ML080460504

License Amendment Request: Methodology for Determining Reactor Coolant System Pressure and Temperature and Low Temperature Over Pressure Limits

ML080460504
Person / Time
Site:	Ginna
Issue date:	02/08/2008
From:	John Carlin Constellation Energy Group, Ginna
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML080460504 (17)

Text

John Carlin R.E. Ginna Nuclear Power Plant, LLC Site Vice President 1503 Lake Road Ontario, New York 14519-9364 585.771.5200 585.771.3943 Fax John.Carlin @constellation.com 0 Constellatilon Energy-Nuclear Generation Group February 8, 2008 U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION: Document Control Desk

SUBJECT:

R.E. Ginna Nuclear Power Plant Docket No. 50-244 License Amendment Request: Methodology for Determining Reactor Coolant System Pressure and Temperature and Low Temperature Over Pressure Limits In accordance with the provisions of 10 CFR 50.90, R.E. Ginna Nuclear Power Plant, LLC (Ginna LLC) is submitting a request for an amendment to change the method of determining Reactor Coolant System (RCS) pressure and temperature and Low Temperature Over Pressure (LTOP) limits as described in Technical Specification (TS) 5.6.6.

TS Section 5.6.6 part C.2 describes the methodology used to develop the Ginna heatup and cooldown limit curves. Currently, the methodology employs 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. The change will allow the methodology described in WCAP-14040-A, Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, Revision 4, to be used in developing the operating curves and limits contained in the Ginna Pressure and Temperature Limits Report (PTLR). As an alternative to Section 3.2 (Cold Over Pressurization Mitigating System Setpoint Determination) df WCAP-14040-A, Revision 4, Ginna is proposing the option of continued use of the previously approved Ginna specific methodology for determining the Ginna Low Temperature Over Pressure (LTOP) pressure setpoint using the RELAP5 computer program.

The enclosure to this letter contains Ginna's evaluation of the proposed change, and Attachment 1 to the enclosure contains the marked up changes being requested.

We have considered the possibility of significant hazards associated with this proposed change and have determined that there are none. We have also determined that operation with the

'o00of I~k10-6

Document Control Desk February 8, 2008 Page 2 proposed change would not result in any significant change in the types or amounts of any effluents that may be released offsite, nor would it result in any significant increase in individual or cumulative occupational radiation exposure. Therefore, the proposed change is eligible for categorical exclusion as set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment is needed in connection with the proposed amendment.

This proposed change to the Technical Specifications and our determination of significant hazards have been reviewed by our Plant Operation Review Committee (PORC), and it has concluded that implementation of these changes will not result in an undue risk to the health and safety of the public. A copy of this letter with attachments has been provided to the appropriate state representative per 10CFR50.91(b)(1).

Ginna's existing heatup and cooldown limit curves contained in the PTLR are acceptable for use until 32 Effective Full Power Years (EFPY). We request that this change be approved by February 27, 2009 to ensure implementation prior to reaching the limiting vessel fluence. Once approved, the amendment shall be implemented within 60 days.

This submittal does not contain any new regulatory commitments.

Should you have questions regarding this matter, please contact Mr. Brian Weaver at (585) 771-5219 or Brian.Weaver @constellation.com.

R.E. Ginna Nuclear Power Plant, LLC

Document Control Desk February 8, 2008 Page 3 STATE OF NEW YORK

TO WIT:

COUNTY OF WAYNE I, John T. Carlin, 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 request 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 befre me, a Notary Public in and for the State of New York and,,County of IMII,4//--- , this day of - , 2008.

WITNESS my Hand and Notarial Seal: /'S.d0 JT Ati z. -c Notary Publi  :-

SHARON L.MILLER "

My Commission Expires: Notary Public, State of New York Registration No. 01M16017755 Monroe County

',onmmission Expires December 21, 20._Q Date JC/MR

Enclosure:

Evaluation of the Proposed Change cc: S. J. Collins, NRC D.V. Pickett, NRC Resident Inspector, NRC (Ginna)

P.D. Eddy, NYSDPS J. P. Spath, NYSERDA R.E. Ginna Nuclear Power Plant, LLC

ENCLOSURE Evaluation of the Proposed Change TABLE OF CONTENTS 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requiremenits/Criteria 4.2 Precedent 4.3 Significant Hazards Consideration 4.4 Conclusions

6.0 REFERENCES

ATTACHMENTS

1. Technical Specification Page Markups R.E. Ginna Nuclear Power Plant, LLC February 8, 2008 Page 1 of 7

ENCLOSURE Evaluation of the Proposed Change 1.0

SUMMARY

DESCRIPTION This evaluation supports a request to amend Operating License No. DPR-18, for the R. E. Ginna Nuclear Power Plant (Ginna).

The proposed change would revise the Operating License to update the method used to develop the Reactor Coolant System (RCS) heatup and cooldown and Low Temperature Over Pressure (LTOP) limits per Technical Specification (TS) 5.6.6.c, utilizing current up to date NRC approved methodology.

2.0 DETAILED DESCRIPTION Ginna's RCS operating limits are developed using the methodology contained in TS 5.5.6.c and are maintained in the Pressure Temperature Limit Report (PTLR) under 10CFR50.59. As the result of approved license renewal (LR) and extended power uprate (EPU) amendments for Ginna, it will be necessary to employ current up to date methodology in support of continued operation. This proposed change will modify the method of determining RCS heatup and cooldown limit curves as described in TS 5.6.6.c. Currently, the methodology employs Reference (a) for heatup and cooldown limit curves, and other site specific analysis for LTOP pressurization analysis. The change will allow use of the methodology described in Reference (b) to develop the Cold Overpressure Mitigating System (referred to as LTOP at Ginna) setpoints and the heatup and cooldown limit curves. Ginna is also requesting to maintain the current previously approved LTOP setpoint analysis methodology described in References (c) and (d) as an alternative to section 3.2 of Reference (b). The limits derived using the methodology described in this proposed change will be controlled in the PTLR under 10CFR50.59.

3.0 TECHNICAL EVALUATION

Ginna's current heatup and cooldown curves are calculated per Reference (a) as described in TS 5.6.6, Section c.2. The existing curves cover the period out to 40 effective full power years (EFPY) of operation, which was the current operating license at the time of development. In March 2004 Ginna was given a renewal to extend the operating license for an additional 20 years per Reference (e). Subsequent to the license renewal, Ginna also received a license amendment authorizing a 16.8 percent power uprate per Reference (f). In order to develop curves that address these amendments, Ginna requests to use the methodology in Reference (b), which has incorporated ASME Code Cases 588, 640 and 641. The analytical method described in Reference (b) has been previously reviewed and approved by the NRC as indicated in Reference (g), which is included in Reference (b). Ginna has reviewed Reference (b) and determined that it R.E. Ginna Nuclear Power Plant, LLC February 8, 2008 Page 2 of 7

ENCLOSURE Evaluation of the Proposed Change is applicable to this facility, and that use of the new methodologies will provide the appropriate operating curves and limits for operation to the end of licensed plant life at the new licensed power level (53 EFPY). Therefore, no further technical analysis is required to validate the use of this methodology at Ginna.

Additionally, Ginna is requesting, as an alternative, continued use of the current methodology for LTOP pressure setpoint determination. The proposed alternative (existing) Ginna specific methodology described in new section 5.6.6.c.2 was reviewed by Ginna and determined to be generally consistent with the corresponding section of Reference (b) with the exception that the existing methodology uses the RELAP5 computer code versus the Westinghouse LOFTRAN code. Ginna's use of LTOP for Residual Heat Removal (RHR) system protection is also addressed. Westinghouse has developed new pressure/temperature limit curves for Ginna using the methodology described in Reference (b). The new Ginna curves are less restrictive than the existing curves. Because the LTOP setpoints are based on not exceeding these heatup and cooldown pressure/temperature (10CFR50 Appendix G) limits, and the new Appendix G curves are less restrictive, Ginna has determined that the existing LTOP pressure setpoints remain conservative and that there is no technical justification to re-perform the LTOP setpoint analysis at this time. The existing Ginna LTOP pressure setpoint methodology was submitted to the NRC in Reference (c) and approved by the NRC in Reference (d). As discussed in Section 2.3.2 of Reference (d), this approval of the Ginna methodology for determining the LTOP pressure setpoint was based on its prior NRC approval in Reference (h). Therefore, no further technical analysis is required to validate the continued use of this methodology at Ginna.

The reactor vessel surveillance program is unchanged from that described in Reference (f),

section 2.1.1, with the exception that capsule N is now scheduled for removal during the spring 2008 refueling outage.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Reguirements/Criteria The methodologies requested by Ginna have been previously reviewed and approved by the NRC in the various documents included by reference in this letter. Specifically, Reference (g) states the following relative to the regulatory requirements used in the NRC review:

"Fourspecific topics are addressed in the context of the development of a PTLR methodology: (1) the calculationof neutronfluences for the RPV and RPV surveillance capsules; (2) the evaluation of RPV materialpropertiesdue to changes caused by neutron radiation;(3) the development of appropriateP-T limit curves based on these RPV materialpropertiesand the establishmentof cold overpressuremitigating system (COMS) setpoints to protect the RPV from brittlefailure; and (4) the development of an RPV material surveillance program to monitor changes in RPV materialpropertiesdue R.E. Ginna Nuclear Power Plant, LLC February 8, 2008 Page 3 of 7

ENCLOSURE Evaluation of the Proposed Change to radiation.Regulatory requirements related to the four topics noted above are addressed in Appendices G and H to Title 10 of the Code of FederalRegulations Part50 (10 CFR Part50). Appendix G to 10 CFR Part 50 provides requirements related to RPV P-T limit development and directly or indirectly addresses topics (1) through (3) above.

Appendix H to 10 CFR Part50 defines regulatory requirements related to RPV material surveillanceprogramsand addressestopic (4) above.

For the staff's review of WCAP-14040, Revision 3, several additionalguidance documents were used. NRC StandardReview Plan (SRP) Sections 5.2.2, "Overpressure Protection," 5.3.1, "Reactor Vessel Materials," and 5.3.2, "Pressure-Temperature Limits, "provide specific review guidance related to RPV materialproperty determination,P-T limit development, and COMS performance. Regulatory Guide (RG) 1.99, Revision 2, "RadiationEmbrittlement of Reactor Vessel Materials," describes analysisprocedures acceptable to the NRC stafffor the purpose of assessing RPV materialproperty changes due to radiation.RG 1.190, "Calculationaland Dosimetry Methodsfor Determining Pressure Vessel Neutron Fluence," addressesNRC staff expectationsfor an acceptablefluence calculationmethodology. American Society for Testing and Materials(ASTM) StandardPracticeE 185, "StandardPracticefor Conducting Surveillance Tests for Light-Water Cooled NuclearPower Reactor Vessels,"

provides guidance on the establishmentof RPV material surveillanceprograms and editions of ASTM E 185 are incorporatedby reference into Appendix H to 10 CFR Part

50. American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI, Appendix G provides specific requirements regardingthe development of P-T limit curves."

Changes to the Ginna PTLR as a result of these methodologies will be made pursuant to 10CFR50.59. The PTLR will be provided to the NRC per Ginna Technical Specification 5.6.6.d.

Therefore, all regulatory requirements are met and no further evaluation is required.

4.2 Precedent WCAP-14040, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, Revision 4 (Reference b) was previously approved for use at Comanche Peak (Reference i), and Vogtle (Reference j) 4.3 Significant Hazards Consideration This proposed change to the Operating Licenses submits changes in the method used to develop LTOP setpoints and RCS heatup and cooldown limit curves. Given that all proposed methodologies have been previously reviewed :and approved for use, R.E. Ginna Nuclear Power R.E. Ginna Nuclear Power Plant, LLC February 8, 2008 Page 4 of 7

ENCLOSURE Evaluation of the Proposed Change Plant, LLC believes it is appropriate to use the described new methodology to develop the appropriate operating limits for Ginna.

R.E. Ginna Nuclear Power Plant, 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 change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

Integrity of the reactor vessel is integral to plant safety. It provides containment and continuity for the reactor core, and as part of the reactor coolant system acts as one of the three fission product barriers to the environment. The purpose of the heatup and cooldown limit curves and LTOP setpoints is to ensure vessel integrity through the spectrum of operating modes. Operating within those limits ensures that brittle failure of the vessel material does not occur due to the thermal and pressure stresses the vessel is subjected to during operation. During power operation, the effects of neutron radiation tend to change the characteristics of the vessel material making it more brittle. To compensate for this the operating limits must be periodically adjusted. The methodology being proposed in this submittal is designed to ensure vessel integrity, is analytically sound, and has been reviewed and approved by the NRC. Therefore, the proposed change does 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.

Due to its thickness and material properties, the reactor vessel is the limiting component for brittle fracture in the reactor coolant system. The proposed methodology appropriately limits the operating parameters to preclude the possibility of vessel failure. No new failure mechanisms or accident precursors are introduced as a result of this proposed change.

Therefore, this change does not create the possibility of a new or different type of accident from any accident previously evaluated.

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

Response: No.

R.E. Ginna Nuclear Power Plant, LLC February 8, 2008 Page 5 of 7

ENCLOSURE Evaluation of the Proposed Change The proposed methodology in WCAP- 14040-A, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit curves, Revision 4, contains appropriate margin and has been reviewed and approved by the NRC.

Since the new methodology for developing heatup and cooldown curves will produce less restrictive curves, use of the existing methodology for LTOP setpoints will continue to provide adequate margin to the 10CFR50 Appendix G limits. Therefore, this proposed change does not involve a significant reduction in the margin of safety.

Based on the above, R.E. Ginna Nuclear Power Plant, LLC concludes that the proposed amendment(s) present 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.

4.4 Conclusions 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 compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c) (9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

(a) 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, LLC February 8, 2008 Page 6 of 7

ENCLOSURE Evaluation of the Proposed Change (b) WCAP-14040-A, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit curves, Revision 4 (c) Letter to Guy S. Vissing (NRC) from Robert C. Mecredy (RG&E), Application for Amendment to Facility Operating License, Revision to Reactor Coolant system (RCS) Pressure and Temperature Limits Report (PTLR) Administrative Controls Requirements, Attachment VI, Section 3.2, dated September 29, 1997 (d) Letter to Robert C. Mecredy (RG&E) from S. Singh Bajwa (NRC), R.E. Ginna - Acceptance for Referencing of Pressure Temperature Limits Report, Revision 2 (TAC No. M96529), dated November 28, 1997 (e) Letter to Robert Mecredy (Ginna) from Pao-Tsin Kuo (NRC), License Renewal Safety Evaluation Report for the R.E. Ginna Nuclear Power Plant, dated March 3, 2004.

(f) Letter to Mary Korsnick (Ginna) from Patrick Milano (NRC), R.E. Ginna Nuclear Power Plant - Amendment Re: 16.8 Percent Power Uprate (TAC No. MC7382), dated July 11, 2006 (g) Letter to Gordon Bischoff (Westinghouse) from Herbert Berkhow (NRC), Final Safety Evaluation for Topical Report WCAP-14040, Revision 3, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves" (TAC No. MB5754), dated February 27, 2004 (h) Letter to Robert Mecredy (Ginna) from Jocelyn Mitchell (NRC), R.E. Ginna - Acceptance for Referencing of Pressure Temperature Limits Report, Revision 1 (TAC No. M94770), dated May 23, 1996 (i) Letter to M.R. Blevins (TXU Power) from Jack Donohew (NRC), Comanche Peak Steam Electric Station, Units 1 and 2 - Issuance of Amendments Re: Revise Technical specification 5.6.6 on Reactor Coolant System Pressure and Temperature Limits Report (TAC Nos. MC9500 and MC9501), dated February 22, 2007 (j) Letter to D.E. Grissette (Southern Nuclear Operating Company, Inc.), from Christopher Gratton (NRC), Vogtle Electric Generating Plant, Units 1 and 2 - Issuance of Exemption and Amendments Re: Request to revise Technical Specifications and Pressure Temperature Limits Report and Relocate the Cold Overpressure Protection System (COPS) Arming Temperature (TAC Nos. MC2225, MC2226, MC2227, MC2228, MC3090, and MC3091)

R.E. Ginna Nuclear Power Plant, LLC February 8, 2008 Page 7 of 7

Attachment 1 Technical Specification Page Markups

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 Radiologiical Environmental Operating Reort 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 10 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 4@&

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 3.2.1, "Heat Flux Hot Channel Factor (FQ(Z))";

LCO 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor (FNAH)";

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 Nucleate Boiling (DNB) Limits"; and LCO 3.9.1, "Boron Concentration."

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

Reporting Requirements 5.6

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-1 0216-P-A, Rev. 1A, "Relaxation of Constant Axial Offset Control / FQ Surveillance Technical Specification,"

February 1994.

(Methodology for LCO 3.2.1 and LCO 3.2.3.)

4. WCAP-1 2610-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-10081-A, "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," August 1985.

(Methodology for LCO 3.2.1.)

7. WCAP-10054-P-A, Addendum 2, Revision 1, "Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection Into the Broken Loop and COSI Condensation Model," July 1997.

(Methodology for LCO 3.2.1)

8. WCAP-11145-P-A, "Westinghouse Small Break LOCA ECCS Evaluation Model Generic Study With the NOTRUMP Code,"

October 1986.

(Methodology for LCO 3.2.1)

9. WCAP-10079-P-A, "NOTRUMP - A Nodal Transient Small Break and General Network Code," August 1985.

(Methodology for LCO 3.2.1)

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.)

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

Reporting Requirements 5.6

11. WCAP-14710-P-A, "1-D Heat Conduction Model for Annular 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.

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 (PIT) 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 3.4.6, "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."

The analytical methods used to determine the RCS press~urej~"

't-ýature and LTOP limits shall be those previousl' S) I "and appod by the NRC in NRC letter, "R.E, na - Acceptance frReferencin-gbWesr Tezea ie"Dits Report, Revision 2 (TAC No. M96529)," "atelr 28, 1997. Specifically, the methodology is descr' 1in t-'e a, owing documents:

1. r from R.C. Mecredy, Rochester Ga Electric Corporation (RG&E), to Document Control Des , C, R.E. Ginna Nuclear Power Plant 5.6-4 Amendment 488

Reporting Requirements

5.6 tention

Guy S. Vissing, "Application for Facility Operati Lice , Revision to Reactor Coolant System (RCS Pressure d Temperature Limits Report (PTL Administrative ntrols Requirements," Aft ment VI, September 29 19 as supplemente y letter from R.C.

Mecredy, RG&E, to Cu Vissi , RC, "Corrections to Proposed Low Temperature rpressure Protection System TcnclSpecificto" ctober* 1997

2. WCAP-1404 -A, "Methodology used to velop Cold Overpre re Mitigating System Setpoints and Heatup and lonLmtCreSetos1ad2Jn

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

5.6.7 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.8, Steam Generator (SG) Program. The report shall include:

a. The scope of inspections performed on each SG,

b. Active degradation mechanisms found,

c. Nondestructive examination techniques utilized for each degradation mechanism,

d. Location, orientation (if linear), and measured sizes (if available) of service induced indications,

e. Number of tubes plugged during the inspection outage for each active degradation mechanism,

f. Total number and percentage of tubes plugged to date, and

g. The results of condition monitoring, including the results of tube pulls and in-situ testing.

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

Proposed Mark-up of T.S. 5.6.6.c Insert (1)

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.

Specifically, the methodology is described in the following documents:

1. WCAP-14040-A, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, Revision 4

2. As an alternative to the use of WCAP-14040"A Section 3.2 methodology, the existing Ginna specific LTOP Setpoint Methodology submitted to the NRC in the letter to Guy S. Vissing (NRC) from Robert C. Mecredy (RG&E), Application for Amendment to Facility Operating License, Revision to Reactor Coolant system (RCS)

Pressure and Temperature Limits Report (PTLR) Administrative Controls Requirements, Attachment VI, Section 3.2, dated September 29, 1997 and approved in letter to Robert C. Mecredy (RG&E) from S.

Singh Bajwa (NRC), R.E. Ginna - Acceptance for Referencing of Pressure Temperature Limits Report, Revision 2 (TAC No. M96529),

dated November 28, 1997, may be utilized.