ML20211G671

Proposed Tech Specs 3.1.4 Re Rod Group Alignment Limits

ML20211G671
Person / Time
Site:	McGuire
Issue date:	08/27/1999
From:	DUKE POWER CO.
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ML20211G670	List: ML20211G663 ML20211G671
References
NUDOCS 9908310329
Download: ML20211G671 (16)

Text

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ATTACHMENT 1 PROPOSED REVISIONS TO THE MCGUIRE NUCLEAR STATION TECHNICAL SPECIFICATIONS l

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Rod Group Alignment Limits 3.1.4 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY i

SR 3.1.4.2 Verify rod freedom of movement (trippability) by moving 9 da s l each rod not fully inserted in the core 2 10 steps in either e n.  !

direction. X p 44 M.GbE 3 upw I uw A i sMup hhtw3

@o oE Cyde G he um4 l e-e hs e SR 3.1.4.3 Verify rod drop time of each rod, from the fully withdrawn Prior to reactor l position, is s 2.2 seconds from the beginning of decay of criticality after I stationary gripper coil voltage to dashpot entry, with: each removal of the reactor head

a. Tog 2 551*F; and )

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b. All reactor coolant pumps operating.  !

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McGuire Units 1 and 2 3.1.4-4 Amendment Nos. 184/166

c ATTACHMENT 2 REVISED McGUIRE NUCLEAR STATION TECHNICAL SPECIFICATIONS )

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, l l Rod Group Alignment Limits l l 3.1.4 I SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.1.4.2 Verify rod freedom of movement (trippability) by moving 92 days

each rod not fully inserted in the core 2 10 steps in either direction. OR '

Prior to entering MODE 3 upon '

Unit 1 startup p following the i Unit 1 End of i Cycle 13 refueling outage, SR 3.1.4.3 Verify rod drop time of each rod, from the fully withdrawn Prior to reactor I position, is s 2.2 seconds from the beginning of decay of criticality after stationary gripper coil voltage to dashpot entry, with: each removal of the reactor head ,

a. Ty2 551*F; and I

b. All reactor coolant pumps operating.

l One time change applicable to Unit 1 only 4

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l McGuire Units 1 and 2 3.1.4 4 Amendment Nos. I I

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' ATTACHMENT 3 DESCRIPTION OF PROPOSED l CHANGES AND TECHNICAL JUSTIFICATION l

e a Attachment 3 Page 1 of 5 Proposed Changes DEC proposes to revise the McGuire Nuclear Station TS's as described below. The changes proposed by DEC in this LAR will provide for a one time extension of the surveillance frequency )

for TSSR 3.1.4.2 beyond the 25% extension allowed by TSSR 3.0.2.

The proposed changes are as follows:

1. The surveillance frequency for TSSR 3.1.4.2 has been revised to provide an option of performing the TSSR prior to entering MODE 3 upon Unit 1 startup following the lEOC13 Refueling Outage.

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2. A note has been added to the bottom of page 3.1.4-4 of the 1 McGuire TS's indicating that the surveillance frequency option described in #1 above is a one time change applicable to Unit 1 only.

Basis.for Proposed Changes i

Background:

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The Rod Control System controls the motion of the control rods (

within the core to modulate reactivity for power and temperature control. The safety function of the reactor control rods is to I fall into the core for safe shutdown of the reactor on receipt of ] '

a reactor trip signal. The major components necessary to convert Rod Control System input signals to actual rod motion are: the Logic Cabinet, the Power Cabinet, and the Control Rod Drive Mechanism Assemblies (CRDM). The Logic Cabinet generates-signals for speed and direction based on input information from the Reactor Control System or the Control Board operator. The Logic Cabinet maintains proper bank alignment and bank overlap. The Power Cabinet receives signals from the Logic Cabinet and generates the currents and signals for holding or moving the rods. The 53 Unit 1 control rods are divided into banks.

Control Banks A, B, C, and D banks contain a total of 25 rods and Shutdown Banks A,.B, C, D, and E contain a total of 28 rods.

Each Control Bank is divided into two groups of rods. Shutdown l Banks A and B are also divided into 2 groups of rods while  !

Shutdown Banks C, D, and E, have only one group of rods. Note l that the affected portions of the Rod Control System do not  ;

provide any inputs or controls associated with the reactor trip l function. Instead, tripping of the control rods is accomplished  ;

by the Reactor Protection System sending a trip signal to the CRDMs. I

Attachment 3 Page 2 of 5 Discussion:

TSSR 3.1.'4.2 provides confidence that the control rods can trip the reactor by verifying individual rod movement. During the Unit 1 performance of this TSSR on August 21, 1999, the Unit 1 Rod Control System experienced an equipment failure which, upon request to move Shutdown Bank A rods, caused an urgent alarm.

Shutdown Bank A was the first bank of rods selected for the testing and there was no prior indication that the affected I equipment was experiencing problems. After initial evaluation of the problem, the alarm was cleared and movement of Shutdown Bank A rods was attempted which again resulted in an urgent alarm.

The surveillance testing was then stopped and Control Bank D was selected for normal reactor control. To ensure that Control Bank D can be used for normal operating control, it was successfully moved in and out a few steps. The equipment problem causing the urgent failure alarm was isolated to the 1AC Power Cabinet

- associated with the Unit 1 Rod Control System. That power cabinet controls the movement of the Group I rods associated with Control Bank A, Control Bank C and Shutdown Bank A (Control Bank D is powered from a different power cabinet). Upon further investigation, it was determined that all three banks associated with the 1AC Power Cabinet were receiving a demand signal for movement. The expected response would be that only the selected  ;

bank would indicate a demand signal. As a result of the failure,  ;

the Group I rods associated with Control Bank A, Control Bank C and Shutdown Bank A may not move in either-the automatic or manual mode. However, this failure does not affect the ability to' trip these rods. The Unit 2 Rod Control System was not affected by the failure with the 1AC Power Cabinet.

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' McGuire proposes to effect repairs to the 1AC Power Cabinet and complete TSSR 3.1.4.2 during the 1EOC13 refueling outage.,

currently scheduled to begin on September 17, 1999. This TSSR was last performed on May 20, 1999 for Unit 1 during which no problems or equipment failures were identified. Given the surveillance frequency of 92 days and factoring in the 25% a extension of the surveillance frequency allowed by TSSR 3.0;2,  !

performance of TSSR 3.1.4.2 is due on September 12, 1999.

Approval of this LAR will allow a one time extension of this late date to allow completion of TSSR 3.1.4.2 during the 1EOCl3 refueling outage. Repair and surveillance testing during power l operations is not desired due to the risk of tripping the unit.

A trip could be initiated through the 1AC Power Cabinet or through the inadvertant actuation of other rod control circuitry.

For example, McGuire has experienced past reactor trips while performing maintenance on rod control and reactor trip circuitry at power (reference LERs 369/98-02 and 370/99-04). In addition, j the anomaly with the 1AC Power Cabinet creates the potential that the logic circuits common to other control rods could contain a T

Attachment 3 Page 3 of 5 1 fault or be affected by the fault in the affected power cabinet.

This increases the possibility that any further rod movement testing at power could cause the observed anomaly to reoccur on

-any bank of control rods, resulting in a reactor trip. McGuire PRA analyses indicate that reactor trips at power do contribute l to the Core Damage Probability. For the proposed outage repairs '

and testing, adequate controls will be in place and the timing of the repairs / testing would be coordinated to prevent any negative consequences' associated with a Rod Control System failure.

Consequently, repair and surveillance testing of the affected circuitry during the 1EOCl3 refueling outage will pose less nuclear safety risk than the inherent reactor trip risks  ;

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associated with online repair and testing. Note that, if a Unit l unscheduled outage occurs prior to the 1EOCl3 refueling outage, McGuire would perform repairs during that outage and complete TSSR 3.1.4.2 prior to entering Mode 3 upon startup from that outage. Upon verification of acceptable rod movement, the McGuire NRC Resident Inspector would be notified.

The equipment failure which caused the movement problems on the Group I. rods associated with Control Bank A, Control Bank C and Shutdown Bank A has not rendered the affected rods inoperable.

The intent of TSSR 3.1.4.2 is to provide-confidence that the control rods can trip the. reactor upon receipt of a reactor trip 7 signal. In lieu of tripping the reactor, this TSSR provides  ;

confidence that the rods are trippable by verifying rod freedom )

of movement. The BASES.for that TSSR state that, if a control rod experiences movement problems, but remains trippable and aligned, the control rod is considered to be operable. All Unit 1 control rods currently satisfy the alignment criteria of TS 3.1.4 and TSSR 3.1.4.1. The BASES for TSSR 3.1.4.2 indicates that confidence as to the trippability of the control rod (s) can be obtained by verification that any movement problem is due to l an electrical related control system failure and not the result of mechanical binding of the rods. The equipment failure which is preventing movement of some of the Unit 1 control rods is an electrical control system failure which does not affect the ability of the control rods to trip the reactor. There is no evidence of any mechanical binding of the rods. As a result, the affected control rods are trippable. Since all Unit 1 control rods are trippable and properly aligned, they are operable.

Note that the inability to move the Group I rods associated with Control Bank A, Control Bank C and Shutdown Bank A will not impact the ability to safely control the reactor during steady state power' operations. The remaining unaffected control rods are sufficient for proper power distribution and temperature control under those plant conditions. In addition, plant procedures and processes will ensure the safe controlled shutdown of Unit 1 at the start of the lEOC13 outage. McGuire PRA

Attachment 3 Page 4 of 5 analyses indicate that a failure'of the rod movement logic for a

-portion'of the control rods does not directly contribute to the failure of any' function modeled in the PRA. The: control rod

-drive system impacts the McGuire PRA model directly only-in its ability to cause'a: reactor trip and to release the rods when a

. reactor. trip when required. .The problems with the 1AC Power Cabinet may cause the' control rods to respond to some transients differently from what would normally be' expected. Consequently, the: failure increases.the probability of a reactor trip for those

-events which :would normally only result' in a' runback in reactor power. However, given the short time. period of the extension and the~relatively low frequency of transients that would be expected to cause a problem, no-meaningful impact on the estimated CDF is anticipated. The problems with the 1AC Power Cabinet do not increase'the-probability of a failure to drop the Group I rods associated with Control Bank A, Control Bank C and Shutdown Bank i A rods upon a reactor trip. Therefore, no increase in the I anticipated transient without scram (ATWS) contribution to the )

core damage frequency (CDF) is expected.

Conclusion The proposed change will provide for the one time extension of the surveillance frequency for TS Surveillance Requirement (TSSR) 3.1.4.2 beyond the 25% extension allowed by TS Surveillance l

-Requirement 3.0.2. This will facilitate repairs of the 1AC Power

. Cabinet and completion of TSSR 3.1.4.2 during the 1EOC13 refueling outage. Performance of the repairs during power operations is not desired.since McGuire has experienced past reacto'r trips while performing maintenance on rod control and reactor trip circuitry at power. Continuation of surveillance testing at power also introduces the potential for a reactor trip. McGuire PRA analyses indicate that reactor trips at power do contribute to the Core Damage Probability These same analyses indicate that the requested surveillance interval extension is expected to have no significant impact on the McGuire core damage frequency. Consequently, repair and testing of the affected circuitry during the 1EOC13 refueling outage is in the~best interests of the overall health and safety of the public since it will pose less nuclear safety risk than the inherent reactor trip risks associated with online repairs. Note that, if a Unit 1 unscheduled outage occurs prior to the 1EOC13 refueling outage, McGuire would perform repairs during that outage and complete TSSR 3.1.4.2 prior to entering Mode 3 upon startup~from that outage. Upon verification of acceptable rod movement, the McGuire NRC Resident Inspector would be notified.

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Attachment 3 l Page 5 of 5 Exigent Criteria Given that the rod movement problem was discovered on August 21, 1999, less than 30 days exists to allow for the normal comment l period under 10 CFR 50.91. The normal surveillance period of TSSR 3.1.4.2.is 92 days. Factoring in 25% grace as allowed by TS i l 3.0.2, the surveillance will be due on September 12, 1999. This

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allows only 22 days from the point of discovery until the point, j absent the performance of the surveillance or an mmendment, the l plant would'be forced to shutdown under the requirements of TS l 3.0.3. There was no prior opportunity to identify the need for a t TS amendment by; Duke Energy. In addition, prior to the l performance of the attempted surveillance on August 21, 1999, the t rod control system had exhibited normal operation. As such, DEC anticipated successful completion of the surveillance well within the 92 day surveillance period. Duke could not have anticipated  !

the need for a license amendment under these circumstances.

Given the above and the determination that-this LAR involves no significant hazards as specified.under 10 CFR 50.92, DEC believes the criteria for exigency as described in 10 CFR 50.91 are met.

Consequently, DEC requests that this LAR be processed as an exigent change pursuant to the requirements of 10 CFR 50.91 (6)

(i) which allows for either (A) a 14 day FEDERAL REGISTER notice l or (B) a local media notice if less than 14 days exists for the 1 FEDERAL REGISTER notice. '

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o ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATIONS l

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Attachment 4 l Page 1 of 3 l l

No Significant Hazards Considerations: j In accordance with the criteria set forth in 10 CFR 50.91 and 50.92, McGuire Nuclear Station has evaluated this proposed Technical Specification change and determined it does not represent a significant hazards consideration. The following is provided-in support of this conclusion.

1. Does the change involve a'significant increase in the probability or consequences of an accident previously evaluated?

No. Performance of the TSSR during power operations would result-in a higher probability of an accident as compared to performing this testing during an outage. If performed at l power, a trip could be initiated through the 1AC Power Cabinet while repairing or replacing components or through the inadvertant actuation of other rod control circuitry.

For example, McGuire has experienced past reactor trips while performing maintenance on rod control and reactor trip circuitry at power. In addition, continuation of surveillance testing at power also introduces the potential for a reactor trip. McGuire PRA analyses indicate that reactor trips at power do contribute to the Core Damage Probability. For the proposed outage repairs and testing, adequate controls will be in place and the timing of the repairs / testing would be coordinated to prevent any negative consequences associated with a Rod Control System failure. j The equipment failure which affected movement of the Group I rods associated with Control Bank A, Control Bank C and Shutdown Bank A has not caused the affected rods to be inoperable. The intent of TSSR 3.1.4.2 is to, in lieu of tripping the reactor, verify rod movement in order to provide confidence that the control rods can trip the reactor. The BASES for that TSSR state that, if a control rod experiences movement problems, but remains trippable and aligned, the control rod is considered to be operable. All Unit 1 control rods currently satisfy the alignment criteria of TS 3.1.4 and TSSR 3.1.4.1. The BASES for TSSR 3.1.4.2 indicates that confidence as to the trippability of the control rod (s) can be obtained by verification that a rod movement problem is due to an electrical related control system failure and not

-the_ result of mechanical binding of the rods. The equipment failure which is preventing movement of some of the Unit 1 control rods is due to an electrical control system failure which does not affect the ability of the control rods to trip the reactor. There is no evidence of any mechanical binding of the rods.

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Attachment 4 Page 2 of 3 As a result, the affected control rods are trippable. Since all Unit 1 control rods are trippable and properly aligned, they are operable and, in the absence of any other failures, should remain operable until repairs and surveillance testing can be effected in the lEOCl3 refueling outage.

The inability to move the Group I rods associated with Control Bank A, Control Bank C and Shutdown Bank A will not impact the ability to safely control the reactor during steady state power operations prior to the 1EOC13 refueling outage. The remaining unaffected control rods are sufficient for proper-power distribution and temperature control under those plant conditions. In addition, plant procedures and processes will ensure the safe controlled shutdown of Unit 1 at the start of the 1EOC13 outage. McGuire PRA analyses indicate that a failure of the rod movement logic for a portion of the control rods does not directly contribute to the failure of any function modeled in the PRA. The control rod drive system impacts the McGuire PRA model directly only in its ability to cause a reactor trip and to release the rods when a reactor trip when required. The problems with the 1AC Power Cabinet may cause the control rods to respond to some transients _ differently from what would normally be expected. Consequently, the failure increases the probability of a reactor trip for those events which would normally only result in a runback in reactor power. However, given the short time period of the extension and the relatively low frequency of transients that would be expected to cause a problem, no meaningful impact on the estimated CDF is~ anticipated. The problems with the LAC Power Cabinet do not increase the probability of a failure to drop the Group I rods associated with Control Bank A, Control Bank C and Shutdown Bank A rods upon a reactor trip. Therefore, continued operation until repair and testing of the failed equipment during the 1EOCl3 outage should not result in an increase in'ATWS contribution to the CDF.

Note that the banks that have not been fully surveillance tested constitute approximately 98% of the total rod worth.

However, the likelihood of any rod, much less all of these rods, not fully inserting is extraordinarily remote. In addition, McGuire Nuclear Station's design incorporates the AMSAC'(ATWS Mitigation System Actuation Circuit) for the worst case such event. .The worst case event is the failure of reactor tip breakers to open when called upon by a valid reactor tip signal. The current condition is clearly bounded by that postulated occurrence.

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Attachment 4 Page 3 of 3

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

No. No changes are being made to actual plant hardware or processes which will result in any new-accident causal mechanisms. -Also, no changes are being made to the way in which the plant is being operated. Therefore, no new accident causal mechanisms will be generated.

l 3. Does this change involve a significant reduction in a margin i of safety?

No. Margin'of safety is related to the ability'of the fission' product barriers to perform their design functions during and'following accident conditions. These barriers  !

include the fuel cladding, the reactor coolant system, and the containment system. Based upon-the response to question

1. 1, the performance of these barriers will not be degraded by the proposed changes.

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I ATTACHMENT 5 i

ENVIRONMENTAL IMPACT ASSESSMENT I

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Attachment 5 Page 1 og 3 Environmental Impact Assessment:

The proposed Technical Specification amendment has been reviewed

-against the criteria of 10 CFR 51.22 for environmental considerations. The proposed amendment does not involve a significant hazards consideration, nor increase the types and amounts of effluents that may be released offsite, nor increase individual or cumulative occupational radiation exposures.

Therefore, the proposed amendment meets the criteria given in 10 CFR 51.22(c)(9) for a categorical exclusion from the requirement for an Environmental Impact Assessment.

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