ML20069L132

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Proposed Tech Specs Re Sealed Source Contamination, Reporting,Fire Protection & Trip Setpoints
ML20069L132
Person / Time
Site: Grand Gulf Entergy icon.png
Issue date: 04/25/1983
From:
MISSISSIPPI POWER & LIGHT CO.
To:
Shared Package
ML20069L101 List:
References
PCOL-83-06, PCOL-83-6, NUDOCS 8304280049
Download: ML20069L132 (14)
v • d • e


Text

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M PLANT SYSTEMS 3/4.7.5 SEALED SOURCE CONTAMINATION LIMITING CONDITION FOR OPERATION -

3.7.5 Each sealed source containing radioactive material either n excess of 100 microcuries of beta and/or gamma emitting material or icrocuries of alpha emitting material shall be free of greater than or equal to 0.005 microcuries of removable contamination.

APPLICABILITY: At all times. .

ACTION:

a. With a sealed source having removable contamination in excess of the above limit, withdraw the sealed source from use and either:
1. Decontaminate and repair the sealed source, or
2. Dispose of the sealed source in accordance with Commission Regulations.
b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS <

4.7.5.1 Test Requirements - Each sealed source shall be tested for leakage and/or contamination by:

a. The licensee, or
b. Other persons specifically authorized by the Commission or an Agreement State.

l The test method shall have a detection sensitivity of at least 0.005 microcuries l

per test sample.

4.7.5.2 Test Fre_g_uencies - Each category of sealed sources, excluding startup sources and fission detectors previously subjected to core flux, shall be tested at the frequency described below.

a. Sources in use - At least once per six acnths for all sealed sources

~containing radioactive material:

1. With a half-life greater than 30 days, excluding Hydrogen 3, and
2. In any form other than gas.

8304280049 830425 PDR ADOCK 05000416 P PDR GRAND GULF-UNIT 1 3/4 7-26

b- -

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

b. Stored sources not in use - Each sealed source and fission detector

' shall be tested prior to use or transfer to anothef licensee unless tested within the previous six months. Sealed sources and fission detectors transferred without a certificate indicating the last test date shall be tested prior to being placed into use.

c. Startup sources and fission detectors - Each sealed startup source and fission detector shall be tested within 31 days prior to being subjected to core flux or installed in the core and following repair I or maintenance to the source.

4.7.5.3 Reports

- A report shall be prepared and submitted to the Come.ission

' '- '- if sealed source or fission detector leakage tests reveal the presence of greater than or equal to 0.005 microcuries of removable contamination.

W:tk'.w 30 do.y l

l 1

1 1

1 l

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t GRAND GULF-UNIT 1 3/4 7-27

3. hiGNS- 682. )

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS 3

4.7.6.1.1 The fire suppression water system shall be demonstrated CPERABLE:

a.
  • At least once per 7 days by verifying the minimum contained water supply volume,
b. At least once per 31 days by starting the electric motor driven fire suppression pump and operating it for at least 15 minutes.
c. At least once per 31 days by verifying that'ench valve, manual, power operated or automatic, in the flow path is in its correct position.
d. At least once per 12 months by cycling each testable valve in the flow path through at least one complete cycle of full travel.
e. At least once per 18 months by performing a system functional test which includes simulated automatic actuation of the system throughout its operating sequence, and:

?. ";rifying th:t :::5 :et::: tic ::1v: '- th: f?re ;:t' ::t;:te:

it i t; 0 0 r r 00 t ;0: i ti: r.,
l. st'. Verifying that each fire suppression pump develops at least .

1500 gpm at a system head of 275 feet, 2L ,4C Cycling each valve in the flow path that is not testable during plant operation through at least one complete cycle of full travel, and Ji,st'. Verifying that each fire suppression pump starts sequentially to maintain the fire suppression water system pressure greater than or equal to 120 psig. ,

f. At least once per 3 years by performing a flow test of the system in accordance with Chapter 5, Section 11 of the Fire Protection Handbook, 14th Edition, published by the National Fire Protection Association.

4.7.6.1.2 The diesel driven fire suppression pump shall be demonstrated OPERABLE:

a. At least once per 31 days by:
1. Verifying the fuel storage tank contains at least 300 gallons of fuel.
2. Starting the diesel driven pump from ambient conditions and I operating for greater than or equal to 30 minutes.

l t

l CRAND GULF-UNIT 1 3/4 7 29 ,

4,GGN5-Mh l PLANT SYSTEMS I SURVEILLANCE REQUIREMENTS (Continued) i

b. At least once per 92 days by verifying that a sample of diesel.. fuel from the fuel storage tank, obtained in accordance with ASTM-D270-75, is within the acceptable limits specified in Table 1 of ASTM 0975-77 when checked for viscosity, water and sediment.
c. At least once per 18 months, during shutdown, by subjecting the diesel to an inspertion in accordance with procedures prepared in conjunction with its manufacturer's recommendations for the class of service.

4.7.6.1.3 The diesel driven fire pump starting 24-volt battery bank and charger shall be demonstrated OPERABLE:

a. At least once per 7 days by verifying that:

pil.t

1. The electrolyte level of each3cell in each battery is above 2.

'"' QlR-The - . r y cell specific gravity, corrected to 77'F a electrolyte level, is greater than or equal to 1.20,

3. Each battery voltage is greater than or equal to 12 volts, and
4. The overall battery set voltage is greater than or equal to 24 volts. / FcR EAcw ccLL '
b. Atleastonceper92daysbyverifyingthatthespecificgravityfis appropriate for continued service of the battery. .
c. At least once per 18 months by verifying that:
1. The battery case and battery racks show no visual indication of physical damage or abnormal deterioration, and
2. Battery terminal connections are clean, tight, free of corrosion and coated with anti-corrosion material.

GRAND GULF-UNIT 1 3/4 7-30

I. .

PLANT SYSTEMS

5. (G G N 5-152.,2 25,,241)

SURVEILLANCE REQUIREMENTS

(.c. THoM H4 NG EXTgM N 4L (rxcgpi omrtmgNTs4L PMCMumt VA4 VES; p, 4.7.6.3.1 Each of the above required CO2 systems shall be strated OPERABLE at least once per 31 days by verifying that each valv manual, power on.

l epe' rated or automatic, in the flow path is in its correct post 4.7.6.3.2 Each of the above required low pressure CO2 systems shall be demonstrated OPERABLE:

a'. At least once per 7 days by verifying the CO2 storage tank level to be greater than 50% and pressure to be greater than 275 psig, and

b. At least once per 18 months by: 4
1. Verifying that the system valves, ssociated ventilation dampers and electro-thermal linksbet r" automatically upon

r'-'d-'

receipt of a'idP.feclot bl.-thr.

e signal, and

2. Flow from each nozzle by performance of a " Puff Test".

J, Hev. femin3 aNisus.\ iN &&tcdION N etu:.k [ir g do.qch ud a.ssocMed Wha.re.

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l GRAND GULF-UNIT 1 3/4 7-33

PLANT SYSTEMS 5. GG NS - 211,211a.) p,g MALON SYSTEMS LIMITJNG CONDITION FOR OPERATION 3.7.6.4 The following Halon systems shall be OPERABLE with the storage tanks having at least 95% of full charge weight and 90% of full charge pressure: .

a. Control Building, elev. 148'0", Computer and Control Panel Room
b. Control Building, elev.166'0", PGCC 4eem omkr Oew o.rea
c. Control Cabinet Room, elev.189'0", PGCC Jteem ond er &cc seu.

APPLICABILITY: Whenever equipment protected by the Halon systems is required i to be OPERABLE.

ACTION:

a. With one or more of the above required Halon systems inoperable, within one hour establish a continuous fire watch with backup fire suppression equipment for those areas in which redundant systems or components could be damaged; for other areas, establish an hourly fire watch patrol. Restore the system to OPERABLE status within 14 days or, in lieu of any other report required by Specification 6.9.1, prepare and submit a Special Report to the Commission pursuant to Specifica-tion 6.9.2 within the next 30 days outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.
b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS .

4.7.6.4 Each of the above required Halon systems shall be demonstrated OPERABLE:

e. At %est-eace-per-M d:y: by ::-44y%-4Aatwech v:he, ::x:1, power ; retedaw-+demat4A-4Ae f hu p:th i '- it: ::rr::t l

500iti r.

Q, )f.

At least once per 6 months by verifying Halon storage tank weight and pressure.

l At least once per 18 months by:

b [.

Verifying the system and associated ventilation dampers 1.

actuation automatically upon receipt of a staulated-ae4*at4+e +c:p signal, a h yt w. da4. A.

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' 2. Performance of a flow test through headers and nozzles to assure no blockage. y

3. We4m:w3g.,,gw Av:wdc,,tu sped;vJ oI e A (;,, y, ,.
a. S s e c :.A ed -

GRAND GULF-UNIT 1 3/4 7-34

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TA8LE 3.7.6.5-1 FIRE H0SE STATIONS HOSE RACK L6 CATION ELEVATION INDENTIFICATION ,

1 AUXILIARY BUILDING .

I 44-6 Q.j 6.0 103'-0" 13A W 3.5 f.7 119'-0" 138 44 '.r 41 - 61 139'-0" 13C 44-54 -6.0 166'-0" 185'-0" 130 13E Q-3. 5-G -S.1 49-54-6.0 208'-0" 93'-0" 13F 14A t11.1 Q - 21.3 P.4-9.0, t 11.1 119'-0" 14B P.4-9.0,

  • 11.1- 139'-0" 14C P. 0 - 0. 0- R 4.g.6 166'-0" 140

? i-0.0 ft 4-7.5 185'-0" 14E 4,40 P-lo 200' IC" 2 08 '-10" 14F P.4-12.5 139'-0" 15A P.4-12.5 166'-0" 158 P.t-12.5 P.4-13.1 185'-0" 15C F13.e R-13.1 208'-10" ISD W15.1 f4 2-El 103'-0" 16A

-M-15-1 #4.7 -f5.1 119'-0" 168 e 15.1 L.7 -18.1 139'-0" 16C ,

  1. l',.-i L.7 IK1 166'-0" 160 ,

-M-i$.1 1 7- El 185'-0" 16E M1- 208'-10" 16F 015.179.7

" 1;.; M.5 -15.9 103*.o" 17A F13.0 7.4 -118 119'-0" 17B H-13.8 139'-0" 17C H-13.8 166'-0" 170 '

G.4 11 103'-0" ISA 119'-0" 188 s2-1164-11.7 c.4-11 4 4 - 12.1 139'-0" 18C GA41Gi.4 113 166'-0" 180 G.4-7.5 103'-0" 19A GA-h6 4 4 -S 3 119'-0" 198 G.4-7.5 139'-0" 19C GA 6 4 'I*f 166'-0" 190 G;4-6-4 q 6 - 6.+ 103'-0" 20A GA-6.4 g. 6-6.4 119'-0" 208 139"-0" 20C H-6.2 H-6.2 166'-0" 200 103'-0" 21A i' L-6.2

  • 119'-0" 218 L-6.2 ,21C L- 6. 2. 139"-0" 166'-0" 210 L-6. 2 GRAND GULF-UNIT 1 3/4 7-36 -

.- . _ . __--_-_l_.

6. (gaNs- Ga.r. )r.2 TABLE 3.7.6.5-1 (Continued)

FIRE HOSE STATIONS e

HDSE RACK ELEVATION INDENTIFICATION

't.0 CATION CONTAINMENT 120'-10" 22A

  • -9-5 M.7 -7 9 23A d-9-t N . 9 -S.1 135'-4" 161'-10" 238 d-9-5 7 1 -I.i 23C K-9-5 L S-7,2. 184'-6" 208'-10" 23D t-6 7.4 -1.K 24A M-9-5 M.2. -7.L 135'-4" 161'-10" 24B
  • -9 M .B -7.T 24C M-t-t t9 2.-F.L 184'-6" 24D M-9-t # -S.t 208'-10" 25A 1; 12. 2 M.6-11.4 135'-4"
- 11.0 N.1-115 161'-10" 25B 208'-10" 25C M-ti-1 W.5-105 26A J-it-f 71-12.0 135'-4" 268 J-ti-0 7 - 11.6 161'-10" 184'-0" 26C K-if-5 K.2 -23.1 26D J-11.8 208'-10" CONTROL BUILDING 133'-0" 53A 4 J.9-18.S 538 4 4 9 K.1-13 3 111'-0" 54B G-19 G.i -19.+ 111'-0" 54C c

G-19 41-28.4 133'-0" -

54D G-19 G.1- 19 7 148'-0" 54E

-G-49 G.2- 18 S 166'-0" 54F .

G-it 4 1. - 2s.7 189'-0" 55A 449 K.1 fS.$ 148'-0" 55B K K.1 88 9 166'-0" 55D 4-19 K.2 JS S 189'-0" DIESEL GENERATOR BUILDING 66A 4-11 R - 88. G 133'-0" 66B

-ft-9 R -S.4- 133'-0" i

l GRAND GULF-UNIT 1 3/4 7-37 l

TABLE 3.7.6.6-1 YARD FIRE HYDRANTS AND ASSOCIATED HYDRANT HOSE HOUSES LOCATION HYDRANT NUMBER /FIRC Y"ED g h noEE LOO? 003:00LL MU";ER HoV5L Wyn8EJt Ncrth Coord. East Coord. Elevation

" 510.00 10,500.00 100'0" 47/0021

-0,070.00 i^,200.33 H62F 40/0023

';.570.00 10,012.50 if62F '1/0024 9,795.00 9,979.00 133 406'0" WD025/NH DoitE 10,112.50 9,753.92 135 -W6' 0" 4W/D010/NH Do794r 9,886.00 9,758.25 j33+26'0" -H6/D009/NM doz 1 g 9,041.00 " 755.25 125/000S

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on ene nn M.

,,r n"..

emw w

,ein-,n ad dl wwes AV4 4 si.Am AV g er w w . w w 0,071.07 10,542.32 M" 3^/0020 e

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6. P.1.

TABLE 3.3.7.9-1 FIRE DETECTION INSTRLHENTATION NINIMUM INSTRUP8ENTS OPERABLE

  • 20NE(1} NEAT (2) FLAME SM0KEI3)

INSTRLMENT LOCATION

a. Containment Building

- 1. Return Duct Mounted NA 3 NA NA Detectors .

R00M NO. ELEV. ROOM NAME

b. Control Building 1-4 6 "NA 4
1. OC202 111' DIV I SWGR RM 1-4 NA NA 1
2. OC207 111' DIV I BATTERY RM 1-27 1 NA 1
3. OC208 111' DIV II REMOTE

$HUTDOWN PANEL ROOM 1-27 1 NA 1

4. OC208A 111' DIV I REMOTE SHUTDOWN PANEL ROOM 1-5 NA NA 1
5. OC209 111' DIV III BATTERY RM 1-5 4 NA 2
6. OC210 2.11' DIV III SWGR RM 1-6 NA NA 1
7. DC211 111' DIV II BATTERY RM 1-6 7 NA 4
8. OC215 111' DIV II SWGR RM 1-10 NA NA 1
9. OC307 133' ELECTRICAL CHASE 1-10 NA NA l'
10. DC308 133' ELECTRICAL CHASE 1-11 NA NA 13
11. OC302 133' HVAL EQUIP. ROOM ,

1-15 7 MA 10

12. OC402 148' CABLE SPREADING RM -

1-14 13 NA 7

13. OC403 148' COMPUTER R00M 1-15 2 NA 1 .
14. OC407 148' INSTR. NDTOR GEN ROOM
15. OC503 1-18 NA NA 17

) OC504 166', CONTROL ROOM 1-23 12 NA 14

16. OC702 189' CABLE SPREADING RM 6

1-24 4 NA

17. OC703 189' CONTROL CAB. ROOM f 1-23 NA NA 1
18. OC707 189' INSTR MOTOR GEN. RM l

O he fire detection instruments located within the primary containment are not required to be OPERABLE during the performance of Type A Containment Leakage Rate Tests.

(1) Zones apply only to smoke detectors.

(2) Heat detectors provide warning and activation of automatic extinguishing

  • systems.

(3) Smoke detectors provide early warning capability.

(4) Rur therrnoccupleo dich mcnitor- ambient air temperature d provide asch twning capabiliq.

GRAND GULF-UNIT 1 3/4 3-77

,, . , - - - ~ , - - - - - - . ,,_n , - . - - . . , - - ,. - - , - - - . - , , - - , - - - - - , -

. TABLE 3.3.7.9-1 (Continued)

FIRE DETECTION INSTRUNENTATION NINIMUM INSTRMENTS OPERABLE

  • INSTRUMENT LOCATION b NO. ELEV.. ROOM NAME ZONE IA) h) FLANE $NOKE(3)
c. Auxiliary 8uilding (Continued) 2-3 NA NA 1
31. 1A220 119' PIPING PENETRATION RM 2-3 2 NA 2
32. 1A221 119' ELECT.'SWGR RM 2-2 NA NA 18
33. IA222 119' WEST CORRIDOR 2-6 NA NA 2
34. IA301 139' NORTHEAST CORRIDOR NA NA 1 139' SOUTHEAST CORRIDOR 2-6
2. IA302 1 2-6 NA NA
36. 1A303 139' RHR 'A' HX RM 2-6 NA NA 1
37. IA304 139' PIPING PENETRATION RM 2-20 NA NA fM
38. IA305 139' STEAM TUNNEL 2-6 NA NA 1
39. IA306 139' PIPING PENETRATION RM NA 1 139' RHR '8' HX RM 2-6 NA
40. IA307 NA 2 2-6 3
41. 1A308 139' ELECT. PENETRATION RM 2-6 3 NA 2
42. IA309 139' ELECT. PENETRATION RM 3 2-6 NA NA 3 139' SOUTH CORRIDOR 2-19
43. 1A314 13 l 2-6 NA NA
44. 1A316 139' NORTH CORRIDOR 2-5 2 NA 2
45. 1A318 139' ELECT. PENETRATION RM f, 2-5 NA NA 1
46. '1A319 139' RiV INSTR. TEST RM 2-5 2 NA 2
47. 1A320 139' ELECT. PENETRATION RM 2-19 NA NA 3 .
48. 1A321 139' NCC AREA NA NA 4 CENTRIFUGAL CHILLER 2-19
49. 1A322 139' .

AREA .

2-19 NA NA 1

50. 1A323 139' SGTS AREA 2-19 NA NA 1
51. 1A324 139' HVAC EQUIP AREA 2-19 NA NA 1
52. 1A326 139' 5GTS AREA 2-8 NA NA 2 >
53. 1A401 166' NORTHEAST CORRIDOR 2-8 NA NA 1
54. SA402 166' STEAM TUNNEL ROOF 2-8 NA NA 2
55. 1A403 166' SOUTHEAST CORRIDOR 2-8 NA NA -1
56. IA404 166' UNA551GNEO AREA 2-8 NA . NA 1
57. 1A405 166' CNTMT VENT. EQUIP RM 2-8 NA NA 1
58. 1A406 166' CNTMT EXKAUST FILTER AND VENT ROOM GRAND GULF-UNIT 1 3/4 3-79 .

N 2.2 LIMITING SAFETY SYSTEM SETTINGS 7. (GCi NS-450)P.1 SASES

2. 2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS

. The Reactor Protection System instrumentation setpoints specified in Table 2.2.1-1 are the values at which the reactor trips are set for each~ para-meter. The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their Safety Limits during normal operation and design basis anticipated operational occurrences

and to assist in mitigating the consequences of accidents. 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 Trip Setpoint and the Allowable Value is equal to or hes than the drift l allowance assumed for each trip in the safety analyses. g
1. Intermediate Range Monitor, Neutron Flux - High The IRM system consists of 8 chambers, 4 in each of the reactor trip sys-tems. The IRM is a 5 decade 10 range instrument. The trip setpoint of 120 divisions of scale is active in each of the 10 ranges. Thus as the IRM is

! ranged up to accommodate the increase in power level, the trip setpoint is l also ranged up. The IRM instruments provide for overlap with both the APRM and SRM systems.

The most significant source of reactivity changes during the power increase is due to control rod withdrawal. In order to ensure that the IRM provides the required protection, a range of rod withdrawal accidents have been analyzed. The results of these analyses are in Section 15.4 of the FSAR. The most severe case l

involve; an initial condition in which THERMAL POWER is at approximately 1% of l

RATED TPERMAL POWER. Additional conservatism was taken in this analysis by assuming the IRM channel closest to the control rod being withdrawn is bypassed.

The results of this analysis show that the reactor is shutdown and peak power is limited to 21% of RATED THERMAL POWER with the peak fuel enthalpy well below the -

fuel failure threshold criterion of 170 cal /gm. Based on this analysis, the IRM provides protection against local control rod errors and continuous withdr_awal of control rods in sequence and provides backup protection for the APRM.

2. Average Power Range Monitor For operation at low pressure and low flow during STARTUP, the APRM scram setting of 15% of RATED THERMAL POWER provides adequate thermal margin between the setpoint and the Safety Limits. The margin accommodates the anticipated maneuvers associated with power plant startup. Effects of increasing pressure at zero or low void content are minor and cold water from sources available during startup is not much colder than that already in the system. Temperature coefficients are small and control rod patterns are constrained by the RPCS.'

Of all the pgssible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power increase. Because'the flux distribution associated with uniform rod withdrawals does not involve high local peaks and because several rods must be moved to change power by a significant GRAND GULF-UNIT 1 8 2-6 .

l

~ __

INSTRUMENTATION

7. (csss- +5oh.2.

BASES ISOLATION ACTUATION INSTRUMENTATION (continued) the A.C. power supply is lost and is restored by startup of the emergency diesel generators. In this event, a time of 13 seconds is assumed before the valve starts to move. In addition to the pipe break, the failure of the O.C. operated valve is assumed; thus the signal delay (sensor response) is concurrent with the 13 second diesel startup. The safety analysis considers an allowable inventory loss in each case which in turn determines the valve speed in conjunc-tir.n with the 13 second delay. It follows that checking the valve speeds and the 13 second time for emergency power establishment will establish the response time for the isolation functions. However, to enhance overall system relia-bility and to monitor instrument channel response time trends, the isolation actuation instrumentation response time shall be measured and recorded as a part of the ISOLATION SYSTEM RESPONSE TIME.

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 Trip Setpoint and the Allowable Value is equal to or i  % , h than the drift allowance assumed for each trip in the safety analyses.

3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION The emergency core cooling system actuation instrumentation is provided t

l to initiate actions to mitigate the consequences of accidents that are beyond I the ability of the operator to control. This specification provides the

! OPERABILITY requirements, trip setpoints and response times that will ensure effactiveness of the systems to provide the design protection. Although the instruments are listed by system, in some cases the same instrument may be used l

to send the actuation signal to more than one system at the same time.

l 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 Trip Setpoint and the Allowable Value is equal to or pum/ar 4ew than the drift allowance assumed for each trip in the safety analyses. -

l<

3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION The anticipated transient without scram (ATWS) recirculation pump trip system provides a means of limiting the consequences of the unlikely occurrence of a failure to scram during an anticipated transient. The response of the l

plant to this postulated event falls within the envelope of study events in General Electric Company Topical Report NEDO-10349, dated March 1971 and NEDO-24222, dated December 1979, and Section 15.8 Appendix 15A of the FSAR.

The end-of-cycle recirculation pump trip (EOC-RPT) system is a part of the Reactor Protection System and is an essential safety supplement to the reactor trip. The purpose of the EOC-RPT is to recover the loss of thermal margin which occurs at the end-of-cycle. The physical phenomenon involved is that the void reactivity feedback due to a pressurization transient can add positive reactivity to the reactor, system at a faster rate than the control rods add negative scram j

reactivity. Each EOC-RPT system trips both recirculation pumps, redpcing coolant l

' flow in order to reduce the void collapse in the core during two of the most limiting pressurization events. The two events for which the EOC-RPT protective GRAND GULF-UNIT 1 8 3/4 3-2

INSTRUMENTATION 7. GG NS - 450) e.3 BASES RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION (Continued) .e ,

feature will function are closure of the turbine stop valves and fast closure of'the turbine control valves.

A fast closure sensor from each of two turbine control valves provides input to the EOC-RPT system; a fast closure sensor from each of the other two turbine control valves provides input to the second EOC-RPT system. Similarly, a closure sensor for 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 other EOC-RPT system. For each EOC-RPT system, the sepsor relay contacts are arranged 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 operation of either logic will actuate the EOC-RPT system and trip both recirculation pueps.

Each EOC-RPT system may be manually bypassed by use of a keyswitch which is administrative 1y controlled. The manual bypasses and the automatic Operating Bypa,ss at less than 40% of RATED THERMAL POWER are annunciated in the control j room.

! The EOC-RPT system response time is the time assumed in the analysis between initiation of valve motion and complete suppression of the electric arc, i.e., 190~ms, less the time allotted from start of motion of the stop valve or turbine control valve until the sensor relay contact supplying the input to the reactor protection system opens, i.e., 70 ms, and less the time allotted for breaker arc suppression datermined by test, as correlated to manufacturer's test results, i.e., 50 ms, and plant pre-operational test results. ,

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 Trip Setpoint and the Allowable Value is equal to or 9 fw [c c Q u than the drift allowance assumed for each trip in the safety analyses. l 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION The reactor core isolation cooling system actuation instrumentation 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 co_re cooling equipment.

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 Trip Setpoint and the Allowable Value is equal to or 9re ler 4e+s than the drif t allowance assumed for each trip in the safety analyses. l 3_/4.3.6 CONTROL ROD BLOCK W 5TRUMENTATION 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 thatatripinanyoneoftheinputswillresultinacontrolroqblock.

Operation with a trip set less conservative than its Trip 5etpoint 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 yre dce 3e u than the drift allowance assumed for each trip in the safety analyses. l GRAND GULF-UNIT 1 8 3/4 3-3 l

1_ _ _ _ _ _ _ , _ _ . ._ - ._

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