ML022770267

Response to Request for Additional Information Regarding Relief Request RR 3-28 for Risk-Informed Inservice Inspection

ML022770267
Person / Time
Site:	Indian Point
Issue date:	10/01/2002
From:	Kansler M Entergy Nuclear Northeast, Entergy Nuclear Operations, Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
IPN-02-079, TAC MB4637
Download: ML022770267 (6)

Text

SEn ter~gy Entergy Nuclear Northeast Entergy Nuclear Operations, Inc 440 Hamilton Avenue White Plains, NY 10601 Tel 9142723200 Fax 914 272 3205 Michael R. Kansler Senior Vice President &

Chief Operating Officer October 1, 2002 IPN-02-079 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop O-Pl-17 Washington, DC 20555-0001

Reference:

Indian Point Nuclear Generating Unit No. 3 Docket No. 50-286 Response to Request for Additional Information Regarding Relief Reauest RR 3-28 for Risk-Informed Inservice Inspection

1. NRC Letter, "Request for Additional Information Regarding Relief Request to Use Risk-Informed Inservice Inspection, Indian Point Nuclear Generating Unit No. 3 (TAC No. MB4637)", dated July 25, 2002

2. Entergy letter to USNRC, IPN-02-007, "Relief Request RR 3-28, Risk Informed Inservice Inspection (RI-ISI) Program", dated February 5, 2002

Dear Sir:

This letter provides the additional information (Attachment 1) as requested by the NRC in Reference 1 regarding the relief request submitted by Entergy Nuclear Operations Inc. (ENO) in Reference 2.

Approval for RR 3-28 is requested by December 16, 2002 to support implementation for the upcoming refueling outage (3R12) scheduled to begin in March 2003.

N,,,

There are no new commitments made in this letter. If you have any questions, please contact Ms. Charlene Faison at 914-272-3378.

Very truly yours Michael R.

nsler Senior Vice President and Chief Operating Officer

Attachment:

1. Response to Request for Additional Information Regarding Relief Request RR 3-28, Rev. 0 for Risk-Informed Inservice Inspection cc:

Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Resident Inspector's Office Indian Point Unit 3 U.S. Nuclear Regulatory Commission P.O. Box 337 Buchanan, NY 10511 Mr. Patrick Milano, Project Manager Project Directorate I Division of Licensing Project Management U.S. Nuclear Regulatory Commission Mail Stop 0-8-C2 Washington, DC 20555

ATTACHMENT 1 TO IPN-02-079 INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 Response to Request for Additional Information Regarding Relief Request RR 3-28, Rev. 0 for Risk-Informed Inservice Inspection ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 DOCKET NO. 50-286 DPR-64

Attachment I to IPN-02-079 Response to Request for Additional Information Regarding Relief Request RR 3-28 for Risk-Informed Inservice Inspection Indian Point Nuclear Generating Unit No. 3

1. Section 3.4, "Risk Characterization", of RI-ISI Program Plan states, in part, that each piping segment was evaluated to determine its potential for failure and impact on containment performance (isolation, bypass and large early release). The program plan does not include a large early release frequency (LERF) estimate for your probabilistic risk assessment nor any other LERF information. Describe how the impact of the segment failure on containment performance (isolation, bypass and large early release) was evaluated.

Entergy Response:

The large, early release frequency (LERF) from the Indian Point 3 (IP3) probabilistic risk assessment is 5.86 x 10;7/year. LERF is dominated by seven types of accidents:

1. accidents initiated by steam generator tube ruptures (SGTR),

2. interfacing systems loss of coolant accidents (ISLOCA),

3. internal flooding induced station blackout accidents,

4. transients,

5. station blackout,

6. loss-of-coolant accidents, and

7. vessel rupture events.

The LERF analysis demonstrated that SGTR and ISLOCA accident progressions dominate early radionuclide releases.

The IP3 RI-ISI evaluations were done consistent with the requirements of EPRI TR 112657, Rev B-A including an assessment of the impact of postulated piping failure on LERF. In assessing the change in risk, the proposed RI-ISI program was shown to be risk neutral with respect to both core damage frequency and LERF. That is, in the high and medium risk regions, the number of inspections were increased or remained the same as compared to the existing Section XI program. In addition, the IP3 RI-ISI application only applies to Class 1 piping which is located entirely inside containment.

2. Provide a discussion of the expected change in LERF due to implementing the RI-ISI program.

Entergy Response:

The accident types and their contribution to internal large early release frequencies are identified in the table below. Implementing the RI-ISI program only affects LOCA and based on the low contribution of LOCA to LERF, as shown in the Table, the change in LERF due to implementing the RI-ISI program is small.

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Attachment I to IPN-02-079 Accident Types and Their Contribution to Internal Large Early Release Frequencies Accident Type Point Estimate

% Contribution to Point Estimate Large Large Early Early Release Frequency Release Frequency (lyr)

Steam Generator 4.19 x 10' 71.47 Tube Ruptures Interfacing Systems 1.32 x 10' 22.60 LOCAs Internal Flooding 2.62 x 10.:

4.47 Transients 5.69 x 10' 0.97 Station Blackout 1.41 x 10-9 0.24 LOCAs 1.28 x 10-9 0.22 Vessel Rupture 1.53 x 10' 0.03

3.

In Section 4, "Implementation and Monitoring," the licensee addresses the procedure to implement and monitor the RI-ISI program. Does the licensee plan, as a minimum, to review and adjust the ranking of piping segments on an ASME ISI interval basis? Will information coming from NRC bulletins or generic letters and from industry and plant specific feedback be evaluated for significance and factored, as necessary, into more frequent adjustments of the ranking?

Entergy Response:

IP3 will review the RI-ISI program, as a minimum, at the ASME interval that also applies to the existing ISI program. In addition, more frequent adjustments may be made based on the review and assessment of relevant information (i.e., Generic Letters, Bulletins, INPO notices, and plant specific feedback) in accordance with the existing Operating Experience Program.

4. Section 3.5.1, "Additional Examinations," states, "additional examinations will be performed on those elements up to a number equivalent to the number of elements required to be inspected on the segment or segments initially." Clarify this statement regarding the sample size selection for augmented inspections.

If additional examinations are needed following the identification of an unacceptable flaw, will all high and medium risk welds in the population of welds (with the same degradation mechanisms) be included in the examination population from which the additional sample will be selected? Discuss the time frame for performing these expanded examinations?

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Attachment I to IPN-02-079 Entergy Response:

Additional examinations will be performed on those elements with the same root cause conditions or degradation mechanisms as the identified flaw or relevant condition. The additional examinations will include high risk significant elements and medium risk significant elements, if needed, up to a number equivalent to the number of elements required to be inspected on the segment or segments during the current outage. If unacceptable flaws or relevant conditions are again found similar to the initial problem, the remaining elements identified as susceptible will be examined. An evaluation shall be performed to establish a schedule for conducting this second sample expansion. No additional examinations will be performed if there are no additional elements identified as being susceptible to the same root cause conditions.

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