BRH/DER 70-2
RADIOACTIVE WASTE DISCHARGES
TO THE ENVIRONMENT
FROM
NUCLEAR POWER FACILITIES
U.S. DEPARTMENT OF HEALTH, EDUCATION, AND WELFARE
Public Health Service
ENVIRONMENTAL HEALTH SERVICE
-------
TECHNICAL REPORTS
Technical reports of the Division of Environmental Radiation,
Bureau of Radiological Health are available from the Clearinghouse
for Federal Scientific and Technical Information, Springfield, Virginia
22151. Price is $3.00 for paper copy and $0,65 for microfiche. The
PB number, if indicated, should be cited when ordering.
BRH/DER 69-1 Evaluation of Radon 222 Near Uranium Tailings
Piles (PB 188 691)
BRH/DER 70-1
Radiological Surveillance Studies at a Boiling
Water Nuclear Power Station
-------
BRH/DER 70-2
RADIOACTIVE WASTE DISCHARGES
TO THE ENVIRONMENT
FROM
NUCLEAR POWER FACILITIES
Joe E. Logsdon
and
Robert I. Chissler
Div ision of Environmental Radiation
MARCH 1970
U.S. DEPARTMENT OF HEALTH, EDUCATION, AND WELFARE
Public Health Service
Environmental Health Service
Bureau of Radiological Health
Rockville, Maryland 20852
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FOREWORD
The Bureau of Radiological Health carries out a national program
designed to evaluate the exposure of man to ionizing and non-ionizing
radiation and to promote development of controls necessary to protect
the public health and safety.
Within the Bureau, the Division of Environmental Radiation conducts
programs relating to 1) public health evaluation of planned and oper-
ating nuclear facilities, 2) field studies at operating nuclear
facilities to develop environmental surveillance technology, and
3) a system of national radiation surveillance projects to evaluate
population exposure from all sources of environmental radioactivity.
The Bureau publishes its findings in the monthly publication
Radiological Health Data and Reports, Public Health Service numbered
reports, appropriate scientific journals, and Division technical
reports.
The technical reports of the Division of Environmental Radiation
allow comprehensive and rapid publishing of the results of intramural
and contract projects. The reports are distributed to State and
local radiological health program personnel, Bureau technical staff,
Bureau advisory committee members, university personnel, libraries
and information services, industry, hospitals, laboratories, schools,
the press, and other interested groups and individuals. These reports
are also included in the collections of the Library of Congress and
the Clearinghouse for Federal Scientific and Technical Information.
I encourage the readers of these reports to inform the Bureau of
any omissions or errors. Your additional comments or requests for
further information are also so
0
Tohn C. Villforth
)irector
Bureau of Radiological Health
iii
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PREFACE
The discharges of radioactivity to the environment from nuclear
power stations contribute to the radiation dose received by the
general population. The Bureau of Radiological Health provides
guidance and recommendations to health agencies for the development
of environmental surveillance programs related to nuclear facilities.
In order to provide a better technical basis for surveillance
recommendations, radiological data relative to discharges of radio-
activity from nuclear power facilities have been compiled.
This report summarizes discharges of radioactive material to the
environment from nine selected nuclear power facilities and relates
the discharges to power produced and plant maintenance operations.
The facilities included in this report represent three basic reactor
types: pressurized water, boiling water, and high temperature gas.
The operating facilities which are not included in this report either
represent unique designs not being constructed in the present genera-
tion of power plants or they are smaller plants similar in design to
those which are included.
Appreciation is expressed to the nuclear facility operators and
the Atomic Energy Commission's Division of Regulation and Division of
Naval Reactors who provided comments and suggestions on the draft
version of this report.
Charles L. Weaver, Director
Division of Environmental Radiation
v
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CONTENTS
Page
FOREWORD ii
PREFACE iii
INTRODUCTION 1
Administrative Controls ........ 1
SOURCES OF LIQUID AND GASEOUS WASTE 2
OPERATING EXPERIENCE ... 3
Liquid Discharges ... 6
Gaseous Wastes 14
SUMMARY 22
TABLES
1. Waste Processing Capability at Operating Nuclear
Power Facilities . ........... 4
2. General Information for Facilities Included in
This Report 5
3. Total Annual Liquid Waste Discharged 7
4. Annual Average Liquid Radioactive Waste Discharge
Concentrations Expressed as Percent of Limit ..... 8
5. Comparison of Radioactive Waste Discharges to
Electrical Power Generation ..... 10
6. Total Annual Liquid Tritium Discharges (Curies) ... 12
7. 1968 Tritium Discharges in Liquid Wastes Compared to
AEC Licensed Limits 13
8. Total Annual Gaseous Waste Discharged . 15
9. Releases of Halogens and Particulates from Power
Reactors in Gaseous Effluents ............ 16
• •
vi r
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CONTENTS
Page
10. Annual Gaseous Radioactive Waste Discharges
Expressed as Percent of Limit 17
FIGURES
1. Comparison of Annual Liquid Radioactive
Waste Discharged to Annual Electrical
Power Generation 11
2. Comparison of Annual Gaseous Radioactive
Waste Discharged to Annual Electrical Power
Generation for BWR Facilities 19
3. Comparison of Annual Gaseous Radioactive
Waste Discharged to Annual Electrical Power
Generation for PWR Facilities 20
A. Comparison of Annual Gaseous Radioactive
Waste Discharged to Annual Electrical Power
Generation for Indian Point Station, Unit 1 21
APPENDICES
I. Big Rock Point Nuclear Power Station 24
11, Connecticut Yankee Atomic Power Plant 30
III. Dresden Nuclear Power Station Unit 1 37
IV. Humboldt Bay Plant Unit 3 46
V. Indian Point Station Unit 1 51
VI. Peach Bottom Atomic Power Station 57
VII. San Onofre Nuclear Generating Station 62
VIII. Shippingport Nuclear Power Station 66
IX. Yankee Atomic Power Station .. 69
REFERENCES 75
ix
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RADIOACTIVE WASTE DISCHARGES TO THE ENVIRONMENT
FROM NUCLEAR POWER FACILITIES
INTRODUCTION
Nuclear power plants produce large quantities of radioactive
material as a by-product of operation of the reactor. Most of the
radioactive material is contained within the fuel elements and remains
there until the fuel is chemically reprocessed at a fuel reprocessing
facility. The relatively small portions of radioactive material that
escape from the fuel or are produced outside the fuel are contained and
processed as radioactive waste at the nuclear power plant. This material
is concentrated and converted to solid form and shipped offsite for
burial at licensed burial sites. The very small amounts of residual
material are discharged to the atmosphere or hydrosphere. The quantities
and types of waste discharged vary from facility to facility depending
primarily on design characteristics of the plant and on waste manage-
ment practices.
Measurements of radioactivity in the environs of nuclear power
plants made by State health departments, nuclear power facility operators,
and the Bureau of Radiological Health, have in most cases revealed little
or no increase in environmental radioactivity resulting from plant
operations. In those cases where increases were measured, the levels
were barely detectable. However, in consideration of the rapid expan-
sion of the nuclear power industry, it is incumbent upon health agencies
and the Bureau of Radiological Health to continually review radioactive
waste disposal practices and to evaluate their potential effect on the
environment.
The purpose of this report is to provide information concerning
discharges of radioactive material to the environment by nuclear power
facilities. Nuclear power facility operators routinely prepare
operating reports which contain information concerning discharges of
radioactive material to the environment. These reports are used as
the principal source of data for this report.
Administrative Controls
Discharge of radioactive waste to the environment by nuclear power
facilities is regulated by the Atomic Energy Commission (AEC) through
an operating license issued to the nuclear facility operator. The license
requires the licensee to operate the plant in accordance with written
Technical Specifications that have been approved by the AEC which include,
among other items, limits for radioactive liquid and gaseous discharges.
Discharge limits presented in the Technical Specifications are based on
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2
effluents. ("Exclusion area" means that area surrounding the reactor
in which the licensee has the authority to determine all activities including
exclusion or removal of personnel and property from the area). Discharge
limits may further be reduced by the AEC to compensate for possible re-
concentration of radionuclides by environmental media. For example, the gaseous
discharge limits now being applied in the licensing proces^ to ^"^1 releases
are reduced by a factor of 700 to compensate for possible reconcentration
through the pasture-cow-milk exposure pathway.
SOURCES OF LIQUID AND GASEOUS WASTE
Radioactivity at nuclear power facilities is produced primarily as a
by-product of the fission process or from neutron activation of structural
material within the pressure vessel and impurities in the primary coolant.
A combination of leakage of fission products through the fuel cladding into
the primary coolant and activation of materials outside the fuel makes the
primary coolant the principal source of liquid and gaseous wastes. However,
leakage of primary coolant into other systems and various plant operations
cause the sources to be numerous. Typical plant operations which result in
liquid
or gaseous radioactive waste include;
1.
Refueling and maintenance
2.
Control of primary coolant chemistry
3.
Sampling
4.
Rejection of non-condensable gases from steam condensers
5.
Blowdown of steam generators
6.
Expansion water when the plant goes from a cold to a hot
operation
7.
Decontamination of clothing, components, tools, and surfaces
8.
Regeneration of demineralizer resins
The activation of impurities in systems other than the primary
coolant has not been a major source of liquid radioactive wastes in light
water reactors. However, at Peach Bottom Nuclear Power Station, a high
temperature gas-cooled reactor, the absence of significant quantities of
liquid radioactive wastes from other sources makes the primary shield
cooling water the principal source of liquid wastes at this reactor.
Measurements of concentrations of specific radionuclides present
in gaseous and liquid wastes are not generally available from nuclear
power plants. Normally, facility operators report only gross beta-
gamma activity and sometimes tritium activity in liquid wastes. Gaseous
waste discharges are generally categorized and reported by facility
operators as being either halogens and particulates or activation and
noble gases. Some facility operating reports include results of specific
radionuclide analyses of primary coolant. However, the relative abun-
dance of radionuclides in the primary coolant may be different than the
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3
relative abundance in liquid or gaseous waste effluents. Relative
abundances of radionuclides in the primary coolant are functions of:
1. Cladding leakage
2. Temperature changes which may cause release of particles
that have been attached to surface of the primary system
3. Use and effectiveness of coolant purification
4. Rate of primary system leakage
5. Chemical additives in the primary coolant
6. Type of coolant
7. Power history
Relative abundances of radionuclides in the waste effluents are primary
functions of:
1. Their abundance in the primary coolant
2. Their respective half-lives
3. Design of the radwaste treatment system
4. Waste treatment practices
Waste treatment capabilities at selected operating nuclear facilities
are summarized in Table 1.
Most radionuclides can be classified as either fission products or
activation products. Tritium (^H) however, is a special case in that
it is produced both from fissioning and from neutron activation. It is
also special because it is not affected by methods presently utilized
in processing radioactive wastes. Therefore, the tritium released to
the primary coolant or produced in the primary coolant is ultimately
discharged to the environment in either liquid or gaseous form.
Additional information on environmental tritium contamination from
nuclear energy sources is provided in Reference 2.
OPERATING EXPERIENCE
Most experience on radioactive waste discharge to date has been with
pressurized water reactors (PWR) and boiling water reactors (BWR).
References 3 and 4 provide descriptions of these types of facilities.
Limited operating experience has been gained from a gas-cooled reactor
through the operation of Peach Bottom-1. Reference 5 describes this type of
facility. Table 2 provides general information for facilities included
in this report.
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TABLE 1
HASTE PROCESSING CAPABILITY AT OPERATING NUCLEAR POWER FACILITIES
GASEOUS WASTE TREATMENT
LIQUID WASTE TREATMENT
REACTOR
HOLDUP
CAPACITY
PARTICULATE
TREATMENT
IODINE
TREATMENT
FWRs
Shippingport
60 days
none
none
gas scrubbing, evaporation,
demineralization, filtration
Yankee
60 days
none
none
gas scrubbing, evaporation,
demineralization
Indian Point-1
120 days
absolute filters
none
evaporation, demineralization,
gas stripping, filtration
San Onofre
60 days
none
none
demineralization
Conn. Yankee
variable
fiberglass filter
none
evaporation, demineralization
BWRs
Dresden-1
20 min.
absolute filters
none
filtration, evaporation,
demineralization
Big Rock Point
30 min.
absolute filters
none
filtration, evaporation,
demineralization
Humboldt Bay
HTGR
18 min. design
40 min. actual
absolute filters
none
filtration, demineralization
Peach Bottom-1
variable
filtration
charcoal filters
demineralization
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TABLE 2
GENERAL INFORMATION FOR FACILITIES INCLUDED IN THIS REPORT
AEC
Power Level^
Stack
Stack
Exhaust
Condenser
Water for
Dilution
Body of Water
Facility
Docket
No.
MWt
MWe
Net
Location
Height
(Feet)
Rate
(CPM)
Flow Rate
(GPM)
Receiving
Liquid Waste
PWRs
Shippingport
None
505
90
Shippingport, Pa.
26a
9,000
114,000
Ohio River
Yankee
50-29
600
175
Rowe, Mass.
150
15,000
138,000
Deerfield River
Indian Point-1
50-3
615
265
Buchanan, N.Y.
400
280,000
300,000
Hudson River
San Onofre
50-206
1,347
430
San Clemente, Calif,
100
40,000
350,000
Pacific Ocean
Conn. Yankee
50-213
1,825
573
Haddam Neck, Conn.
175
70,000
372,000
Connecticut River
BWRs
Dresden-1
50-10
700
200
Morris, 111.
300
45,000
166,000
Illinois River
Big Rock Point
50-155
240
71
Charlevoix, Mich.
240
30,000
50,000
Lake Michigan
Humboldt Bay
HTGR
50-133
240
68
Eureka, Calif.
250
12,000
100,000b
Humboldt Bay
(Pacific Ocean)
Peach Bottom-1
50-171
115
40
Peach Bottom, Pa.
150
20,000
43,000
Susquehanna River
aGas discharge stack; vapor container exhaust stack 116 feet.
bFlow rate for Humboldt Bay Unit 3 is 51,800 gpm. All calculations are based on a flow rate of 100,000 gpm
which is the combined flow rate for Humboldt Bay Units 1, 2, and 3. Units 1 and 2 are fossil fuel plants.
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6
Liquid Discharges
Quantities of gross beta-gamma activity (less tritium) discharged
annually by each facility are shown in Table 3. Operating reports for
Dresden-1 did not indicate total amount of radioactivity discharged,
but gave an average contribution to the radioactivity in the condenser
cooling water discharge canal. Except as noted in the table, the
total annual discharge for Dresden was obtained by multiplying the
facility's contribution to the concentration of radioactivity in the
condenser cooling discharge canal times the annual flow rate of the
canal as calculated from Table 2,
Calculations have been made based on data in Tables 2 and 3 and
the liquid discharge limits for each facility to provide comparisons
to discharge limits. These comparisons are provided in Table 4.
These data do not include tritium which is presented later in Tables
6 and 7.
With the exception of Shippingport,* discharge limits are pre-
scribed by Technical Specifications. All Technical Specifications
limit concentrations in liquid effluents to those listed in Appendix B,
Table II of 10CFR20. Without analysis for specific radionuclides, the
limit is considered to be 10"? juCi/ml. If liquid wastes are analyzed
for specific radionuclides, discharge limits can be based on the
maximum permissible concentration for the radionuclides present. These
limits normally are higher than the limit for unidentified radionuclides.
Most nuclear power plants discharge sufficiently small quantities of
radioactivity in liquid wastes that dilution factors associated with
the condenser cooling canal are sufficient to permit discharge on the
basis of unidentified radionuclides. In most cases, there is no
requirement for reporting radionuclide analyses of wastes in operating
reports and as a result they are not normally reported. Therefore,
the limit for unidentified radionuclides has been used as the basis
for comparison in Table 4 except as noted.
It should be noted that the use of a limit for a mixture including
unidentified nuclides which is adequate to show compliance, involves
the arbitrary assignment of all the activity present to the most
restrictive nuclide present. Liquid wastes necessarily involve a
mixture of many fission and corrosion products. The resulting "percent
of the discharge limit" therefore is artificially high. Limits based
~Shippingport has been developed and operated under AEC sponsorship.
Shippingport radioactivity discharge limits are equal to or less than
radiation protection standards set forth in Title 10, Code of Federal
Regulations, Part 20, AEC Manual Chapter 0524 and a waste discharge
permit from the Pennsylvania Sanitary Water Board.
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TABLE 3
TOTAL ANNUAL LIQUID WASTE DISCHARGED3
GROSS p-7 LESS TRITIUM
(CURIES)
PWRs
1959
1960
1961
1962
1963
1964
1965
1966
1967
1968
Sh ipp ingport ^
0.083
0.21
0.129
0.09
0.19
0.53
0.14
0.06
0.07
0.08
Yankee
0.008
0.008
0.003
0.002
0.029
0.036
0.055
0.008
Indian Point-1
0.130
0.164
13.0
26.3
43.7C
28.0d
34.6d
San Onofre
0.32
1.6
Conn. Yankee
0.216
3.9
BWRs
Dresden-1
0.770
2.095
2.61
2.78
3.82
8.7
11.5
4.3d
6.1
Big Rock Pointe
0.2
0.63
6.22
2.80
6.12
10.1
7.9
Humboldt Bay
0.397
0.664
1.89
2.10
3.13
3.2
HTGR
Peach Bottom-1
0.0017
0.0004
Based on operators' reports except as noted.
TData taken from Radiological Health Data and Reports.
cData from Reference 31.
dData taken from Reference 8.
eData taken from Reference 32.
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00
TABLE 4
ANNUAL AVERAGE LIQUID RADIOACTIVE WASTE DISCHARGE CONCENTRATIONS
EXPRESSED AS PERCENT OF LIMIT a
PERCENT OF DISCHARGE LIMIT EXCLUSIVE OF TRITIUM
1959
1960
1961
1962
1963
1964
1965
1966
1967
1968
PWRs
Shippingport
0.37
0.93
0.57
0.40
0.84
2.34
0.62
0.27
0.31
0.35
Yankee
0.03
0.03
0.01
0.007
0.1
0.13
0.02b
0.03
Indian Point-1
0.22
.26
22
43
70.1
1.55c
1.65e
San Onofre
0.46
2.35
Conn. Yankee
0.01°
5.35
EWRs
Dresden-1
2.33
6.34
7.9
8.42
11.6
26.4
34.8
13.0
18.5
Big Rock Point
4.46
14.2
54.4
42.4
62.5
50.4d
82.3f
Humboldt Bay
2.77
3.37
9.52
12.15
16.89
19.7
HTGR
Peach Bottom-1
0.02
0.005
aPercent of limit calculations were based on the following: (1) 10CFR20 limit for unidentified radionuclides
in water of 10"? juCi/ml except as noted, (2) average flow rates in the discharge canals equal to those given
in Table 2.
^Concentration limit 10x10"^ juCi/ml, based on radionuclide analysis. Concentration limit 35x10"^ jiCi/ml,
based on radionuclide analysis.
Concentration limit 30x10"^ yCi/ml, based on radionuclide analysis.
Concentration limit 1.5x10 mCi/ml,
^Concentration limit 2x10"^ uCi/ml, based on radionuclide analysis. based on radionuclide analysis.
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9
on complete analysis, if performed, would be expected to be substantially
higher than those used; and the percentages in Table 4 would be substan-
tially less.
The data in Table 3 are further expanded in Table 5 to show the
ratio of gross beta-gamma radioactivity discharged to power produced
for each facility. These data indicate that with the exception of
Indian Point-1, BWR's have discharged more gross beta-gamma radioactivity
in liquid waste per unit of power produced than other types. There is
nothing inherent in a BWR which would cause it to produce more radio-
activity in liquid waste than would be produced in PWR. These higher
quantities from BWRs are believed to be primarily a result of fuel
cladding leakage.
Figure 1 compares annual discharges of radioactivity in liquid
wastes to power generation for facilities included in this report.
This figure indicates that both electrical generation and liquid waste
discharges are increasing with time, but that waste discharges are not
increasing as rapidly as electrical generation.
Discharges of tritium in liquid wastes are generally reported
separately from gross beta-gamma discharges primarily because of its
high relative abundance in liquid wastes and its relatively high dis-
charge concentration limit as compared to the concentration limit for
unidentified radionuclides. Table 6 provides a summary of available
data concerning tritium in liquid waste discharges. The number of curies
of tritium discharged in liquid waste is high relative to the number of
curies discharged of other radionuclides. However, the relative hazard
per curie of tritium is low. The data in Table 6 are expanded in Table
7 to show derived average discharge concentrations and percent of the
discharge limit for 1968. By compt^ing the data in Tables 4 and 7, it is
evident that concentrations of gross beta-gamma activity (exclusive of
tritium) in liquid discharges more nearly approach the limit used than do
concentrations of tritium. This is significant because current methods for
treatment of liquid wastes are ineffective in reducing quantities of tritium
discharged.
Table 5 provides a ratio of curies of tritium discharged in liquid
waste to electrical power produced by each facility. The data show that
PWRs discharge much higher quantities than the other two types. Higher
tritium concentrations in PWRs are due in part to neutron reactions with
boron which is added in the form of boric acid to the primary coolant.
Since the boron is in solution with the primary coolant, there is no
cladding barrier to retain the tritium so produced. This is not the case
with a BWR where the boron is used in the form of cladded plates or curtains.
Other sources of tritium which are common to both PWRs and BWRs include
fission product tritium and reactions with lithium, nitrogen, helium-3, poison
material used in control rods or plates, and reactions with structural material.
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TABLE 5
COMPARISON OF RADIOACTIVE WASTE DISCHARGES 10 ELECTRICAL POWER GENERATION
Facility &
Reactor Type
Period
Covered®
Total Waste
Discharges During Period
Total Gross
yCi Discharges/MWe-
•hr Gross
LtQuid
Gaseous
Electrical
Liquid
Gaseous
Gross p-70
(Curies)
Tritium
(Curies)
Gross (3-7
(Curies)
Generation
(MWe-hr)
Gross (3-7
Less 3h
Tritium
Gross P-7
PWRg
Shippingport
1959-68
1.6
281
0.58
3.5x106
0.46
80
0.17
(1968)
(0.08)
(35.2)
(0.001)
(4.1x105)
(0,20)
(86)
(0.002)
Yankee
1961-68
0.15
6,080°
37
8.9x106
0.02
1,220c
4.16
(1968)
(0.008)
(1,170)
(0.68)
(1.2x106)
(0.007)
(950)
(0.57)
Indian Point-1
1964-68
112
1,080^
166
6.2x10®
17.1
319d
26.77
(1968)
(34.6)
(787)
(59.6)
(1.6x106)
(21.6)
(492)
(37.2)
San Onofre
1967-68
1.92
8.8
1.7x10®
1.1
5.18
(1968)
(1.6)
(2,350)
(4.83)
(1.4x106)
(1.1)
(1,680)
(3.45)
Conn. Yankee
1967-68
4.1
1,960
3.75
3.7x10®
1.1
530
1.01
(1968)
(3.9)
(1,740)
(3.74)
(3.2x106)
(1.3)
(544)
(1-17)
BWRs
Dresden-1
1961-68
41.9
2.8x10^
7.6x106
5.5
3.68x105
(1968)
(6.1)
(2.9)
(2.4x105)
(9.7x105)
(6.3)
(3)
(2.47x105)
Big Rock Point
1962-68
33.1
1.33xl06
1.7x10®
19.5
7.82x105
(1968)
(7.5)
(34) e
(2.32x105)
(4.5x105)
(17.6)
(76)
(5.16x105)
Humboldt Bay^
1963-68
11.4
< 248
2.23x10®
1.8x10®
6.3
< 138
1.24x10®
(1968)
(3.20)
< (6.6)
(8.53x105)
(4.7x105)
(6.9)
(15)
(1.83x10®)
HTGR
Peach Bottom-1
1967-68
0.002
117
3-lxlO5
0.006
377
(1968)
(0.0004)
(109)
(1.5xl05)
(0.003)
(727)
al968 data is in parentheses
''Exclusive of tritium
^Based on data for 1965-68 wherein electrical generation was 4.98x10 MWe-hr.
rlased on data for 1967-68 wherein electrical generation was 3.39x10® MWe-hr.
eBased on an upper limit calculation wherein all liquid waste released during 1968 was assumed to contain as
^njuch tritium as was in primary system water.
Data from P.G.&E. records.
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100
90
80
70
60
50
AO
30
20
FIGURE 1
COMPARISON OF ANNUAL LIQUID RADIOACTIVE WASTE DISCHARGED
TO ANNUAL ELECTRICAL POWER GENERATION®
ELECTRICAL POWER
GENERATION
LIQUID RADIOACTIVE
WASTE DISCHARGED
Gross p-7 less
lxlO7
9xl06
8x106
7xl06
6xl06
5xl06
4x10®
u
>>
u
§
w
H
3
«
O
IJ
3
M
e
3x10® ^
, 2x10®
co
to
§
O
10
1x10®
,—
1964
—j—
1967
—I—
1968
"J—
1965
-I 1—
1 1966
aTotals for all facilities included in this report.
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TABLE 6
TOTAL ANNUAL LIQUID TRITIUM DISCHARGES (CURIES)
FWRs
1958
1959
1960
1961
1962
1963
1964
1965
1966
1967
1968
Shippingporta
50.0
64.0
99.0
13.2
1.33
2.17
1.39
3.04
27.3
34.8
35.2
Yankee''
—
--
--
—
1,300
1,920
1,690
1,170
Indian Point-lc
--
--
—
--
125
297
787
£
San Onofre
--
2,350
Connecticut Yankee*'
221
1,740
BWRs
Dresden-lc
--
—
--
--
—
—
--
--
2.9
Big Rock Point0
--
--
—
—
--
--
34e
Humboldt Bay0
< 214
< 100
< 54
< 60
< 166
< 6.6
HTGR
Peach Bottom-lc
--
Neg.
aData for years 1958 through 1960 taken from Shippingport Operations from Power Operation After First Refueling to
Second Refueling. (Hay 6. 1960 to August 16. 1961). DLCS-36402; data for years 1961 through 1968 taken from
Radiological Health Data and Reports.
''Data taken from Yankee Nuclear Power Station Operations Reports; tritium analysis was not included in the
operating reports prior to 1965.
cData taken from Reference 8.
^Data taken from Connecticut Yankee Atomic Power Company operations reports.
eBased on an upper limit calculation wherein all liquid waste released during 1968 was assumed to contain as much
tritium as was in primary system water.
-------
TABLE 7
1968 TRITIUM DISCHARGES IN LIQUID
WASTES COMPARED TO AEC LIMITS3
Pressurized Water Reactors
Average Discharge*3
Concentration juCi/ml
Percent of Limit
Shippingport
1.6xl0~7
0.0053
Yankee
4.5xl0~6
0.15
Indian Point-1
1.56xl0~6
0.045
San Onofre
3.3xl0"6
0,11
Connecticut Yankee
2.4xl0~6
0.08
RMUnp Water Reactors
Dresden-1
9xl0"9
0.0003
Big Rock Point
3.6xl0~7
0.012
Humboldt Bayc
4.1xl0"8
0.0014
Gas Cooled Reactors
Peach Bottom-1
Negligible
Negligible
aBased on 10CFR20 limit for unrestricted areas of 3x10"^ juCi/cc and dilution
in the condenser cooling water discharge canal.
^Calculated based on annual quantity of tritium discharged and the condenser
cooling water flow rate.
cFrom P.G. & E. records.
-------
14
Gaseous Wastes
Gaseous wastes may be in the form of particulates, volatiles (such as
iodine) or gases. The gases constitute the major portion of discharged
radioactivity via the stack and are generally referred to as activation
and noble gases.
Technical Specifications limit average discharge rates for radio-
active gaseous wastes such that the average annual concentrations at
the plants's exclusion boundary will not exceed those listed in Appendix
B, Table II of 10CFR20. Additional limits (usually a factor of 10 higher
than limits for average release rates) are established for maximum release
rates. Limits for gaseous release rates are a function of the atmospheric
dilution available between the point of release and the exclusion boundary.
Since atmospheric dilution factors are affected by stack height, distance
from stack to exclusion boundary, topography, and local meteorology, gaseous
discharge limits vary widely from facility to facility. Limits for gaseous
releases are usually expressed in (j.Ci/sec with limits for iodines and particu-
lates being relatively more restrictive than the limits for activation and
noble gases. The reason for the more restrictive limits for iodines and
particulates is their potential for reconcentration through environmental
media. For example, the discharge limit for l^lj is generally set at a
factor of 700 below the rate that would produce the 10CFR20 concentration
limit of lxl0~^ |iCi/ml at the exclusion boundary. This is to compensate
for possible reconcentration through the pasture-cow-milk chain. The iso-
topic mixtures of noble gas discharges are such that the maximum permissible
concentration in the environment as calculated from values listed in 10CFR20
range from 3x10"® |j.Ci/ml with decay of less than two hours to 3x10"? p,Ci/ral
at ages of three days and longer.^
Table 8 provides an annual summary of discharges of activation and
noble gases. With the exception of Humboldt Bay, the facilities included
in this report do not include separate discharge rates of halogens and
particulates in their operating reports. However, these data have been
reported by the AEC in Reference 8 for 1967 and 1968 and are reproduced
in Table 9 along with the percent of discharge limit for each facility.
Table 10 provides a summary of annual average discharge rates for
activation and noble gases expressed as percent of discharge limits.
Unlike other facilities, the discharge limit for Shippingport does not
take into consideration atmospheric dilution between the point of
discharge and the exclusion boundary. The discharge limit for Shipping-
port is 3xl0"7 |iCi/ml in the stack. The discharge limit 1.26 |iCi/sec
used in Table 10 for calculating percent of limit is derived based on
a stack discharge rate of 9,000 cfm.
Review of the average discharge rate for each of the facilities listed
in Table 8 reveals that boiling water reactors discharge very much larger
quantities of activation and noble gases than pressurized water reactors.
-------
TABLE 8
TOTAL ANNUAL GASEOUS WASTE DISCHARGED*1
NOBLE AND ACTIVATION GASES
(CURIES)
1959
1960
1961
1962
1963
1964
1965
1966
1967
1968
PWRs
Sh ipp ingport*3
0.014
0.029
0.103
0.012
0.351
0.0024
0.032
0.030
0.002
0.001
Yankee
0.00096
21.700
7.4
0.95
1.7
2.4
2.3
0.68
Indian Point-1
—
0.0072
13.2
33.1
36.4
23.4
59.7
San Onofre
4.02
4.83
Connecticut Yankee
0.021
3.74
BWRs
g
Dresden-1
34,800
284,000
71,600
521,000
610,000
736,000
260,000d
240,000
Big Rock Point e
25.6
803
783
132,000
705,000
264,000
232,000
Humboldt Bay^
716
5,975
197,000
282,000
896,000
853,000
HIGR
Peach Bottom-1
0.00126
7.76
109
fBased on operators* reports except as noted,
"Data from Radiological Health Data and Reports. Data corrected for 1963 and 1964 from Reference 33.
c1961 and 1962 based on maximum rate of noble fission gas activity discharged: 1963-1968 based on average
activity discharge rate for the year while plant was operating.
Data taken from Reference 8.
®Data taken from Ref. 32.
'Data from PG&E records
—-indicates plant was operational but discharge data was not available.
-------
TABLE 9
RELEASES OF HALOGENS AND PARTICULATES FROM POWER REACTORS
IN GASEOUS EFFLUENTS^
1967
1968
Facility
and type
Released
(Curies)
Permissible3
(Curies)
Percent of
Permissible
Released
(Curies)
Percent of
Permissible
PRESSURIZED WATER
REACTORS
Yankee
Negligible
0.03
< 1
Negligible
< 1
Indian Point-1
Negligible
7
< 1
Negligible
< 1
San Onofre
Negligible
0.8
< 1
Negligible
< 1
Conn. Yankee
0.001
0.2
0.5
Negligible
< 1
BOILING WATER
REACTORS
Dresden-1
0.039
100
0.04
0.15
0.15
Big Rock Point
0.25
38
0.66
0.09
0.24
Humboldt Bay
0.64
5.7
11
0.45
8
HIGH TEMPERATURE
GAS COOLED
Peach Bottom-1
Negligible
0.09
-------
TABLE 10
ANNUAL GASEOUS RADIOACTIVE WASTE DISCHARGES EXPRESSED AS PERCENT OF LIMIT3
( NOBLE AND ACTIVATION GASES)
PWRs
1959
1960
1961
1962
1963
1964
1965
1966
1967b
1968b
Shippingport
0.035
0.073
0.26
0.03
0.87
0.006
0.08
0.075
0.005
0.0025
Yankee
0.000014
0.32
0.11
0.014
0.025
0.035
0.036
0.008
Indian Point-1
4.5xl0~7
0.00083
0.0020
0.0022
0.0015
0.0037
San Onofre
0.0024
0.00085
Conn. Yankee
0.00003
0.0039
BWRs
Dresden-1
0.158
1.29
0.32
2.37
2.77
3.34
0.87
1.09
Big Rock Point
0.0025
0.43
2.27
0.85
0.74
Humboldt Bay
0.045
0.38
12.5
17.8
56.7
54.0
HTGR
Peach Bottom-1
6.67xl0"6
0.04
0.087
Percent of limit calculations were based on the following:
Table VIII for annual quantities discharged, (2) discharge
noted. In cases where the discharge limit is expressed as
is used.
(1) values as given in Table II for stack flow rates and
limits presented in Appendices I through IX except as
a factor times MPC and no MPC is given, 3x10"® /iCi/cc
^Limits for 1967 and 1968 from Reference 8 except for Shippingport. Shippingport limits from Reference 10 for
1959 through 1968.
-------
18
Th is trend is further analyzed in Table 5 where the ratio of gaseous
waste discharged to electrical power produced is presented. The
greater quantities of radioactive gases discharged from BWRs are a
result of a shorter hold-up for decay prior to discharge to the environ-
ment. Gases in the primary coolant system of a BWR are carried over
with the steam to the condenser air ejectors where they are immediately
ejected as non-condensables and discharged to the environment with a
hold-up time of 20-30 minutes. The radioactive gases generated in a
PWR are retained for longer periods in the primary coolant system.
Those that are released from the coolant system are stored in tanks
for further decay prior to discharge and therefore PWRs discharge less
short-lived gaseous wastes to the atmosphere. As a result, population
exposure to external radiation from gaseous releases will be higher in
the immediate vicinity of a BWR than in the immediate vicinity of a
PWR. However, since the increased quantities discharged from a BWR are
made up of short-lived radionuclides, the contribution by BWRs to
general population exposures should not be greater than for other types
of reactors.
Figures 2, 3, and 4 provide plots of electrical generation and
gaseous waste discharges as a function of time. The graphs have been
separated into BWRs and PWRs in Figures 2 and 3. Since gaseous waste
discharges at Indian Point-1 have been much higher than at other PWRs,
its data were plotted separately in Figure 4. These figures show
relationships for each year, but there is not enough history to
establish definite trends. Several facilities have been involved in
research programs utilizing the reactor to test different fuels and
cladding, in some cases resulting in significant releases of fission
products from the fuel elements to the primary coolant. Humboldt Bay
has experienced a high percentage of leaking fuel cladding resulting
in relatively large amounts of fission products being released to the
primary coolant. Such releases affect the shapes of the discharge
curves in Figures 2, 3, and 4, and therefore reduce their significance
as far as establishing trends.
The Appendices which follow are in alphabetical order by facility
title and present discussions and data pertaining to each facility.
The primary sources of these data are the operating reports issued by
the facility operators with supplemental data from AEC reports and
from direct correspondence with facility operators. TCie lack of
detailed data pertaining to specific radionuclides in waste discharges
is apparent. The best data of this type is found in Appendix 3, which
is a result of special study^ performed around Dresden-1 by the
Bureau of Radiological Health.
-------
FIGURE 2
COMPARISON OF ANNUAL GASEOUS RADIOACTIVE WASTE DISCHARGED TO
ANNUAL ELECTRICAL POWER GENERATION FOR BWR FACILITIES®
2x10
1.8x10®
1.6xl06
^ 1.4x10®
0A
* 1. 2x10®
G4
M
I 1x10®
2"
*
C/l
w
o
8x10
6xl05
4x10^
2xl05
ELECTRICAL POWER
GENERATION
GASEOUS RADIOACTIVE
WASTE DISCHARGED
—1
1964
1—
1965
—r-
1966
1967
1—
1968
2x10®
1.8x10®
1.6x10®
1.4x10®
1.2x10®
1x10®
8xl05
6xl05
4xl05
2x105
0
aIncludes all BWR facilities listed in Table 5.
-------
10
FIGURE 3
COMPARISON of annual gaseous radioactive waste discharged to
ANNUAL ELECTRICAL POWER GENERATION FOR PWR FACILITIES3
GASEOUS RADIOACTIVE
WASTE DISCHARGED
ELECTRICAL POWER
GENERATION
"I
1964
-T
1965
~\
1966
"1
1967
1968
aIncludes all PWR facilities included in this report with the exception of
Indian Point Station, Unit 1
-------
FIGURE 4
comparison of annual gaseous radioactive waste discharged to annual
ELECTRICAL POWER GENERATION FOR INDIAN POINT STATION, UNIT 1
GASEOUS RADIOACTIVE
WASTE DISCHARGED
1964
1965
—T
1966
ELECTRICAL POWER
GENERATION
"~!—
1967
-T—
1968
7xl06
6xl06
5x10^
4x106
)6
2x106
lxlO6
3x10e
5*.
h
t
&
H
2
W
§
o
~4
3
8
CO
CO
§
o
to
-------
22
SUMMARY
Data pertaining to discharges of radioactive liquid and gaseous
wastes from nine selected operating nuclear power facilities are
presented and discussed. The following summary is based on these data:
1. Experience to date with nuclear power plants has shown that
careful waste management practices, engineered safeguards, and proper
operating procedures generally result in radioactivity levels in
waste effluents of a few percent or less of the AEC's licensed dis-
charge limits. Exceptions are mostly associated with either an
unusually high percentage of leaky fuel elements or with liquid dis-
charge limits which are artificially low as a result of not analyzing
liquid wastes for radionuclide content.
2. Technical Specifications for all facilities limit liquid dis-
charges such that average annual concentrations of radioactivity in the
condenser cooling discharge canal will be less than values listed in
Appendix B, Table II, 10CFR20. The limits for gaseous discharges vary
from facility to facility, depending on available dilution factors in
the atmosphere. They have varied in the manner in which they are
expressed; however, the AEC is in the process of developing uniform
reporting requirements. They also have varied in that the limits for
halogens and particulates for some of the early reactors did not in-
clude the "700" factor to account for possible reconcentration through
environmenta1 med ia.
3. Facility operating reports, which are prepared to demonstrate
that the facility operator is in compliance with specific requirements
of the operating license, vary widely as to units used and types of
information concerning discharges of radioactive wastes. In general,
there is a paucity of information in these reports concerning specific
radionuclides discharged. Information which is available indicates
that relative concentrations of specific radionuclides in waste dis-
charges are not constant. Therefore, assumptions which must be made
(in the absence of data) in order to analyze potential exposure path-
ways to man are not necessarily valid.
The AEC has been developing requirements for isotopic analyses
and for reporting of more detailed data on quantities and concentrations
of specific radionuclides in discharges. The selection of nuclides
for analysis will take into account expected release quantities and
possibilities for reconcentration through environmental media.
4. A number of comparisons have been made of power produced
versus liquid or gaseous waste discharges. The most predominant trends
shown in these comparisons are that boiling water reactors discharge
-------
23
relatively large quantities of gaseous waste and pressurized water
reactors discharge relatively high quantities of tritium in liquid
wastes. No obvious trend is discernible concerning quantities of
waste discharged as a function of power generation. This is to be
expected since fuel cladding integrity and waste treatment practices
are major factors in determining the quantity of waste available for
discharge.
5. Information is available concerning the type of waste treat-
ment facilities installed at each nuclear power station. However, in
operating reports, there is generally no indication as to the types of
waste treatment that were used. A comparison of waste treatment
practices to other parameters such as power history, primary coolant
characterization, and quantities and types of waste discharged, would
be useful in analyzing the possible effectiveness of proposed waste
treatment facilities at nuclear power facilities.
-------
24
APPENDIX I
BIG ROCK POINT NUCLEAR POWER STATION
Big Rock Point Nuclear Power Plant is located on the northeast
shore of Lake Michigan near the city of Charlevoix. It uses a boiling
water reactor of General Electric design with an authorized thermal
power level of 240 MWt which is equivalent to about 71 net electrical
megawatts. The plant attained criticality in September 1962 and first
began producing significant amounts of power in January 1963. The
plant is operated by Consumers Power Company.
1 9
Gaseous waste discharges are limited by Technical Specifications
to one curie per second for fission and activation gases. Discharges
of halogens and particulates to the atmosphere are limited to 10CFR20
limits x 1.2 x 10^® cm^/sec. No concentration limit was prescribed;
however, assuming a concentration limit of 10" 10 /iCi/cc which is the
10CFR20 limit for l^ll, the limiting discharge rate for halogens and
particulates is 1.2 uCi/sec. Operating reports did not include
information on quantities of halogens and particulates discharged.
Liquid waste discharge concentrations are limited to concentrations
specified in Appendix B, Table II of 10CFR20. For unidentified gross
beta-gamma activity this would be 10"? juCi/cc, Occasional partial
analyses have indicated that 90% of the activity has consisted of a
combination of ^Zn, ^®Co, 13?Cs, l^OBa, and l^La.
Discharges of radioactivity to the environment are shown on
Figures 1-1 through 1-3. These plots compare primary coolant concen-
trations and waste discharges to electrical power generation.
Since discharge data are not available on a monthly or shorter
basis, it is not practical to compare discharges to plant maintenance
and operations. It can be seen from Figure 1-1 that the primary coolant
activity increased significantly during the first few years of operation
and then began to decrease with further operation. A similar trend is
noted in the gaseous waste discharge as shown in Figure 1-3. The
decrease occurred even though there was no reduction in power generation.
Review of Figure 1-2 indicates no particular trend of this type in
liquid waste discharge.
The Big Rock Point reactor has been used in an extensive R&D
program involving the testing of fuel and fuel cladding.13 Tests have
included different metals for fuel cladding, thick and thin claddings
with various heat treatments, powder and pellet forms, plutonium fuel,
and UO2 molten fuel.14 As a result of the research, the reactor has
-------
25
experienced more fuel failures than would otherwise be expected.
Table 1-1 summarizes fuel failure experience at the Big Rock Point
plant, and relates it to gaseous discharge rates.
Several of the shutdowns were associated with insertion and removal
of fuel rods for research programs. A long shutdown from 11-64 to
4-65 was for testing and repair work on the thermal shield hold-down
assemblies.
-------
26
FIGURE I - 1
BIG ROCK POINT NUCLEAR PLANT
PRIMARY COOLANT ACTIVITY
(GROSS BETA-GAMMA, LESS TRITIUM)
i- o
~ 2
~ Q£
H UJ
— o.
U <
300,000
.001
250,000
— 200,000
—150,000
.0001
100,000
50,000
1-63 1-64 7-64 1-65 7-65 1-66 7-66 1-67 7-67 1-68 7-68
to to to to to to to to to to to
1-64 7-64 1-65 7-65 1-66 7-66 1-67 7-67 1-68 7-68 1-69
-------
27
FIGURE I - 2
BIG ROCK POINT NUCLEAR PLANT
LIQUID WASTE DISCHARGED
(GROSS BETA-GAMMA, LESS TRITIUM)
10
— 300,000
— t/J
JO
— 250,000 §
— 200,000 5
150,000 i—
— 100,000 g
— 50,000
1-63 1-64 7-64 1-65 7-65 1-66 7-66 1-67 7-67 1-68 7-68
to to to to to to to to to to to
1-64 7-64 1-65 7-65 1-66 7-66 1-67 7-67 - 1-68 7 - 68 1-69
-------
FIGURE I - 3
BIG ROCK POINT NUCLEAR PLANT
GASEOUS WASTE DISCHARGED AND ELECTRICAL GENERATION
(ACTIVATION AND FISSION GASES)
100,000
10,000
300,000
— 250,000
—200,000
150,000
—,100,000
50,000
1,000 -
1-63 1-64
to to
1-64 7-64
-------
TABLE I-l14
FUEL FAILURE EXPERIENCE AT THE BIG ROCK POINT NUCLEAR PLANT
Run
Number
Time
Interval
Power
Produced
Mtfh(e) gross
Off-Gas Activity
(fxCi/sec)
Fuel Failures at End of Run
Average
Peak
Bundles
Rods-Known
Rods-Est
I
12/62-4/66
662,287
6,000
50,000
4
9
23
II
5/66-9/66
161,844
40,000
75,000
11
34
112
III
10/66-5-67
260,906
12,000
20,000
1
1
4
IV
6/67-2/68
390,851
10,000
15,000
22
43
110
V
3/68-6-68
124,696
4,000
6,000
9
16
37
VI
7/68-5/69
404,427
25,000
40,000
9
42
82
Total
-
2,005,011
-
-
56
145
368
Note: A normal core loading consists of 84 fuel bundles, containing ~ 11,000 fuel rods. Through 5/69, a
total of 258 fuel bundles, containing 29,066 rods, have been irradiated in the Big Rock Point Reactor.
-------
30
APPENDIX II
CONNECTICUT YANKEE ATOMIC POWER PLANT
Connecticut Yankee Atomic Power Plant is located in Haddam,
Connecticut on the Connecticut River, It has a pressurized water
reactor with an authorized power level of 1,825 megawatts thermal,
which corresponds to 573 net megawatts electrical. The plant is
operated by Connecticut Yankee Atomic Power Corporation.
Primary coolant activity and monthly liquid and gaseous waste
discharges reported in monthly operating reports^ have been plotted
versus time in months (see Figures II-l through II-5) such that these
data may be compared to electrical generation and to other plant
operations. Both liquid and gaseous discharges have been broken down
into gross beta-gamma and tritium.
Technical Specifications^ limit average annual concentrations in
liquid waste discharges to those published in Appendix B, Table II of
10CFR20. Without isotopic analyses, this is 10"' jLtCi/ml. Gaseous
waste discharges are limited as follows:
"When averaged over any calendar year, the release
rate of radioactivity consisting of noble gases and other
isotopes with half-lives less than eight days discharged
at the plant stack shall not exceed 3 x 10^ x (MPC) curies
per second, where MPC is the value in microcuries per cubic
centimeter given in Appendix B, Table II, Column 1 of
10 CFR Part 20. The maximum release rate when averaged
over any one hour shall not exceed 10 times the yearly
averaged limit.
At any time when the average release rate for a
week exceeds 30% of the annual average limit given above,
the licensee shall make provisions for sampling iodine-131
to assure that its release rate averaged over any calendar
year does not exceed 66 (MPC) curies per second."
Assuming an MPC of 3x10"®
-------
31
The discontinuous plots of gaseous waste in Figure II-4 represent
periodic discharges from decay tanks. Gaseous tritium releases
plotted in Figure II-5 are results of purges of the containment
volume. The following notes refer to Figures II-l through II-5:
A Plant shutdown 621 hrs. for repairs
B Plant shutdown 463 hrs. for repairs
C Plant shutdown 80 hrs. for turbine control valve inspection
D Plant shutdown 208 hrs. for repairs, inspection, and modification
E Plant shutdown 456 hrs. for plant maintenance and turbine valve
modification
F Plant shutdown 258 hrs. for turbine valve modification
G Core loading
H Initial criticality
I First power generation
J Lithium 7 hydroxide added
K Lithium 7 concentration reduced
L Boron recovery system not operated
M Purification system shutdown
N Waste liquid processed through evaporator
0 Primary to secondary leakage noted
P Power increase to 600 MWe approved
-------
' FIGURE i - 1
C0MMECT1CUT YANKEE ATOMIC-POWER PLANT
PRIMARY COOLANT ACTIVITY
10 r-
<* ,
in A
1/1
tu
tv
:>
o .
re.
y ,01
001
.0001
M
/'
4
I
Af
f\l
L 4:
( K f
A'
'tiM
l-
-------
3:3
FIGURE II - 2 ...
CONNECTICUT YANKEE ATOMIC POWER PLANT
liquid waste discharged
(GROSS BETA GAMMA LESS TRITIUM)
8
,000
-------
34
FIGURE 11 - 3
CONNECTICUT YANKr-T A'fOMiC POWER PLANT
LIQUID WASTE DISCHARGED ¦
(TRITIUM)
1,000
S
a,
m
m
, . •
Explanatory notes
are on pane. 31.
000 Ul
» j.ii
-------
rfGURt" II • 4-' '¦
CONNECTICUT YANKEE ATOMIC POWER FLAW!
GASEOUS WASTF DISCMARf.fD
(GROSS BETA GAMMA t ESS Tfif 1HJM)
35
T.
6
u ¦
&
f.<|jion>iti'>ry note*
or« oft pag* 31.
vm
400,000 5
f
. .. 1. .
.'l 00,000
11
20a, noo " j?
'i -
100,000
b-
u
U uj
ui
-------
FIGURE 11 - 5
CONNECTICUT -YANKEE; 1.10Mr ftWER. PLANT
GA'>tGU<> WAS rh CfSCH/iPCRt)
(i R! fiUM)
-------
37
APPENDIX 111
DRESDEN MUCLOU; POWRU STATION Will* 1
liro'olen NueUxtr Pm^r .ji I uii i r, Locnrrd on i io filifOMs ftlvi-r
ne^s r >lovi't ^ , 1 i 11 noi v, It nt tl i?i:« a Jkm i 5 ttg wat tr i o-u i or of tienorol
1,1 ('C r r it- dtslf-,11 with an autlmri/.vd puwc i iovei o I >'00 l-*, thi'r-
wa i , which corrt;ud£; t n :t not powor Jew! oi 200 Wt, If » oftor/ii^d
by Oo/witonvt a i tb ftd I «ort uiul It m. boon in «>,»!#!»• re J "i 1 ^polM! hoi lattice
Atigii'.it {%0.
. * y . ...
Haste ci i^cliargiis from i>ro*d»-u-l «>*" / »¦," >t f < d ' *>i< au iiojitul bar i -»
,iihJ ht< plutfni in figures III -1 *mot«i t;jt" (
au h 11, eiftp t. v«» uiddv t" o oorapArf? d I schJ rgt;f« to piiint nui I niouoio,:*; and
cijH'fal jom,. A ii-t/H w j-4 Ki v.myv. TIT-I >'ind ill ¦? i ml fxot .»"*» h n,( n« tn 1
"t iu t'(\i '>», tit t i qii f<1 a red f/tsoons discharges wit.li i£t~veI lug «»i f i'f f cdu<' f j »•.5 * j! N)€fK;fO. 1,'lu* limit Au oi noble i inn ion g.ir.f.;,
i;j 'xl ''r' ju€1/.s»h 4 !'fw t i i f lisail4; arc, C",tahl! for balogoo". n nd
|v
-------
FIGURE 111 - 1
w
00
DRESDEN NUCLEAR POWER STATION, UNIT 1
LIQUID WASTE DISCHARGED
(GROSS BETA GAMMA LESS TRITIUM)
10
ar
<
o
a.
DC.
U
- 2,000,000
- 1,500,000
- 1,000,000
500,000
1960
1961
1962
1963
1964
1965
1966
1967
1968
<
a:
O
_J.
<
U :
cr :
I—
u
-------
FIGURE III - 2
DRESDEN NUCLEAR POWER STATION, UNIT 1
GASEOUS WASTE DISCHARGED
(NOBLE FISSION GASES)
1,000,000
100,000
z
o
2,000,000 <
oc
- 1,500,000 z-2
o<"
- 1,000,000 _J-C
< •
- 500,000 s I
h- ^
-------
TABLE III-l
DRESDEN NUCLEAR POWER STATION UNIT 1
RADIONUCLIDE CONCENTRATION IN PRIMARY COOLANT3
Fission Product
Radionuclides
Concentration, MCi/ml
Feb. 1, 1968
Aug. 22, 1968
51 -d 89Sr
1.2 x 10-4
4.4 x 10 5
28 -yr 90Sr
1.2 x 10"5
1.3 x 10"6
91
9.7 -hr y Sr
2.4 x 10"2
1.2 x 10"2
10.3 -hr 93Y
~7 x 10"3
~2 x 10"3
65 -d 95Zrb
9 x 10"4
2.4 x 10"5
17 -hr 97Zrb
1.0 x 10"2
1.3 x 10"4
35 -d 95Nbb
5 x 10 ^
1.8 x 10 5
66.3 -hr 99Mob
1.4 x 10"3
1.4 x 10"3
6.0 -hr 99mTcb
9.0 x 10"2
4.4 x 10*2
103
39.7 -d Ru
-4
5 x 10
5.6 x 10 5
1.0 -yr 106Ru
NM
1.3 x 10"6
36 -hr 105Rh
~4 x 10"3
~4 x 10"4
78 -hr 132Te
8 x 10"4
1.9 x 10"5
131,
-3
„ „ _3
8.06-d I
7.8 x 10
2.2 x 10 J
20.9 -hr 133I
6.6 x 10"2
2.3 x 10"2
6.7 -hr 135I
1.3 x 10"1
3.6 x 10"2
2.07-yr 134C8c
1.3 x 10"5
2.3 x 10"5
13 -d 136Csc
3 x 10-5
2.4 x 10"5
30 -yr 137Cs
3 x 10"5
4.4 x 10"5
12.8 -d 140Ba
1.4 x 10"3
1.0 x 10"3
32.5 -d 141Ce
9 x 10"4
4.4 x 10"5
-------
TABLE III-l (Cont'd)
DRESDEN NUCLEAR POWER STATION UNIT 1
RADIONUCLIDE CONCENTRATION IN PRIMARY COOLANT3
Concentration, fiCi/ml
Fission Product
Radionuclides
Feb. 1, 1968
Aug. 22, 1968
33 -hr 143Ce
3.2 x 10"3
1.0 x 10"4
284 -d 144Ce
3 x 10~4
8.0 x 10"6
11 -d 147Nd
3 x 10'4
~3 x 10"5
2.34-d 239Npc
2.1 x 10"2
4.1 x 10"3
163 -d 242Cmc
2.2 x 10"7
1.2 x 10"7
Activation Product
Radionuclides
12.3 -yr 3Hd
1.7 x 10'3
1.3 x 10"3
15.0 -hr 24Na
3.0 x 10"3
1.6 x 10'3
27.8 -d 51Cr
NM
-5 x 10"4
313 -d 54Mn
NM
1.6 x 10"6
2.7 -yr
9.5 x 10"5
4.0 x 10"5
270 -d 57Co
~1 x 10 5
-6
1.6 x 10
71 -d 58Co
1.4 x 10"2
1.7 x 10"3
5.26-yr 60Co
2.2 x 10"3
2.6 x 10"4
12.7 -hr 64Cu
~1 x 10~2
2.2 x 10"3
244 -d fi5zn
NM
4.0 x 10"6
253 -d HOmAg
NM
1.9 x 10"6
115 -d 182Ta
2 x 10"5
I
X
1—«¦
o
1
-J
Concentrations at time of sampling. Also activation products,
^Formed by (n,y) reactions with uranium or fission products.
Also from ternary fission. NM: Not measured.
-------
TABLE III-2
DRESDEN NUCLEAR POWER STATION UNIT 1
RADIONUCLIDE CONCENTRATIONS IN HIGH CONDUCTIVITY LIQUID WASTE, pCi/mla
Radio-
Nuclide
Nov. 12
1967
Jan.
13, 1968
March
16, 1968
June 25,
1968
Aug.
20, 1968
Average
3H
900b
(NM)
950
(NM)
520
(NM)
770
(NM)
1,100
(NM)
850
5AMn
NM
(< Db
NM
(< 1)
NM
(< 1)
NM
(< 1)
< 1
(22)
(4)
55Fe
NM
(NM)
NM
(NM)
NM
(NM)
NM
(45)
NM
(50)
(48)
58Co
NM
(65)
4
(34)b
880
(1,800)
3
(85)
18
(260)
230
(450)
60Co
NM
(33)
1
(46)
500
(1,350)
4
(180)
11
(1,400)
130
(600)
89Sr
NM
(140)
220
(320)
140
(170)
85
(89)
24
(34)
120
(150)
90Src
NM
(14)
12
(17)
30
(30)
8
(9)
9
(U)
15
(16)
91y
NM
(66)
1
(26)
NM
(< 1)
NM
(1)
NM
(< 1)
(19)
131^
NM
(5)
6
(6)
NM
(45)
46
(49)
15
(15)
22
(24)
134Cs
NM
(9)
7
(10)
80
(90)
47
(50)
27
(34)
40
(39)
l37Cs
NM
(29)
35
(35)
150
(170)
140
(160)
99
(103)
106
(99)
lA0BaC
NM
(45)
40
(160)
65
(95)
22
(54)
35
(105)
41
(92)
1A4Ce
NM
(5)
NM
(< 1)
NM
(< 1)
NM
(< 1)
8
(16)
(4)
Concentration of radionuclides at indicated date of sampling; 1 pCi/ml = 1 x 10"^ yCi/ml.
^Values without parentheses are for filtered solution, values in parentheses for unfiltered solution.
cSolutions also contained at concentrations equal to its parent; and ^"^La at concentrations
1.15 times those of its 140ga parent.
-------
43
TABLE III-3a
DRESDEN NUCLEAR POWER STATION UNIT 1
RADIONUCLIDE CONCENTRATIONS IN LIQUID LAUNDRY WASTE,b (pCi/ml)
Radionuclide
January 15, 1968
March 12, 1968
3
H
1.3
1.5
58Co
2.2
1.2
60Co
2.3
0.6
89Sr
<0.1°
0.9
90Sr
0.1
< 0.1
131x
< 0.1
< 0.1
134Cs
0.6
< 0.1
137Cs
3.4
0.8
140Ba
< 0.1
< 0.1
141Ce
< 0.1
< 0.1
gross beta
10
5
gross alpha
< 0.1
< 0.1
^able taken from Reference 11.
^Radionuclide concentrations at sampling date; gross values
approximately one week later.
c < values are 3 O counting error.
-------
TABLE III-4a
DRESDEN NUCLEAR POWER STATION UNIT 1
STACK RELEASES OF FISSION PRODUCT NOBLE GASES,^ (jiCi/sec)
Sample Date and Time (Central Time)
Radionuclide
11/15/67
(0600)
11/16/67
(0600)
1/17/68
(1000)
1/18/68
(0930)
1/31/68
(0900)
6/26/68
(1030)
6/27/68
(1630)
8/20/68
(1928)
8/21/68
(0547)
4.4 -hr 85mKr
580
340
280
10.7 -yr 85Kr
0.024
0.25
76 -m 87Kr
~ 540
~ 540
~ 1,300
2.8 -hr 88Kr
~ 1,300
~ 780
~ 640
~ 260
2.3 -d 133n^e
14
11
11
8
5
31
20
5.3 -d 133Xe
450
420
400
410
970
160
110
910
730
9.1 -hr 135Xe
930
1,010
1,660
1,250
2,600
620
520
1,420
17 -m 138Xe
~ 850
~ 2,800
~ 2,400
Gaseous fission
product release
rate
11,000
10,000
14,800
12,600
18,400
7,500
5,700
11,600
11,600
^Table taken from Reference 11.
^Based on sampling off-gas delay line; computed for release at top of stack after radioactive decay
and dilution of 9.4 1/sec off-gas by exhaust air from containment and turbine buildings at flow
rate of 21 m3/sec.
indicates that radionuclide was not measured. In the case of the short-lived radio-
nuclides, their absence in samples was due primarily to delays between sampling and analysis.
-------
45
TABLE III-5a
DRESDEN NUCLEAR POWER STATION UNIT 1
SUMMARY OF STACK RELEASES
OF PARTICULATE RADIONUCLIDES AND GASEOUS IODINE-131
Number of
pCi/
sec
Radionuclide
Measurements
Mean
Range
58C 58Co
15
26
<8 - 140
60Co
16
25
2.5- 95
&9Sr
15
970
220 -2,300
90Sr
15
5
2-25
137Cs
16
35
13 - 55
1^°Ba
17
430
70 - 710
131i
17
920
200 -3,230
Gaseous fission
product release rate,
mCl/sec
17
11.7
5.7 - 53
TABLE III-6a
DRESDEN NUCLEAR POWER STATION UNIT 1
RELEASE RATE OF TRITIUM FROM STACK
Date
1968
Concentration in
Primary Coolant,
pCl/ml
Concentration
1° Delay Line,
pCi/ml
Release Rate. uCl/sec
Measured
Estimated0
Jan. 18
—
1.5
1.4 x 10"2
...
Feb. 1
1,660
...
...
5.0 x 10"3
June 26
...
0.52 + 0,3C
4.8 + 0.3 x 10"3
...
Aug. 22
1,300
~
3.9 x 10-3
aTable taken from Reference 11.
''Based on equivalent water for hydrogen gas of 2.7 ml/sec and for
water vapor of 0.32 ml/sec.
cStandard deviation of duplicate analyses; only one sample was
analyzed on Jan. 18.
-------
46
APPENDIX IV
HUMBOLDT BAY PLANT, UNIT 3
Humboldt Bay Nuclear Power Plant is located on Humboldt Bay near
Eureka, California along with two fossil fuel plants. It utilizes a
boiling water reactor with an authorized power level of 240 megawatts
thermal, which corresponds to 68 net megawatts electrical. The plant
has been in commercial operation since August 1963 and is operated by
Pacific Gas and Electric Company.
Technical Specifications^ limit average annual discharges of
liquid wastes to levels published in Appendix B, Table II of 10CFR20.
Gaseous radioactive waste discharges are limited as follows:
"The annual average stack emission rate for noble and
activated gases shall not exceed 0.05 curies per second;
the instantaneous stack emission rate shall not exceed 0.5
curies per second.
The annual average stack emission rate for halogens
and particulate material based on the isotopes present on
the sampling filters after 48-hour decay period shall not
exceed the permissible air concentrations for unrestricted
areas given in 10CFR20 multiplied by 6 x 10® cubic centi-
meters per second. Until the approximate composition of
the mixtures sampled is indicated by monitoring history
the permissible air concentration shall be assumed to be
3 x 10~10 jnCi/cc for the halogen and particulate groups."
Based on the above guidance, the discharge rate limit for halogens
and particulates would be 0.18 /iCi/sec.
Figures IV-1 and IV-2 represent plots of monthly quantities of
liquid gaseous waste discharged as reported in operation reports.20
The power produced and refueling outages are also plotted for compari-
son. Operating reports include gross beta-gamma analyses less tritium
for liquid waste. Gaseous wastes are reported as noble and activated
gases separate from halogens and particulates. Noble and activated
gases are reported in average monthly discharge rates whereas a range
of discharge rates is reported for halogens and particulates as
summarized in Table IV-1. Discharges of halogens and particulates
are a smaller percentage of the established limit than is the case
for noble and activated gases.
Between initial startup in 1963 and August 1965, the principal
contributors to the radioactivity in the liquid radwaste discharge were
-------
After equilibrium levels had been reached
Dmposition of typical discharges was 87"L
Zn, Mn, and Co.
the average isotopic composition of typical discharges was 87"L UJZn,
97o ^Mn, and 4% ^Co.
Beginning in May 1965, defects began occurring in the Type I
stainless steel fuel. The effect of these failures did not affect
liquid discharges until August 1965 due to hold-up in the radwaste
system. From August 1965 until March 1966, the average isotopic
composition of radwaste discharged was 467« 137co, 297o Zn, 217o
^Cs, 37o -*^Mn, and 17, ^Co. Other fission and corrosion products
are present but their concentrations are sufficiently low to be masked
by the major contributors to the activity. ^®Sr and ®9gr, not being
gamma emitters, are not detected by this analysis; however, radio-
chemical analyses indicate that these nuclides comprise less than 17,
of the gross beta activity.
The feedwater heater tube bundles were changed from Admiralty to
stainless steel in November 1966 and present analyses reflect the
expected decrease in 65zn activity. Typical analyses of liquid rad-
waste now show approximately 407=. 137cs, 257. ^Co, 107> ^Mn,
134Cs, 10% 65Zn, and 5%131I.
The following notes apply to Figures IV-1 and IV-2:
A Six-month average
C Initial core loading
F Commercial operation
R Refueling outage
-------
48
TABLE IV-1
HUMBOLDT BAY NUCLEAR POWER PLANT
AVERAGE WEEKLY STACK DISCHARGE RATES FOR HALOGENS AND PARTICULATES 17
Period
Range of Discharge Rates
uCi/sec a
2-16-63 to 8-15-63
6.0xl0-6 to 2.0xl0"3
8-16-63 to 2-15-64
9.OxlO-5 to 5.5xl0"3
2-16-64 to 8-15-64
2.4xl0"6 to 9.3xl0"6
8-16-64 to 2-15-65
l.OxlO"5 to 4.3xl0"4
8-16-65 to 2-15-66
3.2xl0-5 to 2.8xl0~2
2-16-66 to 8-15-66
2.7xl0~4 to 1.8xl0~2
8-16-66 to 2-15-67
l.OxlO-4 to 7.Ixl0~2
2-16-67 to 8-15-67
4.2xl0™3 to 1.3xl0_1
8-16-67 to 2-15-68
9.OxlO"3 to 9.2xlO"2
1-1-68 to 6-30-68
7.lxlO-3 to 9.2xl0"2
7-1-68 to 12-31-68
3.0x10 ^ to 3.3xlO~2
aBased on an MPC limit of 3x10"^® iiCi/cc^^ the maximum permissible
discharge rate would be 0.18 juCi/sec.
-------
FIGURE IV - 1
HUMBOLDT BAY PLANT, UNIT 3
LIQUID WASTE DISCHARGED
(GROSS BETA-GAMMA LESS TRITIUM)
Explanatory notes
are on page 47.
r 50,000
- 40,000
- 30,000.
w
- 20,000
a. 2
- 10,000 J—
1964
1965
1963
1967
-------
50
FIGURE IV - 2
HUMBOLDT BAY PLANT, UNIT 3
GASEOUS WASTE DISCHARGED
(NOBLE AND ACTIVATION GASES LESS TRITIUM)
100,000 —
10,000 —
z
o
5
9>
a
LO
UJ
QL
3
U
50,000 z
40.000 <
30,000 w e
20,000 < ^
10,000 n
10 iiiiiiiiiiiiiiiniiiTiTimiiiiiTi
11111
1,000
Explanatory notes
are on page 47.
1963 1964 1965
lllllll|lllllllllll
1966 1967
1968
-------
51
APPENDIX V
INDIAN POINT STATION, UNIT 1
Indian Point-1 is located on the Hudson River about 25 miles
north of New York City. It utilizes a pressurized water reactor of
Babcock and Wilcox design and is licensed to a power level of 615
thermal megawatts. The plant is operated by Consolidated Edison and
has been in commercial operation since October 1962.
Discharge of liquid radioactivity to the environment is limited to
concentrations specified in Appendix B, Table II of 10CFR20. Stack
discharges of particulate matter and halogens with half-lives longer
than eight days are limited by the Technical Specifications to average
annual discharge concentrations not exceeding 2.4xl0^xMPC curies per
second where MPC is in jLiCi/cc as listed in Appendix B, Table II of
10CFR20. Discharge concentrations for all other isotopes are limited
to 1.7xl0^xMPC curies per second. Using a limit of 10~10 ju.Ci/cc which
is the MPC for 131l, the discharge rate limit for halogens and
particulates is 0.24 uCi/sec. Assuming an MPC of 3x10"^ JLiCi/cc for
noble and activated gases, the discharge rate limit for gaseous dis-
charges would be 51,000 iiCi/sec.
Radioactivity in primary coolant and in liquid and gaseous dis-
charges has been reported in semiannual operation reports^ and is
plotted for comparison with power generation and maintenance operations
in Figures V-l through V-3. Figure V-l provides a comparison of
tritium concentrations in the primary coolant to power production and
indicates an apparent relationship between shutdowns and tritium
concentrations. Data concerning tritium discharges are plotted for
comparison with power generation in Figure V-2. Review of gross beta-
gamma discharges as compared to power operations plotted in Figure V-2
indicates an increase in quantities of radioactivity discharged following
periods of shutdown. This figure also indicates an increase in dis-
charge rates with increased power history with a tendency to level off
after about three years of operation.
Figure V-3 provides a comparison of gaseous discharges with power
generation and plant operations. The discontinuous plot of discharge
quantities reflect plant outages during which time negligible quantities
of radioactive gases are generated. There is no observable relation-
ship between power history and quantities of gaseous radioactivity
discharged.
Table V-l provides a history of primary to secondary leakage.
These are significant for both liquid and gaseous waste discharges.
With primary to secondary leakage, large quantities of liquid waste
-------
52
result from steam generator blowdown. Gases in the primary coolant
may escape to the steam system where they are scrubbed out through the
air ejectors and discharged to the environment.
The following notes refer to Figures V-l, V-2, and V-3:
A Initial criticality
B First power generation
C Electrical generation based on a six-month average
D Shutdown for maintenance
E Three shutdowns for maintenance totaling 56 days
F Shutdown for refueling and maintenance
G Shutdown for refueling; core "B" installed
-------
TABLE V-l
INDIAN POINT STATION UNIT 1
PRIMARY TO SECONDARY LEAKS
Boiler #
Date Leak Started
Date Fixed
Comments
14
June 3, 1963
October 1963
11
Early 1964
October 1964
14
August 1964
October 1964
11
November 1964
March 1965
11
April 1965
During refueling o
utage
October 20-April 23, 1966
14
April 1966
April 1966
14
April 1966
May 1966
12
May 1966
September 1966
14
May 1966
September 1966
Two tubes leaking
12
October 1966
April 1968
14
April 1968
Four tubes fixed
-------
54
a:
LU
s.
tO
C£
3
U
o
OS
u
FIGURE V - 1
INDIAN POINT STATION, UNIT 1
PRIMARY COOLANT ACTIVITY
(TRITIUM)
mmm
Explanatory notes
are on page 52.
—, 200,000
—I 150,000
—I 100,000 y
—I 50,000
z
o
I-
<
QL
LU
z
O
LU
E
O
\
_c
<
•
u
Q£
J
h"
U
| | |
-J
LU
1966
1967
-------
55
FIGURE V - 2
INDIAN POINT STATION, UNIT 1
LIQUID WASTE DISCHARGED
100
TRITIUM
GROSS 0-y
Explanatory notes
are on page 52.
.01
.001
r 200,000
- 150,000 <
.0001
- 100,000 ~
- 50,000
LU
.00001
I 111*11^1 111 I MM I M
1966 1967 1968
1962
1963
1965
1964
-------
56
FIGURE V - 3
INDIAN POINT STATION, UNIT 1
GASEOUS WASTE DISCHARGED
(GROSS BETA-GAMMA LESS TRITIUM)
100 r-
Explonatory notes
are on page 52.
200,000
150,000
100,000
50,000
| I 11' H I»" I' il j
1966
111 iTtrTrprmt
1962 1963
10
z
o
2
a:
=5
U
.01
.001
-------
57
APPENDIX VI
PEACH BOTTOM ATCMIC POWER STATION
Peach Bottom Atomic Power Station is located on the Susquehanna
River near the town of Peach Bottom, Pennsylvania. It utilizes a
high temperature gas-cooled reactor of Gulf General Atomic design
and is authorized to operate at a net power level of 40 MWe. The
plant is operated by Philadelphia Electric Company and has been in
commercial operation since January 1967.
Technical Specifications^ limit average annual concentrations of
radioactivity in liquid discharges to those published in Appendix B,
Table II of 10CFR20. Average annual discharge rates to the atmosphere
of halogens and particulates with half-lives over eight days is limited
to 28.6xMPC curies per second where MPC is the maximum permissible
concentration in juCi/cc as listed in Appendix B, Table II of 10CFR20.
Annual average discharge rates to the atmosphere of all other radio-
isotopes are limited to 2xl0^xMPC curies per second. Based on an MPC
of 10~10 juCi/cc for halogens and particulates and 2x10"^ fiCi/cc for
other radionuclides, the discharge rate limits are 2.86xl0"3 ^uCi/sec
and 400 jiCi/sec respectively.
Figures VI-1, VI-2, and VI-3 are plots of radioactivity in the
main loop and in liquid and gaseous discharges as compared to power
produced and maintenance operations from Peach Bottom-1 operating
reports.24 Figure VI-1 provides a plot of main coolant activity and
power produced as a function of time in months. The fuel elements
have a purge system designed to remove fission gases prior to their
leaking into the primary coolant. However, cracks have developed in
the sleeves that contain the purge gas permitting fission gases to
leak into the coolant. Figure VI-1 shows an increase in radioactivity
as function of time and power history. The principal isotopes causing
this activity are ^m^r, 88j£rj 87j^r^ 89kj-, l38xe, and 135xe.
Since the primary coolant is gaseous, there is little liquid radio-
active waste discharged at Peach Bottom. The major contributing isotope
in liquid wastes is 24ua which occurs as a result of neutron activation
of sodium salts contained in a rust inhibitor used in the treatment of
the shield cooling water system.25 Figure VI-2 provides a plot of
liquid waste discharges and power generations as a function of time in
months. A similar plot of gaseous discharges is provided in Figure VI-3.
Review of Figure VI-2 reveals no particular trend of liquid waste
discharges as a function of time or reactor maintenance. Review of
Figure VI-3 does reveal a general increase in gaseous discharges as a
-------
58
function of time and power history. Essentially 100% of the gaseous
waste released is 85Kr.25
Footnotes for Figures VI-1, VI-2, and VI-3:
A Small amounts of fission products transported through purification
system
B Breakthrough from Kr-85 trap caused an increase in main loop
impurities
C Small amounts of chemical contaminants entered the primary coolant
and low temperature delay bed system
D Feedwater leak into general containment caused a larger volume of
radioactive waste
F First power supplied to Pennsylvania Electric Company
G Air in-leakage during fuel handling equipment operation
J Unpurged fuel element suspected
K Airborne activity in air room and stack
L Primary system activity increased by a factor of 13 due to fuel
element leakage
R Refueling outage
S Release of contaminated helium
T Leakage of helium from control rod drive housing
-------
FIGURE VI - 1
PEACH BOTTOM ATOMIC POWER STATION, UNIT 1
MAIN LOOP ACTIVITY
(GROSS BETA GAMMA LESS TRITIUM)
Explanatory notes
are on page 58.
50,000
40,000
30,000
20,000 < ^
10,000
1966
1967
1968
1969
-------
60
FIGURE VI - 2
PEACH BOTTOM ATOMIC POWER STATION, UNIT 1
LIQUID WASTE DISCHARGED
(GROSS BETA GAMMA LESS TRITIUM)
.001 i-
.0001 -
Explanatory notes
are on page 58.
X
h-
Z
o
® .00001
to
LLJ
CC
Z>
o
.000001 r
.0000001
10,000
-------
61
FIGURE VI - 3
PEACH BOTTOM ATOMIC POWER STATION, UNIT 1
100 c-
GASEOUS WASTE DISCHARGED
(ACTIVATION AND NOBLE GASES LESS TRITIUM)
10
z
o
<5
Q.
OC.
u .01
.001
.0001
.00001
i i i i i i 1 r
1966
p W l*i
J
>*> I't'l "!
11 M'ci
\\J
Explanatory notes
are on page 58.
1967
1968
J"
-r-r-r
1969
40,000
30,000
2
O
t-
<
QL
LU
Z o
uj g
O <
20,000 •u ^
10,000 £
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62
APPENDIX VII
SAN ONOFRE NUCLEAR GENERATING STATION
San Onofre Nuclear Generating Station is located on the Pacific
coast near San Clemente, California. It utilizes a pressurized water
reactor of Westinghouse design with an authorized net power level of
430 MWe. The plant is operated by Southern California Edison Company
and has been licensed for operation since March 1967.
Technical Specifications^ limit discharges of liquid radioactive
waste as follows: "Averaged over a year, the release rates of liquid
wastes shall not result in concentrations in the circulating water dis-
charge in excess of Part 20 limits for unrestricted areas, except that
the maximum release rate over the period of one hour shall not exceed
10 times the yearly averaged limit."
Gaseous waste discharges are limited as follows: "Averaged over
a year, release rates of gaseous wastes in Ci/sec shall not result in
a value exceeding that calculated from the following formula:
1.8 x 105 (_"I) x SCX (M|i)
sec
Where Cx is the concentration of any radioisotope x, the values of the
concentration of all isotopes discharged shall be such that y; ^x
is less than 1.0. (MPC)x as defined above shall be that (MPC)X
stated in Column 1, Table II of 10CFR20. The maximum release rate
over any one hour shall not exceed 10 times the yearly averaged limit
as stated above."
The limit for gaseous discharges does not specify a separate
standard for halogens and particulates to account for potential
reconcentration through the environment. Operating reports27 provide
information on quantities and concentrations of radioactivity in liquid
and gaseous discharges; however, no isotopic analysis is reported.
Therefore, it is assumed that the concentration limit for liquid dis-
charges is 10-7 /LiCi/cc and the applicable MPC for gaseous releases is
3x10"7 yCi/cc. This limit takes into account a 60-day hold-up for
decay. Based on a concentration limit of 3x10"^ /iCi/cc and the formula
provided in the Technical Specifications, the discharge rate limit
would be 5.4x10^ juCi/sec.
Figures VII-1 and VII-2 provide plots of liquid and gaseous dis-
charges and power generation as a function of time in months. Due to
-------
63
the limited period of operation, trends in discharges of radioactivity
are not yet discernible. The following notes refer to Figures VII-1
and VII-2:
A Initial criticality
B Six-month average
S Shutdown
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FIGURE VII - 1
SAN ONOFRE NUCLEAR GENERATING STATION
LIQUID WASTE DISCHARGED
(GROSS BETA GAMMA LESS TRITIUM)
1 t—
.01
.001
.0001
.00001
1967
I
Explanatory notes
are on page 63.
300,000
z
o
H
<
200,000 £
z
LU
o
_i
<
u
100,000 a
J—
u
LU
I I I I IT
1968
11111111
1969
-------
65
FIGURE VII - 2
SAN ONOFRE NUCLEAR GENERATING STATION
GASEOUS WASTE DISCHARGED
(GROSS BETA GAMMA LESS TRITIUM)
.1
O
Q.
to
a: -01
u
.001
.0001
1967
Explanatory notes
are on page 63.
1968
300,000
z
o
200,000 {*
100,000
o
uj g
o \
< •
y *
a: 2
H
U
UJ
—I
UJ
1969
"I
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66
APPENDIX VIII
SHIPPINGPORT NUCLEAR POWER STATION
Shippingport Nuclear Power Station is located on the Ohio River
about 25 miles northwest of Pittsburgh, Pennsylvania. It utilizes a
pressurized water reactor of Westinghouse design with a power level
equivalent to 150 MWe, and a gross electrical output of 100 megawatts.
Shippingport, unlike other power reactors, has been developed and
operated under AEC sponsorship and, as a result, does not operate under
an AEC license. The plant is operated for the AEC by Duquesne Light
Company and has been producing power since December 1957.
Discharge limits for radioactive wastes are set by the State of
Pennsylvania. "The discharge permit granted by the State stipulated
that the final wastes as discharged to the Ohio River shall at no time
contain more than 10_7juCi/ml above that of the intake water of the Ohio
River (exclusive of tritium). Although the discharge permit allows
10"7jiCi/ml, the normal operating limit at the Shippingport station is
specified as lO-S^ci/ml, the original design basis for the various
liquid waste process streams. Under special conditions, discharges
above 10~®^Ci/ml, but not exceeding 10~7jLiCi/ml, are permitted. "7
Tritium discharges are limited to 10 Ci/day averaged over any 365 day
period with a maximum of 50 Ci/day.33
Gaseous discharges are limited to concentrations of 3xl0"^Ci/ml
in the stack. Based on a flow rate of 9,000 cfm in the stack, this is
equivalent to 1.26uCi/sec.
Figures VIII-1 and VIII-2 provide plots of liquid and gaseous
radioactive waste discharges and power generation as functions of time
in years. Figure VIII-1 shows a significant decrease in tritium dis-
charges beginning in 1961 and a subsequent increase beginning in 1965.
The decrease is attributed to conversion of the LiOH resin in the puri-
fication system from &Li to ?Li. An increase in the station's power
capability from 68 to 150 MWe is the apparent reason for the 1965
increase in discharges. The peak in gross beta-gamma discharges in
1964 is attributed to decontamination of reactor coolant systems that
took place in March of that year. Fluctuations in gaseous discharges
shown in Figure VIII-2 are partly due to the station's practice of
holding up gaseous wastes for decay and releasing these wastes at a
controlled low concentration over a relatively short period of time.
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67
FIGURE VIII - 1
SHIPPINGPORT ATOMIC POWER STATION
LIQUID WASTE DISCHARGED
TOO
10
TRITIUM
az
<
X/
1/5
Ltl
OH
Z>
u
GROSS 0-7
1959 1 1960 ' 1961 ' 1962 1 1963 ' 1964 ' 1965 1966 1967 1968 1969
600,000 2
500,000 2
400,000 z
O \
300,000 _,±
200,000 u ?
a: 5
100,000 [3
0 iU
UJ
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68
FIGURE VIII - 2
10 (—
1.0
cx
<
oc
ra
u
SHIPPINGPORT ATOMIC POWER STATION
GASEOUS WASTE DISCHARGED
(GROSS BETA-GAMMA LESS TRITIUM)
.01
.001
FZ-Z-7.
jmrnmtmz
iv«vAV«vt • • • • • • ! • • • i • • 11 ».* i <
b%Vi%ti%VfSVASSV#VAV,ViV»V*SV,| iV#V»|iV»V»
600,000 p
500,000 a
LU
400,000
300,000 ^
200,000 y ^
£ s
100,000 {3
0 _J
1959 1960 1961 1962 1963 1964 1965 1966 1967 1968 1969
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69
APPENDIX IX
YANKEE ATOMIC POWER STATION
Yankee Atomic Power Station is located on the Deerfield River in
Rowe, Massachusetts. It utilizes a pressurized water reactor of
Westinghouse design with an authorized net power level of 175 MWe.
Yankee is operated by Yankee Atomic Electric Company and has been in
operation since August 1960.
OQ
Technical Specifications^ limit liquid discharges to concentrations
listed in Appendix B, Table II of 10CFR20. Gaseous discharges are
limited as follows: "As determined at the point of discharge from the
primary vent stack and averaged over a period not exceeding one year,
the concentration of radioactive gaseous wastes discharged shall not
be in excess of 1,000 times the limits specified in Appendix B, Table II,
10CFR20." Based on a stack exhaust rate of 15,000 CFM, and a concentra-
tion of 1,000 times MPC* in the stack, the limiting discharge rate for
noble gases is 212 yCi/sec or 6,700 Ci/yr. This does not agree with
the limit of 226 ^Ci/sec used in Table 10 for calculating percent of
limit. The limit of 226 juCi/sec is based on analysis of radionuclides
in 1968 discharges as reported in Reference 8.
Figures IX-1 through IX-4 provide plots of primary coolant activity
and waste discharges as a function of time and also as compared to power
produced. These data are taken from operating reports.30 Some of the
data were not available for Figure IX-1; however, sufficient data are
plotted to indicate that gross beta-gamma activities in the primary
coolant are maintained generally in the vicinity of 0.1 juCi/ml with
rather large reductions during periods of shutdown. Since data in
Figure IX-3 do not show corresponding increases in discharge during
shutdown, it is assumed that the reduction in primary coolant concen-
trations during these periods resulted primarily from decay of short
half-lived radionuclides. The increase in tritium concentrations in
1968 follows a refueling and boron addition to the primary coolant.
The increase would result in increased interaction of neutrons with
to produce tritium.
Figure IX-2 indicates a wide fluctuation in monthly quantities
of liquid waste discharged with peaks occurring either during or
immediately following refueling outages.
*The MPC used is 3x10"® /iCi/ml which is based on a typical noble gas
mixture with less than two hours hold-up for decay.
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70
The following notes refer to Figures IX-1 through IX-4. The notes
in general confirm that the operations listed had little or no effect
on waste discharges.
A NH^OH added to primary coolant
B Boron added to primary coolant
C High carryover from waste disposal evaporator causing an increase
in liquid waste
D Corrosion of control rods increase in primary coolant activity
E Steam generator blowdown to waste disposal system
F Radiochemistry samples released to gaseous waste system
G KOH added to primary coolant
J Crud burst
K During blowdown of surge tank cover gas escaped
L Primary drain collection tank purged
M Outage to repair steam generator leak
N Inadvertent gas release
P Primary to secondary leak
R Refueling outage
S Initial criticality August 19, 1960
W NH4OH removed
X Boron removed
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71
FIGURE IX - 1
YANKEE NUCLEAR POWER STATION
PRIMARY COOLANT ACTIVITY
TRITIUM
=) .01
I'liiiinmiiii
GROSS 0-y
Explanatory notes
ore on page 70.
Z
o
<
200,000
150,000 z o
- 100,000
o
1962
mryrrTmHiiiin
1963 1964 1965
50,000 <5 i
o M
1966
1967
1968
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72
FIGURE IX - 2
YANKEE NUCLEAR POWER STATION
LIQUID WASTE DISCHARGED
(GROSS BETA-GAMMA LESS TRITIUM)
Explanatory notes
are on page 70.
.01
.001
UJ
.0001
.00001
150,000 £ ^
UJ £
100,000 " i
< .
50,000 y|
R
.000001
Tpttt
UJ
1967
1968
1966
1969
1964
1965
1962
1963
1961
-------
73
FIGURE IX - 3
YANKEE NUCLEAR POWER STATION
LIQUID WASTE DISCHARGED
(TRITIUM)
1000
z
o
& 100
CO
LU
a
u
10
Explanatory notes
are on page 70.
Z
o
<
a
Z o
LU E
150,000
_i
100,000 5 i
50,000 £ -
U
0 w
LU
1965
1966
1967
1968
-------
74
FIGURE IX - 4
YANKEE NUCLEAR POWER STATION
GASEOUS WASTE DISCHARGED
(GROSS BETA-GAMMA LESS TRITIUM)
100
.01 — c
Explanatory notes
are on page 70.
A W
.001
200,000 b
150,000 z7
uj g
100,000 J 2
.0001
V|iMrn,vf'Hii'|'i'i'iiiiii i 'iiS v| ¦' 'i ¦¦ 'i *i ¦ i' i' i ¦' 'i v| 'i ¦!1 n 11 'i 11H ¦ j1 i
1964 1965 1966 1967
1963
1962
1969
1961
1968
-------
75
REFERENCES
1. United States Atomic Energy Commission Rules and Regulations
Title 10 Part 20 Code of Federal Regulations. "Standards for
Protection Against Radiation."
2. Peterson, H. T., Jr., et alt "Environmental Tritium Contamination
from Increasing Utilization of Nuclear Energy Sources," U.S. Public
Health Service, Bureau of Radiological Health, March 1969.
3. "A Pressurized Water Reactor Nuclear Power Station," U.S. Public
Health Service, Bureau of Radiological Health, Nuclear Facilities
Branch, NF-67-6, Revision 1, September 196 7.
4. "A Boiling Water Reactor Nuclear Power Station," U.S. Public Health
Service, Bureau of Radiological Health, Nuclear Facilities Branch,
NF-67-3, Revision 1, February 1967.
5. "A High Temperature Gas Cooled Reactor," U.S. Public Health Service,
Bureau of Radiological Health, Nuclear Facilities Branch, NF-67-27,
September 25, 1967.
6. Nuclear Reactors Built. Being Built, or Planned in the United
States as of June 30. 1969. TID-8200 (19th Revision), U.S. Atomic
Energy Commission, Division of Technical Information.
7. WAPD-294, AEC Research and Development Report, "Shippingport
Operations During PWR Core 1 Depletion," December 1968.
8. Joint Committee on Atomic Energy, Congress of the United States,
Selected Materials on Environmental Effects of Producing Electric
Power. (August 1969) Chapter IV, "Background Information on Releases
of Radioactivity in Nuclear Power Effluents," pp. 79-119.
9. Blomeke, J. 0. and Harrington, F. E.; Management of Radioactive
Wastes at Nuclear Power Stations, Oak Ridge National Laboratory 4070,
January 1968.
10. Tash, J. A.; "Shippingport Atomic Power Station Atmospheric
Discharges," February 3, 1958.
11. Radiological Surveillance Studies at a Boiling Water Nuclear Power
Reactor. DER-70-I U.S. Public Health Service, Bureau of Radio-
logical Health, March 1970.
-------
76
12. Appendix "A", "Consumers Power Company Big Rock Point Nuclear
Plant Technical Specifications," appended to Operating License
No. DPR-6. May 1964.
13. Consumers Power Company, "Report of Operation of Big Rock Point
Nuclear Plant," Semiannual Reports covering period May 1964
through October 1968.
14. Walke, Gerald J.; "The Effect of Failed Fuel on the Operations of
a Commercial BWR Plant" from Transactions. Conference on Reactor
Operating Experience. October 1-3, 1969, Reactor Operations
Division, American Nuclear Society.
15. "Connecticut Yankee Atomic Power Company Operating Reports" numbers
67-7 through 69-5 issued monthly.
16. Appendix A to Provisional Operating License DPR-14. "Technical
Specifications for the Connecticut Yankee Atomic Power Company,"
Haddam Neck Plant, Haddam, Connecticut, June 30, 1967.
17. Dresden Nuclear Power Station, Commonwealth Edison Company, "Annual
Reports," 1962 through 1968.
18. Commonwealth Edison Company Appendix "A" to Facility License
No. DPR-2, September 19, 1962.
19. Technical Specifications for Pacific Gas and Electric Company.
Humboldt Bay Power Plant Unit #3. January 21, 1969.
20. Pacific Gas and Electric Company, "Report on the Operation of
Humboldt Bay Power Plant," issued semiannually covering periods
July 1963 through December 1968.
21. Appendix A to Provisional Operating License DPR-5. "Technical
Specifications for the Consolidated Edison Company of New York,
Inc.," October 29, 1965.
22. Indian Point Station. Semiannual Operations Reports, numbers 1
through 12, covering periods August, 1962 through September 1968.
23. Docket #50-171, Philadelphia Electric Company, "Peach Bottom
Atomic Power Station," Appendix A. Technical Specifications.
January 15, 1965.
24. Philadelphia Electric Company, "Peach Bottom Atomic Power Station
Monthly Operating Reports," numbers 1 through 38, covering periods
March 1966 through April 1969.
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77
25. Gazda, N. F. , Philadelphia Electric Company; "Anticipated Radiation
Hazards in a Second Generation Plant Based on Experience from a
Prototype High Temperature Gas Reactor," presented at the
Affiliated Meeting of Power Reactor Health Physics Society Midyear
Topical Symposium on Operation Monitoring, Los Angeles, California,
January 28-31, 1969.
26. Appendix A to Provisional Operating License DPR-13. "Technical
Specifications for the San Onofre Nuclear Generating Station
Unit-1," March 27, 1967.
27. "San Onofre Nuclear Generating Station Semiannual Operating
Reports," numbers 1, 2, and 3, covering periods June 1967 through
December 1968, submitted by Southern California Edison Company
and San Diego Gas and Electric Company.
28. LaPointe, J. R., et al; "Waste Treatment at the Shippingport
Reactor," Journal of the Sanitary Engineering Division Proceedings
of the American Society of Civil Engineers, May 1960.
29. Yankee Atomic Electric Company Docket #50-29. Interim Facility
License. Appendix A, "Technical Specifications," March 4, 1964.
30. Yankee Nuclear Power Station Operating Reports numbers 2 through
101, submitted monthly, January 1961 through May 1969 by Yankee
Atomic Electric Company.
31. Correspondence dated 2/6/70 from Mr. Donald J. McCormick,
Consolidated Edison Company of New York, Inc., to Mr. J. E. Logsdon,
Division of Environmental Radiation, U.S. Public Health Service.
32. Correspondence dated 1/29/70 from Mr. R. W. Sinderman, Consumers
Power Company, to Mr. E. D. Harward, Division of Environmental
Radiation, U.S. Public Health Service.
33. Correspondence dated 3/10/70 from the Naval Reactors Branch,
U.S. Atomic Energy Commission, to Mr. C. L. Weaver, Division of
Environmental Radiation, Bureau of Radiological Health.
-------
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The Coordinate Index
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The cards are
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Accession No.
RADIOACTIVE WASTE DISCHARGES TO THE
ENVIRONMENT FROM NUCLEAR POWER
FACILITIES - Joe E. Logsdon and
Robert I. Chissler;
March 1970; DER 70-2; OAS, RSB, DER, BRH, EHS, PHS, DHEW.
This report summarizes discharges through 1968 of liquid
and gaseous radioactive wastes from nine nuclear power
facilities. Discharges are compared to power history
and maintenance operations.
KEY WORDS:
Nuclear Power, Radioactive Waste, Discharges
Accession No.
RADIOACTIVE WASTE DISCHARGES TO THE
ENVIRONMENT FROM NUCLEAR POWER
FACILITIES - Joe E. Logsdon and
Robert I. Chissler;
March 1970; DER 70-2; OAS, RSB, DER, BRH, EHS, PHS, DHEW.
This report summarizes discharges through 1968 of liquid
and gaseous radioactive wastes from nine nuclear power
facilities. Discharges are compared to power history
and maintenance operations.
KEY WORDS:
Nuclear Power, Radioactive Waste, Discharges
Accession No.
RADIOACTIVE WASTE DISCHARGES TO THE
ENVIRONMENT FROM NUCLEAR POWER
FACILITIES - Joe E. Logsdon and
Robert I. Chissler;
March 1970; DER 70-2; OAS, RSB, DER, BRH, EHS, PHS, DHEW.
This report summarizes discharges through 1968 of liquid
and gaseous radioactive wastes from nine nuclear power
facilities. Discharges are compared to power history
and maintenance operations.
KEY WORDS:
Nuclear Power, Radioactive Waste, Discharges
------- |