ML20197B617

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Forwards Response to 971114 RAI for Clarification of Methodology Used Re 970723 TS Amend Request.Revised Pressure Temperature Limits Repts,Revised TS Pages,Revised Safety Analysis & Revised Significant Hazards Evaluation Included
ML20197B617
Person / Time
Site: Farley  Southern Nuclear icon.png
Issue date: 12/18/1997
From: Dennis Morey
SOUTHERN NUCLEAR OPERATING CO.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20197B622 List:
References
NUDOCS 9712240012
Download: ML20197B617 (10)
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Devs Morey Southern Nucleat We Prosdent Operating Company

- Iadey Project - P0.Ba1295 Dirminghant Alatiama 35201 Tel 205.992.5131 December 18, 1997 SOUTHERN COMPANY Energy ro Serve krWorlP Docket No.: 50-348 10CFR50.90 50-364 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Wastungton, DC 20555 Joseph M. Farley Nuclear Plant Technical Specification Change Request Pressure Temperature Limits Reoort Ladies and Gentlemen:

By letter dated July 23,1997, Southern Nuclear Operating Company (SNC) submitted a Technical Specification (TS) Change Request associated with the relocation of the reactor coolant system (RCS) pressure-temperature (P-T) limits from the TS to the Pressure Temperature Limits Report (PTLR) in accordance with the guidance provided in Generic Letter (GL) 96-03, Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection' System Limits. By letter dated November 14,1997, the NRC issued a Request for Information (RAI) requesting clarification of the methodc!cgy used associated with the SNC TS amendment request. Furthermore, discussions with the NRC Staff resulted in changes to the methodology documented in WCAP-14040-NP-A, Revision 2, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Ileatup and Cooldown Curves, used to develop the Farley PTLR. In addition, the best estimate copper and nickel values determined by the Combustion Engineering Vest.el Owners Group in response to GL 92-01, Revision 1, Supplement 1, have been incorporated in the Farley Nuclear plant reactor vessel integrity analysis.

. Enclosure l' of this letter provides the response to the RAT and the methodology used by SNC

. to generate the P-T limits and setpoints associated with low temperature overpressure protection. Enclosures 2 and 3 provide the revised PTLR for Farley Units I and 2 respectively, Enclosure 4 provides the revised technical specification pages associated with this change. Although there were only minor changes to the technical specification pages submitted with the July 23,1997 submittal, all technical specification pages associated with this technical specification amendment are included for completeness. Enclosure 5 r ovides a revised safety analysis. Enclosure 6 provides a revised significant hazards evaluation. [g h LI 971224001$ 1218 yDR ADOCK 05000348 'I<hllhl]*

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If you have any rf.,estions, plasse advise.

Respectfbily submitted, .

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  • ^'ITHERN NUCLEAR OPERATING COMPANY l

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Dave Morey horn to and.mbxribedbefore me this l$5 day of h e1997

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UNotary Public My Commission Expires: /A,//7./0000 REM /maf.PTLR2NRC. DOC ,

Enclosures:

1. Reuponse to November 14,1997 NRC Staff Request for Additional [

Information j 2i Revised Unit 1 Pressure Temperature Limits Report ,

3. Revised Unit 2 Pressure Temperature Limits Report  !
4. Technical Specification Pages  !
5. Safety Analysis
6. Significant Hazards Evaluation l

cc: Mr. L. A. Reyes, Region II Administrator i Mr. J. I. Zimmerman, NRR Project Manager- i Mr. T. M. Roar. Plant Sr. Resident Inspector Dr. D, E. Williamson, State Department of Public Health  !

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ENCLOSUREl F

ENeLostmE I NRC STAFF REOUEST FOR ADDITIONAL INFORhf ATION PESSURE TEMPERATURE LIMITS REPORI FARLELLICENSE AMENDMENT REOUEST Dantium The proposed license amendments request approval to use a Pressure / Temperature Limits Report (PTLR) in accordance with Generic Letter 96-03, " Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits." Approval of the license amendments will allow the pressure / temperature (PT) limits to be changed without NRC approval. The proposed Technical Specificatioi;(TS) Section 6.9.15 references both the approved topical report, WCAP 14040 NP A, Revision 2, " Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Ileatup and Cooldown Limits Curves," and the approved NRC safety evaluation report, liowever, the approved topical (pg. 3 6) states that the methodology is " applicable only when the pressurizer power-operated relief valves are used for the COMS [ cold overpressure mitigating system]." Beca use the residual heat removal (RHR) relief valves are used for cold overpressure protection at Failey, the topical report is not applicable for developing setpoints for the R11R relief valves. The submittal does not state how the low-temperature overpressure protection (LTOP) setpoints are verified as acceptable.

Although the LTOP or RIIR setpoints are not being changed with this submittal or being moved to the proposed PTLR, each time the PT limits are revised the LTOP system must be reevaluated to ensure the current setpoints are acceptable for meeting the functio;.al requirements of the

  • system and capable of protecting the new PT limits. Because the approved methodology for determining the setpoints is not applicable to Farley, how will you demonstrate that the RdR setpoints are still acceptable relative to the new PT limits?

Because the approved methodology, which is referenced in the TS, is not applicable to RIIR setpoints, please update the submittal and the proposed TS and describe in detail how the approved methodology will be applied If a different methodology is presented to determine the RIIR setpoints, the methodology should be described in detail, explaining how calculations will be performed and how to account for uncertainties and setpoint drift A discussion of the limiting ovemressure transients should be included and how controls are in place and w,ll be maintained to assure the RIIR relief valves will protect the plant for all low temperature overpressure transients.

If a new methodology is used, it should also be referenced in the TS. Please inc'ude sample calculations for the firt,t application (for the proposed PT limits) of the methodo'ogy. The j required elements of a methodology are contained in the generic letter, Note: The above request was subsequently revised to include a request to addccss the seven

" Requirements for Methodology and PTLR" included in Generic Letter 96-03.

Southern Nuclear Response; The response to the NRC requests are included in the following

" Methodology for Determination of Reactor Coolant System Pressure Teruperature Limits and Low Temperatur %rpressure Protection System."

JOSEP11M. FARuiY NUCl.FAR PIANT MimlODO! 00Y FOR Df2iEtMINA110N OF REACTOR COO! ANT SYSTEM PRESSURE TEFERATURE LIMrrS AND Low TEMPERAIURE OVERPRESSURE PROTEC110N SYSTUd The methodology for determirdng the reactor cookt system pressure temperature limits includes the determination orlow temperature overpressure protection setpoints and is best described by addressing the seven " Requirements for Methodology and PTLR" found in Generic Letter 96 03. ,

1. Describe the transport calculation methods including computer codes and formulas used to calculate neutron Ruence. Provide references.

Section 2.2 of WCAP 14040 NP-A, Revision 2, provides the methodology for determining the neutron fluence for the surveillance capsules and the reactor vessel with the exception that, as requested by the NRC, calculated fluence values (&c.a) are used in lieu of best-estimate fluence (ta r,) described in WCAP 14040 NP A, Revision 2.

2. IlrieDy describe the surveillance program. Li.ensee transmittalletter should identify by title and number report conts ag the Reactor Yessel Surveillance Program and surveillance capsule reports. Topical / gen fic report contains place holder only.

Reference Appendix H to 10 CFR 50.

The reactor vessel material surveillance program for Farley Nuclear Plant Unit 1 is described in WCAP 8810. Alabama Power Company Joseph M. Farley Nuclear Plant Unit No. I Reactor Vessel Radiation Surveillance Program, dated December 1976. To date, four surveillance capsules have been removed from Farley Nuclear Plant Unit I as documented in the following test reports submitted to the NRC in accordance with 10 CFR 50, Appendix II:

  • WCAP 14196, Analysis of Capsule W from the Alabama Power Company Farley Unit i Reactor Vessel Radiation Surveillance Program, dated February 1995.
  • WCAP-11563, Revision 1, Analysis of Capsule X from the Alabama Power Company Joseph M. Farley Unit 1 Reactor Vessel Radiation Surveillance Program, dated September 1987.

Farley Unit 1 Reactor Vessel Radiation Surveillance Program, dated February 1984.

  • WCAP 9717, Analysis of Capsule Y from the Alabama Power Company Farley Unit No.1 Reactor Vessel Radiation Surveillance Program, dated June 1980 The reactor vessel material surveillance program for Farley Nuclear Plant Unit 2 is described in WCAP-8956, Alabama Power Company Joseph M. Farley Nuclear Plant Unit No. 2

Enclosure i Page 2 Methodology ,

Reactor Vessel Radiation Surveillance Program, dated Nagust 1977. To date, three surveillance capsules have been removed from Farley Nuclear Plant Unit 2 as documented in

'he following test reports submitted to the NRC in accordance with 10 CFR 50, Appendix II:

  • WCAP 12471, Analysis of Capsule X from the Alabama Power Company Joseph M.

Farley Unit 2 Reactor Vessel Radiation Surveillance Program, dated December 1989.

  • WCAP 11438, Analysis of Capsule W from the Alabama Power Company Joseph M.

Farley Unit 2 Reactor Vessel Radiation Surveillance Program, dated April 1987.

  • WCAP 10425, Analysis of Capsule U from the Alabami Power Company Joseph M.

Farley Unit 2 Reactor Vessel Radiation Surveillance Proytm, dated October 1983.

To assure continued compliance with the requirements of 10 CFR 50, Appendix II, Surveillance Requirement 4.4.10.1.2 for Farley Nuclear Plant Units 1 and 2 associated with the P.T limits requires that the reactor vessel material irradiation surveillance specimens be removed and examined in accordance with 10 CFR 50, Appendix 11.

3. Describe how the LTOP system limits are calculated applying system / thermal hydraulles and fracture mechanics. Reference SRP Section 5.2.2; ASME Code Case N.

514; ASME Code, Appendia G;Section XI as appiled in accordance with 10 CFR 50.55.

Farley Nuclear Plant utilizes the residual heat removal system relief valves (RIIRRVs) for low temperature overpressure protection (LTOP) of the RCS from brittle fracture by assuring that the limits of Appendix G are not exceeded. The RIIRRVs are spring loaded, bellows-type valves which have a setpoint of 450 psig and are designed to provide rated flow at 495 psig (i.e.,10% accumulation). In order to assure that the RIIRRVs are available to protect the RCS from an LTOP event, Technical Specification (TS) 3.4.10.3 requires that the RIIR suction valves be open and the RIIRRVs operable with a lift setting less than or equal to 450 psig or that the RCS be depressurized with a vent of greater than or equal to 2.85 square inches at RCS temperatures less than or equal to 310'F.

The design basis transients for the Farley Nuclear Plant LTOP system consist of a heat input transient and a mass input transient with the RCS in a water-solid condition. The worst-case heat input transient assumes the start of a single reactor coolant pump with a temperature differential of 50'F existing between the RCS and any one steam generator. At RCS temperatures less than or equal to 180*F, the worst-case mass input transient is assumed to be ,

the inadvertent start of one high head safety injection (HilSI) pump with a maximum flow rate of 590 gallons per minute based on the maximum number ofoperable IlliSI pumps allowed by TS 3.1.2.3. For RCS temperatures greater than 180*F, the worst-case mass input transient assumes the inadvertent opert, tion of three liHSI pumps with a maximum total flow rate of 1000 gallons per minute at zero backpressure. These three transients discussed above are utilized to determine the RCS pressure for further analysis.

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Methodology The Farley Nuclear Plant LTOP analysis consists of a detennination ofRCS pressures resulting from each of the design basis LTOP transients based on the relief capacity of the RHRRVs and the following conservative assumptions: l

. Credit is taken for flow through only one RHRRV due to single failure of the 8 other RHRRV; i

  • No flow through the RHRRVs is credited in the analysis until RCS pressure  !

achieves the 10% accumulation pressure for the RHRRVs of 495 psig; e Flashing is assumed to occur at the valve discharge;  !

. No credit is taken for a bubble in the pressurizer; and  ;

  • The analysis is performed at isothermal conditions in the RCS and provides -

protection against the steady-state Appendix G limit.

At RCS temperatures less than or equal to 180'F, the most-limiting design basis transient results in an RCS pressure of 495 psig. The resulting pressure is compared to the proposed Appendix G steady state limit curve to assure that the resulting RCS pressure of 495 psig does not exceed the allowable RCS pressure. The following table provides an Example of i Comparison of Limiting Design Basis Transient (LDDT) to Appendix G Steady State Limit Curve.

Example of Comparison of Limiting Design Basis Transient to Appendix G Steady State Limit Curve for Farley Unit 2 RCS Temperature RCS Pressure Appendix G Steady

('F) (LDBT)(psig) State Limit Curve (psig) 70 495 498 180 495 626 181 562.5 629 260 562.5 1070 261 795 1080 310 795 1749 As stated above, the RCS pressure for each of the above temperatures are compared to the proposed steady state Appendix G curve to assure that the RCS pressure does not exceed the Appendix G allowable pressure for the corresponding temperature. If this criteria is met, the Farley Nuclear Plant LTOP system provides adequate prctection for the proposed Appendix G curver. As can be seen from the above comparison, the Farley Nuclear Plant LTOP system provides adequate protection for the Appendix G curves, if the projected RCS pressure exceeds the Appendix G allowable pressure for the corresponding temperature, changes to the RHRRV characteristics, e.g., capacity, relief setpoint, accumulation, may be required. The te modifications may require a change to TS 3.4,10.3;

Enclosure 1 Page 4 Methodology The Parley Nuclear Plant LTOP enable temperature is the temperature below which the LTOP system is required to be operable in accordance with Section 3.4 of WCAP 14040 NP A, Revision 2. The LTOP enable temperature is compared to the RCS cold leg temperature of 310'F stated in the applicability statement of TS 3.4.10.3 to assure the RCS overpressure protection systems are available at temperatures below the LTOP enable temperature, if 310'F is not an acceptable LTOP enable temperature, a change to Technical Specification 3.4.10.3 will be required.

In order to minimize setpoint uncertainties and drifl, Farley Nuclear Plant tests the RIIRRVs on an accelerated basis from that required by the ASME Code. Bench tests are performed at 18 month intervals on a rotating basis for at least one of the RIIRRVs to verify the setpoint in accordance with TS Surveillance Reo Jtement 4.4.10.3,1(c). This frequency is more stringent than that required by the ASME Code for class 2 relief valves.

Additionally, Farley Nuclear Plant surveillance test procedures currently use an RiiR relief valve setpoint of 44515 psig for the setpoint. This approximately 1% tolerance is more stringent than the ASME Code requirement of 3% tolerance. The use of 1% se: point tolerance for the R11RRV setpoint coupled with the 10% accumulation provide adequate protection against setpoint drift. The increased surveillmce test frequency, the reduced R11RRV setpoint and setpoint tolerance, coupled with the analysis assumption that flow does not start until inlet pressure i : aches 450 psig + 10% accumulation, i.e.,495 psig, provide assurance that the RIIR relief valves will provide adequate protection against the limits of Appendix 0.

ASME Code Case N 514 is not used for Farley calculations.

4. Describe the method for calculating the ART using Regulatory Guide 1.99, Revision 2.

Section 2.4 of WCAP-14040 NP A, Revision 2, provides the methodology for calculating the adjusted reference temperature in accordance with Regulatory Guide 1.99, Revision 2.

5. Describe the application of fracture mechanics in constructing P-T curves based on ASME Code, Appendix G.Section XI, and SRP Section 5.3.2.

. Sections 2.5 and 2.6 of WCAP-14040 NP-A, Revision 2, provides the application of fracture mechanies in constmeting P-T curves. Tne resulting P-T limit curves are adjusted to account for the 60 psi AP between the reactor vessel beltline and the RIIRRVs associated with the operation of three reactor coolant pumps (RCPs) at RCS temperatures greater than or equal to 110*F. At RCS temperatures less than 110'F, the number of operating RCPs is limited to one and the resulting AP correction of 25 psig is applied.

Enclosure ! Page5 Methodology

6. Describe how the minimum temperature requirements in Appendix G to 10 CFR $0 are applied to P-T curves.

Section 2.7 of WCAP-14040-NP.A, Revision 2, provides the methodology for determination of the minimum temperature requirements in 10 CFR 50, Appendix 0. The minimum temperature requirement is adjusted as necessary to assure the RCS pressure resulting from design basis LTOP transients does not exceed the steady state Appendix G limit.

7. Describe how the data from multiple surveillance capsules are used in the ART calculation.

Section 2.4 of WCAP 14040 NP-A, Revision 2, provides the methodology for calculating the adjusted reference .cmperature with multiple surveillance caps iles.

Describe procedure if measured value exceeds predicted value.

As ststed in Section 2.4 of WCAP.14040.NP.A, Revision 2, if the measured value exceeds the predicted value, a supplement to the PTLR must be provided to demonstrate how the results affect the approved methodology.

WiiEN OTHER PLANT DATA ARE USED

1. Identify the source (s) of data when other plant data are used.

Farley Nuclear Plant does not rely on surveillance data from other licensees for its reactor vessel integrity analysi). Therefore, this item is not applicable to Farley Nuclear Plant.

2a. ldentify by title and number the safety evaluation report that approved the use of data for the plant. Justify applicability.

Farley Nuclear Plant does not rely on survell'ance data from other licensees for its reactor vessel integrity analysis. Therefore, this item is not applicable to Farley Nuclear Plant.

OR 2b. Compare licensee data with other plant data for both the radiation environments (e.g., neutron spectrum, irradiation temperature) and the sun'elliance test results.

Enclosure 1 Page 6 -  !

Methodology [

i Farley Nuclear Plant does not rely on surveillance data Rom other licensee, for it: -  !

reactor vessel integrity analysis. Therefore, this item is not applicable to Farley  :

Nuclear Plant. E i

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