ML061180353

R.E. Ginna, License Amendment 96 Regarding Revised Loss-of-Coolant Accident Analyses

ML061180353
Person / Time
Site:	Ginna
Issue date:	05/31/2006
From:	Milano P D Plant Licensing Branch III-2
To:	Korsnick M G Ginna
Milano P, NRR/DLPM , 415-1457
References
TAC MC6860
Download: ML061180353 (28)

Text

May 31, 2006Mrs. Mary G. KorsnickVice President R.E. Ginna Nuclear Power Plant R.E. Ginna Nuclear Power Plant, LLC 1503 Lake Road Ontario, NY 14519

SUBJECT:

R.E. GINNA NUCLEAR POWER PLANT - AMENDMENT RE: REVISED LOSS-OF-COOLANT ACCIDENT ANALYSES (TAC NO. MC6860)

Dear Mrs. Korsnick:

The Commission has issued the enclosed Amendment No. 96 to Renewed Facility OperatingLicense No. DPR-18 for the R.E. Ginna Nuclear Power Plant. This amendment is in response to your application dated April 29, 2005, as supplemented on August 15 and December 9, 2005, and January 11 and 25, and May 9, 2006.The amendment revises Technical Specification (TS) 3.5.1, "Accumulators," and TS 3.5.4,"Refueling Water Storage Tank," to reflect the results of revised analyses performed to accommodate the proposed extended power uprate and revises TS 5.6.4, "Core OperatingLimits Report," to permit the use of approved methodology for large-break and small-break loss-of-coolant accident analyses.A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be incl udedin the Commission's biweekly Federal Register notice. Sincerely,/RA/Patrick D. Milano, Sr. Project ManagerPlant Licensing Branch I-1 Division of Operating Reactor Licensing Office of Nuclear Reactor RegulationDocket No. 50-244

Enclosures:

1. Amendment No. 96 to Renewed License No. DPR-18

2. Safety Evaluationcc w/encls: See next page May 31, 2006Mrs. Mary G. KorsnickVice President R. E. Ginna Nuclear Power Plant R.E. Ginna Nuclear Power Plant, LLC 1503 Lake Road Ontario, NY 14519

SUBJECT:

R.E. GINNA NUCLEAR POWER PLANT - AMENDMENT RE: REVISED LOSS-OF-COOLANT ACCIDENT ANALYSES (TAC NO. MC6860)

Dear Mrs. Korsnick:

The Commission has issued the enclosed Amendment No. 96 to Renewed Facility OperatingLicense No. DPR-18 for the R.E. Ginna Nuclear Power Plant. This amendment is in response to your application dated April 29, 2005, as supplemented on August 15 and December 9, 2005, and January 11 and 25, and May 9, 2006.The amendment revises Technical Specification (TS) 3.5.1, "Accumulators," and TS 3.5.4,"Refueling Water Storage Tank," to reflect the results of revised analyses performed to accommodate the proposed extended power uprate and revises TS 5.6.4, "Core OperatingLimits Report," to permit the use of approved methodology for large-break and small-break loss-of-coolant accident analyses.. A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be incl udedin the Commission's biweekly Federal Register notice. Sincerely,/RA/Patrick D. Milano, Sr. Project ManagerPlant Licensing Branch I-1 Division of Operating Reactor Licensing Office of Nuclear Reactor RegulationDocket No. 50-244

Enclosures:

1. Amendment No. 96 to Renewed License No. DPR-18

2. Safety Evaluationcc w/encls: See next page Accession Number: ML061180353OFFICELPLI-1\PMLPLI-1\LASPWB\BCIOLB\BCCSGB\(A)BCOGCLPLI-1\BCNAMEPMilanoSLittleJNakoskiNO'KeefeEMurphyJMooreRLauferDATE05/31/0605/31/0605/03/0605/17/0602/02/0605/23/0605/31/06Official Record Copy DATED: May 31, 2006AMENDMENT NO. 96 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-18 R.E. GINNA NUCLEAR POWER PLANTPUBLICLPLI-1 R/F R. LauferRidsNrrDorlLpla J. NakoskiRidsNrrDssSpwb F. Orr L. Ward E. MurphyRidsNrrDciCsgb Y. Diaz S. LittleRidsNrrLASLittle P. MilanoRidsNrrPMPMilano T. BoyceRidsNrrDirsItsb G. Hill (2)

OGC RidsOgcRp ACRSRidsAcrsAcnwMailCenter B. McDermott, RIRidsRgn1MailCentercc: Plant Service list R.E. Ginna Nuclear Power Plant cc:

Mr. Michael J. WallacePresident R.E. Ginna Nuclear Power Plant, LLC c/o Constellation Energy 750 East Pratt Street Baltimore, MD 21202Mr. John M. HeffleySenior Vice President and Chief Nuclear Officer Constellation Generation Group 1997 Annapolis Exchange Parkway Suite 500 Annapolis, MD 21401Kenneth Kolaczyk, Sr. Resident InspectorR.E. Ginna Nuclear Power Plant U.S. Nuclear Regulatory Commission 1503 Lake Road Ontario, NY 14519Regional Administrator, Region IU.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406Mr. Peter R. Smith, PresidentNew York State Energy, Research, and Development Authority 17 Columbia Circle Albany, NY 12203-6399Mr. Carey W. Fleming, EsquireSenior Counsel - Nuclear Generation Constellation Generation Group, LLC 750 East Pratt Street, 17th Floor Baltimore, MD 21202Mr. Charles Donaldson, EsquireAssistant Attorney General New York Department of Law 120 Broadway New York, NY 10271Ms. Thelma Wideman, DirectorWayne County Emergency Management Office Wayne County Emergency Operations Center 7336 Route 31 Lyons, NY 14489Ms. Mary Louise MeisenzahlAdministrator, Monroe County Office of Emergency Preparedness 1190 Scottsville Road, Suite 200Rochester, NY 14624Mr. Paul EddyNew York State Department of Public Service 3 Empire State Plaza, 10th Floor Albany, NY 12223 R.E. GINNA NUCLEAR POWER PLANT, LLCDOCKET NO. 50-244R.E. GINNA NUCLEAR POWER PLANTAMENDMENT TO RENEWED FACILITY OPERATING LICENSEAmendment No. 96Renewed License No. DPR-181.The Nuclear Regulatory Commission (the Commission or the NRC) has f ound that:A.The application for amendment filed by the R.E. Ginna Nuclear Power Plant, LLC (the licensee) dated April 29, 2005, as supplemented on August 15 and December 9, 2005, and January 11 and 25, and May 9, 2006, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.The facility will operate in conformity with the application, the provisions of theAct, and the rules and regulations of the Commission;C.There is reasonable assurance: (i) that the activities authorized by thisamendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with theCommission's regulations;D.The issuance of this amendment will not be inimical to the common defense andsecurity or to the health and safety of the public; andE.The issuance of this amendment is in accordance with 10 CFR Part 51 of theCommission's regulations and all applicable requirements have been satisfied. 2.Accordingly, the license is amended by changes to the Technical Specifications asindicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-18 is hereby amended to read as follows: (2)Technical SpecificationsThe Technical Specifications contained in Appendix A, as revised throughAmendment No. 96, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.3.This license amendment is effective as of the date of its issuance and shall beimplemented prior to startup from the fall 2006 refueling outage. Implementation shallinclude revisions to plant procedures and the completion of operator training as described in the licensee's April 29, 2005, application, as supplemented, and asdiscussed in the NRC staff's safety evaluation dated May 31, 2006.FOR THE NUCLEAR REGULATORY COMMISSION/RA/Richard J. Laufer, ChiefPlant Licensing Branch I-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and LicenseDate of Issuance: May 31, 2006 ATTACHMENT TO LICENSE AMENDMENT NO. 96RENEWED FACILITY OPERATING LICENSE NO. DPR-18DOCKET NO. 50-244Replace the following page of the Renewed Facility Operating License with the attached revisedpage. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.Remove Insert 33Replace the following pages of the Appendix A Technical Specifications with the attachedrevised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.Remove Insert3.5.1-13.5.1-13.5.1-23.5.1-2 3.5.4-13.5.4-1 5.6-15.6-1 5.6-25.6-2 5.6-35.6-3 5.6-45.6-4 5.6-55.6-5 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATIONRELATED TO AMENDMENT NO. 96 TO RENEWED FACILITYOPERATING LICENSE NO. DPR-18R.E. GINNA NUCLEAR POWER PLANT, INC.R.E. GINNA NUCLEAR POWER PLANTDOCKET NO. 50-24

41.0 INTRODUCTION

By letter dated April 29, 2005, as supplemented on August 15 and December 9, 2005, andJanuary 11 and 25, and May 9, 2006 (Agencywide Documents Access and Management System Accession Nos. ML051260239, ML052310155, ML053480362, ML060180262, ML060960416, and ML061350375, respectively), R.E. Ginna Nuclear Power Plant, Inc. (the licensee) submitted a request for changes to the R.E. Ginna Nuclear Power Plant (Ginna)

Technical Specifications (TSs). The requested changes would revise: (1) TS 3.5.1 and TS 3.5.4 to reflect the results of revised analyses performed to accommodate the planned power uprate and (2) TS 5.6.4.b to permit the use of methodologies approved by the NuclearRegulatory Commission (NRC) for large-break and small-break loss-of-coolant accident(LBLOCA and SBLOCA) analyses. Specifically, the proposed TS changes would modify the volume and boron concentration requirements for the emergency core cooling system (ECCS)accumulators, revise the boron concentration requirements for the refueling water storage tank (RWST), and revise the list of referenced analytical methods specified in TS 5.6.5.b.The August 15 and December 9, 2005, and January 11 and 25, and May 9, 2006, lettersprovided additional information that clarified the application, did not expand the scope of theapplication as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register on June 7, 2005(70 FR 33219). On July 7, 2005, the licensee provided its application and supporting licensing report requestingan increase in the maximum steady-state thermal power level from 1520 megawatts thermal (MWt) to 1775 MWt, which is an increase of about 16.8% and is considered an extended power uprate (EPU).

2.0 REGULATORY EVALUATION

2.1 Proposed

TS ChangesBecause of the expected changes in the nuclear fuel design associated with the planned EPU,the licensee performed revised analyses using an NRC-approved evaluation methodology atthe uprated operating conditions. Based on the revised analyses, modification of the TS requirements for ECCS accumulator water liquid volume and boron concentration and RWST boron concentration was needed. In addition, the list of approved analysis methodologies specified in TS 5.6.5.b needs to be modified to reflect use of the revised analysis techniques.

In this regard, the licensee proposes the following revisions to the TSs:1. TS 3.5.1, "Accumulators"a. Surveillance Requirement (SR) 3.5.1.2 currently requires that the borated watervolume in each accumulator be verified to be > 1111 cubic feet (50%) and <

1139 cubic feet (82%). The licensee proposes to change SR 3.5.1.2 to state "Verify borated water volume in each accumulator is > 1090 cubic feet (24%) and

< 1140 cubic feet (83%)."b. SR 3.5.1.4 currently requires that the boron concentration in each accumulatorbe verified to be > 2100 parts per million (ppm) and < 2600 ppm. The licenseeproposes to change SR 3.5.1.4 to state "Verify boron concentration in each accumulator is > 2550 ppm and <3050 ppm."2. TS 3.5.4, "RWST"a. SR 3.5.4.2 currently requires that the RWST boron concentration be verified tobe > 2300 ppm and < 2600 ppm. The licensee proposes to change SR 3.5.4.2 to state "Verify RWST boron concentration is > 2750 ppm and < 3050 ppm."3. TS 5.6.5, "Core Operating Limits Report (COLR)"a. In TS 5.6.5.b, the licensee proposes to delete the current references in items 2, 7, 8, and 9.b. The licensee proposes to replace Item 2 with WCAP-16009-P-A, "RealisticLarge-Break LOCA Evaluation Methodology Using the Automated StatisticalTreatment of Uncertainty (ASTRUM)," January 2005.c. The licensee proposes to replace Item 7 with WCAP-10054-P-A, Addendum 2,Revision 1, "Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection Into the Broken Loop and COSI Condensation Model," July 1997. (Methodology for limiting condition for operation (LCO) 3.2.1)d. The licensee proposes to replace Item 8 with WCAP-1 1145-P-A,"WestinghouseSmall Break LOCA ECCS Evaluation Model Generic Study With the NOTRUMP Code," October 1986. (Methodology for LCO 3.2.1) e. The licensee proposes to replace Item 9 with WCAP-10079-P-A, "NOTRUMP - ANodal Transient Small Break and General Network Code," August 1985.

(Methodology for LCO 3.2.1)f. The licensee proposes to add a new Item 11 to include WCAP-14710-P-A, "1-DHeat Conduction Model for Annular Fuel Pellets," May 1998. (Methodology for LCO 3.2.1)2.2 BackgroundThe licensee requested approval to apply the NRC-approved Westinghouse best-estimate (BE)LBLOCA methodology described in WCAP-16009-P-A, "Realistic Large Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM),"January 2005 (Reference 2), at Ginna. The licensee also identified that the SBLOCA analyses for the proposed power uprate were performed using the Westinghouse NOTRUMP Code (Reference 6).The NRC staff reviewed the licensee's evaluation of the ECCS performance analyses for Ginnacompleted in accordance with the ASTRUM and NOTRUMP methodologies, operating at about 117 percent of its current licensed core power of 1520 MWt (the analyses were conducted at the uprated power of 1775 MWt plus 2 percent measurement uncertainty or 1811 MWt). For Ginna, the LOCA analyses were conducted assuming the plant used a mixed core containingWestinghouse 14 x 14 nine-grid Optimized Fuel Assemblies (OFAs) and 422 Vantage

+ fuel(422V+).Ginna is a two-loop, pressurized-water reactor (PWR) of the Westinghouse Electric Companydesign, enclosed within a large, dry containment. The ECCS consists of residual heat removal system (RHR) upper plenum injection (UPI) flow, high-head safety injection (HHSI) flowdelivered to the cold legs, and two accumulators with a cover gas pressure of 714.7 psia, also injecting into the cold legs. The shut-off head of the RHR low-pressure injection pumps isabout 160 psia. The proposed EPU steady-state power level of 1775 MWt (with analysis at 1811 MWt, whichincludes a 2% power uncertainty) represents a core power increase of almost 16.8% above the current core power of 1520 MWt. The addition of 6 MWt for the heat input of the two reactor coolant pumps (RCPs) brings the nuclear steam supply system (NSSS) power level to 1817MWt. The SBLOCA and post-LOCA long-term cooling analyses conducted by the NRC staffwere performed at an NSSS power level of 1817 MWt.Implementation of an EPU requires re-analyzing the LOCA design-basis accident due to thechanges to the accumulator water volume and boron concentration limits and RWST boron concentration limits. Thus, the licensee is also proposing changes to the Ginna TSs under 10 CFR 50.90. The proposed amendment credits the use of sodium hydroxide (NaOH) for maintaining the containment sump pH above 7 for the 30-day period after a LOCA. Guidance for the implementation of this amendment is provided in Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," as applied to this specific change.2.3 Regulatory Basis 2.3.1 LBLOCA and SBLOCA AnalysesThe LBLOCA and SBLOCA analyses are performed to demonstrate that the system designwould provide sufficient ECCS flow to transfer the heat from the reactor core following a LOCA at a rate such that: (1) fuel and clad damage that could interfere with continued effective core cooling would be prevented, and (2) the clad metal-water reaction would be limited to less than enough to compromise cladding ductility and would not result in excessive hydrogengeneration. The NRC staff reviewed the analyses to assure that the analyses reflected suitableredundancy in components and features; and suitable interconnections, leak detection, isolation, and containment capabilities were available such that the safety functions could beaccomplished, assuming a single failure, for LOCAs considering the availability of onsite andoffsite electric power (assuming offsite electric power is not available, with onsite electric power available; or assuming onsite electric power is not available, with offsite electric power available). The acceptance criteria for ECCS performance are provided in Section 50.46 of Title 10 of the Code of Federal Regulations (10 CFR 50.46), and were used by the NRC staff inassessing the acceptability of the Westinghouse ASTRUM and NOTRUMP methodologies forGinna.The NRC staff also reviewed the limitations and conditions stated in its safety evaluation (SE)supporting approval of the Westinghouse ASTRUM and NOTRUMP methodologies and the range of parameters described in the ASTRUM and NOTRUMP topical reports in its assessment of the acceptability of the methodology for Ginna.2.3.2Post-LOCA Containment Sump Conditions Ginna's Updated Final Safety Analysis Report (UFSAR) Section 6.1.2.1.4, "Design ChemicalComposition of the Emergency Core Cooling Solution," states that the minimum value of containment sump liquid pH was revised to a value of 7.0, as detailed in NRC StandardReview Plan (SRP) Section 6.5.2, and as specified in Branch Technical Position MTEB 6-1.

The pH of the aqueous solution collected in the containment sump after completion of the injection of containment spray and ECCS water and all additives for reactivity control, fission product removal, or other purposes should be maintained at a level sufficiently high to provide assurance that significant long-term iodine re-evolution does not occur.The expected long-term partition coefficient is used to calculate the long-term iodine retention. Long-term iodine retention may be assumed only when the equilibrium sump solution pH, aftermixing and dilution with the primary coolant and ECCS injection, is above 7.0. This pH value should be achieved by the onset of the spray recirculation mode. As given in Branch Technical Position MTEB 6-1, experience has shown that maintaining the pH of borated solutions at this level will help to inhibit initiation of stress corrosion cracking of austenitic stainless steel components.

NRC Report NUREG-1465, "Accident Source Terms for Light-Water Nuclear Power Plants,"states that the iodine entering the containment is at least 95% cesium iodide (CsI) with theremaining 5% as elemental and organic iodide plus hydriodic acid, with not less than 1% of each as iodine and hydriodic acid. In order to prevent release of elemental iodine to the containment atmosphere after a LOCA, the sump pH has to be maintained equal or higher

than 7. 2.3.3Operator ActionsThe NRC staff reviewed the operator manual actions using guidance contained in NRCInformation Notice 97-78, "Crediting of Operator Actions in Place of Automatic Actions and Modifications of Operator Actions, Including Response Times," ANSI/ANS 58.8, "Time Response Design Criteria for Safety-Related Operator Actions," and NUREG-0800, "StandardReview Plan, Chapter 18.0, Human Factors Engineering," (Revision 1, 2004).

3.0 TECHNICAL EVALUATION

3.1 LBLOCA

AnalysesIn its April 29, 2005, application (Reference 1), the licensee stated that "Both Ginna LLC and itsanalysis vendor (Westinghouse) have ongoing processes which ensure that the values and ranges of the Best Estimate Large Break LOCA analysis inputs for peak cladding temperature and oxidation-sensitive parameters bound the ranges and values of the as-operated plant parameters." The NRC staff finds that this statement, along with the generic acceptance ofASTRUM, provides assurance that ASTRUM and its LBLOCA analyses apply to Ginna operated at the proposed uprated power. In its application, the licensee provided the results for the Ginna BE LBLOCA analyses at 1811 MWt (about 119 percent of the current licensed power of 1520 MWt) performed in accordance with the ASTRUM methodology. The licensee's results for the calculated peak cladding temperatures (PCTs), the maximum cladding oxidation (local), and the maximum core-wide cladding oxidation are provided in the following table along with the acceptance criteria of 10 CFR 50.46(b). TABLE 1: LBLOCA ANALYSIS RESULTSParameterASTRUM 422V+ ResultsASTRUMOFA Results10 CFR 50.46 LimitsLimiting BreakSize/LocationDEG/PD*DEG/PDN/ACladding MaterialZirloZircaloy(Cylindrical) Zircaloy or ZirloPeak Clad Temperature1870 F1814 F2200 F (10 CFR50.46(b)(1))Maximum Local Oxidation 3.4 % 2.5 %17.0% (10 CFR 50.46(b)(2))

Maximum Total Core-WideOxidation (All Fuel) 0.30 %0.30 %1.0% (10 CFR 50.46(b)(3))* DEG/PD is a double-ended guillotine break at the reactor coolant pump discharge.In its analyses, the licensee also addressed the concern that zircaloy fuel may have pre-existingoxidation that must be considered in its LOCA analyses. In its response to an NRC staff'srequest for additional information, the licensee indicated that it considered that the zircaloy cladfuel has both pre-existing oxidation and oxidation resulting from the LOCA (pre- and post-LOCAoxidation both on the inside and outside cladding surfaces). The licensee also noted that thefuel with the highest LOCA oxidation will likely not be the same fuel that has the highest pre-LOCA oxidation. The licensee indicated that when the calculated pre-LOCA oxidation wasfactored into the licensee's BE LBLOCA analyses for the zircaloy clad fuel, consistent with the Westinghouse ASTRUM methodology, that even during a fuel pin's final cycle in the core thesum of the calculated pre- and post-LOCA oxidation was sufficiently small that the total localoxidation remained less than the 17% acceptance criterion of 10 CFR 50.46(b)(2) as notedabove. The NRC staff finds this appropriately addressed the issue with pre-LOCA oxidation.The concern with core-wide oxidation relates to the amount of hydrogen generated during aLOCA. Because hydrogen that may have been generated pre-LOCA (during normal operation)will be removed from the reactor coolant system throughout the operati ng cycl e, the NRC staffnoted that pre-existing oxidation does not contribute to the amount of hydrogen generated post-LOCA, and therefore, it does not need to be addressed when determining whether the calculated total core-wide oxidation meets the 1.0% criterion of 10 CFR 50.46(b)(3).As discussed previously, the licensee had Westinghouse conduct the BE LBLOCA analyses forGinna at about 119% of the current licensed power level of 1520 MWt using an NRC-approvedWestinghouse methodology (ASTRUM). The NRC staff concludes that the results of theseanalyses (see Table 1) demonstrate compliance with 10 CFR 50.46(b)(1) through (b)(3) for licensed power levels of up to 1775 MWt. Meeting these criteria provides reasonable assurance that at the current licensed power level, the Ginna core will be amenable to coolingas required by 10 CFR 50.46(b)(4). The capability of Ginna to satisfy the long-term coolingrequirements of 10 CFR 50.46(b)(5) will be addressed in SE Section 3.3. LBLOCA ConclusionsBased on its review as discussed above, the NRC staff concludes that the WestinghouseASTRUM methodology, as described in WCAP-16009-P-A, is acceptable for use at Ginna in demonstrating compliance with the requirements of 10 CFR 50.46(b). The NRC staff'sconclusion is based on the assumed (uprated) core power up to 1775 MWt (plus 2.0% margin for measurement uncertainty, i.e., 1811 MWt).The NRC staff's review of the acceptability of the ASTRUM methodology for Ginna focused onassuring that the Ginna specific input parameters or bounding values and ranges (where appropriate) were used to conduct the analyses, that the analyses were conducted within the conditions and limitations of the NRC-approved Westinghouse ASTRUM methodology, and thatthe results satisfied the requirement of 10 CFR 50.46(b) based on a licensed power level of up to 1775 MWt. This SE also documents the NRC staff's review and acceptance of the Westinghouse ASTRUMBE LBLOCA analysis methodology for application to Ginna, and of the LBLOCA analyses discussed above, which were performed with the ASTRUM methodology for reference in the EPU of Ginna.3.2 SBLOCA Analyses 3.2.1Method of Staff Review The purpose of the NRC staff's review is to evaluate the licensee's assessment of the impact ofthe proposed EPU on DBA analyses. The NRC staff evaluated the SBLOCA analyses and post-LOCA long-term cooling analyses.The evaluation also included an audit of Westinghouse calculations pertaining to SBLOCA and post-LOCA long-term cooling, upon which certain accident analyses, presented in the application, were based. The NRC staff performed independent calculations using theRELAP5/MOD3 code to investigate a spectrum of SBLOCAs, as well as the full range of break sizes to assess the timing for boric acid precipitation for both large and small breaks.In areas where the licensee and its contractors used NRC-approved methods in performinganalyses, the NRC staff reviewed relevant material to assure that the licensee/contractor usedthe methods consistent with the limitations and restrictions placed on the methods. In addition,the NRC staff considered the effects of the changes in plant operating conditions on the use ofthese methods to assure that the methods were appropriate for use at the proposed EPUconditions. For these analyses, the licensee provided the statement (Reference 1) that:"Constellation Generation Group and Westinghouse have ongoing processes which assure that the values and ranges of the small break LOCA analysis inputs for peak cladding temperature-sensitive parameters conservatively bound the values and ranges of the as-operated plant for those parameters." The staff finds that this statement, along with the generic acceptance ofNOTRUMP, provides assurance that NOTRUMP and its use in SBLOCA analyses apply to Ginna at the proposed EPU conditions. 3.2.2Evaluation The NRC staff's evaluation consisted of reviewing the results of the licensee's analyses of theSBLOCA spectrum performed at 1811 MWt and a peak linear heat generation rate of 17.5 kw/ft. The NRC staff also reviewed the results of the licensee's post-LOCA long-termcooling analyses to show that the plant's emergency operating procedures (EOPs) could properly deal with and control the build-up of boric acid in the RCS following both LBLOCA and SBLOCAs. These two areas of review are discussed separately below. SBLOCA will bediscussed first.3.2.3SBLOCA Short-Term Behavior and Termination of HHSI Flow The licensee's April 29, 2005, application for SBLOCAs included analysis of the 1.5, 2, and 3inch diameter breaks in the cold leg at the reactor coolant pump discharge leg. The worst break in the licensee's analyses was found to be the 2-inch break with a PCT of 1167 F. TheNRC staff requested additional information from the licensee about the limited nature of thebreak spectrum and requested analyses of additional breaks, particularly those toward the larger end of the small-break spectrum. The larger breaks were of concern because the Ginna plant design requires the operators to terminate HHSI flow when re-aligning injection from the RWST to the containment sump to begin the recirculation phase of LOCA mitigation. Thelicensee's analyses showed a rapid decrease in two-phase level above the top of the core during realignment for the 2 and 3-inch breaks. However, because the two-phase level was well above the top of the active core, core uncovery did not occur. Analysis of these smaller break sizes suggested to the NRC staff that analyses of the larger breaks would be necessaryto show that breaks with potentially less inventory above the top of the core would also notuncover during the realignment period.The licensee assumed the alignment from the RWST to the containment sump could beperformed within about 10 minutes or 600 seconds. In responding to the questions from the NRC staff, the licensee investigated a larger range of break sizes and provided the results ofthe 4, 5, 6, 8.75, and 9.75 inch diameter breaks. The analysis of these break sizes showed thatthe PCT for these breaks remained below 1200 F, due to the high pressure accumulators(714.7 psia) and the high capacity HHSI pumps. In the Westinghouse NOTRUMP (Reference

6) analyses of the 6, 8.75, and 9.75-inch breaks, the results showed that the two-phase levelreceded to very near the top of the core during the 600-second interruption for realignment,then quickly recovered to the hot leg elevation upon re-initiation of HHSI flow. These analyses were performed assuming the break was located on the bottom of the discharge leg. TABLE 2: SBLOCA ANALYSIS RESULTS (LICENSEE ANALYSIS)ParameterNOTRUMP 422V+ ResultsNOTRUMPOFA Results10 CFR 50.46 LimitsLimiting Break Size/Location2-inch2-inchN/ACladding MaterialZirloZircaloy(Cylindrical) Zircaloy or Zirlo Peak Clad Temperature1167 F1167 F2200 F (10 CFR50.46(b)(1))Maximum Local OxidationNegligible* Negligible* 17.0% (10 CFR 50.46(b)(2))

Maximum Total Core-Wide Oxidation (All Fuel) 0.30 %0.30 %1.0% (10 CFR 50.46(b)(3))*Does not include pre-LOCA oxidation, which is expected to be small.The NRC staff performed independent calculations to assess the performance of the GinnaNSSS using the RELAP5/MOD3 code. The core power level was assumed to be 1811 MWt, with the hot rod at the peak linear heat generation rate of 17.5 kw/ft. The model included 24 axial cells to track the two-phase level in the core, which also included a hot bundle parallel channel containing the hot rod and the same level of axial detail. The top skewed power shapeused in the licensee's NOTRUMP (Reference 9) analyses was also input to the RELAP5/MOD3 code. Both reactor coolant loops in the NRC staff's RELAP5/MOD3 model were representedexplicitly in the nodalization of the Ginna NSSS. In the NRC staff's analyses, the ECCS wasalso modeled as well as the steam generator atmospheric dump valves (ADVs) and pressurizer power-operated relief valves (PORVs) to assess the plant cooldown capabilities and limitations. Since the licensee's analysis was conducted consistent with its licensing basis and the NRC-approved methodology, and there is significant margin to the 2200 F PCT acceptance criteriaof 10 CFR 50.46(b)(1) based on the NRC staff's independent analysis, the NRC staff concl udesthat the short-term plant response during termination of HHSI flow following a SBLOCAs at EPU conditions is acceptable.3.2.4Breaks on the Top of the Discharge Leg In its independent calculations, the NRC staff evaluated breaks located on the top of thedischarge piping. Uncovery for these breaks is faster because, with the break located on the top of the discharge leg, loop seal clearing does not occur. The filled loop seals during theLOCA increases the steam pressure and decreases the two-phase level in the upper plenum so that there is less inventory above the top of the core relative to the case with the break at thebottom of the discharge leg. With the breaks in the bottom, the broken loop seal clears of liquid and allows more inventory to accumulate in the upper plenum prior to the realignment interruption. NRC staff calculations showed that for breaks on the top of the discharge leg in the range of 2 to 6 inches in diameter, core uncovery could result if the realignment required more than 15 minutes. As such, the licensee's ability to complete the realignment within the 10 minutes assumed in its analyses is extremely important to provide reasonable assurance that the plant response to SBLOCAs meets the acceptance criteria of 10 CFR 50.46(b). The NRC staff's independent calculations also showed that breaks located on the top of thedischarge leg did not produce more limiting PCTs than the 2-inch break identified as the limiting break by the licensee. Breaks located on the top of the pipe have the potential to be morelimiting for plants with deep loop seals (i.e. when the bottom elevation of the loop seal is well below the top elevation of the core), since the steam pressure in the upper plenum during theSBLOCA is higher and depresses the two-phase level into the core. The NRC staff also notes that the 2-inch break is probably not the worst small break becauseanalysis of integer break sizes produces too coarse of a break spectrum. Staff experience has shown that break sizes intermediate to the integer sizes (for example, break sizes between 2and 3 inches, and between 3 and 4 inches) can result in PCT increases by as much as 150 F. However, the NRC staff concludes that, since the SBLOCA PCTs are very low due to the highcapacity of the HHSI pump relative to the core power level (which sets the core steaming rate during the event) and the high pressure of the accumulators (i.e. 714.7 psia), further analyses of breaks between 2 and 3 inches and 3 and 4 inches is not warranted to support the use of NOTRUMP as the Ginna SBLOCA methodology.3.2.5SBLOCA Conclusion Based on the appropriate application by the licensee of NRC-approved methodologies toanalyze Ginna's response to SBLOCAs and the NRC staff's independent analyses, the NRCstaff concludes that operation of Ginna at EPU conditions is acceptable in being able to mitigate the consequences of SBLOCAs. Therefore, the NRC staff concludes it has reasonableassurance that for SBLOCAs the acceptance criteria of 10 CFR 50.46(b)(1), (2), and (3) related to PCT, local oxidation, and hydrogen generation, respectively, are satisfied for Ginna at EPU conditions.3.3Post-LOCA Long-Term Cooling3.3.1Large-Break Behavior The NRC staff performed assessments of the timing for boric acid precipitation followingLBLOCAs using the staff's models developed for other plant power uprate reviews. NRC staffcalculations using these models showed that without a core flushing flow, precipitation canoccur in 4.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> compared to the 6.2-hours time to precipitation computed by the licensee. The NRC staff utilized the same boundary conditions as the licensee and included:-the mixing volume includes 1/2 of the lower plenum, the core, and the portion of theupper plenum below the bottom elevation of the hot legs.-the boron precipitation limit is assumed to be 29.27 weight percent (wt%) at 14.7 psia.

-the decay heat curve uses the 1971 American Nuclear Society (ANS) Standard with a1.2 multiplier.-mixing into the lower plenum does not begin until the core liquid density, with boric acid,exceeds the density of the water in the lower plenum at the RWST temperature of

120 F. Mixing does not begin in the lower plenum until the concentration in the corereaches 12.3 wt% boric acid. The differences in precipitation timing are due to the licensee's assumption that the boric acidbuild-up does not begin until 24 minutes into the LOCA. NRC staff calculations showed thatwith the 24-minute delay, the 29.27 wt% boron precipitation limit would not be achieved untilabout 5.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, which is reasonably close to the licensee's time of 6.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The NRC staff questioned the delay and requested further analysis and justification from the licensee. In response to the NRC staff's questions and concerns, the licensee performed aWCOBRA/TRAC analysis of the LBLOCA, with 10 CFR Part 50, Appendix K "type assumptions" and showed that within 300 seconds following opening of the break there is sufficient flushing flow to terminate the build-up of boric acid in the core. In fact, at 300 seconds, the HHSI flow into the RCS exceeded the boil-off in the core by 20 lbs/sec. At 300 seconds, the boric acid concentration is about 6.4 wt%. The large flushing flow, which would continue to increase over the first 24 minutes, would reduce the boric acid concentration to very near the source concentration. It is important to note that the limiting large break in this evaluation for Ginna is a hot-leg break.This is the worst break for boric acid precipitation because HHSI is terminated upon depletion of the RWST inventory, which occurs at about 24 minutes into the event for an LBLOCA. TheHHSI pumps must be turned off and re-aligned to take suction from the containment sump to start the re-circulation phase of the LOCA mitigation. It should be noted that the Ginna plant isunique in that the design does not enable the operators to switch the cold side injection tosimultaneous hot and cold side injection. Rather, with the upper plenum injection systemdesign, it must be shown that the RCS pressure can be reduced to a value below 140 psia to enable the RHR low pressure injection to provide water to the upper plenum, simultaneouslywith the HHSI injecting water into the cold legs. Since HHSI is terminated upon drainage of the RWST, analyses of the precipitation timing must be performed to identify the time frame within which HHSI must be re-instituted to flush the boric acid from the system.The operators must realign HHSI prior to the boron precipitation limit being exceeded. ForGinna, this switch time is set at 5.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after depletion of the RWST, or just before the 6.2-hours precipitation time calculated by the licensee. At 300 seconds, the COBRA/TRACcalculation shows the liquid flow out the break to be in excess of 50 lbs/sec, with an HHSI cold leg injection rate of about 80 lb/sec. The NRC staff considers this to be a sufficient flushingflow to reduce the initial build-up and reduce the concentrations to the source concentration prior to termination of HHSI at 24 minutes. It is noted that cold leg breaks are not limiting forthe Ginna NSSS since the lower pressure injection into the upper plenum would provide a flushing flow once RCS pressure decreased below 140 psia. The NRC staff concurs that the LBLOCA analysis for boric acid precipitation timing providessufficient time for the operators to realign HHSI to control the boric acid build-up for all large breaks that depressurize below the shutoff head of the RHR low-pressure safety injectionpumps. Delaying the time to initiate the build-up to 24 minutes following the initiation of the break is justified based on the WCOBRA/TRAC LBLOCA calculation. Smaller breaks that do not depressurize below the shutoff head of the low-pressure pump require additional operator actions to control the boric acid build-up and prevent precipitation. Small breaks and the attendant operator actions are discussed below. 3.3.2Small-Break Behavior In its application, the licensee did not initially provide sufficient information nor analyses todemonstrate boric acid could be controlled following SBLOCAs because the RCS pressure could remain above the shutoff head of the RHR low-pressure safety injection pump for manyhours. The NRC staff issued several requests for additional information that discussed theneed for analyses of the entire small-break spectrum with identification of all the operatoractions and precautions needed to successfully accomplish this function. Since RCS pressure remains above 140 psia for hours for certain SBLOCAs, the NRC staff required analysis of thebreak spectrum to show that the plant could be cooled down below the shutoff head of the RHRpump prior to reaching the boron precipitation limit. For the very small breaks, where cooldown to these low pressures may be difficult, the analysis must show the RCS refills and dispersesthe boric acid throughout the RCS, or another approach to preclude boron precipitation needed to be identified and justified. The NRC staff also expressed concerns for the need to updatethe EOPs, since the EOPs did not provide the timing for the operator actions to use theequipment necessary for cooling down the RCS to initiate RHR low-pressure injection to controlboric acid following SBLOCAs. In response to the NRC staff concerns and need for additionaljustification and analysis for small breaks, the licensee performed analyses of the break spectrum to demonstrate boric acid can be controlled for all break sizes. These results can be summarized in the following manner:-For breaks of 1.0 ft 2 down to the 4-inch diameter break, analyses show that precipitationwill not occur before 6.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and reinstating high pressure injection at 5.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> willcontrol the boric acid buildup and preclude boron precipitation from occurring.-For breaks of 2.0 inch in diameter down to 1.0 inch in diameter, analyses show thatinitiating a cooldown with the ADVs no later than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> into the event will reduce RCSpressure below the shut-off head of the RHR low head safety injection pumps prior toboron precipitation occurring.-For breaks less than 1.0 inch in diameter, analyses show that single-phase naturalcirculation will disperse the boric acid throughout the RCS, reducing the concentration inthe vessel to very low values prior to reaching the RHR cut-in pressure of 140 psia.3.3.3 Enhanced Boron Precipitation Controls Because operator actions are required to control boric acid precipitation following all LOCAs,changes were recommended to the plant EOPs to assure boric acid is controlled and precipitation is prevented during a LOCA. The NRC staff requested that the licensee includethe key operator actions to initiate a timely cooldown of the RCS to assure actuation of the RHR low-pressure safety injection pumps which, in combination with the HHSI pumps, provide a flushing flow through the core for all break sizes that do not refill with ECCS injection water.With a loss of offsite power, it is necessary to initiate a cooldown with the steam generatorADVs. The NRC staff raised a question about boron precipitation impacts, should one of theADVs fail to open. Staff calculations also showed that the RCS can boil for extended periodsduring the cooldown following an SBLOCA. In these situations, the NRC staff requested theGinna EOPs be modified to alert the operators not to suddenly cool the RCS should boilingextend for many hours. As a result of NRC staff calculations for SBLOCAs, the NRC staff raised questions regardingthe failure of an ADV to open and the possible need for the PORVs to be opened to assure a timely cooldown. This condition is not part of the current licensing basis for Ginna. The NRC staff's RELAP5/MOD3 calculations showed that the RCS pressure cannot be reduced below about 120 psia (i.e. the pressure required for sufficient RHR low-pressure injection flow to beginflushing the core) for at least 8.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> when 1 ADVs and 2 PORVs are opened after 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> following the opening of a 0.0125 ft 2 cold leg break. The NRC staff calculations suggest that with the RCS boiling for more than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, largeamounts of boric acid (i.e. in excess of the 29.3 wt% boron precipitation limit at 14.7 psia) can accumulate in the vessel. While the RCS pressure remains above 120 psia, the RCS temperature is sufficiently high to keep the boric acid in solution. As such, the NRC staffexpressed concerns that should the operators regain power to more rapidly depressurize the RCS, boron precipitation could inadvertently occur. Based on the staff's questions and discussions with the licensee, the licensee agreed to enhance its EOPs to provide guidance to caution the operators not to suddenly depressurize the RCS should there be limited cooldowncapability followed by a later restoration of depressurization equipment. The licensee willmodify the EOPs to instruct the operators not to exceed the 100 F/hr cooldown limit followingan SBLOCA. The EOPs will also be updated to alert the operators to use the PORVs to cooldown should one of the ADVs fail to open. While the NRC staff finds that one ADV may notdepressurize the RCS to 120 psia for small breaks for many hours, as noted previously, the high RCS coolant temperature will maintain the boric acid in solution. The proposedenhancements to the EOPs provide the NRC staff with reasonable assurance that there areadequate controls in place that will prevent the operators from causing an inadvertentprecipitation by limiting the depressurization rate during the long-term cooling phase of SBLOCA mitigation in the event boiling persists for extended periods with the RCS pressure above 120 psia. 3.3.4Long-Term Cooling Conclusion The NRC staff considers the analyses and operator actions to be an acceptable approach forcontrolling boric acid precipitation for the Ginna NSSS at the proposed EPU operating conditions. Based on its review, the NRC staff finds the analyses, operator actions, and EOPchanges to facilitate the successful control of boric acid following all LOCAs providesreasonable assurance that the long-term cooling requirements of 10 CFR 50.46(b)(5) are satisfied for Ginna under EPU conditions.3.4SummaryThe NRC staff reviewed the Westinghouse SBLOCA and post-LOCA long-term coolinganalyses for application to the Ginna NSSS operating under the proposed EPU conditions. The NRC staff's review confirmed that the licensee and its vendor have processes to assure that theGinna-specific input parameter values and operator action times (where appropriate) that were used to conduct the analyses will assure that 10 CFR 50.46 limits are not exceeded, and long-term cooling can be assured for all break sizes by providing the means to remove decay heat for extended periods, while also preventing the precipitation of boric acid for all break sizes and locations. Furthermore, the NRC staff finds that the analyses were conducted within theconditions and limitations of the NRC-approved Westinghouse NOTRUMP SBLOCAmethodology, and that the results satisfied the requirements of 10 CFR 50.46(b), based on the proposed EPU conditions. The staff notes that the procedures for assuring boric acid controlfor all breaks for the Ginna NSSS are unique to this system and finds the vendor and licenseeapproach to be a conservative and acceptable approach for demonstrating core cooling during the long term for all break sizes. 3.5 Accumulator and RWST Boron Concentration3.5.1 Containment Sump pH A variety of acids and bases are produced in containment after a LOCA. The pH value of thecontainment sump will depend on the concentration of these chemical species dissolved in thecontainment sump water. The following chemical species are introduced into the containment sump in a post-LOCA environment: hydriodic acid (HI), nitric acid (HNO 3), hydrochloric acid(HCl) and cesium hydroxide (CsOH). CsOH and HI enter the containment directly from thereactor coolant system (RCS). HCl is produced by radiolytic decomposition of cable jacketingand HNO 3 is synthesized in the radiation field existing in the containment. The resultantcontainment sump pH will depend on their relative concentrations and on the buffering action ofNaOH and boric acid.Maintaining sump water in an alkaline condition is needed for preventing dissolved radioactiveiodine from being released to the containment atmosphere during the recirculation containmentspray injection. Most of the iodine leaves the damaged core in an ionic form which is readily dissolved in the sump water. However, in an acidic environment, some of it becomes converted into elemental form which is much less soluble, causing re-evolution of iodine to the containment atmosphere. Per NUREG-1465, "Accident Source Terms for Light-Water NuclearPower Plants," the iodine entering the containment is at least 95% cesium iodide (CsI) with theremaining 5% as elemental and organic iodide plus hydriodic acid, with not less than 1% of each as iodine and hydriodic acid. In order to prevent release of elemental iodine to the containment atmosphere after a LOCA, the sump pH has to be maintained equal or higher

than 7.After a LOCA, the containment sump is mostly filled with water coming from the systemscontaining boric acid: RWST, safety system injection accumulators, and the RCS. This in effectwill cause the sump water pH to become acidic. In order to keep the pH above 7, Ginna usesNaOH as a buffer to maintain the pH above 7 for the 30-day period after LOCA.The licensee utilized the BORDER Code to determine the containment sump pH 30 days after aLOCA. The calculation used the water mass and boron concentrations from the RWST, the accumulators and the RCS, as well as the NaOH spray additive tank volume and concentration.

The code results are calculated in terms of allowable spray additive tank NaOH concentrations such that the sump pH limits, given as inputs, are met. The licensee did not consider the formation of HI, HCl, HNO 3, and CsOH in its calculation. Although HI, HCl, and HNO 3 arestrong acids, the contribution that these acids could have in the pH is minimal due to thebuffering action of NaOH and the higher concentration of boron found in boric acid. In addition, the licensee did not consider the addition of CsOH into containment, which would increase pH, therefore adding conservatism to the calculation.The NRC staff reviewed the licensee's methodology, assumptions, and performed handcalculations to verify the resulting pH value after 30 days. In its computer code calculations, the licensee used the minimum and maximum volumes and concentrations from the borationsources and the spray additive tank. On the basis of these inputs and its computer calcuations,the licensee stated that the minimum sump pH would be 7.8. In addition, the NRC staffperformed an independent verification that also demonstrated the containment sump pH wouldremain above 7 for at least 30 days.3.5.2 Conclusion After an accident, the pH of the containment sump water is determined by the amounts of acidicand basic chemical materials either released from the damaged core or generated in containment and subsequently dissolved in the sump water. It is important to control this pH because if it falls below 7, radioactive iodine could be released to the containment atmosphere.

The addition of a buffering agent, such as NaOH, will keep the water pH above 7, thereforepreventing the iodine from being released. The licensee's analysis has indicated that containment sump water will remain greater than 7 for at least 30 days. The NRC staffreviewed the licensee's methodology for determining pH and performed an independent evaluation of the licensee's calculations. Based on its evaluation, the staff concludes that thelicensee's proposed actions will maintain the sump water pH greater than 7 for 30 daysfollowing a LOCA, thus preventing the release of radioactive iodine into the containmentatmosphere.3.6 Operator ActionsWith respect to boric acid precipitation, the Ginna ECCS design incorporates upper plenuminjection (UPI). The low-head safety injection pumps (RHR pumps) deliver flow directly to theupper plenum. For this reason, the hot-leg switchover procedure that is applied to the typical three-loop and four-loop Westinghouse designs to ensure long-term core cooling is not applied to Ginna. A safety injection (SI) signal starts both HHSI pumps and low-head RHR pumps. When RCS pressure decreases below the low-head RHR injection pressure (140 psia),simultaneous hot (UPI) and cold side (SI) injection will occur. Upon entering the sumprecirculation phase, operators are instructed to establish recirculation flow using the RHRpumps, which will maintain UPI, and terminate flow from the HHSI pumps. After a period oftime, operators will be instructed to restart the HHSI pumps to reestablish simultaneous coldside and hot side (UPI) injection to provide long-term core cooling for all LOCA scenarios.3.6.1 Operator Actions Related to Post-LOCA for Large Breaks For large breaks in the cold leg, the licensee stated that boric acid precipitation cannot occursince the RCS will depressurize quickly and UPI will provide flushing flow through the core. Nooperator actions are required for this scenario, with the exception of identifying the LOCA using the EOPs and manually realigning RHR pump suction to the containment sump whenrecirculation is established.For large breaks in the hot leg, the licensee's calculations show that the boric acid solution willnot approach the solubility limit for atmospheric pressure conditions until approximately 6.2hours after the termination of SI to the cold leg. Under EPU conditions, EOP ES-1.3, "Transfer to Cold Leg Recirculation," will be revised by the licensee to instruct operators to reestablishcold leg SI no later than 5.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after the initial termination of HHSI to prevent boric acid precipitation. The HHSI pumps may be turned off as early as 24 minutes after the LBLOCA because of the depletion of the RWST inventory at that time. The RHR pumps are immediatelymanually realigned to take suction from the containment sump to initiate recirculation using the RHR pumps. The re-establishment of cold-leg injection with the HHSI pumps will take place 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> later, after the containment sump temperature is reduced, to maintain adequate NPSH for the HHSI pumps.The HHSI must be realigned to take suction from the containment sump (via the RHR pumps)and recirculate to the cold leg no later than 5.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after the initial HHSI termination to avoid boric acid precipitation. The realignment involves 3 pairs of valves to be manipulated, which include RWST outlet valves, SI pump recirculation valves, and RHR to SI pump suction valves. The valves are all operated remotely from the control room and take less than 1 minute to operate per set. These actions are the same as the existing steps and will be taken in advanceduring the 4-hour waiting period to lower sump temperature. After the valve realignment, the HHSI pump is then started to begin HHSI injection into the cold leg. The licensee will revise theES-1.3 procedure and provide training to emphasize that the HHSI realignment must take place no later than 5.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after the initial HHSI termination to avoid boric acid precipitation. 3.6.2 Operator Actions Related to Post-LOCA for Small Breaks For small breaks in the cold leg, RCS pressure will stabilize above the UPI initiation pressureand the boric acid concentration in the core is expected. EOP ES-1.2, "Post-LOCA Cooldownand Depressurization," directs the operators in this scenario to depressurize the RCS using the condenser dump valves. In the event that the condenser steam dumps are unavailable for cooldown, the ADVs will be used with a limit on the cooldown not to exc eed 100 F/hr. Whenthe RCS is depressurized through operator action to below 140 psia, UPI using the RHR (low-head SI) pumps will initiate and this will provide immediate core flushing flow. Based upon thelicensee's bounding calculations for LBLOCA, if UPI is initiated within 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, boric acid precipitation is precluded even for RCS at atmospheric pressure for an SBLOCA. For small breaks in the hot leg, RCS pressure will stabilize above the UPI initiation pressure. However, the boric acid concentration in the core will not increase until the cold leg HHSI is terminated. Operators are directed to depressurize the RCS using the mechanisms describedin the paragraph above, maintain UPI using the RHR pumps on recirculation, and terminate theHHSI. Once HHSI to the cold leg is terminated, this scenario is bounded by the large hot leg break scenario where cold leg HHSI will be reestablished no later than 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. This wouldinvolve the operator realigning the HHSI for cold leg injection as described in the LBLOCA scenario. The ES-1.2 procedure will direct the operators to use the revised ES-1.3 proceduredescribed for LBLOCA in order to immediately realign the HHSI pumps to the suction of the containment sump for cold leg injection after the RWST inventory is depleted. 3.6.3 Evaluation Post-LOCA Operator Actions and EOP ChangesThe operator action to immediately realign HHSI to the containment sump (via the RHR pumps)is performed for both LBLOCA and SBLOCA scenarios when the RWST inventory is depleted.

The licensee has determined from previously timed simulator scenarios that the operator action to realign the HHSI pumps to take suction from the containment sump can be accomplished within 10 minutes, which includes 5 minutes for the actual operator action with an additional margin. The NRC staff requested additional information from the licensee regarding thevalidation of this time. The licensee responded that the 10-minute time frame will be validatedupon conclusion of the simulator upgrades, which are being made to reflect the planned EPU modifications. The validation of the operator action time to realign HHSI, as well as the times indicated in the licensee's July 7, 2005, application, as supplemented, for an EPU, will becompleted prior to startup from the fall 2006 refueling outage. The staff agrees that theoperator action time of 10 minutes is reasonable, with the condition that the licensee completesits time validation with the upgraded simulator as well as the appropriate operator training prior to the startup from the fall 2006 refueling outage.The NRC staff also requested additional details regarding the timeframe for initiation of thecooldown and depressurization step in ES-1.2 to ensure that the RCS pressure would be below 140 psia for UPI using the RHR pumps for SBLOCA scenarios. The staff also requestedinformation on how this operator action would be reflected in the existing EOPs to indicate the initiation or completion time of this operator action. The licensee responded that the cooldown and depressurization of the RCS must be initiated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of the accident to ensure that the RHR pumps can be used for UPI at below 140 psia for both small-break scenarios. Thelicensee also committed to adding a cautionary note in the ES-1.2 procedure to indicate that, if the RCS is not depressurized within 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> to the capability of the RHR system, that theplant cooldown rate shall not exceed 100 F/hr to assure that boron will not come out of thesolution in the event of additional equipment (e.g., condenser dump valves) were to become available. The licensee also provided additional details related to performing the cooldown using theADVs. Although one ADV is needed to initiate cooldown within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of the SBLOCA, the operators would have the ability to use the second ADV to perform the cooldown at 100 F/hr ifthe first ADV was unavailable. The licensee also indicated that an additional operator will bedirected to open the ADVs locally in case remote operation fails. In the unlikely event that both ADVs are unavailable, the operators would be able to use the pressurizer PORVs, as an alternative, to provide cooldown until one of the ADVs can be recovered. These subsequent steps will also be reflected in the ES-1.2 procedure prior to EPU implementation. The licenseehas also committed to validating, using the upgraded simulator, that the cooldown anddepressurization can be initiated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, as well as to provide the appropriate operator training reinforcing the importance of performing this operator action in SBLOCA scenarios.

The NRC staff agrees that the operator actions as well as the revisions to the ES-1.2 procedureare acceptable contingent upon the licensee completing the time validation of the cooldownstep. Operator training also must be completed prior to the startup from the fall 2006 refuelingoutage during which the EPU would be implemented. The NRC staff finds that the operator action times in the EOPs as described above arereasonable since the required operator actions have not changed and the licensee has indicated that prior training and testing has shown that the actions can be completed within thetime frames listed. Thus, the staff has reasonable assurance that the actions will beaccomplished as required and margin exists in the time period for action to be completed. In addition, the licensee has committed to complete a time validation with the upgraded simulator as well as the appropriate operator training prior to the startup from the fall 2006 refueling outage to further substantiate that the margins in the action times.3.7 TS Changes3.7.1 TS 3.5.1, "Accumulators" For TS SR 3.5.1.2 on borated water volume, the range was increased by lowering the minimumallowable water volume and increasing the maximum allowable water volume of each of the two Ginna cold-leg ECCS accumulators. SR 3.5.1.2 is consistent with the roles of the accumulators during LBLOCA and SBLOCA events. SRs 3.5.1.3 and 3.5.1.4 provide the ranges of pressure,and boron concentration, respectively, for the cold leg ECCS accumulators consistent with values used in the new LOCA methodologies. Therefore, the proposed TS change is acceptable. 3.7.2 TS 3.5.4, "Refueling Water Storage Tank (RWST)"

For TS SR 3.5.4, the minimum borated water volume of the RWST is specified in SR 3.5.4.1. SR 3.5.4.2 specifies the boron concentration range for the water in the RWST. These valuesare acceptable since they are consistent with the assumptions in the accident analyses.3.7.3 TS 5.6.5, "Core Operating Limits Report (COLR)"

The licensee proposed to revise TS 5.6.5.b as described below:2. WCAP-16009-P-A, "Realistic Large-Break LOCA Evaluation Methodology Usingthe Automated Statistical Treatment of Uncertainty (ASTRUM)," January 2005.7. WCAP-10054-P-A, Addendum 2, Revision 1, "Addendum to the WestinghouseSmall Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection Into the Broken Loop and COSI Condensation Model," July 1997.

(Methodology for LCO 3.2.1)8. WCAP-11145-P-A, "Westinghouse Small Break LOCA ECCS Evaluation ModelGeneric Study With the NOTRUMP Code," October 1986. (Methodology for LCO 3.2.1)9. WCAP-10079-P-A, "NOTRUMP -A Nodal Transient Small Break and GeneralNetwork Code," August 1985. (Methodology for LCO 3.2.1) 11. WCAP-14710-P-A, "1-D Heat Conduction Model for Annular Fuel Pellets," May,1998. (Methodology for LCO 3.2.1) The ASTRUM LBLOCA methodology (Reference 2) was found to apply to all Westinghouseand Combustion Engineering PWR designs in the NRC generic SE of the ASTRUMmethodology. Therefore, ASTRUM is acceptable for application to Ginna which is a PWR of Westinghouse design and inclusion in the Ginna TS and COLR.The NOTRUMP SBLOCA Methodology (References 7, 8, and 9) was found to apply to allWestinghouse designs in NRC generic SEs regarding the NOTRUMP methodology. Therefore,NOTRUMP is acceptable for application to Ginna, which is a PWR of Westinghouse design, and inclusion in the Ginna TS and COLR.WCAP-14710-P-A, "1-D Heat Conduction Model for Annular Fuel Pellets," was found to beapplicable and acceptable for all applications of Westinghouse fuel featuring annual fuel pellets using Westinghouse analytical models. Ginna is a plant of Westinghouse design proposing to use Westinghouse annular pellets and analytical models. Therefore, WCAP-14710-P-A is acceptable for application to Ginna, and inclusion in the Ginna TSs and COLR.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the New York State official was notified of theproposed issuance of the amendment. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facilitycomponent located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The NRC staff has determined that the amendment involves nosignificant increase in the amounts, and no significant change in the types, of any effluents thatmay be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been nopublic comment on such finding (70 FR 33219). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR51.22(b) no environmental impact statement or environmental assessment need be prepared inconnection with the issuance of the amendment.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above that (1) there isreasonable assurance that the health and safety of the public will not be endangered byoperation in the proposed manner, (2) such activities will be conducted in compliance with theCommission's regulations, and (3) the issuance of the amendment will not be inimical to thecommon defense and security or to the health and safety of the public.

7.0 REFERENCES

1.Ginna LLC letter, M. Korsnick to NRC, "License Amendment Request RegardingRevised Loss of Coolant Accident (LOCA) Analyses-Changes to Accumulator, Refuleing Water Storage (RWST), and Administrative Control Technical Specifications,"

April 29, 2005. (Agencywide Documents Access and Management System (ADAMS)

Accession No. ML05126039)2.Westinghouse Report WCAP-16009-P-A,"Realistic Large Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method(ASTRUM)," January 31, 2005. (ADAMS No. ML050910159)3.Ginna LLC letter, M. Korsnick to NRC, "

Subject:

R.E. Ginna Nuclear Power Plant,Docket No. 50-244, License Amendment Request Regarding Extended Power Uprate."

July 7, 2005. (ADAMS No. ML051950123)4.Ginna LLC letter, M. Korsnick to NRC, "Response to Requests for Additional InformationRegarding Topics Described by Letter Dated November 3, 2005," December 19, 2005.

(ADAMS No. ML0536101850)5.Ginna LLC letter, M. Korsnick to NRC, "

Subject:

R.E. Ginna Nuclear Power Plant,Docket No. 50-244, Response for Additional Information Regarding Topics Described in Meeting Minutes," Attachment 4, dated January 25, 2006. (ADAMS No. ML052310155)6.Westinghouse Report WCAP-9236, "NOTRUMP, A Nodal Transient Steam Generatorand General Network Code," P. E. Meyer and G. K. Frick, February 1978.7.Ginna LLC letter, M. Korsnick to NRC, "Supplemental Information Related to SmallBreak (SB) Loss of Coolant Accident (LOCA) and Post-LOCA Boric Acid Precipitation Analysis," August 15, 2005. (ADAMS No.ML052310155)8.Ginna LLC letter, M. Korsnick to NRC, "Supplemental Response to Requests forAdditional Information Regarding Topics Described by Letters Dated August 24, 2005 and October 28, 2005," January 11, 2006. (ADAMS No. ML060180262)9. Ginna LLC letter, M. Korsnick to NRC, "Response to Requests for Additional InformationRegarding Topics Described in Meeting Minutes," January 25, 2006. (ADAMS No.

ML0609604165)10. Ginna LLC letter, M. Korsnick to NRC, "License Amendment Request RegardingAdoption of Relaxed Axial Offset Control (RAOC)," April 29, 2005. (ADAMS No.

ML051300330) 11. Ginna LLC letter, M. Korsnick to NRC, "Response to Requests for Additional InformationRegarding Topics Discussed on Conference Calls for Extended Power Uprate (EPU),"

May 9, 2006. (ADAMS No. ML061350375)Principal Contributors: F. Orr L. Ward Y. Diaz-Castillo G. ArmstrongDate: May 31, 2006