cap de la Hague is reported to have fallen from 6 man
Sv ( G W a)-' in 1975 to I man Sv (GW a)-' in 1985
despite a large increase in reprocessed fuel throughput
over this period [BZZ]. This is in agreement with thedata given in the report of a working group [C6] for aperiod leading up to 1981 and supplemented in areport to the Sizewell B public inquiry in the United
Kingdom [ZI] with data for 1982 and 1983. Theseestimates are in agreement with the detailed results for
C a p de la Hague reported by Henry [HI31 and given

in Table 67. The Table shows that the normalized
collective effective dose equivalent dropped steadily
from 2.2 to 0.9 man Sv (GW a)-' throughout theperiod 1980-1985. The difference of nearly an order ofmagnitude between the normalized values for the two
major installations makes it difficult to make a clear
estimate. It seems, however, that the estimate in the
UNSCEAR 1977 Report of the global collective
effective dose equivalent per unit electric energy
generated is, at 10 man Sv (GW a)-I, too high. Basedon the trends reported for Cap de la Hague, theestimate for Marcoule in 1980. the experience inJapan. and taking into account the predictions for the
new plant at Sellafield, a better estimate for the wholeof the 1980s is about 5 man Sv (GW a)-]. When
allowance is made for the proportion of fuel reprocessed commercially, the normalized contribution
from occupational exposure is 0.25 man Sv ( G W a)-l.
D.

SOLID WASTE DISPOSAL

173. The solid wastes that are generated in thehandling, processing and disposal of spent fuels are of
two broad categories. Most of the activity in the spent
fuel is separated during reprocessing and, after aperiod of storage as a liquid. will be solidified foreventual disposal as high-level waste (HLW), generating
significant decay heat. During the reprocessing operation considerable amounts of solid low-level wastes(LLW) and solid intermediate-level wastes (ILW) areproduced, some streams of the latter being characterized by an appreciable content of actinides. If the
spent fuel is not reprocessed but stored and preparedfor disposal, there will be almost no ILW or LLW.
But the packaged spent fuel is then treated as HLW: itconrains the actinides that would have been separated
for re-use by the reprocessing operation. Since neither
spent fuel nor vitrified HLW have been disposed of.
they are not considered in this assessment of current
operations.
174. Production of other solid wastes in reprocessing
plants has been highly dependent on the operational
characteristics of the particular plant. In particular,much of the waste produced at the Sellafield plant in
the United Kingdom is attributable t o the degradationof Magnos fuel in underwater storage and should not
be taken to be indicative for other plants now o r in
the future.
175. Production of ILW from the British Magnox
reprocessing programme has been estimated [TI I] at
47,000 m' from the reprocessing of 30.000 t uranium
metal. The activity content is estimated to be about
2 10'" Bq alpha and 2 10" Bq beta/gamma a c t ~ \ i t gat
1990. Taking an average fuel requirement for Magnox
reactors of 200 t ( G W a)-I. these correspond to thequantities shown in Table 68. Comparing with thealpha inventory of the fuel throughput, calculated tobe 6.700 TBq ( G W a)-' at six months and 3,800 TBq( G W a)-; at 20 years [G5], the fraction of alphaactivity throughput lost to the ILW is about 0.02-0.03.
Similarly, taking the beta/gamma inventory of thefuel throughput to be 2.4 106 TBq (GW a)-' at six
months and 1.7 loS TBq (GW a)-' at 20 years [G5].
the fraction of beta/gamma throughput lost to the

ILW is about 0.005-0.05. These will be very sensitive
to reprocessing chemical conditions for some nuclides.
especially ?"Np. The generation of ILW from the
proposed oxide fuel reprocessing plant for AGR fuelwas also estimated by Taylor [TI I] to be 11000 m'from the reprocessing of 600 t uranium metal in thethermal oxide reprocessing plant (THORP). The
activity content is 6 l o i 5Bq alpha and 5 1O1%q beta/
gamma activity, assuming a cooling period of five
years. The average fuel requirement of the AGR istaken to be 30 t ( G W a)-) to give the quantities inTable 68. Again. comparing with the alpha and beta/
gamma inventories of the fuel throughput at five
years, calculated t o be 3,100 and 2.5 10' TBq (GW a)-',respectively [G5], the fractions of activity throughput
lost to the ILW are 0.01 for alpha and 0.1 forbeta/gamma activities.
176. The annual rate of waste generation at Marcoulehas been reported [B32] to be about 2,000-3,000 drums
containing a total of about 4 TBq alpha and 4,000 TBq
beta/gamma activity. Assuming the annual fuel throughput of the facility to be 0.4 G W a and the drum
capacity to be 0.2 m3 gives the quantities shown in
Table 68.
177. An alternative method of estimating the activity
content of other solid wastes is to assess it directly as afraction of the throughput of radionuclides in the fuel.
This approach has been used by the United States
Department of Energy [D4] to give the results inTable 69. The quantities are comparable with thoseestimated by Hill et al. [HI I] but considerably lessthan those estimated for an operating plant, as shownin Table 68. The difference is about two orders ofmagnitude for alpha emitters and nearer to threeorders of magnitude for beta/gamma emitters.
178. T o give some estimate of the consequences ofdisposal of such wasres. it is assumed thal the alphalvastes are entirely 2 ' q P and~ the b e t d g a m m a wastes
entirely I3'Cs, and a typical normalized production
from Table 68 is taken to be 100 TBq (GWr a)-I foralpha wastes and 10.000 TBq ( G W a)-I for beta/
gamma wastes. Using the model for a typical ILW
engineered disposal trench from section 1II.D. and thevalues for collective effective dose equivalent commitment per unit activity disposed of from Table 55,
the normalized collective effective dose equivalentcommitment would be 1 man Sv ( G W a)-I. This is
reduced to 0.05 man Sv (GLV a)-' when account istaken of the proportion of fuel reprocessed commercially. Lower losses from throughput to the ILW a n d
LLW wasre streams as estimated in paragraph 176
would significantly reduce this estimate; greater lossesof the very long-lived radionuclides I4C and "91 would
significantly increase it.

V.

COLLECTIVE DOSE COM\IITR.IEIVTS
FROM GLOBALLY DISPERSEDRADIONUCLIDES

179. The nuclides giving rise to a global collecrivedose commitment are sufficiently long-lived and migratethrough the environment, thus achieving widespread
distribution. Those of interest are 'H. ''C. BsKr andI z 9 I . The environmental transfer of 'H. "C and 8sKr is
becoming fairly well established. and reliable estimatesof collective dose commitments were made by theCommittee in the UNSCEAR 1982 Report. Other
long-lived nuclides. such as 2 3 q P ~are, far less mobile
in the environment and therefore become less dispersed after deposition on to soils or sediments,
following release into the local region.
180. The very long-lived nuclides, such as '291,pose a
special problem because of the uncertainty in predicting population size, dietary habits and environmental
pathways over periods of tens of millions of years.
Therefore. little use can be made of these collective
dose commitments for decision-making purposes. The
incomplete collective dose commitment, however. is
useful to demonstrate the time distribution of the dose
commitment and to estimate the per caput doses
arising per year from a finite duration of a practice. In
the following paragraphs, complete and incomplete
dose commitments are given for the globally dispersed
nuclides up to a maximum of lo6 a. The collective dose
commitments per unit release were taken from the
UNSCEAR 1982 Report of the Committee and scaled
for the normalized releases derived for 1980-1984
discharges.
181. In the UNSCEAR 1982 Report, a model reprocessing facility was described and all reactor fuel was
assumed to be reprocessed. In this Report, thecollective doses assessed for reprocessing plant reported
discharges are weighted by the fraction of the energy
value of the total nuclear fuel that is reprocessed,
namely 5% (paragraph 147). The weighted contribution is added to any contribution from reactor
operation to reflect the current normalized exposures.

182. Since krypton is an inert gas, it disperses
throughout the atmosphere and achieves a uniform
concentration in about two years. The Committee, in
the UNSCEAR 1982 Report. estimated the collective
effective dose equivalent commitment from B5Krto be
0.17 man Sv PBq-', assuming a world population of
4 lo9. This must be scaled up to 0.2 man Sv PBq-I for
the world population of 4.6 lo9 during the period
1980-1985. All the dose commitment is delivered
within the first 50 years after release. Paragraph 150
gave normalized production of 8SKr as 11.5 PBq(GW a)-' for LWRs and 14 PBq (GW a)-' for GCRs.
leading to 2.3 man Sv (GW a)-' and 2.8 man Sv( G W a)-' collective effective dose equivalent commitment, respectively. Contributions to the collective
effective dose equivalent commitment come almost
equally from whole-body gamma-radiation and frombeta-irradiation of the skin. Weighting this collective
dose by the fraction of fuel currently reprocessed(0.05) leads to 0.12 man Sv (GW a)-'. The incomplete
collective dose commitments are shown in Table 70.
which indicates that half of the dose from "Kr isdelivered in the first 10 years after discharge.

B.

TRITIUM183. The models used by the Committee in the
UNSCEAR 1982 Report gave a collective effective
dose equivalent commitment of 2.8
man Sv
per TBq released. Because of the short half-life oftritium, this applies to the world population at thetime of release. For the 1980-1985 world population ofman Sv
4.6 lo9, the dose factor is increased to 3.2
per TBq. Releases to the atmosphere and hydrosphere
were not distinguished, since the exchange of water
between the atmosphere and circulating waters of the
globe is rapid. and the models assume immediate
mixing and exchange with the hydrogen content of thecirculating water.
184. The normalized release of tritium to the atmosphere from reactor operations, weighted by electricity
production, is 46 TBq (GW a)-', while for liquid
discharges the data give 40 TBq (GW a)-'. .4veraged
over Sellafield and Cap de la Hague. aquatic andatmospheric releases from reprocessing add up to
about 600 TBq (GW a)-', and since only 5% of thefuel is reprocessed, this adds 30 TBq (GW a)-' to thereactor releases of 86 TBq (GW a)-'. The total collective effective dose equivalent commitment amounts to
0.004 man Sv (GW a)-'. The incomplete collective
dose commitments shown in Table 70 indicate thatessentially all of the dose is received in the first few
years after discharge. The local and regional contribution from tritium releases from reactor operation andthe fractional reprocessing contribution amount toabout 0.6 man Sv (GW a)-', which is a factor of over
100 greater than the global contribution.

185. The Committee used in the UNSCEAR 1982
Report a relatively complex compartment model toassess the environmental distribution and behaviourof I4C. This model allows for two hemispheres, eachcomprising humus, circulating carbon, surface oceanand deep ocean. The circulating carbon represents the
carbon in the troposphere and those sectors of theterrestrial biosphere subject to rapid growth anddecomposition. Humus represents the carbon content of
the terrestrial biosphere which circulates more slou.ly.
Carbon- 14 releases are assumed to be instantaneously
mixed in the compartment to which release occurs.
The results produced by this model are similar to
those produced by more complex models; the mainarea of uncertainty is the rate of transfer of "C to thedeep ocean, from where it is less available.
186. The resulting collective effective dose equivalentcommitment is 67 man Sv TBq-I released, averaged
over both aquatic and atmospheric releases andassuming a future global population of 10l0.Normalized
releases from reprocessing plants are averaged over
the reported figures for Sellafield (Table 59) andcalculated throughput for Cap de la Hague (paragraph 152). Measurements appear to show that all thethroughput is measured in airborne effluents, and i t is
assumed that little is discharged to the sea. The
normalized release is 3.5 TBq (GW a)-I from Sellafield(Table 59): Cap de la Hague is assumed t o give rise toreleases of 0.66 TBq (GW a)-'. The collective effective
dose equivalent commitment for Sellafield is thus
234 man Sv (GW a)-' and 44 man Sv ( G W a)-[ forCap de la Hague. Since about 5% of the annual
energy equivalent of fuel is reprocessed, the weighted
figure averaged over the two sites is 6 man Sv(GW a)-', the remaining fuel being stored and not
giving rise to effluent releases of I4C.
187. HWR releases are about 7.3 TBq (GW a)-' of''C from reactor operations (Table 25). while thosefrom LWGRs and GCRs are about 1 . 1 TBq (GW a)-'(paragraph 72). LWR releases at about 0.3 TBq( G W a)-' (paragraph 70) are small in comparison. The
normalized collective effective dose equivalent commitment from HWR operation is therefore 490 man Sk( G W a)-'. This is nearly a factor of 3 lower than theestimate given in the UNSCEAR 1982 Report, and isentirely due t o lower reported discharge figures.
About 6% of total nuclear generated electric energy
arises from HWRs and about 10% from LWGRs andGCRs, so the electricity production weighted contribution to collective dose is 32 man Sv (GW a)-' from
HWRs and an additional 7.7 man Sv ( G W a)-' from
GCRs and LWGRs. Although LWR releases arc
lower, because of their larger electric production,
LWRs add 17 man Sv (GW a)-'. In summary. thepresent practices of reactor operation and reprocessing
lead to a total collective effective dose equivalen~
commitment of 6 man Sv (reprocessing) plus 57 man
Sv (from HWR, LWR, LWGR and G C R operation),
i.e., 63 man Sv (GW a)-'. This commitment is received
over some 10000 years, while the temporal distribution
is shown in Table 70 to be 3% in 10 years, 10% in
100 years and 19% in 1.000 years.
D.

IODINE- 129

188. When released to the atmosphere. iodine, becauseof its environmental mobility, becomes rapidly incorporated into foodstuffs ingested by individuals. The
highest concentrations of iodine occur in sea water
and, as with "C. the greatest uncertainties surroundthe transfer of Iz9I to deep oceans and any sedimentation that may remove activity from any biologicalchain.
189. Assuming again a future global population of
loL0, the Committee used a collective effective dose
equivalent commitment of 1.4 lo4 man Sv TBq-'
released [UI]; of this, some 0.003% is delivered within
100 years of release. 0.03% in 10,000 years. 5% in
lob years. thus leaving 95% of the collective dose to bedelivered from 1 million years after release, most of i t
coming between 10 million and 40 million years. For
this report incomplete dose commitments to lo6 years
are used so that the value of Iz9I is 700 man Sv TBq-'.
190. The normalized releases from Sellafield and
Cap de la Hague from 1980-1985 averaged about
40 GBq (GW a)-' to the sea and 4 GBq (GW a)-' to
the the atmosphere, giving a total of 44 GBq (GW a)-'.
which, when weighted for the fraction of fuel that is
reprocessed, gives 2.2 GBq ( G W a)-'. The correspond-

ing incomplete collective effective dose conlmitment to
10,000 years is 1.5 man Sv (GW a)-'. The incomplete
value to lo4 vears is 0.0093 man Sv (GW a)-' and for
100 years 0.0008 man Sv (GW a)-'. as shown in
Table 70.

VI.

TRANSPORT

191. Materials of various types are transported
between the installations involved in the entire fuelcycle. The amounts and distances depend on thenumber of facilities and the degree t o which different
facilities are located together. An estimate is given in
Table 71 of the transport needs in a complete nuclear
fuel cycle; this has been adapted from the report of theInternational Fuel Cycle Evaluation (INFCE) [I 151. In
general, mills are located together with mines, andtailings are disposed of close by, so that there is no
significant requirement for transport of \,cry large
quantities of ore or wastes. The other major transport
requirements shown in Table 71 can not be eliminatedby co-location. as other factors will dominate thesiting requirements. IAEA has continued t o work
towards a full assessment of the radiological impact oftransport and have recently published the preliminary
findings of a technical committee [P13]. The general
conclusion was that, although the data available wereincomplete, the indications were that exposures resulting from normal transport operations were low bothfor workers and members of the public.
192. The estimates made during the course of the
IAEA study of occupational collective doses from thetransport of fuel cycle materials were a recognizedcautious estimate of 19 man Sv for the United States
as a projection for 1985 [Nl I] and a more realistic
estimate of 0.14 man Sv for the United Kingdom in
198 1 [GI]. Estimates of less than 0.01 man Sv weremade for selected operations in France, Italy andSweden, but these could not be normalized to energy
production. Using the energy production figures forthe appropriate years gives normalized collective
effective dose equivalents of 0.5 man Sv ( G W a)-' for
the United States and 0.04 man Sv ( G W a)-! for theUnited Kingdom. Noting that the United Statesassessment was pessimistic, but that the United
Kingdom assessment did not include the transport
associated with uranium mining and milling. an
overall estimate of 0.2 man Sv (GW a)-' is probably
reasonable.
193. Doses to members of the public were also
estimated as part of the work of the IAEA committee.
based again on submissions from the United States
and the United Kingdom. The estimate for the UnitedStates was 19 man Sv for 1985 [N 1I], the same as thatfor occupational exposure, whereas that for theUnited Kingdom was several orders of magnitude
lower, at 0.001 man Sv for 1981 [GI]. No estimates
were available for other countries. Based mainly on
the more realistic British assessment, it seems reasonable t o conclude that public exposure from transport
is less than occupational exposure and t o adopt a nestimate for the normalized collective effective dose
equivalent of 0.1 man Sv (GW a)-'.

VII. SUMMARY
194. In the UNSCEAR 1982 Report the Committee
carried out a thorough assessment of the exposures tothe public from nuclear power production. In thisReport the same basic assumptions and environmental
transport models are used to carry out a revised
assessment based on discharge data for the quinquennium 1980- 1984. Some aspects of waste disposal
have been treated here in more detail, especially thelong-term impact of uranium mill tailings and thedisposal of solid low- and intermediate-level wastes by
burial on land. The contribution from reprocessing isbased more closely on the results being obtained at
operating plants rather than on the notional plantused in the previous report. Occupational exposuresfrom the various stages in the fuel cycle are reviewed
in this Annex in association with the other exposuresfrom released radioactive materials.
195. .4 summary of the local and regional normalized
collective effective dose equivalent commitments fromthe nuclear fuel cycle is shown in Table 72. The totalof 4 man Sv (GW a)-' is essentially the same figure asthat derived in the UNSCEAR 1982 Report if thecontribution from uranium mine tailings is excluded,
although in this Annex reprocessing is added explicitly,
whereas a notional plant was used for the UNSCEAR
1982 Report. Contributions other than radon arise
mainly from routine atmospheric releases from reactors and the liquid discharges from reprocessing.
Effectively, all of these dose commitments are received
within one to two years of discharge.

196. The normalized collective effective dose equivalent commitments from the long-term releases fromsolid waste disposal are shown in Table 73. T h edominant contribution, as was recognized in the
UNSCEAR 1982 Report, is from mine and mill tailings.
The numerical estimate is roughly proportional to thelength of time for which release of radon is assumed
to occur. The estimate of 150 man Sv ( G W a)-'corresponds to 10000 years for a tailings pile with areasonable co\nering. The estimate for disposals of
LLW and ILW are for the release from the disposalsites for all time. but a large proportion of the dose isreceived within about 10"earsfrom the date ofdisposal. This applies also for the globally dispersedradionuclides shown in Table 73. as these are dominated in terms of the normalized contribution by ''C.
197. The contributions of the various stages of thefuel cycle to occupational doses are summarized inTable 74. The dominant contribution is from reactor
operation, itself based mainly on recent experience
with LWRs in the United States but with considerable
data from many other countries.

198. The per caput doses from existing nuclear
power production are estimated from the contributions to collective dose commitment in the short term.
This collective dose commitment is from local a n dregional collective doses and from occupational exposure. i.e.. 4 and 12 man Sv ( G W a)-'. respectively(Tables 72 and 74). Assuming a global population of
5 lo9, the per caput dose would be 3 nSv ( G W a)-'.
The energy production from nuclear power in 1987 isabout 190 G W a (Table 2). so that the annual per
caput dose is estimated to be 0.6 uSv.

l d b l e

1

World nuclear qeneratlng CdDaCltV. 1981
(Net capaclty In glgnwdtts and number ot unlts In pdrentheSe5)
1111
Reactor typeCountry

Total
CaDaCtty
PWR

Argent lna
Belglum
Bra211
Bulgzrla
CdnadaChlna (Talwan Provtnce)
Czechorlovak la
FlnlandFranceGerman O m c r a l l c Rep.
Germny. federal Rep.of
nunga ry
lndla
Italy
Japan
Netherlands
Patlstan
Republlc of Korea
South Atrtca
Spat nSwedenSwitzerland

GCR

BWR

0.94
5.49
0.63
2.59

LWGR

H M

f 8R

(ti)

(2)

(7)
(1)
(5)
12.10 (18)

1.81 (2)3.20 (8)0.89 (2)46.46 (46):.69 (5)11.13 (11)1.65 (4)0.26 (I)11.91 (16)0.45 (1)4.75 (6)1.84 (2)4.68 ( 6 )2.63 (3)1.62 (3)16.87 (25)

3.10

(4)

1.47

(2)

6.89

(7)

0.30 (2)0.86 (1)14.64 (19)0.06 ( I )

1.31
6.83
1.31
0.05

(2)
(9)
(2)
(I)

2.01

(4)

1.43

(2)

0.30

(7)

0.02

(I)

0.16

(1)

15.98 (21)

0.10
0.2

(3)
(I)

15.98 (27)

2.38

(7)

0.48

0.85

(4)

0.15

(1)

0.13
0.63

(1)
(1)

0.09

(1)

14.68 (28)

(I)

USSR
Untted Klngdom
Unlted States
Yugosldvla

63.57 (70)0.63 (I)

29.1

(35)

9.90 (36)0.33 (1:Totdl

183.63 (225)68.02 (85)13.01 (45)l a b l e

2

llectrlclty generated by nuclear power. 1987

1111

Llectrlc
energygenerated(GW a)Country-

Unlted States
51 - 9France
28.7
USSR
21.3
Japan
20.8
Germany. Federal
Repubilc of
14. l
Canada8.32
Sweden7.35
5.58
Unlted Klngdom
Belg\um
4.52
4.51
Spa 1 nRepubllc o f Korea
4 .21
Chlna (Taluan Province) 3.58
Sultzerland
2.48

Czechoslovakia

2.36

Flnland
Bulgaria
German Democratic Rep.
HungarySouth Afrlca
Argentlna
1 nd laYugoslavla
Netherlands
BrazllPaklrtan
Italy

2.11
1.31
1.18
1.18
0.71
0.68
0.54
0.49
0.39

Total

0.10
0.03
0.01

189

Percentageo f totalelectrlclty
generatedinstalled
cdpaclty
per capur

0.94
5.49
0.63
2.59
12.10
4.92
3.20
2.31
49.80
1.69
18.95
1,65
1 . I5
1.12
26.90
0.51
0.13
5.38
1 .@4
6.53
9.46
2.93
33.60
10.21
92.98
0.63

0.031
0.55
0.001
6-79
0.41
0.75
0.21
0.41
0.90
0.10
0.31
0.15
0 002
0.02
0.22
0.04
0.001
0.13
0.05
0.16
1.14
0.45
0.12
0.18
0.38
0.03

297.93(411)

0.14

T a b l e

3

Uranium ~ r o d u c t l o nby m l n l n q . 1980-1984[ 04 I

Annual a u a n t l t y of u r a n l u m o x l d e produced ( k t )Country
1980

1981

1982

I983

1984

Argent l n aAustral l aBrazllCanadaFranceGabon
Namlbla
Nlger
South A f r i c a
Spaln
Unlted States

0.18
0.56
0
7.15
2.63
1.03
4.04
4.13
6.15
0.19
16.80

0.12
2.92
0
1.72
2.55
1.02
3.97
4.36
6.13
0.18
14.79

0.15
4.42
0.24
8.08
2.86
0.91
3.78
4.26
1.82
0.15
10.33

0.18
3.21
0.19
7.14
3.27
l.Ol
3.72
3.43
6.06
0.17
8.13

0.13
4.39
0.12
11.17
3.17
0.92
3.70
3.28
5.73
0.20
5.72

a/

44.0

44.0

41.3

36.7

38.7

Total

a/

Not l n c l u d t n g c e n t r a l l y planned economy c o u n t r i e s .

T a b l e

4

Radon e m l s s l o n f r o m sone a c t l v e u r a n l u m mlnes[ E l . J1. JZ. LZ. M4. 16. 17. U3]

Locatlonnlne typeOre grade(X)
New U e x l c o . U n l t e d S t a t e sUnderground

Tennessee,
Unlted States
Unlted States
I l l l o t Lake. Canada
Average: U n l t e d S t a t e sUnderground
Underground
Underground
Underground
Surface
Surface
Surface

Large mlne. U n l t e d S t a t e s
Ranger. A u s t r a l l a

a/
b/
C/

Annualradon
emlsslon(18~1)

0.1
0.1

0.3

Range f o r seven mlnes.
Average f o r seven mlnes.
Average f o r 27 mlnes.

T a h l e

5

A l r b o r n e ernlsslans from a m l l lp r o c e s s l n q 2000 tonnes o f o r e p e r daylU l I

Radlonucllde

Annua 1ernisslons(GW)

Normallredradonemlsslon(G8q t - l )T a b l e

6Radon emanatlon from uranlum mill talllnqs piles[BPS. 834. C10, H14. L1, NS. R2. R3]

Locat ion

Argent lna
ChubutMalargue
Malargue
Cordoba
Salta a/St. Rafael

Talllngs
management

Radon
emanatlon
rate perunlt area- 2 -1
(Bq m
s

Area

Annua 1radon
emanat Ion

(ha)

(TBq)

30
30

< 0.7

(1984)(1984)(1985)( 1985)(1984)(1983)

Austral la
Rum Jungle
Canadallllot Lake
indla
Jaduguda
Unlted States(Temperate)(Seml-arld)
Salt Lake City

Uncovered
3 m capplng< 0.07

1.3

Vegetated
Unvegetated

1.2-4.0
3.5

400

300

Uncovered

1.1

12

4

Uncovered
1 m clay
Uncovered
Impermeable dam
Uncovered
Covered

10
0.3
10
0.1
18
7
4

40
40
90
90

120
4
300
3

Ambrosla Lake

13

a/ This mlne and mlll 1s at an altltude of ZOO0 m.

T a b l e

7

Optlons for treatment o f uranium mlll talllnqs piles
and relatlve effects on radon release rates[ N5 I

OptlonPredlcted radon
exhalatlon rate
at 1000 years
relatlve
to base case a/Inltlal value(0) Bare talllngs pile

1.4
( 1 ) 1 m top cover of silt/sand
0.73(2) 3 m top cover o f sllt/sand
0.58(3) 1 m top cover of clay
0.45(4) As ( 1 ) ulth eroslon protectlon b/
0.29(5) As (2) ulth erosion protectlon b/
0.039(6) As ( 3 ) ulth eroslon protectlon h/
0.0027(7) Bare talllngs below ground level2.7 10-7
covered ulth 3 m clay-shale(8) As (7) ulth rock surround and2.5 10-7
gravel capptng
a/
-

b/

Base case Is the bare talllngs plle.
Crushed rock around exterlor slopes and a gravel cap
over the top surface.

T a b l e

8Collectlve effective dose eaulvalent comnltments
per unlt actlvlty released
In alrhorne releasel from uranluln 111lnesant1 nil 11s

Normallzed collectlve
effectlvc dosee q u l v a l ~ n t comnltment[man Sv (18q)-l]Radlonucllde

T a b l e

9

Truncated collectlve dose comnltments from radionuclides released
from talllnqs plles g/[NSI

Truncated collectlve effectlve dose equivalent comnitment(man Sv)Management
scenarlo

1 0 0 years
ReglonBase case
Optlon 1
OptIon 2
Optlon 3
Optlon 4
Optlon 5
Optlon 6

Contlnent

1000 years
ReglonContlnent

5000 years
ReglonContlnent

1 0 0 0 0 years
ReglonContlnent

46
12
2.8
0.6
12
1.6
0.1

g/ Treatment options a r e llsted ln Table 7. The base case ls the bare talllngs

plle. T h e collectlve effectlve dose eaulvalent comnltments from optlons (7)and ( 8 ) a r e less than 0.01 man Sv.

T a b l e

10

Occupational exposures o f underground u r a n l u m m l n e r s

lA4. E l l . 837. f 3 . S8. 22)

Number o t workers
monl t o r e dCountry

Annual r o l l e r t l v ee t t e c t l v e dosee q u i v a l e n t (man Sv)

Year
t a m
exposedCanada

1980
1981
1982
1983
1984
1985
1980
1981
1982
1983
1984
1985
1980
1981
1987

f rance

Unlted StatesRadon
exposed

7 124
6837
6159
4428
3970
3930

6140
7160
6290
5850
5810
1609
1380
1301
1281
1384
1388
7600

1281
1384
138e
7556,
3790
21 20

E x t e r n a l Radon and I t sradlatlon
daughters- b/

7.0 k/
18.7
18.0
15.6
12.5
6. 2
4.7
5.6
5.0
4.6
4.8
27.0

40
58
51
40
33
29
25
I8

19
16
15
13
68
27
14

a/ Average

Annua 1averageeftectlve
dose
equlvalent

a/

(mSv)
5 b/
10 b/
11
12
11
10
23 c/
21
23 L/
25 E/
19 c/
I8
12

c/

c/

7 d_/
7 d_/

f o r t h o s e exposed t o g a m n a - r a d t a t l o n p l u s average t o r t h o s e
exposed t o r a d o n and I t s d a u g h t e r s , e x c e p t where n o t e d .
T h l r procedurem l g h t l e a d t o an o v e r e s t i m a t e .
b/ There was no m o n l t o r l n g f o r e x t e r n a l r a d l a t l o n I n 1980; d a t a t o r 1981 a r e
n o t f o r a complete year.
c/ Include: t h e t o l l o u l n g v a l u e s o t a n n u a l c o l l e c t l v e e f f e c t i v e dosee q u l v a l e n t (man Sv) f r o m m i n e r a l d u s t . f o r w h i c h a c o n v e r s l o n o f 34 Bq
m ~ v - I was t a k e n : 1980. 6.0; 1981. 5.6; 1982. 5.4; 1983. 10.5; 1984. b.4;
1985. 7.7.
d/ Radon and d a u g h t e r exposure o n l y .

Hadon-- and radon dauqhter exposuresof Canadldn and Australian s u r f a c e u r a n l u m m l n e r s
1414. AH. r 9 ]CountryYear

Canada

lYBU
1981
1982
1983
1984
1985

Number cf
workers
monitored

Australld
N a b a r l e k 1981/82
Ranger1985/86$1

Less t h a n 0.05 mSv.

Annua 1to1lect lve
cttectlve
dose
equlvalent(man Sv)

Annua 1averageeftectlve
dose
equlvalent(msv)

T a b l e

12Effluent dlscharqes
from selected fuel converslon. enrtchment and fabrlcatlon plants. 1980-1985
LA^. 01. 82. 83. Be. 616. 829. 11. n71

Llqu\d dlschdrge: (GBq)
Locatlon and nucllde

Unlted Klngdom
Capenhurst (enrlchment)
U-234
0.64
U-235
0.03
U-238
0.64
Th-234
0.65
1I
Tc-99
Sprlngllelds (converslon. fabrlcatlon)
Uranlc alpha
900
Urantc beta
131000
Canada (fabrlcatlon) (average values per year)
Uranlum
0.09
Port HopeUranlum
3.9
Toronto0.002
Peterborough Uranlum
Varennes
Urantum0
Monc ton
Uran l um< 58
Atrnospherlc dlscharqes (GBq)
Locatlon and nucllde

Unlted Klngdom
Capenhurst (enrlchment)
U-234
0.15
U-235
0.01
U-238
0.15
Th-2340.15
Sprlngflelds (converslon. fabrlcatlon)
Uranlc alpha10.0
Uranlc beta
10.0

0.26
0.01
0.26
0.27

0.20
0.01
0.20
0.20

0.35
0.02
0.35
0.36

0.05
0.002
0.05
0.05

0.003
0.0001
0.003
0.003

2.0
2.0

2.0
2.0

1 .O
1.0

1.0
1.0

1 .O
1.0

Canada (fabrlcatlon) (average values per year)
Uranlum
Port Hope
0.002
Uranlum
Toronto0.004
Peterborough Uranlum
0.0002
Varennes
Uranlum0
Monc ton
Uranlum
4 8

T a b l e

I3

Normallzed effluent dlscharqes from model fuel converslon,
enrlchment and fabrlcatlon facllttles[Mas (~ud)-lI

Atmosphere

Aquatlc

Radlonuc l lde
Convcrslon Cnrlchment

fabrlcatl~n Converslon

Cnrlchment

Fabrlcatlon

T a b l e

14Normallzed collectlve effectlve dose equlvalent comnltment
from the model fuel fabrlcatlon faclllty

Normal \zed collect lve
effectlve dose
equlvalent comnltment(10-3 man Sv (GW a)-)]Rddlonucllde
Inhalatlon

Total (rounded)

2.6

T a b l e

Deposlted
actlvlty

0.2

15

Occupational exporure trom urar~lum tuel fabrlcatlon[A4, 812. U23. H8. 13. 14. 15. 16. 116. 53. NZ. PI41

CountryYear

Argentlna

1981
1982

Number ofworkers
monltored

Annua 1collectlve
eifectlve
dose
equlvalent(man Sv)

AnnualNormal \zedaveragecollectlve
effectlve
effectlve
dose
dose
equlvalentequlvalent(mSv) [man Sv (GW a)-]]CanadaJapan

UnitedKlngdom

Unlted States

1980
1981

5900
5942

11.11
9.40

1.9
1 .b

0.23
0.18

T a b l e

16

Occupatlonal external exposures from plutonium fuel fabrlcatlon. 1977-1982

Ins1

Annualcollective
CountryYear

Number ofworkers
monltored

effectlve
dose
equlvalent(man S V )

Japan

1911
1918
1979
1980
1981
1982

114
134
225
198
200
266

0.9
0.8

T a b l e

Annualaverageeffectlve
dose
equlvalent
(m~v)
0.13
0.10
0.23
0.33
0.50
0.58

I .O
1.1
2.5
2.2

17

Noble ases dlscharqed In alrborne effluents from PWRs. 1980-1985[AI. h6, 85, 819, 0396. C4. FI. F4, 53. KI. 51. 52. 53, 55, 59, 510. T3, 14, 15, 18. 19, 1101

Start-up
year

Country and reactor

Belglum
Doe1 1.2.3
Tlhange 1.2

1974/75
1975

FranceBlayals 1.2.3.4
Bugey 2,3,4.5
Chtnon 81.B2
cnooz
Cruas 1.2.3.4
Damplerre 1.2.3.4
Fessenheim 1.2
Gravellnes 1-6
Paluel 1.2
St. Laurent 81.2
Trlcastln 1.2.3.4
Germany. Federal Rep.
Blblls A.B
Grafenrhelnfeld
Neckarwestheim
Obrlghelm
Stade
Unterweser
GrohndePhlllppsburg 2
Italy
Trlno
Japan
Genkal 1.2
Ikata 1.2
Ylhama 1.2.3
Oh1 1,2
Sendal 1.2
Takahama 1.2

1975/81
1977/82
1970/72/76
1979
1984
1974/75

Netherlands
Borssele

1973

SwedenRlnghals 2
Rlnghals 3
RInghals 4

197 5
1980
1982

Electrlclty
generatlng
capaclty(GW)

Actlvlty (TBq)
1980

1981

1982

1983

1984

1985

Table-17. contlnued

Country and reactor

Start-up
year

tlectrlclty
generatlng
capaclty
( GW)

USSR
Armenlan 1.2
Kalinln 1
Kola 1.2
Kola 3
Nlkolaev 1
Novovoronezh 1.2
Novovoronezh 3 , 4Novovoronezh 5
R o v n o 1.2
Zapororhe 1
United S t a t e s
Arkansas One-l
Arkansas One-2
Beaver Valley
Calvert Cliffs 1.2
Crystal RIver
Davls 8 e s s e
Olablo C a n y o nDonald C. Cook 1 , 2
farley 1
Farley 2
fort C a l h o u n
H.8. R o b i n s o n
H a d d a m Necklndlan Point 1,2
Indian Polnt 3
KewauneeH a l n e Yankee
M c G u l r e 1.2
H i l l s t o n e Pt. 2
tiur t h A r m
DcuneePa I isades
Yolnt Beach 1.2
Pralrle Island I,?
R.E. Glnnd
Rancho Seco 1
Salem 1
Salem 2
San Unof re 1'
S a n O n o f r e 2,3SequoyahSt. L u c l e 1
St. Lucie 2
Surry 1 , 2Three l l l e lsland 1
Three 'ile lsland 2
TrI 2/EPICOR
TrojanTurkey Point'dlrgll C. Sumner"olf Greek
Yankee RodeZ I o n 1.2
Total annual electrtc energy
g e n e r a t e d ( G W a)
Normallzed actlvlty[ r 8 q ( G w a)-I]A v e r a g e normallled a c t l v l t y , 1 9 8 0 - 1 9 8 4[ T B ~( L W a)-lj

Actlvlty (T8q)
1980

1981

1982

1983

1984

1985

T a b l e

I8

Isoto~lc com~osltlon,
of noble gar dlrcharger from PURsln the Unlted States. 1982
[T51

ReactorElectrlc
energyStart-up
generatedyear(CU a)

Arkansas One-l
Arkansas One-2
Beaver Valley 1
Calvert Cllffs 1.2
Crystal Rlver
Oavl5-Besse 1
Oonald Cook 1.2
farley 1
Farley 2
Fort Calhoun I
H.8. Robinson
Haddam Neck
Indlan Polnt 1.2
Indlan Point 3
Keuaunee
nalne Yankee&Gut reMillstone 2
Horth Anna 1.2
Oconee 1,2.3
Pal lsades
Polnt Beach 1.2
Pralrle Island 1.2
R.E. Glnna
Rancho Seco 1
Salem I
Salem 2San Onofre I
San Onofre 2.3
Sequoyah 1.2
St. Lucle 1
Surry 1.2
Three Hlle Island 1
Three Mlle island 2
TMI Z/EPICOR
TrojanTurkey Polnt 3.4
Vlrgll C. SumnerYdntee Roue
Zlon 1.2

0.548
0.868
0.022
0.101
1.125

Total annual electrlc energya)generated (GI21.042Hormallzed actlvlty
Iraq ( w a)-ll

0.424
0.435
0.307
1.182
0.561
0.367
1.409
0.595
0.605
0.397
0.257
0.518
0.508
0.164
0.436
0.516
0.491
0.572
0.736
1.221
0.382
0.720
0.887
0.275
0.384
0.467
0.906
0.058
0.014
0.560
0.773
1.251

Actlvlty (Tea)
Ar-41

Kr-85m

Kr-85

Kr-81

Kr-88

Xe-13lm Xe-133m

Xe-133 Xe-135m Xe-135 Xe-138

-

0.006
0.815

0.168
0.653
0.093

-

-

0.001
0.002

0.010
5.439

2.039
1.180

-

1.001
0.141

0.021
0.451

0.051
0.433
1.084
0.237
0.296
0.001

0.231
6.438
0.002
0.010
0.005
0.042
0.651
0.275

0.134
6.882
0.201
0.004

0.096
6.882
0.074
0.007
0.002
0.031
0.688
0.015

0.285
1 .I900.086
2.520
21.241
0.090
0.437
0.633
0.060
0.181
0.014
0.046
2.287
0.122

-

0.289
0.720
0.028
0.859
1.910
0.030
0.522
1.432
0.238
0.071
0.327
0.240
2.309
0.662
0,001
0.056
0.271
2.017
0.597
6.550
0.008
0.3IO
0.088
0.235
0.336
0.001
0.043
0.005
0.001
3.811
4.700
3.885

111.333 3.978
336.404
25.2264.621 0.004
0.009
290.7422.923
176.585 2.819
12.841 15.133
18.901 0.059
0.385
0.158
119.1202.205
335.069 0.052 895.403 0.981
119.3119.253
11.707 0.001
0.164
0.005
5.588
0.247
22.388 0.008
1.269
0.011
289.346 0.158
6.957
0.002
90.661 0,006
2.802
0.002
0.151
0.095
3.569
0.267
36.1120.012
6.586
289.984 3.152
41.9040.751
178.164 0.013
6.841
0.006
851.126
13.211
272.321 0.037
0.073
15.811 0.899
7.920
2.409
19.580 0.010
0.155 0.005
70.300 0.047
0.525
0.005
51.4302.542
8.329 0.283
0.047
41.159
0.003
2.623
0.058
0.215 0.002
0.009
195.738
10.954147.435 3.959
55.1304.136
751.100
20.646
-

-

-

0.262

-

-

17.464
0.080
0.008
0.004
9.250
0.273
0.004
0.002
0.028
0.001

0.298
5.6
0.032
0.367
0.011
1.806
0.085
0.001
0.142
0.015

0.025
0.548
1.228
0.149

0.001
0.001
0.622
11.507
0.670

-

-

-

0.011
1.706

0.082
0.282

0.079
2.316

0.068
0.944

-

50.690
0.234
0.801
0.958
0.718
0.422
0.006
3.811
5.476
0.514
0.154
0.154
2.938
0.566
16.293
0.154
2.586
0.207
0.581
0.139
0.021
0.011
0.500

0.055
2.087
2.472

-

-

0.013
2.113
0.005

-

0.254

3.474
0.007
0.010
0.011
1.465
0.035
0.004
0.040

-

5.581
0.015

-

0.016
3.226
0.049
0.004
0.018
0.009

-

0.001

0.032
0.418
6.439
0.088
0.012
0.084
0.629
0.002

-

0.44
ll.410

-

-

-

1.289
0.012

12.9870.125

18.05615.762
0.269
0.061
0.622
0.074
5.180
0.013 0.80
0.480 48.470

0.016
0.286

0.196
1.913

0.205
2.605

28.864 0.247
128.900 0.172

1.758
4.144

0.138
0.034

0.126
1.188

0.087
0.153

0.031
13.5071.985
477.514

1.102
10.408

0.154
0.095

-

1.735
2.224

T a b l e

19

noble ares dischar ed In alrborne effluents from BURS. 1980-1985[85. 82O.'rl.
33, K1. z 1 . S2. 53. S5. S9. 510. T3. 14. 15. T8. 79. Ti01

Electrlclty
Start-up generating
Country and reactor
year
capaclty(Gu)

Actlvlty (189)
I980

1981

1982

1983

1984

1985

6.1

25.1

23.4

1.9

30.0
0.16
26.0
0.15
5.3
43.0

19.0
0.021
27 .O
0.95
0.035
1 1 -0F lnlandOlklluoto
Germany. Federal Rep. of
Brunsbittel
1976
Gundrermlngen
1984
Isar
1977
Krumnel1983
1979
Phlllppsburg 1
Wirgassen
1971

0.770
2.49
0.870
1.26
0.864
0.640

28.0

20.0

13.0

21.0
290.0

0.2
118.0

17.0
17 .O

22.0
0.0008
28.0
23.0

1968

0.052

74.4

38.4

38.5

23-68

24.79

1975
1976
1980
I981
1970
1974/85
1974

0.59
0.59
0.90
0.90
0.46
1.65
0.75

2.1
2.3
0

70.1
7.297
0.22
11.5
870
53.51
1500

0.71
0.90
1.941
23.38
680
45
1040

Italy
CaorsoJapan
fukushlma 1-1.2
3.4.5.6
fukushlma 11-1.2
Hamaoka 1.2
Onagawa
ShlmaneTokal 1 1 - 1
Tsuruga 1
Netherlands
Oodevaard
SwedenBarreback 1
Barreback 2
Forsmark 1
Forsmark 2
Oskarshamn 1
Oskarshamn 2.3
Rlnghals 1

5247
309
2800

400.39662.43
1.702
1.405
0.40
11.89
0
0.147
3696.121417.02421.22
110
15000
20000

USSR
VK - 50

Unlted States
819 Rock Polnt
Browns Ferry
Brunswlck
CooperOresden 1
Dresden 2.3
Duane Arnold 1
Fltzpatrlck
Grand Gulf
Hatch 1
Hatch 2
Humbolt Bay
Lacrosse
LasalleMillstone 1
MontlcelloNlne Mlle Polnt
Oyster Creek
Peach Bottom
Pllgrtm
Quad Cltles
Susquehanna
Vermont Yankee

1963
1973/77
1975/77
1974
1960
1971/72
1975
I975
1982
1975
1979
1963
1969
1982
1971
1971
1969
1969
1974
I972
1973
1983
1912

Total annual electrlc energy
generated (GW a)
Normallzed actlvlty[Tap (Gw a ) - 1 )Average normallzed actlvlty. 1980-1984[TBq (Gh' a)-)]2150 i 523

0.16
0.29
71
232
533
47.6
1280

T a b l e

20

I s o t o p l c c o m ~ o s l t l o no f n o b l e pas d l s c h a r q e d f r o m BURS I n t h e U n l t e d S t a t e s . 1982

[is1

ReactorB l g Rock P o l n t
Browns F e r r y
Brunswlck
CooperDresden 1
Dresden 2.3
Duane A r n o l d
fltzpatrlck
Hatch 1
Hatch 2
Humbolt Bay
Lacrosse
LasalleMillstone 1
MontlcelloNlne H t l e Polnt
O y s t e r Creek
Peach Bottom
Pllgrlm
Ouad C l t l e s
Susquehanna 1
Vermont Yankee

Start-up
year

tlectrlc
energygenerated(GW a )

1963
1973/77
1975/77
1974
1960
1971
1975
1975
1975
1979
1963
1968
1982
1971
1971
1969
1969
1974
1972
1973
1983
1972

0.041
1.964
0.551
0.602

0.016
0.053
0.465
0.276
0.129
0.229
1.519
0.375
0.947
0.037
0.476

T o t a l annual e l e c t r l c energy
g e n e r a t e d (GW a )10.293

1.029
0.260
0.566
0.329
0.426

Normalized a c t l v l t y[ T B ~(GW a ) - ) ]ReactorAr-41

418.100
703.000

0.013
11.8032.479
1 .702Start-up
year

Electrlc
energygenerated(GW a)

T o t a l annual e l e c t r l c energy
g e n e r a t e d (GW a )10.293

Kr-83m

Kr-85m

7.067
16.9838.140
950.900
777.000
53.280-19.4620.035
425.500
27.084
0.178

-

0.581

110.505

B l g Rock P o l n t
Browns F e r r y
Brunsulck
CooperDresden 1
Oresden 2.3
Duane A r n o l d
Fltzpatrlck
Hatch 1
Hatch 2
Humbolt Bay
Lacrosrc
Lasalltr
Mlllstone 1
MontlcelloNlne U l l e Polnt
O y s t e r Creek
Peach Oottom
Pllgrtm
Ouad C I t l e s
Susquehanna I
Vermont Yankee

Normallzed a c t l v l t y[TRq (GW a ) - ) ]A c t l v l t y (T8q)

2.334

Kr-85

Kr-87

Kr-88

Kr-89

Kr-90

-

33.485355.940
1883.300
78.07022.8292471.600
1842.600
111.000

19.98022.422

0.008
0.002
0.024

-

6.068
0.006
714.000
1.746
1.173

48.1000.032
880.600
10.138
0.833

-

6.993

-

6.845

15.133

-

6.179
0.592

24.38316.169
3.104

9.731
1.739

56.61036.1492.875
62.16032.0420.002-

120.9900.836
68.45041.4401.136
1.416

548.947

7.511

2.3b3

Xe-137

Xe-138

Xe-139492.100

0.718

-

0.463

0.021
0.002

122.840
0.335
11.618
9.324
1.177
2.201

234.11450.268315.598

0.699

-

-

-

1 .894

0.001

-

-

A c t l v l t y (TBq)
Xe-13lm

Xe-133m

Xe-133

Xe-135m

Xe-135

T a b l e

21

Noble aases dlscharqed In alrborne etfluents from HWRs and LWGRs. 1980-1985[Al, 819, 836, Ll]

Country and reactor

Argent lna
Atucha 1
tmba1.i~
CdndaaBruce A
Bruce B
Gentllly
Plckerlng A
Pickrrlng 8
Point Lepreau

Electrlclty
Start-up generat lng
year
capaclty
(GW)

1980

I981

1982

1983

1984

1985

1974
1983

0.335

250
0.0

46
0.0

19
0.0

47
0.0

4.7
40.9

5.5
156

1976/79
1984
1983
197 1 /73
1983/80
1983

2.96
0.75
0.685
2.91
1.03
0.630

830

61 0

510

670

240.0

250

270
0.1

10.1
350
137.0
3.9

600
29.0
25.9
170
170
0.0

800
106
121
190
270
0.8

5.837

7.37

lut'll annual electric energy
generated (GW a)

Actlvlty (TBq)

4.531

4.810

4.665

5.752

10400

14900

8940
1770

7360
2 190

9490
4700
1400

64 10
6200
3000

-

5830
6300
4100
2030

8300
5600
4000
2b00

6600
5200
2600
2500

3.583

4.355

3.5

3.5

Average normalized activity. 1980-1984[ ! R q ( G w a)-)]USSH
Chernobyl 1.2
Chernobyl 3.4
Iqnallno 1
Kur5k 1 . Z
Leningrad 1.2
Lentngrad 3.4:molcnrk 1

1977/78
1981/83
1983
1976/79
1973/75
1979/81
i982

2.00
2.00
1.50
2.00
2.00
2.00
1 .OO

Total annual electrtc energy
generated (Gu a)

Average normallrt*d activity. 1980-1984
1784 ( t k a)-11

4130

-

-

3.172

3.017

T a b l e

22

Actlvatlon q a s e s dlrcharged In airborne eFfluents f r o m GCRs. 1980-1985
181. 82. 83. 816. 829. Fl. F4. HI. H2, H3. H 4 , 53, K1, P5, P9. 511)

Country and reactor

Start-up
year

Electrlclty
generatlng
capaclty(GW)

Actlvlty (TBq)
1980

198 1

T a b l e

23

1982

1983

1984

FranceChinon 2.3
Bugey 1
St. Laurent A1.2
Italy
LatinaJapan
Tokal
Unlted Klngdom
Berkeley
BraduellChapelcrossOungeness A
Oungeness 8
Hartlepool
HeyshamHlnkley Polnt A
HInkley Polnt 8
Hunterston A
Hunterston 8
Oldbury
SizeuellTrausfynydd
Wylfa
Total dnnud 1 electrlc energy
generated ( G W a )Normalized actlvlty[ T ~ Q( G W a)-)]Average normallzed actlvlty, 1980-1984[TBq ( K i a)-I]

T r l t l u m discharged In airborne effluents from reactors. 1 9 8 0 - 1 9 8 5[85. 8 1 7 , 818, D3, F1. HI, H2. H3. H4. K1. 11. P5, P9. S1. 52. 53, 55,

Actlvlty ( T B q )Country and reactor

Belgium
Ooel 1-3

0.1

f InlandLovl lsa
FranceChooz
Blayalr 1.2

3.5

3.0

11.0

1985

Table 23, continued

Actlvlty (TBq)
Country and reactor

Germany. Federal Rep. of
Blblis A.8
4.6
Grafenrhelnfeld
GrohndeNeckaruesthelm
1.9@.3
Obrlghelm
Phll lppsburg 2
Stade
1.6
Unterueser
0.4
Netherlands
Borssele

0.63

2.5
0.00015
0.60
0.20
0.70
0.40
0.68

0.40

0.59

0.59

Total annual electrlc energy
generated(GWa)
22.29

26.18

28.46

28.99

32.75

24.19

Normallzed actlvlty[Teq (Gw a)-l]

3.310

4.740

9.215

4.702

1.9

Unlted States
4.773
Arkansas 1
Arkansas 2
0.075
Beaver Valley
0.176
Callauay
Calvert Cllffs 1.2
1.032
Crystal Rlver
0.784
Oavls-Besse
0.220
Dlablo Canyon
Donald C Cook 1.2
0.041
Farley 1
22.681farley 2
fort Calhoun
0.048
H.B. Robinson
0.225
Haddam Neck
2.313
Indlan Pojnt 1,2
0.407
Indlan Polnt 3
0.182
0.770
Kewaunee
Halne Yankee
0.117
HcGulreHlllstone Pt. 2
31.450
North Anna
2.076
Oconee0.395
Pal l sddes
0.190
Polnt Beach 1,2
24.161
Pralrle Island
3.193
R.C. Glnna
1.465
Rancho Seco
6.549
Salem 1
Salem 2San Onofre 1
1.365
San Onofre 2.3
Sequoyah 1
St. Lucle 1
13.801
St. Lucle 2
Surry 1.2
0.677
Three rille lsland 1 0.670
Three Hlle lsland 2 14.578
THI ?/EPICOR
21.616
Trojan0.551
0.043
Turkey Polnt 3.4
V l r g l l C Sumner
Wolf Creek
Yankee Roue
0.054
Zlon 1.2

7.453

Average normallzed actlvlty. 1980-1984[TBq (GW a)-I]

5.9

c 2.4

Table 23. continued

kctlvlty ( T 8 q )Country and reactor

rlnlandOlklluoto
Germany. Federal Rep. of
Brunrbuttel
0.01
Gundremnlngen
Isar
3.1
KrumnelPhlllppsburg 1
0.02
wiirgassen
0.6
Italy
Caorso4e?ner lands
Oodeuaard
United S t a t e s
Big Rock Point
Browns F e r r yarunswlck
CooperDresden 2,3Duane Arnold
Fltzpatrlck
Grand Gulf
Hatch 1
Hatch 2
Humbolt Bay
Lacrosse
Ldsd l le
Hlllstone 1
Yont lcel lokine P i l e Polnt
Oyrter Creek
Peach B o t t o m
Ptlgr lm
Quad C l t l e s
Susquehannd
Vermont Y a n k e eg'iP-2
Total a n n u a l electrtc e n e r g y
generated ( G W a)
11.76

9.245

12.28

11.91

12.50

15.94

Normallzed a c t i v i t y[TBq ( G w a)-]]

3.671

2.881

2.791

I. 6 5 9

6.249

5.984

Average norillallzed actlvlty, 1980-1984[TBq (GW a)-)]

3.4

k

1.6

Table 23. c o n t i n u e d

A c t l v l t y (TBq)
C o u n t r y and r e a c t o r

HWRS
Argentina
AtuchaEmbdlse

240

710

300

630

200
7.33

250
29.5

1554

3404

151 1

3700

660

59 2

666

0.69
629
25
25

2553
3
14
430
46
68

I500
99
53.2
390
144
110

T o t a l annual e l e c t r i c e n e r g y
generated(GWa)
4.531

4.810

4.665

5.152

6.408

7.370

riorrnalized a c t l v l t y[ T e q (GU a ) - ] 1

874.5

532.3

871.0

518.3

348.9

CanaclaBruce A
Bruce B
Gentilly
Plckering A
Plckerlng B
P o l n t Lepreau

541.5

Average f i o r m a l i z e d a c t l v l t y . 1980-1984[ l B q (GW a ) - 1 )670 t 190

GCR:
Italy
LatinaU n i t e d Klngdorn
Hunterston A
Hunterston 0
Oldbury
S i zedel lTra.:f
ynydd
Wylta

0.07

0.07

0.90

2.2
2.2
0.5

2.0
4.1
1.6

1 .6
3.4
0.3

1.4
4.6
1.1

3.7

l o t a l annual e 1 e c r r ; cg e n e r a t e d (GW a )energy
1.346

1.252

1.41

1.39

1.45

0.379

tiormallzed a c t i v i t y[TBq (Gw a ) - 1 )

6.44

6.21

4.40

5.10

4.76

1.135

Average n o r m 1 1 z e G a c t l v l t y . 1980-1984[ r e p (CM a ) - I ]5.4 t 0.9

Trltlum dlscharqed In llauld effluents from reactors. 1980-1985[as. 86. 87. B17.818. 829. 03. F4. HI. HZ. H3. H4, KI, LI, PI. PS, P9.
51. 52. S3. 55. 59. 510. 511. T3. T4. T5. T8. 19. 1101

Actlvlty (TBq)
Country and reactor

Belglum
Doe1 1,2
T l hange

24.8
12.4

f InlandLovllsa 1.2

3.7

f ranceBlayals 1.2.3,48ugey 2.3.4.5
53
Chlnon 81 .2
Chooz
110
Cruas 1.2.3.4
Damplerre 1.2.3.4
7
fessenhelm 1.2
28
Gravellnes 1.2.3.4
4
Paluel I,2St. Laurent 01.2
Trlcastln 1,2.3.4
10
Germany. federal Rep. of
Blblls A,B
35
Grafenrhelnfeld
GrohndeNectarwesthelm3
Obr lghelm
3.3
Phl l lppsburg 2
Stade
2.3
Unter~eser
8.8
Italy
Trlno

37.4

Netherlands
Borssele
SwedenRlnghals 2
Rlnghals 3
Rlnghals 4
USSR
ArmenlanUnlted States
Arkansas 1
Arkansas 2
Bedver Valley
Callauay
Calvert Cllffs I
Crystal Rlver
Davls Besse
Olablo Canyon
Donald Cook 1,2
Farley 1
Farley 2
Fort Calhoun
Fort St. Vraln
Grand Gulf
H.8. Roblnson
Haddam Neck
Indlan Polnt 1,2
Indlan Polnt 3
Kewaunee
Ualne Yankee
UcGuireMillstone Pt. 2
North Anna
Oconee

13
0.7

1.9

Table 24. continued

A c t l v l t y (TBq)
Country and r e a c t o r

Palisades
2.764
P o l n t Beach 1.2
28.157
Pralrie Island
20.091
R.E. Glnna
5.920
Rancho Seco
0.0005
Salem 1
Salem 2San Onofre 1
38.110San Onofre 2.3
SequoyaS t . Lucle 1
10.064
St. L u c l e 2
S u r r y 1,2
14.245Three M i l e I s l a n d 1 1.206
Three M l l e I s l a n d 2 0.000022
Trojan4.588
Turkey P o l n t
27.713V l r g i l C. Sumner
WaterFord
Wolf Creek
2.161
Yankee Rove27.565
Zlon 1 .2

10.28624.12420.7948.880
3.089
18.24131.191
10.98912.02519.6470.263
0.0014
3.811
7.215

3.811
32.190T o t a l annual e l e c t r l c energy
generated(GWa)
29.69

40.55

44.97

53.09

61.74

57.84

Normalized a c t l v l t y[T8q (GW a ) - ' ]

28.40

25.60

23.42

25.95

25.00

29.43

Average n o r m a l l z e d a c t l v l t y , 1980-1984[TBq (Gw a ) - ) ]27.0 i 1.8

Table 24, continued

Actlvlty (TBq)
Country and reactor
1980

198 1

1982

I983

1984

1985

0.58

0.84

0.17

0.82

1 .O

I .2

0.8

1.9

1 .I

10.1

14.6

0.05
1.9

0.8

3.1
0.043
1.7
0.43

2.6
0.41
1.8
0.59
2.0
0.79

0.87
1.2
0.47
0.16
0.90
0.71

8 W K sf InlandOlk l luoto

Germany. Federal Rep. of
Brunrbittel
0.09
Gundremnlngen
Isar
8.8
KrumnelPhrllppsburg I
0.4
Wurgassen
2.1

1

Italy
Caorso

0.2

0.1

9.4

Netherlands
Dodevdard

0.6

0.5

0.3

0.22

0.15

SuedrnBarseback 1 ,2
forsmark 1.2
Oskarshamn 1.2
Ringhdls 1

0.79
0.14
0.97
1.7

0.75
1.2
0.6
1 .8

0.52
1.1
17.56
1.9

0.68
1 .0
0.71
1 .0

1 .O
1.2
0.54
0.7

0.58
1.4
0.63
0.525

0.229
0.807
0.474
0.324
0.294

O I L

0.041
1.18
1.25
0.266
1.45
0.00000005
0.116
2.97
0.788
0.0006
4.63
0.041
0.317

0.047
I .23
0.25

Untted Stater
8jg Rock Point
Brodns Ferry
Brunrulck
CooperDresden 2,3Duane Arnold
Fltzpatrlck
Hatch 1
Hatch 2
Humboldt Bay
Lacrosse
LasalleMillstone 1!4ontlcellot d l n ~ nlle Polnt
Oyster Creek
Pmch Bottom
PI lgr lm
Ound Cltles
5u:quehanna
Vermont Yankee
kHP

0.104
0.525
0.396
0.0036
2.664
l .Ol

5.698
1.38
1.48
0.381

0.110
0.884
0.836
1.806
0.309
0.336
0.224
0.05
0.00000008
0.152
0.024
0.429
3.811
0.343
1.362
0.006
0.002
2.864
2.190
0.034
0.097
0.229
0.00002 0.0000001
0.187
0.215
0.988
0.183
1.362
0.877
1.262
0.219
0.44
0.289
0.032
0.111
0.259

0.821
1.18
3.85
0.281
0.00005
0.101
3.50
1.26
0.002
4.59
0.157
0.310
0.000
0.292
0.324
0.747
0.577
0.144
0.332

0.28
0.0013
0.11
1.45
0.67
0.04
4.74
0.014

0.381
1.32
0.544
0.201
0.414

0.16
0.13

0.02

0.055

Total annual electrlc energy
generated ( G W a ) 14.17

14.77

15.85

15.78

16.80

18.0

liormallzed actlvtty[TBq (GW a)-I]

1 .864

7.811

1 .UOl

1.58

l .Ol

2.227

Average normalized actlvlty. 1980-1984[Taq (Gw a)-]]2.1 i 0.5

Table 24, contlnued

Actlvlty (T8q)
Country and reactor

f rdnreMugey 1
Chlnon A2.3
St. Laurent Al .Z

1
4
16

14
4

2

1

1

0.03

0.3

0.5

Total annual electrlc energy
generated (GW a)
3.868

4.847

5.382

Normallzed activity[T8q (GW a)-1)

83.3

95.7

Italy
Latlna1

Unlted Kingdom
Berkeley
BradwellChapelcrossOungenelr A
Oungeness 8
Hartlepool
HeysmanHlnkley Polnt A
Hinkley Polnt 8
Hunterston A
Hunter:ton 8
Oldbury
Sl leuel 1
Trawsfynydd
b4ylta81.5

Average normallzed actlvlty. 1980-1984[TBq (GW a)-1 ]USSRChernobyl
Kursk

103.7

6.223
116.4

5.738
86.75

96 i 13

3.5
2.0

Total annual electrlc energy
0.317
generated (Gu a)
Normallzed actlvlty
1r8q ( ~ a)-']
w

5.885

0.308

0.356

0.376

0.580

1.734

H H R sArgent lna
AtuchalmbalseCanadaBruce A
Bruce B
Gentllly
I'lckerlng A
Plckerlng 0
Polnt Lepreau

290

410

310

240

410
3.48

370
16.1

888

740

962

1600

8.0
276

3 70

604
0.6
14
330
330
68

1060
21.5
30.7
330
380
24

0.4

48 1

Total annual electrlc energy
generated (CW a)
4.531
Normallled actlvlty
Ires ( G W a)-11
365.e

4.810
796.3

0.05

4.665
351.4

Average normdllzed actlvlty, 1980-1984[Teq ( G U a)-]]290 t 68

0.78
3 70
44
9.1

5.752
212.6

6.408
234.6

7.370
296.1

T a b l e

25

Carbon-14 d l r c h a r g e d f r o m r e a c t o r s . 1980-1985[ E l l . 03. R1. W l ]A c t l v l t y (GBq)Country and r e a c t o r

PWRs

a_/

Flnland
Loviisa 1
Germany, F e d e r a l Rep. o f
Blblls A
Blblls 8
Grafenrhelnfeld
GrohndeNeckarwesthelm
11 ( 9 6 )Obrlghelm
22
Phlllppsburg 2
Stade
41 ( 7 8 )Unterweser
26
USSR

Arrnenlan
Kola 3
Kola 4
Novovoronezh 3
Novovoronezh 4400-500

400-500
780

400-500
780

400-500
780

140
140-540140
140-540140
140-540140
140-540

400-500
780
5 10
140
140-540

T o t a l annual e l e c t r l c energy
g e n e r a t e d (GW a)
3.654

4.186

7.285

7.020

7.270

7.856

Normallzed a c t l v l t y[GBq (GW a ) - l ]

482

29 7

333

283

39

0.9

240
300
310
550
220
280

2 60
770
320
190
250
360

329

Average norrnallzed a c t l v l t y . 1980-1984[GBq (GW a ) - l ]

345 i

80

Flnland
TVO 1
Germany. f e d e r a l Rep. o f
Brunsbiittel
30
Gundremnlngen
lsar
KrumnelPhlllppsburg 1
Wirgassen
270
T o t a l annual e l e c t r l c energy
generated (GW a)
1.195
Normalized a c t l v l t y[GBq(GWa)-l]
251.2

240

80

180

4.8

340

6.3
270

91
88

200
15

1.521

457.7

Average n o r m a l l z e d a c t l v l t y . 1980-1984[GBq (GW a ) - ) J

1.275

206.9
330 i 110

1.880

295.7

3.106

434.6

4.117

455.8

Table 25. continued

Actlvlty (GBq)Country and reactor

Argentina
Atucha2300

Total annual electrlc energy
generated (GW a)
0.2488
HormdIized activtty[GBq(Gwa)-l]
9244

2700

0.3021
8938

1800

590

0.20018996

0.26682195

Average normallzed actlvity. 1980-1984
b336 t 3333
[GBq (GW a)-]]USSR

Chernoby 1
Ignallno 1
Kursk 1.2.3
Lenlngrad 1.2.3.4
Smolensk 1
Average normallzed actlvlty. 1980-1983[ 66s (GW a) - 1 )
1300

a/

Carbon dloxlde. carbon monoxide and hydrocarbon
bound values In parentheses.

4 50

0.195
2308

370

0.16788166

T a b l e

26

Iodine-131 discharaed In airborne effluents from reactors. 1980-1985[Al. 85, 817, 818. 819. 836. Dl, f1. HI. H2, H3. H4, 33. K1, L1,
P5. P9. 51. S2. 53. 5 5 . 59. S10. Sll. 13. 74. 15, 18. T9. 1101

Activity ( G B q )Country and reactor

F Inland
Lovi isa

0.002

Germany. Federal Rep.
8lblis A,B
Grafenrhelnfeld
Grohndetdeckarues thelm
Obrlghelm
Phllippsburg 2
Stade
Unterverer
Japan
Genkailkata
MI h a m
00I
Senda iTakahama

SuedenRinghals 2
Rlnghals 3
RInghals 4
USSR
Armenlan 1.2
Kola 1.2
Hikolaev 1~ o v o v o r o n e z h 1 .2
Novovoronezh 3.4
Novovoronezh 5
lcovno 1.2
Unlted States
Arkansas One-l
Arkansas One-2
Beaver Valley
Cal laday
Calvert ClItts 1.2
Crystal CIver
Davls Besse
Dlablo Canyon
Donald Cook 1.2
Farley 1
Farley 2
Fort Calhoun I
H.8. Roblnson
Haddain Neck
Indlan Point 1,2
Indlan Polnt 3
Kewaunee
Maine Yankee
McGulreUlllstone Pt. 2
North Anna
OconeePal lsades
Polnt Beach 1.2
Prairie island 1.2
R.E. Glnna

0.01

7.8
0.257
14
0.424

Table 26, c o n t l n u e d

A c t l v l t y (G8q)
Country and r e a c t o rRancho Seco
0.2557
Salem 10.1447
Salem 20.0020San Onofre 1
0.0086
San Onotre 2.3
Sequoyah0.0015
1.1951
St. L u c l e 1.2
S u r r y 1.2
0.5994Three H l l e I s l a n d 1 Three H l l e I s l a n d 2 THI 2/CPICOR
Trojan0.4218Turkey P o i n t 3.4
1.9203V l r g l l C. Sumner
Wolf Creek
Yankee Roue
0.0023
Zlon I , ?
0.02200.14210.34450.03090.08840.01420.1332
0.0958

.

0.075
0.892
0.009
0.0001
5.77
0.025
3.73
2.76
0.0
0.0
0.0
0.068
5.75
0.00080.870
0.01910.047
0.0003
15.10
0.186
20.05
27.90
0.0
0.0
0.0
0.143
1-01
0.00007

0.237
0.407
0.017

0.0296
2.16451.66500.0006
0.0367
10.1750
2.11271.39491.03600.20358.10300.00620.19050.01110.2409

0.114
0.189

0.231
0.177

T o t a l annual e l e c t r l c energy
generated ( G W a)
29.34

32.90

36.79

41.09

44.94

35.65

Hormallzed a c t l v l t y
LGBq ( G W a ) - ] ]

1.91

1.63

1-37

2.31

1.07

I .54

16.40
4.81

0.002

Average n o r m a l l r e d a c t l v l t y . 1980-1984[G8q (GU a ) - 1 ]

1.75

t

0.33

Table 26, continued

A c t l v l t y (GBq)Country and r e a c t o r
1980

FinlandOlklluoto

0.0042Germany, F e d e r a l Rep.
8runsbuttel
Gundremnlngen
Isar
KrumnelPhlllppsburg 1
Wijrgassen
Italy
Caorso

of
0.04
0.08
2.2
0.01

Japan
Fukushlma I , 1-6
Fukushima 11. 1.2
Hamaoka 1 . 2
ShlmaneTokai 2
Tsuruga

0.010
n/d
0.067
0.027

Netherlands
Oodewaard

0.09

SwedenBarseback 1
Barseback 2
Forstnark 1
Forsmark 2
Forsmark 3
Oskarsharnn 1
Oskarsharnn 2.3
Rlnghals 1
Unlted S t a t e s
819 Rock P o l n t
Browns F e r r y
Brunswlck
CooperDresden 1
Dresden 2.3
Duane Arnold
Fltzpatrlck
Grand Gulf
Hatch 1
Hatch 2
Humboldt Bay
Lacrosse
LasalleUlllstone 1
HontlcelloNIne U l l e P o l n t
Oyster Creek
Peach Bottom
Pllgrlm
Quad C l t l e s
Susquehanna
Vermont Yankee
WNP-2

1981

1.9

-

0.06
0.02
0.02
0.4
0.3
1.4
0.03962.43469.91 60
0.63210.1343130.6100
1.6576
2.837947.3600
0.53280.16177.91800.751 10.440334.9650
1.08783.248612.24700.4107

Total annual e l e c t r l c energy
17.33
g e n e r a t e d (GW a )

17.63

19.19

18.63

19.97

29.18

Normallzed a c t l v l t y[GBq(GWa)-l]

9.690

8.316

9.044

4.230

0.824

15.23

Average normalized a c t l v l t y . 1980-1984[GBq (GW a ) - l ]9 . 3 i 4.9

Table 26, continued

Actlvlty (Gap)Country and reactor

Argent lna
Atucha 1
Emba 1 se
CanadaBruce A
Bruce 8Gentllly
Plckerlng A
Plckerlng 8
Polnt Lepreau

0.130

0.104

1 .I84

0.359

2.801
0.004

0.05
0.052

0.155
-

0.063

0.070

0.070
0.0
0.0096

0.13
0.151
0.0

0.056
0.040
0.0042

Total annual electrlc energy
4.531
generated (GW a)

4.810

4.586

5.645

5.729

7.006

Normallzed actlvlty
IGBo ( W a)-l I

0.12200.27760.10270.53780.14610.1070

Average normallzed actlvlty. 1980-1984[G~Q(GW a)-I]

0.23 i 0.08

USSR

Chernobyl 1.2.3
Kursk 1,2,3
Smolensk 1
Lenlngrad 1.2.3.4

189
25.7

300
28.8

118
112

-

I20

110

Total annual electrlc energy
generated(GWa)
3.112
Normallzed actlvlty[GBq (GW a)-]]

67.69

3.077
150.0

Average normallzed actlvlty. 1980-1984[G~Q(Gw a)-)]

41.4
45.1
72.1
74

40

47

89

40

3.583

4.355

3.5

3.5

94.89

53.27

36.86

24.86

80 t 40

Table 26, c o n t l n u e d

A c t l v l t y (GBq)Country and r e a c t o r

G C R sFranceBugey 1
Chlnon 2.3
S t . L a u r e n t A1 .2
Italy
Latlna0.003

0.0001

0.002

Japan
Tokal 1
Unl ted Klngdorn
Berkeley
Braduel 1
ChapelcrossDungeners A
Oungeness 8
Hartlepool
Heysharn
Hlnkley Polnt A
Hlnkley Point 8
Hunterston A
Hunterston B
Oldbury
SlzeuellTrausf ynydd
wy 1f a

0.518

0.481

0.481

1.7
0.11
0.89

1.8
0.22
1.22

1.9
0.2
0.9

0.518

0.407

0.4

0.37

T o t a l annual e l e c t r l c energy
g e n e r a t e d (GW a )0.82150.88990.8811

1.363

2.372

1.647

Normallzed a c t l v l t y[GBq (GW a ) - 1 ]

0.55

0.9 1

3.21

1.86

2.07

0.64

Average n o r m a l l z e d a c t i v i t y . 1980-1984[Gas ( G W a ) - ] ]1.4 t 1 . 1

T a b l e

27

Isotoplc composltlon of lodlne dlscharged from reactors
In the Unlted States. 1982

IT51

Actlvlty (G8q)
Reactor1 - 1 31

1-132

1-133

1-134

1-135

0.05

6.50

o.oC067

0.41

P W R sArkansas 1
Arkansas 2
Beaver Valley
Calvert Cllffs 1.2
Crystal Rlver
Oavls-Besse
Donald Cook 1.2
Farley 1
Farley 2
fort Calhoun
H.8. Roblnson
Haddam Neck
Indlan Polnt 1.2
lndlan Polnt 3
Kewaunee
Ualne Yankee
UcGulreMillstone Pt. 2
North Anna
OconeePall sades
Polnt Beach 1.2
Pralrle Island 1.2
R.E. Glnna
Rancho Seco
Salem I
Salem 2San Onofre 1
San Onofre 2 , 3SequoyahSt. luc leSurry 1,2
Three nlle lsland 1
lhree Ulle Island 2
TrojanTurkey Polnt
Vlrgll C. SumnerYankee Rove
Zlon 1.2
Total annual electrlc energy
generated (GW a)
21.042
Normallzed actlvlty[ c a ~ccw a)-1 j

2.60

Table 2 7 . c o n t l n u r d

A c t l v l t y (GBq)
ReactorB l g Rock P o l n t
Browns F e r r y
Brunswlck
CooperDresden 1
Dresden 2.3
Duane A r n o l d
fltzpatrlck
Hatch 1
Hatch 2
Humbolt Bay
Lacrosse
LaralleHlllstone 1
HontlcelloNlne n l l e P o l n t
Oyster Creek
Peach B o t t o m
Pllgrlm
Quad C l t l e s
Susquehanna
Vermont Yankee

1-131

1-13?

1-133

1-134

1-135

0.09293.959032.7450
4.14400.000221.49700.206816.0580
6.54902.50120.74000.8695
1 .a241148.3700
1.06930.0001120.6200
0.350868.4500
1.53180.52171.54291.7723
2.105383.9900
4.03300.0796202.02000.305282 .a300
16.2430
0.02634.03306.13400.10750.0892

0.05373.67022.69360.079232.5600
1.11740.873210.6560

16.02103.14870.2875137.2700
28.3050
3,392946.9900

32.59101.72790.3774
216.4500
17.6860
3.9590743.7000

0.03570.7141-0.0053T o t a l annual e l e c t r l c energy
g e n e r a t e d (GW a )10.29
Normalized a c t l v l t y[ G B q (GW a ) - ] ]

13.56

0.46

56.27

0.80

137.03

T a b l e

28Partlculates dlscharqed In alrborne effluents from reactors. 1980-1985[Al, 85, 89. 817, 818. 819. 836. 03, fl. f4, HI, HZ. H3, H4. 33, K1, L1,
P5. P9. S1. 52. 53, 55. 59, S10, 511. T3. T 4 . T5, 18. 19. 1101

Actlvlty (G8q)
Country and reactor
1980

FinlandLovll sa
France a_/Blayals 1,2.3.4
Bugey 2.3.4.5
Chlnon 81.2
Chooz
Cruas 1.2.3.4
Damplerre 1.2.3.4
Fessenhetm 1,2
Gravellnes 1-6
Paluel 1.2
St. Laurent 81.2
Trlcastln 1,2.3.4
Germany, federal Rep.
Btblts A.8
Grafenrhelnfeld
Neckaruesthelm
Obrlghelm
Stade
Unterueser
Netherlands
Borssele
SuedrnRlnghals 2
Rlnghals 3
Rlnghals 4
USSR
Armenlan 1.2
Kallnln 1
Kola 1.2
Kola 3.4
Nlkolaev 1.2
Novovoronezh 1.2
Novovoronezh 3,4Novovoronezh 5
Rovno 1 .2
Zaporozhe 1
Unlted States
Arkansas 1
Arkansas 2
Beaver Valley
Callauay
Calvert Cllffs 1.2
Crystal Rlver
Davls 8esse
Olablo Canyon
Donald Cook 1.2
farley I
farley 2
fort Calhoun
H.B. Roblnson
Haddam Neck
Indlan Polnt 1
lndlan Polnt 3
Keuaunee
Maine Yankee
McGulreUlllstone Pt. 2
North Anna
OconeePal lsades
Polnt Beach 1.2

I981

I982

1983

1984

1985

Table 28. continued

A c t l v l t y (GBq)Country and r e a c t o r
1980

Prairie Island
0.003
R . E . Cinna
0.204
Rancho Seco
0.113
7 . a ~
Salem 1
Salem 2San O n o f r e 1
31.108San Onofre 2.3
Lequoyah
0.094
5:. L u c i e 1
1.099
51. L u c l e 2
Surry 1.7
0.085
Three n i l e I s l a n d 1 0.011
l h r e e H i l e I s l a n d 2 0.021? # I 2/EPICOR
lrojdn0.507
0.b88
Turkey P o i n t
V i r g i l C . Sumner
wolf Creek
0.001
Yankee Rove
Zion 1.2
0.689

1981

1902

1983

0.003
0.176
0.030
17.564
0.703
0.348

0.001
0.471
0.959
0.157
0.048

0.00120.00080.00831.55
1.30
0.00009
0.00270.657b0.0019

1984

1985

0.451
0.681

0.001
4.515
5.180

0.751
0.019
0.001

L.092
U.OOb
0.002

1.443
0.052

0.333
0.037

0.002
0.272

0.010
2.930

0.00310.0148O.OUjO2
0.0046
0.0559
0.000
2.889
0.152
3.294
0.0001
O.bJI
2.263
0.59rl
0.0OSU
0. 0002
7.21
I .05
0.0024U.00000005
0.000009 0.GO02
0 000006 6.600
0.0901
0.05880.0692172.060.001
0.00030.0472
15.95
114.710.813
2.065

i o t a 1 a n n u a l e l e c t r l c energy
generated ( G W a)
29.10

39.35

41.77

49.81

56.54

53.82

Normalized a c t t v l t y[GBq ( G W a ) - I ]

2.703

1.306

8.40

7.80

l .Ol

-

2.514

Average n o r m a l l r e d activity. 1980-1984[GBq ( G d a ) - ] ]4.5 r 2.9

Table 2 8 . continued

A c t l v l t y (GBq)Country and reactor

FlnlandOlklluoto

0.511

Germany. Federal Rep. of
Brunsbuttel
0.47
Isar
0.64
Krumnel
0.14
Phll lppsburg 1
Wirrgassen
1 .6
Netherlands
Dodewaard

0.074

SwedenBarseback 1
Barseback 2
Forsmark 1
Forsmark 2.3
Oskarshamn 1
Oskarshamn 2
Rlnghals 1

0.43
0.11
0.03
0.031
0.025
0.26

Unlted States
8lg Rock Polnt
Browns Ferry
Brunswick
CooperDresden 1
Dresden 2.3
Duane Arnold
Fltzpatrlck
Grand Gulf
Hatch 1
Hatch 2
Humboldt Bay
Lacrosse
LasalleUlllstone 1
UonticelloNine Hlle Polnt
Oyster Creek
Peach Bottom
Pllgrlm
Ouad Cltles
Susquehanna
Vermont Yankee
WNP 2
T o t d l annual electrlc energy

generated (GW a)

13.99

14.53

15.15

15.25

16.29

16.40

Norindlired actlvlty[G8q(GWa)-I]

28.13

26.72

23.14

76.59

61.99

15.85

Avrrdgr normdliied activity. 1980-1904[ G B q (GW ,)-I]

43.3 : 24.4

Table 28. c o n t l n u e d

A c t l v l t y (GBq)Country and r e a c t o rFrance8ugey 1
Chlnon 2.3
S t . l a u r e n t A1.2

0.8
0.3
7.4

0.5
0.1
1.7

0.4
0.3
1 .O

0.3
0.3
0.0

0.3
0.2
1.2

0.2
0.1
1.2

0.111
0.037
0.037

0.037
0.037
0.037

0.037
0.031
0.333

0.333
0.925
0.074
2.96
0.140
0.37
0.296
0.296

0.296
1.132
0.037
2.59
0.333
0.333
0.407
0.259

0.296
0.629
0.037
0.74
0.185
0.222
0.37
0.111

0.037
0.074
0.296
0.111
0.37
0.37
0.333
0.518
0.031
0.74
0.148
0.444
0.333
0.074

0.037
0.185
0.259
0.111
0.037
0.185
0.296
0.518
0.037
0.740
0.111
0.296
0.370
0.074

0.02
0.09
0.24
0.12
0.03
0.04
0.34
0.51
0.23
0.51
0.51
0.14

T o t a l annual e l e c t r l c energy
generated(GWa)
3.868

4.847

5.382

5.985

6.223

5.738

Normallied a c t l v l t y[GBq ( G W a ) - ' ]

1.61

0.87

0.79

0.80

0.74

140
10
1.61

120
6.8

140
9.4

U n l t e d Kingdom
Berkeley
Braduel 1
Oungeness A
Oungeness 0
Hartlepool
HeyshamHlnkleyPolntA
HlnkleyPolntB
Hunterston A
Hunterston 8
Oldbury
Slzewelllrausfynydd
Wylfa

2.86

Average n o r m a l l z e d a c t l v l t y . 1980-1984[GBq (GW a ) - I ]1.39 t 0.79

USSR

Chernobyl 1,2
Kursk 1.2
L e n l n g r a d 1.2.3.4
Smolenks 1

9.3
6.5

8.6

T o t a l annual e l e c t r l c energy
g e n e r a t e d (GW a )
3.172

3.077

3.583

4.355

3.5

3.5

Normallied a c t l v l t y[GBq (GW a ) - l ]

2.11

2.40

34.81

36.22

42.60

2.93

Average n o r m a l l z e d a c t t v i t y .
[ G B (GW~a)-]]

1980-1984
15.69 t 1b.19-

Table 28, continued

A c t l v l t y (GBq)Country and r e a c t o r
1980

1981

1982

1983

1984

1985

0.016

0.014

0.00740.00860.0046

0.022

0.056

0.093

0.100

0.048

0.060
0.117

0.044
0.210

0.118

0.170

0.089

0.020

0.022
0.012

0.027
0.013

T o t a l annual e l e c t r i c energy
generated (GW a )
4.283

4.508

4.386

4.742

4.964

5.956

Normallzed a c t l v l t y[GBq (Gw a ) - ' ]0.05840.04310.01430.04250.0531

HWRs
Argent l n a
Atucha 1
CanadaBruce A
Bruce B
Gentllly
Pickeriog A
Plckerlng 0
P o i n t Lepreau

0.0406

Average n o r m a l l z e d a c t l v l t y . 1980-1984[ G B ~(GW a ) - ) ]0.040 k 0.016

a/ Reported d a t a f o r France l n c l u d e r halogens w l t h p a r t l c u l a t e s .

T a b l e

29

Llauld releases excludlnq trltlum from reactors. 1980-1985(85. 89. 816. 817, B18. 879. C4, F1. fa. HI. HZ. H3. H4, J3. K1, LI, PI. P5,
P9, 51. 52. 53. 55.59. 510, 511, 13, T4. 15. T8. 19, 110, V5J

A c t l v l t y (GBq)Country and reactor

ilnlandLovl i:a17.819Belglum
Doel 1.2,3
T I hange

98.0
56.0

FranceBlayais l,2Buqey 2.3.4.5
Chinon B. 1 ,2Chuoz
Cruds 1-4
Dampierre 1-4
Fessenheim 1.2
Grave:lnes 1-4
Paluel 1.2
it. Laurent 81.2
Irlcastln 1-4

2.757
57.0
294.0

13.47422.31820.46318.167

n/d
n/d
0.04
0.03
n/d
0.009

n/d
n/d
0.034
6.021
n/d
0.0081

33.0
21.5

566
9

59
262
148

17

ter-any. Federal Rep. ot
Biblls A.8
11.1
Grafrnrhelnfeld
Grohndetdeckarwes thetm
0.3
Obrlyhelm
3.0
Philippsburg 2
Stade
3.0
Unterweser
1.8
Italy
Trlno
Japan
Genkal 1.2
Ikata 1.2
riharna 1,2.3
001 1.2
Senda 1
Takahama I,?
Neth~rlands
Borssele
SwedenRlnghals 2
Rinyhals 3
RInghals 4
USSR
Armen l an
Hovovoronerh 5n/d0.01 1
0.17
0.06

n/d
0.0026
0.1
0.20

n/d
0.0004
0.1
0.024

0.052

0.02

0.0078

n/d
n/d
0.096
0.071
n/d
0.0014

4.44

7.03

12.58

7.03

22.2

I20
0.1

78.7
22.2

57.7

69.0
41. I
36.0

153.2
106.2
47.2

11.4

2.1
0.039
0.16

United States
126.540
Arkansas 1
Arkanias 2
152.810Beaver Valley
3.848
Callaway
Calvert Cllffs I,? 167.610
Crystal River
5.402
Davis Besse
7.659
Dtablo Canyon
Donald Cook 1 .2
50.690farley 1
2.287
Farlry 2
F o r t Calhoun 1
18.648H.8. Robinson
13.246
Haddan Neck
10.21246.620indian Polnt I .2

47.72
21 .0959.02

Ta b l e 29, c o n t i n u e d-

A c t l v l t y (GBq)-

C o u n t r y and r e a c t o rlndlan Folnt 3
107.300
Kewaunee
22.829Halne Yankee
10.989
HcGulreHlllstone IJt. 2
103.970
N o r t h Annd
38.850Oconee56.980P d l i:ades
0.323
23.273P o i n t Bedrh 1.7
Pralrle Island
0.488
R . E . Gtnnd
0.725
Rancho Secn
0.140
5alem I
98.050
Salem ?
36.593414.400
San Onofre 1
San Onoire 2.3
SequoyahSt. Lucie 1
67.320
St. L u c i e 2
S u r r y 1,2
142.450
Three M i l e I s l a n d 1 6.771
Three H l l e ]:land 2 0.001
Trojan29.119lurkey Polnt
25.086
VlrgilC.Surrmcr
Wolf C r e e k
Yankee Roue
0.647
17.538
Zlon 1,2

T o t a l annual e l e c t r l c enprgy
g e n e r a t e d (GW a )
35.50

40.55

44.97

53.09

61.74

57.02

Normaltzed a c t l v t t y[ G B Q ( G Wa ) - l ]

111.5

82.87

214.7

160.9

102.7

92.10

Average n o r m a l i z e d a c t l v l t y , 1980-1984[GBq (GH a ) - ] ]132.4 t 49.5

Table 29, c o n t l n u e d

A c t l v l t y (GBq)Country and r e a c t o r
1980

1981

1982

1983

1984

1985

Total annual e l e c t r l c energy
g e n e r a t e d (GW a )
19.05

19.64

21.92

21.35

23.45

23.57

Normalized a c t l v l t y[GBq (GW a ) - l ]

71.31

175.3

95.86

156.5

40.9

Germany. F e d e r a l Rep. of
Brunsbuttel
9.6
Gundremnlngen
Isar
5.9
KrumnelPhlllppsburg 1
4.1
Wijrgassen
10
Italy
Caorso0.44

Japan
Fukushima I
1.2.3,4.5,6Fukushlma 11-1 .2
Hamaoka 1.2
Onagawa
ShlmaneTokal 11-1
Tsuruga

0.037
0.31
0.28

Netherlands
Dodewaard

17.76

Sweden8arseback 1 .2
Forsmark 1 , 2Oskarshamn l .2
Rlnghals 1

57.4
0.25
99
99

1.9

-

2.6

Unlted S t a t e s
Blg Rock P o l n t
Browns F e r r y
Brunswtck
CooperDresden 1
Dresden 2.3
Fltzpatrlck
Grand Gulf
Hatch 1
Hatch 2
Humboldt Bay
Lacrosse
LasalleLlmerlck
Ulllstone 1
UontlcelloNlne U l l e P o l n t
Oyster Creek
Peach Bottom
Pllgrlm
Ouad C l t l e s
Susquehanna
Vermont Yankee
WNP 2

78.23

Average normallzed a c t l v l t y . 1980-1984[GBq (GW a ) - ] ]115 i 47

Table 29. c o n t l n u e d

A c t l v t t y (Gap)Country and r e a c t o r
1980

1981

1982

France8ugey 1
Chlnon A1,2
S t . Laurent A1.2

118
38
407

70
17
237

40
19
200

ltaly
Lat l n a

59.2

86.2

162.8

Japan
Tokal 1

0.31

0.22

0.23

U n l t e d Klngdom
Berkeley
2190
Bradwell
1443
Chapelcross322
Oungeness A
629
Oungeness 8
Hartlepool
HeyshamH l n k l e y P o l n t A 5106
Hlnkley Polnt 8141
Hunterston A
13500
1554
Hunterston 8
Oldbury1406
Slzeuell
1924
5 18
Travsfynydd
74
Wylfa
T o t a l annual e l e c t r l c energy
generated(GWa)
3.868
Normallzed a c t l v l t y[G8q (GW a ) - ) ]

7609

1321
1735
1776
714

662
98 1
4144
840

3367
100
8436
2160
2290
1143
281
51.8

2660
44.4
8695
301 6
1994
1036
485
111

4.847
4890

1983

1984

1985

0.14

0.12

0.1

667.1
962
307 1
864.2
7.4
7.4
7.4
1306.1
812
281 2
2653
2620
658
350.9
78.3

366.3
817.1
481
1750.1
51.8
118.4
59.2
2249.6
928.7
2664
331 9
1713.1
889.1
370
96.2

5.382

5.885

4632

2901

6.223

290
1510
2000
2190
174
218
60
3720
840

1210
1010
430
48

5.738

2582

2406

Average n o r m a l l z e d a c t l v l t y . 1980-1984
4520 t 1790
[G8q (GW a ) - ) ]Argent l n a
Atucha 1
Ernbalse
CanadaBruce A
Bruce 8Gentllly
Plckerlng A
Plckerlng 8P o l n t Lepreau

81

51

37

51

51
25.9

51
1.91

163

81

78

74

7
9.7
32
9
1.6

-

2
8

0.6
18

4
22
11
18.71

72
7
1
27
27
13.10

T o t a l annual e l e c t r l c energy
generated (GW a )
4.283

4.508

4.386

5.283

5.729

7.310

Normallzed a c t l v l t y[GBq (GW a ) - ) ]

19.74

21.89

23.79

22.21

14.86

2
13

41.10

Average n o r m a l l z e d a c t l v t t y . 1980-1984[GBq (GW a ) - 1 )n/d = Discharge n o t d e t e c t e d .

25.7

.t

8.7

T a b l e

30

Radlonucllde c o m o s l t l o n of llauld releases excludtnp trltlum
from reactors in the Unlted States. 1982
(151

Actlvlty (TBq)
Reactor1-131
AfkdnSdS I
ArkanSdS 7
Beaver Valley
Calvert Cll!ts I,?
Crystal klver
Oonalc Cook 1.2
Farley 1
Farley 2
fort Calhoun
H.8. Roblnson
tiaddarn Neck
Indldn Polnt 1.2
lndlan Polnt 3
Kedaunee
nalne Yankee
Mctulre4lllstone Pt. 2
North Anna
OconeePalisades
Poict Beach 1.2
Pralrle Island
R.C. Glnna
Rancho Zeco
Salem 1
Salem 2San Onofre 1
San Onofre 2 , 3SequoyahSt. Lucle
5urry 1.2
lhrer nlle 1:ldnd 1
Three 4i le Illand 2
TrojanTurkey Polnt
Vlrgll C. SumnerYankee Rode
Zlon 1
Zion 2
Total annual electrlc energy
21.042
generated (GW J )

1-13?

1-133

1-134

1-135

Ma-24

Cr-51

Hn-54

Mn-56

Co-57

Co-58

Table 30, contlnued

Actlvlty (TBq)
Reactor
fe-59

Co-60

211-65

Sr-89

Sr-90

Zr-95

Zr-97

Nb-95

Nb-97

no-99

Tc-9910

19.09

0.11

1.26

0.23

1.24

0.04

1.30

0.34

0.47

0.09

ATkdtlSdS 1
Arkansas 2
Beaver Valley
Calvert Cllfts 1,2
Crystal Rlver
Oonald Cook 1.2
Farley 1
Farley 2
Fort Calhoun
H.B. Roblnron
Haddam Neck
Indlan Polnt 1.2
lndlan Polnt 3
Kewaunee
Walne Yankee
McGulreMtllstone Pt. 2
North Anna
OconeePallsades
Polnt Beach 1.2
Pralrte Island
R.E. Glnnd
Rancho Seco
Salem 1
Salem 2San O n o f r e 1
Oan Onotre 2 , 3SeauoyahSt. Lucle
Surry 1.2
Three ulle Island 1
Three Wlle Island 2
TrojanTurkey Polnt
Vlrgll C . S u m e r
Yankee Roue
Zlon 1
Zlon 2
Total annual electrlc energy
generated (Gil a)
21 -042
Normallzed actlvlty[GBq (GW a)-I]

0.50

Table 30. contlnued

Actlvlty (TBq)
ReactorArkansas 1
Arkansas 2
Beaver Valley
CalvertCllffsl.2
Crystal Rlver
Donald took I .2
Farley 1
Farley 2
Fort Calhoun
H.8. Roblnson
Haddam Necklndlan Polnt 1.2
Indldn Polnt 3
Keudunee
Malne Yankee
UcGu l re
Ulllstone Pt. 2
North Anna
OconeePalisades
Polnt Beach 1.2
Pralrle Island
R.E. Glnna
Rancho Seco
Salem 1
Salem 2San Onof re 1
San Onofre 2,3SequoyahSt. Lucle
Surry 1.2
Three Ulle Island 1
Three Mlle Island 2
TrojanTurkey Polnt
Vlrgtl C. SumnerYankee Roue
Zlon 1
Zlon 2

Ru-103

Ru-106Ag-1 lOm

Sb-124

1.380
2.083

1.680
0.596

-

0.299
0.001

0.940

6.031
0.577
0.008
8.066
0.176
0.503
0.020
0.008

0.010
0.002
0.015
0.001
0.096

0.001
0.002

-

0.109

-

0.014

0.265
0.206
5.476

0.001

0.026

0.012

-

0.088
0.004
0.002

0.107
1.066
0.170
1.232
0.026
0.065

0.336

0.147
3.271
1.787
0.331

0.123

0.002

-

0.001

-

0.814
0.236

1.854
0.073
0.714

-

-

0.010

0.174
0.145
0.024
0.367
0.659

-

-

0.295
0.374
0.020
0.002
0.226
2.157
2.087

-

1.221
0.001

0.012

-

Sb-12510.064
0.178

0.183
2.812
4.403
0.692

0.110

0.057
0.035
0.026

Cs-134

Cs-136

Cs-1379.250
32.1900.156
16.317
0.303
6.734
0.072
0.073
0.851
0.031
0.078
9.139
0.290
0.422
0.090

0.009
0.291

0.110

21.09059.9400.259
29.711
0.407
13.246
0.180
0.101
1.912
0.222
0.455
20.498
1.428
1.650
1.273
0.017
46.9905.254
17.353
1.384
11.544

0.021
-

0.792
0.692
2.179
2.827
20.01732.3382.642
11.063
0.640
7.992

1.869
2.446
0.019

0.433
0.356
1.462
1 .876
7.215
1.088
0.733
6.660
72.150
0.305

0.252
0.377

-

-

-

0.009
0.474
0.005
0.015
0.102

1.025
0.060

-

1.388
0.004
0.011

-

0.011

0.019
0.108
1.872

-

Ba/La-I40 Ce-141
1.883
2.364

-

0.071

0.134

3.848
0.240

0.655
0.033

0.014

0.022
0.030
0.165

0.028

0.012
0.241
0.103
0.231

0.094

0.011

0.004

12.284
0.005
0.271
0.249

0.050
0.026

-

0.002

1.561
0.607
0.314

2.757
0.032

0.352

1.502

0.002

0.009

0.06

0.37

-

0.124
0.078

-

0.024

-

-

-

0.006

3.482
11.729107.300
1.532
0.001
1.624
2.250

0.459

0.047
0.021
0.261

-

0.092
0.210

-

-

0.100
0.262

0.084
0.374

0.906
0.662

0.925
1.132

0.123

0.002
2.231
2.164

0.001
0.414
1.709

0.381

-

0.017
0.451
0.588

0.014
0.001

0.038
1.658
0.698

0.004
0.164

1.60

0.62

1.21

10.23

0.27

17.86

1.23

-

2.264
0.573

-

0.906
0.165

Ce-I44-

Total annual electrlc energy
generated (GY a) 21.042
Normallzed actlvlty[GBq(GUa)-I]

0.26

0.20

Table 30. continued

Actlvlty (Tap)
Rear tor

819 Rock Polnt
Browns Ferry
Brunswlck
CooperDresden 1
Dresden 2.3
Ouane Arnold
Fltzpatrlck
Hatch 1
Hatch 2
Humboldt Bay
Lacrosse
LasalleMlllstone 1
MontlcelloNlne Ulle Polnt
Oyster Creek
Peach Bottom
Pllgrlm
Quad Cltles
Suspuehanna
Vermont Yankee

1-131

1-132

1-133

1-134

1-135

Na-24

Cr-51

Mn-54

Mn-56

8.325
4.255
5.143

-

2.451
0.736

-

0.559
0.056

3.885
1.565
0.536

0.012
3.885
20.202
9.842

2.168
3.497
10.582
14.57810.471
0.002
0.002

0.008

0.084

-

0.063
0.182
0.206

0.102
0.548
0.437
2.316
0.175
1.158

2.501
0.313
0.045
0.005
28.416
0.659
1.469

0.001
0.004
-

0.477
0.235
1.502
4.292
0.149

0.011

2.135

-

0.033
2.956
0.777

0.038
0.282
0.074

-

0.004
0.001

-

0.005
0.008
0.002

0.400

0.020

0.255

0.004

0.059

5.402

5.476
0.002
0.429
0.064

-

-

1.054

-

0.381

0.214
0.048

0.235

-

8.695

0.045

2.782

131.720

0.892
0.018

-

0.44

0.335
0.307

0.003
9.176
0.245
0.181
2.830

0.03

1.42

0.21

0.42

13.52

4.97

-

Co-57

Co-58

-

0.071
0.644
I .3888.806-

1 -088
0.085
0.027

0.111

-

-

15.466
0.267
0.189

0.116

-

0.028

-

0.001
4.884
0.199
0.075
0.892

6.90

1.03

0.01

3.31

Zr-97

Nb-95

Wb-97

Mo-99

Tc-99m

-

0.977

-

0.326
0.033
1.487

0.315
0.400
0.492

0.005
0.555
0.025

0.001
0.681
0.062

0.242
0.003
0.028

0.533
0.015
0.195

0.356
0.002
0.165
0.304

2.257
0.216
0.799

0.34

0.58

Total annual electrlc energy
generated (GW a)
10.29
Normallzed actlvlty
I G ~ Q / ( a)]
G~

3.23

Actlvlty ( T B q )ReactorBlg Rock Polnt
Browns Ferry
Brunsulck
CooperDresden 1
Dresden 2.3
Ouane Arnold
Fltzpatrlck
Hatch 1
Hatch 2
Humboldt Bay
Lacrosse
LasalleMlllstone 1
Mont\celloNlne Mlle Polnt
Oyster Creek
Peach Bottom
Pllgrlm
Ouad Cltles
Sus~uehanna
Vermont Yankee

Fe-59

to-60

Zn-b5

Sr-89

Sr-90

Zr-95

1.099
6.771
1.029
0.448

1.869
54.76025.27165.4900.064
27.232
0.111
1.395

0.009
0.385
0.433
7.437

0.145
0.228
0.444
0.264

0.971
0.003
1.251

0.474

-

0.014

0.003

-

0.363
3.533
1.010

0.021
0.174
0.024

0.007
0.014

0.002
0.041
0.038

7.289
0.006
0.246

13.542
0.521
0.356
0.836
98.420
0.107
5.661

-

0.021
0.001
0.143

0.078
2.505
24.01313.283
2.072
0.055

30.16

0.128
0.032
0.003

-

-

-

-

0.001
0.002
0.165

0.014
0.312
0.004
0.051

0.211

72.520
0.206
0.051
0.157

0.403
0.096
0.053
0.005

0.022
0.026
0.021
0.003

0.044
0.009

10.6

0.9

0.15

0.25

3.034
0.076
0.013

-

-

-

O.bb2

0.001

0.083
0.057

0.001
0.008

-

3.504

-

0.011

-

8.288

-

-

2.886

-

0.00

1.54

0.07

0.007

-

-

-

-

Total a n n u a l electrtc energy
generated (GW a)
10.29
Normallzed actlvlty[GBu (GW a)-1 ]1.67

Table 30. continued

A c t l v l t y (Tag)
ReactorB l g Rock P o I n t
Brown: f e r r y
Brunsulck
CooperDresden 1
Dresden 2,3Duane A r n o l drltz~atrlck
Hatch 1
Hatch 2
Humooldt Bay
1acros:e
Ldra l l eMillstone 1
nont l c e l l oNine M i l e Polnt
O y s t e r Creek
Peach Bottom
Pllgrlm
Quad C l t t e r
Susquehanna
Vermont Yankee

Ru-103Ag-llOm

0.011

5.920
0.012
2.401

0.005
0.014
1.543

-

0.020

-

0.003

0.001
0.030
0.021

0.003

Sb-124

Cs-134

Cs-136

Cs-137

Ba/La-140

0.053
0.470

0.063
5.735
5.143
28.151

0.511
0.040
0.332

2.151
8.029
6.438
32.782

0.007
0.407
2.161
1.661

0.024

0.001

0.096

-

-

0.067
0.031
0.002

0.001
0.219
0.007

0.696
0.015
0.001

-

2.575
4.884
1.236
11.06337.000

-

4.329

0.607

2.246

0.026

23.865

0.366

-

-

0.004

1.099
0.003
0.284

0.001
0.810

0.001

-

-

0.004
24.013
4.736
1.654
0.065

-

0.365

0.027
0.105

0.59

13.29

2.88

0.51

0.25

-

0.108

0.020
0.003

-

-

2.024
3.312
0.777
0.685
4.255

0.001
0.001

16.835
0.607
0.124
0.262

0.06

6.61

-

2.864
0.027

-

0.012

Ce-144

Np-2390.308

T o t a l dnnual e l e c t r l c energy
10.29
g e n e r a t e d (GW a )Normallzed a c t l v l t y! G B ~ ( ~ ad) - 1 1

0.15

0.82

T a b l e

31

R a d l o n u c l i d e composition o f liauld r e l e a s e s e x c l u d l n q trltlurnf r o m G C R s ln t h e United Klnqdorn. 1 9 8 2
IH31-Actlvlty ( G B q )Reactor
5-35

Ca-45

nn-54

Fe-55

Co-60

Sr-89

Sr-90

27.41

3.60

29.84

18.13

2.08

292.9

Berkeley
BrdduellChayelcro::Oungenetr A
Hinkley A
Hinkle 8
liunterston A
Hunterrton 8
Oldbur y
S i z r u ~ llTraistynydd
wylta
1 0 1 ~ 1 annual electrlc e n e r g ygerirrated (Gw a)
4.19
Hormdlized dctlvlty[GBq (GW a)-]]
821.0

- - -

-

-

-

Actlvlty ( G B q )Reactor
Ru-I06
Uerk~l~y
Brddwr I 1
Chapel( ro:sA
Dungene::Hiniley A
Iilnkley U
Hunterqton A
Hunterston 8
Oldbury;I~eieil
1rdd:t ynydd
ciyltaTotal drinual elzctrlc energy
generated (GW a) 4.19

Lb-125

Cs-134

Cs-137Ce-144

Pm-I4 1-

T a b l e

32

N o r m a l l z e d l o c a l and r e q l o n a l
c o l l e c t ~ v ~ ~ f f e c t dosel v e e q u l v a l e n t comnltmentf r o m n o b l e Oases r e l e a s e d from t h e model PWR s l t ehormallzed c o l l e c t l v ee f f e c t l v e dosee q u l v a l e n t comnltment

Radlonucllde[lo-'manSv (GW a ) - I ]Ar -4 1
Kr-85m
Kr-85
Kr-81
Kr-88
Xe-13lmXe- 133m
Xe-133xe-135m
Xe-135
Xe-138Total

251

T a b l e

33

Normal l z e d l o c a l and r e q l o n a l
c o t l e c t l v e e f f e c t l v e dose e q u l v a l e n t comnltmentf r o m n o b l e oases r e l e a s e d f r o m t h e model 0WR s l t e

N o r m a l ~ z e dc o l l e c t l v eeffective dosee q u l v a l e n t comnltment[man sv ( G H a ) - ' ]Radlonuc 1 l d e

0.02
0.014
0.0008
0.024
0.32
0.004
0.042
0.004
0.090
0.052

Ar-41
Kr-85m
Kr-85
Kr-87
Kr-88
Xe-13lmxe-133
Xe-135111
xe-135
Xe-138

-

Total

0.56

T a b l e

34

N o r m a l l z e d l o c a l and r e q i o n a l
c o l l e c t l v e e f f e c t l v e dose e q u l v a l e n t c o m l t m e n t strom t r l t l u m released t o the hydrosphere

Reactortype

Normal (zedactlvlty
releaseeffective
dose e q u i v a l e n tcomnl tment1( 1 8 s (GU a ) ' 1 [man sv ( G W a ) - ' ]

GCR

27
2.1
290
97

LWGR

1.7

PWR

BUR
HWR

Collective

Welghted average 0.033 man Sv (GW a ) - 'T a b l e

35

Normallzed local and reqlonalcollectlve effectlve dose equlvalent comnltmentsfrom carbon-14 released to the atmosphere

Reactor
type

PWU
8WR
HWR
GCR
LWGR

Norma I l zed
actlvlty
releaseCollective

effectlve
dose equivalent
comnl tment[ G ~ Q(GW ?)-I] [man Sv (GW a)-']
345
330
6336
1100
1300

Welghted average 1.6 man Sv (GW a)-'T a b l e

3bNormall zed local dnd reqlorial
collectlve effectlve dose eaulvalent comnltmentsfrom releases of lodlne

ReactortY P ~

Normal l zed
lodlne-131
actlvlty

Collectlve effectlve dose e~ulvalent comnltment[man sv (GW a)-']Iodlne-131

Other lodlne

I G B Q( w a)-')
FUR
8WR
GCR
HWU
LWGR

1.7
9.3
1.4
0.73
80

Total

isotopes
0.54
3.61 lo-3
0.54
0.089
30

0.04 l o - 3
I.?
10‘~
0.18
0.013
10

Average welghted collectlve dore:

3.3

T a b l e

37

0.58
4.8
0.72
0.12
40

man Sv (GW a)-'Normallzed local and req:onal
collectlve effectlve dose equlvalent comitmentsfrom particulates released to atmosphere from reactors

Pathway

Collectlve
dose per unlt
actlvlty(
man sv~8q-I)

Olrect cloud
Inhalation
lngestlon
Ground-deposlts
Resuspenslon

0.001
0.12
2.0
3.3
0.004

Collectlve effectlve dose equlvalent comltment
man sv ( G W a)-']

PWR

BUR

HWR

GCR

LWGR

0.005
0.48
9
14.8
0.018

0.043
5.7
87
143
0.17

0.00004
0.005
0.08
0.13
0.00020.0015
0.15
2.4
4.2
0.005

0.016
2.0
33.6
56
0.08

Total

Average welghted collectlve dose:15

man Sv ( W a)-'T a b l e

38

Normallzed collective effectlve dose epulvalent comnltments
for radlonuclldes In Ilgu!d effluents from reactors
dlscharoed to the model rlver

Collectlve effectlve dose equlvalent comltment
man sv ( G W a)-']Reactor Radlotype nucllde

PWR

1-131
Co-58
Co-60
Sr-90
Cs-134
Cs-137

Norma 1 1 zed
released
actlvlty

30
33
20
0.22
10
18

Total: 16.3
BWR

1-131
Co-58
Co-60
Sr -90CS-134
Cs-137

Total: 6.6

3
3
30
0.2

a
15

P a t h w a y
Orlnkln~water

Flsh

External

9.3
0.11
0.93
0.23
1.33
1.67

0.67
0.03
0.07
0.01
0.83
1.03

0.002
0.01

13.63

2.64

0.02

1 .O
0.02
1.4
0.2
1.0
1.2

0.01
0.003
0.1
0.01
0.6
0.8

0.002
0.01

5

I .6

0.02

0.00030.001

0.01

T a b l e

39Normallred collectlve effectlve dose euulvalent comnltments
for radlonuclldes In lluuld effluents from reactors
discharged to coastal waters

Collectlve effectlve dose eaulvalent comnltment[lo-4 man Sv ( W a)-]]Reactor Radlotype nuclide

Norm 1 1 zed
released
actlvlty[GBU (GW a)-]]

PWR

Co-58
Co-60
Ag-llOm
CS - 134
Cs-137

Zn-65
CO-58
Co-60
Ag-1 lOm
Cs-134
Cs-137

10
3
30
0.2
8
15

Total: 375
GCR

Co-60
Sr-90
Ru-106
Sb-125
Cs-134
Cs-131
Ce-144

Total: 1875

~ t s hCrustacea

16.8

1-64

17.30

15
0.02
5
0.3
4
8

1

0.01
1.8
0.1
0.08
0.15

320
0.12
13
5
1
1

32

3

340

3
10
5
7
150
1300
1.5

1
3
1
0.2
4
27
7.5

10
20
40
0.4
30
180
75

1476

44

355

MO~~USCS

33
20
0.2
10
18

Total: 35.1
BUR

P a t h w a y

20
360
50
40
500
2500
150

T a b l e

40

Occupational exposures a t LWRs
[AS. 85. 813. 835. E8. E9, 13. 14, 15. 16, 17, K1. M3. N2. N3. P2.
P3. P7, 51. 52, S 3 . 5 5 , 514, S16, T12. V l ]Country.
reactor typeand y e a r

Numberofunlts

Annualcollective
effectlve
dose
equlvalent

Number o fworkers
monitored

AnnualEnergyaveragegenerated e f f e c t t v e
lnthe
dose
year
equlvalent(man Sv)(GW a )

2.1
1.4
3.3
2.3
3.2
2.2

0.48
1.57
1.80
1.90

F i n l a n d ( 2 PWRs. 2 BURS)
1980
4
1981
4
1982
4
4
1983
1984
4
1985
4

1900
2600
I800
1500

(mSv)

1.7
0.9
l .8
1.4

Normallzed
collective
effectlve
dose
equlvalent[man Sv(GU a ) - ] ]

0.9
1.8
1.2

France (PWR)
1980
1981
1982
1983
1984
1985
Japan ( r a t i o PWRs : BWRs
1980
24
1981
24
1982
27
1983
27
1984
30
1985
31

-

1 t o 1)
134
136
116
112
117
113

72000
88000
82000
87000
1020001 18000

9.1
9.3
11.6
11.9
13.6
14.5

1.9
1.6
1.4
1.3
1.2
1 .O

15
14
10
9.4
8.6
7.8

2.7
6.4
8.9
7.8
5.2

790
1350
1560
1400
1040

0.45
0.39
0.42
0.39

3.5
4.7
5.7
5.6
5.0

5.9
16
21
20

Sweden ( 3 PWRs; 9 BWRs t n 1985)
1981
9
13
1982
10
9.6
1983
10
14.7
I984
10
11.6
1985
12
11.0

4200
3800
4800
4600
5300

4.3
4.3
4.6
5. R
6.5

3.2
2.5
3.1
2.5
2.1

3.0
2.2
3.2
2.0
1.7

S w i t z e r l a n d ( 2 PWRs; 2 BURS)
1980
4
8.9
1981
4
9.1

1900
2050

I. 6

4.6
4.4

5.5

6.7
6.6
6.2
6.6
5.6

18
17
16
17
I5

N e t h e r l a n d s ( 1 PWR; 1 BUR)
1980
2
1981
2
1982
2
1983
2
1984
1 (PWR)

USSR (PWR)
1980s
U n i t e d S t a t e s ( R a t i o PURs : BURS
1980
68
538
1981
70
54 1
1982
74
522
1983
75
565
1984
78
552

- 2 t o 1)

80300
82200
84400
85600
98100

29
31
33
33
37

T a b l e

41

Collective occupatlonal exposures at PURs and BURSIn the Unlted States and Japan[835. TI?]Country
andreactortype

Year

Annualcollectlve
effectlve
dose
equlvalent

Number ofreactors(man Sv)
Unlted States( fJWRJapan(WR)

AnnualEnergycollectlve
generated effectlve
ln thedose
year
equlvalent
per reactor(Wa)(manSv)N o r m 1 1 zedcollectlve
effectlve
dose
equlvalent[man Sv( W a)-11

1980
1981
1982
1983
1984

1980
1981
1982
1983
1984
1985

25.1
28.4
29.2
32.4
34.9
36.4

13
13
13
13
15
15

4.0
4.3
5.4
5.8
5.9
6.7

1.9
2.2
2.2
2.5
2.3
2.4

6.3
6.6
5.4
5.6
5.9
5.5

T a b l e[Ad.

Country.
reactor typeand year42

Occupatlonal exposures at HWRs and GCRs
815. 018. HI. HZ. H3. H4. 13. 14. 15. 16.1
17. K1. N3. P5. P8. P14. 112. W2]

Numberofunlts

Annua 1collectlve Number ofeffective
workers
dose
monltored
equivalent(man Sv)

AnnualEnergyaveragegenerated effective
In thedose
year
equlvalent(GW a)(mSv)

Normallzed
collectlve
effectivedose
equlvalent[man Sv( W a)-l]Argentlna (HWR)
1983
1984
1985
1986

2
2
2
2

5.0
3.6
6.4

980
1000
860

0.34
0.44
0.72

5
4
8
13

13
6
9
18

Canada (HWR)
1980

12

19.1

6780

4.34

2.8

4.4

Japan (GCR)
1980
1981
1982
1983
1984
1985

1 .
1
1
1
1
1

1.1
0.8
0.8
0.8
1 .O
1.4

USSR (LYGR)
1980

2

2.3

-

0.41

United Kingdom (GCR)
22
1980
1981
27
1982
22
1983
25
1984
25

23.0
22.6
19.9
18.7
19.1

11100
17400
17700
19800
20300

3.33
3.48
4.28
4.57
5.00

2.1
1.3
1.1
0.9
0.9

7.1
6.5
4.6
4.1
3.8

Unlted States (GCR)
1980
1
1
1981
1
1982
1
1983

0.03
0.01
0.04
0.01

58
31
22
48

0.08
0.09
0.07
0.09

0.05
0.03
0.02
0.02

0.4
0.1
0.1
0.1

N o r ~ l l l e d occupational exposures at reactors
tor tne aulnquennium 1980-1984

Reactor
type

LWR

HWR
GCR
HTGR
LWGR

Collectlve eftectlve
dose equlvalent per
unlt energy generated[man sv (GW a)-']T a b l e

44

Estimated tvplcal volumes and actlvltles
of condltloned solld wastes from LWRs In t h e 1970s
IESI

Reactor type

[H7]

[ME]

BWR
PWR

3
Annual volume ( m )
1500
1000
1000
1100
400
600

[ES]

[ti301

1000-2000
200- 5 0 0

Annual activlty (TBq)

a/

a/

BUR
PWR

a/

150
70

&/

110
20-55

a/ Not estimatedT a b l e

45

Estlmated volumes and actlvltv concentrationso f condltloned sol:d wastes from LWRs

ILU

Reactor type

PWR
BUR

Volume per unit energy generated[m3 ( G W a)-']50
100
Actlvlty concentratlon

LLW

200
500

PWR
BUR

T a b l e

46

Estlmated typlcal radlonucllde composltlon of condltloned
solld wastes from LWRs after about ten years o f intertm storage

Actlvtty
percentage
Radlonucllde&/ These radionucl ides have daughters that wlll be In equll lbrlum.

Only the percentage of the parent 1s reported.
b/ The radionucllde wlll bulld up as plutonium-241 decays.
-

T a b l e

47

E s t l m d t e d t y p l c a l volumes and a c t t v l t y concentrationso f c o n d l t l o n e d s o l l d wastes From HWRs and GCRs
(831. f 3 ]- - - --

ILW

Reactor t y p e

LLW

Volume p e r u n l t energy g e n e r a t e d[m3 ( G Wa ) - ' ]
HWR
GCR

50
20

250
1000

100
1000

1
10

A c t l v l t y concentration(GBq i 3 )HWR a/
GCRa/

Excludlng t r l t t u m .

T a b l e

48

Some m a j o r c l o s e d and operating s h a l l o u b u r l a l s i t e s[Ca. C9. HIS. N9)

C o u n t r y and s l t e

Unlted States
B e a t t y , Nevada
Maxey F l a t s . Kentucky
West V a l l e y . New YorkR l c h l a n d . Washlngton
Sheffield. I l l l n o l s
B a r n u e l l . South C a r o l l n a
Unt t e d Klngdom
D r l g g . Cumbrla
FranceC e n t r e de l a MancheStartedoperat ton

Status

1962
1962
1963
1965
1967
1971

open
closedclosedopen
closed
open

1971

open

1969

open

T a b l e

Approximate
total
cacac?ty49

Sumnarv o f r e c o r d e d disposals o f packaqed s o l l d w a s t e
t n t o t h e n o r t h - e a s t Atlantic. 1949-1982
IN61

Gross u e l g h tAlpha a c t t v l t y
Beta/gamna a c t t v t t y
T r l t i u m a_/(t)( TBq)(TBq)(T8q)

142 000
680
38 000I 5 000

a/ Recorded s e p a r a t e l y t o r 1975-1982;
I n c l u d e d i n beta/gamna a c t l v l t y f o r
e a r l l e r years.

T a b l e

50Arsumptlons f o r assessment o f l n d l v l d u a l and c o l l e c t l v e doses
f o r r e l e a s e s from l a n d r e p o s l t o r l e s[ ~4 I

foodstuffo r pathway

Maxlmum
annualexposure

Uaxlmum
annualconsumption

(h)

(kg)

food y l e l d

Terrestrial( k g km-2)

Orlnklng water

0.6 a_/

Freshwater f l s h

20

Beef
Cow 1 lv e r
M l lkMuttonSheep 1 l v e r
Green v e g e t a b l e s
Gra I nRoot v e g e t a b l e s
Uarlne f l s h
Crustacea
molluscs
SeaweedBeach occupancy b/
f l s h l n g gear exposure
farm ploughlng
Other l n h a l a t l o n

a/

b/

1000
880
300
8760

m3.
I n h a l a t l o n r a t e : 1 d h-1.

T a b l e

51

C o l l e c t l v e dose e g u l v a l e n t r a t e p e r u n l t a c t l v l t y
a t c l o s u r e I n an e n q l n e e r e d f a c l l l t r
as a f u n c t l o n o f t i m e f r o m 100 t o 2.000 y e a r s

Tlme a f t e r
clorureC o l l e c t l v e dose
equlvalent r a t e
per u n l t a c t l v l t[man Sv (TBq a ) - ]I

Uarlne(kg)

l a b l e

52

Collectlve dose equivalent rate per unlt actlvlty
at closure In an engineered taclltty
a s a functlon of tlme from 10.000 to 250.000 years

Tlme after
closure(years)a/Collectlve dose
equlvalent rate
per unlt actlvlt[man Sv (TBq a)- ]r

2 4 1 ~ u&/

2 4 1 ~ m&/Most of the dose Is dellveredby 2 3 7 ~ pdaughter.

T a b l e

53

Collectlve dose eaulvalent rate ver unlt acttvltf
at closure In an enulneered faclllty
as a functlon o f tlme from 100.000 to 2.000.000 years

Time after
closure(years)Collectlve dose
equivalent rate
per unlt actlvlt[man Sv (T6q a)- ]1

2 3 5 ~3/

2 3 9 ~ u g/-a/ Host of the dose Is From thedaughters 2 3 l ~ aand 2 2 7 ~ ~ .

T a b l e-

54

Collective effectlve dose equlvalent rate and comnltrnent
for disposal o f 1 TBq o f each r a d l o n u c l ~ d e bV shallow burlal (LLUL

Maxlmurn
Radlonucl lde

Collective
effectivedose
equlvalent
comnl tment(man Sv)

collective
effectlve
dose
equlvalentrate(man Sv a-l)

Tline at whlch the speclfled percentage ofthe maxlmum collective effectlve dose
equlvalent rate 1s reached
1%

10%

100%

10%

1X

a/ Global clrculatlon.
b/-Collectlve effectlve dose equlvalent rate less than 10-l2 man Sv a-l,or collectlve dose comnltment less than 10-lO man Sv.

T a b l e

55

Collectlve effectlve dose eaulvalent rate and comnltment
for dlsposal o f 1 TBq of each radlonucllde by enqlneered trench ( I L W L

RadlonuclldeCollecttve
effectivedose
equivalent
comnl tment(man Sv)

Maximum
coilectlve
effectivedose
equlvalentrate(man Sv a-l)

Tlme at which the speclfled percentage ot
the rnaxlmum collective effective d o ~ eequtvalent rate 1 s reached
1%

1 OX

100%

10%

1%

g/ Global clrculatlon.
b/ Collectlve effectlve dose equlvalent rate less than 10-12 man Sv a-l.
or collecttve effective dose equlvalent comnltment less than 10-10 man Sv.

T a b l e

56

Normallzed collectlve etfecttve dose equlvalent comnltment
trom d l s ~ o s a l of lntermedlate-level waste ln a tyD\cal englneered trench

Radlonuclldea/-Normallred
actlvlty
concentratton(10q (GW a)-]]Normallzed collectlve
effective dose
equlvalent comnltment[man Sv ( W a)-)]a/ Only

those radlonuclldes are shown for uhlch thecollectlve effectlve dose equlvalent comnltment perunit actlvtty exceeds 10-10 man Sv (lEiq)-l.

T a b l e

51

Attltudes towards reprocesslnq
In countries wlth nuclear power statlons above 3 0 W

Optlons for dlsposal of spent fuelReprocessing and burlalof solldtfled hlgh-level waste

Geologic burlalof unreprocessed
spent fuelbelng conslderedby:Countrles
reprocesslnq

Canadaf lnland
SwedenSultzerland
Unlted States

-a/
bl
g/

FranceGermany. Fed-Rep. c/
Indla
Japan
USSR
Unlted Klngdom

Countrles
contracted for
reprocess lng
rervlces
8elglum
Germany, Fed.Rep.
Italy
Japan
netherlands
SwedenSwltzerland

Uncomnl tted

Argent lnaa_/Bulgariab_/
Chlna ( Talwan Prov.)
Czechoslovakia b/
Flnlandb/
German ~ e m . ~ e p . L /Hungaryb
Paklstana_/
Republlc of Korea
Spa lnYugoslavia

Pllot-scale reprocesslnq plants reported under construction.
Spent fuel o f Soviet orlgtn ultimately t o be returned t o t h e USSR.
Plant reported to be under malntenance.

T a b l e

58

N a t l o n a l p r o g r a m n e s f o r reprocesslnq spent fuelf r o m c o m n e r c l a l nuclear p o w e r q e n e r a t l o nCountry

Reprocessing c a p a c lty a n d p l a n s

Belglum

The E u r o c h e m l c plant at no1 ( a n n u a l capaclty: 6 0 t o n n e s of uranlum) w a s
operated b e t w e e n 1 9 6 6 and 1974: It w a s t h e n c l o s e d o n t h e g r o u n d s that lt w a s
uneconomic. A d e c l r i o n o n recormencement of o p e r a t l o n s a n d t h e posslblllty o flncreaslng c a p a c l t y ls expected t o b e made. B e l g l u m has c o n t r a c t e d u l t h f r a n c e
for r e p r o c e s s l n g of about 54 tonnes o f u r a n l u m ln fuel.

CanadaResearch s u p p o r t l n g vltrtflcatlon d e v e l o p m e n t .

FranceThe UP1 plant at M a r c o u l e (annual c a p a c l t y : 1,200 t o n n e s o f uranlum) has b e e noperated s l n c e 1 9 5 8 and t h e U P 2 plant at C a p d e la H a g u e ( a n n u a l capaclty: 9 0 0
tonnes of u r a n l u m ) s l n c e 1 9 6 7 for natural u r a n l u m g a s - g r a p h l t e fuel. F o l l o u l n gadaptatlon, t h e U P 2 plant began In 1 9 7 6 r e p r o c e s s l n g L W R fuel at a nomlnal
annual c a p a c l t y of 1 0 0 t o n n e s of uranlum. The c a p a c l t y of t h e U P 2 plant 1sbelng p r o g r e s s l v e l y expanded. and a n e w plant, U P 2 - 8 0 0 ( a n n u a l capaclty: 8 0 0
tonnes o f uranlum) Is scheduled t o b e g l n o p e r a t l o n by 1989. A thlrd plant, U P 3 A(annual capaclty: 8 0 0 t o n n e s of uranlum) b e g a n o p e r a t l o n in 1987. A d u p l l c a t eplant. UP38. Is a l s o under conslderatlon. F r a n c e h a s l n t e r n a t l o n a l reprocesslng
contracts l n v o l v l n g a total of a b o u t 6 , 0 0 0 t o n n e s of u r a n i u m LWR fuel.

Finland

Spent fuel o f Sovlet o r l g l n Is t o b e returned t o t h a t c o u n t r y for reprocesslng.

German Oemocratlc
Republ lc

Spent fuel Is t o be returned to t h e USSR f o r r e p r o c e s s l n g -Germany. Federal
Republtc o f

The e x p e r l m e n t a l WAK plant (annual c a p a c l t y : 3 5 t o n n e s o f uranium) at Karlsruhe.
operattonal s l n c e 1971, w a s reported c l o s e d In H a y 1 9 8 0 d u e t o repalrs.
C o n s t r u c t t o n o f a plant at Wackersdorf In B a v a r l a ( a n n u a l capaclty: 3 5 0 t o n n e sof uranlum) has b e e n announced . A t o t a l o f 1,700 t o n n e s o f uranlum o f spent
LWR fuel Is c o n t r a c t e d f o r reprocesslng In France.
T h e plant at T r o m b a y for reprocesrlng natural u r a n l u m metal fuel b e c a m eoperatlonal In 1 9 6 5 ( a n n u a l capaclty: 6 0 t o n n e s of uranlurn). T h e Tarapur plant
for reprocesslnq HWR a n d LWR Fuel b e c a m e o p e r a t i o n a l In 1 9 7 7 (annual capaclty:
100 t o n n e s o f uranlum).
It ls belleved that a thlrd p l a n t f o r spent o x l d e f u e l
f r o m H W R s w i l l t o b e operatlonal In t h e late 1 9 8 0 s at Kalpakkam (annual
capaclty: 1 0 0 t o n n e s o f uranlum).
2 0 t o n n e s of LWR fuel has been c o n t r a c t e d for r e p r o c e s s l n q ln the U n l t e d
Klngdom. T h e Eurex pllot plant a t S a l u g g l a has a n a n n u a l capaclty of a b o u t
10-20 t o n n e s of u r a n l u m in LWR fuel and ls used For r e s e a r c h and development inreprocesslng.
small demonstration reprocesslng plant at Tokai Mura has been reprocesslnq
LWR fuel lntermlttently s l n c e 1 9 7 7 ( a n n u a l capaclty: 2 1 0 t o n n e s of uranium). A
c o m e r c l a l plant For reprocesrlng LWR fuel ( a n n u a l c a p a c l t y : 8 0 0 t o n n e s o furanlum) Is scheduled t o begin o p e r a t l o n In 1990. J a p a n has contracted f o r
r e p r o c e s s i n g 1.600 t o n n e s o f u r a n i u m in LWR fuel In f r a n c e and 1 6 0 t o n n e s o f
u r a n l u m ln t h e U n l t e d Klngdom.
J a p a n has a l s o renewed a contract f o r
r e p r o c e s s l n g 5 0 0 t o n n e s o f uranlum In g a s - g r a p h l t e fuel tn t h e Unlted Klngdom.

Janan

A

Netherlands

1 2 0 t o n n e s o f u r a n l u m h a s been c o n t r a c t e d for r e p r o c e s s i n g In France.

Sweden7 2 7 a n d 1 4 0 t o n n e s o f u r a n l u m In LWR Fuel h a v e been c o n t r a c t e d for reprocesslng
ln F r a n c e and t h e U n l t e d Klngdom, respectlvely. T h e m a j o r l t y of Suedlsh spent
fuel (6.000 t o n n e s ) 1s t o bp stored In S w e d e n for u p t o 2 0 years pendlng ad e c i s l o n o n Its disposal.

Sultzerland

4 7 0 t o n n e s o f u r a n l u m has been c o n t r a c t e d for r e p r o c e s s l n g In France.

USSR

S p e n t fuel r e p r o c e s r l n g ls being c a r r l e d o u t o n a pllot scale; n o d a t a a r e
a v a l l a b l e o n t h e c a p a c l t l e s or locatlons o f S o v l e t r e p r o c e s s i n g plants. It is
understood, however, that spent fuel of S o v l e t o r l g l n p r o d u c e d In countrles o f
t h e Councll for Mutual Economlc Assistance (e.g. Bulgarla. Czechorlovakla.
G e r m a n O e m o c r a t l c Republlc. kungary) 1s s c h e d u l e d for r e t u r n t o the USSR. T h e
U S S R has a l s o n e g o t l a t e d for t h e r e t u r n of s p e n t fuel of Sovlet orlgln f r o mFlnland.

Table 58. conttnued

Country

Reprocesslng capaclty and plans

Unlted Klngdom

The 8204 plant (annual capacity: 1.000 tonnes of uranlum] reprocessed natural
uranlum gas-graphite fuel ln the 1950s and early 1960s. Ihe 8205 plant (annual
capaclty: 2.000 tonnes o f uranium) has reprocessed thls fuel since 1964 and lsscheduled t o undergo renovation.
The 0204 plant, after modlflcatlon.
reprocessed LWR fuel between 1968 and 1973. A thermal oxlde reprocesrlng plant(THORP) (annual capactty: 1.200 tonnes of uranlum) 1s under constructton and
The Unlted Klngdom has
expected t o begln reprocesslng LWR fuel by 1990.
lnternatlonal contracts for reprocesslng about 3.100 tonnes of uranlum In spent
fuel.

Unlted States

The plant at West Valley. New York, (annual capaclty: 300 tonnes of uranlum)
operated Intermittently from 1966 untll 11s closure in 1972.
Oue tomaintenance problems, a novel plant at Morris. Illlnols. (annual capaclty: 3 0 0
tonnes of uranlum) never began operatlon. Reprocesslng of comnerctal nuclear
power Fuel was deferred IndefInltely In 1977. Constructlon was halted on aplant at Barnwell. South Carolina. whlch could requlre an addltlonal 1 8 0 0
mllllon t o complete (annual capactty: 1,500 tonnes of uranlum).
Its operatlon
In the 1990s has been suggested.

T a b l e

59

Radlonuclldes dlscharqed ln a l r b o r n e e f f l u e n t sfrom fuel reprocesrlnq plants. 1980-1985[el, 82. 83. 87. aa, 816. 829, F I . F ~ I

Actlvlty ( T a g )YearElectric
energyP a r t l c u l a t e release
H-3

C-14

Kr-85
Total
alpha

Sellafleld. Unlted Klngdom

Normallzed actlvlty.
1980-1985[ T B q (GW a)-11
C a p d e la Hague. France

Normallzed actlvlty.
1980- 1985
[ T b q ( G w a)-)]Marcoule. France

Normdllzed actlvlty.
1980[ T E q (Gw a ) - 1 1

Total
beta

Table 59. continued

Year

Electrlc
energyIsotoptc cornpositlon of p a r t l c u l a t e a c t t v l t y (total beta)

S e l l a f l e l d , Unlted K l n g d o mNormallzed actlvlty.
1980-1985[ T e a ( G W a)-] ]C a p d e la Hague. F r a n c eNormallzed actlvlty.
1980-1985[TBq (Gw a)-']T a b l e

60

Radlonuclldes dlscharqed ln ltauld effluentsfrom fuel reprocesslna plants. 1980-1985[el. 82. 83. 88. 816. 829. c5. fi. r4. G ~ I---

Actlvlty (TBq)
Year
Total
alpha

Total beta(otherthan H-3)

H-3

Sr-90

Sellafleld. Unlted Klngdom

Normallzed actlvlty. 1980-1985[TBq (GW a)-']8.0
969(t5.2) (t550)

579
(i35)

17.9
(*49)

Cap de la Hague (France)

Normallzed actlvlty, 1980-1985[TBq ( W a)-']]
0.1
256.9

285.6

18.97

Marcoule, France

Normallzed actlvlty. 1980
[TEq ( G w a)-'] 1
0.063

27

294

3.3

Ru-106

Cs-137

T a b l e

61

Isotoplc composltlon of effluents from the Sellafleld
and Cao de la Hague reorocesslng plants. 1980-1985[El, 02, 83, 08, 016, 029, f4]

Actlvlty (TBq)
Radlonucllde

Sellafleld. United Kingdom
1 .O
5-35
< 0.063
Hn-54
1.1
fe-55
0.78
to-60
0.41
Nl-63
0.033
Zn-65
12
Sr-89
Sr -90
352
60
Zr-95
Nb-95
100
TC-99
57
Ru-103
4.6
Ru-106340
Ag-llOm
0.044
Sb-125211-129< 0.14
Cs-134240
Cs-137
3000
Ce-14437
Pm-14786
Eu-1524.7
Eu-1542.0
Eu-155
4.2
Uranium (kg) 4861
0.67
Np-237
6.9
Pu-23820
Pu-239/240
728
Pu-241
8.3
Am- 24 1
0.33
Crn-242
Crn-243/244
0.19Electrlc energy fromfuel reprocessed(GW a)
2.21

0.51

0.8

30.0

0.7

0.8

< 0.095

< 0.1

< 0.2

< 0.2

< 0.2

1.2
0.74
0.53
< 0.034
11
280
130
200
5.8

0.9
1.1
0.5
< 0.04
< 13
319
212
304
3. 6

0.9
1.3
1.5
< 0.03
< 3.0
72
162
31 2
4.3
8.4
348
< 0.1
12
< 0.1
35
434
9
77
< 0.2
< 0.3
< 0.3
2037
0.3
2.6
8.3
345
2.3
0.1
0.1

0.7
2.3
0.4
< 0.06
< 1.8
52
I8
28
1.9
1 .6
81
< 0.1
11
< 0.1
30
325
< 5
5.9
< 0.1
0.1
0.2
2447
0.2
0.8
2.6
81
I .b
< 0.1
< 0.1

2.65

1.70

2.35

6.11

11

17

530
0.14
26
< 0.19
170
2400
17
32
3.5
1.6
2.6
4499
0.41
5.0
I5
600
8.8
0.19
0.11

419
0.1
23
< 0.1
138
2000
22
32
< 0.6
< 1.0
< 1.2
601 1
0.3
4.7
l i,
485
5.4
0.28
0.14

1 .I
1.7
1 .I
< 0.04
< 8.5
204
21 1
385
4.4
19
553
< 0.1
I8
< 0.2
89
1200
24
25
< 0.2
< 0.5
< 0.6
2602
0.3
2.9
8.7
33 1
2.2
0.4
0.1

3.71

3.14

2.96

Cap de la Hague. FranceElectrlc energy fromfuel reprocessed( G W a)
2.65

3.11

4.50

4.50

T a b l e

62

l o t a l beta/qamna a c t l v l t y I n t l s h f r o m t h e l r l s h Sea
and t h e N o r t h Sea. 1983
[ H7 I

Sarnpllng a r e a /landlng polnts

Number o fsampl l n gobservatlons

SampleMean a c t l v l t y
c o n c e n t r a t l o n ( w e t ) 8q k g - 'Total beta

S e l l a f l e l d s h o r e l l n e area
S e l l a f l e l d offshore area

Ravenglass a/
N o r t h e r n N o r t h Sea

M l d - N o r t h Sea

Southern N o r t h Sea

Iceland area

134-C5131-CsCodflounderPlalceDab
Skate
Whiting
Cod
Cod
PlaicePlalceCod
Haddock
Sal t h e
PlaiceCod
Haddock
Herrlng
Whltlng
PlaiceCod
Whiting
Herr l n g
Cod
Haddock
Plalcea/Landlng p o l n t .
n/d = n o t d e t e c t e d .

T a b l e

63

T r a n s u r a n l c a c t l v l t y I n f l s h and shellfish f r o m t h e l r l s h Sea and N o r t h Sea. I 9 8 3[ H7 I

Sampltng a r e a /landlng potnts

SampleS e l l a t l e l d shorellne area

Cod
Crabs
Lobsters
Wlnkles
PlalceCod
Skate
WhelksCockles
Uussels
PlalceCod
Cod
Nephrops
Nephrops
Mussels
Mussels
CocklesCod

S e l l a f i e l d o f f s h o r e area

Ravenglass a/

N o r t h e r n N o r t h Sea
M l d - N o r t h Sea
S o u t h e r n N o r t h Sea
I c e l a n d area-a/Landlng p o l n t .
n/d = Not d e t e c t e d .

228

Number o fsampl l n gobservations

I
3
3
2
1
1
1
1
1
2
1
1
1
1
1
1
1
1
1

Mean a c t l v l t y concentration ( v e t ) . Bq kg-'
238pU

239,740pu

0.0047
0.71
0.54
6.6
0.00850.0057
0.011
1.7
14
9.9
0.011
0.00400.000670.00190.00075
0.00350.00071
0.0023
O.OC0063

0.025
2.9
2.2
27
0.034
0.026
0.044
7.3
54
41
0.043
0.016
0.00380.00920.00330.019
0.00420.013
0.00077

241 pu

80
63
110

1000

241b

2 4 2 ~ ~2 4 3 , 2 4 4 ~ ~

0.020
7.3
14
37
0.038
0.030
0.045
15
75
55
0.048
0.015
0.00510.00740.00250.00450.00130.00540.00032

0.00044
0.097
0.059
0.45
n/d
0.000810.00054
n /d
1.5
0.45
0.00170.00088n/d
0.00077
n/d
n/d
n/d
n/d
n/d

0.0001 1
0.051
0.067
0.17
n/d
n/d
0.00027
n/d
0.47
0.75
0.000140.000260.000020.00008
n/d
n/d
n/d
n/d
n/d

Normallzed lccal and reulonalcollectlve effectlve dose epulvalent c o m l t n ~ e n t rfrom atrnospherlc releases from fuel reprocesslnq
at Sellafleld and Cap d e la Hasue

Normalized actlvlty
1980-1985
Pathway

Inhalation
H-3
Pu-239

Normal lzed
collectlve effectlve dose
eaulvalent comnltment

iTBq ( ~ . l[man sv (GW~)"]'

Sellafleld

La Hague

120
0.00006

3.5
0.000002

Sellafleld

La Hague

Total

Total

Welghted total

T a b l e

1.3 man Sv ( G W a)-1

65

Normallzed local and reglonalcollectlve effectlve dose eaulvalent c o m l t m e n t sfrom aauatlc discharqes from Sellafleld and Cap d e la Hague

Pathway

Norrnallzed collectlve
effectlve dose equlvalent
comnl tment[man sv ( G M a)-']Sellafleld

LaHague

Mollusc andcrustacea
Ru-106
Sr -90
Pu-239-240

4.7
1
0.2

9.9
0.23
0.06

Total

44

11

T a b l e

66

Occupational exposures at reprocesslns plants
ln the Unlted Klngdom and Japan[A5. 812. 823. 828. HE]

Collectlve
effective
dose equlvalenteffective
comnl tment
dose
per unlt energy
equlvalent
generated(mSv)
lmanSv(GWa)-'1

CountryYear

Numberof
workers
monltored

Annua 1collectlve
effectivedose
equivalent( m n Sv)

Japan

1980
1981
1982

740
940
1170

0.60
0.54
0.71

0.8
0.7
0. 6

1 .O
1.1
1.3

Unlted Klngdom

1980
1981
1982
1983
1984
1985

5200
5400
5600
5300
5600
5600

43
39
38
37
36
32

8.2
7.1
6.7
1 .O
6.7
5.6

19.4
10.4
12.1
12.6
13.4
18.9

AnnualaverageOccuoatlonal exposures at Cap d e la Halue and Uarcoule. France. 1973-1985[Cb. H13, ZI]

Year

Number ofworkers
monltored

Annua 1collectlve
effectlve
dose
equlvalent(man Sv)

Annualaverageeffect\vedose
equlvalent(mSv)

Cap d e la Hague

Karcoulea/ No estimate avallable.

Normallzed
collectlve
effectlve
dose
equtvalent[man Sv( G W a)-1]

T a b l e

68

Solld Intermediate-level waste productlon
at operating reprocesslnq plants[Tll. 8321

Actlvlty per unlt ener y generated[TBq ( G W a ) I

-9

Plant

ILW production[m-3( m a)-']

Sellaf leld Magnox
Sellafleld AGR
Marcoule

Alpha

Beta/gam

300
50
1000

T a b l e

69

Fractlon o f t h e fuel throughput of a reprocesslnq plant
estimated t o arlse as low- or lntermedlate-level vaste
lo41

Volumegenerated
i m3(GWa)-'1

Waste category

fractlon of fuel throughput for radlonuclides

Sr/Cs

Ru/Ce

510-4

510-4

Pu

Am

Cm

Fuel residues. hulls
and hardware

20

5 1 ~ - 4 510-4

5 1 ~ - 4

Non-combustlblewaste
Compressible and

15

lo-6

lo-6

combustible wastes
Concentrated llqulds

70

lo-b

lo-b

and particulate
solldlfled wastes

6

T a b l e

70

Normallzed collective effective dose eaulvalent comnltrnent,
truncated t o different times for globally dispersed nuclides,
welghted for the fractlon o f fuel reprocessed

Years.
- -Radlonucllde
10

100

1000

10.0001.000.000

T a b l e

71

Transport needs In the nuclear fuel cyclefor generatlon of 1 GW a electrical e n e r q y
by a LWR uslng plutonlum recycle[Ill

Material

Amount(tonnes)

From

Uranlum ore
Uranlum yellow-cake

60000
170

Mlne
nlll

Fuel elements
Spent Fuel
Recovered f l ss l lematerlals
Hlgh-level waste
Other solld wastes

37
37
25

Fuel fabrlcatlon
Reactors
Reprocesslng

10
1000

All facllltles

Reprocessing

T a b l e

~lll
Reflnery/enrlchment/
fuel fabrlcatlon
Reactors
Storage/reprocesslng
Converslon/enr\chment/
fuel fabrication
Waste reposltory
Dlsposal sltes

72

Normallzed exposure of members of t h e publlc
from radionuclldes In effluents From t h e nuclear fuel cycle.
local and regional populattons

Operat lon andmain radlonucllde

Normalized
collectlve effective dose
equivalent comnltment[ m a n Sv (GW a)-']MiningRadon
Milling
Uranlum, thorlum, radium
Radon
Mine and mlll tailings piles(releases over Five years)
Radon.
Fuel fabrlcatlon
Uran lum
Reactor operation
Atmospheric
Noble gases
Actlvatlon gases
Tritium
Carbon-1 4
Iodlnes
Particulates (Cs. Ru. Co)
Aquatlc
Trltlum
Others (Cs. Ru. Co)
Reprocesslng
AtmosphericTrltlum
Krypton-85
Carbon-14
Caeslum-137
Iodlne-129
Alpha-emitters
MarlneCaeslurn-134.137
Ruthenium-106
Stront lum-90
Alpha-emltters
Transportatlon
Total( rounded)

0.1
4

T a b l e

73

Normallzed exposure! of members of t h e ~ u b l l cfrom solld waste d l s ~ o ~ aandl globally dlspersed
radlonuclldes in effluents from the nuclear fuel cycle

Norma 1 l zed
collectlve
effectlve
dose equlvalent
comnl tment[man sv (GW a)-']SourceMlne and mlll talllngs
4(releases over 10 years)
and fuel fabrlcatlon
Reactor operatlon
LLW dlsposal
ILW dlsposal
Reprocesslng solld waste dlsposal
Globally dlspersed radlonuclldes

150

Total (rounded)

200

T a b l e

0.00005
0.5
0.05
63

74

Normallzed occupational exposures from t h e nuclear fuel cycle

Operatlon

Normal l zed
collectlve
effectlve
dose equlvalen:[man sv (GW a)-']Uranlum minlng and mllllng
fuel fabrlcatlon
Reactor operat \ o n
Reprocessing
Transportatton

0.7
0.5
10
0.25
0.2

Total (rounded)12

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