PLA-7281, Response to Request for Additional Information on Relief Request 4RR-01

Response to Request for Additional Information on Relief Request 4RR-01

ML15036A505
Person / Time
Site:	Susquehanna
Issue date:	02/05/2015
From:	Franke J Susquehanna
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
PLA-7281
Download: ML15036A505 (16)

Text

Jon A. Franke PPL Susquehanna, LLC Site Vice President 769 Salem Boulevard Berwick, PA 18603 Tel. 570.542.2904 Fax 570.542.1504 jfranke@ pplweb.com FEB 0 5 2015 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 SUSQUEHANNA STEAM ELECTRIC STATION RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON RELIEF REQUEST 4RR-01 Docket No 50-387 PLA-7281 and No. 50-388

References:

I. PPL Letter (PLA-7I93), "Request to Continue Use of a Risk-Informed Inservice Inspection Alternative in a Proposed Relief Request No. 4RR-OI to the Fourth 10-Year lnservice Inspection Program for Susquehanna Units I and 2, " dated October 29, 2014 (Accession ML14302A443).

2. PPL Letter PLA-7I78, "Inservice Inspection (lSI) Program Plan for the Fourth Ten-Year Interval, " dated June 2, 2014.

3. NRC Letter, "Request for Additional Information re: Relief Request 4RR-OI (TAC Nos. MF5097 and MF5098)," dated January 8, 20I5 (Accession ML14343A984).

The purpose of this letter is for PPL Susquehanna, LLC (PPL) to provide the requested additional information (RAI). By Reference 1, PPL submitted a Relief Request 4RR-01 for the Susquehanna Steam Electric Station (SSES), Units 1 and 2, Fourth 10-Year Inservice Inspection Program. Specifically, PPL requested a change to continue use of the Risk-Informed Inservice Inspection (RI-ISI) program as an alternative to the American Society of Mechanical Engineers (ASME)Section XI, lSI Program for Class 1 and 2 piping welds.

The Fourth 10-Year lSI Interval inspection program uses the ASME Section XI, 2007 Edition through the 2008 Addenda for the examination of these components. In Reference 3, the NRC requested additional information. The response to the RAI questions is in the attachment to this letter.

There are no new regulatory commitments associated with this response.

Document Control Desk PLA-7281 If you have any questions or require additional information, please contact Mr. Jeffery N. Grisewood, Manager- Nuclear Regulatory Affairs at (570) 542-1330.

Sincerely, 9-~

Attachment:

Response to Request for Additional Information Copy: NRC Region I Mr. J. Greives, NRC Sr. Resident Inspector Mr. J. Whited, NRC Project Manager Mr. L. Winker, PA DEP/BRP

Attachment to PLA-7281 Response to Request for Additional Information

Attachment to PLA-7281 Page 1 of 13 Response to Request for Additional Information By letter dated December 19, 2012 (I) PPL Susquehanna, LLC (PPL), submitted Relief Request 4RR-01 for review and approval for the Susquehanna Steam Electric Station (SSES), Units 1 and 2. Specifically, PPL requested to continue to use the Risk-Informed Inservice Inspection (RI-ISI) program as an alternative to the American Society of Mechanical Engineers (ASME)Section XI, lSI Program for Class 1 and 2 piping welds.

To complete its review, the U.S. Nuclear Regulatory Commission (NRC) in the Division of Risk Assessment's Probabilistic Risk Assessment (PRA) Licensing Branch (APLA) and the Division of Engineering's Component Performance, Non-Destructive Examination (NDE) and Testing Branch (EPNB) requests a response to the questions below.

APLA RAI-01:

In the Safety Evaluation Report (SER) for Electric Power Research Institute, Inc. (EPRI)

Topical Report TR-112657, "Revised Risk-Informed lnservice Inspection Evaluation Procedure," Revision B-A, the NRC staff stated that "[ t]he scope, level of detail, and quality of a PRA and the general methodology for using PRA in regulatory applications is discussed in [Regulatory Guide] RG 1.174 [, "An Approach for Using Probabilistic Risk Assessment In Risk Informed Decisions On Plant Specific Changes to the Licensing Basis"(2J]. RG 1.178 [, "An Approach for Plant Specific Risk Informed Decisionmaking lnservice Inspection of Piping, ,(J)] provides guidance that is more specific to lSI." An acceptable change-in-risk evaluation requires the use of a PRA of appropriate technical quality that models the as-built and as-operated plant, as discussed in RGs 1.178 and 1.200, Revision 2, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities. ,(4 )

Section 5.3 of EPRI TR-112657 describes that the consequence evaluation portion of the EPRI Methodology utilizes PRA inputs. EPRI TR-112657 further identifies key attributes and/or areas of the PRA where quality is considered relatively important to support a consistent RI-lSI application.

(1) PPL Letter (PLA-7193), "Request to Continue use of a Risk-Informed Inservice Inspection Alternative in a Proposed Relief Request No. 4RR-OI to the Fourth IO-Year Inservice Inspection program for Susquehanna Units I and 2," dated October 29, 2014 (ML14302A443)

(2) NRC RG 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," dated November 29, 2002, (ML023240437)

(3) NRC RG 1.178, "An Approach for Plant-Specific Risk-Informed Decisionmakingfor Inservice Inspection of Piping," dated September 30, 2003, (ML032510128)

(4) NRC RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, " dated March 1, 2009, (ML090410014)

Attachment to PLA-7281 Page 2 of 13 In Table 1 of Attachment 2 to 4RR-01, the licensee provided 24 supporting requirements (SRs) associated with internal flooding not meeting Capability Category [CC] II of the ASME/American Nuclear Society (ANS) PRA standard (ASME/ANS RA-Sa-2009) from a full-scope peer review completed in 2012. Most of these SRs do not meet CCI requirements of the standard. In assessing the impact of these SRs, the licensee stated in Section 1.3 of Attachment 2 to 4RR-01, that "Based on this industry comparison and small contribution of internal flooding to overall [Core Damage Frequency] CDF and [Large Early Release Frequency]

LERF, the Susquehanna internal flooding PRA can be applied to support the fourth 10-year inspection interval based on Code Case N-578-1."

The contribution of flooding to total CDF is irrelevant to judging the acceptability of consequence and change-in-risk evaluations and, therefore, the acceptability of a RI-ISI program. As the full-scope peer-review of the Susquehanna Steam Electric Station (SSES) PRA has identified many Facts and Observation (F&Os) related to the internal flooding, the internal flooding PRA does not seem to be suitable for directly supporting the RI-ISI.

a) Provide an explanation of why the results of the internal flooding PRA, if used in consequence and change-in-risk evaluations to support the relief request, are acceptable given that the large number of SRs related to key areas in flooding identified in EPRI TR-112657, Revision B-A do not meet requirements of the ASME/ANS RA-Sa-2009 and RG 1.200, Revision 2.

b) Describe whether any RI-ISI specific evaluation was performed to overcome the known weaknesses in the flooding evaluation.

PPL's Response:

The requested additional information on use of internal flooding PRA and a specific evaluation are as follows.

a) The internal flooding PRA is not directly used for the consequence and the change-in-risk evaluations to support the relief request. The third interval relief request consequence and change-in-risk evaluations also did not directly use an internal flooding PRA. As identified in the NRC SER for the third interval relief request (Reference 11 ), impacts due to both direct and indirect effects were considered using the guidance provided in EPRI TR-112657, Revision B-A (Reference 9). The EPRI methodology requires consideration of impacts of pipe breaks which were confirmed through plant walkdowns. The impacts considered for the third interval were also considered for the fourth interval relief request which is also based on ASME Code Case N-578-1. In addition, EPRI developed guidance

Attachment to PLA -7281 Page 3 of 13 (e.g., EPRI Report 1021467-A) on PRA Quality requirements for RI-ISI, (Reference 10). As stated in EPRI 1021467-A, RI-ISI applications using the EPRI traditional RI-ISI approach (i.e., ASME Code Case N-578-1) do not use the internal flooding PRA directly. As such, internal flooding supporting requirements (SRs) are not applicable, and Section 3.3 of EPRI Report TR-112657 is the appropriate resource. This is in contrast to the EPRI streamlined RI-ISI approach (ASME Code Case N-716) that uses the internal flooding analysis directly. The NRC acknowledges this approach in the SER contained in EPRI 1021467-A.

Based on the above, the internal flooding PRA analysis was not used for this update and the internal flooding supporting requirements (SRs) do not apply to this application.

b) As discussed in the response to RAI-01 part a), the internal flooding PRA model is not used directly in support of this application and the internal flooding SRs are not applicable. Therefore, a specific evaluation was not performed to overcome weaknesses in the flooding PRA model. Impacts from flooding are considered based on the methodology in EPRI TR-112657, Revision B-A.

APLA RAI-02:

The SER states that EPRI TR-112657, Rev. B does not include a detailed discussion of the specific assumptions to be used to guide the assessment of the direct and indirect effects of segment failures. The SER further states that specific assumptions regarding the direct and indirect effects of pipe segment failure should be developed by the individual licensees and should form part of the onsite documentation.

In Table 1 of Attachment 2 to 4RR-01, the description ofF&O 6-14 states that a "qualitative discussion of additional impacts (jet impingement, pipe whip, humidity) is required for CC !III per RG 1.200 clarification to meet SR IFSN-A6. An evaluation of medium/small bore piping for pipe whip and jet impingement is required to meet SR IFQU-A9."

a) Clarify if specific guidelines and assumptions used for determining direct and indirect effects of flooding have been developed for this application.

b) Clarify if the loss of mitigating ability (discussed in Section 3.2.4 of the SER),

where segment failures that only cause failure of mitigating functions but do not cause a plant trip, has been considered in consequence evaluations for this application in accordance with EPRI TR-112657.

Attachment to PLA-7281 Page 4 of 13 PPL's Response:

The requested additional information on the use of direct and indirect effects of flooding, and consideration of the loss of mitigating ability in consequence evaluations are as follows.

a) As described in response to APLA RAI-01, the internal flooding PRA supporting requirements are not considered to be applicable; and EPRI TR-112657, Revision B-A guidance is used. Specifically, the EPRI guidance explicitly requires and the proposed risk-informed lSI program in 4RR-01 considers both direct and indirect effects of flooding in the consequence evaluation.

b) Also, the EPRI guidance explicitly requires mitigation ability be addressed in the case where a plant trip is not caused. These guidelines were followed during the third interval application and are also considered in this update.

APLA RAI-03:

In Section 4.0 of the SER for EPRI TR-112657, Revision B-A, the NRC staff concluded that a licensee requesting to implement an RI-lSI program pursuant to section 50.55a(a)(3) may incorporate into its application, by reference, the program described in EPRI TR-112657, Rev. B, together with appropriate plant-specific information, provided that the application includes, among other items, a statement that RG principles have been met (or any exceptions) and a summary of any augmented inspections that would be affected.

a) The fifth principle in RG 1.174 states that the impact of the proposed change should be monitored using performance measurement strategies. Clarify whether the implementation and monitoring program of the third 10-year interval lSI program will continue during the proposed fourth 10-year interval. Discuss any changes in the implementation and monitoring program from the third 10-year interval.

b) Identify any changes to the augmented inspection programs from the approved third 10-year interval RI-ISI program and discuss the reason(s) for any changes.

Attachment to PLA-7281 Page 5 of 13 PPL's Response:

The requested additional information on the implementation and monitoring program and augmented inspection programs are as follows.

a) The RI-ISI program is a living program requiring feedback of new relevant information to ensure the appropriate identification of high safety significant piping locations. The implementation and monitoring program of the third 10-year interval lSI program will continue during the proposed fourth 10-year interval with no changes.

b) The augmented inspection program is described in Final Safety Analysis Report (FSAR) Chapters 5.2, "Integrity of Reactor Coolant Pressure Boundary" and Chapter 6.6, "[IS/] of Class 2 and 3 Components," in Sections 5.2.4.7 and 6.6.8, entitled "Augmented Inservice Inspection to Protect Against Postulated Piping Failures." Augmented Examinations Programs are also described in Section 5.1 of the fourth 10-year interval lSI Program Plan, (Reference 8). The program descriptions describe that SSES previously applied the methodology contained in EPRI TR-1006937-A to its Break Exclusion Region (BER) program during the third interval. PPL updated its program for the fourth interval reflecting more recent plant operating practices and experiences as well as reflecting the latest PRA inputs.

The update to the BER program continues to show the existing acceptance criteria for CDF and LERF for system risk and for the total plant level risk impacts continue to be met. A summary of considerations included in the update are provided in the balance of the response to this question. There are no other changes to any augmented inspection programs from the approved third 10-year interval RI-ISI program.

Previous risk-informed BER evaluations remain valid, considering plant service history and failure potential. A summary of considerations from plant operating experience in assessment to the BER scope of piping follows.

• Two welds in feedwater (FW) were re-categorized as susceptible to thermal striping, cycling, and stratification (TASCS) as opposed to thermal transient (TT)

• No change is made in high pressure coolant injection (HPCI) to the degradation assignment for the one BER weld in scope.

• Operating parameters did change in main steam (MS) due to the power uprate, but these changes did not impact the assignment of any degradation mechanism and TT is still applicable to some welds in the MS drain lines.

Attachment to PLA-7281 Page 6 of 13

• More precise evaluation of susceptibility in reactor core isolation cooling (RCIC) to thermal fatigue identified a need to change some welds from a susceptibly ofTASCS to TT, but all of these conditions were related to RCIC pump discharge (that is, flow to the reactor) and the RCIC BER welds experience flow from the reactor (that is, the opposite direction and different lines) and therefore no change to the degradation assignments are needed.

• No change in reactor water cleanup (RWCU) to the degradation mechanism assignment for BER is needed. That is, locations remain identified with a non-degradation mechanism assignment.

In assessing the service history at SSES, the following observations can also be summarized:

• The BER scope consists of the same systems except for some minor continuations to non-ASME piping.

• There has been no inspection history since the previous risk informed BER evaluation to indicate operative degradation.

• Piping failures have not occurred in the BER scope (or outside the BER scope if the findings could be considered applicable to the BER scope) since the previous risk informed BER evaluation.

• Finally, there have been no piping additions/deletions in the BER scope of piping (for example, no piping/welds added, deleted, or modified).

APLA RAI-04:

RG 1.193,<5 ) "ASME Code Cases Not Approved for Use," Revision 4, has listed Code Case N-578-1, "Risk-Informed Requirements for Class 1, 2, or 3 Piping, Method B" as an unacceptable Section XI Code Case. Identify any portions of Code Case N-578-1, which have been used in the development of the proposed RI-ISI program that are not specifically incorporated into, or referenced by, EPRI TR-112657, Revision B-A.

PPL's Response:

Despite the modified wording between the approved third interval relief request 3RR-01 and the proposed fourth interval relief request 4RR-01, the portions of Code Case N-578-1 that have been used in the development of the proposed RI-ISI program for the fourth interval, and which were not specifically incorporated into or referenced by EPRI TR-112657, Revision B-A, are the same with the following exception. There would be (5) NRC RG 1.193, "ASME Code Cases Not Approved For Use, " dated August 30, 2014 (ML13350A001)

Attachment to PLA-7281 Page 7 of 13 differences from use of paragraphs and figures from the 2007 Edition through the 2008 Addenda of ASME Section XI, (e.g., the Code of record for the fourth interval) and the Code of record for the third interval. These instances were defined in the third 10-year lSI interval program plan, and will continue to be true in the fourth 10-year lSI interval.

These instances that are taken from the third interval program descriptions are:

A summary of the ASME Section XI component, support, system pressure testing, and augmented examinations and tests for the third interval are provided in two tables in Reference 6, (e.g., Table 7.0-1, beginning on page 7-4; and Table 7.0-2, beginning of page 7 -17). Note 3 for both the Table 7.0-1, for Unit 1 and common, and Table 7.0-2 for Unit 2, describe that for the third inspection interval, the Class 1 and 2 piping inspection program will be governed by risk-informed regulations.

The Note 3 states further that "The [RI-ISI] Program methodology is described in the EPRI Topical Report TR-112657, Rev. B-A and Code Case N-578-1." Note 5 of both tables state that "Examination requirements within the [RI-ISI] Program are determined by the various degradation mechanisms present at each individual piping structural element. See EPRI TR-112657, Rev. B-A and Code Case N-578-1 for specific exam method requirements. "

Relief Request 3RR-01, "Proposed Alternate Provisions, " states that the third interval RI-lSI program is an EPRI TR-112657, Revision B-A application and will be maintained as a living program.

The following two enhancements implemented in the third interval would also remain applicable in the fourth interval:

a) In lieu of the evaluation and sample expansion requirements in Section 3.6.6.2, "RI-ISI Selected Examinations" ofEPRI TR-112657, SSES will utilize the requirements of Subarticle-2430, "Additional Examinations" contained in Code Case N-578-1. The alternative criteria for additional examinations contained in Code Case N-578-1 provides a more refined methodology for implementing necessary additional examinations.

b) To supplement the requirements listed in the EPRI TR-112657, Table 4-1, "Summary of Degradation-Specific Inspection Requirements and Examination Methods," SSES will utilize the provisions listed in the Code Case N-578-1 Table 1, "Examination Category R-A, Risk-Informed Piping Examinations."

To implement Note 10 of this table, paragraphs and figures from the 1998 Edition through the 2000 Addenda of ASME Section XI [SSES's Code of record for the Third Interval] will be utilized which parallel those referenced in the Code Case for the 1989 Edition. Table 1 of Code Case N-578-1 will be used as it provides risk informed Category I Item Numbers, and a detailed

Attachment to PLA -7281 Page 8 of 13 breakdown for examination method, and also a categorization of parts to be examined where the topical report is either silent or ambiguous.

As previously stated, there would be differences from use of paragraphs and figures from the 2007 Edition through the 2008 Addenda of ASME Section XI, (e.g., Code of record for the fourth interval) and the Code of record for the third interval. However, such instances are defined in the above so that applicable requirements of the previously approved RI-ISI program at SSES from the third interval continue to be true in the fourth 10-year interval for the RI-lSI program.

EPNB RAI-01:

In Section 4.0 of the SER for EPRI TR-112657, the NRC staff concluded that a licensee requesting to implement an RI-ISI program pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Section 50.55a(a)(3) may incorporate into its application, by reference, the program described in EPRI TR-112657, Rev. B, together with appropriate plant-specific information, provided that the application includes, among other items, a summary of the risk impact. The licensee stated in Section 5 of the Attachment 1 to 4RR-01, that "[a]s part ofthe RI-ISI living program update, the delta risk assessment was re-evaluated and was determined to continue to meet the delta risk acceptance criteria of EPRI TR-112657."

a) Provide values for the changes in core damage frequency and changes in large early release frequency for SSES compared to ASME Code,Section XI lSI program values, and to the fourth lSI interval RI-lSI program values.

b) Provide a similar table as shown in Attachment A, Tables 7 and 8, of letter dated September 16, 2003,<6) for your proposed fourth interval program.

PPL's Response:

The risk impact analysis results are in Tables 1 and 2. The risk calculations do not take credit for the RI-ISI improved probability of detection (POD) for thermal fatigue. Both the Section XI program and RI-lSI contributions are shown similar to the original submittal. The acceptance criteria are met for ~CDF and ~LERF per system (1E-07 and 1E-08, respectively) as shown in Tables 1 and 2. The acceptance criteria for each unit are met for Total ~CDF and ~LERF (1E-06 and 1E-07, respectively) as shown in Tables 1 and2.

(6) PPL letter PLA-5662, "Proposed Third Ten-Year Inservice Inspection Interval Inservice Inspection Program Plan for Susquehanna SES Units I and 2," dated September 16, 2003, (ML032670839)

Attachment to PLA-7281 Page 9 of 13 Table 1 - Risk Impact of RI-ISI on CDF and LERF due to Pipe Ruptures for Unit 1 Systems CDF LERF System Description Section XI RI-ISI Delta CDF Section XI RI-ISI Delta LERF CAC Containment Air Control O.OE+OO (1) (1) O.OE+OO (1) (1)

CRD Control Rod Drive 1.5E-14 O.OE+OO 1.5E-14 1.5E-15 O.OE+OO 1.5E-15 cs Core Spray 8.1E-10 2.0E-12 8.1E-10 S.OE-10 2.0E-14 S.OE-10 FW Feedwater 7.5E-10 6.5E-10 l.OE-10 5.1E-12 4.3E-12 8.3E-13 HPCI High Pressure Coolant Injection 6.0E-13 7.0E-13 -l.OE-13 1.1E-14 5.5E-15 5.3E-15 MS Main Steam 1.6E-11 2.5E-12 1.4E-11 2.8E-13 l.OE-14 2.7E-13 RBCW Reactor Building Cooling Water O.OE+OO O.OE+OO O.OE+OO O.OE+OO O.OE+OO O.OE+OO RCIC Reactor Core Isolation Cooling 7.8E-14 O.OE+OO 7.8E-14 4.5E-15 O.OE+OO 4.5E-15 RHR Residual Heat Removal 5.7E-10 1.4E-10 4.3E-10 8.8E-11 4.3E-12 8.4E-11 RPV-E RPV Nozzles 1.3E-09 2.5E-10 l.OE-09 6.2E-12 1.2E-12 5.0E-12 RR Reactor Recirculation 1.8E-09 2.7E-10 1.5E-09 7.1E-12 1.1E-12 6.1E-12 RWCU Reactor Water Cleanup 2.7E-10 1.5E-11 2.6E-10 1.1E-12 6.0E-14 l.OE-12 SBLC Standby Liquid Control O.OE+OO O.OE+OO O.OE+OO O.OE+OO O.OE+OO O.OE+OO Total 5.5E-09 1.3E-09 4.2E-09 9.1E-10 1.1E-11 9.0E-10 Note: (1) New scope with one RI-ISI selection resulting in minor risk reduction

Attachment to PLA-7281 Page 10 of 13 Table 2- Risk Impact of RI-ISI on CDF and LERF due to Pipe Ruptures for SSES Unit 2 Systems CDF LERF System Description Section XI RI-ISI Delta CDF Section XI RI-ISI Delta LERF CAC Containment Air Control O.OE+OO (1) (1) O.OE+OO (1) (1)

CRD Control Rod Drive 1.5E-14 O.OE+OO 1.5E-14 1.5E-15 O.OE+OO 1.5E-15 cs Core Spray 8.1E-10 2.0E-12 8.1E-10 8.0E-10 2.0E-14 8.0E-10 FW Feedwater 5.0E-10 7.0E-10 -2.0E-10 3.6E-12 4.6E-12 -9.9E-13 HPCI High Pressure Coolant Injection 1.1E-12 1.2E-12 -1.1E-13 1.2E-14 7.0E-15 5.2E-15 MS Main Steam 1.2E-11 2.5E-12 9.7E-12 1.7E-13 l.OE-14 1.6E-13 RBCW Reactor Building Cooling Water O.OE+OO O.OE+OO O.OE+OO O.OE+OO O.OE+OO O.OE+OO RCIC Reactor Core Isolation Cooling 5.6E-14 2.0E-13 -1.4E-13 3.3E-15 l.OE-13 -9.7E-14 RHR Residual Heat Removal 4.5E-10 8.5E-11 3.7E-10 8.8E-11 3.8E-12 8.4E-11 RPV-E RPV Nozzles 1.3E-09 2.5E-10 l.OE-09 6.2E-12 1.2E-12 S.OE-12 RR Reactor Recirculation 2.0E-09 2.6E-10 1.7E-09 7.9E-12 1.1E-12 6.8E-12 RWCU Reactor Water Cleanup 1.7E-10 1.3E-11 1.6E-10 6.9E-13 S.OE-14 6.4E-13 SBLC Standby Liquid Control O.OE+OO O.OE+OO O.OE+OO O.OE+OO O.OE+OO O.OE+OO Total 5.2E-09 1.3E-09 3.9E-09 9.1E-10 1.1E-11 9.0E-10 Note: (1) New scope with one RI-ISI selection resulting in minor risk reduction

Attachment to PLA-7281 Page 11 of 13 EPNB RAI-02:

In Section 7 of 4RR-01, the licensee stated, in part, that:

Susquehanna considers both the plant and industry operating experience and updates the RI-lSI program during the re-evaluation process following each inspection period ...

a) Has the operating experience from licensee event report (LER) 50-387/2012-007-01, dated November 20, 2012,(7) been incorporated to the RI-ISI program?

Corrective action number 6 in LER 50-387/2012-007-01 states, in part, that:

The welds are to be inspected each refueling outage (once per cycle) to confirm no surface indications and shall continue until the piping has been modified (or eliminated) to minimize vibrational response.

b) Please summarize the actions taken because of this operational experience.

c) Have the inspections taken place?

d) If so, what are the results?

e) Has the piping been modified (or eliminated) to minimize vibrational response?

f) Has an inspection been performed on the modified piping?

g) If applicable, please explain the plans for further inspections on the piping.

PPL's Response:

Per Section 2.5.2 of EPRI TR-1126557, Revision B-A, one principal finding of the review of the treatment of degradation mechanisms was concurrence that vibrational fatigue should be treated outside the RI-lSI program. The chance of catching a vibration-induced crack in the extremely short time between crack initiation and going through-wall is very small, and therefore this mechanism is not suitable for management under RI-ISI. For this reason, the subject LER does not impact the RI-ISI program.

In response to this operating experience, Susquehanna has modified the 'A' and 'B' loop suction decontamination lines on Unit 1 and Unit 2 that experienced the failure detailed in LER 50-387-2012-007-01 such that the natural frequency of the line is above the range of frequencies associated with the vane passing frequency of the reactor recirculation pumps.

(7) PPL letter PLA-6928, "Susquehanna Steam Electric Station (SSES) Licensee Event Report 50-38712012-007-01, License No. NPF-14," dated November 20, 2012, (ML123250703)

Attachment to PLA-7281 Page 12 of 13 Additionally, as referenced in corrective action number 6 in LER 2012-007-01, six different welds susceptible to high vibration were added to an augmented inspection program for a surface examination each outage. These six welds have each been inspected once since issuance of LER 2012-007-01 with no recordable indications found.

These six locations have not been modified to minimize vibration response. The welds will continue to be inspected each outage or until modified to reduce the effects of vibration.

EPNB RAI-03:

Of the welds not selected for future examinations, have previous examinations of any of these welds identified service induced degradation? If so, what was the degradation mechanism and what was done to mitigate the degradation?

PPL's Response:

There have been no indications of service induced degradation found during the third 10-year inspection interval for any welds within the scope of the RI-ISI program except for the vibration induced failure as discussed in EPNB RAI-02.

EPNB RAI-04:

4RR-01 states the RI-ISI program is a living program monitored periodically for changes, where this monitoring includes numerous facets. Please confirm that vendor issued communications such as General Electric (GE)-Hitachi Safety Communications are included as part of the reviews done for the living aspects of the program.

PPL's Response:

As stated by the SSES lSI Program Plan for the fourth 10-year interval, in Section 10.1.5, entitled "Reevaluation of Risk-Informed Selections," (S)

"The affected portions of the risk-informed inservice inspection program shall be reevaluated as new information [becomes available] affecting implementation of

[the program]. Examples include piping system design changes, industry-wide failure notification, and prior examination results. "

When such valuable information from vendor advisories is received, the Operating Experience (OE) program requires the station to determine their applicability.

(8) PPL Letter PLA-7178, "Inservice Inspection Program Plan for the Fourth Ten-Year Interval, "

dated June 2, 2014

Attachment to PLA-7281 Page 13 of 13 EPNB RAI-05:

The NRC staff notes that in the components affected section on page 1 of 18, of 4RR-01, the B-F welds Code Item number is B5.140. This code item number is no longer in the ASME Code for the applicable inspection interval. Please explain.

PPL's Response:

The use of the B-F weld Code Item number B5.140 of the request 4RR-01 was an oversight. The correct Code Item number is B5.20.

REFERENCES:

1. PPL Letter (PLA-7193), "Request to Continue use of a Risk-/nfonned Inservice Inspection Alternative in a Proposed Relief Request No. 4RR-01 to the Fourth 10-Year Inservice Inspection program for Susquehanna Units 1 and 2," dated October 29, 2014 (ML14302A443)

2. NRC RG 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-

/nfonned Decisions on Plant-Specific Changes to the Licensing Basis," dated November 29, 2002, (ML023240437)

3. NRC RG 1.178, "An Approach for Plant-Specific Risk-Infonned Decisionmaking for Inservice Inspection of Piping," dated September 30, 2003, (ML032510128)

4. NRC RG 1.200, "AnApproachfor Detennining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-/nfonned Activities, " dated March 1, 2009, (ML090410014)

5. NRC RG 1.193, "ASME Code Cases Not Approved For Use," dated August 30, 2014 (ML13350A001)

6. PPL letter PLA-5662, "Proposed Third Ten-Year Inservice Inspection Interval Inservice Inspection Program Plan for Susquehanna SES Units 1 and 2," dated September 16, 2003, (ML032670839)

7. PPL letter PLA-6928, "Susquehanna Steam Electric Station (SSES) Licensee Event Report 50-387/2012-007-01, License No. NPF-14," dated November 20, 2012, (ML123250703)

8. PPL Letter PLA-7178, "Inservice Inspection Program Planfor the Fourth Ten-Year Interval," dated June 2, 2014

9. EPRI Report TR-112657 Revision B-A, "Revised Risk /nfonned Inservice Inspection Procedure," dated February 10, 2000 (ML013470102)

10. EPRI Report 1021467-A, "Nondestructive Evaluation: Probabilistic Risk Assessment Technical Adequacy Guidance for Risk-Infonned In-Service Inspection Programs," June 2012 (ML12171A450)

11. NRC Safety Evaluation Report [related to Relief Request No. 3RR-01], dated July 28, 2005, "Third 10-Year Inservice Inspection (IS/) Interval Program Plan (TAC Nos. MC1181 and 1182)." (ML051990330)