ML20209D026

Proposed Tech Specs,Increasing Steam Generator Tube Plugging Limit to 10% & Changing Heat Flux Hot Channel Factor Limit

ML20209D026
Person / Time
Site:	Farley
Issue date:	08/25/1986
From:	ALABAMA POWER CO.
To:
Shared Package
ML20209D006	List: ML20209D001 ML20209D026 ML20209D039
References
TAC-62283, TAC-62284, NUDOCS 8609090167
Download: ML20209D026 (11)

Text

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ATTACHMENT 1 i

4 Proposed Changes to Technical Specification Pages Unit 1 Revision

- Page 2-2 Replace Page 3/4 2-4 Replace Page 83/4 2-1 Replace Unit 2 Revision Page 2-2 Replace Page 3/4 2-4 Replace Page B3/4 2-1 Replace i

l i

i k

4 1

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.i i

, 8609090167 860825 PDR ADOCK 05000348 P PDR

Es5.-

sse - -

Unacceptable s55- -

Operation 650 ,

400 psia s45 sde " 2250 psia 635

.'s30 2000 psia c's25 .

$ s20 1875 psia sis.

g. s t e l, ,

" ses. ,

seee i 5;5 .I Acceptable 59e " Operation Sa5 "

580 "

575 572 '

5s5

9. .I .2 .5 4 .5 .5 .7 .e '9 1. 1.1 1.2 FRACTION OF RATED THEPSE POWER Figure 2.1-1 Reactor Core Safety Limit Three Loops in Operation Applica ility: .110% Steam Generat:r Tube l

Plugging FARLEY UNIT 1 l Amendment No.

2-2 -

POWER DISTRIBUTION LIMITS 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - Fn(Z)

LIMITING CONDITION FOR OPERATION 2

3.2.2 FQ (Z) shall be limited by the following relationships:

F9 (Z) f [2.32] [K(Z)] for P > 0.5 P

Fg(Z) f [(4.64)] [K(Z)] for P f 0.5 where P = THERMAL POWER RATED THERMAL POWER and K(Z) is the function obtained from Figure (3.2-2) for a given core height location.

APPLICABILITY: MODE 1 ACTION:

With Fg(Z) exceeding its limit:

a. Reduce THERMAL POWER at least 1% for each 1% Q F (Z) exceeds the limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; subsequent POWER OPERATION may proceed provided the Overpower delta T Trip Setpoints have been reduced at least 1% for each 1% Fn(Z) exceeds the limit. The Overpower delta T Trip Setpoint reduction shall be performed with the reactor in at least HOT STANDBY.

b. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER above the reduced limit required by a, above; THERMAL POWER may then be increased provided Fg(Z) is demonstrated through incore mapping to be within its limit FARLEY-UNIT 1 3/4 2-4 AMENDMENT NO.

3/4.2 POWER DISTRIBUTION LIMITS BASES

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The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by: (a) maintaining the minimum DNBR in the core greater than or equal to 1.30 during normal operation and in short term transients, and (b) limiting the fission gas release, fuel pellet temperature and cladding mechanical properties to within assumed design criteria. In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200 F is not exceeded The definitions of certain hot channel and peaking factors as used in these i specifications are as follows:

Fg(Z) Heat Flux Hot Channel Factor, is defined as the maximum local heat flux on the surf ace of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods and measurement uncertainty.

Fgg Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod with the ' highest integrated power to the average rod oower.

Fxy(Z) Radial Peaking Factor, is defined as the ratio of peak power density to average power density in the horizontal plane at core elevation Z.

i 3/4.2.1 AXIAL FLUX DIFFERENCE The limits on AXIAL FLUX DIFFERENCE (AFD) assure that the Fg(Z) upper bound envelope of 2.32 times the normalized axial peaking factor is not exceeded during either normal operation or in the event of xenor redistributien following power changes.

Target flux difference is determined at equilibrium xenon conditions. The full length rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady state operation at high powar levels. The value of the target flux difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER for the associated core burnup conditions. Target flux differences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL POWER level. The periodic updating of the target flux difference value is necessary to reflect core burnup considerations.

FARLEY-UNIT 1 B 3/4 2-1 AMENDMENT NO.

465" saa ' Unacceptable 855' Operation

.ssa, 9400 psia E45 "

640 2250 psia 555- -

.'- 65a. 2000 psia s25 fs2a. -

1875 psia

@E15 -

' 610 m

g685 '

see. -

585 "

Acceptable 590 Operation i

585 -

l 588" 575" 570- -

565 1.1 1.2

6. .I .2 .5 4 .5 .5 .7 .8 ;9 1.

FRACTION OF RATED THERMAL POWER Figure 2.1-1 Reactor Core Safety Limit Three Loops in Operation Applicability: 1ist Steam Generator Tube l Plugging FARLEY UNIT 2 Amendrnent No.

2-2

POWER DISTRIBUTION LIMITS 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - En(Z)

LIMITING CONDITION FOR OPERATION g

3.2.2 Fg(Z) shall be limited by the following relationships:

Fg(Z) f [2.32] [K(Z)] for P > 0.5 P

F9 (Z) 1 [(4.64)] [K(Z)] for P 10.5 where P = THERMAL POWER RATED THERMAL POWER and K(Z) is the function obtained from Figure (3.2-2) for a given core height location.

APPLICABILITY: MODE 1 ACTION:

With Fg(Z) exceeding its limit:

a. Reduce THERMAL POWER at least 1% for each 1% Q F (Z) exceeds the limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; subsequent POWER OPERATION may proceed provided the Overpower delta T Trip Setpoints have been reduced at least 1% for each 1% Fg(Z) exceeds the limit. The Overpower delta T Trip Setpoint reduction shall be performed with the reactor in at least HOT STANDBY.

b. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER above the reduced limit required by a, above; THERMAL POWER may then be increased provided FQ(!) is demonstrated through incore mapping to be within its limit.

FARLEY-UNIT 2 3/4 2-4 AMENDMENT NO.

3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by: (a) maintaining the minimum DNBR in the core greater than or equal to 1.30 during normal operation and in short term transients, and (b) limiting the fission gas release, fuel pellet temperature and cladding mechanical properties to within assumed design criteria. In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200*F is not exceeded.

The definitions of certain hot channel and peaking factors as used in these specifications are as follows:

FQ(Z)

Heat Flux Hot Channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods and measurement uncertainty.

Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of Fh the integral of linear power along the rod with the highest integrated power to the average rod power.

Fxy(Z) Radial Peaking Factor, is defined as the ratio of peak power density to average power density in the horizontal plane at core elevation Z.

3/4.2.1 AXIAL FLUX DIFFERENCE The limits on AXIAL FLUX DIFFERENCE (AFD) assure that the Fn(Z) upper bound envelope of 2.32 times the normalized axial peaking factor is not exceeded during either normal operation or in the event of xenon redistribution following power changes.

Target flux difference is determined at equilibrium xenon conditions. The full length rods may be positioned within the core in accordance with their respective insertion limits and should be inserted rear their normal position for steady state operation at high power levels. The value of the target flux difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER for the i associated core burnup conditions. Target flux differences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL POWER level. The periodic updating of the target flux difference value is necessary to reflect core burnup 4

considerations.

i i

i FARLEY-UNIT 2 B 3/4 2-1 AMEN 0 MENT NO.

i

ATTACHMENT 2 Significant Hazards Evaluation Pursuant to 10CFR50.92 For the Proposed Steam Generator Tube Plugging Limit and FqTechnical Specification Changes Proposed Changes Revise Figure 2.1-1 to show a steam generator tube plugging limit of 10% and revise the F glimit s f Specification 3.2.2 and Bases 3/4.2.1 to be 2.32 for greater than b0% rated thermal power and 4.64 for less than or equal to 50%

rated thermal power.

Background

Farley Nuclear Plant currently has a steam generator tube plugging limit of 5%

as shown on Technical Specification Figure 2.1-1. This limit is based on the Large Break LOCA/ECCS analysis in FSAR Section 15.4 which assumes 5% steam generator tube plugging. Approximately 2.9% of the steam generator tubes have been plugged in Unit 1 and approximately 3.7% of the steam generator tubes have been plugged in Unit 2. This level of steam generator tube plugging includes all row 1 tubes in each steam generator. The 5% technical specification limit is anticipated to be adequate for any tube plugging which may be required as a result of tube inspections performed in the next refueling outage for each unit. However, Alabama Power Company does not want to risk a potential delay of plant startup due to this technical specification limit should more tubes require plugging than currently anticipated. Therefore, a technical specification change is proposed to increase the steam generator tube plugging limit to 10% in order to provide additional margin to the limit.

The current Large Break LOCA analysis in the Farley Nuclear Plant FSAR also assumes a Heat Flux Hot Channel Factor (F limit of 2.31 for greater than 50%

rated thermal power (RTP) and 4.62 for les than or equal to 50% RTP. These limits on Fn were required as a result of penalties assessed by the NRC against the 1978 ver'hion of the Westinghouse ECCS Evaluation Model.

To support the proposed Technical Specification change for 10% steam generator tube plugging, Westinghouse has performed the required Large Break LOCA analysis for Alabama Power Company. This new analysis used the Westinghouse 1981 ECCS Large Break Evaluation Model with BART and assumed an Fo limit of 2.32 (4.64 for

< 50% RTP) since the 1978 Evaluation Model was not being used. The assumed Fo Timit of 2.32 is consistent with the original large Break LOCA analysis, and the design / licensing basis for Farley Nuclear Plant. A description of this new LOCA analysis, including the methodology, assumptions, references and results, is provided in Attachment 3. This analysis has calculated a worst-case peak clad temperature (PCT) of 1973*F and demonstrates that the acceptance criteria of 10CFR50.46 are met. Additionally, the increase in Fg is conservatively bounded by the assumptions of the non-LOCA transient analyses and therefore has no impact on these analyses. The new Fq limits will require a change to Technical Specification 3.2.2.

Attachment 2 Page 2 Alabama Power Company recognizes that Westinghouse has notified the NRC by letter NS-NRC-86-3130 dated June 2,1986 of an assessment regarding the effects of control rod thimble filling during the reflood phase and the removal of a hot assembly power adjustment, originally included to account for control rod thimbles, from the BART code methodology. As discussed in the Westinghouse notification and further described in WCAP-9561-P-A, Addendum 3, the effect of thimbles on core hydraulics and the removal of the inappropriately-applied hot assembly power adjustment was found to be offset by conservatisms currently contained in BART. This generic assessment of the model changes described in WCAP-9561-P-A, Addendum 3 is applicable to the Farley Nuclear Plant analysis documented in Attachment 3. Additional studies have subsequently been performed by Westinghouse which have shown that a revised BART analysis which incorporates the required model changes and removes some of the identified conservatisms should result in a reduction in PCT. However, even if this revised analysis predicted an increase in PCT, the PCT would still remain well within the 10 CFR 50.46 acceptance criterion of 2200*F. Alabama Power Company further recognizes that the resolution of the required model changes is subject to NRC review and approval.

An additional analysis was performed to determine the effects on core flow due to steam generator tube plugging. This analysis determined that 10% steam generator tube plugging would not decrease RCS flow below the thermal design flow (TDF) for Farley Nuclear Plant. Since the non-LOCA transients are based on TDF, a 10%

steam generator tube plugging limit was determined to have no impact on the non-LOCA transients and therefore no impact on DNB.

The steam generator tube plugging limit increase was also evaluated for impact on structural integrity and safety of the reactor coolant system components. This evaluation considered the reduction in flow and the increase in pressure drop across the primary side of the steam generators, the increase in pressure drop across the steam generator tubes from the primary to secondary side of the tubes, and the impact of tube plugging on the various components of the RCS. Since the steam generators are designed to accommodate greater than 10% tube plugging without affecting steam generator performance and the assumptions used to develop the design transients included sufficient conservatism to account for a reduction in RCS flow down to the thermal design flow, there will be no change to the RCS design transients. On this basis, the increase in steam generator tube plugging will have no impact on the structural integrity of the RCS components, including the reactor vessel and internals, the reactor coolant pumps, the pressurizer, the control rod drive mechanisms, or the RCS piping, supports, and nozzle loads.

Based on previous analyses of steam generators similar to those at Farley, sufficient margin in stress levels and fatigue usage exists for the increased pressure drops across the primary side of the steam generators. Additionally, the increased pressure differential from the primary to the secondary side of the steam generator tubes is within the design envelope of the steam generator tubes. Therefore, the structural integrity of the RCS components is not affected by the increase in steam generator tube plugging limits to 10%.

Attachment 2 Page 3 Analysis Alabama Power Company has reviewed the requirements of 10CFR50.92 as they relate to the proposed changes to the steam generator tube plugging limit and FQ technical specifications and considers these changes not to involve a significant hazards consideration. In support of this conclusion, the following analysis is provided:

(1) The proposed changes will not increase the probability or consequer.ces of an accident previously evaluated because the revised ECCS analysis provided in Attachment 3, which was performed to support these changes, has demonstrated that the acceptance criteria for 10CFR50.46 have been met.

The proposed changes have also been demonstrated to have no impact on the non-LOCA DNB transients or RCS structural integrity. Therefore, the probability or consequences of an accident previously evaluated will not be '

increased.

(2) The proposed changes will not create the possibility of a new or different kind of accident from any accident previously evaluated because both changes consist of changes to assumptions in previously evaluated accidents. Additionally, the increase in steam generator tube plugging has been evaluated for impact on RCS average temperature, thermal design flow and secondary side pressure and determined to have no impact on current plant operating limits for these parameters. Furthermore, the increase in the steam generator tube plugging limit will have no effect on RCS structural integrity. Thus, these proposed changes will not create the possibility of a new or different kind of accident from any accident previously evaluated.

(3) The proposed changes will not involve a reduction in a margin of safety because RCS structural integrity is maintained and the revised ECCS analysis has demonstrated the requirements of 10CFR50.46 are met.

Additionally, the calculated peak clad temperature from this revised analysis is even less than the existing analysis and provides additional

, margin to the limit of 2200'F. Therefore, these proposed changes will not involve a reduction in a margin of safety.

[

lt is recognized that the BART analysis documented in Attachment 3 contains an inappropriate hot assembly power adjustment, originally included to account for control rod thimbles, and the error associated with the prediction of thimble filling during the reflood phase. However, sufficient conservatisms exist in the BART analysis which can offset the negative effect of the increased hot assembly power and thimble filling. Additional studies have been performed by Westinghouse which have shown that a revised BART analysis which incorporates the required model changes and removes some of the identified conservatisms should

s Attachment 2 Page 4 result in a reduction in PCT. However, even if this revised analysis predicted an increase in PCT, the PCT would still remain well within the 10 CFR 50.46 acceptance criterion of 2200 F. Therefore, the conclusions of this analysis for a no significant hazards determination remain valid on the basis that the results of a revised analysis which incorporates the required model changes and removes some of the identified conservatisms will still remain well within the bounds established in 10 CFR 50.46.

Conclusion Based upon the analysis provided herewith, Alabama Power Company has determined that the proposed changes to the technical specifications will not increase the probability or consequences of an accident previously evaluated, create the possibility of a new or different kind of accident from any accident previously evaluated, or involve a reduction in a margin of safety. Therefore, Alabama Power Company has determined that these proposed changes meet the requirements of 10CFR50.92(c) and do not involve a significant hazards consideration.