ML042100269

Response to Request for Additional Information Concerning Alternative Repair of Control Rod Drive Housing Interface with Reactor Vessel

ML042100269
Person / Time
Site:	Oyster Creek
Issue date:	07/20/2004
From:	Gallagher M AmerGen Energy Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
2130-04-20157
Download: ML042100269 (8)

Text

fAmerGenSM ArnerGen Energy Company, LLC www.exeloncorp.com An Exelon Company 200 Exelon Way Kennett Square, PA 19348 10 CFR 50.55a 2130-04-201.57 July 20. 20C4 U. S. Nucleai Regulatory Commission Attn

Document Control Desk Washington, DC 20555 Oyster Creek Generating Station Facility License No. DPR-16 Docket No. 50-219

Subject:

Response to Request for Additional Information Concerning Alternative Repair of Control Rod Drive Housing Interface with Reactor Vessel Ref eiences: I1) AmerGen letter 2130-00-20300 dated November 10, 2000, "Alternative Repair of Control Rod Drive Housing Interface with Reactor VessePl

2) AmerGen letter 2130-00-20304 dated November 14, 2000, Modification to Proposed Alternative Repair of Control Rod Drive Housing Interface with Reactor Vessel"

3) USNRC letter dated November 16, 2000, "Request to Use an Alternative Repair of the Control Rod Drive Housing Interface with the Reactor Vessel at the Oyster Creek Nuclear Generating Station (TAC NO.

MB0461)"

4) AmerGen letter 2130-01-20031 dated January 19, 2001, "Alternative Repair of Control Rod Drive Housing Interface with Reactor Vessel Clarification of Leakage Inspection"

5) USNRC letter dated January 8, 2002, "Oyster Creek Nuclear Generating Station - Clarification of Leakage Inspection (TAC NO. MB1 065)"

6) AmerGen letter 2130-02-20214 dated July 26, 2002, "Alternative Repair of Control Rod Drive Housing Interface with Reactor Vessei"

7) AmerGen letter 2130-02-20291 dated Oclober 4, 2002, "Additional Information - Alternative Repair of Control Rod Drive Housing Interface with Reactor Vessel (TAC No. M135700)"

8) USNRC letter dated October 18, 2002, "Oyster Creek Nuclear Generating Station - Alternative Repair of Control Rod Drive Housing Interface with Reactor Vessel (TAC NO. MB5700)"

9) AmerGen letter 2130-03-20271 dated October 21, 2003, 'Alternative Repair of Control Rod Drive Housing Interface with Reactor Vessel" In the Reference 9 letter, in accordance with 10 CFR 50.55a(a)(3)(i), AmerGen Energy Company, LLC (AmerGen) requested continued approval of the proposed alternative to 10 CFR 50.55a(g) as contained in the Reference 6 letter above. This issue was discussed with the NRC staff in a conference call dated May 26, 2004. Attached is our response to questions provided in the call.

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U.S. Nuclear Regulatory Commission July 20, 2004 Page 2 Ifyou should have any questions, please contact Mr. Tom Loomis at 610-765-5510.

Very truly yours, Michael P. Gallagher Director, Licensing and Regulatory Affairs AmerGen Energy Company, LLC cc: H. J. Miller, USNRC, Administrator, Region I P. S. Tam, USNRC, Senior Project Manager, Oyster Creek R. J. Summers, USNRC, Senior Resident Inspector, Oyster Creek File No. 00086

Response to RAI Concerning Alternative Repair of Control Rod Drive Housing Interface with Reactor Vessel Page 1 Oyster Creek Roll Repair Relief Request Questions / Responses from NRC Call dated May 26, 2004 Question:

1. Your October 21, 2003, submittal states that you have pursued an American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME) Section Xl Code Case regarding roll-expansion repair of leaking control rod drive (CRD) and in-core penetrations.

The ASME Section Xl Focus Group on Welding and Special Repair Processes is further developing this Code Case. What are the differences between the proposed Code Case requirements, and the requirements that Oyster Creek intends to use for this outage? Why are these differences acceptable?

Response

The code case is currently in draft form and is at the working group level for review. Current differences exist between the draft code case and the current commitments at Oyster Creek for leakage limits and inservice inspection. The current differences between the code case and the Oyster Creek Generating Station (OCGS) inspection requirements for this outage are as follows:

• The code case includes new allowable leakage rates that are much lower than the BWRVIP-17 leakage rates that OCGS is utilizing now. Note that no leakage has been observed from the two roll repairs at OCGS since the repairs were performed in 2000.

• OCGS is currently committed to a visual inspection of the roll repair each time access is gained to under the vessel, which may include forced outage conditions. Therefore, current inspection requirements are more stringent with regards to inspection for leakage.

Question:

2. Describe your future plans and decision points in time with respect to implementing a permanent welded repair in accordance with the ASME Code should the Code Case accepting this as a permanent repair prove to be unsuccessful.

Response

As discussed in the October 21, 2003 letter, for repair of any additional penetrations that may exhibit leakage, AmerGen intends to employ the roll-expansion repair technique as described in Boiling Water Reactor Vessel and Internals Project (BWRVIP) Report, "Roll/Expansion Repair of Control Rod Drive and In-Core Instrument Penetrations in BWR Vessels (BWRVIP-17)," as used in repair of the two penetrations in Fall 2000. A weld repair according to BWRVIP-58, "BWR Vessel and Internals Project, CRD Internal Access Weld Repair," or an equivalent Code repair, will be employed if roll-expanded penetrations do not meet the new allowable leakage rates (see attached tables) or more than five CRD penetrations continue to leak after roll-expansion repairs.

Response to RAI Concerning Alternative Repair of Control Rod Drive Housing Interface with Reactor Vessel Page 2 The BWRVIP-58 process has not been implemented at a domestic nuclear reactor. We do not plan to have this process ready as a contingency repair plan for the R20 refuel outage this November due to the significant cost and lack of field experience with this process and the success of the roll repairs at OCGS.

Question:

3. Describe in detail all of the inspections that will be performed on CRD housing penetrations 42-43 and 46-39 that were roll repaired. Currently, your October 21, 2003, submittal states that inspections will continue as discussed in References 4 and 5 of your submittal.

However, these references only address the VT-2 visual examination for leakage. The NRC staff's prior approvals of this repair technique, dated October 18, 2002, specified ultrasonic test (UT) examinations of the CRD housings that were roll repaired when normal CRD maintenance activities make access to the housing inside diameter (ID) available, as stated in your letters dated January 19, 2001, and July 26, 2002. Provide a response which re-affirms that the statements made in your January 19, 2001, and July 26, 2002, letters will continue to apply. Finally, will the roll repaired CRD housing IDs be accessible during the R20 refueling outage to allow UT examination to verify cracking has not occurred in the repair during the past two cycles?

Response

As discussed in our July 26, 2002 letter, "during subsequent refueling outages, UT examination of the CRD housings that are roll-expansion repaired will be performed when normal CRD maintenance activities make access to the housing ID available." This statement continues to be our commitment to ongoing UT inspection activities. One of the repaired CRD housings may become accessible this outage due to CRD mechanism maintenance and would therefore be subject to a UT of the housing roll repair area.

Question:

4. AmerGen letter 2130-02-20291, dated October 4, 2002, provided additional information to support the approval of the previous roll/expansion repair of the CRD stub tubes until the R20 refueling outage. Specifically, AmerGen stated in this letter that hydrogen injection had been increased to provide at least 0.5 ppm in the feedwater to meet the current criteria for intergranular stress corrosion cracking (IGSCC) mitigation and protection of the recirculation loop piping and the lower plenum down to the CRD stub tubes. Additionally, AmerGen stated that Oyster Creek had continued zinc injection concentration at approximately 7.0 ppb, and that primary system mitigation will be achieved with the Noble Metals Chemical Application planned for the Ri9 refueling outage. Has the primary system mitigation program been implemented in R19 refueling outage? Describe what mitigation and protection programs have been in place since refueling outage R19. Will these programs continue to refueling outage R21 to support the extension of the approval of the alternative repairs to this time?

Response to RAI Concerning Alternative Repair of Control Rod Drive Housing Interface with Reactor Vessel Page 3

Response

Noble Metals Chemical Addition (NMCA) was injected and Hydrogen Water chemistry (HWC) has been in operation since the R19 outage. NMCA was completed during the R19 outage and is expected to provide adequate mitigation until the R21 outage without re-injection. During the INPO-BWRVIP review it was reported that the HWC availability typically exceeds 99%. No additional mitigation measures are required.

Question:

5. Your October 21, 2003, submittal states that you intend to implement the roll-expansion repair technique as described in BWRVIP-17. As stated in an NRC letter dated March 13, 1998, from G. Lainas, NRC, to C. Terry, BWRVIP Chairman, the NRC staff has not endorsed BWRVIP-17 on a generic basis, and therefore we request the following additional information concerning the applicability of this report to Oyster Creek:

5A. On the assumption that the concept of a non-zero reactor coolant pressure boundary (RCPB) leakage-rate is acceptably justified for Oyster Creek, provide justification why the leakage-rate limits specified in Tables 2-3 and 2-4 of BWRVIP-17 after subsequent re-roll/expansion repairs should not be the same as those triggering the initial roll/expansion repair.

Response

New allowable leakage rates are proposed in the attached Tables 1 and 2 for OCGS to be utilized in the upcoming outage inspections. These proposed tables are significantly more conservative as compared to the BWRVIP-17 tables previously utilized at OCGS. Additionally, more reasonable units of measure have also been utilized, i.e., drops per minute as compared to drops per second.

The proposed tables provide significantly more conservative leakage limits, and ensure that the reroll option has been utilized to the point that no substantial improvement in leakage would be obtained. As discussed in response to Question 2, the BWRVIP-58 repair process has not been implemented at a domestic nuclear reactor. We do not plan to have this process ready as a contingency repair plan for the R20 refuel outage this November 2004 due to the significant cost and lack of field experience with this process and the success with the roll repairs at OCGS.

Question:

5B. Tables 2-3 and 2-4 of BWRVIP-17 specify that if a penetration housing has been rolled a second time, a contingency repair action will be implemented if the leakage-rate is found to exceed the specified limits in these tables. On the assumption that the concept of non-zero RCPB leakage-rate is acceptably justified for Oyster Creek, provide justification why a permanent repair (other than rolling) should not be implemented if the leakage-rate exceeds the initial roll/repair leakage-rate.

Response to RAI Concerning Alternative Repair of Control Rod Drive Housing Interface with Reactor Vessel Page 4

Response

The new proposed tables ensure that less leakage is expected each time a penetration is rolled. The BWRVIP-17 tables allowed more leakage each time the penetration was rolled. Therefore, the new tables reflect the condition that the roll must result in less leakage, and are therefore more conservative than the BWRVIP-17 tables.

Question:

5C. Table 2-4 of BWRVIP-17 states in a footnote that, "secondary signs of leakage, such as dried water stains or dried corrosion products around housings and penetrations, do not necessarily require corrective action. Repairs will be considered as part of preventative maintenance as long as the outage schedule permits." Discuss the action that will be taken to determine the source of these secondary signs of leakage and provide the justification for not requiring corrective action.

Response

Procedure ER-AA-335-015 ("VT-2 Visual Examination") provides guidance with regards to VT-2 examinations. A VT-2 exam is performed on the CRD housings in accordance with our commitments. This examination would look for evidence of water or steam leakage such as water or moisture collected or flowing on walls, structures, etc. and especially for puddles on the floor. If any secondary signs would be identified through the VT-2 inspection, additional examination would be performed to determine if active leakage is present. If there are no signs of active leakage, then the secondary signs of leakage will be dispositioned accordingly.

Unlike PWRs, BWR coolant does not typically have any borate. Thus, borate deposits or borate corrosion are not issues for BWRs. The primary consideration and determination factor for roll repair is leakage; therefore, secondary signs are not necessary indicative of leakage and may result from other sources such as previous maintenance or NDE activities.

Question:

5D. BWRVIP-17 recommends that the structural load-carrying capability of the CRD housing, and the prevention of its ejection under accident conditions, be evaluated on a plant-specific basis. Provide the Oyster Creek plant-specific analysis, which should include the following:

i. An evaluation of the bending stresses induced in the housing by the expansion of the rolled section of the housing through the gap between the outside radius of the housing and the inside radius of the stub tube.

Response

A summary of the BWRVIP-17 Safety Evaluation was provided in the original Relief Request (Letter from R. J. DeGregorio (AmerGen Energy Company, LLC) to NRC, dated November 10, 2000). No evaluation of the bending stresses has been performed on a plant specific basis for OCGS. We expect that this evaluation will be performed as part of the white paper to support the code case.

Response to RAI Concerning Alternative Repair of Control Rod Drive Housing Interface with Reactor Vessel Page 5 Question:

ii. An evaluation of the effect of flow-induced vibration of the housing or the instrument nozzle on the integrity of the roll repair. Alternatively, provide the basis for not considering such flow-induced vibration as a loading condition on the roll repair.

Response

No evaluation of the effect of flow-induced vibration of the housing or the instrument nozzle has been performed on a plant specific basis for OCGS. We expect that this evaluation will be performed as part of the white paper to support the code case.

Question:

5E. BWRVIP-17 states that it has been demonstrated by test that control rod insertion capability is not significantly affected by the roll/expansion repair. However, the test fixture used to demonstrate this did not reflect the as-built configuration of a typical CRD housing. CRDs are generally installed on the reactor vessel head with their axes parallel to the axis of the reactor vessel, in a "hillside" configuration. Provide test-based verification that rod insertion ability will not be affected by a roll/expansion repair for CRD's mounted in a "hillside" configuration.

Response

With regards to plant specific test-based verification for OCGS, control rod scram tests are periodically performed for the roll-repaired penetrations to confirm that the roll repair has not affected the control rod scram times.

All control rod support and guide structures are axisymmetric even for hillside penetrations. This includes the roll repair, which is axisymmetric (i.e., rolled approximately mid-thickness in the penetration) with respect to the CRD housing.

Therefore, it does not matter if the insertion test was performed with a fully axisymmetric configuration.

Response to RAI Concerning Alternative Repair of Control Rod Drive Housing Interface with Reactor Vessel Page 6 TABLE 1 CRD Penetration - Allowable Leakage Rates for Inspections During Plant Outages with a Scheduled Duration of Less than or Equal to 7 Days Condition Allowable Leak Rates Repair Actions 800-1100 psig Depressurized Previously Unrolled 80 drops/min. 20 drops/min. Roll Expand to 4% Nominal (Note 1) (Note 1) Wall Thinning Rolled Once 70 drops/min. 15 drops/min. Reroll to 6% Nominal Wall Thinning Rerolled 60 drops/min. 10 drops/min. Implement Contingency Plan TABLE 2 CRD Penetration - Allowable Leakage Rates for Inspections During Plant Outages with a Scheduled Duration of Greater than 7 Days Condition Allowable Leak Rates (Note 2) Repair Actions 800-1100 psig Depressurized Previously Unrolled No Leakage No Leakage Roll Expand to 4% Nominal Wall Thinning Rolled Once 70 drops/min. 15 drops/min Reroll to 6% Nominal Wall Thinning Rerolled 60 drops/min. 10 drops/min Implement Contingency Plan Notes:

1. Will require roll expansion to 4% nominal wall thinning during the next outage with a scheduled duration greater than 7 days.

2. Allowable leak rates in Table 2 are only acceptable for one-cycle. The one-cycle leakage limit commences when the leakage is first detected.