RADIOACTIVITY RELEASES TO THE ENVIRONMENT
BY NUCLEAR POWER PLANTS-LOCALLY AND FOR THE TOTAL FUEL CYCLE
Robert C. Marlay
MIT Energy Laboratory Report No. MIT-EL 79-014
March 1979
RADIOACTIVITY RELEASES TO THE ENVIRONMENT
BY NUCLEAR POWER PLANTS-LOCALLY AND FOR THE TOTAL FUEL CYCLE
Robert C. Marlay
Dept. of Nuclear Engineering
Massachusetts Institute of
Technology
Cambridge, Massachusetts 02139
March 15,
1979
ABSTRACT
The nuclear fuel cycle is categorized into nine
components. Each component is described with respect
to its operations and radioactive effluent streams.
Engineering estimates of radioactive releases to the
environment are summarized for each component from the
1976 report of the Nuclear Regulatory Commission
entitled, "Final Generic Environmental Statement on
the Use of Recycle Plutonium in Mixed Oxide Fuel and
Light Water Cooled Reactors."
Actual radioactivity release data reported semiannually by licensed facilities in the U.S., plus
actual release data found in the literature for Canadian
and European facilities, are summarized to the extent
that data are available for the years 1970 through
1976. These actual data are compared with the engineerParticular emphasis is
ing estimates of the NRC.
given to a comparison of reactor types, including:
pressurized water reactors, boiling water reactors,
high temperature gas cooled reactors, European gas
cooled reactors and several types of heavy water
Figures showing
cooled and/or moderated reactors.
relative magnitudes of releases for the different reactor
types and trends versus time are drawn.
Estimates of world population exposures for each
fuel cycle component are calculated for the actual
release data from information provided for the estimated
release data. Similarly, total radiological health
effects resulting from the production of one giga-wattyear of power for the various nuclear fuel cycles are
estimated.
Lastly, a comparison is made of these health effects
to the radiological health effects of the fossil fuel
cycles of natural gas, oil and coal. No attempt is
made to characterize the non-radiological health effects
of any fuel cycle.
Table of Contents
page
I.
II.
Introduction
1
"Front End" of the Nuclear Fuel
Cycle
8
A.
B.
C.
D.
E.
III.
V.
VI.
29
A.
31
VIII.
Light Water Reactors-1.
Pressurized Water Reactors (PWR)
2.
Boiling Water Reactors (BWR)
Gas Cooled Reactors (HTGR and GCR)
Heavy Water Reactors (HWR, PHWR, SGHWR)
Liquid Metal Fast Breeder Reactors (LMFBR)
51
61
62
"Back End" of the Nuclear Fuel Cycle
73
A.
B.
C.
73
74
74
Away From Reactor
Waste Disposal
Transportation
(AFR) Spent Fuel Storage
Summary for the Nuclear Fuel Cycle
75
Addendum--Radioactivity Releases
Fuel Fired Power Stations
82
A.
B.
C.
D.
E.
VII.
8
13
19
20
26
Power Reactors
B.
C.
D.
IV.
Uranium Mining
Uranium Milling
UF 6 Conversion
Enrichment
U0 2 Fuel Fabrication
Natural Gas
Oil
Coal
Over What Period of Time?
Coal Impacts
from Fossil
82
83
83
84
88
Conclusion
91
Notes and References
95
-1RADIOACTIVITY RELEASES TO THE ENVIRONMENT
BY NUCLEAR POWER PLANTS--LOCALLY
AND FOR THE TOTAL FUEL CYCLE
Electric power in the United States has been characterized
by remarkably strong and steady growth for more than eight decades, increasing by more than 100 fold since 1900.1
Today,
thermal inputs to the electric utility sector account for thirty
percent of total US energy consumption, and this trend is projected to continue by virtually all econometric models of national energy use.2
By the year 2000, thermal inputs to the electric
utility sector may reach 50 percent of total energy use.
Similar trends have been established abroad in other industrialized nations, and less developed countries are beginning
their own programs for electrification.4
Add to these scenarios
the consequences of the gradual depletion of world oil and gas
reserves,5 and one envisions either an energy future that is
quite different from that of today, or one that is heavily dependent upon electric generation for its energy needs.
The source of this power has yet to be determined, but the
choices are limited.
In the United States, additional increments
to existing generating capacity will likely be from either coal
fired power plants or nuclear reactors.
In Europe and Japan,
where coal is not readily available, nuclear power is currently
the favored choice.
Even oil exporting nations continue to
maintain, despite recent controversies regarding nuclear safety,
waste disposal and proliferation, nuclear power development plans.
In sum, it is quite possible that within a relatively short
period of 50 years fission reactor technology of one form or
-2-
another may be responsible for producing a large fraction of
world power demand.
By the year 2025, it has been estimated
that there may be three- to five-thousand reactors operating
worldwide. 6
Radioactive effluents released to the environment by these
nuclear power plants, and by the supporting facilities in the
fuel cycle, begin to take on a much larger significance when
the annual environmental burden is enlarged by several thousand
reactors over those of today and accumulated in the biosphere
indefinitely.
Consequently, it is important to identify the sources of
these releases, characterize their isotopic compositions, quantify by means of pathway modeling the resulting occupational and
population dose commitments, and estimate their respective contributions to specific world health effects
dences, mortalities and genetic mutations).
(i.e., cancer inciOne can then assess
the relative impacts of various levels of nuclear power development, reflect upon the adequacy of current environmental regulations for the different components of the nuclear fuel cycle
and make appropriate modifications to these standards where
warranted.
The beginning point in these investigations is the identification and measurement of the radioactive effluent releases
to the environment,
both gaseous and liquid, from each of the
nuclear fuel cycle components.
For the purposes of this paper,
the nuclear fuel cycle will be regarded as consisting of the
following discrete components
(Figure 1) and assumes no reprocessing:
-3-
(1) mining
(6) reactor operations
(2) milling
(7) spent fuel storage
(3) UF 6 conversion
(4) enrichment
(8) transportation of radioactive materials
(4) enrichment
(5) U02 fu(9) waste repository and
ultimate disposal
(5) U0 2 fuel fabrication
ultimate disposal
Reprocessing and mixed oxide fuel fabrication are not included within the scope of this paper; however, mention is
made in the Summary section, of the anticipated changes in
effects due to the implementation of the mixed oxide fuel cycle,
relative to the "no recycle" option.
This paper addresses each component in turn, starting
with the mining of uranium ore.
For each component, a brief
description of the process is provided, followed by a characterization of the radionuclides found in the effluent streams and
their sources.
When easily accessible, actual effluent release
data, required by law to be reported semi-annually by all licensed
operators to the Nuclear Regulatory Commission (NRC), has been
used to quantify releases.
Otherwise, engineering estimates of
anticipated releases are shown as found in various Preliminary
Safety Analysis Reports
(PSAR's) or Final Safety Analysis
Reports (FSAR's) for representative operations.
Lacking these,
engineering estimates for "model facilities" are used.
A considerable portion of this effort was devoted to the
statistical comparison of actual release data for several different reactor types.
The reactor types investigated are
characterized as follows:
-4-
iEs
FUEL
SPENT FUEL
LW POWER REACTORS
•iL
REACTOR
STORAGE
1
UO 2 FUEL
FABRICATION
A
s
r
LOW-ENRICHED
UF 6
SPENT FUEL
Ij
-
ENRICHMENT
NATURAL UF 6
FEDERAL WASTE REPOSITORY
CONVERSION
TO UF6
URANIUM MINES & MILLS
Figure
-1
Light Water Reactor Fuel Cycle - No Uranium or Plutonium Recycle
From GES'iO Final Report, Summary Volume, NUREG-0002/1, August 1976, p-. 5-13
-5-
(PWR)
(1) pressurized water reactors
(2) boiling water reactors
(3) gas cooled reactors
(BWR)
(HTGR and GCR)
(4) heavy water reactors
(HWR, PHWR, SGHWR),
(5) liquid metal fast breeder reactors
and
(LMFBR)
Emphasis was also placed on the "front end" of the fuel
cycle;
but due to limitations of time, only a cursory treat-
ment was given to the
"back end" components
(away from reactor
storage of spent fuel and waste repository operations), and to
transportation
of radioactive materials.
Summaries of the re-
lease estimates and the anticipated accumulated dose commitments are provided, nevertheless, for all components, as they
were published in the Final Generic Environmental Statement on
the Use of Recycle Plutonium in Mixed Oxide Fuel in Light Water
Cooled Reactors
(no recycle option)
In reporting effluent
in September 1976.7
releases, dose commitments and health
effects, it is convenient to express this information in terms
of a common unit of productive output.
In the following dis-
cussions, the chosen unit is the "gigawatt-year"
(GWe-yr), repre-
senting an amount of net electrical output equivalent to a
1,000
MWe electric generating station operating continuously at full
power for one year.
Assumptions must be made
for all components
of the fuel cycle except reactors about what level of production
is required to support one GWe-yr of output, and these assumptions are stated where appropriate.
Lastly, in arriving at US and foreign population dose
commitments, complex computer models of radionuclide transport)
-6-
accumulation and decay, and pathways to man must be used.
Schematically, these pathways are represented very simplistically
in Figure 2.
Information on dose commitments stated in this paper are
all quoted from either the GESMO Final Report
(NRC) or from
various publications of the US Environmental Protection Agency
(EPA).
-7-
aqualC
Pathways for exposure of man from atmosphere ana
2
Fig.
releases of radioactive effluents.
Portsmouth Gaseous Diffusion
From: Final Environmental Statement,
1977, pp. 3-65
Plant Expansion, ERDA-1549, Vol. 1, September
-8-
"Front End" of the Nuclear Fuel Cycle
Uranium Mining
The natural uranium requirements for a 1,000 MWe LWR
designed to operate on low enriched fuel
(3.2
percent U-235)
and having burn-up values of 33,000 MWD per tonne are about
550 to 625 short tons
of U 3 0 8
for initial fuel loading and
about 200 short tons for annual refueling.
If one assumes
a 30 year operating lifetime, the average annual fuel requirement is about 215 short tons U 3 0 8 per reactor year. 8
It is
assumed here that a reactor year is equivalent to 0.650 gigawatt years of net electrical output, hence, the annual fuel
requirement for the production of one GWe-yr is about 330 short
tons of U308.
Uranium is mined principally from sandstone deposits containing U 3 08
in concentrations ranging from 800 to 2,000 ppm.9
The average concentration of ore mined in 1975 was about 1600
ppm.
10
As the price of U308 rises, as it has done over the
last decade from about $8/lb to $43/lb, the average concentration of the ore which is economically feasible to mine will
decrease, with a consequent increase in the volume of the
mill tailings.
Ore that is 0.2 percent by weight U 3 0 8 must
be mined in quantities of 165,000 short tons per GWe-yr;
ore
that is 0.1 percent must be mined in quantities of 330,000 tons
per GWe-yr.
In 1974, the domestic uranium mining industry produced
12,000 tons of U308 by two major methods:
(1) open pit mining
-9-
accounting for 58 percent of production from 31 facilities,
and
(2) underground mining accounting for 40 percent from
123 underground mines. 1
Underground mining is normally em-
ployed when the ore lies below 400 feet of overburden.
The radioactive nuclides present in the ore are those of
the U-238 and U-235 decay chains, as shown in Figure 3.
The
principal decay chain, responsible for about 95 percent of
the radioactivity, is that of U-238, having daughter products
U-234, Th-230, Ra-226, Rn-222, Pb-210, and Po-210, before decaying to the stable end product Pb-206.
Assuming that secular equilibrium has been achieved in
all ore deposits
(requiring 100,000 years), ore that is 0.2
percent by weight U 3 0 8 contains 0.00056 grams of Ra-226 per
metric tonne.
is 1 curie
Noting that the activity of 1 gram of radium
(3.7x1010 disintegrationsper second) and that the
daughter products in the decay chain are in secular equilibrium,
hence their activity is also one curie, the ore would contain about
500 pico-curies
meter.
(10
-12
curies = 1 pCi) of
radon gas per cubic
Measured values for ores range typically from 700 to 800
pCi/m3,12 implying a slightly higher grade ore for these measurements.
When the ore is disturbed, as in extraction, the radon gas
diffuses through the exposed mine walls and ore particle surfaces
at rates both estimated and measured to be about 500 pCi/m2-sec, 1 3
thus contaminating both mine environment and global atmosphere.
-10-
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-11-
Hence, uranium mining has several radiological health
effects:
(1) mine workers are exposed to gamma radiation from
the U-238 and U-235 decay chains,
(2) they are also exposed
externally to radon gas in the mine environment and to radon
daughter products deposited on mine dust, (3) they are exposed
internally to these same radionuclides when inhaled, and (4)
both local and global populations are exposed to radon gas
released to the atmosphere and dispersed.
Because radium-226 is relatively long-lived (half-life is
1620 years), the production and escape rate of radon-222 is
essentially constant for hundreds of years.
The effluent re-
lease of radon from a mine, unless it has been well sealed and
stabilized, will continue long after the exhaustion and abandonment of the mine.
Hence, in calculating the healtheffects due
to the production of a single GWe-yr of electricity, one must
make some assumptions regarding the period of consideration.
The GESMO Final Report calculated health effects only over the
period form 1975 through the year 2000 (a 26 year period), during which 35 trillion kwhrs of net electrical output were produced by light water reactors (about 4,000 GWe-yrs).
For the "no recycle" option, the GESMO Final Report estimated that 2.4x107 curies of radon would be released over the
26-year period due to mining operations (excluding the mills
and tailings piles).14
Dividing this by the net production
of electricity over the same period (which is not really
correct because the release figure is an integrated value over
-12-
26 years, accumulating annual releases from past power output),
one gets an accumulated release figure of about 5000 curies
per GWe-yr.
Further, dose commitments due to mining of
uranium ore were estimated to be 910 person-rem per GWe-yr
to the whole body, consisting of 256 occupational, 628 environmental to US residents, and 26 foreign person-rems per GWe-yr.1 5
These are the figures that will be used in a later summary of
the fuel cycle radioactivity releases and health effects.
It is noted that radon-222 has a relatively short halflife of 3.62 days and does not accumulate in large amounts over
time, as do other radionuclides released from the fuel cycle,
such as krypton-85.
Hence, the 26 year period of consideration
will not give a large error in estimating the same "per GWe-yr"
dose commitments due to radon.
There may be some longer term
effects, however, due to uptake by plants and animals of radon
daughters Pb-210 (5 day half-life) and Po-210
(138 day half-
life), the effects of which have not been considered in the GESMO
Final Report.
In summary, radioactivity released in effluents from
uranium mining operations will likely contain small amounts
of all the U-238 and U-235 decay chain radionuclides, but the
release of radon-222 is so overwhelming in comparison to the
combined activity of the other radionuclides, that they are
often ignored in the modeling of dose commitments and health
effects.
Based on radon-222 alone, the activity released to
the environment is figured to be 5,000 Ci/GWe-yr, and the
dose commitment is 910 person-rems/GWe-yr.
-13-
Uranium Milling
In milling operations, uranium is extracted from the
uranium bearing ore and concentrated into a semi-refined product called "yellowcake", which is usually 90 percent U 308.
Two principal methods of extraction are practiced:
(1) sul-
furic acid leach, and (2) sodium carbonate and alkaline leach
processes.
A third method of solution mining and milling is
in an R & D stage of development.1
6
The sulfuric acid leach process is typical of the kinds
of stages by which uranium is extracted:
the ore is crushed
and finely ground; the resulting particles are leached with
sulfuric acid to extract the uranium; the leach liquors are
then purified and concentrated by solvent extraction, after
which the uranium is precipitated as ammonium diuranate by
the addition of ammonia; the yellowcake slurry is washed,
watered, and dried to produce a solid product which is pulverized into a powder and drummed for shipment.l7
This pro-
duct is the feed material to the UF6 conversion plants.
In 1975, 14 conventional mills were operating in the U.S.,
processing about 7 million tons of ore annually containing
12,442 tons U 3 08 (average ore concentrations were 0.18 percent).18
A typical mill in the future is characterized by the GESMO
study as having the capacity to process 3,500 MT of 0.10 percent U 3 0 8 ore per day, operating 300 days per year, recovering
90.5 percent of the U 30 8 for an annual output of 1080 tons of
U 3 0 8 as yellowcake.
In earlier discussions, it was noted that
-14-
330 tons of U 3 0 8 was required for each Gwe-yr of electric
output, given today's LWR plant characteristics; hence, one
such mill can provide enough fuel for approximately three
Gwe-yr of output.
Radionuclides
in the material flows are the same radio-
nuclides mentioned in the section on mining:
Th-230, Ra-226, Rn-220, Po-210 and Pb-210.
found at the Lucky Me
U-238, U-234,
For ore grades
Uranium mine in Freemont County, Wyom-
ing, ranging from 0.19 to 0.26 percent U 3 0 8 , the equilibrium
activity of each of the U-238 chain radionuclides ranged from
460 to 728 pCi per gram of ore.
Activity for the U-235 decay
chain, having only one long-lived radio nuclide, Pa-231 (halflife of 3.43 x 104 years), ranged from 20 to 34 pCiper gram,
amounting to only 5 percent of the total activity.l9
At the mill, the ore is stored on pads to provide a continuous supply for processing.
Storage for ten days results
in the production of 85 percent of the secular equilibrium
activity of Rn-222, which is partially released to the atmosphere by the diffusion process.
Subsequent ore handling,
crushing and pulverizing operations release all the remaining
radon gas.
In addition, dust from these operations becomes
airborne, carrying off-site the U-238 and daughter radionuclides
which have attached themselves to these particles.
Further,
liquid effluent releases also have concentrations of up to
22,000 pCi per liter of Th-230.
After these processes have removed the uranium, 97 percent
-15-
of the original activity in the ore, and virtually all of the
radium-226, remains in the mill tailings, which are ponded
for water evaporation and then piled for ultimate disposal
and stabilization.
The concentration of radium in these piles
for a 0.2 percent ore grade is about 700 pCi per gram.
As
mentioned earlier, because the half-life of radium is 1620
years, concentration of this radionuclide is effectively constant for several centuries, even though its source, U-238,
has been removed by the mill.
Radon-222 reaches its secular
equilibrium concentration within a few weeks and diffuses
through the mill tailings surface at a rate of about 600 pCi/
2
m -sec.
20
Taking as an example a milling operation capable of
supporting one Gwe-yr of electric power, the following numbers
are detailed:
(1) 330 short tons of U 3 0 8 from an ore of grade
0.20 percent extracted at 90 percent efficiency, leaves 185,000
short tons of tailings;
(2) after crushing, grinding and sett-
ling, the average density of these tailings is assumed to be
1.33 short tons per m
(inferred from several data sources),
giving a tailings volume of about 140,000m3;
(3) the tailings
pile is assumed to be shaped to a height of 3 meters giving
a surface area of about 46,000m2;
finally (4) applying the
diffusion rate of 600 pCi/m2-second, one can estimate the
annual radon release rate to the atmosphere to be 870 curies
per Gwe-yr of power per unit.
Integrating this exponentially
decreasing release rate with respect to time from zero to
-16-
infinity until all the radium has decayed, the total consequence of one Gwe-yr of power is the eventual release of
2 million curies of Ra-222.
The U.S. Environmental Protec-
tion Agency has estimated the "100-year" dose commitments,
for example, to be more than 200 times larger than the dose
commitments calculated over the 30 year period of mill operation,21 owing to the continued release of radon and the growing world population.
In this discussion, we take notice of the above considerations, but for the purpose of comparing the relative contributions of the various components of the uranium fuel cycle,
we return to the GESMO Final Report and its 26-year period of
analysis.
The GESMO Final Report estimates the radionuclide releases
to the environment by the uranium milling industry for this
period from 1975 through 2000
(no recycle option).
estimates are displayed in Table 1.
These
The GESMO Final Report
used radiological health assesment models to estimate dose
commitments, and data tables found in Section J of Chapter IV,
Volume 3 of the GESMO Final Report enabled the compiling of
the dose commitments over this period for each Gwe-yr; the
estimates so derived are shown in Table 2.
Recently, the Nuclear Regulatory Commission has been
studying various means for stabilizing the mill tailings piles
and capping them to retain the radon gas within the pile for
a sufficiently long period of time to allow decay.
Typically,
-17-
Table 1: Estimated Radioactive Effuents
Released to the Environment by the U. S.
Uranium Milling Industry for the Period
1975 through 2000
Radionuclide
Release in
Curies
Release in
Curies/Gwe-yr
U-238
260
0.055
U-234
260
0.055
U-235
12
0.003
Th-234
19
0.004
Th-230
24
0.005
Ra-226
13
0.003
Rn-222
4,400,000
930.
Totals
4,400,588
930.
Source:
GESMO Final Report, Chapter IV, Table IV-F-6, Vol. 3,
September 1976.
A factor of 4,732 Gwe-yr of power was assumed
for the 26 year period per Table IV-J(E)-17.
such schemes commence after the six to eight years required
for evaporation
and natural drying of the tailings ponds.
They involve spreading the tailings out over a considerable
land area, applying a clay blanket (impermeable to gas), overlaying the clay by 3.5 feet or more of subsoil, adding topsoil,
grading to natural terrain to limit erosion and revegetating. 2 2
If such stabilization methods were both successful and remain
intact for 10 half-lives of radium (16,200 years), it would
reduce considerably the longer term health effects of the milling operations, ignored by the GESMO Final Report estimates.
-18-
Table 2: Estimated Total Dose Commitment from
the U.S. Uranium Milling Industry for the Period
1975 through 2000 and per Gwe-yr (no recycle option)
Organ
Integrated Dose
Commitment
Person-rem*
Dose Commitment
Person-rem/
Gwe-yr*
Whole body
579,000
248
22,400
52
1,870,000
303
449,000
144
2,130,000
146
1,500
148
Lung
175,000
163
Skin
1,500
165
GI tract
Bone
Liver
Kidney
Thyroid
*Includes occupational, U.S. populations, and foreign exposures
Source:
GESMO Final Report, Chapter IV, Tables IV-F-7 and
IV-J(E)-9 through IV-J(E)-16, Vol. 3, September 1976.
-19-
UF
6
Conversion
The next step in the uranium fuel cycle requires that
the uranium concentrate milled from the ore be converted to a
volatile
(gaseous) compound of uranium hexafluoride
(UF 6 )
in
order to be enriched by the gaseous diffusion processes.
Two industrial processes are used to produce the UF 6 .
The "hydrofluor process" consists of reduction, hydrofluorination and fluorination of the ore concentrates to produce crude
UF 6 ,
followed by fractional distillation to obtain the pure
product.
The
"wet solvent extraction process" uses a wet
chemical solvent extraction step at the beginning of the process to prepare the high purity feed, then proceeds to the
reduction, hydrofluorination and fluorination steps.2
3
The two commercial plants operating in the US are located
at Metropolis, Illinois and Sequoyah, Oklahoma, and have a
combined capacity of converting 10,000 metric tonnes of uranium
metal
(MTU) into UF 6
per year.
24
Recalling that the annual fuel
requirement for 1 GWe-yr is 330 short tons U 3 0 8 ,
containing
255 metric tonnes of uranium metal, the annual conversion
capacity requirement to convert this quantity into UF
be 375 MT.
At 10,000 MTU capacity
utilization factor of 0.65),
6
would
(assuming an average LWR
60 LWR's can be supported by the
existing plants, indicating the need for expansion in the near
future.
Radionuclides handled in the material flows consist, again,
only of the U-238 and U-235 decay chain members.
The distribution
-20-
of radionuclide concentrations have, of course, been altered
by the milling
process which, for example, left behind most
of the radium-226.
Nevertheless, all members of the chains
are present to some degree.
Some decay products appear as
impurities in the mill concentrate; others appear as a result
of the continuing decay processes.
Uranium, present in almost all of the process flows, appears
in liquid effluents and is the predominant source of radioactivity
in the gaseous effluents.
Radium, thorium and decay products
are separated from the uranium in the conversion process and,
hence, also appear in the liquid effluents or in solid wastes.
Gaseous effluents contain natural uranium in several forms:
U 30 8
U0 2 ',UF
4
UF
, 6 , U0 2 F 2 and (NH4 )2 U 2 07.25
Omitting further discussion and simply citing the results
of the GESMO study, the radioactive effluent releases for UF 6
conversion plants were estimated to be those shown in Table 3.
Also, radiological modeling estimates both for the accumulated
dose commitments from these effluents over the 1975-2000 period
and for each GWe-yr were developed as shown in Table 4.
Enrichment
Naturally occurring uranium, containing 0.71 percent of
the fissile isotope U-235, must be slightly enriched to a
concentration of 2 to 4 percent to provide fuel for light water
moderated reactors.
The processes by which this can be
accomplished are several, but the only one currently in use
in the US is gaseous diffusion.
-21-
Table 3:
Estimated Radioactive Effluents Released to the
Environment by UF 6 Conversion Plants for the
Period 1975 through 2000 (no recycle option)
Releases
in Curies
Radionuclide
U-238
U-234
U-235
Th-234
Th-230
Ra-226
Rn-222
Totals
Releases in
Curies/GWe-yr
137.50
137.50
5.95
138.20
40.0
1.51
1.54
0.0344
0.0344
0.0015
0.0346
462.20
0.1016
0.0100
0.0004
0.0004
Source:
GESMO Final Report, Chapter IV, Table IV-F-9.
Volume 3, September 1976.
A factor of 4,551 GWe-yr
of power was assumed for the 26 year period per
Table IV-J(E)-17.
Table 4:
Estimated Dose Commitments from US
UF 6 Conversion
Plants for the Period 1975 through 2000 and per GWE-yr
(no recycle options)
Organ
Whole body
GI tract
Bone
Liver
Kidney
Thyroid
Lung
Skin
Integrated Dose
Commitment
Person-rem*
49,151
8,647
164,291
9,102
25,486
4,551
35,953
13,653
Dose Commitment
Person-rem per
GWe-yr*
10.8
1.9
36.1
2.0
5.6
1.0
7.9
3.0
*Includes occupational, US population and foreign
population exposures.
Source:
GESMO Final Report, Chapter IV, Tables IV-F-11
and IV-J(E)-9 through IV-J(E)-16, Volume 3, September
1976.
-22-
The gaseous diffusion process is based on the principle
that the rate at which a gas (in this case monotomic UF 6)
escapes through a small hole is proportional to the average
speed of the gas molecules.
The average speed of gas mole-
cules at any given temperature is inversely proportional to
the square root of their masses.26
The ratio of the average
velocities of two molecules of UF6 , one consisting of (U-235)F6
and the other of (U-238)F6, is accordingly proportional to the
square root of the ratio of their respective molecular masses-1.0043.
Hence, the maximum theoretical enhancement in the
isotopic content for a single stage is the factor 1.0043.
To enrich natural uranium to 4 percent U-235 requires, given
today's plant efficiencies, 1200 stages. 2 7
All enrichment services in the US are currently provided
by three gaseous diffusion plants owned by the Federal government
and operated by private contractors.
These facilities are
located at Oak Ridge, Tennessee, at Paducah, Kentucky and at
Portsmouth, Ohio.
The capacity of these three plants is 17.2
million kilograms of "separative work units"
(SWU) per year.
In 1972, these plants produced 10.5 million SWU, only a portion
of which was dedicated to commercial nuclear power enrichment
services.28
One GWe-yr of electrical output requires approxi-
amtely 130,000 SWU.29
Uranium, including the isotopes U-234,
U-235, U-236, and U-238, constitutes the major portion by weight
of the radionuclides found in diffusion plant effluents.
The
fission product Tc-99 (technetium), present by way of spontaneous
-23-
fission of U-235, contributes the most radioactivity.
The
uranium daughter Th-234 is the only other radio-nuclide found
in significant concentrations in plant effluents.
Although not
usually detected in effluent samples because of the very short
half-life (1.17 minutes), Pa-234m must be present if Th-234
is present, and is allocated in radiological health effects
studies a portion of the gross beta activity.
is attributed
All alpha decay
to uranium; the beta decay is attributed to
Th-234, Tc-99 and Pa-234m.3 0
Actual release
data for the three US plants is reported
annually and can be found among the publications of the US
Department of Energy.
These data are displayed for the calendar
year 1976 in Table 5.
If one assumes that annual production
from the three plants amounted to a level similar to that for
1972, roughly 10.5 million SWU, and that only about one-third
of this separative work was dedicated to commercial nuclear
power productions, and if one ignores the radionuclides not
directly associated with the gaseous diffusion process (include
only U, Th, Pa and Tc isotopes), one can arrive at an estimate
of about 0.8 Curies released per GWe-yr, most of which is due
to Tc-99.
The GESMO Final Report estimates radioactive release rates
as those shown on Table 6.
These release rates are two orders
of magnitude less than the above estimate.
Either figure is
very low, however, compared to other components of the fuel
cycle.
Table 7 estimates the radiological dose commitments
-24-
Table 5:
Reported Radioactive Effluents Released to the Environment in 1976 by US Gaseous Diffusion Plants
Radionuclide
Uranium
Th-234
Pa-234m
Cs-137
Xe-133
I-131
Pu-106
Tc-99
Su-90
Kr-85
Co-60
H-3
1976
Oak Ridge
Curies
1.55
-.
Totals for U,
Th, Pa & Tc only
1976
Paducah
Curies
1976
Portsmouth
Curies
1.08
1.08
1976
Totals
Curies
3.71
..
--
--
--
0.20
56,000.00
1.33
.20
30.80
4.50
11,500.00
0.90
8,420.00
----16.20
-----
----19.10
-----
32.35
17.28
20.18
--
0.20
56,000.00
1.33
0.20
66.10
4.50
11,500.00
0.90
8,420.00
69.81
Source: Environmental Monitoring Reports for Calendar Year 1976 for each
of the three plants found compiled in ERDA Report No. ERDA-77-104/2.
Table 6:
Estimated Radioactive Effluents Released to the Environment by Gaseous Diffusion Enriched Plants (two sources
of date)
Radionuclide
"no recycle"
GESMO (1)
Release Est.
Ci/26-yr
-
U-232
U-233
U-234
U-235
U-236
U-238
Transneptunium
Np-237
Tc-99
Pu-109
Zr-95 & Nb-95
Cs-137
Ce-144
Other Fission
Products
Totals
3.5
"no recycle"
GESMO (2)
Release Est.
pC i/GWe-yr
-
0.1
760.
20.
0.5
110.
_
-
.0275
.00015
32.5
1.25
.92
5.30
.00000033
.00017
450.
6.0
1.25
.092
.092
.092
_
4.1
890.
"recycle"
1976-EPA (3)
Release Est.
pCi/GWe-yr
Ci
497.524
Source:
(1) GESMO Final Report, Chap. IV, Table IV-F-16, Vol. 3, September 1976; for the
26-year period from 1975 through 2000; (2) a factor of 4.603 GWe-yr of power output for
the 26-year period was applied per Table IV-J(E)-17; and (3) US EPA, Environmental Analysis
of the Uranium Fuel Cycle--Part IV, Supplementary Analysis. Table 3.0-2, pp. 57, July 1976.
-25-
Table 7:
Estimated Dose Commitments from US Gaseous
Diffusion Plants for the Period 1975 through
2000 and per GWe-yr (no recycle option)
Organ
Whole Body
GI Tract
Bone
Liver
Kidney
Thyroid
Lung
Skin
Integrated Dose
Commitment
Person-rem*
3,371
3,820
5,541
5,646
5,751
5,907
6,810
7,448
Dose Commitment
Person-rem per
GWe-yr*
0.7323
0.8299
1.2037
1.2265
1.2495
1.2833
1.4794
1.6181
*Includes occupational, US population and foreign
population exposures.
Source:
GESMO Final Report, Chapter IV.
Tables IV-F-14
and IV-J(E)-9 through IV-J(E)-16, Volume 3, September 1976.
It is noted that there is a considerable discrepancy between
these figures and Tabel S(4)-1 of the same report, the
latter showing whole body dose commitments to be 3,500
person-rems.
-26-
to occupational workers, the US population and foreign populations due to the operation of the US plants for the period
1975 through 2000, as well as for a GWe-yr of production.
Again, these values are extremely small compared to other fuel
cycle components.
It is noted that these estimates are for the "no-recycle"
option.
If uranium and plutonium recycle were allowed, these
estimates are projected by the GESMO Final Report to increase
100 fold due to the presence of more U and Pu radioactive isotopes in the feed material streams. 31
For comparison purposes,
estimates of radioactivity releases under this option per GWeyr, developed by EPA, are also shown on Table 6.
UO 2 Fabrication
The uranium hexafluoride (UF6) from the gaseous diffusion
plants, after having been enriched to 2 to 4 percent U-235, is
shipped in large 2300 kg containers to LWR fuel fabrication
plants.
The UO 2 and fuel fabrication process involves hydrolyz-
ing the UF 6 touranylfluoride (UO2 F 2), precipitating ammonium
diurante
by the addition of ammonium hydroxide, dewatering
the precipitate by centrifuging or filtering, then drying and
reducing the precipitate to U0 2 powder in a cracked ammonia
atmosphere.
The UO2 powder is then pretreated to obtain the
desired consistency, formed into pellets, sintered to the
required density, ground and polished to finished dimensions,
washed and dried and, finally, loaded into zircaloy tubing
and sealed with a welded cap.
These tubes, or fuel elements,
-27-
are then assembled into arrays to be handled as fuel assemblies.3 2 ' 3 3
Scrap material is collected, dissolved in nitric
acid, purified by solvent extraction, calcined and again
reduced to UO2
for recycling. 3 4
Current capacity of the nine fabrication plants in the US
can process 2,700 metric tonnes of uranium metal
year.
(MTU) per
One GWe-yr of electric power requires about 40 MTU of
fabrication throughput per year.35
Radionuclides present in the effluent streams are strictly
those at the beginning of the U-238 decay chain and other isotopes of uranium.
The radionuclides are the result of leakage,
spillage and breakage in the fabrication processes and small
quantities escape to the environment, most probably UO2F
UO 2 .
2
and
Decay of uranium isotopes U-234, U-235 and U-238 in
natural uranium provides about 0.7 Ci per MTU processed; decay
of these same isotopes in 3.2 percent enriched uranium provides about 1.8 Ci per MTU.3 6
Reported effluent releases by the General Electric facility
in Wilmington, North Carolina and the Westinghouse facility in
Columbia, South Carolina for 1972 are 1.36 and 0.50 curies per
year respectively, as scaled to a model 900 MTU plant.37
Using
the 40 MTU per GWe-yr assumption, these two facilities reported
releases in 1972 of 0.0604 and 0.0222 Ci/GWe-yr respectively.
These figures compare well with those estimated from information provided in the GESMO Final Report, and these estimates are shown in Table 8.
The GESMO Final Report radiological
health effects modeling, estimated dose commitments for this
step in the fuel cycle to be those shown in Table 9.
-28Table 8:
Estimated Radioactive Effluents Released to the
Environment by U0 2 and Fuel Fabrication Facilities
for the Period 1975 through 2000 (no recycle option)
Radionuclide
U-234
U-235
U-236
U-238
Th-234
Totals
Release
Curies
237.0
6.4
0.0
29.5
29.5
302.4
Release
Curies/GWe-yr
0.0454
0.0015
0.0000
0.0068
0.0068
0.0696
Source: GESMO Final Report, Chapter IV, Table IV-F-16,
Volume 3, September 1976. A factor of 4346 GWe-yr of
power was assumed for the 26 year period per Table IV-J(E)-17.
Table 9:
Estimated Dose Commitments from US U0 2 and Fuel
Fabrication Facilities for the Period 1975 through
2000 and per GWe-yr (no recycle option)
Organ
Integrated Dose
Commitment
Person-rem*
Dose Commitment
Person-rem per
GWe-yr*
.
Whole body
GI Tract
Bone
Liver
Kidney
Thyroid
Lung
Skin
54,455
57,237
61,583
51,152
51,935
51,935
1,978,647
11,256
12.53
13.17
14.71
11.77
11.95
11.95
455.28
2.59
*includes occupational, US population and foreign population
exposures
Source: GESMO Final Report, Chapter IV, Tables IV-F-20 and
IV-J(E)-9 through IVJ(E)-16, Volume 3, September 1976
-29-
POWER REACTORS
In the process of generating electricity, nuclear power
reactors accumulate very large amounts of radioactivity by irradiating
fuel, structures and coolant with neutron fluxes on
the order of 10
14
neutrons per cm
2
per second, and by the accumu-
lation of a spectrum of fission product elements, and their
decay chain daughter elements, having atomic numbers ranging
from 70 to 160, most of which are found between Br-84 and
Ce-144.
A typical 1000 MWe power reactor will contain hundreds
of millions of curies of radioactivity;
the isotope
I-131 by
itself represents as much as 70 megacuries of the total.
Among the more important radionuclides
are:
(1) the halo-
gens bromine and iodine, which are volatile at reactor operating
temperatures and diffuse as agas out of the ceramic matrix of
the fuel element, and which are chemically active and mobile
along the pathways to man;
(2) the noble gases krypton and xenon,
both of which may be released to the atmosphere and inhaled, are
water soluble and can be ingested;
(3) the alkali metals cesium
and rubidium, which are also water soluble in the form of dissolved salts;
(4) the alkaline earths barium and strontium
(mem-
bers of the same periodic family as calcium), which are water
soluble and bone seekers;
(5) coolant activation products,
which are generally gases, such as the radionuclides
(tritium), argon, fluorine,
nitrogen and oxygen;
and
of hydrogen
(6)
structural activation products having fairly long half-lives,
including zirconium, manganese, nickle, iron, carbon, chromium,
-30-
cobalt and copper, most of which remain fixed within the structural. materials, but some of which may be released by way of
mechanical wear or erosion to the coolant and, hence, to
liquid effluent streams.
The vast majority of the radioactivity is isolated from
the environment by several constructed barriers:
(1) the
fuel cladding, which remains air tight and sealed unless it fails, (2)
the reactor vessel and closed coolant systems, and (3) the
reactor and auxiliary containment buildings.
Nevertheless,
measurable amounts of radioactivity are released to the environment on a routine basis within permissible limits established
by the Nuclear Regulatory Commission (NRC).
examines the nature and
quantities
This section
of these routine releases
for five reactor types:
(1) pressurized water reactors (PWR)
(2) boiling water reactors (BWR)
(3) high temperature gas cooled reactors (HTGR and GCR)
(4) heavy water reactors (HWR, PHWR and SGHWR), and
(5) liquid metal fast breeder reactors (LMFBR)
In each case an attempt is made to identify and briefly explain
the principal sources of the radioactivity releases and to
quantify them in terms that allow comparisons among reactor
types per GWe-yr of net power produced.
Data for these com-
parisons is compiled from the semi-annual radioactive effluent
release reports filed by law with the NRC, when available, or
from engineering estimates as found in the Preliminary Safety
Analysis Reports (PSAR) for reactors not yet operating.
Some
-31-
data on foreign reactors has been analyzed, although it is very
sketchy by comparison to the US data.
Quantities of radioactivity released are expressed in
curies
(10
(Ci), or micro-curies
Ci = 1
Ci), per GWe-yr of
net power output for four classes of radionuclides:
(1) noble gases,
(2) tritium,
(3) halogens and airborne particulates, and
(4) mixed fission (MF) and various coolant and
structural activation products.
Light Water Reactors
In the analysis of the radioactive effluent releases of
light water moderated reactors, thirty PWR's and 25 BWR's were
investigated.
These reactors, their rated power levels and actual
operating thermal power levels for the years 1975 and 1976 are
shown in Tables 10 and 11.
Net electrical output was assumed
to be 32 percent of thermal power.
For boiling water reactors a direct steam cycle is used
whereby the contaminated coolant in its steam phase passes
directly through the turbine.
Entrained radioactive gases, air
which has leaked into the condenser and hydrogen and oxygen
which result from the radiolytic dissociation of water are removed from the main turbine condenser by means of a steam jet
air ejector (off-gas ejector), which is used to maintain the
condensor vacuum.
These gases are removed at a rate of about
300 cubic feet per minute.
In the absence of any failed fuel
cladding, the principal radioactive nuclide released by this
-32-
Pressurized Water Reactors (PWRs) Included in
Effluent Data Analysis for 1975 and 1976
Table 10:
Rated
Name of Unit
(PWR)
Power
Mwe
1975
Thermal
Ojtput
10
Mwd
1976
Thermal
Ogtput
10
Mwd
0.563
1976
Electrical
Output
Gwe-y
0.441
Arkansas One 1
836
642
Calvert Cliffs 1
850
584
1090
608
895
0.533
0.785
Fort Calhoun
501
280
298
0.245
0.261
Ginna
517
404
291
0.354
0.255
Haddem Neck
600
558
540
0.489
0.473
2087
685
561
0.601
0.492
Kewaunee
563
451
450
0.395
0.395
Maine Yankee
830
612
811
0.537
0.711
Millstone Point 2
860
D.C. Cook
Indian Point 1,2
& 3
503
1975
Electrical
Output
Gwe-y
0.512
0.553
631
2766
1960
1650
1.718
1.447
Palisades
811
371
403
0.325
0.353
Point Beach 1 & 2
994
872
946
0.764
0.829
Prairie Island 1 & 2
1076
938
858
0.822
0.752
H.B. Robinson
772
566
661
0.496
0.580
San Onofre
450
417
323
0.366
0.283
1576
1210
1050
1.061
0.921
870
734
580
0.644
0.508
1520
1160
1120
1.017
0.982
185
168
177
0.147
0.155
2100
1370
1290
1.201
1.131
21854
14590
14038
12.791
12.307
Oconee 1,2
& 3
Surrey 1 & 2
Three Mile Island
Turkey Point 3 & 4
Yankee
(Rowe)
Zion 1 & 2
Totals (30 PWR's)
Source:
Data compiled from "Summary of Radioactivity Released in EfU.S.
fluents from Nuclear Power Plants from 1973 thru 1976,"
List is not
Environmental Protection Agency, December 1977.
all inclusive.
Gigawatt years of electrical output assumes
32 percent conversion efficiency.
-33-
Table 11:
Boiling Water Reactors (BWRs) Included in Effluent
Data Analysis for 1975 and 1976
1975
Name of Unit
(BWR)
Rated
Power
Mwe
Thermal
Output
103 Mwd
1976
Thermal
O0tput
10
Mwd
1975
Electrical
Output
Gwe-y
1976
Electrical
Output
Gwe-y
75
41
35
0.036
0.031
Brown's Ferry 1,2 & 3
3195
365
560
0.320
0.491
Brunswick Units 1 & 2
1580
233
332
0.204
0.291
Cooper Nuclear
Station
778
231
494
0.203
0.433
Dresden 1
200
106
143
0.093
0.125
1600
534
1150
0.468
1.008
Duane Arnold 1
569
309
334
0.271
0.293
J.A. Fitzpatrick 2
821
284
527
0.249
0.462
E.I. Hatch 1
786
407
574
0.357
0.503
Humbolt Bay 3
65
55
28
0.048
0.025
La Crosse
53
38
25
0.033
0.022
Millstone Point 1
652
502
485
0.440
0.425
Monticello
545
370
514
0.324
0.451
Nine Mile Point 1
625
430
545
0.377
0.478
Oyster Creek 1
640
409
492
0.359
0.431
2130
1390
1550
1.219
1.359
Pilgrim 1
664
338
317
0.296
0.278
Vermont Yankee
514
469
425
0.411
0.373
1600
964
476
0.845
0.417
17092
7475
9006
6.553
7.896
Big Rock Point
Dresden 2 & 3
Peach Bottom 2 & 3
Quad Cities 1 & 2
Totals
Source:
(25 BWRs)
Data compiled from "Summary of Radioactivity Released in Effluents from Nuclear Power Plants from 1973 thru 1976,"
U.S.
Environmental Protection Agency, December 1977.
List is not
all inclusive.
Gigawatt years of electrical output assumes
32 percent conversion efficiency.
-34-
means would be nitrogen-13, the product of an (n,)
on oxygen-16.
reaction
In the Brown's Ferry (BWR) Final Safety Analysis
Report (FSAR), the annual discharge of this radionuclide was
estimated to be 8,580 curies per year, or about 10,000 curies
per GWe-yr.3 8
Assuming, instead, that 0.25 percent of the fuel elements
developed cladding leaks of some appreciable size to allow
noble gases under pressure to escape, the Brown's Ferry FSAR
estimated that the annual discharge of the noble gases would
be 2,500,000 curies per year, or about 3,000,000 curies per
GWe-yr, overwhelming the N-13 radionuclide. 3 9
Hence, it clear from this comparison that the combination
of BWR direct cycle and failed fuel cladding can result in a
significant source term for radioactive gas effluents.
Off-gas
hold-up systems, however, can reduce the activity of many of
the krypton and xenon radioisotopes; only 5 percent of the predicted activity is represented by radioisotopes having halflives greater than 10 hours.
In pressurized water reactors
compound is added to the coolant for the purposes of
poison
reactivity control.
This boron is a source of tritium by way
of neutron capture in boron-10.
(1) an
an
(PWR's), a soluble boron
Two reactions are possible:
(n,2a)T reaction producingtritium directly, and
(2)
(n,n'a) reaction producing Lithium-6, which in turn has a
large cross section (950 barns
tritium.
) for an (n,a) reaction producing
As in the BWR, the PWR coolant also contains nitrogen-13
and any volatile fission products that may have leaked from failed
fuel.
-35-
Further, in a PWR the boron introduced for reactivity
control at the beginning of a fuel
removed as the fuel is burned up.
cycle, must be gradually
This is accomplished by con-
tinually bleeding off a small portion of the coolant to a boron
recovery system, wherein entrained gases are evolved by a gas
stripper and
routed to a waste gas hold up and treatment system.
Hence, the gas stripper is another source term for noble gases,
halogens, tritium, and other gases.
In both PWR's and BWR's there are some small but inevitable
leaks of the primary coolant directly to the reactor containment
building environment:
from coolant pumps and valve seals, most
of which is captured and returned to the coolant system, from
many smaller valves, from radiological sample taking and
other routes and activities.
Liquids are collected and processed
in the liquid waste systems; noble gases, volatile halogens and some
particulates will escape to the containment building atmosphere.
The containment building atmosphere is vented several times per
year on a periodic basis to reduce temperature and activity
levels prior to plant maintenance and to reduce containment
vessel pressure if excessive steam leakage exists.
In a BWR, some additional sources of radioactivity releases
are:
the gland seal at the turbine generator shaft, the turbine
building atmosphere venting and a mechanical vacuum pump used
in place of the steam powered off-gas ejectors at the time of
reactor start-up.
significant.
Of these, the turbine gland seal is the most
-36-
Tables 12 through 25 compile statistical average releases
per GWe-yr for 30 PWR's and 25 BWR's for the years 1975 and
1976 for each of the aforementioned categories of radionuclides
(a total of 14 tables are presented).
The source of data for
all of these tabels on light water reactors was a 118 page
report published by the US Environmental Protection Agency,
entitled, "Summary of Radioactivity Released in Effluents from
Nuclear Power Plants from 1973 through 1976." 4 0
From this data
a statistical average was obtained and plotted on Figures 4
through 7, shown in the summary of this section on power reactors.
Occasionally, one or two data points representing statistical
extremes
(low or high) were excluded form the reactor type
average, for the purpose of portraying a release figure "representative" of that type of reactor.
These exclusions are noted,
along with the average including all data points, on each of
the tables where this was done.
-37Table 12:
Name of Unit
(PWR)
Noble Gases Released to the Environment in 1975
by Pressurized Water Reactors (PWRs)
Rated
Power
Mwe
1975
Electrical
Output
Gwe-y
Noble
Gases
Airborne
Curies
Noble
Gases
Dissolved
Curies
Noble
Gases
Total
Curies
Curies of
Noble Gases
Per
Gwe-y
185
0.147
25
.2
25
170
2087
0.601
8566
.8
8567
14255
450
0.366
4.7
600
0.489
1795
480
4904
Haddem Neck
1790
480
Ginna
H.B. Robinson
517
772
0.354
0.496
10500
1170
.1
29661
2359
Palisades
811
0.325
2610
.1
Maine Yankee
830
0.537
4120
10500
1170
2610
4120
Fort Calhoun
Kewaunee
501
563
0.245
429
429
0.395
2450
.2
2450
1751
6203
Three Mile Island
870
0.644
3630
1.1
3631
5638
Arkansas One 1
836
0.563
1040
33.1
1073
1906
Calvert Cliffs 1
D.C. Cook 1
850
0.512
13.1
0.533
7733
3
15104
1090
7720
3
Yankee (Rowe)
Indian Point 1,
2 & 3
San Onofre
.3
982
8031
7672
6
Millstone Point 2
860
Point Beach 1 & 2
994
0.764
45300
.5
45301
59295
Surry 1 & 2
Turkey Point
1 & 2
1576
1.061
8040
28.1
8068
7604
1520
1.017
13400
4.8
13405
13181
Oconee 1,2 & 3
Zion 1 & 2
2766
2100
1.718
1.201
15200
45300
2.9
15203
45300
8849
37719
Prairie Island
1 & 2
1076
0.822
2180
2.8
2183
2656
173953
92.8
174046
13607
Totals
Statistical extremes
(#1, 14, 16)
Total excluding statistical extremes
Source:
12.791
45329
1.441
11.350
-
128717
11341
Data was compiled from "Summary of Radioactivity Released in Effluents from Nuclear Power Plants from 1973 through 1976," U.S.
EPA. December, 1977. This list of PWRs is not all inclusive.
-38Table 13:
Noble Gases Released to the Environment in 1976
by Pressurized Water Reactors (PWRs)
Power
Mwe
Unit Name
(PWRs only)
1976
Equivalent
Power
Gwe-y
-
Yankee (Rowe
185
0.155
San Onofre
450
0.283
Haddam Neck
600
Ginna
Noble
Gases
Aerosols
Curies
27
Noble
Gases
Dissolved
Curies
Total Curies of
Noble Noble Gases
Gases
Per
Curies
Gwe-y
-
27
174
430
492
1519
5913
23188
13
0.473
417
492
517
0.255
5520
393
H.B. Robinson
772
0.580
791
791
1364
Palisades
811
0.353
30
85
Maine Yankee
830
1300
Kewaunee
Three Mile Island
563
0.711
0.395
30
1300
0.508
1600
2760
4051
870
1600
2760
Arkansas One I
902
0.441
5690
90
5780
13107
1090
0.785
975
1
1243
Millstone Point 2
860
0.553
1550
4
976
1554
St. Lucie
802
0.013
1790
1790
137692
Beever Valley
833
0.072
1
2010
2425
Point Beach 1 & 2
994
0.829
2010
19230
20879
Surrey 1 & 2
Turkey Pt. 3 & 4
1576
1520
0.921
0.982
19200
30
19230
20879
15600
1
15601
15887
Oconee 1, 2 & 3
Zion 1 & 2
2766
1.447
44000
1
44001
30408
2100
1.131
142000
142000
125553
Prairie Isl. 1 & 2
1076
0.752
1740
2
1742
2316
Indian Pt. 1,2, & 3 2087
Totals
0.492
13
10613
21571
12.392
10600
260243
552
260795
21045
1.724
143848
-
143848
82576
Total excluding statistical
extremes
10.668
Best plant averages (#1 & 6) 0.508
116395
552
-
116947
57
10962
142000
125553
D. C. Cook
Statistical extremes
(#1, 6, 14, 15, 20)
Worst plant averages
Source:
(#20)
1.131
57
142000
-
1040
1828
5433
2810
112
Data compiled from "Summary of Radioactive Released in Effluents
from Nuclear Power Plants from 1973 through 1976," U.S. EPA.
December, 1977. This list of 31 PWRs is not all inclusive.
-39-
Table 14:
Noble Gases Released to the Environment in 1975
by Boiling Water Reactors (BWRs)
Rated
1975
Electrical
Power
Mwe
Output
Gwe-7
Noble
Gases
Airborne
Curies
200
0.093
520,000
Big Rock Point
75
0.036
50,600
Humbolt Bay 3
65
0.048
La Crosse
53
Name of Unit
(BWR)
Noble
Gases
Dissolved
Curies
Noble
Gases
Total
Curies
Curies of
Noble Gases
per
Gwe-y
520,000
5,591,398
0.00 724
50,600
1,405,556
296,000
0.00 800
296,000
6,166,667
0.033
57,100
0.40 700
57,100
1,730,303
640
0.359
206,000
0.42 600
206,000
573, 816
Nine Mile Point 1 625
0.377
1,300,000
0.10 400 1,300,000
3,448,276
Millstone 1
652
0.440
2,970,000
1.11 000 2,970,001
6,750,002
Monticello
545
0.324
155,000
-
155,000
478,395
Vermont Yankee
514
0.411
3,360
-
3,360
8,175
Pilgrim 1
664
0.296
105,000
0 ..00 071
105,000
354,730
Cooper Station
778
0.203
19,700
0.00 550
19,700
97,044
Duane Arnold 1
569
0.271
1,540
1,540
5,683
Hatch 1
786
0.357
1,550
1,550
4,342
Fitzpatrick 2
821
0.249
4,080
4,080
16,386
Brunswick 2
780
0.204
185
185
907
Dresden 2 & 3
1600
0.468
369,000
-
369,000
788,462
Quad Cities
1 & 2
1600
0.845
110,000
-
110,000
128,805
Brown's Ferry
2134
0.320
25,200
25,200
78,750
Peach Botton
2 & 3
2130
1.219
13,000
13,000
10,664
6.553
6,207,315
2.44652 6,207,317
947,248
Dresden 1
Oyster Creek 1
Totals
Source:
-
0.23 400
0.00 107
0.14300
Data was compiled from "Summary of Radioactivity Released in
Effluents from Nuclear Power Plants from 1973 through 1976,"
This list of BWRs is not all incluDecember, 1977.
U.S. EPA.
sive.
-40-
Noble Gases Released to the Environment in 1976
Table 15:
by Boiling Water Reactors
Rated
Power
Mwe
Name of Unit
(BWR)
1976
Electrical
Output
Gwe-y
Noble
Gases
Airborne
Curies
(BWRs)
Noble
Gases
Dissolved
Curies
Noble
Gases
Total
Curies
Curies of
Noble Gases
per
Gwe-y
200
0.125
460,000
460,000
3,680,000
Big Rock Point
75
0.031
15,200
15,200
490,323
Humbolt Bay 3
65
0.025
93,000
0.00415
93,000
3,720,000
La Crosse
53
0.022
124,000
0.09300
124,000
5,636,364
Oyster Creek
640
0.431
166,000
0.04640
166,000
385,151
Nine Mile Point
625
0.478
176,000
0.00558
176,000
368,201
Millstone Point 1 652
0.425
507,000
0.35600
507,000
1,192,941
Monticello
545
0.451
11,400
11,400
25,277
Vermont Yankee
514
0.373
2,870
0.00628
2,870
7,694
Pilgrim 1
664
0.278
183,000
0.00122
183,000
658,273
Cooper Station
778
0.433
38,100
0.04620
38,100
87,991
Duane Arnold 1
569
0.293
5,260
5,260
17,952
Hatch 1
786
0.503
3,110
3,110
6,183
Fitzpatrick 2
821
0.462
46,200
46,200
100,000
Dresden 2 & 3
1600
1.008
32,400
32,400
32,143
Quad Cities
1,2 & 3
1600
0.417
31,500
31,500
75,540
Brown's Ferry
1,2 & 3
3195
0.491
80,400
0.10900
80,400
163,747
Peach Bottom
2 & 3
2130
1.359
209,000
2.84000
209,003
153,792
1580
0.291
18,500
0.70300
18,501
63,577
7.896
2,202,940
4.31403 2,202,944
278,995
Dresden 1
Brunswick 1
Totals
& 2
.79
0.10400
_
-41-
Table 16:
Tritium Released to the Environment in 1975 by
Pressurized Water Reactors (PWRs)
Rated
Power
Mwe
1975
Electrical
Output
Gwe-y
185
0.147
2
247
249
1694
Indian Point
1 & 2
2087
0.601
25
366
391
651
San Onofre 1
450
0.366
34
4000
4034
11022
Haddam Neck
600
0.489
70
5670
5740
11738
Ginna, RE
517
0.354
6
260
266
751
Robinson, HB
772
0.496
193
624
817
1647
Palisades
811
0.325
42
42
129
Maine Yankee
830
0.537
5
177
182
339
Fort Calhoun
501
0.245
2
111
113
461
Kewaunee
563
0.395
37
277
314
795
Three Mile Island
1
Arkansas One 1
870
0.644
40
463
503
781
836
0.563
1
460
461
819
Calvert Cliffs 1
850
0.512
1
263
264
516
1090
0.533
56
56
105
994
0.764
420
1020
1440
1885
Suny 1 & 2
1576
1.061
32
442
474
447
Turkey Point
1 & 2
1520
1.017
3
793
796
783
Oconee 1 & 2
2766
1.718
1660
3550
5210
3033
Zion 1 & 2
2100
1.201
40
40
33
Prairie Island
1 & 2
1076
0.822
10
763
773
940
12.791
2541
19624
22165
1733
2.056
104
9710
9814
4773
Totals excluding statistical extremes
10.735
2437
9914
12351
115 0
Unit Name
(PWRs)
Yankee
(Rowe)
Cook, DC
Point Beach 1 & 2
Totals
Satistical extremes
(#3, 4, 19)
Tritium
Gases
Curies
Tritium
Liquids
Curies
Total
Tritium
Curies
Curies of
Tritium
per
Gwe-y
-42-
Table 17:
Tritium Released to the Environment in 1976
by Pressurized Water Reactors (PWR's)
Rated
Power
Mwe
1976
Thermal
Output
1000 MwL
1976
Equivalent
Power
GWe-y
185
450
600
517
772
811
830
501
563
870
902
1090
860
802
833
994
1576
1520
2766
2100
177
323
540
291
661
403
811
298
450
580
503
895
631
15
82
946
1050
1120
1650
1290
0.155
0.283
0.473
0.255
0.580
0.353
0.711
0.261
0.395
0.508
0.441
0.785
0.553
0.013
0.072
0.829
0.921
0.982
1.447
1.131
1076
858
0.752
2087
561
Ci of
Tritium
per
GWe-y
Tritium
Gases
Curies
Tritium
Liquids
Curies
Total
Tritium
Curies
2
47
793
24
158
156
3380
4850
242
980
10
368
54
214
189
212
192
277
13
9
695
782
770
2190
1
158
3427
5643
266
1138
10
372
57
215
906
219
192
292
15
3729
1090
1154
775
2692
1
1019
12110
11930
1043
1962
29
523
218
544
1783
497
245
528
1154
51792
1315
1253
789
1860
1
33
1930
1963
2610
0.492
24
332
356
724
12.392
6824
17846
24670
1991
Statistical extremes (#2,3,6,14,15,20)
2.325
4562
8263
12825
5516
Totals excluding statistical extremes
10.067
2262
9583
11845
1177
1
1
1
8230
9070
11997
Unit Name
(PWR's only)
1.
2.
3.
4.
5.
6.
7.
8.
9.
10.
11.
12.
13.
14.
15.
16.
17.
18.
19.
20.
21.
22.
Yankee (Rowe)
San Onofre
Haddam Neck
Ginna
H. B. Robinson
Palisades
Marine Yankee
Fort Calhoun.
Kewaunee
Three Mile Island
Arkansas One 1
D. C. Cook
Milstone 2
St. Lucie
Beaver Balley 1
Point Beach 1 & 2
Surrey 1 & 2
Turkey Point 3& 4
Oconee 1,2& 3
Zion 1 & 2
Prairee Islands
1 & 2
Indian Pt.
1, 2 & 3
TOTALS
Best plant averages (#20)
1.131
Worst plant averages (#2 and 3)
0.756
Source:
4
3
1
717
7
15
2
3720
395
372
5
502
840
Data compiled from "Summary of Radioactivity Released in Effluents from
Nuclear Power Plants from 1973 thru 1976," U.S. EPA, December 1977.
This list of 31 PWR's is not all inclusive.
-43-
Table 18:
Tritium Released to the Environment in 1975
by Boiling Water Reactors (BWR's)
Name-of Unit (BWR)
1.
2.
3.
4.
5.
6.
7.
8.
9.
10.
11.
12.
13.
14.
15.
16.
17.
18.
19.
Dresden 1
Big Rock Point
Humbolt Bay 3
La Crosse
Oyster Creek 1
Nine Mile Point 1
Millstone 1
Marticello
Vermont Yankee
Pilgrim 1
Cooper Station
Duane Arnold 1
Hatch 1
Fitzpatrick 2
Brunswick 2
Dresden 2 & 3
Quad Cities 1 & 2
Brown's Ferry
1&2
Peach Bottom
2&3
Rated
Power
Mwe
1975
Electrical
Output
GWe-y
200
75
65
53
640
625
652
545
514
664
778
569
786
821
780
0.093
0.036
0.048
0.033
0.359
0.377
0.440
0.324
0.411
0.296
0.203
0.271
0.357
0.249
0.204
1600
1600
1975
Tritium
Gases
Curies
1975
Tritium
Liquids
Curies
1975
Tritium
Total
Curies
Curies of
Tritium
per
GWe-y
34
7
2
17
3
92
17
1
6
20
127
18
28
80
35
13
22
144
21
120
97
376
361
458
4364
58
318
220
7
74
43
-
19
-
17
311
251
70
22
2
6
7
92
51
19
8
2
3
5
25
0.468
0.845
221
39
54
39
275
78
588
92
2134
0.320
5
10
15
47
2130
1.219
31
31
25
Totals
Statistical extremes
(#4, 16)
Totals excluding
statistical extremes
18
8
6.553
584
449
1033
158
.501
238
181
419
836
6.052
346
268
614
LZ01
x
x
-44-
Table 19:
Tritium Released to the Environment in 1976
by Boiling Water Reactors (BWR's)
Rated
Power
Mwe
1976
Electrical
Output
GWe-y
200
75
65
53
640
625
652
545
514
664
778
569
786
821
0.125
0.031
0.025
0.022
0.431
0.478
0.425
0.451
0.373
0.278
0.433
0.293
0.503
0.462
61
8
1
13
1
19
29
77
14
37
67
16
1
15
2
7
41
39
2
20
2
47
8
9
4
61
10
8
54
40
21
49
77
16
84
75
16
10
19
488
323
320
2455
93
44
115
171
43
302
173
55
20
41
1600
1600
1.008
0.417
14
297
20
23
34
320
34
767
3195
2130
1580
0.491
1.359
0.291
1
27
22
4
74
6
5
101
28
10
74
96
Totals
7.896
720
308
1028
130
Statistical. extreme #16
0.417
297
23
320
767
Totals excluding stat.
extremes
7.479
423
285
708
Name of Unit (BWR)
1.
2.
3.
4.
5.
6.
7.
8.
9.
10.
11.
12.
13.
14.
Dresden
Big Rock Point
Humbolt Bay 3
La Crosse
Oyster Creek
Nine Mile Point
Millstone Point 1
Monticello
Vermont Yankee
Pilgrim 1
Cooper Station
Duane Arnold
Hatch 1
Fitzpatrick 2
15.
16.
17.
Dresden 2 & 3
Quad Cities 1 & 2
Brown's Ferry
1, 2, & 3
Peach Bottom 2 & 3
Brunswick 1 & 2
18.
19.
1976
Tritium
Gases
Curies
1976
Tritium
Liquids
Curies
-
1976
Total
Tritium
Curies
Curies of
Tritium
per
GWe-y
x
-45-
Table 20:
Halogens and Particulates Released to the Environment in 1975 by Pressurized Water Reactors (PWR's)
1975
Total
Halogens &
Particulates
m Ci
1975
m Ci
of H&P
per
GWe-yr
Name of Unit (PWR)
1975
Electrical
Power
Output
GWe-yr
1. Yankee Rowe
0.147
3
8
11
75
1975
Halogens
m Ci
1975
Particulates
m Ci
2.
Indian Point 1 & 2
0.601
410
1279
1689
2810
3.
San Onofre
0.366
246
36
282
770
4.
Haddam Neck
0.489
1
2
3
6
5.
Ginna
0.354
65
65
184
6.
Robinson
0.496
23
1
24
48
0.325
427
1
428
1317
7. Palisades
8.
Maine Yankee
0.537
6
6
11
9.
Fort Calhoun
0.245
7
7
29
10.
Kewanee
0.395
20
666
1686
11.
Three Mile Island 1
0.644
1
1
2
12.
Arkansas One 1
0.563
22
11
33
59
13.
Calvert Cliffs 1
0.512
36
11
47
92
14.
DC Cook
0.533
15.
Point Beach 1 & 2
0.764
189
94
283
370
16.
Suny 1 & 2
1.061
133
78
211
199
17.
Turkey Point 3 & 4
1.017
465
59
494
486
18.
Oconee 1, 2, & 3
1.718
11
11
6
19.
Zion 1 & 2
1.201
217
2
219
182
20.
Praire Island 1 & 2
0.822
21
4
25
30
12.791
2303
2202
4505
Totals
646
0
n0
G352We
-~GWe-yr
-46-
Table 21:
Halogens and Particulates Released to the Environment by Pressurized Water Reactors (PWR's) in 1976
Name of Unit (PWR)
1976
Electrical
Power
Output
GWe-yr
1. Yankee Rowe
1976
Total
Halogens &
Particulates
Ci
p Curies
of H& P
per
GWe-yr
1976
Halogens
Ci
1976
Particulates
I Ci
0.155
134
40
174
1,123
2.
San Onofre 1
0.283
4,470
1,110,000
1,114,470
3,938,057
3.
Haddam Neck
0.473
733
900
1,633
3,452
4.
Ginna, RE
0.255
33,500
90
33,590
131,725
5.
Robinson, H.B.
0.580
254,000
568
254,090
438,086
6.
Palisades
0.353
30,600
14,300
44,900
127,195
7. Maine Yankee
0.711
1,640
433
2,073
2,916
8.
Fort Calhoun
0.261
71,700
1,630
73,330
280,958
9.
Kewanee
0.395
3,270
468
3,738
9,463
10.
Three Mile Island 1
0.508
8,640
4,430
13,070
25,728
11.
Arkansas One 1
0.441
40,800
1,660
42,460
96,281
12.
DC Cook
0.785
1,370
2
1,374
1,750
13.
Millstone Point 2
0.553
49,000
536
49,536
89,577
14.
Point Beach 1 & 2
0.829
6,262
18,200,000
15.
Suny 1 & 2
0.921
545,000
311,000
856,000
929,425
16.
Turkey Point 3 & 4
0.982
552,000
76,500
77,052
78,464
17.
Oconee 1, 2 & 3
1.447
156,000
310,000
466,000
322,046
18.
Zion 1 & 2
1.131
89,200
2,860
92,060
81,397
19.
Prairee Island 1 & 2
0.752
24,600
226
24,826
33,013
20.
Indian Point 1, 2 & 3
0.492
273,000
28,100
301,100
611,992
Totals
12.307
1,594,473
20,063,265
21,657,738
1,759,790 pCi
Totals in mCi
12.307
1,594
20,063
21,658
1,758 mCi
Statistical extremes (#2,
14)
Net excluding statistical extremes
Power apportioned
Average mCi per GWe-yr
18,206,262 21,961,715
19,310
1,594
753
12.307
11.195
130
"'7
I
E17 I mCi
L
l_.I
GWe-y
-47-
Table 22:
Halogens and Particulates Released to the Environment in 1975 by Boiling Water Reactors (BWR's)
Name of Unit (BWR)
1975
Electrical
Power
Output
GWe-yr
1975
Halogens
Released
mCi
1975
Particulates
Released
mCi
1.
Dresden 1
0.093
5,700
359
2.
Big Rock Point
0.036
267
3.
Humboldt Bay 3
0.048
4.
La Crosse
5.
1975
Total
H &P
Released
mCi
1975
mCi of
Released
H&P per
GWe-yr
6,059
65,151
98
365
10,139
1,070
839
1,909
39,771
0.033
133
79,200
79,333
2,404,030
Oyster Creek 1
0.359
41,300
178
41,478
115,538
6.
Nine Mile Point 1
0.377
5,960
441
6,401
16,979
7.
Millstone Point 1
0.440
62,900
188
63,088
143,382
8.
Monticello
0.324
15,200
673
15,873
48,991
9.
Vermont Yankee
0.411
398
2
400
973
10.
Pilgrim 1
0.296
6,220
660
6,880
23,243
11.
Cooper Station
0.203
419
31
450
2,217
12.
Duane Arnold 1
0.271
406
9
415
1,531
13.
Hatch 1
0.357
6
6
17
14.
Fitzpatrick 2
0.249
18
18
72
15.
Brunswick 2
0.204
3
4
7
34
16.
Bresden 2 & 3
0.468
11,700
4,160
15,860
33,889
17.
Quad Cities 1 & 2
0.845
2,920
415
3,335
3,947
18.
Brown's Ferry
1 & 2, 3
0.320
597
59
656
2,050
Peach Bottom 2 & 3
1.219
36
4
40
33
Totals
6.553
155,253
87,320
242,573
37,017
0.033
133
79,200
79,333
2,404,030
6.520
155,120
8,120
163,240
2 5' 0 3 7
19.
Statistical
extreme
(La Crosse #4)
Totals excluding
La Crosse #4
GWe-y
-48-
Table 23:
Halogens & Particulates Released to the Environment in 1976 by Boiling Water Reactors (BWR's)
Name of Unit (BWR)
1976
Electrical
Power
Output
GWe-yr
1976
Halogens
Released
m Ci
1976
Particulates
Released
m Ci
1976
Total
H &P
Released
m Ci
1976
mCi of
H &P
Released
per GWe-yr
1.
Dresden 1
0.125
2,340
470
2,810
22,480
2.
Big Rock Point
0.031
16
58
74
2,387
3.
Humboldt Bay 3
0.025
368
27
395
15,800
4.
La Crosse
0.022
101
95,200
95,301
4,331,864
5.
Oyster Creek 1
0.431
46,400
220
46,620
108,167
6.
Nine Mile Point 1
0.478
8,640
103
8,743
18,291
7.
Millstone Point 1
0.425
36,500
149
36,649
86,233
8.
Monticello
0.451
1,020
50
1,070
2,373
9.
Vermont Yankee
0.373
71
6
10.
Pilgrim 1
0.278
1,980
332
2,312
8,317
11.
Cooper Station
0.433
91
14
105
242
12.
Duane Arnold 1
0.293
108
7
115
392
13.
Hatch 1
0.503
4
4
8
14.
Fitzpatrick 2
0.462
5,770
19
5,789
12,530
15.
Dresden 2 & 3
1.008
12,900
4,550
17,450
17,312
16.
Quad Cities 1 & 2
0.417
4,260
538
4,798
11,506
17.
Brown's Ferry
1, 2 & 3
0.491
598
60
658
1,340
Peach Bottom
2 &3
1.359
1,880
44
1,924
1,416
Brunswick 1 & 2
0.291
463
19
482
1,656
Totals
7.896
123,510
101,866
225,376
28,543
0.022
101
95,200
95,301
4,331,864
7.874
123,409
6,666
1.30,075
18.
19.
Statistical
extremes
(La Crosse #4)
Totals excluding
La Crosse
77
206
16 520GWe-y
'
GWe-y
-49-
Table 24:
Mixed Fission and Activation Products Released
to the Environment in 1975 and 1976 by
Pressurized Water Reactors (PWR's)
1976
1975
Electrical
Output
Name of Unit (PWR) GWe-yr
MF and
Activation
Products
Curies
Curies
per
GWe-yr
1.
Yankee Rowe
0.147
0.010
0.068
2.
Indian Point 1& 2
0.601
6.250
10.399
3.
San Onofre
0.366
1.220
3.333
4.
Haddam Neck
0.489
1.240
5.
Ginna, RE
0.354
6.
Robinson, HB
7.
Electrical
Output
GWe-yr
Curies
per
GWe-yr
0.009
0.058
0.283
7.430
26.254
2.536
0.473
0.130
0.275
0.420
1.186
0.255
0.685
2.686
0.496
0.440
0.887
0.580
0.375
0.647
Palisades
0.325
3.450
10.615
0.353
0.441
1.249
8.
Maine Yankee
0.537
3.200
5,959
0.711
2.840
3.994
9.
Fort Calhoun
0.245
0.133
0.543
0.261
0.549
2.103
10.
Kewaunee
0.395
0.447
1.132
0.395
2.850
7.215
11.
Three Mile Island
1
0.644
0.065
0.101
0.508
0.102
0.201
12.
Arkansas One 1
0.563
3.110
5.524
0.441
13.100
29.705
13.
Calvert Cliffs 1
0.512
1.490
2.910
14.
DC Cook
0.533
0.260
0.488
0.785
0.282
0.359
15.
Millstone Point 2
0.553
0.420
0.759
16.
Point Beach 1 & 2
0.764
3.350
4.385
0.829
3.560
4.294
17.
Suny 1 & 2
1.061
31.900
30.066
0.921
33.600
36.482
18.
Turkey Point 3&4
1.017
3.070
3.019
0.982
8.650
8.809
19.
Oconee 1, 2&3
1.718
5.060
2.945
1.447
6.670
4.610
20.
Zion 1 & 2
1.201
0.009
.007
1.131
21.
Prairie Island
1 & 2
0.822
0.454
0.552
0.752
0.012
0.016
0.492
5.900
11.992
12.307
87.605
I7118
22.
0.020
Indian Point
1, 2 & 3
(see #2)
Totals
12.791
65.578
15I l=TlI
0.155
MF and
Activation
Products
Curies
(see #22)
-50-
Table 25:
Mixed Fission and Activation Products Released
to the Environment in 1975 and 1976 by
Boiling Water Reactors (BWR's)
1976
1975
Name of Unit (BWR)
Electrical
Output
GWe-yr
MF and
Activation
Products
Curies
Curies
per
GWe-yr
Electrical
Output
GWe-yr
MF and
Activation
Products
Curies
Curies
per
GWe-yr
1.
Dresden 1
0.093
0.840
9.032
0.125
0.353
2.824
2.
Big Rock Point
0.036
2.020
56.111
0.031
0.770
24.839
3.
Humboldt Bay 3
0.048
3.470
72.292
0.025
1.080
43.200
4.
La Crosse
0.033
14.100
472.273
0.022
5.680
258.182
5.
Oyster Creek 1
0.359
0.408
1.136
0.431
0.221
0.513
6.
Nine Mile Point 1
0.377
21.000
55.703
0.478
0.214
0.448
7.
Millstone Point 1
0.440
199.000
452.273
0.425
9.650
22.706
8.
Monticello
0.324
Zero
0.451
Zero
Zero
9.
Vermont Yankee
0.411
10.
Pilgrim 1
0.296
2.060
6.959
0.278
2.340
8.417
11.
Cooper Station
0.203
1.730
8.522
0.433
0.653
1.508
12.
Duane Arnold 1
0.271
0.003
0.011
0.293
0.007
0.024
13.
Hatch 1
0.357
0.058
0.162
0.503
0.042
0.083
14.
Fitzpatrick 2
0.249
9.390
37.711
0.462
6.010
13.009
15.
Brunswick 2
0.204
1.920
9.412
16.
Dresden 2 & 3
0.468
0.810
1.731
1.008
1.210
1.200
17.
Quad Cities 1 & 2
0.845
17.100
20.237
0.417
6.990
16.763
18.
Brown's Ferry
1, 2 & 3
0.320
2.790
8.719
0.491
3.950
8.045
19.
Peach Bottom 2 & 3
1.219
0.929
0.762
1.359
2.820
2.075
20.
Brunswick 1 & 21
0.291
3.280
11.271
7.896
45.270
5.733
Totals
6.553
Zero
0.373
277.628
2.367
42.3671
-51-
HTGR and GCR
In gas cooled reactors, a non-reactive gas such as helium
or carbon dioxide is used to transport heat from the core to
the steam generators.
Moderation of the fast neutrons is
accompolished, instead of by water, by the placement of graphite
within the core.
tures
Helium coolants can be heated to higher tempera-
(7750) than carbon dioxide coolants
(3750) and there is
usually a distinction made between the two, naming the former
a
"high temperature gas cooled reactor"
simply a
"gas cooled reactor"
(HTGR) and the latter
(GCR).
The HTGR at Fort St. Vrain utilizes small fuel spheres
(100
micrometers in diameter) coated by two layers of pyrolytic
graphite.
Most fission products in the HTGR fuel will remain
fixed within the fuel pellet ceramic;
however, some may escape
through the two pyrolytic graphite coatings into the graphite
structure of the fuel elements and diffuse into the primary
helium coolant.
Very small amounts of radioactivity also result
from the activation of
the primary coolant.
impurities which may be circulating in
Helium-4 is not easily activated, having
a neutron absorption cross section that is effectively negligible:
a trace amount of helium-3 is found in nature
a large
(n,p)T cross section, forming tritium.
(130 ppm) having
The impurities
found in the helium coolant are most likely to be hydrogen and
oxygen leakages from the
fission products.
steam generators, in addition to escaped
-52-
The principal source of high activity radioactive gaseous
waste originates from the helium coolant purification system.
Small amounts of potentially contaminated gases also come from:
the sampling of the primary coolant, purging of fuel storage
and handling systems, purging of the helium circulator handlingcask and from the reactor vessel support floor vent and liquid
waste tank vent headers.
Additional sources of radioactive
effluent gases containing lower levels of activity are: the
secondary system
and the reheat
steam-jet air ejectors, the deaerator vent
steamline relief valves.41
Gases from all of the above sources are collected and processed.
Low activity gaseous waste is normally vented after
filtration; high activity gaseous waste is treated and held
up for a minimum time of 60 days.
It is interesting to note
on Table 26 that, because of this station's intermittent startup and shut-down record, one can observe from the monthly release
data this 60-day hold-up period.
The Atomic Energy Commission's Environmental Statement
for Fort St. Vrain estimated that noble gas releases, after
the hold-up period, would amount to 993.5 curies per year
(960.8 from the off-gas regeneration systme;
27.5 from the
reactor building leaks; and 5.2 from the air ejector).
42
Fort
St. Vrain was rated at the time of this 1972 report at 330
MWe of net electrical output.
Assuming a yearly utilization
factor of 0.65, one would expect the reactor to produce about
0.215 GWe-yr of power.
Hence, the noble gas release
all Kr-85) was estimated to be about 4,620 Ci/GWe-yr.
(almost
-53-
Fort St. Vrain was scheduled to go into commercial operation in December 1976.
As of December 1978, the reactor was
not yet in commercial operation.
Between January 1977 and June
1978, the station had produced 500,000 kwhrs of electric power,
a tiny fraction (about 0.02 percent) of its expected power
output.
Effluent releases for each of the four categories of
radionuclides
(noble gases, tritium, halogens and particulates,
and mixed fission and activation products) were compiled for
each month of operation from the station's Semi Annual Effluent
Release Reports on file at the NRC's public domument room at
1717 H street, N.W., Washington, D.C., and are shown in Tables
26 through 29.
Actual noble gas releases were on the order of 400 Ci for
the 18-month period.
Because the net power output was so low,
the average noble gas release was calculated to be about 95,000
Ci/GWe-yr.
Liquid radioactive wastes arise principally from decontamination operations.
Smaller quantities accumulate in the
regeneration section of the helium purification system; additional sources of radioactive liquid wastes result from the
reactor vessel linear cooling system and from leaks in the steam
generator feedwater system.
The liner coolant contains neutron
activation products; the steam generator feedwater contains
tritium which has diffused through the walls of the steam
generators from the primary coolant, and contains small amounts
of fission products. 4 3
-54Table 26:
MONTH OF
HTGR
OPERATION
Noble Gases Release to the Environment (1/77 thru
6/78) by HTGR at Fort St. Vrain
Monthly
Electrical
Output
Mwhrs
Noble
Monthly
Gases
Electrical
Airborne
Output
10- bGWe-yrs Curies
Noble
Gases
Dissolved
Curries
Curies of
Noble
Gases
Released per
Gwe-yr
C
DEC 1976
18
+JAN 1977
45
394
0.138
NSA
FEB
0
0
8.600
NSA
MAR
0
0
0.148
NSA
22,553
APR
4
35
5.980
NSA
170,857
MAY
9
79
0.238
8(10) 6
JUN
0
0
JULY
0
0
0.005
0.149
(10)NSA
AUG
10
8
7.600
NSA
91,862
SEP
41
359
9.950
(10)
27,716
OCT
74
648
54.400
<(10)
83,951
NOV
41
359
<(10)
DEC
0
0
7.120
70.010
30
0
263
0
0.488*
19.900
NSA
<(10)
MAR
0
0.003
NSA
APR
MAY
61
93
0
534
+JUN
92
JAN 1978
FEB
JULY
86
NSA
214,847
77,532
<(10)3
138,951
815
74.200
48.100
0.005
59,018
806
97.100
<(10)
120,471
*On Jan 23, 1978 an accident released 113
lbs of primary coolant containing 4.06 Ci
of Noble gases; this release is not reported in the Effluent Release Report submitted
by Ft. St. Vrain for 1-6-78
Totals
(1/77-6/78)
4300
404.129
Totals
(1977)
1882
164.338
(Negl.)
-
93,983
87,321
Totals (1978-1st half
2418
239.791
year)
99,169
Sources:
(1) Monthly electrical output was compiled from the
report summarv published by the NRC ("Grev Books" 12/7610/78):
(2) Release data dompiled from Ft. St. Vrain
semi annual reports on Radioactive Effluent Releases to
the Environment, Public Doc. Room, NRC, DC.
-55-
Table 27:
Tritium Released to the Environment 1977
by HTGR at Fort St. Vrain
Monthly
Electrical
Output
Gwe-yr
10
& 1978
Curies of
Tritium
Released per
Gwe-yr
Tritium
Gases
Curies
Tritium
Liquids
Curies
Tritium
Total
Curies
394
0.104
4.860
4.964
FEB
0
0.180
5.560
5.740
MAR
0
0.054
4.410
4.464
38,497
APR
35
0.026
6.850
6.876
196,457
MAY
79
0.086
1.510
1.596
20,203
JUN
0
0.007
6.790
6.797
JUL
0
0.013
1.800
1.813
AUG
8
0.066
0.293
0.359
103,092
SEP
359
0.221
7.000
7.221
20,114
OCT
648
1.250
2.780
4.030
6,219
NOV
359
0.204
3.590
3.794
DEC
0
0.353
6.180
6.533
263
0.189
9.750
9.939
FEB
0
0.482
33.300
33.782
MAR
0
0.014
12.000
12.014
211,920
APR
534
0.409
13.500
13.909
26,047
MAY
815
59.700
96.400
156.100
191,533
JUN
806
1.01
7.800
8.810
10,931
MONTH OF
HTGR
OPERATION
JAN 1977
JAN 1978
Totals
(1977)
Totals
(1978-
1st half)
Total
6/7 8)
28,766
1882
2.564
51.623
54.187
28,792
2418
61.804
172.750
234.554
97,003
43 0
64.368
224.373
288.741
67,149
(1/77-
-56-
Table 28:
Halogens & Particulates Released to the Environment
1/77 thru 6/78 by HTGR at Fort St. Vrain
Monthly
a-Activity
Gross 3,y
Electrical
in
MONTH OF
Activity
Output in Halogens in Particul. Particulates
HTGR
Curil
Cur'i
Gwe-yr
Curiy
OPERATION 10
X10
x10
X10
Total
Halogens
& Partic.
Curies
X10- 1
394
NSA
NSA
NSA
FEB
0
NSA
NSA
NSA
MAR
0
NSA
NSA
NSA
APR
35
NSA
NSA
NSA
MAY
79
NSA
NSA
NSA
JUN
0
NSA
NSA
NSA
JUL
0
NSA
3.22
0.89
4.11
AUG
8
NSA
3.50
1.31
4.81
SEP
359
NSA
5.48
3.93
9.41
OCT
648
NSA
5.46
1.87
7.33
NOV
359
NSA
4.91
1.01
5.92
DEC
0
NSA
3.50
0.80
4.30
263
NSA
3.95
1.98
5.93
FEB
0
NSA
4.79
0.89
5.68
MAR
0
NSA
3.10
0.84
3.94
APR
534
NSA
4.46
1.42
5.88
MAY
815
NSA
6.02
1.47
7.49
JAN 1979
806
NSA
7.56
1.45
9.01
JAN 1977
JAN 1978
10 - 9 Curies
per Gwe-yr.
7-12/77
1374
26.07
9.81
35.88
26.11
1-6/78
2418
29.88
8.05
37.93
15.69
Totals
(7/77-6/78
only)
3792
55.95
17.86
73 81
19.
-57-
Table 29:
Mixed Fission & Activation Products Released to the
Environment 1/77 thru 6/78 by HTGR at Fort St. Vrain
MF &
Monthly
Electrical
4gutput
10- u Gwe-yr
Gross
Activity
iCi
Gross a
Activity
pCi
Act. Prod.
Total
Activity
pCi
394
8.56
0.18
8.74
FEB
0
32.90
0.59
33.49
MAR
0
22.30
0.49
22.79
APR
MAY
35
79
8.60
37.00
0.35
1.20
8.95
38.20
JUN
0
5.70
0.45
6.15
JUL
0
7.37
0.29
7.66
AUG
8
SEP
359
7.71
54.10
0.30
2.97
8.01
57.07
OCT
NOV
648
359
14.40
4020.00
6.42
0.58
DEC
0
0.29
263
3.84
35.30
FEB
MAR
0
190.00
1.05
9.54
0
54.10
3.00
57.10
APR
MAY
534
815
120.00
49.80
0.38
0.53
120.38
50.33
JUN
806
94.90
27.90
122.80
Month of
HTGR
Operation
JAN 1977
JAN 1978
4034.40 *Exclude this month
as a statistical
7.00 extreme.
4.13
36.35
199.54
Curies per
Gwe-yr
Totals 1977
1882
208.90
4027.69
4236.59
2.251
Total ½ 1978
2418
544.10
42.40
586.50
0.243
Total (18
months)
4300
753.00
4070.09
4823 09
1.122
1234
194.50
7.69
202.19
0.164
3652
738.60
50.09
788.69
0.216
1977 (less
Oct)
18 Month
(less Oct)
-58-
The AEC's Environmental Statement estimated that the
maximum release of all tritium, mxed fission and activiation
products by way of liquid waste systems would be 0.041 curies
per year for an expected release of about 0.2 Ci/GWe-yr. 4 4
Actual data for the 18-month operating period shows that mixed
fission and activation products (Table 29) have amounted to
1.12 Ci/GWe-yr.
Tritium releases (Table 27) amounted to 288
curies; this is the equivalent of 67,000 Ci/GWe-yr.
No esti-
mates were made for halogens and particulates; actual data (Table
23) reveals that these radionuclides are very rarely found in
the releases, having a value of less than 20 nano-curies (10 9
curies) per GWe-yr.
The relatively large amount of tritium
is presumed to arise form (n,p)reactions on helium-3 and from
tri-fission.
Some additional tritium may leak from (n,2a)
reactions on B-10 in the control assemblies.
Some data from European gas cooled reactors was found for
the years 1970 through 1974,45 and the radioactive effluents
for noble gases in tritium were analyzed as shown in Tables
30 and 31.
This data is sketchy and incomplete.
Noble gas
data for GCR's in the United Kingdom, for example, were not
routinely or systematically measured
during these years.
Net power output for each reactor over these years was not
available and had to be estimated from rated power.
Never-
theless, average figures for both noble gases and tritium
were calculated, and found to be 118,760 Ci/GWe-yr and 858
Ci/GWe-yr respectively.
These points are plotted on Figures
4 and 5 among similar data for the other reactors.
-59Table 30:
Name of Unit
(GCR)
Chinon TR-1,
France
Noble Gases Released to the Environment (1971-1974)
by European Gas Cooled Reactors (GCRs)
1971
Noble
Gases
1972
Noble
1973
Noble
Curies
Curies
Curies
Curies
4225
11515
2808
3425
3863
Curies per
1974
Estimate
Gwe-yr.
Noble
Max
Gases
Gases
Gases
of Noble of Elect
Gases
Output Released Released Released Re leased
Output
Mwe
Released*
Gwe-y*
2082
17,000
0.304
4967
4338
= 10,500
0.398
70
Chinon TR-2
210
Chinon TR-3
480
St Laurent des
Eaux, TR-1
480
St Laurent des
Eaux, TR-2
515 I
Bugey TR-1
540
NA
641
3097
4475
17,500
0.216
Latina,
153
2470
3600
2050
3011
=180,000
0.061
-
32000
=400,000
0.079
-
10000
=100,000
0.100
Italy
Calder, UK
200
Chapelcross
198
Brodwell
250
Berkeley
276
For GCRs in the UK
Hunterston A
300
Noble gases were not
Trawsfynydd
390
systematically measured;
40000
=250,000
0.156
Hinkley Point
A
460
these are estimates for
80000
-435,000
0.184
Dungness A
410
Argon-41 (99% of
Sizewell A
420
Noble Gases released).
Oldbury
400
Wylfa
840
=177900
=118,760
1.498
French GCRs
UK GCRs
Italy GCRs
"HTGR"
Total GCRs
= 14,995
309,550
180,000
94,000
=145,770
-
Assumed annual averages
*Estimate of annual electrical
output assumes capacity factor
(utilization factor) of 40%.
-
Mwe
Represented
1535
1298
153
330
3316
"Radioactive Effluents from Nuclear Power Stations in Europe
Source:
1970-1974," appearing in Nuclear Safety, Vol. 18, No. 3, May-June 1977,
pp. 370-380.
-60-
Table 31:
Name of Unit
(GCR)
Tritium Released to the Environment (1970-1974)
by European Gas Cooled Reactors (GCRs)
1970
1971
1972
1973
1974
Avg.
Curies
Est of Tritium
per
Rated Elect. Liquids
Gwe-yr
Power Output Released
of Tritium
Mwe Gwe-yr Curies Curies Curies Curies Curies Curies Released
Chinon TR-1,
France
70
Chinon TR-2
210
Chinon TR-3
480
(Data not available)
St Laurent des
Eaux TR-1
480
St Laurent des
Eaux TR-1
515
Bugey TR-1
540
0.216
Latina, Italy
153
0.061
Calder, UK
200
Chapelcross
Brodwell
198
0.079
5.3
250
0.100
95.3
Berkeley
Hunterston A
276
300
0.110
60.1
Trawsfynydd
390
0.156
67.7
41.9
46.0
Hinkley Point
A
460
0.184
18.6
24.9
38.6
Dungness A
Sizewell A
410
420
0.164
0.168
18.6
20.9
35.5
Oldbury
Wylfa
400
840
0.160
0.336
17.3
-
0.120
824
17
159
13
12.7
102
43.1
17
9.3
251
44.2
33
7
11.7
1.2
198
200
117
56.7
67.0
824
3815
17.4
285
8.0
101
152.7
1527
80.8
102.4
60
66.3
735
853
425
30.0
39
30.2
164
28.9
30.5
20.0
26.7
163
77
53.2
208.0
253.0
729
64.4
30.2
15.0
82.7
13.6
275
37.4
122.4
29.5
162
37.5
86.7
116
134
130.5
184
388
UK Totals
1.577
749.5
475 Avg
European GCR Avg
1.854
1590.9
858 Avg
Source:
"Radioactive Effluents from Nuclear Power Stations in Europe 19701974," appearing in Nuclear Safety, Vol. 18, No. 3, May-June 1977,
pp. 370-380.
-61-
Heavy Water Reactors
The radionuclide inventory in heavy water reactors is
similar to that in light water reactors with respect to the
escape of fission products from failed fuel, neutron induced
activation of coolant and circulating impurities and corrosion
products.
However, a major difference arises from the fact
that heavy water forms tritium upon neutron capture.
The
inventory of tritiated water in the primary heat transport
system at equilibirum into a CANDU reactor is about 1 curie
per kilogram. 4 6
The rates at which fission products escape from failed
fuel depends directly on the rate of fuel failure.
However,
CANDU reactors can replace fuel without shut-down and, when
failed fuel is detected, remove it,
thus keeping leaked fission
products in the coolant to a minimum.
Because heavy water represents a large capital investment,
there are strong economic incentives to reduce heavy water leaks
to a minimum, and to capture all that does leak to the maximum
extent possible.
Consequently, heavy water reactors have complete-
ly closed. ventilation systems fitted with molecular-sieve
dryers to, recover heavy water vapor.
These filters remove
tritiated water as well, and also remove iodine and soluble
cesium radionuclides.
Further, the closed system serves as a
hold-up system for noble gases.
Similarly, all leaks of liquid of heavy water from the low
pressure systems are collected so far as possible in tanks for
-62-
heavy water recovery.
Vacuum cleaning systems are used to
collect smaller amounts of heavy water.
The heavy water
collection tanks and recovery systems, again, serve to contain tritiated water and serve as a hold-up system for dissolved fission products.
Actual release data for heavy water reactors was very
limited.
Some data was obtained, however, for:
at Karlsruhe, Germany;
(1) the PHWR
(2) the HWR at Marts d'Arree, France;
(3) and for four early CANDU reactors: NRX and NRU, WR-1,
NPD and Douglas Point.
From this sketchy information,
figures for both noble gases and tritium were calculated as
shown on Tables 32 and 33;
thesedata points are shown also on
Figures 4 and 5.
It appears from this limited picture that because of the
tight control on heavy water leaks, heavy water reactor systems
enjoy some additional features of radioactivity protection
that light water reactors may not have.
As a consequence, al-
though one would expect a higher tritium release than for light
water reactors, it seems to be about the same as the US PWR;
also the release of noble gases seems to be about the same as
for other closed primary coolant systmes in the US.
Liquid Metal Fast Breeder Reactors
Radioactivity concentrations in the liquid metal coolant
and cover gas are produced from the following sources, presuming a LMFBR design similar to that for Clinch River:
(1)
neutron activation of the sodium coolant, its trace impurities,
-63Table 32:
Noble Gases Released to the Environment (1970-1974)
by Heavy Water Reactors (HWR, PHWR, SGHWR, CAWDU)
1970
1971
1972
1973
1974
Avg
Estimate
Noble
Noble
Gases Curies
Rated of Elect
Gases
Name of Unit Power Output Released
Released
per
Gwe-yr
Curies Curies Curies Curies Curies Curies Gwe-yr
(HWR)
Mwe
Karlsruhe
(PHWR)
51
0.0204
Marts d'Arre
(HWR)
70
0.0280
Winfrith
(SGHWR)
92
0.0368
(Not measured)
NA
NRX & NRU
(CAWDU)
= 10
0.0040
6400
6400 1600000
WR-1
30
0.0120
Y-·mT- z
.,s
NPD
25
0.0100
Estimated
Douglas Point 220
250
Gentilly
0.0880
440
0.1000
Data not available (NA)
-
540
0.2160
Data not available (NA)
-
Pickering
526
955
228
952
665
53810 144450 130051 164460
-_
t- lllla
_---
tUL
>
NA (not avail.)
123192
27708
4.4(10) E
NA
30
2500
45
4500
440
5000
NA
NA
NA
NA
Averages
0.1624
124436
766232
Avg excl
stat extreme
0.1344
1244
9256
Source:
"Radioactive Effluents from Nuclear Power Stations in Europe 19701974," appearing in Nuclear Safety, Vol. 18, No. 3, May-June 1977,
pp. 370-380
-64-
Table 33:
Name of Unit
(HWR)
Rated
Power
Mwe
Tritium Released to the Environment (1970-1974)
by Heavy Water Reactors (PHWR, HWR, SCHWR, CAWDU)
Estimate
of Elect
Output
in
Gwe-yr
Karlsrahe
(MZFR), Ger.
(PHWR
51
0.0204
Maits D'Aree,
France (HWR)
70
0.0280
Winfrith, UK
(SGHWR)
92
0.0368
NRX & NRU,
Canada
10
0.0040
WR-1
30
0.0120
NPD
25
0.0100
Douglas Point
220
0.0880
Gentilly
250
Pickering
540
1970
1971
1972
1973
1974
Curies
Curies
Curies
Curies
Curies
Avg
Curies
Liquid effluent discharged to decontamination center; separate data not available.
5
42
116
116
Liquid effluent data not available
11
NA
NA
NA
NA
NA
NA
NA
NA
30.6
NA
NA
NA
NA
30.6
32.6
NA
NA
NA
NA
32.6
0.1000
NA
NA
NA
0.2160
NA
NA
NA
11
Estimates from
sketchy and old
data
0.1300
Average Curies per Gwe-year
190.2
1463
Sources:
1.
For PHWR, HWR & WGHWR:
"Radioactive Effluents from Nuclear Power
Stations in Europe 1970-1974."
Nuclear Safety, Vol. 18, No. 3.
May-Jun 1977 pp 370-380.
2.
For CANDU reactors: "Some aspects of the Releases of Radioactivity
and Heat to the Environment from Nuclear Reactors in Canada", AECL4156. Atomic Energy of Canada, April 1972.
-65-
the argon cover gas and structural components;
(2) fission
product releases to the primary coolant through cladding failures
in both the reactor core and blanket assemblies;
(3) plutonium
isotope inventory and production in the core and blanket released
to the coolant through cladding failures; and (4) the production
of tritium by neutron capture in boron in the control assemblies,
by tri-fission and by neutron capture in lithium-6 impurities
in the sodium coolant. 4 8
No actual release data for LMFBR's could be readily obtained
for this report.
Data on the Phenix at Gard, France is not
reported among the data on other European reactors.
Estimates
of releases are made for the CRBR, however, in both the FSAR
49
and the Environmental Report.
These data are summarized
for noble gases and tritium in Tables 34 and 35 and indicated
schematically on Figures 4 and 5.
Released rates per GWE-yr for noble gases, halogens and
particulates, and mixed fission and activation products are
estimated to be so low that the data points cannot be plotted
on the Figures comparing other reactor types.
Only tritium is
of a comparable level, about the same as the PWR, about one
order of magnitude greater than the BWR.
Summary for Power Reactors
Figures 4, 5, 6, and 7 summarize graphically the information collected on the five reactor types for noble gases,
tritium, halogens and particulates, and mixed fission and
activation products, respectively.
-66-
Table 34:
Noble Gases Estimated to be Released to the Environment by the CRBR-LMFBR
Noble
Gas
Isotope
Total
Leakage
Ci/day
Ne-23
5.3x10
Ar-39
3.3x10
Ar-41
5
3.300
1.8x10
Kr-85
Kr-88
6.3x10
-4
5.0xlO0
-4
4.2x10
-4
8.7x10
Xe-13lm
1.4x10
5
0.014
Xe-133
1.6x10
3
1.600
Xe-133m
8.9x10
5
0.089
Kr-87
0.180
0.006
0.500
0.420
0.870
-3
Xe-135
Xe-135m
Xe-138
Totals
6.7xl0
-4
1. lxl10
6.700
0.110
1.8x10
0.180
1.44x10
23.3%
0.120
-4
4
Kr-83m
Kr- 85m
Gwe-yr
0.053
-4
1.2x10
Curies of
Noble Gases
Released
Total
Leakage
Ci/d
X10
2
14.142
11.3%
47.4%
82
CRBR rated at 380 Mwe gross
350 Mwe net
Assume 6C) percent utilization
Net power output at Gwe-yr = 350(.60)
100 0
Source:
%
24.58 Curies per Gwe-yr
of Noble Gases
Principally:
Xe-135
Ar-39
Xe-133
0.210 Gwe-yr
Environmental Report, CRBR, Volume II, Table 5.2-1.
-67Table 35:
Tritium (Estimated) to be Released to the
by the CRBR
-
------- - --
Environment
-
Category of tritium
Tritium
Release
Estimate
(units).
8.7x10
Ci/day
3
1.
Released from
gaseous RodWaste System
2.
Released from
1.02x10
Liquid Rodwaste
M Ci/cc
System (low level
activity)
3.
Released from
Liquid Rodwaste System
(Intermediate
level activity)
1.50x10
4
9
Curies of
Tritium
Released
Per
Gwe-yr.
Annyalization
Factor
Annual
Tritium
Released
Curies
365 days yr.
3.2
3.2x10 12 cc
yr
326.4
1554.3
329.6
1570
15.2
3.2x1012 cc
yr
Net power output = 0.210 Gwe-yr.
Source:
Environmental Report CRBR, Tables 5.2-1 and 5.2-2, VD II.
-68-
'10,000
1,000
BWR
s
V;f
L
-cl.`;
'
&-1,A5 lLI
C:
I
0
2
a,
'D
100
4sTE
03
SK\1v
Tweak +;
-
I
I
·4
/0
It
10
\I
1.0
I
II
I
I 1I
,I ,
. I
pI
I
I
III
I
I II .
I
PWR
I
I
/
I
70
71
72
73'
74
YEAR OF EFFLUENT RELEASE
Figure 4
__ I
,
75
1
1
"is
RRr
K
-69-
/
10,000
TFRT ST Vole
.
4%
-t
'"a
CC6R
snrAite
s 1
C.s
1,000
"2]-
-e
'~_-
--
El
PWR
,S
I·
AI
.0
IfJ
a3:
,
100
10
BWR.
....
I .
I
I
I
I
I
I
70
71
72
73
74
YEAR OF EFFLUENT RELEASE
16
14
TI
1q
I
Figure
5
-70100
-BWR
10
w
LU
K
1.0
/
-
z~~~~
/ ~
I~~
/~~
/~~
1
z
w
I
/
/~~
U~~
! II
1:3
-",ll·
\PW
---.-PW R
II
!
0.1
I
CD
irfi
A,
0.01
I
_ ___
I
I
__
I
I
Figure 6
1
__I__
70
71
72
73
74
YEAR OF EFFLUENT RELEASE
I'
o x Ilo
-------
1
-~,
I
I
I-
-71-
IAA
1UU
!-,
I
0
I
01
Ci,
I'
p,
vR
>1
>
OF
I
I,
z
03:
cal
1
0"I
CV,
LL,
x
UJI
_,
/
w
a:'
1.0
i
I
I
I .
74
73
72
71
70
YEAR OF EFFLUENT RELEASE
0.1
-'1
Figure
7
_
I- ----'S
76
11
-
-72-
In general, one observes a gradual trend toward improvement over time for reactors having had higher release rates
relative to other reactors.
BWR's seemed to have improved
dramatically with respect to noble gas releases since 1974,
although they still release about 10 times as much as PWR's.
From a closer inspection of individual BWR plant performances,
shown on Table 15, it can be seen that some BWR plants are
doing about as well as PWR's; witness, for example, Dresden 2
and 3 in 1976.
Similarly, PWR's are improving over time with respect to
tritium releases.
Tables 16 and 17 show an interesting extreme
for Zion 1 and 2, having nearly zero release of tritium in
both 1975
and 1976, while generating at a significant level of
power output.
Both. PWR's and BWR's have shown improvement in the release
of mixedf:Eission and activation products.
-73-
"BACK END" OF THE URANIUM FUEL CYCLE
Spent Fuel Storage
Every 12 to 18 months, depending upon both plant design
and utilization, reactors must be refueled.
CANDU reactors
refuel as well, but do not have to shut down while doing so.
The spent fuel discharged from the reactor is highly radioactive and must be cooled for a period of time to allow for
the dissipation and reduction of decay heat.
This cooling
period is usually not less than 60 or 90 days, and can extend
indefinitely.
The spent fuel is first placed in "spent fuel pools" at
the reactor site, containing water as the cooling and shielding medium.
All radionuclides present in the irradiated fuel
can potentially escape to the atmosphere within the storage
space or building, and may consequently escape to the environment through gaseous and liquid effluent streams.
If no recycling or reprocessing of spent fuel occurs,
as is presently the case, these spent fuel pools must be relied upon indefinitely to safely store the irradiated fuel
assemblies accumulating at the rate of about 35 to 50 metric
tonnes per GWe-yr.
Federal plans call for the construction of
a geologic repository to receive this spent fuel after a
period of 5 to 10 years of cooling at the reactor site, but
due to complex institutional and political factors, recent
estimates of when such a repository may actually receive its
first spent fuel have been extended to the mid-1990's.
In
-74-
the meantime, additional
"away from reactor"
(AFR) spent fuel
pooling capacity is being considered.
Radioactive effluent releases from spent fuel pools at
reactor sites are not itemized in the monthly and semi-annual
reports; releases from such sources are presumably included in
Hence, no data was found specifical-
the overall site reports.
ly relating to spent fuel pool operations.
The GESMO Final Report, however, does estimate the accumulated release of krypton-85 from spent fuel pools over the
period from 1975 to 2000 to be 560,000 curies.50
This would
be a commitment over the 26-year period of an average of 182
curies per GWe-yr of power produced.
Similarly, estimates
for dose commitments are 11,000 person-rem for the 26-year
period;
3.8 person-rem per GWe-yr.
Waste Disposal and Transportation
At present, the spent fuel pool located at the reactor
site or at an AFR location
proposed)
(Barnwell, South Carolina, has been
is the end of the fuel cycle, and will remain as
such until at least 1993
(latest US Department of Energy esti-
mate for the opening date of a full scale geologic repository).
The GESMO Final Report does estimate for the "no recycle"
case, however, what the potential releases and dose commitments
would be if spent fuel were stored for 10 years at the reactor
site and then shipped to a geologic repository.
These estimates
are shown in summary form in Tables 36 and 37 of the concluding
section of this paper.
-75-
SUMMARY FOR THE NUCLEAR FUEL CYCLE
In the preceding sections, each component of the nuclear
fuel cycle has been examined, some in more depth than others,
with respect to the release of radioactivity to the environment.
Both actual data on releases of operating plants and
estimates made for "model" plants have been presented and
compared.
Much use was made of the "Final Generic Environmental
Statement on the Use of Recycle Plutonium in Mixed Oxide
Fuel in Light Water Cooled Reactors," published by the
Nuclear Regulatory Commission (NRC) in 1976, as it provided
a consistent methodology for translating estimates of isotopic releases into "person-rems" for each fuel cycle
component.
By making use of this report, referred to as the GESMO
study, a first approximation of the radiological health
effects resulting from 1 GWe-year of nuclear power can be
made for each of the various reactor types, presuming that
the GESMO modeling is valid, by taking into consideration the
comparison of actual release data to the estimated releases
used in the GESMO analysis.
Table 36 summarizes the radioactivity release estimates
used in the GESMO analysis for each component of the nuclear
fuel cycle.
Table 37 presents the results of the pathway
modeling of these released estimates in terms of total
-76-
Table 36:
Estimated Radioactive Effluents Released to the
Environment by the Fuel Cycle Components for the
Period 1975 through 2000 (no recycle option)
Fuel Cycle
Component
Mining
Milling
UF6 Conversion
Enrichment
U0 2 Fabrication
Reactors
Spent Fuel Storage
Waste Management
Transportation
Totals
Estimated
Effluent
Releases
Curies
Estimated
Equivalent
Power
GWe-yr
24,000,0)00
4,400,0)00
4L62
4
2!71
60,820,0'00
560,0)00
20
4,732
4,732
4,551
4,603
4,346
4,036
3,072
3,857
3,146
5,072.
930.
.102
.001
.062
15,069.
182.
.005
89,780,625
21,253.
Source:
GESMO Final Report, Summary Volume, Table
Volume 3, Table IV-J(E)-17
Table 37:
Estimated
Releases
Ci/GWe-yr
5(A)-l and
Estimated Dose Commitments for the US Light Water
Fuel Cycle for the Period 1975 through 2000 (no
recycle option)
Whole Body Dose
Commitment
26-yr Period
Person-rem
Fuel Cycle
Component
Mining
Milling
UF6 Conversion
Enrichment
U0 2 Fabrication
Reactors
Spent Fuel Storage
Waste Management
Transportation
Totals
Whole Body Dose
Commitment
Person-rem
per GWe-yr
%
49.4
13.5
0.6
3,000
6,900
909.7
248.0
10.8
0.7
12.5
656.3
3.8
0.4
0.5
8,224,800
1842.7
100.0
4,200,000
1,100,000
46,000
3,500
54,400
2,800,000
11,000
0.7
35.6
0.2
Source:
GESMO Final Report, Summary Volume Table 5(A)-l and
Volume 3, Table IV-J(E)-9, September 1976.
Data includes occupational exposures (51.8 percent), US population exposures (45.4
percent) and foreign population exposures (2.8 percent)
-77-
population exposures--U.S., foreign, and occupational.
If the release estimates used to develop these numbers
fairly represent actual operating experience, and if the
pathway modeling is valid, then one can conclude that the
population exposure due to 1 GWe-year of nuclear power is
approximately 1850 person-rems, of which about 63 percent
is due to the mining and milling operations and 35 percent
is due to the operations of the reactors.
Less than 2
percent is attributed to the six other components of the
nuclear fuel cycle.
Table 38 translates these population exposures into
health effects, giving 1.170 health effects per GWe-year,
consisting of:
(a) 0.273 cancer deaths;
(b) 0.372 non-
fetal cancers; and (c) 0.525 genetic defects per GWe-year.
Data on actual releases compares favorably with the
estimates used in the GESMO study, except for the case of
power reactors.
The GESMO study assumes that reactor
releases would be of the order of 15,000 curies per GWeyear.
While this is true for the recent experience of the
PWR, it significantly understates the releases of the other
power reactors, as Table 39 shows.
-78-
Table 38:
Estimated Health Effects from Radioactive Releases
to the Environment by US LWR Industry for the Period
1975 through 2000 (no recycle option)
Type of Health Effect
Health
Effects
Bone cancer deaths
Thyroid cancer deaths
Lung cancer deaths
Other cancer deaths
140
60
390
510
0.035
0.015
0.097
0.126
Total cancer deaths
1,100
0.273
Benign and malignant
thyroid nodules
1,500
0.372
Specific genetic defects
1,300
Defects w/complex etiology
820
0.322
0.203
Total genetic defects
2,120
0.525
Total health effects
4,720
1.170
Health Effects
per GWe-yr
Source:
GESMO Final Report, Chapter IV, Table IV-J-14, Vol.
September 1976.
A factor of 4036 GWe-yr of electric power
was assumed for the 26-year period, per Table IV-J(E)-17 for
LWR reactor operations.
3,
-79-
Table 39:
Reactor
Type
Summary Table of Averages Radioactivity Releases
by Various Power Reactor Types for Recent Years
Noble
Gases
(Ci/GWe-yr)
Tritium
(Ci/GWe-yr)
Approximate
Total
(Ci/GWe-yr)
BWR
300,000
100
300,000
HTGR
100,000
1,800
100,000
GCR
100,000
700
100,000
HWR
10,000
1,800
12,000
PWR
10,000
1,200
11,000
-
-
15,000
GESMO
Source:
See pages 68-71.
Accordingly, one might expect that the health effects predicted by the GESMO study would apply to the PWR and HWR,
but not to the BWR, HTGR or the European GCR's.
In the
latter case, knowing that reactor operations account for
approximately 35 percent of the total fuel cycle estimate
of 1.170 health effects per GWe-yr, one can estimate what
the health effects would be for the larger releases by
scaling up the reactor component of the total.
By doing
this, the impact figures rise to about 9.0 health effects
per GWe-yr for the BWR, and to 3.5 for the HTGR and GCR.
It is interesting to note that in the cases of the PRW and
-80-
HWR, health effects are dominated by the mining and
milling operations; in the cases of the BWR, HTGR and GCR,
health effects are dominated by the operations of the
reactors.
Lastly, one might propose that that uranium and plutonium recycle could reduce the health effects of mining and
milling operations by extending the productivity of already
mined ore, thereby reducing mining operations.
Table 40
presents a comparison of the population dose commitments of
the two options:
plutonium recycle.
(a) no recycle and (b) uranium and
Although the recycle option does reduce
exposures (and the consequent health effects) due to mining
and milling, these reductions are more than offset by
increases due to reprocessing.
The net increase in expo-
sures for the recycle option is about 8 percent, rising
from 1850 to 2000 person-rem per GWe-year.
The breeder
option is not examined here.
In conclusion, the radiological impacts of nuclear
power appear to range between 1 and 10 health effects per
GWe-year, depending upon the type of reactor.
For each
"health effect," there are 0.23 deaths, 0.32 non-fetal
cancers and 0.45 genetic defects.
Mining and milling
operations account for the majority of the health effects
for a fuel cycle involving reactors releasing less than
15,000 Ci/GWe-year (PWR's and HWR's) of radioactive gases,
-81-
Table 40:
A Comparison of Estimated Population Dose Commitments for the US Light Water Reactor Fuel Cycle
for the Period 1975-2000 between the U and Pu
Recycle and the No Recycle Options
Fuel Cycle
Component
Mining
Milling
UF6 Conversion
Enrichment
U02 Fabrication
MOX Fabrication
Reactors
Spent Fuel Storage
Reprocessing
Waste Management
Transportation
Total Person-rem
Person-rem/GWe-yr
Source:
Whole Body Dose
Commitments for
No Recycle Option
(Person-rem)
Whole Body Dose
Commitments for
U and Pu Recycle
(Person-rem)
4,200,000
1,100,000
46,000
3,500
54,400
3,000
6,900
3,240,000
880,000
35,300
2,800
46,200
25,300
2,820,000
3,350
1,848,000
3,000
10,100
8,224,800
8,888,750
1,850
42,000
2,800,000
11,000
GESMO Final Report. Summary Volume, Tables S(A)-l
and S(A)-3, September 1976. Note that the above
estimates are accumulated occupational, US and
foreign population exposures for the 26-year period
1975-2000 based on a US nuclear scenario of 500
gigawatts of installed capacity in the year 2000.
this apportionment changes dramatically for certain other
reactors releasing one to two orders of magnitude more radioactivity per GWe-year (BWR's, HTGR, and GCR's).
Trends in
the release data over the period from 1970 through 1976,
shown graphically in Figures 4, 5, 6 and 7, show significant
improvements being made in the case of the BWR; trend data
for the H:TGR and the European GCR's are not available.
A
0
0
+
+o
-82-
ADDENDUM--RADIOACTIVITY RELEASES
FROM FOSSIL FIRED POWER STATIONS
Little is known about the radiation exposures caused
by the various fossil fuel cycles.
Some data at particular
sites are available and some estimates have been made for
overall population exposures by scaling sparse and sometimes
atypical data upwards, but a comprehensive evaluation has
yet to be made.
For the purposes of perspective, however, a brief
analysis is made here of some of the available data as presented in a recent EPA report entitled, "Radiological Quality
of the Environment in the U.S., 1977."
Natural Gas
As with the rest of our natural environment, all fossil
fuels contain some level of radioactivity.
Natural gas drawn
from sandstone deposits, normally associated with radiumbearing geological strata, has been estimated 5 1 , on the
average, to contain approximately 20 pCi per liter of radon.
Use of this gas in the home has been estimated to expose the
U.S. population to a total dose commitment of 2,730,000
person-rems per year; using the conversion factor of 200
cancer deaths and 400 total radiological health effects per
million person-rem (the same factor applied to the nuclear
fuel cycle),
this use of natural gas results in approximately
-83-
550 cancer deaths and 1100 total health effects per year.
This annual dose commitment is very large by comparison to
any other energy option, and is exceeded in absolute terms
only by three other categories of exposure:
natural back-
ground, medical and dental x-rays and radio-pharmaceuticals.
A natural gas fired generating station producing 1 GWeyear of net electrical power is benign by comparison.
It
requires approximately 1011 cubic feet of natural gas per
GWe-yr, or 2.75 x 1012 liters.
At 20 pCi/liter, the radio-
activity released to the environment would be 55 curies of
radon per GWe-year, resulting in 10 person-rem of exposure
and 0.004 health effects.
Oil
Trace concentrations of natural radioactive decay chains
in oilare generally considered so small as to be negligible. 5 2
Coal
Coal deposits are known to contain concentrations of
uranium and thorium, ranging from 0.001 to 0.100 percent
uranium in Western coal samples from Wyoming and Idaho, with
an average concentration of 0.008 percent. 5 3
Recall, as a
comparison, that conventionally mined uranium ores contain
concentrations of 0.090 percent or higher.
Eastern coals
have been found to contain approximately one-tenth the concentration of Western coals.
-84-
For coal-fired generating stations there are two major
sources of radioactivity releases due to plant operations:
gaseous effluents and flyash residues.
One estimate for a
1000 MWe station, equiped with electrostatic precipitators
of 99.7 percent efficiency, gives the release in gaseous
effluents to be about 2.1 Ci/yr, or about 3.2 Ci/GWe-year,
consisting mostly of Rn-222 and Po-210.54
In the same study
flyash residues were estimated to contain not more than
5.0 pCi/gram of radium-226.
Coal-fired stations operating
at 35 percent efficiency require about 4 million tons of
coal (about 3.7 x 1012 grams) per GWe-year.
Assuming that
the flyash residue is about 4 percent by weight of the
original coal, the radium-226 discharge, given the above
concentration estimate, would be 0.74 Ci/GWe-yr.
By another calculation, assuming the average uranium
concentration for Western Coal is 0.008 percent by weight
and knowing that for every gram of uranium-238 in secular
equilibrium with its decay chain daughters there is 3.59 x
10- 7 grams of radium-226, one can calculate the radium
"throughput" at about 100 grams per GWe-year.
This radium
is the equivalent of 100 Ci/GWe-yr and implies a flyash concentration of about 7000 pCi/gram, which is at considerable
variance with the earlier quoted estimates.
Nevertheless,
the fact remains that a coal facility using Western coal
will receive about 100 grams of radium-226 in its fuel
-85-
stream and this radium must end up somewhere.
The radon-
222 release from the radium-226 is, likewise, about 100
Ci/GWe-yr.
Use of Eastern coal would result in about one-
tenth this amount.
Over What Period of Time?
Before continuing, a significant point must be raised
regarding the period of time into the future over which one
"counts" the accumulating release of radon due to the production of power in some past year.
In the case of nuclear
power, the parent of radon-222 (radium-226) is unearthed by
mining and remains in the tailings piles on the surface for
thousands of years.
Likewise, approximately 10 percent of
the original uranium is left behind in the tailings piles
after milling.
Together, the radium and uranium continue
to decay, continue to produce and release radon to the
environment, and continue to expose future populations to
an incremental amount of radioactivity long after the power
from the ore was produced.
Similarly, the uranium and radium concentrations in
coal are unearthed by mining and dispersed to the environment by stack effluents or deposited on the surface in the
flyash residues.
Worse, the flyash is used in construction
materials, thus enhancing the exposure pathways to man.
Natural gas is different, however, in that the parent
radionuclides of radon are left in the ground.
Once the
-86-
radon concentration in the gas is dispersed by the combustion exhaust, there is no more.
The "total release commitment" of radon to the environment due to 1 GWe-year of power, strictly speaking, would
be the incremental release rate resulting from the unearthing of the ore (uranium or coal), which is decaying function
of time, integrated over all remaining time, from zero to
infinity.
Likewise, the "total dose commitment" would be
the incremental population exposure rate resulting from
these "technologically enhanced pathways," integrated over
time from zero to infinity.
Clearly, these numbers will be
very large, even though both rates are slowly diminishing
over time.
As an example, some numbers have been put together
using the release estimates developed in this paper.
For
the LWR case, it is assumed that 5,000 curies of radon are
released in the mining and milling operations in the first
year, 870 curies per year are released from the remaining
radium as it decays in the piles, and 10 percent of the
original uranium is left in the piles after extraction,
giving a new equilibrium release rate of 87 curies per year
after the original radium decays.
The assumptions for the
fossil fuels are stated in the Table.
-87-
Table 41:
1 GWe-yr
from...
A Comparison of Radon-222 Releases from 1
Year of Power
Peak
Annual
Annual
Release
Rate
Release
Release
Rate
Rate
in year
After
After
of ex2
traction 1st Yr.
0,000y
(curies)
(curies) (curies)
Radon Source
LWR
mines & mills
Coal (Western)
ash piles, stack
Natural gas
exhaust gas
Coal (Eastern)
ash piles, stack
GWe-
Total
Radon Release
Commitment
(curies)
5,870
870
87
565,000,000,000
100
100
100
650,000,000,000
55
0
0
10
10
10
55
65,000,000,000
Over what length of time should today's population be
concerned about the health effects it imposes on future
generations?
This is a very difficult question.
If a specific
length of time is proposed, such as the 26 year period used by
the GESMO study, a non-zero "discount rate" on the value of
human life is implied, meaning that the present value of life
after some period of time is negligible and should not be
included in the health effects analysis.
Conversely, a zero
discount rate would imply that no health effect should be discounted regardless of whether it occurred today or millions
of years in the future.
The ethical arguments surrounding
this issue will not be addressed here, but are fundamentally
important, nevertheless.
-88-
Coal Impacts
Returning to the discussion on coal, it is noted that
stack gas emissions and radon effluents from flyash
storage piles are but two of several exposure pathways to
man.
Three other pathways are:
mine runoff, flyash runoff
and the use of flyash in construction materials
cement and cement block).
(concrete,
Using EPA published data 5 4 '
55
Table 42 has been constructed to show the projected number
of health effects per GWe-year.
According to these estimates, the radiological health
effects of a coal-fired power plant amount to 140 per GWeyr.
This figure seems very large;
it is two orders of
magnitude larger than those estimated for a typical PWR.
It should be noted, however, that the largest contributor
(%130 health effects per GWe-yr)
construction materials.
is the use of flyash in
Neglecting this, the next largest
contributor is stack gas emissions
GWe-yr),
BWR
(12
health effects per
which is the same order of magnitude as a typical
(9 health effects per GWe-yr),
though still ten times
Lastly, the health effects due to the
larger than the PWR.
radon effluents from the flyash storage piles
(0.62) is
nearly identical to those from radon from the uranium
tailings piles
(0.74).
"typical" coal use;
These figures are estimated for
figures would be higher for Western
coal and lower for Eastern coal.
-89-
Table 42:
EPA Estimated Population Dose Commitments and
Health Effects from Coal Fired Power Stations
Per GWe-Year.
Fuel Cycle
Component
Projected Health
Population Dose
Effects
(person-rem x 106 per GWe-yr)
Stack effluents
556,000
11.65
Mine runoff and
acid drainage
1,850
0.06
Runoff from flyash
storage areas
not available
Flyash in construction materials
2,123,000
129.20
615,000
0.62
Radon effluents from
flyash piles
TOTALS
Source:
%3,300,000
not available
%140
Radiological Quality of the Environment in the U.S.,
1977. EPA-520/1-77-009, Table 3-10, pp. 93,
September 1977. All data has been adjusted from
1,000 MWe power plant to 1 GWe-yr by dividing by
utilization factor of 0.65. Ranges have been
expressed as log mean averages. Health effects
include 50 percent fatal and 50 percent non-fatal
at conversion rate of 16 per 106 person-rem.
-90-
To repeat, however, these exposure levels for coal seem
to be very high--so high, in fact, that one should question
whether the data in the EPA source document have been
accurately presented.
If they have been, coal burning,
together with the practice of using flyash in construction
materials, would appear to represent a significantly greater
radiological risk to the general public than nuclear
power--by perhaps as much as two orders of magnitude!
-91-
CONCLUSION
As part conclusion and part perspective, Table 43 is
presented showing the population exposures to the U.S. population from all sources as estimated by the U.S. Environmental Protection Agency.
exposure are:
The three largest categories of
natural background, medical and dental x-
rays and radio-pharmaceuticals.
These three sources
represent more than 90 percent of all radiation exposures
as measured in person-rems.
Surprisingly, natural gas use accounts for the fourth
largest exposure source--more than 2,700,000 person-rems
per year.
The total fuel cycle for coal-fired power stations
would appear to present, based on the discussion of the
preceeding section, a very hazardous situation.
A single
GWe-year of power is said by EPA to cause a 3,300,000
person-rem exposure, resulting in 140 health effects per
year!
Given that this figure is so large, it is presumed
that some error has been made in either the source document
or its interpretation.
For this reason, no exposure level
is given for coal in Table 43.
The use of liquified petroleum gas (LPG) is estimated
by EPA to result in a population exposure of about 70,000
person-rems.
-92-
Table 43:
Estimated U.S. Exposures to
All Sources
Category of U.S.
Population Exposure
1. Ambient exposures
a. cosmic radiation
b. Th-232 series
c. K-40 series
d. U-238 series
2.
Medical and dental
x-rays
3.
Radio-pharmaceuticals
Technologically enhanced
natural radiation
a. Rn-222 in natural gas
b. Coal fired power stations
c. Inactive uranium mines
d. Rn-222 in LPG
e. Oilfired power stations
U.S. Exposures
(Breakdown
Person-rems
per year
Radiation from
U.S. Exposures
Person-rems
per year
Estimated
U.S. Cancer
Deaths per
year
19,800,000
9,700,000
4,600,000
3,100,000
2,400,000
19,800,000
3,960.
14,800,000
14,800,000
2,960.
3,300,000
3,300,000
660.
3,000,000
2,730,000
(see text)
70,000
30,000
15
3,000,000
600.
400,000
400,000
80.
45,000
21,000
24,000
45,000
9.0
4.
5.
Fallout
Uranium fuel cycle (1976)
a. U.S. population only
b. occupational only
6.
7.
Other occupational
4,400
4,400
0.9
8.
Consumer products
6,100
6,100
1.2
TOTALS
Source:
NA
41,355,500
8,270.
"Radiological Quality of the Environment in the U.S., 1977,"
EPA-520/1-77-009, US EPA, September 1977; and the GESMO Final
Report, NUREG-0002, Volume 105, USNRC, September 1976
-93-
The nuclear fuel cycle was estimated by EPA in the
original of Table 43 to result in about 2000 person-rem of
population exposure per year given current levels of
deployment.
These data have been replaced in Table 43 by
higher estimates consistent with the information developed
in this paper.
While it is often said that population exposures resulting from energy technologies are negligible compared to background, this is apparently not true for natural gas use in
general, nor does it appear to be true for large scale deployment
of coal and nuclear power stations.
Further, any statement
which gives a specific number for the expected exposure
level resulting from a single GWe-year of power production,
presumes a specific and limited period of time over which
one "counts" the continuously accumulating total exposure;
otherwise, such a number would be very large, indeed.
This
period of time appears to have been arbitrarily chosen in
the studies reviewed by this paper.
On a more optimistic note, however, recent rulings by
the Nuclear Regulatory Commission indicate that mill tailings piles in the future may release only a small fraction
of that released by the exposed piles of today, due to
stringent stabilization methods, and the release rates from
many of the newer reactors, including BWR's, show continuing
improvements.
-94-
The use of radium-bearing flyash in construction
materials would seem to be inconsistant with the health
standards imposed on the nuclear fuel cycle, and this
practice may be banned.
Flyash accumulation and the stack
releases, however, would remain as problems to be solved
for coal, not to mention the non-radiological health
effects associated with emissions of the oxides of nitrogen,
sulfur and carbon, and of particulates.
-95-
Notes and References
1.
Statistical Yearbooks, 1900 to present, Edison Electric
Institute.
2.
An assessment by Dr. David O. Wood, MIT Energy Lab, of
current econometric models forecasting electricity prices
vs. the cost of fossil fuels and the consequent implications,
November 1978.
3.
For example, the MIT version of the Baughman-Joskow Regional
Electricity model, April 1978.
4.
Giraud, Andre, "World Energy Resources," appearing in
Transactions: Proceedings of the Plenary Sessions, International Conference on World Nuclear Energy, November 1976,
pp. 1.7-23.
5.
Wilson, Carroll L.,
McGraw-Hill, 1977.
6.
Giraud, op. cit., pp. 21.
7.
GESMO Final Report; "Final Generic Environmental Statement
on the Use of Recycle Plutonium in Mixed Oxide Fuel in
Light: Water Cooled Reactors," (GESMO), NUREG-0002, Volumes
1 through 5, September 1976.
8.
GESMO Final Report, Vol.3, Chapter IV, pp. IV-F-1.
9.
"Conventional ores"
are 800 to 1900 ppm and above, GESMO
Final Report, Vol. 3, Chapter IV, pp. IV-F-10.
Energy:
Global Prospects 1985-2000,
10.
"Radiological Quality of the Environment in the US, 1977,"
EPA-520/1-77-009, US Environmental Protection Agency (EPA),
September 1977, pp. 141.
11.
"Statistical Data on the Uranium Mining Industry," GJO-100
(75), ERDA, Grand Junction Office, January 1, 1975, pp. 24
and 27.
12.
"Environmental Analysis of the Uranium Fuel Cycle, Part I-Fuel Supply," Fuel Cycle Part I, EPA-520/9-73-003-B,
US EPA, October 1973, pp 57.
13.
Goodwin, Ansel, "Consequences of Effluent Releases," Nuclear
Safety, Vol. 14, No. 6, Nov-Dec 1973, pp. 643-650.
14.
GESMO Final Report, Summary Volume, Table 5(A)-l, Vol. 1,
pp. 5(4)-3.
-96-
15.
GESMO Final Report, Chapter IV, Table IV-J(5)-9, Vol. 3,
pp. IV-J(E)-9.
16.
EPA, Fuel Cycle Part I, op. cit., pp. 23.
17.
Draft Environmenal Statement for the Lucky Mc Uranium
Mill, Freemont County, Wyoming, Docket No. 40-2259, prepared by the Nuclear Regulatory Commission, Washington,
DC, June 1977, pp. 3-1 through 3-1.
18.
GESMO, Final Report, Chapter IV, VOl1. 3, pp. IV-F-23.
19.
Lucky Mc, op. cit., pp. 3-13.
20.
EPA, Fuel Cycle Part I, op. cit., pp. 54.
21.
Ibid., pp. 73.
22.
Lucky Mc, op. cit., pp. 3-15 and 3-16.
23.
EPA, Fuel Cycle Part I, op. cit., pp. 74.
24.
GESMO Final Report, Chapter IV, Vol. 3, pp. IV-F-33.
25.
EPA, Fuel Cycle, Part I, op. cit., pp. 80.
26.
The velocity distribution for gas molecules is characterized
by a Maxwellian distribution, for which the average kinetic
energy is given by E = 1.5(kT) where k is Boltzmann's
constant. Hence, the average speed of the gas molecules
is given by v =
3kT/m.
27.
GESMO Final Report, op. cit., pp. IV-F-44.
28.
Ibid., pp. IV-F-44.
29.
Ibid., pp. IV-F-45 and Table IV-J(E)-17 (after correction
of typographic error), pp. IV-J(E)-17.
30.
Goodyear Atomic Coropration, "Portsmouth Gaseous Diffusion
Plant Environmental Monitoring Report for Calendar Year
1976, August 1977, pp. 14. Microfisch, ERDA-77-104/2,
pp. 1152.
31.
GESMO Final Report, op. cit., pp. IV-F-54.
32.
GESMO Final Report, op. cit., pp. IV-F-17.
33.
EPA, Fuel Cycle, Part I, op. cit., pp. 114.
34.
Ibid., pp. 114.
35.
GESMO Final Report, op. cit., Table IV-J(E)-17, pp. IV-J(E)-17.
-97-
36.
EPA, Final Cycle, Part I, op. cit., pp. 122.
37.
Ibid., Table 5-4, pp. 123.
38.
Brown's Ferry FSAR as taken from Table 5 in "Environmental
Analysis of the Uranium Fuel Cycle, Part II--Nuclear Power
Reactors," US EPA, EPA-520/9-73-003-C, November 1973, pp.
23.
39.
Ibid., pp. 23
40.
"Summary of Radioactivity Released in Effluents from Nuclear
Power Plants from 1973 through 1976," EPA-520/3-77-012,
US Environmental Protection Agency, December 1977.
41.
"Environmental Statement (Final) for the Fort St. Vrain
Nuclear Generating Station," prepared by the Directorate
of Licensing, AEC, Docket No. 50-267-51, August 1972, pp.
III-23.
42.
Ibid., pp. III-26.
43.
Ibid., pp. III-28 through III-33.
44.
Ibid., Table III-5, pp. III-33.
45.
"Radioactive Effluents from Nuclear Power Stations in
Europe, 1970 - 1974," appearing in Nuclear Safety, Vol.
18, No. 3, May-June 1977, pp. 370-380.
46.
Barry, P.J. and Marko, A.M., "Release of Radionuclides to
the Environment from CANDU-Type Reactors--A Summary of
Canadian Experience," an article appearing in an AECL
report on "Some Aspects of the Release of Radioactivity
and Heat to the Environment from Nuclear Reactors in
Canada," AECL-4156, Atomic Energy of Canada Limited, April
1972, pp. 12-18.
47.
Ibid., pp. 14.
48.
Preliminary Safety Analysis Report (PSAR) for the Clinch
River Breeder Reactor Plant, Vol. 7, Section 11, Project
Management Corp., undated, pp. 11.1-1.
49.
Environmental Report (ER) for the Clinch River Breeder
Reactor Plant, Vol. 2, Section 5, Project Management
Corp., undated, pp. 5.2-9 through 5.2-12.
50.
GESMO Final Report, op. cit., Table 5(A)-l, Summary Volume,
pp. 5(A)-3.
-98-
51.
Johnson, R.H., Jr., Bernhardt, D.E., Nelson, N.S.,
Calley, H.W., Jr., Assessment of Potential Radiological
Health Effects from Radon in Natural Gas, EPA-520/1-73004, 1973.
52.
Natural Radioactivity Contamination Problems, EPA-520/
4-77, February 1978, pp. 31.
53.
Vine, J.D., Uranium Bearing Coal in the US, Geological
Survey Professional Paper 300, 1956.
54.
Boothe, G.F., An Evaluation of the Radiological Aspects
of the Proposed Pioneer Coal-Fired Plant (Pioneer
near Boise, Idaho), Department of Health and Welfare,
State of Idaho, 1976.