W3F1-2008-0061, License Amendment Request to Support Implementation of Extended In-Service Inspection Interval

License Amendment Request to Support Implementation of Extended In-Service Inspection Interval

ML082660037
Person / Time
Site:	Waterford
Issue date:	09/17/2008
From:	Walsh K Entergy Nuclear South, Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
W3F1-2008-0061
Download: ML082660037 (11)

Text

E~nter&g Entergy Nuclear South Entergy Operations, Inc.

17265 River Road Killona, LA 70057 Tel 504 739 6660 Fax 504 739 6678 kwalshl @entergy.com Kevin T. Walsh Vice President, Operations Waterford 3 W3F1-2008-0061 September 17, 2008 U. S. Nuclear Regulatory Commission Attn.: Document Control Desk Washington, DC 20555-0001

REFERENCE:

License Amendment Request NPF-38-280 License Condition to Support Implementation of Extended In-Service Inspection Interval Waterford Steam Electric Station, Unit 3 Docket No. 50-382 License No. NPF-38

1.

Entergy letter dated September 18, 2008, "Request for Alternative W3-ISI-006 Proposed Alternative to Extend the Second 10-Year Inservice Inspection Interval for Reactor Vessel Internal Weld Examinations (W3F1-2008-0060)

2.

Letter from NRC to Mr. Gordon Bischoff, Manager Owners Group Program Management Office, "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),"

dated May 8, 2008

Dear Sir or Madam:

In accordance with the provisions of 10 CFR 50.90, Entergy Operations, Inc. (Entergy) hereby requests an amendment to Waterford Steam Electric Station, Unit 3 (Waterford 3)

Operating License (OL). Entergy has requested a Reactor Vessel Inservice Inspection Relief Request (Reference 1) based on topical report WCAP-16168-NP-A, Revision 2, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval." In Reference 2, the NRC safety evaluation for the topical report required licensees to amend the license to provide the NRC with the information and analyses requested in Section (e) of the final rule for 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) within one year following completion of each ASME Code,Section XI, Category B-A and B-D weld inspection.

W3F1 -2008-0061 Page 2 provides a description of the proposed change, the requested confirmation of applicability, and plant-specific verifications. Attachment 2 provides the OL page marked up to show the proposed change.

The proposed change has been evaluated in accordance with 10 CFR 50.91(a)(1) using criteria in 10 CFR 50.92(c) and it has been determined that this change involves no significant hazards consideration. The bases for these determinations are included in the attached submittal.

Entergy requests approval of the proposed amendment by September 10, 2009, in order to support the Fall 2009 refueling outage. Once approved, the amendment shall be implemented within 60 days. Although this request is neither exigent nor emergency, your prompt review is requested.

There are no regulatory commitments contained in this submittal.

If you have any questions or require additional information, please contact Robert Murillo, Manager, Licensing at (504) 739-6715.

I declare under penalty of perjury that the foregoing is true and correct. Executed on September 17, 2008.

Sincerely, KTW/RLW/ssf Attachments:

1. Analysis of Proposed Operating License Change

2. Proposed Operating License Change (mark-up)

W3F1 -2008-0061 Page 3 cc:

Mr. Elmo E. Collins Regional Administrator U. S. Nuclear Regulatory Commission Region IV 612 E. Lamar Blvd., Suite 400 Arlington, TX 76011-8064 NRC Senior Resident Inspector Waterford 3 P. 0. Box 822 Killona, LA 70066-0751 U. S. Nuclear Regulatory Commission Attn: Mr. N. Kalyanam MS 0-07 D1 Washington, DC 20555-0001 American Nuclear Insurers Attn: Library 95 Glastonbury Blvd.

-Suite 300 Glastonbury, CT 06033-4443 Wise, Carter, Child & Caraway Attn: J. Smith P.O. Box 651 Jackson, MS 39205 Louisiana Department of Environmental Quality Office of Environmental Compliance Surveillance Division P. O. Box 4312 Baton Rouge, LA 70821-4312 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 To W3FI-2008-0061 Analysis of Proposed Operating License Change to W3F1-2008-0061 Page 1 of 4

1.0 DESCRIPTION

Entergy Operations, Inc. (Entergy) requests an amendment to Waterford Steam Electric Station, Unit 3 (Waterford 3) Operating License (OL). The license amendment is needed to support a proposed change to the inservice inspection program that is based on topical report (TR) WCAP-16168-NP-A, Revision 2, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval.". In the NRC safety evaluation (SE) for the TR, the NRC required.

licensees to amend the license to provide the NRC with the information and analyses requested in Section (e) of the final rule for 10 CFR 50.61a, (or the proposed 10 CFR 50.61a, given in 72 FR 56275, prior to issuance of the final 10 CFR 50.61 a) within one year following completion of each ASME Code, Section Xl, Category B-A and B-D weld inspection. Entergy proposes to add a new license condition to provide this information. The NRC also required licensees to submit a request for a proposed alternative, conforming with 10 CFR 50.55a(a)(3)(i), to extend the inservice inspection interval. Entergy is submitting that request for alternative under separate cover, but concurrently with this proposed license amendment request.

2.0 PROPOSED CHANGE

The proposed change will add the following new license condition 2.C item 19:

19.

Reactor Vessel Inservice Inspection Interval Extension Provide the NRC with the information and analyses requested in Section (e) of the final rule for 10 CFR.50.61 a, (or the proposed 10 CFR 50.61 a, given in 72 FR 56275, prior to issuance of the final 10 CFR 50.61 a) within one year following completion of each ASME Code,Section XI, Category B-A and B-D weld inspection for review and approval.

3.0 BACKGROUND

The Pressurized Water Reactor Owners Group (PWROG) submitted Topical Report (TR)

WCAP-16168-NP, Revision 1, "Risk-Informed Extension of the Reactor Vessel In-Service 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. A NRC draft Safety Evaluation (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 via letter dated March 31, 2008.

The NRC issued a final safety evaluation (SE) and approval of TR WCAP-16168-NP, Revision 2 by letter dated May 8, 2008. The NRC Staff's disposition of PWROG comments on the draft SE are also discussed in an attachment to the May 8, 2008 letter. The final SE identified the information requirements to be included in the relief request and requirements for licensees that do not implement 10 CFR 50.61a to amend their licenses to require submittal of information and analyses requested in Section (e) of the final rule for 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 50.61 a) within one year following completion of each ASME Code,Section XI, Category B-A and B-D weld inspection. This information and analyses will be submitted for NRC Staff review and approval.

to W3F1 -2008-0061 Page 2 of 4 The amendment to the license shall be submitted at the same time as the request for alternative. Entergy is not implementing 10 CFR 50.61a since the rule is not final. Therefore, this amendment request satisfies the requirement to submit a license amendment concurrently with submittal of the request for alternative.

4.0 TECHNICAL ANALYSIS

TR WCAP-16168-NP-A, Revision 2, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval," has been approved by the NRC (Reference 5). Entergy meets the Limitations and Conditions contained in the NRC SE as follows:

Within one year of completing each of the ASME Code,Section XI, Category B-A and B-D

[reactor vessel] RV weld inspections required in the proposed ISI interval, the licensee must provide the information and analyses requested in Section (e) of the final 10 CFR 50.61 a (or Alternative fracture toughness requirements for protection against pressurized thermal shock, in Enclosure 1 to'the proposed rulemaking in SECY-07-0104, Reference 12, given in 72 FR 56275 prior to issuance of the final 10 CFR 50.61a). 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 50.61 a (or the proposed 10 CFR 50.61 a, given in 72 FR 56275 prior to issuance of the final 10 CFR 50.61 a) 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.

The proposed license condition 2.C item 19 in this amendment request includes all required elements listed.

The submittal of the proposed request for alternative, pursuant to 10 CFR 50.55a(a)(3)(i), and the proposed license condition in this amendment request conforms to the conditions and limitations imposed by the NRC SE for WCAP-16168-NP-A, Revision 2 (Reference 4).

Therefore, these proposed changes are administrative in nature and will have no adverse impact on plant safety.

5.0 REGULATORY ANALYSIS

5.1 Applicable Requlatory Requirements/Criteria The proposed change has been evaluated to determine whether applicable regulations and requirements continue to be met. The proposed change does not require relief from other regulatory requirements and does not affect conformance with any General Design Criterion differently than described in the Final Safety Analysis Report. There are no rules and regulations requiring the submittal of information and analyses to NRC regarding ASME Code, Section Xl, Category B-A and B-D reactor vessel weld inspections. The information and analyses of Section (e) of the proposed 10 CFR 50.61 a defines the 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 an administrative means selected by the NRC staff to obtain the required information.

to W3F1`-2008-0061 Page 3 of 4 5.2 No Significant Hazards Consideration Entergy Operations, Inc. has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," 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 NRC following completion of each ASME Code, Section Xl, Category B-A and B-D reactor vessel weld inspection. The extension of the ISI interval from 10 to 20 years is being evaluated as part of the relief request independent from this license change. Submission of the information and analyses are administrative in nature and has no impact on any plant configuration or system performance relied upon to mitigate the consequences of an accident.

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?

Response: No.

The proposed change does not involve a physical alteration of any SSC or change the way any SSC is operated. The proposed addition of the license condition has no impact on any plant configurations or on system performance that is relied upon to mitigate the consequences of an accident. The license condition is administrative in nature and does not result in a change to the physical plant or to the modes of operation defined in the facility license. Entergy has demonstrated that the Limitations and Conditions associated with the NRC SE will be met.

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 addition of the license condition is administrative in nature and has no impact on plant operation or equipment or on any margin of safety. The license condition to submit information and analyses is an administrative tool to assure the NRC has the ability to independently review information developed by the Licensee.

to W3F1-2008-0061 Page 4 of 4 Therefore, the proposed change does not involve a significant reduction in a margin of safety.

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

5.3 Environmental Considerations The proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 PRECEDENCE None.

7.0.

REFERENCES

1. Letter from F. P. Schiffley, Westinghouse Owners' Group, "Transmittal of WCAP-16168-NP, Revision 1, 'Risk-Informed Extension of Reactor Vessel In-Service Inspection Interval', MUHP-5097/5098/5099, Tasks 2008/2059," January 26, 2006

2.

Letter from F. P. Schiffley, PWR Owners Group, "Evaluation of NRC Questions on the Technical Bases for Revision of the PTS Rule Relative to Their Effects on the Risk Results in WCAP-16168-NP, Revision 1, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval," June 8, 2006

3. Letter from F. P. Schiffley, PWR Owners Group, "Responses to the NRC Request for Additional Information (RAI) on PWR Owners' Group (PWROG) WCAP-16168-NP, Revision 1, 'Risk-Informed Extension of Reactor vessel In-Service Inspection Interval', MUHP-5097/5098/5099, Tasks 2008/2059," October 16, 2007, and, RAI responses. Enclosure 2, WCAP-16168-NP, Revision 2, 'Risk-Informed Extension of Reactor vessel In-Service Inspection Interval', October 2007

4. Letter from NRC to Mr. Gordon Bischoff, Manager Owners Group Program Management Office, "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),"

May 8, 2008

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

n)

To W3FI -2008-0061 Proposed Operating License Change (mark-up) to W3F1 -2008-0061 Page 1 of 2 d.

Prior to completion of Phase III of the Waterford 3 startup test program, the licensee shall complete corrective actions related to the 23 NRC issues as identified in the LP&L responses.

17.

Basemat The licensee shall comply with its commitments to perform a basemat cracking surveillance program and additional confirmatory analyses of basemat structural strength as described in its letter of February 25, 1985. Any significant change to this program shall be reviewed and approved by the NRC staff prior to its implementation.

18.

Mitigation Strategy License Condition Develop and maintain strategies for addressing large fires and explosions and that include the followi ng key areas:

(a)

Fire fighting response strategy with the following elements:

1.

Pre-defined coordinated fire response strategy and Guidance

2.

Assessment of mutual aid fire fighting assets

3.

Designated staging areas for equipment and materials

4.

Command and control

5.

Training of response personnel (b)

Operations to mitigate fuel damage considering the following:

1.

Protection and use of pers onnel assets

2.

Communications

3.

Minimizing fire spread

4.

Procedures for implementing integrated fire response strategy

5.

Identification of readily-available pre-staged equipment

6.

Training on integrated fire response strategy

7.

Spent fuel pool mitigation measures (c)

Actions to minimize release to include consideration of:

1.

Water spray scrubbing

2.

Dose to onsite responders TJ D.

The facility requires an exemption from certain requirements of Appendices E and' J to 10 CFR Part 50. These exemptions are described in the Office of Nuclear Reactor Regulation's Safety Evaluation Report, Supplement No. 10 (Section 6.1.2) and Supplement No. 8 (Section 6.2.6), respectively. These exemptions are authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest. These exemptions are, therefore, hereby granted pursuant to 10 CFR 50.12. With the granting of these exemptions, the facility will operate, to the extent authorized herein, in conformity with the application, as amended, the provisions of the Act, and the rules and AMENDMENT NO. 434,263 Revised by letter dated 2, 2604 Revised by letter dated July 26, 2007 to W3F1 -2008-0061 Page 2 of 2 Insert 1

19.

Reactor Vessel Inservice Inspection Interval Extension Provide the NRC with the information and analyses requested in Section (e) of the final rule for 10 CFR 50.61 a, (or the proposed 10 CFR 50.61 a, given in 72 FR 56275, prior to issuance of the final 10 CFR 50.61 a) within one year following completion of each ASME Code, Section Xl, Category B-A and B-D weld inspection for review and approval.