ML20065L260

Proposed Tech Specs Re Response Time for RTS & ESFAS Instrumentation

ML20065L260
Person / Time
Site:	Millstone
Issue date:	04/14/1994
From:	NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20065L248	List: B14745, Submits Addl Administrative Clarification to 920210 Proposed License Amend Request to Revise TS Re Response Time RTS & ESFAS Instrumentation to Be Consistent W/Guidance in Generic Ltr 93-08,per 940203 Discussion W/Nrc ML20065L260
References
NUDOCS 9404200363
Download: ML20065L260 (24)

Text

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Qpcket No. 50-423 B14745 Attachment 1 Millstone Nuclear Power Station, Unit No. 3 Proposed Revision to Technical Specifications Response Times for the Reactor Trip System and Engineered Safety Features Actuation System Instrumentation Marked-Up Pages of Technical Specifications April 1994 l

I 9404200363 940414 PDR ADOCK 05000423 P PDR

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. LIMITING' CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE Position Indiction System - Shutdown..................... 3/4 1-24 Rod Drop Time............................................ 3/4 1-25 Shutdown Rod Insertion Limit...... ...................... 3/4 1-26 (

Control Rod Insertion Limits............................. 3/4 1-27

_3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE........... .......................... 3/4 2-1 Four Loops 0perating..................................... '3/42-1 Three Loops 0perating.................................... 3/4 2-3 fI 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F (Z)..................... 3/4 2-5 Four Loops 0perating........... 9......................... 3/4 2-S Three Loops 0perating.................................... 3/4 2-12 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACT 0R................................................... 3/4 2-19 Four Loops 0perating..................................... 3/4 2-19 Three Loops 0perating.................................... 3/4 2-22 3/4.2.4 QUADRANT POWER TILT RATI0................................ 3/4 2-24 l 3/4.2.5 DNB PARAMETERS........................................... 3/4 2-27 l TABLE 3.2-1 DNB PARAMETERS........................................

3/4 2-28 }

3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION...................... 3/4 3-1 TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION................... 3/4 3-2 TABLE 3.3-2 %LE,T@inJ REACTOR SYST0i INSTRU"E"TATIO" RESP 0"SE 3/i 3TIMES.

0 ..

TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE f REQUIREMENTS............................................. 3/4 3-10  ;

3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.......................................... 3/4 3-15 TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM i INSTRUMENTATION.......................................... 3/4 3-17 l

TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM  !

INSTRUMENTATION TRIP SETP0lNTS........................... 3/4 3-26 l 1

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MILLSTONE - UNIT 3 v Amendment No. Ep, Ep,89, 0042

-10/25/9-3 INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS i

SECTION g b( EaQI TABLE 3.3-5 -ENGINEERED-SAFETY-FEATURES-RESPONSE-TIMESr. . . . . . .- . .w - 3/4-3 l TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM l INSTRUMENTATION SURVEILLANCE REQUIREMENTS............... 3/4 3-36 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring for Plant Operations............... 3/4 3-42 TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS.................................... 3/4 3-43 TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS SURVEILLANCE REQUIREMENTS.................... 3/4 3-45 i Movabl e Incore Detectors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 3-46 Seismic Instrumentation................................. 3/4 3-47 TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION................... 3/4 3-48 l TABLE 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................................ 3/4 3-49 Meteorological Instrumentation. . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 3-50 TABLE 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION............ 3/4 3-51 TABLE 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................................ 3/4 3-52 Remote Shutdown Instrumentation......................... 3/4 3-53 ,

TABLE 3.3-9 REMOTE SHUTDOWN INSTRUMENTATION...................... 3/4 3-54 l TABLE 4.3-6 REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................... 3/4 3-58 Accident Monitoring Instrumentation..................... 3/4 3-59 TABLE 3.3-10 ACCIDENT MONITORING INSTRUMENTATION................. 3/4 3-60 TABLE 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................................ 3/4 3-62 TABLE 3.3-11 DELETED Loose-Part Detection

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System............................. 3/4 3-68 Radioactive Liquid Effluent Monitoring Instrumentation.. 3/4 3-69 TABLE 3.3-12 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION ..................................... 3/4 3-70 TABLE 4.3-8 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS............... 3/4 3-72 Radioactive Gaseous Effluent Monitoring Instrumentation. 3/4 3-74 ,

MILLSTONE - UNIT 3 vi Amendment No. 84

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' ' - t- . June-8r1993 3/4.3 INSTRUMENTATIO3 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the Reactor Trip System instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE.with-RESP 0NSE.-T4MES n shown-4n Table-3+2.

APPLICABILITY: As shown in Table 3.3-1.

ACTION:

As shown in Table 3.3-1.

SURVEILLANCE RE0VIREMENTS 4.3.1.1 Each Reactor Trip System instrumentation channel and interlock and the automatic trip logic shall be demonstrated OPERABLE by the performance of the Reactor Trip System Instrumentation Surveillance Requirements specified in Table 4.3-1.

4.3.1.2 The REACTOR TRIP SYSTEM RESPONSE TIME of each Reactor trip function ,

shall be demonstrated to be within its limit at least once per 18 months.* (m~ ;

Each test shall include at least one train such that both trains are tested at \

least once per 36 months and one channel (to include input relays to both t trains) 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 Reactor trip function as shown in the " Total No. of Channels" column ,

of Table 3.3-1.  ;

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• Except that the surveillance requirement due no later than June 13, 1993, may be deferred until the next refueling outage, but no later than September 30, h

1993, whichever is earlier. (

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MILLSTONE - UNIT 3 3/4 3-1 Amendment No. AB, 79, cito

March 11, 1991 -

TABLE 3.3-2 .

gE REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES

P o M TIONAL UNIT RESPONSE TIME m

l.

xMahu 1 Reactor Trip N.A.

2. Power Raitg , Neutron Flux sO second*

3. Power Range, Reutron Flux, High Positive. Rat N.A.

4 Power Range, Neutron F1bx High Negative Rate S 0.5 second*

5. Intermediate Range, Neutron Flu N.A. Y

6. Source Range, Neutron Flux N.A. 8

7. Overtemperature AT i 7 seconds

• w lo 8. Overpower AT .

S 7 seconds *

9. Pressurizer Pressure--Low i 2 seconds

10. Pressurizer Pressur -ligh 1 2 seconds

11. Pressurizer Wa level--High 2 seconds n.

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F

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• )l6utron detectors are exempt from response time testing. Response time of the neutron 4x signal portion of the channel shall be measured from detector output or input of first electronic component in ch nnel.

8 g .$N *

, i-August /. 1 '

TABLE 3.3-2 (Continued) - -

2, .

x .

? REACTOR TRIP SYSTEN INSTRUMENTATION RESPONSE TIMES G ,

8 '

RESPONSE TIME A FUNCT L UNIT .

12. Reac Cgolant Flow--Low ,

• a. Single Loop'(Above P-8)

< 1 second

b. TwoLoops( .ve P-7 and below P-8) ,

11se ,

13. Steam Generator Water 'ev,el--Low-low seconds

\

14. Low Shaft Speed-Reactor Coolan Pumps < 0.6 second**

15. Turbine' Trip /'

a. Low Fluid 011 Pressure N.A.

{ b. Turbine Stop Valve Closure - N.A. f y 16. Safety Injection Input from E$F N.A.

17. Reactor Trip System Interlocks M.A.

18. Reactor Trip Breakers N.A.

19. Automatic Trip and Int ock Logic N. ' '

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1 Three Loop Opera n Bypass Circuitry N.A.

P -

E 9 e

}:e ,

• • Speed sensors are exempt from response time testing. Response time of the speed signal por.tlon the

? channel shall be measured from detector output or first electronic component in the channel.

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, I'NSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2 The Engineered Safety Features Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their Trip Setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4.end with 2 :P0" : TIF23 .. .nu.n :n T;t1; :.0 0, APPLICABILITY: As shown in Table 3.3-3.

ACTION:

a. With an ESFAS Instrumentation or Interlock Trip Setpoint trip less conservative than the value shown in the Trip Setpoint column but more conservative than the value shown in the Allowable Value column of Table 3.3-4, adjust the Setpoint consistent with the Trip Setpoint value.

b. With an ESFAS Instrumentation or Interlock Trip Setpoint less conserva-tive than the value shown in the Allowable Value column of Table 3.3-4, either:

1. Adjust the Setpoint consistent with the Trip Setpoint value of Table 3.3-4, and determine within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that Equation 2.2-1 was satisfied for the affected channel, or

2. Declare the channel inoperable and apply the applicable ACTION statement requirements of Table 3.3-3 until the channel is restored to OPERABLE status with its Setpoint adjusted consistent with the Trip Setpoint value.

Equation 2.2-1 2 + R + 5 < TA Where:

2 = The value from Column Z of Table 3.3-4 for the affected channel, R = The "as measured" value (in percent span) of rack error for the affected channel, S = Either the "as measured" value (in percent span) of the sensor error, or the value from Column 5 (Sensor Error) of Table 3.3-4 for the affected channel, and TA = The value from Column TA (Total Allowance) of Table 3.3-4 for the affected channel.

c. With an ESFAS instrumentation channel or interlock inoperable, take the ACTION shown in Table 3.3-3.

MILLSTONE - UNIT 3 3/4 3-15

April 9, 1987 TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES NITIATION $!GNAL AND FUNCTION RESPONSE TIME IN SECO S

1. Manual Initiation Safety Injection (ECCS) N.A.

b.

Contai.mert Spray N. A.

7

c. hase "A" Isolation N. A.

d. Ph e "B" Isolation N. A.

e. Stea Line Isolation N.A.

f. Feedwat- Isolation N. A.

g. Auxiliary eedwater N.'k.

h. ' Service Wat !N.A.

i. Control Buildi Isolation / N.A.

j. Reactor Trip H. A.

k. Start Diesel Gener or / N.A.

2. Containment Pressure--High r

/

a. Safety Injection (ECCS) .

1 27I3)/12(10)

1) Reactor Trip '

<2 *

2) FeedwaterIsolation/ 6. 8 I3)

3) Phase "A" Isolation # ~

2(2)(6)/12(1)(6)

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4) Auxiliary Feedwater i 60

5) Service Water ' I3) s 90

6) Start Diesel Generator .

< 12

b. ControlBuild[ngIsolation _5 j

3. Pressurizer Pressure--Low

a. Safety Injection (ECCS) i 27I9)/12(10)

1) j Reactor Trip i2 2 < 6. I3)

Jf)/Phase Teedwater "A" Isolation Isolation i2 I2) 3/12(1)(6) .

4) Auxiliary Feedwater i 60

5) Service Water 1 90 II)

6) Start Diesel Generators i 12 I )

MILLSTONE - UNIT 3 3/4 3-32 Amendment No: 3

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, .. b l l April 9, 1987 i

, TABLE 3.3-5 (Continued) l ENGINEERED SAFETY FF ATURES RESPONSE TIMES ITIATING SIGNAL AND FUNCTION RESPONSE TIME IN $ CONDS

4. Steam Line Pressure--Lov /

Safety Injection (ECCS) < 27(5)f37(4)

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1) Reactor Trip <2 /

') Feect ater Isolation 6.8 I) 3 Phase "A" Ifniation 1,2(2)(6)/12(1)(6)

4) uxiliary Feedwater ,

,< 60

5) 5 vice Water / $ SOII)

6) Sta Diesel Generators / < 12

b. Steam Line olation -

/ 6.8 I3)

5. Containment Pressu --High-3 ,

a. Quench Spray .' < 32(2)/42II)

b. Phase "B" Isolatio c 2(2)(6)jy2(1)(6)

c. Motor-Driven Auxilia Feeddater -

< 140 Pumps /

d. . Service Water ,- < 90 II)

6. ContainmentPressure--Hig$-2

a. SteamLineIsolatibn < 6.8 I1)

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7. SteamLinePressurej - Negative Rate-. igh

a. SteamLine,I[olation 1 6.8( )

8. Steam Generato Water level--High-High

a. Turbioe# Trip < 2.5

b. FeecdaterIsolation 6.8 I3)

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9. Steam' Generator Water Level--Low-Low

/a /' Feedwater Motor-Driven Pumps Auxiliary < 60

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!b. Turbine-Driven Auxiliary Feectwater Pump i 60

10. Loss-of-Offsite Power

a. Motor-Driven Auxiliary Feedwater Ptap i 60

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Mill 5 TONE - UNIT 3 3/4 3-33 . Amendment No. 3 1

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February 16, 1908 4

TABLE 3 3-5 (Continued)

,s ENGINEERED SAFETY FEATURES RESPONSE TIMES j

j' INITIATING S AL AND FUNCTION RESPONSE TI}E IN SECONDS

/

11. Loss of Power /

a. 4 kV Bus Under 1tage / 6 13 (Loss of Voltage ,,-

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b. 4 kV Eraergency Bus ,

6 18(7)/310(B)  !

Undervoltage (Grid j  !

Degraded Voltage)/

j .}'

12. Tavg Low Coincident With 4

ReactorTrip({7)

a. Feedwate Isolation f- 12(3) l 13 Control, Building Inlet Ventilation Radiation

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a./ Control Building Isolation .637 MILLSTONE - UNIT 3 3/4 3-34 Amendment No.Ilie72 l4

Ipril 9, 1987 l

\N TABLE 3.3-5 (Continued)

TABLE: NOTATIONS )

esel generator starting and sequence loading delayv neluded.

N((1)

2) Die Offsi ower available.

(3) Air-opera d valves. /,

/

generator starting and sequence loading delay not included.

(4) Diesel gener tor starting and sequencev' leading delays included.

Secuential trah tfer of Charging pump' suction from the VCT to the RHR

• RWST (RWST valveg open, then VCT valves close) is included.

pumps g included - - r'

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(5) Diesel generator sta ting and sequence loading delays not included.

Secuential transfer of Charging pump suction from the VCT to the RHR RWST (RWST valves open,'.then VCT valves close) is included.

pumps not included. /

(6) Time required to c)o'se valve as indicated in Table 3.6-2.

(7) With an ESF signal present.

/

(8) Without an EST signal present.

/ eading delays included.

(9) Diesel generator starting and sequence l'tjon from the VCT to the Sequent,ial transfer of Charging pump suc RWST,(RWST valves open, then VCT valves c1qse) is not included.

Response time assures only opening of RWST valves.-

iesel generator starting and sequence leadin elays not included.

(10) Sequential transfer of charging purs; suction fr(g the VCT to th o

RWST (RWST valves open, then VCT valves close) is%et included. .

RHR puc:ps not included.

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3/4 3-35 Amendment No. 3 MILLSTONE - UNIT 3  ;

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g g 3 INSTRUMEdATION '

A g,gBASES REACTOR TRIP SYSTEM INSTRUMENTATION and ENGINEERED SAFETY-FEATURES ACTUATION

c. g p 5Y5i ? Ih:i W ENTATION (Continueo)

'h dthe sensor from its calibration point or the value specified in Table 3.3-4, p Win percent span, from the analysis assumptions. Use of Equation 3.3-1 allows

~g.

QSfor a sensor drift factor, an increased rack drift factor, and provides a threshold value for REPORTABLE EVENTS.

<S . 'e me:%::::legv to derive the Trip Setpoints is based upon comoining all Inherent to the determination of the 4 of the uncertainties rip Setocints in magnituces are the the enannels.

of these channel uncertainties. Sensor and.

~

' Q., , ack instrumentation utilized in these channels are expected to be capable of W operating within the allowances of these uncertainty magnitudes. Rack drift *

, 4) D in excess of the Allowable Value exhibits the behavior that the rack has not et its allowance. Being that there is a small statistical-chance that this Y ill happen, an infrequent excessive drift is expected. Rack or sensor drift, C in excess of the allowance that is more than occasional, may be indicative of D ,4 $ feore serious problems and should warrant further investigation.-

M T--

M4 The measurement of response time at the specified frequencies provides assurance that the Reactor trip'and the Engineered Safety Features actuation g associated with each cnannel is completed within-the time limit assumed in the- '

thd safety analysesJ No credit was taken in the analyses for those channels with

, response times indicated as not applicable. Response time may be demonstrated d ;5 by any series of sequential, overlapping, or total channel test measurements

.C provided that such tests demonstrate the total channel response time as defined.

~5 Sensor response time verification may be demonstrated by either:- (1) in place, onsite, or offsite test measurements, or (2)'util12ing replacement sensors with certified response time. Detector response times may be measured by the in situ on line noise analys'is-response time degradation method described in the. Westinghouse Topical Report, "The Use of Process Noise Measurements To Determine Response Characteristics of Protection Sensors in U.S. Plants,"

August 1983.

~

ESF response time specified in Table 3.3-5 which include sequential opera - .i tion of the RWST and VCT valves are based on values assumed in the non-LOCA safety analyses. For these analyses, injection of. borated water from the RWST is credited. Injection of borated water is assumed not to occur until the VCT charging pump suction valves are closed following opening of the RWST charging 1 pump suction valves. When the sequential operation of'the RWST and VCT valves is not included in the response time, the values specified'are based on the . i LOCA analyses which credit injection flow regardless of the source. Exceptions.

to this rule are the response times with table notation 10. These response times do not include sequential operation of the RWST and VCT isolation valves f but are cerived from the non-LOCA analyses. Theses exceptions insure that safety injection pumps (except RHR) are started within an appropriate time.when offsite power is present. Since SI functions are identical =regardless of -the '

actuation signal, the individual component' verification will assu w that the.. .

response times specified with and without sequential operation of the VCT and:

RWST valves are met'for LOCA and non-LOCA accidents.

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MILLSTONE - UNIT 3 8 3/4 3-2 'AmendmentrNo. 3 1

l

l Docket No. 50-423 l B14745 l Attachment 2 Millstone Nuclear Power Station, Unit No. 3 Proposed Revision to Technical Specifications Response Times for the Reactor Trip System and Engineered Safety Features Actuation System Instrumentation Retyped Pages of Technical Specifications s

April 1994

r INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANPE REQUIREMENTS SECTION PAGE Position Indiction System - Shutdown..................... 3/4 1-24 Rod Drop Time............................................ 3/4 1-25 .

Shutdown Rod Insertion Limit............................. 3/4 1-26 l Control Rod Insertion Limits . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-27 3/4.2 POWER DISTRIBUTION LIMITS l

3/4.2.1 AXIAL FLUX DIFFERENCE.................................... 3/4 2-1 Four Loops 0perating..................................... 3/4 2-1 Three Loops 0perating.................................... 3/4 2-3 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - Fa(Z)..................... 3/4 2-5 Four Loopt 0perating..................................... 3/4 2-5 ,

Three Loops 0perating.................................... 3/4 2-12 1

3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACT 0R................................................... 3/4 2-19 Four Loops 0perating..................................... 3/4 2-19 Three Loops 0perating.................................... 3/4 2-22 3/4.2.4 QUADRANT POWER TILT RATI0................................ 3/4 2-24 3/4.2.5 DNB PARAMETERS........................................... 3/4 2-27 l 1

TABLE 3.2-1 DNB PARAMETERS........................................ 3/4 2-28 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION...................... 3/4 3-1 TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION................... 3/4 3-2 TABLE 3.3-2 DELETED l TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................................. 3/4 3-10 i

3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.......................................... 3/4 3-15 1

TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.......................................... 3/4 3-17 I TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETP0lNTS........................... 3/4 3-26 MILLSTONE - UNIT 3 v Amendment No. Jip, pp. 77.

0197

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JEQEX LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION EME TABLE 3.3-5 DELETED TABLE 4.3-2 ENGINEERED SAFELY FEATURES ACTUATION SYSTEM l INSTRUMENTATION SURVEILLANCE REQUIREMENTS . . . . . . . 3/4 3-36 3/4.3.3 MONITORING INSTRUMENTATION l Radiation Monitoring for Plant Operations . . . . . . . 3/4 3-42 TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS . . . . . . . . . . . . . . . . . 3/4 3-43 TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS SURVEILLANCE REQUIREMENTS . . . . . . . . . 3/4 3-45 Movable Incore Detectors . . . . . . . . . . . . . . . 3/4 3-46 Seismic Instrumentation . . . . . . . . . . . . . . . . 3/4 3-47 TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION . . . . . . . . . . 3/4 3-48 TABLE 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS . . . . . . . . . . . . . . . . . . . . . 3/4 3-49 Meteorological Instrumentation . . . . . . . . . . . . 3/4 3-50 TABLE 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION . . . . . . . 3/4 3-51 1ABLE 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS . . . . . . . . . . . . . . . . . . . . . 3/4 3-52 Remote Shutdown Instrumentation . . . . . . . . . . . . 3/4 3-53 TABLE 3.3-9 REMOTE SHUTDOWN INSTRUMENTATION . . . . . . . . . . . . 3/4 3-54 TABLE 4.3-6 REMOTE SHUTDOWN HONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS . . . . . . . . . . . . . . . 3/4 3-58 Accident Monitoring Instrumentation . . . . . . . . . . 3/4 3-59 TABLE 3.3-10 ACCIDENT MONITORING INSTRUMENTATION . . . . . . . . . 3/4 3-60 TABLE 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS . . . . . . . . . . . . . . . . . . . . 3/4 3-62 TABLE 3.3-11 Df.LETED Loose-Part Detection System . . . . . . . . . . . . . 3/4 3-68 Radioactive Liquid Effluent Monitoring Instrumentation 3/4 3-69 TABLE 3.3-12 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION . . . . . . . . . . . . . . . . . . . 3/4 3-70 TABLE 4.3-8 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS . . . . . . 3/4 3-72 Radioactive Gaseous Effluent Monitoring Instrumentation 3/4 3-74 MILLSTONE - UNIT 3 vi Amendment No. 77, 0197

3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION l

1TNTTfNo OnNnfTinN FOR OPFDATTON 3.3.1 As a minimum, the Reactor Trip System instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE.

APPLICABILITY: As shown in Table 3.3-1. ,

l ACTION: j As shown in Table 3.3-1.

AllRVFTII ANcF DFntliRFMFMTR 4.3.1.1 Each Reactor Trip System instrumentation channel and interlock and the automatic trip logic shall be demonstrated OPERABLE by the performance of the Reactor Trip System Instrumentation Surveillance Requirements specified in  ;

Table 4.3-1. l 4.3.1.2 The REACTOR TRIP SYSTEM RESPONSE TIME of each Reactor trip function shall be demonstrated to be wi+ bin its limit at least once per 18 months.*

Neutron detectors and speed se.aors are exempt from response time testing. l Each test shall include at least one train such that both trains are tested at least once per 36 months and one channel (to include input relays to both trains) 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 Reactor trip function as shown in the " Total No. of Channels" column of Table 3.3-1.

i s

• Except that the surveillance requirement due no later than June 13, 1993, may be deferred until the next refueling outage, but no later than September 30, 1993, whichever is earlier.

MILLSTONE - UNIT 3 3/4 3-1 Amendment No. JJ. 75 0198

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MILLSTONE - UNIT 3 3/4 3-8 AmendmentNo.JJ.6f seis  ;

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MILLSTONE - UNIT 3 3/4 3-9 0048 AmendmentNo.[

INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2 The Engineered Safety Features Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their Trip Setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4.

APPLICABILITY: As shown in Table 3.3-3.

ACIL0fi:

a. With an ESFAS Instrumentation or Interlock Trip Setpoint trip less conservative than the value shown in the Trip Setpoint column but more conservative than the value shown in the Allowable Value column of Table 3.3-4, adjust the Setpoint consistent with the Trip Setpoint value.

b. With an ESFAS Instrumentation or Interlock Trip Setpoint less conservative than the value shown in the Allowable Value column of Table 3.3-4, either:

1. Adjust the Setpoint consistent with the Trip Setpoint value of Table 3.3-4, and determine within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that Equation 2.2-1 was satisfied for the affected channel, or

2. Declare the channel inoperable and apply the applicable ACTION statement requirements of Table 3.3-3 until the channel is restored to OPERABLE status with its Setpoint adjusted consistent with the Trip Setpoint value.

Equation 2.2-1 Z + R + 51 TA Where:

2 - The value from Column Z of Table 3.3-4 for the affected channel, R - The "as measured" value (in percent span) of rack error for the affected channel, S - Either the "as measured" value (in percent span) of the sensor error, or the value from Column S (Sensor Error) of Table 3.3-4 for the affected channel, and TA - The value from Column TA (Total Allowance) of Table 3.3-4 for the affected channel. l

c. With an ESFAS instrumentation channel or interlock inoperable, take the ACTION shown in Table 3.3-3.  ;

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INSTRUMENTATION BASES REACTOR TRIP SYSTEM INSTRUMENTATION and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued) the sensor from its calibration point or the value specified in Table 3.3-4, in percent span, from the analysis assumptions. Use of Equation 3.3-1 allows for a sensor drift factor, an increased rack drift factor, and provides a threshold value for REPORTABLE EVENTS.

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. Sensor and rack instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes. Rack drift in excess of the Allowable Value exhibits the behavior that the rack has not met its allowance. Being that 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 should warrant further investigation.

The measurement of response time at the specified frequencies provides assurance that the Reactor trip and the Engineered Safety Features actuation associated with each channel is completed within the time limit assumed in the safety analyses. The RTS and ESF response times are included in the Operating Procedure OP-3273 " Technical Requirements--Supplementary Technical Specifications." Any changes to the RTS and ESF response times shall be in accordance with Section 50.59 of 10CFR50 and approved by the Plant Operations Review Committee. No credit was taken in the analyses for those channels with response times indicated as not applicable. Response time may be demonstrated by any series of sequential, overlapping, or total channel test measurements provided that such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either:

(1) in place, onsite, or offsite test measurements, or (2) utilizing replacement sensors with certified response time. Detector response times may be measured by the in situ on line noise analysis-response time degradation method described in the Westinghouse Topical Report, "The Use of Process Noise Measurements To Determine Response Characteristics of Protection Sensors in U.S. Plants," August 1983. ,

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i MILLSTONE - UNIT 3 8 3/4 3-2 Amendment No. 3, 0071

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