ML15166A075

R.E. Ginna - 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)

ML15166A075
Person / Time
Site:	Ginna
Issue date:	06/04/2015
From:	Jim Barstow Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TSTF-425, Rev. 3
Download: ML15166A075 (253)

Text

l&Exelon Generation 200 Exelon WayKennett Square, PA 19348www.exeloncorp.com 10 CFR 50.90June 4, 2015U.S. Nuclear Regulatory Commission ATTN: Document Control DeskWashington, DC 20555R.E. Ginna Nuclear Power PlantRenewed Facility Operating License No. DPR-18NRC Docket No. 50-244

SUBJECT:

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)In accordance with the provisions of Title 10 of the Code of Federal Regulations (10 CFR50.90), "Application for amendment of license, construction permit, or early site permit,"Exelon Generation

Company, LLC (Exelon) is submitting a request for an amendment tothe Technical Specifications (TS), Appendix A of Renewed Facility Operating License No.DPR-18 for R. E. Ginna Nuclear Power Plant (Ginna).The proposed amendment would modify Ginna's TS by relocating specific surveillance frequencies to a licensee-controlled program with the implementation of Nuclear EnergyInstitute (NEI) 04-10, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies."

The changes are consistent with NRC-approved Industry Technical Specifications Task Force(TSTF) Standard Technical Specifications (STS) change TSTF-425, "Relocate Surveillance Frequencies to Licensee Control -Risk Informed Technical Specification Task Force (RITSTF)Initiative 5b, Revision 3," (ADAMS Accession No. ML090850642).

The Federal Register Noticepublished on July 6, 2009 (74 FR 31996), announced the availability of this TS improvement.

Attachment 1 provides a description of the proposed change, the requested confirmation ofapplicability, and plant-specific verifications.

Attachment 2 provides documentation ofProbabilistic Risk Assessment (PRA) technical adequacy.

Attachment 3 provides theexisting Ginna TS pages marked up to show the proposed changes.

Attachment 4 providesthe proposed Ginna TS Bases changes.

Attachment 5 provides a TSTF-425 versus GinnaTS Cross-Reference.

Attachment 6 provides the proposed No Significant HazardsConsideration.

Attachment 7 provides the proposed inserts.

A-U cii License Amendment RequestAdoption of TSTF-425, Rev. 3Docket No. 50-244June 4, 2015Page 2There are no regulatory commitments contained in this letter.Exelon requests approval of the proposed license amendment by June 4, 2016, with theamendment being implemented within 120 days.These proposed changes have been reviewed by the Plant Operations Review Committee and approved in accordance with Nuclear Safety Review Board procedures.

In accordance with 10 CFR 50.91, "Notice for Public Comment; State Consultation,"

a copyof this application, with attachments, is being provided to the designated State Official.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on the 4thday of June 2015.If you should have any questions regarding this submittal, please contact Enrique Villar at610-765-5736.

Respectfully, James BarstowDirector

-Licensing

& Regulatory AffairsExelon Generation

Company, LLCAttachments:

1.2.3.4.5.6.7.Description and Assessment Documentation of PRA Technical AdequacyProposed Technical Specification Page ChangesProposed Technical Specification Bases Page ChangesTSTF-425 (NUREG-1431) vs. Ginna Cross-Reference Proposed No Significant Hazards Consideration Proposed Insertscc: USNRC Region I Regional Administrator USNRC Senior Resident Inspector

-GinnaUSNRC Project Manager, NRR -GinnaA. L. Peterson, NYSERDAw/attachments I1 ATTACHMENT 1License Amendment RequestR. E. Ginna Nuclear Power PlantDocket No. 50-244Application 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)Description and Assessment LAR -Adoption of TSTF-425, Revision 3 Attachment 1Docket No. 50-244 Page 1 of 5DESCRIPTION AND ASSESSMENT

1.0 DESCRIPTION

The proposed amendment would modify the R. E. Ginna Nuclear Power plant (Ginna) Technical Specifications (TS) by relocating specific TS surveillance frequencies to a licensee-controlled program with the adoption of Technical Specification Task Force (TSTF) -425, Revision 3,"Relocate Surveillance Frequencies to Licensee Control -Risk Informed Technical Specification Task Force (RITSTF)

Initiative 5b" (Ref. 1). Additionally, the change would add a new program,the Surveillance Frequency Control Program, to TS Section 5, Administrative Controls.

The changes are consistent with NRC-approved Industry/lTSTF Standard Technical Specifications (STS) change TSTF-425, Revision 3, (ADAMS Accession No. ML090850642).

The FederalRegister notice published on July 6, 2009 (74 FR 31996) (Ref. 2), announced the availability ofthis TS improvement.

2.0 ASSESSMENT

2.1 Applicability of Published Safety Evaluation Exelon Generation

Company, LLC (Exelon) has reviewed the NRC staff's Model SafetyEvaluation for TSTF-425, Revision 3, dated July 6, 2009. This review included a review of theNRC staff's Model Safety Evaluation, TSTF-425, Revision 3, and the requirements specified inNEI 04-10, Revision 1, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies,"

(ADAMS Accession No. ML071360456)

(Ref.3).The traveler and Model Safety Evaluation discuss the applicable regulatory requirements andguidance, including the 10 CFR 50, Appendix A, General Design Criteria (GDC). Ginna was notlicensed to the 10 CFR 50, Appendix A GDC. However, the Ginna's Updated Final SafetyAnalysis Report (UFSAR),

in Section 3.1 "Conformance with NRC General Design Criteria,"

provides an assessment against the GDC. Based on the assessment performed anddescribed in the in the Ginna UFSAR, Exelon believes that the plant-specific requirements forGinna are sufficiently similar to the Appendix A GDC and represent an adequate technical basisfor adopting the proposed change.Attachment 2 includes Exelon's documentation with regard to Probabilistic Risk Assessment (PRA) technical adequacy consistent with the requirements of Regulatory Guide 1.200, Revision1, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities,"

(ADAMS Accession No. ML070240001)

(Ref. 4), Section4.2, and describes any PRA models without NRC-endorsed standards, including documentation of the quality characteristics of those models in accordance with Regulatory Guide 1.200.Exelon has concluded that the justifications presented in the TSTF proposal and the NRC staff'sModel Safety Evaluation prepared by the NRC staff are applicable to Ginna and justify thisamendment to incorporate the changes to the Ginna TS.

LAR -Adoption of TSTF-425, Revision 3 Attachment 1Docket No. 50-244 Page 2 of 52.2 Optional Changes and Variations The proposed amendment is consistent with the STS changes described in TSTF-425, Revision3; however, Exelon proposes variations or deviations from TSTF-425, as identified below, whichincludes differing Surveillance numbers.Revised (clean) TS pages are not included in this amendment request given the number ofTS pages affected, the straightforward nature of the proposed

changes, and outstanding Ginna amendment requests that will impact some of the same TS pages. Providing onlymark-ups of the proposed TS changes satisfies the requirements of 10 CFR 50.90,"Application for amendment of license, construction permit, or early site permit,"

(Ref. 5) inthat the mark-ups fully describe the changes desired.

This is an administrative deviation from the NRC staff's model application dated July 6, 2009 (74 FR 31996) with no impacton the NRC staff's Model Safety Evaluation published in the same Federal RegisterNotice. As a result of this deviation, the contents and numbering of the attachments forthis amendment request differ from the attachments specified in the NRC staff's modelapplication After NRC approval of TSTF-425, it was recognized that surveillance frequencies thathave not been changed under the Surveillance Frequency Control Program (SFCP) maynot be based on operating experience, equipment reliability or plant risk. Therefore, the TSTF and the NRC agreed that the TSTF-425 TS Bases insert, "The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk andis controlled under the Surveillance Frequency Control Program,"

should be revisedto state, "The Surveillance Frequency is controlled under the Surveillance Frequency Control Program."

The existing TS Bases information will be relocated to thelicensee-controlled SFCP.Attachment 5 provides a cross-reference between TSTF-425 versus the GinnaSurveillances included in this amendment request.

Attachment 5 includes a summarydescription of the referenced TSTF-425 TS Surveillances, which is provided forinformation purposes only and is not intended to be a verbatim description of the TSSurveillances.

This cross-reference highlights the following:

a. Surveillances included in TSTF-425 and corresponding Ginna Surveillances havediffering Surveillances numbers,b. Surveillances included in TSTF-425 that are not contained in the Ginna TS, andc. Ginna plant-specific Surveillances that are not contained in TSTF-425 Surveillances and, therefore, are not included in the TSTF-425 mark-ups.

In addition, there are Surveillances contained in TSTF-425 that are not contained in theGinna TS. Therefore, the NUREG-1431 mark-ups included in TSTF-425 for theseSurveillances are not applicable to Ginna. This is an administrative deviation fromTSTF-425 with no impact on the NRC staff's Model Safety Evaluation dated July 6, 2009(74 FR 31996).Ginna TS include plant-specific Surveillances that are not contained in NUREG-1431 and, therefore, are not included in the NUREG-1431 mark-ups provided in TSTF-425.

Exelon has determined that the relocation of the Frequencies for these Ginna plant-specific Surveillances is consistent with TSTF-425, Revision 3, and with the NRC staff's LAR -Adoption of TSTF-425, Revision 3 Attachment 1Docket No. 50-244 Page 3 of 5Model Safety Evaluation dated July 6, 2009 (74 FR 31996), including the scopeexclusions identified in Section 1.0, "Introduction,"

of the Model Safety Evaluation.

Changes to the Frequencies for these plant-specific Surveillances would be controlled under the SFCP. The SFCP provides the necessary administrative controls to requirethat Surveillances related to testing, calibration and inspection are conducted at afrequency to assure that the necessary quality of systems and components ismaintained, that facility operation will be within safety limits, and that the LimitingConditions for Operation will be met. Changes to Frequencies in the SFCP would beevaluated using the methodology and probabilistic risk guidelines contained in NEI 04-10, Revision 1, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies,"

(ADAMS Accession No.ML071360456),

as approved by NRC letter dated September 19, 2007 (ADAMSAccession No. ML072570267).

The NEI 04-10, Revision 1 methodology includesqualitative considerations, risk analyses, sensitivity studies and bounding

analyses, asnecessary, and recommended monitoring of the performance of systems, components, and structures (SSCs) for which Frequencies are changed to assure that reduced testingdoes not adversely impact the SSCs. In addition, the NEI 04-10, Revision 1methodology satisfies the five key safety principles specified in Regulatory Guide 1.177,"An Approach for Plant-Specific, Risk-Informed Decisionmaking:

Technical Specifications,"

dated August 1998 (ADAMS Accession No. ML003740176)

(Ref. 6),relative to changes in Surveillance Frequencies.

Therefore, the proposed relocation ofthe Ginna plant-specific Surveillance Frequencies is consistent with TSTF-425 and withthe NRC staff's Model Safety Evaluation dated July 6, 2009 (74 FR 31996).3.0 REGULATORY ANALYSIS3.1 No Significant Hazards Consideration Exelon has reviewed the proposed no significant hazards consideration (NSHC) determination published in the Federal Register dated July 6, 2009 (74 FR 31996). Exelon has concluded thatthe proposed NSHC presented in the Federal Register notice is applicable to Ginna, and isprovided as Attachment 6 to this amendment

request, which satisfies the requirements of 10CFR 50.91 (a), "Notice for public comment; State consultation" (Ref. 7).3.2 Applicable Regulatory Requirements A description of the proposed changes and their relationship to applicable regulatory requirements is provided in TSTF-425, Revision 3 (ADAMS Accession No. ML090850642) andthe NRC staff's Model Safety Evaluation published in the Notice of Availability dated July 6,2009 (74 FR 31996). Exelon has concluded that the relationship of the proposed changes tothe applicable regulatory requirements presented in the Federal Register notice is applicable toGinna.3.3 Precedence This application is being made in accordance with the TSTF-425, Revision 3 (ADAMSAccession No. ML090850642).

Exelon is not proposing significant variations or deviations fromthe TS changes described in TSTF 425 or in the content of the NRC staff's Model SafetyEvaluation published on July 6, 2009 (74 FR 31996). The NRC has previously approved LAR -Adoption of TSTF-425, Revision 3 Attachment 1Docket No. 50-244 Page 4 of 5amendments to the TS as part of the pilot process for TSTF-425, including but not limited toAmendment Nos. 186 and 147 for Limerick Generating

Station, Amendment No.276 for OysterCreek Nuclear Power Station dated September 27, 2010; Amendment Nos. 200 and 201 forDiablo Canyon Power Plant, Units 1 and 2, respectively, dated October 30, 2008; andAmendment Nos. 188 and 175 for South Texas Project, Units 1 and 2, respectively, datedOctober 31, 2008. The subject License Amendment Request proposes to relocate periodicsurveillance frequencies to a licensee-controlled program and add a new program (theSurveillance Frequency Control Program) to the Administrative Controls section of TS inaccordance with TSTF-425 and as discussed in the previously approved amendments.

3.4 Conclusions

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposedmanner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and securityor to the health and safety of the public.4.0 ENVIRONMENTAL CONSIDERATION Exelon has reviewed the environmental consideration included in the NRC staff's Model SafetyEvaluation published in the Federal Register on July 6, 2009 (74 FR 31996). Exelon hasconcluded that the staff's findings presented therein are applicable to Ginna, and thedetermination is hereby incorporated by reference for this application.

LAR -Adoption of TSTF-425, Revision 3 Attachment 1Docket No. 50-244 Page 5 of

55.0 REFERENCES

1. TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control -RITSTFInitiative 5b," March 18, 2009 (ADAMS Accession Number: ML090850642).

2. NRC Notice of Availability of Technical Specification Improvement to Relocate Surveillance Frequencies to Licensee Control -Risk-Informed Technical Specification Task Force(RITSTF)

Initiative 5b, Technical Specification Task Force -425, Revision 3, published onJuly 6, 2009 (74 FR 31996).3. NEI 04-10, Revision 1, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies,"

April 2007 (ADAMS Accession Number:ML071360456).

4. Regulatory Guide 1.200, Revision 1, "An Approach for Determining the Technical Adequacyof Probabilistic Risk Assessment Results for Risk-Informed Activities,"

January 2007(ADAMS Accession Number: ML070240001).

5. 10 CFR 50.90, "Application for amendment of license, construction permit, or early sitepermit."6. Regulatory Guide 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking:

Technical Specifications,"

dated August 1998 (ADAMS Accession No. ML003740176).

7. 10 CFR 50.91(a),

"Notice for public comment; State consultation."

ATTACHMENT 2License Amendment RequestR. E. Ginna Nuclear Power PlantDocket No. 50-244Application 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)Documentation of PRA Technical Adequacy LAR -Adoption of TSTF-425, Revision 3 Attachment 2Docket No. 50-244 Page i of iDocumentation of PRA Technical AdequacyTABLE OF CONTENTSSection Paqe1 .0 O v e rv ie w .........................................................................................................................

12.0 Technical Adequacy of the PRA Model .......................................................................

22.0.1 PRA Maintenance and Update .......................................................................

32.0.2 Plant Changes not yet Incorporated into the PRA Model ................................

42.0.3 Applicability of Peer Review Findings and Observations

..................................

42.0.4 Consistency with Applicable PRA Standards

..................................................

52.0.5 Identification of Key Assumptions

...................................................................

52.1 External Events Considerations

..................................................................................

52 .2 S u m m a ry .........................................................................................................................

72 .3 R efe re nce s ......................................................................................................................

7 LAR -Adoption of TSTF-425, Revision 3 Attachment 2Docket No. 50-244 Page 1 of 29Documentation of PRA Technical Adequacy1.0 OverviewThe implementation of the Surveillance Frequency Control Program (also referred to asTechnical Specifications Initiative 5b) at the Ginna Nuclear Power Plant will follow the guidanceprovided in NEI 04-10, Revision 1 [Ref. 11 in evaluating proposed surveillance test interval (STI;also referred to as "surveillance frequency")

changes.The following steps of the risk-informed STI revision process are common to proposed changesto all STIs within the proposed licensee-controlled program.* Each STI revision is reviewed to determine whether there are any commitments made tothe NRC that may prohibit changing the interval.

If there are no related commitments, orthe commitments may be changed using a commitment change process based on NRCendorsed

guidance, then evaluation of the STI revision would proceed.

If a commitment exists and the commitment change process does not permit the change, then the STIrevision would not be implemented.

" A qualitative analysis is performed for each STI revision that involves severalconsiderations as explained in NEI 04-10, Revision 1.* Each STI revision is reviewed by an Expert Panel, referred to as the Integrated Decisionmaking Panel (IDP), which is normally the same panel as is used for Maintenance Ruleimplementation, but with the addition of specialists with experience in surveillance testsand system or component reliability.

If the IDP approves the STI revision, the change isdocumented and implemented, and available for audit by the Nuclear Regulatory Commission (NRC). If the IDP does not approve the STI revision, the STI value is leftunchanged.

" Performance monitoring is conducted as recommended by the IDP. In some cases, noadditional monitoring may be necessary beyond that already conducted under theMaintenance Rule. The performance monitoring helps to confirm that no failuremechanisms related to the revised test interval become important enough to alter theinformation provided for the justification of the interval changes.* The IDP is responsible for periodic review of performance monitoring results.

If it isdetermined that the time interval between successive performances of a surveillance test is a factor in the unsatisfactory performances of the surveillance, the IDP returns theSTI back to the previously acceptable STI.* In addition to the above steps, the Probabilistic Risk Assessment (PRA) is used whenpossible to quantify the effect of a proposed individual STI revision compared toacceptance criteria in NEI 04-10. Also, the cumulative impact of all risk-informed STIrevisions on all PRAs (i.e., internal events, external events and shutdown) is alsocompared to the risk acceptance criteria as delineated in NEI 04-10.For those cases where the STI cannot be modeled in the plant PRA (or where a particular PRAmodel does not exist for a given hazard group), a qualitative or bounding analysis is performed to provide justification for the acceptability of the proposed test interval change.The NEI 04-10 [Ref. 1 methodology endorses the guidance provided in Regulatory Guide (RG)1.200, Revision 1 [Ref. 21, "An Approach for Determining the Technical Adequacy ofProbabilistic Risk Assessment Results for Risk-Informed Activities."

The guidance in RG-1.200 LAR -Adoption of TSTF-425, Revision 3 Attachment 2Docket No. 50-244 Page 2 of 29indicates that the following steps should be followed when performing PRA assessments (NOTE: Because of the broad scope of potential Initiative 5b applications and the fact that therisk assessment details will differ from application to application, each of the issuesencompassed in Items 1 through 3 below will be covered with the preparation of each individual PRA assessment made in support of the individual STI interval requests.

Item 3 satisfies one ofthe requirements of Section 4.2 of RG 1.200. The remaining requirements of Section 4.2 areaddressed by Item 4 below.):1. Identify the parts of the PRA used to support the application

-SSCs, operational characteristics affected by the application and how these areimplemented in the PRA model-A definition of the acceptance criteria used for the application

2. Identify the scope of risk contributors addressed by the PRA model-If not full scope (i.e., internal and external),

identify appropriate compensatory measures or provide bounding arguments to address the risk contributors notaddressed by the model.3. Summarize the risk assessment methodology used to assess the risk of the application

-Include how the PRA model was modified to appropriately model the risk impact ofthe change request.4. Demonstrate the Technical Adequacy of the PRA-Identify plant changes (design or operational practices) that have been incorporated at the site, but are not yet in the PRA model and justify why the change does notimpact the PRA results used to support the application.

-Document peer review findings and observations that are applicable to the parts ofthe PRA required for the application, and for those that have not yet been addressed justify why the significant contributors would not be impacted.

-Document that the parts of the PRA used in the decision are consistent withapplicable standards endorsed by the Regulatory Guide ( RG-1.200 Revision 1 wasused for the Ginna Internal Events PRA Peer review).

Provide justification to showthat where specific requirements in the standard are not adequately met, it will notunduly impact the results.-Identify key assumptions and approximations relevant to the results used in thedecision-making process.The purpose of the remaining portion of this attachment is to address the requirements identified in Item 4 above.2.0 Technical Adequacy of the PRA ModelThe GN1 14A-W version of the Ginna PRA model is the most recent evaluation of internal eventrisks. The Ginna PRA modeling is highly detailed, including a wide variety of initiating events,modeled systems, operator

actions, and common cause events. The PRA model quantification process used for the Ginna PRA is based on the event tree / fault tree methodology, which is awell-known methodology in the industry.

Exelon Generation

Company, LLC (Exelon) employs a multi-faceted approach to establishing and maintaining the technical adequacy and plant fidelity of the PRA models for all operating LAR -Adoption of TSTF-425, Revision 3 Attachment 2Docket No. 50-244 Page 3 of 29Exelon nuclear generation sites. This approach includes both a proceduralized PRAmaintenance and update process, and the use of self-assessments and independent peerreviews.

Prior to joining the Exelon nuclear fleet in 2014, comparable practices were in placewhen Ginna was owned and operated by Constellation Energy Nuclear Group (CENG).Because of the similarities between the CENG and Exelon practices, no additional discussion specifically regarding the legacy CENG approach will be provided.

The following information describes the Exelon approach (and by extension the CENG approach) to PRA modelmaintenance, as it applies to the Ginna PRA.2.0.1 PRA Maintenance and UpdateThe Exelon risk management process ensures that the applicable PRA model is an accuratereflection of the as-built and as-operated plants. This process is defined in the Exelon RiskManagement

program, which consists of a governing procedure (ER-AA-600, "RiskManagement")

and subordinate implementation training and reference materials (T&RM's).

• Exelon T&RM ER-AA-600-1015, "Full Power Internal Event (FPIE) PRA Model Update,"delineates the responsibilities and guidelines for updating the full power internal eventsPRA models at all operating Exelon nuclear generation sites.* ER-AA-600-1061 "Fire PRA Model Update and Control" delineates the responsibilities and guidelines for updating the station fire PRA.The overall Exelon Risk Management

program, including ER-AA-600-1015 and ER-AA-600-1061, define the process for: implementing regularly scheduled and interim PRA modelupdates; for tracking issues identified as potentially affecting the PRA models (e.g., due tochanges in the plant, industry operating experience, etc.); and for controlling the model andassociated computer files. To ensure that the current PRA model remains an accurate reflection of the as-built, as-operated plants, the following activities are routinely performed:

" Design changes and procedure changes are reviewed for their impact on the PRAmodel.* New engineering calculations and revisions to existing calculations are reviewed for theirimpact on the PRA model.* Maintenance unavailabilities are captured, and their impact on CDF is trended.* Plant specific initiating event frequencies, failure rates, and maintenance unavailabilities are updated during each model update.In addition to these activities, Exelon risk management procedures provide the guidance forparticular risk management maintenance activities.

This guidance includes:

• Documentation of the PRA model, PRA products, and bases documents.

" The approach for controlling electronic storage of Risk Management (RM) productsincluding PRA update information, PRA models, and PRA applications.

• Guidelines for updating the full power, internal events PRA models for Exelon nucleargeneration sites." Guidance for use of quantitative and qualitative risk models in support of the On-LineWork Control Process Program for risk evaluations for maintenance tasks (corrective maintenance, preventive maintenance, minor maintenance, surveillance tests and LAR -Adoption of TSTF-425, Revision 3 Attachment 2Docket No. 50-244 Page 4 of 29modifications) on systems, structures, and components (SSCs) within the scope of theMaintenance Rule (10 CFR 50.65(a)(4)).

An application specific update of the PRA model was completed in the 4th quarter of 2014 tosupport an update of the Mitigating System Performance Indicator (MSPI) application.

Exelonwill be performing a Full Power Internal Events (FPIE) model update to the Ginna PRA in 2015.As indicated previously, RG 1.200 also requires that additional information be provided as partof the LAR submittal to demonstrate the technical adequacy of the PRA model used for the riskassessment.

Each of these items (plant changes not yet incorporated in to the PRA model,relevant peer review findings, consistency with applicable PRA Standards, and the identification of key assumptions) will be discussed in turn.2.0.2 Plant Changes Not Yet Incorporated into the PRA ModelEach Exelon station maintains an updating requirements evaluation (URE) database to track allenhancements, corrections, and unincorporated plant changes.

During the normal screening conducted as part of the plant change process, if a potential model update is identified a newURE database item is created.

Depending on the potential impact of the identified change, therequirements for incorporation will vary.As part of the PRA evaluation for each STI change request, a review of open items in the UREdatabase for GINNA will be performed and an assessment of the impact on the results of theapplication will be made prior to presenting the results of the risk analysis to the IDP. If a non-trivial impact is expected, then this may include the performance of additional sensitivity studiesor PRA model changes to confirm the impact on the risk analysis.

2.0.3 Applicability of Peer Review Findings and Observations A PRA model update was completed in 2009, resulting in the Ginna PRA Model 6.5. The GinnaPRA model was revised to meet RG 1.200, revision 1, guidance and comply with theASME/ANS PRA Standard RA-Sc-2007[Ref.

3.This model was peer reviewed under the auspices of the PWR Owners Group (PWROG) in the2nd quarter of 2009 [Ref. 71. This peer review was performed following NEI 05-04 [Ref. 51, andNEI 00-002 [Ref. 61. This peer review included an assessment of the PRA model maintenance and update process.Since the 2008 peer review, an application specific PRA model update was completed in 2012to support implementation of NFPA-805.

As part of the development of this model a peerreview of the fire PRA was conducted in June of 2012 [Ref. 8]. This peer review used NEI-07-12 [Ref. 9] to evaluate the model against the ASME PRA Standard (ASME/ANS RA-Sa-2009)

[Ref. 10] along with the NRC clarifications provided in Regulatory Guide 1.200, Rev. 2 [Ref. 22].Since the 2012 peer review, several updates to the Ginna PRA have taken place. Anapplication specific model update was completed in December 2014 to support the Mitigating System Performance Indicator (MSPI) process.

The latest model is the GN1 14A-W. This modelincludes the addition of two diesel generators for providing an alternate source of power to theStandby AFW (SAFW) Pumps and a condensate storage tank to provide a dedicated source ofwater to the SAFW Pumps.

LAR -Adoption of TSTF-425, Revision 3 Attachment 2Docket No. 50-244 Page 5 of 292.0.4 Consistency with Applicable PRA Standards As indicated above there have been two relevant peer reviews conducted on the current PRAmodel.* The 2009 peer review for the PRA ASME model update identified 309 Supporting Requirements (SR) applicable to the Ginna PRA. Of these 29 were not met, 2 metcapability category (cc) 1, 13 partially met cc 2, 31 met cc 2, 22 partially met cc 3, 14met cc 3, and 198 fully met all capability requirements.

There were 24 findings andobservations (F&O's) issued to address the identified gaps to compliance with the PRAstandard.

Subsequent to the peer review, 13 of the findings have been addressed and11 are still open pending the next model update. The F&O's are listed in Table 2-1which includes what, if any impact, there may be to the assessment of STIs for the 5binitiative.

• The 2012 fire PRA peer review for the PRA ASME model update identified 183Supporting Requirements (SR) to be reviewed for the Ginna PRA. Of these 2 were notmet, 2 met capability category (cc) 1, 8 partially met cc 2, 17 met cc 2, 13 partially metcc 3, 7 met cc 3, and 118 fully met all capability requirements and 16 were notapplicable.

There were 19 findings and 22 suggestions issued to address potential gapsto compliance with the PRA standard.

There were 3 Best Practices.

All of the findingsfrom the fire PRA peer review have since been closed. As the results of this peer reviewhave already been communicated to the NRC as part of the NFPA-805 submittal

[Ref.12] and subsequent requests for additional information (RAI), these will not becatalogued in this document.

All remaining gaps will be reviewed for consideration during the 2015 model update but arejudged to have low impact on the PRA model or its ability to support a full range of PRAapplications.

The remaining gaps are documented in the URE database so that they can betracked and their potential impacts accounted for in applications where appropriate.

Each item will be reviewed as part of each STI change assessment that is performed and anassessment of the impact on the results of the application will be made prior to presenting theresults of the risk analysis to the IDP. If a non-trivial impact is expected, then this may includethe performance of additional sensitivity studies or PRA model changes to confirm the impact onthe risk analysis.

2.0.5 Identification of Key Assumptions The overall Initiative 5b process is a risk-informed process with the PRA model results providing one of the inputs to the IDP to determine if an STI change is warranted.

The methodology recognizes that a key area of uncertainty for this application is the standby failure rate utilized inthe determination of the STI extension impact. Therefore, the methodology requires theperformance of selected sensitivity studies on the standby failure rate of the component(s) ofinterest for the STI assessment.

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 2.2.3above for each STI change assessment will be documented and included in the results of therisk analysis that goes to the IDP.2.1 External Events Considerations The NEI 04-10 [Ref. 1] methodology allows for STI change evaluations to be performed in theabsence of quantifiable PRA models for all external hazards.

For those cases where the STI LAR -Adoption of TSTF-425, Revision 3 Attachment 2Docket No. 50-244 Page 6 of 29cannot be modeled in the plant PRA (or where a particular PRA model does not exist for a givenhazard group), a qualitative or bounding analysis is performed to provide justification for theacceptability of the proposed test interval change.External hazards were evaluated in the GINNA Individual Plant Examination of External Events(IPEEE) submittal in response to the NRC IPEEE Program (Generic Letter 88-20, Supplement

4) [Ref. 41. The IPEEE Program was a one-time review of external hazard risk and was limitedin its purpose to the identification of potential plant vulnerabilities and the understanding ofassociated severe accident risks.The primary areas of external event evaluation at GINNA were internal fires and seismic risk.The internal fire events were addressed by using a combination of the EPRI Fire InducedVulnerability Evaluation (FIVE) methodology

[Ref. 141 and fire PSA. The results of the FireAnalysis are documented in the R.E. Ginna Nuclear Power Plant IPEEE Fire Analysistransmitted to the NRC in June 1998 [Ref. 131. The seismic evaluations were performed inaccordance with Generic Implementation Procedure (GIP) developed by the SeismicQualification Utility Group (SQUG) of which Ginna was a member. The GIP provided plants amethod for addressing Unresolved Safety Issue A-46 (Verification of Seismic Adequacy ofMechanical and Electrical Equipment in Operating Reactors (USI A-46). Beyond this, Ginnaperformed a reduced-scope IPEEE for seismic events to close out IPEEE for Seismic Events.The Ginna USI A-46 Seismic Evaluation Report and the IPEEE Seismic Evaluation Report weretransmitted to the NRC in January 1997 [Ref. 151 However, there are no comprehensive CDFand LERF values available from the seismic IPEEE report to support the STI risk assessments.

High Winds, External Floods and Transportation Accidents were reviewed against the StandardReview Plan (SRP) as Ginna was one of the eleven participants in the NRC's Systematic Evaluation Program (SEP). Following plant modifications, it was determined that the Ginnaplant met the Standard Review Plan criteria.

Based on the NRC Safety Evaluation Reports(SERs) for Ginna's SEP results, no further submittals for GL 88-20 Supplement 4 werewarranted for high winds, external floods, or transportation accidents.

Since the performance of the IPEEE, Ginna has submitted a License Amendment Request forconversion from appendix R compliance to NFPA-805 for fire protection

[Ref. 121. Pursuant tothis change, a fire PRA has been created and implemented at GINNA. This Fire PRA modelwas created under the auspices of NUREG/CR-6850

[Ref. 161 and has undergone PWROGpeer review (completed August 2012) [Ref. 111. The Ginna Fire PRA was developed using theNational Institute of Standards and Technology (NIST) Consolidated Model of Fire and SmokeTransport (CFAST) Methodology

[Ref. 181; the Fire Dynamics Simulator, also developed byNIST; NUREG-1805 Fire Dynamics Tools (FDTs) computational Spreadsheets

[Ref. 21];EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities

[Ref. 161 and theassociated NUREG/CR-6850 Frequently Asked Questions (FAQ) Process [Ref. 201; Fire EventsDatabase

[Ref. 191 and plant specific data. This fire PRA has numerous capabilities notconsidered in the IPEEE fire PRA model including explicit analysis of all risk significant fireareas such as the main control room (MCR) and Relay Room (RR). Multiple spuriousoperations (MSO) considerations are also included.

The ignition frequencies for all fire areaswere developed using the guidance in NUREG/CR-6850

[Ref. 16] and also incorporate revisedguidance for ignition frequencies

[Ref. 171.As stated earlier, the NEI 04-10 [Ref. 1 methodology allows for STI change evaluations to beperformed in the absence of quantifiable PRA models for all external hazards.

Therefore, inperforming the assessments for the other hazard groups, a qualitative or a bounding approachwill be utilized in most cases. Where applicable, the results of any STI change will be evaluated LAR -Adoption of TSTF-425, Revision 3 Attachment 2Docket No. 50-244 Page 7 of 29against this model to ensure there is no undue risk associated with a given STI change. Thisapproach is consistent with the accepted NEI 04-10 methodology.

2.2 SummaryThe GINNA PRA technical capability evaluations and the maintenance and update processes described above provide a robust basis for concluding that the GINNA PRA is suitable for use inrisk-informed processes such as that proposed for the implementation of a Surveillance Frequency Control Program.

Also, in addition to the standard set of sensitivity studies requiredper the NEI 04-10 [Ref. 1 methodology, open items for changes at the site and remaining gapsto specific requirements in the PRA standard will be reviewed to determine which, if any, wouldmerit application-specific sensitivity studies in the presentation of the application results.2.3 References

[1] Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control ofSurveillance Frequencies, Industry Guidance

Document, NEI 04-10, Revision 1, April2007.[2] Regulatory Guide 1.200, An Approach for Determining the Technical Adequacy ofProbabilistic Risk Assessment Results for Risk Informed Activities, Revision 1, January2007.[3] American Society of Mechanical Engineers, Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME RA-Sc-2007, New York, New York, July2007.[4] NRC Generic Letter 88-20, "Individual Plant Examination of External Events (IPEEE) forSevere Accident Vulnerabilities

-10 CFR 50.54(f),

Supplement 4," June 28, 1991.[5] NEI 05-04, Process for Performing Internal Events PRA Peer Reviews Using theASME/ANS PRA Standard[6] NEI 00-002 Probabilistic Risk Assessment (PRA) Peer Review Process Guidance, Revision 1, Nuclear Energy Institute (NEI), Washington, DC, May 2006[7] RG 1.200 PRA Peer Review Against the ASME PRA Standard Requirements for R. E.Ginna Station Probabilistic Risk Assessment, Project PA-RMSC-0386, August 2009.[8] Fire PRA Peer Review Against the Fire PRA Standard Supporting Requirements fromSection 4 of the ASME/ANS Standard for L1/LERF PRA for NPP Applications for theGinna Station Fire PRA, Project PA-RMSC-0403, August 2012.[9] NEI 07-12, Fire Probabilistic Risk Assessment (FPRA) Peer Review Process Guidelines, Revision 1, Nuclear Energy Institute (NEI), Washington, DC, June 2010.[10] American Society of Mechanical Engineers, Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME PRA Standard (ASME/ANS RA-Sa-2009),

New York, New York, July 2009.[11] Fire PRA Peer Review Against the Fire PRA Standard Supporting Requirements fromSection 4 of the ASME/ANS Standard for Level l/Large Early Release Frequency Probabilistic Risk Assessments for Nuclear Power Plant Applications for the GinnaStation Fire Probabilistic Assessment, August 2012.[12] Letter from Mr. Joseph E. Pacher (Ginna LLC) to Document Control Desk (NRC), datedMarch 28, 2013, License Amendment Request Pursuant to 10 CFR 50.90: Adoption ofNFPA 805, Performance-Based Standard for Fire Protection for Light Water ReactorElectric Generating Plants (ADAMS Accession No. ML 13093A064).

[13] Letter from Robert C. Mecredy (RG&E) to Guy S. Vissing (NRC), 1. Ginna Station FireIPEEE, RE Ginna Nuclear Power Plant; 2 Hydrogen Storage Facility, dated June 30,1998.

LAR -Adoption of TSTF-425, Revision 3 Attachment 2Docket No. 50-244 Page 8 of 29[14] Fire-Induced Vulnerability Evaluation (FIVE) Methodology Plant Screening Guide, EPRITR-100370, Electric Power Research Institute, Final Report, April 1992.[15] Letter from RC Mecredy (RG&E) to Guy S. Vissing (NRC), Resolution of GL 87-02,Supplement 1 and GL 88-20, Supplements 4 and 5 (Seismic Event Only) RG&E Corp,R.E. Ginna Nuclear Power Plant, dated January 31, 1997.[16] EPRI/NRC-RES, Fire PRA Methodology for Nuclear Power Facilities, EPRI 1011989,NUREG/CR-6850, Final Report, September 2005.[17] Fire Probabilistic Risk Assessment Methods Enhancements Supplement 1 toNUREG/CR-6850 and EPRI 1011989, EPRI 1019259, Electric Power Research Institute, December 2009.[18] National Institute of Standards and Technology's (NIST) Consolidated Model of FireGrowth and Smoke Transport (CFAST) Version (6) (Jones et al., 2004)[19] NSAC/179L, Electric Power Research Institute, Fire Events Database for U.S. NuclearPower Plants, Rev. 1, January, 1993.[20] Letter from John A. Grobe (NRR) to Alexander Marion (NEI), dated June 1, 2009, PathForward in Resolving Frequently Asked Questions Related to NUREG/CR-6850.

Accession No. ML090920045.

[21] NUREG-1 805, Supplement 1, Volumes 1 & 2, Fire Dynamics Tools (FDTs) Quantitative Fire Hazard Analysis Methods for the U.S. Nuclear Regulatory Commission FireProtection Inspection Program.[22] Regulatory Guide 1.200, An Approach for Determining the Technical Adequacy ofProbabilistic Risk Assessment Results for Risk Informed Activities, Revision 2, March,2009 LAR -Adoption of TSTF-425, Revision 3Docket No. 50-244Attachment 2Page 9 of 29Table 2-1 Internal Events PRA Peer Review -FindingsSR Topic Status Finding/Observation Disposition Impact to TSTF-425IE-C12[2005:IE-CIO]URE 845COMPARE results and EXPLAINdifferences in the initiating eventanalysis with generic data sources toprovide a reasonableness check of theresults.Open F&O IE-CIO-01:

The Ginna Initiating EventNotebook (Gl-IE-0001, Rev. 1) Section 4.3provides a cross-reference between theGinna Initiating Events and the NRC Ratesnof Initiating Events in table 4-7. Table 4-7cross-reference includes columns forNUREG/CR-5750 Category and NP-2230EPRI/NUREG/CR-3862 PWR Category.

Documentation only: Providecomparison of core damage resultsbased on generic data cross-referenced in Table 4-7.This item is a documentation issue. No impact on TSTF-425analysis.

Table 4-8 provides a cross-reference between Ginna and similar PWR plants(Point Beach, Prairie Island, and Kewaunee).

An explanation of differences in Initiating Events between Ginna and similar PWRs iscontained in the PRA Quantification (QU)Notebook (G1-QU-0001, Rev. 0) Table 4-5"Comparison of Ginna Core Damage Resultsto 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 QUNotebook (G1-QU-0001, Rev. 0).

LAR -Adoption of TSTF-425, Revision 3Docket No. 50-244Attachment 2Page 10 of 29Table 2-1 Internal Events PRA Peer Review -FindingsSR Topic Status Finding/Observation Disposition Impact to TSTF-425IE-C15 CHARACTERIZE the uncertainty in the Open F&O IE-C13-01:

Gi-IE-000

, PRA Documentation only: Include error This item is a documentation

[2005: initiating event frequencies and INITIATING EVENT (IE) NOTEBOOK, Section factors and brief discussion about IE issue and IE frequency IE-C13] PROVIDE mean values for use in the 5 documents assumptions and sources of frequency uncertainty, distribution evaluation.

Changesquantification of the PRA results.

uncertainty.

However, section 5 does not will not impact the TSTF-425provide or reference the parametric analysisuncertainty initiating event datadistribution.

For example, the distribution for TIGRLOSP is identified in the CAFTAmodel, newauto_65a-w-FId.caf, has havingan EF of 7.39. However, no documentation for the error factor could be found.Therefore, this SR is not met.

LAR -Adoption of TSTF-425, Revision 3Docket No. 50-244Attachment 2Page 11 of 29Table 2-1 Internal Events PRA Peer Review -FindingsSR Topic Status Finding/Observation Disposition Impact to TSTF-425SC-A2 SPECIFY the plant parameters (e.g., highestnode temperature, core collapsed liquidlevel) and associated acceptance criteria(e.g., temperature limit) to be used indetermining core damage. Select theseparameters such that determination ofcore damage is as realistic as practical, in amanner -consistent with current bestpractice.

DEFINE computer code-predicted acceptance criteria with sufficient marginon the code-calculated values to allow forlimitations of the code, sophistication ofthe models, and uncertainties in theresults, in a manner consistent with therequirements specified under HLR-SC-B.

Examples of measures for core damagesuitable for Capability Category Il/111, thathave been used in PRAs, include (a)collapsed liquid level less than 1?3 coreheight or code-predicted peak coretemperature

>2,500°F (BWR) (b) collapsed liquid level below top of active fuel for aprolonged period, or code-pre-dicted corepeak node temperature

>2,200'F using acode with detailed core modeling; or code-predicted core peak node temperature

>1,800°F using a code with simplified (e.g.,single-node core model, lumped para-meter) core modeling; or code-predicted core exit temperature

>1,200°F for 30 minusing a code with simplified core modeling(PWR).OpenF&O SC-A2-01:

The definition of coredamage documented in the Ginna-AS-Notebook-Rev-1 Section 2.2 is consistent with the examples of measures for coredamage suitable for Capability Category I asdefined in NUREG/CR-4550.

For Category IIGinna could use the code-predicted coreexit temperature

>1,200°F for 30 min usingPCTRAN (code with simplified coremodeling (PWR)).We agree with the peer reviewers thatthe approach taken in the Ginna PRA isoverly conservative and not consistent with the requirements of Category II.The peer reviewers suggested using acore exit temperature of 1200°F for 30minutes as the criterion for coredamage, but we would recommend using either that criterion or a peakcore node temperature of 18007F.Based on a review of the PCTRANresults, it is likely that the 18007F peakcore temperature would be reachedearlier than the time at which the coreexit temperature would be greaterthan 1200°F for 30 minutes.Over the typical complete loss ofdecay heat removal timingsuccess criteria, the delta timebetween core uncovery and CETtemperatures reach 12007F for30 minutes or 1800° peak centerline is fairly small. As such, thetiming benefit is not expected tobe large except for the fastmoving events such as largebreak LOCAs. For these events,we use the UFSAR successcriteria.

Although this is notexpected to be a significant effect, we do remain aconservative CAT I. Therefore, the model used for TSTS-425analysis may be conservative.

LAR -Adoption of TSTF-425, Revision 3Docket No. 50-244Attachment 2Page 12 of 29Table 2-1 Internal Events PRA Peer Review -FindingsSR Topic Status Finding/Observation Disposition Impact to TSTF-425SC-A4 IDENTIFY mitigating systems that are Complete F&O SC-A4-02

-Operator action Add RCHFDX1BAF to the Event Tree No impact to TSTF 425. Actionshared between units, and the RCHFDX1BAF (operator fails to align BAF TIU, as appropriate, placed in Event Tree TIU logicmanner 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 appearsthat this event should be added in Eventexperience a common initiating event Tree TIU Sequence 5 Failures under gate(e.g., LOOP). TLFB.This is an omission in the model and mayaffect CDF and LERF.SY-A10 INCORPORATE the effect of variable Complete SY-Al 1-01 -Gate TLFBHRD1 input to gate Review the Bleed and Feed modeling No impact as the Finding has[SY-A11 success criteria (i.e., success criteria TL_FB for failure of Bleed and Feed modelsthat change as a function of plant success as requiring 1 SI pump and 1 PORV. to ensure the system failures been addressed and the logic has-2005] status) into the system modeling.

The logic does not include 75 gpm charging appropriately reflect the success been updated and documented Example causes of variable system flow which is noted in the Success Criteria

criteria, in the Success Criteria Notebook.

success criteria are notebook as required to support single PORV(a) different accident scenarios, success.

This was confirmed throughDifferent success criteria are required discussion with Ginna PRA personnel.

for some systems to mitigate different accident scenarios (e.g., the number of The omission of a needed mitigating systempumps required to operate in some for support of the Bleed and Feed functionsystems is dependent upon the may underestimate the importance of thesemodeled initiating event), sequences for applications.

(b) dependence on other components.

Success criteria for some systems arealso dependent on the success ofanother component in the system (e.g.,operation of additional pumps in somecooling water systems is required ifnoncritical loads are not isolated).

(c) time dependence.

Success criteriafor some systems are time-dependent (e.g., two pumps arerequired to provide the needed flowearly following an accident initiator, butonly one is required for mitigation laterfollowing the accident).

(d) sharing of a system between units.Success criteria may be affected whenboth units are challenged by the sameinitiating event (e.g., LOOP).

LAR -Adoption of TSTF-425, Revision 3Docket No. 50-244Attachment 2Page 13 of 29Table 2-1 Internal Events PRA Peer Review -FindingsSR Topic Status Finding/Observation Disposition Impact to TSTF-425SY-A14[SY-A132005]When identifying the failures in SY-Al 1 CompleteINCLUDE consideration of all failuremodes, consistent with available data and model level ofdetail, except where excluded usingthe criteria inSY-A 15.For example,(a) active component fails to start(b) active component fails to continueto run(c) failure ot a closed component toopen(d) failure of a closed component toremain closed(e) failure of an open component toclose(f) failure of an open component toremain open(g) active component spuriousoperation (h) plugging of an active or passivecomponent SY-A13-02

-Inconsistencies existed in thesystem modeling of the city water system.Where used to support the GE-Betz system,a basic event for unavailability of city waterdue to grid LOOP was added (basic eventCDAACITYWATER).

This same event was notadded to the city water modeling forsupport of the SAFW system.Review the need to add theunavailability event in the SAFWSystem.No impact to TSTF 425. Thedependencies for SAFW havebeen updated in the Ginna PRA.(i) leakage of an active or passivecomponent

6) rupture of an active or passivecomponent (k) internal leakage of a component

(/) internal rupture of a component (in) failure to provide signal/operate (e.g., instrumentation)

(n) spurious signal/operation (o) pre-initiator human failure events(see SY-A 16)(p) other failures of a component toperform its required function62 LAR -Adoption of TSTF-425, Revision 3Docket No. 50-244Attachment 2Page 14 of 29Table 2-1 Internal Events PRA Peer Review -FindingsSR Topic Status Finding/Observation Disposition Impact to TSTF-425SY-A19[SY-A1820051In the systems model, INCLUDE out-of-service unavailability forcomponents in the system model,unless screened, in a mannerconsistent with the actual practices andhistory of theplant for removing equipment fromservice.(a) INCLUDE(1) unavailability caused by testingwhen a component or system train isreconfigured from its required accidentmitigating position such that thecomponent cannot function as required(2) maintenance events at the trainlevel when procedures require isolating the entire trainfor maintenance (3) maintenance events at a sub-train level (i.e., between tagout boundaries, such as a functional equipment group)when directed by procedures (b) Examples of out-of-service unavailability to be modeled are asfollows:(1) train outages during a work windowfor preventive/corrective maintenance (2) a functional equipment group (FEG)removed from service forpreventive/corrective maintenance (3) a relief valve taken out of serviceOpen SY-A18-01

-Ginna PRA System Notebooks provides a list of all the modeled T&M termsin Section 3.4.C. Section 2.9 of thenotebooks provide discussion of procedures and testing that result in Unavailability.

Thereview of these sections found no instances of simultaneous unavailability that can resultfrom planned activities.

However, the PRAengineer noted in a discussion that somesystems are shadowed in plannedmaintenance.

There is not a specificdiscussion on plant maintenance practices within the (a)(4) program that would result inplanned unavailability of multiple systemsOOS (i.e., EDG outages combined with AFWmotor driven pump outages to lower total riskas opposed to performing the workindependently),

or of planned activities resulting in multiple components OOS that donot violate technical specifications (e.g., twoAFW pumps in maintenance or an AFW andSAFW pump in maintenance at the sametime). If work is done in this manner, it maybe appropriate to account for theunavailability of both SSCs in a single term.Modeling of station maintenance practices that result in planned maintenance evolutions with more than a single PRA component OOS (i.e., shadowing equipment outages)can help to minimize the number of randomfailure sequences and ensure there is not"double counting" of unavailability in the PRA.Determine if any maintenance practices are performed that result inoverlapping unavailability of multiplesystems.

If it is determined thatsimultaneous unavailability is possible, model these occurrences as a singleunavailability event in the PRA orjustify why the unavailability is treatedas separate events and include this as apotential source of model uncertainty.

Also, consider adding a specificquestion to the system engineers' questionnaire for each system todetermine if there are plannedevolutions that represent simultaneous unavailability of multiple SSCs.If shadowed unavailability is in-fact significantly affecting theunavailability

numbers, then thiswould conservatively affect TSTF-425 analysis.

The mostsignificant unavailabilities arerelated to MSPI related functions which are less likely to includeconservative data.

LAR -Adoption of TSTF-425, Revision 3Docket No. 50-244Attachment 2Page 15 of 29Table 2-1 Internal Events PRA Peer Review -FindingsSR Topic Status Finding/Observation Disposition Impact to TSTF-425HR-G3 When estimating HEPs EVALUATE theimpact of the following plant-specific and scenario-specific performance shaping factors:(a) quality [type (classroom orsimulator) and frequency]

of theoperator training or experience (b) quality of the written procedures and administrative controls(c) availability of instrumentation needed to take corrective actions(d) degree of clarity of thecues/indications (e) human-machine interface (f) time available and time required tocomplete the response(g) complexity of the requiredresponse(h) environment (e.g., lighting, heat,radiation) under which the operator isworking(i) accessibility of the equipment requiring manipulation (j) necessity,

adequacy, andavailability of special tools, parts,clothing, etc.Complete F&O HR-G3-01:

Details regarding certainelements of the analysis were lacking in theHRA Calculator for a sufficient number ofHFEs to judge that this requirement was notmet. Evidence that the relevant aspectscited in the SR are addressed for each HFE iscritical to assuring that an appropriate analysis has been performed.

This isparticularly important in the case of HRA,for which the methods are lessstraightforward than they are for manyother parts of the PRA.Issue: In item (d) of CC II, Ill, clarifythat 'clarity' refers to the meaning ofthe cues, etc. In item (g) of CC II, IlI,clarify that complexity refers to bothdetermining the need for andexecuting the required response.

Resolution:

Cat I, II, and III: (d) degreeof clarity of the meaning of cues /indications (g) complexity of detection, diagnosis and decision-making, and executing the required response.

No impact to TSTF 425. The HRAshave been reviewed to ensurethe needed parameters for theevaluation have been populated.

CBDM is now used as a maxfunction of CBDT and HCR/ORE.RCHFDMAKEUP as a specificexample has a timing basis fromKey Input 51. When theannunciator model is used, thereis a clear discussion as to theapplicability.

LAR -Adoption of TSTF-425, Revision 3Docket No. 50-244Attachment 2Page 16 of 29Table 2-1 Internal Events PRA Peer Review -FindingsSR Topic Status Finding/Observation Disposition Impact to TSTF-425HR-I1 DOCUMENT the human reliability Complete F&O HR-I1-01:

The bulk of the Documentation only. Same issue as for No impact to TSTF 425. This itemanalysis 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 numerousreview. areas in which the documentation isincomplete.

The documentation shouldinclude a fuller discussion of the cues, basesfor timing, specific procedure steps, andother aspects that could affect the analyses.

QU-B5 Fault tree linking and some other Open F&O QU-B5-01:

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 iscircular logic that must be broken logic checks were performed on the Notebook Section 3.1 of the added to the tree the CAFTAbefore the model is solved. BREAK the integrated top logic model to ensure it did methodology used to address circular software identifies a circular logiccircular logic appropriately.

Guidance not exist. An example is listed, but there is logic, issue. The circular logic is brokenfor breaking logic loops is provided in no further discussion.

System notebooks by inserting as much of the logicNUREG/CR-2728

[2-13]. When reviewed generally state in Section 3.3 what clip into the tree as possible.

resolving circular logic, DO NOT was done when circular logic was identified, Providing more examples of thisintroduce unnecessary conservatisms but no discussion of the methodology was in the documentation is notor non-conservatisms.

provided nor how conservatisms or non- expected to affect the TSTF-425conservatisms are avoided.

No evidence evaluation.

that the required analysis was notperformed.

LAR -Adoption of TSTF-425, Revision 3Docket No. 50-244Attachment 2Page 17 of 29Table 2-1 Internal Events PRA Peer Review -FindingsSR Topic Status Finding/Observation Disposition Impact to TSTF-425LE-C2 INCLUDE realistic treatment of[2005: feasible operator actions following LE-C2a] the onset of core damage consistent with applicable procedures, e.g.,EOPs/SAMGs, proceduralized actions,or Technical Support Center guidance.

Open F&O LE-C2a-01:

It is conservative to NOTtake credit for operator actions post coredamage. This is a requirement of thestandard to move from Category I toCategory II.There are limited operator actions thatcould influence LERF at Ginna, so theeffect of such actions is not likely to besignificant.

Moreover, it is likely thatthere will not be a need for a CategoryII rating in this area to meet therequirements for most risk-informed applications.

One approach toreaching Category II would be toinclude post-core damage operatoractions in the PRA. It is also possiblethat simply identifying operatoractions and showing quantitatively that they will have a negligible impacton LERF will be sufficient to meet therequirements of Category II.There are limited operatoractions that could influence LERFat Ginna, so the effect of suchactions is not likely to besignificant.

If post-core-damage operator actions are credited, LERF estimates could be reduced,but the impact would beminimal.

The omission of theseoperator actions is conservative and does not adversely impactthe use of the model for TSTF-425 analysis.

LAR -Adoption of TSTF-425, Revision 3Docket No. 50-244Attachment 2Page 18 of 29Table 2-1 Internal Events PRA Peer Review -FindingsSR Topic Status Finding/Observation Disposition Impact to TSTF-425LE-C11 JUSTIFY any credit given for N/A F&O LE-C9a-01:

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 acontainment

failure, be impacted by containment failure.

This is by containment failure.

Since no such conservative CAT I.a requirement of the standard to move credit was taken, the SR should havefrom Category I to Category I1. been judged as not applicable (N/A).This is analogous to the assessment ofLE-C7 (old LE-C6) which was judged bythe peer reviewers as N/A becausehuman actions that support theaccident progression analysis were notcredited.

Also, note that, in the CalvertCliffs peer review, the peer reviewers judged this SR as N/A for the samereason. Only if post-containment failure equipment operations orhuman actions are modeled in thefuture would it be necessary to provideengineering analysis and writtenjustification as part of the PRAdocumentation.

Otherwise, noadditional work is needed.LE-C13 PERFORM a containment bypass N/A F&O LE-CIO-01:

Credit for scrubbing was Review the possible credit for release No impact to TSTF 425. A[2005 analysis in a realistic manner. JUSTIFY not taken. A sensitivity for impact of scrubbing to reduce LERF. sensitivity for impact ofLE-ClO] any credit taken for scrubbing (i.e., scrubbing was performed and it was scrubbing was performed and itprovide an engineering basis for the determined that the impact of not was determined that the impactdecontamination factor used). considering scrubbing is negligible.

This is a of not considering scrubbing isrequirement of the standard to move from negligible.

Category I to Category II.

LAR -Adoption of TSTF-425, Revision 3Docket No. 50-244Attachment 2Page 19 of 29Table 2-1 Internal Events PRA Peer Review -FindingsSR Topic Status Finding/Observation Disposition Impact to TSTF-425MU-D1 A PRA Configuration Control Programshall be inplace. It shall contain the following key elements:

(a) a process for monitoring PRAinputs and collecting new information (b) a process that maintains andupgrades the PRAto be consistent with the as-built, asoperated plant(c) a process that ensures that thecumulative impact of pendingchanges is considered when applyingthe PRA(d) a process that maintains configuration control of computercodes used to support PRAquantification (e) documentation of the Programcomplete F&O MU-DI-01

-PRA Configuration Controlprocedure (GNG-CM-1.01-3003)

Step 5.13provides guidance for updating risk-informed applications.

The processdescribed depends upon a databasemaintained by the Fleet PRA ServicesSupervisor to identify current livingapplications requiring change assessment other than those related to maintenance rule performance criteria.

No suchdatabase could be identified for Ginna.The CRMP database has a placeholder This configuration control issuefor a listing of PRA applications.

This has been addressed.

No impactportion of the database has been to TSTF 425.populated to ensure all applications requiring update following a modelrevision can be easily identified.

Without a current list of risk-informed applications, the maintenance and updateprocess is dependent upon the knowledge and experience of the staff to know whichapplications require update. This createsthe possibility that an application could bemissed in the update process.

LAR -Adoption of TSTF-425, Revision 3Docket No. 50-244Attachment 2Page 20 of 29Table 2-1 Internal Events PRA Peer Review -FindingsSR Topic Status Finding/Observation Disposition Impact to TSTF-425IFSO-A4 For each potential source of flooding complete F&O IF-B2-01:

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 Discussion of human causedIF-82] mechanisms that would result in a calculation of flood frequencies, in Section Flooding Study (51 -9100978 -000). floods is discussed in detail influid release.

INCLUDE:

5.2 of document 51-9100978-000.

Failures Describe the situations where a human Section 3.3 and 5.3 of Internal(a) failure modes of components such of components in piping systems other than error could result in flooding (e.g., Flood Notebook (GI-IF-0000-rl) for various systems.

Based onas pipes, tanks, gaskets, expansion tanks are explicitly addressed by the EPRI inadvertent valve opening, inadvertent the analyses pro ed onthe analyses performed, onejoints, 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 thediversion of flow through openings a very limited attempt to address human- be screened.

Consistent with the Aux Building due tocreated to perform maintenance; induced flood mechanisms, as required by Standard, utilize generic data as maintenance, isolated within 65inadvertent actuation of fire item (b) of SR IF-B2. required by SR IFEV-A7 (IF-D6 in 2005 minutes.suppression system Standard)

(c) other events releasing water into Such events have been important causes ofthe area flooding in the operating experience for USnuclear power plants, and as noted abovethe assessment of such floods is explicitly required.

A more systematic consideration should bemade of human-caused floods. This willneed to include an assessment of genericdata related to human-caused floods, perSR IF-D6.

LAR -Adoption of TSTF-425, Revision 3Docket No. 50-244Attachment 2Page 21 of 29Table 2-1 Internal Events PRA Peer Review -FindingsSR Topic Status Finding/Observation Disposition Impact to TSTF-425IFSN-A6[2005:IF-C3]For the SSCs identified in iFSN-A5(2005 text: IF-C2c),

IDENTIFY thesusceptibility of each SSC in a floodarea to flood-induced failuremechanisms.

INCLUDE failure by submergence andspray in the identification process.EITHER:a) ASSESS qualitatively the impact offlood-induced mechanisms that arenot formally addressed (e.g., usingthe mechanisms listed underCapability Category III of thisrequirement),

by using conservative assumptions; ORb) NOTE that these mechanisms arenot included in the scope of theevaluation.

Open F&O IF-C3-01:

There is no discussion offailures due to jet impingement or pipewhip. There is limited consideration offailure due to humidity/high temperature due to failure of heating steam lines. Thereis also no discussion of criteria employed toconsider the potential for spray failures.

To meet Capability Category II, it isnecessary either to provide at least aqualitative assessment of the potential forjet impingement and pipe whip, or to statethat these failure mechanisms were notconsidered.

It is also required that potential spray failures be evaluated.

While sprayfailures are discussed, there are no criteriaspecified that would provide assurance thatthey had been considered in a consistent and adequately comprehensive manner.Provide the requisite discussion of pipewhip and jet impingement to satisfy thestandard.

Specify appropriate criteria forspray impacts, and assure that the potential spray failures adequately reflect thesecriteria.

Cat I1: INCLUDE failure by submergence and spray in the identification process.ASSESS qualitatively the impact offlood-induced mechanisms that arenot formallyaddressed (e.g., using the mechanisms listed under Capability Category III ofthis requirement),

by usingconservative assumptions.

[SAIC note: these mechanisms includesubmergence, spray, jet impingement, pipe whip, humidity, condensation, temperature concerns]

Revise the Internal Flooding Study (51 -9100978 -000) to describe the criteriaused to determine the potential forfailure resulting from spray. Reference Appendix C for a listing of components impacted by spray. Describe howpotential spray impact was addressed in the model. Confirm that theassignment of spray impact isconsistent with the criteria used.In addition, include a qualitative discussion of the potential impact ofjet impingement, pipe whip, humidity, condensation, and temperature effects.Failures due to jet impingement and pipe whip are now discussed in Section 3.3.1 of the InternalFlood Notebook G2-1F-0000 ri.Failures due to Spray arediscussed in Section 3.3.2.Impacts due to spray wereassumed to exist within 10 feetof a break location.

Spray eventsare discussed in the IF Floodnotebook Section 4.5. Twolocations were identified in theAux Building where Fire ServiceWater could impact safetyrelated busses and these areexplicitly modeled (FL-ABM-FSW-BUS15 and FL-ABO-FSW-BUS14).

URE 1179 documents that IFNotebook needs Appendix Ccompleted to completedocumentation of spray impactsand modeling of additional sprayfloods if appropriate.

This wouldbe evaluated for any potential impacts to a surveillance frequency interval extension atthe time of the evaluation but isnot expected to have asignificant impact.

LAR -Adoption of TSTF-425, Revision 3Docket No. 50-244Attachment 2Page 22 of 29Table 2-1 Internal Events PRA Peer Review -FindingsSR Topic Status Finding/Observation Disposition Impact to TSTF-425IFSN-A8[2005:IF-C3b]IDENTIFY inter-area propagation through the normal flow path fromone area to another via drain lines;and areas connected via back flowthrough drain lines involving failedcheck valves, pipe and cablepenetrations (including cable trays),doors, stairwells, hatchways, andHVAC ducts. INCLUDE potential forstructural failure (e.g., of doors orwalls) due to flooding loads.complete F&O IF-C3b-01:

The analysis does notdocument consideration of potential barrierfailures due to flooding loads (structural

failures, failures of doors, etc.) This isrequired to meet capability categories beyond Capability Category I.Review flood barriers and identify andevaluate any whose failures couldcontribute adversely to propagation offloodingCat II, I1l: IDENTIFY inter-area.

A discussion of structural failureof barriers credited as barriershas been added to the IFNotebook rl, Section 4.2.1.INCLUDE potential for structural failure(e.g., of doors or walls) due to floodingloads and the potential for barrierunavailability, including maintenance activities.

Include a discussion of the potential for barrier failure due to flooding, including structures and doors. Forwalls, a qualitative discussion wouldappear to be acceptable.

For doors,however, specific failure criteria shouldbe developed and described.

Floodscenarios should be reviewed andrevised, if necessary, to address thepotential failure of doors.

LAR -Adoption of TSTF-425, Revision 3Docket No. 50-244Attachment 2Page 23 of 29Table 2-1 Internal Events PRA Peer Review -FindingsSR Topic Status Finding/Observation Disposition Impact to TSTF-425IFSN-A16[2005:IF-C8]USE potential human mitigative actions as additional criteria forscreening out flood sources if all thefollowing can be shown:(a) flood indication is available in thecontrol room;(b) the flood source can be isolated; and(c) the mitigative action can beperformed with high reliability for theworst flooding initiator (2005 text:flood from that source).

Highreliability is established bydemonstrating, for example, that theactions are procedurally

directed, that adequate time is available forresponse, that the area is accessible, and that there is sufficient manpoweravailable to perform the actions.Open F&O IF-C8-01:

Only one flood appears tohave been screened based on qualitative consideration of potential human action; forthat action (2000 gpm FSW break in IBN),there doesn't appear to be any justification for the time identified (190 min). Nothingother than time available is cited asrationale for screening the event.Characterize in greater detail thosepotential human actions that couldterminate the event and develop anestimate of the likelihood of failing tomitigate the pipe break using acceptedHRA methods.The screened flood will be addedto the flood model (URE 1176).However, the impact is expectedto be minimal, and is notexpected to have any impact onthe SFCP.To meet Capability Category II, it isnecessary to characterize potential humanactions that could terminate flooding moreexplicitly than was done in this case.Address the required aspects for this andany other human actions used in justifying screening out flood scenarios.

LAR -Adoption of TSTF-425, Revision 3Docket No. 50-244Attachment 2Page 24 of 29Table 2-1 Internal Events PRA Peer Review -FindingsSR Topic Status Finding/Observation Disposition Impact to TSTF-425IFEV-A6[2005:IF-D5a]GATHER plant-specific information onplant design, operating practices, andconditions that may impact floodlikelihood (i.e., material condition offluid systems, experience with waterhammer, and maintenance-induced floods).

In determining the flood-initiating event frequencies for floodscenario groups, USE a combination of the following (2005 text does notinclude "of the following")

(a) generic and plant-specific operating experience; (b) pipe, component, and tankrupture failure rates from genericdata sources and plant-specific experience; (2005 text: and)(c) engineering judgment forconsideration of the plant-specific information collected.

OpenF&O IF-D5a-01:

The current analysis doesnot adequately address plant-specific characteristics that might affect the mannerin which the frequencies of flooding areestimated.

To meet Capability Category II, it is requiredthat plant-specific information be collected and considered on a variety of aspects(including material condition of fluidsystems, experience with water hammer,and maintenance-induced floods).

Thecurrent analysis is limited to the use ofgeneric failure rates. This is consistent withCapability Category I.Address potential issues with materialcondition, experience with water hammer,etc. In particular, further attention shouldbe paid to the possibility of maintenance-induced and other human-caused flooding.

Address potential issues with materialcondition and water hammer usingplant-specific information.

Use thisinformation to revise, if necessary, piping failure frequencies available inindustry-wide

sources, consistent withthe Standard.

For maintenance-induced and otherhuman-caused

flooding, see IFSO-A4.URE 1153 was written to considerupdating flood frequency for agingaffects based on EPRI-302000079 Rupture frequencies.

Plant specific experience withinternal

flooding, water hammeris addressed in the IF Notebookrev 1 in Sections 3.3. Adiscussion of Human-induced floods is contained in Section 5.3.Regarding any effect on floodfrequency due to aging affects, asensitivity evaluation for aparticular STI evaluation wouldshow if there was any impact.

LAR -Adoption of TSTF-425, Revision 3Docket No. 50-244Attachment 2Page 25 of 29Table 2-1 Internal Events PRA Peer Review -FindingsSR Topic Status Finding/Observation Disposition Impact to TSTF-425IFEV-A7[2005:IF-D6]INCLUDE consideration of human-induced floods during maintenance through application of generic data.complete F&O IF-D6-01:

Initiating events that couldresult from human actions were considered only for a small number of possiblemaintenance activities.

These floodcontributors were not evaluated usinggeneric data as required.

See IFSO-A4.Discussion of human causedfloods is discussed in detail inSection 3.3 and 5.3 of InternalFlood Notebook (G1-IF-0000-rl) for various systems.

Based onthe analyses performed, onemaintenance induced flood wasadded to the model, FL-ABO-M-SW -2,000 gpm SW flood in theAux Building due tomaintenance, isolated within 65minutes.Operating experience for nuclear powerplants has provided evidence that human-caused floods can be important.

The SRrequires that such floods be evaluated usingat least generic data to meet Capability Category I or II.Perform a more detailed assessment ofpotential human-caused floods, and applyat least generic data to characterize theirfrequencies.

LAR -Adoption of TSTF-425, Revision 3Docket No. 50-244Attachment 2Page 26 of 29Table 2-1 Internal Events PRA Peer Review -FindingsSR Topic Status Finding/Observation Disposition Impact to TSTF-425IFEV-A8[2005:IF-D7]SCREEN OUT flood scenario groups if(a) the quantitative screening criteriain IFSN-A10 (2005 text: IE-C4), asapplied to the flood scenario groups,are met; OR(b) the internal flood-initiating eventaffects only components in a singlesystem, AND it can be shown that theproduct of the frequency of the floodand the probability of SSC failuregiven the flood is two orders ofmagnitude lower than:the product of the non-flooding frequency for the corresponding initiating events in the PRA, AND therandom (non-flood-induced) failureprobability of the same SSCs that areassumed failed by the flood.complete F&O IF-D7-01:

Quantitative screening ofsome scenarios was performed, but it is notclear what criteria were applied in doing so.The criteria should be defined and appliedin a clear and consistent manner.SRs IF-D7 and IF-E3a provide explicit criteriafor performing quantitative screening offlood scenarios.

The IF Notebookdocuments that some scenarios werescreened on low frequency, but does notinvoke any particular criteria in doing so.Provide a clear set of criteria for performing quantitative screening of flood scenarios, and apply the criteria in a clear andconsistent manner.Update the Internal Flooding Study (51-9100978 -000) to describe thecriteria used to screen flood scenarios.

If current screening criteria are notwell defined, develop such criteria andapply them to scenarios addressed inthe analysis.

No impact to TSTF 425. This issuehas been addressed.

InternalFlood Notebook Section 4.6,Screening Scenarios and Sources,was updated to document thescreening criteria used. Figure4.1, was added which shows theScreening Criteria and Table 4.6was edited to show the screening criterion used for various floodscenarios.

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

LAR -Adoption of TSTF-425, Revision 3Docket No. 50-244Attachment 2Page 27 of 29Table 2-1 Internal Events PRA Peer Review -FindingsSR Topic Status Finding/Observation Disposition Impact to TSTF-425I FQU-A5[2005:IF-ES]If additional human failure events arerequired to support quantification offlood scenarios, PERFORM any humanreliability analysis in accordance withthe applicable requirements described in 2-2.5 (2005 text: Tables4.5.5-2(e) through 4.5.5-2(h)).

complete F&O IF-E5-01:

It was not clear that therequirements were met in all cases. Forexample, interviews to establish aspectssuch as response times were apparently performed as part of the flood analysis, butthe HRA was dramatically changed and newinterviews/changes were not incorporated, nor were any inputs obtained from the HRAperformed as part of the flood analysiscarried forward.It is necessary to perform the assessment ofHFEs associated with internal flooding in thesame manner as for other HFEs. Therequirements to confirm procedure paths,timing, etc. via interviews with operators were not met for a number of events.Re-examine the HFEs associated withinternal

flooding, and either performneeded operator interviews or identify anddocument existing inputs.Re-examine each HFE included in theflooding analysis.

Perform operatorinterviews as needed or identify anddocument previously performed interviews.

Required operator interviews shouldcomprise the following:

1. evaluate the flooding events based onsimilarities to identify a select set ofscenarios to review with the operators (forexample, categorized by the system thatgenerated the flood, e.g., fire protection)

2. schedule interview sessions of about 1/2hour to an hour per each flooding

scenario, conducted separately with two different operators (preferably one experienced, onenovice) to get diverse opinions.

3. include questions on timing consistent with the HRA Calculator Time Windowscreen for time of cue, time to diagnosis, time for execution/manipulation of action(including travel time, with potential flood-related access delays).

Be sure to ask aboutany differences for floods initiated in samesystem but in different rooms.4. document interviews during the sessions(notes and/or tape recordings) and later inthe HRA Calculator screens for OperatorInterviews and Time Window.No impact to TSTF-425.

Ginna Station Flooding HumanReliability Analysis (HRA)documents the flood recoveryactions (Areva Document No.:51-9099406-000 located in GSN0157). The information and HRAvalues in this notebook wereverified to be consistent with theHRA actions being used in theinternal flood model. Noadditional interviews wereidentified as being necessary.

Estimate and document internal floodingHFEs using the same approach as was usedfor other HFEs in the PRA. Recalculate floodscenario frequencies based on the newHFEs.

LAR -Adoption of TSTF-425, Revision 3Docket No. 50-244Attachment 2Page 28 of 29Table 2-1 Internal Events PRA Peer Review -FindingsSR Topic Status Finding/Observation Disposition Impact to TSTF-425IFQU-B1[2005:IF-F1]DOCUMENT the internal floodaccident sequences andquantification in a manner thatfacilitates PRA applications,

upgrades, and peer review.Open F&O IF-FI-O1:

The documentation is comprised primarily of the internal flooding

notebook, supplemented heavily with information providedin a set of Excel worksheets.

The notebook isannotated to provide a link to elements of theworksheets, and an "assumption" provides theformal tie between the notebook and theworksheets.

Some areas in which the links wereindirect or missing were noted.In general, the manner in which important partsof the flood analysis are documented in whatwould usually be characterized as an informal setof worksheets is judged not to meet therequirement that the analysis be documented ina manner that facilitates applications,

upgrades, and peer review.In addition to developing a single integrated setof documentation for the internal flood analysis, there were several areas in which additional documentation would make the analysis moretractable have been provided in connection toother SRs. These include the following:

-Include a set of simplified arrangement drawings to explicate the definition of flood areasand help in understanding aspects such as floodpropagation.

-Tabulate the flood areas and identify clearlywhich are screened and which retained forfurther analysis to make the process moretractable.

Specify clearly which criteria(qualitative or quantitative) are employed inscreening each flood area.-Define explicitly the criteria used to performquantitative screening as noted in Section 6.0.-Define the criteria used to determine whether aPRA component was susceptible to failure due tospray.Documentation only: Revise the InternalFlooding Study (51 -9100978 -000) to meetthe documentation requirements of the2009 Standard.

Address NRC Resolutions asappropriate.

It is recommended that the Study bereformatted to be consistent with the HLRsand SRs of the Standard, integrating appropriate parts of the worksheets intothe primary document.

This will provide adocument that can be easily reviewedagainst the standard and easily followed bypersonnel not involved in the originalanalysis.

Consistent with the F&O, include thefollowing in the revised Study:-Include a set of simplified arrangement drawings to explicate the definition of floodareas and help in understanding aspectssuch as flood propagation.

-Tabulate the flood areas and identifyclearly which are screened and whichretained for further analysis to make theprocess more tractable.

Specify clearlywhich criteria (qualitative or quantitative) are employed in screening each flood area.-Define explicitly the criteria used toperform quantitative screening as noted inSection 6.0.-Define the criteria used to determine whether a PRA component was susceptible to failure due to spray.This documentation item will notimpact the TSTF 425 analysis.

This item has largely beenaddressed by adding tables inSection 5.2 that show thedevelopment of each initiating event frequency, adding anInitiating Event Summary Table(section 5.2.17),

adding asimplified set of arrangement drawings showing each floodarea (Appendix K), defining spraymodeling criteria (Section 3.3.2)and identifying for each floodarea whether it was screenedand the screening criterion used(Table 4.6). The remaining itemis to develop the criteria used toperform quantitative screening, if applicable, in Section 6.0 (URE1177).

LAR -Adoption of TSTF-425, Revision 3Docket No. 50-244Attachment 2Page 29 of 29Table 2-1 Internal Events PRA Peer Review -FindingsSR Topic Status Finding/Observation Disposition Impact to TSTF-425IFQU-B3[2005:IF-F3]DOCUMENT sources of modeluncertainty and related assumptions (as identified in QU-E1 and QU-E2)associated with the internal floodaccident sequences andquantification.

(2005 text: Document the keyassumptions and the key sources ofuncertainty associated with theinternal flooding analysis.)

Open F&O IF-F3-O1:

Section 7 of the IF Notebookprovides a discussion of three areasconsidered to be major sources ofuncertainty in the flood analysis.

This doesnot constitute an adequate characterization of the sources of uncertainty associated with the flood analysis or a comprehensive discussion of the assumptions that couldhave an effect on the results.A reasonably thorough investigation ofsources of uncertainty is necessary forproper characterization of the floodanalyses and results.A more comprehensive characterization ofsources of uncertainty, comparable to thatprovided for other areas of the PRA, shouldDocumentation only: Update thediscussion of assumptions anduncertainty to be consistent with the2009 Standard.

The 2005 Standardrequired the documentation of keyassumptions and key sources ofuncertainty, while the 2009 Standardeliminates the term "key." Theequivalent sections of other PRAtechnical elements provide an exampleof the detail that is required.

Inaddition, the discussion of uncertainty and impact of assumptions in theQuantification Notebook should berevised to include treatment of floodissues (or alternately, a similartreatment should be provided in theFlood Notebook).

This documentation issue will notaffect the TSTF 425 analysis.

Thisissue has been partially addressed by the calculation oferror factors for the floodinitiating events. These havebeen added to table 5-2, floodfrequencies.

Remaining action isto reference any key sources ofuncertainty from the EPRIguideline on the treatment ofuncertainty for ASME PRAStandard SRs related to internalflooding in Section 7 of the PRAInternal Flooding Analysis SystemNotebook (URE 1178)be developed for the internal flood analysis.

ATTACHMENT 3License Amendment RequestR. E. Ginna Nuclear Power PlantDocket No. 50-244Application 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)Pronosed Technical Snecification Paae Chanoes1.1-4 3.3.6-2 3.5.1-2 3.7.12-13.1.1-1 3.4.1-1 3.5.2-2 3.7.14-13.1.2-2 3.4.1-2 3.5.2-3 3.8.1-33.1.4-3 3.4.2-1 3.5.4-1 3.8.1-43.1.5-1 3.4.3-2 3.6.2-4 3.8.1-53.1.6-2 3.4.4-1 3.6.3-6 3.8.3-23.1.8-2 3.4.5-2 3.6.3-7 3.8.4-23.2.1-3 3.4.6-2 3.6.4-1 3.8.6-13.2.1-4 3.4.7-2 3.6.5-1 3.8.6-23.2.2-2 3.4.8-2 3.6.6-2 3.8.7-23.2.3-1 3.4.9-1 3.6.6-3 3.8.8-23.2.4-3 3.4.11-3 3.7.2-2 3.8.9-23.3.1-8 3.4.12-4 3.7.4-1 3.8.10-23.3.1-9 3.4.12-5 3.7.5-3 3.9.1-13.3.1-10 3.4.13-2 3.7.6-1 3.9.2-23.3.2-3 3.4.14-2 3.7.7-2 3.9.3-23.3.2-4 3.4.14-3 3.7.8-2 3.9.4-23.3.3-2 3.4.15-3 3.7.9-2 3.9.5-23.3.4-2 3.4.16-2 3.7.10-1 3.9.6-13.3.5-3 3.5.1-1 3.7.11-1 5.5-13 Definitions 1.1PHYSICS TESTSPRESSURE ANDTEMPERATURE LIMITS REPORT(PTLR)QUADRANTPOWER TILTRATIO(QPTR)RATED THERMALPOWER(RTP)SHUTDOWNMARGIN(SDM)PHYSICS TESTS shall be those tests performed to measure thefundamental nuclear characteristics of the reactor core and relatedinstrumentation.

These tests are:a. Described in Chapter 14, Initial Test Program of the UFSAR;b. Authorized under the provisions of 10 CFR 50.59; orc. Otherwise approved by the Nuclear Regulatory Commission (NRC).The PTLR is the plant specific document that provides the reactor vesselpressure and temperature limits, including heatup and cooldown rates,and the power operated relief valve lift settings and enable temperature associated with the Low Temperature Overpressurization Protection System for the current reactor vessel fluence period. These pressureand temperature limits shall be determined for each fluence period inaccordance with Specification 5.6.6. Plant operation within these limits isaddressed in individual specifications.

QPTR shall be the ratio of the highest average nuclear power in anyquadrant to the average nuclear power in the four quadrants.

RTP shall be a total reactor core heat transfer rate to the reactor coolantof 1775 MWt.SDM shall be the instantaneous amount of reactivity by which the reactoris subcritical or would be subcritical from its present condition assuming:

a. All rod cluster control assemblies (RCCAs) are fully inserted exceptfor the single RCCAof highest reactivity worth, which is assumed tobe fully withdrawn.

With any RCCAs not capable of being fullyinserted, the reactivity worth of the RCCAs must be accounted forin the determination of SDM; andb. In MODES 1 and 2, the fuel and moderator temperatures arechanged to the nominal hot zero power temperature.

A STAGGERED TEST BASIS shale eenii t ef the testing cf .n. .f thesystems, subsystems,

ehanncls, er .the. designated ccmpononts duringthe finteryel speciflcd by the Gurveillenee Frogueney, se that all systems3, subsystems,

.hann.ls,

e. .the. designated

.cmpncnt.

arc testedduring n Su,,,illaon F.equcney intervels, whcrc n is the tetal numbcr efsystems, subsystems,

ehannela, er ether designated eenmpocnts gin theasseciated function.

BA&SIR.E. Ginna Nuclear Power Plant1.1-4Amendment 100 SDM3.1.13.13.1.1REACTIVITY CONTROL SYSTEMSSHUTDOWN MARGIN (SDM)SDM shall be within the limits specified in the COLR.LCO 3.1.1APPLICABILITY:

MODE 2 with keff < 1.0,MODES 3, 4, and 5.ACTIONSCONDITION REQUIRED ACTION COMPLETION TIMEA. SDM not within limit. A.1 Initiate boration to restore 15 minutesSDM to within limit.SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.1.1 Verify SDM is within the limits specified in the COLR. 24 hot',csINSERT 1R.E. Ginna Nuclear Power Plant3.1.1-1Amendment 80 Core Reactivity 3.1.2SURVEILLANCE FREQUENCY YSR 3.1.2.2-NOTE -1. Only required after 60 effective full power days(EFPD).2. The predicted reactivity values must beadjusted (normalized) to correspond to themeasured core reactivity prior to exceeding afuel burnup of 60 EFPD after each fuel loading.Verify measured core reactivity is within +/- 1 % Ak/k ofpredicted values.31 EFPPR.E. Ginna Nuclear Power Plant3.1.2-2Amendment 80 Rod Group Alignment Limits3.1.4SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.4.1Verify individual rod positions within alignment limit.SR 3.1.4.2-NOTE -Only required to be performed if the rod positiondeviation monitor is inoperable.

Verify individual rod positions within alignment limit.1-2 heumsOnce within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />sand eve~y hetUF

-theFeaftef A~- TI T11-4-SR 3.1.4.3Verify rod freedom of movement (trippability) bymoving each rod not fully inserted in the core to aMRPI transition in either direction.

92-,ýSR 3.1.4.4 Verify rod drop time of each rod, from the fully Once prior towithdrawn

position, is < 1.8 seconds from the reactor criticality beginning of decay of stationary gripper coil voltage to after each removaldashpot entry, with: of the reactor heada. Tavg _> 500°F; andb. Both reactor coolant pumps operating.

R.E. Ginna Nuclear Power Plant3.1.4-3Amendment 80 Shutdown Bank Insertion Limit3.1.53.1 REACTIVITY CONTROL SYSTEMS3.1.5 Shutdown Bank Insertion LimitLCO 3.1.5 The shutdown bank shall be at or above the insertion limit specified in theCOLR.-NOTE -The shutdown bank may be outside the limit when required forperformance of SR 3.1.4.3.APPLICABILITY:

MODE 1,MODE 2 with Keff > 1.0.ACTIONSCONDITION REQUIRED ACTION COMPLETION TIMEA. Shutdown bank not within A.1.1 Verify SDM is within the 1 hourlimit. limits specified in the COLR.ORA.1.2 Initiate boration to restore 1 hourSDM to within limit.ANDA.2 Restore shutdown bank to 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />swithin limit.B. Required Action and B.1 Be in MODE 2 with Keff 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />sassociated Completion

< 1.0.Time not met.SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.5.1 Verify the shutdown bank insertion is within the limitspecified in the COLR. n INER.E. Ginna Nuclear Power Plant3.1.5-1Amendment 80 Control Bank Insertion Limits3.1.6SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.6.1 Verify estimated critical control bank position is within Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> priorthe limits specified in the COLR. to achieving criticality SR 3.1.6.2Verify each control bank insertion is within the limitsspecified in the COLR.SR 3.1.6.3-NOTE -Only required to be performed if the rod insertion limitmonitor is inoperable.

Verify each control bank insertion is within the limitsspecified in the COLR.Once within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />sand eveFy 44teF9-t-hefeefte lINSfERT 11SR 3.1.6.4Verify each control bank not fully withdrawn from thecore is within the sequence and overlap limitsspecified in the COLR.R.E. Ginna Nuclear Power Plant3.1.6-2Amendment 80 PHYSICS TESTS Exceptions

-MODE 23.1.8CONDITION REQUIRED ACTION COMPLETION TIMED. Required Action and D.1 Be in MODE 3. 15 minutesassociated Completion Time of Condition C notmet.SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.8.1 Perform a COT on power range and intermediate Once within 7 daysrange channels per SR 3.3.1.7 and SR 3.3.1.8.

prior to criticality SR 3.1.8.2 Verify the RCS lowest loop average temperature is "3 !NStER5300 F. INSERT 1SR 3.1.8.3Verify THERMAL POWER is <5% RTP.30-mifiutes i .1.SR 3.1.8.4Verify SDM is within the limits specified in the COLR.N-wTdINSERT1j R.E. Ginna Nuclear Power Plant3.1.8-2Amendment 80 FQ(Z)3.2.1SURVEILLANCE REQUIREMENTS II-NOTE -During power escalation at the beginning of each cycle, THERMAL POWER may be increased until an equilibrium power level has been achieved, at which a power distribution map isobtained.

SURVEILLANCE FREQUENCY SR 3.2.1.1 Verify FQc(Z) is within limit. Once after eachrefueling prior toTHERMAL POWERexceeding 75%RTPANDOnce within 12hours afterachieving equilibrium conditions afterexceeding, by>_ 10% RTP, theTHERMAL POWERat which FQC(Z) waslast verifiedAND31 E NSFPD th1rIftcIR.E. Ginna Nuclear Power Plant3.2.1-3Amendment 94 FQ(Z)3.2.1SURVEILLANCE FREQUENCY SR 3.2.1.2 -NOTE -If measurements indicate that themaximum over z [FQC(Z) / K(Z)]has increased since the previous evaluation ofFQC(Z):a. Increase FQW(Z) by the greater of a factor of1.02 or by an appropriate factor specified in theCOLR and reverify FQW(Z) is within limits orb. Repeat SR 3.2.1.2 once per 7 EFPD until eithera. above is met or two successive flux mapsindicate that themaximum over z [FQC(Z) / K(Z) ]has not increased.

Verify FQW(Z) is within limit. Once after eachrefueling prior toTHERMAL POWERexceeding 75%RTPANDOnce within 12hours afterachieving equilibrium conditions afterexceeding, by_> 10% RTP, theTHERMAL POWERat which FQW(Z)was last verifiedAND... NSERT1R.E. Ginna Nuclear Power Plant3.2.1-4Amendment 94 FNAH3.2.2SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.2.1Verify FNAH is within limits specified in the COLR.Once after eachrefueling prior toTHERMAL POWERexceeding 75%RTPAND:34 l I IAfe aIteFSIR3.2.2.2

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---NOE-----------------

Only required to be performed if one power rangechannel is inoperable with THERMAL POWER _> 75%RTP.Verify FNAH is within limits specified in the COLR. Once within 24hours and eefy-2.4.. .. A ..L4ýNSERT 1T1R.E. Ginna Nuclear Power Plant3.2.2-2Amendment 80 AFD3.2.33.23.2.3POWER DISTRIBUTION LIMITSAXIAL FLUX DIFFERENCE (AFD)LCO 3.2.3APPLICABILITY:

The AFD in % flux difference units shall be maintained within the limitsspcified in the COLR.-NOTE -The AFD shall be considered outside limits when two or more OPERABLEexcore channels indicate AFD to be outside limits.MODE 1 with THERMAL POWER >_ 50% RTP.IACTIONSCONDITION REQUIRED ACTION COMPLETION TIMEA. AFD not within limits. A.1 Reduce THERMAL POWER 30 minutesto < 50% RTP.SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.3.1 Verify AFD within limits for each OPERABLE excore 1- Iyschannel.

j INSER.E. Ginna Nuclear Power Plant3.2.3-1Amendment 94 QPTR3.2.4SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY IISR 3.2.4.1-NOTE -1. With input from one Power Range Neutron Fluxchannel inoperable and THERMAL POWER< 75% RTP, the remaining three power rangechannels can be used for calculating QPTR.2. SR 3.2.4.2 may be performed in lieu of thisSurveillance.

Verify QPTR is within limit by calculation.

INSERT 1SR 3.2.4.2 -- NOTE -Not required to be performed until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after inputfrom one or more Power Range Neutron Fluxchannels are inoperable with THERMAL POWER> 75% RTP.Perform SR 3.2.1.1, SR 3.2.1.2 and SR 3.2.2.1.R.E. Ginna Nuclear Power Plant3.2.4-3Amendment 94 RTS Instrumentation 3.3.1CONDITION REQUIRED ACTION COMPLETION TIMEW.2 Restore trip mechanism or 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />strain to OPERABLE status.X. Required Action and X.1 Initiate action to fully insert Immediately associated Completion all rods.Time of Condition W notmet. ANDX.2 Place the Control Rod Drive 1 hourSystem in a Condition incapable of rod withdrawal.

SURVEILLANCE REQUIREMENTS

-NOTE -Refer to Table 3.3.1-1 to determine which SRs apply for each RTS Function.

SURVEILLANCE FREQUENCY SR 3.3.1.1 Perform CHANNEL CHECK. 42- h, .....SR 3.3.1.2-NOTE -Required to be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> afterTHERMAL POWER is > 50% RTP.Compare results of calorimetric heat balancecalculation to Nuclear Instrumentation System (NIS)channel output and adjust if calorimetric power is> 2% higher than indicated NIS power.L-ýýý24 heufs4-SR 3.3.1.3-NOTE -1. Required to be performed within 7 days afterTHERMAL POWER is 50% RTP but prior toexceeding 90% RTP following each refueling and if the Surveillance has not been performed within the last 31 EFPD.2. Performance of SR 3.3.1.6 satisfies this SR.Compare results of the incore detector measurements to NIS AFD and adjust if absolute difference is > 3%.INSERT 1]31 effeet-vc ulpower days (EFPD9)R.E. Ginna Nuclear Power Plant3.3.1-8Amendment 112 RTS Instrumentation 3.3.1SURVEILLANCE FREQUENCY SR 3.3.1.4Perform TADOT.STACCERE ITESTB A I ' I ETT&SR 3.3.1.5 Perform ACTUATION LOGIC TEST. ,4, days...a.

S- ACGEREDTEST BASERTSSR 3.3.1.6-NOTE -Not required to be performed until 7 days afterTHERMAL POWER is _ 50% RTP, but prior toexceeding 90% RTP following each refueling.

Calibrate excore channels to agree with incoredetector measurements.

2 INSERT 192-EFPPQSR 3.3.1.7 ------ NOTE -----------------

Not required to be performed for source rangeinstrumentation prior to entering MODE 3 from MODE2 until 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after entering MODE 3. INSERT 1Perform COT. 92 daysSR 3.3.1.8 ------ NOTE -----------------

1. Not required for power range and intermediate range instrumentation until 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> afterreducing power < 6% RTP.2. Not required for source range instrumentation until 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after reducing power < 5E-11amps. INSERT 1Perform COT. 92 daysSR 3.3.1.9 -- NOTE -Setpoint verification is not required.

INSERT 1Perform TADOT.R.E. Ginna Nuclear Power Plant3.3.1-9Amendment 112 RTS Instrumentation 3.3.1SURVEILLANCE FREQUENCY SR 3.3.1.10-NOTE -Neutron detectors are excluded.

24-Riei4he Perform CHANNEL CALIBRATION.

SR 3.3.1.11Perform TADOT.SR 3.3.1.12-NOTE -Setpoint verification is not required.

Perform TADOT.Prior to reactorstartup if notperformed withinprevious 31 daysiSR 3.3.1.13Perform COT.OA -m-i6INtSERT 1R.E. Ginna Nuclear Power Plant3.3.1-10Amendment 112 ESFAS Instrumentation 3.3.2CONDITION REQUIRED ACTION COMPLETION TIMEL. As required by Required L.1Action A. 1 and referenced

---

by Table 3.3.2-1.

-NOTE -The inoperable channelmay be bypassed for up to 4hours for surveillance testing of the otherchannels.

Place channel in trip. 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />sM. Required Action and M.1 Be in MODE 3. 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />sassociated Completion Time of Condition L not ANDmet.M.2 Reduce pressurizer 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />spressure to < 2000 psig.N. As required by Required N.1 Declare associated Auxiliary Immediately Action A. 1 and referenced Feedwater pump inoperable by Table 3.3.2-1.

and enter applicable condition(s) of LCO 3.7.5,"Auxiliary Feedwater (AFW)System."SURVEILLANCE REQUIREMENTS

-NOTE -Refer to Table 3.3.2-1 to determine which SRs apply for each ESFAS Function.

SURVEILLANCE FREQUENCY SR 3.3.2.1Perform CHANNEL CHECK.4hf INSERT 1SR 3.3.2.2Perform COT.92-~ys-----------------------------.

4TSR 3.3.2.3 -NOTE -Verification of relay setpoints not required.

Perform TADOT.L4INSERT 11Amendment 109R.E. Ginna Nuclear Power Plant3.3.2-3 ESFAS Instrumentation 3.3.2SURVEILLANCE FREQUENCY

• SR 3.3.2.4-NOTE -Verification of relay setpoints not required.

Perform TADOT.24-mgenths SR 3.3.2.5Perform CHANNEL CALIBRATION.

SR 3.3.2.6Verify the Pressurizer Pressure-Low and Steam LinePressure-Low Functions are not bypassed whenpressurizer pressure

> 2000 psig.24 Fflenthst1E~TliSR 3.3.2.7Perform ACTUATION LOGIC TEST.OA ---16eL-lJsERýT1lT R.E. Ginna Nuclear Power Plant3.3.2-4Amendment 109 PAM Instrumentation 3.3.3CONDITION REQUIRED ACTION COMPLETION TIMED. One or more Functions D.1 Restore one channel to 7 dayswith two required OPERABLE status.channels inoperable.

E. Required Action and E.1 Enter the Condition Immediately associated Completion referenced in Table 3.3.3-1Time of Condition C or D for the channel.not met.F. As required by Required F.1 Be in MODE 3. 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />sAction E. 1 and referenced in Table 3.3.3-1.

ANDF.2 Be in MODE 4. 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />sG. As required by Required G.1 Initiate action to prepare Immediately Action E.1 and referenced and submit a special report.in Table 3.3.3-1.SURVEILLANCE REQUIREMENTS

-NOTE -SR 3.3.3.1 and SR 3.3.3.2 apply to each PAM instrumentation Function in Table 3.3.3-1.SURVEILLANCE FREQUENCY SR 3.3.3.1 Perform CHANNEL CHECK for each required

'34instrumentation channel that is normally energized.

SR 3.3.3.2 Perform CHANNEL CALIBRATION.

R.E. Ginna Nuclear Power Plant3.3.3-2Amendment 90 LOP DG Start Instrumentation 3.3.4SURVEILLANCE REQUIREMENTS

-NOTE -When a channel is placed in an inoperable status solely for the performance of requiredSurveillances, entry into the associated Conditions and Required Actions may be delayed for upto 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> provided the second channel maintains LOP DG start capability.

SURVEILLANCE FREQUENCY SR 3.3.4.1 Perform TADOT. 31 daysSR 3.3.4.2 Perform CHANNEL CALIBRATION with LimitingSafety System Settings (LSSS)(a) for each 480 V bus t SE Tas follows:

--Ia. Loss of voltage LSSS _> 372.0 V and < 374.8 Vwith a time delay of 2.13 seconds and 2.62seconds.b. Degraded voltage LSSS __ 420.0 V and ___ 423.6V with a time delay of> 68.1 seconds and < 125seconds (@ 420 V) and _> 71.8 seconds and< 125 seconds (@ 423.6 V).(a)A channel is OPERABLE when both of the following conditions are met:1. The absolute difference between the as-found Trip Setpoint (TSP) and the previousas-left TSP is within the CHANNEL CALIBRATION Acceptance Criteria.

The CHANNELCALIBRATION Acceptance Criteria is defined as:las-found TSP -previous as-left TSPI < CHANNEL CALIBRATION uncertainty The CHANNEL CALIBRATION uncertainty shall not include the calibration tolerance.

2. The as-left TSP is within the established calibration tolerance band about the nominalTSP. The nominal TSP is the desired setting and shall not exceed the LSSS.The LSSS and the established calibration tolerance band are defined in accordance with the Ginna Instrument Setpoint Methodology.

The channel is considered operableeven if the as-left TSP is non-conservative with respect to the LSSS providedthat the as-left TSP is within the established calibration tolerance band.R.E. Ginna Nuclear Power Plant3.3.4-2Amendment 109 Containment Ventilation Isolation Instrumentation 3.3.5SURVEILLANCE REQUIREMENTS

-NOTE -Refer to Table 3.3.5-1 to determine which SRs apply for each Containment Ventilation Isolation Function.

SURVEILLANCE FREQUENCY INSERT 1f lSR 3.3.5.1Perform CHANNEL CHECK.24 heu~sfSR 3.3.5.2 Perform COT. 92 daysINSERT 1INSERT 1INSERT 1SR 3.3.5.3Perform ACTUATION LOGIC TEST.24 ments....... .......iSR 3.3.5.4Perform CHANNEL CALIBRATION.

24 ffefitsR.E. Ginna Nuclear Power Plant3.3.5-3Amendment 85 CREATS Actuation Instrumentation 3.3.6CONDITION REQUIRED ACTION COMPLETION TIMED. Required Action and D.1 Suspend movement of Immediately associated Completion irradiated fuel assemblies.

Time of Condition A or Bnot met during movementof irradiated fuelassemblies.

SURVEILLANCE REQUIREMENTS

-NOTE -Refer to Table 3.3.6-1 to determine which SRs apply for each CREATS Actuation Function.

CHMIC11 I AK1111=CDCtHC,.,J v IL...,-,d,,

.I "',',.,'>',

---__-lN~iL-r I 1 ISR 3.3.6.1 Perform CHANNEL CHECK. 12 Iurs k. ,, lNSERT 1-]Zý,K J.J.0.zPerform CU I.ui- eaysSR 3.3.6.3 -- NOTE -Verification of setpoint is not required.

-NSERT 1Perform TADOT. FetSSR 3.3.6.4 Perform CHANNEL CALIBRATION.

24-ment-I SR 3.3.6.5 Perform ACTUATION LOGIC TEST. 'A--e

"]~INET1R.E. Ginna Nuclear Power Plant3.3.6-2Amendment 87 3.43.4.1LCO 3.4.1RCS Pressure, Temperature, and Flow DNB Limits3.4.1REACTOR COOLANT SYSTEMS (RCS)RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB)LimitsRCS DNB parameters for pressurizer

pressure, RCS averagetemperature, and RCS total flow rate shall be within the limits specified inthe COLR.-NOTE -Pressurizer pressure limit does not apply during pressure transients dueto:a. THERMAL POWER ramp > 5% RTP per minute; orb. THERMAL POWER step > 10% RTP.APPLICABILITY:

MODE 1.ACTIONSCONDITION REQUIRED ACTION COMPLETION TIMEA. One or more RCS DNB A.1 Restore RCS DNB 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />sparameters not within parameter(s) to within limit.limits.B. Required Action and B.1 Be in MODE 2. 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />sassociated Completion Time not met.SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1.1 Verify pressurizer pressure is within limit specified in 12- ..U.Sthe COLR.SR 3.4.1.2 Verify RCS average temperature is within limit 12 hO',rsspecified in the COLR. jLINSET1R.E. Ginna Nuclear Power Plant3.4.1-1Amendment 80 RCS Pressure, Temperature, and Flow DNB Limits3.4.1SURVEILLANCE FREQUENCY SR 3.4.1.3 -NOTE -Required to be performed within 7 days after > 95%RTP.Verify RCS total flow rate is within the limit specified in 24 monthsthe COLR.R.E. Ginna Nuclear Power Plant3.4.1-2Amendment 80 3.4RCS Minimum Temperature for Criticality 3.4.2REACTOR COOLANT SYSTEM (RCS)RCS Minimum Temperature for Criticality 2 Each RCS loop average temperature (Tavg) shall be > 540'F.3.4.2LCO 3.4.2APPLICABILITY:

MODE 1,MODE 2 with keff > 1.0.ACTIONSCONDITION REQUIRED ACTION COMPLETION TIMEA. Tavg in one or both RCS A.1 Be in MODE 2 with Keff 30 minutesloops not within limit. < 1.0.SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.2.1 Verify RCS Tavg in each loop > 5400F. Within 30 minutesprior to achieving criticality.

SR 3.4.2.2-NOTE -Only required if any RCS loop Tavg < 5470F and thelow Tavg alarm is either inoperable or not reset.Verify RCS Tavg in each loop > 540'F.Once within 30minutes andthereaftef T;NERIýTR.E. Ginna Nuclear Power Plant3.4.2-1Amendment 80 RCS P/T Limits3.4.3SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 -NOTE -Only required to be performed during RCS heatupand cooldown operations and RCS inservice leak and INSERT 1hydrostatic testing.Verify RCS pressure, RCS temperature, and RCS 39-if motesheatup and cooldown rates are within the limitsspecified in the PTLR.R.E. Ginna Nuclear Power Plant3.4.3-2Amendment 80 3.43.4.4RCS Loops -MODE 1 > 8.5% RTP3.4.4REACTOR COOLANT SYSTEM (RCS)RCS Loops -MODE 1 > 8.5% RTPTwo RCS loops shall be OPERABLE and in operation.

LCO 3.4.zAPPLICABILITY:

MODE 1 > 8.5% RTP.ACTIONSCONDITION REQUIRED ACTION COMPLETION TIMEA. Requirements of LCO not A.1 Be in MODE 1 < 8.5% RTP. 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />smet.SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.4.1 Verify each RCS loop is in operation.

42--IhNSERT F 9HT1R.E. Ginna Nuclear Power Plant3.4.4-1Amendment 80 RCS Loops -MODES 1< 8.5% RTP, 2, and 33.4.5CONDITION REQUIRED ACTION COMPLETION TIMEC. Both RCS loops C.1 De-energize all CRDMs. Immediately inoperable.

ANDORC.2 Suspend operations that Immediately No RCS loop in operation, would cause introduction ofcoolant into the RCS withboron concentration lessthan required to meet theSDM of LCO 3.1.1.ANDC.3 Initiate action to restore one Immediately RCS loop to OPERABLEstatus and operation.

SURVEILLANCE REQUIREMENTS JINSERT 1SURVEILLANCE FREQUENCY SR 3.4.5.1 Verify required RCS loop is in operation.

12 heWFSSR 3.4.5.2 Verify steam generator secondary side water levels 42-.are _> 16% for two RCS loops. INT1SR 3.4.5.3 Verify correct breaker alignment and indicated powerare available to the required RCP that is not inoperation.

R.E. Ginna Nuclear Power Plant3.4.5-2Amendment 112 RCS Loops -MODE 43.4.6CONDITION REQUIRED ACTION COMPLETION TIMEB. One RHR loop -NOTE -inoperable.

Required Action B.1 is notAND applicable if all RCS andRHR loops are inoperable and Condition C is entered.Two RCS loops ---------inoperable.

B.1 Be in MODE 5. 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />sC. All RCS and RHR loops C.1 Suspend operations that Immediately inoperable, would cause introduction ofcoolant into the RCS withOR boron concentration lessthan required to meet theNo RCS or RHR loop in SDM of LCO 3.1.1.operation.

ANDC.2 Initiate action to restore one Immediately loop to OPERABLE statusand operation.

SURVEILLANCE REQUIREMENTS INSERT 1 J-1SURVEILLANCE FREQUENCY SR 3.4.6.1 Verify one RHR or RCS loop is in operation.

12 h4etesSR 3.4.6.2 Verify SG secondary side water level is _> 16% foreach required RCS loop. INSR 3.4.6.3 Verify correct breaker alignment and indicated power daysare available to the required pump that is not in AT-INSERT1 operation.

R.E. Ginna Nuclear Power Plant3.4.6-2Amendment 112 RCS Loops -MODE 5, Loops Filled3.4.7ACTIONSCONDITION REQUIRED ACTION COMPLETION TIMEA. One RHR loop A.1 Initiate action to restore a Immediately inoperable, second RHR loop toOPERABLE status.ANDORBoth SGs secondary sidewater levels not within A.2 Initiate action to restore Immediately limits, required SG secondary sidewater levels to within limits.B. Both RHR loops B.1 Suspend operations that Immediately inoperable, would cause introduction ofcoolant into the RCS withOR boron concentration lessthan required to meet theNo RHR loop in SDM of LCO 3.1.1.operation.

ANDB.2 Initiate action to restore one Immediately RHR loop to OPERABLEstatus and operation.

SURVEILLANCE REQUIREMENTS JINSERT 1SURVEILLANCE REQUENCYSR 3.4.7.1 Verify one RHR loop is in operation.

42--i'usSR 3.4.7.2 Verify SG secondary side water level is > 16% in the 1- -" ....required SG. TfINSERT 1SR 3.4.7.3 Verify correct breaker alignment and indicated power :-daysare available to the required RHR pump that is not in /LINEToperation.

R.E. Ginna Nuclear Power Plant3.4.7-2Amendment 112 RCS Loops -MODE 5, Loops Not Filled3.4.8CONDITION REQUIRED ACTION COMPLETION TIMEB.2 Initiate action to restore one Immediately RHR loop to OPERABLEstatus and operation.

SURVEILLANCE REQUIREMENTS JINSERT 1SURVEILLANCE REQUENCYSR 3.4.8.1 Verify one RHR loop is in operation.

12 h9'8r,SR 3.4.8.2 Verify correct breaker alignment and indicated power 7- dysare available to the RHR pump that is not in/LINSERT 1operation.

R.E. Ginna Nuclear Power Plant3.4.8-2Amendment 112 Pressurizer 3.4.93.4REACTOR COOLANT SYSTEM (RCS)Pressurizer 3.4.9LCO 3.4.9APPLICABILITY:

The pressurizer shall be OPERABLE.

MODES 1, 2, and 3.ACTIONSCONDITION REQUIRED ACTION COMPLETION TIMEA. Pressurizer water level A.1 Be in MODE 3 with reactor 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />snot within limit, trip breakers open.ANDA.2 Be in MODE 4. 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />sB. Pressurizer heaters B.1 Be in MODE 3. 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />scapacity not within limits.ANDB.2 Be in MODE 4. 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />sSURVEILLANCE REQUIREMENTS R.E. Ginna Nuclear Power Plant3.4.9-1Amendment 80 Pressurizer PORVs3.4.11SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.11.1-NOTE -Not required to be performed with block valve closedper LCO 3.4.13.Perform a complete cycle of each block valve.92daysSR 3.4.11.2 Perform a complete cycle of each PORV. 24 Fa nthsL-ýR.E. Ginna Nuclear Power Plant3.4.11-3Amendment 88 LTOP System3.4.12SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.12.1-NOTE -Only required to be performed when complying withLCO 3.4.12.a.

Verify no SI pump is capable of injecting into the RCS.SR 3.4.12.2-NOTE -Only required to be performed when complying withLCO 3.4.12.b.

Verify a maximum of one SI pump is capable ofinjecting into the RCS.SR 3.4.12.3-NOTE -Only required to be performed when ECCSaccumulator pressure is greater than or equal to themaximum RCS pressure for the existing RCS cold legtemperature allowed in the PTLR.Verify each ECCS accumulator motor operatedisolation valve is closed.INSERT 112-heuF sOnce within 12hours and evey- 12;-eI'NSEF T 112 he..s-for unlocked open ventvalve(s):for lockedopen vent valve(s)SR 3.4.12.4-NOTE -Only required to be performed when complying withLCO 3.4.12.b.

Verify RCS vent > 1.1 square inches open.SR 3.4.12.5 Verify PORV block valve is open for each requiredPORV. NTR.E. Ginna Nuclear Power Plant3.4.12-4Amendment 88 LTOP System3.4.12SURVEILLANCE FREQUENCY SR 3.4.12.6

-NOTE -Required to be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> afterdecreasing RCS cold leg temperature to less than orequal to the LTOP enable temperature specified in the ISTPTLR.Perform a COT on each required PORV, excluding 31-daysactuation.

SR 3.4.12.7-NOTE -Only required to be performed when ECCSaccumulator pressure is greater than or equal to themaximum RCS pressure for the existing RCS cold legtemperature allowed in the PTLR.Verify power is removed from each ECCSaccumulator motor operated isolation valve operator.

Once within 12hours and eyeiy 31daysER 1h~e IN T24 Ffentstf7J-TSR 3.4.12.8Perform CHANNEL CALIBRATION for each requiredPORV actuation channel.R.E. Ginna Nuclear Power Plant3.4.12-5Amendment 88 RCS Operational LEAKAGE3.4.13SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.13.1I-NOTE -1. 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 /> afterestablishment of steady state operation.

2. Not applicable to primary to secondary LEAKAGE.Verify RCS operational LEAKAGE is within limits byperformance of RCS water inventory balance.7-2 heWFSR 3.4.13.2

-- --NOTE -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 /> afterestablishment of steady state operation.

Verify primary to secondary LEAKAGE is < 150 72 hUFS,gallons per day through any one SG.IR.E. Ginna Nuclear Power Plant3.4.13-2Amendment 100 RCS PIV Leakage3.4.14CONDITION REQUIRED ACTION COMPLETION TIMEA.2 Isolate the high pressure 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />sportion of the affectedsystem from the lowpressure portion by use of asecond closed manual,deactivated automatic, orcheck valve.B. Required Action and B.1 Be in MODE 3. 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />sassociated Completion Time not met. ANDB.2 Be in MODE 5. 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />sSURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.14.1

-NOTE-1. Not required to be performed until prior toentering MODE 2 from MODE 3.2. RCS PIVs actuated during the performance ofthis Surveillance are not required to be testedmore than once if a repetitive testing loopcannot be avoided.

INSERT 1Verify leakage from each SI cold leg injection line and 24 m,,nthseach RHR RCS PIV is equivalent to < 0.5 gpm pernominal inch of valve size up to a maximum of 5 gpm ANDat an RCS pressure

> 2215 psig and < 2255 psig.Within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />sfollowing valveactuation due toautomatic or manualaction, flow throughthe valve, ormaintenance on thevalveR.E. Ginna Nuclear Power Plant3.4.14-2Amendment 80 RCS PIV Leakage3.4.14SURVEILLANCE FREQUENCY SR 3.4.14.2-NOTE -1. Not required to be performed until prior toentering MODE 2 from MODE 3.2. RCS PIVs actuated during the performance ofthis Surveillance are not required to be testedmore than once if a repetitive testing loopcannot be avoided.Verify leakage from each SI hot leg injection line RCSPIV is equivalent to < 0.5 gpm per nominal inch ofvalve size up to a maximum of 5 gpm at an RCSpressure

> 2215 psig and < 2255 psig.A 11 ..ANDWithin 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />sfollowing valveactuation due toautomatic or manualaction, flow throughthe valve, ormaintenance on thevalveR.E. Ginna Nuclear Power Plant3.4.14-3Amendment 80 RCS Leakage Detection Instrumentation 3.4.15SURVEILLANCE REQUIREMENTS JINSRT1i

-SURVEILLANCE FREQUENCY SR 3.4.15.1 Perform CHANNEL CHECK of containment

,2 h-ewsatmosphere radioactivity monitors.

FINET1SR 3.4.15.2 Perform COT of containment atmosphere radioactivity 92-daysmonitors.

I 1SR 3.4.15.3 Perform CHANNEL CALIBRATION of the required 24 FAIIIcontainment sump monitor.

INSERT 1SR 3.4.15.4 Perform CHANNEL CALIBRATION of containment 24-menths atmosphere radioactivity monitors.

R.E. Ginna Nuclear Power Plant3.4.15-3Amendment 88 RCS Specific Activity3.4.16SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.16.1Verify reactor coolant gross specific activity

< 1 0O/EpCi/gm.SR 3.4.16.2-NOTE -Only required to be performed in MODE 1.Verify reactor coolant DOSE EQUIVALENT 1-131specific activity

< 1.0 pjCi/gm.S SdEyq44-daysANDBetween 2 and 10hours after aTHERMAL POWERchange of _> 15%RTP within a 1 hourperiodSR 3.4.16.3-NOTE -Only required to be performed in MODE 1.Determine E from a reactor coolant sample.Once within 31 daysafter a minimum d 2effective full powerdays and 20 days ofMODE 1 operation have elapsed sincethe reactor was lastsubcritical for > 48hours.AND INSERT 1rhe" eafte .R.E. Ginna Nuclear Power Plant3.4.16-2Amendment 88 Accumulators 3.5.13.53.5.1EMERGENCY CORE COOLING SYSTEMS (ECCS)Accumulators Two ECCS accumulators shall be OPERABLE.

LCO 3.5.1APPLICABILITY:

MODES 1 and 2,MODE 3 with pressurizer pressure

> 1600 psig.ACTIONSCONDITION REQUIRED ACTION COMPLETION TIMEA. One accumulator A.1 Restore boron 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />sinoperable due to boron concentration to withinconcentration not within limits.limits.B. One accumulator B.1 Restore accumulator to 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />sinoperable for reasons OPERABLE status.other than Condition A.C. Required Action and C.1 Be in MODE 3. 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />sassociated Completion Time of Condition A or B ANDnot met.C.2 Reduce pressurizer 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />spressure to < 1600 psig.D. Two accumulators D.1 Enter LCO 3.0.3. Immediately inoperable.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.1.1Verify each accumulator motor operated isolation valve is fully open.27E~qT1SR 3.5.1.2Verify borated water volume in each accumulator is_> 1090 cubic feet (24%) and < 1140 cubic feet (83%).10 6--r,---T.41 lxi.-ý r- "R.E. Ginna Nuclear Power Plant3.5.1-1Amendment 101 Accumulators 3.5.1SURVEILLANCE SR 3.5.1.3Verify nitrogen cover pressure in each accumulator is>_ 700 psig and < 790 psig.12-heHFS': ýINS ýER1-I-ISR 3.5.1.4Verify boron concentration in each accumulator is_ 2550 ppm and < 3050 ppm.,F12 heour (byinleakage monitoring)

IANDf-ienthe (bysape INSERT 131-daysSR 3.5.1.5Verify power is removed from each accumulator motor'operated isolation valve operator whenpressurizer pressure is > 1600 psig.R.E. Ginna Nuclear Power Plant3.5.1-2Amendment 101 ECCS -MODES 1, 2, and 33.5.2SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.1Verify the following valves are in the listed position.

4n --..--Number825A825B826A826B826C826D851A851B856878A878B878C878D896A896BPositionOpenOpenClosedClosedClosedClosedOpenOpenOpenClosedOpenClosedOpenOpenOpenFunctionRWST Suction to SI PumpsRWST Suction to SI PumpsBAST Suction to SI PumpsBAST Suction to SI PumpsBAST Suction to SI PumpsBAST Suction to SI PumpsSump B to RHR PumpsSump B to RHR PumpsRWST Suction to RHR PumpsSI Injection to RCS Hot LegSI Injection to RCS Cold LegSI Injection to RCS Hot LegSI Injection to RCS Cold LegRWST Suction to SI andContainment SprayRWST Suction to SI andContainment SprayI ~I.SR 3.5.2.2Verify each ECCS manual, power operated, andautomatic valve in the flow path, that is not locked,sealed, or otherwise secured in position, is in the correctposition.

'481 daysF-EgKilSR 3.5.2.3 Verify each breaker or key switch, as applicable, for each MY-valve listed in SR 3.5.2.1, is in the correct position.

R.E. Ginna Nuclear Power Plant3.5.2-2Amendment 80 ECCS -MODES 1, 2, and 33.5.2SURVEILLANCE FREQUENCY SR 3.5.2.4 Verify each ECCS pump's developed head at the test flow In accordance point is greater than or equal to the required developed with thehead. Inservice Testing ProgramSR 3.5.2.5Verify each ECCS automatic valve in the flow path that isnot locked, sealed, or otherwise secured in positionactuates to the correct position on an actual or simulated actuation signal.nA ---LL-III I ILI IINSERT 1SR 3.5.2.6 Verify each ECCS pump starts automatically on an actual 24 4:thor simulated actuation signal.SR 3.5.2.7Verify, by visual inspection, each RHR containment sumpsuction inlet is not restricted by debris and thecontainment sump screen shows no evidence ofstructural distress or abnormal corrosion.

24-1TR.E. Ginna Nuclear Power Plant3.5.2-3Amendment 80 RWST3.5.43.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)3.5.4 Refueling Water Storage Tank (RWST)LCO 3.5.4 The RWST shall be OPERABLE.

APPLICABILITY:

ACTIONSMODES 1, 2, 3, and 4.CONDITION REQUIRED ACTION COMPLETION TIMEA. RWST boron A.1 Restore RWST to 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />sconcentration not within OPERABLE status.limits.B. RWST water volume not B.1 Restore RWST to 1 hourwithin limits. OPERABLE status.C. Required Action and C.1 Be in MODE 3. 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />sassociated Completion Time not met. ANDC.2 Be in MODE 5. 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />sSURVEILLANCE REQUIREMENTS F INSERT 1SURVEILLANCE FREQUENCY SR 3.5.4.1 Verify RWST borated water volume is > 300,000 7-daysgallons (88%). INSERT 1ISR 3.5.4.2 Verify RWST boron concentration is > 2750 ppm and 7-days< 3050 ppm.IR.E. Ginna Nuclear Power Plant3.5.4-1Amendment 96 Containment Air Locks3.6.2SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.1-NOTE -1. An inoperable air lock door does not invalidate the previous successful performance of theoverall air lock leakage test.2. Results shall be evaluated against acceptance criteria applicable to SR 3.6.1.1.Perform required air lock leakage rate testing inaccordance with the Containment Leakage RateTesting Program.In accordance withthe Containment Leakage RateTesting ProgramSR 3.6.2.2 Verify only one door in each air lock can be opened at 24 monethsa time.R.E. Ginna Nuclear Power Plant3.6.2-4Amendment 80 Containment Isolation Boundaries 3.6.3SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.3.1 Verify each mini-purge valve is closed, except when 3j"eysthe penetration flowpath(s) are permitted to be open NSERT1under administrative control.SR 3.6.3.2 -NOTE -1. Isolation boundaries in high radiation areas maybe verified by use of administrative controls.

2. Not applicable to containment isolation boundaries which receive an automatic containment isolation signal. INSERT IVerify each containment isolation boundary that is 92 dayslocated outside containment and not locked, sealed,or otherwise secured in the required position isperforming its containment isolation accident functionexcept for containment isolation boundaries that areopen under administrative controls.

SR 3.6.3.3 -NOTE -1. Isolation boundaries in high radiation areas maybe verified by use of administrative means.2. Not applicable to containment isolation boundaries which receive an automatic containment isolation signal.Verify each containment isolation boundary that is Prior to enteringlocated inside containment and not locked, sealed, or MODE 4 fromotherwise secured in the required position is MODE 5 if notperforming its containment isolation accident

function, performed within theexcept for containment isolation boundaries that are previous 92 daysopen under administrative controls.

SR 3.6.3.4 Verify the isolation time of each automatic In accordance withcontainment isolation valve is within limits, the Inservice Testing ProgramSR 3.6.3.5 Perform required leakage rate testing of containment In accordance withmini-purge valves with resilient seals in accordance the Containment with the Containment Leakage Rate Testing Program.

Leakage RateProgram.R.E. Ginna Nuclear Power Plant3.6.3-6Amendment 80 Containment Isolation Boundaries 3.6.3SURVEILLANCE FREQUENCY SR 3.6.3.6 Verify each automatic containment isolation valve that 24,,,,,st,,

is not locked, sealed, or otherwise secured in therequired position actuates to the isolation position on "an actual or simulated actuation signal.R.E. Ginna Nuclear Power Plant3.6.3-7Amendment 80 3.63.6.4Containment Pressure3.6.4CONTAINMENT SYSTEMSContainment Pressuret Containment pressure shall be > -2.0 psig and < 1.0 psig.LCO 3.6.4APPLICABILITY:

MODES 1, 2, 3, and 4.ACTIONSCONDITION REQUIRED ACTION COMPLETION TIMEA. Containment pressure not A.1 Restore containment.

8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />swithin limits, pressure to within limits.B. Required Action and B.1 Be in MODE 3. 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />sassociated Completion Time not met. ANDB.2 Be in MODE 5. 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />sSURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.1 Verify containment pressure is within limits.N TT1R.E. Ginna Nuclear Power Plant3.6.4-1Amendment 80 Containment Air Temperature 3.6.53.63.6.5CONTAINMENT SYSTEMSContainment Air Temperature LCO 3.6.5APPLICABILITY:

Containment average air temperature shall be < 1250F.MODES 1, 2, 3, and 4.ACTIONSCONDITION REQUIRED ACTION COMPLETION TIMEA. Containment average air A.1 Restore containment 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />stemperature not within average air temperature tolimit. within limit.B. Required Action and B.1 Be in MODE 3. 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />sassociated Completion Time not met. ANDB.2 Be in MODE 5. 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />sSURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.5.1Verify containment average air temperature is withinlimit.12 heH9ewR.E. Ginna Nuclear Power Plant3.6.5-1Amendment 116 CS, CRFC, and NaOH Systems3.6.6CONDITION REQUIRED ACTION COMPLETION TIMEF. Two CS trains inoperable.

F.1 Enter LCO 3.0.3. Immediately ORThree or more CRFCunits inoperable.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.6.1 Perform SR 3.5.2.1 and SR 3.5.2.3 for valves 896A In accordance withand 896B. applicable SRs.SR 3.6.6.2 Verify each CS manual, power operated, and/, yautomatic valve in the flow path that is not locked, INERsealed, or otherwise secured in position is in thecorrect position.

SR 3.6.6.3 Verify each NaOH System manual, power operated, 31-daysand automatic valve in the flow path that is not locked,sealed, or otherwise secured in position is in the NSERLTlcorrect position.

SR 3.6.6.4 Operate each CRFC unit for _> 15 minutes.

3 1 dIs-SR 3.6.6.5 Verify cooling water flow through each CRFC unit. 3t-y<SR 3.6.6.6 Verify each CS pump's developed head at the flow In accordance withtest point is greater than or equal to the required the Inservice developed head. Testing ProgramSERT 1SERTISR 3.6.6.7Verify NaOH System solution volume is _> 3000 gal.84 4days ~ --TSR 3.6.6.8 Verify NaOH System tank NaOH solution concentration is _> 30% and _< 35% by weight. INSERT 1SR 3.6.6.9 Perform required CRFC unit testing in accordance In accordance withwith the VFTP. the VFTPSR 3.6.6.10 Verify each automatic CS valve in the flow path that is 24 menthsnot locked, sealed, or otherwise secured in positionactuates to the correct position on an actual orsimulated actuation signal.R.E. Ginna Nuclear Power Plant3.6.6-2Amendment 99 CS, CRFC, and NaOH Systems3.6.6SURVEILLANCE FREQUENCY SR 3.6.6.11 Verify each CS pump starts automatically on an actual OAor simulated actuation signal. SERT1SR 3.6.6.12Verify each CRFC unit starts automatically on anactual or simulated actuation signal.SR 3.6.6.13Verify each automatic NaOH System valve in the flowpath that is not locked, sealed, or otherwise securedin position actuates to the correct position on anactual or simulated actuation signal.7A -- T1-SR 3.6.6.14 Verify spray additive flow through each eductor path. 6-yeeisSR 3.6.6.15 Verify each spray nozzle is unobstructed.

Following maintenance whichcould result innozzle blockageR.E. Ginna Nuclear Power Plant3.6.6-3Amendment 99 MSIVs and Non-Return Check Valves3.7.2SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.2.1 Verify closure time of each MSIV is < 5 seconds under In accordance withno flow and no load conditions, the Inservice Testing ProgramSR 3.7.2.2 Verify each main steam non-return check valve can In accordance withclose, the Inservice Testing ProgramSR 3.7.2.3 Verify each MSIV can close on an actual or simulated 24 ,ionthsactuation signal. INS.ERTR.E. Ginna Nuclear Power Plant3.7.2-2Amendment 80 ARVs3.7.43.73.7.4PLANT SYSTEMSAtmospheric Relief Valves (ARVs)I Two ARV lines shall be OPERABLE.

LCO 3.7.APPLICABILITY:

MODES 1 and 2,MODE 3 with Reactor Coolant System average temperature (Tavg)> 5000F.IACTIONSCONDITION REQUIRED ACTION COMPLETION TIMEA. One ARV line inoperable.

A.1 Restore ARV line to 7 daysOPERABLE status.B. Required Action and B.1 Be in MODE 3 with Tavg 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />sassociated Completion

< 5000F.Time of Condition A notmet.C. Two ARV lines C.1 Enter LCO 3.0.3. Immediately inoperable.

SURVEILLANCE REQUIREMENTS.

ISTSURVEILLANCE FREQUENCY SR 3.7.4.1 Perform a complete cycle of each ARV. 24 methsSR 3.7.4.2 Verify one complete cycle of each ARV block valve. P4 1NS;mei ET 1L-ýNET 1R.E. Ginna Nuclear Power Plant3.7.4-1Amendment 88 AFW System3.7.5SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.5.1 Verify each AFW and SAFW manual, power operated, 14 -aysand automatic valve in each water flow path, and inboth steam supply flow paths to the turbine drivenpump, that is not locked, sealed, or otherwise securedin position, is in the correct position.

SR 3.7.5.2 -NOTE -Required to be met prior to entering MODE 1 for theTDAFW pump.Verify the developed head of each AFW pump at the In accordance withflow test point is greater than or equal to the required the Inservice developed head. Testing ProgramSR 3.7.5.3 Verify the developed head of each SAFW pump at the In accordance withflow test point is greater than or equal to the required the Inservice developed head. Testing ProgramSR 3.7.5.4 Perform a complete cycle of each AFW and SAFW In accordance withmotor operated suction valve from the Service Water the Inservice System, each AFW and SAFW discharge motor Testing Programoperated isolation valve, and each SAFW cross-tie motor operated valve.SR 3.7.5.5 Verify each AFW automatic valve that is not locked, sealed, or otherwise secured in position, actuates to INSERT 1the correct position on an actual or simulated ractuation signal.SR 3.7.5.6 -- NOTE -Required to be met prior to entering MODE 1 for theTDAFW pump.Verify each AFW pump starts automatically on an 24-mentactual or simulated actuation signal.SR 3.7.5.7 Verify each SAFW train can be actuated and 244- negthecontrolled from the control room. /I!NSERT" R.E. Ginna Nuclear Power Plant3.7.5-3Amendment 88 CSTs3.7.63.73.7.6PLANT SYSTEMSCondensate Storage Tanks (CSTs)LCO 3.7.6APPLICABILITY:

The CSTs shall be OPERABLE.

MODES 1, 2, and 3.ACTIONSCONDITION REQUIRED ACTION COMPLETION TIMEA. CST water volume not A.1 Verify by administrative 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />swithin limit, means OPERABILITY ofbackup water supply.ANDA.2 Restore CST water volume 7 daysto within limit.B. Required Action and B.1 Be in MODE 3. 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />sassociated Completion Time not met. ANDB.2 Be in MODE 4. 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />sSURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.6.1 Verify the CST water volume is > 24,350 gal. 2 h...-...-i==i si T 1IR.E. Ginna Nuclear Power Plant3.7.6-1Amendment 97 CCW System3.7.7CONDITION REQUIRED ACTION COMPLETION TIMED.2 Be in MODE 3. 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />sANDD.3 Be in MODE 4. 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />sSURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.7.1 -NOTE -Isolation of CtW flow to individual components doesnot render the CCW loop header inoperable.

Verify each CCW manual and power operated valve 31-daysin the CCW train and heat exchanger flow path andloop header that is not locked, sealed, or otherwise L NERT1secured in position, is in the correct position.

SR 3.7.7.2 Perform a complete cycle of each motor operated In accordance withisolation valve to the residual heat removal heat the Inservice exchangers.

Testing ProgramR.E. Ginna Nuclear Power Plant3.7.7-2Amendment 80 SW System3.7.8ISURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.8.1 Verify screenhouse bay water level and temperature 24She R1sare within limits.SR 3.7.8.2 ------ NOTE -----------------

Isolation of SW flow to individual components doesnot render the SW loop header inoperable.

Verify each SW manual, power operated, and 1 -..YSautomatic valve in the SW flow path and loop headerthat is not locked, sealed, or otherwise secured in L IERT1lposition, is in the correct position.

SR 3.7.8.3 Verify all SW loop header cross-tie valves are locked , 1.daysin the correct position.

--INSERT.lSR 3.7.8.4 Verify each SW automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual orINERT 1simulated actuation signal.SR 3.7.8.5 Verify each SW pump starts automatically on anactual or simulated actuation signal.Z L. E1R.E. Ginna Nuclear Power Plant3.7.8-2Amendment 102 CREATS3.7.9ICONDITION REQUIRED ACTION COMPLETION TIMED. Required Action and D.1 Place OPERABLE CREATS Immediately associated Completion train in emergency mode.Time of Condition A notmet during movement of ORirradiated fuelassemblies.

D.2 Suspend movement of Immediately irradiated fuel assemblies.

E. Two CREATS trains E.1 Suspend movement of Immediately inoperable during irradiated fuel assemblies.

movement of irradiated fuel assemblies.

OROne or more CREATStrains inoperable due toan inoperable CREboundary duringmovement of irradiated fuel assemblies.

F. Two CREATS trains F.1 Enter LCO 3.0.3. Immediately inoperable in MODE 1, 2,3, or 4 for reasons otherthan Condition B.SURVEILLANCE REQUIREMENTS INSERT 1SURVEILLANCE J FREQUENCY SR 3.7.9.1 Operate each CREATS filtration train > 15 minutes.

,4-,e..SR 3.7.9.2 Perform required CREATS filter testing in accordance In accordance withwith the Ventilation Filter Testing Program (VFTP). the VFTPSR 3.7.9.3 Verify each CREATS train actuates on an actual or 24-ment-hs simulated actuation signal. 'LIIN T1R.E. Ginna Nuclear Power Plant3.7.9-2Amendment 105 ABVS3.7.103.7 PLANT SYSTEMS3.7.10Auxiliary Building Ventilation System (ABVS)LCO 3.7.10APPLICABILITY:

The ABVS shall be OPERABLE and in operation.

During movement of irradiated fuel assemblies in the Auxiliary Buildingwhen one or more fuel assemblies in the Auxiliary Building hasdecayed < 60 days since being irradiated.

ACTIONSCONDITION REQUIRED ACTION COMPLETION TIMEA. ABVS inoperable.

A.1-NOTE -LCO 3.0.3 is not applicable.

Suspend movement of Immediately irradiated fuel assemblies inthe Auxiliary Building.

SURVEILLANCE REQUIREMENTS INSERT1SURVEILLANCE SR 3.7.10.1 Verify ABVS is in operation.

SR 3.7.10.2 Verify ABVS maintains a negative pressure with 244rrespect to the outside environment at the Auxiliary Building operating floor level.SR 3.7.10.3 Perform required Spent Fuel Pool Charcoal Adsorber In accordance withSystem filter testing in accordance with the Ventilation the VFTPFilter Testing Program (VFTP).R.E. Ginna Nuclear Power Plant3.7.10-1Amendment 80 SFP Water Level3.7.113.7PLANT SYSTEMSSpent Fuel Pool (SFP) Water Level3.7.11LCO 3.7.11APPLICABILITY:

The SFP water level shall be >! 23 ft over the top of irradiated fuelassemblies seated in the storage racks.During movement of irradiated fuel assemblies in the SFP.ACTIONSCONDITION REQUIRED ACTION COMPLETION TIMEA. SFP water level not within A.1lim it.-NOTE -LCO 3.0.3 is not applicable.

Suspend movement of Immediately irradiated fuel assemblies inthe SFP.SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.11.1 Verify the SFP water level is 23 ft above the top of 7-deysthe irradiated fuel assemblies seated in the storage INET1racks.R.E. Ginna Nuclear Power Plant3.7.11-1Amendment 80 SFP Boron Concentration 3.7.123.7PLANT SYSTEMSSpent Fuel Pool (SFP) Boron Concentration 3.7.12LCO 3.7.12APPLICABILITY:

The SFP boron concentration shall be > 2300 ppm.Whenever any fuel assembly is stored in the SFP.ACTIONSCONDITION REQUIRED ACTION COMPLETION TIMEA. SFP boron concentration

---

-NOTE -not within limit. LCO 3.0.3 is not applicable.

A. 1 Suspend movement of fuel Immediately assemblies in the SFP.ANDA.2 Initiate action to restore SFP Immediately boron concentration towithin limit.SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.12.1 Verify the SFP pool boron concentration is within limit. 7-deysR.E. Ginna Nuclear Power Plant3.7.12-1Amendment 80 Secondary Specific Activity3.7.143.7PLANT SYSTEMSSecondary Specific Activity3.7.14LCO 3.7.14APPLICABILITY:

The specific activity of the secondary coolant shall be <ý 0. 10 PJCi/gmDOSE EQUIVALENT 1-131.MODES 1, 2, 3, and 4.ACTIONSCONDITION REQUIRED ACTION COMPLETION TIMEA. Specific activity not within A.1 Be in MODE 3. 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />slimit.ANDA.2 Be in MODE 5. 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />sSURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.14.1 Verify the specific activity of the secondary coolant is W-ye 0.10 pCi/gm DOSE EQUIVALENT 1-131. INSERT 1/R.E. Ginna Nuclear Power Plant3.7.14-1Amendment 80 AC Sources -MODES 1, 2, 3, and 43.8.1CONDITION REQUIRED ACTION COMPLETION TIMED. Required Action and D.1 Be in MODE 3. 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />sassociated Completion Time of Condition A, B, or ANDC not met.D.2 Be in MODE 5. 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />sE. Two DGs inoperable.

E.1 Enter LCO 3.0.3. Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.1.1 Verify correct breaker alignment and indicated power 7-deysavailability for the offsite circuit to each of the 480 V LtINSERT 1safeguards buses.SR 3.8.1.2 -NOTE -1. Performance of SR 3.8.1.9 satisfies this SR.2. All DG starts may be preceded by an engineprelube period and followed by a warmup periodprior to loading.Verify each DG starts from standby conditions andachieves rated voltage and frequency.

R.E. Ginna Nuclear Power Plant3.8.1-3Amendment 109 AC Sources -MODES 1, 2, 3, and 43.8.1SURVEILLANCE FREQUENCY SR 3.8.1.3-NOTE -1. DG loadings may include gradual loading asrecommended by the manufacturer.

2. Momentary transients outside the load range donot invalidate this test.3. This Surveillance shall be conducted on onlyone DG at a time.4. This SR shall be preceded by and immediately follow without shutdown a successful performance of SR 3.8.1.2 or SR 3.8.1.9.Verify each DG is synchronized and loaded andoperates for _> 60 minutes and < 120 minutes at a load> 2025 kW and < 2250 kW.31 daysIFýT~TSR 3.8.1.4 Verify the fuel oil level in each day tank. 31-daysSR 3.8.1.5 Verify the DG fuel oil transfer system operates to 31daystransfer fuel oil from each storage tank to theassociated day tank.SR 3.8.1.6 Verify transfer of AC power sources from the 50/50mode to the 100/0 mode and 0/100 mode. SESR 3.8.1.7 -- NOTE -1. This Surveillance shall not be performed inMODE 1, 2, 3, or 4.2. Credit may be taken for unplanned events thatsatisfy this SR. IVerify each DG does not trip during and following a 24-menths load rejection of _ 295 kW.R.E. Ginna Nuclear Power Plant3.8.1-4Amendment 109 AC Sources -MODES 1, 2, 3, and 43.8.1SURVEILLANCE FREQUENCY SR 3.8.1.8-NOTE-1. This Surveillance shall not be performed inMODE 1, 2, 3, or 4.2. Credit may be taken for unplanned events thatsatisfy this SR.Verify each DG automatic trips are bypassed on anactual or simulated safety injection (SI) signal except:a. Engine overspeed;

b. Low lube oil pressure; andc. Start failure (overcrank) relay.i17 EýSR 3.8.1.9-NOTE -1. All DG starts may be preceded by an engineprelube period.2. This Surveillance shall not be performed inMODE 1, 2, 3, or 4.3. Credit may be taken for unplanned events thatsatisfy this SR.Verify on an actual or simulated loss of offsite powersignal in conjunction with an actual or simulated SIactuation signal:a. De-energization of 480 V safeguards buses;b. Load shedding from 480 V safeguards buses;andc. DG auto-starts from standby condition and:1. energizes permanently connected loads,2. energizes auto-connected emergency loads through the load sequencer, and3. supplies permanently and auto-connected emergency loads for > 5 minutes.24 menthsR.E. Ginna Nuclear Power Plant3.8.1-5Amendment 109 Diesel Fuel Oil3.8.3SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.3.1 Verify each fuel oil storage tank contains

> 5000 gal ofdiesel fuel oil for each required DG.SR 3.8.3.2 Verify fuel oil properties of new and stored fuel oil are In accordance withtested in accordance with, and maintained within the the Diesel Fuel Oillimits of, the Diesel Fuel Oil Testing Program.

Testing ProgramR.E. Ginna Nuclear Power Plant3.8.3-2Amendment 80 DC Sources -MODES 1, 2, 3, and 43.8.4SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.4.1 Verify battery terminal voltage is > 129 V on floatcharge. / ISRSR 3.8.4.2 -NOTE -1. SR 3.8.4.3 may be performed in lieu of SR3.8.4.2.2. This Surveillance shall not be performed inMODE 1, 2, 3, or4.Verify battery capacity is adequate to supply, and 24 fneothsmaintain in OPERABLE status, the requiredemergency loads for the design duty cycle whensubjected to a battery service test.SIR 3.8.4.3-NOTE -This Surveillance shall not be performed in MODE 1,2, 3, or4.Verify battery capacity is > 80% of the manufacturer's rating when subjected to a performance discharge test.6E) n~enthsAND12 months whenbattery showsdegradation, or hasreached 85% ofexpected life withcapacity

< 100% ofmanufacturer's ratingAND24 months whenbattery has reached85% of theexpected life withcapacity

_> 100% ofmanufacturer's ratingR.E. Ginna Nuclear Power Plant3.8.4-2Amendment 80 3.8Battery Cell Parameters 3.8.6ELECTRICAL POWER SYSTEMSBattery Cell Parameters 5 Battery cell parameters for Train A and Train B batteries shall be withinlimits.3.8.6LCO 3.8.EAPPLICABILITY:

MODES 1, 2, 3, and 4,When associated DC electrical power sources are required to beOPERABLE by LCO 3.8.5, "DC Sources -MODES 5 and 6."ACTIONS-NOTE -Separate Condition entry is allowed for each battery.CONDITION REQUIRED ACTION COMPLETION TIMEA. One or more batteries A.1 Declare associated battery Immediately with one or more battery inoperable.

cell parameters not withinlimits.SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.6.1 Verify electrolyte level of each connected battery cellis above the top of the plates and not overflowing.

SR 3.8.6.2 Verify the float voltage of each connected battery cell 31-daysis > 2.07 V. ---IINSET1SR 3.8.6.3 Verify specific gravity of the designated pilot cell in 31-dayseach battery is >_ 1.195. 'I-IINSERT SR 3.8.6.4 Verify average electrolyte temperature of the 31-daysdesignated pilot cell in each battery is _> 550F. /'-IINSERT 1SR 3.8.6.5 Verify average electrolyte temperature of every fifthcell of each battery is _> 550F. ' 5 INET1R.E. Ginna Nuclear Power Plant3.8.6-1Amendment 80 Battery Cell Parameters 3.8.6SURVEILLANCE FREQUENCY SR 3.8.6.6 Verify specific gravity of each connected battery cell 92-deysis:is:,'LIINSERT 1 1a. Not more than 0.020 below average of allconnected cells, andb. Average of all connected cells is _> 1.195.R.E. Ginna Nuclear Power Plant3.8.6-2Amendment 80 AC Instrument Bus Sources -MODES 1, 2, 3, and 43.8.7CONDITION REQUIRED ACTION COMPLETION TIMED. Two or more required D.1 Enter LCO 3.0.3. Immediately instrument bus sourcesinoperable.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.7.1 Verify correct static switch alignment to Instrument Bus A and C.SR 3.8.7.2 Verify correct Class 1 E CVT alignment to Instrument Bus B. 'L fj jJR.E. Ginna Nuclear Power Plant3.8.7-2Amendment 80 AC Instrument Bus Sources -MODES 5 and 63.8.8SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.8.1 Verify correct static switch alignment to required AC 7-daysinstrument bus(es).SR 3.8.8.2 Verify correct Class 1E CVT alignment to the required 7-0eysAC instrument bus. L-IY 1R.E. Ginna Nuclear Power Plant3.8.8-2Amendment 112 Distribution Systems -MODES 1, 2, 3, and 43.8.9SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.9.1 Verify correct breaker alignments and voltage to 7 dpysrequired electrical power trains.T INSERT 1R.E. Ginna Nuclear Power Plant3.8.9-2Amendment 80 Distribution Systems -MODES 5 and 63.8.10CONDITION REQUIRED ACTION COMPLETION TIMEA.2.1 Suspend COREALTERATIONS.

ANDA.2.2 Suspend movement ofirradiated fuel assemblies.

ANDA.2.3 Suspend operations involving positive reactivity additions that could result inloss of required SDM orboron concentration.

ANDA.2.4 Initiate actions to restorerequired electrical powerdistribution train(s) toOPERABLE status.ANDA.2.5 Declare associated requiredresidual heat removalloop(s) inoperable and notin operation.

Immediately I Immediately Immediately Immediately Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.10.1 Verify correct breaker alignments and voltage to 7 daysrequired electrical power distribution trains. !NSERT 1R.E. Ginna Nuclear Power Plant3.8.10-2Amendment 112 Boron Concentration 3.9.13.9REFUELING OPERATIONS Boron Concentration 3.9.1LCO 3.9.1APPLICABILITY:

Boron concentrations of the Reactor Coolant System, the refueling canal,and the refueling cavity shall be maintained within the limit specified inthe COLR.MODE 6.ACTIONSCONDITION REQUIRED ACTION COMPLETION TIMEA. Boron concentration not A.1 Suspend CORE Immediately within limit. ALTERATIONS.

ANDA.2 Suspend positive reactivity Immediately additions.

ANDA.3 Initiate action to restore Immediately boron concentration towithin limit.SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.1.1 Verify boron concentration is within the limit specified 721h-5sin the COLR. i INSERT 1R.E. Ginna Nuclear Power Plant3.9.1-1Amendment 80 Nuclear Instrumentation 3.9.2CONDITION REQUIRED ACTION COMPLETION TIMEC.2 Suspend positive reactivity Immediately additions.

ANDC.3 Perform SR 3.9.1.1 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />sANDOnce per 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />sthereafter SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.2.1 Perform CHANNEL CHECK. 42 heuFisSR 3.9.2.2 ----------------------------------

---

INSERT 1-NOTE -Neutron detectors are excluded from CHANNELCALIBRATION.

kINSET1Perform CHANNEL CALIBRATION.

R.E. Ginna Nuclear Power Plant3.9.2-2Amendment 112 Containment Penetrations 3.9.3CONDITION REQUIRED ACTION COMPLETION TIMEA.2 Suspend movement of Immediately irradiated fuel assemblies within containment.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.3.1 Verify each required containment penetration is in therequired status. INERTIKSR 3.9.3.2 Verify each required containment purge and exhaust 24 methsvalve actuates to the isolation position on an actual or NSERT 1simulated actuation signal.R.E. Ginna Nuclear Power Plant3.9.3-2Amendment 107 RHR and Coolant Circulation

-Water Level > 23 Ft3.9.4CONDITION REQUIRED ACTION COMPLETION TIMEA.4 Close all containment 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />spenetrations providing directaccess from containment atmosphere to outsideatmosphere.

SURVEILLANCE REQUIREMENTS SURVEILLANCE

{ FREQUENCY SR 3.9.4.1Verify one RHR loop is in operation and circulating reactor coolant.12 heUiFjNS-RT 1R.E. Ginna Nuclear Power Plant3.9.4-2Amendment 112 RHR and Coolant Circulation

-Water Level < 23 Ft3.9.5SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.5.1 Verify one RHR loop is in operation and circulating 1" " .reactor coolant.

-'NSERT1SR 3.9.5.2 Verify correct breaker alignment and indicated power l-ysavailable to the required RHR pump that is not in INSERT 1operation.

R.E. Ginna Nuclear Power Plant3.9.5-2Amendment 112 3.9Refueling Cavity Water Level3.9.6REFUELING OPERATIONS Refueling Cavity Water LevelRefueling cavity water level shall be maintained

_> 23 ft above the top ofreactor vessel flange.3.9.6LCO 3.9.(APPLICABILITY:

During movement of irradiated fuel assemblies within containment, During CORE ALTERATIONS, except during latching and unlatching ofcontrol rod drive shafts.ACTIONSCONDITION REQUIRED ACTION COMPLETION TIMEA. Refueling cavity water A.1 Suspend CORE Immediately level not within limit. ALTERATIONS.

ANDA.2 Suspend movement of Immediately irradiated fuel assemblies within containment.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.6.1 Verify refueling cavity water level is > 23 ft above the 24 hefurStop of reactor vessel flange. INSERT 1R.E. Ginna Nuclear Power Plant3.9.6-1Amendment 80 Programs and Manuals5.5e. The quantitative limits on unfiltered air inleakage into the CRE.These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described inparagraph

c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basisanalyses of DBA consequences.

Unfiltered air inleakage limits forhazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.f. The provisions of SR 3.0.2 are applicable to the Frequencies forassessing CRE habitability and determining CRE unfiltered inleakage as required by paragraph c.ýFýR.E. Ginna Nuclear Power Plant5.5-13Amendment 110 ATTACHMENT 4License Amendment RequestR. E. Ginna Nuclear Power PlantDocket No. 50-244Application 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)Proposed Technical Specification Bases Page Changes(NOTE: TS Bases pages below marked with an asterisk

(*) do not contain any mark-ups.

These pages are provided for completeness and for information purposes only.)B 3.1.1-5 B 3.3.1-46 B 3.4.9-4 B 3.6.2-7 B 3.7.14-3 B 3.9.6-3B 3.1.2-5 B 3.3.1-47 B 3.4.11-7 B 3.6.3-11 B 3.8.1-12B 3.1.4-8 B 3.3.2-31 B 3.4.12-10*

B 3.6.3-12 B 3.8.1-13B 3.1.4-9 B 3.3.2-32 B 3.4.12-11 B 3.6.3-14 B 3.8.1-14B 3.1.5-5 B 3.3.2-33 B 3.4.12-12 B 3.6.4-3 B 3.8.1-15B 3.1.6-6 B 3.3.2-34 B 3.4.12-13 B-3.6.5-3 B 3.8.1-16B 3.1.8-7 B 3.3.3-16 B 3.4.13-4*

B 3.6.6-8 B 3.8.3-3B 3.1.8-8 B 3.3.3-17 B 3.4.13-5 B 3.6.6-9 B 3.8.4-6B 3.2.1-9 B 3.3.4-7 B 3.4.13-6 B 3.6.6-10 B 3.8.4-7B 3.2.1-10 B 3.3.5-8 B 3.4.14-5*

B 3.6.6-11 B 3.8.4-8B 3.2.1-11 B 3.3.5-9 B 3.4.14-6 B 3.7.2-6 B 3.8.6-3B 3.2.2-5*

B 3.3.6-7 B 3.4.14-7 B 3.7.4-4 B 3.8.6-4B 3.2.2-6 B 3.3.6-8 B 3.4.15-5 B 3.7.5-8 B 3.8.7-6B 3.2.3-3 B 3.4.1-4 B 3.4.15-6*

B 3.7.5-9 B 3.8.8-5B 3.2.4-5 B 3.4.1-5 B 3.4.16-4 B 3.7.5-10 B 3.8.9-9B 3.2.4-6 B 3.4.2-3 B 3.4.16-5 B 3.7.6-3 B 3.8.10-6B 3.3.1-39*

B 3.4.3-6 B 3.5.1-6 B 3.7.7-6 B 3.9.1-4B 3.3.1-40 B 3.4.4-3 B 3.5.1-7 B 3.7.8-7 B 3.9.2-3*B 3.3.1-41 B 3.4.5-5 B 3.5.2-11 B 3.7.8-8 B 3.9.2-4B 3.3.1-42 B 3.4.5-6 B 3.5.2-12 B 3.7.9-6 B 3.9.3-4B 3.3.1-43 B 3.4.6-5 B 3.5.2-13 B 3.7.10-4 B 3.9.4-4B 3.3.1-44 B 3.4.7-5 B 3.5.4-4 B 3.7.11-3 B 3.9.5-4B 3.3.1-45 B 3.4.8-4 B 3.5.4-5 B 3.7.12-3 B 3.9.6-2*

SDMB 3.1.1In the determination of the required combination of boration flow rate andboron concentration, there is no unique requirement that must besatisfied.

Since it is imperative to raise the boron concentration of theRCS as soon as possible, the flowpath of choice would utilize a highlyconcentrated

solution, such as that normally found in the boric acidstorage tank, or the refueling water storage tank. The operator shouldborate with the best source available for the plant conditions.

In determining the boration flow rate, the time in core life must beconsidered.

For instance, the most difficult time in core life to increasethe RCS boron concentration is at the beginning of cycle when the boronconcentration may approach or exceed 2000 ppm. Assuming that avalue of 1% Ak/k must be recovered and a boration flow rate of 10 gpmusing 13,000 ppm boric acid solution, it is possible to increase the boronconcentration of the RCS by 100 ppm in approximately 35 minutes.

If aboron worth of 10 pcm/ppm is assumed, this combination of parameters will increase the SDM by 1% Ak/k. These boration parameters of 10 gpmand 13,000 ppm represent typical values and are provided for thepurpose of offering a specific example.SURVEILLANCE REQUIREMENTS SR 3.1.1.1In MODE 2 with Keff < 1.0 and MODES 3, 4, and 5, theSDM is verified bycomparing the RCS boron concentration to a SHUTDOWN MARGINrequirement curve that was generated by taking into account estimated RCS boron concentrations, core power defect, control bank position, RCS average temperature, fuel burnup based on gross thermal energygeneration, xenon concentration, samarium concentration, andisothermal temperature coefficient (ITC).JINSERT 3iD--ýThe ef 24 heU.S is based en the slow inrcquirod beron eeneentratieon and the l6W probab3ility of E1n accidont~curing ithout the Fcguircd 8DM. This allows time for the eporater toeollect thorcquircd data, which ineludcs peoferminig a boronleemccntration

analysis, and eemplcto the ealeuletien.

REFERENCES

1. Atomic Industrial Forum (AIF) GDC 27 and 28, Issued forcomment July 10, 1967.2. "American National Standard Nuclear Safety Criteria for the Designof Stationary Pressurized Water Reactor Plants,"

N18.2-1973.

3. UFSAR, Section 15.1.5.4. UFSAR, Section 15.4.4.R.E. Ginna Nuclear Power PlantB 3.1.1-5Revision 42 Core Reactivity B 3.1.2B._1If the core reactivity cannot be restored to within the 1% Ak/k limit, or ifthe Required Actions of Condition A cannot be completed within theassociated Completion Time, the plant must be brought to a MODE inwhich the LCO does not apply. To achieve this status, the plant must bebrought to at least MODE 2 with Keff < 1.0 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. If the SDM forMODE 2 with Keff < 1.0 is not met, then the boration required by SR3.1.1.1 would occur. The allowed Completion Time is reasonable, basedon operating experience, for reaching MODE 2 with Keff < 1.0 from fullpower conditions in an orderly manner and without challenging plantsystems.SURVEILLANCE SR 3.1.2.1REQUIREMENTS Core reactivity must be verified following operations that could havealtered core reactivity (e.g., fuel movement, control rod replacement, control rod shuffling).

The comparison must be made prior to enteringMODE 1 when the core conditions such as control rod position, moderator temperature, and samarium concentration are fixed or stable.Since the reactor must be critical to verify core reactivity, it is acceptable to enter MODE 2 with Keff > 1.0 to perform this SR. This SR is modifiedby a Note to clarify that the SR does not need to be performed until priorto entering MODE 1.SR 3.1.2.2Core reactivity is verified by periodic comparisons of measured andpredicted RCS boron concentrations.

The comparison is made,considering that other core conditions are fixed or stable, including control rod position, moderator temperature, fuel temperature, fueldepletion, xenon concentration, and samarium concentration.

T:he-Frequemey of 31 EFPE), ia acceptable, based en the slow rate of coreehanges due to fuel depletion and the prcsenee ef ether indicat0rs (QPT-R, AF), et.) fr PrO.Mpt indic.tio

..f an ano.maly.

The SR ismodified by two Notes. The first Note states that the SR is only requiredafter 60 effective full power days (EFPD). The second Note indicates thatthe normalization of predicted core reactivity to the measured value musttake place within the first 60 EFPD after each fuel loading.

This allowssufficient time for core conditions to reach steady state, but preventsoperation for a large fraction of the fuel cycle without establishing abenchmark for the design calculations.

R.E. Ginna Nuclear Power PlantB 3.1.2-5Revision 21 Rod Group Alignment LimitsB 3.1.4D.2If more than one rod is found to be misaligned or becomes misaligned because of bank movement, the plant conditions fall outside of theaccident analysis assumptions.

Since automatic bank sequencing wouldcontinue to cause misalignment, the plant must be brought to a MODE orCondition in which the LCO requirements are not applicable.

To achievethis status, the plant must be brought to at least MODE 2 with Keff < 1.0within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.The allowed Completion Time is reasonable, based on operating experience, for reaching MODE 2 with Keff < 1.0 from full powerconditions in an orderly manner and without challenging plant systems.SURVEILLANCE SR 3.1.4.1REQUIREMENTS VWeifiertion that individual rod poSitionS Wer wthin alignCint limits usingMRPconr the PPCS at a F betweenty nf 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> prvidcs a histFry that.allews the operators to dtct a rod that is beginning t dcviatoe fremn itsmpontod pesitien.

This Frequency takes into account other rod pcsitienDoininformation that is continuously available to the eoratera in the eeotrolnrolom, that du rirg a atual rod m otion, d eviations can imcandietly beimediatelyb eece.IIISR SR 3.1.4.2 moifiedT 3When the rod position deviation monitor (i.e., the PPCS) is inoperable, nocontrol room alarm is available between the normala12 het* Frequency toalert the operators of a rod misalignment.

A reduction of the Frequency te-4 heHF. provides sufficient monitoring of the rod positions when themonitor is inoperable.

This Frequency takes into account other rodposition information that is continuously available to the operator in thecontrol room, so that during actual rod motion, deviations canimmediately be detected.I-NSERT3This SR is modified by a Note that states that performance of this SR isonly necessary when the rod position deviation monitor is inoperable.

SR 3.1.4.3Verifying each control rod is OPERABLE would require that each rod betripped.

However, in MODES 1 and 2with Keff ! 1.0, tripping each controlrod would result in radial or axial power tilts, or oscillations.

Exercising each individual control rod every-92-days provides increased confidence that all rods continue to be OPERABLE without exceeding the alignment limit, even if they are not regularly tripped.

Moving each control rod to aMRPI transition will not cause radial or axial power tilts, or oscillations, tooccur. The 92 rEuRn3 y takes into considcratien ethe informatio R.E. Ginna Nuclear Power PlantB 3.1.4-8Revision 60 Rod Group Alignment LimitsB 3.1.4available t3 the -ep..t..

in the c.ntrol r...m and "SR 3.1.4.1, whihipcrffrmed moroe frquontly and adds to the detcrmfinatien of of thc rod During or between required performances ofSR 3.1.4.3 (determination of control rod OPERABILITY by movement),

ifa control rod(s) is discovered to be immovable, but remains trippable andaligned, the control rod(s) is considered to be OPERABLE.

At any time, ifa control rod(s) is immovable, a determination of the trippability (OPERABILITY) of the control rod(s) must be made, and appropriate action taken.SR 3.1.4.4Verification of rod drop times allows the operator to determine that themaximum rod drop time permitted is consistent with the assumed roddrop time used in the safety analysis.

Measuring rod drop times prior toreactor criticality, after reactor vessel head removal, ensures that thereactor internals and rod drive mechanism will not interfere with rodmotion or rod drop time, and that no degradation in these systems hasoccurred that would adversely affect control rod motion or drop time. Thistesting is performed with both RCPs operating and the averagemoderator temperature

_> 500OF to simulate a reactor trip under actualconditions.

This Surveillance is performed during a plant outage, due to the plantconditions needed to perform the SR and the potential for an unplanned plant transient if the Surveillance were performed with the reactor atpower.REFERENCES

1. Atomic Industrial Forum (AIF) GDC 6, 14, 27, and 28, Issued forcomment July 10, 1967.2. 10 CFR 50.46.3. UFSAR, Chapter 15.4. UFSAR, Section 15.4.6.5. UFSAR, Section 15.1.5.6. UFSAR, Section 15.4.2.R.E. Ginna Nuclear Power PlantB 3.1.4-9Revision 60 Shutdown Bank Insertion LimitB 3.1.5plant to remain in an unacceptable condition for an extended period oftime.B._ 1If Required Actions A.1 and A.2 cannot be completed within theassociated Completion Times, the plant must be brought to a MODEwhere the LCO is not applicable.

To achieve this status, the plant mustbe placed in MODE 2 with keff < 1.0 within a Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.The allowed Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, based onoperating experience, for reaching the required MODE from full powerconditions in an orderly manner and without challenging plant systems.SURVEILLANCE REQUIREMENTS SR 3.1.5.1 JINSERT I-7Since the shutdown bank is positioned manually by the control roomoperator, a verification of shutdown bank position at a Frequency ef-evety12 he'rs is adequate to ensure that the bank is within the insertion limit.Also, the 12 hIur y takes into Ith l inlformtiIn available in the eontrol rMOM for the purpese of moneitering the status ofshutdeWn redS-.REFERENCES

1. Atomic Industrial Forumcomment July 10, 1967.(AIF) GDC 27, 28, 29, and 32, Issued for2. 10 CFR 50.46.3. UFSAR, Chapter 15.4. UFSAR, Section 15.1.5.5. UFSAR, Section 15.4.1.6. UFSAR, Section 15.4.2.7. UFSAR, Section 15.4.6.R.E. Ginna Nuclear Power PlantB 3.1.5-5Revision 60 Control Bank Insertion LimitsB 3.1.6SURVEILLANCE SR 3.1.6.1REQUIREMENTS This Surveillance is required to ensure that the reactor does not achievecriticality with the control banks below their insertion limits. TheFrequency 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 theestimated control bank position is within the limits specified in the COLRshortly before criticality is reached.SR 3.1.6.2..ith an OPERABLE bank inscrtien limit molnitor (i.e., the control boardannunciatorS, Yerification of the control bank insertion limits aFrcgueney of 12 hourS is sufflicint to cnSUrc OPERABILITY of the beni nscrtion limit moneiter and to detect eontrol banks that may beapproaching the insertien limfits sinee, normally, Ycr; littlc red moltionloccur-s In 12 hourS.SR 3.1.6.3When the insertion limit monitor (i.e., the control board annunciatcrsl becomes inoperable, no control room alarm is available between the-norm.al 12 hOUr f" qu.n.y to alert the operators of a control bank notwithin the insertion limits. A reduction of the Frequency to every 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />sprovides sufficient monitoring of control rod insertion when the monitor isinoperable.

Verification of the control bank position at a Frequency-heUFrS is sufficient to detect control banks that may be approaching theinsertion limits. /tINSERT3]

INSERT/ 1This SR is modified by a Note that states that performance of this SR inonly necessary when the rod insertion limit monitor is inoperable.

SR 3.1.6.4When control banks are maintained within their insertion limits asrequired by SR 3.1.6.2 and SR 3.1.6.3 above, it is unlikely that theirsequence and overlap will not be in accordance with requirements provided in the COLR. A Ir...u.ncy of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is .. nsist.nt with the^insertion limi-t abov- in SR 3.1.6.2.R.E. Ginna Nuclear Power PlantB 3.1.6-6Revision 60 PHYSICS TESTS Exceptions

-MODE 2B 3.1.8D.1If Required Action C.1 cannot be completed within the associated Completion Time, the plant must be brought to a MODE in which therequirement does not apply. To achieve this status, the plant must bebrought to at least MODE 3 within an additional 15 minutes.

TheCompletion Time of 15 additional minutes is reasonable, based onoperating experience, for reaching MODE 3 from MODE 2 in an orderlymanner and without challenging plant systems.SURVEILLANCE SR 3.1.8.1REQUIREMENTS The power range and intermediate range neutron detectors must beverified to be OPERABLE in MODE 2 by LCO 3.3.1, "Reactor TripSystem (RTS) Instrumentation."

A CHANNEL OPERATIONAL TEST isperformed on each power range and intermediate range channel within 7days prior to criticality.

This will ensure that the RTS is properly aligned toprovide the required degree of core protection during the performance ofthe PHYSICS TESTS. The 7 day time limit is sufficient to ensure that theinstrumentation is OPERABLE shortly before initiating PHYSICS TESTS.SR 3.1.8.2Verification that the RCS lowest loop Tavg is _> 530'F will ensure that theplant is not operating in a condition that could invalidate the safetyanalyses.

Control board indication for Tavg is available down to 5401Fwhile indication from the plant process computer (PPCS) is available down to 5350F. Between 530°F and 5350F, PPCS cold and hot legindication should be used to determine Tavg.Verifleaticn of the RCS tempefraturo at a Frcgucney ef 30 m~inutes durinlgthe porfformanec of the PHYSICS TESTS will eMSurc that the ntaeendkoecis ef the safety analyses arnt iatdSR 3.1.8.3Verification that THERMAL POWER is: 5% RTP using the NIS detectors will ensure that the plant is not operating in a condition that couldinvalidate the safety analyses.

Verifeatien ef the THERMAL POWERaa Froqueney ef 30 mninutes during the peoreformanc ef the PHYSICSTESTS will onAuro that the minitial eenditieno of the safe" analyses are net16 ele d.e t- ---" .R.E. Ginna Nuclear Power PlantB 3.1.8-7Revision 34 PHYSICS TESTS Exceptions

-MODE 2B 3.1.8SR 3.1.8.4The SDM is verified by comparing the RCS boron concentration to aSHUTDOWN MARGIN requirement curve that was generated by takinginto account estimated RCS boron concentrations, core power defect,control bank position, RCS average temperature, fuel burnup based ongross thermal energy generation, xenon concentration, samariumconcentration, and isothermal temperature coefficient (ITC).The Frcequoney of 24 heurs is based en the gonorally slew chango inrcguirod beren eenecntraticn and en the low probability of an aee6idont

~curing ithout the roquwirod 69M.REFERENCES

1. 10 CFR 50, Appendix B, Section XI.2. 10 CFR 50.59.3. WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology Report,"

July 1985.4. UFSAR, Section 14.6.5. Letter from R. W. Kober (RGE) to T. E. Murley (NRC),

Subject:

"Startup Reports,"

dated July 9, 1984.6. Letter from J. P. Durr (NRC) to B. A. Snow (RGE),

Subject:

"Inspection Report No. 50-244/88-06,"

dated April 28, 1988.R.E. Ginna Nuclear Power PlantB 3.1.8-8Revision 34 FQ(Z)B 3.2.1SR 3.2.1.1Verification that FQC(Z) is within its specified limits involves increasing FQM(Z) to allow for manufacturing tolerance and measurement uncertainties in order to obtain FQC(Z). Specifically, FQM(Z) is themeasured value of FQ(Z) obtained from incore flux map results andFQC(Z) = FQM(Z) 1.0815 (Ref. 4). FQc(Z)is then compared to itsspecified limits.The limit with which FQC(Z) is compared varies inversely with powerabove 50% RTP and directly with a function called K(Z) provided in theCOLR.Performing this Surveillance in MODE 1 prior to exceeding 75% RTPensures that the FQC(Z)limit is met when RTP is achieved, becausepeaking factors generally decrease as power level is increased.

If THERMAL POWER has been increased by > 10% RTP since the lastdetermination of FQc(Z), another evaluation of this factor is required 12hours after achieving equilibrium conditions at this higher power level (toensure that FQC(Z) values are being reduced sufficiently with powerincrease to stay within the LCO limits).The

.f 31 EFPD is adequate t, menitr the ,hangI ef pewredliStributien with cerc burnup bcoausc such ehangco arc slew and we"lecntrolled when the plant is eperotcd in aecordanee with the Tcchnical Spocifloations (TS).~SR 3.2.1.2 t- ERT 3IThe nuclear design process includes calculations performed to determine that the core can be operated within the FQ(Z) limits. Because flux mapsare taken in steady state conditions, the variations in power distribution resulting from normal operational maneuvers are not present in the fluxmap data. These variations are, however, conservatively calculated byconsidering a wide range of unit maneuvers in normal operation.

Themaximum peaking factor increase over steady state values, calculated asa function of core elevation, Z, is called W(Z). Multiplying the measuredtotal peaking factor, FQC(Z), by W(Z) gives the maximum FQ(Z)calculated to occur in normal operation, FQW(Z).R.E. Ginna Nuclear Power PlantB 3.2.1-9Revision 42 FQ(Z)B 3.2.1The limit with which FQW(Z) is compared varies inversely with powerabove 50% RTP and directly with the function K(Z) provided in the COLR.The W(Z) curve is provided in the COLR for discrete core elevations.

Flux map data are typically taken for 61 core elevations.

FQW(Z)evaluations are not applicable for the following axial core regions,measured in percent of core height:a. Lower core region, from 0 to 8% inclusive andb. Upper core region, from 92 to 100% inclusive.

The top and bottom 8% of the core are excluded from the evaluation because of the low probability that these regions would be more limitingin the safety analyses and because of the difficulty of making a precisemeasurement in these regions.This Surveillance has been modified by a Note that may require thatmore frequent surveillances be performed.

If FQW(Z) is evaluated, anevaluation of the expression below is required to account fcr any increaseto FQM(Z) that may occur and cause the FQ(Z) limit to be exceededbefore the next required FQ(Z) evaluation.

If the two most recent FQ(Z) evaluations show an increase in theexpression maximum over z [FQC(Z) / K(Z) ], it is required to meet theFQ(Z) limit with the last FQW(Z) increased by the greater of a factor of1.02 or by an appropriate factor specified in the COLR or to evaluateFQ(Z) more frequently, each 7 EFPD. These alternative requirements prevent FQ(Z) from exceeding its limit for any significant period of timewithout detection.

Performing the Surveillance in MODE 1 prior to exceeding 75% RTPensures that the FQ(Z) limit is met when RTP is achieved, becausepeaking factors are generally decreased as power level is increased.

FQ(Z) is verified at power levels >_ 10% RTP above the THERMALPOWER of its last verification, 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after achieving equilibrium conditions to ensure that FQ(Z) is within its limit at higher power levels.The Gur.....ane.

Fro.uen, y ef 31 EFPD is adequate to monitfr theehange of pewer di~tributien with eero burnup. Thc Surveiilonoo may bedcne maer- frcguently if rcquircd by the rcoults ef F (Z)evluatie.

R.E. Ginna Nuclear Power PlantB 3.2.1-10Revision 42 FQ(Z)B 3.2.1The Frogucncy ef 31 EFPID is adequate to moenitor the changc of pewcrdiStribution beopuse sueh a ehango is sufflciently slew, when the plant isepcraitcd in eccordanco with the TS, to proolude adverse peaking faetefrsbetween 31 day supvcillanccs.

REFERENCES 1 .10 CFR 50.46.2. UFSAR 15.4.5.4.3

3. Atomic Industrial Forum (AIF) GIDC-29, Issued for comment July10, 19674. WCAP-7308-L-P-A, "Evaluation of Nuclear Hot Channel FactorUncertainties,"

June 1988.5. WCAP-1 0216-P-A, Rev. 1lA, "Relaxation of Constant Axial OffsetControl (and) FQ Surveillance Technical Specification,"

February1994.R.E. Ginna Nuclear Power PlantB 3.2.1-11Revision 42 FNAH3.2.2A.3Reduction in the Overpower AT and Overtemperature AT trip setpoints by> 1% for each 1% by which FNAH exceeds its limit, ensures thatcontinuing operation remains at an acceptable low power level withadequate DNBR margin. The Completion Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is sufficient considering the small likelihood of a severe transient in this period, andthe preceding prompt reduction in THERMAL POWER in accordance with Required Action A.1.A.4Verification that FNAH has been restored within its limit by performing SR3.2.2.1 or SR 3.2.2.2 prior to increasing THERMAL POWER above thelimit imposed by Required Action A.1 ensures that the cause that led tothe FNAH exceeding its limit is corrected, and core conditions duringoperation at higher power levels are consistent with safety analysesassumptions.

B. 1If the Required Actions of A.1 through A.4 cannot be met within theirassociated Completion Times, the plant must be placed in a mode inwhich the LCO requirements are not applicable.

This is done by placingthe plant in at least MODE 2 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.The allowed Completion Time is reasonable based on operating experience regarding the amount of time it takes to reach MODE 2 fromfull power operation in an orderly manner and without challenging plantsystems.SURVEILLANCE SR 3.2.2.1REQUIREMENTS The value of FNAH is determined by using the movable incore detectorsystem to obtain a flux distribution map. A data reduction computerprogram then calculates the maximum value of FNAH from the measuredflux distributions.

The measured value of FNAH must be multiplied by1.04 to account for measurement uncertainty before making comparisons to the FN AH limit.After each refueling, FNAH must be determined in MODE 1 prior toexceeding 75% RTP. This requirement ensures that FNAH limits are metat the beginning of each fuel cycle.R.E. Ginna Nuclear Power PlantB 3.2.2-5Revision 21 FNAHJINSERT 3 /3.2.2The Frequency of 31 EFPD) is acceptable becauise the power distribution changS relatively slowly ,ver thiS amouint of fuel burnup. AI J elII.lyl, this Frequency

i. sho^t enough that th ,FNAH limnit .ann.t be e..eeded frany signifi.ant pcri. d f operation.

When the plant is already performing SR 3.2.2.2 to satisfy other requirements, SR 3.2.2.2 does not need to besuspended in order to perform SR 3.2.2.1 since the performance of SR3.2.2.2 meets the requirements of SR 3.2.2.1.SR 3.2.2.2During power operation, the global power distribution is monitored byLCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)," and LCO 3.2.4,"QUADRANT POWER TILT RATIO (QPTR),"

which are directly andcontinuously measured process variables.

With an NIS power range channel inoperable, QPTR monitoring for aportion of the reactor core becomes degraded.

Large tilts are likelydetected with the remaining

channels, but the capability for detection ofsmall power tilts in some quadrants is decreased.

Peforming GR 3.2.2.2at a Frequency of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> provides an accurate altornative mneans foreM5UFt~g-that-FNAH remains within limnits and the core power distribution is..ns stent with the safoty analyseos.

A Frequoncy of 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, s ta Ikes intoeansuderation the rate at which peaking factors are likely to ehange, anthe time required to stabilize the plant and performf ak flu1X MaP....... ......., ,, ,, , .... ...-. +INSERT 3 iThis Surveillance is modified by a Note, which states that it is requiredonly when one power range channel is inoperable and the THERMALPOWER is >_ 75% RTP.REFERENCES

1. 10 CFR 50.46.2. UFSAR, Section 15.4.5.1.

3. Atomic Industrial Forum (AIF) GDC 29, Issued for comment July 101967.4. American National

Standard, "Nuclear Safety Criteria for the Designof Stationary Pressurized Water Reactor Plants,"

N18.2-1973.

R.E. Ginna Nuclear Power PlantB 3.2.2-6Revision 21 AFDB 3.2.3APPLICABILITY The AFD requirements are applicable in MODE 1 greater than orequal to50% RTP when the combination of THERMAL POWER and core peakingfactors are of primary importance in safety analysis.

For AFD limits developed using RAOC methodology, the value of theAFD does not affect the limiting accident consequences with THERMALPOWER < 50% RTP and for lower operating power MODES.ACTIONS A.1As an alternative to restoring the AFD to within its specified limits,Required Action A.1 requires a THERMAL POWER reduction to < 50%RTP. This places the core in a condtion for which the value of the AFD isnot important in the applicable safety analyses.

A Completion lime of 30minutes is reasonable, based on operating experience, to reach 50%RTP without challenging plant systems.SURVEILLANCE SR 3.2.3.1REQUIREMENTS This Surveillance verifies that the AFD, as indicated by the NIS excorechannel, is within its specified limits. The F..quen.y ef 7days is adequate eensidering that the AFD as menitorod by a comnputer and any deviatien frcmA roquircments iz Ealarmcfd.-

REFERENCES

1. WCAP-1 0216-P-A, Revision 1A, "Relaxation of Constant AxialOffset Control/FQ Surveillance Technical Specification",

February1994.2. American National

Standard, "Nuclear Safety Criteria for the Designof Stationary Pressurized Water Reactor Plants,"

N18.2-1973.

3. UFSAR, Section 7.7.2.6.4.

R.E. Ginna Nuclear Power PlantB 3.2.3-3Revision 42 QPTRB 3.2.4assumptions, Required Action A.6 requires verification that FQ(Z) asapproximated by FQC(Z) and FQW(Z), and FNAH are within their specified limits within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reaching equilibrium condition at RTP. As anadded precaution, if the core power does not reach equilibrium condition at RTP within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, but it increases slowly, then the peaking factorsurveillances must be performed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after increasing THERMAL POWER above the limit of Required Action A.1. TheseCompletion Times are intended to allow adequate time to increaseTHERMAL POWER to above the limit of Required Action A.1, while notpermitting the core to remain with unconfirmed power distributions forextended periods of time.Required Action A.6 is modified by a Note that states that the peakingfactor surveillances may only be done after the excore detectors havebeen normalized to eliminate the indicated tilt (i.e., Required Action A.5).The intent of this Note is to have the peaking factor surveillances performed at operating power levels, which can only be accomplished after the excore detectors are adjusted to eliminate the indicated tilt andthe core returned to power.B.1If Required Actions A.1 through A.6 are not completed within theirassociated Completion Times, the plant must be brought to a MODE orcondition in which the requirements do not apply. To achieve this status,THERMAL POWER must be reduced to < 50% RTP within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Theallowed Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is reasonable, based on operating experience regarding the amount of time required to reach the reducedpower level without challenging plant systems.SURVEILLANCE SR 3.2.4.1REQUIREMENTS This Surveillance verifies that the QPTR, as indicated by the NuclearInstrumentation System (NIS) excore channels, is within its limits. TFIe-Frcqucncy of 7 days takes int3 acccun~t ethcr in9fefrmatien and 8olormFIavailable in the control roomSR 3.2.4.1 is modified by two Notes. Note 1 allows QPTR to becalculated with three power range channels if THERMAL POWER is< 75% RTP and one power range channel is inoperable.

Note 2 allowsperformance of SR 3.2.4.2 in lieu of SR 3.2.4.1.For those causes of quadrant power tilt that occur quickly (e.g., adropped rod), there typically are other indications of abnormality thatprompt a verification of the core power tilt.R.E. Ginna Nuclear Power PlantB 3.2.4-5Revision 42 QPTRB 3.2.4SR 3.2.4.2This surveillance is modified by a Note, which states that it is not requireduntil 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the input from one or more Power Range NeutronFlux channel is inoperable and the THERMAL POWER is > 75% RTP.With the input from a NIS power range channel inoperable, tilt monitoring for a portion of the reactor core becomes degraded.

Large tilts are likelydetected with the remaining

channels, but the capability for detection ofsmall power tilts in some quadrants is decreased.

When one NIS power range channel input is inoperable and THERMALPOWER is > 75% RTP, a full core flux map should be performed to verifythe core power distribution instead ef using the thrc ,P-ERB- E poc,range chainncl inputs te Yerif; QPT-R by perfeFrming SR 3.2.1.1, SR3.2.1.2 and SR 3.2.2.1, at a Fr ..un.y of 24 heur.. Performing a fullcore flux map provides an accurate alternative means for ensuring thatFQ(Z) and FN AH remain within limits and the core power distribution isconsistent with the safety analysis.

&t-TINET3 REFERENCES

1. 10 CFR 50.46.2. UFSAR, Section 15.4.5.3. Atomic Industrial Forum (AI) GDC 29, Issued for comment July10,1967.R.E. Ginna Nuclear Power PlantB 3.2.4-6Revision 42 RTS Instrumentation B 3.3.1X.1 and X.2If the Required Action and Associated Completion Time of Condition W isnot met, the plant must be placed in a MODE where the Functions are nolonger required.

To achieve this status, action be must initiated immediately to fully insert all rods and the CRD System must beincapable of rod withdrawal within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. These Completion Times arereasonable, based on operating experience to exit the MODE ofApplicability in an orderly manner.SURVEILLANCE The SRs for each RTS Function are identified by the SRs column of TbleREQUIREMENTS 3.3.1-1 for that Function.

A Note has been added to the SR Table stating that Table 3.3.1-1determines which SRs apply to which RTS Functions.

Note that each channel of process protection supplies both trains of theRTS. When testing Channel 1, Train A and Train B must be examined.

Similarly, Train A and Train B must be examined when testing Channel 2,Channel 3, and Channel 4 (if applicable).

The CHANNEL CALIBRATION and COTs are performed in a manner that is consistent with theassumptions used in analytically calculating the required channelaccuracies (Ref. 8).SR 3.3.1.1A CHANNEL CHECK is required for the following RTS trip functions:

• Power Range Neutron Flux-High;

• Power Range Neutron Flux-Low;

• Intermediate Range Neutron Flux;* Source Range Neutron Flux;* Overtemperature AT;* Overpower AT;* Pressurizer Pressure-Low;

• Pressurizer Pressure-High;

• Pressurizer Water Level-High;

• Reactor Coolant Flow-Low (Single Loop);R.E. Ginna Nuclear Power PlantB 3.3.1-39Revision 61 RTS Instrumentation B 3.3.1" Reactor Coolant Flow-Low (Two Loops); and" SG Water Level-Low LowPerformance of the CHANNEL CHECK encooevcy 12 heaps ensures thatgross failure of instrumentation has not occurred.

A CHANNEL CHECKis normally a comparison of the parameter indicated on one channel to asimilar parameter on other channels.

It is based on the assumption thatinstrument channels monitoring the same parameter should readapproximately the same value. Significant deviations between theinstrument channels could be an indication of excessive instrument driftin one of the channels or of more serious instrument conditions.

ACHANNEL CHECK will detect gross channel failure; thus, it is averification that the instrumentation continues to operate properlybetween each CHANNEL CALIBRATION.

Channel check acceptance criteria are determined by the plant staffbased on a combination of the channel instrument uncertainties, including indication and readability.

If a channel is outside the criteria, itmay be an indication that the sensor or the signal processing equipment has drifted outside its limit.The of 12 houra i* based an -- -. .tin--

e that,iemntratcc ehannel failur i3 raro. The CHANNEL CHECsupplements less f., .m1 , but marc irl qucnt, checks i f .hannelo durin1gnormal8 oporaitienal use of the displays asseeiated with the LCOG roqUirodehannels.

SR 3.3.1.2This SR compares the calorimetric heat balance calculation to the NISPower Range Neutron Flux-High channel output e,,epy.24.heta..

If thecalorimetric exceeds the NIS channel output by > 2% RTP, the NIS is stillOPERABLE but must be adjusted.

If the NIS channel output cannot beproperly

adjusted, the channel is then declared inoperable.

This SR is modified by a Note which states that this Surveillance isrequired to be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after power is _ 50% RTP. Atlower power levels, calorimetric data are inaccurate.

The Fr-lqu y of eve; 24 huFrA i, based en plent eensidering@

in19tru~mnt roliability and epcraiting his9tory data forintumn drift. Tegether these factors dcmenStraltc the ehange in the abselutcdifefcroncc between NIS and heat belenec ealeulated pewcrc rarcly-emeeeds 2% en any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> iporiod.R.E. Ginna Nuclear Power PlantB 3.3.1-40Revision 61 RTS Instrumentation B 3.3.1In addition, control room operators periodically monitor redundant indications and alarms to detect deviations in channel outputs.SR 3.3.1.3 JINSERT 3This SR compares the incore system to the NIS channel output eery 31^ffe.tiv.

full p.w.. days (EFP., ). If the absolute difference is > 3%, theNIS channel is still OPERABLE, but must be readjusted.

If the NISchannel cannot be properly readjusted, the channel is then declaredinoperable.

This surveillance is performed to verify the f(AI) input to theOvertemperature AT Function.

This SR is modified by two Notes. Note 1 clarifies that the Surveillance isrequired to be performed within 7 days after THERMAL POWER is> 50%RTP but prior to exceeding 90% RTP following each refueling and if it hasnot been performed within the last 31 EFPD. Note 2 states thatperformance of SR 3.3.1.6 satisfies this SR since it is a morecomprehensive test.Thc Fr.quen.y of ..v.; 31 EFPD is based en plant .p..ating.

cxpriscc eensidering finstrumcnet rcliability and eperating histery datafc ntru mcnt drift. Also, the slew ehano in ctrn flux during the fueleyele ean be deteeted during thiS imterval.

SR 3.3.1.4 E f lThis SR is the performance of a TADOT evcr; 31 days on aSTACCERED TEST BASIS of the RTB, and the RTB Undervoltage andShunt Trip Mechanisms.

This test shall verify OPERABILITY byactuation of the end devices.The test shall include separate verification of the undervoltage and shunttrip mechanisms except for the bypass breakers which do not requireseparate verification since no capability is provided for performing such atest at power. The independent test for bypass breakers is included inSR 3.3. 1. 11. However, the bypass breaker test shall include a local shunttrip. This test must be performed on the bypass breaker prior to placing itin service to take the place of a RTB.based en industry epcraiting cxperiencc, ccnsidcring inStrumcnt roliability and eperating histery data-.R.E. Ginna Nuclear Power PlantB 3.3.1-41Revision 61 RTS Instrumentation B 3.3.1SR 3.3.1.5This SR is the performance of an ACTUATION LOGIC TEST on the RTSAutomatic Trip Logic .....; 31 days en a STAGCERED TEST BASIS.The train being tested is placed in the bypass condition, thus preventing inadvertent actuation.

All possible logic combinations, with and withoutapplicable permissives, are tested for each protection function.

T-heFroqueney ef eoveo; 31 days en a STACCERED TEST BASIS 09 based eoninduatry epefrating expe ionc cnidering inq~trumcnt roliability andepefrating hok*tFRyF data*-.SR 3.3.1.6This SR is a calibration of the excore channels to the incore channelsevwety 92 E -.. If the measurements do not agree, the excore channelsare still OPERABLE but must be calibrated to agree with the incoredetector measurements.

If the excore channels cannot be adjusted, thechannels are then declared inoperable.

This surveillance is performed toverify the f(AI) input to the Overtemperature AT Function.

A minimum of 2 thimbles per quadrant and sufficient movable incoredetectors shall be operable during recalibration of the excore axial off-setdetection system. To calibrate the excore detector

channels, it is onlynecessary that the movable incore system be used to determine thegross power distribution in the core as indicated by the power balancebetween the top and bottom halves of the core.This SR has been modified by a Note stating that this Surveillance isrequired to be performed within 7 days after THERMAL POWER is 50%RTP but prior to exceeding 90% RTP following each refueling.

The Frogueney of 92 EFPD is adequate based ein industry eperating experuencc, considcrinig ino8trumonet roliability and epefrating histery datafor inStFrumcnt drit.SR 3.3.1.7 E f lThis SR is the performance of a COT e%'e~y-92-days-for the following RTSfunctions:

" Power Range Neutron Flux-High;

" Source Range Neutron Flux (in MODE 3, 4, or 5 with CRD Systemcapable of rod withdrawal or all rods not fully inserted);

" Overtemperature AT;* Overpower AT;* Pressurizer Pressure-Low; R.E. Ginna Nuclear Power PlantB 3.3.1-42Revision 61 RTS Instrumentation B 3.3.1* Pressurizer Pressurizer-High;

• Pressurizer Water Level-High;

" Reactor Coolant Flow-Low (Single Loop);* Reactor Coolant Flow-Low (Two Loops); and* SG Water Level-Low LowA COT is performed on each required channel to ensure the channel willperform the intended Function.

The as-found setpoints must be withinthe COT Acceptance Criteria specified within plant procesures.

The as-left values must be consistent with the setting tolerance used in thesetpoint methodology (Ref. 8).This SR is modified by a Note that provides a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> delay in therequirement to perform this surveillance for source range instrumentation when entering MODE 3 from MODE 2. This Note allows a normalshutdown to proceed without a delay for testing in MODE 2 and for ashort time in MODE 3 until the RTBs are open and SR 3.3.1.7 is nolonger required to be performed.

If the plant is in MODE 3 with the RTBsclosed for greater than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, this SR must be performed within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />safter entry into MODE 3.I111mor- I a rl hp Frequeney ef 92 days isee.nsustent with Refccnee .SR 3.3.1.8This SR is the performance of a COT as described in SR 3.3.1.7 for thePower Range Neutron Flux-Low, Intermediate Range Neutron Flux, andSource Range Neutron Flux (MODE 2), except that this test also includesverification that the P-6 and P-10 interlocks are in their required state forthe existing plant condition.

This SR is modified by two Notes thatprovide a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> delay in the requirement to perform this surveillance.

These Notes allow a normal shutdown to be completed and the plantremoved from the MODE of Applicability for this surveillance without adelay to perform the testing required by this surveillance.

The Frequency INSa p p -li e.y 92 days applies if the plant remains in the MODE of Applicability ISR 1j I "after the initial performances of prior to reactor startup and 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> afterreducing power below P-10 or P-6.R.E. Ginna Nuclear Power PlantB 3.3.1-43Revision 61 RTS Instrumentation B 3.3.1The MODE of Applicability for this surveillance is < 6% RTP for the powerrange low and intermediate range channels and < 5E-1lamps for theSource range channels.

Once the plant is in MODE 3, ihis surveillance isno longer required.

If power is to be maintained

< 6% RTP or < 5E-1lamps for more than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, then the testing required by thissurveillance must be performed prior to the expiration of the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> limit,unless perorm ."-heth prr2d .Four hours is a reasonable time to complete the required testing or place the plant in a MODE wherethis surveillance is no longer required.

This test ensures that the NISsource, intermediate, and power range low channels are OPERABLEprior to taking the reactor critical or after reducing power into theapplicable MODE (< 6% RTP or < 5E-1lamps) for periods > 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.SR 3.3.1.9 JINSERT 3This SR is the performance of a TADOT for the Undervoltage-Bus 11 Aand 11B and Underfrequency-Bus 11A and 11B trip Functions.

T:he-Frequeney 1f evev, 92 izs -ensistnt with Rfer1ncc 9.This SR is modified by a Note that excludes verification of setpoints fromthe TADOT. Since this SR applies to Bus 11A and 11B undervoltage andunderfrequency relays, setpoint verification requires elaborate benchcalibration and is accomplished during the CHANNEL CALIBRATION required by SR 3.3.1.10.Ax SR 3.3.1.10"L-'IJSERT3 This SR is the performance of a CHANNEL CALIBRATION for thefollowing RTS Functions:

• Power Range Neutron Flux-High;

• Power Range Neutron Flux-Low;

" Intermediate Range Neutron Flux;" Source Range Neutron Flux;" Overtemperature AT;* Overpower AT;* Pressurizer Pressure-Low;

" Pressurizer Pressure-High;

" Pressurizer Water Level-High;

" Reactor Coolant Flow-Low (Single Loop);R.E. Ginna Nuclear Power PlantB 3.3.1-44Revision 61 RTS Instrumentation B 3.3.1* Reactor Coolant Flow-Low (Two Loops);* Undervoltage-Bus 11A and 11B;* Underfrequency-Bus 11A and 11B;* SG Water Level-Low Low;* Turbine Trip-Low Autostop Oil Pressure; and* Reactor Trip System Interlocks.

A CHANNE=L CALIBRATIGN is pcrfefrmzd evcr; 24 mcenths, erapp...imat.ly at r"fu:ling.

CHANNEL CALIBRATION is acomplete check of the instrument loop, including the sensor. The testverifies that the channel responds to a measured parameter within thenecessary range and accuracy.

CHANNEL CALIBRATIONS must be performed consistent with theassumptions of the plant specific setpoint methodology (Ref. 8). Thedifference between the current as-found values and the previous test as-left values must be consistent with the drift allowance used in the setpointmethodology.

The Frcqucncy ef 24 months is based en the assumptien of 24 monthealibratifin intewersacin the detefrminatien ef the magnitude of equipment drift in the sctpeint mcethedelegy.

With respect to RTDs, whenever a sensing element is replaced, the nextrequired CHANNEL CALIBRATION of the resistance temperature detectors (RTD) sensors shall include an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel.

This is accomplished by an inplace crosscalibration that compares the other sensing elements with the recentlyinstalled sensing element.This SR is modified by a Note stating that neutron detectors are excludedfrom the CHANNEL CALIBRATION.

The CHANNEL CALIBRATION forthe power range neutron detectors consists of a normalization of thedetectors based on a power calorimetric and flux map performed above50% RTP. The CHANNEL CALIBRATION for the source range andintermediate range neutron detectors consists of obtaining the detectorplateau or preamp discriminator curves, evaluating those curves, andcomparing the curves to the manufacturer's data. This Surveillance is notrequired for the NIS power range detectors for entry into MODE 2 or 1,and is not required for the NIS intermediate range detectors for entry intoMODE 2, because the plant must be in at least MODE 2 to perform thetest for the intermediate range detectors and MODE 1 for the powerrange detectors.

The 24 mo.nth F- qu.n. y i. based en the need teCperfel this Su1, ViI, urlllar-.

und1r the Irnditic,,

that apply during a pllntR.E. Ginna Nuclear Power PlantB 3.3.1-45Revision 61 RTS Instrumentation B 3.3.1,utago and the pet.ntial f- r .19 una`ind transi-nt if the

..,, p-"form--d with the Bt p"wr. Op-rating

,epe;Ciono, hassh.wn these .

usually pass theo Sure..l",n.

when pefnrmoden the 24 month Frogucncy.

SR 3.3.1.11 t-tIN.SERT3 This SR is the performance of a TADOT of the Manual Reactor Trip, RCPBreaker Position, and the Sl Input from ESFAS trip Functions.

This-TA.DOT is p..f....m..d evry 24 months. This test independently verifiesthe OPERABILITY of the undervoltage and shunt trip mechanisms for theManual Reactor Trip Function for the Reactor Trip Breakers and ReactorTrip Bypass Breakers.

The Frcqueney is based ong the known roliability of the Functions andl themult..hanncl r"dundanoy available, and has been shown to boaoscptablc through oporating oxporionos.

SR 3.3.1.12

.- ER3This SR is the performance of a TADOT for Turbine Trip Functions whichis performed prior to reactor startup if it has not been performed within thelast 31 days. This test shall verify OPERABILITY by actuation of the enddevices.The Frequency is based on the known reliability of the Functions and themultichannel redundancy available, and has been shown to beacceptable through operating experience.

This SR is modified by a Note stating that verification of the Trip Setpointdoes not have to be performed for this Surveillance.

Performance of thistest will ensure that the turbine trip Function is OPERABLE prior to takingthe reactor critical because portions of this test cannot be performed withthe reactor at power.SR 3.3.1.13This SR is the pcreffcranooe of a COTF of the RTS interlooks evcr; 24The Froguoncy is based on the known roliability of the intcrlocks and thomnultihanncl rcdandenoy available, and has boon shown to bacccptablc through opefrating exporionoc.

I fl l ITI3i *.1U T. l l l V i [R.E. Ginna Nuclear Power PlantB 3.3.1-46Revision 61 RTS Instrumentation B 3.

3.1REFERENCES

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 Designof 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, Suppl^c.,

nt 2, Rey. 1, Jun .I~eR.E. Ginna Nuclear Power PlantB 3.3.1-47Revision 61 ESFAS Instrumentation B 3.3.2SURVEILLANCE REQUIREMENTS The SRs for each ESFAS Function are identified by the SRs column ofTable 3.3.2-1.

Each channel of process protection supplies both trains ofthe ESFAS. When testing Channel 1, Train A and Train B must beexamined.

Similarly, Train A and Train B must be examined when testingChannel 2, Channel 3, and Channel 4 (if applicable).

The CHANNELCALIBRATION and COTs are performed in a manner that is consistent with the assumptions used in analytically calculating the required charnelaccuracies.

A Note has been added to the SR Table to clarify that Table 3.3.2-1determines which SRs apply to which ESFAS Functions.

SR 3.3.2.1This SR is the performance of a CHANNEL CHECK for the following ESFAS Functions:

SI-Containment Pressure-High; SI-Pressurizer Pressure-Low;

• SI-Steam Line Pressure-Low;

" CS-Containment Pressure-High High;* Steam Line Isolation-Containment Pressure-High High;" Steam Line Isolation-High Steam Flow Coincident with SI and Tavg-Low;* Steam Line Isolation-High-High Steam Flow Coincident with SI;* Feedwater Isolation-SG Water Level-High; and* AFW-SG Water Level-Low Low.Performance of the CHANNEL CHECK oncccvcry 12 heur3 ensures thata gross failure of instrumentation has not occurred.

A CHANNEL CHECKis normally a comparison of the parameter indicated on one channel to asimilar parameter on other channels.

It is based on the assumption thatinstrument channels monitoring the same parameter should readapproximately the same value. Significant deviations betweeninstrument channels could be an indication of excessive instrument driftin one of the channels or of more serious instrument conditions.

ACHANNEL CHECK will detect gross channel failure; thus, it is averification the instrumentation continues to operate properly betweeneach CHANNEL CALIBRATION.

R.E. Ginna Nuclear Power PlantB 3.3.2-31Revision 42 ESFAS Instrumentation B 3.3.2CHANNEL CHECK acceptance criteria are determined by the plant staff,based on a combination of the channel instrument uncertainties, including indication and readability.

If a channel is outside the criteria, itmay be an indication that the sensor or the signal processing equipment has drifted outside its limit.--! .... Lnc -rcu i:Ucncv CIM 1 ii~ or3 zri;N;Z-iac ZAn ci-7; _- ti cxcic ji Jdcmnzrnstmte ehannelf taiurc *9 raro. I he CHANNEL CHECKsupplements less fefrmal, but mcre froguent, ehccks ef ehanncls dluringnrmal8 epcraltienal use ef the displays asseenated with the LCO) rcguircdSR 33.2.2This SR is the performance of a COT eve.y92-days for the following ESFAS functions:

• SI-Containment Pressure-High;

• SI-Pressurizer Pressure-Low;

• SI-Steam Line Pressure-Low;

• CS-Containment Pressure-High High;* Steam Line Isolation-Containment Pressure-High High;Steam Line Isolation-High Steam Flow Coincident with SI and Tavg-Low;* Steam Line Isolation-High-High Steam Flow Coincident with SI;* Feedwater Isolation-SG Water Level-High; and* AFW-SG Water Level-Low Low.A COT is performed on each required channel to ensure the channel willperform the intended Function.

Setpoints must be found to be within theCOT Acceptance Criteria specified in plant procedures.

The as-leftvalues must be consistent with the drift allowance used in the setpointmethodology.

The Froquency 3f 92 days *9Frcguency is adequate base-eeifd-efeens Edcr ME in~tFUment rcl ab litv~sistent with nin Rcfercnec 7.Thaindustry epefrating expcriencc, and hiStr dta..[INSERT3R.E. Ginna Nuclear Power PlantB 3.3.2-32Revision 42 ESFAS Instrumentation B 3.3.2SR 3.3.2.3This SR is the performance of a TADOT evey 92 d-ays. This test is acheck of the AFW-Undervoltage-Bus 11A and 11B Function.

The test includes trip devices that provide actuation signals directly to theprotection system. The SR is modified by a Note that excludesverification of setpoints for relays. Relay setpoints require elaborate bench calibration and are verified during CHANNEL CALIBRATION.

The-Fr-qucn.y ef 92 days is adequate based en i^ndustr; epcratincx lcrIc1 Icnidcring inIstIuI ncl t I cliability and ltirg ; datc.SR 3.3.2.4 [INSERT 31----This SR is the performance of a TADOT every 24 ,,e, the. This test is acheck of the SI, CS, Containment Isolation, Steam Line Isolation, andAFW Manual Initiations, and the AFW-Trip of Both MFW PumpsFunctions.

Each Function is tested up to, and including, the mastertransfer reJINSERT31 perating experienee and *3 eenicsitent with the typieal rcfueling eyele.The SR is modified by a Note that excludes verification of setpoints during the TADOT. The Manual Initiations, and AFW-Trip of Both MFWPumps Functions have no associated setpoints.

SR 3.3.2.5This SR is the performance of a CHANNEL 24meths of the following ESFAS Functions:

SI-Containment Pressure-High; SI-Pressurizer Pressure-Low;

• SI-Steam Line Pressure-Low;

• CS-Containment Pressure-High High;* Steam Line Isolation-Containment Pressure-High High;Steam Line Isolation-High Steam Flow Coincident with SI and Tavg-Low;* Steam Line Isolation-High-High Steam Flow Coincident with SI;Feedwater Isolation-SG Water Level-High; AFW-SG Water Level-Low Low; andAFW-Undervoltage-Bus 11A and 11B.R.E. Ginna Nuclear Power PlantB 3.3.2-33Revision 42 ESFAS Instrumentation B 3.3.2CHANNEL CALIBRATION is a complete check of the instrument loop,including the sensor. The test verifies that the channel responds to ameasured parameter within the necessary range and accuracy.

CHANNEL CALIBRATIONS must be performed consistent with theassumptions of the plant specific setpoint methodology.

The "as left"values must be consistent with the drift allowance used in the setpointmethodology.

The Frcgqucney ef 24 menths as based en the assumptien ef a 24 menthealibraktienitra n the detefrminaticn ef the maegnitude ef equipment drift in the sctpeint mnethedelegy.

SR 3.3.2.6This SR ensures the SI-Pressurizer Pressure-Low and SI-Steam LinePressure-Low Functions are not bypassed when pressurizer pressure> 2000 psig while in MODES 1, 2, and 3. Periodic testing of thepressurizer pressure channels is required to verify the setpoint to be lessthan or equal to the limit.The difference between the current as-found values and the previous testas-left values must be consistent with the drift allowance used in thesetpoint methodology (Ref. 6). The setpoint shall be left set consistent with the assumptions of the current plant specific setpoint methodology.

If the pressurizer pressure interlock setpoint is nonconservative, then thePressurizer Pressure-Low and Steam Line Pressure-Low Functions areconsidered inoperable.

Alternatively, the pressurizer pressure interlock can be placed in the conservative condition (nonbypassed).

If placed inthe nonbypassed condition, the SR is met and the Pressurizer Pressure-Low and Steam Line Pressure-Low Functions would not be considered inoperable.

SR 3.3.2.7 IThis SR is the performance of an ACTUATION LOGIC TEST on allESFAS Automatic Actuation Logic and Actuation Relays Functions eVeiy-24 menths. This test includes the application of various simulated oractual input combinations in conjunction with each possible interlock state and verification of the required logic output. Relay and contactoperation is verified by a continuance check or actuation of the enddevice.The Frequeney ef 24 Fflnths is based en epcrating cxperienee and theneed te perffermf thus testing duFrig a plant shutdewn te prevent 8 reaetrtrip frem eeeurring.

R.E. Ginna Nuclear Power PlantB 3.3.2-34Revision 42 PAM Instrumentation B 3.3.3G.. 1If one channel for Function 7 or 10 cannot be restored to OPERABLEstatus within the required Completion Time of Condition D, the plant musttake immediate action to prepare and submit a special report to the NRC.This report shall be submitted within the following 14 days from the timethe action is required.

This report discusses the alternate means ofmonitoring Reactor Vessel Water Level and Containment Area Radiation, the degree to which the alternate means are equivalent to the installed PAM channels, the areas in which they are not equivalent, and aschedule for restoring the normal PAM channels.

These alternate means must have been developed and tested and maybe temporarily installed if the normal PAM channel(s) cannot be restoredto OPERABLE status within the allotted time.SURVEILLANCE A Note has been added to the SR Table to clarify that SR 3.3.3.1 and SRREQUIREMENTS 3.3.3.2 apply to each PAM instrumentation Function in Table 3.3.3-1.SR 3.3.3.1Performance of the CHANNEL CHECK .n.. c;c ry 31 days ensures thata gross instrumentation failure has not occurred.

A CHANNEL CHECK isnormally a comparison of the parameter indicated on one channel to asimilar parameter on other channels.

It is based on the assumption thatinstrument channels monitoring the same parameter should readapproximately the same value. Significant deviations between the twoinstrument channels could be an indication of excessive instrument driftin one of the channels or of more serious instrument conditions.

ACHANNEL CHECK will detect gross channel failure; thus, it is key toverifying the instrumentation continues to operate properly between eachCHANNEL CALIBRATION.

The high radiation instrumentation should becompared to similar plant instruments located throughout the plant.Channel check acceptance criteria are determined by the plant staff,based on a combination of the channel instrument uncertainties, including isolation, indication, and readability.

If a channel is outside thecriteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit.As specified in the SR, a CHANNEL CHECK is only required for thosechannels that are normally energized.

The Frogueney ef 31 days is based .. .p..ating cxpcriencc thatdefmeonStrates that ehannel failuro is rarc. The CHANNELI=G CHECKsupplements less fefrmal, but ffiro froguent, eheeks of ehannels durin9gR.E. Ginna Nuclear Power PlantB 3.3.3-16Revision 73 PAM Instrumentation B 3.3.3with the LCO) ncral cpefzrtie

,,a use of the displaySR3.3.3.21 S RSScmatfea A (LA NIMIrI ¢A6"IRATII N is mec 24 ,nths, erap...ximatcly at cv,. r'fueling.

CHANNEL CALIBRATION is acomplete check of the instrument loop, including the sensor. The testverifies that the channel responds to the measured parameter with thenecessary range and accuracy.

Whenever a sensing element isreplaced, the next required CHANNEL CALIBRATION of the Core Exitthermocouple sensors shall include an inplace qualitative assessment ofsensor behavior and normal calibration of the remaining adjustable devices in the channel.

This is accomplished by an inplace crosscalibration that compares the other sensing elements with the recentlyinstalled sensing element.

The F-r..u.ncy is based en .p.rating-a -nd as ensistent with the typical industr" rceling eyele.CX .. .-..÷ ÷..^ i .. ....3REFERENCES

1. UFSAR, Section 7.5.2.2. Regulatory Guide 1.97, Rev. 3.3. NUREG-0737, Supplement 1, "TMI Action Items."4. UFSAR, Section 6.2.5.R.E. Ginna Nuclear Power PlantB 3.3.3-17Revision 73 LOP DG Start Instrumentation B 3.3.4significantly reduce the probability that the LOP DG start instrumentation will trip when necessary.

SR 3.3.4.1This SR is the performance of a TADOT eyeoy 3-1days.

This test checkstrip devices that provide actuation signals directly.

For these tests, therelay trip setpoints are verified and adjusted as necessary to ensure theLSSS can still be met. Thc 31 day FFr..u.n.y

i. based en the kn.wnroliability of the r-elays and eentrols and has been shewn to be aeecptable thr.ugh ope,, ting ,xporioene'.

IN E T -SR 3.3.4.2 3This SR is the performance of a CHANNEL CALIBRATION evesy 24months, r appr...fimat"ly at .v..y fu'ling, of the LOP DG startinstrumentation for each 480 V bus.The voltage setpoint verification, as well as the time response to a loss ofvoltage and a degraded voltage test, shall include a single pointverification that the trip occurs within the required time delay.CHANNEL CALIBRATION is a complete check of the instrument loop,including the sensor. The test verifies that the channel responds to ameasured parameter within the necessary range and accuracy.

The .F..u.ncy of 24 ..nths i. based on epcrating cxporicncc eensmstent with the typieal industry rofucling eyelc and is justified by theassufmption of a 24 menth ealibffltift intewe'l win the detrmqinaltien of themnagnitude of equipment drift in the setpeint analysi.REFERENCES

1. UFSAR, Section 8.3.2. UFSAR, Chapter 15.IR.E. Ginna Nuclear Power PlantB 3.3.4-7Revision 37 Containment Ventilation Isolation Instrumentation B 3.3.5SURVEILLANCE A Note has been added to the SR Table to clarify that Table 3.3.5-1REQUIREMENTS determines which SRs apply to which Containment Ventilation Isolation Functions.

SR 3.3.5.1Performance of the CHANNEL CHECK cnccc;vcr; 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ensures thata gross failure of instrumentation has not occurred and theinstrumentation continues to operate properly between each CHANNELCALIBRATION.

The CHANNEL CHECK agreement criteria aredetermined by the plant staff, based on a combination of the channelinstrument uncertainties, including indication and readability.

If a channelis outside the criteria, it may be an indication that the sensor or the signalprocessing equipment has drifted outside its limit.The Frogueney isbased en eperating-cxcI ! that demenstmtcs but moroe frgucnt, cheeks of ehannels during nrmFFal operational use ofthe displays asseeiated with the LCOG Fcqu ird channcls.

SR 3.3.5.2A COT is performed e.e.y .92 ays-.on each required channel to ensurethe channel will perform the intended Function.

The Frequency is basedon the staff recommendation for increasing the availability of radiation monitors according to NUREG-1 366 (Ref. 2). This test verifies thecapability of the instrumentation to provide the containment ventilation system isolation.

The setpoint shall be left consistent with the currentplant specific calibration procedure tolerance.

SR 3.3.5.3 k-.INS E.R1T 3IThis SR is the performance of an ACTUATION LOGIC TEST. Allpossible logic combinations, with and without applicable permissives, aretested for each protection function.

In addition, the master relay is testedfor continuity.

This verifies that the logic modules are OPERABLE andthere is an intact voltage signal path to the master relay coils. Th'is test iperf.rmcd e,-,v; 24 months. The interval

-s aee.ptabl-based on.

t ,, liability and indust,;y operiting ne-"-.4\JINSERT 3R.E. Ginna Nuclear Power PlantB 3.3.5-8Revision 42 Containment Ventilation Isolation Instrumentation B 3.3.5SR 3.3.5.4A ....ANN.I=. GAI...ATII as... pef...mt..d cvcr' 24 eappr..imatcly at ..v... ..fueling.

CHANNEL CALIBRATION is acomplete check of the instrument loop, including the sensor. The testverifies that the channel responds to a measured parameter within thenecessary range and accuracy.

The as based en Ip rtnFgt cxIcII nc and I9 cns.st.nt withthe typieel industry Fefueling eyele.REFERENCES 1 .10 CFR 50.67.2. NUREG-1 366.R.E. Ginna Nuclear Power PlantB 3.3.5-9Revision 42 CREATS Actuation Instrumentation B 3.3.6C.1 and C.2Condition C applies when the Required Action and associated Completion Time of Condition A or B has not been met and the plant is inMODE 1, 2, 3, or 4. The plant must be brought to a MODE thatminimizes accident risk. To achieve this status, the plant must be broughtto MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowedCompletion Times are reasonable, based on operating experience, toreach the required plant conditions from full power conditions in anorderly manner and without challenging plant systems.D.1 and D.2Condition D applies when the Required Action and associated Completion Time of Condition A or B has not been met during movementof irradiated fuel assemblies.

Movement of irradiated fuel assemblies must be suspended immediately to reduce the risk of accidents thatwould require CREATS actuation.

This places the plant in a condition that minimizes risk. This does not preclude movement of fuel or othercomponents to a safe position.

SURVEILLANCE A Note has been added to the SR Table to clarify that Table 3.3.6-1REQUIREMENTS determines which SRs apply to which CREATS Actuation Functions.

SR 3.3.6.1Performance of the CHANNEL CHECK cncccvcry 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> cnsures thatgross failure of instrumentation has not occurred.

A CHANNEL CHECKis normally a comparison of the parameter indicated on one channel to asimilar parameter on other channels.

It is based on the assumption thatinstrument channels monitoring the same parameter should readapproximately the same value. Significant deviations between theinstrument channels could be an indication of excessive instrument driftin one of the channels or of more serious instrument conditions.

ACHANNEL CHECK will detect gross channel failure; thus, it is averification that the instrumentation continues to operate properlybetween each CHANNEL CALIBRATION.

CHANNEL CHECK acceptance criteria are determined by the plant staffbased on a combination of the channel instrument uncertainties, including indication and readability.

If a channel is outside the criteria, itmay be an indication that the sensor or the signal processing equipment has drifted outside its limit.The Frcgucr-ey ef 12 heur3 is based en epcrotirng cxpcriencc hdcmcneS4trotc ehennel failuro iS roro. The CHANNEL CHECsupplements less feFrmal, but moroe froguont, chocks of eharnigeS duringR.E. Ginna Nuclear Power PlantB 3.3.6-7Revision 38 CREATS Actuation Instrumentation B 3.3.6nor al op ,rational use , f the displays ess,,iatd with the I r'" , ,SR 33.6.2~TThis SR is the performance of a COT onco c'er;y 92 days on eachrequired channel to ensure the channel will perform the intendedfunction.

This test verifies the capability of the instrumentation to providethe automatic CREATS actuation.

The setpoints shall be left consistent with the plant specific calibration procedure tolerance.

Thc Frqu.ncy f92 days is based en the knoiwn roliabiliety of the monitoring equipment andhas been shewn to be acooptablo througoh epeffiting epeoneeo.\

SR 3.3.6.3 NSERT 3This SR is the performance of a TADOT of the Manual Initiation Function24.mentl.s.

The Manual Initiation Function is tested up to, andincluding, the master relay coils.The Froguonoy of 24 months is based on the 1(noWn Foliability of theFunetion and the rodundancy available, and has boon shown to beaeooptablo through oporating experienec.

... ,, ...... .,. .. ... ..,,,ok .. .. iNSERT 31The SR is modified by a Note that excludes verification of setpoints because the Manual Initiation Function has no setpoints.

SR 3.3.6.4This SR is the performfanoc ef a CHANNEL GALlBRATION c-vcry 24m.nths, or appr,..imat.ly at c..;y ..fu.ling.

CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The testverifies that the channel responds to a measured parameter within thenecessary range and accuracy.

The Frogquency of 24 months is based on epcraltinq expcrienee and iocrnsistont with the typical industn Fcfucling eyeloeSR 3.3.6.5 L'INSERTThis SR is the performance of an ACTUATION LOGIC TEST. Allpossible logic combinations are tested for the CREATS actuation instrumentation.

In addition, the master relay is tested forcontinuity.

Thisverifies that the logic modules are OPERABLE and there is an intactvoltage signal path to the master relay coils. This test is acceptable based on instrument reliability and operating t3R.E. Ginna Nuclear Power PlantB 3.3.6-8Revision 38 RCS Pressure, Temperature, and Flow DNB LimitsB 3.4.1RCS total flow rate is not a controllable parameter and is not expected tovary during steady state operation.

If the indicated RCS total flow rate isbelow the LCO limit, power must be reduced, as required by RequiredAction B.1, to restore DNB margin and eliminate the potential for violation of the accident analysis bounds.The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time for restoration of the parameters providessufficient time to determine the cause for the off normal condition, toadjust plant parameters, and to restore the readings within limits, and isbased on plant operating experience.

B._1If Required Action A.1 is not met within the associated Completion Time,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 at least MODE 2within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. In MODE 2, the reduced power condition eliminates thepotential for violation of the accident analysis bounds. The Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable to reach the required plant conditions in anorderly manner.SURVEILLANCE REQUIREMENTS SR 3.4.1.1Sinec Requircd Action A.!1 allcws a Completion Time of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to rosteroparametcrS that arc net within limits, the 12 hcur Guryeillanee Frcequcncy fer PrcssUri~zer prcssurce Is sufficiont to cncSUrc the prcSSUrc can berosterod to a normafll cpcratien, steady state eendition fellewing leadeh".g..s and ether .xpc.ted transient Thc 12 heur .ntr..al-has been shown by opcralting pracetics to be sufficicnt to rcgula rly assessfor pctential degradatien and tc Yerify opefrtieon is within safety analysisSR 3.4.1.2 NSERT3Sinec Rcquircd Aetion A.! allows a Completion Time of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> tc Festercpa....mteS that are not within knmits, thc 12 hu" .Su.vCiI an.. Frc.u.n.y for RCS .....g. tc..p..atur' is nt to nsurc the canbe .. tf-cd to a norm.al p...ati-n, steady statc ccnditien following loadchangcs and othcr expcctcd tranfsicnit epcratiens.

The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> intcrwalhas been shown by opcrating practicc to be sufficicnt to rcgulafly assessfor potential degradation and to reif, epefration is within safety anaelysis a.# s, I m I e n1 s,.r't ..A II.# 11 J .l .ll.lI .#.11 11 I IIV l I Ir. ,lR.E. Ginna Nuclear Power PlantB 3.4.1-4Revision 42 RCS Pressure, Temperature, and Flow DNB LimitsB 3.4.1SR 3.4.1.3Measurement of RCS total flow rate oncc cvcry 24 months verifies theactual RCS flow rate is greater than or equal to the minimum requiredRCS flow rate. This verification may be performed via a precision calorimetric heat balance or other accepted means.IINSERT 3 F----.7"--.

... .. -. ^,. -.,.,. ..*.. ,..after a rofucling eutagc when the ccrc has been alterod, whieh maey have.aus.d an altcratine ef f..W rcc.tancc..

Verification of RCS flow rate on ashorter interval is not required since this parameter is not expected tovary during steady state operation as there are no RCS loop isolation valves or other installed devices which could significantly alter flow.Reduced performance of a reactor coolant pump (RCP) would beobservable due to bus voltage and frequency

changes, and installed alarms that would result in operator investigation.

This SR is modified by a Note that allows entry into MODE 1, withouthaving performed the SR, and placement of the plant in the bestcondition for performing the SR. The Note states that the SR shall beperformed within 7 days after reaching 95% RTP. This exception isappropriate since the heat balance requires the plant to be at a minimumof 95% RTP to obtain the stated RCS flow accuracies.

REFERENCES

1. UFSAR, Chapter 15.2. NRC Memorandum from E.L. Jordan, Assistant Director forTechnical

Programs, Division of Reactor Operations Inspection toDistribution;

Subject:

"Discussion of Licensed Power Level (AITSF14580H2),"

dated August 22, 1980.R.E. Ginna Nuclear Power PlantB 3.4.1-5Revision 42 RCS Minimum Temperature for Criticality B 3.4.2ACTIONS A.1If the parameters that are outside the limit cannot be restored, the plantmust be brought to a MODE in which the LCO does not apply. To achievethis status, the plant must be brought to MODE 2 with Keff < 1.0 within 30minutes.

Rapid reactor shutdown can be readily and practically achievedwithin a 30 minute period due to the proximity to MODE 2 conditions.

The allowed time is reasonable, based on operating experience, to reachMODE 2 with Keff < 1.0 in an orderly manner and without challenging plant systems.SURVEILLANCE SR 3.4.2.1REQUIREMENTS This SR verifies that RCS Tavg in each loop is > 540OF within 30 minutesprior to achieving criticality.

This ensures that the minimum temperature for criticality is being maintained just before criticality is reached.

The 30minute time period is long enough to allow the operator to adjusttemperatures or delay criticality so the LCO will not be violated, therebyproviding assurance that the safety analyses are not violated.

SR 3.4.2.2RCS loop average temperature is required to be verified at or above540OF every 30 minutes in MODE 1, and in MODE 2 with keff __ 1.0. The30 minute frequency is sufficient based on the low likelihood of largetemperature swings without the operators knowledge.t-lNSERT 3iThis SR is modified by a Note that only requires the SR to be performed ifany RCS loop Tavg is < 5470F and the low Tavg alarm is either inoperable or not reset. The Tavg alarm provides operator indication of low RCStemperature without requiring independent verification while a Tavg> 5470F in both RCS loops is within the accident analysis assumptions.

Ifthe Tavg 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 PlantB 3.4.2-3Revision 21 RCS P/T LimitsB 3.4.3SURVEILLANCE SR 3.4.3.1REQUIREMENTS Verification that operation is within the PTLR limits is required eveFy--30 nlntkea-when RCS pressure and temperature conditions are undergoing planned changes.

This Freque- i.er asne in viewthe contrOl room findication avaF-ila~ble fto monintor ROS status. AlsIo, sfintermperaturo rate of change limfitS are specified in haurly inecroments, 30minutits permlitS assessment and corroctien for mninor deviations within a.easenable

.Surveillance for heatup, cooldown, or ISLH testing may be discontinued when the definition given in the relevant plant procedure for ending theactivity is satisfied.

This SR is modified by a Note that only requires this SR to be performed during system heatup, cooldown, and ISLH testing.

No SR is given forcriticality operations because LCO 3.4.2 contains a more restrictive requirement.

REFERENCES

1. WCAP-14040, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown LimitCurves,"

Revision 1, December 1994.2. 10 CFR 50, Appendix G.3. ASME, Boiler and Pressure Vessel Code,Section III, Appendix G.4. ASTM E 185-82, July 1982.5. 10 CFR 50, Appendix H.6. Regulatory Guide 1.99, Revision 2, May 1988.7. ASME, Boiler and Pressure Vessel Code,Section XI, Appendix E.R.E. Ginna Nuclear Power PlantB 3.4.3-6Revision 21 RCS Loops -MODE 1 > 8.5% RTPB 3.4.4Operation in other MODES is covered by:LCO 3.4.5, "RCS Loops -MODES 1 8.5% RTP, 2, AND 3";LCO 3.4.6, "RCS Loops -MODE 4";LCO 3.4.7, "RCS Loops -MODE 5, Loops Filled";LCO 3.4.8, "RCS Loops -MODE 5, Loops Not Filled";LCO 3.9.4, "Residual Heat Removal (RHR) and Coolant Circulation-Water Level > 23 Ft" (MODE 6); andLCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation-Water Level < 23 Ft" (MODE 6).ACTIONS A.1If the requirements of the LCO are not met, the Required Action is toreduce power and bring the plant to MODE 1 < 8.5% RTP. This lowerspower level and thus reduces the core heat removal needs andminimizes the possibility of violating DNB limits.The Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, based on operating experience, to reach MODE 1 < 8.5% RTP from full power conditions inan orderly manner and without challenging safety systems.SURVEILLANCE SR 3.4.4.1REQUIREMENTS This SR requires verification eveiy 12 het'. that each RCS loop is inoperation.

Verification includes flow rate, temperature, or pump statusmonitoring, which help ensure that forced flow is providing heat removalwhile maintaining the margin to DNB. Use of control board indication forthese parameters is an acceptable verification.

Thec F..u.n.y

.f 12heura is sufficicnt eensidering ether indicotiecns and 818Frm3 available tet,-he epe tc in the ....c ,nr .... r m -t -m -^itorCS loop p.f... ... la INSERT 3R.E. Ginna Nuclear Power PlantB 3.4.4-3Revision 46 RCS Loops -MODES 1 _< 8.5% RTP, 2, and 3B 3.4.5B.1If restoration of the inoperable loop is not possible within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, theplant must be brought to MODE 4. In MODE 4, the plant may be placedon the Residual Heat Removal System. The additional Completion Timeof 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is compatible with required operations to achieve cooldownand depressurization from the existing plant conditions in an orderlymanner and without challenging plant systems.C.1. C.2, and C.3If two RCS loops are inoperable, or no RCS loop is in operation, exceptduring conditions permitted by the Note in the LCO section, all CRDMsmust be de-energized by opening the RTBs or de-energizing the MGsets. All operations involving introduction of coolant into the RCS withboron concentration less than required to meet the minimum SDM ofLCO 3.1.1 must be suspended, and action to restore one of the RCSloops to OPERABLE status and operation must be initiated.

Borondilution requires forced circulation for proper mixing, and opening theRTBs or de-energizing the MG sets removes the possibility of aninadvertent rod withdrawal.

Suspending the introduction of coolant intothe RCS with boron concentration less than required to meet theminimum SDM of LCO 3.1.1 is required to assure continued safeoperation.

With coolant added without forced circulation, unmixedcoolant could be introduced to the core, however coolant added withboron concentration meeting the minimum SDM maintains acceptable margin to subcritical operations.

The immediate Completion Timereflects the importance of maintaining operation for heat removal.

Theaction to restore must be continued until one loop is restored toOPERABLE status and operation.

SURVEILLANCE SR 3.4.5.1REQUIREMENTS This SR requires verification e-'ey 12 het .that the required RCS loop isin operation.

Verification includes flow rate, temperature, or pump statusmonitoring, which help ensure that forced flow is providing heat removal.Use of the control board indication for these parameters is an acceptable verification. Fc..qu.n.

y .f 12 heur, 0. suffi.i.nt

• .n.idering

.the.

and olIrmH available tI the epeF.af in the e.otrol room t-mcne fitcr R GS lcep pcrfefrmonco.

IR.E. Ginna Nuclear Power PlantB 3.4.5-5Revision 61 RCS Loops -MODES 1 < 8.5% RTP, 2, and 3B 3.4.5SR 3.4.5.2This SR requires verification of SG OPERABILITY.

SG OPERABILITY isverified by ensuring that the secondary side narrow range water level is>_ 16% for two RCS loops. If the SG secondary side narrow range waterlevel is < 16%, the tubes may become uncovered and the associated loop may not be capable of providing the heat sink for removal of reactoror decay heat. The 12 h".. F..qucn.y is ..nsidercd adequate in view " fether indiesticns availableoin the eeontrcl rcen te alert the epeffitc te aless ef GGSC -SR 3.4.5.3Verification that the required RCP is OPERABLE ensures that anadditional RCP can be placed in operation, if needed, to maintain decayheat removal and reactor coolant circulation.

Verification is performed byverifying proper breaker alignment and power available to the requiredpump that is not in operation.

The F..qu.n.y

.f .7 days as ...sidc"reasenable in view of other adminiztrnt~aiy contre's available and hasbeen shewn te be acccptable by cpcralting cxperienee.

REFERENCES

1. UFSAR Section 15.1.5.2. UFSAR Section 15.4.3.3. Letter from D. M. Crutchfield, NRC, to J. E. Maier, RG&E,

Subject:

"SEP Topic XV-9, Startup of an Inactive Loop, R. E. Ginna," datedAugust 26, 1981.4. UFSAR Sections 14.6.1.5.6 and 15.2.5.5. UFSAR Section 14.6.1.5.5.

R.E. Ginna Nuclear Power PlantB 3.4.5-6Revision 61 RCS Loops -MODE 4B 3.4.6SURVEILLANCE REQUIREMENTS SR 3.4.6.1This SR requires verification eveFy 12 heufs that one RCS or RHR loop isin operation.

Verification includes flow rate, temperature, or pump statusmonitoring, which help ensure that forced flow is providing heat removal.Use of control board indication for these parameters is an acceptable verification. F-r..ue...y

.f 12 heur- is suffi...nt

-.

.thc,%,nd:^atins and alarS ..a;ilablc tc the .....ater in the ...t.el rce.. te.AA 4-ROS and RHR leep pcrefffienee.

SR 3.4.6.2This SR requires verification of SG OPERABILITY.

SG OPERABILITY isverified by ensuring that the secondary side narrow range water level is> 16%. If the SG secondary side narrow range water level is < 16%, thetubes may become uncovered and the associated loop may not becapable of providing the heat sink necessary for removal of decay heat.The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frogueney as eensidc rcd adecquate in Yiew ef etherindicati'ns availabl in the "contrcl rcom tc al'^t the .p..ratcr t l ....SR 3.4.6.3Verification that the required pump is OPERABLE ensures that anadditional RCS or RHR pump can be placed in operation, if needed, tomaintain decay heat removal and reactor coolant circulation.

Verification is performed by verifying proper breaker alignment and power available to the required pump that is not in operation.

The ef 7 days is-consodcred rcasenablc in view ef ether admin~iStrativc eentrcls available and has becshew, t. bc e ,,.ptabl" by _pcratitng xpcrienCC.

4JINSERT 3PREFERENCES

1. UFSAR, Section 14.6.1.2.6.

R.E. Ginna Nuclear Power PlantB 3.4.6-5Revision 61 RCS Loops -MODE 5, Loops FilledB 3.4.7SURVEILLANCE REQUIREMENTS SR 3.4.7.1This SR requires verification every 12 haet'r-s-that one RHR loop is inoperation.

Verification includes flow rate, temperature, or pump statusmonitoring, which help ensure that forced flow is providing heat removal.Use of control board indication for these parameters is an acceptable verification.

The F...quen.y of 12 hurFS is suffi.int ensidc;ing

,thor-indcations and a'.ams available to the eperator On the e..tr.l r-cm tomoenitor RHR loop perfcrmanec.

SR 3.4.7.2 t NE T3This SR requires verification of SG OPERABILITY.

Verifying that at leastone SG is OPERABLE by ensuring its secondary side narrow rangewater level is > 16% ensures an alternate decay heat removalmethod inthe event that the second RHR loop is not OPERABLE.

If both RHRloops are OPERABLE, this Surveillance is not needed. Thea 12 h4e' is eensiderod adequate an view of -tho i,, atndiopns available i n the control rc. m to al. I the to the less of IC IYE4LSR 3.4.7.3 [INSERT 3Verification that a second RHR pump is OPERABLE ensures that anadditional pump can be placed in operation, if needed, to maintain decayheat removal and reactor coolant circulation.

Verification is performed byverifying proper breaker alignment and power available to the standbyRHR pump. If secondary side water level is > 16% in at least one SG,this Surveillance is not needed. The 7 days is o ,nsidcr' droasonablo in view of othor administrativ.

ceotrelS aailablo and hasbeen shown to be a.. .ptablc by epcr..ting

.xpcrinco.

REFERENCES

1. UFSAR, Section 14.6.1.2.6

2. NRC Information Notice 95-35R.E. Ginna Nuclear Power PlantB 3.4.7-5Revision 61 RCS Loops -MODE 5, Loops Not FilledB 3.4.8SURVEILLANCE SR 3.4.8.1REQUIREMENTS This SR requires verification ev'eiy 12 hhe's-that one RHR loop is inoperation.

Verification includes flow rate, temperature, or pump statusmonitoring, which help ensure that forced flow is providing heat removal.The Froquoeny ef 12 houro is suffiiont i cnsidering other indldatiens anda'a, H available tek, the epf.e OR.. the ' ee' .. .' .... te .... "t^ RHR '^^pPe~f6efafee.,ý.

.SR 3.4.8.2 /INSERT3Verification that a second RHR pump is OPERABLE ensures that anadditional pump can be placed in operation, if needed, to maintain decayheat removal and reactor coolant circulation.

Verification is performed byverifying proper breaker alignment and power available to the standbypump. The .Fr...u.ny ef 7 days is ".nsiderod roasenab.'

in view .fethcr administativ-

...tr-ls available and has be.n shown tc bcREERptNblE by Nponeting expcricnee.

REFERENCES 1 .None.R.E. Ginna Nuclear Power PlantB 3.4.8-4Revision 61 Pressurizer B 3.4.9B.1 and B.2If the pressurizer heaters capacity is < 100 KW, the ability to maintainRCS pressure to support natural circulation may no longer exist. Bymaintaining RCS pressure

control, a margin to subcooling is provided.

The value of 100 KW is based on the amount needed to support naturalcirculation after accounting for heat losses through the pressurizer insulation during an extended loss of offsite power event.If the capacity of the pressurizer heaters is not within the limit, the plantmust be brought to MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 4 within 12hours. The allowed Completion Times are reasonable, based onoperating experience, to reach the required plant conditions from fullpower conditions in an orderly manner and without challenging plantsystems.SURVEILLANCE SR 3.4.9.1REQUIREMENTS This SR requires that during steady state operation, pressurizer level ismaintained below the nominal upper limit to provide a minimum space fora steam bubble. The Surveillance is performed by observing theindicated level. The Fr..uen.y of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> has been shown by ,per..ting praIti1e to be sufficient tO regulafly assess 'eve' fOr any deviation andverify that operation is within safoty analysos assumptions.

Alarms arealso available for early detection of abnormal level indications.

SR 3.4.9.2This SR is satisfied when the power supplies are demonstrated to becapable of producing the minimum power required.

This may be done bytesting the power supply output by verifying the electrical load on Buses14 and 16 with the respective heater groups on and off. The-Frequeney-of 92 days is considerod adequate to deteet heater degradation and hasbeen show by pfoating experience to be a,,eptable./I\

,R.E. Ginna Nuclear Power PlantB 3.4.9-4Revision 21 Pressurizer PORVsB 3.4.11SURVEILLANCE REQUIREMENTS ISR 3.4.11.1 JINSER- 3Block valve cycling verifes that the valve(s) can be closed if needed. Thebasis fer the F-.quc..y

.f 92 days is the ASME Cod- (Ref. .If theblock valve is closed to isolate a PORV that is OPERABLE and is notleaking in excess of the limits of LCO 3.4.13, "RCS Operational LEAKAGE,"

then opening the block valve is necessary to verify that thePORV can be used for manual control of reactor pressure.

If the blockvalve is closed to isolate an otherwise inoperable PORV, the maximumCompletion Time to restore the PORV and open the block valve is 72hours, which is well within the allowable limits (25%) to extend the blockvalve Frequency-ef92--dxys.

Furthermore, these test requirements wouldbe completed by the reopening of a recently closed block valve uponrestoration of the PORV to OPERABLE status (i.e., completion of theRequired Actions fulfills the SR).The Note modifies this SR by stating that it is not required to beperformed with the block valve closed per LCO 3.4.13. This prevents theneed to open the block valve when the associated PORV is leaking > 10gpm creating the potential for a plant transient.

SR 3.4.11.2This SR requires a complete cycle of each PORV using the nitrogenaccumulators.

Operating a PORV through one complete cycle ensuresthat the PORV can be manually actuated for mitigation of an SGTR. T-he-Frcqueney of 24 molnths is based en a typical rcfucling eyelc and industryeeeoptcd PFeieREFERENCES

1. UFSAR, Section 15.2.2. ASME Code for Operation and Maintenance of Nuclear PowerPlants.IR.E. Ginna Nuclear Power PlantB 3.4.11-7Revision 58 LTOP SystemB 3.4.12disabling of a charging pump is necessary since RV 203 cannot mitigatea charging/letdown mismatch event if RHR is providing decay heatremoval above MODE 5 and three charging pumps are operating.

The passive vent must be sized _> 1.1 square inches to ensure that theflow capacity is greater than that required for the worst case mass inputtransient reasonable during the applicable MODES. This action isneeded to protect the RCPB from a low temperature overpressure eventand a possible brittle failure of the reactor vessel and to protect the RHRsystem from overpressurization.

The Completion Time of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to depressurize the RCS and establish avent considers the time required to place the plant in this Condition andthe relatively low probability of an overpressure event during this timeperiod due to increased operator awareness of administrative controlrequirements.

SURVEILLANCE SR 3.4.12.1REQUIREMENTS To minimize the potential for a low temperature overpressure event bylimiting the mass input capability, all SI pumps must be verified incapable of injecting into the RCS when the PORVs provide the RCS vent path(LCO 3.4.12.a) and a minimum of two SI pumps must be verifiedincapable of injecting into the RCS when the RCS is depressurized andan RCS vent > 1.1 square inches is established (LCO 3.4.12.b).

The SIpumps are rendered incapable of injecting into the RCS throughremoving the power from the pumps by racking the breakers out underadministrative control.

An alternate method of LTOP control may beemployed using at least two independent means to prevent a pump startsuch that a single failure or single action will not result in an injection intothe RCS. This may be accomplished through the following:

a. placing the pump control switch in the pull-stop position and closingat least one valve in the discharge flow path;b. locking closed a manual isolation valve in the injection path; orc. closing a motor operated isolation valve in the injection path andremoving the AC power source.The flowpaths through the test connections associated with the ECCSaccumulator check valves (i.e., lines containing air operated valves 839A,839B, 840A, and 840B) and the ECCS accumulator fill lines (i.e., linescontaining air operated valves 835A and 835B) do not have to be isolatedfor this SR since the potential mass addition from a single SI pumpthrough these six lines is limited by the installed orifices to less than thatassumed for the charging/letdown mismatch analysis.

R.E. Ginna Nuclear Power PlantB 3.4.12-10 Revision 52 LTOP SystemB 3.4.12The ECCS accumulator motor operated isolation valves can be verifiedclosed by use of control board indication for valve position.

Thisverification is only required when the accumulator pressure is greaterthan or equal to the maximum RCS pressure for the existing RCS coldleg temperature allowed by the P/T limit curves provided in the PTLR. Ifthe accumulator pressure is less than this limit, no verification is requiredsince the accumulator cannot pressurize the RCS to or above the PORVsetpoint.

The Froqueney of 12 heurs is suffleient, eensidoring other indicationis and-alars available to the eperator in th- control ro-:om, to Ycrif; the Fequir.dstatus of the cguiprnent.

The Froquency of eyer; 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thecroafter ferGR 3 4 12 3 enGUre that the- EGGfS aeu ijmlteF nete e~r~ier~atnd mselatirm

-valves are maintained eloseSetRatm°n.21ý S R 3.4. 1 INSERTd and de noet rosult in a petcntial L-TOPSee SR 3.4.12.1SR 3.4.12.3See SR 3.4.12.1SR 3.4.12.4The RCS vent of > 1.1 square inches is proven OPERABLE by verifying its open condition eitheR.*a. Oncc evcr; 12 hourS for a vent (e.g., valve) that cannot be lockcd.b. Onco cvcr; 31 days for a vent (e.g., Yalyc) that ic looked sealed, orsccurcd in positien.

A d 1ros11riz1 eafety volv fits thisThe passive vent arrangement must be > 1.1 square inches and be opento be OPERABLE.

This Surveillance is required to be performed if thevent is being used to satisfy the pressure relief requirements of the LCO3.4.12.b.

ISR 3.4.12.5The PORV block valve must be verified open evoey 72 hoet sto providethe flow path for each required PORV to perform its function whenactuated.

The valve may be remotely verified open in the main controlroom. This Surveillance is performed if the PORV satisfies the LCO.The block valve is a remotely controlled, motor operated valve. Thepower to the valve operator is not required to be removed, and themanual operator is not required to be locked in the inactive position.

Thus, the block valve can be closed in the event the PORV developsR.E. Ginna Nuclear Power PlantB 3.4.12-11 Revision 52 LTOP SystemB 3.4.12excessive leakage or does not close (sticks open) after relieving anoverpressure situation.

The 72 heur Frcqueney is censidercd adequate in view ef ethr

...t. 'r available

t. the epolratoer in thc eentrcl reem, suchals valve pesitien indieation, that Yerify that thoe PORY bleek valve rcmnainsepen.T --INSERT 31SR 3.4.12.6Performance of a CHANNEL OPERATIONAL TEST (COT) is requirede.e.y. dys-on each required PORV to verify and, as necessary, adjustits lift setpoint.

The COT will verify the setpoint is within the allowedmaximum limits in the PTLR. PORV actuation could depressurize theRCS and is therefore not required.

A Note has been added indicating that this SR is required to beperformed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after decreasing RCS cold leg temperature toless than or equal to the LTOP enable temperature specified in the PTLRif it has not been performed the pr,;ieus 31 days. Depending onthe cooldown rate, the CO ay not have been performed before entryinto the LTOP MODES. The est must be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> afterentering the LTOP MODES. he 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> considers the unlikelihood of alow temperature overpressu event during this time.SR 3.4.12.7 FINSERT 1Verification once within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and every 31 days-thereafter that poweris removed from each ECCS accumulator motor operated isolation valveensures that at least two independent actions must occur before theaccumulator is capable of injecting into the RCS. "-iee peweP-le-romeyed under administrative control 81nd -valve position is Yerificd cvcry12 hourS, the peorfflrmnec of this surwcillanee enco within 12 heurs andcvcr; 31 days thercafter will proevide assur-ancc that peweFrI is cmovcd.This SIR is modified by a Note which states that the Surveillance is onlyrequired when the accumulator pressure is greater than or equal to themaximum RCS pressure for the existing cold leg temperature allowed inthe PTLR. If the accumulator pressure is below this limit, the LTOP limitcannot be exceeded and the surveillance is not required.

SR 3.4.12.8 ISRPerformance of a CHANNEL CALIBRATION on each required PORVactuation channel is required eveFy--24 mfenths to adjust the wholechannel so that it responds and the valve opens within the required rangeand accuracy to knownREFERENCES 1 .10 CFR 50, Appendix G.R.E. Ginna Nuclear Power PlantB 3.4.12-12 Revision 52 IDeleted LTOP SystemB 3.4.122. Gencrie Lettcr8 1 "NRC Posfiticn on Em~brittlement cf ReaeterVesselI Meaetria 81nd its Impaet en Plant I tiens."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 LicenseNo. DPR-1 8," dated July 26, 1979.7. Generic Letter 90-06, "Resolution of Generic Issue 70, "Power-Operated Relief Valve and Block Valve Reliability,"

and GenericIssue 94, "Additional Low-Temperature Overpressure Protection forLight-Water Reactors."

R.E. Ginna Nuclear Power PlantB 3.4.12-13 Revision 52 RCS Operational LEAKAGEB 3.4.13valves leak and result in a loss of mass from the RCS, the loss must beincluded in the allowable identified LEAKAGE.ACTIONS A..1Unidentified LEAKAGE or identified LEAKAGE in excess of the LCOlimits must be reduced to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. This Completion Time allows time to verify leakage rates and either identify unidentified LEAKAGE or reduce LEAKAGE to within limits before the reactor mustbe shut down. This action is necessary to prevent further deterioration ofthe RCPB.B.1 and B.2If any RCS pressure boundary LEAKAGE exists, or primary to secondary LEAKAGE is not within limits, or if the Required Action of Condition Acannot be completed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the reactor must be brought to lowerpressure conditions to reduce the severity of the LEAKAGE and itspotential consequences.

It should be noted that LEAKAGE past sealsand gaskets is not pressure boundary LEAKAGE.

The reactor must bebrought to MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Thisaction reduces the LEAKAGE and also reduces the factors that tend todegrade the pressure boundary.

The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full powerconditions in an orderly manner and without challenging plant systems.In MODE 5, the pressure stresses acting on the RCPB are much lower,and further deterioration is much less likely.SURVEILLANCE SR 3.4.13.1REQUIREMENTS Verifying RCS LEAKAGE to be within the LCO limits ensures the integrity of the RCPB is maintained.

Pressure boundary LEAKAGE which is notallowed by this LCO, would at first appear as unidentified LEAKAGE andcan only be positively identified by inspection.

Unidentified LEAKAGEand identified LEAKAGE are determined by performance of an RCSwater inventory balance.The RCS water inventory balance must be met with the reactor at steadystate operating conditions (stable temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows). The Surveillance is modified by two Notes. Note 1states that this SR 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 /> afterestablishing steady state operation.

The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowance providesR.E. Ginna Nuclear Power PlantB 3.4.13-4Revision 52 RCS Operational LEAKAGEB 3.4.13sufficient time to collect and process all necessary data after stable plantconditions 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 isdefined as stable RCS pressure, temperature, power level, pressurizer and volume control tank levels, makeup and letdown, and RCP sealinjection and return flows.An early warning of pressure boundary LEAKAGE or unidentified LEAKAGE is provided by the automatic systems that monitor thecontainment atmosphere radioactivity and the containment sump level. Itshould be noted that LEAKAGE past seals and gaskets is not pressureboundary LEAKAGE.

Leakage detection systems are specified in LCO3.4.15, "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 bemeasured accurately by an RCS water inventory balance.The heur F-,, uee,,y is a trnd ILE-AGAE andrcCegngizes the impcrtoncc ef corly leakage.

de~teetken in the ffevnti en efaeedentR.2 SR 3.4.132 INSERT3This SR verifies that primary to secondary LEAKAGE is less or equal to150 gallons per day through any one SG. Satisfying the primary tosecondary LEAKAGE limit ensures that the operational LEAKAGEperformance criterion in the Steam Generator Program is met. If this SRis not met, compliance with LCO 3.4.17, "Steam Generator TubeIntegrity,"

should be evaluated.

The 150 gallons per day limit ismeasured at room temperature as described in Reference

5. Theoperational LEAKAGE rate limit applies to LEAKAGE through any oneSG. If it is not practical to assign the LEAKAGE to an individual SG, allthe primary to secondary LEAKAGE should be conservatively assumedto 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 ofsteady state operation.

For RCS primary to secondary LEAKAGEdetermination, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeupand letdown, and RCP seal injection and return flows.The Sure."llln FFr ,u.ncy ,f 7-2 h-e..- --r .iabl- int"-val t- tr ,ndprimfary te secondar; LEAKAGE and rcccgnizes the impectanee ef carlyleakage dcteetion in the provcntien ef eeeidcnts.

The primarfy teeseeendar; LEFAKACE=

is detefrminod using eentinueuo prccooa radiateion R.E. Ginna Nuclear Power PlantB 3.4.13-5Revision 52 RCS Operational LEAKAGEB 3.4.13meniteS 6r radieehcm.ieal grab o8.nampling in a"eerdancc with thc EPRIguidelines (Refcrcgeeýn).

REFERENCES

1. Atomic Industry Forum (AIF) GDC 16, Issued for comment July 10,1967.2. Generic Letter 84-04, "Safety Evaluation of Westinghouse TopicalReports Dealing with Elimination of Postulated Pipe Breaks in PWRPrimary Main Loops."3. UFSAR, Chapter 15.4. NEI 97-06, Steam Generator Program Guidelines

5. EPRI, Pfessurized Water Rea~ete. PrmleIy t= Se ,nday Loa(iR.E. Ginna Nuclear Power PlantB 3.4.13-6Revision 52 RCS PIV LeakageB 3.4.14Required Action A.2 specifies that the double isolation barrier of twovalves be restored by closing some other valve qualified for isolation.

The use of a valve other than the previously leaking PIV must includeconsideration that the plant may no longer be in an analyzed condition.

The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time after exceeding the limit considers the timerequired to complete the Action and the low probability of a second valvefailing during this time period.B.1 and B.2If leakage cannot be reduced, the system isolated, or the other RequiredActions accomplished, the plant must be brought to a MODE in which 1herequirement does not apply. To achieve this status, the plant must bebrought to MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. ThisAction may reduce the leakage due to reduced RCS pressure whilereducing the potential for a LOCA outside the containment.

The allowedCompletion Times are reasonable based on operating experience, toreach the required plant conditions from full power conditions in anorderly manner and without challenging plant systems.SURVEILLANCE SR 3.4.14.1REQUIREMENTS Performance of leakage testing on each RCS PIV or isolation valve usedto satisfy Required Action A.1 and Required Action A.2 is required toverify that leakage is below the speciled limit and to identify each leakingvalve. The leakage limit of 0.5 gpmper inch of nominal valve diameter upto 5 gpm maximum applies to each valve and should be based on anRCS pressure of +/- 20 psig of normal system operating pressure.

Leakage testing requires a stable pressure condition.

For multiple in-series PIVs, the leakage requirement applies to eachvalve individually, except as noted below, and not to the combinedleakage across both valves. If the PIVs are not individually leakagetested, one valve may have failed completely and not be detected if theother in-series valve meets the leakage requirement.

In this situation, theprotection provided by redundant valves would be lost.R.E. Ginna Nuclear Power PlantB 3.4.14-5Revision 58 IRCS PIV LeakageB 3.4.14The SI hot leg injection lines are each configured with two check valvesand a motor operated valve in series. Each of these components independently is considered a qualified pressure boundary.

The twocheck valves function as a single pressure isolation barrier and the motoroperated valve serves as the second pressure isolation barrier to preventan intersystem LOCA. Both barriers need to be tested. Testing of thecheck valves (877A, 877B, 878F, and 878H) and the motor operatedvalves (878A and 878C) identified as PIVs in the SI hot leg injection linesis to be performed at least once every 40 months. This surveillance interval is allowed since the two SI hot leg injection lines are maintained closed to address pressurized thermal shock (PTS) concerns (Ref. 7 andRef. 11).Testing ^f the RCS ,IVs On #ic GI cld lcg ::cctien.

lin.. an"d RHR systemN to be

....y 24 ,months, a typical rofu'ling

' y.l. .The 24eentefined in the Insefviee Testing Programf, ic within. the frogueney allowed by the American Seeioty ef Meeheinicol Enginecr3 (ASMVE) Code-,(Ref. 9), and i. based en the need to peform .u h sur-illane--

undorthe eendlieion that apply durfing an eutage and the petential fr BAnunplanned trnnsient if the Suryelllanee wero perferrncd with the roaeter at-peweF.In addition to the periodic testing requirements, testing must beperformed once after the valve has been opened by flow, exercised, orhad maintenance performed on it to ensure tight reseating.

Thismaintenance does not include minor activities such as packingadjustments which do not affect the leak tightness of the valve. PIVsdisturbed in the performance of this Surveillance should also be testedunless documentation shows that an infinite testing loop cannotpractically be avoided.

Testing must be performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> afterthe valve has been reseated.

A limit of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is a reasonable andpractical time limit for performing this test after opening or reseating avalve.The leakage limit is to be met at the RCS pressure associated withMODES 1 and 2. This permits leakage testing at high differential pressures with stable conditions not possible in the MODES with lowerpressures.

Entry into MODES 3 and 4 is allowed to establish the necessary differential pressures and stable conditions to allow for performance ofthis Surveillance.

SR 3.4.14.2See SR 3.4.14.1R.E. Ginna Nuclear Power PlantB 3.4.14-6Revision 58 RCS PIV LeakageB 3.4.14REFERENCES

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 Risksin U.S. Commercial Nuclear Power Plants,"

Appendix V, October1975.5. NUREG-0677, "The Probability of Intersystem LOCA: Impact Dueto 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 CoolantSystem Pressure Isolation Valves,"

and associaled SER on PrimaryCoolant System Pressure Isolation Valves (WASH-1400, Event V),dated April 20, 1981. (ML010542030)

8. EG&G Report, EGG-NTAP-6175.

9. ASME Ccde feF Gpcraitien and Maintemnenc ef Nuelefr Pewcr-I~el-toe'e-ed

10. 10 CFR

11. Letter from D. M. Crutchfield, NRC, to J.E. Maier, RGE,

Subject:

"TMI-2 Category "A" Items" and associated SER for Amendment No. 42 to Provisional Operating License No. DPR-18, dated May11, 1981. (ML010540356)

R.E. Ginna Nuclear Power PlantB 3.4.14-7Revision 58 RCS Leakage Detection Instrumentation B 3.4.15Completion Time ensures that the plant will not be operated in a reducedconfiguration for a lengthy period of time.E.1 and E.2If a Required Action of Condition A, C, or D cannot be met, the plant mustbe brought to a MODE in which the requirement does not apply. Toachieve this status, the plant must be brought to at least MODE 3 wilhin 6hours and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Timesare reasonable, based on operating experience, to reach the requiredplant conditions from full power conditions in an orderly manner andwithout challenging plant systems.F.1With all required monitors inoperable, no automatic means of monitoring leakage are available, and immediate plant shutdown in accordance withLCO 3.0.3 is required.

SURVEILLANCE SR 3.4.15.1REQUIREMENTS This SR requires the performance of a CHANNEL CHECK of thecontainment atmosphere radioactivity monitors.

The check givesreasonable confidence that the channels are operating properly.

T-he-Froqueney 3f 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is based en inStrumonet roliability and is-reasenable for deteoting 3ff nRmF1al eonditions.

SR 3.4.15.2 3This SR requires the performance of a CHANNEL OPERATIONAL TEST(COT) on the containment atmosphere radioactivity monitors.

The testensures that the monitors can perform their function in the desiredmanner. The test verifies the alarm setpoint and relative accuracy of theinstrument string Tc. Th F ..quny of 92 days ..nsidc. S .nStruf...nt

..liabil.it,,

and .p... ting cxpori.noC has Shown. thalt it is propor ferSR 3.4.15.3

&-'INSERT 31These SRs require the performance of a CHANNEL CALIBRATION foreach of the RCS leakage detection instrumentation channels.

Thecalibration verifies the accuracy of the instrument string, including theinstruments located inside containment.

The Fr.quonc.y

.f 24 mnthsconsiders ehegnnl rcliabifilty and IpIrting has pfvIn thatthis Fr.qu.n.y is a: .pt.... {eR.E. Ginna Nuclear Power PlantB 3.4.15-5Revision 62 RCS Leakage Detection Instrumentation B 3.4.15SR 3.4.15.4See SR 3.4.

15.3REFERENCES

1. Atomic Industry Forum (AIF) GDC 16 and 34, Issued for commentJuly 10, 1967.2. Regulatory Guide 1.45.3. IE Bulletin No. 80-24, "Prevention of Damage Due to WaterLeakage Inside Containment."

4. NUREG-0609, "Asymmetric Blowdown Loads on PWR PrimarySystems,"

1981.5. Generic Letter 84-04, "Safety Evaluation of Westinghouse TopicalReports Dealing With Elimination of Postulated Pipe Breaks inPWR Primary Main Loops."6. Letter from D. C. Dilanni, NRC, to R. W. Kober, RG&E,

Subject:

"Generic Letter 84-04," dated September 9, 1986.7. NUREG-0821, "Integrated Plant Safety Assessment, Systematic Evaluation

Program, R. E. Nuclear Power Plant," December 1982.8. Letter from Guy S. Vissing (NRC) to Robert C. Mecredy (RG&E),"Staff Review of the Submittal by Rochester Gas and ElectricCompany to Apply Leak-Before-Break Status to Portions of the R.E.Ginna Nuclear Power Plant Residual Heat Removal SystemPiping",

dated February 25, 1999.R.E. Ginna Nuclear Power PlantB 3.4.15-6Revision 62 RCS Specific ActivityB 3.4.16C.1If the gross specific activity is not within limit, the change within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> toMODE 3 and RCS average temperature

< 500OF lowers the saturation pressure of the reactor coolant below the setpoints ofthe main steamsafety valves and prevents automatically venting the SG to theenvironment in an SGTR event. The allowed Completion Time of 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />sis reasonable, based on operating experience, to reach MODE 3 below500OF from full power conditions in an orderly manner and withoutchallenging plant systems.SURVEILLANCE SR 3.4.16.1REQUIREMENTS This SR requires performing a gamma isotopic analysis as a measure ofthe gross specific activity of the reactor coolant at least en .. .v,,y 7While basically a quantitative measure of radionuclides with halflives longer than 15 minutes, excluding

iodines, this measurement is thesum of the degassed gamma activities and the gaseous gamma activities in the sample taken. This Surveillance provides an indication of anyincrease in gross specific activity.

Trending the results of this Surveillance allows proper remedial action tobe taken before reaching the LCO limit under normal operating conditions.

The Surveillance is applicable in MODES 1 and 2, and inMODE 3 with Tavg >_ 500OF. The 7 day Frtqu.n.y

..n.ide.S theunilkoliheed of a grooo fuel failuro duringM this timo.4\F.ý SR 3.4.16.2This SR is only performed in MODE 1 to ensure iodine remains withinlimits during normal operation and following fast power changes when7R 3 fuel failure is more likely to occur. The 14 day is adequate t,...........

., I ..The Frequency, between 2 and 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> after apower change > 15% RTP within a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period, is established becausethe iodine levels peak during this time following fuel failure; samples atother times would provide inaccurate results.SR 3.4.16.3A radiochemical analysis for E determination is required within 31 daysafter a minimum of 2 effective full power days and 20 days of MODE 1[INSERT 1 I operation have elaps since the reactor was last subcritical for at least48 hours and evey , --g (6 ,mnt,,) thereafter.

This ensures that theradioactive materials are at equilibrium so the analysis for E isrepresentative and not skewed by a crud burst or other similar abnormalevent. The E determination directly relates to the LCO and is required toverify plant operation within the specified gross activity LCO limit. TheR.E. Ginna Nuclear Power PlantB 3.4.16-4Revision 42 RCS Specific ActivityB 3.4.16analysis for E is a measurement of the average energies perdisintegration for isotopes with half lives longer than 15 minutes, JINSERT 3excluding iodines .... The u.....y Fc.g.. i. -E .... .. ....haigc. d, V,,.This SR is modified by a Note that indicates sampling is only required tobe performed in MODE 1 such that equilibrium conditions are presentduring the sample.REFERENCES

1. 10 CFR 50.67.2. Design Analysis DA-NS-2001-084, Steam Generator Tube RuptureOffsite and Control Room Doses.IR.E. Ginna Nuclear Power PlantB 3.4.16-5Revision 42 Accumulators B 3.5.1power to the valve, or restore the proper water volume or nitrogen coverpressure ensures that prompt action will be faken to return the inoperable accumulator to OPERABLE status. The Completion Time minimizes thepotential for exposure of the plant to a LOCA under these conditions.

The 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> allowed to restore an inoperable accumulator toOPERABLE status is justified in WCAP-1 5049-A, Rev. 1 (Ref. 10).C.1 and C.2If the accumulator cannot be returned to OPERABLE status within theassociated Completion Time, the plant must be brought to a MODE inwhich the LCO does not apply. To achieve this status, the plant must bebrought to MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and pressurizer pressure reduced to<_ 1600 psig within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The allowed Completion Times arereasonable, based on operating experience, to reach the required plantconditions from full power conditions in an orderly manner and withoutchallenging plant systems.D.1If both accumulators are inoperable, the plant is in a condition outside theaccident analyses; therefore, LCO 3.0.3 must be entered immediately.

SURVEILLANCE SR 3.5.1.1REQUIREMENTS Each accumulator motor-operated isolation valve shall be verified to befully open eveyr 12 het ,. Use of control board indication for valveposition is an acceptable verification.

This verification ensures that theaccumulators are available for injection and ensures timely discovery if avalve should be less than fully open. If an isolation valve is not fully open,the rate of injection to the RCS would be reduced.

Although a motoroperated valve position should not change with power removed, a closedvalve could result in not meeting accident analyses assumptions.

Tie-y is ..n.id. .d ras.n. bl. in M ;vicw of ,the, administrativ.

.i1t .lIs that e,,,uer a isi,.c iiccd isolation va, c, unlikely..

JINSERT 3PR.E. Ginna Nuclear Power PlantB 3.5.1-6Revision 44 Accumulators B 3.5.1SR 3.5.1.2The borated water volume and nitrogen cover pressure shall be verifiedevery 12 "he.... for each accumulator.

This Fr..qucn.y is suffi...

nt toi nSur i adequate injctiin during n8 Ll.A. B 'eause .f the stati. dUsignief the accumulator-,

a 12 heur Froguoinoy usually allew th e~pcrater teidentif; changcs b.f... lim^itS " rc rcachcd.

Main control b'ard alarm.sWe else available fer a...umulater paramctorS.

The leveltransmitters for the accumulators measure the level over a 14" span forthe corresponding 0-100% level indicated on the main control board.O~porating experienee h~as shown this Froqueney te be appropriate foreEarly d eteetien and eorrootioni ef off normal1 tronds. -ýSR 3.5.1.3See SR 3.5.1.2SR 3.5.1.4The boron concentration shall be verified to be within required limits foreach accumulator cvcr; 12 h9Ur3 by Me .. it...n inlcakegc.

This isaccomplished by monitoring the level ineach accumulator evefy 12 .het.sand comparing to the previous level readings.

An unexplained increasein level could be an indication of inleakage and, therefore, dilution of theboron concentration.

If an unexplained increase in level is detected, theongoing change in boron concentration shall be determined bycalculation.

If the calculation indicates that the boron concentration haddecreased to within 100 ppm of the lower limit, the affected accumulator shall be sampled to confirm boron concentration.

In additin,accumulators shall be samgplcd eyer; 6 months to eeonfirmf that the borongconccentratien, infcrrcd frcm, inlakag t n limits.Six mneiths is roaseinablo for Ycrifioation by sampling to dctcrmnine thateaeh aeoumnulator's borong eonccn~trationl i within the roquirod limgits,bcoausc the static design of the aocumulaIterS limits the ways In whiehthc-eonccntratien ean be changcd.

This Frcqucnoy is adequate to identifyohanggos that could occur from mocehanismns, such as stratifloation or~ifeakage.

SR3...Verification eveiy-8-1-days that power is removed from each accumulator isolation valve operator when the pressurizer pressure is > 1600 psigensures that an active failure could not result in the undetected closure ofan accumulator motor operated isolation valve. If this were to occur, noaccumulators would be available for injection if the LOCA were to occurin the cold leg containing the only OPERABLE accumulator.

Siflee-pewet-i s romovod undcr administrativc eontrol and Yalvo position is Ycrificde-ve; 12 hoeurs, the 31 day Froguency will provide adequate asSUranee that power is rcmoved.R.E. Ginna Nuclear Power PlantB 3.5.1-7Revision 44 ECCS -MODES 1, 2, and 3B 3.5.2B.1 and B.2If the inoperable train cannot be returned to OPERABLE status within theassociated Completion Time, the plant must be brought to a MODE inwhich the LCO does not apply. To achieve this status, the plant must bebrought to MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Theallowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full powerconditions in an orderly manner and without challenging plant systems.C.1If both trains of ECCS are inoperable, the plant is in a condition outsidethe accident analyses; therefore, LCO 3.0.3 must be immediately entered.

With one or more component(s) inoperable such that 100% ofthe flow equivalent to a single OPERABLE ECCS train is not available, the facility is in a condition outside the accident analysis.

Therefore, LCO3.0.3 must be immediately entered.SURVEILLANCE SR 3.5.2.1REQUIREMENTS Verification of proper valve position ensures that the flow path from theECCS pumps to the RCS is maintained.

Use of control board indication for valve position is an acceptable verification.

Misalignment of thesevalves could render both ECCS trains inoperable.

The listed valves aresecured in position by removal of AC power or key locking the DC cortrolpower. These valves are operated under administrative controls suchthat any changes with respect to the position of the valve breakers or keylocks is unlikely.

The verification of the valve breakers and key locks isperformed by SR 3.5.2.3.

Mispositioning of these valves can disable thefunction of both ECCS trains and invalidate the accident analyses.

A-12-heur Froequ Ic is c~dercd rea3CnabC vin w V eWf ethcr adMini~traltive ccntroks that enGurc a mtspesitiened valve i3 ulikety.SR 3.5.2.2Verifying the correct alignment for manual, power operated, andautomatic valves in the ECCS flow paths provides assurance that theproper flow paths will exist for ECCS operation.

This SR does not applyto valves that are locked, sealed, or otherwise secured in position, sincethese were verified to be in the correct position prior to locking,

sealing, orsecuring.

A valve that receives an actuation signal is allowed to be in anonaccident position provided the valve will automatically reposition within the proper stroke time. This Surveillance does not require anytesting or valve manipulation.

Rather, it involves verification that thosevalves capable of being mispositioned are in the correct position.

The 8day Frcequcnc i3 ppciate beeause the valves arc epcraited underadminictratiye

ecntrcl, and an improper valve pesitien in mcest eases,R.E. Ginna Nuclear Power PlantB 3.5.2-11Revision 58 ECCS -MODES 1, 2, andB 3.5.w.uld only affect a single train. This F..qu.n.y has been sh.wn t. beaeeoptabic through epcralting experienee.

43.2LIINSE ý[JýSR 3.5.2.3Verification e.v.ey..

31 AC or DC power is removed, asappropriate, for each valve specified in SR 3.5.2.1 ensures that an activefailure could not result in an undetected misposition of a valve whichaffects both trains of ECCS. If this were to occur, no ECCS injection orrecirculation would be available.

Since power is romovo.d und.admfinistrativo oontrol and valve position is YE~ified evor; 12 hourS, the 31day Frequen^y will pro.vide adequate aSSUra.nc.

SR 3.5.2.4that power i 0 remv.MVed.

LiNSERT 3tIIPeriodic surveillance testing of ECCS pumps to detect gross degradation caused by impeller structural damage or other hydraulic component problems is required by the ASME Code. This type of testing may beaccomplished by measuring the pump developed head at a single pointof the pump characteristic curve. This verifies both that the measuredperformance is within an acceptable tolerance of the original pumpbaseline performance and that the performance at the test flow is greaterthan or equal to the performance assumed in the plant safety analysis.

SRs are specified in the Inservice Testing Program, which encompasses the ASME Code. The ASME Code provides the activities andFrequencies necessary to satisfy the requirements.

SR 3.5.2.5These Surveillances demonstrate that each automatic ECCS valveactuates to the required position on an actual or simulated SI signal andthat each ECCS pump starts on receipt of an actual or simulated SIsignal. This surveillance is not required for valves that are locked,sealed, or otherwise secured in the required position under administrative controls.

The 24 month Frequency is based on the need to thseSurf'eillanees under the conditions that apply during a plant cutagc andthe potential for unplanned plant tralnsicnts if the Swr~eillanees wercperformned with the reactor at power-. The 24 molnth Frequenoy is elseaeceptable based en considcration of the design roliability (anconlfirmning epcrating cxperienee) of the eqjpmonet.

The aetuation logic istested as pa.. of ESF Aetuation System. testing, and equipmenperformifanec is monitored as part of the lnseryiee Testing Programff.,

[INSERT 3R.E. Ginna Nuclear Power PlantB 3.5.2-12Revision 58 ECCS -MODES 1, 2, and 3B 3.5.2SR 3.5.2.6See SR 3.5.2.5SR 3.5.2.7Periodic inspections of the containment sump suction inlet to the RHRSystem ensure that it is unrestricted and stays in proper operating condition. 24 m^centh Frc..uen.y

i. based en the need t. Fm this'Survc"Ilnec undcr the eenditions that apply during a plant eutage, andu,- v....,, ,,.... n ,.r..., ,,,,. ,.v,,.,,,,,.

VrtknThF-u irn~h- r~ r. rto be 'uffie"^nt te dcteet abnormal dr,*pfitn -,x ..... enNSE Tdntlpnand is byREFERENCES

1. Letter from R. A. Purple, NRC, to L. D. White, RG&E,

Subject:

"Issuance of Amendment 7 to Provisional Operating License No.DPR-1 8," dated May 14, 1975.2. Branch Technical Position (BTP) ICSB-1 8, "Application of theSingle Failure Criterion to Manually-Controlled Electrically Operated Valves."3. Letter from A. R. Johnson, NRC, to R. C. Mecredy, RG&E,

Subject:

"Issuance of Amendment No. 42 to Facility Operating License No.DPR-18, R. E. Ginna Nuclear Power Plant (TAC No. 79829),"

datedJune 3, 1991.4. Letter from D. M. Crutchfield, NRC, to J. E. Maier, RG&E,

Subject:

"SEP Topic VI-7.B: ESF Switchover from Injection to Recirculation Mode, Automatic ECCS Realignment, Ginna," dated December 31,1981.5. NUREG-0821.

6. UFSAR, Section 6.3.7. Not Used8. Atomic Industrial Forum (AIF) GDC 44, Issued for comment July10, 1967.9. 10 CFR 50.46.10. UFSAR, Section 15.6.11. UFSAR, Section 6.2.R.E. Ginna Nuclear Power PlantB 3.5.2-13Revision 58 RWSTB 3.5.4ACTIONS A._1With RWST boron concentration not within limits, it must be returned towithin limits within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Under these conditions neither the ECCS northe CS System can perform its design function.

Therefore, prompt actionmust be taken to restore the tank to OPERABLE condition.

The 8 hourlimit to restore the RWST boron concentration to within limits wasdeveloped considering the time required to change the boronconcentration and the fact that the contents of the tank are still available for injection.

B. 1With the RWST water volume not within limits, it must be restored toOPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. In this Condition, neither the ECCS northe CS System can perform its design function.

Therefore, prompt actionmust be taken to restore the tank to OPERABLE status or to place theplant in a MODE in which the RWST is not required.

The short time limitof 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to restore the RWST to OPERABLE status is based on thiscondition simultaneously affecting redundant trains.C.1 and C.2If the RWST cannot be returned to OPERABLE status within theassociated Completion Time, the plant must be brought to a MODE inwhich the LCO does not apply. To achieve this status, the plant must bebrought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36hours. The allowed Completion Times are reasonable, based onoperating experience, to reach the required plant conditions from fullpower conditions in an orderly manner and without challenging plantsystems.SURVEILLANCE SR 3.5.4.1REQUIREMENTS The RWST water volume should be verified evey" 7,,,.. to be above therequired minimum level in order to ensure that a sufficient initial supply isavailable for injection and to support continued ECCS and CS Systempump operation on recirculation.

Sin:^ the RWST^ vel- m ,. i* .....,,Ystablc and the RWSTF is lccatcd in the Auxiiiar;y Building which prcViedcs suffieicnt Icak deteetien eapability, a :7 daY Froquonc is pppiate andhas been shewn to be acooptable throlugh epefrating cxpeRionoo.

INSERTl 3'... 7 p. -R.E. Ginna Nuclear Power PlantB 3.5.4-4Revision 42 RWSTB 3.5.4SR 3.5.4.2The boron concentration of the RWST should be verified evrey 74 .dys tobe within the required limits. This SR ensures that the reactor will remainsubcritical following a LOCA. Further, it assures that the resulting sumppH will be maintained in an acceptable range so that boron precipitation in the core will not occur and the effect of chloride and caustic stresscorrosion on mechanical systems and components will be minimized.

S*ien the RW"AIT vlu,, i" n,1.lly stable, a 7 day ytc Ycrify beron conoontratiein ia 8ppropriato and has been shown to beacccptable through epcrating expcrienee.

REFERENCES

1. UFSAR, Section 3.11.2. 10 CFR 50.49.3. UFSAR, Section 6.3 and Chapter 15.R.E. Ginna Nuclear Power PlantB 3.5.4-5Revision 42 Containment Air LocksB 3.6.2SR 3.6.2.2The air lock interlock is designed to prevent simultaneous opening of bothdoors in a single air lock. Since both the inner and outer doors of an airlock are designed to withstand the maximum expected post accidentcontainment

pressure, closure of either door will support containment OPERABILITY.

Thus, the door interlock feature supports containment OPERABILITY while the air lock is being used for personnel transit in andout of the containment.

Periodic testing of this interlock demonstrates that the interlock will function as designed and that simultaneous openingof the inner and outer doors will not inadvertently occur. -Due-te-the-purely mocehanical naturo of this interlock, and given that the interleeck mocehanism is only challenged when the containment airlock door i5epencd, this test is enly rcgUircd tO be pe~fefmed onco every 24 moenths.The 24 mnth n. y i. base ein g judgment and is~nidered adequate in view of ethor indication

-fdo aditelm~echanis~m status available to operaitiOns personolREFERENCES

1. UFSAR, Section 6.2.1.1.2. 10 CFR 50, Appendix J, Option B.R.E. Ginna Nuclear Power PlantB 3.6.2-7Revision 21 Containment Isolation Boundaries B 3.6.3E.2Required Action E.2 requires that the mini-purge valve leakage must berestored to within limits, or the affected penetration flow path must beisolated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The method of isolation must be by the use of atleast one isolation barrier that cannot be adversely affected by a singleactive failure (including a single human error). For automatic valves, thisrequires two independent means to prevent the valve from re-opening.

Isolation barriers that meet this criterion are a closed and de-activated automatic valve, closed manual valve, or blind flange. A purge valve withresilient seals utilized to satisfy Required Action E.2 must have beendemonstrated to meet the leakage requirements of SR 3.6.3.5.

Thespecified Completion Time is reasonable, considering that onecontainment purge valve remains closed so that a major violation ofcontainment does not exist.Following completion of Required Action E. 1, verification that the affectedpenetration flow path remains isolated must be performed in accordance with Required Action D.2.F.1 and F.2If the Required Actions and associated Completion Times are not met,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 at least MODE 3within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach therequired plant conditions from full power conditions in an orderly mannerand without challenging plant systems.SURVEILLANCE SR 3.6.3.1REQUIREMENTS This SR ensures that the mini-purge valves are closed except when thevalves are opened under administrative control.

The mini-purge valvesare capable of closing in the environment following a LOCA. Therefore, these valves are allowed to be open for limited periods of time. Thevalves may be opened for pressure

control, ALARA or air qualityconsiderations for personnel entry, maintenance activities, operational requirements, or for Surveillances that require the valves to be open. Tobe opened, the valves must be capable of closing under accidentconditions, the containment isolation signal to the valves must beOPERABLE, and the effluent release must be monitored to ensure that itremains within regulatory limits. 31 day Fr..qu.n.y is based en th^c.rolative im~pertanee of these Yalyes sinec they provide a dircoet path to the-eutsfidc cnyergnment and the administrativc ccngtrols that are in plaec.I I N S I,R.E. Ginna Nuclear Power PlantB 3.6.3-11Revision 64 Containment Isolation Boundaries B 3.6.3SR 3.6.3.2This SR requires verification that each containment isolation boundarylocated outside containment and not locked, sealed or otherwise securedin the required position is performing its containment isolation accidentfunction.

Containment isolation boundaries located beneath Appendix Rfire wrap may be considered secured in the required position due to theadministrative controls in place provided that a verification of theboundary position was made prior to securing the fire wrap. The SRhelps to ensure that post accident leakage of radioactive fluids or gasesoutside of the containment barrier is within design limits. This SR doesnot require any testing or valve manipulation.

Rather, it involvesverification, through a system walkdown, that those isolation boundaries outside containment and capable of being mispositioned are in thecorrect position.

This includes manual valves, blind flanges, pipe andend caps, and closed systems.

.ine. e.ntainmcnt is.lati.

n b8un.a.io.

Wrc maintained under admfini~tratiye eentrols with eentainmmnt uselation beundary tags installed, the probability of their miselignment is low and a92 day Frogucney to verif' thoir corroct pesition is appr.ria.

' t".e SRspecifies that isolation boundaries that are open under admin=istra t Ncontrols are not required to meet the SR during the time the boundarie) are open. JINET3The SR is modified by two notes. The first Note applies to containment isolation boundaries located in high radiation areas and allows theseboundaries to be verified closed by use of administrative means.Allowing verification by administrative means (e.g., procedure control) isconsidered acceptable, since access to these areas is typically restricted during MODES 1, 2, 3 and 4 for ALARA reasons.

Therefore, theprobability of misalignment of these isolation boundaries, once they havebeen verified to be in the proper position, is small. The Second Notestates that this SR is not applicable to containment isolation boundaries which receive an automatic signal since the isolation times of thesevalves are verified by SR 3.6.3.4 and the boundaries are required to beOPERABLE.

SR 3.6.3.3This SR requires verification that each containment isolation boundarylocated inside containment and not locked, sealed or otherwise securedin the required position is performing its containment isolation accidentfunction.

The SR helps to ensure that post accident leakage ofradioactive fluids or gases outside of the containment barrier is withindesign limits. This SR does not require any testing or valve manipulation.

Rather, it involves verification, through a system walkdown, that thoseisolation boundaries inside containment and capable of beingmispositioned are in the correct position.

This includes manual valves,blind flanges, pipe and end caps, and closed systems.

Sincecontainment isolation boundaries are maintained under administrative controls with containment isolation boundary tags installed, theR.E. Ginna Nuclear Power PlantB 3.6.3-12Revision 64 Containment Isolation Boundaries B 3.6.3SR 3.6.3.6Automatic containment isolation valves close on a containment isolation signal to prevent leakage of radioactive material from containment following a DBA. This SR ensures that each automatic containment isolation valve will actuate to its isolation position on a containment isolation signal. This surveillance is not required for valves that arelocked, sealed, or otherwise secured in the required position underadministrative controls. 24 m-nth Fr..u.ncy is based on the needto pci-formf this Sur~'cillanee under the conditions that apply during a planteutagc and the pctential for an9 unplanned transient if the -Sur~'cillancc wcrc peffefficd with the roacter Bt pewcr. Operating cxpcricncc hasshown that these compenents usually pass this Surveillanee whenpecfeFmcd at the 24 mcn~th Frogueney.

Thercfcrc, the Frcgucney wasnudd to be acc.ptab.

fr.m a liability standp.i..t..

INS.ERT 3JREFERENCES

1. Atomic Industry Forum GDC 53 and 57, issued for comment July10, 1967.2. Branch Technical Position CSB 6-4, "Containment Purging DuringNormal Operation."

3. UFSAR, Section 6.2.4 and Table 6.2-15.4. Regulatory Guide 1.4, Revision 2.5. 10 CFR 50, Appendix A, GDC 55, 56, and 57.6. Ginna Station Procedure A-3.3.7. NUREG-0800, Section 6.2.4.R.E. Ginna Nuclear Power PlantB 3.6.3-14Revision 64 Containment PressureB 3.6.4to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. However, due to the largecontainment free volume and limited size of the containment Mini-Purge System, 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is allowed to restore containment pressure to withinlimits. This is justified by the low probability of a DBA during this timeperiod.B.1 and B.2If containment pressure cannot be restored to within limits within therequired Completion Time, the plant must be brought to a MODE in whichthe LCO does not apply. To achieve this status, the plant must bebrought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36hours. The allowed Completion Times are reasonable, based onoperating experience, to reach the required plant conditions from fullpower conditions in an orderly manner and without challenging plantsystems.SURVEILLANCE REQUIREMENTS SR 3.6.4.1Verifying that containment pressure is within limits ensures that plantoperation remains within the limits assumed in the containment analysis.

This verification should normally be performed using PI-944. The--, Frcquenc';

ef thus SR wasdevelped based en eperoting

.xprincrclt.d te trcnding

.f containment prcaUr. vYrBitions durng th'^applicable MO)DES. FwuthefRmcr, the 12 heur Froquenoy is considcrod-adequate min vie f ether indieatwions available lin the eentrol roomfi,9including

alarmis, to alert the opeffater to an9 abnormfal containmoent Calibration of PI-944 or other containment pressure monitoring devicesshould be performed in accordance with industry standards.

REFERENCES

1. UFSAR, Section 6.2.1.2.2. 10 CFR 50, Appendix K.R.E. Ginna Nuclear Power PlantB 3.6.4-3Revision 72 Containment Air Temperature B 3.6.5containment average air temperature within the limit is not required inMODE 5 or 6.ACTIONS A.1When containment average air temperature is not within the limit of theLCO, it must be restored to within the limit within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This RequiredAction is necessary to return operation to within the bounds of thecontainment analysis.

The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Completion Time is acceptable considering the sensitivity of the analysis to variations in this parameter and provides sufficient time to correct minor problems.

B.1 and B.2If the containment average air temperature cannot be restored to withinits limit within the required Completion Time, the plant must be brought toa MODE in which the LCO does not apply. To achieve this status, theplant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, basedon operating experience, to reach the required plant conditions from fullpower conditions in an orderly manner and without challenging plantsystems.SURVEILLANCE SR 3.6.5.1REQUIREMENTS Verifying that containment average air temperature is within the LCO limitensures that containment operation remains within the limit assumed forthe containment analyses.

In order to determine the containment average air temperature, an arithmetic average is calculated usingmeasurements taken at locations within the containment selected toprovide a representative sample of the overall containment atmosphere.

There are 6 containment air temperature indicators (TE-6031, TE-6035,TE-6036, TE-6037, TE-6038, and TE-6045) such that a minimum of threeshould be used for calculating the arithmetic average.

The 4 2 Froquoney of this SR is eenoidorod aeooptablo based en ebserved slewFates .f t,.mpr,,tu,.

in......

within oontainm.

nt as a ..sult , fonvionmota!

heat Sourocs (due to the lorgo velumo of eontainment).

FufthefRmor, the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frogquonoy is eensiderod adequate in view efethor indioctiefns available in the ecntrcl rccm to alc~t the operater te anabnormI9al eentafinmont teffpefraturoeni

^" .. .^-'" .. " .... ...co .'.ilc INSERT 3/Calibration of these temperature indicators shall be performed inaccordance with industry standards.

R.E. Ginna Nuclear Power PlantB 3.6.5-3Revision 72 CS, CRFC and NaOH SystemsB 3.6.6D._ lWith one or two CRFC units inoperable, the inoperable CRFC unit(s)must be restored to OPERABLE status within 7 days. The inoperable CRFC units provided up to 100% of the containment heat removal needs.The 7 day Completion Time is justified considering the redundant heatremoval capabilities afforded by combinations of the CS System andCRFC System and the low probability of DBA occurring during thisperiod.E.1 and E.2If the Required Action and associated Completion Time of Condition D ofthis LCO are not met, the plant must be brought to a MODE in which theLCO does not apply. To achieve this status, the plant must be brought toat least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Theallowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full powerconditions in an orderly manner and without challenging plant systems.F. 1With two CS trains inoperable, or three or more CRFC units inoperable, the plant is in a condition outside the accident analysis.

Therefore, LCO3.0.3 must be entered immediately.

SURVEILLANCE SR 3.6.6.1REQUIREMENTS The applicable SR descriptions from Bases 3.5.2 apply. This SR isrequired since the OPERABILITY of valves 896A and 896B is alsorequired for the CS System.SR 3.6.6.2Verifying the correct alignment for manual, power operated, andautomatic valves in the CS flow path provides assurance that the properflow paths will exist for CS System operation.

This SR does not apply tovalves that are locked, sealed, or otherwise secured in position, sincethese were verified to be in the correct position prior to locking,

sealing, orsecuring.

This SR does not require any testing or valve manipulation.

Rather, it involves verification, through a system walkdown, that thosevalves outside containment (there are no valves inside containment) andcapable of potentially being mispositioned are in the correct position.

R.E. Ginna Nuclear Power PlantB 3.6.6-8Revision 72 CS, CRFC and NaOH SystemsB 3.6.6SR 3.6.6.3Verifying the correct alignment for manual, power operated, andautomatic valves in the NaOH System flow path provides assurance thatthe proper flow paths will exist for NaOH System operation.

This SRdoes not apply to valves that are locked, sealed, or otherwise secured inposition, since these were verified to be in the correct position prior tolocking,

sealing, or securing.

This SR does not require any testing orvalve manipulation.

Rather, it involves verification, through a systemwalkdown, that those valves outside containment (there are no valvesinside containment) and capable of potentially being mispositioned are inthe correct position. SR 3.6.6.4 NOperating each CRFC unit for _ 15 minutes onco cvcry 31 days ensuresthat all CRFC units are OPERABLE and that all associated controls arefunctioning properly.

It also ensures that blockage, fan or motor failure,damper failures, or excessive vibration can be detected for corrective action. The A and C CRFC units must be operated with their respective charcoal filter train in the post accident alignment.

The 31 day F..qu.n.y was devcleped considoring the knoewn Fliability of the fan units andeentrols, the rodundanoey available, and the low probability of significant dcgradation of the CRFC units occurring0 bcevccn survcillanees.

It haselso been shown to be acccptable through opefrating expcriicnco.,

SIR 3.6.6.5Verifying cooling water (i.e., SW) flow to each CRFC unit providesassurance that the energy removal capability of the CRFC assumed inthe accident analyses will be achieved (Ref. 11). The minimum andmaximum SW flows are not required to be specifically determined by thisSR due to the potential for a containment air temperature transient.

Instead, this SR verifies that SW flow is available to each CRFC unit.The 31 day Froguency was develeped congsidcring the known roliability of-the GW Systcmn, the twe CRFC train rcdundaney available, and the lowprobability of a significant degradlation of flew occurring bcevocnswefIfiaees-SIR 3.6.6.6 'INER3TVerifying each CS pump's developed head at the flow test point is greaterthan or equal to the required developed head ensures that spray pumpperformance has not degraded during the cycle. Flow and differential pressure are normal tests of centrifugal pump performance required bythe ASME Code (Ref. 12). Since the CS pumps cannot be tested withflow through the spray headers, they are tested on recirculation flow.This test confirms one point on the pump design curve and is indicative ofoverall performance.

Such inservice testing confirms component OPERABILITY, trends performance, and detects incipient failures byR.E. Ginna Nuclear Power PlantB 3.6.6-9Revision 72 CS, CRFC and NaOH SystemsB 3.6.6abnormal performance.

The Frequency of the SR is in accordance withthe Inservice Testing Program.SR 3.6.6.7To provide effective iodine removal, the containment spray must be analkaline solution.

Since the RWST contents are normally acidic, thespray additive tank must provide a sufficient volume of spray additive toadjust pH for all water that is injected.

This SR is performed to verify theavailability of sufficient NaOH solution in the spray additive tank. Fhe484-day Frcqucncy was d...lpcd based .n the l.w , r... i-ity cf anundeteeted ehanggo in tank velume eeeurring during the SR interval 3inccthe tanqk as ,.R,,lly 1 .Tank level is also indicated and alarmed inthe control room, so that ther is high confidence that a substantial change in level would be dete ted.SR 3.6.6.8 INSERT 3This SR provides verification of the NaOH concentration in the sprayadditive tank and is sufficient to ensure that the spray solution beinginjected into containment is at the correct pH level. The 484 dayFrogueney is suffiefient te encUrc that the eeneentratien lcvcl ef NeOH inthe SPpraY additive tanik rcmain3 within the established lifnits.

This asbased on the lew likcliheed ef an uneentrclled ehag incccctratien sencc the tank ic nrm~fally iselated and the Prcbabii tha ay ubqtantial varianee

• in tank velumc will be detCeCe4\.SR 3.6.6.9This SR verifies that the required CRFC unit testing is performed inaccordance with the VFTP. The VFTP includes testing HEPA filterperformance.

The minimum required flow rate through each of the fourCRFC units is 33,000 cubic feet per minute at accident conditions (or38,500 cubic feet per minute at normal operating conditions).

Specifictest frequencies and additional information are discussed in detail in theVFTP. However, the maximum surveillance interval for refueling outagetests is based on 24 month refueling cycles and not 18 month cycles asdefined by Regulatory Guide 1.52 (Ref. 13).S R 3.6.6.10These SRs require verification that each automatic CS valve in theflowpath (860A and 860D) actuates to its correct position and that eachCS pump starts upon receipt of an actual or simulated actuation of acontainment High pressure signal. This surveillance is not required forvalves that are locked, sealed, or otherwise secured in the requiredposition under administrative controls.

The 24 moenth Froguoncey is baseden the need to pcrferM these Safveillanees undor the eonditicns thatthe Survo"ieneop wero perfermed with the roaotfr at pewcr. Opefrating expericnee has shcwn that thesc eempenents usually pass thioR.E. Ginna Nuclear Power PlantB 3.6.6-10Revision 72 CS, CRFC and NaOH SystemsB 3.6.6Survcillanecs when perfefrmcd at the 24 moenth F-r-qucney.

Thcrofero, the Froequeney was ccncludcd to be acocptable frcmA a roliability-Stf,,.dp ei,.t. 11 d I SE T 3SR 3.6.6.11

,INSERT3See SR 3.6.6.10SR 3.6.6.12This SR requires verification that each CRFC unit, and the charcoal filtertrain associated with the A and C units, actuates upon receipt of an actualor simulated safety injection signal. The 24 month Fc..qu.n.y is basedn ig judgment and has been shown to *b,,-he aeeptablo thr.ughEprtn cPcricncc.

See SR 3.6.6.10 and SR 3.6.6.11, abeve-,6for further diseussion of the basis for the 24 mcnth Froequeney.

,SR 3.6.6.13This SR provides verification that each automatic valve in the NaOHSystem flow path that is not locked, sealed, or otherwise secured inposition (836A and 836B) actuates to its correct position upon receipt ofan actual or simulated actuation of a containment Hi-Hi pressure signal.The 24 mo.nth frogu.ncy is based...

on cginccrng judge.m.t and hasbeen shewn to be aeecptable through oprtnexpeReiono.

See SR3.6.6.10 and oSR 3.6.6.11, abov, for futh'"r^

dis"-ussion of the baiS forthe 24 mcenth Frequeonoy.

,SR 3.6.6.14To ensure that the correct pH level is established in the borated watersolution provided by the CS System, flow through the eductor is verified,onoc cvey 5 yceFs. This SR in conjunction with SR 3.6.6.13 providesJINSERT 1 "assurance that NaOH will be added into the flow path upon CS initiation.

A minimum flow of 20 gpm through the eductor must be established asassumed in the accident analyses.

A flow path must also be verified fromthe NaOH tank to the eductors.

Due to the ur* of the SpFayadditive flew controls, the 6 ycar Frouoc Is .ufficiont to identify.d....adation that mayf.. t flew injeetion.

SR 3.6.6.15 3With the CS inlet valves closed and the spray header drained of anysolution, low pressure air or smoke can be blown through testconnections.

As an alternative, a visual inspection (e.g. boroscope) ofthe nozzles or piping could be utilized in lieu of an air or smoke test if avisual inspection is determined to provide an equivalent or a moreeffective post-maintenance test. A visual inspection may be moreeffective if the potential for material intrusion is localized and the affectedarea is accessible.

This SR ensures that each spray nozzle isunobstructed and provides assurance that spray coverage of thecontainment during an accident is not degraded.

Due to the passiveR.E. Ginna Nuclear Power PlantB 3.6.6-11Revision 72 MSIVs and Non-Return Check ValvesB 3.7.2SR 3.7.2.3This SR verifies that each MSIV can close on an actual or simulated actuation signal. This Surveillance is normally performed upon returning the plant to operation following a refueling outage. The MSIVs should notbe tested at power, since even a partial stroke exercise increases the riskof a valve closure and plant transient when the plant is above MODE 4.As the MSIVs are not tested at power, they are exempt from the ASMECode (Ref. 5), requirements during operation in MODES 1, 2 and 3.The~ ffequejnew ef SR'V tes-toinn asrr ev94 Fflenths The 24 menthFrogueney fer testing is based encxperienee has shown that theseSurveillamee when performoed attthe rcfucling eyele. Gperatineemponcnts usually pass the-e- 24 month

.

thi-liability stardp3Reaueney is aeeemahle ffem a FREFERENCES

1. UFSAR, Section 5.4.4.2. UFSAR, Section 15.1.5.3. UFSAR, Section 3.6.2.5.1.

4. 10 CFR 50.67.5. ASME Code for Operation and Maintenance of Nuclear PowerPlants.IR.E. Ginna Nuclear Power PlantB 3.7.2-6Revision 58 ARVsB 3.7.4SURVEILLANCE SR 3.7.4.1REQUIREMENTS To perform a cooldown of the RCS, the ARVs must be able to be openedeither remotely or locally.

This SR ensures that the ARVs are testedthrough a full control cycle at least once per fuel cycle. Performance ofinservice testing or use of an ARV during a plant cooldown may satisfythis requirement

.... ....ting expi'n" has that thesecompenents usually pass the Suryeillanee whein pcrfermed at the 24menth Froquenoy.

The Froquoney is aeccptable fromn a roliabilt SR 3.7.4.2The function of the block valve is to isolate a failed open ARV. Cyclingthe block valve both closed and open demonstrates its capability toperform this function.

Performance of inservice testing or use of theblock valve during plant cooldown may satisfy this requirement.

Operating cxpericnco has shewn that these eompenonts usually pass theSeu..illanoc when p...r.fo..d at the 24 mone^th TheFrogquonoy is eeecptable Yfrom Et rliability standpoinA-REFERENCES

1. UFSAR, Section 10.3.2.5.

2. UFSAR, Section 15.6.3.3. UFSAR, Section 15.1.6.R.E. Ginna Nuclear Power PlantB 3.7.4-4Revision 69 AFW SystemB 3.7.5plant should not be perturbed by any action, including a power change,that might result in a trip. The seriousness of this condition requires thataction be started immediately to restore one MDAFW, TDAFW, or SAFWtrain to OPERABLE status. For the purposes of this Required Action,only one TDAFW train flow path and the pump must be restored to exitthis Condition.

Required Action H.1 is modified by a Note indicating that all requiredMODE changes or power reductions are suspended until one MDAFW,TDAFW, or SAFW train is restored to OPERABLE status. In this case,LCO 3.0.3 is not applicable because it could force the plant into a lesssafe condition.

SURVEILLANCE SR 3.7.5.1REQUIREMENTS Verifying the correct alignment for manual, power operated, andautomatic valves in the AFW and SAFW System water and steam supplyflow paths provides assurance that the proper flow paths will exist forAFW operation.

This SR does not apply to valves that are locked,sealed, or otherwise secured in position, since they are verified to be inthe correct position prior to locking,

sealing, or securing.

This SR alsodoes not apply to valves that cannot be inadvertently misaligned, such ascheck valves. This Surveillance does not require any testing or valvemanipulation; rather, it involves verification, through a system walkdown, that those valves capable of being mispositioned are in the correctposition.

The 31 day Frcqueney 09 based an cngnccn judgment, us eensistcnt with the procadural eeontrols gavarning Wev pcain, and ecnzUrcscorrcct valve pesitiens.

SR 3.7.5.2 3Periodically comparing the reference differential pressure and flow ofeach AFW pump in accordance with the inservice testing requirements ofthe ASME Code (Ref. 4) detects trends that might be indicative of anincipient failure.

The Frequency of this surveillance is specified in theInservice Testing Program, which encompasses the ASME Code. TheASME Code provides the activities and Frequencies necessary to satisfythis requirement.

This SR is modified by a Note indicating that the SR is only required to bemet prior to entering MODE 1 for the TDAFW pump since suitable testconditions have not been established.

This deferral is required becausethere is insufficient steam pressure to perform the test.R.E. Ginna Nuclear Power PlantB 3.7.5-8Revision 66 AFW SystemB 3.7.5SR 3.7.5.3Periodically comparing the reference differential pressure and flow ofeach SAFW pump in accordance with the inservice testing requirements of the ASME Code (Ref. 4) detects trends that might be indicative of anincipient failure.

Because it is undesirable to introduce SW into the SGswhile they are operating, this testing is performed using the testcondensate tank. The Frequency of this surveillance is specified in theInservice Testing Program, which encompasses the ASME Code. TheASME Code provides the activities and Frequencies necessary to satisfythis requirement.

SR 3.7.5.4This SR verifies that each AFW and SAFW motor operated suction valvefrom the SW System (4013, 4027, 4028, 9629A, and 9629B), each AFWand SAFW discharge motor operated valve (4007, 4008, 9701A, 9701B,9704A, 9704B, and 9746), and each SAFW cross-tie motor operatedvalve (9703A and 9703B) can be operated when required.

TheFrequency of this Surveillance is specified in the Inservice Test Programand is consistent with the ASME Code (Ref. 4). The TDAFW discharge motor operated valve (3996) is maintained open and not required to beclosed for the DBA's and transients described within the Applicable Safety Analyses section.

Therefore, testing of the TDAFW discharge motor operating valve is not required.

SR 3.7.5.5This SR verifies that AFW can be delivered to the appropriate SG in theevent of any accident or transient that generates an actuation signal, bydemonstrating that each automatic valve in the flow path actuates to itscorrect position on an actual or simulated actuation signal. ThisSurveillance is not required for valves that are locked, sealed, orotherwise secured in the required position under administrative controls.

The 24 menth Froequonoy is based on the need to pcrferm thKihSRveianSe undee the etnditions that apply durt ing a plant outage andthe potontial for an unplanned transicnt if the Sur~'ci~ancc worpcrfefrmod with the roaetcr at pewor. The 24 molnth Froequcnei aeeeptable based on eperating experienee and thec doi_ roiblity etthe eettipmenle.

SR 3.7.5.6 -ISR3TThis SR verifies that the AFW pumps will start in the event of anyaccident or transient that generates an actuation signal by demonstrating that each AFW pump starts automatically on an actual or simulated actuation signal. The 24 menth Fr..u.n.y is based en the needto porforFM this Gurveillancc undor the conditions that apply during 8 planteutage.R.E. Ginna Nuclear Power PlantB 3.7.5-9Revision 66 AFW SystemB 3.7.5This SR is modified by a Note indicating that the SR is only required to bemet prior to entering MODE 1 for the TDAFW pump since suitable testconditions may have not been established.

This deferral is requiredbecause there is insufficient steam pressure to perform the test.SR 3.7.5.7This SR verifies that the SAFW System can be actuated and controlled from the control room. The SAFW System is assumed to be manuallyinitiated within 14.5 minutes in the event that the preferred AFW Systemis inoperable.

This Surveillance includes the verification of the automatic response of the motor operated discharge valves (9701 A and 9701 B)and the recirculation valves (9710A and 9710B). The cf 24menths Is based en the need te perfeFRm thiS Suryeillanee under theecnditions that apply dur*ing a plant eutagc and the petential fer eanunplannced tramseeint of the Survcfillancc wcrc pcrfcjfedat pewecj.REFERENCES 1 .UFSAR, Section 10.5.2. UFSAR Chapter 15.3. American National

Standard, "Nuclear Safety Criteria for the Designof Stationary Pressurized Water Reactor Plants,"

N18.2-1 973.4. ASME Code for Operation and Maintenance of Nuclear PowerPlants.R.E. Ginna Nuclear Power PlantB 3.7.5-10Revision 66 CSTsB 3.7.6ACTIONS A.1 and A.2If the CST water volume is not within limits, the OPERABILITY of thebackup supply should be verified by administrative means within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.OPERABILITY of the backup feedwater supply must include verification that the flow paths from the backup water supply to the preferred AFWpumps are OPERABLE and immediately available upon AFW initiation, and that the backup supply has the required volume of water available.

Alternate sources of water include, but is not limited to, the SW Systemand the all-volatile-treatment condensate tank. In addition, the CSTsmust be restored to OPERABLE status within 7 days, because thebackup supply may be performing this function in addition to its normalfunctions.

Continued verification of the backup supply is not required dueto the large volume of water typically available from these alternate sources.

The 7 day Completion Time is reasonable, based on anOPERABLE backup water supply being available, and the low probability of an event occurring during this time period requiring the CSTs.B.1 and B.2If the backup supply cannot be verified or the CSTs cannot be restored toOPERABLE status within the associated Completion Time, the plant mustbe placed in a MODE in which the LCO does not apply. To achieve thisstatus, the plant must be placed in at least MODE 3 wilhin 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and inMODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The allowed Completion Times arereasonable, based on operating experience, to reach the required plantconditions from full power conditions in an orderly manner and withoutchallenging plant systems.SURVEILLANCE SR 3.7.6.1REQUIREMENTS This SR verifies that the CSTs contain the required volume of coolingwater. The 24,350 gal minimum volume is met if one CST is >_ 22.8 ft(including instrument uncertainty) or if both CSTs are >_ 13.6 ft (including instrument uncertainty)..

12 ^- bcxpericncc and the need fer epcr-ater awarcness ef plant evelutiens that.- ...... hint... ............

1q A ,Qq. ,t== Q iF-e'uen^y is :-nsiderTd adequate in view 3f ethi teentrol room, ineluding alarms, to alert the epefrater te abnormadcviations in the CST- lvel-.the1R.E. Ginna Nuclear Power PlantB 3.7.6-3Revision 60 CCW SystemB 3.7.7SURVEILLANCE SR 3.7.7.1REQUIREMENTS Verifying the correct alignment for manual and power operated valves inthe CCW flow path provides assurance that the proper flkw paths exist forCCW operation.

This SR does not apply to valves that are locked,sealed, or otherwise secured in position, since these valves are verifiedto be in the correct position prior to locking,

sealing, or securing.

This SRalso does not apply to valves that cannot be inadvertently misaligned, such as check valves. This Surveillance does not require any testing orvalve manipulation; rather, it involves verification, through a systemwalkdown, that those valves capable of being mispositioned are in thecorrect position.

The 31 day F-r..u...y is based en cngir ^ crig judgment, as "ntwith the p....dur.l c.ntr.ls gc,;cing

'alvc ....at'"n, and ..S... seerroct valve pesitiens.

This SR is modified by a Note indicating that the isolation of the CCWflow to individual components may render those components inoperable but does not affect the OPERABILITY of the CCW loop header.SR 3.7.7.2This SR verifies that the two motor operated isolation valves to the RHRheat exchangers (738A and 738B) can be operated when required sincethe valves are normally maintained closed. The Frequency of thisSurveillance is specified in the Inservice Test Program and is consistent with the ASME Code (Ref. 2).REFERENCES

1. UFSAR, Section 9.2.2.2. ASME Code for Operation and Maintenance of Nuclear PowerPlants.IR.E. Ginna Nuclear Power PlantB 3.7.7-6Revision 58 SW SystemB 3.7.8C.1 and C.2If the SW pumps cannot be restored to OPERABLE status within theassociated Completion Time, the plant must be placed in a MODE inwhich the LCO does not apply. To achieve this status, the plant must beplaced in at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full powerconditions in an orderly manner and without challenging plant systems.D.1With three or more SW pumps or the loop header inoperable, the plant isin a condition outside of the accident analyses; therefore, LCO 3.0.3 mustbe entered immediately.

Required Action D.1 is modified by a Note requiring that the applicable Conditions and Required Actions of LCO 3.7.7, "CCW System,"

beentered for the component cooling water heat exchanger(s) madeinoperable by SW. This note is provided since the inoperable SW systemmay prevent the plant from reaching MODE 5 as required by LCO 3.0.3 ifboth CCW heat exchangers are rendered inoperable.

SURVEILLANCE SR 3.7.8.1REQUIREMENTS This SR verifies that adequate NPSH is available to operate the SWpumps and that the SW suction source temperature is within the limitsassumed by the accident analyses and the system design. The-24-heL*-

Frcgqucncy iz based ein epcroting experienee rclated te tronding cf theparometcr YariotieHS dluring the applieeble MO)DES.SR 3.7.8.2 3Verifying the correct alignment for manual, power operated, andautomatic valves in the SW flow path provides assurance that the properflow paths exist for SW operation.

This SR does not apply to valves thatare locked, sealed, or otherwise secured in position, since they areverified to be in the correct position prior to being locked, sealed, orsecured.

This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves. This SR does not require any testingor valve manipulation; rather, it involves verification, through a systemwalkdown, that those valves capable of being mispositioned are in thecorrect position.

R.E. Ginna Nuclear Power PlantB 3.7.8-7Revision 68 SW SystemB 3.7.8The 31 day is based ,n ..gin ..i judgment, is t. n.i.t.ntwith the procodural eeontrolS geyerningi vov oatien, and cnSUroscorroct -vaiye positions.

This SR is modified by a Note indicating that the isolation of the SW flowto individual components or systems may render those components inoperable, but does not affect the OPERABILITY of the SW System.SR 3.7.8.3This SR verifies that all SW loop header cross-tie valves are locked in thecorrect position.

This includes verification that manual valves 4623,4639, 4640, 4665, 4668B, 4669, 4756, and 4760 are locked open andthat manual valves 4610, 4611,4612, and 4779 are locked closed. Thediesel generator cross-tie valves (4665, 4760, 4669, and 4668B) may beindividually (one at a time) closed intermittently under administrative

controls, such as during surveillance

testing, as described in the LCOBases. The 31 day is based en _ngncring Judge'mnt, isc .nsistent with the proi.dural go.o i cd o, andcnsuros corroct valve eDositlens N S fERTý3SR 3.7.8.4This SR verifies proper automatic operation of the SW motor operatedisolation valves on an actual or simulated actuation signal (i.e., coincident safety injection and undervoltage signal).

SW is a normally operating system that cannot be fully actuated as part of normal testing.

ThisSurveillance is not required for valves that are locked, sealed, orotherwise secured in the required position under administrative controls.

The 24 mointh Froqueney is based en the nooed tc pecfewm thiTuhvnisvanee undpr thoe anditions that apply duiong a plant putago andthe portntial fae an unplanned tignsianthif inc l act of the ropsafetmd with the neartor at pswgn. OpSWating apinormall has shownthat thcs eompenents usually pass the Surillano when poormaestingd atthe 24 moenth Frog~uoncy.

Thercfero, the Frcguceney is aepocpal fr~m. aMroliability standpoint.

4SR 3.7.8.5 t-INE.RI3ýT This SR verifies proper automatic operation of the SW pumps on anactual or simulated actuation signal. This includes the actuation of theSW pumps following an undervoltage signal and following a coincident safety injection and undervoltage signal. SW is a normally operating system that cannot be fully actuated as part of normal testing duringnormal operation.

The 24 moenth F-Fequcncy is based en the need to-p.......

this .ur.illane under the e"nditions that apply during a planteutagc and the potential for anH unplanned transiont if the Sufveffillanee wc.. peio.....d with the roacter at power. GOpcating oexperiene hasshown that these eempencnts usually pass the SurveJllanee whenpeofefrmed at the 24 moenth Frogqueney.

Thcroforc, the Froequon-y iacooptablo fromH a roliability standpoint{.

R.E. Ginna Nuclear Power PlantB 3.7.8-8Revision 68 CREATSB 3.7.9a condition outside the accident analyses.

Therefore, LCO 3.0.3 must beentered immediately.

SURVEILLANCE SR 3.7.9.1REQUIREMENTS Standby systems should be checked periodically to ensure that theyfunction properly.

As the environment and normal operating conditions on this system are not too severe, testing each CREATS filtration trainoncc cver,' 31 days for _> 15 minutes provides an adequate check of thissystem. The 31 day Frcgueney is based en the Feliabilit' ef theequipment, and the twe train rcdundancy

.,.SR 3.7.9.2This SR verifies that the required CREATS testing is performed inaccordance with the Ventilation Filter Testing Program (VFTP). TheVFTP includes testing the performance of the HEPA filter, charcoaladsorber efficiency, flow rate, and the physical properties of the activated charcoal.

The required flowrate through each CREATS filtration train is6000 cubic feet per minute (+/-10%). Specific test Frequencies andadditional information are discussed in detail in the VFTP.The value of 1.5% methyl iodide penetration was chosen for thelaboratory test sample acceptance criteria

because, even though the newsystem contains 4-inch charcoal beds, the design face velocity is 61 fpm.Regulatory Guide 1.52, Revision 3 (Ref. 9), Table 1, provides testingcriteria assuming a 40 fpm face velocity.

The value of 1.5% wasinterpolated between the two values listed because of the higher facevelocity of Ginna's system. The face velocity is listed in the specification because it is a non standard number. Testing at 61 fpm or greater satifiesthe criteria.

SR 3.7.9.3This SR verifies that each CREATS train starts and operates and thateach CREATS automatic damper actuates on an actual or simulated actuation signal. The Fr.quen.y

.f 24 mon,.,th.

is based en lnduIt.;-

op....ting cxpcricncc.

SR 3.7.9.4This SR verifies the OPERABILITY of the CRE boundary by testing forunfiltered air inleakage past the CRE boundary and into the CRE. Thedetails of the testing are specified in the Control Room EnvelopeHabitability Program.The CRE is considered habitable when the radiological dose to CREoccupants calculated in the licensing basis analyses of DBAR.E. Ginna Nuclear Power PlantB 3.7.9-6Revision 51 ABVSB 3.7.10ACTIONSA._1When the ABVS is inoperable, action must be taken to place the plant ina condition in which the LCO does not apply. Action must be takenimmediately to suspend movement of irradiated fuel assemblies in theAuxiliary Building.

This does not preclude the movement of fuel to a safeposition.

Required Action A.1 is modified by a Note indicating that LCO 3.0.3 is notapplicable.

If moving irradiated fuel assemblies in the Auxiliary Buildingwhich have decayed < 60 days since being irradiated, the fuel movementis independent of reactor operations.

Therefore, inability to suspendmovement of irradiated fuel assemblies is not sufficient reason to requirea reactor shutdown.

SURVEILLANCE REQUIREMENTS SR 3.7.10.1This SR verifies the OPERABILITY of the ABVS. During fuel movementoperations, the ABVS is designed to maintain a slight negative pressurein the Auxiliary Building to prevent unfiltered LEAKAGE.

This SRensures that Auxiliary Building exhaust fan C, and either Auxiliary Building main exhaust fan A or B are in operation and that the ABVSinterlock mode switch is in the correct position.

The Frcqu.ncy ef 24heurS is based en enginccring judgcmcent and shewn te be aeccptable through cpefroting experienee.

."L--NSERT 3RTSR 3.7.10.2This SR verifies the integrity of the Auxiliary Building enclosure.

Theability of the Auxiliary Building to maintain negative pressure with respectto the uncontaminated outside environment must be periodically verifiedto ensure proper functioning of the ABVS. During fuel movementoperations, the ABVS is designed to maintain a slight negative pressurein the Auxiliary Building to prevent unfiltered leakage.

This SR ensuresthat a negative pressure is being maintained in the Auxiliary Building.

The Frcequency ef 24 heura i5 based Fnognoing judgement andshewn te be eeeeptable through eporain oxpriooSR 3.7.10.3This SR verifies that the required SFP Charcoal Adsorber System testingis performed in accordance with the Ventilation Filter Testing Program(VFTP). The SFP Charcoal Adsorber System filter tests are in generalaccordance with Regulatory Guide 1.52 (Ref. 5). The VFTP includesR.E. Ginna Nuclear Power PlantB 3.7.10-4Revision 62 SFP Water LevelB 3.7.11ACTIONS A. 1When the initial conditions assumed in the fuel handling accident analysiscannot be met, steps should be taken to preclude the accident fromoccurring.

When the SFP water level is lower than tle required level, themovement of irradiated fuel assemblies in the SFP is immediately suspended.

This action effectively precludes the occurrence of a fuelhandling accident.

This does not preclude movement of a fuel assemblyto a safe position (e.g., movement to an available rack position).

Required Action A.1 is modified by a Note indicating that LCO 3.0.3 doesnot apply since if moving irradiated fuel assemblies while in MODE 5 or 6,LCO 3.0.3 would not be applicable.

If moving irradiated fuel assemblies while in MODES 1, 2, 3, and 4, the fuel movement is independent ofreactor operations.

Therefore, inability to suspend movement ofirradiated fuel assemblies is not sufficient reason to require a reactorshutdown.

SURVEILLANCE SR 3.7.11.1REQUIREMENTS This SR verifies sufficient SFP water is available in the event of a fuelhandling accident.

The water level in the spent fuel pool must bechecked periodically during movement of irradiated fuel assemblies toensure the fuel handling accident assumptions are met. The 7 day-Froquenoy is appropriate bcoa use the vo Iume in the peel iS normaElly stable and the SFP is designed t3 provont drainago bclew 23 ft. Wator!eyol changes arc controllcd by plant procodurco and arce eeooptflblc based ong eporating cxpericnco.

Verification of SEP water level can be accomplished by several means.The top of the upper SEP pump suction line is 23 ft above the fuel storedin the pool. If there is ! 23 ft of water above the reactor vessel flange (asrequired by LCO 3.9.6), with equal pressure in the containment and theAuxiliary

Building, then at least 23 ft of water is available above the top ofthe active fuel in the storage racks.In addition to the physical design features, there are two SEP levelalarms (LAL 634) which are available to alert the operators of changingSEP level. A low level alarm will actuate when the SFP water level falls 4inches or more from the normal level while a high level alarm will actuatewhen the SEP water level rises 4 inches or more from the normal level.These alarms must receive a calibration consistent with industrypractices before they are to be used to meet this SR.R.E. Ginna Nuclear Power PlantB 3.7.11-3Revision 59 SFP Boron Concentration B 3.7.12APPLICABILITY This LCO applies whenever fuel assemblies are stored in the SFP toensure the SFP keff remains 0.95 at all times.ACTIONS A.1 and A.2When the concentration of boron in the SFP is less than required, immediate action must be taken to preclude the occurrence of anaccident or to mitigate the consequences of an accident in progress.

This is most efficiently achieved by immediately suspending themovement of fuel assemblies.

The initiation of actions to restoreconcentration of boron is simultaneous with suspending movement of fuelassemblies.

The Required Actions are modified by a Note indicating that LCO 3.0.3does not apply since if the LCO is not met while moving irradiated fuelassemblies in MODE 5 or 6, LCO 3.0.3 would not be applicable.

Ifmoving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuelmovement is independent of reactor operation.

Therefore, inability tosuspend movement of fuel assemblies is not sufficient reason to require areactor shutdown.

SURVEILLANCE SR 3.7.12.1REQUIREMENTS This SR verifies that the concentration of boron in the SFP is within thelimit. As long as this SR is met, the analyzed accidents are fullyaddressed.

The 7 day Fr .i. appr"p.iate sin" e the b"-^ n isr *edited with F, subcr.itical.

Alo,, the vlure andberen eeneentration in the peel is ncrrnally stable and all water levelchanges and bercn eeneentfatien changes are centrelled by plantpfI eedul f Ve IIIs .l4J~s~l~ll R.E. Ginna Nuclear Power PlantB 3.7.12-3Revision 20 Secondary Specific ActivityB 3.7.14ACTIONS A.1 and A.2DOSE EQUIVALENT 1-131 exceeding the allowable value in thesecondary

coolant, is an indication of a problem in the RCS andcontributes to increased post accident doses. If the secondary specificactivity is not within limits the plant must be placed in a MODE in whichthe LCO does not apply. To achieve this status, the plant must be placedin at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Theallowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full powerconditions in an orderly manner and without challenging plant systems.SURVEILLANCE SR 3.7.14.1REQUIREMENTS This SR verifies that the secondary specific activity is within the limits ofthe accident analysis.

A gamma isotopic analysis of the secondary

coolant, which determines DOSE EQUIVALENT 1-131, confirms thevalidity of the safety analysis assumptions as to the source terms in postaccident releases.

It also serves to identify and trend any unusualisotopic concentrations that might indicate changes in reactor coolantactivity or LEAKAGE.

The 31 de, Fr..qu.n.y

0. based en the d.tcoti:n incrcasing trcnd. of the 'eye' of DOSE EQUIVALENT 1'131, and all'wf8F apprcpriate actien to be talton to Fmaintain levels below the LCO) limnit..INSERT 31REFERENCES

1. 10 CFR 50.67.2. Design Analysis DA-NS-2002-007, Main Steam Line Break Offsiteand Control Room Doses.R.E. Ginna Nuclear Power PlantB 3.7.14-3Revision 42 AC Sources -MODES 1, 2, 3, and 4B 3.8.1SURVEILLANCE REQUIREMENTS The AC sources are designed to permit inspection and testing of allimportant areas and features, especially those that have a standbyfunction (Ref. 2). Periodic component tests are supplemented byextensive functional tests during refueling outages (under simulated accident conditions).

SR 3.8.1.1This SR ensures proper circuit continuity for the independent offsitepower source to each of the onsite 480 V safeguards buses andavailability of offsite AC electrical power. Checking breaker alignment and indicated power availability verifies that each breaker is in its correctposition to ensure that distribution buses and loads are connected to theirqualified power source. The Fr, uen.y 7- days ,- d,,,t, cincc'reaker p.siti.n is net likely to ch.nge withut thc pcratoros kn.wl.dgc and be"..use r.'m--- and indi"atficn

.f broakor Stu We arvailable in thc rocFR.-SR 3.8.1.2This SR verifies that each DG starts from standby conditions andachieves rated voltage and frequency.

This ensures the availability of theDG to mitigate DBAs and transients and to maintain the plant in a safe.shutdown condition.

The DG voltage control may be either in manual orautomatic during the performance of this SR. The F...u.n.y of 31 days-is adequate to provide nce of DG .PERABIIT.Y, whil. minmzingdegradation reulting from. testing.This SR is modified by two Notes. Note 1 indicates that performance ofSR 3.8.1.9 satisfies this SR since SR 3.8.1.9 is a complete test of the DG.The second Note states that all DG starts may be preceded by an engineprelube period and followed by a warmup period prior to loading.

Thisminimizes the wear on moving parts that do not get lubricated when theengine is not running.SR 3.8.1.3This SR verifies that the DGs are capable of synchronizing with the offsiteelectrical system and accepting loads greater than or equal to theequivalent of the maximum expected accident loads. A minimum runtime of 60 minutes is required to stabilize engine temperatures.

Amaximum run time of < 120 minutes minimizes the time that the DG isconnected to the offsite source.R.E. Ginna Nuclear Power PlantB 3.8.1-12Revision 74 AC Sources -MODES 1, 2, 3, and 4B 3.8.1Although no power factor requirements are established by this SR, theDG is normally operated at a power factor between 0.85 lagging and 0.95lagging.

The upper load band limit of < 2250 kW is the DG two-hourrating and is provided to avoid routine overloading of the DG which mayresult in more frequent inspections in accordance with vendorrecommendations in order to maintain DG OPERABILITY.

The lowerband limit of 2025 kW bounds the maximum expected load following aDBA, based on worst case loading during the injection phase of theaccident.

The diesel generator loading will be below the long-term ratingof 1950 kW within two hours.In addition to verifying the DG capability for synchronizing with the offsiteelectrical system and accepting loads, the DG ventilation system shouldalso be verified during this surveillance.

The Frcequeney of 31 days is adequate to proVido ac.Uranco of DGOPERABILITY, while minimizing degradetion rcsulting from tccting.JINSERT 3This SR is modified by four Notes. Note 1 indicates that diesel engineruns for this Surveillance may include gradual loading, as recommended by the manufacturer, so that mechanical stress and wear on the dieselengine are minimized.

Note 2 states that momentary transients outsidethe load band (e.g., due to changing bus loads), do not invalidate thistest. Similarly, momentary power factor transients above or below theadministrative limit do not invalidate the test. Note 3 indicates that thisSurveillance shall be conducted on only one DG at a time in order toavoid common cause failures that might result from offsite circuit or gridperturbations.

Note 4 stipulates a prerequisite requirement forperformance of this SR. A successful performance of SR 3.8.1.2 or SR3.8.1.9 must precede this surveillance to prevent unnecessary starts ofthe DGs.S 13-38. 1.4This SR provides verification that the level of fuel oil in each day tank is ator above the minimum level, including instrument uncertainty, at whichfuel oil is automatically added when the fuel oil transfer pump is in autoand the DG is operating.

This level ensures adequate fuel oil for aminimum of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of DG operation at 110% of full load. A level of 8.75inches, as read on the local sight glass, achieves these requirements.

TeFrcguency ef 31 days is adequato to ensurc that a sufficicnt supplyof fuel ofil is available, sinee low lcycl alarmnS arc praVided and opcratoru SR 3.8.1.5This SR demonstrates that each DG fuel oil transfer pump operates andtransfers fuel oil from its associated storage tank to its associated daytank. This is required to support continuous operation of the DGs. ThisR.E. Ginna Nuclear Power PlantB 3.8.1-13Revision 74 AC Sources -MODES 1, 2, 3, and 4B 3.8.1Surveillance provides assurance that the fuel oil transfer pump isOPERABLE, the fuel oil piping system is intact, the fuel delivery piping isnot obstructed, and the controls and control systems for automatic ormanual fuel transfer systems are OPERABLE.

The Frequency of 31 doys is ade quatc to provide assuranc of DGOPERABILITY, since the dcsign of thc fuel ill transfcr system is suchthat pumps .peratc automatically o must bc sta.ted manually in-ordor tomaintain an adequate vojume OT fuel om in the day tank flK9 cuinFoSR 3.8.1.6This SR involves the transfer of the 480 V safeguards bus power suprfrom the 50/50 mode to the 100/0 mode and 0/100 mode whichdemonstrates the OPERABILITY of the alternate circuit distribution network to power the required loads. Thc Fr.qucnoy of 24 mo.nths ibasen eig e dgment, taking into oonsidcration the plantly,.iti, nrquirca to pcncrm te .urvciiiance, and is intcnaed to meconsistent with cxpcct.d fuel ccylo lengths.

Operating cxporionc hasshown that thesc components usually pass the SR when porformoed at=iec -44 mnin rrcuucncv

.1erfcr, the.n rrU luency was cunliuonu to u4aeceptagic tram a rSR 3.8.1.7eiiaaiiity standpoint.

This SR verifies that each DG does not trip during and following a loadrejection of > 295 kW. Each DG is provided with an engine overspeed trip to prevent damage to the engine. Recovery from the transient caused by the loss of a large load could cause diesel engine overspeed, which, if excessive, might result in a trip of the engine. This SRdemonstrates the DG load response characteristics and capability toreject the largest single load on the buses supplied by the DG (i.e., asafety injection pump).In order to ensure that the DG is tested under load conditions that are asclose to design basis conditions as possible, testing must be performed using a power factor _< 0.9 lagging.

This power factor is chosen to berepresentative of the actual design basis inductive loading that the DGwould experience.

.-- =The Freauc nov ot24 moenthsinto consideration plantand is intended to be cceen^ itiens to p 'rform the INSERT, 3This SR is modified by two Notes. The first Note states that this L-Surveillance shall not be performed in MODE 1, 2, 3, or 4. The reason forthe Note is that during operation in these MODES, performance of thisSR could cause perturbations to the electrical distribution systems thatcould challenge continued steady state operation and, as a result, plant......R.E. Ginna Nuclear Power PlantB 3.8.1-14Revision 74 AC Sources -MODES 1, 2, 3, and 4B 3.8.1safety systems.

The second Note acknowledges that credit may betaken for unplanned events that satisfy this SR.SR 3.8.1.8This SR demonstrates that DG noncritical protective functions (e.g.,overcurrent, reverse power, local stop pushbutton) are bypassed on anactual or simulated S! actuation signal. The noncritical trips arebypassed during DBAs but still provide an alarm on an abnormal enginecondition.

This alarm provides the operator with sufficient time to reactappropriately.

The DG availability to mitigate the DBA is more criticalthan protecting the engine against minor problems that are notimmediately detrimental to emergency operation of the DG. The DGcritical protective functions (engine overspeed, low lube oil pressure, andstart failure (overcrank) relay) will be tested periodically per the stationperiodic maintenance program.The Frzquency vf 24 .....th. ia based en i"g judgment, takingiN to ecnsodortien plant cenditionB rcqUircd to pcrfcrm the Gufveillanoc, and is intended to be eensistcnt with cxpeotcd fuel cycle lcngh.Operating cxpcricncc has shown that these compencnts usually pass theSR when perftrmed at th24-moenth FrFequency.

T-herefore, thisFrcqucney is aeceptablc fromn a reliability standpoit This SR is modified by two Notes. The first Note states that thisSurveillance shall not be performed in MODE 1, 2, 3, or 4. The reason forthe Note is that performing the Surveillance would remove a required DGfrom service which is undesirable in these MODES. The second Noteacknowledges that credit may be taken for unplanned events that satisfythis SR.SR 3.8.1.9In the event of a DBA coincident with a loss of offsite power, the DGs arerequired to supply the necessary power to ESF systems so that the fuel,RCS, and containment design limits are not exceeded.

This SR demonstrates the DG operation during an actual or simulated loss of offsite power signal in conjunction with an actual or simulated SIlactuation signal. In lieu of actual demonstration of connection andloading of loads, testing that adequately shows the capability of the DGsystem to perform these functions is acceptable.

This testing mayinclude any series of sequential, overlapping, or total steps so that theentire connection and loading sequence is verified.

Since it is not possible to operate all sequenced motors at their DBAloadings, a transient simulation program is used to demonstrate acceptable DG governor and voltage regulator operation.

To successfully validate the testing data with the transient simulation

program, the largestloads (with respect to both kW and current) must be sequenced on theR.E. Ginna Nuclear Power PlantB 3.8.1-15Revision 74 AC Sources -MODES 1, 2, 3, and 4B 3.8.1DG during performance of this test. This includes two SI pumps, a CSIINSERT 3J- ,, \and RHR pump, and safety-related motor control centers, as a minimum.The Frcquency of 24 moentha is based on cnginccrin~g judgement, takinginte considcratien plant conditiona9 rcguircd to performn the Surveillancc, and us intendled to be ecnsiotent with cxpccted fuel cyele lcnghaThis SR is modified by three Notes. Note 1 states that all DG starts maybe preceded by an engine prelube period which is intended to minimizewear and tear on the DGs during testing.

For the purpose of this testing,the DGs must be started from standby conditions, that is, with the enginelube oil continuously circulated and temperature maintained consistent with manufacturer recommendations for the DGs. Note 2 states that thisSurveillance shall not be performed in MODE 1, 2, 3, or 4 sinceperforming the Surveillance during these MODES would remove arequired offsite circuit from service, cause perturbations to the electrical distribution

systems, and challenge safety systems.

Note 3acknowledges that credit may be taken for unplanned events that satisfythis SR.REFERENCES

1. UFSAR, Chapter 8.2. Atomic Industrial Forum (AIF) GDC 39, Issued for comment July10,1967.3. UFSAR, Section 9.4.9.5.4. UFSAR, Chapter 6.5. UFSAR, Chapter 15.6. 10 CFR 50, Appendix A, GDC 17.7. "American National

Standard, Nuclear Safety Criteria for the Designof Stationary Pressurized Water Reactor Plants,"

N 18.2-1973.

8. Generic Letter 84-15, "Proposed Staff Actions to Improve andMaintain Diesel Generator Reliability,"

July 2, 1984.9. UFSAR Section 3.11R.E. Ginna Nuclear Power PlantB 3.8.1-16Revision 74 Diesel Fuel OilB 3.8.3time to correct high particulate levels prior to reaching the limit ofacceptability.

Poor sample practices (bottom sampling),

contaminated sampling equipment, or errors in laboratory analysis can produce failuresthat do not follow a trend. Since the presence of particulates does notmean failure of the fuel oil to burn properly in the diesel engine, andparticulate concentration is unlikely to change significantly betweenSurveillance Frequency intervals, and proper engine performance hasbeen recently demonstrated (within 31 days), it is prudent to allow a briefperiod prior to declaring the associated DG inoperable.

The 7 dayCompletion Time allows for further evaluation, resampling and re-analysis of the DG fuel oil.C..1With the new fuel oil properties defined in SR 3.8.3.2 not within requiredlimits, a period of 30 days is allowed for restoring the stored fuel oilproperties.

This period provides sufficient time to test the stored fuel oilto determine that the new fuel oil, when mixed with previously stored fueloil, remains acceptable, or to restore the stored fuel oil properties.

Thisrestoration may involve feed and bleed procedures, filtering, orcombinations of these procedures.

Even if a DG start and load wasrequired during this time interval and the fuel oil properties were outsidelimits, there is a high likelihood that the DG would still be capable ofperforming its intended function.

D..1With a Required Action and associated Completion Time not met, or oneor more DG's fuel oil notwithin limits for reasons other than addressed byConditions A, B, or C (e.g., cloud point temperature reached),

theassociated DG may be incapable of performing its intended function andmust be immediately declared inoperable.

SURVEILLANCE SR 3.8.3.1REQUIREMENTS This SR verifies an onsite supply of >_ 5000 gal of fuel oil is available foreach required DG. This ensures that there is an adequate inventory offuel oil in the storage tanks to support each DG's operation for 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />swhile providing maximum post-LOCA loads. The 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> period issufficient time to place the plait in a safe shutdown condition and to bringin replenishment fuel from an offsite location.

The Frcqueney ef 31 days 09 adequate te ensuro that a sufficicnt supplyef fuel eel is available, s~incc indicaticnS arc available te einsur thatepefrltefr Would be aWBro ef BAnY larg uses of fuel eil durin9g thiS pcriod.R.E. Ginna Nuclear Power PlantB 3.8.3-3Revision 48 DC Sources -MODES 1, 2, 3, and 4B 3.8.4SURVEILLANCE SR 3.8.4.1REQUIREMENTS Verifying battery terminal voltage while on float charge for the batteries helps to ensure the effectiveness of the charging system and tle ability ofthe batteries to perform their intended function.

Float charge is thecondition in which the charger is supplying the continuous chargerequired to overcome the internal losses of a battery (or battery cell) andmaintain the battery (or a battery cell) in a fully charged state. Thevoltage requirements are based on the nominal design voltage of thebattery and are consistent with the initial voltages assumed in the batterysizing calculations.

The elevated equalize charge capability of thebattery chargers is not an OPERABILITY requirement of the batterychargers and is not to be in service during the surveillance.

The voltagedrop when changing from the equalize conditions to the normal floatconditions occurs relatively quickly.

The 7 day ny is ..nsistwntwith ma.enufacture.r r nmm-ndaticns and IEEEr 450 (Rcf. 8)- ,SR 3.8.4.2 3TThis SR verifies that the capacity of each battery is adequate to supplyand maintain in OPERABLE status, the required emergency loads for thedesign duty cycle when subjected to a battery service test. A batteryservice test is a special test of battery capability, as found, to satisfy thedesign requirements (battery duty cycle) of the DC electrical powersystem. The discharge rate and test length corresponds to the designduty cycle requirements specified in Reference 2.The Sur~oillanee Frogueney of 24 months is eensistcnt with the-rocommendatiens of Rogulater; Cuide 1.32 (Ref. 9) and Rcgulater;y Guide 1. 129 (Ref. 10), whieh state that the battery zorwiee test should bepcrfermcd duarig rcfuelnlg oporaltiens or at seme other outagc, withintcrwais

between, tests Het to exeeedd 24 moinths.

t ISETýThis SR is modified by two Notes. Note 1 states that SR 3.8.4.3 may beperformed in lieu of SR 3.8.4.2.

This substitution is acceptable becauseSR 3.8.4.3 represents a more severe test of battery capacity than doesSR 3.8.4.2.

Note 2 states that this surveillance shall not be performed inMODE 1, 2, 3, or 4 because performing the Surveillance would perturbthe electrical distribution system and challenge safety systems.SR 3.8.4.3This Surveillance verifies that each battery capacity is ! 80% of themanufacturer's rating when subjected to a performance discharge test. Abattery performance test is a test of constant current capacity of a battery,normally done in the as found condition, after having been in service, todetect any change in the capacity as determined by specified acceptance R.E. Ginna Nuclear Power PlantB 3.8.4-6Revision 41 DC Sources -MODES 1, 2, 3, and 4B 3.8.4criteria.

The test is intended to determine overall battery degradation dueto age and usage.A battery should be replaced if its capacity is below 80% of themanufacturer rating. A capacity of 80% shows that the battery rate ofdeterioration is increasing, even if there is ample capacity to meet theload requirements.

,-,, I.NERT 1The Frequency for this SR is 6A menths when the battery is < 85% of itsexpected life with no degradation and 12 months if the battery showsdegradation or has reached 85% of its expected life with a capacity< 100% of the manufacturer's rating. When the battery has reached 85%of its expected life with capacity

_ 100% of the manufacturer's rating, theFrequency becomes 24 months. Battery degradation is indicated whenthe battery capacity drops by more than 10% relative to its capacity onthe previous performance test or when it is > 10% below themanufacturer rating. These Frequencies are considered acceptable based on the testing being performed in a conservative manner relativeto the battery life and degradation.

This ensures that battery capacity isadequately monitored and that the battery remains capable of performing its intended function.

This SR is modified by a Note stating that this SR shall not be performed in MODE 1, 2, 3, or 4. The reason for the Note is that during operation inthese MODES, performance of this SR could cause perturbations to theelectrical distribution system and challenge safety systems.R.E. Ginna Nuclear Power PlantB 3.8.4-7Revision 41 DC Sources -MODES 1, 2, 3, and 4B 3.8.41. Atomic Industrial Forum (AIF) GDC 39, Issued for comment JulyREFERENCES

1. Atomic Industrial Forum (AIF) GDC 39, Issued for comment July10,1967.2. UFSAR, Section 8.3.2.3. UFSAR, Section 9.4.9.3.4. UFSAR, Chapter 6.5. UFSAR, Chapter 15.6. UFSAR, Section 8.3.1.7. 10 CFR 50, Appendix A, GDC 17.8. IEEE-450-1980.

__,,1/eleeleted

9. R.gul.t'r-y Guide 1.32, Fcbruor; 7...'10. R.gul.ety Cuide 1..129, Dc......r 1974.R.E. Ginna Nuclear Power PlantB 3.8.4-8Revision 41 Battery Cell Parameters B 3.8.6SURVEILLANCE SR 3.8.6.1REQUIREMENTS This SR verifies that the electrolyte level of each connected battery cell isabove the top of the plates and not overflowing.

This is consistent withIEEE-450 (Ref. 4) and ensures that the plates suffer no physical damageand maintain adequate electron transfer capability.

The F,-,qucn.y of 31days is eensistcnt with IEEE 450.SR 3.8.6.2 3This SR verifies that the float voltage of each connected battery cell is> 2.07 V. This limit is based on IEEE-450 (Ref. 4) which slates that a cellvoltage of 2.07 V or below, under float conditions and not caused byelevated temperature of the cell, indicates internal cell problems and mayrequire cell replacement.

of 31 days is else ..nsistcntwith IEEE 450.SR 3.8.6.3 -iINSERT 3TThis SR verifies the specific gravity of the designated pilot cell in eachbattery is _> 1.195. This value is based on manufacturer recommendations.

According to IEEE-450 (Ref. 4), the specific gravityreadings are based on a temperature of 77°F (251C). The specific gravityreadings are corrected for actual electrolyte temperature.

For each 30F(1.671C) above 770F (250C), 1 point (0.001) is added to the reading; 1point is subtracted for each 30F below 77°F. The specific gravity of theelectrolyte in a cell increases with a loss of water due to electrolysis orevaporation.

Because of specific gravity gradients that are produced during therecharging

process, delays of several days may occur while waiting forthe specific gravity to stabilize.

A stabilized charger current is anacceptable alternative to specific gravity measurement for determining the state of charge. This phenomenon is further discussed in IEEE-450.

The Frogueney ef 31 days as eensistent with IEEE 450.SR 3.8.6.4 INERT 3This SR verifies the average electrolyte temperature of the designated pilot cell in each battery is > 550F. This temperature limit is an initialassumption of the battery capacity calculations.

The Fr.qu.n.y

.f 31days is eensistent with IEEE 450 (Ref-. 4.R.E. Ginna Nuclear Power PlantB 3.8.6-3Revision 40 Battery Cell Parameters B 3.8.6SR 3.8.6.5This SR verifies that the average temperature of every fifth cell of eachbattery is > 550F. This is consistent with the recommendations of IEEE-450 (Ref. 4). Lower than normal temperatures act to inhibit or reducebattery capacity.

This SR ensures that the operating temperatures remain within an acceptable operating range. The FFr.u.ncy

.f 92 days-0 s ccnsistent with IEEE 450.SR 3.8.6.6 3This SR verifies the specific gravity of each connected cell is not morethan 0.020 below average of all connected cells and that the average ofall connected cells is > 1.195. This value is based on manufacturer recommendations and IEEE-450 (Ref. 4) which ensures that the effect ofa highly charged or new cell does not mask overall degradation of thebattery.

The temperature correction for specific gravity readings is thesame as that discussed for SR 3.8.6.3.

The Fr..u.n.y ef 92 del. iCccnsste.nt with I-EEE 450-REFERENCES

1. UFSAR, Section 3.8.2.2. UFSAR, Chapter 6.3. UFSAR, Chapter 15.4. IEEE-450-1980.

R.E. Ginna Nuclear Power PlantB 3.8.6-4Revision 40 AC Instrument Bus Sources -MODES 1, 2, 3, and 4B 3.8.7SURVEILLANCE REQUIREMENTS SR 3.8.7.1This SR verifies correct static switch alignment to Instrument Bus A andC. This verifies that the inverters are functioning properly and ACInstrument Bus Aand C are energized from their respective inverter.

Theverification ensures that the required power is available for theinstrumentation of the RPS and ESF connected to the AC instrument buses. The F..qucn.y

f. 7 days takes int' ae-e"nt the ,-dundant capability ef the invclter.

and .thc1 indi tinsI ayailabic in the I rcamn that alert the epeffltfr te inverter mfalfuncticems.~

SR 3.8.7.2This SR verifies the correct Class 1 E CVT alignment to Instrument Bus B.This verifies that the Class 1 E CVT is functioning properly and supplying power to AC Instrument Bus B. The verification ensures that the requredpower is available for the instrumentation of the RPS and ESF connected to the AC instrument bus. The F.qu.n.y

.f 7- days takes int, a.. .untthe rdundint Umnt buses and ther in theccntrcl rccm that alert the epefrater to the Glass 1 E CYT mc~lfuctc REFERENCES

1. UFSAR, Chapter 8.3.2.2. UFSAR, Chapter 6.3. UFSAR, Chapter 15.4. UFSAR, Section 8.3.1.5. 10 CFR 50, Appendix A, GDC 17.R.E. Ginna Nuclear Power PlantB 3.8.7-6Revision 41 AC Instrument Bus Sources -MODES 5 and 6B 3.8.8The Completion Time of immediately is consistent with the required timesfor actions requiring prompt attention.

The restoration of the required ACinstrument bus power source should be completed as quickly as possiblein order to minimize the time the plant safety systems may be withoutpower or powered from an alternate power source.SURVEILLANCE SR 3.8.8.1REQUIREMENTS This SR verifies correct static switch alignment to the required ACinstrument buses. This SR verifies that the inverter is functioning properly and the AC instrument bus is energized from the inverter.

Theverification ensures that the required power is available for theinstrumentation connected to the AC instrument bus. The F.rqu.n.y cf 7days takes inte aeeeunt the rodundant capability of the finycrtcr and etheri ndfieaticns available in the control rocm that alert the epefrater to inverterSR 3.8.8.2 3This SR verifies the correct Class 1E CVT alignment when Instrument Bus B is required.

This verifies that the Class 1 E CVT is functioning properly and supplying power to AC Instrument Bus B. The verification ensures that the required power is available for the instrumentation of theRPS and ESF connected to the AC instrument bus. The Fr.qu.ncy cf 7days takes into aeecwunt the rcdundant inotrument buses and ethcrindieations available in the contrOl room that alert the epefrater to theClass! E C r- l Falfuneticc,-A:^

REFERENCES

1. None.R.E. Ginna Nuclear Power PlantB 3.8.8-5Revision 61 Distribution Systems -MODES 1, 2, 3, and 4B 3.8.9D.1 and D.2If the inoperable distribution subsystem cannot be restored toOPERABLE status within the required Completion Time, the plant mustbe brought to a MODE in which the LCO does not apply. To achieve thisstatus, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> andto MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times arereasonable, based on operating experience, to reach the required plantconditions from full power conditions in an orderly manner and withoutchallenging plant systems.E._1With two trains with inoperable electrical power distribution subsystems, the potential for a loss of safety function is greater.

If a loss of safetyfunction exists, no additional time is justified for continued operation andLCO 3.0.3 must be entered.

This Condition may be entered with the lossof two trains of the same electrical power distribution subsystem, or withloss of Train A of one electrical power distribution subsystem coincident with the loss of Train B of a second electrical power distribution subsystem such that a loss of safety function exists.SURVEILLANCE SR 3.8.9.1REQUIREMENTS This SR verifies that the electrical power trains are functioning

properly, with all required power source circuit breakers closed, tie-breakers open,and the buses energized from their allowable power sources.

Requiredvoltage for the AC electrical power distribution subsystem is _ 420 VAC;for the DC electrical power distribution subsystem,

_ 108.6 VDC and< 140 VDC; and for AC instrument bus electrical power distribution subsystem, between 113 VAC and 123 VAC at the instrument buses.Required voltage for the instrument distribution panels is between 110VAC and 123 VAC. Required voltage for inverter MQ-483 is between 107volts and 129.8 volts. The loss of inverter MQ-483 is addressed in LCO3.3.2, "Engineered Safety Feature Actuation System (ESFAS)Instrumentation" and LCO 3.3.3, "Post Accident Monitoring (PAM)Instrumentation" for the affected individual containment wide rangepressure and steam generator B pressure instrumentation (PT-950 andPT-479).

The verification of proper voltage availability on the busesensures that the required power is readily available for motive as well ascontrol functions for critical system loads connected to these buses. T-he-F...u.n.y of 7 days int" account the "edundent

.apability

' f theAC, D, n.. ACG nO.trumnt bus ,lctr..al pewe. di.t.ibu.tien

-ub-systcm, and ether in~dicatiens available in the control reoom that alcrt the eperaterto stubsystcm malfunctions.

R.E. Ginna Nuclear Power PlantB 3.8.9-9Revision 68 Distribution Systems -MODES 5 and 6B 3.8.10Therefore, Required Action A.2.5 requires declaring RHR inoperable, which results in taking the appropriate RHR actions.The Completion Time of immediately is consistent with the required timesfor actions requiring prompt attention.

The restoration of the requireddistribution subsystems should be completed as quickly as possible inorder to minimize the time the plant safety systems may be withoutpower.SURVEILLANCE SR 3.8.10.1REQUIREMENTS This Surveillance verifies that the electrical power distribution trains arefunctioning

properly, with all the required power source circuit breakersclosed, required tie-breakers open, and the required buses energized from their allowable power sources.

Required voltage for the AC powerdistribution electrical subsystem is >_ 420 VAC, for the DC powerdistribution electrical subsystem

> 108.6 VDC and _< 140 VDC, and for ACinstrument bus power distribution electrical subsystem is between 113VAC and 123 VAC at the instrument buses. Required voltage for theinstrument distribution panels is between 110 VAC and 123 VAC.Required voltage for inverter MQ-483 is between 107 volts and 129.8volts. The verification of proper voltage availability on the buses ensuresthat the required power is readily available for motive as well as controlfunctions for critical system loads connected to these buses. Trhe-Frcgqucrny ef 7 days takes into account the eapability of the AC, DCG, andAC instrumonet bus eleetrical pewer diStributien subsystems, and etherinmdications available in the control roomf that alcrt the epcratert subsystem mtalfunctions-.

REFERENCES 1 .None.R.E. Ginna Nuclear Power PlantB 3.8.10-6Revision 61 Boron Concentration B 3.9.1There are no safety analysis assumptions of boration flow rate andconcentration that must be satisfied.

The only requirement is to restorethe boron concentration to its required value as soon as possible.

Inorder to raise the boron concentration as soon as possible, the operatorshould begin boration with the best source available for plant conditions.

Once action has been initiated, it must be continued until the boronconcentration is restored.

The restoration time deperds on the amount ofboron that must be injected to reach the required concentration.

SURVEILLANCE REQUIREMENTS SR 3.9.1.1This SR ensures the coolant boron concentration of the refueling canal,the refueling cavity, and the portions of the RCS that are hydraulically

coupled, is within the COLR limits. The boron concentration of thecoolant is determined by chemical analysis.

The sample should berepresentative of the portions of the RCS, the refueling canal, and therefueling cavity that are hydraulically coupled with the reactor core.A FFeueieyefene vecr;7 heLurS iS Faseel e metunt ef timneYerify the bercn eenecntratien ef the rcprcsentative sample(s).

TheFrcquency is based on epcroiting cxpcrienee, whieh has shewn :72 hourste be I l d. q ued.lI y V J -~i.'q

,llilq.REFERENCES

1. Atomic Industrial Forum (AIF) GDC 27, Issued for comment July10, 1967.2. UFSAR, Section 15.4.4.2.

3. NUREG-0800, Section 15.4.6.R.E. Ginna Nuclear Power PlantB 3.9.1-4Revision 61 Nuclear Instrumentation B 3.9.2The Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is sufficient to obtain and analyzecoolant samples for boron concentration.

The Frequency of once per 12hours ensures unplanned changes in boron concentration would beidentified.

The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is reasonable, considering the lowprobability of a change in core reactivity during this time period.C.1, C.2, and C.3With no audible count rate available, only visual indication is available and prompt and definite indication of a boron dilution event has been lcst.Therefore, CORE ALTERATIONS and positive reactivity additions mustbe suspended immediately.

Performance of Required Actions C.1 andC.2 shall not preclude completion of movement of a component to a safeposition (i.e., other than a normal cooldown of the coolant volume for thepurpose of system temperature control within established procedures).

Since CORE ALTERATIONS and positive reactivity additions are not tobe made, the core reactivity condition is stabilized until the audible countrate capability is restored.

This stabilized condition is determined byperforming SR 3.9.1.1 to ensure that the required boron concentration exists.The Completion time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is sufficient to obtain and analyze coolantsamples for boron concentration.

The Frequency of once per 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />sensures unplanned changes in boron concentration would be identified.

The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is reasonable, considering the low probability of achange in core reactivity during this time period.SURVEILLANCE SR 3.9.2.1REQUIREMENTS This SR is the performance of a CHANNEL CHECK, which is acomparison of the parameter indicated on one monitor to a similarparameter on another monitor.

It is based on the assumption that the twoindication channels should be consistent with core conditions.

Changesin fuel loading and core geometry can result in significant differences between source range monitors, but each monitor should be consistent with its local conditions.

The inoperability of one source range neutron flux channel preventsperformance of a CHANNEL CHECK for the operable channel.

However,the Required Actions for the inoperable channel requires suspension ofCORE ALTERATIONS and positive reactivity addition such that theCHANNEL CHECK of the operable channel can consist of ensuringconsistency with known core conditions.

R.E. Ginna Nuclear Power PlantB 3.9.2-3Revision 61

.Nuclear Instrumentation B 3.9.2The Frcgueney 3f 12 heurs is eensistent with the CHANNEL G CHECKIIFrcuenly spc.ificd simIlrly for thc Ioll iilFthelsate ii C -3.3.4,'Reaeter TrFip System (RTSG) Instrumentatien."

SR 3.9.2.2This SR is the performance of a CHANNEL CALIBRATION every24-tenlnhs.

This SR is modified by a Note stating that neutron detectors areexcluded from the CHANNEL CALIBRATION.

The CHANNELCALIBRATION for the source range neutron flux monitors consists ofobtaining the detector plateau or preamp discriminator curves, evaluating those curves, and comparing the curves to baseline data. The 24 m-nthFroguenL y is based en the need te perform this _urv1illanee undrr theeenditiens that apply durfing a plant outage. O~perating experienee hasshown these eempenents usually pass the Survoillanco when perfefrmod Rim1 TFRl -.46 mFRRR3 I-FequefiyREFERENCES

1. UFSAR, Section 7.7.3.2.2. Atomic Industrial Forum (AIF) GDC 13 and 19, Issued for CommentJuly 10, 1967.RE. Ginna Nuclear Power PlantB 3.9.2-4Revision 61 Containment Penetrations B 3.9.3ACTIONSA.1 and A.2If the containment equipment hatch (or its closure plate or roll up doorand associated enclosure building),

air lock doors, or any containment penetration that provides direct access fromthe containment atmosphere to the outside atmosphere is not in the required status, including theContainment Ventilation Isolation System not capable of automatic actuation when the purge and exhaust valves are open, the plant must beplaced in a condition where the isolation function is not needed. This isaccomplished by immediately suspending CORE ALTERATIONS andmovement of irradiated fuel assemblies within containment.

Performance of these actions shall not preclude completion of movement of acomponent to a safe position.

SURVEILLANCE REQUIREMENTS SR 3.9.3.1This SR demonstrates that each of the containment penetrations are inthe required status. The Surveillance on the open purge and exhaustvalves will demonstrate that the valves are not blocked or otherwise prevented from closing (e.g., solenoid unable to vent).The is ooe.Formd eve.y 7 days during CORE.........

L OAJAL 1LLjAI of .neveffcnt e4 odiatcd assemblies wine-ntainent.

The Surmil lan;e ."craI is selected t3 be eemmensurtc with the nirmal durati n ,f fi.fl tc fuel handling atic;s. Assuch, this Surveillanee cn9Surcs that a postulated fuel handling eeeidentthat rolcases fission product radioactivity within the eontainmcnt will noetI ,sUilt in a e, f fissiu n produ.tSR 3.9.3.2.... iradlieaotlvity to trec cnVIronment.

JINSERT3This SR demonstrates that each containment purge and exhaust valveactuates to its isolation position on manual initiation or on an actual orsimulated high radiation signal. The 24 month Fro......y mainta::ins unsistny with othnr snim ..ilar .inSt ....ntati. and v .lv. t .stigroguiromolnts.

In LCO) 3.3.5, the Containmcnt Vcntlation Isolatioinstrumentatien roguirca a CHANNEL CHECK cvcr; 24 heurs an~d a COT:ever; 92 days to cnsurc the chanrncl O)PERABILITY during r-efucling

ý ý--! Mvý 'A ý 4ký^ At-r! WArlk K! ! f'Ir' =0'r -AI II,... .ll Jl *~VSl 3. I.,V * *'J ,L Ic ll% v,J V%# l lV.l k y .l l l l.tl lCHANNE6 is pIFrfeord.

These s will ensUthat the valves arc capablc of closing after a postulated fuel handlingeeewdent to limfit 8 rolcase of fission product radioactivity fromg the:'aT-pR.E. Ginna Nuclear Power PlantB 3.9.3-4Revision 53 RHR and Coolant Circulation

-Water Level 2! 23 FtB 3.9.4A.4If RHR loop requirements are not met, all containment penetrations providing direct access from the containment atmosphere to the outsideatmosphere must be closed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. With the RHR looprequirements not met, the potential exists for the coolant to boil andrelease radioactive gas to the containment atmosphere.

Closingcontainment penetrations that are open to the outside atmosphere ensures dose limits are not exceeded.

The Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is reasonable, based on the lowprobability of the coolant boiling in that time.SURVEILLANCE SR 3.9.4.1REQUIREMENTS This SR requires verification e;e'y 12 heotr that one RHR loop is inoperation and circulating reactor coolant.

Verification includes flow rate,temperature, or pump status monitoring, which help ensure that forcedflow is providig decay heat removal capability and mixing of the boratedcoolant to prevent thermal and boron stratification in the core. The-Frcqueney ef 12 heUFS 69 suffienent eensidering ethcr indmeatiens and,-8FH ..........

te the epf.e ....

"e "ete ,ee te'^ Fne,4:^

• H e, pREFERENCES

1. UFSAR, Section 5.4.5.2. UFSAR, Section 15.4.4.2.

R.E. Ginna Nuclear Power PlantB 3.9.4-4Revision 61 RHR and Coolant Circulation

-Water Level < 23 FtB 3.9.5SURVEILLANCE SR 3.9.5.1REQUIREMENTS This SR requires verification evety 12 hlret, that one RHR loop is inoperation and circulating reactor coolant.

Verification includes flow rate,temperature, or pump status monitoring, which help ensure that forcedflow is providing decay heat removal capability and mixing of the boratedcoolant to prevent thermal and boron stratification in the core. T-he-Frcequeney of 12 heurs is suffricgnt considcring ether indicatiens andalarmS available tc the eporatr in the ro;m tc m.nite- RHR pe~fefffle~ee SR 3.9.5.2 IVerification that a second RHR pump is OPERABLE ensures that anadditional pump can be placed in operation, if needed, to maintain decayheat removal and reactor coolant circulation.

Verification is performed byverifying proper breaker alignment and power available to the standbypump. The F...qucn.y of 7 days is ..nsidorcd r.as.nabl.

in Yi.w .fether adFmignstrativc controls available and has been shown tobooeptable by operating oxporion~oo.

REFERENCES

1. UFSAR, Section 5.4.5.2. UFSAR, Section 15.4.4.R.E. Ginna Nuclear Power PlantB 3.9.5-4Revision 61 Refueling Cavity Water LevelB 3.9.6LCO A minimum refueling cavity water level of 23 ft above the reactor vesselflange is required to ensure the radiological consequences of apostulated fuel handling accident inside containment are withinacceptable limits and preserves the assumptions of the fuel handlingaccident analysis (Ref. 1). As such, it is the minimum required levelduring movement of fuel assemblies wthin containment.

Maintaining thisminimum water level in the refueling cavity also ensures that > 23 ft ofwater is available in the spent fuel pool during fuel movement assumingthat containment and Auxiliary Building atmospheric pressures are equal.APPLICABILITY This LCO is applicable when moving irradiated fuel assemblies withincontainment.

This LCO is also applicable during CORE ALTERATIONS, except during latching and unlatching of control rod drive shafts. TheLCO ensures a sufficient level of water is present in therefueling cavity tominimize the radiological consequences of a fuel handling accident incontainment.

Requirements for fuel handling accidents in the spent fuelpool are covered by LCO 3.7.11, "Spent Fuel Pool (SFP) Water Level."ACTIONS A.1 and A.2When the initial condition assumed in the fuel handling accident cannotbe met, steps should be taken to preclude the accident from occurring.

With a water level of < 23 ft above the top of the reactor vessel flange, alloperations involving CORE ALTERATIONS or movement of irradiated fuel assemblies within the containment shall be suspended immediately to ensure that a fuel handling accident cannot occur.The suspension of CORE ALTERATIONS and fuel movement shall notpreclude completion of movement of a component to a safe position.

SURVEILLANCE SR 3.9.6.1REQUIREMENTS Verification of a minimum refueling cavity water level of 23 ft above thetop of the reactor vessel flange ensures that the design basis for theanalysis of the postulated fuel handling accident during refueling operations is met. VAter at the required level above the top of the reactorvessel flange limits the consequences of damaged fuel rods that arepostulated to result from a fuel handling accident inside containment (Ref. 1).R.E. Ginna Nuclear Power PlantB 3.9.6-2Revision 59 Refueling Cavity Water LevelB 3.9.6The F-rc.ucn.y

.f 24 hurs is bascd cn coing judgment and .aeemsidered adequate in view ef the large Yealuin ef water and the neFrrnalp....du.al

...ntr'sl 'f valve p..ition:,

which m9o" :ignifi-ant unplan'..d level ehanges unlikely.

REFERENCES

1. UFSAR, Section 15.7.3.2. 10 CFR 50.67.3. Regulatory Guide 1.183.R.E. Ginna Nuclear Power PlantB 3.9.6-3Revision 59 ATTACHMENT 5License Amendment RequestR. E. Ginna Nuclear Power PlantDocket No. 50-244Application 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)TSTF-425 (NUREG-1431) vs. Ginna Cross-Reference LAR -Adoption of TSTF-425, Revision 3Docket No. 50-244Attachment 5Page 1 of 10TSTF-425 (NUREG-1431) vs. Ginna Cross-Reference Technical Specification Section Title/Surveillance Description*

TSTF-425 GINNADefinitions 1.1 1.1Staggered Test TestingShutdown margin (SDM) 3.1.1 3.1.1Verify SDM within limits 3.1.1.1 3.1.1.1Core Reactivity 3.1.2 3.1.2Verify core reactivity within predicted value 3.1.2.1 3.1.2.2Rod Group Alignment Limits 3.1.4 3.1.4Verify individual rod position within alignment 3.1.4.1 3.1.4.1Verify rod freedom of movement (trippability) 3.1.4.2 3.1.4.3Verify individual rod position within alignment with rod position 3.1.4.2monitor inoperable Shutdown Bank Insertion Limits 3.1.5 3.1.5Verify shutdown bank within insertion limit per COLR 3.1.5.1 3.1.5.1Control Bank Insertion Limit 3.1.6 3.1.6Verify control bank within insertion limit per COLR 3.1.6.2 3.1.6.2Verify sequence and overlap limits per COLR 3.1.6.3 3.1.6.4Verify control bank within insertion limit per COLR when insertion 3.1.6.3limit monitor inoperable Physics Tests (Exceptions Mode 2) 3.1.8 3.1.8Verify RCS lowest loop temperature 3.1.8.2 3.1.8.2Verify thermal power 3.1.8.3 3.1.8.3Verify SDM within COLR 3.1.8.4 3.1.8.4Fq(Z) 3.2.1Verify measured Fq(Z) 3.2.1.1Verify measured FWxV <Fx' 3.2.1.2F,(Z) (RAOC-W(Z)

Methodology 3.2.1B 3.2.1Verify F c%(Z) is within limit 3.2.1.1 3.2.1.1Verify F Wq(Z) is within limit 3.2.1.2 3.2.1.2Fq(Z) (CAOC-W(Z)

Methodology 3.2.1CVerify F cq(Z) is within limit 3.2.1.1Verify F Wq(Z) is within limit 3.2.1.2Nuclear Enthalpy Rise Hot Channel Factor (FN dh) 3.2.2 3.2.2Verify Fdh within limits per COLR 3.2.2.1 3.2.2.1Verify Fdh within limits per COLR (only when one power range 3.2.2.2inoperable)

AFD 3.2.3NBVerify AFD is within limits 3.2.3.1Update target flux difference 3.2.3.2Determine by measurement target flux difference 3.2.3.3AFD(RAOC Methodology) 3.2.3B 3.2.3Verify AFD is within limits 3.2.3.1 3.2.3.1QPTR 3.2.4 3.2.4Verify QPTR within limits by calculation 3.2.4.1 3.2.4.1Verify QPTR within limits by incore detectors 3.2.4.2 3.2.4.2 LAR -Adoption of TSTF-425, Revision 3Docket No. 50-244Attachment 5Page 2 of 10Technical Specification Section Title/Surveillance Description*

TSTF-425 GINNARTS Instrumentation 3.3.1 3.3.1Channel Check 3.3.1.1 3.3.1.1Compare calorimetric to power range channel output 3.3.1.2 3.3.1.2Compare results of incore to NIS 3.3.1.3 3.3.1.3Perform.

TADOT 3.3.1.4 3.3.1.4Perform ACTUATION LOGIC TEST 3.3.1.5 3.3.1.5Calibrate excore channels to agree with incore 3.3.1.6 3.3.1.6Perform COT 3.3.1.7 3.3.1.7Perform COT 3.3.1.8 3.3.1.8Perform TADOT 3.3.1.9 3.3.1.9Channel Calibration 3.3.1.10Channel Calibration (Neutron detectors excluded) 3.3.1.11 3.3.1.10Channel Calibration (Include resistance temperature detector) 3.3.1.12Perform COT 3.3.1.13 3.3.1.13Perform TADOT 3.3.1.14 3.3.1.11Perform TADOT (verification of setpoint is not required) 3.3.1.15Verify RTS time response 3.3.1.16ESFAS Instrumentation 3.3.2 3.3.2Channel Check 3.3.2.1 3.3.2.1Perform Actuation Logic Test 3.3.2.2 3.3.2.7Perform Actuation Logic Test (continuity may be excluded) 3.3.2.3Perform Master Relay Test 3.3.2.4 3.3.2.7Perform COT 3.3.2.5 3.3.2.2Perform Slave Relay Test 3.3.2.6 3.3.2.7Perform TADOT 3.3.2.7 3.3.2.3Perform TADOT (setpoint verification not required for manual 3.3.2.8 3.3.2.4functions)

Channel Calibration 3.3.2.9 3.3.2.5ESFAS Time Response 3.3.2.10 3.3.2.7Verify pressurizer pressure low and steam line pressure low NOT 3.3.2.6bypassedPAM Instrumentation 3.3.3 3.3.3Channel Check 3.3.3.1 3.3.3.1Channel Calibration 3.3.3.2 3.3.3.2Remote Shutdown System 3.3.4Channel Check 3.3.4.1Verify control circuit and transfer switch functional 3.3.4.2Channel Calibration 3.3.4.3Perform TADOT (reactor trip breakers) 3.3.4.4LOP DG Start Instrumentation 3.3.5 3.3.4Channel Check 3.3.5.1Perform TADOT 3.3.5.2 3.3.4.1Channel Calibration 3.3.5.3 3.3.4.2Containment Purge and Exhaust Isolation Instrumentation 3.3.6Channel Check 3.3.6.1Perform Actuation Logic Test 3.3.6.2 LAR -Adoption of TSTF-425, Revision 3Docket No. 50-244Attachment 5Page 3 of 10Technical Specification Section Title/Surveillance Description*

TSTF-425 GINNAPerform Master Relay Test 3.3.6.3Perform Actuation Logic Test (applicable to ESFAS 3.3.6.4instrumentation)

Perform Master Relay Test 3.3.6.5Perform COT 3.3.6.6Perform Slave Relay Test 3.3.6.7Perform TADOT (validation of setpoint not required) 3.3.6.8Channel Calibration 3.3.6.9Containment Ventilation Isolation Instrumentation 3.3.5Channel Check 3.3.5.1Perform COT 3.3.5.2Perform Actuation Logic Test 3.3.5.3Channel Calibration 3.3.5.4CREFS Actuation Instrumentation 3.3.7Channel Check 3.3.7.1Perform COT 3.3.7.2Perform Actuation Logic Test 3.3.7.3Perform Master Relay Test 3.3.7.4Perform Actuation Logic Test (applicable to ESFAS 3.3.7.5instrumentation)

Perform Master Relay Test (applicable to ESFAS instrumentation) 3.3.7.6Perform Slave Relay Test 3.3.7.7Perform TADOT (validation of setpoint not required) 3.3.7.8Channel Calibration 3.3.7.9CREATS Actuation Instrumentation 3.3.6Channel Check 3.3.6.1Perform COT 3.3.6.2Perform TADOT 3.3.6.3Channel Calibration 3.3.6.4Perform Actuation Logic Test 3.3.6.5FBACS Actuation Instrumentation 3.3.8Channel Check 3.3.8.1Perform COT 3.3.8.2Perform Actuation Logic Test 3.3.8.3Perform TADOT (validation of setpoint not required) 3.3.8.4Channel Calibration 3.3.8.5BDPS 3.3.9Channel Check 3.3.9.1Perform COT 3.3.9.2Channel Calibration 3.3.9.3RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 3.4.1Verify pressurizer pressure per COLR 3.4.1.1 3.4.1.1Verify RCS average temperature per COLR 3.4.1.2 3.4.1.2Verify RCS total flow rate per COLR 3.4.1.3 3.4.1.3Verify by heat balance that RCS total flow rate per COLR 3.4.1.4RCS Minimum Temperature for Criticality 3.4.2 3.4.2 LAR -Adoption of TSTF-425, Revision 3Docket No. 50-244Attachment 5Page 4 of 10Technical Specification Section Title/Surveillance Description*

TSTF-425 GINNAVerify RCS average temperature in each loop :[540] 3.4.2.1Verify RCS average temperature in each loop ->[540] Only required

---

3.4.2.2if Tave alarm is inoperable and RCS loop Tave <[547]RCS P/T Limit 3.4.3 3.4.3Verify RCS pressure, temperature, heat up and cooldown rates per 3.4.3.1 3.4.3.1PTLRRCS Loops -MODES 1 and 2 3.4.4Verify each RCS loop is in operation 3.4.4.1RCS Loops -MODES 1 >8.5% RTP 3.4.4Verify each RCS loop is in operation 3.4.4.1RCS Loops -MODES 1 < 8.5% RTP, 2 and 3 3.4.5Verify required RCS loop is in operation 3.4.5.1Verify steam generator secondary side water level 3.4.5.2Verify correct breaker alignment and power to required RCP pump --------------

3.4.5.3RCS Loops -MODE 3 3.4.5Verify required RCS loops are in operation 3.4.5.1Verify steam generator water level 3.4.5.2Verify correct breaker alignment and power to pumps 3.4.5.3RCS Loops -MODE 4 3.4.6 3.4.6Verify required RCS or RHR loop is in operation 3.4.6.1 3.4.6.1Verify steam generator water level 3.4.6.2 3.4.6.2Verify correct breaker alignment and power to pump 3.4.6.3 3.4.6.3RCS Loops -MODE 5, Loops Filled 3.4.7 3.4.7Verify required RHR loop is in operation 3.4.7.1 3.4.7.1Verify steam generator secondary side water level 3.4.7.2 3.4.7.2Verify correct breaker alignment and power to required RHR pump 3.4.7.3 3.4.7.3RCS Loops -MODE 5, Loops Not Filled 3.4.8 3.4.8Verify required RHR loop is in operation 3.4.8.1 3.4.8.1Verify correct breaker alignment and power to RHR pump 3.4.8.2 3.4.8.2Pressurizer 3.4.9 3.4.9Verify pressurizer water level 3.4.9.1 3.4.9.1Verify capacity of each group of pressurizer heaters 3.4.9.2 3.4.9.2Verify required pressurizer heaters capable being powered by 3.4.9.3emergency powerPressurizer PORVs 3.4.11 3.4.11Cycle each block valve 3.4.11.1 3.4.11.1Cycle each PORV 3.4.11.2 3.4.11.2Cycle each solenoid air control valve 3.4.11.3Verify PORVs and block valves capable being powered by 3.4.11.4emergency powerLTOP System 3.4.12 3.4.12Verify no SI pump capable of injecting 3.4.12.1 3.4.12.13.4.12.2Verify maximum of [one] [HPI] pump capable of injecting 3.4.12.1 3.4.12.23.4.12.2Verify maximum of charging pump capable of injecting 3.4.12.2 3.4.12.2Verify each accumulator is isolated 3.4.12.3 3.4.12.3 LAR -Adoption of TSTF-425, Revision 3Docket No. 50-244Attachment 5Page 5 of 10Technical Specification Section Title/Surveillance Description*

TSTF-425 GINNAVerify RHR suction valve is open 3.4.12.4Verify RCS vent path 3.4.12.5 3.4.12.4Verify (required)

PORV block valve open 3.4.12.6 3.4.12.5Verify RHR suction valve is locked open and power removed 3.4.12.7Perform COT (excluding actuation) 3.4.12.8 3.4.12.6Verify power removed from each ECCS accumulator MOV 3.4.12.7Channel Calibration 3.4.12.9 3.4.12.8RCS Operational Leakage 3.4.13 3.4.13Verify RCS operational leakage within limits 3.4.13.1 3.4.13.1Verify primary to secondary leakage within limits 3.4.13.2 3.4.13.2RCS PIV Leakage 3.4.14 3.4.14Verify leakage from each RCS PIV 3.4.14.1 3.4.14.13.4.14.2Verify RHR interlock functionality (opening) 3.4.14.2Verify RHR interlock functionality (closure) 3.4.14.3RCS Leakage Detection Instrumentation 3.4.15 3.4.15Channel Check 3.4.15.1 3.4.15.1Perform COT (atmosphere radioactivity monitor) 3.4.15.2 3.4.15.2Channel Calibration (containment sump monitor) 3.4.15.3 3.4.15.3Channel Calibration (containment atmosphere monitor) 3.4.15.4 3.4.15.4Channel Calibration (containment air cooler flow rate monitor) 3.4.15.5RCS Specific Activity 3.4.16 3.4.16Verify gross specific activity 3.4.16.1 3.4.16.1Verify Dose Equivalent 1-131 3.4.16.2 3.4.16.2Determine E 3.4.16.3 3.4.16.3RCS Loop Isolation Valves 3.4.17Verify loop isolation valve is open and power removed 3.4.17.1RCS Loops -Test Exceptions 3.4.19Verify thermal power is <P-7 3.4.19.1Accumulators 3.5.1 3.5.1Verify accumulator isolation valves fully open 3.5.1.1 3.5.1.1Verify borated water volume 3.5.1.2 3.5.1.2Verify Nitrogen cover pressure 3.5.1.3 3.5.1.3Verify boron concentration 3.5.1.4 3.5.1.4Verify power removed from isolation valve 3.5.1.5 3.5.1.5ECCS -Operating 3.5.2 3.5.2Verify (listed) valves in proper position with power removed 3.5.2.1 3.5.2.1Verify valves in flow path in the proper position with power 3.5.2.2 3.5.2.2removedVerify breakers, for each valve listed in 3.5.2.1, in correct position 3.5.2.3Verify ECCS piping full of water 3.5.2.3Verify ECCS automatic valves actuates to its proper position on an 3.5.2.5 3.5.2.5actual or simulated signalVerify ECCS pumps starts automatically on an actual or simulated 3.5.2.6 3.5.2.6signalVerify ECCS throttle valves (listed) in the correct position 3.5.2.7Visual inspection of ECCS train 3.5.2.8 3.5.2.7 LAR -Adoption of TSTF-425, Revision 3Docket No. 50-244Attachment 5Page 6 of 10Technical Specification Section Title/Surveillance Description*

TSTF-425 GINNARWST 3.5.4 3.5.4Verify RWST borated water temperature 3.5.4.1Verify RWST borated water volume 3.5.4.2 3.5.4.1Verify RWST boron concentration 3.5.4.3 3.5.4.2Seal Injection Flow 3.5.5Verify manual seal injection valves adjusted to proper flow 3.5.5.1BIT 3.5.6Verify BIT borated water temperature 3.5.6.1Verify BIT borated water volume 3.5.6.2Verify BIT boron concentration 3.5.6.3Containment Air Locks 3.6.2 3.6.2Verify one door can be opened at a time 3.6.2.2 3.6.2.2Containment Isolation Valves / Boundaries 3.6.3 3.6.3Verify purge valve [42 inch] closed 3.6.3.1Verify purge valve [8 inch] closed 3.6.3.2 3.6.3.1Verify manual isolation valves, blind flanges outside containment 3.6.3.3 3.6.3.2closedVerify isolation times 3.6.3.5Cycle each testable check valves through one full cycle 3.6.3.6Leakrate purge valves with resilient seals 3.6.3.7Verify automatic valves actuate to their correct position on an 3.6.3.8 3.6.3.6actual or simulated signalCycle each testable check valves through one full cycle 3.6.3.9Verify purge valve is blocked to restrict flow 3.6.3.10Containment Pressure 3.6.4A 3.6.4Verify containment pressure within limits 3.6.4A.1 3.6.4.1Containment Temperature 3.6.5A 3.6.5AVerify containment average temperature within limits 3.6.5A.1 3.6.5.1Containment Spray and Cooling Systems / (CS,CRFC and NaOH 3.6.6A 3.6.6System)Verify valves in the flow path in the correct position 3.6.6A.1 3.6.6.2Operate containment cooling system (->15 minutes) 3.6.6A.2 3.6.6.4Verify containment cooling train water flow rate(->700 gpm) 3.6.6A.3 3.6.6.5Verify containment spray valves actuate to their correct position on 3.6.6A.5 3.6.6.10an actual or simulated signalVerify containment spray pumps starts on an actual or simulated 3.6.6A.6 3.6.6.11signalVerify containment cooling train starts on an actual or simulated 3.6.6A.7 3.6.6.12signalSpray Additive System 3.6.7Verify valves in the flow path in the correct position 3.6.7.1 3.6.6.3Verify spray tank solution volume 3.6.7.2 3.6.6.7Verify NaOH concentration 3.6.7.3 3.6.6.8Verify containment spray additive tank valves actuate to their 3.6.7.4 3.6.6.13correct position on an actual or simulated signalVerify spray additive flow from each solution flow path 3.6.7.5 3.6.6.14MSIV / (MSIV and Non-Return Check Valves) 3.7.2 3.7.2 LAR -Adoption of TSTF-425, Revision 3Docket No. 50-244Attachment 5Page 7 of 10Technical Specification Section Title/Surveillance Description*

TSTF-425 GINNAVerify MSIV actuate to its isolation position on an actual or 3.7.2.2 3.7.2.3simulated signalMFIVs and MFRVs, and Associated Bypass Valves 3.7.3 3.7.3Verify the isolation time of each MFIV, MFRV and associated 3.7.3.2bypass valveAtmosphere Dump (Relief)

Valves (ADVs)(ARVs) 3.7.4 3.7.4Verify once complete cycle of each ADV(ARV) 3.7.4.1 3.7.4.1Verify once complete cycle of each ADV (ARV) block valve 3.7.4.2 3.7.4.2AFW System 3.7.5 3.7.5Verify valves in the water and steam flow path in their correct 3.7.5.1 3.7.5.1positionVerify each AFW automatic valve actuates to the isolation position 3.7.5.3 3.7.5.5on an actual or simulated signalVerify each AFW pump starts automatically on an actual or 3.7.5.4 3.7.5.6simulated signalVerify each SAFW train can be operated from control room 3.7.5.7Condensate Storage Tank (CST) 3.7.6 3.7.6Verify CST level 3.7.6.1 3.7.6.1Component Cooling Water System (CCW) 3.7.7 3.7.7Verify each CCW valve is in the correct position 3.7.7.1 3.7.7.1Verify each CCW valve in the flow path actuates to the correct 3.7.7.2position on an actual or simulated signalVerify each CCW pump starts automatically on an actual or 3.7.7.3simulated signalService Water System (SWS) 3.7.8 3.7.8Verify screen house bay water level and temperatures 3.7.8.1Verify each SWS valve is in the correct position 3.7.8.1 3.7.8.2Verify each SWS valve in the flow path actuates to the correct 3.7.8.2 3.7.8.4position on an actual or simulated signalVerify each SWS pump starts automatically on an actual or 3.7.8.3 3.7.8.5simulated signalVerify SW loop header cross-tie valves in correct position 3.7.8.3Ultimate Heat Sink (UHS) 3.7.9Verify water level in the UHS 3.7.9.1Verify average water temperature in the UHS 3.7.9.2Operate each cooling tower fan (>15 minutes) 3.7.9.3Verify cooling fan starts automatically on an actual or simulated 3.7.9.4signalControl Room Emergency Filtration System (CREFS) 3.7.10Operate each CREFS train (>15 minutes) with the heaters on (>15 3.7.10.1minutes)Verify each CREF train actuates on an actual or simulated signal 3.7.10.3Verify each CREF train maintain a positive pressure 3.7.10.4Control Room Emergency Air Temperature Control System 3.7.11 3.7.9(CREATCS)

Verify the CREATMS removes the assume heat load 3.7.11.1Operate CREATS filtration train (>15 minutes) 3.7.9.1 LAR -Adoption of TSTF-425, Revision 3Docket No. 50-244Attachment 5Page 8 of 10Technical Specification Section Title/Surveillance Description*

TSTF-425 GINNAVerify each CREATS train actuates on an actual or simulated signal -------------

3.7.9.3Auxiliary Building Ventilation System 3.7.10Verify ABVS in operation 3.7.10.1Verify ABVS maintains negative pressure with respect to Aux. Bldg. -------------

3.7.10.2ECCS PREACS 3.7.12Operate each PREAC for ->10 hours with heaters on or for -15 3.7.12.1minutes for systems without heatersVerify each PREAC train actuates on an actual or simulated signal 3.7.12.3Verify PREAC can maintained pressure 3.7.12.4Verify each ECCS PREAC filter bypass damper closed 3.7.12.5FBACS 3.7.13Operate each FBACS for -10 hours with heaters on or for >15 3.7.13.1minutes for systems without heatersVerify each FBACS train actuates on an actual or simulated signal 3.7.13.3Verify FBACS can maintained pressure 3.7.13.4Verify each FBACS filter bypass damper closed 3.7.13.5PREACS 3.7.14Operate each PREAC for ->10 hours with heaters on or for 15 3.7.14.1minutes for systems without heatersVerify each PREAC train actuates on an actual or simulated signal 3.7.14.3Verify PREAC can maintained pressure 3.7.14.4Verify each ECCS PREAC filter bypass damper closed 3.7.14.5Fuel Storage Pool Water Level 3.7.15 3.7.11Verify fuel storage pool water level (-23 feet) 3.7.15.1 3.7.11.1Spent Fuel Pool Boron Concentration 3.7.16 3.7.12Verify spent fuel boron concentration within limits 3.7.16.1 3.7.12.1Secondary Specific Activity 3.7.18 3.7.14Verify specific activity of Dose Equivalent 1-131 (50.10) 3.7.18.1 3.7.14.1AC Sources -Operating 3.8.1 3.8.1Verify correct breaker alignment 3.8.1.1 3.8.1.1Verify each diesel starts from standby conditions 3.8.1.2 3.8.1.2Verify each diesel is synchronized and loaded (for -60 minutes) 3.8.1.3 3.8.1.3Verify each day tank contains proper fuel quantity

(- 220 gallons) 3.8.1.4 3.8.1.4Check and remove accumulated water 3.8.1.5Verify fuel oil transfer operation (from storage tank to day tanks) 3.8.1.6 3.8.1.5Verify each diesel starts from standby conditions in proper time 3.8.1.7 3.8.1.2(510 sec)Verify transfer of AC power sources (Normal to Alternate) 3.8.1.8 3.8.1.6Load rejection test (largest post-accident load) 3.8.1.9 3.8.1.7Verify diesel does not trip and voltage is maintained during and 3.8.1.10 3.8.1.7following the load rejection Verify diesel performs properly on an actual or simulated loss of 3.8.1.11 3.8.1.9offsite power signalVerify on an actual or simulated ESF actuation each diesel auto 3.8.1.12 3.8.1.9starts from standby conditions Verify non critical trips are bypassed 3.8.1.13 3.8.1.8Verify each diesel operates for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 3.8.1.14 LAR -Adoption of TSTF-425, Revision 3Docket No. 50-244Attachment 5Page 9 of 10Technical Specification Section Title/Surveillance Description*

TSTF-425 GINNAVerify diesel starts and performs properly within 5 minutes of 3.8.1.15operating for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> a maximum loadVerify diesel synchronizes with offsite power while loaded with 3.8.1.16emergency loadsVerify an actual or simulated ESF signal overrides a test signal 3.8.1.17Verify interval between each sequenced load block 3.8.1.18Verify on an actual or simulated loss of offsite power in conjunction 3.8.1.19with an actual or simulated ESF signal the diesel performs properlyVerify when started simultaneously from standby conditions each 3.8.1.20diesel performs properlyDiesel Fuel Oil, Lube Oil, and Starting Air 3.8.3 3.8.3Verify each fuel oil storage tank volume (gallons) 3.8.3.1 3.8.3.1Verify lube oil inventory 3.8.3.2Verify diesel start receiver pressure 3.8.3.4Check and remove accumulated water 3.8.3.5DC Sources -Operating 3.8.4 3.8.4Verify battery terminal voltage 3.8.4.1 3.8.4.1Verify battery charger can recharge the battery 3.8.4.2Verify battery capacity 3.8.4.3 3.8.4.23.8.4.3Battery Parameters 3.8.6 3.8.6Verify battery float current 3.8.6.1Verify each pilot cell voltage 3.8.6.2Verify connected batteries electrolyte level 3.8.6.3 3.8.6.1Verify average electrolyte temperature (fifth cell of each battery) 3.8.6.5Verify specific gravity of pilot cell 3.8.6.3Verify specific gravity of each connected cell 3.8.6.6Verify each pilot cell temperature 3.8.6.4 3.8.6.4Verify connected battery cell voltage 3.8.6.5 3.8.6.2Verify battery capacity 3.8.6.6 3.8.4.3Inverters

-Operating 3.8.7Verify inverter voltage and alignment to AC buses 3.8.7.1AC Instrument Bus Sources -MODES 1, 2, 3 and 4 3.8.7Verify correct static switch alignment 3.8.7.1Verify correct Class 1 E CVT alignment 3.8.7.2Inverters

-Shutdown 3.8.8Verify inverter voltage and alignment to AC buses 3.8.8.1AC Instrument Bus Sources -MODES 5 and 6 3.8.8Verify correct static switch alignment 3.8.8.1Verify correct Class 1 E CVT alignmnet 3.8.8.2Distribution Systems -Operating 3.8.9Verify correct breaker alignment and voltage of required buses 3.8.9.1Distribution Systems -MODES 1, 2, 3 and 4 3.8.9Verify correct breaker alignment and voltage of required buses 3.8.9.1Distribution Systems -Shutdown 3.8.10Verify correct breaker alignment and voltage of required buses 3.8.10.1Distribution Systems -MODES 5 and 6 3.8.10 LAR -Adoption of TSTF-425, Revision 3Docket No. 50-244Attachment 5Page 10 of 10Technical Specification Section Title/Surveillance Description*

TSTF-425 GINNAVerify correct breaker alignment and voltage of required buses 3.8.10.1Boron Concentration 3.9.1 3.9.1Verify boron concentration within the COLR 3.9.1.1 3.9.1.1Unborated Water Source Isolation Valves 3.9.2Verify each valve that isolates unborated water source is closed 3.9.2.1Nuclear Instrumentation 3.9.3 3.9.2Channel Check 3.9.3.1 3.9.2.1Channel Calibration 3.9.3.2 3.9.2.2Containment Penetrations 3.9.4 3.9.3Verify containment penetration in required status 3.9.4.1 3.9.3.1Verify purge and exhaust valves isolate on an actual or simulated 3.9.4.2 3.9.3.2signalRHR and Coolant Circulation

-High Water Level (>23 Ft) 3.9.5 3.9.4Verify one RHR loop in operation circulating reactor coolant 3.9.5.1 3.9.4.1RHR and Coolant Circulation

-Low Water Level (< 23 Ft) 3.9.6 3.9.5Verify one RHR loop in operation 3.9.6.1 3.9.5.1Verify correct breaker alignment and power to operating RHR 3.9.6.2 3.9.5.2pumpRefueling Cavity Water Level 3.9.7 3.9.6Verify refueling cavity water level above flange 3.9.7.1 3.9.6.1__ __ _ 4+4+-1-* The Technical Specification Section Title/Surveillance Description portion of this attachment isa summary description of the referenced TSTF-425 (NUREG-1431)/Ginna TS Surveillances which is provided for information purposes only and is not intended to be a verbatimdescription of the TS Surveillances.

ATTACHMENT 6License Amendment RequestR. E. Ginna Nuclear Power PlantDocket No. 50-244Application 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)Proposed No Significant Hazards Consideration LAR -Adoption of TSTF-425, Revision 3 Attachment 6Docket No. 50-244 Page 1 of 2PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION Description of Amendment Request:

This amendment request involves the adoption ofapproved changes to the standard technical specifications (STS) for Westinghouse Plants,WOG STS (NUREG-1431),

to allow relocation of specific TS surveillance frequencies to alicensee-controlled program.

The proposed changes are described in Technical Specification Task Force (TSTF) Traveler, TSTF-425, Revision 3 (ADAMS Accession No. ML090850642) related to the Relocation of Surveillance Frequencies to Licensee Control -RITSTF Initiative 5band are described in the Notice of Availability published in the Federal Register on July 6, 2009(74 FR 31996).The proposed changes are consistent with NRC-approved Industry/

TSTF Traveler, TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control -RITSTF Initiative 5b."The proposed changes relocate surveillance frequencies to a licensee-controlled

program, theSurveillance Frequency Control Program (SFCP). The changes are applicable to licensees using probabilistic risk guidelines contained in NRC-approved NEI 04-10, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies,"

(ADAMS Accession No. 071360456).

Basis for proposed no significant hazards consideration:

As required by 10 CFR 50.91 (a),the Exelon analysis of the issue of no significant hazards consideration is presented below:1. Do the proposed changes involve a significant increase in the probability or consequences ofany accident previously evaluated?

Response:

No.The proposed changes relocate the specified frequencies for periodic surveillance requirements to licensee control under a new Surveillance Frequency Control Program.Surveillance frequencies are not an initiator to any accident previously evaluated.

As aresult, the probability of any accident previously evaluated is not significantly increased.

Thesystems and components required by the technical specifications for which the surveillance frequencies are relocated are still required to be operable, meet the acceptance criteria forthe surveillance requirements, and be capable of performing any mitigation functionassumed in the accident analysis.

As a result, the consequences of any accident previously evaluated are not significantly increased.

Therefore, the proposed changes do not involve a significant increase in the probability orconsequences of an accident previously evaluated.

2. Do the proposed changes create the possibility of a new or different kind of accident fromany previously evaluated?

Response:

No.No new or different accidents result from utilizing the proposed changes.

The changes donot involve a physical alteration of the plant (i.e., no new or different type of equipment willbe installed) or a change in the methods governing normal plant operation.

In addition, the LAR -Adoption of TSTF-425, Revision 3 Attachment 6Docket No. 50-244 Page 2 of 2changes do not impose any new or different requirements.

The changes do not alterassumptions made in the safety analysis.

The proposed changes are consistent with thesafety analysis assumptions and current plant operating practice.

Therefore, the proposed changes do not create the possibility of a new or different kind ofaccident from any accident previously evaluated.

3. Do the proposed changes involve a significant reduction in the margin of safety?Response:

No.The design, operation, testing methods, and acceptance criteria for systems, structures, andcomponents (SSCs), specified in applicable codes and standards (or alternatives approvedfor use by the NRC) will continue to be met as described in the plant licensing basis(including the final safety analysis report and bases to TS), since these are not affected bychanges to the surveillance frequencies.

Similarly, there is no impact to safety analysisacceptance criteria as described in the plant licensing basis. To evaluate a change in therelocated surveillance frequency, Exelon will perform a probabilistic risk evaluation using theguidance contained in NRC approved NEI 04-10, Rev. 1, in accordance with the TS SFCP.NEI 04-10, Rev. 1, methodology provides reasonable acceptance guidelines and methodsfor evaluating the risk increase of proposed changes to surveillance frequencies consistent with Regulatory Guide 1.177.Therefore, the proposed changes do not involve a significant reduction in a margin of safety.Based upon the above, Exelon concludes that the requested changes do not involve asignificant hazards consideration as set forth in 10 CFR 50.92(c),

"Issuance of Amendment."

ATTACHMENT 7License Amendment RequestR. E. Ginna Nuclear Power PlantDocket No. 50-244Application 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)Proposed Inserts LAR -Adoption of TSTF-425, Revision 3 Attachment 7Docket No. 50-244 Page 1 of 1INSERT 1In accordance with the Surveillance Frequency Control ProgramINSERT 25.5.17 Surveillance Frequency Control proqramThis program provides controls for the Surveillance Frequencies.

The programshall 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 ofFrequencies of the Surveillance Requirements for which the Frequency iscontrolled by the program.b. Changes to the Frequency 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 provision of Surveillance Requirement 3.0.2 and 3.0.3 are applicable to the Frequencies established in the Surveillance Frequency ControlProgram.INSERT 3The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.