Text

Amend 97 to License NPF-29,revising TS & Associated Bases to Extend STIs & AOTs for Instrumentation Supporting ECCS, Control Rod Block Function & Isolation Actuation Instrumentation
	ML20097A659
Person / Time
Site:	Grand Gulf
Issue date:	05/20/1992
From:	Larkins J; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20097A662	List: ML20097A659; ML20097A679;
References
	NUDOCS 9206030090
	Download: ML20097A659 (33)
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.E UNITED STATES s

f NUCLEAR REGULATORY COMMISSION

/

s 'ASHINGTON, D.C, MM JTERGY OPERATInS. INC.

SYSTEM ENERGY RESOURCES. INC_t SOUTH MISSISSIPPI ELECTRIC POWER ASSOCIATI MIS $1SSIPPI POWER AND LIGHT COMPANY

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DOCKET NO. 50-41_6 GRAND CULF NUCLEAR STATION. UNIT l_

AMEt'OMENT TO FACILITY OPERATING LICENSE

(

Amendment No. 97 License No. NPF-29 1.

The Nuclear Regulatory Commission (the Commission) ha s found that:

A.

The application for amendment by Entergy Operations licensee) dated June 26, 1991,

, Inc. (the Act of 1954, as amended (the Act), and the Comm as supplemented April 22, 1992, regulations set forth in 10 CFR Chapter I; on's rules and nergy B.

provisions cf t' ' Act, and the rules and regulatiTh cation, the Commission; ons of the C.

this amerdmer,t can be conducted without endan 4

es authorized by conducted in compliance with the Commission's r ng the health and e

D.

The issuance of.this amendment will not be inimical t defense and security or to tne health and safety of th o the common E.

e public; and of the Commission's regulations and all appli R Part 51 been satisfied.

rements have 4,

9206030090 920520 PDR P

ADOCK 05000416 PDR

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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 Soecifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No.

97, 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 V *"".

~~a' John T. Larkins, Director Project Directorate IV-1 Division of Reactor Projects - l!1/IV/V Office of Nuclear Reactor Regulaticri

Attachment:

Changes to the Technical Speci'4 cations Date of Issuance:

May 20, 1992 I

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,s ATTACHMENT *TO LICENSE AMENDMENT NO. 97 FACIL11Y OPERATING LICENSE NO. NPF-29 DOCKET NO. 50-416 Replace the following pages of the Appendix A Technical Specifications with the attached pages.

The revised pages are identified by amendment number and cont.in vertical lines indicating the area of change. The corresponding overleaf pages are also provided to maintain docur+1nt completeness, i

REMOVE PAGES INSERT PAGES 3/4 1-5 3/4 1-5 3/4 3-9 3/4 3-9 3/A 3 9a 3/4 3-10 3/4 3-10 3/4 3-11 3-11 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 3-22 3/4 3-22 3/4 3-23 3/4 3-23 3/4 3-24 3/4 3-24 3/4 3-25 3/4 3-25 3/4 3-26 3/4 3-26 3/4 3-29 3/4 3-29 l-3/4 3-30 3/4 3-30 3/4 3-34 3/4 3-34 3/4 3-34a 3/4 3-34a 3/4 3-35 3/4 3-35 3/4 3-36 3/4 3-36 3/4 3-55 3/4 3-56 B 3/4 3-1 B 3/4 '-1 B 3/4 3-2 B 3/4-3-2 B 3/4 3-3 B 3/4 3 3 B 3/4,3-3a B 3/4 3-3a B 3/4 3-4 3 3/4 3-4 b

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.t REACTIVITV CONTROL 5(STEMS SURVEILLANCE REQUIREMENTS (Continued) 4.1.3.1.4 The scram discharge velume shall be determined OPERABLE by demonstrating:

a.*

The scram discharge volume drain anc vent valses OPERABLE, when cont '>1 rods are scram tested from a normal control rod configuration of less the or equal to 50% ROD DENSITY at least once per 18 months, by verifying that the d"3in and vent valves:

1.

Close within 30 seconds after receipt of a signal for control rods to scram, and 2.

Open when the scram signal is reset.

i b.

Proper level sensor response by performance of a CHANNEL FUNCTIONAL TEST of the scraT4 discharge volume scram and control rod block level instrumentation at least once per 92 days.

l i

l-

The presisions of Specification 4.0.4 are not applicable provided this surveillance is perforced at least once per 16 months.

GRAND GULF-bNIT 1 3/4 1-5 Amendment No. 22, 97 9

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REA TIV17v CONTROL SYS7 EMS CONTROL ROD Mar! MUM SCRAM INSERTION TIMES LIMITING CON 01T10N FOR OPERATION i

3.1.3.2 The maximum scram insertion time of each control rod from the fully withdrawn position, based on de energitation of the scram pilot valve solenoids as time zero, shall not exceed the following limits:

Maximum Insertion Times to Netch Pesition (Seconds)

Reactor Vessel Dome Pressure (ptic)*

43 29 13 550 C

FTI 3

1050 0.32 0.E6 1.17 APPL IC A$1L,1)): OPERATIONAL CONDITIONS 1 and 2.

ACTION:

Witr the eaximum scram insertion time of one or more control rods a.

exceeding the maximum scram insertion time limits of Specification 3.1.3.2 as determined by Surveillance Requirement 4.1.3.2.a or b, operation may continue provided that:

~

1.

For all " slo " control rocs, i.e., those which exceed the limits of Specification 3.1.3.2, the individual scram insertion times do not exceed the following limits:

Maximum Insertion Tires to Notch Position (Seconds)

Reacter Vessel Deer Pressurt tr s ic P 43 29 13 550 G

M U

1050 0.35 1.14 2.22 2.

For *f ast* conttc1 reds, i.e., those which satisfy the limits of Specification 3.1.3.2, the average scram insertion times do not exceed the following limits:

Maximum Average Insertion Tiees to Notch Position-(Seconds)

Reactor Vessel Dome Pressure (esic)"

43 29 13 550 O

U M

1050 0.31

0. B4 1.53 3.

The sum of " fast" control rods with individual scram insertion times in excess of the limits of ACTION 4.2 and of "siow" control rods does not exceed 7.

4.

No " slow" control rod, " fast" control red with individual scram insertion time in excess of the lirits of ACTION a.2, or other-wise inoperable control rod occupy adjacent locations in any direction, including the diagonal, to another such control rod.

Otherwise, be in at least HDT SHUTDOWN vithin 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

'f or iET~ uate reacter vessel dome pressure, the scra; time criterie is ce te rni:

'y linear interpolation at each~ notch position, GRAND GULF-UNIT 1 3/4 1-6 t

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INSTRUMENTATION 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2 The isolation actuctico instremvntation channels shown in Table 3.3.2-1 shall be OPERABLE with their trip 5+tpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.2-2 and with ISOLATION SYSTEM RESPONSE TIME as shown in Table 3.3.2-3.

APPLICABILITY:

As chown in Table 3.3.2-1.

ACTION:

a.

With an isolation actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.2-2, declare the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.

b.

With the number of OPERABLE channels less than required by the minimum OPERABLE channels per trip tystem requirement for one trip system:

1.

If placing the inoperable channel (s) in the tripped condition would cause an isolation, the inoperable channe1(s) shall be restored to OPERABLE ttatus within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or the ACTION required by Table 3.3.2-1 for the affected trip function shall be taken.

OR 2.

If placing the inoperable channel (s) in the tripped condition would not cause an isolation, the inoperable thanne1(s) and/or that trip system shall be placed in the tripped condition within a) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for trip functiont common to RPS instrum2ntation#;

and b) 24 hou.'s for trip functions not common to RPS instrumentation #.

c.

With the number of OPERA 3LE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for both trip systems, place at laast one trip system

in the tripped condition within one hour and '

.e the ACTION required by Table 3.3.2-1.

- SURVEILLANCE REOUlamMENTS 4.3.2.1 Eech isolation actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.2.1-1.

4.3.2.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic. operation of i

all channels shall be perfor ad at least once per 18 months.

4.3.2.3 The ISOLATION SYSTEM RESPONSE TIME of each isolation trip function shown in Table 3.3.2-3 shall be demonstrated to be within its limit at least once per l

16 ronths.

Each test shall include at least once channel per trip system such that all channels are tested at least once every N times 18 months, where A is the total number of redundant channels in a specific isolation trip syster GRAND GULF-UNIT 1 3/4 3-9 Amendment No.

ff, 97 L

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INSTRUMEFTATION 3/4.3.2 ISOLATION ACTUA710N INSTRUMENTATION SURVEILLANCE REQUIREMENTS 4.3.2.1 Each isolation actuation instrumentation channel s:all be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.2.1-1, 4.3.2.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months.

4.3.2.3 The ISOLATION SYSTEM RESPONSE TIME of each isolation trip function shown in Table 3.3.2-3 shall be demonstrated to be within its limit at least once per 18 months.

Each test shall include at least one channel per trip system such that all channels are tested at least once every N times 18 months, where N is the total number of redundant channels in a specific 1 solation trip system.

GRAND GULF-UNIT 1 3/4 3-9a Amendment No. 97

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ISOLATION ACTUATION INSTRilMENTATION 5

c7 VALVE GROUPS HINIMUM APPLICABLE g

OPERATED BY OPERABLE CilANNELS OPERATIONAL 3

TRIP filNCTION 5IGNAL (a) PER TRIP SYSTEM (b)

CONDITION ACTION 2.

MAIM STEAM LINE ISOLATION (Continued) f.

Main Steam Line Tunnel Temperature - liigh 1

2 1,2,3 23 q.

Main Steam Line Tunnel A Temp.- High 1

2 1,2,3 23 h.

Manual Initiation

1. 10 2

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SECONDARY CONTAINMENT ISOLATION a.

Reactor Vessel Water Level-Low low, level 2 N.A.(c)(d)(h) 2 1, 2, 3, and #

25 y

b.

Drjwell Pressure - High"*

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Manual Initiation N.A.(c)(d)(h) 2 1,2,3 26

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REACTOR WATER CLEANUP SYSTEM ISULATION a.

A Flow - High 8

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VALVE GROUPS MINIM!!M APPLICABfF 4

OPERATED BY OPERABLE CHANNELS OPERATION L 5

_T_R_IP FUNCTION SIGNAL (a) PER Trap SYSTEM (b)

CONDIT!!n1 ACTION I

5.

REACTOR CORE ISOLATION C00 LING SYSTEM ISOLATION i.

RilR Equipment Room Arrlavant Temperature - High 4

1/ room 1, 2, 3 27 l

j.

RHR Equipment Room a Temp. -

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1/ room 1, 2, 3 27 k.

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1 1,2,3 27 1.

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m.

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1/ room I, 2, 3 28 b.

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1/ room 1, 2, 3 28 S

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I TA$LE 3.3.2-1 (Continued)

ISOLATION ArTUATION INSTRUMENTATION

'~

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

ACTION 21 Close the affected system isolation vaive(s) within one hour or:

In OPERATIONAL CONDITION 1, 2, or 3, be in at least HOT a.

SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTOOWN 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 operations with a potential for draining the reactor vessel ACTION 22 Restore the manual initiation function to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT SHUTOOWN 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 .

ACTION 23 Be in at least STARTUP with the associated isolation valves closed within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least HOT SHUTOOWN 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 .

ACTION 24 Be in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

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

ACTION 26 Restore the manual initiation function to OPERABLE status 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 +he effected system inoperable.

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

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

ACT!0N 29 Close the affected system isolation valves within one hour and declare the affected system or component inoperable or:

In OPERAlIONAL CONDITION 1, 2 or 3 be in at least HOT SHUTDOWN a.

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 /4 hours, b.

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

ACTION 30 Declare the affected SLCS pump inoperable.

ACTION 31 Isolate the shutdown cooling common suction line within one iiour if it is not needed for shutdown cooling or initiate action within one hour to establish SECONDARY CONTAINMENT INTEGRITY, HOTES Wher. handling irradiated fuel in~the prirrsry or secondary containment and during CORE ALTERATIONS and operations with a potential for draining the i

reactor vessel.

The low condenser vacuum MSIV closure may be manually bypassed during reactor SHUTDOWS or for reactor STARTUP when condenser vacuum is below the trip set-l point to allow opening of the MSIVs.

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

i Trip function commom to RPS Instrumentation.

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

A 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.

l GRAND GULF-UNIT 1 3/4 3-14 Amendment No. 70, 97 l

y9' o-

.me,9 y

g y.

,-g

,,.,,.,p

-..7n 7

,,,.7 y

c q_

.s TABLE 3.3.2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION NOTES (Continued)

(b) A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for l

required surveillance without placing the trip system in the tripped con-dition provided at least one other OPERABLE channel in the same trip system is monitoring that parameter.

(c) Also actuates the standby gas treatment system.

(d) Also actuatus the control room emergency filtration system in the isolation f c;eraticn.

endr (e) Two u'.scaie-hi Hi, one upsca'e-hi Hi and one cow > scale, or two c;..ascale signals from the same trip system actuate the trip system and initiate 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 va*ves G33-F001, G33-F004, and G33-F251.

--- ( j ) Actuates the Standby Cas Treatment System and isolates 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-Los 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, E61-f 056, 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 !eip system in the tripped condition provided the diesel generator associatec with the OPERABLE isolation valve is OPERAB E.

l l

l l-GRAND GULF-UNIT 1 3/4 3-15 Amendment hc. 70, 97

+ -,.

-,,c,

...w

.w-,

y-3.,,

,-5

, -. +

m

TABLE 3.3.2-2 ISOLATION ACTUATION INSTRUMEhTATION SETPOINTS 5

c, TRIP FUNCTION

^

TRIP SETPOINT g

1.

PRIMARY CONTAINMENT ISOLATION O

a.

Reacter Vessel Water Level -

Low Low, Level 2 w

1 -41.6 inches *

> -43.8 inches b.

Reactor Vessel Water level-l Low Low, level 2 (ECCS -

-> -41.6 inches *

> -43.8 inches

~

Division 3) c.

Reactor Vessel Water Level-

> -150.3 inches

low Low Low, level 1 (ECCS

> -152.5 inches Division 1 and Division 2) d.

Drywell Presst:re - High

$ 1.23 psig i 1.43 psig Dryse11 Pressure-High (ECCS -

w c.

a Division 1 and Division 2)

~< 1.39 psig

< 1.44 psig

~

m f.

Drywell Pressure-High (ECCS -

< 1.39 psig i 1.44 psig Division 3) g.

Containment and Drywell Ventilation Exhaust Radiation - High High

< 3.6 mR/hr**

1 4.0 mR/hr**

Manual Initiation NA NA 2.

MAIN STEAM LINE ISOLATION a.

Reactor Vessel Water Level -

Low low Low, level 1 1 -150.3 inches *

> -152.5 inches b.

Main Steam line Radiation - High

$ 3.0 x full power

< 3.6 x full power background background c.

Main Steam Line Pressuie - Low 3 849 psig

> 837 psig d.

Main Steam Line Flow - High

$ 169 psid i 176.5 psid c.

Condenser Vacuum - Low 19 inches Ug. Vacuuw.

> 8.7 inches Hg. Vacuum f.

Main Steam Line Tunnel Temperature - High

$ 185'F**

_ 191*F**

~

.s INSTRUMENT AT10N TABLE.3.3.2-3 (Continued)

ISOLATION SYSTEM INSTRUME*"i ATION RESPONSE TIME TRIP FUNCTION RESPONSE TIME (Seconds)#

5.

REACTOR CORE ISOLATION CWLINC SYSTEM ISOLATION a.

RCIC Steam Line Flow - High

<10(af##

b.

RCiC Steam Supply Pressure - Low I 10(as c.

RCIC Turbine Exhaust Diaphragm Pressure - High RA d.

RCIC Equipment Room Ambient Tempe-ature - High NA e.

RC?C Equipment Room a Temp. - H ph NA f.

Hun Steam Line Tunnel Ambient Temp. - High NA g.

Main Steam ;.ine Tunnel A Temp. - High NA h.

Main $5eam Line Tunnel Temperature Timer 1A 1.

RHR Equipment Room Ambient Temperature - High NA j.

RHR Eouipment Room a Temp. - High NA k.

RHR/RCIC Steam Line Flow - High NA 1.

Manual Initiation HA m.

Dry. ell Pressure - High (ECCS Division 1

< 10(a) and Division 2) 6.

RHR SYSTEM 150Lt.T10N a.

RHR Equipment Room Ambient ~:'perature - High NA b.

RHR Equipment Room o Temp. - High NA c.

Reactor Vessel Water Level - Low, Level 3

< 10

~

d.

Reactor Vessel (RHR Cut-in Permissive)

Pressure - High NA e.

Dry. ell Pressure - High NA f.

Manual Initiation NA (a) ine isolation syster instrumentation response time shall be measured and recorded as a part of the ISOLATION SYSTEM RESPONSE TIME.

Isolation syster instrumentation response time specifitd includes the delay for diesel generator starting a55umed in the accident analysis.

(b) Radiation detectors are exempt from response time testing.

Response time shall be measured from detector output or the input of the first electronic component in the channel.

Isolation system instrumentation rasponse time for MSIVs only.

No diesel generator delays assumed.

- lsolation system instrumentation response time for associated valves except MSIVs.

- - 1 solation system instrumentation response time for air operated dampers.

No diesel generator delays assumed.

Isolation system instrumentation response time specified for the Trip Function actuating each valve group shall be added to isolation time shown in Tables 3.6.4 1 and 3.6.6.2-1 for valves in each salve group to obtain ISOLATION SYSTEM RESPONSE TIME for each valve.

1. 1. 1ncludx time delay of 3 to 7 seconds.

CRAC CULF-UNIT 1 3/4 3-21

o 2

TABLE 4.3.2.1-1 ISOLATION ACTUATION INSTRUMENTATION SURVEILLAieCE REQUIREMENTS a

I E

CHANNEL OPERATIONAL j

5 CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH i

2; TRIP FUNCTION CHECK TEST CALIBRATION S'fRVEILLANCE REQUIRED

~

1.

PRIMARY CONTAINMENT ISOLATION a.

Reactor Vessel Water Leus:1 -

i Low low Level 2 S

Q R

1, 2, 3 and #

b.

Reactor Vessel Water Level-5 Q

R 1, 2, 3 and #

i Low Low, tevel 2 (ECCS -

Divisian 3) i c.

Reactor Vessel Water Level-5 Q

R(c) 1, 2, 3 and #

low low low, level 1 (ECCS -

Division I and Division 2) d.

Drywell Pressure - High 5

Q R

1, 2, 3 w

j k

e.

Drywell Pressure-High (CCCS -

S Q

R 1,2.3 Division I and Division 2) w f.

Dryweil Pressure-High (ECCS -

S Q

R(c)

1. /, 3 Division 3)

N i

g.

Containment and Dryerell j

Yentilation Exhaust Radiation - High High 5

Q(a)

A 1, 2, 3 and

h.

Manual Initiation NA Q

NA 1, 2, 3 and *#

I l

2.

MAIN STEAM LINE ISOLATION j

a.

Reactor Vessel Water Level -

l Low low Low, level 1 3

0 p(c) 2, 3

~

b.

Main Steam Lis.e Radiatiori -

High 5

Q R

1,2,3 j'

c.

Main Steam Line Pressure -

l n

Low 5

Q R(c) l 1

k d.

Main Steam Line Flow - High 5

Q R

1, 2, 3 l

g e.

Condenser Vacuum - Low Q

R 1,

2**, 3**

S F

i 8

i i

~

4

o E

TABLE 4.3.2.1-1 (Continued)

ISOLATION ACTl!ATION INSTRtIMENTATION SURVEILLANCE REQUIREMENTS T

CHANNEL OPERATIONAL SE CilANNEL FUNCTIONAL CllANNEL CONDITIONS IN WHICH

=

TRIP FUNCTION CllECK TEST CALIBRATIJN SURVEILLANCE REQUIRED 2.

MAIN STEAM LINE ISOLATION (Continued)

I.

Main Steam Line Tunnel Temperature - High Q

A 1,2,3 9

Main Steam Line Tunnel i

A Temp. - High S

Q(a)

A 1,2,3 h.

Manual Initiation

!!A Q

NA 1,2,3 3.

SECONDARY CONTAINMENT ISOLATION a.

Reactor Vessel Water e

level - Low low, level 2 5

Q R

1, 2, 3 and #

w A

b.

Drywell Pressure - High 5

Q R

1,2,3 i

4:

c.

Fuel Handling Area Ventilation m

J, Exhaust Radiation - High High 5 Q

A 1, 2, 3 and

w d.

Fuel riandling Area Pool Sweep l

A 1, 2, 3 and

i Exhaust Radiation - High liigh S Q *)

I e.

Manual Initiation NA Q

NA 1, 2, 3 and

l 4.

REACTOR WATER CLEANUP SYSTEM ISOLATION a.

A Flow - High 5

Q R

1,2,3 b.

A Flo-Timer NA Q

Q 1,2,3 c.

Equipment Area Temperature -

High Q

A 1, 2, 3 S

d.

Equipment Area Ventilation A Temp. - High S

Q A

1, 2, 3 e.

Reactor Vessel Water i

k Level - Low low, level 2 5

Q R(c) 1, 2, 3 h

4 3

O

'S j

i

.i -

og i AM F

'.3.2.1-1 (Continued)

I 6

ISOLATION ACTUAL D*: INpMFNTATION ' SURVEILLANCE REQUIREMENTS m

]U CHANNEL E

OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH TRIP FUNCTION CHECK TEST CALIBRATION St!RVEILLANCE REQUIRED w

4.

4 REACTOR WATER CLEANUP SYSTEM ISOLATION (Continued) t.

Ma m Steam Line Tunne: Ambient lemperature - High 5

Q A

q.

Main Steam Line Tunnel 1,2,3 a Temp. - High 5

Q

^

I* 2 -

i h.

3LCS Initiation NA Q(b)

NA 1, 2' 5##

i.

Manual Initiation NA Q

NA 1,2,3 S.

REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION i

D RCIC Steam Line How - High w

a.

1.

Pressure S

Q RIC) 1, 2, 3 -

l y

2 Time Delay NA Q

Q 1,2,3 j

b.

RCIC Steam Supply Pressure -

1 Low S

Q RfCI RCIC Tu.-bine Exhaust Diaphragm 1, 2, 3 I

c.

Pressure - High S

Q RfC) 1, 2, 3 d.

RCIC Equipment Room Ambient Temperature - High 5

Q A

RCIC Equipment Room a Temp. -

1, 2, 3 i

e.

High S

Q A

1, 2, 3 f.

Main Steam Line Tunnel Ambient Temperature - High S

Q A

1, i, 3 l

E q.

Main Steam Line Tunnel 3

a Temp. - High S

Q A

1, 2, 3 S

u 5

I i

h TABLE 4.3.2.1-1 (Continued) 5 ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS L

E' f

r; CHANNEL OPERATIONAL a

CHANNEL CUNCTIONAL.

CHANNEL CONDITIONS IN MIICH

_z TRIP FUNCTION CHECK TFST.

CALIBRATION SURVEILLANCE REQUIRED

[

S.

REACTOR CORE ISOLATION C00 TING SYSTEM ISOLATION (Continued) f h.

Main Steam Line Tunnel Temperature Timer NA Q

Q 1, 2, 3 i.

RHR Equipment Room Ambient Temperature - High 5

Q A

1,2,3 j.

RHR Equipment Room A Te-p. -

High S

Q A

1, 2, 3 l

w k.

RHR/RCIC Steam Line Flow -

IC) 1 High S

Q R

1, 2, 3 I

w Q(a)

NA 1,2,3 L

E 1.

Manual Initiation NA

?

Drywell Pressure-High 5

Q R

1, 2, 3 f

IC) m.

(ECCS Division 1 and Division 2)

I 6.

RHR SYSTEM ISOLATION a.

RHR Equipmant Room Ambient 4

3-Temperature - High 5

Q A

1 ;, 3 b.

RHR Equipment Room.

j y

a Temp. - High S

Q A

1, 2, 3 S

c.

Reactor Vessel Water Level -

,\\C)

Low, level 3 5

Q R

1,2,3,4,5 i

z O

4 1

~

d.

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

Q R(c 1, 2, 3 l

1.

w i

1

l og TABl.E 4.3.Z_1-1 (Continued) 5 ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS E

CHANNEL OPERATIONAL 4

CHANNEL-FUNCTIONAL CHANNEL CONDITIONS IN WHICH z

TRIP IUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED d

6.

RHR SYSTEM ISOLATION (Continued)

E Drywell Pres ure - High S

Q b

g 1,2,3 i

f Manual initiation NA Q,)

NA 1, 2, 3

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

- The low condenser vacuum MSIV closure may be manually bvpassed during reactor SHUTDOWN or for reactor STARTUP when condenser vacuum is below the trip setpoint to allow opening of the MSlvs.

I U

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

During CORE ALTERATION and operations with a putential for draining the reactor vessel.

T

1. With any control rod withdrawn.

or 3.9.10.2.

Not applicable to control rods removed per Specification 3.9.10.1 (a) Manual initiation switches shall be tested at least once per 18 months dueing shutdowm All other circuitry associated with manual initiation shall receive a CHANNEL FUNCTIONAL TEST at least once per 92 days as part of circuitry required to be tested for automatic system isolation.

I (b) Each train or logic channel shall be tested at least every other 92 days.

[

(c) Calibrate trip unit at least once per 92 days.

l l

I I

i l

i i

N i

?e n

l

T

'S t

TABLE 3.3.3-1 (Continued)

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION e,

x MINIMUM OPERABLE APPLICsBLE CHANNELS PER OPERATIONAL g

TRJP FUNCTION TRIP FUNCTION (#)

CONDITIONS ACTION T

C.

DIVISION 3 TRIP SYSTEM C

25 1.

HPCS SYSTEM a.

Reactor Vessel Water level - Low, Low, Level 2 4

1, 2, 3, 4*, 5*

33

~

b.

Drywell Pressure - High##

4 1, 2, 3 33

)

c.

Reactor Vessel Water Level-11igh, tevel 8 2(d) 1, 2, 3, 4*, 5*

31 d.

Condensate Storage Tank Level-tow 2

I, 2, 3, 4*, 5*

34 Id) e.

Suppression Pool Water Level-High 2

1, 2, 3, 4*, 5*

34 f.

Manual Initiation ##

1 1, 2, 3, 4 *, 5*

32 D.

LOSS OF POWER i

1.

Division 1 and 2 3.

4.16 kV Bus Undervoltaqa 4

1, 2, 3, 4**, 5**

30 '

(Loss of Voltage) t' b.

Deleted

[

L c.

4.16 kV Bus Undervoltage 4

1, 2, 3, 4**, 5**

30 Y

(Degraded Voltage) f m

'o 2.

Division 3 4

a.

4.lf kV Bus Undervoltage 4

1, 2, 3, 4**, 5**

30 (Loss of Voltage)

(

b.

4.16 kV Bus Undervoltage 4

1, 2, 3, 4**, 5**

30 j

(Oegraded Voltage) l (a) A channel may be placed in an inoperabie status for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months during periods of required surveillance l

r l

g without placing the M p system in the tripped condition provided at least one other OPERABLE channel in g

the same trip system is monitoring that parameter.

cy (b) Also actuates the associated division diesel ger.erator.

g (c) Provides signal to close HPCS pump discharge valve only.

(d) Provides signal to HPCS pump suction valves only.

f y

Applicable when the system is required to be OPERABLE per Specification 3.5.2 or 3.5.3.

Required when applicable ESF equipment is required to be OPERABLE.

t N

Not required to be OPERABLE when reactor e am dome p-ssure is less than or equal to 135 psig.

1ho injection function of Drywell Pressur - High and Manual Initiation are not required to be

!=,

OPERABLE with indicated reactor vessel water level on the wide range instrument greater than level 8 setpoint coincident with the reactor pressure less than 600 psig.

- ~ ~

c e.

_ INSTRUMENTATION TABLE 3.3.3-1 (Continued)

EMERGENCYCOREC00gNGSYSTEMACTUATIONINSTRUMENTATION ACTION ACTION 30 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement:

With one channel inoperable, place the inoperable channel a.

in the tripped condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or declare the l

associated system (s) inoperable.

b, With more than one channel inoperable, declare the associated system (s) inoperable.

ACTION 31 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months declare the associated ADS trip system or ECCS l

inoperable.

ACTION 32 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, restore the inoperable channel to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months l

l or declare the associated ADS trip system or ECCS inoperable.

ACTION 33 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, place the inoperable channel (s) in the tripped condition within 24

. hours or declare the HPCS system inoperable.

ACT*0N 34 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels por Trip Function recuirement, place at least one inoperable channel in the tripped condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or declare the HPCS system inoperable.

l l

ACTION 35 -

With the nucer of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, place the inoperable channel (s) in the tripped condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or declare the associated system (s) inoperable.

l l

GRAND GULF-UNIT 1 3/4 3-30 Amendment No. Ei, 97

l i

i g

TABtE 4.3.3.1-1 EMERGENCY CORE COOLING SYSTEH ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS h

CHANNEL OPERATIONAL 7

CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH s

TRIP FUNCTION CHECK TEST

[ALIBRATION SURVEILLANCE REQUIRED A.

DIVISION'I TRIP SYS.'IM 1.

RHR-A (LPCI MODE) AND LPCS SYSTEM a.

Re'ctor Vessel Water level -

Low low Low, level 1 5

Q R

1, 2, 3, 4*, 5*

b.

Drywell Pressure - Ilich 5

Q R

1,2,3 c.

LPCI Pump A Start Time Delay Relay NA Q(b)

Q 1, 2, 3 4*, 5*

d.

Manual Initiation NA R

NA 1, 2, 1 4*, 5*

e.

Reactor Vessel Pressure -

R *)

1, 2, 3, 4*, 5*

I Low (Injection Permissive) S Q

{

2.

AUTOMATIC DEPRESSUkfZATION SYSTEM i

IRIP SYSIEM "A"#

a.

Reactor Vessel Water Level -

Low Low Low, levei 1 5

Q R(a)

1. 2, 3 b.

Drywell Pressure-High S

Q R(a) 1, 2, 3 c.

ADS Initiation Timer NA Q

Q 1, 2, 3 d.

Reactor Vessel Water Level -

Low, level 3 5

Q R(a) 1, 2, 3 e.

LPCS Pump Discharge P'ressure-High S

Q R(a) 1, 2, 3 f.

LPCI' Pump A Discharge a)

Pressu e-High S

Q(b)

R 1, 2, 3 i

g.

Marual Initiation NA R

NA 1, 2, 3 S

h.

ADS Bypass Timer (High Drywell Pressure)

NA Q

Q 1,2,3 l

M i.

Manual Inhibit NA R

NA 1, 2, 3

~

2 i

l

L I

.og TABLE 4.3.'3.1'1 (Continued)

3 i

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE I.

(

T-CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICil g

TRIP TUNCTION.

CHECK TEST

CALIBRATION SURVEILLANCE REQUIRED

-4 B.

DIVISION 2 TRIP SYSTEM w

t 4.

1.

RHR B AND C (LPCI MODE) i i

i

'a.

Reactor Vessel Water level -

t Low Low Low, Level 1 5

Q R

1, 2, 3, 4*, 5*

{

b.

Drywell Pressure - High 5

Q R

1, 2, 3 c.

LPCI Pimp B Start Time

[

Delay' Relay NA l

Q(g d.

Manual Initiation NA R -

Q 1, 2, 3, 4*, 5*

i HA 1, 2, 3, 4*, 5*

i e.

Reactor Vessel' Pressure -

L Low (Injection Permissive) S 0

IR ")

1, 2, 3, 4*, 5*

l, y

Z.,

1-E' 3

k 5

i a

i i

g i

=

+

y 8

f 1

5 i

h.

S TABLE 4.3.3.1-1 (Continued)

EMERGENCY CORE COOLING SYSTEM ACIDAI10N INSTRUMENTATION SURFfjlLLANCE REQUIREMENTS I

CilANNEL OPERATIONAL o'E CilANNEL FUNCTIOrlAL CHANNEL CONDITIONS FOR WHICH TRIP 1 UNCTION CliECK TEST CALIBRATION SURVEILLANLE REQUIRED

[

z B.

_ DIVISION 2 TRIP SYSTEM (Continue <f) 2.

AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "B"#

1 a.

Reactor Vessel Water Level -

Rfa 1, 2, 3 Low Low Low, level 1 5

Q b.

Drywell Pressure-liigh 5

Q R

1, 2, 3 c.

ADS Initiation Timer NA Q

Q 1, 2, 3

~

g i

d.

Reactor Vessel Water Level -

R(a) 1, 2, 3 I

Low, level 3 5

Q l

i w1 e.

LPCI Pump B and C Discharge 9) 1, 2, 3 R

Pressure-High 5

Q(b) w f.

Manual Initiation NA R

NA 1,2,3 l

w g.

ADS Bypass Timer l

(High Drywell Pressure)

NA Q

Q 1, 2, 3 h.

Manual Inhibit NA R

NA 1,2,3 C.

DIVISION 3 TRIP SYSTEM I.

HPCS SYSTEM a.

Reactor Vessel Water Level -

R('))

1, 2, 3, 4*, 5*

f Low Low, level 2 5

Q a

R ')

1, 2, 3 M

b.

Drywell Pressure-High##

5 Q

I S

c.

Reactor Vessel Water S

Q R

1, 2, 3, 4*, 5*

M Level-High, level 8 E

d.

Condensate Storage Tank R(a) 1, 2, 3, 4*, S*

Level - Low S

Q

r e.

Suppression Pool Water R(a) 1, 2, 3, 4*, 5*

level - High 5

Q

)

f.

Manual Initiation ##

NA R

NA 1, 2, 3, 4 *, 5*

m

~

"3 i

- -. ~.

e.

O 5-.ee TD W

3 O' J

W W

m m

m ett W M to EO.

H QwW 2

>==

u 4

W HME W

et v

tir m.

att E eQ:

W cc o J

=

e e

E W *=*==8 M

M M

M CL M a==

8' Q ow W e

o e

e N

N N

N uJ CC K

D e

o e

o W

M M

emd e=J M

Z W

C J===

CC WW CC CC CC E3:

D 2 c<e E

M

< CD E

Z

^O u.J

U

e 4 >=

U D elC C*

ee= E oa wk Ow J

u C4 J<

v >=

L;Z M

ZQ W Z Z=H m

m a

=

et: >== M 0

0 M

%UW v

a::"

at 2

W Z >-

2*

E E

z M

Q D

e

=

w M

e <

9er 3 W J W <

CD a

< 2*

LJM

-s ZV eC" n't Q

gg Z

Z Z

Z

=W M

W I

2 a

m J

e.)

e*

eJ C

w m

e O

C O

O O

W

>^

>m Le 6 @

1. m b 0 W

4 c-0 Ch

&# @ 0 Ch 2

V C:n tG "O CMV W O

C to C ed C N C ea u

D *J De D eJ D m O

m o

N eO m>

mom >

W 2>

3 D>D E

t EC ED T JD SC W

C w

pa w

e m

> C *e d,Y

> O>

Le M

G.m

.M Mn kJ Pa4 m eJ t-M M

L E

WM @ @ Clh US M W O'l 6J CC C

e=4 O-M 0 C

M C r=1 0 C

eJ G eO O

eJ eO WvQWv

e=

wVWv w

m G

.e=

ee-Z Q

u,.

e

. e

- =

s e

Q C

5

.O U Q

.O

,_==

u m

Z m

b c

e e

1 w

J M

N A

.ec e

l GRAO GULF UN:T 1 3/4 3-35a Acendment Noe 2f,85

v TABLE 4.3.3.1-1 (Continued)

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENi' NOTATION f

Not required to be OPERABLE when reactor steam dome pressure is less than or equal to 135 psig.

The injection function of Drywell Pressure - High and Manual :nitiation 3

are not required to be OPERABLE with indicated reactor vessel-water level on the wide range instrument greater than Level 8 setpoint coincident with the reactor pres-sure less than 600 psig.

Applicable when the system is required to be OPERABLE per Specification 3.5.2 or 3.5.3.

Required when'-ESF equipment is required to be OPERABLE.

(a) Calibrate trip unit at least once per 92 days.

l (b) Manual initiation switches shall be tested at least once per 18 months during shutdown. All other circuitry associated with manual initiation shall receive a CHANNEL FUNCTIONAL TEST at least or.co per 92 days as a j

part of circuitry required to be tested for automatic system actuation.

(c) DELETED-(d) DELETED (e) _ Functional Testing of_ Time Delay Not Required t

GRAND GULF-UNIT 1 3/4 3-36 Amendment No. 21, 97

jg TABLE 4.3.6-1 5

CONTROL ROD Bl.0CK INSTRUMENTATION SURVEILLANCE REQUIREMENTS a

i E

1 T

CHANNEL OPERATIONAL E

CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH j

TRIP FUNCTION CllECK TEST.

CALIBRATION (a)

SURVEILLANCE REQUIRED 1.

ROD PATIERN CONTROL SYSTEM

}

a.

Low Power Sotpoint NA S/U

,Q Q

1, 2

}

b.

liigh Power Sotpoint NA S/U

,Q Q

1**

{

2 APRM I

a f low fliased Neutron Flux-i lipscale NA Q

W( }(9), SA 1

l l

b.

Inoperative NA S/ti,Q Nh) 1, 2, 5 c.

Downscale NA Q

W

, SA 1

l d.

Neutron Flux - Upscale, Startup NA S/U

,Q Q

2, 5 l

l g

m I

A 3.

SOURCE RANGE MONITORS' Y

a.

Detector not full in NA S/U,W NA 2,.

(

b.

Upscale NA S/U,W Q

2, 5 c.

Inoperative NA S/U,W NA 2, 5

(

l d.

Downstale NA S/II,W Q

2, 5 l

4.

INTERMEDIATE RANGE MONITORS a.

Detector not full in NA S/U,W NA 2, 3 b.

Upscale NA S/U,W Q

2, 5 j

c.

Inoperative NA S/II,W NA 2, 5

(

d.

Downscale NA S/U,W Q

2, 5 p

5.

SCRAM DISCilARGE VOLUME

{

a.

Water Level-liigh NA Q

R 1, 2, 55 l

l; M

6.

REACTOR COOLANT SYSTEM RECIRCULATION FLOW

{

a.

"pscale NA Q

Q 1

l t

I c

7.

REACTOR MODE SWITCH SliUIDOWN y

Fusill0N NA R

NA 3, 4

[

.o i

i I

o 3/4.3 INSTRUMENTATION BASES 3/4.3.1 REACTOR PdOTECTION SYSTEM INSTRUMENTATION The reactor protection system automatically initiates a reactor scram to:

a.

Preserve the integrity of the fuel cladding.

b.

Preserve the integrity of the reactor coolant system.

c.

Minimize the energy which must be absorbed following a loss-of-coolant accident, and d.

Prevent inadvertent criticality.

This specification provides the limiting conditions for operation necessary to preserve the ability of the system to perform its intended function even during periods when instrument channels may be out cf service because of main-tenance. When necessary, one channel may bc snaae inoperable for brief intervals to conduct required surveillance.

The reactor protection system is made up of two independent trip systems.

There are usually four channels to monitor each parameter with two channels in each trip system.

The outputs of the channels in a trip system are combined in a logic so that either channel will trip that trip system.

The tripping of both trip systems will produce a reactor scram.

Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with NEDC-30851P, " Technical Specification Improvement Analyses for BWR Reactor Protection System," as approved by the NRC and documented in the SER (letter T. A. Pickens from A. Thadani dated July 15,1987).

The bases for the trip settings of the RPS are discussed in the bases for Specification 2.2.1.

The measurement of response time at the specified frequencies provides assurance that the protective functions associated with each channel are com-pleted within the time limit assumed in the accident analysis.

No credit was taken for those channelt with response times indicated as not applicable.

Response time may be demonstrated by any series of sequential, overlapping or total channel test measurement, provided such tests demonstrate the total channel response time as defined.

Sensor response time verification may be demonstrated by either (1) ir. place, onsite or offsite test measurements, or (2) utilizing replacement sensors with certified response times.

3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION This specification ensures the effectiveness of the instrumentation used to mitigate the consequences of accidents by prescribing the OPERABILITY trip setpoints and response times for isolation of the reactor systems.

When neces-sary, one channel may be inoperable for brief intervals to conduct required surveillance.

Some of the trip settings may have tolerances explicitly stated where both the high and low values are critical and may have a substantial effect on safety.

Negative baroretric pressure fluctuations are accounted for in the trip setpoints and allowable values specified for drywell pressure-high.

The setpoints of other instrumentation, where only the high or low end of the setting have a direct bearing on safety, are established at a level away from the normal operating range to prevent inadvertent actuation of the systems involved.

GRA C GULF-UNIT 1 B 3/4 3-1 Amendment No. H, 97 4

s INSTRUMENTATION BASES ISOLATION ACTUATION INSTRUMENTATION (Continued)

Specified surveil'ance intervals and surveillance and maintenance outage times have been determined in accordance with:

(1) NEDC-30851P-A, Supplement 2, " Technical Specification Improvement Analysis for BWR Isolation Instrumen-tation Common to RPS and ECCS Instrumentation" as approved by the NRC and docu-

~

mented in the NRC Safety Evaluation Repurt (letter to D. N. Grace frorr C. E.

Rossi dated January 6,1989) and (2) NEDC-31677P-A, " Technical Specification j

Improvement Analysis for EWR Isolation Actuation Instrumentatien" es approved i

t, the NRC anc d:cc tntea in tne hr.; Safety Evaluation Report (letter to S. D.

Floyo from C. E. Rossi cated June 16, 1990).

Except for the MSIVs, the safety analysis does not address individual sensor respon>e times or tr.e response times of the logic systems to which the sensors are a nected.

For D.C. operated valves, a 3 second delay is assumed before the starts to move.

For A.C. operated valves, it is assumed that the A.C. power supply is lost and is restored by startup of the emergency diesel generators.

In this event, a time of 10 seconds is assumed before the valve starts to move, in addition to the pipe break, the failure of the D.C. operated valve is assumed; thus the signal dF ay (sensor response) is concurrent with the 10 second diesel startup.

The safety analysis considers an allowable inventory loss in each case which in turn determines the valve speed in conjunction with the 10 second delay.

It follows that checking the valve speeds and the 10 second time for emergency pewer establishmerit will establish the response timt for the isolation functions.

However, ta enhance overall system re,lia-biliv ei to monitor instrument channel response time trends, the isolation actuat i irstrumentation response time hall be measured and recorded as a part of he I:0LATION SYSTEM RESPONSE TIME.

Opero...

with a trip set less conservative than its Trip Setpoint but

'within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or greater than the drift allowance assumed for each trip in the safety analyses.

3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION t

The e~ergency core cooling system actuation instrumentation is provided to initiate actions to mitigate the consequences of accidents that are beyond the ability of the operator to control.

This specification provides the OPERABILITY requirements, trip setpoints and response times that will ensure effectiveness of the systems to provide the design protection.

Negative baro-metric pressure fluctuations are accounted for in the trip setpoints and allow-able values specified (or dryaeli pressure-high.

Although the instruments are listed by system, in some cases the_same instramen' may bs used to send the actuation signal to more than one system at the sa n time.

Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with NEDC-30936P-A, Parts 1 and 2,

" Technical Specification Improvement Methodology (With Demonstration for BWR ECCS Actuation Instrumentation)" as approved by the NRC and documented in the NRC Safety Evaluation Reports (letter to D. N. Grace from A. C. Thadani dated l

December 9, 1988 (Part 1) and letter to D. N. Grace from C. E. Rossi cated December 9, 1988 (Part 2)).

l GRAND GULP UNIT I B 3/4 3-2 Amendment No. U 97 l

l

.-- a

3-INSTRUPDT'710N L

hb ~; _

EMERGEh*3 CORE COOLING SYSTEM ACTUATION INSTRUMENTATION (Continued)

Operation with a trip set less conservative than its Trip setpoint but Ek4 specified Allowable Value is acceptable on the basis that the i between each Trip Setpoint and the Allowable Value 11 equal to or

p..

Se drift allowance assumed for each rip in the W ety analyses.

3; ecVLATION PUMP TRIP ACTUATION INSTRUMENTATION k

cipated transient without scram recirculation pump trip (ATWS-RPT)

/5 V

, des a means of limiting the consequences of the unlikely occurrence of 6

.- to scram during an anticipated transient.

The response of.the plant t..

- postult.ted event has baen evaluated in General Electric Company repcrt NEDs 32408 dated March 1987.

The results of the analysis show that the Grand Gelf AIWS-RPT design provides adocaate protection for these events in

+ici. the normal scram paths fail.

The ATVS-RPT providas fully redundant trip of the recirculation pump motors so that the pumps c.oast down to zero speed.

This trip ionction reduces core flow creating steam voids in the core, thereby decreasing power generation and limiting cny power or pressure excursions.

The Grand Gulf ATWS-RPT design prcvides comp 1bnce with the requirements oi' the NRC FWS Rule r:FR50.62.

The ATWS-RPT and Alternate Rod Insertion (ARI) system use common setpoints and trip channels L re W t*ers and trip systems).

Therefore, the ARI trip function m. the RPl trJ anction will be initiated simultaneously.

The 5

instrumer@ fon setpoints for the RPV pres ure and water level trip channels are estabi: Ahed such that the normal scram paths for these variables would already be initiated.

The end-of-cycle recirculation pump trip (EOC-RPT) system is a part of tne I:eactor Protection System and is an essential safety supplement to the reactor trip.

The purpose of the EOC-RPT is to reccver the loss of thermal margin which occurs at the end-of-cycle.

Tha physica' phenomenon involi ed it that the void reactivity feedback due to a pressurf zation transient can add positive reactivity to the reactor system at a faster rate-than the control rods add esgetive scram reactivity.

Each EOC-RPT system trips both recirculation pumps, ieducing coolant flow in order to reduce the void collapse in the core during

o of the most

' uiting p*essurization eve

.s.

TM ;wo events fx which t:. E?C-RPT pectective l

-feature will function are closure of the turbine stop valves and fast closure l

of the turbine control valves.

A fast closure so : r frso each of two turbine control valves provides input'to the EOC-RPT syste:;

fast closure sensor from each of the other two turbine control valves pros:ces input to the second EOC-RPT system.

Similarly, a closure _senser fer each of two turbine stop valves provides input to one EOC-RPT system; a closure sensor from each of the other two stop valves provides input to the othee EOC-RPT system.

For each EOC-RPT system, the sensor relay contacts are arr&nged to form a 2-out-of-2 logic for the fast closure of turbine control valves and a 2-out-of-2 logic for the turbine stop valves.

The opera-tion of either logic will actuate the EOC-RPT system and trip both recircula-l tion pump.

GRAND GULF-UNIT 1 B 3/4 3-3 Amendment No. 97

INSTRUMENTATION BASES RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION (Continued)

Each EOC-RPT system may be n.:nually bypassed by use of a keyswitch which is admi.11stratively controlled.

The manual bypasses and the automatic Operating Bypass at less than 40% of RATED THERMAL. POWER are annunciated in the control The automatic bypass setpoint is feedwater temperature dependent due to room.

the subcooling changes thst affect the turbine first-stage pressure-reactor power relationship.

For RATED THERMAL POWER operation with feedwater tempera-ture greeter than or equal to 420'F, an allowable setpoint of < 26.9% of control valve wide open turbine first-stage pressure is provided for tee bypass func-tion.

This setpoint is also applicable to operation at less than RATED THERMAL POWER with the correspondingly lower feedwater temperature.

The allowable set-point is reduced to < 22.5% of control valve wide cpen turbine first-stage pres-sure for RATED THERMAL POWER operation with a feedwater temperature between 370 F and 420'F.

Similarly, the reduced setpoint is applicable to operation at less than RATED THERMAL POWER with the corresponding lower feedwater u.a.perature.

The EOC-RPT system response time is the time assumed in the analysis between initiation of valve motion and comDiete suppression of the electric arc, i.e., 190 ms.

Included in this time are:

the response time of the sensor, the response time of the system logic and the breaker interruption time.

Breaker interruption time includes both breaker response time and the manufacturer's design arc suppression time of 12 ms.

Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Tric Setpoint and the Allowable Value is equal to or greater then the drift alicence assumed for each trip in the safety analyses.

3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION The reactor core isolation cooling system actuation instrumentatior. is provided to initiate actions to assure adequate core cooling in the event of reactor isolation from its primary heat sink and the loss of feedwater flow to the reactor vessel without providing actuation of any of the emergency core cooling equipment.

Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on tne basis that the dif ference between each Trip Setpoint and the Miowable Vali e is equal to or greater than the drif t allowance assumed for eech trip in tne safety analyses.

3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION The control rod block functions are provided consistent with the require-ments of the specifications in Section 3/4.1.4, Control Rod Program Controls and Section 3/4.2 Power Distribution Limits.

The trip logic is arranged so that a trip in any one of the inputs wii; results in a control rod block.

The OPERABILITY of the control rod block instrumentation in OPERATIONAL CONDITION 5 is to provide diversity of rod block protection to the one-rod-out interlock.

GRAND GULF-UNIT 1 B 3/4 3-3a Amendment No. 97

INSTRUMENTAT10N BASES 3/4.3.6 - CONTROL ROD BLOCK INSTRUMENTATION (Continued)

Specified surveillance intervals have been determined in accordance with NEDC-30B51P-A, Supplement 1, " Technical Specification Improvement Analysis ft BWR Control Rod Block Instrumentation" as approved by the NRC and documented in the NRC Safety Evaluation Report (letter to D. N. Grace from C. E. Rossi dated September 22, 1988).

Operation with a trip set less conservative than its i. p Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or greater that the drift allowance assumed for each trip in the safety analyses.

3/4.3.7 MONITOPING INSTRUMENTATION

~

'3/4.3.7.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring instrtmentation ensures that:

(1) the radiation levels are continually measured in the araes served by the individual channels; (2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded; and (3) sufficient information is available on selected clant parr'?ters to monitor and assess these variables following an accident.

'nis ca,;bility is consistent with the recommendations of NUREG-0737, " Clarification of TMI Action Plan Requirements," November, 19S0.

'/4.3.7.2 SEISMIC MONITORING INSTRUMENTATION The OPERABILITY lof the seismic monitoring instrumentation ensures that sufficient capability is available tn promptly determine the magnitude of a seismic event and evaluate the response of ttose features important to safety.

This capability is requireo to permit comparison of the measured response to that used in the design basis for the un.

3/4.3.7.3 METEOROLOGICAL MONITORING INSTRUMENTATION The OPERABILITY of the meteorological monitoring instrumentation ensures that sufficient meteorological data are available for estimating potential radiation doses to the public as a result of routine or accidensal release of radioactive materials to the atmosphere.

This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public.

This instrumentation is consistent with the recommenda-nl tions of Regulatory Guide 1.23 "Onsite Meteorological Programs," February, 1972.

3/4.3.7.e REMOTE SHUTDOWN SYSTEM INSTRUMENTATION AND CONTROLS The OPERABILITY of the remote shutdown system instrumentation and controls ensures that sufficient capability is available to permit shutdown and main-tenance of HOT SHJTDOWN of the unit from locations outside of the control room.

This capability is required in the event control room habitability is lost and is consistent with General Design Criterlon 19 of 21 CFR 50.

GRAN: GULF-L.:IT 1 B 3/4 3-4 Amendment No. 41, 97 l

_ _ _ _ _ _ - _ _ _

ML20097A659

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