Text

Amends 210 & 88 to Licenses DPR-66 & NPF-73,respectively, Relocating Reactor Trip Sys & ESFAS Response Times from TS Tables 3.3-2 & 3.3-5 to Section 3 of Plant Licensing Requirements Manual
	ML20199F292
Person / Time
Site:	Beaver Valley
Issue date:	01/20/1998
From:	Stolz J; Office of Nuclear Reactor Regulation
To:	; Duquesne Light Co, Ohio Edison Co, Pennsylvania Power Co
Shared Package
ML20199F298	List: ML20199F292; ML20199F305;
References
	DPR-66-A-210, NPF-73-A-088 NUDOCS 9802030126
	Download: ML20199F292 (23)
	v • d • e

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DUQUESNE LIGHT COMPANY OHIO EDISDN COMPANY PENNSYLVANIA POWER COMPANY i

DOCKET No. 50-334 BEAVER VALLEY POWER STATION. UNIT No. I b

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AMENDMENT TO FACILITY OPERATING LICENSE 1

Amendment No.210 License No. DPR-66 1.

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

A.

The application for amendment by Duquesne Light Com>any, et al. (the licensee dated September 11, 1997, complies with tie standards and requireme)nts of the Atomic Energy Act of 1954, as amended (the i

and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will o erate in conformity with the a!! ions of the

lication, 4

the provisions of t e Act, and the rules and regul Commission; C.

There is reasonable assurance (i) that the activities authorized 4

by this amenoment can be conducted without endangering the health i

and safety of the public, and (ii) that such activities will be i

conducted in compliance with the Commission's regulations; J

D.

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

The issuance of this amendaent is in accordance with 10 CFR Part 4

51 of the Comission's regulations and all applicable requirement:

j have been-satisfied.

i 9902030126 990120 PDR ADOCK 05000334 P

PDR

e l 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating Lic3nse No. DPR-66 is hereby amended to read as follows:

(2) Technical Sancifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 210, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

In addition, the license is amended by changes to paragraph 2.C.(10) to the Facility Operating License No. DPR-66 as follows:

(10) Additional Conditions The Additional Conditions contained in Appendix C, as revised through Amendment No. 210

, are hereby incorporated into this license. Duquesne Light Company shall operate the facility in accordance with the Additional Conditions.

3.

This license amendment is effective as of the date of its issuance, to be implemented within 30 days.

Implementation of this amendment shall include the relocation of these technical specification requirements to the appropriate documents, as described in the licensee's application dated September ll, 1997, and evaluated in the staff's safety evaluation attached to this amendment.

FOR THE NUCLEAR REGULATORY C0m!SS10N kb Jopn'F. Stolz, Director Project Directorate I-2 ivision of Reactor Projects - I/II office of Nuclear Reactor Regulation Attachments:

1.

Page 1 of Ap>endix C of License

No. DPR-66 2.

Changes to tie Technical Specifications 9 ate of Issuance: January 20, 1998

Page 1 of Appendix C is attached, for convenience, for the composite license to reflect this change.

I ATTAC MENT TO LICENSE AMENDMENT ND R FACILITY OPERATING LICENSE NO. DPR-66 D0tKET No. 50-334 1.

Revise Appendix C of the License as follows:

Remove Page Insert Paan 1

1 2.

Replace the following pages of Appendix A Technical Specifications, with the enclosed > ages as indicated. The revised pages are identified by amendment num>er and contain vertical lines indicating the areas of change.

Remove Insert XVI XVI 3/4 3-1 3/4 3-1 3/4 3-7 3/4 3-7 3/4 3-9 3/4 3-10 3/4 3-14 3/4 3-14 3/4 3-24b 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 B3/43-la

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APPDWIX C ADDIT 10hAL CON 01T10hf OPERATING L. CENSE.to. D >t-66 Duquesne Light Company, Ohio Edison Company, and Pennsylvania Power Company shall comply with the following conditions on the schedules noted below:

Amendment Additional Condition Implementation Number Date 202 The licensee is authorized to relocate certain The amendment Technical Specification requirements to shall be licensee-controlled documents.

Implementation implemented of this amendment shall include the relocation within 60 days of these technical specification requirements from April 14, to the appropriate documents, as described in 1997 the licensee's application dated September 9, 1996, and evaluated in the staff's safety evaluation attached to this amendment.

208 The licensee commits to perform the mst weld The amendment heat treatment of sleeve welds and tse shall be NRC-reconnended inspections for repaired tubes implemented as described in the licensee's application within 60 days dated March 10, 1997, as supplemented July 28 from and September 17, 1997, and evaluated in the November 25, staff's safety evaluation attached to this 1997 amendment.

209 The licensee is authorized to relocate certain The amendment Technical Specification requirements to shall be licensee-controlled documents.

Implementation implemented of this amendment shall include the relocation within 60 days of these technical specification requirements from to the appropriate documents,.as described in December 10, the licensee's application dated March 14, 1997 1997, as supplemented July 29 and August 13, 1997, and evaluated in the staff's safety evaluation attached to this amend.nent.

210 The licensee is authorized to relocate certain The amendment Technical Specification requirements to shall be licensee-controlled documents.

Implementation implemented of this amendment shall include the relocation within 30 days of these technical specification requirements from to the appropriate documents, as described in January 20, 1998 the licensee's application dated Septec6er 11, 1997, and evaluated in the staff's safety evaluation attached to this amend:nent.

1 Ameadment No. 403, 98, 409.210

s DPR-66 1

M Z&BLE 21TLE PAGE 2.2-1 Reactor Trip System Instrumentation Trip 2-6 Setpoints-3.1-1 Accident Analyses Requiring Reevaluation 3/4 1-19a in the event of an Inoperable Pull or Part Length Rod 3.2-1 DNB' Parameters 3/4 2-13.

3.3-1 Reactor Trip System Instrumentation 3/4 3-2 1

4.3-1

-Reactor Trip System Instrumentation 3/4 3-11 surveillance Requirements 3.3-3 Engineered Safety Featurou Actuation System 3/4-3-15 Instrumentation 3.3-4 Engineered Safety Features Actuation System 3/4 3-22 Instrumentation Trip 8etpoints I-4.3-2 Engineered Gafety Feature Actuation System 3/4 3-29 Instrumentation Surveillance Requirements 3.3-6 Radiation Monitoring Instrumentation 3/4.3-34 4.3-3 Radiation Monitoring Instrumentation 3/4 3-36 surveillance Requirements 3.3-7 seismic Monitoring Instrumentation 3/4 3-39 4.3-4 seismic Monitoring Instrumentation 3/4 3-40 surveillance Requirements 3.3-8 Meteorological Monitoring Instrumentation 3/4 3-42 4.3 Meteorological Monitoring Instrumentation 3/4 3-43 Burveillance Requirements 3.3-9 Remote Shutdown Panal-Monitoring 3/4 3 Instrumentation 4.3-6 Remote Shutdown Monitoring Instrumentation 3/4 3-46 Surveillance Requirements

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BEAVER VALLEY - UNIT 1 XVI Amendment No. 210 d

DPR-66 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP BYSTEM INSTRUMENTATION 1

LIMITING CONDITION FOR OPERATION 3.3.1.1 As a minimum, the reactor trip systcm instrumentation channels and-interlocks of Table 3.3-1 shall be OPERABLE.

APPLICABILITY:

As shown in Table 3.3-1.

ACTION:

As shown in Table 3.3-1.

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SURVEILLANCE REQUIRENENTS 4.3.1.1.1 Each reactor trip system instrumentation channel shall be demonstrated OPERABLU by the performance - of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNC'IONAL TEST operations during the modes and at the frequencies shown ;.a 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 per 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.. Neutron detectors are exempt fron response time testing.. Each test shall include at least one logic train such that both logic trains are tested at least once per 36 months and - one channel per function such that all channels are tested at least once every N times 18 months-where N-is the total number of redundant channels.in a specific reactor trip function as shown in the " Total No. of Channels" column of Table 3.3-1.

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-BEAVER VALLEY - UNIT 1 3/4 3-1 Amendment No. 210

DPR-66 TABLE 3.3-1 (Continuedl 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 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months st operation may continue until performance of the next required CHANNEL FUNCTIONAL TEST.

ACTION O -

Not applicable.

l 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 Uithin 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 braskers.

ACTION 39 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirainunt, 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

hours;

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 OPEPABLE.

BEAVER VALLEY - UNIT 1 3/4 3-7 Amendment No. 210 (Next page is 3/4 3-11) l J

DPR-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 the values shown in the Trip setpoint column of Table 3.3-4.

APPLICABILITY:

As shown in Table 3.3-3.

ACTIQHi a.

With an engineered safety feature actuation system instrumentation channel trip setpoint less conservative 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 syrdam 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 aftected by interlock operation.

4.3.3.1.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESP 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 Amendment No. 210

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DPR-66 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 PROTECTIVE AND ENGINEERED SAFETY FEATURES (ESP) l}iSTRUMENTATION fContinued)

ESF response times which include sequential operation of the RWST and VCT valves are based on values assumed in the Non-LOCA safety analyses and are provided in section 3 of the Licensing Requirements Mandal.

These analyses take credit for injection of borated water.

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

When sequential operation of the RWST and VCT valves is not ingluded in the response times, the values l

specified are based on the LOCA analyses.

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

Verification of the response times will assure that the assumptions used for the LOCA l

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 repantial, 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 followir.g 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 Open), and enables auto actuation of the pressuriser PORVs.

BEAVER VALLEY - UNIT 1 B 3/4 3-la Amendment No. 210

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UNITED STATES s

j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 30MH001

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DUQUESNE LIGHT COMPANY OHIO EDISON COMPANY THE CLEVELAND ELECTRIC ILLUMINATING COMPANY THE TOLEDO EDISON COMPANY DOCKET NO. 50-412 BEAVER VALLEY POWER STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 88 License No. NPF-73 1.

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

A.

The application for amendment by Duquesne Light Com)any, et al. (the licrasee) dated September 11, 1997, complies with tie standards and recutrements of the Atomic Energy Act of 1954, as amended (the Act),

anc the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will o>erate u. conformity with the application, the provisions of tie Act, and the rule; and regulations of the Comission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the hesith and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

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

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

Y 4

-t-2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2)llows:of Facility Operatthg License No. NPF-73 is hereby amended to read as fo (2) Technical snacifications The Technical Specifications contained in Appendix A, as revised through Amendment No.88, and the Environmental Protection Plan contained in Apg.endix B, both of which are attached hereto are hereby incorporatcd in the license. DLC0 shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

In addition, the license is amended by changes to naragraph 2.C.(ll) to l

the Facility Operating License No. DPR-73 as follows:

(11) Additional conditions -

The Additional Conditions contained in Appendix D, as revised through Amendment No. 88

, are hereby incorporated into this license. - Duquesne Light Company shall operate tire facility in accordance with the Additional Conditions.

3.

This license amendment is effective as of the date of its issuance, to be implemented within 30 days, implementation of this amendment shall include the relocation of these technical specification requirements to the appropriate documents, as described in the licensee's application dated September 11, 1997, and evaluated in the staff's safety evaluation attached to this amendment.

FOR YHE NUCLEAR REGULATORY COMISSION L

Jh F. Stolz, Directdr Project Directorate I-2 Division of Reactor Projects - I/II office of.'h:cInr donctor Regulation Attachments:

1.

Page 1 of Appendix D of License

No. NPF-73 2.

Changes to the Technical Specifications Date of Issuance: January 20, 1998 i

Page 1 of Appendix D is attached, for convenience, for the composite license to reflect this change.

ATTACMENT 10 LICENSE AMENDMENT N0.88 FACILITY OPERATING LICENSE N0. NPF-73 DOCKET No. 50-412 1.

Revise Appendix D of the License as follows:

Remove Page Inmart Pane 1

1 Replace the following pages of Appendix A Technical Specifications,by 2.

with the enclosed sages as indicated.

The revised pages are identified amendment num>er and contain vertical lines indicating the areas of change.

Remove Insert 3/4 3-1 3/4 3-1 3/4 3-)

3/4 3-7 3/4 3-8 3/4 3-9 3/4 3-14 3/4 3-14 3/4 3-15 3/4 3-15 3/4 3-28 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 8 3/4 3-2 B 3/3 3-3 8 3/3 3-3 8 3/4 3-4 B3/43-4

APPEND 11 D I

ADDIT 10hR CONDITIGAf OPERATING L CENSE ND. N U-73 i

Duquesne Light Company, Ohio Edison Company. The Cleveland Electric Illuminating Company, and The Toledo Edison Compan following conditions on the schedules noted below:y shall comply with the Amendment Additional Condition Implementation Number Date 83 The licensee is e nhorized to relocate certain The amendment Technical Specification requirements to shall be licensee-controlled documents.

Implementation implemented of this amendment shall include the relocation within 60 days of these technical specification requfrements from April 14, to the appropriate documents, as described in 1997 the licensee's application dated September 9, 1996, and evaluated in the staff's safety evaluation attached to this amendment.

87 The licensee is authorized to relocate certain The amendment Technical Specification requirements to shall be licensee-controlled documents.

Irplementation implemented of this amendment shall include the relocation within 60 days of these technical specification requiraments from to the appropriate documents, as described in December 10, the licensee's application dated March 14, 1997 1097, as supplemented July 29 and August 13, 1997, and evaluated in the staff's safety evaluation attached to this amendment.

88 The-licensee is authorized to relocate certain The amendment Technical Specification requirst nts to shall be licensee-controlled documents.

amplementation implemented-of this amendment shall includ? the relocation within 30 days of these technical specification requirements from to the appropriate documents, as described in January 20, 1998 the licensee's application dated September 11, 1997, and evaluated in the staff's safety evaluation attached to this amendment.

1 Amendment No. SS, 87, 88

NPF-73 3/4.3 INBTRUMENTATION I

I 3/4.3.1 REACTOR TRIP BYSTEM INBTRUMENTATION LINITING COND:: TION 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.

I APPLICABILITY:

As shown in Table 3.3-1.

ACTION:

As shown in Table 3.3-1.

l SURVEILLANCE REQUIREMENTS 4.3.1.1.1 Each reactor trip system instrumentation channel and interlock and automatic trip logic shall be demonstrated OPERABLE by the performance of th Reactor Trip System Instrumentation Surveillance Raquirements(j) during the MODES and at the frequencies l

shown in Tabin 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 per 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.

Neutron detectors are exempt from response time testing.

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 avrxy 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.

(1) For the automatic trip logic, the surveillance requirements shall l

be the application of various simulated input combinations in conjunction with each possible interlock logic state and verification of. the' required logic output including, as a

minimum, a continuity check of output devices.

-BEAVER VALLEY - UNIT 2 3/4 3-1 Amendment No. 88

NPF-73 TABLE 3.3-1 (Continuedi 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 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; operation may continuc 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 a 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 - hours or declare the breaker incperable and be in HOT STANDBY within the Dext 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

hours;

however, oas channel may be bypassel 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 OPERABIE.

M JION 44 - With less than the Minhum Number of Channels OPERABLE, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 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.

BEAVER VALLEY - UNIT 2 3/4 3-7 Amendment No. 88 (Next page is 3/4 3-10) l

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NPF-73 l

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 (ESFAS) instrumentation channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their Trip Setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4.

APPLIC)d}2LITY:

As shown in Table 3.3-3.

ACTION:

a.

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

b.

With an ESFAS Instrumentation or Interlock Trip Setpoint less conservative than the value shown in the Al]owable 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 ifbperable and apply the applicable ACTION statement requirements of Table 3.3-3 until the channel is restored to OPERABLE status with its Setpoint adjusted consistent with the Trip Setpoint value.

EQUATION 2.2-1 Z + R + S 5 TA where:

Z=

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

channel, R=

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

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

c.

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

BEAVER VALLEY - UNIT 2 3/4 3-14 Amendment No. 88

NPF-73 l

INSTRUMENTATION t

3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 4.3.2.1.1 Each engineered safety feature actuation system instrumentation channel and interlock and the automatic actuation l

logic with master and slave relays shall be demonstrated OPERABLE by the performa:.ca of the ESFAS Instrumentation Surveillance Requisements W during the MODES and at the frequencies shown in l

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.

'ue total interlock function shall be demonstrated OPERABLE at least Unce 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 demons' rated to be within the limit at least once c

per 18 months.

Each test shall includo at least one logic train such l

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 numbJ of redundant channels in a specific ESF function as shown in the rotal No. Of Channels" Column of Table 3.3-3.

(1) For the automatic actuation logic, the surveillance requirements l

shall be the application of various simulated input conditions in conjunction witt.

each 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 continuity check of associated actuation devices that are not testable.

BEAVER VALLEY - UNIT 2 3/4 3-15 Amendment No. 88 l

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TABLE 3.3-4 (Continued)

NPF-73 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM IN&i-RUni.niATION TRIP SETPOINTS TOTAL SENSOR TRIP ALLOWABLE FUNCTIONAL UNIT ALLOWANCE (TA)

Z DRIFT (S)

SETPOINT VALUE 8.

ENGINEERED SAFETY FEATURE INTERLOCKS a.

Reactor Trip, P-4 N.A.

N.A.

b.

Pressurizer Pressure, N.A.

N.A.

5 2000 psig 5 2010 psig P-11 c.

Low-Low Targ, P-12 4.0 0.82 0.87 2 541*F 2 538.5'F BEAVER VALLEY - UNIT 2 3/4 3-28 Amendment No. 88 (Next page is 3/4 3-33) l

WPF-73 3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM AND ENGINErnun gAFtrY FEATUnan ACTUATION SYSTEM INsiktfriENTATION (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 uncertainty-magnitudes.

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

Being that. there is a small statistical chance that this will happen, an infrequent excessive drif t --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 possibility of 'an Anticipated Transient Without Scram (ATWS) event by ensuring OPERABILITY of the divene trip features (

Reference:

Generic Letter-85-09).

The measurement of. response time at the specified frequencies provides - assurance that the protective. and ESF action function associated with each channel is completed within the tiaw limit assumed - in the accident-analyses.

No credit was taken. in _ the analyses for those channels with response times indicated as_not applicable.

ESF response times:which include sequential operation of the RWST.

and VCT valves are based on values-assumed in the non-LOCA safety analyses and are provided-in Section 3

of the _ Licensing Requirements Manual.

Thsse analyses take credit for 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.

When sequential operation of. the RWST and VCT valves is not included in the response times, the values specified are based l

on the LOCA analyses.

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

Verification of the response times 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.

The maximum response time for control room isolation on high radiation is - based on ensuring that the control room remains habitable following a small line brea" outside the containment.

From a control room habitability aspect, the worst case accident that does not initiate a Containment Isolation - Phase B signal is s

BEAVER VALLEY - UNIT 2 B 3/4 3-2 Amendment No. 88

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NPF-73 3/4.3' INSTRUMENTATION BASES 3/4.3.1 and' 3/4.3.2 REACTOR TRIP SYSTEM AND ENGINRRnRn - SAFETY FEATURRE ACTUATION SYSTEM INSTRUMENTATION (Continued)-

the small line break outside the containment.

This response time includes radiation monitor-processing delays - associated with the

nonitor averaging techniques.

Diecal Generator starting and sequence loading delays are not included since these delays occur prior to the control room environment exceeding 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.

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

P-4 Reactor tripped ' -

Actuates; turbine

trip, ses-main feedwater valves on T.,

below setpoint, prevents the opening of the main feedw'ater 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, and enables safety injection and - steamline isolation (with Loop Stop Valve Open) on low steamline pressure.

Below the setpoint, P-11 allows the manual block of safety injection actuation on low pressurizar

pressure, allows manual block' of safety injection and

-steamline isolation (with Loop Stop Valve Open) on low steamline pressure and enables steamline isolation on high steam pressure rate.

P-12 Above the setpoint.

P-12 automatically reinstates an arming signal to t.? e steam dump system.

Below the setpoint P-12 blocks steam dump and allows manual bypass of the steam dump block to cooldown condenser dump valves.

BEAVER VALLEY - UNIT 2 B 3/4 3-3 Amendment No. 38

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4 NPF-73 3/4.3 INSTRUMENTATION BASES 3 / 4. 3.'1 and ' 3 / 4. 3. 2 REACTOR TRIP SYSTEM AND ' ENGINRRDFn SAFETY'

- FEATUDR81 ACTUATION SYSTEM INSTRUMENTATION.(Continued)

Table 3.3-1 Action 2 has been modified _ by two notes.

Note (4)

- allows placing the inoperabla channel in the bypass-condition for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months while-performing:

a) routine surveillance testing of other channels, and b) setpoint adjustments of other channels when required-to rednce the setpoint in accordance with other technical specifications.-

The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months time limit is justified in accordance with WCAP-10271-P-A, Supplement 2,

Revision 1,

June 1990.

Note ' (5) only requires SR 4.2.4 to be performed if a Power Range High Neutron Flux channel input to QPTR becomes inoperable.

Failure of a' component in the Power Range High Neutron Flux channel which renders the High Neutron Flux trip function inoperable may not affect the capability to monitor QPTR.

As such, determining QPTR using the movable incore detectors once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months may not be necessary.

3/4.3.3 -MONITORING-INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring channels ensures that:

1) the radiation levels are continually measured in the areas served by the individual. channels;: 2) the alarm or automatic

- action is initiated when' the _ radiation level trip setpoint is exceeded; and 3) sufficient information is available on selected plant-- parameters to monitor and assess these variables following an accident.

This capability is consistent with the recommendations of NUREG-0737, " Clarification of TMI Action Plan Requirements," October, 1980.

3/4.3.3.2 MOVABLE INCORE DETECTORS The OPERABILITY of the movable incore detectors with the specified minimum complement of equipment ensures that the measurements.

obtained from use of this system accurately represent-the spatial neutron flux distribution of the reactor core.

The OPERABILITY of this system is demonstrated by irradiating each detector used and y

determining the acceptability of itu voltage curve.

For the purpose of measuring Fg(Z) or Fh, a full incore flux map is used.

Quarter-core flux maps, as defined in WCAP-8648, June 1976, may be used in re-calibration of the excore neutron flux detection system, and full incore flux maps or symmetric incore thimbles may be used for monitoring the Quadrant Power Tilt Ratio when one Power Range Channel is inoperable.

BEAVER VALLEY - UNIT 2 B 3/4 3-4 Amendment No. 88

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ML20199F292

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