ML051260239

License Amendment Request Regarding Revised Loss of Coolant Accident (LOCA) Analyses-Changes to Accumulator, Refueling Water Storage Rwst), and Administrative Control Technical Specifications

ML051260239
Person / Time
Site:	Ginna
Issue date:	04/29/2005
From:	Korsnick M Constellation Energy Group
To:	Skay D Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML051260239 (75)

Text

Maria Korsnick 1503 Lake Road Vice R-esident Ontario, New York 14519-9364 57.

585.771.3494 585.771.3943 Fax maria.korsnick~constellation.com Constellation Energy R.E. Ginna Nuclear Power Plant April 29, 2005 Ms. Donna M. Skay Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Attn:

Document Control Desk Washington, D.C. 20555-0001

Subject:

License Amendment Request Regarding Revised Loss of Coolant Accident (LOCA) Analyses-Changes to Accumulator, Refueling Water Storage (RWST),

and Administrative Control Technical Specifications R.E. Ginna Nuclear Power Plant Docket No. 50-244

Dear Ms. Skay:

In accordance with the provisions of 10 CFR 50.90 of Title 10 of the Code of Federal Regulations (10 CFR), R.E. Ginna Nuclear Power Plant, LLC (Ginna LLC) is submitting a request for a license amendment to modify the Technical Specifications (TS) for the R.E. Ginna Nuclear Power Plant.

The proposed amendment would revise the TS to reflect revised analyses performed in support of the planned power uprate. The changes modify the volume and boron concentration requirements for the accumulators, revise the boron concentration requirements for the RWST and revise the list of referenced analytical methods specified in TS 5.6.5.b.

It has been determined that this amendment application does not involve a significant hazard consideration as determined by 10 CFR 50.92. Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of this amendment. provides a description and assessment of the proposed changes. Enclosure 2 provides the existing TS pages marked up to show the proposed changes. Enclosure 3 provides revised (clean) TS pages. Enclosure 4 provides the existing TS Bases pages marked up to reflect the proposed changes (for information only). Changes to the TS Bases will be provided in a future update in accordance with the Bases Control Program. Enclosure 5 provides additional information regarding Westinghouse Best-Estimate Large Break LOCA Methodology.

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a, provides additional information regarding the Westinghouse Small Break Loss-of-Coolant Accident analysis. Enclosure 7 provides additional information regarding the Westinghouse Post LOCA Analysis. Enclosure 8 provides a list of regulatory commitments associated with this license amendment request.

Approval of this amendment application is requested by April 30, 2006 to provide adequate time to prepare for implementation. Once approved, the amendment will be implemented prior to startup from the fall 2006 refueling outage.

In accordance with 10 CFR 50.91, a copy of this application, with enclosures, is being provided to the designated New York State Official.

If you have any questions regarding this submittal, please contact George Wrobel, Nuclear Safety and Licensing at (585) 771-3535.

VeW truly yo s ary G.

o nic

Enclosures:

1. Evaluation of Proposed Change

2. Proposed Technical Specification Changes (markup)

3. Revised Technical Specification Pages, (retyped)

4. Marked-up Copy of Technical Specification Bases

5. Application of Westinghouse Best-Esi"mate Large Break LOCA Methodology to the R. E. Ginna Nuclear Plant

6. Technical Evaluation-Small Break Loss-of-Coolant Accident

7. Ginna EPU Post LOCA Analysis

8. List Of Regulatory Commitments 2

9 STATE OF NEW YORK

TO WIT:

COUNTY OF W'AYNE I, Mary G. Korsnick, being duly sworn, state that I am Vice President - R.E. Ginna Nuclear Power Plant, LLC (Ginna LLC), and that I am duly authorized to execute and file this response on behalf of Ginna LLC. To the best of my knowledge and belief, the statements contained in this document are true and correct. To the,extent that these statements are not based on my personal knowledge, they are based upon information provided by other Ginna LLC employees and/or consultants. Such information has been reviewed in accordance with company practice and I believe it to be reliable.

Subscribed and sworn before me, a Notary Public i and for the State of New York and County of,b ev

,this 02 "dayof _1:

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2005.

WITNESS my Hand and Notarial Seal:

My Commission Expires:

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Notary Public MICHALENE A BUNTS Notary Public, State of New York Registration No. 01 BU6018576 Monroe County Commission Expires Jan I/ 7-o

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Ms. Donna M. Skay (Mail Stop 0-8-C2)

Project Directorate I Division of Licensing Project Management Office of Nuclear Regulatory Regulation U.S. Nuclear Regulatory Commission:

One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738 Regional Administrator, Region 1 U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 U.S. NRC Ginna Senior Resident Inspector Mr. Peter R. Smith New York State Energy, Research, and Development Authority 17 Columbia Circle Albany, NY 12203-6399 Mr. Paul Eddy NYS Department of Public Service 3 Empire State Plaza, 10th Floor Albany, NY 12223 4

ENCLOSURE 1 R.E. Ginna Nuclear Power Plant Description and Assessment of Proposed Change

Subject:

Revision to Technical Specifications 3.5.1, 3.5.4, and 5.6.5.b to Reflect the Results of Revised Analyses to Accommodate the Planned Power Uprate

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration 5.2 Applicable Regulatory Requirements/Criteria

1.0 DESCRIPTION

This letter is a request to amend Operating License No. DPR-18 for the R. E. Ginna Nuclear Power Plant (Ginna). This proposed change would revise the Operating License to change Technical Specification (TS) 3.5.1 and TS 3.5.4 to reflect the results of revised analyses performed to accommodate a planned power uprate for the facility and revise TS 5.6.5.b to permit the use at Ginna of NRC approved methodology for large break and small break LOCAs. Approval of the amendment is requested by April 30, 2006 to provide adequate time to prepare for implementation. Once approved, the amendment will be implemented prior to startup from the fall 2006 refueling outage.

2.0 PROPOSED CHANGE

S This request proposes to modify TS 3.5.1, 3.5.4 and 5.6.5 by revising the volume limits specified for the accumulators in SR 3.5.1.2, revising the boron concentration limits for the accumulators specified in SR 3.5.1.4, revising the boron concentration limits for the refueling water storage tank (RWST) specified in SR 3.5.4.2, and revising the listing of approved analytical methods specified in 5.6.5.b.

The proposed changes revise TS 3.5.1, Accumulators, as follows:

a. SR 3.5.1.2 currently states,'Verify borated water volume in each accumulator is

> 1111 cubic feet (50%) and < 1139 cubic feet (82%)"

SR 3.5.1.2 is being revised 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 states,"Verify boron concentration in each accumulator is

>2100 ppm and < 2600 ppm" SR 3.5.1.4 is being revised to state,'Verify boron concentration in each accumulator is

> 2550 ppm and <3050 ppm' The proposed changes revise TS 3.5.4, RWST, as follows:

a. SR 3.5.4.2 currently states,'Verify RWST boron concentration is > 2300 ppm and

< 2600 ppm"

b. SR 3.5.4.2 is being revised to state,'Verify RWST boron concentration is > 2750 ppm and < 3050 ppm" 2

The proposed changes revise TS 5.6.5, Core Operating Limits Report (COLR), as follows:

a.

Subsection b, the references in items 2, 7, 8, and 9 are being deleted.

NOTE: The new replacement items in the COLR reference list (Items 2, 7, 8, and 9) do not represent a like for like replacement.

b.

Item 2 is being replaced with,'WCAP-16009-P-A,'Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty (ASTRUM),'January 2005"

c.

Item 7 is being replaced 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 LCO 3.2.1)

d.

Item 8 is being replaced 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.

Item 9 is being replaced with,"WCAP-10079-P-A,'NOTRUMP - A Nodal Transient Small Break and General Network Code'August 1985 (Methodology for LCO 3.2.1)

f.

A new item 11 is being added as,'",WCAP-14710-P-A,"l-D Heat Conduction Model for Annular Fuel Pellets'May, 1998 (Methodology for LCO 3.2.1)'

Changes to the TS Bases for Specification for TS 3.5.1 and TS 3.5.4 are provided in markup form as Enclosure 4. The changes to the Bases include necessary revisions that incorporate the proposed revised Specifications. There is no Bases markup for the changes to 5.6.5 since there is no Bases section for the Administrative Controls Section of the TS.

3.0 BACKGROUND

In order to accommodate expected changes in the nuclear fuel (e.g., higher enrichment) that are associated with the planned power uprate, revised analyses has been performed for Ginna by Westinghouse using NRC approved methodology at the uprated operating conditions. These revised analyses indicate the necessity to modify the required accumulator water liquid volume and the associated boron concentrations as well as the RWST boron concentrations. In addition, the list of approved analysis methodologies specified in TS 5.6.5.b must be modified to reflect Ginna's use of these techniques.

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4.0 TECHNICAL ANALYSIS

The proposed changes to the accumulator water volume and boron concentration limits and RWST boron concentration limits have been evaluated for acceptability at the planned uprate power level using some analysis methodologies not previously used at Ginna. The analyses have been performed at 1811 MWt (inclusive of calorimetric uncertainty). The 1811 MWt core thermal power represents a conservatively high value that incorporates a 2% measurement uncertainty. The analyses at the planned power uprate level bounds operation at lower power levels including the current power level.

The change in the accumulator volume limits is still within the capability of the installed level instrumentation. The accumulator level instrumentation measure the level over a narrow 14 inch span which corresponds to 0-100% indicated level, not the entire height of the accumulators. The increase in the boron concentration limits for the accumulator and the RWST is acceptable with respect to the tanks and associated piping material specifications.

The analyses with respect to large break LOCAs uses the Westinghouse best-estimate large break loss of coolant accident (BELOCA) analysis methodology with ASTRUM, The small break LOCA analysis was performed using the 1985 Westinghouse SBLOCA Evaluation Model with NOTRUMP.(NQTRUMP-EM) including changes to the methodology described in WCAP-I054-P-A, Addendum 2, Revision 1, July 1997 and WCAP-14710-P-A, May 1998. The analyses also include post-LOCA long-term cooling evaluations using current methodology, including evaluation of the potential for boron precipitation. A summary description of each of these analyses is provided in this section A detailed description of each analyses is provided in Enclosures 5, 6, and 7.

Best Estimate Large Break LOCA Analysis This license amendment request (LAR) proposes to apply the Westinghouse best-estimate large break loss of coolant accident (BELOCA) analysis methodology which has been completed for the R. E. Ginna Nuclear Plant.

Westinghouse obtained generic NRC approval of its original topical report (WCAP-12945-P-A, Volume 1, Revision 2 and Volumes 2 through 5, Revision 1, and WCAP-14747 (Non-Proprietary)) describing best-estimate large break LOCA methodology in 1998. NRC approval of the methodology is documented in the NRC safety evaluation report appended to the topical report!. This methodology was later extended to 2-loop Westinghouse plants similar to Ginhna4iith Upper Plenum Injection (UPI) in 1999 as documented in the NRC safety evaluation report appended to the UPI topical report (WCAP-14449-P-A, Revision 1).

Westinghouse recently underwent a program to revise the statistical approach used to develop the Peak Cladding Temperature (PCT) and oxidation results at the 95th percentile. This method is still based on the Code Qualification Document (CQD) 4

methodology and follows the steps in the Code Scaling, Applicability, and Uncertainty (CSAU) methodology. However, the",uncertainty analysis (Element 3 in CSAU) is replaced by a technique based on orderi'statistics. The ASTRUM methodology replaces the response surface technique with a' statistical sampling method where the uncertainty parameters are simultaneously sampled 'for each case. The ASTRUM evaluation model is documented in WCAP-1 6009-P-A, January 2005 and has received NRC approval for referencing in licensing calculations. (Only WCAP-1 6009-P-A, January 2005 is proposed to be added to the listing of methodologies in TS 5.6.5 since WCAP-12945-P-A, WCAP-14747, and WCAP-14449-P-A are referenced by WCAP-16009-P-A.)

The WCOBRA/TRAC model was developed to support a fuel change from OFA to 422Vantage + (422V+) fuel. Ginna will transition to 422V+ in association with the planned power uprate. Additional information regarding the transition to 422V+ fuel will be included in the extended power uprate submittal. The BELOCA analysis assumed an increased reactor power level of <1811 MWt and the revised accumulator water volumes specified in the proposed change to TS SR 3.5.1.2. The BELOCA analysis used a lower minimum accumulator boron concentration than the proposed changes to TS SR 3.5.1.4. The analysis at lower boron concentration than that specified in TS SR 3.5.1.4 ensures the specified minimum boron concentration is bounding. The analysis at the increased power level bounds operation at lower power levels. These analyses demonstrate acceptable results at the higher power level, revised accumulator water volume, and revised accumulator boron' concentrations.

The results of the BELOCA analysis determined the peak cladding temperature of 1,870'F (95/95) is well below the 10 CFR 50.46 acceptance limit of <2200'F, and the maximum local oxidation of 3.4 percent (95/95) is well below the 10 CFR 50.46 acceptance limit of <17 percent. The core-wide hydrogen generation of 0.30 percent (95/95) remains well below the 10 CFR 50.46 acceptance limit of 1 percent, and the core geometry remains amenable to cooling. Enclosure 5 provides additional details regarding the BELOCA analysis methodology, assumptions, and results.

Small Break LOCA (SBLOCA) Analysis The SBLOCA analysis was performed at the planned uprate conditions using the 1985 Westinghouse SBLOCA Evaluation Model with NOTRUMP (NOTRUMP-EM) [WCAP-10079-P-A, August 1985, WCAP-10054-P-A, August 1985, and WCAP-11145-P-A, October 1986.] including changes to the methodology described in WCAP-10054-P-A, Addendum 2, Revision 1, July 1997 and WCAP-14710-P-A, May 1998 The analysis used an increased reactor power level of 181 1 MWt and modeled the plant-specific design features of Ginna such as an accumulator water volume range specified in the proposed change to TS SR 3.5.1.2, a flat upper support plate, a downflow barrel-baffle region, and an upper head temperature equal to the hot leg temperature (THOT).

Prior 10 CFR 50.46 assessments (included in UFSAR 15.6.4.1.4.5) were incorporated into the analysis, primarily through the use of corrected code versions and selection of 5

input values. The SBLOCA methodology does not assume any specific boron concentration in the accumulators or RWST.

The results of the SBLOCA analysis determined the peak cladding temperature is 1,1670F, and the maximum local oxidation is 0.07 percent. The design limit 95% upper bound pre-transient oxidation value for each of the fuel designs that will be included in the EPU cores is < 16%. The calculations performed to support the extended power uprate studies demonstrated that the u per bound oxidation values are wvell below 16%.

It is anticipated that future cycle specific calculations will continue to show significant margin with respect to the 95% upper bound pre-transient oxidation limit. Because the transient oxidation is so low, the sum of the transient and pre-transient oxidation remains below 16% at all times in life. The core-wide hydrogen generation remains well below the 10 CFR 50.46 acceptance limit of 1 percent, and the core geometry remains amenable to cooling.

The SBLOCA analysis results meet the pertinent acceptance criteria of 10 CFR 50.46.

The peak cladding temperature is less than 2,200'F; the maximum local oxidation is less than 17 percent; the core-wide hydrogen generation is less than 1 percent; and, the core geometry remains amenable to cooling. provides additional details regarding the SBLOCA analysis methodology, assumptions, and results.

New Peak Clad Temperature (PCT) Results The use of the new analysis methodology for BELOCA (with ASTRUM) and the revised analysis methodology for small break LOCA will establish new analysis of record peak clad temperatures (PCTs) for these events. The new licensing basis analysis of record for the BELOCA will be 18700F. The new licensing basis analysis of record for the small break LOCA will be 11670F. These analyses are based on the analysis for 422 Vantage+

(422V+) fuel and are effective beginning with Cycle 33 with a mixed core containing both OFA and 422V+ fuel. Transition cycles containing OFA fuel are bounded by the analysis for 422V+ fuel. After the startup from the fall 2006 refueling outage, these changes will be included in the subsequent annual 10 CFR 50.46 report to satisfy the requirements of 10 CFR 50.46 paragraph (a)(3)(ii).

Post LOCA Analysis An evaluation of the effects of the proposed changes in boron (H3BO3) concentrations in the accumulators and RWST on containment sump pH has been completed. The results of this evaluation indicate existing NaOH addition requirements are adequate to control sump pH under post LOCA conditions.

To support the planned power uprate, the Ginna Station Technical Specification RWST and accumulator boron concentration limits (TS SR 3.5.4.2 and SR 3.5.1.4 respectively) are proposed to be changed. To support the changes to the Technical Specification boron 6

concentrations, post-LOCA long-term cooling evaluations were performed. These evaluations demonstrated that at the planned uprated core power level the proposed Technical Specification boron concentration changes are acceptable with respect to post-LOCA long-term cooling.

The Ginna licensing position for satisfying the requirements of 10 CFR 50.46 Paragraph (b), Item (5), is documented in WCAP.-8339, June 1974. This document states that the core will remain subcritical post-LOCA by borated water from various injected Emergency Core Cooling System (ECCS) water sources. Post-LOCA sump boron calculations demonstrate that the core will remain subcritical upon entering the sump recirculation phase of ECCS injection. Containment sump boron concentration calculations were used to develop a core reactivity limit that was confirmed as part of the Westinghouse Reload Safety Evaluation Methodology described in WCAP-9272-P-A, July 1985 (TS 5.6.5.b.1).

There are no specific acceptance criteria when calculating the minimum post-LOCA sump boron concentration other than assuring the core remains subcritical. However, the resulting sump boron concentration, which is calculated as a function of the pre-LOCA RCS boron concentration, is reviewed for each cycle-specific core design to confirm that adequate boron exists to maintain subcriticality for the long-term post-LOCA. With respect to post-LOCA criticality, a post-LOCA subcriticality boron limit curve was developed for the expected plant conditions. The curve will be included in the Reload Safety Analysis Checklist (RSAC) for each reload cycle.

The Westinghouse ECCS evaluation model relies on precluding boric acid precipitation for satisfying the requirements of 10 CFR 50.46 Paragraph (b), Item (4) and 10 CFR 50.46, Paragraph (b), Item (5). The potential for boric acid precipitation during post-LOCA conditions is evaluated by calculations that consider the buildup of boric acid solution in the core for limiting scenarios. The limiting scenarios are large breaks where the break location and the point of safety injection are such that the water in the core is stagnant. Borated ECCS water boils off due to decay heat, leaving behind boric acid.

The concern is the potential for the boric acid solution in the core to reach the boric acid solubility limit.

There were no specific acceptance criteria for the results of the calculations to preclude boric acid precipitation. The Emergency Operating Procedures will address the maximum time permissible to establish simultaneous hot leg / cold leg injection.

With respect to boric acid precipitation, Ginna is an upper plenum injection (UPI) design.

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 signal starts both high head SI 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 7

the RHR pumps which will maintain UPI, and terminate flow from the high head SI pumps. After a period of time (less than six hours), operators will be instructed to restart the high head safety injection pumps to re-establish simultaneous cold side and hot side (UPI) injection to provide long term core cooling for all LOCA scenarios.

Three categories of LOCA break sizes were considered for the boric acid precipitation evaluation: (1) large breaks (greater than approximately 6" in diameter) where the RCS pressure rapidly decreases to the UPI initiation pressure (140 psia) with no operator action, (2) intermediate breaks (between approximately 1" and 6" in diameter) where RCS pressure decreases but stabilizes above the UPI initiation pressure, and (3) small breaks (approximately 1" in diameter and smaller) where high head safety injection refills the RCS and natural circulation is established.

For large breaks in the cold leg, boric acid precipitation cannot occur since the RCS will depressurize quickly and upper plenum injection will provide flushing flow through the core.

For large breaks in the hot leg, the core region boric acid concentration will only begin to increase with the termination of high head safety injection to the cold legs. Calculations (Enclosure 7) have shown that the boric acid solution will not approach the solubility limit for atmospheric pressure conditions until approximately 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after the termination of SI to the cold leg. For the EPU, the Ginna Emergency Operating Procedure ES-1.3, Transfer to Cold Leg Recirculation will be revised to instruct operators to re-establish cold leg SI (i.e. simultaneous injection) no later than 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after the termination of SI in the cold leg. In this case boric acid precipitation will be prevented.

There are no limitations on early switchover to simultaneous injection.

For intermediate breaks in the cold leg, RCS pressure will stabilize above the UPI initiation pressure and the core region boric acid concentration will begin to increase prior to upper plenum injection. Emergency Operating Procedure ES-1.2, Post-LOCA Cooldown and Depressurization directs the operators in this scenario to depressurize the RCS using the condenser steam dumps or SG atmospheric relief valves. Since the core boil-off rate would decrease at higher RCS pressures, the previously discussed calculations that assume atmospheric conditions conservatively predict the rate of increase in boron concentration. 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. Operators will depressurize the RCS to less than 140 psia within six hours. Based on the bounding calculations for large break LOCA, 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 intermediate breaks in the hot leg, RCS pressure will again stabilize above the UPI initiation pressure, however the core boric acid concentration will not increase until the high head cold leg SI is terminated. Operators are again directed to depressurize the RCS, and maintain UPI using the RHR pumps on recirculation and terminate the high head SI as necessary. Once high head SI to the cold leg is terminated, this scenario is 8

bounded by the large hot leg break scenario where cold leg SI (i.e. simultaneous injection) will be re-established no later than 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after termination. Note that there are no differences in the required operator actions for hot or cold leg breaks.

For small hot leg or cold leg breaks the RCS remains pressurized with the high head SI pumps. Natural circulation will be established and will maintain flushing flow through the core for either hot or cold side breaks. Emergency Operating Procedure (EOP) actions will depressurize and cooldown the reactor under controlled conditions with eventual realignment to RHR normal shutdown cooling. Natural circulation or RHR normal shutdown cooling will dilute any buildup of boric acid in the core.

A post-LOCA subcriticality boron limit curve was developed for the EPU plant conditions. Cycle-specific reload safety evaluations will ensure that the core will remain subcritical post-LOCA, and decay heat can be removed for the extended period required by the remaining long-lived radioactivity.

Summary LOCA Analyses using NRC approved methodology have been performed at the planned uprate power level using the proposed revised limits specified in the TS for accumulator volume, accumulator boron concentrations, and RWST boron concentration. A post LOCA analysis using a current conservative methodology has been performed including consideration of post LOCA criticality and the potential for boron precipitation. These analyses demonstrate the acceptability of the proposed revised limits with respect to satisfying 10 CFR 50.46 requirements including post LOCA precipitation considerations.

5.0 REGULATORY ANALYSIS

5.1 NO SIGNIFICANT HAZARDS CONSIDERATION Ginna LLC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed changes include revising accumulator volume and boron concentration requirements and Refueling Water Storage Tank (RWST) boron concentration requirements that are necessary to accommodate expected changes in the nuclear fuel (e.g., higher enrichment) that are associated with the planned power uprate. Additionally, thechange would allow Ginna to utilize analysis methodologies that have been previously approved for use at Westinghouse 9

nuclear plants. The changes to the TS are necessary to ensure the acceptability of these systems to perform their intended function in the event of an accident.

The proposed changes do not adversely affect accident initiators or precursors nor significantly alter the design assumptions, conditions, and configuration of the facility or the manner in which the plant is operated and maintained. The proposed changes do not adversely alter or prevent the ability of structures, systems, and components (SSCs) from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. The proposed changes do not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. Further, the proposed changes do not increase the types and amounts of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational/public radiation exposures. The proposed changes cannot affect the probability of an accident occurring since they reflect a necessary change in plant design consistent with current design which is not an accident initiator. The proposed changes cannot increase the consequences of postulated accidents since they reflect a change in plant design that will continue to mitigate the effects of potential accidents. Therefore, the changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No The proposed changes include revising accumulator volume and boron concentration requirements and RWST boron concentration requirements that are necessary to accommodate expected changes in the nuclear fuel (e.g., higher enrichment) that are associated with the planned powver uprate. Additionally, the change would allow Ginna to utilize analysis methodologies that have been previously approved for use at Westinghouse nuclear plants. The changes to the TS are necessary to ensure the acceptability of these systems to perform their intended function in the event of an accident.

The proposed changes involve changes to accumulator volume and boron concentration requirements and RWST boron concentration requirements to ensure the continued acceptability of LOCA and post LOCA analysis results. The changes to the Technical Specifications (TS) are necessary to properly accommodate the changes in plant design. The changes ensure applicable acceptance criteria will continue to be met. The changes do not involve a significant change in the methods governing normal plant operation. The proposed TS changes do not create the possibility of a new or different type of accident from those previously evaluated since they reflect a change that will 10

ensure the accumulators and RWST will continue to perform their intended function in the event of an accident.

Therefore, the changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No The proposed changes include revising accumulator volume and boron concentration requirements and RWST boron concentration requirements that are necessary to accommodate expected changes in the nuclear fuel (e.g., higher enrichment) that are associated with the planned power uprate. Additionally, the change would allow Ginna to utilize analysis methodologies that have been previously approved for use at Westinghouse nuclear plants. The changes to the TS are necessary to ensure the acceptability of these systems to perform their intended function in the event of an accident.

The level of safety of facility operation is not significantly affected by the proposed changes since there is no change in the intent of the TS requirements of assuring proper plant response 'ini the event of an accident. The response of the plant systems to accidents anid transients reported in the Updated Final Safety Analysis Report (UFSAR) is not adversely affected by this change. Therefore, the capability to satisfy accident analysis acceptance criteria is not adversely affected. The proposed TS change cannot involve a significant reduction in the margin of safety since it is based upon changes that will maintain a substantial margin of safety with respect to accumulators and RWST functions.. Therefore, the changes do not involve a significant reduction in a margin of safety.

Based on the above, Ginna LLC concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

5.2 APPLICABLE REGULATORY REQUIREMENTS/CRITERIA The regulatory requirements/criteria applicable to this request for amendment is 10 CFR 50.46. In brief summary, the pertinent portions of 10 CFR 50.46 requires that ECCS cooling performance be calculated in accordance with an acceptable evaluation model and the results'of the analysis must demonstrate to a high level of probability that the criteria specified in paragraph b are satisfied. 10 CFR 50.46 paragraph b requires (1) the calculated maximum fuel element cladding temperature shall not exceed 22000 F; (2) the calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before 11

oxidation; (3) the calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel,,excluding the cladding surrounding the plenum volume, were to react; (4) calculated changes in core geometry shall be such that the core remains amenable to cooling; and (5) after any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core.

The revised analyses have been performed using acceptable methodologies. The analyses have been performed using conditions and assumptions (e.g., power level) that are consistent with the planned power uprate. Although this license amendment request does not permit an increase in rated thermal power, operation at lower power levels is bounded by the analysis performed for the uprated conditions. The proposed changes to TS 5.6.5 allow Ginna to adopt analysis methodologies previously reviewed and accepted by NRC in similar applications.

The results of the new analyses indicate the criteria specified in 10 CFR 50.46 continue to be satisfied.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the'proposed manner, (2) such activities will be conducted in accordance with 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.

6.0 ENVIRONMENTAL CONSIDERATION

Ginna LLC has evaluated the proposed changes and determined that:

1. The changes do not involve a significant hazards consideration; and

2. The changes do not involve a significant change in the types or significant increase in the amounts of any effluent that may be released offsite since there is no change in the type or quantities of material available for release than that previously analyzed; and

3. The changes do not involve a significant increase in individual or cumulative occupational radiation exposure since the change in plant configuration does not significantly increase overall operations and maintenance requirements nor is any different type of equipment required to be installed.

12

Accordingly, the proposed changes meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), an environmental assessment of the proposed changes is not required.

7.0 REFERENCES

None

.13

ENCLOSURE 2 R.E. Ginna Nuclear Power Plant Proposed Technical Specification Changes (markup) l.

Accumulators 3.5.1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.1 Accumulators LCO 3.5.1 Two ECCS accumulators shall be OPERABLE.

APPLICABILITY:

MODES I and 2, MODE 3 with pressurizer pressure > 1600 psig.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

One accumulator A.1 Restore boron 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Inoperable due to boron concentration to within concentration not within limits.

limits.

B.

One accumulator B.1 Restore accumulator to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> inoperable for reasons OPERABLE status.

other than Condition A.

C.

Required Action and C.1 Be in MODE 3.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A or B AND not met.

C.2 Reduce pressurizer 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> pressure to s 1600 psig.

D.

Two accumulators D.1 Enter LCO 3.0.3.

Immediately inoperable.

SURVEILLANCE REQUIREMENTS SURVEILLANCE l

FREQUENCY SR 3.5.1.1 Verify each accumulator motor operated isolation 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> valve is fully open.

SR 3.5.1.2 Verikborated wa 2t 6);cubic feet accumulator is cubic feet (e).

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />

-:7 R.E. Ginna Nuclear Power Plant 3.5.1 -1 Amendment i

Accumulators 3.5.1 SURVEILLANCE FREQUENCY SR 3.5.1.3 Verify nitrogen cover pressure in each accumulator is 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 2 700 psig and

• 790 psig.

SR 3.5.1.4 Ve boron concentration In each accumulator Is 31 days on a 2

nppmand: s ppm.

STAGGEREDTEST BASIS SR 3.5.1.5 Verify power Is removed from each accumulator 31 days motor operated isolation valve operator when pressurizer pressure Is > 1600 psig.

Amendmenr R.E. Ginna Nuclear Power Plant 3.5.1-2

RWST 3.5.4 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.4 Refueling Water Storage Tank (RWST)

LCO 3.5.4 The RWST shall be OPERABLE.

APPLICABILITY:

ACTIONS MODES 1, 2, 3, and 4.

CONDITION REQUIRED ACTION COMPLETION TIME A.

RWST boron A.1 Restore RWST to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> concentration not within OPERABLE status.

limits.

B.

RWST water volume not B.1 Restore RWST to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> within limits.

OPERABLE status.

C.

Required Action and C.1 Be In MODE 3.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met.

AND C.2 Be In MODE 5.

36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS Amendment e R.E. Ginna Nuclear Power Plant 3.5.4-1

'"M Reporting Requirements 5.6 5.6.5 CORE OPERATING LIMITS REPORT (COLR)

The following administrative requirements apply to the COLR:

a.

Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented In the COLR for the following:

2.1, Safety Umim (SLs)

LCO 3.1.1.

'SHUTDOWN MARGIN (SDM)";

LCO 3.1.3.

"MODERATOR TEMPERATURE COEFFICIENT (MTC)";

LCO 3.1.5.

"Shutdown Bank Insertion Umir';

LCO 3.1.6.

"Control Bank Insertion Umits";

LCO 3.2.1.

'Heat Flux Hot Channel Factor (FO(Z))';

LCO 3.2.2.

'Nuclear Enthalpy Rise Hot Channel Factor (FNH)";

LCO 3.2.3.

-AXIAL FLUX DIFFERENCE (AFD)";

LCO 3.3.1.

"Reactor Protection System (RPS) Instrumentation";

LCO 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nudeate Boiling (DNB) Umits" and LCO 3.9.1, "Boron Concentration."

b.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1.

WCAP-9272-P-A, WVstinghouse Reload Safety Evaluation Methodology," July 1985.

(Methodology for 2.1, LCO 3.1.1, LCO 3.1.3. LCO 3.1.5, LCO 3.1.6. LCO 3.2.1, LCO 3.2.2, LCO 3.2.3, and LCO 3.9.1.)

2.

VVCAP-i 3671$-A, "10 CFR 50.46 Evaluatioh Model Report: )

I---C c1BRUrRAO Two-Loop Upper Plenrain Injecti 7odel/

Updates to Support ZIRLOTM Claddfng Option" ' ebruary-1994.

/

<(Methodology for LCO 3.2.1)/

3.

WCAP-8385, "Power Distribution Control and Load Following Procedures - Topical Report," September 1974.

(Methodology for LCO 3.2.3.)

R.E. Ginna Nuclear Power Plant 5.6-2 Amendment

Insert A WCAP-1 6009-P-A, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty (ASTRUM)," January 2005.

Reporting Requirements 5.6

4.

WCAP-12610-P-A, VANTAGE + Fuel Assembly Reference Core Report," April 1995.

(Methodology for LCO 3.2.1.)

5.

WCAP 11397-P-A. 'Revised Thermal Design Procedure:'

April 1989.

(Methodology for LCO 3.4.1 when using RTDR)

6.

WCAP-1 0054-P-A and WCAP-10081-A, "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code." August 1985.

(Methodology for LCO 3.2.1.)

7.

AP-10924-P-A,.Volume 1, Revision 1, 'Westinghouse BgBreaJ A Best-EstimateMetfiodology,-Vbfume1 1

HeodeLO iption and Validation Responses to NRC uestions," and Addenddai,2,3, December 1988..-

Methodology for LCO 3.2.1.-

8.

EAP-0924-P-A, Y me 2, Revision 2 nghouse Large-Bre~0CA Best-Estimate ology, Volum\\

Applicatkdhi to Two-Loop PtPe ESuipped with Urper Plenum lnjection," and Addenduml, December 198g.

.2 Methodology for LCO 3.2.1.)

9.

WCAP-109244P-i~olume 1, RevisiotI..1Addendu A,\\

P g

1 b"Wstingh6Euse Large-Break LOGA-est-Estimet\\

Methodology, Volume 1; Mo-el Descriptiol6and Validation, Addendum 4: Model Revisions," Mafch 1991.,- -

(Methodology fdr LCO 3.2.1.)'

10.

WCAP-8745, "Design Basis for the Thermal Overpower Delta T and Thermal Overtemperature Delta T Trip Functions,"

March 1977.

It S se

-(Methodology for LCO 3.3.1.)

c.

The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d.

The COLR, Induding any midcycle revisions or supplements, shall be provided upon Issuance for each reload cycle to the NRC.

R.E. Ginna Nuclear Power Plant 5.6-3 Amendment

Insert B 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)

Insert C WCAP-11145-P-A, "Wcstinghouse Small Break LOCA ECCS Evaluation Model Gencric Study With the NOTRUMP Code," October 1986.

(Methodology for LCO 3.2.1)

Insert D WCAP-1 0079-P-A, "NOTRUMP - A Nodal Transient Small Break and General Network Code," August 1985 (Methodology for LCO 3.2.1)

Insert E

11.

WCAP-14710-P-A, "1-D Heat Conduction Model for Annular Fuel Pellets," May, 1998 (Methodology for LCO 3.2.1)

ENCLOSURE 3 R.E. Ginna Nuclear Power Plant Revised Technical Specification Pages (retyped)

Accumulators 3.5.1 3.5 3.5.1 EMERGENCY CORE COOLING SYSTEMS (ECCS)

Accumulators 1

Two ECCS accumulators shall be OPERABLE.

LCO 3.5.1 APPLICABILITY:

MODES 1 and 2, MODE 3 with pressurizer pressure > 1600 psig.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

One accumulator A.1 Restore boron 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> inoperable due to boron concentration to within concentration not within limits.

limits.

B.

One accumulator B.1 Restore accumulator to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> inoperable for reasons OPERABLE status.

other than Condition A.

C.

Required Action and C.1 Be in MODE 3.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A or B AND not met.

C.2 Reduce pressurizer 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> pressure to < 1600 psig.

D.

Two accumulators D.1 Enter LCO 3.0.3.

Immediately inoperable.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.1.1 Verify each accumulator motor operated isolation 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> valve is fully open.

SR 3.5.1.2 Verify borated water volume in each accumulator is 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 2 1090 cubic feet (24%) and < 1140 cubic feet (83%).

I R.E. Ginna Nuclear Power Plant 3.5.1-1 Amendment

Accumulators 3.5.1 SURVEILLANCE FREQUENCY SR 3.5.1.3 Verify nitrogen cover pressure in each accumulator is 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 2 700 psig and

• 790 psig.

SR 3.5.1.4 Verify boron concentration in each accumulator is 31 days on a 2 2550 ppm and

• 3050ppm.

STAGGEREDTEST BASIS SR 3.5.1.5 Verify power is removed from each accumulator 31 days motor operated isolation valve operator when pressurizer pressure is > 1600 psig.

I R.E. Ginna Nuclear Power Plant 3.5.1-2 Amendment

RWST 3.5.4 3.5 3.5.4 EMERGENCY CORE COOLING SYSTEMS (ECCS)

Refueling Water Storage Tank (RWST) 4 The RWST shall be OPERABLE.

LCO 3.5.e APPLICABILITY:

MODES 1, 2,3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

RWST boron A.1 Restore RWST to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> concentration not within OPERABLE status.

limits.

B.

RWST water volume not B.1 Restore RWST to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> within limits.

OPERABLE status.

C.

Required Action and C.1 Be in MODE 3.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met.

AND C.2 Be in MODE 5.

36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.4.1 Verify RWST borated water volume is 2 300,000 7 days gallons (88%).

SR 3.5.4.2 Verify RWST boron concentration Is 2 2750 ppm and 7 days s 3050 ppm.

I R.E. Ginna Nuclear Power Plant 3.5.4-1 Amendment

Reporting Requirements 5.6 5.0 ADMINISTRATIVE CONTROLS 5.6 Reporting Requirements The following reports shall be submitted in accordance with 10 CFR 50.4.

5.6.1 Deleted 5.6.2 Annual Radiological Environmental Operating Report The Annual Radiological Environmental Operating Report covering the operation of the plant during the previous calendar year shall be submitted by May 15 of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the radiological environmental monitoring activities for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C.

The Annual Radiological Environmental Operating Report shall include the results of analyses of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible.

5.6.3 Radioactive Effluent Release Report The Radioactive Effluent Release Report covering the operation of the plant shall be submitted in accordance with 1 0 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the plant. The material provided shall be consistent with the objectives outlined in the ODCM and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B.1.

5.6.4 Deleted R.E. Ginna Nuclear Power Plant 5.6-1 Amendment

Reporting Requirements 5.6 5.6.5 CORE OPERATING LIMITS REPORT (COLR)

The following administrative requirements apply to the COLR:

a.

Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

2.1,

'Safety Limits (SLs)";

LCO 3.1.1.

-SHUTDOWN MARGIN (SDM)";

LCO 3.1.3, "MODERATOR TEMPERATURE COEFFICIENT (MTC)";

LCO 3.1.5, "Shutdown Bank Insertion Limit";

LCO 3.1.6, "Control Bank Insertion Limits";

LCO 32.1, "Heat Flux Hot Channel Factor (Fm(Z))";

LCO 32.2, "Nuclear Enthalpy Rise Hot Channel Factor (FN t M)";

LCO 3.2.3 "AXIAL FLUX DIFFERENCE (AFD)";

LCO 3.3.1.

"Reactor Protection System (RPS) Instrumentation";

LCO 3A.1.

"RCS Pressure, Temperature, and Flow Departure from Nucleate Boling (DNB) Limits"; and LCO 3.9.1.

"Boron Concentration."

b.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1.

WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985.

(Methodology for 2.1, LCO 3.1.1, LCO 3.1.3, LCO 3.1.5, LCO 3.1.6, LCO 3.2.1, LCO 3.2.2, LCO 3.2.3, and LCO 3.9.1.)

2.

WCAP-16009-P-A, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty (ASTRUM)," January 2005.

3.

WCAP-8385, "Power Distribution Control and Load Following Procedures - Topical Report," September 1974.

(Methodology for LCO 3.2.3.)

R.E. Ginna Nuclear Power Plant 5.6-2 Amendment

Reporting Requirements 5.6

4.

WCAP-12610-P-A, 'VANTAGE + Fuel Assembly Reference Core Report," April 1995.

(Methodology for LCO 3.2.1.)

5.

WCAP 11397-P-A, "Revised Thermal Design Procedure,"

April 1989.

(Methodology for LCO 3.4.1 when using RTDP.)

6.

WCAP-10054-P-A and WCAP-1 0081-A, "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," August 1985.

(Methodology for LCO 3.2.1.)

7.

WCAP-1 0054-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-1 0079-P-A, "NOTRUMP - A Nodal Transient Small Break and General Network Code," August 1985.

(Methodology for LCO 3.2.1)

10.

WCAP-8745, "Design Basis for the Thermal Overpower Delta T and Thermal Overtemperature Delta T Trip Functions,"

March 1977.

(Methodology for LCO 3.3.1.)

11.

WCAP-14710-P-A, "1-D Heat Conduction Model forAnnular Fuel Pellets," May, 1998.

(Methodology for LCO 3.2.1)

c.

The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d.

The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

R.E. Ginna Nuclear Power Plant 5.6-3 Amendment

Reporting Requirements 5.6 5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

The following administrative requirements apply to the PTLR:

a.

RCS pressure and temperature limits for heatup, cooldown, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following.

LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits'

b.

The power operated relief valve lift settings required to support the Low Temperature Overpressure Protection (LTOP) System, and the LTOP enable temperature shall be established and documented in the PTLR for the following:

LCO 3A.6, RCS Loops - MODE 4";

LCO 3.4.7,

'RCS Loops - MODE 5, Loops Filled";

LCO 3.4.10, "Pressurizer Safety Valves"; and LCO 3.4.12,

'LTOP System."

c.

The analytical methods used to determine the RCS pressure and temperature and LTOP limits shall be those previously reviewed and approved by the NRC in NRC letter, "R.E. Ginna - Acceptance for Referencing of Pressure Temperature Limits Report, Revision 2 (TAC No. M96529)," dated November 28,1997. Specifically, the methodology is described in the following docum ents:

1.

Letter from R.C. Mecredy, Rochester Gas and Electric Corporation (RG&E), to Document Control Desk, NRC, Attention: Guy S. Vissing, "Application for Facility Operating License, Revision to Reactor Coolant S ystem (RCS)

Pressure and Temperature Limits Report (PTLR)

Administrative Controls Requirem ents," Attachment VI, September 29, 1997, as supplemented by letter from R.C.

Mecredy, RG&E, to Guy S. Vissing, NRC, "Corrections to Proposed Low Temperature Overpressure Protection System Technical Specification," October 8, 1997.

2.

WCAP-14040-NP-A, "Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," Sections 1 and 2, January, 1996.

R.E. Ginna Nuclear Power Plant 5.6-4 Amendment

Reporting Requirements 5.6

d.

The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for revisions or supplement thereto.

R.E. Ginna Nuclear Power Plant 5.6-5 Amendment

ENCLOSURE 4 R.E. Ginna Nuclear Power Plant Mlarked-up Copy of Technical Specification Bases

Accumulators B 3.5.1 The accumulator size, water volume, and nitrogen cover pressure are selected so that one of the two accumulators Is sufficient to partially cover the core before significant clad melting or zirconium water reaction can occur following a LOCA. The need to ensure that one accumulator is adequate for this function is consistent with the LOCA assumption that the entire contents of one accumulator will be lost via the RCS pipe break during the blowdown phase of the LOCA.

APPLICABLE The accumulators are assumed OPERABLE In both the large and small SAFETY break LOCA analyses at full power (Ref. 4). These are the Design Basis ANALYSES Accidents (DBAs) that establish the acceptance limits for the accumulators. Reference to the analyses for these DBAs is used to assess changes In the accumulators as they relate to the acceptance limits.

In performing the LOCA calculations, conservative assumptions are made concerning the availability of ECCS flow. In the early stages of a large break LOCA, with or without a loss of offsite power, the accumulators provide the sole source of makeup water to the RCS. The assumption of loss of offsite power Is required by regulations and conservatively Imposes a delay wherein the ECCS pumps cannot deliver flow until the emergency diesel generators start, come to rated speed, and go through their timed loading sequence. In cold leg bkeak l t

\\

scenarios, the entire contents of one accumulator are assumed to be lost through the break.

r-.-- A.

Th sarge break LOCA is a double ended guillotine break at the

/

on r discharge of the reactor coolant pump. During this event, the accumulators discharge to the RCS as soon as RCS pressure decreases to below accumulator pressure. As a conservative estimate, no credit is taken for ECCS pump flow until an effective delay has elapsed. This delay accounts for SI signal generation, the diesels starting, and the pumps being loaded and delivering full flow. During this time, the accumulators are analyzed as providing the sole source of emergency core cooling. No operator action is assumed during the blowdown stage of a large break LOCA.

The worst case small break LOCA analyses also assume a time delay before pumped flow reaches the core. For the larger range of small breaks, the rate of blowdown is such that the increase in fuel clad temperature is terminated solely by the accumulators, with pumped flow then providing continued cooling. As break size decreases, the accumulators and safety injection pumps both play a part in terminating the rise in clad temperature. As break size continues to decrease, the role of the accumulators continues to decrease until they are not required and the safety Injection pumps become solely responsible for terminating the temperature Increase.

R.E. Ginna Nuclear Power Plant B 3.5.1-2 RevisionQ

Accumulators B 3.5.1 This LCO helps to ensure that the following acceptance criteria established for the ECCS by 10 CFR 50.46 (Ref. 5) will be met following a LOCA:

a.

Maximum fuel element cladding temperature is

• 22000F;

b.

Maximum cladding oxidation is S 0.17 times the total cladding thickness before oxidation:

c.

Maximum hydrogen generation from a zirconium water reaction is s 0.01 times the hypothetical a mount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react, and

d.

Core is maintained in a coolable geometry.

Since the accumulators discharge during the blowdow<phase of a LOCA, they do not contribute to the long term cooling requirements of 10 CFR 50.46.

Fori thef~~small break LOCA a nominal contained accumulator water volume Is used. The contained water volume is the same as the deliverable volume for the accumulators, since the accumulators are emptied, once discharged. For small breaks, an increase in water volume Is a peak clad temperature penalty due to the reduced gas volume. A peak clad temperature penalty is an assumed increase in the calculated peak clad temperature due to a change in an Input parameter. For large breaks, an increase in water volume can be either a peak clad temperature penalty or benefit, depending on 5 ie W,

I, ret 1A downcomer filling and subsequent s ill through the break during the core Ct LI

-S V C S reflooding portion of the transient a

u

~~~~~ac utatl v s cuuapolum Inc ud-h ather IiK me fro the r+

a

-r..L cf u

tor toi check yalve a

1JC) ha r

a Vok l

S (Temnimum boron concentration c

4

.s; w

ri boron concentration calculation. The calculation is performed to assure o

ve

• 0*

z

• reactor subcriticality in a post LOCA environment. Of particular Interest Is

',the large break LOCA, since no credit Is taken for control rod assembly tic l

'i a

r" einsertion. A reduction in the accumulator minimum boron concentration to-I would produce a subsequent reduction in the available containment L o o k sump concentration for post LOCA shutdown and an increase in the L

maximum sump pH. The maximum boron concentration is used in cor ddr5 4 I *t; determining the time frame in which boron precipitation is addressed post 1

^

LOCA. The maximum boron concentration limit is based on the coldest expected temperature of the accumulator water volume and on chemical l

f effects resulting from operation of theCndtheContainmentSprai n3O d

(CS) System. The maximum value o wouldnoc

'e.,

6 J-r potential for boron precipitation in the accumulator assuming a containment temperature of 600 F (Ref. 6). Analyses perfrede R.E. Ginna Nuclear Power Plant B 3.5.1-3 Revision

Accumulators 1 CFR 50.49 (Ref. 7) assumed a chemical spray solutione pppm boron concentratioqRef. 6). The chemical spray

,anirrrpacts suMp -pTT-and 11i6 gulting effect of chloride and caustic I

ID tress corrosion on mechanical systems and components. The sump pH t) 23so affects the rate of hydrogen generation within containment due to the t e "

p *

%D Ct 4

.('Interaction of CS and sump fluid wi mponents.

The(EiE small break LOCA s

Z§ r performed at the minimum 5 I A L b o.k.

nitrogen cover pressure, since sensitivity analyses have dem onstrated that higher nitrogen cover pressure results in a computed peak clad L ocA4 c

0 AL 45 temperature benefit The maximum nitrogen cover pressure limit

CAI, prevents accumu or relief valve actuation at 800 psig, and ultimately Co~j r5 1preserves acc ulator Integrity.

t FA C C1-A\\

JVV%

The effects on containment mass and energy releases from the

{ tJ Ir "

o

¢

',accumulators are accounted for In the appropriate analyses (Refs. 8 and

,7 COy a> F

~;~e VThe accumulators satisfy Criterion 3 of the NRC Policy Statement.

The LCO establishes the minimum conditions required to ensure that the accumulators are available to accomplish their core cooling safety function following a LOCA. Two accumulators are required to ensure that 100% of the contents of one accumulator will reach the core during a LOCA. This is consistent with the assumption that the contents of one accumulator spill through the break. If less than one accumulator Is injected during the blowdown phase of a LOCA, the ECCS acceptance criteria of 10 CFR 50.46 (Ref. 5) could be violated.

For an accumulator to be considered OPERABLE, the motor-operated isolation valve must be fully open (see Figure B 3.5.2-1a), power removed above 1600 psig, and the limits established In the SRs for contained volume, boron concentration, and nitrogen cover pressure must be met.

APPLICABILITY In MODES 1 and 2. and In MODE 3 with RCS pressure > 1600 psig, the accumulator OPERABILITY requirements are based on full power operation. Although cooling requirements decrease as power decreases, the accumulators are still required to provide core cooling as long as elevated RCS pressures and temperatures exist.

This LCO is only applicable at pressures > 1600 psig. At pressures s 1600 psig, the rate of RCS blowdown Is such that the ECCS pumps can provide adequate injection to ensure that peak clad temperature remains below the 10 CFR 50.46 (Ref. 5) limit of 22000F.

R.E. Ginna Nuclear Power Plant B 3.5.1-4 Revision@

ECCS - MODES 1, 2, and 3 B 3.5.2 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

B 3.5.2 ECCS-MODES 1. 2, and 3 BASES BACKGROUND The function of the ECCS is to provide core cooding and negative reactivity to ensure that the reactor core Is protected after any of the following accidents:

a.

Loss of coolant accident (LOCA) and coolant leakage greater than the capability of the normal charging system;

b.

Rod ejection accident;

c.

Loss of secondary coolant accident, Including uncontrolled steam release or loss of feedwater; and

d.

Steam generator tube rupture (SGTR).

The addition of negative reactivity Is designed primarily for the loss of secondary coolant accident where primary cooldown could add enough positive reactivity to achieve criticality and return to significant power.

There are two phases of ECCS operation: injection and recirculation. In the Injection phase, water is taken from the refueling water storage tank (RWST) and injected into the Reactor Coolant System (RCS) through the cold legs and reactor vessel upper plenum. When sufficient water Is removed from the RWST to ensure that enough boron has been added to maintain the reactor subcritical and the containment sump has enough water to supply the required net positive suction head to the ECCS pum

, suction is switched to Containment Sump B for recirculation.

l 1eJ~hours`simultaneous ECCS injection is used to ce the potential for bollifig In the top of the core and any resulting ron precipitation.

The ECCS consists of two separate subsystems: safety Injection (SI)

.AI <,,0>

and residual heat removal (RHR) (see Figure B 3.5.2-1A). Each L rA K'subsystem consists of two redundant, 100% capacity trains. The ECCS accumulators and the RWST are also part of the ECCS, but are not lJ

'J considered part of an ECCS flow path as described by this LCO.

The ECCS flow paths which comprise the redundant trains consist of piping, valves, heat exchangers, and pumps such that water from the RWST can be Injected into the RCS following the accidents described in this LCO. The major components of each subsystem are the RHR pumps, heat exchangers, and the SI pumps. The RHR subsystem consists of two 100% capacity trains that are Interconnected and R.E. Ginna Nuclear Power Plant B 3.5.2-1 Revision 21

ECCS - MODES 1, 2, and 3 B 3.5.2 For LOCAs that are too small to depressurize the RCS below the shutoff head of the Si pumps, the steam generators provide core cooling until the RCS pressure decreases below the SI pump shutoff head.

During the recirculation phase of LOCA recovery, RHR pump suction Is manually transferred to Containment Sump B (Refs. 4 and 5). This transfer is accomplished by stopping the RHR pumps. isolating RHR from the RWST by closing motor operated Isolation valve 856, opening the Containment Sump B motor operated Isolation valves to RHR (850A and 8508) and then starting the RHR pumps. If motor operated Isolation valve 856 fails to close, check valve 854 provides necessary isolation of the RWST. The SI and CS pumps are then stopped and the RWST isolated by closing motor operated Isolation valve 896A and 896B for the SI and CS pump common supply header and closing motor operated Isolation valve 897 or 898 for the SI pumps recirculation line.

The RHR pumps then supply the SI pumps If the RCS pressure remains above the RHR pump shutoff head as correlated through core exit temperature, containment pressure, and reactor vessel level indications (Ref. 6). The RHR pumps can also provide suction to the CS pumps for containment pressure control. This high-head recirculation path is provided through RHR motor operated Isolation valves 857A, 8578, and 857C. These Isolation valves are Interlocked with valves 896A, 896B, 897, and 898. This Interlock prevents opening of the RHR high-head recirculation isolation valves unless either 896A or 896B are closed and either 897 or 898 are closed. If RCS pressure is such that RHR provided adequate core and containment cooling, the Si and CS pumps remain in) pull-stop. During recircullb-Is discha h the same Paths, as the injection phase.

a hiouou=Nmultaneous ection by the SI and RHR pumps is used to prevent boron precipitation

r. \\

This consists of providing SI through the RCS cold legs and Into f 1 fieer plenum while providing RHR through the 6ore deluge valves Into the upper plenum.

X

£e The two redundant flow paths from Containment Sump B to the RHR

.< > ~. (~

pumps also contain a motor operated Isolation valve located within the 1

4 :,. *sump (851A and 851 8). These Isolation valves are maintained open with l. :r epower removed to Improve the reliability of switchover to the recirculation phase. The operators for isolation valves 851A and 851B are also not qualified for containment post accident conditions. The removal of AC power to these isolation valves is an acceptable design against single failures that could result In an undesirable actuation (Ref. 2).

The Si subsystem of the ECCS also functions to supply borated water to the reactor core following increased heat removal events, such as a steam line break (SLB). The limiting design conditions occur when the negative moderator temperature coefficient Is highly negative, such as at the end of each cycle.

R.E. Ginna Nuclear Power Plant B 3.5.2-3 Revision 21

ECCS - MODES 1. 2. and 3 B 3.

5.2 REFERENCES

1.

Letter from R. A. Purple, NRC, to L. D. White, RG&E,

Subject:

"Issuance of Amendment 7 to Provisional Operating Ucense No.

DPR-1 8,* dated May 14, 1975.

2.

Branch Technical Position (BTP) ICSB-18, 'Application of the Single Failure Criterion to Manually-Controlled Electrically Operated Valves."

3.

Letter from A. R. Johnson, NRC. to R. C. Mecredy, RG&E,

Subject:

'Issuance of Amendment No. 42 to Facility Operating License No.

DPR-1 8, R. E. Ginna Nuclear Power Plant (TAC No. 79829),- dated June 3,1991.

4.

Letter from D. M. Crutchfield, NRC, to J. E. Maler, RG&E,

Subject:

'SEP Topic VI-7.B: ESF Switchover from Injection to Recirculation Mode, Automatic ECCS Realignment, Ginna," dated December 31.

1981.

5.

NUREG-0821.

6.

UFSAR, Section 6.3.

7.

Lter f

. M. CrutchfieldN<C, to J. E.

r. RG&

Jed:

y

, Boron Optlon SystemVE2Ginn;

!dated J

gus 198

8.

Atomic Industrial Forum (AIF) GDC 44, Issued for comment July 10, 1967.

9.

10 CFR 50.46.

10.

UFSAR, Section 15.6.

11.

UFSAR, Section 6.2.

12.

NRC Memorandum to V. Stello, Jr., from R.L. Baer, "Recommended Interim Revisions to LCOs for ECCS Components,' December 1, 1975.

R.E. Ginna Nuclear Power Plant B 3.5.2-1 2 Revision 21

RWST B 3.5.4

b.

Sufficient water volume exists in the containment sump to support continued operation of the ECCS and CS pumps at the time of transfer to the recirculation mode of cooling; and

c.

The reactor remains subcritical following a LOCA.

Insufficient water In the RWST could result In inadequate NPSH for the RHR pumps when the transfer to the recirculation mode occurs.

Improper boron concentrations could result In a reduction of SDM or excessive boric acid precipitation In the core following the LOCA, as well as excessive caustic stress corrosion of mechanical components and systems Inside the containment.

APPLICABLE SAFETY ANALYSES During accident conditions, the RWST provides a source of borated water to the ECCS and CS pumps. As such, it provides containment cooling and depressurization, core cooling, and replacement Inventory and is a source of negative reactivity for reactor shutdown (Ref. 3). The design basis transients and applicable safety analyses concerning each of these systems are discussed in the Applicable Safety Analyses section of LCO 3.5.2, "ECCS-MODES 1, 2, and 3"; LCO 3.5.3, "ECCS-MODE 4"; and LCO 3.6.6, "Containment Spray (CS), Containment Recirculation Fan Cooler (CRFC), NaOH, and Containment Post-Accident Charcoal Systems." These analyses are used to assess changes to the RWST in order to evaluate their effects in relation to the acceptance limits in the analyses.

The RWST must also meet volume, boron concentration, and temperature requirements for non-LOCA events. The volume Is not an explicit assumption in non-LOCA events since the volume required for Reactor Coolant System (RCS) makeup is a small fraction of the available RCS volume. The deliverable volume limit is set by the LOCA and containment analyses. For the RWST, the deliverable volume Is selected such that switchover to recirculation does not occur until sufficient water has been pumped into containment to provide necessary NPSH for the RHR pumps. The minimum boron concentration is an explicit assumption in the steam line break (SLB) analysis to ensure the required shutdown capability. The maximum boron concentration is an explicit assumption in the evaluation of chemical effects resulting from the operation of the CS System.

F 5 ra lar e k LOCA analysis, the minimuyrwater volum Woof c130 gallons and thrwer boroncpnc otration liiare used to 7 A /compute the post LOCXmp boron concentration necessary to assure subcriticality. The large break LOCA is the limiting case since the safety analysis assumes that all control rods are out of the core.

R.E. Ginna Nuclear Power Plant B 3.5.4-2 Revision

Insert A Minimum mass and minimum boron concentrations for significant boron sources and maximum mass and minimum boron concentration for significant dilution sources

RWST B 3.5.4 I

The upper limit on boron concentration Is used to determine the time frame In which boron precipitation is addressed post LOCA. The maximum boron concentration limit is based on the coldest expected temperature of the RWST water volume and on chemical effects resulting from operation of the ECCS and the CS System. A value -~

would not create the potential for boron precipitation In the RWST ssumln an Auxiliary Building temperature of 500F (Ref. 1). Analyses l1+ p 6 Jdrss performe ei

!10 CFR 50.49 (Ref. 2) assumed a chemical pm boron concentratioA(Ref. 1). The chemical s r olution Impacts sump pH and th e

ing effect of o 5 a 0~

c oude and caustic stress corrosion on mechanical systems and components. The sump pH also affects the rate of hydrogen generation

a. cc a within containment due to the interaction of CS and sump fluid with 1

i "

IH d

~"

e WS satisfiouminum components.

PpAi So roo Cvop-Cc e RWST satisfies Criterion 3 of the NRC Policy Statement.

% 4,L Sr LCO The RWST ensures that an adequate supply of borated water Is available to cool and depressurize the containment in the event of a Design Basis Accident (DBA), to cool and cover the core In the event of a LOCA, to maintain the reactor subcritical following a DBA, and to ensure adequate level in the containment sump to support ECCS and CS pump operation In the recirculation mode.

To be considered OPERABLE, the RWST must meet the water volume and boron concentration limits established in the SRs.

APPLICABILITY In MODES 1, 2, 3, and 4, RWST OPERABILITY requirements are dictated by ECCS and CS System OPERABILITY requirements. Since both the ECCS and the CS System must be OPERABLE In MODES 1, 2, 3, and 4, the RWST must also be OPERABLE to support their operation.

Core cooling requirements In MODE 5 are addressed by LCO 3.4.7,

'RCS Loops - MODE 5, Loops Filled," and LCO 3.4.8, "RCS Loops -

MODE 5. Loops Not Filled." MODE 6 core cooling requirements are addressed by LCO 3.9.4, "Residual Heat Removal (RHR) and Coolant Circulation-Water Level 2 23 Ft," and LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation-Water Level c 23 Ft."

ower Plant B 3.5.4-3 Revision R.E. Ginna Nuclear Pc

ENCLOSURE 5 R.E. Ginna Nuclear Power Plant Application of Westinghouse Best-Estimate Large Break LOCA Methodology to the R. E. Ginna Nuclear Plant

APPLICATION OF WESTINGHOUSE BEST-ESTIMATE LARGE BREAK LOCA METHODOLOGY TO THE R. E. GINNA NUCLEAR PLANT A best-estimate loss of coolant accident analysis has been completed for the R. E. Ginna Nuclear Plant.

This license amendment request (LAR) for R. E. Ginna requests approval to apply the Westinghouse best-estimate large break loss of coolant accident (LOCA) analysis methodology.

Westinghouse obtained generic NRC approval of its original topical report describing best-estimate large break LOCA methodology in 1998. NRC approval of the methodology is documented in the NRC safety evaluation report appended to the topical report [1]. This methodology was later extended to 2-loop Westinghouse plants with Upper Plenum Injection (UPI) in 1999 as documented in the NRC safety evaluation report appended to the UPI topical report [2].

Westinghouse recently underwent a program to revise the statistical approach used to develop the Peak Cladding Temperature (PCT) and oxidation results at the 95*' percentile. This method is still based on the CQD methodology [1],[2] and follows the steps in the CSAU methodology. However, the uncertainty analysis (Element 3 in CSAU) is replaced by a technique based on order statistics. The ASTRUM methodology replaces the response surface technique with a statistical sampling method where the uncertainty parameters are simultaneously sampled for each case. The approved ASTRUM evaluation model is documented in WCAP-16009-P-A [3].

This report summarizes the application of the Westinghouse ASTRUM BELOCA evaluation model to the R. E. Ginna Nuclear Plant for analysis of the large break LOCA.

The WCOBRAITRAC model for the R. E. Ginna Nuclear Plant was developed to support a fuel change from OFA to 422V+ fuel. This analysis was also performed at an increased core power level to support an expected future extended power uprate. 'The 'analysis at the increased power level bounds operation at a lower power level. Table 1 lists the major plant pjarameter assumptions used in the BE LOCA analysis for the R. E. Ginna Nuclear Plant. Table 2 summarizes the results of the ASTRUM analysis. Table 3 contains a sequence of events for the limiting PCT transient.

The ASTRUM methodology requires the execution of 124 transients to determine a bounding estimate of the 95th percentile of the Peak Clad Temperature (PCT), Local Maximum Oxidation (LMO), and Core Wide Oxidation (CWO) with 95% confidence level. These parameters are needed to satisfy the 10 CFR 50.46 criteria with regard to PCT, LMO, and CWO. From these 124 calculations, run 106 proved to be the limiting PCT transient and the limiting LMO transient, and run 069 the limiting CWO transient.

This analysis is in accordance with the applicability limits and usage conditions defined in Section 13-3 of WCAP-16009-P-A [3] as applicable to the ASTRUM methodology. Section 13-3 of WCAP-16009-P-A

[3] was found to acceptably disposition each of the identified conditions and limitations related to WCOBRA/TRAC and the CQD uncertainty approach per Section 4.0 of the ASTRUM Final Safety Evaluation Report appended to this WCAP.

The scatter plot presented on Figure 2 shows the Ieffect of the effective break area on the analysis PCT.

The effective break area is calculated by multipfyimg the discharge coefficient CD with the sample value of the break area, normalized to the cold-leg cross' sctional area. Figure 2 is provided because the break area is a significant contributor to the variation in PCT.

A BNFL Group company

Figure 3 shows the predicted clad temperature transient at the PCT limiting elevation for the limiting PCT case. Figure 4 presents the clad temperature transient predicted at the LMO elevation for the limiting LMO case. Figure 5 shows the PCT trace for the CWO limiting transient.

Based on the results as presented in Table 2, it is concluded that the R. E. Ginna Nuclear Plant continues to maintain a margin of safety to the limits prescribed by 10 CFR 50.46.

References

[I] Bajorek, S. M., et. al., 1998, "Code Qualification Document for Best Estimate LOCA Analysis,"

WCAP-12945-P-A, Volume 1, Revision 2 and Volumes 2 through 5, Revision 1, and WCAP-14747 (Non-Proprietary).

[2] Dederer, S. I., et. Al., 1999, "Application of Best Estimate Large Break LOCA Methodology to Westinghouse PWRs with Upper Plenum Injection," WCAP-14449-P-A, Revision 1.

[3] Nissley, M. E., et.al., 2005, "Realistic Large Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)," WCAP-16009-P-A.

A BNFL Group company

Table I Mlajor Plant Parameter Assumptions Used in the BE LOCA Analysis for Ginna Parameter Value A. Plant Physical Description SG Tube Plugging l510%

B. Plant Initial Operating Conditions Reactor Power lS100% of 1811 MWt Peaking Factors l F0H s 1.72 Axial Power Distribution l See Figure 1 C. Fluid Conditions TAVG 564.6 5 TAvG S 576.0 F TAVG Uncertainty

+/- 4 'F Pressurizer Pressure 2190 psia 5 PRCS S 2310 psia Reactor Coolant Flow 2 85100 gpm/loop Accumulator Temperature 60 OF 5 TACC S 120 OF Accumulator Pressure 714.7 psia 5 PACC 5 804.7 psia Accumulator Water Volume 1090 ft3 S VAcc S 1140 ft3 Accumulator Boron Concentration a 2100 ppm D. Accident Boundary Conditions Single Failure Assumptions Loss of one ECCS train Safety Injection Flow Minimum Safety Injection Temperature 50 °F s Ts, s 104 °F Low Head Safety Injection 5 19 sec (with offsite power)

Initiation Delay Time 5 30 sec (without offsite power)

High Head Safety Injection s 21 sec (with offsite power)

Initiation Delay Time Is 32 sec (without offsite power)

Containment Pressure Bounded (minimum)

A BNFL Group company

Table 2 Ginna Best Estimate Large Break LOCA Results 10 CFR 50.46 Requirement Value Criteria 95/95 PCT (0F) 1870

< 2200 95/95 LMO (%)

3.4

<17 95/95 CWOt (%)

0.30

< 1 Note, these results are applicable to both the new fuel (422V+) and the resident fuel (OFA)

Table 3 Ginna Best Estimate Large Break Sequence of Events for Limiting PCT Case Event Time (sec)

Start of Transient 0.0 Safety Injection Signal 4.0 Accumulator Injection Begins 13.0 Low Head Safety Injection Begins 34.0 High Head Safety Injection Begins 36.0 End of Blowdown 38.0 Bottom of Core Recovery 46.0 Accumulator Empty 49.0 PCT Occurs 242.0 End of Transient 450.0 Local Maximum Oxidation t

Core Wide Oxidation A BNFL Group company

0.45 0.305, 0.44 0.4 0.435. 0.37 I..

° 0.35 0.3 0.305, 0.29 0.25

>0.435, 0.215 0.2 0.2 0.25 0.3 0.35 0.4 0.45 0.5 PMID Figure 1 Ginna BELOCA Analysis Axial Power Shape Operating Space Envelope A BNFL Group company

2000 1800 -

1600 -

b.

I-1a 0o 2Ci140)

C)

C--)

~3_

1200 -

1000 -

800 -

U U

IN U

U a

U U

Ul Wj is s*

0"*U U.*

• i UU*

0 0 4y.

a 0

N IU 0 0U U

I I

.4'.

I I

6 1.5 Effective Break Area 25 Figure 2 Ginna BELOCA Analysis WCOBRA/TRAC hot Rod PCT v. Effective Break Area Scatter Plot A BNFL Group company

2000 1800-1600-1400-CL.

E1200 -

/

100 800 2

100 260 300 400 500 Time After Break (s)

Figure 3 Ginna BELOCA Analysis Hotspot Clad Temperature Transient at the Limiting Elevation for the Limiting PCT Case A BNFL Group company

2000

-1400-CL E 1200-1000-800 600-0 10 200 300 400 500 Time After Break (s)

Figure 4 Ginna BELOCA Analysis Hotspot Clad Temperature Transient at the Limiting Elevation for the Limiting LAIO Case A BNFL Group company

1800 1600 1400 -

1200 -

L.

-31000 800 600 400 -

200 -

0 100 200 300 400 500 Time After Break (s)

Figure 5 Ginna BELOCA Analysis WCOBRA/TRAC Hot Assembly PCT Transient for the Limiting CWVO Case A BNFL Group company

ENCLOSURE 6 R.E. Ginna Nuclear Power Plant Technical Evaluation - Small Break Loss-of-Coolant Accident

Small Break Loss-of-Coolant Accident Introduction The small break loss-of-coolant accident (SBLOCA) analysis is described in Ginna UFSAR section 15.6.4.1. A LOCA is defined as a rupture of the reactor coolant system (RCS) piping or of any line connected to the system. The SBLOCA includes all postulated pipe ruptures with a total cross-sectional area less than 1.0 ft2. The SBLOCAs analyzed in this section are for those breaks beyond the capability of a single charging pump resulting in the actuation of the emergency core cooling system (ECCS). The analysis was performed to demonstrate conformance with the 10 CFR 50.46 requirements for the conditions associated with the EPU.

Input Parameters, Assumptions, and Acceptance Criteria Key input parameters are summarized in Tables I through 3. 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. Furthermore, the SBLOCA analysis is based on Ginna Station specific models.

The acceptance criteria for the SBLOCA analysis are specified in 10 CFR 50.46, as follows:

I. The calculated maximum fuel element cladding temperature shall not exceed 2,2001F.

2.

The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation.

3. The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.

4.

Calculated changes in core geometry shall be such that the core remains amenable to cooling.

5. Afler any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core. (Note that this criterion is not addressed as part of the short-term SBLOCA analysis. Please refer to Post-LOCA Section.)

Description of Analyses and Evaluations The SBLOCA analysis was performed for the EPU using the 1985 Westinghouse SBLOCA Evaluation Model with NOTRUMP (NOTRUMP-EM) [References I through 3], including NRC approved changes to the methodology as described in References 4 and 5. Westinghouse obtained generic NRC approval of the NOTRUMP computer code's modeling capabilities and solution techniques [ I ] and the use of the NOTRUMP computer code for licensing applications [2] in 1985. NRC approval of additional modeling details [3], such as limiting break location was obtained in 1986. The NOTRUMP-EM was later revised A BNFL Group company

[4] and granted generic NRC approval for an improved condensation model and related changes in safety injection modeling assumptions for safety injection to the reactor coolant system (RCS) cold legs. Most recently, the NRC generically approved updates to the NOTRUMP-EM to include the ability to model annular fuel pellets [5] in the fuel rod heat-up calculations.

The analysis modeled the appropriate plant-specific design features of Ginna Station such as an accumulator water volume range of 1090 to 1140 fR3, a flat upper support plate, a downflow barrel-baffle region, and an upper head temperature equal to the hot leg temperature (Tiior). All prior 10 CFR 50.46 assessments were incorporated into the analysis, primarily through the use of corrected code versions and selection of input values.

Results Tables 4 and 5 provide the NOTRUMP and SBLOCTA results for the SBLOCA analysis, respectively.

The peak cladding temperature is 1,1670F, and the maximum local transient oxidation is 0.07 percent.

The design limit 95% upper bound pre-transient oxidation value for each of the fuel designs that will be included in the EPU cores is < 16%. The actual upper bound values predicted for each of the fuel designs are expected to be well below this value. Because the transient oxidation is so low, the sum of the transient and pre-transient oxidation remains below 16% at all times in life. The core-wide hydrogen generation remains well below the 10 CFR 50.46 acceptance limit of 1 percent, and the core geometry remains amenable to cooling. The transient results for the limiting analysis case are provided in Figures I to S.

Conclusions The SBLOCA analysis results meet the pertinent acceptance criteria of 10 CFR 50.46. The peak cladding temperature is less than 2,2001F; the maximum local oxidation is less than 17 percent; the core-wide hydrogen generation is less than I percent; and, the core geometry remains amenable to cooling.

References

1. WCAP-10079-P-A, "NOTRUMP - A Nodal Transient Small Break and General Network Code,"

Meyer, P. E., August 1985.

2.

WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," Lee, N., et al., August 1985.

3.

WCAP-1 1145-P-A, "Westinghouse Small Break LOCA ECCS Evaluation Model Generic Study with the NOTRUMP Code," Rupprecht, S. D., et al., October 1986.

4.

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," Thompson, C. M., et al., July 1997.

5.

WCAP-I4710-P-A, "I-D Heat Conduction Model for Annular Fuel Pellets," Shimeck, D. J., May 1998.

A BNFL Group company

Table I Input Assumptions and Initial Conditions A. Core Parameters Analyzed Core Power Level 1811 MWt Calorimetric Uncertainty 0%

Fuel Type 422V+, OFA Total Core Peaking Factor, FQ 2.60 Channel Enthalpy Rise Factor, Fm 1.72 Axial Offset

+ 25%

K(z) Limit 1.0 everywhere B. Reactor Coolant System Thermal Design Flow 85,100 gpm/loop Nominal Vessel Average Temperature Range 564.6 - 576.0 0F Pressurizer Pressure 2250 psia Pressurizer Pressure Uncertainty 60 psi C. Reactor Protection System Reactor Trip Setpoint 1730 psia Reactor Trip Signal Processing Time (Includes Rod Drop Time) 5.0 seconds D. Auxiliary Feedwater System Maximum AFW Temperature 104 OF Minimum AFWV Flow Rate 0 - 170 gpm/SG Initiation Signal ILow Pressurizer Pressure Si Signal AFW Delivery Delay Time 600 seconds E. Steam Generators Steam Generator Tube Plugging

• 10%

MFW Isolation Signal Low Pressurizer Pressure SI Signal MFW Isolation Delay Time 2.0 seconds MFW Flow Coastdown Time 10.0 seconds Feedwater Temperature 390 - 435 'F Steam Generator Safety Valve Flow Rates Table 2 F. Safety Injection Limiting Single Failure Loss of one Emergency Diesel Generator Maximum Si Water Temperature 104 "F A BNFL Group company

Table I Input Assumptions and Initial Conditions Low-Low Pressurizer Pressure Signal 1715 psia Si Delay Time 32 seconds Safety Injection Flow Rates Table 3 G. Accumulators Water/Gas Temperature 120 OF Initial Accumulator Water Volume' 111 fi3 Minimum Cover Gas Pressure 714.7 psia II. RWN'ST Draindown Input Maximum Containment Spray Flow 1800 gpm per pump Minimum Usable RWST Volume 184,950 gal Maximum Delay Time for Switchover to Cold Leg Recirculation, sec 600 seconds (HHSI)

Minimum Si Flow Rate During Switchover No Si flow is modeled during switchover.

Minimum Si Flow Rate After Switchover Table 3 Maximum SI Water Temperature After Switchover to Cold Leg 212 'F Recirculation Signal is Generated

1. Corresponds to a range of 1090 ft3to 1140 ft3.

A BNFL Group company

Table 2 Steam Generator Safety Valve Flows Per Steam Generator MISSV Set Pres (psig)

Uncert. (%)

Accum. (%) Rated Flow at Full Open Pressure (lbm/hr) 1 1085 1

3 797,700 2

1140 1

3 837,600 3

1140 1

3 837,600 4

1140 1

1 i

3 837,600 A BNFL Group company

Table 3 Safety Injection Flows vs. Pressure Minimum Safeguards, Spill to RCS and Containment Pressure RCS Pressure (psia)

Intact Loop Injection Flow Broken Loop Injection Broken Loop Spill Flow (gpm)

Flow (gpm)

(gpm)

(breaks < 8.75 in. diameter)

(breaks > 8.75 in. diameter) 14.7 300.0 300.0 385.0 114.7 300.0 300.0 385.0 214.7 300.0 300.0 385.0 314.7 300.0 300.0 385.0 414.7 300.0 300.0 385.0

.514.7 300.0 300.0 385.0 614.7 288.9 288.9 385.0 714.7 272.7 272.7 385.0 814.7 252.9 252.9 385.0 914.7 229.1 229.1 385.0 1014.7 200.7 200.7 385.0 1114.7 166.6 166.6 385.0 1214.7 125.1 125.1 385.0 1314.7 62.0 62.0 385.0 1389.7 0.0 0.0 385.0 A BNFL Group company

Table 4 NOTRUNIP Transient Results Event Time (sec)

IS5 Inch 2 Inch 3 Inch Break Initiation 0

0 0

Reactor Trip Signal 51.0 25.6 11.4 S-Signal 58.6 26.7 11.9 Safety Injection Begins 90.6 58.7 43.9 Loop Seal Clearing*

985 511 236 Core Uncovery 2820 1157 415 Accumulator Injection Begins 8544 2832 673 RWN'ST Low Level N/A 5357 2683 Core Recovery 4750 2570 897

• Loop seal clearing is defined as break vapor flow > I lb/s A BNFL Group company

Table 5 Beginning of Life (BOL) Rod Ileatup Results 1.5 Inch 2 Inch Case 3 Inch Time-in-Life BOL BOL BOL PCT(0F) 1011 1167 1117 PCT Time (s) 3578 1650 748 PCT Elevation (ft) 11.25 11.50 11.00 IIR Burst Time (s)

N/A N/A N/A

}1R Burst Elevation (ft)

N/A N/A N/A Maximum Local ZrO2 (%)

0.02 0.07 0.02 Maximum Local ZrO2 Elev (ft) 11.25 11.25 11.25 Total Hydrogen Generation (%)

<< 1.0

<< 1.0

<< 1.0 A BNFL Group company

2500 200 0

0 1 0 -

- - -- - - --5 C,,

C.

10 0

500 -

I 02600 4d00 Ti me (s)

Figure 1: Pressurizer Pressure 2-Inch Break A BNFL Group company

Core Mixture Level Top of Core 32 30 28 26

_" 24 22 20 18 16 Time (s)

Figure 2: Core Mixture Level 2-Inch Break A BNFL Group company

B Broken Loop Pumped Si Flow Intact Loop Pumped Si Flow 50 Cn C.

10 0

2000 4000 6000 Time (s)

Figure 3: Broken Loop and Intact Loop Pumped SI Flow Rate 2-Inch Break A BNFL Group company

1200 -

1100 -

- I- - -

1000 -------------

900 -

_I I

2 500- -

400-,

1 I

40 2

4d00 6000 Time (s)

Figure 4: Peak Cladding Temperature at PCT Elevation 2-Inch Break A BNFL Group company

120 -

-00 80 --

U, 40-0 2

22 0

4000 Time (s)

Figure 5: Core Exit Vapor Flow 2-Inch Break A BNFL Group company

ENCLOSURE 7 R.E. Ginna Nuclear Power Plant Ginna EPU Post LOCA Analysis

Ginna EPU Post LOCA Analyses Introduction As part of the EPU, the Ginna Station Technical Specification refueling water storage tank (RWST) and accumulator boron concentration limits (Technical Specifications Sections 3.5.4 and 3.5.1) are being changed. To support the uprated core power and the changes to the Technical Specification boron concentration specifications, post-loss-of-coolant accident (LOCA) long-term cooling evaluations were performed. These evaluations demonstrated that the EPU and Technical Specification boron concentration changes are acceptable with respect to post-LOCA long-term cooling. The injection and recirculation ECCS modes are described in the Ginna UFSAR, Sections 6.3.2.3.3 and 6.3.2.3.4. The switchover from injection mode to sump recirculation mode is described in the Ginna UFSAR, Section 6.3.3.3.

The Constellation Generation Group (CGG) licensing position for satisfying the requirements of IOCFR50.46 Paragraph (b), Item (5), is documented in WCAP-8339 (Reference 1). Reference I states that the core will remain subcritical post-LOCA by borated water from various injected Emergency Core Cooling System (ECCS) water sources. Post-LOCA sump boron calculations demonstrate that the core will remain subcritical upon entering the sump recirculation phase of ECCS injection. Containment sump boron concentration calculations were used to develop a core reactivity limit that was confirmed as part of the Westinghouse Reload Safety Evaluation Methodology (Reference 2).

The Westinghouse ECCS evaluation model relies on precluding boric acid precipitation for satisfying the requirements of IOCFR50.46 Paragraph (b), Item (4) and IOCFR50.46, Paragraph (b), Item (5). The potential for boric acid precipitation during post-LOCA is evaluated by calculations that consider the buildup of boric acid solution in the core for limiting scenarios. The limiting scenarios are large breaks where the break location and the point of safety injection are such that the water in the core is stagnant.

Borated ECCS water boils off due to decay heat, leaving behind boric acid. The concern is the potential for the boric acid solution in the core to reach the boric acid solubility limit. Boric acid precipitation during long-term cooling is addressed in the Ginna Updated Final SafetyAnalysis Report (UFSAR),

Section 6.3.3.4. Actions required to preclude boric acid precipitation are described in Technical Specification Section B 3.5.2 Long-term core cooling also requires adequate ECCS flow during the sump recirculation period. ECCS pump availability and specific flow path alignments may reduce ECCS recirculation flow as compared to the flows available during the injection phase.

Input Parameters, Assumptions, and Acceptance Criteria Post-LOCA Sump Boron Calculations The input parameters used in the sump boron calculations are given in Table 1.

The sump boron concentration calculational model is based on the following assumptions:

The calculation of the sump mixed mean boron concentration assumes minimum mass and minimum boron concentrations for significant boron sources and maximum mass and minimum boron concentration for significant dilution sources.

• Boron is mixed uniformly in the sump. The post-LOCA sump inventory is made up of constituents that are equally likely to return to the containment sump; i.e., selective holdup in containment is neglected.

A BNFL Group company

The sump mixed mean boron concentration is calculated as a function of the pre-trip reactor coolant system (RCS) conditions.

There are no specific acceptance criteria when calculating the post-LOCA sump boron concentration.

However, the resulting sump boron concentration, which is calculated as a function of the pre-LOCA RCS boron concentration, is reviewed for each cycle-specific core design to confirm that adequate boron exists to maintain subcriticality in the long-term post-LOCA.

Potential for Boric Acid Precipitation The major inputs to the boric acid precipitation calculation include core power assumptions and assumptions for boron concentrations and water volume/masses for significant contributors to the containment sump. The input parameters used in the Ginna EPU boric acid precipitation calculations are given in Table 2.

The boric acid precipitation calculation model is based on the following assumptions:

• A boric acid concentration level is computed over time for a core-region mixing volume.

• The boric acid concentration limit is the experimentally determined boric acid saturation concentration with a four weight-percent uncertainty factor.

• The decay heat generation rate is based on the 1971 American Nuclear Society Standard for a finite operating time.

• The boron concentration of the make-up safety injection water during recirculation is a calculated sump mixed mean boron concentration. The calculation of the sump mixed mean boron concentration assumes maximum mass and maximum boron concentrations for significant boron sources, and minimum mass and maximum boron concentrations for significant dilution sources.

• The methodology items listed above are consistent with, or otherwise are conservative with respect to, the methodology described in CLC-NS-309 (Reference 3).

There were no specific acceptance criteria for the results of the calculations that preclude boric acid precipitation. However, the UFSAR, the Tech Specs, and the Emergency Operating Procedures (EOPs) must be consistent with the maximum time to establish simultaneous hot leg / cold leg injection.

ECCS Recirculation Flows ECCS recirculation flows are evaluated by comparing minimum safety injection pump flows to the flows necessary to dilute the core, and the flows necessary to replace core boiloff, thus keeping the core covered.

The evaluation of ECCS recirculation flows was based on the following assumptions:

The decay heat generation rate for core boil-off calculations used in evaluating ECCS recirculation flows is based on the 1971 American Nuclear Society Standard for an infinite operating time with 20%

A BNFL Group company

uncertainty. The assumed core power includes a multiplier to address instrument uncertainty as identified by Section l.A of 1 OCFR50, Appendix K.

The small break LOCA (SBLOCA) analysis considers ECCS flow and enthalpy changes that may occur during the switchover from injection mode to sump recirculation mode.

Table 1 Post-LOCA Sump, Boron Calculation Input Parameters Parameter Current EPU Analyzed Core Power (MWt) 1550 1811 RWVST Boron Concentration, Minimum (ppm) 2300 2750 Accumulator Boron Concentration, Minimum (ppm) 2100 2550 RWST Volume, Minimum (gallons) 184,950 229,475 Total Sump Water Mass (Ibm) 2,202,904 1,979,197 Table 2 Post-LOCA Boric Acid Precipitation Evaluation Inp t Parameters Parameter Value Analyzed Core Power (MWt) 1811 Decay Heat Standard 71 ANS H-3BO3 Solubility Limit (w/o) 23.53 RWST Boron Concentration, Maximum (ppm) 3050 Accumulator Boron Concentration, Maximum (ppm) 3050 RWST Volume, Maximum (gallons) 323,000 Total Sump Water Mass (Ibm) 3,027,271 Description of Analyses and Evaluations With respect to post-LOCA criticality, a post-LOCA subcriticality boron limit curve was developed for the EPU plant conditions. The curve will be included in the Reload Safety Analysis Checklist (RSAC) for each reload cycle. Provided that the cycle-specific maximum critical boron concentration remains below the post-LOCA sump boron concentration limit curve (for all rods out, no Xenon, 680 - 2121F), the core will remain subcritical post-LOCA, and decay heat can be removed for the extended period required by the remaining long-lived radioactivity.

With respect to boric acid precipitation, Ginna is an upper plenum injection (UPI) design. 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 signal starts both high head SI 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 high head SI pumps. After a period of time (less than six hours), operators will be instructed to restart the high head safety injection pumps to re-establish simultaneous cold side and hot side (UPI) injection to provide long term core cooling for all LOCA scenarios.

Three categories of LOCA break sizes were considered for the boric acid precipitation evaluation: (1) large breaks (greater than approximately 6" in diameter) where the RCS pressure rapidly decreases to the A BNFL Group company

UPI initiation pressure (140 psia) with no operator action, (2) intermediate breaks (between approximately I" and 6" in diameter) where RCS pressure decreases but stabilizes above the UPI initiation pressure, and (3) small breaks (approximately I" in diameter and smaller) where high head safety injection refills the RCS and natural circulation is established.

For large breaks in the cold leg, boric acid precipitation cannot occur since the RCS will depressurize quickly and upper plenum injection will provide flushing flow through the core.

For large breaks in the hot leg, the core region boric acid concentration will only begin to increase with the termination of high head safety injection to the cold legs. Calculations have shown that the boric acid solution will not approach the solubility limit for atmospheric pressure conditions until approximately 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after the termination of SI to the cold leg. For the EPU, the Ginna Emergency Operating Procedure ES-I.3, Transfer to Cold Leg Recirculation will be revised to instruct operators to re-establish cold leg SI (i.e. simultaneous injection) no later than 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after the termination of SI in the cold leg. In this case boric acid precipitation will be prevented. There are no limitations on early switchover to simultaneous injection.

For intermediate breaks in the cold leg, RCS pressure will stabilize above the UPI initiation pressure and the core region boric acid concentration will begin to increase prior to upper plenum injection.

Emergency Operating Procedure ES-1.2, Post-LOCA Cooldown and Depressurization directs the operators in this scenario to depressurize the RCS using the condenser steam dumps or SG atmospheric relief valves. Since the core boil-off rate would decrease at higher RCS pressures, the previously discussed calculations that assume atmospheric conditions conservatively predict the rate of increase in boron concentration. 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.

Operators will depressurize the RCS to less than 140 psia within six hours. Based on the bounding calculations for large break LOCA, 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 intermediate breaks in the hot leg, RCS pressure will again stabilize above the UPI initiation pressure, however the core boric acid concentration will not increase until the high head cold leg SI is terminated.

Operators are again directed to depressurize the RCS, and maintain UPI using the RHR pumps on recirculation and terminate the high head SI as necessary. Once high head SI to the cold leg is terminated, this scenario is bounded by the large hot leg break scenario where cold leg SI (i.e.

simultaneous injection) will be re-established no later than 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after termination.

For small hot leg or cold leg breaks the RCS remains pressurized with the high head SI pumps. Natural circulation will be established and will maintain flushing flow through the core for either hot or cold side breaks. Emergency Operating Procedure (EOP) actions will depressurize and cooldown the reactor under controlled conditions with eventual realignment to RHR normal shutdown cooling. Natural circulation or RHR normal shutdown cooling will dilute any buildup of boric acid in the core.

Results A post-LOCA subcriticality boron limit curve was developed for the EPU plant conditions. Cycle-specific reload safety evaluations will ensure that the core will remain subcritical post-LOCA, and decay heat can be removed for the extended period required by the remaining long-lived radioactivity.

Calculations with EPU conditions have shown that boric acid precipitation is not a concern during the injection phase, since the cold leg safety injection and upper plenum injection will be sufficient to promote flow through the core. For large hot leg breaks with no cold leg safety injection during an extended period in sump recirculation, boric acid precipitation will be prevented if cold leg safety A BNFL Group company

injection (i.e., simultaneous injection) is re-established 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after the termination of safety injection in the cold leg.

ECCS flows during sump recirculation were shown to be adequate to prevent boric acid precipitation and to provide long-term core cooling for EPU plant conditions.

References

1. WCAP-8339, "Westinghouse Emergency Core Cooling System Evaluation Model - Summary," June 1974.

2.

WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985.

3.

Letter from C. L. Caso to T. M. Novak-, Chief, Reactor Systems Branch, NRC, from Manager, Safeguards Engineering, Westinghouse Corporation Power Systems, CLC-NS-309, April 1, 1975.

A BNFL Group company

ENCLOSURE 8 R.E. Ginna Nuclear Power Plant List of Regulatory Commitments The following table identifies those actions committed to by R.E. Ginna Nuclear Power Plant, LLC in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.

REGULATORY COMMITMENT DUE DATE Once approved, the amendment will be Prior to startup from the fall 2006 refueling implemented prior to startup from the fall 2006 outage.

refueling outage.

A post-LOCA subcriticality boron limit curve Prior to startup from the fall 2006 refueling was developed for the planned uprated plant outage.

conditions. Cycle-specific reload safety evaluations will ensure that the core will remain subcritical post-LOCA, and decay heat can be removed for the extended period required by the remaining long-lived radioactivity.

Ginna Emergency Operating Procedure ES-1.3, Prior to startup from the fall 2006 refueling Transfer to Cold Leg Recirculation will be outage.

revised to instruct operators to re-establish cold leg SI (i.e. simultaneous injection) no later than 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after the termination of SI in the cold leg.

After the startup from the fall 2006 refueling Subsequent annual 10 CFR 50.46 report after outage, these changes (new PCT results) will the startup from the fall 2006 refueling outage.

be included in the subsequent annual 10 CFR 50.46 report to satisfy the requirements of 10 CFR 50.46(a)(3)(ii).

Additional information regarding the transition Extended Power Uprate submittal currently to 422V+ fuel will be included in the extended planned by June 30, 2005.

power uprate submittal.

I