ML11312A245

Response to Request for Additional Information License Renewal Application

ML11312A245
Person / Time
Site:	Columbia
Issue date:	11/04/2011
From:	Swatzke B Energy Northwest
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
G02-11-177, TAC ME3058
Download: ML11312A245 (9)

Text

Bradley J. Sawatzke ENERGY Columbia Generating Station P.O. Box 968, PE08 Richland, WA 99352-0968 Ph. 509.377.4300 1 F. 509.377.4150 bjsawatzke@energy-northwest.com November 4, 2011 G02-11-177 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001

Subject:

COLUMBIA GENERATING STATION, DOCKET NO. 50-397 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION LICENSE RENEWAL APPLICATION

References:

1) Letter, G02-1 0-11, dated January 19, 2010, WS Oxenford (Energy Northwest) to NRC, "License Renewal Application"

2) Letter dated February 3, 2011, NRC to SK Gambhir (Energy Northwest),

"Request for Additional Information for the Review of the Columbia Generating Station, License Renewal Application, for Metal Fatigue (TAC NO, ME3058" (ADAMS Accession No. ML110240426)"

3) Letter, G02-11-165, dated October 6, 2011, DA Swank (Energy Northwest to NRC, "Response to Request for Additional Information, License Renewal Application"

4) Letter dated February 3, 2011, NRC to Energy Northwest, "Summary of Telephone Conference Call held on January 20, 2011, Between the U.S.

Nuclear Regulatory Commission and Energy Northwest, Concerning the Request for Additional Information Pertaining to the Columbia Generating Station, License Renewal Application (TAC NO, ME 3058),"

(ML110240202)

Dear Sir or Madam:

By Reference 1, Energy Northwest requested the renewal of the Columbia Generating Station (Columbia) operating license.

A request for additional information (RAI) was transmitted to Energy Northwest via Reference 2. Reference 3 provided the initial response to the RAI. As indicated in Reference 3, this letter provides additional information. Rather than providing the additional information to reflect the revised calculations, the RAI 4.3-09 response in the attachment replaces the earlier response provided in Reference 3.

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION LICENSE RENEWAL APPLICATION Page 2 of 2 The initial analyses of additional locations for limiting cumulative usage factor (CUF) were provided to the NRC in Reference 3. In Reference 3, Energy Northwest indicated that initial calculations had shown four locations with a CUF greater than 1.0. Using refined analytical techniques, Energy Northwest has established a CUF of less than 1.0 for these locations. A new Table 4.3-7, to show the CUF for the locations beyond those discussed in NUREG/CR-6260, Application of NUREG/CR-5999 Interim Fatigue Curves to Selected Power Plant Components, is provided in the enclosure as Amendment 46.

Additionally a correction to table 4.3-6 is provided in Amendment 46.

No new or revised commitments are included in this response.

If you have any questions or require additional information, please contact Abbas Mostala at (509) 377-4197.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the date of this letter.

Respectfully, BJ Sawatzke Vice President, Nuclear Generation & Chief Nuclear Officer

Attachment:

Response to Request for Additional Information

Enclosure:

License Renewal Application Amendment 46 cc: NRC Region IV Administrator NRC NRR Project Manager NRC Senior Resident Inspector/988C EFSEC Manager RN Sherman - BPA/1 399 WA Horin - Winston & Strawn AD Cunanan - NRC NRR (w/a)

John Daily - NRC NRR (w/a)

BE Holian - NRC NRR MA Galloway - NRC NRR RR Cowley - WDOH

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION LICENSE RENEWAL APPLICATION Attachment Page 1 of 2 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION "Request for Additional Information for the Review of the Columbia Generating Station, License Renewal Application,"

(ADAMS Accession No. ML110240426)

RAI 4.3-09 Background and Issue:

In the response to RAI 4.3-06, dated November 11, 2010, the applicant provided the basis for not selecting the core shroud supports, main steam shell nozzles, and LPCI nozzle thermal sleeves as additional environmentally-assisted fatigue analysis locations. The staff noted that the applicant's plant-specific configuration may contain additional locations (including but not limited to those provided in LRA Tables 4.3-3 and 4.3-5) that may need to be analyzed for the effects of the reactor coolant environment other than those identified in NUREG/CR-6260. This may include locations that are limiting or bounding for the applicant's particular plant-specific configuration, or that have calculated environmentally-adjusted CUF values that are greater than those calculated by the applicant for locations that correspond to those identified in NUREG/CR-6260.

Request:

Confirm and justify that the locations selected for environmentally-assisted fatigue analyses in LRA Table 4.3-6 consists of the most limiting locations for the plant (beyond the generic components identified in the NUREG/CR-6260 guidance). If these locations are not bounding, clarify the locations that require an environmentally assisted fatigue analysis and the actions that will be taken for these additional locations. If the limiting location identified consists of nickel alloy, state whether the methodology used to perform the environmentally-assisted fatigue calculation for nickel alloy is consistent with NUREG/CR-6909. If not, justify the method chosen.

Response

This response completely replaces the response to RAI 4.3-09 provided to the NRC in letter, G02-11-165, dated October 6, 2011, DA Swank (Energy Northwest) to NRC, "Response to Request for Additional Information, License Renewal Application".

However, Amendment 45 submitted under letter G02-11-165 is still valid.

The locations selected for environmentally-assisted fatigue (EAF) analysis are presented in the License Renewal Application (LRA) Table 4.3-6 for the locations that correspond to NUREG/CR-6260. The nickel-alloy components in Table 4.3-6 were calculated consistent with NUREG/CR-5704, Effect of LWR Coolant Environment on the Fatigue Design Curves of Austenitic Stainless Steel. This approach is consistent with revision 1 of NUREG-1800. The environmentally-assisted highest cumulative usage factor (CUF) for the locations beyond-NUREG/CR-6260 locations are provided in a new

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION LICENSE RENEWAL APPLICATION Attachment Page 2 of 2 LRA Table 4.3-7, provided as Amendment 46 in the enclosure. Consistent with revision 2 of NUREG-1800, the CUF for the nickel-alloy location in this table was calculated consistent with the methods of NUREG/CR-6909, Effect of LWR Coolant Environment on the Fatigue life of Reactor Materials. All calculated values are less than 1.0.

Selection of beyond-NUREG/CR-6260 locations Since the EAF locations selected for analysis in the LRA Table 4.3-6 do not necessarily contain the limiting locations for the plant, Energy Northwest performed additional analyses, discussed below.

The locations listed in LRA Tables 4.3-3 and 4.3-5 were selected by reviewing the design stress reports for all of the class 1 systems and identifying those locations with the highest CUF in air. The additional locations (beyond-NUREG 6260 locations) were selected based on the highest cumulative usage locations listed in LRA Table 4.3-3 (for reactor vessel locations) and Table 4.3-5, (for reactor pressure boundary piping and piping components). This selection addressed the highest CUF, by system, from the design stress reports. The selection provides assurance that limiting fatigue usage locations in air were evaluated for the impact of environment. As discussed in a conference call with the staff on January 20, 2011, locations in the tables that met any of the following criteria were screened out:

o Exposed to air or dry steam environment o Non-pressure retaining items o Vessels internals Non-wetted locations, such as those only exposed to dry steam, were not evaluated for EAF. Non-wetted locations are in the upper reactor vessel above the water seal skirt and are only exposed to steam that has been dried by passing through the moisture separator and steam dryer. Non-pressure retraining items, such as thermal sleeve extensions were not evaluated. The vessel internals aging management is addressed in the Boiling Water Reactor Vessel Internals Program (BWRVIP).

The review and associated calculations (for the beyond-NUREG/CR-6260 locations) looked at every class 1 system and included a variety of materials and locations in Tables 4.3-3 and Table 4.3-5. Some locations in these tables were not evaluated because there was a similar bounding location. For example, Reactor Feed Water (RFW) piping line B is bounded by RFW piping line A.

Locations as discussed above, which did not screen out, were evaluated and results are tabulated in LRA Table 4.3-7.

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION LICENSE RENEWAL APPLICATION Enclosure Page 1 of 1 LICENSE RENEWAL APPLICATION AMENDMENT 46 Section Page RAI Number Number Number Table of Contents Contents xxxvii 4.3-09 Table 4.3-6 4.3-16 4.3-09 Table 4.3-7 4.3-16a 4.3-09 Table 4.3-7 4.3-16b 4.3-09

Table 4.3-7 CUFs for components beyond NUREG/

CR-6260 locations .................. 4.3-16a Columbia Generating Station License Renewal Application TABLE OF CONTENTS TABLES Table Pagqe Number Table 4.3-4 CUFs for Reactor Vessel Internals .............................................. 4.3-8 Table 4.3-5 CUFs for Reactor Pressure Boundary Piping and Piping C om ponents .................................................................. 4.3-11 able 4.3-6 CUFs for NUREG/CR-6260 Locations ....................................... 4.3-15 Table A-1 Columbia License Renewal Commitments ................................... A-42 Table B-1 Correlation of NUREG-1801 and Columbia Aging Management P rogram s ................................................................................ . . B -11 Table B-2 Consistency of Columbia Aging Management Programs with NUR EG -1801 ........................................................................... B-19 Table C-1 BWRVIP-1 8-A, BWR Core Spray Internals Inspection and Flaw Evaluation Guidelines ....................................................... C-5 Table C-2 BWRVIP-25, BWR Core Plate Inspection and Flaw Evaluation G uidelines ................................................................................ . . C -7 Table C-3 BWRVIP-26-A, BWR Top Guide Inspection and Flaw Evaluation G uidelines .................................................................... C -9 Table C-4 BWRVIP-27-A, BWR Standby Liquid Control System/Core Plate DP Inspection and Flaw Evaluation Guidelines .............. C-1i1 Table C-5 BWRVIP-38, BWR Shroud Support Inspection and Flaw Evaluation Guidelines ............................................................. C-1 3 Table C-6 BWRVIP-41, BWR Jet Pump Assembly Inspection and Flaw Evaluation Guidelines ............................................................. C-15 Table C-7 BWRVIP-42-A, LPCI Coupling Inspection and Flaw Evaluation Guidelines ............................................................. C-17 Table C-8 BWRVIP-47-A, BWR Lower Plenum Inspection and Flaw Evaluation G uidelines .................................................................. C -20 Table of Contents Page xxxvii iAmendmenti46 24

Columbia Generating Station License Renewal Application Technical Information Table 4.3-6 (continued)

CUFs for NUREG/CR-6260 Locations Columbia Material Revised Per NUREGICR-5704 and NUREGICR-6583 NUREG/CR-6260 generic plant-specific CUF in locations type (2) Min. Average Max. Environmentally locationsairFen_(3) Fen (3) Fen(3) assisted CUF Core spray line reactor 4 vessel nozzle and LPCS piping CS 0.155 1.74 5.22 7.33 0.809 associated Class 1 piping Core spray line reactor 4 vessel nozzle and HPCS piping CS 0.321 1.74 2.25 2.49 0.723 1 associated Class 1 piping Residual Heat Removal RHRILPCI Nickel 5 (RHR) nozzles and 0.139 2.55 6.16 6.94 0.856 associated Class 1 piping nozzle safe end Alloy Residual Heat Removal RHR/LPCI 5 (RHR) nozzles and nozzle safe end CS 0.190 1.74 2.39 2.75 0.455 associated Class 1 piping extension Residual Heat Removal 5 (RHR) nozzles and RHR/LPCI piping CS 0.001 20.49 20.49 20.49 0.02 associated Class 1 piping 6 Feedwater line Class 1 I piping RFW/RWCU Tee (1)

CS 1____1.4.8 0 210 1.74 1.85 2.85 2.85_ I 0389 3 Note: CS is carbon steel, LAS is low alloy steel, SS is stainless steel '\'-0"097 (1) Assumed NWC dissolved oxygen concentration equaled to 150 ppb for the RFW nozzle and RFW/RWCU Tee Fe,, calculation.

r-, ^frcrd

.= prFv8Y 9W !*, ntfc Table 4.3 3 and Table 4.3 5.

Efeciv F, ee la arbased on a tirne weighted average for HWC and NWC for 60 years of operation.

Average F,, is the reported environmen 'a Faen divided by the non-environmentally assisted CUF.

Replace footnote 2 with the following footnote:

(2) The "Revised CUF in air'is the maximum computed CUF (in air) for the wetted surface of interest for the evaluation of the effect of the reactor water environment. The CUF of record was previously identified in Table 4.3-3 and Table 4.3-5.

Time-Limited Aging Analyses Page 4.3-16 januaw--.3 Amendment 46 Ilnsert A from pages 4.3-16a through 4.3-16b I

Columbia Generating Station License Renewal Application Technical Information Insert A:

Table 4.3-7 CUFs for components beyond NUREG/CR-6260 locations LRA Table 4.3-3 CGS T or 4.3-5 Specific Component 60-year Uair Fen Environmentally Component Location Core DP Cell Stub Tube NiCrFe 0.218 2.259 0.494 Forging LAS 0.0008 3.979 0.005 HPCS Core Spray 0.0008 2.455 Nozzle Safe End 0.10431 3.584 Extension 0.00631 1.740 CRD Return Nozzle Forging LAS 0.093 3.565 0.330 Safe End CS 0.162 2.527 0.410 Min 2 = 2.4 FW Nozzle Forging LAS 0.0398 Max = 5.34 0.140 RHR/LPCI Nozzle Forging LAS 0.001 10.51 0.0103 RRC Inlet Nozzle Forging LAS 0.0351 4.363 0.153 RRC Outlet Nozzle Cladding SS 0.00487 12.902 0.063 Vessel Head Spray Nozzle LAS 0.0013 3.106 0.004 RFW Piping Line A CS 0.284 Min = 1.0 0.385

______________Max = 1 .897038 Bounded by RFW Piping Line B CS RFW Line A Calculation 1 For event group 1 and 2 2 Highest and lowest Fen for multiple load pairs Time-Limited Aging Analyses Page 4.3-16a Amendment 46

Columbia Generating Station License Renewal Application Technical Information LRA Table 4.3-3 CGS Environmentally or 4.3-5 Specific Material 60-year Uair Fen assisted CUF Component Location RWCU Piping CS 0.164 Max= 4.266 0.193 RCIC Piping CS Dry steam environment -

No environmental effects RPV Head Spray Piping CS 0.259 1.74 0.451 RPV Vent to MS Piping CS Dry steam environment -

RPVVenttoMSPipingCSNo environmental effects RPV Level Condensing Pot SS 0.245 2.547 0.624 Min' = 1.0 SLC Piping CS 0.424 Max = 1.74 0.737 RPV Head Spray Zone 1 Check Valve CS 0.386 2.439 0.941 Zone 2 0.331 2.503 0.828 HPCS/LPCS Valve CS 0.326 1.74 0.558 3 Highest and lowest Fen for multiple load pairs Analyses Aging Analyses Page 4.3-16b Amendment 46 Time-Limited Aging Time-Limited Page 4.3-16b Amendment 46