Text

Proposed Section 3 Tech Spec Changes for Facility.License Conditions Encl
	ML18085A974
Person / Time
Site:	Salem
Issue date:	02/17/1981
From:	; Public Service Enterprise Group
To:	;
Shared Package
ML18085A972	List: ML18085A971; ML18085A973; ML18085A974;
References
	NUDOCS 8103030012
	Download: ML18085A974 (37)
	v • d • e

-

w I

N 0

A TABLE 3.3-3 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION FUNCTIONAL UNIT

8.

AUXILIARY FEEDWATER

a.

Automatic Actuation Logic**

b.

Stm. Gen. Water Level-Low-Low

c.

d.

i.

Start Motor Driven Pumps ii. Start Turbine-Dri ven Pumps

Undervoltage-RCP Start Turbine-Driven Pump S. I.

Start Motor-Driven Pumps

- Applies to items b. and c.

TOTAL NO.

OF CHANNELS 2

3/stm. gen 3/stm. gen.

4-1/bus CHANNELS TO TRIP

'. 1 MINIMUM CHANNELS OPERABLE 2.

2/stm. gen.

2 stm. gen.

any.stm. gen.

2/stm. gen.

2 stm. gen.

any 2 stm. gen.

1/2 x 2 3

APPLICABLE HODES 1, 2. 3 l, 2. 3 1, 2, 3

l. 2 ACTION 20 141111:

19 See 1 above (All S.J. initiating functions and requirements)

TABLE 3.3-3 (Continued)

ACTION 17 With less than the Minimum Channels OPERABLE. operation may continue provided the contai111Dent purge and exhaust valves are maintained closed.

ACTION 18 With the number of OPERABLE Channels one less than the Total Number of Channels. restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or 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 .

ACTION 19 With the number of OPERABLE Channels one less than the Total Number of Channels. STARTUP and/or POWER OPERATION ~ay proceed provided the following conditions are satisfied:

DESIGNATION P-11 P-12

a.

The inoperable channel is placed in the tripped condition within l hour.

b.

The Minimum Channels OPERABLE requirements is met; however.

one additional channel may 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 SAFETY FEATURES INTERLOCKS CONDITION AND SETPOINT With 2 of 3 pressurizer pressure channels ~ 1925 psig.

With 3 of 4 Tavg ~hannels

~ 545°F.

With 2 of 4 Ta channels

< 541°F.

vg FUNCTION P-11 prevents or defeats manual block of safety injection actuation on low pressurizer pressure.

P-12 prevents or defeats manual block of safety injection actuation high steam line flow and low steam line pressure.

Allows manual block of safety injection actuation on high steam line flow and low steam line pressure.

Causes steam line isolation on high ste~m flow.

Affects steam dump blocks.

ACTION 20 With the number of OPERABLE channels one less than the Total Number of Channels, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; how-ever. one channel may be bypassed for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months for surveillance testing.

SALEM - UNIT 1 cJ.,1_ AC'TlONS l~ ~ 2.0 ~

~..t... ESF !~Lc.S

~

'Or~

AC.TJOIV J~.

-~- --

3/4 3-U.

c z 4

l.J

~

w I

N en TA_BLE 3..:I-~_J ~-~!1 ti flUe!:f )_

£NGINUREn SAHlY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP, TPOltlTS

./

TRIP SHPOINT FUNCTIONAL UNIT

. '*+.

.r*'*

/

5.

TURBINE TRIP ANo**w;;nowATER ISOLATION a

Steam Genera tor ~

Leve 1--

< 67% of na'~::w range

6.

High-High Tnstrumeht span each steam genera tor UNOERVOLTAGE VITAL BUS

.**->' 70X of bus voltage I

~"\\

'//

.~*'

.. /

ALLOWABLE VALUES

< 68% of narrow range Tnstrument span each steam generator

> 65% of bus voltage J

TAOl.E l. 3-4 (Continued)

ENGINEERFO SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS c :z:

~ FUNCTIONAL UNIT S.

TURBINE TRIP ANO FEEDWATER ISOLATION

a. Stea111 Generator Water Level--

Hgh-Htgh

6.

SAFEGUARDS EQUIPMENT CONTROL SYSTEM (SEC)

7.

UNOERVOLTAGE, VITAL eus.

~

a.

Loss of Voltage

'i'

8.

AUXILIARY FEEOWATER N.,

a.

AutoMattc Actuatfon logtc

b.

St~a* Generator Water leve1-1ow-1ow

c.

Undervo1tage - RCP

d.

s. I.

TRIP SETPOINT

< 67% of narrow range Tnstrument span each steaM generator Not*.Appl fcable

> 7UI Not Appltcable

> ld of nctrrow range Tnslrtm1enl span each stea111 generator

~ 1~ RCP bus voltage ALLOWABLE VALUES

< 681 of narrow range TnstrUHnt 1pan each stea11 generator Not Appltcable

) 651 Nol Applicable

> 171 of narrow range Tnslru.ent span each stea11 generator

~ 651 RCP bus voltage See 1 Above (All S.I. setpo1nts)

~-

1

]

TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION

1.

Manual

2.

1.

Safety Injection (ECCS)

Feedwater Isolation Reactor Trip (SI)

Containment Isolation-Phase "A" Containment Ventilation Isolation Auxiliary Feedwater Pumps Service Water System Containment Fan Cooler

b.

Containment Spray Containment Isolation-Phase "B" Containment Ve~tilation Isolation

c.

Containment Isolation-Phase "A" Containment Ventilation Isolation

d.

Steam Line Isolation Containment Pressure-High

a.

Safety Injection (ECCS)

b.

Reactor Trip (from SI)

c.

Feedwater Isolation

d.

Containment Isolation-Phase "A"

e.

Containment Ventilation Isolation

f.

Auxiliary Feedwater Pumps

g.

Service Water System SALE~ - UNIT 1 3/4 3-27 RESPONSE TIME IN SECONDS Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not applicable Not Applicable Not Applic~ble Not Applicable

< 27.0'*

i.ori *--

~ 2..o

<iQ

.C:7.0

- j:

1C=

~

..:. Hl d~*/e8. e -

~ 11.0 /21.0 Not Applicable Not App 1 i cable *

~ 13.0#/4e.o"=

. e.

TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEAT~ES ~ESPONSE TIMES INITIATING SIGNAL ANO FUNCTION

3.

Pressurizer Pressure-Low RESPONSE TIPIE IN SECONDS

~

1'2..0

1.

Safety Injection {ECCS)

b.

Reactor Trip (fran SI)

c.

Feedwater Isolation

d.

Containment Isolation-Phase "A"

e.

Contairanent Ventilation Isolation

f.

Auxfliary Feedwater Pumps

g.

Service Water System

4.

Differential Pressure Between Steam Lines-High

a.

Safety Injection (ECCS)

5.

b.

Reactor Trip (fran SI)

c.

Feedwater Isolation

d.

Containment Isolation-Phase "A"

e.

Containment Vent11 ati on I sol at ion

f.

Auxiliary Feedwater Pumps

. g.

Se.rvi ce Water System Steam Flow in Two Steam Lines - H1gh Coincident with T

--Low-Low avg

a.

Safety Injection (ECCS}

b.

Reactor Trip (from SI)

c.

Feedwater Isolation

d.

Containment Isolation-Phase "A"

e.

Containment Ventilation Iso1ation

f.

Auxiliary Feedwater Pumps

g.

Service Water System

h.

Steam Line Iso1ation

!. 27.O* /J4 I 9t-

!..a..o. z.o

!.~ 7.0

< 18.0f.

-Not Applicable Not Applicable

48 01/13.01

~ +9.o"/

!.J* ;~/23.e;.*~11.o 1 z.2.o

<.a.t-0 z..o

<~ 1.0

!..e.ef./28. e*** ~ 11.0~;.o#

Not.A.pp1 icab1 e Not Applicable

~ 13.0U48.0#i=

=*/ #.

~.,..e ns..o-14.o;.z.~c

< i-ra*.i\\.O

< liQ..-0 Cf,O

~ ~

,11,,'~* ". o:lj z 9.,,-fl::#:

Uot App1icab1e Not Applicable

< ~

Cf,O SALEM - UNIT l 3/4 3-28 Amendment No.

17

I

-~

e.

_. e TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION

6.

Steam Flow in Two Steam Lines-High Coincident with Steam Line Pressure-Low

a.

Sa f e*ty Inject ion ( ECCS)

b.

Reactor Trip (from SI)

c.

Feedwater Isolation

d.

Containment Isolation-Phase "A"

e.

Containment Ventilation Isolation

f.

Auxiliary Feedwater Pumps

g.

Service Water System

h.

Steam Line Isolation

7.

Containm~nt Pre~sure--High~Hi9h

a.

Containment Spray

b.

Containment Isolation-Phase "B"

c.

Steam Line Isolation

d.

Containment Fan Cooler

8.

Steam Generator Water Level--High-High

a.

Turbine Trip-Reactor Trip

b.

Feedwater Isolation

9.

Steam Generator Water Level --Low-Low

a.

Motor-Driven Auxiliary Feedwater Pumps :ti=#;#:

b.

Turbine-Driven Auxiliary Feedwater Pumps**

SALEM - UNIT 1 3/4 3-29 RESPONSE TIME IN SECONDS

~ 13 O';'BO.O

~....,._ 2,C

~..., 7,0

~ J.i Or:'Q0 :~ 'f Not Applicable Not Applicable

~ 14.0#/48.0;;ri:

< 8.0

< 45.0 Not Applicable

< 7. 0.

< 40.0

< 2.5

< 11. 0

~ 60.0 s 60.0

TABLE 3.3*5 (Continued)

EMGINEEREO SAFET'f FEATURES RESPONSE TIMES lNITlATlHG SIGNAL AHO FUNCTION RESPONSE TIME IM SECOHOS

10. Undervo1tage RCP Bus

a.

Turb1ne*Dr1ven Aux11ia'IJ Fe~dwater P1,11ps

11. Containment Radioactivity

High

a.

Purge and Exhaust Isolation

12.

UndeT"Va1tage. Vital Bus

a.

Loss of Voltage Note:

Response time for Motor-driven Auxi1iary Feedwater Pumps on all S.I.

signa1 starts SALEM

UNIT 1 3/4 3-30

! 60.0

~ 4.0

< 60.0

e.

TABLE 3.3-5 (Continued)

TABLE NOTATION

Diesel generator starting and sequence loading delays included.

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

Diesel generator starting and sequence loading delays not included.

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

1. Diesel generator.starting and sequence loading delays included.

Response time limit includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal char;~n~

pumps.

~

On 2/3 in any steam generator.

""* On 2/3 in 2/4 steam generators.

- - ' RAOlArt'\\"DN l:>erec:."l'OttS ~R.rr e'>Cf!Mit"T"'" F44-te-s11:>...,s~ T*ME 'Te'S"'n..,-.

R..w;sf>t,.,.;E "T*Mt:Clf TIE 2~-o**T10...a F*..,_~ 'S1ce~"'- R>a-r10"' o i= 'T"'c-c:.~iwN5'- Sr'Au.. as,..,.-~su4.Q ~""' "nL* ?>~12. ou-rPur oR

~llOM "Ne ~T OF "71-lfi: FtltSI e'UiCT'26N"- ~MPo"'~T IN "T'\\-V."rC.AAAIAJf:L SAL EM -

UN IT 1

w -

~

w I w N

~

TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION su~_yEl((~NCUtQUIREMENTS FUNCTIONAL UNIT

3.

CONTAINMENT ISOLATION

a. Phase "A" Isolation

1) Manual

2)

From Safety Injection Automatic Actuation Logic

b.

Phase "B" Isolation

c.

---~

1) Manual

2) Automatic Actuation Logic

3) Containment Pressure--

High-High Containment Ventilation Isolation

1) Manual

2) Automatic Actuation Logic CHANNEL CtlECK N. A.

N.A.

s N.A.

N.A.

CHANNEL CALI BRAT ION N.A.

N.A.

R N.A.

N.A.

CHANNEL FUNCTIONAL TEST R

M(2)

R M(2)

M(3}

R H(2)

~

~~~_0 Containment Radio-s R

H acUvity-Htgh MODES IN WHICH SURVEILLANCE

~QUIRED

1. 2. 3. 4 1,2,3,4 1,2,3,4 1, 2. 3, 4 1, 2. 3 1, 2. 3, 4 1, 2, 3, 4 1,2,3,4

]

TABLE 4.3-2 (Continued}

V1

):>

r ENGINEERED SArETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 3:

c CHANNEL MOO ES I H ~IHI CH

z:

CHANNEL CHANNEL FUNCTIONAL SURVEILLl'lNCE

-I FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED

4.

STEAM LINE ISOLATION

a. Manual N.A.

N.A.

R

1. 2, 3

b. Automatic Actuation Logic N. I\\. '

N.A.

M(2)

1. 2, 3

c. Containment Pressure--

s R

M(3)

1. 2, 3 w

High-High

d.

Steam Flow in Two Steam s

R M

1' 2' 3 w

Lines--High Coincident with I w T

-- Low or Steam Line lJ P~~~sure--Low

5.

TURBINE TRIP ANO FEEOWATER ISOLATION

a.

Steam Generator Water s

R M

1, 2' 3 Level--High-High

6.

SAFEGUARDS EQUIPMENT CONTROL SYSTEM (SEC) LOGIC

a.

Inputs N. I\\.

N.A.

M 1 t 2 I 3. 4

b.

Logic, Timing and N.A.

N. I\\.

M( 1) 1 ** 2. 3, 4 cJA Outputs

]

7.

lltlll[ HVOL TAG(, VITAL BUS s

R M

1. 2. 3.... ~

w w

I....

TAOlE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SORVE1tr1'Rcr-REQOIRF~tNlS FUNCTIONAL UNIT

8.

AUXILIARY FEEOWATER

a.

Aut0111altc Actuation logtc

b.

Stea~ Generator Water Level-low-low CHANNEL CHECK H.A.

s s

CHANNEL CALIBRATION H.A.

R R

CHANNEL FUNCTIOHAL TEST M(2)

M N.A.

t<<>OES IN WHICH SURVEILLANCE REQUIRED 1, 2, 3

'* 2, J 1, 2 Undervoltage - RCP

c.

See 1 above (All S.1. survetllance requtre.ents) d.

S.1.

(

e

i.._,

TABLE 4.3-2 {Continued)

TABLE NOTATION (1) Each logic channel shall be tested at least once per 62 days on a STAGGERED TEST BASIS.

The CHANNEL FUNCTION TEST of each logic channel shall verify that its associated diesel generator automatic load sequence timer is OPERABLE with the interval between J each load block within I~ of its design interval.

it 1 Se.c:ar.J~

(2) Each train or loqic channel shall be tested at least every

, days/ o" a. ~ERED 1 Est 8JISIS,

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

SALEM -

U~IT 1 3/4 3-34

'.. ~

INSTRUMENTATION ACCIDENT J<<lNITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.7 The 1ccfde~t 110nitoring fnstrumentltion channels shown in Table 3.3-11&&1\\cl Table 3.3-llb 1hal1 be OPERABLE.

APPLICABILITY:

MODES 1, 2 and 3.

ACTION:

- As slo.o""'"' ~~ To..b I& S.3-11, ~cl. T.. ~1.. ~.~ -ll b

b.

The provisions of Specification 3.0.4 are not applicable.

SU~VEILLANCE REQU!RE~ENTS 4.3.3.7 Each accident incnitoring instrumentation channel sha11 be demonstra-ted OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-11.

SALEM

UNIT 3/4 3-5°3

w w *

-t-Tl\\RtE l. l-11 ta.

~CCIDENT MONITORING 1HSTRUHENTATION TOTAL NO.

OF CHANNELS 1 NSTRUHENT

1.

Reactor Coolant Outlet Teiwperature - lifoy (Wide RanQe) 4 (l/loop)

l. Re1ctor Coolant Inlet T~erature - lcolD (Wide Ainge) 4 (l/loop)

I

]. Ae1ctor Coolant Pressure - Wtde Range 2

Pressurt1er Water level 3 (hot)

5.

SteaM ltne Pressure 3/Stm.Gen.

6.

Stea* Generator \\liter level - Narrow Range 3/Stm.Gen.

1.

Stea* Generator Water level - Wtde Range 4 (l/Stm.Gen.)

8.

Refueltng Water Storage Tank Water level 2

REQUIRED NO. OF CHANNELS 2

2 2

2 2/Stm.Gen.

l/Stm.Gen.

4 (l/Strn.Gen.)

2 ACTION 1

1 1

1 Boric Actd Tank Solutton level 1/tank

.< 2 tanks) 1 /tank 3

9.

10.

Au*tllary Feedwater Flow Rate 4 (1/Stm. Gen.)

4 (1/Stm.Gen.)

4

11.

Reactor Coolant SysteM Suhcooltng Margin Monttor l

l 5

11.

PORV Posttton 1ndtcator l/valve N. A.

1J PORV Block Valve Posttton 1 nd tea tor l/valve N.A.

14.

Saf~ly Va1v~ Po~itton 1ndtc~tor l/valve N.A.

cJJ_~~

TARlE l.]-11~

TOTAL NO.

MINIMUM CHANNF.LS OPERABLE w

1 NSTR\\IM[Hl I. R**ctor Cool*nt OUtl*l T*""'*r*tur*

THOT (Vldo Rong*)

R*octor Coolonl lnl*t T""1'*r1ture

lcoto (Wide Ronq*l

z.

]. Reactor Coo1ant Pressure

w*de Range

'ressurlzer Water level

6.

~ 7.

SteaM Generator Water level - Marrow Range Stea* Generator Water level - Vtde Range Refue11ng Water Storage Tant Vater level

8.

CJ.

Bor*c Actd lank Solutton Level

10.

11.

Aux\\1\\ary Feedwater Flov Rate Reactor Coolant Sysl*M Suhcool \\ng M~rg\\n MonUor

11.

PORV Po~*t*on 1nd*cator 11 PO~V Rlnc~ ValvP Po~*t\\on 1nd*cator OF CHANNELS 4 (1/loopl 1

4 (l/loopl 1

1 2

l (hot) 1 l/Stm.Gen.

l/Stm.Gen

l/Stm.Gen

1/Stm.Gen.

4 (l/Stm.Gen.)

N.A.

l 2

l/tanlt (2 tanks) 1 4 ( 1/Stm. Gen.)

3 1

1 l/valve N.A.*

1/valve N.A.

1/valve N. A.

ACTION 2

2 2

J 4

5 r-

1 I.

ACTION 1 AC'l'ION 2 ACTION 3 ACTION 4 ACTION 5 SALEM -

UNIT 1 9.

TABLE 3.3-lla&b (continued)

TABLE NOTATION With the nmber of OPERABLE accident ncni tor:i.nq channels less than the Required Number of Channels s.ha.m *in.

Table 3.3-lla, restore the ~able channel(s) to OPERABLE status within 7 days, or be in at least HC7I'

~

within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

With the nl.II1ber of OPERAB!E accident :rroni toring channels les:s than the.MllWmml Channels OPERABLE requirarents of Table 3.3-llb, restore the inoperable channel(s) to OPERABLE status within 48 :OOurs or be in at least HC7I' SHU'l'I:O'm within the next 12 :OOurs.

With the number of OPERABLE channels less than th' Total Number of Channels shewn in Tables 3.3-lla&b, ope.ration nay p:rcx::eed provided. that the Boric Acid Tank associated with the OPERABLE channel satisfies the requirarents of Specification 3.1.2.8.a.

With the' number of OPERABLE channels less than the Total Number of Channels shewn in Tables 3.3-lla&b, operation nay proceed. provided that an OPERABLE Stearn Generator level channel is available as an alternate m:ans of in:li-cation for the Steam Generator with no OPERl'BLE Allxiliary Feedwa.ter Flc:M Rate channel.*

With the number of OPERABLE channels less than the Total Number of Charmels shown in Tables 3.3-lla&b, operation nay proceed proVide>d that the follCMing Required. Channels shcwn*on Table 3 *.3-lla are OPERABLE to provide an al-ternate neans.of calculating Reactor Coolant System sub--

cooling margin*:

a. Reactor Coolant Outlet Tatp:rature - THC71'

~Wide Range)

b. Reactor Coolant Pressure - Wide Range

St.earn Tables available in Control ~

3/4 3-t'

_____~

TAULE 4. l-lt ACCIOENT MONIJORING INSlRllHCHTATIOH SllRV£1UANCE REQUIREMENTS c:

z -

-t INSTRUHCNT I. Reactor Coolant Outlet Teeperature - T1101 (Wtde Range)

l. Reactor Coolant.Inlet T~erature - TCOLO.(Wfde Range)

]. Reactor Coolant Pressure - Wide Range

4. Pressurizer Waler Level

5. Sleat11 line Pressure w '

6. Slea11 Generator Water Level - Narrow Range w

I

1. Stea* Generator Waler Level - Wide Range 4

8. Refueliny Waler Storage Tank Water Level

9. Boric Acid Tank Solution level

10. Auxiliary feedwater Flm1 Rate

11. Reactor Coolant Sy~te~ Subcooling Margin Monitor

12. PORV Position lnclfcator

~

13. PORV Block Valve Position Indicator Jll.

Saf~ly Valve ru~ilion Indicator t

{

CHANNEL CtflCK M

M M

H M

H N,A.

N.A.

N.A, Q

CHAHHEL CAL IDRA TION I

R R

It R

A A

N.A.

Mk N.A.

N.A.

IJ.A.

N.A. *

"-'A JJ.A.

tJ.A.

N.A.

JJ.A.

N./l.

Q Q

R

" ltEACTOR COOLANT SYSTcM

'J/4.4. 2 SAFETY VALVES SAFETY VALVES * ~DOWN LIMITING CONDITION FOR OPERATION

.u.

1 3.4.2.I A ainiaum of one pressurizer code safety valve shall be OPERABLE witi'I a

~

lift setting of 2485 psig t ti.*~. """l.

1 1 ~

o.-.l.~ $~

APPLICABILITY:

MODES 4 1nd S.

~

ACTION:

With no pressurizer code safety valve OPERABLE, illll'lediately suspend a11 operations involving positive reactivity changes and place an OPERABLE RHR loop into operation in the shutdown cooling mode..

SURVEILLANCE REQUIREMENTS

$~

- -rn-e-.....1....

if,..t-se....,t-t-ing pressure sha 11.correspond to ambient conditions of the \\la 1 ve at nominal operating temperature and pressure.

SALEM - UNIT !

3/4 4-4*

J J

J

REACTOR COOLANT SYSTEM n....-...1 3/4. 4.2 SAFffi VALVES

l.

J.-.r-~ l*"*6 "p-

~

SAFETY VALVES -

OPERATIN~.J LIMITING CONDITION FOR OPERATION

~--...,

3.4.t.Z All pressurizer code safety valves shall be OPERABLE with a lift se~ting of 2485 psig :t 1%.*.,

<.__ ~.J.. ~

~~o.)f:-~

APP Ll CAB I LITY:

MODES 1, 2 and 3.

ACTION:

With one pressurizer code safety valve inoperable, either restore the inoperab1e valve to OPERABLE status within 15 minutes or be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS J.,.f 1

4.4,t.1. No additiona1 Surveillance Requirements other than those required by Specification 4.0.5.

3 "The lift setting pressure shall correspond to ambient conditions of the 11al11e

]

at nominal operating temperature and pressure.

SALEM - UNIT t 3/4 4-'f-*a..

]

REACTOR COOLANT SYSTEM 3/4.4.3 RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.3 Two power relief valves (PORVs) and their associated block valves shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3.

ACTION:

a.

With one or more PORV(s) inoperable, within l hour either restore the PORV(s) to OPERABLE status or close the associated block valve(s) and remove power from the block valve(s); otherwise, 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 .

b.

With one or more block valve(s) inoperable, within l hour either restore the block valve(s) to OPERABLE status or close the block valve(s) and remove power from the block valve(s); otherwise, 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 .

c.

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.4.3.l In addition to the requirements of Specification 4.0.S, each PORV shall be demonstrated OPERABLE at least once per 18 months by performance of a CHANNEL CALIBRATION and operating the valve through one complete cycle of full travel.

4.4.3.2 Each block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel.

SALEM - UNIT f 3/4 4-S'.

REACTOR COOLANT SYSTEM PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.4 The pressurizer shall be OPERABLE with a steam bubble.

APPLICABILITY:

MODES 1 and 2 ACTION:

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

SURVEILLANCE REQUIREMENTS J

4.4.4 No additional Surveillance Requirements other than those required]

by Specification 4.0.5.

SALEM - UN IT l 3/4 4-6

/

/.;

REACTOR COOLANT SYSTEM

~

PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.4 The pressurizer shall be OPERABLE with at least 150 kw of pressu~izer heaters and a water volume of less than or equal to 1650 cubic feet (92%

indicated level).

APPLICABILITY:

MODES 1, 2, and 3.

ACTION:

a.

With the pressurizer inoperable due to an inoperable emergency power supply to the pressurizer heaters either restore the inoperable emergency power supply within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or 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 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 othentise inoperable, be in at least HOT STANDBY*

.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 .

SURVEILLANCE REQUIREMENTS 4.4.4.

The pressurizer water volume shall be determined to be within its

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

SALEM - UNIT I 3/4 4-<D

ti_'

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 OPERABLE with isolation times as shown in Table 3. 6-1.

APPLICABILITY:

MODES 1, 2, 3 and 4.

ACTION:

With one or more of the isolation valve(s) specified in Table 3.6-1 inoperable,]

o,d.A.:tl~,i-maintain at least one isolation valve OPERABLE in each affected penetration

~

l,.that is open and either:

a.

Restore the inoperable valve(s) to OPERABLE status 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 COLO SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.3. l. 1 The isolation valves specified in Table 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 performance of a cycling test and verification of isolation time.

SALEM - UNIT 1 3/4 6-lt I.

I Ii i I I

. I 11 SUR\\':'.ILLMICE REQU!RH~ENTS (Continued) 4.6.3.1.2 Eac~ isc1a:icn va~ve scec~~ie: ~n Ta::e 3.5-1 s~?~: ::

demonstrated OFERASLE during tne CG:..:i S:-!!.:7:::.;~; er ~:::=-::::~::;:; ~~:::~ a:.~::.!:

once per 18 months by:

a.

Verifying :~a:'" a Phase~ co~ta~~-=~t is:1~t~:~ test eac~ Fhase A iso1~tic~ va~ve ac:~a:es t: ~ts ~s:~?t~:~

Ver~f1in; tha~ o~ ~ ?~as= 5 =~~~~~:.~=~~ is:~=~~=- ::s:

eac~ Phase 5 is:la:~~~ va~ve a:t~a:es t~ its is::a:~:~

s ~ :.. ? **

s~:'."'a.,

,,,~~ 1 *I-***

c.

VerH,:.tin; t~a: OI"\\ a fee:wate!" is:~ati:n t:s: s~;'."'a~, ea:'":

f:edw~~e!" 1s~1atic~ Y!~ve iso1a:es :~ its is:::.:~:~ ::s~:~:~.

c.

'/!~i~y~r.g tha~ 011 !

C=ri:!~~~~: ?

1~r;: ar~ ?.,.es:;:.:-:-'.':::;:..~

e~ief is:~a:i::. t:s: s~~.. :*, :::~~"..:.. ;:.a~: =--e~;_,..~-*.':::..... -

~=~ie= va~ve a::~a:es t: ~:s is:~a:~=~ ~=s~:~:'."'.

4.6.3.l.3 At least once per 13 ~ont~s. ve~ify th:: c~:

-- :o:i-

) ---*

isola:ion test signal, ea:~ mafo ste::.: is:i::~:~ v:>.-e s:e:~.:o~e: ir.

3.6-1 a::Ja::s t: its is:i!!.:i:~ ;i~s~:~:...

ll4.6.3.l.4The isolation time of each power operated or automatic valve of

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

!Specification 4.0.5.

I 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 wh~n the measured leakage rate is added to the leakage rates determined pursuant to Specification 4.6.1.Zd. for all other Type B and C penetrations, the combined leakage rate is less than or equal to 0.60L1.

SALEM - UNIT 1 3/4 6-13 Amendr.:er.~ ~lo. 1 S

V1

)>

r fll

.::c c:

Vf\\l.Vr m1~mrn

-~

- I

r.

,,neit.T 't//+LV N11r'l8ets_.

..,..,0

~-t.JiJM&d.. C.O'-"'~""

"'5 ON... "f'TlO.C.I/.t> P-4-C.C w --i*

())

I

?j It>

~.,

n

1 IU

.1 **

()

Mf\\Nlll\\l

1.

(2 valves)#

2.

11 CV 9B#

). 12 CV 9U#

4.

13 CV 91\\#

5.

1'1 CV 90#

6.

1 SJ 71#

I.

11 SS 9J.* I

u.

12 SS 93*1 IJ.

13 SS 93*1

10.

14 SS 9:J*H

11.

1 SA llU#

1 £1

1 Wl 190#

1"1.

1 SF 36#

14.

1 \\.IL 191#

1 '

~-

1 SF 22#

1 G.

1 vc 9*n

11.

1 vc 1o*1 111.

1 vc IJ*N 111.

1 vc Jt1

H t'll.

T/\\U~L_3. Ci** 1 _ ( C:on U _!lt1ed )_

CONTl\\INMENT l.... Ol l\\T ION Vl\\1.VES rtltlf. 11 llN P1*es-.11r i 1er 1lPa1l-W1* iqhl c.il i hril tor eves - ltll' St!it Is lVCS - HCP

~1t'1llS eves - HCP SP.ils eves l<LP Sc!illS eves r 111'\\lii 111J c onnc* c ti on o.., 1... 1111 1;1*111~1*,1 tor S.1111p Ii 11q

~ t e.1111 1;1*111*1*..i t.11r Samp I i 11q

~ii ('11111 1,1*11cr.1tor

~.111111 I i 11q

~; L t*.1111 1;1*11e1*..i L.or Sa111p I i 11')

C11111pn*~~1!tl I\\ i r Supp I y Ill' I 111* I i 11~1 r.rn.1 l

.... 111'1' 1 y H1*l 1w I i11q l.i111.1l Supp I y 1!1* I 111* I i 111_1 r.111.1 l ll i <-.dl1llll!

1!1*1111* I i11q C.111.1 I lli*.d11ll"ljP I 1111 t.1 i llllll'lll H.111ii1I.i1111 * ** 1111111 i ll'I t:1111l.1 illllll!lll II.id i 11 I i 1111 '1111111' 1 i ""

t:11111.1i111111'111 l!.11 I i.1 t i 1111 '1111111' 1 i 11'1 I 1111 I 'I i 111111 *II I 11.111 i <1 I i 1111 *.. 1111p Ii 1111 I 111

1 1 r.111511*1* l 11 lll' I SOLATI Ort TIME Not Applicllble Not Applicdble Not A1111lic11ble Not ApplicJble Not Applicable e

Not Applicable tlot l\\pp 1 it db 1 e Not Applicable Not Awlicable Not Applict1ble Not /\\pp 1icah1 e Not Applicable Not Applicable Not ApplicJhle Nut App I fr,1h le

. Nut Appl ic.1hle Not Appl ic.1L>le Not /\\ppl iL.1bl1!

Not Al'Pl irt.1hl<~

Not l\\ppl it.1hle

"'i V1 TABLE 3. 6-1 (Continue~

~

r-

,(

fT1 CONTAINMFNT ISOLATION VALVfS x

c:

I

z VALVE NUMBER FUNCTION ISOLATION TIME (Seconds)

f.

HANUl\\l

1.

SS900#

Pressurizer Dead-Weight Calibrator Not Appl 1cable J r

SS90l#

Pressurizer Dead-Weight Calibrator Not AppHcable

3.

I CV 98#

eves - RCP Seals Not App 1i cab 1 e

4.

2 CV 981 eves - RCP Seals Not App 1i cab 1 e

5.

l CV 98#

eves - RCP Seals Not Applicable

6.

4 CV 98#

CVCS - RCP Seals Not Applicable e

1.

SJ 71#

eves flushing Connection Not Applicable

8.

1 SS 93 111

~team Generator Sampling Not Applicable

9.

2 SS 93 111 Steam Generator Sampling Not Applicable

10.

l SS 93 11#

Steam Generator Sampling Not App I icab 1 e

~

11.

4 SS 93 11#

Steam Generator Sampling Not Applicable en ll.

SA 118#

Compressed Air Supply Not Applicable I

13.

WL 190#

Refueling Canal Supply Not Applicable

.::\\

14.

SF 36#

Refueling Canal Supply Not Applicable

15.

WL 191#

Refueling Canal Discharge Not Applicable

16.

Sf 2211 Refueling Canal Discharge Not App 1i cab 1 e I 7.

vc 9"#

Containment Radiation Sampling Not Applicable

18.

VC IO"I Containment Radiation Sampling Hot Applicable

19.

vc 13"#

Containment Radiation Sampling Not App I icab le

20.

vc 14"#

Containment Radiation Sampling Not App I icab le

21.

fuel lransfer Tube Not Applicable e

e

,..OOt F1 'TC AG It: '-"",, "

An~c. w~~,,. CL(*

PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 At lea.st three independent steam generator auxiliary feedwater pumps and associated flow paths shal~ be OPERABLE with:

a.

Two feedwater pumps, each capable of being powered from separate vital busses, and

b.

One feedwater pump capable of being powered from an OPERABLE steam supply system.

APPLICABILITY:

MODES l, 2 and 3.

ACTION:

With one auxiliary feedwater pump inoperable, restore at least three auxiliary feedwater pumps (two capable of being powered from separate vital busses and on*e capable of being powered by an OPERABLE steam supply system) to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-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 .

SURVEILLANCE REQUIREMENTS 4.7.1.2 Each aux111ary feedwater pump shall be demonstrated OPERABLE:

a.

At least once per 31 days by:

l.

Verifying that the steam turbine driven pump develops a discharge pressure of > 1500 psig on recirculation flow when the secondary steam supply pressure is greater than 750 psig.

2.

Verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed or otherwise secured in position, is in its correct position.

SALEM-UNIT l 3/4 7-5

9.

e*

PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7*.1.2 At least three independent steam generator auxilfary feedwater _pumps

]

and associated manual activation switches in the control room and flow paths shall be OPERABLE with:

a.

Two feedwater pumps, each capable of being powered from separate vital busses, and

b.

One feedwater pump capable-of being powered from an OPERABLE steam supply system.

APPLICABILITY:

MODES 1, 2 and 3.

ACTION:

a.

With one auxiliary feedwater pump inoperable, restore the sequired auxiliary feedwater pumps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or 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 HOT SHUTDOWN within the follo~ing 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

With two auxiliary feedwater pumps inoperable be in at least HOT STANDBY 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 .

c.

With three auxiliary feedwater pumps inoperable, immediately initiate corrective action to restore at least one auxiliary feedwater pump*

to OPERABLE status as soon as possible.

SURVEILLANCE REQUIREMENTS 4.7. 1.2 Each auxiliary feedwater pump shall be demonstrated OPERABLE:

a.

At least once per 31 days by:

1.

Verifying that each motor-driven pump develops a discharge pressure of greater than or equal to 1275 psig on recirculation flow.

2.

Verifying that the steam turbine-driven pump develops a discharge pressure of greater than or equal to 1500 psig on recirculation J

flow when the secondary steam supply pressure is greater than 750 psig.

The provisions of Specification 4.0.4 are not applicable. J'

3.

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

SALEM - UNIT I 3/4 7-5

EC; a*tr PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

b.

At least* once per 18 months during shutdown by:

1.

Verifying ~teach 1utomatic valve in the snotor driven pump flow path actuates to its correct position on a pump discharge pressure test signal.

2.

Verifying that each motor driven pump starts automatically upon receipt of each of the following test signals:

a)

Loss of main feedwater pwnps.

b)

Safeguards sequence signal.

c)

Steam Generator Water Level -- Low-Low from one steam generator.

3.

Verifying that*the steam turbine driven pump starts auto-matically upon receipt of each of the following test SALEM-UNIT 1 signals:

a)

Loss of offsite power.

b)

Steam Generator Water Level -- Low-Low from two steam generators.

3/4 7-6

PL.ANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

4.

Verify that valves 11AF3, 12AF3, 13AF3, 11AF20, 12AF20, 13AF20, 14AF20, 11AF22, 12AF22, 13AF22, 14AF22, llAFlO, 12AF10, 13AF10, 14AF10, 11AFB6, 12AF86, 13AFB6, and 14AFB6 are locked open.

b.

At least ance per 18 110nths during shutdown by:

c.

1.

Verifying that each automatic valve in the motor driven pump flow path actuates to its correct position on a pump discharge pressure test signal.

2.

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

lhe auxn iary fetdwat1r S)'SUe thall bt dfttonstnttd OPERIJ!l~ prior to ~ntry into Modr J foll~fng,ach COLO SlfJTDO'ta( by ~r~or~1ng a flow te~t to verify tile ooM!'al flcv pith!r f~ u~ ~tl hary Feec-wattr Storage Tan~ to Hth of the 1ua.a qt-Mrators.

SALEM

UNIT I 3/4 7-6

i...._

BASES 3/4.3.3.6 FIR£ DETECTION INSTRUMENTATION OPERABILITY of the fire detection 1nstrtrnentation ensures that adequate warning capability 1s 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 to safety related equipment and is an integral element in the overall facility fire protection program.

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

3/4.3.3.7 ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor an assess these variables following an accident.

This capability is consistent with the Recommendations of Regulator Guide 1.97, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Fo 11 owing and Accident, 11 December 1975.

SALEM - UN!T 1 B 3/4 3-3

L...i

~01>1 F1 T'.\\Gt.Cc w '1l{

,..,.,..,.., ct( t l) ** "~

3/4.4 REACTOR COOLANT SYSTEM BASES 3/4. 4. 2.and 3/4.4. 3 SAFETY VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig.

Each safety va1ve is designed to relieve 420,000 lbs per hour of saturated steam at the valve set point. The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown.

In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, provides overpressure relief capability and wi11 prevent RCS overpressurization.

During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2735 psig. The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss of load assuming no reactor trip until the first Reactor Protective System trip set point is reached (i.e., no credi~ is taken for a direct reactor trip on the loss of 1oad) and also assuming no operation of the power operated relief valves or steam dump valves.

Demonstration of the safety valves' lift settings wi11 occ~r only during shutdown and wi11 be perfonned in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code.

SALEM - UNIT 1 B 3/4 4-la Amendment No. 2.l

3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.2 SAFETY VALVES J

The pressurizer code safety valves operate to prevent the RCS from 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 sing1e safety valve is adequate to re1ieve any overpressure condition which cou1d occur during shutdown.

In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, provides overpressure relief capability and wi11 prevent RCS overpressurization.

During operation, a11 pressurizer code safety va1ves must be OPERABLE to prevent the RCS from being pressurized above its safety 1imit of 2735 psig. The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss of 1oad assuming no reactor trip until tpe first Reactor Protective System trip set point is reached (i.e., no credit is taken for a direct reactor trip on the loss of load) and also assuming no operation of the power operated relief valves or steam dump valves.

Demonstration of the safety valves' lift settings will occur only during shutdown and will be perfonned in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code.

3/4.4.! RELIEF VALVES 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 decrease with steam* dump.

Operation of the power operated relief valves minimizes the undesirable opening of the spring-loaded pressurizer code safety valves.

Each power operated relief valve has a remotely operated block valve to provide positive shutoff capability should a relief valve become inoperable.

SALEM - UNIT 1 B 3/4 4-la

REACTOR COOLANT SYSTEM BASES 3/4.4.4-PRESSURIZER 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 the RCS will be maintained. The program for inservice inspection of steam generator 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 tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion.

Inservice inspection of steam generator tubing also provides 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 within those chemistry limits found to result in negligible corrosion of the steam generator tubes.

If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant operation wou1d be limited by the limitation of steam generator tube leakage between the primary coo1ant system a.nd the secondary coolant system (primary-to-secondary leakage =

500 gallons per day per steam generator).* Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during norma1 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 shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged.

SALEM - UNIT 1 B 3/4 4-2

--~------------~~

REACTOR COOLANT SYSTEM BASES

- 3/4.4.4 PRESSURIZER The limit on the maximum water volume in the pressurizer assures that the parameter is maintained within*the normal steady-state envelope of operation assumed in the SAR.

The limit is consistent with the initial SAR assumptions.

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 is restored to within its limit following expected transient operation.

The maximum water volume also ensures that a steam bubble is formed and thus the RCS is not a hydraulically solid system.

The requirement that a minimum number of pressurizer heaters be OPERABLE assures that the plant will be able to establ hh 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 the RCS will be maintained.

The program for inservice inspection of steam generator tubes is based on a ~edification 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 tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion.

Inservice inspection of steam generator tubing also provides 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 within those chemistry limits found to result in negligible corrosion of the steam generator tubes.

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

The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage =

500 gallons per day per steam generator).

Cracks having a primary-to-secondary leakage less than this limit during operation wi11 have an adequate margin of safety to withstand 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 shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged.

SALEM - UNIT 1 B 3/4 4-2

)

CO~DITIONS FOR LICENSE

. q/Q BE ADDED TO

FACILITY OPERATING LICENSE DPR-70 The following License Conditions shall be added to conform with NRC letter to all PWR licensees dated July 2, 1980:

A. Systems Integrity The licensee shall f*plement 1 program to reduce leakage from systems outside*

containment that would or could contain highly rad~oactive fluids during a serious transient or accident to as low as practical levels.

This progra~*

shall include the following:

l.

Provisions establishing preventive maintenance and periodic visua1 inspection requirements, and

2.

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

B. Iodine Monitoring The licensee shall implement a program which will ensure the capability to accurately determine.the airborne iodine concentration in vital areas under accident conditions... This program shall include the follo,..ing:

l.

Training of personne1,

2.

Procedures for monitoring, and

3.

Provisions for maintenance of sampling and analysis equipment.

c. Ba:kw~ Metnod for Determining SJbcooling Margin The 1icersee shall implement a program which wil1 ensure the capability to accurately monitor the Reactor Coolant System subcooling margin.

This progra~

sr,a11 include the followirtg:

l.

Training of personnel, and

2.

Procedures for monitoring.

ML18085A974

Text

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