Text

Amend 39 to License DPR-70,incorporating Requirements for Implementing TMI-2 Lessons Learned Category a Items
	ML20031G776
Person / Time
Site:	Salem
Issue date:	10/08/1981
From:	Varga S; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20031G777	List: ML20031G776; ML20031G778;
References
	NUDOCS 8110230651
	Download: ML20031G776 (36)
	v • d • e

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

k3 Kf 2g

.,(

o, UNITED STATES j

NUCLEAR REGULATORY COMMISSION

(

3 E

WASHINGTON, D. C. 20555 E

o s.,

/

PUBLIC SERVICE ELECTRIC AND GAS COMPANY PHILADELPHIA ELECTRIC COMPANY DELMARVA POWER AND LIGHT COMPANY ATLANTIC CITY ELECTRIC COMPANY DOCKET N0. 50-272 SALEM NUCLEAR GENERATING STATION, U:4IT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE r

Amendment No. 39 License No. DPR-70 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment by Public Service Electric and Gas Company, Philadelphia Electric Company, Delmarva Power and Light Company and Atlantic City Electric Company (the licensees) dated September 29, 1981, complies with the standards and require-ments of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulntions of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public and (ii) that such activities will be conducted in compliance, th the Comission's regulations; D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical' Specifications as indicated in the attachment to this license amendment, and paragra 2.C.(2) of Facility Operating License No. DPR-70 is hereby amended to read as follows:

8110230651 811000 PDR ADOCK 05000272 P

PDR

. (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 39, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

The license is also amended by the addition of new paragraphs 2.C.(7),

2.C.(8) and 2.C.(9) that read as follows:

(7) Systems Integrity The licensee shall implement a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. This program shall include the following:

1.

Provisions establishing preventive maintenance and periodic inspection mquirements, and 2.

Integrated leak test requirements for each system at a fmquency not to exceed refueling cycle intervals.

(8)

Iodine Monitoring The licensee shall implement a program whicn will ensum the capability to accurately detemine the airborne iodine concentra-tion in vital areas under accident conditions. This program shall include the following:

1.

Training of personnel; 2.

Procedums for monitoring, and 3.

Provisions for maintenance of sampling and analysis equipment.

l (9) Backup Method for Determining Subcooling Margin The licensee shall implement a program which will ensure the tapability to accurately monitor the Reactor Coolant System subcooling margin. This program shall include the following:

1.

Training of personnel, and 2.

Procedures for monitoring.

D i

4.

This license amendment is effective as of the date of its issuance.

FIR THE NUCL F

GULATORY COMMISSION l l l

O even A.

a, hie <

. Operating Reactors Branch #1 Division of Licensint

Attachment:

Changes to the Technical Speci fications Date of Issuance:

October 8,1981 i

(

l l

l

l0 t

ATTACHMENT TO LICENSE AMENDMENT NO. 39

?

FACILITY OPERATING LICENSE NO. DPR-70 DOCKET NO. 50-272 Revise Appendix A as follows-.

Remove Pages Insert Pages IV IV X

X 3/4 3-20a 3/4 3-22 3/4 3-22 3/4 3-26 3/4 3-26 3/4 3-27 3/4 3-27 3/4 3-28 3/4 3-28 3/4 3-29 3/4 3-29 3/4 3-30 3/4 3-30 3/4 3-31 3/4 3 -31 3/4 3-32 3/4 3-32 3/4 3-32a 3/4 3-33 3/4 3-33 3/4 3-34 3/4 3-34 3/4 3-53 3/4 3-54 3/4 3-55 3/4 3-56 3/A 3-57 3/4 4-4 3/44-4 3/4 4-4a 3/4 4-5 3/4 4-5 3/4 4-6 3/4 4-6 3/4 6-12 3/4 6-12 3/4 6-13 3/4 6-13

uw m.Wc

-.de J

J4J hL.L 4

.#J.

%A2A oJA w--.

J A

A+E.4 D3m-L.

4 A

t i

g.

ATTACHMENT (CONTINUED) 4 I

Remove Pages Insert Pages a

f 3/4 6-17 3/4 6-17 l

3/4 7-5 3/4 7-5 1

3/4 7-6 3/4 7-6 I

B3/4 3-3 B3/4 3-3 B3/4 4-la B3/4 4-la B3/4 4-2 83/4 4-2 e

i, 1*

I I

l f

1 i

s r

I i

l l

l 1

i-i l

8 i

f f

1 i

i

{

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS SECTION Pace 3/a.2 POWER DISTIR30 TION LIMITS 3/4.2.1 AXI AL FLU X DI FFEREN CE...................................... 3/4 2-1 3/4.2.2 H EAT FLUX HOT CH ANN EL FACT 0 R...............................3/4 2-5 3/4.2.3 NUCLEAR ENTHALPY HOT CHANNEL FACT 0R........................ 3/42-9 3/4.2.4 Q UA D RANT POW ER TI LT RATI 0.................................. 3/4 2-11 3/4.2.5 DNS PARAMETERS.............................................

3/4 2-13 3/4.3 INSTRUMET ATION 3/4.3.1 REACTC 3'P SYSTEM INSTRUMENTATION........................

3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION............................................

3/4 3-14 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation.......................

3/4 3-35 Mov a bl e Inco re Det ectors................................... 3/4 3-39 Seisnic Instrumentation.......,............................

3/4 3-40 Meteorol ogical Instrumentation.............................

3/4 3-43 Remote Shutdown Instrumentation............................

3/4 3 46 Fire Deteccion Instrumentation.............................

3/1 3-49 Acci dent Mon i torin g In st rumentation........................ 3/4 3-53

\\

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS N o rm a l O p e r a t i o n........................................... 3/4 4-1 3/4.a.2 S A F E TY VA LV E S.............................................. 3/A 4-4 3/4.4.3 REL I E F V ALV E S..............................................

3/4 4-5 1

3/4.4.4 PRESSURIZER................................................

3/4 4-6 3/4.4.5 STEAM GENERATORS...........................................

3/4 a-7 3/a.a.6 REACTOR CCOLANT SYSTEM LEAKAGE Lea ka ge Det ecti on Sys t ems.................................. 3/4 4-14 Operational Leakage....................................... 3/4 4-15 Pressure Isolation Valves..........

....................... 3/4 a-16a S 'LE." - U'!! T I IV Amendment No. 39

(

INDEX BASES SECTION PJGE 3/4.3 INSTRUMENTATION 3/4.3.1 P R311 TIVE INSTRUMENTATION..............................

B 3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURES (ESF) INSTRUMENTATION........ B 3/4 3-1 3/4.3.3 MONITORIN G INSTRUMENTATI ON..............................

B 3/4 3-1 3/4.4 REACTOR COOLANT SYSTEM

~

3/4.4.1 REACTOR COOLANT L00PS...................................

B 3/4 4-1 3/4.4.2 S A FET Y V AL V E S............................................. B 3/4 4-la 3/4.4.3 RE L I E F V AL VE S.............................................. B 3 / 4 4 -l a 3/4.4.4 PRESSURIZER.............................................

B 3/4 4-2 3/4.4.5 STEAM GENERATORS........................................

B 3/4 4-2 3/4.4.5 REACTOR COOLANT SY STEM LEAKAGE..........................

B 3/4 4-3 3/4.4.7 CHEMISTRY...............................................

B 3/4 4-4 3/4.4.8 SPECIFIC ACTIVITY.......................................

B 3/4 4-5 3/4.4.9 PRESSURE / TEMPERATURE LIMITS..........................

B 3/4 4-6 3 / 4. 4.10 STRU CTU RAL I NT E G R ITY..................................

B 3/4 4 SALEM - UNIT I X

Amencment No. 39

I y

TABLE 3.3-3 (Continued)

ENGINEtREO SAFETY FEATURE ACTUAIION SYSifM INSTRUNENTATION i

E HINIHi#l TOTAL No.

CilANNELS CilANNELS APPLICAHLE FDHCIIONAL UNil 0F CitANNftS 10 1 RIP OPERABLE HDDES ACTION 8.

AUXILIARY TEEDWATER a.

Automatic Actuation logic **

2 1

2 1, 2, 3 20 b.

Sim. Gen. Water i

tevel-tow-tow I

i.

Start Motor mg Driven Pumps 3/sta. gen 2/sta. gen.

2 sta. gen.

1, 2, 3 14*

i any sim. gen.

E

11. Start Turlaine-O Driven Pumps 3/sta. gen.

2/sta. gen.

2 sim. gen.

1, 2, 3 14" any 2 sim. gen.

c.

Undervoltage-NCP Start Turbisie-l Driven Pus,)

4-1/ bus 1/2 x 2 3

1, 2 19 i

d.

S.1.

I Start Motor-Driven Pimps See 1 above (All S.I. Initiating functions and requirements) e.

Emergency Trip of Steam Generator s

Feedwater Pumps - start

[

Motor Driven Pumps 2-1/ pump 2

2 -1/ pump 1

21 O

f.

Station Blackout See 6 and 7 above (SEC and U/V Vital Bus)

"m

- Applies to items b. and c.

TA8LE3.3-3(Continuedl ACTICN 17 With less than the Minfaus Channels OPERA 8LE, operation may continue provided the containment purge and exhaust valves are maintained closed.

ACTION'18 With the number of CPERA8LE Channels one 16ss than the Total Nuster of Channels, restore the inoperable channel to CPERA8LE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least uCT STANOSY within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and in COLD $NUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

ACTION 19 With the ausser of CPERA8LE Channels one less than the Total Number of Channels, STAATUP and/or POWER OPERATICN say proceed provided the following conditions are satisfied:

The inoperable channel is placed in the tripped condition a.

within I hour, b.

The Minfaum Channels CPERA8LE requirements is met; however, one additional channel say be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.2.1.1.

ENGINEERED '5AFETY FEATURES INTERLCCxs CESIGNATTCN CONCITICN AND SETPOINT FUNCTICN P-11 With 2 of 3 pressurizer 9-11 prevents or defeats pressure channels 3 1925 sanual block of safety psig.

infection actuation on low

~

pressurizer pressure.

P-12 With 3 of 4 T"'I channels F-12 prevents or defeats

> 545'F.

annual block of safety

~

fnjection actuation high staas ifne flow and low staan Ifne pressure.

With 2 of 4 T channels Allows annual block of safety

< 541*F.

avg infection actuation on hign staas line flow and low steam ifne pressure. Causes staam ifne isolation on high steam flow. Affects steam dump blocks.

ACTION 20 With the number of CPERA8LE channels one less than the Total Nv-3er'of Channels, be in at least HOT STANC8Y within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and Ic at least HOT SHUTDChN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; how-ever, s'io channel say $4 bypassed for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months for surveillanca tasting.

ACTION 21 With the nunber-ef TERABIZ channels cme less than the Mini:msn Nmber of Channels, coeration may croceed prev.ided that either:

a. The incperable channel is restored to CFERABIZ within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , cr

b. If the affected Steam Generator Feedwater Purp is expected to be cut of ser/ ice for more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , the incperable channel is jmpered so as to enable the Start Circuit of the Auxil.tarf Feed-

)

water Pungs upcn the loss of the other Steam Generator Feedwater Purp.

3/4 3-22 Amendment No. 39

y IAntE 3.3-4 (Continueil)

)

[NGINEFRfD SAFEIY fEAIUNE ACIUATION SYSIEM INSINUHfHIAIION INIP SEIPOINIS Eq IUNCil0NAl 1) Nil IRIP SEIPolNI Att0WABIE VAtHES

~

S.

IURBINE IRIP AND FEEDWATER 150tATION 4

a. Steam Generator Water level--

< 67% of narrow range

< l'8% of narrow range liigh-liigh Instrumesit spasi eacle steam Isistrimient span each Uenerator s t east gerierator 1

6.

SAFEGUANDS IQUIPHLNI CONIN01.

SYSitH (SIC)

Not Applicable Not Applicable t

7.

UNDfHVOLIAGE, VIIAt. BUS

{

a.

loss of Voltage 1 70%

3 65%

[

8.

AUMit IARY IELDWAIEN o,

4.

Aistomatic Actuation logic Not Appilcable Not Applicable b.

Steam Generator

> 18% of starrow range

> 17% of narrow range Water level-low-low instrusesit span eacli Tnstrimient span each steam gesierator steam gesterator c.

Undervoltage - RCP 1 70% RCP bus voltage 1 65% RCP bus voltage d.

S.I.

See 1 Above (All S.I. setpoints) e.

Dnergency Trip of Steam Generator Not Applicable Not Applicable s

Feedwater Pumps

<+

f.

Station Blackout See 6 and 7 above (SEC and Undervoltage, Vital Bus) l l

l l

l

i TA8tE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND F1)MCTION RESPONSE TIME IN SECONDS

-.--. ~.....

1.

Manual a.

Safety Injection (ECC5)

Not Applicabte Feedwater Isolation Not Applicable Reactor Trip (51)

Not Applicable Containment Isolation-Phase "A" Not Applicable Containment Ventilation Isolation Not Applicable Auxiliary Feedwater Pumps Not Applicable Service Water System Not Applicable Containment Fan Cooler Not Applicable b.

Containment Spray Not Applicable Containment Isolation-Phase "S" Not Applicable Containment Ventilation Isolation Not applicable Containment Is'olation-Phase "A" Not Applicable c.

Containment Ventilation Isolation Not Applicable d.

Steam Line Isolation Not Acplicable 2.

Containment Pressure-Hich a.

Safety Injection (ECCS) 3 27.0*

10 2

b.

Reactor Trip (from SI) 10 7

Feedwater Isolation d.

Containment Isolation-Phase "A" 1 17.0#/27.0##

~

Cetainment Ventilation Isolation Net Acclicable e.

f.

Auxi'siary Feedwater Pumps Not Acclicable 13.0 / 8.0" g.

Service Water System 1

UNIT I 3/4 3-27 Amendment No. 39 f

SALE'd l

l l

O m

e

9 TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES RESPONSE TI1E IN SECONDS INITIATING SIGNAL AND FUNCTION 3.,

Pressurizer Pressure-Low Safety Injection (ECCS) 1 27.0*/,12.0#

a.

b.

Reactor Trip (from SI) 12.0 7

1 0-Feedwater helation c.

d.

Containment Isolation-Phase "A" 1 18.0!

Not Applicable Containment Ventilation Isolation e.

Not Applicable

[

f.

Auxiliary Feedwater Pumps 149.0"/13.0#

g.

Service Water System Differential Pressure Between Steam Lines High_

-4.

Safety Injection (ECCS) 112.0#/22.04 a.

b.

Reactor Trip (from SI) 12.0 17.0 Feedwater Isolation c.

Containment Isolation-Phase "A' 117.0#/27.04 d.

Net Applicable Containment Ventilation Isolatten e.

Not Accitcable f.

Auxiliary Feedwater Pu=;s 1 13.0U48.08, g.

Service Water System Steam Flew in Two Stea Lines - Mich Coincident 5.

witn I,yq--tow-Low 114.0#/24. nae Safety Injection (ECCS) a.

b.

Reactor Trip (from SI) 11.0 19.0 '

Feedwater Isolation

~

c.

Containment Isolation-Phase "A" i 19.0#/29.0##

d.

tiet Applicable Containment Ventilation Isolation e.

Not Applicable f.

Auxiliary Feedwater Pumps ij4.0#/49.0##

g.

Service Water System i 9,n h.

Steam Line Isolation SALEM - UNIT 1 3/4 3-23 Amencent Nc. 39

TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES RESPONSETIMEINSEC0b5 INITIATING SIGNAL AND FUNCTION 6.

Steam Flew in Two Steam t.ines-High Coincicent witn Steam Line Pressure-Low a.

Safety injection (ECCS) 1 12.0#/22.0##

b.

Reactor Trip (from SI) 12,0 710 c.

Feedwater Isolatic,n d.

Containment Isolation-Phase "A" 117.0#/27.0de e.

Containment Ventilation Isolation Not Applicacie f.

Auxiliary Feedwater Pumps Not Acolicable g.

Service Water System i 14.Or/48.0==

l h.

Steam Line Isolation 1 8.0

{

7.

Containment Pressu m--Hich Wich a.

Containment Spray 1 45.0 b.

Containment Isolation-Phase "S" Not Acolicable c.

Steam Line Isolation 1 7.0 d.

Containment Fan Cooler 1 40.0 8.

Stean Generator Water Level--Hich Wich a.

Turoine Trip-Reactor Tri; 1 2.5 Feecwater Isolation

< 11.0 9.

St.eam Generator Water Level --t.ow-t.ow a.

Motor-Oriven Auxiliary Feedwater 1 60.0 Puaos b.

Turbine-Oriven auxiliary Fetcwater 5 60.0 Punos 1

SALEM - UMT 1 3/4 3-29 Amendment No. 39

TA8LE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONGS

10. Undervoltage RCP Bus a.

Turbine-Oriven Auxiliary Feedwater 1 60.0 Pumps

11. Containment Radioactivity - High a.

Contairment Pressure-Vacuum Relief 5 5.0 (" *)

System Isolation

12. Trip of Feedwater Pures a.

Auxiliary Feedwatar Punos Not Applicable

13. Undervoltane, Vital Bus a.

Loss of Voltage 1 4.0

[

14.

Station Blackout a.

Motor Driven Auxiliary n60.0 Feedwater Pumps t

Nota:

Response time for Motor-driven Auxiliary Feedwater Pumps on all 5.I.

signal star *.s 1 60.0 l

l SALEM - UNIT I 3/4 3-30 Arendment No. 39

TABLE 3.3-5 (Continued)

TABLE NOTATION

(* )) Diesel generator startin; and sequence loading delays included.

Response

time limit includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal enarging pumps, SI and RHR pumps.

(#)) Diesel generator starting and sequence loading delays not included.

l Offsite power available. Response time limit includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging pumos.

(**))

Diesel generator starting and sequence loading delays included.

Response

time limit includes opening of valves to establish SI path and attainment of discharga pressure for centrifugal charging pumps.

(***))

On 2/3 in any steam generator.

(**))

On 2/3 in 2/4 steam generators.

(***))

The response time is the time the isolation circuitry input reaches the isolation setpoint to the time theContainment Pressure-Vacuum Relief l

valves are fully shut.

l SALEM - UNIT I 3/4 3-31 Amendment No. 39

SI TABLE 4.3-2(Continuedl r-9 ENGINEERID SAffiY FEATURE ACTUATION SYST[M INSTRUMENTATION c

~~

~ '~' ' " _511NY$$I.4NCE_,R_,00_{Rf Mf MI S f

M CHANNEL MODES IN NHICH CilANNEL CHANNEL IUNCil0NAL SURVEILLANCE ClfCK CAllBHA110N TESI REQUIRED FUNCTIONAL UNIT i

3.

CONTAINMENT 150LATl04 a.

Phase "A" Isolation M.A.

M.A.

R

1. 2. 3. 4

1) Manual id

2) From Safety injection N.A.

N.A.

M(2) 1.2.3.4 Automatic Actuation Logic ya ki b.

Phase "B" Isolation

1) Manual N.A.

N.A.

R

1. 2. 3. 4

2) Aatomatic Actuation N.A.

N.A.

M(2)

1. 2. 3. 4 l[

Logic

3) Containment Pressure--

.5 R

M(3)

1. 2. 3 s

liigh-liigh sa Containment Ventilation 5I c.

Isolation M

I) Hanual N.A.

N.A.

R 1, 2, 3, 4

2) Automatic Actuation logic N.A.

N.A.

M(2) 1, 2. 3, 4

3) Containment Radio-5 R

H 3, 2, 3, 4 ac t iv i ty-fligh

i i

I_ABlf 4.3-2 (Continued)

U INGINflRID SAf f fV ffA10Hf ACillATI0li SYSTEM IN51RilMENTATION ShlBliXQNcf7t#iikfRTR15 l~.

~ ~ ~ ~ -

i l

CMANNIL MODES IN 6418CH E

CilANNEL CHAftNfL FUNCil"WAL SURVEILLANC.I U

fill CK CAllBRATION Tf5(

SEI)UIRf D FUNCil0NAL bNil 4.

51[AM LINE 150LAil0N N.A.

N.A.

R I. 2. 3 i

a.

Manual N.A.

N.A.

M(2)

1. 2. 3 b.- Automatic Actuation Logic l

S R

M(3)

1. 2. 3 Containment Pressure--

c.

liigh-itigh 5

R H

1. 2. 3 l

l d.

Steam Flow in two Steam Y

Lines--High Colnr.ident with M

i

-- Low or Steam iIne lilIsure--tow i

IllRBINE 1 RIP ANG fiffMAi[R 5.

150 tall 0N 5

R M

1. 2. 3 Steam Generator Stater l

g Level -liigh-liigh a.

s c2.!

6.

5AFIGUARDS [fRil M NI ON11ROL SY5itM (5tC) LOGIC l

!?

N.A.

M 1, 2. 3. 4 i

c.

Intusts N.A.

N.A.

M{l)

' l.- 2. 3. 4 logic. Ilming anil Hutputs i

I. 2. 3. 4 5

R H

7.

litiD(HV0t1 AGC VITAL 8115

1 g

TA8tE4.3-2(Contlaued) h

[NGINEERfD SAFETY FEAIUME ACIUATI.)N SYSilN INSTRuNENTATION e

SUAW1FilPfENf0llIRflWATS C

E CHANNEL N0 DES I'N WHICH 2

CHANNE L CHANNEL FUNCil0NAL SURVEILLANCE FUNCIl0NAt UNil EllECK CA11BNA110N 1E51 REQUIRED 8.

AUXILIARY FELDWATER a.

Automatic Actuation logic N.A.

N.A.

M(2) 1, 2, 3 b.

Steam Generator Water S

R H

1, 2, 3 i

level-tow-low I

c.

Undervoltage - RCP S

R M (2) 1, 2 1

d.

5.1.

See 1 above (All S.I. surveillance.equirements) a4 e.

Emergency Trip of Steam Cen-N.A.

N.A.

R 1

crator Feedwater Pumps ta 6

f.

Station Blackout See 6b and 7 above (SEC and U/V Vital Bus) a 2

E E.

TA8LE 4.3 2 (Continued)

TA}(ENOTATION (1) Esch logic channel shall be tested at lease ence per 62 days The CHANNEL FUNCTION TEST of each.

on a STAGGERED TEST SA5!5.

logic channel shall verify that its associated diesel generator, automatic load sequence. timer is OPERA 8LE with the interval between j

each load block within,1 second of its desian interval.

J (2) Each train or locic channel shall be tasted at least every J

62 days on a staggered basis.

The CHANNEL FUNCTIONAL TEST shall include exercising the trans:.itter (3) by applying either a vacuum or pressure to tM appropriate side of the transmitter.

l t

l SALE * - Uf;;* 1 3/4 3-34 Amendment No. 39

INSTEWEMTATION_

AC:10EMT MCNITORING INSTRUptMTATION LIMITING CONDITION FOR OPERATION 3.3.3.7 The accident monitoring instrumentation channels shown in Table 3.3-114.med Table 3.3-Mb shall be OPERA 8LE.

AP8LICABILITY: McCES 1, 2 ano 3.

ACTION:

As shown in Table 3.3-11a and Table 3.3-lib.

a.

I b.

The provisions of Specification 3.0.4 are not applicaele.

SURVEILLANCE REOUIR98ENTS Each ac:ident monitoring instrumentation channel shall be demonstra-4.3.3.7 tec OPERA!LE by performance of the CHANNEL CNECX and CHANNEL CALIlitAI!ON acerations at the frequencies shown in Table 4.3-11.

SALIN - UNIT l 3/4 3-53 Amenc: rent No. 39

TABLE 3.3-Ila y

ACCIDEUT MONITORING INSTRUMENTATION

.g TOTAL NO.

REQUIRED g

OF HO. OF INSTRUNENT CHANNELS CHANNELS ACTION H

F3

1. Reactor Coolant Outlet Temperature - T (Wide Range) 4 (1/1 p) 2 1

110T

2. Reactor Coolant Inlet Temperature - TCO2 (Wide Range) 4 (1/ loop) 2 1

3. Reactor Coolant Pressure (Wide Range) 2 2

1

4. Pressurizer Water Level 3 (hot) 2 1

5. Steam Line Pressure 3/ Steam Generator 2/ Steam Generator 1

6. Steam Generator Water Level (Narrow Range) 3/Stean Generator 2/ Steam Generator 1

na Ns

7. Steam Generator Water Level (Wide Range) 4 (1/ Steam Generator) 4 (1/ Steam Generator) 1

8. Refueling Water Storage Tank Water Level 2

2 1

9. Boric Acid Tank Solution Level 2 (1/ tank) 2 (1/ tank) 3 g

le

10. Auxiliary Feedwater Flow Rate 4 (1/ Steam Generator) 4 (1/ Steam Generator) 4 S

11. Reactor Coolant System Subcooling Margin Monitor 2*

2*

5

12. PORV Position Indicator 2/ valve **

2/ valve **

1

13. PORV Block Valve Position Indicator 2/ valve **

2/ valve **

1

14. Pressurizer Safety Valve Position Indicator 2/ valve **

2/ valve **

1

(*) 1btal number of channels is considered to be two (2) with one (1) of the channels being manual calculation by licensed control room personnel using data from OPERABLE wide range Reactor Coolant Pressure and Temperature along with steam Tables as described in ACTION 5.

(**) Total nuaber of Channels is considered to be two (2) with one (1) of the channels beAng any one (1) of the following alternate means of determining PORV, PORV Block, or Safety Valve position: Tallpipe Temperatures for the valves, Press-urizer Relief Tank Temperature, Pressurizer Relief Tank Level OPERABLE.

e TABLE 3.3-llb y

ACCIDENT MONITORING INSTRUMENTATION 1)

TOTAL NO.

MINIMUM g

OF NO. OF INSTRUNENT CllANNELS CHANNER.S ACTION 4

1. Reactor Coolant Outlet Temperature - T (Wide Range) 4 (1/ loop) 1 2

HO'1,

2. Reactor Coolant Inlet Temperature - TCO 2 (Wide Range) 4 (1/ loop) 1 2

1 "l.

Reactor Coolant Pressure (Wide Range) 2 1

2

4. Pressurizer Water Level 3 (hot) 1 2

5. Steam Line Pressure 3/Stease Generator 1/ Steam Generator 2

w

6. Steam Generator Water Level (Narrow Range) 3/ Steam Generator 1/ Steam Generator 2

N

7. Steam Generator Water Level (Wide Range) 4 (1/ Steam Generator) 3 (1/ Steam Generator) 2

8. Refueling Water Storage Tank Water Level 2

1 2

9. Boric Acid Tank Solution Level 2 (1/ tank) 1 2

10. Auxiliary Feedwater Flow Rate 4 (1/ Steam Generator) 3 (1/ Steam Generator) 6 s

11. Reactor Coolant System Subcooling Margin Monitor 2*

1 6

m f

12. PORV Position Indicator 2/ valve **

I 2

13. PORV Block Valve Position Indicator 2/ valve **

I 2

14. Pressurizer Safety Valve Position Indicator 2/ valve **

1 2

(*) Total number of channels is considered to be two (2) with one (1) of the channels being manual calculation by licensed control room personnel using data froan OPERABLE wide range Reactor Ccolant Pressure and Temperature along with Steam Tables as described in ACTION 5.

(**) 'Ibtal number of Channels is considered to be two (2) with one (1) of the channels being any one (1) of the following alternate means of determining PORV, PORV Block, or Safety Valve positions Tailpipe Temperatures for the valves, Press-urizer Relief Tank Temperature, Pressurizer Relief Tank Level OPERABLE.

TABLE 3.3-11asb (continued)

TABLE NOTATION ACTION 1 With the number of OPERABLE accident monitoring channels less than the Bequired Number of Channels shown in Table 3.3-11a, restore the inoperable channel (s) to OPERABLE status within 7 days, or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 2 With the number of OPERABLE accident monitoring channels less than the Minimum Number of Channels shown in Table 3.3-llb, restore the inoperable channel (s) to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 3 With the number of OPERABLE channels one less than the Required Number of Channelt shown in Table 3.3-11a, operation may proceed provided that the Boric Acid Tank associated with the remaining OPERABLE channel satisfies all requirements of Specification 3.1.2.8.a.

ACTION 4 With the number of OPERABLE channels one less than the Required Number of Channela shown in Table 3.3-11a, operation may proceed provided that an OPERABLE Steam Generator Wide Range Level channel is available as an alternate maans of indication for the Steam Generator with no OPERABLE Auxiliary Feedwater Flow Rate channel.

ACTION 5 With the number of OPERABLE channels less than the Required Number of Channels shown in Table 3.3-11a, operation may proceed provided that Steam Tables are available in the Control Room and the following Required Channels shown in Table 3.3-11a are OPERABLE to provide an alternate means of calculating Reactor Coolant System subcooling margins

a. Reactor Coolant Outlet Temperature - THOT (Wide Range)

b. Reactor Coolant Pressure (Wide Range)

ACTION 6 With the number of OPERABLE channels less than the Minimum Number of Channels shown in Table 3.3-11b, restore the inoper-able channel (s) to OPERABLE status within 7 days, or be in HOT SHUTDCWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SALEM - UNIT 1 3/4 3-56 Amendment No. 39

TAllLE 4.3-11 SUHVEILLANCE REQUIREtiENTS FOR un ACCIDENT MONITORING INSTHUHFNTATION b

CHANNEL a

CHANNEL CHANNEL FUNCTIONAL INSTRUMENT CHECK CALIBRATION TEST 2:

Hd

1. Reactor Coolant Outlet Temperature - Tg (Wide Range)

M R

NA H

2. Reactor Coolant Inlet Temperature - Tg (Wide Range)

M R

NA

3. Reactor coolant Pressure (Wide Range)

M R

NA

4. Pressuriser Water Level M

R NA

5. Steam Line Pressure M

R NA

6. Steam Generator Water level (Narrow Range)

M R

NA i

w N

4.

7. Steam Generator Water Level (Wide Range)

M R

NA s

8. Refueling Water Storage Tank Water Level M

R NA y

9. Boric Acid Tank Solution 34 vel M

R NA

10. Auxiliary Feedwater Flow Mate SUI R

NA

11. Reactor Coolant System Subcooling Margin Monitor M

MA NA O

12. PORV Position Indicator M

NA Q

[

13. PokV Block Valve Position Indicator M

NA Q

~w

14. Pres.surizer Safety Valve Position Indicator M

HA R

I Auxiliary Feedwater System is ussd on each Startup and Flow Rate indication is verified at that time.

The instruments used to develop RCS Subcooling Margin are calibrated on an 18 Month cycles the Monitor will be compared Quarterly with calculated subcooling margin for known input values.

s REACTOR COOLANT SYSTEM 1

' 3/4.4.2 SAFETY VALVE 5 1

-5AFETY VALVES - SMUTDOW LIMITING CONDITION FOR OPERATION

. j 3.4.2.1 A sinima of one pressurizer code safety valve shall be OPERA 8LE lift setting of 2485 psig

IL" AF*t.ICA8It.ITY: M00E5 4 and 5.

ACTION:

With no pressurme :sde safety valve C7ERABLE, imm 1000 ints operation in the shutdown cooling mode.

Si)Rv!ILLANCE RE0'JITEWEWS j

i 4.a.2.1 No ad::itional Surveillance Recuirements other than those requ rec Scecification 4.0.5.

l

Ine lif t se.ing pressure shall. correspond to amcient conditions of the at nominal operating temperature and pressure.

l 3/4 a-4 Amendtnent No. 39 SALEM - UNIT !

1 REACTOR COOLANT SYSTEM

?

3/4.4.2 SAFETY VALVES SAFETY VALVES - OPERATING

@ TING CONDITION FOR OPERATION 3.4l2.2 A11 pressurizer code safety valves shall be OPERA 8LE with a lift setting j

of 2485 psig 2 1%."

APDLICAB!LITY: MCDES 1, 2 and 3.

ACTICN:

With one pressurizer code safety valve inoperable, either restore the inoceracle valve to OPERASLE status within 15 minutes or he in HOT SHUT 00hN witnin 12 ncu SURVE:LLANCE RECUIREMENTS 4.42.2 No accitional Surveillanca Recuirements other tnan taase recuitec Oy

)

. eification 4.0.5.

I l

set ing pressure sna11 correscond to ameient coneiticas of tne valve "ine ist at nc=inal ccerating temcerature and pressure.

I l

l i

SALEM - UNIT !

3/4 4-4a Amendment No. 39 l

l I

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

1 gw-,

j i

ItEACTOR COOLANT SYSTEM 3/4.4.3 RELIEF val.YES LIMITING CONDITION FOR OPERATION 3.4.3 Two power relief valves (PCRVs) and their associated block valves shall be OPERA 8LE.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a.

With one or more PORV(s) inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the PORV(s) to CPERA8LE status or close the a3sociated block valve (s) i and remove power from the block valve (s); othe.wis:,, be in at laast HOT STANOSY within tre 7. ext 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDCWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b.

With one or more block valve (s) inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the block valve (s) to CPERABLE status or close the block valve (s) and remove power from the block valve (s); otherwisa, ce in at least HOT STAN08Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUIDCWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

The previsions of' Specification 3.0.4 are not acclicable.

c.

SURVEILLANCE REQUIP.EWENTS 4.4.3.1 In addition to the requirements of Specification 4.0.5, each PCRV shall be comonstrated CPERABLE at least once per 18 sonths by performance of a CH/NNEL CALIBRATION and operating the valve through one complete cycle of full travel.

4.4.3.2 Each blocx valve shall be demonstrated CPERABLE at least once per 92 days by ocerating the valve through one comoleta cycle of full travel.

i l

l l

l SALEM - UNIT l 3/44-5 Amendment No. 39

,.,..._,.-,-n.

,n,-_.,

O REACTOR CCOLANT SYSTEM 3/4.a.4 PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.4 The pressurizar shall be OPERA 8LE with a water volume e less than or equal to 1650 cubic feet (92% indicated level), and at least two groups of pressurizer heaters each having a capacity of 150 kw.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

With the pressurizer inoperable due to an inoperable emergency power a.

supply to the pressurizer heaters either restore the inoperable emergency power sucaly within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STAN08Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

With the pressurizer othenvise inoperable, be in at least HOT STAN08Y with the reactor trip breakers open within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

I SURVEILLANCE REOUIREMENTS 4.4.4.1 The pressurizer water volume shall be determined to be within its I

limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.4.4.2 The capacity of each of the abovo required groups of pressurizer heaters shall be verified by measuring circuit current at least once per 92 days.

4.4.4.3 The emergency power supply for the pressurizer heaters shall be demonstrated CPERA8LE at least once per la months by sanually transferring power from the normal to the emergency power supply and energizing the heaters.

l l

l SALEM - UNIT 1 3/44-6 Anendment No. 39 l

l

CONTAINMENT SYSTEMS 3/4.6.3 CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.3.1 The containment isolation valves specified in Table 3.6-1 shall be OPERA 8LE with isolation times as shown in Table 3.6-1.

APPLICA3ILITY: MODES 1, 2, 3 and 4.

ACTION:

With one er more of the isolation valve (s) specified in Table 3.6-1 inoperablej maintM n at least one isolasion valve OPERABLE in each affected penetration that is open and either:

a.

Restore the inoperable valve (s) to OPERABLE s'.atus within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or b.

Isolate each affected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one deactivated automatic valve secured in the isolation position, or c.

Isolate each affected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one closed manual valve or blind flange; or d.

Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVETLLANCE REOUIREMENTS 4.6.3.1.1 The isolation valves specified in Taele 3.6-1 shall be demonstrated OPERABLE prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by perfomance of a cycling test and veri *ication of isolation time.

SALEM - UNIT I 3/4 6-12 Amendment No. 39

f CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 1 4.6.3'.1.2 Each isolation valve specified in Table 3.6-1 shall be den.onstrated OPERABLE during the COLD SHUTDOWN or REFUELING MODE at least once per 18 months by:

a.

Verifying that on a Phase A containment isolation test signal, each Phase A isolation valve actuates to its isolation position, i

b.

Verifying that on a Phase B containment isolation, test signal, each Phase B isolation valve actuates to its isolation position, c.

Verifying that on a feedwater isolation test signal, each feedwater isolation valve actuates to its isolation position.

i d.

Verifying that on a Containment Purge and Pressure-Vacuum Relief isolation test signal, each Purge and Pressure-Vacuum Relief valve actuates to its isolation position.

i 4.6.3.1.3 At least once per 18 month, verify that on a main steam isolation test signal, each main steam isolation valve specified in Table 3.6-1 actuates to its isolation position.

4.6.3.1.4 The isolation time of each power operated or automatic valve of f

Table 3.6-1 shall be determined to be within its limit when tested pursuant to Specification 4.0.5.

4.6.3.1.5 Each containment purge isolation valve shall be demonstrated OPERABLE within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after each closing of the valve, except when the valve is being used for multiple cyclings, then at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , by verifying that when the measured leakage rate is added to the leakage rates determined pursuant 1

to Specification 4.6.1.2d. for all other Type B and C penetrations, the combined leakage rate is less than or equal to 0.60L,.

SALEM - UNIT 1 3/4 6-1S Amendment No. 39

~- -,.+---

e-

--y

,,.e r--t

=t-

"-+v~*-~d*-'=~=

=-e-+---*

= -

v=-,-*

-'---"ww-'-e----r----'--

w

+w

+ er, - - + -v-

= - - - - - - - - * + - - - ' -

w--

. e v...

O IAlltf 3.6-l (Continued) g CONIAlHHINI 1501 All0N VAlvis 2

1501All0N IIME (Seconds) l VALVE NUHetR lUNCI10N

~'

I.

MANU4L I.

1559008 Pressuriser Dead-Weight Calibrator Not Applice d. l 2.

1559018 Pressurizer Dead-Weight Calibratar Not Applicatm 5 3.

Il CV 98#

CVCS - NCP 5eals Not Applicable Not Applicable 4.

12 CV 98#

CVCS - NCP 5eals 5.

13 CV 90s CVC5 - HCP 5eals Not Applicable 6.

14 CV 9ns CVC5 - HCP 5eals Not Applicable 7.

l SJ 71#

CVCS, Flushing Connection Not Applicable 8.

Il 55 91"#

Steam Generator Sampling Not Applicable 9.

12 55 91"#

Steam Generator Sampling Not Applicable 10.

13 55 93*#

Steam Generator Sampling Not Applicable i

ll.

14 $$ 93*#

Steam Generator Sampling Not Applicable 12.

1 5A 1188 Compressed Air Supply Not Applicatie i

13.

I WL 1908 Refueling Canal Supply Not Applia ble in 14.

I 5F 36#

Nefueling Canal Supply Not Applicable 15.

I WL 1918 Refueling Canal Discharge Not Applicable 16.

t $f 22#

Hefueling Canal Discharge Not Applicable l

I /.

I Vf 9'#

Containment Radiatiosi Samplirg Not Appilcable l

18.

l VC 10*#

Contaisisnent Nadiation 54apling Not Applicable 19.

I VC ll*#

Containment Radiation Samplitig Not Applicable g

20.

l VC 14*#

Contaisinesit Radiation Sanspling Not Applicable a

21.

f uel Irarister lidae Not Applicable S

?+

E.

l I

.e PLANT SYSTEMS AUXILIARY FEECWATER SYSTEM u

LIMITING CONDITION FOR OPERATION

3. 7.1. 2 At least three independent steam generator auxilfan feedwater. pumps j

and associated manual activation switches in the control room and flow paths j

shall be OPERABLE with:

Two feedwatar pumps, eaca capule of being powered from separata a.

vital busses, and 1

b.

One feedwater pump capable of being powered _ from an CPEAABLE steam supply systas.

APPLICA8ILITY: MODES 1, 2 and 3.

ACTION:

a With one auxiliary feedwater pump inoperable, restore the sequired a.

auxiliary feedwatar pumps to CPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANOBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTD01=N within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

With two auxiliary feedwater pumos inoperable be in at least HOT STANOBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

With three auxiliary feedwater pumps inoperatie, immeciately initiate c.

corrective action to restore at least one auxiliary feetwater pump to OPERABLE status as soon as possible.

SURVEILLANCE REQUIREMENTS l

4.7.1.2 Each auxiliary feedwater pump shall be demonstrated OPERABLE:

I l

a.

At least once per 31 days by:

1.

Verifying that each actor-driven pump ceveicos a disenarge pressure of greater than or equal to 1275 psig on recirculation flow.

l 2.

Verifying that the steam turoine-criven aumo develops a cischarge pressure of greater than or equal to 1500 psig on recirculation flow when the secondary steam supoly pressure is greater than 750 psig. The provisions of Specification 4.0.4 are not apolicaele.

3.

Verifying that eaca non-automatic valve in the flow path that is not locked, sealed or otherwise secured in position, is in its correct position.

1 SALEM - UNIT I 3/4 7-5 Arnendment No. 39

0 O

9LANTSYsiEks SUWEILLANCE REQUIREMENTS (Continued)

~

4.

Verify that valves 11M3,12M3,13M3,11M20,12M20, 13M20, 14M20, 11M22, 12M22, 13M22, 14M22, 11M10, 12M10, 13M10, 14M10, 11v86, 12M86, 13M86, and 14M86 are locked open.

h.

At least once per 18 months during shutdown by:

1.

Verifying that esca automatic valve in the motor driven pum::

flow path actuatas to its correct, position on a pump discharge pressert tast signal.

2.

Verifying that each auxiliary feedwater ' pump starts as designed automatically upon receipt of each auxiliary feedwater actuation tast signal.

.s The availlary fecewater systae shell be demonstrated OPfitAEL! prior to entry into Mode 3 following eacn COLO SMifDen by perforsing a c.

flow test to verify the nomal flow paths free the Ausiliary lete-water Storage Tank to enth of the staas generators.

l l

SALD4 - UNIT 1 3/4 7-6 Amendment No. 39

v BASE 3 3/4.3.3.5 FTRI DE'

'1JMSTRLMENTATTON OPERABILITY of

.. fire detection instrumentation ensures that adequate warning tapability is available for the prompt detection of fires. This capability is required in order to detect and locate fires in their early stages. Prompt detection of fires will reduce the poten-tial for damage te safety related equipreent and is an integral element in the overall facility fire protac' Ton program.

In the event that a portion of the fire detection instrunentation is inoperable, the establishment of frequent fire patrols in the affected areas is required to provide detection capability until the inoperable instrunentation is restored to CPERASILITY.

i 3/4.3.3./ A00! DENT MONITORING INSTRUMENTATICN The OPERA 8I ITi of the accident monitoring instrumentation ensures that sufficient information is availanle on selected plant parameters to monitor and assess these variables following an accident. This capacility is consistent with the Rscommendations of Requiator Guide 1.97, "Instrumentxtion for Light-Water-Cocied Nuclear Power Plants to Asssss Plant Concitions Daring and Followiss and Accioent," Decenter 1975.

i l

l l

52 - WIT 1 5 3/4 3-3 Amendment No. 39

e i

3/4.4 REAC10R C00LAN* SYS*EM psEs 3/4.4.2 SAFETY VALVES The pressurizer code safety valves operate to prevent the RCS frcm being pressurized above its Safety Limit of 2735 psig. Each safety valve is designed to relieve 420,000 lbs per hour of saturated steam at the valve set point. The relief capacity of a single safety ealve is adecuate toIn relieve any overpressure condition which could occur during shutdown.

the event that no safety valves are C7ERASLE, an operating RHR loop, connected to the RCS, provides overpetssure relief cacability and will prevent RCS overpressurization.

During ope *ation, all pressurizer code safety valves must be OPERASLE to prevent the RCS from being pressurized above its safety limit of 2735 The combined relief capacity of all of these valves is greater than asig.

the maximum surge rate resulting from a complete loss of load assuming no reactor trip until the first Reactor Protective Sys:em trip set scint is reached (i.e., no credit is taken for a direct reactor trio en the loss of lead) and also assuming no coeration of the power operated relief valves or i

l steam duma valves.

Demonstration of the safety valves' lift settings will occur only f

during shutdown and will be cerfor ec in accordance witn the provisions of Section XI of the ASME Soiler and Pressurt Code.

3/4.4.3 RELIEF VALVE 5 The power operated relief valves and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load cecrease with steam dumo. Operation of the power coerated relief valves sinimizes the undesirable opening of the spring-loacec pressurizer coce safety valves.

Eacn pcwer ocerated relief valve has a remotely coerated block valve I

to provice positive snutoff capacility should a relief valve become inoceracie.

l l

1 SALEM - UNIT 1 B 3/4 4-la Amendment No. 39

REACTOR COOLANT SYSTEw BASES

- 3/4.4.4 PRE 55URI2ER The limit on the maximum water volume in the pressurizer assures that the parameter is maintained within the normal staady-state envelope of speration The limit is consistant with the initial SAR assumotions.

assumed in the SAR.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months periodic surveillance is sufficient to ensure that the parameter The is restored to within its limit following a:pected transient coeration.

maximum water volume also ensures that a steam bucole is formed and thus the RCS is not a hydraulically solid system. The requirement that a sinimum nuncer of pressuri:er heatart be OPERA 8LE assures that the plant will be tale to estaalish natural circulation.

3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure,that the structural integrity of this portion of tne RCS The program for inservice inspection of steam generator will be maintained.

tubes is based on a modification of Regulatory Guide 1.83. Revision 1.

Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tuees in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to Inse-vice inspection of steam generator tubing also provides corrosion.

a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.

The plant is expected to be operated in a manner such that the secondary coolant will be maintained witnin those chemistry limits found If the I

to result in negligible corrosion of the steam generator tubes.

secondary coolant chemistry is not maintained within these limits, The localized corrosion may likely result in stress corrosion cracking.

l extent of cracking during plant operat on would be limited by the i

limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage =

50C gallons ser day per steam generator). Cracks having a primary-to-secondary leakage 1e:s than tnis limit during operation will nave an adequate margin of safety to witnstand the loads imposed during normal operation and by postulated accidents. Operating plants have demonstrat-ed that primary-to-secondary leakage of 500 gallons per day per steam generator can readily be detected by radiation monitors of steam generator blowdown. Leakage in excess of this limit will require plant snutdown c

l and an unscheduled inspection, during which the leaking tubes will be f

located and plugged.

t

,5ALEM - UNIT 1 B 3/4 4-2 Amendment No. 39 o

l i

l l

--.. _ -.. _ _ _

ML20031G776

Contents

Text

Attachment:

Response

Response

Navigation menu

ML20031G776

Text

Attachment:

Response

Response

Navigation menu

Search