ML20082G431

Safety Evaluation Supporting Amends 114 & 105 to Licenses NPF-2 & NPF-8,respectively

ML20082G431
Person / Time
Site:	Farley
Issue date:	04/07/1995
From:	Office of Nuclear Reactor Regulation
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Shared Package
ML20082G426	List: ML20082G423 ML20082G431
References
NUDOCS 9504130249
Download: ML20082G431 (7)

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' NUCLEAR REGULATORY, COMMISSION

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.114 T0 FACILITY OPERATING LICENSE NO. 'NPF-2 4

1 AND AMEN 0 MENT NO.105 TO FACILITY OPERATING LICENSE NO. NPF-8

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SOUTHERN NUCLEAR 0PERATING COMPANY. INC.

JOSEPH M. FARLEY NUCLEAR PLANT. UNITS 1 AND'2

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l DOCKET NOS. 50 -348 AND 50-364 I

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1.0 INTRODUCTION

By. letter dated June.10, 1994 (Reference 1),..the Southern Nuclear Operating Company (SNC or the licensee) submitted proposed changes to the Technical:

i Specification (TS) for the Joseph M. Farley Nuclear Plant (Farley) Units 1 and 2, to allow modifications to relocate the lower level steam generator;,

water level taps to be made during the upcoming refueling outages to both' units. These modifications affect the TS associated with the Reactor Trip,

. System (RTS) and Engineered Safety Feature Actuation System (ESFAS) setpoints.

In support of these amendments, the licensee submitted, as an enclosure to the June 10, 1994, letter, a proprietary report prepared by Westinghouse, WCAP-13992 (Reference 2). A non-proprietary version of this report,.

1, WCAP-13993, was also submitted.

4 The following TS changes are proposed for Farley, Units 1 and 2:

(1)

In Table 2.2-1, Reactor Trip System Instrumentation Trip Setpoints, Functional Unit 13 (Steam' Generator Water Level Low-Low), the trip

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setpoint was changed from 215% to 2 25% of the narrow range instrument.

span and the allowable value was changed-from 214.4% to 2 23.3% of. the narrow range instrument span of each steam generator.

(2).

In Table 3.3-4, Engineered Safety Feature Actuation System Instrumentation Trip Setpoints, for Functional Unit 5.a (Steam Generator Water Level High-High) the turbine trip and feed water.

isolation trip setpoint was changed from 5 75% to s 79.2%~of the narrow range instrument span and the allowable value was changed from s 76% to s 80.5% of the narrow range instrument span of each steam generator.

(3)

For Functional Unit 6.b,in Table 3.3-4 (Steam Generator Water Level Low-Low) the trip actuation setpoint was changed from 2 15% to 2 25% of the narrow range instrument span and the allowable value was changed from 2 14.4% to 2 23.3% of the narrow range instrument spari of each steam-

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2.0 BACKGROUND

Both Farley units currently has reactor trip and safeguards actuation on low-low steam generator water level and reactor trip on steam /feedwater mismatch and low steam generator water level for protection for loss of heat sink caused by postulated events such as loss of normal feedwater, feedline rupture, and loss of all ac power to station auxiliaries. The steam generator low-low water level trip setpoint change reflects the proposed Farley modification to the lower level taps, including Farley specific instrumentation, procedures, calibration practices and uncertainties, and accounts for the increased span due to lowerinq of the steam generator lower level taps.

In addition, the steam generator high-high level setpoint for j

turbine trip and feedwater isolation has been revised to be consistent with the increased narrow range span.

Reference 2 provides the basis of the revised setpoints.

The purpose of the revision to these setpoints is to allow increased operational flexibility and to reduce spurious reactor trips due to feedwater system transients.

3.0 EVALUATION Because the relocation of the level taps results in a reduced steam generator inventory, the applicable transients and accidents related to the Farley design basis required review and/or revision.

SNC stated that all non-LOCA analyses that credit low-low steam generator level as primary protection were re-analyzed.

In addition, the most limiting steamline breaks for environmental qualification which credit low-low steam generator level as primary protection were re-analyzed.

Setpoint uncertainty calculations were also performed for steam generator level low-low level reactor trip and ESFAS high-high level turbine trip and feedwater isolation. The instrumentation uncertainties associated with each of these protective system functions were calculated using the Westinghouse statistical setpoint methodology.

3.1 Transient and Accident Safety Evaluations 3.1.1 Non-LOCA Evaluation The steam generator level tap relocation and the low-low level setpoint reduction required that the following Final Safety Analysis Report (FSAR)

Chapter 15 accident analyses be re-analyzed: (1) the loss of normal feedwater (FSAR Section 15.2.8), (2) loss of non-emergency ac power to 11 ant auxiliaries (FSAR Section 15.2.9), and (3) the feedwater system pipe brea( (FSAR Section 15.4.2.2).

These were re-analyzed by SNC as discussed below.

3.1.1.1 Loss of Normal Feedwater and Loss of Non-Emergency AC Power to the Plant Auxiliaries The loss of normal feedwater and the loss of non-emergency ac power to the plant auxiliaries are American Nuclear Society (ANS) Condition II events that are analyzed to demonstrate adequate heat removal capability exists to remove core decay heat and stored energy following reactor trip.

The acceptance criteria for these events include demonstrating there is no overpressurization of the primary or secondary side and that pressurizer filling does not occur.

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t The physical relocation of the steam generator level tap and:the corresponding reduction in the low-low level setpoint reduces the amount of mass.available following reactor trip to remove the core decay heat and stored energy, resulting in a potentially more limiting transient.

The; licensee stated that the analysis method and analysis assumptions were the same as'used in the current Farley FSAR. SNC stated that the results of the transient analyses showed that the capacity of the auxiliary feedwater system is adequate to provide ~ sufficient heat removal. from the Reactor Coolant System j

(RCS) following a reactor trip. The criterion that the pressurizer does not fill was met, assuring that the integrity of the primary system is not adversely affected.

For the. case without offsite power available, the results verified that the natural circulation capacity ~of.the RCS provides. sufficient heat removal capability to prevent fuel or clad damage following reactor -

coolant pump coastdown.

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3.1.1.2 Feedwater System Pipe Break The feedwater system pipe break is an ANS Condition IV event which is analyzed to demonstrate that the peak primary and secondary side pressures do not exceed allowable limits and the core remains adequately covered with water.

The relocation of the steam generator low-low level setpoint reduces the amount of mass available following a reactor trip to remove the core decay heat and stored energy, resulting in a potentially more limiting transient.

The licensee stated that two cases were analyzed in the Farley FSAR which vary the auxiliary feedwater (AFW) delivered to the intact steam generators following actuation on a steam generator low-low level signal. The first (Case A) assumes a total AFW flow rate of 350 gpm from the two motor driven pumps delivered to two steam generators 10 minutes following an actuation signal on-low-low level. The second (Case B) assumes a total 150 gpm is fed to the intact steam generators on a low-low level signal following a 60-second delay. The flow is then increased to 350 gpm 30 minutes from the time of-the actuation signal. Additional key assumptions are: (1) the initial power is assumed to be at 102% of the NSSS design power rating (2,790 MWt); (2) a conservative core residual heat generation model based on the 1979 version of -

ANS-5.1 is used; (3) the steam generator low-low. water level setpoint is

. conservatively assumed to be at 0% of the new narrow range span; and (4) a 20%

steam generator tube plugging level is also assumed. The method of analysis and assumptions used are otherwise in accordance with those presented in the FSAR.

The licensee stated that the transient results show that the capacity of.the auxiliary feedwater system is adequate to provide sufficient heat removal from the RCS to prevent overpressurization of the RCS and the main steam system and to prevent core uncovery. The reactor coolant remains subcooled, assuring that the core remains adequately covered with water. The analysis results also verify that the natural circulation capacity of the RCS provides sufficient heat removal capability following reactor coolant pump coastdown.

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.. 3.1.1.3 Non-LOCA Transients Not Requiring Any Reanalysis

' The licensee stated that'the following transients were not reanalyzed since-either the transients are not affected by safety analysis assumptions or any-change to secondary side analysis assumptions will not adversely affect the results of the. analyses.

Uncontrolled RCCA Bank Withdrawal ~ from a Subcritical Condition (FSAR Section 15.2.1)

Uncontrolled RCCA Bank Withdrawal at Power (FSAR Section 15.2.2)

RCCA Misalignment'(FSAR Section-15.2.3)

Uncontrolled Boron Dilution (FSAR Section 15.2.4)

Partial Loss of Forced Reactor Coolant Flow (FSAR Section 15.2.5)

Startup of an Inactive Reactor Coolant Loop (FSAR Section'15.2.6)

Loss of External Electrical Load and/or Turbine-Trip (FSAR Section 15.2.7)

Excessive Heat Removal Due to Feedwater System Malfunctions (FSAR Section 15.2.10)

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Excessive Load Increase Incident (FSAR Section 15.2.'11)

Accidental Depressurization of the RCS (FSAR Section 15.2.12)

Accidental Depressurization of the Main Steam System-(FSAR Section 15.2,13)

Inadvertent Operation of ECCS During Power Operation (FSAR Section 15.2.14)

Complete loss of Forced Reactor Coolant Flow (FSAR Section 15.3.4)

Rupture of Main Steam Line (FSAR Section-15.4.2.1)

Single Reactor Coolant Pump Locked Rotor (FSAR Section 15.4.4)

Rupture of a Control Rod Drive Mechanism Housing (RCCA Ejection) (FSAR Section 15.4.6)

Main Steam Line Ruptures Inside Containment (FSAR Section 6.2.1.3.11)

In addition, SNC stated that all non-LOCA analysis, including main steam line break mass and energy releases for containment response, which were not specifically reanalyzed, were evaluated and found not to be impacted by the proposed modification.

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3.1.2 LOCA Related Evaluations The licensee ~ stated that all LOCA, LOCA forces, steam generator tube rupture,-

and LOCA-related analyses were reviewed and were found to be unaffected by this proposed modification.

3.1.3 Steam Generator Water Level Control The licensee stated that the steam generator water level control system uses inputs from narrow range level instruments. Therefore, the control system programmed setpoint will be revised to account for the increased fluid velocity effect and the increased span resulting from the' relocated lower i

level tap.

3.2 Setooint Methodoloav Evaluations The Farley Units 1 and 2 steam generator level instrumentation lower tap l

modification moved the tap approximately 68 inches (from 443 inches to 375 inches).. This relocation resulted in an increased level span from 144 inches.

to 212 inches. The functions affected by the change are; steam generator low-low level reactor trip and ESF actuation and the steam generator high-high -

level turbine trip and feedwater isolation. The instrument uncertainties associated with each of these protection system functions were calculated using the Westinghouse statistical setpoint methodology. The calculations accounted for all known instrument uncertainties associated with the. level transmitters, signal processing equipment, and calibration methods that are applicable to these functions.

In addition, process measurement accuracy allowances were included to account for process pressure changes and reference leg ambient temperature changes from the reference conditions,-as well as fluid velocity effects and downcomer subcooling effects associated with the-new lower tap location.

Environmental allowances were also included in the level setpoint calculations to account for the potential effects induced on the level transmitter, signal cable and reference leg by adverse containment environmental conditions. This calculation resulted in the proposed Nominal Trip Setpoints of 25% narrow range span (NRS) for. low-low level and 79.2% NRS for high-high level which provides positive margin to the Safety Analysis Limits, after accounting for all known uncertainties. The allowable values for the low-low and high-high steam generator level protection functions have-been calculated to be 23.3% NRS and 80.5% NRS, respectively.

Comparison of the existing setpoint (15% NRS for low-low and 75% NRS for high-high level trip) with the proposed values indicate that the proposed setpoints are conservative, will allow increased operational flexibility, and will reduce spurious reactor trips due to feedwater system transients.

The licensee further committed to revise the Control System programmed setpoint to account for the fluid. velocity effect and the increased narrow range span.

The staff compared SNC's calculations with those contained in WCAP-13751 (Reference 3) which were previously reviewed and approved by the staff.

. 3.3 Evaluation Summary The staff has reviewed the safety analyses and setpoint evaluations performed by SNC to support the proposed level tap modification and subsequent setpoint changes. Since the results of the reanalyzed' accidents are within' allowable limits and the proposed setpoint methodology calculations are consistent with those contained in WCAP-13751, which were previously. approved by the staff, the staff concludes that the proposed TS changes resulting from the modification to lower the steam generator level taps are acceptable.

4.0 STATE CONSULT &I1QM In accordance with the Commission's regulations, the State of Alabama official was notified of the proposed issuance of'the amendnents. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types,..

of any effluents that may be released offsite, and that there is no s gnificant increase in individual or cumulative occupational radiation i

exposure.

The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (60 FR 12253 dated March 6, 1995).

Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR 51.22(b) no environmental impact statement or environment:11 assessment need be prepared in connection with the issuance of the amendments.

6.0 CONCLUSION

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

Principal Contributors:

H. Balukjian l

I. Ahmed Date:' April 7,1995 J

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REFERENCES 1.

Letter from D. Morey, SNC, to USNRC, June 10, 1994.

2.

WCAP-13992, " Steam Generator Level Tap Relocation Assessment for J. M.

Farley Nuclear Plant Units 1 and 2," R. J.- Morrison and J. Srinivasan, March'1994.

3.

WCAP-13751 " Westinghouse Setpoint Methodology For Protectio'n Systems for-Farley Nuclear: Plant Units '1 and 2," S.V. Andre,; June 1993.

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X AMENDMENT NO. 114 LTO FACILITY OPERATING LICENSE NO. NPF FARLEY, UNIT 1-

. AMEN 0 MENT NO.-105

-T0 FACILITY OPERATING LICENSE NO. NPF FARLEY,-UNIT.2 DISTRIBUTION:

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4 PD Il-2' Reading File S. Varga.

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'J. Zwolinski OGC

.G. Hill (4)

C. Grimes - DOPS/0TSB J. Wermiel

'R. Jones H. - Balukjian I.-Ahmed

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OPA OC/LFDCB E..Merschoff, R-II-cc:

Farley Service List 9

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