ML16034A139

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..

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 Kennett Square. PA 19348 www exeloncorp.com Exelon Generation 10 CFR 50.90 February 3, 2016 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 A. E. Ginna Nuclear Power Plant Renewed Facility Operating License No. DPR-18 NRC Docket No. 50-244

SUBJECT:

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 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. 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,

,JL-.')r~ J~

David T. Gudger Manager - Licensing & Regulatory Affairs Exelon Generation Company, LLC Attachments: 1. Response to Request for Additional Information

2. Revised Technical Specifications Bases Pages

3. Revised INSERT 2 cc: USNRC Region I Regional Administrator w/attachments USNRC Senior Resident Inspector - Ginna USNRC Project Manager, NRR - Ginna A. L. Peterson, NYSERDA

ATTACHMENT 1 License Amendment Request R. E. Ginna Nuclear Power Plant Docket No. 50-244 Application for Technical Specification Change Regarding Risk-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 Attachment 1 Response to Request for Additional Information Page 1 of 46 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 In a letter dated June 4, 2015 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML15166A075), 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 Attachment 1 Response to Request for Additional Information Page 2 of 46 Docket No. 50-244 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 Attachment 1 Response to Request for Additional Information Page 3 of 46 Docket No. 50-244 Updated Table 2-1 Internal Events PAA Peer Review - Findings SR Topic Status Finding/Observation Disposition Impact to TSTF-425 IE-C12 COMPARE results and EXPLAIN Open F&O IE-Cl0-01: The Ginna Initiating Event During the 2015 model update, the The difference in the loss of bus

[2005: differences in the initiating event Notebook {Gl-IE-0001, Rev. 1} Section 4.3 Initiating Event notebook is updated initiating events is captured in IE-ClO] analysis with generic data sources to provides a cross-reference between the with a comparison of the frequencies of URE 1202, which will be reviewed the plant-specific initiating events and URE 845 provide a reasonableness check of Ginna Initiating Events and the "NRC Rates for applicability for each ST/

generic initiating events.

the results. of Initiating Events" in table 4-7. Table 4-7 change evaluation as required by cross-reference includes columns for In most cases, the plant-specific IE Exelon procedural guidance.

NUREG/CR-5750 Category and NP-2230 frequencies are comparable to generic EPRl/NUREG/CR-3862 PWR Category. frequencies. However, in some cases, electrical bus failures were lower for Table 4-8 provides a cross-reference the site-specific modeling, due to between Ginna and similar PWR plants crediting operator actions for

{Point Beach, Prairie Island, and Kewaunee). recoveries prior to trip.

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).

License Amendment Request Attachment 1 Response to Request for Additional Information Page 4 of 46 Docket No. 50-244 Updated Table 2-1 Internal Events PRA Peer Review - Findings SR Topic Status Finding/Observation Disposition Impact to TSTF-425 IE-ClS CHARACTERIZE the uncertainty in the Complete F&O IE-C13-01: Gl-IE-0001, PRA INITIATING Assumptions and Uncertainties have This F&O has been addressed

[2005: initiating event frequencies and EVENT (IE) NOTEBOOK, Section 5 documents been added to the Ginna Initiating with the current PRA model and IE-C13] PROVIDE mean values for use in the assumptions and sources of uncertainty. Event notebook. Where generic data documentation, and does not quantification of the PRA results. However, section 5 does not provide or was used, the error factors from the impact the TSTF-425 analysis.

reference the parametric uncertainty generic data are used as input to the initiating event data distribution. For Bayesian update process and updated example, the distribution for TIGRLOSP is accordingly with plant-specific identified in the CAFTA model, evidence.

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.

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

collapsed liquid level less than 1?3 core height or code-predicted peak core Although no other cases are temperature >2,SOO"F (BWR) (b) identified where core uncovery collapsed liquid level below top of active fuel for a prolonged period, or code-pre-does not lead to core damage in dicted core peak node temperature short order, this issue is captured

>2,200"F using a code with detailed core as URE 0838 which will be modeling; or code-predicted core peak reviewed for applicability for node temperature >1,800"F using a code each ST/ change evaluation as with simplified (e.g., single-node core required by Exelon procedural model, lumped para-meter) core guidance.

modeling; or code-predicted core exit temperature >l,200"F for 30 min using a code with simplified core modeling (PWR).

License Amendment Request Attachment 1 Response to Request for Additional Information Page 6 of 46 Docket No. 50-244 Updated Table 2-1 Internal Events PRA Peer Review - Findings SR Topic Status Finding/Observation Disposition Impact to TSTF-425 SC-A4 IDENTIFY mitigating systems that are Complete F&O SC-A4 Operator action Add RCHFDXlBAF to the Event Tree No impact to TSTF 425. Action shared between units, and the RCHFDXlBAF (operator fails to align BAF TIU, as appropriate. placed in Event Tree TIU logic manner in which the sharing is given 1 of 2 PORVs and no charging) is not and Finding addressed.

included in the fault tree model. It appears performed should both units that this event should be added in Event experience a common initiating Tree TIU Sequence 5 Failures under gate event (e.g., LOOP). TL_FB.

This is an omission in the model and may affect CDF and LERF.

License Amendment Request Attachment 1 Response to Request for Additional Information Page 7 of 46 Docket No. 50-244 Updated Table 2-1 Internal Events PAA Peer Review - Findings SR Topic Status Finding/Observation Disposition Impact to TSTF-425 SY-AlO INCORPORATE the effect of variable Complete SY-All Gate TL_FBHRDl input to gate Review the Bleed and Feed modeling to No impact as the Finding has success criteria (i.e., success criteria that TL_FB for failure of Bleed and Feed models ensure the system failures been addressed and the logic has

[SY-All change as a function of plant status) into success as requiring 1 SI pump and 1 PORV.

-2005] the system modeling. Example causes of appropriately reflect the success been updated and documented The logic does not include 75 gpm charging variable system success criteria are criteria. in the Success Criteria Notebook.

flow which is noted in the Success Criteria (a) different accident scenarios. Different success criteria are required for some notebook as required to support single systems to mitigate different accident PORV success. This was confirmed through scenarios (e.g., the number of pumps discussion with Ginna PRA personnel.

required to operate in some systems is The omission of a needed mitigating system dependent upon the modeled initiating for support of the Bleed and Feed function event). may underestimate the importance of these (b) dependence on other components.

sequences for applications.

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).

License Amendment Request Attachment 1 Response to Request for Additional Information Page 8 of 46 Docket No. 50-244 Updated Table 2-1 Internal Events PRA Peer Review - Findings SR Topic Status Finding/Observation Disposition Impact to TSTF-425 SY-A14 When identifying the failures in SY-All Complete SY-A13 Inconsistencies existed in the Review the need to add the No impact to TSTF 425. The INCLUDE consideration of all failure system modeling of the city water system. unavailability event in the SAFW dependencies for SAFW have

[SY-A13 modes, consistent 2005] with available data and model level of Where used to support the GE-Betz system, System. been updated in the Ginna PRA.

detail, except where excluded using the a basic event for unavailability of city water criteria in due to grid LOOP was added (basic event SY-AlS.

CDAACITYWATER). This same event was not For example, (a) active component fails to start added to the city water modeling for (b) active component fails to continue to support of the SAFW system.

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

License Amendment Request Attachment 1 Response to Request for Additional Information Page 9 of 46 Docket No. 50-244 Updated Table 2-1 Internal Events PRA Peer Review - Findings SR Topic Status Finding/Observation Disposition Impact to TSTF-425 SY-A19 In the systems model, INCLUDE out-of- Complete SY-A18-0l - Ginna PRA System Notebooks The Ginna maintenance scheduling This issue has been addressed service unavailability for components in provides a list of all the modeled T&M terms in

[SY-A18 practices are, when two functional with the current PRA model and the system model, unless screened, in a Section 3.4.C. Section 2.9 of the notebooks 2005] manner consistent with the actual provide discussion of procedures and testing that equipment groups are scheduled to be documentation, and has no practices and history of the result in Unavailability. The review of these out-of-service in the same week, that impact on TSTF-425 analysis.

plant for removing equipment from sections found no instances of simultaneous the FEGs are sequenced rather than service. unavailability that can result from planned activities. However, the PRA engineer noted in a working them simultaneously.

(a) INCLUDE Exceptions are rare and are risk-discussion that some systems are shadowed in (1) unavailability caused by testing when a planned maintenance. There is not a specific assessed.

component or system train is discussion on plant maintenance practices within reconfigured from its required accident the (a)(4} program that would result in planned mitigating position such that the unavailability of multiple systems OOS (i.e., EDG The Ginna PRA model will include some component cannot function as required outages combined with AFW motor driven pump random combinations of maintenance (2) maintenance events at the train level outages to lower total risk as opposed to configurations. Certain overlapping when procedures require isolating the performing the work independently), or of planned activities resulting in multiple configurations are explicitly excluded entire train components OOS that do not violate technical from the model, such as taking out of for maintenance specifications (e.g., two AFW pumps in service both trains of a two-train Tech.

(3) maintenance events at a sub-train maintenance or an AFW and SAFW pump in level (i.e., between tagout boundaries, Spec. system.

maintenance at the same time). If work is done in such as a functional equipment group) this manner, it may be appropriate to account for when directed by procedures the unavailability of both SSCs in a single term.

(b) Examples of out-of-service Modeling of station maintenance practices that unavailability to be modeled are as result in planned maintenance evolutions with follows: more than a single PRA component OOS (i.e.,

shadowing equipment outages) can help to (1) train outages during a work window minimize the number of random failure for preventive/corrective maintenance sequences and ensure there is not "double (2) a functional equipment group (FEG) counting" of unavailability in the PRA.

removed from service for preventive/corrective maintenance (3) a relief valve taken out of service

License Amendment Request Attachment 1 Response to Request for Additional Information Page 10 of 46 Docket No. 50-244 Updated Table 2-1 Internal Events PRA Peer Review - Findings SR Topic Status Finding/Observation Disposition Impact to TSTF-425 HR-G3 When estimating HEPs EVALUATE the Complete F&O HR-63-01: Details regarding certain This F&O was addressed in the Fire HRA No impact to TSTF 425. The HRAs impact of the following plant-specific elements of the analysis were lacking in the Notebook for the fire related human have been reviewed to ensure and scenario-specific performance HRA Calculator for a sufficient number of actions. This included almost all of the the needed parameters for the non-fire related HRA events as most of shaping factors: HFEs to judge that this requirement was not evaluation have been populated.

the non-fire related HRAs are included (a) quality [type (classroom or met. Evidence that the relevant aspects CBDM is now used as a max in the fire model as well.

simulator) and frequency] of the cited in the SR are addressed for each HFE is Consideration of cue clarity and function of CBDT and HCR/ORE.

operator training or experience critical to assuring that an appropriate complexity were considered as part of RCHFDMAKEUP as a specific (b) quality of the written procedures analysis has been performed. This is the 2015 internal events model update example has a timing basis from and administrative controls particularly important in the case of HRA, for Ginna. Any and all additions to cue Key Input 51. When the (c) availability of instrumentation for which the methods are less clarity and complexity have been annunciator model is used, there incorporated into the HRA Calculator needed to take corrective actions straightforward than they are for many is a clear discussion as to the database file for the FPIE model, and (d) degree of clarity of the other parts of the PRA. applicability.

will also be incorporated in Appendix I cues/indications of the 2015 Ginna FPIE HRA Notebook.

(e) human-machine interface As such, the 2015 internal events PRA (f) time available and time required model update is consistent with HR-G3 to complete the response including the clarifications provided in (g) complexity of the required RG 1.200, Revision 2.

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.

License Amendment Request Attachment 1 Response to Request for Additional Information Page 11 of 46 Docket No. 50-244 Updated Table 2-1 Internal Events PAA Peer Review - Findings SR Topic Status Finding/Observation Disposition Impact to TSTF-425 HR-11 DOCUMENT the human reliability Complete F&O HR-11-01: The bulk of the Documentation only. Same issue as for No impact to TSTF 425. This item analysis in a manner that facilitates documentation for the HRA is provided in HR-G3. has been addressed. See HR-G3.

PRA applications, upgrades, and peer the HRA Calculator. There are numerous review. 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.

QU-BS Fault tree linking and some other Open F&O QU-BS-01: In Section 3.1 of the QU Documentation only: Provide a The circular logic process is self-modeling approaches may result in Notebook, a mention is made that circular discussion in the Quantification revealing when a support gate is circular logic that must be broken logic checks were performed on the Notebook Section 3.1 of the added to the tree the CAFTA before the model is solved. BREAK integrated top logic model to ensure it did methodology used to address circular software identifies a circular logic the circular logic appropriately. not exist. An example is listed, but there is logic. issue. The circular logic is broken Guidance for breaking logic loops is no further discussion. System notebooks by inserting as much of the logic provided in NUREG/CR-2728 [2-13]. reviewed generally state in Section 3.3 what clip into the tree as possible.

When resolving circular logic, DO was done when circular logic was identified, Providing more examples of this NOT introduce unnecessary but no discussion of the methodology was in the documentation is not conservatisms or non-conservatisms. provided nor how conservatisms or non- expected to affect the TSTF-425 conservatisms are avoided. No evidence evaluation.

that the required analysis was not performed.

License Amendment Request Attachment 1 Response to Request for Additional Information Page 12 of 46 Docket No. 50-244 Updated Table 2-1 Internal Events PRA Peer Review- Findings SR Topic Status Finding/Observation Disposition Impact to TSTF-425 LE-C2 INCLUDE realistic treatment of Open F&O LE-C2a-Ol: It is conservative to NOT There are limited operator actions that There are limited operator

[2005: feasible operator actions following take credit for operator actions post core could influence LERF at Ginna, so the actions that could influence LERF LE-C2a] the onset of core damage consistent damage. This is a requirement of the effect of such actions is not likely to be at Ginna, so the effect of such with applicable procedures, e.g., standard to move from Category I to significant. Moreover, it is likely that actions is not likely to be EOPs/SAMGs, proceduralized actions, Category II. there will not be a need for a Category significant. If post-core-damage or Technical Support Center II rating in this area to meet the operator actions are credited, guidance. requirements for most risk-informed LERF estimates could be reduced, applications. One approach to but the impact would be reaching Category II would be to minimal. The omission of these include post-core damage operator operator actions is conservative actions in the PRA. It is also possible and does not adversely impact that simply identifying operator actions the use of the model for TSTF-and showing quantitatively that they 425 analysis/or risk increases or will have a negligible impact on LERF baseline risk. URE 0835 is open will be sufficient to meet the to develop a post core-damage requirements of Category II. RCS depressurization recovery.

As part of the Exelon program, open UREs are evaluated for impact on the TSTF risk assessments.

License Amendment Request Attachment 1 Response to Request for Additional Information Page 13 of 46 Docket No. 50-244 Updated Table 2-1 Internal Events PRA Peer Review - Findings SR Topic Status Finding/Observation Disposition Impact to TSTF-425 LE-Cll JUSTIFY any credit given for Open F&O LE-C9a-Ol: It does not appear that The requirement is to justify credit As no equipment or HRA is

[2005: equipment survivability or human credit was taken for continued operation of taken for equipment survivability or credited post-containment LE-C9a] actions that could be impacted by equipment and operator actions that could human actions that could be affected failure, the PRA model remains a containment failure. be impacted by containment failure . This is by containment failure. conservative CAT I. This does not a requirement of the standard to move from In the Ginna Level 2 Analysis, early adversely impact the use of the Category I to Category II. containment failure after core- model for TSTF-425 analysis for damage and vessel breach is the risk increases or baseline risk.

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.

LE-C13 PERFORM a containment bypass Open F&O LE-Cl0-01: Credit for scrubbing was Review the possible credit for release A sensitivity for impact of

[2005 analysis in a realistic manner. not taken. A sensitivity for impact of scrubbing to reduce LERF. scrubbing was performed and it LE-ClO] JUSTIFY any credit taken for scrubbing was performed and it was was determined that the impact scrubbing (i.e., provide an determined that the impact of not of not considering scrubbing is engineering basis for the considering scrubbing is negligible. This is a negligible for base LERF decontamination factor used). requirement of the standard to move from conditions. However, a review Category I to Category II. 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.

License Amendment Request Attachment 1 Response to Request for Additional Information Page 14 of 46 Docket No. 50-244 Updated Table 2-1 Internal Events PRA Peer Review- Findings SR Topic Status Finding/Observation Disposition Impact to TSTF-425 MU-Dl A PRA Configuration Control Program Complete F&O MU-Dl PRA Configuration Control The CRMP database has a placeholder This configuration control issue shall be in procedure (GNG-CM-1.01-3003) Step 5.13 for a listing of PRA applications. This has been addressed. No impact place. It shall contain the following provides guidance for updating risk- portion of the database has been to TSTF 425.

key elements: informed applications. The process populated to ensure all applications (a) a process for monitoring PRA described depends upon a database requiring update following a model inputs and collecting maintained by the Fleet PRA Services new information Supervisor to identify current living revision can be easily identified.

(b) a process that maintains and applications requiring change assessment upgrades the PRA other than those related to maintenance to be consistent with the as-built, as rule performance criteria. No such database operated plant could be identified for Ginna.

(c) a process that ensures that the cumulative impact of pending Without a current list of risk-informed changes is considered when applying applications, the maintenance and update the PRA process is dependent upon the knowledge (d) a process that maintains and experience of the staff to know which configuration control of computer applications require update. This creates codes used to support PRA the possibility that an application could be quantification missed in the update process.

(e) documentation of the Program

License Amendment Request Attachment 1 Response to Request for Additional Information Page 15 of 46 Docket No. 50-244 Updated Table 2-1 Internal Events PRA Peer Review - Findings SR Topic Status Finding/Observation Disposition Impact to TSTF-425 IFSO-A4 For each potential source of flooding complete F&O IF-B2-0l: Failure mechanisms are Address the potential for human- No impact to TSTF 425.

[2005: water, IDENTIFY the flooding addressed in conjunction with the caused flooding in the Internal Flooding Discussion of human caused IF-B2] mechanisms that would result in a calculation of flood frequencies, in Section Study (51 - 9100978 - 000). Describe floods is discussed in detail in fluid release. INCLUDE: 5.2 of document 51-9100978-000. Failures the situations where a human error Section 3.3 and 5.3 of Internal (a) failure modes of components of components in piping systems other than could result in flooding (e.g., Flood Notebook (Gl-IF-0000-rl) for various systems. Based on such as pipes, tanks, gaskets, tanks are explicitly addressed by the EPRI inadvertent valve opening, inadvertent the analyses performed, one expansion joints, fittings, seals, etc. pipe failure data base. This was the source train realignment, doors left open) and maintenance induced flood was (b) human-induced mechanisms that employed to characterize the frequencies of estimate the probabilities of such added to the model, FL-ABO-M-could lead to overfilling tanks, floods for Ginna. There has, however, been events. Model such floods that cannot SW - 2,000 gpm SW flood in the diversion of flow through openings a very limited attempt to address human- be screened. Consistent with the Aux Building due to created to perform maintenance; induced flood mechanisms, as required by Standard, utilize generic data as maintenance, isolated within 65 minutes.

inadvertent actuation of fire item (b) of SR IF-B2. required by SR IFEV-A7 (IF-06 in 2005 suppression system Standard)

(c) other events releasing water into Such events have been important causes of the area 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.

License Amendment Request Attachment 1 Response to Request for Additional Information Page 16 of 46 Docket No. 50-244 Updated Table 2-1 Internal Events PRA Peer Review - Findings SR Topic Status Finding/Observation Disposition Impact to TSTF-425 IFSN-A6 For the SSCs identified in I FSN-A5 Open F&O IF-C3-0l: There is no discussion of Cat II: INCLUDE failure by submergence Failures due to jet impingement

[2005: (2005 text: IF-C2c), IDENTIFY the failures due to jet impingement or pipe and spray in the identification process. and pipe whip are now discussed IF-C3] susceptibility of each SSC in a flood whip. There is limited consideration of ASSESS qualitatively the impact of in Section 3.3.1 of the Internal area to flood-induced failure failure due to humidity/high temperature flood-induced mechanisms that are not Flood Notebook Gl-IF-0000 rl.

mechanisms. due to failure of heating steam lines. There formally Failures due to Spray are INCLUDE failure by submergence and is also no discussion of criteria employed to addressed (e.g., using the mechanisms discussed in Section 3.3.2.

spray in the identification process. consider the potential for spray failures. listed under Capability Category Ill of Impacts due to spray were EITHER: this requirement), by using assumed to exist within 10 feet a) ASSESS qualitatively the impact of To meet Capability Category II, it is conservative assumptions. of a break location (modeled?).

flood-induced mechanisms that are necessary either to provide at least a [SAIC note: these mechanisms include Spray events are discussed in the not formally addressed (e.g., using qualitative assessment of the potential for submergence, spray, jet impingement, IF Flood notebook Section 4.5.

the mechanisms listed under jet impingement and pipe whip, or to state pipe whip, humidity, condensation, Two locations were identified in Capability Category Ill of this that these failure mechanisms were not temperature concerns] the Aux Building where Fire requirement), by using conservative considered. It is also required that potential Service Water could impact assumptions; OR spray failures be evaluated. While spray Revise the Internal Flooding Study (51- safety related busses and these b) NOTE that these mechanisms are failures are discussed, there are no criteria 9100978 - 000) to describe the criteria are explicitly modeled (FL-ABM-not included in the scope of the specified that would provide assurance that used to determine the potential for FSW-BUS15 and FL-ABO-FSW-evaluation. they had been considered in a consistent failure resulting from spray. Reference BUS14). URE 1179 documents and adequately comprehensive manner. Appendix C for a listing of components that IF Notebook needs Appendix impacted by spray. Describe how C completed to complete Provide the requisite discussion of pipe whip potential spray impact was addressed documentation of spray impacts and jet impingement to satisfy the standard. in the model. Confirm that the and modeling of additional spray Specify appropriate criteria for spray assignment of spray impact is floods if appropriate. This would impacts, and assure that the potential spray consistent with the criteria used. be evaluated for any potential failures adequately reflect these criteria. impacts to a surveillance In addition, include a qualitative frequency interval extension at discussion of the potential impact of jet the time of the evaluation but is impingement, pipe whip, humidity, not expected to have a condensation, and temperature significant impact.

effects.

License Amendment Request Attachment 1 Response to Request for Additional Information Page 17 of 46 Docket No. 50-244 Updated Table 2-1 Internal Events PRA Peer Review - Findings SR Topic Status Finding/Observation Disposition Impact to TSTF-425 IFSN-A8 IDENTIFY inter-area propagation complete F&O IF-C3b-01: The analysis does not Cat II, Ill: IDENTIFY inter-area. No impact to TSTF 425.

[2005: through the normal flow path from document consideration of potential barrier A discussion of structural failure IF-C3b] one area to another via drain lines; failures due to flooding loads (structural INCLUDE potential for structural failure of barriers credited as barriers and areas connected via back flow failures, failures of doors, etc.) This is (e.g., of doors or walls) due to flooding has been added to the IF through drain lines involving failed required to meet capability categories loads and the potential for barrier Notebook rl, Section 4.2.1.

check valves, pipe and cable beyond Capability Category I. unavailability, including maintenance penetrations (including cable trays}, activities.

doors, stairwells, hatchways, and Review flood barriers and identify and HVAC ducts. INCLUDE potential for evaluate any whose failures could Include a discussion of the potential for structural failure (e.g., of doors or contribute adversely to propagation of barrier failure due to flooding, walls) due to flooding loads. 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.

License Amendment Request Attachment 1 Response to Request for Additional Information Page 18 of 46 Docket No. 50-244 Updated Table 2-1 Internal Events PRA Peer Review- Findings SR Topic Status Finding/Observation Disposition Impact to TSTF-425 IFSN- USE potential human mitigative Complete F&O IF-CS-01: Only one flood appears to The FSW breaks in the IBN are no This F&O has been addressed in A16 actions as additional criteria for have been screened based on qualitative longer screened in the current model. the current model and

[2005: screening out flood sources if all the consideration of potential human action; for New flood initiator FL-/BN-FSW-2K has documentation and has no IF-CB] following can be shown: that action (2000 gpm FSW break in IBN), been added to the model and to the impact on TSTF-425 analysis.

there doesn't appear to be any justification Flood Notebook.

(a) flood indication is available in the for the time identified (190 min). Nothing control room; other than time available is cited as rationale for screening the event.

(b) the flood source can be isolated; and To meet Capability Category II, it is necessary to characterize potential human (c) the mitigative action can be actions that could terminate flooding more performed with high reliability for explicitly than was done in this case.

the worst flooding initiator (2005 text: flood from that source). High Address the required aspects for this and reliability is established by any other human actions used in justifying demonstrating, for example, that the screening out flood scenarios.

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.

License Amendment Request Attachment 1 Response to Request for Additional Information Page 19 of 46 Docket No. 50-244 Updated Table 2-1 Internal Events PRA Peer Review - Findings SR Topic Status Finding/Observation Disposition Impact to TSTF-425 IFEV-AG GATHER plant-specific information Complete F&O IF-DSa-01: The current analysis does In the current updated internal flood Plant specific experience with

[2005: on plant design, operating practices, not adequately address plant-specific analysis, a review was conducted to internal flooding, water hammer IF-D5a] and conditions that may impact flood characteristics that might affect the manner assess potential issues with material is addressed in the IF Notebook likelihood (i.e., material condition of in which the frequencies of flooding are condition, water hammer, and aging rev 1 in Sections 3.3. A fluid systems, experience with water estimated. management strategies. The plant discussion of Human-induced hammer, and maintenance-induced specific information has been floods is contained in Section 5.3.

floods). In determining the flood- To meet Capability Category II, it is required considered and use of generic data is initiating event frequencies for flood that plant-specific information be collected found to be appropriate for Ginna. This F&O has been addressed and scenario groups, USE a combination and considered on a variety of aspects does not impact TSTF-425 of the following (2005 text does not (including material condition of fluid For maintenance-induced and other analysis.

include "of the following") systems, experience with water hammer, human-caused flooding, see IFSO-A4 and maintenance-induced floods). The Which is statused as "Complete.".

(a) generic and plant-specific current analysis is limited to the use of operating experience; generic failure rates. This is consistent with (b) pipe, component, and tank Capability Category I.

rupture failure rates from generic data sources and plant-specific Address potential issues with material experience; {2005 text: and) condition, experience with water hammer, (c) engineering judgment for etc. In particular, further attention should consideration of the plant-specific be paid to the possibility of maintenance-information collected. induced and other human-caused flooding.

License Amendment Request Attachment 1 Response to Request for Additional Information Page 20 of 46 Docket No. 50-244 Updated Table 2-1 Internal Events PRA Peer Review - Findings SR Topic Status Finding/Observation Disposition Impact to TSTF-425 IFEV-A7 INCLUDE consideration of human- Complete F&O IF-06-01: Initiating events that could See IFSO-A4. No impact to TSTF 425.

[2005: induced floods during maintenance result from human actions were considered Discussion of human caused IF-D6] through application of generic data. only for a small number of possible floods is discussed in detail in maintenance activities. These flood Section 3.3 and 5.3 of Internal contributors were not evaluated using Flood Notebook {Gl-IF-0000-rl) generic data as required. for various systems. Based on the analyses performed, one Operating experience for nuclear power maintenance induced flood was plants has provided evidence that human- added to the model, FL-ABO-M-caused floods can be important. The SR SW - 2,000 gpm SW flood in the requires that such floods be evaluated using Aux Building due to at least generic data to meet Capability maintenance, isolated within 65 Category I or II. minutes.

Perform a more detailed assessment of potential human-caused floods, and apply at least generic data to characterize their frequencies.

License Amendment Request Attachment 1 Response to Request for Additional Information Page 21of46 Docket No. 50-244 Updated Table 2-1 Internal Events PRA Peer Review - Findings SR Topic Status Finding/Observation Disposition Impact to TSTF-425 IFEV-A8 SCREEN OUT flood scenario groups if Complete F&O IF-07-01: Quantitative screening of Update the Internal Flooding Study (51 No impact to TSTF 425. This issue

[2005: (a) the quantitative screening criteria some scenarios was performed, but it is not - 9100978 - 000} to describe the criteria has been addressed. Internal IF-D7] in IFSN-AlO (2005 text: IE-C4}, as clear what criteria were applied in doing so. used to screen flood scenarios. If Flood Notebook Section 4.6, applied to the flood scenario groups, The criteria should be defined and applied in current screening criteria are not well Screening Scenarios and Sources, are met; OR a clear and consistent manner. defined, develop such criteria and was updated to document the (b) the internal flood-initiating event apply them to scenarios addressed in screening criteria used. Figure affects only components in a single SRs IF-D7 and IF-E3a provide explicit criteria the analysis. 4.1, was added which shows the system, AND it can be shown that the for performing quantitative screening of Screening Criteria and Table 4.6 product of the frequency of the flood scenarios. The IF Notebook documents was edited to show the screening flood and the probability of SSC that some scenarios were screened on low criterion used for various flood failure given the flood is two orders frequency, but does not invoke any scenarios.

of magnitude lower than : particular criteria in doing so.

the product of the non-flooding frequency for the corresponding Provide a clear set of criteria for performing initiating events in the PRA, AND the quantitative screening of flood scenarios, random (non-flood-induced) failure and apply the criteria in a clear and probability of the same SSCs that are consistent manner.

assumed failed by the flood.

If the flood impacts multiple systems, DO NOT screen on this basis.

License Amendment Request Attachment 1 Response to Request for Additional Information Page 22 of 46 Docket No. 50-244 Updated Table 2-1 Internal Events PRA Peer Review - Findings SR Topic Status Finding/Observation Disposition Impact to TSTF-425 IFQU-AS If additional human failure events are complete F&O IF-ES-01: It was not clear that the Re-examine each HFE included in the No impact to TSTF-425.

[2005: required to support quantification of requirements were met in all cases. For flooding analysis. Perform operator Ginna Station Flooding Human interviews as needed or identify and IF-ES] flood scenarios, PERFORM any example, interviews to establish aspects Reliability Analysis (HRA) document previously performed interviews.

human reliability analysis in such as response times were apparently documents the flood recovery accordance with the applicable performed as part of the flood analysis, but actions (Areva Document No.:

Required operator interviews should requirements described in 2-2.5 the HRA was dramatically changed and new comprise the following: 51-9099406-000 located in GSN (2005 text: Tables 4.5.5-2(e) through interviews/changes were not incorporated, 1. evaluate the flooding events based on 0157). The information and HRA 4.5.5-2(h)). nor were any inputs obtained from the HRA similarities to identify a select set of values in this notebook were performed as part of the flood analysis scenarios to review with the operators (for verified to be consistent with the carried forward. example, categorized by the system that HRA actions being used in the generated the flood, e.g., fire protection) internal flood model. No

2. schedule interview sessions of about 1/2 It is necessary to perform the assessment of additional interviews were hour to an hour per each flooding scenario, HFEs associated with internal flooding in the identified as being necessary.

conducted separately with two different same manner as for other HFEs. The operators (preferably one experienced, one requirements to confirm procedure paths, novice) to get diverse opinions.

timing, etc. via interviews with operators 3. include questions on timing consistent were not met for a number of events. with the HRA Calculator Time Window screen for time of cue, time to diagnosis, Re-examine the HFEs associated with time for execution/manipulation of action (including travel time, with potential flood-internal flooding, and either perform related access delays). Be sure to ask about needed operator interviews or identify and any differences for floods initiated in same document existing inputs. 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.

License Amendment Request Attachment 1 Response to Request for Additional Information Page 23 of 46 Docket No. 50-244 Updated Table 2-1 Internal Events PRA Peer Review - Findings SR Topic Status Finding/Observation Disposition Impact to TSTF-425 IFQU-Bl DOCUMENT the internal flood Open F&O IF-Fl-01: The documentation is comprised Documentation only: Revise the Internal This documentation item will not primarily of the internal flooding notebook, Flooding Study (51- 9100978 - 000) to meet

[2005: accident sequences and impact the TSTF 425 analysis.

supplemented heavily with information provided the documentation requirements of the IF-Fl] quantification in a manner that in a set of Excel worksheets. The notebook is This item has largely been 2009 Standard. Address NRC Resolutions as facilitates PRA applications, annotated to provide a link to elements of the addressed by adding tables in appropriate.

upgrades, and peer review. worksheets, and an "assumption" provides the Section 5.2 that show the formal tie between the notebook and the It is recommended that the Study be development of each initiating worksheets. Some areas in which the links were indirect or missing were noted. reformatted to be consistent with the HLRs event frequency, adding an and SRs of the Standard, integrating Initiating Event Summary Table In general, the manner in which important parts appropriate parts ofthe worksheets into the (section 5.2.17), adding a of the flood analysis are documented in what primary document. This will provide a simplified set of arrangement would usually be characterized as an informal set document that can be easily reviewed of worksheets is judged not to meet the drawings showing each flood requirement that the analysis be documented in a against the standard and easily followed by area (Appendix K), defining spray manner that facilitates applications, upgrades, personnel not involved in the original and peer review. modeling criteria (Section 3.3.2) analysis.

and identifying for each flood In addition to developing a single integrated set area whether it was screened Consistent with the F&O, include the of documentation for the internal flood analysis, following in the revised Study: and the screening criterion used there were several areas in which additional documentation would make the analysis more

• Include a set of simplified arrangement (Table 4.6). The remaining item tractable have been provided in connection to drawings to explicate the definition of flood is to develop the criteria used to other SRs. These include the following: areas and help in understanding aspects perform quantitative screening,

• Include a set of simplified arrangement such as flood propagation.

drawings to explicate the definition of flood areas if applicable, in Section 6.0 (URE

• Tabulate the flood areas and identify and help in understanding aspects such as flood 1177).

clearly which are screened and which propagation.

retained for further analysis to make the

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

(qualitative or quantitative) are employed in

• Define explicitly the criteria used to screening each flood area. perform quantitative screening as noted in

• Define explicitly the criteria used to perform Section 6.0.

quantitative screening as noted in Section 6.0.

• Define the criteria used to determine whether a

• Define the criteria used to determine PRA component was susceptible to failure due to whether a PRA component was susceptible spray. to failure due to spray.

License Amendment Request Attachment 1 Response to Request for Additional Information Page 24 of 46 Docket No. 50-244 Updated Table 2-1 Internal Events PRA Peer Review - Findings SR Topic Status Finding/Observation Disposition Impact to TSTF-425 IFQU-B3 DOCUMENT sources of model Complete F&O IF-F3-01: Section 7 of the IF Notebook In estimating the event mean frequency This F&O has been addressed and

[2005: uncertainty and related assumptions provides a discussion of three areas for each internal flood initiator, the does not impact TSTF-425 IF-F3] (as identified in QU-El and QU-E2) considered to be major sources of initiating event uncertainty parameters analysis.

associated with the internal flood uncertainty in the flood analysis. This does from the EPRI 1013141 data were used accident sequences and not constitute an adequate characterization and error factors reported in the quantification. of the sources of uncertainty associated Internal Flood analysis notebook (Gl-with the flood analysis or a comprehensive PRA-012). These parametric (2005 text: Document the key discussion of the assumptions that could uncertainty values propagate to the assumptions and the key sources of have an effect on the results. end results using the CAFTA PRA uncertainty associated with the software. Modeling uncertainty for the internal flooding analysis.) A reasonably thorough investigation of internal flood portion of the PRA was sources of uncertainty is necessary for also addressed and documented in Gl-proper characterization of the flood PRA-012 using the guidance found in analyses and results. EPRI 1016737. The finding for IF-F3-01 is considered to be resolved.

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.

License Amendment Request Attachment 1 Response to Request for Additional Information Page 25 of 46 Docket No. 50-244 Table 2-2 Internal Events PRA Peer Review - Selected Suggestions SR Topic Status Finding/Observation Disposition Impact to TSTF-425 AS-A9 USE realistic, Open USE realistic, applicable (i.e., from Consistent with Capability Category Ill for SR For this suggestion, no adverse applicable (i.e., from similar plants) thermal hydraulic As-A9, Ginna uses realistic, plant-specific impact is expected on the use of the similar plants) thermal analyses to determine the accident thermal hydraulic analysis. Results are model for TSTF-425 analysis for risk hydraulic analyses to progression parameters (e.g., reviewed for reasonableness; however, the increases or baseline risk.

determine the accident timing, temperature, pressure, review for RCP Seal LOCAs is not documented, progression parameters steam) that could potentially affect and should be re-performed. This issue is captured as URE 0850 (e.g., timing, the operability of the mitigating which will be reviewed for temperature, pressure, systems. The Seal LOCA results are applicability for each ST/ change steam) that could inconsistent with respect to generic evaluation as required by Exelon potentially affect the industry data. Seal LOCA Cases procedural guidance.

operability of the RPSL364CD and RPSL960CD should mitigating systems.

be reviewed and compared with WCAP 16141.

DA-Dl CALCULATE realistic Complete There is no explanation of how the The data notebook has been updated to This F&O has been addressed with parameter estimates for composite data located in the third support the latest PRA model update using the current PRA model and significant basic events and second last columns of Bayesian update techniques or best available documentation, and does not impact based on relevant "Component Generic Failure Data" generic data sources including specific the TSTF-425 analysis.

generic and plant-specific are determined from the three references for the generic data sources.

evidence. generic sources listed.

The data appears to be reasonable, but the method used to develop it is not documented.

License Amendment Request Attachment 1 Response to Request for Additional Information Page 26 of 46 Docket No. 50-244 Table 2-2 Internal Events PRA Peer Review - Selected Suggestions SR Topic Status Finding/Observation Disposition Impact to TSTF-425 DA-D6 USE generic common Open The current modeling uses the This suggestion F&O has not been For this suggestion, no adverse cause failure probabilities failure to run alpha factor for the incorporated into the updated PRA model or impact is expected on the use of the consistent with available failure to run in the first hour. documentation. model for TSTF-425 analysis for risk plant experience. NUREG/CR-6268, Rev. 1, Table 5-7 increases or baseline risk.

EVALUATE the common indicates that events are coded as cause failure probabilities failure to run if the "component fails This issue is captured as URE 0820 in a manner consistent which will be reviewed for to continue running at rated with the component conditions after reaching rated applicability for each ST/ change boundaries ..

conditions." This implies that the evaluation as required by Exelon failures to run in the first hour procedural guidance.

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.

DA-D6 USE generic common Complete The data used for the latest update The data analysis and notebook have been This F&O has been addressed with cause failure probabilities was from an INEEL/NRC report updated using current CCF generic data. the current PRA model and consistent with available published in 2002 (Key Input 1). documentation, and does not impact plant experience. the TSTF-425 analysis.

EVALUATE the common Update the generic common cause cause failure probabilities data to a more current version.

in a manner consistent with the component boundaries ..

License Amendment Request Attachment 1 Response to Request for Additional Information Page 27 of 46 Docket No. 50-244 Table 2-2 Internal Events PRA Peer Review - Selected Suggestions SR Topic Status Finding/Observation Disposition Impact to TSTF-425 DA-E3 DOCUMENT the sources Complete G1-DA-OOOO, Revision 1, Section The previous assumption applicable to This F&O has been addressed with of model uncertainty and 4.1 states that the "Ginna PRA staggered testing is no longer applicable. The the current PRA model and related assumptions (as assumed staggered testing for all model includes staggered and non-staggered documentation, and does not impact identified in QU-E1 and components subject to CCF." testing for CCF updates, as applicable. A data the TSTF-425 analysis.

QU-E2) associated with This is not captured as an update has recently been completed for the data analysis. assumption which could common cause factors. MSIVs are modeled as contribute to uncertainty in non-staggered.

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.

IE-C12 In the ISLOCA frequency Open PRA Modeling methodology for This suggestion remains open as it has not As the suggestion remains open until

[Now IE- analysis, INCLUDE the pipe rupture analysis (for been formally closed-out formally closed-out, this issue is C14] features of plant and ISLOCA frequency estimation) captured as URE 0822 which will be procedures that influence has changed over the past reviewed for applicability for each ST/

the ISLOCA frequency: decade. Recommend updating to change evaluation as required by the latest methodology has Exelon procedural guidance.

documented in latest EPRI tech report.

IF-Bl [Now For each flood area, Complete The screening process appears This documentation suggestion has been This F&O has been addressed with IFSO-Al] IDENTIFY the potential to be adequate, but the manner addressed. The internal flood notebook was the current PRA model and sources of flooding in which criteria for screening are updated to summarize all the flooding areas, documentation, and does not impact applied and the degree to which and if/why they were screened.

the TSTF-425 analysis.

such criteria have been employed in a systematic manner is not clear.

License Amendment Request Attachment 1 Response to Request for Additional Information Page 28 of 46 Docket No. 50-244 Table 2-2 Internal Events PRA Peer Review - Selected Suggestions SR Topic Status Finding/Observation Disposition Impact to TSTF-425 LE-Cl Develop (LEAF) accident Open The definition of LEAF should This documentation suggestion has not yet This documentation suggestion sequencestothelevelof include the basis for the been addressed. would have no adverse impact on detail to account for the declaration of General TSTF analysis for risk increase or Emergency.

potential contributors. baseline risk.

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.

QU-E4 For each source of model Complete Table H-1 in the Quantification The updated 2009 revision of the standard This suggestion is not applicable to uncertainty and related Notebook addresses each key does not contain the requirement to perform the 2009 ASME PRA standard and assumption identified in source of uncertainty however no sensitivity studies to meet Cat 11/111 as it is does not impact the TSTF-425 QU-E1 and QU-E2, sensitivity studies related to expected that specific sources of uncertainty analysis.

respectively, IDENTIFY these key sources of uncertainty will be addressed on an application specific how the PRA model is are addressed with sensitivity basis.

affected (e.g., studies. The Quantification introduction of a new notebook describes that basic event, changes to uncertainties that are associated basic event probabilities, with scope and level of detail will change in success only be addressed for specific criterion, introduction of a applications. Cat 11 requires that new initiating event). sensitivity analysis be performed to address key assumptions.

License Amendment Request Attachment 1 Response to Request for Additional Information Page 29 of 46 Docket No. 50-244 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, 11 Joint 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 Attachment 1 Response to Request for Additional Information Page 30 of 46 Docket No. 50-244

i. In SR FSS-A4, RG 1.200, Revision 2 changes "one of more" to 11 sufficient. 11 ii. In Fire PRA F&O FSS-F1-01, RG 1.200, Revision 2 changes SR FSS-F1 from 11 one 11 11 or more fire scenarios that could to a 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 post-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 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 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 Attachment 1 Response to Request for Additional Information Page 31of46 Docket No. 50-244 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 Attachment 1 Response to Request for Additional Information Page 32 of 46 Docket No. 50-244 Exelon Response to RAJ 2.d.iv 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 Attachment 1 Response to Request for Additional Information Page 33 of 46 Docket No. 50-244 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 Attachment 1 Response to Request for Additional Information Page 34 of 46 Docket No. 50-244 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 Attachment 1 Response to Request for Additional Information Page 35 of 46 Docket No. 50-244 CDF Basic Event Event Name Description (Cutset) Value 1.03E-02 AXTMSAFSGB SAFW TRAIN D TO SIG 0.0.S. DUE TO T/M 9.41 E-01 MODE1 MODE1 1.00E+OO NOSBO TAG - NO STATION BLACKOUT (SBO) 8.94E-06 SWCCFPUMPR_ALL CCF OF ALL COMPONENTS IN GROUP 1

SWCCFPUMPR 1 1.00E+OO TRANSX 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 11 back-referenced 11 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 Attachment 1 Response to Request for Additional Information Page 36 of 46 Docket No. 50-244 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 Attachment 1 Response to Request for Additional Information Page 37 of 46 Docket No. 50-244 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 11 that the impact 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 1.174, Revision 2, which provides quantitative risk acceptance guidelines for changes to CDF and LERF. Revision 2 of RG 1.174 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 1.174, Revision 2, which provides for changes to CDF and LERF.

License Amendment Request Attachment 1 Response to Request for Additional Information Page 38 of 46 Docket No. 50-244 Exelon Response to RAJ 9.a 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 Attachment 1 Response to Request for Additional Information Page 39 of 46 Docket No. 50-244 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/ 1O.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/ 1O.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/ 1O.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 Attachment 1 Response to Request for Additional Information Page 40 of 46 Docket No. 50-244 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 CAFTA 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 IF-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 Response to Request for Additional Information Page 41 of 46 Docket No. 50-244 11 that have been peer reviewed. If not, confirm whether a focused Peer Review had been 11 performed 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 Attachment 1 Response to Request for Additional Information Page 42 of 46 Docket No. 50-244 Confirm that any gap assessment and, if identified as required due to model upgrades, focused-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 Attachment 1 Response to Request for Additional Information Page 43 of 46 Docket No. 50-244 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 Attachment 1 Response to Request for Additional Information Page 44 of 46 Docket No. 50-244 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 1a 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, 11 A 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 11 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 11 reference in the bases description for this SR.

Exelon response to RAI 1b 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 Attachment 1 Response to Request for Additional Information Page 45 of 46 Docket No. 50-244 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 11 30 minute 11 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 11 30 minute 11 references in the bases description for this SR.

Exelon response to RAI 1c 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 1d 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 Attachment 1 Response to Request for Additional Information Page 46 of 46 Docket No. 50-244 References

[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 ML15166A075.

[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, Risk-Informed Method for Control of Surveillance Frequencies, Industry Guidance Document, NEl 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 ML13093A064.

AITACHMENT2 License Amendment Request R. E. Ginna Nuclear Power Plant Docket No. 50-244 Application for Technical Specification Change Regarding Risk-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

RTS Instrumentation B 3.

3.1 REFERENCES

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, "Power-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

i::~~~t !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

RCS Operational LEAKAGE B 3.4.13 monitors or radiochemical grab sampling in accordance with the EPRI guidelines (Reference 5). If\ ~

~

REFERENCES 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

RCS PIV Leakage B 3.4.14 REFERENCES 1. 10 CFR 50.2.

2. 10 CFR 50.55a(c).

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

~Plants.

~'V

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

Control Bank Insertion Limits 8 3.1.6 SURVEILLANCE SR 3.1.6.1 REQUIREMENTS 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

RCS Minimum Temperature for Criticality 8 3.4.2 ACTIONS 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.

SURVEILLANCE SR 3.4.2.1 REQUIREMENTS 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 tem~erature 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 lil~elinood 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.

REFERENCES 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 October 2, 2015 _E_C_C_S~p-ip-i-ng~a-n_d_c_o_m_p_o_n_e_nt_s_h_a_v_e_t-he~p-ot_e_n-tia-1~

Docket No. 50-244 . . Attachment 5 o develop voids and pockets of entrained gases.

Page 4 of 7 Preventing and managing gas intrusion and INSERTD accumulation is necessary for proper operation of the ECCS and may also prevent water hammer, SR 3.5.2.8 pump cavitation, and pumping of non-condensible gas into the reactor vessel.

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 s~stem 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 Risk-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 Response to Request for Additional Information Page 1 of 1 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, "Risk-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.