-------
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. Ihe 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 (13'14> and from the plant operators records.* '
3he 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
-------
No. 2 ^ ' and that reported for an unnamed PWR l ' 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 radioiretric 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 Cs showed a variation
greater than 30 fold during this period, while the concentration of
6-2
-------
TABLE 6-1
Radionuclide Analysis of PWR Spent Resin Samples
Measured Under Other Analytic Programs
Radionuclide Reactor
RE Ginna(13) Indian Pt.2<12>
Relative Relative
Concentration Proportion* Activity Proportion*
(UCi/ml) (%) (Ci) (%)
Cs137 28-23 30.2 5.77 72.0
Cs134 7.38 7.9 1.31 16.0
Sb125 N.R. N.R.
Co60 37.29 39.9 0.91 11.0
Co58 N.R. N.R.
Mn54 20.47 21.9 0.03 1.0
Unidentified
Concentration P
(yci/ml)
14.32
3.89
0.137
5.23
3.51
2.97
PWR<14)
Relative
'roportio
(%)
47.6
12.9
0.5
17,4
11.7
9.9
*of only those radionuclides analyzed.
-------
TABLE 6-2
Radionuclide Analysis of PWR Evaporator Concentrate Samples
Measured Under Other Analytic Programs
Radionuc1ide
RE Ginna<13)
Concentr at ion
(yCi/ml)
Cs137 0.542
Cs134 0.272
I131 0.013
Co60 0.045
Co58 N.R.
Mn54 0.021
Relatxve
Proportion*
60.4
30.3
1.3
5.0
2.2
Indian Pt.
Concentration**
(yCi/ml)
0.718
0.532
N.R.
0.020
0.388
0.014
2 (12)
Relative
Proportion*
43.3
28.5
2.6
22.2
1.4
Unidentified
Concentration
(yci/tal )
0.000532
N.R.
N.R.
0.005S9
0.000523
0.000514
PWR(14>
Relative
Proportion*
7.0
.__
...
79.3
6.9
6.8
*of only those radionuclides analyzed.
**average concentration July-rDecember 1975.
NR = Not Reported
-------
TABLE 6-3
Radiorvuclide Analysis of Evaporator Concentrate Samples From Indian Point No. 2 Reactor
July, September, October, December, 1975
Radionuclide Month
July
Cone en- Relative
tration Proportion*
(UCi/ml) (%)
Cs137 2.350
Cs134 1.810
Co60 0.017
Co58 0.019
Mn5^ 0.020
55.5
42.8
0.4
0.5
0.5
September
Concen- Relative
tration Proportion*
(yci/ml) {%)
0.169
0.113
0.017
0.113
0.007
40.4
27.0
4.1
27.0
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
(VlCi/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
pattern of variation. For the same two radionuclides, Cs134 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 (12) and from the unnamed PWR (14) 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
Radionuclide Analysis of PWR Filter Sludge Samples Measu
Under Other Analytic Programs
Radionuclide
Indian Pt.2(12)
RCS
Cone en- Relative
tration Proportion
(uCi/gm)x!03 (%)
Sb125 N.R.
Cd115m N.R.
CdU3in N.R.
Ag110m N.R.
Zr95 N.R.
7_65 N p
£iii n*x\«
Co60 1.89 15.1
Co58 3.65 29.2
Co^7 -N.R.
Mn54 0.362 2.9
Cr51 6.60 52.8
SFP
Concen-
tration
(uCi/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
Unident
red
ified PWR(14)
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
M &
it • I\*
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 = Haste Holding Tank
NR = Not Reported
-------
7.0 Comparisons, Interpretations, and Reccnmsndations
This study provides a preliminary data base on the radionuclide
composition, and actual and relative concentrations to be found in the
oredoninant waste forms generated by the four UWRs, which are processed, nackaged
and shipped to commercial 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 Conposition of Waste Sartples
The composition and relative radionuclide concentration in the
samples of waste generated at LHPs 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 -jnat 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
Cs137, Cs134, Co60 and Mn54.
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 gamma 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
' 1 "^7
CsXJ Concentration
Relative Prop.
Cs
134
Co
60
Mn
54
Sample of
9/75
0.169uCi/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
Sample of
2/76
0.542yCi/ml 0.102yCi/gm
61.6%
0.272
30.1
0.045
5.1
0.021
2.3
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
-------
TflBUE 7-2
Comparison of Concentrations of Gamma Emitting Radionuclides
(Tsj>300 days) in Samples of Evaporator Concentrate From PWRs 6 BWRs
Radionuclide
£S137 concentration
Relative Prop.
PWR
BWR
CS
134
Indian Point No.2 R.E.. Ginna
0.
64.2%
0.037
23.2
54.0%
0.190
34.2
Mine Mile Point J.A. Fitzpatrick
0.229pCi/gm
44.3%
0.169
32.7
0.0004yCi/ml
1.9%
0.0001
0.5
Co1
60
0.035
6.3
0.019
11.9
0.096
18.6
0.0089
42.0
0.031
5.6
0.001
0.6
0.023
4.4
0.0118
55.7
Total Concentration in
Selected Sample
0.556
0.159
0.517
0.0212
7-5
-------
TABLE 7-3
Comparison of Concentrations of Gamma Emitting Radionuclides
days) in Samples of Spent Ion Exchange Resins From PWRs & BWRs
Radionuclide
R.E. Ginna
Indian Point No.2
Unidentified PWR
Nine Mile Point
o\
Cs^-37 Concentration
Relative Prop.
Cs*34
Co^O
Mn*4
*
28,23UCi/ml
30.2%
7.38
7.9
37.29
39.9
20.47
21.9
**
21.9nci/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
•**
31.7viCi/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 conparisons 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 itiore
supportable conclusions to be drawn with regard to 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 PWRs (where data was
available) essentially of the same order as a function of time. This
may iitply 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 fron all the reactors, with the exception of
the samples frcm Fitzpatrick, are Cs137 , Cs134, Co60, and Co58, generally
in that order. This agrees with the information provided in the lit-
erature (Itef. 4). Oh the basis of the half lives of the gairma emitters,
Cs137, Cs134, Co60, and Mi54 will generally be predominant in the buried
waste. Furthermore, Fe55, Ni6 , 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. TMs 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
. -. t
(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 samples.
7-8
-------
(b) The predominant radionuclides present in spent icn exchange
resin sarrples fron all of the reactors are Cs^- , Ccfi®f and Mn^4, which
occur in varying proportions in each sanple. Since these radionuclides
are all relatively long lived (T*5>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 Cs137, Cs134,
Co60, Co58, Co57, Fe55, and Mn54.
7.4 Recommendations
Ihe 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 establishing 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 Industry, QRNL-4790.
2. Kibbey, A.H. and Godbee, H.W., March, 1974, A Critical Review of
Solid Radwaste Practices at Nuclear Power Plants, OFNL-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, Lew 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.C., Docket No. 50-220.
6. Duell, J., 1976, Nine Mile Point Nuclear Station, Personal com-
munications.
7. Janes A. Fitzpatrick Final Environmental Statement, March 1973, U.S.A.E.G.,
Docket No. 50-333.
8. James A. Fitzpatrick Final Safety Analysis Report, Volute 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 ccmrunications.
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 conncmications.
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 197^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 comnunications.
16. Hutchinson, J.A., June, 1977, Associate Radiochemist,
Radiological Safety Laboratory, N.Y.S.D.H., Personal
communications.
8-1
-------
APPENDIX A
WASTE TREATMENT SYSTEMS AT REACTORS
FROM WHICH SAMPUES WERE CX3LEJBCTED
-------
APPENDIX A
A. Waste Treatment Systems at Reactors Fran Which Samples Were Collected
The following sections describe the liquid and solid radwaste systems
in use at the four comercial nuclear power plants at the time the samples
were collected. Ihe 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 Padwaste System
Ihe 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. Ihe 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. Ihe ion exchange
filters are regenerated and the regeneration solutions are processed by the
regenerant chemical subsystem.
Ihe 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 dernineralizers. Ihe 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)i
WASTE COLLECTOR
LOW CONDUCTIVITY WASTE
EQUIPMENT DRAINS FROM
DRYWELL AND REACTOR,
RADWASTE AND TURBINE
BUILDING. CONDENSATE
DEMtNERALIZER RINSE,
CONCENTRATOR DISTILLATE,
AND DRYWELL FLOOR SUMP.
FLOOR DRAIN
HIGH CONDUCTIVITY WASTE
FLOOR DRAINS FROM REACTOR.
TURBINE AND RADWASTE BUILDINGS.
REOENERANT
CHEMICAL WASTE
RESIN REGENERATION CHEMICALS,
LABORATORY DRAINS, SAMPLE
DRAINS AND EQUIPMENT
DECONTAMINATION.
MISCELLANEOUS WASTE
REGENERANTS
+ RINSE
REGENERATION
STATION + URC
DEMINERALIZERS
I CONDENSATE
[STORAGE TANK
WASTE
DEMINERALIZER
WASTE SAMPLE
TANKS 25,000 gal (2)
RADIATION
MONITOR
FLOOR DRAIN SAMPLE
TANKS 10,000 gal 12)
LIQUID EFFLUENT TO
RADWASTE BLDG.
FLOOR DRAIN.
WASTE CONCENTRATOR
20 gpm
WASTE CONCENTRATOR
12 9pm
SOLID RADIOACTIVE WASTE
SYSTEM (SRWS)
SPENT RESIN AND FILTER
SLUDGE TANKS, CENTRIFUGE
AND DRUVMING STATION
CONCENTRATED WASTE
TANKS 5000 gsl <2>
DRUMMED WASTE TO
OFF-SITE DISPOSAL
DISCHARGE 100%
LAKE ONTARIO
LAUNDRY DRAINS
CASK CLEANING
PERSONNEL DECONTAMINATION
NOTE:
1. SHWS DENOTES THE SOLID RADIOACTIVE WASTE SYSTEM,
2. URC DENOTES THE ULTHASONiC RESIN CLEANER.
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 sairpled in the waste neutralizes tank.
From this tank the wastes are punped 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.
Ihe 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 3pCi/ml is reached. Upon
reaching the operational limit, the concentrate is punped either to a con-
centrate waste tank from which it is subsequently punped 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
and burial.(6>
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 demineralizer, 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 tiite, 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, and
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 sanple 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.
She 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. She distillate fran the evaporation process is sent
to ibhe waste collector tank (high purity waste system). The concentrate
is either subject to further concentration in 0.8-gpm evaporator or sent
directly to oneof the two concentrate waste tanks.
The detergent waste .system collects laundry, personnel decon-
tanination 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. Chce 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-5
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REACTOR
t
1
REACTOR WATCR
OEMINERALUER
(POWDFX)
U_
wAsft
SURGE TANK i
ss.ooc c.i. i ,
DISH PURITT WAlTfj i
EQUIPMENT DRAINS, LtAKOFFS
PHASE SEPARATOR OCCAMTS
CCNTSIfuGC EFFLUENTS
LOW PURITY WASTES - •
FLOOR OKAINS.COOLINO WATER LEAKS. ETC.
WASTE SLUDGE DECANT
CHEMICAL WASTES — » •
DCUIhERALIZER REGENEOANT.LAi DRAINS,
NOkOETERGENT DECONTAMINATION
SOLUTIONS
IACHWASH FROM CLEANUP FILTER- •
OEMINERALIKR
CONDENSATE AND WASTE DEMINERALKER -•
SPENT RESINS
SPENT FILTERS, SACK WASH AND SLUOW — *
FROM WASTE, FLOOR DRAIN AND
(Ull fOOL IILTEN
3ETEROENT DECONTAMINATION SOLUTIONS
LAUNDRY WASTE
WASTE
COLLECTOR TAHK
(1) 30.OOO •„.
I
CVAPODATOR
O.ltt"
TE
IALIZFR
200,000 (•!.
*»H.
WASTE
«> 14,000 5
-------
A.2.2 Solid Kadwaste System ^8)
The solid radwaste system at Fitzpatridc 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 gpra). 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*9) ,
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 fron which they are purped 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.^10' Once the
concentrate reaches either of these limits, it is pumped to the solid waste
system.
The distillate from the evaporation process is purped 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*9^
The solid radioactive waste generated at R.E. Ginna is composed
primarily of evaporator concentrates and spent ion exchange resins.
The evaporator concentrates are puttped 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 demineral i zers 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 dnnming 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
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A.4 Indian Point 2 (PWR)
A.4.1 Liquid Radweste 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. Prom
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
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LAUNDRY &
SHOWER
TANKS
REVERSE
OSMOSIS
UNIT
C)
RADIATION
MONITOR
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. GIMNA
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
("\2\
distillate storage tanks.l '
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. Fran this tank, the condensate is punped 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. Tine 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. Ihey
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, ttiese 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 f6)
boric acid filters.
A-12
-------
APPENDIX B
ANALYTICAL METHODS USED BY THE RADIOLOGICAL
SCIENCE LABORATORY (RSL)
-------
APPENDIX B
B. Analytical Methods Used By the Radiological Science Laboratory (RSL) f:16'
B.I Sample Preparation
Measurements of evaporator concentrate sairples were performed on a
dilution of each sample. Most of the other sanples 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 NaCH for the
C14 and l129measurements.
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 counted 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 uranium
B-l
-------
B.3 Gross Alpha - Spectrometric
A small aliquot of the liquid sample was evaporated to dryness and
digested with nitric acid. After electrodeposition 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
2 ^2
U"* 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 HN03. 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 HN03 .
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 mm? silicon surface barrier detector. The
B-2
-------
system amplifier was biased to cover an energy range of about 4 l*fev to 6
Mev.
the net cpm in each region were calculated and the values corrected
for interference from higher energy alpha peaks, if necessary. Ohe 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 HN03. 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 HN03 and HC1 solutions, then the plutonium was eluted with a 0.36
N HCL - 0.01 N HP solution. Plutonium was electrodeposited from an am-
monium sulphate solution onto a stainless steel disc.
The electroplated disc samples were counted on an alpha-spectronetry
system using a 450 mm2 silicon surface-barrier detector. Ohe 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. Ihe Pu238
and Pu239'240 activity were then calculated by applying the ap-
propriate chemical recovery and counting efficiency factors.
B-3
-------
B.6 Am241 Analysis
was added as 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^- Sate sludge saitples 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 HN03.
Uranium and any remaining plutonium were oxidized to the (IV) valence
state with sodium nitrite. The plutonium nitrate corplex formed in the
strong nitric acid solution was removed on an anicn exchange column. The
effluent was evaporated to dryness, taken up in 9 N HC1 and the uranium
chloride complex adsorbed on an anlon 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 NH4I, and the solution was ccnbined
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.f Dowex 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 ammonium sulphate solution onto a
stainless steel disc.
The electroplated disc samples were counted on an alpha-spectro-
metry 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.
B-4
-------
The net cpm in each region was calculated. The ?\m^4-^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 sattple was
determined by using the efficiency factor for the photopeak of the isotope.
The efficiency was obtained frcm a gamma-ray efficiency curve, prepared
by measuring selected standards , in the standardized geometry, and usinp-
their known gamma ray intensities to determine photon efficiencies.
B.9 Sr90 Analysis
p (-
Sr85 tracer and stable strontium were added to the sample. The Sr
tracer was used to radiometrically determine the chemical recovery of
B-5
-------
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 sanples were dried and strontium removed by
leaching twice with 6 N HNO^. 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 Y90 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-flow 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 iimediately 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 Y9^as a known, performed a
least-squares-fit to the counting data to calculate the Sf ?® activity.
B-6
-------
B. 10 Radioiodine Analysis
Stable iodine carrier was added to the sairple to determine chemical
recovery. Sanples were treated to convert all iodine in the sample to a
caiman 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 solunn.
Sludges were fused with a NaOH-Na2°°3 mixture. The melt was cooled,
dissolved in distilled water and sodium hypochlorite added to oxidize the
iodine to iodate. Hydroxylamine hydrochloride then reduced the iodate
to elemental iodine for OC1. extraction.
After sanples had been treated to convert all iodine in the sample
to a common oxidation state, the iodine was isolated by solvent ex-
traction or a combination of ion exchange and solvent extraction steps.
Iodine, as the iodi.de, 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 determijied gravemetrically.
The Pdl2 precipitate was counted on an intrinsic-germanium detector
and the intensities of the Kot X-rays from Te125andXe129 measured.
The decay of I131 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 I 31 in-
terference or the saitple allowed to decay until the j!31 activity no
B-7
-------
longer seriously interfered with the I129 measurement.
The germanium detector is standardized for both I125 and I129as a
function of weight of the PdI2 precipitate. The Kct X-rays at 27.5 and
27.2 keV for Te125and 29.7 and 29.4 keV forXe129 are vised 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 I131 interference are determined
from I2Pd131standards.
The counts in the I125 region and the I129 region were sunned separately.
The net counting rate in each region was computed. A matrix calculation
was used to correct the I125 net counting rate and the I129net counting
rate for mutual interference from Compton interactions and I133interference.
The appropriate decay /volume, counting efficiency and chemical recovery
corrections were then applied to compute the I125and I129 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-flohr-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
-------
The extraction of G02 and CH4 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
002 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 H2EO4- (NH4) 2003 solution onto a
polished copper disc, and the 5.9 keV X-ray was then measured with an
instrinsic-germanium detector.
B. 14 Ni63 Analysis
Nickel was isolated from water samples, fusion extracts and acid
leachates by forming nickel dimethylgloximate which was extracted into
B-9
-------
chloroform. Nickel carrier, measured spectrqphotometrically, was used to
determine the chemical recovery. The nickel dimethylgloximate was de-
colorized with hydrochloric acid and the 67-keV beta of Ni*> ^counted 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.
B-10
*UA GOVERNMENT PRINTING OFFICE:1978 260-880/1 1-3
-------