ML090430400

Request to Extend Inservice Inspection Interval for the Reactor Vessel Weld Examination and Request for License Amendment for Submittal of ISI Information and Analyses - Response to Request for Additional Information

ML090430400
Person / Time
Site:	Cook
Issue date:	01/14/2009
From:	Weber L Indiana Michigan Power Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
AEP-NRC-2009-2, TAC MD9934
Download: ML090430400 (9)

Text

Indiana Michigan Power Company INDIA*NA Nuclear Generation Group One Cook Place MICHIG-AN Bridgman, Ml 49106 PORI R aep.com January 14, 2009 AEP-NRC-2009-2 10 CFR Part 50.55a DoCket'No0.: 50-316 U. S -NS'tlIear_ Regulatory-Co-mmission . , .. Y.

ATTN: Document Control Desk Washington, DC 20555-0001,'.

Sujddt: . Donald C. Cobk Ni~clar Plant Unit.2

'..- *'.Recl6estfo& Relief td,-Extend the Unit,2clhsrvibeIlispectiohteValfoi the Ren.tor.

.'Vessel Welcd Ekamiinatibn,ahd Request for" License Amehdn'ment fori:submittal-bf ISI Information and Analyses - Response to Request for Additional Information (TAC NO. MD9934). :. *...

References:

1. Letter from L. J. Weber, Indiana:Michigan Power C ompany (I&M), to Nuclear Regulatory Commission (NRC) Document Control Desk,, "Donald C. Cook Plant Unit 2, Request for Relief to Extend the Unit 2 Inservice Inspection Interval for the Reactor Vessel Weld Examination and Request for License Amendment for Submittal of iSI Information and Analysis," AEP-NRC-2008-41, dated October 9, 2006-(ML082980354).

2. Letter from T. A. Beltz, NRC, to M. W. Rencheck, I&M, "Donald C. Cook Nuclear Plant, Unit 2 - Request for.Additional Information Regarding Relief Request (ISIR-29) for use of Risk-Inf*rmed'Extension of the Inservice Inspection Interval for the Reactor Pressure Vessel Weld Examination, (TAC No. MD9934)," dated December 10, 2008 (ML083430018).

In Reference 1;" Indiana Midhig-an.P6wer: Com-npany"(I&M):subrfiitted a re~dlest for. relief to extend the Unit 2 lnservice". hsp&6ction .*(ISI) :InterVabl 'for-the -Reabto(V ssbl Weld :Exarnihation (ISI R-29)!"and request, for license:. aren'ndment'-for:subrmittalT'of ISI Information and Analysis. Reference 2 transmitted the Nuclear Regulatory Commission's request for additional information (RAI) regarding the requestfor relief. The'attachment to this letter provides I&M's response to the RAI.

There are no nbew0r revised commitments made in this letter. Should you have any questions, please contact John A. Zwolinski, Manager of Regulatory Affairs, at (269) 466-2478.

Sincerely',-

Lawrence J. Weber Site Vice President Ac10

U. S. Nuclear Regulatory Commission AEP-NRC-2009-2 Page 2 RP/rdw

Attachment:

Request for Relief to Extend the Unit 2 Inservice Inspection Interval for the Reactor Vessel Weld Examination and Request for License Amendment for Submittal of ISI Information and Analysis, Response to Request for Additional Information c: T. A. Beltz - NRC Washington, DC J. L. Caldwell - NRC Region III K. D. Curry - AEP Ft. Wayne J. T. King - MPSC MDEQ - WHMD/RPS NRC Resident Inspector

U. S. Nuclear Regulatory Commission AEP-NRC-2009-2 Page 4 AFFIRMATION I, Lawrence J. Weber, being duly sworn, state that I am Site Vice President of Indiana Michigan Power Company (I&M), that I am authorized to sign and file this request with the Nuclear Regulatory Commission on behalf of I&M, and that the statements made and the matters set forth herein pertaining to I&M are true and correct to the best of my knowledge, information, and belief.

Indiana Michigan Power Company Lawrence J. Weber Site Vice President SWORN TO AND SUBSCRIBED BEFORE ME THIS DAY OF 42009 1

Notary Public My Commission Expires 1//1/e/

Attachment to AEP-NRC-2009-2 Request for Relief to Extend the Unit 2 Inservice Inspection Interval for the Reactor Vessel Weld Examination and Request for License Amendment for Submittal of ISl Information and Analyses, Response to Request for Additional Information In Reference 1, Indiana Michigan Power Company (I&M) submitted a request for relief to extend the Unit 2 Inservice Inspection (ISl) Interval for the Reactor Vessel Weld Examination and request for license amendment for submittal of ISI Information and Analysis. Reference 2 transmitted the Nuclear Regulatory Commission's (NRC) request for additional information (RAI) regarding the request for relief. The requested information is provided below.

NRC RAI 1 In Section 3.4 of the final safety evaluation report issued May 8, 2008 (ADAMS Accession No. ML081060045), the NRC staff notes that licensees submitting a request for an alternativebased on Topical Report WCAP-16168-NP, "Risk-Informed Extension of the Reactor Vessel In-Service.

Inspection Interval," must submit the following plant-specific information:

1. Licensees must demonstrate that the RTmax-x and the shift in the Charpy transition temperature produced by irradiationdefined at the 30 ft-lb energy level AT 30 must be calculated using the latest approved methodology documented in Regulatory Guide (RG) 1.99, "Radiation Embrittlement of Reactor Vessel Materials," or other NRC approved methodology.

"Other NRC-approved methodology" includes equations 5, 6, and 7, as described in paragraph ,.

(g) of the proposed rule, Title 10 of the Code of Federal Regulations (CFR) 50.61 a, as published in the Federal Register, Vol. 72, No. 191, dated October 3, 2007. However, paragraph(0(6) of the proposed rule contains a prescriptiveapproach to determining the validity *:

of implementingequations5 through 7 of paragraph(g) for the calculation of AT 30 values. Data from a plant specific or integrated surveillance program shall be evaluated in accordance with 10 CFR 50.6la(0(6) to validate use of the associatedembrittlement model.

The licensee has implemented the use of equations prescribed in 10 CFR .50.61a(g),

NUREG-1874, and WCAP-16168-NP to determine AT 30 values for the beltline materials.

Please submit one of the following:

1) Information consistent with the requirements of proposed rule 10 CFR 50.6la(f)(6) that establishes the applicability of equations 5 through 7 as provided in 10 CFR 50.61a(g) for calculatingAT 30 values for a plant's beltline materials.

2) Recalculation of AT 30 using RG 1.99. This would require reporting of recalculated values of both AT 30 and all through wall cracking frequency (TWCF) calculations consistent with the format of Table 3 provided in the proposedrelief request.

Attachment to AEP-NRC-2009-2 Page 2 I&M Response to NRC RAI 1 I&M has chosen to respond to option 2 of RAI 1. Delta T30 (AT 30) has been recalculated based on Regulatory Guide (RG) 1.99, Revision 2. The through-wall cracking frequency (TWCF) was recalculated using the new AT 30 values. The inputs to these calculations and the results are provided in Table 1 (next page). It should be noted that the chemistry factors in Table 1 are consistent with those used in the calculation of the Donald C. Cook Nuclear Plant (CNP) Unit 2 heatup and cooldown limit curves and RTPTS values. These calculations were provided to the NRC in WCAP-15047, Revision 2, and WCAP-13517, Revision 1, respectively. These two WCAP reports were approved by the NRC via the SER as noted in the response to RAI 3. As shown in Table 1, the TWCF value for CNP Unit 2 based on AT30 values calculated using RG 1.99, is 3.80E-10 events per year. This value is less than the Westinghouse pilot plant value of 1:76E-08 events per year in WCAP-16168-NP-A, Revision 2. Therefore, the application of the extended IS[ interval to CNP Unit 2 remains acceptable.

Attachment to AEP-NRC-2009-2 Page 3 Table 1 Details of Revised TWCF Calculation - D.C. Cook Unit 2 at 60 EFPY Inputs 542.5 Reactor Coolant System Temperature, TRCS[°F]: 5 T,,,, [inches]: 8.50 R.G. Un- Fluencel at Region/Component

1. Material Cu Ni C.F. 1.992 Irradiated 60 EFPY Description [wt%] [wt%] [OF] RTNDT(u) [n/cm2, S[IF E > 1.0 MeV]

1 Int. Shell Axial Weld Linde 124 0.056 0.956 66.3 2.1 -35 1.38E+19 2 Int.. Shell Axial Linde 124 0.056 0.956 66.3 2.1 -35 1.38E+19 Weld 3 Low. Shell Axial Linde 124 0.056 0.956 66.3 2.1 -35 9.58E+18 Weld 4 Low. Shell Axial Linde 124 0.056 0.956 66.3 2.1 -35 9.58E+18 Weld t 1 5 Int./Low. Circ Weld Linde 124 0.056 0.956 66.3 2.1 -35 3.08E+19 6 Inter. Shell Plate A 533B 0.125 0.580 102.3 2.1 38 3.08E+19 7 Inter. Shell Plate A 533B 0.150 0.570 108.4 1.1 58 3.08E+19 8 Lower Shell Plate A 533B 0.110 0.640 74.6 1 1.1 -20 3.08E+19 9 Lower Shell Plate A'533B 0.140 0.590 99.5 1.1 -20 3.08E+19 Outputs Methodology Used to Calculate AT30 : Regulatory Guide 1.99, Revision 2 Controlling Material Fluence 1 FF Region # RTMAxxx [n/cm 2, (Fluence AT30 ['F] TWCF95-XX (From E > 1.0 MeV] Factor)

Above)

Axial Weld - AW 7 635.7 1.38E+19 1.09 118.1 3.49E-11 Circumferential Weld - 7 3.08E+19 1.30 140.6 CW 658.3 3.96E-14 Plate - PL 7 658.3 3.08E+19 1.30 140.6 1.28E-10 TWCF95-TOTAL (cOAwTWCFe5-AW + QXPLTWCF95-PL + cccwTWCF 95-cw): 3.80E-10 Note 1: Values are for fluence at the cladding-to-base metal interface.

Attachment to AEP-NRC-2009-2 Page 4 NRC RAI 2 In Table 1 of the submittal, the license states that they are bounded by seven heatup/cooldowns per year.

Please cite the plant design basis for heatup/cooldownsper year.

I&M Response to NRC RAI 2 According to the Updated Final Safety Analysis Report (UFSAR) Section 4.1.4. Cyclic Loads:

"A renewed operating license extends the license term an additional 20 years for Cook Nuclear Plant (CNP), Units 1 and 2. This extension was justified based on design transient cyclic loads defined in Table 4.1- 10. The reactor coolant system was originally qualified using a conservative estimate of design cycles for a 40 year life. However, design life is dependent in part on fatigue cycles, not years of service. In evaluations performed for CNP, the actual number of cycles was extrapolated to 60 years. For the major reactor coolant system components, the extrapolated numbers of cycles over a 60-year life will not exceed the design cycles noted in the UFSAR. The actual transient cycles are tracked and documented to ensure they remain below the allowable number of design cycles, as further discussed in Chapter 15 of the UFSAR."

For Unit 2 there have been 61.75 heatup and .60.75 cooldown cycles over the operating life of the plant. The number of design heatup and. cooldown cycles is 200 for the life of the plant.

Based on this, the remaining number of design 'cycles for the remaining life of the plant is approximately 140 cycles.

The renewed license permits operating Unit 2 until December 23, 2037 (DPR-74). The number of heatup and cooldown cycles allowed in the remaining 29 years of operation for Unit 2 would be 140/29 = 4.8 cycles per year. Given the previous 31 years of operation for Unit 2, the unit has averaged approximately 2 heatups and cooldowns per year. The analyses noted in the I&M submittal based on seven cycles per year would envelop the plant's remaining design basis number of cycles.

NRC RAI 3 In Table 3 of the submittal, the licensee cites compositions in terms of Cu, Ni, P, and Mn for items 1-9. These compositions are not consistent with NRC records in Reactor Vessel Integrity Database.

Please explain where the Cu, Ni, and P compositions were accepted by the NRC or provide new data.

Please explain on what basis the reported Mn compositions were determined.

Attachment to AEP-NRC-2009-2 Page 5 I&M Response to NRC RAI 3 The composition of Copper (Cu), Nickel (Ni), and Phosphorus (P) elements for the reactor vessel materials are taken from the Electric Power Research Institute (EPRI) reactor vessel database called RPVDATA. This database was specifically developed to include the pressurized water reactor (PWR) reactor vessel material chemical and mechanical properties into a single integrated database. The RPVDATA database was developed by EPRI with input on the reactor vessel material ir~formation supplied by all three PWR Owners groups, i.e.,

Westinghouse, Combustion Engineering, and Babcock & Wilcox.

Information from the RPVDATA database has been used in surveillance capsule analysis performed by Westinghouse for CNP Unit 2 in 2002. These analyses have been documented in Westinghouse WCAP reports as noted below:

1. WCAP-13515, Revision 1, "Analysis of Capsule U from the Indiana Michigan Power Company D.C. Cook Unit 2 Reactor Vessel Radiation Surveillance Program," April 2002.

2. WCAP-15047, Revision 2, "D.C. Cook Unit 2 WOG Reactor Vessel 60-Year Evaluation Minigroup Heatup and Cooldown Limit Curves for Normal Operation," May 2002.

3. WCAP-13517, Revision 1, "Evaluation of Pressurized Thermal Shock for D.C. Cook Unit 2,"

May 2002.

The above WCAP reports have been submitted to the NRC as part of the technical specification amendment for the revised heatup-cooldown curves. The NRC approval is documented in their Safety Evaluation Report (SER) for Amendment No. 255 for Unit 2, dated March 20, 2003.

The Manganese (Mn) composition values identified in Table 3 of the submittal, and used in the TWCF calculations, are based on the conservative percent weight estimates in Table 4 of the proposed alternate PTS Rule, 10 CFR 50.61a.

It should be noted that the P and Mn compositions are no longer relevant since TWCF has been recalculated in the response to RAI 1 using the AT 30 shift correlations of RG 1.99, Revision 2, which do not consider P and Mn composition.

Attachment to AEP-NRC-2009-2 Page 6

References:

1. Letter from L. J. Weber, I&M, to NRC Document Control Desk, "Donald C.

Cook Plant Unit 2, Request for Relief to Extend the Unit 2 Inservice Inspection Interval for the Reactor Vessel Weld Examination and Request for License Amendment for Submittal of ISl Information and Analysis,"

AEP-NRC-2008-41, dated October 9, 2008 (ML082980354).

2. Letter from T. A. Beltz, NRC, to M. W. Rencheck, I&M, "Donald C. Cook Nuclear Plant, Unit 2 - Request for Additional Information Regarding Relief Request (ISIR-29) for use of Risk-Informed Extension of the Inservice Inspection Interval for the Reactor Pressure Vessel Weld Examination (TAC' No. MD9934)," dated December 10, 2008 (ML083430018).