ML20214S210

Proposed Tech Specs,Increasing Steam Generator Tube Plugging Limit to 10% & Incresing Fq Coefficient to 2.32 for Greater than 50% & to 4.64 for Less than or Equal to 50% Rated Thermal Power

ML20214S210
Person / Time
Site:	Farley
Issue date:	06/02/1987
From:	ALABAMA POWER CO.
To:
Shared Package
ML20214S183	List: ML20214S179 ML20214S210 ML20214S233
References
TAC-62283, TAC-62284, NUDOCS 8706090182
Download: ML20214S210 (11)

Text

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ATTACHMENT 1 Proposed Changed Pages Unit 1 Revision Page 2-2 Replace Page 3/4 2-4 Replace Page B3/4 2-1 Replace ~

Unit 2 Revision Page 2-2 Replace Page 3/4 2-4 Replace Page B3/4 2-1 Replace t

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FRACTION OF RATED THERMAL POWER

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Figure 2.1-1 Reactor Core Safety Limit l

Three Loops in Operation Appl icability: < 10% Steam Generator Tube

( Plugging l

FARLEY UNIT 1 Amendment No.

2-2

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

LIMITING CONDITION FOR OPERATION

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3.2.2 F9 (Z) shall be limited by the following relationships:

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

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F9 (Z) $ [4.64] [K(Z)] for P $ 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 ()(Z) exceeding its limit:

a. Reduce THERMAL POWER at least 1% for each 1% Q (Z) exceeds the limit within 15 minutes and similarly reduce the Pow)re 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% FQ (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 F (Z) is demonstrated through incore mapping to be within its limit 9 FARLEY-UNIT 1 3/4 2-4 AMENDMENT NO.

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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 the fission gas release, fuel pellet temperature and cladding m properties to within assumed design criteria.

linear power density during Condition I events provides assurance that theIn ad initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200*F is not exceeded.

The definitions specifications are asoffollows:

certain hot channel and peaking factors as used in these Fg(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.

Ffg 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 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 thatgthe F (Z) upper bound envelope of 2.32 times the normalized axial peakir.g factor is not exceeded during either normal operation or in the event of xenon redistribution following power changes.

4 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 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 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.

flux difference value is necessary to reflect core burnup considerations.The peri FARLEY-UNIT 1 B3/4 2-1 AMENDMENT N0.

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9. .I .2 .5 4 .5 .6 .7 .8 ;9 1. 1.1 1.2 FRACTION OF RATED THERMAL POWER Figure 2.1-1 Reactor Core Saf'ety Limit Three Loops in Operation Applicability: < 10s Steam Generator Tube Plugging FARLEY UNIT 2 Amendment No.

2-2

1 POWER DISTRIBUTION LIMITS l

3/4.2.2

~ HEAT FLUX HOT CHANNEL FACTOR - Fn, (Z) i LIMITING CONDITION FOR OPERATION 3.2.2 =

Fq (Z) shall be limited by the following relationships:

Fq (Z) 5 [2.32] [K(Z)] for P > 0.5 P

Fq (Z) 1 [4.64] [K(Z)] for P 1 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 Qeach 1% 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% FQ (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 providedQ F (Z) is demonstrated through incore mapping to be within its limit FARLEY-UNIT 2 3/4 2-4 AMENDMENT NO.

d 3/4.2 POWER DISTRIBUTION LIMITS s

- 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 th'e 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:

F(Z)

Q 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 measurenent uncertainty.

FfH 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 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 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 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 near 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 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 2 B3/4 2-1 AMENDMENT NO.

ATTACHNENT 2 i Significant Hazards Evaluation Pursuant to 10 CFR 50.92 for the Proposed Steam Generator Tube Plugging Limit and FQ Technical Specification Changes Proposed Changes Revise Figure 2.1-1 to show a steam generator tube plugging limit of 10% and revise the FQ coefficient of Specification 3.2.2 and Bases 3/4.2.1 to be 2.32 for greater than 50% 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. Based on degradation identified during the last Unit 2 inspection (LER 86-004-00), expected tube pluggings during the Fall 1987 Unit 2 refueling outage could exceed the current margin to 5%, especially if F* and sleeving options are not available. Alabama Power Company does not desire to risk a potential delay of plant startup due to this 5% Technical Specification limit should defective tubes require plugging that exceeds the 5% limit. 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 assumes 5% steam generator tube plugging and a full power Heat Flux Hot Channel Factor (FQ ) of 2.32. The Technical Specification Fq coefficient of 2.31 for greater than 50% RTP and 4.62 for less than or equal to 50% RTP was required as a result of penalties assessed by the NRC against the 1978 version 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 (WCAP-9220-P-A and WCAP-9221) with BASH (WCAP-10266, Rev. 2) and assumed an Fq of 2.40 and an FdeltaH of 1.62 which bound the small break FQ of 2.32 (4.64 for 50% RTP) and FdeltaH of 1.55. 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 2013*F and demonstrates that the acceptance criteria of 10 CFR 50.46 are met. Additionally, the increase in Fq 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.

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ATTACHMENT 2 Page 2 Subsequent to the completion of the Farley Large Break LOCA analysis with BASH Westinghouse notified Alabama Power Company of enhancements to the BASH code and methodology that were made to improve the reliability and performance of the code in certain circumstances. The modifiations to the BASH methodology which incorporate these enhancements are described in Addendum 2 to .

WCAP-10266, Revision 2 and were submitted to the NRC via letter NS-NRC-87-3212 dated March 26, 1987. An evaluation of the impact of the BASH code modifications on the Farley Large Break LOCA analysis with BASH is provided in . This evaluation concludes that the Farley Large Break LOCA analysis with BASH remains conservative and bounding.

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. Pump coastdown characteristics are based on TDF which does not change for 10% steam generator tube plugging. Therefore the modeled pump coastdown in the current non-LOCA analyses will not become more severe. The pump coastdown for Farley is modeled using the PHOENIX computer code. A description of the model may be found in WCAP-7973,

" Calculation of Flow Coastdown After Loss of Reactor Coolant Pump (PHOENIX Code)."

Since the non-LOCA DNB transients are based on TDF, a 10% steam generator tube plugging limit was determined to have no impact on the non-LOCA DNB transients. The effect of 10% steam generator tube plugging upon those non-LOCA accidents which are not DNB related or for which DNB is not the only safety criteria were also evaluated. The only accident of this group which is affected by 10% steam generator tube plugging is the boron dilution analysis.

An input to the boron dilution analysis for Modes 1 and 2 is the RCS active volume, i.e. , the total RCS volume minus the volumes of the pressurizer, the pressurizer surge line, the dead volume of the reactor vessel head, and plugged steam generator tubes. Reduction of the RCS active volume is directly proportional to the reduction in operator response time for the boron dilution event described in the Farley FSAR. It is estimated that 10% tube plugging will reduce the Farley active volume by approximately 4%. However, from the boron dilution analysis done for Farley, it can be shown that the RCS active volume can be reduced by more than 4% and operator action time (of at least 15 minutes) will still be met. Therefore the 10% tube plugging level for Farley will not change the conclusions of the safety analysis.

As stated above, the licensing basis small break LOCA analysis is based on a limiting power shape contained within an envelope of peaking factors with a maximum allowable total peaking factor (FQ ) of 2.32 and an FdeltaH of 1.55.

At low steam generator tube plugging levels (up to 20%), small break LOCA transients would not be affected by the tube plugging. The proposed changes in Technical Specifications will not impact or invalidate the current licensing basis small break LOCA analysis as represented in the Farley FSAR.

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- ATTACIDENT 2 Page 3 i

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% .

l 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 l pressure differential from the primary to 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%.

3

Analysis Alabama Power Company has reviewed the requirements of 10 CFR 50.92 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 consequences 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 10 CFR 50.46 have been met. The proposed changes have also been demonstrated to have no impact 4

on the conclusions of the small break LOCA analysis and all the non-LOCA 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 iapact on RCS average temperature, thermal design flow and secondary side pressure and determined to have no impact on p

4

ATTACEMENT 2 Page 4 i 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 10 CFR 50.46 are met.

I Additionally, the calculated peak clad temperature from this revised analysis is even less than the present Farley 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, a Conclusion Based upon the analysis provided herein, 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 saft.ty.

Therefore, Alabama Power Company has determined that these proposed changes

seet the requirements of 10 CFR 50.92(c) and do not involve a significant hazards consideration.

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