ML11354A097

Response to Request for Additional Information on License Renewal Application

ML11354A097
Person / Time
Site:	Columbia
Issue date:	12/14/2011
From:	Javorik A Energy Northwest
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
G02-11-195
Download: ML11354A097 (22)

Text

Alex L. Javorik ENERGY ColumbiaP.O. Box 968,Station Generating PE04 Richland, WA 99352-0968 Ph. 509-377-8555 F. 509-377-2354 aljavorik@ energy-northwest.com December 14, 2011 G02-11-195 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 April 22, 2011, DA Swank (Energy Northwest) to NRC, "Response to Request for Additional Information License Renewal Application," (G02-11-084)

3) Letter dated September 26, 2011, NRC to DA Swank (Energy Northwest), "Request for Additional Information for the Review of the Columbia Generating Station, License Renewal Application," (ADAMS Accession No. ML11269A014)

4) Letter dated November 1, 2011, DA Swank (Energy Northwest) to NRC, "Request for Additional Information License Renewal Application, (G02-11-175)

Dear Sir or Madam:

By Reference 1, Energy Northwest requested the renewal of the Columbia Generating Station (Columbia) operating license. Via Reference 2, Energy Northwest provided a response to a Nuclear Regulatory Commission (NRC) Request for Additional Information (RAI) regarding upper shelf energy (USE). Reference 3 requested additional information to support NRC review of the response provided via Reference 2. Via Reference 4, Energy Northwest provided a response to this NRC request for additional information. This response contained proprietary information. Based on General Electric-Hitachi Nuclear Energy (GEH) request that this information provided via Reference 4 be withheld from public disclosure in accordance with 10 CFR 2.390(a)(4),

similar information provided via Reference 2 should be redacted.

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION LICENSE RENEWAL APPLICATION Page 2 of 2 Transmitted herewith is a revised version of the letter supplied in Reference 2 with the proprietary information redacted with the understanding that this revised letter will be used in place of the original letter (Reference 2) in the ADAMS Public Documents.

Also, please note that the proprietary information being redacted is contained in two locations of the draft Safety Evaluation Report (SER). Therefore, Energy Northwest requests that this proprietary information be removed from SER section 1.5 discussion of 0I 4.2-1 and section 4.2.2.2.

No new or revised commitments are included in this letter.

If you have any questions or require additional information, please contact John Twomey at (509) 377-4678.

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

Respectfully, AL Javonik Vice President, Engineering

Enclosure:

Letter dated April 22, 2011, DA Swank (Energy Northwest) to NRC, "Response to Request for Additional Information License Renewal Application," (G02-11-084), Redacted version 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)

MA Galloway - NRC NRR RR Cowley - WDOH

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION LICENSE RENEWAL APPLICATION Enclosure Page 1 of 1 G02-11-084 "Response to Request for Additional Information License Renewal Application" Redacted Version

N0- David A. Swank ENERGY Acting Vice President, Engineering NORTHW ESTRichland, P.O. Box 968, Mail Drop PE04 WA 99352-0968 Ph. 509-377-2309 F. 509-377-4173 daswank@energy-northwest.com April 22, 2011 G02-11-084 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, G02-10-11, dated January 19, 2010, WS Oxenford (Energy Northwest) to NRC, "License Renewal Application"

2) Letter, G02-11-031, dated January 28, 2011, Energy Northwest to NRC "Response to Request for Additional Information License Renewal Application"

3) Letter dated March 23, 2011, NRC to SK Gambhir (Energy Northwest),

"Request for Additional Information for the Review of the Columbia Generating Station, License Renewal Application," (ADAMS Accession No. ML110630360)

Dear Sir or Madam:

By Reference 1, Energy Northwest requested the renewal of the Columbia Generating Station (Columbia) operating license. Via Reference 3, the Nuclear Regulatory Commission (NRC) requested additional information related to Energy Northwest submittal, Reference 2, an earlier response to a request for additional information.

Transmitted herewith in the Attachment is the Energy Northwest response to the Request for Additional Information (RAI) contained in Reference 3. Enclosure 1 contains Amendment 33 to the Columbia License Renewal Application (LRA). One revised commitment is included in this response.

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION LICENSE RENEWAL APPLICATION Page 2 of 2 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.

Respectully, L DA Swank Acting Vice President, Engineering

Attachment:

Response to Request for Additional Information

Enclosure:

License Renewal Application Amendment 33 cc: NRC Region IV Administrator NRC NRR Project Manager NRC Senior Resident Inspector/988C EFSEC Manager RN Sherman - BPA/1399 WA Horin - Winston & Strawn AD Cunanan - NRC NRR (w/a)

BE Holian - NRC NRR RR Cowley - WDOH

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION LICENSE RENEWAL APPLICATION Page 1 of 8 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION "Request for Additional Information for the Review of the Columbia Generating Station, License Renewal Application,"

(ADAMS Accession No. ML110630360)

RAI 4.2-3

Background:

The Columbia Generating Station (Columbia) reactor vessel (RV) N12 water level instrument nozzles (N12 nozzles) are in the beltline region of the RV because they are projected to experience neutron fluence greater than 1 x 1017 n/cm 2 (E > 1.0 MeV) at the end of the period of extended operation, corresponding to 54 effective full power years (EFPY). Based on its January 28, 2011, RAI response, the staff determined that the applicant had provided an acceptable adjusted nil-ductility reference temperature (ART) evaluation for the N12 nozzles that is valid through 54 EFPY. The staff noted that the N12 nozzles from Heat No. 219972 are the limiting RV beltline material because the 54 EFPY ART value (149 OF) exceeds that for all other RV beltline materials. Furthermore, based on the bounding values for copper and nickel content, the 40 OF initial reference temperature nil-ductility transition (RTNDT) value and the 34 OF margin term value, the staff determined that the N12 nozzles from Heat No. 219972 are the limiting RV beltline material, with respect to ART, for all fluence values greater than or equal to 1 x 1017 n/cm 2 (E > 1.0 MeV).

Therefore, based on the limiting nature of the N12 nozzles, the staff identified a concern regarding the impact of these nozzles on the current 33.1 EFPY pressure-temperature (P-T) limit curves in the Columbia Technical Specifications (TSs). This concern was based on the fact that the P-T curve calculations documented in General Electric (GE)

Report NEDO-33144, "Pressure-Temperature Curves for Energy Northwest, Columbia,"

April 2004, did not account for the limiting 33.1 ART value for the N12 nozzle in establishing the current TS P-T limit curves for 33.1 EFPY.

However, the GE-Hitachi Report NEDO-33178-A, "GE Hitachi Nuclear Energy Methodology for Development of Reactor Pressure Vessel Pressure-Temperature Curves," Appendix J, "Water Level Instrumentation Nozzle LEFM [linear elastic fracture mechanics] Evaluation," June 2009, was reviewed and approved by the staff by letter dated April 27, 2009. This report documents a LEFM evaluation of the water level instrument nozzles in all boiling-water reactors (BWRs) based on bounding assumptions for RV and water level instrument nozzle geometry, postulated flaw configuration, operating pressures, and thermal transients. The report documents calculations of Mode I applied stress intensity factors (K1) due to pressure loads and thermal transients.

The report also calculates bounding "T-RTNDT" values for the BWR water level instrument nozzle using the acceptance criteria for total applied K, values (including safety factors) that are based on the lower bound of the static critical (or reference)

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION LICENSE RENEWAL APPLICATION Page 2 of 8 stress intensity factor, KIc, curve, as specified in the ASME Code,Section XI, Appendix G.

Issue:

The staff noted that the results of the LEFM analysis documented in NEDO-33178-A, Appendix J could be used to calculate P-T limit curves specifically for Columbia's N12 instrument nozzles. However, in order to determine how these methods can be applied for determining P-T limits specifically for Columbia's N12 nozzles, the staff determined that the applicant should provide additional information concerning the plant-specific applicability of the postulated flaw configuration used for calculating the applied K, values, as described in the subject report.

The staff also noted that the NEDO-33178-A, Appendix J analysis postulated a 2.276 inch flaw that originates at the blend radius of the instrument nozzle and extends through the nozzle into the adjacent RV shell plate. The tip of the postulated flaw in this analysis is apparently located in the adjacent RV shell plate. Accordingly, Section 5.0 of NEDO-33178-A, Appendix J states that for BWR instrument nozzles located in the beltline region of the RV, "the ART from the adjacent [reactor pressure vessel] RPV shell material is used to create a component-specific P-T curve."

Request:

a. State whether the 2.276 inch postulated flaw for the N1 2 nozzle, as described in the subject report, extends into and terminates in the adjacent RV shell plate material.

b. Ifthe postulated flaw described in the report terminates in the adjacent RV shell plate material, identify the RV beltline shell plate material that surrounds this nozzle and the ART value used for determining the component-specific P-T limits for the N12 nozzles.

Energy Northwest Response:

a. The 2.276 inch postulated flaw assumed in the NEDO-33178-A, App. J, Water Level Instrument (N1 2 for Columbia) nozzle analysis, does extend into and terminates in the adjacent reactor vessel shell plate material.

b. The limiting reactor vessel shell plate material adjacent to the N12 nozzle is heat C1 336-1 and the corresponding adjusted reference temperature (ART) is 44 deg.

F (for current license P-T curves).

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION LICENSE RENEWAL APPLICATION Page 3 of 8 RAI 4.2-4

Background:

In the applicant's RAI response by letter dated January 28, 2011, calculations of the 54 EFPY upper-shelf usage (USE) values for the N6 residual heat removal (RHR)/Iow pressure coolant injection (LPCI) nozzle forgings (SA-508, Class 2) were provided.

These calculations are based on an initial USE value of 70 ft-lbs.

Issue:

The applicant did not provide a basis to the 70 ft-lbs initial USE value. The staff believes that initial USE value based on a lower bounding value of Charpy USE test data for SA-508, Class 2 forging material is conservative.

Request:

Provide a basis the 70 ft-lbs initial USE value and justify that the value is based on the lower bounding value of available Charpy USE test data for SA-508, Class 2 forging material, or other means.

Energy Northwest Response:

The initial (pre-irradiated) upper shelf energy (USE) value of 70 ft-lb is based on Columbia's review of available data from Electric Power Research Institute (EPRI) and NRC databases for SA-508 Class 2 forging material and previous utility submittals related to SA-508 Class 2 material. The databases are referenced in Altran Technical Report 96124-TR-01, "N-1 6 Nozzles Upper Shelf Energy Evaluation" December 1996.

The Altran report was approved by the NRC in a letter to Carolina Power & Light (CP&L) (Brunswick), "Evaluation of the January 17, 1992 Operating Transient at the Brunswick Steam Electric Plant, Unit I and Evaluation of Carolina Power & Light Company's Equivalent Margins Analysis of the N-16A/B Instrument Nozzles at the Brunswick Steam Electric Plant, Units 1 and 2 (TAC Nos. MA0399/400)," October 16, 1998. The initial USE data in the Altran report was updated in a later GEH report 0000-01 14-0580-RO-NP to include additional SA-508 Class 2 forging data from the current NRC database RVID2. The GEH report (non-proprietary version), entitled Limerick Generating Station, Units 1&2, Upper Shelf Energy Evaluation for LPCI Nozzle Forging Material, August 2010, is attachment 3 to Exelon Nuclear letter RS-1 0-144 (ML102440265). Columbia has reviewed all the tabulated data for SA-508 Class 2 cited above and concluded that the lowest value in these reports for initial USE is 72 ft-lb.

The 72 ft-lb value of initial USE can be considered to be a lower bound value for the data presented. However, the initial USE value of 70 ft-lb is conservatively used for Columbia's calculation of USE in order to be consistent with the initial USE previously accepted by NRC for this material.

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION LICENSE RENEWAL APPLICATION Page 4 of 8 RAI 4.2-5

Background:

In its RAI response by letter dated January 28, 2011, the applicant stated that the 54 EFPY percentage decrease in USE (% USE decrease) for the N6 RHR/LPCI nozzle-to-RV welds is bounded by the equivalent margins analysis (EMA) acceptance criteria for RV shell welds from BWRVIP-74-A, "Boiling Water Reactor Vessel and Internals Project Reactor Pressure Vessel Inspection and Flaw Evaluation Guidelines."

The % USE decrease acceptance criteria in BWRVIP-74-A are based on minimum USE requirements derived from EMAs performed in GE NEDO-32205-A, "10 CFR 50 Appendix G Equivalent Margin Analysis for Low Upper-Shelf Energy in BWR-2 through BWR-6 Vessels," for shell plates and shell welds, and a conservative estimate of initial USE values based on a statistically significant set of Charpy USE data for each type of RV shell material.

The staff notes that the NEDO-32205-A EMAs developed minimum USE acceptance criteria for shell plates and shell welds. The NEDO-32205-A EMAs and minimum USE acceptance criteria are based on the ASME Code Case N-512 procedures, which are now codified in Appendix K of the ASME Code,Section XI.

The procedures include (1) the selection of an appropriate J-integral fracture resistance curve for the class of material being analyzed, (2) the calculation of J-integrals due to applied loads for RV shell components based on a postulated flaw configuration, and (3) the application of the acceptance criteria for (a) the applied J-integral at a ductile flaw extension of 0.1 inch and (b) flaw stability due to ductile tearing.

Calculations of J-integrals due to applied loads are very component specific - for example, the applied J integrals for RV shell components differ significantly from the applied J integrals for nozzles, even if the two types of components are fabricated from the same class of material.

Issue:

In order to demonstrate that the BWRVIP-74-A acceptance criteria for shell welds can be used to determine the acceptability of the N6 nozzle welds, it is necessary to confirm that:

(1) The N6 nozzle weld material is of the same class as the shell weld material analyzed in NEDO-32205-A, with respect to weld filler metal, welding flux, and weld fabrication technique.

(2) The N6 nozzle weld configuration, postulated flaw configuration, and loading is identical to (or bounded by) the RV shell weld configuration, postulated flaw

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION LICENSE RENEWAL APPLICATION Page 5 of 8 configuration, and loading, with respect to the applied J integral values, as calculated using ASME Code Case N-512 and Appendix K procedures.

Request:

Provide justification for the use of the BWRVIP-74-A shell weld EMA acceptance criteria to determine the acceptability of the N6 RHRILPCI nozzle welds, based on (1) N6 nozzle weld material and weld fabrication technique; and (2) the N6 nozzle weld configuration, postulated flaw configuration, and loading 1 , relative to the RV shell welds.

Energy Northwest Response:

(1) The percent decrease in USE acceptance criteria in BWRVIP-74-A is based on NEDO-32205-A. Columbia's reactor vessel beltline shell plate, weld materials and welding processes were included in the development of NEDO-32205-A.

The welding procedure specification (WPS) used to fabricate Columbia's reactor vessel beltline shell plate welds specifies the weld materials and welding processes. Columbia's N6 nozzle welds were fabricated using the same WPS that was used for the shell plate welds, thus the same weld materials and welding processes were used for the N6 nozzle welds.

Therefore, Columbia's N6 nozzle weld material is of the same class as the shell weld material analyzed in NEDO-32205-A with respect to filler metal, welding flux and weld fabrication technique.

(2) The calculation of projected USE for the N6 RHR/LPCl nozzle welds is provided in the Table-1. The projected USE exceeds 50 ft-lb and thus an Appendix G of 10 CFR 50 equivalent margin analysis using the methodology (applied J-integral) of BWRVIP-74-A (NEDO-32205-A) is not required. NEDO-32205-A, Table 2, page xviii (NRC Safety Evaluation) provides an initial USE for "non-Linde 80" weld material of 70 fl-lb. This value is based on the staffs statistical analysis for predicting 95% of the entire population with a one-sided 95% confidence for the "non-Linde 80" type of beltline weld material.

Columbia's N6 RHR/LPCI nozzle weld material is of this type (see response to (1) above). The 54 EFPY 1/4 T fluence for the N6 RHRPLPCI nozzle welds is 5.OOE+17 n/cm 2 . The BWRVIP Integrated Surveillance Program (ISP) best estimate chemistry copper content for Columbia's N6 weld material, heat 5P6214B, is 0.019%. Using the above data the Regulatory Guide (RG) 1.99 Rev. 2 projected decrease in USE is 7.8%. This results in a projected USE value of 64 ft-lb for weld heat 5P62144B at 54 EFPY. Therefore, the 10 CFR 50 Appendix G requirement for 50 ft-lb minimum USE at end of vessel life is met for the current license period and through the end of the Period of Extended Operation (PEO) for the N6 welds.

If the N6 nozzle welds are located in the shell of the RV, as depicted in the ASME Code,Section XI, Figures IWB-2500-7(a) and IWB-2500-7(b), then the applied J integral values calculated in NEDO-32205-A for the RV shell welds, based on the ASME Code, Section Xl, Appendix K calculation procedures, may be applicable to the N6 nozzle welds.

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION LICENSE RENEWAL APPLICATION Page 6 of 8 RAI 4.2-6

Background:

In its RAI response by letter dated January 28, 2011, the applicant added License Renewal Commitment Item No. 70, "TLAA - Embrittlement of Reactor Vessel," to the final safety analysis report (FSAR) supplement. The commitment states that the applicant will "[p]erform a 54 EFPY equivalent margins analysis for the embrittlement (upper shelf energy) of the reactor vessel N12 (instrumentation) nozzle forgings ...

[p]rior to the period of extended operation."

Issue:

The staff is concerned that the N12 nozzles' USE may drop below 50 ft-lbs prior to the period of extended operation. The staff believes that a commitment to include submittal of the EMA for NRC staff review and approval either (i) at least 2 years prior to the estimated date 2 the N12 nozzles' USE would drop below 50 ft-lbs; or (ii) at least 2 years prior to the period of extended operation; whichever timeframe is earlier, is the conservative approach.

Request:

The applicant needs to include in its LRA commitment that the N12 nozzle EMA will be submitted for NRC staff review and approval either (i) at least 2 years prior to the estimated date the N12 nozzles' use would drop below 50ft-lbs, or (ii)at least 2 years prior to the period of extended operation, whichever timeframe is earlier.

Energy Northwest Response:

The calculation of USE projected to 54 EFPY for the N12 water level instrument nozzle forgings is provided in the Table-1. The N12 nozzles are fabricated from SA-508 Class 1 material. The unirradiated (initial) transverse USE of 62 ft-lb and copper content of 0.27% used in the calculation of o USE for the N12 nozzles are based on the results of a statistical analysis of -i data by the original equipment manufacturer (OEM) for SA-508 Class 1 forging material. The RG 1.99 Rev. 2 decrease in USE projected to 54 EFPY for the N12 nozzles is 18%. This results in a USE projected to 54 EFPY of 51 ft-lb for the limiting N12 forging. Therefore, the 10 CFR 50 Appendix G requirement for 50 ft-lb minimum USE at end of vessel life is met for the current license period and for the PEO for the N12 forgings.

2 Estimates concerning the time when the N12 nozzles' USE is expected to drop below 50 ft-lbs can be made based on (a) calculations of projected % decrease in USE based on bounding copper content and projected fluence, as specified in RG 1.99, Rev. 2; and (b) a determination of a reasonably conservative initial USE value for the N12 nozzles based on available Charpy USE test data for SA-508, Class 1 forgings (the forging specification for the N1 2 nozzles), or possible application of the lower bound 70 ft-lb initial USE value from Charpy test data for SA-508, Class 2 forgings, if it can be ascertained, based on metallurgical principles, that 70 ft-lb is a reasonably conservative initial USE for SA-508, Class 1 forgings.

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION LICENSE RENEWAL APPLICATION Page 7 of 8 The above information provides reasonable assurance that Columbia's N12 nozzle forging USE will not drop below 50 ft-lb prior to the PEO. Discussions between the staff, Columbia and the OEM have confirmed that the evaluation methodology (applied J-integral) of NEDO-32205-A may not be bounding for beltline nozzles.

License Renewal commitment Item No. 70 has been revised by Energy Northwest to perform the necessary equivalent margin analysis for the N12 forgings no later than 2 years prior to the PEO.

RAI B.2.46-2

Background:

The first sentence of LRA Section B.2.46 states: "The Reactor Vessel Surveillance Program manages the reduction of fracture toughness due to radiation embrittlement for the low alloy steel reactor vessel shell and welds in the beltline region."

Issue:

The Columbia Reactor Vessel Surveillance Program manages the reduction in fracture toughness due to radiation embrittlement tor all low alloy steel RV beltline components, including all ferritic RV beltline nozzles and nozzle-to-RV welds.

Request:

The applicant needs to include all low alloy steel RV beltline components, including all ferritic RV beltline nozzles and nozzle-to-RV welds, into its Reactor Vessel Surveillance Program FSAR supplement and program description.

Enerav Northwest Response:

LRA section B.2.46 and the Final Safety Analysis Report (FSAR) supplement, A.1.3.1 have been revised to address all reactor vessel beltline components, including all ferritic beltline nozzles and nozzle to vessel welds.

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION LICENSE RENEWAL APPLICATION Page 8 of 8 Table-1 USE projected to 54 EFPY for the N6 and N12 nozzles and their associated nozzle to vessel welds.

54 EFPY 1/4T Unirradiated Fluence  % Drop USE Sub-Component Material Heat or Heat/Lot %Cu USE n/cm in USE (1/4 T)

FORGINGS Q2Q55W 790S-1, N6 (RHR / LPCI) SA508 C12 2, 3 0.11 70 4.48E+17 9.6% 63.3 N12 (Instrumentation) SA508 C11 219972 0.27 62 4.48E+17 18% 51 WELDS: I N6 (RHR / LPCl) RAC01NMM 5P6214B / 0331 0.019 70 5.OOE+17 7.8% 64 N12 (Instrumentation) Inco-82/182 NA NA NA NA NA NA

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION LICENSE RENEWAL APPLICATION Page 1 of 1 LICENSE RENEWAL APPLICATION AMENDMENT 33 Section Page Number RAI Number Number 4.2.2 4.2-4a RAI 4.2-6 Table 4.2-2 RAI 4.2-5 Insert Line Items RAI 4.2-6 Table 4.2-2 RAI 4.2-5 Footnote RAI 4.2-6 Table 4.2-8 4.2-14b RAI 4.2-5 A.1.3.1.1 A-28 RAI B.2.46-2 A.1.3.1.2 A-29a RAI 4.2-6 Table A-I TbeA1A-68d RAI 4.2-6 Line Item 70 B.2.46 B-175 RAI B.2.46-2

Columbia Generating Station License Renewal Application Technical Information Insert A for Page 4.2-4: and The projected EMAs are listed in Table 4.2- T ble The projected EMAs in Table 4.2-3 Table 44.22-,,,.2 used the projected 54 EFPY fluence listed i\able 4.2-1, and the curves provided in RG 1.99 Figure 2. The predicted values were compared to the minimum 54 EFPY USE limits in BWRVIP-74-A.

and Insert B for Page 4.2-4 P-eFinfe-vessef-9euine noezzics-, mfe-maxfimmacac in ubt: was roune to Be per.er.-(see-Tab.es-4.2-8-and 4.2 9). This is lss than the assumed decrease oef 23.5 pcrccnt in the e .ui.alnt margi* -analysis. Thc.reo. the mximum', predicted d.., ase.

in-USE-4r-the-nezzes-in-the-vessel bein arc boUnd by thc generic 54 EFPY equivalent a-maianalysis--deueimneted in BWRVIP 74 '. The --- 'ced USE- for th.

rie-FK)zz-ieb-is-aecep,.el.-...e-pe- - -0. - '*J i all beltline materials, including I instrumentation nozzl eAmust be considered when the licensee develops pressure-temperature limits for Columbia in accordance with 10 CFR Part 50, Appendix G and the ASME Code,Section XI, Appendix G. Columbia will continue to develop future pressure-temperature limit curves considering all beltline plates, welds, and nozzles.

The I: ;- R,-, I6.-,,-, AG,- EMA-,.*, ,f^_ht-l,* ;II 6 (-i-f,--**.e., theI ,,;;

N,.Re zes

• Ci *,w iG 4 Amendmentl2 Time-Limited Aging Analyses Time-Limited Aging Analyses Page 4.2-4a Page 4.2-4a Amendment 2 Arn e n d m e n I

4.2.2 Upper Shelf Energy Evaluation Columbia Generating Station License Renewal Application Technical Information Table 4.2-2 USE Projections for 54 EFPY 1 1

/'T Drop /,T I.D. Heat Initial Fluence in USE 2

Sub-Component(l) No. (Single/Tandem wire)  % Cu USE nlcm USE (ft-lb)

PLATES:.

Lower-intermediate Mk22- B5301-1 0.13 98 8.10E+17 12.1% 86.1 Shell (Course #2) 1-1 Lower Vertical BA- 3P4966 (S) 0.025 98 2.71E+17 7.0% 91.1 (Axial/Longitudinal) BD 3P4966 (T) 0.025 98 2.71E+17 7.0% 91.1 Lower-Intermediate Vertical BE- 3P4986 (S) 0.025 98 8.10E+17 9.1% 89.1 (Axial/Longitudinal) BH 3P4966(T) 0.025 98 8.10E+17 9.1% 89.1 Lower to Lower-Intermediate Girth AB 5P6756 (S) 0.080 91 3.30E+17 9.8% 82.1 (Circumferential) AB 5P6756 (T) 0.080 97 3.30E+17 9.8% 87.5 3P4955 (S) 0.027 90 3.30E+17 7.4% 83.3 AB 3P4955 (T) 0.027 95 3.30E+17 7.4% 87.9 (1) -The sub--oT ipiofl =L 2= L L ____;_ =*! __ J. __ L I  ! I/*'*p

~IILO I I'JI VI I tI 1I LOUI IOZV1 I IJ 1JJ WJjVY.LIVI I JUUV LI LI IV IIOI'Js.;L.

11110 b ayes-fe--the--i

~ate-ant~wet&

(2) Heat 5P6214B is a surveillance weld in the BWRVIP Integrated Surveillance program. All seven data points for this heat show less reduction in USE than predicted by RG Replace footnote with Insert A from Page 1.99, thus the RG 1.99 prediction will be used without 14.2-5a correction.

N6 Nozzle forgings SA-508 SA58 012 I Q2Q55W 790S-1,2,3 10.11 I 70 II4.48E+17 9.6% 63.3 N12 Nozzle Forqinqs I. ....

SA-508 CI1 219972 0.27 62'4.48E+17 18% 51 I.

Time-Limited Aging Analyses Page 4.2-5 FAmendment 33 Am, en,_28

Columbia Generating Station License Renewal Application Technical Information Insert A for Page 4.2-5: and (1) The sub-components not on thisitable have no projection due to the initial USE being unknown. See Table 4.2-37 Table 4.2-4, Tablc 4.2 8 aRnd Table 4.2 9 for the equivalent margin analyses for the limiting plate limiting weldT and "Itline nozzloc.

Time-Limited Aging Analyses Page 4.2-5a Ameidpment-12

[Amendment 33 4

Columbia Generating Station License Renewal Application Technical Information Trable--4--8 WpeY -ntve .i4 w "Stirveillance Weld: ',° Heait*k6214Bseven program.-*Al is a surveillance weld data points for thisin heat the BWRVIP show lessIntegrated reduction Surveillance/

in USE ta predicted by-RN 1.99, thus the RG 1.99 prediction will be used withou,,t.,o~rrection.

N6 weld USE:"*

_ "  % Cu = 0.02 54 EFPY 'T Fluen = 5.00 17 RG 1.99 Predicted Decrease = 8%

Adjusted Decr se= N/A 7.8% _< 39% (bounding value omr SER for BWRVIP-74-A)

Therteore the N6 nozzle-to-vessel weld is bounded by the equivalent ma n lysis in BWRVIP-74-A Time Limited Aging Analysis Page 4.2-14.b -Ameftdmeni-2 JAmendment 33

A.1.3.1.1 Neutron FPuence Columbia Generating Station License Renewal Application Technical Information oFF]

I nozzles and the associated Beltline Evaluation Inozzle to vessel welds For the extended operating peri rritic materials for vessel beltline shells, welds, and assembly components are required to be evaluated for neutron irradiation embrittlement if high energy neutron fluence is greater than a threshold value of 1E+17 n/cm 2 (E >1 MeV) at the end of the 60 years. The only vessel assembly items, other than the shells and welds of the beltline region that would experience neutron fluence greater than the 1E+17 n/cm 2 during the period of extended operation are instrumentation nozzle N12 (and and residual heat removal/low pressure coolant injection (RHR/LPCI) nozzle N6* associated S-- _nozzle-to-Instrumentation nozzle N12 has a thickness less than 2:5 inches and therefore does not vessel require a fracture toughness evaluation per ASME Code Appendix G. Section G2223. welds)

Nozzle N6 is evaluated for ART below. The ART for this nozzle is less than that for the highest weld and plate. Consequently, nozzle N6 is not the limiting material for the vessel, and thus is not a beltline component However, as nozzle N6 was evaluated for ART it meets the definition of a bettline component per 10 CFR 50, Appendix G.

The beltline definition for the period of extended operation includes the lower shell (Course #1 / Ring #21), lower-intermediate shell (Course #2 / Ring #22), associated vertical (Iongitudinal) welds, the girth (circumferential) weld that connects the lower and lower-intermediate shells,*and nozzle N6. < (and its associated

-::-7nozzle-to-vessel Disposition Iweld), and nozzle N12 Neutron fluence is not a TLAA, it is a time-limited asstimptier, used in various neutron embrittlement TLAAs.

A.1.3.1.2 Upper Shelf Energy Evaluation 10 CFR 50 Appendix G requires the upper shelf energy (USE) of the vessel beltline materials to remain above 50 ft-lb at all times during plant operation, including the effects of neutron radiation. If USE cannot be shown to remain above this limit, then an equivalent margin analysis (EMA) must be performed to show that the margins of safety against fracture are equivalent to those required by Appendix G of Section XI of the ASME Code.

The initial (unirradiated) USE is not known for all the Columbia vessel plates and welds.

For those plates and welds for which the initial USE is known, USE was projected using Regulatory Guide 1.99, Revision 2 methods. For the vessel plates and welds for which the initial USE is not known, USE equivalent margin analyses were performed using the Boiling Water Reactor Owners Group (BWROG) equivalent margin analysis (EMA) methodology. Results from the testing and analysis of surveillance materials were used in the EMA analyses.

Final Safety AnalysisFinal Report Supplement afet Page A-28 J r -40 E ',L:-:.L 3 :-*~*'d-*'?

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Columbia Generating Station License Renewal Application Technical Information Insert A for Page A-29: , nozzle forg ings, All of the projected USE values for the vessel beitline plates and welds for which the initial USE is known remain above 50 ft-lbs through tl'e end of the period of extended _-__

operation (54 EFPY). For the vessel beltline plates- *elds and-rnozzles for which the initial USE is not known, the maximum decrease in USE was found to be less than the assumed decrease in the associate equivalent margin analyses. The maximum predicted decreases in USE for 54 EFPY for these beltline plates-, elds .-.-,nezzf a are bounded by the fquivalent margin analyses. Therefore, the projected USE for the vessel beltline plates,"elds and nozzles is acceptable for the period of extended operation. generic lall beltline materials, including I CKI eeprd-nyeeN2Rh 4in. -S 011101110 7A A

] entin thc analyzcd -epper c,,tct. Energy Northwest agrees thatthe N12 instrumentation nozzles must be considered when the licensee develops pressure-temperature limits for Columbia in accordance with 10 CFR Part 50, Appendix G and the ASME Code, Section Xl, Appendix G. Columbia will continue to develop future pressure-temperature limit curves considering all beltline plates, welds, and nozzles.

There ic no eqUiValent Margin analYSic aVailable for feogings for Whieh

nwai,,R e.,;tkR1wn kc 114 n42 ,Ie). 6ensec,.eily, tcRG hY

- No~iwst ill Peaf6FM a zpcc5-4 EFPY oguivalent margin analysis

,-. t N1ie 2 no-zlo foFrgn;"*q- c fit*t'd iR Table A 1 f* t1c 1 ;iE",nce Penawal Application.

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ITbleA-1 Columbia Generating Station License Renewal Application Technical Information Insert into page A-68c FSAR Enhancement Item Number Commitment Supplement or Location Implementation (LRA App. A) Schedule

68) Ensure that the condenstate (COND) and reactor feedwater (RFW) Prior to the period Flow-Accelerated systems are screened and evaluated for cavitation prior to entering A. 1.2.28 of extended Corrosion (FAC) the period of extended operation (PEO). If the in-scope portion of operation.

either systemerosion, to cavitation is determined then a to be susceptible program(s) will betomodified loss of material dueto or created manage the loss of material

69) Re-evaluate the aws for the period of extended operation (54 Prior to the period Inservice EFPY), in accordance with the requirements of the ASME Code, A. 1.2.33 of extended Inspection (ISI) Section XI, IWB-3600 based on the results of 2015 inservice operation.

Program inspection.

portions of the reactor pressure vessel beltline welds BG and BM

70) TLAA - Perform a 54 EFPY equivalent margin analysis for the embrittlement A. 1.3.1.2 rior to the period Embrittlement of (upper shelf energy) of the reactor vessel N1 2 (instrumentation) nozzle of extended reactor vessel forgings. operation No later than 2 years prior Fll Final Safety Analysis Report Supplement Page A-68d AeMRdmemo- 26 jAmendment 33

Columbia Generating Station License Renewal Application Technical Information B.2.46 Reactor Vessel Surveillance Program shell, welds, ferritic nozzles and the Program Description associated nozzle to vessel welds The Reactor Vessel Surveillance Program manages the reduction of fracture toughness due to radiation embrittlement for the low alloy steel reactor vessel '*heI-an-weld in the beltline region. The Reactor Vessel Surveillance Program is a condition monitoring program developed in response to 10 CFR 50 Appendix H.

The Columbia program is part of the BWRVIP Integrated Surveillance Program (ISP) that includes multiple BWR vessels. The BWRVIP ISP is an NRC-approved program that appropriately implements the requirements of Appendix H to 10 CFR Part 50.

Testing and reporting done by the BWRVIP ISP is performed in accordance with the requirements of ASTM E 185 (1982). The NRC has approved the use of the BWRVIP ISP in place of a unique plant program for Columbia. The BWRVIP ISP has been revised for License Renewal, as documented in BWRVIP-1 16, to ensure representative capsules are irradiated to fluence levels corresponding to the end of the period of extended operation.

The BWRVIP ISP uses material surveillance capsules in BWR plants, as well as supplemental capsules irradiated in host plants, to provide data which bounds all operating BWR plants. No surveillance capsules from Columbia are included in the BWRVIP ISP; however, the Columbia surveillance capsules will continue to be maintained in the reactor vessel in standby (deferred) status as required by the ISP.

Capsules from host plants will be removed and tested in accordance with the ISP implementation plan defined in BWRVIP-86-A. Results from these tests that are applicable to Columbia will provide the necessary data to monitor embrittlement for the Columbia reactor pressure vessel (RPV). EN will apply the results of the ISP capsule testing to Columbia.

The neutron fluence values used for the projections of neutron embrittlement effects are determined using NRC-approved methodology. The exposure conditions of the reactor vessel are monitored to ensure that they continue to be consistent with those used to project the effects of embrittlement to the end of the license term. If the reactor vessel exposure conditions (neutron flux, spectrum, irradiation temperature, etc.) are altered, then the basis for the projection to 60 years is reviewed; and, if deemed appropriate, a revised fluence projection is prepared and the effects of the revised fluence analysis on neutron embrittlement calculations will be evaluated.

The determination of neutron embrittlement effects for Columbia fully complies with NRC Regulatory Guide 1.99, Revision 2. Projections for neutron embrittlement effects have been adjusted to account for the specific nickel and copper contents of the Columbia materials. The extent of reactor vessel embrittlement for upper-shelf energy (USE) and adjusted reference temperature for nil-ductility transition (ART) is projected for 60 years in accordance with Regulatory Guide 1.99, Revision 2. These projections Aging Management Programs Page B-1 75 Janueay-1--t

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