TECHNICAL NOTE
ORP/TAD-77-3
CHARACTERIZATION OF
SELECTED LOWLEVEL
RADIOACTIVE WASTE
GENERATED BY FOUR
COMMERCIAL LIGHTWATER
REACTORS
December 1977
U.S. ENVIRONMENTAL PROTECTION AGENCY
Office of Radiation Programs
Washington, D.C. 20460
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Technical Note
ORP/TAD-77-3
CHARACTERIZATION OF SELECTED LOW-LEVEL RADIOACTIVE
WASTE GENERATED BY FOUR COMMERCIAL
LIGHT-WATER REACTORS
BY
DAMES AND MOORE
White Plains, New York
through
The New York State Energy Research
and Development Authority
DECEMBER 1977
This report was prepared as an account of work sponsored by the
Environmental Protection Agency of the United States Government
under Contract No. 68-01-3294
PROJECT OFFICER
WILLIAM F. HOLCOMB
Radiation Source Analysis Branch
Technology Assessment Division
OFFICE OF RADIATION PROGRAMS
U.S. ENVIRONMENTAL PROTECTION AGENCY
WASHINGTON, D.C. 20460
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EPA REVIEW NOTICE
This report has been reviewed by the Office of Radiation Programs,
U.S. Environmental Protection Agency (EPA) and approved for publication.
Approval does not signify that the contents necessarily reflect the views
and policies of the EPA. Neither the United-States nor the EPA makes any
warranty, expressed or implied, or assumes any legal liability or responsibility
of any information, apparatus, product or process disclosed, or represents that
its use would not infringe privately owned rights.
II
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PREFACE
The Office of Radiation Programs of the U.S. 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 the public health and safety. This report
was prepared in order to determine the radioactivity source terms associated
with the low-level wastes generated by light-water reactors and subsequently
shipped to commercial shallow-land burial facilities. Readers of this report
are encouraged to inform the Office of Radiation Programs of any omissions
or errors. Comments or requests for further information are also invited.
David S. Smith
Director
Technology Assessment Division (AW-459)
Office of Radiation Programs
III
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ABSTRACT
An investigation was made of the radionuclide makeup of light-water
nuclear reactors' radioactive wastes presently being consigned to shallow
land burial. The studies were contracted through the New York State
Energy Research and Development Authority and consisted of radiochemical
analyses of spent ion exchange resins, evaporator concentrates and filter
sludges for specific radionuclides including activation products, fission
products and transuranics . Ten waste samples were obtained from two BWRs
and two PWRs.
IV
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TABLE OF CONTENTS
EPA Review Notice II
Preface Ill
Abstract IV
Table of Contents V
List of Tab!es VII
List of Figures \/III
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TABLE OF CONTENTS
SECTION PAGE
1.0 Introduction 1-1
2.0 Summary 2-1
3.0 PWR and BWR Radioactive Waste Treatment Systems 3-1
and Components
3.1 General Systems Comparison 3-1
3.2 Ion Exchange Resin Characteristics. ... 3-6
3.3 Evaporator Characteristics 3-8
4.0 Waste Treatment Systems at Sampled Reactors 4-1
5.0 Radionuclide Analyses of Waste Samples Collected
Under This Program. - 5-1
5.1 Sample Definition and Collection Procedures 5-1
5.2 Spent Ion Exchange Resins ...... 5-4
5.3 Evaporator Concentrates 5-7
5.4 Filter Sludges. 5-10
6.0 Radionuclide Analyses of LWR Wastes Performed
Under Other Programs 6-1
6.1 Spent Ion Exchange Resins 6-1
6.2 Evaporator Concentrates 6-2
6.3 Filter Sludges 6-6
7.0 Comparisons, Interpretations, and Recommendations . . . 7-1
7.1 Variables Influencing Composition of Waste Sample . . 7-1
7.2 Comparison of Radionuclide Analyses 7-2
7.3 Interpretations of Data 7-7
7.4 Recommendations 7-9
8.0 Bibliography 8-1
Appendix A - Waste Treatment Systems at Reactors From
Which Samples Were Collected A-l
Appendix B - Analytical Methods Used by the Radiological
Science Laboratory (RSL) B-l
VI
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LIST OF TABLES
TABLE PAGE
3-1 Liquid Radwaste Classification 3-2
5-1 Description of Collected Samples 5-2
5-2 Radionuclide Analysis of Spent Ion Exchange
Resin Samples Measured Under This Program 5-5
5-3 Radionuclide Analysis of Evaporator Concentrate
Samples Measured Under This Program 5-8
5-4 Radionuclide Analysis of Filter Sludge Samples
Measured Under This Program 5-11
6-1 Radionuclide Analysis of PWR Spent Resin Samples
Measured Under Other Analytic Programs 6-3
6-2 Radionuclide Analysis of PWR Evaporator
Concentrate Samples Measured Under Other
Analytic Programs 6-4
6-3 Radionuclide Analysis of Evaporator Concentrate
Samples From Indian Point No. 2 Reactor - July,
September, October, December, 1975 6-5
6-4 Radionuclide Analysis of PWR Filter Sludge Samples
Measured Under Other Analytic Programs 6-7
7-1 Comparison of Concentrations of Gamma Emitting
Radionuclides (T%>3QQ days) In Samples of
Evaporator Concentrate From PWRs 7-4
7-2 Comparison of Concentrations of Gamma Emitting
Radionuclides (T^>300 days) In Samples of
Evaporator Concentrate From PWRs & BWRs. ...... 7-5
7-3 Comparison of Concentrations of Gamma Emitting
Radionuclides (TJj>300 days) In Samples of
Spent Ion Exchange Resins From PWRs & BWRs 7-6
VII
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LIST OF FIGURES
FIGURE PAGE
1 Typical System for Treatment of Liquid and
Solid Radioactive Wastes at a Boiling Water
Reactor. ..... 3-4
2 Typical System for Treatment of Liquid and
Solid Radioactive Wastes at a Pressurized
Water Reactor 3-5
3 Schematic Diagram of Mixed-Bed and Separate-Bed
Ion Exchange Systems 3-7
4 Typical Evaporators Used in Processing Liquid
Radwaste 3-9
A-l Upgraded Liquid Radwaste System, Nine Mile
Point Nuclear Station, Unit 1. A-2
A-2 Flow Chart of the Liquid Waste System A-6
A-3 Liquid Waste System at R.E. Ginna A-10
VIII
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1.0 Introduction
The purpose of this study was to provide data on the radionuclide
composition and concentration in spent ion exchange resins, evaporator
concentrates, and filter sludges which result from waste management
operations at commercial nuclear power plants and contribute to the
radioactive source term in the burial site. The characteristics of
'the radioactive source term are of major importance in evaluating po-
tential movement in groundwater after emplacement in a shallow land
burial site. The development of this data was to be accomplished by
the analysis of systems of two PWRs and two EWRs operating in New York
State, the evaluation of relevant information in the literature,and
the compilation and interpretation of the available data.
The samples of BWR radwaste were obtained from the James A.
Fitzpatrick and Nine Mile Point Power Stations. The PWR samples were
obtained from the Indian Point 2 and R.E. Ginna Stations.
Dames & Moore reviewed the waste processing systems at the four
facilities, recommended a sampling program, and compiled and analyzed
the results of the radionuclide analyses performed under this study and
reported in the literature.
The Radiological Science Laboratory (RSL) of the New York State
Department of Health collected the samples at the reactor facilities, and
performed the laboratory analyses.
1-1
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2.0 Summary
This study considered the spent ion exchange resins, evap-
orator concentrates, and filter sludges produced at commercial
nuclear power plants and disposed of by burial at shallow land
burial sites. The dry solid rad-waste, which was not included,
is a major contributor to the total volume of low-level waste
generated, but a relatively minor contributor to the total
activity in the waste. The liquid radioactive waste collection
and treatment systems, in which the ion exchange units, evap-
orators, and filters are components, differ for BWRs and PWRs.
Generalized systems for each type of reactor,and the types of
liquid wastes treated ,are described in Secion 3.1. In addition,
the waste removal characteristics of the components are described
in Section 3.2 and 3.3.
The waste treatment systems in use at the four commercial
nuclear power plants at the time of sample collection were
reviewed and documented. The systems are described in Section
4.0 and Appendix A. In several instances on-site modifications
to the systems had been made since initial installation to
improve operations.
Samples were collected at the reactors early in 1976. There
were significant variations in plant operating history and the
size of the samples obtained at each of the reactors. These
variations are described in Section 5.1. The results of the
radiometric analyses of the samples are tabulated and dis-
cussed in Sections 5.2 through 5.4. The analytic methods used
to analyze the samples are described in Appendix B.
2-1
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The available literature describing reactor generated
waste was reviewed* so that the relevant data on radionuclide
analyses of similar types of waste could be extracted and
compared with the data developed under this program. This
data proved to be extremely limited, and the data that was
available did not include analyses for all the constituents
in the waste. The radiometric data that was extracted from
the literature and obtained from the plant operator's records
is presented in Sections 6.2 through 6.3.
An attempt to draw definitive conclusions from the
radionuclide analyses performed under this program and from
prior laboratory analyses was not feasible due to the lack
of a sufficient number of similar samples, and of information
on the operating experience pertinent to the samples collected.
However, certain analyses of selected segments of the data
were made (see Section 7.2) in an attempt to determine pre-
liminary trends. The interpretations as to radionuclide comp-
osition of the evaluated types of waste that can be supported
by the available data are presented in Section 7.3.
This study does provide preliminary indications of trends of the
radionuclide composition, and relative concentrations of radionuclides
to be found in three types of waste generated by LWRs and disposed of at
shallow land burial sites. In
*
A number of relevant references have become available since
the completion of this study and, with increasing interest
in this and related subjects, continue to be published.
These include:
1. M.J. Steindler and L.E. Trevorrow "Wastes from the Light
Water Fuel Cycle" presented at Waste Management - '76
(continued)
2-2
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addition, examination of the characteristics and operating
modes of the reactor waste processing system in conjunction
with the analytical results provides an insight into the
factors that need to be considered in developing an expanded
program of sampling and analysis.
Any further confirmatory programs should be designed
to permit collection of a sufficient number of samples having
the same parameters so as to be statistically reliable. The
parameters that need be considered are reactor and processing
system characteristics, type of waste, duration of reactor
operation, age of sample since generation, the location at
which the sample is collected within the waste system, age
of reactor and previous history. Uniformity in sample size
and procedures employed in analyzing the sample need be
maintained.
1.(continued) Tucson, Arizona Oct. 1976, to be published.
2. T.B. Mullarkey et. al., "A Survey and Evaluation of
Handling and Disposal of Solid Low-Level Nuclear Fuel
Cycle Wastes" Atomic Industrial Form, Inc. - Executive
Summary Oct. 1976.
2-3
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3.0 PWR and BWR Radioactive Waste Treatment Systems and Components
3.1 General Systems Comparison
In both a Pressurized Water Reactor (PWR) and a Boiling Water Reac-
tor (BWR) , the primary coolant is circulated through the reactor core
and takes up heat, This, in turn, produces steam for turning the tur-
bines which generate electrical power. The primary coolant in a BWR is
the source of steam while in a PWR the primary coolant is passed through
a heat exchanger and steam is produced in a secondary system. The primary
coolant, in most PWRs, contains boric acid which is used as a chemical shim
to control reactivity.
The primary coolant, in both a PWR and a BWR, picks up radioactive
corrosion products. Additional contamination of the coolant system re-
sults from the release of fission products from defective fuel elements
and diffusion of certain relatively mobile fission products (i.e. tritium)
through the intact fuel element cladding. These particulates and dissolved
solid contaminants are removed from the coolant stream by ion exchange
resins, filters and evaporators. Other contaminated solutions generated
at the reactor facility, particularly decontamination solutions, floor and
laboratory drain liquids and laundry water are treated in a similar manner.
Evaporators are utilized in both BWRs and PWRs to remove those par-
ticulates and dissolved solids that are not removed or compatible to re-
moval by ion exchange or filtration. In addition, most PWRs use evap-
orators to recover a portion of the boric acid. The evaporate is either
reused or discharged while the concentrate is sent to the radwaste building
for immobilization by incorporation into a matrix prior to packaging and
3-1
-------
shipment for disposal.
Generalized schematics of "typical" liquid and solid radwaste
treatment ^ystems at a BWR and PWR are shown in Figure 1 and 2 res-
pectively.
The liquid radwaste produced in nuclear power plants are cate-
gorized according to their physical and chemical properties. These cate-
gories vary between reactor types (PWR and BWR) and are shown in Table
3-1. W Within the reactor types, differences in design and operational
features also exist.
TABLE 3-1
LIQUID RADWASTE CLASSIFICATION
PWR BWR
Clean Wastes; low solid content High Purity Waste; liquids
liquids from controlled releases of low-electrical conductivity
and leaks from the primary and low solids content.
coolant loop. Primarily reactor coolant water.
Dirty Wastes; high solids Low Purity Wastes; Liquids
content and high electrical of intermediate electrical conductivity.
conductivity liquids including Primarily water collected from
those liquids collected from floor drains.
the containment buildings,
auxiliary buildings and Chemical Wastes; solutions of
chemical laboratory. caustic and sulfuric acid
which are utilized to
3-2
-------
Blew Down Wastes; continuous or
intermittent stream that is re-
moved fron the "bottoms" in the
stream side of the stream gen-
erator.
Detergent Waste; includes
liquids from laundry, personnel
and equipment decontamination
facilities.
Turbine Building Drain Waste;
leakage from secondary system that
is collected in the turbine building
floor sump.
regenerate spent resins as
well as solutions from
laboratory drains and
equipment drains.
Detergent Waste; laundry
and personnel and equipment
decontamination solutions.
A study comparing the volume and the activity of solid radwaste pro-
duced per thermal megawatt-hour of operation of BWRs and PWRs for the time
(2)
period 1959-1972 has shown that BWRs generated a significantly higher
volume of solid radwaste than PWRs, 1.50 x 10 ft3 per MW-hrt) and
0.56 x 10" 3 ft3 per MW-hr/. % respectively. However, the rate at which
activity was produced was essentially the same, 3.0 x 10~^Ci per
for BWRs and 3.04 x 10~5ci per MW-hr, .for PWRs. The specific activity
_ o
of the PWR waste therefore is much higher than BWR wastes, 5.5 x 10 zCi
per ft3 to 2.03 x 10~2Ci per ft3 respectively.
3-3
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T
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-T! R
REACTOR
COOLA
CLEAN
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WASTE
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-[SPECIAL TREATMENT]-,
POWDEX-
SOLKA-FLOC
3LUDOE
IBACKWASHI
STORAGE
TANK
HIGH-PURITY WASTES:
EQUIPMENT DRAINS,
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CONDENSATE DEMINERALI2ER
BACKWASH
MISCELLANEOUS
DRY
COMPRESSIBLE
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V//////A
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SHIELDED
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u a
FIGURE 1
Typical System for Treatment of Liquid and Solid
Radioactive Wastes at a Boiling Water Reactor (Ref.2)
3-4
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(BORIC ACID]
STORAGE
(REUSE)
L^N
ITOR]
ORNL DWG 73- 9938R2
W_ASTE CORE COMPONENTS
SPECIAL TREATMENT]
STORE SPENT
FILTER CAR-
TRIDGES AND
SLUDGES IN
DRUMS FOR
DECAY
TO ION
EXCHANGE
-REGENERANT «
' NEUTRALIZER ^
TANK J
t
MICAL
ITION
DECONTAM-
INATION
SOLUTIONS
_ FLOOR
DRAINS
LABORATORY
WASTES
SOME
-LAUNDRY
WASTES
I DIRTY OR
I MISCELLANEOUS
[ WASTE
TANK
V
V7777A
RAOWASTE BUILDING
I INCORPORATION IN
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yFORMALDEHYDE, E
DEWATER
(PUMP OR
CENTRIFUGE)
DECONTAMINATION
HYDRAULIC BALER
DRUM
STORAGE
SHIELDED
TRUCK
o a
FIGURE 2
Typical System for Treatment of Liquid and Solid
Radioactive Wastes at a Pressurized Water Reactor (Ref.2)
3-5
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3.2 Ion Exchange Resin Characteristics
The process of ion exchange is, essentially, a stoich-
iometric exchange between a resin and an electrolytic solution
of ions of the same sign and size as those in the resins. The
process is applicable only to those radionuclides in an ionic
state. Non-ionic nuclides or complexes (i.e. , insoluble,
neutral molecules and neutral complexes) show only a minor
response to treatment due primarily to a physical sorption
rather than an ion exchange process.
Strong-acid cation and strong-base anion exchange resins
of a polystyrene matrix are the types of resins most frequently
utilized by nuclear power stations. Mixed bed units ( a strong-
acid cation resin and a strong-base anion) are the most widely
used. Diagrams of the two types of exchange systems are shown
in Figure 3.
The liquid streams amenable to ion exchange in a BWR
are the primary coolant, the steam condensate and the liquid
radwaste system (including the fuel pool clean-up system) .
PWR liquid waste streams treated by ion exchange include
the primary coolant, the secondary coolant, the liquid radwaste
and the boron recycle (feed and concentrate). The treatment
of these streams varies from that of a BWR in that the letdown
from the primary coolant loop is treated by both separate and
mixed bed units. The boron recycle system uses a cation exchange
resin.
3-6
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(Ref. 3y
ORNL DWG. 72-13546
Co
I
WATER OR WASTE SOLUTION
INLET
DISTRIBUTOR
REGENERANT
INLET AND
BACKWASH
WATER
OUTLET
AIR VENT
DEIONIZED
SOLUTION
OUTLET AND
BACKWASH
WATER INLET
( Q) SEPARATE BED SYSTEM
WATER OR WASTE SOLUTION
INLET
DISTRIBUTOR
INLET FOR
CAUSTIC
REGENERANT FOR
ANION EXCHANGER
AND BACKWASH
WATER OUTLET
SPENT
REGENERANT
EFFLUENT
COLLECTOR
AIR VENT
AIR IN
SEPARATE
RESIN
LAYERS
REPRESENT
CONDITION IN
REGENERATION
DEIONIZED SOLUTION
OUTLET AND ACID
REGENERANT INLET
(AND BACKWASH WATER)
(b) MIXED-BED SYSTEM
Fig. 3. Schematic Diagram of Mixed-Bed and Separate-Bed Ion Exchange Systems.
-------
The radionuclides removed from the PWR waste streams by
ion exchange are essentially the same as those removed in
a BWR.
The life expectancy of an ion exchange system in a PWR
is lower than that of a comparable system in a BWR. This
is attributed to the fact that the chemicals added to the
primary and secondary coolant systems, for the purpose of
controlling reactivity and pH, will compete with activation
and corrosion products for available exchange sites within
the resin.
3.3 Evaporator Characteristics
Evaporators are used to treat those wastes which, due
to their physical and/or chemical characteristics, are not
compatible to treatment by filtration or ion exchange. In
PWRs, evaporators are used primarily on the clean and dirty
waste streams and in the boron recycle system. Evaporators
in BWEs handle, primarily, the chemical and low purity waste
streams.
An evaporator consists, basically, of two devices;
the first is a heating apparatus which transfers heat for
boiling to the solution or slurry; and the second is a
mechanism which separates the liquid and vapor phases. The
basic principles used in evaporator design are those of heat
transfer, vapor-liquid separation, volume reduction and
energy utilization.(D Diagrams of the two most commonly
utilized types of radwaste evaporators are shown in Figure 4
3-1
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(Ref .1 )
ORNL DWG. 73-8589 R
VAPOR
I
VD
.-.-MS 4--OEMISTER
APPROXIMATE
LIQUID LEVEL-
VENT4-
CUTAWAY VIEW OF
SHELL-ANO-TUBE
HEAT EXCHANGER
LIQUOR BOILING
INSIDE TUBES
DRIPS
VAPOR
«STEAM
(CONOENSINO
OUTSIDE TUBES)
CUTAWAY VIEW OF
SHELL-ANO-TUBE
HEAT EXCHANGER
LIQUOR BOILING
INSIDE TUBES
VENT-<
STEAM
(CONDENSING
OUTSIDE TUBES)
DEMISTER
+-FLASH CHAMBER
IMPINGEMENT
BAFFLE
APPROXIMATE
LIQUID LEVEL
FEED
THICK
LIQUOR
CALANDRIA-TYPE EVAPORATOR LONG-TUBE RECIRCULATION EVAPORATOR
Fig. 4 Typical Evaporators Used in Processing Liquid Radwaste.
-------
Und^r routine operating conditions, these radwaste
evaporators operate on a continuous or semi-continuous mode
as compared to a batch mode used at facilities with a low
volume of waste. When in a continuous mode of operations,
the waste is introduced into the evaporator in a predetermined
volume, boiling occurs and the vapors are continuously
removed, condensed, collected and treated. The evaporation
process is continued until the feed is expended or a pre-
determined concentration in the concentrate is obtained.
Once this concentration is reached, the concentrate is
transferred to the solid radwaste handling facility for
processing and packaging.
PWR evaporator concentrates, excluding the concentrate
from the boron recovery system, are primarily sodium borate
which results from the neutralization of boric acid from
primary coolant leakage. Those y-emitting radionuclides
present in the concentrate as reported in the literature
are predominantly Co58, Co60, Cs134, and Cs137at a total
concentration of approximately 0.2 yCi/ml.^4)
BWR evaporator concentrates, in comparison, are primarily
soldium sulfate which results from the use of sulfuric acid
and sodium hydroxide to regenerate ion exchange resins. The
Y-emitting radionuclides present in the concentrate as
reported in the literature are predominantly Co58, Co60,
Cs134, and Cs137 and at a concentration in the range of 2.0-
3.0 yCi/ml. <4>
3-10
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4.0 Waste Treatment Systems at Sampled Reactors
The liquid radwaste systems at a commercial nuclear power plant
are designed to collect, monitor and process for reuse or disposal, all
potentially radioactive liquid wastes. The residues of the processing of
the liquid waste streams, are evaporator concentrates, filter sludges, and
spent ion exchange resins. These materials are immobilized by different
techniques, packages and shipped to a burial site for disposal.
The nuclear power plants participating in this study provided
representative radwaste samples from both PWR and BWR systems having a
range of operating lifetimes. The power plants sampled were;
(1) Indian Point No. 2- A PWR operated by the Consolidated
Edison Company having a net capacity of 873 MWe that began commercial
operation in August, 1973.
{ 2) R.E. Ginna- A PWR operated by the Rochester Gas &
Electric Company having a net capacity of 420 Mtfe that began commercial
operation in July, 1970.
(3) Nine Mile Point- A BWR operated by the Niagara Mo-
hawk Power Corporation having a net capacity of 610 MWe that began com-
mercial operation in December, 1969.
(4) James A. Fitzpatrick- A BWR operated by the Power
Authority of the State of New York having a capacity of 821 Mfle that
began correnercial operation in July, 1975,
The liquid and solid radwaste systems used at these four power
plants at the time of sample collection were reviewed from the available
4-1
-------
literature (5,7,8,9,11) and through personal visits to each of the
plants and conversations with the knowledgeable plant personnel.
(6,10,12) In several instances, the systems had been modified from
the published descriptions. The radwaste systems at each of the fac-
ilities are described in Appendix A.
4-2
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5.0 Radionuclide Analyses of Waste Samples Collected During This Program
5.1 Sample Definition and Collection Procedures
The program provided funds for the collection and analysis of ten
(10) samples distributed among four (4) reactors. This approach, it
was considered, would provide analyses of the waste constituents from
two (2) PWRs and two (2) BWRs of varying periods of accumulated operat-
ing time since start up, and permit comparison of the differences in
radionuclide concentrations resulting from these factors.
The reactor facilities were visited by Dames & Moore personnel and
discussion held with plant personnel to determine the accessibility of
the waste processing and packaging, and the availability of each type
of sample. Based on the information obtained, a sampling program was
recommended consisting of two (2) evaporator concentrate samples from
each facility. Resin samples and filter sludges would not be collected
because the reactor operators had indicated that these samples could
not readily be made available. When the Radiological Science Laboratory
(RSL) collected the samples at the facilities, it was necessary to
revise this program because the reactor operators were able to make
available certain filters and resin samples and could not provide all
the specified evaporator concentrate samples. The samples collected
from each reactor, and the conditions under which they were collected,
are described in Table 5-1.
5-1
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TABLE 5-1
Description of Collected Samples
Reactor
R.E. Ginna
Waste Type
Evaporator Concentrate
Filter Sludge
Spent Resin
Indian Point No. 2
Evaporator Concentrate
Filter Sludge
Spent Resin
Description
<20 ml sample collected; high
undissolved solids and salts
content which hampered titration;
small sample size and solids
content prevented volumetric
conversion and required reporting
of results on a weight basis.
Sample consisted of 3 surface
smears of the Primary Coolant
Filters which had been in-line
for approximately 1 year; Station
Health Physicist considered
collection of an actual filter
sample to be inadvisable due to
>100mr potential personnel
exposures; nature of sample
required reporting of results
on a per filter basis.
~1 ml sample of resin beads (wet)
collected from spent resin storage
tank; length of time resin in-line
is unknown; results reported on
a weight basis because volume of
beads could not readily be measured
25 ml sample collected from com-
posite evaporator in Station No.1
and diluted to 500 ml; results
expressed on volumetric basis.
Sample obtained from filter placed
in the tap line of the primary
coolant system thru which 304
liters of primary coolant was
passed; Station Health Physicist
considered collection of an actual
filter sample to be inadvisable
due to >100mr potential exposures;
results reported on a per filter
basis.
No sample collected.
5-2
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TABLE 5-1 (cont'd.)
Reactor
Nine Mile Point
Waste Type
Evaporator Concentrate
Filter Sludge
Spent Resin
James A. Fitzpatrick
Evaporator Concentrate
Filter Sludge
Spent Resin
Description
~1 ml sample of unknown age and
which had been previously collected
and stored at site; high, undis-
solved solids and salt content
which hampered titration; results
reported on a weight basis.
~1 ml sample of unknown age
collected from the sludge storage
tank; results reported on a weight
basis.
~1 ml sample of resin beads (dry)
collected from spent resin storage
tank; length of time resin in-line
is unknown; results reported on a
weight basis.
1 liter sample collected; solids
content unknown; results reported
on volumetric basis.
~45 ml of dry centrifuge waste
(powder) collected; results
reported on weight basis.
No sample collected.
5-3
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As can be noted from examination of Table 5-1, the samples of
each type of waste collected varied in size, prior history, and in the
case of the filter sludge, in the type of sample collected. Thus many
of the factors that may influence the radionuclide composition of the
waste types vary from sample to sample, making a comparison among
sample analyses difficult. The comparative analyses that can be made
between similar waste types from the different reactors are provided
in Section 7.0.
The analytical procedure employed by RSL to analyze the samples
of each type of waste collected are described in Appendix B for the
various radionuclides evaluated.
5.2 Spent Ion Exchange Resins
The results of the radiometric analyses of spent ion exchange
resins performed by RSL are presented in Table 5-2. Both the concen-
trations and the relative percent of the individual radionuclides in
each sample are provided. Samples were available from only the Nine
Mile Point (BWR) and R.E. Ginna (PWR) facilities.
Although the percentages of the radionuclides present vary
between the two samples, in each instance three of the radionuclides,
Cs , Cs and Co , account for approximately 90% of the
total concentration, with Cs being the predominant radionuclide
in both samples. The concentrations of each sample are quite compar-
able, 43.23y Ci/gm for Nine Mile Point, and 41.03U Ci/gm for Ginna.
5-4
-------
TABLE 5-2
Radionuclide Analysis of Spent Ion Exchange
Resin Samples Measured Under This Program
Radionuclide Nine Mile Point R.E. Ginna
* 241
Am
Pu239,240
Pu238
U238
U235
u234
Ce144
Csl37
Cs"4
I131
I129
125
Sb
Sb124
106
Ru
99
Tc
Zr95
Nb95
90
Sr
Zn65
Ni63
r, 60
Co
59
Fe
(f.p.).
(f.p.)
(f.p.)
(f.p.)
(f.p.)
(f.p.)
(f.p.)
(f.p.)
(f.p.)
(f.p.)
(f.p.)
(f.p.)
(a. p.)
(f.p.)
(f.p.)
(f.p. or a. p.)
(f.p. or a. p.)
(f.p.)
(a. p.)
(a. p.)
(a. p.)
(a. p.)
Concen- Relative
tration Proportion
(VJCi/gm) (%)
6 x 10~5 <.l
3 x 10~5 <.l
4 x 10~5 <.l
<3 x 10~6 <.l
<5 x 10~6 <.l
<7 x 10~6 <.l
0.12 0.3
31.7 73.4
2.9 6.7
ND
<2 x 10~6 <.l
0.11 0.3
ND
0.3 0.7
9 x 10~3 <.l
ND
ND
7.2 x 10~2 0.2
8 x 10~2 0.2
9.7 x 10~ <.l
6.24 14.5
ND
Concen- Relative Pro-
tration portion
(yCi/gm) (%)
7 x 10 4 <.l
8 x 10~4 <.l
4 x 10~4 <.l
4.5 x 10~5 <.l
<1.2 x 10~5 <.l
2.3 x 10~5 <.l
0.3 0.7
21.9 53.4
12.4 30.2
ND
6 x 10~4 <.l
0.2 0.5
ND
0.7 1.7
2 x 10~3 <.l
ND
ND
8.5 x 10~2 0.2
0.15 0.4
1.39 3.4
2.06 5.0
ND
5-5
-------
TABLE 5-2 (cont'd.)
Rad i onuc1id e
Co
58
Co
57
Fe
55
Mn
54
(a.p.)
(a.p.)
(a.p.)
(a.p.)
Cr51 (a.p.)
cl4(co2)
-------
It should be noted that the major constituents delineated in this
and succeeding sections on the basis of measured concentration in the
samples will not necessarily be the major constituents remaining over
the long term after radionuclide decay has occurred.
5.3 Evaporator Concentrates
The results of the radiometric analyses of the evaporator concen-
trate samples are presented in Table 5-3. The concentrations and
relative percent of each radionuclide are provided. Samples were
analyzed from all four of the reactors.
The radionuclide concentration vary from sample to sample within
each reactor type, and between reactors. In the sample from Indian
Point No. 2, the major constituents in order of predominance are I ,
Cs , Cs , Co , and Fe , comprising approximately 90% of
the total concentration. In the evaporator sample from Ginna, H ,
Cs , Cs , Co , and Co , in that order, are the major con-
stituents, comprising approximately 95% of the total concentration.
In the case of Nine Mile Point, Fe55, Cs137, Cs134, Co60, Mn54
represent 95% of the total concentration of the sample. While
in the sample from Fitzpatrick the major constituents in order of
predominance are Mn , Co , Co , Cr , and Zn comprising
approximately 87% of the total concentration.
5-7
-------
Table 5-3
Padionuclide Analysis of Evaporator Concentrate Sairples
Measured under this Program
Radionuclide
Am241
Pu239,240
Pu238
u238
u235
u234
144
Ce
Cs137
Cs134
jl31
j-129
sb125
Sb124
Ru106
To"
Zr95
Indian Point No. 2
Concen- Relative
tration Proportion
(yci/ml) (%)
3.0 x 10"7 <.l
8.0 x IO-8 <.l
2.0 x 10"7 <.l
1.9 x icf <.l
8.0 x 10"8 <.l
1.2 x 10"7 <.l
ND
0.3 21.8
0.19 13.8
0.41 29.8
2.0 x 10"5 <.l
ND
ND
0.007 0.5
2.0 x 10~5 <.l
ND
R.E Ginna
Concen- Relative
tration Proportion
(yci/gm) (%)
ND
1.8 x 10~6 <.l
1.0 x 10~6 <.l
1.88 x 10"6 <.l
2.0 x 10"7 <.l
3.0 x 10~7 <.l
8.0 x 10"4 0.2
0.102 29.9
-2
3.7 x 10 10.8
ND
4.0 x 10"6 <.l
1.0 x 10~2 0.3
1.0 x 10"4 <.l
2.0 x 10"3 0.6
7.0 x 10~5 <.l
6.0 x 10~4 <.l
Nine Mile Point
Concen- Relative
tration Proportion
(yci/gm) (%)
5.0 x 10~6 <.l
8.0 x 10"6 <.l
1.3 x 10"5 <.l
1.5 x 10~6 <.l
2.0 x 10~6 <.l
3.0 x 10~6 <.l
6.0 x 10~3 0.7
0.229 27.0
0.169 19.9
ND
1.0 x 10"4 <.l
6.0 x 10"3 0.7
ND
1.9 x .10~2 2.2
1.0 x 10"3 0.1
ND
J.A. Fitzpatrick
Concen-
tration
(yCi/ml)
ND
5.5 x 10"8
-8
1.6 x 10
7.0 x 10~9
1.0 x 10~8
1.6 x 10~8
2.0 x 10~4
4.0 x 10"4
-4
1.0 x 10
ND
4.0 x 10"7
2.0 x 10"4
4.0 x 10~4
8.0 x 10"4
1.6 x 10~6
5.0 x 10~4
Relative
Proportion
(%)
<.l
<.l
<.l
<.l
<.l
0.4
0.9
0.2
<.l
0.2
0.9
1.8
<.l
1.1
5-j
-------
TABLE 5-3 (cont'd)
Radionuclide Indian Point No. 2
~ -i_ii\-i j-cni ITU-LUC 1NO . Z
R.E. Ginna
Nine Mile Point
,T . A . Pi -t-smat-i-i nlr
Concen- Relative Concen- Relative Concen- Relative Concen- Relative
tration Proportion tration Proportion tration Proportion tration Proportion
(yCi/ml) (%) (yci/gm) (%) (yCi/gm) (%) (yCi/ml) (%)
Nb95
Sr90
Zn65
Ni"
Co60
59
Fe
Co58
Co57
55
Fe
Mn54
Cr51
P14
C (C02)
r14
C (CH4)
H3
ND
7.0 x 10~5 <.l
ND
1.91 x 10~2 1.4
3.5 x 10~2 2.5
_ o
3.0 x 10 0.2
0.1890 13.7
3.0 x 10~4 <.l
0.1280 9.3
3.1 x 10~2 2.3
-2
3.64 x 10 2.6
2.1 x 10~5 <.l
2.1 x 10~7 <-l
2.72 x 10~2 1.97
ital Concentration 1.3759
7.0 x 10~4 .1
7.6 x 10~5 .1
3.0 x 10~4 .1
6.1 x 10~3 1.8
1.89 x 10~2 5.5
ND
-2
3.46 x 10 10.1
1.2 x 10~4 <.l
4.4 x 10~3 1.3
1.0 x 10~3 0.3
ND
6 x 10~5 <.l
4.0 x 10~7
0.132 38.6
0.3417
ND
1.3 x 10~3 0.2
4.0 x 10~3 0.5
2.2 x 10~3 0.3
9.6 x 10~2 11.3
ND
ND
5 x 10~4 <.l
0.2900 34.1
2.3 x 10~2 2.7
ND
1.8 x 10~6 <.l
9.0 x 10~ <.l
2.5 x 10~3 0.3
0.8492
9.0 X 10~4
7.0 X 10~?
3.4 x 10~3
1 x 10~4
8.9 x 10~3
ND
-2
.09 x 10
3 x 10~5
7 x 10~4
1.18 x 10"2
3.7 x 10~
7.1 x 10~6
6 x 10~9
1.7 x 10~3
0.0447
2.0
.1
7.6
0.2
19.9
24.4
< .1
.1
26.4
8.3
<.l
<.l
3.8
5-9
-------
5.4 Filter Sludges
The results of the radiometric analyses of the filter sludge or
se-tected "equivalent" samples are presented in Table 5-4. The data is
here again reported as concentrations and relative percent of each
radionuclide of the total concentration for each sample. Samples, of
varying origin (see Table 5-1), were collected from each of the
reactors.
Again substantial variation in relative concentrations of the
various radionuclides can be noted among the four samples. In the
filter sludge sample from Indian Point No. 2 the major constituents in
order of predominance are Cr , Co , and Co which together comprise
in excess of 95% of the total specific activity of the sample. In the
analysis of the surface smears of the primary coolant filter from
Ginna, containing sludge particles representative of the material
collected for packaging, Fe , Co , and Ni in that order are
the major constituents comprising approximately 86% of the total
specific activity. In the sample from the sludge storage tank
at Nine Mile Point the major constituents are Fe , Cs , and
134
Cs comprising approximately 88% of the total specific activity of
the sample. And in the powdered dry centrifuge waste sample from
Fitzpatrick, Co , Mn , Fe and Co comprise approximately
88% of the activity in the sample. As in the case of the evaporator
samples, the various types of filter sludge samples exhibit a wide
range in total activity.
5-10
-------
Table 5-4
Radionuclide Analysis of Filter Sludge
Samples Measured Under This Program
Rad ionuc 1 ide
* 241
Am
Pu239,240
Pu238
U238
U235
u234
Ce144
Cs137
Cs134
I131
I129
Sb125
Sb124
RU106
Tc"
Zr95
Indian Point No. 2
Concen- Relative
tration Proportion
(yCi/f liter) (%)
ND
5.5 x ID"5 <.l
1.3 x 10~5 <.l
3.0 x 10"6 <. 1
4.0 x 10"6 <.l
6.0 x 10~6 <. 1
3.0 x 10-2 o.l
0.1510 0.6
0.1260 0.5
ND
-5
8.0 x 10 <. 1
3.9 x 10~2 0,2
ND
0.1 0.4
1.4 x 10~3 <.!
5.3 x 10"2 0.2
R.E. Ginna
Concen- Relative
tration Proportion
(yci/filter) (%)
3.07 x 10~4 <.l
5.9 x 10"4 <.l
2.35 x 10" <.l
6.0 x 10"7 <.l
-7
4.0 x 10 <. 1
2.6 x 10"6 <.l
1.4 x 10"2 0.9
4.1 x 10"3 0.3
9 x 10"4 <. 1
ND
1.8 x 10 <. 1
4.1 x icf 0.3
ND
3.9 x 10"2 2.5
9 x 10~4 <. 1
1.3 x 10"2 0.8
Nine Mile Point
Concen- Relative
tration Proportion
(VCi/gm) (%)
1.8 x 10~5 <.l
1.5 x 10~4 <. 1
2.8 x 10~4 <.l
2.0 x 10"5 <. 1
1.8 x 10"5 <. 1
3.0 x 10~ <. 1
6 x 10~ 0.5
1.130 9.0
0.7400 5.9
ND
1.1 x 10~4 <.l
4 x 10~2 0.3
ND
8 x 10"2 0.6
8 x 10"2 <.l
ND
J.A. Fitzpatrick
Concen-
tration
(VCi/gm)
2.5 x lo"6
5.0 x 10"7
1.2 x 10"6
1.9 x 10~6
5.0 x 10~7
7.0 x 10"
2 x 10"3
8 x 10"4
8 x 10~4
ND
3.0 x 10~6
1.9 x IO3
ND
1.1 x 10"3
6 x 10
6 x 10"3
Relative
Proportion
<.l
<.l
<.l
<.l
<.l
<.l
.1
<.l
<.l
<1
0.1
<1
<.!
0.3
5-U
-------
TABLE 5-4 (cont'd.)
Radionuclide
Indian Point No. 2
Nb95
Sr9°
Zn65
63
Ni
Co60
Fe59
Co58
Co"
55
Fe
Mn54
51
Cr
c14
(C02)
/ .'-ITT \
Concen-
tration
(yCi/filter)
0.1090
7 x 10
8.5 x 10~2
1.1 x 10-2
1.9100
0.1880
9.4000
1.9 x 10~2
4.38 x 10~2
0.3230
10.5000
2.36 x 10~2
7 x 10~
Relative
Proportio:
0.5
<.l
0.4
<.l
8.4
0.8
40.7
<.l
0.2
1.4
45.9
0.1
R.E. Ginna
Nine Mile Point
J.A. Fitzpatrick
ND
Concen-
tration
(yCi/filter)
2.38 x 10~2
-2
1 x 10
2.5 x 10~3
9.3 x 10~2
0.2550
ND
3.9 x 10~2
4 x 10~4
0.9800
1.25 x 10~2
1.2 x 10~2
3.9 x 10~2
8.0 x 10~8
-3
1.3 x 10
Relative
Proportion
(%)
1.5
0.6
0.2
6.0
16.5
2.5
<-l
63.4
0.7
0.8
2.5
<.l
0.1
Concen-
tration
(yCi/gm)
ND
5.7 x 10~2
-2
6.9 x 10
2.8 x 10~2
1.5300
ND
6.4 x 10~2
2 x 10"3
7.7000
0.5300
0.5000
1 x 10~3
7.0 x 10~5
2.0 x 10~3
Relative Concen-
Proportion tration
(%) (yCi/gm)
1.13 x 10"2
0.5 <1 x 10~4
0.6 3.1 x 10~2
-3
0.2 7.4 x 10
12.2 0.1230
2.9 x 10~2
0.5 0.262
<.l 0.800
61.4 0.28
4.2 0.346
4.0 ND
<.l 1 x 10~4
<.l 2.5 x 10~7
<.l ND
Relative
Proportion
0.6
<.l
1.6
0.4
6.4
1.5
13.8
41.8
14.6
18 J
<.l
<.l
5- 12
-------
6.0 Radionuclide Analyses of LWR Waste Performed Under Other Programs
The limited amount of available literature containing specific
analyses of similar waste types generated at LWRs was reviewed to
determine the extent of available data that could be used as a basis
for comparison with the data developed under this program. The review
revealed that the amount of relevant data is limited, and that which
is available does not provide a complete breakdown of all the consti-
tuents to be found in the various waste forms. Most of the reported
analyses of waste, as concentrates, filter sludge, or resins, or as
solidified wastes, were of gross activity levels measured to assure
that the plant systems were continuing to operate within specifications
and that no anomalies were present.
Of the available semi-annual Effluent and Waste Disposal Reports
prepared for the four reactors whose wastes were analyzed under this
study, only those for Ginna provided any relevant radionuclide
breakdown.
The data of relevance to this study that was available from prior
work is presented in the succeeding sections. It was obtained both
from published reports ' and from the plant operators records.
The data is limited to waste generated at PWRs; comparable BWR
data is not available in the literature.
6.1 Spent Ion Exchange Resins
The results of the radiometric analyses of spent resin samples
performed by the laboratory personnel at Ginna and Indian Point
6-1
-------
/ -i 2\ (14)
No. 2 v ' and that reported for an unnamed PWR are presented
in Table 6-1. The analyses are limited to certain significant gamma
emitting radionuclides which, it can be noted, vary considerably among
the samples analyzed. Both the concentrations (or total activities)
and relative percent of the individual radionuclides analyzed in each
sample are provided.
6.2 Evaporator Concentrates
The results of the radiometric analyses of evaporator concentrates
from the same three reactors reported on for spent resin constituents
are presented in Table 6-2. The analyses are similarly limited,
covering only certain gamma emitting radionuclides which also show
substantial variation from sample to sample.
Data available from monthly analyses of evaporator concentrate
samples collected from the Indian Point No. 2 reactor during four of
the months in the period from July to December 1975 provides an
indication of the variation of concentrations of certain of the
radionuclides with time. The concentrations of the measured radio-
nuclides and their relative proportion of the total activity are
presented in Table 6-3.
The concentration of each constituent in the total sample vary
during the six month sampling period, (corresponding to semi-annual
reporting period for the reactors), but no consistent pattern can be
found. For example, the concentration of Cs134 showed a variation
greater than 30 fold during this period, while the concentration of
6-2
-------
TABLE 6-1
en
i
UJ
Radionuclide
Cs137
Cs134
Sb125
Co60
Co58
Mn54
Eadionuclide Analysis of PWR Spent Resin Samples
Measured Under Other Analytic Programs
Reactor
RE Ginna^13' Indian Pt.2<12) Unidentified PWR^14)
Relative Relative Relative
Concentration Proportion* Activity Proportion* Concentration Proportio
(yCi/ml) (%) (Ci) (%) (yci/ml) (%)
28.23 30.2 5.77 72.0 14.32 47.6
7.38 7.9 1.31 16.0 3.89 12.9
N.R. N.R. 0.137 0.5
37.29 39.9 0.91 11.0 5.23 17.4
N.R. N.R. 3.51 11.7
20.47 21.9 0.03 1.0 2.97 9.9
*of only those radionuclides analyzed.
-------
TABLE 6-2
Radionuclide Analysis of PWR Evaporator Concentrate Samples
Measured Under Other Analytic Programs
Radionuclide Reactor
RE Ginna^13) Indian Pt.2^12'
Relative Relative
Concentration Proportion* Concentration** Proportion*
(yCi/ml) (%) (yci/ml) (%)
Cs137 0.542 60.4 0.718 43.3
Cs134 0.272 30.3 0.532 28.5
I131 0.013 1.3 N.R.
Co60 0.045 5.0 0.020 2.6
Co58 N.R. 0.388 22.2
Mn54 0.021 2.2 0.014 1.4
Unidentified
Concentration
(yci/ml)
0.000532
U.R.
N.R.
0.00599
0.000523
0.000514
PWR^4)
Relative
Proportio:
7.0
79.3
6.9
6.8
*of only those radionuclides analyzed.
**average concentration July-December 1975.
NR = Not Reported
-------
TABLE 6-3
Radionuclide Analysis of Evaporator Concentrate Samples From Indian Point No. 2 Reactor
July, September, October, December, 1975
Radionuclide Month
July
Concen- Relative
tration Proportion*
(yCi/ml) (%)
Cs137 2.350 55.5
Cs134 1.810 42.8
Co60 0.017 0.4
Co58 0.019 0.5
Mn54 0.020 0.5
September
Concen- Relative
tration Proportion*
(yCi/ml) (%)
0.169 40.4
0.113 27.0
0.017 4.1
0.113 27.0
0.007 1.6
October
Concen-
tration
(yci/ml)
0.249
0.149
0.024
0.147
0.017
Relative
Proportion*
42.5
25.4
4.0
25.1
3.0
December
Concen-
tration
(yCi/ml)
0.105
0.057
0.020
0.109
0.011
Relative
Proportioi
34.8
18.8
6.6
36.1
3.6
*of only those radionuclides analyzed.
-------
Co remained essentially constant. On a relative basis, the
concentrations in each sample also tended to show an inconsistent
134
pattern of variation. For the same two radionuclides, Cs ranged
from a high of 43% of the total sample measured to a low of 19%, while
Co ranged from a high of almost 7% to a low of less than 1%.
6.3 Filter Sludges
The results of the radiometrc analyses of filter sludges collected
from Indian Point No. 2 ^ ' and from the unnamed PWR ^ ' are
presented in Table 6-4. The filter samples analyzed from the latter
include samples from the spent fuel pool, the reactor coolant system,
and the waste holding tank, while the sample from Indian Point No. 2
is from a single unidentified location in the system. The tabulation
shows both the concentration and relative percent of the individual
radionuclide in each sample.
6-6
-------
TABLE 6-4
01
i
Radionuclide
Sb
125
Cd
115m
Cd
113m
Ag
110m
Zr
95
Zn
65
Co
60
Co
58
Co
57
Mn
54
Cr
51
Indian Pt. 2
RCS
(12)
Concen- Relative
tration Proportion
(uCi/gm)xlQ3 (%)
N.R.
N.R.
N.R.
N.R.
N.R.
N.R.
1.89 15.1
3.65 29.2
N.R.
0.362 2.9
6.60 52.8
ysis of PWR Filter Sludge Samples Measured
der Other Analytic Programs
Unidentified PWR^14'
SFP
Concen-
tration
(yci/gm)
0.581
7.750
0.049
0.327
0.131
N.R.
2.490
19.4
0.036
1-32
14.4
Relative
Proportion
(%)
1.2
16.6
0.1
0.7
0.3
5.3
41.8
0.1
2.9
30.9
RCS WHT
Concen-
tration
(yci/gm)
N.R.
N.R.
N.R.
1.851
N.R.
N.R.
6.80
28.6
0.109
2.27
N.R.
Relative Concen-
Proportion tration
(%) (yci/gm)
.0124
N.R.
N.R.
4.7 .0144
N.R.
.0099
17.1 .509
72.2 .464
0.3 -0021
5.7 .0707
N.R.
Relative
Proportion
(%)
1.1
1.3
0.9
47.0
42.9
0.2
6.5
SFP = Spent Fuel Pool
RCS = Reactor Coolant System
WHT = Waste Holding Tank
NR = Not Reported
-------
7.0 Comparisons, Interpretations, and Recomnrendations
This study provides a preliminary data base on the radionuclide
composition, and actual and relative concentrations to be found in the
oredominant waste forms qenerated by the four LWRs, which are processed, packaged
and shipped to conmercial radioactive waste burial sites. In addition,
examination of the characteristics of the waste processing systems used
to process and package the waste, and the analytical results provides
an insight into the factors that need be considered in establishing a
future expanded program of sampling and analysis.
7.1 Variables Influencing Composition of Waste Samples
The composition and relative radionuclide concentration in the
samples of waste generated at LWRs is influenced by the following
factors;
(a) Type of reactor and waste processing systems.
(b) Extent of release of fission products from failed fuel
elements in the reactor core into the primary coolant (primarily a function
of reactor operating time).
(c) Extent of corrosion products in the primary coolant
(primarily a function of reactor operating time).
(d) Type of waste form sampled (i.e., filter sludge, resins,
or evaporator bottoms).
(e) Location in waste processing system sample is drawn from
(e.g., in individual waste streams vs. mixture in collection tanks).
7-1
-------
(f) Age of sample from time of initial generation of the
waste to time of analysis (concentrations of radionuclides will
change as a function of half lives).
In addition to the above noted factors, the ability to
accurately determine the composition of the sample is a func-
tion of sample size, solids content, and analytic procedures
followed in the laboratory.
7.2 Comparisons of Radionuclide Analyses
A study of LWR wastes would be most useful if the pattern
of radionuclide concentrations could be ascertained for the
types of waste examined, so that information could be developed
about concentrations of the radionuclide in the processed waste
shipped to the burial site.
An attempt to draw definitive conclusions from the data
obtained under this program and from prior laboratory analyses
was impossible due to the lack of a sufficient number of similar
samples necessary to provide statistical accuracy, and due to a
lack of information on the operating experience pertinent to the
samples collected. However, analyses can be performed to determine
preliminary trends from the selective examination of classes of
radionuclides in specific waste forms.
The data from the evaporator concentrate and spent ion
resin waste form was used for this comparison. The filter sludge
analyses were not considered due to the variability in sample
form, lack of information on sample history and wide range in
reported analyses.
7-2
-------
In the case of the evaporator concentrates the variables effecting the
data are further limited by considering the gamma emitting radionuclides
reported in the literature (See Section 6.0) for the two PWRs, R.E.
Ginna and Indian Point No. 2, for which data was compiled under this
program. In addition, the effect of variation in sample composition as
a result of the differential decay of the radionuclide inventory
in the period between generation of the waste and sample analysis is
minimized by further limiting the comparison to those radionuclides
having half lives greater than 300 days. With the restrictions, it is
felt unat direct comparison of the selected radionuclide concentrations
can be made. Table 7-1 presents the radionuclide concentrations and
relative proportion of total activity of the selected nuclides for
^ 137 134 n 60 , M 54
Cs , Cs , Co and Mn .
A similar restricted comparison was then made of the evaporator
concentrate analyses determined under this program for the two PWRs
and two BWRs. This data is presented in Table 7-2.
In the case of the spent ion exchange resin, the same type of
analysis was applied to all of the long lived gainma emitters reported
in the sample analyses from both this program and all those reported
in the literature. This data is presented in Table 7-3.
7-3
-------
TABLE 7-1
Comparison of Concentrations of Gamma Emitting Radionuclides
(T!j>300 days) In Samples of Evaporator Concentrate From PWRs
Radionuclide
Reactor
Indian Point No. 2
R.E. Ginna
Cs Concentration
Relative Prop.
Cs
134
Co
60
Mn
54
Sample of
9/75
0.169yci/ml
55.2%
0.113
36.9
0.017
5.6
0.007
2.3
Sample of
3/76
0.300yCi/ml
54.0%
0.190
34.2
0.035
6.3
0.031
5.6
Sample of
1975
0.542yCi/ml
61.6%
0.272
30.1
0.045
5.1
0.021
2.3
Sample of
2/76
0.102jjCi/gm
64.2%
0.037
23.2
\
\
0.019
11.9
0.001
0.6
Total Concentration
in Selected Sample
0.306
0.556
0.880
0.159
7-4
-------
TABLE 7-2
Comparison of Concentrations of Gamma Emitting Radionuclides
(Tis>300 days) in Samples of Evaporator Concentrate From PWRs & BWRs
Radionuclide
Concentration
Relative Prop.
PWR
BWR
Cs
134
Co
60
54
Mn
Total Concentration in
Selected Sample
Indian Point No.2 R.E. Ginna
0.300yci/ml
54.0%
0.190
34.2
0.035
6.3
0.031
5.6
0.556
0.102yci/gm
64.2%
0.037
23.2
0.019
11.9
0.001
0.6
0.159
Nine Mile Point J.A. Fitzpatrick
0.229yCi/gm
44.3%
0.169
32.7
0.096
18.6
0.023
4.4
0.517
0.0004yCi/ml
1.9%
0.0001
0.5
0.0089
42.0
0.0118
55.7
0.0212
7-5
-------
TABLE 7-3
Comparison of Concentrations of Gamma Emitting Radionuclides
(TJ5>300 days) in Samples of Spent Ion Exchange Resins From PWRs & BWRs
Radionuclide
R.E. Ginna
Indian Point No.2
Unidentified PWR
Nine Mile Point
Cs
134
Co
60
Mn
54
Concentration
Relative Prop.
I
CTi
28.23UCi/ml
30.2%
7.38
7.9
37.29
39.9
20.47
21.9
21.9yci/ml
6.0%
12.4
34
2.06
5.6
.16
.4
5.77Ci
30.2%
1.31
7.0
11.0
57.7
1.0
5.2
1.43yCi/ml
54.2%
0.39
14.8
0.52
19.7
0.30
11.4
31.7yCi/gm
77.5%
2.9
7.1
6.24
15.2
0.09
.2
Sample from other programs
Sample from this program
-------
7.3 Interpretations of Data
Interpretations can be made of the radionuclide analyses determined
under this and prior programs, and comparisons made between selected portions
of the data, with the proviso that these interpretations are of preliminary
trends (or patterns) and certainly cannot be considered to be definitive.
The following interpretations appear to be justifiable.
I. Evaporator Concentrates
(a) Of the three waste forms examined, the consistency of the
sample sources and of identifiable patterns in the data permits ^nre
supportable conclusions to be drawn with regard Lo evaporator concentrate
compositions.
(b) The comparison of the relative concentrations of long half
lived gamma emitting radionuclides (see section 7.2) shows that, with
the exception of the sample from Fitzpatrick, the relative proportion
of the constituents appears to be essentially of the same order for each
reactor sampled under this program; and for the FWRs (where data was
available) essentially of the same order as a function of time. This
may imply a pattern in the relative concentrations of all the radio-
nuclides in the evaporator concentrate samples. This initial pattern
should serve as a reference point for future more detailed studies.
(c) The predominant gamma emitting radionuclides present in
evaporator concentrates from all the reactors, with the exception of
the samples fron Fitzpatrick, are Cs , Cs^-34f CQ , and Co , generally
in that order. This agrees with the information provided in the lit-
erature (Ref. 4). On the basis of the half lives of the gamma emitters,
Cs , Cs , Co , and Mn will generally be predominant in the buried
waste. Furthermore, Fe , Ni , and H , because of their long half lives
7-7
-------
must also be considered as potential major constituents of the buried
waste. It is reiterated that significant concentrations of individual
radionuclides in the analyzed samples are not necessarily indicative of
the relative long term importance of the radionuclides in terms of re-
lease and migration potential.
(d) The lack of agreement between the radionuclide analysis
in the sample from Fitzpatrick and the other reactors along with its
significantly lower total activity may be attributed to the short period
of system operation at Fitzpatrick. It would be anticipated that the
contribution from corrosion and fission products would be minimal during
the early stages of reactor operation. Thus, the majority of the radio-
nuclides present are activation products, while at the older plants,
fission products tend to predominant. This can be related to the greater
integrity of the fuel cladding in the early phases of plant operation.
(e) The data from radionuclide analyses of evaporator con-
centrate samples taken over a period of months from Indian Point No. 2
(see section 6.2) show appreciable variations in actual and relative con-
centrations of the radionuclides which cannot be correlated with reactor
operations.
II. Spent Ion Exchange Resins
(a) The results of the various radionuclide analyses reported
herein are too inconsistent to permit any trends to be discussed in the
actual or relative concentrations of radionuclides. The comparison of
the relative concentrations of the long lived gamma emitting radio-
nuclides (see section 7.2) does not show, as it did in the case of the
evaporator concentrates, any repeatable pattern among the sanples.
7-8
-------
(b) The predominant radionuclides present in spent ion exchange
resin samples from all of the reactors are Cs134, Co60, and Mn54, which
occur in varying proportions in each sample. Since these radionuclides
are all relatively long lived (1^>300 days), they will generally be pre-
dominant in the buried waste.
Ill Filter Sludges
(a) The results of the various radionuclide analyses reported
herein are too inconsistent to permit any trends to be discerned in the
actual or relative concentrations of radionuclides.
(b) The predominant radionuclides present in the samples of
filter sludges or "equivalent" vary, but are inclusive of Csl37f Cs ,
Co60, Co58, Co57, Fe55, and Mh54.
7.4 Recommendations
The following is recommended with regards treatment of the results
of this study and for future work.
(1) The radionuclide analyses and their interpretations re-
ported herein should be considered as preliminary indicators of trends
and should be used as a tool in establshing the parameters for a more
definitive program.
(2) In any future program, the sampling program must permit
collection of a sufficient number of samples having the same parameters
so as to be statistically reliable. To achieve this, samples similar in
waste form, duration of reactor operation, age since generation, and lo-
cation within the waste system should be obtained. Samples of sufficient
size must be taken to permit standard laboratory analyses to be made and
reported in consistent units. The radionuclide analyses should cover the
full spectrum of radionuclides present.
7-9
-------
8.0 Bibliography
1. Godbee, H.W., September 1973, Use of Evaporation for the Treatment
of Liquids in the Nucelar Indus-try, ORNL-4790.
2. Kibbey, A.H. and Godbee, H.W., March, 1974, A Critical Review of
Solid Radwaste Practices at Nuclear Power Plants, ORNL-4924.
3. Lin, K.H., December 1973, Use of Ion Exchange for the Treatment of
Liquids in Nuclear Power Plants, ORNL-4792.
4. Duckworth, J.P., et. al., September 1974, Low Level Radioactive
Waste Management Research Projects, Nuclear Fuel Services Inc.
5. Nine Mile Point Nuclear Station, Unit 1, June 1972, Niagara Mo-
hawk Power Corporation, U.S.A.E.G., Docket No. 50-220.
6. Duell, J. , 1976, Nine Mile Point Nuclear Station, Personal com-
munications .
7. James A. Fitzpatrick Final Environmental Statement, March 1973, U.S.A.E.G.
Docket No. 50-333.
8. James A. Fitzpatrick Final Safety Analysis Report, Volume 5, Docket No.
50-333.
9. DeMeritt, E.L., May 1971, Waste Control at Ginna Station, R.G. & E
Company, Presented at 69th National Meeting of AICHE.
10. Quinn, B., 1976, R.E. Ginna Station, Personal communications.
11. Indian Point Station, Unit No. 1, February 1976, System Description
No. 27, Liquid Disposal System, Revision No. 1.
12. Kelly, J., 1976, Director of Radiation Chemistry,Indian Point
Station, Personal communicatiors.
13, Effluent and Waste Disposal; Semiannual Report, No. 10, January
1975, July to Dec. 1974, Docket No. 50-244, (RE Ginna).
14. Cooley, C.R., and Lerch, R.E., May 1976, Nuclear Fuel Cycle and
Production Program Report, July to December 1975, HEDL-TME 76-22.
15. Hutchinson, J.A., 1976, Associate Radiochemist, Radiological
Safety Laboratory, N.Y.S.D.H., Personal communications.
16. Hutchinson, J.A., June, 1977, Associate Radiochemist,
Radiological Safety Laboratory, N.Y.S.D.H., Personal
communications.
3-1
-------
APPENDIX A
WASTE TREATMENT SYSTEMS AT REACTORS
FROM WHICH SAMPLES WERE COLLECTED
-------
APPENDIX A
A. Waste Treatment Systems at Reactors From Which Samples Were Collected
The following sections describe the liquid and solid radwaste systems
in use at the four commercial nuclear power plants at the time the samples
were collected. The participating facilities were the Nine Mile Point,
James A. Fitzpatrick, R.E. Ginna, and Indian Point No. 2 nuclear power
stations.
A.I Nine Mile Point (BWR)
A.1.1 Liquid Radwaste System
The liquid radwaste system at Nine Mile Point is subdivided into (1) the
waste collector subsystem, (2) the floor drain subsystem and (3) the regenerant
chemical subsystem. A diagram of the system is presented in Figure A-l.
The waste collector subsystem processes those potentially radioactive
liquid wastes which are characteristic of low conductivity. The wastes
collected by this subsystem includes liquid waste from the reactor cooling
system, the condensate system , the feedwater system, the reactor water
clean-up system, the condensate demineralizer regeneration system and waste
evaporator distillate. Any radioactive materials in these wastes are re-
moved by filtration and ion exchange. The processed liquids are either
reprocessed or sent to the condensate storage tank for in-plant reuse. The
filter sludge is processed by the solid radwaste system. The ion exchange
filters are regenerated and the regeneration solutions are processed by the
regenerant chemical subsystem.
The floor drain subsystem collects all potentially radioactive high
conductivity waste liquids from floor drains, laboratory drains, radwaste
building sumps and decontamination drains. The collected liquids are passed
through filters and then through demineralizers. The filtrate is either re-
covered or discharged while the sludge is processed by the solid radwaste
A-l
-------
>
REACTOR
CLEAN-UP SYSTEM
FILTERS (2) AND
DEMINERALIZERS (2)
WASTE COLLECTOR
LOW CONDUCTIVITY WASTE
EQUIPMENT DRAINS FROM
DRYWELL AND REACTOR,
RADWASTE AND TURBINE
BUILDING, CONDENSATE
DEMINERALIZER RINSE,
CONCENTRATOR DISTILLATE,
AND DRYWELL FLOOR SUMP.
FLOOR DRAIN
HIGH CONDUCTIVITY WASTE
FLOOR DRAINS FROM REACTOR,
TURBINE AND RADWASTE BUILDINGS.
REGEN6RANT
CHEMICAL WASTE ^__
RESIN REGENERATION CHEMICALS,
LABORATORY DRAINS, SAMPLE
DRAINS AND EQUIPMENT
DECONTAMINATION.
MISCELLANEOUS WASTE
LAUNDRY DRAINS
CASK CLEANING
PERSONNEL DECONTAMINATION
FLOOR DRAIN SAMPLE
TANKS 10,000 gal (2)
LIQUID EFFLUENT TO
RADWASTE BLDG.
FLOOR DRAIN.
WASTE CONCENTRATOR
12 gpm
It
*
CONCENTRATED WASTE
TANKS 5000 gal (21
T
SOLID RADIOACTIVE WASTE
SYSTEM (SRWS)
SPENT RESIN AND FILTER
SLUDGE TANKS, CENTRIFUGE
AND DRUMMING STATION
1
DRUMMED WASTE TO
OFF-SITE DISPOSAL
DISCHARGE 100%
1. SRWS DENOTES THE SOLID RADIOACTIVE WASTE SYSTEM.
2. UHC DENOTES THE ULTFIASONIC RESIN CLEANER.
268,000 gpm
LAKE ONTARIO
FIG. A-l UPGRADED LIQUID RADWASTE SYSTEM,
NINE MILE POINT NUCLEAR STATION, UNIT 1.
-------
system,
The regenerant chemical subsystem collects those chemical wastes
which result from the regeneration of the condensate demineralizers. These
wastes are collected, neutralized and sampled in the waste neutralizer tank.
From this tank the wastes are pumped to the waste evaporators, which are
of 12 and 20 gpm capacity, where they are processed. The distillate is
collected and is routed, eventually, to the waste collector subsystem.
The waste concentrate is pumped to the solid radioactive waste system.
A.1.2 Solid Radwaste System
The wastes handled by this system include (1) evaporator concentrates,
(2) filter sludges, (3) spent ion exchange resins, and (4) miscellaneous trash.
The evaporator concentrates are the solid wastes which remain from the
processing of those wastes collected in the waste neutralizer tank and pro-
cessed by the system's two waste concentrators.
The waste evaporator concentrates are routinely monitored in order to
determine when the normal operational limit of 3yCi/ml is reached. Upon
reaching the operational limit, the concentrate is pumped either to a con-
centrate waste tank from which it is subsequently pumped to the mixer or
directly to the mixer where it is mixed with urea formaldehyde under the
correct physio-chemical conditions. The mix is then pumped into a 150
cubic foot disposal Hittman liner for storage and subsequent transportation
, , . , (6)
and burial.
Filter sludges result from the filtration of those liquid wastes collected
in the waste collector subsystem and floordrain subsystem. The filters are
travelling belt-type filters which are designed to (1) reduce backwash water
and (2) permit utilization of ultrasonic resin drains to remove resin crud
A-3
-------
thus increasing the length of time between resin regeneration. In both
systems the filter is designed to discharge a damp solid crud which is then
handled by the solid waste system. This crud is incorporated with urea
formaldehyde and the mix is pumped into the shipping cask for storage,
transporation and burial.
Spent resins from the mixed bed demineralizef, are flushed directly
to a 165 cubic foot capacity spent resin tank for storage. After a suf-
ficient decay period has elapsed, or if more volume is required, the spent
resins are pumped directly to the disposable Hittman shipping cask where they
are dewatered prior to shipment. At the present time, solidification '"of
the spent resin is being considered. " * '
A. 2 James A. Fitzpatrick (BWR)
A.2.1 Liquid Radwaste System
The wastes collected by the liquid radwaste system at Fitzpatrick are
classified as high purity, low purity, chemical, detergent and sludge wastes.
A flow chart of the liquid radwaste system showing the steps in processing
each type of waste is provided in Figure A-2.
The high purity liquid wastes from the reactor coolant clean-up:system,
the residual heat removal system, waste and turbine buildings, are brought to the
waste collector tank (30,000 gallons) . The wastes are processed by fil-
tration and demineralization. After processing, the filtrate is analyzed to
determine whether the filtrate should be reused, reprocessed or discharged.
The filters, filter sludges, and demineralizers are processed by the
solid waste system.
Low purity liquid wastes, from the dry well, reactor, radwaste, arid
turbine building floor drains, are collected in a floor drain tank (8,500
gallon). These wastes are processed by filtration prior to transfer; to'one
A-4
-------
of the floor drain sample tanks (17,000 gallons each). In these tanks
the processed waste is sampled and subsequently analyzed. Based on the
results of the analysis performed, these wastes are either discharged to the
environment or subjected to additional processing in the chemical waste
system or the high purity waste system.
The chemical wastes, collected from condensate demineralizer re-
generation solutions, non-detergent decontamination and laboratory
drains, are collected, neutralized, and sampled in one of the waste
neutralizer tanks (17,000 gallons each). After sampling,these wastes
are pH adjusted (7.0 to 9.0) prior to transfer to one of the two 20-gpm
waste evaporators. The distillate frcm the evaporation process is sent
to the waste collector tank (high purity waste system). The concentrate
is either subject to further concentration in 0.8-gpm evaporator or sent
directly to one of the two concentrate waste tanks.
The detergent waste system collects laundry, personnel decon-
tamination and other detergent wash down wastes. These wastes are
filtered prior to discharge. If activites higher than expected occur, the
waste is transferred to the chemical waste system.
The waste sludge system is designed to collect waste filter, floor drain
filter, and fuel-pool filter backwash and sludges in a filter sludge
tank (11,000 gallons). The sludges are permitted to settle prior to
decanting to the low purity waste system. Once decanted the sludge is
transferred to the centrifuges for dewatering.. The backwash from the
reactor waste cleanup filter demineralizer precoat is collected in two
phase separator tanks. The backwash is permitted to settle. The sup-
ernate is decanted to the high purity waste system and the sludges to the
centrifuges before being sent to the solid waste system.
A-
-------
>
L
L.
WASTE
SU1CE TANK
5!, 000 col.
EQUIPMENT DRAINS, LEAXOFTS
FLOOR DRAINS, COOLING WATER LtAKS
ASTES
SOLUTIONS
SPENT HES NS
SPENT flLTEHt, BACKWASH ANO JLUOQt
FROM WASTE, FLOOR DHAIK AND
HJtL ^OOL FILTEN
PERSONNEL DECONTAMINATION,
5CTEHGENT
DCCONTAMIWATIOH SOLUTION
1
REACTOR WATER
DEMINERALI2ER
(POWOFK)
i r
WASTE
(1} 30,000 qoi
l_ J
T
{ FLOOR ORA;N
ETC. }TANK("M *. 500901
CONDENSATE
OFMINERALIZEP
obli)
WASTE
, . Of MINER ALIZfR
i cz^
1 WASTE '^ ^"1 , , WASTE I
, NEUTrtALiZER frJ FILTER ) " CONCENTRATORS f
CLEANUP PHASE
» SEPARATORS (2 )
20 qpm to
TO HIOH PUBITY
SYSTEM
.uuujgo'" ./-14-1-1 -1"-" Trt w*5TE
COLLECTOR
TANK
[ SPFNT RESIN
R H TANK
-In* I'U
i *" t » . i r w
« TANK
DfTdlGCNT
S I,OOO qol EACH
CENTRIFUGE J
, "" ^ <21 2°'f-
J 1 1
Miir
U.....-,-*. TO DRi.'U)
. . . >4urtAiN ULUR -
CONOENSATE
200,000 90!
WASTE
14,000 gol
[ FLOOR DRAIN
\ 17,000 gal «a
1
TURBINES s
L CONDENSER
90%
10%
^_^_J STRUCTURE
I ?5 gp<*i
EVAPORATOR
-
*J
1 INTAKE
,_j 5 T nuc i unr
/f"'
_ ALTERNATE
SLURRY
Fig. A-2 Flow Chart of the Liquid Waste System.
James A. Fitzpatrick Nuclear Fewer Station
-------
A.2.2 Solid Radwaste System <8)
The solid radwaste system at Fitzpatrick is divided into two sub-
sections. The first subsection is designed to handle dry solid wastes
(rags, paper, solid wastes, etc.) These wastes are compressed when pos-
sible in 55 gallon drums prior to transportation to a burial facicility.
The second subsection is designed to handle wet, solid wastes i.e.,
precoat materials, ion exchange resins and concentrate materials.
Precoat materials are discharged from filter-demineralizers into
one of the two phase separator tanks. After settling has occurred the
liquid is transferred to the waste collector tank for subsequent treat-
ment and reuse. Precoat filters from waste, floordrains and fuel pool
filters are discharged to the waste sludge tank. After permitting solids
to settle the liquid is pumped to the floordrain sample tank. When the
concentration of solids in the waste sludge tank reaches 1-5%, the con-
centrates are pumped to one of the two centrifuges and subsequently to the
radwaste building for solidification.
Spent resins from the radioactive waste and condensate demineralizers
are sluiced to a spent resin tank (3000 gallons) for storage prior to
being fed to one of the two centrifuges (20 gpm) . Spent resins are dis-
charged directly from the centrifuges to the waste solidification facility.
A. 3 R.E. Ginna (PWR)
A.3.1. Liquid Radwaste System^
All liquid wastes processed by this system whether collected by floor
drains, equipment drains, laboratory drains or personnel decontamination
drains are brought to the Waste Holdup Tank. A generalized schematic of
the liquid system is shown in Figure A-3.
A-7
-------
These collected liquids are then transferred to the evaporator feed
tank from which they are pumped into the evaporator. The contents of the
evaporator and the evaporator feed tank are circulated together and sampled
every 4 hours. This analysis is conducted to determine when the operational
limit of 2yCi/ml or 10% boric acid concentration is reached.* ' Once the
concentrate reaches either of these limits, it is pumped to the solid waste
system.
The distillate from the evaporation process is pumped first to the
distillate tank and then to a waste condensate tank where it is analyzed
and its release rate calculated.
A.3.2 Solid Radwaste System( *
The solid radioactive waste generated at R.E. Ginna is composed
primarily of evaporator concentrates and spent ion exchange resins.
The evaporator concentrates are pumped from the evaporator feed
tank to the drumming station where the concentrate is mixed into verm-
iculite-cement mixture in 55-gallon drums. These drums are then moved to
the drum storage area to await transportation to the disposal site.
The majority of the primary coolant system demineralizers are not
designed to be regenerated. Under routine operating conditions, the
spent resins are replaced by flushing and new resins installed.
The flushed resin is transferred to the spent resin storage tank
where it remains until sufficient decay has occurred or more storage room
is required. The flushed resin is then pumped to the drumming station
where it is dewatered and placed in a 100 .cubic foot Atcor shipping cask.
Any regenerant solutions, from the regeneration of the polishing
demineralizers, are pumped to the waste holdup tank and then processed
by the waste evaporator.
A-8
-------
A. 4 Indian Point 2 (PWR)
A.4.1 Liquid Radwsste System '^
The liquid radioactive waste processing system at Indian Point-1 was
being used to handle the liquid radwaste produced by Indian Point-2 at the
time the samples were collected. The liquid radwaste handling system is
designed to collect, treat, process and store all potentially radioactive
liquid wastes generated on-site.
The collection center for these liquids consists of four waste col-
lection tanks. The collected waste is subsequently transferred to the
waste gas stripper. The removed waste gases are vented to the waste gas
condenser and then processed by the gaseous waste system. The stripped
liquid waste is pumped to the waste evaporator by means of an evaporator
feed pump system.
The distillate from the evaporation process is passed through a pol-
ishing waste demineralizer and collected in the waste distillate storage
tank. The collected distillate is sampled and, depending on the activity
levels, Is either transferred to the clear water storage tank or dis-
charged to the environment. The concentrate is pumped to a sludge storage
tank where it is held until transferred to the solidification processing
facility.
The liquid waste handling facility at Indian Point Station is currently
being improved. In the improved system the waste, after initial waste gas
stripping, will be passed through filters into a feed pre-heater. From
the pre-heater; the waste will be passed through a second gas stripper.
After gas-stripping, the waste will be processed by two larger capacity
evaporators. The concentrate will be pumped directly to the solidification
station. The distillate will be passed through an absorption tower and
A-9
-------
LAUNDRY &
SHOWER
TANKS
-x-
REVERSE
OSMOSIS
UNIT
RADIATION
MONITOR
Y
CONDENSER WATER
CANAL
RADIO
CHEMISTRY
LAB DRAIN
TANK.
CONTROLLED AREA
EQUIPMENT
& FLOOR
DRAINS
WASTE HOLD UP TANK
EVAPORATOR
DEMINERALIZER
WASTE
CDNDENSATE
TANKS
DRUMMING
STATION
FIGURE A-3
LIQUID WASTE SYSTEM AT R.E. CHINA
A-10
-------
a distillate cooler and then to two large volume distillate tanks. Oper-
ation of this new system has been initiated with the exception of the
distillate storage tanks. ^12^
The chemical and volume control system at Indian Point 2 is functional
and is designed to handle and process reactor coolant letdown water.
The coolant letdown water passes through both regenerative and non-
regenerative heat exchangers and a mixed bed coolant filter before storage
in a volume control tank. From this control tank the coolant water is
either pumped directly into the reactor coolant system or indirectly,
by injection, into the seals of the reactor coolant pumps.
Liquid effluents from the reactor coolant system, containing boric
acid, are collected in hold-up tanks for the purpose of recovering boric
acid and reactor make-up water. Liquid from the hold-up tanks is passed
through the evaporator feed ion exchanger, and the ion exchange filter
before entering the waste gas stripper. The effluent from the stripper
is transferred to the boric acid evaporator where the dilute boric acid
is concentrated. The gases from the evaporator are condensed and cooled,
passed through an evaporator condensate demineralizer and filter, and
collected in a monitor tank. From this tank, the condensate is pumped to
the primary water storage tank. The evaporator concentrates are discharged
through a concentrate filter and into a concentrate holding tank before
transfer to a boric acid tank.
A, 4.2 Solid Radwaste System
The predominant portion of the waste handled by this system reaults
from the treatment and processing of liquid, radioactive wastes. These
wastes are essentially evaporator concentrates, spent resins, filter
sludges and filters.
A-ll
-------
Evaporator concentrates from the liquid radwaste systems are col-
lected and mixed with urea formaldeyde and a catalyst in either 35 or 55
gallon drums. The future use of a 1500 gallon cask is under consideration.
There are no regenerable resins in the liquid radwaste system or the
chemical and volume control system, except for the boric acid evaporator
condensate demineralizer. All resins are sluiced to a spent resin stor-
age tank, where they are held until sufficient decay has occurred. They
are then pumped to the solidification facility where they are mixed with
urea formaldehyde and a catalyst in a 200 ft3 Atco cask.
Filters and filter sludges are handled in a manner similar to the way
in which the concentrates are handled. These filters and sludges are
(1) reactor coolant filters, (2) seal water filters, (3) seal injection
filters, (4) spent fuel pool filters, (5) ion exchange filters, and (6)
boric acid filters.
A-12
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APPENDIX B
ANALYTICAL METHODS USED BY THE RADIOLOGICAL
SCIENCE LABORATORY (RSL)
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APPENDIX B
B. Analytical Tfethods Used By the Radiological Science Laboratory (RSL) (]6)
B.I Sanple Preparation
Measurements of evaporator concentrate samples were performed on a
dilution of each sample. Most of the other samples were fused with NaOH
and the melt dissolved in distilled water. Two sludge samples were dis-
solved in acid, but portions of these samples were fused with NaOH for the
C14 and * measurements.
B.2 Gross Alpha/Gross Beta Analysis
An aliquot of a water sample or fusion extract was evaporated and the
residue quantitatively transferred to a planchet. The sample planchets
were counted on a gas flow proportional counter. Sample planchets which
required only gross-beta analysis were covered with saran wrap and counted
on the alpha/beta plateau. Sample planchets which required both gross-
beta and gross-alpha analyses were left uncovered and counte d first on
the alpha/beta plateau then on the alpha plateau.
The method is only adequate for screening purposes. Loss of volatile
radionuclides, such as radioiodine and tritium, is one problem. Another
drawback is the difficulty in radiometric standardization for a mixture
of unknown alpha and beta emitters.
The radionuclides used as standards in the Radiological Science
Laboratory are:
a. For gross beta - Sr90-Y90
b. For gross alpha - natural viranium
B-l
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B.3 Gross Alpha - Spectroiretric
A small aliquot of the liquid sample was evaporated to dryness and
digested with nitric acid. After electrcdeposition from an ammonium suplhate
solution onto a stainless steel disc, the radioactivity was measured with a
silicon surface-barrier detector, using only the counting efficiency to
calculate the activity on the disc at each energy region. The plating
efficiency for specific radionuclides was not known, so the method served
only to qualitatively identify a-emitters present in the sample.
B.4 Isotopic-Uranium Analysis
U2^2 Was added as a tracer to determine chemical and electrodeposition
recovery. Water samples and fusion extracts were evaporated to dryness
then taken up in 7.2N HNO^. Some sludge samples were leached with aqua
regia extracting plutonium, uranium, americium, cerium and iron.
Plutonium was collected from the leach solution by a batch ion exchange
process, leaving uranium and iron in the leach solution, which was then
evaporated to dryness. The residue from the pre-treatment and evaporation
of the sample was dissolved in 7.2 N KNO^ .
Uranium and any remaining plutonium were oxidized to the (IV) valence
state with sodium nitrite. The plutonium nitrate complex formed in the
strong nitric acid solution was removed on an anion exchange column.
The effluent was evaporated to dryness, taken up in 9 N HC1 and the
uranium chloride complex adsorbed on an anion exchange column. Iron
was removed from the column with a solution of 9 N HC1 - 0.25 M NH4I.
The uranium was then eluted with 1.2 N HC1 and electroplated onto a
stainless steel disc from an ammonium sulphate solution.
The electroplated disc samples were counted on an alpha spec-
trometry system using a 450 mm2 silicon surface barrier detector. The
B-2
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system amplifier was biased to cover an energy range of about 4 Nfev to 6
Mev.
The net cpm in each region were calculated and the values corrected
for interference from higher energy alpha peaks, if necessary. The U235
and U23Activity levels were then calculated by applying the appropriate
chemical recovery and counting efficiency factors.
B.5 Isotopic-Plutonium Analysis
Pu242 was added as a tracer to determine chemical and electrodeposition
recovery. Water samples and fusion extracts were evaporated to dryness,
then taken up in the 7.2 N HNCX. Some sludge samples were leached with
aqua regia, plutonium collected from the leach solution by a batch ion
exchange process, then eluted and the eluate evaporated to dryness.
The residue from the pre-treatment of the sample was dissolved in
7.2 N HNC>3. Plutonium was oxidized to the (IV) valence state with sodium
nitrite and the plutonium nitrate complex formed in the strong nitric acid
solution was absorbed on an anion-exchange column. The column was washed
with HISD-j and HC1 solutions, then the plutonium was eluted with a 0.36
N HCL - 0.01 N HF solution. Plutonium was electrodeposited from an am-
monium sulphate solution onto a stainless steel disc.
The electroplated disc samples were counted on an alpha-spectrometry
system using a 450 mm2 silicon surface-barrier detector. The system
amplifier was biased to cover an energy range of about 4 Mev to 6 Mev.
The net cpm in each region were calculated and the values corrected
for interference from higher energy alpha peaks, if necessary. The P
and pu239'240 activity were then calculated by applying the ap-
propriate chemical recovery and counting efficiency factors.
B-3
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B.6 Am241 Analysis
was 3^^ 33 a tracer to determine chemical and electro-
deposition recovery. Water samples and fusion extracts were evaporated
to dryness, then taken up in 7.2 N HNO.,. Some sludge samples were
leached with aqua regia extracting plutonium, uranium, americium, curium,
and iron. Plutonium was collected from solution by a batch ion exchange
process leaving uranium, americium, curium and iron in the eluent which
was then evaporated to dryness. The residue from the pre-treatment of
the sample was dissolved in 7.2 N HNC>3.
Uranium and any remaining plutonium were oxidized to the (IV) valence
state with sodium nitrite. The plutonium nitrate complex formed in the
strong nitric acid solution was removed on an anion exchange column. The
effluent was evaporated to dryness, taken up in 9 N HC1 and the uranium
chloride complex adsorbed on an anion exchange column. The effluent was
collected for separation of americium. Iron was removed from the column
with a solution of 9 N HC1 - 0.25 M NH4l, and the solution was combined
with the effluent just previous to be used for the americium separation.
The combined solution was evaporated to dryness, oxidize iodine with con-
centrate HN03, and the residue dissolved in 0.5 N HC1.
Americium was separated from the solution on a cation exchange
resin, Dcwex 50 x 8 (H+) . The column was washed with 0.5N HC1 and the
americium eluted with 12 N HC1. The eluent was taken to dryness and
americium was electroplated from anmonium sulphate solution onto a
stainless steel disc.
The electroplated disc samples were counted on an alpha-spectro-
metry system using a 450 nm2 silicon surface-barrier detector. The system
amplifier was biased to cover an energy range of about" 4 Mev to 6 Mev.
B-4
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The net cpm in each region was calculated. The Am24-Activity was
then calculated applying the appropriate chemical recovery and counting
efficiency factors.
B.7 Analysis of Tritium as HTO
Samples were vacuum distilled and the distillate collected to sep-
arate tritium from other interfering radionuclides and to remove chemical
and/or physical quenching agents. An aliquot of the distillate was mixed
with an organic scintillator and counted in a liquid scintillation spec-
trometer. Water known to be of low tritium content was used as a back-
ground sample.
The degree of quenching in a sample was determined by external stan-
dardization. The quench factor obtained was used to determine the counting
efficiency for calculation of the tritium activity in the sample. Analysis
of a 10 ml aliquot of the distillate resulted in a sensitivity of approx-
imately 500 pCi/1.
B.8 Isotopic Gamma Analysis Ge(Li)
The liquid or solid sample in a standardized geometry, was analyzed
with a Ge(Li) detector system. The system utilized a 4096-channel anal-
yzer with an energy calibration of 0.5 keV/channel.
The activity of each gamma-emitting radionuclide in the sample was
determined by using the efficiency factor for the photopeak of the isotope.
The efficiency was obtained from a gamma-ray efficiency curve, prepared
by measuring selected standards , in the standardized geometry, and using-
their known gamma ray intensities to determine photon efficiencies.
B.9 Sr90 Analysis
Sr85 tracer and stable strontium were added to the sample. The Sry-"
tracer was used to radiometrically determine the chemical recovery of
B-5
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strontium, while the stable strontium acted as a carrier. Water samples
and fusion extracts were acid digested and strontium precipitated as the
carbonate. Some sludge samples were dried and strontium removed by
leaching twice with 6 N HN03. The leach solutions were evaporated to
dryness and the residue taken up in HC1. Iron was removed on an anion
exchange column and strontium in the effluent was precipitated as the
oxalate then converted to the oxide.
The carbonate or the oxide from the sample pretreatment was dissolved
in nitric acid. The rare earths, ruthenium and any remaining calcium was
removed by precipitation of strontium nitrate from concentrated HN03. Yttrium
carrier was added and the sample set aside 10-14 days for Y"° ingrowth.
At the end of the ingrowth period, yttrium was precipitated as the
hydroxide, purified by repeated extractions into TBP and back-extractions
into water. Yttrium was collected as the hydroxide, reprecipitated as the
oxalate, converted to the oxide and mounted in a filter paper disc. The
yttrium recovery was determined gravimetrically. The yttrium oxide was
mounted on a nylon planchet and counted in an end-window, gas-flew propor-
tional counter.
The chemical recovery for strontium was determined by gamma counting
the Sr 8-> tracer on a Nal detector.
Three or more measurements, beginning immediately after the chemical
separation of yttrium from strontium and continuing at approximately 2-day
intervals, were made on the y90 fraction in order to follow its decay.
A computer program, using the half-life of Y90as a known, performed a
least-squares-fit to the counting data to calculate the Sf ^ activity.
B-6
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B. 10 Radioiodine Analysis
Stable iodine carrier was added to the sample to determine chemical
recovery. Samples were treated to convert all iodine in the sample to a
common oxidation state prior to chemical separation and purification.
Water samples were taken through an oxidation-reduction step using
hydroxylamine hydrochloride and sodium bisulfite to convert all iodine
to iodide suitable for processing through an anion exchange solumn.
Sludges were fused with a NaOH-Na2CO3 mixture. The melt was cooled,
dissolved in distilled water and sodium hypochlorite added to oxidize the
iodine to iodate. Hydroxylamine hydrochloride then reduced the icdate
to elemental iodine for CC1. extraction.
After samples had been treated to convert all iodine in the sample
to a coirmon oxidation state, the iodine was isolated by solvent ex-
traction or a combination of ion exchange and solvent extraction steps.
Iodine, as the iodide, was concentrated by adsorption on an anion
exchange column. Following a NaCl wash, the iodine was eluted with
sodium hypochlorite. Iodine, as iodate, was reduced to elemental iodine
for extraction as palladium iodine.
Chemical recovery of the added carrier was determined gravemetrically.
The PdlT precipitate was counted on an intrinsic-germanium detector
and the intensities of the Ka X-rays from Te125and Xe129 measured.
The decay of I-1-3-1-also results in the production of xenon X-rays.
Consequently 1131 constituted an interference in the procedure. Prior
to the X-ray measurement, all samples were counted for 100 minutes on a
Nal well-detector to check for the presence of I131- A second measurement
on the intrinsic diode after two weeks decay provided further verification
of I131. If I131was present, the X-ray data was corrected for I131 in-
terference or the sample allowed to decay until the jJ-31 activity no
B-7
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1 9 Q
longer seriously interfered with the I ^measurement.
"IOC l o g
The germanium detector is standardized for both I 3 and I as a
function of weight of the PdI2 precipitate. The Ka X-rays at 27.5 and
27.2 keV for Te125and 29.7 and 29.4 key forXe129 are used to quantitate
the data. The matrix coefficients to correct for the interference of one
spectral region to the other are also determined from the standard spectra
for I12^nd I129.Correction factors for I^3^-interference are determined
from 12Pd131 standards.
The counts in the I region and the I-*-29 region were summed separately.
The net counting rate in each region was computed. A matrix calculation
1 -, r I 90
was used to correct the I-1-"net counting rate and the I net counting
rate for mutual interference from Ccmpton interactions and I -*- ^ -knterference.
The appropriate decay/volume, counting efficiency and chemical recovery
19^ 129
corrections were then applied to compute the I-L/-3and I activities.
QQ
B.ll Te Analysis
Technetium was separated by solvent extraction with nitrobenzene.
Stable rhenium was added to the sample to determine the chemical recovery .
The rhenium was oxidized to the perrhenate and technetium was oxidized to
the pertechnetate. An extraction was performed from dilute nitric acid
into nitrobenzene, using tetraphenyarsonium chloride as the extracting
agent. The pertechnetate and perrhenate were then back-extracted into
concentrated nitric acid. Tetraphenylarsonium-pertechnetate and per-
rhenate were then reprecipitated. The precipitate was filtered and the
rhenium recovery is determined gravimetrically. Te was counted in
an end-window gas-flow-proportional counter.
B.12 C14 Analysis
Sludge and resin samples were first fused with NaCH and the resulting
melt dissolved in distilled water. Water samples were analyzed directly.
B-8
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The extraction of CO and CH^ was carried out in a closed vacuum system.
A sample volume of 50-100 ml was spiked with 0.1 g of sodium carbonate,
introduced into the vacuum system, and 50 ml of concentrated hydrochloric
acid was added under vacuum. The sample was constantly purged with He
containing a total of 25 ml (STP) of methane carrier gas. The evolved
CO2 and the stripped methane were then collected and separated cryo-
genically after removal of the water vapor in a series of cold traps.
Subsequently, the gases were purified in a gas chromatograph and the
extraction yield determined volumetrically. The purified gas was
loaded into an internal gas-proportional counter and diluted in the
counter with P-10 counting gas. Spectral analysis was performed by
pulse-height analysis under controlled conditions in a massive iron shield,
where the counting tube was operated inside an anticoincidence guard
counter.
B. 13 Fe55 Analysis
Stable iron was added as a carrier to determine chemical recovery.
Water samples, fusion extracts, and acid leachates were evaporated to dry-
ness and the residue dissolved in a 50% acetone-water solution. The sample
was then passed through a chromatographic column containing AG50W-X8 cation-
exchange resin which had been equilibrated with 50% acetone-water sol-
ution. The iron (III) was eluted with 80% acetone-0.5 M HC1 solution.
Iron was electrodeposited from a NH H2PO4-(NH4)2CO3 solution onto a
polished copper disc, and the 5.9 keV X-ray was then measured with an
instrinsic-germanium detector.
B. 14 Ni^ Analysis
Nickel was isolated from water samples, fusion extracts and acid
leachates by forming nickel dimethylgloximate which was extracted into
B-9
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chloroform. Nickel carrier, measured spectrophotometrically, was used to
determine the chemical recovery. The nickel dimethylgloximate was de-
colorized with hydrochloric acid and the 67-keV beta of Ni63counted on
a liquid scintillation spectrometer.
B.15 Detection Limits
The detection limits varied for each sample measurement inasmuch as
these limits are a function of the quantity of sample used, counting time,
and processing recovery, which varied. The detection limits of radio-
nuclides measured by isotopic gamma analyses also vary with the gamma
composition of the sample. The deviations on the measured samples ranged
from + 5% to greater than +_ 80%, without any consistent pattern for in-
dividual radionuclides.
*U.S. GOVERNMENT PRINTING OFFICE:1978 260-880/1 i-3
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