AEP-NRC-2008-41, Request for Relief to Extend the Inservice Inspection Interval for the Reactor Vessel Weld Examination and Request for License - Amendment for Submittal of ISI Information and Analyses

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

ML082980354
Person / Time
Site:	Cook
Issue date:	10/09/2008
From:	Weber L American Electric Power Co, Indiana Michigan Power Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
AEP-NRC-2008-41
Download: ML082980354 (12)

Text

Indiana Michigan Power Company Nuclear Generation Group INDIANA One Cook Place MICHIGAN Bridgman, MI 49106 POWER aep.com October 9, 2008 AEP-NRC-2008-41 10 CFR 50.55a Docket No.: 50-316 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop O-Pl-17 Washington, DC 20555-0001

SUBJECT:

Donald C. Cook Nuclear 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 Analyses

Reference:

Letter from Ho K. Nieh, Nuclear Regulatory Commission, to Gordon Bischoff, Westinghouse Owner's Group, regarding Final Safety Evaluation for PWROG Topical Report WCAP-16168-NP, Revision 2, (TAC NO. MC9768), dated May 8, 2008 .(ML081060051).

Dear Sir or Madam:

Pursuant to 10 CFR 50.55a(a)(3)(i), Indiana Michigan Power Company (I&M), the licensee for Donald C. Cook Nuclear Plant Unit 2, hereby requests Nuclear Regulatory Commission (NRC) approval of the following request for the third ten-year interval inservice inspection testing program:

Relief Request ISIR-29 for use of risk-informed extension of the Reactor Pressure Vessel Inservice Inspection (ISI) Interval from 10 to 20 years. This relief request also applies to future ISI inspection intervals up to the end of license, DPR-74.

The NRC approved WCAP-16168-NP-A, Revision 2, "Risk-Informed Extension of The Reactor Vessel In-Service Inspection Interval," in the above referenced letter. This WCAP provides for extension of the ISI interval for certain pressure retaining welds in the reactor vessel from 10 to 20 years. I&M proposes to implement this extended ISI interval for Unit 2. The plant-specific information identified by the above letter needed to support this request is included in Enclosure 1 to this letter. I&M has concluded that the proposed alternative provides an acceptable level of quality and safety in accordance with 10 CFR 50.55a(a)(3)(i).

As required by the referenced letter, I&M is also requesting an amendment to the Unit 2 license that will require that the information and analyses requested in the final rule for 10 CFR 50.61a, Section (e) or, prior to issuance, the proposed rule (72 FR 56275) for 10 CFR 50.61a, Section (e), be submitted within one year of completing each of the American Society of Mechanical Engineers Code,Section XI, Category B-A and B-D Reactor Vessel weld inspections. The requested amendment and evaluation are contained in Enclosure 2 to this letter.

U. S. Nuclear Regulatory Commission AEP-NRC-2008-41 Page 2 I&M requests approval of the relief request and license amendment by March 6, 2009, to support the Spring 2009 Unit 2 refueling outage U2C18. In a telephone conference with the NRC staff on July 9, 2008, it was agreed that an expedited review of the Unit 2 relief request would be provided if the submittal was provided in early October 2008 and there are no significant deviations from the approved NRC safety evaluation documented in the referenced letter. The enclosed request contains no significant deviations from the approved NRC safety evaluation documented in the referenced letter.

This letter contains no new or revised commitments. Should you have any questions, please contact John A. Zwolinski, Manager of Regulatory Affairs, at (269) 466-2478.

Sincer ly, Lawrence J. Weber Site Vice President RSP/rdw Unit 2 Request for Relief to Extend the Third 10-year Reactor Vessel Inservice Inspection Interval Relief Request No: ISIR-29 Proposed License Amendment Regarding ASME Relief Request 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-2008-41 Page 3 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 (CAD Pf- ,2008 My Commission Expires too W, do,,,nFA My'vu I; to AEP-NRC-2008-41 Page 1 Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)

-Alternative Provides Acceptable Level of Quality and Safety-

1. ASME Code Component(s) Affected The affected component is the Donald C. Cook Nuclear Power Plant (CNP) Unit 2 reactor vessel (RV), specifically the following American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code Section XI (Reference 1) examination categories and item numbers covering examinations of the RV. These examination categories and item numbers are from IWB-2500 and Table IWB-2500-1 of the ASME BPV, Code Section X1.

Examination Category Item No. Description B-A BI.11 Circumferential Shell Welds B-A B1.12 Longitudinal Shell Welds B-A B1.21 Circumferential Head Welds B-A B1.22 Meridional Shell Welds B-A B1.30 Shell-to-Flange Weld B-D B3.90 Nozzle-to-Vessel Welds B-D B3.100 Nozzle Inner Radius Areas

2. Applicable Code Edition and Addenda

ASME Code Section X1, "Rules and Inservice Inspection of Nuclear Power Plant Components,"

Code 1989 Edition (no Addenda).

3. Applicable Code Requirement

IWB-2412, Inspection Program B, requires volumetric examination of essentially 100% of RV pressure retaining welds identified in Table IWB-2500-1 once each ten-year interval. The CNP Unit 2 third ten-year inservice inspection (ISI) interval is scheduled to end on or before February 28, 2010.

to AEP-NRC-2008-41 Page 2

4. Reason for Request

An alternative is requested from the requirement of IWA-2412, Inspection Program B, that volumetric examination of essentially 100% of RV pressure retaining, Examination Category B-A and B-D welds, be performed once each ten-year interval. Extension of the inspection interval for Examination Category B-A and B-D welds from 10 years to up to 20 years will result in a reduction in man-rem exposure, estimated savings of 1.5 man-rem per outage, and examination costs.

5. Proposed Alternative and Basis for Use Indiana Michigan Power Company (I&M), the licensee for CNP, proposes to defer the ASME Code required volumetric examination of the CNP Unit 2 RV full penetration pressure retaining Category B-A and B-D welds for the third ISI until 2019. Further, the fourth period ISI will be performed on a twenty-year inspection interval, instead of the currently required ten-year inspection interval. Therefore, the subsequent period ISI is proposed to be performed in 2039.

This schedule is a deviation from the dates provided in Pressurized Water Reactor (PWR)

Owners Group letter OG-06-356 (Reference 2). However, the dates proposed above for Unit 2 are consistent with those proposed in OG-06-356 for CNP Unit 1. It is the intention of I&M to perform the third interval inspections for CNP Unit 1 in 2009 and submit a separate relief request for CNP Unit 1 to be inspected on the dates indicated in OG-06-356 for CNP Unit 2. This change in proposed inspection dates for CNP Units 1 and 2 (i.e., swapping of the future inspection dates between Unit 1 and Unit 2) will not affect the planned distribution of inspections over the next 40 years as presented in OG-06-356 for the PWR fleet and will ensure continued collection and submittal of the inspection data to the Nuclear Regulatory Commission (NRC).

In accordance with 10 CFR 50.55a(a)(3)(i), an alternate inspection interval is requested on the basis that the current inspection interval can be extended based on a negligible change in risk by satisfying the risk criteria specified in Regulatory Guide 1.174 (Reference 3).

The methodology used to demonstrate the acceptability of extending inspection intervals for Category B-A and B-D welds based on a negligible change in risk is contained in WCAP-16168-NP-A, Revision 2 (Reference 4). This methodology was used to develop a pilot plant analysis for Westinghouse, Combustion Engineering, and Babcock and Wilcox RV designs and is an extension of the work that was performed as part of the NRC Pressurized Thermal Shock (PTS)

Risk Re-Evaluation (Reference 5). The critical parameters for demonstrating that this pilot plant analysis is applicable on a plant-specific basis, as identified in WCAP-16168-NP-A, Revision 2, are identified in Table 1. By demonstrating that each plant-specific parameter is bounded by the corresponding pilot plant parameter, the application of the methodology to the CNP Unit 2 RV is acceptable as shown in Table 1. Additional plant-specific information as required per Section 3.4 of NRC Safety Evaluation to WCAP-16168-NP, Revision 2, dated May 8, 2008, are provided in Tables 2 and 3 (Reference 9).

Enclosure 1 to AEP-NRC-2008-41 Page 3 Table 1 Critical Parameters for Application of Bounding Analysis Additional Evaluation Parameter Pilot Plant Basis Plant-Specific Basis Required?

Dominant PTS Transients in the NRC PTS Risk Study PTS Generalization No NRC PTS Risk Study are applicable (Reference 5) Study (Reference 6)

Through Wall Cracking Frequency 1.76E-08 Events per year 3.32E-10 Events per No (TWCF) (Reference 4) year (Calculated per Reference 5)

Frequency and Severity of Design 7 heatup/cooldowns per year Bounded by 7 No Basis Transients (Reference 4) heatup/cooldowns per year Cladding Layers (Single/Multiple) Single Layer (Reference 4) Single Layer No Additional information relative to the CNP Unit 2 RV inspection is provided in Table 2. This information confirms that satisfactory examinations have been performed on the CNP Unit 2 RV.

Table 2 Additional Information Pertaining to RV Inspection Inspection methodology: ASME Section XI and Appendix iii (Reference 7).

Number of past inspections: All welds have been inspected at least twice.

Number of indications found: One planar indication was detected in the most recent inservice inspection. The indication is acceptable in accordance with IWB-3500 of Section Xl of the ASME Code. The indication is not located in the RV beltline region. The "Allowable Number of Flaws" requirements for the proposed voluntary PTS Rule (10 CFR 50.61a) in SECY-07-0104 (Reference 8) are satisfied.

Proposed inspection schedule The third inservice inspection is currently scheduled for 2009. The third for balance of plant life: inservice inspection is proposed to be performed in 2019. The fourth inservice inspection is proposed to be performed in 2039.

to AEP-NRC-2008-41 P age 4 Table 3 provides additional information relative to the calculation of the TWCF parameter for CNP Unit 2.

Table 3 Details of TWCF Calculation at 60 EFPY Inputs Reactor Coolant System Temperature, TRcs[°F]: 1 542.55 Twa,, [inches]: 8.50 1

Un- Fluence at Region/Component Material Cu .Ni P Mn Irradiated 60 EFPY Description [wt%] [wt%] [wt%] [wt%] rradted [n/cm2, RTNDT(u) [0F] E > 1.0 MeV]

1 Int. Shell Axial Weld Linde 124 0.056 0.956 0.019 1.63 -35 1.38E+19 2 Int. Shell Axial Weld Linde 124 0.056 0.956 0.019 1.63 -35 1.38E+19 3 Low. Shell Axial Weld Linde 124 0.056 0.956 0.019 1.63 -35 9.58E+18 4 Low. Shell Axial Weld Linde 124 0.056 0.956 0.019 1.63 -35 9.58E+18 5 Int./Low. Circ Weld Linde 124 0.056 0.956 0.019 1.63 -35 3.08E+19 6 Inter. Shell Plate A 533B 0.125 0.580 0.013 1.45 38 3.08E+19 7 Inter. Shell Plate A 533B 0.150 0.570 0.014 1.45 58 3.08E+19 8 Lower Shell Plate A 533B 0.110 0.640 0.011 1.45 -20 3.08E+19 9 Lower Shell Plate A 533B 0.140 0.590 0.012 1.45 -20 3.08E+19 Outputs Methodology Used to Calculate AT3o: NUREG-1874 Controlling Material RT Fluence Region # [R]x-x [n/cm 2 , 4 (flux) AT 3o ['F] TWCF 95-xx (From E > 1.0 MeV]

Above)

Axial Weld - AW 7 634.86 1.38E+19 7.28E+09 117.17 2.71E-11 Circumferential Weld - CW 7 656.63 3.08E+19 1.63E+10 138.94 2.76E-14 Plate - PL 7 656.63 3.08E+19 1.63E+10 138.94 1.15E-10 TWVCF95-TOTAL ((xAwTWCF95-AW + cLPLTWCF95-PL + (xcwTWCF95-cw): 3.32E-10 Note 1: Values are for fluence at the cladding-to-base metal interface.

to AEP-NRC-2008-41 Page 5

6. Duration of Proposed Alternative This request is applicable to the CNP Unit 2 inservice inspection program up to the end of the license period including the period of extended operation, through December 23, 2037.

7. References

1. ASME Boiler and Pressure Vessel Code,Section XI, 1989 Edition, American Society of Mechanical Engineers, New York.

2. OG-06-356, "Plan for Plant Specific Implementation of Extended Inservice Inspection Interval per WCAP-16168-NP, Revision 1, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval." MUHP 5097-99, Task 2059," October 31, 2006.

3. NRC Regulatory Guide 1.174, Revision 1, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," November 2002.

4. WCAP-16168-NP-A, Revision 2, "Risk-Informed Extension of Reactor Vessel In-Service Inspection Interval," June 2008.

5. NUREG-1874, "Recommended Screening Limits for Pressurized Thermal Shock,"

March, 2007.

6. NRC Letter Report, "Generalization of Plant-Specific Pressurized Thermal Shock (PTS)

Risk Results to Additional Plants," December 14, 2004.

7. ASME Section XI, 1995 Edition, with 1996 Addenda.

8. SECY-07-0104, "Proposed Rulemaking - Alternate Fracture Toughness Requirements for Protection against Pressurized Thermal Shock," Enclosure 1, June 25, 2007.

9. NRC Safety Evaluation to WCAP-16168-NP, Revision 2 dated May 8, 2008, titled "Final Safety Evaluation for Pressurized Water Reactor Owners Group (PWROG) Topical Report (TR) WCAP-16168-NP, Revision 2, Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval (TAC No. MC9768)."

to AEP-NRC-2008-41 Page 1 Proposed License Amendment Regarding ASME Relief Request Information

1. DESCRIPTION Pursuant to 10 CFR 50.90, Indiana Michigan Power Company (I&M) hereby requests an amendment to the Donald C. Cook Nuclear Plant Unit 2 license, DPR-74. I&M has requested a Reactor Vessel Inservice Inspection (ISI) Relief Request for Unit 2 based on the Nuclear Regulatory Commission (NRC) approved Topical Report (TR) WCAP-16168-NP-A, Revision 2, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval." The NRC safety evaluation (SE) report approving the WCAP, required licensees requesting the relief to the reactor vessel ten-year inspection program to submit a request to amend the license. The purpose of this request is to comply with that requirement.

2. PROPOSED CHANGE The proposed change to Unit 2 license will add item (ee) to Section 2.C(3), Additional Conditions, that will read as follows:

(ee) I&M shall provide the NRC with the information and'analysis requested in Section (e) of the final 10 CFR 50.61a (or the proposed 10 CFR 50.61a, given in 72 FR 56275, prior to issuance of the final 10 CFR 50.61a) following completion of each ASME Code,Section XI, Category B-A and B-D Reactor Vessel weld inspection. The information must be submitted within one year of the inspection.

3. BACKGROUND The Pressurized Water Reactor Owners Group (PWROG) submitted TR WCAP-16168-NP, Revision 1, "Risk-Informed Extension of the Reactor Vessel In-Servce Inspection Interval," to the NRC staff by letter dated January 26, 2006, and supplemented by letter dated June 8, 2006.

PWROG letter dated October 16, 2007, submitted TR WCAP-16168-NP, Revision 2, and responses to the NRC staff's request for additional information (RAI) for NRC staff review. An NRC draft SE regarding approval of TR WCAP-16168-NP, Revision 2, was provided to the PWROG for review and comments by letter dated March 6, 2008. Comments were provided by the PWROG by letter dated March 31, 2008.

The NRC issued a final SE and approval of TR WCAP-16168-NP, Revision 2, by letter dated May 8, 2008, as identified in section 6 of this attachment. The NRC staff's disposition of PWROG comments on the draft SE are discussed in an attachment to the May 8, 2008, letter.

The SE attached to the May 8, 2008, letter identifies the information requirements to be included in the relief request and requires that licensees submit the information and analyses requested in Section (e) of the final 10 CFR 50.61a (or the proposed 10 CFR 50.61a, given in 72 FR 56275, prior to issuance of the final 10 CFR 50.61a) within one year of completing each American Society of Mechanical Engineers (ASME) Code,Section XI, Category B-A and B-D Reactor Vessel weld inspection. To administratively control the submission of this information the SE also requires that "Licensees that do not implement 10 CFR 50.61a must amend their licenses to require that the information and analyses requested in Section (e) of the final 10 CFR 50.61a (or the proposed 10 CFR 50.61a, given in 72 FR 56275, prior to issuance of the to AEP-NRC-2008-41 Page 2 final 10 CFR 50.61a) will be submitted for NRC staff review and approval. The amendment to the license shall be submitted at the same time as the request for alternative." I&M is not implementing 10 CFR 50.61a since the rule is not final. This amendment request implements the requirement to submit a license amendment request at the time of submitting a request for the alternative.

4. TECHICAL ANALYSIS The addition of a license condition to require the submittal of the information and analysis requested in Section (e) of the final 10 CFR 50.61a (or the proposed 10 CFR 50.61a, given in 72 FR 56275, prior to issuance of the final 10 CFR 50.61a) following completion of each ASME Code,Section XI, Category B-A and B-D Reactor Vessel weld inspection is an administrative change with no effect on the public safety. The change provides the NRC assurance that I&M will submit defined information and analyses to the NRC every time that a specific ISl is done.

The relief request to extend the Unit 2 IS[ from 10 to 20 years is separate from this license change and is reviewed and approved independently.

5. REGULATORY ANALYSIS 5.1 No Significant Hazards Consideration Indiana Michigan Power Company (I&M) has evaluated the safety significance of the proposed change regarding the addition of a license condition to submit the information and analysis requested in Section (e) of the 10 CFR 50.61a (or the proposed 10 CFR 50.61 a, given in 72 FR 56275, prior to issuance of the final 10 CFR 50.61a) following completion of each ASME Code,Section XI, Category B-A and B-D Reactor Vessel weld inspection according to the criteria of 10 CFR 50.92, "Issuance of Amendment." I&M has determined that the subject change does not involve a Significant Hazards Consideration as discussed below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No. The proposed change will revise the license to require the submission of information and analyses to the Nuclear Regulatory Commission (NRC) following completion of each American Society of Mechanical Engineers (ASME) Code, Section Xl, Category B-A and B-D Reactor Vessel weld inspection. Submittal of the information and analyses can have no effect on the consequences of an accident or the probability of an accident because the submission of information is not related to the operation of the plant or any equipment, the programs and procedures used to operate the plant, or the evaluation of accidents Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

to AEP-NRC-2008-41 Page 3 Response: No. The proposed change will only affect the requirement to submit information and analyses when specified inspections are performed. There are no changes to plant equipment, operating characteristics or conditions, programs or failures. There are no new accident initiators or precursors.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No. The proposed change will revise the license to require the submission of information and analyses to the NRC following completion of each ASME Code,Section XI, Category B-A and B-D Reactor Vessel weld inspection which does not affect any Limiting Conditions for Operation used to establish the margin of safety. The requirement to submit information and analyses is an administrative tool to assure the NRC has the ability to independently review information developed by the licensee. The proposed change does not involve a significant reduction in the margin of safety.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, I&M concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

5.2 Applicable Regulatory Requirements/Criteria The proposed change has been reviewed to evaluate the potential effect on regulatory requirements and criteria. There are no rules and regulations requiring the submittal of information and analyses to NRC regarding NRC ASME Code, Section Xl, Category B-A and B-D Reactor Vessel weld inspection. 10 CFR 50.55a compliance with ASME Code Section Xl, requires a summary of these inspections. The information and analyses of Section (e) of the proposed 10 CFR 50.61a defines requirements for verifying that the pressurized thermal shock screening criteria of the proposed rule are applicable to the reactor vessel. The final rule will be the same or modified as a result of comments. The amendment is the administrative means chosen by the NRC staff to obtain this information.

5.3 Environmental Considerations The proposed amendment relates to changes in reporting requirements. Accordingly, these amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(10). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

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6. REFERENCES

1. Letter from Ho K. Nieh, NRC, to Gordon Bischoff, WOG, regarding Final Safety Evaluation for PWROG TR WCAP-16168-NP, Revision 2, (TAC NO. MC9768), dated May 8, 2008 (ML081060051).