ML20210H916
| ML20210H916 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 06/30/1985 |
| From: | Leitch G PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | |
| Shared Package | |
| ML20210H897 | List: |
| References | |
| 2, NUDOCS 8609260367 | |
| Download: ML20210H916 (65) | |
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i SEMI-ANNUAL EFFLUENT RELEASE REPORT s-NO. 2 JANUARY 1, 1985 THROUGH JUNE 30, 1985 t
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Facility Operating License NPF-27
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TABLE OF CONTENTS I.
Summary of Radioactive Liquid and Gaseous Effluents II.
Supplemental Information III.
"D" RHR Service Water Radiation Monitor Failure IV.
Offsite Dose Calculation Manual - Revision 2 O
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SUMMARY
OF RADIOACTIVE LIQUID AND GASEOUS EFFLUENTS In accordance with Limerick Generating Station (LGS) i Technical Specification 6.9.1.8, this report summarizes the effluent release data from LGS for the period January 1, 1985 through June 30, 1985.
There were no measurable releases of radioactive gaseous effluents during the report period.
There i
were no solid radwaste shipments during the report period.
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Radioactive liquid effluent releases, from the LGS liquid radwaste discharge point at the Schuylkill River, are summarized f
following the format of Regulatory Guide 1.21, " Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and 4
Releases of Radioactive Materials in Liquid and Gaseous Effluents i
from Light-Water-Cooled Nuclear Power Plants", Appendix Bi i
Revision 1, June, 1974.
Liquid effluent radiation units of f
"O.000E+00" denote less than minimum detectable levels.
The i
resultant of f-site doses from this pathway to members of the public are not required (per the aforementioned Technical Specification) to be included in this report, but will be evaluated in the January 1, 1986 submittal.
Supplemental Information, following the format of Regulatory Guide 1.21, Appendix B, is also included.
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NOTE 1 TO THE LIQUID EFFLUENT
SUMMARY
REPORT i
Release limits for liquid effluents are based on doses per LGS Technical Specification 3.11.1.2.
The percent of the quarterly limits of the liquid releases are as follows:
QUARTER 1 TOTAL BODY ORGAN 1/41y limit 1.5 mrem 5 mrem Calc. dose 2.919E-05 mrem 2.919E-05 mrem j
1.946E-03 5.838E-04 i
i QUARTER 2 TOTAL BODY ORGAN G,
1/41y limit 1.5 mrem 5 mrem j
Calc. dose 1.325E-05 mrem 2.491E-05 mrem i
8.84 E-04 4.982 E-04 f
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t' j
II. SUPPLEMENTAL INFORMATION Facility: Limerick Generating Station - Unit 1 License:
NPF-27 1.
Regulatory Limits (Technical Specification Limits)
A.
Noble Gases:
i 1 1 500 mRems/Yr - total body
" instantaneous" limits per 1 3000 mRems/yr - skin Tech Spec 3.11.2.1 2,
i.
2 1 5 mRads - air gamma
- quarterly air dose limits per
't
< 10 mRads - air beta Tech Spec. 3.11.2.2 i
3 1 10 mRads - air gamma
- yearly air dose limits per 1 20 mRads - air beta Tech Spec. 3.11.2.2 B.-
Iodines, tritium, particulates with half life > 8 days:
1 1 1500 mRems/yr - any organ
" instantaneous" limits per (inhalation path)
Tech Spec. 3.11.2.1 2
1 7.5 mRems - any organ
- quarterly dose limits per Tech. Spec. 3.11.2.3
- 3. I 15 mRems - any organ
- yearly dose limits per Tech. Spec. 3.11.2.3 C.
Liquid Effluents:
- 1. Concentration < 10CFR20
" instantaneous" limits per Appendix B, TaEle II, Col. 2 Tech. Spec. 3.11.1.1
- 2. I 1.5 mRems - total body
- quarterly dose limits per 1 5 mRems - any organ Tech. Spec. 3.11.1.2 3
1 3 mRems - total body
- yearly dose limits per
[
1 10 mRems - any organ Tech. Spec. 3.11.1.2 2.
Maximum Permissible Concentrations
(
MPCs are not used to calculate permissible release rates and concentrations for gaseous releases.
3 The MPCs specified in 10CFR20, Appendix B, Table II, Column 2 for identified nuclides are used to calculate permissible
.A D
release rates and concentrations for liquid releases per LGS oh' L
Technical Specification 3.11.1.1.
j 7
e es I
i 5'i r.
1 3.,
i o
s>
'W 4
2
r 3.
Avnrrge Enargy E determination will not be done until the reactor has operated for 20 Full Power Days.
4.
Measurements and Approximations of Total Radioactivity A.
Fission and Activation Gases The method used is the Canberra Series 90 counting System; GS
- Gas Marinelli B.
Iodine:
The method used is the Canberra Series 90 counting System; CH
- Charcoal Cartridge C.
Particulate:
The method used is the Canberra Series 90 Counting System; PT
- Air Particulate Sample, 47 mm filter.
D.
Liquid Effluents:
The method used is the Canberra Series 90 Counting System and the Radwaste Liquid Discharge Pre-Release Method with a 3.5 i
Marinelli.
5.
Batch Releases e
A.
Liquid Q1 02 4 of Batch Releases:
181 95 Total Time for batch releases, minutes 12748 7032 i
Maximum time period for a batch release, minutes 132 100 Average time period for batch release, minutes 70 74 Minimum time period for a batch release, minutes 47 46 Average stream flow (blowdown) p during periods of release of 1
effluents into a flowing stream, gpm 5984 5557
,1 B.
Gaseous
.i l h.
ik t
/
j 6.
Abnormal Releases o
s A.
Liquid i
None II d
f B.
Gaseous J.
None J
III.
"D" Rasidunl Heat Rnmoval Servico Water Rndiction Monitor On June 18, 1985, the "D" Residual Heat Removal Service Water Radiation Monitor Sample Flow Switch (FISHL-026-121D) failed high.
A subsequent investigation determined that the inoperable switch could not be repaired.
The specified required delivery date of the replacement device, ordered through General Electric Company,
)
was July 12, 1985.
The long-lead (ordering) time, of approximately 18 weeks, mandatory for this flow switch impacted restoration of the operability of the monitor within 30 days - the time specified in Technical Specification 3.3.7.11.
Replacement
+
and repair alternatives are currently being evaluated to prevent recurrence.
Sampling and analysis of the Service Water are being performed by the Chemistry Department, per the ACTION Statement within the aforementioned Technical Specification, during periods of Residual Heat Removal Service Water Loop B operation.
1 I
i 3
5 i
8 j Y M
y I
d
4 IV. Offnite Dosa Calculation Manuel - Revision 2 Revision 2 of the Offsite Dose Calculation Manual (ODCM) is submitted within.
These changes, which have been generated in concurrence with Limerick Generating Station Technical Specification 6.14, are licensee-initiated; contain sufficiently
, detailed information to totally support the rationale for the I
change without the benefit of additional or supplemental information; and do not degrade the integrity of dose calculations
.or setpoint determinations.
A.
Table VI.A.I. and Figure VI.A.I were revised to more precisely reflect the environmental sampling locations.
B.
Parameters have been provided to enable the " assessment of the radiation doses from radioactive liquid and gaseous effluents to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY", pursuant to Technical Specifications Bases 3/4 11.2.1.
9 j
r
)
. J
OFFSITE DOSE CALCULATION MANUAL Revision 2 i
l LIMERICK GENERATING STATION l
I
(
UNITS 1 AND 2 i
l I
PHILADELPHIA ELECTRIC COMPANY Docket Nos. 50-352 & 50-353 b
PORC Approval:
/ Ft-l'iTJtd 6!d df Station SQperintendent Nuclear and Environmental
/
Section Approval:
Mk ll E
I(
(
LGS Health Physics Representative:Uk d
h.
i 1
o
{
Nuclear and Environmental
/
/
[ /2e/(v3 / /// I /d /
Representative:
/
0
/
~
e
)
Y,
.1' s
Table of Contents d
4-p
.L Page I.
Purpose
~
II.
Liquid Pathway Dose Calculations A.
Surveillance Requirement 4.11.1.1.2 1
B.
Surveillance Requirement 4.11.1.2 2
C.
Surveillance Requirement 4.11.1.3.1 3
III.
Gaseous Pathway Dose Calculations A.
Surveillance Requirement 4.11.2.1.1 5
B.
Surveillance Requirement 4.11.2.2 8
C.
Surveillance Requirement 4.11.2.3 11 D.
Surveillance Requirement 4.11.2.5.1 12 IV.
Nuclear Fuel Cycle Dose Assessment - 40 CFR 190 A.
Surveillance Requirement 4.11.4.1 14 B.
Surveillance Requirement 4.11.4.2, 14 V.
Calendar Year Dose Calculations A.
Unique Reporting Requirement 6.9.1.8 15 VI.
Radiological Environmental Monitoring Program A.
Surveillance Requirement 4.12.1 15 VII.
Effluent Radiation Monitor Setpoint Calculations 22 VIII. Bases 33 IX.
Liquid and Gaseous Effluent Flow Diagrams 38 g
P
[.)
1 l
Y];4 l
~
~~~
~~
c
, _ ~.
e I.
Purpoaa The purpose of the Offsite Dose Calculation Manual is i
to establish methodologies and procedures for j
calculating doses to individuals in areas at and t
beyond the SITE BOUNDARY due to radioactive effluent i
from Limerick Generating-Station and establishing I*
setpoints for radioactive effluent monitoring instrumentation.
The results of these calculations are required to determine compliance with Appendix A to Operating License NPF-27, " Technical Specification and Bases for Lime' rick Generating Station Units No. 1 and 2.
- IIL, Liquid Pathway Dose Calculations A.
Surveillance Requirement 4.11.1.1.2 - Liquid Radwaste Release Compliance with 10CFR20 Limits Limerick Generating Station Units 1 and 2 have one common discharge point for liquid releases under normal circumstances.
In the event of heat exchanger leakage, additional release pathways are possible through the plant service water system and the RHR service water system.
The following calculation assures that the radwaste release limits are met.
The flow rate of liquid radwaste released from the site to areas at and beyond the SITE BOUNDARY
+
shall be such that the concentration of radioactive material af ter dilutibn shall be limited to the concentration specified in 10 CFR 20.106(a) for radionuclides other than the dissolved or entrained noble gases and the concentration listed in Technical Specification Table 3.11.1.1-1 for all dissolved or entrained noble gases as specified in Technical Specification 3.11.1.1.
Each tank of radioactive waste is sampled prior to release and is quantitatively analyzed for identifiable gamma emitters as specified in Table 4.11-1 of the Technical Specification.
From this gamma isotopic analysis the maximum permissible release flow rate is determined as follows:
Determine a. Dilution Factor by:
Dilution Factor uCi/ml i q
=
i MPCi
}
(
uCi/ml i = the activity of each identified gamma i
n emitter in uCi/ml j !
)
[
., j
~
\\
MPCi = The MPC cp0cificd in 10 CFR 20, Appendix B, Table II, Column 2 for radionuclides other t
7 than dissolved or entrained noble gases or the concentrations listed in Technical Specification Table 3.11.1.1-1 for dissolved' or entrained noble gases.
Any unidentified concentration is assigned an MPC value of lE-07 uCi/ml.
s Determine the Maximum Permissible Release Rate with this Dilution Factor by:
=
A l
Release Rate (gpm)
D (Dilution Factor) i A = The cooling tower blowdown volume which will provide dilution.
Maximum flow rate is 10,000 gpm.
l B = margin of assurance which includes consideration of the maximum error in the activity setpoint and the maximum error in the flow setpoint and the possibility of multiple release pathways.
B'.
Surveillance Requirement 4.11.1.2 9
The primary method of calculating dose contributions from liquid effluents released to-areas at or beyond the SITE BOUNDARY will be by using a computer-based calculational program developed using the equations and parameters of R.G.
1.109, Rev.
1, October, 1977*(see bases Note 1
- 4) for all organs and age groups.
The Ai values used for this calculation are located in the Appendix, Table 1.
Dose contributions from liquid effluents released to areas at and beyond the SITE BOUNDARY shall be calculated using the equation below.
This dose calculation uses as a minimum those appropriate radionuclides listed in Table II.A.l.
These radionuclides account for virtually 100 percent of the total body dose and bone dose from liquid effluents.
Ri i(Ai b tl Cil F1)
D
=
1=1 t
[ek;.
j Ut where:
f
i
.t' the cumulative dose commitment to the total'Vt 4
D 7 = body or any organ, T, f rom liquid ef fluents) f'
?
for the total time period md in mrom i
=1 tl
)
i
~2-f.
\\
s a
f i
a
i t
j l
l Ri reported release points
=
I i
ght tl the length of the first time period over which
=
I Cil and F1 are averaged for the liquid release, in hours.
the average concentration of radionuclide, i, Cil
=
in undiluted liquid effluent during time period I
ll t from any liquid release, (determined by the 4
effluent sampling analysis program, Technical Specification Table 4.11.1.1-1), in uCi/ml.
Alf the site related ingestion dose commitment
=
g factor to the total body or organ,T, for L
each radionuclide listed in Table II.A.1, in mrem-ml per hr-uCi.
See Site Specific Data.**
the near field average dilution factor for F1
=
l Cil during any liquid effluent release.
Defined as the ratio of the maximum undiluted liquid waste flow during release to the average flow from the discharge structure to the Schuylkill River.
II.C Surveillance Requirement 4.11.1.3.1 Projected dose contributions from liquid effluents shall be calculated using the methodology described in Section II.B.
To estimate expected concentration of the various radionuclides (Cil) in the undiluted liquid effluent, 3
the duration of liquid release (4t), and the near field average dilution factor (F1), the expected plant operating status shall be reviewed.
If no operational changes are expected which would affect Cil,A t, or F1 the same values as used to evaluate Section II.B may
(
be used.
If any operational changes are expected during the following 31 days which could affect Cil,4 t or F1, the values used shall be based on plant history.
During the initial stages of plant operation, the values for Cil, sit, and F1 as given in LGS FSAR Section 11.2 and,EROL Section 5.2 may be used.
See Note 1 in Bases
, c
TA?;LE II.A.1 LIQUID EFFLUENT INGESTION DOSE FNCTORS (Decay Corrected)
Al Dose Factor (mrem-ml per br-uci)
Radionuclide Total Body Bone Cs-137 3.42E+05 3.82E+05 Cs-134 5.79E+05 2.98E+05 P-32 5.llE+04 2.05E+05 Cs-136 8.42E+04 2.97E+04 Zn-65 3.32E+04 2.31E+04 SrJp0 1.35E+05 5.52E+05 H-3 3.29E-01 Na-24 1.35E+02 1.35E+02 l
I-131 1.16E+02 1.40E+02 Co-60 5.70E+02 I-133 1.23E+01 2.31E-01 Fe-55 1.06E+02 6.61E+02 Sr-89 6.36E+02 2.21E+04 Te-129m 1.70E+03 1.08E+04 Mn-54 8.34E+02 8.34E+02 Co-58 2.00E+02 Fe-59 9.26E+02 1.02E+03 Te-131m 3.88E+02 9.53E+02 Ba-140 1.33E+01 2.03E+02 Te-132 1.21E603 1.99E+03 NOTE: The listed dose factors are for radionuclides that may be detected in liquid effluents and have significant dose consequences.
These factors are decayed for one day to account for the time between effluent release and ingestion of fish by the maximum exposed individual, an adult.
There is no bone dose factor given in R.G.
1.109 for these nuclides.
l J
j III.
Grecous Pathwny Doua Calculations The controling receptor locations for the gaseous
. pathway dose calculations are based on a land-use census performed in 1975 to 1976 which has been periodically updated.
The most recent update was in 1983.
A.
Surveillance Requirement 4.11.2.1.1 The dose rate in areas at and beyond the SITE BOUNDARY due to radioactive materials released in 4
gaseous effluents shall be determined by the i
expressions below:
I 1.
Noble Gcses The dose rate from radioactive noble gas releases shall be determined by either of two methods.
Method (a), the Isotopic Analysis Method, utilizes the results of noble gas analysis required by specification 4.11.2.1.1 and 4.11.2.1.2.
Method (b), the Gross Release Method, assumes that all noble gases released are the most limiting nuclide-Kr-88 for total body dose and Kr-87 for skin dose.
For normal operations, it is expected that method (a) will be used.
However, if isotopic release data are not available method (b) can be used.
Method (a) allows more operating flexibility by using data that more accurately r~eflect actual releases.
i a.
Isotopic Analysis Method i
=)(i(Ki (X/Q)v Qiv)
DTB
=}{i((Li + 1.lMi) (X/Q)v) l Ds where:
The location is the site boundary, 790m NE from the vents.
This location results in the highest calculated dose to an individual from noble gas releases.
DTB
= total body dose rate, in mrem /yr.
Ds
= skin dose, in mrem /yr.
Ki
= the total body dose factor due to gamma emissions for each identified noble gas radionuclide.
Values are listed on Table III.A.1 and are taken from R.G.
1.109, l,
E. _ _ _
in arcm/yr per uCi/m3.
(X/Q)v
= 1.lE-05 sec/m3; the highest calculated annual average relative concentration for any area at or beyond the SITE BOUNDARY for all vent releases (NE boundary).
Qiv
= the release rate of noble gas radionuclide, i, in gaseous effluents from all vent releases determined by isotopic analysis averaged over one hour, in uCi/sec.
i Li
= the skin dose factor due to beta emissions for each identified noble gas radionuclide.
i Values are listed on Table III.A.1 and are taken from R.G.
1.109, in mrem /yr per uCi/m3.
Mi
= the air dose factor due to gamma emissions for each identified noble gas radionuclide.
Values are listed on Table III.A.1 and are taken from R.G.
1.109, in mrad /yr per uCi/m3.
1.1
= unit conversion, converts air dose to skin dose, mrem / mrad.
b.
Gross Release Method DTB
=K (X/Q)V Onv Ds
= (L + 1.lM) (X/Q) Onv where:
The location is the site boundary, 790m NE from the vents.
This location results in the highest calculated dose to an individual from noble gas releases.
DTB
= total body dose rate, in mrem /yr.
Ds
= skin dose rate, in mrem yr.
.t K
= 1.47E04 mrem /yr per uCi/m3; the total body dose factor due to gamma emissions fo,r Kr-88 (Reg. Guide 1.109, Table B-1).
(X/Q)v
= 1.lE-05 sec/m3; the highest calculated annual average relative concentration for any area at or beyond the SITE BOUNDARY for all vent releases (NE boundary).
i i
1 J
~
- = _. - -
(b)ny
= the gross release rate of noble gases in gaseous effluents from vent releases determined by gross activity vent monitors averaged over one hour, in uCi/sec.
L
= 9.73E03 mrem /yr per Ci/m3; the skin dose factor due to beta emissions for Kr-87 (Reg. Guide 1.109, Table B-1).
M
= 6.17E03 mrad /yr per'uC1/m3; the air dose factor due to gamma emissions for Kr-87 (Reg. Guide 1.109, Table B-1).
I.
2.
The primary method of calculating dose contribution from Iodine-131, Iodine-133, tritium, and radioactive material in particulate form, other than noble gases, with half-lives greater than eight days will be by using a computer-based calculational program developed using the equations and parameters of R.G.
1.109, Rev.
1, October, 1977 (see bases Note 4) for all organs and age groups.
If the computer model is not available, the dose contributions from Iodine-131, Iodine-133, tritium, and radioactive materials in particulate form, other than noble gases, with half-lives greater than eight days will be calculated using the equation below:
~
DT
= (CF)i Pi (WV (Qiv))
where:
The location is the site boundary, 762m ESE from the vents.
DT
= dose rate to the thyroid, in mrem /yr.
= 1.02; the correction factor accounting for the use of iodine-131 and iodine-133 in lieu of all radionuclides released in gaseous effluents.
P
= 1.62E07 mrem /yr per uCi/m3; the inhalation I-131 do'se parameter for I-131 inhalation pathway.
The dose factor is based on the critical
- 1 individual organ, thyroid, and most restrictive t]
age group, child.
All values are from Reg.'.
jj Guide 1.109 (Tables E-5 and E-9).**
',J
-P
= 3.85E06 mrem /yr per uCi/m3; the inhalation I-133 dose parameter for I-133 inhalation pathway.
The dose factor is based on the critical 1 (
J
individual organ, thyroid, cnd moot rectrictiva age group, child.
All values are from Reg.
Guide 1.109 (Tables E-5 and E-9).**
WV
= 1.00E-05 sec/m3; the highest calculated annual average relative concentration for any area at or beyond the SITE BOUNDARY for all vent releases (NE boundary).
hiv
= the release rate of iodine-131 and/or l'odine-133 in gaseous effluents from all vent releases, determined by the effluent I,
sampling and analysis program (Technical 4
Specification Table 4.8.2) in uCi/sec.
u III.B Surveillance Requirement 4.11.2.2 The air dose in areas at and beyond the SITE BOUNDARY due to noble gases released in gaseous effluents shall be determined by the expressions below.
i The' dose rate from radioactive noble gas releases shall be determined by either of two methods.
Method
'l (a), the Isotopic Analysis Method, utilizes the resu'.ts of noble gas analysis required by specification 4.11.2.1.1 and 4.11.2.1.2.
Method (b),
the Gross Release Method, assumes that all noble gases released are the most limiting nuclide
- Kr-88 for total body dose and Kr-87 for skin dose, i
For normal operations, it is expected that method (a) will be used.
However, if isotopic release data are not available, method (b) can be used.
Method (a)
)
allows more operating flexibility by using data that more accurately reflects actual releases.
i See Note 2 in Bases
\\
l.
for gamma radiation b
p i
a)
Isotopic Analysis Method t,
- L p
I s..
t
= 3.17E-08 )
i(Mi(X/Q)v Qiv)
' y., ;
]
DG 3~
f i
\\
'i t
- t 1
a
<3%
F_ _ _.
I where:
The location is the SITE BOUNDARY, 762m ESE from i
the vents.
This location results in the highest j
calculated gamma air dose from noble gas releases.
i 1
where:
l -
= gamma air dose, in mrad.
I
'3.17E-08= years per second.
i I
J l
Mi
= the air dose factor due to gamma emissions for each identified noble gas radionuclide.
Values are listed on Table III.A.1 and are i
taken from R.G.
1.109 in mrad /yr per uCi/m3.
(X/0)V
= 1.lE-05 sec/m3; the highest calculated average relative concentration from vent releases for any area at or beyond the SITE 4
BOUNDARY.
the release of noble gas radionuclides, i, Qiv
=
in gaseous effluents from all vents as determined by isotopic analysis, in uCi.
Releases shall be cumulative over the calendar quarter or year, as appropriat'e.
b.
Gross Release Method j
3.27E-08 (M
(X/Q)v Qv )
=
where:
The location is the SITE BOUNDARY 790m NE from the vents.
This location results in the highest calculated gamma air dose from noble gas releases.
gamma' air dose, in mrad.
a DG
=
3.17E-08= years per second.
~
1.52204 mrad /yr per uCi/m3; the air d
=
t j
s -
dose factor due to gamma emissions for Kr-88 (Reg. Guide 1.109, Table B-1).
.4 g
1.lE-05 sec/m3; the highest calculated I
,)
(X/Q)v
=
3 d
sg annual average relative concentration v
from vent releases for any area at or g
,m
'['
g l'
g\\
?
'beyond the SITE BOUNDARY.
\\
f 6i.-
the gross release of noble gas radio-2
,'i nuclides in gaseous effluents from all t
I
W 1
vsnte, datorminsd by grote cctivity vant monitors, in uCi.
Releases shall be 3
cumulative over the calendar quarter or year as appropriate.
2.
for beta radiation j*
a.
Isotopic Analysis
=3.17E-08)i(Ni (X/Q)v Qiv)
DB where:
l The location is the SITE BOUNDARY 790m NC from I
the vents.
This location is the highest calculated gamma air dose from noble gas releases.
3.17E-08 = years per second.
Ni
= the air dose factor due to beta emissions for sach identified noble gas radionuclide.
s j
Values are listed on Table III.A.1 and are taken from Reg. Guide 1.109, in mrad /yr t
per uCi/m3.
{
(X/0)v
= 1.lE-05 sec/m3; the highest calculated annual average relative concentration from vent releases for any area at or beyond the SITE BOUNDARY.
Qiv
= the release of noble gas radionuclide, 1
1, in gaseous effluents from all vents as determined by isotopic analysis, in uC1.
Releases shall be cumulative over the calendar quarter or year, as appropriate.
b.
Gross Release Method DB
= 3.17E-08 N (X/Q)v Qv where:
The location is the SITE BOUNDARY 790m NE from the vents.
This location results in the highest calculated gamma air dose from noble gas releases.
= beta air dose, in mrad.
3.17E-08 = years per second.
bk l
ES' N
= 1.03E04 mrad /yr per uCi/m3; the air dose i t.}
factor due to beta emissions for Kr-87 (Reg.
7 Guide 1.109, Table B-1).
- ~
g;
,9 I
3 p
e L
(X/Q)v
= 1.lE-05 sec/m3; the highest calculated annual average relative concentration from vent releases for any area at or beyond the SITE BOUNDARY.
Qv
= the gross ~ release of noble gas radionuclides in gaseous effluents from all vents determined by gross activity vent monitors, in uCi.
Releases shall be cumulative over the. calendar quarter or year, as appropriate.
III.C Surveillance Requirement 4.11.2.3 1
The primary method of calculating dose to an individual from Iodine-131, Iodine-133, tritium, and radioactive materials in particulate form, other than noble gases, with half-lives greater than eight days in gaseous effluents released to areas at and beyond the SITE BOUNDARY, will be by using a computer-based calculational program developed using the equations and pararaeters of R.C.
1.109, Rev.
1, October, 1977 (see Bases Note 4) for all organs and age groups.
If' the computer model is not available, the following expression will be used:
D = 3.17E-09 (CF)
(0.5))f (Ri WV Qiv) where:
Location is the critical pathway dairy 1770m ESE from vents.
D
=
critical organ dose, thyroid, from all pathways, in mrem.
3.17E-08 = years per second.
= 1.00; the correction factor accounting for the use of Iodine-131 and Iodine-133 in lieu of all radionuclides released in gaseous effluents.
0.5
= fraction of iodine releases which are nonelemental.
4 I
R
= 9.51 Ell m2 (mrem /yr) per (uci/sec); the dose
!l]f I-131 factor for Iodine-131.
The dose factor is based on the critical individual organ, thryoid, and fQ, most restrictive age group, infant.
See Site Specific Data.**
2
.i s
i
_J
T R
= 8.13E09 m2 (mrcm/yr) par uCi/ccc; the does I-133 factor for Iodine-133.
The dose factor is based on the critical individual organ, thyroid, and most restrictive age group, infant.
See Site
}
Specific Data.**
?
WV
= 1.82E-9/m2 (D/Q) for the food pathway for vent releases.
1 Qiv
= the release of Iodine-131 and/or Iodine-133 determined by the effluent sampling and analysis program (Technical Specification Table 4.11.2.1.2-1) in uCi.
Releases shall be cumulative over the calendar quarter or year, as appropriate.
i III'.D Surveillance Requirement 4.11.2.5.1 The projected doses from releases of gaseous effluents to areas at and beyond the SITE BOUNDARY shall be calculated in accordance with the following sections of this manual:
a.
gamma air dose - III.B.1 bs beta air dose
- III.B.2 c.
organ dose
- III.C The projected dose calculation shall be based on expected releases from plant operation.
The normal release pathways result in the maximum releases from the plant.
Any alternative release pathways result in lower releases and therefore lower doses.
To estimate the expected releases of noble gases and radioiodines in gaseous effluents, the expected plant operating status shall be reviewed.
If no operational changes are expected which would affect the magnitude or type of releases the same values u3ed to evaluate Sections III.B.1, III.B.2 and III.C mav be used, if any operational changes are expected during the following 31 days which could affect the magnitude or type of releases, the values used shall be based on plant history.
During the initial stages of plant operation the values for releases expected as given in LGS FSAR Section,11.3 may be used.
i See Note 3 in Bases h
'l!
,l1 l
.f<
I!
t'
'I i E
(_..
TABLE III.A.1 DOSE FACTORS FOR EXPOSURE TO A SEMI-INFI ITE CLOUD OF NOBLE GASES Nuclide B-air *(Ni)
B-Skin **(Li[
G-Air *(Mi)
G-Body **(Ki)
Kr-83m 2.88E-04 1.93E-05 7.56E-08 Kr-85m 1.97E-03 1.46E-03 1.23E-03 1.17E-03 Kr-85 1.95E-03 1.34E-03 1.72E-05 1.61E-05 Kr-87 1.03E-02 9.73E-03 6.17E-03 5.92E-03 i
Kr-88' 2.93E-03 2.37E-03 1.52E-02 1.47E-02 Kr-89 1.06E-02 1.01E-02 1.73E-02 1.66E-02 Kr-90 7.83E-03 7.29E-03 1.63E-02 1.56E-02 Xe-131m 1.11E-03 4.76E-04 1.56E-04 9.15E-05 Xe-133m 1.48E-03 9.94E-04 3.27E-04 2.51E-04 Xo-133 1.05E-03 3.06E-04 3.53E-04 2.94E-04 Xe-135m 7.39E-04 7.llE-04 3.36E-03
}.12E-03 Xe-135 2.46E-03 1.86E-03 1.92E-03 1.81E-03 Xe-137 1.27E-02 1.22E-02 1.FlE-03 1.42E-03 Xc-138 4.75E-03 4.13E-03 9.21E-03 8.83E-03 Ar-41 3.28E-03 2.69E-03 9.30E-03 8.84E-03 i
- mrad-m3 pCi yr l
- mrem-m3 pC1 yr
REFERENCE:
Regulatory Guide 1.109, Revision 1, October 1977 b.
! ' fl
[
Sil e
r
+
IV.
TOTAL DOSE 2
A.
Surveillance Requirement 4.11.4.1
.l i
If the doses as calculated by the equations in this manual do not exceed the limits given in Technical Specifications 3.llil.2.a, 3.11.2.b, 3.ll.2.a, 3.11.2.2.b, 3.ll.2.3.a, or 3.11.2.3.b by more than two times, the conditions of Technical Specification 3.11.4.2 have been met.
B.
Surveillance R'equirement 4.11.4.2 If the doses as calculated by the equations in this manual exceed the limits given in Technical Specifications 3.ll.l.2.a, 3.ll.l.2.b, 3.11.2.2.a, 3.11.2.2.b, 3.11.2.3.a, or 3.ll.2.3.b by more than two times, the maximum dose or dose commitment to a real individual shall be determined utilizing the methodology provided in Regulatory Guide 1.109,
" Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I",
Revision 1, October 1977.
Any deviations from the methodology provided in Regulatory Guide 1.109 shall be documented in the Special Report to be prepared in accordance with Technical Specification 3.11.4.1.
The cumulative dose' contribution from direct radiation from the two reactors at the site and from radwaste storage shall be determined by th,e following method:
Cumulative dose contribution from direct radiation =
Total dose at the site of interest (as evaluated by TLD measurement) -
Mean of background dose (as evaluated by TLD's at background sites) -
Effluent contribution to dose (as evaluated above).
The method provided in the second paragraph above is used only to evaluate the contribution from direct radiation dose.
The direct radiation dose is then added to the dose or dose commitment determined in accordance with the methods in the first paragraph above to determine total dose from all pathways.
p-This evaluation is in accordance with ANSI /ANS 6.6.1-1979 Section 7.
The error using this method is y'
estimated to be approximately 8%.
9
?
k ;
g n
i V.A Unique Reporting Requirement for the Radioactive Effluent (6.9.1.8) - Dose Calculations Release Report The assessment of radiation doses for the radiation dose i
assessment report shall be performed utilizing the methodology provided in Regulatory Guide 1.109, of Reactor Effluents for the Purpose of Evaluating"Calculrti Compliance with 10 CFR Part 50, Appendix I",
Revision 1, October 1977.
in Regulatory Guide 1.109 shall be documented in theAny deviations fro radiation dose assessment report.
l The meteorological conditions concurrent with the time of 1
8, release of radioactive materials frequency of measurement)
(as determined by sampling or approximate methods shall be used as input to the dose model.
l within 60 days after January 1 of each year.The Radioactive Eff VI.A SurveiJlance Requirement 4.12.1 collected pursuantThe radiological environmental monitoring samples shall be to Table VI.A.1 from the locations shown on Figures VI.A.1, to the requirements of Table 3.12-1 of the LGSVI.A.2 and VI.A pursuant Technical Specifications.
VII.A Surveillance Requirement 4.12.3 Pursuant Specifications,to Section 4.12.3 of the LGS Te*chnical environmental analyses shall participate in anthe laboratory performing the 4
4 interlaboratory comparison program which has been approved by the NRC.
This program is the Environmental Protection
{
Agency's (EPA's) Environmental Laboratory Intercomparison Studies (cross check) Program.
(sample medium-radionuclide combination) Participation inc CJ.
by the EPA and that are also included in the monitoring that are offered program.
The results of the analysis of these (cross check)
Environmental Operating Report. samples will be included in the Annual Radio 5
I _
I TABLE VI.A.1 1
RADIOIOGICAL EWIRON'4ENT'AL PONI'IORING PROGRAM EXPOSURE PATHWAY NUMBER OF SAMPLES AND S' PATI (N STATION DISTANCE AND/OR SAMPLE SAMPLE STATION NAME CODE SIDUR (MILES)
COPMENTS i
1 Direct 40 Iocations (a) TLD sites were chosen in accordance Radiation (a)
INNER RING IOCATIONS with Limerick Generating Station's i
- 1) Evergreen & Sanatoga 90 ads 36S1 N
0.6
'nechnical Specifications Table
- 2) Sanatoga Road 3S1 NNE 0.6 3.12-7, Item 1.
The inner ring
- 3) Possum Hollow Road SSI NE 0.4 and outer ring stations cover
- 4) TES Training Center 7S1 ENE 0.5 all sectors.
- 5) Keen Road 10S3 E
0.5
- 6) ISS Information Center llS1 ESE 0.5
'The control and special interest
- 7) Longview Road, SE Sector 14S1 SE 0.6 stations provide information on Site Boundary population centers and other
Site Boundary i
- 9) Railroad Track Along 1891 S
0.3 i
Tongview Road
i
- 11) Transmission 'Ibwer, WSW 23S2 WSW 0.5 l
Sector Site Boundary
- 13) Meteorological 'Ibwer 2 Site 26S3 M
0.4
- 14) WNW Sector Site Boundary 29S1 WNW 0.5
- 15) NW Sector Site Boundary 32S1 NW 0.6
- 16) Meteorological '1bwer 1 Site 34S2 NNW 0.6 OLTPER RING IOCATIONS
- 1) Ringing Rock Substation 35F1 N
4.2
- 2) Laughing Waters GSC 2El NNE 5.1
- 3) Neiffer Hoad 4El NE 4.6
- 4) Pheasant Road, Game Farm 7El ENE 4.2 Site
- 5) Transmission Corrider 10El E
3.9
- 6) Trappe Substatico 10F3 ESE 5.5
- 7) Vaughn Substation 13El SE 4.3
- 8) Pikeland Substation 16F1 SSE 4.9
- 9) ~Showden Substation 19D1 S
-r
-. -MO), qhander Substation - '
- - 20F1 SSW ~, _ _3,.6
~5.2 ENGR.
/Jgg
-~
H.P.
JA([,
DATE f/d. j'_y I
.m
~ ~- **=~* w
- Q g 3y__
_-___ y _=
e
OUTER RING IOCATICNS (Cont'd) t
- 11) Porter's Mill Substation 24D1 SW 3.9
- 12) Transmission Corrider, 25D1 WSW 4.0 Hoffecker and Keim Streets l
t J
- 13) 'n ansmission Corrider, 28D2 W
3.8 W. Cedarville Road i
- 14) Prince Street 29El WNW 4.9
- 15) Poplar Substation 31D2 NW 3.9 i
- 16) Yarnell Road 34E1 NNW 4.6 OOTPROL AND SPECIAL INJwe6I' f
IOCATIONS
?
- 1) Birch Substation (control)
SN1 NE 25.8 j
- 2) Pottstown Landing Field 6C1 ENE 2.1
- 3) Reed noad 001 E
2.2 i
- 4) King Road 13Cl SE 2.9
- 5) Spring City Substation 15D1 SE, 3.2
- 6) Linfield Substation 17D1 S
1.6
- 7) Ellis Woods Road 20D1 SSW 3.1
- 8) Lincoln Substation 31D1 NW 3.0 5 IoCATIONS
- 2. Airborne
- 1) Keen Road 10S3 E
0.5 (b) These stations provide for l
- 2) ISS Information Center llS1 ESE 0.5 coverage of the highest annual j
Radiolodine,and 3) Iongview Road 14S1 SE 0.6 ground level D/Q, and a j
Particulates
- 4) King Road 13Cl SE 2.9 control location. Radio-(b)
- 5) 2301 Market Street, 13N4 SE 28.8 iodine cartridges which have Philadelphia, PA (control) been tested for performance by 3
the manufacturer are used l
at all times.
- 3. Waterborne (c) 9 IOCATIONS (c) All surface and drinking stations
~
have continuous samplers.
l Surface
- 1) Limerick Intake (control) 24S1 SSW 0.3
- 2) Linfield Bridge 16B2 SSE -
1.1 Ground
- 1) ISS Information Center 1191 ESE 0.5
- 2) South Sector Farm Near Site 18Al S
1.0 Drinking
- 1) Phoenixville Water Works 15F7 SSE 5.2
- 2) Pottstown Water Authority 28F3 WNW 5.9 (control)
- 3) Philadelphia Suburban Water 15F4 SE 7.8 Ccupany
~
'"" 3)MEehs 'Home" Water Company.? 16CE,d "$* A C-SSE 2.4
..16C4 1.9 ENGR. M SedimentTuun
'lbViricent' Dam Pool' Area.
H.P. (A/LkC,
' ~ ~ ~ ~ ~ ~ ' '
Shoreline tL=
DATE 444r
- 4. Ingestion 6 IGCATIONS Milk (d)
- 1) Control Station 22P1 (d) Milk samples are taken frcan l
2)
SC1 several farms surrounding IGS.
3) 10B1
%ese farms include those with the 4) 25B1 highest dose potential from which j'
as well as a control station.
samples are routinely available, W e locations of the farms are not listed herein due to a long-standing agreement with the farmers involved. In return for being allowed to sample and analyze the milk, PFID has agreed not to divulge the location of the farms.
Fish (e)
- 1) Middle of Vincent Pool 16C3 SSE 1.9 (e) Two species of recretionally im-upstream to Pigeon Creek porant fish, sunfish and brown bullhead will be sampled if i
available.
I
- 2) Upstream of IGS, Keim Street 29Cl WNW 3.2 i
Bridge to Hanover Street Bridge (control) f Fbod Products
- 1) IGS Information Center llS1
.ESE 0.5 (f) Food products are to be samples (f) as part of the IGS Technical Specification Program only if milk sampling is not performed.
%e milk pathway, which results in a higher maximum dose to humans than the vegetation path-way is monitored at a location near the site, and is a better indicator than vegetation-samples. In addition, no crops grown in the vicinity of IGS are irrigated with water in I
which liquid plant wastes have i
been discharged. -
r,-
7.._
., g
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e,m 1
1 9 enums 8
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s s.
345$et.
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i 4,144ERICK GENERATING STATION ENGR. M
.1 2
H.PN _
8 DATE g/g/J.s-
' O.2,-ilss '
ENVIRCNMENTAL SAMPLING STATIONS SITIBOUNDARY j-l FIGURI 72.A.1
- yo
[
--.,,-r-
_.1,.,_
.,e.'
i soonTx yd "kt
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p' J s l
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i LiedERICE CINERATING STATICN UnnT31 AND 2 l
1 141 <.:
'ENVIRDNMENTAL SAMPLING STATsONS i.
INTERMEDIATE DISTANCE.
a i
TICURI 71.A.2 20
[
1 P
ac,,
acad amt ht 1
i I
we i
M u.
INl:
e {1 il 6.
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or LRtERICK GENERAT1kG STAT)Cet Lact 51 AND 2 1
ENVIRONMENTAL SAMPLING STATIONS 3
OlsTANT_LCtcATIONS j
_ _.. l
/0 n,;le.s 8
}
/
FIGURE VI.A.3 21 i
b i
w VII.
Effluent Rediation Monitor Satpoint Calculations A.
Liquid Effluents e
1.
Radwaste Discharge Line Radiation Monitor -
Monitor alarm setpoints will be determined in order to assure compliance with 10CFR20.
The setpoints will indicate if the concentration of a
radionuclides in the liquid effluent at the site boundary is approaching the concentrations specified in 10CFR20, Appendix B, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases.
The setpoints will also assure that a concentrations listed on i
Technical Specification Table 3.11.1.1-1 for dissolved or entrained noble gases is not exceeded.
The following method applies to liquid releases from the plant via the cooling tower blowdown line when determining the high-high alarm setpoint for the Liquid Radwaste Effluent Monitor during all operational conditions.
When the high-high alarm setpoint is reached or exceeded, the releases will be automatically terminated.
a.
The setpoint for the Liquid Radwaste Effluent monitor will be calculated as follows:
1)
Determine Ct Ct = h CiD
( FL) 5 (Ci/MPCi) where:
Ct
= concentration at the liquid radwaste discharge line monitor prior to dilution (to assure 10CFR20.106 limits are not exceeded); uCi/cc
)[Ci total concentration of liquid effluent discharge
=
prior to dilution with cooling tower blowdown; uCi/cc 5
= margin of safety factor including F uncertainty, to assure that the high-high alarm will terminate the disch,arge before 10CFR20 limits are exceeded.
)h
[
Ci
= sum of the ratio of the isotopic concentrations MPCi divided by their respective MPC.
,,,,./-
l*
1 i
k l
_~
m.
D
= dilution factor due to blowdown from the cooling tower; calculated by dividing the total flow (cooling tower blowdown plus radwaste discharge flow) by the radwaste discharge flow.
Fi
= Ratio of MPC-weighted releases in the liquid radwaste effluent monitor flow path divided by the total MPC-weighted liquid releases; a
e.g.
Ci release of flow path of interest Ci I
all release flow paths MPCi 2.
Determine C.R.
7 C.R.
=Ct E
where:
C.R.
= the calculated monitor count rate above background attributable to the radionuclides; CPS the detection efficiency of the monitor; E
=
uCi/cc/ cps.
3.
The monitor high-high alarm setpoint above background should be set at the C.R.
value.
b.
The monitor high-high alarm setpoint will be calculated monthly.
The calculation will be based on isotopes detected in the liquid radwaste sample tanks during the previous month.
If there were no isotopes detected during the previous month then the annual average concentrations (EROL Table 3.5-3) of those isotopes listed in Table II.A.1 will be used to determine the setpoint.
If the calculated setpoint is less than the existing monitor setpoint, the setpoint will be geduced to the new value.
If the calculated setpoint is greater than the j
4-b{
existing monitor setpoint, the setpoint may
{
j l,i; remain at the lower value or increased to c.
I
's the new value.
!' f. I ; J '
ip i i.i 2.
Plant Service Water Monitor - Monitor alarm id I
setpoint will be determined in order to be able '
dj to identify and rectify any potential problem du'e
/
l 9
to excessive leakage of heat exchangers.
This
] 13 setpoint results in concentrations at the site
.j boundary far below 10CFR20, Appendix B, Table II t
e ;
(
l i
l limits.
Tha cervico water cida of tha fusi pool haat exchangers is kept at highor pressure than l
the shell side to prevent potential radioactive 1
contamination of the service water.
t a.
The setpoint for the Plant Service Water monitor will be calculated as follows:
1)
Determine C.R.s C.R.s = Z (CRb) where:
C.R.s
= the calculated monitor setpoint count rate l
3 attributable to system leakage plus background; CPM Z
= multiplier to establish monitor setpoint count rate above background count rate C.R.b
= monitor count rate attributable to background j
radiation; CPM b.
The monitor high alarm setpoint will be calculated monthly.
The calculation will be based on the background count rate during the previous month.
If the calculated P
c setpoint is less than the existing moni' tor setpoint, the setpoint will be reduced to the new value.
If the calculated setpoint is greater than the existing setpoint, the U'
setpoint may remain at th*e lower value or increased at the new value.
3.
RHR Service Water Monitor - Monitor alarm setpoints will be determined in order to be able to identify and rectify any potential problem due to excessive lea': age of heat exchangers.
This setpoint results in concentrations at the site boundary far below 10CFR20, Appendix B, Table II limits.
The following method applies to liquid releases from the plant to the spray pond when determining the high-high alarm setpoint for the RHR Service Water Monitor during all operational conditions.
When the high-high alarm setpoint is reached or exceeded, the releases will be automatically terminated.
a.
The setpoint for the RHR Service Water monitor will be calculated as follows:
1)
Determine C.R.
C.R.s = Z x C.R.b where:
l
-~
C.R.s
= the calculatsd monitor count rate above background attributable to system leakage plus background; CPM' I
Z
= multiplier to establish monitor setpoint count i
rate above background count rate.
C.R b
= monitor count rate attributable to background radiation; CPM
~
E
= the detection efficiency of the monitor; uCi/cc/ CPM.
3)
The monitor high-high alarm setpoint above background should be set at the C.R.
value.
b.
The monitor high-high alarm setpoint will be calculated monthly.
The calculation will be based on the background count rate during the previous month.
If the calculated setpoint is less than the existing monitor setpoint, the setpoint will be reduced to the new value.
If the calculated setpoint is greater than the existing monitor setpoint, the setpoint may remain at the lower value or increased to the new value.
i B.
Gaseous Effluents 1.
North and South 5 tack Vent Radiation Monitors -
Monitor alarm setpoints will be determined in order to assure compliance with 10CFR20.
The setpoints will indicate if the* dose rate at or beyond the site boundary due to radionuclides in the gaseous effluent released from the site is J
approaching 500 mrem /yr to the whole body and i
3000 mrem /yr to the skin from noble gases, or 1500 mrem /yr to the thyroid from I-131 and I-133 (inhalation pathway only).
The alarm setpoint for the gaseous effluent radiation monitors will be calculated as follows:
4 a.
North and South Stack Vent Noble Gas Channel i
j 1)
Determine Ct Ct = 2.12E-03 Qt F
2 4,
thy,,h i
i where:
,i f.
l ',
O 3
.j; -
I Ct
= the concentration at the vent noble gas radiation li i
monitor which indicates that the 10CFR20 dose p
rate limit at the site boundary has been reached; l
uCi/cc 4
l I
i i
2.12E-03= unit convareion factor to convort uC1/ esc /CFM to uCi/cc.
Qt
= the total release rate of all noble gas radio-nuclides in the gaseous effluent (uci/sec) based on the lower of either the whole body exposure limit (500 mrem /yr) or the skin exposure (3000 mrem /yr) Qt will be calculated as shown in Attachment 1.
F
= anticipated maximum vent flow rate; CFM
~
2)
Determine the noble gas channel alarm setpoint (Sn)
I Sn =
VFi (Ct) where:
VFi
= fractional contribution to site boundary dose rate from the release point of interest; i.e. noble gas dose rate contribution 7
from North Vent divided by the total noble gas dose rate contribution from
?
the North and South Vents.
j Normally the VFi values will be 4
determined on a monthly basis but may be performed more often in response to plant conditions.
{
b.
North and South Stack Vent Iodine Channel 1)
Determine Ct i
s Ct =
2.12E-03 Ot F
1 I
where:
Ct
= the concentration at the vent iodine radiation monitor which indicates that the 10CFR20 dose rate limit at the site boundary has been reached;,a uCi/cc.
l l
2.12E-03= unit converstion factor to convert uCi/sec/CFM to uCi/cc.
Qt
= the total release rate of radiciodines in the f
{
O l i C
gaseous effluents (uCi/sec) Qt will be y
..,{ g calculated as shown in Attachment 1.
pq, F
= maximum antcipated vent flow; CFM.
'y i
2)
Determine the iodine channel alarm i
4 setpoint (Si)
Si =
VFi (Ct) f 1
L.
1 wharo:
4 VFi
= fractional contribution to site boundary dose rate from the release point of interest; i.e.
iodine dose rate contribution from North Vent divided by total iodine dose rate contribution from the North and South Vents.
Harmally, the VFi values will be determined on a monthly basis but may be performed more often in response to plant conditions.
i i
2.
The monitor alarm setpoints will be calculated a
monthly.
These routine calculations will be i
based on isotopic analysis of the first scheduled sample of the month.
The monitor alarm setpoint calculations may be performed more often in response to plant conditions.
If there were no I
isotopes detected in the sample, then isotopic concentrations calculated from the expected i
annual average noble gas and iodine-131 and 133 isotopic release rates (EROL Table 3.5-6) will be used to determine the setpoint.
If any calculated setpoint is less than the existing monitor setpoint, the setpoint will be reduced to the new value.
If the calculated setpoint i greater than the existing value, the setpoin,st may remain at the lower value or increased to the new value.
f Due to the fact that I-131 and" I-133 comprise I
98.5% of the total dose based on expected annual average releases (LGS FSAR Table 11.3-1) and particulates contribute a minor fraction of the total dose, a particulate channel cetpoint will not be calculated for purposes of the ODCM.
3.
Containment Purge Isolation a.
Monitor alarm setpoints will be determined for the North Stack Vent Wide Range Gas Monitor to initiate closure of the containment purge supply and exhaust lines in the event that high radioactivity releases are detected.
The setpoint will be dete'rmined to alarm and isolate containment at the minimum release rate from the North Vent which corresponds to a value less than h.
or equal to 2.1 uCi/cc.
The total effluent tj high alarm setpoint for the Wide Range Gas 21.:
Monitor will be calculated as follows:
d&
p 4
1)
Determine Si I'*
('
Si = Ci x F (min.) x 472
' i; O
I whare:
Si
= containment purge isolation setpoint (uCi/sec)
Ci
= a value < 2.1 uCi/cc determined by the plant staff.
minimum anticipated vent flow F(min)
=
rate during purge 472
= units conversion factor to convert uCi/cc per CFM to uCi/sec 4.
Containment Purge During Routine Operations a.
Monitor alarm setpoints will be determined for the North Stack Vent Wide Range Gas Monitor to indicate to Control Room personnel that unanticipated high radioactivity releases are detected.
The setpoint will be determined to alarm in the event that 10CFR20 dose rates at the site boundary are approached or exceeded.
The total effluent alert alarm setpoint for the l
Wide Range Gas Monitor will be calculated as follows:
1)
Determine Sn Sn = VFiQt where:
Sn = Containment purge 10CFR20 alert alarm limit (uCi/sec)
Qt = the total release rate of all noble gas radionuclides in the gaseous effluent (uCi/sec) based on the lower of either the whole body exposure limit (500 mrem /yr) or the skin exposure limit (3000 mrem /yr) g fractional contribution to site boundary VFi
=
dose rate from the release point of interest i.e. noble gas dose rate contribution from the north vent divided by total noble gas dose rate contribution from the North and South Vents Normally the VFi values will be determined on a monthly basis but may be performed more often in response to plant conditions.
b)
Prior to containment purge and venting, the monitor setpoint will be recalculated.
The calculations will be based on the noble 4
gases detected by isotopic analysis of the containment atmosphere.
If the calculated 28 -
E sotpoint in less than the exiciting monitor setpoing, the setpoint will be reduced to the new value.
If the calculated setpoint is greater than the existing value,-the setpoint may remain at the lower value or increased to the new value.
5.
Hot Maintenance Shop Setpoint Determination a.
The Hot Maintenance Shop Particulate and Iodine setpoints are based on a worst case isotope assumption.
Although the application of the worst case isotope results in a highly conservative setpoint, I
releases from the Hot Maintenance Shop are expected to be small by comparison.
In addition, a sufficient margin of safety factor is built in to the calculation, to preclude the application of a VFi for the release point.
1.
The iodine high alarm setpoint is set to alarm in the event that 10CFR20 dose rates at the site boundary are approached or exceeded.
The methodology is as follows:
C 1500 mR/hr
=
t (1.0E-05 sec/m3)(7000CFM)(472)(1.62E07 mrem /yr) uCi/m3 where:
Ct = the concentration at the iodine monitor which indicates that the 10CFR dose rate limit at the site boundary has been reacned, 2.8E-06 uCi/cc 1500mR/yr=10CFR20 dose rate limit for iodine, tritium and particulates with half lives greater than 8 days.
1.0E-05 sec/m3 = annual average depleted Chi /Q 7000 CFM = maximum vent flow rate 472= conversion factor to convert uC1/sec per CFM to uCi/cc l
1.62E07 mrem /yr = inhalation dose factor, I - 131 for uCi/m3 Child, per Reg. Guide 1.109 6
n
2.
Determine the Hot Maintenance Shop high alarm setpoint for iodine as follows:
q ;.
Si = 0.01(2.8E-06 uCi/cc) t !
whgros i
Hot Maintenance Shop Iodine high alarm Si
=
setpoint; 2.8E-08 uCi/cc
.0l= Margin of Safety Factor to encompass possible contribution from all other release points.
3.
The particulate high alarm setpoint is set to alarm in the event that 10CFR20 dose rates at the site bo'undary are approached or exceeded.
The i
methodology is as follows:
s Sp =.01(2.8E-06 uCi/cc)
6.1 where
Sp = Hot Maintenance Shop Particulate high alarm setpoint; 4.59E-09 uCi/cc Q
.01F Margin of Safety Factor to encompass possible i
contribution from all other release points 2.8E-06 uCi/cc = The concentration at the iodine monitor which indicates that 10CFR20 dose rate limits at the site boundary have been reached.
[I 6.1= ratio of the adult inhalation dose factor for 1
Sr-90 x breathing rate for adult to the child s'
inhalation dose factor for I-131 x breathing rate for child.
This ratio may be modified i
by plant personnel if the isotopes available for release are identified and a new ratio based on dose weighted averages is established t
4
+
',, 1 6
k ', '
o 0
l I
4 J i
=
ATTACHMENT 1 Qt Calculations 1.
Ot(whole body 500
=
(X/Q)v KiSi where:
Qt
= the total release rate of all noble gas radionuclides in the gaseous effluent; uCi/sec.
i
'(X/Q)v
= 1.lE-05 sec/m3; the hichest calculated annual average relative concentration for an area at or beyond the site boundary for all vent releases (NE boundary).
r Ki
= whole body gamma dose factors due to noble gases listed on Table III.A.1 (from Reg.
k Guide 1.109, Table B-1).
v-Si -
= the fraction of the total radioactivity in the b
- (4 gaseous effluent comprised by noble gas I
radionuclide "i".
~
2.
Q(t(skin))=
3000 (X/Q)vj(i((Li + 1.lMi)Si) y (X/Q)v
= 1.lE-05 sec/m3; the highest calculated annual average relative concentration for an area at or beyond the site boundary for all vent releases (NE boundary).
7, Li
= beta skin dose factor due to noble gases, listed on Table III.A.1 (from Reg. Guide 1.109, l
Table B-1).
i I
Mi
= air dose factor due to noble gases, listed on Table III.A.1 (from Reg.
Guide 1.109, Table B-1).
i Si
= the fraction of the total radioactivity in the gaseous effluent comprised by noble gas
~
radionuclide "i".
h., A
. \\'
(X/Q)djg'PiAi lf [,s,) ~I.k 4 l'/'
3.
Ot(thyroid)=
1500 5,,;;
.'l y!A 1
1
( ;,h
I-f A
.a
< <, : i,, I
? - h ',
3.1 h-f p' !
f a.
whsro Qt
= the total release rate of radiciodines in the gaseous effluent; uCi/sec.
(X/Q)d
= 1.0E-05 sec/m3; the highest calculated annual average depleted concentration for an area at or beyond the site boundary for all vent releases (NE boundary).
Pi
= inhalation dose factor for child thyroid for radioicdines mrem-m3/uCl-yr; 1.62E07 for I-131 and 3.85E06 for I-133 Ai
= the fraction of the total radioactivity in the 3
gaseous effluent (iodine channel) comprised by radionuclide "i".
e s
e I
e 4
b 4
' j H
i t
g s'
s s
i i l
[
/
4s VII.
BASES 4
3 Site Specific Data Note 1:
Liquid dose factors, Al, for section III.A were developed using the following site specific data.
The liquid pathways involved are drinking water
' 9_
and fish.
The maximum exposed individual is an adult.
r Ai7 =
(Uw/Dw + UF x BFi) KO x DFi 730 liters per year; maximum adult usage of
[,
Uw
=
drinking water (Reg. Guide 1.109, Table 3-5).
TQ in i
- .[E -
85; average annual dilution at Phoenixville Water Dw
=
a Authority intake.
r.
21 kg per year; maximum adult usage of fish (Reg.
UF
=
Guide 1.109, Table E-5).
BFi =
bicaccumulation factor for nuclide, i, in fresh-
]
water fish.
Reg. Guide 1.109, Table A-1, except j
P-32 which uses a value of 3.0E03 pCi/kg per pCi/ liter.
I 1.14E05 (lE06pCi/uCi)(lE03 ml/Kg)8760 hr/yr KO
=
j units conversion factor.
j DFi =
dose conversion factor for nuclide, i, for adults in total body or bone, as applicable.
Reg. Guide s
1.109, Table E-ll, except P-32, bone which uses a value of 3.0E-05 mrem /pCi ingested.
^
The data for D was taken from data published in Limerick i
Generating Station Units 1 and 2 Environmental Report Operating License Stage, Volume 3.
All other data except j
P-32 BF and DFi were used as given in Reg. Guide 1.109, j
Revision 1, October 1977.
A P-32 BFi value was taken from i
- Kahn, B. and K.
S. Turgeon, "The Bioaccumulation Factor for Phosphorus-32 in Edible Fish Tissue", NUREG-CR-1336, March, 1980.
A P-32 DF value was taken from Limits for Intakes of Radionuclides by Workers, International Commission on Radiological Protection ICRP Publication 30, Supplement to j
Part 1, 1979.
Note 2:
To develop constant P(I-131) for Section III.A, the following data'were used:
i P(I-131)
K' (BR) (DFA)
=
U 10E06 pCi/uCi; unit conversion factor K'
=
3700 m3/yr; child's inhalation rate.
BR
=
4.39E-03 mrem /pCi; the thyroid inhalation J
9 DFA
=
I-131 dose factor for I-131 in the child.
1 I 4
L_ _.._ _ _
Tho pathway in tha inholation pnthway for a child.
All valucs are teken from Rtgulatory Guido 1.109, Ravision 1, Octobsr 1977.
Note 3:
To develop constant R for section III.C, the following site specific data were used:
RGi (D/Q) = K'QF (Uap) (Fm)(r) (DFLi)a fp(1-fs)(exp-itf) j.
A i+Aw Yp lE06pCi/uCi unit conversion factor K'
=
6Kg/ day; goat's consumption rate I
QF
=
Uap =
330 1/yr; yearly milk consumption by an infant
',)(i (9.97E-07)/sec decay constant for I-131;
=
9.48E-06 for I-133.
f h00 (5.73E-07)/sec decay constant for removal
=
of activity in leaf and plant surfaces.
(6.0E-02) day / liter, the stable element Fm
=
l transfer coefficient for I-131.
r
,=
1.0 fraction of deposited radiciodine retained in goat's feed grass.
DFLi=
(1.39E-02) mrem /pCi - the thyroid ingestion dose factor for I-131 in the infant; 3.31E-03 mrem /
pCi for I-133.
0.75; the fraction of the year the goat is on fp
=
pasture (average of all farms)".
0.0; the fraction of goat feed that is stored fs
=
feed while the goat is on pasture (average of all farms).
t I
Yp 0.7 Kg/m2 - the agricultural productivity of
=
pasture feed grass.
2 days - the transport time from pasture to goat, t
=
f to milk, to receptor.
The pathway is the grass-goat milk ingestion pathway.
These data were derived from data published in Limerick Generating Station Units 1 and 2 Environmental Report Operating Stage, Volume 3.
All other data were used as given in Reg. Guide 1.109, Revision 1, October 1977.
Similar data were used to develop the constant R for I-133.
I Note 4:
The methodology described herein will be implemented i
via computer codes.
These codes have been verified as documented in:
1.
G.A. Technologies, RM-21A Computational Models, Document No. E-ll5-1241, June 1984..
2.
G. A. Technologies, Mateorological Monitoring, Dinplay and Reporting System /RM-21A, Document No.
0375-9032, January, 1984.
' Surveillance Requirement 4.11.1.2 Liquid Pathway Dose Calculations The equations for calculating the doses due to the actual release rates of radioactive materials in liquid effluents were developed from the. methodology provided in Regulatory Guide 1.109,
" Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFRPart 50, Appendix I",
Revision 1, October 1977 and NUREG-0133 " Preparation of Radiological Effluent Technical Spscifications for Nuclear Power Plants". October 1978.
~
I Surveillance Requirement 4.11.2.1.1 and 4.11.2.1.2 - Dose Noble Gases The equations for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents were developed from the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I",
Revision 1, October 1977, NUREG-0133 " Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants", October 1978, and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases l
from Light-Water-Cooled Reactors," Revision 1, July 1977 with site specific dispersion curves and disperion methodology.
The specified equations provide for determining the air doses in areas at and beyond the SITE BOUNDARY based upbn the historical average atmospheric conditions.
The dose due to a noble gas release as calculated by the Gross Release Method is much more conservative than the dose calculated by the Isotopic Analysis Method.
Assuming the release rates given in Limerick Generating Station Units 1 and 2 Environmental Report Operating License Stage, Volume 3, the values calculated by the Gross Release Method for total body dose rate and skin dose rate are 4.8 times and 3.25 times, respectively, the values calculated by the Isotopic Analysis Method.
For the Gross Release Method, Kr-87 and Kr-88 are used for the limiting skin and total body dose factors respectively, due to half life considerations.
Kr-89, the nuclide with the highest dose factors per Regulatory Guide 1.109 Table B-1 has a half-life of 3.2 minutes while the half-lives of Kr-87 and Kr-88 are 76 r
minutes and 2.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> respectively.
Therefore, by the time that gaseous effluents have been transported offsite, Kr-89 will have j
dacayedenoughsothatKr-87andKr-8gareeffectivelythemost limiting nuclides.
l The model Technical Specification LCO for all radionuclides and radioactive materials in particulate form and etdionuclides other than noble gases requires that the instantaneous dose rate be less than the equivalent of 1500 mrem per year.
For the purpose i c
of calculating this instantaneous doso rate, thyroid dosa from icdins-131 and iodine-133 through tho inhalation pathway will be used.
Since the expected annual releases pr.esented its LGS FSAR Table 11.3-1 indicate that iodine-131 and iodine-133 releases have the major dose impact this approach is appropriate.
The value calculated is multiplied by 1.02 to account for the thyroid dose from all other nuclides.
This allows for expedited analysis and calculation of compliance with the LCO.
Surveillance Requirement 4.11.2.2 and 4.11.2.3 - Dose Noble Gases The equations for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents were developed from the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I",
Revision 1, October 1977, NUREG-0133 " Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants", October 1978; and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors", Revision 1, July 197E with site specific dispersion curves and dispersicn methodology. 'The i
specified equations provide for determining the air doses in areas at and beyond the SITE BOUNDARY based upon the historical average atmospheric conditions.
1 MEMBERS OF THE PUBLIC, who may at times be within the SITE BOUNDARY, will be subject to lower annual average concentr'atiotis' than those calculated at the SITE BOUNDARY.
The maximum expected \\
occupancy factor is a working year (or 25% of the year) along the railroad tracks.
The maximum chi /q along the railroad tracks is 2.09 E-06 in the West sector.
Both chi /q an'd the occupancy factor are lower for this case than the NE sector SITE BOUNDARY.
The dose due to noble gas releases as calculated by the Gross Release Method is much more conservative than the dcse calculated by the Isotopic Analysis Method.
Assuming the release ra'tes given in Limerick Generating Station Units 2 and 3 Envidonmental Report Operating License Stage, Volume 3, the values calculated 3
by the Gross Release Method for total body dose rate and skin dose rate are 4.8 times and 3.7 times, respectively, the values calculated by the Isotopic Analysis Method.
Dose, Iodine-131, Tritium, and Radioactive Material in Particulate Form The equations for calculating the doses due to the actual release rates of radiciodines, radioactive material in particulate form,
(
and radionuclides other than noble gases with half-lives greater 4
than 8 days were developed using the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I",
Revision 1, October 1977; NUREG-0133, " Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants",
1 October 1978; and Regulatory Guide 1.111, " Methods for Estimating, Atmospheric Transport and Dispersion of Gaseous Effluends in ENGR.D H P. I AAA C_ _
t Y,/4@
DATE
'l
s s
.x \\ ?
w e
+
.A s
s Routine ReleasetJ, frm Light-Water-Cooled Reactors", Revisicm 1, July 1977 with site specific dispersion curves and dispersicm method 61ogy. '1hese equations provide for determining the actual doses based upon the historimi average
,e ctmospheric conditions, t
MENBERS OF,'IHE PUBLIC, who may at times be within the SITE BOUNDARY, will 1
be subject to lower annual average concentrations than those calmlated
{
at the SITE BOUNDARY. N ma;cimum expected m eancy factor is a working 4
g" year (or 25% of the calendar year) along the railroad tracks. The maximm depleted chi /q alc..g the railroad tracks is 1.97 E-06 in the West sector.
% Both depleted chi /q and the cecupancy factor are lower for this case than the NE sector SITE BOUNDARY.
- \\
Canp[ lance with tle' 10 TR 50 limits for radiciodines, radioactive materials t-in particulate fora and radionuclides other than noble gases with half lives gr6ater than eight days is to be determined by calculating the thyroid dose frm iodine-131 and iodine-133 releases. Since the iodine-131 and iodive-13~1 dose accounts for 99.97 percent of the total dose to the thyroid, the value calculated is not increased.
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MAP DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADIOACTIVE GASEOUS AND LIQUlO EFFLUENTS "I
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November 22, 1985 l
Docket No. 50-352 y'i Admini strator
~
Dr.. Thomas E. Murley, 3
, Region I U.S. Nuclear Regulatory Commission
'631; Park Avenue
' King of Prussia, PA 19406 J
~
SUBJECT:
Correction to our November 13, 1985 letter submitting a revision to the Sem1-Annuni Ef fluent Release Report r
s
No. 2 for Limerick Generating Station
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J
Dear Dr. Murley:
Y This letter is being submitted to correct an error, due 3
to.an; administrative oversight, in that the correct Dodket Number
.*W' was not provided on our transmittal letter for a revision to the "i,
Semi-Annual Releases Recort No.2 for the period January 1, 1985 lthrough June 30, 1985 for Limerick Generating Station, our 4W letter dated November 13, 1985, which submitted the revision, m
ui >
iincorrectly referrenced Docket Nos. 50-277 and 50-278 (Peach ff.'Q,
.Botton Units 2 and 3) and is being corrected to Docket No, 50-352 f
]/[/l(LimerickGeneratingStation).
wfCl Additionally, this lj ter
' (correet's. the designated Site Inspector to receive a rhoncjpy
..}
jgjgf of sthe ; revision from 'the Peach Bottoe to the Limeric Site
'W
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Inspeotor.
wsm p-s MP,
As a matter of completeness, enclosed is a ; corrected
- m., 3 copy: of the November 13, 19R5 letter complete with its original
- R%:.: '
attachments.
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If you have any questions, planse do not hesitate to contact us.
n; 1
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Engineer-in-Charge Licensing Section 7,
Nuclear Generation Di. vision 9
closure
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censDocument Control Desk U.S. Nuclear Regulatory Commission LWashington, DC 20555 w,
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LE.-M. Kelly, Senior Site Inspector
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'See Service List' i
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l-s PHILADELPHIA ELECTRIC COMPANY 2301 MARKET STREET f
P.O. BOX 8699 PHILADELPHIA. PA.191o1 (21si e41-4ooo November 13,.1985 j
Docket No. 50-352 l
i Dr. Thomas E. Murley, Administrator j
Region I j
U.S. Nuclear Regulatory Commission 631 Park Avenue i
King of Prussia, PA 19406 1
i
SUBJECT:
Revision to Semi-Annual Effluent Releases Report No. 2 January 1, 1985 through June 30, 1985 Limerick Generating Station Dear Dr. Murley Enclosed is a revised " Liquid Ef fluent-Summation of all Releases" page to the Semi-Annual Ef fluent Releases Report. 2 for e
Limerick Generating Station Unit 1.
The Semi-Annual Effluent Releases Report was submitted on August 29, 1985 in compliance with the Annual Reporting requirements of Technical Specification 6.9.1.8 of Operating License NPF-27 in accordance with Regulatory Guide 1.21 and Regulatory Guide 10.1.
t A revised page is being submitted to correct errors, due to an administrative oversight, in the values for the " Average Diluted Concentration During Period" line for the first and second quarters of 1985 as presented under Sections A.2 and B.2.
i Limerick Generating Station uses an in-house computer system to correlate and compute certain of the data submitted in the Semi-Annual Ef fluent Release Reports.
At the time of the original submittal of the Semi-Annual Effluent Release Report No.
2 for Limerick, the, computer program capability was limited to exponents to the base ten (10) of less than or equal to nine (09), i.e. E+00 through E+09.
For the original submittal, the computer expressed the data under Section F of this page with exponents of E+01 when the actual exponents should have been E+11.
(Examples the data for Section F for first and second quarters as expressed by the computer were 0.523E+01 and 0.141E+01 respectively rather than 0.523E+11 and 0.141E+11 respectively).
This progam limitation was recognizied and the exponents for the data under Section F were changed to E+11 by hand before the Semi-Annual Ef fluent Release Report No. 2 was submitted.
k' s
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N-N+h-
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November 13, 1985 Page 2
)
Subsequent to the submittal of Report No. 2, it was realized that the computer utilizied its values in Section F to 1
computo the values in Sections A.2 and B.2 (i.e. A.1/F = A.2 and B.1/F = B.2) and had used an exponent of E+01 instead of E+11 in *
/
order to compute the values for Sections A.2 and B.2.
The values 7
in Sections A.2 and B.2 have been changed on this revised page to provide the proper data.
L To pr'eclude a future occurrence, the program for the in-1 l
house computer has been changed to accept values with exponents up to E+99.
The corrected data is indicated by a vertical bar in the margin of the revised page.
If you have any questions, please do not hesitate to contact us.
Very truly yours, W. M. Alden Engineer-In-Charge Licensing Section Attachments i
cc:
Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C.
20555 E. M.
Kelly, Senior Site Inspector t
i See Service List i, j l
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lI
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cc:
Troy B. Conner, Jr., Esq.
Ann P. Hodgdon, Esq.
Mr. Frank R. Romano Mr. Robert L.. Anthony Ms. Phyllis Zitzer a
Charles W. Elliott, Esq.
Zori G.
Forkin, Esq.
Mr. Thomas Gerusky g
Director, Penna. Emergency Management Agency Angus Love, Esq.
f David Warsan, Esq.
Robert J. Sugarman, Esq.
Kathryn S. Lewis, Esq.
Spence W. Perry, Esq.
Jay M. Gutierrez, Esq.
Atomic Safety & Licensing Appeal Board i
i Atomic Safety & Licensing (Board Panel Docket & Service Section 3 Copies) l E. M. Kelly 5
Timothy R. S. Campbell 4
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September, 1985
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SITES LIMERICE UNITS Ut s
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8/I4/85 I3sR4 i
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EFFLUENT AMD WASTE DISPOSAL REPORT
~~
LidOID EFFLUENTS - SUMMATION OF ALL RELEASES I
i di V
O a UNITS a. QUARTER OUARTER sEST. TOTAL:
e s
I s
E
- ERROR. k A. FISSION AND ACTIUATION PRODUCTS i'
- 1. TOTAL RELEASE (EXCL.s CI
- 0;449E-05 s 0.373E-03 s e.230E+et :.
8 TRIT. CASES, ALPH4):
s s
e i
i i
8 e 2. AVERACE DILUTED -
IUCI/ML a 0.858E-13 i8 0.254E-10 8
CONC. DURING PERIOD s
e a 3. PERCENT OF 3
s 0.'estE+0e a e.90eE+ee s s
APPLICABLE LIMIT a
~
- 3. TRITIUM l
8 1. TOTAL RELEASE
- CI IE352E-et e.57eE-01 : e.30eE+et I 0.673E-09's0.404E-04'\\
s E. AUERACE DILUTED UCI/ML :
8 i
CONC. DURING PERIeb :
s 3. PERCENT OF s
2 3't.eetE+0e e.000E+04 :
j s
APPLICABLE LIMIT s
s 8
. s J
1 C. DISSOLbED AND ENTRAINED CASES I 1. TOTAL REttASE
- CI
- 6.cetE+0e : e.teeE+00
- 4. tete +04 8 l
i
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s E. AUERAGE DILUTED UCI/RL s 9.eetE+0e : 0.000E+0e e 8
CONC. DURING PERIOD 8 8
a 1
s 3. PERCENT OF 3
R s e.SeeE+et 9.eetE+0e a I
APPLICA8tE LIMIT e
e a
a
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D. GROSS ALPHA RADIDACTIVITV i
2-l
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- e 1. TOTAL RELEASE
- CI e.eeeE+et s.t.3 Set.e4 s e.T3eE+et s i,
l E. U0 Lune UASTE RELEASES sLITERS e e.116t+es e 0 503C+et : e.IceE488 s l
8 (PRIOR TO DILUTIONS e
e s '
s a
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F. UOLUIE DILUTION UATER sLITERS 8 s e.IceE403 e
! ;...._E_.t M E nt PERIOD 8,_
{ 0*523s+11s 0*141x+11 s
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1 November 13, 1985 1
Docket No. 50-277 50-278
(
t
-i Dr. Thomas E. Murley, Adninistrator Region I U.S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, PA 19406 SUEJECT:
Revi sion to Semi-Annual Ef f 3uent Releaser Report No. 2 January 1, 1985 through Junc 30, 1985 Limerick Generating Station
Dear Dr. Murley:
~
4-Enclosed is a revised " Liquid Effluent-Summation of all Releases" page to the Semi-Annual Ef fluent Releases Report. 2 for
' Limerick Generatina Station Unit 1.
The Semi-Annual Effluent Releasen Report was submitted on August 29, 1985 in compliance
'with the Annual Reportino requirements of Technical Specification 6.9.1.8 of Operatino License NPF-27 in accordance with Regulatory l Guide 1.71 and Regulatory Guide 10.1.
A revised page i s being submitted to correct errors, due GF
- to an administrative oversicht, in the values for the "Averag'e
') 4 -
Diluted Co tcentration During Period" line for the first and i g* -U, second ouarters of 1985 as presented under Sections A.2 and B' 2.
4
- " '[',
Limerick Generating Station uses an in-house computer system to correlate and compute certain of the data submitted in
-f' the Semi-Annual Effluent Release Reports.
At the time of the original submittal of the Semi-Annual Ef fluent Release Peport No.
2 for Limerick, tho' conputer program capability was limited to exponents to the base ten (10) of less than or equal to nine (09), i.e. E+00 throuch F+09.
For the original submittal, the computer expressed the data under Section F of thin page with 5
exponents of E+01 when the actual exponents should have been
.E+11.
(Example the data for Section F for first and second quarters as expressed by the computer were 0.523E+01 and
- . y_
0.141E+01 respectively rather than 0.523E+11 and 0.141E+11
_ respectively).
This progam limitation wan recognizied and the so "exponento for the data under Section F were changed to E+11 by hand before the Semi-Annual Effluent Release Report No. 2 was f'
j'i su mitted.
1L
- pl?H 50(
[
yp -
t _
November 13, 1985 Page 2 Subsecuent to the submittal of Report No. 2, it was realized that the computer utilizied its values in Section F to '
A.1/F = A.2 an[n !
d compute the values in Sections A.2 and D.2 (i.e.
E+11 B.1/P = B. 2 ) and had used an exponent of E+01 instead ofs order to compute the values for Sections A.? and D.2.
The val ues.1 in Sections A. 2 and B. ? have been chanced on this revised pace 'to.
I provide the proper data.
s To preclude a future occurrence, the program for the ind house computer has been chanqed to accept values with exponents up to E+99.
t' The corrected data is indicated by a vertical bar in th margin of the revised page.
If vou have any questions, please do not hcsitate to contact us.
Very truly yours, W.
M.
Alden i
Engineer-In-Charge Licensino Section
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Attachments x
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" !#1 U.S. Nuclear Reculatory Commission it m
. "' b l Washington, D.r.
20555 fi i[
s T.
P.
Johnson, Site Inspector
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See Service List l;
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UNITS us USER: MsJtT DATES 3/14/85 13:24
't EFFLUENT AND WASTE DISPOSAL REPORT e f,
LIQUID EFFLUENTS - SUMMATION OF ALL RELEASES UNITS
. QUARTER CUARTER sEST.
TOTALS l
1 3
'2
- ERROR. 4:
A. FISSION AND ACTIUATION PRODUCTS
- 1. TOTAL RELEASE (EXCL.s CI 0.449E-95 0.37X-93 8 e.230E+42 TRIT., CASES, ALPHA) s 8
8 t 2. AVERACE DILUTED UCI/ML s 0.assE-13. : 0.264E-10 CONC. DURING PERIOD :
s
's s
3 3. PERCENT OF x
- 9.'estE+00 s 8.000E+44 s 3
APPLICABLE LIMIT s
a s
- 3. TRITIUM
~
s 1. TOTAL RELEASE
- CI
- 0.3SEE-41 0.S74E-01 : e.300E+42 8 0.404E-04'ls
- 2. AVERACE DILUTED sWCI/ML : 0.673E- 09 CONC. DURING PERIOD s a
s
- 3. PERCENT OF 2
3't.990E+0e : 0.940E+94 :
("
APPLICABLE LIMIT s
e C. DISSOLVED AND ENTRAINED CASES
.E.
s..
E.
s..;.E+.. s ii.701AtREtEA.E 3 CI
- 2. AVERAGE DILUTED SUCI/ML
- 9.040E+et 0.096E+00 s
CONC. DURING PERIOD s s
- 3. PERCENT OF 4
9.000E+00 8 0.949E+00 s g%7 APPLICASLE LIMIT 3
to
- 9. GROSS ALPHA RADIDACTIVITY 4.354E-04s4.730Ehtas 3
s 1. TOTAL RELEASE
- CI
- 9.000E+0e :
l e
C. UOLUME WASTE RELEASED sLITER$ e 0.11SE+43 0.59X+06 : G.100E493 :
c-3 (PR100t TO DILUTIOM) s s ~
s e
p 8
8 O*IOOE*0I I F. UOLDIE DILUTION UATER SLITER$ e 0.523E+11 0.141E+11 s USED DURING PERIOD s
-s s
e s
_