ML090760866

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Perry Nuclear Power Plant, Application for Technical Specification Change Regarding Revision of Control Rod Application for Technical Specification Change Regarding Revision of Control
ML090760866
Person / Time
Site: Perry FirstEnergy icon.png
Issue date: 03/11/2009
From: Bezilla M B
FirstEnergy Nuclear Operating Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
L-09-031
Download: ML090760866 (30)
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Text

FENOC FirstEnergy Nuclear Operating Company Perry Nuclear Power Station 10 Center Road Perry, Ohio 44081 MarkB. Bezilla Vice President March 11,2009 L-09-031 440-280-5382 Fax: 440-280-8029 10 CFR 50.90 ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Perry Nuclear Power Plant Docket No. 50-440, License No.

NPF-58 Application for Technical Specification Change Regarding Revision of Control Rod Notch Surveillance Test Frequency and a Clarification of a Frequencv Example In accordance with the provisions of 10 CFR 50.90, FirstEnergy Nuclear Operating Company (FENOC) is submitting a request for an amendment to the technical specifications (TS) for the Perry Nuclear Power Plant (PNPP). The proposed amendment would: (1) revise TS surveillance requirement (SR) frequency in TS 3.1.3, "Control Rod OPERABILITY," and (2) revise Example 1.4-3 in Section 1.4 "Frequency" to clarify the applicability of the 1.25 surveillance test interval extension. The enclosure provides the evaluation for the proposed amendment.

Approval of the license amendment is requested prior to August 29, 2009, with the amendment to be implemented within 90 days following its effective date.

Regulatory commitments associated with this submittal are included in the attachment.

If there are any questions or if additional information is required, please contact Mr. Thomas A. Lentz, Manager - Fleet Licensing, at (330) 761-6071.

I declare under penalty of perjury that the foregoing is true and correct. Executed on March L\, 2009. Mark B. Bezillal/

March 11, 2009 L-09-031 10 CFR 50.90

ATTN: Document Control Desk

U. S. Nuclear Regulatory Commission

Washington, DC 20555-0001

SUBJECT:

Perry Nuclear Power Plant

Docket No. 50-440, License No. NPF-58

Application for Technical Specificati on Change Regarding Revision of Control Rod Notch Surveillance Test Frequency and a Clarification of a Frequency Example

In accordance with the provisions of 10 CFR 50.90, FirstEnergy Nuclear Operating Company (FENOC) is submitting a request for an amendment to the technical

specifications (TS) for the Perry Nuclear Power Plant (PNPP).

The proposed amendment would: (1) revise TS surveillance requirement (SR) frequency in TS 3.1.3, "Control Rod OPERABILITY," and (2) revise Example 1.4-3 in Section 1.4 "Frequency" to clarify the applicability of the 1.

25 surveillance test interval extension.

The enclosure provides the evaluation for the proposed amendment.

Approval of the license amendment is r equested prior to August 29, 2009, with the amendment to be implemented within 90 da ys following its effective date.

Regulatory commitments associated with this subm ittal are included in the attachment. If there are any questions or if additional information is required, please contact Mr. Thomas A. Lentz, Manager - Fl eet Licensing, at (330) 761-6071.

I declare under penalty of perjury that the foregoing is true and correct. Executed on March ___, 2009.

Sincerely,

Mark B. Bezilla

Perry Nuclear Power Plant L-09-031 Page 2 of 2

Attachment:

Regulatory Commitment List

Enclosure:

Application for Technical Specificati on Change Regarding Revision of Control Rod Notch Surveillance Test Frequency and a Clarification of a Frequency Example

cc: NRC Region III Administrator NRC Resident Inspector NRR Project Manager

Executive Director, Ohio Emergency Management Agency, State of Ohio (NRC Liaison)

Utility Radiological Safety Board Attachment L-09-031 Regulatory Commitment List Page 1 of 1 The following list identifies those actions commi tted to by FirstEnergy Nuclear Operating Company (FENOC) for Perry Nuclear Power Pl ant in this document. Any other actions discussed in the submittal represent intended or planned actions by FENOC. They are described only as information and are not R egulatory Commitments. Please notify Mr.

Thomas A. Lentz, Manager - Fleet Licensi ng, at (330) 761-6071 of any questions regarding this document or asso ciated Regulatory Commitments.

Regulatory Commitment Due Date

1. FENOC commits to revising Technical Specification Bases based on TSTF-475, Revision 1 as proposed in Attachment 2 to the

Enclosure.

Concurrent with amendment implementation.

Application for Technical Specification Change Regarding Revision of Control Rod Notch Surveillance Test Frequency and a Clarification of a Frequency Example

1. DESCRIPTION
2. ASSESSMENT

2.1 APPLICABILITY

OF PUBLISHED SAFETY EVALUATION

2.2 OPTIONAL

CHANGES AND VARIATIONS

3. REGULATORY ANALYSIS

3.1 NO SIGNIFICANT HAZARDS DETERMINATION

3.2 VERIFICATION

AND COMMITMENTS

4. ENVIRONMENTAL EVALUATION

Attachments:

1. Proposed Technical Specification Changes (Mark-Up)
2. Proposed Changes to Technical Specifications Bases
3. Proposed Technical Specification Changes (Retyped)

Page 2 of 3

1.0 DESCRIPTION

The proposed amendment would: (1) revise the Technical Specification (TS)

Surveillance Requirement (SR) 3.1.3.

2 frequency in TS 3.1.3, "Control Rod OPERABILITY," and (2) revise Example 1.4-3 in Section 1.4, "Frequency," to clarify the applicability of the 1.25 surveillanc e test interval extension.

The changes are consistent with Nuclear Regulatory Commission (NRC) approved Industry/Technical Specification Task Forc e (TSTF) Standard Technical Specification (STS) change TSTF-475, Revision 1. The Federal Register notice published on November 13, 2007 announced the availability of this TS improvement through the consolidated line item impr ovement process (CLIIP).

2.0 ASSESSMENT

2.1 Applicability

of Published Safety Evaluation FirstEnergy Nuclear Operating Company (FEN OC) has reviewed the safety evaluation dated November 5, 2007 as part of the CLIIP.

This review included a review of the NRC staff's evaluation, as well as the background information provided to support

TSTF-475, Revision 1. FENOC has concluded t hat the justifications presented in the TSTF proposal and the safety evaluation pr epared by the NRC staff are applicable to Perry Nuclear Power Plant (PNPP), Unit 1 and justify this amendment for the incorporation of the changes to the PNPP TS.

2.2 Optional

Changes and Variations TSTF-475, Revision 1 proposes three changes to the STS. The proposed change: (1) revises the TS control rod notch surveillance frequency in TS 3.1.3, (2) clarifies the TS 3.3.1.2 requirement for fully inserting contro l rods for one or more inoperable SRMs in Mode 5, and (3) revises one example in Se ction 1.4, "Frequency," to clarify the applicability of the 1.25 surveillance test interval extension.

Only two of the changes are proposed for th is amendment application. The change to TS 3.3.1.2, which is not included in this amendment application, would have incorporated the word "fully" into Requi red Action E.2, "Source Range Monitoring Instrumentation," as follows:

Initiate action to fully insert all insertable control rods in core cells containing one or more fuel assemblies.

The word "fully" is already in PNPP TS 3.

3.1.2, Required Action E.2, "Source Range Monitoring Instrumentation." Therefore, no change is necessary to incorporate this aspect of TSTF-475 into the PNPP TS.

Page 3 of 3

FENOC is not proposing any other variati ons or deviations from the TS changes described in the modified TSTF-475, Revision 1 and the NRC staff's model safety

evaluation dated November 5, 2007.

FENOC is proposing a variation relative to the proposed TS Bases changes contained

in TSTF-475, Revision 1. The TSTF-475 pr oposed TS Bases discussion for SR 3.1.3.3 (previously identified as SR 3.1.3.4) woul d remove SR 3.1.4.4 from the listed SRs performed in conjunction with SR 3.1.3.3.

Incorporation of this change would be inconsistent with TS SR 3.1.3.3, which in cludes SR 3.1.4.4 in the Frequency column.

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration Determination FENOC has reviewed the proposed no signific ant hazards consideration determination (NSHCD) published in the Federal Register as part of the CLIIP. FENOC has concluded that the propos ed NSHCD presented in the Federal Register notice is applicable to PNPP without need for modification despite the deviation described in

Section 2.2 of this application and is hereby incorporated by reference to satisfy the requirements of 10 CFR 50.91(a).

3.2 Verification

and Commitments As discussed in the notice of availability published in the Federal Register on November 13, 2007 for this TS improvem ent, FENOC verifies the applicability of TSTF-475 to PNPP, and commits to revising Technical Specification Bases based on

TSTF-475, Revision 1, as proposed in Attachment 2.

These changes are based on TSTF change trav eler TSTF-475, Revision 1, which proposes revisions to the STS by: (1) revising the frequency of SR 3.1.3.2, notch testing of fully withdrawn control rod, from "7 days after the control rod is withdrawn and THERMAL POWER is greater than the [Low Power Setpoint] LPSP of [Rod Pattern Control System] RPCS" to "31 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP of the RPCS," and (2) revising Example 1.4-3 in Section 1.4, "Frequency," to clar ify that the 1.25 surveillance test interval extension in SR 3.0.2 is applicable to time periods discussed in NOTES in the "SURVEILLANCE"

column in addition to the time per iods in the "FREQUENCY" column.

4.0 ENVIRONMENTAL

EVALUATION FENOC has reviewed the environmental eval uation included in the model safety evaluation dated November 5, 2007 as part of the CLIIP. FENOC has concluded that the staff's findings presented in that eval uation are applicable to PNPP without need for modification despite the deviation described in Section 2.2 of this application, and the evaluation is hereby incorporated by reference for this application.

Page 1 of 9

PROPOSED TECHNICAL SPECIFICATION CHANGES (MARK-UP)

Frequency 1.4 1.4 Frequency EXAMPLES EXAM PLE 1.4 -2 (cont i nued) "Thereafter" lndicates future performances must be established per SR 3.0.2, but only after a specified condition is ffrst met (Le., the "once" performance in this example).

measurement of both intervals stops. New intervals start upon reactor power reaching 25% RTP. If reactor power decreases to < 25% RTP, the EXAMPLE 1, 4-3 The interval continues whether or not the unit operation is < 25% RTP between perfownces.

As the Note modifies the required performance of the Surveillance, it is construed to be part of the "specified Frequency." Should the 7 day interval be exceeded while operation is < 25% RTP, this Note allows 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after power reaches 2 25% RTP to perform the Surveillance.

The Surveillance is still consldered to be within the "specffied Frequency.

Therefore, if the Surveillance were not performed within the 7 day (plus the extension allowed by SR 3.0.2) Interval, but operatton was < 25% RTP, 3t would not constjtute a failure of the SR or fai'lure to meet the LCO. Also, no violation of SR 3,0.4 occurs when changing MDU, even with the 7 day Frequency not met, provided operation does not exceed 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ith power 2 25% RTP. tCORt5W$dl I PERRY - UNIT 1 1.0-27 Amendment No, 69 Frequency 1.4 1.4 Frequency EXAMPLES EXAMPLE 1.4-3 (continued)

Once the unit reaches 25% RTP, 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> would be allowed for ompleting the Survei 1 lance. If the Survei 1 lance were not hour i@erVtiJp there would then be erformed within thj tal lure to perrorinsalhver I lance wqthin the specified Frequency, and the provisions of SI? 3.0.3 would apply. EXAMPLE 1.4-4 SURVEILLANCE REQUIREMENTS SURVEILLANCE Verify leakage rates are within limits. FREQUENCY 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Example 1.4-4 specifies that the requirements of this Surveillance do not have to be met until the unit is in MODE 1. The interval measurement for the Frequency of this Surveillance continues at all times, as described in Example 1.4-1. However, the Note constitutes an "otherwise stated" exception to the Applicability of this Surveillance.

Therefore.

if the Surveillance were not erformed within the but the unit was not in MODE 1, there would be no failure of the SR nor failure to meet the LCO. Therefore, no violation of SR 3.0.4 occurs when changing MODES, even with the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Fre uency exceeded, provided the MODE change was not that the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency were not met), SR 3.0.4 would re uire satisfying the SR, except as provided by SR 3.0.3 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (plus the extension allowed by S R 3.0.2) interval, made into d DE 1. Prior to entering MOD* 1 (assuming again an 1 LCO 3.0.4. 1.0-28 Amendment No. 131 Control Rod OPERABILITY

'112 3.1 REACTIVITY CONTROL SYSTEMS 3.1.3 Control Rod OPERABILITY LCO 3.1.3 Each control rod shall be OPERABLE.

APPLICABILITY:

MODES 1 and 2. ACTIONS --------------------___c_____________

NOTE-------------------------------------

Separate Condition entry is allowed for each control rod. CONDITION A. One withdrawn control rod stuck. PERRY - UNIT 1 REQUIRE0 ACTION ------------

NOTE-------------

A stuck rod may be bypassed in the Rod Action Control System (RACS) in accordance with SR 3.3.2.1.9 if required to allow continued operation.

A. 1 Verify stuck control rod separation criteria are met. A.2 Disarm the associated control rod drive (CRD) . - AND COMPLETION TIME Inpnedi ately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (continued) 3.1-7 Amendment No. 120 Control Rod OPERABILITY

3.1.3 ACTIONS

CONDITION A. (continued)

B. Two or more withdrawn control rods stuck. C. One or more control rods inoperable for reasons other than Condition A or B. REQUIRED ACTION A.3 3.2 for each withdrawn OPERABLE control rod. - AND A.4 Perform SR 3.1.1.1. B.l Be in MODE 3. c.1 --_----- NOTE---------

Inoperable control rods may be bypassed in RACS in accordance with SR 3.3.2.1.9.

if required.

to a1 low insertion of i nopera bl e control rod and continued operation, ---------------------

Fully insert inoperable control rod. COMPLETION TIME 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from discovery of Cond-ition A concurrent wi t h THERM1 POWER greater than or equal to the 1 ow power setpoint (LPSP) of the Rod Pattern Control System (RPCS). 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 12 hours 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> 4 hours C.2 Disarm the associated CRD . (continued)

PERRY - UNIT 1 3.1-8 Amendment No. 1 20 Control Rod OPERABILITY 3.1.3 Two or more inoperable control rods not in compl i ance with banked pos i ti on wi t hdraw 1% sequence (BPWSJ and not separated by twu or more OPERABtE control rods. E. Required Action and a ssoci at ed Compl et i un Time of Condition A. C, or D not met. Nine or more cantm3 rods i noperabl e,. PERRY - UNIT 1 D. 1 Restore cornpli&nc-dth BPMS . - OR D.2 Restore control rod to OPERABLE status. E.1 Be in MODE 3. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 4 hours 3.1-9 Amendment No. 112 Control Koa UYLKH~Z~LIII

3.1.3 SURYEICLANCE

REQUIREM,ENTS SURVE I lLAElCE Insert each wi thdrawn control rod at least one no c SR 3.1,3@ Verify each control rad scram time frum fully wjthdrann to notch positton 13 is d L 7 seconds. PERRY - UNIT 1 3.1-10 FREQUENCY 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 31 days In accordance w# th SR 3.1.4*1, SR 3.1-4,2, SR J,Z,4.3, and SR 3.1.4.4 (continued)

Amendment No. 69 PERRY - 3.1-11 Amendment No, 69 Table 3.1.4-1 Control Rod Scram Times (a) Maximum scram time from fully withdrawn position, based on de-energization of scram pilot valve solenoids as time zero- (b) Scram times as a functlon of reactor steam dome pressure when < 950 PSig are within established limits. (c) For intermediate reactor steam dome pressures, the scram time criteria are deterrained by 1 inear interpol ation. PERRY - UNIT 1 3.1-14 Amendment No. 69 Page 1 of 7

PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS BASES (PROVIDED FOR INFORMATION)

Control Rod OPERABILITY B 3.1.3 BASES satisf the intended reactivity control re uirements, strict control rods is required to satisfy the assumptions of the DBA and transient analyses. (cont i nued) contro r over the number and distribution o 8 inoperable LCO APPCICABI L IN In MODES 1 and 2. the control rods are assumed to function durin a DBA or transient and are therefore required to be OPEdLE in these MODES. In MODES 3 and 4, control rods are not able to be withdrawn since the reactor mode switch is in Shutdown and a control rod block is ap lied. This provides these conditions.

Control rod re uirements in PKIDE 5 are located in LCO 3.9.5. "Control Ro 1 OPERABl[LITY-Refueling.

adequate requirements for control rod E PERABILITY during ACTIONS The ACTIONS table is modified by a Note indicating that a separate Condition entry is allowed for each control rod. This is acceptable, since the Required Actions for each Condition provide appropriate compensator actions for each may a1 low for continued operation, and subsequent inoperable control rods are governed by subsequent Condition entry and application of associated Required Actions. inoperable control rod. Complying with t z e Required Actions A.1, A.2. A.3, and A.4 A control rod is considered stuck if it will not insert (using a71 available insertion methods) by either CRD drive water or scram pressure.

With a fully inserted control rod stuck. no actions are required as long as the control-rod remains fully inserted.

The Required Actions are modified b a Note that allows a stuck control rod to be bypassed in t K e Rod Action Control System (RACS) to allow cont-inued o erati on. SR 3 3.2.1.9- provides addi ti onal requi rements den control rods are bypassed in RACS to ensure compliance with the CRDA analysis.

With one withdrawn control rod stuck. the local scram reactivity rate assumptions may not be met if the stuck control rod separation criteria are not met. Therefore. verification that the separation criteria (continued)

Control Rod OPERABILITY 8 3.1.3 BASES ACTIONS A.l. A.2, A.3. and A.4 (continued) are met must be performed immediately.

The stuck control rod separation criteria are that the stuck control rod ma not occupy a location adjacent to a 'slow" control rod. {he description of "slow" control rods is provided in LCO 3.1.4 "Control Rod Scram Times". In addition, the control rod must be disarmed within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The allowed Completion Time of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is acceptable, considerin the reactor can to insert. and provides a reasonable amount of time to perform the Re uired Action in an orderly manner. Isolating control rod can be still be shut dorm. assuming no additiona 3 control rods fail the control r 4 from scram prevents damage to the CRDM. The (cont i nued 1 PERRY - UNIT 1 B 3.1-15a Revision No. 4 Control Rod OPERABILITY B 3.1.3 BASES ACTIONS A.1. A.2. A.3. and A.4 (continued) isolated from scram by isolating the hydraulic control unit from scram and normal drive and withdraw pressure.

yet still maintain cooling water to the CRD. A control rod can be hydraulically disarmed by closing the drive water and exhaust water isolation valves. Electrically.

the control rod can be disarmed by dlsconnecting power from all four directional control valve solenoids.

Monitoring of the insertion capability for each withdrawn_

control rd must alsoHrformed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. A7 3" erforflperiodic tests of the control rod( insert% caDa E ility of withdrawn control rods. SR 3.1.3.2 Testing each withdrawn control rod ensures that a. generic problem does not exist. The allowed Completion Time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> provides a reasonable time to test the control rods, considering the potential fur a need to reduce power to perform the tests. Required Action A.2 has a modifjed time zero Completion Time. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Completion Time for this Required Action starts when the withdrawn control rod is discovered to be stuck and THERMAL POWER is greater than the actual low Ower set int (LPSP) of the rod pattern control fer (RPC P , since !? he notch insertions may not be co atible with the re uirements of rod pattern control (LE 3.1.6) and the RP 2 (LCO 3.3.2.1, "Control Rod Block Instrumentation"

1. To allow continued operation with a withdrawn control rod stuck. an evaluation of adequate SDM is also required within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Should a DBA or transient require a shutdown.

to preserve the single failure criterion an additional control rod would have to be assumed to have failed to insert when required.

Therefore.

the ori inal SM demonstration may not be valid. The SOM must there s ore be evaluated (by measurement or analysis) with the stuck control rod at its stuck position and the highest worth OPERABLE control rod assumed to be fully withdrawn.

The allowed Completion Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to verify SDM is adequate, considering that with a single control rod stuck In a withdrawn position, the remaining OPERABLE control rods are capable of providing the re uired scram and shutdown . Failure to reach DE 4 is on1 likely if an 51 d additiona reactivitl control rod adjacent to the stuc control rod (continuedl PERRY - UNIT 1 B 3.1-16 Revision No. 4 Control Rod OPERABILITY B 3.1.3 BASES ACTIONS A.1. A.2. A.3. and A.4 (continued) also fails to insert durin a required scram. Even with the rod to insert, sufficient reactivity control remains to reach and maintain MODE 3 conditions (Ref. 7). postulated additional sing 9 e failure of an adjacent control With two or more withdrawn control rods stuck, the plant should be brought to WDE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Isolating the control rod from scram prevents damage to the CRDM. The occurrence of more than one control rod stuck at a withdrawn sition increases the probabi 1 ity that the reactor cannot control rods eliminates the possibilit .of an additional failure of a control rod to insert. Tie allowed Completion Time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems. C.l and C.2 With one or more control rods inoperable for reasons other than being stuck in the withdrawn position, operation may continue, provided the control rods are fully inserted within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and disarmed (electrically or hydraulically) shutdown and scram Inserting capabi ities are not adversely affected.

wi thi n 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The control rod is disarmed to prevent inadvertent withdrawal during subsequent o erations.

The control rods can be hydraulically disarmed E y closing the drive water and exhaust water isolation valves. Electrically, the control rods can be disarmed b disconnecting ower from all four directional control va 7 ve solenoids.

R ith a control rod not coupled to its associated drive mechanism, insert the control rod drive mechanism to accomplish recoup1 ing. Verify recouplin by withdrawing the control rod and instrumentation and demonstrating that the control rod drive will not go to the overtravel position.

Required Action C.1 is modified b a Note that allows control rods to be ino erable control rods an continued operation.

SR g.3.2.1.9 provides additional requirements when the control rods are bypassed to ensure compliance with the CRDA analysis.

shut down if required.

Insertion of all insertable a control rod ensures the observing any in i icated response of the nuclear bypassed in t z e RACS if re uired to allow insertion of the 1 (continuedl Control Rod OPERABILITY B 3.1.3 SASES (continued)

SURVEILLANCE REQUIREMENTS SR 3.1.3.1 The position of each control rod must be determined, to ensure adequate information on control rod position is available to the operator for determining control rod OPERABILITY andecontrol 1 ing rod patterns.

Control rod osition may be determined by the use of at least one gPERA6i-E position indicator, by moving control rods to a position with an OPERABLE indicator, or by the use of other appropriate methods. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency of this SR is based on operating experience related to expected changes in control rod position and the availability of control rod position indications in the control rm. SR 3.1.3.21ki;6&$R fll.Y&J Control rod insertion capabilit is demonstrated b least one notch and observing that the control rod moves. The control rod may then be returned to its original position.

Observation of changes in indicated control rod position provides evidence that the control rod osition stuck and is free to insert on a scram signal. When plant rocedures permit, these SRs may also be met by rod scram. hese Surveillances are not required when THEWL POWER is less than or equal to the actual LPSP of the RPC since the notch insertions may not be cmatible with the reauirements inserting each partially or ful T y withdrawn contro T rod at indication is OPERABLE.

This ensures the contro P rod is not ew@iede relawd to thdchamedin CRD drfordce an&he 'awn control rods are t i? Frequenck based on the Dotential -mer rxxhctim ntwco~dd;imfJ Furthermore.

the atins exDerience related to'changes in CRD performance'.

At any the, if a control rod is imnovable.

a determination of that control rod's tri pability (OPERABILITY) must be made and SR 3.1.3.w appropria e e action taken. Verifying the scram time for each control rod to notch position 13 is s 7 seconds provides reasonable assurance that the control rod will insert when required during a DBA or transient, thereby completing its shutdown function. (continued)

~ZnformatM?

+! j PERRY - UNIT 1 B 3.1-19 Revision No. 3 BASES (continued) in conjunction with the control rod SR 3.1.4.1. SR 3.1.4.2, SR 3.1.4.3, SYSTEM FUNCTIONAL TEST in SURVEILLANCE REQUIREMENTS LCO 3.3.1.1. "Reactor Protection System (Rps) InstrumentatJon." and the functional testing of SDV vent and drain valves in LCO 3.1.8. "Scram Discharge Volume WIV) Vent and Drain Valves." overlap this Surveillance to provide complete testing of the assunred safety function.

The associated Frequencies are acceptable, considering the more frequent testing erformed to demonstrate other aspects of control rod OPE RA[1 ILIiv and operati ty eyience. which shows scram times do not slgnlficant y c ange over an operating cycle. Coupling verification is performed to ensure the control rod is connected to the CRDM and will perform its internled function when necessary.

The Survei 11 ance requi res verifying that a control rod does not go to the withdrawn overtravel positi it is fully withdrawn.

The ture rovides a positlve check on the coupling int pity, since on P y an uncoupled CRD can reach overtravel posf ti the overtrave 7 position.

The verification is required to be performed anytime a control rod is withdrawn to the "full out" position (notch,posltion

48) or prior to declaring the control rod OPERABtEafter work on the control rod or CRD System that could affect coupling.

This includes control rods inserted one notch and then returned to the "full out" position durtng the performance of SR 3.1.3.2. Until the control rod reaches the "full out" position where coupling can be verified, the nuclear Instrumentation is observed for any indicated response duri withdrawal.

This Frequency 1s acceptable, consideri the 7 ow probability that a control rod will become uncoup 7 ed when it is not king moved and operati ng experi me re1 ated to uncoupl i ng events . PERRY - UNIT 1 8 3.1-20 Revision No. 1 Page 1 of 7

PROPOSED TECHNICAL SPECIFICATION CHANGES (RETYPED)

Frequency 1.4 1.4 Frequency EXAMPLES EXAMPLE 1.4-2 (continued) "Thereafter" indicates future performances must be established per SR 3.0.2, but only after a specified condition is first met (i.e., the "once" performance in this example).

measurement of both intervals sto

s. New intervals start If reactor power decreases to < 25% RTP, the upon reactor power reaching 25% R 4! P. EXAMPLE 1.4-3 SURVEILLANCE REOUIREMENTS PERRY - UNIT 1 SURV E I L LANCE Perform channel adjustment.

FREQUENCY 7 days The interval continues whether or not the unit operation is 25% RTP between performances.

As the Note modifies the required erformance of the Frequency." Should the 7 day interval be exceeded while operation is c 25% RTP, this Note allows 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after ower reaches 2 25% RTP to perform the Surveillance.

The ! urveillance is still considered to be within the "specified Fre uency . 'I Therefore, if the Survei 1 7 ance were not SR 3.0.2) interval, but operation was < 25% RTP, it would not constitute a failure of the SR or failure to meet the LCO. Also, no violation of SR 3.0.4 occurs when changing MODES, even with the 7 day Frequency not met, provided operation does not exceed 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (plus the extension allowed by SR 3.0.2) with power 2 25% RTP. Surveillance, it is construed to be + part o the "specified per 9 ormed within the 7 day (plus the extension allowed by (continued) 1.0-27 Amendment No. I Frequency 1.4 Only required to be met in MODE 1. Verify leakage rates are within limits. .......................................

1.4 Frequency

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> EXAMPLES EXAMPLE 1.4-3 (continued)

Once the unit reaches 25% RTP, 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> would be allowed for completing the Surveillance.

If the Surveillance were not performed within this 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> interval (plus the extension allowed by SR 3.0.2). there would then be a failure to perform a Survei 7 1 ance within the speci f i ed Frequency, and the provisions of SR 3.0.3 would apply. EXAMPLE 1.4-4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Example 1.4-4 specifies that the requirements of this Surveillance do not have to be met until the unit is in MODE 1. The interval measurement for the Frequency of this Surveillance continues at all times, as described in Example 1.4-1. However, the Note constitutes an "otherwise stated" exception to the Applicability of this Surveillance.

Therefore, if the Surveillance were not erformed within the but the unit was not in MODE 1. there would be no failure of the SR nor failure to meet the LCO. Therefore, no violation of SR 3.0.4 occurs when changing MODES, even with the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Fre uency exceeded, provided the MODE change was not made into M 8 DE 1. Prior to entering MODE 1 (assuming again that the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency were not met), SR 3.0.4 would 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (plus the extension allowed by S g 3.0.2) interval, re uire satisfying the SR-, except as provided by SR 3.0.3 an ! LCO 3.0.4. PERRY - UNIT 1 1.0-28 Amendment No.

ACTIONS CONDITION A. (continued)

B. Two or more withdrawn control rods stuck. C. One or more control rods inoperable for reasons other than Condition A or B. REQUIRED ACTION ~~~ A.3 Perform SR 3.1.3.2 for each withdrawn OPERABLE control rod. - AND A.4 Perform SR 3.1.1.1. B.1 Be in MODE 3. c.1 -------- NOTE---------

Inoperable control rods may be bypassed in RACS in accordance with SR 3.3.2.1.9, if required, to allow insertion of inoperable control rod and continued opera t i on. ..................... Fully insert i noperabl e control rod. -- COMPLETION TIME 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from Condition A concurrent with THERMAL POWER greater than or equal to the low power setpoint (LPSP) of the Rod Pattern Control System (RPCS). discovery of t 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> , 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> , 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> 4 hours c.2 Disarm the associated CRD . (cont i nued 1 PERRY - UNIT 1 3.1-8 Amendment No. I Control Rod OPERABILITY 3.1.3 Not required to be performed until 31 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP of the RPCS. Insert each withdrawn control rod at least 31 days one notch. SR 3.1.3.3 Verify each control rod scram time from fully withdrawn to notch position 13 is I 7 seconds. In accordance with SR 3.1.4.1. SR 3.1.4.2, SR 3.1.4.3, and SR 3.1.4.4 PERRY - UNIT 1 3.1-10 (cont i nued 1 Amendment No.

Control Rod OPEM8ILITY 117 3.1.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE SR 3.1.3.4 Verify each control rod does not go to the withdrawn overtravel position.

FREQUENCY Each time the control rod is withdrawn to "full out" position Prior to declaring control rod OPERABLE after work on control rod or CRD System that could affect coup1 i ng PERRY - UNIT 1 3.1-11 Amendment No.

Control Rod Scram Times 3.1.4 Table 3.1.4-1 Control Rod Scram Times ------------------__-----------------

NOTES------------------------------------

1. OPERABLE control rods with scram times not within the limits of this Table are considered "slow." 2. Enter a plicable Conditions and Required Actions of LCO 3.1.3, "Control position 13, These control rods are inoperable, in accordance with SR 3.1.3.3, and are not considered "slow." Rod OPE i ABILITY," for control rods with scram times > 7 seconds to notch --------------------_______c____________--------------------------------------

NOTCH POSITION SCRAM TIMES(a) (b) (seconds 1 REACTOR REACTOR 950 psig 1050 psig STEAM DOME PRESSURE(c)

STEAM DOME PRESSURE (C) (a> Maximum scram time from fully withdrawn position, based on de-energization of scram pilot valve solenoids as time zero. (b) Scram times as a function of reactor steam dome pressure when < 950 psig are within established limits. (c) For intermediate reactor steam dome pressures, the scram time criteria are determined by linear interpolation.

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