In Reference 3, the RAI includes two questions for which responses are provided in Attachment 1.

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1 Jon A. Franke Site Vice President Susquehanna Nuclear, LLC 769 Salem Boulevard Berwick, PA Tel Fax TALEN ~ ENERGY SEP U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC CFR SUSQUEHANNA STEAM ELECTRIC STATION RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON TECHNICAL SPECIFICATION CHANGES TO ADOPT TRAVELER TSTF -425 PLA-7381 Docket Nos and References: 1. Letter PLA-7119, [Proposed Amendments to License NPF-14 and NPF-22} Adoption of Technical Specification Task Force Traveler TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control- Risk Informed Technical Task Force (RITSTF) Initiative 5, "dated October 27, 2014 (AccessionML14317A052). 2. Letter PLA-7334, "Response to Request for Additional Information on Technical Specification Changes to Adopt Traveler TSTF-425, "dated July 2, 2015 (Accession ML15183A248). 3. NRC Letter, "Request for Additional Information Regarding License Amendment Request to Adopt Technical Specifications Task Force Traveler (TSTF)-425, (TAC Nos. MF5151 andmf5152), "dated August 24, 2015 (AccessionML15209A974). The purpose of this letter is for Susquehanna Nuclear, LLC to provide the requested additional information (RAI). By Reference 1, and as supplemented by additional information in Reference 2, Susquehanna Nuclear, LLC submitted a license amendment request (LAR) to modify Susquehanna Stearn Electric Station, Units 1 and 2 (SSES) Technical Specifications (TS) by relocating specific surveillance frequencies to a licensee-controlled program. The program will implement Nuclear Energy Institute (NEI) 04-10, "Risk-Informed Technical Specifications Initiative 5B, Risk-Informed Method for Control of Surveillance Frequencies," (Accession ML ). The changes adopt an NRC approved Technical Specification Task Force (TSTF) traveler, TSTF-425, Revision 3, (Accession ML ). In Reference 3, the RAI includes two questions for which responses are provided in Attachment 1. Susquehanna Nuclear, LLC has reviewed the information supporting a finding of no significant hazards consideration and the environmental consideration provided to the NRC in Reference 1. The additional infmmation provided by this submittal does not

2 - 2- Document Control Desk PLA-7381 affect the bases for concluding that the proposed license amendment does not involve a significant hazards consideration. Furthe1more, the additional information also does not affect the bases for concluding that neither an environmental impact statement nor an environmental assessment needs to be prepared in connection with the proposed amendment. There are no new regulatory commitments associated with this response. If you have any questions or require additional information, please contact Mr. Jeffery N. Grisewood (570) I declare under penalty of pe1jury that the foregoing is tme and conect. Executed on: 7 ranke Attachment 1: Response to Requested Additional Information Copy: NRC Region I Mr. J. E. Greives, NRC Sr. Resident Inspector Mr. J. A. Whited, NRC Project Manager Mr. M. Shields, PA DEP/BRP

3 Response to Requested Additional Information

4 Page 1 of 4 Response to Requested Additional Information By letter dated October 27, 2014,( 1 ) and as supplemented by additional infmmation in a letter dated July 2, 201s,CZ> Susquehanna Nuclear, LLC submitted a license amendment request (LAR) for the Susquehanna Steam Electric Station (SSES), Units 1 and 2. The proposed amendment would modify the SSES Technical Specifications by relocating specific frequencies to a licensee-controlled program with the implementation of Nuclear Energy Institute (NEI) 04-10, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies. "< 3 > The NRC requested additional information (RAI) in a letter dated August 24, 2015.( 4 ) This Attachment provides the requested additional information. RAI8: In the response to request for additional information (RAI) 3, regarding Supporting Requirement (SR) HR-B2, the licensee clarified that the intent of the finding and observation (F&O) resolution was to indicate that there is no longer any reliance on staggered testing/maintenance principles for screening purposes, and that the pre-initiator process was used for screening. The licensee's response also seems to suggest that all common mode eltors were screened out. Discuss the justification and conclusions on screening pre-initiator common mode eltors. Also, discuss the considerations and the bases for the inclusion or exclusion of modeling of common-mode eltors in the probabilistic risk assessment (PRA). SSES Response to RAI 8: Both common mode miscalibration events and common mode misalignment events were considered for the SSES pre-initiator Human Reliability Analysis (HRA). (1) Letter (PLA-7119), [Proposed Amendments to License NPF-14 and NPF-22] Adoption of Technical Specification Task Force Traveler TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control- Risk Informed Technical Specification Task Force (RITSTF) Initiative 5, "dated October 27,2014 (Accession ML14317A052). (2) Letter (PLA-7334), "Response to Request for Additional Information on Technical Specification Changes to Adopt Traveler TSTF-425, "dated July 2, 2015 (Accession ML15183A248). (3) Nuclear Energy Institute (NEI) 04-10, Revision 1, "Risk-Informed Technical Specifications Initiative 5B, Risk-Informed Method for Control of Surveillance Frequencies," dated April30, 2007 (Accession ML ). ( 4) NRC Letter, "Request for Additional Information Regarding License Amendment Request to Adopt Technical Specifications TaskForce Traveler (TSTF)-425, (TAC Nos. MF5151 and MF5152)," dated August 24, 2015 (Accession ML15209A974).

5 Page 2 of4 For common mode miscalibrations, the pre-initiator identification process identified those calibration activities during which a common error could occur that would prevent the automatic actuation of a function required in the PRA. For example, common mode miscalibration events are included in the PRA for the critical Reactor Pressure Vessel (RPV) pressure switch channels that would fail the low pressure permissive logic required for Low Pressure Coolant Injection and Core Spray injection (i.e., A and B, A and D, Band C, etc.). Common mode calibration activities were screened from the analysis if: The activities could lead to calibration errors in non-pra systems. The activities could lead to calibration errors for equipment within a component's boundary (i.e., miscalibrations of sensors/instruments that are within the component boundary are inherent in the component failure data and do not require additional failure events). The combination of miscalibration errors does not result in critical failure of the logic. For example, for an actuation signal that requires operation of channel "A" OR "B" AND channel "C" OR "D" for success, it is not necessary to include events of common cause miscalibration of channels "A" and "D" because the signal is not failed by that combination. For common mode misalignment errors, the pre-initiator identification process identified single activities that could result in the misalignment of a component that would prevent operation of multiple redundant trains of a system or diverse systems. For example, a misalignment was included for a spray pond retum valve, which if left closed, would result in the failure of the "A" train of Emergency Service Water (for both Units 1 & 2) and the "A" train of Residual Heat Removal Service Water (also for both Units 1 & 2). Activities leading to misalignment/restoration errors were screened if: The activities were related to non-pra systems. The activities were related to n01mally operating systems and operation would reveal a misalignment. Components in multi-train systems in which one train may be assumed to be normally running and the other( s) assumed to be in standby were not screened. All trains were assumed to be in standby for the pre-initiator Human Error Probability (REP) identification process. Position indication is available in the main control room, automatic re-alignment occurs on system initiation, a status check is perf01med on a shiftly basis, or a signal exists that would identify the misalignment. This screening rule was not applied to misalignments/restoration errors that could simultaneously impact multiple redundant trains or diverse systems. Common mode events impacting redundant system trains or diverse systems were specifically treated in accordance with HR-A3. This SR requires the identification of work practices that "involve a mechanism that simultaneously affects equipment in either

6 Page 3 of4 different trains of a redundant system or diverse systems (e.g., use of common calibration equipment by the same crew on the same shift, a maintenance or test activity that requires realignment of an entire system (e.g., SLCS)." The intent of the SR was viewed to focus on capturing single activities that impact redundant trains of a system or diverse systems, not multiple, separate activities that impact redundant or diverse systems, even if they are performed in an outage. Any components that could be manipulated in a way that would disable different trains of a redundant system or diverse systems should be identified through the system review and developed as events. These activities may not be screened. Quantification of the common mode pre-initiator events was performed using the preinitiator Accident Sequence Evaluation Program (ASEP) methodology, which is identified in Section 3.2 of NUREG-1842, "Evaluation of Human Reliability Analysis Methods Against Good Practices," as a valid approach for evaluating risk-significant preinitiator events. RAI9: The response to RAI 4 does not discuss reflecting the current plant configuration and operating experience when considering extemal events using NEI guidance. (The response does confirm this for the intemal events analysis.) Please explain whether evaluation of the fire risk and other extemal events supporting this application reflects, or considers, the current plant configuration and operating experience. SSES Response to RAI 9: By following Steps 1 Oa and 1 Ob of the NEI guidance, the evaluation of fire risk and other extemal events risk supporting this application will reflect and consider the current plant configuration and operating experience. The Individual Plant Examination for Extemal Events (IPEEE) is not a living document and has not been updated to the present plant configuration and operating experience. As a result, the fire risk and other extemal event risk information from the IPEEE is limited to qualitative insights. For the surveillance test interval (STI) change evaluations, the intent is not to directly use any numerical results from the IPEEE fire studies or other extemal events, but to qualitatively assess any available information to determine the impact on the proposed surveillance interval changes, consistent with Step 1 Oa of the NEI methodology. This qualitative assessment of fire risk and other extemal event risk will include a review of applicability to the current plant configuration and operating experience. Additionally, for some STI change evaluations, per Step lob of the NEI methodology, qualitative reasoning and very low changes to core damage frequency (~CDF) and large

7 Page 4 of4 early release frequency (L1LERF) results from the internal events analysis may be sufficient to support the STI change evaluation where Step 1 Ob reads in part: "Alternative evaluations for the impact from external events and shutdown events are also deemed acceptable at this point. For example, if the!:l.edf and MERF values have been demonstrated to be very small from an internal events perspective based on detailed analysis of the impact of the SSe being evaluated for the STI change, and if it is known that the edf or LERF impact from external events (or shutdown events as applicable) is not specifically sensitive to the sse being evaluated (by qualitative reasoning), then the detailed internal events evaluations and associated required sensitivity cases (as described in Step 14) can be used to bound the potential impact from external events and shutdown PRA model contributors. " Qualitative evaluation of fire and external events risk in support of Step 1 Ob would also include consideration of applicability to the cunent plant configuration and operating expenence. Therefore, by following Steps 1 Oa and 1 Ob of the NEI guidance, the evaluation of fire risk and other external events will reflect and consider the cunent plant configuration and operating experience.

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