Text

Amend 70 to License NPF-29,revising Tech Specs & Bases by Adding Requirements for Operation & Use of Alternate DHR Sys During Future Outages
	ML20028G960
Person / Time
Site:	Grand Gulf
Issue date:	09/24/1990
From:	Quay T; Office of Nuclear Reactor Regulation
To:	;
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ML20028G961	List: ML20028G960; ML20028G962;
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	NUDOCS 9010040305
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l UNITED STATES h

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NUCLEAR REGULATORY COMMISSION 4

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,*4 ENTERGY OPERATIONS. INC.

SYSTEM ENERGY RESOURCES. INC.

SOUTH M1551551PPI ELECTRIC POWER ASSOCIATION MISS!$$1PPI POWER AND LIGHT COMPANY DOCKET NO. 50-416 GRAND GULF NUCLEAR STATION UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 70 License No. NPF-29 1.

The Nuclear Regulatory Comission (the Comission) has found that A.

The application for amendmen* 'y the licensee dated April 27,1990, and revised July 5,1990. August 6 1990 August 9, 1990 August 20, 1990, and September 11, 1990, complies wIth the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

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

Thereisreasonableassurance(i)thattheactivitiesauthorizedby this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

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

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied, gkooggggg0gggj416 PNV p

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2.

Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. NPF.29 is hereby amended to read as follows.

(2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised i

through Amendment No. 70, are hereby incorporated into this license. Entergy Operations, Inc. shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION u ech w h

.^^ o p Theodore R. Quay, Acting Director Project Directorate IV-1 l

Division of Reactor Projects - III l

IV, Y, and Special Projects Office of Nuclear Reactor Regulation

Attachment:

l-Changes to the Technical l-Specift:ations Date of Issuance: September 24, 1990 l

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U ATTACHMENT TO LICENSE AMENDMENT h0. 70 l

FACIllTY OPERATING LICENSE NO. NPF-29 DOCKET NO. 50-416 Replace the following > ages of the Appendix A Technical Specifications with the attached pages. T1e revised areas are indicated by marginal lines.

Remove Insert 3/4 3-13 3/4 3-13 3/4 3-14 3/4 3-14 3/4 3-15 3/4 3-15 3/4 3-25 3/4 3-25 3/4 4-27 3/4 4-27 3/4 5-6 3/4 5-6 3/4 6-28 3/4 6-28 3/4 9-18 3/4 9-18 3/4 9-19 3/4 9-19 B 3/4 5-2 B 3/4 5-2 B 3/4 5-2a B 3/4 9-2 B 3/4 9-2 B 3/4 9-2a i

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k TA8LE 3.3.2-1 (Continued)

E ISOLATION ACTUATION INSTRUMENTATION E

VALVE ~ GROUPS MINIMUM APPLICABLE OPERATED BY OPERASLE CHANNELS OPERATIONAL J.

z TRIP FUNCTION SIGNAL-(a) PER TRIP SYSTEM (b)

CONDITION ACTION 5.

REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION 1.

RHR Equipment Room Ambient Temperature - High 4

1/ room 1, 2, 3 27 j.

RHR Equipment Room a Temp. -

High 4

1/ room 1, 2, 3 27 k.

RHR/RCIC Steam Line Flow -

High 4

1 1,2,3 27 I) 1 1,2,3 26 1.

Manual Initiation 4

m Drywell Pressure-High 9(")

1 1,2,3 27 m.

J, (ECCS-Division 1 and Division 2) w 6.

RNR SYSTEM ISOLATION RHR Equipment Room Ambient a.

Temperature - High 3

1/ room 1, 2, 3 28 b.

RHR Equipment Room a Temp. - High 3

1/ room 1, 2, 3 28 h

c.

Reactor Wassel Water Level - Low, Level 3 3

2 1,2,3 28 g

3(p) 2(p) 4, 5 31 I

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d.

Reactor Vessel (RHR Cut-in p

Permissive) Pressure -

3(1) 2 1,2,3 28 High II)

Drywell Pressure - High 3

2 1, 2, 3 28 e.

f.

Manual Initiation 3

2 1, 2, 3 26 l.4

TABLE 3.3.2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION ACTION Be in at least HOT SHUT 5013 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN ACTION 20 within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Close the affected system isolation valveb) within one hour or:

ACTION 21 a.

In OPERATIONAL CONDITION 1, 2, ce b Le in at least HOT SHUTDOWN within the next 12 nours and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

In OPERATIONAL CONDITION *, suspend CORE ALTERATIONS, handling of irradiated fuel in the primary containment and oW rations with a potential for draining the reactor vessel.

Restore the manual initiation function to OPERABLE status within ACTION 22 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Be in at least STARTUP with the associated isolation valves closed ACTION 23 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Be in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 24 Establish SECONDARY CONTAINMENT INTEGRITY with the standby gas ACTION 25 tr6atment system operating within one hour.

Restore the manual initiation function to OPERABLE status ACTION 26 within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or close the affected system isolation valves within the next hour and declare the affected system inoperable.

Close the affected system isolation valves within one hour ACTION 27 and declare the affected system inoperable.

Within one hour lock the affected system isolation valves closed, ACTION 28 or verify, by remote indication, that the valve is closed and electrically disarmed, or isolate the per.atration(s) and declete the affected system inoperable.

Close the affected system isolation valves within one hour and ACTION 29 declare the affected system or component inoperable or:

a.

In OPERATIONAL CONDITION 1, 2 or 3 be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

In OPERATIONAL CONDITION # suspend CORE ALTERATIONS and opera-tions with a potentini for draining the reactor vessel.

Declare the affected SLCS pump inoperable.

ACTION 30 Isolate the shutdown cooling common suction line within one hour ACTION 31 if it is not needed for shutdown cooling or initiate action within one hour te establish SECONDARY CONTAINMENT INTEGRITY.

NOTES When handling irradiated fuel in the primary or secondary containment and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel.

Yi.3 low condenser vacuum MSIV closura may be manually bypassed during reactor SHUTDOWN or for reactor STARTUP when condenser vacuum is below the trip set-point to allow opening of the MSIVs.

The manual bypass shall be removed when condenser vacuum exceeds the trip setpoint.

During CORE ALTERATIONS and operations with a potential for draining the reactor vessel.

With any control rod withdrawn.

Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

(a) See Specification 3.6.4, Table 3.6.4-1 for valves in each valve group.

GRAND GULF-UNIT 1 3/4 3-14 Amendment No. 70

1 TABLE 3.3.2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION NOTES (Continued)

(b) A channel may be placed !; an inoperable status for ur to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for required surveillance without placing the trip system in the tripped con-dition provided at *. east one other OPERABLE channel in the same trip system is monitoring thu parameter.

(c) Also actuates the standby gas treatment system.

(d) Also actuates the control room emergency filtration system in the isolation mode of operation.

I (e) Two upscale-Hi Hi, one upscale-Hi Hi and one downscale, or two downscale signals from the same trip system actuate the trip system and initiate t

isolation of the associated containment and drywell isolation valves.

(f) Also trips and isolates the mechanical vacuum pumps.

(g) Deleted.

(h) Also actuates secondary containment ventilation isolation dampers and valves per Table 3.6.6.2-1.

(i) Closes only RWCU system isolation valves G33-F001, G33-F004, and G33-F251.

(j) Actuates the Standby Gas Treatment System and isolater Auxiliary Building penetration of the ventilation systems within the Auxiliary Building.

(k). Closes only RCIC outboard valves.

A concurrent RCIC initiation signal is required for isolation to occur.

(1) Valves E12-F037A and E12-F037B are closed by high drywell pressure.

All other Group 3 valves are closed by high reactor pressure.

(m) Valve Group 9 requires concurrent drywell high pressure and RCIC Steam Supply Pressure-Low signals to isolate.

(n) Valves E12-F042A and E12-F042B are closed by Containment Spray System initiation signals.

(o) Also isolates valves E61-F009 E61-F010 E62-F056, and E61-F057 from Valve Group 7.

(p) Only required to isolate RHR system isolation valves E12-F008 and E12-F009.

One trip system and/or isolation valve may be inoperable for up to 14 days without placing the trip system in the tripped condition provided the diesel generator associated with the OPERABLE isolation valve is OPERABLE.

GRAND GULF-UNIT 1 3/4 3-15 Amendment No. 70 l

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TABLE 4.3.2.1-1 (Continued) zo ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS E

CHANNEL OPERATIONAL 4

CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH 5

TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED

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5.

REACTOR CORE ISOLATION COOLING SYSTLn ISOLATION (Continued) h.

Main Steam Line Tunnel Temperature Timer NA M

Q 1, 2, 3 i.

RHR Equipment Room Ambient Temperature - High 5

M A

1,2,3 j.

RHR Equipment Room a Temp. -

High S

M A

1, 2, 3 R

k.

RHR/RCIC Steam Line Flow -

IC)

Y High S

M R

1, 2, 3 1.

Manual Initiation MA M(a)

NA 1,2,3 m.

Drywell Pressr.re-High 5

M R(C) 1, 2, 3 (ECCS Divisien 1 and Division 2) 6.

RHR SYSTEM ISOLATION a.

RHR Equipment Room Ambient Tercerature - High 5

M A

1, 2, 3 5g b.

RHR Equipment Room g

a Temp. - High S

M A

1,2,3 5

c.

Reactor Vesse! Water Level -

P Low..Levei 3 S

M R(c) 1,2,3,4,5 z

b d.

Reactor Vessel (RHR Cut-in Permissive) Pressure - High S

M R(c) 1, 2, 3

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i REACTOR COOLANT SYSTEM COLD SHUTDOWN QMITINGCONDITIONFOROPERATION i

3.4.9.2 Two# shutdown cooling mode loops of the residual heat removal (RHR) system shall be OPERABLE ** and, unless at least one recirculation pump is in l

operation, at least one shutdown cooling mode loop shall be in operation *'N with each loop consisting of at least:

a.

One OPERABLE RHR pump, and b.

One OPERABLE RHR heat exchanger.

APPLICABILITY:

OPERATIONAL CONDITION 4.

ACTION:

a.

With less than the above required RHR shutdown cooling mode loops OPERABLE, within one hour and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, demonstrate the OPERABILITY of'at least one alternate method capable of decay heat l

removal for each inoperable RHR shutdown cooling mode loop.

I b.

With no RHR shutdown cooling mcde loop in operation, within one hour establish reactor coolant circulation by an alternate method and monitor reactor coolant temperature and pressure at least once per hour.

SURVEILLANCE REQUIREMENTS 4.4.9.2 At least one shutdown cooling mode loop of the residual heat removal system or alternate method shall be determined to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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ne RHR shutdown cooling modo loop may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for 0

surveillance testing provided the other loop is OPERABLE and in operation.

The shutdown cooling pump may be removed from 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 provided the other loop is OPERABLE.

1. The shutdown cooling mode loop may be removed from operation during hydrostatic testing.

- 0ne of the two shall have an 0FERABLE associated diesel generator, i

GRAND GULF-UNIT 1 3/4 4-27 Amendment No.70 i

EMERGENCY CORF COOLING SYSTEMS 3/4 5.2 ECCS - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.5.2 At least two** of the following shall be OPERABLE :

l a.

The low pressure core spray (LPCS) system with a flow path capable i

of taking suction from the suppression pool and transferring the water through the spray sparger to the reactor vessel.

b.

Low pressure coolant injection (LPCI) subsystem "A" of the f;HR system with a flow path capable of taking suction from the suppression pool and transferring the water to the reactor vessel.

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c.

Low pressure coolant injection (LPCI) subsystem "B" of the RHR system with a flow path capable of taking suction from the suppression pool and transferring the water to the reactor vessel.

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Low pressure c0olant injection (LPCI) subsystem "C" of the RHR system with a flow path capable of taking suction from the suppression pool and transferring the water to the reactor vessel.

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The high pressure core spray (HPCS) system with a flow path capable l

of taking suction from one of the following water sources and trans-ferring the water through the spray sparger to the reti ir vessel:

1.

From the suppression pool, or 2.

When the suppression pool _ level is less than the limit or is drained, from the condensate storage tank containing at least 170,000 available gallons of water, equivalent to a level of 18 feet.

APPLICABILITY:

OPERATIONAL CONDITION 4 and 5*.

ACTION:

a.

With one of the above required subsystems / systems inoperable, pro-vided an automatic subsystem / system is OPERABLE, restore at least two subsystems / systems to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or suspend all operations that have a potential for draining the reactor vessel.

Otherwise, with no autcmatic subsystem / system OPERABLE, suspend all operations that nave a potential for draining the reactor vessel, b.

With both of the above required subsystems / systems inope able, suspend CORE ALTERATIONS and all operations that have a potential for draining the reactor vessel.

Restore at least one subsystem / system to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or establish SECONDARY CONTAINMENT INTEGRITY within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

OPERATIONAL CONDITION changes per Specifica-tion 3.0.4 are not permitted.

The 'iCCS is not required to be OPERABLE provided that the reactor vessel head is removed, the cavity is flooded, the reactor cavity and transfer canal gates in the upper conta nment pool are removed, and water level is maintained d

within the limits of Specifications 3.9.8 and 3.9.9.

- 0ne of the two required ECCS subsystems / systems shall have an OPERABLE assoc-iated diesel generator.

0ne of the two required ECCS subsystems / systems may require manual realignment prior to initiation and injection.

GRAND GULF-UNIT 1 3/4 5-6 AmendmentNo.70]

C0h...!4 MENT SYSTEMS 3/4.6.4 CONTAINMENT AND ORYWELL ISOLATION VALVES LIMITING CONDITION FOR OPERATION t

I 3.6.4 The containment and drywell isolation valves shown in Table 3.6.4-1 shall be OPERA 8LE with isolation times less than or equal to those shown in Table 3.6.4-1.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, 3, and #.

ACTION:

With one or more of the containment or drywell isolation valves shown in Table 3.6.4-1 inoperable, maint.in at least one isolation valve OPERABLE in each affected penetration t;iat is open and within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either:

a.

Restore the inoperable valve (s) to OPERABLE status, or b.

Isolate each affected penetration by use of at least one deactivated automatic valve secured in the isolated position,* or c.

Isolate each affected penetration by use of at least one closed manual valve or blind flange *,

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .**

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Isolation valves, except MSIVs, closed to satisfy these requirements may be reopened on an intermittent basis under administrative controls.

OPERATIONAL CONDITION changes, as provided by Specification 3.0.4, are not allowed wnile isolation valves are open under these administrative controls.

Isolation valves shown in Table 3.6.4-1 are also required to be OPERABLE when their associated actuation instrumentation is required to be OPERABLE per l

Table 3.3.2-1.

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- Except for E12-F008 and E12-F009 in OPERATIONAL CONDITIONS 4 and 5 take action per Specification 3.3.2 Table 3.3.2-1, Trip Function 6.c.

GRAND GULF-UNIT l' 3/4 6-28 Amendment No. 70

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. REFUELING OPERATIONS 3/4.9.11 RESIDUAL HEAT REMOVAL AND C00LA'JT CIRCULATION HIGH WATER LEVEL i<

LIMITING CONDITION FOR-OPERATION h

3.9.11.1 One of two of the following shall be OPERABLE # and,.unless the alter-E nate decay heat removal system (ADHRS) is in operation, at least one of the following two shal1 be in operation *:

7 c.

residual heat removal (RHR) system shutdown cooling mode train "A",

or b.

RHR system shutdown cooling mode train "B".

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p APPLICABILITY:

OPERATIONAL CONDITION 5, when irradiated fael is in the reactor L

vessel and the water level is greater than or equal to 22 feet 9 inches above 1-the top of the t actor pressure vessel flange.

ACTION:

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a.

With no RHR shutdown cooling mode train OPERABLE, within one hour and at 1 east once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, demonstrate the OPERABILITY of at leas l

one alternate method capable of decay heat removal.. Otherwise, suspend L

all operations-involving an increase in the reactor decay heat load and establish SECONDARY CONTAINMENT INTEGRITY within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b.

With no RHR shutdown cooling mode train in operacion, within one hour establish reactor coolant circulation by an alternate method and monitor reactor coolant temperature at least once per hour.

TRVEILLANCEREQUIREMENTS 4.9.'.1.1 At least one shutdown coc11ng mode train of the residual heat removal l

syster, ADHRS or alternate nethod shall be verified to be in operation and circu i ting reactor coolant at-least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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m The shutdown cooling pump or ADHRS may be removed from operation fe' up to l

2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8-hour period.

The one required shall.have an OPERABLE associated diesel ganerator.

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GRAND GULF-UNIT 1 3/4 9-18 Amendment No. 70 l a

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RXI.LEpNG OPERATIONS-

-LOW WATER LEVEL-LIMITING CONDITION FOR OPERATION

-3.9.11.2 Two of three of the following shall be OPERABLE

,id at least one shall be in operation *:

s a.

residual' heat removal (RHR)' system shutdown cooling mode train "A",

or b.

RHR system shutdown cooling ude train "B", or l

c.

alternate decay heat removal system (ADHRS) in the reactor cooling' mode.

APPL.ILt3ILITY:

OPERATIONAL CONDITION b, when irradiated fuel is in the reactor vessel and the water level is less then 22 feet 8 inches above the top of the.

reactor pressure vessel flange.

(

ACTION:

a.

With less than the above required shutdown cooling mode trains of the RHR r

system OPERABLE, within one hour and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, demonstrate the OPERABILITY cf at least one alternate method capable of l

decay heat removal for each inoperable RHR' shutdown cooling mode train.

-Otherwise, either raisa water level to greater than or equal to 22 feet 8 inches above the top tf the reactor pressure vessel flange within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months t

of discovery of the iaoperable system or alternate or suspend all operations

. involving: an increat-it, the reactor decay heat' load and initiate action within one hour to establish SECONDARY CONTAINMENT INTEGRIry, b.-.

With no RHR shutdown cooling mode train in operation, withir. one hour establish reacter coolant circulation by an titernate method and monitor reactor coolant temperature at least once per hour.

SURVEILLANCE REQUIREMENTS 4.9.11.2 At least one shutdown cooling mode train of the residual heat removal system, ADHRS in the reactor cooling mode or alternate method shall be ykrified l~

.to be.in' operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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SThe shutdevn coding pump or ADHRS may be removed from operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months l

,per 8-hour period.

ne of the'two required shall have an OPERABLE associated diesel generator.

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ORANDGULF-UNIT 1 3/4 9-19 Amendment No.70

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.c 3/4.5 EMERGENCY CORE' COOLING SYSTEM t

' BASES

ECCS-OPERATING and SHUTOOWN (Continued)

The capacity of the system is selected to provide the required core cooling.

The HPCS pump is designed te deliver greater than or equal to 1650/7115 gpm at differential pressures of 1147/200 psid.

Initially, water from the condensate storage tank is used instead of injecting water from the suppression pool into

.the reactor, but no credit is taken in the safety analyses for the condensate storage tank water.

With the HPCS sys e inoperable, adequate core cooling is assured by the OPERABILITY of the red uidant and diversified automatic depressurization system and both the LPCS anc.PCI systems.

In addition, the reactor core isolation cooling (RCIC) system, a system for which no credit is taken in the safety analysis, will automatica11y' provide makeup at reactor operating pressures'on a racctor low water level condition.

The HPCS out-of-service period of 14 days is based on the demonstrated OPERABILITY of redundant and diversified low pres-sure-core cooling systems.

The surveillance reouirements provide adequate assurance that the HPCS system will be OPERABLE when required.

Flow and total developer' head values for surveillance tesi.ing include system losses to ensure design requirements

-are met.

Although all active components are testable and full flow can be demonstrcted by recirculation through a test loop during reactor operation, a complete functional test with reactor vessel injection requires reactor shutdown.

The pump discharge piping is maintained full to prevent water hammer damage and to provide cooling at the earlieat moment.

Upon failure of the HPCS system to function properly after a small break loss-of-coalant accident, the automatic depressurization system (ADS) automati--

L cally causes selected safety relier valves to open, depressurizing the reactor so that flow from the low pressure core coolina systems can enter the core in time to limit fuel cladding temperature to less than 2200'F.

ADS is conserva-tively required to be OPERABLE whenever reactor vessel pressure exceeds 135 psig even though LPCS has incipient flow into the reactor pressure vessel at 295 psid and 7115 gpm rated flow at 128 psid, and LPCI has incipient flow into the reac-tor pressure vessel at 229 psid and 7450 gpm rated flow at'24 psid.

ADS automatically controls eight selected safety-relief valves although the safety analysis only takes credit for seven valves.

It is therefore appropriate to permit one valve to be out-of-service for up to 14 days without materially-reducing' system reliability.

In OPEFATIONAL CONDITIONS 4 and 5 this specification permits one ECCS to

~ be-capable of manual realignment in order to perform its vessel injection func-tion.

The ECCS requiring manual realipment shall be capable of.being realigned from control room panels tvithu 20 minutes.

In' OPERATIONAL CONCS JS 4 and 5, LPCI "C" is prohibited from simultane-ously operating with the alternate decay heat removal system (ADHRS).

GRAND GULF-UNIT 'l B 3/4 5-2 Amendment No. ]_Q

3/4.5 = EMERGENCY' CORE COOLING SYSTEM BASES 3/4.5.3 SUPPRESSION POOL-The supression pool is required to be OPERABLE as'part of the ELCS to ensure that a sufficient supply of water is available to the HPCS, LPCS and LPCI systems-in the' event of a LOCA.

This limit on suppression pool minimum water volume en-sures that sufficient water is available to permit recirculation cooling flow to the core.

The OPERABILITY of the suppression pool in OPERATIONAL CONDITIONS-1 J

2 or 3 is-required by Specification 3.6.3.1.

Repair work might require making the suppression pool-inoperable.

This specification will permit those repairs to be made and at the same time give assurance that the irradiated fuel has an adequate cooling water supply when the suppression pool must be made inoperable, including' draining,- in OPERATIONAL CONDITION 4 or 5.

GRAND GULF-UNIT 1 B 3/4 5-2a Amendment No. 10 l

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cy REFUELING OPERATIONS.

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LBkSES 3/4.9.7 CRANE TRAVEL'- SPENT FUEL AND UPPER CONTAINMENT FUEL STORAGE POOLS-The restriction on movement of loads in e ess of the nominal weight of a fuel assembly over other fuel assemblies in the storage pools ensures that in the event this load is dropped (1) the activity release will-be limited to that contained in a single fuel assembly, and (2) any possible distortion of fuel in the storage racks will not result in a critical array.

This assumption is _-

consistent with the activity release assumed in the safety analyses.

3/4.9.8 and 3/4.9.9 WATER LEVEL - REACTOR VESSEL and WATER LEVEL - SPENT FUEL AND UPPER CONTAINMENT FUEL STORAGE POOLS The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly.

This minimum water depth is consistent with the assumptions of the accident analysis.

3/4.9.10 CONTROL ROD REMOVAL These specifications ensure that maintenance or repair of control rods or -

control rod drives will be performet. Wer conditions that limit the probability of.it

'rtent criticality.

The requirements for simultaneous removal of more-thar

ontrol rod are more stringent since the SHUTDOWN MARGIN specification pro for the core to remain subcritical with only one control rod fully withor6wn.

3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION An OPERABLE residual heat removal (RHR) system shutdown cooling mode train consists of at least one OPERABLE RHR pump'and one OPERABLE RHR heat exchanger

. train.

The requirement that at~least one residual heat removal loop be OPERABLE and in operation or that an alternate method capable of decay heat removal be demonstrated and that an alternate method of coolant mixing be in operation

~

-ensures that (1) sufficient cooling capacity is available to remove. decay heat and maintain the water in the reactor pressure vessel below 140*F as required during REFUELING, and (2) sufficient coolant circulation would be availabl.e Y

through the reactor core to assure accurate temperature indication and to distribute and prevent stratification of the poison in the event it becomes necessary to actuate the standby liquid control system.

The requirement to have two shutdown cooling mode loops OPERA 3LE when there

-is-less than 22 feet 8 inches of water above the reactor vessel flange ensures that a single failure of the operating loop will not result in a complete loss of residual heat removal capability. With the reactor vessel head removed and 22 feet 8 inches of. water above the reactor vessel flange, a large heat sink is available for core cooling.

Thus, in the event a failure of the operating RHR loop, adequate time is provided to initiate alternate methods capable of decay heat removai or emergency procedures to cool the core.

GRAND GULF-UNIT 1 B 3/4 9-2 Amendment No. 20

REFUELING OPERATIONS BASES 3/4.9.11-RESIDUAL HEAT REMOVAL AND COOLANT-CIRCULATION (continued).

The alternate decay heat removal system (ADHRS) is designed to provide decay heat removal via the plant service water system.

ADHRS is capable of maintaining reacter coolant temperatures below technical specification limits during REFUELING operations.

For specification 3.9.11.2 additional require-ments are imposed during ADHRS operation since ADHRS is not designed as a safety-related system ard has no onsite power supply capability.

An OPERABLE ADHRS in the reactor cooling mode consists of two OPERABLE ADHR5 pumps taking suction from reactor recirculation system loop "B", passing the water through-two OPERABLE ADHRS heat exchangers and returning the water to the reactor vessel via the low pressure coolant injection "C" injection line.

In OPERATIONAL' CONDITION 5, simultaneous operation'of the ADHRS in the reactor cooling mode and RHR system shutdown cooling mode trains "A" and "B" is prohibited for certain alignments of these systems.

3/4.9.-12 HORIZONTAL' FUEL TRANSFER SYSTE.i The purpose of the horizontal fuel transfer system specification is to control personnel access to those potentially high radiation areas immediately adjacent to the system and to assure safe operation of the Lystem.

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i GRAND GULF-UNIT 1 B 3/4 9-2a Amendment No. 70

ML20028G960

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