ML20236R190

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Proposed Tech Specs Extending TS Surveillances TS 4.20.2(a), Table 4.20-1,items 1,2 & 4,TS 4.1-1,Table 4.1-1,items 8,9,10 & 11 & TS 4.5.1.2.1,item b(1)
ML20236R190
Person / Time
Site: Oconee  Duke Energy icon.png
Issue date: 07/16/1998
From:
DUKE POWER CO.
To:
Shared Package
ML15112A665 List:
References
NUDOCS 9807210289
Download: ML20236R190 (25)


Text

.

1 I

l ATTACHMENT 1 V

TECHNICAL SPECIFICATION Remove Page Insert Page 4.1-3 4.1-3 4.1-3a 4.1-4 4.1-4 4.5-2 4.5-2 4.20-5 4.20-5

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4.5.1.1.3 Core Flooding System l

l a.

.Every 18 months, a system test shall be conducted to demonstrate proper operation of the system.

During pressurization of the Reactor Coolant System, verification shall be made that the check and isolation valves in the core flooding tank discharge lines operate properly.

b.

The test will be considered satisfactory if control board indication of core flood tank level verifies that all valves have opened.

4.5.1.2 Component Tests 4

4.5.1.2.1 Valves - Power Operated q

Valves LP-17. -18, shall only t e tested every cold shutdown unless previously tested during the a.

current quarter.

b.

Every 18 months, the following LPI system valves shall be cycled manually to verify the manual operability of these power operated valves:

6 (1) L1:, ump discharge (ES) LP-17,-l8 l

1 (2) LPI discharge throttling LP-12,-14 (3) LPI discharge header crossover LP-9.-10 (4) LPI discharge to liPl/RBS LP-15,-16 4.5.1.2.2 Check Valves Periodic individual leakage testing

  • of valves CF-12, CF-14, LP-47 and LP-48 shall be accomplished prior to power operation after every time the plant is placed in the cold shutdown condition for refueling, after each time the plant is placed in a cold shutdown condition for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if testing has not been accomplished in the preceding 9 months, and prior to returning the valve to service after maintenance, repair or replacement work is performed. Whenever integrity of these valves cannot be demonstrated, the integrity of the remaining valve in each high pressure line having a leaking valve shall be determined and reco Jed daily. In addition, the position of the other closed valve located in the high pressure piping shall be recorded daily. For the allowable leakage rates and limiting conditions for operation, see Technical Specificat!on 3.1.6.10.

PJsci The Emergency Core Cooling Systems are the principle reactor safety features in the event ofloss of coolent accident. The removal of heat from the core provided by these systems is designed to limit core damage.

The liigh Pressure Injection System under normal operating conditions has one pump operating. The liPI system test required by Specification 4.5.1.1.1 verifies that the llPI system responds as required to actuation of ES channels 1 and 2.

(a) To satisfy ALARA requirements, leakage may be measured indirectly (as from the performance of pressure indicators)if accomplished in acccrdance with approved procedures nd supported by computations showing that the method is capable of demonstrating valve comp ance with the leakage o

criteria.

(b) A one-time extension of the LPI pump discharge valves LP-17 and LP-18 manual cycle test frequency is allowed to a maximum of 24 months for Oconee Unit 3 during operating cycle 17.

1 Oconee 1,2, and 3 4.5-2 Amendment No.

(Unit 1)

Amendment No.

(Unit 2)

Amendment No.

(Unit 3)

TABLE 4.20-1 SSFINSTRUMENTATION SURVEILLANCE REQUIREMENTS Check Calibrate Remarks 1.

RCS Pressure (3)

WE 18 months (4)

Loop A, B l 2.

SSF RC Mckeup Pump (3)

Suction Pressure QU(1) 18 months Discharge Pressure QU(1) 18 months Suction Temperature QU(1) 18 months Discharge Flow QU(1) 18 months (4) l 3.

RC System Temperature (3)

NA(2) 18 months.

Loop A, B Hot, Cold

. 4.

Pressurizer Water Level (3)

WE 18 months (4) 5.

SSF Auxiliary Service

-- Water Pump Suction Pressure QU(l)

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~ Unit 3 Discharge Pressure NA AN Discharge Test Flow QU(1)

AN Suction Temperature QU(1)

AN 6.

Steam Generator Levels (3)

WE 18 months A,B 7.

Underground Fuel Oil Storage Tank NA AN Inventory l

8.

' D/G Service Water Pump Discharge Flow QU(1)

AN Discharge Pressure QU(1)

AN l

9.

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Pressum (1)

Check when pump operated / tested per IST.

l (2)

This instrumentation is normally aligned through a transfer / isolation device to each Unit Control Room and is thus checked in accordance with Specification 4.1, Table j

4.1-1, item 7. Every 18 months, the instmment string to the SSF Control Room will be checked and calibrated.

1 (3)

Units 1,2,3.

I (4).

A one-time extension of the test frequency to a maximum of 24 months is allowed for Oconee Unit 3 during operating cycle 17.

l Oconee 1,2, and 3 4.20-5 Amendment No.

(Unit 1)

Amendment No.

(Unit 2)

Amendment No.

(Unit 3)

ATTACHMENT 2

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(1)

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I Insert to Technical Specification items 10 and 11.

1 (1)

A one-time extension of the test frequency to a maximum 4

of 24 months is allowed for Oconee Unit 3 during

)

operating cycle 17.

I I

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415.1.1. CorePloodingSystem Every 18==aad><, a system test shall be conducted to demonstrate proper operation atthe system.

l a.

Dming pressurization of the Reactor coolant Sysican, verification shall be made that the check and isolation valves in the core flooding tank discharge lines operate properly.

b.

'Ibe test will be ra==hed

4 w y if contml board 3=dic=rian of core flood tank level verifies that allvalves have opened.

i 4.5.1.2 ComponentTests 4.5.1.2.1 Valves-PowerOperated Valves 2-17. -18, shall only be tested every cold shutdown unless previously tested dwing the a.

cunent quarter.

b.

Every 18 months, the following 21 system valves shall be cycled snannally to verify the unanual l

operability of these power operated valves:

(1)

LPIpump discharge (ES)2-17.-18 (2)

.LPIdischarge throttling 2-12,-14 (3) 21 discharge header crossover 2-9,-10 (4) 21 discharge to HPl/RBS U-15,-16.

4.5.1.2.2 CheckValves Periodic individual leakage testing

  • of valves CF 12, CF-14.11-47 and 2-48 shall be accomplished prior to power operation after every time the plant is placed in the cold shutdown condition for refueling, after each time the plant is placed in a cold shutdown condition for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if testing hasmotheen accomplished in the preceding 9 -h and prior to retuming the valve to service after mainta=aare, repair or Er'n..m wetk is performed. Whenever; Vdj of these valves cannot be hed, the integrity of the rer==ining valve in each high pressure line having a lesking valve shall be determined and recorded daily. In addition, the position of the other closed valve located in the high pressee piping shall be tocorded daily. For the allowable leakage rates and limiting conditions for operation, see Techaical Specification 3.1.6.10.

Basm

'Ihe E.T sy Core Cooling Systems are the principle reactor safety features in the event of loss of coolant accident. 'Ihe removal of heat from the core provided by these systems is as:irc4 to limit core damage.

The High Pressure injection System under normal operating conditions has one pump operating. 'the HPI system test required by Specification 4.5.1.1 1.*enfies that the HPI system responds as required to actuation of ES channels I and 2.

j (a)

To satisfy ALARA requirements, leakage may be measured indirectly (as from the perfonnance of pressure indicators) if accomplished in accordance with approved procedures and supported by computations showing piat the method is capable of demonstrating valve compliance with the leakage criterit.

g x-yGitT~ MMMPort 6M ATritoM b (AGG Oconee 1,2, and 3 D.5C Amendment No.228(Unit 1)

Amendment No.229(Unk 2) n _ m nu M&-

Insert for page 4.5-2 (b)

A one-time extension of the LPI pump discharge valves LP-17 and LP-18 manual cycle test frequency is allowed to a maximum of 24 months for Oconee Unit 3 during operating cycle 17.

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TABLE (20-1 SSFINSTRUMENTA'ITON SURVEILLANCE REQUIREMENTS

_Chai Calibrate Remarks 1..

RCSPressure(3)

. WE Imp A,B

[

2.

SSFRCMa%gPump(3)-

' Suction Pressure

" U(1) 18 maa %

~

Q Discharge Pressure QU(1) 18 months SuctionTemperature QU(1) 18 months DischargePiow QU(1) Mg 3.

RCSystemTemperature(3)

NA(2) 18 months loop A,B

[

Hot, Cold L

Pressurizer WaterInvel(3)

WB 18 months b l

S.

SSFAuxiliaryService WaterPump i

Suction Pressure QU(1)

AN Discharge Pressure QU(1)

AN Unit 1 Discharge Pressure NA AN Unit 2 Disdiarge Pressure NA AN Unit 3 Discharge Pressure NA AN DuichargeTest Flow QU(1)

AN SuctionTemperature QU(1)

AN 6.

SteamGeneratorlevels(3)

WE 18 months A,B l

7.

Undeyciux!FuelOilStorage NA AN TanhInventory

-- t 8.

D/G ServiceWaterPump DischargeFlow QU(1)

AN Discharge Pressure QU(1)

AN 9.

. D/G AirStart System WE AN Pressure (1)

Check when pump operated / tested per IST.

(2)

This instrumentation is nonnally aligned through a transfer / isolation device to each Unit Control Room and is thus checked in accordance with S edification 4.1, Table 4.1-1, Item 7. Every I8 months, the instrument string

[

l P

l to the SSF Control Room will be checked and calibrated.

(3)

Units 1, E 3.

~ = __ ~

R) fl onc. Trnc C *.Tcamou of TH E TrS T C CCGilEN C Y TO A t1AMr<uM of 2 4 m c.u-rus & Anow60 ret CcCNEf l)wT 3 bugou s, c WUT&&

cycut 17 Ocorue 1,2, and 3 4.20-5 Amendment No.228(Unit 1)

Amendment No. 22% Unit 2)

~

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' Technical Justificati'on

- Background

!This -license mnendment request became necessary due to two

~ forced: outages.and one prolonged shutdown of-Oconee Unit 3.

.These; delays moved the refueling' outage window for Oconee

- Unit'3 to the. period between late August 1998 and mid-October 1998.

Delaying the beginning of the outage until October 13',

1998,-involves a coast'down but avoids a-concurrent' refueling outage with Catawba Unit 2 allowing

- better utilization of' resources needed for both refueling outages.

The outage begins.when.the generator breaker is

- opened.

As a result of the' outage schedule change'and the lengthened duration of the operating cycle, the Technical Specification.

- surveillance ~ requirements were. reviewed to ensure compliance with the Technical' Specifications.

This review indicated that seven instrument channel calibrations and a manual stroke' test'of.two valves were required to be completed

- before.the' refueling-outage.

Technical-Specification Section 4.0-specifies an 18 month surveillance frequency as having a maximum interval of 22

- months land 15. days.

The maximum interval will be first

- exceeded on: August 31, 1998, for the SSF RCS Pressure

' Instrument Calibration.

The lengthened operating cycle will result in calibrations and tests described in the table below'being exceeded before the start of the refueling outage.-

Considering the previous instrument calibration date and the need to' operate until October 13,'1998, the maximum allowed interval of 22 months and 15 days would be exceeded by one month and:13 days.

Therefore, an extension is needed.

The following. table summarizes the Technical Specification surveillance for which an extension is being requested and the effective time being added by the extension to the

- beginning of_the outage.

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Technical Justification L

Description of Component and l

TS Section Surveillance Added Time Standby Shutdown Facility (SSF)

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4.20.2 (a),

Reactor Coolant System (RCS) l Table 4.20-1, Pressure Instrument Loops A and 1 Month, 12 l

item 1 B Calibration Days 4.20.2 (a),

Table 4.20-1, SSF RCS Pressure Instrument 1 Month, 13 item 4 Calibration Days 4.20.2 fa),

Table 4.20-1, SSF RCS Makeup Pump Flow 1 Month, 12 item 2 Instrument Calibration Days Reactor Protective System (RPS) 4.1.1, Table RCS Flow Instrument Calibration L

4.1-1, item 10 for Channels C and D 26 Days 4.1.1, Table RPS RCS Pressure Instrument 4.1-1, items Calibration for Channels C and 8,

9 and 11 D

23 Days Low Pressure Injection System l

4.5.1.2.1, Pump Discharge Valves LP-17 and i

item b(1)

LP-18 Manual Cycle 6 Days Test data from prior surveillance tests and calibrations i

were reviewed.

The review indicated that a one-time extension of the listed calibrations and tests would result l

in an extremely low probability of instrument drift beyond the allowed tolerance or affect the capability to manually open valves LP-17 and LP-18.

The bases for these l

conclusions are provided in the following technical i

justification.

The additional background information for this amendment request is broken down into three parts to separately address the SSF instrument calibrations, the RPS instrument calibrations, and the manual stroke test of the Low Pressure Injection (LPI) System Pump Discharge Isolation Valves.

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Standby Shutdovn Facility (SSF) Instrumentation The SSF provides the capability to shutdown the three Oconee nuclear reactors from outside the main control room in the event of a fire, flood, or sabotage related event.

The SSF provides additional " defense-in-depth" providing a backup to I

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Technical Justification safety-related systems.

The SSF is designed to maintain hot shutdown conditions on all three units for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following the above events.

It is designed to maintain Reactor Coolant System (RCS) inventory, maintain RCS

. pressure, remove decay heat, and maintain shutdown margin.

The SSF is provided with instrumentation to monitor several RCS parameters that include the SSF RCS Pressure and SSF Pressuri'er Lesel.

Technical Specification Table 4.20-1 specifies the frequency and type of surveillance for SSF instrumentation.

Technical Specification Table 4.20-1, item 1 (SSF RC Pressure), item ~2 (SSF RC Makeup Pump

.. Discharge Flow), and item 4 (SSF PZR Level) have a specified calibration frequency of 18 months.

The maximum allowable frequency of 18 months is defined in Technical Specification 4.0.2 as 22 months and 15 days.

Portions of the SSF RCS Pressurizer Level instrumentation and SSF RCS Pressure instrumentation surveillance must be performed in containment.

Due to the location of the SSF RCS Pressurizer Level instrumentation and the high dose rate that.would be received, it is not justifiable to perform this surveillance at power.

The efforts required to perform the SSF RCS Pressure instrumentation surveillance at power and the'dese that would be received do not justify the additional confidence that would be obtain by performing this surveillance within the current allowable interval of 22' months, 15 days.

Considering the previous SSF instrument calibration dates and the need to operate until October 13, 1998, the maximum allowed interval of 22 months and 15 days would be exceeded by.one month and 13 days for the earliest due date.

Therefore, an extension is needed.

Reactor Protective System Instrumentation The Reactor Protective System (RPS) monitors parameters related to safe operation and trips the reactor to protect the reactor core from fuel rod cladding damage.

It also assists in protecting against Reactor Coolant System (RCS) damage caused by high system pressure by limiting energy input to the system through reactor trip action.

The RPS is a 2-out-of-4 coincidence logic system.

The RPS inputs to the system are provided to monitor reactor flux,

'various RCS parameters, Peactor Building pressure, Main 3

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Technical Justification Turbine crip and loss of Main Feedwater.

This portion of the amendment request involves only the RCS pressure and flow instrumentation discussed below.

The RPS has "our independent input channels for monitoring RCS pressure and four independent input channels for monitoring RCS flow.

RCS flow is an input to the Flux / Imbalance / Flow trip and provides operators with flow indication for each RCS Loop.

RCS pressure is an input to the High RCS Pressure trip, Low RCS Pressure trip, Shutdown Bypass High Pressure trip, and the Variable Low RCS Pressure trip.

This portion of the amendment request involves only channels C and D of the RCS pressure and flow instrumentation discussed below.

The channel A and channel B calibrations for both the RCS pressure and flow instruments will remain within the allowable surveillance frequency.

Technical Specification Table 4.1-1 includes the frequency und type of surveillance required for RPS instrumentation.

Technical Specification Table 4.1-1, item 8 (High Reactor Coolant Pressure), itean 10 (Flux-Reactor Coolant Flow Comparator) and item 11 (Reactor Coolant Pressure Temperature Comparator).have a specified interval between tests of 18 months.

The maximum allowable interval between 18 month frequency surveillance is 22 months and 15 days.

It is undesirable to calibrate the RPS RCS Flow and Pressure instrument channels with the reactor at power due to the potential for a reactor trip during these calibrations.

Considering the previous RPS instrument calibration dates and the need to operate until October 13, 1998, the maximum allowed interval of 22 months and 15 days would be exceeded by 26 days for the earliest due date.

Therefore, an extension is needed.

Low Pressure Injection (LPI) System Pump Discharge Valves LP-17 and LP-18 The LPI-System provides post-accident core cooling injection for larger RCS break sizes.

The LPI System removes decay heat from the reactor core and sensible heat from the RCS l

during the later stages of normal plant cooldown.

Two redundant trains of LPI are provided with pumps and heat exchangers outside containment.

The containment penetration for each LPI flow path through the containment wall to the 4

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l Technical' Justification RCS is isolated by a Motor Operated Valve (MOV)' located outside containment and a check valve inside containment.

The. containment isolation MOVs (LP-17'and'LP-18) located outside containment and the check valves located inside

- containment, isolate the high pressure RCS from the lower design pressure LPI system-during normal operation.

During normal operation, LP-17 and LP-18 must be maintained operable and closed.

LP-17 and'LP-18 must be capable of being opened and closed by their motor operators post-accident. For certain scenarios they must be capable of being manually opened in the event of a failure.

LPI System Pump Discharge Valves LP-17 and LP-18 are boundary valves between high' pressure and low pressure design piping..As such, testing of valves LP-17 and LP-18 should only.be performed during Cold Shutdown conditions when the RCS pressure is below the design pressure of the LPI System piping.

Testing at RCS temperature and pressure greator than that specified fer Cold Shutdown would constitute a Technical Specification violation.

Technical Specification 4.5.1.2.1,. item b(1), specifies that the LPI System pump discharge valves LP-17 and LP-18 ba cycled manually every 18 months to verify the manual operability of these MOVs.

The maximum allowable interval between tests for an 18' month frequency is defined in Technical Specification 4.0.2 as 22 months and 15 days.

Considering the previous manual stroke test date for the valves and the need to operate until October 13, 1998, the maximum allowed interval of 22 months and 15 days would be exceeded by 6-days.

Therefore, an extension is needed.

Description of Technical Specification Change This Technical Specification amendment request involves one time changes tc the calibration and test frequencies of the SSF RCS Pressure Instrument Loops A and B, SSF RCS Pressurizer Level and Pressure Instruments, SSF RCS Makeup Pump Flow Instrument, RPS RCS Flow Instruments for Channels C and D, RPS RCS Pressure Instrument Calibration for Channels C and D, and LPI System Pump Discharge Valves LP-17 and LP-18 Manual Stroke Test.

Additional information about the individual changes is provided in the following

. paragraphs.

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Technical Justification To support the operation of.Oconee Unit 3 to the scheduled refueling outage, a one-time extension of the interval between surveillance to a maximum of 24 months is necessary for the:

1.

Instrument calibrations for Technical Specification Table 4.20-1, item numbers 1, 2 and 4, and for Technical Specification Table 4.1-1, items 1, 8,

9 and 11.

And, 2.

Manual stroke cest for Technical Specification Table 4.5.1.2.1, Item b(1).

Technical Justification The Technical Justification for this amendment request is broken into.three parts to separately address the channel calibration of the SSF instruments and RPS instruments, and the LPI Makeup Pump Discharge Valve manual cycle test.

SSF Instruments Procedure IP/0/A/0370/002A is used to calibrate and verify

proper operation of the Loops-A and B SSF Reactor Coolant Pressure Instrumentation. fA review of the previous two performances of procedute IP/0/A/0370/002A did not indicate any adverse trends and found that this pressure

' instrumentation remained within' allowable tolerances.

~ Procedure IP/0/A/0370/002C is uced tas calibrate and verify proper oparation of'the SSF Pressurizer Water Level

' Instrumentation.

A review of the previous two performances of procedure IP/0/A/0370/002C did not indicate any adverse trends and found that this level instrumentation remained within allowable tolerances.

Procedure IP/0/A/0370/001C is used to calibrate and verify proper operation of the SSF Reactor Coolant Makeup Pump Discharge Flow instrumentation.

A review of the previous two performances of procedure IP/0/A/0370/001C did not I

indicate any adverse trends and found that this flow

-instrumentation remained within allowabte tolerances.

The as found data for the above surveillance were well within the specified tolerance of the procedures.

There is no indication of significant calibration drift to suggest a degradation of this instrumentation due to a surveillance 6

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Technical Justification l

extension of up to 24 months.

A review of historic instrument drift for these instruments indicates that a 6 month extension beyond the normal 18 month surveillance

' interval should not result in drift beyond their allowable tolerances.

Thus, there is a high level of confidence that a one-time extension of the calibration frequency to a maximum of 24 months would result in an extremely low probability of instrument drift beyond the allowed tolerances.

RPS Instruments Procedure IP/3/A/0305/001L is used to calibrate and verify proper operation of the Channel C RPS RCS flow instrumentation, and procedure IP/3/A/0305/00lK is used to calibrate and verify proper operation of the Channel D RPS RCS flow instrumentation.

A review of the previous two performances of these two procedures did not indicate any adverse trends and found that these flow instruments remained within allowable tolerances.

Procedure IP/0/A/0305/00lO is used to calibrate and verify proper operation of the Channel C RPS RCS pressure instrumentation, and procedure IP/0/A/0305/00lP is used to calibrate and verify proper operation of the Channel D RPS RCS pressure instrumentation.

A review of the previous two performances of these two procedures did not indicate any adverse trends and found that these pressure instruments remained within allowable tolerances.

The RPS RCS instrumentation discussed above is highly reliable based on performance history.

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confidence in the continuing accuracy of the instrumentation is achieved by performance of an Instrument Channel Check each shift (i.e.,

twice per day).

The Instrument Channel Check verifies acceptable performance of each instrument channel by observation of.its behavior and/or state; this v% ification includes comparison of output and/or state of each of the four independent instrument channels for RCS flow and the four independent channels for RCS pressure.

There is no indication of significant calibration drift to suggest that a degradation of this instrumentation would occur due to a surveillance extension of up to 24 months.

A review of historic instrument drift'for these instruments indicates that a 6 month extension to the normal 18 month surveillance interval should not result in drift beyond 7

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Technical Justification their allowable tolerances.

Thus, there is a high level of confidence that a one-time extension of the calibration frequency to a maximum of 24 months should not result in any instrument drift beyond the allowed tolerance.

LPI Pump Discharge Valves Procedure PT/3/A/0152/12 was used for the previous performance, and will be used for the next valve manual cycle tests of LPI System Pump Discharge Valves LP-17 and l

LP-18.

Prior surveillance tests were performed using l

procedure PT/0/A/203/05.

Review of the test procedures f

documenting the last two performances of the manual cycle tests of valves LP-17 and LP-18' indicates that the valves were manually opened and no restriction or binding during manual operation was documented. Additionally, the last four years of maintenance history for valves LP-17 and LP-18 was reviewed.

This review found no record of valve problems that would suggest difficulties in manually opening these valves during the 6 day surveillance interval extension requested by this amendment.

The LPI System Pump Discharge Valves LP-17 and LP-18 have been demonstrated to be reliably capable of being manually opened.

A review of industry operating experience with-similar valves and actuators did_.not find any failures of a type which would prevent the valves from being manually cycled.

Based on the past performance of these valves and the short duration of the requested extension, it is reasonable to conclude that no adverse effects on the manual operability of.the valves should occur as the renult of this extension. Thus, a one-time extension of the manual cycle test frequency to a maximum of 24 months should not result in degradation of the capability to manually open LP-17 and LP-18.

Based on the information in this attachment and the Bases of the Technical Specifications, Duke Energy Corporation concludes that the proposed amendment will not present an undue risk to public health and safety.

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ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION EVALUATION This proposed change has been evaluated against the standards in 10 CFR 50.92 and has been determined to involve no significant hazards, in that operation of the facility in accordance with the proposed amendment would not:

a 1.

Involve a significant increase in the probability or i

consequences of an accident previously evaluated?

I No.

A review of the previous two instrument channel tests and calibrations, and two manual valve cycle tests discussed in this amendment request concluded that no adverse effects should occur as a result of the one-time extension.

There is a high level of confidence that the instruments and valves should be available to perform their intended function during the requested extension period.

.Thus, the probability and consequences of an accident previously evaluated will not be significantly increased.

2.

Create the possibility of a new or different kind of accident from the accidents previously evaluated?

No.

Since the one-time extension should not cause any

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adverse effects on Standby Shutdown Fac4.lity, Reactor Protective System or the Low Pressure ?njection system, a new or dif ferent kind of accident frua the accidents which were previously evaluated will not occur.

The Standby Shutdown Facility, Reactor Protective System or the Low Pressure Injection system should be available to perform their intended function during the requested extension period.

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ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONr

'dRATION EVALUATION 3.

Involve a significant reduction in a margin of safety?

No.

The margin of safety will not be significantly reduced by this amendment request because the Standby

' Shutdown Facility, Reactor Protective System or the Low Pressure Injection system should be available to perform their intended function during the requested extension period'. In addition, the review of the previous tests and calibrations which are discussed in the amendment request concluded that no adverse effects should occur as a result of the one-time extension.

Duke has concluded, based on the above information, that there are no significant hazards involved in this amendment request.

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ATTACHMENT 5 ENVIRONMENTAL IMPACT ANALYSIS Pursuant to 10 CFR 51.22 (b), an evaluation of the proposed amendment has been performed to determine whether or not it meets the criteria for categorical exclusion set forth in 10 CFR 51.22 (c) 9 of the regulations. The proposed amendment does not involve:

1)

A significant hazards consideration.

This conclusion is supported by the determination of no significant hazards.

2)

A significant change in the types or significant increase in the amounts of any effluents that may be released offsite.

This amendment will not change the types or amounts of any effluents that may be released offsite.

3)

A significant increase in the individual or cumulative occupational radiation exposure, This amendment will not increase the individual or cumulative occupational radiation exposure.

In summary, this amendment request meets the criteria set forth in 10 CFR 51.22 (c) 9 of the regulations for categorical exclusion from cn environmental impact statement.

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