ML061180353

License Amendment 96 Regarding Revised Loss-of-Coolant Accident Analyses

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

Text

May 31, 2006 Mrs. Mary G. Korsnick Vice 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 Operating License 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 Operating Limits 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 included in the Commission's biweekly Federal Register notice.

Sincerely,

/RA/

Patrick D. Milano, Sr. Project Manager Plant Licensing Branch I-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-244

Enclosures:

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

2. Safety Evaluation cc w/encls: See next page

May 31, 2006 Mrs. Mary G. Korsnick Vice 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 Operating License 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 Operating Limits 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 included in the Commission's biweekly Federal Register notice.

Sincerely,

/RA/

Patrick D. Milano, Sr. Project Manager Plant Licensing Branch I-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-244

Enclosures:

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

2. Safety Evaluation cc w/encls: See next page Accession Number: ML061180353 OFFICE LPLI-1\PM LPLI-1\LA SPWB\BC IOLB\BC CSGB\(A)BC OGC LPLI-1\BC NAME PMilano SLittle JNakoski NOKeefe EMurphy JMoore RLaufer DATE 05/31/06 05/31/06 05/03/06 05/17/06 02/02/06 05/23/06 05/31/06 Official Record Copy

DATED: May 31, 2006 AMENDMENT NO. 96 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-18 R.E. GINNA NUCLEAR POWER PLANT PUBLIC LPLI-1 R/F R. Laufer RidsNrrDorlLpla J. Nakoski RidsNrrDssSpwb F. Orr L. Ward E. Murphy RidsNrrDciCsgb Y. Diaz S. Little RidsNrrLASLittle P. Milano RidsNrrPMPMilano T. Boyce RidsNrrDirsItsb G. Hill (2)

OGC RidsOgcRp ACRS RidsAcrsAcnwMailCenter B. McDermott, RI RidsRgn1MailCenter cc: Plant Service list

R.E. Ginna Nuclear Power Plant cc:

Mr. Michael J. Wallace Ms. Thelma Wideman, Director President Wayne County Emergency Management R.E. Ginna Nuclear Power Plant, LLC Office c/o Constellation Energy Wayne County Emergency Operations 750 East Pratt Street Center Baltimore, MD 21202 7336 Route 31 Lyons, NY 14489 Mr. John M. Heffley Senior Vice President and Ms. Mary Louise Meisenzahl Chief Nuclear Officer Administrator, Monroe County Constellation Generation Group Office of Emergency Preparedness 1997 Annapolis Exchange Parkway 1190 Scottsville Road, Suite 200 Suite 500 Rochester, NY 14624 Annapolis, MD 21401 Mr. Paul Eddy Kenneth Kolaczyk, Sr. Resident Inspector New York State Department of R.E. Ginna Nuclear Power Plant Public Service U.S. Nuclear Regulatory Commission 3 Empire State Plaza, 10th Floor 1503 Lake Road Albany, NY 12223 Ontario, NY 14519 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. Peter R. Smith, President New York State Energy, Research, and Development Authority 17 Columbia Circle Albany, NY 12203-6399 Mr. Carey W. Fleming, Esquire Senior Counsel - Nuclear Generation Constellation Generation Group, LLC 750 East Pratt Street, 17th Floor Baltimore, MD 21202 Mr. Charles Donaldson, Esquire Assistant Attorney General New York Department of Law 120 Broadway New York, NY 10271

R.E. GINNA NUCLEAR POWER PLANT, LLC DOCKET NO. 50-244 R.E. GINNA NUCLEAR POWER PLANT AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 96 Renewed License No. DPR-18

1. The Nuclear Regulatory Commission (the Commission or the NRC) has found 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 the Act, and the rules and regulations of the Commission; C. There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated 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 Specifications The Technical Specifications contained in Appendix A, as revised through Amendment 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 be implemented prior to startup from the fall 2006 refueling outage. Implementation shall include revisions to plant procedures and the completion of operator training as described in the licensees April 29, 2005, application, as supplemented, and as discussed in the NRC staffs safety evaluation dated May 31, 2006.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Richard J. Laufer, Chief Plant Licensing Branch I-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and License Date of Issuance: May 31, 2006

ATTACHMENT TO LICENSE AMENDMENT NO. 96 RENEWED FACILITY OPERATING LICENSE NO. DPR-18 DOCKET NO. 50-244 Replace the following page of the Renewed Facility Operating License with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove Insert 3 3 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert 3.5.1-1 3.5.1-1 3.5.1-2 3.5.1-2 3.5.4-1 3.5.4-1 5.6-1 5.6-1 5.6-2 5.6-2 5.6-3 5.6-3 5.6-4 5.6-4 5.6-5 5.6-5

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 96 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-18 R.E. GINNA NUCLEAR POWER PLANT, INC.

R.E. GINNA NUCLEAR POWER PLANT DOCKET NO. 50-244

1.0 INTRODUCTION

By letter dated April 29, 2005, as supplemented on August 15 and December 9, 2005, and January 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 Nuclear Regulatory 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, letters provided additional information that clarified the application, did not expand the scope of the application 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 requesting an 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 Changes Because 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 at the 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 water volume 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 accumulator be verified to be > 2100 parts per million (ppm) and < 2600 ppm. The licensee proposes 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 to be > 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, Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment 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,"Westinghouse Small 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 - A Nodal 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-D Heat Conduction Model for Annular Fuel Pellets, May 1998. (Methodology for LCO 3.2.1) 2.2 Background The 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 licensees evaluation of the ECCS performance analyses for Ginna completed 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 containing Westinghouse 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 Company design, 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) flow delivered 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 is about 160 psia.

The proposed EPU steady-state power level of 1775 MWt (with analysis at 1811 MWt, which includes 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 1817 MWt. The SBLOCA and post-LOCA long-term cooling analyses conducted by the NRC staff were performed at an NSSS power level of 1817 MWt.

Implementation of an EPU requires re-analyzing the LOCA design-basis accident due to the changes 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 Analyses The LBLOCA and SBLOCA analyses are performed to demonstrate that the system design would 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 hydrogen generation. The NRC staff reviewed the analyses to assure that the analyses reflected suitable redundancy in components and features; and suitable interconnections, leak detection, isolation, and containment capabilities were available such that the safety functions could be accomplished, assuming a single failure, for LOCAs considering the availability of onsite and offsite 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 in assessing the acceptability of the Westinghouse ASTRUM and NOTRUMP methodologies for Ginna.

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.

2.3.2 Post-LOCA Containment Sump Conditions Ginnas Updated Final Safety Analysis Report (UFSAR) Section 6.1.2.1.4, Design Chemical Composition 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 Standard Review 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, after mixing 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 the remaining 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.

2.3.3 Operator Actions The NRC staff reviewed the operator manual actions using guidance contained in NRC Information 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, Standard Review Plan, Chapter 18.0, Human Factors Engineering, (Revision 1, 2004).

3.0 TECHNICAL EVALUATION

3.1 LBLOCA Analyses In its April 29, 2005, application (Reference 1), the licensee stated that Both Ginna LLC and its analysis 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 of ASTRUM, 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 licensees 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 RESULTS ASTRUM ASTRUM 10 CFR 50.46 Limits Parameter 422V+ Results OFA Results Limiting Break DEG/PD* DEG/PD N/A Size/Location Cladding Material Zirlo Zircaloy (Cylindrical) Zircaloy or Zirlo Peak Clad Temperature 1870 EF 1814 EF 2200 EF (10 CFR 50.46(b)(1))

Maximum Local Oxidation 3.4 % 2.5 % 17.0% (10 CFR 50.46(b)(2))

Maximum Total Core-Wide 0.30 % 0.30 % 1.0% (10 CFR 50.46(b)(3))

Oxidation (All Fuel)

• 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-existing oxidation that must be considered in its LOCA analyses. In its response to an NRC staffs request for additional information, the licensee indicated that it considered that the zircaloy clad fuel has both pre-existing oxidation and oxidation resulting from the LOCA (pre- and post-LOCA oxidation both on the inside and outside cladding surfaces). The licensee also noted that the fuel 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 was factored into the licensees BE LBLOCA analyses for the zircaloy clad fuel, consistent with the Westinghouse ASTRUM methodology, that even during a fuel pins final cycle in the core the sum of the calculated pre- and post-LOCA oxidation was sufficiently small that the total local oxidation remained less than the 17% acceptance criterion of 10 CFR 50.46(b)(2) as noted above. 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 a LOCA. Because hydrogen that may have been generated pre-LOCA (during normal operation) will be removed from the reactor coolant system throughout the operating cycle, the NRC staff noted 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 for Ginna at about 119% of the current licensed power level of 1520 MWt using an NRC-approved Westinghouse methodology (ASTRUM). The NRC staff concludes that the results of these analyses (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 cooling as required by 10 CFR 50.46(b)(4). The capability of Ginna to satisfy the long-term cooling requirements of 10 CFR 50.46(b)(5) will be addressed in SE Section 3.3.

LBLOCA Conclusions Based on its review as discussed above, the NRC staff concludes that the Westinghouse ASTRUM 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 staffs conclusion 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 staffs review of the acceptability of the ASTRUM methodology for Ginna focused on assuring 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 that the 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 staffs review and acceptance of the Westinghouse ASTRUM BE 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 3.2.1 Method of Staff Review The purpose of the NRC staffs review is to evaluate the licensees assessment of the impact of the 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 the RELAP5/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 performing analyses, the NRC staff reviewed relevant material to assure that the licensee/contractor used the 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 of these methods to assure that the methods were appropriate for use at the proposed EPU conditions. 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 of NOTRUMP, provides assurance that NOTRUMP and its use in SBLOCA analyses apply to Ginna at the proposed EPU conditions.

3.2.2 Evaluation

The NRC staffs evaluation consisted of reviewing the results of the licensees analyses of the SBLOCA 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 licensees post-LOCA long-term cooling analyses to show that the plants 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 be discussed first.

3.2.3 SBLOCA Short-Term Behavior and Termination of HHSI Flow The licensees April 29, 2005, application for SBLOCAs included analysis of the 1.5, 2, and 3 inch diameter breaks in the cold leg at the reactor coolant pump discharge leg. The worst break in the licensees analyses was found to be the 2-inch break with a PCT of 1167 EF. The NRC staff requested additional information from the licensee about the limited nature of the break 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. The licensees 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 necessary to show that breaks with potentially less inventory above the top of the core would also not uncover during the realignment period.

The licensee assumed the alignment from the RWST to the containment sump could be performed 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 of the 4, 5, 6, 8.75, and 9.75 inch diameter breaks. The analysis of these break sizes showed that the PCT for these breaks remained below 1200 EF, 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 level receded 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)

NOTRUMP NOTRUMP 10 CFR 50.46 Limits Parameter 422V+ Results OFA Results Limiting Break Size/Location 2-inch 2-inch N/A Cladding Material Zirlo Zircaloy (Cylindrical) Zircaloy or Zirlo Peak Clad Temperature 1167 EF 1167 EF 2200 EF (10 CFR 50.46(b)(1))

Maximum Local Oxidation Negligible* Negligible* 17.0% (10 CFR 50.46(b)(2))

Maximum Total Core-Wide 0.30 % 0.30 % 1.0% (10 CFR 50.46(b)(3))

Oxidation (All Fuel)

• Does not include pre-LOCA oxidation, which is expected to be small.

The NRC staff performed independent calculations to assess the performance of the Ginna NSSS 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 shape used in the licensees NOTRUMP (Reference 9) analyses was also input to the RELAP5/MOD3 code. Both reactor coolant loops in the NRC staffs RELAP5/MOD3 model were represented explicitly in the nodalization of the Ginna NSSS. In the NRC staffs analyses, the ECCS was also 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 licensees analysis was conducted consistent with its licensing basis and the NRC-approved methodology, and there is significant margin to the 2200 EF PCT acceptance criteria of 10 CFR 50.46(b)(1) based on the NRC staffs independent analysis, the NRC staff concludes that the short-term plant response during termination of HHSI flow following a SBLOCAs at EPU conditions is acceptable.

3.2.4 Breaks on the Top of the Discharge Leg In its independent calculations, the NRC staff evaluated breaks located on the top of the discharge 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 the LOCA 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 the bottom 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 licensees 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 staffs independent calculations also showed that breaks located on the top of the discharge 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 more limiting 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 the SBLOCA 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 because analysis 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 2 and 3 inches, and between 3 and 4 inches) can result in PCT increases by as much as 150 EF.

However, the NRC staff concludes that, since the SBLOCA PCTs are very low due to the high capacity 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.

3.2.5 SBLOCA Conclusion Based on the appropriate application by the licensee of NRC-approved methodologies to analyze Ginnas response to SBLOCAs and the NRC staffs independent analyses, the NRC staff 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 reasonable assurance 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.

3.3 Post-LOCA Long-Term Cooling 3.3.1 Large-Break Behavior The NRC staff performed assessments of the timing for boric acid precipitation following LBLOCAs using the staffs models developed for other plant power uprate reviews. NRC staff calculations using these models showed that without a core flushing flow, precipitation can occur 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 the upper 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 a 1.2 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 EF. Mixing does not begin in the lower plenum until the concentration in the core reaches 12.3 wt% boric acid.

The differences in precipitation timing are due to the licensees assumption that the boric acid build-up does not begin until 24 minutes into the LOCA. NRC staff calculations showed that with the 24-minute delay, the 29.27 wt% boron precipitation limit would not be achieved until about 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 licensees 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 staffs questions and concerns, the licensee performed a WCOBRA/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. The HHSI 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 is unique in that the design does not enable the operators to switch the cold side injection to simultaneous hot and cold side injection. Rather, with the upper plenum injection system design, 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, simultaneously with 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. For Ginna, 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/TRAC calculation 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 flushing flow 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 for the 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 provides sufficient 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 injection pumps. 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.

3.3.2 Small-Break Behavior

In its application, the licensee did not initially provide sufficient information nor analyses to demonstrate 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 many hours. The NRC staff issued several requests for additional information that discussed the need for analyses of the entire small-break spectrum with identification of all the operator actions 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 the break spectrum to show that the plant could be cooled down below the shutoff head of the RHR pump 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 disperses the 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 update the EOPs, since the EOPs did not provide the timing for the operator actions to use the equipment necessary for cooling down the RCS to initiate RHR low-pressure injection to control boric acid following SBLOCAs. In response to the NRC staff concerns and need for additional justification 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 ft2 down to the 4-inch diameter break, analyses show that precipitation will 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 /> will control 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 that initiating 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 RCS pressure below the shut-off head of the RHR low head safety injection pumps prior to boron precipitation occurring.

- For breaks less than 1.0 inch in diameter, analyses show that single-phase natural circulation will disperse the boric acid throughout the RCS, reducing the concentration in the 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 include the 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 generator ADVs. The NRC staff raised a question about boron precipitation impacts, should one of the ADVs fail to open. Staff calculations also showed that the RCS can boil for extended periods during the cooldown following an SBLOCA. In these situations, the NRC staff requested the Ginna EOPs be modified to alert the operators not to suddenly cool the RCS should boiling extend for many hours.

As a result of NRC staff calculations for SBLOCAs, the NRC staff raised questions regarding the 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 staffs 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 begin flushing 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 ft2 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 />, large amounts 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 staff expressed concerns that should the operators regain power to more rapidly depressurize the RCS, boron precipitation could inadvertently occur. Based on the staffs 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 cooldown capability followed by a later restoration of depressurization equipment. The licensee will modify the EOPs to instruct the operators not to exceed the 100 EF/hr cooldown limit following an SBLOCA. The EOPs will also be updated to alert the operators to use the PORVs to cool down should one of the ADVs fail to open. While the NRC staff finds that one ADV may not depressurize 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 proposed enhancements to the EOPs provide the NRC staff with reasonable assurance that there are adequate controls in place that will prevent the operators from causing an inadvertent precipitation 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.

3.3.4 Long-Term Cooling Conclusion The NRC staff considers the analyses and operator actions to be an acceptable approach for controlling 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 EOP changes to facilitate the successful control of boric acid following all LOCAs provides reasonable assurance that the long-term cooling requirements of 10 CFR 50.46(b)(5) are satisfied for Ginna under EPU conditions.

3.4 Summary The NRC staff reviewed the Westinghouse SBLOCA and post-LOCA long-term cooling analyses for application to the Ginna NSSS operating under the proposed EPU conditions. The NRC staffs review confirmed that the licensee and its vendor have processes to assure that the Ginna-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 the conditions and limitations of the NRC-approved Westinghouse NOTRUMP SBLOCA methodology, 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 control for all breaks for the Ginna NSSS are unique to this system and finds the vendor and licensee approach 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 Concentration 3.5.1 Containment Sump pH A variety of acids and bases are produced in containment after a LOCA. The pH value of the containment sump will depend on the concentration of these chemical species dissolved in the containment sump water. The following chemical species are introduced into the containment sump in a post-LOCA environment: hydriodic acid (HI), nitric acid (HNO3), hydrochloric acid (HCl) and cesium hydroxide (CsOH). CsOH and HI enter the containment directly from the reactor coolant system (RCS). HCl is produced by radiolytic decomposition of cable jacketing and HNO3 is synthesized in the radiation field existing in the containment. The resultant containment sump pH will depend on their relative concentrations and on the buffering action of NaOH and boric acid.

Maintaining sump water in an alkaline condition is needed for preventing dissolved radioactive iodine from being released to the containment atmosphere during the recirculation containment spray 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 Nuclear Power Plants, the iodine entering the containment is at least 95% cesium iodide (CsI) with the remaining 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 systems containing boric acid: RWST, safety system injection accumulators, and the RCS. This in effect will cause the sump water pH to become acidic. In order to keep the pH above 7, Ginna uses NaOH 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 a LOCA. 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, HNO3, and CsOH in its calculation. Although HI, HCl, and HNO3 are strong acids, the contribution that these acids could have in the pH is minimal due to the buffering 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 licensees methodology, assumptions, and performed hand calculations 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 boration sources 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 staff performed an independent verification that also demonstrated the containment sump pH would remain 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 acidic and 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, therefore preventing the iodine from being released. The licensees analysis has indicated that containment sump water will remain greater than 7 for at least 30 days. The NRC staff reviewed the licensees methodology for determining pH and performed an independent evaluation of the licensees calculations. Based on its evaluation, the staff concludes that the licensees proposed actions will maintain the sump water pH greater than 7 for 30 days following a LOCA, thus preventing the release of radioactive iodine into the containment atmosphere.

3.6 Operator Actions With respect to boric acid precipitation, the Ginna ECCS design incorporates upper plenum injection (UPI). The low-head safety injection pumps (RHR pumps) deliver flow directly to the upper 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 sump recirculation phase, operators are instructed to establish recirculation flow using the RHR pumps, which will maintain UPI, and terminate flow from the HHSI pumps. After a period of time, operators will be instructed to restart the HHSI pumps to reestablish simultaneous cold side 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 occur since the RCS will depressurize quickly and UPI will provide flushing flow through the core. No operator 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 when recirculation is established.

For large breaks in the hot leg, the licensees calculations show that the boric acid solution will not approach the solubility limit for atmospheric pressure conditions until approximately 6.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 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 reestablish cold 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 immediately manually 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 advance during 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 the ES-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 pressure and the boric acid concentration in the core is expected. EOP ES-1.2, Post-LOCA Cooldown and 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 exceed 100 EF/hr. When the 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 the licensees 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 described in the paragraph above, maintain UPI using the RHR pumps on recirculation, and terminate the HHSI. 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 would involve 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 procedure described 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 Changes The 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 the validation of this time. The licensee responded that the 10-minute time frame will be validated upon 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 licensees July 7, 2005, application, as supplemented, for an EPU, will be completed prior to startup from the fall 2006 refueling outage. The staff agrees that the operator action time of 10 minutes is reasonable, with the condition that the licensee completes its 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 the cooldown 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 requested information 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. The licensee 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 the plant cooldown rate shall not exceed 100 EF/hr to assure that boron will not come out of the solution 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 the ADVs. 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 EF/hr if the first ADV was unavailable. The licensee also indicated that an additional operator will be directed 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 licensee has also committed to validating, using the upgraded simulator, that the cooldown and depressurization 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 procedure are acceptable contingent upon the licensee completing the time validation of the cooldown step. Operator training also must be completed prior to the startup from the fall 2006 refueling outage during which the EPU would be implemented.

The NRC staff finds that the operator action times in the EOPs as described above are reasonable 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 the time frames listed. Thus, the staff has reasonable assurance that the actions will be accomplished 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 Changes 3.7.1 TS 3.5.1, Accumulators For TS SR 3.5.1.2 on borated water volume, the range was increased by lowering the minimum allowable 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 values are 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 Using the Automated Statistical Treatment of Uncertainty (ASTRUM)," January 2005.

7. 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 LCO 3.2.1)

8. WCAP-11145-P-A, "Westinghouse Small Break LOCA ECCS Evaluation Model Generic 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 General Network 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 Westinghouse and Combustion Engineering PWR designs in the NRC generic SE of the ASTRUM methodology. 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 all Westinghouse 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 be applicable 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 the proposed 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 facility component 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 no significant increase in the amounts, and no significant change in the types, of any effluents that may 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 no public 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 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (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 security or to the health and safety of the public.

7.0 REFERENCES

1. Ginna LLC letter, M. Korsnick to NRC, License Amendment Request Regarding Revised 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 Information Regarding 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 Generator and General Network Code," P. E. Meyer and G. K. Frick, February 1978.

7. Ginna LLC letter, M. Korsnick to NRC, "Supplemental Information Related to Small Break (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 for Additional 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 Information Regarding Topics Described in Meeting Minutes," January 25, 2006. (ADAMS No.

ML0609604165)

10. Ginna LLC letter, M. Korsnick to NRC, "License Amendment Request Regarding Adoption 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 Information Regarding 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. Armstrong Date: May 31, 2006