Text

Proposed Tech Specs,Allowing Limited Inoperability of Bast Level Channels & Transfer Logic Channels to Provide for Required Testing & Maint of Associated Components
	ML20196B972
Person / Time
Site:	Prairie Island
Issue date:	11/25/1998
From:	; NORTHERN STATES POWER CO.
To:	;
Shared Package
ML20196B954	List: ML20196B949; ML20196B957; ML20196B972;
References
	NUDOCS 9812010217
	Download: ML20196B972 (21)
	v • d • e

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l EXHlBIT B PRAIRIE ISLAND NUCLEAR GENERATING STATION l 1

License Amendment Request dated November 25,1998 l l l

l 1

Appendix A, Technical Specification Pages l Marked Up Pages (shaded material to be added, strikethrough material to be removed)

TS.3.2-1 !

Table TS.3.5-2B (Page 6 of 9) l Table TS.3.5-2B (Page 9 of 9)

B.3.5-1 B.3.5-4

! B.3.5-5 i

B.3.5-6 l

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l 9812010217 981125 l- PDR P pon ADDCK 05000282{1

s TS.3.2-1 REV 111 10/27/80 i 3.2 CHEMICAL AND VOLUME CONTROL SYSTEM Applicability

. Applies to the operational status of the chemical and volume control system. ;

Qbjective To define those conditions of the chemical and volume control system i necessary to assure safe reactor operation and safe COLD SHUTDOWN. '

Snecification A. When fuel is in a reactor and reactor coolant system average temperature is at or below 200*F there shall be at least one flow path to the core for boric acid injection. If no OPERABLE flow path exists. suspend all operations involving CORE ALTERATIONS or 7

positive reactivity changes.

B. - A ' reactor shall not be made or maintained critical nor shall the reactor coolant system average temperature exceed 200'F unless the following conditions are satisfied (except as specified in 3.2.C or ;

3.2.D belo ): '

1. Two of the three charging pumps shall-be OPERABLE.

2. At least one boric acid tank shall be aligned to the unit and l shall contain a minimum of 2000 gallons of 11.5% to 13% by weight boric acid solution at a temperature of at least 145'F.

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3. System piping, valves and pumps shall be OPERABLE to the extent l of establishing two independent flow paths for boric acid injection -- one flow path from the boric acid tanks to the core and one flow path from the refueling water storage tank to the core.

4. Two channels of heat tracing shall be OPERABLE for the flow paths from the boric acid tanks required to meet the requirements of

-Specification-3.2.B.3. i

5. Automatic valves, piping, and interlocks associated with the above components which are required to operate for the steam line break

accident are OPERABLE.

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E TABLE TS.3.5-2B (Page 6 of 9)

ENGINEERED SALETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINlMUM

. TOTAL NO. CHANNELS CHANNELS A_PPLICABLE EMNCTIONAL UNIT OF CHANNELS TO TRI_E OPERABLE MODES ACTION

8. LOSS OF POWER

a. Degraded Voltage 4/ Bus 2/ Bus 3/ Bus 1, 2, 3, 4 31, 32, 33 4kV Safeguards Bus (2/ phase on (1/ phase 2 phases) on 2 phases)

b. Undervoltage 4/ Bus 2/ Bus 3/ Bus 1. 2, 3, 4 31. 32, 33 4kV Safeguards Bus (2/ phase on (1/ phase 2 phases) on 2 phases)

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_ . _ _ _ _ . _ _ _ . _ _ . _ _ _ _ ______________.m_________..______% __ . _ _ _ _ _ _ _ _ _ _ ______m_ -

_ . _ _ _ _ _ u..s__ _ .-we- e =.______-m__._.u. _ _ . _ < . .-_-.-__._____m_._-.u-

-_______.________e--

TABLE 3.5-2B (Page 9 of 9)

Action Statements ACTION 30: With the number of OPERABLE channels ACTION 33: If the requirements of ACTIONS 30 or 31 one less than the Total Number of cannot be met within the time Channels, declare the associated specified, or with the number of auxiliary feedwater pump inoperable and OPERABLE channels three less than the take the action required by Total Number of Channels, declare the Specification 3.4.2. However, one associated diesel generator (s) .

channel may be bypassed for up to 8 inoperable and take the ACTION required hours for surveillance testing per by Specification 3.7.B.

Specification 4.1 provided the other channel is OPERABLE. &#M$$18'EW ACTION 31: With the number of OPERABLE channels ' tigiberf .

N F

MC_ _ fW one less than the Total Number of Ll l T d M _ > Mitt %1ii!

Channels. operation in the applicable HODE may proceed provided the i

hhmaneE W itt $ 1e L::onetitients93591(gli :bsMWebaMtled

"_lf$neh M inoperable channel is placed in the Minianna C " ' '

bypassed condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Wegg . LFM . W ' ' ' 'Ttbl idopdrab. "i .

ACTION 32: With the number of OPERABLE channels two less than the Total Number of stutuksJ' Aasit?

- M[te}

h aijidtfbalWiliM

'iwishinXtifsGnest Channels, operation in the applicable Jonist SL _ NS1littDOWal MODE may proceed provided the following . L hi '~ "glhD1hougii2 conditions are satisfied:

W@IME41IhiotietheinopentaMU

a. One inoperable channel is placed in EastorettMiineperen16hentslito the bypassed condition within 6 DPEltAB14 hours, and. )ies in Estd(4tMustitJalsn@

eDbMS8WIDC111%Mitiiisi 25bsbrs Ssi thissutT&henka?stidfl6JCdLD

b. The other inoperable channel is hBUTDOWN!sithiniths%fdlio0inIDd placed in the tripped condition hodr,s3 h 2 !s!

within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and. <

h@CTIDNNE'N35 cb"asEElsiasyi SeYia~op'7sb1Rf6EES $@

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c. All of the channels associated with itdT1thbucatori servell haseMes tisagjpe s [ *e r the redundant 4kV Safeguards Bus Sp6Eificatida W If33Restord$atklaast ,0,,."

are OPERABLE. Onlitidhria151?ifd 7098tAttf atstus Mthis u ithisf1thoetrYoilinitietStheishtism cx 6

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IlOTr. .icdadibeilm"at31eesi40T c to hug 90NW%il.*MVth inantEbhonisMand3ii 'E COLIP3801!BlNBijtithis3th41fd11ownsidQ homesid

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f. , 4 B.3.5-1 REV-441 S A10/94 3.5 INSTRUMENTATION SYSR.M DBr&ti Instrumentation has been provided to sense accident conditions and to initiate reactor trip and operation of the Engineered Safety Features (Reference 1). The OPERABILITY of the Reactor Trip System and the Engineered Safety System instrumentation and interlocks ensures that: (1) the associated ACTION and/or reactor trip will be initiated when the parametet monitored by each channel or combination thereof reaches its setpoint. (2) the specified coincidence logic and sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance consistent with maintaining an appropriate level of reliability of the Reactor Protection and Engineered Safety Features instrumentation and, (3) sufficient system functions capability is available from diverse parameters.

The OPERABILITY of these systems is required to provide the overall reliability redundancy and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions. The integrated operation of each of these systems is consistent with the assumptions used in the safety analysis.

Specified surveillance and maintenance outage times have been determined in accordance with WCAP-10271, " Evaluation of Surveillance Frequencies and

, Out of Service Times for the Reactor Protection Instrumentation System",

l and supplements to that report. Out of service times were determined '

i based on maintaining an appropriate level of reliability of the Reactor i Protection SyJtem and Engineered Safety Features instrumentation'.ET(BErid I XElF5EESTTrEEs~EEumEEEElWYSidh3Eoilde^dfiEEdiEf3Dr'anhfer24f h a fetElinj e cti$$suc tionivasdiquodelkcIlinithis sWCAP , j The evaluation of surveillanes recquenc4es and out of service times for l the reactor protection and engineered safety feature instrumentation described in WCAP-10271 included the allowance for testing in bypass. The evaluation assumed that the average amount of time the channels within a I given trip function would be in bypass for testing was 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

l Safety Injection l

l The Safety Iniection System is actuated automatically to provide emergency j j cooling and ruduction of reactivity in the event if a loss-of-coolant ;

accident or a eteam line break accident.

Safety injection in response to a loss-<!-c: nn. accident (LOCA) is provided by a high containment pressurs s -! 3nal wacked up by the low pressurizer pressure signal. These cvrc.H ict s would accompany the depressurization and coolant loss during 9 LOCA.

Safety injection in response to a stsam itne break is provided directly by a low steam line pressure signal, backc3 up by the low pressurizer pressure signal and, in case of a break within the containment, by the high containment pressure signal.

The safety injection of highly borated water will offset the temperature-induced reactivity addition that could otherwise result from cooldown following a steam line break.

B.3.5-4 REV 111 S/20/04 3.5 INSTRUMENTATION SYSTEM Basas continued M EN E NN **F7ih393cENUE60 The( MM '

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hesocistudi isEssidimanc$$#astsidiM$NasT4Eis mditpibe3NMalas Limiting Instrument Setpoints

1. The high containment pressure limit is set at about 10% of the maximum internal pressure. Initiation of Safety Injection protects against loss of coolant (Reference 2) or steam line break accidents as discussed in the safety analysis.

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2. The Hi-Hi containment pressure limit is set at about 50% of the maximum internal pressure for initiation of containment spray and at about 30% for initiation of steam line isolation. Initiation of Containment Spray and Steam Line Isolation protects against large loss of coolant (Reference 2) or steam line break accidents (Reference 3) as discussed in the safety analysis.

3. The pressurizer low pressure limit is set substantially below system operating pressure limits. Fowever, it is sufficiently high to protect against a loss of coolant accident as shown in the safety analysis (Reference 2). ,

4. The steam line low pressure signal is lead / lag compensated and its set-point is set well above the pressure expected in the event of a l large steam line break accident as shown in the safety analysis l (Reference 3). )

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5. The high steam line flow limit is set at approximately 20% of nominal j full-load flow at the no-load pressure and the high-high steam line I flow limit is set at approximately 120% of nominal full-load flow at

! the full load pressure in order to protect against large steam break

[ accidents. The coincident low T.,,, setting limit for steam line isolation initiation is set below its hot shutdown value. The safety analysis shows that these settings provide protection in the event of a large steam break (Reference 3). '

I 6. Steam generator low-low water level and 4.16 kV Bus 11 and 12 (21 and i 22 in Unit 2) low bus voltage provide initiation signals for the Auxiliary Feedwater System. Selection of these setpoints is discussed in the Bases of Section 2.3 of the Technical Specification.

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B.3.5-5 REV 111 S/10/94 .

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3.5 INSTRUMENTATION SYSTEM Bases continued Limiting Instrument Setpoints (continued)

7. High radiation signals providing input to the Containment Ventilation ;

Isolation circuitry are set in accordance with the Radioactive l Effluent Technical Specifications. The setpoints are established to l prevent exceeding the limits of 10 CFR Part 20 at the SITE BOUNDARY. ,

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8. The degraded voltage protection setpoint is 294.8% and 596.2% of nominal 4160 V bus voltage. Testing and analysis have shown that all ;

safeguards loads will operate properly at or above the minimum l degraded voltage setpoint. The maximum degraded voltage setpoint is l chosen to prevent unnecessary actuation of the voltage restoring l scheme at the minimum expected grid voltage. The first degraded voltage time delay of 8 i 0.5 seconds has been shown by testing and ;

analysis to be long enough to allow for normal transients (i.e., motor j starting and fault clearing). It is also longer than the time required to start the safety injection pump at minimum voltage. The second degraded voltage time delay is provided to allow the degraded i voltage condition to be corrected within a time frame which will not ;

cause damage to permanently connected Class IE loads.

The undervoltage setpoint is 75 i 2.5% of nominal bus voltage. The j minimum setpoint ensures equipment operates above the limiting value of 75% (of 4000 V) for one minute operation. The 75% maximum setpoint is chosen to prevent unnecessary actuation of the voltage restoring j scheme during voltage dips which occur during motor starting. The l undervoltage time delay of 4 1.5 seconds has been shown by testing j and analysis to be long enough to allow for normal transients and j short enough to operate prior to the degraded voltage logic, providing I a rapid transfer to an alternate source.

Instrument Operating Conditions During plant operations, the complete instrumentation systems will normally be in service. Reactor safety is provided by the Reactor Protection System, which automatically initiates appropriate action to prevent exceeding established limits. Safety is not compromised, however, by continuing operation with certain instrumentation channels out of service since provisions were made for this an the plant design. This specification outlines limiting conditions for operation necessary to preserve the effectiveness of the Reactor Control and Protection System when any one or more of the channels is out of service.

Almost all reactor protection channels are supplied with sufficient l redundancy to provide the capability for CHANNEL CALIBRATION and test at '

power. Exceptions are backup channels such as reactor coolant pump breakers. The removal of one trip channel on process control equipment is accomplished by placing that channel bistable in a tripped mode: e.g.. a l two-out-of-three circuit becomes a one-out-of-two circuit. The source and I intermediate range nuclear instrumentation system channels are not

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intentionally.placed in a tripped mode since these are one-out of-two !

trips, and the trips are therefore bypassed during testing. Testing does not trip the system unless a trip condition exists in a concurrent channel.

1 References

1. USAR, Section 7.4.2 I

2. USAR, Section 14.6.1

3. USAR, Section 14.5.5

4. FSAR, Appendix I l i

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l EXHIBIT C j

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PRAIRIE ISLAND NUCLEAR GENERATING STATION License Amendment Request dated November 25,1998 Appendix A, Technical Specification Pages Revised Pages i

TS.3.2-1 Table TS.3.5-2B (Page 6 of 9)

Table TS.3.5-2B (Page 9 of 9)

B.3.5-1 B.3.5-4 B.3.5-5 ,

B.3.5-6 I

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TS.3.2-1 1

l 3.2 CHEMICAL AND VOLUME CONTROL SYSTEM i

Applicability Applies to the operational status of the chemical and volume control '

system.

Qbjective To define those condition of the chemical and volume control system necessary to assure safe reactor operation and safe COLD SHUTDOWN.

Specification

]

A. When fuel is in a reactor and reactor coolant system average temperature is at or below 200*F there shall be at least one flow path to the core for boric acid injection. If no OPERABLE flow path exists, suspend all operations involving CORE ALTERATIONS or positive reactivity changes. l l

B. A reactor shall not be made or maintained critical nor shall the reactor coolant system average temperature exceed 200 F unless the following conditions are satisfied (except as specified in 3.2.C or 3.2.D below, or Table TS.3.5-2B):

1. Two of the three charging pumps shall be OPERABLE.

2. At least one boric acid tank shall be aligned to the unit and shall contain a minimum of 2000 gallons of 11.5% to 13% by weight boric acid solution at a temperature of at least 145F.

3. System piping, valves and pumps shall be OPERABLE to the extent l

of establishing two independent flow paths for boric acid

injection -- one flow path from the boric acid tanks to the core and one flow path from the refueling water storage tank to the Core.

4. Two channels of heat tracing shall be OPERABLE for the flow paths from the boric acid tanks required to meet the requirements of Specification 3.2.B.3.

l S. Automatic valves, piping, and interlocks associated with the above

! components which are required to operate for the steam line break l accident are OPERABLE.

,. e TABLE TS.3.5-2B (Page 6 of 9)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATIO_H M_INIMUM TOTAL NO. CHANNELS CHANNELS APPLICABL_E FUNCTIONAL UNIT OJ CHANNELS TO TRIP OPERABLE MODES ACTION

8. LOSS OF POWER

a. Degraded Voltage 4/ Bus 2/ Bus 3/ Bus 1, 2, 3, 4 31, 32, 33 4kV Safeguards Bus (2/phace on (1/ phase 2 phases) on 2 phases)

b. Undervoltage 4/ Bus 2/ Bus 3/ Bus 1, 2. 3, 4 31, 32, 33 4kV Safeguards Bus (2/ phase on (1/ phase 2 phases) on 2 phases)

9. BORIC ACID STORAGE TANK

a. Lo-Lo Level 2 channels 1 sensor 2 sensors 1. 2, 3, 4 34 with 2 per in one sensors per channel channel channel in both channels

b. Automatic Actuation Logic 2 1 2 1. 2. 3. 4 35, 36 and Actuation Relays E$

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b

t IABLE 3.5-2B (Page 9 of 9)

Action Statem_qrtt_a ACTION 30: With the number of OPERABLE channels ACTION 33: If the requirements of ACTIONS 30 or 31 one less than the Total Number of cannot be met within the time Channels, declare the associated specified, or with the number of auxiliary feedwater pump inoperable and OPERABLE channels three less than the take the action required by Total Number of Channels, declare the Specification 3.4.2. However, one associated diesel generator (s) channel may be bypassed for up to 8 inoperable and take the ACTION required hours for surveillance testing per by Specification 3.7.B.

Specification 4.1. provided the other channel is OPERABLE. ACTION 34: With the number of OPERABLE ,

channels less than the Total ACTION 31: With the number of OPERABLE channels Number of Channels, operation may one less than the Total Number of proceed provided the inoperable Channels, operation in the applicable channel is placed in the tripped MODE may proceed provided the condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and the inoperable channel is placed in the Minimum Channels OPERABLE bypassed condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . requirement is met. Restore the inoperable channel to OPERABLE ACTION 32: With the number of OPERABLE channels status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at two less than the Total Number of least HOT SHUTDOWN within the next Channels, operation in the applicable 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN MODE may proceed provided the following within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

conditions are satisfied:

ACTION 35: With one channel inoperable,

a. One inoperable channel is placed in restore the inoperable channel to the bypassed condition within 6 OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or hours, and, be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD

b. The other inoperable channel is SHUTDOWN within the following 30 placed in the tripped condition hours. 2 52 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and. Qp ACTION 36: Two channels may be inoperable for up M

c. All of the channels associated with to I hour for surveillance testing per s the redundant 4kV Safeguards Bus Specification 4.1. Restore at least &Pw are OPERABLE. one channel to OPERABLE status within this I hour or initiate the action *6 necessary to place the affected unit in a HOT SHUTDOWN, and be in at least HOT W SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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B.3.5-1 3.5 INSTRUMENTATION SYSTEM BAFEJi Instrumentation has been provided to sense accident conditions and to initiate reactor trip and operation of the Engineered Safety Features (Reference 1). The OPERABILITY of the Reactor Trip System and the Engineered Safety System instrumentation and interlocks ensures that: (1) {

the associated ACTION and/or reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its setpoint, (2) the specified coincidence logic and sufficient redundancy is maintained to permit a channel to be out of service for testing or '

maintenance consistent with maintaining an appropriate level of I reliability of the Reactor Protection and Engineered Safety Features l l

instrumentation and, (3) sufficient system functions capability is available from diverse parameters.

1 The OPERABILITY of these systems is required to provide the overall ;

i reliability, redundancy and diversity assumed available in the facility {

l design for the protection and mitigation of accident and transient !

conditions. The integrated operation of each of these systems is I consistent with the assumptions used in the safety analysis. l Specified surveillance and maintenance outage times have been determined in accordance with WCAP-10271, " Evaluation of Surveillance Frequencies and l Out of Service Times for the Reactor Protection Instrumentation System",

and supplements to that report. Out of service times were determined based on maintaining an appropriate level of reliability of the Reactor Protection System and Engineered Safety Features instrumentation. (Boric Acid Storage Tank instrumentation which provides automatic transfer of ;

safety injection suction was not modeled in this WCAP.)

The evaluation of surveillance frequencies and out of service times for the reactor protection and engineered safety feature instrumentation described in WCAP-10271 included the allowance for testing in bypass. The evaluation assumed that the average amount of time the channels within a given trip function would be in bypass for testing was 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . ,

i Safety Injection l The Safety Injection System is actuated automatically to provide emergency

! cooling and reduction of reactivity in the event of a loss-of-coolant j accident or a steam line break accident, i

Safety injection in response to a loss-of-coolant accident (LOCA) is

( provided by a high containment pressure signal backed up by the low l

pressurizer pressure signal. These conditions would accompany the j

depressurization and coolant loss during a LOCA.

I Safety injection in response to a steam line break is provided directly by

a low steam line pressure signal, backed up by the low pressurizer pressure signal and, in case of a break within the containment, by the

'. high containment pressure signal.

The safety injection of highly borated water will offset the temperature-induced reactivity addition that could otherwise result from

. cooldown following a steam line break, i

)

B.3.5-4 i

3.5 INSTRUMENTATION SYSTEM l Hagga continued i Automatic Transfer of Safety Injection Suction I The plant is equipped with three boric acid storage tanks for the two units. One tank is normally aligned to the safety injection system for each unit. Following initiation of the Engineered Safety Features, the safety injection pumps take suction from the aligned boric acid storage tank. When the boric acid storage tank level falls to the lo-lo level, an interlock automatically transfers the safety injection pumps suction from the boric acid storage tank to the refueling water storage tank. The boric acid storage tank that is not aligned to either unit, including its associated piping and interlocks, is not required to be OPERABLE.

Limiting Instrument Setpoints

1. The high containment pressure limit is set at about 10% of the maximum internal pressure. Initiation of Safety Injection protects against loss of coolant (Reference 2) or steam line break accidents as discussed in the safety analysis.

2. The Hi-Hi containment pressure limit is set at about 50% of the l maximum internal pressure for initiation of containment spray and at l about 30% for initiation of steam line isolation. Initiation of

Containment Spray and Steam Line Isolation protects against large loss of coolant (Reference 2) or steam line break accidents (Reference 3) as discussed in the safety analysis.

3. The pressurizer low pressure limit is set substantially below system operating pressure limits. However, it is sufficiently high to protect against a loss of coolant accident as shown in the safety i

analysis (Reference 2).

4. The steam line low pressure signal is lead / lag compensated and its l set-point is set well above the pressure expected in the event of a l large steam line break accident as shown in the safety analysis l (Reference 3).

5. The high steam line flow limit is set at approximately 20% of nominal full-load flow at the no-load pressure and the high-high steam line flow limit is set at approximately 120% of nominal full-load flow at the full load pressure in order to protect against large steam break

( accidents. The coincident low T ,, setting limit for steam line isolation initiation is set below its hot shutdown value. The safety i

analysis shows that these settings provide protection in the event of a large steam break (Reference 3).

6. Steam generator low-low water level and 4.16 kV Bus 11 and 12 (21 and 22 in Unit 2) low bus voltage provide initiation signals for the Auxiliary Feedwater System. Selection of these setpoints is discussed in the Bases of Section 2.3 of the Technical Specification.

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B.3.5-5 1

3.5 INSTRUMENTATION SYSTEM i i

Bases continued Limiting Instrument Setpoints (continued)

7. High radiation signals providing input to the Containment Ventilation ;

Isolation circuitry are set in accordance with the Radioactive Effluent Technical Specifications. The setpoints are established to prevent exceeding the limits of 10 CFR Part 20 at the SITE BOUNDARY. ;

8. The degraded voltage protection setpoint is 294.8% and $96.2% of nominal 4160 V bus voltage. Testing and analysis have shown that all safeguards loads will operate properly at or above the minimum degraded voltage setpoint. The maximum degraded voltage setpoint is chosen to prevent unnecessary actuation of the voltage restoring scheme at the minimum expected grid voltage. The first degraded voltage time delay of 8 i 0.5 seconds has been shown by testing and i analysis to be long enough to allow for normal transients (i.e., motor starting and fault clearing). It is also longer than the time required to start the safety injection pump at minimum voltage. The second degraded voltage time delay is provided to allow the degraded voltage condition to be corrected within a time frame which will not cause damage to permanently connected Class IE loads.

l The undervoltage setpoint is 75 1 2.5% of nominal bus voltage. The 2 minimum setpoint ensures equipment operates above the limiting value j l of 75% (of 4000 Vi for one minute operation. The 75% maximum setpoint j is chosen to prevent unnecessary actuation of the voltage restoring .

I scheme during voltage dips which occur during motor starting. The undervoltage time delay of 4 1.5 seconds has been shown by testing and analysis to be long enough to allow for normal transients and short enough to operate prior to the degraded voltage logic, providing a rapid transfer to an alternate source.

l Instrument Operating Conditions During plant operations, the complete instrumentation systems will normally be in service. Reactor safety is provided by the Reactor Protection System, which automatically initiates appropriate action to prevent exceeding established limits. Safety is not compromised, however.

by continuing operation with certain instrumentation channels out of service since provisions were made for this in the plant design. This ,

specification outlines limiting conditions for operation necessary to )

preserve the effectiveness of the Reactor Control and Protection System l when any one or more of the channels is out of service.

Almost all reactor protection channels are supplied with sufficient a redundancy to provide the capability for CHANNEL CALIBRATION and test at power. Exceptions are backup channels such as reactor coolant pump l breakers. The removal of one trip channel on process control equipment is accomplished by placing that channel bistable in a tripped mode: e.g., a

! two-out-of-three circuit becomes a one-out-of-two circuit. The source and intermediate range nuclear instrumentation system channels are not l _ __ __ - _ ,

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I. B.3.5-6 3.5 . INSTRUMENTATION SYSTEM Bases. continued Instrument Operating Conditions (continued)

'ntentionally i placed in a tripped mode since these are one-out of-two trips,'and the trips are therefore bypassed during testing. Testing does not. trip'the' system unless a trip condition exists in a concurrent channel. i References

1. USAR Section 7.4.2 !

2. USAR, Section 14.6.1

3. USAR, Section 14.5.5 ;

4. FSAR, Appendix.I- !

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FIGURE 1 BAST to RWST Transfer Logic Diagram (Two Sheets)

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FIGURE 2 Safety injection System Simplified Flow Diagram i

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