ML20211H294

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Rev 4 to Allowable Leakrate Calculation for SG Interim Plugging Criteria
ML20211H294
Person / Time
Site: Byron, Braidwood  Constellation icon.png
Issue date: 09/30/1997
From: Passmore K
COMMONWEALTH EDISON CO.
To:
Shared Package
ML20211H271 List:
References
CON-9044-057, CON-9044-57 ATD-0410, ATD-0410-R04, ATD-410, ATD-410-R4, NUDOCS 9710060318
Download: ML20211H294 (28)
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Text

___________ _ _______ ___- __ _ ..

Exhibit C NEP 12-02 Revision 5 Page 1 of 2 PREPARATION, REVIEW AND APPROVAL OF CALCULATIONS CALCULATION TITLE PAGE Calculation Not Aro-0410 DESCRIPTION CODE: ROS DISCIPLINE CODE: H BYRON STATION UNIT 1_ SYSTEM cot,E: RC, ns I'

TITLE: ALLowAntz tzAxnATE CsLCutArrou ron arrAu oEnEnAron zurEnts PLuoan o '

CRITERIA

(@ Safety Related O Augmented Quality C Non-Safety Related REFERENCE NUMllERS Type Number Type Number I 1 COMPONENT EPN: DOCUMENT NUM HERS:

EPN Compt Type Doc Type / Subtype Document Number 1RC01BA B10 Steam Generator CALC /ENG Bi M 7- 3 32 1RC01BB B10 Steam Generator 1RC01BC B10 Steam Generator IRC01BD B10 Steam Generator REMARKS:

REY. REVISING APPROVED DATE ORGANIZ,ATION PRINT / SIGN O comed C502 D. Palmer / (See original for signature) 7/31/94 1 Comed C502 W. Perchiazz1/ (see original for signature) 8/21/97 2 Comed C502 W. Perchiazzi/ (See original for signature) 8/2f/T/

v.

3 comed C502 K. Passmore/ (see original for signature) 9/22/97 4 comed C502 /( t ys,1 9 n m occ / f( g_ 9/m b 9710060318 971001 PDR ADOCK 05000454 P PDR ,

Exhibit C NEP.12-01 Revision 5 Page 2 cf 2 COhlh10NWEALTil EDISON COh1PANY CALCULATION llEVISION PAGE CALCULATION NO, ATD-0410 PAGE NO.: 2 IlEVISION SUh1h1AltlES REV: 3 l REVISION SUhihiARY:

Pevise calculation to reflect changes to the iodine concentrations listed in UFSAR Table 15.0-9, per Westinghouse Letter CAE-97-185 dated 9/19/97. Revise l to reflect changes to BYR97-332. See page 2.1 for more detail.

Electronic Calculation Data Files:

(hopam Name, Version, rite name eeneutehour/: min)

J. Smith / Isee otagnal for angnatucel 9/22/97 Prepared by: M. Marchionda/ tree orignal f o r s t ana t u r e t 9/22/97 Print / Sign Date Reviewed by: 9._leht,11Q.. oragt.nal tsr g gnatu m 9/22f97 _ _..

Print /Sinn Date Type of Review

[@] Detailed [ ] Alternate [0] Test DO ANY ASSUMFDONS IN1111S cal.CULA110N REQUIRElAllR YERIHCA110N [O] YES [0] NO Tracked by:

REV: 4 REVISION SUhihiARY:

This revision supersedes all previous revisions of this calculation. See page 2.2 for detail.

Electronic Calculation Data Files:

(hopam Name, Verskm, role nanw cu sinate hour / min)

J. Smith / N N Prepared by:._M. Matchiondh/ ,; 11.( { [}@dl. ,_

Pri 'gn j Date Reviewed by: _o. tteN. , , eI<f j-3r 97 Lahti/,Pnnt v /Sinn , - - - ,

Date Type of Review

[X) Detailed [C] Alternate { } Test DO ANY ASSUMFi1ONS IN11IIS CAlfUIAT10N REQUIRE 1A11R \1RIRCAT10N 10] YES (D)NO Tracked by:

I

l l

CALCULATib,' NO: ATD-0410 l PAGE NO: 2.1 1

Revision summary ior Revision 3 (con't)

- Pages 5 and 6: Replaced Design Inputs 5 and 6 with Design Inputs 19 and 20.

Renumbered remaining inputs accordingly.

( -

Page 6: Added Design Input 22.

Pages 6 and 7: Deleted Reference 5 (BOP CV-17 and BOP CV-9). Renumbered remaining references accordingly.

Page 7: Added References 23 and 24.

Page 8, 10-18, 20, 22,-24: Revised Reference numbers.

Page-8: Revised values in Table 1.a.

Page 84 Added. unit conversion factor.

- Page 9 Revised values in Table 1.b.

Page 104 Revised values in Table 1.c and revised text accordingly.

- Page 16: Revised values in Table 2.f.3.

- Page 21: Revised values in Table 3.f.3.

- Pages 21, 22, 24 and 25: Revised values in equations and revised text accordingly.

REVISION NO. : 4

. a- -

- Site Appendices NEPol2-02 Revision 5 COMMONWEALTH EDISON COMPANY CALCULATION NO: ATD-0410 PAGE NO: 2. 2 Revision Sunenary for Revision 4 (con' t_ f rom p.2) :

Revise calculation to reflect total activity released with 12.8 gpm primary to secondary leak in the time period 0-2 hours to determine the low population tone thyroid dose.

-- The letdown purification system parameters were revAsea to reflect UFSAR Table-11.1-1 instead of UFSAR Table 9.3-2.

Revisions to Design Inputs, References, variable names and text for clarity and

-e ditorial purposes.

r REVISION NO.: 4 -

Exhibit D NEP 12-02 Revision 5 COMMONWEALTil EDISON COMPANY CALCULATION TABLE OF CONTENTS CALCULATION NO. ATD-0410 REV. NO. 4 PAGE No. 3 SECTION PAGE NO. SUB PAGE NO.

l .-

TITLE PAGE 1 2 2.1, 2.2 REVISION

SUMMARY

TABLE OF CONTENTS 3 PLKPOSE/0BJECTIVE 4 METil0DOLOGY AND ACCEPTANCE CRITERIA 4 s ASSUMPTIONS 5 DESIGN INPUT 5 REFERENCES 6 CALCULATIONS e

SUMMARY

AND CONCLUSIONS 25 ATTACHMENTS "/*

Exhibit E N EP-l>02 Revision 5 COMMONWEALTH EDISON COMPANY l CALCULATION NO. : ATD-0410 PROJECT NO. PAGE NO. 4 l PURPOSE AND OBJECTIVEt The purpose of this calculation is to generate the maximum allowable primary to secondary steam generator tube leak rate during a postulated main steam line break using 24% plugging criteria deM data. The evaluation was performed for both a pre accident and accident initiated lodine sptf.t,. The release of lodine and the resulting thyroid dose at the Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) were considered in the leak rate determination. Whole body dose due to noble gas immersion is less limiting than thyroid dose as documented in UFSAR Table 15.011, Given the large margin to the 25 rem whole body dose limit, whole body dose was not re eva'uated.

This calculation also determines the resulting thyroid dose at the Exclusion Area Boundary, Low Population Zone and Main Control Room for the actual predicted end-of cycle 8 steam generator tube leakage during a postulated Main Steam Line Break.

METHODOLOGY AND ACCEPTANCE CRITERIA:

The Main Steam Line Break (MSLB) accident is evaluated because the event causes a sustained large pressure difference across the steam generator tubes providing a motive force for Reactor Coolant System (RCS) release. The dose attributed to a 1 gpm leak rate from the reactor coolant system was calculated. This value was then used to determine the allowable leak rate without exceeding the Standard Review Plan dose criteria.

The activity released to the environment due to a MSLB is analyzed in two distinct releases:

1. The release of the iodine act'vity that has been established in the secondary coolant prior to the accident, and
2. The release of the primary coolant lodine activity due to tube leakage.

The methodology used for calculating the Radiological Consequences of a MSLB with primary to secondary leakage is consistent with the Standard Review Plan (NUREG 0800),15.1.5 Appendix A.

TID 14844 dose conversion factors were used to determine dose equivalent lodine concentrations for the RCS, which is the Technical Specification definition of dose equivalent lodine. The TfD values are based on ICRP 2, ' Permissible Dose for Internal Radiation,1959.*

The off site dose assessment uses ICRP 30< ' Limits for intakes of Radionuc1 des by Workers, 1979' dose conversion factors. ICRP 30 is also the basis for Federal Guidance Report No.11,

  • Limiting Values of Radionuclide intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingection,' dated 1988.

The dose Acceptance Criteria are based on the guldance of Standard Review Plan (NUREG-0800) Section 15.1.5, Appendix A. For a MSLB with a postulated pre accident lodine spike, the calculated doses should not exceed the guideline values of 10CFR Part 100 Section 11. The numerical values used for these doses are 25 rem to the whole body and 300 rem to the thytold from lodine exposure for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> following the accident. For a MSLB with an accident initiated lodine spike, the calculated (oses should not exceed a small fraction of the 10 CFR 100 guideline values, i.e. 2.5 rem and 30 rt m respectively for the whole body and thyroid doses.

l REVISION NO.: 4 l

Exhibit E NEPol2 02 Revision 5  ;

COMMONWEALTH EDISON COMPANY l CALCULATION NO. : ATD-0410 PROJECT NO. PAGE NO. 5 l ASSUMPTIONS:

1) The effect of boron on the RCS density is assumed to be negligible since the boron trass  ;

is less than 1% of the total RCS mass at the beginning of core life.

1 PE81GN INPUT 8t

1. The total volume of the RCS is 12,062 ft'. (Reference 1)
2. The full power RCS temperature and pressure are 586.2 *F and 2250 psla. (Ref 1 and 2)
3. The RCS specific volume at full power is 0.02258 ft'/lbm. (Ref. 3)
4. Deleted
5. Letdown Purificsica System temperature is 130*F and 2300 psig. (Ref 1)
6. Letdown Purifictiion Sytdem specific volume is 0.01613 ft'/lbm. (Ref. 3)
7. Breathing rates are taken from B/B UFSAR Table 15A 1. (Ref. 5)
8. Atmospheric DLtion Factors, X/Q, are the fifth percentile values taken from UFSAR l Table 15.013. (Ref.6)
9. RCS iodine concentrations are based on UFEAR Table 11.12. (Ref. 7)
10. The initial steam release from the defective and intact steam generators is taken from .

UFSAR Table 15,13. (Ref. 8)

11. The secondary side faulted steam generator has a partition coefficient of 1.0 and the l Intact steam generators have partition coefficients of 0.1. (Reference 14) l 12. The half life for 1131 is 8.04 days,1132 is 2.30 hrs,1 133 is 20.8 hrs,1134 is 52.6 min, and 1135 is 6.61 hrs (VO.693/ half life). (Ref.20)
13. The initial primary coolant activity for the pre accident spike is 60 Cl/g DE l 131 and 1 Cl/g DE l 131 for the accident initiated spike. (Ref.11 and 13)
14. The initial secondary coolant activity is 0,1 pCl/g DE l 131. (Ref.11 and 13) t 15. The duration of the spike is 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. (Ref. 25) 16.- No fuel failure is attributable to the accident since DNB design basis is met. (Ref.11 and 26) 17, lodine partition coefficients for all SGs are 1.0 for primary-to-secondary leakage. (Ref.

14) l REVISION NO.: 4 l l

a e Exhibit E NEPol2002 Revision. %

COMMONWEALTH EDISON COMPANY l CAL CULATION NO t ATD 0410 PROJECT NO. PAGE NO. 6 l

18. Normal letdown purification flow Is 75 gpm. (Ref 1) i
19. Domineralizer Decon Factor, DF, for lodine is 10. (Ref 1)
20. The lodine release rate spike factor is 500. (Ref 11) i 21. The main control room MSLB thyrold dose is 7.4 rem for an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> release. (Ref 21)
22. Dose conversion factore are based on ICRP 30 values using the significant digits from Feders Guidance Report No.11. (Ref.16 and 19)
23. The allowable primary to secondary leak rate is 150 gallons per day (0.1 gpm) per steam generator. (Ref.13)

I

24. The total projected end of cycle 8 MSLB teak rate is 63.1 gpm at operating conditions.

(Ref. 27)

REFERENCE 31

1) B/B UFSAR Table 11.1 1, Revision 0 superseded by Westinghouse i.ctter CAE 97185,

' Byron and Braldwood Primary Coolant Source Term,' dated September,19,1997

2) B/B UFSAR Table 5.1 1, Revision 0
3) ASME Steam Table, Fifth Edition
4) Deleted
5) B/B UFSAR Table 15A 1, Revision 0
6) D/B UFSAR Table 15.0-13. Revision 0
7) B/B UFSAR Table 11.12, Revision 0
8) B/B UFSAR Table 15.13, Revision 6
9) Introductory Nuclear Physics by Kenneth S. Krane,1988
10) Deleted
11) Standard Review Plan (NUREG 0800),15.1.5 Appendix A
12) Deleted
13) Technical Specifications 3.4.8 (Amendment 77),3.7.1.4 (Original), 3.4.6.2 (Amendment 67)
14) WCAP 14046, 'Braldwood 1 Technical Support for Cycle 5 Steam Generator Interim Plugging Criteria,' dated May,1994 l REVISION NO.: 4 l

Exhibit E NEN2 02 l Revision 5 COMMONWEALTH EDISON COMPANY i l CALCULATION NO. : ATD 0410 PROJECT NO. PAGE NO. 7 l

15) ICRP Publication 2, Report of Committee ll on Permissible Dose for Intemal Radiation, 1959
16) ICRP Publication 30, Limits for Intakes of Radionuclides by Workers,1979
17) J.P. Adams and C.L. Atwood,'The lodine Spike Release Rate During a Steam Generator Tube Rupture,' Nuclear Technology, Vol. 94, pp 361371. June 1991; and EGG NERD-8648 Technical Report,' Probability of the lodine Spike Release Rate During a SGTR,'

September 1989

18) Westinghouse Letter CAE 97171, dated July 21,1997, pertaining to the Reactor Coolant Water Density Used in Determining Byron and Braidwood Attemate Tube Plugging Limit
19) Federal Guidance Report No.11," Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors For Inhalation, Submersion, and Ingestion, 1988
20) Radioactive Decay Data Tables: A Handbook of Decay Data For Application to Radiation Dosimetry and Radiological Assessments, DOE / TIC 11026,1981 by David C. Kocher
21) Byron Calcelation BYR97 332, Revision 2
22) B/B UFSAR Table 6.41, Revision 0
23) B/B UFSAR Table 15.14a, Revision 6
24) B/B UFSAR Table 15.0-9, Revision 0, superseded by Westinghouse Letter CAE 97185

' Byron and Braidwood Primwy Coolant Source Term,' dated September, 19,1997

25) B/B UFSAR Section 15.1.5.3, Revision 6
26) B/B UFSAR Section 15.1.5.1, Revision 4
27) Byron Operability Assessment 97-044, dated July 22,1997
28) TID-14844, ' Calculation of Distance Factors for Power and Test Reactor Sites," March 23, 1962 VARIABLE AND CONSTANT DEFINITIONS:

M RCS mass [g]

M.% Steam Generator steam release mass [lb]

S V RCS volume [ft ]

v RCS specific volume [ft*/lbm 4 RCS leak rate constant [sec',))

6 Fuel Release constant [Ci/sec]

Ao isotope Decay Constant [sec"]

Aot Letdown Purification Removal Constant [sec"]

4 Total lodine Removal Rate [sec"]

t Time [sec)

A, RCS lod!ne activity [Cl]

l REVISION NO.: 4 l

4 Exhibit E NEP-12-02 Revision 5 COMMONWEALTH EDISON COMPANY l CALCuthTION NO. : ATD 0410 PROJECT NO. PAGE NO. 8 l C, lodine Concentration [Cl/g or pCl/g]

C. In:'tlal lodine Concentration [Cl/g or pCl/g)

F, Letdown Punfication Flow [g/sec) j R. ActivKy Released of nuclide,I[Cl) i D Thyroid Inhalation Dose [ rem)

B Breathing Rate [m'/sec)

X'O Atmospheric Dilution Factor (sec/m'l W. DCF Weighted Activity [ Rem]

DE l 131 Dose Equivalent lodhe 131 OEFINE UNITS:

pCi = IE-6 Ci 1 1 Ibm = 454 g t

i ft 3= 7.48 gal 1 min a 60 sec 1 Sv/Bq = 3.7E12 Rem /Cl 1.

CALCULATION OF DOSE DUE TO STEADY STATE ACTIVITY IN SECONDARY SIDE The first dose component to be calculated will be the dose from the secondary side. The l secondary side activity is conservatively taken as the Technical Specification limit of 0.1 pCl/g (Design input 14). This value is the same for both the pre accident and accident initiated events. The steam release for the faulted steam generator (SG) is 96.000 lbs (Design input 10) which is the entire initial SG water mass. The faulted SG is assumed to steam dry in 10-15 minutes so all of the lodine is available for release. The combined 0-2 hr steam release for the three intact steam generators is 406,716 lbs (Design input 10).

The combined 2 8 hr steam release for the three intact SGs is 939,604 lbs (Design input 10). For the three intact SGs, a partition coefficient of 0.1 is used (Design input 11),

a.

The lodine concentrations a e obtained from Reference 24 and are converted to Cl/lb, sint* the steam release is defined in Ibs.

C = C. x454 x 1E - 6 Equation 1.a TABLE 1.a Nuclide - lodine Concentration, C.,- lodine Concentration, C, ~

'(Reference 24)- ' (Equation 1.a)

[gCile) - (Cillb]

l131 0.0645 2.93E 5 l132 0.0723 3.28E 5 l133 0.1032 4.69h-5 l-134 0.0155 7.04E-6 l135 0.0567 2.57E-5 l REVISION NO.: 4 l

Exhibit E NEP-12-02 Revision 5 -

COMMONWEALTH EDISON COMPANY l CALCULATION NO. : ATD-0410 PROJECT NO. 23GE NO. 9 l

b. The iodine concentration for each nuclide, C. from Table 1.a la multiplied by the mass of steam released (96,000 lbs for the faulted SG and 406,716 lbs for the three intact SGs) and then multiplied by the partition coefficient (1.0 for faulted steam generator and 0.1 for intact steam generators) to obtain the total curies released, R., for 0-2 hours.

Rl*""*d(Cl} = C i r 'd,*","*d(Ib] x 1.0 Equation 1.b.1 i

Rf'*d (Cl) = C, x M',",',*d (ib) x 0.1 Equation 1.b.2 TABLE 1.b

. Nuclide : Activity Released from.  : Activity Released from-Faulted SG, R ", intact SGs (0-2 hrs), Ri '"'*", :

-(Equation 1.b.1) [Ci] (Equation 1.b.2) [Cil-l131 2.81E0 1.19E0 1132 3.15E0 1.33E0 1133 4.50E0 1.91Ec l134 6.76E 1 2.86E-1 1135 2.47E0 1,05E0

c. The activity released. R, determined above, is multiplied by the ICRP 30 Dose Conversion Factor, DCF,, (Design input 22) for each lodine isotope and then summed separately for the faulted SG and intact SGs. The DCF weighted activity released is:

Wf*""*d(rem) = Rf*""*d(Ci]x DCF, - Equation 1.c.1 Wl"'**(.em) = Rf'*d(Cl)x DCFi Equation 1.c.2 l REVISION NO.: 4 , l

Exhibit E NEP.12 02 Revision 5 COMMONWEALTH EDISON COMPANY I CALCULATION NO. : ATD 0410 PROJECT NO. PAGE NO.10 l TABLE 1.c Nuclide ~ ICRP-30 Dose - = Weighted Activity . Weighted Activity .

Conversion Factor. '

from Faulted SG, from Intact SGs, DCFi,e #'", #",(0 2 hrs) ~

(Design input 22) . (Equation 1.c.1) [ rom) (Equation 1.c.2) [ rem]

[ rom /Cl] '

l131 1.08E6 3.03E6 1.29E6 l132 b.44E3 2.03E4 8.57E3 1133 1.80E5 8.10E5 3.44E5 l134 1.07E3 7.23E2 3.06E2 1135 3.13E4 7.73E 4 3.29E4 Total I(R xDCFi) i 3.94E6 1.68E6 The 0 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> weighted activity released from the faulted ano the three intact SGs Ic 5.62E6 rem (3.94E6 + 1.68E6 rem). This total weighted activity can also be defined as I(R,xDCF).

The DE l-131 activity released from the faulted and intact steam generators is the total weighted activity from Table 1.c divided by the 1 131 dose conversion factor.

Numerically this is 5.20 Cl DE l 131 (5.62E6 rem /1.08E6 rem /Cl)

d. The off sito thyroid inhalation dose at the exclusion area boundary, DeAs, is calculated in accordance with UFSAR equation 15A-2. The low population zone dose for both the preaccident and accident initiated lodine spike cases includes the activity released from the secondary side and is calculated in steps 2.f and 3.f, respectively. The breathing rate is based on the 0-8 hour time period.

Exclusion Area Boundary Dose (0-2 hours) Due to Secondary Side Release y

DeAs(rem} = -

xBx (R, x DCFi ) Equation 1.d.1 sO /EA8 sec~ *

= 5.7E - 4 x 3.47E - 4

.m.3 sec x 5.62E6[ rem}

= 1.11 (rem}

l REVISION NO.: 4 l

Exhibit E NEP-12 02 Revision 5 COMMONWEALTH EDISON COMPANY 1 l CALCULATION NO. : ATD 0410 PROJECT NO. PAGE NO.11 l

2. CALCULATION OF DOSE DUE TO PRIMARY-TO-SCCONDARY LEAKAGE DURING PRE-ACCIDENT INITIATED SPIKE in accordance with Reference 11, the pre-accident case occurs when the reactor is operating at the maximum value permitted by the Technical Specifications prior to the postulated MSLB. The radioactive isotopes are assumed to be evenly distributed throughout the RCS The lodine activity changes over time due to radioactive decay and the rate at which activity leaves the RCS due to primary to-secondary tube leakage,
a. The RCS mass inventory, M, will be calculated given the hot full power volume and specific volume.

RCS Volume: V=12062 ft' (Design input 1)

RCS specific volume v=0.02258 ft'/lbm (Design input 3)

Mg) = x 454 U-- Equation 2.a ft 3 . lbm, v

t 1206$ft5 9 x 454 ft 3 ~lbm' 0.02258 lbm

= 2.42E8 (g]

b. The RCS activity needs to be calculated for 60,.Ci/g DE l 131. UFSAR Table 11.1-2 is used to obtain RCS activity, which is based on 1% fuel clad defects per UFSAR Table 11.1 1. The total initial RCS activity is calculated by multiplying the initial concentration by the RCS mass. The initial DE l-131 actidy is then determined by multiplying each isotope's activity by its dose conversion factor, summing the values for each nuclide and dividing the sum by the I-131 dose conversion factor to normalize the activity to 1 131. This DE l 131 activity is the contribution due to 1% fuel clad defects. To determine the activity at 1 Cl/g DE l-131, the fraction of each isotopes contribution to the DE l-131 is calculated and then multiplied by the RCS mass to obtain the corrected total activity in the RCS at 1 pCl/g DE l-131. To obtain the total activity at 60 Cl/g DE l 131, each isotope activity is multiplied by 60.

l l REVISION NO.: 4 l

Exhibit E NEP-12-02 Revision 5 I COMMONWEALTH EDISON COMPANY I CALCULATION NO. : ATD-0410 PROJECT NO. PAGE NO.12 l 1

A i (Cl)= C, b x M(g) Equation 2b.1 9

W,[ rem)= A i[Cl)x DCF, Equation 2b.2

' 1.39E9{ rem)

DE 1131[Cl)= .=

    • = 939.2 (Cl)

DCFnu 1.48E6 Nuclide Concentration at i b

^

Equation 2b.3 E

RCS Activityat 1 {Cl)= Equation 2.b.3 x M(g)x1 x Equation 2.b.4 RCS Activityat 60 b{Cl)= Equation 2b.4[Cl)x 60 Equation 2b.5 g

TABLE 2.b Nuclide . RCS- RCS- . ICRP 2 Weighted Nuclide RCS Total RCS Total Concent., Activity. - . Dose -. . Activity,E : Concent. - Activity at Activity at C. LA, ' Conversion .WJ at- . 1 pCilg . 60 pCilg (UFSAR (Eq. Factor DCFi (Eq.- :1 Cilg '(Eq. 2.b.4) (Eq. 2.b.5)

Table 2.b.1) - [ rem /Ci] 2.b.2)1 - (Eq. [Ci]- [Cl]

11.1-2) - [Ci] ~ (Ref. 28) . [ rem] ;2.b.3)

[Cilg]

l131 2.5E-6 605 1.48E6 8.95E8 0.645 156.1 9.36E3 1132 2.8E-6 678 5.35E4 3.63E7 0.722 174 8 1.05E4 l-133 4,0E-6 968 4.00E5 3.87E8 1.032 249.7 1.50E4 l134 6.0E-7 145 2.50E4 3.63E6 0.155 37.5 2.25E3 1-135 2.2E-6 532 1.24E5 6.60E7 0.566 137.3 8.23E3

EW i- 1.39E9 l REVISION NO.: 4 l

Exhibit E NEP-12-02 Revision 5 COMMONWEALTH EDISON COMPANY l CALCULATION NO. : ATD-0410 PROJECT NO. PAGE NO.13 l I

c. The two removal mechanisms for this accident are due to decay and leak rate to the secondary side of 1 gpm. The time dependent activity after two hours with l the removal constants can be calculated using the basic decay equation methodology (Reference 9),

dAt) = -A A(t)- A, A(t) dA(t) = -

t A, A(t) o 8 - ** )

A(t) = A e" Where : t = 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> = 7200 sec 1 gpm Volume of RCS

~ gal'

, . min . .

1 ft 3 _ 1[ min) _

12062 ft 3 7.48[ gal), ,60[sec).

= 1.8SE - 7 sec d

d. Since the isotope activity is assumed to remain evenly distributed throughout the RCS volume, then the rate at which the isotope activity leaks from the RCS, R(t),

is simply the RCS leak rate times the activity. The total activity released during a given time interval is the integration of the release rate over that Interval, in this case,2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

R(t)= A, x A(t)

R(t) = A, x A,ed ** * ** )  :

I jR(t)dt = f A, A,ed (*****I dt o o

, W 0

-(Ad + ke)o(A +A,%-I(As**e)dt R= "^* Equation 2.d Ao + A, (1- e"(****'I) l REVISION NO.: 4 I

Exhibit E NEP-12-02 Revision 5 COMMONWEALTH EDISON COMPANY l CALCULATION NO. : ATD-0410 PROJECT NO. PAGE NO.14 l TABLE 2.d Nuclide RCS Activity at 60 . lsotope Decay.'

~

Activity

pCilg _

Constant, A4 , Released, Ri -

(Table 2.b) [Ci] : (Equation 2.d)

(Design

' - [sec'inp]ut_12) [Ci] -

l131 9.36E3 9.97E-7 1.24E1 1-132 1.05E4 8.37E 5 1.05E1 1133 1.50E4 9.25E-6 1.93E1 1134 2.25E3 2.20E-4 1.50E0 1-135 8.23E3 2.91 E-5 9.88E0

e. Calculate the 0-2 hour thyroid inhalation dose at the Exclusion Area Boundary (EAB) in accordance with UFSAR equation 15A 2.

TABLE 2.e

. Nuclide Activity ICRP-30 Dose- Weighted Activity

Released, Ri
Conversion Factor,'  : Released -

(Table 2.d) [CI) _

'DCFi,1 'Ri x DCF

= (Design input 22)  :[ rem]

+

[ rem /Ci] '

l131 1.24E1 1.08E6 1.34E7 l132 1.05E1 6.44E3 6.76E4 l-133 1.93E1 1.80E5 3.47E6 l134 1.50E0 1.07E3 1,60E3 1135 9.88E0 3.13E4 3.09ES

- Total I(RixDCFi)us 1.72E7 The total DE l-131 activity released is the total weighted activity from Table 2.e divided by the 1131 dose conversion factor. Numerically this is 15.9 Ci (1.72E7 rem /1.08E6 rem /Cl). The breathing rate is based on the 0-8 hour time period.

Exclusion Area Boundary Dose for 0-2 hours for a 1 com Leak Rate y

Dua(rem)= -

x B x E(Ri x DCF)as i Equation 2.e.1

<O,ys

= 5.7E - 4 x 3.47E - 4 tr.3 . ,

sec x 1.72E7[ rem)

= 3.40 (rem) l REVISION NO.: 4 l

Exhibit E NEP-12-02 Revision 5 COMMONWEALTH EDISON COMPANY l CALCULATION NO. : ATD-0410 PROJECT NO. PAGE NO.15 l- l f, Calculate the thyroid inhalation dose at the Low Population Zone (LPZ) using the equation from UFSAR Section 15A.4. The activity released during the accident from 2-40 hours was obtained from UFSAR Table 15.14a. This activity includes ,

the dose contribution from 12.8 gpm primary to secondary leakage and the secondary side release, l j

TABLE 2.f.1 Nuclide  : RCS lodine Activity.. .. . .lCRP-30, Dose -  ; 2-40 Hour Weighted t

. Released, RR ' . Conversion Factor, - Activity Released,-

(UFSAR Table - . . DCFL . J Ri x DCFE

.15,14a); [Ci]; ' (Design input 22) . (rem]:

[ rem /Ci]

1131 2.4E3 1.08E6 2.59E9 l-132 5.1E1 6.44E3 3.28E5 l-133 2.3E3 1.80E5 4,14E8

! l-134 5.1E0 1.07E3 5.46E3 l

l135 5.0E2 3.13E4 1.57E7 i

i Total E(R x DCF,) -

4 3.02E9 The total 2-40 hour weighted activity calculated above in Table 2.f.1 is separated into specific time periods of 2-8 hrs,8 24 hrs,24-40 hrs. This is based on scaling the total 2-40 hour weighted activity by the fraction of steam released during the same time period. The 2 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> steam release was obtained from UFSAR Table 15.13.

TABLE 2.f.2 Time Period - Steam Release, Fraction of Total-: Total 2-40 Hr ; Weighted -

(UFSAR Table ~ _ Steam Release = -Weighted _ Activity

.15.1-3) [ib] . for Time Period Activity , Released in :

Released (Table - Time Period..

- 2.f.1) : [ rem] -  :-[ rem]

2 8 hr 939,604 0.30 3.02E9 9.06E8 8-24 hr 1,234,515 . 0.39 3.02E9 1.18E9 24-40 hr 980,806 0.31 3.02E9 9.36EB Total Steam .: Total Weighted -

Release 3,154,925 . Activity l 3.02E9

Released The atmospheric dilution factors (X/Q) for 0-8 hrs, 8-24 hrs, and 24 40 hrs were obtained from UFSAR Table 15.0-13. The breathing rates for 0-8 hrs,8-24 hm and 24-40 hrs were obtained from UFSAR Table 15A-1. Calculate the thyroid inhalation dose at the Low Population Zone (LPZ) using the equation from UFSAR Section 15A.4.

e X'

D Lez[ rem) = x B x E(R, x DCFi ) Equation 2.f.1 h)LPZ l REVISION NO.: 4 l

Exhibit E NEP-12-02 Revision 5 COMMONWEALTH EDISON COMPANY l CALCULATION NO. : ATD-0410 PROJECT NO. PAGE NO.16 l TABLE 2.f.3

- Time . - Atmos. = . Breathing Rate, Weighted : . LFZ Dose, DLPz, Period i utspersion -- ' B (UFSAR- . Activity ; w/ '12.8 gpm Leak -

T Factor, X/Q, ; Table 15A-1) ' Released.1  : rate (Equation;

'(UFSAR Table

' [m*/sec) s

~

E(RxDCF) 2.f.1) [ rem];

~~ ' 15.013) i . (Table 2.f.2)-

i [sec/mi [ rem]!

0-2 hr 1.70E 5 3.47E 4 2.26E8* 1.33 2-8 hr 1.70E 5 3.47E 4 9.00E8 5.34 6-24 2.4E 6 1.75E-4 1.18E9 0.50 24-40 1.1E 6 2.32E 4 9.36E8 0.24 Total LPZ Dose w/12.8 gpm 7.41 Leak rate [ rem) i

  • From Tables 2.e and 1.c [(12.8x1.72E7) + (3.94E6+1.68E6) =2.26E8]
3. CALCULATION OF DOSE DUE TO PRIMARY TO SECONDARY LEAKAGE DURING ACCIDENT INITMTED SPIKE The accident initiated spike model is the same as the pre-accident model except an additional iodine appearance rate term is added for fuel release rate into the RCS. In accordance with the Standard Review Plan, the reactor trip and/or primary system depressurization associated with the MSLB creates an iodine spike in the primary system.

The spiking model assumes that the iodine release rate from the fuel rods to the primary coolant increases to a value 500 times greater than the Technical Specification limit.

This factor adds an additional release rate factor for iodine activity, hue.

a. Calculate the total removal rate of iodine,4, through letdown purification and radioactive decay. Equation 2 of Reference 17 defines this total as:

A, sec- j= 1dec"}+ A (sec") d Equation 3.a.1 F, -

Where : Ao t sec d = x 1- Equation 3.a.2 l REVISION NO.: 4 l

. l Exhibit 0 NEP.12 02 Revision 5 COMMONWEALTH EDISON COMPANY l CALCULATION NO. : ATD 0410 PROJECT NO. PAGE NO.17 l The 75 opm letdown purification flow F,, is converted from opm to grams /sec at letdown design parameters (Design inputs 5 and 6).

~ g' . gal' i ft 3 ' ilb ~454g' ' 1 min '

b ,sec, * ' ((E, *

,7.48 gal,

,0.01613 ft ,

lb ,

60 sec,

= 4704 9.

.sec, Substituting the values of F,,M and DF int o Equation 3.a.2 gives:

4704 , s 4

" 2.42E8[g]' 'k

= 1.75E - 5 sec-'

TABLE 3.a Nuclide Letdown Purif. Isotope Decay  : Total lodine Removal Constant,~ : Constant, le ' Removal Rate,1 Aw,(Equation 3.a.2) - (Design input  : Equation 3.a.1

[sec] 12)' [.ec] - [sec) .'

l-131 1.75E 5 9.97E-7 1.85E 5 l132 1.75E 5 8.37E 5 1.01 E-4 l133 1.75E-5 9.25E 6 2.67E-5 l134 1.75E-5 2.20E-4 2.38E-4 l-135 1.75E-5 2.91 E-5 4.66E-5

b. The equilibrium fuel release rate,6, is defined as the product of the RCS activity at 1 pCl/g DE l-131 (from Table 2.b) and the total lodine removal rate for each isotope:

6 [Cl/sec) = A,[Ci] x A [sec] Equation 3.b l REVISION NO.: 4 l

1 Exhibit E NEP-12-02 Revision 5 COMMONWEALTH EDISON COMPANY l CALCULATION NO. : ATD-0410 PROJECT NO. PAGE NO.18 l l Each equilibrium fuel release rate is multiplied by 500 (Design input 20) to obtain the spiked release rate.

. TABLE 3.b Nuclide . Total lodine - - Fuel Release - - Spiked Removal Rats, Ai ~ Rats, Aw - Release Rate -

(Table 3.a) (Equation 3.b) 500 x Aw

- [sec] [Cl/sec) ' [Cl/sec) 1131 1.85E 5 2.89E 3 1.45 l132 1.01E-4 1.77E 2 8.85 l133 2.67E 5 6.67E-3 3.34 l134 2.38E-4 8.92E 3 4.46 1135 4.66E 5 6.39E 3 3.20

c. Neglecting the loss terms (radiodecay, leakage, and letdown), the time dependent RCS activity due to the accident initiated spike may be written as:

dA(t)

= 500Aw dt A t j dA(t) = j500Awdt A. O A(t) = A, + 500Awt Equation 3.c

d. Since the isotope activity, A(t) is assumed to remain evenly distributed throughout the RCS volume, then the rate at which the isotope activity leaks from the RCS, R(t), is the RCS leak rate constant, A., multiplied by the activity determined by Equation 3.c. The total activity released during the event is calculated by integrating the release rate over the time interval.

R(t) = AvA(t)

= A,(A, + 500Aut) t i JR(t)dt = J A,(A, + 500Awt)dt o o R = Ay At+ " Equation 3.d o

l REVISION NO.: 4 l

1 Exhibit E NEP-12-02 Revision 5 COMMONWEALTH EDISON COMPANY

] CALCULATION NO. : ATD-0410 PROJECT NO. PAGE NO.19 l TABLE 3.d Nuclide - RCS Activity at Spiked Release Rate _

0 2 hour:.

1 pcl/g,' A. - (Table 3.b) [Cl/sec) : Activity Relocsed, Ri

- (Table 2.b) [Ci] (Equation 3.d) [Ci]

l131 156.1 1.45 7.16 l132 174.8 8.85 4.27E1 1133 249.7 3.34 1.63E1 1134 37.5 4.46 2.14E1 1135 137.3 3.20 1.55E1

e. Calculate the thyroid inhalation dose at the Exclusion Area Boundary and Low Population Zone in accordance with UFSAR equation 15A-2.

TABLE 3.e Nuclide : 0-2 hour DCFi R,xDCFi

' Activity Released, R [ rem /CI) . J[ rem)

- (Table 3.d) [Cil-I131 7.16 1.08E6 7.73E6 l132 4.27E1 6.44E3 2.75ES l133 1.63E1 1.80E5 2.93E6 1134 2.14E1 1.07E3 2.29E4 l-135 1.55E1 3.13E4 4.85E5

. I(R ixDCFi) 1.14E7 The total DE l-131 activity released in 0-2 hours is the total weighted activity released, I(R, x DCF,), from Table 3.e divided by the 1-131 dose conversion factor. Numerically this is 10.6 C DE l 131 released in the first two hours (1,14E7 rem /1.08E6 rem /CI). The breathing rate is based on the 0-8 hour time period.

Exclusion Area Boundary Dose for a 1 com Leak Rate:

1 g

Dm(rem)= x B x E(R, x DCFi ) Equation 3.e.1 sQsm sec' *

- 5.7E - 4 x 3.47E - 4

.m.3 sec x 1.14E7[ rem]

= 2.25 (rem]

l REVISION NO.: 4 l

Exhibit E NEP-12-02 F evision 5 COMMONWEALTH EDISON COMPANY l CALCULATION NO. : ATD 0410 PROJECT NO. PAGE NO. 20 l

f. Calculate the thyroid inhalation dose at the Low Population Zone (LPZ) using the i trquation from UFSAR Section 15A.4. The activity released during the accident '

from 2-40 hours was obtained from UFSAR Table 15.1-4a. This activity includes the dose contribution from a 12.8 gpm primary to secondary leakage and secondary side release.

TABLE 3.f.1

- Nuclide - ' RCS lodine Activity ICRP-30, Dose : 2 40 Hour Weighted .

Released. Ri, Conversion Factor,  ; Activity Released - '

L(UFSAR Table DCFi , (Design input . R x DCFi 115.1 4a) [Ci]1 22) [ rem /Ci] - [ rem]

l-131 2.7E3 1.08E6 2.92E9 l132 1.4E3 6.44E3 9.02E6 l133 3.8E3 1.80E5 6.84E8 l-134 1.8E2 W E3 1.93E5 1-135 1.6E3 3."i 3E4 5.01E7 1 Total E(Ri x DCF) 3.66E9 The total 2-40 hour weighted activity released calculated above in Table 3.f.1 is separated into specific time periods of 2 8 hrs,8 24 hrs, and 24-40 hrs. This is based on scaling the total 2-40 hour weighted activity released by the fraction of steam released during the same time period. The 2-40 hour steam release was obtained from UFSAR Table 15.13.

TABLE 3.f.2 Time Period = Steam Release, Fracticn of Total - ' Total 2-40 Hr. Weighted

(UFSAR Table - Steam Release: Weighted = . Activity 15.13) [1b)

' ~

for Time Period Activity L  ; Released For Released (Table. Time Period.

3.f.1) [ rem] [ rem]-

2 8 hr 939,604 0.30 3.66E9 1.10E9 8-24 hr 1,234,515 0.39 3.66E9 1.43E9 24-40 hr 980,806 0.31 3.66E9 1.13E9 Total Steam : Total Weighted j Release ! 3,154,925 . Activity . 3.66E9 Released The atmospheric dilution factors (X/Q) for 0-8 hrs, 8-24 hrs, and 24-40 hrs were obtained from UFSAR Table 15.0-13. The breathing rates for 0-8 hrs,8-24 hrs and 24-40 hrs were obtained from UFSAR Table 15A 1. Calculate the thyroid inhalation dose at the Low Populaticn Zone (LPZ) using the equation from UFSAR Section 15A.4.

l REVISION NO.: 4 l

Exhibit E NEP-12-02 Revision 5 COMMONWEALTH EDISON COMPANY l CALCULATION NO. : ATD 0410 PROJECT NO. PAGE NO. 21 l I

Otpz(rem]=

x' x B x E(R; x DCF) a i Equation 3.f.1 A O /LPZ TABLE 343 Time :.. 1. .- Atmos. . Breathing Rate, .. Weighted c . LPZ Dose. D ez, t s

[ Period i Dispersion: B,(UFSAR : . .- Activity Release w/ a 12.8 gpm Leak -

! Factor, XIQ, Table 15A 1)- t I(RxDCF) Rate (Equation :

Y F(UFSAR Table - l [m'/sec); .;(Table 3.f.2)i i 3.f.1) [ rem](

.18.0-13) r > [ rem] .

  • ~

[sec/m*] '

j 0-2 hr 1.7E-5 3.47E-4 1.52E8* 0.90 2 8 hr 1.7E-5 3.47E 4 1.10E9 6.49 8-24 2.4E-6 1.75E-4 1.43E9 0.60 24-40 1.1 E-6 2.32E-4 1.13E9 0.29

/ Total LPZ Dose .

l L w/ a 12.8 Leak ; 8.28 Rate [ rem] '

  • From Tables 3.e and 1.c [(12.8x1.14E7) + (3.94E6+1.68E6) =1.52E8]
4. CALCULATION OF SITE ALLOWABLE LEAK RATE

~

a. Results of the Pre-Accident lodine Spike Model j The total EAB dose due to a 12.8 gpm leak rate and secondary side activity is 1

44.63 rem [(12.8x3.40)+1.11]. The total LPZ dose calculated in Table 2.f.3 is 7.41 rem. Therefore, the EAB dose is more limiting.

The thyroid dose due to the release of activity in the secondary side of all four steam generators is 1.11 rem (page 10). The dose due to 1 gpm primary to secondary leakage in 4 steam generators with a concentration of 60 Cl/g is 3.40 rem (page 14). Given that the dose limit in the Standard Review Plan is 300 rem for the pre-accident model, the maximum allowable leak rate without exceeding 300 rem is:

r 3 Allowable Leak Rate =

    • -I' *

3.40 **

s gpm ,

= 87.91 gpm l REVISION NO.: 4 l

Exhibit E NEP-12 02 Revision 5 COMMONWEALTH EDISON COMPANY l CALCULATION NO. : ATD 0410 PROJECT NO. PAGE NO. 22 j Consequently, the total EAB dose due to a 87.91 gpm leak during a MSLB is 300 rem. Allowing 0.1 gpm room temperature leakage per each of the three intact steam generators leaves 87.49 gpm (87.91 0.3x1.406) for the faulted loop.

Note that the 87.91 gpm allowable leak rate is calculated at RCS operating conditions. Should the allowable leak rate be desired to be expressed at room temperature conditions, the 87.91 gpm must be divided by 1.406 (Reference 18) to account for RCS density differences. Therefore, the room temperature allowable leak rate is 62.52 gpm.

b. Results of the Accident initiated lodine Spike Model The EAB dose due to a 12.8 gpm leak rate and secondary side activity is 29.9 rem [(12.8x2.25)+1.11). The total LPZ dose calculated in Table 3.f.3 is 8.28 rem. Therefore, the EAB dose is more limiting.

The thyroid dose due to the release of activity in the secondary side of all four steam generators is 1.11 rem (page 10). The dose due to 1 gpm primary to secondary leakage in 4 steam generators with a concentration of 1 C;/; is 2.25 rem (page 19). Given that the dose limit in the Standard Review Pla is 30 rem for the accident initiated spike model, the maximum allowable leak iate without exceeding 30 rem is:

f 5 Allowable Leak Rate =

2.25 **

s opm ,

= 12.84 gpm Allowing 0.1 gpm room temperature leakage per each of the three intact steam generators leaves 12.42 gpm (12.84-0.3x1.406) for the faulted loop.

Note that the 12.84 gpm allowable leak rate is calculated at RCS operating conditions. Should the allowable leak rate be desired to be expressed at room temperature conditions, the 12.84 gpm must be divided by 1.406 (Reference 18) to account for RCS density differences. Therefore, the room temperature allowable leak rate is 9,13 gpm.

l REVISION NO.: 4 l

i Exhibit E l NEP-12 01 Revision 5 COMMONWEALTH EDISON COMPANY -

l CALCULATION NO. : ATD-0410 PROJECT NO. PAGE NO. 23 l

5. CALCULATION OF END-OF-CYCLE 8 PREDICTED DOSES in accordance with the requirements for the Byron Unit i voltage based repair criteria (IPC) for outer diameter stress corrosion cracking at tube support plates, the potential tube leakage during a MSLB event with containment bypass must be predicted at the end of the next operating period in addition to the predicted IPC leakage, the MSLB leakage contribution from circumferential cracking at the top of the tubesheet must also be factored into the end of cycle leakage assessment. This combined predicted leak rate must be compared to and shown to be less than the maximum site allowable leak rate determined in Section 4 above.

Via a January 31,1997, transmittal to the NRC, Byron Station requested a Technical Specification change to lower the RCS Dose Equivalent lodine 131 limit to 0.2 pCl/g. As documented in Section 3, the site allowable leak rate of 12.8 gpm is based on an RCS DE l-131 limit of 1 pCl/g. The site allowable leak rate can be increased proportional to a reduction in RCS DE l 131. Therefore, by reducing the RCS DE l 131 limit to 0.2 Cl/g, the allowable leak rate is increased to 64 gpm (12.8 gpm/0.2). This amendment request is currently being reviewed by the NRC.

The Byron Unit 1 Attachment C Operability Assessment 97 044, which was transmitted to the NRC on July 30,1997, via Byron letter 97-0184, determined that the total predicted end-of cycle 8 leakage is 63.1 gpm operating conditions. This predicted leakage includes leakage due to IPC, circumferential indications and operational leakage (0.1 gpm) from each of the three intact SGs. This is bounded by the requested 64 gpm site allowable leakage limit at operating conditions. The end of cycle date used in the evaluation is the scheduled Unit 1 shutdown date of 11/07/97.

This section of the calculation determines the EAB, LPZ, and main control room thyroid dose for the predicted end-of-cycle leaksge of 63.1 gpm to validate that the current operating conditions are bounded by existing calculations. The EAB and LPZ dose is bounded by Section 3 of this document, which showed that the accident initiated spike is the limiting accident,

a. The most restrictive EAB thyroid dose limit is 30 rem per Section 4.b (page 22).

This dose limit corresponds to an allowable leak rate of 12.8 gpm at an RCS DE l 131 concentration of 1 Cl/g, The calculated EAB dose remains the same when allowable leakage is increased to 64 gpm because RCS DE l-131 is reduced a proportional amount. The EAB dose, XeAs, due to current cycle projected leakage of 63.1 gpm is calculated by performing a ratio of calculated values to projected values.

63.1 gpm , Xeas 64 gpm 30 rem XeAs(64 gpm) = (63.1 gpm)(30 rem)

Xsas = 29.6 rem at a 0.2 Cl/g RCS DE I-131 concentration l REVISION NO.: 4 l

Exhibit E NEP-12-02 Revision 5 COMMONWEALTH EDISON COMPANY l CALCULATION NO. : ATD-0410 PROJECT NO. PAGE NO. 24 l Therefore, the end of cycle 8 predicted EAB dose is within the 30 rem dose limit under end of cycle 8 operating conditions.

b. The LPZ calculated thyiold dose is 8.28 rem per Section 3.f (page 21). This dose value corresponds to an allowable leak rate of 12.8 gpm at an RCS DE l 131 concentration of 1 pCl/g, which again remains the same under the proposed allowable leak rate of 64 gpm because DE l.131 was reduced to 0.2 Cl/g DE l.

131. The LPZ dose for projected end of-cycle conditions, Xtrz, is calculated by performing a ratio of calculated values to projected values.

63.1 opm , X tpz 64 gpm 8.28 rem (X tpz)(64 gpm)= (63.1 gpm)(8.28 rem)

X tpz = 8.16 rem at a 0.2pCl/g RCS DE l-131 concentration The refore, the end-of cycle 8 predicted LPZ dose is within the 30 rem dose limit under end-of-cycle 8 operating conditions.

c. The control room dose methodology follows the same strategy as 5.a and 5.b above.

Calculation BYR97 332 (Reference 21) estimates the MSLB control room dose for an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> release. The dose was calculated by comparing the LOCA 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> release and corresponding dose to the MSLB event. The control room dose was determined to be 7.4 rem, which is less than the LOCA control room dose of 8.27 rem.

To ensure the control room dose, Xuca, with a projected leak rate of 63.1 gpm is bounded by the control room dose calculation, BYR97-332, the calculated values are ratio'd to the projected end-of cycle values.

63.1 gpm , Xuen 64 gpm 7.4 rem Xuca (64 gpm) = (63.1 gpm)(7.4 rem)

Xuca = 7.3 rem Therefore, the end-of cycle main control room dose remains bounded by the LOCA control room dose as presented UFSAR Table 6.41 and is less than the calculated dose in BYR97-332, revision 2.

l REVISION NO.: 4 l

i Exhibit E NEP-12-02 Revision 5 COMMONWEALTH EDISON COMPANY l CALCULATION NO. : ATD 0410 PROJECT NO. PAGE NO. 25 l

SUMMARY

AND CONCLUSIONS It is concluded from Section 4 that the accident initiated spike is more limiting, therefore the maximum site allowable SG leak rate during a postulated MSLB is 12.8 gpm at RCS operating conditions (9.1 gpm at room temperature) with a RCS DE l 131 concentration of 1 pCilg. This value includes the 0.1 gpm contribution from each of the three intact SGs.

Section 5 determined that the Unit 1end-of-cycle 8 predicted MSLB tube leakage results in off site thyroid doses that are less than a small fractior- (10%) of 10CFR100 limits. The resulting EAB and LPZ doses, with a 0.2 pCl/g RCS DE l 131 limit, are 29.6 rem and 8.16 rem, respectively, which are less than the 30 rem limit for the limiting accident initiated spike case. The control room dose resulting from the predicted end of cycle 8 MSLB leakage,7.3 rem, is continued to be bounded by the LOCA control room dose case.

FINAL -

l REVISION NO.: 4 I

_ - . -

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