Text

Proposed Tech Specs,Incorporating Many Generic Improvements Agreed Upon Between Industry & NRC Since GGNS Conversion to ITS
	ML20211E715
Person / Time
Site:	Grand Gulf
Issue date:	08/20/1999
From:	; ENTERGY OPERATIONS, INC.
To:	;
Shared Package
ML20211E711	List: ML20211E706; ML20211E715;
References
	NUDOCS 9908300082
	Download: ML20211E715 (131)
	v • d • e

SLs Attachment 3 2.0 Page 2

( ,' , 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow < 10% rated core flow:

THERMAL POWER shall be s 25% RTP.

2.1.1.2 With the reactor steam dome pressure m 785 psig and core flow = 10% rated core flow:

MCPR shall be = 1.11 for two recirculation loop operation or a 1.12 for single recirculation loop operation.

2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel..

2.1.2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall be s 1325 psig.

{

2.2 SL Violations wMW 2 ?

With any SL violation, the following actions shall be completed:

g 2.2.1 Within I hour, notify the NRC Operations Center, in accordance)

8th 10 Cro 50.7?. -

q2.2.2 Within2 hours:

2.2 % 1 Restore compliance with all SLs; and 2.2. 2 Insert all insertable control rods.

7;,; ungu4. 34 km,,o notifv the plan", mana<1er and the corporap executive responsible for overal' plan; nucie.1 . . 4. ,

MCPR values in T.S. 2.1.1.2 are applicable only for cycle 10 operation. l (continued)

GRAND GULF 2.0-1 Amendment No. 120,131,136 9908300082 990820

!, DR ADOCK 0 y 46

bb5 I Attaclunent 3 Page 3 2.0 l

1 2.0 SLs SL Violations Tc M _

2.2.4 30 days, a Licensee Event Report (LER) shall be prepared pursuant 0.73. The LER shall be submitted to the NRC, the plant manager an orate executive responsible for overall plant nuclear safety.

2.2.5 ion of the unit shall not be resumed until au d

~

l l

l

(

i

(

GRAND GULF 2.0-2 Amendment No. 120

^ '

pj[c f"* 3 LCO Applicability 3.0

(

3.0 LCO APPLICABILITY LCO 3.0.4 specified conditions in the Applicability'that are required (continued) to comply with ACTIONS or that are part of a shutdown of the unit.

Exceptions to this Specification are stated.in the individual Specifications.fThese ns allow entry

-te " F5 or other spec 1ried conditions in Applicab b the associated ACTIONS to be en e allow unit operation in other specified conau.im. i- th- l oplicability only o . it=d neriod of i

-LCO 3.0.4 is only applicable for entry into a MODE or other specified condition in the Applicability in MODES 1, 2, and 3.

~

LCO 3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to

( demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to LC0 3.0.2 for the system returned to service under administrativ6 centrol to perform the testing required to demonstrate OPERABILITY.

LCO 3.0.6 When a supported system LCO is not met solely due to a support system LCO not being met, the Conditions and -

Required Actions associated witti this supported system are not required to be entered. Only the support system LCO ACTIONS are required to be entered. This is an exception to LCO 3.0.2 for the supgorted system. In_this eve _nt, _

-- . u...

%.JI on evaluatione e ; s. -- u.;; =n rt in be_

accordance with Specification 5.5.10,' Safety : unction e.cCcM ,

Determination Program (SFDP)." If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

When a support system's Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LCO 3.0.2.

t (continued) l GRAND GULF < 3.0-2 Amendment No. 120 j

f Attachment 3 Page5 SDM 3.1.1

[

3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 SHUTDOWN MARGIN (SDM) wMM n -Ne. \tm:4s pen nsD CO G7 LCO 3.1.1 SDM shal h

a. a 6.30" * /k. with the highest worth control rod i analytically de e -

r i b. 2: . with the highest worth co d determined by 1

APPLICABILITY: MODES 1, 2, 3, 4, and 5.

ACTIONS CONDITION REQUIRED ACTION COMRLETl0N TIME A. SDM not within limits A.1 Restore SDM to within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

( in MODE 1 or 2. limits.

i B. Require:1 Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months I

associbed Completion Time of Condition A not met.

I C. SDM not within limits C.1 Initiate action to Immediately in MODE 3. fully insert all insertable control rods.

D. SDM not within limits D.1 Initiate action to Immediately in MODE 4. fully insert all insertable control rods.

AND

(

(continued)

GRAND GULF 3.1-1 Amendment No. 120 J

Attachment 3 Page 6 SDM 3.1.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.1.1 Verify SDM s: Prior to each in vessel fuel gf a: 0.38% thehighestwork movement during I

control rod analytica rmined; : fuel loading ,

o sequence

b. /k with the h g th' AND k control riithfeteratined by tC Once within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after do be.wIkkIlm'+y) criticality following fuel movement within the reactor pressure vessel or control rod replacement

(

GRAND GULF 3.1-4 Amendment No. 120

^" C'f"*"' 3 p Control Rod OPERABILITY 1 3.1.3 i . ,.

', 3.1 REACTIVITY CONTROL SYSTEMS 3.1.3 Control Rod OPERABILITY LCO 3.I.3 Each control rod shall be OPERABLE.

APPLICABILITY: MODES I and 2.

ACTIONS

..............................__.....N0TE----------------..---.-...-..- ------

Separate Condition entry is allowed for each control rod.

CONDITION REQUIRED ACTION COMPLETION TIME A. One withdrawn control ------------NOTE-------------

rod stuck. A stuck rod may be bypassed

( in the Rod Action Control System (RACS) in accordance with SR 3.3.2.1.9 if required to allow continued operation.

k Disarm the associated control rod drive 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months (CRD).

AND (continued) bTn:5=.d

3. I ~7 A L i

GRAND GULF ,. 3.1-7 Amendment No. 120

Attachment 3 Insert 3.1-7A Page 8 1

A.1 Verify stuck control rod immediately separation criteria are met.

AND l

- . j C

i I

=

I

Attaciunent 3 Pagc 9 Control Rod OPERABILITY 3.1.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A. h Perform SR 3.1.3.2 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from and SR 3.1.3.3 for discovery of each withdrawn t- THERMAL POWER OPERABLEcontrolrod.( greater than the low power g g. ), n A setpoint (LPSP)

Cen4. m,.h of the Rod con c. Pattern Control l System (RPCS)

AND A. N Perform SR 3.1.1.1. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months I B. Two or more withdrawn B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months control rods stuck.

(.

C. One or more control --------NOTE---------

rods inoperable for Inoperable control reasons other than rods may be bypassed Condition A or B. in RACS in accordance <

with SR 3.3.2.1.9, if required, to allow insertion of inoperable control rod and continued operation.

C.1 Fully insert 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months inoperable control rod.

AND C.2 Disarm the asso::iated 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months CRD.

(continued)

GRAND GULF 3.1-8 Amendment No. 120

Anachment 3 Page 10 Control Rod Scram Times 3,g,4 4

3.1 REACTIVITY CONTROL SYSTEMS 3.1.4 Control Rod Scram Times LCO 3.1.4 a. No more than 14 OPERABLE control rods shall be " slow,"

in accordance with Table 3.1.4-1; and

, b. No OPERABLE control rod that is " slow" shall occupy a l location adjacent to another OPERABLE control rod that l 1s " slow" @ e it;,d d .. ce,,tr;T ret if id d. L l

l APPLICABILITY: MODES 1 and 2.

l l

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME

~

i A. Requirements of the A.1 Be in MODE 3. 12 ho6rs' '

LCO not met.

I l

GRAND GULF 3.1-12 Amendment No. 120

l 1

Attachment 3 Page 11 RCS Leakage Detection Instrumentation 3.4.7 e

3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.7 RCS Leakage Detection Instrumentation 1

I LCO 3.4.7 The following RCS leakage detection instrumentation shall be OPERABLE:

a. Drywell floor drain sump monitoring system;

b. One channel of either drywell atmospheric particulate or atmospheric gaseous monitoring system; and

c. Drywell air cooler condensate flow rate monitoring system.

1

. l

, APPLICABILITY: MODES 1, 2, and 3.

. ACTIONS _ )

CONDITION REQUIRED ACTION COMPLETION TIME

~

A. Drywell floor drain -------------NOTE------------

sump monitoring system , LCO 3.0.4 is not applicable.

inoperable. (-----------------------------

A.1 Restore drywell floor drain sump monitoring 30 days system to OPERABLE status.

l B. Required drywell B.1 Analyze grab samples Once per atmospheric monitoring of drywell 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months system inoperable. atmosphere.

(continued) l l- GRAND GULF 3.4-16 Amendment No. 120

Attachment 3 Page 12 RCS Leakage Detection Instrumentation 3.4.7 l (

, ACTIONS (continued) l CONDITION REQUIRED ACTION COMPLETION TIME C. Drywell air cooler -------------NOTE------------

condensate flow rate Not applicable when the monitoring system required drywell atmospheric inoperable. monitoring system is inoperable.

C.1 Perform SR 3.4.7.1.

Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months D. Required drywell -------

---NOTE------ ----

atmospheric monitoring L .0.4 is licab system inoperable. Q - _-------------- --

AND D.1 Restore required . .

drywell atmospheric 30 days !

Drywell air cooler monitoring system to 1 condensate flow rate OPERABLE status.

i monitoring system ;

inoperable. QR j D.2 Restore drywell air cooler condensate 30 days flow rate monitoring system to OPERABLE status.

E. Required Action and E.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A, AND B, C, or D not met.

E.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months F. All required leakage F.1 Enter LC0 3 0.1. Immediately i

~ '

detection systems inoperable.

l i GRAND GULF

3.4-17 Amendment No. 120 1

l i

Attachment 's Page 13 RHR Shutdown Cooling System-Hot Shutdown 3.4.9

(

, 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.9 Residual Heat Removal (RHR) Shutdown Cooling System-Hot Shutdown LC0 3.4.9 Two RHR shutdown cooling subsystems shall be OPERABLE, and, with no recirculation pump in operation, at least one RHR shutdown cooling subsystem shall be in c,peration.

............................N0TES-.-------------------------

1. Both RHR shutdown cooling subsystems and recirculation -

pumps mayqs rusved T@oper tion for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period, g g

2. One RHR shutdown cooling subsystem may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for performance of Surve111ances.

APPLICABILITY: MODE 3 with reactor steam dome pressure less than the. RHR cut in permissive pressure.

ACTIONS

{

..................................----NOTFS------------------------------------

1. LCO 3.0.4 is not applicable.

2. Separate Condition entry is allowed for each RHR shutdown cooling subsystem.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or two RHR A.1 Initiate action to Immediately shutdown cooling restore RHR shutdown subsystems inoperable. cooling subsystem (s) to OPERABLE status.

AND (continued)

GRAND GULF 3.4-21 Amendment No. 120 l

o l-Attachment 3 Page 14 RHR Shutdown Cooling System-Cold Shutdown 3.4.10

\'

l 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.10 Residual Heat Removal (RHR) Shutdown Cooling System-Cold Shutdown LCO 3.4.10 Two RHR shutdown cooling subsystems shall be OPERABLE, and, with no recirculation pump in operation, at least one RHR shutdown cooling subsystem shall be in operation.

...........................-N0TES---------------------------

1. Both RHRfutdown coogsubsystems and recirculation pumps may s t a.g fr e operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . g g

2. One RHR shutdown cooling subsystem may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for the performance of Surveillances.

3. Both RHR shutdown cool __ing subsystems and recirculation pumps ma?-n 7:- ;;; 77;4 eration during RCS inservice

. leak and hydrostatic testi')g. i y gQ .

(- APPLICABILITY: MODE 4.

ACTIONS

....................................-NOTE---------------------------..--------

Separate Condition entry is allowed for each RHR shutdown cooling subsystem.

CONDITION REQUIRED ACTION COMPLETION TIME

! A. One or two RHR A.1 Verify an alternate I hour I

shutdown cooling method of decay heat l subsystems inoperable. removal is available AND for each inoperable RHR shutdown cooling Once per subsystem. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

thereafter (continued) 4 GRAND GULF 3.4-24 Amendment No. 120 .

Attaclunent 3 Page 15 Primary Containment Air Locks 3.6.1.2 ti..

SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.1.2.3 --------------NOTE--- -------

Only re to be performed u ry l r exit throu rimary containmen a l Q............ ..................

gG fif%

Verify only one door in the primary containment air lock can be opened at a Galk time.

SR 3.6.1.2.4 Verify, from an initial pressure of 18 months 90 psig, the primary containment air lock seal pneumatic system pressure does

- not decay at a rate equivalent to ' ~ ~ ~

> 2 psig for a period of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

c

(,

l l

1 GRAND GULF 3.6-8 Amendment No. 120

p i

Attachenent 3 PCIVs Page 16 3.6.1.3 j SURVEILLANCE REQUIREMENTS

. SURVEILLANCE FREQUENCY I l

l l SR 3.6.1.3.1 ------------------NOTES------------------

1. Only required to be met in MODES 1, 1 2, and 3.

I

2. Not required to be met when the 20 inch primary containment purge i valves are open for pressure control, ALARA, or air quality considerations ;

for personnel entry. Also, not required to be met during Surveillances or special testing on l the purge system that requires the valves to be open. The 20 inch primary containment purge valves shall not be open with the 6 inch !

primary containment purge or the '

drywell vent and purge supply and exhaust lines open. - --

Verify each 20 inch primary containment 31 days

(~, purge valve is closed.

SR 3.6.1.3.2 ------------------NOTES------------------ ,

1. Valves and blind flanges in high radiation areas may.be verified by use of administrative means.

, 2. Not required to be met for PCIVs that l

are open under administrative controls.

Verify each primary containment isolation 31 days manual valve and blind flange that is located outside primary containment,

'drywell, to be closed andduring steam tunnelconditions accident g )d is required is closed.

u y (c,EM se.nla.cS (continued) g g bat-w & :kcu M.S ;

(

GRAND GULF ' 3.6-14 Amendment No. 120

PCIVs Anaciunen 3 Page n 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued) I l

SURVEILLANCE FREQUENCY SR 3.6.1.3.3 ------------------NOTES------------------

1. Valves and blind flanges in high radiation areas may be verified by use of administrative means.

2. Not required to be met for PCIVs that are open under administrative controls.

Verify each primary containment isolation Prior to manual valve and blind flange that is entering MODE 2 located inside primary containment, or 3 from drywell, or steam tunnel and is required MODE 4, if not to be closed during acci enticonditions performed is closed. _ _

L --- within the na no4Tcksc0;suleb or. 8 previous 92 days-obewhs _ _

Se c.oesd

( SR 3.6.1.3.4 Verify the isolation time of each power In accordance operated and each automatic PCIV, except with the MSIVs, is within limits. Inservice Testing Program (continued)

GRAND GULF 3.6-15 Amendment No. 440 128 l e

E Attachment 3 Secondary Containment Page 18 3.6.4.1

(, ACTIONS (continued)

. CONDITION RE0VIRED ACTION COMPLETION TIME C. Secondary _ C.1 --------NOTE---------

containment LCO 3.0.3 is not inoperable during applicable.

movement cf irradiated ---------------------

fuel assemblies in the primary or secondary Suspend movement of Immediately containment, during irradiated fuel CORE ALTERATIONS, or assemblies in the during OPDRVs. primary and secondary ;

containment. l i

AND l C.2 Suspend CORE Immediately l ALTERATIONS. l AND

~

C.3 Initiate action to Immediately' suspend OPDRVs.

(,'

l SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.1.1 Verify all auxiliary building and 31 days enclosure building equipment hatches and blowout panels are closed and sealed.

- I SR 3.6.4.1.2 Verif auxiliary building and 31 days enclosure building access door.is closed, except when the access openinglis bein used for entry and exit. Afgh o#o (continued) t GRAND GULF

3.6-43 Amendment No. 120

Attachment 3 AC Sources-Operating Page 19 3.8.1 ACTIONS (continued) ;

CONDITION REQUIRED ACTION COMPLETION TIME H. Three or more required H.1 Enter LCO 3.0.3. Immediately AC sources inoperable.

i SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.1.1 Verify ccrrect breaker alignment and 7 days indicated power availability for each required offsite circuit.

SR 3.8.1.2 -------------------NOTE--------------------

All DG starts may be preceded by an engine

( prelube period and followed by a warmup period prior to loading. l i

Verify each DG starts from standhv 31 days Tn s 10 seconds, k conditions > 3744 andV ana- achievesf4576 s~

frequency a 58.8 Hz and s oi.2 A.

V and N

N -gn.,cc+ (continued)

, 3.1i- S A)

GRAND GULF 3.8-5 Amendment No. 440, 134

Attachment 3 Page 20

' insert 3.8-5A 6.5.

. a,- in s 10 seconds, voltage 2 3744 V and frequency 2 58.8 Hz ; and

b. steady state voltage 23744 V and s 4576 V and frequency 2 58.8 Hz and s 61.2 Hz.

l l

i

(

l

Attachment 3 AC Sources-Operating Page 21 381

/

SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.12 -------------------NOTES-------------------

1. All DG starts may be preceded by an engine prelube period.

2. This Surveillance shall not be performed in MODE 1 or 2. However, credit may be taken for unplanned events that satisfy this SR.

Verify on an actual or simulated Emergency 18 months Core Cooling System (ECCS) initiation signal each DG auto-starts from standby condition and:

a. In s 10 seconds after auto-start and during t sts, achieves voltage Tnd a 3744 ,,,e a $7C C, .y,4 '?o $

b. L 4 Au w w d 2't 2"" d a r' 2rS ,

du, um t::t:,AchieveQequency

a: 58.8 Hz and s 61.2 Hz; F g 3.5-Io O h

c. Operates for a 5 minutes; and

d. Emergency loads are auto-connected to the offsite power system.

(continued) i GRAND GULF 3.8-10 Amendment No. 120 1

f I

(

l Insert 3.8-10A ^*CI"" cat 3 Page 22 l .,

r.

frequency 2 58.8 Hz j insert 3.8-10B steady state voltage 23744 V and s 4576 V and i

e t

($

, 1 l

t l l l

l I

l l

l t

l L.m

Attachment 3 AC Sources-Operating Page 23 3.8.1

.' SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.15 ------------------NOTES-------------------

1. This Surveillance shall be performed within 5 minutes of shutting down the DG after the DG has operated a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or until operating temperatures stabilized loaded a 5450 kW and s 5740 kW for DG 11 and DG 12, and a 3300 kW for DG 13.

Momentary transients outside of the load range do not invalidate this test.

2. All DG starts may be preceded by an engine prelube period.

I lng 3.t0A[

~

Verify each DG starts and achieve , in 8 months' 510 _ii3Wtace 2 3744 V and s 4576 V a_nd frequency a 58.E nz enu s 61.2 rit.y y SR 3.8.1.16 -------------------NOTE--------------------

This Surveillance shall not be performed in MODE 1, 2, or 3. However, credit may be taken for unplanned events that satisfy this SR.

Verify each DG: 18 months

a. Synchronizes with offsite power source while loaded with emergency loads upon a simulated restoration of offsite power;

b. Transfers loads to offsite power source; and

c. Returns to ready-to-load operation.

(continued)

(

GRAND GULF 3.8-13 Amendment No. 120 c

- Insert 3.8-13A Attaciunent 3

- - Page 24 l

l a- in s 10 seconds, voltage 2 3744 V and frequency 2 58.8 Hz ; and

b. steady state voltage 23744 V and s 4576 V and frequency 2 58.8 Hz and s 61.2 Hz.

[

9 4

6 g fe l

f Attachment 3 AC Sources-Operating Page 25 3.8.1

( ,- SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.20 -------------------NOTE--------------------

All DG starts may be preceded by an engine prelube period.

Verify, when started simultaneously fro 10 years standby enndition. each DG achieve , n ~

7 :--ands, voltage a 3/** v and s 4576 Y bndfraninencya58.8Hzands01.4nt.

L 7n%tN-3,1-10o A 3

(,

1 j

i 1

(

GRAND GULF 3.8-16 Amendment No. 120 4 i

l Attachment 3 Page 26 Insert 3.8-16A

( .-

a. In s 10 seconds, voltage 2 3744 V and frequency 2 58.8 Hz ; and

b. steady state voltage 23744 V and s 4576 V and frequency 2 58.8 Hz and s 61.2 Hz. ,

9

. 6 e

.e

~

l l

Attachment 3 Diesel Fuel Oil, Lube Oil, and Starting Air Pagc 2'1 3.8.3 (i.. _

, SURVEILLANCE REQUIREMENTS (continued) l SURVEILLANCE FREQUENCY

.8.3.6 For each fuel oil storage tank: 10 years a.

] Drain the fuel o ;

b. Remove the sediment; and

c. Clean the tank.

l

(' . l GRAND GULF 3.8-25 Amendment No. 120 i-

f

d DC Sources-Operating p,'l', j'nent g 3 3.8.4 i l

l [ SURVEILLANCE REQUIREMENTS (continued)

. SURVEILLANCE FREQUENCY ;

1 SR 3.8.4.3 Verify battery cells, cell plates, and 18 months racks show no visual indication of physical j damage or abnormal deterioratiog 6 4 too\J ~dtwea A - hdkscu m

~ dEm .M, SR 3.8.4.4 Remove visible corrosion and verify battery 18 months cell to cell-and terminal connections are coated with anti-corrosion material.

SR 3.8.4.5 Verify battery connection resistance is 18 months s 1.5 E-4 ohm for inter-cell connections, s 1.5 E-4 ohm for inter-rack connections, s 1.5 E-4 ohm for inter-tier connections,

. and s 1.5 E-4 ohm for terminal connections. ,,

SR 3.8.4.6 Verify each Division 1 and 2 required 18 months (T- battery charger supplies a 400 amps at a 125 V for a 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months ; and the Division 3 battery charger supplies a 50 amps at a 125 V for a 4 hcurs.

(continued) i GRAND GULF ' 3.8-28 Amendment No. 120 1

I

]

g.=

Attachment 3 Battery Cell Parameters Page 29 3.8.6 r

I 3.8 ELECTRICAL. POWER SYSTEMS 3.8.6 Battery Cell Parameters LCO 3.8.6 BatterycellparametersfortheDivgon1.2.and3 batteries shall be withidigh limitswi T.a 0.0.0 ".

APPLICABILITY: When associated DC electrical power subsystems are required to be OPERABLE.

l ACTIONS

....................................-NOTE--------------------- ---------------

l Separate Condition entry is allowed for each battery.

l l

. CONDITION REQUIRED ACTION COMPLEU0tl TIME l

l A. One or more batteries A.1 Verify pilot cell's I hour with one or more electrolyte level and

(+? battery cell float voltage meet

~

parameters not within Table 3.8.6-1 Table 3.8.6-1 Category Category C limits.

A or B limits.

AND

.l A.2 Verify battery cell 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months parameters meet Table 3.8.6-1 AND Category C limits.

Once per 7 days thereafter AND A.3 Restore battery cell 31 days parameters to Category A and B limit N Table 3.5.53 (continued) t, GRAND GULF 3.8-34 Amendment No. 120 e j 4

~

r Attachmen 3 Battery Cell Parameters Page 30 3.8.6 4

t' ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME I

B. Required Action and B.1 Declare associated Immediately l

associated Completion battery inoperable.

l Time of Condition A i not met.

One or more batteries with average electrolyte temperature of the representative cells

< 60*F.

E One or more batteries --

with one or more j

l battery cell l parameters not within l Category C limits.

l

% blo S L O l

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.6.1 Verify battery cell parameters meet 7 days Table 3.8.6-1 Category A limits.

(continued) l I

GRAND GULF 3.8-35 Amendment No. 120

p [3j""*"'3 RHR-High Water Level 3.9.8 3.9 REFUELING OPERATIONS 3.9.8 Residual Heat Removal (RHR)-High Water Level LC0 3.9.8 One RHR shutdown cooling subsystem shall be OPERABLE and one decay heat removal subsystem shall be in operation.

.......................-----N0TE-------------.--.---_--------

The required decay heat removal subsystem may@ rupp %

nof,c,M s operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period.

APPLICABILITY: MODE 5 with irradiated fuel in the reactor pressure vessel (RPV) and the water level 2: 22 ft 8 inches above the top of the RPV flange.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME

( A. Required RHR shutdown A.I Verify an alternate I hour i cooling subsystem method of decay heat !

inoperable. removal is available. AND Once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter i

B. Required Action and B.I Suspend loading Immediately associated Completion irradiated fuel Time of Condition A assemblies into the not met. RPV.

AND B.2 Initiate action to Immediately restore secondary containment to OPERABLE status.

AND (continued) i GRAND GULF 3.9-10 Amendment No. 120

I l Attachment 3

- ow Water Level Page 32 399

~

1

( 3.9 REFUELING OPERATIONS j l 1 3.9.9 Residual Heat Removal (RHR)-Low Water Level )

LC0 3.9.9 Two decay heat removal subsystems shall be OPERABLE, and one decay heat removal subsystem shall be in operation.

........................----N0TE-------------.--------------

The required operating decay heat removal subsystem mayW 00, O t = =:f E operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period.

s r1 ............................................................

APPLICABILITY: MODE 5 with irradiated fuel in the reactor pressure vessel (RPV) and the water level < 22 ft 8 inches above the top of the RPV flange.

ACTIONS

.........................................N0TE-------------.-..----. 7.-------

$$I[ $. $" $$. " $5. I. ..$$$ . $$.$IS . $$I5. $I .5$$$!I..I $5! $I...

p' CONDITION REQUIRED ACTION COMPLETION TIME l

A. One or two required A.1 Verify an alternate 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months i decay heat removal method of decay heat subsystems inoperable. removal is available AND for each inoperable required decay heat Once per removal subsystem. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter l

(continued)

(

GRAND GULF 3.9-12 Amendment No. 120 t

Anaclunent 3 Progrcms and Manuals Page 33 5.5

\

l

( 5.5 Programs and Manuals (continued) 5.5.5 Component Cyclic or Transient Limit This program provides controls to track the cyclic and transient occurrences identified on UFSAR Table 3.9-35 to ensure that the reactor vessel is maintained within the design limits.

5.5.6 Inservice Testina Program This program provides controls _for <nservicettesting of ASME Code Class 1, 2, and 3 componenM.r.tM M idh : ;;xft3. The program shall include the followTng:

a. Testing frequencies specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Mdenda as follows:

ASME Boiler and Pressure Vessel Code and . -.

applicable Addenda terminology for Required Frequencies inservice testing for performing inservice

( activities testinc activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every At least once per 92 days 3 months Semiannually or At least once per 184 days every 6 months J Every 9 months At least once per 276 days '

Yearly or annually At least once per 366 days Biennially or every At least once per 731 days 2 years

b. The provisions of SR 3.0.2 are applicable to the above required frequencies for performing inservice testing i activities; !

c. The provisions of SR 3.0.3 are applicable to inservice testing activities; and

d. Nothing in the ASME Boiler and Pressure Vessel Code shall be i construed to supersede the requirements of any TS. 1 (continued) !

i GRAND GULF 5.0-11 Amendment No. 120 I

i 1

j

Anacimiem 3 Programs and Manuals Page 34 5.5

(' 5.5 Programs and Manuals 5.5.8 Explosive Gas and Storace Tank Radioactivity Monitorina Procram (continued)

b. A surveillance program to ensure that the quantity of radioactive material contained in any outside temporary tank not including liners or shipping radwaste is 5 10 curies, excluding tritium and dissolved or entrained noble gases.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies.

5.5.9 Diesel Fuel Oil Testino Procram A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. The purpose of the program is to establish the following:

g:

a. Acceptability of new fuel oil for use prior to addition to s

storage tanks, and acceptability of stored fuel oil every 92 days, by determining that the fuel oil has:

1. a water and sediment contents within limits, and

2. a kinematic viscosity within limits for ASTM 2D fuel oil;

b. Total particulate concentration of the new fuel is 5 2 mg/100 m1 when tested in accordance with ASTM D-2274-70 within 7 days after addition of the new fuel to the storage tank; and

c. Total particulate concentration of the fuel oil in the storage tanks is 5 2 mg/100 ml when tested every 92 days in 1

-[g g , accordance with ASTM D-2274-70.

Jo-NA m 5.5.10 Safety Function Determination Proaram (SFDP)

This program ensures loss of safety function is detected and appropriate actions taken. Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other appropriate limitations and remedial (continued)

GRAND GULF 5.0-14 Amendment No. 120 e

Attaciunent 3 Page 35 Insert 5.0-14A 0 -

' The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program

. testing frequencies.

t ' 6 m

(

I l.

k.

I L_

f Attachment 3 Reporting Requirements

, Page 36 5.6 5.6 Reporting Requirements

{

5.6.2 Annual Radiological Environmental Operatino Report (continued) results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible.

5.6.3 Radicactive Effluent Release Report The Radioactive Effluent Release Report covering the operation of the unit during the previous calendar year shall be submitted by May 1 of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCM and process control program and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B.1.

5.6.4 Monthly Operatino Reports Routine reports of operating statistics and shutdown experience, ,

( including documentation of all challenges to the main steam i safety / relief valves, shall be cubmitted on a monthly basis no l later than the 15th of each month following the calendar month covered by the report.

5.6.5 Core Operatino limits Report (COLR) l

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

1) LCO 3.2.1, Average Planar Linear Heat Generation Rate (APLHGR),

2) LCO 3.2.2, Minimum Critical Power Ratio (MCPR), an

3) LC0 3.2.3, Linear Heat Generation Rate (LHGR P (continued)

@ Lco '31.1, 5ks%m Mar @ (5Orn) i GRAND GULF 5.0-18 Amendment No. 120 e

F i

Reactor Core SLs l Attachment 3 Page 37 -

8 2.1.1 I l

BASES (continued) 1 SAFETY LIMIT VIOLATIONS k2.

j If any SL is violaleC-the 0 erations Center must be N , notified within I hour, in accor an CFR 50.72 i

j i

NHML i

1 v

Execeding an SL may cause fuel damage and create a potential for radioactive releases in excess of 10 CFR 100, " Reactor l Site Criteria," limits (Ref. 4). Therefore, it is required to insert all insertable control rods and restore compliance with the SL within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . (The required actions for a violation of the reactor water level SL include manually initiating ECCS to restore water level and depressurizing the reactor vessel, if necessary, for ECCS operation.) The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time ensures that the operators take prompt remedial action and also ensures that the probability of an accident occurring during this period is' miiiiiial.

2.2.3 any SL is violated, e General Manager, P t Operatio an he Vice President, Op tions GGNS shall be tified within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months riod provides time plant operators d staff to take the repriate immediate a ion {

and assess t ondition of the uni efore reporting to t senior managemen

{2.2.4 any SL is violated, a License vent Report shall be '

prep and submitted within 30 da to the NRC in accordanc ith 10 CFR 50.73 (Ref. 5). The report will N describe the icable circumstances pre ing the violation, the ef of the violation upon t components, systems, or structures, d the corrective acti taken to prevent recurrence. A copy the report shall al be submitted to the General Manage , lant Operations an he e President, Operations GGNS.

(continued)

GRAND GULF B 2.0-5 Revision No. I

r l

Att itment 3 Pag,s 311 Reactor Core SLs l B 2 d .1 l l

BASES

(..

SAFETY LIMI VIOLATIONS g (continued) I violated, restart of the unit shall not commence unti the NRC. This requirement es the NRC that all necess s, analyses, and i action completed before the unit beg n tart to Qrmal operat v -

REFERENCES 1. 10 CFR 50, Appendix A, GDC lu.

2. XN-NF524(A), Revision 2, April 1989.

'3 IO OTR 00.7Q :

i ydpp

4. 10 CFR 100.

Sbi." 50.7h ' ~~

6. NEDE-24011-P-A, GESTAR-II. l l

7. GNRI-96/00230, Amendment 131 to the Operating License.

l l'

l 1

i l

l 1

i GRAND GULF B 2.0-6 Revision No. 2

Attaclunent 3 Page 39 RCS Pressure SL B 2.1.2

( ._.

BASES APPLICABLE The RCS pressure SL has been selected such that it is at a SAFETY ANALYSES pressure below which it can be shown that the integrity of (continued) the system is not endangered. The reactor pressure vessel is designed to ASME, Boiler and Pressure Vessel Code,Section III, 1971 Edition, including Addenda through the winter of 1972 (Ref. 5), which permits a maximum pressure transient of 1107,, 1375 psig, of design pressure 1250 psig.

The SL of 1325 psig, as measured in the reactor steam dome, is equivalent to 1375 psig at the lowest elevation of the RCS. The RCS is designed to ASME Code,Section III, 1974 Edition (Ref. 6), for the reactor recirculation piping, which permits a maximum pressure transient of 1107,of design pressures of 1250 psig for suction piping,1650 psig for discharge piping between the pump and the discharge valve, and 1550 psig beyond the discharge valve. The RCS pressure SL is selected to be the lowest transient overpressure allowed by the applicable codes.

SAFETY LIMITS The maximum transient pressure allowable in the RCS pressure vessel under the ASME Code,Section III, is 1107,of design pressure. The maximum transient pressure allowable in the

( RCS piping, valves, and fittings is 1107, of design pressures .

of 1250 psig for suction piping,1650 psig for discharge piping between the pump and the discharge valve, and 1550 psig beyond the discharge valve. The most limiting of these allowances is the 1107, of the suction piping design pressure; therefore, the SL on maximum allowable RCS pressure is established at 1325 psig as measured at the reactor steam dome.

APPLICABILITY SL 2.1.2 applies in all MODES.

SAFETY L.2.d VIOLATIONSLIMIT 4 any SL is violated, the NRC Operat enter must be

' notu within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , in accordance with R 50.72 (Ref.7).

Exceeding the RCS pressure SL may cause immediate RCS l

, failure and create a potential for radioactive releases in !

(continued)

GRAND GULF B 2.0-8 Revision No. O i

i

r i

AHachment 3 RCS Pressure St Page 40 B 2.1.2

,. BASES 0

SAFETY LIMIT C 'a ntin nd) V VIOLATIONS excess of 10 CFR 100, " Reactor Site Criteria," limits (Ref. 4). Therefore, it is required to insert all insertable control rods and restore compliance with the SL within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time ensures that the operators take prompt remedial action and also ensures that the probability of an accident occurring during this period is minimal.

2.2.3' -

f any SL is violated, the Gen 1 Manager,PlantOperat' ion)s a the Vice President, Operation GNS shall be notified I wit in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months peri rovides time for plant opera rs and staff to take the approp ate immediate action and ass s the condition of the unit bef reporting to the

. senior ma ement. -.

2.2.4 a l l If any SL is violated, Licensee Event Report shall b prepared and submitted w (hin 30 days to the NRC in ,

accordance with 10 CFR 50.7 (Ref. 8) . The report will describe the applicable circ tances preceding the l

\

violation, the effect of the y ation upon unit components, stems, or structures, and the c rective actions taken to pr nt recurrence. A copy of the ort shall also be i subai d to the General Manager, Pla Operations and the l Vice Pres nt, Operations GGNS.

2.2.5 If any SL is violated, r art of the unit shall n t commence until authorized b he NRC. This require nt ensures the NRC that all neces reviews,- analyses, nd actions are completed before the un t begins its restar to normal operation.

(continued)

GRAND GULF B 2.0-9 Revision No. 1

RCS Pressure St Attachment 3 Page 41 B 2.1.2 BASES (continued) 4 REFERENCES 1. 10 CFR 50, Appendix A, GDC 14, GDC 15, and GDC 28.

2. ASME, Boiler and Pressure Vessel Code,Section III.

3. ASME, Boiler and Pressure Vessel Code,Section XI, Article IWA-5000.

4. 10 CFR 100. -

5. ASME, Boiler and Pressure Vessel Code, 1971 Edition, Addenda, winter of 1972.

6. ASME, Boiler and Pressure Vessel Code, 1974 Edition.

10CFR50.7D _

8. 10 CFR ,

GRAND GULF' B 2.0-10 Revision No. I

Attachment 3 Page 42 LCO Applicability B 3.0 j

, 4 BASES LCO 3.0.4 provisions of LCO 3.0.4 shall not prevent changes in MODES !

(continued) or other specified conditions in the Applicability that result from any unit shutdown.

Exceptions to LCO 3.0.4 are stated in the individual F

Imek- }

L g 9 (,.A \ Knecifications.4 Exceptions may apply to all the ACTIONS or i

to a specific Required Action of a Specification.

LC0 3.0.4 is only applicable when entering MODE 3 from MODE 4, MODE 2 from MODE 3 or 4, or MODE 1 from MODE 2.

Furthermore, LCO 3.0.4 is applicable when entering any other specified condition in the Applicability only while operating in MODE 1, 2, or 3. The requirements of LCO 3.0.4 do not apply in MODES 4 and 5, or other specified conditions of the Applicability (unless in MODE 1, 2, or 3) because the ACTIONS of individual Specifications sufficiently define'the remedial measure to be taken. . _ _

The ACTIONS for an inoperable required battery charger in LCO 3.8.4, "DC Sources - Operating," and LCO 3.8.5, "0C Sources - Shutdown," include a Note explicitly precluding I. entry into specific MODES or other specified conditions of the Applicability while relying on the ACTIONS. With an inoperable required battery charger this Note in LCO 3.8.4 ,

prohibits entry in MODE 1, 2, or 3, except during power decrease and in LCO 3.8.5 prohibits starting movement of irradiated fuel, entering MODE 4 from MODE 5, or loading fuel into the vessel if the vessel is defueled.

Surveillances do not have to be performed on the associated inoperable equipment (or on variables outside the specified limits), as permitted by SR 3.0.1. Therefore, changing MODES or other specified conditions whila in an ACTIONS Condition, either in compliance with LCO 3.0.4, or where an exception to LCO 3.0.4 is stated, is not a violation of SR 3.0.1 or SR 3.0.4 for those Surveillances that do not have to be performed due to the associated inoperable equipment. However, SRs must be met to ensure OPERABILITY

, prior to declaring the associated equipment OPERABLE (or variable within limits) and restoring compliance with the affected LCO.

LC0 3.0.5 LCO 3.0.5 establishes the allowance for restoring equipment to service under administrative controls when it has been removed from service or declared inoperable to comply with ACTIONS. The sole purpose of this Specification is to (continued)

GRAND GULF ,..

B 3.0-6 Revision No. O

Attaciunent 3 Page 43 Insert B3.0-6A These exceptions allow entry into MODES or other specified conditions in the Applicability when the associated ACTIONS to be entered do not provide for continued operation for an unlimited period of time.'

M e i

l t

1

Atta:hment 3 Control Rod OPERABILITY {

Page 44 8 3.1.3

-(

. BASES , ,

LCO satisfy the intended reactivity control requirements, strict (continued) control over the number and distribution of inoperable control rods is required to satisfy the assumptions of the DBA and transient analyses.

APPLICABILITY In MODES I and 2, the control rods are assumed to function during a DBA or transient and are therefore required to be i In MODES 3 and 4, control rods are OPERABLE in these MODES.

not able to be withdrawn since the reactor mode switch is in Shutdown and a control rod block is applied. This provides adequate requirements for control rod OPERABILITY during i these conditions. Control rod requirements in MODE 5 are i located in LCO 3.9.5, " Control Rod OPERABILITY-Refueling."

ACTIONS The ACTIONS table is modified by a Note indicating that a

'~

separate Condition entry is allowed for each control rod. !

This is acceptable, since the Required Actions for-each Condition provide appropriate compensatory actions for each inoperable control rod. Complying with the Required Actions may allow for continued operation, and subsequent inoperable control rods are governed by subsequent Condition entry and i application of associated Required Actions.

h.l.A.2 A.

A control rod is considered stuck if it will not insert by ;

either CRD drive water or scram pressure. With a fully inserted control rod stuck, no actions are required as long ,

as the control rod remains fully inserted. The Required !

Actions are modified by a Note that allows a stuck control rod to be bypassed in the Rod Action Control System (RACS) to allow continued operation. SR 3.3.2.1.9 provides ,

additional requirements when control rods are bypassed in i be*

g,f fyA (

l RACS to ensure compliance with the CRDA analysis. With one withdrawn control rod stuck4 the control rod must be disarmed within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . The allowed Completion Time of i

L 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months is acceptable, considering the reactor can still be shut down, assuming no additional control rods fail to insert, and provides a reasonable amount of time to perform the Required Action in an orderly manner. Isolating the control rod from scram prevents damage to the CRDM. The j (continued)

GRAND GULF B 3.1-14 Revision No. 0

(

Attachment 3 Page 45 insert B3.1 14A I:

the local scram reactivity rate assumptions may not be met if the struck control rod separation criteria are not met. Therefore, a verification that the separation criteria are met must be performed immediately. The separation criteria are not met if the struck control rod occupies a location adjacent to a " slow" control rod. The description of " slow" control rods is provided in LCO 3.1.4, " Control Rod Scram Times." In addition, i

1

. ~ .

i f

t

r Attachment 3 "E " Control Rod OPERABILITY l

B 3.1.3 BASES ACTIONS A.1. A.2 M A.3 [(continued) cwJL control rod can be isolated from scram by isolating the hydraulic control unit from scram and normal drive and withdraw pressure, yet still maintain cooling water to the CRD. nj '

E6 3.1-13 4 '

Monitoring of the insertion capability for each withdrawn !

control rod must also be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . )

SR 3.1.3.2 and SR 3.1.3.3 perform periodic tests of the control rod insertion capability of withdrawn control rods.

Testing each withdrawn control rod ensurn that a connric F problem does_ not exist.1 The allowed Completion Time of qw-t !

3,p;g "/ '24 nours,provides a reasonable time to test the control rods, considering the ootential for a need to reduce cowK l to oerform the tests.) Required ^.:ti:r. A.: b:: : :dt' icd

. -tfr: zere CeJ+1stien Tir:. Th: 2' h:;r C:;; h.t'en " r: f0r

-th;. % ;;;,ed A u ivu nou hs th: .:ithdr M : Mtfe L M

".fns't.xY i s di o..e - L tud .;d Tii;;".",L POT". i: ;. ute. thm i 63.1-is'qNthe actuai%ed iv ;x:r seiguint -(LPSP)*of the r:d att:ra"-

4 ent,wil.,l RPCfi since the notch insertions may not be l

compatible with the requirements of red pattern control (LCO 3.1.6) and the RPC (LCO 3.3.2.1, " Control Rod Bio Instrumentation"). 7 To allow continued operation with a withdrawn control rod stuck, an evaluation of adequate SDM is also required within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Should a DBA or transient require a shutdown, to 4 preserve the single failure criterion an additional control rod would have to be assumed to have failed to insert when required. Therefore, the original SDM demonstration may not be valid. The SDM must therefore be evaluated (by measurement or analysis) with the stuck control rod at its stuck position and the highest worth OPERABLE control rod assumed to be fully withdrawn.

The allowed Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to verify SDM is adequate, considering that with a single control rod stuck in a withdrawn position, the remaining OPERABLE control rods are capable of providing the required scram and shutdown reactivity. Failure to reach MODE 4 is only likely if an additional control rod adjacent tc' the stuck control rod also fails to insert during a required scram. Even with the postulated additional single failure of an adjacent control rod to insert, sufficient reactivity control remains to reach and maintain MODE 3 conditions (Ref. 7).

(continued)

GRAND GULF B 3.1-15 Revision No. O

[~T i

Attachment 3 Page 47 Insert B3.1-15A i

from discovery of Condition A concurrent with THERMAL POWER greater than the low power l

setpoint (LPSP) of the rod pattem controller (RPC)

Insert B3.1-15B from discovery of Condition A concurrent with THERMAL POWER greater than the LPSP of the of the RPC Insert B3.1-15C This Completion Time allows for an exception to the normal " time zero" for beginning the allowed outage time " clock." The Required Action A.2 Completion Time only begins upon discovery of Condition A concurrent with THERMAL POWER greater than l

l l

l

"*"3 Control Rod Scram Times B 3.1.4 i

BASES LCO To ensure that local scram reactivity rates are maintained (continued) within acceptable limits, no " slow" control rod may occupy a location adjacent to another " slow" control rod er ;dja;;r,t'

w . withdrawn dua centr:1 red.

Table 3.1.4-1 is modified by two No'tes, which state control rods with scram times not within the limits of the Table are considered " slow" and that control rods with scram times

> 7 seconds are considered inoperable as required by SR 3.1.3.4.

This LCO applies only to OPERABLE control rods since inoperable control rods will be inserted and disarmed (LC03.1.3). Slow scramming control rods may be conservatively declared inoperable and not accounted for as

" slow" control rods.

APPLICABILITY In MODES I and 2, a scram is assumed to function' d'uring transients and accidents analyzed for these plant conditions. These events are assumed to occur during startup and power operation; therefore, the scram function

( of the control rods is required during these MODES. In MODES 3 and 4, the control rods are not able to be withdrawn since the reactor mode switch is in shutdown and a control I rod block is applied. This provides adequate requirements for control rod scram capability during these conditions.

Scram requirements in MODE S are contained in LC0 3.9.5,

" Control Rod OPERABILITY-Refueling."

ACTIONS A.1 When the requirements of this LCO are not met, the rate of negative reactivity insertion during a scram may not be within the assumptions of the safety analyses. Therefore, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.

l i

(continued) i GRAND GULF B 3.1-23 Revision No. I

RCS Leakage Detection Instrumentation Attachmerit 3 Page 49 B 3.4.7 BASES

' I, APPLICABLE . RCS leakage detection instrumentation satisfies Criterion 1 SAFETY. ANALYSES of the NRC Policy Statement.

(continued) 1 LC0 The drywell floor drain sump monitoring system is required to quantify the unidentified LEAKAGE from the RCS. Thus, :

for the system to be considered OPERABLE, the sump level monitoring portion of the system must be OPERABLE. The other monitoring systems provide qualitative indication to the operators so closer examination of other detection ,

i systems will be made to determine the extent of any l corrective action that may be required. With the leakage !

detection systems inoperable, monitoring for LEAKAGE in the I RCPB is degraded. 1 APPLICABILITY In MODES 1, 2, and 3, leakage detection systems are required

~

to be OPERABLE to support LCO 3.4.5. This Applicability is

~ - l consistent with that for LCO 3.4.5.

1 ACTIONS A.1

{ . Tn3w-\- ) With the drywell floor drain sump monitoring system 6 3*V-3 p in perable, no other form of sampling can provide the equivalent information to quantify leakage. However, the drywell atmospheric activity monitor and the drywell air cooler condensate flow rate monitor will provide indications of changes in leakage. j With the drywell floor drain sump monitoring system !

inoperable, but with RCS unidentified and total LEAKAGE i being determined every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (SR 3.4.5.1), operation may continue for 30 days. The 30 day Completion Time of Required Action A.1 is acceptable, based on operating experience, considering the multiple forms nf laakaoa k(detection that by modified area note still tnat availtble/ Required Action states tha ons ofA.1 is u,6 0.0.1 '-- aat applicable. As a result, a M s allowed when the dryweli iiva d-un sump monitoring system is 1 - his allowance is provided oet.au a : tsar nstrumentation is to monitor RCS leakage.

(continued) i GRAND GULF. B 3.4-34 Revision No. O c

j

Att chment 3 Page 50 Insert B3.4-34A The Actions are modified by a Note that states that the provisions of LCO 3.0.4 are not ;

applicable. As a result, a MODE change is allowed when the drywell floor drain sump monitoring system and required radiation monitors are inoperable. This allowance is provided because other instrumentation is available to monitor RCS leakage.

1 l

1 9 6 e Y**

1 l

O j

i i.

1 I

i Attachment 3 RCS Leakage Detection Instrumentation Page 51 B 3.4.7 BASES-(,

l ACTIONS B.1 (continued)

With both ' gaseous and particulate drywell atmospheric l monitoring channels inoperable, grab samples of the drywell atmosphere shall be taken and analyzed to provide-periodic leakage information. Provided a sample is obtained and analyzed every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , the plant may continue operation since at least one other form of drywell leakage detection (i.e., air cooler condensate flow rate monitor) is available. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval provides periodic information that is adequate to detect LEAKAGE.

- C.1 With the required drywell air cooler condensate flow rate monitoring system inoperable, SR 3.4.7.1 is performed every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to provide periodic information of activity in the drywell'at a more frequent interval than the routine Frequency of SR 3.4.7.1. The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months interval prov4 des

. periodic information that is adequate to detect LEAKAGE and recognizes that other forms of leakage detection are available. However, this Required Action is modified by a i Note that allows this action to be not applicable if the l

(- required drywell atmospheric monitoring system is inoperable. Consistent with SR 3.0.1, Surveillances are not required to be performed on inoperable equipment.

l 0.1 and 0.2 1 4

With both the gaseous and particulate drywell atmospheric monitor channels and the drywell air cooler condensate flow ,

rate monitor inoperable, the only means of detecting LEAKAGE !

is the drywell floor drain sump monitoring system. This Condition does not provide the required diverse means of leakage detection. The Required Action is to restore either :

of the inoperable monitoring systems to OPERABLE status within 30 days to regain the intended leakage detection diversity. The 30 day Completion Time ensures that the plant will not be ope d in a degraded configuration for_a lenathy time Deriod. The Required Actions are moottled by al ,

Tote that states tha )

of LC0 3.0.4 are not M As a result, a MODE change both ;

r thg.,drywell gasevu. .r.d ; rtienlate atmospheric monitor n channess ved-dcjantQ, air cooler condensate i6 r;t:

(continued)

GRAND GULF ' B 3.4-35 Revision No. O

Attachment 3 RCS Leakage Detection Instrumentation Page 52 8 3.4.7 BASES 4

ACTIONS D.1 and 0.2 (continued) b b ring system are inoperable.

' This allowance is nenvided because otner instru=ut. Man h available to

<mnnitor KL3 i t:a kap . __

~

E.1 and E.2 If any Required Action of Condition A, B, C, or D cannot be met within the associated Completion Time, the plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the plant must be brought to at least .

MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions in an orderly manner and without challenging plant systems.

F.1 - - i With all required monitors inoperable, no required automatic i means of monitoring LEAKAGE are available, and immediate '

( plant shutdown in accordance with LC0 3.0.3 is required.

SURVEILLANCE SR 3.4.7.1 REQUIREMENTS This SR requires the performance of a CHANNEL CHECK of the required drywell atmospheric monitoring system. The check gives reasonable confidence that the channel is operating properly. The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is based on instrument reliability and is reasonable for detecting off normal conditions.

SR 3.4.7.2 This SR requires the performance of a CHANNEL FUNCTIONAL TEST of the required RCS leakage detection instrumentation.

The test ensures that the monitors can perform their function in the desired manner. The test also verifies the relative accuracy of the instrumentation. The Frequency of 31 days considers instrument reliability, and operating experience has shown it proper for detecting degradation.

(continued) i GRAND GULF B 3.4-36 Revision No. O l

1*

F 1

l RHR Shutdown Cooling System-Hot Shutdown f Attachment 3 g 3,4,9 Page 33 BASES l

LC0 OPERABLE RHR pump, two heat exchangers in series, and the (continued) associated piping and valves. Each shutdown cooling subsystem is considered OPERA 3LE if it can be manually aligned (remote or local) in the shutdown cooling mode for removal of decay heat. In MODE 3, one RHR shutdown cooling subsystem can provide the required cooling, but two subsystems are required to be OPERABLE to provide ,

redundancy. Operation of one subsystem can maintain or I reduce the reactor coolant temperature as required.

However, to ensure adequate core flow to allow for accurate average reactor coolant temperature monitoring, nearly continuous operation is required. ,

g Note 1 permits both RHR shutdowr cooling subsystems and recirculation pumps tWE._ for a period of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in an 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period. Note 7 allows one RHR shutdown cooling subsystem to be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for performance of surveillance tests. These tests may be on the affected

~

RHR System or on some other plant system or component that necessitates placing the RHR System in an inoperabl~e status during the performance. This is permitted because the core l heat generation can be low enough and the heatup rate slow I enough to allow some changes to the RHR subsystems or other i

( operations requiring RHR flow interruption and loss of redundancy.

APPLICABILITY In MODES 1 and 2, and in MODE 3 with reactor steam dome pressure greater than or equal to the RHR cut in permissive pressure, this LCO is not applicable. Operation of the RHR System in the shutdown cooling mode is not allowed above this pressure because the RCS pressure may exceed the design i pressure of the shutdown cooling piping. Decay heat removal J at reactor pressures greater than or equal to the RHR cut in permissive pressure is typically accomplished by condensing the steam in the main condenser. Additionally, in MODE 2 l below this pressure, the OPERABILITY requirements for the i Emergency Core Cooling Systems (ECCS) (LCO 3.5.1, )

"ECCS-Operating") do not allow placing the RHR shutdown cooling subsystem into operation.

In MODE 3 with reactor steam dome pressure below the RHR cut in permissive pressure (i.e., the actual pressure at which the interlock resets) the RHR System may be operated in the l (continued)

GRAND GULF B 3.4-43 Revision No O

i Anachment 3 RHR Shutdown Cooling System-Cold Shutdown Page 54 - B 3.4.10 BASES .

I LC0 aligned (remote or local) in the shutdown cooling mode for 1 (continued) removal of decay heat. In MODE 4, one RHR shutdown cooling ;

subsystem can provide the required cooling, but two !

subsystems are required to be OPERABLE to provide l redundancy. Operation of one subsystem can maintain and !

reduce the reactor coolant temperature as required.

However, to ensure adequate core flow to allow for accurate average reactor coolant temperature monitoring, nearly continuous operation is required. f gg Note 1 permits both RHjt s_hutdow cooling subsystems and recirculation pumps tdW Wt %# for a period of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in an 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period. Note 2 allows one RHR shutdown cooling subsystem to be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for performance of surveillance tests. These tests may be on the affected RHR System or on some other plant system or component that necessitates placing the RHR System in an inoperable status during the performance. This is permitted because the core i heat generation can be low enough and the heatup rate slow enough to allow some changes to the RHR subsystems er-other operations requiring RHR flow interruption and loss _of _

redundancy.

6 9.gt-(

( Note 3 permits both RHR shutdown ooling subsystems and l recirculation pumps t846CHEFile@ during performance of inservice leak testing and during hydrostatic testing. This is permitted because RCS pressures and temperatures are being closely monitored as required by LC0 3.4.11. I i

APPLICABILITY In MODES 1 and 2, and in MODE 3 with reactor steam dome ;

pressure greater than or equal to the RHR cut in permissive i pressure, this LC0 is not applicable. Operation of the RHR System in the shutdown cooling mode is not allowed above this pressure because the RCS pressure may exceed the design pressure of the shutdown cooling piping. Decay heat removal ,

at reactor pressures greater than or equal to the RHR cut in 1 permissive pressure is typically accomplished by condensing the steam in the main condenser. Additionally, in MODE 2 below this pressure, the OPERABILITY requirements for the Emergency Core Cooling Systems (ECCS) (LCO 3.5.1, "ECCS-Operating") do not allow placing the RHR shutdown cooling subsystem into operation.

l (continued)

GRAND GULF B 3.4-48 Revision No. 0

m Primary Containment Air Locks Attachment 3 B 3.6.1.2 Page 55 BASES

[

SURVEILLANCE SR 3.6.1.2.2 (continued)

REQUIREMENTS following access through the air lock, which occurs L

regularly. The 7 day Frequency has been shown to be acceptable through operating experience and is considered adequate in view of the other indications available to operations personnel that the seal air flask pressure is -

low.

SR 3.6.1.2.3 The air lock interlock mechanism is designed to prevent simultaneous opening of both doors in the air lock. Since both the inner and outer doors of an air lock are designed to withstand the maximum expected post accident primary containment pressure (Ref. 3), closure of either door will support primary containment OPERABILITY. Thus, the *

. interlock feature supports primary containment OPERABILITY

~

while the air lock is being used for persorinel transit in and out of the containment. Periodic testing of this interlock demonstrates that the interlock will function as designed and that simultaneous inner and outer door opening

%$M will not inadvertently occur. Due to the nature of his 6 3,tf i$ interlock, and given that the interlock mechanism i hallenged when the primary containment air lock door s s

4 , this test is only required to be performed upon

-[ .--

ScB~ N (n..._:r ; e entite; :ot reauired7-b; r-m- ,t air.locli, but more frequently than once oer 18 i

e i p p,,- M" Q Frequency is Dased on engineering judgment and is !

consideFed adequatet .. ... vi v.... . _ . . . . . . . . . . -

% sed M e-t r! O o j

\ 6.7Jo-QC ,

nam &

SR 3.6.1.2.4 A seal pneumatic system test to ensure that pressure does not decay at a rate equivalent to > 2 psig for a period of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months from an initial pressure of 90 psig is an effective leakage rate test to verify system performance. The 18 month Frequency is based on the fact that operating experience has shown these components usually pass the ;

Surveillance when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the Frequency l was concluded to be acceptable from a reliability standpoint.

(continued) 4 GRAND GULF B 3.6-12 Revision No. O g'.

r '

Attachment 3 Page 56 insert B 3.6-12A used for entry and exit (procedures require strict adherence to single door opening)

Insert B 3.6-128 '

every 24 months. The 24 month Frequency is based on the need to perform this Surveillance under conditions that apply during a plant outage, and the potential for loss of primary

. containment OPERABILITY if the Surveillance was performed with the reactor at power. The 24 month frequency for the interlock is justified based on generic operating experience.

i insert B 3.6-12C given that the interlock is not challenged during the use of the airlock.

e

9 O G

1

F \

Attachment 3 PCIVs Pagc 57 l 8 3.6.1.3

)

( BASES SURVEILLANCE SR 3.6.1.3.1 (continued)

REQUIREMENTS purge valves are capable of closing in the environment following a LOCA. Therefore, these valves are allowed to be open for limited periods of time. The 31 day Frequency is consistent with other PCIV requirements. _

enA nM IockcM, Su=l<.S , ;

SR 3.6.1.3.2 " k* M MWM 1 This SR verifies that each primary containment isolation 1

manual valve and blind flange that is located outside primary containment, drywell, and steam tunnel, and is required to be closed during accident conditions, is closed. I The SR helps to ensure that post accident leakage of ,

radioactive fluids or gases outside of the primary '

containment boundary is within design limits. This SR does not require any testing or valve manipulation. Rather, it

~ involves verification that those devices outside primary containment, drywell, and steam tunnel, and capabfe 6f being mispositioned, are in the correct position. Since verification of valve position for devices outside primary containment is relatively easy, the 31 day Frequency was

( ~

chosen to provide added assurance that the devices are in the correct positions.

Two Notes are added to this SR. The first Note applies'to valves and blind flanges located in high radiation areas and allows them to be verified by use of administrative controls. Allowing verification by administrative controls is considered acceptable, since access to these areas is typically restricted during MODES 1, 2, and 3 for ALARA reasons. Therefore, the probability of misalignment of these devices, once they have been verified to be in the proper position, is low. A second Note is included to clarify that PCIVs open under administrative controls are !

not required to meet the SR during the time the PCIVs are open.

cm3 odI M ,5(d k SR 3.6.1.3.3 Y This SR verifies that each primary tontainment manual isolation valve and blind flange locatejd inside primary containment, drywell, or steam tunnel,fand required to be (

closed during accident conditions, is closed. The SR helps i l l (continued) l l

l GRAND GULF B 3.6-22 Revision No. O

AtlaChment 3 page 3g B 3.6.4.1 BASES ACTIONS C.I. C.2, and C.3 (continued)

Movement of irradiated fuel assemblies in the primary or secondary containment, CORE ALTERATIONS, and OPDRVs can be postulated to cause fission product. release to the secondary containment. In such cases, the secondary containment is the only barrier to release of fission products to the environment. CORE ALTERATIONS and movement of irradiated fuel assemblies must be immediately suspended if the secondary containment is inoperable.

Suspension of these activities shall not preclude completing an action that involves moving a component to a safe position. Also, action must be immediately initiated to suspend OPDRVs to minimize the probability of a vessel i draindown and subsequent potential for fission product i release. Actions must continue until OPDRVs are suspended.

Required Action C.1 has been modified by a Note stating that LCO 3.0.3 is not. applicable. If moving irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in

, MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.

SURVEILLANCE SR 3.6.4.1.1 and SR 3.6.4.1.2 in uchc e REQUIREMENTS one b '

Verifying that Auxiliary Building and Enclosure Building ,

equipment hatches, blowout panels, an access door ar '

closed ensures that the infiltration of outside air of such a magnitude as to prevent maintaining the desired negative ,

pressure does not occur. Verifying that all such openings '

are closed provides adequate assurance that exfiltration from the secondary containment will not occur. In this '

i application the term " sealed" has no connotation of leak tightness. Maintaining secondary containment OPERABILITY :

requires verityingp ash door in the access opening is !

closed, except whEn the access opening is being used for .

e entry and exit The 31 day Frequency for these SRs has been I ggg shown to De adequate based on operating experience, and is considered adequate in view of the other controls on p%; p}GA4KC secparycontainmentaccessopenings. ,

WQ W O W Y & GN qcCts o@N

-. ___ (continued) l GRAND GULF B 3.6-86 Revision No. 2 e

7_

i Attaclunent 3 AC Sources-Operating l Page 59 8 3.8.1 i BASES SURVEILLANCE SR 3.8.1.2 REQUIREMENTS (continued) This SR helps to ensure the availability of the standby electrical power supply to mitigate DBAs and transients and maintain the unit in a safe shutdown condition.

To minimize the wear on moving parts that do not get lubricated when the engine is not running, this SR is modified by a Note to indicate that all DG starts for this Surveillance may be preceded by an engine prelube period and followed by a warmup period prior to loading.

For the purposes of this testing, the DGs are started from standby conditions. Standby conditions for a DG mean that the diesel engine coolant and oil are being continuously circulated and temperature is being maintained consistent with manufacturer recommendations for DG 11 and DG 12. For DG 13, standby conditions mean that the lube oil is heated

- by the jacket water and continuously circulated.throu.gh a portion of the system as recommended by the vendor. Engine jacket water is heated by an immersion heater and circulates through the system by natural circulation.

SR 3.8.1.2 requires that the DG starts from standby conditions and achieves required voltage and frequency within 10 seconds. The DG's ability to maintain the required voltage and frequency is tested by those SRs which require DG loading. The 10 second start requirement r r the assumptions in the design basis LOCA analysis yj3 6 33-ISA A The DGs are started for this test by using one of the following signals: manual, simulated loss of offsite power by itself, simulated loss of offsite power in conjunction with an ESF actuation test signal, or an ESF actuation test signal by itself.

The 31 day Frequency for SR 3.8.1.2 is consistent with the industry guidelines for assessment of diesel generator performance (Ref. 14). This Frequency provides adequate assurance of OG OPERABILITY, while minimizing degradation resulting from testing.

(continued)

GRAND GULF B 3.8-15 LDC 98004

r-l

' Attachment 3 l- Page 60 l

L . Insert B 3.8-15A b

in addition to the SR requirements,~ the time for the DG to reach steady state operation is periodically monitored (data is taken at least once per 6 months during the performance of SR i 3.8.1.2) and the trend evaluated to identify degradation of govemor and voltage regulator performance.

e

(.

C .

i Diesel Fuel Oil, Lube Oil, and Starting Air Attachment 3 Pagc61 8 3.8.3 BASES 1

SURVEILLANCE SR 3.8.3.5 REQUIREMENTS !

I (continued) Microbiological fouling is a major cause of fuel oil degradation. There are numerous bacteria that can grow in fuel oil and cause fouling, but all must have a water environment in order to survive. Removal of water from the storage tanks once every 92 days eliminates the necessary environment for bacterial survival. This is the most effective means of controlling microbiological fouling. In addition, it eliminates the potential for water entrainment in the fuel oil during DG operation. Water may come from any of several sources, including condensation, ground water, rain water, contaminated fuel oil, and from breakdown of the fuel oil by bacteria. Frequent checking for and l removal of accumulated water minimizes fouling and provides data regarding the watertight integrity of the fuel oil system. The Surveillance Frequencies are established by Regulatory Guide 1.137 (Ref. 2). This SR is for preventive maintenance. The presence of water does not necessarily

' represent a failure of this SR provided that accumulated water is removed during performance of the Surveillance.

6R 3.8.3 h __

raining of the fuel oil stored in the supply tanks, remova o

mulated sediment, and tank cleaning are required at 10 year rvals by Regulatory Guide 1.137 (Ref. 2), )

paragraph 2. . is SR is typically performed in i conjunction with t ME Boiler and Pressure Vessel Code,Section XI (Ref. 7), exa ' tions of the tanks. To preclude the introduction of surfactan ; ,

n the fuel oil system, the l cleaning should be accomplished us - sodium hypochlorite solutions, or their equivalent, rather n soap or detergents. This SR is for preventive main ance. The presence of sediment does not necessarily repres a !

failure of this SR provided that accumulated sedimen d during performance of the Surveillance-REFERENCES 1. UFSAR, Section 9.5.4.

2. Regulatory Guide .1.137.

3. ANSI N195, Appendix B, 1976.

(continued)

GRAND GULF B 3.8-49 Revision No. I

Attaclunent 3 Diesel Fuel Oil, Lube Oil, and Staiting Air Page 62 8 3.8.3 BASES REFERENCES 4. UFSAR, Chapter 6.

(continued) .

5. UFSAR, Chapter 15.

6. ASTM Standards: D4057-88; D975-92a; 02274-70.

b Q. d!":,Boiies and F. 22u,=

'!was.1 Ced;, SecLv-c .

(' '

9 GRAND GULF '

B 3.8-50 l Revision No. 0

Auadunent 3 DC Sources-Operating Page 63 8 3.8.4 BASES- (continued) 4 SURVEILLANCE SR 3.8.4.1 REQUIREMENTS Verifying battery terminal voltage while on float charge helps to ensure the effectiveness of the charging system and I the ability of the batteries to perform their intended I function. Float charge is the condition in which the charger is supplying the continuous charge required to overcome the internal losses of a battery (or battery cell) and maintain the battery (or battery cell) in a fully charged state. The voltage requirements are based on the nominal design voltage of the battery and are consistent with the initial voltages assumed in the battery sizing calculations. The 7 day Frequency is consistent with manufacturer's recommendations and IEEE-450 (Ref. 8).

SR 3.8.4.2 Visual inspection to detect corrosion of the battery cells and connections, or measurement of the resistance of each inter-cell, inter-rack, inter-tier, and terminal connection, provides an indication of physical damage or abnormal deterioration that could potentially degrade battery

( performance.

The Surveillance Frequency for these inspections, which can detect conditions that can cause~ power losses due to resistance heating, is 92 days. This Frequency is considered acceptable based on operating experience related to detecting corrosion trends.

SR 3.8.4.3

' Visual inspection of the battery cells, cell plates, and

-g A battery racks provides an indication of physical damage or 6 35-%g abnormal deterioration that could potentially degrade q battery performance. 3 The 18 month Frequency of the Surveillance is based on engineering judgement, taking into consideration the desired unit conditions to perform the Surveillance. Operating experience has shown that these components usually pass the SR when performed at the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

(continued)

GRAND GULF ' B 3.8-56 Revision No. 0 i

Attachment 3 j Page 64

' Insert B 3.8-56A The presence of physical damage or deterioration does not necessarily represent a failure of this SR, provided an evaluation determines that the physical damage or deterioration does not affect the OPERABILITY of the battery (its ability to perform its design function).

9 e up ,

.48 i

l l

\,

RHR-High Water Level Attachment 3 B 3.9.8 Page 65 i

BASES APPLICABLE Although the RHR System does not meet a specific criterion SAFETY ANALYSES of the NRC Policy Statement, it was identified in the NRC !

(continued) Policy Statement as an important contributor to risk reduction. Therefore, the RHR System is retained as a Specification. The ADHRS is included in the Specification to provide requirements for decay heat removal capability during an outage while the RHR System is out of service.

LCO Only one RHR shutdowr, cooling subsystem is required to be OPERABLE in MODE 5 with irradiated fuel in the RPV and the I water level 2. 22 ft 8 inches above the RPV flange. Only one subsystem l's required because the volume of water above the RPV flange provides backup decay heat removal capability.

An OPERABLE RHR shutdown cooling subsystem consists of an RHR pump, two heat exchangers, valves, piping, instruments, and controls to ensure an OPERABLE flow path. The required RHR shutdown cooling subsystem must have a OPERABLE diesel generatoi capable of supplying electrical power.

Additionally, each RHR shutdown cooling subsystem is considered OPERABLE if it can be manually aligned (remote or local) in the shutdown cooling mode' for removal of decay heat. Operation (either continuous or intermittent) of one decay heat removal subsystem (either'RHR or ADHRS) can maintain and reduce the reactor coolant temperature as required. However, to ensure adequate core flow to allow

.for accurate average reactor coolant temperature monitoring,

%i 4 be nearly continuous equired. A Note is prn ided operatio@Glwe the operating

Igm to allow a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months exception w subsystem every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

APPLICABILITY One RHR shutdown cooling subsystem must'be OPERABLE in MODE 5, with irradiated fuel in the RPV and the water level 2 22 'ft 8 inches above.the top of the RPV flange, to provide decay heat removal. RHR System requirements in other MODES are covered by LCOs in Section 3.4, Reactor Coolant System (RCS); Section 3.5, Emergency Core Cooling Systems (ECCS) and Reactor Core Isolation Cooling (RCIC)

System; and Section 3.6, Containment Systems. RHR Shutdown Cooling System requirements in MODE 5, with the water level

< 22 ft 8 inches above the RPV flange, are given in LCO 3.9.9, " Residual Heat Removal (RHR)-Low Water Level."

(continued)

GRAND GULF B 3.9-26 Revision No. I 1

Attachment 3 RHR-Low Water Level Pagc 66 B 3.9.9 BASES I

i APPLICABLE Specification.

SAFETY ANALYSES The ADHRS is included in the Specification (continued) to provide requirements for decay heat removal capability during an outage while the RHR System is out of service.

LCO In MODE 5 with irradiated fuel in the reactor pressure vessel (RPV) and the water level < 22 ft 8 inches above the RPV flange both RHR shutdown cooling subsystems must be OPERABLE, subsystems.

or the ADHRS may be substituted for one of the RHR An OPERABLE RHR shutdown cooling subsystem consists of an RHR pump, two heat exchangers, valves, piping, instruments, and controls to ensure an OPERABLE flow path. An OPERABLE ;

ADHRS consists of two pumps, two heat exchangers, valves, ;

piping, instruments and controls to ensure an OPERABLE flow path.

At least one of the reqaired RHR shutdown cooling i subsystems must have a OPERABLE diesel generator capable of supplying electrical power.

Additionally, each RHR shutdown cooling subsystem is considered OPERABLE if it can be manually aligned (remote or heat. in the shutdown cooling mode for removal of decay local)

< Operation (either continuous ~or intermittent) of one i

decay heat removal subsystem (either RHR or ADHRS) can maintain and reduce the reactor coolant temperature as required. However, to ensure adequate core flow to allow BPM k 4* for accurate average reactor coolant temperature monitoring, "O nearly continuous operation is re uired. A Note is provided to allow a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months exceptio -the operating subsystegevery8 hours.

APPLICABILITf Two decay heat removal subsystems are required to be OPERABLE in MODE 5, with irradiated' fuel in the RPV and the water level < 22 ft 8 inches above the top of the RPV flange, to provide decay heat removal. RHR System i

requirements in other MODES are covered by LCOs in Section 3.4, Reactor Coolant System (RCS); Section 3.5, Emergency Core Cooling Systems (ECCS) and Reactor Core Isolation Cooling (RCIC) System; and Section 3.6, Containment Systems. RHR Shutdown Cooling System requirements in MODE 5, with the water level 2: 22 ft 8 inches above the RPV flange, are given in LCO 3.9.8,

" Residual Heat Removal (RHR)-High Water Level."

(continued)

GRAND GULF B 3.9-30 Revision No. 1

GNRO-99/00060 i

Attachment 4 Mark-up of Affected Technical Specifications and Bases with ;

Reference to TSTF Numbers l for l Grand Gulf Nuclear Station TS pages TSTF TS Bases TSTF )

Pages j

~ZO-1 5 B 2.0-5 5 2.0-2 5 B 2.0-6 5 3.0-2 104,166 B 2.0-8 5 l 3.1-1 9 8 2.0-9 5 3.1-4 9 B 2.0-10 5 3.1-7 32 B 3.0-6 104 3.1-8 32,33 B 3.1-14 32

{

3.1-12 32 B 3.1-15 32,33 1 3.4-16 60- B 3.1-23 32-3.4-17 60 B 3.4-34 60 3.4-21 153 B 3.4-35 60 3.4-24 153 B 3.4-36 60 l 3.6-8 17 B 3.4-43 153 3.6-14 45 B 3.4-48 153 3.6-15 45 B 3.6-12 17 3.6-43 18 B 3.6-22 45 3.8-5 163 B 3.5-86 18 3.8-10 163 8 3.8-15 163 3.8-13 163 B 3.8-49 2 3.8-16 163 B 3.8-50 2 3.8-25 2 B 3.8-56 38 T B~28 38 B 3.9-26 153 3.8-34 278 8 3.9-30 153 <

3.8-35 278 3.9-10 153 3.9-12 153 5.0-11 279 l 5.0-14 118 5.0-18 9 I

SLs Attachment 4 Page 2 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow < 10% rated core flow:

THERMAL POWER shall -be s 25% RTP.

2.1.1.2 With the reactor steam dome pressure a 785 psig and core I flow a 10% rated core flow:

MCPR shall be a 1.11 for two recirculation loop operation or a 1.12 for single recirculation loop operation.- l 2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel, i 2.1.2 Reactor Coolant System Pressure SL

\

Reactor steam dome pressure shall be s 1325 psig.

l 2.2 SL Violations , wMW 2 With any SL violation, the following actions shall be completed:

has '

o S.2.1 Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , notify 77, the NRC Operations Center, in accordance) i i .4 + h i. n. f.' r. D C h.

i

'2.2.2 Within 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months s:

2.2% 1 Restore compliance with all SLs; and

[ k i

2. 2 Insert all insertable control rods.

. ,; "4 + h h 24 hmere.

nntifv'theplantmanagerandthecorporap executive responsible for overall plant nucie:4. s o i c s.7 . ,

MCPR values in T.S. 2.1.1.2 are applicable only for cycle 10 operation. l (continued) l GRAND GULF 2.0-1 Amendment No. 42OrM1,136

E 1 1-i l Attaclunent 4 gg3 Page 3 2.0

(

I 2.0 SLs SL Violations (contin _ TF - E 'CR ,

2.2.4 30 days, a Licensee Event Report (LER) shall be prepared l

pursuant 0.73. The LER shall be submitted to the NRC, the plant manager an orate executive responsible for ,

overall plant nuclear safety.

2.2.5 Operation of the unit shall not be resumed until au dbythe)

NR V --

~

Q l

1 I

i i

GRAND GULF ,

2.0-2 Amendment No. 120

c Attachment 4 LCO Applicability Page 4 3.0

3.0 LCO APPLICABILITY LC0 3.0.'4 specified conditions in the Applicability that are required

-(continued) to comply with ACTIONS or that are part of a shutdown of the unit.

Exceptions to this Specification are stated in the l

' individual Specifications.fThese s allow entry i

@te unn 5 or other specified conditions i Applicabi he associated ACTIONS to be ent

'5T- -

allow unit operation in

'cI5nHii. ivi, k %sJpplicability only or other specified d naciod of I (04 ~ di"*:/

I LCO 3.0.4 is only applicable for entry into a MODE or other b I I

specified condition in the Applicability in MODES 1, 2, and

. 3. ,

LCO 3.0.5 Equipment removed from service or declared inoperable to l comply with ACTIONS may be returned to service under j administrative control solely to perform testing required to i

( demonstrate its OPERABILITY or the OPERABILITY of other !

equipment. This is an exception to LCO 3.0.2 for the system l

returned to service under administrative control to perform the testing required to demonstrate OPERABILITY.

LCO 3.0.6 When a supported system LCO is not met solely due to a support system LCO not being met, the Conditions and -

Required Actions associated with this supported system are i not required to be entered. Only the support system LCO ACTIONS are required to be entered. This is an exception to In this event, W// l n 4 _LCO 3.0.2 for the sup

__ mm,-evaluation rted 6,v . system.

u. . c =.x :n r x ri =- i n be- '

accordance with Specification 5.5.10,'" Safety function pebd Determination Program (SFDP)." If a loss of safety function !

/ is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

When a support system's Required Action directs a supported system to be declared inoperable or directs entry into ,

Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LCO 3.0.2.

(continued)

GRAND GULF c 3.0-2 Amendment No. 120 !

F :

l Attaciuncut 4 j Page 5 SDM I 3.1.1 i.

3.1 REACTIVITY CONTROL SYSTEMS l 3.1.1 SHUTDOWN MARGIN (SDM)

GTF7-C3 :

i wi4 W % .ilm:45 peow ygo i LCO 3.1.1 SDM shalig 6 M C.0 L R.

AC {

a. .a 6. E ^6/k. with the highest worth control rod !

analytically de e -

r

b. a: . with the highest worth co dj !

determined by y j APPLICABILITY: MODES 1, 2, 3, 4, and 5.

ACTIONS . j

~

CONDITION REQUIRED ACTION COMPLETI0N' TIME A. SDM not within limits A.1 Restore SDM to within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months b- in MODE 1 or 2. limits.

B. ' Required Action and 8.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A not met.

I C. SDM not within limits C.1 Initiate action to Immediately in MODE 3. fully insert all insertable control rods.

I D. SDM not within limits D.1 Initiate action to Immediately in MODE 4. fully insert all insertable control I rods. l AND

, (continued) i GRAND GULF 3.1-1 Amendment No. 120

Attachment 4 SDM Page 6 3.1.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.1.1 Verify SDM s: Prior to each in vessel fuel f.~ - 2: 0.38% thehighestworM movement during I

control rod analytica ermined; fuel loading or sequence b /k.with the hig cQh AND k.

control ro ined by te h >

Once within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after is be w (bl )lm'4 y) criticality following fuel Tf I

I--h movement within the reactor .

pressure vessel or control rod replacement c

GRAND GULF 3.1-4 Amendment No. 120

p 3 Attaclunent 4 Page 7 Control Rod OPERAB!LlTY 3.1.3 3.1 REACTIVITY CONTROL SYSTEMS 3.1.3 Control Rod OPERABILITY LCO 3.1.3 Each control rod shall be OPERABLE.

APPLICABILITY: MODES 1 and 2.

ACTIONS

...........................----------NOTE----------------------------..------.

. Separate Condition entry is allowed for each control rod. ;

.............................................................................. 4 CONDITION REQUIRED ACTION COMPLETION TIME A. One withdrawn control ------------NOTE-------------

rod stuck. A stuck rod may be bypassed

(,. in the Rod Action Control Systera (RACS) in accordance with SR 3.3.2.1.9 if required to allow continued operation.

h Disarm the associated2 hours control rod drive (CRD).

AND (continued) dTn:5e.d

' 3.1 ~7 A i -

S.F32-Clo l

i GRAND GULF ,. 3.1-7 Amendment No. 120

r.

- Attachment 4 insert 31-7A' 4

A.1 - Verify stuck control rod immediately separation criteria are met.

,. l j

^*

TSTF 32- C/a I

, - == e

(:. l

\

i i

l I I l

l t;- ;

\ l 1

i i

L.

I Attachment 4 Pagc 9 Control Rod OPERABIL!TY ,

3.1.3 ]

\

ACTIONS i

CONDITION- REQUIRED ACTION COMPLETION TIME A. '(continued) A. h Perform SR 3.1.3.2 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from and SR 3.1.3.3 for discovery of each withdrawn THERMAL POWER OPERABLE control rod.p* greater than the low power T9T DCW wconcorr de m& .

3*daio!LPSP)

Pattern Control System (RPCS)

A. Perform SR 3.1.1.1. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months l B. Two or more withdrawn B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months control rods stuck.

('

C. One or more control --------NOTE---------

rods inoperable for Inoperable control reasons other than rods may be bypassed Condition A or B. in RACS in accordance with SR 3.3.2.1.9, if required, to allow insertion of inoperable control rod and continued operation.

C.1 Fully insert 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months inoperable control rod.

AND C.2 Disarm the associated 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months CRD.

(continued)

' I GRAND GULF 3.1-8 Amendment No. 120

Attachment 4 Page lo Control Rod Scram Times 3.1.4 l 3.1 REACTIVITY CONTROL SYSTEMS 3.1.4 Control Rod Scram Times i

LCO 3.1.4 a. No more than 14 OPERABLE control rods shall be " slow,"

in accordance with Table 3.1.4-1; and

b. No OPERABLE control rod that is " slow" shall occupy a !

location adjacent to another OPERABLE control rod that !

is " slow"6F e withdd r. eer.treT r~J 15 d@.L T9TF02-C6 APPLICABILITY: MODES 1 and 2. (SR tJgf4cr 1934

! ) OWfh CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months LCO not met.

5.. ,

==

GRAND GULF 3.1-12 Amendment No. 120

Attachment 4 Page 11

'RCS Leakage Detection Instrumentation

. 3.4.7 e

3.4 REACTOR COOLANT SYSTEM (RCS) .

3.4.7 RCS Leakage Detection Instrumer.tation LCO 3.4.7 The following RCS leakage detection instrumentation shall be OPERABLE:

a. Drywell floor drain sump monitoring system;

b. One channel of either drywell atmospheric particulate or atmospher.ic gaseous monitoring system; and

c. Drywell air cooler condensate flow rate monitoring system.

l i

, APPLICABILITY: MODES 1, 2, and 3.

' ~ ~ ~

ACTIONS _ T5TT-C.o - C / o

' CONDITION REQUIRED ACTION COMPLETION TIME A. Drywell floor drain -------------NOTE------------

sump monitoring system , LCO 3.0.4 is not applicable.

inoperable. (_----------------------------

A.1 Restore drywell floor drain sump monitoring 30 days system to OPERABLE status.

B. Required drywell B.1 Analyze grab samples Once per atmospheric monitoring of drywell 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months system inoperable. atmosphere.

(continued)"

GRAND GULF 3.4-16 Amendment No. 120

Attachment 4 i Page 12 RCS Leakage Detection instrumentation 3.4.7 -

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Drywell air cooler -------------NOTE------------

condensate flow rate Not applicable when the monitoring system required drywell atmospheric inoperable. monitoring system is inoperable. l C.1 Perform SR 3.4.7.1.

Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 1 D. Required drywell ------- ---NOTE------ ----

atmospheric monitoring system inoperable.

L licab l I W 6VC/O Q - _ -----_-----_-- ]

AND D.1 Restore required -. I drywell atmospheric 30 days l Drywell air cooler monitoring system to condensate flow rate OPERABLE status. r

(' monitoring system f' inoperable. OR D.2 Restore drywell air cooler condensate 30 days flow rate monitoring system to OPERABLE status.

E. Required Action and E.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A, AND B, C, or D not met.

E.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

F. All required leakage F.1 Enter LC0 3 0.3. ~

Immediately detection systems inoperable.

GRAND GULF ,

3.4-17 Amendment No. 120

+

Attachment 4 P11ge l3 RHR Shutdown ' Cooling- System--Hot Shutdown 3.4.9 3.4 REACTORCOOLANTSYSTEM(RCS) 3.4.9 Residual Heat Removal (RHR) Shutdown Cooling System--Hot Shutdown LC0 3.4.9 Two RHR shutdown cooling subsystems'shall be OPERABLE, and, with no recirculation pump in operation, at least one RHR

-shutdown cooling subsystem shall be in operation.

....................--------N0TES----...----------.....-....

1. Both RHR shutdown cooling subsystems and recirculation -

pumps ma,q: r;.T.sset byoper tion for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period. ^,4 b' '*

% Kl$P(h 2. -One RHR shutdown cooling subsystem may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for performance of Surve111ances.

. APPLICABILITY: MODE 3 with reactor steam dome pressure less than the.RHR cut in permissive pressure.

ACTIONS

.(

....................................-NOTES------------------------------------

1. LCO 3.0.4 is not applicable.

2. Separate Condition entry is allowed for each RHR shutdown cooling subsystem.-

CONDITION REQUIRED ACTION COMPLETION TIME A. One or two RHR A.1 Initiate action to Immediately shutdown cooling restore RHR shutdown subsystems inoperable. cooling subsystem (s) to OPERABLE status.

AND (continued)

GRAND GULF ,

, 3.4-21 Amendment No. 120

Attachment 4 Page 14 i

RHR Shutdown Cooling System-Cold Shutdown

, 3.4.10

(

3.4 REACTOR COOLANT SYSTEM (RCS)

)

3.4.10 Residual Heat Removal (RHR) Shutdown Cooling System-Cold Shutdown LCO 3.4.10 Two RHR shutdown cooling subsystems shall be OPERABLE, and, with no recirculation pump in operation, at least one RHR shutdown cooling subsystem shall be in operation.

..-----------------NOTES---------------------------

1. subsystems and recirculation BothRHRshutdowncoog:poperationforupto2 pumps ma . reg.e . g . r hours gi :- l53-CB2.

M One RHR shutdown cooling subsystem may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for the performance of Surveillances.

3. Both RHR shutdown cooling subsystems and recirculation pum)s ma K re:ft:r ffed eration during RCS inservice

~

......!!..$$ .. $'.. .$. $$..$ *.. D

~~

APPLICABILITY: MODE 4.

ACTIONS

....................................-NOTE-------------------------------------

Separate Condition entry is allowed for each RHR shutdown cooling subsystem.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or two RHR A.1 Verify an alternate 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months shutdown cooling method of decay heat subsystems inoperable. removal is available AND for each inoperable RHR shutdown cooling Once per subsystem. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter (continued)

GRAND GULF 3.4-24 Amendment No. 120

Attacliment 4 se 15 Primary Containment Air Locks 3.6.1.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.1.2.3 --------------NOTE---

3 y g7, h Only re to be performed up ry l #

r exit throug rimary containmen f

..h qfn Verify only one door in the primary ca; {4 containment air lock can be opened at a time, i

SR 3.6.1.2.4 Verify, from an initial pressure of 18 months l 90 psig, the primary containment air lock seal pneumatic system pressure does not decay at a rate equivalent to . ..

> 2 psig for a period of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

( ;

)

O GRAND GULF 3.6-8 Amendment No. 120

E Auachment 4 PCIVs Page 16 3.6.1.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY l

}

SR 3.6.1.3.1 ------------------NOTES------------------ j

1. Only required to be met in MODES 1, )

-2, and 3.

t

2. Not required to be met when the )

20 inch primary containment purge valves are open for pressure control, ALARA, or air quality considerations for personnel entry. Also, not required to be met during Surveillances or special testing on the purge system that requires the valves to be open. The 20 inch primary containment purge valves shall not be open with the 6 inch primary containment purge or the drywell vent and purge supply and exhaust lines open. * ~-

..,.- Verify each 20 inch primary containment 31 days

( purge valve is closed.

SR 3.6.1.3.2 ------------------NOTES------------------ 3

1. Valves and blind flanges in high l radiation areas may be verified by j use of administrative means. ;

2. Not required to be met for PCIVs that are open under administrative controls.

......................................... l Verify each primary containment isolation 31 days manual valve and blind flange that is located outside primary containment, drywell, and steam tunnelg and is required to be closed during accident conditions is closed.

an ook loc.h O 9E,\,.c(

i (continued) i g d be-w h 3(c e d GRAND GULF . 3.6-14 TM C7 Amendment No. 120 c

l

Attaciunent 4 PCTVs Page 17 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY l SR 3.6.1.3.3 ------------------NOTES------------------

1. Valves and blind flanges in high ;

radiation areas may be verified by '

use of administrative means. 1

2. Not required to be met for PCIVs that are open under administrative controls. i Verify each primary containment isolation Prior to manual valve and blind flange that is entering MODE 2 located inside primary containment, or 3 from drywell, or steam tunnel and is required MODE 4, if not to be closed during acci en ) conditions performed is closed. L _

within the L

. R f'}-(g a g g [9 {h g g previous 92 days -

l i

e4%=v why. 53 c-otS, i

f. SR 3.6.1.3.4 Verify the isolation time of each power In accordance operated and each automatic PCIV, except with the MSIVs, is within limits. Inservice Testing Program (continued) l GRAND GULF 3.6-15 Amendment No. 424 128 l c -

r Secondary Containment

- Attachment 4 Page 18 3.6.4.1

, ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C.. Secondary- C.1 --------NOTE---------

containment LCO 3.0.3 is not inoperable during- applicable, movement of irradiated ---------------------

fuel assemblies in the primary or secondary Suspend movement of Immediately containment, during irradiated fuel CORE ALTERATIONS, or -assemblies in the during OPDRVs. primary and secondary containment.

AND C.2 Suspend CORE Immediately ALTERATIONS.

AND C.3 Initiate action to Immedi'ately ~

. suspend OPDRVs.

I SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.I.1 Verify all auxiliary building and 31 days enclosure building equipment hatches and blowout panels are closed and sealed.

SR 3.6.4.1.2 Verif auxiliary building and 31 days enclosure building access door is closed, '

exceptwhentheaccessopeninglisbein - l used for entry and exit. g4ce.h acess {

f oWW '

i I

(continued) t T5TF-Ig- c5 1 GRAND GULF ' 3.6-43 Amendment No. 120 1

,. I

r Attachment 4 AC Sources-Operating Page 19 3.8.1 ACTIONS (continued)

CONDITION REQUIRED ACTION

, COMPLETION TIME H. Three or more required H.1 Enter LCO 3.0.3. Immediately AC sources inoperable.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.1.1 Verify correct breaker alignment and 7 days indicated power availability for each required offsite circuit.

SR 3.8.1.2 -------------------NOTE--------------------

All DG starts may be preceded by an engine

( prelube period and followed by a warmup period prior to loading.

V 31 days k c_erifyonditions each DG starts and from standbv achievesfTh s 10 seconds,

> 3744 V ana s 4576 V and frequency 2: 58.8 Hz and s 01.2 a.

N N-gmc4 (continued)

\ 3.5 -S A ;

\ l l

3 l f "I 0 -l l

l l

GRAND GULF 3.8-5 Amendment No. 440, 134 l

Attachmcat 4 Page 20

. Insert 3.8-5A

.( .

s. . In s 10 seconds, voltage 2 3744 V and frequency 2 58.8 Hz ; and

b. steady state voltage 23744 V and s 4576 V and frequency 2 58.8 Hz and s 61.2 Hz.

%TF S3-CJ4

, - an. e

(. ..

f I

l I

Attachment 4 Page 21 AC Sources-Operating 3.8.1 A

SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.12 -------------------NOTES-------------------

1. All DG starts may be preceded by an engine prelube period.

2. This Surveillance shall not be performed in MODE 1 or 2. However, 4 credit may be taken.for unplanned l events that satisfy this SR. l Verify on an actual or simulated Emergency 18 months Core Cooling System (ECCS) initiation signal each DG auto-starts from standby 1 condition and- l i

a. In s 10 seconds after auto-start and l during t sts, achiev_es voltage Tnnd 2 3744 w,; a $70 3 .s;5-Io A

b. n = Au w w,-h 2fte rt s t r' r2 Ju . . ..y t::t:,f%chieveQequency F

(. a 58.8 Hz and s 61.2 Hz; In 3.15 -to O \

c. Operates for a 5 minutes; and

d. Emergency loads are auto-connected to the offsite power system.

l (continued)

TST: - (,3- CS I

1 9

i GRAND GULF 3.8-10 Amendment No. 120 t*

l 1

I" l:

L.~

Attachment 4 Page 22 Insert 3.8-10A frequency 2 58.8 Hz Insert 3.8-108 steady state voltage 23744 V and s 4576 V and T5TF -163 - CR

'O 4

9 e

e aan e 9

5 t

Attachment 4 Pagc 23 AC Sources-Operating i 3.8.1 1

SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY l

l l SR 3.8.1.15 ------------------NOTES-------------------

l 1. This Surveillance shall be performed within 5 minutes of shutting down the i DG after the DG has operated a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or until operating temperatures i stabilized loaded a 5450 kW and I s 5740 kW for DG 11 and DG 12, and 2 3300 kW for DG 13.

Momentary transients outside of the load range do not invalidate this test.

2. All DG starts may be preceded by an engine prelube period.

I

~105'34 1 g - l Q -C l 3 .T' O ( I Verify each DG starts and achieve , in 18 months -

=,10 unah, xaltac e 2 3744 y and s 4576 V Qand frequency a 58.0 ttz and s 61.2 rky y

(,

SR 3.8.1.16 -------------------NOTE--------------------

This Surveillance shall not be performed in MODE 1, 2, or 3. However, credit may be taken for unplanned events that satisfy this SR.

Verify each DG: 18 months

a. Synchronizes with offsite power source while loaded with emergency loads upon a simulated restoration of offsite power;

b. Transfers loads to offsite power source; and

c. Returns to ready-to-load operation.

(continued) 6 GRAND GULF 3.8-13 Amendment No. 120 e

Attachment 4 Insert 3.8-13A

- a. In s 10 seconds, voltage 2 3744 V and frequency 2 58.8 Hz ; and

b. steady state voltage 23744 V and s 4576 V and frequency 2 58.8 Hz 'and s 61.2 Hz.

T5T -/l,3-c/4

. g e

9 9

?.

Attaciunent 4 AC Sources-Operating Page25 3,8,1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.20 -------------------NOTE--------------------

All DG starts may be preceded by an engine prelube period.

Verify, when started simultaneously from 10 years A -standby condition. each DG achieve , in 60r.seconhvoltagey 3/94 y and s 4576 V land MnJtency :a: 58.8 Hz and s 61.4 rit._~

T 3

-~~Tn5Lt-Y 31~ Ib b j -

-/ ]-(/

I.

t.

GRAND GULF 3.8-16 Amendment No. 120 g'.

r Attachment 4 l Page 2(>

- Insert 3.8-16A 1

I

. a. In s 10 seconds, voltage 2 3744 V and frequency 2 58.8 Hz ; and

b. ' steady state voltage 23744 V and s 4576 V and frequency 2 58.8 Hz and s 61.2 Hz.

~fb~b e

4 aus e t

l 1

i

~

Attachment 4 -

Page 27 Diesel Fuel Oil, Lube Oil, and Starting Air 3.8.3

, SURVEILLANCE REQUIREMENTS (continued) l fl ~ Sk /

SURVEILLANCE FREQUENCY

.8.3.6 For each fuel oil storage tank: 10 years

] a. Drain the fuel oi ;

b. Remove the sediment; and

c. Clean the tank.

O (e I I

l l

i I

-t GRAND GULF 3.8-25 Amendment No. 120

'4 At ajn' DC Sources-Oper ng 8

SURVEILLANCE REQUIREMENTS (continued) T WF -W- cg ,

SURVEILLANCE FREQUENCY SR 3.8.4.3 Verify battery cells, cell plates, and 18 months racks show no visual indication of physical damage or abnormal deterioratiog 6s__M too\J b e o A EM5ecu m b7[m,M w

SR 3.8.4.4 Remove visible corrosion and verify battery 18 months cell to cell and terminal connections are coated with anti-corrosion material.

SR 3.8.4.5 Verify battery connection resistance is 18 months s 1.5 E-4 ohm for inter-cell connections, s 1.5 E-4 ohm for inter-rack connections, s 1.5 E-4 ohm for inter-tier connections, and s 1.5 E-4 ohm for terminal connections. .. ,

)

(i SR 3.8.4.6 Verify each Division 1 and 2 required 18 months battery charger supplies a 400 amps at 2125 V for a 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months ; and the Division 3 battery charger supplies a 50 amps at ;

a 125 V for a 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . '

(continued) i GRAND GULF ' 3.8-28 Amendment No. 120

Battery Cell Parameters

^{l[unenu 3.8.6 3.8 ELECTRICAL POWER SYSTEMS 3.8.6 Battery Cell Parameters LCO 3.8.6 Battery cell parameters for the Divi ion 1, 2. and 3 batteries shall be withi@ limits . Tobic 3.0.5y.

T57F-2 % cg APPLICABILITY: When associated DC electrical power subsystems are required to be OPERABLE.

ACTIONS

...............................------NOTE------------------------------------- i Separate Condition entry is allowed for each battery.

. CONDITION REQUIRED ACTION COMPLETION. TIME A. One or more batteries A.1 Verify pilot cell's I hour

(, with one or more electrolyte level and battery cell float voltage meet parameters not within Table 3.8.6-1 Table 3.8.6-1 Category Category C limits.

A or B limits.

AND A.2 Verify bat'tery cell 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months parameters meet Table 3.8.6-1 AND Category C limits.

Once per 7 days thereafter AND A.3 Restore battery cell 31 days parameters to L3. 8. 63 J$y -)) _{ / ( (continued)

GRAND GULF 3.8-34 Amendment No. 120 y'.

Attachment 4 Battery Cell Parameters Page 30 3.8.6

, ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Declare associated Immediately associated Completion battery inoperable.

Time of Condition A not met.

0.R One or more batteries with average electrolyte temperature of the representative cells

< 60*F.

01

. One or more batteries _,

with one or more battery cell parameters not within T9T:mg - a6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.6.1 Verify battery cell parameters meet 7 days Table 3.8.6-1 Category A limits.

(continued)

GRAND GULF 3.8-35 Amendment No. 120 e

P f

""4 p,'gI 1 RHR-High Water Level 3.9.8

~3.9 REFUELING OPERATIONS 3.9.8 Residual Heat Removal (RHR)-High Water Level LC0 3.9.8 One RHR shutdown cooling subsystem shall be OPERABLE and one decay heat removal subsystem shall be in operation.

............................N0TE---------............-......

The required decay heat removal subsystem may

.M G operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months pe$e ,um,J g %

Co[ct i

riod.

TS7F-19? cl3 .

APPLICABILITY: . MODE 5 with irradiated fuel in the reactor pressure vessel (RPV) and the water level a: 22 ft 8 inches above the top of the RPV flange.

ACTIONS ]

CONDITION REQUIRED ACTION COMPLETiONTIME ;

i

. 1

(

A. Required RHR shutdown A.1 Verify an alternate 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months !

cooling subsystem method of decay heat )

inoperable. removal is available. AND l

l Once per {

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months j thereafter l B. Required Action and B.1 Suspend loading Immediately associated Completion irradiated fuel Time of Condition A assemblies into the not met.~ RPV. !

AND l B.2 Initiate action to Immediately restore secondary containment to OPERABLE status. 4 AND i (continued)

GRAND GULF 3.9-10 Amendment No. 120

l Attachment 4 RHR-Low Sater Level Page 32 3.9.9 3.9 REFUELING OPERATIONS 3.9.9 Residual Heat Removal (RHR)-Low Water Level LCO 3.9.9 Two decay heat removal subsystems shall be OPERABLE, and one decay heat removal subsystem shall be in operation.

........................----N0TE--------.-.------- ------.--

operating decay heat removal subsystem may<5ite-00, O 2 e ='!c_ .r u operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period, i ra ............................................__..............

TSTP-l 53-Cl 3 i APPLICABILITY: MODE 5 with irradiated fuel in the reactor pressure vessel (RPV) and .the water level < 22 ft 8 inches above the top of the RPV flange.

ACTIONS

...............................------.---NOTE-------------.------------------

Separate Condition entry is allowed for each decay heat removal subsystem.

4

( ~

CONDITION REQUIRED ACTION COMPLETION TIME i

A. One or two required A.1 Verify an alternate I hour decay heat removal method of decay heat subsystems inoperable. removal is available AND for each inoperable required decay heat Once per removal subsystem. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter (cortinued)

' GRAND GULF 3.9-12 Amendment No. 120 t'

Attaciunent 4 Pagc 33 Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.5 Component Cyclic or Transient Limit This program provides controls to track the cyclic and transient occurrences identified on UFSAR Table 3.9-35 to ensure that the reactor vessel is maintained within the design limits.

5.5.6 Inservice Testino Program 7p g .]

This program provides controls for inservicttesting of ASME Code C1 ass 1, 2, and 3 componenlfff=M'"9 =0 i :bli:0;=(t3. The program shall include the following:

a. Testing frequencies specified in Section XI of the ASME Boiler and Pressure Vessel Code and' applicable Addenda as follows:

ASME Boiler and Pressure Vessel Code and - --

applicable Addenda terminology for Required Frequencies inservice testing for performing inservice

{ activities testino activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every At least once per 92 days 3 months Semiannually or At least once per 184 days every 6 months Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every At least once per 731 days 2 years

b. The provisions of SR 3.0.2 are applicable to the above required frequencies for performing inservice testing activities;

c. The provisions of SR 3.0.3 are applicable to inservice testing activities; and

d. Nothing in the ASME Boiler and Pressure Vessel Code shall be i construed to supersede the requirements of any TS.

I 4

(continued)

GRAND GULF 5.0-11 Amendment No. 120 c \

Attachment 4 Programs and Manuals Page 34 5.5 5.5 Programs and Manuals 5.5.8 Explosive Gas and Storage Tank Radioactivity Monitorino Program (continued)

b. A surveillance program to ensure that the quantity of radioactive material contained in any outside temporary tank i not including liners or shipping radwaste is 110 curies, !

excluding tritium and dissolved or entrained noble gases.

l The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies.

5.5.9 Diesel Fuel Oil Testina Proaram A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM-Standards. The purpose of the program is to establish the following:

a. Acceptability of new fuel oil for use prior to addition to

( storage tanks, and acceptability of stored fuel oil every 92 days, by determining that the fuel oil has:

1. a water and sediment contents within limits, and

2. a kinematic viscosity within limits for ASTM 20 fuel oil;

b. Total particulate concentration of the new fuel is s 2 mg/100 ml when tested in accordance with ASTM D-2274-70 within 7 days after addition of the new fuel to the storage tank; and

[$$-y Kgc. Total particulate concentration of the fuel oil in the storage tanks is 1 2 mg/100 ml when tested every 92 days in I

gg q accordance with ASTM D-2274-70.

K0 - 14 A 5.5.10 Safety Function Determination Program (SFDP)

This program ensures loss of safety function is detected and appropriate actions taken. Upon entry into LC0 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other appropriate limitations and remedial (continued)

GRAND GULF 5.0-14 Amendment No. 120

0 !

Attachment 4 Page 35 Insert 5.0-14A The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program testing frequencies.

TSTT~ ~ l18'- G J2, a

l I

l l

(

1 l

l l ,

j l

i.* 1 i

i

l l

Attachment 4 Reporting Requirements Page 36 5.6 5.6 Reporting Requirements 5.6.2 Annual Radiolooical Environmental Operatino Report (continued) results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible.

5.6.3 Radioactive Effluent Release Report l The Radioactive Effluent Release Report covering the operation of the unit during the previous calendar year shall be submitted by May 1 of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCM and process control program and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B.1.

5.6.4 Monthly Operatina Reports Routine reports of operating statistics and shutdown experience,

( including documentation of all challenges to the main steam :

safety / relief valves, shall be submitted on a monthly basis no later than the 15th of each month following the calendar month l covered by the report.

5.6.5 Core Operatino limits Report (COLR)

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

1) LCO 3.2.1, Average Planar Linear Heat Generation Rate (APLHGR),

2) LCO 3.2.2, Minimum Critical Power Ratio (MCPR), an

3) LC0 3.2.3, Linear Heat Generation Rate (LHGR f (continued) q)Lcod.).1, SW%em rv%.egl m (50 rn *

/ Gei, GRAND GULF

. N ~9) 5.0-18

/ AMM6-llGV

\ ou u sm 2,8/Ah1 Amendment No. 120 e

r-Attachment 4 Reactor Core SLs Page 37 B 2.1.1 BASES (continued)

SAFETY LIMIT k. 2.

$[ff-CQ If We-N erations Center must be

Q notified within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , in a 72 9

-2.2.2 Execedir.g an SL may cause fuel damage and create a potential for radioactive releases in excess of 10 CFR 100, " Reactor Site Criteria," limits (Ref. 4). Therefore, it is required to insert all insertable control rods and restore compliance with the SL within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . (The required actions for a violation of the reactor water level SL include manually initiating ECCS to restore water level and depressurizing 1 the reactor vessel, if necessary, for ECCS operation.) The l 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time ensures that the operators take l

- prompt remedial action and also ensures that the probability 1 of an accident occurring during this period is' minimal. '

h 75T'F-9Q, any SL is violated, e General Manager, P t Operatio an he Vice President, Op tions GGNS shall be tified within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months riod provides time lant operators d staff to take the opriate immediate on and assess t condition of the uni senior managemen efore reporting to hg l i

l2.2.4 any SL is violated, a Licenbrent Report shall be i prep and submitted within 20 da accordanc ith 10 CFR 50.73 (Ref.

to the NRC in 5). The report will Ns'N describe the icable circumstances pre ing the violation, the ef of the violation upon t components, systems, or structures, d the corrective acti taken to prevent recurrence. A copy the report shall al be submitted to the General Manage , lant Operations an he ent, Operations GGNS.

(continued)

GRAND GULF B 2.0-5 Revision No. 1 l

l .

r Anachment 4 Reactor Core SLs Page 38 B 2.1.I BASES i

ATI (continued) I a violated, restart of the unit shall not commence unti tn4zad. the NRC. This requirement T w@s g es the NRC that all necessa action N completed before the unit beg s, ns analyses, and tart to i

Qal operati y REFERENCES 1. 10 CFR $0, Appendix A, GDC lu.

2. XN-NF524(A), Revision 2, April 1989.

F I3 4.

0 CTR 50.7h : "

D TSVFC 10 CFR 100.

45 f", 50.ID

~

6. NEDE-240ll-P-A, GESTAR-II.

l

7. GNRI-96/00230, Amendment 131 to the Operating License.

l

[:,

t GRAND GULF B 2.0-6 Revision No. 2

Attachment 4 Page 39 RCS Pressure SL f B 2.1.2 l

. BASES I APPLICABLE The RCS pressure SL has been selected such that it is at a SAFETY ANALYSES pressure below which it can be shown that the integrity of (continued) the system is not endangered. The reactor pressure vessel is designed to ASME, Boiler and Pressure Vessel Code,Section III, 1971 Edition, including Addenda through the winter of 1972 (Ref. 5), which permits a maximum pressure transient of 110%, 1375 psig, of design pressure 1250 psig.

The.SL of 1325 psig, as measured in the reactor steam dome, is equivalent to 1375 psig at the lowest elevation of the RCS. The RCS is designed to ASME Code,Section III, 1974 Edition (Ref. 6), for the reactor recirculation piping, which permits a maximum pressure transient of 110% of design pressures of 1250 psig for suction piping,1650 psig for discharge piping between t'ie pump and the discharge valve, and 1550 psig beyond the discharge valve. The RCS pressure SL is selected to be the lowest transient overpressure allowed by the applicable codes.

SAFETY LIMITS The maximum transient pressure allowable in the RCS pressure vessel under the ASME Code,Section III, is 110% of design l pressure. The maximum transient pressure allowable in the

RCS piping, valves, and fittings is 110% of design pressures .._..

of 1250 psig for suction piping,1650 psig for discharge piping between the pump and the discharge valve, and 1550 psig beyond the discharge valve. The most limiting of these allowances is the 110% of the suction piping design pressure; therefore, the SL on maximum allowable RCS pressure is established at 1325 psig as measured at the reactor steam dome.

APPLICABILITY SL 2.1.2 applies in all MODES.

SAFETY 2.2.1 DD$

VIOLATIONSLIMIT 4 M any SL is violated, the NRC Operat enter must be not t within I hour, in accordance with R 50.72 (Ref.7).

2 Exceeding the RCS pressure SL may cause immediate RCS

, failure and create a potential for radioactive releases in (continued) en.e*

GRAND GULF B 2.0-8 Revision No. O

Attachment 4 Page 40 RCS Pressure SL !

B 2.1.2 BASES SAFETY LIMIT VIOLATIONS 4.2.2

~ 'ar.tira d; h excess of 10 CFR 100, " Reactor Site Criteria," limits !

(Ref. 4). Therefore, it is required to insert all !

insertable control rods and restore compliance with the SL within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time ensures that the i operators take prompt remedial action and also ensures that the probability of an accident occurring during this period is minimal.

2.2.3' N h-Q f any SL is violated, the Gen 1 Manager, Plant Operat' ions

[awit inthe24Vice President, Operation GNS shall be notified hours. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months perio

[ rovides time for plant opera rs and staff to take the'approp ate imediate . action '

and ass s the condition of the unit bef reporting to the senior ma ement. --

2.2.4 q

{ If any SL is violated, Licensee Event Report shall b prepared and submitted w hin 30 days to the NRC in accordance with 10 CFR 50. (Ref. 8). The report will describe the applicable circ tances preceding the violation, the effect of the vi ation upon unit components,

stems, or structures, and the c rective actions taken to p(rewtnt recurrence. A copy of the ort shall also be submitt d to the General Manager, Pla Operations and the Vice Pres ent, Operations GGNS.

2.2.5 If any SL is violated, r start of the unit shall n t comence until authorized bh he NRC. This require nt ensures the NRC that all neces reviews, analyses, nd actions are completed before the t begins its restar to normal operation.

(continued)

GRAND GULF B 2.0-9 Revision No. 1 I

Attachment 4 RCS Pressure St Page 41 8 2.1.2 BASES (continued)

REFERENCES 1. 10 CFR 50, Appendix A, GDC 14, GDC 15, and GDC 28.

2. ASME, Boiler and Pressure Vessel Code,Section III.

3.

ASME, Boiler and Pressure Vessel Code,Section XI,

. Article IWA-5000.

4. 10 CFR 100. -

5.

ASME, Boiler and Pressure Vessel Code, 1971 Edition, Addenda, winter of 1972.

6. ASME, Boiler and Pressure Vessel Code, 1974 Edition.

10CFR50.72)h

8. 10 CFR .

I t

i GRAND GULF' B 2.0-10 Revision No. 1 s-

' Attachment 4 Page 42 LC0 Applicability B 3.0 BASES i

LCO 3.0.4 provisions of LC0 3.0.4 shall not prevent changes in MODES (continued) or other specified conditions in the Applicability that result from any unit shutdown.

F h ef k- Exceptions to LCO 3.0.4 are stated in the individual

(.-A ecifications.4 Exceptions may apply to all the ACTIONS or Lgy*e to a specific Required Action of a Specification.

[Q g M LCO 3.0.4 is only applicable when entering MODE 3 from MODE

4. MODE 2 from MODE 3 or 4, or MODE 1 from MODE 2.

Furthermore, LCO 3.0.4 is applicable when entering any other specified condition in the Applicability only while operating in MODE 1, 2, or 3. The requirements of LCO 3.0.4 do not apply in MODES 4 and 5, or other specified conditions of the Applicability (unless in MODE 1, 2, or 3) because the ACTIONS of individual Specifications sufficiently define'the remedial measure to be taken. - - -

The ACTIONS for an inoperable required battery charger in LCO 3.8.4, "DC Sources - Operating," and LCO 3.8.5, "DC Sources - Shutdown," include a Note explicitly precluding entry into specific M00Es or other specified conditions of the Applicability while relying on the ACTIONS. With an inoperable required battery charger this Note in LCO 3.8.4 prohibits entry in MODE 1, 2, or 3, except during power decrease and in LC0 3.8.5 prohibits starting movement of irradiated fuel, entering MODE 4 from MODE 5, or loading fuel into the vessel if the vessel is defueled.

Surveillances do not have to be performed on the associated inoperable equipment (or or variables outside the specified limits), as permitted by SR 3.0.1. Therefore, changing MODES or other ;pecified conditions while in an ACTIONS Condition, either in compliance with LCO 3.0.4, or where an exception to LCO 3.0.4 is stated, is not a violation of SR 3.0.1 or SR 3.0.4 for those Surveillances that do not have to be performed due to the associated inoperable equipment. However, SRs must be met to ensure OPERABILITY ,

prior to declaring the associated equipment OPEPABLE (or l variable within limits) and restoring compliance with the i affected LCO.

LCO 3.0.5 LCO 3.0.5 establishes the allowance for restoring equipment to service under administrative controls when it has been i removed from service or declared inoperable to comply with l ACTIONS. The sole purpose of this Specification is to

. (continued) ;

l GRAND GULF. ,

B 3.0-6 Revision No. 0

Attaciunent 4 .

Page 43 insert B3.0-6A i

These exceptions allow entry into MODES or other specified conditions in the Applicability when the associated ACTIONS to be entered do not provide for continued operation for an unlimited period of time.

TsTF-I 04- C / /

eN s

i I

i

l Attachment 4 Control Rod OPERABILITY Page 44 8 3.1.3 BASES LC0 satisfy the intended reactivity control requirements, strict (continued) control over the number and distribution of inoperable control rods is required to satisfy the assumptions of the DBA and transient analyses.

APPLICABILITY In MODES 1 and 2, the control rods are assumed to function during a DBA or transient and are therefore required to be OPERABLE in these MODES. In MODES 3 and 4, control rods are not able to be withdrawn since the reactor mode switch is in Shutdown and a control rod block is applied. This provides adequate requirements for control rod OPERABILITY during i these conditions. Control rod requirements in MODE 5 are I located in LCO 3.9.5, " Control Rod OPERABILITY-Refueling."

ACTIONS The ACTIONS table is modified by a Note indicating that- a !

. separate Condition entry is allowed for each control rod. l This is acceptable, since the Required Actions for each '

Condition provide appropriate compensatory actions for each inoperable control rod. Complying with the Required Actions may allow for continued operation, and subsequent inoperable control rods are governed by subsequent Condition entry and application of associated Required Actions.

h.1,A.2 o 3 4.

A control rod is considered stuck if it will not insert by either CRD drive water or scram pressure. With a fully i inserted control rod stuck, no actions are required as long '

as the control rod remains fully inserted. The Required Actions are modified by a Note that allows a stuck control i TSTF:32-G k rod to be bypassed in the Rod Action Control System (RACS) to allow continued operation. SR 3.3.2.1.9 provides additional requirements when control rods are bypassed in RACS to ensure compliance with the CRDA analysis. With one Mur* 1 withdrawn control rod stuck4 the control rod must be 1 C.NyA

% ( disarmed within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . The allowed Completion Time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months is acceptable, considering the reactor can still be shut down, assuming no additional control rods fail to insert, and provides a reasonable amount of time to perform the Required Action in an orderly manner. Isolating the control rod from scram prevents damage to the CRDM. The (continued)

GRAND GULF B 3.1-14 Revision No. 0

Attaciunent 4 Page 45 Insert B3.1-14A the local scram reactivity rate assumptions may not be met if the struck control rod separation criteria are not met. Therefore, a verification that the separation criteria are met must be performed immediately. The separation criteria are not met if the struck control rod occupies a location adjacent to a " slow" control rod. The description of " slow" control rods is provided in LCO 3.1.4, " Control Rod Scram Times." In addition, T57F-3R-C6 G

l l

l

Control Rod OPERABILITY Attachment 4 g 3,t,3 l Page 46 BASES ACTIONS A.l.A.2MdA.3[(continued e JL control rod can be isolated from scram by isolating the hydraulic control unit from scram and normal drive and withdraw pressure, yet still maintain cooling water to_the CRD.

__ 3 nj '

onitoring of the insertion capability for each hdr l

' control rod must also be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 2

'C7 SR 3.1.3.2 and SR 3.1.3.3 perform periodic tests of the f

control rod insertion capability of withdrawn control rods. i Testing each withdrawn control rod ensuras that a geneti c F

ww-t _ problem does not exist.1 The allowed Completion Time of 3,j ,, s g (24 nours>provides a reasonable time to test the control Prods, considering the potential for a need to reduce oowerk to cerform the testsd Iequir:d ^:ti;n A.: h;; ; ::d F M 5 k -ti= ;ere Completier. T!=. The 24 h ur C=phti= T!= for

-th ; 5 ",;;u . red Au ivi4 naL. hc'n "- ""-'- %^- M =d bM $5 diaccii ed Lu 63.1-l5 $Nthe actualA w ;=bs atuCk Or.d II5EId'Al ECL7.

r setpcint--(LPSPftof the r:dI s. estir thl" patternt l

^cer.trello.yRPC-)t since the notch insertions may not be compatible with the requirements of rod pattern control 1 (LCO 3.1.6) and the RPC (LCO 3.3.2.1, " Control Rod Blo Instrumentation"). 7 To allow continued operation with a withdrawn control rod stuck, an evaluation of adequate SDM is also required within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Should a DBA or transient require a shutdown, to preserve the single failure criterion an additional control rod would have to be assumed to have failed to insert when required. Therefore, the original SOM demonstration may not be valid. The SDM must therefore be evaluated (by measurement or analysis) with the stuck control rod at its stuck position and the highest worth OPERABLE control rod assumed to be fully withdrawn.

The allowed Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to verify SDM is adequate, considering that with a single control rod stuck in a withdrawn position, the remaining OPERABLE control rods are capable of providing the required scram and shutdown reactivity. Failure to reach MODE 4 is only likely if an additional control rod adjaceni. tc' the stuck control rod also fails to insert during a required scram. Even with the postulated additional single failure of an adjacent control rod to insert, sufficient reactivity control remains to reach and maintain MODE 3 conditions (Ref. 7).

(continued)

GRAND GULF B 3.1-15 Revision No. 0

@%"'"" T5TF C (o insert B3.1-15A from discovery of Condition A concurrent with THERMAL POWER greater than the low power setpoint (LPSP) of the rod pattern controller (RPC)

Insert B3.1-15B from discovery of Condition A concurrent with THERMAL POWER greater than the LPSP of the of the RPC Insert B3.1-15C j

This Completion Time allows for an exception to the normal " time zero" for beginning the allowed outage time " clock." The Required Action A.2 Completion Time only begins upon discovery of Condition A concurrent with THERMAL POWER greater than l

4

Anachment 4 Control Rod Scram Times Page 48 B 3.1.4 BASES LCO To ensure that local scram reactivity rates are maintained (continued) within acceptable limits, no " slow" control rod may occupy a location adjacent to another " slow" control rod er ;dj: car *

i.v a withdrewr. n u d ceritrel d .

T5TF-he Tabie 3.i.4-2 is modified by two notes, which state control rods with scram times not within the limits of the Table are considered " slow" and that control rods with scram times-

> 7 seconds are considered inoperable as required by SR 3.1.3.4.

This LCO applies only to OPERABLE control rods since inoperable control rods will be inserted and disarmed (LC03.1.3). Slow scramming control rods may be conservatively declared inoperable and not accounted for as

" slow" control rods.

APPLICABILITY ~

In MODES I and 2, a scram is assumed to function du' ring transients and accidents analyzed for these plant conditions. These events are assumed to occur during startup and power operation; therefore, the scram function

' of the control rods is required during these MODES. In MODES 3 and 4, the control rods are not able to be withdrawn since the reactor mode switch is in shutdown and a control I rod block is applied. This provides adequate requirements for control rod scram capability during these conditions.

Scram requirements in MODE 5 are contained in LCO 3.9.5,

" Control Rod OPERABILITY-Refueling."

ACTIONS A.1 1

When the requirements of this LCO are not met, the rate of l negative reactivity insertion during a scram may not be I within the assumptions of the safety analyses. Therefore, the plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is reasonable, based on operating l experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems. {

l (continued)

GRAND GULF B 3.1-23 Revision No. I

f' Attaciunent 4 Leakage Detecdon Inshumentation Page 49 8347 BASES APPLICABLE RCS leakage detection instrumentation satisfies Criterion 1 SAFETY ANALYSES of the NRC Policy Statement.

(continued)

LC0 The drywell floor drain sump monitoring system is required to quantify the unidentified LEAKAGE from the RCS. Thus, for the system to be considered OPERABLE, the sump level monitoring portion of the system must be OPERABLE. The other monitoring systems provide qualitative indication to the operators so closer examination of other detection systems will be made to determine the extent of any corrective action that may be required. With the leakage detection systems inoperable, monitoring for LEAKAGE in the RCPB is degraded.

APPLICABILITY In MODES 1, 2, and 3, leakage detection systems are required

, to be OPERABLE to support LC0 3.4.5. This Applicability is consistent with that for LCO 3.4.5.

ACTIONS A.1

{ 1:se-N l With the drywell floor drain sump monitoring system O 3.4 -3'/ A in perable, no other form of sampling can provide the equivalent information to quantify leakage. However, the drywell atmospheric activity monitor and the drywell air cooler condensate flow rate monitor will provide indications of changes in leakage.

'5TF60CID With the drywell floor drain sump monitoring system inoperable, but with RCS unidentified and total LEAKAGE being determined every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (SR 3.4.5.1), operation may continue for 30 days. The 30 day Completion Time of Required Action A.1 is acceptable, based on operating experience, considering the multiple forms of laaka detection that are still available.fRequired Action A.1 is k (modified by a Note Enat states thaT' % e t - h'ons of us 0.0.1 r= nnt applicable. As a result aM t , is allowed when the drywell-W,onitoring system is i This allowance is providea oeca , .

j nstrumentation is '

e to monitor RCS leakage. 1 (continued)

GRAND GULF B 3.4-34 Revision No. O I

e

Attachment 4 Page 50 Insert B3.4-34A The Actions are modified by a Note that stat'es that the provisions of LCO 3.0.4 are not applicable. As a result, a MODE change is allowed when the drywell floor drain sump monitoring system and required radiation monitors are inoperable. This allowance is provided because other instrumentation is available to monitor RCS leakage.

T5TF40 -Cr0 l

q

.J Attaciunent 4 RCS Leakage Detection Instrumentation Pagc 51 B 3.4.7 BASES ACTIONS B.1 (continued)

With both gaseous and particulate drywell atmospheric monitoring channels inoperable, grab samples of the drywell atmosphere shall be taken and analyzed to provide periodic leakage information. Provided a sample is obtained and analyzed every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , the plant may continue operation since at least one other form of drywell leakage detection (i.e., air cooler condensate flow rate monitor) is available. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval provides periodic information that is adequate to detect LEAKAGE.

i l

C.1 I With the required drywell air cooler condensate flow rate I monitoring system inoperable, SR 3.4.7.1 is performed every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to provide periodic information of activity in the drywell at a more frequent interval than the routine Frequency of SR 3.4.7.1. The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months interval provities' periodic information that is adequate to detect LEAKAGE and recognizes that other forms of leakage detection are ,

available. However, this Required Action is modified by a l i

Note that allows this action to be not applicable if the required drywell atmospheric monitoring system is inoperable. Consistent with SR 3.0.1, Surveillances are not required to be performed on inoperable equipment.

D.1 and D.2 With both the gaseous and particulate drywell atmospheric monitor channels and the drywell air cooler condensate flow rate monitor inoperable, the only means of detecting LEAKAGE is the drywell floor drain sump monitoring system. This Condition does not provide the required diverse means of leakage detection. The Required Action is to restore either of the inoperable monitoring systems to OPERABLE status within 30 days to regain the intended leakage detection diversity. The 30 day Completion Time ensures that the plant will not be opepted in a degraded configuration for_a lenathy time oeriod.3The Required Actions are moattied by al FYote that states that s of LC0 3.0.4 are not - l M . As a result, a MODE change i nA when both Qedrywellga'seuua ar.d p rt h hte atmospheric monito W channeT5 and-dr.yg11 air cooler condensate 7tcw rate

/ (continued)

I GRAND GULF '

gp-wcv B 3.4-35 Revision No. 0 1'

Attachment 4 RCS Leakage Detection Instrumentation Page $2 B 3.4.7 BASES ACTIONS 0.1 and 0.2 (continued) b ring system are inoperable. This allowance is nrovided because otner instrumvideMnn h available to 6 1 tor HL5 iceka3C. _-

El1 and E.2 T5TF40-sto If any Required Action of Condition A, B, C, or D cannot be met within the associated Completion Time, the plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions in 2n orderly manner and without challenging plant systems.

p,3 - --

With all required monitors inoperable, no required automatic means of monitoring LEAKAGE are available, and immediate

( plant shutdown in accordance with LC0 3.0.3 is required.

SURVEILLANCE SR 3.4.7.1 REQUIREMENTS This SR requires the performance of a CHANNEL CHECK of the required drywell atmospheric monitoring system. The check gives reasonable confidence that the channel is operating properly. The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is based on instrument reliability and is reasonable for detecting off normal conditions.

SR 3.4.7.2 This SR requires the performance of a CHANNEL FUNCTIONAL TEST of the required RCS leakage detection instrumentation.

The test ensures that the monitors can perform their function in the desired manner. The test also verifies the relative accuracy of the instrumentation. The Frequency of 31 days considers instrument reliability, and operating experience has shown it proper for detecting degradation.

(continued)

GRAND GULF B 3.4-36 Revision No. O e

Attachment 4 RHR Shutdown Cooling System-Hot Shutdown Page 53 8 3.4.9 BASES LC0 OPERABLE RHR pump, two heat exchangers in series, and the (continued) associated piping and valves. Each shutdown cooling subsystem is considered OPERABLE if it can be manually aligned (remote or local) in the shutdown cooling mode for removal of decay heat. In MODE 3, one RHR shutdown cooling subsystem can provide the required cooling, but two subsystems are required to be OPERABLE to provide redundancy. Operation of one subsystem can maintain or reduce the reactor coolant temperature as required. ,

However, to ensure adequate core flow to allow for accurate average reactor coolant temperature monitoring, nearly _ )

continuous operation is required. ,

g, Note 1 permits both RHR shutdow cooling subsystems and TM--lWGO recirculation pumps to w..a w for a period of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in an 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period. Note 7 allows one RHR shutdown cooling subsystem to be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for performance of surveillance tests. These tests may 6e on the affected

. RHR System or on some other plant system or component .that necessitates placing the RHR System in an inoperable status during the performance. This is permitted because the core heat generation can be low enough and the heatup rate slow enough to allow some changes to the RHR subsystems or other operations requiring RHR flow interruption and loss of redundancy.

i 1

APPLICABILITY In MODES I and 2, and in MODE 3 with reactor steam dome pressure greater than or equal to the RHR cut in permissive i pressure, this LC0 is not applicable. Operation of the RHR System in the shutdown cooling mode is not allowed above this pressure because the RCS pressure may exceed the design pressure of the shutdown cooling piping. Decay heat removal at reactor pressures greater than or equal to the RHR cut in permissive pressure is typically accomplished by condensing !

the steam in the main condenser. Additionally, in MODE 2 below this pressure, the OPERABILITY requirements for the Emergency Core Cooling Systems (ECCS) (LC0 3.5.1, "ECCS-Operating") do not allow placing the RHR shutdown cooling subsystem into operation.

In MODE 3 with reactor steam dome pressure below the RHR cut in permissive pressure (i.e., the actual pressure at which the interlock resets) the RHR System may be operated in the (continued)

GRAND GULF B 3.4-43 Revision No. 0

I AnacInnent a RHR Shutdown Cooling System-Cold Shutdown Pasc 54 , B 3.4.10

, BASES LCO aligned (remote or local) in the shutdown cooling mode for (continued) removal of decay heat. In MODE 4, one RHR shutdown cooling subsystem can provide the required cooling, but two subsystems are required to be OPERABLE to provide redundancy. Operation of one subsystem can maintain and reduce the reactor coolant temperature as required.

However, to ensure adequate core flow to allow for accurate average reactor coolant temperature monitoring, nearl continuous operation is required. g .

Mim Note 1 permits both RH3 shutdow cooling subsystems and recirculation pumps tdW Mt 1:M for a period of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in an 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period. Note 2 allows one RHR shutdown cooling subsystem to be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for performance of surveillance tests. These tests may be on the affected k " lg /g } RHR System or on some other plant system or component that necessitates placing the RHR System in an inoperable status during the performance. This is permitted because the core

' heat generation can be low enough and the heatup rate slow enough to allow some changes to the RHR subsystem's or o~ther operations requiring RHR flow interrup_tiog loss _of !

redundancy.

g g p edf j Note 3 permits both RHR shutdown, ooling subsystems and recirculation pumps t&O "EC'#ES during performance of inservice leak testing and during hydrostatic testing. This is permitted because RCS pressures and temper:tures are being closely monitored as required by LC0 3.4.11. I i APPLICABILITY In MODES 1 and 2, and in MODE 3 with reactor steam dome pressure greater than or equal to the RHR cut in permissive pressure, this LC0 is not applicable. Operation of the RHR System in the shutdown cooling mode is not allowed above this pressure because the RCS pressure may exceed the design pressure of the shutdown cooling piping. Decay heat removal at reactor pressures greater than or equal to the RHR cut in i permissive pressure is typically accomplished by condensing l the steam in the main condenser. Additionally, in MODE 2 i below this pressure, the OPERABILITY requirements for the Emergency Core Cooling Systems (ECCS) (LC0 3.5.1, "ECCS-Operating") do not allow placing the RHR shutdown ;

cooling subsystem into operation. -

l (continued)

GRAND GULF B 3.4-48 Revision No. O {

c j l

l Attachment 4 Primary Containment Air Locks

[

Pagc 55 8 3.6.1.2 BASES ,

)

SURVEILLANCE SR 3.6.1.2.2 (continued)

REQUIREMENTS following access through the air lock, which occurs regularly. The 7 day Frequency has been shown to be ;

acceptable through operating experience and is considered l adequate in view of the other indications available to operations personnel that the seal air flask pressure is -

low.

SR 3.6.1.2.3 The air lock interlock mechanism is designed to prevent simultaneous opening of both doors in the air lock. Since both the inner and outer doors of an air lock are designed to withstand the maximum expected post accident primary containment pressure (Ref. 3), closure of either door will support primary containment OPERABILITY. Thus, the interlock feature supports primary containment -0PERABILITY while the air lock is be!ng used for personnel transit in and out of the containment. Periodic testing of this interlock demonstrates that the interlock will function as i

designed and that simultaneous inner and outer door opening I

% #i will not inadvertently occur. Due to the nature of his '

6 9-G A interlock, and given that the interlock mechanism is

- nged when the primary containment air lock door is

-QA- * , c , this test is only required to be performed Qupon

_6 34-I28^"kn'o't reauired more frequently than once per 184-days,J ihe#"'E 2I 2m t ptim.g%

C degFrequency is based on engineering judgment and is considered adequate'hn ;;; vi uus ac "'";; rat we

)

%=8 \ Jtc-tre?D h y+

616-J tC } ~ noemJq f j SR 3.6.1.2.4 TSTF -17 A seal pneumatic system test to ensure that pressure does not decay at a rate equivalent to > 2 psig for a period of

{l)I 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months from an initial pressure of 90 psig is an effective leakage rate test to verify system performance. The 18 month Frequency is based on the fact that operating experience has shown these components usually pass the Surveillance when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the Frequency l was concluded to be acceptable from a reliability standpoint.

(continued)

GRAND GULF B 3.6-12 Revision No. O c

I Attachment 4 Page 56 Insert B 3.6-12A g y l

used for entry and exit (procedures require strict adherence to single door opening) insert B 3.6-128 '

every 24 months. The 24 month Frequency is based on the need to perform this Surveillance under conditions that apply during a plant outage, and the potential for loss of primary containment OPERABILITY if the Surveillance was performed with the reactor at power. The 24 i

month frequency for the interlock is justified based on generic operating experience.

{

insert B 3.6-12C given that the interlock is not challenged during the use of the airlock.

(

l l

1 l

l l

k

1 Attaciunent 4 Pagc 57 PClVs B 3.6.1.3 p

I-BASES SURVEILLANCE SR 3.6.1.3.1 (continued)

REQUIREMENTS purge valves are capable of closing in the environment following a LOCA. Therefore, these valves are allowed to be open for limited periods of time. The 31 day Frequency is consistent with other PCIV requirements. - ---

SA oA lockO,% h&,

SR 3.6.1.3.2 *C b*

MNM -

This SR verifies that each primary containment isolation manual valve and Llind flange that is located outside primary containment, drywell, and steam tunnel, and is required to be closed during accident conditions, is closed.

The SR halps to ensure that post accident leakage of radioactive fluids er gases outside of the primary containment boundary it within design limits. This SR does not require any testing or valve manipulation. Rather, it involves verification that those devices outside pnimary containment, drywell, and steam tunnel, and capable of being mispositioned, are in the correct position. Since verification of valve position for devices outside primary containment is relativo1y easy, the 31 day Frequency was

( chosen to provide added assurance that the devices are in the correct positions.

Two Notes are added to this SR. The first Note applies'to valves and blind flanges located in high radiation areas and allows them to be verified by use of administrative controls. Allowing verification by administrative controls is considered acceptable, since access to these arehs is ,

typically restricted during MODES 1, 2, and 3 for ALARA reasons. Therefore, the probability of misalignment of

.these devices, once they have been verified to be in the proper position, is low. A second Note is included to clarify that PCIVs open under administrative controls are not required to meet the SR during the time the PCIVs are open.

TSTF-M a~9 a M " M SR 3.6.1.3.3 Ch bW+

This SR verifies that each primary containment manual isolation valve and blind flange locatejd inside primary l containment, drywell, or steam tunnel,4and required to be j closed during accident conditions, is closed. The SR helps (continued)

GRAND GULF ' B 3.E-22 Revision No. O c

I Secondary Containment I Attachment 4 g 3,6,4,} l Page 58 l

BASES C.l. C.2. and C.3

~

ACTIONS (continued)

Movement of irradiated fuel assemblies in the primary or secondary containment, CORE ALTERATIONS, and OPDRVs can be ,

postulated to cause fission product release to the secondary containment. In such cases, the secondary containment is the only barrier to release of fission products to the environment. CORE ALTERATIONS and movement of irradiated I fuel assemblies must be immediately suspended if the i secondary containment is inoperable.

Suspension of these activities shall not preclude completing an action that involves moving a component to a safe l position. Also, action must be immediately initiated to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release.. Actions must continue until OPDRVs are suspended.

. Req'uired Action C.1 has been modified by a Note stat.ing that LCO 3.0.3 is not. applicable. If moving irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.

SURVEILLANCE SR 3.6.4.1.1 and SR 3.6.4.1.2 in u4 s Can REQUIREMENTS one 4 Verifying that Auxiliary Building and Enclesure Bullding equipment hatches, blowout panels, an access door are closed ensures that the infiltration of outside air of such a magnitude as to prevent maintaining the desired negative pressure does not occur. Verifying that all such openings T5TF-tT-C5 are closed provides adequate assurance that exfiltration from the secondary containment will not occur. In this application the term " sealed" has no connotation of leak tightness. Maintaining secondary containment OPERABILITY requires verityingpad door in the access opening is

=

closed, except whEn the access opening is being used for v entry and exit;,. The 31 day Frequency for these SRs has been I gag -

shown to be adequate based on operating experience, and is considered adequate in view of the other controls on tvu l 0 }C P M L C, secondary containment access openings.

W ] W o W Y & GN qcCt%S o m _ _ (continuJ)

GRAND GULF B 3.6-86 Revision No. 2 3

... .;.~. - - -

3

{'

AC Sources-Operating Attachment 4 g 3,g,}

Page 59

. BASES SURVEILLANCE SR 3.8.1.2 REQUIREMENTS (continued) This SR helps to ensure the availability of the standby electrical power supply to mitigate DBAs and transients and maintain the unit in a safe shutdown condition.

To minimize the wear on moving parts that do not get lubricated when the engine is not running, this SR is modified by a Note to indicate that all DG starts for this Surveillance may be preceded by an engine prelube period and followed~by a warmup period prior to loading.

For the purposes of this testing, the DGs are started from standby conditions. Standby conditions for a DG mean that the diesel engine coolant and oil are being continuously circulated and temperature is being maintained consistent with manufacturer recommendations for DG 11 and DG 12. For DG 13, standby conditions mean that the lube oil is heated by the jacket water and continuously circulated through a portion of the system as recommended by the vendor. Engine jacket water is heated by an immersion heater and circulates

'through the system by natural circulation.

^

SR 3.8.1.2 requires that the DG starts from standby conditions and achieves required voltage and frequency within 10 seconds. The DG's ability to maintain the required voltage and frequency is tested by those SRs which require DG loading. The 10 second start requirement r supports the assumptions in the design basis LOCA analysis 2* \ (Ref. 5)S u83,1-MA -

The DGs are started for this test by using one of the following signals: manual, simulated loss of offsite power by itself, simulated loss of offsite power in conjunction with an ESF actuation test signal, or an ESF actuation test TI _fI L signal by itself.

The 31 day Frequency for SR 3.8.1.2 is consistent with the industry guidelines for assessment of diesel generator performance (Ref. 14). This Frequency provides adequate assurance of DG OPERABILITY, while minimizing degradation resulting from testing.

(continued) l GRAND GULF B 3.8-15 LDC 98004 e

7,.

' Attachment 4 j Page 60 -

l Insert B 3.8-15A In addition to the SR requirements, the time for the DG to reach steady state operation is periodically monitored (data is taken at least once per 6 months during the performance of SR 3,8.1.2) and the trend evaluated to identify degradation of govemor and voltage regulator performance.

% TF-(0 -c/lf l

l l

N.

Diesel Fuel Oil, Lube Oil, and Starting Air Attacllment 4 '

Page 6i BASES a

SURVEILLANCE SR 3.8.3.5 REQUIREMENTS -

1 (continued) Microbiological fouling is a major cause of fuel oil degradation. There are numerous bacteria that can grow in fuel oil and cause fouling, but all must have a water environment in order to survive. Removal of water from the storage tanks once every 92 days eliminates the necessary environment for bacterial survival. This is the most effective means of controlling microbiological fouling. In addition, it eliminates the potential for water entrainment in the fuel oil during DG operation. Water may come from any of several sources, including condensation, ground water, rain water, contaminated fuel oil, and from breakdown of the fuel oil by bacteria. Frequent checking for and removal of accumulated water minimizes fouling and provides data regarding the watertight integrity of the fuel oil system, The Surveillance Frequencies are established by RegulatoryGuide1.137(Ref.2). This SR is for preventive maintenance. The pret.snce of water does not necessarily represent a failure of this SR previded that accumulated water is removed during performance of the 5drveillance,

/TR 3.8.3.6 N aining of the fuel oil stored in the upply tanks, remova o mulated sediment, and tank cleaning are required at 10 year rvals by Regulatory Guide 1.137 (Ref. 2), 3 paragraph 2. . is SR is typically performed in conjunction with t ME Boiler and Pressure Vessel Code,Section XI (Ref. 7), exa tions of the tanks. To preclude the introduction of surfactan n the fuel oil system, the I cleaning should be accomplished us sodium hypochlorite solutions, or their equivalent, rather n soap or detergents. This SR is for preventive main ance. The presence of sediment does not necessarily repre a failure of this SR provided that accumulated sedimen Q ed during performance of the Surveillance._

REFERENCES 1. UFSAR, Section 9.5.4.

2. Regulatory Guide.1.137.

3. ANSI N195, Appendix B, 1976.

(continued)

GRAND GULF B 3.8-49 Revision No. 1 4'

. l

r Diesel Fuel Oil, Lube Oil, and Starting Air Attachment 4 Page 62 8 3.8*3 BASES REFERENCES .4. UFSAR, Chapter 6.

(continued).

5. UFSAR, Chapter 15.

6. ASTM Standards: D4057-88; D975-92a; D2274-70.

A Q. 10".0, Builer . id Ti ..u, =

. .l C e i , S a i.iurr X h '

GTF C

p. . .

i i

1 l

GRAND GULF ' B 3.8-50 Revision No. O e

DC Sources-Operating Attachmeru 4 8 3.8.4 Pagc 63 I

(

BASES (continued)

~

SURVEILLANCE SR 3. 8. 4. I' REQUIREMENTS i Verifying battery terminal voltage while on float charge I helps to ensure the effectiveness of the charging system and the ability of the batteries to perform their intended I function. Float charge is the condition in which the L charger is supplying the continuous charge required to overcome the internal losses of a battery (or battery cell) and maintain the battery (or battery cell) in a fully charged state. The voltage requirements are based on the nominal design voltage of the battery and are censistent with the initial voltages assumed in the battery sizing calculations. The 7 day Frequency is consistent with manufacturer's recommendations and IEEE-450 (Ref. 8).

SR 3.8.4.2 Visual inspection to detect corrosion of the battery cells and connections, or measurement of the resistance of each inter-cell, inter-rack, inter-tier, and terminal connection, provides an indication of physical damage or abnormal

, deterioration that could potentially degrade battery I- performance.

The Surveillance Frequency for these inspections, which can detect conditions that can cause' power losses due to resistance heating, is 92 days. This Frequency is considered acceptable based on operating experience related to detecting corrosion trends.

SR 3.8.4.3 Visual inspection of the battery cells, cell plates, and g,-& battery racks provides an indication of physical damage or

{6 35-5Gq A abnormal deterioration that could potentially degrade battery performance. 3 The 18 month Frequency of the Surveillance is based on engineering judgement, taking into consideration the desired l unit conditions to perform the Surveillance. Operating l experience has shown that these components usually pass the '

SR when performed at the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

(continued)

GRAND GULF ' B 3.8-56 Revision No. O

e Attachment 4 Page 64 I

Insert B 3.8-56A ThIf "3 [~bh The presence of physical damage or deterioration does not necessarily represent a failure o'f this SR, provided an evaluation determines that the physical damage or deterioration does not affect the OPERABILITY of the battery (its ability to perform its design function), i i

I I

RHR-High Water Level Attachment 4

Page 65 BASES APPLICABLE Although the RHR System does not meet a specific criterion SAFETY ANALYSES of the NRC Policy Statement, it was identified in the NRC (continued) Policy Statement as an important contributor to risk reduction. Therefore, the RHR System is retained as a Specification. The ADHRS is included in the Specification to provide requirements for decay heat removal capability during an outage while the RHR System is out of service.

LCO Only one RHR shutdown cooling subsystem is required to be OPERABLE in MODE 5 with irradiated fuel in the RPV and the water level a 22 ft 8 inches above the RPV flange. Only one subsystem i~s required because the volume of water above the RPV flange provides backup decay heat removal capability.

An OPERABLE RHR shutdown cool!ng subsystem consists of an RHR pump, two heat exchangers, valves, piping, instruments, and controls to ensure an OPERABLE flow path. The required

~ RHR shutdown cooling subsystem must have a OPERABLE diesel generator capable of supplying electrical pbwer. ~

Additionally, each RHR shutdown cooling subsystem is considered OPERABLE if it can be manually aligned (remote or local) in the shutdown cooling mode' for removal of decay heat. Operation (either continuous or intermittent) of one decay heat removal subsystem (either RHR or ADHRS) can T6TT~-19-Q3="intain and reduce the reactor cooiant temperature as required. However, to ensure adequate core flow to allow for accurate average reactor coolant temperature monitoring, 6[ g%g nearly continuous operatioqis required. A Note is provided to allow a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months exception C a ^- b the operating subsystem every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

APPLICABILITY One RHR shutdown cooling subsystem must be OPERABLE in MODE 5, with irradiated fuel in the RPV and the water level a 22 'ft 8 inches above.the top of the RPV flange, to provide decay heat removal. RHR System requirements in other MODES are covered by LCOs in Section 3.4, Reactor Coolant System (RCS); Section 3.5, Emergency Core Cooling Systems (ECCS) and Reactor Core Isolation Cooling (RCIC)

System; and Section 3.6', Containment Systems. RHR Shutdown l

Cooling System requirements in MODE 5, with the water level

< 22 ft 8 inches above the RPV flange, are given in LCO 3.9.9, " Residual Heat Removal (RHR)-Low Water Level."

(continued)

GRAND GULF B 3.9-26 Revision No. I

V RHR-Le Water Level Attachment 4 8 3,g,g Page 66 BASES APPLICABLE Specification. The ADHRS is included in the Specification SAFETY ANALYSES to provide requirements for decay heat removal capability (continued) during an outage while the RHR System is out of service.

LCO In MODE 5 with irradiated fuel in the reactor pressure vessel (RPV) and the water level < 22 ft 8 inches above the RPV flange both RHR shutdown cooling subsystems must be OPERABLE, or the ADHRS may be substituted for one of the RHR subsystems.

An OPERABLE RHR shutdown cooling subsystem consists of an RHR pump, two heat exchangers, valves, piping, instruments.

and controls to ensure an OPERABLE flow path. An OPERABLE ADHRS consists of two pumps, two heat exchangers, valves, piping, instruments and controls to ensure an OPERABLE flow path. At least one of the regoired RHR shutdown cooling subsystems must have a OPERABLE diesel generator capable of supplying electrical power.

~

Additionally, each RHR shutdown cooling subsystem is considered OPERABLE if it can be manually aligned (remote or local) in the shutdown cooling mode for removal of decay heat. Operation (either continuous or intermittent) of one 1

p .l g]. C G decay heat removal subsystem (either RHR or ADHRS) can maintain and reduce the reactor coolant temperature as required. However, to ensure adequate core flow to allow g o3 g for accurate average reactor coolant temperature monitoring, a Pg,g nearly continuous operatior,is reguired. A Note is provided to allow a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months exceptior the operating subsystegevery 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

APPLICABILITY Two decay heat removal subsystems are required to be OPERABLE in MODE 5, with irradiated ~ fuel in the RPV and the water level < 22 ft 8 inches above the top of the RPV flange, to provide decay heat removal. RHR System requirements in other MODES are covered by LCOs in Section 3.4, Reactor Coolant System (RCS); Section 3.5, Emergency Core Cooling Systems (ECCS) and Reactor Core ,

Isolation Cooling (RCIC) System; and Section 3.6, Containment Systems. RHR Shutdown Cooling System requirements in MODE 5, with the water level a: 22 ft 8 inches above the RPV flange, are given in LCO 3.9.8,

" Residual Heat Removal (RHR)-High Water Level."

(continued)

GRAND GULF, B 3.9-30 Revision No. 1 9

0

ML20211E715

Text

Navigation menu

Search