Text

Proposed Tech Spec Pages Re App A,NUREG-0578 TMI Lessons Learned Category a Items
	ML18086A949
Person / Time
Site:	Salem
Issue date:	09/30/1981
From:	; Public Service Enterprise Group
To:	;
Shared Package
ML18086A946	List: ML18086A945; ML18086A947; ML18086A949;
References
	RTR-NUREG-0578, RTR-NUREG-578 NUDOCS 8110020337
	Download: ML18086A949 (34)
	v • d • e

TABLE 3.3-3 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO.

CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

8.

AUXILIARY FEEDWATER

a.

Automatic Actuation Logic**

2 1

2 1, 2, 3 20

b.

Stm. Gen. Water Level-Low-Low

i.

Start Motor Driven Pumps 3/stm. gen 2/stm. gen.

2 stm. gen.

1, 2, 3 14*

any stm. gen.

ii. Start Turbine-Driven Pumps 3/stm. gen.

2/stm. gen.

2 stm. gen.

1, 2. 3 14*

any 2 stm. gen.

c.

Undervoltage-RCP Start Turbine-Driven Pump 4-1/bus 1/2 x 2 3

1, 2 19

d.

s. I.

Start Motor-Driven Pumps See 1 above (All S.I. initiating functions and requirements)

e.

Emerqencv Trip of Steam Generator Feedwater Pumps - start Motor Driven Pumps 2-1/p:ump 2

2 -1/pump

f.

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

- Applies to items b. and c.

1 21 e

e

ACTION 17 ACTION 18 ACTION 19 DESIGNATION P-11 P-12 ACTION 20 TABLE 3.3*3 (Continued)

With less than the Minfmum Channels OPERABLE, operatf on may continue provided the containment purge and exhaust valves are aaf ntained closed.

l____.

. With the number of OPERABLE Channels one less than the Total

Number of Channels, restore the fnoperable channel to OPERABLE

/status withfn 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be fn at least HOT STANDBY within the

11ext* 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and fn COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

- 'With the number of OPERABLE Channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a.

The f noperable channel f s placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

b.

The Minimum Channels OPERABLE requirements f s 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 ANO SETPOIHT FUNCTION With 2 of 3 pressurf zer pressure channels ~ 1925 psi g.

With 3 of 4 Tavg channels

> 54S°F.

With 2 of ~ T1 g channels

< 541°F.

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

P-12 prevents or defeats

anual block of safety injection actuation high steam line flow and low steam line pressu!e.

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

Causes steam line isolation on high steam flow.

Affects steam dump blocks.

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 u~ tc l hour for surveillance testing.

With the number-of OPERABLE channels one less than the Mininrum Number r

. *of Channels, operation may proceed provided that either:

a. The inoperable channel is restored to OPERABIB within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or

b. If the affected Steam Generator Feed.water Pump is expected to be SALEM - UNIT 1 out of service for rrore than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , the inoperable channel is jurrpered so as to enable the Start Circuit of the Auxiliary Feed.-

water Pumps upon*the loss of the.other Steam Generator Feed.water Pump.

3/4 3-21

c::

z

-I TABLE 3.3-4 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT

5.

TURBINE TRIP AND FEEDWATER ISOLATION

a. Steam Generator Water Level--

High-High

6.

SAFEGUARDS EQUIPMENT CONTROL SYSTEM (SEC)

7.

UNOERVOLTAGE, VITAL BUS

8.

a.

Loss of Voltage AUXILIARY FEEOWATER

a.

Automatic Actuation Logic

b.

Steam Generator Water Level-low-low

c.

Undervoltage - RCP

d.

s. I.

e.

Emergency Trip of Steam Generator Feedwater Pumps

f.

Station Blackout TRIP SETPOINT

< 67% of narrow range Tnstrument span each steam generator Not Applicable

> 70%

Not Applicable

> 18% of narrow range Tnstrument span each steam generator

~ 70% RCP bus voltage ALLOWABLE VALUES

< 68% of narrow range Tnstrument span each steam generator Not Applicable

> 65%

Not Applicable

> 17% of narrow range Tnstrument span each steam generator

> 65% RCP bus voltage See 1 Above (All S.I. setpoints)

Not Applicable Not Applicable See 6 and 7 above (SEC and Undervoltage, Vital Bus)

I

]

---.~,~*-

........ ~. '-"':- "1:.*.*

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

1.

Manual

2.

a.

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 Ventilation Isolation

c.

Containment Isolation-Phase "A" Containment Ventilation Isolation r

d.

Steam Linelisolation r1-Containment Pressure-Hioh d

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 Applicable Not Applicable

< 27.0'*

~ 2..o

,~g -.c::.7.0

- j:

~

.: 10,0;;/ee.o -

~,1.ojlz1.o Not Applicable Not Applicable'

~ l3.o#;4e.o"=

TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATmtES ~ESPONSE TIMES INITIATING SIGNAL ANO FUNCTION RESPONSE Tifl'! tN SECONDS

3.

Pressurizer Pressure-Low I.

Safety* Injection (ECCS}

b.

Reactor Trip (frcm SI}

c.

Feedwater Iso1at1on

d.

Containment Iso1at1on-Phase "A"

e.

Containment Ventilation Isolation

f.

Auxiliary Feedwater Pumps

g.

Service Water System

4.

Differential Pressure Between Steam Lines-High

5.

a.

Safety Injection (ECCS)

b.

Reactor Trip (from SI)

c.

Feedwater Iseli.ti on

d.

Containment Isolation-Phase "A"

e.

Containment Ventilation Isolation

f.

Auxiliary Feedwater Pumps

g.

Service 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.

Conta foment Isolation-Phase "A"

e.

Containment Ventilation Isolation

f.

Auxiliary Feedwater Pumps

g.

Service Water System

h.

Steam Line Isolation 12.0

!. 27 o O* /J 3 I Qt-

!..a...o. z.o

~~ 7.0

< 18. Of.

Not Applicable Not Applicable ft45 Ot/13.0#

~ +9.oi'

~.l:i filil'/23. 8~~12..0 /t.2..0

<~ z..o

<~ 1.0

~.9.~/28.e** !.11.ojlz.1.o#

Not fl.pplicab1e Not Applicable

~ 13.0l./48.o=~

!kl

a

~ 14.8 f2S.'1-14.0/z.c+.c

< iori}. "'

c

< liQ..-0 Cf I 0

~*

1'{.0~Z'l.o#

tlot App1icab1e Not App1icab1e

~ 35 g; l i avcw-lcf. 0 ~4-'1. 0 ff

< ~

~.o SALEM - UNIT l 3/4 3-28 Amendment No. 17

9 TABLE 3.3-5 (Continued)

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

6.

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

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

h.

Steam Line Isolation

7.

Containment Pressure--High-High

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~

b.

Turbine-Driven Muxiliary Feedwater Pumps *"If SALEM - UNIT 1 3/4 3-29

~ ia a )Ba.o

~iM-2.c

~.M /.0

~ J.i Ot:'08.'@*

Not Applicable Not Applicable

~ 14.0#/4s.o~:

< 8.0

< 45.0 Not *Applicable

< 7. 0.

< 40.0

< 2.5

< 11. 0

~ 60.0 s 60.0

TABLE 3.3-5 (Continued)

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

10.

Undervoltage RCP Bus

a.

Turbine-Driven Auxiliary Feedwater Pumps

11.

Containment Radioactivity - High

a.

Containment Pressure-Vacuum Relief System Isolation

12.

Trip of Feedwater Pumps

a.

Auxiliary Feedwater Pumps

13.

Undervoltage, Vital Bus

a.

Loss of Voltage

14.

Station Blackout

a.

Motor Driven Auxiliary Feedwater Pumps Note:

Response time for Motor-dri~en Auxi1i*ry Feedwater Pumps on a11 S.I.

signal starts SALEM - UNIT I 3/4 3-30

< 60.0

< 5.0 (***)

Not Applicable

< 4.0

~ 60. 0

~ 60.0

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.

(##)) 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.

(###))

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 thecontainrnent Pressure-Vacuum Relief valves are fully shut.

SALEM - UNIT t 3/4 3-3t.

(/1 TABLE 4.3-2 (Continued}

~

)>

I f"T1 3:

c ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION

z

~UR~Ei[L~~Cf:J!~_QUIREMENTS

-4 CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CllECK CALIBRATION TEST

~QUIRED

3.

CONTAINMENT ISOLATION

a. Phase "A" Isolation e

1) 1, 2' 3, 4 Manual N.A.

N.A.

R w

2)

From Safety Injection N.A.

N.A.

M(2) 1, 2' 3, 4

~

Automatic Actuation Logic w

I w

b.

Phase "8" Isolation N

1) Manual N.A.

N.A.

R 1

2, 3, 4

2) Automatic Actuat;on N.A.

N.A.

M(2) 1, 2, 3, 4 Logk

3) Containment Pressure--

s R

M(3) 1

2, 3 High-H;gh

c.

Containment Vent;lation e

I sol a Uon

1) Manual N.A.

N.A.

R 1,2,3,4

~ -

---~ 2) Automatic Actuation Logic

]

N.A.

M(2) 1, 2, 3, 4

~

~-4~.0 Conta;nment Radio-s R

H 1,2,3,4 act iv ;ty-Ht gh

V1 TABLE 4.3-2 (Continued)

)>

I

£NGJNE£R[0 SAFETY FfATURE ACTUATION SYSTEM INSTRUMfNTATIOff 3:

SURVEILLANCE REgUIREMENTS c

CHANNEL MODES IN *IH ICH z

CHANNEL CHANNEL FUNCTIONAL SURVEILLllNCE FUNCTIONAL UNIT CllECK CALI BRA Tl ON TEST REgUIREO

4.

STEAM LINE ISOLATION

a. Manual N.A.

N.A.

R

1. 2, 3 e.

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

N.A.

M(2) 1, 2, 3

c. Containment Pressure--

s R

M(3) 1, 2, J w

Jfigh-High J:>

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.

TURBIN[ TRIP ANO FHOWATER ISOLATION

a.

Steam Generator Water s

R M

1, 2. J level--High-High e

6.

SAF£GUAROS EQUIPMENT CONTROL SYSTEM (SEC) LOGIC

a.

Inputs N.A.

N.A.

M

1. 2. 3, 4

'1.

toqic, Timing an

r tTl 3:

w

~

w I..,

~,

TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS FUNCTIONAL UNIT

8.
AUXILIARY FEEDWATER
a.
Automatic Actuation Logic
b.
Steam Generator Water Level-Low-Low
c.
Undervoltage - RCP CHANNEL CHECK N.A.
s s
CHANNEL CALIBRATION N.A.
R R
CHANNEL FUNCTIONAL TEST M(2)
M M (2)
MODES IN WHICH SURVEILLANCE REQUIRED 1, 2, 3 l, 2, 3 l, 2
d.
s. I.
See 1 above (All 5.1. surveillance requirements) e.
Emergency Trip of Steam Gen-N.A.
N.A.
R 1
erator Feedwater Pumps
f.
Station Blackout See Gb and 7 above (SEC and U/V Vital Bus)
'.*I e \\
r:
l
~:
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 each load block within/~ of its design interval.
't!: 1 Se.co...icl.
(2)
Each train or loqic channel shall be tested at least every ** 1
,days/ on a. ST~ER.£0 1EST S!tSIS.
(3)
The CHANNEL FUNCTIONAL TEST shall include exercising the transnitter by applying either a vacuum or pressure to the appropriate side of the transmitter.
SALEM - ur;n 1 314 3-34 J
=i
~~--:-.
.. ~-~---~
INSTRUMENTATION ACCIDENT "'NITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.7 The 1ccfde~t 90nfto~1ng fnstruDtntatfon channels shown in Tlb1e 3. 3*111. al\\cl Table J. J*Ua.
1h111 bt OPERABLE.
APPLICABILITY:
~OES 1, z and 3.
ACTION:
. As*"°"""' '"'To.\\,I& S.3-ll-. ~ct T.. 1.1. "!>.~-u b
b.
The provisions of Specification 3.0.4 are not a;>p1icab1e.
SU~VEILLANC! REQUIRE~ENTS 4.3.3.7 Each accident aonit.oring instrumentation channel sna11 be demonstra-ted OPERABLE by perfor111ance of the CHANNEL CHECK 1nd CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-11.
SALEM
UNli I 3/4 3-53
TABLE 3.3-lla ACCIDENT MONITORING INSTRUMENTATION TOTAL NO.
INSTRUMENT OF CHANNELS REQUIRED NO. OF CHANNELS ACTION c:: z H
8 -
1.
2.
3.
4.
s.
6.
7.
8.
9.
10.
11.
12.
13.
14.
Reactor Coolant Outlet Temperature - THOT (Wide Range)
Reactor Coolant Inlet* Temperature - TCOLD (Wide Range)
Reactor Coolant Pressure (Wide Range)
Pressurizer Water Level Steam Line Pressure Steam Generator Water Level (Narrow Range)
Steam Generator Water Level (Wide Range)
Refueling Water Storage Tank Water Level Boric Acid Tank Solution Level Auxiliary Feedwater Flow Rate Reactor Coolant System Subcooling Margin Monitor PORV Position Indicator PORV Block Valve Position Indicator Pressurizer Safety Valve Position Indicator 4 (I/loop) 2 4 (I/loop) 2 2
2 3 (hot) 2 3/Steam Generator 2/Steam Generator 3/Steam Generator 2/Steam Generator 4 Cl/Steam Generator) 4 Cl/Steam Generator) 2 2
2 Cl/tank) 2 (I/tank) 4 Cl/Steam Generator) 4 (I/Steam Generator) 2*
2*
2/valve**
2/valve**
2/valve**
2/valve**
2/valve**
2/valve**
(*) 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 from OPERABLE wide range Reactor Coolant Pressure and T~mperature along with Steam Tables as described in ACTION 5.
1 1
1 1
1 1
1 1
3 4
5 1 e 1
1
(**) Total 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 position: Tailpipe Temperatures for the valves, Press-urizer Relief Tank Temperature, Pressurizer Relief Tank Level OPERABLE.
~~
~*~
. ~
~.;*
.r.,
f;
n.
H J1
1i

1
( i
~I t:
t
",I :
~ r
.:f
'.*t
.. ~.i 1}
TABLE 3.3-llb ACCIDENT MONITORING INSTRUMENTATION TOTAL NO.
INSTRUMENT OF CHANNELS MINIMUM NO. OF CHANNELS ACTION
1. Reactor Coolant Outlet Temperature -
THOT (Wide Range) 4 Cl/loop) 1
2. Reactor Coolant Inlet Temperature -
TCOLD (Wide Range) 4 Cl/loop) 1
3. Reactor Coolant Pressure (Wide Range) 2 1
4. Pressurizer Water Level 3 (hot) 1
5. Steam Line Pressure 3/Steam Generator l/Steam Generator
6. Steam Generator Water Level (Narrow Range) 3/Steam Generator 1/Steam Generator
7. Steam Generator Water Level (Wide Range) 4 Cl/Steam Generator) 3 Cl/Steam Generator)
8. Refueling Water Storage Tank Water Level 2
1
9. Boric Acid Tank Solution Level 2 Cl/tank) 1
10. Auxiliary Feedwater Flow Rate 4 (!/Steam Generator) 3 (!/Steam Generator)
11. Reactor Coolant System Subcooling Margin Monitor 2*
1
12. PORV Position Indicator 2/valve**
1
13. PORV Block Valve Position Indicator 2/valve**
1
14. Pressurizer Safety Valve Position Indicator 2/valve**
1
(*) Total nwnber 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.
2 2
2 2
2 2
2 6
2 2
2
(**) Total 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 position: Tailpipe Temperatures for the valves, Press-urizer Relief Tank Temperature, Pressurizer Relief Tank Level OPERABLE.
ACTION 1 ACTION 2 ACTION 3 ACTION 4 ACTION 5 e
TABLE 3.3-lla&b (continued)
TABLE NOTA~ION With the number of OPERABLE accident monitoring channels less than the Required Number of Channels shown in Table 3.3-lla, 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 .
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 .
With the number of OPERABLE channels one less than the Required Number of Channels shown in Table 3.3-lla, 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.
With the number of OPERABLE channels one less than the Required Number of Channels shown in Table 3.3-lla, operation may proceed provided that an OPERABLE Steam Generator Wide Range Level channel is available as an alternate means of indication for the Steam Generator with no OPERABLE Auxiliary Feedwater Flow Rate channel.
With the number of OPERABLE channels less than the Required Number of Channels shown in Table 3.3-lla, operation may proceed provided that Steam Tables are available in the Control Room and the following Required Channels shown in Table 3.3-lla are OPERABLE to provide an alternate means of calculating Reactor Coolant System subcooling margin:
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-llb, restore the inoper-able 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 .
SALEM -
UNIT i 3/4 3-56
en
):ii t'i l'll 3:
I c: z H
~
w w
I c.n TABLE 4.3-11 SURVEILLANCE REQUIREMEUTS FOR ACCIDENT MONITORING INSTRUMEN'l'ATION CHANNEL CHANNEL CHANNEL FUNCTIONAL INSTRUMENT CHECK CALIBRATION TEST
1. Reactor Coolant Outlet Temperature - THOT (Wide Range)
M R
NA
2. Reactor Coolant Inlet Temperature - 'l'coLD (Wide Range)
M R
NA
3. Reactor Coolant Pressure (Wide Range)
M R
NA
4. Pressurizer Water Level M
R NA
5. Steam Line Pressure M
R NA
6. Steam Generator Water Level (Narrow Range)
M R
NA
7. Steam Generator Water Level (Wide Range)
M R
NA
e. Refueling Water Storage Tank Water Level M
R NA
9. Boric Acid Tank Solution Level M
R NA
10. Auxiliary Feedwater Flow Rate SU#
R NA
11. Reactor Coolant System Subcooling Margin Monitor M
NAfl NA
12. PORV Position Indicator M
NA Q
13. PORV Block Valve Position Indicator M
NA Q
14. Pressurizer Safety Valve Position Indicator M
NA R
Auxiliary Feedwater System is used 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 cycle; the Monitor will be compared Quarterly with calculated subcooling margin for known input values.
e
'\\ REACTOR COOLANT SYSTEM
'3/4. 4. 2 'SAFffi VALVES SAFITT VALVES
SHUTDOWN
LIMITING CONDITION FOR OPERATION
.u.
71
]
3.4.Z.I A *ini*um of one pressurizer code safety valve sha11 be OPERABLE witil a
~
Hft setting of 2485 psig t 1:.*,... L 11 ~
~~$~
APPLICABILITY:
MODES 4 1nd 5.
~
ACTION:
With no pressurizer code safety valve OPERABLE, immediately suspend a11 operations involving positive reactivity changes and place an OPERABLE RHR loop into operation in the shutdo~n cooling mode.
SURVEILLANCE REQUIREMENTS l
o))..~
-*-rn-e--.-1.... if.,..,t--s-et_t __ ing pressure shall.correspond to ambient conditions of the va:ve J at nominal operating temperature and pressure.
SALEM
UNIT !
3/4 4-4*
'J
REACTOR COOLANT SYSTEM n_~.--.;
3/4.4.Z SAFETY VALVES
\\.J-.f-~~--
SAFETY VALVES - OPERATINg_J LIMITING CONDITION FOR OPERATION
~- "1 3.4.t.Z All pressurizer code safety valves shall be OPERABLE with a 1ift setting of 2485 psig +/- 1%.*...
~----~o.J.1. ~
a..-J._~o:;f:-~
APPLICABILITY:
MODES 1, ACTION:
Z and 3.
With one pressurizer code safety valve inoperable, either restore the inoperable 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--.r 1
4.4,t.'2. No additional Surveillance Requirements other than those required by J
Specification 4.0.5.
rne lift setting pressure shall correspond to ambient conditions of the valve
]
at nominal operating temperature and pressure.
SALEM - UNIT l 3/4 4-'f-*c..
]
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); othenitise, 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 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the block valve(s) to OPERABLE status or close the block valve(s) and remove power from the block valve(s); othef"Wise, 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 4.4.3.1 In addition to the requirements of Specification 4.0.5, each PORV shall be demonstr?ted 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 I 3/4 4-S'
REACTOR COOLANT SYSTEM 3/4.4.4 PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.4 The pressurizer shall be OPERABLE with a water volume of 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:
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 otherwise 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 4.4.4.1 The pressurizer water volume shall be determined to be within its 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 above required groups of pressurizer heaters shall be verified by measuring circuit current at least once per 92 days.
4.4.4.3 The emergency po~er supply for the pressurizer heaters shall be demonstrated OPERABLE at least once per 18 months by manually transferring power from the normal to the emergency power supply and energizing the heaters.
SALEM - UNIT f 3/4 4-I
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:~.f"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 COLD 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. 1. l 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 11 SURV:ILLA~ICE REqurRn~ENTS (Continued) 4.6.3.1.2 Eac!i ~sc1a<;ien va~ve soec~'fie: 1!1 Ta:::e 3.5-1 s~:~: :-:
demonstrated OrERASLE durins the COL:J SHt:7:::.:~~ er ~~::-*.:~:..::;~ ~~::~ c.:. ~=~~:
once *per 18 months by:
a.
b.
Verifying :~a:'" a ?has:~ cent:~~-=~: is:1a:~:r t:s:
eac~ rhase ~ isoi~:icr. va~ve ac:~a::s :: ;~s ~s:~a:~:~
Ver~7yin; tha~ on !
?~as: 5 ::i~!~:.~=~~ is:~~~~:~ :~~:
each Phase 5 isola:~~~ va~ve a:t~ates t~ its is:~a:~:~
s,: ;- ?,
I
.~:..::-
5'": ;:~.,
---~,;..; _,...
pt/W~ I,. I w** *
c.
VerHJin; t~a: o., a fee:*.... ate'!" is:~ati:n t:s: s*;~a~, :::~
faedwa~e'!" isolatic., va~ve iso1a:es :~its is:~::*:~ ::s~:~:~.
c.
4.6.3.1.3 At least once per 13 ~ont~s. verif; tha: =~a isola:ion test signal, ea:~ mafo stea:.'. is:ia:~:~ n".-.-: s:::~.:~e: ir.
3.6-1 a::i.la::s t: its is~:!:i:~ ;~s~:~:r.
ll4.6.3.t.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 jSpecification 4.0.5.
I 4.6.3.l.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 to Specification 4.6.l.2d. 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 Amend~en: ~o. 15
V1
)>
r-Pl 3:
c:
V/\\1.VF NllMnrn
-~
. I
r.
~ S8/lt. T' t,/,.L V N 111"\\bs __,.
....,0 ~-t.rtJ~ C.OC..UI"\\~
"5 ON 4"7T>><-il£0 P.+GC
}*
-I rt>
~.,
0 : J rlJ,
()
Ml\\Nlll\\l
1.
2.
).
4.
5.
6.
I.
u.
9.
10.
11.
1 l:'
1 *1.
M.
l ~.
IG.
1 / -
111 _
111.
i' 11 *
(2 valves)#
11 CV 90#
12 CV 90#
lJ CV 90#
Jt1 CV 90#
1 SJ 71#
11 SS 9J.*I 12 SS 93*1 lJ SS 9J*#
111 SS 9:t*/I 1 SA llU#
1 WL 190#
1 SF 36/I 1 Wl 191#
1 SF 22#
1 VC 9 *II 1 vc JO*#
1 VC lJ*N 1 VC 111 *II Tl\\ll~L_].fi--1. (Cont.i!'ucd)_
CONTl\\INMHH l'>Ol /\\TION VAl.VES rum: 111 JN 1'1-es~11t'i1er llPcl<l-Wl'iqhl C,il ih1*alor eve 5 - 11u* Se.tis lVCS llU' ~)p,ils eve~
11rl' SPals rvcs 1<L11 S1!il 1 s CVCS Tl11<>hi111J _Conm~d ion
'>I 1*,1111 1;1*111!1'c1Lor s.111111 I i11*1
'>le.1111 t;1*111*1*.it.11r s.1111111 ill'J
~.I p,1111 f11'11er.1lor
~.1111pl i11q SI e<1111 lil'11c1*.i Lor S11111p I i 1111 C11111pn*~~Pd /\\ir Supply Ill' I 111* i ll~J C.111.1 I ~""" 1 v ll1* I 1w i 11q t.111.1 l 51111p I y 111*1111*
i111_1 C:.111.11 lli'>dl11l'ljl~
llt*l111*
i1111 C.111.11 lli~.d1.11*qp
11111L.1 111111*111 ll.11lit11.i1111 *,,1111plin*1 t:1111l.1 11111t!lll ll.11li,11 i1111 *.. unpl inq 1:1111 t.1 111111*11 I 11.111 i,1 t i 1111 '.. 1111p 1 i 11'1 11111 I.1 111111*11t 11.ul i.it i 1111 *.. 111111Ii1111 1111*1 1"1111511'1" 111hp I SOLA Tl Ori Tl M[
Not Ap1> licab le Not App 1icdb1 e Not /\\p1il ic,1b le Not Ap111i cJb 1 e Not Applicable Not Applicable Not /\\p11litdble Not Ap1>licable Not Aflplit:ab1e Not /\\p11liciJble Not Applicable Not App 1 icah le Not Applicable Not Ap1ilicclhle Not App I c.1h le Not Appl c.1hle Not Appl c.11Jle Not /\\j1p I t.1h le!
Not /\\ppl rahle Not /\\pjll L*1hle
~*...
TABLE 3.6-1 (Continued}
)>
r fT'1 CONTAINMFNT ISOLATION VALVES x
c I
z:
VALVE NUMBER FUNCTION ISOLATION TIME (Seconds}
~
F.
HANU~l
1.
SS900#
Pressurizer Dead-Weight Calibrator Not Applicable J r
SS90l#
Pressurizer Dead-Weight Calibrator Not App 1tcab1 e
3.
I CV 98#
eves - RCP Sea 1s Not App 1icab1 e
4.
2 CV 981 CVCS - RCP Seals Not Applicable
5.
3 CV 98#
eves - RCP Seals Not Applicable
6.
4 CV 98#
CVCS - RCP Seals Not Appl tcable
1.
SJ 71#
CVCS Flus~ing Connection Not Applicable
8.
1 SS 93"1
~team Generator Sampling Not Applicable
9.
2 SS 93"#
Steam Generator Sampling Not Applicable w
10.
l SS 93*#
Steam Generator Sampling Not Applicable
11.
4 SS 93*#
Steam Generator Sampling Not Applicable 0\\
12.
SA 118#
Compressed Air Supply Nol Applicable I
13.
WL 190#
Refueling Canal Supply Not Applicable 4
14.
SF 36#
Refueling Canal Supply Not Applicable
15.
WL 191#
Refueling Canal Discharge Nol App li cab I e
16.
SF 22#
Refueling Canal Discharge Not Applicable 1 7.
vc 9*#
Containment Radiation Sampling Not Applicable
18.
vc 10*1 Containment Radiation Sampling Not Appl tcable
19.
vc 13*#
Containment Radiation Sampling Not App I tcab le
20.
vc 14*#
Containment Radiation Sampling Not App 1i cab I e
21.
fuel Transfer Tube Not Applicable
e
"'oO.F1 TO AGlt:._.,,, \\.\\
ATt~C.U~~ P1tCLf*
PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 At least three independent steam generator auxiliary feedwater pumps and associated 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:
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:
1.
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 1 3/4 7-5
PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION
3. 7*. 1. 2 At least three independent steam generator auxi lf ary feedwater,pumps J
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 ;equired 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. 1.
3.
Verifying that each non-automatic valve in the fl ow path that is not locked, sealed or otherwise secured in position, is in its correct position.
SALEM - UNIT I 3/4 7-5
PLANT SYSTEMS SURVEILLANCE REQUIREMEHTS (Continued)
4.
Verify that valves 11AF3, 12AF3, 13AF3, 11AF20, 12AF20, l3AF20, 14AF20, 11AF22, 12AF22, l3AF22, 14AF22, llAFlO, 12AF10, 13AF10, 14AF10, 11AF86, 12AF86, 13AF86, and 14AF86 are locked open.
b.
At least once per 18 months during shutdown by:
c.
1.
Verifying that 1ach au~matic 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 feectwater pump starts as designed automatically upon receipt of each auxiliary feectwater actuation test signal.
lhe auxiliary feedll'atgr 1)'5Ue tMll k dfttonstrlttd OPHUJ!L~ prior to tntry into Modr 3 fotl0tt!ng Pach COLO SlfJ!DO'.m by ~r~or~1n9 a fl0ti te~t to verify the ooM'al flow pith!. fr tt\\f ~.:' hary feed*
~ater Storage Tao~ to ~ath of tf\\e ~t#aa qttn11rater~.
SALEM - UNIT I 3/4 7-6
BASES 3/4.3.3.6 FIRE DETECTION INSTRUMENTAT!ON OPERABILITY of the fire detection 1nstr1J11entat1on ensures that adequate warning capability fs ava11 able for the prompt detection of fires. This capability is required fn order to detect and locate f1res in their early stages. Prompt detectfon of fires will reduce the poten-tial for damage to safety related equipment and 1s an 1ntegra1 element in the overall facility fire protection program.
In the event that 1 portion of the fire detection instrumentation is inoperable. the establishment of frequent fire patrols in the affected areas is required to provide detection capabii1ty unt11 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 Following and Accident," December 1975.
SALEM - UNIT l B 3/4 3-3
{
e I
~.J
~01>1 F1 T" ~tee w mt
'\\"1"TlltCtfCA
~~
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 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 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-t is taken for a direct reactor trip on the loss of 1oad) and a1so 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 will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code.*
SALEM - UNIT l B 3/4 4-la Amendment No. 24
l
3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.2 SAFID 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 ~er 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 will 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 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 wi11 occur only during shutdown and will be performed 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 l B 3/4 4-la
-~.
REACTOR COOLANT SYSTEM BASES 3/4.4.4-PRESSURIZER press 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 wi11 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 stea#m 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 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 1 oca ted 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 establish 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 wi11 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 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 will 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 SOD 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 l B 3/4 4-2
CONDITIONS FOR LICENSE TO BE ADDED TO FAC~.Il'Y OPERATING LICENSE DPR-70 1
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 a 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:
1.
Provisio"s establishing preventive maintenance and periodic visual 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 1 program which will ensure the capability to accurately determine.the airborne iodine concentration in vital areas under accident conditions.** This program shall include the following:
1.
Training of personnel,
2.
Procedures for monitoring, and
3.
Provisions for maintenance of sampling and analysis equipment.
c. Ba:k~: Method for Determining SJbcooling Margin The lice~see shall implement a program which will ensure the capability to accurately monitor the Reactor Coolant System subcooling margin.
This progra~
st'1all include the followin-g:
l.
Training of personnel, and
2.
Procedures for monitoring.
'_:_

ML18086A949

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