Text

Proposed Tech Specs Implementing line-item Improvements Provided in GL 93-08 for Removal of Response Times from TS
	ML20217A980
Person / Time
Site:	Beaver Valley
Issue date:	09/11/1997
From:	; DUQUESNE LIGHT CO.
To:	;
Shared Package
ML20217A970	List: L-97-009, Application for Amends to Licenses NPF-66 & NPF-73,changing TS to Implement line-item Improvements Provided in GL 93-08 for Removal of Response Times from TS; ML20217A980;
References
	GL-93-08, GL-93-8, NUDOCS 9709230111
	Download: ML20217A980 (50)
	v • d • e

_ _ _ _ _ _ _ _ _ _ _ _ _ _

n ATTACHMENT A-1 Beaver Valley Power Station, Unit No. 1 Proposed Technical Specification Change No. 217 The following is a list of the affected pages:

Affected Pages:

3/4 3-1 3/4 3-7 3/4 3-9 3/4 3-10 3/4 3-14 3/4 3-24b 3/4 3-25 3/4 3-26 3/4 3-27 3/4 3-27a 3/4 3-28 B 3/4 3-la 9709230111 970911 PDR ADOCK 05000334 P

PDR

DPR-66 if4.3 INSTRUMENTATION 3 / 4. 3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1.1 As a

minimum, the reactor trip system instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE.with RE;PON;; TIM 0 es ehern in Table 3.3"Fr APPLICABILITY:

As shown in Table 3.3-1.

ACTION:

As shown in Table 3.3-1.

SURVEILLANCE REQUIREMENTS 4.3.1.1.1 Each reactor trip system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the modes and at the frequencies shown in Table 4.3-1.

4.3.1.1.2 The logic for the interlocks shall be demonstrated OPERABLE during the at power CHANNEL FUNCTIONAL TEST of channels affected by interlock operation.

The total interlock function shall be demonstrated OPERABLE at least once po't 18 months during CHANNEL CALIBRATION testing of each channel affected by interlock operation.

4.3.1.1.3 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, Each test shall include at least one logic l

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 very N times 18 months where N is the total number of redundant ch nnels in a specific reactor trip function as shown in the " Total No.

f Channels" column of Table 3.3-1.

ron detectors are exempt from response time BEAVER VALLEY - UNIT 1 3/4 3-1 (Proposed Wording)

s

'DPR-66 TABLE 3.3-1 (Continued)

ACTION 8 -

With the number of OPERABLE channels one less than the Total Number of Channels and with the THERMAL POWER level above P-7, place the inoperable channel in-the tripped condition within 6

hours; operation may continue until performance of the next required CHANNEL FUNCTIONAL TEST.

ACTION 9 -

Not applicable.

ACTION 10 -

Not applicable.

ACTION 11 -

With less than the Minimum Number of Channels OPERABLE, operation may continue provided the inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 12 -

With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and/or open the reactor trip breakers.

ACTION 39 -

With the number of OPERABLE channels one-less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the reactor trip breakers within the next hour.

ACTION 40 -

a.

With one of the diverse trip features (undervoltage or shunt trip attachment) of a reactor trip breaker inoperable, restore the diverse trip feature to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or declare the breaker inoperable and be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Neither breaker shall be bypassed while one of the diverse trip features is inoperable except for the time required for performing maintenance to restore the breaker to OPERABLE status.

b.

With one reactor trip breaker inoperable as a result of - something other than an inoperable diverse trip feature; be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

however, one channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months - for surveillance testing per Specification 4.3.1.1.1, provided the other channel is OPERABLE.

BEAVER VALLEY - UNIT 1, 3/4 3-7

//

Amendment No. -t99-(next page is 3/4 3-)()

(htcf M NMkl

o TABLE 3.3-2 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES g

e FUNCTIONAL UNIT RESPONSE TIME g

1.

' Manual Reactor Trip NOT APPLICABLE 2.

Power Range, Neutron Flux

$ 0.5 seconds -*[O 3.

Power Range, Neutron Flux, fffjjtj"j1/fgftg High Positive Rate 4.

Power Range, Neutron Flux, High Negative Rate 5 0.5' seconds -*[U 5.

Intermediate Range, Neutron Flux NOT APPLICABLE 6.

Source Range, Neutron Flux NOT APPLICABLE 7.

Overtemperature Delta T

$ 6.0 seconds -*d) 8.

Overpower Delta T 5 6.0 secondss*k) 9.

Pressurizer Pressure -- Low 1 2.0 seconds 10.

Pressurizer Pressure -- High 5 2.0 seconds 11.

Pressurizer Water Level -- High NOT APPLICABLE

[0-+

Neutron detectors are exempt from response time testing. Response time shall be measured from detector output or input of first electronic component in channel.

I$EAVER VALLEY - UNIT 1 3/4 3-9 (delete this page)

-- - = - - - --

a.

TABLE 3.3-2 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES w

ee*

FUNCTIONAL UNIT RESPONSE TIME 1

12.

Loss of Flow - Single Loop (Above P-8)

$ 1.0 seconds 13.

Loss of Flow - Two Loops (Above P-7 and below P-8)

$ 1.0 seconds 14.

Steam Generator Water Level--Low-Low 1 2.0 seconds 15.

Steam /Feedwater Flow Mismatch and Low Steam Generator Water Level NOT APPLICABLE 16.

Undervoltage-Reactor Coolant Pumps' i 1.2 seconds 17.

Underfrequency-Reactor Coolant Pumps 1 0.6 seconds 18.

Turbane Trip l

I A.

Auto Stop Oil Pressure NOT APPLICABLE l

B.

Turbine Stop Valve NOT APPLICABLE I

19.

Safety Injection Input from ESP NOT APPLICAB13 20.

Reactor Coolant Pump Breaker Position Trip NOT APPLICABLE 13EAVER VALLEY - UNIT 1 3/4 3-10 (delete this page)

_ _ =.-

s OPR-66 INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2.1 The engineered safety feature actuation system instrumentation channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their trip setpoints set consistent with th6 values shown in the, Trip setpoint column of Table 3.3-4. :.md uith RESPC" E APPLICABILITY:

As shown in Table 3.3-3.

ACTION:

a.

With an engineered safety feature actuation system instrumentation channel trip setpoint less consersative than the value shown in the Allowable Values column of Table 3.3-4, declare the channel inoperable and apply the applicable ACTION requirement of Table 3.3-3 until the channel is restored to OPERABLE status with the trip setpoint adjusted consistent with the Trip Setpoint Value, b.

With an engineered safety feature actuation system instrumentation channel inoperable, take the action shown in Table 3.3-3.

SURVEILLANCE REQUIREMENTS 4.3.2.1.1 Each engineered safety feature actuation system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the modes and at the frequencies shown in Table 4.3-2.

4.3.2.1.2 The logic for the interlocks shall be demonstrated OPERABLE during the at power CHANNEL FUNCTIONAL TEST of channels affected by interlock operation.

The total interlock function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by interlock operation.

4.3,2.1.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESF function 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 per N

times 18 months where N is the total number of redundant channels in a specific ESF function as shown in the " Total No. of Channels" Column of Table 3.3-3.

BEAVER VALLEY - UNIT 1 3/4 3-14 (Proposed Wording)

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DPR-66 Tintt 3.3-s ENCTNEF9FB EAFETY FEATURks RESPONSE TIMrs INITTATING STCNAL AND FUNCTION REEDONSF TTMF TN SECONDS 1.

Manual a.

Safety Injection (ECCS)

Not Applicable Feedwater Isolation Not Applicable Reactor Trip (SI)

Not Applicable containment Isolation-Phase "A"

Not Applicable Containment Vent and Purge Isolation Not Applicable Auxiliary Feedwater Pumps Not Applicable Rx Plant River Water System Not Applicable b.

Containment Quench Spray Pumps Not Applicable Containment Quench Spray Valves Not Applicable containment Isolation-Phase "B"

Not Applicable c.

Containment Isolation-Phase "A"

Not Applicable d.

Control Rcom Ventilation Isolation Not Applicable 2.

cantain==nt o===ura-vish r

a.

Safety Injection (ECCS) s 27.

b.

Reactor Trip (from SI) 1 3.0 c.

Feedwater Isolation 1)

Feedwater Regulating Valves 1 10. 0 (+)6 2)

Feedwater Bypass Valves s 30.0(4)s d.

Containment Isolation-Phase "A" 122.0h/33.0h e.

Auxiliary Feedwater Pumpe Not Applicable f.

Rx Plant River Water System 177.0h)/110.0th BEAVER VAILEY - UNIT 1 3/4 3-25 Amendment No. 4 Ueleh (L.i pp)

0 DPR-66

~

TABLE 3.3-5 (continued)_

ENOINEEDFB SAFETY FEATURES REspoNsr TIMES INITIATING SIGNAY. AND FUNCN'IOJ REspoNEE TIMF IN SECONDS 3.

Pramaurirar Drasmura-Low Safety Injection (ECCS) a.

s 27.

/27.

b.

Reactor Trip (from SI) s 3.0 c.

Feedvater Isolation 1)

Feedvater Regulating Valves d

2)

Feedwater Bypass Valves s 10.0(t)6 s 30.0(+)

d.

Containment Isolation-Phase "A"

1 22.0(4)f Auxiliary Feedwater Pumps e.

Not Applicable f.

Rx Plant River Water System 1 77. 0 h /110. 0 [)

i e

BEAVER VALLEY - UNIT 1 3/4 3-26 Amendment No. 4

~

6ltltktYluij'aye)

DPR-66 TAar r 3. 3-s (continuedt ENGTNEFRED SAFETY FEATUREE REEDONSE TIMEE INITTATING STCNAL AND FUNCTION REEDONCE TTME IN SECONDS

4. Stamm Line Pressura-Lew a.

Safety Injection (EccS) s 27.0 37.0 b.

Reactor Trip (from SI) s 3.0 c.

Feedwater Isolation 1)

Feedwater Regulating Valves 1 10. 0 (-i-) $

2)

Feedwater Bypass Valves s 30.0(+)4 r

d.

Containment Isolation-Phase "A" 1 22.0(4)/33.0(4)

Auxiliary Feedwater Pumps Not Applicable e.

f.

Rx Plant River Water System 5 77.0

/110.0h g.

Steam Line Isolation s 4.0

5. cantain= ant Pramaure--Hiah-Minh a.

Containment Quench Spray s45.0(k b.

Containment Isolation-Phase "B"

Not Applicable c.

control Room Ventilation Isolation 1 22.0(

/77.0(h)

6. Stan= canavatar Watar Taval--Minh-Minh a.

Turbine Trip-Reactor Trip s 2.5 (Above P-9) b.

Feedwater Isolation 1)

Feedwater Regulating Valves s 10.0(414 2)

Feedwater Bypass Valves s 30.0(t)4

7. CgTAin= ant Pr===ure--Intm H inta wish-Minh a.

Steam Line Isolation 1 4.0

8. Rema=1ina Pr===ura Rata--Minh Maantiva a.

Steaaline Isolation s 8.0 9.

Lama of power a.

4.16kv Emergency Bus Undervoltage i 1.3 (Loss of voltage) b.

4.16kV and 480v Emergency Bus s 95 Undervoltage (D1 graded voltage)

BEAVER V.'t. LEY - UNIT 1 3/4 3-27 Amendment No. 4 5-Gelde A;yye)

DP,R-66 TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

/S g. Stdam Generator Water Level-Low-Low Motor-driven Auxiliary 60.0 a.

Feedwater Pumps **(N b.

Turbine-driven Auxi ary 60.0 Feedwater Pumps ***- (

//p. Undervoltace RCP a.

Turbine-driven Auxiliary 60.0 Feedwater Pumps

/,2y.Emercancy Bus Undervoltage a.

Motor-driven Auxiliary 60.0 Feedwater Pumps 9

/,I J4. Trio of Main Feedwater Pumns a.

Motor-driven Auxiliary 60.0 Feedwater Pumps NOTE:

Response time for Motor-driven Auxiliary 60.0 Feedwater Pumps on all S.I.

signal starts (I) **'

on 2/3 any Staan Generator (y

on 2/3 in 2/3 Steam Generators BEAVER VALLEY - UNIT 1 3/4 3-27a Amendment No. M--

(clelete % fyd

OPR-66 TARI,I 3.3-5 (Continued 1 TABIE NOTATTON

{3)-

Diasel generator starting and sequence loading delays included. Response time limit includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging pumps and Low Head Safety Injection Sequential transfer of charging pump suction from pumps.

the volume control tank (VCT) to the refueling water storage tank (RWST)

(RWST valvos open, then VcT valves close) is not included.

[4;/).-#-

Diesel gen rator starting and sequence loading delays nat included.

Offsite power available.

Response time limit includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging Sequential transfer of charging pump suction from pumps.

the volume control tank (VcT) to the refueling water storage tank (RWST)

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

{@

Diesel generator starting and sequence loarting delays included. Response time limit includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging pumps.

!;equential transfer of charging pump suction from the volume control tank (Ver) to the refueling water storage tank (RNST) (RWST valves open, then VCT valves close) is included.

h Feedwater isolation includes signal responra *:.0 valve closure time.

h Diesel generator starting and sequence loading delays inclu/Jed.

T (t) Diesel generator starting and sequence loading delays nat included.

BEAVER VALLEY - UNIT 1 3/4 3-28 Amendment No. 4 6/e/e4 Ai pye)

4 DPR-66 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 PROTECTIVE AND 'ENGINEERED SAFETY FEATURES (ESF)

INSTRUMENTATION (Continued) fdng q pry vigt Qitensope Kt9m%of sk Sec A )., a af 4 41 )%uf ESF response times :; :i'i:d 1. T:M: 2.7 :~whic operation of the RWST and VCT valves (!!CT:0 # g include sequential l values assumed in the Non-LOCA safety analysesf

..d ff; are based on These analyses take l credit-for injection of borated water.

Initial borated water is supplied by the BIT, however, injection of borated water from the RWST is assumed not to occur until-the VCT charging ptag suction valves are closed following opening of the RWST charging pump auction valves.

When sequential operation of the RWST and VCT valves is not

- included in the response times -( ::::

^),

the values specified are based on the LOCA analyses.

The LOCA analyses take credit for injection flow regardless of the source.

Verification of the response times -:;;;ifi:4 i r.

T:M: -.:

will assure that the assumptions used for the LOCA and Non-LOCA analyses with respect to operation of the VCT and RWST valves are valid.

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 times.

The Engineered Safety Feature Actuation System interlocks perform the following' functions:

P-4 Reactor tripped Actuates turbine trip, closes main feedwater valves on Tavg below setpoint, prevents the opening of the main feedwater valves which were closed by a safety injection or high steam generator water level signal, allows safety injection block so that components can be reset.or tripped.

Reactor not tripped - prevents manual block of safety injection.

P-11 Above the setpoint P-11 automatically reinstates safety injection actuation on low pressurizer-

pressure, automatically blocks steamline isolation on high steam pressure rate, enables safety injection and steamline isolation on low steamline pressure (with Loop Stop valves y

open), and enables auto actuation of the pressurizer PORVs.

BEAVER VALLEY - UNIT 1 B 3 3-la Amendment No. :'

(l%pese/ 4d Wod@)

ATTACHMENT A-2 Beaver Valley Power Station, Unit No. 2 Proposed Technical Specification Change No. 84 The following is a list of the affected pages:

Affected Pages:

3/4 3-1 3/4 3-7 3/4 3-8 3/4 3-9 3/4 3-14 3/4 3-15 3/4 3-28 3/4 3-29 3/4 3-30 3/4 3-31 3/4 3-32 B 3/4 3-2

NPF-73 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1.1 As a minimum, the reactor trip system instrumentation channels and interlocks of Table 3.3-1 shall be OPERABt.E.with RCL^^ZC TIMCS ;; ehear, ir, APPLICABILITY: As shown in Table 3.3-1.

ACTION:

As shown in Table 3.3-1 SURVEILLANCE RE0VIREMENTS 4.3.1.1.1 Each reactor trip system instrumentation channel and interlock and automatic trip logic shall be demonstrated OPERABLE by the p ormance of the Reactor Trip Systen Instrumentation Surveillance Requirement during the MODES I

and at the frequencies shown in Table 4.3-1.

(p 4.3.1.1.2 The logic for the interlocks shall be demonstrated OPERABLE during the at power CHANNEL FUNCTIONAL TEST of channels affacted by interlock opera-tion.

The total interlock function shall be demonstrated OPERABLE at least once per 18 months durin by interlock operation.'g CHANNEL CALIBRATION testing of each channel affected de n 4.3.1.1.3 The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip fu n

shall be demonstrated to be within its limit at least once per 18 month *.

Each I

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 reactor trip function as shown in the " Total No. of Channels" column of Table 3.3-1.

htrondetectorsareexemptfromresponsetimetesting dele.ke 64For the automatic trip logic, the surveillance requirements shall be the I

application of various simulated input combinations in conjunction with each possible interlock logic state and verification of the required logic output ludin k as a sinimu_s, a continuity check of output devices.

"The specified 18-month s_urfeillance ditervai~ durEg the first fuel cycle may be extended to coincide with completion of the first refueling outage.

BEAVER VALLEY - UNIT 2 3/4 3-1 Amendment No.

(Proposed Wording)

4 NPF-73 TABLE 3.3-1-(Continued)

ACTION 8 -

With the number of OPERABLE channels one less than the Total Number of Channels and with the THERMAL POWER level above P-9, place the inoperable channel in the tripped condition within 6

hours; operation may-continue until performance of the next required CHANNEL FUNCTIONAL TEST.

ACTION 9 -

This Action is not used.

ACTION 10 -

This Action is not used.

ACTION 11 -

With less than the Minimum Number of Channels OPERABLE, operation may continue provided the inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 12 -

With the number of channels OPERABLE one lesw than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and/or open the reactor trip breakers.

ACTION 39 -

With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement ~,

restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the reactor trip breakers within the next hour.

ACTION 40 -

a.

With one of the diverse trip features (undervoltage or shunt trip attachment) of a reactor trip breaker inoperable,- restore the diverse trip feature to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or declare the breaker inoperable and be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Neither breaker shall be bypassed while one of the diverse trip features.is inoperable except for the time required for performing maintenance to restore the breaker to OPERABLE status.

b.

With one reactor trip ' breaker inoperable as ' a result of something other than an inoperable diverse trip feature, be in at least NOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.1.1.1, provided the other channel is OPERABLE.

ACTION 44 -

With less than the Minimum Number of Channels OPERABLE, within

-1 hour determine by observation of the associated permissive annunciator window (s) that the interlock is in its required state for the existing plant condition, or apply Specification 3.0.3.

3 Amendment No. (*123$ pag /4 3:-7e a 3/V 3-/c)

BEAVER VALLEY - UNIT 2 (Ngm/ Ad@)

NPF-73 T_ABLE 3.3-2 REACTOR TRIP SYSTEM INSTRUMENTAT!0N RESPONSE TIMES FUNCTIONAL UNIT RESPONSE TIME 1.

Manual Reactor Trip NOT APPLICABLE 2.

Power Range, Neutron Flux

<0.5secondsk) 3.

Power Range, Neutron Flux, High Positive Rate NOT APPLICABLE 4

Power Range, Neutron Flux, High Negative Rate 10.5secondsM 5.

Intermediate Range, Neutron Flux NOT APPLICABLE 6.

Source Range, Neutron Flux (Below P-10)

NOT APPLICA8LE 7.

Overtemperature AT 1 6.0 second M 8.

Overpower AT 16.0secondM) 9.

Pressurizer Pressure--Low

~ 2.0 seconds (Above P-7)

10. Pressurizer Pressure--High 1 2.0 seconds

11. Pressurizer Water Level--High (Above P-7)

NOT APPLICA8LE

12. Loss of Flow - Single Loop (Above P-8) i 1.0 seconds

13. Loss of Flow - Two Loop (Above P-7 and below P-8)

< 1.0 seconds

14. Steen Generator Water Level--Low-Low (Loop Stop Valves Open)

< 2.0 seconds

15. DELETED.

16. Undervoltage-Reactor Coolant Pumps (Ateve P-7) i 1.5 seconds

17. Underfrequency-Reactor Coolant Pimps (Above P-7) 1 0.9 seconds

[O Wout,ron detectors are exempt free response time testing.

Response

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

BEAVER VALLEY - UNIT 2 3/43-8 (delete this page) l

NPF-73 4

TABLE 3 3 0 (Continued)

REACTOR TRIP SYSTEM IN5'EUMENTATION RESPONSE TIMES FUNCTIONAL UNIT RESPONSE TIME

18. Turbine Trip (Above P 9)

A.

Emergency Trip Header Low Pressure NOT APPt.ICABLE B.

Turbine Stop Valve Closure NOT APPLICABLE

19. Safety Injection Input from ESF NOT APPLICABLE

20. Reactor Coolant Pump Breaker Position Trip NOT APPLICABLE (Above P-7)

21. Reactor Trip Breakers NOT_ APPLICABLE

22. Automatic Trip Logic NOT APPLICABLE

23. Reactor Trip System Interlocks NOT APPLICABLE BEAVER VALLEY - UNIT 2 3/4 3-9 (delete this page)

NPF-73 INSTRUMENTATf0N 3f4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2.1 The Engineered Safety Feature Actuation System (ESFAS) instrumentstion channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their Trip Setpoints set consistent with the values show Table 3.3-4.- -.....,,...........,. -..-.. n in the Trip Setpoint column of I

APPLICABILITY:

As shown in Table 3.3-3.

ACTION:

With an ESFAS Instrumentation or Interlock Trip Setpoint less conservative a.

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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that Equation 2.2-1 was satisfied for the affected channel, or 2.

Declare the channel inoperable and apply the applicable ACTION state-ment 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 + S < TA where:

Z

=

The value for Column Z of Table 3.3-4 for the affacted

channel, R

=

The "as measured" value (in percent span) of rack error for the affected chanrel, S

=

Either the "as measured" value (in percent span) of the sensor error, or the value from Column S (Sensor Orift) 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, With an ESFAS instrumentation channel or interlock inoperable, take the c.

ACTION shown in Table 3.3-3.

BEAVER VALLEY - UNIT 2 3/4 3-14 (Proposed Wording)

PF-73 INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REOUIREMENTS 4.3.2.1.1 Each engineered safety feature actuation system instrumentation channel and interlock and the automatic actuation logic with master and slave relays shall be demonstrated OPERA by the performance of the ESFAS Instru-mentation Surveillance Requirement during the MODES and at the frequencies I

shown in Table 4.3-2.

o 4.3.2.1.2 the at power CHANNEL FUNCTIONAL TEST of channels affected tion.

The total interlock function shall be demonstrated OPERABLE at least once per 18 months durin by interlock operation, g CHANNEL CALIBRATION testing of each channel affected gg 4,3.2.1.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESF funct shall be demonstrated to be within the limit at least once per 18 months **

Each I

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 per N times 18 months where N is the total number of redundant chann61s in a specific ESF function as shown in the " Total No. of Channels" Column of Table 3.3-3.

M For the automatic actuation logic, the surveillance requirements shall be the I

applicationofvarioussimulatedinputconditionsinconjunctionwitheach possible interlock logic state and verification of the required logic output including as a minimum, a continuity check of output devices.

For the actuation, relays, the surveillance requirements shall be the energization of each master and slave relay and verification of OPERABILITY of each relay.

The test of master relays shall include a continuity check of each associated slave relay. The test of slave relays (to be performed at least once per 92 days in lieu of at least once per 31 days) shall include, as a minimum, a tinuity check of associated actuation devican that are not testable.

- The spec 171ed 18-month surveillance interval during the~first fuel cycle may be extended to coincide with completion of the first refueling outage.

% elete.

d BEAVER VALLEY - UNIT 2 3/4 3 15 Amendment No.

(Proposed Wording)

5 IABLE 3.3-4 (Continued) en h

ENGINEERED SAFETY FEATURES ACTUATION SYSTEN INSTRUMENIAIION TRIP SEIPOINIS c

TOIAL SOtSOR 1 RIP N FUNCTIONAL UNIT ALLOWANCE (TA)

_Z DRIFT (5)

SEIPOINI Att0WABil VALUE

.s

8.

ENGINEERED SAFETY FEATURE INTERLOCKS E

1 Q

a.

Reactor Trip, P-4

,N.A.

N.A.

)

l b.

Pressurizer Pressure, i

P-11 N.A.

N.A.

< 2000 psig

$ 2010 psig Low-Low I,,, P-12 4.0 0.82 0.87

> 541*f

> 538.5*l g

c.

e 22 o o e, O

gp g

&;"~

N (J o.)

8:t 6

d'

NPF-73 TABLE 3.3-5 ENGINEERED SAFETY FEATURES REw0NSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECON05 1.

Manual a.

Safety Injecti:,n (ECCS)

Not Applicable Feedwater Isolation Not Applicable Reactor Trip (SI)

Not Applicable Containment Isolation-Phase "A" Not Applicable Containment Vent and Purge Isolation Not Applicable Auxiliary Feedwater Pumps Not Applicable Service Water System Not Applicable b.

Containment Quench Spray Pumps Not Applicable Containment Quench Spray Valves Not Applicable Containment Isolation-Phase "B" Not Applicable c.

Containment Isolation-Phase "A" Not Applicable d.

Control Roon Ventilation Isolation Not Applicable 2.

Containment Pressure-High a.

Safety Injection (ECCS) 1 27.

b.

Reactor Trip (from SI) 1 2.0 c.

Feedwater Isolation 6

1 7.0(t) 9

/f d.

Containment Isolation-Phase "A" 1 61.5(+)/115.5(+)

e.

Auxiliary Feedwater Pumps 1 60.0 7

2 f.

Service Water System 1 72.5(t)/181.5(4) 3.

Pressurizer Pressure-Low (Z)

[W a.

Safety Injection (ECCS) 1 27.0*/27.0#-

b.

Reactor Trip (from SI) 1 2.0 c.

Feedwater Isolation 17.0(h 1 61.0

/115.0($)

d.

Coritainment Isolation-Phase "A" BEAVER VALLEY - UNIT 2 3/4 3-29 (delete this page)

N,PF-73 TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AN9 FUNCTION RESPONSE TIME IN SECONOS 3.

Pressurizer Pressure-Low (Continued) e.

Auxiliary Feedwater Pumps 1 60.0 7

Y f.

Service Water Systee 1 72.0(t)/181.0(+)

4.

Steam Line Pressure-Low (D

(4) a.

Safety Injection (ECCS) 1 37.0 M/27.0F b.

Reactor Trip (from SI) 12.0 c.

Feedwater Isolation 1 7.0(+)

d.

Containment Isolation-Phase "A"

$ 61.0( /115.0h) e.

Auxiliary Feedwater Pumps 1 60.0 f.

Service Water System 172.0d)/181.0($)

g.

Steam Line Isolation 1 7.0 5.

Containment Pressure--Hich-High

/C a.

Containment Quatth Spray 1 85.5(+)

b.

Containment Isolation-Phase "B" Not Applicable 9

to c.

Control Room Ventilation Isolation 122.0R)/77.0(fr) 3 6.

Steam Generator Water Level--High-High a.

Turbine Trip 1 2.5 b.

Feedwter Isolation 17.0(h 7.

Contairumnt Pressure--Intermediate High-High a.

Stees Line Isolation 1 7.0 8.

Steamline Pressure Rate--High Negative a.

Staanline Isolation 1 7.0 BEAVER VALLEY - UNIT 2 3/4 3-30 (delete this page)

NPF-73 TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTIOk

. RESPONSE TIME IN SECONOS 9.

LossjfPower a.

4.16kv Emergency Bus Unt'ervoltage 1 1 C.1 sec.

(Loss of Voltage) (Trip Feeder) b.

4.16kv and 480v Emergency Bus Under-90 t 5 sec.

voltage (Degraded Voltage)

-10. 'Intenti:.:!!y bl ek)

/4F[. Steam Generator Water Level-Low-low Motor-drivon Auxi.d11ary

< 60.0 a.

Faedwater Pump W b.

Turbine-drivenAuxiliary

< 60.0 Feedwater Pump W a>

~

//g. Undervoltage RCP a.

Turbine-driven Auxiliary 1 60.0 Feedwater Pump

/'.i)3'. Trip of Main Feedwater Pumps a.

' Motor-driven Auxiliary

< (0.0 '

Feedwater Pumps

~

/tJ4' Cor. trol Room High Radiation ll a.

Control Room Ventilation Isolation i 180(+)

b M n 2/3 in 2/3 Steam Generators

@) ***on 2/3 atiy Steam Generator BEAVER VALLEY - UNIT 2 3/4 3-31 (delete this page)

Tintr 3.3-5 (continuedt TABLE NOTATION (3)-+-

Diesel generator starting and sequence loading delays included.

Response time limit includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging pumps and Iow Head Safety Injection pumps.

Sequential transfer of charging pump suction from the volume control tank (VCT) to the refueling water storage tank (RWST) (RWST valves open, then VCT valves close) is not included.

h-#- Diesel generator starting and sequence loading delays nas included.

offsite power available.

Response time limit includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging pumps. Sequential transfer of charging pump suction from the volume control tank (VCT) to the refueling water storage tank (RWST) (RWST valves open, then VCT valves close) is included.

[Ei# Diesel generator starting and sequence loading delays included.

3 Response time limit includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal 9

charging pumps.

Sequential transfer of charging pump suction from the volume control tank (VCT) to the refueling water storage tank (RNST) (RNST valves open, then VCT valves close) is included.

6

(+) Feedwater system overall response time shall include verification of valve stroke times applicable to the feedwater containment isolation valves for Train A and the main feedwater regulating valves and bypass valves for Train B.

(h) Diesel generator starting and sequence loading delays included.

Response time. limit includes attainment of discharge pressure for service water pumps.

T(-S-) Diesel generator starting and sequence loading delays n3 included.

Response time limit only includes opening of valves to establish the flowpath to the diesel coolers, h Diesel generator starting and sequence loading delays nat included.

Offsite power available.

Response time limit includes operation of' valves / dampers.

/4

(+) Diesel generator starting and sequence loading delays included.

Response time limit incl es operation of valves / dampers.

II(4) Diesel generator starting and sequence loading delays nat included.

Response time limit includes operation of dampsrs.

BEAVER VALLEY - UNIT 2 3/4 3-32 Amendment No.4 dele h dd tye)

9/4.3 INS?RUMEN m !ON hASES x

3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued)

~

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 uncert.finty magnitudes.

i Rack drift in excess of *he Allowable Va:ue exhibits the behavior that the rack has not cet its allowance.

Being that there is a small statistical chance that this till 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 surveillance requirements for the Manual Trip Function, Reactor Trip Breakers, and Reactor Trip Bypass Breakers are provided to reduce the possi-bility of an Anticipated Transient Without Scram (ATWS) event by ensuring OPERABILITY of the diverse trip features (

Reference:

Generic Letter 85-09).

The measurew nt of response time at the specified frequencies provides assurance that the protective and ESF action ft. tion associated with each channel is coseleted within the time limit ast i in the accident analyses.

No credit was taken in the analy g(altN &cM J 6 @

U n ta$ pants times r

L.

indicated not applicable.

4Fm rh.a/

onse times ;;;;i"!:.Ococ en9 Ach~.t nuts in ESF

- T:t -

operation of the RWST and VCT valves 7 b : n wnicn incluae secuential e-4 #0) are based on values the non LOCA safety analysesI These analyses take credit for assumed injection of borated water from the RWST.

Injection of borated water is assumed not to occur until the VCT charging pump suction valves are closed following opening of the RWST charging pump suction valves.

Whun sequential operation of the RWST and VCT valves is not included in the response times (Met: *), the values specified are based on the LOCA analyses.

The LOCA analyses take credit fcr injection flow regardless of the source.

Verification of the response timss ;;;;if kt. k Teth 3.3 twill assure that the assumptions used for the LOCA and Non-LOCA analyses with respect to operation of the VCT ano RWST valva are valid.

The maximum response time for cnntrol room isolation on hign radiation is based on ensuring that the control room remains habitable following a small line break outside the containment.

From a control room habitability aspect, the worst case accident that does not initiate a Containment Isolation - Phase B si;nal is the small line break outside the containment.

This response time includes radiation monitor processing delays associated with the monitor aver :ing techniques.

Diesel Generator starting and sequence loading delays are t included since these delays occur prior to the control room environ-men

txceeding the high radiation setpoint.

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 times.

BEAVER VALLEY - UNIT 2 8 3/4 3-2 (tmpsed Wmlj)

t s

ATTACHMENT B Beaver Valley Power Station, Unit Hos. 1 and 2 Proposed Technical Specification Change No. 217 and 84 REMOVAL OF RESPONSE TIME TABLES A.

DESCRIPTION OF AMENDMENT REQUEST n

The proposed amendment is a line-item technical specification (TS) improvement described in Generic Letter 93-08 which would remove response time Table 3.3-2 and 3.3-5 from the TSs and relocate a

footnote from Table 3.3-2 into the curveillance requirements.

Several editorial changes are also described in the submittal.

B.

BACKGROUND The Nuclear Regulatory Commission (NRC) issued the generic letter as guidance in preparing a TS change to rolocate the tables of response time limits for the reactor trip system (RTS) and the engineered safety features actuation system (ESFAS) instruments to the Updated Final Safety Analysis Report (UFSAR).

The NRC has already implemented this line-item TS improvement in the Improved Standard Technical Specifications (ISTS).

C.

JUSTIFICATION The proposed amendment incorporates the NRC guidance provided in Generic Letter 93-08 for relocating the subject response time tables from the TS to the UFSAR.

The Licensing Requirements Manual (LRM) was developed and issued as a vehicle to control and maintain those items removea from the TS.

The core Operating Limits Report (COLR) and containment penetrations table (containment isolatior. valves) are controlled and maintained in the LRM.

Like the UFSAR, changes to the LRM are controlled in accordance with the requirements of 10 CFR 50.59; therefore, the iRfS and ESFAS response times will be maintained in the LRM.

This is consistent with the intent of the generic letter where the response time requirements will be controlled and maintained to satisfy operability and accident analysis assumptions.

The surveillance reonirements will not be reduced since they will continue to require verification that the response time of each function is within the required limit.

Future changes to the response times will be in accordance with 10 CFR 50.59 and the tables will be controlled by the same controls applied when changing the UFSAR.

D.

SAFETY ANALYSIS The limiting conditions for operation (LCOs) for the RTS and ESFAS instruments require that these systems be operable with response times as specified in the TS tables for these systems.

The surveillance requirements specify that these systems be tested to verify that the response time of each function is within its limits.

Relocating the tables of the RTS and ESFAS

ATTACHMENT B, c ntinutd Prep::cd Tcchnic31 Sp:cificStien Chongo Nco. 217 and 84 Page 2 instrument response time limits from the TS to the LRM will not alter these surveillance requirements.

The response time limits for the RTS and ESFAS instruments will be addressed in the LRM, including those channels for which the response time limit is indicated as "NA",

that is, a

response time limit is not applicable.

The clarification provided in the applicable TS footnote will also be relocated to the LRM with the response times to describe how the response time limits are to be applied.

This change will allow the plant to administrative 1y control changes to the response time limits for RTS and ESPAS instruments in accordance with 10 CFR 50.59 without the need to process a technical specification change.

The LCos for the RTS and ESPAS specify that the associated instruments "...shall be OPERABLE with RESPONSE TIMES as shown in Table 3.3-2" for the RTS or "... Table 3.3-5" for the ESFAS.

The generic letter grovides an ecceptable alternative by removing reference to response times and simply stating that this instrumentation "...shall be OPERABLE."

This is compatible with relocating the response time tables.

An unrelated change is also addressed in the Unit 2 RTS and ESFAS specifications where the

note has been changed to (1) and the ** note has been deleted.

Changing the

note to (1) is an editorial change and is not a safety concern.

The ** note only applied during the first fuel cycle and has since expired.

Also for both units, the page preceding the response time tables has been revised to include wording so that the reader is aware that pages have been deleted, i.e.,

next page is The surveillance requirements specify that the response time for each trip function is to be demonstrated to be-within its limit at the specified frequency and do not reference the tables of response time limits.

Therefore, the surveillance requirements need not be modified to implement this change.

However, a

footnote in the table of response time limits for the RTS states that neutron detectors are exempt from response time testing.

To retain this exception, which is stated in the table being removed from the TS, surveillance requirement 4.3.1.1.3 has been modified by adding

" Neutron detectors are exempt from response time testing."

In addition, the Bases have been modified by deleting reference to the response time tables and the applicable notes for compatibility with relocating the tables.

Attachments C-1 and C-2 provide suggested LRM changes to incorporate the response time tables into Unit 1 and Unit 2 LRMs, respectively.

These changes will be included in the LRM following approval of this amendment.

The proposed amendment incorporates a technical specification line-item improvement by relocating the RTS and ESFAS response time tables to the LRM.

This change is in agreement with Generic Letter 93-08 and does not reduca any response time testing requirements.

Plant response time procedures will be modified to B-2 1

I

ATTACHMlWiT D, c ntinund Proposed Technical Specification Change No1. 217 and 84 Page 3 i

reference the LRM in lieu of the technical specifications.

Future changes to response times will be in accordance with 10 CFR 50.59.

This does not change any of the response time testing requirements or affect the UFSAR accident analyses; therefore, this change has been determined to be safe and will not reduce the safety of the plant..

E.

NO SIGNIFICANT HAZARDS EVALUATION The no significant hazard considerations involved with the proposed amendment have been evaluated, focusing on the three standards set forth in 10 CFR 50.92/c) as quoted belov The commission may make a final determination, pursuant to the procedures in paragraph 50.91, that a proposed amendment to an operating license for a

facility licensed under paragraph 50.21(b) or paragraph 50.22 or for a

testing facility involves no significant hazards consideration, if operation of the facility in accordance with the proposed amendment would nott (1)

Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2)

Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3)

Involve a significant reduction in a margin of safety.

The following evaluation is provided for the no significant hazards considsration standards.

1.

Does the change involve a

significant increase in the probability or consequences of an accident previously evaluated?

The proposed amendment relocates the instrument response time limits for the reactor trip system (RTS) and engineered safety feature actuation system (ESFAS) from the technical specifications to the Licensing Requirements Manual (LRM).

The Core Operating Limits Report (COLR) and containment penetrations table (containment isolation valves) are controlled and maintained in the LRM.

The LRM was developed to control and maintain those items removed from the technical specifications.

The proposed amendment conforms to the guidance given in Enclosures 1 and 2 of Generic Letter 93-08.

Neither the response time limits nor the surveillance requirements for performing response time testing will be altered by this sabmittal.

The overall RTS and ESFAS functional capabilities will not be changed and assurance that action requirements of the protective and engineered safety features systems are completed ~within the time limits assumed in the accident analyses is unaffected by the B-3 m

ATTACHMENT B, ctntinutd Proposed Technical Specification Change Nos. 217 and 84 Page 4 proposed amendment.

Therefore, operation of the facility in accordance with the proposed amendment will not involve a significant increase in the probability or consequences of an accident previously evaluated.

2.

Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?-

The proposed amendment will not change the physical plant or the modes of plant operation defined in the operating license.

The change does not involve the addition or modification of equipment nor does it alter the design or operation of plant systems.

Therefore, operation of the f acility in accordance with the proposed amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated.

3.

Does the change involve a significant reduction in a margin of safety?

The measurement of instrumentation response times at the frequencies specified in the technical specification provides assurance that actions associated with the protective and engineered safety features systems are accomplished within the time limits assumed in the accident analyses.

The response time limits, and the measurement frequencies remain unchanged by the proposed amendment.

The proposed changes do not alter the basis for any other technical specification that is related to the establishment of or maintenance of a nuclear safety margin.

Therefore, operation of the facility in accordance with the proposed amendment will not involve a significant reduction in the margin of safety.

F.

NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION Based on the considerations expressed above, it is concluded that the activities associated with this license amendment request satisfy the no significant hazards consideration standards of 10 CFR 50.92(c)

and, accordingly, a

no significant hazards consideration finding is justified.

G.

ENVIRONMENTAL CONSIDERATION The proposed amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20.

It has been determined that the proposed amendment involves no significant increase in the amounts, and no significant change in the types of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR

51. 22 (c) (9).

Pursuant to 10 CFR 51.22(b) no environmental impact B-4 I

i l

AT'fACHMENT B, centinu:d l

Proposed Technical Specification Change Nos. 217 and 84 Page 5 statement or environmental assessment need be prepared in connection with the issuance of this proposed amendment.

H.

UFSAR CHANGES No UFSAR changes are required.

B-5

=

ATTACHMENT C-1 Beaver Valley Power Station, Unit No. 1 Proposed Technical Specification Change No. 21.7 LICENSING REQUIREMENTS MANUAL CHANGES Applicable Licensing Requirements Manual Changes

BVPS-1 LICENSING REQUIREMENTS MANUAL 3.0 INSTRUMENTATION 3.0-1 1

BVPS-1 LICENSING REQUIREMENTS MANUAL 3.1 Reactor Trin System Instrumentation Resnonse Times Each reactor trip system instrumentation response time listed in Table 3.1-1 shall be maintained in the manner specified in Technical Specification 3/4.3.1.

APPLICABILITY:

As specified in TS 3/4.3.1 3.1-1

BVPS-1 LICENSING REQUIREMENTS MANUAL TABLE 3.1-1 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES FUNCTIONAL UNIT RESPONSE TIME 1.

Manual Reactor Trip NOT APPLICABLE 2.

Power Range, Neutron Flux 5 0.5 seconds (II 3.

Power Range, Neutron Flux, High Positive Rate NOT APPLICABLE 4.

Power Range, Neutron Flux, III High Negative Rate 1 0.5 seconds S.

Intermediate Range, Neutron Flux NOT APPLICABLE 6.

Source Range, Neutron Flux NOT APPLICABLE 7.

Overtemperature AT s 6.0 seconds (l)

III 8.

Overpower AT

$ 6.0 seconds 9.

Pressurizer Pressure -- Low 5 2.0 seconds 10.

Pressurizer Pressure -- High 1 2.0 seconds 11.

Pressurizur Water Level -- High NOT APPLICABLE 12.

Loss of Flow - Single Loop (Above P-8)

$ 1.0 seconds 13.

Loss of Flow - Two Loops (Above P-7 and below P-8) 5 1.0 seconds 14.

Steam Generator Water Level -- Low-Low

$ 2.0 seconds 15.

Steam /Feedwater Flow Mismatch and Low Steam Generator Water Level NOT APPLICABLE 16.

Undervoltage-Rear. tor Coolant Pumps 5 1.2 seconds 17.

Underfrequency-Reactor Coolant Pumps s 0.6 seconds 18.

Turbine Trip A.

Auto Stop Oil Pressure NOT APPLICABLE B.

Turbine Stop Valve NOT APPLICABLE 19.

Safety Injection Input from ESF NOT APPLICABLE 20.

Reactor Coolant Pump Breaker Position Trip NOT APPLICABLE TABLE NOTATION (1)

Neutron detectors are exempt from response time testing.

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

3.1-2 e

BVP8-1 LICENSING REQUIREMENTS MANUAL I

3.2 Enaineered Safety _ Feature Resnonse Tlaas Each engineered safety feature response time listed in Table 3.2-1 shall be maintained in the manner specified in Technical Specification 3/4.3.2.

APPLICABILITY:

As specified in TS 3/4.3.2 3.2-1

(

BVPS-1 LICENSING REQUIREMENTS MANUAL TABLE 3.2-1 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 1.

Manual a.

Safety Injection (ECCS)

Not Applicable Feedwater Isolation Not Applicable Reactor Trip (SI)

Not Applicable Containment Isolation-Phase "A" Not Applicable Containment Vent and Purge Isolation Not Applicable Auxiliary Feedwater Pumps Not Applicable Rx Plant River Water System Not Applicable b.

Containment Quench Spray Pumps Not Applicable Containment Quench Spray Valves Not Applicable Containment Isolation-Phase "B"

Not Applicable c.

Containment Isolation-Phase "A" Not Applicable d.

Control Room Ventilation Isolation Not Applicable 2.

Containment Pressure-Hiah a.

Safety Injection (ECCS) 5 27.0(3) b.

Reactor Trip (from SI)

$ 3.0 c.

Feedwater Isolation 1)

Feedwater Regulating Valves 5 10.0(6) 2)

Feedwater Bypass Valves 5 30.0(6) d.

Containment Isolation-Phase "A"

s 22.0(8)/33.0(7) e.

Auxiliary Feedwater Pumps Not Applicable f.

Rx Plant River Water System

$ 77.0(8I /110.0(7) 3.2-2

BVPS-1 LICENSING REQUIREMENTS MANUAL

+

TABLE 3.2-1 (Continued)

ENGINEERED SAFETY FEATUREE RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 3.

Pressurizar Pressure-Low a.

Safety Injection (ECCS) 5 27.0(3I /27.0(4I b.

Reactor Trip (from SI) s 3.0 c.

Feedwater Isolation 1)

Feedwater Regulating Valves 6 10.0('I 2)

Feedwater Bypass Valves s 30.0('I d.

Containment Isolation-Phase "A"

$ 22.0(s) e.

Auxiliary Feedwater Pumps Not Applicable f.

Rx Plant River Water System 5 77.0(8I /110.0(7) 4.

Steam Line Pressure-Low a.

Safety Injection (ECCS) 5 27.0(4I /37.0(5) b.

Reactor Trip (from SI)

$ 3.0 c.

Feedwater Isolation 1)

Feedwater Regulating Valves 5 10.0(6) 2)

Feedwater Bypass Valves 5 30.0(6) d.

Containment Isolation-Phase "A" 5 22.0(s)/33.0(7I a.

Auxiliary Feedwater Pumps Not Applicable f.

Rx Plant River Water System

$ 77.0(8)/110.0(7I g.

Steam Line Isolation 5 0.0 5.

Containment Pressure--Hiah-Hiah a.

Containment Quench Spray 5 85.0(7I b.

Containment Isolation-Phase "B" Not Applicable c.

Control Room Ventilation Isolation S 22.0(8)/77.0(7) 3.2-3

BVPS-1 LICENSING REQUIREMENTS MANUAL

+

TABLE 3.2-1 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 6.

Steam Generator Water Level--Hiah-Hiah a.

Turbine Trip-Reactor Trip 5 2.5 (Above P-9) b.

Feedwater Isolation 1)

Feedwater Regulating Valves 5 10.0 'I I

2)

Feedwater Bypass Valves s 30.0(6) 7.

Containment Pressure--Intermediate Hiah-Hiah a.

Steam Line Isolation s 8.0 8.

Steamline Pressure Rate--Hiah Neantive a.

Steamline Isolation s 8.0 9.

Loss of Power a.

4.16kV Emergency Bus Undervoltage s 1.3 (Loss of Voltage) b.

4.16kV and 480v Emergency Bus s 95 Undervoltage (Degraded voltage) 10.

Steam Generator Water Level-Low-Low a.

Motor-driven Auxiliary 60.0 Feedwater Pumps (2)

Turbine-drivenp)uxiliary b.

60.0 Feedwater Pump (

11.

Undervoltaae RCP a.

Turbine-driven Auxiliary 60.0 Feedwater Pump 12.

Emercency Bus Undervoltace a.

Motor-driven Auxiliary 60.0 Feedwater Pumps 3.2-4

BVPS-1 LICENSING REQUIREMENTS MANUAL TARLE 3.2-1 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 13.

Trin of Main Feedwater P,qmpa a.

Motor-driven Auxiliary 60.0 Feedwater Pumps NOTE:

Response time for Motor-driven Auxiliary 60.0 Feedwater Pumps on all S.I.

signal starts TABLE NOTATION (1) on 2/3 any Steam Generator (2) on 2/3 in 2/3 Steam Generators (3)

Diesel generator starting and sequence loading delays included.

Response time limit includes opening of valves to establish SI path and attainment of discharge

pressure for centrifugal charging pumps and Low Head Safety Ingection pumps.

Sequential transfer of charging pump suction from the volume control tank (VCT) to the refueling water storage tank (RWST)

(RWST valves open, then VCT valves close) is not included.

(4)

Diesel generator starting and sequence loading delays Dnt included.

Offsite power available.

Response

time limit includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging pumps.

Sequential transfer of charging pump suction from the volume control tank (VCT) to the refueltng water storage tank (RWST)

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

(5)

Diesel generator starting and sequence loading delays included.

Response time limit includes opening of valves to establish SI psLn and attainment of discharge pressure for centrifugal charging pumps.

Sequential transfer of charging pump suction from the volume control tank (VCT) to the refueling water storage tank (RWST)

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

(6)

Feedwater isolation includes signal response and valve closure time.

(7)

Diesel generator starting and sequence loading delays included.

(8)

Diesel generator starting and sequence loading delays nnt included.

3.2-5

]

ATTACHMENT C-3

]

Beaver Valley Power Station, Unit No. 2 Proposed Technical Specification Change No. 84 LICENSING REQUIREMENTS MANUAL CHANGES Applicable Licensing Requirements Manual Changes l

.o BVPS-2 LICENSING REQUIREMENTS MANUAL 3.0 INSTRUMENTATION 3.0-1

.. o BVPS-2 LICENSING REQUIRENENTS MANUAL 3.1 Reactor Trio system Instrumentation Resnonse Times Each reactor trip system instrumentation response time listed in Table 3.1-1 shall be maintained in the manner specified in Technical specification 3/4.3.1.

APPLICABILITY:

As specified in TS 3/4.3.1 3.1-1

BVPS-2 LICENSING REQUIREMENTS MANUAL TABLE 3.1-1 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES FUNCTIONAL UNIT RESPONSE TIME 1.

Manual Reactor Trip NOT APPLICABLE 2.

Power Range, Neutron Flux s 0.5 seconds (l) 3.

Power Range, Neutron Flux, High Positive Rate NOT APPLICABLE 4.

Power Range, Neutron Flux, High Negative Rate 5 0.5 secondsIII 5.

Intermediate Range, Neutron Flux NOT APPLICABLE 6.

Source Range, Neutron Flux (Below P-10)

NOT APPLICABLE 7.

Overtemperature AT s 6.0 seconds (l) 8.

Overpower AT S 6.0 seconds (II 9.

Pressurizer Pressure--Low (Above P-7)

$ 2.0 seconds 10.

Pressurizer Pressure--High s 2.0 seconds 11.

Pressurizer Water Level--High NOT APPLICABLE (Above P-7) 12.

Loss of Flow cingle Loop (Above P-8) s 1.0 seconds 13.

Loss of Flow - Two Loop (Above P-7 and below P-8)

$ 1.0 seconds 14.

Steam Generator Water Level--Low-Low (Loop Stop Valves Open) 5 2.0 seconds 15.

DELETED 16.

Undervoltage-Reactor Coolant Pumps (Above P-7)

$ 1.5 seconds 17.

Underirequency-Reactor Coolant Pumps (Above P-7) 5 0.9 seconds 3.1-1

BVPS-2 LICENSING REQUIREMENTS MANUAL TARLE 3.1-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES EUECTIONAL UNIT RESPONSE TIME 18.

Turbine Trip (Above P-9)

A.

Emergency Trip Header NOT APPLICABLE Low Pressure B.

Turbine Stop Valve Closure NOT APPLICABLE 19.

Safety Injection Input from ESF NOT APPLICABLE 20.

Reactor Coolant Pump Breaker Position Trip NOT APPLICABLE (Abovo P-7) 21.

Reactor Trip Breakers NOT APPLICABLE 22.

Automatic Trip Logic NOT APPLICABLE-23.

Reactor Trip System Interlocks NOT APPLICABLE TABLE NOTATION x

(1)

Neutron detectors are exempt from response time testing.

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

3.1-2

o BVPS-2 LICENSINS REQUIREMENTS MANUAL 3.2 Enaineered safety Feature Reanonsa Times Each engineered safety feature response time listed in Table 3.2-1 shall be maintained in the manner specified in Technical specification 3/4.3.2.

APPLICABILITY:

As specified in TS 3/4.3.2 3.2-1

o BVPS-2 LICENSING REQUIREMENTS MANUAL TABLE 3.2-1 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONQS 1.

Manual a.

Safety Injection (ECCS)

Not Applicable Feedwater Isolation Not Applicable Reactor Trip (SI)

Not Applicable containment Isolation-Phase "A" Not Applicable Containment Vent and Purge Isolation Not Applicable Auxiliary Feedwater Pumps Not Applicable Service Water System Not Applicable b.

Containment Quench Spray Pumps Not Applicable containment Quench Spray Valves Not Applicable Containment Isolation-Phase "B"

Not Applicable c.

Containment Isolation-Phasz "A"

Not Applicable d.

Control Room Ventilation Isolation Not Applicable 2.

Containment Pressure-Hiah 0I a.

Safety Injection (ECCS)

$ 27.0 b.

Reactor Trip (from SI) 5 2.0 c.

Feedwater Isolation s 7.0(6) 03 d.

Containment Isolation-Phase "A" s 61.5

/115.500) e.

Auxiliary Feedwater Pumps 5 60.0 f.

Service Water System 5 72.5(7)/181.5(8) 3.2-2

BVPS-2 LICENSING REQUIREMENTS MANUAL TABLE 3.2-1 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 3.

Pressurizer Pressure-Low a.

Safety Injection (ECCS) 5 27. 0d 3I /27.0(4I b.

Reactor Trip (from SI) s 2.0 c.

Feedwater Isolation s 7.0(6) d.

Containment Isolation-Phase "A" s 61.0('I /115.0(10I e.

Auxiliary Feedwater Pumps s 60.0 f.

Service Water System 5 72.0(7I /181.0(8) 4.

Steam Line Pressure-Low a.

Safety Injection (ECCS)

$ 37.0J5)/27.0(4I b.

Reactor Trip (from SI)

$ 2.0 c.

Feedwater Isolation s 7.0(6) d.

Containment Isolation-Phase "A" s 61. ad'I /115.0(10) e.

Auxiliary Feedwater Pumps 5 60.0 f.

Service Water System 5 72.0d7I /181.0(8) g.

Steam Line Isolation s 7.0 5.

Containment Pressure--Hiah-Hiah a.

Containment Quench Spray 5 85.5(10) b.

Containment Isolation-Phase "B"

Not Applicable c.

Control Room Ventilation Isolation s 22.0('I /77. 0(10) 6.

Steam Generator Water Level--Hiah-Hioh a.

Turbine Trip 5 2.5 b.

Feedwater Isolation s 7.0(6) 3.2-3

__ o

BVPS-2 LICENSING REQUIREMENTS MANUAL TABLE 3.2-1 (Continued)

EHQINEERED SAFETY FEATURES RESPONSE TIMER INITIATING SIGNAL AND FUNCTIDH RESPONSE TIME IN SECONDS 7.

Containment Pressure--Intermediate Hiah-High a.

Steam Line Isolation s 7.0 8.

Steamline Pressure Rate--Hlah Neaatiya a.

Steamline Isolation J 7.0 9.

Lggs of Power a.

4.16kV Emergency Bus Undervoltage 1 i 0.1 sec.

(Loss of Voltage) (Trip-Feeder) b.

4.16kV and 480v Emergency Bus 90 i 5 sec.

Undervoltage-(Degraded Voltage) 10.

Steam Generator Water Level-Low-Low Motor-drivenAuxp)liary

$ 60.0 a.

Feedwater Pumps (

Turbine-drivenp)uxiliary 1 60.0 b.

Feedwater Pump (

11.

Undervoltaae RCP a.

Turbine-driven Auxiliary 5 60.0 Feedwater Pump 12.

Trio of Main Feedwater_Pumos a.

Motor-driven Auxiliary 1 60.0 Feedwater Pumps 13.

Control Room Hiah Radiation a.

Control Room Ventilation Isolation s 180(11) 3.2-4

,o g

DVPS-2 o

LICENSING REQUIREMENTS MANUAL TABLE 3.2-1 (Continued)

ENGINEERED SAFETY PEhTURES RESPONSE TIMES TABLE NOTATION (1) on 2/3 in 2/3 Steam Conorators (2) on 2/3 any Stoam Generator (3)

Diosol generator starting and sequence loading delays included.

Rosponse tino limit includes opening of valvos to establish SI path and attainment of dischargo prosauro for contrifugal charging pumps and Low Ilead Safety In:,0ction pumps.

Sequential transfor of charging pump suction from the volume control tank (VCT) to the refueling water storage tank (RWST)

(RWST valvos open, then VCT valvos closo) is not included.

(4)

Diosol generator starting and sequenco loading dolays D21 includod.

Offuito power available.

Response

timo limit includos opening of valvos to establish SI path and attainment of dischargo prosauro for contrifugal charging pumps.

Soquential transfor of charging pump suction from the volume control tank (VCT) to the refueling water storage tank (RWST)

(RWST valvos opon, than VCT valvos closo) is included.

(5)

Diosol gonorator starting and sequence loading delays included.

Rosponse timo limit includos opening of valvos to establish SI path and attainmont of dischargo pressure for contrifugal charging pumps.

Soquential transfer of charging pump suction from tho volume control tank (VCT) to the refueling water storage tank (RWST) (RWST valvos open, then VCT valvos closo) is included.

(6)

Foodwater system ovorall response timo shall includo verification of valvo stroke times applicable to the foodwater containment isolation valves for Train A and the main foodwater regulating valvos and bypass valves for Train B.

(7)

Diesel generator starting and sequence loading delays included.

Response timo limit includes attainment of dischargo pressure for service water pumps.

(8)

Diosol generator starting and sequence loading delays DQt included.

Responso timo limit only includes opening of valvos to establish the flowpath to the diosol coolors.

(9)

Diosol generator starting and sequence loading delays D21 included.

Offsito power available.

Responso tino limit includes operation of valves / dampers.

(10) Diosol generator starting and sequenco loading delays included.

Rosponso timo limit includes cporation of valves /dampora.

(11) Diosol generator starting and sequence loading delays D2t included.

Response timo limit includos operation of dampors.

3.2-5 l

k

_ - _ _ _ _ _

ML20217A980

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