Text

Amend 67 to License NPF-5,revising Allowable Values to Provide for Use of Rosemount,As Well as Barton Transmitters for Certain Instrumentation Channels Associated W/Analog Transmitter Trip Sys
	ML20207B057
Person / Time
Site:	Hatch
Issue date:	11/06/1986
From:	Muller D; Office of Nuclear Reactor Regulation
To:	; City of Dalton, GA, Georgia Power Co, Municipal Electric Authority of Georgia, Oglethorpe Power Corp
Shared Package
ML20207B062	List: ML20207B057; ML20207B063;
References
	TAC 62029 NUDOCS 8611110453
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_ _ _ _ _ - _ _ _ _ _ _ _

/

'o UNITED STATES 8

~g NUCLEAR REGULATORY COMMISSION o

U WASHINGTON. D. C. 20666

~s.,...../

GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET N0. 50-366 EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE l

Amendment No. 67 License No. NPF-5 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment by Georgia Power Company, et al.,

(thelicensee)datedJuly 18, 1986 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 l

Comission; l

C.

There is reasonable assurance (1) that the activities authorized by l

this amendment can be conducted without endangering the health and i

safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and l

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.

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-5 is hereby amended to read as follows:

Nk NooO$66 PDR

i ATTACHMENT TO LICENSE AMENDMENT NO.67 FACILITY OPERATING LICENSE N0. NPF-5 DOCKET NO. 50-366 i

)

\\'

Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendmentinumber and contain vertical lines indicating the area of change. The overleaf pages are provided for convenience.

Remove Insert 2-4

,2-4 3/4 3-2 3/4 3-2 3/4 3-16 3/4 3-16 3/4 3-17 3/4-3-17 4

3/4 3-18 3/4 3-18 3/4 3-26 3/4 3-26 3/4 3-28..

3/4 3-28 3/4 3-29 3/4 3-29 j

3/4 3-31 3/4 3-31 3/4 3-35 3/4 3-35

'1 B 3/4 3-6

.B 3/4 3-6

/

--_r-,.--

_m

. (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amenment No. 67, are hereby incorporated in.the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 60 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Daniel

. Muller, Director BWR Project Directorate #2 Division of BWR Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: November 6,1986 9

1 f

l SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The reactor protection systen instrumentation setpoints shall be set consistent with the, Trip Satpoint values shown in Table 2.2.1-1.

O APPLICABILITY: As shown for each channel in Table 3.3.1-1.

ACTION:

With a reacter protectien system instrumentation setpoint less conservative tr.an the valve shown in the Allowble Values column of Table 2.2.1-1. declare the channel inoperable and apply the applicable ACTION statement requirement cf Specification 3.3.1 until the channel is restored to OPERA 8LE status with its trip setpoint adjusted consistent with the Trip 5etpoint value.

2 o

u t

s jiATCH - UNIT 2 2-3 a.

a w

TABLF 2.2.1-1 5

P.4 RfACTOR PROTECil001 SYSTEM INSTRtEE8tTATIOlt SETPOIIITE s

gn 3 Z s

FUNCTIONAL UNIT TRIP SETPolNT ALLOWA8tf VALUES E

1.

Intermediate Range Monitor, Neutron flux-High 5 120/125 divisions s 120/125 divisions 3

p (2C51-k601 A,B,C,0,E,r,C,II) of full scale or rull scale

?M 2.

Average Power Range Monitor:

j (2C51-K605 A,8,C D,E,r) y l

[

,s.

Neutron Flux-Upscale, 15%

5 15/125 divisions s 20/125 divisloce 1

or rull ssels or full scale m

b.

Flow Referenced Simulated Thornet n (0.58 W + 59%),

1 (0.58 W + 62%),

Powe r-Upsca le with a maximum with a maximum m

s 113.5% or RATED 5 115.5% or RATED N

J THERNAL POWER 1HERMAL POWER c.

Fixed Neutron Flux-Upscale, 118%

s 118% of RATED 5 120% of RATED 7

IllERNAL PCHER TilERMAL POWER 3.

Reactor Vessel Steen Dome Pressure - High 5 1054 psig s 1954 psig.

(2B21-N678 A,8,C,D) 4.

Reactor Vessel Water Level - Low (Level 3) 2 10 Inches above 2 10 laches above l

(?s21-N680 A,e,C,0) instrimment zero*

Instrument zero' Y

5.

Mein Stese I.ine Isolation Velve - Closure 5 10% closed s 10% closed (NA) 6.

Main Steen Line Redletion - High 5 3 x rull power s 3 x rull power j

(2011-M603A,8,C,D) background background 7.

Drywell Pressure - High 5 1.92 psig S 1.92 psig (2C71-N650A,8,C,D) t a

4 a

6 e

i eSee Bases figure 8 3/4 3-1.

i

m. -

- ~sa f

3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION i

I LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the reactor protection system instrumentatien channels shown in Table 3.3.1-1 shall bi OPERAELE with the REACTOR PROTECTION SYSTEM RESPONSE TIME as shown in Table 3.3.1-2.

Set points and interlocks. are given in Table 2.2.1-1.

APPL:CASILITY: As shown in Table 3.3.1-1.

ACTION:

a.

With the requirements for the minimum number of OPERA 3LE channels not satisfied for one trip system, place at least one incperable channel l

in the t-ipDed condition within one hour, g

b.

1ith the requirerents for the minimum number of CFER*BLE channels not satisfied for both trip systems, place at least one inoperable channel in at least one trip system

in the tripped concition within one hour and take the ACTI0t required by Table 3.3.1-1.

c.

The provisions of Specification 3.0.3 are not applicable in OPERA-l TIO!;AL C0!.;31T10N -5.

SURYEILLANCE RE001REMEi:T5 4.3.1.1 Each reactor protection systa= instrumentation channel shall be demenstrated OPEPAELE by the performar.ce of the CHANNEL CHECK, CHA::NEL FUNCTION TEST and CHA!!NEL CALIBRATION operations during the OPER*T10tiAL COND:TIO :S and at the frequencies shown in Table 4. 3.1-1.

4.3.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated auto.atic operation of all channelt shall be performed at least once per 18 months and shall includa calibration of time delay relays and timers necessary for proper functioning of the trip system.

4.3.1.3 The REACTOR PROTECTION SYSTEM RESPONSE TIME of each reactor trip function of Table 3.3.1-2 shall be demonstrated to be within its limit at least once per 18 months.

Each test shall include at least one logic train such that both logic trains are tested at least once per 36 months and.one

. char.nel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundar.t channels in a specific reactor trip function.

f

'If Octn channels are inoperable in ene trip system, select at least one C

ine erable channel in that trip systes to place in the tripped conditiens, exce:t when this could cause the Tric Function to occur.

F HLTCH - UNIT 2 3/4 3 1

'A.cnd er.: :;r.

i TABLE 3.3.1-1

" b$

_ REACTOR PROTECTION SYSTEM INSTRUMENTAftoff e

APPLICABLE MINIMUM NUMBER s

FUNCTIONAL UNIT OPERATIONAL OPLRA8LE CHANNELS "2

CONolflDNS 24 Intermediate Range Monitors:

PER 1 RIP SYSTEM (a)

ACTIDN 1.

o-(2C51-M601, A, B, C, D, E, F, C, H) ha i

sg a.

Neutron Flux - liigh l

1h 2,

5'*'

3 1

b.

Inopera tive 3, 4 2

gg 2,5

2 3

3, 4 1

2 2.

Average Power Range Monitort 2

C3 (2C51-N605 A, 8, C, D, E, F) i i

s.

Neutron flux - Upscale,15%

l.

b.

Flow Referenced Simulated 2, 5 1hermal Power - Upscale 2

1 1

c.

Flxed Neutron Flux -

1 2

Upscale, 118%

3 d.

Ipopera tive 1

2 e.

Downscele 1, 2, 5 3

2 F.

LPRM 1

N 2

3 1, 2, 5 (d)

MA i

3.

Reactor Vesset Steam Dome Pressure -

High (2821-N678 A, 8, C, 0) g, 1,

2

(J. 2821-N045 5

A, 8, C, D)

I*

4.

Reactor Yessel Water Level -

ha low (Level 3) (2B21-N680 A, 8, C, D) 1, 2 1

(J, 2821-N024A, 8 5

1 2821-N025A, 8)

I 5.

Main Steen Line isoletion Valve -

Closure (NA)

I

4 3

6 Main Steam Line Radletion - High 1,

2

(2Dit-k603 A, 8, C, 0) 2 6

7.

Opf t Pressure - High 1,

2

(2C71-N650 A, 8, C, D) 2 5

r

. _....... - ~... ~.... - - '

a.

2-

.........s i

_. ~

~.

TAB E 3.3.2-1 (Continued)

.ISOI.ATION ACTUATION INSTRWrSITATION ACTION ACTION 20 Be in at least HOT SHUIDOW within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and, in COI.D SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

ACTION 21 Be in at least STARTUP with the main steam line isolation valves closed within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at lesst HOT SHUIDOW within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COID SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

ACTION 22 Be in at least STARTUP within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

i ACTION 23 Be in at least STARTUP with the Group 1 isolation valves closed within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or in at least HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 24 Establish SECONDARY CONTAINMENT INIIGRITI with the standby gas treatment system operating within one hour.

ACTION 25 Isolate the reactor water cleanup system.

ACTION 26 Close the'affected system isol.: tion valves and declare the affected system inoperable.

ACTION 27 Verify power availability to the bus at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or close the affected system isolation valves and declare the affected system inoperable.

ACTION 28 Close the shutdown cooling supply and reactor vessel head spray isolation valves unless reactor steam dome pressure < 145 psig.

I i

NOTES Actuates operation 'of the main control room environmental..+ trol system in the pressurization mode of operation.

Actuates the standby gas treatment system.

4en handling irradiated fuel in the secondary containment.

See Specification 3.6.3, Table 3.6.3-1 for valves in each valve group.

a.

b.

A channel may be placed in an inoperable status 'for up 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 condition provided at least one other OPERABE channel in the same trip 3

system is monitoring that parameter.

~'

With a design providing only one channel per trip system, an inoperable c.

channel need not be placed in the tripped condition where this would cause the Trip Function to occur.

In these cases, the inoperable channel shall be restored to OPERABE status within.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or the ACTION required by Table 3.3.2-1 for that Trip Function shall be taken.

(.

}

d.

Trips the mechanical vacuum pumps.

A channel is OPERABE if 2 of 4 instruments in that' channel are OPERABE.

e.

f.

May be bypassed with all turbine stop valves closed.

l 3.-

Closes only RWCU outlet isolation valve 2G31-F004.

h.

Alarm only.

i.

Adjustable up to 60 minutes.

L HATCH - UNIT 2 3/4 3-15 Amendment No. 9, 39 l

i TA8LE 3.3.2-2 ISOLAT10ll ACTUATION INSTRUMENTATIOM SETPolNTS 3@nx 5,

3 TRIP FUNCTION ALLOWA9tr

- c TRIP SETPOINT VALUE 23 1.

PRIMARY CONTAINMENT ISOLATION O H g

e.

Reactor Vessel Water Level M

1.

Low (Level 3) 2 10 inches' 2 10 Inchr:'

2.

Low Low (Level 2) 2 -47 inches' 2 -47 incnes' g

3.

Low Low Low (Level 1) 2 -713 Inches

2 -113 Inches

b.

Drywell Pressure - High 5 1.92 psig 5 1.92 psig c.

Main Steam Line 1.

Radletion - High 5 3 x' Full power background 5 3 x rull power background 2.

Pressure - Low 1 825 dsig 1 825 psig 3.

Flow - High 5 138% rated flow s 130% rated Flow d.

Mein Steen Line Tunnel w

Temperature - High 5 194*F s 194*F e.

Condenser Vacuus - Low k 7" th! vacuum 1 7" Hg vacuum w

/,

r.

Turbine Building Aree Temp.-High 5 200*F S 200*F ch 2.

SECONDARY CONTAINMENT ISO (Afl0N e.

Reactor Buildig Exhaust Redletion - Hlgh 5 60 mr/hr 5 60 nr/hr b.

Dryvell Pres'sure - High 5 1.92 psig s 1.92 psig c.

Reactor Vessel Water Level - Low Low (Level 2) 2 -47 inches

2 -47 inches

d.

Refueling Floor Exhaust Redletion - High 5 20 ar/hr 5 20 nr/hr

'See Bases Figure B 3/4 3-1.

1 e

.TAat r 3. 3_2-2 (Conti;-^ ed t i

3= >r

.lSOLAT1000 ACTUATION INSTRUMENTATION SETPolNTS

@N E'

r g,

TRIP FUNCTION ALLOWASLE TRIP SETPolNT VALUE t<

$z C

3.

REACTOR WATER CLl'ANUP SYSTEM ISOLATION n

z -4 e.

A Flw - High o"

5 79 gpa

$ 79 gpa b.

Area Temperature-High

,5 124'F

$ 124*F w7

. c.

Ares Ventilation A Temperature - liigh 5 67'F s 6T'F Q

d.

SLCS Initiation MA NA e.

Reactor Vessel Water Level-Lw LW 2 -47 leches

2 -47 inches

i (Level 2) l 4.

HICH PRESSURE COOLANT INJECTION SYSTEM ISOLATI'ON s.

HPCI Steam Line Flw-High 5 303% OF rated Flow

$ 303% of" rated Flw b.

HPCI Steam Supply Pressure - Low 2 100 psig 2 100 psig l

HPCI Turbine Exhaust Diephragm c.

Pressure-High d.

HPCI Pipe Penetration Room 5 20 psig 5 20 psig w

Temperature - High 5 169'r 5 169'F N

i e.

Suppression Pool Area Ambient Temperature-High 5 169'F

$ 169'F w

F.

Suppression Pool Area: AT - High

< 42*F 4

g.

Suppression Pool Area Temperature

< 42*F l

w Timer Relays MA h.

Emergency Area Cooler Temperature -

NA High 5 169'r 1

Drywell Pressure - High 5 169'F i

J.

Logic Power sus Monitors 5 1.92 psig 5 1.92 psig l

NA t

NA i

l

'See Bases Figure B 3/4 3-1.

N

_ TABLE 3.3.2-2 (Continued) 3-i' 2$

_ISotATION ACTUATION INSTRUMENTATION SETPOINTS E2

_ TRIP FUNCTION ALLOWA8tt TRIP SETPolNT VALUE c:

3.

RfACfDR CDRf ISOLATION fGOfING SYSIEM ISOLAff0N i

-4 j

p, s.

RCIC Steen Line Flow - High 5 307% or rated flow 5 307% of rated flow =

l b.

RCIC Steam Supply Pressure - Low 2 60 psig 2 60 psIg c.

RCIC Turbine Exhaust Diaphrage Pressure - Illgh 5 20 psig i 20 psig j

d.

Emergency Area Cooler Temperature-High 5 169'r 5 169'F e.

Suppression Pool Area Ambient Temperature 5 169'F s 169'F liigh k

r.

Suppression Pool Area AT - High 5 42'F 5 42*F l

g.

Suppression Pool Area Temperature Timer Relays NA NA h.

Drywell Pressure - High 5 1.92 psig 5 1.92 palg l

sm I.

Logle Power Monitor NA

,~

NA b

6.

SHufbOWN COOllho SYSTEM l50LATION I

03 i

a.

Reactor Vessel Water Level - Low 2 10 inches' 2 10 inches

l l

(Level 3) b.

Reactor Steam Dome Pressure - fil h 5 145 psig 5 145 psig 9

l

\\

'See Basss figure 8 3/4 3-1.

)

Proposed TS/00414/188 W"

f INSTRUMENTATION t

SURVEILLANCE REOUIREMENTS (Continued) 4.3.3.3 The ECCS RESPONSE TIME of each ECCS function shown in Table 3.3.3-3 shall be demonstrated to be within the limit at least once per 18 months.

Each test shall include at least one logic train such,that both logic trains are tested at least once per 36 months and one channel per function 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 ECCS function.

e MTCH - UNIT 2 3/4 3-25

-s-s ew e *~, + ' * * * ' ' ' ' ' ' ~ '

TABLE 3.3.3-1 k g, iifM EMERCENCY CORE C00LINC SYSTEM ACTUATION INSTRUMENTATION UC

@e MINIMUM NUM8ER APPLICABLE TRIP FUNCTION OPERABLE CHANNELS OPERATIONAL

" c-z PER TRIP SYSTEM CONDITIONS setj 1.

CORE SPRAY SYSTEM o

63 2.

Reactor Vessel Water Level - Low Low Low (Level 1) 2 te (2821-N691A,8,C Drywell Pressure,D) 1,2,3,4,5

n b.

High (2E11-N694 A,8,C,0) 2 Reactor Steam Dome Pressure - Low (Injection Permissive) 1,2,3

= c.

ch (2821-N690A,8,C,D)

'd d.

Logic Power Monitor (2E21-K1A,8) 2 1,2,3,4,5 1/ bus

1,2,3,4,5 2.

LOW PRESSURE COOLANT INJECTION MODE OF RifR SYSTEM Drywell Pressure - High (2E11-NC94A,8 a.

Reactor vessel Water Level - Low Low [C.D) 2 b.

1,2,3 ow (Level 1) 2 (2821-N691A,8,C,0) 1,2,3,4*,5*

c.

Reactor Vessel Shroud Level- (Level 0) (Drywell Spray Parmissive) (2821-N685A, 8 l

Reactor Steam Dome Pressure -) Low (I,njection Permissive) 1 d.

1,2,3,4',5*

(2821-N690A,8,C,0) 2 La e.

Reactor Steam Dome Pressure - Low (Recirc. Olscharge Velve 1,2,3,4',5*

];

Peralasive) (2821-N6418,C and 2821-N690E.F) 2 f.

RHR Pump Start - Time Delay Relay 1,2,3,4e,$a La 1)

Pump A (2E11-K70A, 2E11-K1258) 1/ pump 1,2,3,4',5*

d3 2)

Pump 8 (2E11-K708, 2E11-M125A) cn 3)

Pump C (2E11-K758) 4)

Pump D (2 Ell-K75A, 2E11-K126) g.

Logle Power Monitor (2E11-K1A,8) 1/ bus'**

1,2,3,4',5*

\\

Not applicable when two core sprey system subsystems are OPERABLE per Spectrication 3.5.3.1. -

l i

(a) Ale rm only. When inoperable, vertry power eyeliability to the bus et least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months a

or declare the system inoperable, i

7 9

i e

e

.C.

.F\\

.C.-

TABm 3.3.3-1> w...ttrued)

M DEmGENCY 00RE ODOLING SYSUM ACIIRTION INS'mtMNPATION

?

L H

MINIRM NGEER APPLICAB M w

5tIP PUICTIO4 OFERABIE O{Ap4EIS OPERATIONAL IER TRIP SYSTEM CDNDITIONSS

3. HIGH HESSURE (DOLANP INJECTION SYSTEM

'a.

Reactor Virssel Water Ievel - Iow Ist (2821 N692 A,B,C,D)

(Ievel 2) 2 1,2,3 1

- b.

Drywell Presaire - High (2 Ell-N694 A,B,C,D) 2 1, 2, 3 Condensate Storage Tank Invel-Ima (2E41-N002, 2E41-N003) 2(b) (c) c.

d.

Rappression Owsuber Water level-Iligh (2E41-N662B,D) 2(b) (c) 1, 2, 3 1, 2, 3 Iogic Power Monitor (2E41-K])

e.

1(a) f.

Reactor Vessel Water Ietel-High (Imiel 8) (2B2A-N693 B,D) 2 1, 2, 3 1,2,3

.s:

4. AUIOMATIC DEPRESSURIZATION SYSHM Drywell Pressare - High (Permissite) (2 Ell-N694A,B,C,D) 2 a.

u b.

1 heactor Wssel Water Ievel - Iow Ima Iow (Ievel 1) 1, 2, 3 (2B21-N691 A,B,C D)

ADS' Timer (2821-K/52 A, B) 2 c.

1,2,3 Y

d.

1 ADS tow Water Iesel Actuation Timer (2B21-K754A,B; 2B21-K756A,B) 1,2,3 I

2 1,2,3 Reactor Wasel Water level-tow (Invel 3) (Permissive) 1 1, 2, 3 e.

(2B21-N695A,B) f.

Cbre Spray Rag Discharge Presaire - High (Permissive) i' (2E21-N655A,Br 2E21-N652A,B) 2 RHR (LPCI DODE) Rag Discharge Presaare - liigh (Permissive) 1,2,3 g.

i (2E11-N655A,B,C,D; 2 Ell-N656A,B,C,D) h.

Control Power Monitor (2821-K1A,B) 2/ loop 1, 2, 3 Y

l/tusta) 1, 2, 3 g

5. IOW IDW SET S/RV SYSTEM I

'k l

5 ReFctor Steam Dome PressJre - Illgh (Permissive) a.

(2B21-N620A,B,C,D) k 2

E 1,2,3 s

F>

g tal Alarm only.

declare the system inoperable.Nhe-Inoperable, verify power availabilit/ to the tus at least once per 12 hairs or

-h (b)

?rovides signal to IIPCI pug saction valves only.

(c)

When either channel of the aatomatic transfer logic is inoperable, align HFCI paq aiction to g

the asppression pool.

I IlPCI and NXI are not rego ired to tw> Ol'lGunfJ: with rea,: Lor utnasia,kwwr prerunern

$ 150 pni,3 IS m

m m--

r e-

TABLE 3.3.3-2 g

EMERCENCY CORE C00LINC SYSTEM ACTUATION INSTRUMENTATION SETPolNTS i

ALLOWABLE

_ TRIP FUNCTION TRIP SETPOINT VALUC

.Q 1.

. CORE SPRAY SYSTEM

%. g s.

Reactor Vessel Water Level - Low LW Low (Level 1) 2 -113 inches

2 -113 inches *-

b.

Drywell Pressure - High 5 1.92 psig 5 1.92 psig c.

Reactor Steam Dome Pressure - Low y

2 390 psig**

m d.

Logic Power Monitor NA NA y

2.

LOW PRESSURE COOLANT INJECTION MODE OF Ri1R SYSTEM w

a.

Drywell Pressure - liigh 5 1.92 psly 5 1.92 psig cn b.

Reactor Vessel Water Level - Low Low Low (Level 1) 2 -113 incnes*

2 -113 i nctie s'

)

N c.

Reactor Vessel Shroud Level (Level 0) - liigh 2 -202 Inches

2 -202 Inches

d.

Reactor Steam Dome Presstere-Low g

2 390 psig**

2 390 psig**-

l e.

Reactor Steam Dome Pressure-Low

2 335 psig 2 335 psig f.

RitR Pump Start - Time Delay Relay 2)

Piemps A, 8 and D 10 1 1 seconds 10 1 1 seconds I

1

)

Pump C 0.5 1 0.5 seconds 0.5 1 0.5 seconds

(

i g.

Logic Power Monitor MA NA N

b tas

$3 m

l I

i l

l l

'See Bases figure 8 3/81 3-1.

l

- This trip function shall be less than or equal to 500 psig.

l l

l 9

e i

t=

l TABLE 3.3.3-2 (Continued)

IMERCENCY CORE COOLING SYSTEM ACTUAfl0N INSTRUMENTATION SETPolNTS n

3*

e TRIP FUNCTION TRIP SETPolNT ALLOWABLE d

3.

_HICH PRES $t/RE COOLANT llLIECTIOlt SYSTEN VALUE

[N Reactor Vessel Water Level - Low Low (Level 2) a.

b.

2 -47 inches *

+

Drywell Pressure-fligh 2 -47 inches

y c.

Condensate Storage Tank Level - Low 5 1.92 psig Suppression Chamber Water Level - liigh 2 0 Inches **

5 1.92 psig 3'

w d.

Logic Power Monitor s 154.2 inches ***

2 0 inches **

.e.

NA 5 154.2 inches ***

I 4

w r.

Reactor Vessel Water Level-fligh (Level 8)*

NA s 56.5 inches 5

5 56.5 inches

^

3 AUTOMATIC DEPRESSURlZAfl0N SYSTEM w

,N a.

Drywell Pressure-liigh b.

Reactor Vessel Water Level - Low Low Low (Level 5 1.92 psig

3) 2 -113 inches

5 1.92 psig m'

1 ADS Timer c.

N d.

ADS Lov Water Level Actuation Timer 5 120 seconds 2 -113 inches

5 120 seconds Resctor vessel Water Level - Low s 13 minutes e.

Core Spray Plump Discharge Pressure (Level 3) 2 to inches 5 13 minutes r.

e 2 10 Inches *

- liigh 2 137 psig RilR (LPCI MODE) Pump Discharge Pressure - tilgh g.

2 13T psig 2 112 psig h.

Control Power Monitor NA 2 112 psig NA 5.

LOW LOW SET S/RV SYSTEM Reactor Steam Dome Pressure - liigh a.

5 1054 psig 5 1054 psig to to

See Bases Figure 8 3/4 3-1.

- Equivalent to 10

- - Measured above,000 gallons or water in the CST.

torus invert.

g j

i l

4

..n.-

-~ -

.w-

-:~:.

+.:>.-

4 TABl! 3.3.3 3 E."IRGINCY COE CCOIING ST3 TIM ESPONs! TI.wr.3

?

.s.

ESPONEI ~!S (Secones) 1.

CCRI SPRAY 373 23

< 27 2.

I.C'4 PRII3t:RI CCCI. Art I:CICT.*oM "CDI OT 33R 37a.4d

< 10 3.

32C3 ??"'*2I CCCI. ANT IXJIC.'0N ST3IIM

< 30 a.

A C C.".ATIC ::IPRESSIRI'.A!!ON 375:I3 NA 5.

ARM.',0*J I,0W EI".T.42 t

O MAT *H

.: NIT 2 3 ; *..':-30

.imenenen-Nc. 33

-~

e e s

.e ' < -,. gw M-r*v -,

g g-e- w m* p e g

-e

,-e--

TNirJi 4.3.3-1 (Cosit-laun1) til IMmGEICY QJtE (. DOLING SYSitM AC1UATION INS 111tMWl'AT104 SURVEIIANK2 REQUIRIMNrs e

OIANP&L OETRATIONAL g

OlNGEL FUNCTIONAL OIANDEL (DNDITIJNS IN WilIOl s

u TRIP IUCTION QECK TEST CALIBRATION SUlWEIIJANCE IEX)UllEDI 3.

HIQ1 PRESSURE (DOIANP ISUFICTION SYSPIM l

a. Reactor Vessel Water Invel -

Iow Iow (Ievel 2)

S M

R 1, 2, 3

b. Drywell Presaare-High S

M R

1, 2, 3

c. Oordensate Storage Tank level -

Iow NA M

Q 1, 2, 3 d

d. Rappression Chamber Water s

Invel - High S

M R

1, 2, 3

e. Icgic Power 2 nitor NA' R

NA 1, 2, 3 4

f. Reactor Vessel Water Imm1 - High S

M

,R 1, 2, 3 i

(Ievel 8)

I r u1 4.

AUIDETIC DEPHESSURIZATION SYSTIM u

I' 4

a. Drywell Presaare - High S

M R

1, 2, 3 u

b. Reactor Vessel Whter Ievel -

Iow Iow low (Invel 1)

S M

R 1, 2, 3

c. ADS Timer NA NA R

1, 2, 3

d. ADS Iow Water Invel Actination Timer NA NA R

1, 2, 3 l

e. Reactor Vessel Water level - Inw S

M R

1, 2, 3 i

(Ievel 3)

I. Core Spray Rangi Discharge p

Presaire - High S

M R

1, 2, 3 f.

g. RIR (IKI MODE) Rang Discharge" 4

Pressare - High S

M R

1, 2, 3

=

(

h. Control Power Monitor NA R

NA 1, 2, 3,

l;.

.lI m

N

'5.

ION IDW SET S/RV SYSTEM j.

=

E

a. Reactor Stease Dame Presaire -

6 High S

M R

1, 2, 3 h

.K Ccacroc h ppar' M ard ADS are not recnired to be OPfRABLE with tor steam dome presaire 5 150 psig.

l-i

'~~~ ~., -

g N

TABLE 4.3.3-1 1

8o

_EMERCEllCY CORE COOLIIIC SYS1EM ACTUATilNI INSTRtMENTATION SURVEI 4

3Z lj f,

CllAllMEL

_ TRIP FUllCTION CilANNEL FUNCTIONAL CllANNEL CONotilONS IM wilCN OPERAT 1001AI.

_ CllECR TEST I

_CAllBRATION SURVEILLAIICE REQUIREO 1.

CORE SPRAY SYSTEM j i z -4 Pm a.

Reactor Yassel Water Level -

L w Low Low (Level 1)

S M

i g

b.

Drywell Pressure - High R

- t S

M 1,2,3.4,5 c.

Reactor Steam Dome

{

R Pressure - tow 1, 2, 3 S

M y

d.

Logic Power Monitor R

NA R

1, 2, 3, 4, 5 NA 2.

_l0W PRESSURE C001. ANT INJECTION MODE Ol' RilR SYSTEM 1,2,3,4,5

[

s.

Drywell Pressure - liigh S

M l

b.

Reactor Vessel Water Level -

R I'

1,2,3 Low Low Low (Level 1)

S M

}'

c.

Reactor Vessel Shroud Level R

(Level 0) 1, 2, 3, 4 ', 5 '

S M

d.

Reactor Steem nome R

Pressure - tow 1, 2. 3, 4*.

5' l

S Reactor Steen Dome Pressure - Low M

R e.

S M

1,2,3,4*.5' l'.

RilR Pump Start-Time Delay Reley NA NA 1, 2, 3, 4 *, 5' R

R g.

Logic Power Monitor R

A NA R

1, 2, 3, 4 ', 5 '

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

'I W

.w

'Not applicable when two cGro spray subsystems are OPERABLE per Spectrication 3 5 31 4

9 5

I e

g S

e d

4 4

e e

TAntE 3.3.4-2 gh REACTOR CORE ISOLAfl006 C00LillC SYSTEM ACTUAil006 INSTRUMENTATION

-4 s9 Q,

FUNCTIONAL UNITS ALLOWABLE TRIP SETPOINT VALUE

}

$c s.

Reactor Vessel Water Level - Low Low (Level 2) 2 -47 inches

2 -47 Inches

l 2

b.

Condensate Storage Tank Level - Low o

--a 2 0 Inches" 2 0 inchesee c.

Suppression Pool Water Level - High 5 151 inches 5 151 Inches

?

See Bases Figure B 3/4 3-1

- Equivalent to 10,000 gallons or water in the CST.

w:"

~

e, N*

r e

h w

UI I

\\\\

9 e

9

..-...... - - - - - - - ~ ~ - - - ^-

4 ml.

=,, s y-g,'

N.

,4

$5

=

E _,2 3

=

a a

g5 D4 01 E,

E B

=

h I

lUh1 g

_5 E

9 DE" z

C 5

i!

E l

lli g

a m

g I

i-E 5

c 1

3

=

b 3

3 E

4 p ' A n a%

i 5

5

-ls= 3 I.

1 i

=

i e

m

.s _s A.

s o

lE B:

6 b

h1

.E l

E

$1 $$ 'aT$

i i

31*.3 - 7.T:7 2 3/A 3-36 Amendment No. 7), 39 1

m 7

~ "

s NOTE. SCALEsNeNCHES 3

A83VE VE53EL 2ERO e

^

WATER LEVgg wcuggggg7ggg MEIGHT A80VE 800 * -

VESSEL 2EAO No.

IINCMEsl mEActNC INETauMENT ISI 573.5

44.5 750 -

BARTON 171 559

+42 CE/W&C

, VE5SEL _

14)

$dt

32 G E/MAC t

FLANCE 131 527

+10 i

BARTON/ROSEMOUNT (2J 470

-47

~~

SARTON/ROSEMOUNT til 404

- 113 BARTON/ROSEMOUNT 101 315

- 202 BARTON/ROSEMOUNT BSO - =

" # ~ MAIN STEAM LINE f

s00 - -

577

3rifl,83 0--

a --

550 -, ga g e4 483 13, 171 ' 42 MI ALAnu I

80TTOM OF STE AM 527 (3) 7m PS LD ALAmu

-ORYER Suset 517 sN57mDMENT (33 - 10 (2) 10 LOW (LEVEL 31 -

FEED 500 - -

0 N 0--

CONTalal[r[TO ACS OR a

3

,,y$ _ CORE

- -47 LOW LOW (LEVEL 21'

- 470 (2) SPRAY g

INITIATE MPCI, I

450 -

RCIC 404(1) 400

- -113 LOW LOW LOW (LEVEL 11 4 3g7

-150 -. INITIATE RMR. C.E 35o -w 352.54 START DIEEEL AND CONTRIBitTE TO A.D.E.

2/3 CC A E CLCSE Msivs

-202 -

MEICMT 515 (0)

PE R Missivt 300 - -

[

(LEVELDJ ACTIVE FUEL 250

-217 - -

200 - -208.54 AECimC

' RECIRCULATION PUMP TRIP mECant

- 173.58 Ol5CH A Agg ANALYTICAL LIMITIS -58 INCHES SuCT son - 1s1.3 NO22LE NCZZLE 150 - -

300 - -

50 - -

0 -,-

BASES FIGURE B 3/4 31 REACTOR VESSEL WATER LEVELS HATCH - Utili 2 B 3/4 3-6 Admendment No. AE, 67

- -- +

=

-r*

' ' ' ' ' ' * ' ' ' ' * * - - - ~ ~ ' ~ ~ ~ ' ~ ~ ^

" ' ' ~ '

'

v t e Letter Sequence Other
TAC:62029, (Open)
Initiation Request, Request, Request Acceptance... Administration Questions Answers Results Approval, Approval Other: ML20090G560, ML20207B057, ML20207J604, ML20207J614
MONTHYEAR1986-10-15 [Table View]

ML20207B057

Text

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