EPA-520/3-74-007
RADIOLOGICAL SURVEILLANCE
STUDY AT THE HADDAM NECK
PWR NUCLEAR POWER STATION
OFFICE OF RADIATION PROGRAMS
U.S. ENVIRONMENTAL PROTECTION AGENCY
NATIONAL ENVIRONMENTAL RESEARCH CENTER
CINCINNATI, OHIO 45268
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EPA-520/3-74-007
RADIOLOGICAL SURVEILLANCE
STUDY AT THE HADDAM NECK
PWR NUCLEAR POWER STATION
Bernd Kahn
Richard L. Blanchard
William L. Brinck
Herman L. Krieger
Harry E. Kolde
William J. Averett
Seymour Gold
Alex Martin
Gerald L. Gels
December 1974
U. S. ENVIRONMENTAL PROTECTION AGENCY
Office of Radiation Programs
Radiochemistry and Nuclear Engineering Facility
Cincinnati, Ohio 45268
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The Office of Radiation Programs, USEPA, has reviewed this report and approved its publication. Mention of
trade names or commercial products does not constitute endorsement or recommendation for use.
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FOREWORD
The Office of Radiation Programs of the Environmental Protection Agency carries out a national
program designed to evaluate population exposure to ionizing and non-ionizing radiation and to promote
development of controls necessary to protect public health and safety. In order to carry out these
responsibilities relative to the nuclear power industry, the Environmental Protection Agency has
performed field studies at nuclear power stations and related facilities. These field studies have required
the development of means for identifying and quantifying radionuclides as well as the methodology for
evaluating reactor plant discharge pathways and environmental transport.
Electrical generation utilizing light-water-cooled nuclear power reactors is experiencing rapid growth
in the United States. The growth of nuclear energy has been managed so that environmental
contamination is minimal at the present time. The Environmental Protection Agency has engaged in
studies at routinely operating nuclear power stations to provide an understanding of the radionuclides in
reactor effluents, their subsequent fate in the environment, and the real or potential population
exposures.
A previous study at the Yankee Rowe reactor (185 MWe) provided an initial base for evaluating the
environmental effects of operating pressurized water reactors. This particular field study was performed
at the Haddam Neck (formerly called Connecticut Yankee) nuclear power station, a 590 MWe
pressurized water reactor. Results from this study have allowed the evaluation of the operational and
environmental effects of larger pressurized water reactors, and will provide a better basis on which to
evaluate even larger reactors not yet operating.
Comments on this report would be appreciated. These should be sent to the Director, Technology
Assessment Division of the Office of Radiation Programs, Environmental Protection Agency, 401 M
Street, S.W., Washington, D.C. 20460.
.V.
W. D. Rowe, Ph.D.
Deputy Assistant Administrator
for Radiation Programs
111
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Contents
Page
1. INTRODUCTION . . 1
1.1 Need for Study ... .... 1
1.2 The Station . . 2
1.3 The Study ... 2
1.4 References . . 3
2. RADIONUCLIDES IN WATER ON SITE 5
2.1 Water Systems and Samples ... 5
2.1.1 General . . 5
2.1.2 Reactor coolant system 5
2.1.3 Chemical and volume control system 5
2.1.4 Secondary coolant system . 5
2.1.5 Paths of radionuclides from reactor, CVCS, and secondary systems ... .... 9
2.1.6 Other liquids on site 11
2.1.7 Samples .... 12
2.2 Analysis . . . ... 12
2.2.1 General 12
2.2.2 Radiochemistry . . . . 13
2.3 Results and Discussion 13
2.3.1 Radioactivity in reactor coolant water 13
2.3.2 Tritium in reactor coolant water 17
2.3.3 Fission products in coolant water .... . . . . . 19
2.3.4 Activation products in coolant water 19
2.3.5 Radioactivity in secondary coolant water 20
2.3.6 Radionuclides in refueling cavity water . . . ... 21
2.4 References 21
3. AIRBORNE RADIOACTIVE DISCHARGES 23
3.1 Gaseous Waste System and Samples 23
3.1.1 Gaseous waste system 23
3.1.2 Radionuclide release . . 25
3.1.3 Sample collection 26
3.2 Analysis . . 27
3.2.1 Gamma-ray spectrometry ... . . 27
3.2.2 Radiochemical analysis . . .27
3.3 Results and Discussion ... 27
3.3.1 Radioactive gases in reactor coolant . . . . .27
3.3.2 Radionuclides in the waste gas surge sphere 29
3.3.3 Radionuclides in vapor container air 31
3.3.4 Radionuclides in primary auxiliary building air 33
3.3.5 Radionuclides discharged from secondary coolant system
at main condenser steam jet air ejector 34
3.3.6 Radionuclides in turbine building air 36
3.3.7 Radionuclides discharged from air ejector at
turbine gland seal condenser 37
3.3.8 Radionuclides discharged at blowdown flash tank vent 37
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Page
3.3.9 Radionuclides in fuel building air 37
3.3.10 Radioactive gases discharged through the vent stack 37
3.3.11 Radioactive particles discharged through the vent stack 41
3.3.12 Iodine-131 discharged through the vent stack 41
3.3.13 Estimated annual radionuclide discharges 44
3.3.14 Estimated population radiation dose 45
3.4 References ... • • • • 46
4. RADIONUCLIDES IN LIQUID WASTES ... 47
4.1 Liquid Waste System .... 47
4.1.1 Waste processing • .... 47
4.1.2 Radionuclide release 50
4.2 Samples and Analyses . . . . . 51
4.2.1 Samples . . .... 51
4.2.2 Analysis of waste solutions .51
4.2.3 Analysis of circulating coolant water 51
4.3 Results and Discussion 52
4.3.1 Radionuclides in the boron recovery system ... .... 52
4.3.2 Radionuclides in aerated liquid waste . 52
4.3.3 Radionuclide discharge to circulating coolant water ... ... 54
4.3.4 Radionuclides in circulating coolant water 57
4.4 References . . . ... 59
5. RADIONUCLIDES IN THE AQUATIC ENVIRONMENT 63
5.1 Introduction .... .... ... 63
5.1.1 Studies near Haddam Neck .... 63
5.1.2 Connecticut River hydrology 63
5.2 Tritium in River Water ... ... . .... 64
5.2.1 Sampling and analysis 64
5.2.2 Results and discussion . ... 66
5.3 Other Radionuclides in River Water . . ... . ... 66
5.3.1 Gross activity measurements . ... ... 66
5.3.2 Average radionuclide concentrations in the discharge canal . . .... . . 67
5.4 Radionuclides in Vegetation, Plankton, and Algae . ... .... 67
5.4.1 Sampling and analysis . . 67
5.4.2 Results and discussion 68
5.5 Radionuclides in Fish 71
5.5.1 Introduction . . . 71
5.5.2 Collection and analysis . . . . 71
5.5.3 Results and discussion . . . . ... . 72
5.5.4 Estimated average radionuclide concentrations in fish . . . . 77
5.5.5 Estimated population radiation dose . .... 78
5.6 Radionuclides in Shellfish . . . 78
5.6.1 Collection and analysis . . 78
5.6.2 Results and discussion . . 79
5.7 Radionuclides in Sediment . 80
5.7.1 Sampling and measurement . . . . .... . . 80
5.7.2 Description of sediment samples . . . . . 80
5.7.3 Radioactivity measurement . . . . 85
5.7.4 Results and discussion of analyses ... . . ... • ... 85
5.7.5 Results and discussion of probe measurements . . . .... 85
5.7.6 Significance of radioactivity in sediments . . . ,88
5.8 References . ... . • • • .... . . . 89
6. RADIONUCLIDES IN ENVIRONMENTAL AIR 93
VI
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Page
6.1 Introduction .... . . .... 93
6.1.1 Purpose . 93
6.1.2 Environment and meteorology . . • -93
6.2 Measurement of Short-term Radiation Exposure and
Radionuclide Concentration . . . .... . . 95
6.2.1 Air sampling . . . ... 95
6.2.2 Scintillation detector for low-energy photons 95
6.2.3 Measurements . . . . . . ... -95
6.2.4 Estimation of atmospheric dispersion . . 97
6.2.5 Results and discussion . . . . . 97
6.3 References . . . 100
7. RADIONUCLIDES AND RADIATION IN THE TERRESTRIAL ENVIRONMENT ... .101
7.1 Introduction . . ... . 101
7.1.1 Sampling . ... . . 101
7.1.2 Environment of Haddam Neck ... -101
7.2 Tritium in Well Water ... .103
7.2.1 Sampling and analysis . . ... 103
7.2.2 Results and discussion . . . ... .103
7.3 Radionuclides in Food Crops . .104
7.3.1 Sampling and analysis . . . . 104
7.3.2 Results and discussion . 104
7.4 Radionuclides in Milk . . . 105
7.4.1 Sampling and analysis . 105
7.4.2 Results and discussion . . 105
7.5 Iodine-131 in Bovine Thyroids 105
7.5.1 Sampling and analysis .105
7.5.2 Predicted concentration in bovine thyroids .106
7.5.3 Results and discussion .106
7.6 Radionuclides in Deer . . 106
7.6.1 Sampling and analysis ... .... . .106
7.6.2 Results and discussion ... 108
7.6.3 Estimated radiation dose from eating deer meat 108
7.7 External Gamma Radiation .... . 109
7.7.1 Detection instruments 109
7.7.2 Measurements . .109
7.7.3 Results and discussion 109
7.7.4 Estimated external radiation exposure to persons in the environs ... . . . .113
7.8 References 113
8. SUMMARY AND CONCLUSIONS 115
8.1 Radionuclides in Effluents from the Haddam Neck Station 115
8.2 Radionuclides in the Environment at the Haddam Neck Station ... 116
8.3 Monitoring Procedures . . 117
8.4 Recommendations for Environmental Surveillance . ... 117
APPENDICES.
Appendix A Acknowledgements . . . . .119
Appendix B Data Reported by the Haddam Neck Plant 120
Appendix C Radionuclide Calculations . 126
Appendix D Airborne Dispersion Calculations 131
Appendix E Radiation Dose Calculations 137
vn
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Figures
Page
2.1 Coolant Flow Schematic .... 6
2.2 Haddam Neck Electrical Production, August 1967 through December 1969 .... ... 7
2.3 Haddam Neck Electrical Production, January 1970 through December 1971 . . . . . 8
2.4 Effluent Release Pathways . 10
2.5 Gamma-ray Spectrum of Radionuclides from Reactor
Coolant Retained on Cation Exchange Paper . 14
2.6 Gamma-ray Spectrum of Radionuclides from Reactor
Coolant Retained on Anion Exchange Paper . . . . . 15
2.7 Gamma-ray Spectrum of Radionuclides from Reactor Coolant
not Retained on Cation or Anion Exchange Papers . ... ... 16
3.1 Gaseous Waste Disposal System . . . 24
3.2 Gamma Ray Spectrum of Off-gas from Sampling Reactor Coolant .... 28
4.1 Sources of Liquid Waste . 48
4.2 Boron Recovery and Liquid Waste Disposal System . 49
5.1 Sites for Aquatic Sampling 65
5.2 Gamma-ray Spectrum of Algae Collected 69
5.3 Sites for Sediment Sampling and Gamma-ray Probe Measurements .... .81
5.4 Gamma-ray Spectrum of Sediment Sample . . . 84
6.1 Locations for Plume Sampling and Measurement 94
6.2 Gamma-ray Spectrum of Waste Surge Sphere Gas . . ... .96
6.3 Scintillation Detector Response During Air Sampling 99
7.1 Terrestrial Sampling Locations 102
7.2 Locations of Off-site Radiation Exposure Measurements with Survey Meters 110
7.3 Locations of On-site Radiation Exposure Measurements with Survey Meters Ill
Vlll
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Tables
Page
2.1 Radionuclide Concentration in Reactor Coolant Water . . -18
2.2 Radionuclide Concentration in Steam Generator Water . . . 21
2.3 Radionuclide Concentration in Refueling Cavity Water . . . . .22
3.1 Radioactive Gases Released to Stack from Sampling Reactor Coolant . , . . 29
3.2 Radioactivity Contents of Waste Gas Surge Sphere .... . 30
3.3 Radioactivity in Vapor Container Atmosphere . .... 32
3.4 Gaseous Radioactivity in Primary Auxiliary Building Atmosphere . .33
3.5 Radioactivity Contents of Discharge from Main Condenser
Air Ejector in Secondary Coolant System . . . 35
3.6 Gaseous Radioactivity in Turbine Hall Atmosphere . 36
3.7 Gaseous Radioactivity in Fuel Building Atmosphere 38
3.8 Radionuclide Concentrations in Primary Vent Stack Effluents 38
3.9 Average and Annual Estimated Radioactivity Releases from the
Primary Vent Stack 39
3.10 Comparison of Gaseous Radionuclide Release Rates Measured
in Plant Pathways and Stack 40
3.11 Particulate Radionuclide and Gaseous Iodine-131 Concentrations
in Stack Effluent 42
3.12 Summary of Stack Release Rates and Estimated Annual Releases
of Particulate and Gaseous Iodine Radionuclides . 43
4.1 Radionuclide Concentrations in Boron Recovery System 53
4.2 Decontamination Factors for Demineralizing, Demineralizing
plus Filtering, and Evaporating . .... . . 54
4.3 Radionuclide Concentration in Aerated Liquid Waste ... 55
4.4 Radionuclide Discharge in Reactor System Liquid Waste . 56
4.5 Radionuclide Discharge in Secondary System Liquid Waste ... . 57
4.6 Radionuclide Concentrations in Circulating Coolant Water . 58
4.7 Radionuclide Concentrations in Circulating Coolant Water 60
5.1 Concentration of Stable Substances in Connecticut River Water . . 64
5.2 Tritium Concentrations in Connecticut River . . 66
5.3 Radionuclide and Stable Ion Concentrations in Aquatic Plants . 70
5.4 Radionuclide and Stable Ion Concentration in Plankton and Algae 70
5.5 Fish Collected at Haddam Neck 73
5.6 Radionuclide and Stable Ion Concentration in Fresh Water Fish 74
5.7 Average Radionuclide Concentrations in Bone and Muscle in Fresh Water Fish . 75
5.8 Radionuclide and Stable Ion Concentrations in Shad . . 76
5.9 Radionuclide and Stable Ion Concentrations in Shellfish . . . .79
5.10 Mineralogical Analyses of Sediment Samples . 82
5.11 Comparative Analyses of Sediment ... . 83
5.12 Concentration of Radionuclides in Sediment ... .86
5.13 Concentration of Radionuclides in "Core" vs "Top" Samples . . 87
5.14 Concentration of Radionuclides in Core Sample 32 as Function of Depth . . .87
5.15 Net Count Rate of 58Co and 60Co with Nal(Tl) Underwater Probe . . . 88
IX
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Page
5.16 Ratio of Count Rate by Underwater Probe to Radionuclide Concentration
in Sediment Samples .... . . .,. .... 89
6.1 Test Conditions for Sampling Haddam Neck Stack Effluents at Ground Level on Site . . . - 97
6.2 Airborne Radionuclide Concentrations and Radiation Exposure Rates Measured
at Ground Level on Site During Waste Release from Surge Sphere 98
7.1 Radionuclides in Vegetation . . . ... . ... ... .104
7.2 Radionuclide Concentrations in Milk Samples . . . .105
7.3 Estimated Levels of ml in Cow Thyroids 107
7.4 I31I in Bovine Thyroids . ... 107
7.5 Description of Sampled Deer . .107
7.6 Radionuclide and Stable Ion Concentration in Deer Samples .108
7.7 External Radiation Exposure Rates near Haddam Neck .. . .. . ....112
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1. INTRODUCTION
1.1 Need for Study
Determining the potential radiation exposure
beyond the station boundary due to routine operation is
one of several important aspects in evaluating the
impact of a nuclear power station on the environment
and the public. This determination requires detailed
knowledge of radioactive discharges at the station and
environmental pathways leading to radiation exposure.
A useful approach is to quantify this knowledge in
terms of a "model" station in a "model" environment.
The calculational models must then be confirmed or
appropriately modified by observation and measure-
ment at each station under consideration. Several
aspects of these models are in their early stages and
require the collection of additional radioactivity data at
nuclear power stations. In some cases, even the
procedures for obtaining these data must be developed
and tested.
The Office of Radiation Programs, U. S.
Environmental Protection Agency (EPA), has, for
these reasons, undertaken a program of studies at
commercially operated nuclear power stations. The
U.S. Atomic Energy Commission (AEC), state health
or environmental protection agencies, and station
operators are cooperating in this program. This report
describes the third of four projects—two at boiling
water reactors (BWR's) and two at pressurized water
reactors (PWR's). Results of the first two projects, at
the Dresden BWR and the Yankee-Rowe PWR, have
been published."'2'
The four stations were selected for study so as to
provide generally applicable information. Because the
program was begun during the initial expansion in
nuclear power production when only a few relatively
small stations were at full operation in the U. S., care
must be taken in applying observations to newer
stations that are larger and inevitably different in
design and operation. To make the study at this third
station generically applicable, results have been related
to the calculational models presented by the AEC and
EPA, <3'4) although these became available only after the
study was completed. The pertinence of observations at
this station for estimating exposures at other stations
can also be evaluated by comparing the amounts of
discharged radionuclides and the pattern of these
discharges. Gross radioactivity values are given for this
purpose in Section 1.2; annual discharges of individual
radionuclides are now routinely reported by station
operators and compiled by the AEC;(5> and the
separation of data for effluent sources within the
station, as shown in Section 8.1, may well become
available in future compilations.
Guidance for evaluating environmental radiation
exposure by emphasizing the observation of critical
radionuclides, pathways, and exposed populations in
the environment has been available for some time.'6'
This approach concentrates efforts on the few most
important ("critical") causes of exposure in the
presence of many potential ones. Models for computing
radionuclide transfers—for example, from water to
fish, stack to vegetation, and stack to cows' milk for
"'I—have been utilized in the two earlier reports0'2' and
are described fully in the AEC and EPA models.'3'4'
The two agencies have recently published
environmental monitoring guides/1'8' and appreciable
information concerning environmental transfers at
nuclear facilities, beyond that cited in the earlier
reports, has been presented in the past few years.<9"15)
At least two other detailed studies at commercial
nuclear power stations in the U. S. are available, one of
radionuclides in the aquatic ecosystem"" and the other
of terrestrial radiation exposure.1171
The methods for collecting and analyzing samples
have been described in this series of reports with some
care because test measurements at and beyond the
station are important in validating computed
exposures. The problems encountered in sampling and
analysis at the station are very different from those in
the environment. Samples taken at the station usually
show numerous radionuclides, including some that
decay rapidly or are not readily detectable by the
preferred method of gamma-ray spectrometry. Samples
from the environment, on the other hand, usually
contain only a few radionuclides at very low
concentrations, distinguishable from background
radiation only with difficulty, if at all. Analytical
efforts for in-plant samples, therefore, must be focused
-------
on effectively resolving complex mixtures, while
environmental monitoring requires informed sample
selection and ultra-low-level analyses.
1.2 The Station
The study was undertaken at the Haddam Neck
(also called Connecticut Yankee) Nuclear Power
Station, a PWR station built by the Westinghouse
Electric Corp. for the Connecticut Yankee Atomic
Power Co. The station began operating in 1967 and
reached its present maximum power level of 1825
megawatts thermal (MWt) in 1969; the corresponding
gross electrical output is approximately 590 megawatts
(MWe). It had produced 15 terawatt-hours (TW-hr) of
electricity at the end of 1971—more than any other
commercial nuclear power station in the U. S.
Operation of the station is described in several
publications.08'22'
At the time the study began in June 1970, the
reactor had been partially refueled once, in April 1970.
Most fuel elements inserted in 1970 consisted of
uranium oxide (UCh) pellets enriched to 3.67 percent in
235U, and clad in stainless steel. The 235U enrichment
was lower in fuel elements loaded earlier, and higher in
the subsequent loading on April, 1971. A few Zircaloy-
clad test elements were also in the reactor during the
study period.
The station is located in a shallow valley on the east
bank of the Connecticut River. It is in the town of
Haddam Neck, Connecticut, 35 km south-southwest of
Hartford and 26 km north of Long Island Sound.
The study was undertaken at Haddam Neck
because it was one of the two larger PWR stations in
the U. S. that had been in operation for more than a
year in 1970. For comparison, the commercial PWR
stations that had been operated a full year in 1972 are
listed below with their radioactive discharges in curies
difference in those that began operating after 1967 is
the use of fuel clad in Zircaloy instead of stainless steel.
The gross activity at Haddam Neck in both liquid and
airborne waste is shown by the above table to have been
median to values at other stations. The relatively high
amounts of 3H in liquid waste at Haddam Neck and
other older stations are attributed to fission-produced
3H leaking through stainless steel cladding.
1.3 The Study
The study was undertaken in seven field trips to the
station and its environs between June, 1970 and April,
1971. The trips were scheduled to observe radionuclide
concentrations throughout the station operating cycle
at various seasons. Such observations, under ordinary
circumstances, were considered to approximate
average or total radionuclide values for sources and
pathways sufficient for the generic purpose of the
study. The computed averages or totals from this study
are compared, when possible, with values obtained by
the station operator, often by much more frequent
analyses, to evaluate the applicability of the mea-
surements during the field trips. The field trips were
not intended to be inspections of operating practices at
the station.
The study was performed by the Radiochemistry
and Nuclear Engineering Facility at the EPA National
Environmental Research Center, Cincinnati, with the
support of the Technology Assessment Division, ORP-
EPA, and other EPA laboratories. Cooperating in
these studies were the persons listed in Appendix A
from the Connecticut Department of Environmental
Protection, the operating utility, and AEC. The
utility's contractor for aquatic studies, the Essex
Marine Laboratory, directed by Professor Daniel
Merriman, was particularly helpful in collecting
aquatic samples and giving guidance in their field.
WM uuimg ilia
Station
Yankee Rowe
Indian Pt. I
Palisades
R. E. Ginna
San Onofre I
Point Beach I
Haddam Neck
H. B. Robinson
LU ycai :
Year of
initial
operation
1960
1963
1972
1969
1967
1970
1967
1970
Rated
power ,
MWe
175
265
710
420
430
497
590
663
Liquid
1972 power Ci
generation ,
TW(t)-hr
2.4
2.7
5.9
7.7
8.5
10.0
13.8
15.0
Gross
beta
0.02
25.4
6.8
0.38
13.2
1.53
4.8
0.83
waste ,
'H
803
574
208
119
3,480
563
5,890
405
Airborne waste,
Ci
Gases
18
543
505
11,800
19,100
2,810
645
257
Particles
& iodine
0.0008
0.93
0.0097
0.035
0.0005
0.030
0.018
0.027
All stations were built by Westinghouse except Indian
Point I (built by Babcock and Wilcox) and Palisades
(built by Combustion Engineering). A notable
The study had been planned on the basis of results
obtained at the similar but smaller PWR station at
Yankee-Rowe.'
In addition, the
following
-------
information provided guidance: publications describing
the Haddam Neck station,08"22' monthly station
operating reports, reports by the operator's contractor
for environmental surveillance,'23' and the State's
environmental surveillance reports/24'
This information suggested that:
(1) several sources at the station would emit
gaseous and liquid effluents of comparable
dosimetric import;
(2) critical radiation exposure pathways would
include consumption of fish caught near the
outfall, direct radiation from waste storage
tanks, and external radiation from effluent
gases;
(3) bottom sediment and aquatic vegetation near
the liquid waste outfall would be among the
few environmental media to contain readily
detectable radionuclides from the station;
(4) radioiodine might be at detectable levels in the
thyroid of cattle grazing near the station.
The measurement program accordingly emphasized
these aspects of the station and its environment.
1.4 References
1. Kahn, B., et al, "Radiological Surveillance
Studies at a Boiling Water Nuclear Power Reactor," U.
S. Public Health Service Rept. BRH/DER 70-1
(1970).
2. Kahn, B., et al., "Radiological Surveillance
Studies at a Pressurized Water Nuclear Power
Reactor," EPA Rept. RD 71-1 (1971).
3. Directorate of Regulatory Standards,
"Numerical Guides for Design Objectives and Limiting
Conditions for Operation to Meet the Criterion 'As
Low As Practicable' for Radioactive Material in Light-
Water-Cooled Nuclear Power Reactor Effluents,"
AEC Rept. WASH-1258 (1973).
4. Office of Radiation Programs, "Environmental
Analysis of the Uranium Fuel Cycle, Part II - Nuclear
Power Reactors," EPA Rept. EPA-520/9-73-003-C
(1973).
5. Directorate of Regulatory Operations, "Report
on Releases of Radioactivity in Effluents and Solid
Waste from Nuclear Power Plants for 1972," AEC,
Washington, D. C. (1973); in Nuclear Safety 15, 311
(1974).
6. Committee 4, International Commission on
Radiological Protection, "Principles of Environmental
Monitoring Related to the Handling of Radioactive
Materials," ICRP Publication #7, Pergamon Press,
Oxford (1965).
7. "Environmental Radioactivity Surveillance
Guide," EPA Rept. ORP/SID 72-2 (1972).
8. Directorate of Regulatory Standards,
"Regulatory Guide 4.1. Measuring and Reporting
Radioactivity in the Environs of Nuclear Power
Plants," AEC, Washington, D. C. (1973).
9. Jinks, S. M. and M. Eisenbud, "Concentration
Factors in the Aquatic Environment," Rad. Health
Data Rept. 13, 243 (1972).
10. Thompson, S. E., etal., "Concentration Factors
of Chemical Elements in Edible Aquatic Organisms,"
AEC Rept. UCRL-50564 Rev. 1 (1972).
11. Radioecology Applied To Man and His
Environment, International Atomic Energy Agency,
Vienna (1972).
12. Radioactive Contamination of the Marine
Environment, International Atomic Energy Agency,
Vienna (1973).
13. Peaceful Uses Of Atomic Energy, Proceedings
of the Fourth International Conference, Vol. 2 and 11,
United Nations, New York, and IAEA, Vienna (1972).
14. Environmental Behaviour of Radionuclides
Released in the Nuclear Industry, International
Atomic Energy Agency, Vienna (1973).
15. Environmental Surveillance Around Nuclear
Installations, International Atomic Energy Agency,
Vienna (1974).
16. Lentsch, J. W., et al., "Manmade Radionuclides
in the Hudson River Estuary," in Health Physics
Aspects of Nuclear Facility Siting, P. J. Voilleque and
B. R. Baldwin, eds., B. R. Baldwin, Idaho Falls, Idaho
(1971) p. 499.
17. Lowder, W. M. and C. V. Gogolak,
"Experimental and Analytical Radiation Dosimetry
Near a Large BWR," IEEE Transactions NS-21, 423
(1974).
18. Connecticut Yankee Atomic Power Co.,
"Haddam Neck Nuclear Power Plant, Environmental
Report, Operating License Stage," AEC Docket No.
50-213 (July 1972).
19. Directorate of Licensing, "Final Environmental
Statement Related to the Haddam Neck (Connecticut
Yankee) Nuclear Power Plant," AEC Docket No.
50-213'(1973).
20. Graves, R. H., "Coolant Activity Experience at
Connecticut Yankee," Nuclear News 13, 66 (1970).
21. Chave, C. T., "Waste Disposal System for
Closed Cycle Water Reactors," Nuclear Tech. 15, 36
(1972).
22. Coe, R., "Nuclear Power Plants in Operation. 5
Case Histories," Nuclear News 12,41 (1969).
23. Combustion Engineering Combustion Division,
"Operational Environmental Radiation Monitoring
-------
Program, Connecticut Yankee Atomic Power
Company Summary Report 1970"
24. Connecticut Department of Environmental
Protection, "Radiological Data of Environmental
Surveillance - Year 1970," Hartford (1971); also for
Year 1971.
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2. RADIONUCLIDES IN WATER ON SITE
2.1 Water Systems and Samples
2.1.1 General. The power-producing systems at the
Haddam Neck plant are typical of PWR's. Water flows
in the reactor coolant, secondary coolant and chemical
and volume control systems shown in Figure 2.1. Other
water systems on site include boron recovery and waste
disposal, spent fuel pit cooling, safety injection,
component coolant, circulating water, service water,
and sanitary water.
2.1.2 Reactor coolant system.^ During routine
operation, reactor (primary) system water under a
pressure of 2,000 psig is heated in the reactor by the
fission process in nuclear fuel. The water is pumped to
four steam generators in parallel at a total flow rate of
4.6 x 107 kg/h (1.0 x 108 Ib/h). Approximately 1.6 x 105
kg (3.6 x 10s Ib) of water are in a system volume of 2.5 x
105 liters (8,782 ft3). The water temperature is 318° C at
the reactor outlet and 291° C at the inlet at full power.
Reactor coolant water contains lithium hydroxide
and, during most of the fuel cycle, boron in the form of
boric acid. The boron is added to supplement the
control rods for maintaining criticality. It is gradually
decreased from 800 mg/kg immediately after refueling
to 0 mg/kg after 8 months of operation during a cycle
of 10-11 months of operation and 1-2 months for
refueling (see Figures 2.2 and 2.3). The concentration
of lithium is approximately 1 mg/kg throughout the
operating cycle; additional lithium hydroxide is added
when the lithium concentration goes below 0.5 mg/kg.
The lithium hydroxide maintains an acidity for
corrosion control at approximately pH 6 (measured at
25° C) in the presence of boric acid and at pH 10 in its
absence. Monthly averages of these values, reported by
the station operator, are summarized in Appendix B. 1.
The water is under nitrogen gas to provide an inert
atmosphere. Hydrogen gas is added to keep the amount
of oxygen from the radiation-induced decomposition of
water below 0.1 mgAg. The concentration of hydrogen
is usually 30-35 cc/kg; that of nitrogen is
approximately 1 cc/kg at standard temperature and
pressure.
2.1.3 Chemical and volume control system.w The
chemical and volume control system (CVCS) is used to
adjust the pressure, volume, purity and chemical
content of the reactor coolant system (see Figure 2.1).
Makeup water is obtained from two shallow wells on
site, demineralized by passing it through mixed-bed
ion-exchange resin, stored in a 570,000-liter (150,000-
gal) tank, and heated before it is added to the reactor
coolant. One stream of the reactor coolant flows
continuously at the rate of 300 kg/min through the
purification demineralizer—a mixed-bed ion-
exchange-resin column with 25-micron filter—to
maintain the purity of the water in order to prevent
deposition on heat-transfer surfaces and in flow
channels.
For reactor shutdown, water at high boron
concentration (several thousand mg/kg) is added
through the system; at startup, the boron concentration
is lowered to the appropriate operating values by feed-
and-bleed replacement with water. To remove the last
amounts of boron (below 30 mg/kg) near the end of the
operating cycle, reactor coolant water is circulated
through the deborating demineralizer.
2.1.4 Secondary coolant system.™ The secondary
coolant system contains 2.6 x 10s kg of water. The
water is converted to steam at a nominal pressure of
690 psia and temperature of 261° C in four steam
generators. Each steam generator contains
approximately 3.0 x 104 kg (8 x 103 gal) of water.'" The
steam operates a high pressure turbine and, after
passage through four moisture separator reheaters, two
low-pressure turbines. The used steam is condensed at
an absolute pressure of 38 mm (1.5 in) Hg in two
condensers that have a hot-well capacity of 82,000
liters (21,600 gal) each.
The condensed water, together with steam and
water from the moisture separators, is pumped through
feedwater heaters (omitted from Figure 2.1 for the sake
of simplicity) back to the steam generators at a
feedwater temperature of 222° C. Makeup water is
pumped from the wells on site, demineralized, stored in
a 380,000-liter (100,000 gal) tank, and delivered to the
condenser hot-wells as needed.
-------
REACTOR COOLANT SYSTEM
4.6x10 kg/hr
Reactor
PRIMARY LOOPS
(4 Loops , 2OOO psi)
Regenerative
Heat
Exchanger
-(Pressurizer
Oeborating
Oemineralizer
Purification
Oemineralizer
(30O l./min.)
Charge
Pumps
Makeup
-*- SECONDARY COOLANT
SYSTEM
/""""•^ * 3.5xl06kg/hr
^
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CHEMICAL AND VOLUME
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SECONDARY
LOOP
Circulating
Water
(from
River)
POWER - 1825 MWt
WATER -
REACTOR SYSTEM - 1.6x10 kg
SECONDARY SYSTEM - 2.6x IO5kg
Figure 2.1 Coolant Flow Schematic
-------
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Figure 2.2 Haddam Neck Electrical Production, August 1967 Through December 1969
(From Monthly Operation Reports)
-------
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Ffgure 2.3 Haddam Neck Electrical Production, January 1970 Through December 1971
(From Monthly Operation Reports)
-------
2.1.5 Paths of radionuclides from reactor, CVCS,
and secondary systems."^ The radionuclides in the
reactor 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—are fuel that
contaminates the surface of new fuel elements ("tramp
uranium") and fuel that passes into reactor coolant
water from failed fuel elements. The activation
products in reactor coolant water are formed by
neutron irradiation of the water and its contents
(including gases and dissolved or suspended solids) and
of materials in contact with the coolant (container and
structural surfaces, fuel and control rod cladding) that
subsequently corrode or erode.
The radionuclides in the reactor coolant water
circulate and decay within the system, deposit as
"crud" (which may later recirculate), are retained by
the purification and deborating demineralizers (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 2.4.
During routine operation, water and associated
gases leave the reactor coolant system through leaks,
intentional discharge from the volume control tank,
and sample collection. At the time of the study, the
total water leakage rate (see Appendix B.I) averaged
2,200 kg/day. At this rate, intentional discharge
("shim bleed") was not required for routinely reducing
the boron concentration.
Of the total leakage, 75 to 150 kg/day (20 to 40
gal/day) were reported to leak into the secondary
system through steam generator tubes. <3'4) Various
leaks into the vapor containment were estimated13' to
total 110 kg/day (40 gal/day at specific gravity of 0.7),
of which approximately 35 percent initially would turn
to steam. After several weeks, the building atmosphere
will be saturated with water vapor and an amount of
water equal to that leaking will collect in the
containment building sump; the measured rate of water
transfer to the aerated liquid waste system averaged
280 kg/day.'" Leakage into the primary auxiliary
building has been estimated to occur at the rate of 75
kg/day (20 gal/ day),'3' of which 2.6 kg/day is steam,""
and the remainder is water collected as aerated liquid
waste. The largest amount of leaking
water—approximately 1900 kg/day—is collected in
reactor coolant drains in the containment and primary
auxiliary buildings and passes to the boron recovery
system.
The following amount of reactor water was
discharged in 1971 (see Appendix B.3):
boron recovery ("hydrogenated"
liquid waste) system 1.8x10' kg/year
"aerated" liquid waste
system 3.0x10'
total 4.8x10' kg/year
The leakage rate of 1,900 kg/day contributes 0.6 x 10"
kg/year to the boron recovery system in 330 days of
operation. Most of the water enters the boron recovery
system during reactor system shut-down and start-up,
refueling, and boron removal toward the end of reactor
operation. The amount of water from the containment
building sump (9 x 10" kg/yr) is only a small fraction of
the aerated liquid waste.
Radionuclides in the leaking water are expected to
be at the concentrations observed in samples of reactor
coolant water, except for the effects of steam flashing
(i.e., higher concentrations of nonvolatile radionuclides
and lower values for volatile radionuclides). Volatile
radionuclides accumulate in the containment building
until that building is vented. This occurs at least once
and possibly a few times each year (see Section 3.1). In
the primary auxiliary building, ventilating air, which
contains volatile radionuclides, is discharged
continuously. Radionuclides in the shim bleed are at
lower concentrations than in reactor water samples to
the extent that these radionuclides are removed by the
purification demineralizer (see Figure 2.1).
Radionuclides enter the secondary coolant system
through steam-generator leaks. Average leakage rates
were reported to be as follows in 1971 :<5>
February
September
October
November
December
110 kg/day
55
150
150
150
Occasionally, the leakage rate increases rapidly until
the faulty tubes are plugged. In February, 1970, the
reactor-to-secondary leakage reached 6,400 kg/day,
whereupon the loop at fault was isolated, and the
leaking tubes were plugged during refueling in April.
Water leaves the secondary system through steam
generator blowdown, discharge of moisture-saturated
noncondensable gases and system leakage. The steam
generators are blown down continuously during the
day at the rate of 2,300 kg/day. The additional nightly
blowdown is 21,000 kg/day during 8 hours if inleakage
from the reactor system is appreciable, or 7,000 kg/day
during 4 hours for negligible inleakage. An average
blowdown rate of 19,000 kg/day has been reported.'5'
It was estimated that 35 percent of this amount is
-------
Vapor Container Vent
VAPOR
Turbine Blag. Air
to Roof Vent
. '%i
s %
Gases
Liquids
Figure 2.4 Effluent Release Pathways
-------
vented to the atmosphere from the flash tank as
steam;'4' the remainder condenses and is discharged to
the coolant canal. All but the volatile radionuclides are
expected to remain in the water.
Noncondensable gases are removed continuously
from the secondary system by the air ejectors at the
main condensers and through the turbine gland seals.
The gases include volatile radioisotopes of krypton and
xenon. The discharge rate of moisture is approximately
30 kg/day at the main condenser (Section 3.3.5) and
may be somewhat more at the gland seal condenser
(Section 3.3.7).
Leakage from the secondary system plus discharge
with noncondensable gases was approximately 38,000
kg/day at the time of the study.'" The rate of steam
leakage, if it is one-half of the rate for a "model" 3,500
MWt plant,"" would be 9,300 kg/day. Of this, 3,300
kg/day (35 percent) would remain as steam and 6,000
kg/day would condense.(4) The steam is discharged
with turbine-building air, while the water is collected in
turbine-building drains together with water leaking
from the secondary system at the rate of 28,700 kg/day,
(i.e., 38,000 minus 9,300 kg/day). The water is
discharged directly to the coolant canal. Both steam
and water are expected to contain mostly tritium and
volatile radionuclides.
The annual discharge from the secondary coolant
system for 330 days of operation, based on the above
values, is:
blowdown water
blowdown steam
leakage water
leakage steam
total
4.1 x 10' kg/year
2.2 x 10'
11.5 x 10°
1.1 x 10'
18.9 x 10' kg/year
The computed annual discharge rate of water plus
steam from the reactor and secondary coolant system is
thus 4.8 x 106 + 18.9 x 106 = 23.7 x 10' kg/yr. This
sum is almost the same as the 1970 total reported by the
station operator, but is less than the 1971 discharge (see
Section 4.1.1). Considerable fluctuations may occur
from year to year due to changes in water use and leak
rates.
2.1.6 Other liquids on site.m Several ancillary
water systems exist at the station, but only the first
three of the following are believed to result in
radioactive discharges:
(1) Boron recovery and waste disposal system.
The system for liquids is described in Section
4.1.1, and for gases, in Section 3.1.1.
(2) Spent-fuel pit. Fuel pit water consists of
approximately 1.3 x 10' kg (3.5 x 10s gal).
Demineralized water from wells on site is
pumped through the fuel pit in which used fuel
elements are stored. During reactor operation,
the water is circulated continuously through a
mixed-bed ion-exchange demineralizer at 4
kg/s. This water is not discharged during
reactor operation.
(3) Refueling water. The reactor cavity and
refueling canal connecting the reactor vessel
with the fuel pit are flooded during refueling
with 8.9 x 10' kg (2.35 x 10' gal) of borated
(1820 mg/kg) water from the 9.8 x 105-kg (2.6
x 105-gal) refueling water storage tank. The top
of the reactor vessel is opened and fuel
elements are moved from the reactor vessel to
the fuel pit while submerged in the water.
After storage for decay of short-lived fission
products, the fuel is transferred to casks and
shipped off-site. The water from the refueling
cavity is circulated through the purification
demineralizer during refueling. Water leaking
or pumped from this system is transferred to
the boron recovery and waste treatment
system. After refueling, the water in the
reactor cavity is returned to the refueling water
storage tank via the residual heat removal
system.
At the beginning of the long, cold shutdown
for maintenance in April 1969, soluble and
insoluble 58Co wefe released to the refueling
water,(6> causing high surface exposure rates on
many auxiliary plant components. A possible
source of the 58Co is the large amount of nickel-
bearing alloys in contact with the water.
Provision was made to pass the water through
the purification demineralizer during
shutdown. During refueling in 1970 and 1971,
increased 58Co concentrations in the water
were reduced satisfactorily by the purification
system.
(4) Safety injection system. During reactor
operation, the safety injection and core deluge
systems would use the water in the refueling
water storage tank if these systems were
needed. The borated water can be rapidly
pumped into the reactor core in case of a major
loss-of-cooling accident.
(5) Component cooling system. Water for the
component cooling system consists of 30,000
kg (8,000 gal). Potassium chromate (175-225
mgAg) is added as a corrosion inhibitor. The
system is an intermediate cooling system to
transfer heat from components containing
11
-------
reactor coolant to the service water cooling
system. This water is not normally discharged.
(6) Circulating water system. Circulating cooling
water is pumped from the Connecticut River
through the main condenser by 4 pumps at the
rate of 1.4 x 10' kg/min (327,000 gal/min). It
is returned through a canal to the river,
carrying with it the heat extracted from the
steam. The maximum temperature increase in
the circulating cooling water is 12.4° C (22.3°
F).
(7) Service water. Connecticut River water is
pumped into the service water system to cool
several systems at the station, clean the
traveling screens in the circulating coolant
water intake, and inject hypochlorite
periodically into the circulating coolant water
for eliminating bacterial slime. Service water
cools most station auxiliary equipment,
including the component cooling system, spent
fuel pit water, and miscellaneous oil and air
coolers. Three of the four 23,000 kg/min
(6,000 gal/min) pumps provide the required
flow from the circulating coolant water intake.
The service water is discharged into effluent
circulating coolant water.
(8) Sanitary waste. This water is discharged into
two 11,000 kg (3,000 gal) septic tanks, one at
the plant and the other at the information
center.
2.1.7 Samples. To identify potential radioactive
effluents, liquids at the Haddam Neck Nuclear Power
Plant 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 in
plastic bottles by station personnel:
1. reactor coolant, 2 liters, collected July 24, 1970
at 0900;
2. reactor coolant, 1.5 liters, collected Nov. 20,
1970 at 0830;
3. reactor coolant, 1 liter, collected March 16,
1971 at 1000;
4. steam generator blowdown, 3.5 liters (pH 6.8),
collected July 23, 1970 at 1500 hours;
5. composite steam generator blowdown, 3.5
liters, collected Sept. 15,1970 at 1000;
6. steam generator blowdown, 3.5 liters, collected
Nov. 20, 1970 at 0830;
7. steam generator blowdown, 3.5 liters, collected
Mar. 15,1971 at 1000;
8. steam generator blowdown, 3.5 liters, collected
Mar. 16,1971 at 1000;
9. steam generator blowdown, 3.5 liters, collected
April 14,1971 at 1945;
10. refueling cavity water, 1 liter, collected May 7,
1971,atlll5.
One liter each of samples #1,2, and 6, and 500 ml
of sample #3 were acidified with cone. HNCb (10%
v/v) to minimize deposition of radionuclides on the
walls of the bottle. The unacidified portion was
reserved for radioiodine, tritium, and 14C analyses.
2.2 Analysis
2.2.1 General. Aliquots of all samples were counted
for gross alpha and beta activity, examined with
gamma-ray spectrometers and analyzed
radiochemically. Analyses were performed for high-
yield fission products and common activation products
in reactor water. Because radioactive decay between
sampling and analysis was usually between 1 and 2
days, radionuclides with half-lives less than 6 hours,
and in some cases, 24 hours, could not be measured.
Aliquot volumes ranged from 1 to 200 ml.
Radionuclide concentrations were computed from
count rates obtained with detectors calibrated with
radioactivity standards. Values were corrected for
radioactive decay and are given as concentrations at
sampling time. Half-lives and branching ratios are from
recent publications/7"101 The concentration of
radioactive progeny such as 133Xe, 133mXe, and 13SXe was
corrected for ingrowth in the sample between collection
and analysis.
The difficulty of retaining radionuclides in solution
reported earlier00 was also observed during this study,
in that radionuclides remained on empty plastic sample
containers when the liquid samples were poured out
after contact periods of days to weeks. Even with
acidification, losses of 10-50 percent were observed for
radionuclides such as 5lCr, 54Mn, 58Co, 60Co, and 59Fe.
The following techniques were applied to prevent
underestimating the radionuclide content of liquid
samples:
12
-------
(1) Cutting the empty sample bottle into small
pieces, placing it in a container of known
counting efficiency, and measuring gamma-
ray emitters.
(2) Collecting the liquid sample on a dry sponge in
a container to saturate the sponge with the
liquid at a volume calibrated for the counting
efficiency of gamma-ray emitters.
(3) Passing samples of low ionic content
immediately through cation- and anion-
exchange membrane filters* to collect
particulate and ionic radionuclides on the
filters02' for analysis by a gamma-ray
spectrometer. The filtrate was also analyzed.
(4) Leaching the empty sample bottle with three
25-ml portions of hot aqua regia to collect for
analysis those radionuclides that do not emit
gamma rays. The completeness of the leaching
was checked by assuring that no gamma-ray
emitters remained in the bottle. Results
obtained for samples treated with procedures
(1) or (4) were corrected by including the
amounts retained on container walls.
2.2.2 Radiochemistry. Radionuclides that emit
gamma rays were identified by their characteristic
gamma-ray energies with a Ge(Li) detector and 1600-
channel spectrometer in aliquots of reactor coolant and
liquid wastes. Spectral analyses were obtained at
intervals to eliminate interference by shorter-lived
radionuclides and to confirm the identity of the
measured radionuclides by observing their half lives.
The large number of radionuclides and their
concentration differences in coolant water made
identification after collection on ion-exchange papers
particularly convenient. This technique also identified
the ionic form of the radionuclides. Sample #3
(Section 2.1.7) was analyzed in this manner by filtering
35 ml in a suction apparatus through 3 cation- and 2
anion-exchange papers in series. The papers were then
separated and transferred individually to containers for
spectral analysis. The filtrate was also analyzed.
Figures 2.5, 2.6, and 2.7 show the Ge(Li) spectra of
each fraction 2 days after collection.
The radionuclides "Mo-"mTc, '33I-mXe, and 135I-
135Xe were on the anion paper, and '33Xe and l35Xe were
in the filtrate. All other radionuclides were on the top
(cation-exchange) filters. These included longer-lived
ones at lower concentrations that could only be
measured after many of the radionuclides seen in
Figure 2.5 had decayed. Chromium-51, although
expected to be anionic when dissolved, was retained on
the top, cation-exchange filter, possibly because it was
in particulate form; some gaseous 133Xe was retained on
the same filter, possibly absorbed on the resin.
In reactor coolant samples, gamma rays of energies
below 160 keV from relatively short-lived
radionuclides were obscured by the radiations from
relatively large amounts of 133Xe. The 133Xe was
removed by boiling and stirring a 35 ml aliquot of the
coolant water with 5 ml cone. HC1. Replicate tests
indicated that less than 1 percent of the 131I was
volatilized by this process.
A 10-cm x 10-cm Nal(Tl) detector with 200-
channel spectrometer was used to analyze samples that
contained only a few radionuclides at low levels of
radioactivity. The better energy resolution of the
Ge(Li) detector was generally unnecessary for these
samples, and the higher counting efficiency of Nal(Tl)
detectors was advantageous.
A 400-mm2 Si diode with 400-channel spectrometer
was used to identify radionuclides that emit alpha
particles in samples that showed detectable gross alpha
activity. The alpha-particle energies were determined
within +30 kev. The amount of each alpha-particle-
emitting radionuclide was computed from the gross
alpha activity and the relative counts at each energy
peak.
Radionuclides were separated chemically to
confirm gamma-ray spectral identification, 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.113' After chemical
separation, the following detectors were used: Nal(Tl)
crystal plus spectrometer for photon-emitting
radionuclides; low-background end-window Geiger-
Mueller (G-M) counter for 14C, 32P, 35S, 89Sr, '°Sr, IWI,
147Pm, and 185W; liquid scintillation detector plus
spectrometer for 3H, I4C, and "Ni; and xenon-filled
proportional counter plus spectrometer for 55Fe.
Measurements with the G-M detector included
observation of the effect of aluminum absorbers on
count rates to determine maximum beta-particle
energies and thus confirm radionuclide
identification.
-------
0
800
50
850
100
900
150
950
200
1000
550
1350
600
1400
650
1450
700
1500
750
1550
800
1600
250 300 350 400 450 500
1050 1100 1150 1200 1250 1300
CHANNEL NO. (1.00 keV/channel)
Figure 2.5 Gamma-ray spectrum of radionuclides from reactor coolant retained on cation exchange
paper, 0-1600 keV
Detector: Ge(Li), 10.4 cm2x 11 mm, trapezoidal.
Sample: Cation exchange paper containing activity from 35 ml, collected March 16, 1971 at 1000.
Count: March 18, 1971; 50 minutes (background not subtracted).
-------
800
1600
CHANNEL NO. (1.00 keV/chonne I)
Figure 2.6 Gamma-ray spectrum of radionuclides from reactor coolant retained on anion exchange
paper, 0-1600 keV.
Detector: Ge(Li) , 10.4 cm2 * 11 mm, trapezoidal.
Sample: Anion exchange paper containing activity from 35 ml, collected March 16, 1971 at 1000.
Count : March 18,1971: 50 minutes (background not subtracted).
-------
10'
105
10'
o io3
u
IO
10 -
0)
V
X
0) en
X
co
O
CO
I
100 150 200 250 300 350 400 450 500
CHANNEL NO. (1.00 keV/channel)
550 600 650 700 750 800
Figure 2.7 Gamma-ray spectrum of radionuclides from reactor coolant not retained on cation
or anion exchange papers, 0-1600 keV.
Detector: Ge(Li), 10.4 cm2 x 11 mm, trapezoidal .
Sample: 35 ml effluent from ion-exchange column, collected March 16, 1971 at 1000.
Count: March 18., 19.71; 16.6 minutes (background not subtracted).
-------
131I, 133I, 135I, I33mXe, and 135Xe. The sum of all other
measured radionuclides ranged from 0.002 to 0.01
uCi/ml in the three samples. The sum of all listed
radionuclides except 3H and the noble gases was
approximately 0.1 uCi/ml. In comparison, the average
monthly gross radioactivity excluding 3H (and
probably also excluding noble gases) reported by the
station during the sampling periods (see below) was
between 0.2 and 0.5 pCi/ml. These gross radioactivity
measurements may include relatively abundant short-
lived radionuclides such as the fission products 132I and
134I and the activation products 18F and "Mn.
Fission products other than radioiodine and
radioxenon were at relatively low concentrations, and
several high-yield fission products could not be
detected at the limiting sensitivity of approximately 1 x
IG* uCi/ml (see footnote 3 to Table 2.1). Most of the
other radionuclides are neutron activation products
that have been reported earlier."1'15' They are formed
in water, steel, antimony (in the Sb-Be neutron source),
and zirconium (in Zircaloy-2 cladding of 2 fuel rods).
The activation products 14C, 3SS, and "Ni were at
relatively low concentration, as previously reported for
the Yankee-Rowe reactor."4'
Alpha activity was found in two of the three
samples. Isotopic analysis by alpha spectrometer of the
highest-level sample of March 16, 1971, showed the
following components three years after sample
collection:
86-yr 2MPu
"'Pu and
24,400-yr
6,580-yr
17.6-yr ""Cm
gross alpha activity
9.1 x 10" iiCi/ml
2.4 x 10'
0.7 x 10'
12.2 x 10' uCi/ml
Based on the calculated production of the
transuranium elements,06' the concentration of MOPu
would be slightly greater than that of 23'Pu in the
unresolved sum of the two isotopes. The decrease of 3.7
x 10"7 uCi/ml in three years from the initial gross alpha
activity shown in Table 2.1 is attributed to the
radioactive decay of 162-d M2Cm. These five
radionuclides constitute more than 95 percent of the
calculated total alpha activity in the fuel."6'
Radionuclide concentrations in the three samples
were considerably different. This is to be expected
because a number of factors change during a fuel cycle;
notably, radionuclides accumulate in the fuel, while in
the coolant the pH value increases as the boric acid
concentration decreases (see Appendix B.I). The
monthly average values reported by the Haddam Neck
Plant at the sampling periods are:
July 1970 Nov. 1970 March 1972
Month of core II cycle 1st 5th 9th
Power level, MWe 551 586 571
Boron, mg/kg 571 278 0.6
PH 6.2 6.8 10.0
Gross activity, pCi/ml 0.20 0.26 0.50
Radionuclide concentrations in coolant water are also
affected by many other variables, especially the quality
of the fuel elements, the rate and effectiveness of
coolant-water purification, and the extent of
radionuclide accumulation within and loss from the
coolant system.
2.3.2 Tritium in reactor coolant water. The
measured 3H concentrations in Table 2.1 are consistent
with the average concentrations of 5.43, 3.68, and 4.52
uCi/ml, respectively, reported for those months by the
station operator (see Appendix B.I). The sources of the
tritium in coolant water are believed to be, in order of
importance: (1) ternary fission in the fuel, (2) 10B (n,2
alpha) reaction in the boron dissolved in coolant water,
and (3) 6Li (n.alpha) reaction with the
lithium—containing 0.1 percent 6Li—in coolant
water.04'17' The calculated generation rate by fission
(see Appendix C.I) is 210 uCi/s or 6,600 Ci/yr. Its
production from boron and lithium computed for a
station at a power level of 1473 MWt (which is the
equivalent at Haddam Neck to full power at 0.8
capacity factor) is approximately 400 and 30 Ci/yr,
respectively.08' Accordingly, of the tritium discharge
of 5,800 Ci during 1971, the first full year of operation
after the initial core had been partially replaced, 5,400
Ci would be from fission. This suggests that
approximately 80 percent of the tritium formed by
fission during the year had leaked through the
stainless-steel cladding. The actual leakage fraction
would be somewhat less because some additional
tritium from previous years would have accumulated in
fuel elements remaining from core I.
Concentrations of tritium in the coolant (see
Appendix B.I) were generally higher after the first
refueling in May-June 1970 than before refueling. This
may have been due to the new fuel elements or
operation at higher power levels. Calculations for the
first core had indicated that only approximately two-
thirds of the fission-produced tritium had leaked
through the cladding."8' Immediately after refueling,
and tapering off during a 2-week period, a much higher
level of tritium than usual was observed in the coolant;
this is attributed to a higher leakage rate caused by
redistribution of power and temperature in the fuel
elements remaining from the preceding core.08' If,
according to the annual tritium production values
given above, almost all of the tritium in coolant water is
17
-------
Table 2.1
Radionuclide Concentration in Reactor Coolant Water, pCi/ml*
Radionuclide
12.3 -yr
50.5 -d
28.5 -yr
65 -d
35.1 -d
66.2 -hr
39.6 -d
8.06-d
20.9 -hr
6.7 -hr
5.29-d
2.3 -d
9.1 -hr
2.07-yr
13 -d
30 -yr
12.8 -d
32.4 -d
284 -d
2.34-d
5730 -yr
15.0 -hr
14.3 -d
88 -d
27.7 -d
313 -d
2.7 -yr
44.6 -d
270 -d
71.3 -d
5.26-yr
92 -yr
60.2 -d
115 -d
3Hf
89Sr
90Sr
95Zr|
9SNbt
"Mof
103Ru
I31I
133I
,J5j
l33Xe
133m v
AI
135Xe
l34Cst
13°Cst
l37Cs
140Ba
141Ce
144Ce
239Np
gross
14C
24Na
32p
35S
51Cr
54Mn
5SFe
59Fe
"Co
58Co
60Co
63Ni
l24Sbt
182Ta
* Concentration at
July 24, 1970
from fuel
6.0
1.0x10"
3 xlO"7
1.1x10"°
1.5x10-°
NA
<1 xlO'7
8.9x10"
NA
NA
8.8xlO"2
e NA
NA
6.2x10-'
7.6x10""
8.5x10-'
1.0x10"
<1 xlO"
<1 xlO"
NA
alpha <3 xlO"'
from activation of water, cladding,
9.7xlQ-6
NA
4.2x10"
<1 xlO"7
3.5x10"'
1.3xlO"3
3.9x10"
4.0x10"'
1.6x10"
3.6xlO"3
5.6x10-"
4.9x10"'
1.6x10"
<1 xlO-7
November 20, 1970
3.5
5.0x10-°
1.6x10"
8 xlO"
2 xlO"7
2.0x10"
1.3x10"°
2.4x10-'
3.3x10-'
2.5xlQ-2
1.1
9.3x10"
2.8x10"
2.7x10"
8.2xlO"5
2.9x10""
3 xlO"
-1.0x10"°
<2 xlO"
2.7x10"'
2 xlO"
and construction materials
3.4xJO"
9.0x10"
l.lxlO"5
1.2x10'"
1.0x10"
1.3x10"
1.3x10-"
4.2x10"°
5 xlO"
1.5x10"'
1.8x10"
NA
2 xlO~7
1.5x10-°
March 16, 1971
4.5
4.5x10"
2.7x10"
1.9x10"
1.0x10-'
4.7x10"
1.1x10-'
3.7x10"
7.2x10"'
5.6xlO"2
5.5x10"'
1.8x10"
7.2x10-'
7.7x10""
2.4x10"
7.2x10""
1.0x10"
2.3x10-'
2.2x10-'
5.2xlO"
4.9x10"
NA
1.0x10"
NA
NA
2.2x10-'
2.0x10"
<2 xlO"
<2 xlO"'
<4 xlO"7
7.1x10"
5.5x10"°
NA
<1 xlO-'
<3 xlO"
time of sampling; water at standard temperature and pressure.
t 3H is also an activation product; 95Zr, its daughter 95Nb
, and "Mo may also
be activation products; '3"Cs and '3°Cs are produced by (n,gamma) reactions
with fission-produced 133Cs and '35Cs, respectively, and 124Sb may also be
produced by (n,gamma) reactions with fission-produced 123Sb.
Notes:
1. NA = not analyzed
2. < values are 3 sigma counting error.
3. The following fission products were not detected (usually < 1x10"' uCi/1):
93Y, "Zr, 106Ru, 127Sb, I29I, 132Te, l43Ce, 147Nd. The radionuclides
"Zn, 1K""Ag, '83Ta, and "5W were also not observed at this minimum
detectable level.
18
-------
from the fuel, then the major variations in the tritium
concentration would arise from refueling, sudden
power changes, and changes in the turnover rate of
coolant water.
2.3.3 Fission products in coolant water. The
measured concentrations of U'I on the three occasions
recorded in Table 2.1 were reasonably consistent with
average monthly values from Haddam Neck operating
reports (see Appendix B.I); 131I/'33I atom ratios were
also in agreement:
July 1970 Nov. 1970 March 1971
"'I concentration, )iCi/ml
this report
operator's report
8.9 x 10° 2.4 x 102 3.7 x 10'
4.6 x 10° 2.8 x 10' 3.0 x 101
'!/'"! atomic ratio
this report — 6.7 4.8
operator's report 4.6 5.3 3.7
The concentrations of the radioiodine and many of
the other fission products listed in Table 2.1 are
consistent with calculated radionuclide release rates
from the fuel when the fraction of fuel elements that
leak radioactivity is 2 x 10"4, as reported by the
operator.'3' The fuel release rate, Rf (in uCi/s), is the
product of the accumulation in fuel, A (in uCi, see
Appendix C.I, last column), the escape rate coefficient,
E, for the element'4' (in s"1, see footnote to Appendix
C.2), and the above mentioned fraction, F, of fuel
elements releasing radioactivity:
Rf = AEF (2.1)
The calculated values of Rf are given in column 3 of
Appendix C.2. At equilibrium, they are equal to the
rates at which the radionuclides leave coolant water.
Thus, the concentration of a radionuclide, Cr (in
uCi/g), in the reactor coolant water is related to Rf by:
Cr = Rf/(Vr)(2Xr) (2.2)
where
Vr : amount of reactor coolant water (1.6
x 108 g)
ZXr : sum of radionuclide turnover coefficients
in reactor coolant water (s"1).
The turnover coetficients summed in column 4 of
Appendix C.2, are as follows:
(1) The average coolant water loss reported by
the operator was 2,200 kg/day, hence the
water turnover coefficient, Xwater turnover =
(2,200 kg/ day)/(1.6 x 105 kg x 8.64 x 104
s/day)= 1.6xlOV
(2) Radioactive decay is characterized by the
decay constants, Xa, from column 3 of
Appendix C. 1 that range from 3 x 10"3 to 8
xlO-'V.
(3) The removal of ions by the demineralizer is
the product of the removal fraction, fa,
and the water flow through the
demineralizer relative to the total amount
of coolant water, hence the turnover
^coefficient, Xacminerniizer = (300 kg/min x
fd)/ (1.6 x 10s kg x 60 s/min) = 2.8 x 10'5
s"' when fa is 0.9 (for all ions except
cesium)'4' or 1.6 x 10'5 s'1 when fa is 0.5
(for cesium).
Radionuclides may also be removed from water by
processes such as surface deposition or volatilization.
The concentration values calculated in Appendix C.2
will be too high whenever such additional removal
processes occur.
For a few radionuclides, these computed
concentrations in Appendix C.2 are considerably
different than the measured values in Table 2.1.
Computed values are higher for "Nb, radioxenon, 134Cs,
and >37Cs, and lower for 124Sb and 144Ce. Most of the
measured 124Sb probably was formed by neutron
activation (see Table 2.1). Radioxenon isotopes are
volatile (see Section 3.3.1). In view of the approximate
nature of the values of A, E, F, Xwater turnover, and fa on
which the calculations are based, the agreement with
many of the measured values suggests that the
calculations yield useful estimates of reactor coolant
concentrations.
The ratios of the measured concentrations to the
computed values in Appendix C.2 were similar for the
three iodine isotopes of widely different half lives:
Measured concentrations/computed concentrations
Radioiodine Nov. 1970 March 1971
1.1
1.0
1.7
2.2
'"I 1.1 2.5
This suggests that the iodine isotopes in the coolant are
in an "equilibrium mixture" that is formed when the
passage of the isotopes from the fuel to the coolant is
slow compared to the longest half-life of the isotopes (8
days, in this case).
Radioisotopes of tellurium (e.g., 132Te) were not
detected in the reactor water samples, possibly because
tellurium is retained within the cladding or
immediately removed from coolant water by deposition
on surfaces. For "Mo, which was observed to be
anionic (see Section 2.2.2), the ion-exchange removal
fraction of 0.9 that was applied for other anions yielded
computed values consistent with measured
concentrations, although surface deposition
("plateout") had been indicated to be the major
removal mechanism.<4>
2.3.4 Activation products in coolant water. The
measured concentrations of the activation products in
Table 2.1 were generally lower than had been
calculated:
19
-------
Radionuclide Concentration in Reactor Coolant Water,
uCi/liter
51Cr
"Mn
"Fe
!'Fe
!'Co
'"Co
Model
calculation141
1.9
0.31
1.6
1
6
2
Pre-operational
calculation'"
4.4
—
1.6
8.5
1.9
Measured
0.01-0.04
0.002-1.3
< 0.0002-3.9
< 0.0002-0.04
0.07-3.6
0.006-0.6
The highest measured concentrations of S4Mn, 58Co,
and 60Co approached the calculated values, however,
and the measured concentration of 55Fe exceeded the
calculated value.
Following the observation of 14C at Yankee-
Rowe,<14) its annual production at PWR stations was
estimated to be 30.4 Ci per 1,000 MWe in one report0"
and 6.3 Ci per 1,000 MWt in another.'20' According to
these calculations, it is produced mostly by the
reactions 14N(n,p) and 17O(n,alpha) in both the fuel and
the coolant, while little 14C is formed by 13C(n,gamma)
and ternary fission. The amount of 14C generated at the
Haddam Neck station at a power level of 1825 MWt
and an 80-percent use factor would be either 14 or 9
Ci/yr.
2.3.5 Radioactivity in secondary coolant water.
Many of the radionuclides observed in reactor coolant
water were measured at concentrations lower by two
orders of magnitude in blowdown water from the steam
generators. These values, given in Table 2.2, were taken
to represent concentrations in the 1.2 x 105 kg of steam
generator water. The remaining 1.4 x 105 kg of
secondary system water are steam and condensate.
Radionuclide concentrations (except for 3H) are
believed to be lower in these then in reactor coolant
water by factors of 1,000 (nonvolatile radionuclides) or
100 (volatile radionuclides in steam).<4)
The rate at which water leaks from the reactor
coolant into the secondary coolant system at the steam
generators can be determined by measuring
radionuclide concentrations in the two systems and one
other parameter. In analogy to equation 2.2, at
equilibrium in the secondary system,
Cs = Rr/Vs2X (2.3)
where Cs is the radionuclide concentration in steam
generator water, Rr is the rate of radionuclide
inleakage, vs is the volume of water, and 2 X is the sum
of the turnover coefficients of the radionuclide in the
water. The rate of radionuclide inleakage can be taken
to be the product of the radionuclide concentration in
the reactor coolant, Cr, and the rate of leakage of water
from the reactor coolant to the secondary coolant
system, wr. The value of SX for nonvolatile
radionuclides is the sum of the radioactive decay
constant, Xd, and Ws/vs, where ws is the water discharge
rate. Thus,
Q = Wr (2.4)
Cr Ws + (Vs)(Xd)
As defined in Appendix C.3, vs and Ws refer respectively
to the water volume and blowdown rate at the steam
generators for nonvolatile radionuclides. The same
equation can be used for 3H by applying the respective
values for water in the entire secondary system and
blowdown plus leakage rates from the entire system.
The calculation is simplified if the half life of the
measured radionuclide is either very long or very short,
because (vs)(Xd) or ws will then be negligible relative to
the other.
From 3H measurements in the two systems on three
occasions and water discharge rates from the secondary
system reported by station staff,<21) the following
reactor-to-secondary leakage rates, Wr, were computed
with equation 2.4:
Sampling date
July 23/24, 1970
Nov. 20, 1970
Mar. 15/16, 1971
(VG
2.3xl03
3.4xl03
4.9x10°
Ws,
kg/day
55,000
53,000
64,000
Wr,
kg/da;
130
180
310
The 3H concentrations are from Tables 2.1 and 2.2 on
the indicated dates; values of Ws are the average 38,000-
kg/day leakage rate from the secondary system
(Section 2.1.5) plus the daily blowdown rate (Appendix
C.3). The first inleakage rate lies within the range of
75-150 kg/day reported by the station operator,<4)
while the other two are somewhat higher. Possible
sources of error are the use of an average leakage rate
for the secondary system because the specific values for
these days were not available, and sample collection on
two occasions with a one-day interval instead of
simultaneously.
The concentrations of the nongaseous
radionuclides in steam generator water, calculated by
equation 2.4 from the reactor coolant concentrations in
Table 2.1 and the above-listed reactor-to-secondary
leakage rates, are given in Appendix C.3. The
calculated values for November 20, 1970 are all within
a factor of two of the measured concentrations in Table
2.2. On March 15/16, three of the values—for S8Co,
9SZr, and 135I—are different, and on July 23/24, almost
every calculated value differs from the measured one.
On these two occasions, also, some radionuclides
predicted to be at measurable levels were not detected.
Such differences can arise from fluctuations in the
radionuclide concentrations and rates of water
turnover, or from a sample that does not represent all
four steam generators.
20
-------
Table 2.2
Radionuclide Concentration in Steam Generator Water, uCi/ml
July 23,
Radionuclide 1970
3H
14C
24Na
32P
51Cr
MMn
"Fe
MCo
"Co
"Sr
"Zr
"Mo
131I
133I
13SI
134Cs
13SCs
137Cs
1.4xlO-2
<1 xlO~7
NA
3 xlO-7
<1 xlO-'
1 xlO-7
<1 xlO-'
2 xlO-7
1 xlO-7
1 xlQ-"
1 xlQ-7
NA
8.0x10-'
NA
NA
2.6x10-'
5 xlO-7
3.0x10-'
Sept. 15,
1970
3.3xlQ-3
<1 xlO-'
NA
2.2x10-'
1.6x10-°
1.1x10-°
1.8x10-'
2.0x10-'
7 xlO-'
< 1 xlO"8
1 xlO-7
9.1x10-°
4.6x10-
NA
NA
4.3xlQ-5
1.3x10-'
4.9xlQ-5
Nov. 20,
1970
1.2xlQ-2
2 xlQ-'
NA
<1 xlO"7
<1 x!Q-°
3 xlO-7
1.9x10-'
1.7x10-°
2 xlO"7
1 xlO-8
< 1 xlQ-'
<2 x!0~7
1.2x10-
8.9x10-'
NA
3.0x10-'
6 xlO-7
4.0x10"°
Mar. 15,
1971
2.2xlQ-2
3 xlQ-7
NA
.4xlO~°
< xlO-'
< xlO-7
< xlO"
< xlO'7
xlO-7
5 xlO-
< 1 xlO"7
4.0x10-°
5.5x10-
3.8x10-
NA
.l.lxlQ-5
2.5x10-°
l.lxlQ-5
Mar. 16,
1971
NA
NA
3.0x10-°
NA
NA
<1 xlO 7
NA
1 xlO'7
1 xlO-'
NA
NA
1.8x10-'
3.3x10-
3.3x10-'
1.4x10-
8.1x10-°
1.6x10-'
8.0x10-°
April 14,
1971
2.1xlO'2
< 1 xlO"7
NA
1.9x10-°
<1 xlO'6
1 xlQ-7
<1 xlO"'
< 1 xlO"7
< 1 xlO"7
1 xlO-
<1 xlO'7
7.7x10"°
3.6x10-
5.4xlQ-4
NA
8.9x10-'
1.8x10-°
8.4x10'°
Notes:
1. NA: not analyzed
2. < values are 3-sigma counting error
The following radionuclides were not detected:
3.
(< 1 x ID'7 uCi/ml)
"Fe, "Co, "Ni, "Sr, "Nb, 1J-°- Ag, 140Ba, mCe, 182Ta, and "!W;
(< 5 x 10-' uCi/ml) 144Ce, 23'Np; (< 1 x 10-9uCi/ml) gross alpha.
2.3.6 Radionuclides in refueling cavity water. While
the reactor was shut down for refueling (April 16 to
May 21), a sample of refueling cavity water was
analyzed radiochemically with the results shown in
Table 2.3. All radionuclides except 3H and relatively
short-lived ones such as "Mo and 131I were at higher
concentrations in this water than in the most recent
reactor coolant sample (Table 2.1, last column).
Radioisotopes of cobalt and iron were at particularly
high concentrations relative to reactor coolant water,
and 10*Ru and "*" Ag were found in this sample
although below the detection limit in reactor coolant
water.
2.4 References
1. Connecticut Yankee Atomic Power Company,
"Facility Description and Safety Analysis," Vol. 1 and
2, AEC Docket No. 50-213-5 and 50-213-6, Haddam
Neck, Conn. (1966), and R. Graves, personal
communication.
2. Brinck, W. L., "Monitoring of Effluents from a
Nuclear Power Plant," M.S. Thesis, Dept. of Chemical
and Nuclear Engineering, University of Cincinnati
(1971).
3. Directorate of Licensing, "Final Environmental
Statement Related to the Haddam Neck (Connecticut
Yankee) Nuclear Power Plant," AEC Docket No.
50-213 (1973), pp 3-23 to 3-25.
4. Directorate of Regulatory Standards,
"Numerical Guides for Design Objectives and Limiting
Conditions for Operation to Meet the Criterion 'As
Low as Practicable' for Radioactive Material in Light-
Water-Cooled Nuclear Power Reactor Effluents,"
AEC Rept. WASH-1258 (1973).
5. Connecticut Yankee Atomic Power Company,
"Haddam Neck Plant Monthly Operational Reports,"
Nos. 70-1 to 71-12, Haddam, Conn. (1970,1971).
6. Graves, R. H., "Coolant Activity Experience at
Connecticut Yankee," Nuclear News 13, 66 (1970).
7. Lederer, C. M., J. M. Hollander, and I. Perlman,
Table of Isotopes, John Wiley, New York (1967).
21
-------
Table 2.3
Radionuclide Concentration in Refueling Cavity Water on May 7, 1971
Radionuclide
3H
MC
32P
51Cr
54Mn
"Fe
S9Fe
57Co
58Co
"Co
"Sr
90Sr
95Zr
Concentration,
uCi/ml
3.6
3
1.2
1.6
1.5
1.4
1.8
3.6
1.4
8.5
1.0
1.6
8.9
X
X
X
X
X
X
X
X
X
X
X
X
X
1C"2
10"
io-4
io-3
io-4
10~J
io-4
10-'
io-2
io-4
io-4
io-5
io-5
Radionuclide
95Nb
"Mo
103Ru
106Ru
1100 Ag
124Sb
,„!
134Cs
136Cs
137Cs
140Ba
141Ce
144Ce
Concentration,
uCi/ml
1.4
1.3
6.8
2.0
1.4
6.8
2.3
6.4
5.2
5.5
7.8
1.5
2.0
X
X
X
X
X
X
X
X
X
X
X
X
X
io-4
io-s
io-5
io-5
io-5"
io-5
io-J
io-3
io-4
io-3
io-s
io-'
io-4
Notes:
1. Concentrations at time of sampling.
2. Reactor was shut down for refueling on April 16, 1971.
8. McKinney, F. E., S. A. Reynolds, and P. S.
Baker, "Isotope User's Guide," AEC Rept. ORNL-
IIC-19(1969).
9. Martin, M. J. and P. H. Blichert-Toft,
"Radioactive Atoms," Nuclear Data Tables A8, 1
(1970).
10. Wakat, M. A., "Catalogue of Gamma-Rays
Emitted by Radionuclides," Nuclear Data Tables A8,
445(1971).
11. Kahn, B., et al., "Radiological Surveillance
Studies at a Boiling Water Nuclear Power Reactor,"
Public Health Service Rept. BRH/DER 70-1 (1970).
12. Gilbert, R., General Electric Company,
Vallecitos, California, personal communication (1970).
13. Krieger, H. L. and S. Gold, "Procedures for
Radiochemical Analyses of Nuclear Reactor Aqueous
Solutions," EPA Rept. EPA-R4-70-014 (1973).
14. Kahn, B., et al., "Radiological Surveillance
Studies at a Pressurized Water Nuclear Power
Reactor," EPA Rept. RD-71-1 (1971).
15. Rodger, W. A., "Safety Problems Associated
with the Disposal of Radioactive Wastes," Nuclear
Safety 5, 287 (1964).
16. "Siting of Fuel Reprocessing Plants and Waste
Management Facilities," AEC Rept. ORNL-4451
(1970), pp 3-25.
17. Mountain, J. E. and J. H. Leonard, "Tritium
Production in a Pressurized Water Reactor," Dept. of
Chemical and Nuclear Engineering Rept., University
of Cincinnati (1970).
18. Locanti, J. and D. D. Malinowski, "Tritium in
Pressurized Water Reactors," in Tritium, A. A.
Moghissi and M. W. Carter, eds., Messenger Graphics,
Phoenix, (1973), p 45.
19. Bonka, H., K. Brussermann, and G. Schwarz,
"Umweltbelastung durch Radiokohlenstoff aus
kerntechnischen Anlagen," to be published.
20. Hayes, D. W. and K. W. MacMurdo, "Carbon-
14 Production by the Nuclear Industry," Health Phys.,
to be published.
21. Graves, R. H. and D. W. Lenth, Haddam Neck
Nuclear Power Station, personal communication
(1974).
22
-------
3. AIRBORNE RADIOACTIVE DISCHARGES
3.1 Gaseous Waste System and Samples
3.1.1 Gaseous waste system. The airborne
radionuclides from the reactor system are removed
principally by radioactive decay and by discharge to
the atmosphere. The airborne radionuclides are either
gaseous—fission-produced tritium, krypton, xenon, and
iodine and activation-produced tritium, carbon, and
argon—or on particles. The pathways for gaseous
waste are depicted in Figure 3.1."'10) A program to
augment the system for reducing the amounts of
radioactive waste effluents is now under way.<8)
Gases separate from reactor coolant water after
discharge through the chemical and volume control
system (CVCS) and when the reactor-vessel cover is
removed after shutdown. Haddam Neck reports that
the reactor is opened once each year for refueling/8'
during which the gases are vented to the vapor
container. Gases from the CVCS are transferred
directly through the head space in the primary drain
tank to the main waste gas system. Flow proceeds in
cascade through the head space in the first and second
boron waste storage tanks to a 570-m3 waste gas surge
sphere. A polyurethane-coated nylon cloth diaphragm
("Wiggins balloon") attached at the sphere equator
restores waste gas to the boron waste storage tanks as
liquid drains from them.
When the surge sphere is full, a portion of the
stored gas is pumped at a constant rate between 0.03
and 0.3 mVmin through a glass fiber filter (for
removing particles) to a plenum where the gas is
diluted before reaching the primary vent stack.
Analyses of stored gas by station staff indicate that its
main radioactive constituent is '33Xe, with some 83Kr
and 135Xe. From January 1968 to December 1970, 28-
to 210-m3 volumes of gas were released on eight
occasions. The average discharge was 100 m3/" No
routine releases are reported for 1971, but gas was
released in March and April of 1971 in connection with
this study to test methods for measuring this effluent in
the environment (see Section 6). Leakage, if any,
through the sphere diaphragm is exhausted
continuously to the stack duct.
Leaks in valve stems, pump shaft seals and other
system equipment allow some reactor coolant and
associated gases to enter the atmospheres of the vapor
container and the primary auxiliary building. The
station has reported that these coolant leakage rates to
the two buildings are negligible. The Environmental
Statement'8' assumed leakage rates during routine
operation of 110 kg/day (40 gal/day at a specific
gravity of 0.7) to the vapor container and 75 kg/day (20
gal/day) to the auxiliary building/8' These are also the
rates estimated for "model" PWR plants by the AEC<10)
and EPA/1"
Airborne radioactivity accumulates and decays in
the vapor container because its air is exhausted only for
refueling or major maintenance. From January 1968
through December 1971, the 63,000-m3 structure was
purged four times for repairs and twice for refueling/6'
The air is exhausted to the discharge plenum initially at
1,000 to 2,000 mVmin. When radioactivity levels
become lower, the exhaust rate is reduced; it was, for
example, 140 mVrriin to remove gases from the open
reactor vessel during the April-May 1971 refueling
operation/3'
Gases that leak from reactor coolant into the
primary auxiliary building are removed continuously
by the building ventilation system. Air is exhausted at
the rate of 570 mVmin (20,000 cfm) to the discharge
plenum. In addition, approximately 30 cc of gas from
sampling and analyzing reactor coolant are discharged
weekly through a hood to the plenum.13' These direct
discharges of reactor-system gases contain short-lived
radionuclides.
Gases from two aerated drain tanks that collect
leaking reactor coolant water from the vapor container
and from hoods are discharged through headers as the
tanks fill, and pass through a filter to the plenum. The
gas volume is negligible, an estimated 6 mVweek/3'
Small amounts of off-gas from the boron recovery
evaporators are also processed to the plenum.
When reactor coolant water leaks into the
secondary coolant system at the steam generators (see
Section 2.3.5), it is accompanied by radioactive gases.
Noncondensable gases are removed from the secondary
23
-------
Surge Sphere Diaphragm Leakage
Hydrogenofed
Vent
S 1
1
Primary
Drain
Tank
Boron
Recovery
Evaporators
1
Be
We
Stc
T(
(2)
(0.03-0.28
Waste Gas
Surge Sphere
Waste Gas
Blowers(2)
m3/mln.)
Aerated
Displaced Gas (~ 6 m3/week)
Vents
Reactor Coolant Gas
3
(30 cc once weekly)
Ventilation Air
(570 m3/min.)
Containment Purge (2000 m /mln. max, during venting)
Reactor Vapor
Container
Turbine Building Vent
(2820m3/min.
Gland Seal
Leakage
->- To Atmosphere
Off- gas
(0.5m3/min. max.)
Fan
(I000m3/mln,)
Fan
(I000m3/min.)
Outside Air
(430 m3/min.)
o
o
>*
k.
E
Figure 3.1 Gaseous Waste Disposal System
-------
coolant at the main condenser by a steam jet air ejector
(SJAE) at a rate between 0.2 and 0.5 mVmin. Off-gas is
discharged to the plenum and monitored continuously
by a Geiger-Mueller (G-M) radiation detector. This is
expected to be the major pathway for releasing short-
lived radioactive gases to the environment.
Gaseous waste discharged to the plenum is diluted
by outside air and by ventilation air exhausted from the
primary auxiliary building. A fan operating at 1,000
mVmin blows the air through a discharge duct to the
1.8-m-dia., 53-m-high cylindrical primary vent stack.
A second fan is operated for releasing gases with higher
radioactivity concentrations, such as surge sphere
contents. Radiation levels in the duct are monitored
continuously by a flow-through G-M detector. A
stream of this gas is withdrawn isokinetically at a rate
of approximately 50 liters/min and passed through two
sample collectors—a glass fiber filter followed by a
charcoal-impregnated filter. These are changed daily
and analyzed for radioactive particles and
radioiodine.(3)
Three other continuous pathways for gaseous
radionuclides from the secondary coolant to the
environment are known:
1) Air inleakage to the turbine is prevented by
passing 0.1 percent of the steam through the shaft
gland seal annulus. The steam is then condensed and
returned to the system. Noncondensable gases are
removed by an air ejector at the condenser and vented
to the atmosphere through a small stack atop the
turbine building, 23 m above ground.
2) Leaks from many small sources occur in the
turbine building. Haddam Neck has reported that
leakage to the building is negligible/8' For calculating
radionuclide discharges, the steam leakage rate was
assumed to be 9,300 kg/d, one-half the rate (1,700
Ib/hr) for the "model" 3,500-MWt PWR system.00'
The building air is discharged through a roof vent at
the rate of 2,800 mVmin (100,000 cfm),(" and also
reaches the outside through doorways and other
openings.
3) Steam generator blowdown water is pumped to a
flash tank where approximately 35 percent flashes as
steam that is discharged directly to the primary vent
stack at an average rate of 6,700 kg/day (see Section
2.1.5). The flash tank discharge is considered to be the
major pathway of radioiodine to the atmosphere.'8'
A minor source in the secondary system is the
pumping of gases from the condenser to establish a
vacuum for start-up.<10> Steam discharged to the
atmosphere at relief valves during abnormal operations
has also been mentioned as an occasional effluent.""
Radioactive gases from fuel pool water in which
used fuel elements are stored in the fuel building diffuse
into the building atmosphere. Ventilation air from the
fuel building is exhausted continuously to the primary
vent stack at a rate of 70 mVmin (2,500 cfm).
3.1.2 Radionuclide release. Radioactive gases
discharged by Haddam Neck are limited by the AEC as
follows:"2'
When averaged over any calendar
year, the release rate of radioactivity
consisting of noble gases and other
isotopes with half lives less than eight
days discharged at the plant stack shall
not exceed 3 x 10" x (MFC) curies per
second, where MFC is the value in
microcuries per cubic centimeter given in
Appendix B, Table II, Column 1 of 10
CFR 20. The maximum release rate when
averaged over any one hour shall not
exceed 10 times the yearly averaged limit.
At any time when the averaged
release rate for a week exceeds 30 percent
of the annual average limit given above,
the licensee shall make provisions for
sampling iodine-131 to assure that its
release rate averaged over any calendar
year does not exceed 66 x (MFC) curies
per second.
The values in the cited Table II are derived from
Section 20.105 of 10 CFR 20,"3' which limits the added
radiation dose to an individual in an unrestricted area
to 500 mrem/yr. The permissible limits given in Table
II, Column 1 of 10 CFR 20 have been increased by a
factor of 1000 by the AEC in consideration of
atmospheric dilution of effluents before reaching the
site exclusion boundary."' The limits of average
annual effluent concentrations and annual releases
allowed to the Haddam Neck station for individual
radionuclides are as follows:
Radionuclide
Gases
12.3 -yr
5730 -yr
1.83-h
4.48-h
10.7 -yr
1.27-h
2.80-h
11.9 -d
2.25-d
5.29-d
9.15-h
JH (as HT)
(as HTO)
HC (s)
(as CCh)
"Ar
""Kr
"Kr
"Kr
"Kr
ul"Xe
m"Xe
131Xe
'"Xe
Other fission gases,
half-lives <2 hr
Effluent
concentration
limit,
uCi/cc
4 x
2 x
1 x
1 X
4 x
1 x
3 x
2 v
2 x
4 x
3 x
3 x
1 x
3 x
Annual
release
limit,*
Ci
10" 2.1 x 10'
10" 1.1 x 10!
10' 5.3 x 10'
10' 5.3 x 10'
10s 2.1 x 10'
10' 5.3 x 10*
10'
10'
10J
10'
10'
10"
.6 x 10s
.1 x 10'
.1 x 10'
.1 x 10!
.6 x 10s
.6 x 10'
10' 5.3 x 10'
10s 1
.6 x 10'
25
-------
Particles and Radioiodines
313 -d
2.7 -yr
71.3 -d
5.26-yr
17.8 -m
50.5 -d
28.5 -yr
8.06-d
20.9 -h
2.07-yr
30.0 -yr
MMn (s & i)
"Fe (s & i)
!"Co (i)
MCo (i)
MRb (i)
"Sr (s)
"Sr (s)
1J1I (s)
'"I (s)
1MCs (i)
"7Cs (i)
1 x 10^
3 x 10s
2 x 10"
3 x 10 7
3 x 10s
3 x 10'
3 x 10'
1 x 10 '
4 x 10 '
4 x 107
5 x 107
5.3 x 102
1.6 x 104
1.1 x 103
1.6 x 102
1.6 x 10'
1.6 x 102
1.6 x 10'
5.3 x 10'
2.1 x 102
2.1 x 102
2.6 x 102
occasions just before stored gas was discharged. These
releases were conducted under the following
conditions:
*Based on a continuous stack discharge of 1.66 x 10' cc/s
(5.25xlO"cc/yr). ~~
Notes:
1. The individual limits apply in the absence of other
radionuclides; if several radionucHdes^ are present,
the sum of individual percentages of the limit may
not exceed 100.
2. s = soluble; i = insoluble.
The station has reported the following annual
airborne releases between 1967 and 1973:
Noble and Halogens
Year
1967
1968
1969
1970
1971
1972
1973
activation
gases, Ci
0.02
3.7
190
641
3251
645
31.8
and Fraction of
particles, a 3H,Ci allowable, %*
0.001
Negligible
0.0004
0.001
0.231
0.018
0.029
..
9.0
2.52
0.082
0.88
6.55
50.61
„
0.006
0.07
0.25
1.35
1.47
0.01
*Based on allowable concentrations of individual radionuclides
measured in plant effluent.
Monthly values reported by the station of volume
and radioactivity of gaseous effluent are tabulated in
Appendices B.2, B.3, and B.4. Volumes up to 104 m3 per
month reflect routine discharges and occasional stored
gas releases. Greater volumes indicate (except in
September 1971) that the vapor container was purged.
Higher releases of radioactivity are usually associated
with discharges of stored gas. Individual radionuclides
are given in Appendices B.3 and B.4, and their sources
are identified in B.4. Data are included for 17.8-min
88Rb, 20.9-h 133I, and 15.6-min 13!mXe, which, because of
their relatively short half-lives or small quantities were
not measured in this study.
3.1.3 Sample collection. Samples of gas flashed
during reactor coolant sampling were collected in
duplicate on November 20, 1970, and February 9,
1971. Aliquots were contained at atmospheric pressure
in 9-cc serum bottles sealed with rubber stoppers held
by crimped aluminum holders. Off-gas discharged
through the exhaust line from the SJAE was collected
in 1.8-liter metal bottles on eight occasions from July
1970 to April 1971.
The contents of the gas surge sphere were sampled
on July 27, 1970, and April 14, 1971, and on three
Date
(sampling
time)
Release
interval,
hrs
Discharge
rate,
mVmin
Volume,
m1
Sept. 16, 1970
(0845 hrs) 1152-1530 0.100 21.8
March 15, 1971
(1005 hrs) 1100-1340 0.142 48.7
1340-1530 0.227
April 16, 1971
(0745 hrs) 0845-1100 0.113 27.2
1100-1200 0.198
Samples were collected at atmospheric pressure in 100-
or 850-cc bottles. Aliquots were transferred to 12-cc
glass bottles with rubber stoppers or 9-cc glass bottles
sealed with rubber stoppers held by crimped aluminum
holders.
Primary vent stack effluents were sampled with an
air pump connected to a single-nozzle probe, centered
in the stack discharge duct, that is routinely used for
stack monitoring. Gaseous emissions were collected in
8.2-liter evacuated metal bottles on September 15,
1970, March 16, 1971, and April 14, 1971, during
routine discharges and at the times of stored gas
releases. Particulate emissions were sampled by means
of a Unico filter holder for 5-cm-dia MSA 1106 glass
fiber or HV-70 particulate filters. Behind the filter, a
3.2-cm-dia container for 26 g bed of activated charcoal
from Cesco type B cartridges was inserted for sampling
gaseous radioiodine. A second 26-g bed of activated
charcoal and a 62-g bed of Kl-impregnated charcoal
(Mine Safety Appliances type 85851) were placed
behind the first bed on one occasion to observe the
collection efficiency of the sampler, but no additional
information was obtained because all of the charcoal
was inadvertently combined for analysis. Flow rates
varied from 40 to 50 liters/min. Samples were obtained
July 27 to August 4, 1970, September 15-16, 1970,
March 15-16, 1971, and April 14-15, 1971. Separate
samples were obtained while gas from the waste surge
sphere was released. In addition, 3 gaseous radioiodine
samples were obtained on June 1-4, 1971, shortly after
start-up with core HI.
Vapor container atmosphere was sampled by filling
evacuated 8.2-liter containers inside the structure near
the personnel access lock. On November 20, 1970, the
ambient container temperature was 29° C and the
relative humidity was 40 percent of saturation; on
March 16, 1971, they were 35° C and 21 percent. One
liter of atmospheric moisture condensate was obtained
on November 20, 1970, and 100 cc, on March 16,1971.
The latter sample was collected from the condensate
drain. Another air sample was obtained on May 5,
26
-------
1971, when the vapor container building was open to
the outside during refueling.
Gas samples of 8.2-liter volumes were collected in
three other buildings to measure ambient radionuclide
concentrations: 1) turbine hall, on April 15, 1971, near
the ventilation air exhaust grate; 2) fuel storage
building, on February 9, 1971, on the upper level near
the spent fuel pit; and 3) primary auxiliary building, on
February 9, 1971, in the corridor outside the sampling
hood area. Samples could not be collected in
ventilation discharge ducts.
Various difficulties prevented sampling of flashed
steam generator blowdown, off-gas from the condenser
air ejector for the turbine gland seals, gas from the
boron recovery evaporators, and air vented from the
aerated liquid waste tanks.
3.2 Analysis
3.2.1 Gamma-ray spectrometry. Analytical
measurement systems and procedures were similar to
those described in Section 3.2.1 of the study at the
Yankee-Rowe station/1"' Xenon-133 values are based
on results of detector efficiency calibrations conducted
with the 133Xe standard issued by the National Bureau
of Standards in October 1973. Photon intensities for all
krypton and xenon radionuclides were taken from a
recent data summary.<15) Samples obtained on
February 9, 1971, were analyzed by NaI(Tl) and
Ge(Li) detector systems (see Fig. 3.2) within 4 to 6
hours after collection to measure short-lived
radionuclides.*
Radioiodine adsorbed on charcoal was analyzed
with the Nal(Tl) detector system. The results were
corrected for 92-percent collection efficiency."" This
has been confirmed by observations in measuring stack
gases at BWR nuclear power stations, where
efficiencies of 88 and 90 percent were observed*17'18'
although most of the 131I was in an organic form08' such
as methyl iodide.
3.2.2 Radiochemical analysis. Most samples were
analyzed for 3H, 14C, 85Kr, and radiostrontium as
described in the Yankee-Rowe report/"" with the
modifications indicated below. Samples containing low
concentrations of 85Kr were analyzed only by using
plastic scintillator spheres. Beginning with the gas
samples obtained on February 9, 1971, the tritiated
moisture fraction was separated at -76° C in a freeze
trap located at the beginning of the 3H-I4C gas analysis
train. A bubbler containing Ba(OH)2 was inserted
behind the trap to collect the I4CO2 fraction. The
remaining gases were then catalytically oxidized and
the resulting HiO and CO2 were collected in another
freeze trap and Ba(OH)2 bubbler, respectively. The
chemical forms of 14C in the non-CCh fractions and of
gaseous 3H have not been identified.
3.3 Results and Discussion
3.3.1 Radioactive gases in reactor coolant. All
krypton and xenon radioisotopes with half-lives longer
than one hour that are produced at high yields by
fission were measured in gas released from reactor-
coolant samples, as shown in Table 3.1. Tritium and
the activation products 14C and "Ar were also detected.
Tritium was found as a gas (not in water vapor) and I4C
was in a chemical form other than CCh. Argon-41,
formed by the neutron activation of argon in air, was
barely detectable, as expected because the reactor
coolant is de-aerated before operation/3'
The station reported the following noble gas
measurements/"
Date,
1970
Concentration in gas,
nCi/cc
1MXe
11!Xe
Sept. 4 50.7 3.05
Oct. 30 30. 1.61
NOT. 20 35.8 2.65
The values on Nov. 20 are comparable to those given
for 133Xe and 13!Xe in Table 3.1.
The concentrations of the measured radionuclides
would be as follows in reactor coolant water at a gas
concentration of 35 cc/kg water (see Section 2.1.2):
Concentration in reactor coolant water
based on gas-phase measurements, nCi/g water
Radionuclide
!!°Kr
"Kr
"Kr
"Kr
'""Xe
'"Xe
"5Xe
Nov. 20, 1970
1.3 x 10°
1.8 x Iff1
1.1
6.0 x 10!
Feb. 9, 1971
4.6 x 10!
2.4 x 101
6.0 x Iff1
8.8 x Iff2
2.4 x Iff2
1.8
3.0 x Iff1
Compared to the values measured directly in the water
(see the Nov. 20, 1970, sample in Table 2.1), the 133Xe
concentration is identical while the concentrations of
133mXe and 13!Xe in the gas samples are twice as high.
Radioxenon values measured in the water may be low
because the sample, not being intended for gas analysis,
was not maintained air-tight.
The concentrations of the shorter-lived xenon and
krypton radioisotopes are reasonably consistent with
the values computed in Appendix C.2. This suggests
*We thank Messrs. Christopher Nelson and Gerald Karches, formerly with the Northeastern
Radiological Health Laboratory, U. S. Public Health Service, for facilitating these analyses.
27
-------
e
i
\
vt
C
O
2
Z)
O
u
40
80
120 160 200 240
CHANNEL NO. (keV-4 x channel no.)
280
3 20
360
400
Figure 3.2 - Gamma-ray spectrum of off-gas from sampling reactor coolant, 0-1600 keV
Detector: Ge (Li) 10.4 cm2x llmm, trapezoidal.
Sample: 9cc bottle containing Ice gas; collected Feb. 9, 1971 at 0755 hour.
Count: 1O minutes on Feb. 9, 1971 (14O5-1415 hr) at Winchester, Mass.
-------
Table 3.1
Radioactive Gases Released to Stack from Sampling Reactor Coolant
Concentration, uCi/cc
Radionuclide
12.3 -yr
5730 -yr
1.83-hr
4.48-hr
10.7 -yr
76.3 -m
2.80-hr
2.25-d
5.29-d
9.15-hr
Nov. 20, 1970
3H (gas)
"C (non-CO2)t
"Ar
83mKr
85Kr
87Kr
"Kr
mmXe
133Xe
'"Xe
1.0
1.7
3.8
5.1
3.2
1.7
+ 0.1 x
± 0.4 x
NA
NA
± 0.1 x
NA
NA
+ 0.1 x
+ 0.1 x
+ 0.1
10-'
10 J
10-'
10"
10'
Feb. 9
3.8
3.2
1.1
1.3
6.8
1.7
2.5
7.0
5.1
8.5
±
+
±
+
±
±
+
±
+
±
, 1971
0.9 x
0.1 x
0.8 x
0.1
0.2 x
0.3
0.1
0.4 x
0.1 x
0.6
lO'5
10'4
10"
10"
10"
10'
Estimated annual
release,* Ci
3.5
1.4
1.6
1.9
7.7
2.4
3.6
8.7
6.0
7.3
x
x
X
X
X
X
X
X
X
X
10'"
lO'6
io-4
io-3
10'4
10°
10°
io-4
io-2
io-3
*Based on release of 30 cc of gas during weekly sampling operation for 48 weeks per year.
fThe concentration of 14C in CCh was < 1 x IO'6 uCi/cc on Nov. 20, 1970.
Notes:
1. + values indicate analytical error expressed at 2-sigma; < values are minimum
detectable levels at 3-sigma counting error.
2. NA - not analyzed.
3. 3H as water vapor was not measured.
that the concentrations of the noble gas radionuclides
that could not be measured because of their short half
lives and/or low abundances—83l"Kr, 89Kr, 13""Xe,
13!mXe, '"Xe, and '38Xe—are of the magnitude given in
Appendix C.2.
The concentrations based on the measured gas
values were, on the average, twice as high as the
concentrations computed in Appendix C.2 for the
short-lived krypton isotopes, one-half as high for the
xenon isotopes, and one-eighth as high for 85Kr. The
differences for the short-lived noble gases may be due to
the use of a mean escape rate coefficient in computing
the leakage of both krypton and xenon from the fuel,
when the escape rate is actually greater for the smaller
krypton atoms. The relatively low measured
concentration of 85Kr may indicate that the gases leave
the reactor coolant system more rapidly than inferred
from the turnover of water according to the model
discussed in Section 2.3.3. Of the measured noble gas
radionuclides, only 8SKr leaves the reactor coolant
system mainly by discharge and leakage; the other
noble gases are removed mostly by radioactive decay
within the system.
The annual discharge of gaseous radionuclides
during sampling estimated in Table 3.1 is a minute
fraction of the total discharges summarized in Section
3.3.13. The unmeasured noble gas radionuclides listed
above would add only approximately 0.001 Ci/yr to
this total, as estimated from the annual discharge of
1440 cc of gas (see footnote to Table 3.1)—i.e., the gas
in 41 kg reactor coolant water at the computed
concentrations in Appendix C.2.
The concentration of 3H in water vapor, although
not measured in these samples, is expected to be at low
but detectable levels. At a concentration of 3 pCi/g
water (see Table 2.1 and Appendix B.I) when the two
samples were collected, and a saturation water content
of 2 x 10'5 g/cc in the gas samples at room temperature,
the 3H content would be 6 x 10"5 uCi/cc. The annual
discharge, in l,440cc, is 9 x IO'8 Ci.
3.3.2 Radionuclides in the waste gas surge sphere.
All of the longer-lived gaseous radionuclides in reactor
coolant were also observed in the stored gas: 3H, UC,
85Kr, '33l"Xe, and l33Xe. Their concentrations, shown in
Table 3.2, were lower than in the reactor coolant by
factors of approximately 5 for "5Kr, 100 for l33Xe, and
1000 for '33mXe. This reduction is expected from
dilution by nitrogen purge and decay of the shorter-
lived radionuclides.
Xenon-135 was not measurable at this laboratory
(<9 x 10'" uCi/cc in the optimum sample) because of
its low initial abundance and its decay by several half-
lives before analysis. The Haddam Neck staff provided
the following measurements:
29
-------
Table 3.2
Radioactiv
ity Contents of Waste Gas Surge Sphere
Concentration, pCi/cc
Radionuclide
'H (gas)
'H (HaO)
"C (non-COi)
"C (COz)
"Kr
'""Xe
'"Xe
'"Xe
July 27, 1970
1.7+0.1x10'**
NA
7.3+0.2x10'**
NA
1.3+0.1x10''
7 +2 xlO'4
1.8+O.lxlO'1
NA
Sept. 16, 1970
5 +1 xlO"'**
NA
1.4+O.lxlO4**
NA
1.1+O.lxlO'1
2.1+0.2x10'
1.1+0.1
<9xl04
March 15, 1971
NA
NA
NA
NA
1.3+O.lxlO'1
<9xlO'
7.2+O.lxlO'1
NA
April 14, 1971
1.6+0.4x10'
<3xlO'
2.6+0.1x10'
<2xlO'°
1.2+O.lxlO'
7 +5 xlO''
3. 1 +0.1x10''
NA
April 16, 1971
<3xlO'
<2xlO''
1.1+O.lxlO4
-------
system liquid waste—according to the values given in
Appendix B.4.
The releases of radionuclides from the gas
processing systems estimated in the Environmental
Statement<8) on the basis of a model station adjusted for
operating parameters at Haddam Neck are
considerably higher for 85Kr and 133Xe:
"-Kr 13 Ci/yr n""Xe 40 Ci/yr
"Kr 480 '""Xe 3
"Kr 8 '"Xe 545
'"Xe 2
These values, however, are based on 0.25 percent
leakage of fission products from fuel rather than the
actual leakage of 0.02 percent (see Section 2.3.3)
reported by the station and a holdup time in the gas
storage tank of only 30 days.
3.3.3 Radionuclides in vapor container air. Samples
of air from the vapor container during reactor
operation showed the presence of 85Kr, 133Xe, and
133"Xe, as well as 3H and 14C in various chemical forms
(see Nov. 20, 1970, and March 16, 1971 samples in
Table 3.3). Samples of condensed water vapor from the
vapor container collected at the same time (see Table
3.3) showed some of the same long-lived radionuclides
that were observed in reactor coolant water (see Table
2.1).
The ambient concentrations of 3H in water from
vapor container air is assumed to be the value
computed from the value for condensed water vapor;
that measured directly in air is believed to be low
because non-radioactive steam is customarily injected
into the air to reduce the 3H concentration before
personnel entry into the building. The two sets of
values compare as follows:
Date 3H in air (condensed water vapor)
Nov. 20 1.14x10* ml/cc x 2.2x10' pCi/ml = 2.5xlO' uCi/cc
Mar. 16 8.3 xW* ml/cc x 1.2 uCi/ml =9.5x10' pCi/cc
Date 3H in air (direct)
Nov. 20
Mar. 16
1.7 x 10' uCi/cc
3.2 x 10' uCi/cc
The concentration of 3H in the condensed water vapor
was lower than in reactor coolant water (see Table 2.1)
but higher than in secondary coolant water (Table 2.2).
In the air sample of May 7, 1971 (after the container
had been open to the outside for 3 weeks), ambient
concentrations of the two most abundant radioactive
gases were, on the average, 1,000-fold lower than in the
two samples collected earlier, during reactor operation.
These two earlier samples were taken to represent
vapor container air at the time of shutdown.
Annual releases from venting the vapor container
atmosphere depend upon the number and duration of
reactor shutdowns for major maintenance and
refueling. Plant reports indicate that the building is
exhausted an average of once per year each for
maintenance and for refueling."' The amount of
activity discharged annually was calculated as the
average of the concentrations found on November 20,
1970 and March 16, 1971, times 2 shutdowns, times an
assumed air volume of 31,500 m3 per shutdown. To this
value was added the amount of radionuclides released
during refueling after the container is initially purged
of the accumulated radionuclides: the radionuclide
concentrations on May 7, 1971, multiplied by 8.7 x 10'
m3 (for an exhaust rate of 142 mYmin during a 43-day
period from April 19 to June 1, 1971). The total
releases of the two relatively abundant radionuclides
were, thus:
"Kr, Ci l]]Xe, Ci
accumulated radionuclides
discharged immediately after
reactor shutdown 72 78
radionuclides discharged
continuously during refueling 6 43
annual total 78 121
The annual releases of all gaseous radionuclides,
estimated by this procedure, are shown in Table 3.3.
Tritium appears to be discharged mostly during
refueling, while the other radionuclides are discharged
mostly at the time of shutdown.
For comparison, the amounts of radionuclides that
would accumulate in the containment vessel and be
discharged twice yearly were calculated. At the average
concentrations in reactor coolant water of gaseous
radionuclides (see Section 3.3.2) and radioiodine (see
Table 2.1), and the assumed leakage rate into the
containment of 110 kg/day for 330 days, the discharges
would be:
Average concentration Calculated
in reactor Calculated annual
coolant water, daily leakage, discharge,
Radionuclide uCi/g Ci Ci
3.0 x 10s
1.3 x 10°
6.6 x 10'
1.5 x 10'
2.3
2.2 x 102
5.8 x 104
1.4 x 104
The annual discharge values were calculated as were
those for the gas processing system in Section 3.3.2. In
addition, the I3II and I33I values are based on an air-
water partition factor of 0.1,(8> where that is the ratio of
the iodine in air to the iodine in air plus water in the
building."0' The discharges based on samples of
JH (gas)
I4C (non-CCM
"Kr
'""Xe
'"Xe
'"Xe
Ulw
'"I
8.4
3.5
1.8
2.1
1.4
1.8
2.3
5.2
x
x
X
V
X
X
X
10'
10'
w
Iff'
10'
102
10 '
9.2
3.8
2.0
2.3
1.5
2.0
2.5
5.7
x
x
X
X
X
X
X
X
10
10
10
10
10
10
10
10
-8
-6
-J
•J
-'
•1
-4
-4
31
-------
Table 3.3
Radioactivity in Vapor Container Atmosphere
Radionuclide
3H (gas)
3H (mo)
14C (non-CCh)
I4C (C02)
54Mn
58Co
60Co
S5Kr
89Sr
9°Sr
,3,j
mmXe
'"Xe
134Cs
U7Cs
Nov. 20, 1970
2.7+0.3x10-'
1.7+0.5xlO'7
1.9+O.lxlO'6
7.9+0.7x10-"
NA
NA
NA
8.9+O.lxlO'4
NA
NA
<2xlO'7
3.5+0.9x10''
7.7+O.lxlQ-4
NA
NA
In air, uCi/cc
March 16, 1971
1.6+0.3xlO'7
3.2+0.2xlO'7
2.5+O.lxlO'6
NA
NA
NA
NA
1.4+0.1x10°
NA
NA
NA
5 +1 xlO'6
1.7+0.1x10''
NA
NA
May 7, 1971
< 3x10''
1.6+0.2x10''
< 1x10''
4 +2 xlO'9
NA
NA
NA
7.1+0.2xlO'7
NA
NA
NA
< 1x10"
4.9+0. IxlO'6
NA
NA
In condensate
Nov. 20, 1970
NA
2.2+0. IxlO'1
4.2+0.4xlO'7
NA
1.8+0.6xlO'7
3.0+0.8xlO'7
2 +1 xlO'7
NA
< IxlO"8
8 +3. xlO'9
1.8+O.lxlO'5
NA
NA
1.6+0.5xl07
2.6+0.6xl07
, pCi/ml
March 16, 1971
NA
1.2+0.1
2.4+0.3xlO'7
NA
5 +2 xlO'7
6 +2 xlO'7
1.1+0.3x10''
NA
<3xlO'7
3.8+O.lxlO'7
1.6+0.4xlO'5
NA
NA
6 ±2 xlO'7
4 ±3 xlO'7
Estimated annual
release of gases, Ci
1.9xlO'2
1.6x10'
1.4x10'
4.0xl02
--
--
-
7.8x10'
--
-
--
3 xlO'1
1.2xl02
-
___^
Notes:
1. + values indicate analytical error expressed at 2-sigma; < values are minimum detectable
concentrations at 3-sigma.
2. NA - not analyzed
3. Ambient water vapor concentration: 11.4 g/m3 on Nov. 20
8.3 g/m' on Mar. 16
-------
containment air are higher than these calculated values
by approximately two orders of magnitude. Possibly
some of the calculational assumptions are erroneous;
for example, the leakage rate of these gases may exceed
that of the water, as suggested in Section 3.3.1.
The annual discharge of radionuclides by purging
the containment, assumed to occur four times yearly,
was estimated to be as follows in the Environmental
Statement:'8'
"Kr 15 Ci/yr B3Xe 130 Ci/yr
u"°Xe 2 ml 0.11
'"™Xe 1 '"I 0.02
These estimated values agree with the discharge values
in Table 3.3 for 133Xe, are higher for 133mXe, and lower
for85Kr.
3.3.4 Radionuclides in primary auxiliary building
air. The long-lived gases 3H (in water vapor), 85Kr, and
133Xe were observed in a single sample of air from the
primary auxiliary building (see Table 3.4). These
airborne radionuclides presumably leaked from reactor
coolant water and from liquid wastes.
Emission rates to the stack plenum, computed for
the ventilation rate given in the note to Table 3.4, yield
an annual discharge of approximately 700 Ci, almost
entirely 133Xe. The ratio of 71:1 for short-lived I33Xe to
long-lived 85Kr in the sample of building air is similar to
that in the reactor coolant gas sample of February 9,
1971 (see Table 3.1).
The amounts of gaseous radionuclides
accompanying reactor coolant water leaking into the
building at the rate of 75 kg/day (see Section 3.3.1)
would be:
Calculated
annual leakage,
Radionuclide Ci/year
3H (gas) 2.1 x 10s
14C(non-CCh) 8.6 x 10J
"Kr 4.5 x 10 '
Radionuclide
'""Xe
'33Xe
13SXe
Calculated
annual leakage,
Ci/year
6.1 x 10'
3.5 x 10'
4.4
The above are the discharges into the vapor container
associated with 75 x 330 = 2.5 x 10" kg of water per
year at the concentrations computed in Section 3.3.2. If
the unmeasured noble gas radionuclides listed in
Section 3.3.1 are at the concentrations computed in
Appendix C.2, then their amounts in 2.5 x IO4 kg of
water would total 0.8 Ci/yr. The values computed
above for 85Kr and 133Xe, however, are 20-fold less than
the annual releases estimated in Table 3.4 from the
measurement of ventilating air; hence the other
computed values may also be low.
The amount of water vapor 3H in building air would
be 8 x 103 pCi/day, if 2.6 kg of leaking reactor coolant
water flash daily into the building atmosphere (see
Section 2.1.5) at a concentration of 3 pCi/ml (see
Appendix B.I for February, 1971). This equals 3 Ci/yr,
which is within a factor of 2 of the estimated value in
Table 3.4.
The amounts of radioiodine discharged with
building air per year would be 3.1 x 10'3 Ci 131I, 7.0 x 10'3
Ci 133I, and 5.4 x 10'3 Ci 135I, if 2.5 x 104 kg of reactor
coolant water leak each year at the averages of the
radioiodine concentrations given in Table 2.1. This is
based on an iodine partition factor of 0.005<8) that takes
into account leakage of both hot and cold reactor
coolant water into the building.00'
The annual release of radioactive gases from the
auxiliary building was estimated as follows in the
Environmental Statement:""
"5Kr 7 Ci/yr
mmXe 5
l33™Xe 9
mXe 760
"!Xe 13
'"I 0.02
'"I 0.04
Sum of short-lived radiokrypton 16
Sum of short-lived radioxenon 3
Table 3.4
Gaseous Radioactivity in Primary Auxiliary Building Atmosphere,
Sample of February 9, 1971
Radionuclide
3H (gas)
3H (HzO)
MC
85Kr
133Xe
Concentration,
uCi/cc
1
3
2.
<3
.6 +
<6
.4 +
•4 ±
x 10
0.2
x 10
1.0
0.1
-9
X
-9
X
X
10'"
10"'
10'6
Emission rate,
uCi/s
<3
1.
<6
3.
2.
,5
2
3
x
x
X
X
X
lO'2
ID'1
io-2
10"
10'
Estimated annual
release,* Ci
<9
4.
<2
9.
6.
x 10"'
.3
.1
.5 x IO2
Notes:
1. Emission rates calculated for discharge rate of 9.5 x IO6 cc/s.
Annual release based on 330 operating days per year.
2. + values indicate analytical error expressed at 2-sigma; < values
are minimum detectable levels at 3-sigma.
33
-------
These estimated values of 85Kr and U3Xe agree with the
annual releases computed in Table 3.4.
3.3.5 Radionuclides discharged from secondary
coolant system at main condenser steam jet air ejector.
All radionuclides observed in reactor coolant gas
except 41Ar were measured in gas discharged from the
SJAE, as shown in Table 3.5. The concentrations of
noble gases were approximately 10,000-fold lower than
in the reactor coolant (Table 3.1) on November 20,
1970, and February 9, 1971. Radionuclide
concentrations in the 8 secondary coolant samples
varied within a factor of ten due to changes in such
factors as reactor coolant concentrations and the
leakage rate from reactor to secondary coolant.
Flow rates in the air ejector exhaust line at times of
sampling are given in the last line of Table 3.5 for
calculating radionuclide release rates during sampling.
Annual releases of each radionuclide were estimated by
averaging these release rates and then multiplying the
average by an operating period of 330 days per year. Of
the annual discharge of 1,200 Ci, 90 percent is
contributed by 133Xe.
The discharge data for off-gas at the SJAE reported
by the station for July-December 1970 (see Appendix
B.4) are of the same magnitude as the values for 133Xe
and 135Xe in Table 3.5. In addition, 0.4 Ci 41Ar was
reported discharged in one month.
Concentrations of tritium in off-gas water vapor
were computed from the tritium concentrations in
steam-generator water shown in Table 2.2. The
temperature of off-gas at the air ejector is given as 43° C
(110° F),<9) at which saturated air contains 60 g of water
vapor per cubic meter. The tritium concentrations at
this water content would be:
Calculated 3H concentration in
Sampling date off-gas water vapor, uCi/cc
July 24, 1970 8.5 x 107
Sept.15, 1970 2.0 v 10'
Nov. 20, 1970 7.2 v 10'
Mar. 15, 1971 1.3 x 10'
Apr. 14, 1971 1.3 x 10'
The single measured value on these dates—that of
March 15—is in agreement with these values (see Table
3.5). The annual release of tritiated water vapor
computed from these concentrations is 1.3 x 10"'
Ci/year, compared to 6.1 x 10~2 Ci/year on the basis of
the three measured values in Table 3.5.
The discharges of short-lived radionuclides of
krypton and xenon and of 131mXe were estimated by
assuming that these radionuclides were in the same
amounts relative to 133Xe as computed for the reactor
coolant water in Appendix C.2:
C"Xe
"3mKr
"Kr
'3'mXe
13""Xe
n'Xe
138Xe
Total
1100 Ci/yr from Table 3.5, last column)
1.6
3.2
15.6
0.4
0.6
2.1
23.5 Ci/yr
Because of radioactive decay in transit, the amounts of
short-lived gases may be less. The measured amounts of
the krypton and xenon isotopes given in Table 3.5 agree
with the amounts computed as above within a factor of
2 or better, except that 85Kr is 4-fold lower. The amount
of 41Ar was below the value of 2 Ci/yr corresponding to
the detection limit; if the concentration relative to the
other short-lived gases were similar to the values in
Table 3.1, approximately 0.7 Ci of 41Ar would be
discharged annually.
The discharge of 131I was calculated from the
average concentration in steam generator water of 3.2 x
10"7 Ci/kg (from Table 2.2), the steam flow rate of 3.5 x
10° kg/h(see Figure 2.1), and the air/water partition
factor for iodine in the model plant00' of 5 x 10"6 This
factor is composed of partition factors of 0.01 at the
steam generators and 0.0005 at the SJAE. The 131I
release, therefore, would be 3.2 x 10"7 x 3.5 x 10° x 5 x
10'6 = 5.6 x 10'6 Ci/h,or 4.4 x 10'2 Ci/yr for 330 days of
operation. The corresponding discharge values for 133I
and 135I, at respective average concentrations in steam
generator water of 3.3 x 10"7 and 1.4 x 10"7 Ci/kg (from
Table 2.2), are 4.5 x 10'2 and 1.9 x 10'2 Ci/yr.
The following annual discharges of radioactive
gases were estimated in the Environmental Statement
at the SJAE of the main condenser in the secondary
system:'8'
""Kr
""Kr
8sKr
"Kr
"Kr
13"°Xe
133mXe
1 Ci/yr
5
8
2
8
5
9
3Xe
s"Xe
5Xe
!Xe
'I
770
1
14
2
0.03
0.03
Ci/yr
For the seven radioactive noble gases listed in Table
3.5, these estimates are approximately two-fold lower
than the annual releases based on measured values.
This difference is not excessive in view of the
uncertainties of the estimate and the variability of
measured values.
The main difficulty in relating the measured
concentrations of radioactive gases to their turnover in
the secondary system is that the discharge rate at the
SJAE exceeds the inleakage rate at the steam
generators. On an annual basis, the estimated releases
of the noble gases in Table 3.5 divided by the computed
concentrations in reactor coolant given in Appendix
34
-------
Table 3.5
Radioactivity Contents of Discharge from Main Condenser Air Ejector in Secondary Coolant System
Concentration,jiCi/cc
Radionuclide
3H(gas)
3H(H2O)
14C (non-CCh)
UC (COa)
4'Ar
85°Kr
85Kr
87Kr
"Kr
133mXe
133Xe
I35Xe
Flow rate, -cc/sec
July 24, 1970
1 3.5+0.6 x 10"
2.6+0.4 x 10"
NA
NA
NA
6 +2 x ID'5
NA
NA
NA
6.2+0.1 x 10°
NA
6170
Sept. 16, 1970
<1 x 10"
NA
2.1+0.3 x 10"
NA
NA
NA
1.1+0.1 x 10"
NA
NA
1.1+0.1 x 10"
7.7+0.1 x 10'3
3.0+0.2 x 10"
6700
Nov. 20, 1970 Feb. 9, 1971
<4 x
NA
1.1+0.3 x
NA
NA
NA
1.1+0.1 x
NA
NA
3.4+0.2 x
2.4+0.1 x
1.3+0.2 x
6170
10'" 3 ±1 x 10'8
6 +1 x 1Q-8
10" 2.1+0.2 x 10"
NA
<2 x 10'5
<2 x 10'5
10'5 4.4^-0.6 x 10'5
<4 x 10'5
<5 x ID'5
10'5 1.2+0.1 x W
10'3 5.5+0.1 x 10'3
1Q-" 4.9+0.1 x 10'"
8030
Mar. 15, 1971
1.0+0.1 x 10"
1.1+0.1 x 10'6
5.3+0.2 x 10"
5+1 x 10'8
NA
NA
1.1+0.1 x 10"
NA
NA
1.9+0.2 x 10"
1.3+0.1 x 10'2
2.1+0.8 x 10"
3780
Radionuclide
3H(gas)
3H(H2O)
14C (non-COi)
14C (CC-2)
41Ar
8S-Kr
85Kr
87Kr
88Kr
133mXe
l33Xe
l35Xe
Flow rate, cc/sec
Mar. 16, 1971
1.3+0.1 x 10"
5 +3 x 10"
6.2+0.2 x 10"
NA
<4 x 10"
4.2+0.3 x 10'5
1.5+0.1 x 10"
8 +1 x 10'5
1.1+0.1 x 10"
2.5+0.2 x 10"
1.5+0.1 x 10'2
1.7+0.1 x 10"4
3530
Concentration,
Apr. 14,
NA
NA
NA
NA
NA
NA
uCi/cc
1971
8 +2 x 10'5
NA
NA
1.4+0.1 x ID"4
7.3+0.1 x 10'3
NA
4250
Apr. 16, 1971
NA
NA
NA
NA
NA
NA
7 +1 x 10'5
NA
NA
1.1+0.1 x 10"
5.2+0.1 x 10'3
9.6+0.9 x 10"
4250
Estimated annual
release,* Ci
2.0 x 10"
6.1 x 10"
4.6 x 102
5 x 10'3
<2
6.6
1.2 x 10'
1 x 10'
1.7 x 10'
1.8 x 10'
1.1 x 103
5.8 x 10'
—
*Based on the average of the emission rates in pCi/s multiplied by 330 days (2.85 x 101 s) of reactor operation per year.
Notes:
1. + values indicate analytical error expressed at 2ff ;
< values are minimum detectable concentrations at 3
-------
C.2 require an average inleakage rate of 3 x 10s kg of
reactor coolant water. This rate (900 kg/day) is several
times higher than the reported value of 75 to 150
kg/day.18> On the two occasions when gas samples
were collected both from reactor coolant and at the
SJAE, the inleakage and discharge rates compare as
follows:
Nov. 20, 1970
The annual discharges calculated for these two
radionuclides at an assumed steam leakage rate from
the secondary system of 9,300 kg/day (see Section
3.1.1) are considerably lower. If 35 percent of the
leaking steam remains as vapor (see Section 2.1.5), the
3H discharge at an average 3H concentration in steam of
Radionuclide
3H (gas)
"Cdion-CCh)
""Kr
"Kr
"Kr
"Kr
'"-Xe
mXe
13SXe
Leakage
into secondary
system, pCi/s
Discharge
at SJAE,
MCi/s
4.5xl07
7.6xlOs
1.7xl02
< 4.5x10'
6.8x10'
6.8xl02
5.7x10' 2.1x10'
1.4 1.5x10'
7.6xl02 8.0x10'
The gas inleakage rate was calculated at the gas
concentrations of Table 3.1, for 35 cc gas per kg reactor
coolant water and an average water inleakage rate of
110 kg reactor coolant per day; the discharge at the
SJAE is at the concentrations and gas flow rates given
in Table 3.5 for the two sampling dates. Even a 2-fold
higher rate of water inleakage (see computed values in
Section 2.3.5) would not bring the radionuclide flow
rates into balance. The apparently higher discharges
could be due to nonrepresentative samples or a greater
leakage rate for gases than for water.
3.3.6 Radionuclides in turbine building air. The
only radionuclides observed in a single sample of air
collected in the turbine building were 3H (in water
vapor) and 85Kr at the concentrations given in Table
3.6. During 330 days of operation per year at an air
turnover of 4.7 x 107 cc/s (100,000 cfm),(9) 150 Ci of 3H
and 43 Ci 85Kr would be discharged per year.
Feb. 9,
Leakage
into secondary
system , uCi/s
1.7x10'
1.4x10'
5.9xl02
3.1xl02
7.6xl02
1.1x10'
3.2xl02
2.3
3.8x10'
he eas
1971
Discharge
at SJAE,
uCi/s
2.4x10'
1.7xlOJ
<2. xlO'
3.5x10'
<3. xlO'
<4. xlO'
9.6x10'
4.4x10'
3.9
1.4 x ID'5 Ci/kg (from Table 2.2) would be 15 Ci/yr.
For 85Kr and all other noble gases, the annual discharge
can be taken to be 1 x 10"" of the values at the SJAE (see
Table 3.5), that being the ratio of the steam leakage rate
to the steam flow of 3.5 x 106 kg/h in the secondary
system. Hence, the 85Kr discharge would be 1.2 x 10'3
Ci/yr. The ten-fold higher measured value of 3H may
be due to incomplete mixing of air at the sampling
point, steam leakage greater than estimated, or both.
The 36,000-fold higher measured value of 85Kr suggests
that the radioactive gas was from a source other than
secondary system leakage. Additional samples should
be collected to check these values.
The annual discharges of 131I, '"I, and 135I,
computed from their average concentrations in steam
generator water (see Section 3.3.5), a steam/water
partition factor of 1.0 relative to the steam—i.e., 0.01
Table 3.6
Gaseous Radioactivity in Turbine Hall Atmosphere,
Sample of April 15, 1971
Radionuclide
3H (gas)
3H (HiO)
UC
85Kr
l!3Xe
Concentration,
uCi/cc
< 3 x 10''
1.1+0.1 x 10"'
< 8 x 10'9
3.2+0.8 x ID"'
< 3 x 10'7
Release rate,
uCi/s
<2 x 10''
5.2
<4 x 10'
1.5
<1 x 10'
Estimated annual
release, Ci
<4
1.5 x 10J
<1 x 10'
4.3 x 10'
<4 x 102
Notes:
1. Release rates calculated for a building discharge rate of 4.7 x 10' cc/s.
Annual release computed for 330 operating days per year.
2. + values indicate analytical error expressed at 2-sigma; < values are
minimum detectable levels at 3-sigma.
36
-------
relative to steam-generator water—and an annual
leakage rate of 3.1 x 106 kg, would be 0.010, 0.010, and
0.004 Ci, respectively. The estimated annual discharges
scaled down by a factor of two from the 2-fold larger
model plant1101 are similar: 0.012 Ci for ml, and 0.008 Ci
for '"I. According to the model, the release of each
noble gas radioisotope is less than 0.5 Ci/yr.
3.3.7 Radionuclides discharged from air ejector at
turbine gland seal condenser. Samples of this effluent
were not available, but should be collected to check
estimated radionuclide discharges. Calculations
suggest that 3H in water vapor may be the radionuclide
at highest concentration. At a temperature of 52° C
(125° F)(9>, for which the saturation concentration of
water vapor is 90 g/m3, and an average 3H
concentration of 1.4 x 10'2 uCi/g in steam-generator
water (from Table 2.2), the effluent gas would have a
3H concentration of 1.3 uCi/m3 For each 4,100 m3 of
gas flowing per day (100 cfm), the annual (330-day)
discharge of 3H would be 1.8 Ci. The actual flow rate of
noncondensable gases was not known, but is believed to
be of this magnitude.
The discharge of radioactive noble gases is inferred
to be 1 x 10"3 of that at the SJAE (see Table 3.5), in that
0.1 percent of the steam flows through the gland seal
system.<8> This would include 1 Ci 133Xe per year, and
considerably lesser amounts of the other noble gas
radionuclides.
The 13'I discharge was calculated from the average
concentration in steam generator water of 3.2 x 107
Ci/kg (see Section 3.3.5), the steam flow rate in the
gland seal system of 3.5 x 103 kg/h,and a partition
factor for iodine of 1 x 10"5 relative to steam-generator
water (a steam/water ratio of 0.01 in the steam
generators, 0.01 in the gland seal condenser, and 0.1 in
the gland seal air ejector).'8'"" In 330 days of operation,
this yields an annual 131I discharge of 9 x 10"5 Ci. An
equal amount of 133I and less 135I (see Section 3.3.5)
would accompany the 131I.
3.3.8 Radionuclides discharged at blowdown flash
tank vent Samples from this source also were not
available for analyzing the radioactive effluent;
however, it is discharged through the vent stack, hence
the samples described in Section 3.3.10 through 3.3.12
include this effluent. Calculations suggest that 3H in
water vapor is the radionuclide at highest
concentration, and that most of the airborne
radioiodine is released from the station via this
pathway.
The discharge of 3H in water vapor was computed
from the annual amount of 2.2 x 10* kg blowdown
steam estimated in Section 2.1.5, at an average 3H
concentration (from Table 2.2) of 1.4 x 10"' Ci/kg. This
yields an annual 3H discharge of 31 Ci.
Th£ discharge of 13II in the gaseous effluent was
computed from the total annual blowdown (water plus
steam) of 6.3 x 10' kg (see Section 2.1.5), an average 131I
concentration of 3.2 x 10'1 Ci/kg (see Section 3.3.5),
and a vapor/water partition factor of 0.05 for iodine in
the flash tank.18' Accordingly, the annual 131I discharge
would be 0.04 Ci. Discharges of 133I and 13!I, at the
concentrations given in Section 3.3.5, would be 0.04
and 0.018 Ci/yr, respectively. The estimated values for
13II and I33I in the Environmental Report are 0.20 and
0.18 Ci/yr, respectively.(8>
The concentrations of the radioactive noble gases in
blowdown steam and water are not known. If it is
assumed for an upper limit that their concentrations
are the same as in turbine steam, and that all of the
noble gases will accompany the flashing steam, the
ratio of the discharge rate of radioactive noble gases at
the blowdown vent to that at the SJAE would equal the
ratio of the blowdown rate to the steam flow rate. This
is approximately 790 kg/h divided by 3.5 x 106 kg/h =
2 x 10"4. Relative to the values in Table 3.5, the sum of
the radioactive noble gases amounts to less than 1
Ci/yr. The estimates for the model plant, adjusted for
the two-fold smaller size of the Haddam Neck station,
are less than 0.5 Ci/yr for each radioactive noble gas.(10)
3.3.9 Radionuclides in fuel building air. Only JH in
water vapor, 14C, and 85Kr were observed in a single
sample of ventilating air collected in the fuel building.
The annual releases shown in Table 3.7 were computed
on the basis of these measurements and a ventilation
rate of 70 mVmin for 365 days. These long-lived
radionuclides would be expected to be associated with
fuel stored more than 200 days since the previous
refueling; short-lived radionuclides would appear
during and immediately after refueling. If the moisture
content of the ventilating air was approximately 1 x 10"5
g/cc, the measurement in air would reflect an 3H
concentration of 6.1 x 10"8 uCi/cc divided by 1 x 10"s
g/cc = 6.1 x 10'3 uCi/g water in the fuel pool.
3.3.10 Radioactive gases discharged through the
vent stack. Radioactive gases were measured in stack
samples on three occasions during routine continuous
discharges and twice when surge sphere gases were also
being released. The measured concentrations in Table
3.8 and the calculated release rates in Table 3.9 indicate
the presence of 3H, 85Kr, and '33Xe during continuous
discharges, and larger amounts of the same
radionuclides, as well as some 14C, when gas from the
surge sphere was added. The largest fraction of the
annual discharge of 3H and 133Xe was due to continuous
sources.
37
-------
Table 3.7
Gaseous Radioactivity in Fuel Building Atmosphere,
Sample of February 9, 1971
Radionuclide
3H (gas)
3H (HiO)
I4C
85Kr
l33Xe
Concentration,
uCi/cc
< 2 x 10"'
6.1+0.2 x 10'8
9 +2 x 10''
2 +1 x 10''
< 3 x 10''
Emission rate,
uCi/s
<2 x 10'2
7.3 x 10"2
1.1 x 10"2
2.4 x 10'2
<4 x l(y'
Estimated annual
release, Ci
< 6 x 10''
2.3
3 x 10"'
8 x 10'1
<12
Notes:
1. Emission rates calculated for discharge rate of 1.2 x 10' cc/s.
Annual release based on 365 days per year.
2. + values indicate analytical error expressed at 2-sigma;
< values are minimum detectable levels at 3-sigma.
Table 3.8
Radionuclide Concentrations in Primary Vent Stack Effluents, uCi/cc
Radionuclide
3H (gas)
3H (H2O)
"C (non-CCh)
14C (CCh)
85Kr
"3Xe
Sept. 15, 1970
)
> 1.3+0.8xlO"8t
1
}
\ < 3xlO-'t
I
7 +1 xlO'9
3.0+O.lxlO'6
Sept. 16, 1970*
2.5+0.9xlO-8f
1.7+0.2xlO''t
9.0+0. IxlO'6
6.2+O.lxlO'5
Mar. 16, 1971
< 3xlO'9
4.3+0. IxlQ-1
< 3xlO'9
4.2+0.7xlQ-8
2.3+0.3xlO'6
Apr. 14, 1971
< 3xlO'9
3.9+0.4x10-'
< 3x10-"
< 4x10"'
4+1. xlQ-8
3.5+0.4x10-'
Apr. 16, 1971*
1.5+0.3x10"'
l.l+0.2xlO'8
7.0+0.4x10"'
< 2x10"'
8.7+0.1x10"'
5.1+0.2xlO'5
* Samples were obtained during release of waste surge sphere gas when stack flow
rate was 33.3 m3/s. Flow rate for other samples was 16.6 m3/s.
fGaseous 3H and I4C measurements of Sept. 15 and 16, 1970, samples include all forms.
Note:
+ values indicate analytical error expressed at 2-sigma; < values are minimum
detectable concentrations at 3-sigma counting error.
38
-------
Table 3.9
Average and Annual Estimated Radioactivity Jteleases from the Primary Vent Stack
Average release rate, uCi/s
Estimated annual release, Ci
Radionuclide
3H
UC
95Kr
133Xe
Notes:
1 . Continuous
Continuous
0.43
<0.1
0.9
53
values are averages oi
Surge sphere
plus continuous
0.85
1.4
290
1,900
P results from Table 3.8 for S
Continuous
12
< 3
26
1,500
lept. 15, 1970, and Ap
Surge sphere
0.05
0.17
35
220
iril 14, 1971;
surge sphere plus continuous values are averages of results for Sept. 16, 1970,and April 16, 1971;
the values for 3H in gas and FhO have been combined, as are the values for '4C in CCh and
non-CCh gas.
2. Average release rates are calculated from average of concentrations for continuous or stored gas
releases (Table 3.8) times stack discharge rate of 1.67 x 10' cc/s for continuous releases or 3.33 x
10' cc/s for stored gas releases.
3. Estimated annual releases for continuous discharge are given by multiplying average release rate
by 330 days (2.85 x 107 s) of reactor operation. Annual stored gas releases are given by subtracting
the continuous release rate from the release rate during discharge of stored gas and multiplying
this difference by 33.3 hrs (1.20 x 10s s), the estimated time required to release 300 m3 of waste per
year at average flow rate of 2500 cc/s.
The averages of measured stack values were used to
check some of the previously described measurements
at the sources of the effluent gases. For releases from
the surge sphere, annual discharges compare as follows:
Estimated annual discharge from surge sphere, Ci
Surge sphere samples Stack samples
Radionuclide (Table 3.2) (Table 3.9)
'H
"C
"Kr
'"Xe
0.007
0.032
29
130
0.05
0.17
35
220
For continuous releases, the four contributing sources
(see Figure 3.1) were summed: (1) SJAE (Table 3.5); (2)
primary auxiliary building air (PAB, Table 3.4); (3)
fuel building air (FB, Table 3.7); and (4) blowdown
flash tank vent (BFT, Section 3.3.8). These compare to
the continuous discharge values from stack samples in
Table 3.9 as follows:
Considerable uncertainty is introduced into the
comparison by such factors as sample collection at
different times; possibly unrepresentative samples, as of
building air; use of approximate flow rates (e.g., daily
averages); and fluctuating discharges (e.g., much of the
blowdown occurs during a few hours each day).
Specific comparisons of radionuclide discharge
rates measured in the stack with the total of those
measured on the same day in the principal pathways to
the stack are presented in Table 3.10. The pathways
include off-gas from the secondary steam condenser air
ejector (Section 3.3.5), ventilation air exhausted from
the primary auxiliary and fuel buildings (Sections 3.3.4
and 3.3.9, respectively) and waste surge sphere
discharge (Section 3.3.2). Radionuclide concentrations
Radio-
Estimated annual continuous discharge, Ci
nuclide SJAE
3H
14C
"5Kr
1MXe
0.081
0.051
12
1,100
PAB
4.3
<2
9.1
650
FB
2.3
0.3
0.8
12
BFT*
(31)
nc
(0.01)
(1)
Sum
7 to 38
0.35 to <2.4
22
1,800
Stack samples
12
<3
26
1,500
'Values were calculated, not measured; nc: not calculated, but believed
to be smaller than the SJAE value.
The values are comparable except for the 5-to-7
times higher discharge values of 3H and 14C from the
surge sphere on the basis of the stack samples.
were converted to discharge rates for the air flow rates
given in footnotes to the utilized tables.
39
-------
Table 3.10
Comparison of Gaseous Radionuclide Release Rates Measured in Plant Pathways and Stack
Plant Pathways, uCi/sec
Sampling
Radionuclide date
Secondary steam Primary auxiliary Fuel building Waste surge
condenser bldg. exhaust* exhaust* sphere discharge
Total
Primary vent
stack discharge,
uCi/s
Ratio,
pathway/stack
Routine waste release
3H (HiO) Mar. 16,
85Kr Sept. 15,
Mar. 16,
Apr. 14,
133Xe Sept. 15,
Mar. 16,
Apr. 14,
85Kr Sept. 16,
'33Xe Sept. 16,
Apr. 16,
1971
1970
1971
1971
1970
1971
1971
1970
1970
1971
2
7.4
5.3
3
5.2
5.3
3.1
7.4
5.2
2.2
X
X
X
X
X
X
X
X
X
X
10'3
Iff't
Iff1
lo-1
io't
10'
10'
Iff1
10'
10'
1.5 x 10"'
3.2 x 10''
3.2 x Iff1
3.2 x 10''
2.3 x 10'
2.3 x 10'
2.3 x 10'
Stored gas
3.2 x Iff1
2.3 x 10'
2.3 x 10'
7.3 x
2.4 x
2.4 x
2.4 x
<3x
<3x
<3x
Iff2
Iff1
ID'2
lO'2
Iff1
10-'
Iff'
2.2 x Iff'
1.1
8.7 x 10'1
6.4 x 10''
7.5 x 10'
7.6 x 10'
5.4 x 10'
7.1
1.2
7.0 x
7 x
5.1 x
3.9 x
5.7 x
Iff'
Iff'
10'
10'
10'
0.03
0.9
1.2
0.9
1.5
2.0
1.0
releases plus routine waste releases
2.4 x
<3x
<3x
Iff2 1.8 x 102
10"1 1.8 x 103
10" 8.3 x 102
1.8 x 102
1.9 x 103
8.8 x 102
3.0 x
2.1 x
1.7 x
102
103
103
0.6
0.9
0.5
* Samples of building ventilation exhaust air collected on Feb. 9, 1971.
t Samples of off-gas from secondary steam condenser air ejectors collected on Sept. 16, 1970.
-------
Most results agree within a factor of two as shown
by the ratios of pathway and stack values given in the
last column of Table 3.10. The release rate of 3H in
water vapor was 30 times higher in the stack than in the
summed pathways on March 16, 1971, but that sum
does not include flashed steam from steam generator
blowdown. This was computed to carry 3H at an
average rate of approximately 1 pCi/s (see Section
3.3.8). The higher mXe stack effluent rate on April 16,
1971, may have arisen from operations for a planned
shutdown later in the day.
3.3.11 Radioactive particles discharged through the
vent stack. Airborne particles collected in the stack
sampler contained the 9 long-lived activation and
fission products listed in Table 3.11. Iron-55 usually
was the most abundant radionuclide; only 55Fe, 58Co,
and 90Sr were detected in all samples. Radionuclide
concentrations fluctuated among samples by
approximately two orders of magnitude. Iodine-131
was found on the charcoal samplers and, in two
instances, on filters, as discussed in Section 3.3.12. The
main continuous sources of the particulate
radioactivity are unfiltered air from the primary
auxiliary and fuel buildings, and the unfiltered gases
from the SJAE and the blowdown flash tank.
Concentrations of particulate radionuclides in stack
samples taken during the three releases of waste surge
sphere gas were higher than during continuous
releases, suggesting that some radioactive particles are
discharged with surge sphere gas despite filtering. The
average concentrations during continuous release and
the incremental concentrations while the stored gas
from the surge sphere was released are shown in Table
3.12. The amounts released annually from the surge
sphere were about an order of magnitude lower than
those released continuously.
The total discharge of all long-lived radioactive
particles in stack effluents listed in Table 3.12 is 3 x 10'3
Ci per year, of which 55Fe constitutes 40 percent.
Haddam Neck staff reported short-lived 17.8-min 88Rb
as the major particulate stack emission, in that 0.2 Ci
was released in 1971 in addition to 1.8 x 10'2 Ci of other
radionuclides (see Appendix B.3). The longer-lived
radionuclides in Table 3.11 were, therefore, only a
small fraction of the particulate effluent.
3.3.12 Iodine-131 discharged through the vent
stack. Iodine-131 was found in all charcoal samplers,
indicating that it is being discharged continuously (see
Tables 3.11 and 3.12). Sample concentrations varied by
approximately two orders of magnitude. The highest
measured continuous release rate was 4 x 10'3 uCi/s, on
September 15 to 16, 1970. This value may be related to
reactor startup on the previous day.
Iodine-131 stack emissions were predominantly
gaseous, although radionuclides could be found on the
filters that preceded the charcoal during two relatively
high releases. On these two occasions, approximately 4
percent of I31I was collected on the filter.
The higher emission rates during waste surge
sphere releases than during routine continuous plant
discharge indicate that I3'I was present in the waste
surge sphere although never measurable directly (see
Section 3.3.1), as shown by the following sets of
samples:
Discharge rate, pCi/s
Date routine stored gas
Sept. 16, 1970 4.0 x 10J 2.2 x 10!
March 15, 1971 1.3 x 10'' 1.7 x 10'
April 16, 1971 8.1 x 10! 3.3 x 10'
The charcoal samplers used with large volumes of stack
effluent are more sensitive for detecting 13II than the
relatively small samples of surge sphere gas.
The annual I3'I release was 1.3 x 10'2 Ci from
continuous sources, as shown in Table 3.12, estimated
by the method described in Section 3.3.11. Iodine-133
is also emitted from the stack but, because of its short
(20.9-hr) half-life, was not measured in samples
collected during this study. The station reported a
discharge of 1.2 x 10'2 Ci 13II and 1 x 103 Ci 133I in 1971
(see Appendix B.3).
The estimated continuous discharges of I31I by the
various pathways leading into the vent stack, except for
the unknown but expectedly small contribution from
the fuel building (see Sections 3.3.9), compare as
follows:
Source
Primary auxiliary bldg.
(Section 3.3.4)
SJAE (Section 3.3.5)
Blowdown flash tank
(Section 3.3.8)
Total
Estimated annual I3'I discharge, Ci/yr
based on I3II concentration from Environmental
in secondary coolant water Statement""
0.0015
0.04
0.04
0.08
0.02
0.03
0.20
0.25
41
-------
Table 3.11
Participate Radionuclide and Gaseous Iodine-131 Concentrations in Stack Effluent, uCi/m
1970
Radionuclide
Particles on filter
.1 1 .1 -day "Mn
2.7 -yr "Fe
71 3 -day '"Co
5.26-yr *°Co
"50.5 -day "*Sr
28.5 -yr ""Sr
806-day '""I
2 07-yr '"Cs
30 -yr '"Cs
Gaseous iodine on charcoal
8.06-day 1J'It
Sample volume,
m
Radionuclide
Particles on filter
3 1 3 -day "Mn
2.7 -yr "Fe
71.3 -day "Co
5.26-yr "Co
50.5 -day "Sr
28.5 -yr "Sr
8.06-day "'I
2.07-yr IMCs
30 -yr '"Cs
Gaseous iodine on charcoal
8.06-day °'If
Sample volume,
m1
July 27-28
2 +1 xlO"
1.6 + 0.1x10'
7.5+0.3x10'
1.6 + 0.2x10'
<8 xlO"
1.4 + 0.4x10"*
<2 xlO'7
<2 xlO"
<2 xlO""
1.4+0.1x10"
65
1970
Sept. 15-16
7 +3 xlO'
5.4+0.1x10''
1.8+0.2x10"'
7 +1 xlO'7
<5 xlO""
1.7+0.4x10-"
9.3+0.1x10'
2.0+0. 1x10"'
2.9+0.5x10-'
2.4+0.1x10"'
55
July 28-29
1.9 + 0.2x10"'
2.8 + 0.1x10""
1.9 + O.lxlO'"
4.1+0.2x10'
< 1 xlO'
3.0 + 0.4x10"*
<5 xlO'
<4 xlO*
<5 xlO''
1.5 + 0.1x10"'
57
Sept. 16*
6 +3 xlO''
4.5 + 0.4x10"'
1.4 + 0.5x10''
<7 xlO"'
<4 xlO''
1.7+0.2x10-'
3.0+0.5x10''
2.3+0.5x10"'
4.3+0.5x10"'
6.6+0.1xlOJ
11
July 29-31
4.0+0.6x10'
2.8 + 0.1x10'
3.7 + 0.3x10""
6.9 + O.lxlO'
<5 xlO""
2.8 + 0.4x10'*
<4 xlO'
<9 xlO'"
<9 xlO'"
1.4 + 0.1x10''
134
March 15*
7.0+0.6x10"
8.3+0.6x10"'
4.3+0.3x10"
1.3 + 0.1x10'
<8 xlO"
5 +2 xlO'"
<3 xlO'
4.1+0.6x10-'
8.7+0.5x10'"
5.1+0.2x10'*
18
July 31-Aug. 3
1.2+0.1x10"'
1.4 + 0.1x10"'
1.1+0.1x10*
2.0+0.1x10'
<3 xlO"
1.0+0.1x10"
<7 xlO'"
<3 xlO"
<3 xlO'"
9.8 + 0.1x10''
220
1971
March 15-16
<2 xlO'
4.0+0.8x10''
1.7 + 0.2x10"'
1.0+0.2x10'
<4 xlO"
4.9+0.6x10'*
<5 xlO'"
<1 xlO'
<2 xlO'
7.9+0.6x10''
56
Aug. 3—4
1 +1 xlO''
2.7+0 1x10''
1.3+0.1x10'"
2.7 + 0.2x10''
<9 xlO"
2.5 + 0.4x10""
<3 xlO'
<3 xlO'
<3 xlO'
1.2 + 0.1x10'"
63
April 14-15
<9 xlO'"
5 +1 xlO"'
2.8 + 0.2x10''
1.4+0.1x10"'
2.7+0.5x10''
].3 + 0.4xlO'"
<2 xlO''
<7 xlO""
<8 xlO"*
4.9+0.6xlO'6
75
April 16*
1.5+0.3x10""
1.9+0.1x10'
1.5+0.8x10''
3.4+0.9X10'1
<2 xlO'7
6 +3 xlO'"
<7 xlO'"
1.0+0.5x10'
3.1+0.7x10''
1.0+0.4x10''
9
Samples obtained during release of waste surge sphere gas when stack flow rate was 33.3mVs.
For other samples, flow rate was 16.7 mVs.
Additional values: gaseous IJ'I June 1-2, 1971: 8.2+0.2 pCi/mJ, 24 m' air volume
2-3, 1971: 1.3+0.1 pCi/m], 34 m5 air volume
3^t_ 1971: 1.0+0.1 pCi/m3- 42 m5 air volume
Note: < values are minimum detectable concentrations at 3 Sigma counting error; + values are 2 sigma analytical error.
-------
Table 3.12
Summary of Stack Release Rates and Estimated Annual Releases of
Particulate and Gaseous Iodine Radionuclides
Continuous
Avg. concentration,*
Radionuclide uCi/m3
Particles on filter
54Mn
S5Fe
58Co
60Co
89Sr
'°Sr
,3,j
134Cs
U7Cs
Gaseous iodine on charcoal
,3,j
2.0 x
3.0 x
1.5 x
3.3 x
2.8 x
2.0 x
7.1 x
2.5 x
2.1 x
2.8 x
ID'7
10''
10-'
io-7
io-8
10"8
10-'
IO7
io-7
io-5
release
Avg. flow rate,
uCi/s
3.4
4.7
2.3
5.6
4.6
3.3
1.2
4.2
5.8
4.6
x
X
X
X
X
X
X
X
X
X
1Q-6
io-5
io-5
io-6
ID'7
io-7
io-5
10'6
ID'6
10"
Stored gas release
Estimated
annual
release,** Ci
9.7 x
1.3 x
6.6 x
1.6 x
1.3 x
9.4 x
3.4 x
1.2 x
1.7 x
1.3 x
10'5
10'3
ID'4
io-4
io-s
10'6
ID'4
10""
10'4
ID'2
Avg. concentration
above continuous
release," uCi/m3
3.6 x
4.2 x
1.3 x
6.6 x
<2 x
6.7 x
7.2 x
2.6 x
5.9 x
1.7 x
10""
io-5
IO"
1Q-6
ID'7
10"8
10"6
10"
IO"
io-4
Estimated
annual
release,** Ci
1.4 x
1.7 x
5.2 x
2.6 x
<8 x
2.7 x
2.9 x
1.0 x
2.4 x
6.8 x
ID'5
10"4
10-'
ID'5
10'7
io-7
io-5
io-5
ID'5
io-4
* Values below minimum detectable concentration levels in Table 3.11 were taken to be zero for averaging.
All concentrations were weighted by sample volume.
**Estimated annual releases for continuous discharge are given by multiplying average release rate by 330 days (2.85 x
IO7 s) of reactor operation. Annual stored gas releases are given by subtracting the continuous release rate from the
release rate during discharge of stored gas and multiplying this difference by 33.3 hrs (1.20 x 10s s), the estimated
time required to release 330 m3 of waste per year at average flow rate of 2500 cc/s.
-------
The far lower measured value suggests either that 13II
discharge from the blowdown flash tank is grossly
over-estimated, or that the charcoal sampler was not
effective. The latter situation could be due to the
chemical form of the iodine or to interferences, as by
steam from the relatively large nightly blowdown (see
Section 2.1.5).
3.3.13 Estimated annual radionuclide discharges.
The measured effluent values discussed in the
preceding parts of Section 3.3 provide the radioactivity
source terms for planning environmental
measurements. The total discharged radioactivity and
the associated radiation doses (discussed in Section
3.3.14)based on these measured values are as follows:
Radionuclide
3H
"C
1SmKr
!sKr
"Kr
"Kr
'"•Xe
IJ1Xe
'"Xe
'"I
Annual discharge, Ci
Haddam Neck
reports
0.88
21
3,308
225
0.012
Environmental Statement
estimate
23
510
4
24
20
,200
30
0.36
The station also reported values for 135mXe, and the
Environmental Statement also contains estimates for
83raKr, "Kr, 131mXe, 135inXe, 137Xe, 138Xe, and 133I.
Radionuclide
Gases
85"Kr
"Kr
12.3 -yr 3H (as HT)
(as HTO)
5730 -yr 14C (total)
4.48-h
10.7 -yr
1.27-h "Kr
2.80-h "Kr
2.25-d '""Xe
5.29-d U3Xe
9.15-h '"Xe
Particles and I3'I
S4Mn
sFe
313 -d
2.7 -yr
71.3 -d "Co
5.26-yr "Co
50.5 -d "Sr
28.5 -yr
8.06-d
2.07-yr
'Sr
"I
l4Cs
Estimated annual
release,* Ci
4.6 x 10''
1.6 x 10'
5.6 x 10'
6.6
1.7 x 10'
1.0 x
1.7 x
1.9 x
10'
10'
10'
2.0 x
5.8 x
103
10'
30.0 -yr IJ'Cs
1.1 x 10"
1.5 x Iff3
6.7 x 10"
1.9 x 10"
1.3 x 10'
9.7 x 10'
1.4 x 102
1.3 x 10"
1.9 x 10"
Estimated annual dose at
nearest residence, mrem
3.1 x 107
3.5 x Iff'
2.5 x 10"
3.0 x 10°
2.6 x Iff2
2.2 x 10'
2.7 x 10' ,
3.0 x Iff3
3.1 x 10'
2.5 x 102
1.5 x Iff5
6.5 x 10'
4.4 x Iff5
8.5 x 10s
3.6 x 10'
2.6 x 10"
4.1 x 10"
4.5 x 10s
5.1 x Iff'
* For 3H, HC, kryptons and xenons, the annual release represents
the sum of measured individual pathways; the annual release
of particles and 131I was computed from the sum of measured
primary vent stack discharges during continuous and stored
gas releases.
Because these release values are based on
occasional—sometimes single—measurements, they
can only approximate the total discharges. Whether
they are at all applicable was checked by comparing (1)
measurements of the same pathway at several points, as
in Section 3.3.10; (2) discharge data reported by the
station for the two semi-annual periods from July 1,
1970, to June 30, 1971 (see Appendix B.3); and (3)
discharge estimates in the Environmental Statement.<8>
The latter two are as follows:
The annual releases computed from the
measurements in this study are similar to the 131I and
133Xe values reported by the station, but 3H and 8SKr
releases are higher and 135Xe releases are lower. The 3H
and 85Kr totals from this study are associated mostly
with ventilating air discharged from the turbine and
vapor containment buildings, respectively. Additional
measurements to resolve the differences between
measured and reported values would be desirable.
The release values from this study for the
radioactive noble gases are all within a factor of four of
44
-------
the Environmental Statement estimates, and are almost
identical for the major constituent, 133Xe. Such
agreement may be fortuitous, however: on the one
hand, the leakage rate from fuel into the reactor coolant
assumed in the Environmental Statement estimate was
0.25 percent compared to the actual rate of 0.02
percent;181 on the other hand, the leakage rates of the
radioactive noble gases from the reactor coolant
calculated from the measurements reported in Section
3.3.5 are higher than the assumed water leakages. The
much higher estimate of 131I releases is mainly due to
the noncondensable component of the blowdown flash
tank, as described in Section 3.3.12. Direct
measurements of U'I in this effluent and of the
radioactive noble gases leaking from the reactor
coolant would be useful in checking the computational
model.
Annual 14C discharges at PWR stations of 11 and 6
Ci per 1,000 MWe have been estimated on the basis of
formation rates <19) and measurements/20' respectively.
When reduced 2-fold for the smaller size of the
Haddam Neck station, these are still larger than the
value of 0.56 Ci given above. The discharge measured
in liquid waste (see Section 4.3.3) adds only 0.03 Ci to
this value. Some additional gaseous 14C may have been
discharged, however, in ventilating air from the
primary auxiliary building and the turbine building, for
which measured values were <2 and < 10 Ci/yr,
respectively.
In comparable measurements of airborne effluents
at the Yankee-Rowe reactor,"4' discharges of 3H and '"C
were generally consistent with the 3-fold smaller size of
that reactor. Discharges of the radioactive noble gases
were higher by factors of 102 to 10" than at Yankee-
Rowe. These higher values suggest higher leakage rates
from the fuel into the reactor coolant at the Haddam
Neck station.
3.3.14 Estimated population radiation dose. The
annual whole-body dose to an adult at the residence
nearest the Haddam Neck station was 0.5 millirem
(mrem) from airborne effluents, according to the values
listed in Section 3.3.13. Of the total, 0.3 mrem was from
l33Xe. Less than 0.001 mrem was to specific organs
from I31I and airborne radioactive particles. The annual
dose from inhaling 131I at the maximum ratio of dose to
intake—for the 4-year-old—would be four times the
listed value, i.e., 0.0016 mrem.01' Added to the dose
values based on measurements of the indicated
radionuclides at the station should be small increments
due to other radionuclides that are expected to be
present, although they were not measured. These,
discussed in Sections 3.3.1 to 3.3.12, include 4lAr,
I31I"Xe, relatively short-lived radioactive noble gas
fission products, particulate progeny of the noble gases,
and iodine isotopes.
The annual dose was obtained for each listed
radionuclide by computing the centerline
concentration in ground-level air at the point of interest
and then converting from concentration in air to tissue
dose. To compute the centerline concentration in
ground-level air, the estimated annual discharge was
divided by 3.15 x 107 s/yr to obtain the average
discharge rate. This rate was multiplied by the relative
concentration X/Q (see Appendix D.I). The relative
concentration was computed from the value of Xu/Q
plotted as a function of distance in Appendix D.2, the
annual average wind velocity, u, and the directional
frequency given in Appendix D. 1. The curve for Xu/Q
was derived by the station operator from the Gaussian
plume model; the meteorological values for the model
and the calculations were obtained by the operator
during a 15-month measurement period before the
station was built. The conversion factors from annual
average radionuclide concentration in ground-level air
to annual dose are given in Appendix E. 1.
The dose to persons would be lower than the
computed value at a given location because no
adjustment was made for the distribution of the effluent
across wind-rose sections, for shielding, and for
occupancy factors. Other potential sources of error are
associated with the effluent, which is neither at
constant concentration nor discharged from a single
location; the variable meteorological conditions; and
the uneven terrain. These may be reflected (see
Appendix D.I) in the 2-to 11-fold lesser dispersion
factors used in the Environmental Statement and the
greater dispersion factors obtained by the operator in
tracer tests and given in the Environmental Report.
The EPA dose estimate for a model PWR station with
no waste-gas storage holdup in a river valley, reduced
by a factor of four because of the lower power level of
the Haddam Station, is 0.9 mrem/yr at the nearest
residence,"1' within a factor of two of the reported dose.
The radiation doses at other nearby locations listed
in Appendix D. 1 are all lower than the values given in
Section 3.3.13. The annual doses relative to the dose at
the nearest residence are:
Location
nearest residence 0.7 km WNW
nearby residence 0.8 km NW
nearby population group 1.2 km E
" 1.2 km SSE
fishing in canal 0.5 km SE
Annual dose relative to
Distance & unity dose at nearest
direction residence
1.00
0.82
0.57
0.32
0.15
Included above is the relative dose to persons fishing
for 500 hours (0.057 yr) on site at the coolant water
45
-------
effluent canal; doses for similarly brief occupancy
factors can be computed from the values in Appendix
D. 1 for locations at the station boundary and on or by
the Connecticut River.
3.4 References
1. "Management of Radioactive Wastes at Nuclear
Power Plants," Safety Series No. 28, International
Atomic Energy Agency, Vienna (1968).
2. Connecticut Yankee Atomic Power Company,
"Facility Description and Safety Analysis," Vol. 1 and
2, AEC Docket No. 50-213-5 and 50-213-6, Haddam
Neck, Conn. (1966).
3. Lenth, D. W. and J. Kangley, Haddam Neck
Nuclear Power Plant, personal communication, 1970
and 1971.
4. Coe, R., "Nuclear Power Plants in Operation, 5
Case Histories," Nuclear News 12, 41 (1969).
5. Connecticut Yankee Atomic Power Co.,
"Technical Data - Haddam Neck Nuclear Generating
Station," company brochure (1969).
6. Connecticut Yankee Atomic Power Co.,
"Haddam Neck Plant Monthly Operation Reports,"
Nos. 67-7 to 72-12, Haddam, Conn. (1970-1971); and
"Haddam Neck Plant Semiannual Operating
Reports," Nos. 73-1 and 73-2, Haddam,
Conn.(1967-1974).
7. Connecticut Yankee Atomic Power Co.,
"Haddam Neck Nuclear Power Plant, Environmental
Report, Operating License Stage," AEC Docket No.
50-213 (July 1972).
8. Directorate of Licensing, "Final Environmental
Statement Related to the Haddam Neck (Connecticut
Yankee) Nuclear Power Plant," AEC Docket No.
50-213(1973).
9. Graves, R., Haddam Neck Nuclear Power Plant,
personal communications, 1974.
10. Directorate of Regulatory Standards,
"Numerical Guides for Design Objectives and Limiting
Conditions for Operation to Meet the Criterion 'As
Low As Practicable' for Radioactive Material in Light-
Water-Cooled Nuclear Power Reactor Effluents,"
AECRept. WASH-1258 (July 1973).
11. Office of Radiation Programs, "Environmental
Analysis of the Uranium Fuel Cycle. Part II-Nuclear
Power Reactors," EPA Rept. EPA-520/9-73-Q03-C
(1973).
12. Connecticut Yankee Atomic Power Co.,
"Provisional Operating License DPR-14, Appendix A,
Technical Specifications" (June 30, 1967).
13. 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. (1971).
14. Kahn, B., et al., "Radiological Surveillance
Studies at a Pressurized Water Nuclear Power
Reactor," EPA Rept. RD 71-1 (1971).
15. Martin, M. J., "Radioactive Atoms -
Supplement I," AEC Rept. ORNL-4923 (November
1973).
16. Dillow, W. D., "Radioiodine Measurements of
the Stack Effluent from the CP-5 5.0-MW Heavy-water
Reactor," AEC Rept. ANL-7429 (1968).
17. Kahn, B., et al., "Radiological Surveillance
Studies at a Boiling Water Nuclear Power Reactor," U.
S. Public Health Service Rept. BRH/DER 70-1
(1970).
18. Pelletier, C. A., "Results of Independent
Measurements of Radioactivity in Process Systems and
Effluents at Boiling Water Reactors" (May, 1973),
unpublished.
19. Hayes, D. W. and K. W. MacMurdo, "Carbon-
14 Production by the Nuclear Industry," Health Phys.,
to be published.
20. Kunz, C., W. Mahoney, and T. Miller, "C-14
Gaseous Effluent from Pressurized Water Reactors,"
in Symposium on Population Exposure, J. C. Hart, R.
H. Ritchie, and B. S. Varnadore, eds., AEC Rept.
CONF-74108(1974),p. 229.
46
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4. RADIONUCLIDES IN LIQUID WASTES
4.1 Liquid Waste System
4.1.1 Waste processing"' Four categories of
radioactive liquid waste are processed at the Haddam
Neck station: hydrogen-bearing ("hydrogenated") and
air-bearing ("aerated") liquid from the reactor system,
and steam generator blowdown and leakage water from
the secondary system. The sources of these wastes are
listed in Figure 4.1; the processing systems at the time
of the study (an augmented system is being installed)'21
are shown in Figure 4.2; and the amounts of liquid
handled per year are estimated in Section 2.1.5.
The hydrogenated liquid waste is processed by the
boron recovery system. It is mostly "letdown" from the
chemical and volume control system, in amounts that
fluctuate with reactor operation. Because the
"letdown" first flows through the reactor coolant
purification demineralizer, radionuclides are removed
from the reactor coolant water by demineralizing and
filtering. Other portions of reactor coolant water,
however, pass untreated into the boron recovery system
from reactor coolant drains that collect equipment
leaks. The hydrogenated water flows into the primary
drain collecting tank and from there is pumped to the
waste storage tanks for batch processing. Radioactive
decay for 30 days during accumulation has been
assumed.<2) The liquid is then passed through a filter
(5-micron pore size) and a boric-acid-conditioned
cation demineralizer for purification, and evaporated
for boron recovery.
The two-stage boron recovery evaporator is
designed to operate normally as two units in series at
the rate of 75 liters/min (20 gpm) or, occasionally when
needed, in parallel at 150 liters/min. Each stage
contains a cyclone separator and an entrainment
eliminator. The system size is based on its ability to
store and process liquid wastes produced in the
following plant operations: (a) a refueling startup for
each core cycle; (b) a cold shutdown and restart
immediately following initial full power operation of
each core cycle; (c) one 50 percent, 60-hour load
reduction per week; (d) one hot shutdown of at least 60
hours duration every four weeks; (e) one cold shutdown
at the end of full power life and one at refueling; and (0
core stretch-out of three months duration at zero boron
concentration. The system must also be capable of
handling the dilution from two sequential hot
shutdowns, with at least 60 hours at maximum power
between each shutdown, occurring at any time over the
core life. The volume of liquid waste to be processed
under extreme conditions includes 45,000 liters from
dilution for cold shutdown to full power at the
beginning of core life and 55,000 liters from dilution for
cold shutdown to full power just before the end of core
life.
The steam from the evaporator is condensed and
collected in the distillate accumulator. From this small
tank, the liquid is pumped to the two test tanks.
Conditions in the evaporator are maintained so that
most of the boron remains in the evaporator bottoms.
These are usually recovered to reuse the boric acid;
infrequently, the bottoms are packaged as solid waste
for disposal off-site.
The aerated liquid waste contains reactor coolant
water that had leaked from the system and collected in
various sumps and drains, considerably diluted by
process water. Waste water from the reactor and fuel
cavity during refueling is also processed as aerated
waste. The liquid is collected in two small aerated drain
tanks, and then purified by pumping through a prefilter
(2 5-micron pore size) and mixed-bed demineralizer
into the same test tanks used for evaporator distillate.
Waste liquids from the reactor system are held in
the test tanks for periodic discharge. Radioactive decay
for four days during accumulation has been
assumed.<2) The liquid is sampled and analyzed for
radionuclide content before discharge, and can be
recycled for additional treatment, if necessary. For
discharge, it is pumped at a selected flow-rate by two
150-liter/min (40 gpm) pumps into the service water
discharge header, and then into the circulating coolant
water discharge canal.
Blowdown water from the four steam generators in
the secondary system is monitored for radioactivity as
it flows into the blowdown flash tank, where
approximately 35 percent of the water is vented to the
47
-------
HYDROGENATED DRAINS
Letdown to Waste Disposal
Sampling System
Volume Control Tank Drain
Pressurizer Relief Tank
Valve Stem Leak-Off
Reactor Coolant Pump Seal Leak-Off
Loop Drains
Pressurizer Drain
AERATED DRAINS (UNCONTAMINATED)
Safety Injection System Drain
Chemical Addition Tank Drain
AERATED DRAINS (CONTAMINATED)
Aerated Liquid Strainers Discharges
Boron Waste Storage Tank Moats
Reactor Containment Sump Pumps
Aerated Liquid Strainers Overflow
Steam Generator Slowdown Tank Drain
Charging Pump Drain
Purification System Drain
Boric Acid Tank Drain and Overflow
Ion Exchange Pit Sump Pump
Spent Fuel Pit Filter Drain
Reactor Coolant Filter Drain
Waste Liquid Transfer Filter Drain
'Primary
Drain
Collecting
Tank
Drains from Deminerolizers Q Ion Exchangers
Aerated
Drain
Tanks
SECONDARY SYSTEM DISCHARGES
Steam Generator Blowdown _
Secondary System
Leakage
f -N
k S
Slowdown
Flash
Tank
'
Circulating
Coolant
Water
Discharge
Figure 4.1 Sources of Liquid Waste
48
-------
(15)
Hydrogenated_
Drains
Aerated
Drains )
^- Primary^
Drain
Tank
(28,000 1.)
x^_
^-— . —
^
_^
__- ^^
x^
Taste -"Y f 1 V / [ B0r°n ] Distil
Storogp >., "t * 1 ^ v >• RpoovBry > Accum
Tanks"(2) ~ £ ' £
ate
0)
L.
o ****
.2 E
a x
L.
S O
*
* m
u 00
^ «j
-------
atmosphere as steam."' The flow consists of a
relatively large nightly blowdown plus a small
continuous blowdown (see Section 2.1.5). The liquid
waste containing the nonvolatile radionuclides is
pumped into the service water discharge header, and
from there into the circulating coolant water discharge
canal.
The relatively large volume of water leaking from
the secondary system (see Section 2.1.5) is collected in
turbine building drains. The water is assumed to be
mostly from leaking steam and condensate leakage,
primarily around pump seals, and to contain all
radionuclides except 3H at much lower concentrations
than blowdown water. The water flows directly to the
circulating coolant water discharge canal.
4.1.2 Radionuclide release. The following
radionuclides were discharged at Haddam Neck
between 1970 and 1973:(3'4)
During these two years, the average gross beta
radioactivity at the point of discharge was, therefore,
approximately 1 x 10"" pCi/ml, and the 3H
concentration, 1 x 10"5 uCi/ml. The average
concentrations apply after complete mixing in the
canal; further dilution occurs beyond the mouth of the
coolant discharge canal in the Connecticut River.
Instantaneous concentrations would have fluctuated
considerably, depending on the liquid waste being
discharged.
Concentrations of radionuclides in effluents to
unrestricted areas are limited by the AEC according to
paragraph 20.106 of 10 CFR 20. Concentrations above
background in water, averaged over 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
Amount in liquid waste, Ci/yr
Radionuclide
!H
MMn
"Co
MCo
]'I
"I
"Xe
)5Xe
37Cs
1970
7380
0.097t
3.94
0.013
0.76
0.13f
15.4
0.041
0.10
Gross beta-gamma* 6.7
Volume
(liters)
23,5x10'
1971
5830
0.40
0.80
0.81
2.1
1.0
29.9
0.16
0.62
5.7
28.7x10'
1972
5890
0.02
0.97
1.15
0.30
0.57
7.35
0.17
0.71
4.8
34.4x10'
1973
3900
0.02
0.76
0.59
0.05
0.15
1.16
0.05
0.31
3.0
26.8x10'
Limit, Ci/yr**
2 x 10'
7 x 10*
7 x 10'
2 x 10*
2 x 101
7 x 102
..
—
2 x 10'
7 x 10'
—
* Does not include 3H, '"Xe, and '3!Xe; in addition, 0.025 Ci "Fe was discharged in January - June, 1971.
"Discharge into circulating cooling water flowing at the rate of 7.3 x 10" ml/yr.
t Does not include values for the first 6 months.
The releases for 1970 and 1971 are typical of the
station, except that values were very low during the
first year of operation in 1967. For the period 1967
through 1973, annual discharges of 3H ranged from 120
to 7380 Ci, and of gross beta activity, from 0.4 to 12
Ci.(3-4)
The average concentration of radionuclides in the
circulating coolant discharge canal due to station
releases can be calculated at dilution volumes for the
years 1970 and 1971 of 6.5 x 10" liters and 7.3 x 10"
liters, respectively, as follows:'3'4'
1970: jiCi/ml = amount released (Ci/yr)xl.5xlO'
1971:uCi/ml = amount released (Ci/yr)xl.4xlO'
is 1 x 10"7 pCi/ml for an unidentified mixture
containing no ml, 2KRa, and 228Ra. Limits for
individual radionuclides are 3 x 10"3 uCi/ml for 3H, the
radionuclide at highest concentration in Haddam Neck
effluent, and 3 x 10"7 uCi/ml each for soluble 90Sr and
"'!, 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).
The limits at the coolant canal discharge for an
annual flow rate of 7.3 x 1014 ml are tabulated above
with the radionuclide discharges. The gross beta values
were less than 10 percent of the limits and the
individual radionuclides were at or below 1 percent.
50
-------
4.2 Samples and Analyses
4.2.1 Samples. The following samples* of water
from the station were provided by station staff:
11. test tank during discharge, 1 liter, collected
Sept. 15, 1970 at 0900;
12. test tank during discharge, 3.5 liters, collected
Mar. 15,1971 at 1000;
13. aerated drain tank, 3 liters, collected July 23,
1970 at 1510;
14. evaporator feed from boron waste storage
tank, 1 liter, collected Feb. 5, 1971 at 1030;
15. evaporator feed after demineralizer, 1 liter,
collected Feb. 5, 1971 at 1035;
16. evaporator distillate before test tank, 1 liter,
collected Feb. 5, 1941 at 1045;
17. boron product from evaporator (boric acid
mixture), 1 liter, collected Feb. 5, 1971 at 1100;
these evaporator bottoms were from previous
batches;
18. evaporator feed from boron waste storage
tank, 1 liter, collected May 25, 1971 at 1125;
19. evaporator feed after demineralizer and filter,
1 liter, collected May 25, 1971 at 1130;
20. evaporator distillate after first stage, 4 liters,
collected May 25, 1971 at 1145;
21. aerated drain tank water after demineralizer, 4
liters, collected May 25, 1971 at 1135.
The efficacy of waste treatment in the boron
recovery system was observed with two sets of samples.
On February 5, 1971, samples of liquid waste were
collected before treatment (#14), after passage
through the cation-exchange resin (#15), and from the
condensate after evaporation (#16). The concentrated
borate solution—the evaporator bottoms—could not
be collected, but an earlier boron batch (#17) was
sampled. On May 25, 1971, samples of liquid waste
were collected before treatment (#18), after ion
exchange and filtration (#19), and from the
condensate after evaporation in the first stage (#20).
Aliquots of liquid from the aerated drain tanks were
obtained on July 23, 1970, and May 25, 1971 (samples
#13 and 21). Samples of liquid waste just before
discharge were obtained on Sept. 15, 1970, and March
15, 1971 (samples #11 and 12). The wastes were
discharged from the test tank into the circulating
coolant water discharge canal at flow rates of 19
liters/min (5 gpiri) and 76 liters/min (20 gprn),
respectively. Six samples of steam generator blowdown
'Samples No. 1-10 are described in Section 2.1.7.
water—samples #4 to 9 listed in Section 2.1.7— were
also analyzed. No samples of water leaking from the
secondary system were obtained.
To confirm discharge values computed from
radionuclide measurements in liquid wastes, the
following large samples of river water were collected
and analyzed for radionuclide content:
Volume, Suspended Hardness,
Location Time liters solids, mg/1 mg CaCOa/l
Discharge on Sept. 15, 1970, 0905 - Sept. 16, 1130
Service water intake 0930 164 60 50
Discharge weir 1030 204 130
Canal mouth 1115 204 140
Discharge on March 15, 1971, 1020 - 2400
Service water intake 0915 148 6.3 40
Discharge weir 1000 194 7.3 —
Discharge weir 1045 201 5.4 —
Canal mouth nso 194 6.1 —
Service water intake 1245 144 4.0 —
Service water is taken from influent circulating coolant
water; the canal mouth is 1.7 km downstream from the
weir in the circulating coolant water canal. The first
and second samples on March 15 were obtained to
measure radionuclides in blowdown and leakage from
secondary system; the others, to measure the combined
radionuclides discharged from the secondary system
and reactor system test tanks.
4.2.2 Analysis of waste solutions. The liquids from
the station were analyzed spectrometrically with a
Ge(Li) gamma-ray detector. 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 samples were analyzed radiochemically
for 3H, 14C, 32P, 55Fe, "Ni, 89Sr, and 90Sr.<5) If short-lived
radionuclides such as 24Na and U5I were present, they
were not detected because of the relatively long period
between collection and analysis.
4.2.3 Analysis of circulating coolant water. The
water volumes shown in Section 4.2.1 (148-204 liters)
were collected in 210-liter drums and all but 4 liters
were passed through 5-section ion-exchange columns'"
at flow rates of approximately 100 ml/min. Each of the
columns was then separated into 3 cation-exchange
resin sections, 2 anion-exchange resin sections, and a
glass-wool filter.16' Each part was analyzed with a
Nal(Tl) gamma-ray spectrometer for 1,000-minute
counting periods. The anion-exchange resins were
recounted at weekly intervals to confirm the 131I
51
-------
measurements. Every cation-exchange resin section
was eluted with 1,200 ml 6N HC1. The elutriants were
analyzed radiochemically in sequence for strontium,
cesium, and cobalt. The columns appeared to be
sufficiently large to retain the ionic radionuclides, in
that only 5 to 10 percent of the radionuclide amounts
measured in the top cation- or anion-exchange resin
section were found on the second section. No
radionuclides could be detected on the third cation-
exchange resin section.
The remaining 4 liters of water were analyzed for
hardness (calcium plus magnesium) and some
radionuclides. Ten-mi aliquots were used to determine
tritium (see Section 5.2.1) and hardness. Most of the 4-
liter sample was acidified with 10 ml cone. HNOs and
evaporated to 45 ml. One-third of this sample was
evaporated to dryness and analyzed with a Nal(Tl)
gamma-ray spectrometer. The other 30 ml were
analyzed sequentially for radioactive strontium,
cesium, and cobalt, but radionuclide concentrations
were usually too low for comparing results with those
obtained with the cation-exchange resins. These
radionuclides were counted for 100- or 1000-min
periods with G-M detectors at a background of
approximately 1.5 counts/min.
Solids that had settled in each drum were collected
by slurrying with the 4 liters of water, and combined
with solids flushed from the glass wool filter. They were
filtered, dried at 110° C, and weighed to determine the
amount of suspended solids. These samples were then
analyzed by a Nal(Tl) gamma-ray spectrometer.
4.3 Results and Discussion
4.3.1 Radionuclides in the boron recovery system.
The fraction of radionuclides removed by successively
filtering, demineralizing, and evaporating two batches
of hydrogenated liquid waste was smaller than is
generally reported for such processes. The overall
decontamination factor (DF, defined as the
concentration in the influent stream divided by the
concentration in the effluent) for the liquid on Feb. 5,
1971, based on the concentration measurements shown
in Table 4.1, was less than 100 for all radionuclides
except radiocesium. Overall DF values were much
higher for the batch of May 25, 1971, but only the DF
values for radio-manganese and -cobalt exceeded
10,000.
In contrast, the AEC has computed removals of
nonvolatile radionuclides from waste streams with the
following DF values:"'91
Decontamination factor
Radioelement
iodine
other anions
cesium, rubidium
other cations
cation-exchange
resin
1
1
10
100
boric acid
evaporator
100
1,000
1,000
1,000
Surveys of waste treatment practices in the nuclear
industry also report claims of much higher
decontamination factors by these processes/8"1" (Note,
however, that decontamination factors for evaporators
are sometimes reported as the ratio of the concentrate
to the distillate, which may be much higher than.the
DF relative to the feed).
The DF values computed separately in Table 4.2 for
the demineralizer and the evaporator are, with a few
exceptions, distinctly different for the two batches of
waste. The DF value of unity (no decontamination) for
3H and >4C is to be expected if these radionuclides are in
the form of water and carbon dioxide, respectively. The
DF values for radio-iron, -cobalt, and -iodine at the
evaporator were also consistently low. One reason for
the differences in DF values was probably an
unpremeditated change in sampling-point selection
that resulted in including filtration with evaporation in
one batch, and with demineralization, in the other. The
two liquid wastes were also different in the content, and
possibly the chemical form, of radionuclides. The
liquid processed on Feb. 5 contained 3H at
concentrations similar to reactor coolant water. Most
other radionuclides were at lower concentrations (see
Tables 4.1 and 2.1), presumably due to the removal of
ionic radionuclides by the reactor coolant purification
demineralizer and radioactive decay during waste
accumulation. The feed solution on May 25 had a lower
3H concentration and higher values for all other
radionuclides. It was probably from reactor shut-down
for refueling. Process operating parameters also may
have been changed between the two batches.
The concentrations of radionuclides in the boron
product of Feb. 5 (see Table 4.1) refer to a batch treated
earlier, and are, therefore, only qualitative indications
of the extent to which radionuclides are recirculated to
the reactor coolant with recovered boron. In this
instance, 3H, followed by 55Fe, 58Co, and 60Co, were the
radioactive constituents at highest concentrations.
4.3.2 Radionuclides in aerated liquid waste. The
sample of July 23, 1970, contained, in very different
proportions, the radionuclides of the July 24 sample of
reactor coolant water. Concentrations of 3H, 89Sr, and
"'I, for example, were lower by approximately two
orders of magnitude, while 51Cr, 60Co, 106Ru, and 110mAg
were at considerably higher concentrations (see Table
4.3). Reduction of some of these radionuclide
52
-------
Table 4.1
Radionuclide Concentration in Boron Recovery System, pCi/ml
Nuclide
3H
"C
"P
51Cr
"Mn
55Fe
5'Fe
"Co
58Co
'°Co
89Sr
MSr
95Zr
95Nb
103Ru
106Ru
»«-Ag
124Sb
U,j
'"Cs
137Cs
140Ba
14lCe
144Ce
Evaporator Feed
From Storage
Tank (14)*
2.8 x 10'
13
<0.1
<1
2.9
7.4
0.2
0.21
15
6.4
<0.05
<0.05
<1
<1
<0.1
<1
0.47
<0.1
0.45
3.6
3.6
0.2
<0.1
<0.2
February 5, 1971
Evaporator Feed
After Demineralizer
(IS)
2.8 x 106
18
—
—
0.5
1.9
<0.2
<0.03
0.7
1.0
—
—
—
—
—
—
<0.1
—
0.30
1.3
1.3
<0.1
—
—
Evaporator Distillate Boron
Before Test Tank Product
(16) (17)
2.8 x 10s
17
0.14
1.1
0.18
0.5
—
0.14
<0.03
<0.03
.—
—
—
3.6 x 10s
3.1
0.2
150
4800
20
10
620
520
<0.05
0.19
~3
~3
~2
—
11
13
0.12
16
21
<0.1
—
—
May 25, 1971
Evaporator Feed Evaporator Feed After Evaporator Distillate
From Storage Tank Demineralizer and Filter j First Stage
(18) (19) (20)
1.6 x 10s
3.7
18
180
1400
2000
170
44
6300
2700
75
2.8
120
140
28
12
3.6
57
1400
13,000
11,000
100
39
37
1.6 x 105
0.9
NA
<0.5
~0.6
2.3
0.4
<0.04
4.4
1.0
<0.03
<0.03
<0.4
<0.1
0.2
<1
<0.1
<0.1
51
27
24
<0.1
<0.2
<0.4
1.5 x 105
1.2
<0.1
<0.5
<0.1
0.7
0.14
<0.03
0.3
0.4
<0.03
<0.03
0.4
<0.1
<0.2
<1
<0.1
<0.1
48
15
14
<0.1
<0.2
<0.4
* Numbers in parentheses refer to sample numbers in Section 2.1.7 and in Figure 4.2.
Notes:
1. '33Xe was detected in all feed and distillate samples.
2. 1 pCi/ml = ID'6 uCi/ml
3. NA - not analyzed
-------
Table 4.2
Decontamination Factors for Demineralizing,
Demineralizing plus Filtering, and Evaporating
February 15, 1971
May 25, 1971
Nuclide
3H
uc
5'Cr
54Mn
5SFe
59Fe
"Co
58Co
"Co
"Sr
90Sr
95Zr
9SNb
I03Ru
""Ru
"0mAg
124Sb
,3,j
134Cs
'"Cs
140Ba
14lCe
144Ce
Demineralizer
1
1
—
6
4
>1
>7
21
6
—
—
—
—
—
—
>5
—
2
2
3
2
.„
—
Filter plus
Evaporator
1
1
—
4
2
—
...
4
2
—
—
—
—
—
...
—
—
2
>40
>40
—
—
—
Demineralizer
plus filter
1
4
> 400
2300
900
400
>1100
1400
2700
>2500
> 90
> 300
>1400
140
> 12
> 360
> 600
30
500
500
>1000
>• 190
> 90
Evaporator,
First stage
1
1
—
>6
3
3
—
15
3
—
--
-
-
>1
-
--
-
1
2
2
—
—
-
Note:
Values computed from concentrations in Table 4.1.
A value of 1 indicates no decontamination.
concentrations before the liquid is discharged would be
expected from the demineralization treatment that
follows collection in the drain tank. The demineralized
sample of May 25, 1971, was from liquid waste
collected during refueling, and resembles the
evaporator feed in radionuclide content after
demineralization (see column 7 of Table 4.1).
4.3.3 Radionuclide discharge to circulating coolant
water. In the two samples of reactor system liquid
waste collected just before the test tanks were
discharged (see Table 4.4), radionuclide concentrations
were somewhat lower, on the average, than in reactor
coolant water (see Table 2.1). Although the
concentration of radionuclides in reactor coolant water
and liquid waste are not directly comparable because
the samples were collected at different times, 3H
concentrations suggest a several-fold dilution of reactor
coolant water; 133I and 131I concentrations indicate
radioactive decay for about 1 week; and concentrations
of radio-carbon, -manganese, -iron, -cobalt, -strontium,
and -iodine suggest relatively low DF values in
processing the reactor system liquid waste, as observed
in Section 4.3.1.
The amount of radionuclides discharged annually
from the reactor system in liquid waste was estimated
in Table 4.4 to be 8,000 Ci 3H, 20 Ci 133Xe, 4 Ci 131I, and
1.3 Ci of all others. The totals were computed by
multiplying the averages of the two sets of measured
radionuclide concentrations in test tank liquid by the
waste liquid discharge of 5 x 106 liters reported for 1971
(Appendix B.3). The discharge would be
approximately 10 percent less in 3H and 131I if lower
concentrations during refueling were taken into
account (see samples of May 25, 1971 in Tables 4.1 and
4.3). Discharges of 134Cs and U7Cs, on the other hand,
would be somewhat higher.
Although these estimated values are based on only
two sets of measurements, they are consistent with
discharge data reported by the station operator for July
- December 1970 (Appendix B.4) except that the 131I
54
-------
Table 4.3
Radionuclide Concentration in Aerated Liquid Waste, uCi/ml
Radionuclide
3H
UC
31P
MCr
"Mn
"Fe
"Fe
!8Co
MCo
"Sr
wSr
"Zr
"Nb
"*Ru
"°°Ag
124Sb
U1J
134Cs
137Cs
'"Ba
Gross alpha
Drain Tank
July 23, 1970
3.2
5.5
3.1
1.3
3.5
2.8
9.6
3.0
9.0
1.0
2.0
2.0
1.5
1.0
1.7
5.8
1.0
3.5
5.0
1.3
1.8
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
10"
io-6
10'6
10°
10°
10'
Iff'
io-2
10°
io-7
io-7
lO'7
io-5
10s
io-5
io-6
Iff*
10!
10J
10"'
ID'8
After
May
1.4
1.0
<5
<5
1.6
8.0
2.9
1.8
3.5
4.0
3.0
2.6
5.0
<4
3
1.5
3.7
4.8
4.4
<1
Demineralizing
25, 1971
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
10"
10
1 fl-
lC-
10"
10"
10'
10"
10"
10'
10-
10"
10"
10
1
6
7
7
6
1
1
S
6
8
8
7
8
6
io-7
1C'7
10'
10'
10'
10
6
5
5
7
NA
Notes:
1. Not detected (<1 x IO'6
2. NA: not analyzed
pCi/ml): "Co, '"Ru, l36Cs, '"Ce, and 1<4Ce.
value is approximately twice as high as the operator's.
The operator also reported discharging small amounts
of 135Xe. In the Environmental Statement, the estimated
liquid "radwaste" discharge has been normalized to a
total (not including 3H and gases) of 5 Ci/yr, but
contains radionuclides in very different proportions:"'
"Mo
""Tc
0.38 Ci/yr
-------
Table 4.4
Radionuclide Discharge from Reactor System Liquid Waste
Concentration in test tank, uCi/ml
Radionuclide
3H
14C
32P
"Cr
!4Mn
"Fe
"Fe
"Co
S8Co
'"Co
63Ni
""Sr
"Zr
"Nb
"-Ag
131j
'"I
133Xe
134Cs
137Cs
Notes:
1. < values
2. The folk
September 15, 1970
1.8
2.9
2
1.2
5.0
1.3
3.4
1.3
1.6
7.5
1.1
1.3
2
<2
1.1
1.8
2
4.8
1.0
9
are 3-sigma
x
x
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
10"
io-7
10"
Estimated
March 15, 1971 discharge, Ci/yr
1.
.5
9.5
<2
<5
1Q-5 2.2
10* 1.0
10"
10"
lo-4
ID'5
<2
1.0
8.1
7.6
x
x
X
X
X
X
X
X
X
10"
lO'7
lO'7
10"
10'*
io-7
10"
10"
10"
10" NA
io-7
io-7
io-7
10"
lO'3
lO'5
io-3
10"
lO'7
5
<1
1
2
1.
2,
1.
3.
5.
.1
.6
,2
.8
.7
.1
X
X
X
X
X
X
X
X
X
lO'8
lO'7
10"
io-7
io-5
IO"
io-3
ID'6
10"
8
3
5
3
1
4
8
6
4
2
6
4
5
3
3
4
1
2
1
2
x
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
IO3
io-2
io-4
io-3
10-'
10"
10°
lO'3
10"
10"
io-3
io-4
IO"4
io-3
10°
10"'
10'
io-2
ID"2
counting error.
•wing radionuclides were not
detected: ( < :
2 x 10"' uCi/ml,
approximately) "Sr, "Mo, '03Ru, U4Sb, 136Cs(
144Ce; (<1 x 10"' uCi/ml) gross alpha.
3. The annual discharge was estimated by multiplying the average of
the two radionuclide concentrations by the annual liquid waste
discharge from the reactor system of 5 x 10' ml (5 x IO6 kg—see
Section 2.1.5). The estimated discharge of each radionuclide listed
in note 2 would be < 1 x IO"3 Ci/yr.
4. NA - not analyzed.
The above estimates are similar to the values of "'I, '"I,
and I35I measured in this study; are much higher for
S8Co, "Mo, 132Te, 134Cs, and V37Cs; and provide values for
very short-lived 88Rb and 132I, which could not be
measured.
The annual discharge of radionuclides to the
Connecticut River during the period of study was taken
to be the sum of the estimated discharges in Table 4.4
and 4.5. It was assumed that water leaking from the
secondary system contained negligible amounts of
radionuclides other than 3H, and that all other
discharged liquids were nonradioactive. Of the
discharged radionuclides, the short-lived ones, 32P and
radiocesium were observed mostly in blowdown water;
all others were more in reactor-system wastes. The
estimated values for the study period during the last
half of 1970 and the first half of 1971 are reasonably
consistent with the discharges reported by the station
operator for 1970 and 1971 (see Section 4.1.2). The
amounts of 54Mn, 6°Co, 133Xe, and 137Cs are intermediate
to the two sets of annual values reported by the
operator, those for 3H, 131I, and 133I were higher, and the
amount of 58Co was lower. Compared to the Yankee-
Rowe station,'12' the same radionuclides generally were
discharged in amounts higher by one to two orders of
magnitude. Only 51Cr was discharged in higher
amounts at Yankee-Rowe.
These amounts of radionuclides in water have no
direct health implication because the Connecticut
56
-------
Table 4.5
Radionuclide Discharge from Secondary System Liquid Waste
Average concentration,
* Estimated discharge, Ci/yr
Radionuclide
3H
"C
"Na
32P
51Cr
"Mn
"Fe
59Fe
"Co
MCo
60Co
°Ni
90Sr
"Zr
"Nb
"Mo
I31I
,MI
'35I
13"Cs
13SCs
137Cs
uCi/ml
1.4 x 10'2
1 x 10''
3 x 10'6
1.2 x 10''
3 x 10'7
2.5 x 10'7
7.4 x lO"6
(1 x 10'7)tt
(3 x 10-')
3.7 x lO'6
2.0 x 10'7
(4 x 10'7)
1.6 x 10'8
4 x 10-'
(4 x 10'')
4.5 x 10''
3.2 x 10"
3.3 x 10"
1.4 x 10"
1.3 x 10'5
3.3 x 10'*
1.4 x 10'5
Blowdown f Leakage**
6 x 10' 2 x 102
6 x 10"
2 x 10'2
7 x 10°
2 x 10°
2 x 10'3
4 x 10'2
(6x1 Q-4)
(2x10")
2 x 10'2
1 x 10°
(2xlO'3)
1 x 10"
2 x 10"
(2x10")
3 x 10'2
2
2
8 x lO"
8 x 10'2
2 x ID'2
8 x ID'2
* Concentrations are averages of measured values from Table 2.2.
** Leakage from the secondary system at the estimated annual rate of 1.2 x 1010 ml.
t The average concentration multiplied by an estimated annual blowdown of 6 x 10' kg
(6 x 10' ml) for all radionuclides, except 4 x 10' kg for 3H.
ft Concentrations in parentheses are averages of computed values from Appendix C.3.
River at and below the station is not a source of public
water supply. The intake of radionuclides through
eating fish caught in these waters is discussed in Section
5.5.4.
4.3.4 Radionuclides in circulating coolant water.
The three radionuclides—3H, 131I, and 58Co—computed
to be at highest concentration in the liquid effluent
were detected on Sept. 15, 1970, in cooling canal water
at the weir and mouth, as shown in Table 4.6. The
measured concentrations, however, were inconsistent
with effluent values. Compared to the predicted totals
in Table 4.6, measured concentrations of 3H, 58Co, 60Co,
and 131I were approximately 10-fold lower. These low
values can be explained if the test tank containing
reactor system waste initially was discharged more
slowly than reported; this hypothesis is supported by
the higher 3H concentrations measured downstream in
the Connecticut River later in the day (see Section
5.2.2). Measured concentrations of 134Cs and 137Cs in the
canal, and of 131I at the station water intake before flow
from the test tank would have reached it, however,
require a source other than the test tank. One
possibility is that these radionuclides remained from
the much more voluminous nightly blowdown;
another, that they were in water leaking from the
secondary system.
Some of the 3H and 137Cs in the water and most of
the 90Sr are attributed to fallout. The following
radionuclide concentrations are reported in U. S.
surface waters:
57
-------
Table 4.6
Radionuclide Concentrations in Circulating Coolant Water in September 15, 1970, pCi/liter
Radio-
nuclide
3H
58Co
MCo
90Sr
,3,j
""Cs
137Cs
Intake
<0.6xl03
<0.1
+0.1**
<0.1
0.5
1.0
<0.1
<0.1
Measured
Weir
1.2xl03
0.2
+ 0.1
<0.1
0.6
3.5
+ 0.1
0.3
0.4
Calculated from discharge
Canal
mouth
3.7xl03
0.3
+ 0.1
<0.1
0.7
3.0
+0.1
0.4
0.3
Secondary*
system
O.OSxlO3
0.02
<0.01
<0.01
0.5
0.05
0.06
Reactorf
system
24 x 103
2.2
1.0
0.002
24
0.01
0.01
Total
24 xlO3
2.2
1.0
<0.01
24
0.06
0.07
* Concentrations in Table 2.2 on the indicated date divided by the
following dilution factors:
3H—circulating coolant water flow rate of 1.4 x 10" I/day divided
by average discharge (see Section 2.1.5) of 20 1/min (28,700
I/day) from leakage plus 1.0 1/min (0.65 x 2,300 I/day) from
continuous blowdown = 1.4 x 106/21 = 6.7 x 10"
other radionuclides—1.4 x 10' I/day divided by continuous blowdown
flow of 1.6 1/min (2,300 I/day) = 8.8 x 105
**Value after plus indicates radionuclide on suspended solids, calculated
per volume of water from amount of suspended solids.
f Concentrations on the indicated date in Table 4.4 divided by dilution
factor of 1.4 x 106 1/min 4- 19 1/min = 7.4 x 104.
Note: < values are 3-sigma counting error; the 2-sigma counting errors
were 0.6|pCi/ml for 3H and 0.1 pCi/1 for all other radionuclides.
-------
Concentration, pCi/Iiter
JH
»QCI
Sr
'"Cs
Surface water, July 1970 -
March 1971"" <200-10,300
Surface water, January -
March 1971"4' — 0.4-2.1
Lake Michigan, August -
September 1970"" 400- 900 0.4-2.3 <0.
At least the lower extremes of these ranges reflect
concentrations of fallout-related radionuclides in
surface water.
The detection of U1I in the Connecticut River at the
station water intake suggests that some of the cooling
water recirculates. River flow near the station reverses
during flood tides (see Section 5.1.2), and temperature
measurements have shown that warm water discharged
into the coolant canal reaches to and beyond the water
intake.'21
To distinguish between radionuclides from
blowdown and test tank discharges, a second set of
radionuclide measurements in effluent water,
undertaken on March 15, 1971, included collection of
samples both before and during the test tank discharge.
The former samples were expected to contain
measurable amounts of 3H and 131I from continuous
blowdown; the latter, higher concentrations of these
two radionuclides as well as detectable amounts of !5Fe
(which was not analyzed for), S8Co, 6°Co, 134Cs, and
137Cs. The results shown in Table 4.7 also indicate
radionuclides from fallout at the following levels: 3H,
<700 pCi/liter; 90Sr, 0.4-0.6 pCi/liter; and 137Cs,
<0.1-0.2 pCi/liter.
When blowdown and leakage from the secondary
system were the only known sources at the station of
radionuclides in the coolant canal water, 58Co and 131I
were measured at the weir at higher concentrations
than predicted from discharge values (see Table 4.7).
The two radionuclides were also at the station water
intake. These observations support the above-cited
possibilities that (1) either the nightly blowdown or
leakage water resulted in higher concentrations than
computed from only the continuous blowdown, or (2)
some cooling water recirculates.
Increased levels of 3H, 58Co, 131I, 134Cs, and l37Cs
were associated with the discharge of the test tank.
Except for 58Co and 60Co, the concentration increases in
canal water were consistent with predicted values. A
possible reason for the lower radiocobalt
concentrations is sorption on suspended material or
other surfaces. The presence of some radiocobalt, as
well as radioiodine, on suspended material is shown in
Table 4.6. The movement of coolant water upstream is
indicated by elevated values of 3H at the coolant water
intake, as well as the detection there of 13!I and 134Cs.
1-0.8
The above observations demonstrate the feasibility
of measuring some of the radionuclides in liquid
effluents after discharge to observe effects such as the
recirculation of water and the removal of radionuclides
from water. Improved coordinations with effluent
measurements would be needed to compare measured
with predicted concentrations. Possible sources of error
in the environmental measurements include incomplete
mixing of the waste liquid with the circulating coolant
water, deposition of radionuclides on surfaces between
discharge and sampling locations, formation of
chemical forms not retained by the collectors, and
uncertainty concerning flow rates.
4.4 References
1. Brinck, W. L., "Monitoring of Effluents from a
Nuclear Power Plant," M. S. Thesis, Dept. of Chemical
and Nuclear Engineering, University of Cincinnati
(1971).
2. Directorate of Licensing,"Final Environmental
Statement Related to the Haddam Neck (Connecticut
Yankee) Nuclear Power Plant," AEC Docket No,
50-213(1973)
3. Connecticut Yankee Atomic Power Co.,
"Haddam Neck Plant Monthly Operation Reports,"
Nos. 70-1 to 72-12, and "Haddam Neck Plant
Semiannual Operating Reports," Nos. 73-1 and 73-2,
Haddam, Conn. (1970-1974).
4. Directorate of Regulatory Operations, "Report
on Releases of Radioactivity in Effluents from Nuclear
Power Plants for 1971," AEC, Washington, D. C.
(1972).
5. Krieger, H. L. and S. Gold, "Procedures for
Radiochemical Analysis of Nuclear Reactor Aqueous
Solutions," EPA Rept. EPA-R4-73-014 (1973).
6. Krieger, H. L. and G. W. Frishkorn, "Evaluation
of Ion-Exchange Surveillance Sampler for Analyzing
Radioactive Liquid Effluents," Health Phys. 21, 591
(1971).
7. Directorate of Regulatory Standards,
"Numerical Guides for Design Objectives and Limiting
Conditions for Operation to Meet The Criterion 'As
Low As Practicable' for Radioactive Material in Light-
Water-Cooled Nuclear Power Reactor Effluents,"
AEC Rept. WASH-1258 (1973).
59
-------
Table 4.7
Radionuclide Concentrations in Circulating Coolant Water on March 15, 1971, pCi/liter
Measured during
continuous discharge
from secondary system
Radio-
nuclide
3H
58Co
60Co
MSr
131I
l34Cs
137Cs
Intake
l.OxlO3
0.2
<0.1
0.5
0.5
<0.1
0.2
Weir
1.3xl03
0.2
<0.1
0.4
2.6
+ 0.1*
<0.1
0.2
Intake
4.2xl03
0.2
<0.1
0.5
0.6
0.2
0.1
Measured during
continuous and
test-tank discharge
-
Weir
65 xlO3
0.5
<0.1
0.5
3.4
+ 0.1
0.3
0.3
Calculated from
discharge*
Canal
mouth
20 x 103
0.4
<0.1
0.6
3.9
0.2
0.3
Secondary
system
O.SxlO3
<0.01
0.01
0.01
0.6
0.01
0.01
Reactor
system
81 xlO3
0.44
0.41
0.003
0.87
0.20
0.28
*See footnotes to Table 4.6, except that dilution factor from reactor system was 1.4x10*
1/min -=-76 1/min = 1.8 x 104
-------
8. Lin, K. H., "Use of Ion Exchange for the
Treatment of Liquids in Nuclear Power Plants," AEC
Kept. ORNL-4792 (1973).
9. Goodbee, H. W., "Use of Evaporation for the
Treatment of Liquids in the Nuclear Industry," AEC
Rept.ORNL-4790(1973).
10. Leonard, J. H., T. R. Thorton, and R. K.
Mosavi, "Performance Evaluation of Radioactive
Liquid Effluent Treatment Systems," University of
Cincinnati Report (July 1973).
11. Hittman Associates, Inc., "Radioactive Waste
Management - A Survey," EPA Contract No.
68-04-0052 (1972), pp. IV-52, -76, and -78.
12. Kahn, B., et al., "Radiological Surveillance
Studies at a Pressurized Water Nuclear Power
Reactor," EPARept. RD 71-7 (1971)
13. Office of Radiation Programs, "Tritium
Surveillance System, July-December 1970," and
"Tritium Surveillance System, Jan.-March, 1971,"
Rad. Health Data Repts. 12, 111 and 384(1971).
14. Office of Water Programs, "Gross
Radioactivity in Surface Waters of the United States,"
Rad. Data Repts. 13, 361 (1972).
15. Office of Water Programs, "Radioactivity of
Lake Michigan, August-September 1970," Rad. Data
Repts. 13, 559 (1972).
61
-------
5. RADIONUCLIDES IN THE AQUATIC ENVIRONMENT
5.1 Introduction
5.1.1 Studies near Haddam Neck. The
measurements described in Section 4.3.4 showed that
detectable concentrations of radionuclides from the
Haddam Neck station were in the water of the
discharge canal and the Connecticut River near the
station. Efforts to measure effluent radionuclides in the
aquatic environment were, therefore, concentrated in
these areas. The following studies are described in
detail in Sections 5.2 to 5.7:
(1) Tritium was measured in the water of the
Connecticut River between Haddam Island
State Park and the East Haddam bridge during
two releases of radioactive liquid waste by the
station. Tritium concentrations above
background were found at the mouth of the
discharge canal and as far as 2 km downstream
and 3 km upstream. Other radionuclides in
river water are reported in Section 4.3.4.
(2) Aquatic plants, plankton, and algae from the
mouth of the discharge canal and the river
near the water intake contained several
radionuclides discharged by the station at
approximately 10" times the computed water
concentration. Aquatic plants were not found
in the discharge canal, nor algae in the river
near the station.
(3) Radionuclides discharged by the station were
detected in fish from the discharge canal, but
not in fish collected in the river 9 and 18 km
upstream from the station. No radioactivity
attributable to the station except 3H was
observed in shad that swim past the station to
spawn upriver, and none in shellfish collected
from the Connecticut River estuary at Old
Saybrook.
(4) The benthos of the discharge canal and the
Connecticut River from Haddam Island State
Park to Salmon Cove was monitored with an
underwater Nal(Tl) probe connected to a
portable gamma-ray spectrometry system.
Sediment samples, mostly bottom sand, were
collected by a diver and with dredges dropped
from a boat where the probe showed
radionuclides from the station, and these were
analyzed for photon-emitting radionuclides
and 90Sr. The radionuclides 54Mn, "Co, 58Co,
MCo, 134Cs, and 137Cs from the station were
found at the mouth of the discharge canal and
at a few nearby locations along the east bank of
the Connecticut River.
The utility's contractor for environmental
surveillance and the Connecticut Department of
Environmental Protection found gross activity
attributed to the station in water and in sediment
samples at the mouth of the discharge canal in 1970
and 1971."'2) Gross-beta activity in excess of
background had also been observed in plankton
collected from the mouth of the discharge canal, but
not in fin-fish or shellfish. No specific radionuclides
were reported but 54Mn, 58Co, and 60Co were believed to
be in some samples.<2)
For comparison, at the Indian Point 1 PWR low
levels of 24Na, S6Mn, and I3'I were observed in discharge
water with an immersed Nal(Tl) detector;'31 and 54Mn,
!8Co, 60Co, 134Cs, and 137Cs were in sediment, aquatic
vegetation, and fish below the outfall.(4) At the
Yankee-Rowe PWR, 3H was found in effluent coolant-
canal water during waste discharges; 54Mn, 58Co, and
60Co were in aquatic vegetation; and 54Mn, 60Co, 90Sr,
l25Sb, and 137Cs, in sediment.'5'
Extensive and detailed ecological studies have been
conducted on the discharge canal and the Connecticut
River since 1965 by The Essex Marine Laboratory to
determine the effect of thermal pollution on all phases
of aquatic life.'6' These studies did not include
radiological measurements, but aided considerably the
ones described in this section.
5.1.2 Connecticut River hydrology. The Haddam
Neck station is located on the Connecticut River
approximately 29 km* by river above the mouth of the
*1 mile = 1.61 km; 1 cubic foot per second (cfs) = 28.3 liters/s
63
-------
river in Long Island Sound. The station discharges its
wastes into a circulating coolant water canal that
empties into the Connecticut River 1.7 km below the
station and 1.2 km above the mouth of the Salmon
River (see Figure 5.1).
While the site experiences a semidiurnal reversal of
tidal flow, saline water from Long Island Sound only
extends to 3.2 km south of the station.'7' The results of
water analyses for stable substances are shown iri Table
5.1. The average flow of fresh water in the river is
approximately 18,000 cfs , with average monthly flows
as low as 2,000 cfs and as high as 70,000 cfs. Because of
the tide, the net minimum daily average flow is 15,000
cfs, and the tidal range in the river at the site is about 1
meter.(7)
5.2 Tritium in River Water
5.2.1 Sampling and analysis. Water was collected
to measure tritium concentrations beyond the point of
release during and after the release of waste liquid from
the test tanks (see Section 4.2.1). On Sept. 15, 1970, and
March 15, 1971, samples were taken at the locations
and times listed in Table 5.2. At each location, water
was collected from the surface in 50-ml portions
generally at three points across the river.
Table 5.1
Concentration of Stable Substances in Connecticut River Water
Concentration, ug/liter
Discharge canal at weir,
Maromas power plant,
Substance
Calcium + Magnesium
(as CaCOs)
Iron
Boron
Strontium
Barium
Zinc
Copper
Aluminum
Arsenic
Beryllium
Cadmium
Chromium
Cobalt
Lead
Manganese
Molybdenum
Nickel
Phosphorus
Silver
Vanadium
Haddam Neck, 3/15/71
4.2 x 104
98
94
53
39
21
7
<17
<43
< 1
< 9
< 4
< 9
<17
< 4
<17
< 9
<43
< 1
<17
Middletown, 3/1/71
4.5 x 104
70
89
47
29
19
5
<17
<43
< 1
< 9-
< 4
< 9
<17
< 4
<17
< 9
<43
< 1
<17
Notes:
1. We thank John F. Kopp, NERC-Cincinnati for these analyses.
2. Concentrations for the stable elements were measured by
emission spectrographic analysis, except that calcium +
magnesium were determined by complexometric titration.
3. Concentrations of Ca and Mg in the Connecticut River at Haddam were
12 and 2.0 mg/1, respectively, based on averages of 12 monthly
analyses beginning Oct. 1971. We thank W. H. Oldaker, Needham
Heights Laboratory, EPA, Boston, Mass., for these values.
4. The Maromas power plant is 17 km upstream from the Haddam Neck station.
64
-------
ON
kilometer
Rgure 5.1 Sites for Aquatic Sampling
-------
Table 5.2
Tritium Concentrations at Sampling Points in Connecticut River, pCi/ml
Sampling Point
Hlgganum L.ighl (1)*
Haddam Wand (2)
Water intake (3)
0.3 km above canal moulh (7)
Canal mouth at log harrier (4)
Mouth of Salmon River (5)
Haddam Bridge (6)
1300 hrs
NSt
1.3
NS
4.0
3.7
8.4
10.6
Sept. 15, 1970
1500 hrs
<0.6
NS
NS
5.8
5.6
3.9
10.7
1700 hrs
NS
<0.b
<0.b
1.9
2.9
6.1
5.9
1300 hrs
NS
<0.9
1.0
NS
63
I 4
NS
tviarcn 15. 1971
1500 hrs
NS
1.0
2.8
NS
63
1.2
NS
1700 hrs
NS
1.0
1.0
NS
64
2.5
NS
'Numbers in parenthesis refer In the locations in Figure 5 1.
tNS not sampled.
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 3H. For routine analysis, the minimum
detectable concentration was 0.6 pCi/ml.*
5.2.2 Results and discussion. The 3H
concentrations in the discharge canal on September 15,
1970, and on March 15, 1971, computed from analyses
of station effluents, were 24 and 81 pCi/ml,
respectively (Table 4.6 and 4.7). The measurement
results on September 15, listed in Table 5.2, were lower
than predicted at the mouth of the discharge canal, but
concentrations were higher at Haddam Bridge, after at
least some additional dilution. The pattern of
concentrations (see also Table 4.6) suggests that the
discharge rate was lower than the value given in Section
4.2.1 during part of the release period. Water from the
canal appears to have moved on the surface either up-
or down-stream in response to tidal conditions without
very much dilution. On March 15,3H concentrations in
the canal mouth were reasonably consistent with
discharge values. The 3H up- and down-stream in the
Connecticut River was much diluted, although
measurable.
The observations suggest that 3H in reactor system
liquid wastes, although at low concentrations relative
to the discharge limits, can be used as a tracer for flow
and dispersion studies. Studies for this purpose would
require much more sampling as a function of distance,
cross-section, and time than was possible in this study.
5.3 Other Radionuclides in River Water
5.3.1 Gross activity measurements. Measured
radionuclide concentrations in canal water and
Connecticut River water at the station water intake are
reported in Tables 4.6 and 4.7. Upper concentration
limits of these radionuclides in the Connecticut River
at the other locations listed in Table 5.2 can be inferred
from their measured values in Tables 4.6 and 4.7
relative to those of 3H.
The utility's contractor for environmental
surveillance and the Connecticut Department of
Environmental Protection measured the following
gross activity values in Connecticut River water:0'2'
Gross beta activity, pCi/liter
Location*
1969'"
3.8
6.8
4.0
4.0
1970'"
4.3
9.9
3.9
5.3
1971"'
3.4
3.9
4.3
4.1
Higganum Light (l)f
Discharge canal (4)
Salmon River (5)
HaddamBridge (6)
*Numbers in parentheses refer to map locations in Figure 5.1.
fContro! (background) sample relative to the Haddam Neck
plant.
The gross beta activity was usually above the
background level at the mouth of the discharge canal,
and occasionally so at the downstream locations. The
highest concentration, at location #4 in 1970, was
probably due to test tank discharges during sample
collection. The annual average gross beta activity in the
discharge canal, based on analyses in liquid waste by
the station operator (see Section 4.1.2), was 9 pCi/liter
in 1970 and 1971. Concentrations beyond the mouth of
the discharge canal would be less to the extent of
dilution in the Connecticut River. Average annual
gross alpha activities were below the minimum
*We thank R. Lieberman, EERF, ORP-EPA, for analyzing some of the samples.
66
-------
detection limit, at the 3-sigma level, of 1 pCi/liter or
less.01
5.3.2 Average radionuclide concentrations in the
discharge canal. Because concentrations of
radionuclides discharged by the plant were in many
cases near or below minimum detectable levels in canal
and river water, average concentrations in canal water
were calculated from concentrations in samples of
liquids before discharge. Below are the average
radionuclide concentrations calculated to be in the
discharge canal during the study period:
Calculated radionuclide concentration, pCi/liter
from values reported
Radionuclide by station
12.3 -yr
5730 -yr
15 -h
14.3 -d
27.7 -d
313 -d
2.7 -yr
44.6 -d
270 -d
71.3 -d
5.26-yr
28.5 -yr
65 -d
35.1 -d
66.2 -h
253 -d
8.06-d
20.9 -h
6.7 -h
2.07-yr
13 -d
30 -yr
3H
"C
"Na
3'P
"Cr
"Mn
"Fe
"Fe
"Co
^Co
"Co
"Sr
"Zr
9!Nb
"Mo
"-Ag
13II
1J3I
11SI
1MCs
1MCs
'"Cs
9,700
NR
NR
NR
NR
0.42
NR
NR
NR
3.6
0.56
NR
NR
NR
NR
NR
2.0
0.94
NR
NR
NR
0.50
from measured in-plant
samples, this study
11,000
0.04
0.03
0.01
0.007
0.1
0.6
0.01
0.008
0.6
0.3
0.0007
0.001
0.004
0.04
0.004
8
3
1
0.1
0.03
0.1
Notes: approximately 23 Ci of '"Xe and 0.10 Ci of '"Xe were
discharged annually in the water, but aeration would be
expected to expel these nuclides; "Sr was not detected
in waste liquids.
NR - Not Reported.
The average concentrations listed in the first data
column are based on discharges in 1970 and 1971 and
the total available dilution reported by the station (see
Section 4.1.2). The average concentrations listed in the
second data column are the summed annual discharges
listed in Tables 4.4 and 4.5, divided by the total dilution
volume in the canal of 7.3 x 10" liters/yr. The values in
the first data column, based on monitoring all
discharges, should be superior to those based on the
occasional samples in this study. Actual concentrations
would fluctuate about these averages because reactor
system wastes are discharged periodically.
Concentrations in the Connecticut River, after the
canal water had been thoroughly mixed in it, would be
lower, on the average, by a factor of 25.
5.4 Radionuclides in Vegetation,
Plankton, and Algae
5.4.1 Sampling and analysis. Four samples of
aquatic plants were collected from the Connecticut
River on September 15, 1970—one sample each of
Vallisneria americana and Potamogeton sp. 8 km
upstream from the mouth of the discharge canal at
location 1 (see Figure 5.1), and two samples of V.
americana collected at the boat dock near the plant
water intake (location 3, Fig. 5.1).* Samples at location
1 were considered as controls. Neither V. americana
nor Potamogeton was observed growing in the
discharge canal.
Two samples of plankton, including both phyto-
and zoo- plankton with some detritus, were collected
on September 15, 1970.* One sample was collected at
the mouth of the discharge canal by hanging a 1-m-
diameter plankton net from the log barrier for 12.2
hours. Based on a flow rate of 0.33 m/s in the canal, the
sampled water volume was estimated to be 11,500 m3.
The control sample was collected by towing a 0.5-m-
diameter plankton net near the mouth of Higganum
Creek (location 1, Figure 5.1). The volume of sampled
water, measured with a flow meter, was 75 m3. The
species of collected plankton were not identified.
Two algae samples were collected from the wooden
barrier across the mouth of the circulating coolant
discharge canal. The first was obtained on September
15, 1970; the second, on March 2, 1971. The algae
samples consisted mostly of blue-green algae (Lyngbya,
Oscillatoria, Phormidium) with some green algae
(Cosmarium) and diatoms (Nitzschia, Melosira).**
Attempts were made to locate algae upriver from the
station, but no adequate quantities of growth were
found.
Gamma-ray analyses were initially performed on
the fresh samples to measure any photon-emitting
radionuclides that might be volatile during ashing,
particularly radioiodine and radioruthenium. The
samples were then dried at 100° C and ashed at 400° C.
* We thank R. R. Massengill and associates, Essex Marine Laboratory, for collecting and identifying these
samples.
**We thank Dr. William Brungs and associates, Newtown Fish Toxicology Laboratory, EPA, for
identifying the algae samples.
67
-------
Samples were analyzed for photon emitters by
spectrometry with an 11-cm3 Ge(Li) detector (see
Figure 5.2), and with a Nal(Tl) gamma-ray
coincidence/anticoincidence system. Radiochemical
analyses were performed to measure 90Sr, 89Sr and 2P;
3H and 14C were determined by treating samples in a
combustion train, collecting thO and CCh, and
measuring the radioactivity with liquid scintillation
and gas counting techniques.! Stable strontium and
calcium were measured by an atomic absorption
spectrophotometer.
5.4.2 Results and discussion. The concentrations of
radionuclides measured in the four samples of aquatic
plants are listed in Table 5.3. The control samples
contained only 55Fe, 95Nb, 137Cs, and 144Ce from
atmospheric fallout. Both samples collected near the
station contained measurable quantities of
radionuclides discharged by the station; these included
32P, 54Mn, 58Co, '"Co, 90Sr, and 13T. The fallout
radionuclides 95Nb, 137Cs, and 144Ce in these samples
were at somewhat lower concentrations than in the
controls. Radionuclides not detected in any of these
samples at the 3-sigma detection limit of 25 pCi/kg
were "Co, 89Sr, 95Zr, and 134Cs.
The V. americana found floating at flood tide near
the boat dock contained higher radionuclide levels than
the plants that grew at the sampling location. It is
possible that the floating plants had grown nearer the
mouth of the discharge canal, were uprooted by carp or
catfish searching for food, and forced upriver by the
rising tide.00' The detection of discharged
radionuclides in plants growing upstream from the
mouth of the discharge canal is in accord with
observations of the tidal movement of water upstream,
past the station water intake, in Sections 4.3.4 and
5.2.2.
The plankton from the discharge canal contained a
number of radionuclides indicative of station wastes
that were not detected in the control sample—54Mn,
58Co, 60Co, 95Nb, 13T, and 134Cs (see Table 5.4). Observed
in both the canal and control samples were l37Cs and
144Ce from fallout; 95Nb, which may be from fallout, was
only in the canal sample. Although 14C was not
measured in the control plankton sample, the specific
activity of 14C in the canal sample is higher than the
usual specific activity of 6.1 pCi/g carbon in
contemporary samples. The plankton is exposed to
plant discharges for a relative short time—about 80
min during transit in the 1.8-km canal, hence
equilibrium may not be attained between the
radionuclide content of the plankton and the water in
the canal. The results illustrate the strong tendency of
plankton, a principal food for some species of fish, to
concentrate certain radionuclides.
Algae samples from the barrier at the mouth of the
discharge canal contained relatively high
concentrations of the following radionuclides
discharged from the station: 54Mn, S5Fe, 58Co, '"Co, 90Sr,
131I, and 134Cs (see Table 5.4). The higher levels observed
in the earlier sample may have resulted from higher
radionuclide concentrations in effluent, faster algae
growth, a possibly greater uptake of nuclides during the
higher temperatures of summer (35-40° C vs. 12-13° C
in winter),"" the age of the algae,02' or a longer period
of growth. The first sample had grown in the discharge
canal for an undetermined but probably lengthy period,
while the later sample had grown only during the 5-
month interval since the first collection. Some or all of
the S5Fe, 90Sr, 95Zr, 95Nb, 137Cs, and 144Ce may be from
fallout. In the first sample only, the 14C specific activity
exceeded the normal value. Tritium, measured only in
the later sample, was at a higher concentration than the
approximately 500 pCi/liter water of combustion
usually observed in environmental samples at the
time.03' The observed 3H level in the algae sample is
consistent with the average 3H concentration in canal
water of 10 nCi/1 calculated in Section 5.3.2.
Unlike the plants and plankton, the algae samples
were submerged in the canal water continuously during
their entire growing period. Hence, radionuclide
concentration factors (CF) were calculated for the first
algae sample based on the concentrations measured in
the sample and the average water concentrations
derived in Section 5.3.2 in this study. The CF (pCiAg
algae per pCi/kg water) for this mixture of green and
blue-green algae are:
I4C = I x 104 6°Co = 7 x 10'
"Mn = 1 x 10s I3'I = 1 x 10J
"Co = 5 x 104 134Cs = 1 x 104
s"Co = 6 x 104 131Cs = 1 x 10*
To compute these CF values the following background
concentrations for radionuclides in fallout were added
to the concentrations in water from station operation
given in Section 5.3.2: 0.1 pCi 137Cs/liter and 0.067 pCi
14C/liter. The latter is based on the assumption of 11
ppm carbon in fresh water with a specific activity of 6.1
pCi 14C/g carbon.'14'
All of these concentration factors are between 103
and 105. The three cobalt isotopes showed similar
values, as did the two cesium isotopes. The large
concentration factors may make algae an important
fWe thank E. J. Troianello, EPA, Winchester, Massachusetts, for the 14C and 3H analyses of all aquatic
samples.
68
-------
400
1200
450
1250
500
1300
550
1350
600
1400
650
1450
700
1500
750
1550
800
1600
CHANNEL NO. (1.00 keV/channel)
Figure 5.2 Gamma-ray spectrum of algae
Detector: Ge(Li), 10.4 cm2x 11 mm, trapezoidal.
Sample: 467 gm (400 cc) fresh wt., collected March 2, 1971 from log barrier at mouth of
circulating coolant discharge canal.
Count: March 10-11, 1971 (1460 min., background not subtracted); Ra and Th refer to 226Ra
and 232Th plus progeny.
-------
Table 5.3
Radionuclide (pCi/kg)* and Stable Ion (g/kg)* Concentrations in Aquatic Plants
Nuclide
"P
"Mn
"Fe
s'Co
"Co
MSr
"Nb
M,j
'31Cs
L4'Ce
K
Ca
Sr
ash wt./wet wt.
near water
V. americana\
31 +
140 +
NA
120 +
25 +
26 +
85 +
250 +
14 +
150 +
0.36
0.47
17
22
18
9
9
20
24
7
50
0.0032
0.017
intake (location
3)
V. americana
<15
49 +
600 +
72 +
<20
< 10
45 +
160 +
16 +
110 +
0.36
0.53
0.0029
0.015
8
200
9
9
90
10
26
Higganum li
Potamogeton sp.
<15
<10
370 + 60
<10
<20
< 10
120 + 40
<20
31 + 11
270 + 40
0.32
0.40
0.0021
0.010
ght (location 1)
V. americana
<10
<20
<250
<20
<30
< 10
85 + 30
<25
30 + 16
230 + 60
0.40
0.72
0.0043
0.019
*kg wet weight
tfound floating at flood tide; all other samples were growing at the sampling
Notes:
+ values are 2-sigma and < values are 3-sigma counting error.
NA - not analyzed
location
Table 5.4
Radionuclide (pCi/kg)* and Stable Ion (g/kg)* Concentration in Plankton and Algae
Nuclide
'H
"C
5JMn
"Fe
"Co
"Co
6°Co
*"Sr
"Zr
"Nb
,.„,
l3'Cs
IJ7Cs
M4Ce
K
Ca
Sr
Fe
ash wt./
wet wt.
Plankton
Canal mouth
(location 4)
NA
10.4 + 0.3(71)**
1,200 + 150
NA
<40
3,700 + 200
1,600 + 200
<30
<90
560 + 160
8,900 + 600
570 + 150
420 + 110
820 + 300
2.1
0.53
0.0044
NA
0.047
(Sept. 15, 1970)
Higganum Light
(location 1)
NA
NA
<150
NA
< 40
<120
<200
< 50
<190
<120
< 100
<150
340 + 110
1,100 + 600
2.3
0.50
0.0040
NA
0.079
Algae at canal
Sept. 15, 1970
NA
30.0 + 1.0(390)**
11,000 + 1,100
41,000 + 4,000
370 + 150
39,000 + 1,500
21,000 + 1,400
42 + 5
1,800 + 1,200
1,300 + 900
11,000 + 1,100
1,400 + 700
2,000 + 800
4,600 + 800
0.6
1.23
0.0044
15
0.123
mouth (location 4)
March 2, 1971
7,400 + 800f
6.4+ 0.5(5)**
560 + 50
7,000 + 2,000
<30
730 + 60
640 + 80
28 + 4
970 + 100
870 + 70
2,300 + 80
1,200 + 70
1,900 + 80
3,800 + 180
0.8
1.10
0.0048
11
0.22
*kg wet weight
**"C concentrations are in pCi/g carbon, and the values in parentheses are
the percent excess "C (natural concentration = 6.1 pCi/g C); algae samples contained
3.9% carbon by weight.
fConcentration in units of pCi/liter of water of combustion (measured 78%
of fresh sample weight)
NA- not analyzed
70
-------
link in the chain of radionuclide transfer from water
through fish to man. Relative ease of analysis makes
them useful indicators of radionuclides in the aqueous
environment.
5.5Radionuclides in Fish
5.5.1 Introduction. The only commercial fish in the
Connecticut River is shad in the lower 43 km (27 miles)
of the river, but the river is used extensively for sport
fishing.'7' Popular fishing locations on the river in the
vicinity of Haddam Neck are near the mouth of the
Salmon River (Figure 5.1, location 5) 1.2 km
downstream from the discharge canal, and the warm
water near the mouth of the discharge canal. At the
time of the study, fishing was not permitted beyond the
barrier at the mouth of the canal. More recently, the
station operator has provided fishermen access to the
banks of the canal.00'
The discharge canal offers an unusual habitat for
studying radionuclide uptake by fish."5'16' Certain
species of fish are attracted to and remain active in the
heated effluent canal, whose water is as much as 12.4° C
above the ambient temperature during the colder
months of the year. In the colder river water, the fish
become inactive, cease eating, and some may even
burrow in the bottom sediments when the water
temperature reaches 6-8° C. Occupancy of this "warm-
water" habitat results in an abnormal seasonal
metabolism as well as feeding and activity behavior.
Brown bullheads in a control area (a cove 8 km
downstream from the discharge canal, and 3.2 km
below any detectable heated water from the plant) all
had empty stomachs in December when the water
temperature was 2° C; a comparable sample from the
canal at this time showed that 30 percent of the fish had
fed. In March, when the control area was 4° C, 27
percent of the bullheads had fed, while 51 percent of
those in the 13° C canal water were feeding actively."6'
During the spring and early summer months when the
water temperature increases, fish begin to leave the
canal. The physical condition of the canal fish is
somewhat impaired because: (1) high population
densities in the canal result in overcrowding and
increased competition for food (an average of 30,000
fish was estimated to be in residence during the
winter),"6* (2) metabolic rates of the canal fish in winter
are higher than for fish in waters at ambient
temperature, and (3) more energy is required for fish to
maintain themselves in the canal at a flow rate of 0.3
m/s.<15> Hence, the uptake of radionuclides by these
fish may be atypical.
The main foods normally consumed by the various
fish species in the canal are:
Carp (bottom scavengers)
Juvenile - live organisms; crustaceans, larvae,
and worms
Adult - (not discrete in food selection); silt,
leaves, roots, sticks, decaying fish,
and sometimes insects.
White catfish (selective)
Juvenile - small crustaceans (copopods,
cladocearns), fly pupae, and larvae.
Adult - small fish, worms, larvae, plankton,
and some vegetation. Catfish will
feed on the bottom when food is not
available in water. They do not
actively feed in winter.
Bullhead (scavengers)
Juvenile - crustaceans (copopods,
cladocearns), insect larvae and
vegetation; they feed about 50
percent of the time in the water
column and 50 percent of the time on
the bottom.
Adult - small fish, detritus, insect larvae,
vegetation, roots, stems and decayed
material on bottom.
Although these foods were the basic diet of the fish, a
wide variety of small amounts of other matter was
observed in the stomachs of the fish.""'
Because commercial fishing for the American shad
(Alosa sapidissima) could be affected by the aqueous
discharges from the nuclear power station, the fish
have been a major subject for study by the Essex
Marine Laboratory. Shad is a marine fish that enters
the Connecticut River estuary in April, swims
upstream past the nuclear power station, and spawns
far upriver beyond Hartford, Connecticut. After
spawning, the shad return to the ocean, followed in late
summer or early autumn by the juvenile fish. The adult
shad do not eat while in the river, but the juvenile fish
will eat on their passage to the ocean. During their
migration from the mouth of the river to the station,
the shad are exposed to aqueous discharges from the
station for approximately 1 day, but are in the
immediate neighborhood of the station—say 10 km
downstream—for only approximately 6 hours.00'
5.5.2 Collection and analysis. Fresh-water fish were
collected from the discharge canal on two occasions by
trawling. At the same time, background fish were
collected from the Connecticut River at Higganum and
Middletown, about 9 and 18 km above the discharge
71
-------
canal, respectively.* The collected fish are listed in
Table 5.5. The fish species from the discharge
canal—white catfish, brown bullhead, and
carp—composed more than 95 percent of the fish
species population of the canal in winter,"7' Control
fish for the March sampling period were obtained from
the intake screens at the fossil-fueled Maromas power
plant.
Shad that had migrated past the station were
collected on two occasions about 30 km upriver at
Rockyhill, Connecticut. Control shad were taken from
the estuary.
Samples were frozen immediately after collection
and returned to the laboratory on dry ice. For analysis,
the fish were thawed, weighed, and dissected for the
following tissues that were expected to concentrate the
radionuclides of interest:
muscle - n
-------
Table 5.5
Fish Collected at Haddam Neck
Date
Dec. 1, 1970
Dec. 4, 1970
Mar. 2, 1971
Apr. 15, 1971
Apr. 26, 1971
June 1, 1972
Location
Conn. R.
discharge
Conn. R.
discharge
Conn. R.
Conn. R.
Conn. R.
Conn. R.
at Higganum
canal
at Middletown
canal
estuary
at Rockyhill
estuary
at Rockyhill
Fish Species
White catfish (Ictalurus catus)
White catfish (Ictalurus catus)
Carp (Cyprinus carpio)
Crappie (Pomoxis sp.)
Yellow perch (Perca flavescens)
Brown bullhead (Ictalurus nebulosus)
White catfish (Ictalurus catus)
Carp (Cyprinus carpio)
Brown bullhead (Ictalurus nebulosus)
Shad (Alosa sapidissima)
Shad (Alosa sapidissima)
Shad (Alosa sapidissima)
Shad (Alosa sapidissima)
Number
8
10
6
3
4
2
4
2
5
2
2
2
2
Average wt., g
500
850
300
250
850
5,300
600
1,900
1,500
3,000
1,600
-------
Table 5.6
Radionuclide (pCi/kg)* and Stable Ion (gAg)* Concentrations in Fresh Water Fish
Sample Nuclide
Bone Ash wt./wet wt.
90Sr
Sr
Ca
Muscle Ash wt./wet wt.
3H
HC(pCi/g C)**
'"Sr
134Cs
137Cs
Sr
Ca
K
Fe
Liver plus
kidney 3H
"C(pCi/g C)
55Fe
,3,jt
Fe
Gut 58Co
<0Co
"'If
134Cs
137Cs
Carp
0.17
840 + 30
0.083
51
0.019
NA
NA
29 + 2
140 + 8
320 + 8
0.0029
2.13
3.76
NA
1300 + 400
12.5 + 0.3
NA
1800 + 130
NA
290 + 60
140 + 10
<60
200 + 50
380 + 60
December 1-4,
Catfish
0.15
610 + 4U
0.090
50
0.018
NA
NA
34 + 2
57 + 6
190 + 6
0.0032
2.41
3.46
NA
700 + 400
8.0 + 0.1
NA
<30
NA
<50
<40
<80
110+20
280 + 50
1970
Catfish
Background
0.14
620 + 40
0.091
51
0.015
NA
NA
15 + 2
<10
120 + 8
0.0017
0.92
3.76
NA
NA
NA
NA
<50
NA
<70
<30
<180
<50
74 + 10
March 2, 1971
Carp
0.12
440 + 40
0.059
45
0.013
4500 + 400
7.0 + 0.3
3.5 + 0.4
<30
100 + 10
0.00057
0.45
3.72
0.021
NA
NA
3500 + 400
460 + 20
0.194
<20
<20
220 + 30
<20
100 + 50
Catfish
0.13
460 + 40
0.091
47
0.019
2400 + 300
5.9 + 0.4
30 + 1
<30
150 + 10
0.0047
2.97
3.70
0.0093
NA
NA
600 + 200
200 + 20
0.091
<30
60 + 16
670 + 60
80 + 18
190 + 20
Bullhead
0.13
650 + 30
0.099
49
0.015
5700 + 400
8.7 + 0.3
20 + 1
43 + 10
190 + 10
0.0021
1.50
3.60
0.0073
NA
NA
1300 + 200
250 + 20
0.094
<30
80 + 10
1500 + 40
130 + 20
260 + 10
Mixed Fish
Background
0.15
580 +
0.105
53
0.013
600 +
6.3 +
NA
<30
49 +
50
160
0.3
9
0.0014
0.78
3.36
NA
NA
NA
4300 +
<160
NA
<80
<50
<70
<60
90 +
300
40
kg wet weight
**11.0 + 0.8 percent carbon by weight (average measured value)
t Confirmed by decay measurements
NA — not analyzed
Notes: + values are 2-sigma and < values are 3-sigma counting error.
-------
Table 5.7
Average Radionuclide Concentrations in Bone and Muscle
in Fresh Water Fish, pCi/kg*
Radionuclide
Canal fish
Background fish
Bone
'°Sr
3H
I4C
'°Sr
1MCs
13'Cs
600 ±
Muscle
160
2900 ± 2100
900 +
23 +
50 +
190 ±
300
12
50**
80
600 ±
600
670
15
< 20
80 ±
30
40
*kg wet weight; + values are 1-sigma of individual
concentrations.
**Values below the detection limit were averaged as one-
half of this limit.
Notes:
1. Concentrations of JH and "C measured in kidney plus
liver samples were averaged with concentrations in muscle,
on the assumption that the two radionuclides were uniformly
distributed in these tissues.
2. The concentration of "C was computed from the specific
activity on the basis of 110 g C per kg muscle.
3. The concentrations of 3H and 90Sr in background fish
are from single samples; the specific activity of "C in
background fish was taken to be 6.1 pCi/g C.
concentration of 60 pCi/kg fresh weight (3-sigma
counting error).
Measured values of stable strontium and 90Sr in the
muscle are believed to be due, in part, to small bones
analyzed with the meat, although they would normally
be removed when eaten. Note that the large carp
collected on March 2, 1971, for which meat and bone
could be separated efficiently, contained very little
calcium, strontium, and '°Sr in muscle. A previously
reported'5' ratio of strontium concentration in muscle to
that in bone of 0.01 is consistent with the results of
stable strontium and 90Sr in this carp sample. This
factor, applied to the average concentration of 90Sr in
Table 5.7 of 600 pCi/kg bone, yields an average of 6
pCi/kg muscle, which is considered more applicable
than the average measured value for fish muscle in
Table 5.7.
The concentrations of 137Cs in fish muscle were
higher for fish collected from the canal than for those
collected upriver in each of the two sets of samples,
although the 1-sigma values of the averages overlap
(see Table 5.7). The average '"Cs concentration relative
to potassium was 52 + 21 pCi/g K in the flesh of the
canal fish, and 22+12 pCi/g K in the two control
samples. The average 137Cs concentration in the canal
fish was not high in comparison to some other values in
fish from waters that do not contain discharges from
nuclear power reactors,'24""1 and less than in fish from
the Sherman and Harriman reservoirs'5' The presence
of 134Cs in some canal fish samples but not in
background fish (see Table 5.6), also suggests that part
of the 137Cs in fish from the canal is due to effluents
from the station. The average U4Cs/'37Cs ratio in fish
flesh was 0.3 + 0.1, about 1000-fold greater than in
fallout.'27' The ratio in water was approximately 0.5,
based on average 134Cs and '37Cs concentrations of 0.1
pCi/liter due to effluents (see Section 5.3.2) and the
l37Cs concentration in water of approximately 0.1
pCi/liter from fallout (see Section 4.3.4).
Cesium-134 was also detected in the stomach and
gut of all but one of the canal fish samples. The average
134Cs/'37Cs ratio was 0.5 + 0.1 in these samples. Due to
the unusual eating habits of the fish in winter discussed
above, many (60-75 percent) of the stomachs and guts
of these fish were empty. Therefore, the detection of
radiocesium in these samples may reflect to a large
degree the concentration in the stomach and gut wall,
rather than in the food. The amount of stomach
75
-------
contents was too small for analysis.
The specific activity of HC in muscle was higher in
two of the three canal fish collected on March 2 than in
the background sample. Values for the background
sample and the third catfish from the canal agreed
within the uncertainty of measurement with the usual
specific activity. Specific activity values above this
value were also measured in the liver and kidney (the
only tissue remaining for analysis) of fish caught in the
canal on December 1,1970.
The 3H concentration in three muscle samples and
in one of the two kidney plus liver samples was greater
than in the background sample. The excess of 3H
concentration in canal fish over background fish of
2,300 pCiAg (see Table 5.7), at a water content of 0.8
liter/kg in fish corresponds to a 3H concentration of
2,900 pCi/liter, approximately one-third of the annual
average value from station discharges in canal water
computed in Section 5.3.2. The concentration of 3H in
the background fish, at a water content of 0.8 liter/kg,
was 700 pCi/liter, consistent with background values
for river water reported in Section 4.3.4.
No 131I, 60Co, 58Co, or 55Fe was detected in muscle.
The minimum detectable levels were approximately 20
Table 5.8
pCi '"Co/kg, 40 pCi 58CoAg, 80 pCi 131I/kg, and 100
pCi "Fe/kg (10 pCi 5SFe/mg iron).
Radiocobalt was detected in the gut, 13II, in gut and
in kidney plus liver samples, and 55Fe, in kidney plus
liver samples. The highest concentration of S5Fe was in
the background sample, suggesting that this
radionuclide originated in fallout. The average S5Fe
concentration was 2,400 -j- 1,700 pCiAg wet weight or
13 + 5 pCi/mg iron. This specific activity falls within
the range of 3-50 pCi/mg iron in liver of fresh water
fish collected in Finland during 1955.<28) That the gut,
kidney, and liver offish contain these radionuclides has
no health implications because they are not eaten by
man, but these tissues may, within limits, be useful
indicators for radionuclides in edible tissue. For
example, the 55Fe specific activity of 13 -j- 5 pCi/mg
iron may be a better value for muscle than the value
from direct measurement of < 10 pCi/mg iron. The
specific activity in muscle may be lower, however, if
equilibrium has not been attained.
In shad, only 3H and 137Cs were higher upstream
from Haddam Neck than in the estuary (see Table 5.8).
The higher 137Cs concentrations may have resulted
from exposure to 137Cs in fresh water with its much
Radionuclide (pCi/kg) and Stable Ion (g/kg) Concentrations in Shad
Sample
Bone
Muscle
Kidney +
liver
Gut
Nuclide
Ash/wet wt.
MSr
Ca
Sr
Ash/wet wt.
JH
'"Sr
""Ru
mCs
Ca
Sr
K
S!Fe
Fe
"7Cs
April 15 +
Estuary
0.13
16 + 10
38
0.045
0.014
540 + 60
1.4 + 0.6
15 + 8
<25
0.35
0.0005
3.9
4700 + 200
0.14
<30
26, 1971
River
0.10
8 + 6
35
0.036
0.014
3600 + 100
2.1 + 0.5
20 + 10
50 + 20
0.31
0.0004
3.9
4500 + 200
0.14
<25
June 1, 1972
Estuary
0.12
12 + 7
36
0.042
0.012
NA
NA
20 + 10
19+6
0.29
0.0004
3.5
NA
NA
<20
River
0.11
10 + 6
38
0.040
0.012
NA
NA
18 + 6
16 + 6
0.33
0.0006
3.4
NA
NA
70 + 30
Notes:
1. + values are 2-sigma counting error; < values are 3-sigma counting error.
2. NA - not analyzed
3. The following were not detected in muscle, kidney + liver, and gut;
(< 30 pCi/kg) s"Co, "Co, 65Zn, and 134Cs; (< 100 pCi/kg) '"I.
4. kg values are wet weight.
76
-------
lower potassium content than sea water."4' The
concentration of 3H measured in the shad caught
upriver is similar to that measured in fish from the
discharge canal, despite the short period of time the
shad normally remain in the river just below the station
(see Section 5.5.1). Additional sampling is needed to
reach any conclusions regarding the levels of these
radionuclides in shad.
5.5.4 Estimated average radionuclide
concentrations in fish. The following radionuclide
concentrations and percents of intake limits (discussed
in Section 5.5.5) for fish from the coolant canal were
computed from average radionuclide concentrations in
canal water to indicate the magnitude of and major
contributors to the radiation dose from eating these
fish:
Radio-
nuclide
'H
UC
"Na
MP
"Cr
"Mn
"Fe
"Fe
"Co
"Co
"Co
"Sr
"Zr
"Nb
"Mo
"°-Ag
"I
"I
"I
"Cs
MCs
17Cs
Annual average
concentration in
canal water , *
pCi/1
9,700
0.04
0.03
0.01
0.007
0.42
0.6
0.01
0.008
3.6
0.56
0.0007
0.001
0.004
0.04
0.004
2.0
0.94
1
0.1
0.03
0.50
Concentration
factor"41
0.90
4,500
20
100,000
40
100
100
100
20
20
20
5
3.3
30,000
10
2.3
15
15
15
400
400
400
Hypothetical
concentration
in canal fish.t
pCi/kg
8,700
200
0.6
1,000
0.3
42
60
1
0.2
72
11
0.004
0.003
120
0.4
0.009
30
14
15
40
12
200
Percent of
limit**
0.01 TB
<0.001 TB
< 0.001 GI
2 B
<0.001 GI
0.004 GI
< 0.001 S
< 0.001 GI
< 0.001 GI
0.008 GI
0.002 GI
< 0.001 B
<0.001 GI
0.02 GI
<0.001 GI
< 0.001 GI
1.0 T
0.13 T
0.06 T
0.02 TB
< 0.001 TB
0.04 TB
* From Section 5.3.2, using values reported by the station
operator when available.
t The product of the values in columns 2 and 3. The concentration
in river fish would be l/25th of these values, due to dilution.
**The limit is based on an intake of 50 g fish per day12" that
will result in an exposure equal to the Radiation Protection
Guides recommended by the FRC™': The RPG are 500 mrem/yr
for thyroid (radioiodine) and 170 mrem/yr for all other
critical organs. The critical organs are: (TB) total body;
(GI) gastrointestinal tract; (B) bone; (S) spleen; (T) thyroid.
The only radionuclide detected in samples of shad
stomach, intestine, thyroid and roe was naturally
occurring 40K. In the kidney plus liver, equal amounts
of the fallout radionuclide 55Fe (33 + 3 pCi/mg iron)
were observed in river and control fish. The average
'°Sr concentration in shad bone of 12 + 4 pCi/kg was
lower than in fresh water fish. All other radionuclides
were below the limits of detection given in the note to
Table 5.8.
These radionuclide concentrations were computed
with the listed concentration factors for edible portions
of fresh water fish041 from the estimated annual average
concentrations of radionuclides in the discharge canal
water given in Section 5.3.2. Concentrations in water
were based on the station's effluent data (first data
column of the table in Section 5.3.2) when available;
when not, data from this study were used. For the
purpose of the calculations, it was assumed that
77
-------
radionuclides in the edible portions of all consumed
fish had reached equilibrium with radionuclide
concentrations in the discharge canal water. In reality,
these calculated radionuclide concentrations in water
and concentration factors from water to fish flesh are
only approximate, and radioactive equilibrium may not
have been attained in the fish.
The computed concentrations in fish agree with the
average measured values in canal fish minus those in
background fish (see Table 5.7) for 14C, 134Cs, and 137Cs,
but are several-fold higher for 3H. The computed
concentrations of 55Fe, 60Co, and 131I were below the
limits of detection. The computed values for 58Co and
"Nb appear to be too high by at least a factor of two in
that the indicated concentration would have been
detected if present.
5.5.5 Estimated population radiation dose.
Phosphorus-32 and the three radioisotopes of iodine
are the critical radionuclides according to the table in
Section 5.5.4, being at the highest percent of the limit in
fish caught in the coolant canal. The annual doses from
the listed radionuclides would be 3 mrem/yr to bone
(mostly from 32P), 6 mrem/yr to the thyroid (mostly
from "'!), 0.4 mrem/yr to the GI tract (mostly from
32P) and 0.3 mrem/yr to the total body. The dose rates
from eating 18 kg fish per year estimated in the
Environmental Statement are higher for all organs
except the thyroid:'33'
Total body 10.0 mrem/yr
GI tract 0.66
Thyroid 2.8
Bone 7.5
A total-body dose of 0.7 mrem/yr, almost entirely from
134Cs and I37Cs, is derived from the dose calculations for
a model 2-reactor 2,000 MWe PWR station,'32' adjusted
4-fold downward for the lower power level of the
Haddam Neck station. This value, however, is based on
the different concentration factors in an earlier version
of the report04' used here.
The percentages of intake limits in Section 5.5.4
(last column in the table) are based on the indicated
radionuclide concentrations in water. Daily
consumption of 50 g offish caught 6 months per year in
the canal (concentrations in column 4 x 0.5 yr x 0.05
kg/day) and 6 months per year in the river
(concentrations in column 4 x 0.5 yr x 0.05 kg/day x
0.04 dilution) is assumed. The calculations took the
maxitnum permissible daily occupational drinking-
water intake listed by the ICRP to correspond to 5
rem/yr to the total body, 15 rem/yr to the GI tract, and
30 rem/yr to bone,<33> or directly applied FRC guidance
for radiostrontium and radioiodine.<30) For these
calculated dose rates, the limit is taken to be the
Radiation Protection Guides for a "suitable sample" of
the exposed population as recommended by the FRC:
500 mrem/yr for the thyroid and 170 mrem/yr for all
other critical organs.'30' The applied ratios04' of
pCi/day per rem/yr are tabulated in Appendix E.2.
Of the critical radionuclides, 131I in fish muscle at
the indicated concentrations may be barely detectable
by gamma-ray spectrometry of fresh fish, but was not
measured because of the time that elapsed between
sampling and analyses. Measurements of 32P by beta-
particle detection after chemical separation were not
undertaken. Both of these radionuclides should be
measured in future studies.
Annual doses based on measured radionuclide
concentrations in fish are tabulated below. The average
amounts from Table 5.7 of 3H, 14C, 134Cs and 137Cs
measured in canal fish muscle, and of 90Sr in muscle
inferred from fish bone analyses, are given for a 50-g
sample. The amount in the canal fish due to the station
was obtained by subtracting the concentration values in
control fish. The annual radiation doses in the next
column are based on the ratios in Appendix E.2 for a
daily intake of 50 g fish caught 6 months of the year
from the canal and 6 months of the year from the river.
Critical organ
Total body
Total body
bone
Total body
Total body
When fallout is included, the dose is 0.13 mrem/yr
to the total body, and 0.25 mrem/yr to bone. The
additional radiation dose from eating shad that
contains 160 pCi 3H per 50 g during 3 months would be
0.003 mrem.
5.6Radionuclides in Shellfish
5.6.1 Collection and analysis. Shellfish were
reported to have been in the Connecticut River near the
station, but could not be found.'10' Shellfish were
sampled from the mouth of the Connecticut River,
however, because they are harvested from this location
for human consumption. Clams (Ellito complanatus)
Average concentration
Radio-
nuclide
measured in
pCi/50 g
fish,
Radiation dose
from station,
mrem/yr
Total From station
]H
"C
'°Sr
134Cs
l]'Cs
145
45
0.3
3
10
120
12
<0.1
3
6
0.005
0.002
<0.05
0.04
0.04
78
-------
and oysters (Crassostrea virginica) were collected on
April 16, 1971, from the mouth of the Connecticut
River. At the same time, control samples were collected
from brackish water near Elihu Island, Stonington,
Connecticut, approximately 38 km east of the mouth of
the Connecticut River.*
The shellfish were returned to the laboratory on dry
ice. The meat was thawed, removed from the shell, and
analyzed for photon-emitting radionuclides and
radiostrontium as described in Section 5.5.2. The shells
were analyzed similarly after removing all organic and
chitinous material from their exterior.
5.6.2 Results and discussion. No radionuclides
attributed to the nuclear power station were observed
in any shellfish samples, as shown in Table 5.9. Only
40K, 5!Fe, '°Sr and 137Cs were detected, and these were
approximately at the same concentrations in samples
from the Connecticut River estuary and from
-Stonington.
Cesium-137 was below the detectable concentration
in oyster meat, but was measured in clam meat. This
agrees with a report by Chipman who observed in most
cases higher '"Cs levels in clams than in oysters."5' For
"Fe, the concentration was higher in oyster meat: the
specific activity was 4.5 + 0.3 nCi/mg iron. The '°Sr
levels were similar in clam and oyster meat. The
average concentration was 70+15 pCi/kg, about 10
times the concentration in the fish muscle.
The only radionuclides in the shells were 5SFe, at a
specific activity of 6.4 + 1.9 pCi/mg Fe in the clam,
and '°Sr in oyster and clam. The concentration of S5Fe
was approximately 20 times lower than in meat, but the
90Sr concentration was 10 times higher.
The average concentrations of "Fe, '°Sr and 137Cs
measured in shellfish muscle correspond to the
following annual radiation exposure at a daily
consumption of 50 g:<36)
Table 5.9
Radionuclide (pCi/kg) and Stable Ion (g/kg) Concentrations in Shellfish
Oysters
Nuclide
Meat
5SFe
"Sr
137Cs
Ca
Sr
Fe
K
Shell
"Fe
MSr
Ca
Sr
Fe
River mouth
127,000+38,000
80 + 20
<25
1.6
0.029
0.029
0.8
<300
640 + 60
150
0.34
0.09
Control
200,000+36,000
80 + 15
<30
0.48
0.010
0.045
0.7
<330
620 + 60
170
0.50
0.12
Clams
River mouth
30,000+20,000
60+15
60 + 20
8.2
0.090
0.025
1.1
1500 ± 400
560 + 50
400
0.70
0.30
Control
50,000+25,000
50 + 20
60 + 30
0.93
0.013
0.051
0.6
2000 + 500
980 + 70
440
1.9
0.26
Notes:
1. + values are 2-sigma counting error; < values are 3-sigma counting
error.
2. The following were not detected:
in meat—'"Co (< 40 pCiAg), '3II (< 110 pCi/kg), ™Cs (< 40 pCi/kg)
in shell—MCo (< 40 pCiAg), "Zn (< 70 pCiAg).
3. kg values are wet weight.
*We thank R. Massengill for collecting the shellfish samples.
79
-------
Average concentration Radiation dose,
Radionuclide in shellfish, pCi/50 g mrem/yr Critical organ
"Fe
MSr
"7Cs
5,100*
3.5*
3**
4
3
0.03
Spleen
Bone
Total body
*Average concentration in clam and oyster meat.
"Average concentration in clam meat only.
According to this tabulation, a person who consumes
50 g shellfish per day will receive a dose to the spleen
and bone of 4 and 3 mrem/yr, respectively. Because no
significant differences were observed between the
radionuclide concentrations in shellfish collected from
the mouth of the Connecticut River and in the control
samples, these exposures are attributed to fallout and
not to the Haddam Neck station.
On the assumption that some clams can be obtained
for consumption from the discharge canal, the
Environmental Statement attributed to station
operation doses of 0.1-0.2 mrem/yr to the total body,
thyroid, GI tract, and bone.(31)
5.7 Radionuclides in Sediment
5.7.1 Sampling and measurement. Bottom
sediments in the coolant water discharge canal and
near the banks of the Connecticut River between
Haddam Island and the Salmon River were surveyed
with a submersible gamma-ray probe on September
14-16, 1970. The 10-cm x 10-cm-dia. cylindrical
Nal(Tl) detector in a water-tight container was
connected to a multichannel analyzer operated at 10
keV/channel.* The analyzer and associated
equipment, including portable electric generator, were
operated in a boat. The probe was lowered over the side
of the boat and then placed by a diver. Most of the
locations at or below the station (see Figure 5.3) were
selected because staff from the Essex Marine
Laboratory had observed that suspended material
accumulated there. Three locations near Haddam
Island were selected as controls, to measure the natural
background radiation in the vicinity.
Probe readings were screened initially by observing
dead-time values; those that were above 1 percent
suggested the presence of radiation levels higher than
usual, possibly from effluents at the station. A gamma-
ray spectrum was obtained for a 10-min period at each
of these locations and at the background sites. The
background spectra showed gamma rays from 4QK and
the progeny of radium and thorium. Gamma-rays
attributed to 58Co and <0Co discharged at the station
were observed in spectra that showed the higher dead
times.
At the locations indicated by the letter B in Figure
5.3, the diver collected sediment samples. Two samples
were generally taken side by side—the top 3 cm scraped
with a specially devised scoop, and a 10-cm-dia. core,
13 cm deep. These samples were identified by the letters
T and C, respectively. The wet samples were stored in
plastic containers for analysis.
Two additional sediment samples were collected on
March 2, 1971, from the mouth of the discharge canal
(B-31) and from the mouth of the discharge canal at the
Maromas fossil-fueled plant near Middletown, 18 km
upstream (B-30). These two samples were obtained
with a Petersen dredge. A core sample (B-32), 86 cm
deep and 4.5 cm in diameter, was collected on April 28,
1971, from the accumulated sediment at the mouth of
the canal. The sampling location was intended to be
that of sample B-7. This core sample contained five
separate zones and was later sectioned into these five
parts.
All samples were returned to the laboratory, air
dried at room temperature (20° C), homogenized and
passed through a 2.0-mm (No. 10) sieve. Material
larger than 2 mm, mostly twigs and pebbles, was
discarded.
5.7.2 Description of sediment samples. To define
geochemically the nature of the collected sediment
samples, some were analyzed for pH, organic content,
particle size distribution, cation exchange capacity, and
clay mineralogy.t The results of these analyzes are
presented in Table 5.10.
The analytical techniques are standard procedures
of ASTM and the American Society of Agronomy.071
In brief, the pH was determined by stirring a weighed
aliquot of dried, homogenized sample with an equal
volume of water, and reading the pH of the supernatant
*We thank Mr. Charles Phillips, Eastern Environmental Radiation Facility, EPA, for providing the
instrument and participating in the measurements
|We thank Professor L. Wilding, Department of Agronomy, Ohio State University, for performing these
analyses.
80
-------
B- Benthal samples plus gamma-ray probe
(except benthal sample only at 81,233)
G- Gamma-ray probe only
Log
/^Barrier
meter
100 200 300
Figure 5.3 Sites for Sediment Sampling and Gamma-Ray Probe Measurements
-------
oo
to
Table 5.10
Mineralogies! Analyses of Sediment Samples
Sample No.
Category
Texture
pH
Organic carbon
Organic matter
Clay (< 2 u dia.)
Silt (2-50 u dia.)
Sand (50-2000 u dia.)
2
Top
sand
5.8
Organic
0.33
0.57
Particle
<0.1
1.8
98.2
5
Top
sand-to-
loamy sand
5.3
Content, % of Total
1.06
1.82
Size Distribution, %
4.3
8.4
87.3
12
Core
loamy
sand
5.4
Dry Weight
0.60
1.03
of Total Weight
1.4
18.4
80.2
12
Top
loamy
sand
5.4
0.68
1.17
1.4
16.2
82.4
17
Core
sandy
loam
5.8
2.50
4.30
2.0
44.6
53.4
21
Core
sand
6.0
0.55
0.95
1.6
4.3
94.1
31
Dredge
loamy sand
5.6
0.79
1.36
4.0
13.0
83.0
Cation Exchange Capacity (CEO, meq/100 g
Total, Direct- Method
Exchangeable H+
Exchangeable Bases
Sum exchange H+ + Bases
2.2
1.4
0.6
2.0
2.3
...
...
—
Clay Minerals and Amorphous
Illite (mica)
Vermiculite
Quartz
Chlorite
Kaolinite
Amorphous Material**
...
...
—
—
...
49(37)
18(13)
26(20)
<5(<4)
7(5)
25
3.6
4.0
1.1
5.1
Material, % of
—
...
...
—
...
...
4.2
3.2
1.4
4.6
Crystalline
—
—
...
—
...
...
8.5
8.8
5.0
13.8
Fraction*
60(44)
17(12)
13(9)
<5(<4)
10(7)
28
2.2
3.6
1.4
5.0
53(42)
19(15)
22(17)
<5(<4)
6(5)
21
1.8
-
—
-
—
-
-
—
-
-
* values in parentheses are % of crystalline fraction, that is % of
total clay fraction minus organic matter and allophane contents.
**% of non-organic matter.
-------
liquid with a standard glass electrode. The carbon
content of the samples was determined by a gravimetric
measurement of the CCh produced from thermal
decomposition of the dried material. The carbon values
were converted to equivalent organic matter by
multiplying by 1.72. Particle-size distribution was
determined for the sandy fractions (> 50 micron) by
dry sieving, while the finer silt and clay fractions (< 50
micron) were digested with HzCh and the Calgon-
dispersed aliquots were separated by sedimentation.
The total cation exchange capacity (CEC) was
measured directly by saturating a weighed sample with
K+ (as KC1), eluting the excess salt with 95-percent
ethanol, displacing the K+ ion with NH4+ (NH4OH at
pH 7), and measuring the K+ ion by an atomic
absorption spectrophotometer. To confirm these
values, the CEC of five samples was also measured
indirectly as the sum of exchangeable bases (saturating
with sodium acetate) and exchangeable acids (barium
chloride-triethanolamine method). The exchangeable
acid is H+, while the exchangeable bases are defined as
the alkali and alkaline earth metals.
Sample B-17 was unusual in that the alcohol wash
was highly colored, suggesting removal of some organic
matter that could account for a fraction of the total
CEC. For this sample, the CEC obtained by the sum
technique was considered more valid, although, even
for the other samples, values obtained by the different
techniques are not necessarily comparable.
The clay mineral composition of three samples was
determined by X-ray crystallography of preferred-
oriented aggregate clay fractions on ceramic plates.'37'
The sample particles were dispersed with NaaCOs
without prior removal of carbonates and iron oxide,
and the <2-micron clay fractions were flocculated
with NaCl. Amorphous matter was taken as the
difference between total sample weight and the sum of
the crystalline clay fractions.
It was not possible, on the basis of these data alone,
to distinguish CEC due to organic matter from that
arising from the mineral component. However, data in
Tables 5.10 and 5.11 for sample #31 suggests that
most of the exchange capacity was associated with the
clay fraction.
To determine the effects of sample preparation on
particle size distribution, percent of clay minerals and
organic matter, and total cation exchange capacity,
sample B-31 was analyzed in the following three forms;
(1) air-dried and electrolyte dispersed, (2) air-dried and
water dispersed, (3) wet and electrolyte dispersed. The
results, given in Table 5.11, indicate that evaluations of
the clay fraction are dependent on sample preparation.
Hence, carefully defined standardized analytical
techniques must be used in studies of this type.
Table 5.11
Comparative Analyses of Sediment Sample B-31
Analysis
Air Dried
Electrolyte H2O
dispersed dispersed
Wet
Electrolyte
dispersed
Particle Size Distribution
Clay (< 2u dia.), %
Silt (2-50p dia.), %
Sand (5—2000u dia.), %
CEC, meq/100 g
CEC, clay fraction only,
meq/100 g
Organic carbon,
clay fraction, %
Organic matter*, %
4.0
13.0
83.0
1.8
59
7.0
12.6
2.8
16.4
80.7
3.8
14.4
81.8
2.7
32
5.9
10.6
*% carbon x 1.72
83
-------
0
800
50
850
100
900
150
950
200
1000
500
1300
550
1350
600
1400
250 300 350 400 450
1050 1100 1150 1200 1250
CHANNEL NO. (1.0 keV/channel)
Gamma-ray spectrum of sediment sample, B-7 (top), 0-1600 keV.
650
1450
700
1500
750
1550
800
1600
Figure 5.4
Detector: Ge(Li), 10.4 cm2 x 11 mm, trapezoidal.
Sample: 564 gm (400 cc) dry wt., collected Sept. 13, 1970 from mouth of discharge canal.
Count: Nov. 9-10, 1970 (1000 min., background not subtracted); Ra and Th refer to 226Ra and
232Th plus progeny.
-------
5.7.3 Radioactivity measurement. The count rates
due to Co and 60Co in the gamma-ray spectra obtained
with the submersible probe were computed by
subtracting interferences from other radionuclides. For
58Co, net count rates were derived by summing the
energy range from 0.76 MeV to 0.88 MeV, and
subtracting the Compton interference estimated
graphically under each 58Co peak by tracing a straight
line over the energy interval from 0.50 to 1.00 MeV.
Interference from traces of 54Mn (0.84 MeV) and '"Cs
(0.80 MeV) was not considered. This interference was
determined later from laboratory analyses of sediment
collected at the probe points to have resulted in 58Co
values overestimated by no more than 25 percent. Net
count rates due to 60Co were calculated for its 1.17-
MeV and 1.33-MeV photopeaks together by summing
from 1.08 MeV to 1.37 MeV and subtracting as
background over the same energy range the average
count rates for probe measurements at locations 1, 2, 3,
5, and 6. The major correction is for the 1.46-MeV
gamma ray of naturally occuring 40K.
Radionuclides in the sediment samples were
measured by analyzing 400 ml volumes(360-680 g) of
air-dried, homogenized sample by gamma-ray
spectrometry with a 10- x 10-cm Nal(Tl) detector or a
Ge(Li) detector (see Figure 5.4). The Nal(Tl) spectra
were solved by a computer (matrix technique)
program. The better-resolved Ge(Li) spectra were
solved by direct calculation. Although sample densities
ranged from 1.3 to 1.7 g/cm3, no correction for self-
absorption was applied because the effect of absorption
in this density range on measuring the radiations of
interest was found to be no larger than 10 percent.
Concentrations of 90Sr were determined by
separating the strontium from either acid-leached or
fused samples with an ion-exchange procedure.09' The
MSr was determined by counting the beta particles of
the ingrown 90Y.
The initial analyses of the sediment samples for
naturally occurring 226Ra were with the Nal(Tl)
detector, based on the 1.76-MeV photon of the 214Bi
progeny. Radioactive equilibrium between parent and
daughter was assumed. Later analyses of most of these
samples with the Ge(Li) detector by measuring directly
the 0.186-MeV photon of 226Ra indicated that earlier
results yielded only about 40 percent of the actual 226Ra
concentration. This presumably was due to the
emanation of 222Rn from the dried samples. Samples 2,
3, 5, and 6 were only measured by Nal(Tl) detector
and, therefore, have 22'Ra results that are too low to this
extent.
5.7.4 Results and discussion of analyses. Sediment
collected at the mouth of the discharge canal and at
selected locations immediately upstream and
downstream along the eastern shore of the Connecticut
River (see Figure 5.3) contained radionuclides
attributable to plant operation as shown in Table 5.12.
The highest levels were associated with particulate
sediment deposited in the mouth of the discharge canal.
The four background samples (1, 2, 3 and 30) collected
in the river upstream from the station contained 54Mn,
Sr, and '"Cs—presumably from fallout—at
concentrations not exceeding 0.2 pCi/g. They also
contained naturally occurring 40K, 226Ra plus progeny
and 232Th plus progeny. No "Co, S8Co, 60Co, or n4Cs
(< 0.1 pCi/g) was detected in these 4 samples and in
sample 6.
Cobalt-58 was the predominant radionuclide
indicative of contamination from plant discharges in
the sediment samples. It was usually accompanied by
60Co, 134Cs, and 137Cs. Some of the samples also
contained 54Mn and "Co. Only samples 5 and 31
contained some of these radionuclides without 58Co.
Concentrations of 90Sr above background were not
found, except possibly in sample 17C.
The ratio in Table 5.13 of radionuclide
concentrations in the top 3 cm to those in the top 13 cm
of sediment from the mouth of the discharge canal
indicate approximately 2-fold higher levels of effluent
radionuclides near the top. For the naturally occurring
radionuclides 40K, 226Ra and 232Th, the ratio was near
unity. Concentrations of effluent radionuclides in the
sections of core sample 32 (Table 5.14) also were
highest near the surface. None of these radionuclides
were found below the 29-cm depth.
5.7.5 Results and discussion of probe
measurements. The probe was a convenient device for
surveying benthic matter in situ because results were
immediately available to indicate the location and
distribution of radioactivity. Analysis of sediment
collected at the locations examined by the probe then
provided precise information concerning radionuclide
identity and quantity. An approximate value of the
counting efficiency of the probe was obtained by
comparing the probe count rate with radionuclide
concentrations in samples from the same locations.
The probe measurements in the Connecticut River
and discharge canal detected 58Co and 60Co in the
mouth of the discharge canal and for short distances
up- and down-stream along the east shore of the
Connecticut River (see Table 5.15). Neither 58Co nor
'"Co was detected by the probe within the discharge
canal. The 58Co and 60Co count rates were highest at the
85
-------
Table 5.12
Concentration of Radionuclides in Sediment, pCi/g dry weight
Sample No.
1C
IT
2C
2T
3C
3T
4C
4T
5C
5T
6C
6T
7C
7T
8C
8T
IOC
10T
11C
11T
12C
12T
13C
14C
17C
18C
21C
23C
30D
31D
Sample Wt.,
g/400 cc
621
603
595
662
622
572
633
608
500
479
623
645
592
564
481
459
428
444
590
629
545
546
443
483
364
562
585
447
515
484
"Mn
<0.1
<0.1
0.2
0.2
0.1
<0.1
<0.3
<0.2
<0.1
<0.1
<0.1
<0.1
0.5
<0.2
<0.3
<0.3
<0.4
<0.3
0.3
<0.2
<0.1
1.5
0.1
<0.3
1.4
0.5
0.7
<0.1
<0.1
<0.2
"Co
<0.1
<0.1
<0.1
<0.1
<0.1
<0.1
0.4
0.6
<0.1
<0.1
<0.1
<0.1
4.1
5.6
1.3
3.3
0.5
2.5
2.8
1.4
1.9
13.5
8.1
2.5
8.5
4.3
5.5
1.4
<0.1
<0.1
"Co
<0.1
<0.1
<0.1
<0.1
<0.1
<0.1
<0.1
<0.1
<0.1
<0.1
<0.1
<0.1
0.9
1.1
0.2
0.4
0.2
<0.1
0.6
0.3
0.5
3.2
1.6
0.6
1.5
0.8
1.2
0.2
<0.1
0.1
1MCs
<0.1
<0.1
<0.1
<0.1
<0.1
<0.1
<0.1
<0.1
0.1
<0.1
<0.1
<0.1
0.6
0.6
<0.1
0.3
<0.1
<0.1
0.5
0.1
0.4
1.9
1.1
<0.1
1.0
0.4
0.6
0.4
<0.1
0.3
'"Cs
0.2
0.1
<0.1
0.2
0.2
0.1
<0.1
<0.1
0.5
0.6
0.1
0.1
1.3
1.4
0.6
0.8
0.6
0.6
0.7
0.2
0.8
3.1
1.9
0.7
2.0
0.7
1.3
0.9
0.2
0.4
"«Sr
0.2
0.1
0.1
0.2
0.1
0.1
<0.1
0.1
0.1
0.1
0.1
0.1
0.1
0.1
0.2
0.2
NA
NA
0.1
0.1
NA
0.1
0.2
0.1
0.3
0.2
0.1
NA
0.1
0.1
4«K
10.8
10.8
10.5
9.0
10.8
11.0
9.8
11.0
9.9
9.9
9.7
10.1
10.1
9.5
10.0
9.2
14.3
11.3
11.7
8.6
12.5
10.5
13.5
10.6
11.7
9.2
9.5
11.1
14.5
13.1
"6Ra
0.2
0.3
0.3
0.7
0.2
0.3
1.0
0.6
0.5
0.5
0.2
0.2
1.1
1.5
1.5
1.3
1.8
2.3
1.0
0.5
1.5
1.0
1.0
1.2
1.0
0.7
1.5
0.7
0.7
1.5
"2Th
0.3
0.3
0.3
0.5
0.2
0.3
0.3
0.3
0.5
0.5
0.3
0.3
0.4
0.4
0.6
0.6
0.7
0.6
0.3
0.3
0.3
0.5
0.5
0.5
0.8
0.5
0.3
0.7
0.3
0.4
Notes:
1. Analyses of gamma-ray emitters are by Nal(Tl) spectroscopy for samples 2, 3, 5, and 6 and by Ge(Li)
for all other samples [except that all !32Th and s*Co analyses are based on the initial Nal(Tl) scan,
corrected for long-lived interference with "Co by a later Ge(Li) scan].
2. T = top 3 cm (0-3 cm depth); C = top 13 cm (0-13 cm depth); D = by Petersen dredge.
3. Detection limits at 2-sigma counting error were + 0.1 pCi/g except: "Mn by Ge(Li) = + 0.2 pCi/g,
"K by Nal(Tl) = ± 0.3 pCi/g, 4°K by Ge(Li) = + 1.2 pCi/g, I2'Ra by Ge(Li) = + 0.4
4. "Co was found at a concentration of 0.04 pCi/g in sample 7T and 0.09 pCi/g in 12T.
86
-------
Table 5.13
Concentration of Radionuclides in "Core" vs. "Top" Samples"
Average concentration, pCi/g
Radionuclide
HCo
"Co
'"Cs
137Cs
«K
"sRa
U2Th
13-cm depth (O
2.7
0.44
0.28
0.60
11.7
1.4
0.50
3-cm depth (T)
5.9
1.0
0.46
1.1
9.8
1.3
0.48
Ratio, T/C
2.2
2.3
1.6
1.8
0.8
0.9
1.0
'Samples B-7, B-8, B-10, B-ll, and B-12 from mouth of
discharge canal (see Figure 5.2).
Table 5.14
Concentration of Radionuclides in Core Sample (B-32)
as Function of Depth, pCi/g dry weight
Depth, cm
0-6
6-18
18-29
29-44
71-79
Weight, g MCo
55.4 0.3
113.6 <0.1
120.0 <0.1
153.9 <0.1
122.0 <0.1
MCo
0.38
0.08
0.15
<0.02
<0.03
1MCs
0.4
0.10
0.09
<0.03
<0.03
137Cs
1.1
0.26
0.24
<0.02
<0.02
40K
12.4
11.7
9.5
9.7
10.0
"
-------
Table 5.15
Net Count Rate of 58Co and 60Co with Nal(Tl) Underwater Probe
Location
B-l
B-2
B-3
B-4
B-5
B-6
B-7
B-8
G-9
B-10
B-ll
B-12
B-13
58Co,
Counts/min
<20
<20
<20
20 ± 10
<20
<20
1110 + 40
140 ± 20
<20
<20
410 ± 30
690 ± 30
920 + 40
60Co,
Counts/min
<20
<20
<20
<20
<20
<20
690 ± 30
80 ± 20
<20
<20
290 + 20
560 ± 20
640 + 20
Location
B-14
G-15
G-16
B-17
B-18
G-19
G-20
G-21
B-21
G-22
G-24
G-25
58/-i_
Co,
Counts/min
550 ± 30
1020 + 40
570 + 30
500 + 30
1100 + 40
880 + 40
<20
note 4
770 ± 30
note 4
<20
<20
Co,
Counts/min
430 + 20
680 + 20
420 ± 20
570 + 20
670 + 20
610 ± 20
<20
note 4
310 ± 20
note 4
<20
<20
Notes:
1. See Figure 5.3 for locations.
2. All counting times were 10 min.
3. + values are 2-sigma counting error; < values are 3-sigma counting error.
4. Gamma-ray spectra were not recorded because counter dead-time indicated zero.
mouth of the discharge canal.
The counting efficiencies of the probe for 58Co and
6°Co, given in Table 5.16 in terms of the ratio of probe
count rate of Table 5.15 to the concentrations of Table
5.12, varied considerably among locations. This would
be expected from nonuniform vertical and horizontal
distributions of radionuclides in the sediment. The
response of the probe did not differ significantly with
respect to the radionuclide content in samples from a
depth of 0-3 cm ("top") or 0-13 cm ("core"). The
average ratios for 60Co of 500 and 600 c/min per pCi/g
are consistent with values between 290 and 800 c/min
per pCi/g observed in the reservoir at Yankee-Rowe.(s)
The sensitivity of the probe is 0.15 pCi/g for 58Co
and 0.04 pCi/g for <0Co, based on the minimum
detectable count rate above background of 20 c/min
(see Table 5.15) and the respective average ratios in
Table 5.16 of 130 and 550 c/min per pCi/g. Minimum
detectable values based on comparing the relatively low
S8Co and 60Co concentrations in samples 4, 8, and 10
with net count rates by the probe are approximately
three times as large, and probably more realistic.
5.7.6 Significance of radioactivity in sediments. The
observation of radioactivity at only a few locations,
mainly in the mouth of the discharge canal, suggests
that fine particles with associated radionuclides are
swept down the discharge canal at a rate too high for
deposition. Upon reaching the broad mouth of the
canal, the flow rate is slowed and the fine particles are
deposited. The Essex Marine Laboratory has observed
the growth of the deposit since the station began
operating.<17) Radionuclides from the station were also
found in lower amounts at places where silt
accumulated near the east bank of the Connecticut
River extending from the station's water intake
upstream from the canal to the mouth of the Salmon
River, downstream.
The concentrations of radionuclides in the largest
deposit, at the mouth of the canal, were as follows,
based on the averages of values measured in core
samples?, 11,12,13,14, 17, 18, and21 (Table5.12):
Average concentration, Estimated total,
Radionuclide
S4Mn
"Co
'"Co
1MCs
"7Cs
pCi/g
0.4
4.7
0.8
0.6
1.2
pCi/cm!
9
100
18
13
26
mCi
0.3
3.5
0.6
0.5
0.9
The concentrations per unit area were computed from
the average density of 1.3 g/cm3 and the estimate that
the amount of radionuclide in the entire silt column
was approximately 1.3 times the amount in the top 13
cm core sample (see Table 5.13). The totals were
estimated for an area of 3.5 x 101 cm2; divers from the
Essex Marine Laboratory had observed the deposit to
be in the shape of a lens approximately 90 m x 60 m just
-------
Table 5.16
Ratio of Count Rate by Underwater Probe to Radionuclide Concentration
in Sediment Samples
Location
range
Average (4- Iff)
lsCo,(c/min)/(pCi/g)
•'core" "top"
50-360
20-210
170+100 80+80
>0Co,(c/inin)/(pCi/g)
"core"
260-1120
"top"
B-4
B-7
B-8
B-ll
B-12
B-13
B-14
B-17
B-18
B-21
50
270
110
150
360
120
220
60
260
140
30
120
20
210
40
—
770
400
480
1120
400
720
380
840
260
-
630
200
970
180
180-970
600+280 500+380
Note: Count rates are from Table 5.15 and radionuclide concentrations
from Table 5.12.
outside the canal barrier, with a neck extending
downriver near the east bank.
The amounts in the sediment are approximately 0.1
percent of these radionuclides in liquid discharges if it
is assumed that the effluents since operation began
contained twice the discharges reported for the years
1970 and 1971 in Section 4.1.2. This is unlike the
situation at the Yankee-Rowe station,<5) where a large
fraction of these effluent radionuclides remained in the
sediment.
The radiation exposure from these radionuclides in
sediment is believed to be minute as long as the deposit
is covered by 0.5 m or more of water to absorb the
radiation, persons stay at a distance, and the
radionuclides remain in place. In the absence of
covering water, direct radiation doses to adults 1 m
above the benthal radioactivity are related to the
surface concentration by factors of 0.042 to 0.17
urem/hr per pCi/cm2 for the listed radionuclides, if an
infinite plane source is assumed.'38' The accumulating
sediment at the mouth of the canal was covered by 4 m
of water at the time of the study, and neither fishermen
nor swimmers came within its immediate vicinity.
The possibility of radionuclides in benthal material
entering the food chain through uptake by fish has been
suggested.'"" The extent and rate of such entry into the
food chain in the river are not known. Consequently,
the deposits are a potential source of radionuclides to
the aquatic environment and should be evaluated
periodically.
5.8 References
1. Combustion Engineering Combustion Division,
"Operational Environmental Radiation Monitoring
Program, Connecticut Yankee Atomic Power
Company Summary Report 1970," also for year 1971.
2. Connecticut Department of Environmental
Protection, "Radiological Data of Environmental
Surveillance - Year 1970," Hartford, Conn. 06117
(1971); also for Year 1971.
3. Riel, G. K. and R. Duffey, "Monitoring of
Radionuclides in Environmental Water," Trans. Am.
NucLSoc. 77,52(1968).
4. Lentch, J. W. etal., "Manmade Radionuclides in
the Hudson River Estuary," in Health Physics Aspects
of Nuclear Facility Siting, P. G. Voilleque and B. R.
Baldwin, eds., B. R. Baldwin, Idaho Falls, Idaho
(1971), p. 499.
5. Kahn, B. et al., "Radiological Surveillance
Studies at a Pressurized Water Nuclear Power
Reactor," EPA Rept. RD 71-1 (1971).
6. Merriman, D. et al., "The Connecticut River
Study," Essex Marine Laboratory Semi-Annual
Reports Nos. 1-10, Essex, Connecticut (1965-1970).
89
-------
7. Connecticut Yankee Atomic Power Co.,
"Haddam Neck Nuclear Power Plant, Environmental
Report, Operating License Stage," AEC Docket No.
50-213 (July 1972).
8. U. S. Atomic Energy Commission, "Standards
for Protection Against Radiation," Title 10, Code of
Federal Regulations Part 20, U. S. Government
Printing Office, Washington, D. C. (1965).
9. Krieger, H. L., and S. Gold, "Procedures for
Radiochemical Analysis of Nuclear Reactor Aqueous
Solutions," EPA Rept., EPA-R4-73-014 (1973).
10. Massengill, R. R., personal communication,
Essex Marine Laboratory, Essex, Connecticut (1974).
11. Harvey, R. S., "Temperature Effects on the
Sorption of Radionuclides by Fresh Water Algae,"
Health Phys. /£ 293 (1970).
12. Rice, T. R., "The Accumulation and Exchange
of Strontium by Marine Plank tonic Algae," Lim.
Ocean. 7,123(1956).
13. Office of Radiation Programs, EPA, "Tritium
Surveillance System, Jan. - Dec. 1970; Jan. - Mar.
1971; April - June 1971," Rad. Health Data Repts. 12,
272,384,576(1971).
14. Thompson, S. E., C. H. Burton, D. J. Quinn,
and Y. C. Ng, "Concentration Factors of Chemical
Elements in Edible Aquatic Organisms," AEC Rept.
UCRL-50564, Rev. 1 (1972).
15. Marcy, B. C. and R. C. Galvin, "Winter-Spring
Sport Fishery in the Heated Discharge Canal of a
Nuclear Power Plant," J. Fish Biol., In Press.
16. Massengill, R. R., "Change in Feeding and
Body Condition of Brown Bullheads Overwintering in
the Heated Effluent of a Power Plant," Chesapeake Sci.
74,133(1973).
17. Merriman, D. etaJ., "Connecticut River Study
Tenth Semi-Annual Rept.," Essex Marine Laboratory,
Essex, Connecticut (1970).
18. Chavin, W., "Thyroid Distribution and
Function in the Goldfish, Carassius Auratus L." J.
Exper. Zool. 133,259 (1956).
19. Porter, C. R., B. Kahn, M. W. Carter, G. L.
Rehnberg and F. W. Pepper, "Determination of
Radiostrontium in Food and Other Environmental
Samples," Environ. Sci. Technol. /, 745 (1967).
20. Templeton, W. L. and V. M. Brown,
"Accumulation of Strontium and Calcium by Brown
Trout from Waters in the United Kingdom," Nature
198, 198 (1963).
21. Nelson, D. J., etal., "Clinch River and Related
Aquatic Studies," AEC Rept. ORNL-3697, 95 (1965).
22. Ophel, I. L. and J. M. Judd, "Skeletal
Distribution of Strontium and Calcium and Strontium
Release Ratios in Several Fish Species," in Strontium
Metabolism, J. Lenihan, J. Loutit and J. Martin, eds.,
Academic Press, New York (1967), p. 103.
23. > Ruff, M., "Radioaktivitat in
Susswasserfischen," Zeit. Veterinarmed. 12, 605
(1965).
24. Gustafson, P. F., A. Jarvis, S. S. Brar, D. N.
Nelson and S. M. Muniak," Investigations of 137Cs in
Freshwater Ecosystems," AEC Rept. ANL-7136, 315
(1965).
25. Gustafson, P. F., "Comments on Radionuclides
in Aquatic Ecosystems," in Radioecological
Concentration Processes, B. Aberg and F. P. Hungate,
eds., Pergamon Press, New York (1967), p. 853.
26. Kolehmainen, J3., E. Hasenen and J. K.
Miettinen, "137Cs Levels in Fish of Different
Limnological Types of Lakes in Finland During 1963,"
Health Phys. 72,917(1966).
27. Hanson, W. C., D. G. Watson and R. W.
Perkins, Concentration and Retention of Fallout
Radionuclides in Alaskan Arctic Ecosystems," in
Radioecological Concentration Processes, B. Aberg
and F. P. Hungate, eds., Pergamon Press, New York
(1967), p. 233.
28. Jaakkola, T., "55Fe and Stable Iron in Some
Environmental Samples in Finland," ibid., p. 247.
29. Cowser, K. E. and W. S. Snyder, "Safety
Analysis of Radionuclide Release to the Clinch River,"
AEC Rept. ORNL-3721, Supp. 3(1966).
30. "Background Material for the Development of
Radiation Protection Standards," Fed. Rad. Council
Rept. #2, U.S. Government Printing Office,
Washington, D. C. 20402, (1961).
31. Directorate of Licensing, "Final Environmental
Statement Related to the Haddam Neck (Connecticut
Yankee) Nuclear Power Plant," AEC Docket No.
50-213(1973).
32. Office of Radiation Programs, "Environmental
Analysis of the Uranium Fuel Cycle, Part II—Nuclear
Power Reactors," EPA Rept. EPA-520/9-73-003-C
(1973).
33. International Commission on Radiological
Protection, "Report of Committee II on Permissible
Doses for Internal Radiation," Health Phys. 3 (1960).
34. Blanchard, R. L. and B. Kahn, "The Fate of
Radionuclides Discharged from a PWR Nuclear
Power Station into a River," in Environmental
Behavior of Radionuclides Released in the Nuclear
Industry, International Atomic Energy Agency,
Vienna, 195(1973).
35. Chipman W. A., "Accumulation of Radioactive
Material by Fishery Organisms," 11th Annual Meeting
of the Gulf and Caribbean Fisheries Institute, Miami
Beach, Florida, Nov. 17-21 (1958).
90
-------
36. Weaver, C. L., "A Proposed Radioactivity
Concentration Guide for Shellfish," Rad. Health Data
Rept. £, 491(1967).
37. Black, C. A., et al., Methods of Soil Analysis,
Amer. Soc. of Agronomy, Monograph No. 9, Vol. 1
and 2, Madison, Wisconsin (1965), pp. 545-567,
653-698, 891-923,1353-1365, 1397-1400, and 1413.
38. Directorate of Regulatory Standards,
"Numerical Guides for Design Objectives and Limiting
Conditions for Operation to Meet the Criterion 'As
Low As Practicable" for Radioactive Material in
Light-Water-Cooled Nuclear Power Reactor
Effluents," AEC Rept. WASH-1258 (1973), p. F-53.
91
-------
6. RADIONUCLIDES IN ENVIRONMENTAL AIR
6.1 Introduction
6.1.1 Purpose. Studies were conducted in the
environs of the Haddam Neck station to test techniques
for measuring ambient concentrations of, and
radiations from, radionuclides in airborne emissions
from a PWR. Tests were performed during three
periods of elevated radioactivity release when stored
gas was discharged from the waste surge sphere.
Radionuclides routinely discharged to air at the
Haddam Neck station are usually dispersed to
concentrations near or below detection limits before
reaching the ground. For example, the continuous
release from the vent stack of the most abundant
radionuclide, 133Xe, at a rate of 53 uCi/s (see Table 3.9)
would lead to an average concentration of only 2 x 10"*
uCi/m3 at the nearest exclusion fence. The average
dispersion value of 3.5 x 10"* s/m3 (see Appendix D.I)
used for this station is probably too large because the
elevation of the point of release is not considered. The
concentration value is below the minimum detectable
level of 4 x 1CT1 uCi/m3 by the procedure described in
this section. During releases of stored gas from the
surge sphere, however, the rate of 133Xe discharges was
approximately 40-fold higher. On those occasions, the
radionuclide should be readily detected under usual
conditions.
The radioactive gases in the waste surge sphere
consist primarily of 10.7-year 85Kr and 5.3-day 133Xe,
with relatively small amounts of other radionuclides
(see Table 3.2). Because 8!Kr emits its 514-keV photon
with an abundance of only 0.43 percent, almost the
entire gamma-ray flux from this gas is produced by
133Xe. The latter emits 32-keV cesium XK rays and 81-
keV gamma rays with 47 and 37 percent abundances,
respectively.0'
/
Optimum sampling locations were selected by
detecting the radiation from the plume of stack effluent
during discharge of the sphere with portable
scintillation detectors responsive to low-energy
radiation. Ambient 85Kr and 133Xe were then sampled
by compressing approximately 0.5 m3 of air into
cylinders for analysis at the laboratory. The direct
radiation was measured continuously during the
collection periods with the scintillation detector
coupled to a chart recorder.
6.1.2 Environment and meteorology. The
immediate environment is described in Section 7.1.2.
The following aspects are pertinent to these
measurements: the local region is hilly and heavily
forested; the plant is on the east bank of the
Connecticut River; a steeply rising wooded hill borders
the plant on the northwest side and lies within the
exclusion area to the east; and an access road from the
northwest leads to the site entrance, 530 m northwest of
the reactor. Overall, few roads exist in the area. A
contour map of the area is given in Figure 6.1.
A site meteorological study before plant
construction began showed that winds are affected
locally by the hills and tend to follow the river valley.
Surface winds, therefore, are predominantly
northwesterly or east-southeasterly. Atmospheric
dilution from stack to ground level beyond the plant
boundary is calculated by the station operator with the
diffusion model'2' described in Appendices D.I and
D.2.
A meteorological tower is located on site
approximately 200 m south of the stack. Its base is 6 m
above mean sea level (MSL). During this study, wind
speed and direction were measured 39 m (129 ft) above
MSL, and temperature, 31 m (101 ft) above MSL. The
data are recorded continuously in the reactor control
room.
The 53-m stack is at the same elevation and stands
adjacent to the reactor containment structure. The
plume will, therefore, be influenced by mechanical
turbulence around the building to an extent dependent
on meteorological conditions. Stack effluents are most
likely to be drawn into the reactor building wake
during southeasterly winds. At equal distances near the
stack, ground level concentrations nearby would then
be higher toward the northwest.
93
-------
Hodtfdnt
Musfear
Connecticut River
Haddom Meadows
State Park
Shai erviMe
Contours show ground elevation
measured in feet.
Figure 6.1 Locations for Plume Sampling and Measurement
(Numerals Indicate Test Desior.afii'n]
-------
6.2 Measurement of Short-term
Radiation Exposure and
Radionuclide Concentration
6.2.1 Air sampling. Samples were obtained with an
air compressor (27-V DC Cornelius Model 32-R-300)
connected to a 34-liter low-pressure gas bottle rated to
contain 0.9 m3 at maximum pressure. Each cylinder
was filled with about 0.5 m air in a period of 40 min.
The pump is powered by an 115-V AC motor generator
with output converted to 27V DC by a full-wave
rectifier.
For U3Xe analysis, sampled air was released from
the tank at a rate of 6 liters/min for 16.7 min. It was
passed through beds of Linde 13X molecular sieve and
Ascarite for removal of water vapor and CCh, then
through a 1-cm-dia. x 80-cm copper cooling coil, and
finally through a 3.2-cm-dia. x 66-cm copper U-tube
containg 180 g of Columbia 6GC (10-20 mesh)
charcoal. Both tubes were immersed in a -76° C dry-
ice-acetone refrigerant bath. The charcoal under these
conditions collected all 133Xe from 1 m3 or less of air.
After passage of 100 liters, the U-tube was opened
and the charcoal transferred to 10-cm-dia., 450-cc
plastic containers. The charcoal was allowed to warm
for 1 hour to room temperature to eliminate pressure
build-up. The container was then sealed with a rubber
gasket and a bolted lid. It was found that 36 percent of
the 133Xe on the charcoal is lost due to warming. The
charcoal was analyzed for 1000 min with a 10- x 10-cm
NaI(Tl) gamma-ray detector connected to a 200-
channel spectrometer. The analyzer was calibrated
with a 133Xe gas standard provided by the National
Bureau of Standards in October 1973.
The remainder of the air sample was analyzed for
85Kr, adding 1.86-hr 83°Kr to determine the krypton
yield. Krypton was separated and purified by cryogenic
fractionation. The krypton fraction was dissolved in
liquid scintillator solution, and 85Kr and 83l°Kr were
measured in a liquid scintillation counter. *<3)
6.2.2 Scintillation detector for low-energy photons.
A portable thin Nal(Tl) detector connected to a single-
channel analyzer with count rate meter (FIDLER) was
tested at Haddam Neck and proved to be a sensitive
detector under ambient conditions of the low-energy
photons emitted by '"Xe.'4' The hand-held detector
consists of a cylindrical 13-cm-dia. x 1.6-mm-thick
crystal optically coupled through a quartz light pipe to
a 13-cm-dia. photomultiplier tube. The pulse rate meter
(Eberline Instr. Co. Model PRM-5-3) is battery
powered and contains a linear-log readout meter and
three independent energy discrimination settings. Each
of the three provides single-channel pulse-height
analyzer capability. Before use, the meter was attached
to a multichannel analyzer and one energy setting was
adjusted to the center of the 81-keV spectral peak with
a range corresponding to one half-width. The gamma-
ray spectrum of 133Xe in the 10 to 160 keV energy range
with this arrangement is shown in Figure 6.2.
The scintillation detector was calibrated for 81 keV
photon response by means of 400 uCi of 133Xe in a 3-cc
glass vial. Count rate meter readings were related to
exposure rates by calibrating at the same time with a
Shonka muscle-equivalent ionization chamber and
electrometer used for low level radiation
measurements. The relationship between count and
gamma-ray exposure rates was linear for the source
located at three distances from the two instruments. A
counting rate with the scintillation detector of 15,000
counts/min was equivalent to 1 uR/h, i.e., 9 x 10s
counts = 1 uR.
One survey meter was connected to a strip chart
recorder (Texas Instrument Servo/Writer #2, Model
PSO-W/6A), to plot readings continuously at a chart
speed of 2.5 cm/min. Meter and chart-recorder
performances were tested periodically during field
operation with a 10.7-yr 133Ba source, whose gamma
rays include one of 81 keV energy.
6.2.3 Measurements. Three field tests were
conducted 0.5 to 0.6 km distant from the stack at the
locations shown in Figure 6.1. Atmospheric stability
conditions from slightly unstable to neutral were
selected so that the plume was likely to be measurable
at ground level. The scintillation detectors were used to
locate the plume centerline. Three or four compressed
air samples were collected in each test. Background air
samples were obtained upwind of the plant during tests
1 and 2.
The first test was conducted on September 16, 1970,
from 1227 to 1429 hours, approximately 0.55 km
WNW of the stack on the north side of the plant access
road. The '33Xe concentration in the surge sphere was
1.1 uCi/cc (see Table 3.2). Three air samples were
obtained but l33Xe was found only in the third. Because
the winds shifted constantly, results could not be
interpreted in terms of dispersion factors. The
usefulness of the scintillation detectors for detecting the
*We thank Mr. Sam Cummings of Eastern Environmental Radiation Facility, EPA, for performing these
analyses.
95
-------
10,000
1000
E
LJ
I-
2
3
O
100
20
40
CHANNEL
60
(KeV=~
80 100
l.02x channel number)
120
140
160
Figure 6.2 Gamma-ray spectrum of waste surge sphere gas.
Detector: 1.6mm x 12.7cm dia Nal(Tl) scintillation detector.
Sample: 9 cc bottle of gas collected 1005 hrs, March 15, 1971
Count: 1 min at 1710hrs, March 15, 1971
96
-------
plume was demonstrated, and 133Xe concentrations at
ground level near the station boundary were found to
be detectable while the sphere was being discharged
under these conditions.
The second test, on March 15, 1971, from 1339 to
1615 hours, did not yield useful results. Sampling was
conducted 0.55 km NW of the stack on the hillside
north of the access road. Traverses of the hillside from
WNW to ENE with scintillation detectors showed only
background values of 2,700 to 3,200 counts/min in the
81 keV channel. The mXe concentration in the surge
sphere was 7.2 x 10"2 uCi/cc. The sample of stack
effluent for determining 133Xe discharge rates was not
satisfactory. It was determined later that the
discharged 133Xe was at levels too low to be detected in
the environment with the survey meters.
During the third test, on April 16, 1971, a sampling
station was located on site 0.6 km ESE of the stack at
the same elevation as the plant grounds. The location
was bounded on the NE by the discharge canal and
beyond that by a steep hillside. The sky was overcast in
advance of an air mass moving toward the southeast.
Wind velocities averaged for 10 min intervals varied
from 2 to 8 m/s, with gusts to 11 m/s. Variations in
wind direction for 10 min averages ranged from +10°
to + 80°, with an overall average of + 45°. The
temperature was 7.5° C (45.5° F) at the surface and 6.7°
C (44° F) at the top of the tower. The temperature
gradient, overcast skies, and fast winds in very irregular
terrain characterize the atmosphere as slightly unstable
(Pasquill Category C).(5)
Table 6.1 summarizes the test conditions on April
16. Gas from the surge sphere was released from 0845
to 1200 hours. Krypton-85 and l33Xe emission rates
were determined from a stack effluent gas sample taken
shortly after the release started (see Section 3.1.3). At
1100 hours, the release rate from the sphere was
increased from 1,890 cc/s to 3,300 cc/s. Release rates
for samples*No. 3 and 4 in Table 6.2 were computed
from the ratio of these flow rates and, in the case of
sample No. 3, the fraction of time during the flow rates.
Sampling was twice suspended briefly when it was
evident that the plume had shifted significantly. After
1135 hrs, the plume was no longer detected due to a
major wind shift.
6.2.4 Estimation of atmospheric dispersion
Atmospheric dispersion along the plume centerline at
the sampling point on April 16 was estimated by the
Pasquill-Gifford dispersion equation'5' in Appendix
D.3. Vertical and horizontal plume dispersion
coefficients apply for slight atmospheric instability.
These coefficients, however, are for open and level
ground and sampling times of about 10 min.
Calculations of plume rise above the stack were based
on ASME recommendations,16' also given in Appendix
D.3. These atmospheric dispersion values are listed in
Table 6.2.
6.2.5 Results and discussion. Xenon-133 was
measured in all air samples collected during the test at
levels well above the limits of detection, as shown in
Table 6.2. The highest concentration was in sample No.
2. Krypton-85 was found in both samples analyzed for
that radionuclide, at the same concentration relative to
133Xe as in the sample from the primary vent stack (see
Table 3.8).
The atmospheric dispersion values obtained by
dividing the measured concentrations in ground-level
air by the stack release rates in Table 6.1 agreed with
values for the centerline of the plume calculated by the
Pasquill-Gifford technique for samples No. 1, 3, and 4.
The higher measured concentration value in sample 2
was confirmed by the elevated radiation exposure rates
observed with the survey meters (see Table 6.2). No
Table 6.1
Test Conditions for Sampling Haddam Neck Stack Effluents
at Ground Level on Site, April 16, 1971
Sample number
Sampling interval, hrs.
Duration of plume detection, min
Average wind speed, m/s
Stack release rate, uCi/s
85Kr
133Xe
* except 1003-1005
**except 1050-1055
1
0906-0942
15
3.2
290
1700
2
0951-1029*
35
5.0
290
1700
3
1040-1118**
31
5.1
445
2600
4
1120-1154
16
5.4
500
2900
97
-------
Table 6.2
Airborne Radionuclide Concentrations and Radiation Exposure Rates Measured at
Ground Level on Site During Waste Surge Sphere Release, April 16, 1971
Sample number
Air sample volume, m3 0.47
Measured concentration, uCi/m3
85Kr NA
133Xe 1.4+O.lxlO'2
Atmospheric dispersion (X/Q), s/m3
Measured
8SKr
133Xe
Calculated
Total net counts (81 keV) 24,000
Radiation exposure rate*, uR/hr 0.11
0.47
0.44
0.47
g.lxlCT6
5.5+0. IxlO'3 NA
3.5+O.lxlO'2 1.6+O.lxlO'2
1.9xlO'!
2.0x10-'
7.3x10-'
72,000
0.14
6.3x10-'
7.1x10-'
46,000
0.10
2.9+0. IxlO'3
1.7+0. lxlO'2
5.7x10''
5.7x10"'
7.0x10-*
34,000
0.14
* Normalized to hourly rates, based on duration of plume detection
+_ values are analytical precision at 2-sigma
NA-not analyzed
reason can be given for the reduced dispersion during
this period.
Ground-level air concentrations computed with the
Haddam Neck diffusion model given in Appendices
D. 1 and D.2 were higher than sampling results, from
slightly higher for sample No. 2 to more than 5-fold for
No. 1. The relative concentration value normalized for
wind speed (Xa/Q) of 1.4 x 10"" m'2 for a 0.6-km
distance from the stack, from Appendix D.2, at the
average wind speeds given in Table 6.1 results in
dispersion values (X/Q) from 2.6 x 10"s to 4.5 x 10"'
s/m3. The computed dispersion is increased four-fold
by including the factor for elevated discharges.
The plume was observed with the scintillation
survey meters during the entire test on April 16, 1971.
The count rates given in Figure 6.3 indicate the extent
to which radiation levels fluctuated at the air sampling
location. The net counts in Table 6.2 were calculated by
determining the mean count rate less background in
each 10-second chart interval, averaging the 6 values
for each minute, and then integrating the indicated
periods. Background activity remained constant at
approximately 2,500 counts/min. In general, the
variations in net counts were proportional to the
measured !33Xe concentrations, in that 1000 counts
corresponded to 5 x 10"4 pCi/m3.
The average gamma-ray exposure rate due to the
plume of stack gas while it was overhead was
approximately 0.13 uR/h (see Table 6.2). The rates
were derived from the scintillation detector counting
results given in the table by applying the conversion
factor indicated in Section 6.2.2 and adjusting for the
duration of plume detection during each sampling
interval (see Table 6.1). Thus, for the first sample:
(2.4xl04 counts/0.25 h)/(9x!0s counts/uR) = 0.11 pR/h
The average gamma-ray exposure rate from 133Xe of
0.13 uR/h, at the average 133Xe stack release rate of
2,200 uCi/s during the period of plume detection,
yields a ratio of 5.8 x 10"5 uR/h per uCi/s, or 1 uR per
61 Ci 133Xe. A discharge of 2,000 Ci 133Xe per year
would thus result in an annual gamma-ray exposure of
33 uR to an individual submerged continuously in the
plume 0.6 km distant from the stack under similar
atmospheric conditions. The annual dose from all the
98
-------
6OOO
. 4000
2000
AIR SAMPLING
Period I
_L
Average Background Level
Period 3
Period 4
_L
0900
0920
0940
1000
IO20 1040
Time, hours
1100
1120
1140
1200
Figure 6.3 Scintillation detector response during air sampling on April 16, 1971. (Count rates are
averaged for one-minute periods)
-------
radiations of 133Xe, including conversion electrons, beta
particles, and X-rays, would be 0.2 mrem because the
gamma rays contribute only 17 percent of the dose.a)
This is consistent with the annual dose of 0.3 mrem
given in Section 3.3.13 for the nearest habitation, 0.7
km WNW, based on measured discharges and the
station operator's meteorological model.
The one ground level air sample showing positive
results in the test on September 16, 1970, contained
133Xe at a concentration of 4.1 + 0.2 x 10"3 uCi/m3.
Thus, because of the variability in wind direction, the
133Xe concentration in a sample collected at
approximately the same discharge rate (see Table 3.8)
and distance was 5-fold lower than on April 16. Most
radiation exposure rates measured with the scintillation
detector on September 16 were at background values,
but a few brief elevated readings at location 1 (see
Figure 6.1) showed net values as high as 4,200
counts/min (0.3 uR/h) due to the stack release. The
following brief exposure rates from the plume were also
observed on September 16 at the locations shown in
Figure 6.1:
Location
la
Ib
Ic
Id
le
Distance, km
1.3
1.4
1.7
1.7
1.7
Brief net exposure
rate,uR/h
0.04
0.07
0.17
0.25
0.02
No elevated count rates were observed to the east or
west of the above locations, or at greater distances from
the stack.
The tests demonstrate that, under some conditions,
1) a portable scintillation detector (FIDLER) can be
used to measure directly even the relatively low
radioactivity levels discharged from PWR stacks and 2)
133Xe and 85Kr can be measured in samples of ground-
level air. Under meteorological conditions leading to
the observed dispersion factors, the scintillation
detector can detect, at the fence-line, 133Xe releases as
low as 1,000 pCi/s. The 133Xe analysis is approximately
an order of magnitude more sensitive.
6.3 References
1. Martin, M. J., "Radioactive Atoms - Supplement
I," AEC Rept. ORNL-4923 (1973).
2. Connecticut Yankee Atomic Power Co.,
"Facility Description and Safety Analysis," Vol. I,
AEC Docket No. 50-213-5(1966).
3. Cummings, S. L., R. L. Shearin, and C. R.
Porter, "A Rapid Method for Determining 85Kr in
Environmental Air Samples," in Rapid Methods for
Measuring Radioactivity in the Environment,
International Atomic Energy Agency, Vienna (1971),
p. 163.
4. Karches, G. J. et al, "Field Determination of
Dose from l33Xe in the Plume from a Pressurized Water
Reactor," ibid., p. 515.
5. Turner, D. B., "Workbook of Atmospheric
Dispersion Estimates," EPA Rept. AP-26 (1970).
6. Smith, M., "Recommended Guide for the
Prediction of the Dispersion of Airborne Effluents,"
American Society of Mechanical Engineers, New
York, N.Y.( 1968).
100
-------
7. RADIONUCLIDES AND RADIATION IN
THE TERRESTRIAL ENVIRONMENT
7.1 Introduction
7.1.1 Sampling. Release data by the Haddam Neck
station (see Appendices B.2-B.4) and radioactivity
measurements during this study of airborne effluents
(Section 3) and ground-level air in the immediate
environment (Section 6) suggest that radionuclide
concentrations in ground-level air and deposition on
ground and vegetation due to station operation were
very low. Environmental samples are analyzed for
radioactivity by the Connecticut Department of
Environmental Protection (CDEP) and the station's
contractor for environmental surveillance. a'2) During
1970 and 1971, CDEP performed gross alpha, gross
beta, and some gamma-ray spectral analyses in potable
well water, vegetation, milk and fodder (hay and silage)
and measured MSr and '"Sr in milk and 3H in well water.
The station's contractor measured gross beta activity in
airborne particles; gross alpha and beta activity,
gamma-ray emitters and 3H in well water; gross beta
activity and gamma-ray emitters in vegetation; and
external radiation exposure with thermoluminescent
dosimeters (TLD's).
The following samples and measurements were
obtained in the neighborhood of the Haddam Neck
station (see Figure 7.1):
(1) Four samples of potable well water were
collected for 3H analysis on August 6, 1970,
from homes located 0.8 to 1.5 km from the
station. A control sample was obtained from a
home 4.6 km from the station.
(2) Three food crops—blueberries, lettuce plus
cabbage leaves (the control sample contained
only lettuce), and sweet corn—were collected
near the station and at background locations
for radionuclide analysis.
(3) Milk samples were provided for radionuclide
analysis by CDEP from three dairy farms
located 2 to 4 km from the station, and from a
9.5-km-distant dairy farm as control.
(4) Bovine thyroids for I31I analysis were collected
at slaughter from two nearby cows that had
grazed on a pasture 3.3 km distant from the
station. Control thyroids were obtained from
cattle that had grazed 28 and 18 km distant.
(5) Tissue of two deer killed near Haddam Neck
and two deer killed at distant locations were
compared for radionuclide content.
(6) External radiation exposure was measured
with NaI(Tl) survey meters at 17 points on-site
and at 19 points off-site within 5 km of the
station.
None of the radionuclides found in well water,
milk, food, or deer tissue is attributed to Haddam Neck
station, as described in detail in Sections 7.2, 7.3, 7.4,
and 7.6. The only radionuclide believed to be from the
station was 131I found in one cow's thyroid.
Calculations of expected concentrations of 131I in
bovine thyroids and milk are presented in Section 7.5.2
and Appendices D.4 and D.5. Although grossly
approximate, they demonstrate the procedure, indicate
the magnitude of radionuclide concentrations that may
be attributed to Haddam Neck station, and illustrate
the higher sensitivity of measuring 131I in bovine thyroid
than in milk. The calculations are based on 131I values
measured in airborne effluents (Section 3.3) and
meteorological data from the nearest station of the
National Oceanic and Atmospheric Administration
(NOAA).
The external radiation exposure rate above
background was estimated to be about 1 microroentgen
per hour (uR/h) at a location on the Haddam Neck
station exclusion perimeter, and 0.3 uR/h or less at the
nearest residences, as described in Section 7.7. This
radiation was attributed to gamma rays from
radioactive waste stored at the station.
7.1.2 Environment of Haddam Neck.^ The site
consists of 223 hectare (525 acres) and is bounded by
the Salmon River to the east and the Connecticut River
to the south and west (See Figure 6.1). The elevation of
the plant site is about 3 m (10 ft) at the river and 6.5 m
(21 ft) at the plant yard. Heavily wooded hills, cresting
near 90 m MSL, rise steeply just beyond the northeast
101
-------
Figure 7.1 Terrestrial Sampling Locations
- Vegetables
~ Bovine Thyroid*
- Well Water
-------
perimeter of the yard fence. The elevation of the hills
approaches 120 m northwest of the plant site.
The plant is located in a region with a relatively low
population density. Haddam, about 1.6 km (1 mile)
west across the river, is the nearest community and has
a population of 130.* The largest nearby town is
Middletown, 16 km (10 miles) northwest of the site,
with a population of 37,000. The rural nature of the site
vicinity is suggested below by a population tabulation:'41
Distance from
site (miles)t
1/2
1
5
10
15
20
25
30
40
50
Accumulated
population
11
278
8,550
55,239
160,055
481,111
1,187,352
1,697,308
2,292,932
2,937,764
tl mile = 1.61 km
The nearby directional population within 3 km of the
sitewas:'4'**
Direction Population Percent
5
10
7
11
3
2
7
7
5
6
8
7
15
1
4
4
The population near the Connecticut River fluctuates
because of summer homes and riverside cottages in the
area. Several small resorts are located near East
Haddam, 5 km east of the plant.
The nearest dairy farms in the area are located 2.6
km (1.6 miles) NW and 3.3 km (2.1 miles) ESE of the
site. Other dairy farms in the general area are located
near East Hampton, Killingworth, and Middletown.
Some goat milk is also reported to be produced in the
N
NNE
NE
ENE
E
ESE
SE
SSE
S
SSW
sw
wsw
w
WNW
NW
NNW
98
194
128
220
51
32
134
131
92
108
152
129
288
15
79
86
area. Hay and silage are harvested in the local area; the
fields nearest to the station are at East Haddam. A few
small private gardens, but no truck farms, are located
in the vicinity of the station. The Connecticut River
near and below the station is not used as a public water
supply; drinking water for local residents is taken from
wells. The nearest wells are located at private homes
along Indian Hollow Road, the access road to the
station. Except for UARCO, Inc. at Deep River, 11 km
(7 miles) SE, the nearest industry is located near
Middletown.
7.2 Tritium in Well Water
7.2.1 Sampling and analysis. Duplicate samples of
well water were collected at five private homes on
August 6, 1970 (see Figure 7.1). Three of the samples
were from the homes nearest the station along Indian
Hollow Road, 0.8, 1.1 and 1.5 km NW of the station.
The fourth sample was from a well about 10m from the
far bank of the Connecticut River, 1.2 km SE of the
station. The fifth was a control sample, from a well on
high ground 4.6 km ESE of the station. The samples
were divided and analyzed for tritium by liquid
scintillation techniques at this laboratory15' and by
CDEP.
7.2.2 Results and discussion. All well water samples
contained less than 0.8 pCi/ml of tritium, the 3-sigma
detection limit. The CDEP reported similar values (0.6
to <0.3 pCi/ml). The only elevated 3H values in this
type of sample had been found in one well, located near
the discharge canal on station property, which
contained approximately 4 pCi/ml according to
CDEP"1 and the station's contractor."' Gross
radioactivity levels in well water samples from homes
nearby and at background locations were reported to be
5 to < 1 pCi/liter for alpha and 5 to about 0.5 pCi/liter
for beta."'2' These data suggest that diffusion, if any, of
radioactivity from the discharge canal or river into the
water table supplying off-site wells was below the limit
of detection.
The upper limit of the radiation dose from drinking
3H in well water can be estimated from Appendix E.2.
If the tritium concentration in the well water had been
at the detection limit of 0.8 pCi/ml, an individual
drinking one liter of water per day would have received
a dose to the total body of 0.06 mrem/yr.'"
*Populations are from the 1970 census.
**The population in this area is 1937 and the sum of the percentages is 102 due to rounding off.
103
-------
7.3 Radionuclides in Food Crops
7.3.1 Sampling and analysis* Collecting an
adequate quantity of vegetables for analysis was
difficult because truck farming does not exist near the
station. Small quantities of home-grown produce are
available at roadside stands, although some items for
sale are brought from distant farms.
On August 6, 1970, fresh blueberries were obtained
from a field near the Salmon River, 2.3 km east of the
station, at an elevation of 15 m MSL. On the same date,
sweet corn and lettuce mixed with some cabbage leaves
were collected from a private garden 1.8 km SW of the
station near the top of the ridge, about 75 m MSL (see
Figure 7.1). Because of the late date, the lettuce plants
were in poor condition and approximately an equal
weight of immature cabbage plants had to be added to
obtain a sufficient sample for analysis. Control samples
of blueberries, lettuce and sweet corn that had been
harvested August 5, 1970, at farms 22-34 km NW of
the station were obtained at vegetable stands. No
cabbage was available from regional truck farms at that
time.
The corn was husked and the kernels were removed
from the ears. Prior to analysis, all samples were
washed with tap water, as for food preparation. The
corn husks were also analyzed to ascertain possible
airborne deposition on their surfaces. Samples in both
fresh and ashed (400° C) states were analyzed with a
10- x 10-cm Nal(Tl) detector and multichannel
analyzer for radionuclides that emit gamma rays, and
the ash was analyzed for 90Sr.(S) Tritium and 14C were
determined by treating samples in a combustion train,
collecting water and CO2, and measuring the
radioactivity with liquid scintillation and gas counting
techniques.**
7.3.2 Results and discussion. No photon-emitting
radionuclides attributable to the station were detected
in any of the samples. Only naturally occurring "°K,
226Ra and 232Th and traces of the fallout nuclides U7Cs,
95Zr and 95Nb were detected. As shown in Table 7.1,
there were no significant differences in the content of
3H and 14C in samples collected near the station and in
those from a distance, or, with one exception, in the
content of 90Sr. The 90Sr content was higher in the
nearby sample of mixed lettuce and cabbage than in the
control sample of lettuce, but this may be due to a
higher 90Sr content in the cabbage. The "C specific
activity was consistently higher than in the aquatic
background samples (Tables 5.4 and 5.6).
Analyses of nearby vegetation (weeds, grass, hay,.
silage) in 1970-1971 show only 40K and traces of the
fallout radionuclides 95Zr-95Nb, 103Ru, 106Ru, 137Cs, 140Ba-
140La and 144Ce. The average gross beta activity of
these samples was reported to be about 30 pCi/g ash,<2)
similar to that of the controls.
Table 7.1
Radionuclides in Food Collected August 6, 1970
Sample
Blueberries, near
Blueberries, distant
Lettuce + Cabbage near
Lettuce, distant
Corn kernels, near
Corn kernels, distant
Corn husks, near
Corn husks, distant
Distance
2.3
22
1.8
34
1.8
27
1.8
27
km
km
km
km
km
km
km
km
E
NNW
SW
NNW
SW
NNW
SW
NNW
3K,
pCi/ml HhO
2.6 + 0.6
1.2 + 0.6
NA
NA
<0.9
<0.7
<0.7
<0.9
"C,
PCi/g
7.8
7.8
+
+
C
0.2
0.2
NA
NA
7.4
7.9
8.5
8.1
+
±
+
+
0.2
0.2
0.2
0.2
"Sr,
pCi/g ash
6.6
5.9
5.5
1.5
0.64
0.53
1.4
3.7
+ 0.3
± 0.3
+ 0.1
+ 0.2
+ 0.05
+ 0.04
+ 0.1
+ 0.1
+ values are 2-sigma of counting error; < values are 3-sigma error.
NA - not analyzed.
*We thank Messrs. E. W. Prout and G. G. Curtis, University of Connecticut Cooperative Extension
Service, Haddam, Connecticut, for their advice on regional agricultural practices.
* * We thank E. J. Troianello, EPA, Winchester, Massachusetts, for the "C and JH analyses.
104
-------
7.4 Radionuclides in Milk
7.5Iodine-131 in Bovine Thyroids
7.4.1 Sampling and analysis. On August 4, 1970,
raw milk samples were collected from the dairy farms
described in Table 7.2, at locations shown in Figure 7.1.
Photon-emitting radionuclides were measured by
analyzing 3.5 liters of milk with a 10- x 10-cm Nal(Tl)
detector, and 89Sr and 90Sr were determined by
radiochemical analysis.
-------
paraformaldehyde in plastic bags. The samples were
dissected into small pieces, placed in 9-cm-dia. x 1.5-cm
plastic Petri dishes and counted between two 10- x 10-
cm Nal(Tl) detectors within an annular Nal(Tl)
anticoincidence shield. The detection limit, based on
the 3-sigma deviation observed in a series of
background counts was 2 to 5 pCi per thyroid when
adjusted for radioactive decay.
7.5.2 Predicted concentration in bovine thyroids.
Iodine-131 levels in cows' thyroids and milk were
predicted from release rates at the station stack, the
station's model for dispersion in air,<9) and reported
values for iodine deposition velocity and metabolic
transfer factors for the cow. The dry deposition, D, in
pCi/m2, (it did not rain while the cows were on pasture
just before slaughter) was computed by:
D = Q'o(T/u)(Xn/Q)vd(XL/X) (7.1)
where:
Q'o = release rate of 131I at the stack during
the period of interest, pCi/s
(averaged from Table 3.11)
T = duration of wind toward sampling
location, s
u = velocity of wind toward sampling
location, m/s
Xu/Q = relative concentration normalized for
wind velocity, rri2 (from Appendix
D.2)
vd = deposition velocity of 131I on grass, m/s
Xi/X = ratio of average long-term
concentration in sector ground-level
air to that at centerline.
The values of T and u were obtained from
climatological data at the NOAH meteorological
station at Bradley Field, Hartford.(10) To simplify the
wind frequency calculations, a 90° sector (270° to 360°)
was used, allowing for channeling effects of the valley
while also encompassing both pasture locations, but the
duration value was then divided by 4 to apply the
exposure to a 22.5° sector. A deposition velocity of 0.01
m/s was used."1'12' The ratio XL/X was taken to be 0.5,
based on the discussion in Appendix D.I. The average
value of Xu/Q at a 3.3-km distance is 2.5 x 10"5 m"2
according to Appendix D.2. The 13II in daily deposition
was computed to decay with a 5-day environmental
half-life04' to estimate day-by-day concentrations of 131I
on grass shown in Appendix D.4.
The net cumulative concentration of 131I in the two
thyroids was calculated by assuming that each cow
grazed effectively 45 m2 of pasture each day,"3' and that
20 percent of the iodine intake was taken up by the
thyroid, where it decayed with an effective half-life of 7
days."3' These calculations are shown in Appendix
D.5 and the results, in Table 7.3.
Although no milk was collected from these cows at
this time, hypothetical concentrations of 131I in the milk
were computed in Appendix D.5 for a series of single
doses according to the graph by Garner and
Russell."3' Thyroid and milk detection sensitivities for
131I are compared for these circumstances in Table 7.3.
The ratios of I31I per liter of milk to 131I per gram of
thyroid, at thyroid weights of 50 g, are less than the
reported ratio of 1:12."5>
7.5.3 Results and discussion. Iodine-131
attributable to reactor operation was detected only in
the-thyroid of the cow that had grazed 3.3 km ESE of
the station until September 14, 1970 (see Table 7.4).
The measured 131I burden in this thyroid was 21 + 3 (2-
sigma) pCi, or 0.44 pCi/g. The level predicted from the
station diffusion model was six times as great (see Table
7.3). The undetectably low 131I content of the thyroid
collected June 17, 1971, was also less than the predicted
value. The 131I concentration would be overestimated
because the elevation of the discharge point was not
considered (see Section 6.2.5). Another reason for the
discrepancy is the use of meteorological data from the
distant and dissimilar location at Hartford. The
computed concentrations of 131I in milk from these
cows were well below the detection limits shown in
Table 7.2.
According to the Federal Radiation Council, a
daily intake of 80 pCi 13II will result in a dose to a
child's thyroid of 500 mrem/yr."61 At a daily intake of
1 liter of milk with the estimated 131I concentration of
0.44 x 1/12 = 0.04 pCi/liter, the dose to the thyroid
would be 0.2 mrem/yr.
7.6Radionuclides in Deer
7.6.1 Sampling and analysis. To begin evaluation of
the radionuclide content in wildlife, four white-tail deer
(Odocoileus virginianus) were collected, two within 5
km of the reactor, and two at a distance of 21 km. The
deer are described in Table 7.5. Deer D-l and D-2 were
killed in automobile accidents on Highway 2, D-4 was
also killed in an automobile accident at the intersection
of Highways 196 and 151, and D-3 was shot at Haddam
Neck. Samples of bone (femur), muscle, liver and
kidney were preserved in plastic bags on dry ice. No
bone sample was supplied for deer D-3, and, although
requested, rumen content—a direct indicator of recent
radionuclide intake with food—was not collected from
any deer. Muscle, liver and kidney samples were ashed
at 400°C and analyzed for photon-emitting
radionuclides by spectrometry with a 10- x 10-cm
106
-------
Table 7.3
Estimated Levels of 13'I in Cow Thyroids
Location/
collection Qo'*
period pCi/sec
3.3 km ESE,
Sept. 1970 820
Thyroid level,** pCi
Hypothetical milk
•"•
concentration, pCi/1
Concentration ratio
thyroid/milk,t
Generalized Qo' as shown Generalized Qo' as shown pCi/g per pCi/liter
0.16 Qo'
130
7.8 x 10's Qo' 6.4 x 10''
41
3.3 km SE,
June 1971
58
0.062 Q«'
3.6
7.6 x 10'5 Qo' 4.4 x 10-3
16
* The Sept. 1970 value is the average of 6 I31I concentrations (Table 3.11) measured during continuous
discharges in July-Sept. 1970, multiplied by the flow rate of 16.7 mVsec; the June 1971 value is the
average of 3 '"I concentrations (see footnote to Table 3.11) measured in June 1971, multiplied
by 16.7 mVsec.
** X u/Q = 2.5 x 10'5 m'2 (from Appendix D.2).
t A thyroid was assumed to weigh 50 g.
Table 7.4
"I in Bovine Thyroids
Animal
type
Cow
Cow
Cow
Cow
Ox
Ox
location
3.3
28
28
3.3
18
18
km
km
km
kin
km
km
ESE
WSW
WSW
SE
E
E
Last day on
pasture
Sept.
Sept.
Sept.
June
June
June
14,
14,
14,
3,
14,
14,
1970
1970
1970
1971
1971
1971
Thyroid
weight, g
49
50
33
75
29
52
"'I content,
pCi/ thyroid
21 ±
<3
<5
<2
<2
<2
*
3
Notes:
1. Thyroid samples contained variable amount of fat.
2. + value based on 2-sigma of observed standard deviation in a
series of background counts; < values are + 3-sigma.
*Corrected for physical and biological decay of 131I to last day on pasture.
Table 7.5
Description of Sampled Deer*
Distance and
Deer
No.
D-l
D-2
D-3
D-4
direction from
Location
E. Gastonbury
E. Gastonbury
Haddam Neck
Wopowog
reactor, km
21
21
1.2
4.8
NNW
NNW
N
NNW
Date
collected,
1970
Nov.
Nov.
Aug.
Sept.
13
13
15
12
Age,
yrs
3.5
0.5
2.5
2.5
Sex
F
F
M
F
*We thank Edward Goldin, Game Biologist, Connecticut Board of Fisheries and Game, for
collecting the deer and determining their ages.
107
-------
Nal(Tl) detector or an 11-cm3 Ge(Li) detector.
Samples were also analyzed with a Nal(Tl) gamma-ray
coincidence/anticoincidence system. Bone samples
were ashed at 600° C. Bone and muscle samples were
analyzed for radiostrontium by radiochemical analysis,
and for stable calcium and strontium with an atomic
absorption spectrophotometer.
7.6.2 Results and discussion. The only
radionuclides detected in the deer samples were 90Sr,
I37Cs, and naturally occurring 40K, as shown in Table
7.6. The 58Co and '°Co content was found to be <40
pCi/kg each (3-sigma counting error) in liver and
kidney samples. The average concentrations of 90Sr and
131 Cs in the deer collected near the plant site were lower
than those in the deer collected at a distance.
The average 137Cs concentration in muscle was
2,400 + 1,000 (+ 1-sigma) pCi/kg or 740 + 140
pCi/g potassium. No 134Cs was detected; the minimum
detectable concentration was 20 pCi/kg at the 3-sigma
level. In the muscle, the average MSr concentrations
were 8+1 pCi/kg, 43 + 4 pCi/g calcium and 69+14
pCi/mg strontium. The concentrations relative to
calcium and strontium are similar to those in bone.
The 137Cs concentrations in the muscle of these deer
are somewhat higher than observed previously in deer
collected in the vicinity of the Yankee-Rowe and
Dresden nuclear power stations/5'8' but the levels are
not high compared to deer muscle collected at some
areas distant from nuclear power stations.07"1" Jenkins
and Fendley reported numerous cases in which levels of
137Cs in the muscle of white-tail deer from the
southeastern United States approach 150,000 pCi/kg
as a result of the concentration of fallout 137Cs in certain
types of vegetation.'17'1"
The average '°Sr concentration in deer bone was
10,000 + 8,000 pCi/kg, 62 + 32 pCi/g calcium and 83
+ 54 pCi/mg strontium. These concentrations are
similar to values observed in deer collected in 1969
from western Massachusetts/5' and also to
concentrations reported to be in deer from South
Carolina, Colorado and California/18'20"22'
The average '"Sr concentration of 8 pCi/kg deer
meat is only one-seventh of that reported for Alaskan
caribou or reindeer meat/23' but is about 8 times that in
meat sampled as a typical component of New York
City diets on three occasions in 1970/24' The average
stable and radiostrontium concentrations in deer
muscle were approximately 1/1,200 of the
concentration found in bones, similar to the ratio of
1/1,500 observed in deer samples collected from
western Massachusetts in 1969/5'
Because of the wide range of measured
concentrations among samples, more samples would
have to be collected and analyzed if radionuclide
concentrations nearby and at a distance had to be
compared with better precision. This problem of a wide
range of concentration values is common among
environmental samples, but is especially serious in
animals whose radionuclide contents can be affected by
their mobility.
7.6.3 Estimated radiation dose from eating deer
meat. The radiation dose a person could receive from
eating deer meat was estimated from daily intake-dose
rate relationships as discussed in Section 5.5.5.
According to these, a dose rate of 170 mrem/yr to the
bone or whole body will result from a daily intake of
200 pCi 90Sr or 15,000 pCi 137Cs. At the average "Sr
Table 7.6
Radionuclide (pCi/kg)* and Stable Ion Concentration (g/kg)* in Deer Samples
Sample
type
Muscle
Bone
Distant
Nuclide
"7Cs
'"Sr
K
Sr
Ca
ash wt./wet wt.
'°Sr
Sr
Ca
ash wt./wet wt.
D-l
3740 + 30
8.4 + 0.8
3.96
0.00012
0.18
0.009
4300 + 120
0.075
117
0.34
D-2
2,030 + 20
8.8 + 0.8
3.21
0.00010
0.19
0.010
15,700 + 400
0.108
159
0.40
Nearby
D-3
1320 + 50
6.0 + 0.6
2.00
0.00010
0.16
0.010
NS
NS
NS
NS
D-4
2550 + 20
6.9 + 0.6
3.59
0.00012
0.17
0.009
9600 + 200
0.211
186
0.47
kg wet weight
Note: + values are 2(7 counting error.
NS - no sample was available.
108
-------
concentration of 8 pCi/kg meat and a maximum
annual consumption by an individual of 45 kg (0.12
kg/day),(25) the radiation dose to bone marrow is 0.8
mrem/yr. At the same consumption rate and an
average 137Cs concentration of 2,400 pCi/kg muscle, the
radiation dose from mCs to the whole body is 3.3
mrem/yr. To the average deer hunter who consumes an
estimated 2.7 kg/yr deer meat/251 the doses are 6
percent of those calculated above. These doses are
believed to be from radionuclides in fallout, but provide
an upper limit if some of the radioactivity in deer were
from the station.
7.7External Gamma Radiation
7.7.1 Detection instruments. Radiation exposure
rates were measured with cylindrical Nal(Tl) gamma-
ray detectors (5-cm diameter x 5-cm length) connected
to portable count-rate meters. The instruments had
been calibrated by comparing their count rates for
gamma rays above 80 keV in the natural radiation
background at Cincinnati with measurements by a
muscle-equivalent ionization chamber and Shonka
electrometer. Radiation levels during calibration
ranged from 5 pR/h over water in a lake to 19 uR/h
over granite. The count rate, C (in counts/min), of the
survey instruments varied linearly with the radiation
exposure rate, R (in uR/h), of the ionization chamber;
a typical calibration curve had the equation R = 7.0 x
lO^C + 3.3. Radiation exposure rates at the
measurement locations near Haddam Neck 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 had been found
applicable in a variety of natural radiation
backgrounds. In numerous measurements, the
standard error of the survey meters was + 0.35 uR/h,
and the exposure values computed from the readings
were within 4 percent of the values measured with the
ionization chamber in 95 percent of the
measurements.'26'
7.7.2 Measurements. The 36 radiation
measurement locations shown in Figures 7.2 and 7.3
were selected for the following reasons:
(1) Point No. 1 was investigated as a source of
high background radiation due to a large
granite outcropping at a location remote from
the station;
(2) Sixteen points, Nos. 2, 4-9 and 11-19, 0.5 km
to 2.4 km distant from the center of the station,
provided radiation exposures in the immediate
environment of the station;
(3) Point Nos. 3 and 10 were considered to be
sufficiently distant from Haddam Neck station
but similar in natural radiation to yield
terrestrial background values for comparing
with and subtracting from exposure rates near
the station;
(4) Seventeen points on site, but outside the
security fence, Nos. 20-36, were intended to
aid in identifying the source of external
radiation from Haddam Neck station and to
check off-site exposure values by extrapolating
from these higher, more precise
measurements.
Most exposure rates were measured on two
occasions while the station was operating at full power.
Detectors were held 1 m above ground surface. Count
rates ranged from 5,000 to 100,000 counts/min.
7.7.3 Results and discussion. The gross radiation
exposure rates (which include the natural radiation
background) in Table 7.7 range from 6.9 to 10.2 uR/h
in the immediate environs, and up to 69 uR/h on-site at
the measurement locations. Radiation values at
background locations 3 and 10 averaged 8.3 + 0.5
uR/h, and all off-site values averaged 8.1 -j- 0.8 (1-
sigma) uR/h. Because of differences in the natural
background at various locations, none of the measured
exposure rates off-site could be shown to include
radiation from the station.
The higher radiation exposure rates on-site are
attributed primarily to direct radiation from stored
radioactive waste. This explanation was supported
qualitatively by the general decrease in exposure rates
with distance from the boron waste storage tanks, and
the lower values where buildings provided shielding.
Since all measurements were taken while the reactor
was operating, it is not possible to eliminate direct
radiation from the reactor as a possible source. The
responses of the survey meters in moving from location
to location suggested that the higher exposure rates
were not due to higher radiation background,
radionuclides from the station deposited on the ground,
or radiation from airborne radioactive effluents.
An attempt was made to evaluate the exposure rates
off-site by extrapolating from the higher values
measured on-site. The distance of each location from
the boron waste storage tanks, at the north corner of
the plant, appeared to be the critical measurement in
correlating the dose and distance measurements. The
values were extrapolated to off-site locations by the
equation:
109
-------
Figure 72 Locations of Off-Site Radiation Exposure Measurements
with Survey Meters
-------
\
\
SALMON
RIVER
CONNECTICUT RIVER
0 100 200 300 400
i 1 1 1 1
meters
Figure 7.3 Locations of On-Site Radiation Exposure Measurements with Survey Meters
-------
Table 7.7
External Radiation Exposure Rates Near Haddam Neck
Location
No.
Distance "
Exposure rate on
July 22, 1970, uR/hr
Location
No.
Exposure rate on
Distance March 16, 1971, nR/hr
Off-site
1
2
3
4
5
6
7
8
9
10
11
12
13
14
15
16
17
18
19
5.0 km
0.5 km
3.7 km
1.0 km
1.2 km
1.6 km
2.3 km
2.0 km
2.1 km
3.0 km
2.4 km
1.3 km
1.0 km
1.0 km
1.5 km
0.7 km
0.8 km
1.5 km
0.6 km
NNW
NW
NNW
NNE
N
N
NNW
ENE
E
ESE
SE
SE
SSE
ssw
w
wsw
w
WNW
NW
10.2
8.1
9.0
8.7
8.8
8.0
8.3
8.6
8.2
7.9
7.8
7.1
7.2
6.9
8.6
8.3
7.9
7.5
7.0
8.3
±
+
+
±
+
+
±
±
±
+
+
±
±
+
±
±
+
+
±
±
0.4"
0.1
0.5 c
.0.4
0.0
0.1
0.2
0.3
0.1
0.4
0.5
0.2
0.3
0.2
0.2
0.2
0.3
0.1
0.3
0.2°
20
21
22
23
24
25
26
27
28
29
30
31
32
33
34
35
36
0.18 km
0.20 km
0.22 km
0.21 km
0.24 km
0.28 km
0.28 km
0.31 km
0.38 km
0.20 km
0.14 km
0.13 km
0.12 km
0.18 km
0.31 km
0.39 km
0.47 km
On-site
NW
WNW
NW
NW
WNW
NW
NW
WNW
WNW
W
WSW
sw
ssw
s
SSE
SSE
SE
69
38
34
30
21
24
18.3
15.0
12.1
10.1
26
32
24
41
55
31
14.8
10.1
+
±
±
±
±
±
±
+
±
±
±
±
+
+
±
±
±
±
9
7
3
4
2
1"
0.8
0.4
0.2
0.5
1
1
1
1
3
2
0.5
0.3
" Distance from center of reactor containment
b Exposure rates are averages of 2 to 8 measurements;
the range for 2 measurements or 2ff values for
c Measured on March 16, 1971
" Measured on July 22, 1970
R = 1.4DJexp(-4D) (7.2)
where R is the net radiation exposure rate (background
subtracted) in uR/h and D is the distance in kilometers
from the boron waste storage tanks. The constant of 1.4
in this equation was obtained from the net exposure
rates at locations 21-28 by a least-squares evaluation of
the data. The exponential constant of 4 accounts for the
attenuation of the gamma radiation in air.
The extrapolated values are as follows:
J; values are 1/2 of
more than 2 measurements.
The extrapolated radiation rates were 1 uR/h or less at
the site boundary (except on the Connecticut River)
and 0.2-0.3 uR/h at the nearest residences, both NW of
the site and across the river in Haddam. These values
are consistent with the measured gross values, but are
uncertain because the extrapolation is overly simple in
assuming the same relation in all directions and
ignoring radiation shielding and scattering. The
exposure rate on the river, 100 meters offshore from the
No.
Location
2 (NW perimeter)
19 (NW nearest residence)
14 (Haddam)
River in front of plant
(100 m from shore)
15 (Haddam)
16 (Haddam)
17 (Haddam)
Distance from boron waste
waste storage tanks, m
460
640
1000
240
1500
700
800
Extrapolated
exposure rate, iiR/h
1.1
0.3
0.1
9.3
0.03
0.2
0.2
112
-------
station, was estimated to be approximately 9 uR/h
above background.
The station's contractor for environmental
surveillance has reported12' gross gamma radiation
exposures averaging from 22 to 26 uR/h at the station
and eight neighboring towns. These TLD values seem
questionable in view of the lower values measured in
this study and the generally encountered lower levels of
background radiation.
It would be of interest to obtain additional
measurements of external radiation exposure in the
environs of the station to check the presented data with
regard to instrument calibration, background
subtraction, and possible correlation with station
operation, including the radiation levels of the stored
radioactive waste. Long-term measurements of
continuous gamma-ray sources to detect levels 0.3
uR/h above background do not appear feasible even
with sensitive detectors, since it would be difficult to
distinguish between radiation from the station and the
natural background.
7.7.4 Estimated external radiation exposure to
persons in the environs. The instantaneous exposure
rate from the station of 0.3 uR/h at the nearest
location, computed in Section 7.7.3, equals 2.6 mR/yr.
The estimated exposure to individuals in Haddam
would then be 2 mR/yr or less. These values are subject
to considerable uncertainties in measurement and
calculation, and will be reduced by shielding by house
walls and time spent by persons at less exposed
locations. In comparison, the natural radiation
background was approximately 70 mR/yr and its
variation was much greater than the exposure
attributed to the station. Persons in boats on the river
less than 1 km from the plant would be subject to
higher exposure rates, but for shorter periods of time.
In 100 hours per year at a location 100 meters from the
nearest shore line in front of the station, the annual
dose would be 1 mR. The set of measurements suggests
that the radiation exposure from radioactive wastes
stored at the station was essentially zero at distances of
2 km and more.
The exposure due to direct radiation from the boron
waste storage tanks was estimated in the
Environmental Statement to be 5.5 mrem/yr at the site
boundary and about 0.9 mrem/yr at the nearest
residence.'4' This was calculated from an assumed
radionuclide inventory in the tanks and the shielding
thickness of the tank walls. A similar calculation for
direct exposure from the waste gas surge sphere
resulted in an estimate of 6 mrem/yr at the nearest
boundary. Due to the different locations of the storage
tanks and the waste gas surge sphere, no off-site
location would be exposed to the sum of these maxima.
7.8 References
1. Connecticut Department of Environmental
Protection, "Radiological Data of Environmental
Surveillance - Year 1970," Hartford, Conn. 06115
(1971); also for year 1971.
2. Combustion Engineering Combustion Division,
"Operational Environmental Radiation Monitoring
Program, Connecticut Yankee Atomic Power
Company Summary Report, 1970."
3. Connecticut Yankee Atomic Power Company,
"Haddam Neck Plant Environmental Report-
Operating License Stage," AEC Docket No. 50-213
(1972).
4. Directorate of Licensing, "Final Environmental
Statement Related to the Haddam Neck (Connecticut
Yankee) Nuclear Power Plant," AEC Docket No.
50-213(1973).
5. Kahn, B., et al., "Radiological Surveillance
Studies at a Pressurized Water Nuclear Power
Reactor," EPARept. RD 71-1 (1971).
6. Blanchard, R. L. and B. Kahn, "Pathways for the
Transfer of Radionuclides from Nuclear Power
Reactors Through the Environment to Man," in
Radioecology Applied to the Protection of Man and
His Environment, Commission of European
Communities, Luxembourg, (1972), p. 175.
7. "Milk Surveillance, July 1970" and "Milk
Surveillance, August 1970," Radiol. Health Data Rept.
11, 617 and 673 (1970).
8. Kahn, B. et al., "Radiological Surveillance
Studies at a Boiling Water Nuclear Power Reactor," U.
S. Public Health Service Rept. BRH/DER 70-1
(1970).
9. Connecticut Yankee Atomic Power Company,
"Preliminary Safety Analysis Report," NYO-3250-5,
Vol. II, Fig. 2.2-11(1966).
10. U. S- Department of Commerce, ESSA, "Local
Climatological Data, Hartford, Connecticut (Bradley
International Airport)," Aug. - Sept. 1970 and May -
June 1971.
11. Van der Hoven, I., "Deposition of Particles and
Gases," in Meteorology and Atomic Energy, D. H.
Slade, ed., AEC Rept. TID-24190 (1968), p. 206.
12. Bryant, P. M., "Derivation of Working Limits
for Continuous Release Rates of 90Sr and 137Cs to
Atmosphere in a Milk Producing Area," Health Phys.
72,1393(1966).
113
-------
13. Garner, R. and R. S. Russell, "Isotopes of
Iodine," in Radioactivity and Human Diet, R. S.
Russell, ed., Pergamon Press, Glasgow, (1966), pp.
302-303, 305.
14. Koranda, J. J., "Agricultural Factors Affecting
the Daily Intake of Fresh Fallout by Dairy Cows,"
AEC Rept. UCRL-12479 (1965), pp. 20 and 3la.
15. Falter, K. H. and G. Murray, "Measurement of
ml in Bovine Thyroids," Radiol. Health Data Rept. 6,
451(1965).
16. "Background Material for the Development of
Radiation Protection Standards," Fed. Rad. Council
Rept. No. 2, US Government Printing Office,
Washington, D. C. 20402, (1961).
17. 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).
18. Rabon, E. W., "Some Seasonal and
Physiological Effects on 137Cs and 89'90Sr Content of the
White-Tailed Deer, Odocoileus virginianus," Health
Phys. 15, 37 (1968).
19. Whicker, F. W., G. C. Farris, E. E. Remmenga
and A. H. Dahl, "Factors Influencing the
Accumulation of Fallout 137Cs in Colorado Mule
Deer," Health Phys. 11, 1407(1965).
20. Whicker, F. W., G. C. Farris and A. H. Dahl,
"Concentration Patterns of 90Sr, 137Cs and 131I in a Wild
Deer Population and Environment," in
Radioecological Concentration Processes, B. Aberg
and F. P. Hungate, eds., Pergamon Press, Oxford
(1967), p. 621.
21. Longhurst, W. M., M. Goldman and R. J. Delia
Rosa, "Comparison of the Environmental and
Biological Factors Affecting the Accumulation of 90Sr
and I37Cs in Deer and Sheep," ibid, p. 635.
22. French, N. R. and H. D. Bissell, "Strontium-90
in California Mule Deer," Health Phys. 14, 489 (1968).
23. Chandler, R. P. and D. R. Snavely, "Summary
of 137Cs and 90Sr Concentrations Reported in Certain
Alaskan Populations and Foodstuffs, 1961-1965,"
Radiol. Health Data Rept. 7, 675 (1966).
24. Health and Safety Laboratory, "Strontium-90
in Tri-city Diets, January-December 1970," Radiol.
Health Data Rept. 12, 568 (1971).
25. Magno, P. J., "Studies of Dose Pathways from a
Nuclear Fuel Reprocessing Plant," in Environmental
Behavior of Radionuclides Released in the Nuclear
Industry, International Atomic Energy Agency,
Vienna, 537 (1973).
26. Levin, S. G., R. K. Stoms, E. Kuerze and W.
Huskisson, "Summary of National Environmental
Gamma Radiation Using a Calibrated Portable
Scintillation Counter," Radiol. Health Data Rept. 9,
679(1968).
114
-------
8. SUMMARY AND CONCLUSIONS
8.1 Radionuclides in Effluents from the
Haddam Neck Station
Radionuclides were discharged by numerous
pathways in small amounts relative to effluent limits.
The largest constituents among radioactive effluents
were 3H, mostly in liquid waste, and 133Xe, mostly in
airborne waste. These observations appear to be
generally applicable to large PWR nuclear power
stations, except that less 3H is discharged when the fuel
is clad in Zircaloy instead of stainless steel. Lesser
discharges of many radionuclides, including 133Xe, have
been predicted when additional waste treatment is
discharges reported by the station operator and the
discharge estimates presented in the Environmental
Statement for the station; any differences in individual
values are discussed in Sections 3 and 4. The largest
discrepancy concerns calculations based on a 'model'
plant: the amounts of discharged gaseous radionuclides
are similar to those predicted for 0.25 percent of fuel
elements releasing radionuclides to the coolant,
although this value was only 0.02 percent at the
Haddam Neck station.
The estimated amounts of radionuclides in airborne
effluents during the second half of 1970 and the first
half of 1971 are as follows:
Radionuclides in airborne effluents, Ci/yr
Radionuclide
3H
"C
"-Kr
"Kr
"Kr
"Kr
1M-Xe
'"Xe
'"Xe
'"I
Long-lived
particulate
note: na =
(1)
Waste gas
surge sphere
0.007
0.032
na
29.
na
na
0.4
130.
0.50
0.0007
0.0003
not analyzed
(2)
Vapor
container
air
0.18
0.18
na
77.
na
na
0.3
120.
na
na
na
(3)
Primary
auxiliary
bldg air
4.3
<2.
na
9.1
na
na
na
650.
na
na
na
(4)
Steam jet
air ejector
0.081
0.051
6.6
12.
10.
17.
18.
1,100.
58.
na
na
(5)
Fuel bldg
air
2.3
0.3
na
0.8
na
na
na
<12.
na
na
na
(6)
Primary
vent
stack
12.
<3.
na
26.
na
na
na
1,500.
na
0.013
0.003
(7)
Turbine
bldg air
150.
<10.
na
43.
na
na
na
<400.
na
na
na
applied to meet 'as low as practicable' criteria for
design objectives.
Results of the effluent measurements in this study
are summarized below, based on the information in
Sections 3 and 4. For simplicity, they are given as
annual releases. Because these values were obtained by
occasional sampling, they should be considered only
indications of the magnitude of radionuclide
discharges. Exact values must be derived from frequent
or continuous measurements at the many waste
streams or discharge locations. The totals (see Sections
3.3.13 and 4.3.3) are comparable to the annual
The values in data column 6 include the same
wastes as those in columns 3, 4, and 5, and also
discharges from the blowdown flash tank, which could
not be measured separately. The 3H and 14C values are
for all forms of the radionuclides, but distinctions
between tritiated water and gases, and between 14C in
CCh and other gases, are made in Section 3 for several
of the waste streams. A small amount of41 Ar was also
observed in one of the gaseous wastes. The presence of
83mKr, 89Kr, 13""Xe, 135mXe, 137Xe, and 138Xe in curie
amounts was inferred, although these radionuclides
could not be measured. Short-lived progeny of noble
115
-------
gases, such as 88Rb, and relatively short-lived iodine
isotopes, such as 133I and 135I, were also indicated to be
present. Some of the listed values in ventilating air were
noted to be uncertain because too few samples were
collected. A few totals are incomplete because a
contributing waste stream could not be sampled.
The tabulation suggests that the usual program of
radioactive gas measurements at the air ejector and in
waste tanks can account for a large fraction of the
discharges. Discharged ventilating air, however, also
carries radioactivity by a variety of pathways. The
radionuclide compositions of the various effluents
differ widely.
The estimated amounts of radionuclides in liquid
effluents during the same period are as follows:
Radionuclides in liquid effluents, Ci/yr
Reactor system Secondary system waste
Radionuclide
]H
"C
"Na
"P
s'Cr
"Mn
"Fe
!9Fe
"Co
"Co
"Co
"Ni
90Sr
"Zr
"Nb
"Mo
"°™Ag
"'I
'"I
'"I
133Xe
1MCs
136Cs
"'Cs
waste
8 x 103
3 x 102
ND
5 x 10"
3 x 103
1 x 10'
4 x 10'
8 x 10°
6 x 10 J
4 x 10 '
2 x 10'
6 x 10 3
4 x 10"
5 x 10"
3 x 10J
<1 x 10J
3 x 103
4
1 x 10'
ND
2 x 10'
1 x 10 2
<1 x Iff3
2 x 102
blowdown leakage
6 x 10' 2 x 102
6 x 10"
2 x Iff2
7 x 103
2 x Iff3
2 x Iff3
4 x Iff'
(6 x 10")
(2 x Iff4)
2 x Iff2
1 x 10°
(2 x 103)
1 x 10"
2 x 10"
(2 x 10")
3 x 102
ND
2
2
8 x 10 '
ND
8 x 102
2 x Iff2
8 x 10'!
note: ND-not detected
Tritium was assumed to be the only radionuclide in
water leaking from the secondary system, although no
confirmatory measurements were made. The values in
parentheses were inferred as described in Section 4.3.3.
Unlisted radionuclides, such as 89Sr, were not detected
in any effluent samples; less-than values for some are
given in Section 4.3.3.
The bulk of the effluent radioactivity was
discharged from the reactor waste system. A few
radionuclides, however, were in higher amounts in
secondary system blowdown, which is discharged
without storage or treatment.
8.2 Radionuclides in the En vironment at
the Haddam Neck Station
Radionuclides at low concentrations from the
station were found in various media sampled in the
aquatic environment:
(1) The radionuclides JH and 131I were in water at
concentrations of approximately 10 pCi/ml
and 1 pCi/liter, respectively. The
radionuclides S8Co and 134Cs were at
concentrations between 0.1 and 1 pCi/liter.
Samples with these contents were obtained in
the coolant water discharge canal and within a
few kilometers of its mouth in the Connecticut
River (see Sections 4.3.4 and 5.2).
(2) Numerous radionuclides were in algae,
plankton, and aquatic plants collected at the
mouth of the canal and nearby in the
Connecticut River. Iron-55 had the highest
concentration in these samples, at 41 pCi/g
wet weight (see Section 5.4).
(3) Fish caught in the canal contained 3H, 14C, 13T,
134Cs, and 137Cs. The highest concentration, of
JH, was 2.9 pCi/g wet weight of tissue,
compared to a background value of 0.6 pCi/g.
One sample of shad, which swim up the
Connecticut River for a brief period to spawn,
showed a similar increase in 3H concentration,
but contained no other radionuclides
attributed to the station (see Section 5.5).
Oysters and clams from the mouth of the
Connecticut River had no elevated levels of
radionuclides (see Section 5.6). No shellfish
were found in or near the coolant canal,
although their presence had been reported.
(4) Sediment from some locations that accumulate
silt along the east bank of the Connecticut
River at and just above and below the mouth of
the canal contained 54Mn, "Co, 58Co, 60Co,
134Cs, and 137Cs attributed to station effluents.
The presence of 58Co was usually most
apparent; its highest concentration was 13.5
pCi/g (see Section 5.7).
Radioactive effluents and direct radiation from the
station could not be readily detected in the terrestrial
environment. No radionuclides attributed to the station
were found in well water from just beyond the station
boundary, in vegetables from nearby gardens, in milk
from nearby dairy farms, or in the meat of deer killed
near the station (see Sections 7.2, 7.3, 7.4, and 7.6). The
following special measurements showed radionuclides
or radiations from the station in the environment:
116
-------
(1) The thyroid of one dairy cow that had grazed
on a hill 3.3 km distant from the station
contained 21 pCi of ml (0.4 pCi/g thyroid) at
the time when the U1I discharge from the stack
was approximately 10° uCi/s (see Section 7.5).
A second thyroid, from a cow that had grazed
at the same distance but at a lower elevation
and during lesser U1I discharges, contained no
detectable 131I(<2pCi).
(2) While gas from the surge sphere was
discharged for this purpose, the presence of
Xe in ground-level air was observed with a
large, thin NaI(Tl) detector (FIDLER) with
pulse height discrimination to count the
characteristic 81-keV gamma rays.
Measurements of 133Xe at a stack discharge
rate of approximately 2,000 uCi/s were
possible on site. By collecting 0.5 m3 of air,
Xe and 85Kr were detected in ground level air
at concentrations of 2 x 10"2 uCi/m3 and 4 x 10"3
uCi/m3, respectively. Indications of 133Xe
during releases from the waste gas surge
sphere could occasionally be obtained off-site
with the FIDLER survey instrument. Such
measurements were not sufficiently sensitive to
detect the much lower amounts of gaseous
radioactivity discharged continuously from
the stack (see Section 6).
(3) Measurements with survey meters beyond the
station boundary showed no observable
increase over the background radiation
exposure of approximately 8 pR/h (70
mR/yr). Extrapolation of elevated radiation
exposures within the boundary suggested that
the highest exposure rate at nearby habitations
was 0.3 uR/h (2.6 mR/yr) due to direct
radiation from the station. The exposure rate
was estimated to be lower at the nearest
population center, but higher on the
Connecticut River, where persons would be
exposed only briefly (see Section 7.7). The
source of the radiation is believed to be stored
radioactive waste.
On the basis of the observed effluent and on-site
measurements, the highest population radiation doses
were computed to be from consuming fish caught in
and near the coolant-water discharge canal and from
external radiation due to stored wastes and gaseous
discharges:
(1) fish consumption (Section 5.5.5) may have
resulted in 3 mrem/yr to bone, 6 mrem/yr to
thyroid, 0.4 mrem/yr to GI tract, and 0.3
mrem/yr to the total body;
(2) direct radiation (Section 7.7.4) may have
resulted in 3 mrem/yr to the total body;
(3) airborne discharges (Section 3.3.14) may have
resulted in 0.5 mrem/yr to the total body.
The computations—particularly those of the dose from
fish consumption—utilized several assumptions that
require checking. The external radiation doses would
be lower if adjusted for shielding and occupancy factor.
8.3 Monitoring Procedures
The following procedures were demonstrated in
this and previous studies for monitoring effluents and
environments of PWR stations:
(1) measurement of effluent radionuclides other
than the long-lived ones readily detected by a
gamma-ray spectrometer; of particular
interest, in addition to usually measured 3H,
"Sr, and 90Sr, are 14C, 32P, and "Fe;
(2) measurement of 3H and 14C in their various
gaseous species. Other recent studies suggest
the inclusion of species measurements for
radioiodine in air and radiocobalt in the
aquatic environment;
(3) surveillance of sediment with submersible
gamma-ray detectors to indicate "hot spots"
for detailed sampling and analysis;
(4) use of concentration devices to collect ionic
and insoluble radionuclides from water for
measurements at concentrations of 10'10
uCi/ml;
(5) use of bovine thyroids to detect 13T at very low
concentrations (equivalent to 0.02 pCi/liter
milk) in the terrestrial environment;
(6) use of specialized survey instruments for
detecting low levels of 133Xe (400 pCi/m3) in
ground-level air;
(7) collection of fish that are under conditions of
relatively restricted mobility in the
environment to study their uptake of
radionuclides.
8.4 Recommendations for
En vironmental Surveillance
The observations in this study support previously
presented recommendations that assessment of
population radiation exposures from routine facility
operation be based on measuring the radioactive
effluent and radiation flux at the station.
Environmental radionuclide and radiation levels
attributable to station operation were generally too
117
-------
variable, obscured by the radiation background, or near
instrumental detection limits to be measured precisely
for evaluating exposure. Measurements at the source
must include all significant pathways and radionuclides
during the entire period of operation; critical
radionuclides can be missed by monitoring only the
obvious effluents and the easily measured
radionuclides, or by ignoring the effects of changes in
the operating cycle. Detailed studies of in-plant
radionuclide pathways are needed for selecting an
optimum program for sampling and analysis.
Environmental measurements were found to be
useful for supporting and confirming the population
radiation exposures computed from on-site monitoring,
and for providing these computations with numerical
factors applicable to the site. Such measurements, if
performed reliably, can also be reassuring in
demonstrating that no unexpected radioactivity is in
the environment. For a station and site such as
Haddam Neck, the following measurements provide
useful information:
(1) confirmation of critical pathways
a) measure critical radionuclides in fish caught
in and near the coolant canal
b) measure external radiation exposure rates
on site and the decrease of the exposure
rate as function of distance to off-site
locations
(2) determination of numerical factors for
computing doses
a) compute X/Q values by measuring 133Xe
concentration in ground-level air relative
to the release rate at the station
b) observe long-term accumulation of
radionuclides in aquatic environment and
possible transfer to the food chain
(3) assurance that no significant exposure exists
from unforeseen sources or occasional
operational occurrences
a) measure radiation exposure at nearby
habitations
b) measure radionuclides in milk, food, and
drinking water obtained in immediate
vicinity of station.
This program by EPA and cooperating groups will
be concluded with a similar study at a commercially
operated BWR nuclear power station. Generic
radiological surveillance studies at other facilities in the
nuclear fuel cycle are under consideration.
118
-------
Appendix A
Acknowledgments
This report presents the work of the staff of the Radiochemistry and Nuclear Engineering Facility, USEPA,
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
Participation by the following is gratefully acknowledged:
Joseph Smolen, Connecticut Dept. of Environmental Protection
Leo Higginbotham, USAEC
Ronald Massengill, Essex Marine Laboratory
Floyd Galpin, Office of Radiation Programs, USEPA
Charles Phillips, Eastern Environmental Radiation Facility, USEPA
Ronald Shearin, Eastern Environmental Radiation Facility, USEPA
Sam Windham, Eastern Environmental Radiation Facility, USEPA
Gerald Karches, formerly Northeastern Radiological Health Laboratory, USPHS
Chris Nelson, formerly Northeastern Radiological Health Laboratory, USEPA
Joseph Cochran, formerly Northeastern Radiological Health Laboratory, USEPA
Carl Rosenberg, formerly Northeastern Radiological Health Laboratory, USEPA
David Lenth, Haddam Neck Nuclear Power Station
John Kangley, Haddam Neck Nuclear Power Station
Edward Goldin, Connecticut Board of Fisheries and Game
Assistance by W. A. Mills, E. D. Harward, and J. E. Martin, ORP, USEPA, in planning the study is gratefully
acknowledged. We wish to thank Prof. Daniel Merriman, Sears Foundation of Marine Research, Yale University,
and William Boyd, Essex Marine Laboratory, for their guidance of the aquatic aspects of this study, and Prof. Larry
Wilding, Ohio State University, for guidance and geochemical analysis of sediment samples. We thank Messrs.
James Gruhlke and Paul Magno, USEPA; Messrs. Bernard Weiss and Leo Higginbotham, USAEC; Richard
Graves, Haddam Neck Nuclear Power Station; Joseph Smolen, Connecticut Department of Environmental
Protection; and Profs. Conrad Straub, U. Minnesota; Hoyt Whipple, U. Michigan; James Leonard, U. Cincinnati;
and Daniel Merriman, Yale U., for reviewing the report.
119
-------
Appendix B.I
Haddam NeclT Average Monthly Power and Reactor Coolant Chemistry Statistics from Monthly Operating Reports
Average Power
Year
1967
Month
MWt
MWe
(gross)
Criticality achieved 1:04
No significant
1968
1969
1970
Oct.
Nov.
Dec.
Jan.
Feb.
Mar.
Apr.
May
June
July
Aug.
Sept.
Oct.
Nov.
Dec.
Jan.
Feb.
Mar.
Apr.
May
June
July
Aug.
Sept.
Oct.
Nov.
Dec.
Jan.
Feb.
Mar.
Apr.
May
June
July
Aug.
Sept
Oct.
Nov.
Dec.
240
627
1413
969
1448
81
512
1436
1188
1227
954
1193
1466
1249
1239
1420
1402
1577
648
771
1148
1274
1309
1476
1595
1436
1747
1799
1654
1337
740
0
79
1724
1543
1483
1724
1825
1739
operation
76
209
471
322
482
27
170
479
396
409
318
398
492
420
420
482
476
532
218
258
376
414
427
482
530
478
584
596
545
433
234
0
22
551
500
481
568
586
576
Boron,
ppm
am July
Tritium,
uCi/ml
24, 1967.
during August and
<1657
<1633
1630
~ 1650
- 1650
1850
1600
<1530
1490
1415
1375
1295
1211
1144
1080
993
918
806
710
728
678
610
533
425
312
234
155
51
0.1
<1
<1
2300
0.05
0.30
0.77
0.97
0.14
0.70
0.36
1.19
1.30
1.46
1.58
1.29
2.39
3.45
3.00
2.87
3.81
3.67
4.54
1.57
1.23
1.59
2.03
1.43
2.61
1.63
2.16
2.11
1.71
0.87
0.96
0.10
776** 0.57
574
516
447
371
287
207
5.43
5.38
1.77
2.94
3.68
3.39
Main Coolant*
"'I,
10 3 nCi/ml
September.
N.S.
0.39
0.52
2.47
0.26
0.29
0.47
0.15
0.18
0.09
3.68
0.39
0.51
0.13
0.27
0.23
0.10
0.13
0.18
10.96
0.29
0.26
0.26
0.40
1.31
1.54
9.10
3.78
2.98
1.48
13.4
N.S.
0.14
4.65
5.86
13.8
56.0
27.9
85.
pH
5.6
5.5
5.5
5.6
6.0
5.7
5.6
6.3
6.2
6.3
6.3
6.3
6.3
6.4
6.5
6.6
6.6
6.7
6.7
6.6
6.7
6.8
6.9
7.1
6.9
7.2
6.7
6.5
7.7
8.5
8.1
N.S.
5.4
6.2
6.9
6.3
6.3
6.8
6.9
Crud,
ppm
0.39
0.18
0.07
0.17
0.08
0.06
0.20
0.06
0.07
0.04
<0.01
-0.01
-0.01
0.06
0.34
<0.02
<0.10
<0.01
<0.01
0.08
0.03
0.13
<0.01
0.03
0.20
0.04
0.02
0.01
0.04
<1.00
<0.11
N.S.
N.S.
0.03
<0.02
0.03
<0.02
<0.02
<0.01
Gross Activity,
Iff1 uCi/ml
0.5
1.2
1.3
1.3
1.4
0.5
0.5
2.0
1.8
1.8
2.2
1.5
2.3
1.7
1.5
.6
.5
.7
.6
.0
.1
.0
.0
.1
.4
.5
1.5
1.5
1.6
1.0
1.1
1.5
0.1
2.0
2.0
2.2
2.6
2.6
3.5
Primary Plant
Leak Rate,
1/min Remarks
0.4
0.4
0.6
0.4
0.6
0.6
0.5
0.5
0.6
0.7
2.1
0.8
0.8
1.2
1.3
3.9
1.9
1.8
1.8
1.
1.
2.
3.
4.
5.
5.
5.
120
-------
Appendix B.I (cont'd)
Haddam Neck Average Monthly Power and Reactor Coolant Chemistry Statistics from Monthly Operating Reports
Average Power
MWe Boron, Tritium,
Year Month MWt (gross) ppm uCi/ml
1971 Jan. 1792 596 102 3.29
Feb. 1763 588 15 3.07
Mar. 1720 571 0.6 4.52
Apr. 739 275 0.9 3.41
May 231 78 1243** 0.14
June 1582 517 930 1.9
July 1713 549 875 3.67
Aug. 1728 555 798 5.17
Sept. 1760 574 709 5.63
Oct. 1692 556 620 4.14
Nov. 1813 601 511 7.00
Dec. 1733 575 446 6.23
* Average of reported values
** While reactor was operating
N.S. - No Sample
Remarks:
1. Shutdown for turbine modification 4/11/69 to
2. Bleed and feed operation toward end of month
Main Coolant*
'"I,
10J uCi/ml
34.
45.3
29.5
19.
25.4
17.4
15.8
16.5
15.6
21.3
17.5
17.6
5/11/69.
Crud, Gross Activity,
pH
7.1
8.5
10.0
9.9
6.3
6.7
6.6
6.6
6.8
6.8
6.8
6.8
to reduce fission
ppm 10' pCi/ml
0.01
<0.01
<0.01
<0.01
<0.01
0.01
0.11
<0.01
0.03
0.01
0.01
0.01
product gases
4.1
5.0
5.0
3.7
1.7
4.5
4.9
6.4
6.7
6.1
5.3
4.4
in coolant.
Primary
Plant
Leak Rate,
1/min
1.8
1.6
1.9
2.0
1.5
1.5
1.4
0.9
0.7
0.7
0.6
0.9
Remarks
3.
6.
7.
7.
3. Bleed and feed operation to reduce boron in main coolant.
4. Coastdown began 2/7/70.
5. Shutdown for refueling 4/17/70 to 6/24/70.
6. Coastdown began 3/5/71.
7. Shutdown for refueling 4/16/71 to 5/21/71.
121
-------
Appendix B.2
Haddam Neck Radioactive Waste Discharges from Monthly Operating Reports
Year
1967
1968
Month
Oct.
Nov.
Dec.
Jan.
Feb.
Mar.
Apr.
May
June
July
Aug.
Sept.
Oct.
Nov.
Dec.
Volume,
liters
2.1 x
3.5 x
3.6 x
1.4 x
3.1 x
3.2 x
105
10s
105
105
10=
10s
Liquid
Gross /3- 7,
mCi
20
71
125
131
47
356
1,619
162
31
34
142
1,102
233
13
99
Tritium, Volume,
Ci m3
2
123
95
215
60
337
119
85
137
27 1.4 x 103
170 8.5 x 101
159 2.1 x 10'
66
166
195
Gaseous
Gross /3-7,
Ci
<0.1
none
0.6
none
<0.1
<0.1
none
<0.1
2.6
<0.1
0.3
none
none
none
Tritium,
mQ
9
9,060
1968 Total
1969
3,969
1,735
3.7
9,069
Jan.
Feb.
Mar.
Apr.
May
June
July
Aug.
Sept.
Oct.
Nov.
Dec.
2.5 x 105
2.0 x 10s
1.9 x 10s
2.9 x 105
4.0 x 10!
2.6 x 105
3.7 x 105
1.6 x 10s
4.5 x 105
6.3 x 105
3.9 x 105
6.4 x 10'
28
16
4
51
10,740
1,320
105
33
5
3
12
525
269
160
261
624
669
138
181
156
360
943
560
730
2.1 x 102
6.1 x 10s
4.7 x 102
4.7 x 102
6.6 x 102
5.7 x 101
9.2 x 101
1.3 x 103
6.3 x 102
9.3 x 101
none
none
<0.1
<0.1
1.0
0.9
2.9
0.3
0.1
129.9
44.7
5.1
1
2,500
1
3
5
30
0
5
3
0
1969 Total 4.3 x 10' 12,842 5,051
1969 Total Reported*
4.3 x 106 12,170 5,163
185.0
185.2
2,548
2,520
122
-------
Appendix B.2 (cont'd)
Haddam Neck Radioactive Waste Discharges from Monthly Operating Reports
Volume,
Year Month liters
1970 Jan.
Feb.
Mar.
Apr.
May
June
July
Aug.
Sept.
Oct.
Nov.
Dec.
1970 Total
1970 Total
1971 Jan.
Feb.
Mar.
Apr.
May
June
July
Aug.
Sept.
Oct.
Nov.
Dec.
1971 Total
1971 Total
6.9 x~10!
2.3 x 10'
2.9 x 10'
1.0 x 10'
6.7 x 10s
1.2 x 10'
4.6 x 10*
4.9 x 10'
2.9 x 10'
2.5 x 10'
2.7 x 10'
2.5 x 10'
2.0 x 107
Reported**
1.6 x 10'
2.5 x 10'
2.6 x 10'
2.3 x 10'
2.9 x 10'
2.6 x 10'
2.2 x 10'
. 2.5 x 10'
2.3 x 10'
2.5 x 10'
2.4 x 10'
2.4 x 10'
2.9 x 107
Reportedf
2.9 x 107
Liquid
Gross fi-y
mCi
638
1,410
474
4,640
1,040
3,187
573
2,749
10,863
1,150
2,347
1,867
30,938
22,085
2,016
7,509
1,380
18,854
1,765
1,436
294
142
409
304
1,334
597
36,040
35,896
t Tritium,
Ci
1,078
"826
434
441
112
56
138
828
1,494
441
751
753
7,354
7,377
879
993
550
1,576
291
189
134
221
88
487
140
280
5,830
5,832
Volume,
m3
1.1 x
9.9 x
2.8 x
7.9 x
2.4 x
5.1 x
1.7 x
1.6 x
1.6 x
1.9 x
1.9 x
2.0 x
1.7 x
1.9 x
1.9 x
1.5 x
6.3 x
2.9 x
1.2 x
1.2 x
5.3 x
1.4 x
1.7 x
1.2 x
1.3 x
2.3 x
2.3 x
ior
103
10'
105
10s
10s
104
104
104
104
104
104
10'
104
104
104
10'
10'
104
104
104
107
104
104
104
107
107
Gaseous
Gross ft-y.
Ci
18.7
70.4
1.4
326.3
16.9
2.7
8.7
12.8
234.2
46.4
56.3
80.4
875.2
701.5
93.0
137.8
133.1
2,760.
19.
11.6
13.2
15.7
16.0
31.9
23.6
34.7
3,289.6
3,251.0
, Tritium,
mCi
0
n-r-ft
0
n.r.
n.r.
81
n.r.
n.r.
n.r.
n.r.
n.r.
n.r.
81
123
0
n.r.
n.r.
605
271
n.r.
n.r.
n.r.
n.r.
n.r.
n.r.
n.r.
876
876
* Connecticut Yankee Atomic Power Company, Operation Report No. 70-1, Jan. 1970.
"Connecticut Yankee Atomic Power Company, Operation Report No. 70-12, Dec. 1970.
t Connecticut Yankee Atomic Power Company, Operation Reports 71-6, June 1971, and
71-12, Dec. 1971.
tfn.r. - none reported
Note:
Total liquid gross 0-y discharged, excluding '"Xe and '"Xe was:
1970 6.67 Ci
1971 5.9 Ci
123
-------
Appendix B.3*
Haddam Neck Radionuclide Discharges
1970
Total volume (liters)
Reactor system volume -
borated (liters)
aerated (liters)
Not identified (mCi)
3H (Ci)
"Mn (mCi)
"Co (mCi)
"Co (mCi)
131I (mCi)
I33I (iriCi)
133Xe (mCi)
135Xe (mCi)
137Cs (mCi)
Total % Allowable (based on
isotopic analysis)
Total Available Dilution (liters)
Volume (m3)
3H (mCi)
41 Ar (Ci)
83Kr (Ci)
88Rb (mCi)
131I (mCi)
133I (mCi)
133Xe (Ci)
135Xe (Ci)
^ Xe (Ci)
Participates (mCi)
Total (Ci)
Total % Allowable (Based on
isotopic analysis)
Total Available Dilution (m3)
January-
June
Liquid
8.7 x 10'
n.r.**
n.r.
702
2,970
n.r.
3,444
6
72
n.r.
8,269
7
20
4.04
2.7 x 10"
Gaseous
n.r.
120ft
n.r.
7
n.r.
n.r.
n.r.
188
7
n.r.
1,410ft
219ft
0.25
n.r.
July
December
1.5 x 10'
14. x 10!
19. x 105
883
4,406
97
494
7
687
178
7,103
33
84
3.99
3.8 x 10"
n.r.
n.r.
3ft
3
n.r.
n.r.
n.r.
415
21
n.r.
n.r.
483ft
2.55
n.r.
1971
January-
June
1.5 x 107
15. x 105f
20.9 x 105t
21
4,478
3
764
79
1,750
716
28,800
110
575
2.44
3.5 x 10"
9.8 x 10'
876
n.r.
18
41
12
0.9
2,893
204
0.7
8
3,116
0.18
5.1 x 10"
July-
December
1.4 x 107
3.1 x 10!f
9.6 x 105f
32
1,352
404
37
734
373
326
1,080
48
43
0.62
3.9 x 10"
1.4 x 107
n.r.
n.r.
n.r.
159
0.03
0.05
109
16
10
10
135
5.3 x 10"
* Connecticut Yankee Atomic Power Company, Operation Reports 70-12, 71-6, and 71-12,
Dec. 1970, June 1971, and Dec. 1971.
**n.r. - none reported
f Connecticut Yankee Atomic Power Company, Operation Reports 71-1 to 71-12, Jan-Dec
1971.
ffExpressed as 133Xe equivalent
124
-------
Appendix B.4*
Sources of Waste at Haddam Neck—July-December, 1970
Radionuclide or Volume
Liquid waste from reactor system
Gross /3-7, mCi (unidentified)
3H, Ci
"Mn, mCi
"Co, mCi
"Co, mCi
I3'I, mCi
133I, mCi
133Xe, mCi
'3SXe, mCi
137Cs, mCi
July
test tanks
435
111
n.r.**
8
n.r.
n.r.
n.r.
86
7
n.r.
Aug.
5
826
n.r.
80
n.r.
n.r.
n.r.
2,653
2
n.r.
Sept.
208
1,485
n.r.
60
n.r.
479
37
10
4
n.r.
Oct.
109
429
56
n.r.
n.r.
31
n.r.
925
2
n.r.
Nov.
n.r.
730
28
319
7
6
n.r.
1,760
17
74
Dec.
n.r.
720
13
27
n.r.
n.r.
n.r.
1,670
2
10
Volume, liter
4.6xl05
4.9xl05
9.0xl05 3.8x10'
7.3xlOs
3.5x10'
Liquid waste from secondary system: leakage and blowdown
Gross /J-7,mCi (unidentified)
3H, Ci
"'I, mCi
133I, mCi
34
27
n.r.
n.r.
9
3
n.r.
n.r.
55
9
n.r.
n.r.
21
12
21
12
2
23
66
69
1
32
84
61
Volume, liter
2.0x10'
2.1x10'
2.0x10'
2.1x10'
Gaseous waste from air ejector
133Xe, Ci
135Xe, Ci
4IAr, Ci
Volume, m3
Gases from waste gas sphere
5.2 12.1 44.6 42.5 49.8 71.6
3.1 0.4 2.6 4.0 6.4 4.1
0.4 n.r. n.r. n.r. n.r. n.r.
1.7xl04 1.6x10' 1.6xl04 1.9xl04 1.9x10" 2.0xl04
133Xe, Ci
1MXe, Ci
85Kr, Ci
Volume, m3
3.1x10-'
n.r.
n.r.
none 8.5x10'
1.8x10'
2.4x10-'
1.8x10"
1.3xl02 none
5.2xlO-2
1.0x10"
n.r.
1.4x10'
3.7x10°
n.r.
9.8x10"
9.1x10°
*Connecticut Yankee Atomic Power Company, Operation Reports 70-7 through 70-12, July-December 1970.
**n.r. - not reported
125
-------
Appendix C.I
Calculated Generation Rate of Fission Products in Fuel at 1825 MWt Power(a)
Fission
product
3H
83"Kr
85mKr
85Kr
87Kr
88Kr
8'Kr
8'Sr
'°Sr
"Zr
"Nb
"Mo
I03Ru
124Sb
,,,j
'"I
"'I
131mXe
"3™Xe
133Xe
"""Xe
135Xe
137Xe
138Xe
134Cs
136Cs
"'Cs
140Ba
u'Ce
144Ce
Yield, y(b)
9.5 x 10-!(c)
5.8xlOJ
1.3xlO'2
2.9xlO'3
2.4xlQ-2
S.SxlO'2
4.2xlO'2
4.5xlQ-2
5.5xlO'2
6.4xlO'2
6.4xlO'2
5.7xlO'2
3.3xlO'2
3.0x10-'
3.2X10'2
6.5xl02
6.0xl02
4.5x10"
1.9x10°
6.5xlO'2
l.lxlO'2
6.3xlO"2
6.0xlO"2
5.8xlO"2
1.2x10"'
1.6x10"
6.2xl02
6.0xlO"2
6.0xl02
5.2xlO"2
Decay constant
X, s'1
1.78x10''
1.04xlO'4
4.30x10-'
2.05x10"'
LSlxlO'4
6.88x10-'
3.66xlO'3
1.57x10''
7.82xlO-10
1.23x10-'
2.29x10"'
2.90x10""
2.03x10"'
1.33x10"'
9.96x10"'
9.21x10"'
2.87x10"'
6.74x10"'
3.57x10"'
1.52x10"'
7.38x10"
2.10x10"'
3.02xlO"3
8.15xlO"4
1.06xlO"8
6.17x10''
7.30x10"'°
6.26x10'
2.48x10"'
2.82xlO"8
Generation rate
nCi/s
2.1xl02
7.3x10'
6.8x10'
7.2xl03
4.5x10'
3.0x10'
1.9x10"
8.6x10"
5.2xl04
9.6x10'
2.1xlO'(d)
2.0x10"
8.2x10'
4.9x10'
3.9x10'
7.3xl08
2.1x10'
3.7x10'
8.1x10'
1.2x10"
9.8x10'
1.6x10'
2.2x10"
5.8x10'°
~4. xlO'(e)
1.2x10'
5.5xl04
4.6x10'
1.8x10'
1.8x10'
Accumulation in
500 days, uCi
8.6x10'
7.1xl012
1.6x10"
2.9x10"
3.0x10"
4.3x10"
5.2x10"
5.5x10"
2.2xl012
7.8x10"
1.7xl014
7.0x10"
4.0x10"
3.7x10"
3.9x10"
8.0x10"
7.3x10"
5.6x10"
2.3xl012
7.9x10"
1.3x10"
7.7x10"
7.3x10"
7.1x10"
~ 1.5xlO'2(e)
2.0x10"
2.3xl012
7.3x10"
7.3x10"
4.5x10"
Based on 100% uranium fission; actually there is an increasing
fraction with time related to the fission of generated plutonium.
b Meek, M. E. and B. F. Rider, "Compilation of Fission Product Yields",
General Electric, Vallecitos Nuclear Center Kept. NEDO-12154-1 (1974)
c Albenesius, E. L. and R. S. Ondrejcin, "Nuclear-Fission
Produces Tritium", Nucleonics 18 (9), 199 (1960).
d Equilibrium with longer-lived parent is assumed.
e Mountain, J. E., L. E. Eckart and J. H. Leonard, "Survey
of Individual Radionuclide Production in Water-Cooled
Reactors", University of Cincinnati Rept. (1968).
Notes:
1. Generation rate = thermal power x Fission rate x use factor x y x X =
MWt
1825 MWt x 3.1 x 10"
fission/s x 0.8 x y x X x
uCi
2.
3.
MWt 3.7 x 104 dis/s
Use factor of 0.8 is average for 1970 and 1971
Accumulation = thermal power x fission rate x use factor x y (l-e'^1)
MWt
126
-------
Appendix C.2
Calculated Fission Product Concentrations in Reactor Coolant Water
Fission
Product
83mKr
85"Kr
8SKr
87Kr
88Kr
89Kr
89Sr
'°Sr
"Zr
"Nb
"Mo
'°3Ru
[2"Sb
"'I
,MI
'"I
'31mXe
133mXe
'"Xe
1JSmXe
'"Xe
l]'Xe
T38Xe
134Cs
'3'Cs
137Cs
u°Ba
u'Ce
'"'Ce
Notes:
1. Amounts
Amount in 0.02%
of fuel elements,
uCi
1.4x10'
3.2x10'
5.8xl07
6.0x10'
8.6x10'
1.1x10'°
1.1x10'°
4.4x10"
1.6x10'"
3.4x10'°
1.4x10'°
8.0x10'
7.3x10*
7.8x10'
1.6x10'°
1.5x10'"
l.lxlO8
4.6xl08
1.6x10'°
2.6x10'
1.5x10'°
1.5x10'°
1.4x10'°
~ 3.0x10"
4.0x1 07
4.7x10"
1.5x10'°
1.5x10'°
9.0x10'
are 2xlO'4 x values
Leakage
rate,
uCi/s
9.2x10'
2.1x10'
3.8
3.8x10'
5.6x10'
6.7xl02
l.lxlO'1
4.4xlO'3
2.6x10''
5.4xlO'2
2.8x10'
1.3x10'
1.2x10-'
l.OxlO2
2.1xl02
2.0xl02
7.2
3.0x10'
l.OxlO3
1.7xl02
l.OxlO3
9.5xl02
9.2x10'
-3.9
5.2x10''
6.1
1.5x10''
2.4xlO'2
1.4xlO'2
in column 5 of t
Summed turn-
over constants,
s1
1.0x10'"
4.3xlO'!
1.6x10-'
1.5x10'"
6.9xlOs
3.7xlO'3
2.8xlO's
2.8xlO'5
2.8xlOs
2.8xlQ-5
3.1xlO'5
2.9xlO's
2.8xlO'5
2.9xlO'5
3.8xlO-s
5.7xlO'5
8.3x10''
3.7xlQ-6
1.7xlO-6
7.4x10'"
2.1xlO's
3.0xlO'3
8.2x10'"
1.6xlO's
1.6xlO'5
1.6xlO's
2.9xlQ-5
2.9xlO'5
2.8xlO'5
Appendix C.I.
Concentration,
uCi/ml
5.5xlO'3
S.OxlO'2
1.5x10''
1.6xlO'2
5. 1x10 2
l.lxlO'2
2.5xlO'5
9.8x10''
5.8xlO'5
1.2xlO'5
5.6xlO'3
2.8x10''
2.7x10'"
2.2x10-'
3.5x10''
2.2xlO'2
5.4xlO'2
S.OxlO'2
3.8
l.SxlO'3
2.9x10''
2.0xlO'3
7.1x10°
l.SxlO'3
2.0x10'"
2.4xlO'3
3.2xlO'5
5.3xlO'6
3.1x10''
2. Leakage rates are amounts x the following escape rate coefficients:
6.5 x 10" s'for Xe, Kr
1.3 x 10" s'for I, Cs
2.0 x 10-' s'for Mo
1.0 x 10"" s'for Sr, Ba
1.6 x 10'12 s'for all others
3. Summed turnover COnStantS = Xwalcr loss + Xdemineraliier + Xdecay
Where Xwater loss = 1.6 X 10'' S''
Xdemineralizer = 1.56 X 10"* S'' for CS
0 s'1 for Xe, Kr
2.82 x 10'5 s'1 for all others
Xdecay are values in column 3 of Appendix C.I
127
-------
Appendix C.2 (cont'd)
4. Concentration = leakage rate/(1.6 x 10" x summed turnover constants)
where 1.6 x 10" is the reactor coolant volume in ml.
5. The values of XdemineraUzer are for a flow rate of 300 kg/min
through a mixed bed demineralizer that has decontamination factors
of 2 for cesium and 10 for all other ions. These decontamination
factors and the escape rate coefficients are from reference 3 in
Section 1.4.
6. Values are for 1825 MWt reactor after operation with 80 percent
capacity factor for 500 days, with 0.020 percent of fuel leaking
radioactivity.
128
-------
Appendix C.3
Calculated Radionuclide Concentration in Steam Generator Water
Concentration, Cs, uCi/ml
Radionuclide July
24, 1970
Nov.
20, 1970
March 16, 1971
from fuel
"Sr
"Sr
"Zr
"Nb
"Mo
1MRu
"'I
133 j
IJJj
1J4Cs
lMCs
"7Cs
140Ba
I41Ce
MCe
™Np
gross
7
2
9
1
<7
4
.5
4
6
6
<6
<8
alpha < 2
X
X
X
X
—
X
X
--
—
X
X
X
X
X
X
-
X
from activation
14C
24Na
32P
3!S
51Cr
"Mn
"Fe
"Fe
"Co
58Co
MCo
63Ni
124Sb
182Ta
1
2
<7
2
I
3
3
1
3
4
4
1
<7
X
—
X
X
X
X
X
X
X
X
X
X
X
X
io-8
10-'
io-9
lO'8
-
Iff10
10"
-
-
10"
io-8
10"
lO'8
10"°
10'10
-
10-"
of water,
ID'8
-
10'"
10''°
IO"
10J
10'!
lO'7
10"
ID'5
IO"
lO'7
io-8
ID'10
5
2
9
2
8
1
2
5
1
3
7
3
3
1
<2
1
2
cladding,
4
1
9
1
1
2
2
4
6
2
2
2
2
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
10'8
ID'8
10-'
10-'
JO'8
io-8
lO"4
10"'
10'!
10"
10"
10"
io-8
io-8
ID'8
10"
10'10
5
3
2
1
3
1
3
2
5
9
2
9
1
2
3
3
6
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
10"
io-8
10"
10"
10"
10"
10"4
io-4
ID'5
io-6
10"
10"
10"
10"
10"
10"
10-'
and construction materials
X
X
X
X
X
X
X
X
X
X
X
—
X
X
10"
10"
io-8
io-8
10"
10"
10"
io-9
10-'
10"
10"
10-'
io-8
2
2
2
<2
<2
<5
8
7
<1
<3
__
X
--
-
X
X
X
X
X
X
X
--
X
X
10"
10"
io-8
10'
10'
10-'
10"
io-8
10-'
io-8
Notes:
1. Calculated by Cs = ( C,w, )/( w* + v»Xd )
where
Cs = radionuclide concentration in steam generator water, uCi/ml
Cr = radionuclide concentration in reactor coolant water (Table 2.1), uCi/ml
Wr = rate of water leakage from reactor into 'secondary coolant
water, kg/day
129
-------
Appendix C.3 cont'd)
Ws = rate of water discharge (blowdown) from steam generators, kg/day
Vs = amount of water in steam generators (1.2 x 10s kg), kg
Xd = radioactive decay constant (column 3, Appendix C.I, multiplied
by 86,400 s/day), day'
2. The following values in kg/day were used in the calculation:
July 24, 1970 Nov. 20, 1970 March 16, 1971
Wr 110 180 310
Ws 17,000 15,000 26,000
130
-------
Appendix D.I
Estimation of Airborne Radioactivity in the Environment
The atmospheric diffusion model for the Haddam Neck site is derived from the following Gaussian dispersion
equation (see Connecticut Yankee A tomic Po wer Company, Facility Description and Safety Analysis, Vol. I):
XU 1
TT oy az
where
X = ground-level centerline concentration, Ci/m3
u = average wind velocity, m/s
Q = source release rate, Ci/s
0y = crosswind plume standard deviation, m
0r = vertical plume standard deviation, m
The equation estimates the plume centerline concentration at ground level from a source released at ground level.
Meteorological observations were conducted at the site before building construction began for approximately a
15-month period from the winter of 1962. Wind speed and direction frequency data were provided by instruments
mounted 1.6 m and 33 m above ground level. Temperatures were measured 7.6 m and 37 m above ground level to
obtain the temperature difference for classifying atmospheric stability. By releasing chemical tracers and measuring
their concentrations in air, oy was determined as a function of downwind distance. Values of ov were obtained from
data published by Sutton. The values ofay and
-------
Appendix D.I (cont'd)
Characteristic
Nearest residence
Nearby residence
Nearby residence
Nearby population
of 70
Nearby population
of 89
Nearby population
of 46
Nearest security
fence
Nearest exclusion
fence
Nearest dairy
farm
Fishing in canal
Fishing in canal
Fishing in river
State park
Location
0.7 km WNW
0.8 km NW
0.8 km S
1.2 km S
1.2 km E
1.2 km SSE
0.1 km NW
0.5 km NW
3.4 km ESE
0.5 km SE
1.7 km SE
0.2 km SW
0.7 km SW
Relative
concentration
normalized to
wind velocity
(Xu/Q), m2
1.2 x 10"
1.1 x 10"
1.1 x 10"
7.1 x 10s
7.1 x 10s
7.1 x 10'5
6.8 x 10"
1.7 x 10"
2.4 x 105
1.7 x Iff'
4.9 x 10s
4.0 x 10"
1.2 x 10"
Wind
frequency
%
8.2
6.6
2.6
2.6
6.9
4.5
6.6
6.6
13.2
15.6
15.6
2.2
2.2
Avg. wind
velocity,
if, m/s
3.5
3.2
3.9
3.9
3.1
3.5
3.2
3.2
3.5
3.6
3.6
4.0
4.0
Avg. annual
relative
concentration
(X/Q), s/mj
2.8 x 10'
2.3 x 10"
7.3 x 10'
4.7 x 10 -'
1.6 x 10'
9.1 x 10'
1.4 x 10s
3.5 x 10'
9.1 x 10'
7.4 x 10'
2.1 x 10'
23. x 10'
6.6 x 10'
Values of X/Q calculated by the site diffusion model are about one-fourth as large as the annual average
relative concentrations for various sectors and distances given in the Haddam Neck Plant Environmental Report,
Operating License Stage, Section 2.7 (1972). On the other hand, estimates of X/Q in the Environmental Statement
indicate less available dispersion, ranging from a factor of 11 at the fence to a factor of 2 at the dairy, than those
given by the site model. The values for listed locations are as follows:
Location X~/Q, s/m3
0.1 km NW (nearest security fence) 1.6 x 10"
0.7 km WNW (nearest residence) 1.8 x Iff"
3.4 km ESE (nearest dairy) 2 x 10"'
Centerline concentration values at a receptor location overestimate annual average concentrations in a sector
because the effluent is actually distributed across the sector. The average long-term concentration, XL (see equation
3.144 in Meteorology and Atomic Energy 1968, D. Slade, ed., USAEC Kept. TID-24190), for a 16-sector wind rose
is related to the centerline value by XL/X = 6.38(Ty/x, where x is distance from the stack. Use ofoy values developed
for the Haddam Neck site leads to ratios of XL/X below unity at distances beyond 2.5 km from the source: at 3.4
km, the ratio is 0.9. Generalized ov values (ibid. pg. 102) weighted according to atmospheric stability frequencies
given above yield a ratio of approximately 0.5 for distances of 0.1 to 3.4 km.
132
-------
10
-3
CM
I
I ICT4
o
x.
it
CL
W
?
i
p
E
o
c
.o
o
I
o
-------
Appendix D.3
Atmospheric Dispersion and Plume Rise Estimates for Short-term Sampling
Plume dispersion during the tests described in Section 6 was estimated by the equation given in Appendix D. 1,
modified to take into account the elevation of the discharge from the stack:
Xu 1 f 1 H ,2
- - (—)
Q - a a ~* | 2
where H = effective stack height (53 + Ah), m
Plume rise (Ah) was estimated by the ASME technique:
where:
Ah = height of plume centerline above stack height, m
Vs = stack effluent velocity = 13.0m/s
u = average wind speed at stack height, m/s
d = stack diameter = 1.8m
Parameters used to estimate dispersion during Test 3 (see Table 6.2) for slight atmospheric instability and 0.6
km distance to stack were:
n Ah, ay,
-------
Appendix D.4
Estimated Deposition of "'I at Farms Where Thyroids were Collected
Date
3.3 km
Aug.
Sept.
3.3 km
May
June
Wind frequency in Avg. wind
270°-360° speed,* *(u)
sector,*(T)hr/day m/s
ESE
25, 1970
26
27
28
29
30
31
1
2
3
4
5
6
7
8
9
10
11
12
13
14
SE
26, 1971
27
28
29
30
31
1
2
3
9
3
15
12
24
6
18
21
18
9
0
12
15
12
3
0
0
21
12
0
9tt
3
12
15
3
3
9
12
—
—
3.09
1.55
2.88
1.80
4.32
1.80
6.59
4.86
3.61
1.55
—
4.51
4.43
3.73
1.55
—
—
4.34
2.19
...
4.29
2.58
3.35
2.58
1.55
5.15
3.61
2.96
...
—
Deposition "'I on
ground f, (D), pCi/m2
3.29 x KTQ'o
2.19
5.89
7.53
6.28
3.77
3.09
4.88
5.63
6.56
—
3.01
3.83
3.64
2.19
—
—
—
—
—
2.37
1.31
4.05
6.57
2.19
0.66
2.82
4.58
...
—
* U. S. Dept. of Commerce, ESS A Local Climatological Data Sheets, Hartford, Conn.,
Aug. 1970, Sept. 1970, May 1971, June 1971.
** As reported from Hartford, for 90° sector. Divided in calculation by 4 to
allow for 22.5° sector.
f For (Xu/Q) = 2.5 x 10'' m2, vd = 10'2 m/sec,(Xi/X) = 0.5;
T = hrs. favorable wind direction in 270°-360° sector multiplied
by fraction (22.5790°) for smaller sector and by factor 3.6 x 103 s/hr.
Thus, D = 1.13 x 10"4 Q'o (T/n)
ft Deposition considered only for first half of day, since cow sent to market during this day.
135
-------
Appendix D.5
Calculations of Estimated '"I Levels in Cow Thyroids and Milk
Date
Aug.
Sept.
Total
3.3 km
May
June
25
26
27
28
29
30
31
1
2
3
4
5
6
7
8
9
10
11
12
13
14
SE
26
27
28
29
30
31
I
2
3
Deposition
From
Previous
Day
2.86
4.39
8.64
14.33
17.93
18.88
19.12
20.87
23.06
25.77
22.51
22.20
22.65
2287
21 80
19.04
16.54
14.28
12.53
10.78
1.14
4.52
9 64
10.29
9.53
10.74
13.33
11.64
, (10-' Qo')'
Fresh
Deposition
3.29
2.19
5.89
7.53
6.28
3.77
3.09
4.88
5.63
6.56
...
3.01
3.83
3.64
2.19
...
...
237
1.31
4.07
6.57
220
068
2.80
4.58
...
,pCi/m!
Total
3.29
5.05
10.28
16.47
20.61
21.70
21.97
23.99
26.50
29.62
25.77
25.52
26.03
26.29
25.06
21.80
19.04
16.54
14.28
12.53
13.15
1.30
5.19
11.08
11.83
10.95
12.35
15.32
13.33
11.64
Daily
Intake
(Ifl-'Qo')1
pCi/d
148
227
463
741
927
977
989
1080
1190
1330
1160
1150
1170
1183
1130
981
857
744
643
564
592
59
234
499
532
493
556
689
600
524
Decay
Factor
to
Sept. 14
0.14
0.15
0.17
0.19
0.21
0.23
0.25
0.28
0.30
0.34
0.37
0.41
0.45
0.50
0.55
0.61
0.67
0.74
0.82
0.91
0.95'
0.45
0.50
0.55
0.61
0.67
0.74
0.82
0.91
1.00
In thyroid,
20% daily
uptake
30
45
93
148
185
195
198
216
238
266
232
230
234
237
226
196
171
149
129
113
118
12
47
100
106
99
111
138
120
105
UO'Qo')1, pCi
Cumulative
amount to
Sept. 14
4
7
16
28
39
45
50
60
71
90
86
94
105
119
124
120
115
110
106
103
112
1604
= O.I6Q,/
5
24
55
65
66
82
113
109
105
Hypothetical cone, in
% intake
per liter
0.01
0.01
0.02
0.04
0.08
0.22
0.35
milk
Concentration
on Sept. 14
(10'Q,') ', pCi/1
0.0 1
0.01
0.02
0.04
0.08
0.23
0.39
= 0.78 x lO^Qo'
0.01
0.01
0.02
0.04
0.08
0.22
0.35
0.01
0.01
0.01
0.03
0.09
0.24
0.37
Total 624 =0.76 x 10"Qo'
= 0.062Q,,'
Note:
See Seclion 7.5 for factors (environmental half-life ~ 5 days, effective half-life in cow — 7 days,
effective daily grii/ing area = 45mJ, ralio of '"I concentration in milk to daily intake by cow from
Curve "a" for sequential single intakes of "'I in reference 13).
' For half day.
136
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Appendix E.I
Relation of Airborne Radionuclide Concentration to Dose Rate
Maximum permissible concentration
Radio-
nuclide
3H (HTO)
(HT)
"C (CO2)
"Ar
"•Kr
MKr
"Kr
MKr
U1-Xe
1UmXe
133Xe
135Xe
Critical organ
Total body (In)'2'
Skin (Sub)'3'
Fat (In)
Total body (In)
Total body (Sub)
Total body (Sub)
Total body (Sub)
Total body (Sub)
Total body (Sub)
Total body (Sub)
Total body (Sub)
Total body (Sub)
Total body (Sub)
per maximum
(pCi/cc) /
Gases
2/5
400/30
1/5
2/5
0.4/5
1/5
3/5
0.2/5
0.28/5'"
4/5
2.8/5'41
3/5
1/5
permissible dose,"'
(rem/yr)
0.4
= 13
= 0.2
0.4
= 0.08
= 0.2
= 0.6
= 0.04
0.06
0.8
0.6
= 0.6
= 0.2
other fission gases
with half-lives
hrs
"Mn
55Fe
"Co
MCo
"Sr
MSr
131I
133I
USj
1MCs
137Cs
< 2
Total body (Sub)
Airborne particles
Lung (I)(5)
Spleen (S)"1
Lung (I)
Lung (I)
Lung (I)
Bone (S)
Lung (I)
Bone (S)
Total body (S)
Thyroid (S)
Thyroid (S)
Thyroid (S)
Lung (I)
Lung (I)
0.27/5'"
and iodine by inhalation
0.01/15
0.3/15
0.3/15
0.02/15
0.003/15
0.01/30
0.01/15
0.0001/30
0.0003/5
0.003/30
0.01/30
0.04/30
0.004/15
0.005/15
0.05
0.00067
0.020
0.020
0.0013
= 0.00020
= 0.00033
= 0.00067
0.0000033
0.000060
= 0.0001
= 0.00033
0.0013
0.00027
= 0.00033
137
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Appendix E.I (cont'd)
1. ICRP, Report of Committee 2 on Permissible Dose for Internal Radiation,
ICRP Publication 2, Pergamon Press, Oxford (1959); concentrations based on
168-hour limits.
2. (In) - Inhalation
3. (Sub) - Submersion
4. Based on ICRP Publication 2, equation 21, divided by 4 for a 168 hour week:
(MPC)a = 2.6 x 1/4 = pCi/cc,
S(E)
where 2(E), the total effective energy per disintegration (j3, y, e', x-rays),
has the values:
88Kr = 2.33 MeV
'33mXe = 0.234 Me"
Short-lived nuclides
(Ti/2 < 2 hrs) = 2.42 MeV (based on "Kr, the
radionuclide of highest disintegration energy with a
half-life less than 2 hours)
5. (I) - Insoluble
6. (S) - Soluble
138
-------
Appendix E.2
Relation of Daily Radionuclide Intake in Water to Dose Rate
Radionuclide
3H
UC
24Na
"P
"Cr
"Mn
5!Fe
!'Fe
57Co
58Co
MCo
MSr
"Zr
9!Nb
"Mo
11 Om A
Ag
13lj
I33I
,35j
134Cs
136Cs
'37Cs
Critical organ
Total body
Total body
GI(LLI)
Bone
Total body
GI(LLI)
GI(LLI)
GI(LLI)
spleen
GI(LLI)
Spleen
GI(LLI)
GI(LLI)
GI(LLI)
Bone
GI(LLI)
GI(LLI)
GI(LLI)
GI(LLI)
Thyroid
Th., roid
Thyroid
Total body
Total body
Total body
Maximum permissible intake
per maximum permissible dose,"1
(pCi/day) / (rem/yr)
66,000,000/5 =
17,600,000/5 =
660,000/15
440,000/30
1,980,000/5 =
1,980,000/15 =
44,000,000/15 =
2,200,000/15
17,600,000/15 =
1,320,000/15 =
2,200,000/15 =
11,000,000/15
2,200,000/15 =
1,100,000/15
200/0.167™ =
1,320,000/15 =
2,200,000/15 =
4,400,000/15 =
660,000/15 =
80/0. 500'2' =
280/0.500'3' =
800/0.500"" =
198,000/5 =
1,980,000/5
440,000/5 =
13,200,000
3,520,000
44,000
14,700
396,000
132,000
2,930,000
147,000
1,170,000
88,000
147,000
733,000
147,000
73,300
1,200
88,000
147,000
293,000
44,000
160
560
1,600
39,600
396,000
88,000
ICRP Report of Committee 2 on Permissible Dose for Internal Radiation,
ICRP Publication 2, Pergamon Press, Oxford (1959): Intake based on 168-
hour concentration limits.
2. based on recommendations by Federal Radiation Council in "Background
Material for the Development of Radiation Protection Standards",
Report No. 2, U. S. Government Printing Office, Washington, D. C. (1961).
3. based on ICRP, 'UI/'31I ratio: 80 x 7 x lQ-'/2 x 10"' = 280.
4. based on ICRP, I35I/'3'I ratio: 80 x 2 x 10V2 x 10'' = 800.
139
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Abstract
RADIOLOGICAL SURVEILLANCE STUDY AT THE HADDAM NECK PWR NUCLEAR POWER STATION
B. Kahn. R. L. Blanchard, W. L. Brinck. H. L. Krieger, H. E. Kolde. W. J. Avcrclt, S. Gold, A. Martin, G. Gels; Jan.
1975; EPA-520/3-74-007:ENVIRONMENTAL PROTECTION AGENCY.
A radiological surveillance study, one of a series at commercially operated nuclear power stations, was undertaken
at the Haddam Neck (Connecticut Yankee) PWR plant. Radionuclide concentrations and external raoiation were
measured in the immediate vicinity of the 590-MWe station. The radionuclide contents and pathways of gases and
liquids at the station, including points of discharge, were also measured to estimate radionuclide levels in the
environment.
The radionuelides in airborne effluents were mostly '"Xc. JH. and "
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