W3F1-2006-0008, Supplement to Amendment Request NPF-38-260 Tubesheet Inspection Depth for Steam Generator Tube Inspections for Waterford, Unit 3

Supplement to Amendment Request NPF-38-260 Tubesheet Inspection Depth for Steam Generator Tube Inspections for Waterford, Unit 3

ML060860331
Person / Time
Site:	Waterford
Issue date:	03/22/2006
From:	Tankersley T Entergy Nuclear South, Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NPF-38-260, W3F1-2006-0008
Download: ML060860331 (12)

Text

Entergy Entergy Nuclear South Entergy Operations, Inc.

17265 River Road Killona, LA 70057-3093 Tel 504-739-6780 Fax 504-739-6698 ttanker@entergy.com Tom Tankersley Acting Director, Nuclear Safety Assurance Waterford 3 Contains 10 CFR 2.390(a)(4) Proprietary Information W3F1 -2006-0008 March 22, 2006 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

REFERENCES:

Supplement to Amendment Request NPF-38-260 Tubesheet Inspection Depth for Steam Generator Tube Inspections Waterford Steam Electric Station, Unit 3 Docket No. 50-382 License No. NPF-38

1.

Entergy letter dated March 15, 2005, License Amendment Request NPF-38-260 Proposed Technical Specification Change Regarding Tubesheet Inspection Depth for Steam Generator Tube Inspections (W3F1 -2005-0009)

2.

Entergy letter dated July 21, 2005, License Amendment Request NPF-38-262 Proposed Technical Specification Change to Waterford-3 Steam Generator Tube Inservice Inspection Program Using Consolidated Line Item Improvement Process (W3F1 -2005-0040)

3.

Entergy letter dated February 15, 2006, Supplement to Amendment Request NPF-38-262 Steam Generator Tube Inservice Inspection Program (W3F1-2006-0007)

Dear Sir or Madam:

By letter (Reference 1), Entergy Operations, Inc. (Entergy) proposed a change to Waterford Steam Electric Station, Unit 3 (Waterford 3) Technical Specifications (TSs) Section 4.4.4.4 to modify the steam generator tube inspection Acceptance Criteria for the "Plugging or Repair Limit" and the "Tube Inspection," as contained in the Waterford 3 Surveillance Requirements 4.4.4.4.a.7 and 4.4.4.4.a.9, respectively. The purpose of these changes was to define the depth of the required tube inspections and to clarify the plugging criteria within the tubesheet region.

On November 17, 2005, Entergy received an NRC Staff Request for Additional Information (RAI) to support the review of the proposed change. Entergy agreed with the NRC Staff on a tentative 90-day response date based on interaction with Florida Power & Light Company's (FPL) St Lucie Unit 2 and the on-going Westinghouse response to their RAI questions related to pO1

W3F1 -2006-0008 Page 2 WCAP-1 6208-P, Revision 0. On February 13, 2006, Entergy and a member of your Staff held a conference call to discuss the status of the RAI response and Waterford-3's need to revise the reply date to March 22, 2006 to incorporate supplemental information from Westinghouse. *The revised reply date was granted as requested. Entergy's response is contained in Attachment 1 with several of the responses referencing information contained in Attachment 2., Responses to NRC Requests for Additional Information on WCAP-1 6208-P, Rev.1, NDE Inspection Length for CE Steam Generator Tubesheet Region Explosive Expansions for Waterford Steam Electric Station Unit 3, LTR-CDME-06-16-P, contains information proprietary to Westinghouse Electric Company LLC. Therefore, it is requested that be withheld from public disclosure in accordance with the provisions of 10 CFR 2.390. Attachment 3 is a nonproprietary copy of Attachment 2 for the public document room. is the supporting Westinghouse affidavit for withholding the proprietary information in Attachment 2. Correspondence with respect to the copyright or proprietary aspects of the!

items listed above or the supporting Westinghouse Affidavit should reference CAW-06-2118 and should be addressed to B.F. Maurer, Acting Manager of Regulatory Compliance and Plant Licensing, Westinghouse Electric Company, P.O. Box 355, Pittsburgh, Pennsylvania 15230.

0355.

Based on Westinghouse's analysis of more conservative tube pull-out forces, a change to the proposed inspection depth from 10.4 inches to 10.6 will be required. The tubesheet inspection depth defined in both the proposed license amendments included in References 1, 2, and 3 will be revised to reflect this new inspection depth. Attachment 5 provides a new marked up TS page for the proposed license amendment regarding tubesheet inspection depth for steam generator tube inspections in Reference 1. These marked up pages replace the pages provided in Attachment 2 of the original submittal (Reference 1) in their entirety. Additionally, the proposed license amendment on the steam generator tube inservice inspection program using consolidated line item improvement process contained two TS markup pages (References 2 and 3) that are also impacted by the change in inspection depth. Attachment 6 provides these two marked up TS pages due to the change in inspection depth. Note that in Attachment 6, changes to marked up TS page 16 of 19 replaces the page provided in Attachment 2 of the original submittal (Reference 2 ) and TS page 17 of 19 replaces the page provided in of the response to an RAI (Reference 3) in their entirety. provides a new commitment to address the examination inspection depth into the tube sheet for sleeved steam generator tubes.

The coiclusions of the original no significant hazards consideration included in References -l and 2 remains bounding and are not affected by this change.

Entergy previously requested approval of the proposed amendment for Reference I by March 1, 2006. However, this response to your Request for Information as well as the RAI response (Reference 3) associated with Entergy's proposed Waterford-3 license amendment (Reference

2) on Steam Generator Tube Inservice Inspection Program will prevent your staff from meeting this requested date. Therefore, Entergy requests NRC approval of the amendment by August 1, 2006 to support implementation plans prior to the next refueling outage scheduled in November 2006.

W3Fl -:2006-0008 Page 3 If you have any questions or require additional information, please contact Ron Williams or Steve Bennett at (504) 739-6255 and 479-858-4626, respectively.

I declare under penalty of perjury that the foregoing is true and correct. Executed on March 22, 2006.

Sincerely, r39 T u rs ey U

TET/Rl W/cbh Attachrnents:

1. Response to Request for Additional Information

2. Proprietary Copy of Westinghouse Report LTR-CDME-06-16-P

3. Nonproprietary Copy of Westinghouse Report LTR-CDME-06-16-NP

4. Westinghouse Affidavit Regarding Proprietary Information

5. Revised Markup of Corrected TS Pages for Reference 1

6. Revised Markup of Corrected TS Pages for References 2 and 3

7. List of Regulatory Commitments

W3FI-:2003-0008 Page 4 cc:

(w/Attachments)

Dr. Bruce S. Mallett Regional Administrator U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064 NRC Senior Resident Inspector Waterford Steam Electric Station Unit 3 P.O. Box 822 Killona, LA 70066-0751 U.S. Nuclear Regulatory Commission Attn: Mr. Mel B. Fields MS O-7E1 Washington, DC 20555-0001 (w/o Attachment 2)

Wise, Carter, Child & Caraway ATTN: J. Smith P.O. Box 651 Jackson, MS 39205 Winston & Strawn ATTN: N.S. Reynolds 1700 K Street, NW Washington, DC 20006-3817 Morgan, Lewis & Bockius LLP ATTN: T.C. Poindexter 1111 Pennsylvania Avenue, NW Washington, DC 20004 Louisiana Department of Environmental Quality Office of Environmental Compliance Surveillance Division P. O. Box 4312 Baton Rouge, LA 70821-4312 American Nuclear Insurers Attn: Library Town Center Suite 300S 29th S. Main Street West Hartford, CT 06107-2445

Attachment I To W3F1 -2006-0008 Response to Request for Additional Information W3FI-2006-0008 Page 1 of 7 Response to Request for Additional Information Question 1:

Throughout the submittal, WCAP-1 6391-P is referenced. To the staffs knowledge, this was never formally submitted to the NRC. Please confirm that the information in WCAP-16391-F is fully ccnsistent with WCAP-1 6208-P, Revision 1, or with WCAP-1 6208-P, Revision 0, as supplemented by the FPL letter dated March 31, 2005. Alternatively, please provide a copy of WCAP-1 6391 -P for the staffs review.

In the following questions, the staff assumes your proposed TS changes, which are based cn WCAP-16208-P, Revision 0, and WCAP-16391-P, are fully consistent with WCAP-16208-P, Revision 1.

Response 1:

The rererence to WCAP-1 6391-P was an error in Entergy's submittal. The referenced document was a Westinghouse response to Florida Power & Light's (FPL) St Lucie Unit 2 associated with NRC Requests for Additional Information (RAI) on WCAP-16208-P, Rev 0.

Westinghouse changed this referenced document to LTR-CDME-05-14 'Responses to NRC Requests for Additional Information on WCAP-1 6208-P, Rev 0, 'NDE Inspection Length for CE Steam Generator Tubesheet Region Explosive Expansions" and transmitted it to Florida Power

& Light Company in February 2005. Westinghouse incorporated LTR-CDME-05-14 RAI responses into revision 1 of the final WCAP-1 6208-P report and issued it in May 2005. This error ir, our submittal was entered into our Corrective Action Program (CR-WF3-2006-0216).

The Waterford 3 proposed TS changes are fully consistent with WCAP-1 6208-P, Revision 1.

Question 2:

Please confirm that your operating parameters will always be bounded by the conditions for which the C* distance was determined in WCAP-16208-P, Revision 1 (e.g. temperature, pressure; etc.). If the conditions will not always be bounded, what controls are in place to ensure an adequate depth of inspection in the tubesheet?

Response 2:

.WCAP-16208-P RI Paragraph 4.3 "Selection of Elevated Temperature" used a conservative test value of 600 F as a conservative approved leak rate temperature. The Waterford 3 existing operating parameters for T-hot bound the conditions for which the C* distance was determined.

At current 100% Core Power, 3716 MWth, the nominal T-hot bulk is 602 F as determined by CECOR (Combustion Engineering Full Core Instrumentation Analysis System) during our surveillance. The T-hot Bulk Temperature during the operating cycle is stable. The maximum leakrate described in WCAP-16208-P RI Paragraph 4.6 "Effect of Temperature and Pressure on Leakrate" is bounded by the leakrate at 2560 psi. Waterford 3 existing operating parameters for primary pressure bound (less than 2560 psi) the conditions for which the C* distance was determined.

W3F1 -2006-0008 Page 2 of 7 Procedure NE-004-006, RCS Flow with COLSS Operable, monthly surveillance determines the T-hot E3ulk Temperature as part of the flow surveillance. The T-hot bulk value determined from this procedure has no acceptance criteria.

Waterford 3 experiences hot leg streaming which causes the RTD T-hot indication at the beginning of the operating cycle to be higher than at the end of cycle by several degrees F for the same reactor power conditions. Additionally, Waterford 3's T-hot RTDs are all mounted on the top of the RCS piping. These conditions are addressed in the monthly RCS flow surveillances by the use of the Hot Leg Delta T Stratification (DELTSTRAT) Program which compensates and determines a T-hot Bulk Temperature.

RCS pressure is maintained at 2250 psia (2175 to 2265 psia) per OP-010-004 Step 9.1.1.2.

Changes to the primary operating conditions require full analysis that is performed normally under our reload process for each operating cycle. The impact of a change to the operating temperature would have to be reviewed under the 10 CFR 50.59 process.

Questilon 3:

Please discuss the expected condition of the tube-to-tubesheet joint. For example, discuss the amount of corrosion expected at the top of the tubesheet (similar to what may have been present in some of the test specimens) and whether there is sludge buildup at the top of the tubesheet.

Response 3:

The tube-to-tubesheet joint of the Waterford 3 Steam Generators (SG) is expected to be consistent with or exceed the extent of corrosion that may have been present in the test specimens.

• Waterford 3 has conducted two SG chemical cleanings to reduce the amount of corrosion products within the SG. The chemical cleanings were performed in RF10 (Fall 2000) and RF12 (Fall 2003).

Waterford 3 also conducts periodic sludge lancing, SG blowdown operation and chemistry controls to minimize and remove sludge.

However, the SGs still contain sludge pile heights of up to 3 inches, as documented in the plant's RF13 Field Summary Report on Sludge Height Maps.

Question 4:

The letter dated March 15, 2005, compares the Nuclear Energy Institute (NEI) report 97-06 primary-to-secondary accident-induced leakage limit to the 720 gallon per day (gpd) operational leakage limit in TS 3.4.5.2. Since you are comparing the NEI accident-induced leakage limit to your TS operational leakage limit, the staff assumes that, at the time this application was W3Fl-2006-0008 Page '3 of 7 submitted, your operational leakage limit was the same as your accident induced leakage limit.

Please! confirm the staffs understanding.

The lel:ter dated March 15, 2005, discusses a change in the assumed accident-induced leakage rate from 720 gpd to 540 gpd. The staff understands this to mean that, although the accidelt-induced leakage rate in the licensing basis was 720 gpd at the time the C* amendment was submitted, the accident analyses was in the process of being revised in support of your extended power uprate and alternative source term amendments. This revised analyses would require! that you limit the amount of accident-induced leakage to 540 gpd. Please confirm the staffs understanding.

The staff notes that your current TS operational leakage limit is 75 gpd. Assuming 540 gpd (0.375 gallons per minute (gpm)) is your current accident-induced leakage limit, it is the staffPs understanding that no more than 0.275 gpm could come from sources other than implementation of C* (since implementation of C* assumes that accident-induced leakage i; 0.1 gprn). Other sources could include sleeves, plugs, and other flaws in the SG. Please confirm the staff's understanding.

Assuming (1) you were to operate at your TS operational leakage limit of 75 gpd (0.05 gpm), (2) that none of the operational leakage was a result of implementation of C*, and (3) that there!

was no accident-induced leakage expected from other sources, it is the staffs understanding that you would continue to have margin to your accident-induced leakage limit even after accounting for the increase in the amount of operational leakage, as a result of the higher differential pressures associated with various postulated accident conditions. Please confirm the staff's understanding.

Response 4:

The staff's understanding is correct. Waterford 3 has established an operational leakage value of 75 gpd per TS 3.4.5.2.c with the issuance of License Amendment 199 on April 15, 2005.

Additionally, for accident analyses with a faulted Steam Generator, the faulted SG is assumed to have a 540 gpd primary-to-secondary leakage. This is described in the TS Bases section 3/4.4.5.2, Reactor Coolant System Operational Leakage, page B3/4 4-4e via Change 38.

Although not analyzed, the staffs position concerning sufficient margin to the accident-induc~ed leakaga of 540 gpd for operation at the operational leakage of 75 gpd can be confirmed by performing the following simplistic analysis. The change in leakage is expected to be proportional to the square root of the differential pressure. Under operational conditions the nominal pressure difference is 2250 psi - 800 psi =1450 psi. Under accident conditions (MS'LB) the nominal pressure difference is 2250 psi - 0 = 2250 psi. Ratio of the differential pressures is 2250/1450 = 1.55. The square root of the ratio is 1.25. The nominal expected flow increase would be 75 gpd

• 1.25 = 93.75 gpd which is well below 540 gpd.

Waterford 3 has sufficient margin between the accident-induced leakage and operational leakage.

W3Fl-2006-0008 Page 4 of 7 Question 5:

Please clarify whether the load at first slip was reported and plotted in Figures 5-1 through 5-3 of WCAP-1 6208-P, Revision 1, or whether the maximum load was plotted. If the load at first slip was not used in all cases, please discuss the effect on the required inspection distance if the load at first slip was used.In addition, if the load at first slip was not used in Table 6-8 of WCAP-16208-P, Revision 1 ("Burst Based Inspection Length"), please provide Table 6-8 values to confirm that the 10.4 inch proposed inspection distance is still bounded when the most limiting specimen is evaluated using load at first slip.

Response 5:

The response is provided in Attachment 2, LTR-CDME-06-16-P.

Question 6:

Please discuss your plans to revise your TS to include the reporting requirements listed below.

(a) Number of total indications, location of each indication, orientation of each indication, severity of each indication, and whether the indications initiated from the inside or outside surface.

(b) The cumulative number of indications detected in the tubesheet region as a function of elevation within the tubesheet.

(c) ProJected end-of-cycle (EOC) accident-induced leakage from tubesheet indications. This leakage shall be combined with the postulated EOC accident-induced leakage from all other source3. If the preliminary estimated total projected EOC accident-induced leakage from all sources exceeds the leakage limit, the NRC staff shall be notified prior to unit restart.

Response 6:

The Waterford 3 existing TS requirements do not contain the majority of the "reporting requirements" listed in your question. The existing Waterford 3 TS reporting requirements include the following information:

1. Number and extent of tubes inspected.

2. Location and percentage of wall-thickness penetration for each indication of an imperfection.

3. Identification of tubes plugged or sleeved.

In addition, Waterford 3's SG tube inspection 12-Month Special Report also contains the following additional information:

4. Scope of inspections performed.

5. Active degradation mechanisms found.

6. NDE techniques utilized for each degradation mechanism.

7. Repair methods and tubes repaired by each method.

W3Fl-2006-0008 Page 5S of 7

8. Total number/percentage of tubes plugged and/or repaired to date and plugged percentage in each SG.

9. Tube integrity assessment.

10. Corrective actions implemented, if any.

11. Evaluation of circumstances if monitoring results exceeded previous cycle operational assessment.

Entergy has submitted the proposed license amendment NPF-38-262 Proposed Technical Specification Change to Waterford-3 Steam Generator Tube Inservice Inspection Program Using Consolidated Line Item Improvement Process, as stated in Reference 2, that is consistent with the NRC reviewed and approved TSTF-449, Revision 4.

The TESTF-449 reporting requirements are performance based. These reporting requirements remove the burden of unnecessary reports from both the NRC and the licensee, while ensuring that critical information related to problems and significant tube degradation is reported more completely and, when required, more expeditiously than under the current technical specifications. These reporting requirements are specified as follows:

STEAM GENERATOR TUBE INSPECTION REPORT 6.9.1.5 A report shall be submitted within 180 days after the initial entry into Hot Shutdown following completion of an inspection performed in accordance with the Specification 6.5.9, Steam Generator (SG) Program. The report shall include:

a.

The scope of inspections performed on each SG,

b.

Active degradation mechanisms found,

c.

Nondestructive examination techniques utilized for each degradation mechanism,

d.

Location, orientation (if linear), and measured sizes (if available) of service induced indications,

e.

Number of tubes plugged or repaired during the inspection outage for each active degradation mechanism,

f.

Total number and percentage of tubes plugged or repaired to date,

g.

The results of condition monitoring, including the results of tube pulls and in-situ testing,

h.

The effective plugging percentage for all plugging and tube repairs in each SG, and

i.

Repair method utilized and the number of tubes repaired by each repair method."

Waterford 3 will continue to submit the SG tube inspection 12 month special report containing the exiling TS reporting requirements (1 thru 3) and the additional information (4 thru 11) unfil the NRC Staff completes their review for approval of the proposed TS change submitted via Reference 2.

Question 7:

In WCAP-16208-P, Revision 1, it is not clear whether all of the available data were used to support the analytical adjustment to account for the axial load resistance provided by interns I pressure. For example, specimens 8 and 12 from the Task 1154 program were run at room temperature with internal pressure; however, an analysis of this data (similar to what was done for the elevated temperature data point) was not provided. Please evaluate all data in which internal pressure (above ambient pressure) was applied to support the basis for the analytical adjustments to account for the internal pressure. With respect to the analysis of the pressure

Attach nent 1 W3Fl-2006-0008 Page E6 of 7 effects, please provide additional details on how the axial force resistance due to the internal pressure of 1435 pounds per square inch was calculated and discuss how the effect of the residual contact pressure was taken into account in your analysis. (The actual pullout force was nearly the same as the pullout resistance expected analytically from the internal pressure effects. As a result, if the residual contact pressure was not included in this assessment, it would appear that the analytical adjustments for internal pressure are too high.)

Response 7:

The response is provided in Attachment 2, LTR-CDME-06-16-P.

Question 8:

It is the NRC staff's understanding that not all data was included in Appendix B of WCAP-16208-*P, Revision 1 (i.e., some data was not included since it was well outside the targeted temperatures and pressures). It is also the staff's understanding that some data in Appendix B was not included in Table 4-1 of WCAP-16208-P, Revision 1 (which was used in determining the leak rate as a function of joint length). Please confirm the staffs understanding and discuss the basis for not including all of the Appendix B data in Table 4-1. For example, was data from Appendix B not included in Table 4-1 when steady state was never reached although the temperatures and pressures were within the desired range?

Response 8:

The response is provided in Attachment 2, LTR-CDME-06-16-P.

Question 9:

The Waterford 3 TS (4.4.4.4.b) currently allow installation of leak-tight sleeves according to CENS Report CEN-605-P. Since sleeves could extend into the tubesheet below the C*

distance, the proposed TS would not require an inspection of this portion of the sleeve (including the lower sleeve joint). Sleeves were not addressed in the testing and analysis to justify excluding part of the tube from inspection (WCAP-16208-P, Revision 1). What plans do you have to ensure the lower ends of sleeves (i.e., those within the tubesheet below the C*

distance) will be inspected?

Response 9:

Waterford-3 steam generators do not presently have sleeves installed. However, if sleeves are installed, Entergy plans to inspect inservice sleeves over their full length plus 5 inches beyond the sleeve-to-tube rolled joint in the tube sheet in accordance with the requirements of the EPRI Guidelines using appropriate examination methodology. The tube shall be plugged upon detection of any service induced imperfection, degradation or defect in the sleeve or pressure boundary portion of the original tube wall in the sleeve-to-tube rolled joint. Entergy will periodically inspect sleeves as a minimum in accordance with the existing TS requirements.

Entergy has submitted a proposed license amendment on the steam generator tube inservice inspection program using the consolidated line item improvement process (Reference 2 and 3) that implements TSTF-449 R4 requirements.

W3F -2006-0008 Page A" of 7 A prior NRC regulatory commitment was made in support of NRC Staffs issuance of license amendment 117 dated December 14, 1995 to use welded tube sleeves This commitment required Waterford 3 to perform baseline examinations and periodic examinations per the EPRI "PWR Steam Generator Examination Guidelines" using the state-of-the-art examination methodology. This commitment remains. The approval of the proposed license amendment to implement TSTF-449 (Reference 2 and 3) will continue to require the use of the EPRI Guidelines as described in NEI 97-06. The examination guidelines provide sufficient requirements for sleeve examination.