RD71-1
RADIOLOGICAL
SURVEILLANCE STUDIES
AT A PRESSURIZED WATER
NUCLEAR POWER REACTOR
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This uic!v, tcnl report of the Office of Radiation Programs, USEPA, is available
fr..>IT> ;lie Clearinghouse for Federal Scientific and Technical Information,
Sprl n>it ield, Virginia 22151, under the namber PB 205-640.
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RADIOLOGICAL
SURVEILLANCE STUDIES
AT A PRESSURIZED WATER
NUCLEAR POWER REACTOR
Bernd Kahn
Richard L. Blanchard
Harry E. Kolde
Herman L. Krieger
Seymour Gold
William L. Brinck
William J. Averett
David B. Smith
Alex Martin
U. S. ENVIRONMENTAL PROTECTION AGENCY
Radiochemistry and Nuclear Engineering Branch
National Environmental Research Center
Cincinnati, Ohio 45268
August 1971
Second Printing August 1973
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Foreword
The Environmental Protection Agency has the responsibility of carrying out a
national program for measuring the population exposure to ionizing and nonionizing
radiation and for assessing the radiological quality of the environment. The
Radiation Research group conducts a program to determine the presence and
examine the effects of radiation in order to form the scientific base for protecting
man and his environment. Part of this research includes the development of means
for identifying radionuclides, and methods for performing field studies at nuclear
power stations and related facilities to quantitate discharged radionuclides and to
measure radionuclides in the environment.
The projected increase in the use of nuclear power for generating electricity has
placed an increased emphasis on nuclear surveillance programs at both the state and
federal levels. The Environmental Protection Agency is engaged in studies at
routinely operating nuclear power stations to provide information on the
concentration of radionuclides in effluents and throughout the environment.
The data for this study were obtained at the pressurized water nuclear power
reactor operated by the Yankee Atomic Electric Company at Rowe, Massachusetts.
The results reported here are intended to provide an initial base for performing
radiological surveillance at pressurized water nuclear power stations. Additional
studies are planned at newer and larger stations to provide applicable information
and to evaluate the effect of other operational and environmental conditions on
radiation exposures to the population.
William A. Mills
Acting Chief
Radiation Research
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Contents
Page
1. INTRODUCTION !
1.1 Need for Study !
1.2 Description of Study 1
1.3 References 2
2. RADIONUCLIDES IN WATER ON SITE 3
2.1 Water Systems and Samples 3
2.1.1 General 3
2.1.2 Main coolant system 3
2.1.3 Secondary coolant system 3
2.1.4 Paths of radionuclides from main and secondary systems 3
2.1.5 Other liquids on site 5
2.1.6 Samples 7
2.2 Analysis 7
2.2.1 General approach 7
2.2.2 Gamma-ray spectrometry 7
2.2.3 Radiochemistry 11
2.3 Results and Discussion 11
2.3.1 Radioactivity in main coolant water 11
2.3.2 Tritium in main coolant water 13
2.3.3 Fission products in main coolant water 15
2.3.4 Activation products in main coolant water 16
2.3.5 Radionuclides in secondary coolant water 17
2.3.6 Radionuclides in other liquids 18
2.4 References 19
3. RADIONUCLIDES RELEASED FROM STACK 21
3.1 Gaseous Waste System and Samples 21
3.1.1 Gaseous waste system 21
3.1.2 Radionuclide release 21
3.1.3 Sample collection 23
3.2 Analysis 24
3.2.1 Gamma-ray spectrometry 24
3.2.2 Radiochemical analysis 26
3.3 Results and Discussion , 26
3.3.1 Gaseous release in sampling main coolant 26
3.3.2 Gaseous effluent from secondary coolant 27
3.3.3 Gas surge drum contents 28
3.3.4 Radionuclide concentrations in the vapor container 29
3.3.5 Particulate radioactivity and radioiodine in the primary vent stack 30
3.3.6 Gaseous radioactivity in the primary vent stack 31
3.3.7 Particulate effluent from incinerator 31
3.3.8 Release limits and estimated annual radionuclide releases 32
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3.4 References 33
4. RADIONUCLIDES IN LIQUID EFFLUENT 35
4.1 Liquid Waste System and Samples 35
4.1.1 Liquid waste system 35
4.1.2 Radionuclide release . . 35
4.1.3 Samples 37
4.2 Analysis 37
4.2.1 Test tank solution . . . . . 37
4.2.2 Circulating coolant water 38
4.2.3 Yard-drain samples 38
4.3 Results and Discussion 38
4.3.1 Radionuclides discharged to circulating coolant water 38
4.3.2 Radionuclides in circulating coolant water 40
4.3.3 Performance of the ion-exchange columns for collecting radionuclides 42
4.3.4 Radionuclides in yard-drain effluent 42
4.3.5 Release limits and estimated annual radionuclide releases 43
4.4 References 44
5. RADIONUCLIDES IN THE AQUATIC ENVIRONMENT . 45
5.1 Introduction 45
5.1.1 Studies near Yankee 45
5.1.2 Deerfield River and Sherman Reservoir '..'.' 45
5.2 Tritium in Water 45
5.2.1 Sampling and analysis 45
5.2.2 Results and discussion 45
5.3 Other Radionuclides in Water 52
5.3.1 Unfiltered samples 52
5.3.2 Suspended solids 53
5.4 Radionuclides in Vegetation v . . . 54
5.4.1 Sampling and analysis 54
5.4.2 Results and discussion 54
5.5 Radionuclides in Fish 58
5.5.1 Collection and analysis 58
5.5.2 Results and discussion 60
5.5.3 Hypothetical radionuclide concentration in fish 61
5.6 Radionuclides in Benthal Samples . 62
5.6.1 Sampling and on-site measurements 62
5.6.2 Description of benthal samples 63
5.6.3 Analysis 64
5.6.4 Results and discussion of sample analyses '."."..' 67
5.6.5 Distribution of radionuclides in benthal
samples as function of particle size 68
5.6.6 Results and discussion of probe measurements 68
5.6.7 Significance of radioactivity in sediment . . 69
5.7 References 71
6. RADIONUCLIDES IN THE TERRESTRIAL ENVIRONMENT 73
6.1 Introduction 73
6.1.1 Sampling 73
6.1.2 Environment of Yankee 74
6.1.3 Meteorology and climatology 74
6.2 Estimation of Radioactivity Concentrations 74
6.2.1 Dispersion of 85Kr in air 74
6.2.2 Accumulation of 90§r in snow 75
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6.2.3 Accumulation of 90sr in vegetation 75
6.2.4 Accumulation of 90Sr on soil 75
6.2.5 Iodine-131 in cows' milk : 75
6.3 Radionuclides in Snow 76
6.4 Radionuclides in Vegetation and Soil 77
6.5 Radionuclides in Milk 78
6.6 Radionuclides in Deer 79
6.6.1 Sampling and analysis 79
6.6.2 Results and discussion 79
6.6.3 Hypothetical radiation dose from eating deer meat 81
6.7 External Radiation 81
6.7.1 Detection instruments 81
6.7.2 Measurements 81
6.7.3 Results and discussion . 84
6.7.4 Estimated external radiation exposure to persons in the environs 85
6.8 References 85
7. SUMMARY AND CONCLUSIONS 87
7.1 Radionuclides in Yankee Effluents 87
7.2 Radionuclides in Environment of Yankee '...'.' 89
7.3 Monitoring Procedures 90
7.4 Recommendations 90
APPENDICES:
Appendix A 93
Appendix B.I 94
Appendix B.2 95
Appendix B.3 96
Appendix B.4 97
Appendix B.5 97
Appendix C.I 98
Appendix C.2 98
Appendix C.3 93
Appendix C.4 99
Appendix C.5 99
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Figures
Page
2.1 Coolant Flow Schematic of Yankee PWR 4
2.2 Paths of Effluents at Yankee PWR 6
2.3 Gamma-ray Spectra of Main Coolant Water 8
2.4 Gamma-ray Spectra of Main Coolant Water 9
2.5 Gamma-ray Spectrum of Liquid in Waste Holdup Tank 10
2.6 Aluminum Absorber Curves of C and %i Separated from
Yankee Waste Holdup Tank Liquid ,11
2.7 Yankee Electrical Loading, April, 1968 Through February, 1971 14
3.1 Sources of Airborne Effluent 22
3.2 Gamma-ray Spectra of Gas Surge Drum Samples 24
3.3 Gamma-ray Spectra of Gas Released in Sampling Main Coolant 25
4.1 Liquid Waste Sources and Treatment 36
4.2 Aluminum Absorber Curve of Yankee Test Tank Sample 38
4.3 Gamma-ray Spectrum of Sand and Gravel from East Yard Drain 39
5.1 Deerfield River Near Yankee Nuclear Power Station 47
5.2 Yankee Nuclear Power Station 48
5.3 Yankee Nuclear Power Station Detailed Plan 49
5.4 Gamma-ray Spectrum of Water Moss 55
5.5 Gamma-ray Spectrum of Water Moss 56
5.6 Gamma-ray Spectrum of Dead Leaves from Sherman Reservoir 57
5.7 Gamma-ray Spectra of Bottom of Sherman Reservoir 65
5.8 Gamma-ray Spectra of Benthal Samples from Sherman Reservoir 65
6.1 Locations of Radiation Exposure Measurements with Survey Meters 83
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Tables
Page
2.1 Radionuclide Concentration in Main Coolant Water 12
2.2 Radionuclide Concentration in Secondary System Water „ 17
2.3 Radionuclide Concentration in Waste Holdup Tank on Oct. 4,1968 18
2.4 Radionuclide Concentration in Safety Injection Water, June 10,1970 18
3.1 Radioactive Gases Released to Stack during Depressurizing
Main Coolant for Sampling 27
3.2 Radioactivity Contents of Off-gas from Air Ejector at Main
Condenser in Secondary Coolant System 28
3.3 Gas Surge Drum Contents 28
3.4 Radioactivity in Vapor Container 29
3.5 Stack Releases of Particulate Radionuclides and Gaseous Iodine-131 30
3.6 Stack Effluent Release Rates During and After Gas Surge Drum Release 31
3.7 Particulate Radioactivity Emitted from Incinerator Stack, June 9,1970 32
4.1 Radionuclide Concentration in Test Tank before Discharge at Yankee 40
4.2 Radionuclide Concentration in Main-Condenser Circulating
Coolant Water on June 3,1969 41
4.3 Radionuclide Concentration in Yard Drains 42
5.1 Concentration of Stable Substances in Water from Deerfield River 50
5.2 Tritium Sampling Points 50
5.3 Tritium Concentration in Sherman Reservoir and Deerfield River 51
5.4 Gross Beta Activity and 90sr Concentration in Water from
Sherman Reservoir and Deerfield River 52
5.5 Gross Beta Activity and Concentration of 90$r and 13?Cs in
Suspended Solids from Surface Water in Sherman Reservoir 53
5.6 Radionuclides in Water Moss and Dead Leaves from Sherman Reservoir 54
5.7 Radionuclide Concentration in Water Moss and Dead Leaves 58
5.8 Fish Collected in Sherman and Harriman Reservoirs 59
5.9 Radionuclide and Stable Ion Concentration in Fish Tissue 59
5.10 Benthal Sampling Points 63
5.11 Mineralogical Analysis of Benthal Samples 64
5.12 Concentration of Radionuclides in Benthal Samples from
Sherman Reservoir and Deerfield River 66
5.13 Radionuclide Distribution in Dredged Benthal Samples as
a Function of Particle Size 69
5.14 Net Count Rate of 60co and 137cs with Nal (Tl) Underwater
Probe in Sherman Reservoir 70
5.15 Ratio of Count Rate by Underwater Probe to Radionuclide
Concentration in Benthal Samples 70
6.1 Radionuclides in Snow 76
6.2 Radionuclide and Stable Ion Concentration in Vegetation, June 4,1969 77
6.3 Radionuclide and Stable Ion Concentration in Soil, June 4,1969' 78
6.4 Radionuclide Concentration in Milk 78
6.5 Description of Sampled Deer 79
6.6 Radionuclide and Stable Ion Concentration in Deer Samples 80
6.7 External Radiation Exposure Rate Measurements near Yankee 82
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1. Introduction
J.I
for Studg
Each of the many nuclear power stations that will
soon be operating in the United States requires an
effective radiological surveillance program to assure
that radiation exposure to the population is within
acceptable limits. The Radiochemistry and Nuclear
Engineering Branch of the Environmental Protection
Agency (EPA)-formerly a part of the Bureau of
Radiological Health, Public Health Service -- has,
therefore, undertaken a program of studies at
commercial nuclear power stations to suggest
surveillance guidelines. The studies are intended to
provide the following information: (1) identity and
amount of radionuclides in effluents, (2) influence of
station operation on radionuclide discharges, (3)
degree of dispersion or concentration of radionuclides
in the environment, (4) relative importance of
specific radionuclides and vectors in exposing
population groups, (5) magnitude of radiation
exposure in the environment, and (6) applicability of
various monitoring and measuring techniques.
In the future, much of this information should be
available in response to the recent requirements by
the Atomic Energy Commission (AEC) that nuclear
power stations report semiannually the quantities of
discharged radionuclides and the environmental levels
of radiation and radioactivity that result from plant
operation.'- ' Until now, stations have reported
discharges mostly in terms of gross activity and
tritium. ' ' Few of the environmental surveillance
reports by the stations are publicly available, and
most of these, while indicating the absence of
significant radiation exposure through "less-than"
values, provide little guidance in planning other
monitoring programs. On the other hand, much
general information is available on environmental
surveillance for radionuclides (see footnote, Section
1.3), including several recent publications concerning
nuclear facilities. ' > '
The work described here was performed at the
Yankee Nuclear Power Station, a pressurized water
reactor (PWR). Yankee was built at Rowe,
Massachusetts by the Westinghouse Electric Corp.,
and operates at a maximum power of 185 megawatts
electric (MWe) and 600 thermal megawatts (MWt). It
had produced more than 1 x 10? megawatt-hours
between 1960 and 1969, and had passed through
seven fuel cycles. The fuel is enriched (4.9 percent
2350) uranium oxide (UO2) pellets, clad in stainless
steel. The operation of the station has been described
by several authors. (10-12)
The study at Yankee follows one performed at the
Dresden Nuclear Power Station, U3)a boiling water
reactor (BWR) that began operation in 1959 and has
been producing power at a rated capacity of 210
MWe. At present, a study is in progress at one of the
newer and larger PWR's, and one is being planned at a
large new BWR. In the meantime, it is believed that
many of the reported observations are applicable to
planning radiological surveillance at the newer BWR's
and PWR's. Caution should be exercised, however, in
applying the reported discharge data to newer
stations, because aspects of both design and operation
tend to differ among stations. For example, even
gross activity values indicate that, among commercial
nuclear power stations, Yankee discharges unusually
small amounts of radionuclides other than tritium/')
1.2 Description of
The study at the Yankee Nuclear Power Station
was planned and performed by the Radiochemistry
and Nuclear Engineering Branch, supported by staff
of the Divisions of Surveillance and Inspection, and
of Technology Assessment, in the Office of Radiation
Programs, EPA. The Yankee Atomic Electric
Company, which operates the station, the
Massachusetts Department of Public Health (MDPH),
and the Division of Compliance of the AEC
cooperated in the study. A field trip to Yankee was
undertaken on June 3-4, 1969; other samples were
obtained on October 4, 1968, April 1, 1969, July 10
1
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and 29, 1969, June 4 and 10, 1970, November 19,
1970, and February 9,1971. Participants in the study
are listed in Appendix A.
As in the study at Dresden, measurements of
radionuclides at the station, in effluents, and in
environmental media were coordinated to attempt to
show relative magnitudes among these three
categories, critical radionuclides or pathways, and
indicator radionuclides or media. Detailed
descriptions are provided to demonstrate monitoring
procedures. At Yankee, however, the amounts of
discharged radionuclides were so small that only in
the Sherman Reservoir, which receives liquid wastes,
could radionuclides attributable to station effluents
be detected. Although results of radionuclide analyses
in other environmental media are reported, the most
detailed discussion of environmental sampling,
therefore, pertains to the aquatic environment.
Planning was guided by the available data on
radionuclides in effluents and the environment, and
an attempt was made to avoid duplicating ongoing
programs. Monthly operating reports by Yankee
Nuclear Power Station contain gross beta-gamma and
tritium discharge values. Gross alpha activity, gross
beta activity, tritium concentrations, and some
gamma-ray spectral analyses are reported annually by
Yankee's contractor for environmental surveillance.
The MDPH reports gross beta activity in water and
concentrations, of photon-emitting radionuclides in
benthal deposits. These data are cited in the
appropriate sections of this report.
1.3 References*
1. U. S. Atomic Energy Commission, "Standards
for Protection Against Radiation", Title 10, Code of
Federal Regulations, Part 50, Federal Register 35,
18388(1970)
2. Blomeke, J. O. and F. E. Harrington,
"Management of Radioactive Wastes at Nuclear
Power Stations", AEC Rept. ORNL-4070 (1968).
3. Brinck, W. L. and B. Kahn, "Radionuclide
Releases at Nuclear Power Stations", in
Environmental Surveillance in the Vicinity of Nuclear
Facilities, W. C. Reinig, ed., C. C. Thomas,
Springfield, 111., 226-233 (1970).
4. "Management of Radioactive Wastes at Nuclear
Power Plants", Safety Series No. 28, International
Atomic Energy Agency, Vienna (1968).
5. Logsdon, J. E. and R. I. Chissler, "Radioactive
Waste Discharges to the Environment from Nuclear
Power Facilities", Public Health Service Rept.
BRH/DER 70-2 (1970).
6. Thompson, T. J., "Statement on the
Environmental Effects of Producing Electric Power",
in Environmental Effects of Producing Electric
Power, Part 1, Hearings of the Joint Committee on
Atomic Energy, U. S. Gov't. Printing Office,
Washington, D. C., 175-194(1970).
7. "AEC Report on Releases of Radioactivity
from Power Reactors in Effluents During 1969", in
Environmental Effects of Producing Electric Power,
Part 1, Hearings of the Joint Committee on Atomic
'Energy, U.S. Gov't. Printing Office, Washington,
D.C., 2316-2317 (1970).
8. Voilleque, P. G. and B. R. Baldwin, eds.,Health
Physics Aspects of Nuclear Facility Siting, B. R.
Baldwin, Idaho Falls, Idaho (1971).
9. Environmental Aspects of Nuclear Power
Stations, International Atomic Energy Agency,
Vienna (1971).
10. Coe, R., "Nuclear Power Plants in Operation.
4. Yankee-Rowe", Nuclear News 12, No. 6, 54 (June
1969).
11. Coe, R. J. and W. C. Beattie, "Operational
Experience with Pressurized-water Systems", in
Proceedings of the Third International Conference on
the Peaceful Uses of Atomic Energy, Vol. 5, United
Nations, New York, 199-206(1965).
12. Kaslow, J. F., "Yankee Reactor Operating
Experience", Nuclear Safety 4, 96 (1962).
13. Kahn, B. et al, "Radiological Surveillancf
Studies at a Boiling Water Nuclear Power Reactor"
Public Health Service Rept. BRH/DER 70-1 (1970)
*References that provide guidance for environmental surveillance, information on waste.management, at
nuclear facilities, and discussion of the transfer of radionuclides in the environment are listed in Section
1.3 of Reference 13.
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2. Radionuclides in Water on Site
2.1 Water Sgstems amd
Samples
2.1.1 General. A PWR such as Yankee has three
consecutive cooling systems, shown schematically in
Figure 2.1. In the main or primary system, water is
heated under pressure in the reactor, circulates
through four parallel steam generators, and returns at
lower temperature to the reactor. In the secondary
system, steam formed in the steam generators passes
through the turbine to produce power and is then
cooled to form water in the condenser. The water is
then returned to the steam generators. In the third
system, circulating coolant water is pumped from the
bottom (25-m depth) of Sherman Reservoir through
the secondary-system condenser at the rate of
530,000 liters/min (140,000 gal/min), and returned
to the surface of the reservoir.O)
2.1.2 Main coolant system. U > 2) The main coolant
is 64,000 kg of water that circulates approximately
once every 12 seconds. In addition to the four high
pressure loops for steam generation, the system
includes the lines, shown in Figure 2.1, by which
water is added or withdrawn for pressure control,
chemical adjustments, continuous purification, and
sample collection. At the time of the study, the flow
rate through the purification filter and demineralizer
was 113 kg/min (30 gal/min), (2) which results in a
mean turnover period of 64,000 kg 4 113 kg/min =
570 min (3.4 x 1Q4 sec) for main coolant water.
The water for the system is taken from Sherman
Reservoir and demineralized. During two-thirds of the
operating cycle of approximately 16 months, the
main coolant water contains boron (in the form of
boric acid) to control the neutron flux, and is, as a
consequence, at a pH value of approximately 5. The
boron concentration is decreased gradually during
this period from 1,300 parts per million (ppm) to 0
ppm. During the final "power stretchout" period of
the operating cycle, the water contains no boron, but
ammonium hydroxide is added to maintain the pH
value at approximately 9 for corrosion control.
Nitrogen gas is added to the system to maintain
the concentration of oxygen from ambient air at a
low concentration, and hydrogen gas is added to
depress below 0.1 ppm the concentration of oxygen
formed by the radiation-induced decomposition of
water. The concentration of nitrogen is
approximately 4 cc/kg of water at standard
temperature and pressure (STP) in the absence of
ammonium hydroxide, and 12 cc/kg of water at STP
with ammonium hydroxide in the coolant. The
concentration of hydrogen is approximately 35 cc/kg
of water at STP.
During refueling, the space above the opened
reactor vessel is flooded, and fuel is carried to the fuel
transfer pit on a carriage through the transfer chute.
Upper and lower lock valves in the transfer chute
reduce the movement of water from reactor to fuel
transfer pit. The water is continuously purified by
passage through the shutdown demineralizer.
2.1.3 Secondary coolant system. 0>2) The
secondary coolant is 190,000 kg (420,000 Ib) of
water that circulates approximately once every 10
minutes. The water is obtained from Sherman Res-
ervoir and demineralized. Accumulation of salt in the
steam; generators is minimized by a relatively small
continuous blowdown and a more massive blowdown
for one to two hours every night. The rate of
continuous blowdown depends on the salt content of
water in the steam generators, and averages
approximately 2,000 liters/day; the once-nightly
blowdown is approximately 12,000 liters/day.
2.1.4 Paths of radionuclides from main and
secondary systems. The radionuclides in the main
coolant water are fission products and activation
products. The fission products in the water are
formed within the uranium oxide fuel and enter the
water through small imperfections in the stainless
steel cladding of the fuel elements. Other possible
sources of fission products-apparently minor at
Yankee-are fuel that contaminates the surface of new
fuel elements ("tramp uranium") and fuel that reaches
the main coolant water from failed fuel elements.
Activation products in water are formed by neutron
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18.3 X to6 Kg/hr
PRIMARY
LOOPS
REACTOR
FEED AND
BLEED HEAT
EXCHANGER
O CHARGE
PUMPS
— MAKEUP
LOW
PRESSURE
SURGE TANK
(21.280 I.)
TO
WASTE HOLDUP
-». TANKS AND
PRIMARY DRAIN
COLLECTING TANK
COOLANT PURIFICATION SYSTEM
38-380 hg/min
1.1 X ID6 kg/hr
AIR EJECTOR
SECONDARY
LOOP
CIRCULATING
WATER
POKIER - 600 MNt
WATER VOLUME
PRIMARY SYSTEM - 64,000 kg
(83,000 I at 2000 psia
and 263 to 284°C)
SECONDARY SYSTEM —190,000 kg
Figure 2.1. Coolant Flow Schematic of Yankee PWR.
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irradiation of the water and its contents (including
gases and dissolved or suspended soils) and of reactor
materials that subsequently corrode or erode.
The radionuclides in main coolant water circulate
and decay within the system, deposit as crud (which
may later recirculate), are retained by the purification
filter and demineralizer (which are periodically
replaced and shipped off-site as solid waste), or leave
the system with gases and liquids. Paths from the
system to the environment are shown schematically
in Figure 2.2.
Under routine operation, water and associated
gases leave the main coolant system at leaks and by
intentional discharge from the low-pressure surge
tank for pressure control, boron-concentration
adjustment and sample collection. The water passes
through the primary drain collecting tank and is
stored in a waste holdup tank. When a sufficient
volume has accumulated, the liquid waste is treated in
an evaporator. The distillate is discharged to effluent
circulating coolant water, and the residue is shipped
as solid waste. The water released during refueling is
also processed in the evaporator. Gas from the liquid
waste system is collected in the gas surge drum. The
gaseous waste system is described in Section 3.1 and
the liquid waste system, in Section 4.1.
Liquid waste from the reactor plant was estimated
to consist of the following constituents in 1969:
main system leakage
toutine opeiation of main system
refueling
incineiator rotoclone (2)
decontamination (2)
tank moats w
total reactor plant
(see Section 4.1)
0.2 x 106 liters/year
0.9
0.8
0.4
0.1
02
2.6 x 106 liters/year
The main system leakage was an average of 1 percent
per day of the system contents. (2) The discharge
during routine operation includes of the order of 1 x
105 liters for each major reduction of boron con-
centration during startup or for power stretchout.
Radionuclides enter the secondary coolant system
through leaks in the steam generators. Normally, the
leakage rate is only a few liters per day in each of the
four loops, but occasionally the leakage rate increases
rapidly until the faulty tubes are plugged. As a rule of
thumb, tubes are plugged before the leakage rate
approaches 4,000 liters/day; high leakage rates, re-
quiring loop isolation and tube repair, have occurred
on four occasions in the period 1960-1969.(2).
Secondary system liquid waste consisted of 15 x
106 liters/year in 1969 and 1970, of which
approximately 37 percent was blowdown, 57 percent
was leakage, and 6 percent, discharge from the spent
fuel pit and waste tank moats. Through combined
leakage and blowdown, 19 percent per day of the
system water was discharged. The water is discharged
directly, without storage or treatment, to effluent
circulating coolant water.
2.1.5 Other liquids on site. During and after
refueling, used fuel elements are stored in
demineralized water in the fuel transfer pit. The
water is circulated through the fuel-pit ion-exchanger
for purification, and through a cooler to control the
temperature. After use, water from the fuel transfer
pit is discharged with secondary system liquid wastes.
The waste holdup system consists of two tanks
with a total capacity of 570,000 liters (150,000 gal).
Reactor plant liquid wastes are pumped into these
tanks from the primary drain collecting tank, and are
stored until treated in the evaporator. An 18,000-liter
(5,000-gal) gravity drain tank collects other reactor
plant liquid wastes for treatment in the evaporator.
Component cooling water is circulated in the
neutron shield tank and other components at the
station. The water contains approximately 400 ppm
potassium chromate as corrosion inhibitor. It has not
been discharged from the station. (2)
Safety injection water, containing 1 percent (by
weight) boric acid, is stored in a 470,000-liter
(125,000-gal) tank to be available for cooling the
reactor core during a major loss-of-coolant accident.
It is used to flood the shield-tank cavity during
refueling and is returned to the storage tank after
/*\
refueling. It is not normally discharged Az/
The incinerator at Yankee utilizes a mechanical
centrifugal scrubber (rotoclone) to moisten and retain
dust particles from the exhaust air steam. In 1969,
approximately 4 x 10-> liters of water were used in
the scrubber. W This water is stored and processed
by evaporation with main coolant liquid waste (see
Section 2.1.4).
The sanitary-system water at Yankee is passed into
a septic tank on the site. Normally, it would not be
contaminated with radioactivity.
Because Yankee is located between a ridge and the
Sherman Reservoir, rain water runs off across the site.
Two yard drains lead into the reservoir (see Section
4.2.3). In addition, some water collects in storage
tank moats and is treated as necessary (see Section
2.1.4). In 1969, approximately 200,000 liters of
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Effluent Gas
Effluent Liquid
PRIMARY
VENT
STACK
PRIMARY
AUXILIARY
BUILDING
Circulating <
Coolant
Wafer
Incinerator
Vent.
Stack
SHERMAN RESERVOIR
Figure 2.2. Paths of Effluents at Yankee PW.
-------
water from the moats were stored and evaporated
with reactor plant wastes, and 300,000 liters were
discharged with secondary plant liquid waste. (2)
2.1.6 Samples. To identify potential radioactive
effluents, liquids at the Yankee Nuclear Power
Station were sampled within the plant, where
radionuclides were at much higher concentrations and
therefore more easily detected than at the point of
release. The following water samples were provided
by Yankee staff in plastic bottles:
(1) main coolant,! liter, collected Oct. 4,
1968 at 1300;
(2) main coolant, 2 liters, collected July 10,
1969 at 0840;*
(3) main coolant, 2 liters, collected June 10,
1970 at 0945;
(4) waste holdup tank, 1 liter, collected Oct.
4,1968 at 1000;
(S) continuous steam generator blowdown, 1
liter, collected Oct. 4,1968 at 0950;
(6) continuous steam generator blowdown,
3.5 liters, collected June 10, 1970 at
1000;
(7) secondary system condensate, 2 liters,
collected July 10,1969 at 0855;*
(8) secondary system condensate discharge, 1
liter, collected June 10, 1970 at 1120;
(9) component cooling system, 3.5 liters,
collected June 10,1970 at 1100; and
(10) safety injection tank, 3.5 liters, collected
June 10,1970 at 1045.
One liter each of Samples (2), (3), and (7) was
acidified with 100 ml cone. HNO3 to reduce
deposition of radionuclides on the walls of the bottle;
the other liter of sample remained unacidified to
prevent loss of radioiodine. Sample (6) was filtered at
collection, and the filter and filtrate were analyzed
separately.
2.2 Amalgslm
2.2.1 General approach. Aliquots of all samples
were first counted for gross alpha and beta activity,
then examined with gamma-ray spectrometers, and
finally analyzed radiochemically. Analyses were
performed for high-yield fission products and
common activation products in reactor water.
Relatively short-lived radionuclides could not be
measured because of radioactive decay between
sampling and analysis: the main coolant water of Oct.
4, 1968 was first analyzed 20 hours after sampling;
the other two main-coolant samples, after 2 days; and
the other samples, usually after 1 week. Aliquot
volumes ranging from less than 1 ml to 200 ml were
used.
A special effort was made to measure
radionuclides that, because they emit only weak beta
particles, tend to be underestimated by gross beta
counting and are not detected by gamma-ray
spectrometry. These radionuclides were 12.3-yr %
(maximum beta particle energy, 18 keV), 5,730-yr
14C (158 keV), 88-d 35s (167 keV), 92-yr 63tfi (67
keV), and 1.6 x 10?-yr 129i (150 keV).
Radionuclide concentrations were computed from
count rates obtained with counters calibrated as
functions of gamma-ray or average beta-particle
energies. All values were corrected for radioactive
decay and are given as concentrations at sampling
time. The values of decay rates and branching ratios
are from recent publications. (4,5,6)
Loss of radionuclides from the samples by
volatilization and adsorption on container walls was
given special consideration. C^) Concentration values
for radioiodine and *-^C were obtained with
unacidified samples. For other radionuclides in
main-coolant water, data from analyses of the
acidified samples and aqua-regia leaches of the
containers were combined. Results of analyzing the
filtered blowdown water and the filter (sample No. 6)
were also combined.
2.2.2 Gamma-ray spectrometry. Radionuclides
that emit gamma rays were identified in aliquots of
main coolant water by multichannel spectrometry
with a Ge(Li) detector (see Figures 2.3 and 2.4).
*We thank G. J. Karches, Northeastern Radiological Health Laboratory, Public Health Service (NERHL,
PHS), for obtaining these samples.
-------
1,000
100 -
10
1.0
0.1
0.01
Oct. 5, 1968 (0952-1132)
Oct. 13-14, 1969 (1,000 minutes)
June 6-7, 1969 (1,000 minutes)
*-"-CMOCM
oc
_
to eo o>-- CM CM >»
^-100
-COC°° ^^-05 ^^^CMW
. CD o
CM CMCMCM ^
to n eor- »
in S —« *
03 (0*0 M
CO CD ^
CD C3 CM
to r^ r-
ura ra
___.._ _. ^ H- ^ O CD H-H- ^ ^-Xh-
encscsi mr^E cq » Ujco CM CM en in CM
SSS «£ g 2 SS22 2»!2»
JIKJi
M6 Z
in c3m
ri i I'M si"
CM ^. CO •- ° CD O
CM fs| e*3 CO ^ ^* *~
x t-
*- c»>
CO CO
o o
*rco
cp in
I
100
200
300 400 500
CHANNEL NO. (1.01 keV/channel)
600
700
Figure 2.3. Gamma-ray Spectra of Main Coolant Water, 40 - 808 keV.
Detector: Ge(Li), 10.4 cm2 x 11 mm, trapezoidal
Sample : 35 ml (including 1 ml acid), collected Oct. 4, 1968 at 1300.
Counts : At times indicated on spectra; background (bkgd) not subtracted.
800
-------
O
U
0.001
0.01
1,000
1,100
1,200
1.300
1,400
1,500
1,600
CHANNEL NO. (1.01 keV channel)
Figure 2.4. Gamma-ray Spectra of Main Coolant Water, 808 - 1,616 keV.
n
Detector: Ge(Li), 10.4 cm x 11 mm, trapezoidal
Sample : 35 ml (including 1 ml acid), collected Oct. 4, 1968 at 1300.
Counts : At times indicated on spectra; background (bkgd) not subtracted.
-------
800
100
900
200
1,000
300 400
1,100 1,200
CHANNEL NO. (1.006 keV/channel)
500
1,300
600
1,400
700
1,500
800
1,600
Figure 2.5. Gamma-ray Spectrum of Liquid in Waste Holdup Tank.
Detector: Ge(Li), 10.4 cm2 * 11 mm, trapezoidal
Sample : 35 ml (including 1 ml acid) collected Oct. 4, 1968.
Count : Oct. 10-11, 1968 (1,000-minute background not subtracted).
-------
Spectral analyses were repeated at appropriate
intervals to measure long-lived radionuclides in the
main coolant without interference by short-lived ones
and to measure half lives for confirming the identity
of the radionuclides. The minimum half life of the
measured radionuclides was 6 hours, and the
maximum, 30 years. Minimum detectable
concentrations at these half lives were approximately
1 x 10-4 and 1 x 10-5 microcurie per milliliter
), respectively.
10,
5,
2,
1,
000
000
000
000
500
200
100
50
20
10
5
2
1
0.5
0.2
0.1
•TT
i i i
LEGEND
• 63
U
Ni SAMPLE
C SAMPLE
'4C AND 63Ni STANDARDS
0
8 12 16 20 24
SURFACE DENSITY, mg/cm2
28
Figure 2.6. Aluminum Absorber Curves of
14C and 6%i Separated from
Yankee Waste Holdup Tank Liquid.
Detector: Low-background G-M end-window.
Samples : C'.unacidified 5 ml aliquot
of sample collected Oct. 4,
1968; 63Ni:acidified 25 ml
aliquot.
Counts : April U-15. 1970. 200 min. at
each point.
The main-coolant sample of Oct. 4, 1968, was
analyzed by obtaining 6 spectra in the interval from
0.9 to 250 days after sample collection. Main coolant
samples collected on July 10, 1969, and June 10,
1970, were analyzed similarly, but gamma rays from
relatively short-lived radionuclides were obscured
because the initial spectrum could not be obtained so
O/i
soon after sampling and the z^Na content was
relatively high. Analysis by Ge(Li) spectrometry was
also performed for the sample from the waste holdup
tank (see Figure 2.5).
All other water samples were analyzed by
multichannel spectrometry with a 10 cm x 10 cm
Nal(Tl) detector. These samples contained fewer
radionuclides at much lower levels of radioactivity.
Hence, the higher energy resolution of the Ge(JJ)
detector was generally unnecessary, and the higher
counting efficiency of the NaI(Tl) detectors was
advantageous.
2.2.3 Radiochemistry. Radiochemical analysis was
performed to confirm spectral identification by
gamma-ray energy and half life, measure
radionuclides more precisely and at lower
concentrations than by instrumental analysis of a
mixture, and detect radionuclides that emit only
obscure gamma rays or none at all. After chemical
separation, the following detectors were used: NaI(Tl)
crystal plus spectrometer for photon-emitting
radionuclides; low-background end-window Geiger-
Mueller (G-M) counter for 14c, 32p, 35s, 89sr,
90Sr, 129J, and 185)^; liquid scintillation detector
plus spectrometer for 3R and 63>Ji; and xenon-filled
proportional counter plus spectrometer for 55pe.
Measurements with the G-M detector includes
observation of the effect of aluminum absorber on
count rates to determine maximum beta-ray energies
(see Figure 2.6) and thus confirm radionuclide
identification.
2.3 Rv*mlt* and Di»cm*»iim
2.3.1 Radioactivity in main coolant water. Tritium
was by far the most abundant of those radionuclides
with half lives longer than 6 hours (see Table 2.1). At
second highest concentration was ^^Na, some of
which may have been formed in sodium salt added at
times by Yankee staff as leak tracer. (2) The sum of
all other measured radionuclides was between 0.003
and 0.005 MCi/ml. The average gross activity
(without ^H) reported by Yankee (see below) was
II
-------
Table 2.1
Radionuclide Concentration in Main Coolant Water, fjCi/ml*
Radionuclide
12.3 -yr3Ht
50.5 -d 89Sr
28.5 -yr90Sr
9.7 -hr91Sr
65 -d 95Zrt
QC
35.1 -d 95Nbt
66.2 -hr"Mot
8.06-d 131I
20.9 -hr 133I
6.7 -hr 135I
5.29-d 133Xe
9.1 -hr135Xe
2.07-yr134Cst
30 -yr137Cs
12.8 -d 140Ba
5730 -yr 14C
15.0 -hr24Na
14.3 -d 32P
88 -d 35S
27.7 -d 51Cr
313 -d 54Mn
2.7 -yr55Fe
44.6 -d 59Fe
270 -d 57Co
71.3 -d 58Co
5.26-yr 60Co
92 -yr 63Ni
12.8 -hr64Cu
253 -d llomAg
2.7 -d 122Sb
60.2 -d 124Sb
42.5 -d 181Hf
115 -d 182Ta
5.1 -d 183Ta
75 -d 185W
24 -hr187W
Oct. 4, 1968
5.0
7 xlO-6
2 x 10-7
9.8 x 10-5
4.5 x 10-5
5.0 x 10-5
1.1x10^
5.2 x 10-5
6.6 x 10^
9.1 x 10-4
2.4 x 10-4
2.5 x 10-4
3 x 10-7
2 x 10-7
1.2 x 10-5
July 10, 1969
from fuel
4-OxlO-1
<5 x 10-7
8 x 10'7
NA
6 xlO"6
6 xlO-6
4.6 x ID'5
3.0 x 10'5
4.0 x 10^
NA
1.8 x 10-5
NA
8 x 10-7
3 xlO-6
<1 xlO-6
June 10, 1970
1.8
1 xlO-6
2 x lO'7
NA
2 xlO-6
1 xlO-6
1.9 x 10-4
5.5 x 10-5
8.0 x 10^
NA
<1 xlO-5
NA
3 x lO'7
1 x lO'7
1 xlO-6
from activation of water, cladding, and construction materials
NA
1.6 x 10-3
1.0x10-5
NA
8.5 x 10-4
5.4 x 10-4
l.OxlO-4
1.9 x 10-4
~6 xlO'7
3.4 x 10-4
9.0 x 10-5
1.9 x 10-5
~2 xlO-4
1.9x10-5
4.9 x 10-5
2.0 x 10-5
~8 x 10-6
~6 x lO-5
l.lxlO"4
1.0 x 10-5
4.8 x 10^
1.5 x 10-5
1.8 x 10'2
2.5 x 10-5
3 xlO-6
2.1 x 10-4
2.4 x ID"4
3.7 x 10-4
8.9 x 10-5
2 xlO-6
1.0x10-3
1.3x10-4'
4 xlO-6
1 x 10-5
<1 xlO-6
2 x 10-6
1 xlO-6
~4 x 10-6
~1. 1x10-5
1.5 x 10-5
1.4x10-5
2.6 x 10^
8 xlO-6
3.8 x 10"2
6.5 x 10-5
3 xlO-6
5.0 x 10-5
4.0 x 10-5
1.0 x 10^
2.0 x 10-5
< 1 x 10-6
8.0 x 10-5
1.2 x 10-5
1 xlO-6
1.4 x 10-3
1.7 x 10-5
2.5 x 10^
4.0 x 10-5
NA
~3 xlO-6
NA
NA
NA
*Concentrations at time of sampling.
3H is also an activation product; 95zr, its daughter 95Nb, and 99Mo may aiso be activation products;
*34Cs is produced by the (n,y) reaction with fission-produced l"Cs.
Notes: 1. NA = not analyzed.
2. < values are 3& counting error.
3. The following fission products were not detected « 1 x lfl-6 AtCi/ml): 93Y, 97zr, 103RUj 106RU>
127Sb, 129I, 132Te, 141Ce, 143Ce, l44Ce, 14?Nd. The radionuclides 65zn, 136cs, and
239Np were also not observed at this minimum detectable level.
4. No gross alpha activity could be detected «\ x 10'9
12
-------
approximately 0.1 MCi/ml; this presumably consists
mostly of radionuclides with half lives shorter than 6
hours.
Fission products contributed only a small fraction
of the non-tritium radioactivity, and many relatively
long-lived high-yield fission products could not be
detected at the limiting sensitivity of approximately 1
x lO"6 MCi/ml (see footnote 3 to Table 2.1). Most of
the other radionuclides are neutron activation
products that have been reported earlier .(7,8) They
are formed in water, steel, copper, silver (in the
original Ag-In-Cd control rods), antimony (in the
Sb-Be neutron source), hafnium (in new Hf control
rods) and zirconium (in Zircaloy-2 cladding of
control rods). In addition to activation products
previously reported in power reactors, 14C, 35s, and
63Ni were found at relatively low concentration. No
radionuclides that emit alpha particles were detected
( <1 x 10-9/iCi/ml).
Considerable differences among the samples in
radionuclide concentrations were expected because,
during a fuel cycle, the boron concentration
decreases, the pH value increases, and the power level
decreases toward the end (see Figure 2.7). The
monthly average values reported by Yankee (9> 10>
* 1) at the sampling periods are:
October 1968 July 1969 June 1970
core
month of cycle
power level, MWe
boron, ppm
pH
radioactivity,
MCi/ml
Vll
6th
182
585
5.3
0.084
VII
15th
130
0
9.4
0.085
VIII
9th
169
183
6.7
0.108
Radionuclide concentrations are also affected by
many other variables, including the quality of the fuel
elements, the occurrence of shutdowns, the rate of
coolant-water purification and turnover, and the
extent of accumulation of radioactive material within
the coolant system.
2.3.2 Tritium in main coolant water. Measured
concentrations are consistent with the reported
monthly average values shown in Appendix B.I. The
sources of tritium at Yankee are believed to be
known, but the contribution of each has not been
quantified. These sources are:
(1) ternary fission within the fuel. The gen-
eration rate of 85 /tCi/sec at 600 MWt
was computed from a fission yield of 9.5
x 10-5 (see Appendix B.3). Other values
of the fission yield, (12) ranging from 8 x
10-5 to 13 x id"5 would change the
computed generation rate proportionally.
(2) the fast neutron reaction lOfi (n,2«) 3H
in main coolant water. A generation rate
of 4/tCi/sec at 600 MWt (185 MWe) and
a boron concentration of 1,300 ppm was
derived from predicted values for a
reactor at 1,000 MWe and 1,500 ppm
B,(13) and also from values for a reactor
at 1,000 MWt and 1,200 ppm B. (14)
(3) the two-step reaction 10B (n,a) 7Li (n,
not) 3H in main coolant water. This
reaction appears to contribute only a
small fraction of the •% produced in the
above-cited direct reaction, 0^) although
Ray(13) indicates that it is important if
^Li accumulates in the coolant.
(4) the reaction 2H (n,7 ) 3H in main
coolant water. This was computed to
produce 3H at the relatively low rate of
0.06 MCi/sec in a slightly larger boiling
water reactor. (')
(5) reactions (2) and (3) in the eight shim
rods that are located near the periphery
of the reactor vessel and consist of
Ziracaloy-clad steel with 1.2 percent
boron. The remote location, small
amounts of boron relative to that initially
in the coolant, and the cladding, which
appears to be a good barrier against
3H,(7) suggest that this source is minor.
(6) other reactions, such as 3He (n,p) 3n
(where the 3He is produced by the
beta-decay of 3H ) and lOfi (n,d) 9fie (n,
a) 6Li (n,«) 3H. The former accounts for
less than 0.5 percent of reactions (2) and
(3); (15) the latter is believed to produce
a negligible amount of 3H. (13)
On the basis of the above evaluation, only the first
two-ternary fission combined with high fractional
transfer from fuel to water, and the l^B (n,2a) 3H
reaction-are important sources of tritium at Yankee.
The 1,400 curies of tritium discharged annually in
1970 and 1969 (see Section 4.1.2), if produced
during approximately 320 days (2.8 x 10? sec) of
operation per year, indicate an average generation
rate of 50 /"-Ci/sec. This value suggests that ternary
fission produced more than 90 percent of the
discharged 3H, and that approximately one-half of
13
-------
(0
S
UI
o
o
UI
~<
g UJ
§
8
u.
20CL
mg
1 ww
~ 150.
V
Z
0)
o
o:
o 100,
"*"^
ir
UJ
o
Q.
uj 50.
o
(£
UJ
^*
^*
0
<• 3 cc o: a: a: 5 Q: &
-------
the 3H formed by ternary fission moved from fuel
into coolant water. Alternatively, either the
generation rate of 4/iCi/sec by the 10u (n, 2«^) 3H
reaction is vastly underestimated, at least one of
sources (3) to (6) is not negligible, or another source
of tritium exists in the reactor.
The monthly average 3H concentrations measured
in main coolant water by Yankee (see Appendix B.I)
yield conflicting results concerning the major source
of the 3H. The concentration of 3H is related to its
rate of production in, or transfer to, the main coolant
by:
C=R(1 -exp-At)(VX)-1
- At
(2.1)
where C
R
radionuclide concentration
in main coolant, /xCi/ml
rate of production in or
transfer to main coolant,
water turnover constant
(1.2 x 10'7 sec-1 at 1
percent per day) +
radionuclide decay constant
(1.78 x lO^sec'1 for 3H)
reactor operating period,
sec
main coolant water volume
(6.4 x 107 ml)
radionuclide concentration
at t = 0,
t
V'
cr
Thus, after continuous operation for, say, five
months at maximum power and boron concentration,
the ^H concentration from dissolved boron would be
CB = 4 (1 - 0.21) (6.4 x!07x 1.2 xlO'7)-1
= 0.4 ,x Ci/ml (2.2)
while that from ternary fission at 50 percent leakage
from fuel would be
Cfp = 85 x 0.5 (1 - 0.21) (6.4 x 107 x 1.2 x lO'7)'1
= 4.4 /i Ci/ml (2.3)
These computed values are consistent with the
magnitude of measured averages toward the beginning
of the fuel cycle. Concentrations at the end of the
fuel cycle-when no boron was in the coolant and
special efforts had been made to change to fresh
water-were considerably lower in 1969 and 1970,
however, than would be computed by equation 2.3,
even when the- lowered power levels and briefer
period of operation are taken into account.
This inconsistency may be due to the previously
discussed alternative modes of tritium formation, or
to the influence on the fractional transfer of 3H from
fuel to water by other factors, such as %
accumulation, local temperature, and surges of %
through cladding during start-up. Continuing
observations of 3fl[ levels at Yankee, and a current
study at the Ginna PWR (16) - where the fuel
elements are Zircaloy-clad-may provide quantitative
information on the sources of %[.
2.3.3 Fission products in main coolant water. The
concentrations and atom ratios of 131j relative
to 133j jn the three samples were comparable to
average monthly values reported by Yankee: (9-11)
Oct, 1968
July, 1969
June, 1970
Oct. 1968
July, 1969
June, 1970
This report
5.2 x 10-5
3.0 x 10-5
5.5 x 10-5
Yankee
monthly avg.
2.3 x 10-5
2.0 x 10-5
3.4 x 10-5
1311/1331, atom/atom*
This report
0.73
0.69
0.64
Yankee
monthly avg.
0.83
0.44
0.51
x half life ) 13 lj/ (&. x half life ) 133(
The 6-hour j COuld be measured only in the one
sample that was analyzed promptly after collection.
The concentration of 1.6 x 107-yr 1*9 j was fog low
detectable levels in all three samples, as is expected
from the low production rate (more than a
billion-fold lower than ^Ij).
According to the very low ratios of turnover rates
of the fission products in main coolant water to
production rates in fuel (see Appendix B.4), an ex-
tremely small fraction of these radionuclides moved
from fuel to coolant. In the Oct. 4, 1968 sample,
ratios ranged from 0.5 x 10-9 to 22 x lO*9 with an
average of 8.2 x 10-9; in the two other samples, most
ratios were equal or lower. The similar ratios for 131^
133], and 13$!, despite their different half lives,
suggest that the composition approached that of a
"recoil mixture". This results from the rapid transfer
of fission products from fuel to water.
The main-coolant water samples also contained the
radioactive gases l^Xe an^ 135xe. The values
shown in Table 2.1 refer to concentrations in excess
of those due to the decay of 133j ^j 135j m ^
15
-------
133i
135i
samples, and show that a fraction of these radioactive
gases remains in water despite its high temperature
and rapid movement. The values are not quantitative
in view of the probable loss of xenon from water
during sample collection and aliquot removal.
Concentrations of 133xe and 135Xe in the gas phase
are given in Section 3.3.1 for the same June 10,1970
sample whose aqueous phase is described in Table
2.1.
The measured concentrations of fission products
were 102 - to 106- fold lower than was predicted by
Yankee for 1 percent of fuel rods with pin holes or
small cracks, at a 38-liter/min flow rate through the
purification system:
Predicted Measured (Oct. 4)/
Radionuclide concentration, (1 )/u Ci/ml predicted, percent
89Sr 0.036 0.019
0.029 0.38
1.6 0.003
2.1 0.031
0.94 0.097
0.088 0.0002
Predictions for the other measured fission products
are not available; on the other hand, the
concentration of 78-hour 132je was predicted to be
2.2 /tCi/rnl, but none (<1 x 10"6 MCi/ml) was
detected.
That some fission products were found and others,
of similar fission yields, were not (see Table 2.1), may
be attributable to the volatility of the detected ones
or their radioactive precursors. Relatively little
removal by the main coolant demineralizer or
relatively high solubility in coolant water may also be
responsible for the presence of these radionuclides in
the water. Thus, radioiodine, the radiokrypton
precursors of 89sr) 90sr, and 91Sr, and the
radioxenon precursors of 13?cs and 140ga may have
passed through the fuel cladding at higher rates than
other radionuclides; or the radioisotopes of the rare
earths, ruthenium, etc., may have been removed from
the water very effectively by the demineralizer or by
crud formation in the coolant system. In the case of
95Zr, its daughter 95]\fb, and 99Mo, neutron
activation of Zircaloy and steel, respectively, may be
responsible for the presence of these radionuclides in
coolant water.
2.3.4 Activation products in main coolant water.
The turnover rates of the longer-lived activation
products, considering only their removal by
demineralizing and decay (not by crud formation or
coolant water discharge, for example), ranged from
0.001 to 1.6 nCi/sec according to calculations in
Appendix B.5. These are in the same range as fission
product turnover rates (see Appendix B.4).
The highest concentrations of measured activation
products were between 0.1 and 0.5 percent of
predicted concentrations for most radionuclides. The
maximum concentration relative to predicted values
was 25 percent for 24Na. These predictions by
Yankee were based on an overall monthly corrosion
rate of 10 milligrams per square decimeter and a
38-liter/min flow rate through the purification
system:iO)
Predicted Highest measured/
Radionuclide concentration, /uCi/ml predicted, percent
51Cr
55Fe
59Fe
58Co
64Cu
0.15
0.8
0.116
0.12
0.052
0.84
0.077
0.019
25.
0.11
0.46
0.31
0.37
0.12
0.17
7.4
The measured concentrations presumably include
suspended (insoluble) radioactive material. The
influence of the pH value in main coolant water on
these radionuclide concentrations is suggested by
comparing measured totals with Yankee data on
average radionuclide concentrations in crud
multiplied by crud concentrations of 0.4 ppm:(9,10)
October 4, 1968
Radionuclide
51Cr
54Mn
110mAg
Radionuclide
59Fe
in crud,MCi/ml
1.1 x 10-4
5.3 x 10*
1.3 x 10-5
2.7 x 10-5
1.2 x 10-5
7.8x10*
July 10,
incrud,/iCi/ml
3.2x10-4
2.8 x 10-4
1.8x10-4
1.6 x 10-3
5.5x10-4
1.2 x 10*
crud/ total
0.13
0.01
0.07
0.08
0.13
0.41
1969
crud/total
1.5
1.2
2.0
1.6
4.2
<1.2
The listed radionuclides are relatively soluble at the
low pH value of the Oct. 4 sample, but insoluble at
the higher pH in the July 10 sample. The ratio is only
qualitative, as indicated by ratios that exceed unity,
16
-------
because crud radionuclide concentrations are average
monthly values.
Analyses for the three activation products that
emit only low-energy beta particles -- *'*C, -^S, and
63]\{i - showed that all three are present at relatively
low concentrations (see Table 2.1). All may be
formed by thermal-neutron activation of the
elements. In addition, l^C is formed by the
(n, ft ) reaction (in water, for example) and the
(n,p) reaction (in ammonia and nitrogen gas, for
example).
2.3.5 Radionuclides in secondary coolant water.
Samples of steam generator blowdown water and
condenser water contained %H at concentrations
between 0.02 and 0.002 juCi/ml, and several other
fission and activation products at much lower con-
centrations (see Table 2.2). Except in the sample of
Oct. 4, 1968, only a few radionuclides other than ^H
could be detected. The blowdown and condensate
water samples of June 10, 1970 contained the same
concentrations of 3jj and 131i; other radionuclides
could not be measured with sufficient sensitivity for
comparison. Some of the radionuclides may have
been in insoluble form, as suggested by the obser-
vation that all of the 54\in in the June 10, 1970
sample was removed from the water by filtering.
The leakage rate of water from the main into the
secondary system was estimated by the equation:
main-to-secondaiy leakage rate _ 3H concentration, secondary
secondary turnover rate 3n concentration, main
Based on the tritium concentrations in Tables 2.1 and
2.2 and a makeup volume for secondary coolant
water of 40,000 liters/day, leakage rates were as
follows:
date of sample
Oct. 4,1968
July 10,1969
June 10, 1970
Leakage rates between 370 and 580 liters reported by
Yankee 00 for June, 1970, are consistent with the
value for June 10, 1970. The calculation assumes
equilibrium and is applicable only to constant or
slowly changing leakage rates. That several other
radionuclides have lower ratios than 3H can be
JH ratio,
secondary / main
5 xlO-4
1.6 x lu-2
1.0x10-2
Calculated leakage
rate, liters/day
20
640
400
Table 2.2
Radionuclide Concentration in Secondary System Water, ^Ci/ml
Continuous steam generator blowdown
Radionuclide
3H
14c
32P
51&
54Mn
55Fe
59Fe
58Co
60co
63Ni
89Sr
90Sr
95Zl
95Nb
110mAg
124Sb
1311
134Q,
137Cs
gross beta
Notes: 1.
2.
3.
Oct. 4, 1968
2.5 x 10-3
<2 xlO^
<2 xlO-7
3 xlO-6
1 xlO-6
5 x 10-8
1 xlO-6
2 x 10-6
5 x 10-7
2 x 10-7
-------
attributed to their removal by deposition either in the
main or the secondary system; higher ratios-found
only in the Oct. 4, 1968 sample-may indicate
residues from earlier leakage.
2.3.6 Radionuclides in other liquids. The %
concentration in one of the two waste holdup tanks
on Oct. 4, 1968 (see Table 2.3) was approximately an
order of magnitude lower than in main coolant water;
the other radionuclides were all relatively long-lived,
and mostly at higher concentrations than in the main
coolant at the same date. The specific sources of the
waste are not known. The low gross beta activity in
Table 2.3 indicates how misleading this measurement
can be in the presence of radionuclides that emit few
or no beta particles.
Safety injection water contained, at relatively low
concentrations, some of the long-lived radionuclides
detected in the main coolant, as shown in Table 2.4.
These radionuclides presumably entered the water
while it was in the shield tank cavity during refueling.
None of the radionuclides listed in Table 2.4 was
found in component cooling water. The minimum
detectable concentration was 1 x 10'6MCi/mlfor 3R
and between 1 x ICh7 and 1 x 10-8/*• Ci/ml for all
others.
Table 2.3
Radionuclide Concentration in Waste Holdup Tank on Oct. 4, 1968
Radionuclide
Concentration, ju Ci/ml
14c
32P
51Cr
55Fe
57Co
58Co
63Ni
89Sr
90Sr
134Cs
137Cs
182Ta
gross beta
3.8 x 10-1
1.2 x 10"4
< 2 x 10-7
<5 xlO-7
1.4 x 10-3
8.2 x 10-4
1.0 x 10-*
6 xlO-6
2.9 x 10-4
6.4 x 10-4
2.3 x 10-4
3 x 10-7
7 x 10-8
3.1 x 10-4
< 3 x 10-8
2.0 x 10-5
5.4 x 10-5
1.6 x 10-5
5.0 x 10-4
Notes: 1. Radionuclide concentrations are at time of sampling, gross beta activity is
on Oct. 9, 1968.
2. < values are 3 ff counting error.
Table 2.4
Radionuclide Concentration in Safety Injection Water,
June 10,1970
Radionuclide
Concentration, /uCi/ml
55Fe
59pe
57Co
124Sb
137Cs
2.2 x 10-2
9.5 x 10-5
5 xlO-6
1 xlO-6
6 x 10-7
1.5 x 10-5
4.5 x 10-5
8 xlO-6
2 xlO-6
2 x 10-7
Note: The following radionuclides were not detected ( < 1 x 10"7 MCi/ml):
14C, 32P> 51Cr, 89Sr, 90Sr> 131r> and 134Cs
18
-------
£.4 References
1. Yankee Nuclear Power Station-Yankee Atomic
Electric Co., "Technical Information and Final
Hazards Summary Report", AEC Docket No. 50-29
(1960).
2. Heider, Louis, Yankee Nuclear Power Station,
personal communication (1970).
3. Blomeke, J. O. and F. E. Harrington,
"Management of Radioactive Wastes at Nuclear
Power Stations", AEC Rept. ORNL-4070 (1968).
4. Lederer, C. M., J. M. Hollander, and I. Perlman,
Table of Isotopes, John Wiley, New York (1967).
5. McKinney, F. E., S. A. Reynolds, and P. S.
Baker, "Isotope User's Guide", AEC Rept.
ORNL-IIC-19(1969).
6. Martin, M.J. and P.H. Blichert-Toft, "Radio-
active Atoms", Nuclear Data T ablest,1 (1970).
7. Kahn, B. et al., "Radiological Surveillance
Studies at a Boiling Water Nuclear Power Reactor",
Public Health Service Rept. BRH/DER 70-1 (1970).
8. Rodger, W. A., "Safety Problems Associated
with the Disposal of Radioactive Waste", Nuclear
Safety 5, 287(1964).
9. "Yankee Nuclear Power Station Operation
Report No. 94 for the Month of October 1968",
Yankee Atomic Electric Co., Boston, Mass. (1968).
10. "Yankee Nuclear Power Station Operation
Report No. 103 for the Month of July 1969",
Yankee Atomic Electric Co., Boston, Mass. (1969).
11. "Yankee Nuclear Power Station Operation
Report No. 114 for the Month of June 1970",
Yankee Atomic Electric Co., Boston, Mass. (1970).
12. Dudley, N. D., "Review of Low-Mass Atom
Production in Fast Reactors", AEC Rept. ANL-7434
(1968).
13. Ray, J. W., "Tritium in Power Reactors",
Reactor Fuel-Processing Tech. 12, 19 (1968).
14. Weaver, C. L., E. D. Harward, and H. T.
Peterson, "Tritium in the Environment from Nuclear
Power Plants", Public Health Repts. 84, 363 (1969).
15. Mountain, J. E. and J. H. Leonard, "Tritium
Production and Release Mechanisms in Pressurized
Water Reactor Coolant", Trans. Am. Nucl. Soc.
13, 220 (1970), and Mountain, J. E., Master's Thesis,
University of Cincinnati (1969)
16. Locante, John, Westinghouse Electric Corp.,
personal communication (1970).
17. "Yankee Nuclear Power Station Operation
Report No. 98 for the Month of February 1969",
Yankee Atomic Electric Co., Boston, Mass. (1969).
19
-------
3. Radionuclides Released from Stack
3.1 Gaseous Waste System
and Samples
3.1.1 Gaseous waste system. Gaseous radioactive
wastes generated at Yankee are discharged to the air
as depicted in Figure 3.1, which is based on
descriptions by several authors.' ^) Yankee wastes
are classified as hydrogen-bearing and air-bearing.
Hydrogen-bearing waste originates in the main
coolant system; with one exception, it is collected in
the gas surge drum at a compression of several
atmospheres and held for radioactive decay. Three
decay tanks are available to store additional gas under
pressure. Transfer from main coolant to storage is
either direct, at the low-pressure surge tank, or
through venting the hydrogen-bearing liquid waste
from collection tanks and the waste-evaporator
condenser. The storage tanks are blanketed with
nitrogen to prevent mixing hydrogen with air.
Gas from the surge drum is released at a nominal
rate of 0.425 standard nP/min through a deep-bed
glass-fiber filter to the base of the 1.1-m dia., 46-m
high, cylindrical primary vent stack. In the stack, the
gas is diluted with ventilating air from the Primary
Auxiliary Building, which is discharged at the
nominal rate of 425 m^/min. Surge drum gas is
usually released once each year, (3) although releases
were reported in February, March and April 1969, vO
and a special release was made for the measurements
during the field trip on June 3,1969.
Air-bearing waste consists of gases from the air
ejector at the main condenser in the secondary
coolant system, the gland seal condenser in the
secondary system, tanks that contain secondary-
system liquid wastes, and the evaporator when air-
bearing liquid waste (from the gravity drain tank) is
being processed. These gases are released directly into
the primary vent stack for dilution by the ventilating
air from the Primary Auxiliary Building. Vapor
container air is also discharged to the stack whenever
the containment building is opened; this occurred 15
times in a 4-year period (see footnote to Appendix
B.2). Air from the Primary Auxiliary Building is
discharged continuously through the stack. Air from
the Turbine Building is discharged to outside air
without passing through the stack.
Also released directly to the stack are
two liters per day of hydrogen-bearing gas that pass
from main-coolant sampling ports into the laboratory
hoods when aliquots of main coolant water are
collected for analysis. This usually occurs once daily.
Yankee routinely reports values of 41 Ar, 133xe,and
135xe concentrations in the main coolant.(4)
Gases from burning solid waste in the incinerator
are discharged through a wet-gas scrubber and
deep-bed glass-fiber filter through a 20-cm dia., 2.4-m
high, stack on top of the Primary Auxiliary Building.
This effluent is reported to contain negligible
radioactivity A1)
The major components of the radioactivity
released from the surge drum would be expected to
be the fission-produced long-lived radioisotopes of
krypton and xenon (see Appendix B.3 of the Dresden
study) (5) and tritium. Any other gaseous or
relatively volatile fission and activation products in
this effluent would also be long-lived because of the
long retention period.
The radionuclide content of the continuously
discharged stack gases depends on the leakage rate
from the main-coolant system and the extent of
specific releases such as main coolant sampling and
vapor container venting. Radioactive gases from the
air ejectors and main coolant sampling would contain
relatively short-lived isotopes. Some of these gas
streams are unfiltered and may carry radioactive
particles. Off-gas emissions from the air ejector at the
condenser are monitored continuously by an
anthracene detector; all stack discharges are
monitored by 4 G-M tubes.
3.1.2 Radionuclide release. Radioactive gas
discharges by Yankee are limited by the AEC as
follows: "As determined at the point of discharge
from the primary vent stack and averaged over a
21
-------
to
VAPOR
CONTAINER
OTHER
LIQUI OS
Figure 3.1. Sources of Airborne Effluent.
-------
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, 10 CFR 20." (6) The values in Table II
derive from Section 20.105 of 10 CFR 20, which
limits the added radiation dose to individuals in
unrestricted areas to 0.5 rem/year. The factor of
1,000 is allowed in consideration of atmospheric
diffusion from the stack (1) to the boundary (at the
300-m perimeter) of the Yankee exclusion area.
Limits for discharging individual radionuclides to air
by Yankee are given in Section 3.3.8.
Yankee has reported the following annual releases:
(1,4,7)
Radioactivity 1970 1969 1962 to 1968
/? y in gas, Ci 17.2 4.13 0.7-22
3H Ci 9.0 9.19 8-16*
/8Y in particles. /xCi l-82 2-51 ?.89+
*1965-1968;+1968
The highest annual release of gross beta-gamma
activity represents 0.5 percent of the release limit of
4,500 Ci/yr for 87Kr and 88Kr, the most hazardous
noble gas fission products. For •%, the highest annual
release represents 0.04 percent of the 45,000 Ci/yr
limit for tritiated water vapor (HTO). The particulars
radioactivity is an extremely small component of the
total radioactive discharge.
3.1.3 Sample collection. Samples of gas surge
drum contents were obtained on October 4, 1968,
April 1, 1969, and June 3,1969. The first sample was
withdrawn in triplicate at the sampling port of the
surge drum into evacuated 9-cc glass serum bottles,
sealed with rubber stoppers held by crimped
aluminum holders. On the other two occasions,
duplicate samples were collected in evacuated
0.85-liter gas cylinders.
A 144-m^ volume of gas was discharged from the
surge drum through the primary vent stack on June 3,
1969, from 1500 to 2145 hours. During this period, a
sampling system was attached to a single-nozzle
probe, centered in the stack. The system components
were in the following sequence:
(1) membrane filter (Millipore Filter* type
AA, 5-cm dia., in Unico holder) for
sampling particles;
(2) carbon bed (26.6 g Columbia 6GC
activated charcoal, type 10/20, 3.2-cm
dia.) for sampling gaseous iodine;
(3) pressure-vacuum gauge;
(4) calibrated flowmeter (F and P
Flowrator);
(5) vacuum pump (Cast Model 0406).
This sampling procedure was repeated on June 4,
1969, from 0910 to 1530, to measure radionuclide
concentrations when no gas surge drum contents were
being discharged. The sample volumes that passed
through the filter and carbon bed were computed to
be 2.0 m^ on June 3 and 3.6 m-* on June 4. At the
beginning of each of the 2 sampling periods, an
evacuated 8.2-liter gas bottle was filled with gas at the
stack probe to measure the concentration of
radioactive gases.
Beginning June 5, 1970, at 0930 hours, five
consecutive 24-hr filter and carbon bed samples were
obtained in the primary vent stack to measure the
variability of particulate and radioiodine emissions.
The sampling system was the same as that described
above, except that the carbon bed was 5.0 cm in
diameter. Sampling flow rates varied between 12 and
20 liters/min; the typical sample volume was 27 np.
The following samples were collected on June 10,
1970:
(1) air ejector off-gas from the main
condenser in the secondary coolant
system before dilution in the stack, 8.2
liters.
(2) vapor container atmosphere , 8.2 liters.
Ambient temperature was 31°C; the
relative humidity was 43 percent of
saturation. (-0
(3) water from the dehumidifer in the vapor
container, 4 liters. Yankee operates the
dehumidifier to collect water samples for
^H analysis, and reports ^H
concentrations in discharged vapor
container air on the basis of these
analyses. (3)
(4) main-coolant gas, during depressurizing
for routine liquid sampling, 17 cc in two
9-cc serum bottles.
* Mention of commercial products does not constitute endorsement by the Environmental Protection
Agency.
23
-------
(5) 5-cm-dia. glass fiber filter used by Yankee
to sample particulate emissions in
incinerator stack, during operation from
2030 to 2130 hours, June 9, 1970.
Sampling flow rate was 10 liters/min.
Other samples were as follows:
(6) vapor container atmosphere, 8.2 liters, on
Nov. 19. 1970. Ambient temperature was
14°C; the relative humidity was 47
percent of saturation. (3) The vapor
container was open to outside air during
refueling.
(7) water from the dehumidifier in the vapor
container, 100 ml, on Nov. 19,1970.
(8) water from the dehumififier in the vapor
container, 1 liter, on Nov. 30, 1970.
Ambient temperature was 27°C; the
relative humidity was 53 percent of
saturation. (3) The vapor container had
been sealed for 10 days prior to sampling.
(9) main coolant gas, two 9-cc serum bottles,
on Feb. 9,1971.
(10) air ejector off-gas from the main
condenser in the secondary coolant
system before dilution in the stack, 8.2
liters on February 9, 1971.
3.3 Analysis
3.2.1 Gamma-ray spectrometry. Radionuclides
that emit gamma rays were routinely analyzed with a
10-cm x 10-cm cylindrical Nal(Tl) detector coupled
to a 400-channel spectrometer (see Figures 3.2 and
3.3). Identification was confirmed by spectral
analysis with a high-resolution 10.4-cm^ x 1.1-cm
Ge(Li) detector and a 1600-channel spectrometer or
with a low-background dual 10-cm x 10-cm Nal(Tl)
detector system in various coincidence/anticoinci-
dence modes. Iron-55 was measured with a
xenon-filled x-ray proportional counter and a
200-channel spectrometer.
The samples of main coolant gas and air ejector gas
obtained on February 9, 1971, were first analyzed
within 5 hours of collection to detect relatively
short-lived 41Ar, 87Kr, and 88 Kr.* All other samples
were counted one day after collection, hence only
radionuclides with longer half lives (>6 hours) could
be detected. As an exception, ^Ar and 8^mKr were
detected in the sample of main coolant gas obtai ed
on June 10, 1970, because of their relatively high
initial concentrations.
Primary coolant gas and waste surge drum samples
were analyzed in 9-cc glass serum bottles. Aliquots of
8.2-liter gas samples were counted in 209-cc
100R
BACKGROUND
(INSTRUMENT PLUS
GLASS CONTAINERS)
0.6 0.8
ENERGY, MeV
1.4
Figure 3.2.Gamma-ray Spectra of Gas Surge
Drum Samples.
detector: 10 * 10-cm Nal(Tl)
sample : 27 cc of gas in glass serum bot-
tles, collected 1200 EUT October
k, 1968, and 1125 EOT, June 3,
1969.
count : Oct. 4 sample - 1600 Oct. 10 to
0840 EDT, Oct. 11, 1968.
June 3 sample - 1700 June 6 to
0940 EUT, June 7, 1969.
*We thank Messrs. G. J. Karches and C. Nelson, NERHL, PHS, for making possible the prompt analysis
of these samples.
24
-------
10,000
1,000
100
s 10.0
1.0
0.1
I =
_ co E
I Si "if
i si
200 400 600 800 1,000
ENERGY, keV
,200
1,400 1,600
Figure 3.3, Gamma-ray Spectra of Gas Released in Sampling Main Coolant.
J)etector: Nal(Tl). 10 X 10 cm
Sample : 9 cc bottle of gas collected 1030 hours EUT, June 10, 1970
Counts : #1 - 1515 hours EDT, June 11. 1970 (10 min)
#2 - 1044 hours EffT. June 16, 1970 (50 min)
25
-------
volumetric flasks. The serum bottles and flasks were
sealed with rubber stoppers held by crimped
aluminum seals. Activated charcoal and 440-ml
aliquots of vapor container water were counted in
plastic containers.
Detection efficiencies for the radionuclides,
containers, sample volumes, and media of interest
were determined with standardized radioactivity
solutions or 85jCr gas. Because glass contains ^^K and
charcoal contains 40K and 226Ra, distinct
backgrounds were measured for these materials.
Counting intervals and techniques were selected to
provide, when possible, counting precision of + 10
percent or better at the 95 percent confidence level.
The usual counting duration for low-level activity was
1000 min. Samples were re-analyzed periodically to
confirm identification of radionuclides by
determining half lives, and to look for longer-lived
radionuclides.
3.2.2 Radiochemical analysis. Strontium was
chemically separated from one half of each
particulate filter and from aliquots of the
dehumidifier condensates. The radiostrontium
content was measured by counting for 100-min
intervals with low background G-M beta particle
detectors. Strontium-90 was distinguished from 89$r
by separating and counting the ^"Y daughter.
Krypton-85 at relatively low concentrations was
determined by liquid scintillation counting.
Approximately 3-cc aliquots of the gas surge drum
samples were mixed with degassed PPO and bis-MSB
liquid scintillator solution and measured for 50-min
periods in a liquid scintillation counter with
spectrometer. *(&) Aliquots of all samples obtained in
June of 1969 and thereafter were analyzed by
counting 85Ki with 1-mm-dia. plastic scintillator
spheres occupying 15 cc of the 25-cc vial volume.
Samples of 85Kr were either transferred directly to
the counting vial or concentrated from 0.5 - to 1 -liter
aliquots by passing gas through charcoal at -78°C and
then heating the charcoal to transfer 8$Ki to the
counting vial.
During liquid scintillation analysis of the first gas
surge drum sample for °^Ki, ^ unexpected gaseous
radionuclide at relatively high concentration was
detected. This gas was identified as ^^C in the form
of CO or an organic compound, but not CC>2,by
observing its disintegration mode, beta-particle
spectrum, and chemical behavior. Gas samples
collected later were analyzed in duplicate for 14, •; by
passing aliquots mixed with CO carrier gas through
an alumina-platinum (0.5 percent) catalyst heated to
550°C to convert the sample carbon to CO2, and into
a bubbler containing BaCOs. The 14C activity in the
precipitate was counted for 10-to 100-min intervals in
low-background beta counters. Identification was
confirmed by aluminum absorber curves. Aliquot
sizes ranged from 10 cc to 1 liter, depending on the
14c concentration.
Tritium in HT, HTO vapor, or other gaseous form,
was separated in the samples collected in June, 1969,
and thereafter by passing aliquots mixed with \\i
carrier gas through a copper oxide bed heated to
550°C to oxidize hydrogen, and collecting the water
in a trap at -78°C. The -^H activity in the condensate
was measured by liquid scintillation counting for
200-min periods. Aliquot sizes were the same as for
*4c analyses.
Tritiated water vapor concentrations in the gas
samples were determined by liquid scintillation
counting. To collect 3H in this form, distilled water
equivalent to 5 percent of the gas sample volume was
injected into containers previously used for
gamma-ray analysis and intermittently swirled for 2
to 3 days. The water was then removed and distilled,
and aliquots were mixed with liquid scintillator for
analysis.
3.3 Results and Discussion
3.3.1 Gases released by sampling main coolant.
The radionuclides found in main-coolant gas (see
Table 3,1) include all high yield fission-produced
krypton and xenon isotopes whose half lives were
longer than 1 hour. In addition, 3H and the activation
products 14C and 41 Ar were detected. The 41Ar was
probably formed from argon in air within the system;
production of 3H and ^C is discussed in Sections
2.3.2 and 2.3.4, respectively. Measurements by
Yankee staff of the June 10, 1970, sample were as
follows:
41Ar 1.18 MCi/cc
133xe 4.05 x 10-3
135xe 4.74 x 1Q-3
The values for 41Ar and 133Xe in Table 3.1 are in
agreement, while the concentration of ' 3^Xe is more
than two-fold higher.
*We thank Dr. A. A. Moghissi, Mr. R. Shaping and staff at the Southeastern Radiological Health
Laboratory (SERHL, EPA) for analyzing these samples.
26
-------
Table 3.1
Radioactive Gases Released to Stack during Depressurizing Main Coolant for Sampling
Concentration, juCi/cc
Radionuclide
12.3 -yr 3H
5730 -yr 1*C
1.83-hr 4lAr
4.4 -hr SSmjc,
10.7 -yr SSfcj
76 -m 87Kr
2.8 -hr SSjcr
2.3 -d 133mxe
5.29-d 133Xe
9.1 -hr!35Xe
June 10, 1970
1.9 + O.l'xlO-4**
2.6 ±0.1x10-3
1.0 + 0.2
5.4 + 0.6 x ID'3
9+4 x 10-5
NA++
NA
1.5 + 0.3 x 10-4
4.8 + 0.1 x ID'3
1.2 ±0.1 x 10-2
Feb. 9, 1971
1.4 + 0.1 x 10-3
3.7 + 0.3 x 10"3
3.8 + 0.1 x 10'1
6.7 + 0.1 x 10'2
1.8 + 0.2x10-3
7.2 + 0.7 x 10-2
8.8 + 0.3 x 10-2
5.4 + 0.3 x 10-3
4.2 + 0.1 x 1C-1
2.1 ±0.1x10-!
Release per Sample, yuCi*
June 10, 1970
3.6 x 1C-1
4.9
1.9 x 103
1.0 xlO1
1.7 x lO-1
—
3.0 x lO-1
9.2
2.2 x 101
Feb. 9, 1971
2.6
7.0
7.1 xlO2
1.3 xlO2
3.4
1.4 xlO2
1.7 xlO2
1.0 xlO1
7.9 x 102
4.1 xlO2
Estimated
Average Annual
Release, + Ci
5 x 10-4
2 x 10-3
4 x 10"1
2 x 10-2
6x10-4
2 x lO-2
3 x 10-2
2 x 10-3
1 x 10-1
7 x lO'2
* based on the release of 1900 cc of gas during sampling operation
+ Average of two release values (approx. 10 MCi each assumed for 87^ an) relative to a value of unity for
135Xe:
9.1-hr
3.2-min
15.6-min
3.8-min
17. -min
135Xe
89Rj
135mxe
137xe
138Xe
1.0
3.2
1.4
4.4
5.0
The .sum of the four short-lived radionuclides is thus
estimated to be 14 times the amount of * 3^Xe, or 14
x 0.07 = 1.0 Ci/yr. Short-lived 10-min 13N probably
is also in this effluent gas.
3.3.2 Gaseous effluent from secondary coolant.
Tritium, 14C, 133Xe, and 135Xe were observed in
off-gas from the air ejector at the secondary-coolant
main condenser (see Table 3.2). The concentrations
of these radionuclides were so low in the sample of
Feb. 9, 1971, that the shorter-lived xenon and
krypton isotopes, if present in the proportions
indicated in Table 3.1, would not have been detected
after the 5-hour interval between sampling and
counting. Only a small fraction of the tritium was in
the form of water vapor; the largest fraction probably
was associated with molecular hydrogen or an organic
compound.
The concentrations of these radionuclides on June
10, 1970, were three to four orders of magnitude
lower than in main-coolant gas; on Feb. 9,1971, they
were as much as six orders of magnitude lower (see
Table 3.1). The lower ratios in the later sample
suggest a much lower main-to-secondary leakage rate.
Differences among the ratios of concentrations in the
secondary system to those in main-coolant water may
be caused also by differences in gas turnover rates in
the secondary system, the solubility of the various
gases in water, and the occurrence of chemical
reactions.
Radionuclide release rates at the times of sampling
were computed from the concentrations and gas flow
27
-------
Table 3.2
Radioactivity Contents of Off-gas from Air Ejector at Main Condenser in Secondary Coolant System
Concentration before dilution
. in stack. AtCi/cc
Radionuclide
3R (total)
3H (water vapor)
14C
4Ur
85Kr
133mXe
133Xe
135Xe
June 10, 1970
2.210.1x10-7**
8 ± 6 x 10-9
5.9±0.8xlO-7
NA++
-------
Table 3.4
Radioactivity in Vapor Container
In ail, JUCi/cc
Radionuclide
3R (total)
June 10, 1970 Nov. 19, 1970
1.8+0.2x10-6
3fl {water vapor) 5 .3+0.8x1 0'7
14C
24Na
51Cr
54Mn
57Co
58Co
60Co
59Fe
85Kr
89St
90Sr
1 lOrriAg
124Sb
131i
133Xe
134Cs
137Cs
182Ta
Notes: 1.
2.
3.
l.l±0.1xlO-6
NA
NA
NA
NA
NA
NA
NA
<5 xlO-9
NA
NA
NA
NA
NA
<3 xlO-7
NA
NA
NA
8 +2 xlO-7
5.5+0.6x10-7
4 ±2 xlO-9
NA
NA
NA
NA
NA
NA
NA
1.5±0.1xlO-7
NA
NA
NA
NA
NA
<3 xlO-7
NA
NA
NA
+ values indicate analytical error expressed at 2
at 3 O counting error.
NA - Not analyzed.
Water vapor concentration,
g/m3; June 10 -
Nov. 19 -
Nov. 30 -
In dehumidifiei condensate. uCi/ml
June 10, 1970
9.8+0.2x10-!
NA
1.9+0.1x10-6
8 ±2 xlO-6
NA
4 ±1 xlO-8
NA
NA
2 +1 xlO-7
NA
NA
NA
NA
NA
NA
4 +1 xlO-7
NA
NA
NA
NA
a , < values are
14.1
5.6
14.0
Nov. 19,1970
2.1+0.01x10-3
NA
1.8+0.6 xlO-6
NA
1.3+0.1 xlO4
1.6x0.03x10-4
1.6x0.02x10-6
2.5+0.02x10-4
2.6+0.03x10-*
5.5+0.6 xlO-5
NA
<1 xlO-7
3.2±0.3 xlO-6
1.410.2 xlO-5
1.0+0.2 xlO-5
<1 xlO-5
NA
3 +1 xlO-7
6 +2 xlO-7
1.7+0.2 xlO-5
minimum detectable
Nov. 30, 1970
1.0+O.OlxlO-1
NA
2 ±1 xlO-7
NA
<8 xlO-7
2.3+0.6 xlO-7
<4 xlO'7
1.7+0.5 xlO-7
4.3+0.9 xlO'7
<3 xlO-7
NA
NA
NA
<1 xlO-7
<1 xlO-7
<8 xlO-7
NA
NA
NA
<2 xlO'7
concentrations
drum.(3) Gaseous 14c was observed at
concentrations above 1 x 10'4 MCi/cc in all three
samples, but was measured accurately only in the last
sample. The 3H in the sample of June 3, 1969, was
mostly ( > 99 percent) in the form of hydrogen gas
(HT) or a gaseous organic compound.
3.3.4 Radionuclide concentrations in the vapor
container. The only radionuclides found in vapor
container air were 3H (both as water vapor and gas),
l^C, and ^Kr, at the concentrations in columns 2
and 3 of Table 3.4. The minimum detectable level of
other radionuclides by gamma-ray spectrometry is
indicated by the "less-than" value of 133Xe.
Condensed water vapor, from a dehumidifier which
collects water samples for tritium analysis by Yankee,
contained 3H and relatively low concentrations of
many of the fission and activation products found in
main coolant water (see Table 3.4, columns 4 to 6).
The 3H concentrations in air, computed from
concentrations in the condensed water vapor and the
moisture content in air (see note 3 in Table 3.4), do
not agree with directly measured values:
Date
3H in air (condensed water vapor)
1.4x10-5 jnlx 9.8x10-1 MCJ_= 1.4*
cc ml
5.6X10"6 x 2.1x10-3
ml cc
= 1.2x10-8
3H in ail (direct)
June 10
Nov. 19
June 10
Nov. 19 8 xlO'7
On June 10, during reactor operation, the condensed
water vapor indicated an 8-fold higher 3H
concentration than was found in air; on Nov. 19,
1970, while the building was open during refueling, it
indicated a 70-fold lower concentration. The two
types of samples were obtained at the same location,
but the air was collected for a much shorter interval
than the condensed water vapor. The presence of
29
-------
and °^Kr in air and the differences in 3H values
suggest that gas samples should be analyzed by
Yankee to determine radionuclide releases while
ventilating the vapor container. The detection of the
other, nonvolatile, radionuclides in condensed water
vapor indicates their presence, but air filter samples
would be required to quantify their concentrations.
The two sets of measurements in air were used to
estimate annual releases: the amount discharged
immediately after a shutdown was taken to be the
product of the concentration on June 10, the air
volume in the vapor container (24,000 m3) (3), and
the number of shutdowns per year (say 4); the
amount discharged during refueling was taken to be
the product of the concentration on Nov. 19, the
ventilation rate (425 m-Vmin), the refueling period
(say, 30 days per year). Thus, the annual release
would be:
accumulated radionuclides
discharged immediately
after reactot shutdown
radionuclides discharged
continuously during
refueling
yearly total
3H, Ci 14C, Ci 85jcr, Ci
0.17 0.11 <,0.0005
13. 0.066 2.8
13. 0.18 2.8
These calculations suggest that most of the
effluent gaseous radioactivity at Yankee is released
from the vapor container during refueling (see totals
in Section 3.3.8), and that monitoring this effluent
provides a significant portion of the annual release
data. The above values, based on one sample each,
serve only to indicate the magnitude of radionuclide
releases from the vapor container.
3.3.5 Particulate radioactivity and radioiodine in
the primary vent stack. The activation products ^Mn
and "0(To and the fission product 9®Sr were the only
particulate radionuclides detected on the stack
sampler (see Table 3.5 and 3.6). All three
radionuclides were at extremely low concentrations.
Except possibly 60co, these radionuclides appear to
be associated with continuous release, rather than
surge-drum gas (see Table 3.6). No particulate or
gaseous 131j was detected in any sample. The
24-hour samples of June 1970 (see Table 3.5)
provided a more sensitive test of 131i concentrations
during continuous discharge than those of June 1969
because gamma-ray spectrometry was initiated sooner
(within 31 hours) after sampling.
The average release rates according to the seven
values in Tables 3.5 and 3.6 (and less-than values for
131I based on Table 3.5 only) were:
Radionuclide
54Mr.
60co
90Sr
Average stack release
5 pCi/sec
8
8
<9
To compute the annual discharge of these
radionuclides, multiply the release rates by 2.8 x 10 ^
sec/yr.
The amounts of °"Sr and ^3'Cs that are formed
in environmental air by radioactive decay of their
Table 3.5
Stack Releases of Paiticulate Radionuclides and Gaseous Iodine-131, pCi/sec
Date, 1970
Radionuclide
Particles on membrane filter
313 -d 54Mn
2.7 -yr 55pe
71.3 -d 58co
5.26-yr 60rj0
50.5 -d 89sr
28.5 -yr 90sr
8.06-d 131j
June 5
7±6
< i
<2
10 + 4
< 1
3 + 1
<11
June 6
^ i
< 1
<2
3 ±2
< i
0.5 + 0.2
<4
June 7
5
< 1
< 2
8
< 1
0.?
< 9
+ 2
±3
i + 0.2
JuneS
< 1
< 1
<2
2 ±1
< 1
<0.2
<2
June 9
<• i
<; 1
<2
2 ±1
<; 1
0.8 + 0.1
<3
Gaseous iodine on charcoal
8.06-d
<3
<3
< 3
<3
<3
Notes: 1. Nominal stack flow rate is 7.1 m3/sec.
2. < values are minimum detectable concentrations at 3 a counting error; ± values are 2(7 counting error.
3. 106pCi/sec=l j/Ci/sec.
30
-------
Table 3.6
Stack Effluent Release Rates During and After Gas Surge Drum Release, MCi/sec
Calculated from
surge drum contents,*
Measured during release,
Measured after release.
Radionuclide
Gas
12.3 -yr 3ft
5730 -yr 14C
10.7 -yr 85 Kr
5.3 -d 133Xe
Gaseous iodine on charcoal
8.06-d 131l
Particles on filter
5.26-yr 60Co
313 -d 54Mn
50.5 -d 89sr
28.5 -yr90Sr
June 3, 1969
0.62
5.6
0.52
0.37
...
...
June 3, 1969
0.46 + 0.02
3.2 + 0.4
0.42 ± 0.09
<0.7
< 50x10-6
20 ± 5 x ID'6
18 ±4 x ID"6
<1 xlO'5
25 ± 2 x 10"6
June 4, 1969
<0.15
0.01010.002
<0.03
<0.7
< 30x10-6
14±3x 10-6
3± 1x10-6
< 1 x ID'5
28 + 3 x 10-6
*Calculated for concentrations in Table 3.3 and release rate of 0.425 m3 (STP) per minute from surge drum.
Notes:
1. Nominal stack flow rate is 7.1 m3/sec (15,000 cfm).
2. < values are minimum detectable concentrations at 3 ff counting error; + values are 2 cr counting error of single
analysis or the difference of results of duplicate analyses, whichever is greater.
respective gaseous precursors, ^Kr and
would be even smaller. The maximum amounts per
curie of precursor are 42 ^Ci 89sr and 0.25 /j. Ci
137cs. Approximately 1 Ci each of the two gases
may be released annually at the air ejector and
through sampling the primary coolant system (see
Section 3.3.1 and 3.3.2). Hence the amount of 89sr
reaching the environment through this mode may be
42 MCi/yr, or 1.4 pCi/sec; and the amount of 137cs,
0.25/xCi/yr or 0.008 pCi/sec.
3.3.6 Gaseous radioactivity in the primary vent
stack. As shown in Table 3.6, ^C was measured both
during and after the release from the gas surge drum.
Its emission rate was considerably higher during the
release than afterwards. Tritium and **%r were
detected only during the release.
The measured surge drum release rates shown --
i.e., differences between values in the third and
fourth columns of Table 3.6 - were approximately
two-thirds of the rates computed from the surge
drum gas analyses (Table 3.3, last column). The lower
value may have been caused by imperfect sampling or
a slower gas release than the rated 0.425 m-Vmin.
This rate pertains to conditions at initial discharge,
when the stack sample was obtained; as the internal
surge drum pressure decreases due to discharge, the
flow rate and hence the emission rate to the stack
may decrease. The total released radioactivity, based
on the discharged volume of 144 m^ and the
concentrations listed in the last column of Table 3.3
was as follows:
Radionuclide
14c
133Xe
Total release, Ci
1.3x10-2
1.1x10-1
1.1 x 10-2
< 3 x 10^
7.5 x 10-3
This indicates the magnitude of the annual release if
the waste was typical and is discharged once per year.
Compared to the (3-y gaseous release for June
1969 reported by Yankee (see Appendix B.2) of
0.445 Ci, the radionuclides from the surge drum
accounted for approximately one-third of the
activity. The tritium release from the surge drum was
10 percent of the reported monthly release of 0.13
Ci.
The l^C release rate in the stack on June 4,1969,
was 12-fold higher than measured at the main
condenser air ejector (Table 3.2) on two other
occasions. The difference was probably due to a
higher radionuclide release at the air ejector on June
4, 1969, because the leakage from main to secondary
coolant was higher.(3)
3.3.7 Particulate effluent from incinerator. The
only airborne particulate radionuclides observed in
the effluent from the incinerator during combustion
of solid wastes were 58(]o, 60co and 90sr, as shown
31
-------
Table 3.7
Participate Radioactivity Emitted from Incinerator Stack, June 9, 1970*
tedionuclide
55Fe
58Co
60Co
89Sr
90Sr
1311
Concentration,
MCi/cc
< 10 x 10-12**
4 ± 2 x 10-!2
10 ±3 x 10-12
<2 x 10-12
2.5 ±0.6x1 0-1 2
<4 xlO'12
Emission rate,
MCi/sec
—
6 x lO'1 0
2 x ID"9
—
4x10-1°
...
Estimated
annual release, + Ci
5 x 10-10
2 x 10-9
3 x 10-10
...
* operation from 2030 to 2130 hrs.; sampling rate and stack exhaust flow assumed to be 167 cc/sec.
+ computed for 241 hours of operation in 1970.(3)
** <£ values indicate minimum detectable concentrations at 3 a counting error; + values are 2 ff counting error.
in Table 3.7. For computing emission rates, the radionuclide releases. The totals of the release values
sampling flow rate was assumed to be identical to the
stack exhaust rate. Radionuclide concentrations and
the estimated annual release were very low on the
basis of these data.
3.3.8 Release limits and estimated annual
in Sections 3.3.1, 3.3.2, 3.3.4, 3.3.5, and 3.3.6
compare as follows with the limits established by the
AEC at the Yankee stack (1,000 times the limits
given in 1 0 CFR 20, (9) Appendix B, Table Il.column
1 for unrestricted areas):
Radionuclides
Gases
12.3 -yr SH (as HTO)
(as HT)
5730 -yr itC(s)
(as CO )
1.83-hr 4lAr
4.4 -hr ssmKr
10.7 -yr 8SKj
76 -min 87Kr
2.8 -hr 8»Kr
2.3 -d i33mXe
5.29-d i33Xe
9.1 -hr i35Xe
Other fission gases,
half lives < 2 hr
Particles and 131I
313 -d 54Mn(s&i)
5.26-yr 60Co (i)
50.5 -d 89Sr(s)
28.5 -yr 90Sr(s)
8.06-d i3H(s)
30 -yr i37Cs(i)
Yankee
limit,
/xCi/cc
2x10-4
4 x 10-2
1 xlO-4
1x10-3
4x10-5
1 X 10-4
3 x 10-4
2 x 10-5
2 x 10-5
3x10-4
3 x 10-4
1 X lO-4
3x10-5
1x10-*
3 x 10-7
3xlO"7
3 x ID'8
IxlO-7
5 x lO'7
Annual
release
limit,* Ci
4.5 x 104
8.9 x 10*
2.2 x 104
2.2 x 105
8.9 x 103
2.2x104
6.7 x 104
4:5 x 103
4.5 x 103
6.7 x 104
6.7 x 104
2.2 x 104
6.7 x 103
2.2 x 102
6.7 x 10J
6.7xlOi
6.7
2.2 x 101
1.1x102
Estimated
annual
release, Ci
1.3x101
3 x 10-1
xlO-i
XlO-2
X10-2
X10-2
XlO-3
xlO-i
xlO-i
2x10-4
(4 x 10-s)+
2x10-4
< 3x10-4
(2 x lO'7)"1"
Percent
of
limit
0.029
0.001
0.004
< 0.001
0.004
< 0.001
< 0.001
< 0.001
< 0.001
0.001
(0.04)
,<0.001
<0.001
K0.001)
0.003
< 0.001
K0.001)
*Based on a continuous stack discharge rate of 425 m-'/min
"These values were estimated from 135xe measurements (see Sections 3.3.1, 3.3.2, and 3.3.5).
Notes:
1. The individual limits apply in the absence of other radionuclides; if several radionuclides are
present, the sum of individual percentages of the limit may not exceed 100.
2. s = soluble, i = insoluble.
32
-------
The estimated annual releases of ^H and the sum
of all other radionuclides shown above are within
better than a factor of two of the 1969-1970 values
in Section 3.1.3 reported by Yankee. The annual
values by Yankee are based on many more
measurements than the ones in this study; on the
other hand, the station reports isotopic analyses only
for %.
The whole-body radiation dose to persons who
remained at the exclusion boundary throughout the
year would have been 0.08 percent of 500 mrem/yr-
i.e., 0.4 mrem/yr - according to the above estimates
from measured radionuclide releases. At highest
fraction of the limit were ^H (assuming the worst
case-that all tritium was in the form of water vapor)
and the very short-lived noble gas fission products
among gases, and 90§r among particles.
The actual population exposure would probably
be lower than the estimated value at the boundary
because the nearest town, Monroe Bridge, is
approximately 1 km distant. A better value of the
annual dose rate could be obtained by performing
isotopic analyses of the various airborne effluents at
the station and measuring with a tracer the degree of
dispersion from the stack to ground-level air.
3.4 References
1. Blomeke , J. O. and F. E. Harrington,
"Management of Radioactive Wastes at Nuclear
Power Stations", AEC Kept. ORNL-4070, 89-97
(1968).
2. Shoupp, W. E., R. J. Coe and W. C. Woodman,
"The Yankee Atomic Electric Plant", in Proceedings
of the Second United Nations International
Conference on Peaceful Uses of Atomic Energy, Vol.
8, United Nations, Geneva, 492-507 (1958)
3. Pike, D. and J. A. MacDonald, Yankee Atomic
Electric Co., personal communication (1969, 1970).
4. Yankee Nuclear Power Station Monthly
Operation Reports, Yankee Atomic Electric Co.,
Boston, Mass.
5. Kahn, B. et al, "Radiological Surveillance
Studies at a Boiling Water Nuclear Power Reactor",
Public Health Service Rept. BRH/DER 70-1 (1970).
6. Yankee Atomic Electric Company Docket No.
50-29, Interim Facility License, Appendix A,
"Technical Specifications" (March 4,1964).
7. "Management of Radioactive Wastes at Nuclear
Power Plants", Safety Series No. 28, International
Atomic Energy Agency, Vienna (1968).
8. Shuping, R. E., C. R. Phillips, and A. A.
Moghissi, "Low-Level Counting of Environmental
85Kr by Liquid Scintillation", Anal. Chem.41, 2082
(1969)
9. U. S. Atomic Energy Commission, "Standards
for Protection Against Radiation", Title 10, Code of
Federal Regulations, Part 20, U. S. Gov't. Printing
Office, Washington, D. C. (1965).
33
-------
4. Radionuclides in Liquid Effluent
4.1 Liquid Waste System
and Samp If ft
4.1.1 Liquid waste system. (1-3) Two classes of
liquid waste are discharged by Yankee: reactor plant
liquid waste, which may contain hydrogen gas added
to the main coolant to minimize the decomposition
of water in the reactor, and secondary plant water,
which does not contain added hydrogen gas. The
usual sources and directions of flow of these wastes
are indicated in Figures 4.1 and 2.2. Interconnections
in the storage and treatment system provide other
options.
Reactor plant liquid waste consists mostly of
water that had been used in the main coolant system
or in refueling the reactor (see Section 2.1.4). It is
stored in two Waste Holdup Tanks and a Gravity
Drain Tank before treatment by batch evaporation.
The condensed water from the evaporator is collected
in the Test Tanks, analyzed by Yankee for gross
beta-gamma activity and tritium concentration, and
discharged into effluent circulating coolant water.
The dilution factor during this discharge is 530,000
liters/min-r 113 liters/min = 4,700.
Secondary-plant liquid wastes are mostly system
leakage and once-daily steam-generator blowdown
water; some steam-generator blowdown water is also
discharged continuously (see Sections 2.1.3 and
2.1.4). The water isj>assed through two 5,300-liter
(1,400 gal) monitored waste tanks and discharged at
rates up to 113 liters/min into effluent circulating
coolant water.
4.1.2 Radionuclide release. The following liquid
waste was discharged at Yankee during 1970 and
1969: (4,5)
Class
1970.
reactor plant
secondary plant
Volume,
liters
3.1 x 106
15.8 x 106
Gross beta-
gamma, Ci
0.69 x 10-3
33.15 x 10-3
Tritium, Ci
1,212
280
1969
reactor plant
secondary plant
total
2.6 x 106 0.89 x 10-3 1,048
13.8 x 106 18.34x10-3 173.
16.4 x 106 19.23 x 10-3" 1,221
The total discharge of gross activity and ^H was
typical of operations at Yankee. W The volume of
liquid waste and the radioactivity varied considerably
from month to month, as shown in Appendix B.2.
Average concentrations of gross activity were
higher in secondary-plant waste, probably because it
is usually untreated while reactor-plant waste is
usually evaporated before discharge:
Pass
Gross beta-gamma,
/jtCi/ml
Tritium,
1970
1969
1970
reactor plant 2.2x10-7 3.4x10-7 3.9xlO-1 4-OxlO'1
secondary plant 2.1xlfr* 1.3x10-6 1.8x10-2 1.3x10-2
Tritium concentrations were considerably lower in
secondary-plant waste.
Average radioactivity concentrations in effluent
circulating coolant water during waste discharge in
1969 and 1970, based on the annual release data
given above and the coolant water flow rate of
530,000 liters/min, were:
source
assumed
release rate, gross
liters/min
reactor plant
secondary plant
113
28
6xlO'n
9x10-11
3H,/*Ci/ml
8x10-5
8x10-7
total
18.9 x 10^ 33.84 x 1Q-3 1,492
The assumed rates imply discharge of reactor plant
(Test Tank) waste during 4.8 percent of the year, and
continuous release of secondary plant waste.
Concentrations of radionuclides in effluents to
unrestricted areas are limited by the AEC according
to paragraph 20.106 of 10 CFR 20. (7)
Concentrations above background in water, averaged
35
-------
REACTOR PLANT LIQUID WASTES
(CONTAINING HYDROGEN)
SYSTEM LEAKAGE
SAMPLING DRAINS
NEUTRON POISON SOLUTION
EQUIPMENT DRAINS
COOLANT EXPANSION
ACTIVITY DILUTION 8 FLUSHING
CHEMICAL WASTES
(CONTAINING AIR)
EQUIPMENT DECONTAMINATION
SPECIAL WASTES
(CONTAINING AIR)
LABORATORY SINKS
INCINERATOR FLUE GAS
SCRUB WATER
BUILDING DRAINS
SECONDARY SYSTEM WASTES
SECONDARY SYSTEM SLOWDOWN
SECONDARY SYSTEM LEAKAGE
PRIMARY
DRAIN
COLLECTING
TANK
(Z8.8OO I.)
GRAVITY
DRAIN
TANK
(18,OOP I.)
WASTE GAS HEADER
T0_ WASTE
GAS SYSTEM
TO SHERMAN
CIRCULATING COOLANT WATET
RESERVOIR
FROM MAIN CONDENSER
Figure 4.i. Liquid Waste Sources and Treatment.
-------
over no more than 1 year, as listed in Appendix B,
Table II, column 2 of 10 CFR 20, are applied at the
boundary of the restricted area. The limit is 1 x 10"^
/jCi/ml for an unidentified mixture containing no
129I; 226Ra> an 3228^. Limits for ^^^31
radionuclides are 3 x 10~3 yuCi/ml for 3H, the
radionuclide at highest concentration in Yankee
effluent, and 3 x 10 '7 /uCi/ml each for soluble 90Sr
and 1^1, which are usually the radionuclides with
the lowest limits in reactor effluent. Higher limits are
permissible under conditions of Subsection (b) of
paragraph 20.106, or more stringent limits may be
applied under Subsection (e).
Massachusetts has given temporary approval for
daily releases of ^H by Yankee at amounts not to
exceed 10 Ci on the average, or 75 Ci at any time. 00
This is considerably lower than the limit of 2300
Ci/day computed at the normal flow rate of
circulating coolant water according to Appendix B in
10 CFR 20.
4.1.3 Samples. Two 4-liter samples of a
27,400-liter (7,236-gal) reactor-plant waste solution
in one of the two Test Tanks were obtained from the
Yankee staff on June 3, 1969. This waste solution
was condensate from the evaporator. One of the
samples was acidified with 100 ml concentrated HCI
to reduce possible sorption of radionuclides on the
sides of the plastic container. The waste solution was
released as usual into the effluent circulating cooling
water at the flow rate of 113 liters/min (30 gal/min)
between 1130 and 1530 on June 3,1969.
To measure directly the radionuclide content of
the effluent circulating coolant water, 200 liters were
collected at the outlet weir in a steel drum at
1150-1200. The water was passed through an
ion-exchange resin column at a flow rate of 100
ml/min to concentrate the ionic radionuclides on the
column. A 4-liter aliquot of water from the drum was
retained for measuring water hardness and
radioactivity. For comparison, a 200-liter sample of
service water, which is obtained at the same location
in Sherman Reservoir as circulating coolant water,
was collected in a steel drum from a tap in the pump
house at 1015-1020 on June 3, 1969. A 180-liter
volume of this water flowed through an ion-exchange
resin column in a 29-hour period, and a 4-liter aliquot
was retained for further analysis.
A second set of reactor-plant waste solution
samples~4 liters acidified (10 percent HNC^) and 1
liter unacidified-was obtained on Nov. 19,1970. The
reactor had been shut down for refueling on Oct. 24,
hence most or all of the waste was from the refueling
operation.
Four samples of water from the secondary
plant-samples No. (5) to (8) in Section 2.1.6--were
obtained from Yankee staff for analysis. Samples (5)
and (6) were taken to represent blowdown discharges,
and (7) and (8), secondary-plant leakage water.
Samples of flowing water and of a mixture of sand
and gravel were collected on two occasions from the
two yard drains that carry run-off water from the
plant area:
(1) 4 liters water and 0.8 kg sand and gravel from
east yard drain on June 3,1969 at 1700;
(2) 0.8 kg sand and gravel from east yard drain on
June 10,1970 at 1000;
(3) 4 liters water and 0.8 kg sand and gravel from
west yard drain on June 10,1970 at 1000.
The west drain is located near the parking area and
discharges into No. 5 Reservoir; the east yard drain is
to the east of the pump house and discharges into
Sherman Reservoir (see Section 5.1.2). Flow rates at
the time of sampling were estimated to be 3
liters/min in the east drain and ten times as much at
the west drain.
4.3 A*atg*i*
4.2.1 Test Tank solution. The unacidified (at pH
6.1) and acidified solutions of the waste were
analyzed spectrometrically with Ge(li) and Nal(Tl)
gamma-ray detectors. The samples were first counted
within a week after collection and again several
months afterwards to identify radionuclides by
combining observations of gamma-ray energies and
decay rates. The identified radionuclides were
quantified by computing disintegration rates from
count rates under characteristic photon peaks on the
basis of prior counting efficiency calibrations of these
detectors. The unacidified sample was analyzed
radiochemically for 3H, 14C, 129I, and 131I, and the
acidified sample, for 55Fe, 63Ni, 89Sr, and 90Sr.
Thirty-ml aliquots of the samples were evaporated,
measured with a low-background G-M counter to
determine gross beta activity, and analyzed by
counting with aluminum absorbers of increasing
37
-------
thickness to indicate the beta energy of the major
component and the effective counting efficiency (see
Figure 4.2).
100
50
20
10
5
2
1
0.5
0.2
0.1
TOTAL
COUNT RATE
OTHER
ADIONUCLIDES ~4
14,
i I
02 4 6 8 10 12 14 16 18 20 22 2426 28
SURFACE DENSITY, mg/cm2
Figured.2. Aluminum Absorber Curve of
Yankee Test Tank Sample.
Detector: Low-background G-M end-window.
Sample : 30-ml aliquot of sample col-
lected June 3, 1969 evaporated
on stainless steel planchet.
Counts : April 20, 1970, 100 min. at
each point.
4.2.2 Grculating coolant water. Each of the two
ion-exchange resin columns was separated into 6
parts: 3 cation-exchange resin sections, 2
anion-exchange resin sections, and a glass wool filter.
(9) Each part was analyzed with a Nal(Tl) gamma-ray
spectrometer for 1,000-minute counting periods.
Every cation-exchange resin section was eluted with
1,200 ml 6 N HC1. The elutriants were analyzed
radiochemically in sequence for strontium, cesium,
and cobalt.
The two water samples were analyzed for
hardness, gross beta activity, photon-emitting
radionuclides, and a few individual radionuclides.
Ten-mi aliquots were used to determine hardness and
tritium. The tritium sample was distilled, and 4 ml
were counted with a liquid scintillation detector. The
remaining 4 liters of water were acidified with 10ml
cone. HNO3 and evaporated to 45 ml, of which 15 ml
were further evaporated to dryness for gross beta
measurement and gamma-ray spectrometry with
Nal(Tl) detectors, and 30 ml were analyzed
sequentially for radiostrontium and radiocesium.
These radionuclides and the gross-beta-activity
samples were counted for 100 or 1000-min periods
with G-M detectors at a background of approximately
1.5 counts/minute.
4.2.3 Yard-drain samples. The water samples were
analyzed in the same manner as circulating coolant
water by gamma-ray spectrometry and for tritium.
radiostrontium, and radiocesium.
The sand and gravel were dried at I25°C, mixed.
and analyzed in weighed 100-cc and 400-cc aliquots
by gamma-ray spectrometry with Ge(Li) (see Figure
4.3) and Nal(Tl) detectors. The material was then
separated with a U. S. No. 10 sieve and the larger and
smaller particles were analyzed separately with a
Nal(Tl) detector. Ten-gram samples of the larger and
smaller particles were analyzed for ^%r and 90<}r
content by leaching with two 25-ml portions of hot
6N HNO3 chemically separating strontium, and
counting first total radiostrontium and, after 2 weeks,
radiostrontium plus 90y.
4.3 Result* amd Di*em**iom
4.3.1 Radionuclides discharged to circulating
coolant water. The -*H concentration in the Tes
Tank waste solution of June 3, 1969, was the same as
in main coolant water on July 10, 1969 (see Tables
4.1 and 2.1), 14C was approximately 3-fold lower in
the waste, and all other radionuclides were lower ir
the waste by two to four orders of magnitude
Presumably, distillation in the evaporator reduced the
concentrations of all other radionuclides. Tritium was
the main radioactive component both during reactor
operation on June 3, 1969, and during refueling on
Nov. 19, 1970. Of the other radionuclides, ^C and
"pe were at highest concentrations.
In this type of sample, analysis for specific
radionuclides is particularly desirable because many
of the radionuclides are not effectively counted with
the usual beta-particle detectors, hence the gross
activity value may be considerably lower than the
actual radionuclide content. For example, the tritium
concentration measured in the June 3 sample by
Yankee staff (3) was 0.375 fiCi/ml, in agreement with
the value in Table 4.1, but the gross beta-gamma
concentration was only 5.63 x 10-7 ^Ci/ml. For
38
-------
0.001 -
0
800
100
900
200
1,000
400
1,200
500
1,300
600
1,400
700
1,500
VO
300
1,100
CHANNEL NO. (1.008 keV/channel)
Figure 4.3. Gowno-ray Spectrum of Sand and Gravel from East Yard Drain.
Detector: Ge(Li), 10.4 cm2 x 11 am, trapezoidal
Sample : 633 g dried wt (400 cc), collected June 3, 1969.
Count : July 1-2, 1969 (1,000 min.); Th refers to 2327Yi and progeny, Bfegrf refers to
counter background (see background in Figures 5.4 and 5.5).
800
1,600
-------
comparison, gross-beta concentrations in 5 batches of
waste in the Test Tanks during the first seven months
of 1968 ranged from 4 x 10-8/iCi/ml to 1 x 10-6
The radionuclide concentrations in secondary
plant wastes (see columns 2 to 5 of Table 2.2) were
generally two to three orders of magnitude lower
than in the main coolant. As indicated in Section
2.3.5, the concentration of ^H in secondary coolant
water depends on the rates at which water leaks into
and from the secondary system; hence, it may be one
to four orders of magnitude below the main-coolant
concentration. Several other radionuclides had the
same secondary/main concentration ratios as ^H.
Compared to the 1970 annual average concentration
in Section 4.1.2, 3H values were the same on June 10,
1970, but the sum of other radionuclide
concentrations was lower.
4.3.2 Radionuclides in circulating coolant -water.
Tritium was the only radionuclide measured in
effluent circulating coolant water, while Test-Tank
contents were discharged, that was attributable to
release of this waste, as indicated in Table 4.2. The
measured concentration of 7.9 x 10-5 /tCi/ml was in
excellent agreement with the value of 8.5 x 10-5
juCi/ml computed from the concentration in the Test
Tank on June 3, 1970 (Table 4.1) and the dilution
factor of 4,700. No tritium could be detected in
intake water.
Even after concentrating ionic radionuclides from
200 liters of water on the ion-exchange resin column,
only 90sr and 137cs could be detected. The two
radionuclides were at the same concentrations in
influent and effluent water, suggesting that these
radionuclides originated in fallout from atmospheric
nuclear weapon tests. As indicated by the calculated
discharge values - measured concentrations in the
Test Tank divided by the dilution factor for
circulating coolant water - in Table 4.2, the
concentrations added by Yankee to the effluent were
within analytical uncertainty and thus not noticeable.
All other radionuclides discharged by Yankee were
below minimum detectable concentrations (<1 x
10-10MCi/ml).
The 90sr concentrations measured directly in the
water samples (see upper half of Table 4.2) were also
the same in effluent and influent, but were higher
than values obtained with the ion-exchange resins.
The difference between the total and the ionic 90sr
concentrations may be due to 90gr in suspended
Table 4.1
Radionuclide Concentration in Test Tank before Discharge
at Yankee,*
Radionuclide
June 3, 1969
3n 4.0 x 10-1
14C
32P
54Mn
55Fe
59Fe
58Co
60Co
90Sr
110mAg
124Sb
131l
137Cs
gross beta (unacidified)
5
5
2
1
< 1
2
7
7
2
< 1
5
< 1
5
xlO-6
xlO-8
xlO-7
xlO-7
xlO-7
xlO-7
xlO-8
xlO-9
xlO-7
xlO-7
xlO-8
xlO-7
xlO-6
Nov. 19, 1970
7.3
1.4
< 1
1.1
8
4
9
6
1.0
< 1
5
6
9
1.4
xlO-3
xlO-6
xlO-8
xlO-6
xlO-6
xlO-7
xlO-8
xlO-8
xlO-8
xlO-8
xlO-8
xlO-8
xlO-8
xlO-6
* Radionuclide concentrations are at time of sampling; gross beta activity .was.obtained 5 days later.
Note:
51cr, 63Njf 89sr, 95%, 95Nt>; 1&1, and 134Cs were not detected. Minimum detectable levels were
5 x 10'7jttCi/ml for 51Cr, 1 x lO^/xCi/ml for 89Sr, and 1 x 10'7/iCi/ml for all others.
40
-------
Table 4.2
Radionuclide Concentration in Main-Condenser Circulating Coolant Water on June 3, 1969
Radionuclide
Water analysis
3H
14C
32P
55Fe
90Sr
110mAg
137Cs
gross beta
Intake,
MCi/ml
< 2 x lO'6
NM
NM
NM
1.5+0.6x10-9
NM
<3x 10-10
2.4 + 0.5x10-9
Effluent,
/tCi/ml
7.9 x ID"5
NM
NM
NM
1.4 ±0.6x10-9
NM
<3 xlO40
2.4 ± 0.5 x 10-9
Calculated discharge,
MCi/ml
8.5 x 10-5
1.1 x 10-9
l.lxlO-H
2 x 10-11
1.5 x 10-12
4 x 10-11
<2 x 10-11
Ion-exchange resin analysis
54Mn
58Co
60co
89Sr
90Sr
131r
134Cs
!37Cs
Notes: 1.
2.
3.
-------
Among these radionuclides, only 24f4a could have
been readily detected. During 1966, a gamma-ray
detector (with spectrometer) tested as an underwater
monitor at the point of cooling-water discharge
showed the presence of only one radionuclide -
24Na--at the concentration of 1.3 x 10-1°
4.3.3 Performance of the ion-exchange columns
for collecting radionuclides. Relatively large volumes
of water were passed through the columns because
the hardness of the water was very low - 9 mg/liter in
terms of CaCO3 in both inlet and outlet samples. On
each cation-exchange resin, approximately 60 pCi
90sr and 9 pCi 13?cs were retained. The
distribution of these radionuclides on each column
was:
section
top
middle
bottom
90S,
81 ± 5*%
14 ±4
5 + 1
It was not necessary to wash suspended solids
from these columns as was done during continuous
sampling of coolant water (9) because the solids had
settled in the barrels that held the water prior to
passage through the columns. It would be desirable in
future studies to collect and analyze the associated
suspended solids.
4,3A Radionuclides in yard-drain effluent. Tritium
and 6OCo were found in water samples from both
drains; 54Mn, 90sr, and 95Zi were detected in water
from the east yard drain only (see Table 4.3),
possibly because water from the west yard drain was
analyzed with lesser sensitivity. Of these
radionuclides, ^H and probably 54Mn and 60Co came
from Yankee operations, and the others from fallout.
Average concentrations of radionuclides in rainwater
at Cincinnati during May and June, 1969, were:
81 ±5*%
12±3
7±2
* + values are one-half range of percent
values for influent and effluent
The sequential percentages suggest that, at most, 2
percent of the ionic strontium and 4 percent of the
cesium were not retained on the columns. The devices
are therefore useful for concentrating these
radionuclides under the indicated conditions.
54Mn
6°Co
9<>Sr
106Ru 2 x 10-8
125Sb<2xlO-9
3x10-9
4x10-8
<2 x 10-9
<2xlO-9
3x10-9
jxlo-8
Tritium concentrations in rain at nine locations in the
U.S. were all below 2 x 10-6/xCi/ml during 1969 .((12)
Concentrations of 90sr, 95zr + 95Nb, 106RU, and
13?Cs in Cincinnati rainwater were considerably
higher than detected or minimum detectable
concentrations for the yard-drain water, possibly
Table 4.3
Radionuclide Concentration in Yard Drains
East yard drain
Water, MCi/ml Sand, pQ/c
Radionuclide June 3, 1969 June 3, 1969 June
3H
54Mn
58Co
60Co
89Sr
90sr
95Zr
106Ru
125Sb
137Cs
144Ce
Notes: 1.
2.
3.
4.
5.
1.4 x lO-5 NA
4 x 10-9 1.7
< 3 x 10-9 0.1 <
5 x 10-9 3.0
< 1 x 10-9 < o.l
1.5 x 10-9 0.1
2 x 10-9 1.2
< 2 x 10-9 0.4
< 2 x 10-9 o.l
<,2 xlO-? 1.0
< 2 x 10-9 0.7
West yard drain
10, 1970
NA
3.2
:o.i
5.1
0.4
0.5
NA
NA
NA
3.1
NA
Water, /x Ci/ml
June 10, 1970
7.5 x 10-6
< 5 x 10-8
< 5 x 10-8
3 xlO-8
< 2 x 10-8
< 1 xlO-8
< 1 x 10-8
NA
< 5 xlO-8
< 1 x 10-8
NA
Sand, pCi/g
June 10, 1970
NA
0.2
'< 0.1
0.3
< 0.1
0.3
0.9
NA
NA
0.5
NA
radionuclide concentrations are at time of sampling.
1 pCi/g = 1 x 10-6 /iCi/g
NA: not analyzed
< values are 3
-------
because these radionuclides were retained on soil
during runoff of rainwater.
Tritium concentrations above background have
been reported in the east and west storm drains by
Yankee's contractor for environmental surveillance
on several occasions during 1968 and 1969. 03)
Values ranged from 2.2 x 10"4 to < 2 x 10-6juCi/ml.
A single beta activity value -- 2.38 x 10-8juCi/ml --
was above background. The variation in reported 3R
values suggests that radionuclides from Yankee were
only occasionally in the yard drain.
The 54Mn, 58Co and 60c0 in the sand and gravel
over which the water flows (see Table 4.3) are
attributed to Yankee, and were undoubtedly
deposited from the water. The other radionuclides in
the solids are at similar or higher concentrations in
soil at other locations - several were found in the
3H (s,i)
14C(s)
24Na (i)
32p (S;i)
51Cr (s,i)
54Mn (s,i)
55pe (s)
59pe (i)
58Co (i)
60Co (i)
63Ni (s)
64Cu (i)
9<>Sr (s)
95Zr(s,i)
95Nb (s,i)
99Mo (i)
110mAg(S)i)
124Sb (s,i)
1311 (s)
133j (s)
135i (S)
137Cs (s)
10 CFR 20
limit,* /uCi/ml
3 x ID'3
8x10-4
3 x lO-5
2 x ID'5
2 x ID'3
IxlO-4
8 x 10-4
5 x 10-5
9 x 10-5
3 x 10-5
3 x 10-5
2x ID"4
3 x 10-7
6 x 10-5
IxlO-4
4 x ID'5
3 x 10-5
2 x 10-5
3 x 10-7
1 x 10'6
4 x 10-6
2 x 10-5
Annual release
limit,** Ci
8 x 105
2xlfl5
8xlfl3
6xlfl3
6x105
3xl04
2x105
1 xlO4
3 x 104
8xl03
8xlfl3
6xlfl4
8x101
2x104
3xl04
IxlO4
8x103
6x103
SxlQl
3xl02
IxlO3
6x103
Estimated annual
release, + Ci
8x102
1 x 10-2
(3)
8 x 10'5
2 x ID'2
1 x ID'2
1 x 10-2
4 x 10-3
1 x 10-2
2 x 10-3
1 x 10-3
(7 x 10'2)
9 x 10"5
4 x 10-3
3 x ID'3
(1 x 10'2)
1 x 10-3
2 x 10-3
4 x 10-3
(7 x 10'2)
(9 x 10'2)
2xl04
samples listed in Table 6.3 - and are attributed to
fallout from atmospheric nuclear weapon tests. The
radionuclides were found in both gravel and sand, but
at somewhat higher concentrations in the smaller
particles.
4.3.5 Release limits and estimated annual
radionuclide releases. Amounts of individual
radionuclides in liquid wastes were calculated by
multiplying concentrations in reactor-plant liquid
waste (Section 4.3.1) and secondary system steam
generator blowdown (Section 2.3.5) by the volumes
of waste water discharged annually (Section 4.1.2).
The yard drains did not contribute significantly to
these totals, according to Table 4.3. The releases
compare as follows with the AEC limits for aqueous
discharges:
Percent
of limit
0.1
< 0.001
(0.04)
< 0.001
< 0.001
< 0.001
< 0.001
< 0.001
< 0.001
< 0.001
< 0.001
«0.001)
< 0.001
<0.001
< 0.001
(< 0.001)
-------
The estimated annual release of 3R was 0.1
percent of the limit, that of 131j was Q.005 percent,
and all other measured radionuclides were at
considerably lower percentages of the limit. Of the
shorter-lived radionuclides whose concentrations in
effluent water from the secondary system was
inferred from analyses of main coolant water (see
Section 4.3.2), the amounts of released 24Na and
133i were estimated to be, respectively, 0.04 and
0.02 percent of the limits; the others were at much
lower percentages. The estimated annual release of
3H is almost 2-fold lower than reported by Yankee
(see Section 4.1.2), and the sum of all other
radionuclides is several-fold higher. Note that these
calculations are based on only a few sets of
radioactivity data, and are therefore indications of
the magnitude of individual radionuclide discharges
rather than exact values.
These amounts of radionuclides in water at
Yankee have no direct health implication because the
Sherman Reservoir, the Deerfield River downstream
from Yankee, and the Connecticut River below its
confluence with the Deerfield River are not sources
of public water supplies. The intake of radionuclides
through eating fish caught in these waters is
considered in Section 5.5.3.
Federal Regulations Part 20, U. S. Gov't. Printing
Office, Washington, D. C. (1965).
8. Taylor, Worthen H., Massachusetts Department
of Public Health, Division of Sanitary Engineering,
letter to Yankee Atomic Electric Co. (April 5,1968).
9. Kahn, B. et al, "Radiological Surveillance
Studies at a Boiling Water Nuclear Power Reactor",
Public Health Service Kept. BRH/DER 70-1 (1970).
10. Simmons, W. A., Massachusetts Department of
Public Health, private communication (1969).
11. Riel, G. K. and R. Duffey, "Monitoring of
Radioisotopes in Environmental Water", Trans. Am.
Nucl. Soc. 11, 52 (1968).
12. Bureau of Radiological Health, "Tritium in
Precipitation", Radiol. Health Data Rep. 11, 313,
354(1970).
13. "1968 Annual Report, Environs Monitoring
Program, Yankee Atomic Nuclear Power Station";
"1969 Annual Report, Environs Monitoring Program,
Yankee Atomic Nuclear Power Station", Isotopes,
Westwood, N.J. (1968,1969).
4.4 References
1. Yankee Nuclear Power Station-Yankee Atomic
Electric Co., "Technical Information and Final
Hazards Summary Report". AEC Docket No. 50-29
(1960).
2. Blomeke , J. O. and F. E. Harrington,
"Management of Radioactive Wastes at Nuclear
Power Stations", AEC Rept. ORNL4070 (1968).
3. Pike, David, Yankee Nuclear Power Station,
personal communication (1969).
4. "Yankee Nuclear Power Station Operation
Report No. 121 for the Month of January 1971",
Yankee Atomic Electric Co., Boston (1971).
5. "Yankee Nuclear Power Station Operation
Report No. 109 for the Month of January 1970",
Yankee Atomic Electric Co., Boston (1970).
6. Logsdon, J. E. and R. I. Chissler, "Radioactive
Waste Discharges to the Environment from Nuclear
Power Facilities", Public Health Service Rept.
BRH/DER 70-2 (1970).
7. U. S. Atomic Energy Commission, "Standards
for Protection Against Radiation", Title 10, Code of
44
-------
5. Radionuclides in the Aquatic Environment
S.I Introduction
5.1.1 Studies near Yankee. A preliminary
examination of release data in Sections 3.1.2 and
4.1.2 suggested that the only location in the
environment where radionuclides from Yankee might
be found was Sherman Reservoir near the circulating
coolant water outlet. Efforts to detect and measure
effluent radionuclides were therefore concentrated in
this area. These studies are described in detail in
Sections 5.2 to 5.6. In brief, they consisted of the
following:
(1) Tritium concentrations in the Sherman
Reservoir and Deerfield River below the
Reservoir were measured during and after
release of a batch of radioactive liquid
waste by Yankee. As indicated in Table
4.2, tritium was the only radionuclide in
this waste that could be detected at the
point of discharge. Tritium
concentrations above background were
found just beyond the point of discharge
and in the Deerfield River below the
Sherman Reservoir.
(2) Water samples were also analyzed by
gross beta, gamma-ray spectrometric, and
radiostrontium measurements. Plankton
samples were collected throughout
Sherman Reservoir and analyzed in the
same way, but only very small samples
could be obtained. No radioactivity
attributable to Yankee was detected in
any sample.
(3) Radiostrontium and photon-emitting
radionuclides were measured in a sample
of water moss and a sample of dead leaves
from Sherman Reservoir near Yankee.
Both media apparently collected some of
the radionuclides discharged by Yankee.
Algae were looked for on June 3 and July
19, 1969, but were not found growing,
presumably because the water was too
cold.
(4) Radionuclide contents were compared in
fish from Sherman Reservoir and from
Harriman Reservoir, upstream from
Yankee. Only 90sr, 13?Cs and traces
of 22Na were found in both sets of fish
samples.
(5) Benthal samples - mostly bottom mud -
were collected both by diver and with
dredges from a boat, and examined for
radionuclide content by gamma-ray
spectrometry and 90sr analysis. The
bottom of the southern end of Sherman
Reservoir was monitored with an
underwater Nal(Tl) probe connected to a
portable gamma-ray spectrometry system.
Radionuclides attributable to Yankee
were found in the samples and with the
probe throughout the southern end of
Sherman Reservoir.
Radioactivity attributed to Yankee in water
(tritium) and in benthal samples had been observed
previously by Yankee's contractor for environmental
surveillance/!) Such radioactivity had also been
detected in sediment samples by the Massachusetts
Department of Public Health (MDPH). (2) Gross beta
activity measured in Sherman Reservoir water during
previous years 0>2) showed no increase due to
Yankee, in accord with the observations in this study.
Gamma-ray spectra from a Nal(Tl) detector
immersed in water at the circulating coolant outlet on
June 16, 1966, had shown naturally occurring 40K,
226Ra, and 232jh, and a trace of 24Na (0.13 ± 0.1
pCi/liter) from Yankee. (3)
At the Indian Point I PWR, low levels of 24Na>
56Mn and 131i were observed in discharge water with
the immersed NaI(Tl) detector; (3) and 54Mn, 58c0)
60co, 134cs, and 13?cs were detected in sediment,
aquatic vegetation, and fish below the outfall. (4) At
the Dresden I BWR, 58Co 60co, 89Sr, 90Sr, 131i,
134cs, 13?Cs, and l^Ofia were found in effluent
45
-------
coolant-canal water during waste discharges. (5)
5.1.2 Deerfield River and Sherman Reservoir. The
Deerfield River is formed by several branches that
arise in the Green Mountains of southern Vermont. It
empties into the Connecticut River near Greenfield,
Mass., 40 river miles* below the Sherman Dam. The
river is used intensively for generating power, and its
flow is closely controlled for this purpose at the large
Somerset and Harriman Reservoirs, upstream from
Sherman Reservoir. Water flows into the northern
end of Sherman Reservoir from discharge at the
Harriman hydroelectric station and/or Harriman
Dam; it flows out of Sherman Reservoir through the
intake of Sherman hydroelectric station, and/or the
sluice and spillway of Sherman Dam. Approximately
0.7 miles below Sherman Dam is Dam No. 5, which
impounds water for use by hydroelectric station No.
5 and a paper (glassine) manufacturer at Monroe
Bridge. The Deerfield River is used for sport fishing
but not for public water supply. (6) Flow data for the
Deerfield River at the USGS Charlemont Gaging
Station on the left bank near Deerfield River Mile 26
(DRM 26) from 1913 to 1966 are as follows: (7)
maximum daily, Sept 21,1938
minimum daily, June 17,1921
mean daily in 1965
56,300 cfs (flood)
5
528
During the field trip described here, the average dairy
river flow was as follows:
Date, 1969
June 2
June 3
June 4
Sherman Station^8*
701 cfc
728
560
Chartemont
gaging station
831 cfs
993
750
(9)
The Sherman Reservoir, located at the
Vermont-Massachusetts border, is approximately
rectangular with a narrow neck at its northern end, as
shown in Figure 5.1. The rectangle is approximately
8,000 ft (2,400 m) long and 850 ft (260 m) wide, and
the lake extends to a depth of 80 ft (24 m). On the
basis of these dimensions, it was estimated to have a
capacity of 3 x 10^ ft3. The water is cold; on June 3,
1969, it was 54»F at the surface and 47°F at a depth
of approximately 8 m; on July 29, 1969, it was 60°
to 63°F at the surface. Effluent circulating coolant
water from Yankee is approximately 15°F warmer
than influent water. The water is very soft (i.e., low
calcium plus magnesium content), according to the
analytical data in Table 5.1.
Yankee is on the southern shore of Sherman
Reservoir, as shown in Figure 5.2. The locations of
the intake and outlet for circulating coolant water
and of nearby sampling points are shown in Figure
5.2, and in greater detail in Figure 5.3. Note the
proximity of the Yankee Station water outlet to the
Sherman Station water intake.
5.2 Tritimm in Water
5.2.1 Sampling and analysis. Water was collected
to measure tritium concentrations beyond the point
of release during and after the discharge of
reactor-plant waste solution (see Section 4.1.3) into
effluent circulating coolant water. Samples were
obtained at the locations and times listed in Table
5.2. Water was collected in 50-ml portions at the
water surface and, in some instances, 2.5 m below the
surface (see Note 2 to Table 5.2). All of the samples
at the south end of Sherman Reservoir were collected
while the waste solution was being released. The
Sherman hydroelectric power station was operating
during the entire period, and a distinct pattern of
water flow from the Yankee outlet to the Sherman
water intake was visible.
The water samples were prepared for tritium
analysis by distilling at least 10 ml of water to
separate tritium from nonvolatile radionuclides. The
distilled water was then mixed with scintillating
solution to measure the tritium in a
liquid-scintillation counter. The energy-response
settings of the counter were adjusted to optimize
detection of the low-energy beta particles of ^H. For
routine analysis, the minimum detectable
concentration was 2 pCi/ml. Some samples were
counted with an improved detection limit of 0.2
pCi/ml in a modified liquid-scintillation apparatus, t
Results at the higher concentrations were confirmed
by analyzing several samples with both detectors.
5.2.2 Results and discussion. The 3H
concentration at the circulating coolant water outlet
was 79 pCi/ml (Table 4.2). Values from the traverses
in front of the coolant-water outlet-jsamples 22 A to
E and 23 A to E in Table 5.3 -- conform to the
observed flow pattern in that tritium concentrations
were relatively high near the water intake for
1 mile =1.61 km; 1 cubic foot per second (cfs) = 28.3 liter/sec, t We thank R. Lieberman, SERHL, EPA, for these analyses.
46
-------
NORTH BRANCH
DEERFIELD R.
VEST BRANCH
DEERFIELD R.
SAMPLE KEY
Grass (?) Benthos
Milk (§) Snow
S3 Soi I tfS later
SOUTH BRANCH
OEERFIELO R.
YANKEE NUCLEAR POWER STATION
VERMONT^
MASSACHUSETTS
0 *
1 I
3
I t 1
6
I 1
9
1 > 1
FLORIDA BRIDGE
KILOMETERS
COLD RIVER
STATE RT. 2
EAST
CHARLEMONT
Figure 5.1. Deerfield River Near Yankee Nuclear Power Station.
47
-------
ELEVATION
IM FEET
BARBED WIRE
EXCLUSION FENCE
Figure 5.2. Yankee Nuclear Power Station. Note: Elevations refer to New England Power
Co. datum; add 106 ft to obtain U3GS elevation above mean sea level.
48
-------
SHERMAN
RESERVOIR
Figure 5.3. Yankee Nuclear Power Station Detailed Plan.
49
-------
Table 5.1
Concentration of Stable Substances in Water from Deerfield River
Substance
Sodium
Magnesium
Potassium
Calcium
lion
Aluminum
Boron
Manganese
Zinc
Nickel
Barium
Strontium
Copper
Arsenic
Beryllium
Cadmium
Chromium
Cobalt
Lead
Molybdenum
Phosphorus
Silver
Vanadium
Concentration, /JL g/liter
3,500
2,900
900
700
140
80
57
55
36
21
17
6
5
<13
< 0.03
< 3
< 1
< 3
< 5
< 5
<13
< 0.3
< 5
Notes:
1. Sample was collected at location #27, below outflow of Sherman Station, on June 3,1969 at 1345.
2. We thank Robert Kroner, Water Quality Office, EPA, Cincinnati, Ohio, for this analysis.
3. Concentrations were measured by emission-spectrographic analysis, except that sodium, magnesium,
potassium, and calcium were by atomic absorption spectrometry.
Table 5.2
Tritium Sampling Points
#
1
20
21
22 A
B
C
D
E
23 C
23 A
B
C
D
E
26
27
Location
Sherman Reservoir near 300-m station perimeter
north of Harriman Station (backgrou' d)
Deerfield River west of Charlemont (DRM 27)
Sherman Reservoir, 8 m north of outlet weir at west shore
3 m from west shore
at centerline
3 m from east shore
at east shore
(2) 16 m north of outlet weir at centerline
16m north of outlet weir at west shore
3 m from west shore
at centerline
3 m from east shore
at east shore
northwest of Sherman Station intake
#5 Reservoir below Sherman Station outlet (DRM 40)
Collection date and time,
1969
June 3, 1200
June 4, 1000
June 4, 1600
June 3, 1200-1210
June3, 1300
June 3, 1215-1225
June 3, 1230
June 3, 1345
Notes:
1. numbers indicate locations shown on Figures 5.1, 5.2 and 5.3.
2. all samples were collected at surface; in addition, samples were collected at 2.5-m depth for #1, 22, 23 and 26.
50
-------
Table 5.3
Tritium Concentration in Sherman Reservoir and Deerfield River
Depth
Concentration, pCi/ml
22
23(2)
23
26
surface
2.5-m depth
surface
surface
2.5-m depth
surface
2.5m depth
Sherman Reservoir, near weir
A _B
72 65
69 58
41
18
4
<2
42
32
C_
65
60
16
1.5
16
62
12
0.6
<2
4
<2
<2
3
1 surface
2.5-m depth
20
27
21
surface
surface
surface
Sherman Reservoir, near 300-m perimeter
<2
< 2
Sherman Reservoir, background
0.4 ± 0.2
Deerfield River, below Sherman Station
27
2.4 ± 0.4
Notes:
1. Sampling points are described in Table 5.2. See Figure 5.1 for locations #20 to 21; Figure 5.2 for locations #1 and
27; and Figure 5.3 for locations #22, 23, and 26.
2. + values are 2
-------
Thus, tritium in radioactive liquid wastes
discharged at Yankee can be used as a tracer to
determine dispersion near the point of release and
dilution in the Deerfield River. The dispersion would
be different from the observed pattern when Sherman
Station does not operate, so that the water is retained
in Sherman Reservoir or released at the dam. The
short-term concentrations of 3n at the point of
discharge and beyond were below 3 percent of the
limiting annual average of 3,000 pCi/ml (3 x 10-3
/tCi/ml) given in 10 CFR 20. Because the water is not
ingested by humans, there is no direct radiation
exposure to humans by this route.
5.3 Other Radiommcll***
in Water
5.3.1 Unfiltered samples. Water samples (3.5 liters)
were collected at locations No. 20, 21, and 27 at the
same time as the tritium samples, and also at the
Yankee outlet weir on June 3, 1969, at 1000, before
liquid waste was released from the Test Tank. The
samples were acidified with 10 ml (Concentrated
HC1, evaporated to 45 ml, and analyzed with a
Nal(Tl) gamma-ray spectrometer. Thirty ml of each
concentrated solution were then analyzed
radiochemically for 89sr, 90sr, and 137Cs, and 15 ml
were evaporated to dryness and counted with a
low-background G-M detector for gross beta activity.
The average gross beta activity of the four samples
in Table 5.4 and the two in Table 4.2 was 2.3 ± 0.2 x
10-9 ^Ci/ml, and the average 90sr co.itent, 1.1 ± 0.3
x 10 -9 /liCi/ml; no individual sample. had significantly
higher values than the averages, hence the 90$r in
water is attributed to fallout from atmospheric
nuclear weapon tests, and the gross beta activity, to
fallout plus naturally occurring radionuclides. No
89Sr ( < 2 x lO^Ci/ml) or 13?cs ( < 5 x 10-10
/tCi/ml) was detected by radiochemical analysis, and
no radionuclides (generally < 2 x 10-9 /iCi/ml)
were found by gamma-ray spectrometry. These
results are consistent with the calculated discharges in
Table 4.2.
The gross beta activity measured in Sherman
Reservoir and the Deerfield River is within the
ranges of the most recent published data by the
MDPH and Yankee's contractor for environmental
surveillance, but is considerably below maximum
values reported by the latter:
MDPH (2) Yankee contractor
Location Mav-Nov.. 1968 Jan. -Dec..
Harriman Station
Sherman Reservoir
Sherman Dam Sluiceway
Station #5
Monroe Bridge
2-6 pa/liter*
2-8
1-6
1-5
< 4.5-377 pCi/liter
< 4.5-21
< 4.5-12
* 1 pCi/liter = 1 x 10-9/*Ci/ml.
The highest concentration reported by Yankee's
contractor - 377 pCi/liter at Harriman Station on
Oct. 31, 1969 - was found by the contractor to be
due to dissolved 60co;0) the source of this
radionuclide, at a location upstream from Yankee, is
unknown, but laboratory contamination may be a
.possibility. The gross alpha activity during 1969,
Table 5.4
Gross Beta Activity and 9°Sr Concentration in Water
from Sherman Reservoir and Deerfield River, pCi/liter
I
20
27
21
Notes:
1.
2.
3.
Sample Gross beta 90Sr
Yankee outlet, no waste discharged 2.4±0.5 I.I ±0.5
Sherman Reservoir water (background) 2.5 + 0.5 1.0 + 0.5
#5 Reservoir, DRM 40 2.2 + 0.5 0.7 + 0.5
Deerfield River, DRM 27 1 .9 ± 0.5 1.1 + 0.5
pCi/liter = 1 x 10"' juCi/ml;± values are 2 o-counting error.
See Table 5.2 for sampling locations and times; water at outlet was sampled on June 3, 1969 at 1000.
Values are based on 1 liter unfiltered water for gross beta and 2 liters unfiltered water for 90Sr.
52
-------
measured by the contractor, (1) was <2.3 pCi/liter
except for one or two values near the minimum
detectable level at all three sampling locations. The
most recent radioactivity concentrations reported for
raw surface water in the general area by the Federal
Water Pollution Control Administration are for the
Connecticut River:
gross beta activity (Wilder, Vt, Dec. 1968)d3): 4 pa/liter
90Sr (Narthfldd, Mass., July-Sept 1967)(14>: 1-1
5.3.2 Suspended solids. An 11.4-liter sample of
water, collected at location 23 C(2) (see Table 5.2)
during release of Test Tank waste by Yankee, was
immediately passed through a membrane filter (8-n
pore diameter) to separate suspended solids for
radiometric analysis. The filter was counted for 1000
minutes with a Nal(Tl) gamma-ray spectrometer, and
was then ashed, weighed, and analyzed chemically for
radiostrontium and radiocesium content.
Three macroplankton samples were collected from
Sherman Reservoir on July 29, 1969, by towing a
10-cm-dia. plankton net at a depth of 1.5 m behind a
slowly moving boat. The samples were obtained in
front of the Yankee outlet, in the bay east of the
Yankee pump house, and just upstream from the
outlet at Harriman Station, to provide a background
value. The volume of sampled water was estimated to
be between 250 and 500 liters in each collection.
Because of heavy rains, the Reservoir was
approximately 2 m higher than during the June field
trip, and the water was muddy. Very little plankton
was observed. The plankton samples were separated
on Whatman No. 41 filter paper from the 50 -100 ml
of water in which they were suspended. They were
then ashed and analyzed in the same way as the filter
sample described above.
The sample collected on the membrane filter,
which appeared to be mostly silt, contained a small
amount of 90sr but no detectable 13?cs (see Table
5.5). The samples collected with the plankton net
contained some silt -- especially the heaviest sample --
and showed 137cs but no 90sr. The concentration of
the two radionuclides per liter of water, based on the
values in Table 5.5, were:
Sample
filter-Yankee outlet
net-Yankee outlet
Yankee bay
Harriman Station
90Sr, pa/liter
0.12
< 0.0005
< 0.0008
< 0.0005
137Cs, pCi/liter
<0.05
0.004
0.002
0.004
The filtered sample had approximately 10 percent of
the 90sr concentration in unfiltered water (see Table
5.4), while the other samples had less than 0.1
percent. The 13?Cs concentration in the samples
collected with the plankton net was approximately 7
percent of the 13?Cs concentration in Reservoir
water (see Table 4.2). No 89Sr (<1 pCi/sample) or
long-lived photon-emitting radionuclides (generally
<5 pCi/sample) were detected in these samples. The
90sr and 13?Cs in these samples are attributed to
fallout.
Table 5.5
Gross Beta Activity and Concentrations of 90$r and
from Surface Water in Sherman Reservoir
in Suspended Solids
Sample
Water volume, Ash wt, Radionuclide content pCi/sample
liter mg Gross beta
June 3,1969, membrane filter
near Yankee outlet 11.4
July 19,1969, plankton net
near Yankee outlet ~ 400
bay east of pump house ~400
near Harriman Station
outlet ~400
20.4
60.1
12.7
233.
NA
4.0
1.4
9.0
1.4 + 0.3
<0.5
<0.2
<0.3
<0.2
1.4 + 0.5
0.8 ± 0.4
1.6 + 0.4
Notes:
1. NA - not analyzed.
2. ± values are 2 a error of counting; < values are 3 a
error of counting.
53
-------
fm detector plus 1600-channel analyzer and with a
Nal(Tl) detector plus 200-channel analyzer to
identify and quantify photon-emitting radionuclides.
Spectra obtained with the Ge(Li) detector are shown
5.4.1 Sampling and analysis. Dead leaves that were " Fi&urf 5"4; 5'5> md 5.6. Radiochemical analysis
barely submerged at the edge of Sherman Reservoir was Perf™d to measure 90Sr and to confirm the
near location No. 2 (see Figure 5.2) were collected on gamma-spectral identification of 137Cs and 106Ru.
June 2,1969. Common water moss, Fontinalissp.,* In addltlon' stable calcium and strontium were
was collected on June 3, 1969 from rocks at a depth measured by atomic absorption spectroscopy, and the
of 2.5 m near the shore at location No. 23 E (in front silica content was determined by gravimetric analysis.
of Yankee outlet weir - see Figure 5.3). The samples 5-4-2 Results and discussion. Longer-lived fission
were weighed while wet, after drying at 100°C, and and activation products were detected in the two
again after ashing at 400°C. Both samples contained samples at the concentrations given in Table 5.6. The
silt. 60co in both samples and 54Mn and 58co in moss
The ashed samples were analyzed with a Ge(Li) are attributable to Yankee; all other radionuclides are
Table 5.6
Radionuclides in Water Moss and Dead Leaves
from Sherman Reservoir, pCi/g ash weight
Radionuclide Water moss Dead leaves
54Mn
58Co
60co
90Sr
95Zr
95Nb
103Ru
106Ru
125Sb
137Cs
141Ce
144ce
40K
226Ra
232Th
?8e
Sr++ (mg/gash)
Ca"1"1' (mg/gash)
Si°2 (mg/gash)
Ash wt/wet wt, %
Dried wt/wet wt, %
Notes:
1 . Water moss was collected on
26
~4
13
4 .
12
18
~12
28
1
2
~9
58
180
15
~70
~38
0.14
13.7
124
7.1
14.5
June 3, 1969, from rock near
2
NM
4
3
<4
NM
NM
4
2
6
NM
10
27
3
~6
NM
0.15
12.2
278
6.7
15.5
weir (location 23 E) at 2.5-m
depth; dead leaves were collected on June 2, 1969, at water-line on shore of Sherman
Reservoir (near location 2).
2. < values are 3 (T counting error.
3. Concentrations pertain to collection time.
4. NM - not measured.
5- The following radionuclides were not detectable: 51Cr «15 pCi/g); 59Fe «5 pCi/g);
89Sr and 11(Hvg «1 PCi/g),
* We thank M. C. Palmer, Environmental Protection Agency, Cincinnati, for identifying the moss.
54
-------
10
1.0
£' 0.1
0.01
co o> CM o in
o co -^ r» o
CN CN CN (N CM
•* P- COCO CD CO
C* CM lOCD CD OS
O 101 2
100
200 300 400 500
CHANNEL NO. (1.00 keV/channel)
600
700
800
Figure 5.4. Gamma-ray Spectrum o/ Water Moss, 0 - 800 feeV.
Detector: Ge(Li), 10.lt cm2 x 14 mm, trapezoidal
Sample : 18 g (35 cc) ash, collected June 3, 1969, in Sherman Reservoir near outlet for
Yankee cooling water.
Counts : (upper curve) July 9-10, 1969 (1,000 minutes, background not subtracted);
(lower curve) counter background; Ra and Th refer to 2%?a and 23277i plus progeny.
-------
1.0
0.
0.01
0.001
800
900
1,000
1,100
1,200
1,300
1,400
1,500
1,600
CHANNEL NO. (1.00 keV channel )
Figure 5.5, Gamma-ray Spectrum of Water Moss, 800 - 1,600 keV.
Detector: Ge(Li), 10.4 cm X 11 mm, trapezoidal
Sample : 18 g (35 cc) ash, collected June 3, 1969, in Sherman Reservoir near outlet for
Yankee cooling water.
Counts : (upper curve) July 9-10, 1969 (1,000 minutes, background not subtracted);
(lower curve) counter background; Ra and Th refer io226fla and 23277i plus progeny.
-------
.600
CHANNEL NO. (1.004 keV/channel)
Figure 5.6. Gamma-ray Spectrum of Dead Leaves from Sherman Reservoir.
Detector: Ge(Li), 10.4 cm2 x 11 mm, trapezoidal
Sample : 210 g (450 cc) ash, collected June 2, 1969, at east shore near 300-m perimeter.
Count : Nov. 12-13, 1969 (1,000 min.); Ra, Th, and Bkgd refer to 226Ra plus progeny.
2327>i plus progeny,and counter background (see Figures 5.4 and 5.5), respectively.
-------
probably from fallout (see Table 6.2 for radionuclide
content of vegetation samples collected on land) or
occur naturally. The radionuclides may be in both the
organic material and the accompanying silt (see
Section 5.6 for the radionuclide content of sand, silt
and clay in sediment samples). Because of its
proximity to the outlet, the water moss would be
expected to collect radionuclides discharged in
circulating coolant water. The dead leaves were
collected within 200 m of the east yard drain, and
may have retained radionuclides from effluents at
that drain.
In terms of wet weight, the concentrations of
, 58co, and °0co in the moss and leaves range
from 1 x 10-7 to 2 x 10-6 MCi/g (see Table 5.7).
Accumulation factors for these radionuclides-defined
as concentration in the media divided by the
concentration in water-can not be calculated because
neither the average radionuclide concentrations in
water near the media nor the exposure periods of the
media are known. The accumulation factors of 90sr
and 13?cs from fallout and stable strontium and
calcium from Sherman Reservoir water are
approximately 2000 in moss, as shown in Table 5.7.
Thus, despite the extremely low concentrations of
radionuclides discharged by Yankee, some of these
radionuclides could be detected in organic material in
Sherman Reservoir, at concentrations considerably
higher than in the water (see Table 4.2, last column).
The moss was seen only near the discharge weir, and
may be confined to that area because the water is
colder everywhere else; dead leaves are found at many
locations near the edge of Sherman Reservoir.
Concentration of radionuclides in these media does
not appear to have any consequence as a health
hazard to humans through consumption, external
radiation, or return of radionuch'des to water after
concentration. In future studies, it would be of
interest to analyze these media again, both at the
indicated sites and at background locations, to check
attribution of the noted radionuclides to Yankee, and
to examine the use of these media as convenient
indicators of discharged radionuclides.
S.S Ra4io*ttetldes im FiaJk
5.5.1 Collection and analysis. Fish were collected
on June 18, 1969, from both the Sherman and
Harriman Reservoirs by the electro-shocking
method.' As shown in Figure 5.1, the two reservoirs
are well separated, hence movement of fish between
them is unlikely and the fish from Harriman
Reservoir can serve as the background sample.
The collected fish are listed in Table 5.8. The fish
from each reservoir were combined in three categories
according to their feeding habits: bottom feeders,
insect eaters and predators. Catfish, a bottom feeder,
were analyzed separately since this type of fish was
available from both reservoirs. Also listed are the
numbers of fish, total wet weight, and age as
determined by annular scale marks. Ages of the
crappie and catfish are unknown.
TaWe 5.7
Radionuclide Concentration in Water Moss and Dead Leaves
Water moss
Dead leaves
Substance
Amount per wet wt. Accumulation factor*
Amount per wet wt
54Mn(fiCi/g)
58co
60Co
9°Sr
137Cs
Si++( n g/g)
Ca^ (mg/g)
1.8 x 10-6
3 x 10-7
9 x 10-7
3 x 10-7
1.4 x 10-7
9.9
0.97
—
—
—
9 x Ifl2
3 xlQ3
1.6 x 103
1.4 x 103
1.4 x 10-7
—
3 x 10-7
2 x 10-7
4 x 10-7
10.0
0.82
*Calculated by dividing values in preceding column by concentration in water. Concentrations in water for
90Sr and 137^5 are average values from analysis of ion-exchange resin in Table 4.2j for Sr"1""1" and Ca"1""1", values
are from Table 5.1.
t We thank Colton H. Bridges and associates, Bureau of Wildlife Research and Management, Division of
Fisheries and Game, State of Massachusetts, for collecting these samples and providing data on fish ages.
58
-------
Samples were frozen immediately after collection.
For analysis, the fish were thawed, weighed, and
dissected into the following tissues that were
expected to concentrate the radionuclides of interest:
muscle - 134Cs arid 13?Cs analysis
kidney + liver -5^ Fe, 58co and 60Co analysis
bone - 8"Sr and '^Sr analysis
No analyses were performed for 131] jp the thyroid
because of the lapse of time between sample
collection and analysis.
liver plus kidney were analyzed directly by
gamma-ray spectrometry with a NaI(Tl) detector, and
also with a NaI(TlJ "-gamma-ray
coincidence/anticoincidence spectrometer system.
The iron fraction was separated, and analyzed for
55pe with an x-ray proportional detector, and for
stable iron with an atomic absorption spectrometer.
Bone was ashed at 600°C, and strorttium was then
separated chemically. Radiostrontium was measured
by counting total strontium and 90y. Stable
strontium and calcium were determined by atomic
Table 5.8
Fish Collected in Sherman and Harriman Reservoirs
Reservoir
Sherman
Harriman
Category
Bottom Feeder
Insect Eater
Predator
Bottom Feeder
Insect Eater
Predator
Type
White sucker
Catfish, bull head
Rock bass
Golden shiner
Crappie
Yellow perch
Small mouth bass
Common sucker
Catfish, bull head
Rock bass
Yellow perch
Lake trout
Brown trout
Chain pickerel
Total weight,
kg (number)
7.2 (11)
1.9 (17)
0.55 (16)
0.30 ( 3)
0.20 ( 4)
0.65 (19)
1.3 ( 3)
3.9 (10)
0.65 (10)
1.3 (11)
1.10 ( 9)
0.60 ( 1)
0.50 ( 1)
1.8 ( 2)
Average age,
yr (range)
4.7 (2-8)
...
3.4 (2-7)
5.2 (2-7)
—
6.2 (2-9)
4.8 (2-7)
3.6 (2-5)
...
5.8 (3-7)
4.7 (4-6)
4
3
8.5 (8-9)
Table 5.9
Radionuclide (pCi/kg)3 and Stable Ion (g/kg)a Concentration in Fish Tissue
Bone
Category
Sherman Reservoir
Bottom Feeder-sucker
Bottom Feeder-catfish
Insect Eater
Predator
Harriman Reservoir
Bottom Feeder-sucker
Bottom Feeder-catfish
Insect Eater
Predator
90sr
2230
3510
3070
2950
2370
3530
2320
2790
Ca
32
36
40
41
38
36
32
32
Ash/wet
Sr weight 22jja 13?Cs
0.059
0.091
0.072
0.068
0.072
0.091
0.058
0.043
0.
0.
0.
0.
0.
0.
10
11
12
12
12
11
0.11
0.
10
3.1
3.1
2.0
3.0
1.9
NM
0.5
1.2
250
120
110
650
170
210
520
460
Muscle
K
3.42
2.82
3.56
4.42
3.09
3.62
3.60
4.12
Ca
0.59
0.35b
0.86
2.45
1.15
0.89
0.72
0.96
1.09
Si
Ash/wet
weight
0.0011 0.016
0.00064b 0.009b
0.0010
0.0019
0.0011
0.0021
0.0020
0.0010
0.0005
0.014
0.013
0.018
0.012
0.026
0.027
0.014
aAll kg values are wet weights.
"Bone was removed very thoroughly.
Notes:
1. ± values (2 a counting errors) are: 90Sr, 90 pCi/kg; 22Na, 0.3 pCi/kg; and 137Cs, 10 pCi/kg.
2. NM - not measured.
59
-------
absorption spectroscopy.
Muscle was ashed at 400°C and then analyzed by
gamma-ray spectrometry. Cesium-137 and 40K in
muscle were determined by gamma-ray spectrometry,
and the potassium content was calculated from the
40jC measurement. A gamma ray at 0.51 MeV energy
was observed by coincidence/anticoincidence
spectrometry and the emitting radionuclide was
identified as 22fta by its photon spectrum and by
chemical separation. Stabk strontium and calcium
were also measured in these samples. To evaluate the
contribution of incompletely separated bone to the
radiostrontium content in muscle, an additional
sample--muscle of sucker from the Sherman
Reservoir-was prepared for stable strontium and
calcium analysis with special care to remove all bone.
5.5.2 Results and discussion. The 90sr and *37Cs
concentrations in fish were not consistently higher in
Sherman Reservoir than in Harriman Reservoir (see
Table 5.9), hence these radionuclides in all of the fish
are attributed to fallout. The average 90sr
concentration in bone for all fish was 2840 pCi/kg
wet weight, 42 pCi/mg of strontium, and 79 pCi/g of
calcium. The average 137cs concentration in muscle
for all fish (adjusted for the number of fish in each
category), was 235 pCi/kg wet weight, and 67 pCi/g
of potassium. These concentrations fall within the
range of previously reported values. (20-22) The
average observed ratio for radiostrontium
[ORbone/water = (Sr/Ca) bone -KSr/Ca) water] was
0.17 ± 0.05 (2o) based on a 90sr concentration of
0.32 pCi/liter of water (Table 4.2) and a calcium
concentration of 0.7 mg/liter of water (Table 5.1).
The average OR bone/water f°r stable strontium was
0.22 ± 0.06 (2
-------
occurring 40j£ were found in any sample. The
minimum detectable level for '34cs in muscle was 2
pCi/kg, and for radiocobalt in kidney plus liver, 15
pCi/kg. The iron content of the kidney plus liver
samples ranged' from 0.04 to 0.26 g/kg; but no
5.5.3 Hypotheticalffadionuclide concentration in
fish. The concentrations of radionuclides in fish
exposed to radioactive effluent from Yankee was
computed to demonstrate the procedure and indicate
possible critical radiomiclides:
-
Radio-
nuclide
3H
14C
24Na
32?
siCr
S4Mn
sspe
S9Fe
58Co
60Co
63Ni
64Cu
90Sr
95Zr
95Nb
99Mo
i lOmAg
124Sb
1311
1331
1351
137CS
Annual average
concentration
in water, * /j. Ci/ml
3 x 1Q-6
4x 10'11
(1 x 10'8)
3xlO'13
7X10'11
4xlOM1
4 x 10'1 1
1 x 10"12
4x 10'11
7X10'11
4x 10-12
(2x 10'10)
3x 10'13
1 x 10'11
IxlO'11
(5 x 10'11)
4 x 10'12
7x 10'12
IxlO'11
(3 x 1(T10)
(SxlO'10)
7xlO'13
Concentration
factor, (29)
ml/g
0.9
4,550
31.7
100,000
200
25
300
300
500
500
40
200
40
100
30,000
100
3,080
40
1
1
1
1,000
Hypothetical
Concentration
in fish, +
yuCi/lOOgwetwt.
3 x 10'4
2 x 10's
(3 x 10'5)
3 x 10"6
1 x 10"6
1 x 10"7
1 x 10'6
3 x 10~7
2 x 10"6
4 x 10"7
2 x 10'8
(4 x 10"6)
1 x 10'9
1 x 10"7
3 x 10"5
(5 x 10'7)
1 x 10"6
3 x 10'8
1 x 10'9
(3 x 10"8)
(3 x 10'8)
7 x 10~8
Percent of
intake guide**-
0.005
0.001
(0.05)++
0.007
< 0.001
< 0.001
< 0.001
< 0.001
0.001
0.001
< 0.001
(0.001)
< 0.001
< 0.001
0.01
<(0.001)
0.002
< 0.001
< 0.001
(0.001)
«0.001)
< 0.001
*The estimated annual discharge (Section 4.3.5) divided by the flow of circulating coolant water of
2.8 x 1014 ml/yr.
+The product of the values in columns 2 and 3, multiplied by the estimated intake of 100 g/day.
The limiting concentrations from Section 4.3.5 multiplied by the water intake of 2,200 ml/day
on which the concentration limits are based, (30) except for limits for 90Sr (200 pCi/day)
and 131I (80 pCi/day) from Federal Radiation Council guidance. (31)
++Values in parentheses are based on inferred, not measured, concentrations.
( < 400 pCi/kg wet weight, < 10 pCi/mg Fe) was
found. Concentrations of 55pe between 3 and 50
pCi/mg iron have been reported for freshwater fish
collected in Finland during 1965. (28)
In summary, the only radionuclide in fish that
might be attributable to Yankee was 22Na at a
concentration above background of approximately 2
pCi/kg wet weight. Since there is fishing in Sherman
Reservoir, however, it appears reasonable to check
radionuclide concentrations periodically in the edible
portions of food fish.
This tabulation is based on an average daily intake
of 100 g fish (values of 50 g (32) and 100 g (33)
have been reported), a tabulation of concentration
factors for edible portions of freshwater fish, (29) the
radionuclide release estimates in Section 4.3.5' and
the assumption that the radionuclides in the edible
portions of all consumed fish had reached equilibrium
with radionuclide concentrations in circulating
coolant water at the point of discharge. Of these, the
radionuclide release estimates and many of the
concentration factors are quite approximate, and it is
61
-------
improbable that radioactive equilibrium is attained in
all fish.
The total estimated intake of radionuclides by
eating fish is, therefore, below 0.1 percent of the
intake guide. The dose rates from the listed
radionuclides are 0.3 mrem/yr to the gastrointestinal
tract (mostly from 95]%), 0.2 mrem/yr to bone
(from 32p); and less than 0.1 mrem/yr to the thyroid
and whole body. These values were computed by
comparing the hypothetical daily intakes (column 4,
above) to the maximum permissible daily
occupational drinking-water intakes listed by the
NCRP that correspond to 5 rem/yr to the total body,
15 rem/yr to the GI tract, and 30 rem/yr to bone,
(30) or directly applying FRC guidance for
radiostrontium and radioiodine. (31)
Of the listed radionuclides, 3ft, 24Na, and 95Nb
would be readily detected in fish muscle at the
indicated concentrations, and should be looked for in
fish samples from Sherman Reservoir near the Yankee
outfall or the Deerfield River below Sherman Dam. In
the analyzed samples (which were collected
throughout Sherman Reservoir, however), no 95Nb
was found (<2 x 10-6^Ci/100 g wet weight).
The average concentrations given in Section 5.5.2
of 22Na and 137Cs measured in fish muscle, and of
90Sr in muscle inferred from fish bone analyses,
correspond to the following annual radiation
exposure at a daily fish consumption of 100 g:
Radionuclide
22
!Na
90
137
'Sr
Average
concentration
in fish,
MCi/lOOg
2.1 x 10-7
3.2 x ID"6
2.4 x 10-5
Radiation
dose,
mrem/yr
0.005
2.7
0.3
Critical organ
GI tract (30)
bone (3D
whole body (31)
'Cs
As indicated in Section 5.5.2, the 90sr and 13?cs are
attributed to fallout, but most of the 22jsja may be
from Yankee.
5.6 Radio* mclidv* J»
Bvmtkal Sam
5.6.1 Sampling and on-site measurements. The
MDPH in 1965 found radionuclides from Yankee in
at least one of five benthal samples collected in
Sherman Reservoir. An effort was therefore made to
confirm this observation and to evaluate the extent of
the contamination. Three methods of determining
radioactivity in benthal samples were compared with
respect to sensitivity and ability to define the extent
and magnitude of the contamination. The methods
were:*
Use of a 10-cm x 10-cm NaI(Tl) detector
as submergible probe; gamma-ray spectra
were obtained for 4 - to 20-minute periods
while the detector rested on the bottom of
Sherman Reservoir.
Collection of core samples by a diver; the
core sample, 10 cm in diameter and 13.4
cm in depth, was separated into equal
upper and lower fractions and analyzed
for radionuclide content.
Collection of samples with an Ekman or a
Petersen dredge, and analyses for
radionuclide content.
The location and number of probe measurements
and samples are given in Table 5.10. Samples were
collected from a boat by lowering the diver into the
water or dropping a dredge to the Reservoir bottom;
the probe was lowered from a second boat which
contained the multichannel analyzer with associated
power supply (motor-generator) and recording
system. The probe was positioned on the Reservoir
bottom by the diver, who later collected benthal
samples by hand at the same location. Measurements
were taken and/or samples collected at three
locations across the Reservoir near the 300-m
perimeter relative to the containment sphere; at 7
points along the south shore of the Reservoir near
(1)
(2)
(3)
*We thank the MDPH, SERHL, and NERHL for making this study possible; especially Cornelius J.
O'Leary, MDPH, for providing equipment and guiding the sampling, Edw^' Crockett, MDPH, for
performing the diving, Charles Phillips, SERHL, for providing and operating th£4nderwater probe, and
Raymond H. Johnson, Jr.,NERHL, for providing sample collection equipment and advising on sampling
procedures.
62
-------
Yankee; at 2 locations in front of the Yankee
circulating coolant water outlet; at the north end of
the Reservoir to indicate background values; and at
one location in the Deerfleld River below the
Reservoir. Brief probe measurements were obtained
in the relatively shallow water along the south shore
of the Reservoir until the area of highest radionuclide
concentration was identified. In that area, 10 probe
measurements and 5 samples were taken to define the
extent of the contamination and the response of the
probe.
5.6.2 Description of benthal samples. * Five of the
samples were characterized as shown in Table 5.11.
(34) in brief, organic carbon was determined by
measuring the carbon dioxide formed in ashing the
samples, and the weight of organic matter was
estimated by multiplying the organic carbon content
by 1.72. Particle-size separation, was by wet sieving
and sedimentation (for clay). Cation-exchange
capacity was determined by saturating the sample
with sodium acetate. Mineral constituents were
identified by x-ray crystallographic analysis of
preferred-oriented aggregation specimens prepared on
ceramic plates.
The background sample (No. 20) and sample No.
24 are sandy, while samples No. 19 and 25 are loamy
Table 5.10
Benthal Sampling Points
Approximate location
#*
1
2
3
4
5
6
7
8
9
10
11
12
13
14
15
16
17
18
19
24
25
20
21
Probe
Depth, m Distance fiom shore, m measuiemen
21
2.5
2
2.5
2.5
3
4
14
9
3
7
7
5
7
6
6
7.5
6
4.5
12
20
1.5
0.2
Sherman
140
20
20
Sherman
11
10
6
9
15
9
6
9
12
6
9
9
10.5
9
7.5
9
Sherman
Number of samples
Hand- Dredge-
t collected collected4
Reservoir near 300-m station perimeter. June 2, 1969
1
1
1
Reservoir near south shore,
1
1
1
1
1
1
2
1
1
1
1
1
1
1
1
1
2
2
1
June 3, 1969
0
0
0
0
0
0
0
0
0
0
0
2
2
2
2
2
2
2
2
0
.0
0
0
0
0
0
0
0
0
0
1
1
1
1
1
Reservoir north of outlet weir, June 3, 1969
30** 0
60** 0
Sherman
10
Deerfield
0.5
0
0
1
1
Reservoir north of Harriman Station, June 4, 1969
1
River west of Charlemont,
0
2
June 4, 1969
1
1
0
* numbers refer to locations in Figures 5.1,5.2 and 5.3.
+ samples # 1-3 and 15-20 were collected with an Ekman dredge; # 24 and 25 were collected with
' a Petersen dredge.
distances are from Yankee circulating water outlet.along centerline.
*We thank Profs. Clyde R. Stahnke and Larry Wilding, Agronomy Dept., Ohio State University, for
performing these analyses.
63
-------
and have much higher fractions of silt, clay, and
organic material. Samples No. 1 to 3 and No. 15 to
1.8 also appear to be loamy, while sample No. 21 is
sandy. The loamy samples showed a cation-exchange
capacity of approximately 16 milliequivalents per
100 gram (meq/100 g) and the sandy samples, 2
meq/100 g. The two samples that were examined in
detail consisted mostly of illite, with some
vermiculite, quartz, and kaolinite. Stahnke's estimate
of the contribution of the various components to the
total cation-exchange capacity of sample No. 19,
hand, top, is in agreement with the measured value of
16.8meq/100g:
Component
organic matter
allophane (amor-
phous constituents)
illite
vermiculite
Approximate
capacity,
meq/100 g,
200
150
40
150
Fraction of
total sample
0.050
0.44 x 0.071 x 0.95
0.39x0.071x0.95
0.08 x 0.071 x 0.95
The listed cation-exchange capacities are commonly
used values. The contributions to the total capacity
are the products of the individual capacities and the
sample fraction, obtained from Table 5.11. The
contribution by the fractions of quartz and kaolinite
in the analyzed samples was considered negligible.
5.6.3 Analysis. The spectra obtained with the
probe were plotted as shown in Figure 5.7, and the
gross count rates of 13?Cs and 60Co were obtained at
the energy ranges of 0.63 - 0.71 MeV and 1.10-1.40
MeV, respectively. The 60co count rate of each
spectrum was the difference between the gross
Contribution to
total capacity,
meq/100 g
10
4.5
1.1
0.8
16.4
Texture
Table 5.11
Mineralogical Analysis of Ben thai Samples*
Number :
Collected by :
Core fraction :
19
hand
top
19
hand
bottom
20
dredge
24
dredge
25
dredge
loam sandy loam
sand
sand
Organic material, % of total dried weight
silt loam
Organic carbon
Organic matter
2.92 2.88
5.02 4.95
0.23 0.46
0.40 0.79
2.92
5.02
Particle size distribution, % of total mineral weight
Clay «2 /tdia.)
Silt (2-50 /udia.)
Sand (50-2,000 n&a.)
7.1 6.9
48.4 43.4
44.5 49.7
Cation exchange capacity,
Total
Clay & organic material"1"
Clay**
Illite (mica)
Vermiculite
Quartz
Chlorite
Kaolinite
16.8 14.8
143 129
94 71
Clay mineral,
55(39)
10 ( 8)
25(6)
<5«3)
10(3)
2.0 1.2
1.1 4.1
.96.9 94.7
meq/100 g of individual fraction
1.83 2.44
77 123
51 72
% of total clay ++
—
—
—
... —
— —
6.4
52.9
40.7
16.4
148
99
60(41)
15 ( 8)
2,0 ( 6)
<5 «3)
5( 3)
* By C.R. Stahnke and L. Wilding, Agronomy Dept., Ohio State University.
+ The cation-exchange capacity was assumed to be entirely due to clay (including allophane) and organic
material.
** A cation-exchange capacity of 200 meq/100 g was assumed for organic material.
++ Values in parentheses are percentages of total clay that are crystalline; the amorphous constituents were
removed with 0.5 M NaOH.
64
-------
10.C
1,000
j 100
10
«J CO
ec o
co r-
CM i—
#13
#20
I
I
0.4
0.8 1.2
ENERGY, MeV
1.6
2.0
Figure 5.7. Gamma-ray Spectra of Bottom of
Sherman Reservoir.
detector: 10 X 10-cm. Nal(Tl) Probe
location: see Table 5.10
counted : June 3. 1969.
reading in this energy range and the value at the
background location (No. 20). The background
location showed gamma rays of naturally occurring
40jC, 226Ra plus progeny, and 232jh plus progeny,
and also 13?Cs from fallout. To calculate the count
rate of 137cs in each spectrum, (1) the background
(No. 20) spectrum was subtracted, (2) the low count
rate attributed to 13?Cs at the background location
(50 c/m) was added, and (3) the Compton continuum
attributed to 60co on the basis of the net count rate
of 60co and a typical 60Co spectrum was subtracted.
At locations No. 1 and 3, the count rates in the
energy region of the 13?Cs gamma ray were actually
lower than in sample No. 20.
The benthal samples were either placed directly
into sample containers or were first separated in the
field with a U. S. No. 10 sieve (2-mm-dia. mesh). At
the laboratory, the samples were air dried and
thoroughly mixed. The samples were analyzed
gamma-spectrometrically with a Nal(Tl) detector and
200-channel analyzer as shown in Figure 5.8. Three
of the samples were separated with a standard No.
270 sieve (53-/x-dia. mesh) into sand and silt plus clay
fractions. They were further separated into silt and
clay fractions by extracting the clay into water as a
1.000 C
0.4
2.4 2.8
0.8 1.2 1.6 2.0
ENERGY, MeV
Figure 5.8. Gamma-ray - Spectra of Benthal
Samples from Sherman Reservoir.
detector: 10 X 10-cm.NaI(Tl)
#2. 89 g$ #19. 101 g; #20,
137 g. i
#2, June 26, i959 f200 min)
#19. Aug. ±k, 1969 (100 min)
#20. Nov. 12, 1969 (1000 min).
samples
counts
65
-------
suspension in 11 successive extractions. Sodium
carbonate was added as flocculating agent and the pH
was adjusted to a value of 9. (35) These separated
samples were also analyzed by gamma-ray
spectrometer.
Some of the samples were analyzed in duplicate
for 90sr content by leaching strontium from 10-g
portions of the ben thai material, separating it
chemically, and counting radiostrontium and 90y
with a low-background G-M counter. (36) Sample No.
17 was analyzed radiochemically for antimony to
confirm the 125sb results obtained by gamma-ray
spectrometry. Several of the samples were also
analyzed with a Ge(Li) detector and 1600-channel
spectrometer to identify the photon-emitting
radionuclides through precise measurement of
characteristic gamma-ray energies (+ 1 keV). This was
especially necessary for 54Mn and 125sb, whose
Table 5.12
Concentration of Radionuclides in Benthal Samples from Sherman Reservoir and Deerfield River
Sample
Weight/volume of
Concentration,
pCi/g dried weight
# Collection analyzed sample, g/cc 60co 137^ 'Ogr ^^Mn l^Sb
1
2
3
15
16
17
18
19
20
21
24
25
Notes:
1.
hand, top
hand, lower
dredge, s
dredge, u
hand, top
hand, lower
dredge, s
dredge, u
hand
dredge, s
dredge, s
dredge, p
hand, top
hand, lower
dredge, u
hand, top
hand, lower
dredge, u
hand, top
hand, lower
dredge, u
hand, top
hand, lower
dredge, u
hand, top
hand, lower
dredge, u
hand, top
hand, lower
dredge, u
hand
dredge, s
dredge, s
46/100
85/100
350/400
52/100
89/100
71/100
300/400
265/400
640/400
144/100
122/100
142/100
416/400
107/100
118/100
493/400
428/400
500/400
396/400
412/400
417/400
440/400
500/400
384/400
101/100
103/100
505/400
593/400
600/400
560/400
600/400
674/400
350/400
1.9
<0.1
1.2
0.8
0.6
0.2
1.0
1.0
<0.1
0.3
0.5
<0.1
6.0
6.0
4.6
0.9
0.5
7.0
4.2
1.5
10.6
0.5
0.1
32.0
20.2
18.6
9.1
<0.1
<0.1
<0.1
<0.1
1.8
1.6
3.3
0.7
4.5
3.4
3.6
1.8
5.0
4.3
0.2
1.2
0.8
0.3
3.4
4.9
3.2
1.4
1.4
4.6
4.3
3.4
5.7
1.4
1.0
5.2
6.4
6.1
3.7
0.4
0.8
0.5
0.3
2.5
4.2
—
...
0.6
—
—
—
0.4
—
0.1
...
—
—
...
—
0.2
—
—
...
...
...
0.2
—
—
0.2
0.1
—
...
...
—
0.1
0.1
0.1
0.4
0.7
<0.1
0.3
<0.1
<0.1
< 0.1
0.3
0.2
< 0.1
0.2
0.2
<0.1
0.4
0.5
0.3
< 0.1
< 0.1
0.4
0.3
0.1
0.5
< 0.1
<0.1
2.0
1.5
0.9
0.8
< 0.1
< 0.1
< 0.1
•C 0.1
0.4
0.2
Sample collection definitions:
hand =
top =
lower =
10-cm-dia core collected by hand
0 cm to 6.7 cm
6.7 cm to 13.4
from surface
cm from surface
u
s
p
=
=
=
< 0.1
< 0.1
0.4
<0.1
0.4
< 0.1
0.3
0.2
0.2
0.3
0.3
<0.1
0.5
0.9
0.6
0.3
0.2
0.7
0.6
0.6
0.7
0.2
0.2
0.5
0.8
0.9
0.5
<0.1
0.2
<0.1
0.2
0.4
0.6
unscreened
40K
15
19
16
12
18
18
19
17
15
16
16
18
21
20
20
21
22
20
20
22
23
23
25
18
21
19
18
10
14
14
11
14
22
passed-through US
retained on
226Ra
0.9
1.4
0.6
0.6
0.9
1.2
1.0
0.9
0.5
0.9
0.8
1.1
0.9
1.2
1.0
0.7
0.9
1.0
0.7
0.7
0.9
0.8
0.8
0.7
0.8
1.5
1.0
0.5
0.4
0.5
0.6
0.5
0.9
sieve #10
232™
0.8
0.8
1.2
0.7
0.8
0.6
0.9
0.9
0.6
0.7
0.4
0.8
0.9
0.7
0.7
0.7
0.6
0.8
0.8
0.9
1.4
0.8
0.8
1.0
0.9
0.9
0.9
0.5
0.4
0.4
0.4
0.6
1.0
US sieve #10
2.2o- values are approximately + 0.1 pCi/g for 54Mn, 60co, 125Sb, and 13?Cs; ± 0.02 pCi/g for 90Sr; + 0.3 pCi/g
for 226Ra and 232xh; and + 1 pCi/g for 40jC; < values are 3 a counting error.
3.
/iCi/g.
00
-------
gamma rays could not be as clearly identified by the
Nal(Tl) spectrometer as those of 6QCo and 13?Cs.
The concentrations of photon-emitting
radionuclides were computed from count rates
accumulated in 100- and 300-min periods. Calibration
curves had been established with 100- and 400-cc
solutions of standardized radionuclides at specific
gravity 1.00. At higher specific gravity (1.25 - 1.75),
the results were multiplied by the factor 1.1 to
correct for the observed lower counting efficiencies.
The 226Ra and 232xh values were computed on the
assumption that radioactive progeny were in
equilibrium.
5.6.4 Results and discussion of sample analyses.
The 90$r and gamma-ray spectral results summarized
in Table 5.12 show 60Co and 13?cs attributable to
Yankee operations at every sample location in the
southern end of Sherman Reservoir. The background
sample from the north end of Sherman Reservoir
(No. 20) contains 137cs attributed to fallout at
concentrations of 0.4 to 0.8 pCi/g, but no 60Co (<
0.1 pCi/g). The radionuclide content of sample No.
21 collected in the Deerfield River at DRM 27 is
similar to that of the background sample. The highest
concentrations of 60co and 137cs are in samples No.
18 and 19, in the small bay east of the pump house.
The highest concentration of radionuclides was found
at the same location by MDPH in 1965.
The samples collected at the south end of Sherman
Reservoir also contained relatively low concentrations
of 54Mn) 90sr, and 125sb. All three radionuclides
occur in fallout, but the background values in sample
No. 20 suggest that they are from Yankee if their
concentrations are considerably larger than 0.1 pCi/g.
Sample No. 20 provides an appropriate background
only for samples No. 21 and 24, however, because
these three are set apart by their relatively sandy
nature, as reflected in their high specific gravity and
low concentrations of naturally occurring
radionuclides. For all other samples, No.
1-hand-lower (see Table 5.12) may serve as
background: its radionuclide concentrations are
lowest among these samples, and similar to the
concentrations in sample No. 20. The concentration
of 90sr from fallout on land ranged from 0.1 to 1.5
pCi/g soil, and that of 54Mn was approximately 0.1
pCi/g soil (see Section 6.4).
Radionuclide concentrations were generally
highest in the dredge samples, intermediate in
hand-top samples, and lowest in hand-lower samples.
The differences are in most cases not large, and are
reversed in a few samples. The largest differences
between the hand-top and hand-lower concentrations
occur at locations No. 1 and 2, where relatively little
radioactivity was found in the lower sample. The
largest differences between dredge- and
hand-collected samples are at locations No. 16, 17
and 18, where the dredged samples contained
approximately an order of magnitude more ^Co.
These values suggest that radionuclides attributable to
Yankee are dispersed! throughout the bottom
deposits at the south end of Sherman Reservoir, even
below the depth of 6.7 cm, but that concentrations
are highest near the surface.
Collection by hand appears preferable in view of
the better sample definition as to location and depth
than for a dredged sample. In Sherman Reservoir,
however, the dredged samples provided the most
sensitive indication of radioactivity on the bottom,
possibly because they contained mostly the surface of
the sediment.
The benthal samples (see Section 5.2.1) collected
by the Sanitary Engineering Division, MDPH, on
November 2, 1965, were taken at the following
locations:(2)
(1) middle of the reservoir, 100 m S of the
Vermont state line;
(2) 50 m from the east shore, 300 m S of the
Vermont state line;
(3) 50 m from the west shore and 500 m upstream
from the dam;
(4) 50 m from the east shore and 500 m northeast
of the condenser coolant discharge;
(5) 10 m from the south shore and 100 m east of
the dam.
The last of these locations is in the same general area
as locations No. 7 and No. 11-19 in the present study.
The MDPH samples were dried, and analyzed by
Nal(Tl) gamma-ray spectrometry with an 11-isotope
matrix at NERHL.
Results were as follows:(2)
MDPH Benttul Simples of Nov. 2,1965, pCi/g
«OCo
137c,
"Ml.
«SZn
106Ru
125sb
<34&
144ce
«<*
226R,
MJl*
Station 15
2.8
6.5
1.9
ND
6.4
2.0
OJ
7.9
17
0.9
5.1
Station f 1-4
ND»- 0.1
ND • IS
ND • 0.4
ND - 0.05
0.2 - 2.4
0.07- 0.9
ND - 0.1
0.7 - 3.7
14. -20.
05 - 0*
i-j - 5.4
Possible orijn
Yankee
Mont » Yankee
faBout + Yankee
not significant
falout* Yankee
fallout •(• Yankee
Yankee
faflout + Yankee
natunl
natural
lutural
»ND: not detected
67
-------
Compared to the measurements near MDPH
station No. 5, values at location No. 15 in this study
are similar for 60co, 13?cs, 40fc, and 226Ra, and
somewhat lower for 54Mn, 125sb, and 232xh, The
radionuclides 106RU and 144ce were detected in
1965, but not in this study, possibly because these
radionuclides decayed in the 3.5-year interval
between measurements and were not replaced. The
radionuclides 65Zn and 134cs were very low or
undetectable in 1965, and undetectable in 1969.
Radionuclides measured in 1965 at concentrations
above 0.1 pCi/g at MDPH stations No. 1 to 4 are all
attributable to fallout or naturally occurring
radioactivity, while values of 0.1 pCi/g or less are
highly uncertain.
The benthal samples collected by Yankee staff for
analysis by their contractor for environmental
surveillance at 3 locations in Sherman Reservoir, 2
locations in No. 5 Reservoir just below Sherman
Dam, and 9 downstream locations in the Deerfield
River between Mohawk Park and Red Mill Dam, all
contained the following radionuclides: 0)
Yankee Benthal Samples of Dec. 14-15,1967, pCi/g
40K
gross alpha
Sherman Reservoir
and #5 Reservoir
0.5 - 1.4
0.4 - 1.5
1.7 - 3.4
6 - 15
3.5- 4.2
Deerfield River
0.2 - 0.6
0.1- 0.4
0.4 - 1.6
5 - 11
0.8 - 5.2
No 234u, 235u, 238u, or 239pu was detected. The
maximum concentrations of 54\in, 60co, and 13?Cs
were in one sample from No. 5 Reservoir.
Concentrations in the samples from the Deerfield
River were very low, but showed a general downward
trend with increased distance from Yankee.
These concentrations of 54\in, 60Co, and 137rjs
in Sherman Reservoir are within the range of values
measured in this study. The data by Yankee's
contractor show that the three radionuclides
discharged at Yankee were also deposited in No. 5
Reservoir, and possibly farther downstream in the
Deerfield River.
5.6.5 Distribution of radionuclides in benthal
samples as function of particle size. Three samples
were separated into sand, silt plus clay, and clay
fractions to observe the distribution of radionuclides
among particle-size ranges (see Table 5.13). It had
been noted in a tracer study with 65zn that the
radionuclide concentration is relatively high in the
fine fraction ( < 43-/n dia.), and that radionuclide
contents in a variety of samples are more readily
comparable if the fine fraction is analyzed in each.
(37)
The smaller particles generally, but not invariably,
contained higher concentrations of the deposited
fission and activation products. This trend suggests
that samples containing a relatively high fraction of
silt plus clay should be selected for more sensitive
detection of radionuclides in sediment. For accurate
background subtraction, the background samples
must contain a similar particle-size distribution as the
samples of interest. Radionuclide analysis of the clay
fraction did not improve the analytical sensitivity,
however, because only a small amount of clay was
separated. Separation of silt plus clay from sand in
wet media rather than after drying, as in this study,
was recommended to fractionate the radionuclides
more accurately .(38)
5.6.6 Results and discussion of probe
measurements. The probe measurements identified
13?Cs at the background location (No. 20) at the
level of 50 c/m, but did not detect any 60Co,
according to Table 5.14. In the southern end of
Sherman Reservoir, 13?Cs at count rates distinctly
above background was found at all locations except
No. 1, 3, 4, 8, and 12, and 60Co, at all locations
except No. 3. The highest count rates, at locations
No. 7, 11, and 13-19, coincided with the highest
concentrations of 60co and 13?Cs in benthal samples
(see samples No. 15-19 in Table 5.12). The probe
data from this area indicate that count rates decrease
toward the east, at locations No. 16 and 12, but do
not clearly delineate the distribution of the
radioactivity. Duplicate measurements at location No.
10 are consistent, but the three different values at
locations No. 7,11, and 15 (which were intended to
be at the same spot) suggest the difficulty of locating
exactly the same spot (note also the differences in
recorded depth in Table 5.10).
The counting efficiencies of the probe for 60Co
and 137cs> given in Table 5.15 in terms of the ratio
of probe count rate to benthal concentration, varied
considerably among locations. This would be
expected from an uneven vertical and horizontal
distribution of radionuclides in the sediment. The
averages in Table 5.15 may indicate the magnitude of
the ratios of count rates to radionuclide
concentrations.
The sensitivity of the probe is shown by the low
and "less-than" readings in Table 5.14 as compared to
68
-------
Table 5.13
Radionuclide Distribution in Dredged Benthal Samples as a Function of Particle Size,
pCi/gdry weight
Radionuclide
60co
-I37Cs
90Sr
54Mn
125Sb
,#19 dredged
Sand
2.6(12)
4.5(52)
NM
0.2(15)
0.5(37)
4<>K 18. (41)
226Ra
232Th
Fraction by wt.,
separated**
analysis!
Particle diameter
mm >
0.8(35)
0.8(35)
0.45
0.44
0.053
Silt & Clay
* 14.9(88)
3.4(48)
NM
0.9(85) <
0.7(63) ,
21. (59) «
1.2(65) <
1.2(65) <
0.55
0.56
<0.053
Clay
40(3)
10(2)
NM
C 3
^ 3
C 13
C 13
C 13
0.0072
0.07
< 0.002
Sand
0.8(50)
2.1(86)
NM
0.2(65)
0.3(87)
13. (93)
0.4(89)
0.4(82)
0.95
0.95
> 0.053
#24
SOt & Clay
16.4(50)
7.2(14)
NM
2.3(35)
0.9(13)
20. (7)
1.0(11)
1.8(18)
0.047
0.053
< 0.053
Clay
27(2)
11(0.5)
NM
3(1)
< 3
<10
<10
<10
Sand
0.7(20)
3.4(31)
NM
0.2(30)
0.6(34)
20. (29)i
0.6C22)
0.6(20)
0.0010 0.30
0.012 0.41
< 0.002
>0.053
#25
Silt & Clay
1.3(80)
3.2(69)
0.37
0.2(70)
0.5(66)
21. (71)
0.9(78)
1.0(80)
0.70
O.S9
< 0.053
Clay
4.5(6)
14. (6)
NM
< 2
< 2
< 9
and ^ 'Cs = 1 pCi/g, each
total amount = 6 x 108 xlOxl.lxl-7x 10' pCi = 7 mCi, each
69
-------
Table 5.14
Net Count Rate of 60o> and 137& with NalfTl) Underwater Probe
in Sherman Reservoir
Location
Counting
tune, nun
count/min
count/min
at 300-m perimeter
# 1
2
3
within 300-m perimeter
4
5
6
7
8
9
10
11
12
13
14
15
16
17
18
19
20
background
20
10
10
4
4
4
4
4
4
8
4
4
4
4
4
10
10
10
10
10
20
40 ±10
60 ±20
<20
90 + 30
170+30
220130
1,650+50
40130
70+30
490±30
530 + 40
2,130 ±50
630 + 40
2,220+50
1,940+50
1,960130
780 + 30
2,270+30
1,290+30
4,860+50
<20
20 ±10
170 + 10
<20
60+20
140 + 30
160 ± 30
230 + 40
70+20
90 + 30
110+20
110+30
200 + 40
40130
320 + 50
180 + 40
370+30
120 + 20
300 + 30
130 + 20
300 + 40
50110
Notes: 1. See Figure 5.3 for location of #6-19, Figure 5.2 for #lr5", and Figure 5.1 for #20.
2. + values are 2 a counting error; < values are 3
-------
In comparison, discharges during the 10 years of
operation at the annual 60co and 13?Cs releases
estimated in Section 4.3.5 would be approximately
20 mCi 60Co (of which 9 mCi would have decayed)
and 2 mCi 13?Cs. Both sets of estimates are highly
uncertain, but suggest that a considerable portion of
the discharged 60co and 137'Cs remained in benthal
material.
The radionuclide concentrations in the sediment
are too low to result in any detectable direct
radiation exposure to humans. The possibility of
radionuclides in benthal material entering the food
chain through uptake by fish, however, has been
suggested. (4) Although,, at the observed
concentrations, tJie uptake by fish would be
expected to be very low, this potential exposure
pathway should be evaluated periodically by
comparing radionuclide levels in benthal material and
fish.
5.7 Reference*
1. "Annual Report, Jan. 1, 1967 - Dec. 31,1967,
for the Environs Monitoring Program, Yankee Atomic
Electric Company, Rowe, Mass."; "1968 Annual
Report, Environs Monitoring Program, Yankee
Atomic Nuclear Power Station"; "1969 Annual
Report, Environs Monitoring Program, Yankee
Atomic Nuclear Power Station"; Isotopes, Westwood,
NJ. (1968,1969,1970).
2. Scally, N. J., "The Pollutional Effects of
Nuclear and Fossil Fuel Power Plants on the
Environment", Master's Thesis, Northeastern
University, 1968; also Simmons, W. A., Massachusetts
Department of Public Health, personal
communication (1969).
3. Riel, G. K. and R. Duffey, "Monitoring of
Radionuclides in Environmental Water", Trans. Am.
NucLSoc. 11, 52(1968).
4. Lentch, J. W., et al., "Manmade Radionuclides
in the Hudson River Estuary", in Health Physics
Aspects of Nuclear Facility Sting, P. G. Voilleque
and B. R. Baldwin, eds., B. R. Baldwin, Idaho Falls,
Idaho, 499-528 (1971).
5. Kahn, B. et al, "Radiological Surveillance
Studies at a Boiling Water Nuclear Power Reactor,"
PHS Rept. BRH/DER 70-1 (1970).
6. O'Leary, Cornelius, Massachusetts Department
of Public Health, personal communication.
7. U. S. Department of The Interior, Geological
Survey, "1966 Water Resources Data for
Massachusetts, New Hampshire, Rhode Island,
Vermont", Water Resources Div., U. S. Geological
Survey, JFK Federal Bldg., Boston, Mass. (1967).
8. Robinson, J., Yankee Atomic Electric Co.,
Hourly flow data, personal communication (1969).
9. Knox, C., U.S. Geological Survey, personal
communication (1969).
10. Bureau of Radiological Health, "Tritium in
Surface Water Network, January - June, 1969",
Radiol. Health Data Rep. 10, 513 (1969).
11. Jaske, R. T., "A Test Simulation of the
Temperatures of the Deerfield River", AEC Rept.
BNWL-628(1967).
12. Heider, Louis, Yankee Nuclear Power Station,
personal communication (1970).
13. "Gross Radioactivity in Surface Waters of the
United States, November - December 1968", Radiol.
Health Data Rep. 10, 312 (1969).
14. "Gross Radioactivity, May 1968, and 90Sr,
July 1966 - September 1967, in Surface Waters of the
United States", Radiol. Health Data Rep. 9, 660
(1968).
15. Templeton, W. L. and V. M. Brown,
"Accumulation of Strontium and Calcium by Brown
Trout from Waters in the United Kingdon", Nature
798,198 (April 13,1963).
16. Nelson, D. J. et al, "Clinch River and Related
Aquatic Studies", AEC Rept. ORNL-3697, 95-104
(1965).
17. Ophel, I. L. and J. M. Judd, "Skeletal
Distribution of Strontium and Calcium and
Strontium/Calcium Ratios in Several Species of
Fish", in Strontium Metabolism, J. Lenihan, J.
Loutit and J. Martin, eds., Academic Press, New
York, 103-109(1967).
18. Ruf, M., "Radioaktivitat in
Siisswasserfischen", Zeit. Veterinarmed. 12, 605
(1965).
19. Krumholz, L. A. and R. A. Foster,
"Accumulation and Retention of Radioactivity from
Fission Products and Other Radiomaterials by Fresh
Water Organisms", in The Effects of Atomic
Radiation on Oceanography and Fisheries, NAS-NRC
Pub. No. 551, National Academy of Science-National
Research Council, Washington, D. C., 88-95 (1957).
20. Gustafson, P. F., A. Jarvis, S. S. Brar, D. N.
Nelson and S. M. Muniak, "Investigation of !37Csin
Freshwater Ecosystems", AEC Rept. ANL-7136,
315-327(1965).
21. Gustafson, P. F., "Comments on
Radionuclides in Aquatic Ecosystems", in
Radioecological Concentration Processes, B. Aberg
71
-------
and F. P. Hungate, eds., Pergamon Press, Oxford,
853-858(1967).
22. Kolehmainen, S., E. Hasenen and J. K.
Miettinen, "137cs Levels in Fish of Different
Limnological Types of Lakes in Finland During
1963", Health Phys. 12,917 (1966).
23. Agnedal, P.O., "Calcium and Strontium in
Swedish Waters and Fish, and Accumulation of
Strontium-90", AEC Kept. AE-224 (1966).
24. Templeton, W. L. and V. M. Brown, "The
Relationship Between the Concentrations of Calcium,
Strontium and Strontium-90 in Wild Brown Trout,
Salmo Trutta L. and the Concentrations of the Stable
Elements in Some Waters of the United Kingdom,
and the Implications in Radiological Health Studies",
Int. J. Air Water Poll. 8, 49 (1964).
25. Nelson, D. J., "The Prediction of 90sr Uptake
in Fish Using Data on Specific Activities and
Biological Half Lives", in Radioecological
Concentration Processes, B. Aberg and F. P. Hungate,
eds., Pergamon Press, Oxford, 843-851 (1967).
26. Perkins, R. W. and J. M. Nielsen, "Sodium-22
and Cesium-134 in Foods, Man and Air", Nature 205,
866 (Feb. 27, 1965).
27. Jenkins, C. E., "Radionuclide Distribution in
Pacific Salmon", Health Phys. 17, 507 (1969)
28. Jaakkola, T., "55pe and Stable Iron in Some
Environmental Samples in Finland", in
Radioecological Concentration Processes, B. Aberg
and F. P. Hungate, eds., Pergamon Press, Oxford,
247-251 (1966).
29. Chapman, W. H., H. L. Fisher, and M. W. Pratt,
"Concentration Factors of Chemical Elements in
Edible Aquatic Organisms", AEC Rept. UCRL-50564
(1968).
30. National Committee on Radiation Protection
and Measurement, "Maximum Permissible Body
Burdens and Maximum Permissible Concentrations of
Radionuclides in Air and Water for Occupational
Exposure", NBS Handbook 69, U.S. Gov't. Printing
Office, Washington (1959).
31. Federal Radiation Council, "Background
Material for the Development of Radiation Protection
Standards", Report No. 2, U. S. Gov't, Printing
Office, Washington (1961).
32. Cowser, K. E. and W. S. Snyder, "Safety
Analysis of Radionuclide Release to the Clinch
River", AEC Rept. ORNL-3721, Supp. 3 (1966).
33. Essig, T. H., ed., "Evaluation of Radiological
Conditions in the Vicinity of Hanford for 1966",
AEC Rept. BNWL-439 (1967).
34. Wilding, L. P. et al., "Mineral and Elemental
Composition of Wisconsin-age Till Deposits in West
Central Ohio," in Symposium on Till, R. P.
Goldthwait, ed., Ohio State University Press,
Columbus (1971).
35. Jackson, M. L., "Soil Chemical Analysis
Advanced Course", U. of Wisconsin, Madison 1956
(unpublished).
36. Kahn, B., "Procedures for the Analysis of
Some Radionuclides Adsorbed on Soil", AEC Rept.
ORNL-1951(1955).
37. Abrahams, J. H., Jr. and R. H. Johnson, Jr.,
"Soil and Sediment Analysis: Preparation of Samples
for Environmental Radiation Surveillance", Public
Health Service Publ. 999-RH-19 (1966).
38. Johnson, R. H., Jr., Northeastern Radiological
Health Laboratory, personal communication (1970).
72
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6. Radionuclides in the Terrestrial Environment
6.1 introduction
6.1.1 Sampling. Release data by Yankee (see
Appendix B.2) and radioactivity measurements in
airborne effluents during this study (Section 3.3)
suggest that radionuclide concentrations in
ground-level air and deposition on ground and
vegetation attributable to Yankee were extremely
low. Because so little airborne radioactivity is
released, few radioactivity measurements are
performed on land by the Yankee contractor for
environmental surveillance. In 1969, they consisted
only of gross alpha and beta activity analyses in soil
from 9 locations.O) The Yankee Hazards Summary
Report also mentions the collection of airborne
particles (on filters and gummed trays) and hay, but
states that "with complete information available on
the amount of radioactivity released from the plant,
the need for an extensive post-operational survey will
be limited".(2)
The following samples and measurements were
obtained in the neighborhood of Yankee:
(1) Air was collected in 96-liter Saran-plastic
bags at four locations 300 to 500 m NE
to E of the Yankee stack to measure
radionuclides in ground-level air during
the release of gas from the gas surge
drum. Air was pumped by hand at the
rate of approximately 1 liter/min. The
collection technique was satisfactory, but
the bags leaked, hence no samples were
available for the intended analysis of
85Kr. As indicated in Section 6.2.1, the
estimated concentration of ^Ki in the
collected air was, in any case, too low for
detection. A release rate higher by an
order of magnitude than the one
described in Section 3.1.3 had been
anticipated, but was not attained because
of the limited size of the orifice at the
discharge into the stack .(3)
(2) One sample of snow was collected from
the ground on site, and a background
sample was collected at a distance of 8
km.
(3) One set of grass and soil samples was
collected at the on-site location (0.2 km
west of the stack), two were collected
just beyond the 0.3-km station perimeter
and one was collected at the background
location, 8 km distant (see locations No.
201-204 in Figures 5.1 and 5.2).
(4) Two samples of milk from cows that
grazed on a pasture 3.1 km SE of Yankee
were collected at the dairy in Rowe (see
Figure 5.1). One sample was obtained
before gas was released from the surge
drum, and the other, one day after the
gas release.
(5) Three deer that had died in accidents near
Yankee and three that had died similarly
at distant locations were compared for
radionuclide content.
(6) External radiation exposure was
measured with survey meters at the
following number of points: 10 on site at
Yankee, 5 at the 0.3-km station
perimeter, 8 in the immediate environs,
and 3 at background locations.
Calculations of expected concentrations of
radionuclides from Yankee in the environment are
presented in Section 6.2 and Appendices C.I to C.5
to demonstrate the procedure and indicate the
magnitude of radionuclide concentrations that may be
attributed to Yankee. The computed concentrations
in snow, vegetation, soil, and milk were several orders
of magnitude below detectability. Values measured in
airborne effluent (Section 3.3) were used as source
terms; meteorological data were taken from a
summary of short-term measurements on site or from
U. S. Weather Bureau data for Albany, N.Y.; and
dispersion over a flat terrain was assumed. As a
consequence, results of these calculations are gross
approximations. They are considered useful guides
73
-------
for planning environmental surveillance, however, as
long as radionuclide releases are so low that
calculated radiation doses are far below AEC limits.
Sample analyses and measurement results are
described in detail in Sections 6.3 to 6.7. The
detected radionuclides are believed to have been
deposited as fallout from atmospheric nuclear tests,
or to occur naturally. Their concentration varied so
much among samples, however, that careful sample
selection and numerous samples are needed to
determine with assurance whether any of these
radionuclides should be attributed to effluents from
the station. The external radiation exposure rate
above background was 1 to 3 microroentgen per hour
0*R/hr) at the Yankee exclusion perimeter, and was
estimated to be 0.7 jiR/hr at the nearest habitation
and 0.3 /uR/hr at the town of Monroe Bridge. This
radiation was attributed to gamma rays from stored
radioactive wastes at the station.
6.1.2 Environment of Yankee. The plant lies in the
deep narrow valley of the Deerfield River in the
Berkshire Mountains of northwest Massachusetts. The
elevation of the plant is approximately 350 m(1150
ft); within 1.5 km to the east, south, and west, the
mountains rise to elevations between 550 and 640 m
(1800 and 2100 ft). The slopes of the mountains are
wooded, and there are few open spaces or roads in
this area within a 3-km radius. Sherman Reservoir is
immediately to the north of the Yankee Plant (see
Figures 5.1 and 5.2).
Populated areas within a 3-km radius include the
town of Monroe Bridge (1.2 km SW, pop. 200 in
1960), a few houses on Main Road in Monroe
township approximately 2 km west, part of Rowe
township (4.3 km SE, pop. 230), and a few houses
above the valley in southern Vermont (the Vermont
border is 1.2 km north of the station). The Sherman
hydroelectric station is immediately west of Yankee,
Harriman hydroelectric station is 2.6 km to the
north, and the Readsboro Road carries traffic along
the west bank of Sherman Reservoir, 0.4 km to the
northwest. Towns at slightly greater distances from
Yankee include Charlemont and Florida in
Massachusetts and Readsboro and Whitingham in
Vermont. Nearby cities are North Adams and
Greenfield, Mass.; the Albany-Troy area, 60km west,
is the nearest large population center.
There appears to be no farming within 3 km of the
plant. The dairy at which milk was collected
apparently is the only one in Rowe. A herd of three
cows was seen in Whitingham township, and a few
cows were reputed to be in Monroe township. The
only edible crops from the immediate area are said to
be apples and maple syrup.(3) The only nearby
industry is a glassine paper plant in Monroe Bridge,
which uses part of the water that is retained by No. 5
Dam, just below Sherman Dam.
6.1.3 Meteorology and Climatology. An aerovane
and an anemometer mounted at the station indicate
that winds in the valley are predominantly along the
axis of the valley, but that appreciable turbulence
occurs/2) Under unstable conditions, the air within
the valley would be expected to mix with the air
above the ridges and to flow in the direction of the
wind at these higher elevations. Under stable
conditions, the air in the valley would be isolated
from the general airflow unless the airflow is along
the axis of the valley, but dispersion is expected to be
increased by the air turbulence at the plant site .(2)
Structures near the stack, at or just below stack
height, enhance dispersion.
6.2 Estimation of
ttadioaetirity
Conceit tr attorns
6.2.1 Dispersion of 85Kr in air. Calculation of the
dispersion of radioactive gas from the stack can only
be approximate in view of the complex terrain and air
turbulence near the stack. An approximate value of
the normalized dispersion at ground level on the
plume center-line, Xu/Q (in nr2), was derived from
curves of dispersion vs. distance as a function of stack
height and stability categories/4) Values were taken
from Figures 3-5B and 3-5C in this reference for the
46-m height of the Yankee stack and a distance of
300 m between stack and sampling points. Other
sample-collection information and the calculated
results are given in Appendix C.I.
The concentration of 85j(j m ajr at ground level,
X (in pCi/m^), was computed from the graphic values
of Xu/Q, values in Appendix C.I of the measured
release rate, Q (in pCi/sec), and the mean wind
speeds, u (in m/sec). The calculated concentrations (7
and 3 pCi/m^) are lower than the %$Ki background
(approximately 1 \ pCi/m3) in air.(5) Actual
concentrations from Yankee may be even lower
because of greater air turbulence than was considered
in the computation. The computed concentrations
could have been detected in several cubic meters of
74
-------
air, but not in the 96-liter volumes that were
collected.
6.2.2 Accumulation of90sr in snow. The sampling
locations had been covered with snow for several
months prior to sampling on April 1, 1969. At
Albany, rainfall totalled 0.3 cm between March 26
and 31^6) this is believed to have fallen as snow in
the mountains where Yankee is located. To calculate
the radionuclide content in the 2-cm-deep snow
samples at and near Yankee, washout in the recently
precipitated snow was added to dry deposition during
the period in which this snow had been on the
surface.
Deposition by washout, W(in pCi/m2), was
computed by the following equation, derived from
equation 5.64 of Slade:(7)
Q*nLTexp(-Lx/u)
~BTTx
(6.1)
where Q'
o
L
T
0
u
X
virtual release rate at stack, pCi/sec
washout coefficient, sec
duration of washout, sec
sector width, radians
wind speed at release height, m/sec
distance from stack to sampling point, m
Dry deposition, D (in pCi/m2), was computed by
integrating equation 5.44 of Slade(8) with respect to
the cross-wind direction and then distributing depo-
sition across the appropriate 20 sector:
vdQ'xT
D =
where:
T
e
u
h
(6.2)
deposition velocity, m/sec
depletion-corrected release rate at point
of interest, pCi/sec
total duration of deposition, sec
sector width, radians
distance to point of interest, meters
standard deviation of vertical concentration
distribution, meters
release height wind speed, m/sec
effective release height, meters
The results of the calculation are shown in
Appendix C.3. The washout of 90Sr from the
atmosphere was computed for 2 cm of snow that was
assumed to have fallen during a 34-hour period with
midpoint on March 29, 1969. It was also assumed
that the wind blew from the stack to both of the
sampling points during the entire snowfall. Dry
deposition was summed for the period March 29-31,
at the average wind frequency to the sector shown in
Appendix C.2. The source term was used without
correction for depletion. Values of vj and Lfor 90$r
were taken to be 3 x 10-3 m/sec and 1 x 10'5 sec'1,
respectively .(9)
6.2.3 Accumulation of 90& jn vegetation.
Deposition under neutral and unstable atmospheric
stability and during precipitation was computed with
equations 6.1 and 6.2 for five sampling locations (see
Appendix C.4). Deposition parameters on which
these values are based are listed in Appendix C.2.
Atmospheric stability and wind data in Appendix C.2
are from instruments at Sherman Dam (for locations
No. .201 to 203)1 and on the hillside (for location
No. 204 and the dairy farm). They were obtained in
April, May, and June of 1959.(2) The data for the
valley are probably representative despite the brevity
of the record because air flow patterns within a deep
valley are highly recurrent. Deposition was summed
for April and May, 1969, because the samples were
collected on June 3, and it was assumed that the
growing season consisted of these two months.
Rainfall data are from the April and May 1969
summaries for Albany.(6) To convert deposition per
square meter to concentration per gram ash, the
average grass density was taken to be 0.33 kg dry
weight per square meter ,(10) and the ash/dry weight
ratio, 0.07,01) for an overall ratio of 23 g ash/m2. A
half-time for radio strontium in vegetation of 14 days,
(12) taken from the mid-point of the 2-month period,
was used to correct for removal of the radionuclide
from the grass before sampling.
6.2.4 Accumulation of 90$r on soil The
deposition calculations described in Sections 6.2.2
and 6.2.3- were used to compute average annual
accumulation of 90§r in soil, as shown in Appendix
C.5. The average of annual precipitation values in
1968 and 1969 at Albany(6) was used, deposition
values were corrected for decay according to the
28-year half life of 90Sr, and the parameters affecting
deposition calculations that are given in Appendix
C.2 were applied. It was assumed that one-half of the
deposited activity remained in the soil, and that
one-half of the activity in the soil was in the
2-cm-deep soil-layer collected for analysis.
6.2.5 Iodine-131 in cows'milk. The concentration
75
-------
of 131j in milk from cows at the dairy 3.1 km from
Yankee was estimated by combining calculations of
dry deposition and transfer from grass to milk. It was
assumed that 131l was released from the gas surge
drum at concentration Q'o (in pCi/sec) for the 6.75
hour period following 1500 on June 3, 1969, that the
wind blew continuously throughout that period from
the Yankee stack toward the pasture, and that
atmospheric stability was one-third neutral and
two-thirds unstable. According to equation 6.2, the
deposition parameters in Appendix C.2, and a
deposition velocity for 131i of 1 x 10-2 m/sec,(8) the
total deposition was 1.04 x 10-4 Q'o.
The 131l concentration in milk, M, was computed
by multiplying the deposited amount of activity by
the effective daily grazing areaOO) and the ratio of
concentration in milk 1.25 days after initial ingestion
of 13Ij to the average daily intake of 13Ij. (13) The
latter was taken from curve C in Figure 14.2 of
reference 13. Thus,
M = 1.04 x KT4 Q' pCi/m2 x 45 m2/day
x 2.9 x 10'3 day /liter
= 1.4 x lO-5 Q' pCi/liter (6.3)
For Qf0 <50 pCi/sec, the minimum detectable
release rate (see Table 3.6), the computed
concentration is <7 x 10-4 pCi/liter. Even if 131]
were discharged continuously at the same rate, its
concentration in milk would be higher by only an
order of magnitude, far below detectable levels.
Concentrations of 89sr, 90sr, and 137cs in milk
would be similarly far below detectable levels because
their release rates were less than 50 pCi/sec (see
Section 3.3.5) and their transfer from cows' feed to
milk are within a.factor of five of the 131l transfer.
€.3 Radionurlid** tm SHOW
Two snow samples were collected on April 1,
1969; locations and amounts of sample are given in
Table 6.1. The snow was melted and passed through
0.45-ju-dia. membrane filters. For both samples, 7.6
liters of filtrate were evaporated to 35 ml. The
membrane filters and the 35-ml samples were
analyzed by gamma-ray spectrometry with Nal(Tl)
detectors and 200-channel analyzers. The membrane
filter from the on-site sample was also analyzed with
the Ge(Li) detector plus 1,600-channel spectrometer.
Aliquots of the four samples were then analyzed
radiochemically for radiostrontium and radiocesium.
The radionuclide content in on-site sample No.
202 is so similar to that in background sample No.
Table 6.1
Radionuclides in snow, April 1,1969
Radionuclide
3H
54Mn
60Q,
89Sr
90Sr
95Zr
95Nb
103Ru
106Ru
131i
137Q
HlCe
144Ce
#202, 0.2
Soluble
700 + 200
< 2
< 3
< 0.2
1.1
< 2
< 2
NM
< 8
< 2
2
NM
<10
Concentration.
kmW
Insoluble
—
< 1
< 1
< 0.1
0.2
4
16
3
<2
< 1
3
13
26
, pCi/liter
#204, 8 km S
Soluble
500 + 200
<2
<3
<0.2
0.8
< 2
<2
NM
<8
<2
1.3
NM
<10
Insoluble
—
< 1
<2
<0.2
0.6
8
18
NM
15
< 1
4
12
39
Notes:
1. The sample at location #202 consisted of 16.6 1 of water, and was taken from the top 2 cm
of snow in a 3-m2 area; the sample at location #204 consisted of 7.6 1 of water, and was taken
from the top 2 cm of snow in a 2.3-m2 area.
2. + values are 2 o-counting error; < values are 3cr counting error; NM - not measured.: ; .
76
-------
204 (see Table 6.1) that the entire radioactivity is
attributed to fallout from atmospheric nuclear tests.
The computed 90Sr concentrations from Yankee in
Appendix C.3 are two orders of magnitude lower
than the measured concentrations in the on-site
sample. To detect particulate radionuclides that are
released from the Yankee stack at similar rates, are
deposited to approximately the same degree as 90sr,
and moreover do not occur in fallout, it would appear
necessary to collect 500-fold larger samples.
6.4 Radiommrlidea In
Vegetation amd Soil
Four samples were collected on June 4, 1969, at
the locations shown in Table 6.2. Several kilograms of
vegetation were obtained by cutting grass and weeds
in an area of approximately 10 m?. Soil samples of
approximately 500 cc were taken from the top 2 cm
at the same locations after removing the covering
vegetation.
The vegetation samples were dried at 110°C in
cloth bags and then ashed at 50(PC. The dried
weights could not be measured because the samples
accidentally ignited during drying. The ashed samples
were analyzed gamma-spectrometrically in 400-cc
volumes with Nal(Tl) and Ge(Li) detectors plus
multichannel analyzers. Aliquots were then analyzed
radiochemically for strontium, cesium, ruthenium,
and antimony. Soil samples were dried at 1 10°C and
then analyzed with gamma-ray spectrometers, and
radiostrontium was determined in aliquots with the
leaching procedure referred to in Section 5.6.3.
The radionuclides measured in the vegetation and
soil samples are listed in Tables 6.2 and 6.3,
respectively. The radionuclides usually found in
fallout were observed. The much lower
concentrations in some of the nearby samples (No.
201 , 202, and 203) than in the background sample
(No. 204) suggest considerable variability in
deposition or accumulation of these radionuclides. To
determine if the higher concentrations of some
radionuclides in nearby vegetation-90sr (No. 201),
106Ru (No. 202), and 13?cs (No. 201)--could
possibly be attributable to Yankee would require
analysis of several samples at each location to
establish standard deviation values. The wide
differences of radionuclide concentrations
vegetation are also indicated by the 90$r and
contents of six deer rumen, which ranged from 1 1 to
51 pCi/g ash and from 9 to 86 pCi/g ash, respectively
(converted from pCi/kg wet weight in Section 6.6.2).
in
Table 6.2
Radionuclide (pCi/g ash) and Stable Ion (mg/g ash) Concentration
in Vegetation, June 4,1969
Substance
54Mn
89Sr
9<>Sr
95zt + 95Nb
106Ru
125Sb
137Cs
144Ce
calcium
strontium
potassium
silica
#201
0.3 km NE
<2
<5
64 + 3
34+1
<8
<2
13+1
18 + 2
56
0.40
550
NM
#202
0.2 km W
<2
<5
20+ 1
35 + 1
38 ±5
<2
3±1
NM
36
0.20
230
103
#203
0.4kmNW
<2
<5
19+1
52+2
<8
<2
5±1
38 + 4
33
0.30
250
NM
#204
8kmS
<2
< 5
57 + 3
36+2
<8
<2
6±1
NM
39
0.27
590
18
Notes:
1. + values are 2 a counting error; < values are 3 a counting error ;NM-not measured.
2. see locations in Figures 5.1 and 5.2.
77
-------
Table 6.3
Radionuclide (pCi/g dried) and Stable Ion (mg/g dried) Concentration
in Soil, June 4,1969
Substance
90Sr
95Zr
137Cs
l*»Ce
calcium
strontium
potassium
#201
0.3 tan NE
0.51 ±0.05
0.13 + 0.02
1.5 ±0.1
1.4 +0.5
2.0
0.042
11.4
#202
0.2 km W
0.49 + 0.05
0.66 ± 0.03
1.4 ±0.1
2.0 ±0.5
2.3
0.054
11.5
#203
0.4kmNW
0.14 ± 0.02
0.40 + 0.06
3.5 ±0.2
1.0 ±0.5
1.7
0.053
17.9
#204
8kmS
1.54 ±0.05
NM
5.4 +0.3
1.3 ±0.5
1.1
0.060
8.7
Note: see footnotes to Table 6.2.
The samples also contained naturally occurring
, U plus progeny, and Th plus progeny. No 89$r
( < 0.2 pCi/g) or photon-emitting radionuclides other
than listed (generally < 1 pCi/g) were found in soil.
Computed concentrations of 90sr in grass from
deposition of airborne particles released by Yankee
are lower than measured values by four orders of
magnitude, according to Appendix C.4. Strontium-90
concentrations in soil computed for long-term
deposition of airborne particles from Yankee
(Appendix C.5) are also lower than measured values
by four orders of magnitude.
G.5 Radionmctidem i* Milk
Two four-liter samples of raw milk were collected
at the dairy in Rowe, one at the morning milking on
June 3,1969, and the other at the evening milking on
June 4. A 3.54iter sample of milk was analyzed with
a NaI(TT) detector plus multichannel analyzer for
photon-emitting radionuclides, and a 1-liter aliquot
was analyzed radiochemically for radiostrontium*.
Analyses were by the procedure for routinely
collected milk-network samples, but the counting
periods were longer to improve precision of
measurement.
The radionuclides were at essentially the same
concentrations in both samples, as shown in Table
6.4. Average radionuclide concentrations in
pasteurized milk at nearby cities during June 1969
were:(H)
Radionuclide Albany, N. Y. Boston, Mass. Hartford, Conn.
8pCi/l 11 pCi/1 8pCi/l
<20
19
13
Table 6.4
Radionuclide Concentration in Milk, pCi/liter
Radionuclide
89Sr
90Sr
1311
137Cs
140Ba
June 3, 1969
morning
6± 2
17 ± 1
<3
48+2
<3
June 4, 1969
evening
7+ 2
13+ 1
<3
53+2
<3
Notes:
1. Gas was released from surge drum at Yankee Nuclear Power Station on June 3, 1969,
at 1500-2145.
2. Milk is from a dairy at Rowe, see Figure 5.1.
3. Analysis was by NERHL, PHS.
4. ± values are 2 a counting error; < values are 3 cr counting error.
* We thank NERHL, PHS for analyzing these samples.
78
-------
Strontium-89 concentrations between 5 and 11
pCi/liter were reported at 15 stations throughout the
country, but no 89
-------
The differences are in no case significant because of
the relatively large standard deviations, suggesting
that all of the measured radioactivity was from
fallout.
Because of the large differences in radionuclide
concentrations among samples, collection of more
samples appears necessary to determine adequate
mean values and standard deviations. This problem
was also encountered with vegetation and soil samples
(see Section 6.4), but is especially serious in animals
because their radionuclide contents are affected by so
many variables-e.g., environment, location, season,
age, food supply, and individual differences.
Even the highest 13?Cs concentration in muscle is
not unusually high compared to deer muscle collected
in areas distant from nuclear power stations/I 5,16)
Jenkins and Fendley reported numerous cases in
which levels of 13?Cs ijn the muscle of Whitetail deer
from the southeastern United States, collected during
winter and early spring, approach 150,000
pCi/kg.(15) The 13?cs levels in both rumen and
muscle in this study are higher than those in four deer
obtained in June 1969 from the vicinity of Dresden,
Illinois.O7)
The rumen contained mostly grasses and leaves. If
these samples can be considered typical, the
accumulation factor (AF), in pCi 137Cs/kg muscle
per pCi 13?Cs/kg rumen content, ranges from 0.64 to
3.23, with an average AF value of 1.5 ± 0.4 (1 (r).
These values are similar to those reported
previously .(1^-17)
The observation of 22Na jn the muscle and rumen
of deer parallels that in fish muscle (see Table 5.9).
The AF for 22fta from rumen to muscle is 0.5 ± 0.2.
The mean 90sr concentration in bone was 9,400 ±
3,100 pCi/kg, 95 + 11 pCi/g calcium, and 132 ± 40
pCi/mg strontium. The mean concentration in bone is
approximately 4 times higher than in deer collected
in Illinois/17) but is similar to concentrations
reported for deer from South Carolina,(16)
Colorado,(18) and California.(19,20) The average AF
values from diet to bone for the six deer, assuming
rumen content to be a typical diet, are as follows for
90sr, strontium and calcium:
Table 6.6
Radionuclide (pCi/kg)* and Stable Ion (g/kg)* Concentration in Deer Samples
Sample
Type
Bone
Muscle
Rumen
Content
Background samples
Nuclide
90Sr
Sr
Ca
ash wt./
wet wt.
22Na
90Sr
137Cs
K
ash wt./
wet wt.
22Na
90Sr
137Cs
K
Sr
Ca
ash wt./
wet wt.
D-l
12,500 ±340
0.090
106
0.28
7.1 + 1.1
5.3 ± 1.9
990 ± 40
3.77
0.013
10.5 + 0.7
339 ±9
920 + 40
4.23
0.0027
0.33
0.014
I>2
10,000 + 260
0.060
100
0.26
10.5 ± 0.5
9.9 ±2.1
550 ±20
3.39
0.015
30.3 + 1.3
480 ±11
170 ± 10
3.68
0.0043
1.65
0.019
D-3
3,650+140
0.069
87
0.23
3.110.6
4.4+1.4
160 + 10
3.31
0.010
8.3 ±0.7
171 +6
250 +10
4.89
0.0019
1.09
0.016
Samples from Vicinity of Yankee
D4
10,600 ± 320
0.078
117
0.31
3.4 + 0.4
4.4+1.4
1,470 +60
3.76
0.012
4.9 + 0.3
370 ± 14
1,440 ± 60
5.29
0.0030
0.44
0.017
D-5
11, 000 ±260
0.081
102
0.26
1.9 + 0.3
10.2 ±1.8
2,060 + 90
2.94
0.009
8.9 ± 0.6
760 +20
1,210 ±50
3.07
0.0029
0.53
0.014
I>6
8,600 ± 220
0.053
82
0.22
Not anal.
4.5+1.9
270 + 10
2.24
0.011
20.0+0.1
810 +20
210 ± 10
2.26
0.0086
2.06
0.016
*Kg wet weight
Note: + values are 2
-------
= 23
AFSr = 24
= 160
The average observed ratio from diet to bone for
strontium relative to calcium (ORbone/diet) is 0.20 ±
0.05 for both 90sr and stable strontium. The AF and
ORbone/diet f°r ^Sr agree with those previously
reported for deer collected in Illinois, OT) but the
ORbone/diet and AF from diet to bone are slightly
smaller than for Alaskan caribou (0.31 and 37,
respective ly).(21)
The mean 90$r concentration of 6 pCi/kg deer
meat is one-tenth of that reported for Alaskan
caribou or reindeer meat, (22) but is 6 times that in
meat taken to be a typical component of New York
City diets during Jan.-March 1 969. (23) The average
90Sr concentration in deer muscle was approximately
1/1500 of the concentration in bone, a much smaller
ratio than that reported for Alaskan caribou.(21) This
ratio undoubtedly varies because muscle reflects
recent dietary intake of 90§r more directly than does
bone. The average AF for 90$r from rumen to muscle
is 0.016; in 25 Alaskan caribou and reindeer, the AF
ranged from 0.004 to 0.21, with an average of
0.036/24,25)
Strontium-89 was not detected in rumen content,
muscle, or bone. Minimum detectable levels at the
3-sigma confidence limit were 20 pCi/kg wet weight
in rumen content and muscle, and 400 pCi/kg wet
weight in bone. No 134cs was detected in deer
muscle at the minimum detectable concentration of 2
pCi/kg. The fission products 106RU (m deer D-2 and
D-3) and 95zr (traces) were identified by their
characteristic gamma rays in rumen contents, as were
naturally occurring 40K, 226Ra plus progeny, and
232 jh plus progeny.
6.6.3 Hypothetical radiation dose from eating deer
meat. The radiation dose a person might receive from
eating deer meat was estimated from Federal
Radiation Council values, according to which 170
mrem/year is equivalent to a daily intake of 200 pCi
90Sr. (26) At the average 90§r concentration of 6
pCi/kg deer meat and an annual consumption of 79
kg meat (0.22 kg/day)(23) applied entirely to deer
meat, the radiation dose to bone marrow is 1.1
mrem/year. The average 1 37cs concentration of 920
pCi/kg muscle was 1 50 times more than that of 90§r
and the limit for 1 37cs is 1 50 times higher, hence the
radiation dose from 137cs to the whole body is also
1.1 mrem/yr. The additional radiation dose to the
whole body from 22Na at an average concentration
of 5 pCi/kg is negligible (0.002 mrem/yr). These
doses are believed to be entirely from radionuclides in
fallout, but provide an upper limit if some of the
radioactivity in deer were attributed to Yankee.
6.7 External Radiation
6.7.1 Detection instruments. Radiation exposure
rates were measured with cylindrical NaI(Tl)
gamma-ray detectors (5-cm diameter x 5-cm length)
connected to portable count-rate meters.(27) The
instruments had been calibrated by comparing their
count rates in the natural radiation background at
Cincinnati with measurements by a muscle-equivalent
ionization chamber and Shonka electr.eme.ter.(28)*
Radiation levels during calibration ranged from 5
jaR/hr over water in a lake to 19 /^.R/hr over granite.
The count rate, C (in count/min), of the survey
instruments varied linearly with the radiation
exposure rate, R (in /u.R/hr), of the ionization
chamber; a typical calibration curve had the
equation:
R = 7.0 x 10-4 C +3.3 (5.4)
Radiation exposure rates at the measurement
locations near Yankee were computed by applying
these calibration curves to the observed count rates.
Despite the dependence of the counting efficiency
of the detectors on the energy distribution of the
gamma-ray flux, the calibration curves were
applicable in a variety of natural radiation
backgrounds. In numerous measurements, the
standard error of the survey meters was ± 0.35 /u.R/hr,
and the exposure values computed from the readings
were within 4 percent of the values measured with
the ionization chamber during 95 percent of the time.
(27) Similar calibration curves could also be applied
to readings within or beneath mixtures of noble-gas
fission products that were emitted from the stack of a
boiling water reactor. (17)
6.7.2 Measurements. The 26 radiation
measurement locations listed in Table 6.7 were
selected for the following reasons:
(1) points 0 and P were considered to be
sufficiently distant from Yankee but
similar in natural radiation to yield
*We thank Richard Stoms, PHS, Cincinnati, for making the two sets of instruments available.
-------
terrestrial background values for
comparing with and subtracting from
exposure rates near the station; point M
yielded the background value over water;
(2) eight points, 0.37 to 1.1 km distant from
the center of the station, provide values
for computing potential radiation
exposure of persons in the environment;
(3) five points indicate the radiation
exposure at the 0.3-km exclusion
perimeter of the station; and
(4) ten points on site were intended to aid in
identifying the source of external
radiation from Yankee and to check
off-site exposure values by extrapolating
from these higher, more accurately
measured values.
The first and third sets of measurements were taken
while the station was operating at full power; the
second set was obtained during refueling, when the
reactor was not operating and neither short-lived nor
stored radioactive gases were being discharged.
Table 6.7
External Radiation Exposure Rate Measurements near Yankee
Location Exposure Rate,c;u.R/hr
Point*
A
B
C
D
E
F
G
H
I
J
K
L
M
N
O
P
Q
R
S
T
U
V
W
X
Y
Z
a.
Distanceb June 4, 1970
0.30 km NE 10.4 ± 0.0d
0.29 km NE
0.30 km NNE
0.40 km NNW
0.40 km NW
0.37 km NW 7.5 ± 0.1
0.39 km WNW
0.25 km NW
0.18 km W
0.14 km W 11.0 + 0.0
0.14 km W 11.5 ±0.1
0.30 km WSW
2.0 kmN
1.1 km SW
8 km S 7.5 + 0.0
17 kmSE 7.410.1
0.12 km NW
0.1 8 km NW
0.31 kmNW
0.16 km W
0.21 km W
0.30 km WSW
0.22 km WSW
0.39 km WNW
0.44 km W
1.1 kmSW
Nov. 18, 1970
12.2 ± 0.2
10.5 + 0.7
9.3 + 0.4
7.0 + 0.6
9.5 ± 0.2
9.2 ±0.1
9.3 ±0.1
9.5 + 0.2
13.4 ±0.2
14.5 ± 0.1
14.3 ±0.1
8.8 + 0.1
6.5 ± 0.2
9.2 + 0.1
8.5 ± 0.1
8.4 + 0.1
Feb. 8, 1971
5.9 ±0.1
7.1 ±0.0
7.2 + 0.0
9.4 ±0.1
8.7 + 0.1
6.0 ± 0.0
5.7 + 0.1
5.6 + 0.1
14.0 + 0.1
9.2 ± 0.0
5.7 ±0.1
12.4 + 0.0
8.5 + 0.1
6.8 ±0.1
6.9 ±0.1
6.7 ±0.0
5.7 + 0.1
6.0 ± 0.0
All points are shown in Fig. 6.1 except those more than 0.4 km distant from Yankee:
point M is in Sherman Reservoir; N,
near west end of Dam #5
(Fig. 5.1); O, at location
#204 (see Fig. 5.1); P, in East Charlemont; Y, on Readsboro Rd; and Z, near the Monroe
b.
c.
d.
Bridge school.
Distances are from the center of the
exclusion area shown in Fig. 6.1.
Measurements at points B, C, D, and M were taken 1 m above
Reservoir; all others were obtained 1
m above ground.
water level in Sherman
Exposure rates are averages of 2 to 10 measurements; ± values are 1/2 of the range for
2 measurements or 2 a values for more than 2 measurements.
82
-------
SHERMAN DAM
SCALE-METERS
SHERMAN RESERVOIR
300 METER RADIUS
PRIMARY AUXILIARY
BUILDING
BARBED WIRE
EXCLUSION FENCE
•WAREHOUSE
WASTE DISPOSAL
'BUILDING
CHAINLINK
PLANT FENCE
4
\
I
Figure 6.1. Locations of Radiation Exposure Measurements with Survey Meters.
83
-------
For these measurements, detectors were held 1 m
above the surface of the ground or water. The count
rates were between 3,400 and 16,OOOcount/min.On
land, locations over grassy terrain were selected to
minimize variations in the background; however, a
rock fill near point G and cuts in hillsides near points
K and Z may have increased the background radiation
at these locations. The snow cover to a depth of 0.3
to 1 m on Feb. 8, 1971, undoubtedly lowered
background values at all locations. The lower
background over water also shows the effect of
shielding material (i.e., the water) between the
detector and soil or rock.
To observe the effect of distance and direction
from the center of the station on the external
radiation field, the on-site measuring points were
located along several radii toward northeast (NE),
northwest (NW), west (W), and west-southwest
(WSW), as shown in Figure 6.1. In the opposite
directions steep and unpopulated hillsides adjoin the
station.
6.7.3 Results and discussion. The gross radiation
exposure rates in Table 6.7 (which include the natural
radiation background) range from 5.7 to 14.5/xR/hr
at the station and its immediate environs. The
terrestrial background radiation values at locations 0
and P agreed within 0.1 ju.R/hr, but the average value
Location
was different during each of the three measurement
periods.
All radiation exposure rates except three at or near
Yankee were above the background values obtained
on the same days. These higher values are attributed
to direct radiation from radioactive waste stored at
the station. This explanation was supported
qualitatively by (a) the general decrease in exposure
rates with distance from the station, (b) the lower
values where there was shielding by buildings or
topographical features (see Fig. 6.1), and (c) no
change in radiation exposures during reactor
shut-down or changes in wind directions. For these
reasons, the higher radiation exposure rates were not
attributed to higher natural radiation background,
deposition of radionuclides from the station on the
ground, radionuclides in the plume of airborne
radioactive effluents from the station, or direct
radiation from the Yankee reactor.
Because the net radiation exposure rates beyond
the station were so low and the radiation
backgrounds could not be measured directly at these
points, an attempt was made to check these exposure
rates by extrapolating from the higher values
measured on site. The resulting sets of values compare
as follows:
Exposure rate, //. R/hr
Point
A~
C
L
V
D
E
F
G
X
Y
N
Z
Distance
0.30 km NE
0.30 km NNE
NE perimeter
0.30 km WSW
0.30 km WSW
WSW perimeter
0.31 km NW
NW perimeter
0.40 km NNW
0.40 km NW
0.37 km NW
0.39 km WNW
0.39 km WNW
0.44 km W
Readsboro Road
1.1 km SW
1.1 km SW
Monroe Bridge
Measurements
minus
background
3.3
2.8
average 3.0 ± 0.3
0.3
1.1
average 1.0 ± 0.5 (la)
0.0
average 0.6 + 0.6
0.5
0.6
0.4
1.1
1.0
0.0
average 0.7 ± 0.3 (1 er)
0.7
0.3
average 0.3 ± 0.3 (ler)
Extrapolation
.of on-site
measurements
1.3
1.3
1.2
0.8
0.8
0.9
0.8
0.8
0.6
0.1
0.1
84
-------
Exposure rates in the first data column were
obtained by subtracting background from the gross
values in Table 6.7 and then averaging the net values
at each location. The natural radiation background
was 7.5 iiR/hr on June 4, 1970, 8.5 xtR/hr overland
and 6.5 /xR/hr over water on Nov. 18, 1970, and 5.7
/xR/hr on Feb. 8,1971.
Values in the last column were computed by the
equation:
R=0.12D-2 (6.5)
where R is the radiation exposure rate in /xR/hr and D
is the distance in km from the center of the exclusion
area. The constant in this inverse-square relation
between radiation exposure and distance from the
source was obtained from the net radiation exposures
at points Q and R (i.e., the values in Table 6.7 minus
5.7 tiR/hr). Radiation exposures along a radius
through these two points were intermediate to one
through points T, I, and G (where radiation exposures
were higher by a factor of 1.4), and through points J,
K, V, and W (where radiation exposures were lower
by 1.4). No effort was made to extrapolate in the NE
direction because directly measured exposure rates at
the exclusion perimeter were significantly above
background.
The extrapolated exposure rates are generally
consistent with the directly measured values along
Readsboro Road, where the largest number of
measurements are available for comparison, but differ
considerably at some other locations. On the whole,
the two sets of values suggest that external radiation
exposure rates were 3 /xR/hr or less at points on the
exclusion perimeter, and approximately 0.7 AtR/hr
along Readsboro Road. The average radiation
exposure rate at the town of Monroe Bridge was 0.3
/xR/hr, but the value is so low as to be very uncertain.
It would be of interest to obtain more accurate
data on the external radiation exposure in the
environs of the station, and to check on possible
errors in the presented data due to counter
calibration, background subtraction, or periodic
variations in the intensity of the radioactive waste.
Long-term exposure measurements with sensitive
detectors such as thermoluminescent dosimeters
appear feasible at locations A and C. At other off-site
locations, it would be difficult to distinguish between
radiation from the station and the natural
background by long-term measurements, but it may
be possible to do so by instantaneous measurements
at selected locations. On the other hand, it may be
simpler to reduce off-site exposure rates to natural
background levels by improved shielding of stored
wastes.
6.7.4 Estimated external radiation exposure to
persons in the environs. The average instantaneous
exposure rates listed in Section 6.7.3, multiplied by
8,760 hours per year, yield a radiation exposure value
at the nearest habitation (location E in Figure 6.1)
due to Yankee of 6 + 3 milliroentgen per year
(mR/yr), and 3 ± 3 mR/yr near the center of the
town of Monroe Bridge. These values are subject to
the uncertainties discussed in Section 6.7.3 and have
not been corrected for shielding by house walls and
time spent by persons at other locations. In
comparison, the natural radiation background
averaged 64 mR/yr, and its variation was greater than
the radiation exposure attributed to Yankee.
Exposures of travelers on Readsboro Road, fishermen
on the southern end of Sherman Reservoir, and those
who approached Yankee from the SSW or NE
directions would undoubtedly have occurred during
only a small fraction of the year, and accordingly
been relatively low. The set of measurements suggests
that distance and shielding by the terrain reduced
radiation exposure from radioactive wastes stored at
Yankee effectively to zero at distances of 2 km or
more.
1. "1969 Annual Report, Environs Monitoring
Program, Yankee Atomic Nuclear Power Station",
Isotopes, Westwood, N.J. (1970).
2. Yankee Nuclear Power Station-Yankee Atomic
Electric Co., "Technical Information and Final
Hazards Summary Report", AEC Docket No. 50-29
(1960).
3. Pike, David, Yankee Nuclear Power Station,
personal communication (1969).
4. Turner, D. B., "Workbook of Atmospheric
Dispersion Estimates", PHS Rept. 999-AP-26 (1967).
5. Sax, N.I., R.R. Reeves, and J. D. Denny,
"Surveillance for Krypton-85 in the Atmosphere",
Radiol. Health Data Rep. 10, 99 (1969).
6. "Local Climatological Data, 1969, Albany, New
York", U.S. Dept. of Commerce, U.S Gov't. Printing
Office, Washington, D.C. (1969).
7. Engelmann, R.J., "The Calculation of
Precipitation Scavenging", in "Meteorology and
Atomic Energy 1968", D.H. Slade, ed., AEC Rept.
85
-------
TID-24190, 208-221 (1968).
8. Van der Hoven, I., "Deposition of Particles and
Gases" ibid., 202-208.
9. Bryant, P.M., "Derivation of Working Limits for
Continuous Release Rates of 9Qsr and l-*7Cs to
Atmosphere in a Milk Producing Area", Health Phys.
12, 1394(1966).
10. Koranda, J. J., "Agricultural Factors Affecting
the Daily Intake of Fresh Fallout by Dairy Cows,"
AEC Kept. UCRL-12479, pp. 20 and 31a (1965).
11. Nay, U., "The Adsorption of Fallout 90sr at
the Surface of Different Grass Species", in
Radioecological Concentration Processes, B. Aberg
and F. P. Hungate, eds., Pergamon Press, Oxford,
489-491 (1967).
12. Russell, R. S., Radioactivity and Human Diet,
Pergamon Press, Oxford, 189-211 (1966).
13. Garner, R. J. and R. S. Russell, "Isotopes of
Iodine", in Radioactivity and Human Diet, R.S.
Russell, ed., Pergamon Press, Oxford, 301-303
(1966).
14. "Milk Surveillance, June 1969", Radiol.
Health Data Rep. 10, 435 (1969).
15. Jenkins, J. H. and T. T. Fendley, "The Extent
of Contamination, Detection, and Health Significance
of High Accumulations of Radioactivity in
Southeastern Game Populations", presented at the
22nd Annual Conference of the Southeastern
Association of Game and Fish Commissions,
Baltimore, Oct. 22, 1968.
16. Rabon, E. W., "Some Seasonal and
Physiological Effects on 13?cs and 89,90sr Content
of the White-Tailed Deer, OdoceUeus virginianus",
Health Phys. 15, 37(1968).
17. Kahn, B. et al, "Radiological Surveillance
Studies at a Boiling Water Nuclear Power Reactor",
PHS Rept. BRH/DER 70-1 (1970).
18. Whicker, F. W., G. C. Farris, and A. H. Dahl,
"Concentration Patterns of 90Sr, 137cs and 131j jn a
Wild Deer Population and Environment", in
Radioecological Concentration Processes, B. Aberg
and F. P. Hungate, eds., Pergamon Press, Oxford,
621-633(1967).
19. Longhurst, W. M., M. Goldman and R. J. Delia
Rosa, "Comparison of the Environmental and
Biological Factors Affecting the Accumulation of
90sr and 137Cs in Deer and Sheep", ibid., 635-648.
20. French, N. R. and H. D. Bissell, "Strontium-90
in California Mule Deer", Health Phys. 14, 489
(1968).
21. Watson, D. G., W.C. Hanson, and J. J. Davis,
"Strontium-90 in Plants and Animals of Arctic
Alaska, 1959-1961", Science 144,1005 (1964).
22. Chandler, R. P. and D. R. Snavely, "Summary
of Cesium-137 and Strontium-90 Concentrations
Reported in Certain Alaskan Populations and
Foodstuffs, 1961-1966", Radiol. Health Data Rep. 7,
675 (1966).
23. Rivera, J., "HASL Diet Studies: First and
Second Quarters 1969", in AEC Rept. HASL-214,
II-4 to 11-7(1969).
24. Schulert, A. R., "Strontium-90 in Alaska",
Science 136,146(1962).
25. "Radionuchdes in Alaskan Caribou and
Reindeer, 1963-1964", Radiol. Health Data Rep. 6,
277(1965).
26. Federal Radiation Council, "Background
Material for the Development of Radiation Protection
Standards", Report No. 2, U.S. Gov't Printing Office,
Washington, D.C. (1961).
27. Levin, S. G., R. K. Stoms, E. Kuerze, and W.
Huskisson, "Summary of Natural Environmental
Gamma Radiation Using a Calibrated Portable
Scintillation Counter", Radiol. Health Data Rep. 9,
679 (1968).
28. Kastner, J., J. Rose and F. Shonka,
"Muscle-Equivalent Environmental Radiation Meter
of Extreme Sensitivity." Science 140, 1100 (1963).
86
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7. Summary and Conclusions
7.1 Radionuclides in
Yankee Effluents
Almost the entire radioactive content of effluents
from Yankee consisted of 3H, in accord with the
station's operating reports. The other radionuclides
discharged to the environment were mostly noble gas
fission products in airborne effluents and activation
products in liquid effluents. As points of interest,
14c was found in relatively low amounts in both
gaseous and liquid wastes, and no 131l was detected
in either gaseous or airborne particulate form.
Radionuclides other than 3R were discharged only in
collected at the same points and multiplying these
averages by the volumes released annually, as
reported by the station. The annual release estimates
are presented to indicate magnitudes arid permit
comparison with other data, but are not based on
sufficient samples to be considered accurate records
of annual releases. It is expected that detailed and
continuous isotopic discharge data will, in the future,
be obtained by station operators in response to recent
AEC regulations.
The estimated amounts of radioactive gases
discharged annually through the stack from four
sources at the station were as follows:
Radionuclides in Gaseous Effluent, Ci/yr
3H
14c
4lAr
SSmjd
85Kr
87Kr
88Kr
133mXe
133Xe
135Xe
Main coolant
sampling
5xl(H
2x10-3
4x10-1
2x10-2
6xlO-»
2x10-2
3x10-2
2x10-3
1x10-1
7x10-2
*NA = not analyzed; ND =
Air ejector
4x10-3
IxlO-2
-------
container is opened for inspection, repairs, or
refueling.
Most of the gaseous radioactivity was released
while ventilating the vapor container. Gaseous
radioactive effluents from the third and fourth
sources consist mostly of longer-lived radionuclides,
while the radionuclides from the first and second
sources were relatively short-lived. In addition to the
measured radionuclides, it was estimated that
approximately 3 Ci of short-lived noble gas fission
products (89Kr, 135mXe, 137xe, and 138xe) were
released annually at the first and second sources.
Particulate radionuclides were at very low
concentrations; the total based on analyses of filters
in ventilating-air and incinerator stack effluents was
less than 1x10-3 Ci/yr. No. 131l (< 3x10-4 Ci/yr)
in gaseous or particulate form was detected.
The estimated total annual releases of 13 Ci 3H
and 4 Ci of all other radionuclides are consistent with
releases reported by Yankee for 1969 of 9.19 Ci 3H
and 4.13 Ci gross beta-gamma activity. These
amounts are considerably below the most restrictive
annual release limit of 4.5 x 103 Ci for individual
radionuclides.
The estimated amounts of radionuclides
discharged annually into effluent circulating coolant
water were as follows:
Reactor plant wastes are treated by evaporation and
then discharged in batches; secondary plant effluents
are mostly steam-generator blowdown and leakage
water, discharged directly and without delay. Some
radionuclides were also in effluent yard-drain water.
Tritium was at highest concentration in both
wastes. The most prevalent radionuclides after 3H, at
far lower concentrations, were the activation
products 14C, 51Q, 54\in, 55pe, and 58Co.
Shorter-lived radionuclides (half life < 8 days) than
those measured could also be in secondary plant
effluent; 24Na> for example, was estimated to be
released at the rate of 3 Ci/yr. In yard-drain water, if
observed concentrations and flow rates were typical,
annual releases were of the magnitude of 0.1 Ci 3H, 5
x 10-4 Ci 60co, and lower for other radionuclides.
The sum of 3H releases-800 Ci/yr-agrees with the
value of 1,048 Ci/yr in 1969 reported by Yankee, but
the sum of all other radionuclides-0.08 Ci/yr-is
higher than the reported gross beta-gamma activity of
0.019 Ci/yr. The AEC limit of 82 Ci/yr for
discharging the most hazardous radionuclides listed
above--90sr and 131i-.js many orders of magnitude
higher than the indicated releases of all radionuclides
except 3H. Tritium releases approach the limits most
closely; these are 84,000 Ci/yr according to AEC
regulations and 3,650 Ci/yr for the station according
Radionuclides in Liquid Effluent, Ci/yt
3H
14C
32P
51Cr
54Mn
55Fe
59?e
58co
60co
63Ni
90Sr
95Zr
95Nb
110mAg
124Sb
131,
137cs
Reactor plant
6xlfl2
1x10-2
8x10-5
ND
2xl(T3
1x10-2
6x10-4
4x10-4
2x10-4
ND
3 x 10'5
ND
ND
3x10-4
8x10-5
2x10-4
1x10-4
Secondary plant
2X102
1 x 10'3
ND*
2x10-2
9 x 10'3
2x10"*
4 x 10'3
1 x 10'2
3 x 10'3
1 x 10'3
6 x 10'5
4 x 10'3
3 x 10'3
SxlO"4
2 x 10'3
4 x 10"3
IxlO-4
*ND = not detected.
88
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to the Massachusetts Department of Public Health.
Thus, all radionuclides were released in quantities
well below AEC limits. These releases at Yankee of
radionuclides other than 3H are approximately two
orders of magnitude lower than in liquids and gases at
other commercially operated full-scale PWR nuclear
power stations. Tritium releases appear to be typical
of the power level at PWR stations with stainless steel
fuel cladding.
Radiommrlidett in the
Environment of
Radionuclides from Yankee were found in the
aquatic environment at low concentrations (see
Section 5), but not at all in the terrestrial
environment (Section 6). That these radionuclides
could be detected in sediment and aquatic vegetation
at Yankee, despite the relatively low radioactivity
level in its liquid effluent, suggests thai they can be
found at most other nuclear power stations.
At the point of discharge of circulating coolant
water into the Sherman Reservoir, ^H was at a
concentration of 79 pCi/ml during release of
reactor-plant liquid waste. The measured
concentration agreed with the value computed from
the measured concentration in the waste and the
4,700-fold dilution by circulating coolant water. At
the same time, 3JJ was also measured downstream
from the outfall at considerably lower
concentrations. Other radionuclides in the waste
could not be detected at the point of discharge
because their concentrations were too low.
Benthal material (sediment) from Sherman
Reservoir within approximately 200 m of the outfall
of circulating coolant water has accumulated the
following radionuclides from Yankee liquid wastes:
Radionuclides in Sediment, pCi/g dry wt.
highest concentration" background"
5.26-yr 60Co 32 <0.1
28.5 -yr 90Sr 0.6 0.1
30 -yrl37Cs 6 0.7
313 -d 54Mn 2 <0.1
2.77-yr l25Sb 0.9 0.2
probably a considerable fraction of the total release
of the two radionuclides during the 10-year life of the
station.
Water moss on rock at the outfall and dead leaves
submerged in water at the nearby shore of Sherman
Reservoir also contained radionuclides attributed to
Yankee. Radionuclide concentrations were higher by
four or five orders of magnitude than estimated
concentrations in water. The values in the one sample
of each that was collected were as follows:
Radionuclides in Aquatic Vegetation, pCi/g wet wt.
313 -d 54Mn
71.3 -d 58Co
5.26-yi 60Co
water moss
1.8
0.3
0.9
dead leaves
0.1
not detected
0.3
The sediment was estimated to contain
approximately 7 mCi each of 60Co and
In fish from Sherman Reservoir, the average
concentration in muscle ranged from 2.0 to 3.1
pCi/kg wet weight among four sampling categories,
compared to averages of 0.5 to 1.9 pCi/kg in
background samples. The difference in 22f4a
concentration may be due to waste discharges by
Yankee, although fish with higher 22^ja
concentrations than in Sherman Reservoir fish have
been found elsewhere.
No radioactivity attributed to Yankee could be
observed in suspended solids, including plankton,
from Sherman Reservoir. These samples were of
relatively small volume, however, because the water
was low in suspended solids.
No radioactivity attributed to Yankee was found
in the following terrestrial samples:
snow within the station perimeter at Yankee
vegetation and soil just beyond the Yankee
perimeter
milk from a dairy at Rowe
deer that had died accidentally within 3 km of
Yankee.
Computations based on measured effluent
concentrations and a simple model of dispersion in air
indicated that radionuclide concentrations in air and
on the ground near Yankee were so low that they
could not be detected with the available sample
volumes and analytical procedures.
External radiation measurements with survey
meters yielded an exposure rate above background of
1 to 3 /xR/hr at the 0.3-km perimeter at Yankee, 0.7
± 0.3 /i R/hr at the nearest habitation (0.4 km distant
on the west side of Sherman Reservoir), and 0.3 ± 0.3
at Monroe Bridge (1.1 km distant). The natural
89
-------
radiation background at somewhat greater distances
ranged from 5.7 to 8.5yuR/hr, depending on the time
of year. The radiation flux above background is
believed to have been gamma rays from radioactive
waste stored at Yankee (Section 6.7).
On the basis of these measurements in the
environment, the radiation exposure from Yankee to
persons living approximately 1 km distant was 3 + 3
mR/yr due to direct radiation. The natural
background radiation in the area is approximately 64
mR/yr. Radiation exposure from this source to
persons living at greater distances would be essentially
zero because of the terrain and distance. The
radiation dose from Yankee to avid fishermen and
fish eaters through ingesting fish caught at the
southern end of Sherman Reservoir was 0.3 mrem/yr
as inferred from effluent radioactivity data, and was
considerably less on the basis of direct radionuclide
analyses of fish muscle. The radiation dose from stack
effluent was estimated to be 0.4 mrem/yr at the
Yankee exclusion boundary. Thus, operation of the
Yankee nuclear power station under the observed
conditions had an extremely small impact on the
radiation dose in the environment. The direct
exposure rate was so far below the natural radiation
background that it could not be measured with
certainty, while inferred radiation doses by two other
pathways were each only a fraction of one i mrem/yr.
No other exposure pathway was observed.
7.3 Monitoring Procedures
The following techniques, in addition to those
reported earlier in the study at Dresden, were
demonstrated:
(1) measurement of radionuclides that emit
only low-energy beta particles,
specifically 14C and 63Ni;
(2) measurement of total 3n in air, as
distinguished from ^H as water vapor;
(3) measurement of radionuclides in aquatic
vegetation;
(4) use of a Nal(Tl) detector plus
multichannel analyzer as survey
instrument for detecting photon-emitting
radionuclides in the benthos, and
comparison of survey data with measured
concentrations in silt;
(5) comparison of benthal sample collection
by hand (diver) and by dredge.
(6) measurement of radiation exposure at
low levels from gamma rays emitted at
the station.
7.4 Recommendations
The fundamental recommendation for radiological
surveillance programs by nuclear power stations,
based on observations in this study and the one at the
Dresden I BWR, is that all radioactive effluents be
analyzed to obtain in detail their radionuclide
content. After the radioactive constituents have been
identified, analyses can be limited to the
radionuclides at highest abundance and of greatest
health significance. Once the pattern of radionuclide
discharges has been observed, the frequency of
analysis can also be reduced. Significant changes in
station operation or the radionuclide content of
effluents require at least a brief return to more
detailed analyses. These radionuclide discharge data
provide the basis for estimating population exposure,
planning environmental surveillance, and treating
wastes at the station. Such data will, in the future, be
available from the stations in response to recent AEC
regulations.
At a nuclear power station such as Yankee, where
few pathways for population exposure exist because
of the remote location and very low amounts of
discharged radionuclides (except 3jj in liquids), a
small-scale surveillance program will provide
sufficient information if effluent radioactivity is
rigorously monitored. The following environmental
(offsite) measurements can be suggested:
(1) external radiation exposure measured
continuously at off-site locations of
potential personal exposure;
(2) radionuclide analyses of fish caught in the
southern end of Sherman Reservoir and
immediately below Sherman Reservoir, at
times when fishermen are active.
Radiochemical analysis of edible portions
for 3n and 1 *C content and
gamma-ray spectral analysis of large
amounts of the same sample are of
particular interest.
(3) occasional analyses of other foods from
the immediate vicinity of Yankee,
including wild life, milk, fruit and
vegetables (if any), and maple syrup;
(4) occasional measurements of the
radionuclide content of benthal samples
90
-------
and aquatic vegetation for comparison
with radionuclide concentrations in fish,
to determine whether the radionuclides
deposited in the sediment enter the food
chain.
The program of environmental surveillance must be
evaluated periodically to consider modifications in
response to changes in effluent radioactivity, new
patterns of population distribution and
environmental use, and increased knowledge of the
behavior of radionuclides in the environment.
At nuclear power stations that discharge more
radioactivity than Yankee, more extensive
environmental surveillance will usually be found
desirable. In addition, studies to relate concentrations
of radionuclides along critical environmental
pathways for human radiation exposure to the release
rates of these radionuclides "will often be useful. The
radionuclide acts as tracer to quantify transfer
coefficients from station to man, providing a better
and more pertinent basis for calculating exposures at
the site than most published values. Such studies may
need to be performed only once. At Yankee, the only
radionuclide that appears feasible as such a tracer is
3H in water.
The EPA research program of which this study is a
part is being continued through field studies at the
newer and larger nuclear power stations. In total,
these field studies should indicate the degree to which
release data are generally applicable, the influence of
the environment and station size, design, and
operating practices on human radiation exposure, and
the need for studying specific environmental
pathways for radionuclides.
91
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Appendix A
Acknowledgment*
This report presents the work of the staff of the Radiochemistry and Nuclear Engineering Branch, EPA,
consisting of the following:
William J. Averett Seymour Gold B. Helen Logan
Richard L. Blanchard Betty J. Jacobs Alex Martin
William L. Brinck Bernd Kahn Eleanor R. Martin
Teresa B. Firestone Jasper W. Kearney Elbert E. Matthews
George W. Frishkorn Harry E. Kolde James B. Moore
Gerald L. Gels Herman L. Krieger David B. Smith*
Participation by the following is gratefully acknowledged:
Cornelius J. O'Leary, Massachusetts Department of Public Health
Edward Crockett, Massachusetts Department of Public Health
William Simmons, Massachusetts Department of Public Health
Colton H. Bridges, Massachusetts Bureau of Wildlife Research and Management
David Pike, Yankee Atomic Electric Company
John Connelly, Yankee Atomic Electric Company
Carroll D. Hampelmann, Division of Compliance, AEC*
Charles Phillips, Southeastern Radiological Health Laboratory, EPA
Raymond H. Johnson, Northeastern Radiological Health Laboratory, PHS
James Murphy, Northeastern Radiological Health Laboratory, PHS
Assistance by C. L. Weaver, E. D. Harward, and J. E. Martin, EPA, in planning the study is gratefully
acknowledged. We wish to thank Prof. G. Hoyt Whipple, U. of Michigan, for his valuable suggestions, especially
those that led to use of a gamma-ray probe to measure radioactivity in benthal deposits and to analysis of gas
for l^c. For reviewing this report, we thank the above, and also F. Galpin and J. Russell, EPA, J. A.
MacDonald, Yankee Atomic Electric Co., H. R. Denton, AEC, G. J. Karches, PHS, Prof. J. Leonard, U. of
Cincinnati, and Prof. C. P. Straub, U. of Minnesota.
*Affiliation at the time of this study.
93
-------
Appendix B.I
Main Coolant Data from Yankee Nuclear Power Station Monthly Operating Reports*
Month
June 1967+
July
August
September
October
November
December
January 1968
February
March
April
May
June
July
August
September
October
November
December
January 1969
February
March
April
May
June
July
August
September
October
November
December
January 1970
February
March
April
May
June
July
August
September
October
November
December
January 1971
February
Average
Power,
MWe
176
73
173
174
170
175
171
156
141
94
0
175
179
165
177
163
182
143
143
179
177
181
143
156
130
115
4
34
173
185
185
185
175
127
183
167
169
177
116
158
112
3
181
184
176
Average**
Boron
Concentration,
ppm
586
477
375
235
125
4
0
0
0
0
~2800
<1500
1103
978
853
721
585
445
317
187
61
0
9
0
0
0
2893
2417
1101
979
870
722
610
538
395
305
183
2
~330
1
0.
2760
1525
1270
1110
Average
Tritium
Concentration,
jiiCi/ml
2.92
~1.
1.36
2.40
1.56
1.30
1.82
1.67
1.42
1.22
0.08
<3.61
3.81
4.47
4.86
4.36
4.21
2.07
2.25
2.24
1.20
1.29
~0.9
0.531
0.429
0.364
0.040
2.20
3.05
3.93
3.61
2.57
1.93
1.63
1.16
1.80
1.29
0.83
1.15
0.30
n.r.++
0.03
1.37
3.12
3.92
Comments
Maintenance shutdown 7/8-7/25
Dilution for boron removal 1 1/3
Refueling shutdown 3/23-5/1
Maintenance shutdown 11/8-11/15
Dilution for boron removal 2/18
Maintenance shutdown 4/11-4/16
Primary-Secondary leakage
Primary-Secondary leakage
Refueling shutdown 8/2-9/25
Maintenance shutdown 3/21-3/29
Primary-Secondary leakage
Primary-Secondary leakage
Dilution for boron removal 7/1
Maintenance shutdown 8/21-8/31
Primary-Secondary leakage
Refueling shutdown 10/24-11/30
*Data reviewed and corrected by Yankee staff.
**Calculated weighted average from reported concentrations for given portions of the month, or as a mean where only
maximum and minimum values were reported.
+Tritium concentration in main coolant only reported sporadically before June, 1967.
++n.r. - not reported
94
-------
Appendix B.2
Radioactive Waste Discharge Data from Yankee Nuclear Power Station Monthly Operating Reports
Liquid Releases
Waste Disposal
Month
November 1966
December
January 1967
February
March
April
May
June
July
August
September
October
November
December
January 1968
February
March
April
May
June
July
August
September
October
November
December
January 1969
February
March
April
May
June
July
August
September
October
November
December
Secondary Plant
Volume, Gross/?-?, Tritium* "Volume, Gross /?-% Tritium,
105l fid Ci IflSl AiCi O
2.5
7.4
8.8
10.1
7.5
10.8
9.8
7.3
4.9
7.0
11.5
8.2
5.4
6.1
1.1
0.7
2.1
3.8
2.4
1.2
1.1
1.2
0.6
1.2
2.3
2.2
1.9
1.2
2.1
3.4
2.0
1.6
2.2
4.5
3.1
2.1
0.9
0.8
370
22,410
5,000
19,460
11,360
786
2,330
5,150
2,380
574
3,010
4,420
195
112
53
31
73
164
104
17
27
11
5
18
53
74
35
19
87
168
61
59
153
124
60
36
29
55
32
194
219
65
53
78
132
180
44
94
126
51
225
195
34
27
96
100
57
42
66
126
48
150
229
184
98
101
191
205
88
28
26
45
30
76
75
86
4.2
6.4
10.5
8.1
6.0
6.5
9.5
11.1
6.6
9.0
11.2
8.6
7.2
7.0
14.8
11.0
8.2
12.7
10.7
7.1
10.6
8.9
10.9
11.8
11.4
8.7
12.2
10.4
19.7
16.0
13.3
18.0
9.4
2.1
4.7
15.5
7.8
8.8
<5
35.9
48
52
32
10.9
99
47
28
7.5
11.6
6.9
6.3
10.2
15.1
5.0
5.1
7,640
31
4
6
5
17
4
11
3
9.9
90
478
458
6,000
4,760
1,150
450
130
466
1,440
2,900
2.5
30.5
6.7
14.7
9.9
2.3
16.4
45.9
29.5
0.3
0.2
0.1
0.1
0.2
0.4
0.3
0.3
0.1
0.2
0.3
0.9
0.9
1.2
3.4
3.8
0.7
1.7
1.4
16.6
10.5
22.0
17.2
6.3
2.8
0.7
12.8
35.3
48.1
Gaseous Releases
/?-y,mCi
495
134
196
148
902
166
150
421
90
17
78
111
7
26
14
20
33
nj.
20
14
28
99
76
84
107
168
147
158
162
358
817
445
311
24
135
523
263
790
Tritium, Ci **
0.3
n.r.+
nj.
n.r.
8.97
0.06
n.r.
ri.f.
6.Q5
n.r.
n.r.
n.r.
n.r.
n.r.
n.r.
0.10
4.09
1.90
n.t..
n.r.
0.05
n.r.
n.r.
n.r.
1.91
n.r.
n.r.
0.16
0.06
4.96
n.r.
0.13
4.5 x 10-*
2.23
1.39
1.16
0.01
2.9 x 1(K
*Monthly reports of tritium liquid waste discharge began in March, 1965.
" *Comments on gaseous releases of tritium:
Reported as "gaseous waste releases" in March, 1967, and as "gaseous releases" in February, March, April, June,
August, November, and December, 1969; January, February, March, June, August, September and October, 1970.
Reported as "a vapor, from the vapor container" in November, 1966; March and July, 1967; March, April, July, and
November, 1968; April, July, August, September, and October, 1969; March, August, October, and November, 1970.
Reported as "an inadvertent gas release" in April, 1967, and February, 1968.
A 62-mCi release during main steam line safety valve test was reported in June, 1969.
n.r. - not reported
Notes: 1. Core lifetimes:
Core VI - November 8,1966 - March 23,196 8
Core VII - May 1,1968 - August 2,1969
Core VIII - September 5,1969 - October 24,1970
2. Data reviewed and corrected by Yankee staff.
95
-------
Radioactive Waste Discharge Data from Yankee Nuclear Power Station Monthly Operating Reports (cont'd).
January 1970
February
March
April
May
June
July
August
September
October
November
December
January 1971
February
1.4
2.2
2.8
3.4
1.0
2.5
4.0
3.0
4.5
2.5
2.0
1.5
2.3
1.7
25
34
46
60
15
41
65
44
161
87
59
55
56
25
143
96
86
239
47
126
174
90
103
48
14
47
146
145
14.9
15.8
17.4
16.2
12 A
14.2
9.9
11.5
15.1
12.6
3.5
15.0
15.3
14.8
4,390
4,240
10,590
265
1,590
3,790
2,758
4,432
261
316
382
125
35
191
46.6
44.7
40.8
3.0
8.5
18.3
33.8
73.8
4.1
5.1
0.5
0.3
0.5
1.1
1,462
2,054
1,731
714
1,628
1,549
4,659
1,743
304
145
25
439
392
508
1.9 x ID"5
4.5 x 10-5
3.79
n.r.
n.r.
0.08
n.r.
3.94
0.03
0.54
0.62
n.t.
n.r.
n.r.
Appendix B.3
Estimated Generation Rate of Fission Products in Fuel
at 600 MWt Power
Product
3H
85Kr
SSnifcr
89Sr
90Sr
91Sr
95Zr
95Nb
99Mo
131i
133i
135j
133Xe
133mXe
135Xe
137Cs
140Ba
Fission
yield,Y*
9.5 x 1C-5 +
2.9 x 10-3
1.3xlO-2
4.5 x 10-2
5.9x10-2
5.8 x 10-2
6.3 x 10-2
6.3 x 10-2
6.1 x 10-2
2.9 x ID'2
6.5 x 10-2
6.0 x 10-2
6.6 x 10-2
1.6 x 10-3
6.3 x 10-2
5.9 x 10-2
6.6 x 10-2
Decay constant,
A,sec-l
1.78x10-9
2.05 x 10-9
4.37 x 10-5
1.57 x 10-?
7.82 x 10-!0
1.98xlO-5
1.23x10-7
2.29 x lO-7
2.90 x 10-6
9.96 x ID'7
9.21 x 10-6
2.87 x 10-5
1.52xlO-6
3.5 x 10-6
2.11 x 10-5
7.30 xlO'10
6.26 x 10-7
Generation rate,
H Ci/sec
8.5x101
3.0 xlO3
2.9 x 108
3.6 x 106
2.3 x 10*
5.8 x 108
3.9 x 106
8.5 x 106**
8.9 x 107
1.5x107
3.0 x 108
8.6 x 108
5.0x107
2.8 xlO6
6.7 x 108
2.2 x 104
2.1 xlfl7
Accumulation in
2 years, fid
5.9 x 109
1.8 xlO11
6.5x!0!2
2.7 xlO13
1.7 xlO12
3.4x!0!3
3.7 xlO13
8.1 xlO13
3.6x!0!3
1.7 xlO13
3.8 xlO13
3.5 \IQ13
S.SxIO13
8.0 xlO11
3.2 xlO13
1.5 xlO12
3.9 xlO13
* Harley, N., I. Fisenne, L. D. Y. Ong, and J. Harley, "Fission Yield and Fission Product Decay" in AEC Kept. HASL
164 (1965), p. 251; Russeil, I. J. and R._V. Griffith, "The Production of 109cd and H3mcd in a Space Nuclear
Explosion" in AEC Rept. HASL 142 (1964) p. 306.
+Albenesius, E. L. and R. S. Ondrejein, "Nuclear Fission Produces Tritium", Nucleonics 18 (9), 199 (1960).
**Equilibrium with longer-lived parent is assumed.
Notes:
1. Generation rate = thermal power x
fission rate
MWt
MWt
2. Accumulation = thermal power x
MWt
3.1 x 1Q16 fission/sec
MWt
fission rate
xYx X
xYx A
MCi
3.7 x 104 dis/sec
MWt
x Y x
96
-------
Appendix B.4
Estimated Turnover Rate of Ionic Fission Products in Main Coolant Water Based on
Concentration Measurements, and Ratio of Turnover Rate to Generation Rate
Fission
Product
89Sr
90$r
91$r
9Szr
95jjjj
99Mo
131,
133i
135i
137Cs
140Ba
Avg.
Notes:
1. A tttrn
A decay + A turnover,
sec-1
3.0 x 10-5
3.0 x 10-5
5.0 x 10-5
3.0 x 10-5
3.0 x 10-5
3.3 x 10-5
3.1 x 10-5
3.9 x 10-5
5.9 x 10-5
3.0 x 10-5
3.1 x 10-5
Turnover rate,
MCi/sec
1.3x10-2
3.8 x 10-*
3.1 x 10-1
8.6 x 10-2
9.6 x 10-2
2.3 x 10-1
1.0x10-1
1.6
3.4
3.8 x 10-*
2.4 x 10-2
1.0 x to'5 sec 'I
Turnover rate/
Generation rate
3.6 x 10-9
1.6 x 10-8
5.3 x 10-10
2.2 x 10-8
1.1 x 10-8
«2.6 x 10-9
6.7 x 10-9
4.0 x 10-9
4.0 x 10-9
1.7 x 10-8
1.1 x 10-9
8.2 x 10-9
54,0
2. Turnover rate = Concentration x coolant amount x ( A^g^y + A turnover)
= Concentration in ft Ci/g x 6.4 x 107 g x ( A decay + 3.0 x 10-5) sec'l
3. Concentrations from Table 2.1 for sample of Oct. 4, 1968.
4. Generation rate and A^ from Appendix B.3.
Appendix B.5
Estimated Turnover Rate of Longer-lived Ionic Activation Products in
Main Coolant Water Based on Concentration Measurements
Activation product Turnover rate,P€i/sec
32p 1.9 x 10-2
51Cr 1.6
54Mn 1.0
55pe 1.9 x 10-1
59pe 3.6 x 10"!
57co ~ 1 xlO-3
58co 6.5 x 10-1
60c0 1.7 x 10-1
3.6 x 10-2
3.8 x 10-2
181Hf —2 xlO-2
182Ta ^1 x 10-1
185y/ 1.9 x 10-2
Notes:
1. Concentrations fromTable 2.1 for sample of Oct. 4,1968.
2- turnover = 3'° x 10~5 sec'~1; A decay <5 x 10-7sec-l for aU
listed radionuclides.
3. See footnote 2 to Appendix B.4 for calculation of turnover rate.
97
-------
Appendix C.I
Test Conditions and Calculations for Sampling Yankee Stack Effluent
in Environment, June 3, 1969
Period
Hours (EOT)
Mean temperature (°F)
Mean wind speed, m/sec
Solar radiation
Stability category (Pasquill-Gifford)
Normalized concentration, m"2
8sKr Release rate (Table 3.6), pCi/sec
Computed 85 Kr concentration, pCi/m3
1
1600-1700
73
4.0
moderate
B
7 x ID'5
4.2 x 105
7
2
1700-1800
72
5.2
moderate
C
4 x 10-5
4.2 x 105
3
Appendix C.2
Radionuclide Deposition Parameters
Location
Distance, m
Azimuth, deg.
Wind azimuth, deg.
Mean wind frequency in 20° sector
unstable
neutral
Mean wind speed, m/sec
unstable
neutral
Standard deviation a, m
unstable (A)
neutral (C)
#201
260
046
226
0.060
0.045
3.6
3.6
35
16
#202
230
270
090
0.012
0.010
2.2
3.1
35
16
#203
450
336
156
0.008
0.003
2.2
1.8
90
30
#204
8000
180
0
0.012
0.019
4.4
4.0
1000^
160b
Dairy
farm
3100
135
315
0.034
0.018
4.9
4.0
iooo|>
160b
* Letters A and C refer to Pasquill-Gifford Stability Class.
b Estimated
Appendix C.3
Computed Accumulation of 90sr jn Snow during March 1969
Location
Deposition, pCi/m^
Snow
Dry, unstable
Dry, neutral
Total
Area sampled, m2
Sampled activity, pCi
Volume of melted snow, 1
Concentration3, pCi/1
2.45
0.52
0.03
3.0
3.0
9.0
16.6
5.4
4.3
#202
xlO'3
xlO-3
xlO-3
xlO'3
x 10~3<
xlO"4'
xlO'3
QO'
Qo'
Qol
Qo'
QO
n_i
7.03
0.005
0.001
7.0
16.1
2.1
1.7
#204
x 10"5QJ
x 10-5 Qy
x 10-5 Q^j
x 10-5 Q^
2.3
x 10'5 QJ
7.6
x 10-5 Q(j
xlO-4
a Value based on average Qo = 8 pCi/sec in Section 3.3.5.
98
-------
Appendix C.4
Conputed ^Sr Accumulation in Grass,
April-May 1969
Location
Deposition, pCi/m2
Dry, unstable
Dry, neutral
Precipitation, April
Precipitation, May
Total
#201
2.8 xlO-2Qo'
0.21 x 10-2 QO'
0.05 x 10-2 Qcj
0.03 x 10-2 Qo1
3.2 xlO-2Q0'
#202
1.06xlO-2Q0'
0.06 x ID'2 Qo1
0.02 x 10'2 Qo'
0.01 x 10-2 Qo'
1.2 xlO-2Q0'
#203
2.82 x ID"3 Qo1
3.81 x 10-3 Qo'
0.06 x ID'3 Qo1
0.03 x ID'3 Qo1
6.7 xlO-3Q0'
#204
0.12 x 10-4 Qo'
1.27xlO-4Q0'
0.04 x 10-4 QO,
0.02 x 10-4 Qn'
1.5 x!0-4Q0'
Dairy farm
1.78 x 10-4 Q0'
3.11 xlO-4Q0'
0.17 x 10-4 Qo'
5.1 x 10-4 Qo1
Remaining at sampling
time, a pCi/m2
In grass ash, pCi/g
6.1 xlO-3Q0' 2.3 xlO-3Q0' 1.3 x 10'3 Qo' 2.9 x 10'5 Qo'
2.7 x
2.2 xlO-3
1.0 xlO-4Q0'
8.0 x 10-4
5.7 xlO-5Q0'
4.6 x 10-4
1.3 xlO-6Q0'
1.0 x 10-5
9.7
4.2 xlO-6Q0'
3.4 x 10'5
aBased on 14-day environmental half life from mid-time of period (May 1) to collection date (June 3), the decay factor is 0.19
t>l m2 of vegetation was assumed to yield 23 g ash; Qo = 8 pCi/sec in Section 3.3.5
Note: There were 90 hr of precipitation in April and 44 hr in May.
Appendix C.S
Computed Long-term 90$r Accumulation in Soil
Location
Annual Deposition, pCi/m2
Dry, unstable
Dry, neutral
Precipitation3
Total
8-yr accumulation, ^ pCi/m2
In top 2-cm layer, c pCi/m2
Insoil,d pCi/g
#201
0.175 Qo'
0.013 Qo1
0.004 Qo'
0.19 Qo'
1.4 Qo1
0.35 Qo'
1.2 x 10-5 Qo1
9.6 x 10-5
#202
0.065 Qo'
0.004 Qo1
0.001 Qo'
0.070 Qo1
0.51 Qo1
0.13 Qo'
4.3 x 10-6 QJJ.
3.4 x 10-5
#203
0.0172 Qo
0.0089 Qo1
0.0005 Q0'
0.027 Qo1
0.19Q0'
0.048 Qo'
1.6 x 10-6 Qo'
1.3x10-5
#204
0.000067 Qo1
0.000774 Qo'
0.000037 Q0'
0.00088 Qo'
0.0064 Qo'
0.0016 Qo'
5.3 x 10-8 QJ
4.2 x 10-7
aBased on precipitation of 754 hr/yr during 1968 and 1969.
^Corrected for decay of 90Sr.
cOne-fourth of the accumulation, assuming that one-half is removed from the soil and one-half is below the top 2 cm.
dpor dry density of 1.5 g/cm3 and 2-cm sampling depth, 1 m2 surface area corresponds to 3 x Ifl4 gms soil; value
based on average Qo' = 8 pCi/sec in Section 3.3.5.
99
-------
KEY WORDS:
Nuclear
Power
Radiological
Surveillance
Radionuclide
Analysis
Radiation
Exposure
Reactor
Effluents
RADIOLOGICAL SURVEILLANCE STUDIES AT A PRESSURIZED WATER NUCLEAR
POWER REACTOR. B. Kahn, R.L. Blanchard, H.E. Kolde, H. L. Krieger, S. Gold, W.L. Brinck,
W.J. Averett, D.B. Smith, and A. Martin; Aug. 1971; RD 71-1; ENVIRONMENTAL PROTECTION
AGENCY.
A radiological surveillance study was undertaken at the Yankee Nuclear Power Station to make
available information for calculating population radiation exposures at routinely operating
commercial PWR stations and to demonstrate effective monitoring procedures. Radionuclide
concentrations and external radiation were measured in the immediate environment of the station.
At the same time, the radionuclide contents of liquids and gases at the station and of effluents at
points of discharge were measured, and levels of environmental radioactivity were estimated from
these values.
The radioactivity in effluents at Yankee consisted mostly of 3H, in amounts typical of PWR
stations that use fuel clad in stainless steel. The amounts of other radionuclides discharged to the
environment from the reactor plant were very small, apparently because of effective containment
of fission products other than 3H within the fuel elements and treatment of wastes by storage (for
radioactive decay) and evaporation. A considerable fraction of the effluent radioactivity was
discharged at the secondary coolant system because these effluents are released without treatment.
In the environment, radionuclides from Yankee were found only in the aquatic environment, at
low concentrations. The detected radionuclides do not appear to constitute significant direct
radiation exposure to the population; and radiation doses inferred from radionuclide measurements
in liquid and gaseous wastes were less than 1 mrem/year through all pathways that were
considered. Measurements of external radiation exposure in the environment suggested that a small
increment above the natural radiation background was due to gamma rays emitted by wastes stored
at Yankee.
RADIOLOGICAL SURVEILLANCE STUDIES AT A PRESSURIZED WATER NUCLEAR
POWER REACTOR. B. Kahn, R.L. Blanchard, H.E. Kolde, H. L. Krieger, S. Gold, W.L. Brinck,
W.J. Averett, D.B. Smith, and A. Martin; Aug. 1971; RD 71 -1; ENVIRONMENTAL PROTECTION
AGENCY.
A radiological surveillance study was undertaken at the Yankee Nuclear Power Station to make
available information for calculating population radiation exposures at routinely operating
commercial PWR stations and to demonstrate effective monitoring procedures. Radionuclide
concentrations and external radiation were measured in the immediate environment of the station.
At the same time, the radionuclide contents of liquids and gases at the station and of effluents at
points of discharge were measured, and levels of environmental radioactivity were estimated from
these values.
The radioactivity in effluents at Yankee consisted mostly of 3H, in amounts typical of PWR
stations that use fuel clad in stainless steel. The amounts of other radionuclides discharged to the
environment from the reactor plant were very small, apparently because of effective containment
of fission products other than 3H within the fuel elements and treatment of wastes by storage (for
radioactive decay) and evaporation. A considerable fraction of the effluent radioactivity was
discharged at the secondary coolant system because these effluents are released without treatment.
In the environment, radionuclides from Yankee were found only in the aquatic environment, at
low concentrations. The detected radionuclides do not appear to constitute significant direct
radiation exposure to the population; and radiation doses inferred from radionuclide measurements
in liquid and gaseous wastes were less than 1 mrem/year through all pathways that were
considered. Measurements of external radiation exposure in the environment suggested that a small
increment above the natural radiation background was due to gamma rays emitted by wastes stored
at Yankee.
RADIOLOGICAL SURVEILLANCE STUDIES AT A PRESSURIZED WATER NUCLEAR
POWER REACTOR. B. Kahn, R.L. Blanchard, H.E. Kolde, H. L. Krieger, S. Gold, W.L. Brinck,
W.J. Averett, D.B. Smith, and A. Martin; Aug. 1971; RD 71-1; ENVIRONMENTAL PROTECTION
AGENCY.
A radiological surveillance study was undertaken at the Yankee Nuclear Power Station to make KEY WORDS:
available information for calculating population radiation exposures at routinely operating
commercial PWR stations and to demonstrate effective monitoring procedures. Radionuclide Nuclear
concentrations and external radiation were measured in the immediate environment of the station. Power
At the same time, the radionuclide contents of liquids and gases at the station and of effluents at
points of discharge were measured, and levels of environmental radioactivity were estimated from Radiological
these values. Surveillance
The radioactivity in effluents at Yankee consisted mostly of 3H, in amounts typical of PWR
stations that use fuel clad in stainless steel. The amounts of other radionuclides discharged to the Radionuclide
environment from the reactor plant were very small, apparently because of effective containment Analysis
of fission products other than 3H within the fuel elements and treatment of wastes by storage (for
radioactive decay) and evaporation. A considerable fraction of the effluent radioactivity was Radiation
discharged at the secondary coolant system because these effluents are released without treatment. Exposure
In the environment, radionuclides from Yankee were found only in the aquatic environment, at
low concentrations. The detected radionuclides do not appear to constitute significant direct Reactor
radiation exposure to the population; and radiation doses inferred from radionuclide measurements Effluents
in liquid and gaseous wastes were less than 1 mrem/year through all pathways that were
considered. Measurements of external radiation exposure in the environment suggested that a small
increment above the natural radiation background was due to gamma rays emitted by wastes stored
at Yankee.
KEYWORDS:
Nuclear
Power
Radiological
Surveillance
Radionuclide
Analysis
Radiation
Exposure
Reactor
Effluents
•&TJ.S. GOVERNMENT PHINTING OFFEE: 1973—757-561/3302 5-H
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