ML16034A139

R.E. Ginna - Response to Request for Additional Information - Application for Technical Specifications Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled

ML16034A139
Person / Time
Site:	Ginna
Issue date:	02/03/2016
From:	David Gudger Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
CAC MF6358
Download: ML16034A139 (60)

Text

200 Exelon Way Exelon Generation Kennett Square. PA 19348 www exeloncorp.com 10 CFR 50.90 February 3, 2016 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 SUBJECT: A. E. Ginna Nuclear Power Plant Renewed Facility Operating License No. DPR-18 NRC Docket No. 50-244 Response to Request for Additional Information -Application for Technical Specifications Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3) REFERENCES: 1. Letter from James Barstow (Exelon) to U.S. Nuclear Regulatory Commission, "Application for Technical Specifications Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3)," dated June 4, 2015. 2. Letter from the U.S. Nuclear Regulatory Commission to Bryan C. Hanson (President and Chief Nuclear Officer-Exelon), "A. E. GINNA NUCLEAR POWER PLANT -REQUEST FOR ADDITIONAL INFORMATION REGARDING: RISK-INFORMED TECHNICAL SPECIFICATIONS INITIATIVE SB (CAC NO. MF6358), dated January 7, 2016. 3. Letter from James Barstow (Exelon) to U.S. Nuclear Regulatory Commission, "Supplemental Information Regarding TSTF-425 License Amendment Request," dated October 2, 2015. By letters dated June 4, 2015 (Reference 1) and supplemented in October 2, 2015 (Reference 3) Exelon Generation Company, LLC (Exelon) requested to change the R. E. Ginna (Ginna) Technical Specifications (TS). On November 12 and December 4, 2015, the U.S. Nuclear Regulatory Commission (NRC) identified areas where additional information was necessary to complete the review. On January 7, 2016, (Reference 2) NRC issued its final Request for Additional Information (RAI).

U.S. Nuclear Regulatory Commission Response to Request for Additional Information Docket No. 50-244 February 3, 2016 Page 2 Attachment 1 to this letter contains the NRC's request for additional information as documented in the January 7, 2016 letter immediately followed by Exelon's response. Attachment 2 contains the revised TS Bases pages. Additionally, Attachment 7 from the initial submittal; "INSERT 2" contained editorial errors in parts 5.5.17.b and 5.5.17.c. Attachment 3 contains a revised "INSERT 2." Exelon 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 information provided in this response does not affect the bases for concluding that the proposed license amendment does not involve a significant hazards consideration. Furthermore, the additional information provided in this response 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 commitments contained in this response. If you should have any questions regarding this submittal, please contact Enrique Villar at 61 0-765-5736. I declare under penalty of perjury that the foregoing is true and correct. Executed on the 3rd day of February 2016. Respectfully, David T. Gudger Manager -Licensing & Regulatory Affairs Exelon Generation Company, LLC Attachments: 1. Response to Request for Additional Information cc: 2. Revised Technical Specifications Bases Pages 3. Revised INSERT 2 USNRC Region I Regional Administrator USNRC Senior Resident Inspector -Ginna USNRC Project Manager, NRR -Ginna A. L. Peterson, NYSERDA w/attachments ATTACHMENT 1 License Amendment Request R. E. Ginna Nuclear Power Plant Docket No. 50-244 Application for Technical Specification Change Regarding Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3) Response to Request for Additional Information License Amendment Request Response to Request for Additional Information Docket No. 50-244 REQUEST FOR ADDITIONAL INFORMATION REGARDING ADOPTION OF TSTF-425 EXELON GENERATION COMPANY, LLC R. E. GINNA NUCLEAR POWER PLANT DOCKET NO. 50-244 Attachment 1 Page 1 of 46 In a letter dated June 4, 2015 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 15166A075), Exelon Generation Company, LLC, (Exelon, the licensee), submitted an application for a proposed amendment to the Technical Specifications (TSs) (or license or licensing basis) for R. E. Ginna Nuclear Power Plant (Ginna), which would modify TSs by relocating specific surveillance frequencies to a licensee-controlled program with the implementation of Nuclear Energy Institute 04-10, "Risk-Informed Technical Specifications Initiative Sb [RITS-5b], Risk-Informed Method for Control of Surveillance Frequencies." The Nuclear Regulatory Commission (NRC) staff is reviewing the submittal and has the following questions: Division of Risk Assessment/PAA Licensing Branch In Attachment 2 of the license amendment request (LAR), assessment of the technical adequacy of the Ginna Internal Events Probabilistic Risk Assessment (PRA) is based primarily on the 2009 peer review. As required by Regulatory Guide (RG) 1.200, Revision 2, document all the individual findings and selected suggestions, i.e., those suggestions for which the reference supporting requirements (SR) changed between the 2007 version of the American Society of Mechanical Engineers/ American Nuclear Society (ASME/ANS) PRA Standard, as clarified by Revision 1 to RG 1.200, and the 2009 version of the Standard, as clarified and qualified by Revision 2 of RG 1.200, resulting from the 2009 internal events peer review, and their disposition, whether or not they have been closed (unless closed via a subsequent peer review, full or focused-scope). Include discussion as to whether the disposition applies to changes in risk as well as the base-line risk, since the peer review is against the latter, but the application involves the former as well. Exelon Response to RA/ 1 Table 2-1 from the TSTF-425 LAR submittal [2] has been updated to disposition the findings with respect to their current status and to note the potential impact on changes in risk as well as base-line risk. The updated table 2-1 is provided below. Changes compared to the original Table 2-1 are shown in italics. Table 2-1 provided in this submittal supersedes the original Table 2-1 submitted on June 4, 2015.

License Amendment Request Response to Request for Additional Information Docket No. 50-244 Attachment 1 Page 2 of 46 Of the 34 suggestions from the Ginna peer review, excluding formatting and very minor editorial changes, only 9 of the suggestions were associated with changes to the reference supporting requirements (SR) which changed between the 2007 version of the American Society of Mechanical Engineers/ American Nuclear Society (ASME/ ANS) PRA Standard, as clarified by Revision 1 of RG 1.200, and the 2009 version of the Standard, as clarified and qualified by Revision 2 of RG 1.200. The current disposition of the applicable suggestions is provided in Table 2-2.

License Amendment Request Response to Request for Additional Information Docket No. 50-244 Updated Table 2-1 Internal Events PAA Peer Review -Findings SR IE-C12 [2005: IE-ClO] URE 845 Topic COMPARE results and EXPLAIN differences in the initiating event analysis with generic data sources to provide a reasonableness check of the results. Status Open Finding/Observation F&O IE-Cl0-01: The Ginna Initiating Event Notebook {Gl-IE-0001, Rev. 1} Section 4.3 provides a cross-reference between the Ginna Initiating Events and the "NRC Rates of Initiating Events" in table 4-7. Table 4-7 cross-reference includes columns for NUREG/CR-5750 Category and NP-2230 EPRl/NUREG/CR-3862 PWR Category. Table 4-8 provides a cross-reference between Ginna and similar PWR plants {Point Beach, Prairie Island, and Kewaunee). An explanation of differences in Initiating Events between Ginna and similar PWRs is contained in the PRA Quantification (QU) Notebook (Gl-QU-0001, Rev. O) Table 4-5 "Comparison of Ginna Core Damage Results to Similar Plants". However, no explanation of differences between plant-specific initiating events and generic initiating events was located in either the Initiating Event Notebook {Gl-IE-0001, Rev. 1) or QU Notebook (Gl-QU-0001, Rev. O). Disposition During the 2015 model update, the Initiating Event notebook is updated with a comparison of the frequencies of the plant-specific initiating events and generic initiating events. In most cases, the plant-specific IE frequencies are comparable to generic frequencies. However, in some cases, electrical bus failures were lower for the site-specific modeling, due to crediting operator actions for recoveries prior to trip. Attachment 1 Page 3 of 46 Impact to TSTF-425 The difference in the loss of bus initiating events is captured in URE 1202, which will be reviewed for applicability for each ST/ change evaluation as required by Exelon procedural guidance.

SR IE-ClS [2005: IE-C13] License Amendment Request Response to Request for Additional Information Docket No. 50-244 Topic CHARACTERIZE the uncertainty in the initiating event frequencies and PROVIDE mean values for use in the quantification of the PRA results. Updated Table 2-1 Internal Events PRA Peer Review -Findings Status Finding/Observation Disposition Complete F&O IE-C13-01: Gl-IE-0001, PRA INITIATING Assumptions and Uncertainties have EVENT (IE) NOTEBOOK, Section 5 documents been added to the Ginna Initiating assumptions and sources of uncertainty. Event notebook. Where generic data However, section 5 does not provide or was used, the error factors from the reference the parametric uncertainty initiating event data distribution. For example, the distribution for TIGRLOSP is identified in the CAFTA model, newauto_65a-w-Fld.caf, has having an EF of 7.39. However, no documentation for the error factor could be found. Therefore, this SR is not met. generic data are used as input to the Bayesian update process and updated accordingly with plant-specific evidence. Attachment 1 Page 4 of 46 Impact to TSTF-425 This F&O has been addressed with the current PRA model and documentation, and does not impact the TSTF-425 analysis.

SR SC-A2 License Amendment Request Response to Request for Additional Information Docket No. 50-244 Updated Table 2-1 Internal Events PAA Peer Review -Findings Topic SPECIFY the plant parameters (e.g., highest node temperature, core collapsed liquid level) and associated acceptance criteria (e.g., temperature limit) to be used in determining core damage. Select these parameters such that determination of core damage is as realistic as practical, in a manner -consistent with current best practice. DEFINE computer code-predicted acceptance criteria with sufficient margin on the code-calculated values to allow for limitations of the code, sophistication of the models, and uncertainties in the results, in a manner consistent with the requirements specified under HLR-SC-B. Examples of measures for core damage suitable for Capability Category 11/111, that have been used in PRAs, include (a) collapsed liquid level less than 1?3 core height or code-predicted peak core temperature >2,SOO"F (BWR) (b) collapsed liquid level below top of active fuel for a prolonged period, or dicted core peak node temperature >2,200"F using a code with detailed core modeling; or code-predicted core peak node temperature >1,800"F using a code with simplified (e.g., single-node core model, lumped para-meter) core modeling; or code-predicted core exit temperature >l,200"F for 30 min using a code with simplified core modeling (PWR). Status Open Finding/Observation F&O SC-A2-01: The definition of core damage documented in the Notebook-Rev-1 Section 2.2 is consistent with the examples of measures for core damage suitable for Capability Category I as defined in NUREG/CR-4550. For Category II Ginna could use the code-predicted core exit temperature >l,200°F for 30 min using PCTRAN (code with simplified core modeling (PWR)). Disposition It is acknowledged that the approach taken in the Ginna PRA is conservative and not fully consistent with the requirements of Category II. The peer reviewers suggested using a core exit temperature of 1200°F for 30 minutes as the criterion for core damage, but we would recommend using either that criterion or a peak core node temperature of 1800°F. Based on a review of the PCTRAN results, it is likely that the 1800°F peak core temperature would be reached earlier than the time at which the core exit temperature would be greater than 1200°F for 30 minutes. Attachment 1 Page 5 of 46 Impact to TSTF-425 Over the typical complete loss of decay heat removal timing success criteria, the delta time between core uncovery and CET temperatures reach 1200°F for 30 minutes or 1800° peak center line is fairly small. As such, the timing benefit is not expected to be large except for the fast moving events such as large break LOCAs. For these events, we use the UFSAR success criteria. Although this is not expected to be a significant effect, this SR remains CAT/, with potentially conservative overall risk results. Although no other cases are identified where core uncovery does not lead to core damage in short order, this issue is captured as URE 0838 which will be reviewed for applicability for each ST/ change evaluation as required by Exelon procedural guidance.

SR SC-A4 License Amendment Request Response to Request for Additional Information Docket No. 50-244 Updated Table 2-1 Internal Events PRA Peer Review -Findings Topic IDENTIFY mitigating systems that are shared between units, and the manner in which the sharing is performed should both units experience a common initiating event (e.g., LOOP). Status Finding/Observation Complete F&O SC-A4-02 -Operator action RCHFDXlBAF (operator fails to align BAF given 1 of 2 PORVs and no charging) is not included in the fault tree model. It appears that this event should be added in Event Tree TIU Sequence 5 Failures under gate TL_FB. This is an omission in the model and may affect CDF and LERF. Disposition Add RCHFDXlBAF to the Event Tree TIU, as appropriate. Attachment 1 Page 6 of 46 Impact to TSTF-425 No impact to TSTF 425. Action placed in Event Tree TIU logic and Finding addressed.

License Amendment Request Response to Request for Additional Information Docket No. 50-244 Updated Table 2-1 Internal Events PAA Peer Review -Findings SR SY-AlO [SY-All -2005] Topic INCORPORATE the effect of variable success criteria (i.e., success criteria that change as a function of plant status) into the system modeling. Example causes of variable system success criteria are (a) different accident scenarios. Different success criteria are required for some systems to mitigate different accident scenarios (e.g., the number of pumps required to operate in some systems is dependent upon the modeled initiating event). (b) dependence on other components. Success criteria for some systems are also dependentonthesuccessofanother component in the system (e.g., operation of additional pumps in some cooling water systems is required if noncritical loads are not isolated). (c) time dependence. Success criteria for some systems are time-dependent (e.g., two pumps are required to provide the needed flow early following an accident initiator, but only one is required for mitigation later following the accident). (d) sharing of a system between units. Success criteria may be affected when both units are challenged by the same initiating event (e.g., LOOP). Status Finding/Observation Complete SY-All-01-Gate TL_FBHRDl input to gate TL_FB for failure of Bleed and Feed models success as requiring 1 SI pump and 1 PORV. The logic does not include 75 gpm charging flow which is noted in the Success Criteria notebook as required to support single PORV success. This was confirmed through discussion with Ginna PRA personnel. The omission of a needed mitigating system for support of the Bleed and Feed function may underestimate the importance of these sequences for applications. Disposition Review the Bleed and Feed modeling to ensure the system failures appropriately reflect the success criteria. Attachment 1 Page 7 of 46 Impact to TSTF-425 No impact as the Finding has been addressed and the logic has been updated and documented in the Success Criteria Notebook.

License Amendment Request Response to Request for Additional Information Docket No. 50-244 Updated Table 2-1 Internal Events PRA Peer Review -Findings SR SY-A14 [SY-A13 2005] Topic When identifying the failures in SY-All INCLUDE consideration of all failure modes, consistent with available data and model level of detail, except where excluded using the criteria in SY-AlS. For example, (a) active component fails to start (b) active component fails to continue to run (c) failure of a closed component to open (d) failure of a closed component to remain closed (e) failure of an open component to close (f) failure of an open component to remain open (g) active component spurious operation (h) plugging of an active or passive component (i) leakage of an active or passive component (j) rupture of an active or passive component (k) internal leakage of a component (I) internal rupture of a component (m) failure to provide signal/operate (e.g., instrumentation) (n) spurious signal/operation (o) pre-initiator human failure events (see SY-A16) (p) other failures of a component to perform its required function Status Finding/Observation Complete SY-A13-02 -Inconsistencies existed in the system modeling of the city water system. Where used to support the GE-Betz system, a basic event for unavailability of city water due to grid LOOP was added (basic event CDAACITYWATER). This same event was not added to the city water modeling for support of the SAFW system. Disposition Review the need to add the unavailability event in the SAFW System. Attachment 1 Page 8 of 46 Impact to TSTF-425 No impact to TSTF 425. The dependencies for SAFW have been updated in the Ginna PRA.

License Amendment Request Response to Request for Additional Information Docket No. 50-244 Updated Table 2-1 Internal Events PRA Peer Review -Findings SR SY-A19 [SY-A18 2005] Topic In the systems model, INCLUDE service unavailability for components in the system model, unless screened, in a manner consistent with the actual practices and history of the plant for removing equipment from service. (a) INCLUDE (1) unavailability caused by testing when a component or system train is reconfigured from its required accident mitigating position such that the component cannot function as required (2) maintenance events at the train level when procedures require isolating the entire train for maintenance (3) maintenance events at a sub-train level (i.e., between tagout boundaries, such as a functional equipment group) when directed by procedures (b) Examples of out-of-service unavailability to be modeled are as follows: (1) train outages during a work window for preventive/corrective maintenance (2) a functional equipment group (FEG) removed from service for preventive/ corrective maintenance (3) a relief valve taken out of service Status Finding/Observation Complete SY-A18-0l -Ginna PRA System Notebooks provides a list of all the modeled T&M terms in Section 3.4.C. Section 2.9 of the notebooks provide discussion of procedures and testing that result in Unavailability. The review of these sections found no instances of simultaneous unavailability that can result from planned activities. However, the PRA engineer noted in a discussion that some systems are shadowed in planned maintenance. There is not a specific discussion on plant maintenance practices within the (a)(4} program that would result in planned unavailability of multiple systems OOS (i.e., EDG outages combined with AFW motor driven pump outages to lower total risk as opposed to performing the work independently), or of planned activities resulting in multiple components OOS that do not violate technical specifications (e.g., two AFW pumps in maintenance or an AFW and SAFW pump in maintenance at the same time). If work is done in this manner, it may be appropriate to account for the unavailability of both SSCs in a single term. Modeling of station maintenance practices that result in planned maintenance evolutions with more than a single PRA component OOS (i.e., shadowing equipment outages) can help to minimize the number of random failure sequences and ensure there is not "double counting" of unavailability in the PRA. Disposition The Ginna maintenance scheduling practices are, when two functional equipment groups are scheduled to be out-of-service in the same week, that the FEGs are sequenced rather than working them simultaneously. Exceptions are rare and are assessed. The Ginna PRA model will include some random combinations of maintenance configurations. Certain overlapping configurations are explicitly excluded from the model, such as taking out of service both trains of a two-train Tech. Spec. system. Attachment 1 Page 9 of 46 Impact to TSTF-425 This issue has been addressed with the current PRA model and documentation, and has no impact on TSTF-425 analysis.

SR HR-G3 License Amendment Request Response to Request for Additional Information Docket No. 50-244 Updated Table 2-1 Internal Events PRA Peer Review -Findings Topic When estimating HEPs EVALUATE the impact of the following plant-specific and scenario-specific performance shaping factors: (a) quality [type (classroom or simulator) and frequency] of the operator training or experience (b) quality of the written procedures and administrative controls (c) availability of instrumentation needed to take corrective actions (d) degree of clarity of the cues/indications (e) human-machine interface (f) time available and time required to complete the response (g) complexity of the required response (h) environment (e.g., lighting, heat, radiation) under which the operator is working (i) accessibility of the equipment requiring manipulation (j) necessity, adequacy, and availability of special tools, parts, clothing, etc. Status Finding/Observation Complete F&O HR-63-01: Details regarding certain elements of the analysis were lacking in the HRA Calculator for a sufficient number of HFEs to judge that this requirement was not met. Evidence that the relevant aspects cited in the SR are addressed for each HFE is critical to assuring that an appropriate analysis has been performed. This is particularly important in the case of HRA, for which the methods are less straightforward than they are for many other parts of the PRA. Disposition This F&O was addressed in the Fire HRA Notebook for the fire related human actions. This included almost all of the non-fire related HRA events as most of the non-fire related HRAs are included in the fire model as well. Consideration of cue clarity and complexity were considered as part of the 2015 internal events model update for Ginna. Any and all additions to cue clarity and complexity have been incorporated into the HRA Calculator database file for the FPIE model, and will also be incorporated in Appendix I of the 2015 Ginna FPIE HRA Notebook. As such, the 2015 internal events PRA model update is consistent with HR-G3 including the clarifications provided in RG 1.200, Revision 2. Attachment 1 Page 10 of 46 Impact to TSTF-425 No impact to TSTF 425. The HRAs have been reviewed to ensure the needed parameters for the evaluation have been populated. CBDM is now used as a max function of CBDT and HCR/ORE. RCHFDMAKEUP as a specific example has a timing basis from Key Input 51. When the annunciator model is used, there is a clear discussion as to the applicability.

SR HR-11 QU-BS License Amendment Request Response to Request for Additional Information Docket No. 50-244 Attachment 1 Page 11 of 46 Updated Table 2-1 Internal Events PAA Peer Review -Findings Topic DOCUMENT the human reliability analysis in a manner that facilitates PRA applications, upgrades, and peer review. Fault tree linking and some other modeling approaches may result in circular logic that must be broken before the model is solved. BREAK the circular logic appropriately. Guidance for breaking logic loops is provided in NUREG/CR-2728 [2-13]. When resolving circular logic, DO NOT introduce unnecessary conservatisms or non-conservatisms. Status Finding/Observation Complete F&O HR-11-01: The bulk of the documentation for the HRA is provided in the HRA Calculator. There are numerous areas in which the documentation is incomplete. The documentation should include a fuller discussion of the cues, bases for timing, specific procedure steps, and other aspects that could affect the analyses. Open F&O QU-BS-01: In Section 3.1 of the QU Notebook, a mention is made that circular logic checks were performed on the integrated top logic model to ensure it did not exist. An example is listed, but there is no further discussion. System notebooks reviewed generally state in Section 3.3 what was done when circular logic was identified, but no discussion of the methodology was provided nor how conservatisms or conservatisms are avoided. No evidence that the required analysis was not performed. Disposition Impact to TSTF-425 Documentation only. Same issue as for No impact to TSTF 425. This item HR-G3. has been addressed. See HR-G3. Documentation only: Provide a discussion in the Quantification Notebook Section 3.1 of the methodology used to address circular logic. The circular logic process is revealing when a support gate is added to the tree the CAFTA software identifies a circular logic issue. The circular logic is broken by inserting as much of the logic clip into the tree as possible. Providing more examples of this in the documentation is not expected to affect the TSTF-425 evaluation.

SR LE-C2 [2005: LE-C2a] License Amendment Request Response to Request for Additional Information Docket No. 50-244 Updated Table 2-1 Internal Events PRA Peer Review-Findings Topic Status Finding/Observation INCLUDE realistic treatment of Open F&O LE-C2a-Ol: It is conservative to NOT feasible operator actions following take credit for operator actions post core the onset of core damage consistent damage. This is a requirement of the with applicable procedures, e.g., standard to move from Category I to EOPs/SAMGs, proceduralized actions, Category II. or Technical Support Center guidance. Disposition There are limited operator actions that could influence LERF at Ginna, so the effect of such actions is not likely to be significant. Moreover, it is likely that there will not be a need for a Category II rating in this area to meet the requirements for most risk-informed applications. One approach to reaching Category II would be to include post-core damage operator actions in the PRA. It is also possible that simply identifying operator actions and showing quantitatively that they will have a negligible impact on LERF will be sufficient to meet the requirements of Category II. Attachment 1 Page 12 of 46 Impact to TSTF-425 There are limited operator actions that could influence LERF at Ginna, so the effect of such actions is not likely to be significant. If post-core-damage operator actions are credited, LERF estimates could be reduced, but the impact would be minimal. The omission of these operator actions is conservative and does not adversely impact the use of the model for TSTF-425 analysis/or risk increases or baseline risk. URE 0835 is open to develop a post core-damage RCS depressurization recovery. As part of the Exelon program, open UREs are evaluated for impact on the TSTF risk assessments.

SR LE-Cll [2005: LE-C9a] LE-C13 [2005 LE-ClO] License Amendment Request Response to Request for Additional Information Docket No. 50-244 Updated Table 2-1 Internal Events PRA Peer Review -Findings Topic JUSTIFY any credit given for equipment survivability or human actions that could be impacted by containment failure. PERFORM a containment bypass analysis in a realistic manner. JUSTIFY any credit taken for scrubbing (i.e., provide an engineering basis for the decontamination factor used). Status Finding/Observation Open F&O LE-C9a-Ol: It does not appear that credit was taken for continued operation of equipment and operator actions that could be impacted by containment failure. This is a requirement of the standard to move from Category I to Category II. Open F&O LE-Cl0-01: Credit for scrubbing was not taken. A sensitivity for impact of scrubbing was performed and it was determined that the impact of not considering scrubbing is negligible. This is a requirement of the standard to move from Category I to Category II. Disposition The requirement is to justify credit taken for equipment survivability or human actions that could be affected by containment failure. In the Ginna Level 2 Analysis, early containment failure after damage and vessel breach is the end-state for the LERF accident progression. There are no equipment dependencies or human actions that are identified that could be reasonably credited to prevent a release through a failed containment. Review the possible credit for release scrubbing to reduce LERF. Attachment 1 Page 13 of 46 Impact to TSTF-425 As no equipment or HRA is credited post-containment failure, the PRA model remains a conservative CAT I. This does not adversely impact the use of the model for TSTF-425 analysis for risk increases or baseline risk. A sensitivity for impact of scrubbing was performed and it was determined that the impact of not considering scrubbing is negligible for base LERF conditions. However, a review identified that the importance of feedwater during SGTR cases may be masked due to this conservatism. URE 0834 is open to credit scrubbing in the LERF analysis. While this URE remains open, it will be evaluated for any potential impacts to a surveillance interval evaluation.

SR MU-Dl License Amendment Request Response to Request for Additional Information Docket No. 50-244 Updated Table 2-1 Internal Events PRA Peer Review-Findings Topic A PRA Configuration Control Program shall be in place. It shall contain the following key elements: (a) a process for monitoring PRA inputs and collecting new information (b) a process that maintains and upgrades the PRA to be consistent with the as-built, as operated plant (c) a process that ensures that the cumulative impact of pending changes is considered when applying the PRA (d) a process that maintains configuration control of computer codes used to support PRA quantification (e) documentation of the Program Status Finding/Observation Complete F&O MU-Dl-01 -PRA Configuration Control procedure (GNG-CM-1.01-3003) Step 5.13 provides guidance for updating informed applications. The process described depends upon a database maintained by the Fleet PRA Services Supervisor to identify current living applications requiring change assessment other than those related to maintenance rule performance criteria. No such database could be identified for Ginna. Without a current list of risk-informed applications, the maintenance and update process is dependent upon the knowledge and experience of the staff to know which applications require update. This creates the possibility that an application could be missed in the update process. Disposition The CRMP database has a placeholder for a listing of PRA applications. This portion of the database has been populated to ensure all applications requiring update following a model revision can be easily identified. Attachment 1 Page 14 of 46 Impact to TSTF-425 This configuration control issue has been addressed. No impact to TSTF 425.

License Amendment Request Response to Request for Additional Information Docket No. 50-244 Updated Table 2-1 Internal Events PRA Peer Review -Findings SR IFSO-A4 [2005: IF-B2] Topic For each potential source of flooding water, IDENTIFY the flooding mechanisms that would result in a fluid release. INCLUDE: (a) failure modes of components such as pipes, tanks, gaskets, expansion joints, fittings, seals, etc. (b) human-induced mechanisms that could lead to overfilling tanks, diversion of flow through openings created to perform maintenance; inadvertent actuation of fire suppression system (c) other events releasing water into the area Status Finding/Observation complete F&O IF-B2-0l: Failure mechanisms are addressed in conjunction with the calculation of flood frequencies, in Section 5.2 of document 51-9100978-000. Failures of components in piping systems other than tanks are explicitly addressed by the EPRI pipe failure data base. This was the source employed to characterize the frequencies of floods for Ginna. There has, however, been a very limited attempt to address induced flood mechanisms, as required by item (b) of SR IF-B2. Such events have been important causes of flooding in the operating experience for US nuclear power plants, and as noted above the assessment of such floods is explicitly required. A more systematic consideration should be made of human-caused floods. This will need to include an assessment of generic data related to human-caused floods, per SR IF-06. Disposition Address the potential for caused flooding in the Internal Flooding Study (51 -9100978 -000). Describe the situations where a human error could result in flooding (e.g., inadvertent valve opening, inadvertent train realignment, doors left open) and estimate the probabilities of such events. Model such floods that cannot be screened. Consistent with the Standard, utilize generic data as required by SR IFEV-A7 (IF-06 in 2005 Standard) Attachment 1 Page 15 of 46 Impact to TSTF-425 No impact to TSTF 425. Discussion of human caused floods is discussed in detail in Section 3.3 and 5.3 of Internal Flood Notebook (Gl-IF-0000-rl) for various systems. Based on the analyses performed, one maintenance induced flood was added to the model, SW -2,000 gpm SW flood in the Aux Building due to maintenance, isolated within 65 minutes.

License Amendment Request Response to Request for Additional Information Docket No. 50-244 Updated Table 2-1 Internal Events PRA Peer Review -Findings SR IFSN-A6 [2005: IF-C3] Topic For the SSCs identified in I FSN-A5 (2005 text: IF-C2c), IDENTIFY the susceptibility of each SSC in a flood area to flood-induced failure mechanisms. INCLUDE failure by submergence and spray in the identification process. EITHER: a) ASSESS qualitatively the impact of flood-induced mechanisms that are not formally addressed (e.g., using the mechanisms listed under Capability Category Ill of this requirement), by using conservative assumptions; OR b) NOTE that these mechanisms are not included in the scope of the evaluation. Status Finding/Observation Open F&O IF-C3-0l: There is no discussion of failures due to jet impingement or pipe whip. There is limited consideration of failure due to humidity/high temperature due to failure of heating steam lines. There is also no discussion of criteria employed to consider the potential for spray failures. To meet Capability Category II, it is necessary either to provide at least a qualitative assessment of the potential for jet impingement and pipe whip, or to state that these failure mechanisms were not considered. It is also required that potential spray failures be evaluated. While spray failures are discussed, there are no criteria specified that would provide assurance that they had been considered in a consistent and adequately comprehensive manner. Provide the requisite discussion of pipe whip and jet impingement to satisfy the standard. Specify appropriate criteria for spray impacts, and assure that the potential spray failures adequately reflect these criteria. Disposition Cat II: INCLUDE failure by submergence and spray in the identification process. ASSESS qualitatively the impact of flood-induced mechanisms that are not formally addressed (e.g., using the mechanisms listed under Capability Category Ill of this requirement), by using conservative assumptions. [SAIC note: these mechanisms include submergence, spray, jet impingement, pipe whip, humidity, condensation, temperature concerns] Revise the Internal Flooding Study (51-9100978 -000) to describe the criteria used to determine the potential for failure resulting from spray. Reference Appendix C for a listing of components impacted by spray. Describe how potential spray impact was addressed in the model. Confirm that the assignment of spray impact is consistent with the criteria used. In addition, include a qualitative discussion of the potential impact of jet impingement, pipe whip, humidity, condensation, and temperature effects. Attachment 1 Page 16 of 46 Impact to TSTF-425 Failures due to jet impingement and pipe whip are now discussed in Section 3.3.1 of the Internal Flood Notebook Gl-IF-0000 rl. Failures due to Spray are discussed in Section 3.3.2. Impacts due to spray were assumed to exist within 10 feet of a break location (modeled?). Spray events are discussed in the IF Flood notebook Section 4.5. Two locations were identified in the Aux Building where Fire Service Water could impact safety related busses and these are explicitly modeled FSW-BUS15 and BUS14). URE 1179 documents that IF Notebook needs Appendix C completed to complete documentation of spray impacts and modeling of additional spray floods if appropriate. This would be evaluated for any potential impacts to a surveillance frequency interval extension at the time of the evaluation but is not expected to have a significant impact.

License Amendment Request Response to Request for Additional Information Docket No. 50-244 Updated Table 2-1 Internal Events PRA Peer Review -Findings SR IFSN-A8 [2005: IF-C3b] Topic IDENTIFY inter-area propagation through the normal flow path from one area to another via drain lines; and areas connected via back flow through drain lines involving failed check valves, pipe and cable penetrations (including cable trays}, doors, stairwells, hatchways, and HVAC ducts. INCLUDE potential for structural failure (e.g., of doors or walls) due to flooding loads. Status Finding/Observation complete F&O IF-C3b-01: The analysis does not document consideration of potential barrier failures due to flooding loads (structural failures, failures of doors, etc.) This is required to meet capability categories beyond Capability Category I. Review flood barriers and identify and evaluate any whose failures could contribute adversely to propagation of flooding Disposition Cat II, Ill: IDENTIFY inter-area. INCLUDE potential for structural failure (e.g., of doors or walls) due to flooding loads and the potential for barrier unavailability, including maintenance activities. Include a discussion of the potential for barrier failure due to flooding, including structures and doors. For walls, a qualitative discussion would appear to be acceptable. For doors, however, specific failure criteria should be developed and described. Flood scenarios should be reviewed and revised, if necessary, to address the potential failure of doors. Attachment 1 Page 17 of 46 Impact to TSTF-425 No impact to TSTF 425. A discussion of structural failure of barriers credited as barriers has been added to the IF Notebook rl, Section 4.2.1.

SR A16 [2005: IF-CB] License Amendment Request Response to Request for Additional Information Docket No. 50-244 Updated Table 2-1 Internal Events PRA Peer Review-Findings Topic USE potential human mitigative actions as additional criteria for screening out flood sources if all the following can be shown: (a) flood indication is available in the control room; (b) the flood source can be isolated; and (c) the mitigative action can be performed with high reliability for the worst flooding initiator (2005 text: flood from that source). High reliability is established by demonstrating, for example, that the actions are procedurally directed, that adequate time is available for response, that the area is accessible, and that there is sufficient manpower available to perform the actions. Status Finding/Observation Complete F&O IF-CS-01: Only one flood appears to have been screened based on qualitative consideration of potential human action; for that action (2000 gpm FSW break in IBN), there doesn't appear to be any justification for the time identified (190 min). Nothing other than time available is cited as rationale for screening the event. To meet Capability Category II, it is necessary to characterize potential human actions that could terminate flooding more explicitly than was done in this case. Address the required aspects for this and any other human actions used in justifying screening out flood scenarios. Disposition The FSW breaks in the IBN are no longer screened in the current model. New flood initiator FL-/BN-FSW-2K has been added to the model and to the Flood Notebook. Attachment 1 Page 18 of 46 Impact to TSTF-425 This F&O has been addressed in the current model and documentation and has no impact on TSTF-425 analysis.

License Amendment Request Response to Request for Additional Information Docket No. 50-244 Updated Table 2-1 Internal Events PRA Peer Review -Findings SR IFEV-AG [2005: IF-D5a] Topic Status Finding/Observation GATHER plant-specific information Complete F&O IF-DSa-01: The current analysis does on plant design, operating practices, not adequately address plant-specific and conditions that may impact flood characteristics that might affect the manner likelihood (i.e., material condition of fluid systems, experience with water hammer, and maintenance-induced floods). In determining the initiating event frequencies for flood scenario groups, USE a combination of the following (2005 text does not include "of the following") (a) generic and plant-specific operating experience; (b) pipe, component, and tank rupture failure rates from generic data sources and plant-specific experience; {2005 text: and) (c) engineering judgment for consideration of the plant-specific information collected. in which the frequencies of flooding are estimated. To meet Capability Category II, it is required that plant-specific information be collected and considered on a variety of aspects (including material condition of fluid systems, experience with water hammer, and maintenance-induced floods). The current analysis is limited to the use of generic failure rates. This is consistent with Capability Category I. Address potential issues with material condition, experience with water hammer, etc. In particular, further attention should be paid to the possibility of induced and other human-caused flooding. Disposition In the current updated internal flood analysis, a review was conducted to assess potential issues with material condition, water hammer, and aging management strategies. The plant specific information has been considered and use of generic data is found to be appropriate for Ginna. For maintenance-induced and other human-caused flooding, see IFSO-A4 Which is statused as "Complete.". Attachment 1 Page 19 of 46 Impact to TSTF-425 Plant specific experience with internal flooding, water hammer is addressed in the IF Notebook rev 1 in Sections 3.3. A discussion of Human-induced floods is contained in Section 5.3. This F&O has been addressed and does not impact TSTF-425 analysis.

License Amendment Request Response to Request for Additional Information Docket No. 50-244 Updated Table 2-1 Internal Events PRA Peer Review -Findings SR IFEV-A7 [2005: IF-D6] Topic INCLUDE consideration of induced floods during maintenance through application of generic data. Status Finding/Observation Complete F&O IF-06-01: Initiating events that could result from human actions were considered only for a small number of possible maintenance activities. These flood contributors were not evaluated using generic data as required. Operating experience for nuclear power plants has provided evidence that caused floods can be important. The SR requires that such floods be evaluated using at least generic data to meet Capability Category I or II. Perform a more detailed assessment of potential human-caused floods, and apply at least generic data to characterize their frequencies. Disposition See I FSO-A4. Attachment 1 Page 20 of 46 Impact to TSTF-425 No impact to TSTF 425. Discussion of human caused floods is discussed in detail in Section 3.3 and 5.3 of Internal Flood Notebook {Gl-IF-0000-rl) for various systems. Based on the analyses performed, one maintenance induced flood was added to the model, SW -2,000 gpm SW flood in the Aux Building due to maintenance, isolated within 65 minutes.

License Amendment Request Response to Request for Additional Information Docket No. 50-244 SR IFEV-A8 [2005: IF-D7] Topic SCREEN OUT flood scenario groups if (a) the quantitative screening criteria in IFSN-AlO (2005 text: IE-C4}, as applied to the flood scenario groups, are met; OR (b) the internal flood-initiating event affects only components in a single system, AND it can be shown that the product of the frequency of the flood and the probability of SSC failure given the flood is two orders of magnitude lower than: the product of the non-flooding frequency for the corresponding initiating events in the PRA, AND the random (non-flood-induced) failure probability of the same SSCs that are assumed failed by the flood. If the flood impacts multiple systems, DO NOT screen on this basis. Updated Table 2-1 Internal Events PRA Peer Review -Findings Status Finding/Observation Disposition Complete F&O IF-07-01: Quantitative screening of Update the Internal Flooding Study (51 some scenarios was performed, but it is not -9100978 -000} to describe the criteria clear what criteria were applied in doing so. used to screen flood scenarios. If The criteria should be defined and applied in current screening criteria are not well a clear and consistent manner. defined, develop such criteria and apply them to scenarios addressed in SRs IF-D7 and IF-E3a provide explicit criteria the analysis. for performing quantitative screening of flood scenarios. The IF Notebook documents that some scenarios were screened on low frequency, but does not invoke any particular criteria in doing so. Provide a clear set of criteria for performing quantitative screening of flood scenarios, and apply the criteria in a clear and consistent manner. Attachment 1 Page 21of46 Impact to TSTF-425 No impact to TSTF 425. This issue has been addressed. Internal Flood Notebook Section 4.6, Screening Scenarios and Sources, was updated to document the screening criteria used. Figure 4.1, was added which shows the Screening Criteria and Table 4.6 was edited to show the screening criterion used for various flood scenarios.

License Amendment Request Response to Request for Additional Information Docket No. 50-244 Updated Table 2-1 Internal Events PRA Peer Review -Findings SR IFQU-AS [2005: IF-ES] Topic If additional human failure events are required to support quantification of flood scenarios, PERFORM any human reliability analysis in accordance with the applicable requirements described in 2-2.5 (2005 text: Tables 4.5.5-2(e) through 4.5.5-2(h)). Status Finding/Observation complete F&O IF-ES-01: It was not clear that the requirements were met in all cases. For example, interviews to establish aspects such as response times were apparently performed as part of the flood analysis, but the HRA was dramatically changed and new interviews/changes were not incorporated, nor were any inputs obtained from the HRA performed as part of the flood analysis carried forward. It is necessary to perform the assessment of HFEs associated with internal flooding in the same manner as for other HFEs. The requirements to confirm procedure paths, timing, etc. via interviews with operators were not met for a number of events. Re-examine the HFEs associated with internal flooding, and either perform needed operator interviews or identify and document existing inputs. Disposition Re-examine each HFE included in the flooding analysis. Perform operator interviews as needed or identify and document previously performed interviews. Required operator interviews should comprise the following: 1. evaluate the flooding events based on similarities to identify a select set of scenarios to review with the operators (for example, categorized by the system that generated the flood, e.g., fire protection) 2. schedule interview sessions of about 1/2 hour to an hour per each flooding scenario, conducted separately with two different operators (preferably one experienced, one novice) to get diverse opinions. 3. include questions on timing consistent with the HRA Calculator Time Window screen for time of cue, time to diagnosis, time for execution/manipulation of action (including travel time, with potential related access delays). Be sure to ask about any differences for floods initiated in same system but in different rooms. 4. document interviews during the sessions (notes and/or tape recordings) and later in the HRA Calculator screens for Operator Interviews and Time Window. Estimate and document internal flooding HFEs using the same approach as was used for other HFEs in the PRA. Recalculate flood scenario frequencies based on the new HFEs. Attachment 1 Page 22 of 46 Impact to TSTF-425 No impact to TSTF-425. Ginna Station Flooding Human Reliability Analysis (HRA) documents the flood recovery actions (Areva Document No.: 51-9099406-000 located in GSN 0157). The information and HRA values in this notebook were verified to be consistent with the HRA actions being used in the internal flood model. No additional interviews were identified as being necessary.

License Amendment Request Response to Request for Additional Information Docket No. 50-244 Updated Table 2-1 Internal Events PRA Peer Review -Findings SR IFQU-Bl [2005: IF-Fl] Topic DOCUMENT the internal flood accident sequences and quantification in a manner that facilitates PRA applications, upgrades, and peer review. Status Open Finding/Observation F&O IF-Fl-01: The documentation is comprised primarily of the internal flooding notebook, supplemented heavily with information provided in a set of Excel worksheets. The notebook is annotated to provide a link to elements of the worksheets, and an "assumption" provides the formal tie between the notebook and the worksheets. Some areas in which the links were indirect or missing were noted. In general, the manner in which important parts of the flood analysis are documented in what would usually be characterized as an informal set of worksheets is judged not to meet the requirement that the analysis be documented in a manner that facilitates applications, upgrades, and peer review. In addition to developing a single integrated set of documentation for the internal flood analysis, there were several areas in which additional documentation would make the analysis more tractable have been provided in connection to other SRs. These include the following:

• Include a set of simplified arrangement drawings to explicate the definition of flood areas and help in understanding aspects such as flood propagation.

• Tabulate the flood areas and identify clearly which are screened and which retained for further analysis to make the process more tractable. Specify clearly which criteria (qualitative or quantitative) are employed in screening each flood area.

• Define explicitly the criteria used to perform quantitative screening as noted in Section 6.0.

• Define the criteria used to determine whether a PRA component was susceptible to failure due to spray. Disposition Documentation only: Revise the Internal Flooding Study (51-9100978 -000) to meet the documentation requirements of the 2009 Standard. Address NRC Resolutions as appropriate. It is recommended that the Study be reformatted to be consistent with the HLRs and SRs of the Standard, integrating appropriate parts ofthe worksheets into the primary document. This will provide a document that can be easily reviewed against the standard and easily followed by personnel not involved in the original analysis. Consistent with the F&O, include the following in the revised Study:

• Include a set of simplified arrangement drawings to explicate the definition of flood areas and help in understanding aspects such as flood propagation. *Tabulate the flood areas and identify clearly which are screened and which retained for further analysis to make the process more tractable. Specify clearly which criteria (qualitative or quantitative) are employed in screening each flood area.

• Define explicitly the criteria used to perform quantitative screening as noted in Section 6.0.

• Define the criteria used to determine whether a PRA component was susceptible to failure due to spray. Attachment 1 Page 23 of 46 Impact to TSTF-425 This documentation item will not impact the TSTF 425 analysis. This item has largely been addressed by adding tables in Section 5.2 that show the development of each initiating event frequency, adding an Initiating Event Summary Table (section 5.2.17), adding a simplified set of arrangement drawings showing each flood area (Appendix K), defining spray modeling criteria (Section 3.3.2) and identifying for each flood area whether it was screened and the screening criterion used (Table 4.6). The remaining item is to develop the criteria used to perform quantitative screening, if applicable, in Section 6.0 (URE 1177).

License Amendment Request Response to Request for Additional Information Docket No. 50-244 Attachment 1 Page 24 of 46 Updated Table 2-1 Internal Events PRA Peer Review -Findings SR IFQU-B3 [2005: IF-F3] Topic DOCUMENT sources of model uncertainty and related assumptions (as identified in QU-El and QU-E2) associated with the internal flood accident sequences and quantification. (2005 text: Document the key assumptions and the key sources of uncertainty associated with the internal flooding analysis.) Status Finding/Observation Complete F&O IF-F3-01: Section 7 of the IF Notebook provides a discussion of three areas considered to be major sources of uncertainty in the flood analysis. This does not constitute an adequate characterization of the sources of uncertainty associated with the flood analysis or a comprehensive discussion of the assumptions that could have an effect on the results. A reasonably thorough investigation of sources of uncertainty is necessary for proper characterization of the flood analyses and results. A more comprehensive characterization of sources of uncertainty, comparable to that provided for other areas of the PRA, should be developed for the internal flood analysis. Disposition Impact to TSTF-425 In estimating the event mean frequency This F&O has been addressed and for each internal flood initiator, the does not impact TSTF-425 initiating event uncertainty parameters analysis. from the EPRI 1013141 data were used and error factors reported in the Internal Flood analysis notebook (Gl-PRA-012). These parametric uncertainty values propagate to the end results using the CAFTA PRA software. Modeling uncertainty for the internal flood portion of the PRA was also addressed and documented in Gl-PRA-012 using the guidance found in EPRI 1016737. The finding for IF-F3-01 is considered to be resolved.

SR AS-A9 DA-Dl License Amendment Request Response to Request for Additional Information Docket No. 50-244 Attachment 1 Page 25 of 46 Table 2-2 Internal Events PRA Peer Review -Selected Suggestions Topic Status USE realistic, Open applicable (i.e., from similar plants) thermal hydraulic analyses to determine the accident progression parameters (e.g., timing, temperature, pressure, steam) that could potentially affect the operability of the mitigating systems. CALCULATE realistic Complete parameter estimates for significant basic events based on relevant generic and plant-specific evidence. Finding/Observation USE realistic, applicable (i.e., from similar plants) thermal hydraulic analyses to determine the accident progression parameters (e.g., timing, temperature, pressure, steam) that could potentially affect the operability of the mitigating systems. The Seal LOCA results are inconsistent with respect to generic industry data. Seal LOCA Cases RPSL364CD and RPSL960CD should be reviewed and compared with WCAP 16141. There is no explanation of how the composite data located in the third and second last columns of "Component Generic Failure Data" are determined from the three generic sources listed. The data appears to be reasonable, but the method used to develop it is not documented. Disposition Impact to TSTF-425 Consistent with Capability Category Ill for SR For this suggestion, no adverse As-A9, Ginna uses realistic, plant-specific impact is expected on the use of the thermal hydraulic analysis. Results are model for TSTF-425 analysis for risk reviewed for reasonableness; however, the increases or baseline risk. review for RCP Seal LOCAs is not documented, and should be re-performed. This issue is captured as URE 0850 which will be reviewed for applicability for each ST/ change evaluation as required by Exelon procedural guidance. The data notebook has been updated to This F&O has been addressed with support the latest PRA model update using the current PRA model and Bayesian update techniques or best available documentation, and does not impact generic data sources including specific the TSTF-425 analysis. references for the generic data sources.

SR DA-D6 DA-D6 License Amendment Request Response to Request for Additional Information Docket No. 50-244 Table 2-2 Internal Events PRA Peer Review -Selected Suggestions Topic Status USE generic common Open cause failure probabilities consistent with available plant experience. EVALUATE the common cause failure probabilities in a manner consistent with the component boundaries .. USE generic common Complete cause failure probabilities consistent with available plant experience. EVALUATE the common cause failure probabilities in a manner consistent with the component boundaries .. Finding/Observation The current modeling uses the failure to run alpha factor for the failure to run in the first hour. NUREG/CR-6268, Rev. 1, Table 5-7 indicates that events are coded as failure to run if the "component fails to continue running at rated conditions after reaching rated conditions." This implies that the failures to run in the first hour would be included in the failure to run group. Modeling of common cause for components separately where the database includes them in the boundary may result in slightly conservative results. The data used for the latest update was from an INEEL/NRC report published in 2002 (Key Input 1). Update the generic common cause data to a more current version. Disposition This suggestion F&O has not been incorporated into the updated PRA model or documentation. The data analysis and notebook have been updated using current CCF generic data. Attachment 1 Page 26 of 46 Impact to TSTF-425 For this suggestion, no adverse impact is expected on the use of the model for TSTF-425 analysis for risk increases or baseline risk. This issue is captured as URE 0820 which will be reviewed for applicability for each ST/ change evaluation as required by Exelon procedural guidance. This F&O has been addressed with the current PRA model and documentation, and does not impact the TSTF-425 analysis.

License Amendment Request Response to Request for Additional Information Docket No. 50-244 Table 2-2 Internal Events PRA Peer Review -Selected Suggestions SR DA-E3 IE-C12 [Now C14] IF-Bl [Now IFSO-Al] Topic DOCUMENT the sources of model uncertainty and related assumptions (as identified in QU-E1 and QU-E2) associated with the data analysis. In the ISLOCA frequency analysis, INCLUDE the features of plant and procedures that influence the ISLOCA frequency: For each flood area, IDENTIFY the potential sources of flooding Status Complete Open Complete Finding/Observation G1-DA-OOOO, Revision 1, Section 4.1 states that the "Ginna PRA assumed staggered testing for all components subject to CCF." This is not captured as an assumption which could contribute to uncertainty in Section 6.0 of the notebook. In addition, it was identified in discussion with Ginna PRA personnel that the MSIVs are not tested on a staggered basis. PRA Modeling methodology for pipe rupture analysis (for ISLOCA frequency estimation) has changed over the past decade. Recommend updating to the latest methodology has documented in latest EPRI tech report. The screening process appears to be adequate, but the manner in which criteria for screening are applied and the degree to which such criteria have been employed in a systematic manner is not clear. Disposition The previous assumption applicable to staggered testing is no longer applicable. The model includes staggered and non-staggered testing for CCF updates, as applicable. A data update has recently been completed for common cause factors. MSIVs are modeled as non-staggered. This suggestion remains open as it has not been formally closed-out This documentation suggestion has been addressed. The internal flood notebook was updated to summarize all the flooding areas, and if/why they were screened. Attachment 1 Page 27 of 46 Impact to TSTF-425 This F&O has been addressed with the current PRA model and documentation, and does not impact the TSTF-425 analysis. As the suggestion remains open until formally closed-out, this issue is captured as URE 0822 which will be reviewed for applicability for each ST/ change evaluation as required by Exelon procedural guidance. This F&O has been addressed with the current PRA model and documentation, and does not impact the TSTF-425 analysis.

SR LE-Cl QU-E4 License Amendment Request Response to Request for Additional Information Docket No. 50-244 Table 2-2 Internal Events PRA Peer Review -Selected Suggestions Topic Status Develop (LEAF) accident Open sequencestothelevelof detail to account for the potential contributors. For each source of model Complete uncertainty and related assumption identified in QU-E1 and QU-E2, respectively, IDENTIFY how the PRA model is affected (e.g., introduction of a new basic event, changes to basic event probabilities, change in success criterion, introduction of a new initiating event). Finding/Observation The definition of LEAF should include the basis for the declaration of General Emergency. Review of the EAL demonstrated that the General Emergency would be declared at CRFCs failure and the failure of containment heat removal does not contribute to LEAF. Table H-1 in the Quantification Notebook addresses each key source of uncertainty however no sensitivity studies related to these key sources of uncertainty are addressed with sensitivity studies. The Quantification notebook describes that uncertainties that are associated with scope and level of detail will only be addressed for specific applications. Cat 11 requires that sensitivity analysis be performed to address key assumptions. Disposition This documentation suggestion has not yet been addressed. The updated 2009 revision of the standard does not contain the requirement to perform sensitivity studies to meet Cat 11/111 as it is expected that specific sources of uncertainty will be addressed on an application specific basis. Attachment 1 Page 28 of 46 Impact to TSTF-425 This documentation suggestion would have no adverse impact on TSTF analysis for risk increase or baseline risk. This suggestion is not applicable to the 2009 ASME PRA standard and does not impact the TSTF-425 analysis.

License Amendment Request Response to Request for Additional Information Docket No. 50-244 Attachment 1 Page 29 of 46 The LAR indicates that a Fire PRA, associated with transition to NFPA-805, was performed and Peer Reviewed in August 2012. However, the facts and observations (F&Os) identified from the NFPA-805 Fire Peer Review were not provided for consideration in the LAR associated with RITS-5b changes to TS Surveillance Frequencies. The LAR states: The 2012 fire PRA peer review for the PRA ASME model update identified 183 Supporting Requirements (SR) to be reviewed for the Ginna PRA. Of these 2 were not met, 2 met capability category (CC) 1, 8 partially met CC 2, 17 met CC 2, 13 partially met CC 3, 7 met CC 3, and 118 fully met all capability requirements and 16 were not applicable. There were 19 findings and 22 suggestions issued to address potential gaps to compliance with the PRA standard. There were 3 Best Practices. All of the findings from the fire PRA peer review have since been closed. As the results of this peer review have already been communicated to the NRC as part of the NFPA-805 submittal and subsequent requests for additional information (RAI), these will not be catalogued in this document. Previous responses described above and in the NFPA-805, submittals are associated with assessing the PRA technical adequacy to address fire-related hazards. To the extent that there were deficiencies in the Fire PRA models associated with systems, structures, and components for which changes to TS Surveillance Frequencies are being sought, there is no equivalent clarification of how the Fire PRA related F&Os will not have an impact on the Technical Specifications Task Force (TSTF)-425, Revision 3. It is the NRC's position that Fire PRA related F&Os must be considered when evaluating TS Surveillance Frequency changes. Therefore provide the following: a. An assessment of how the 2012 Fire Peer Review F&Os have been resolved to assure PRA Technical Adequacy with respect to TSTF-425, not NFPA-805. Include discussion as to whether the disposition applies to changes in risk as well as the base-line risk, since the peer review is against the latter, but the application involves the former as well. b. For those Fire PRA related F&Os, which are dispositioned as not having an impact on TSTF-425, Revision 3, provide the technical basis for this determination. c. Discussion of how the licensee plans to incorporate updates to fire PRA state-of-the-art enacted since the 2012 peer review, including but not limited to updated fire ignition frequencies and non-suppression probabilities (as per NUREG-2169, "Nuclear Power Plant Fire Ignition Frequency and Non-Suppression Probability Estimation Using the Updated Fire Events Database") and updated spurious operation occurrence probabilities and probabilities for duration exceedance (as per NUREG/CR-7150, Volume 2, 11Joint Assessment of Cable Damage and Quantification of Effects from Fire"). d. Consistent with the requirements in Table A-4 of RG 1.200, Revision 2, clarify how the Fire PRA addresses the following requirements with regard to differential risk evaluations related to TSTF-425, Revision 3:

License Amendment Request Response to Request for Additional Information Docket No. 50-244 Attachment 1 Page 30 of 46 i. In SR FSS-A4, RG 1.200, Revision 2 changes "one of more" to 11sufficient.11 ii. In Fire PRA F&O FSS-F1-01, RG 1.200, Revision 2 changes SR FSS-F1 from 11one or more fire scenarios that could11 to 11a sufficient number of fire scenarios to characterize." iii. In Fire PRA F&O FSS-G5-01: Is potential failure of the wall water spray system to provide structural integrity of the boundary addressed? This includes the probability that the system does not perform its function such that the boundary could be breached and result in a multi-compartment fire scenario (e.g., the assumption of perfect reliability versus high reliability, is non-conservative). iv. In Fire PRA F&O SF-A 1-02, provide a disposition that addresses the item of concern, namely failure of the analysis to fully assess the potential impact of a seismically induced failure (rupture or spurious operation of fire protection features on the earthquake response). v. The disposition of SR FSS-G5 partly justifies reclassifying the F&O as CC II based on the disposition cited for F&O FSS-G5-01 discussed previously. The concern discussed previously needs to be resolved in order for the CC 11 assignment to be fully justified. Exelon Response to RA/ 2.a and 2.b Although the Peer Review was focused on the baseline risk, the National Fire Protection Association (NFPA) -805 submittal required both acceptable baseline risks as well as an acceptable delta risks. Knowing this requirement, the closure of the fire PRA findings focused on addressing the finding versus using a conservative argument that could mask delta risk calculations. As such, the NFPA-805 dispositions would also support TSTF-425. All of the findings in Table V-1 of the NFPA-805 LAR submittal were reviewed again to specifically assess if the finding closure could introduce conservatisms that could significantly affect the TSTF-425 delta risk calculations. There are some inherent conservatisms associated with the NFPA-805 methods, but only conservatisms beyond those are identified. This is appropriate given the NFPA-805 methods are deemed acceptable. Of the findings listed in Table V-1, there are only two findings listed with conservative closure practices. These two findings are dispositioned for TSTF-425 acceptability:

• FSS-A3-01 -bounding cable routes used -Although bounding routes were used for some conduits, bounding routes that significantly affected risk calculations were further walked down to refine the routing. Due to the limited role the remaining bounding routes play in the analysis, this will not significantly affect delta risk calculations.

• FSS-A6-01 -conservative Main Control Room (MGR) frequency development -This approach does make the control room risk more important than a traditional NUREG/CR 6850 Appendix L approach. But, this is only a frequency issue and does not mask any equipment impacts that would be evaluated in a TSTF-425 delta risk calculation. This MGR modeling just increases the calculated delta risks.

License Amendment Request Response to Request for Additional Information Docket No. 50-244 Attachment 1 Page 31of46 The remaining finding closures were largely documentation improvements or findings that were directly resolved without introducing any conservatisms that could potentially mask a TSTF-425 delta risk. Exelon Response to RAJ 2.c The NUREG/CR-7150 Vol 2 information was already included in the NFPA-805 analysis. The next revision of the fire model will incorporate NUREG/CR 6850 Appendix L, NUREG-2169, and NUREG-2178. All three of these changes affect frequency development. NUREG-2169 will cause an increase in the Main Control Board (MCB) and electrical cabinet frequencies. NUREG-2178 will reduce the heat release rates reducing the electrical cabinet high risk scenario frequencies. Appendix L will lower the MCB frequencies. These three changes in aggregate should result in a net reduction of the higher risk scenarios. As a result, the existing NFPA-805 model is conservative with regard to TSTF-425 delta risk calculations. Further as these changes are all frequency related no masking issues are introduced or removed by these updates. Exelon Response to RAJ 2.d.i Sufficient targets have been identified for the ignition sources in the unscreened Physical Analysis Units (PAUs) such that the credible range of system and function impacts has been represented. A typical range of impacts is the ignition source, the ignition source plus a set of raceways and adjacent equipment, and a full compartment burn. If the full compartment burn contribution is too large, then intermediate scenarios are developed if the key target raceways are fairly well removed from the ignition source. Exelon Response to RAJ 2.d.ii A sufficient number of fire scenarios has been developed to characterize the damage leading to collapse of the exposed structural steel for each identified scenario. All of the ignition sources in non-full-compartment-burn PAUs that have a high enough heat release rate to damage exposed structural steel are included in the evaluation. This primarily includes oil fire scenarios. Exelon Response to RAJ 2.d.iii and 2.d. v As discussed in the NFPA-805 Table V-1, the water spray system is not credited as a boundary under NFPA-805 and is not allowed per the standard. However, this water spray system is a design requirement for Ginna. The NFPA-805 analysis uses the standard NUREG/CR 6850 approach of plant partitioning PAUs. There is a concrete wall between the turbine building and the control room which is an adequate barrier per NUREG/CR 6850. The spray system was installed from a design perspective for defense-in-depth given all combustibles in the turbine building are engaged in a fire. Although purely a design issue, it was requested by the peer review team that this additional information be provided. The suppression system is credited in the multi-compartment analysis with the appropriate reliability and availability factors considered.

License Amendment Request Response to Request for Additional Information Docket No. 50-244 Exelon Response to RAJ 2.d.iv Attachment 1 Page 32 of 46 As discussed in the NFPA-805 Table V-1, all of the areas in the global analysis boundary were assessed and dispositioned as not having a significant seismic impact that is not already bounded by existing fire scenarios. Revision 2 of RG 1.200 adds clarification to IE-C12 regarding resolution of F&O IE-C10-01 [SR IE-C12 in the current version of the ASME/ANS Risk Standard], including its accompanying Note. Since this peer review finding was against RG 1.200, Revision 1, explain whether there is any change to the disposition or impact on TSTF-425 as a result of the Revision 2 update. If none, justify why not. Exelon Response to RAJ 3 The Ginna Full Power Internal Events (FPIE) PAA has detailed modeling for several initiating events, with logic built into Support System Initiating Event (SSIE) fault trees. In several instances, events such as loss of Component Cooling Water (CCW) could have been modeled with a single initiating event rate. These annualized initiating event rates are obtained from the NRC's 2012 update to NUREG/CR-6928 values, which include industry data through 2010. SR I E-C12 from the ASME PAA standard states to compare these generic industry values with the equivalent quantified gate in the PAA. RG 1.200, Revision 2 includes guidance to "COMPARE results and EXPLAIN differences in the initiating event analysis with generic data sources to provide a reasonable check of the results." During the 2015 FPIE PAA Update, the Initiating Event (IE) Notebook was updated with a comparison of the IE values provided in Section 4.4.4. In most cases, quantified Ginna SSIE values were comparable to generic data. The main differences were with electrical bus failures. Several SSIE fault trees have operator action recoveries. These recoveries, with the addition of more detailed modeling, caused the lower event frequencies in some initiating events. In some cases with electrical bus SSIE fault trees, logic could be simplified by just using an IE with the generic event rate. Beyond the electrical bus initiating events, no significant differences were found and this should have little impact on TSTF-425 analysis. The difference in the loss of bus initiating events is captured as a URE which will be reviewed for applicability for each STI change evaluation as required by Exelon procedural guidance. The current PAA model was assessed to only be Capability Category (CC) I, whereas expectations are that all SR be met at the CC II level (or justification be provided for the adequacy of Capability Category I for the specific application) regarding resolution of SC-A 12-01 [Also SR SC-A-2 in the current version of the ASME/ANS Risk Standard], which remains unresolved. The LAA takes the position that the SR is conservative and that differential risk evaluations for the TS Surveillance Frequency changes will thus also be conservative. Verify by example, or analysis, that this presumed conservatism is such that it ensures the differential risk for the application is also conservative, (i.e., the risk estimated for the before versus after License Amendment Request Response to Request for Additional Information Docket No. 50-244 Attachment 1 Page 33 of 46 condition is not overestimated such that subtracting it from the after value could underestimate the risk increase). Exelon Response to RA/ 4 For most of the end-state success criteria cases using the thermal-hydraulic analysis software, core uncovery alone was used as a surrogate for core damage. A benchmarking analysis includes cases of core uncovery as well as core damage. In those cases, the difference between core uncovery and core damage was fairly short (e.g. a few minutes). For cases that lead to core uncovery, core heat removal is clearly lost or clearly maintained except for larger loss-of-coolant accidents. In the case of Large Loss of Coolant Accidents (LOCAs), it is identified that core uncovery can initially occur, but the core can be quickly re-covered with the accumulators and residual heat removal (AHR) pumps providing makeup, mitigating a core damage event. A differential risk calculation could be impacted if 1) if a valid system recovery is not credited for re-covering the core prior to core damage, or 2) the time between core uncovery and core damage is significant where a previously un-identified system recovery could take place that is not credited in the model due to timing restraints. Although no other cases are identified where core uncovery does not lead to core damage in short order, this issue is captured as an Updating Requirements Evaluation (URE) which will be reviewed for applicability for each surveillance test interval (STI) change evaluation as required by Exelon procedural guidance. Resolution of SY-A18-01 [SR SY-A19 in the current version of the ASME/ANS Risk Standard] involves use of a systematic approach to consider maintenance unavailability, some of which may be overlapping, or not precluded by operating procedure limitations, which remains unresolved. The standard requires: In the systems model, INCLUDE out-of-service unavailability for components in the system model, unless screened, in a manner consistent with the actual practices and history of the plant for removing equipment from service ... The LAA states that the possibility of partially overlapping component unavailability has not yet been resolved, but is in all cases conservative because component unavailability combinations that would normally not be possible are being added into the Core Damage Frequency (CDF) and Large Early Release Frequency (LEAF) differential quantifications. The disposition does not determine the extent of such overlapping unavailability, but rather a-priori assumes that, if there are, modeling would be less conservative than currently failing to model. Given that risk changes, before versus after, are the subject of concern, such conservatism, if applied to the before risk, could actually generate non-conservative risk increases when a larger risk is subtracted from the after risk than a more accurate smaller risk. Provide: a. A further discussion of plant practices and the modeling of these practices relevant to overlapping simultaneous Test and Maintenance unavailability. b. A verification by example, or analysis, that this presumed conservatism is such that it ensures the differential risk for the application is also conservative (i.e., the risk License Amendment Request Response to Request for Additional Information Docket No. 50-244 Attachment 1 Page 34 of 46 estimated for the before versus after condition is not overestimated, such that subtracting it from the after value could underestimate the risk increase). Response to RAJ 5.a A review of schedules and practices indicated that when two Functional Equipment Groups (FEGs) are scheduled in the same week at Ginna, the current practice is to sequence the FEGs rather than work them simultaneously. Exceptions to this practice are very rare and are carefully discussed, risk assessed, and unavailability recorded. This minimizes the concern of shadowing of maintenance unavailability. However, since overlap of these combinations is not procedurally excluded, coincident maintenance may occur and this is allowed via random combinations of maintenance events in the PRA model. Exelon Response to RAJ 5.b As discussed in the response to RAI 5.a, the Ginna PRA model does not preclude overlapping maintenance of certain Structures Systems and Components (SSCs) even though overlapping maintenance is not typically done at Ginna. However, certain overlapping maintenance configurations that are explicitly excluded by Technical Specifications (TS) are removed from cutsets through the use of the mutually exclusive file. These typically include disallowed maintenance, such as both trains of a two-train TS system. As such, the Ginna PRA model will include some cutsets with random combinations of maintenance configurations which will tend to increase the 'base' (average maintenance) CDF. However, higher risk combinations (such as both trains of Emergency Core Cooling Systems (ECCS) or both trains of Auxiliary Feed Water (AFW)) are precluded (through the mutually exclusive file). Therefore, the combinations of maintenance events that are in the cutsets but are not routinely entered should not be significant contributors to base CDF. Additionally, test and maintenance (TM) basic event probabilities are calculated from all unavailability events, both planned and emergent. It is possible that configurations can occur in the plant due to one train being in planned maintenance and the failure of another train. The Average Test and Maintenance (ATM) model base case results are conservative due to the possibility of cutsets which contain random overlapping maintenance unavailability events. This unavoidable conservatism in the base result however has little effect on the delta risk for this given application. An example of this in the ATM model is the highest cutset with two TM terms that could potentially have a conservative effect. This cutset is shown in the table below. CDF Basic Event Event Name Description (Cutset) Value 6.53E-09 3.65E+02 TIOOOOSW TOTAL LOSS OF SERVICE WATER 1.29E-02 AFTMOTDAFW TDAFW PUMP TRAIN OUT-OF-SERVICE FOR MAINTENANCE 1.60E-02 AXHFR04084 SAFW TRAIN C FT-4084 RESTORATION ERROR AFTER CALIBRATION License Amendment Request Response to Request for Additional Information Docket No. 50-244 CDF Basic Event Event Name (Cutset) Value 1.03E-02 AXTMSAFSGB 9.41 E-01 MODE1 1.00E+OO NOSBO 8.94E-06 SWCCFPUMPR_ALL 1.00E+OO TRANSX Description Attachment 1 Page 35 of 46 SAFW TRAIN D TO SIG 0.0.S. DUE TO T/M MODE1 TAG -NO STATION BLACKOUT (SBO) CCF OF ALL COMPONENTS IN GROUP 1SWCCFPUMPR1 TAG -TRANSIENT EVENT In the case of a hypothetical STI risk evaluation, suppose that the STI change involves the service water pumps such that the value of the SWCCFPUMPR_ALL basic event in the above cutset is increased consistent with the surveillance frequency change program methodology (e.g., by a factor of 2). For each STI evaluation, the delta risk is always driven by changes to a specific set of basic event values unique to a specific surveillance test. As such, any basic events that appear in cutsets with overlapping maintenance unavailability events would correspondingly increase for the STI evaluation. This would be conservative in all cases (i.e., the risk estimated for the before versus after condition is included in both cases, such that subtracting it from the after value will not underestimate the risk increase). For our hypothetical case (factor of 2) the cutset value increases to 1.31 E-08, with a delta CDF of -6.53E-09. This is conservative, yet is far below the acceptance criteria for STI changes showing the insignificant impact of the ATM model conservatism. The standard requires that a PRA model regarding resolution of F&O IE-C13-01 [SR IE-C15 in the current version of the ASME/ANS Risk Standard] does the following: "CHARACTERIZE the uncertainty in the initiating event frequencies and PROVIDE mean values for use in the quantification of the PRA results." The sources of uncertainty which are 'considered' versus 'not considered' in estimation of mean values of any cutset element form an important input in judging the technical adequacy of a PRA model. The original peer review noted that "Section 5 [of the Initiating Event Notebook] does not provide or reference the parametric uncertainty initiating event data distribution [with a specific example cited]." The LAR treatment of this F&O expresses an opinion that while this 'documentation only' is still unresolved, this issue would not impact TSTF-425 PRA evaluations. The sources of uncertainty that were actually considered are an integral part when assessing PRA technical adequacy. Therefore: a. Characterize what types of uncertainties are actually considered in the estimation of each initiating event mean frequency in the current PRA model of record. b. Clarify if this currently unresolved F&O IE-C13-01 was subsequently re-evaluated in the 2012 Fire PRA Peer Review as a 11back-referenced11 SR item. Exelon Response to RAJ 6.a Assumptions and uncertainties are addressed in the Section 5.0 of the Ginna IE Notebook. This section addresses uncertainties such as only using industry Loss of Offsite power (LOOP) data from 1997-2013 due to deregulation, and not Bayesian updating LOCA values as they are industry expert's best estimates.

License Amendment Request Response to Request for Additional Information Docket No. 50-244 Attachment 1 Page 36 of 46 Error Factors (EFs) were added to Table 4-1 in the IE Notebook as part of the 2015 PAA update. The EFs from the generic data are used as an input into the Bayes update process and updated accordingly with the plant-specific evidence. For SSIE fault tree quantification, uncertainty was captured at the basic event level. In some cases, split fractions were applied to generic initiating event frequencies. In these cases, the Jeffreys non-informative prior alpha factor of 0.5 was used. The EFs were then estimated from the corresponding statistical distribution. Exelon Response to RA/ 6.b This F&O was addressed in the Fire Uncertainty Notebook for the fire related initiating events. This Fire Uncertainty Notebook did not address non-fire related initiating events, but as discussed in response to RAI 6.a, this F&O has been addressed as part of the 2015 internal events PAA model update. Resolution of F&O HR-G3-01 was based upon conformance with AG 1.200, Revision 1. The assessment of PAA Technical Adequacy must address conformance with AG 1.200, Revision 2. Revision 2 of AG 1.200 has added a number of specific clarifications to the ASME/ANS Risk Standard regarding SR HR-G3, which are noted below: Cat I: (a) The complexity of detection, diagnosis, decision-making and executing the required response. Cat II, and Ill: (d) Degree of clarity of the cues/indications in supporting the detection, diagnosis, and decision-making give the plant-specific and scenario-specific context of the event. (g) Complexity of detection, diagnosis and decision-making, and executing the required response. Provide a gap assessment of the current Human Reliability Analysis in the PRA model of record against the additional clarifications in RG 1.200, Revision 2 noted above. Exelon Response to RA/ 7 This F&O was addressed in the Fire Human Reliability Analysis (HRA) Notebook for the fire related human actions. This included almost all of the non-fire related HRA events as most of the non-fire related HRAs are included in the fire model as well. Consideration of cue clarity and complexity were considered as part of the 2015 internal events model update for Ginna. Any and all additions to cue clarity and complexity have been incorporated into the HRA Calculator database file for the FPIE model, and will also be incorporated in Appendix I of the 2015 Ginna FPIE HRA Notebook. As such, the 2015 internal events PRA model update is consistent with HR-G3 including the clarifications provided in AG 1.200, Revision 2.

License Amendment Request Response to Request for Additional Information Docket No. 50-244 Attachment 1 Page 37 of 46 Resolution of F&O IF-CB-01 [IFSN-A 16 in the current version of the ASME/ANS Risk Standard] involves a flooding source that was screened based on qualitative consideration of potential human action; but for that action (in response to a 2,000 gallons per minute fire service water break in IBN), there doesn't appear to be any justification for the time identified (190 min). Nothing other than time available is cited as rationale for screening the event. The LAR states that the 11impact is expected to be minimal, and is not expected to have any impact on the Surveillance Frequency Control Program." Without having corrected the PRA model of record to address the specific internal flood source issue it is not readily obvious how the conclusion of minimal impact was obtained. Therefore, provide the technical bases for assuring this omitted flood source in fact does not have any impact on the TSTF-425 based Surveillance Frequency Control Program. Exelon Response to RA/ B The FSW breaks in the Intermediate Building North (IBN) are no longer screened, since they are now being represented by the internal flood initiator FL-IBN-FSW-2K. Revision 2 of RG 1.177 provides guidance for changing TS Surveillance Frequencies. However, for allowable risk changes associated with Surveillance Frequency extensions, it refers to RG 1.17 4, Revision 2, which provides quantitative risk acceptance guidelines for changes to CDF and LERF. Revision 2 of RG 1.17 4 invokes RG 1.200, Revision 2 to address PRA Technical Adequacy. Revision 2 of RG 1.200 endorses, with clarifications, portions of the ASME/ ANS RA-Sa-2009 standard. The RITS-5b LAR is based upon TSTF-425, Revision 3 and a PRA Model, which was assessed in a Peer Review for conformance with RG 1.200, Revision 1. Conformance with the requirements of RG 1.200, Revision 2 is a requirement. Therefore: a. Provide a gap analysis to Identify any areas where the current PRA model of record does not conform to the PRA Technical Adequacy requirements of RG 1.200, Revision 2, and the ASME/ ANS RA-Sa-2009 standard. b. Clarify how the PRA applications associated with RITS-Sb will not be impacted by the gaps in the PRA model conformance with RG 1.200, Revision 2. c. Clarify that there have been no PRA model upgrades as defined in Appendix 1-A of ASME/ANS RA-Sa-2009, which would require a focused Peer Review. Specifically, discuss whether the addition of two diesel generators as an alternate source of power to the standby auxiliary feedwater pumps and a condensate storage tank as a dedicated water source for these pumps in model GN114A-W constitutes an upgrade. If so, has there been a focused-scope peer review? If not, justify. d. Confirm that the total baseline risk is consistent with the quantitative risk acceptance guidelines of RG 1.17 4, Revision 2, which provides for changes to CDF and LERF.

License Amendment Request Response to Request for Additional Information Docket No. 50-244 Exelon Response to RAJ 9.a Attachment 1 Page 38 of 46 A gap assessment was performed for the internal events PRA between RG 1.200, Revision 1 and RG 1.200, Revision 2 [3]. This gap assessment did not lead to the identification of any new "Not Mets" or changes to the original capability category ranking from the 2009 peer review. The results of the 2011 gap assessment provided the origin for the dispositions provided in Table 2-1 of the LAR which has subsequently been updated in the response to RAI 1 above. Therefore, the identification and disposition of the internal events model gaps provided in RAI 1 is consistent with the PRA Technical Adequacy requirements of RG 1.200, Rev. 2, and the ASME/ ANS RA-Sa-2009 standard. Exelon Response to RA/ 9.b Refer to updated Table 2-1 provided in response to RAI 1. Response to RA/ 9.c The addition of the two diesel generators as an alternate source of power to the SAFW pumps and a condensate storage tank as a dedicated water source in the updated PRA model utilized methods consistent with the peer-reviewed PRA model. Additionally, other changes to the PRA model were also developed consistent with the methods employed in the peer-reviewed PRA model. As such a focus-scope peer review of the internal events PRA model is not currently warranted. However, as discussed in the response to RAI 2.c, Ginna plans on transitioning to Appendix L of NUREG/CR-6850 for determining revised MCB fire frequencies. This will require a focused scope peer review. Exelon Response to RA/ 9.d As provided in Attachment W of the NFPA-805 LAR submittal [5], the RG 1.174 guidelines are met. It should be noted that the internal event CDF and LERF values in the most recent version of the PRA model are lower than that reported in the NFPA-805 LAR submittal. It is understood that those guidelines must continue to be met to allow the use of risk informed applications. RAI 10 Revision 2 of RG 1.200 defines a significant model change as follows: "Whether a change is considered significant is dependent on the context in which the insights are used. A change in the risk insights is considered significant when it has the potential to change a decision being made using the PRA.11 F&Os IF-D5a-01 (unresolved), IF-07-01, [IFEV-A6, IEFV-A8 in the current version of the ASME/ANS Risk Standard], in the current PRA model of record, involve: a. Not adequately addressing plant-specific characteristics that might affect the manner in which the frequencies of flooding are estimated (e.g., material condition, aging degradation, and water-hammer potential). b. Inappropriate screening (out) of certain internal flood scenarios without applying consistent screening criteria, as required in SRs IF-07 and IF-E3a.

License Amendment Request Response to Request for Additional Information Docket No. 50-244 Attachment 1 Page 39 of 46 If the frequencies of specific internal floods are improperly evaluated, the importance of specific flood scenarios and how they impact the unavailability of specific components will be inappropriate, and this will impact the technical adequacy of the PRA model of record. The RITS-5b LAR indicates that a sensitivity evaluation for a particular surveillance test interval evaluation will be performed to determine if there is any impact. Within the scope of TSTF-425, Revision 3, clarify: a. The specific sensitivity studies which are to be performed with the PRA model of record in order to demonstrate technical adequacy of the internal flooding frequencies without correcting the identified deficiency noted in the peer review F&O IF-D5a-01. b. The impact on the unavailability of specific components evaluated in the Surveillance Frequency Control Program of "screening in" internal flood sources which were eliminated in the current PRA model of record. Exelon Response to first RA/ 1 O.a A review of plant-specific operating experience for Ginna determined that there were no significant flooding events that have occurred in the past 16 years (since August 1998). Based on this, the generic industry failure rates developed by EPRI are acceptable for use, and a plant-specific update of these frequencies is not deemed warranted as a Bayes update with no events will have a very minimal impact on the flooding frequencies. This is documented in the latest revision of the Internal Flood Notebook (G1-PRA-012}. The material condition and aging management strategies for plant piping are addressed via the Risk Informed In-Service Inspection (RI-ISi) programs that are implemented at the site. The effects of water hammer on plant piping are inherently included as a part of the calculated rupture frequencies developed by EPRI based on industry experience. Exelon Response to first RA/ 10.b Screening criteria based on the ASME/ ANS PRA Standard has now been consistently applied to flood sources and areas. This is documented in the latest revision of the Internal Flood Notebook (G1-PRA-012}. Exelon Response to second RA/ 1 O.a For each SFCP analysis, a review will be made to see if the adjusted basic events for the components or system of interest appear in cutsets concurrent with a particular flood initiator, of notable significance, e.g., greater than a 10% contribution to the calculated change in CDF or LERF. If so, a specific sensitivity analysis will be performed related to the flood frequency to see if it could influence the acceptability of the STI change evaluation consistent with the guidance in Step 14 of NEI 04-10 [4]. The potential need for this sensitivity will be controlled via the PRA model Updating Requirement Evaluation (URE) database that is reviewed any time the PRA model is used for a documented risk application. Response to second RA/ 1 O.b The most recent PRA model update carefully considered the criteria in the ASME/ ANS PRA Standard with regard to being able to screen flood areas and water sources. Based on the License Amendment Request Response to Request for Additional Information Docket No. 50-244 Attachment 1 Page 40 of 46 current PRA update, there are no longer any water sources that were inappropriately screened, and no scenarios were numerically screened using Supporting Requirement IFQU-A3 of the PRA Standard. Because of this, there is no risk of "screening in" any new internal flood scenarios. RAI 11 Similar to F&O IE-C13-01 dealing with internal events, Internal Flooding F&O IF-F3-01 [IFQU-83 in the current version of the ASME/ANS Risk Standard], and which is still unresolved, identified deficiencies in the consideration of uncertainties and that the treatment "did not constitute an adequate characterization of the sources of uncertainty associated with the flood analysis or a comprehensive discussion of the assumptions that could have an effect on the results." The LAR treats this F&O as a "documentation only F&O" which will not impact evaluation of specific components in the Surveillance Frequency Control Program. Without knowing what sources of uncertainty were actually considered, and how such uncertainties propagate to the end results, it was not possible for the original peer review to assess the required technical adequacy. Therefore: a. Characterize what types of uncertainties are actually considered in the estimation of each initiating event mean frequency in the current PRA model of record. b. Clarify if this currently unresolved F&O I F-F3-01 was subsequently re-evaluated in the 2012 Fire PRA Peer Review as a "back-referenced" SR item. Exelon Response to RAJ 11.a In estimating the event mean frequency for each internal flood initiator, the initiating event uncertainty parameters from the EPRI 1013141 data were used and error factors reported in the Internal Flood Notebook (G1-PRA-012). These parametric uncertainty values propagate to the end results using the CAFT A PRA software. Modeling uncertainty for the internal flood portion of the PRA was also addressed and documented in G1-PRA-012 using the guidance found in EPRI 1016737. The finding for IF-F3-01 is considered to be resolved. Exelon Response to RAJ 11.b Per the response to RAI 11.a, IF-F3-01 is now considered to be resolved. RAI 12 F&O I F-E5-01 [I FQU-A5 in the current version of the ASME/ ANS Risk Standard], involved use of HRA methods, which were not consistent with the methods used elsewhere in the PRA model. The LAR indicates the issue has been resolved. The ASME/ANS Risk Assessment Standard, Non-Mandatory Appendix 1-A, would require a focused peer review if there was an underlying PRA model upgrade (e.g., application of new methods which were different than those in the original model) but not for PRA model maintenance, where PRA model maintenance is specifically defined: "plant modifications, procedure changes, plant performance (data)." Confirm that the revised HRA performed for the internal flooding portion of the PRA model of record uses HRA methods that are consistent with other portions of the PRA License Amendment Request Attachment 1 Page 41 of 46 Response to Request for Additional Information Docket No. 50-244 that have been peer reviewed. If not, confirm whether a 11focused Peer Review had been performed11 for the internal flooding HRA consistent with the requirements of ASME/ANS RA-Sa-2009, Appendix 1-A. Exelon Response to RA/ 12 As discussed under PRA RAI 7 related to F&O HR-G3-01 an improved HRA method was implemented as part of the fire analysis. This method was peer reviewed as part of the fire evaluation. The internal flooding HRA document was reviewed to ensure consistency between the HRA methodology applied in the analyses of both internal flooding and FPIE operator actions. The internal flooding analyses were determined to be consistent with the methodology applied throughout the FPIE HRA. Additionally, the internal flood HFE analyses were included into the HRA Calculator database to ensure consistency in future updates of the FPIE HRA. The results of these analyses are included in the 2015 Ginna FPIE HRA Notebook. RAI 13 The LAR states in Section 2.0.5 of Attachment 2: The results of the standby failure rate sensitivity study plus the results of any additional sensitivity studies identified during the performance of the reviews as outlined in 2.2.1 and 2.2.3 above for each STI change assessment will be documented and included in the results of the risk analysis that goes to the IDP. The LAR does not contain any Section 2.2.1 or 2.2.3. Correct the LAR to address the missing Sections 2.2.1 and 2.2.3. Exelon Response to RAI 13 The LAR does not contain any missing sections. The paragraph quoted above was submitted with a typographical error. Specifically, the reference to sections 2.2.1 and 2.2.3 should have read 2.0.2 and 2.0.4, respectively. RAI 14 The LAR states in Section 2.0.4 of Attachment 2 with regard to the most recent PRA model GN114A-W and peer reviews conducted for the internal events model in 2009 and fire PRA model in 2012: All remaining gaps will be reviewed for consideration during the 2015 model update but are judged to have low impact on the PRA model or its ability to support a full range of PRA applications. The remaining gaps are documented in the URE database so that they can be tracked and their potential impacts accounted for in applications where appropriate.

License Amendment Request Response to Request for Additional Information Docket No. 50-244 Attachment 1 Page 42 of 46 Confirm that any gap assessment and, if identified as required due to model upgrades, scope or full-scope peer review will be performed in accordance with the then latest version of the ASME/ANS PRA Standard as endorsed, clarified and qualified by the then latest revision (currently Revision 2) of RG 1.200. Exelon Response to RAJ 14 The current status of the gaps to RG 1.200, Revision 2 based on the most recent internal events PRA model update are provided in response to RAI 1. As noted in response to RAI 9, other changes to the PRA model were also developed consistent with the methods employed in the peer-reviewed PRA model. As such, a focus-scope peer review of the internal events PRA model is not currently warranted. RAI 15 F&O LE-C2a-01 addressed the need for realistic treatment of feasible operation actions after core damage, noting it is conservative not to credit these. The cited impact to TSTF-425 stated that there are limited operator actions that could influence LERF, such that their effect is unlikely to be significant, possibly even lowering LERF estimates. Therefore, the omission of these actions is conservative and does not adversely impact the PRA model used for TSTF-425 analysis. Conservatism in the before versus after risk when performing a risk increase calculation does not guarantee a conservative estimate of the risk increase, since a more realistic estimate of the before risk, being lower, would lead to a more conservative estimate of the risk increase when before is subtracted from after. Either demonstrate essentially no effect on the before risk by excluding credit for these actions or reassess the before risk, and therefore the risk increase, after incorporating credit for these actions. Exelon Response to RA/ 15 Two human actions are identified in the Level 2 analysis that may be credited in the LERF PRA model for human action post-core damage, but prior to vessel breach: 1) late recovery of offsite power in station blackout scenarios where core damage is arrested prior to vessel breach and 2) late depressurization of the reactor coolant system. Late recovery of offsite power is explicitly modeled in the LERF PRA. In the Ginna Level 2 Analysis, the probability of an early Containment failure is dependent on the loads on the Containment at vessel breach. One factor that can affect Containment loads is Reactor Coolant System (RCS) pressure at vessel breach. RCS depressurization prior to core damage is credited in the PRA LERF model. Given that the RCS is not depressurized early, a late depressurization action is feasible. However, the system responses that would be measured by a STI change are already credited in the early depressurization action. Failure of those systems early would fail the late action and would not non-conservatively impact the delta risk calculation. In addition, in the LERF accident progression, the late RCS depressurization action would only impacts containment failure probabilities.

License Amendment Request Response to Request for Additional Information Docket No. 50-244 Attachment 1 Page 43 of 46 A URE is open to develop and implement a human error probability for the late depressurization action. As with all TSTF-425 related assessments, the delta risk results with be reviewed to ensure that no conservatisms are significantly masking the delta risk evaluation. RAI 16 F&O LE-C9a-01 addressed survivability credit for equipment or human actions that could be impacted by containment failure, stating that it did not appear such credit was taken, leaving this SR as CC I, acknowledged as not applicable in the disposition and impact on TSTF-425. If crediting equipment survivability in the before versus after risk condition would lead to a more conservative estimate of the risk increase, then it may not only be non-conservative to have ignored this, but also may fail to meet even CC-I for the application where it is the risk increase that is the key, not the base risk. Further, N/A may not be an appropriate disposition. Address this F&O in light of the potential effect on risk increase, not only base risk, with regard to TSTF-425. Exelon Response to RAJ 16 In the Ginna Level 2 Analysis, early containment failure after core-damage and vessel breach is the end-state for the LERF accident progression. There are no equipment dependencies or human actions that are identified that could be reasonably credited to prevent a release through a failed containment. There are no credited equipment, systems, or human actions that would be impacted by the adverse environment impacted by containment failure. Therefore, this issue would not impact delta-risk calculations. A URE is open to capture that this F&O will remain unresolved and the SR will remain Category I. RAI 17 F&O LE-C10-01 addressed realistic containment bypass analysis, including justification for any scrubbing credit, stating that no such credit was taken, although there was a sensitivity analysis determining any impact would be negligible. As a result, no impact on TSTF-425 was cited. Verify that the impact of not considering scrubbing is negligible with respect to the risk increase from the before vs. after risk calculation, not just negligible with respect to the base risk. Exelon Response to RAJ 17 In the Ginna Level 2 analysis, no credit is given for scrubbing of release paths. However, the Ginna Level 2 analysis identifies that scrubbing may be applicable to the following three containment bypass conditions: 1) a steam generator tube rupture event with feedwater available, or 2) internal flood scenarios with an interfacing system LOCA and the affected auxiliary building room flooded, or 3) sequences where the interfacing system LOCA break is in the RH R pits, thus resulting in the break potentially being submerged under a substantial water level.

License Amendment Request Response to Request for Additional Information Docket No. 50-244 Attachment 1 Page 44 of 46 Of the three above conditions, the steam generator tube rupture event with feedwater available is identified as a candidate for potential impact for SI evaluation delta risk calculations, as the feedwater system is under the surveillance program. The delta risk impact is expected to be minimal as many, but not all, SGTR LERF accident sequences involve a loss of auxiliary feedwater. There may be cases where the status of feedwater is not assessed in the accident cutsets, and the impact of feedwater may be masked by the delta risk calculations. These cases should be considered in the SI analysis. A URE is open to credit scrubbing of release paths for the LERF analysis. As with all TSTF-425 related assessments, the delta risk results with be reviewed to ensure that no conservatisms are significantly masking the delta risk evaluation. Division of Safety Systems!Technical Specification Branch 1. As required by section 50.36 of Title 10 of the Code of Federal Regulations (1 O CFR 50.36), "Technical Specifications," the licensee must provide a summary statement of the bases or reasons for such specifications as part of the LAR submittal. Although the NRC staff does not approve TS bases changes, this information is utilized by the staff during the review of the LAR. The following issues associated with the TS bases were identified during the LAR review: a. The licensee provided proposed revisions to the TS bases pages in Attachment 4 of the initial submittal on June 4, 2015. During the NRC staff's review, it was noted that several references cited throughout the bases pages were being deleted due to revisions associated with the adoption of TSTF-425, but it appeared that the deleted references were also cited in other parts of the TS bases; therefore, the deletions would be incorrect. The pages with deleted references that are in question from Attachment 4 include: B 3.3.1-47, B 3.4.12-13, 8 3.4.13-6, and 8 3.4.14-7. Please verify the deletion of these references is accurate. Exelon Response RAI 1 a Exelon has reviewed the affected pages with the deleted references and has determined that all of the references need to be retained, with the exception of Reference 10 on page B 3.4.14-7. Reference 10 is not mentioned beyond the text identified for deletion; therefore deleting Reference 10 is appropriate. Attachment 2 contains the revised TS Bases pages. b. On TS bases page B 3.1.6-6 of the initial licensee submittal, the description for SR 3.1.6.3 states, 11A reduction of the Frequency to every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> .... 11 Since the LAR is proposing to transfer the periodic frequency for this SR to the Surveillance Frequency Control Program (SFCP), please explain why there is a 114 hour11 reference in the bases description for this SR. Exelon response to RAI 1 b Ginna TS SR 3.1.6.3 is unique to Ginna. The TSTF 425 does not have an equivalent SR for when the rod insertion limit monitor is inoperable. The current License Amendment Request Response to Request for Additional Information Docket No. 50-244 Attachment 1 Page 45 of 46 stated frequency in the Ginna TS is "Once within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> thereafter." The first 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />) is an event driven time and therefore outside the scope of TSTF 425. The phrase "every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> thereafter" is a standard surveillance frequency interval, and only this time is moved to the SFCP. However, the NRC Staff is correct and the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> reference in the TS Bases is a standard surveillance frequency interval and should have been removed. Attachment 2 contains the revised TS Bases page. c. On TS bases page B 3.4.2-3 of the initial licensee submittal, there are references to the 1130 minute11 SR frequency for SR 3.4.2.2. Since the LAR is proposing to transfer the periodic frequency for this SR to the SFCP, please explain why there are 1130 minute11 references in the bases description for this SR. Exelon response to RAI 1 c Ginna TS SR 3.4.2.2 is unique to Ginna. The TSTF 425 does not have an equivalent SR for when the Taveg alarm is inoperable or not reset. The current stated frequency in the Ginna TS is "Once within 30 minutes and every 30 minutes thereafter." The first 30 minutes (within 30 minutes) is an event driven time and therefore outside the scope of TSTF 425. The phrase "every 30 minutes thereafter is a standard surveillance frequency interval, and only this time is moved to the SFCP. However, the NRC Staff is correct and the 30 minute reference in the TS Bases is a standard surveillance frequency interval and should have been removed. Attachment 2 contains the revised TS Bases page. d. In the LAR supplement submitted by the licensee on October 2, 2015, TS bases information associated with the adoption of TSTF-425 was provided in Attachment-5. On page 4 of this attachment, the new proposed description for SR 3.5.2.8 was provided. The first paragraph of this description is not written in a coherent manner. Please correct the language. Exelon response to RAI 1 d The first paragraph on page 4 of Attachment 5 was not properly transferred from the description provided in TSTF 523. The first paragraph should have read: " ECCS piping and components have the potential to develop voids and pockets of entrained gases. Preventing and managing gas intrusion and accumulation is necessary for proper operation of the ECCS and may also prevent water hammer, pump cavitation, and pumping of non-condensible gas into the reactor vessel." This paragraph, as stated above, replaces and supersedes the first paragraph on page 4 of Attachment 5. Attachment 2 contains the revised page.

License Amendment Request Response to Request for Additional Information Docket No. 50-244 References Attachment 1 Page 46 of 46 [1] Letter from Diane Render, U.S. Nuclear Regulatory Commission to Mr. Bryan C. Hanson, Exelon, R.E. Ginna Nuclear Power Plant-Request for Additional Information Regarding: Risk-Informed Technical Specifications Initiative 58 (GAG No. MF6358), January 7, 2016. [2] Exelon Generation Company, LLC, Application for Technical Specifications Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3), June 4, 2015, ADAMS Accession Number ML 15166A075. [3] SAIC, R.E. Ginna Probabilistic Risk Assessment (PRA) Gap Assessment Work Plan, Revision 0, September 2011. [4] Nuclear Energy Institute, Risk-Informed Technical Specifications Initiative Sb, Informed Method for Control of Surveillance Frequencies, Industry Guidance Document, N El 04-10, Revision 1, April 2007. [5] Letter from Mr. Joseph E. Pacher (Ginna LLC) to Document Control Desk (NRC), License Amendment Request Pursuant to 10 CFR 50.90: Adoption of NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, March 28, 2013, ADAMS Accession Number ML 13093A064.

AITACHMENT2 License Amendment Request R. E. Ginna Nuclear Power Plant Docket No. 50-244 Application for Technical Specification Change Regarding Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3) Revised Technical Specifications Bases Pages REFERENCES RTS Instrumentation B 3.3.1 1. Atomic Industry Forum (AIF) GDC 14, Issued for comment July 10, 1967. 2. 10 CFR 50.67. 3. American National Standard, "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants," N18.2-1973. 4. UFSAR, Chapter 7. 5. UFSAR, Chapter 6. 6. UFSAR, Chapter 15. 7. IEEE-279-1971. 8. EP-3-S-0505, "Instrument Setpoint/Loop Accuracy Calculation Methodology". 9. WCAP-10271-P-A, Supplement 2, Rev. 1, June 1990. R.E. Ginna Nuclear Power Plant B 3.3.1-47 Revision 61 LTOP System 83.4.12 2. Generic Letter 88-11, "NRC Position on Embrittlement of Reactor Vessel Materials and its Impact on Plant Operations." 3. UFSAR, Section 5.2.2. 4. 10 CFR 50, Section 50.46. 5. 10 CFR 50, Appendix K. 6. Letter from D. L. Ziemann, NRC, to L. D. White, RG&E, Subject: "Issuance of Amendment No. 28 to Provisional Operating License No. DPR-18," dated July 26, 1979. 7. Generic Letter 90-06, "Resolution of Generic Issue 70, Operated Relief Valve and Block Valve Reliability," and Generic Issue 94, "Additional Low-Temperature Overpressure Protection for Light-Water Reactors." R.E. Ginna Nuclear Power Plant B 3.4.12-13 Revision 52 RCS Operational LEAKAGE 83.4.13 sufficient time to collect and process all necessary data after stable plant conditions are established. Steady state operation is required to perform a proper inventory balance; calculations during maneuvering are not useful. For RCS operational LEAKAGE determination by water inventory balance, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and volume control tank levels, makeup and letdown, and RCP seal injection and return flows. An early warning of pressure boundary LEAKAGE or unidentified LEAKAGE is provided by the automatic systems that monitor the containment atmosphere radioactivity and the containment sump level. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. Leakage detection systems are specified in LCO 3.4.15, "RCS Leakage Detection Instrumentation." Note 2 states that this SR is not applicable to primary to secondary LEAKAGE because LEAKAGE of 150 gallons per day cannot be measured accurately by an RCS water inventory balance. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Frequency is a reasonable interval to trend LEAKAGE and recognizes the importance of early leal<age detection in the prmrention of !INSERT 3 j This SR verifies that primary to secondary LEAKAGE is less or equal to 150 gallons per day through any one SG. Satisfying the primary to secondary LEAKAGE limit ensures that the operational LEAKAGE performance criterion in the Steam Generator Program is met. If this SR is not met, compliance with LCO 3.4.17, "Steam Generator Tube Integrity," should be evaluated. The 150 gallons per day limit is measured at room temperature as described in Reference 5. The operational LEAKAGE rate limit applies to LEAKAGE through any one SG. If it is not practical to assign the LEAKAGE to an individual SG, all the primary to secondary LEAKAGE should be conservatively assumed to be from one SG. The Surveillance is modified by a Note which states that the Surveillance is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation. For RCS primary to secondary LEAKAGE determination, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows. The Surveillance Frequency of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is a reasonable interval to trend primary to secondary LEAKAGE and recognizes the importance of early 1eakage detection in the prevention of aeci6effis.:. The primary to secondary LEAKAGE is determined using continuous process radiation R.E. Ginna Nuclear Power Plant B 3.4.13-5 Revision 52 REFERENCES RCS Operational LEAKAGE B 3.4.13 monitors or radiochemical grab sampling in accordance with the EPRI guidelines (Reference 5). If\ 1. Atomic Industry Forum (AIF) GDC 16, Issued for comment July 10, 1967. 2. Generic Letter 84-04, "Safety Evaluation of Westinghouse Topical Reports Dealing with Eliminaion of Postulated Pipe Breaks in PWR Primary Main Loops." 3. UFSAR, Chapter 15. 4. NEI 97-06, Steam Generator Program Guidelines 5. EPRI, Pressurized Water Reactor Primary-to-Secondary Leak Guidelines R.E. Ginna Nuclear Power Plant B 3.4.13-6 Revision 52 REFERENCES 1. 10 CFR 50.2. 2. 10 CFR 50.55a(c). RCS PIV Leakage B 3.4.14 3. Atomic Industry Forum (AIF) GDC 53, Issued for comment July 10, 1967. 4. WASH-1400 (NUREG-75/014), "An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants," Appendix V, October 1975. 5. NUREG-0677, "The Probability of lntersystem LOCA: Impact Due to Leak Testing and Operational Changes," May 1980. 6. Generic Letter, "LWR Primary Coolant System Pressure Isolation Valves," dated February 23, 1980. 7. Letter from D. M. Crutchfield, NRC, to J. E. Maier, RG&E, Subject: "Order for Modification of License Concerning Primary Coolant System Pressure Isolation Valves," and associa1ed SER on Primary Coolant System Pressure Isolation Valves (WASH-1400, Event V), dated April 20, 1981. (ML010542030) 8. EG&G Report, EGG-NTAP-6175. 9. ASME Code for Operation and Maintenance of Nuclear Power 10. 10 GFR 50.55a(f). 11. Letter from D. M. Crutchfield, NRC, to J.E. Maier, RGE, Subject: "TMl-2 Category "A" Items" and associated SER for Amendment No. 42 to Provisional Operating License No. DPR-18, dated May 11, 1981. (ML010540356) R.E. Ginna Nuclear Power Plant B 3.4.14-7 Revision 58 SURVEILLANCE REQUIREMENTS SR 3.1.6.1 Control Bank Insertion Limits 8 3.1.6 This Surveillance is required to ensure that the reactor does not achieve criticality with the control banks below their insertion limits. The Frequency of within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to achieving criticality ensures that the estimated control bank position is within the limits specified in the COLR shortly before criticality is reached. SR 3.1.6.2 With an OPERABLE bani< insertion limit monitor (i.e., the control board annunciators, verification of the control bani< insertion limits at a Frequeney of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is sufficient to ensure OPERABILITY of the bank insertion limit monitor and to detect control banl<s that may be ap13roaching the insertion limits since, normally, 'iePJ little rod motion occurs in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. If' SR3.1.6.3 When the insertion limit monitor (i.e., the control board annunciators becomes inoperable, no control room alarm is available between the normal 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> frequet=tey to alert the operators of a control bank not within the insertion limits. A reduction of tf:le Frequency to every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> pro1t*ides suffieient monitoring of control rod insertion when the monitor is inoperable. Verification of the control bank position at a Frequency flet:tfS is sufficient to detect control banks that may be approaching the insertion limits. A\ INSERT1 This SR is modified by a Note that states that performance of this SR in only necessary when the rod insertion limit monitor is inoperable. SR 3.1.6.4 When control banks are maintained within their insertion limits as required by SR 3.1.6.2 and SR 3.1.6.3 above, it is unlikely that their sequence and overlap will not be in accordance with requirements provided in the COLR. A Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is consistent Viith the iAseFlieA liA'til eAeek aBeve ill SR 3.1.6.2. R.E. Ginna Nuclear Power Plant 8 3.1.6-6 Revision 60 ACTIONS SURVEILLANCE REQUIREMENTS REFERENCES RCS Minimum Temperature for Criticality 8 3.4.2 If the parameters that are outside the limit cannot be restored, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 2 with Keff < 1.0 within 30 minutes. Rapid reactor shutdown can be readily and practically achieved within a 30 minute period due to the proximity to MODE 2 conditions. The allowed time is reasonable, based on operating experience, to reach MODE 2 with Keff < 1.0 in an orderly manner and without challenging plant systems. SR 3.4.2.1 This SR verifies that RCS T avg in each loop is 540°F within 30 minutes prior to achieving criticality. This ensures that the minimum temperature for criticality is being maintained just before criticality is reached. The 30 minute time period is long enough to allow the operator to adjust temperatures or delay criticality so the LCO will not be violated, thereby providing assurance that the safety analyses are not violated. SR 3.4.2.2 RCS loof3 a*,erage is required to be verified at or above 540°F e*o*ery 30 mitlutes in MODE 1, aRd in MODE 2 *witM keff 1.0. TMe 30 minute frequency is sufficient basee Otl tMe low of large tempef!ltt1re swiflgs withet1t the epefl!ltefS itflewledge. This SR is modified by a Note that only requires the SR to be performed if any RCS loop T avg is < 54 7°F and the low T avg alarm is either inoperable or not reset. The T avg alarm provides operator indication of low RCS temperature without requiring independent verification while a T avg > 547°F in both RCS loops is within the accident analysis assumptions. If the T avg alarm is to be used for this SR, it should be calibrated consistent with industry standards. This surveillance is replaced by SR 3.1.8.2 during PHYSICS TESTING. 1. None. R.E. Ginna Nuclear Power Plant 8 3.4.2-3 Revision 21 Supplement to License Amendment Request Adoption of TSTF-425, Rev. 3 October2,2015 Docket No. 50-244 . . INSERTD SR 3.5.2.8 o develop voids and pockets of entrained gases. Preventing and managing gas intrusion and accumulation is necessary for proper operation of the ECCS and may also prevent water hammer, pump cavitation, and pumping of non-condensible gas into the reactor vessel. Attachment 5 Page 4 of 7 EGGS pipiAg and componeRts have H'le VtlitM tMe exeeptioA of tMe operatiflg eentrifugal eha1 ging pump, the EGGS pumps are nermall:r in a standby, non operating mode. As such, flow path piping has tMe potential to de*velop 1q*oids and pockets of entraiMed gases. Preventing and rnaAaging gas intrusion and accumulation is necessary for MaiRtaining the piping from the EGGS pumps to the ACS full of *water pro15er operation of the EGGS and may also ensures that the oi'ill perform 15roperl:yi, if1jeeting its full capacity iRto the ACS upon demand. This will also pre*tent water hammer, pump ea-vitatiof\, and pumping of nof1eondensible gas (e.g., air, nitrogeFI, or t'lydroger=i) iRto the reactor vessel fellovo*iAg an SI sigRBI or during shutdouvn cooling. Selection of ECCS locations susceptible to gas accumulation is based on a review of system design information, including piping and instrumentation drawings, isometric drawings, plan and elevation drawings, and calculations. The design review is supplemented by system walk downs to validate the system high points and to confirm the location and orientation of important components that can become sources of gas or could otherwise cause gas to be trapped or difficult to remove during system maintenance or restoration. Susceptible locations depend on plant and system configuration, such as stand-by versus operating conditions. The ECCS is OPERABLE when it is sufficiently filled with water. Acceptance criteria are established for the volume of accumulated gas at susceptible locations. If accumulated gas is discovered that exceeds the acceptance criteria for the susceptible location (or the volume of accumulated gas at one or more susceptible locations exceeds an acceptance criteria for gas volume at the suction or discharge of a pump), the Surveillance is not met. If it is determined by subsequent evaluation that the ECCS is not rendered inoperable by the accumulated gas (i.e., the system is sufficiently filled with water), the Surveillance may be declared met. Accumulated gas should be eliminated or brought within the acceptance criteria limits. ECCS locations susceptible to gas acrumulation are monitored and, if gas is found, the gas volume is compared to the acceptance criteria for the location. Susceptible locations in the same system flow path which are subject to the same gas intrusion mechanisms may be verified by monitoring a representative sub-set of susceptible locations. Monitoring may not be practical for locations that are inaccessble due to radiological or environmental conditions, the plant configuration, or personnel safety. For these locations alternative methods (e.g., operating parameters, remote monitorhg) may be used to monitor the susceptible location. Monitoring is not required for susceptiije locations where the maximum potential accumulated gas void volume has been evaluated and determined to not challenge system OPERABILITY. The accuracy of the method used for rronitoring the susceptible locations and trending of the results should be sufficient to assure system OPERABILITY during the Surveillance interval. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The Surveillance Frequency may vary by location susceptible to gas accumulation.

ATTACHMENT 3 License Amendment Request R. E. Ginna Nuclear Power Plant Docket No. 50-244 Application for Technical Specification Change Regarding Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3) Revised INSERT 2 License Amendment Request Attachment 3 Page 1 of 1 Response to Request for Additional Information Docket No. 50-244 INSERT2 5.5.17 Surveillance Frequency Control program This program provides controls for the Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met. a. The Surveillance Frequency Control Program shall contain a list of Frequencies of the Surveillance Requirements for which the Frequency is controlled by the program. b. Changes to the Frequencies listed in the Surveillance Frequency Controlled Program shall be made in accordance with NEI 04-10, Informed Method for Control of Surveillance Frequency," Revision 1. c. The provisions of Surveillance Requirement 3.0.2 and 3.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.