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Text

Monthly Operating Repts for May 1983
	ML20077A839
Person / Time
Site:	San Onofre
Issue date:	06/15/1983
From:	Mayweather L, Ray H; SOUTHERN CALIFORNIA EDISON CO.
To:	; NRC OFFICE OF RESOURCE MANAGEMENT (ORM)
References
	NUDOCS 8307220465
	Download: ML20077A839 (25)
	v • d • e

~

AC40. R A NRC MONTHLY OPER ATING REPORT DOCKET NO 50-361 DATE 6/15/1983 COMPLETED BY I m yweather TELEPHOSE 71a/ao? 7700 EXT. 56223 OPE RA TING ST A TL'S San Onofre Nuclear Generating Station, Unit 2

1. Urut.Name 1 May,1983 through 31 May,1983

2. Repomr.g Penod 3,390

3. Licensed Thermal Power (Mht:

I.1 27 4 Nameplaie Rating (Gron Mw l e

5 Deugn Electne.a! Rating (Net Mwe) 1, (19 7 1.177

6. Maurr.un Dependable Capacity (Gron Mhe 1 (10 7

7. Maximum Deper6dable Capacey (Net Mhe

8. If Changes Occur in Capoeity Ratags (tienu Number 3 Throigh 7) Since Lm Report.Gr t Reasonr NA 9 Pome Leie' To w h.d Reuritied. lf 4e3 iNet Mw #

e Na 10 Re..m. F e.r Restr.:reen If An3 Tha Mor.th i r -ic-Da t e C uz-a:n e 744 3,623 6,033.2 11 Ho l-Re>m; Femud 1,453.13 2.436 &

M7 17 1: %mN 0 Hour R e s.-toe h as Cnt.:4.

0 0

0 13 Rear:o' Reer Shutse r Heun 598.48 1.075.38 1.899.48 14 Houn Gerierator Or trie 0

n 0

15 Urut Resene Shardeme Hom 1.142.700 1 Aa1 lon 9 771 72n 16 Gron Tbmna' Energi Generated (Mwm 17 G% Eecmes? Erie 3 Genersted (Mh%

335,600 500,'200

'695',212 299,900 417.000 543.020 lb Net Ek.v>ca' E nc gi Gemeested iMw Hi NA NA Na 19 trut Sen.ee F actor NA NA Na 20 Last Aiadame Factor 0

0 0

21 L'nat C apa m F ac toe e l ur.g M DC Net,

0 0

0 2.'

L ni C s;m:st s F a:10: 4L eg DER Neil 0

0 0

23 L'ast Furred Outage Rai, 24 Shuido.m Scheduled Oser Nes 6 Month iType. Dele. and Duration of Eachi NOME N3 25 Pf Shot Duur. 4: End Of Report Pmod Emmated Dete of Startup h Ues la Tes Starm IPruar se C===, mal Operatesi Facema Actaned 7/17/82 7/26/82 INIT14L NKm 9/

9/20/82 INITML Et.Ecntrin Under review pngggg gp B307220465 830615 PDR ADOCK 05000361 R

pon

AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.

50-361 UNIT SONGS - 2 DATE 6/15/83 p

CCEPIETED _BY L.Mayweather TELEPHONE 714/492-7709 Ext. 56223 Mott H MAY 1983 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL 3

( N Net)

(MWe-!kt) 1 0

17 348.92 2

0 18 63.23 3

0 19 581.96 4

102.10 20 453.58 5

59.96 21 433.04 6

325.13 22 437.81 7

408.65 23 438.88 8

434.54 24 440.56 9

146.48 25 454.67 10 0

26 515.85 11 336.06 27 584.02 12 430.48 28 622.60-4 13 424.96 29 839.10 14 486.40 30 858.13 15 725.21 31 862.60 16 855.71 i

i 0557u l

t

S0- M T

g

_p UNIT SHUTDOWNS AND POWER RFI)UCTIONS DATE 6/15 /R1 COeWLETED SY i

u m ms+ hop RFPORT MON rH MAV, 1083 TELEPHONE 714/492-7700 Ext. 56223

_h,,

} ri uen.-

c.

c--ac-,- -

1!

4(v e.....

w.

Inse, I.

g I.cne 2y 2

t Rec. e

AO Pievens Mrsvesense S

j$

.2 ggg d

i

~

10 830211 F

76.60 0

4 NA NA NA Unit awaiting removal of regulatory restrictions on reactor trip breakers and PASS following extended outage for maintenance on various systems.

11 830504 F

1.30 G

NA NA IIA VALVEX Turbine trip due to error in attemptin g to raise load with IIP stop valves at an 18% limit with bias applied.

12 830505 F

3.25 A

NA NA IIA VALVEX Torbine manually tripped due to #4 HP governor valve oscillations.

Reset Mod F on #4 HP governor valve.

13 830509 F

37.00 A

3 NA IA CKTBRK Failure of Channel B vital bus input power switch caused momentary loss of power to vital bus B.

i 14 830517 F

27.37 A

3 NA IIH VALVEX Reactor trip on low steam generator level caused by failure of feedwater concensate isolation vaive.

1 4

I 2

I.minhie f Insisiestonis Methen!

i: F.wced fleeve 8 Manuel Ine Prepar.sii.m i I Da.ta 5 $ctieduled A l.guipment FellweeIL=plein)

II.Me6eitensiace os Test 1 Manusi kram.

Lnin Sheets i n Iisrnser i Aui.un.ei

% eam.

1,cni Nep. e ti l Rll eic tNilHl(,

(*.Reinefing 4 Continuation from previous one.ii l

11.Resul. inn Meetilcel.m I.Oprias..e lenining A i1.enic I naminettaa month F Asinunitiensi"

!). Reduction of 20% or greater' "h d U S'"" * """'

64)pressi. ial l ie.e ll ipteint in the past 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months il4tiier. : = ri.in i 9.0ther (Explain)

.eec t.:.......,

SUPMARY OP (PERATIllG EXPERIENCE POR ' HIE MONni DOCKET to.

50-361 UNIT SONGS 2 DATE 6/15/1983 001PLETED BY L.Mayweather TELEPHONE 714/492-7700 Ext. 56223 May 1, 0001 Unit is in Mode 3 ct 5300, 2250 psia.

F Preparations for Mode 2 entry are in progress.

May 2, 0521 Entered Mode 2 May 2,

-0530 Reactor critical.

May 2, 0535 Entered Mode 3 to re-calculate Estimated Critical Position (ECP).

May 2, 0600 Conpleted ECP review and consnenced CEA withdrawal.

May 2, 0604 Entered Mode 2.

May 2, 0617 Reactor critical.

May 2, 1622 Safety Evaluation Report issued by NRC and administrative reactor power limit of 5%

removed.

May 3, 1850 Entered Mode 1.

May 4, 0436 Unit synchronized.

May 4, 1250

'Iurbine tripped during weekly valve exercise test as a result of an error in attempting to raise turbine load with !!P stop valves at an 18% limit with bias applied.

May 4, 1410 Unit synchronized.

May 4, 2300 Reactor power raised to 50%, turbine load at 450 MWe gross.

May 5, 0100 Tripped turbine as a result of erratic performance of 94 governor valve.

May 5, 0415 Unit synchronized after repair of #4 governor valve.

Sur nary of Opertting Expericnce for the Month of May,1983 Page 2 of 2 May 6, 0520 Reactor power raised to 50%, turbine load at 450 Kde gross.

May 9, 0950 Reactor and turbine trip due to failure Channel B vital bus input power switch.

May 10, 1818 Entered Mode 2.

May 10, 1834 Reactor critical.

May 10, 2122 Entered Mode 1.

May 10, 2255 Unit synchronized.

May 11, 0655 Reactor power at 50%, turbine load at 445 MWe gr oss.

May 14, 1300 Conpleted 50% power startup program testing.

May 15, 1830 Reactor power at 80%, turbine load at 880 Kde gr oss.

May 17, 1020 Reactor ad turbine trip on low steam generator level caused by failure of feedwater isolation valve.

May 18, 0932 Entered Mode 2.

May 13, 0945 Reactor eritical.

May 18, 1220 Entered Mode 1.

May IS, 1345 Unit synchroni ud.

May 18, 2120 Raised reactor power to 50%, turbine load at 470 Kde gross.

May 19, 0320 Turbine load 700 tide.

May 19, 0400 Experienced a large inflax of tuna crabs in the circulating water system causing high differential pressure on the :ain condenser and Corponent Cooling Water (CCW) heat exchangers.

May 19, 1030 Decreased reactor power to 50% and turbine load to 500 MWe gross to clean the main condenser water boxes and a CCW heat exchangers.

May 26, 1530 Raised reactor power to 60%, turbine load to 640 Kde gross.

May 28, 2045 Raised reactor power to 50%, turbine load to 920 MWe gross.

May 31, 2359 Unit is in Mode 1 at 80% power. Turbine load is 920 MWe gross.

9 o

cn p

we.

a-y,-.

q

.--e-

REFUELItG INFORMATION DOCKET 10.

50-361 UNIT SOtGS 2 DATE 6/15/83 CCMPLETED BY L.Mayweather TELEPHONE 714/492-7700 y

Ext. 56223 1.

Scheduled date for next refueling shutdcw1.

Not yet determined.

2.

Scheduled date for restart following refueling.

Not yet determined.

3.

Will refueling or resumption of operation thereafter require a Technical Specification change or other license amendment?

tbt yet determined.

What will these be?

Not yet datermined.

4.

Sdieduled date for submitting proposed licensing action and supporting informtion.

tbt yet determined.

5.

Important Licensing considerations associated with refueling, e.g. new or different fuel dasign or supplier, unreviewed design or performance analysis methods, significant changes in fuel design, new operating procedures.

Not yet determined.

6.

We nunber of fuel asse::blies.

a) In the core.

217 b) In the spent fuel storage pool. O.

7.

Licensed spent fuel storage capr. city. 800 Intended change in spent fuel storage capacity. NA 8.

Projected date of last refueling that can be discharged to spent fuel storage poole assuming present capacity.

g 0554u

O

\\

AC40 R A NRC MONTHLY OPER ATING REPORT DOCKET NO _50-362 DATE 6/15/1983 COMPLETED BY L.Ma vwea ther TELEPHONE 71a/dQ2-7700 Ext. 56223 OPERATING ST4 TUS

1. Uni Name _

SAN ONOFRE NUCLEAR GFNFPATfNC RTaTION, UNIT 3 1 Pbv,1983 throuah 31 Mav.1983

2. Reportmg Penod.

3. Licensed Thermal Power (Mht1 3.390 1,127

4. Nameplate Rating (Gron MWei 1,087 5 Devg: Electrical Ratmg(Net MheL 1,127 6 Masimum Dependable Capacity (Gron Mket.

1,087

7. Maximum Dependable Capacity (Nc MWei:

8. If Changes Occur in Capocity Ratings (Items Number 3 Through 7)Since Last Report. Gbe Reasons:

s NA 9 Power Leiel To which Restricted. lf Any (Net Mke

A 10 Re.,nn, For Restnet orts if An3 Na This Moe.th i r.to-Da r e C ur.da tiv e 744 3.623 4.751 11 Hou Ir Repu-ttn; Penod 0

0 0

12 Number Of Houn Reactor has Cntica!

0 0

0 13 Reactor Rewne Shutdowm Houn 0

0 0

14. Houn Generator On Law

15. Unit Rewne Shutdo.n Houn 0

o n

0 0

0 16 Gron Thermal Energs Generated (Mw}l, 17 (,rtm Electrical Energs Genersted (MWH 0

n n

O n

n Ib Net E6ectncal Energs Generated iMWH.

NA N3

.: n 19 Lmt Senece Factor NA NA NA 20 Laut Asadability Factor 0

0 0

21 Crut Capacits Factor tUsm; MDC Net, 0

r, 9

2? Umt C apeats Factor ilmg DER Net) 0 0

0 23 Unal Foered Outage Rate 24 Shutdoom Scheduled Over Nest 6 Month tTypc.Dete.and Dwatson of Eachi NONE 25 ff Shot Dunn 4: End Of Repoet Pmod.Esemated Date of Startup 36 Umsts la Test Starw> IPnme to Commemmi Operatioel Facecan Actoned Under review INITI A L CRITICAUTY Under review INITI4L ELECTRICITY COesMERCLAL OPER 4T10N Under review

,,.. i. i.......

AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.

50-362 UNIT SONGS - 3 DATE 6/15/83 COMPLETED BY L.Mayweather TELEPHONE 714/492-7700 Ext. 56223 MONTH MAY 1983 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)

(MWe-Net) 1 0

17 0

2 0

18 0

3 0

19 0

4 0

20 0

5 0

21 0

6 0

22 0

7 0

23 0

8 0

24 0

9 0

25 0

10 0

26 0

11 0

27 0

12 0

28 0

f 13

_0 29 0

14 0

30 0

i 15 0

31 0

16 0

UNIT SMUTDOWN% AND Potet.R RFIMsCTIONS DOCKET NO.

60 167 UMeiNAME SONGS - 3 DATE June 15 1983 RFFORi MON f}f f4AY l983 COMPLETtu sY L. h~^9 her TELEP900NE 714 /4 f)7.7700 Ext. 56223 "6

E

'F i ke,cc F-rei,= a r.ee m.

)sj 1-

'i W

ther E

Eh 4

I w as E5 Asin...

4 3$

elf g di p Rer - =

.A u

}'

Pie.e. sic....c 6

NA NA NA NA NA NA NA NA NA NA l

I i

4 i F.wced stees..n M rs h.=1 I.i. en F.so,ussi..is S Sdtedwied A l.gwepnient Fellwee il nplein) i Mennel t..e reepeeati.m..I Data R.Meletenes.ce.w Test

. Mannel % :sm I.ntes Shert. sin I s.rmec e ster.. clan, e Ans...n.e

%.4.n i.. serp, ogl M 31.sceNtexpg.

1) stepulos..eg Mess kilen 4 Continuat ton from prevlous Ult.:

I.Opress... T eetning A t i.enee i.seninsel.m rnonth F Adunnistreil<c s.3.Reduttlon of 20% or greater t.4 pre se..,e i..... il.pt.in g ei o.lir. e i.c.i.

in the past 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months I "'"' I l 5'" '""' #

i s

9.0ttier (f xpla tn) esec t.) en... et.e

I

)

SLNMARY OF OPERATI!C EXPERIE?CE FOR CIE MOtCH DOCKET to.

50-362 UNIT SONGS 3 DATE 5/13/1983 CCMPL3t. BY L.Maywether i

TELEPH0t:E 714/492-7700 Ext. 56223 the Unit is in Mode 5 in an extended outage.

1 i

i l

l-l l

l l

l f

I l

J 0617u

REFUELItG INFORMATION DOQ r to.

50-362 UNIT SotES-_3 DATE f /15/83 CCMPLETED BY L.Mayweather TELEPHONE 714/492-7700 I

Ext. 56223 1.

Scheduled date for next refueling shutdown.

!bt yet determined.

Scheduled date for restart following refueling.

2.

Not yet determined.

Will refueling or resumption of operation thereafter 3.

require a Technical Specification change or other license amendment?

!bt yet determined.

What will these be?

lbt yet determined.

Scheduled date for submitting proposed licensing action and 4.

supporting information.

!bt yet determined.

Irportant Licensing considerations associated with 5.

refueling, e.g. new or different fuel design or supplier, unreviewed design or performance analysis mthods, significant changes in fuel design, new oparating procedures.

!bt yet determined.

6.

We nurbar of fuel assemblies.

a) In the ccre. 217 b) In the spent fuel storage pool. O.

800 7.

Licensed spent fuel storage capacity.

Intended change in spent fuel storage capacity. NA Projected date of last refueling that can be discharged to 8.

E spent fuel storage poole assuming present capacity.

0554u g.

T t

r

-+--W--

= -

e ATTACHMENT G

OFFSITE DOSE CALCULATION MANUAL REVISION 10 0996u 5/18/83

OFFSITE DOSE CALCULATION MANUAL SAN ONOFRE NUCLEAR GENERATING STATION UNITS 2&3 LIST OF EFFECTIVE PAGES i

PAGE N00 REV. DATE PA'JE NO.

REV. DATE PAGE NO.

REY. DATE

TitlO Page 11-15-82 2-1 7-1-82 2-23 7-1-82

I 11-15-82 2-2 11-15-82 2-24 7-1-82 iII 7-1-82 2-3 11-15-82 2-25 7-1-82 III 11-15-82 2-4 11-15-82 2-26 7-1-82

. IY 7-1-82 2-5 11-15-82 2-27 7-1-82 1-1 7-1-82 2-6 11-15-82 2-28 7-1-82 1-2 7-1-82 2-7 7-1-82 2-29 7-1-82 1-3 7-1-82 2-8 7-1-82 2-30 7-1-82 1-4 5-18-83 2-9 7-1-82 2-31 7-1-82 1-5 7-1-82 2-10 5-18-83 2-32 7-1-82 1-6 5-18-83 2-11 7-1-82 2-33 7-1-82 1-7 11-15-82 2-12 7-1-82 2-34 7-1-87 1-8 5-18-83 2-13 7-1-82 2-35 7-1-82 1-9 5-18-83 2-14 7-1-82 2-36 7-1-82 1-10 5-18-83 2-15 7-1-82 2-37 7-1-82 1-11 5-18-83 2-16 7-1-82 2-38 7-1-82 l1-12 5-18-83 2-17 7-1-82 2-39 7-1-82 1-13 5-18-83 2-18 7-1-82 2-40 7-1-82 1-14

,5-18-83 2-19 7-1-82 2-41 7-1-82 1-15 5-18-83 2-20 7-1-82 3-1 7-1-82 1-16 7-1-82 2-21 7-1-82 4-1 7-1-82 1-17 7-1-82 2-22 7-1-82 4-2 1-29-82 1

1 Rev. 10 5/18/82 Page 1 of 2 l

L, h

[6,

'l s Step 2) 1he adjustment factne, A, f or each batch tank (or

[:

'~\\

3 s

x'\\

'h susp) is determinid using:j i

N' g

5' A=

IC

+C

+

C

+C

+ C (1-3)

MC, y

i

/

A

~

the limiting concentrations of tha appropriate MP0(g, HPC,, MPC,

=

g h*,

radionuclide from 10CTR20, Appendix B,. Table II, MPCg,, }PC a Column 2.

For dissolve.d or *ntrained noble

-z

' gases, the concentration shall be ' limited ' to -

^

2.0 E-4pci/mi total activity.

"e l,

i

~

r Step ~3) The radioactivity sonitor setpoint may now be

(

specified based <:n the values of -

{Cy, T, A and R to provide compliance with the g

' limits of 10CFR20... Appendix B. Table 11, Column

~~

n 2.

The monitor setpoint (cpm) is taken from the w 'spplicable calibration constants given in Table 1-1 to correspond to the calculated monitor s

.- limit C,.

6, s...

-~

_s

(

/

I-2/3 RT - 7813 t $$s

[ '-

i:

0.7 Y C

~~~

ff

- 4)

R C

<R Ag+R2 2 + *** +

C' A

^

a g

n n

i.

lkfg 3.-

1-4 Rev. 10 j;

05/18/83 y.

a.

e'

. u

s A, A, etc. = Value of A from equation (1-3) for first 3

2 r

tank, second tank, etc.

The 0.7 is an administrative value used to R

s account for the potential activity for other releases. This assures that the total concentra-s i

tion from all release points to the plant discharge will not result in a release of concen-1

~-

trations exceeding the limits of 10CFR20, Appendix B. Table II, Column 2 from the site.

t i

NOTE:

If C, j[ C,gg then no release is t

possible. To increase C, increase dilution flow F (by running more f

circulating water pumps in the C

applicable discharge structure), and/or t

decrease the effluent flow rates R, R '

1 2

i etc. (by throttling the combined flow as measured on 2/3 FI-7643), and i

recalculate C, using the new F, R and i

equation (1-4).

P l

d 1

i l

9 k

f 1-6 Rev. 10 05-18-83 l

1

F then no release is NOTE:

If C,<{C g possible.

To increase C,,

increase dilution flow F (by running more f

circulating water pumps), and/or

[

~

)

decrease the effluent flow rate R (by i,

l' throttling the flow as measured on f-h 2F1 3772 and 3F1 3772), and recalcula*.e C,using the new F. R and equation (1-6).

If there is no release associated v.ith this monitor, the monitor serpoint should be established as cicae to background as practical to prevent spurious alarms and yet assure an alarm r

should an inadvertant release occur.

1.1.2 Continuous Release Setpoint Determination h

c

(

l Step 1) The isotopic concentration for the continuous releases are obtained for each release stream i

l (steam generator blowdown, and steam generator R

blowdown bypass, and turbine building sump) from the sum of the respective measurad l

concentrations as determined by analysist I

P I

(

1-8 Rev. 10 05-18-83

~. - -.

II~ )

C={CY g + C, + Cg + C, + CFe i

I where:

,e i

the total gamma activity ( pCi/cc) associated IC yg I

with each radionuclide, i, in the weekly I

i composite analysis for the release stream.

C, =

The total measured gross alpha concentration l

(pCi/cc) determined f rom the previous monthly composite analysis for the releans stream.

I i

l f

The total Fe-55 concentration as determined in CFe

j l

the previous quarterly composite sample for the i

release stream.

1 i

the total measured H-3 concentration (pCi/cc)

C

=

j e

if determined from the previous monthly composite I

i analysis for the release stream.

.I j

4 1

C, =

the total measured concentration (pCi/cc) of Sr-89 and Sr-90 as determined f rom the previous j

quarterly composite analysis for the release stream.

a i

1-9 Rev. 10 06/18/83 l

1

-m y

-w--

ww s

,w~,

e wm

w---

2-w

+

3 Step 2) The adjuetment factor, B, for each release stream (steam generator blotidown or turbine I,

building su=p) is determined using:

t I-C, (1-8)

B=

IC

+

C,

+

C C

+

p yg t

a b4PC MPC MFC MPC MFC i

e t

a Fe l.

i Step 3) The setpoint for each continuous release

I' radioactivity monitor may now be specified based on the respective values of {CYg, F, B t

Ir and R to provide compliance with the limits of r

10CFR50, Appendix B, Table II, Column 2.

The t.

monitor setpoint (cpm) in taken from the applicable calibration constants given in g

Table 1-1 to correspond to the calculated I

L monitor limit, C,.

c.

2RT - 7817, 3RT - 7817 b

(0.5)(0.1)F{ ctg (g_9,)

C, <

g RB t.

$q Where I

= values of IC ac4 d (as defined in Steps ICYg, B Yg i

i 1 and 2 above) for the steam generator blowdown.

?

1-10 Rev. 10 t-05-18-83 i.

I

s Table 2-1 Caseous Effluent Radiation Monitor Calibration Constants r

Monitor Kr-85*

Xe-133*

2RT-7804-1C 3.07 E-8 3.86 E-8 R

3RT-7804-1C 2.05 E-8 1.67 E-8 2/3RT-7808C

2. 76 E-8

3. 72 E-8 e

2/3RT-7814A 3.37 E-8 4.30 E-8 lR 2/ 3RT-7814B 4.24 E-5 3.61 E-5 7

2RT-7818A 3.06 E-8 5.30 E-8 2RT-7818B 6.92 E-5 5.11 E-5 lR 3RT-7818A 3.14 E-8 4.56 E-8 3RT-7818B 3.00 E-5 2.83 E-5 l

2RT-7865-1 (low) 1.41 E-8 3.02 E-8 p

2RT-7665-1 (mid) 5.33 E-5 2RT-7865-1 (high) 6.81 E-2

[

e 3RT-7865-1 (low) 1.41 E-8 3.02 E-8 c'

3RT-7865-1 (mid) 8.02 E-5 i

- 3RT-7865-1 (high) 2.39 E-2 g

L

[4 2RT-78 70-1 (low) 1.41 E-8 3.02 E-8 2RT-78 70-1 (mid) 1.07 E-4 f-2RT-7870-1 (high) 2.87 E-2 l

3RT-78 70-1 (low) 1.41 E-8 3.02 E-8 5,

3RT-78 70-1 (mid) 1.08 E-4 i

3RT-7870-1 (high) 2.17 E-2

(j T1

pCi/ce/cpa

{.

I

$~

2 y

l'.-H il 2-10 Rev. 10 f.

05-18-83

~ " " " ' -

w...

1 5

R = 400 gpa where R is the effluent flow rate at the radiation monitor as defined in Step 2.

The 0.1 is an administrative value to account for the potential activity in other release pathways. This d

assures that the total concentration from all release points to the plant discharge will not result in a release of concentrations exceeding the limits of 10CFR20, Appendix B, Table II, Coluen 2 from the site.

0.5 is an administrative value used to account for t

simultaneous releases from both SONGS 2 and SONGS 3.

i t

F then no release is possible. To NOTE:

C,<{C7g l-l increase C, increase the dilution flow F (by m

running more circulating water pumps), and/or decrease the effluent flow rate R (by throttling the flow as measured on 2FI-3772 or 3FI-3 772, as appropriate) and recalculate C,

(

using the new values of F, R and equation i

(1-9a).

b L

i f,

1-11 Rev. 10 h

05-18-83

[.;

r

,,-._,,,------.--_,c,

.--.____,--,,,,.r_,_

2RT - 6753, _2RT - 6 759, 3RT - 6753, 3RT - 6759

(

(0.5)(0.05)F{CYg q _

RB Where:

t.

h and B (as defined in B = values of {Cyg

{Cyg, Steps 1 and 2 above) for the steam generator blowdown bypass.

Vl R = 200 spa N

f

[

where R is the maximum per steam generator blewdown t-bypass ef fluent flowrate.

}

br The 0.05 is an administrative value to account for the lj potential activity in other release pathways. This assures that the total concentration from all release

{

points to the plant discharge will not result in a v

release exceeding the limits of 10CFR20, Appendix B, Table II, Column 2 f rom the site. The 0.5 is an administrative value used to account for simultaneous l

L-releases from both SONGS 2 and SONGS 3.

f' l'

[.

l _

l-12 Re v. 10 05-18-83

e.

Q t.

then no release is possible.

Note:

If q, jC {C7g 6

To increase C,, increase the dilution flow F

[

(by running more circulating water pumps),

i and/or decrease the ef fluent flow rate R (by N

4 throttling the flow as measured on 2FIC-4055, y;

2FIC-4056, 3FIC-4055 or 3FIC-4056, as i

appropriate), and recalculate q,in equation

+

1-9b) using the new values of F, R.

?

2RT - 7821, 3RT - 7821 l

(0.5) (0.1) F " C T

(1-10)

C <

a RB

[.!

Wheret

.;e IC g,B=

values of IC and B (as defined in j

g i

i steps 1) and 2) above) for the turbine building sump i:t b

R

= 50 gpm/ pump (x no. sump pumps to be run) t.;

!4, i,

i,c

.i N.

l' 1-13 Rev. 10 05-18-83

d

)

The 0.1 is an administrative value to account for the t

I, potential activity in other release pathways. This assures that the total concentration from all release points to the plant discharge will not result in a e

release of concentrations exceeding the limits of l

i, 10CFR20, Appendix B, Table II, Column 2 from the site.

0.5 is an administrative value used to account for simultaneous releases from both SONG 3 2 and SONGS 3.

I then no release is possible. To j.

NOTE:

If C,<{CYg r-increase C,, increase the dilution flow F (by running more circulating water pumps) and y

f^

recalculate C,using the new value of F and equation (1-10).

La I

i'

.~

i I

k L-3-

l-

?

5 hi E'

I+A c:

c li 1-14 Re*

10 f.

05/18/83 f

.,c.

r-

t h

f Table 1-1 Liquid Ef fluent Radiation Monitor Calibration Constants

'1 0

'tT Monitor Co-60*

Ba-133*

Cs-137*

I 1

2RT-6753 1.86 E-8 1.96 E-8

.M V

2RT-6759 1.79 E-8 1.91 E-8 E

f.

y 3RT-6753 Note 1 Note 1 3RT-6759 Note 1 Note 1 2/3RT-7813 2.08 E-9 3.14 E-9 4.59 E-9 2RT-7817 2.11 E-9 3.20 E-9 4.71 E-9 2RT-7821 2.08 E-9 3.17 E-9 4.61 E-9 3RT-7817 2.24 E-9 2.99 E-9 4.63 E-9 i

3RT-7821 2.15 E-9 3.30 E-9 4.72 E-9

}

)Ci/cc/ cpm i.

6-Note 1 -- These calibration constants will be provided prior to utilizing Unit 3 Steam Generator blowdown bypass lines.

p i:

t:

5-t i.I l

1-15 Rev. 10 l

05-18-83

.

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