ML041340394

License Amendment Request: Control Rod Drive Mechanism Surveillance

ML041340394
Person / Time
Site:	Palisades
Issue date:	05/10/2004
From:	Domonique Malone Nuclear Management Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML041340394 (20)

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Comfedto dNu MCr Palisades Nuclear Plant Operated by Nuclear Management Company, LLC May 10, 2004 10 CFR 50.90 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Palisades Nuclear Plant Docket 50-255 License No. DPR-20 License Amendment Request: Control Rod Drive Mechanism Surveillance Pursuant to 10 CFR 50.90, Nuclear Management Company, LLC (NMC) requests Nuclear Regulatory Commission (NRC) review and approval of a proposed license amendment for the Palisades Nuclear Plant. The proposed amendment requests NRC approval to replace existing license condition 2.C.(5) and its corresponding table, Table 2.C.(5), which are outdated, with a new license condition stating that performance of Technical Specification (TS) surveillance requirement (SR) 3.1.4.3 is not required, for control rod drive (CRD) 19 only, until the next refueling outage, but no later than September 30, 2004.

SR 3.1.4.3 is normally performed every 92 days to verify control rod freedom of movement. This testing is believed to be aggravating the seal leakage of two CRDs, CRD-19 and CRD-29. TS limiting condition for operation (LCO) 3.1.4, "Control Rod Alignment," condition D, which allows one full-length control rod to be immovable provided it is still trippable, is currently being applied to CRD-29.

SR 3.1.4.3 does not apply to CRD-29 because of the application of LCO 3.1.4 condition D. Condition D would continue to be applied to CRD-29 for the duration of the proposed change. Without the proposed change, testing of CRD-1 9 must continue. Accelerated seal degradation is expected following the next test, which could result in a forced shutdown due to excessive seal leakage. NMC has determined that operation without the proposed change creates a potential challenge to reactor safety without providing an overall benefit in safety.

27780 Blue Star Memorial Highway

• Covert, Michigan 49043-9530 Telephone: 269.764.2000

Document Control Desk Page 2 provides a detailed description of the proposed change, background and technical analysis, No Significant Hazards Consideration Determination, and Environmental Review Consideration. Enclosure 2 provides annotated operating license pages showing the changes proposed. Enclosure 3 provides the revised operating license page reflecting the proposed change. Enclosure 4 provides a figure of a control rod drive mechanism. Enclosure 5 provides CRD leakoff rates for the current operating cycle. Enclosure 6 provides a determination of CRD leakoff effect on trippability.

NMC requests approval of this proposed license amendment by July 15, 2004, with the amendment being implemented within 15 days. The approval date was selected to avoid exercising CRD-19 at the next required SR due date.

A copy of this request has been provided to the designated representative of the State of Michigan.

Summary of Commitments This letter contains one new commitment that would apply if the proposed amendment request is approved. There are no revisions to existing commitments.

• NMC will inspect the seals for CRD-19 and CRD-29 and make appropriate repairs prior to entering Mode 2 following the next entry into Mode 3.

I declare under penalty of perjury that the foregoing is true and accurate. Executed on May 10, 2004.

Daniel J. Malone Site Vice President, Palisades Nuclear Plant Nuclear Management Company, LLC Enclosures (6)

CC Regional Administrator, Region Ill, USNRC Project Manager, Palisades, USNRC NRC Resident Inspector, Palisades USNRC

ENCLOSURE I DESCRIPTION OF REQUESTED CHANGES

1.0 DESCRIPTION

Nuclear Management Company, LLC (NMC) requests to amend Operating License DPR-20 for the Palisades Nuclear Plant. The proposed amendment requests Nuclear Regulatory Commission (NRC) approval to replace existing license condition 2.C.(5) and its corresponding table, Table 2.C.(5), with a new license condition stating that performance of Technical Specification (TS) surveillance requirement (SR) 3.1.4.3 is not required, for control rod drive (CRD) 19 only, until the next refueling outage, but no later than September 30, 2004.

2.0 PROPOSED CHANGE

NMC proposes to revise the Palisades Operating License to replace existing license condition 2.C.(5) and its corresponding table, Table 2.C.(5), with a new license condition stating that performance of TS SR 3.1.4.3 is not required, for CRD-1 9 only, until the next refueling outage, but no later than September 30, 2004.

3.0 BACKGROUND

Description of current license condition 2.C.(5)

Amendment 206 to Operating License No. DPR-20, dated December 19, 2001, revised the Palisades Operating License. This amendment added license condition 2.C.(5) and Table 2.C.(5) to extend certain TS surveillance requirement intervals, on a one-time basis, to permit them to be performed during a refueling outage, but no later than April 30, 2003.

Description of CRDs The control rod drive mechanisms (CRDMs) at Palisades are of the vertical rack-and-pinion type with the drive shaft running parallel to the rack and driving the pinion gear through a set of bevel gears. The design of the drive is shown in Enclosure 4. These drives have a drive package, containing components both inside and outside of the primary coolant system (PCS) pressure boundary. The drive motor, position indication system, and a magnetic clutch, are outside the PCS boundary. The drive shaft, right angle gear set, pinion gear, and rack, are inside the PCS boundary. Part-length control rods, which maintain their position during a reactor trip, have a solid shaft assembly instead of a magnetic clutch.

The drive package is connected to the drive shaft through a face-type rotating seal, which forms part of the PCS pressure boundary. Leakage through the face-type rotating seal enters a cavity which is vented to a collection header, and which is sealed at the top by a vapor seal. Each face-type rotating seal has a Page 1 of 8

thermocouple to measure the temperature of the leakoff. The leakoff from all forty-five CRDMs is collected in a normally unpressurized common header and routed to the containment sump.

An increase in leakage of a CRDM seal is detected by an indicated increase in the CRDM leakoff temperature. Individual CRDM seal leakoff temperatures are available for review and trending on a chart recorder in the control room.

Leakage measurement from individual seals is not possible. Combined leakage from all seals, collected in the common seal leakoff header, can be measured locally inside containment. The current combined leakage from all seals is approximately 680 ml/min. Since the seal leakoff header flow is directed to the containment sump, observing the rate of containment sump level rise can also be used to approximate the combined seal leakoff flow rate if there is not significant leakage from other sources.

The full-length CRDMs provide two functions. First, a reactor trip signal de-energizes the clutch and allows the control rod to drop by gravity into the core.

This is the only CRDM safety function assumed in the safety analyses. Second, when a reactor trip signal is generated, a rod rundown signal energizes all full-length CRDM motors to drive their control rods in case they should not trip freely into the core. The rod rundown signal is terminated when that control rod nears full insertion. The clutch is designed to allow the motor to apply torque in the "IN" direction, even when the clutch is released. Functioning of the rod rundown feature is not assumed in the safety analyses.

Description of the Current TS Requirements Limiting Condition for Operation (LCO) 3.1.4, 'Control Rod Alignment," requires that all control rods be operable. SR 3.1.4.3 requires operability to be demonstrated at least once per 92 days by moving each individual full-length control rod that is not fully inserted into the reactor core 2 6 inches in either direction. SR 3.1.4.6 verifies that each full-length control rod drop time is

< 2.5 seconds prior to reactor criticality, after each reinstallation of the reactor head.

LCO 3.1.4, condition D, allows one control rod to be immovable, provided it is still trippable. The required action for this condition is to restore the control rod to operable status prior to entering Mode 2 following the next entry into Mode 3.

TS LCO 3.4.13, PCS Operational Leakage, limits unidentified leakage to one gallon per minute (gpm) and identified leakage to 10 gpm. SR 3.4.13.1 requires verification of PCS operational leakage every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. In addition, a plant procedure directs a reactor shutdown when confirmed CRDM seal leakage is in excess of two gpm.

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Bases for the Current CRD Requirements Verifying each full-length control rod is trippable would require that each full-length control rod be tripped. In Modes 1 and 2, tripping each full-length control rod would result in radial or axial power tilts, or oscillations. Therefore, individual full-length control rods are exercised every 92 days to provide increased confidence that all full-length control rods continue to be trippable, even if they are not regularly tripped. A movement of 6 inches is adequate to demonstrate motion without exceeding the alignment limit when only one control rod is being moved. The 92-day frequency takes into consideration other information available to the operator in the control room and other surveillances being performed more frequently, which add to the determination of operability of the control rods. At any time, if a control rod(s) is inoperable, a determination of the trippability of the control rod(s) must be made, and appropriate action taken.

Condition 3.1.4 D would apply whenever it is discovered that a single full-length control rod cannot be moved by its operator, either functionally or administratively, yet the control rod is still capable of being tripped (or is fully inserted).

Condition D is entered whenever it is discovered that a single full-length control rod cannot be moved by its operator, either functionally or administratively, yet the control rod is still capable of being tripped (or is fully inserted). Although the ability to move a full-length control rod is not an initial assumption used in the safety analyses, it does relate to full-length control rod operability. The inability to move a full-length control rod by its operator may be indicative of a systemic failure (other than trippability) that could potentially affect other rods. Thus, declaring a full-length control rod inoperable in this instance is conservative since it limits the number of full-length control rods that cannot be moved by their operators to only one. The completion time to restore an inoperable control rod to operable status is stated as prior to entering Mode 2 following next Mode 3 entry. This completion time allows unrestricted operation in Modes 1 and 2, while conservatively preventing a reactor startup with an immovable full-length control rod.

Need for Revision of CRD Requirement In July 2003, NMC observed increased temperature on seal leakoff from CRD-19 at the Palisades Nuclear Plant. NMC chose to declare CRD-19 immovable but trippable due to the elevated seal leakage. LCO 3.1.4, condition D, was entered for CRD-19. Therefore, for CRD-19, performance of SR 3.1.4.3 was not required until Mode 2 following the next Mode 3 entry.

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Following performance of SR 3.1.4.3 in January 2004, CRD-29 exhibited increased seal leakage. For the subsequent performance of SR 3.1.4.3 in April 2004, NMC determined that exercising CRD-19 would be less likely to negatively impact seal performance than exercising CRD-29. On April 6, 2004, SR 3.1.4.3 was performed successfully for CRD-19 and LCO 3.1.4 condition D was exited. Subsequently, LCO 3.1.4 condition D was entered for CRD-29, since it was not exercised within the surveillance frequency. NMC noted CRD-1 9 exhibited increased seal leakage after performance of SR 3.1.4.3.

Refer to Enclosure 5 for CRD leakoff rate data for the current operating cycle.

In accordance with the TS frequency of 92 days, SR 3.1.4.3 must be performed no later than July 30, 2004, for CRD-19. This date includes the 25% extension allowed by SR 3.0.2. The SR is not required for CRD-29 since LCO 3.1.4 Condition D is being applied to this CRD. Therefore, a maximum of 177 days is being requested to perform SR 3.1.4.3 for CRD-19, which is 192% of the 92-day frequency. Note that LCO 3.1.4 condition D would remain applied to CRD-29 for the duration of the extended surveillance interval.

If the proposed license amendment is approved, the expected rate of seal degradation may be reduced, which may facilitate continued operation until the next scheduled refueling outage. Excessive seal leakage may still force a maintenance outage prior to the refueling outage, but such a shutdown will be less probable with the reduced testing requirement.

4.0 TECHNICAL ANALYSIS

Operating License DPR-20, condition 2.C.(5) and Table 2.C.(5) are no longer required since the allowed extension expired on April 30, 2003. Therefore, it is appropriate to delete them.

Review of the performances of related test procedures for the current operating cycle showed that CRD-1 9 was moved Ž 6 inches in either direction without failure during the reactor startup for the current operating cycle, in April 2003, and again on April 6, 2004, during performance of SR 3.1.4.3 testing. Therefore, CRD-19 was successfully moved after more than 11 months of being in a fully withdrawn condition. In addition, the part-length control rods, which are fully withdrawn prior to reactor startup, are not exercised, and typically remain motionless throughout the operating cycle. Operating experience at Palisades has shown that these rods will insert into the core during a normal shutdown.

Therefore, NMC has determined that CRD-19 can be expected to continue to move freely when needed.

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NMC has determined that CRD seal leakage does not increase the likelihood of an untrippable control rod. Enclosure 6 provides the details of this determination, which provides confidence in trippability.

Plant operating experience has shown that completion of SR 3.1.4.3, for CRD-19, would most likely result in increased seal leakage. Increased seal leakage could require a plant shutdown prior to the next refueling outage. An unscheduled plant shutdown during the summer could have an adverse impact on grid stability. Therefore, performance of this test, for CRD-1 9, creates a potential challenge to reactor safety without providing an overall benefit in safety.

NMC calculates PCS operational leakage on odd dates of the calendar month and CRD seal leakoff on a monthly frequency. Plant procedures require initiation of an action request when unidentified PCS operational leakage exceeds 0.1 gpm. Unidentified PCS operational leakage levels at or above 0.15 gpm will result in increased management involvement. These action levels are well below the Technical Specification value of 1.0 gpm for unidentified PCS operational leakage. NMC is also required to shutdown the reactor if confirmed CRD seal leakage exceeds two gpm, in accordance with plant procedures.

The root cause of the leakage is currently unknown. NMC believes, based on seal design and previous operating experience, that it is unlikely that any of the CRD seal leakage is being deposited on the reactor pressure vessel head. NMC will inspect the seals for CRD-19 and CRD-29 and make appropriate repairs prior to entering Mode 2 following the next entry into Mode 3.

Based on the information described above, the proposed extension would not significantly impact CRD performance, reliability, or monitoring.

5.0 REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration Nuclear Management Company, LLC (NMC) has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below:

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

Response: No.

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The proposed license amendment deletes outdated information from the operating license and adds a license condition to delay testing of one control rod from the Palisades Technical Specification surveillance requirement for partial movement every 92 days. The proposed License Condition does not affect or create any accident initiators or precursors.

As such, the proposed license condition does not increase the probability of an accident.

The proposed license amendment does not significantly increase the consequences of an accident. The safety analyses assume full-length control rod insertion, except the most reactive rod, upon reactor trip. The proposed surveillance requirement (SR) extension request does not increase the allowed outage time of any required operable structures, systems, or components (SSCs), and does not reduce the requirement to know that the deferred SR could be met at all times. Deferral of testing does not, by itself, increase the potential that the testing would not be met. The ability to move a full-length control rod by its drive mechanism is not an initial assumption used in the safety analyses. Control rod drop times are verified during performance of a surveillance that is normally performed during refueling outages. NMC has determined that control rod drive (CRD) seal leakage does not increase the likelihood of an untrippable control rod. Therefore, the assumptions of the safety analyses will be maintained, and the consequences of an accident will not be increased significantly.

Deleting the existing license condition 2.C.(5) and Table 2.C.(5) is administrative, since the provision has expired, and has no impact on plant operation or equipment.

Therefore, operation of the facility in accordance with the proposed License Condition would not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed license condition does not involve a physical alteration of any SSC or change the way any SSC is operated. The proposed license condition does not involve operation of any required SSCs in a manner or configuration different from those previously recognized or evaluated. No new failure mechanisms will be introduced by the SR deferral being requested.

Deleting the existing license condition 2.C.(5) and Table 2.C.(5) is administrative, since the provision has expired, and has no impact on plant operation or equipment.

Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The safety analyses assume full-length control rod insertion, except the most reactive rod, upon reactor trip. The proposed License Condition does not, by itself, introduce a failure mechanism. Past performance of the SR in question has demonstrated reliability in passing the deferred SR. The proposed license condition does not involve any physical changes to the plant or manner in which the plant is operated. The ability to move a full-length control rod by its drive mechanism is not an initial assumption used in the safety analyses. Control rod drop times are verified during performance of a surveillance that is normally performed during refueling outages. NMC has determined that CRD seal leakage does not increase the likelihood of an untrippable control rod. Therefore, the assumptions of the safety analyses will be maintained, and the margin of safety is not reduced significantly.

Deleting the existing license condition 2.C.(5) and Table 2.C.(5) is administrative, since the provision has expired, and has no impact on plant operation or equipment.

Therefore, the proposed amendment would not involve a significant reduction in a margin of safety.

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Based on the evaluation above, NMC concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

5.2 Applicable Regulatory Requirements/Criteria NUREG-1432, "Standard Technical Specifications Combustion Engineering Plants," contains a SR to verify freedom of movement of control rods every 92 days. The proposed amendment extends this SR interval for CRD-19 only. The proposed surveillance requirement (SR) extension request does not increase the allowed outage time of any required operable SSCs, and does not reduce the requirement to know that the deferred SR could be met at all times. Deferral of testing does not, by itself, increase the potential that the testing would not be met.

In conclusion, based on the considerations described above, (1)there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

6.0 ENVIRONMENTAL CONSIDERATION

NMC has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

7.0 REFERENCES

1. Amendment 206 to Facility Operating License No. DPR-20, 'Palisades Plant - Issuance of Amendment to Extend Surveillance Requirement Intervals," dated December 19, 2001.

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ENCLOSURE 2 PROPOSED OPERATING LICENSE CHANGES (mark-up) 2 Pages Follow

- For SRs that existed prior to this amendment whose intervals of performance are being reduced, the first reduced surveillance interval begins upon completion of the first surveillance performed after implementation of this amendment.

- For SRs that existed prior to this amendment that have modified acceptance criteria, the first performance is due at the end of the first surveillance interval that began on the date the surveillance was last performed prior to the implementation of this amendment.

- For SRs that existed prior to this amendment whose intervals of performance are being extended, the first extended surveillance interval.

begins upon completion of the last surveillance performed prior to the implementation of this amendment.

(5) In lieu of the specified frequencyies, NMC may complete Ohe-surveillance requirements (SRs) noted in Table 2.C.(5) on Page 4a f.1.4;.3forcontr rod drive 196onIy, during the next refueling outage, but no later than APS epteber 30,2004.

D. The facility has been granted certain exemptions from the requirements of Section III, G of Appendix R to 10 CFR Part 50, uFire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979." This section relates to fire protection features for ensuring the systems and associated circuits used to achieve and maintain safe shutdown are free of fire damage. These exemptions were granted and sent to CPCo* in letters dated February 8, 1983, July 12, 1985, and July 23, 1985.

In addition, the facility has been granted certain exemptions from Appendix J to 10 CFR Part 50, "Primary Reactor Containment Leakage Testing for Water Cooled Power Reactors." This section contains leakage test requirements, schedules and acceptance criteria for tests of the leak-tight integrity of the primary reactor containment and systems and components which penetrate the containment. These exemptions were granted and sent to CPCo in a letter dated December 6, 1989.

These exemptions granted pursuant to 10 CFR 50.12, are authorized by law, will not present an undue risk to the public health and safety, and are consistent with the common defense and security. With these exemptions, the facility will operate, to the extent authorized herein, in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission.

On March 11, 1997, the name "Consumers Power Company" was changed to "Consumers Energy Company." Nuclear Management Company, LLC, hereinafter referred to as NMC, succeeds Consumers Energy Company as operator of the Palisades Plant. Consequently, NMC is authorized to act as agent for Consumers Energy Company and has exclusive responsibility and control over the physical construction, operation, and maintenance of the facility.

Amendment No.474,476, 492,201, 26,

Surveillance Requirnment Deseription SR3.3.3.3 Channel Calibration of Safety Injction and (for Table 3.3.3 1, ltem 3.a) Rcfucling Watcr Tank Low Level.

SR 3.3.4.3 Channcl Functional Tcst of Safety Injection (for Table 3.3.4 1, Item 1) Signal (I-SeS)efuneton.

SR 3.3.4.3 Channel Functional Tcst of Recirculation (for Table 3.3.4 1, Item 3) Aetuatien Signal-funetien-.

SR- 3.3.6.1 Channel Functional Tcst of Dicecl Gcncrator Undervoltagc Start logic.

SR 3.5.2.8 (High Prcssure Safety Injection Throttle valve position stop is in correct to Hat Leg 1-Valves M)-3082 and M) 3083 pesitien.

SR 3.7.8.2 (Non Critical Scrvic Watcr Automatic valve actuates to the correct Header Iselation Valyc CV 1359 only) position on actual or simulated actuation SR 3.E8r..gency AC power performs as required on actual or simulated loss of offsite power (LGGP) signal.

SRt3.8.1.9 .Efmgcncy AC powcer performs as required on actual or simulated restoration of offsitc pewer.

SR 3.8.1.1 0 Load sequencing for each automatic load sequenecr.

SRl3.8.1 Emergency AC power performs as required on actual or simulated LOOP signal in conjunction with actual or simulated SIS.

Amendment No. 206

ENCLOSURE 3 PROPOSED OPERATING LICENSE CHANGES (typed) v 1 Page Follows

- For SRs that existed prior to this amendment whose intervals of performance are being reduced, the first reduced surveillance interval begins upon completion of the first surveillance performed after implementation of this amendment.

- For SRs that existed prior to this amendment that have modified acceptance criteria, the first performance is due at the end of the first surveillance interval that began on the date the surveillance was last performed prior to the implementation of this amendment.

- For SRs that existed prior to this amendment whose intervals of performance are being extended, the first extended surveillance interval begins upon completion of the last surveillance performed prior to the implementation of this amendment.

(5) In lieu of the specified frequency, NMC may complete surveillance requirement (SR) 3.1.4.3, for control rod drive 19 only, during the next refueling outage, but no later than September 30, 2004.

D. The facility has been granted certain exemptions from the requirements of Section III, G of Appendix R to 10 CFR Part 50, "Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979." This section relates to fire protection features for ensuring the systems and associated circuits used to achieve and maintain safe shutdown are free of fire damage. These exemptions were granted and sent to CPCo* in letters dated February 8, 1983, July 12, 1985, and July 23,1985.

In addition, the facility has been granted certain exemptions from Appendix J to 10 CFR Part 50, "Primary Reactor Containment Leakage Testing for Water Cooled Power Reactors." This section contains leakage test requirements, schedules and acceptance criteria for tests of the leak-tight integrity of the primary reactor containment and systems and components which penetrate the containment. These exemptions were granted and sent to CPCo' in a letter dated December 6, 1989.

These exemptions granted pursuant to 10 CFR 50.12, are authorized by law, will not present an undue risk to the public health and safety, and are consistent with the common defense and security. With these exemptions, the facility will operate, to the extent authorized herein, in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission.

On March 11, 1997, the name "Consumers Power Company" was changed to Consumers Energy Company." Nuclear Management Company, LLC, hereinafter referred to as NMC, succeeds Consumers Energy Company as operator of the Palisades Plant. Consequently, NMC is authorized to act as agent for Consumers Energy Company and has exclusive responsibility and control over the physical construction, operation, and maintenance of the facility.

Amendment No. 4-74,1476, 492, 204, 26,

ENCLOSURE 4 CONTROL ROD DRIVE MECHANISM FIGURE I Page Follows

. FSAR CHAPTER 3 - REACTOR FIGURE 3-13 Revision 21 CONTROL ROD DRIVE MECHANISM

• DRIVE SHAFT A LETDOWN NUT HARD STOP RACK B

THERMAL SLEEVE DETAIL "A" DETAIL "B"

ENCLOSURE 5 CONTROL ROD DRIVE LEAKOFF RATES FOR CURRENT OPERATING CYCLE Measured Control Rod DATE Drive (CRD) Seal Comments Leakoff Rate (ml/min) 5/20/03 5 6/19/03 25 7/3/03 45 CRD-19 leakoff temperature alarms in morning. SR 3.1.4.3 performed for all full-7/14/03 N/A length control rods except CRD-19 in evening. CRD-29 noted to rise after performance of SR 3.1.4.3.

7/15/03 50 9/9/03 40 SR 3.1.4.3 performed for all full-length 10/7/03 110 control rods except CRD-1 9. CRD leakoff rate measured after SR 3.1.4.3.

11/4/03 60 12/2/03 70 1/4/04 N/A CRD-19 leakoff temperature alarms.

SR 3.1.4.3 performed for all full-length control rods except CRD-1 9. CRD-29 noted 1/6/04 275 to rise following performance of SR 3.1.4.3 (still below alarm setpoint). CRD leakoff rate measured after SR 3.1.4.3.

1/9/04 230 1/27/04 200 2/26/04 320 3/16/04 330 3/23/04 338 SR-3.1.4.3 performed for all full-length 4/6/04 N/A control rods except CRD-29. CRD-19 leakoff temperature alarms after performance of SR 3.1.4.3.

4/20/04 680 Page 1 of I

ENCLOSURE 6 CONTROL ROD DRIVE LEAKOFF EFFECT ON TRIPPABILITY Nuclear Management Company, LLC (NMC) has determined that control rod drive (CRD) seal leakage does not increase the likelihood of an untrippable control rod. In order to do so, the seal leakage would have to cause the clutch to fail to release, or cause mechanical binding of the driveshaft between the lower clutch face and the face-type rotating seal. All components above the lower clutch face are disengaged from the drive shaft on a reactor trip. Components inside the primary coolant system (PCS) pressure boundary are normally wetted, and therefore, will not be mechanically bound by the leakage effects. The following text addresses seal components and their effect on trippability.

Clutch In order to hinder trippability, the lower section must either fail to disengage or jam between the shaft and some stationary component. Plausible failure modes cause the clutch to disengage (thus causing a rod trip), not remain engaged. The clutch uses a spring bellows and jaw faces that do not depend upon sliding action. When electrical power is removed, the jaw faces separate, an action that is not prone to mechanical jamming. Even if the vapor seal failed, leakage would not prevent rotation of a disengaged lower clutch element.

Bearings There are three sets of ball bearings between the clutch and vapor seal. To prevent a rod trip, one or more of these sets would have to bind sufficiently to resist dropping of a weight in excess of 200 pounds, or degrade badly enough to allow gross driveshaft misalignment. The vapor seal protects the bearings from a corrosive atmosphere, and the two gallons per minute procedural CRD seal leakage limitation reduces the likelihood of vapor seal failure.

Vapor seal The vapor seal is an elastomeric cup seal with a metal backing ring. The steam impingement washer protects it from erosion, 'and the vapor seal in turn protects drive components above the vapor seal from leakage. Operating temperature is dependent upon seal leakoff pressure as long as flashing occurs in the leakoff cavity. The collection header is normally unpressurized. The elastomer is designed for high temperature operation, and there is no metal-to-metal contact between stationary and rotating parts. If the vapor seal were to fail, itwould not itself prevent shaft rotation.

Steam Impingement Washer The steam impingement washer is a thin stainless washer fit loosely around the driveshaft immediately below the vapor seal, at the top of the seal leakoff cavity. It cannot bind between the shaft and housing while remaining around the shaft, and plausible leaks will not break it.

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i Seal Assembly The rotating element is inside the PCS boundary, so leakage will not corrode or bind small internal parts. There is clearance between the stationary assembly and driveshaft. Shear forces will prevent binding at the seal boundary itself, as the seal contact area is very small and materials were selected for low friction operation. A leak-induced temperature increase can degrade the three static O-rings, but this will not prevent rotation.

Driveshaft One end of the driveshaft is inside the PCS boundary, so component material was selected to withstand PCS effects. Driveshaft upper end alignment is maintained by the lower clutch shaft that rides in three sets of ball bearings above the vapor seal. The drive shaft lower end bearings are within the PCS boundary.

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