ML082880033

License Amendment Request: Reporting of Reactor Vessel In-Service Inspection Information and Analyses in Support of Code Relief Request for Extension of Reactor Vessel In-Service Inspection Interval

ML082880033
Person / Time
Site:	Ginna
Issue date:	10/07/2008
From:	John Carlin Constellation Energy Nuclear Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
WCAP-16168-NP, Rev 2
Download: ML082880033 (14)

Text

John Carlin R.E. Ginna Nuclear Power Plant, LLC Site Vice President 1503 Lake Road Ontario, New York 14519-9364 585.771.5200 585.771.3943 Fax John.carlin @constellation.com Constellation Energy Nuclear Generation Group October 7, 2008 U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION: Document Control Desk

SUBJECT:

R.E. Ginna Nuclear Power Plant Docket No. 50-244 License Amendment Request: Reporting of Reactor Vessel In-Service Inspection Information and Analyses in Support of Code Relief Request for Extension of Reactor Vessel In-Service Inspection Interval

REFERENCES:

(a) Federal Register Notice (72 FR 56275), dated October 3, 2007, Alternative Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events (b) Letter from Mr. J.E. Pacher (Ginna LLC) to Document Control Desk (NRC), dated October 3, 2008, Fourth Ten-Year Interval Inservice Inspection Program Re-submittal of Relief Request Number 18 (c) WCAP-16168-NP, Revision 2, dated June 2008, Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval (d) Final Safety Evaluation for Pressurized Water Reactor Owners Group Topical Report WCAP-16168-NP, Revision 2, dated May 8, 2008, Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval Pursuant to 10 CFR 50.90, the R.E. Ginna Nuclear Power Plant, LLC (Ginna LLC) hereby requests an amendment to the Renewed Operating License No. DPR-18 to require the reporting of specified Reactor Vessel In-Service Inspection information and analyses as specified in Reference (a). This amendment request is a required part of a code relief request (Reference b) to extend the Reactor Vessel In-Service Inspection 10-year inspection interval. The code relief AO4-7

Document Control Desk October 7, 2008 Page 2 request is based on the analysis contained within Reference (c). The methodology in Reference (c) was evaluated by the Nuclear Regulatory Commission (NRC), in Reference (d), as being acceptable for referencing in license amendment requests. Reference (d) requires that licensees, which do not implement 10 CFR 50.61a, must amend their licenses to require the submittal of the information and analyses requested in paragraph (e) of the final 10 CFR 50.61a [or the proposed 10 CFR 50.61a, as listed in Reference (a), prior to the issuance of the final 10 CFR 50.61a]. Currently, Ginna LLC has not implemented 10 CFR 50.61a because the final rule has not been approved. Therefore, in accordance with Reference (d), we are requesting to amend our operating license as described in Attachment (1). Attachment (2) contains the marked up Operating License pages.

There are no new regulatory commitments identified as a result of this submittal.

Ginna LLC requests approval of this proposed amendment by April 3, 2009 in order to support planning for the fall 2009 refueling outage. In addition, Ginna LLC requests an implementation period of 60 days.

Document Control Desk October 7, 2008 Page 3 Should you have questions regarding this matter, please contact Mr. David F. Wilson at (585) 771-5219.

Very truly yours, STATE OF NEW YORK TO WIT:

COUNTY OF WAYNE I, John Carlin, being duly sworn, state that I am Vice President, R.E. Ginna Nuclear Power Plant, LLC (Ginna LLC), and that I am duly authorized to execute and file this request on behalf of Ginna LLC. To the best of my knowledge and belief, the statements contained in this document are true and correct. To the extent that these statements are not based on my personal knowledge, they are based upon information provided by other Ginna LLC employees and/or consultants. Such information has been reviewed in accordance with company practice and I believe it to be reliable.

Subscribed and sworn before me, a Notary Public in pnd, for th State of New York and County of **n.Oa* ,this I day of L1O. i.9L , 2008.

WITNESS my Hand and Notarial'Seal:

Notary Pubi ".-

-HAIRONL.MILLER " "

Notary Public;'State of'New York*.>'

My Commission Expires: MonrýCo flnty, ¢, ."

• e, . L Attachments: (1) Evaluation of the Proposed Change (2) Marked up License Page cc: D. V. Pickett, NRC Resident Inspector, NRC S. J. Collins, NRC P.D. Eddy, NYSDPS J. P. Spath, NYSERDA

ATTACHMENT (1)

EVALUATION OF THE PROPOSED CHANGE TABLE OF CONTENTS 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 Significant Hazards Consideration 4.4 Conclusions

6.0 REFERENCES

R.E. Ginna Nuclear Power Plant, LLC October 7, 2008

ATTACHMENT (1)

EVALUATION OF THE PROPOSED CHANGE 1.0

SUMMARY

DESCRIPTION This letter requests an amendment to Renewed Operating License DPR- 18 for the R.E. Ginna Nuclear Power Plant (Ginna). The proposed change will insert into the license a requirement involving the reporting of specified Reactor Vessel (RV) In-Service Inspection (ISI) information and analyses as specified in Reference (1). This amendment request is a required element of code relief request numberl8 (Reference 2) to extend the RV ISI 10-year inspection interval.

The code relief request is based on the analysis contained within Reference (3). The methodology in Reference (3) was evaluated by the Nuclear Regulatory Commission (NRC), in Reference (4), as being acceptable for referencing in license amendment requests. Reference (4) requires that licensees desiring to extend the frequency of the RV ISI interval, which do not implement 10 CFR 50.61 a, must amend their licenses to require the submittal of information and analyses requested in paragraph (e) of the final 10 CFR 50.61a [or the proposed 10 CFR 50.61a, as listed in Reference (1), prior to the issuance of the final 10 CFR 50.61a]. Currently, Ginna has not implemented 10 CFR 50.61a because the final rule has not yet been approved. Therefore, in accordance with Reference (4), we are requesting to amend our licenses as described below.

Acceptance of this amendment request would allow the RV ISI interval to be extended such that Ginna's fourth inservice inspection would be completed during the 2011 refueling outage and its fifth interval inspection would be completed during the 2031 refueling outage. This is consistent with the proposed inspection schedule in Reference (5).

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ATTACHMENT (1)

EVALUATION OF THE PROPOSED CHANGE 2.0 DETAILED DESCRIPTION The Ginna proposed change will add, as item (10) to Section 2.C of the Ginna license, the following requirement:

Reactor Vessel In-Service InspectionInterval Extension As a requiredpart of the Relief Request to extend the reactorvessel In-Service Inspection interval, Ginna LLC shall submit to the NRC the information and analyses specified in paragraph(e) of the proposed 10 CFR 50.61a, given in 72 FR 56275,following completion of each American Society of MechanicalEngineers (ASME) Code,Section XI, Category B-A and B-D reactor vessel weld inspection. The information and analyses shall be submitted within one year of the completion of each inspection. Once the final 10 CFR 50.61a is effective, the requiredinformation and analyses specified in paragraph(e) of the final 10 CFR 50.61a shall be the information submitted to the NRC.

This proposed change implements the requirement to submit a license amendment request at the time of submitting a relief request as specified in Reference (4).

3.0 TECHNICAL EVALUATION

The proposed change is administrative in nature and is limited to the submittal of requested information and analyses that provide the ASME code required RV ISI results. Submittal of the data is necessary to demonstrate the RV maintains its ability to meet code fracture toughness requirements and is used to justify continued use of the approved code relief request for an extended RV ISI interval. Since the proposed change involves only a requirement for the submittal of information, it clearly does not have any impact on the operation of the plant or on 2

ATTACHMENT (1)

EVALUATION OF THE PROPOSED CHANGE any design basis accident. The proposed change cannot have a significant effect on the margin of safety because it is not related to any margin of safety.

The relief request to extend the RV ISI interval is separate from this license change and is reviewed separately.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria This amendment request is a required part of a code relief request submitted to extend Ginna' s RV ISI 10-year inspection interval. Reference (4) provides the requirements a station must meet in order for the NRC to approve such an alternative to Section XI of the ASME code. This amendment request meets the requirement, listed in Section 3.4 of Reference (4), that the licensee will submit identified information and analyses to the NRC.

The proposed change, to report results from RV examinations required by Section XI of the ASME code, is being submitted in support of demonstrating Ginna's ability to justify extending its RV ISI interval frequency.

4.2 Precedent While there is no existing approved precedent for this proposed change, Indian Point Units 2 and 3 and Palisades, have recently submitted license amendment requests for this proposed change that are similar in content and approach. Submittal of this proposed change is consistent with the requirements stated in Reference (4).

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ATTACHMENT (1)

EVALUATION OF THE PROPOSED CHANGE 4.3 Significant Hazards Consideration Ginna is submitting a code relief request from American Society of Mechanical Engineers (ASME) code,Section XI that will allow Ginna to extend the frequency of the Reactor Vessel (RV) In-Service Inspection (ISI) beyond the required 10-year interval. The code relief request is based on the analysis of Topical Report WCAP-16168-NP, Revision 2 which was found to be an acceptable approach by a Nuclear Regulatory Commission (NRC) Safety Evaluation. The Safety Evaluation required that utilities not implementing 10 CFR 50.61a (Ginna is not implementing 10 CFR 50.6 la at this time) must submit a license amendment request that requires the submittal of information and analyses specified in Section (e) of 10 CFR 50.61a (or the proposed 10 CFR 50.61a, given in 72 FR 56275, for the time period prior to the issuance of the final 10 CFR 50.6 l a). This proposed change adds a statement in the Ginna license to provide the required information and analyses.

Ginna has evaluated whether or not a significant hazards consideration is involved with the proposed change by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment" as discussed below:

1. Operation of the facility in accordancewith the proposed amendment would not involve a significant increase in the probability or consequences of an accidentpreviously evaluated The proposed change, which adds a requirement within the Ginna license, to provide required information and analyses as a supporting condition for extending the allowed reactor vessel ISI interval, only involves the commitment to provide data obtained from the reactor vessel ISI. This proposed change involves only the submittal of generated data that will be used to verify the reactor vessel has more than sufficient margin to prevent any pressurized thermal shock event from occurring. This proposed change does not involve 4

ATTACHMENT (1)

EVALUATION OF THE PROPOSED CHANGE any change to the design basis of the plant or of any structure, system, or component.

Therefore, the proposed change does not involve a significant increase in the probability or consequence of an accident previously evaluated.

2. Operation of the facility in accordancewith the proposed amendment would not create the possibility of a new or different kind of accidentfrom any accidentpreviously evaluated.

The proposed change, which adds a requirement within the Ginna license to provide required information and analyses as a supporting condition for extending the reactor vessel ISI interval, only involves the commitment to provide data and analyses obtained from the reactor vessel ISI. As such this proposed change does not result in physical alteration to the plant configuration or make any change to plant operation. As a result no new accident scenarios, failure mechanisms, or single-failures are introduced. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Operation of the facility in accordancewith the proposed amendment would not involve a significant reduction in a margin of safety.

The proposed change, which adds a requirement within the Ginna license, to provide required information and analyses as a supporting condition for extending the allowed reactor vessel ISI interval, only involves the commitment to provide data and analyses obtained from the reactor vessel ISI. The submitted data will be used to verify the condition of the reactor vessel meets all required standards to ensure sufficient safety margin is maintained against the occurrence of a pressurized thermal shock event during the expanded time interval between reactor vessel ISIs. The proposed change is administrative in nature and is not related to any margin to safety. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

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ATTACHMENT (1)

EVALUATION OF THE PROPOSED CHANGE Based upon the above evaluation, Ginna has concluded that the proposed amendment does not involve a 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.

4.4 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission' regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

The proposed amendment is confined to (i) changes to surety, insurance and/or indemnity requirements, or (ii) changes to recordkeeping, reporting or administrative procedures, or requirements. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(10). Therefore, pursuant to 10 CFR 51.22(b),

no environmental impact statement or environmental assessment need be prepared in conjunction with the proposed amendment.

6.0 REFERENCES

(1) Federal Register Notice (72 FR 56275), dated October 3, 2007, Alternative Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events (2) Letter from Mr. J.E. Pacher (Ginna LLC) to Document Control Desk (NRC), dated October 3, 2008, Fourth Ten-Year Interval Inservice Inspection Program Re-submittal of Relief Request Number 18 (3) WCAP-16168-NP, Revision 2, dated June 2008, Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval 6

ATTACHMENT (1)

EVALUATION OF THE PROPOSED CHANGE (4) Final Safety Evaluation for Pressurized Water Reactor Owners Group Topical Report WCAP-16168-NP, Revision 2, dated May 8, 2008, Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval (5) PWR Owners Group letter 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," dated October 31, 2006 7

ATTACHMENT (2)

MARKED UP LICENSE PAGE Renewed Operating License Page 5 R.E. Ginna Nuclear Power Plant, LLC October 7, 2008

.5-

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 personnel 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 (9) Control Room Envelope Habitability Upon Implementation of Amendment No. 105 adopting TSTF-448, Revision 3, the determination of control room envelope (CRE) unfiltered air inleakage as required by SR 3.7.9.4, in accordance with TS 5.5.16.c.i and the assessment of CRE habitability as required by 5.5.16.c.0i, shall be considered met. Following Implementation:

(a) The first performance of SR 3.7.9.4 in accordance with Specification 5.5.16.c.i shall be within the specified Frequency of 6 years, plus the 18-month allowance of SR 3.0.2, as measured from February 8, 2005, the date of the most recent successful tracer gas test, as stated In the April 6, 2007 letter response to Generic Letter 2003-01, or within the next 18 months if the time period since the most recent tracer gas test is greater than 6 years.

(b) The first performance of the periodic assessment of CRE habitability, Specification 5.5.16.c.ii, shall be within 3 years,' plus A. the 9-month allowance of SR 3.0.2 as measured from February 8, 2005, the date of the most recent successful tracer gas test, as stated in the April 6, 2007 letter response to Generic Letter 2003-01, or within the next 9 months Ifthe time period since the

--most recent successful tracer gas test is greater than 3 years.

D. The facility requires an exemption from certain requirements of 10 CFR 50.46(a)(1). This Includes an exemption from 50.46(a)(1), that emergency core cooling system (ECCS) performance be calculated In Amendment No. 105

Insert "A" (10) Reactor Vessel In-Service Inspection Interval Extension As a required part of the Relief Request to extend the reactor vessel In-Service Inspection interval, Ginna LLC shall submit to the NRC the information and analyses specified in paragraph (e) of the proposed 10 CFR 50.61a, given in 72 FR 56275, following completion of each American Society of Mechanical Engineers (ASME) Code,Section XI, Category B-A and B-D reactor vessel weld inspection. The information and analyses shall be submitted within one year of the completion of each inspection. Once the final 10 CFR 50.61a is effective, the required information and analyses specified in paragraph (e) of the final 10 CFR 50.61a shall be the information submitted to the NRC.