ML20082L879

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Rev 4 to ODCM
ML20082L879
Person / Time
Site: Peach Bottom  Constellation icon.png
Issue date: 06/06/1991
From:
PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:
Shared Package
ML20082L872 List:
References
PROC-910606, NUDOCS 9109040429
Download: ML20082L879 (46)
v • d • e


Text

. _ _ _ _ _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ .

i

.t Offsite Dose Calculation Manual oevision 4 Peach Bottom Atomic Po er Station Units 2 and 3 Philadelphia Electric Company Docket Nos. 50-277 L E0-278 u .-r ~

PORC Approval  : -: /kdA f( - '

.:hf l[,/'e '[

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/ Date PORC Chairman PORC Meeting # : 6// - /3,Z /; o_ /

/Date i

9109040429 910929 PDR ADOCK 05000277 R PDR

i- ' . - . -

. . . . -. = -  ;

. Table of Contents  !

}

l.. Purpose-  ;

I' '  ?-  !

11. Instrument Setpoints f l

111. _ Liquid Pathway Dose Calculations .!

t U

A. ; Liquid Radwaste Release _ Flow Rate Determination  !

B. . Surveillance Requirement 4.8.B.2 i C. Surveillance Requirement 4.8.B.4a  !

n  !

IV. Gaseous Pathway Dose Calculations -[

i A. Surveillan'ce Requirement d.8.C.1 B. Surveillance Requirement 4.8.C.2- .

C. Surveillance Requirement 4.8.C.3  !

0. Surveillance Requirement 4.8.C.5a E. Surveillance _ Requirement 4.8.C.6b

~V.- Nuclear _ Fuel Cycle Dose Assessment - 40 CFR 190 i

. A. Surveillance Requirement 4.8.0 [

t  !

VI. Calendar Year. Dose-Calculations i t

A. _ Unique Reporting Requirement 6.9.2.h i

Vll. Radiological Environmental Monitoring Program i l

A ~. Surveil' lance Requirement 4.8.E i

VI!!. Bases j t

i h

5 i

t I

2

\

l-' .

Page 1 of 44, Rev. 4 i

1. Purcose ,

The purpose of the Offsite Dose Calculation Manual is to establish methodologies and procedures for calculating, doses to individuals in areas at and beyond the SITE BOUNDARY due to radioactive effluents from Peach Bottom Atomic Power Station. The results of these calculations are required to determine compliance with Appendix A to Operating Licenses OPR-44 and DPR-56, " Technical Specification and Bases for Peach Bottom Atomic Power Station Units No. 2 and 3".

11. Setcoint Determination for Liauid L Gaseous Monitors A. Liouid Radwaste Activity Monitor Setcoint Each tank of radioactive waste is sampled prior to release. A small liquid volume of this sample is analyzed for gross gamma (well count) activity. This analysis is performed in a Na[ well counter. This well counter has a ccunting efficiency similar to the liquid radweste discharge gross activity monitor. The well counter and lig.id radwaste discharge gross activity thonitor are calibrated against the same liquid radioactivity source in the geometry to be used by each detector. Ar, efficiency is determined for each rads:aste tank to be released. Exceeding the expected response would indicate that an incorrect sample had been obtained for that release and the release is automatically stopped.

S.P. = (Net CPM /ml(well) X Eff W/RW) + Background CPS S.P. = Liquid Radwaste o~ss activity monitor setpoint in CPS Net CPM /ml(well) = gross gamma activity for the radwaste sample tank determined by the well counter.

Eff W/RW = conversion factor between well counter and liquid radwaste gross activity monitor (CPS (R/W monitor) - CPM /ml(well)).

Background CPS = Background reading of the liquid radwaste gross activity monitor (CPS).

The alarm and trip pot setcoints for the liquid radsaste activity monitor are cetermined from a calibration curve for the alarm pot and trip pot. The alarm pot setting includes a factor of 1.25 to allow for analysis error, pot setting error, instrument error and calibration error. The trip pot setting includes a factor of 1.35 to i allow for analysis error, pot setting error, instrument error and calibration error. The flow rate determination includes a margin of assurance which includes consideration of these errors such that the instantaneous release limit cf 10 CFR 20 is not exceeced.

. Page 2 of 44, Rev. 4 -

B. Liouid Radwaste Release Flowrate Setpoint Determination The trip' pot setpoint for the liquid radsaste release flowrate is determined b'y multiplying the liquid radweste flowrate determined above by 1.2 and using this value on the appropriate calibration curve for the discharge flow meter to be used. The Peach Bottom redwaste ,

system has two flow monitors (high flow (5 to 300 gpm) and low flow .

(0.8 to 15 gpm).- The factor of 1.2 allows for pot setting error and instrument error. The flow rate determinstion includes a margin of assurance which includes consideration of this error such that the instantaneous release limit of 10 CFR 20 is not exceeded.

C. Setooint Determination for Gaseous Radwaste The high and high-high alarm setpoints for the main stack radiation monitor, Unit 2 roof vent radiation monitor and Unit 3 roof vent radiation monitor are determined as follows:

Hioh Alarm - the high alarm setpoint is set at approximately 3 x the normal monitor reading.

Hich-Hioh Alarm - the high-high alarm setpoint-is set at a release rate from this vent of.approximately 30% of the instantaneous release limit of 10 CFR 20 as specified in-Technical Specification 3.8.C.l.a for the most restrictive case.(skin or total body)onanunidentifiedbasis.

To determine these setpoints, solve the gaseous effluent dose rate equations in section IV.A of the ODCM to determine what main stack release rate and roof _ vent release rate will produce a dose rate of 150 mrem /yr to the total-body and a dose rate of 900 mrem /yr to theskin(307.ofthelimitof n 3000 mrem /yr)fromeachrelease point. Using the smallest (most restrictive) release rate for each release point determine l monitor response required to I produce this release rate l assuming a normal vent flow l rate and pressure correction '

l factor. Set the high-high alarm for approximately this monitor response.

. Dage 3 of 44. Rev, 4 D. Setooint Determination for Gaseous Radcaste l

Flow Monitors lhe alarm setpoints for the main stack flow fronitor is as follows:

Low Flow Alarm - 10,000 CFM. - This se', ling insures that the main stack minimum dilution flow as specified in Technical Specification 3.8.C.4.a is maintained.

The alarm setpoints for the roof vent flow monitors are as follows:

5 Low Flow Alarm - 1.5 x 10 cfm High Flow Alarm - 5,4 x 100 cfm 111. Licuid Pathway Dose Calculations A. Liouid Radwaste Release Flow Rate Determination Peac'h Bottom Atomic Power Station Units 2 and 3 have one common discharge point for licuid releases. 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 after dilution shall be limited to the concentration specified in 10 CFR 20.106(a) for radionuclides other than noble gases and 2E-4 uCi/ml total activity concentration for all noble gases as specified in Technical Specification 3.8.B.1. Each tank of radioactive waste is sampled prior to release and is quantitatively analyzed for identifiable gamma emitters as specified in Table t S.1 of the Technical Specifications. From this gamma isotopic analysis the maximum permissible release flow rate is determined as follows:

Determine a Dilution Factor bn Dilution Factor = [ uCi/ml i

~

MPC1 i

uCi/m'. i = the activity of each identified gamma emitter in uCi/ml MPCi = The MFC specified in 10 CFR 20, Appendix B, Table II, Column 2 for radionuglides other than noble gases or 2 x 10" uCi/ml for noble gases.

l

- g 4 7g g ,

Determine the Maximum Permissible Release Rate with this Dilution f actor by:

Release Rate (gpm) = A x 2.0 x 10 5 l B x C x Dilution factor

! A = The number of circulating water pumps running which will provide dilution ,

2.0 X 105 = the flow rate in gpm for each circulating water pump running B = margin of assurance which includes consideration of the maximum error in the activity setpoint, the maximum error in the flow setpoint, and possible loss of 5 out of the 6 possible circulating water pumps during a release. The value used for B is 10.0.

C = concentration gradient factor. The value used for C is 5.0 for discharge canal water levels less than 104' and 3.0 for canal water levels greater than 104'.

B. Surveillance Requirement 4.8.B.2 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 those appropriate radionuclides listed in Table Ill.A.I. These radionuclide account for virtually 100 percent of the total body dose and organ dose from liquid effluents.

The dose for each age group and each organ should be calculated to determine the maximum total body dose and organ dose for each quarter and the year, as appropriate. Cumulative dose files for quarterly and yearly doses should be maintained separately and the maximum total body and organ dose reported in each case.

. Page 5 of 44, Rev. 4 -'

D =- AT A' t C f

,i,i 1=1 -1 il 1 where:

0 -Thecumulativedosecombitmenttothetotal body or any organ, from liquid effluents for the tott.: time pariod m , in mrem E dt 1=1 1 At = The length of the Ith time period over which.

1 Cj) and F1 are averaged for the liquid release, in hours.

C =.The average _ concentration of radionuclide, i, 11- in undiluted liquid effluent during time period

A t from any liquid release (determined by l the effluent sampling analysis program, Technical _SpecificationTable4.8.1),inuCi/ml.

A = The site related ingestion dose commitment i factor to the total body or organ, T , for 4 each radionuclide listed in Table III.A.1, in mrem-m1 per br-uCi. See Site Specific Data.**

F = The near field average dilution factor for 1 Cj] 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 Conowingo Pond.

III.C Surveillance Requirement 4.8.B.4a

, Projected dose contributions from liquid effluents shall be calculated using the methodology cescribed in section Ill.B.

l

[

    • See Note 1-in Bases 5

l l

, - - - , , -,,w--- ~ w v a n -- -,._v

' Page 6 of 44, Rev. 4 I TABLE III.A.1 i

! LIQUID EFFLUENT INGESTION DOSE FACTORS (DECAY CORRECTED)

A j Cl DOSE FACTOR (MREM-ML PER HR-UCI)

RADIO! TOTAL BODY NUCLIDE ADULT TEEH CHILD H-3 2.13E+00 1.53E+09 2.70E+00 NA-24 1.65E+02 1.70E+02 1.98E+02

, P-32 5.93E+04 6.49E+04 8.33E+04 ,

i l MN-54 9.82E+02 1.00E+03 1.08E+03 FE-55 1.31E+02 1.40E+02 1.96E+02 FE-59 1.14E+03 1.17E+03 1.36E+03 I CO-58 2.59E+02 2.62E+02 3.17E+02 l

! CO-60 7.40E+02 7.48E+02 9.07E+02 I

ZH-65 3.87E+04 3.95E+04 4.16E+04 SR-89 8.83E+02 9.45E+02 1.48E+03

! SR-90 1.88E+05 1.56E+05 1.72E+05 l

1 TE-129M 2.01E+03 2.17E+03 2.79E+03 TE-131M 4.57E+02 4.81E+02 5.74E+02

TE-132 1.40E+03 1.44E+03 1.65E+03 I-131 1.86E+02 1.79E+02 2.36E+02 1-133 1.97E+01 2.03E+01 3.20E+01 CS-134 6.74E+05 3.88E+05 1.49E+05 CS-136 9.79E+04 9.15E+04 7.30E+04 CS-137 3.98E+05 2.20E+05 8.49E+04 BA-140 3.66E+01 3.62E+01 7.42E+0!

NOTE: The listed dose factors are for radionuclides that may be detected in liquid effluents and have significant dose consequences. The f actor s are decayed for one day to dccount for the time between effluent release and ingestions of fish by the maximum exposed  :

individual.

~

Page 7 of 44, Rev. 4 TABLE lll.A.1 LIQUID EFFLUENT INGE5i10N 005E FACTORS

. (DECAY CORRECTED)

AjT DOSE FACICR (MREM-ML PER HR-UCI)

RADIO LIVER NUCLIOE ADULT TEEN CHILD H-3 2.13E400 1.53E+00 2.70E+00 E NA-24 1.65E+02 1.70E402 1.98E+02 P-32 9.55E404 1.04E+05 1.01E+05 MN-54 5.15E+03 5.06E+03 4.03E+03 FE-55 5.62E+02 6.01E+02 6.33E+02 1

FE-59 2.96E+03 3.02E+03 2.73E+03 CO-58 1.16E+02 1.14E+02 1.04E+02 CO-60 3.35E+02 3.32E+02 3.07E+02 ZN-65 8.55E+04 8.46E+04 6.69E+04 SR-89 no data no data no data SR-90 no data no data no data TE-129M 4.74E+03 5.09E+03 5.02E+03 TE-131M 5.48E+02 5.77E+02 5.40E+02 TE-132 1.4SE+03 1.53E+03 1.36E+03 1-131 3.25E+02 3.32E+02 4.16E+02 I-133 6.48E+01 6.66E+01 8.45E+01 CS-134 8.25E+05 8.36E+05 7.06E+05 1

CS-136 1.36E+05 1.36E+05 1.13E+05 CS-137 6.07E+05 6.32E+05 5.75E+05 BA-140 7.00E-01 6.90E-01 1.11E+00 NOTE: The listed dose factors are for radionuclides that may be detected in liquid effluents and have significant dose conse'uences.

q The f actors are decayed for one day to account for the time between effluent release and ingestions of fish by the maximum exposed individual.

~h

.:.!, ., Pa9e 8 of 44, Rev. 4- -

l

-TABLE III.A.1:

LIQUID EFFLUENT INGESTION 00SE: FACTORS (DECAYCORRECTED) 1 A iT_DOSEFACTOR(MREM-MLPERHR-UCI) i I

RADIO- BONE  !

NUCL10E ADULT . TEEN CHILC i H-'3 no data -no data no data i I

.NA 1.65E+02 1.70E+02 1.98E+02 P-32 2.38E+05 2.58E+05 3.35E+05 l l

MN-54 no data no data no data i FE 8.12E+02 8.47E+02 1.19E+03 i FE-59 1.26E+03 1.30E+03 1.68E+03_  !

CO-58 no data no data no' data l' C0-60 no o'ata- no data no data- f ZN-65 2.69E+04 2.43E+04 2.51E+04 SR 3.08E404 3.30E+04 5.19E+04  ;

I

-SR 7.67E+05 6.31E+05 6.78E+05 .j TE-129M 1,27E+04 1.37E+04 1.80E+04 i

TE-131M 1.12E+03 1.21E+03 1.56E+03  !

TE-132' 2.29E+03 2.42E+03 3.07E+03

'I-131 2.28E+02 2.38E+02' 4,13E+02' [

i

.I-133 3.72E+01 3.92E+01 6.84E+01 j CS-134 3.47E+05 3.55E+05 4.30E+05 j CS-136 3.45E+04 3.46E+04 4.10E+04 CS-137 4.44E+05 - 4.75E+05 6.01E+05 BA-140 'E.57E+02 5.63Et02 1.27E+03 i NOTE: The listed dose factors are for-radionuclides that may be detected in liquid efflu,ents and have significant dose consequences. The factors are dedayed _for.one day to account for.the time between i effluent release and in9estions of fish by the maximum exposed }

individual. l i

I t

I f

t

~_ - _ _ _ _ _ _ . . . _ , _ _ , _ . _ _ . . _ . - - _ . . _ . . .- _ - _ . ,, .

.7_.

' ~

.. To .= " Page_9 of 44, Rev. 4' 4-TABLE III.A.1-

. LIQUID EFFLUENT INGESTION DOSE FACTORS

-(DECAYCORRECTED)

A.jlI DOSE FACTOR (MREM-ML PER HR UCI).

RADIO - 'KIONEY NUCLIOE ADULT TEEN CHILD H-3 2.13E+00 1.53E+0.0 2.70E+00 NA-24 1.65E+02 1.70E+02 1.98E+02 ,

P-32 no data no data no data MN-54 1.53E+03 1.51E+03 -1.13E+03 FE-55 no data no data no data 1 FE-59 no data no data no data -

CO-58 no data- no data- no data C0-60 no data no data rio data ZN-65 5.72E+04 5.41E+04 4.22E+04 i SR-89 no data no data no-data SR-90: no data no data no data .

-TE-129M 5.31E+04 5.74E+04 5.29E+04

TE.131M 5.55E+03 6.01E+03 5.22E+03 TE-132 1.43E+04 '1.47E+04 1.27E+04 I-131. 5.57E+02- 5.73E+02 ~ 6.82E+02 l 1-133 1.12E+02 1.16E+02 1.41E+02

~

CS-134 2.67E+05 2.66E+05- 2.19E+05-CS-136 7.57E+04 7.42E+04 6.00E+04 CS-137. 2.06E+05 2,15Es05 1.87E+05 BA-140 2.38E-01 2.34E-01 3.62E-01 NOTE: The listed dose factors are for radionuclides that may be detected  ;

in liquid, effluents and have significant dose consequences. The factors are decayed for one day to account for the time between effluent release and ingestions of fish by the maximum exposed I

individual.

i i

i

i Page 10 of 44, Rev. 4 TABLE Ill.A.1 LIQU10 EFFLUENT INGESTION 00SE FACTORS (DECAY CORRECTEDI A $ 7I DOSE FACTOR (MREM-ML PER HR-UCI)

RADIO-GI-LLI NUCLICE ADULT TEEN CHILD H-3 2.13E+00 1.53E400 2.70E+00 NA-24 1.65E+02 1.70E+02 1.98E+02 P-32 1.73E+05 1.41E+05 5.9BE+04 MN-54 1.58E404 1.04E+04 3.38E+03 FE-55 3.22E+02 2.60E+02 1.17E402 FE-59 9.90E+03 7.15E+03 2.84E+03 CO.58 2.35E+03 1.56Et03 6.04E402 CO-60 6.30E+03 4.33E+03 1.70E+03 ZN-65 5.38E+04 3.58E+04 1.18E+04 SR-89 4.94E+03 3.93E+03 2.01E+03 SR-90 2.22E+04 1.77E+04 9.13E+03 TE-129M 6.40E+04 5.15E+04 2.19E+04 TE-131M 5.44E+04 4.53E+04 2.19E+04 TE-132 7.02E+04 4.85E+04 1.37E404 I-131 8.5BE+0i 6.57E+01 3.70E+01 1-133 5.82E+01 5.03E+01 3.40E+01 CS-134 1.44E+04 1.04E+04 3.80E+03 i

! CS-136 1.55E+04 1.09E+04 3.96E+03 CS-137 1.lBE+04 9.00E+03 3.60E+03 BA-140 1.15E+03 S 69E+02 6.43E+02 NOTE: The listed dose factors are for radionuclides that may be detected in liquid effluents and have significant dose consequences. The factors are decayed for one day to account for the time between effluent release and ingestions of fish by the maximum exposed individual.

_. . . .. ._.m. _ _ _ . _ . _ _ - _ - _ _ . . _ . _

~

Page 11 of 44, Rev. 4

-IV. Gaseous Pathway Dose Calculation _s_

A. Surveillance Recuirement 4.8.C.1 The dose rate in areas at and beyond the SITE BOUNDARY due to radioactive materials released in caseous effluents shall be determined by the expressions below:!

1. Noble Gases:

The dose rate from radioactive noble gas releases shall be determined by either of two methods. Method (a), 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. Method (b), the_-

Isotopic-Analysis Method, utilizes the results of noble _ gas analyses required by specification 4.8.C.la.

I 1

For_ normal operations, it is expected that method (a) will be used, i L' However, if noble gas releases are close to the limits as calculated by method (a), method (b) can be used to allow more operating flex,ibility by using data that more accurately reflect actual releases.

a'. Gross Release Method D =Vh +K(X/Q) h

-TB NS V NV I

D. = (L (X/Q) +1.18)h +(L+1.lM)(X/Q)h s s NS _. V NV where:

The location is the site boundary, 1097m SSE from the vents. This l location results in the highest calculated dose to an individual from noble gas releases.

D = total body dose rate, in mrem /yr.

, -TB l

!- D = skin dose rate, in mrem /yr.

I s V = 4.72 X 10-4 mrem /yr per uCi/sec; the constant for Kr-88 accounting for the gamma radiation from the elevated finite plume. This constant was developed ustrg MARE program with plant specific inputs for PBAPS.

h = The gross release rate cf noble gases from the NS stack determined by gross activity stack-monitors averaged over one hour, in uCi/sec.

l

ci, .

Page 12 of 44, Rev. 4 -

. c a

K = 1.47~X 104 mrem /yr per uCi/m 3

the total bpdy dose factor due to gamma emissions for Kr-88 (Reg. Guide 1.109, Table B-1). ,

f

= 5.33 X 10~7 sec/m 3

the highest calculated I (Y7Q) y annual average relative concentration for any >

area at or beyond the SITE B0VHDARY for all vent releases.

h = The gross release rate of noble gases in gaseous NV effluents from vent releases determined by gross activity vent monitors averaged.over one hour,

. in'uCi/sec.

s 3 3 L = 9.73 X 10 mrem /yr per uCi/m ; the skin dese.

factor due-to-beta emissions for Kr-87. (Reg.

Guide 1.109, Table B-1). ,

-(X/Q) = 9.97 X 10-8 sec/m 3

the highest calculated s' annual average relative concentration from tha stack releases for any area at or beyond the SITE B0UNDARY.

B = 1.74 X 10~4 mrad /yr per uCi/sec; the constant for-Kr-87 accounting for the gamma radiation from

-the. elevated finite plume. This-constant was-developed using MARE program with plant specific inputs for-PBAPS.

3 M =.6.17 X 103 mrad /yr per uCi/m  ; the air dose i factor due-to gamma emissions for Kr-87. 3' (Reg. Guide 1.109,. Table B-1).

b. Isotopic Analysis Method' L 1

0 = )_ ( V - h +K (X/Q) h )-

TB -- i i is i v iv

_, i 0 =). '(L(i7Q) + -.18- ) h + (L- + 1.lM ) (i76) - (h - )

s. i

.i s i is i i V iv j where: ,

I The location is the site boundary, 1097m SSE from the vents. This

- location results in the highest calculated dose to an individual from noble _ gas releases. I i 0 = total body dose rate, in mrem /yr. j

. 'TB  !

L I

r 3

9 v _ _ - - . _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ , - _ _ - - - -

, , --...m., , _ _ _ . , , . . _ , ~ . . . , , - -

. - . ~ . - . - - . . - . - - . . . - . . - .--- ._, _ . -- . . -

' ^ ^

g L

  • Page 13 of 44, Rev.-4 th,_'-

o .

-e +

+

0  :=skindos'e.inmrem/yr,.

! S.

1 V- = The constant for each-identified noble gas p i radionuclide for the gamma radiation from

- -the elevated. finite plume. The constants I

were developed using the MARE program With

i. plant specific inputs for PBAPS. Values are listed on Table IV.A.1, in mrem /yr per

~

, uCi/sec.

1 i Q = The release rate of noble gas radionuclide,

! is i,;in gaseous effluents from the-stack

! determined by isotopic analysis averaged-

!- over one hour, in uCi/sec.

i -

[ K = The total body. dose factor due to gamma

- i emissions for each identified noble gas- -f i- radionuclide. Values are listed:en Table ,

IV.A.1, in mrem /yr per uCi/m3 .

{ ,

',' (X/Q) -= 5.33 X 10-7 sec/m3 ; the highest y calculated annual average relative' concentration

' for any area at or beyond the SITE BOUNDARY for all vent releases, t>.

! a

( Q .= The release rate of noble gas radientclide, i' iv i, in gaseous effluents from all vent

? releases determined by isotopic analysis l-

-averaged over one hour, in uCi/sec.-

$- L' = The skin dose f actor due to beta emissions i for each identified noble gas radionuclide, j . Values-are listed cn. Table IV.A.1, in i mrem /yr per uCi/m3. - ,

3

-(X/Q) = 9.97 L 10~8 sec/m  ; the highest calculated s annual average relative concentration from the stack releases for any area at or beyond the SITE BOUNDARY.

i T

  • + e- s _. <_m ..__-_m _ ___ _.

e

~

~

Page 14 of t4, Rev. 4 B

= The constant f or each identified noble gas i radionuclide accounting for the gamma

-radiation from the elevated finite plume.

The constants were developed using MARE program with plant specific inputs for PBAPS. Values are listed on Table N. A.1, e I in mrad /yr per uCi/sec.

M = The air dose factor due to gamma emissions i for each identified noble gas radionuclice.

ValueserglistedonTableIV.A.1,inmrad/yr per uCi/m 1.1 = Unit conversion, coverts air dose to sk :n dose, mrem / mrad.

i

TABLE IV.A.1 - Constants for Isotocic Analysis Method (corrected for decay during transit)

Total Plume-Bocy Skin Gamma Beta Air Bocy Plume-Air Oose Dose Air Dese Dose 00te Oose Facter Factor Factor Factor ractcr Factor B K L M N V i i i i i i (mrad /yr (mrem /yr (trem/yr (rrad/yr (mrad /yr (mrem /yr pcr per. per. per, per . per d 3 d Radionuclide uC1/sec) uCi/m;) uCi/m ) uC1/m) uCi/m ) uCi/sec)

Kr-85m c.02E-05 1.17E+03 1.46E+03 1.23E+03 1.97E403 3.76E-05 Kr-87 1.74E-04 5.92E+03 9.73E+03 6.17E+03 1.03E404 1.56E-04 Kr-89 4.90E-04 1.47E+04 2.37E+03 1.52E+04 2.93E+0a 4.72E-04 Xe-133 1.19E-05 2.94E+02 3.06E+02 3.53E+02 1.05E+03 1.11E-05 Xe-133m 1,09E-05 2.SiE+02 9.94E+02 3.27E+02 1.48E403 1.01E-DE Xe-135 6.37E-05 1.81E+03 1.86E+03 1.92E+03 2.46E+03 5.95E-05 Xe-135m 6.61E-05 2.53E+03 5.76E+02 2.72E+03 L39E+02 6.17E-05 Xe-138 1.52E-04 7.33E+03 3.43E+03 7.54E+03 3.94E+03 1.46E-04 The values K j , Lj , Mj , and N j are taken f rom Reg. Guide 1.109, Table B-1. The values Bj and Vj were developed using the MARE program with plant specific inputs for PBAPS.

J

Page 16 of 41, Rev. 4

2. lodine-131, iodine-133, tritium and radioactive materials in particulate form, other than noble caset, with half-lives greater than eight days. .

The dose rate shall be determined by either of tno methods. Method (a), the lodine-131 Method, uses tne iodine-131 releases and a '

correction f actor to calculate the dcse rate from all nuclides released. Method (b), the Isotopic Analysis Method, utilizes ali applicable nuclices.

For normal cperations, it is expected that Method (a) will be used since iodine-131 dominates the critical pathway - thyroid. However, in the event iodine-131 releases are minimal (e.g., during long term shutdown) Method (b) will be used to provide accurate calculations.

In the absence of iodine-131 releases, the lung is the critical organ.

a. Iodine-131 Method D =

(CF) P W h +W h T I $ IS V IV where:

The location is the site boundary,1097m SSE f rom the vents.

0 = dose rate to the thyroid, in mrem /yr.

T CF = 1.09; the correction factor accounting for the use of iodine-131 in lieu of all radionuclides released in gaseous effluents including iodine-133.

, P = 1.624 X 107 mrem /yr per uCi/m3 ; the dose

'. parameter for I-131 via the inhalation pathways. The dose factor is based on the critical individual organ, thyroid, and most restrictive age group, child. All v lues are from Reg. Guide 1.109 (lables E-5 and E-9).

) W = 1.03 X 10-7 sec/m3 ; the highest calculated

.I 5 annual average relative concentration for any area at or beyond the SITE ECUNDARY frem stack releases. (5SE boundary) h = The release rate of iodine-131 in gaseous 15 effluents from the stack determined by the effluent samoling and analysis program (Technical Specification Table 4.8.2) in uCi/sec.

I ,

W = a.73 x 10~7 sec/m"; the highest calculated v annual average relative concentration for arc, area at cr beyond the $1TE 80UNDARY for all vent releases (SSE boundary).

. -* Page 17 of 44, Rev.l4 -!

l

+ t i

-Q =_The release-rate.cf iodine-131 in gaseous l IV effluents from all vent releases, determined i by the effluent-sampling and analysis program  !

(TechnicalSpecificationTakle4.8.2)inuCi/sec. l l

b' . Isotopic; Analysis Method {

I 0- = P. W h + 'W h L. i S; iS- V iV {

I

-where: .i The -Iocation is the site boundary,1097m SSE from the vents. )

D . = dose rate to the lung, in mrem /yr.

L- l

-t P. . = The dose parameter for radionucli?s other

i. than n'oble gases;for the inhalation pathway, f The dose f actors are based on the critical i individual organ-lung, and most restrictive age group-child. All values are from Reg.  ;

l Guide 1.109 (Tables E-5 and E-9). Values j are '1jsted on Table IV.A.2, in mrem /yr per _

uCi/m -l 3

.W = 1.03 X 10-7 -sec/m  ; the highest calculated _!

5 annual average. relative contentration for any  !

area at or beyond the SITE 80VHDARY from I stack releases. (SSE boundary) l i

i

= The release rate of radionuclides; i, in  !

h _

is gaszous effluents from the stack determined 1 by the effluent sampling'and analysis program ii (TechnicalSpecificationTable4.8.2)in j uCi/sec.  ;

. W = 4.78 X 10-7 sec/m3 ; the highest calculated  :

y annual average relative concentration for any [

area at or beyond the SITE BOUNDARY for all  ;

. vent releases. (SSEboundary) i

'l Q = The release rate of radionuclides, i, in  ;

iv gaseous effluents from all vent releases, i determined by the effluent sampling anc j analysis program-(Technical Specification i 1 Table 4.8.2) in uCi/sec. [

i l

i c s k

L  !

TABLE IV.A.2 - CONSTANTS FOR ISOTOPIC ANALYSIS METHOD (crem/yr. per uCi/m")

] PI - Inhalation -

Radionucli'e c Luno Dose Factor 6

i Mn-54 1.5BX10 4

Cr-51 1.70x10 6

Co-58 1.11x10 Co-60 7.07x10 6 Zn-65 9.95x10 5 Sr-89 2,16x10 6 7

Sr-90 1.48x10 5

Ce-141 5.44x10 5

Cs-134 1.21x10

s-137 1.04x10 5 6

Ba-140 1.74x10 i

- Page 19 of 44, Rev. 4 IV B. Surveillance Recuirement 4.8.C.2 The air dose in areas at _and beyond the SITE SOUNDARY due to noDie gases released in gaseous effluents shall be determined by the expressicns below.

- The air dose shall be determined by either of two methods. Method (a), the Gross Release Method, assumes that all noble !;ases released are the most limiting nuclide - Kr-8B for gamma radiation and Kr-87 for beta radiation.

Method (b), the Isotopic Analysis Methcd, utilizes the results of noble gas analyses required by specification 4.8.C.la.

For normal operations, it is expected that Method (a) will be used.

However, if noble gas releases are close to the limits as calculated by Method (a), Method (b) can be used to allow more operating flexibility by using data that more accurately reflect actual releases.

1. for gamma radiation:

a) Gross Release Method Dy = 3.17 x 10-0 (M (X/Q) Q + BQ) v v s where:

The lccation is the SITE BOUNDARY 1097m SSE from the vents. This location results in the highest calculated gamma air dose from noble gas releases.

Dy = gamma air dose, in mrad.

3.17 x 10~a = years per second.

M = 1.52 x 104 mrad /yr per LCi/m 3

the air dose factor due to gamma emissions for Kr-88. (Reg. Guide 1.109, Table B-1)

(5) = 5.33 x 10-7 sec/m3 ; the highest calculated V annual average relative concentration from vent releases for any area at or beyond the SITE BOUNDARY.

Q = The gross release of nocle gas v radionuclides in gaseous effluents from all vents, determined by gross activity vent monitors, in uCl. Releases shall be cumulative over the calendar cuarter or year as appropriate.

IW Page 20 of 44, Rev. 4 B = 3.15 x 10-4 mrad / year per uCi/sec; the constant for Kr-S8 accounting for the gamma radiation from the elevated finite plume.

The constant was developed using the MARE program with plant specific inputs for PBAPS.

5 Q = The gross release of noble gas S radionuclides in gastaus releases from the stack determined by gross activity stack monitor in uCi. Releases shall be cumulative over the calendar quarter or year as appropriate.

b) Isotopic Analysis Method Dy = 3.17 x 10-3 M (X/Q) Q +B Q ,

1 1 v iv i is where:

The location is the SITE BOUNDARY, 1097m SSE from :ne vents. This location results in the highest calculated gamma air dose from nobla gas releases.

Oy = gamma air dose, in mead.

3.17 x 10-8 = years per second.

M = The air dose factor due to gamma emissions i for each identified noble gas radionuclide.

Values ara listeJ on Table IV.A.1, in mrad /yr cer uCi/m3 .

(T/Q) = 5.33 x 10-7 sec/m3 ; the highest calculated V average relative concentration from vent releases for any area at or beyond the SITE BOUNDARY.,

Q = The release of noble gas radionuclides, i, iV in gaseous effluents from all vents as determined by isotopic analysis, in uCi.

Releases shall be cumulative over the calendar quarter or year, as appropriate.

B = The constant for each identified noble gas i radionuclide accounting for the gamma radiation for the elevated finite plume. .

The constants were developed using the MARE program with plant specific inputs for PBAPS.

!alues are listed cn Tacle IV.A.1, in mrad /yr per uti/sec.

Page 21 of M , Rev. 4 4

Q

= ine release of noble gas radionuclides, i, ,

is in gaseous effluents from the stack determined by isotopic analysis, in uCi. Releases shall l be cumulative over the calendar quarter or year, as! appropriate.

2. for beta radiation:

a) Gross Release Method D = 3.17 x 10-8 N (X/Q) Q + (X/0) Q A v v s s whcrc:

This The location is the SITE BOUNDARY 1097m SSE from the vents.

location results in the highest calculated gamma air dose from noole gas releases.

Og

= beta- air dose, in mrad.

3.17 x 10-8 = years per second.

3 N = 1.03 x 104 mrad /yr per uCi/m  ; the air dose f actor due to beta emissions for Kr-87. (Reg. Guide 1.109,TableB-1) 3 (X /Q', = 5.33 x 10-7 sec/m ; the highest calculated v annual average relative concentration from vent releases for ar.y area at or beyond the SITE BOUNDARY.

Q = The gross release of noble gas v radionuclides in gaseous effluents from all vents determined by gross activity vent monitors, in uCi . Releases shall be cumulative over the calendar auarter or year, as appropriate.

= 9.97 x 10-8 sec/m3 ; the hignest calculated (T/Q) s annual average relative concentration from the stack releases for any area at or beyond the SITE BOUNDARY.

U = The gross release of noble gas

  • s radionuclides in gaseous releases from the stack determined by cross activity stack monitors, in uCi . Releases shall be cumulative over tne calendar cuarter ~ {

or yea , as a:prc:riate.  ;

. Page 22 of 44. Rev. 4 b) Isotopic Analysis Method _

C, = 3.17 x 10'b N (X/Q) Q + (X/Q) Q l " i i , v iv s is !

3.17 x 10-8 = years per second.

N s The air dose factor due to beta i emissions for each identified noble cas radionuclide. Values are listed 3 on Table IV.A.1, in nrad/yr per uCi/m ,

5.33 x 10-7 sec/m ; the highest calculated 3

=

(x/Q) v annual average relative concentration from vent releases for any area at or beyond the SITE BOUNDARY.

Q

= The release of noble gas radionuclide, i, iv in gaseous effluents from~all vents as determined by isotopic analysis, in uCi.

Releases shall be cumulative over the calendar quarter or year, as appropriate.

(X/Q) = 9.97 x 10-8 sec/m3 ; the highest calculated s annual average relative concentration from the stack releases for any area at or beyond tne SITE BOUNDARY.

Q r The release of noble gas radionuclide, i, is in gaseous effluents from the stack as determined by isotopic analysis, in uCi.

Releases shall be cumulative over the calendar quarter or year, as appropriate.

IV.C Surveillance Recuirement 4.8.CJ The dose to an individual frcm iodine-131, iodine-133, tritium and radioactive materials in particulate form and radionuclides otner than noble gases with half-lives greater than eignt days in gaseous effluents released to areas at and beyond the SITE BOUNDARY.

The dose shall be determined by one of two methods. Method (a), the lodine-131 Method, uses the iodine-131 releases and a correction factor to calculate the dose from all nuclides released. Method (b), the Isotopic Analysis Method, utilizes all applicable nuclides.

For normal operatiori, it is expected that Method (a) will be used since iodine-131 dominates the critical pat'hway - thyroid. However, in the event iodine-131 releases are minimal (e.g. during long term shutdown) Method (b) will be used to provide accurate calculations. In :ne ebsence of iodine- i 131 releases, the liver is the critical organ.

1

~

-'- .Page 23 of 44, Rev. 4

a. Iodine'--131-Method ,

1 0=3.17x-'10'8(CI)(0.5)R 'W Q '+ _' W Q.

Ti S IS V IV where:  :

Location is the critical pathway dairy 2103m SSW from vents.

0- = critical organ dose, thyroid, from all T pathways,.in mrem.

3.17 x 10-6 = years per second.

CF = 1.09; the correction factor accounting for the use of Iodine-131 in lieu of 'all- radio-nuclides released in gaseous effluents including Iodine-133.

0.5 '

= fraction ofLiodine releases which are nonelemental. .,

1 R = 3.08 x 10 11m2(mrem /yr)peruCi/sec;the dose factor for iodine-131. The' dose-factor is I based on the critical individual organ, thyroid, and most restrictive age group, infant. See Site Specific Data.**

1 i

W = 4.95 x lo do meters-2; (g/Q) for the food s pathway for stack releases.

I

_Q = The release-of iodine-131 fq m the stack I

IS- -determined by the effluent sampling and analysis program (Technical Specification '

Table'4.8.2),'inuCi. Releases shall be cumulative over the calendar quarter or

-year,_as appropriate.

l-W = 1.14 x 10~9 meters-2; (57Q) for the food v pathway for vent-releases.

Q = The release of iodine-!?1 frcm the vent IV' determined by the effluent sampling and analysis program (Technical Specification-Table 4.8.2), in uCi. Releases shall be cumulative over the calendar quarter or year, as appropfiate.

See Note 2 in Bases. .

l l

l i

I

.- , - - c . . . , - , . ~ - . - _ _,_.____ __ _ _ _ _ _ ______ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ . _ _ _ _ _ _

Page 24 of 44, Rev. 4 b, Isotcpic Analysis Method a N 0 = 3.17 x 10 * /__ R WQ_ + WQ_

i 4

5 is v iV i

- a shere:

Location is the critical pathway dairy 2103m $$W from vents.

D = critical organ dose, liver, from all pathways, in mrem.

Q 3.17 x 10 * = years per second.

R = The dose factor for each identified i radionuclide, i, basec on the critical individual organ, liver and most restrictive age group, infant.

2 Values are listed on Table IV.C.1, in M (mrem /yr) per uCi/sec.

W = 4.95 x 10'10 meters'2; (D/Q) for the food s pathway for stack releases.

Q = The release of radionuclides, i, in gaseous is effluents from the vents determined by the eftiuent sampling (Technical Specification Table 4.8.2), in uCi. Releases shall be cumulative over the calendar quarter or year, as appropriate, W = 1.14 x 10 c' meters; (D/Q) for the food v for vent releases.

0  : The release of radionuclides, i, in gaseous iv effluents from the vr.nt! determined by the ef fluent samDling and analysis program (Technical Specification Table 4.3.2) in uCi Release shall be cumulative over the Calendar cuarter or year, as appropriate.

i

s Pege 2+., cf 44, Lev. "

TABL! IV.u.) - C0t1STANTS FOR ISOTOPIC ANflV5]5 b'ETHOD (m2 (crem/yr) per uti/ set) f Radionuclide RI Mn-54 1.14 x 10 7 l

t Cr-51 4.72 x 10'*

Co 58 2.13 x 10 7 Co-60 2.58 x 10 7 In-65 5.56 x 10 9 Sr-69 1.06 x 108*

Sr . 9.06 x 109 '

Ce-la: 7.73 x 10' C -134 1.49 x 10 10 Cs-137 1.76 x 10 10 Bn-140 7.04 x 10 4

  • There it no liver dose f actor given in R.G.l.109 f or these nuclides. Therefore, ti.e whole body dose facter was used.

i

.. . . Page 26 of 44, Rev. 4 .. !

! IV.0 Surveil _ lance Requirement 4.8.C.5a  !

The projected doses from releases of gateous effluents-to areas at and I beyond the 511E B0VHDARY'shall be calculated in accordance with the following sections of this manual:

}

a. gamma air dose - IV.8.1 ,
b. beta air dose - IV.B.2
c. organ dose - IV.C The projected dose calculation sna11 be based on expected release from plant operation. .The normal release pathw6ys result in the maximum releases from the plant. Any alternative release pathways result in lower I releases and, therefore, lower doses. [

l IV.E Surveillance Recuirement 4.8.C.6.b lL - 1. - The two types of recombiner hydrogen analyzers used at Peach Bottom are:

a. Hays Thermal Conductivity type (Analyzers 205192L, 205192H, 205222, 205223,  !

305192L 305192H, 30S222, 305223)  !

b. ScottSeries9000 Helium-Immunetype(Analyzers [

205192L, 20S222, 30S192L and 305222)

2. The calibration gases for the two types are:  ;

i

a. Hays Analyzers l Zero Gas - Air j

.g - Calibration Gas - 4% Hydrogen, Balance Nitrogen-  ;

i

^

b. Scott Analyzers Zero Gas - Air Calibration Gas - 2% Hydrogen, Balance Air ,
l V.A. Sur.'eillance Recuirement 4.8.0~ j i

i If tre doses as calculated by the equations in this manual do not exceed  !

the limits given in Technical Specifications 3.8.8.2, 3.8.C.2, or 3.8,C.3  :

by more than two times, ...e conditions of Tachnical Specification 3.8.0  ;

have been met.

I If the doses as calculated by-the ecuations-in this-manual exceed the {

limits given in Technir i Specifications 3.8.B.2, 3.8.C.2, or 3.8.C.3 by  !

more than two times, the maximum dose or-dose commitment to a real j individual shall be determined utilizing the methodology provided in  !

Regulatory Guide 1.109, "Calculatinn of Annual Doses to Man from Routine j 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 Regu.latory Guide 1.109 shall be documented in '

the Special Report to be prepared in accordanse with Tee nical ' -l Specification 3.8.D. j I

e l  ?

I..,... _,_.,,,.,.4., _ . , ~ - , , - , _ , ~ , . - _ .m..c..._-__ _ _,.,_,~-..----__.---~,m--------- -

race a us

.m . nt,. ~

lhe c$mulative dote contribution f rom dirtct radiation f rom the two reactors at the site and f rcm racoat'.e storage shall be determined by tM following methods:

Cumulative dose contributicn from direct radiation

  • Total dose at the site of interest (as evaluated by Ttt tnessurement) -

Pean of background dose (as evaluatea by TLD's et background tites) -

Effluent contribution to dose (as evaluated by surveillance requirement 4.8.D).

ine This evaluation is in accordance with AN51/ANS 6.0.1-19/9 Section 7.

error using this method it estimated to be approximately R%.

VI. A. Uniove Peocrtina Pecuirement 6.9.2.h(3) Dosn Calculations for the Radiation Dose Assrssment Peport The assessment cf radiation doses f or the radiation dose assessment rcport shall be perf ormed utilizing the rtethodology provided in Regulatory Guide 1.109, " Calculation of Annual Doset To Man f rom Routine Releases of Reactcr Effluents for the Purpose of Evaluating Compliance with 10 CFR Part E0, Apper. dix 1", Revis ion 1. October 1977. Any deviations from the methodology prov'ded in Regulatory Guide 1.109 shall be documented in the radiation doae assessment report.

The meteorological conditions concurrent with the time of releate of radioactive materials (as determined by sampling frequency of measurement) or approximate methods shall be used as input to the dose model.

The Radiation Dose Assessment Report shall be submitted within 120 days af ter January 1 of each year in order to allow time for the calculation of radiation doses folicwing publication of radioactive releaces in the Radioactive Effluent Release Report. There is a very short turnarcund tir.e betwcen the determination of all radioactive releases and publicatirn of the Radioactive Effluent Release Report. 'ihis would not allow time for calculation of radiation doses in time for publication in the same report.

Vll . A Surveillance Raquirement 4.8.E The radiological environnent monitcring samples shall be collected purnant to lable Vll . A.1 f rom the locations shown on figures Vll. A.1, V:1. A.? cr4 Vll.A.3 and shall be analyzed purtuant to the requirements of Tatile Vll.A.l.

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    • 7-21 T O R I P AC. PR O'ir4 A M P A D I C L OG 7 C Al, E T4V j PO MCtJ i At ov ter nf 5 a s* !
  • s a<rf Statten Le mtion-Diretttorv 945;tr%5ica e cos 3r v s'a t u. a ,
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the roc

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r. p n a s otscharp, -

(OmDJsite UNI?1 2 & 3 C,res t i nocus s m eap t e r in the er sov to, .: L (on 3 t r:co D a n. -

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N ** '" -. P m ' * ' " " w . . , ~ . _ _ - , , , , ~ ~ ~ , , , - - . _ . . . . , _ _ _ , - - - - _ - ~ _ _ _ _ _ _ . _ _ . ,, face 41 of 44, ke4 4 . ,, s , e \ , e l Vill. EASES Sit., Sce:ific Data Note 1: Licuid cose f a:tus, As, t er secticn Ill.E oe e ceal::e: using the folicain; site 5;e:ific data. The licaid tit %1ys , involved are crir. ing water and fisn.

s a (U /D 4 U x EF )K x LF x RC i w w F 'i 0 i U = liters per year; maximum age group usage of w crinking water (Reg. Guice 1.109, Table E-5)

D = 5.4; average annual diluticn at Conc ing0 intake w U = kg per year; maximum age grcuo usa;e cf fith F (Reg. Guide 1.109, Table E-5) BF = bicaccumulation factor for nuclice, i, in fresn.ater i fish. Reg. Guide 1.109, Table A-1, except P-32 which uses a value of 3.0E03 pCi/kg ;er pCi/ liter. 1.14 x 106 = (105 pCi/uti x 10 ml/l - 8760 hr/yr) 3 K = 0 units conversion factor. DF = dose conversion f actor for nuclide, i, for the age i group in total body or organ, as applicable. Reg. Guice 1.109, Tacle E-ll, except P-32 bone wh ch uses a value as indicated below. 3.0 x 10 . RC = 1.16; reconcentration frcm PEAPS ciscnarge back through PEAPS intake. The octa for Dw and RC were derived frcm data published in Peacn Ecttom Atomic Power Station Units 2 and 3 (Docket Ncs. 50-277 and 50-278) Radioactive Ef fit ent Ocse Assessnent , Enclosure A, Septerer 3D, 1976. All other cata except P-32 Si ana DF *ere used as given in Ee;. Guice 1.109, Revision 1, Oct; Der 1977. Tne P-32 EF ana DF were used in accordance with i nformation supplied in Branagan, E.F., Nichols, C.R., and Willis, C.A., "The Impcriance of P-32 in Nuclear Reacter Licuid Effluents", NRC, 6/EE. The teen and child dose factors were derived by the ratic of the adult ocne dose factors in Reg. Guide 1.109 arc Eranagan, et al. Note 2: To develop constant R for section !V.C, tne fclicsing site specific data were used: . RC (0/0) = K'O (U )F x r x CFLi )a f (1-f s) -At i F ao m P e 1f Y A. ,. + A K' s 10 ;[i/uCi U",it Ccnversicn fa:tcr 1 i '**~' i fage 42 of 44, Rev. 4 ~ l I r Q = 50 Kg/ day; cow's consumption rate  ! f . f V 3301/yr; yearly milk consum; tion by an inf ant ao A = 9.97 x 10-7 sec'l decay constant for 1 131 i , A = 5.73 x 10*I sec -1' decay constant for removal of f w activity in leaf and plant surfaces.  ! F = 6.0 x 10~3 day / liter, the stable element transfer m coefficient for 1-131. r = 1.0 fraction of deposited radiciodine retained in cow's feed grass. DFL = 1.39 x 10-2 mrem /pCi - the thyroid ingestion dose f factor'for ,1-131 in the infant. ^ f = 0.6; the fraction of the year the cow is on pasture , - p. -(average o_f all farms) [ i f = 0.513; the fraction of cow feed that is stored feed i s while the cow is on pasture (average of all f arms).  ! 2 -Y = 0.7 Kg/m - the agricultural productivity of pasture p feed grass.  ; t = 2 days - the transport time from pasture to cow, to f milk, to receptor,  ; The pathway is the grass-cow-milk ingestion pathway. These data were } derived from data published in Peach Bottom-Atomic Power Station Units 2 _ and 3 (Docket Nos. SP-277 and 50-278) Radioactive Effluent Dose Assessment2 Enclosure A 30, 1976, All other cata were used as given in Reg. Guice 1.107,, Revision September 1, October 1977. 5 Surveillance Recuirement 4.8.8.2 Licuid Pathway Dose Calculations , The equations for' calculating the ' doses due to the actual release rates of } radioactive materials in liquid effluents were oeveloped from the methodology  ! provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from f Routine Releases of_ Reactor Effluents for the Purpose of Evaluating Compliance l -with 10 CFR Part 50, Appendix 1", Revision 1, October 1977 and NUREG-0133  : " Preparation of Radiological Effluent Technical Specifications for Nuclear Power  ! ~ Plants", October 1978. l ,5urveillance Recuirement 4.8.C.1 Dose Noble Gases j t The equations for calculating the doses due to the actual release rates of [ radioactive noble gases in gaseous effluents were ceveloped from the methodolooy  ! t E i 1 -_ u._______.___ _ ...__-- _.._ _ _ _ _ _ _ . _ _ - _ _ _ _ _ . - . 1 Page 43 of 44. Rev. 4 .s* t l l provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man f rom Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix 1", Revision 1, October 1977. NUREG-0133 " Preparation of Radiological !f fluent Technical Specifications for Nuclear Power Plants", August 1978, and the atmospheric dispersion model presented in information Recuested in Enclosure ? to letter from Georae Lear to E. G. Bauer I hated febraary 17, 1976, SeptemDer 30, 1976. The specified equations provide for determining the air doses in areas at and beyond the SITE _ BOUNDARY based upon the historical. Average atmospheric conditions, 'The dose due to 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 Radioactive Eff_lgent Dose Assessment, September 30, 1976, the values calculated by the Gross Release Method for total I body dose rate and skin dose rate are 6.0 times and 5.7 times, respectively, the I values calculated by the Iso'opic Analysis Method.; ' i The model Technical specification LCO for all radionuclides and radioactive  ; l_ materials in particulate form _ and radionuclides other than noble gases requires l i 'that the instantaneous. dose rate be less than the equivalent of 1500 mrem per -year, for the purpose of calculating this instantaneous dose rate, thyroid dose 'from iodine-131 through the inhalation pathway will b6 used. . Since the operating history to date-indicates that iodine-131 releases have had the major dose' impact, this approach is appropriate. The value calculated is increased by-  ; nine per cent to account for the thyroid dose from all other nuclides. This allows for expedited analysis and calculation of compliance with the LCO.  ; 'In the event that the plant is shutdown long enough so that iodine-131 is no longer present in gaseous effluents, an isotopic Analysis Method is available. I Since no iodines are present, the critical organ changes from the thyroid to the lung. Surveillance Pecuirement 4.8.C.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 ytth 10 CFR Part 50, Appendix 1", Revision 1, October 1977, NUREG-0133 " Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants", August 1978, and the atmospheric dispersion model presented in l Information Recuested in Enclosure 2 to letter from Georae Lear to E. G. Bauer dated February 17,'1976, September 30, 1976. The specified equations provide for determin} lng the air doses in areas at and beyeno the SilE BOUNDARY based ypon the historical average atmospheric conditions. _ The dote due to noble gas releases as calculated by the Cross Release Method is much more conservative than the_ dose calculated by tne isotopic Analysis Method. Assuming the releases rates given in Radioactive Effluent Dose Assessment, September 30, 1976, the values calculated by the Gross Release Method for total body dose rate and skin dose rate are 4.3 times and 7.2 times, respectively, the values calculated by the isotopic Analysis Method. l l-

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