Text

Amends 122 & 120 to Licenses DPR-80 & DPR-82,respectively, Revise Combined TS for Dcpp,Units 1 & 2 to Change Surveillance Frequencies from Once Every 18 Months to at Least Once Per Refueling Interval for RTS & ESFAS Channels
	ML20203L479
Person / Time
Site:	Diablo Canyon
Issue date:	02/17/1998
From:	Steven Bloom; NRC (Affiliation Not Assigned)
To:	;
Shared Package
ML20203L483	List: ML20203L479; ML20203L488;
References
	NUDOCS 9803060085
	Download: ML20203L479 (35)
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.

/g urg)t UNITED STATE 8 Oj-

,}

NUCLEAR REGULATORY COMMISSION WASHINGTON D.C. 3004H001 e,s**.,*/

PACIFIC GAS AND ELECTRIC COMPANY DOCKET NO. 50 275 DIABLO CANYON NUCLEAR POWER PLANT. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.122 License No. OPR 80 1.

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

A.

The application for amendnent by Pacific Gas and Electric Company (the licensee) dated December 9. 1996, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's 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 d the Commission:

C.

There is reasonable assurance (I) that the' activities authorized by this amendment can be conducted without endangering the health and safety of the public and (ii) that such activities will be conducted-in compliance with the Commission's regulations:

D.

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

The-issuance of this amendment is in accor. nce with 10 CFR Part 51 of the Commission'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. DPR-80 is hereby amended to read as follows:

$0'O$o0!o$o$o275 p

PDR

/4

a.

2.=

(2)

Technical SDecifications The Technical Specifications contained in Appendix A and the.

Environmental Protection flan contained in Appendix B. as revised through Amendment No.122. are hereby incor) orated in the license.

Pacific Gas and Electric Company s1all operate the facility-in accordance with the Technical Specifications and the Environmental. Protection Plan. except where otherwise stated in specific license conditions.

3.

This license amendment is effective as of.its date of issuance to be implemented within 90 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

(

.(m c-i l

.teven D. Cloom. Project Manager Project Directorate IV-2 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical L

Specifications Date of Issuance: February 17, 1998 h

y

[jonog *t UNITED STATES g

,g NUCLEAR REGULATORY COMMISSION g

WASHINGTON. D.C. 30eeH001

%,*...+/

E,CIFIC GAS AND ELFCTRIC COMPANY A

DOCKET NO. 50s DIABLO CANYON NUCLEAR 00WER PLANT. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.120 License No. DPR 82 1.

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

A.

The application for amendment by Pacific Gas and Electric Company (the licensee) dated December 9. 1996, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I:

B.

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

C.

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations:

D.

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

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifica"ons as indicated in the attachment to this license amendment.

and parar ' ', ?.C.(2) of Facility Operating License No. DPR-82 is hereby amended t :-ad as follows:

4-2 (2)

Technical Soecifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B. as revised through Amendment No.120. are hereby incorporated iti the license.

Pacific Gas and Electric Company shall operate _the facility in accordance with the Technical Specifications and the-Environmental Protection Plan, except where otherwise stated in specific license conditions.

3.

This license amendment is effective as of its date of issuance to be implemented within 90 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION 4-Steven D. Bloom. Project Manager Project Directorate IV-2 Division of Reactor-Projects III/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical

- Specifications Date of Issuance:

February 17, 1998 5

4 4-ATTACHMENT TO LICENSE AMENDMENTS

.t -

AMENDMENT NO.122 TO FACILITY OPERATING LICENSE NO. DPR-80 AND AMENDMENT NO.120 TO FACILITY OPERATING LICENSE N0l DPR-82 DOCKET NOS. 50-275 AND 50-323 4

Revise Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by

- Amendment nunber and contain marginal lines indicating the areas aof change,

.The corresponding overleaf pages are also provided to maintain document

-completeness.

REMOVE INSERT e

2-4 2-4

.2-5 2-5 2-6 2-6 3'

2-7 2-7 2-8 2-8 4

i 29 2-9 2.10 2-10 B 2-3

'B 2-3 B24 B 2-4 3/4 3-10 3/4 3-10 i ~

3/4 3-11 3/4 3.

3/4 3-12 3/4 3-12 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 3-27 3/4~3-27 3/4 3-32 i

-3/4 3-32 3/4:3-33 3/4 3-33 3/4-3-34 3/4 3-34 F

3/4 3-35 3/4 3-35 B 3/4 3 B 3/4 3-1 B 3/4 3-la B 3/4 3-la B 3/4 3-lb r

1 e

v

~m

--w T

se

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.)

The Reactor Trip System Instrumentation and Interlock Setpoints shall be set consistent with the Trip Setpoint values shown in Tcble 2.2-1.

APPLICABILITY: As shown for each channel in Table 3.3-1.

AGIlgli:

With a Reactor Trip System Instrumentation or Interlock Satpoint less a.

conservative than the value shown in the Trip Setpoint column but more conservative than the value shown in the Allowable Values column of Table 1

2.2-1, adjust the Setpoint consistent with the Trip setpoint value.

b.

With a Reactor Trip System Instrumentation or Interlock Setpoint less conservative than the value shown in the Allowable Values column of Table 2.2-1, declare the channel inoperable and apply the a)plicable ACTION statement requirements of Specification 3.3.1 until tie channel is restored to OPERABLE status with its Setpoint adjusted consistent with g

the Trip Setpoint value, i

n 4

i 4

DIABLO CANYON - UNITS 1 & 2 2-3 Unit 1 - Amendment No. 84,103 Unit 2 - Amendment No. 83.102

~

JUN 2 1%6s

E TABLE 2.2-1 iE G

REACTORTRIPSYSTEMINSTRUMENTATIONTRIPSETP0lNH FUNCTIONAL UNII TRIP SETPOINT ALLOWABLE VALUES

1. Manual Reactor Trip N.A.

N.A

2. Power Range. Neutron Flux 5 26.2% of RATED THERMAL POWER g

a.

Low Setpoint 5 25% of RATED THERMAL POWER b.

High Setpoint 5109% of RATED THERMAL POWER 5110.2% of RATED THERMAL POWER

[

3. Power Range. Neutron Flux 5 5% of RATED THERMAL POWER with 5 5.6% of RATED THERMAL POWER l

High Positive Rate a time constant 2 2 seconds with a time constant 1 2 seconds i

4. Power Range. Neutron Flux 5 5%'of RATED THERHAL POWER with 5 5.6% of RATED THERMAL POWER l

High Negative Rate a time constant 2 2 seconds with a time constant 2 2 seconds

5. Intermediate Range. Neutron 5 25% of RATED THERMAL POWER 5 30.6% of RATED THERMAL POWER l

Flux l

m 5

5

6. Source Range. Neutron Flux 510 counts per second 51.4 x 10 counts per second

7. Overtemperature AT See Note 1 Sea Note 2 l

gg

8. Overpower AT See Note 9 See Note 4 23

9. Pressurizer Pressure-Low a 1950 psig 2 1947.5 psig l

m _.

h h

10. Pressurizer Pressure-High 5 2385 psig 5 2387.5 psig l

aa

11. Pressurizer Water Level-digh 5 90% of instrument span 5 90.2% of instrument span l

ma 88

%5

12. Reactor Coolant Flow-Low 2 90% of minimum measured 2 89.8% of minimum measured l

flow ** per loop flow ** per loop.

zz

.O O Et

~ hi 3'.

EsM

- Minimum measured flow is 89.800 gpm per loop for Unit 1 and 90.625 gpm per loop for Unit 2.

v s

r

,r-.-

v.

e IABLE 2.2-1 (Continued) o-7 g

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS.

$ FUNCTIONAL UNIT TRIP SETP0 INT ALLOWABLE VALUES

$:13.SteamGeneratorWater 2 7.2% of r; arrow range instrument 2 7.0% of ' narrow range instrument

'l Level-Low-Low span-each steam generator span-each steam generator l

E Coincident with:

j d

a..RCS Loop AT Equivalent RCS Loop AT variable input RCS Loop AT variable input l

to Power s50% RTP-550% RTP s50.7% RTP l

[

With a time delay (TD)-

sTD (Note 5) 5(1.01)TD (Note 5)

Or b.

RCS Loop AT Equivalent RCS f.oop AT variable input RCS Loo) AT variable-input.

to Power >50% RTP.

>50% RTP

>50.7% RTP With no time delay TD = 0 TD = 0

'14. DELETED m

& 15. Undervoltage-Reactor 2 8050 volts-each bus t 7877 volts-each bus l

Coolant Pumps

16. Underfrequency-Reactor 2 54.0 Hz - each bus a 53.9 Hz - each bus gg Coolant Pumps EN 17. Turbine Trip m-a.

Low Autostop Oil 2 50 psig 2 45 psig Pressure gg b.

Turbine Stop Valve 1 1% open 2 1% open gg Closure

$ $ 18. Safety Injection Input N.A.

N.A.

ga from ESF

= 0 19. Reactor Coolant Pump N.A; N.A.

J.

Bieaker Position Trip

-[-[ 20. Reactor Trip Breakers N.A.

N.A.

-$-2 21. Automatic Trip and N.A.

N.A.

ES Interlock Logic a

L _ __ __

S TABLE 2.2-1 (Continued)

G REN TOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS hFUNCTIONALUNIT TRIP SETPOINT

' ALLOWABLE VALUES o

22. Reactor Trip System Interlocks E

a Intermediate Range Neutron 2 1 x 10.ia amps 2 8 x 10~22 amps l

3 Flux. P-6

b. Low Power Reactor Trips

[

Block P-7

1) P-10 Input 10% of RATED 8.8%. 5 11.2% of RATED l

'l4ERMAL POWER THERMAL POWER i

2) P-13 Input 5 10% RTP Turbine Impulse 5 10.2% RTP Turbine Impulse l

l Pressure Equivalent Pressure Equivalent

c. Power Range Neutron Flux. P-8 5 35% of RATED 5 36.2% of RATED I

THERMAL POWER THERMAL POWER

d. Power Range Neutron Flux P-9 5 50% cf RATED 5 51.2% of RATED l

THERMAL. POWER THERMAL POWER

e. Power Range Neui.ron Flux. P-10 10% of RATED E 8.8%. 5 11.2% of RATED l

THERMAL POWER THERMAL F0WER cc EE

((

f. Turbine Impulse Chamber Pressure. P-13 5 10% RTP Turbine Impulse 5 10.2% RTP Turbine Impulse l

Pressure Equivalent Pressure Equivalent

[ [ 23. Seismic Trip 5 0.35 g 5 0.43 9 l

EE 88

%a E.E 5

V has 4

o h

15 TABLE 2.2-1 l

E L

G REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS l'

a E

TABLE NOTATIONS 3

NOTE 1:

OVERTEMPERATURE 4T f

c

' $ aT (1 + r,5 )'. s AT, K -K (1 + r S) [T - T*] + K (P '- P' ) - f ( AI)'.

i i 2 3

i 1 + r S.

1 + r5 3

3 a.

'" Where:

1 +.7S

- Lead-lag compensator on measured aT 4

i 1+r5 s

- Time constants utilized in the lead-lag compensator for aT. 7 - 0 seconds.: 73 - O seconds l

r.. r3 aT,

- Loop specific indicated 4T at RATED-' THERMAL POWER l

i, K.

- 1.2

=

K

-.0.0182/*F 2

n 1 + rs

- The function generated by the lead-lag compensator for T,, dynamic compensation

l i

ru,

1+T5 2

>g t

jg gg r = 4-seconds i

- l

[

- Time constants utilized in'the lead-lag compensator for T.,. r - 30 seconds.

r,r2 i

2 AE T

- Average temperature.

F-

.E.E

^

.E.E

=

==

o ru

[

Wt

TABLE 2.2-1 (Continued) o L~

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS g

l 9

TABLE NOTATIONS

!i! NOTE 1:

(continued)

T'

- Nominal loop specific indicated T at RATED THERMAL POWER l

m c"

I d

K

- 0.000831/psig 3

i A

P

= Pressurizer pressure, psig m

P"

- 2235 psig (Nominal RCS operating pressure)

S

- Laplace transform operator, s-2 and f (AI) is a function of the indicated difference between top and between detectors of the power t

i range nuclear ion chambers: with gains to be selected based on measured instrument response during ma plant startup tests such that:

l for gt - q, between -19% and +7%. f,(AI) = 0 (where q and q, are percent RATED THERMAL POWER in (i) the top and bottom halves of the core respectively,tand q, + q, is total THERMAL POWER in cc()

percent of RATED THERMAL POWER).

(ii) for each percent that the magnitude of (qt - q,) exceeds -19%. the AT Trip Setpoint shall be

"^

automatically reduced by 2.75% of its value at RATED THERMAL POWER.

j RF (iii) for each percent that the magnitude of (qt - q,) exceeds +7%. the AT Trip Setpoint shall be

((

automatically reduced by 2.38% of its value at RATED THERMAL POWER.

oo

" " NOTE 2:

The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 0.46% AT span for hot leg or cold leg temperature inputs. 0.14% AT span for pressurizer pressure

[

input, or 0.19% AT span for AI inputs.

TP

-e* N

=

Yr; o~

A m/

.3

O P

TABLE 2.2-1 (Continued)

O

-9

. REACTOR TRIP SYSTEM lNSTRUMENTATION TRIP SETPOINTS 2.c E

TABLE NOTATIONS

,h NOTE 3:

Overpower AT m

3(7S

) T - K, [T - T"] - f,(AI).

AT (1 + T S ) s AT, K -K 3

4 o-1+rS 1+r5 3

3 N

i 1+rS

= Lead-lag t.21pensator on measured AT Where:

1+T53

- Time constants utilized in the lead-lag compensator for AT. r, = 0 seconds.

l 7

r 4

s r - 0 seconds 3

AT,

= Loop Specific: Indicated 5 at RATED TurRNAL POWER l

,'?

K

= 1.072 s

K

= 0.0174/*F for increasing average temperature. and 0 for decreasing _ average 3

temperature cc hb 75

- The function generated by the. rate-lag cogensator for T, dynamic compensation N7 3

i 1+T53 FF

- Time constar.ts utilized in the rate-lag compensator for T,. r = 10 secs.

l

$[

3 r3

&a j@

K,

= 0.00145/*F for T > T". and 0 for T s T" fy T

= Average temperature. *F E"

T"

= Nominal loop specific indicated T,at RATED THERMAL POWER l

NL 5

= Laplace transform operator, s

EE

'g*g f (AI)

= O for=all Al 2

m=

NON

-g

TABtE 2.2-1 (Centinued) o

~>

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETP0lJi5

[

TABLE NOTATIONS The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 0.46%

E NOTE 4:

AT span for hot leg or cold leg temperature inputs.

NOTE 5:

Steam Generator Water Level Low-Low Trip Time Delay TD

- B1(P)' + B2(P)2 + B3(P) + B4 e

Where: P

- RCS Loop AT Equivalent to Power (%RTP). P s 50% RTP TD

- Time delay for Steam Generator Water Level Low-Low Reactor Trip (in seconds).

B1

-0.007128

'?

B2 - +0.8099 1

m 83

-31.40 B4 - +464.1 EE 32 m-t i a> cs 33 88 on i

mkn b

IU M

~

O u

L2 L IMllite W t li 515ttM 5t ilitm BASFS 2 2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS The Reactor Trip Setpoint Limits specified in Table 2.2-1 are the nominal values at which the Reactor trips are set for each functional unit.

The Allowable Values are considered the Limiting Safety System Settings (LSSS) as identified in 10 CFR 50.36.

The LSSS settings 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 the Engineered Safety Features Actuation System in mitigating the consequences of accidents.

The various reactor trip circuits automatically open the reactor trip breakers whenever a condition monitored by the Reactor Trip System reaches a preset or calculated level.

In addition to redundant channels and trains. the design ap3 roach 3rovides a Reactor Tr1p System which monitors numerous system variaales, t1erefore, providing protection system functional diversity.

The Reactor Trip System initiates a Turbine trip signal whenever Reactor trip 1s initiated This prevents the reactivity insertion that would other-wise result from excessive Reactor Coolant System cooldown and thus avoids unnecessary actuation of tne Engineered Safety Features Actuation System.

The setpoint for a reactor trip system or interlock function is considered to be adjusted consistent with the nominal value when the "as left" setpoint is within the band allowed for calib-ation tolerance.

There is a band allowed for calibration tolerance only for those setpoints which use analog hardware.

For example, the Power Range, Neutron Flux High setpoint is properly adjusted when it is set at 109%

0.3% (0.25% of 120% power span).

The calibration tolerance, after appropriate conversion, should correspond to the rack comparator setting accuracy defined in the latest setpoint study.

The setpoints which use digital hardware are set at the nominal value in the system.

Trip setpoints may be administratively redefined in the conservative direction for several reasons including startup, testing, process error accountability, or even a conservative response fcr equipment malfunction or inoperability.

Some trip functions have historically been redefined at the beginning of each cycle for purposes of startup testing, e.g.. Power Range Neutron Flux High and Overtemperature AT.

Calibration to within the defined calibration tolerance of an administratively redefined, conservative Trip Setpoint is acceptable.

Redefinition at full power conditions for these functions is expected and acceptable.

To accommodate the instrument drift that may occur between operational tests and the accuracy to which setpoints can be measured and calibrated.

Allowable Values for the Reactor Trip Setpoints have been specified in Table 2.2-1.

Operation with a trip set less conservative than its Trip Setpoint.

but within its specified Allowable Value, is acceptable.

Rack drift in excess of the Allowable Value exhibits the behavior that the rack has not met its allowance.

Since there is a small statistical chance that this will happen, an infrequent excessive drift is expected.

Rack or sensor drift in excess of the allowance that is more than occasional may be indicative of more serious problems and warrants further investigation.

The methodology to derive the Trip Setpoints is based upon combining all of the uncertainties in the channels.

Inherent to the determination of the Trip Setpoints are the magnitudes of these channel uncertainties.

Sensors and other instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes.

Manual Reactor Trio The Reactor Trip System includes manual Reactor trip capability.

DIABLO CANYON - UNITS 1 & 2 B 2-3 Unit 1 - Amendment No. M.122 Unit 2 - Amendment No. 43.120

llMITING SAFETY SYSTEM SETTINGS BASES

'l Power Rance. Neutron Flux In each of the Power Range Neutron Flux channels there are two indepen-dent bistables, each with its own trip setting used for a High and Low Range trip setting.

The Low Setpoint trip provides protection during subcritical and low power operations to mitigate the consequences of a power excursion beginning from low power and the High Setpoint trip provides protection during power operations to mitigate the consequences of a r activity excursion from all power levels.

The Low Setpoint trip may be manually blocked above P-10 (a power level of approximately 10% of RATED THERMAL POWER) and is automatically reinstated below the P-10 setpoint.

Power Rance. Neutron Flux. High Rates The Power Range Positive Rate trip provides protection against rapid flux increases which are characteristic of a rupture of a control rod drive housing. Specifically, this trip complements the Power Range Neutron Flux High and Low trips to ensure that the criteria are met for rod ejection from mid-power.

The Power Range Negative Rate Trip provides protection for control rod drop accidents.

At high power, a rod drop accident could cause local flux Jeaking which could cause an unconservative local DNBR to exist.

The Power Range Negative Rate Trip will prevent this from occurring by tripping the reactor.

No credit is taken for operation of the Power Range Negative Rate Trip for those contrcl rod drop accidents for which the DNBRs will be greater than or clual to the DNBR limits.

Intermediate and Source Range. Neutron Flux The Intermediate and Source Range Neutron Flux trips provide core protec-tion during reactor STARTUP to mitigate the consequences of an uncontrolled rod cluster control assembly bank withdrawal from a subcritical condition.

These trips provide redundant protection to the Low Setpoint trip of the Power Range Neutron Flux channels.

The Source Range channels will initiata a Reactor trip at about 10+5 counts per second unless manually blocked when P-6 becomes active.

The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to a) proximately 25% of RATED THERMAL POWER unless manually blocked when P-10 3ecomes active.

No credit was taken for operation of the trips associated with either the Intermediate or Source Range channels in the accident analyses; however, their functional capability at the specified trip settings is required by this specificat;on to enhance the overall reliability of the Reactor Trip System.

Overoower AT The Overpower AT trip arovides anurance of fuel Integrity, e.g., no fuel pellet cracking or me' ting, under all possible overpower conditions, limits the required range for Overtemperature AT protection. and provides a backup to the High Neutron Flux trip.

The Setpoint is automatically varied DIABLO CANYON - UNITS 1 & 2 B 2-4 Unit 1 - Amendment No. N.122 Unit 2 - Amendment No. M.120

7 Alt! 3.b2 (Continued)

Rt ACT0a TRIP Sv5 TEM INSTRu".!NTATION tt$PON$f TIMI$

.1 FUN **10w.ON!T RESPON$f TIMI

T.

Turtine Trip

a.

Low Fluis 011 Preas sa N.A.

t.

Turbine Stop Valve N.A.

11.

Safety 2njection Input from [5F N.A.

al.

Restter Coolant Pump Breaker Position Trip N.A.

2t.

Ree' :r Trip Breakers N.A.

21.

Auton.atic 1 rip and Interlock Logic N.A.

22 Rea:ter 1 rip $yster Ir.teriseks N.A.

23.

Se's-4C Trip k.A.

e G

e DIABLO CANYON = UNITS 1 & 2-3/4 3-9 w,

w=._

TABLE 4.3-1 o

5~.

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 9

TRIP 5

ACTUATING.

MODES FOR E

' CHANNEL DEVICE WHICH CHANNEL CHANNEL GPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE g FUNCTIONAL (WIT CHECK CALIBRATION TEST TEST LOGIC TEST 15 REQUIRED-

1. Manual Reactor Trip N.A.

N.A.

R24(14)

N.A.

1. 2. 3*.4*.5*

2. Power Range. Neutron Flux N

a.

High Setpoint S

D(2. 4).

0 N.A.

N.A.

1. 2 M(3. 4).

0(4. 6).

R24(4. 5) b.-

Low Setpoint S

R24(4)

S/U(1)

N.A.

1#ff. 2

3. Power Range. Neutron Flux.

N.A.

R24(4) 0 N.A.

N.A.

1. 2 l

R High Positive Rate h 4. Power Range. Neutron Flux.

N.A.

R24(4) 0 N.A.

N.A.

1. 2 l

e; High Negative Rate 5.

Intermediate Range.

S R24(4. 5)

S/U(1)

N.A.

1##f. 2 l

Neutron Flux EE 22 6. Source Range. Neutron Flux 5

R24(4. 5)

S/U(1).0(8)

N.A.

2#f. 3. 4. 5 l

7. Overtemperature 4T S

R24 0

N.A.

1. 2 i

m-II 8. Overpower AT S

R24 0

N.A.

1. 2 l

EE R R 9. Pressurizer Pressure-Low:

S R24 0

N.A.

1 l

AA

10. Pressurizer Pressure-High S

R24 0

N.A.

1. 2 l

zz

??

11. Pressurizer Water Level-High S R24 0

N.A.

1 l

E 12. Reactor Coolant Flow-Low S

R24 0

N.A.

1 l

3 NE o~

g

S TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE RE0VIREMENTS O

TRIP g

ACTUATING MODES FOR z

CHANNEL DEVICE WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL' ACTUATION SURVEILLANCE:

E FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED a

" 13. Steam Generator Water Level-Low-Low a.

Steam Generator S

R24 0

N.A.

1. 2 l

Water Level-tow-Low b.

RCS Loop AT Equivalent N.A.

R24 0

N.A.

1. 2 to Power

~

14. DELETED w 15. Undervoltage-Reactor N.A.

R24 N.A.

Q N.A.

1 l

l 2

Coolant Pumps w

. 16. Underfrequency-Reactor N.A.

R24 N.A.

Q N.A.

1 l

Coolant Pumps

17. Turbine Trip a.

Low Fluid Oil Pressure N.A.

N.A.

S/U(1. 9)

N.A.

I b.

Turbine Stop Valve N.A.

N.A.

S/U(1. 9)

N.A.

1

((

Closure

18. Safety Injection Input N.A.

N.A.

R24 N.A

1. 2 from ESF kk19.ReactorCoolantPump N.A.

N.A.

R24 N.A.

1 Breaker Position Trip m.

E E'20. Reactor Trip System Interlocks ee a.

Intermediate Range N.A.

R24(4)

R24 N.A.

N.A.

2ff l

??

Neutron Flux. P-6 M

b.

Low Power Reactor Trips Block. P-7 N.A.

R24(4)

R24 N.A.

N.A.

1 l

~'m=

?*

c.

Power Range Neutron jj Flux. P-8 N.A.

R24(4)

R24 N.A.

N.A.

1 l

L

t o

. TABLE 4.3-1 B

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS' o

. TRIP i

-g ACTUATING MODES FOR :

CHANNEL DEVICE WHICH l'

. CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION

-SURVEILLANCE ~'

!iii FUNCTIONAL UNIT

' CHECK

^ CALIBRATION TEST TEST LOGIC TEST IS REQUIRED

.l A

. 20. Reactor Trip System Interlc_ks (Continued) d.

Power Range Neutron Flux P-9 N.A.

R24(4)

R24 N.A.

N.A.

.1

.l e.

Low Setpoint Power Range

'l Neutron Flux P-10 N.A.

R24(4)

'R24-N.A.

N.A.

1. 2 f.

Turbine Impulse Chamber t

Pressure. P-13 N.A.

R24 R24 N.A.

N.A.

I g

j

21. Reactor Trip Breaker N.A.

N.A.

M(7. 10)

N.A.

1.2.3*.4*.5*.

~

w a

i

22. Automatic Trip and N.A.

N.A.

M(7) 1.2.3*.4*.5*

i w

h Interlock Logic

23. Seismic Trip N.A.

R24 N.A.

R24 M(7)

1. 2 l

j

24. Reactor Trip Bypass N.A N.A.

N.A.

M(7.15).R24(16)N.A.

.1.2.3*.4*.5*

gg Breaker 33 m-l r&

88 t

&R 88 t

4 oo a

22 1

f

%we (M e (M (b L

??

i em -e*

r

'i O

^

e TABLE 3.3-4 c

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS 53FUNCTIONALUNIT TRIP SETPOINT ALLOWABLE VALUES 5

s:

1.

Safety Injection (Reactor Trip. Feedwater i

Isolation. Start Diesel Generators.

Si Containment fan Cooler Units, and a

Component Cooling Water) w a.

Manual Initiation N.A.

N.A o.

^)

b.

Automatic Actuation Logic N.A.

N.A and Actuation Relays c.

Containment Pressure-High s 3 psig s 3.12 psig l

1847.5 psig l

d.

Pressurizer Pressure-Low a 1850 psig 2

e.

DELETED 3:

f.

Steam Line Pressure-Low a 600 psig (Note 1)

= 197.6 psig (Note 1) l w

w EE M

h) "

e I BB 99 88

~

.O hh

==

ON w_.

=-

TABLE 3.3-4 (Continued) o 5*

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS g

9 FUNCTIONAL UNIT TRIPSETPOIrplT ALLOWABLE VAlij[S 59 2.

Containment Spray (coincident with SI signal)

C5 a.

Manual Initiation N.A.

N.A

'ni b.

Automatic Actuation Logic and fl. A.

N.A i

l Actuation Relays a.

c.

Containment Pressure-High-High s 22 psig s 22.12 psig l

~

I l

3.

Containment Isolation l

l a.

Phase "A" Isolation 1)

Manual N.A.

N.A 5

2)

Automatic Actuation Logic N.A.

N.A.

and Actuation Relays wk 3)

Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values.

(

gg b.

Phase "B" Isolation l

33 1)

Manual N.A.

N.A m,

2)

Automatic Actuation Logic N.A.

N.A

[g and Actuation Relays hh 3)

Containment Pressure-High-High s 22 psig s 22.12 psig

%A E.E aa 25 LL WN

TABLE 3.3-4 (Continued)

'o5 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM IN!TRUMENTATION TRIP SETPOINTS E FUNCTIONAL UNIT TRIP SETP0 INT ALLOWABLE VALUES m3z 3.

Containment Isolation (Continued) h c.

Containment Ventilation Isolation a

m 1)

Automatic Actuation Logic N.A.

N.A.

and Actuation Relays 2)

Deleted o

3)

Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values.

4)

Containment. Ventilation Per the 00CP Exhaust, Radiation-High (RM-44A and 448) 4.

Steam Line Isolation b

a.

Manual N.A.

N.A.

b.

Automatic Actuation Logic' N.A.

N.A.

and Actuation, Relays bh Containment Pressure-High-High 5 22 psig 5 22.12 psig I

c.

d.

Steam Line Pressure-Low a 600 psig (Note 1)

= 597.6 psig (Note 1)

.l

@R aa RR 88 hh sa hh

..=

ru fV

2 TABLE 3.3-4 (Continued)

E; E

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRLM NTATION TRIP SETPOINTS O FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES g

l i

e.

Negative Steam Line Pressure Rate-High 5 100 psi (Note 3) s 102.4 psi (Note 3) l E

5.

Turbine Trip and feedwater Isolation

~--.

l a.

Automatic Actuation Logic N.A.

N.A.

and Actuation Relays N

b.

Steam Generator Water level-s 75% of narrow range s 75.2% of narrow range l

High-High instrument span each steam instrument span each steam generator.

generator.

6.

Auxiliary Feedwater a.

Nanual N.A.

N.A.

b.

Automatic Actdation Logic N.A.

N.A.

and Actuation Relays w

s c.

Steam Generator

= 7.2% of narrow range

= 7.0% of narrow range l

[

Water Level-Low-Low instrument span each instrument span each steam generator.

steam generator.

m Coincident with:

CC El

1) RCS Loop AT Equivalent to RCS Loop AT variable input RCS Loop AT variable input Power s 50% RTP s 50% RTP s 50.7% RTP l

m-With a time delay (TD) s TD (Note 2) s (1.01)TD (Note 2) kk Or aoyy

2) RCS Loop AT Equivalent to RCS Loop AT variable input RCS Loo) AT variable input gg Power > 50% RTP

>50% RTP

>50.7%

RTP.

l With no tiime delay TD - 0 TD - O h

d.

Undervoltage,- RCP

= 8050 volts

= 7877 volts l

33 e.

Safety Injection See : tem 1. above for all Safety In.jection Trip Setpoints and 2%

Allowable Values.

sC

_=."

CG o m.

..u.

j

?.

o TABLE 3.3-4 (Continued) 5 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES

>57.

Loss of Power 9

(4.16 kV Emergency Bus Undervoltage) a.

First Level g

1)

Diesel Start

= 0 volts with a

= 0 volts with a a

s 0.8 second time delay s 0.8 second time delay and and w

= 2583 volts with a a 2583' volts with a s 10 second time delay s 10 second time delay a-2)

Initiation of Load Shed One relay One relay N

2 0 volts with a

= 0 volts with a 5 4 second time delay 5 4 second time delay and and a 2583 volts with a a 2583 volts with a 5 25 second time delay s 25 second time delay with one relay with one relay

= 2870 volts, instantaneous a 2870 volts. instantaneous w

b.

Second Level a

1)

Diesel Start a 3785 volts with a a 3785 volts with a w

s 10 second time delay s 10 second time delay

',o 2)

Initiation of Load Shed a 3785 volts with a a 3785 volts with a 5 20 second time delay 5 20 second time delay

$$ 8.

Engineered Safety Features Actuation System Interlocks a.

Pressurizer Pressure P-11 s 1915 psig s 1917.5 psig l

b.

DELETED c.

Reactor Trip. P-4 N.A.

N.A.

((

h h NOTE 1:

Time constants utilized in the lead-lag compeasator for Steam Pressure - Low are 7 = 50 seconds l

3 gg and 72 - 5 seconds.

" " NOTE 2:

Steam Generator Watgr Level Low-Lcw Trip Time Delay y=

TD = B1(P) + B2(P) + B3(P) + B4 Lt Where:

P = RCS Loop AT Equivalent to Power (%RTP). P s 50% RTP

..P TD = Time delay for Steam Generator Water Level Low-Low (in seconds)

+'

.E,_f B1

-0.007128

$=

B2 - +0.8099 Le B3 = -31.40 B4 = +464.1

$$ NOTE 3:

Time constants utilized in the rate-lag compensator for Negative Steam Line Pressure Rate - High l

L _.

are r - 50 seconds and r4 - 50 seconds.

3 L

4 TAatt 3.5 1 i

gg1NttRfD 1AFETY FIATURtt RtsP0N5t TIMt1 1NITIATING IIGNAL AND FUNCTION fgigg$t TIME IN 1ECONDS 1.

Manual Initiation a.'

Safety Injection (ECCS)

N.A..

1)i Feedwater Isolatten N.A.

2 Reacter Trip N.A.

3h Phase 'A' Isolatioa N.A.

4h Containment Ventilation Isolatten N.A.

i 5p Auxiliary Feedwater N.A.

8)

CMonent Cooling Water N.A.

73 Containment Fan Cooler Units N.A.

8)

Auxiliary Saltwater Pumps N.A.

b.

Phase '8" Isolation 1)

Containment!) ray (CoinciG$twith i

31 Signal Conta<nmen)tVentilationIsolation j

I)

N.A.

l c.

Phase 'A' Isolation 1)

C6ntainment Ventilation Isolation N.A.

d.

Steam L% !selation M.A.

2.

Containment Pressure High I

a.

SafetyInjection(ECCS) s 17*/25

l 1)

F setor Trip s!

IJ Feedwater Isolation s 43 1

3)

Phase 'A' !solatten s 18"'

41 Containment Ventilatten Isolation M. A. */28*

Eh Auxiliary Feedwater s 60 4l)

Component Cooling Water s 38'p44*

71 Containment Fan Cooler Units as 40 8l)

Auxiliary Saltwater Pumps s 48"'/58 "

~

3.

Pressurizer Pressure Low a.

Safety injection (ICCS) s 17*/25*/35*

2 11 Reactor Trip Il Feedwater Isolatten aI s 43 l

31 Phase 'A' Isolatten a 18"8 41 Containment Ventilation Isolatten El Auxiliary Feedwater N.A.

8 l Component Coeling Water s 60

38"'

s48g 7

\\

Containment Fan Cooler Units s 40 8)

Auxiliary Saltwater Pumps s 58 */48"'

01MLO CANYON - UNITS 1 & 2 3/4318 Amendment Nos. W & 4h M & M, 77 & 7C

\\

TAatf 1.1 1 fcanti M 5

TM1.I lefATlW5 (1) Diesel generator starting delay not included because offsite power eva1121e.

(2) hotation deleted.

l (3) Diesel generater starting and leading delays included.

(4) Diesel generater starting delay not included because offsite toer is eveilele, Response time limit includes opent of valves to este11sh 5J path and attainment of discherse pressure for centrif I chargins sums (urere asolicele). Seemn.

i tial transfer of cherring puq 1an free the'K7 to the 467 (MT valves spen, then VCT valves elese) is included.

(5) Diesel eenerster starting and seemec 1sding delays included. Offsite power is not available. Response time limit includes apening of valves to estelish 5! path and attairment of discharge pressure for contrifussi charpi

. Seesential trensfer of che ng pug suction from the VCT to the IRS

(

v61ves spen then WCT valves close is included.

(6)

.The anxisma response time of 48.5 seconds is ths time from when the containment pressure exceeds the Nigh.Nigh 5etpoint until the spray pep is started and the discharge valve travels to the fully open position assains off. site power is net available. The time of 48.5 seconds includes the 23.second maximum delay related to E5F loeding sequence. Spriy riser piping fill time is not include 1.

The 80 second maximum spray delay time does not include the time from LOCA start to *P' signal.

(7) Diesel generator starting and sessence leading delaps included. Sesential trans.

fer of chargin valves clone) g pump suction from the VCT to the it5T (1667 valves open, then VCT is not included. Response time limit includes opening of valves to establish 51 flow charg1ng pumps 5! path and attainment of discharge pressure for centrifugal

. and Ibet pumps (where applicable).

(8) Does not include Trip Time Delays.. Response times include the transmitters. Eegle-21 Process Pro uction cabinets. Solid State Protection System cabinets and actus.

tion dertces only. This reflects the response times necessary for TElpAL POWER in excess of 504 RTP.

DIABLO CANYON. WIT 51 & 2 3/4 3 31 Acadment hos. 3e & eg. M & h.

NSD m

m,..-

,_ _~--- -.-

1 I

TABLE'4.3-2 o

5' ENGINEERED

.* L EATURES ACTUATION SYSTEM INSTRUE NTATION aORVEILLANCE REQUIREMENH 95 TRIP 9

ACTUATING CHANNEL DEVICE MODES FOR c.

CHANNEL OPERA-OPERA-MASTER SLAVE WHICH 5

CHANNEL CALI-TIONAL

.TIONAL ACTUATION-RELAY RELA" SURVEILLANCE d FUNCTIONAL UNIT CHECK BRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED

' 1. Safety Injection.'(Reactor Trip

[

Feedwater Isolation. Str.rt Diesel Generators. Contai!1 ment Fan Cooler Units. and Cogonent 3

Cooling Water) w2 a.

Manual Initiation N.A.

N.A.

R24 N.A.

N.A.

1. 2. 3. 4 Y

b.. Automatic Actuation N.A.

N.A.

M(1)

R

1. 2. 3. 4 M

Logic and Actuation Relays

~

c.

Containment Pressure-High S

R24 0

N.A.

N. A..

N.A.

1. 2. 3. 4 l j

cc El j

d.

Pressurizer Pressure-Low S

R24 0

N.A.

1. 2. 3 l

m-i i

e.

DELETED

g>

gj f.

Steam Line S

R24 Q

N.A.

1. 2. 3 l

f j gg Pressure-Low t

a a 2. Containment Spray (coincident with SI signal) q U.

a.

Manual Initiation N.A.

N.A.

R24 N.A.

N.A.

1. 2. 3. 4

[

1 ws La b.

Autcmatic Actuation Logic N.A.

N.A.

M(1)

R

1. 2. 3. 4 i

and Actuation Relays

[

r

??

c.

Containment Pressure-S R24 Q

N.A.

1. 2. 3. 4 l

}

I High-High l

l I

i I

F 1

r s.

.= --.

_.,..._-,,..mm

-j om

L t

.~.

o i

o TABLE 4.3-2 (Continued) 5;-

t E

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION r

SURVEILLANCE REQUIREMENTS

. 2-i 5

TRIP i

E ACTUATINF a

~

CHANNEL DEVICE.

MOUES FOR e

CHANNEL OPERA-OPERA-PCSTER SLAVE WHICH

$ [UNCTIONAL UNIT CHANNEL CALI-TIONAL TIONAL ACTUATION RELAY RELAY SURVEILLANCE l

m CHECK BRATION TEST TEST LOGIC TEST TEST TEST.

IS REQUIRED l

3

3. Containment Isolation a-m a.

Phase "A" Isolation i

1) M6r.ual N.A.

N.A.

R24 N.A.

N.A.

1. 2. 3. 4

2) Automatic Actuation N.A.

N.A.

M(1)

MM R

1. 2. 3. 4 Logic and Actuation Relays i

wD

3) Safety Injection See Item 1. above for all Safety Injection Surveilla ce Requirements.

w b.

Phase "B" Isolation g-

1) Manual N.A.

N.A.

R24 N.A.

N.A.

1. 2. 3. 4 2).. Autom6 tic Actuation N.A.

N.A.

M(1)

R I. 2. 3. 4 Logic and Actuation

}

cc aa Relays

3) Containment S

R24 Q

N.A.

1. 2. 3. 4 1

Pressure-High-High l

m-Containment Ventilation 3

a '

c.

gg Isolation i

gg

1) Automatic Actuation N.A.

N.A.

M(1)

R

1. 2. 3. 4

[

gg logic and Actuation gg Pelays j

i

2) Deleted i

l

==

3) Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.

I g;

4) Containment Ventilation

.??

Exhaust Rrhation-High gg (RM-44A and 448)

S R

Q N.A.

N.A.

1. 2. 3. 4 1*

2E i

11 i

EE

~1 I

!=

t i

Em t

i

l i

TABLE 4.3-2'(Continued) 1 g.

G ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRLMENTATION

n, SURVEILLANCE REQUIREE NTS y

Ei-TRIP ACTUATING CHANNEL DEVICE MODES FOR.

E CHANNEL OPERA-OPERA-MASTER SLAVE. WICH G.

CHANNEL CALI-TIONAL TIONAL ACTUATION-RELAY RELAY SURVEILLANCE i FUNCTIONAL UNil CHECK BRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED

4. Steam Line Iso 7ation a.

Manual N.A.

N.A.

R24 N.A.

N.A.

1. 2. 3 b.

Automatic Actuation Logic N.A.

N.A.

M'"

R

1. 2. 3 and Actuation Relays t'

c.

Containment Pressure-S R24 Q

N.A.

1. 2. 3

[

High-High d.

Steam Line Pressure-Low S

R24 0

N.A.

1. 2. 3 l

e.

Negative Steam Line S

R24

- 0 N.A.

N.A.

3*

l Pressure Rate-High ce 1 1 5. Turbine Trip and Feedwater Isolation ru -

a.

Automatic Actuation N.A.

N.A.

R

1. 2 Logic an' Actuation Relays p

b.

Steam Generator Water S

R24 0

N.A.

1. 2 l

a, 2 KR Level-High-High 88 E E 6. Auxiliary Feedwater ee

?.".

a.

Manual N.A.

W.A.

N.A.

R N.A.

N.A.

1. 2. 3 em

??

b.

Automatic Actuation N.A.

N.A.

R

1. 2. 3 Logic and Actuation Relays il c.

Steam Gmerator Water, T.f '

Level-Low-Low OG 1). Steam Generator S

E'4 0

N.A.

1. 2. 3*

g Water Level-Low-Low

2) RCS Loop AT Eq"ivalent N.A.

R24 0 N.A. N.A. N.A. N.A.

1. 2 3

.-gg to Power l s-e

i e-o TABLE 4.3-2 (Continued) l ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE RE0VIREMENTS [ n E i g TRIP 2

== ACTUATING t CHANNEL DEVICE MODES FOR I g CHANNEL OPERA-OPERA-MASTER SLAVE WHICH CHANNEL CALI-TIONAL TIONAL ACTUATION RELAY RELAY SURVEILLANCE

FUNCTIONAL UNIT CHECK BRATION TEST TEST LOGIC TEST TEST TFST IS REQUIRED

6. Auxiliary Feedwater (Continued) a-m d.

Undervoltage - RCP N.A. R24 N.A. R24 N.A. N.A. N.A. 1 l[ e. Safety Injectica See Item 1. above for all Safety Injection Surveillance Requirements. wk I 7. Loss of Power w i a t a. 4.16 kV Emergency 3us N.A. R N.A. R N.A. N. A. - N.A.

1. 2. 3. 4 Level 1 E E-3%

b. 4.16 kV Emergency Bus N.A. R N.A. R N.A. N.A. N.A.

1. 2. 3. 4 l

Level 2 m-

8. Engineered Safety Feature gg Actuation System Interlocks i

..EE R2 a. Pressurizer Pressure. N.A. R24 0 N.A. N.A. N.A. N.A.

1. 2. 3 l!

5E P-11 $2 b. Deleted "I La c. Reactor Trip. P-4 N.A. N.A. N.A. R24 N.A. N.A. N.A.

1. 2. 3

\\ EG TABLE NOTATIONS I 77 l i Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS. Ti "' For the Containment Ventilation Exhaust Radiation-Pigh monitor only, a CHANNEL FUNCTIONAL TEST shall be perfor: sed c at least once every 31 days. i ? T "' Trip function automatically blocked above P-11 (Pressurizer Pressure Interlock) setpoint and is automatically blocked below P-11 when Safety Injection on Steam Line Pressure-tow is not blocked. i Deleted. GC

For Mode 3. the Trip Time Delay associated with the Steam Generator Water Level-Low-Low channel must be less than or equal to 464.1 seconds.

I i l m. amu n i a u osi i. 3/4.3.3 MWITOR1WS INSTttMINTAY!ON SAbl& TION @Nffbk!WC FOR PLAWT D# tut 1DNS I t 11enTIN3 CONB1710W FOR OPIUTIDW 8.3.3.1 The radiation monitori instrueentation channels for plant operations shown in Table 3.3 8 ahall be DP 4A8LE with their Alarm / Trip Setpoints within j the specified 11eits. j )#LICABILTTY: As shown in Table 3.9-8. ) ,M: 2 a. With a radiation monitoring thannel Alarm / Trip Setpoint for plant l eparations escoeft the value shown in Table 3.3 5 ad wet the i detpoint to within he limit within 4 hours or declare thannel j inoperable. 4 j 6. With one er more radiation monitoring channels for plant operations inoperable. take the ACTION shown in Table 3.3-8. a. The previsions of Specificatten 3.0.3 are not applicable. l / svevrsti.aw:t ateutttutWTs t i 4 3.3 1 !a:6. radiatten monitoring instrumentation thannel for plant operattens shall be semenstrates OttnAlt! by the performance of the MANNtl CHICK, CHANNEL CAL 34MT30W and CMANNIL FUNCTIONAL TEST for the 20t8 and at the frequencies chown in Table 4.5 3. i i i DIARLD taNYDW x1TS 1 & 3 3/4 3-38 Amendment Nos. Il and 54 4 ---a--.--. - - - - - - - - - -. - - - - -. - - - - - -. - - - - - - - ~ ..m-,-, -w--

3/4.3 INSTRUMENTATION BASES 3/4.3.L and 3/4.3.2 REACTOR " RIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUAT. ON SYSTEM INSTRUM ENTAT ON ~ The OPERABILITY of the Reactor Trip System and Engineered Safety Features Actuation System (ESFAS) instrumentation and interlocks ensure that: (1) the I associated ACTION and/or Reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its Setpoint. (2) the specified coincidence logic and sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance consistent with maintaining an appropriate level of reliability of the Reactor-Protection and Engineered Safety Features instrumentation, and (3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and (4) sufficient system functional capabi.lity is available from diverse parameters. The OPERABILITY of these systems is recuired to provide the overall reliability, redundancy, and diversity assumed available in the facility design for the protectice and mitigation of accident and transient conditions. The integrated operation of each of these systems is consistent with the assumptions used in the accident analyses. The Surveillance Requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards. The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability. Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with WCAP 10271. " Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System." and supplements to that report. Surveillance intervals and out of service times were determined based on maintain 1hg an appropriate level of reliability of the Reactor Protection System. The Process Protection System is designed to permit any one channel to be tested and maintained dt If a channel has been bypassed for any purpose, power in a bypassed mode.the bypass is continuously room as required by applicable codes and standards. As an alternative to testing in the bypass mode. testir.g in the trip mode is also possible and permitted. The ESFA1 senses selected plant parameters and determines whether or not l predetermined limits are being exceeded. If they are, the signals are combined into logic matrices sensitive to combinations indicative of various l accidents, events, and transients. Once the required logic combination is completed. the system sends actuation signals to those engineered safety features components whose aggregate function best serves the requirements of the condition. As an example, the following actions may be initiated by the ESFAS to-mitigate the consequences of a steam line break or loss of coolant l accident: (1) safety injection pumps start and automatic valves position,- (?) Reactor trip. (3) feedwater isolation. (4) startup of the emergency diesel generators. (5) containment spray pumps start and automatic valves position. -(6) containment isolation. (7) steam line isolation. (8) Turbine trip. (9) at 111ary feedwater pumps start and automatic valve position, (10) containment fan cooler units start, and (11) component cooling water pumps start and automatic valves position. DIABLO CANYON UNITS 1 & 2 B 3/4 3-1 Unit 1 - Amendment No, M-84.122 Unit 2 Amendment No. W-83.120 i l

4 4 4 4 e4h a b f 7 b S B ? .P t t 1 1 5 u I a J h h .--4,. ... - ~ _,

h INSTRUMENTATION BASES REACTOR >ROTECTION SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM I VSIRUMEliTATION (Continued) The ESFAS Instrumentation Trip Setpoints specified in Table 3.3 4 are the nominal values at which the trips are set for each functioral unit. The Allowable Values are considred to be the Limiting Safety System Settings (LSSS) as identified in 10 CFR 50.36 and have been selected to mitigate the consequences of accidents. If the functional unit is based on analog hardware.thesetpointisconsideredtobeadjustedconsistentwiththe nominal value when the "as left" setpoint is within the band allowed for calibration tolerance. The calibration tolerance, after appropriate conversion, should correspond to the rack comparator setting accuracy defined in the latest setpoint study. For all setpoints in digital hardware, the setpoints are set at the nominal values. The ESFAS Trip Setpoints may be administrative 1y redefined in the conservative direction for several reasons including startup, testing, process error accountability, or even a conservative response for equipment malfunction or inoperability. ESFAS functions are not historically redefined at the beginning of.each cyc.le for purposes of startup or testing as several Reactor Trip functions are. However, calibration to within the defined calibration tolerance of an administratively redefined, conservative Trip Setpoint is acceptable. Redefinition at full power conditions for these functions is expected and acceptable. To accommodate the instrument drift that may occur between o>erational tests and the accuracy to which set >oints can be measured and ca'ibrated. Allowable Values for the setpoints lave been specified in Table 3.3 4. 0)eration with setpoints less conservative than the Trip Setpoint, but within tie Allowable Value is acceptable. Rack drift in excess of the Allowable i Value exhibits the behavior that the rack has not met its allowance. Since there is a small statistical chance that this will happen, an infrequent excessive drift is expected. Rack or sensor drift in excess of the allowance l that is more tha'1 occasional may be indicative of more serious problems and warrants further investigation. The methodology to derive the Tri of the uncertainties in the channel. p Setpoints is based upon combining all Inherent to the determination of the Tri) Setpoints are the magnitudes of these channel uncertainties. Sensor and ract instrumentation utilized in these channels are expected to be capable of operating within the allowances of' these uncertainty magnitudes. DIABLO CANYON - UNITS 1 & 2-B 3/4 3-la Unit 1 - Amendment No. 86-93.122 Unit 2 - Amendment No. BM2.120

INSTRUMENTAfl0N ) BASES REACTOR )R0"ECTION SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM I iSTRUME iTA"10N (Continued) ESF response times specified in Table 3.3 5. which include sequential operation of the RWST and VCT valves (Table Notations 4 and 5), are based on i values assumed in the non LOCA safety analyses. These analyses take credit l for injection of borated water from the RWST. Injection of borated water is assumed not to occur until the VCT charging pump suction isolation valves are closed following opening of the RWST charging pump suction isolation valves. When the sequential operation of the RWST and VCT valvet :s not included in the response times (Tat'le Notation 7), the values specified are based on the LOCA analyses. The LOCA analyses takes credit for injection flow regardless of the source. Verification of the response times specified in Table 3.3 5 will assure that the assumptions used for the LOCA and non LOCA analyses with i respect to the operation of the VCT and RWST vahes are valid. For slave relays in the ESF actuation system circuit that are Potter & Brumfield type MDR relays, the SLAVE RELAY TEST is performed on a refueling presented in WCAP-1387B.quency is based on relay reliability assessments" frequency. The test fre Series Relays." WCAP-13900. " Extension of Slave Relay Surveillance Test Intervals," and WCAP-14117. " Reliability Assessment of Potter and Brumfield MOR Series Relays." These reliability assesroents are relay specific and apply only to Potter and Brumfield HDR series relays. Note that for normally energ1:ed applications, the relays may have to be replaced periodically in accordance with the guidance given'in WCAP-13878 for MDR relays. Undervoltage protection will generate a loss of pow n diesel generator start in the event a loss of voltage or degraded voltage condition occurs. The diesel generators provide a source of emergency power when offsite power - is either unavailable or is insufficiently stable to allow safe unit operation. The ft/st level undervoltage relays (FLURs) detect the loss of bus L voltage (less than G t bus voltage). The second level undervoltage relays (SLURS) provide a second level of undervoltage protection which protects all Class 1E loads from short or long term degradation in the offsite power system. The SLUR allowaole value is the minimum steady state voltage needed on the 4160 volt vital bus to ensure adequate-voltage is available for safety related equipment at the 4160 volt. 480 volt, and 120 volt levels, 4 DIABLO_ CANYON - UNITS 1 & 2 B 3/4 3-lb Unit 1 - Amendment No. 4M.122 l Unit 2 - Amendment No.1-9.120 --}}

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