ML17353A907

Safety Evaluation Supporting Amends 191 & 185 to Licenses DPR-31 & DPR-41,respectively

ML17353A907
Person / Time
Site:	Turkey Point
Issue date:	09/26/1996
From:	NRC (Affiliation Not Assigned)
To:
Shared Package
ML17353A906	List: ML17353A905 ML17353A907
References
NUDOCS 9610030108
Download: ML17353A907 (54)

Text

~it A500

~o a

daO I

C V/

n

~O S

UNITED STATES NUCLEAR REGULATORV COMMISSION WASHINGTONa D.C. 2055&0001 A

T 0 Y T OFF C

NUC R

R ACTOR REGULATION YOP T

CNSE 0.

D

-31 AN AM NOME T NO.

185 TO FAC I Y OPERAT G

CENS NO.

DPR-41 F ORID OW R

ND GHT COMPANY U

OIN OS.

3 OC T

OS. 50- 50 AND 50-25 1.0

~INT QO I

N By letter dated December 1&, 1995, as supplemented on May 3, June ll, July 1',

July 3, and August 22, 1996 (hereafter, collectively referred to as power uprate submittal) Florida Power and Light Company (FPL or the licensee) requested changes to the Facility Operating License (FOL) and Technical Specifications (TS) to increase rated thermal power from 2200 Megawatt thermal (MWt) to 2300 Mwt (approximately, 4.5 percent) for Turkey Point units 3 and 4.

The results of the uprate evaluations and analyses were documented in Westinghouse WCAP-14276, Revision 1, "florida Power

& Light Company Turkey Point Units 3 and 4 Uprating Licensing Report,"

(WCAP-142T6) dated December 1995 and submitted by the licensee with the December 18, 1995 request.

The original ~Feder 1 geee ster notice included information from the licensee's December 18,

1995, May 3 and June 11, 1996 letters.

The July 1, July 3, and August 22, 1996 letters provided clarification and amplification of the analysis in the previously noticed letters and were not outside the scope of the original federal

~ge ister notice.

2.0 BACKGROUND

Detailed evaluation of the Nuclear Steam Supply System (NSSS)

(including Loss of Coolant Accident (LOCA), non-LOCA, Containment Responses and Dose Consequences),

engineered safety features, power conversion, emergency

power, support systems and environmental issues were performed by the licensee and Westinghouse.

The licensee stated that the results of these evaluations and analyses confirmed that Turkey Point Units 3 and 4 can safely operate't the increased power level.

The capability of Turkey Point Units 3 and 4 to operate at uprated conditions was verified by the licensee in accordance with guidelines contained in Westinghouse topical report WCAP-10263, "A Review Plan for Uprating the Licensed Power of a Pressurized Water Reactor Power Plant".

This WCAP methodology, although not formally approved by the

NRC, was followed by North Anna,

Salem, Indian Point Unit 2, Callaway and Vogtle for their core power upratings.

9620030208 960926 PDR ADOCK 05000250 P

PDR

0

The licensee stated that Turkey Point Units 3 and 4 have as-designed equipment and system capability to accommodate steam flow rates of at least 5 percent above the original rating and the increase to higher power is obtained by effective utilization of existing systems and equipment.

3.0 ACC 0 NT ANALYSES A UATION The accident analyses were reanalyzed or evaluated to support operation at the uprated NSSS power level as discussed in the following sections.

3. 1.

valu t o o

No -

OC vents and Standb Safet Features Anal s's The licensee has reviewed all of the Updated Final Safety Analysis Report (UFSAR) Chapter 14 non-LOCA analyses for Turkey Point Units 3 and 4 to determine their continued acceptability based on plant operation at the uprated power level.

The following non-LOCA events were either evaluated or reanalyzed for plant conditions at uprated power level.

The licensee's reanalyses were performed using NRC-approved methods and computer codes.

The analyses incorporated a Revised Thermal Design Procedure (RTDP), which is a

part of the current.licensing basis for Turkey Point Units 3 and 4.

3. 1. 1 Uncontrolled Rod Cluster Control Assembl RCCA Bank Withdrawal from a Subcritical Condition The uncontrolled RCCA bank withdrawal from a subcritical condition is analyzed to ensur e that the core and the reactor coolant system (RCS) are not adversely affected.

This has been demonstrated since the results of the analysis show that the minimum departure from nucleate boiling ratio (DNBR) remains greater than the safety analysis limit and that the maximum fuel temperatures predicted to occur are much less than those required for clad damage (2700'F) or fuel melting (4800'F) to occur.

The staff considers that the effect of the power uprate on this event is, therefore, acceptable.

3. 1.2 Uncontrolled RCCA Bank Withdrawal at Power The uncontrolled RCCA bank withdrawal at power is analyzed to ensure that the core and the RCS are not adversely affected.

This has been demonstrated since the results of the analysis at the uprated conditions show that the high neutron flux and overtemperature hT reactor trip functions provide adequate protection to ensure that the minimum DNBR remains greater than the safety analysis limit and that the RCS and main steam systems are maintained below 110 percent of the design pressures.

The staff considers that the effect of the power uprate on this event is, therefore, acceptable.

3.1.3

~RCCA Dro Dropping of a full length RCCA into the core is analyzed to ensure that any resulting adverse power distribution does not violate the DNB design basis.

The analysis shows that following a dropped RCCA event, without automatic rod withdrawal, the plant will return to a stabilized condition at less than or equal to the initial power.

The staff considers

that, since the DNBR remains

II'

above the limit value, the event does not adversely affect the core and the resu1ts due to the power uprate are acceptable.

3. 1.4 C

i d Volume ontrol S stem CVCS Malfunction Unborated water can be inadvertently added to 'the RCS via the CVCS and cause a

reactivity increase.

The event is analyzed to ensure that there is sufficient time for mitigation of an inadvertent boron dilution event prior to the complete loss of shutdown margin (criticality).

The results show that the maximum reactivity addition due to the dilution is slow enough to allow the operator sufficient time to determine the cause of the addition and take corrective action before shutdown margin is completely lost.

For Mode I, at least 15 minutes are available for operator action from the time of alarm to preclude a complete loss of shutdown margin.

For Modes 2 and 6, at least 15 minutes and 30 minutes, respectively, are available for operator action from the time of initiation of the dilution.

This meets the Turkey Point licensing basis for the inadvertent dilution event and is, therefore, acceptable to the staff.

3. 1.5 Star tu of an Inactive Reactor Coolant Loo This event is precluded by the current Turkey Point TS, which do not allow operation with an inactive loop.

3. 1.6 Excessive Heat Re oval Due to Feedwater S stem Malfunctions An example of this type of event is one of the feedwater control valves is inadvertently fully opened while the reactor is operated at full power.

The reactor protection

systems, including power range high neutron flux, overpower hT, and turbine trip on high-high steam generator water level, are available for mitigating this event.

The reanalysis results indicate a transient minimum DNBR of 2.0, which is above the minimum DNBR limit, and a transient peak RCS pressure of 2300 psia, which is less than the maximum allowable limit.

Therefore, the results of this transient analysis are acceptable.

3. 1.7 Excessive Load Increase Incident This event assumes a rapid increase of steam demand that causes a power mismatch between the reactor power and steam load.

If the load increase exceeds the capability of the

RCS, the transient would be terminated by the reactor protection system to keep the transient DNBR above the minimum DNBR.

The reactor protection systems reactor trip setpoints, including overtemperature bT, overpower hT, power range high neutron flux, and low pressurizer

pressure, are available for mitigating this event.

The results of the reanalysis show a transient minimum DNBR of 2. I, which is above the minimum DNBR limit, and a transient peak RCS pressure of 2260 psia which is less than the maximum.allowable limit.

Therefore, the results of the transient are acceptable.

3.1.8 t Flow The licensee has performed reanalysis of both a partial and complete loss of forced reactor coolant flow and compared the results to the American Nuclear Society (ANS) condition II criteria.

These incidents may result from a, mechanical or electrical failure in one or more of the reactor coolant pumps (RCPs).

The transient would be terminated by the reactor protection systems to keep the transient ONBR above the minimum ONBR.

The reactor protection systems reactor trip setpoints, including undervoltage or underfrequency on RCP power supplies, underfrequency RCP breaker trips, low reactor coolant loop flow, and pump circuit breaker opening, are available for mitigating this event.

The results of the reanalysis for the limiting case (complete loss of flow) show a transient minimum DNBR of 1.55, which is above the minimum DNBR limit and a transient peak RCS pressure of 2370 psia, which is less than the maximum allowable limit.

The RCP locked rotor/shaft break events were reanalyzed as ANS condition IV events.

In this analysis, the off-site power is assumed available, which is consistent with the original licensing basis of the plant.

While the consequences of a locked rotor are very similar to those of a pump shaft

break, the analysis considers a scenario which represents the most limiting condition for the locked rotor. and pump shaft break event.

Following this

event, a reactor trip will be actuated on a low RCS flow signa'1.

For this

event, DNB is assumed to occur in the core.

The number of rods in DNB are conservatively calculated for use in dose consequences evaluations.

The results of the reanalysis show that the number of rods in ONB is less than 10 percent and the radiological consequences are within a small fraction of the 10 CFR 100 guideline values.

The peak transient RCS pressure is 2700

psia, which is less than the maximum allowable limit.

The results of the analysis meet the acceptance criteria for the condition IV event and are, therefore, acceptable to the staff.

3. 1.9 Loss of xternal Electrical Load and or Turbine Tri The licensee has performed a reanalysis of this event for the cases with and without pressure control and with maximum and minimum reactivity feedback.

For this event, the reactor protection systems reactor trip setpoints, including overtemperature hT, high pressurizer

pressure, and low-low steam generator water level, are available for mitigating this event.

The results of the reanalysis for the most limiting case (without pressure.control) show that the peak transient RCS pressure is 2700 psia, which is less than the maximum allowable limit.

Since this is a heatup event, transient DNBR generally remains above the initial point for all cases analyzed.

Therefore, the results of this transient are acceptable to the staff.

3. 1. 10 oss of Normal Feedwater and Loss of Non-emer enc Power to 'the Plant Auxiliaries In these

events, plant protection is provided by either the reactor trip setpoints for the low-low steam generator water level or the steam flow and feed flow mismatch coincident with low steam generator water level in any loop.

The results of the reanalysis show that the consequences of these

events are bounded by the loss of external electrical load and/or turbine trip

event, which were found acceptable to the staff as indicated above.

3.1.11 reak HSLB Co e

Res onse An MSLB could cause excess cooldown of the RCS.

with a negative moderator temperature coefficient, the RCS cooldown results in a reduction of core shut down margin.

Assuming the most reactive control rod is stuck in its fully withdrawn position, it is possible that the core will return to critical.

However, the core will be ultimately shut down by the injection of borated water from the refueling storage tank via the safety injection pumps.

The licensee states that the most limiting HSLB event is performed at hot zero power (HZP) conditions, which did not change for the power uprating.

In a letter dated June 11, 1996, the licensee provided its results of an analysis which reflected the uprated power conditions.

The transient minimum DNBR for a typical cell is 1.48 which is above the minimum allowable DNBR limit of

'.45.

Therefore, the results of the HSLB analysis meet the acceptance criteria for this event and are acceptable to the sta'ff.

3. 1.12 Ru ture of a Control od Orive Mechanism CROM RCCA E 'ection The mechanical failure of a CRDM pressure.housing could cause the ejection of the RCCA and drive shaft, resulting in a rapid reactivity insertion and possible localized fuel damage.

The results indicate that the radially averaged enthalpy remains'well below 280 cal/gm at any axial fuel location and, therefore, there is no danger of sudden fuel dispersal into the coolant.

Since the peak pressure does not exceed that which would cause stresses to exceed the Service Limit C, as described in the American Society of Mechanical Engineers (ASHE) Code,Section III, there is no danger of further

'onsequential damage to the RCS.

Therefore, the effect of the power uprate on the results of the RCCA ejection accident are acceptable to the staff.

The radiological consequences are evaluated in section 3.6 of this SE.

3.2 va uation of LOCA and LOCA Related Events 3.2. 1 Lar Break oss-of-Coolant Accident LBLOCA Anal sis The licensee has performed a reanalysis of LBLOCA to demonstrate conformance with the 10 CFR 50.46 requirements for the conditions associated with the uprating.

Peak cladding temperature (PCT) of 2103'F and 2082'F were calculated for the RCS low (562.7'F) and high (585.7'F)

Tavg conditions respectively.

After assessing the PCT effect for top skewed power shapes and containment purge on the most limiting case, the resulting maximum PCT for a

.LBLOCA is 2144'F.

The results of the reanalysis of a LBLOCA show that all requirements of 10 CFR 50.46 are met and are, therefore, acceptable to the staff.

.2.2 ~k The small break LOCA analysis utilizes the NOTRUHP computer code to calculate the transient depressurization of the RCS as well as to describe the mass and energy release of the fluid flow through the break.

The 3-inch equivalent

diameter cold leg break, high nominal vessel average temperature, was found to be the limiting case with a PCT of 1688 F.

The small break LOCA analysis for the uprate condition was previously approved by the NRC by Amendment numbers 184 and 190 on August 13; 1996, for implementation pending approval. of WCAP-10054-P, Addendum 2, Revision 1 (proprietary),

"Addendum to the Westinghouse Small Break LOCA ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection in the Broken Loop and Improved Condensation Model," October 1995.

3.2.3 Hot wite ove 0

The licensee has performed a calculation to determine the new HLSO time and minimum hot leg recirculation flow based on an uprated core power of 2300 MWt.

The new HLSO time is 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

The new hot leg recirculation minimum flow is 33 ibm/sec.

This hot leg recirculation minimum flow has been shown to be available.

The licensee has concluded that with the above HLSO time and flow

rate, the core geometry will remain acceptable.

The staff finds these results acceptable.

3.2.4 Post-OCA on Term Coolin The licensee has performed an evaluation to determine the effects of power uprating to post-LOCA long term cooling.

It is concluded that the Tavg range has a negligible effect on the post-LOCA sump boron concentration.

Therefore, the core will remain subcritical post-LOCA and that decay heat can be removed for the extended period of time required by the long-lived radioactivity remaining.

The revised post-LOCA long term core cooling boron limit curve is used to qualify the fuel on a cycle-by-cycle basis during the fuel reload process.

The staff finds the results acceptable.

3.3 Evaluation of Steam Generator Tube Ru ture SGTR Event The licensee has performed a reevaluation of the SGTR event using the methodology consistent with that used in the UFSAR.

This method does not include a computer analysis to determine the plant transient behavior following an SGTR.

Rather, simplified calculations were performed, based on the expected SGTR transient

response, to determine the primary to secondary break flow and the steam release to the atmosphere for use in calculating the offsite doses during the event.

Also, a single failure was not assumed in this analysis.

Although no single failure is explicitly modeled, the licensee considered the analysis provides a conservative estimate of the offsite doses following an SGTR.

The analysis assumes that the primary to secondary break flow is terminated at 30 minutes after the event initiation.

The residual heat removal (RHR) system is operating at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the SGTR and steam release is terminated at this time.

The radiological consequences are evaluated in section 3.6 of this SE.

3.4 Containment Inte rit Anal sis The licensee has performed containment integrity analyses at uprated power to ensure that the maximum pressure inside the containment will remain below the containment building design pressure of 55 psig if a design basis LOCA or MSLB inside containment should occur during plant operation.

The analyses also

established the pressure and temperature conditions for env'ironmenta1 qualification and operation of safety related equipment located inside the containment.

The peak pressure is also used as a basis for the containment leak rate test pressure to ensure that dose limits will be met in the event of a release of r'adioactive material to containment.

The licensee indicated that although the current licensed power is 2200 HWt, safety related systems (with the exception of the emergency core cooling system) were originally evaluated for core power level of 2300 HWt.

The emergency core cooling system was analyzed at the higher power as part of the uprate request:

The licensee indicated that the containment functional analyses included the assumption of the most limiting single active failure and the availability or unavailability of offsite power, depending on which resulted in the highest

'containment temperature and pressure.

Bounding initial temperatures and pressures for analyses were selected to envelop the limiting conditions for operation.

Previously, all three emergency containment cooling (ECC) units were automatically started on a safety injection (SI) signal.

The licensee indicated that to support post-LOCA long-term containment pressure/temperature

analyses, a minimum of one ECC is required to start immediately with a second ECC unit starting within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the event.

The revised design and TS would require only two ECCs.units to automatically start on SI signal and that the third (swing)

ECC unit be maintained in an operational condition and available for manual starting.

This change is required to limit the component cooling water system (CCWS) operating temperature du'ring injection and/or recirculation phase of the LOCA at uprated conditions.

3.4. 1 Ha Steam ine e

Containment Inte rit Anal sis:

The licensee has performed analyses to determine. the containment pressure and temperature response during postulated MSLBs inside containment for limiting conditions for operation at uprated power.

As in the current licensing basis FSAR, the uprated analyses were evaluated for initial power. levels of 102

percent, 70 percent, 30 percent, and zero percent and spectrum of break sizes similar to that in the current FSAR.

The MSLB mass and energy release and the pressure and temperature analyses have included the effects of various single failures.

The HSLB mass and energy releases were calculated using the LOFTRAN computer code and Containment temperature and pressure using the COCO computer code.

The LOFTRAN and COCO computer codes were used in the current design bases analyses and the staff has found the use of these codes acceptable.

As in the current analysis, the licensee indicated that the most limiting case with respect to peak containment pressure was determined to be a full double-ended rupture (DER) downstream of the flow restrictor'in main steamline at hot zero power (1.4 ft DER at HZP).

The most limiting single failure was found to be a failure of the main steam check valve (MSCV) on the faulted loop with offsite power available.

Initial containment pressure and temperature conditions for this limiting case were assumed to be +3.0 psig and 130'F.

For the MSLB, the uprating analyses calculated a peak containment pressure of

48. 1 psig and a peak temperature of 269.4'F for the limiting case.

The current FSAR had calculated a peak containment pressure of 42.8 psig for MSLB case.

The peak containment pressure and temperature at uprated conditions remains below the containment design pressure of 55 psig and temperature of

283 F.

It also remains below the FSAR transient analysis which calculated a

peak accident pressure of 49.9 psig and a peak accident temperature of 276 F.

Based on its review, containment pressure acceptable since the and temperatures are the staff finds the proposed change due to uprate in peak and temperature as a result of postulated HSLB is containment design and original peak accident pressures not exceeded.

3.4.2 CA C

ment I te rit Anal ses 1

The licensee has performed analyses to determine the containment pressure and temperature response during postulated LOCAs using mass and energy releases which incorporate revised design parameters corresponding to 2300 HWt,with updated computer modeling.

As in the current licensing basis FSAR, the postulated LOCA analyses were performed for the double ended hot leg (DEHL) guillotine break of reactor coolant pipe and the double ended pump suction (DEPS) break.

The cold leg break (between pump and vessel) has been found in previous studies to be much less limiting in terms of overall containment energy releases. 'he analyses were performed for a diesel failure, a

containment spray pump failure, no failure with minimum and maximum initial containment pressures.

These cases are shown to result in maximum pressure and temperature response.

The licensee indicated that the mass and 'energy releases in the containment are calculated using methods described in Westinghouse Topical Report WCAP-10325-A and the containment pressure and temperature response is calculated using the COCO computer codes.

Westinghouse Topical Report WCAP-8312A and COCO code were used for the current design bases analyses.

The updated Westinghouse WCAP-10325 computer code with same methodology and assumptions (except the Turkey Point specific data) have been used for Catawba,

McGuire,

. Sequoyah, Watts Bar, Surry, Hillstone Unit 3, and Beaver Valley Unit 2 and Indian Point Unit 2.

For the DEHL break, the Turkey Point uprating analyses calculated a

containment peak pressure of 48. 1 psig and peak temperature of 273.9'F.

For the DEPS breaks, the uprating analyses calculated a containment peak pressure of 46.2 psig and a peak temperature of'71. 1 F with loss of offsite power and initial containment pressure of 0.3 psig.

The uprated calculated LOCA peak pressure and temperature of 48. 1 psig and 273.9'F remains below the FSAR transient analysis peak accident pressure and temperature of 49.9 psig and 276'F and containment design pressure and temperature of 55 psig and 283'F.

In addition, all long-term cases were well below 50 percent of the peak value within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The licensee indicated that the reductions in the calculated peak pressure and temperature for the uprate power analyses were due to the use of revised methods for calculating the mass and energy releases to the containment and updated plant parameters.

The updated calculated pressure and temperature curves for LOCA and HSLB cases will remain bounded by the curves used for equipment qualifications'nd,for containment leak rate test pressure.

The licensee indicated that a

CCW thermal performance analysis was performed for the thermal uprate program.

This analysis also considered the LOCA and MSLB

transients.

When only one or two ECCs are assumed to start in a postulated

accident, CCWS acceptance criteria are met.

Based on the above discussion, the staff finds the licensee analyses for determining the containment peak pressure and temperature for design basis LOCA acceptable as the methodology and assumptions used for calculating mass and energy release and for calculating pressure and temperature transients have been used previously for plants of similar design to meet the requirements of Standard Review Plan (SRP) Section 6.2. 1.3 for mass and energy analyses and Section 6.2.1. I.A for dry pressurized water reactor (PWR) containment integrity peak pressure analyses.

The proposed change for power uprate will not affect the containment integrity as the calculated peak containment pressure of 48. 1 psig remains below the containment design pressure of 55.0 psig and containment leak rate test pressure of 49.9 psig.

3.4.3 S

o t-e Subcom artment Anal sis The licensee has indicated that the original design basis short-term LOCA mass and energy releases resulting from DERs of the primary loop piping for the subcompartment analyses will remain bounding for uprated power.

This is due to the application of the Leak-Before-Break (LBB) Technology to the short-term LOCA mass and energy releases.

Under LBB, the most-limiting break would be a

DER of one of the largest RCS loop branch lines (pressurizer surge line, accumulator/SI'ine, or RHR suction line).

Based on the above review, the staff concludes that the uprating is acceptable as the subcompartment pressure loading analysis from high-energy-line ruptures remain bounded by the current FSAR analysis.

The staff notes that use of LBB methodology has'een prevously approved by the NRC for use at Turkey Point.

3.5 Add t onal es' Bas s

and Pro rammatic Evaluations 3.5. 1 H d o en Generation Rates The licensee indicated that an analysis of containment post-LOCA hydrogen generation rate was performed for the uprated core thermal power of 2336 HWt (102 percent of 2300 HWt).

The analysis showed that with no recombiner in

service, the hydrogen concentration will not exceed four percent by volume for 17 days following a LOCA.

Placing a hydrogen recombiner in service prior to the 18th day following a LOCA will maintain containment hydrogen levels below the lower flamoability limit of 4 percent.

Based on the above review, the staff finds that the power uprate will not impact the post-LOCA hydrogen control system.

.5.2 The licensee performed evaluations to determine the impact of plant operations at the proposed power level on the following generic issues/programs:

10 3.5.2.1 o

wit 0

CFR 50 A

endix R

The licensee evaluated the analyses which were performed in support of the Appendix R evaluation for potential impact resulting from plant operations at the proposed power level.

The licensee stated that the evaluation did not identify changes to design or operating conditions that will adversely impact the ability to provide post-fire safe shutdown in accordance with Appendix R.

Since there are no physical plant configuration or combustible load changes resulting from the uprated power level operations, the staff concurs with the licensee that plant operations at the proposed power level will have no impact on the Appendix R evaluation previously performed.

3.5.2.2 t

B c o SBO The licensee performed evaluations of the impact resulting from plant operations at the proposed uprated power level on system response and coping capabilities for SBO events.

The licensee stated that with the exception of the minimum inventory of condensate required to be stored in the condensate storage tank (CST) to provide safe shutdown following an SBO event, no other changes to system design or operating conditions were identified.

The licensee stated that the CST minimum required volume would be higher.

The minimum usable volume which is required to support the design basis that the plant be maintained at hot standby for 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> followed by a 4-hour cooldown to RHR cut-in temperature (350'F) was determined to be 199,000 gallons for plant operations at the proposed power level.

Consequently, the license'e proposed to revise the TS to increase the CST minimum required volume from 185,000 gallons to 210,000 gallons.

Based on our review and the experience gained from our review of power uprate applications for similar PWR plants, the staff finds that the impact on system response and coping capabilities for an SBO event resulting from plant operations at the proposed uprated power level will be insignificant, and that the licensee's proposal to increase the TS CST minimum required volume from 185,000 gallons to 210,000 gallons is acceptable.

3.5.2.3 Generic Letter 89-13 "Service Water S stem Problems Affectin

-Related ui ment" The licensee performed evaluations of the effects of plant operations at the proposed power level on component cooling water (CCW) system.

In the CCW heat exchanger thermal analysis, revised heat exchanger parameters (e.g.,

fouling

factor, CCW and intake cooling water flow rates, etc.)

were used.

These revised heat exchanger parameters are to be included in the licensee's Generic Letter 89-13 program for monitoring the system and heat exchanger performance.

Based on our review, the staff finds that the above licensee's commitment meets the intent of Generic Letter 89-13 and, therefore, is acceptable.

3.6 11 The licensee reevaluated the effect of the power uprate on design basis accident (DBA) radiological consequences.

The original licensing DBA source terms for Turkey Point were considered.

The licensee also reevaluated the control room habitability under DBA conditions.

The licensee stated that the original radiological consequence analyses could not be exactly reconstituted.

Therefore, the licensee reconstituted analyses performed using methodology described in the UFSAR with the original licensing basis assumption at 2346 MWt (102 percent of requested power level).

The analyses also considered changes that had occurred since the original analyses were performed, including burnups, enrichments, fuel masses, and operating times.

The licensee's reconstituted analyses indicate that, for all DBAs, the calculated offsite radiological consequences doses are within the dose acceptance criteria stated in the SRP and 10 CFR Part 100 and also comply with the dose acceptance criteria for control room operators given in General Design Criterion (GDC) 19 of Appendix A to 10 CFR Part 50.

The staff independently performed confirmatory evaluations at the uprated power level of 2400 MWt by increasing the previously calculated doses in the original safety evaluation'eport'y 4.3X (from 2300 to 2400 MWt, 104 percent of requested power level).

The events reevaluated by the staff for the uprated power were the LOCA, MSLB,

SGTR, and the fuel handling accident (FHA).

The whole body and thyroid dose were calculated for the exclusion area boundary (EAB), the low population zone (LPZ), and the control room.

The following table contains the results of the staff. calculations compared to the licensees results.

Ac ide t a

rea Boundar sion

~FP 8 2346 MWt 102X NRC 8 2400 MWt 104X Steam Generator Tube Ru ture 0.068 rem thyroid 0.4 rem whole bod Fuel Handling Accident 33 rem thyroid

0. 1 rem whole bod 44 rem thyroid

<1 rem whole bod 1.5 rem thyroid

<1 rem whole bod Steam Line Break LOCA 0.042 rem thyroid 0.5 rem whole bod 24 rem thyroid 1.4 rem whole body 1.5 rem thyroid

<1 rem whole bod 66 rem thyroid 2 rem whole body Safety Evaluation by the Division of Reactor Licensing U.S. Atomic Energy Commission in the matter of Florida Power and Light Company, Turkey Point Units 3 and 4, Dade County, Florida, Docket Nos.

50-250 and 50-251.

March 15, 1972.

12 Po ul at on o

Fuel Handling accident 3.2 rem thyroid 0.24 rem whole bod 4.3 rem thyroid

<1'rem whole'od Steam Generator Tube Ru ture Steam Line Break LOCA 0.01 rem thyroid 0.002 rem whole bod 0.01 rem thyroid 0.00005 rem whole bod 2.7 rem thyroid 0.02 rem whole body

<1 rem thyroid

<1 rem whole bod

<1 rem thyroid

<1 rem whole bod 14 rem thyroid 1 rem whole body Control Room 15 rem thyroid 0.5 rem whole body 14 rem thyroid

<.2 rem whole body The staff finds that the offsite radiological consequences and control room operator doses at the uprated power level of 2300 MWt will continue to remain within the acceptance criteria stated in the SRP and within the 10 CFR Part 100 and the GDC 19 dose reference values for all DBAs.

Therefore, the staff concludes that the licensee's request to uprate the authorized maximum'reactor core power level by 4.5 percent to 2300 MWt from its current limit of 2200 MWt is acceptable.

4.0 S

RUC URE AND COMPONENTS EVA UATION 41 Re t e

t rit

4. 1. 1 P ess i ed he l Shock PTS 10 CFR 50.61 Assessment The staff reviewed FPL's PTS assessments of the Turkey Point reactor pressure vessels (RPVs) under the current and uprated power conditions for., the plants.

The current PTS calculations for Intermediate Shell'-to Lower'hell Circumferential Weld SA-1101 (the limiting material in the Turkey Point RPVs) are based on a chemistry factor (CF) value of 180 F and a margin term ("M") of 56.00, as determined in accordance with Regulatory Position

1. 1 and Table 1 of Regulatory Guide (RG) 1.99, Revision 2, for a ferritic weld containing 0.26 percent copper and 0.60 percent nickel.

Tables

2. 1-1 and 2. 1-2 provide a

comparison of the calculated end-of-life (EOL) RT>>, values for Weld No.

SA-1101 before and after the uprated power levels are licensed for the plants.

The data in Row 2 of the Tables correspond to the values for the current power levels (2200 MWt); the data in Row 3 of the Tables correspond to the values for the uprated power conditions (2300 MWt).

The RT>> values in Tables 2.2-1 and 2.2-2 are based on a

CF of 180 F, as determined Prom Table 1 of RG 1.99, Revision 2, for a ferritic weld material containing 0.26 percent copper and 0.60 percent nickel.

10 CFR 50.61 requires that the RT>>

values for ferritic circumferential weld materials in RPVs be less than 300 F at EOL.

Tables 2.2-1 and 2.2-2 indicate

13 that the RT>>, values at EOL for limiting material SA-1101 in the Turkey Point RPVs have increased by 2-3'F.

However, the new RT>>~ values for the limiting materials in the Turkey Point RPVs will still be within the PTS screening criteria even under the uprated conditions for the plant Therefore, the staff concludes that, in regard to the integrity of the Turkey Point

RPVs, FPL will continue to comply with the requirements of 10 CFR 50.61 under the uprated conditions for the plants.

4.1.2 8

l atin R-T Limit Curves The staff evaluates the P-T Limit Curves used for heatup,

cooldown, and normal operation of PNRs based on the following NRC regulations and guidance:

10 CFR Part 50, Appendix G; GL 88-11; GL 92-01, Revision 1;

GL 92-01, Revision 1,

Supplement 1;

RG 1.99, Revision 2; and Standard Review Plan (SRP) Section 5.3.2.

In GL 92-01',

Revision 1, the staff requested that licensees submit the RPV data for their plants to the staff for review.

This data is used by the staff as the basis for the staff's review of P-T Limit submittals, and as the basis for, the staff's review of PTS assessments,(10 CFR 50.61 assessments).

Appendix G to 10 CFR Part 50 requires that R-T Limits for the reactor pressure vessel (RPV) be at least as conservative as.those obtained by applying the'ethodology of Appendix G to Section XI of the ASME Boiler and Pressure Vessel Code.

SRP 5.3.2 provides an acceptable method of calculating the P-T Limits for ferritic materials in the beltline of the RPV based on the linear elastic fracture mechanics (LEFM) methodology of Appendix G to Section XI of the ASME Code (Appendix G).

The basic parameter of this methodology is the stress intensity factor K, which is a function of the stress state and flaw configuration of t(e material in question.

The methods of Appendix G

postulate the existence of a sharp surface flaw in the RPV that is normal to the direction of the maximum stress.

The flaw in the RPV is postulated to have a depth that is equal to one-fourth of the RPV beltline'hickness and a

length equal to 1.5 times the RPV beltline thickness.

The critical locations in the RPV beltline region for this methodology are the 1/4 thickness (1/4t) and 3/4 thickness (3/4t) locations, which= correspond to the depth of the maximum postulated flaw, if initiated and grown from the inside and outside surfaces of the RPV, respectively.

4.1.3 Su t e P evious Basis for A rovin the Current Set of R-T vsTS FPL's current set of P-T Limit Curves were approved by the staff in the License Amendments Nos.

134 and 128 to the respective Facility Operating Licenses DPR-31 and DPR-41, and the staff's Safety Evaluation (SE) to FPL, dated January 10, 1989.

In the staff's SE.of January 10, 1989, the staff concluded that the current P-T Limit Curves for Heatup and Cooldown would be acceptable until 20 effective full power, ars (EFPY).

The staff based its assessment of the P-T Limit Curves on the methods of SRP 5.3.2, and on the plant-specific RPV data.

It should be noted that FPL's current set of P-T Limit Curves for the Turkey Point units are based on the adjusted reference temperature (ART) values for

0 I

14 the 1/4t and 3/4t RPV locations (252.5'F and 200.4 F, respectively),

and on a

chemistry factor (CF) of 200.2'F.

The CF of 200.2 F was determined in accordance with the criteria of Position

2. 1 of RG 1.99, Revision 2, as determined from data obtained from Turkey Point Surveillance Capsules "T" and "V" from Unit 3, and Capsule "T" from Unit 4.

The surveillance capsules were removed in accordance with FPL's Integrated Surveillance Program for the Turkey Point units, which was previously determined by the staff to meet the requirements of 10 CFR Part 50, Appendix H, "Reactor Vessel Material Surveillance Program Requirements."

The staff approved the Turkey Point Integrated Surveillance Program by letter dated April 22, 1985.

Table 2.3-1.

provides a summary of the Turkey Point surveillance capsule data used to establish the CF (200.2 F) for the limiting RPV beltline material.

It should be noted that the staff's SE of January 10, 1989, did not address the credibility of the Turkey Point surveillance capsule data.

For this review, the staff performed a review of the Turkey Point surveillance capsule data.

The staff determined that, for the Turkey Point Unit 4 Surveillance Capsule T, the measured value of MT >> differed from the calculated value of MT

~ by 74'F.

This value exceeds tie scatter of MTQ>> data for ferritic welf materials by 46'F.

With the existing surveillance information, the staff considers that the data from Turkey Point Unit 4 Surveillance Capsule T would not be used.

However, the use of the surveillance data is within the Turkey Point licensing basis

and, as discussed

below, use of the surveillance data does not have a significant impact on the results.

Use of surveillance capsule data for establishing the adjusted reference temperatures listed in the Turkey Point P-T Limit Curves (TS Figures 3-4.2, 3-4.3, and 3.4-4) differs from the methodology used for FPL's latest PTS assessment (use of Position

1. 1 and Table 1 of RG 1.99, Revision 2).

The staff performed an independent calculation to determine what the adjusted reference temperatures would be at the 1/4t and 3/4t vessel locations if methodology of Position 1.1 and Table 1 of RG 1.99, Revision 2, were used for the calculation.

Tables 2.4-1 and 2.4-2 compare the adjusted reference temperatures, at the 1/4t and 3/4t locations, respectively, if the methodology of Regulatory Position 2.1 of RG 1.99, Revision 2 and surveillance data are

used, and if the methodology of Regulatory Position

1. 1 and chemical composition data are used.

Tables 2.4-1 and 2.4-2 indicate that, for Weld Heat No. 71249 (the limiting material in the Turkey Point RPVs), the use of

. chemical composition data and the Table 1 in RG 1.99, Revision 2; yields more conservative ART values than does surveillance capsule data.

However, since the P-T Limit curves incorporate the margins of Appendix G to Section XI of the ASIDE Code, the small differences in the ART values (as summarized in Tables 2.4-1 and 2.4-2, respectively) are not considered to be significant and the staff finds the results acceptable.

4. 1.4 Staf val ti n of FPL's Pro osed Chan es to the R-T Limit eat nd Cooldow C rves FPL's proposed changes to the current P-T Limit Heatup and Cooldown Curves for the Turkey Point units do not involve changes to the actual curves.

Instead, to account for the slight increase in the vessel neutron fluence levels, FPL proposed that the P-T Limit Curves be scaled back from 20 EFPY to 19 EFPY.

i

15 FPL justified the new expiration date based on the results of a plant specific calculation.

The licensye determined the length of time to amass a neutron fluence of 2.022E19 n/cm at'he inner surface of the

RPVs, based on the new uprated conditions for the units.

This calculation was based on a limiting neutron fluence of 1.80E19 n/cm at the RPV jnner surface after 16 EFPY, and an uprated neutron flux rate of 2.31E10 n/cm -sec.

FPL's calculations indicated that the current R-T Limit Heatup and Cooldown Curves would be applicable until 19 EFPY for Turkey Point Unit 3 and until 19.7 EFPY for Turkey Point Unit 4.

FPL has conservatively set the amended expiration date for the Turkey Point P-T Limit Curves to the more conservative value from the calculation (i.e.,

19 EFPY).

This is acceptable to the staff.

4.1.5 e t FP Com i

ce with 10 CFR Part 50 A

endi G:

U er Shelf o

ider tions In its letter to FPL dated July 24, 1995, the staff requested that 'FPL assess the effect of the proposed thermal power uprate on EOL USEs and FPL's equivalent margin analyses (EHAs) for the limiting USE materials in the Turkey Point RPVs.

On Hay 3,

1996, FPL responded that the EHA for the Turkey Point RPVs was performed using an'nner wall EOL fluence of 2.7E19 n/cm~,

as provided in BN Proprietary Report BAW-2118P (November

1991, Ref. 17).

The staff approved the EHA analysis for the Turkey Point units on October 19,

1993, as supplemented on Harch 29, 1994.

FPL's estimates for the uprated EOL fluences for the limiting USE mat~rials in the Turkey Point RPVs hyve been estimated to increase to 2.74E19 n/cm for Unit 3 and 2.68E19 n/cm for Unit 4, respectively.

Therefore, since the EOL neutron fluences for Welds SA-1101 will not change significantly as a result of the proposed power uprate, the staff concludes that the proposed power uprate will not affect FPL',s EHA for the Turkey Point RPVs, nor any of the conclusions stated in the staff's SEs of October 19, 1993 and Harch 29, 1994 and is, therefore, acceptable to the staff.

4. 1.6 Con usions - Vessel Inte rit Considerations The EHCB staff has reviewed the FPL submittals and determined that FPL will still comply with the requirements of 10 CFR 50.61 and the requirements of 10 CFR Part 50, Appendix G under the uprated power conditions for the plants.

The staff has also determined that FPL's proposed scaling back of the current set of P-T Limit Curves to 19 EFPY is acceptable.

The staff, therefore, concludes that, with respect to the structural integrity of the Turkey Point reactor pressure

vessels, the proposed thermal power uprate is acceptable.

The staff notes that, by letter dated July 1,

1996, FPL committed to provide a

new P-T limit curve analysis for NRC review a minimum of 6 months prior to the expiration of the Turkey Point P-T Limit Curves.

'FPL also committed, by the same letter, to include in the limiting material property evaluation, (1) the data from the three surveillance capsules previously removed from Turkey Point Units 3 and 4, and (2) supplemental surveillance data from capsules being irradiated in the Davis Besse Reactor Vessel.

The licensee st'ated that an evaluation of the temperature and fluence environment between the host plant (Davis Besse) and Turkey Point will be provided demonstrating the

16 applicability of the surveillance data to the Turkey Point limiting materials.

FPL indicated that it plans to utilize the Linde 80 generic initial RTNDT lower bound value of -27 F.

The licensee reported that the power increase will result in changing the design parameters given in Table 2.1-1 of WCAP-14276.

Table 2. 1-1 provides various cases that were developed for use in"the power uprate analysis.

There are no significant changes in thermal transients and LOCA blowdown forces as a

result of the power uprating.

The licensee evaluated the design and operation of the regions of the reactor vessel affected by the temperature change and

fluence, based on the proposed uprated core power.

The evaluation included a

review of the reactor vessel design specifications, stress report and fracture mechanics analyses.

The regions of the reactor vessel affected by the temperature change include the RPV (main closure head flange, studs, and vessel shell),

CROM nozzles, core support pads, vent nozzles and the instrumentation tubes.

The licensee evaluated the maximum ranges of stresses and cumulative fatigue usage factors for the critical components at the core power uprated conditions.

The evaluation w'as performed in accordance with the ASME Boiler and Pressure Vessel

Code,Section III, 1965 Edition, with addenda through the Summer 1966 to assure compliance with the code o'f record.

The licensee indicated that the core power uprate does not affect the maximum stress ranges in the existing reactor vessel stress reports for Turkey Point Units 3 and 4, and the maximum cumulative fqtigue usage factors remain significantly below the allowable ASME Code limit of 1.0.

On the basis of its review, the staff concurs with the licensee's conclusion that the reactor vessel is acceptable for the proposed core power uprate.

4.3 Reactor Core Su ort Structure and Vessel Internals By letters dated June 11, 1996, the licensee provided the additional information requested, by the staff, with regard to the evaluation of the reactor vessel core support and internal structures.

The limiting reactor internal components evaluated include lower core plate, core barrel, baffle plates and baffle/barrel region bolts.

The licensee evaluated the upper and lower internals considering the worst case set of operating parameters provided in Table 2. 1-1 of WCAP-14276.

Stresses and cumulative fatigue factors for the limiting internal components

'at the power uprate conditions are below the allowable limits of the original design basis which had been previously reviewed by the staff.

Further, the licensee performed the flow-induced vibration analysis on the

. guide tubes and the upper support column at the uprated power level.

The evaluation indicated that the existing analysis provides sufficient margins to accommodate the increase in the flow-induced vibration loads due to the power uprate.

17 On the basis of the above evaluation, the staff concluded that the reactor internal components at Turkey. Point Units 3 and 4 will remain wi.thin the allowable limits of stress and fatigue usage factor for operation at the proposed uprated power conditions.

4.4 Re ct r C

o ant Pum CPs The licensee evaluated the RCPs by reviewing the design specifications in comparison with the proposed uprated conditions.

.At the core power uprate, the reactor coolant system pressure remains unchanged.

There are no significant changes to the design thermal transients.

The small fluctuation (6 F) in the RCP inlet temperature has an insignificant effect on the pressure boundary stresses.

On the basis of its review, the staff concurs with the licensee's conclusion that the current Model 93

RCPs, when operating at the proposed power uprated conditions, will remain in compliance with the requirements of the codes and standards under which the Turkey Point Units 3 and 4 were originally licensed.

4.5 Control od Drive echanisms The licensee evaluated the adequacy of the CRDMs by reviewing the Turkey Point current Model L106B CROM design specifications and stress report to compare the design basis input parameters against the operating conditions at the uprated core power.

Based on this evaluation, the licensee concluded that the original design basis thermal and structural analyses are bounding for the core power uprate.

On the basis of its review, the staff concurs with the licensee's conclusion that the current design of CRDMs continues to be in compliance with codes and standards under which the plant was licensed, for the power uprated conditions.

4.6 NSSS i i a d Pi e

Su orts The proposed power uprate of Turkey Point Units 3 and 4.involves the increase of temperature difference across the Reactor Coolant System (RCS).

The design input parameters that define the various temperature conditions associated with the full power operating conditions of the plant were given in Table

2. 1-1 for both the current and the power uprated conditions.

The licensee does not project a change in the RCS loop pressure as a result of the proposed core power uprate.

At Turkey Point, the existing design basis thermal analyses of the NSSS piping and supports were reviewed by the licensee, in comparison with the uprated power conditions, with respect to the design system parameters and transients.

The licensee concluded that the existing design basis stress analyses for the RCS system piping and supports and systems connecting to the RCS system, remain valid for the power uprated conditions.

The evaluation was performed in accordance with the American Standards Association (ASA) B31. 1 Power Piping Code to assure'ompliance with the code of record at Turkey Point Units 3 and 4.

The staff finds that the increase in temperature difference across the RCS system, will have an insignificant effect on the NSSS piping, and will

18 minimally impact the design basis analysis of the piping and pipe support.

Therefore, the existing NSSS piping and supports, the primary equipment nozzles',

the primary equipment supports, and the branch lines connecting to the primary loop piping will remain in compliance with the requirements of the design bases criteria as defined in the

FSAR, and are acceptable for the power uprate.

4.7 The licensee evaluated the adequacy of the pressurizer and components including the pressurizer spray nozzle, safety and relief nozzle, upper head/upper shell, manway and instrument nozzle, the pressurizer surge nozzle, lower head/heater well, and support skirt for operation at the uprated conditions.

The evaluation was done by modifying the existing Turkey Point pressurizer stress report and design basis analyses of the pertinent pressurizer components.

The licensee found that the uprate conditions are bounded by those used in the original pressurizer stress analyses.

However, the original fatigue analyses were updated to account for the uprated power conditions.

The licensee.concluded that stresses and cumulative fatigue usage factors remain in compliance with the requirements of the ASME Code,Section III; 1965 Edition through Summer 1965 Addendum.

On the basis of its review, the staff concurs with the lic'ensee's conclusion that the existing pressurizer and components remain adequate for the plant operation at the proposed uprated core power.

4.8 Ste m Generato SGs The licensee evaluated the SGs by comparing the power uprate conditions with the design parameters of the Westinghouse Model 44F SGs at Turkey Point.

The comparison shown in Table 2. 1-1 of WCAP-14276 indicates that critical design system parameters such as the primary and secondary side pressures, as well as the vessel outlet and secondary side temperatures, are not significantly affected by the uprated power conditions.

The variation in the primary-to-secondary pressure differential is within about 3 percent.

The licensee indicated that there are no significant changes to the design transients as a

result of the core power uprate.

The stress level and cumulative fatigue usage factors of the. critical SG components continue to remain in compliance with the requirements of the 1965 Edition of the ASME Code,Section III through the Suaeer 1965 Addenda.

On the basis of its review, the staff concurs with the licensee's conclusion that the current Turkey Point Units 3

and 4

SGs are acceptable for the proposed core power uprate.

4.8. 1 SG T b nte rit Review 4.8. 1. 1 ffe t of the Power U rate on SG Tube Inte rit FPL contracted with the Westinghouse Corporation to evaluate the structural integrity of SG tubes under the uprated power conditions.

The effects of the power uprate on the SG tube integrity are summarized in Westinghouse Topical

Report, WCAP-14276, Revision 1.

Westinghouse evaluated the effects of the uprated power conditions on structural integrity of the SG tubesheets, tubesheet junctions, tube to tubesheet

welds, tubes secondary shell, minor

19 shell penetrations, and feedwater nozzles.

Westinghouse evaluated the SG tubes for two.different plugging cases:

(1) no tube plugging occurs in the SG; and (2) 20 percent of the tubes in the SGs are plugged.

Each case used three different multiplying factors as input parameters to account for variations under increased power conditions.

Westinghouse estimated that variations in the primary system pressure under uprating conditions were within 1 percent of the reference conditions.

Variations in the secondary side pressure were about 6 percent.

Variations in the primary-to-secondary pressure differential were about 3 percent.

From these variations, a factor of 1.01 was used for the primary side pressure, 1.06 for the secondary side

pressure, and 1.03 for the primary to secondary pressure differential.

These factors were incorporated in the evaluation to adjust pressure stresses under steady-state conditions to the corresponding pressure stresses under the uprating conditions.

4.8.1.2 m

T c ness onsiderations FPL stated that the current plugging limit of 40 percent through wall in the TS would still satisfy the minimum Code wall thickness requirements, even under the uprated power conditions.

Using conservative allowances for eddy current measurement uncertainty and continued crack growth, FPL established that an unflawed wall thickness of 0.020 inches would satisfy the minimum wall thickness requirements of the ASNE Code for the SG tubes.

The average tube wall thickness in the Turkey Point SGs is 0.050 inches.

Therefore, FPL concluded that the 40 percent through wall SG plugging limit in the Turkey Point technical. specifications would continue to provide adequate margin to the minimum required wall thickness.

The staff finds FPL's assessment on this issue acceptable.

'4.8. 1.3 Tube W a Cons derations FPL evaluated the U-bend region of the tubes in order to determine whether the uprated conditions would induce additional tube wear from anti-vibration bars.

FPL stated that the increase in steam flow and concurrent increase in void fraction could increase vibration in the U-bend region.

FPL stated that the additional vibration in the small radius U-bends would not lead to significant increases in fatigue-type degradation o'r tube wear.

FPL also evaluated the larger radius U-bends for increased wear from the anti-vibration bars.

FPL stated that the number of U-bends that are subject to wear at the anti-vibration bar intersections as a result of the uprated power conditions would constitute less than 0.3 percent of the total tube count over the life of the SGs.

This number is insignificant in contrast to the total number of tubes in the SGs.

The staff concludes that the number of plugged tubes from additional wear by the anti-vibration bars is insignificant under the uprated power conditions.

4.8. 1.4 Cor osion and Foulin Considerations FPL stated that the increase in average heat flux resulting from the power uprating could increase the potential for corrosion and long-term fouling.

However, FPL also stated that Turkey Point SGs have not experienced

t

20 significant corrosion or fouling.

The staff reviewed FPL's inservice inspection reports for the Turkey Point Units 3 and 4

SG tubes, dated January 17, 1996 and October 6, 1995, respectively.

The inspection reports did not indicate any evidence of significant degradation in the Units 3 and 4

SG tubes.

Even'if additional corrosion were to occur in any of the Turkey Point SG tubes, the inservice inspection requirements and plugging limit in the TS would provide adequate assurance of the structural integrity of the SG tubes.

4.8.1.5 Re l tor Guide 1.

21 Anal sis Considerations RG 1. 121 is a staff guidance for the assessment of the structural integrity of degraded SG tubes.

Because no active corrosion or other degradation phenomena are occurring within the Turkey Point SGs, a plant specific RG 1. 121 analysis is not necessary.

The staff concurs that the SG tubes in Turkey Point Unit 3 and 4 have not shown significant degradation to date; therefore, no RG F 121 analysis is required at this time.

4.8. 1.6 SG Tub Survei lance Considerations FPL stated that the scope of its SG inspections has exceeded TS requirements in each of the past three refueling outages.

These inspections included bobbin coil inspection for 100 percent of full length tubes and motorized rotating pancake coil inspection of tube manufacturing anomalies on a sampling basis.

FPL has determined that manufacturing anomalies affect a limited number of tubes in each SG.

The anomalies include minor denting at support intersections and minor over-expansion of the tube expansion transition at the top of the tubesheet.

The tubes with these anomalies may be more susceptible to inter-granular attack or stress corrosion cracking than tubes without the anomalies.

However, FPL added that corrosion has not been experienced in any of Turkey Point SG tubes and no significant amounts of degradation or wear is expected in the future.

In addition, FPL has stated that it will follow the protocol in the report, "PWR Steam Generator Tube Examination Guidelines," for future SG tube inspections.

The scope of this report covers inspection

methods, equipment, personnel training and qualifications.

Based on this information and the current status of corrosion in the SGs (i.e.,

no corrosion mechanisms to date),

the staff concludes that the scope of the inspection is sufficient to provide assurance to the structural integrity of the tubes.

4.8. 1.7 Con io s e

rdin SG Tube Inte rit The staff has reviewed FPL proposed license amendment in regard to the effect of the uprated power on the SG tube integrity.

The staff has determined that the proposed uprated power will not affect the 40 percent through wall plugging limit required by the TS, nor significantly increase the wear of tubes by the anti-vibration bars.

The staff has also determined that the uprated power is not expected to cause a significant increase in the corrosion of the SG tubes.

Because the corrosion of the Turkey Point SG tubes is insignificant, the staff has determined that a

RG 1. 121 analysis is not needed at this time, and that the current scope for the inspection of the SG tubes is sufficient to monitor for degradation of the tubes at this time.

Therefore, the staff concludes that FPL has provided reasonable assurance that the

0 t

l

21 structural integrity of Turkey Point SG tubes will be maintained under the uprated power conditions.

4.9 S

-of-1 OP Interface S stems 4.9.1 i

Feedwate S stem Condensate Stora e Tank The licensee performed evaluations of the effects of plant operations at the proposed uprated power level on auxiliary feedwater (AFW) system/condensate storage tank.

It was determined that the AFW system components have sufficient margin to provide the required flow and pressure.

The minimum usable.CST volume required during an SBO event to maintain the plant at hot standby for 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> followed by a 4-hour cooldown to RHR cut-in temperature (350'F) would be higher and was determined to be 199,000 gallons for plant operations at the proposed power level.

Consequently, the licensee proposed to revise the TS to increase the CST minimum required volume from 185,000 gallons to 210,000 gallons.

Based on our review and the experience gained from our review of power uprate applications for similar PWR plants, the staff concludes that the AFW system and the proposed TS.,CST minimum required volume of 210,000 gallons are acceptable for plant. operations at the proposed.power level.

4.9.2 Com e

Coo n

W ter The CCW system provides cooling water to various safety systems including three emergency containment coolers (ECCs) and non-safety systems during all phases of plant operations.

The CCW system is a closed loop system which serves as an intermediate barrier between the plant ultimate heat sink and systems which contain radioactive or potentially radioactive fluids in order to eliminate the possibility of an uncontrolled release of radioactivity.

Ultimate heat sink cooling flow is provided by the intake cooling water (ICW) system.

The licensee stated that the CCW system heat loads resulting from plant operations at the proposed uprated power level will increase slightly.

The increases in heat loads are from the spent fuel pool (SFP) cooling system during both power and refueling operations, and RHR system during plant shutdown.

The licensee performed evaluations of the effects of plant operations at the proposed power level on CCW system.

Results of the evaluations indicate that when all three ECCs are allowed to operate following a loss-of-coolant accident (LOCA),

CCW system operating temperature can exceed its maximum allowable limits.

When only one (following a LOCA, only one ECC is required to keep the containment temperature and pressure from exceeding design limits) or two ECCs are assumed to start, CCW system acceptance criteria are met.

Therefore, the licensee concluded that the CCW system has adequate capacity to perform its intended cooling function providing that no more than two ECCs are allowed to start automatically following a LOCA.

The licensee stated that as part of power uprate

program, design changes will be made to assure that no more than two ECCs vill automatically start in response to an accident.

Based on our review, the licensee's commitment to the CCW design changes

above, and the experience gained from our review of power uprate applications

t

22 for similar PWR plants, the staff concludes that the CCW system is acceptable for plant operations at the proposed uprated power level.

4.9.3 S

e ool Coo in S ste The spent fuel pool cooling system (SFPCS) was designed to remove the decay heat released from the spent fuel assemblies stored in the SFP; to maintain the SFP wat'er temperature at or below the design temperature of 150'F during plant operations and refueling; to maintain its cooling function.during and after a seismic event; and to structurally withstand a design temperature of 212'F.

The decay heat released from irradiated fuel will increase slightly following plant operations at the proposed power level.

Turkey Point routinely offloads the full core during refueling outages.

The licensee analyz'ed this condition for the uprate condition and concluded that adherence to the current administrative limit of 140'F (i.e., stopping the offload if the SFP temperature reaches 140'F) will maintain the peak pool temperature below 150'F'.

The analysis assumed that eight fuel assemblies are transferred to the spent fuel pool each hour.

The licensee stated that this offload rate exceeds the capacity of the fuel transfer equipment and maximizes heat input into the spent fuel pool.

The staff concludes that operation in the uprated condition is acceptable since the SFP temperature will remain below 150'F for normal refueling.

The licensee also performed an analysis for the case of full core offload following a forced shutdown with a I/2 core recently offloaded (36 days after shutdown) and a complete loss of SFP cooling.

The analysis indicated that with a complete loss of SFP cooling', the SFP water temperature will rise and eventually reach boiling.

The calculated minimum time from the loss-of-pool

'cooling until the pool boils is 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> and the maximum boil-off rate is 76.3 gpm.

makeup water in excess of the boil-off rate can be provided to the pool from the refueling water storage tank or via temporary connections from the fire water system or the primary water storage tank.

The minimum time to boil allows ample time to restore the SFP cooling function or align makeup water supplies.

Therefore, the minor increase in decay heat resulting from the power uprate does not impair the ability of, operators to recover from a loss of cooling and the staff concludes that operation in the uprated condition is acceptable.

Overall, based on our review, evaluations described

above, and the experience gained from our review of power uprate applications for similar PWR plants, we conclude that plant operations at the proposed power level will have an insignificant impact on the SFPCS.

Also, an issue associated with spent fuel pool cooling adequacy was identified in NRC Information Notice 93-83 and its Supplement I, "Potential Loss of Spent Fuel Pool Cooling Following a Loss of Coolant Accident (LOCA)," dated October 7, 1993 and August 24, 1995, respectively, and in a 10 CFR Part 21 notification, dated November 27, 1992.

The staff is evaluating this issue, as well as broader isCues associated with spent fuel storage

safety, as part of the NRC generic issue evaluation process.

If the generic review concludes that additional.requi} ements in the area of spent fuel pool safety are

23 warranted, the staff will address those requirements to the licensee under separate cover.

4. 10 T rbi rator S stems The licensee performed evaluations on turbine operations with respect to design acceptance criteria to verify the mechanical integrity under the conditions imposed by plant operations at the proposed uprated power level.

Results of the evaluations showed that there would be no increase in the probability of'urbine overspeed nor associated turbine missile production due to plant operations at the proposed uprated power level.

Therefore, the licensee concluded that the turbine could continue to be operated safely at the proposed uprated power levels.

Based on our review and the experience gained from our review of power uprate applications for similar PWR plants, the staff agrees. with the licensee that operation of the turbine at the proposed uprated power level is acceptable.

4. 11 ui ment uglification E

Inside and Outside Containment The licensee evaluated the effects of plant operations at the proposed power level on qualified equipment including safety-related electrical equipment and mechanical components.

With regard to the radiological dose used for'g, the licensee stated that the existing dose used for Eg was calculated based on a reactor power level of 2300 NWt.

The licensee reperformed the dose analyses for Eg evaluation based on a reactor power level of 2346 HWt (2300 HWt plus 2 percent) and concluded that the eXisting Eg is still valid for plant operations at the proposed power level.

With regard to the temperatures and pressures used for qualifying equipment inside containment, the licensee stated that results of the revised containment analysis indicate that containment temperatures and pressures are within the existing Eg profiles, except for the long-term temperature at 31 days.

The revised analysis indicates an increase of 2.4'F at 31 days.

However, this is within the normal range for containment temperature (104'F-130'F).

Therefore, the temperature profile for the accident duration of 31 days is still acceptable and plant operations at the proposed power level will not have. an adverse impact on the Eg program.

With regard to high energy line break analyses which support equipment environmental qualification outside containment, the licensee stated that the existing calculations remain bounding for plant operations at the proposed power level.

Since the Eg parameters affected'y the proposed changes remain bounded by the values used in the existing Eg program, and based on the experience gained from our review of power uprate applications for similar PWR plants, the staff concludes that plant operation at the proposed uprated power level will have an insignificant or no 'impact on the Eg of electrical equipment and mechanical components inside and outside containment and, therefore, is acceptable.

5.0 AN UTON 24 The licensee performed an evaluation of the effects r'esulting from plant operations at the proposed uprated power level on the main steam system including the main steam isolation valve (HSIV), main steam check valve (HSCV), main steam bypass valve (HSBV) and main steam safety valve (HSSV).

The licensee stated. that the steam flow resulting from plant operations at the proposed uprated power level will be 10,061,000 lb/hr which is approximately 5 percent above the design 'flow of 9,6000,000 lb/hr.

The main steam design conditions of 1085 psig and 600'F remain unchanged and bound all predicted operating conditions for the system and components.

The licensee concluded that, with the exception of HSSV discharge piping, plant operations at the proposed uprated power level will have an insignificant or no impact on the main steam system and its associated components.

\\

The licensee stated that HSSV discharge pipe backpressure will be higher at the uprated conditions and a modificati'on to the HSSV discharge piping will be required to ensure adequate margin for plant operations at the proposed uprated power level.

Based on our review and the experience gained from our review of power uprate applications for similar PWR plants, the staff considers that plant operations at the proposed uprated power level will have an insignificant or no impact on the main steam system.

The licensee evaluated the steam dump system for the plant operations at 2300 HWt reactor power level and stated that all of the system operating conditions are bounded by the existing design conditions.

Based on the experience gained from our review of power uprate applications for similar PWR plants, we find that plant operations at the proposed uprated power level do not change the design aspects and operations of the steam dump system. 'herefore, the staff concludes that operation of the steam dump system at the proposed uprated power level is acceptable.

5.3 Conden t

and eedwater S stem The licensee evaluated the condensate and feedwater systems for the plant operations at 2300 HWt reactor power level and stated that all of the system operating conditions are bounded by the existing desi'gn conditions.

Since these systems do not perform any safety related function, the staff has not reviewed the impact of plant operations at the proposed uprated power level on the design and performance of these systems.

5.4 Extraction Steam S stem The extraction steam system is designed to provide steam at various pressures and temperatures to preheat condensate and feedwater as it flows from the main condensers to the SGs.

Since the extraction steam system does not perform any

25 safety related function, the staff has not reviewed the impact of plant operations at the proposed uprated power level on the extraction steam system.

5.5 Ci c 1

Mater S stem The circulating water system is designed to remove the heat rejected to the condenser by turbine exhaust and other exhausts over the full range of operating loads, thereby maintaining adequately low condenser pressure.

The licensee stated that performance of this system was evaluated for power uprate and determined that the system is adequate for uprated power level operation.

Since the circulating water system does not perform any safety function, the staff has not reviewed the impact of plant operations at the proposed uprated power level on the designs and performances of this system.

5.6 Tu b ne Plant poli Mat r S stem TPCW The TPCM system is a closed-loop cooling water system and provides cooling water during normal operation to various non-. safety related equipment coolers.

The licensee stated that performance of this system was evaluated for power uprate and determined that the system is adequate for uprated power level operation.

Since the circulating water system does not perform any safety function, the staff has not reviewed the impact of plant operations at the proposed uprated power level on the designs and performances of this system.

5.7 nt Coolin Wate CW S stem The ICW system is designed to supply cooling water to safety-related CCW system equipment during a station blackout event and a

LOCA or main steam line break accident, and non-safety related TPCW system during normal plant operation.

The licensee performed evaluations of the effects of plant operations at the proposed uprated power level on the ICM system and concluded that the ICW system as designed will supply sufficient water to remove the additional heat loads resulting from plant operations at the proposed uprated

'ower level.

. Based on our review and the experience gained from our review of'power uprate applications for similar PWR plants, the staff finds that plant operations at the proposed uprated power level do not change the design aspects and operations of.the ICW system.

Therefore, the staff concludes that plant operations at the proposed uprated power level have an insignificant or no impact on the ICW system.

'.8 Heati Ventilatio and Air Conditionin The following heating, ventilating, and air conditioning (HVAC) systems were evaluated to ensure that they are capable of supporting the plant uprate conditions:

26 control room DC equipment/invertor rooms Cable spreading 5 computer equipment rooms Radwaste building Fuel handling building 480 V load centers 8 4. 16 kV. switchgear rooms Auxiliary building Unit 4 emergency diesel generator building Electrical equipment room Containment penetrations During normal plant operation, these HVAC systems cool, heat, and ventilate plant areas to maintain a suitable environment for plant personnel and equipment, as appropriate.

The licensee stated that these HVAC systems will continue to maintain normal operating temperatures at or below their maximum normal.operating temperatures.

In addition, regarding the control room emergency ventilation system, the existing TS requires a methyl iodide removal efficiency of 90 percent.

The licensee stated that the required methyl iodide removal efficiency is being increased to 99 percent to assure consistency between testing efficiency and analysis assumptions for post accident control room doses.

This increase is consistent with the recommendations of RG 1.52, "Design, Testing, and Maintenance Criteria for Post Accident Engineered-Safety-Feature Atmosphere Cleariup System Air Filtration and Absorption Units of Light-Water-Cooled Nuclear Power Plants,"

and is more conservative.

The staff considers it acceptable.

The TS change associated with this area is described in section 6.13 of this SE.

Based on our review and the experience gained from our review of power uprate applications for similar PWR plants, we find that plant operations at the proposed uprated power level do not change the design aspects and operations of the HVAC systems (except as previously discussed).

Therefore, we concur with the licensee that plant operations at the proposed uprated power level will have an insignificant or no impact on these HVAC systems.

5.9 Miscell neous S stems The licensee stated that various systems were evaluated and found not affected by the power uprate.

The following are major plant systems that were not affected by power uprate:

Instrument air system Auxiliary steam and condensate recovery system Feedwater heaters Condensate polishing system

Heater, moisture separator and reheater drain system Main condenser Since plant operations at the proposed uprated power level do not change the design aspects and operations of these

systems, and these systems do not perform any safety function, the staff did not review the impact of the

27 uprated power level operation on the designs and performances of these

'ystems.

5.10 te s

id and Gaseous The liquid and gaseous radwaste activity is influenced by the reactor, coolant activity which is a function of the reactor core power.

The licensee stated that the existing design of the radwaste systems is based on the core power level of 2300 HNt.

Therefore, plant operations at the proposed uprated power level will have an insignificant or no impact on the radwaste systems.

Based on our review, the staff agrees with the licensee that plant operations at the proposed uprated power level will have an insignificant or no impact on the radwaste systems.

5. 11 Add't' Ba a ce o

Pla t BOP Reviews The impact of plant operations at the proposed uprated power level on High Energy Line Break (HELB) Outside Containment and Equipment Environmental gualification is addressed in section 4.11.

5.12

~BP~Pi )~n The licensee evaluated the adequacy of the BOP piping systems based on comparing the existing design-bases parameters with the core power uprate conditions.

The code of record for BOP piping at Turkey Point is ASA B31. 1-1955.

In its letter dated June 11, 1996, the licensee indicated that the American National Standards Institute (ANSI) 831. 1 Power Piping Code, 1973 Edition with addenda through Summer 1976 (the code) was used for the power uprate at Turkey Point.

The staff finds the methodology to be acceptable considering that the stress limits in the code are generally conservative in comparison with the stress limits specified in the Turkey Point UFSAR.

On the basis of its analysis, the licensee concluded that the BOP piping, pipe supports and equipment nozzles remain acceptable and continue to satisfy design basis requirements for the power uprate.

In addition, the design bases pipe break analyses were also reviewed by the licensee to evaluate the effects of the upr'ate conditions on the pipe break locations, jet thrust and jet impingement forces which were used in the plant hazard analyses and the design of pipe whip restraints.

The review verified that the existing postulated pipe break locations are not affected by the power uprate since the design bases piping analyses will not change due to the power uprate.

The current design bases for jet thrust and jet impingement forces due to postulated pipe breaks for these systems are not affected by the

uprate, since the systems do not experience. pressure increase as a result of the core power uprate.

Based on its review, the staff concurs with the licensee's conclusion that the original design analyses for the pipe break locations, jet thrust, jet impingement and pipe whip restraints are unaffected by the power uprate.

28 On the basis of the above evaluation, for all the secondary-side systems

reviewed, the staff concurs with the licensee's conclusion that the power upr ate will have no significant impact on the BOP design bases.

6.0 EV UAT 0 CHANG 0 TS AND FACILITY OPERATING LICENSE FOL The FOL and TS changes requested by the licensee in their power uprate submittal are:

6.1 ce o d't The licensee proposes to change License Condition 3.A for Operating License DPR-31 and DPR-41 "Maximum Power Level" from 2200 HWt to 2300 HWt.

As documented in WCAP-14276, the licensee has provided the results 'of its reanalyses or evaluation including LOCA and Non-LOCA-transients and,accidents, containment

response, radiological consequences, NSSS and BOP systems and components to support the operation of Turkey Point Units 3 and 4 at an uprated power level.

The staff has reviewed the licensee's submittal and concluded that, for the reasons stated in this SE, both Turkey Point Units can safely operate at a core power of 2300 MWt.

6.2 TS 4

efinition o "RATED THERMAL POWER" The license proposed changing 2200 HWt to 2300 HWt to reflect the new uprated power level.

The staff finds this change acceptable as specified previously.

6.3 TS i

r

-1 and T ble 2.2-1 For TS Figure 2. 1-1, "Reactor Core Safety Limits Three Loops in Operation,"

the licensee proposed revising Figure 2.1-1 to reflect changes associated with

, the new operating conditions at the uprated power level.

For TS Table 2.2-1, "Reactor Trip System Instrumentation Trip Setpoints,"

Functional Unit 5, Overtemperature hT and Functional Unit 6 - Overpower hT, the revised core safety limits required changes to the overtemperature hT and Overpower bT setpoints.

Use of the Revised Thermal Design Procedure (RTDP) methodology and the inclusion of site specific instrument uncertainties resulted in changes to the other values associated with overtemperature hT and

, Overpower bT.

For TS Table 2.2-1, "Reactor Trip System Instrumentation Trip Setpoints" Functional Unit 10 - Reactor Coolant Flow-Low, FPL proposed changing the loop design flowrate from "89,500 gpm" to "85,000 gpm" for analyzed increase in the percentage of plugged steam generator tubes.

For TS Table 2.2-1, "Reactor Trip System Instrumentation Trip Setpoints" Functional Unit 11 - Steam Generator Water Level - Low-Low and Functional Unit 12 - Steam Generator Water. Level - Low, the licensee proposed changing the allowable value to incorporate plant specific uncertainties.

Core safety limits for three loops in operation (TS Figure 2. 1-1) have been revised to account for the proposed power uprating using the Revised Thermal

29 Design Procedure (RTDP) methodology.

The RTDP. methodology has been previously approved by the NRC and implemented at Turkey Point by FPL.

The increased power level as well as increased peaking factors and a loop design flow reduction of 4500 gpm were included in the revised safety limits.

In

addition, new overtemperature hT (OTAT) and overpower hT (OPBT) trip setpoints were generated based on the new core safety limits.

Each transient that is sensitive to the changes in these setpoints (i.e.,

rod withdrawal at power, boron dilution, and loss of load) has been analyzed by the licensee and in all

cases, the applicable acceptance criteria, as stated in NUREG-0800 (Standard Review Plan),

were met.

The revised trip setpoints provided adequate protection to maintain the minimum value of departure from nucleate boiling ratio (DNBR) larger than the safety analysis limit and to maintain the reactor coolant system (RCS) pressure below 110 percent of the design pressure.

Therefore, we find the revised core safety limits acceptable.

RTDP Instrument Uncertainties

Use of the RTDP methodology requires that variances in the plant operating parameters, pressurizer

pressure, primary coolant temperature, reactor power, and reactor coolant system flow, be justified.

Therefore, in support of the power uprate, FPL submitted Revision 2 to the RTDP methodology (WCAP-13719) which addressed the changes to the instrument uncertainties for, the primary system operating parameters as a

result of the increase in power level.

These uncertainty values are acceptable and they, or more conservative

values, have been used in the RTDP analysis.

The revised core safety limits of TS Figure 2.1-1 required changes to the OTBT and OP4T.setpoints.

The use of the RTDP methodology and the inclusion of Turkey Point specific instrument uncertainties have resulted in revisions to the values associated with these trip function.

These revised setpoints in the proposed TS were used in the accident analysis with acceptable results which are documented in WCAP-14276.

The reduced RCS loop flow accounts for an analyied increase in the percentage of steam generator tubes plugged (20 percent).

The effects of the reduced RCS flow have been factored in the revised core safety limits.

The reduced RCS flow has been assumed in the accident analysis with acceptable results which are documented in. WCAP-14276.

The steam generator level setpoints are revised using the Turkey Point specific instrument uncertainties in accordance with the NRC approved setpoint methodology of WCAP-12745.

The revised setpoints have been used in the loss of normal feedwater transient with acceptable results which are. documented in WCAP-14276.

The licensee indicated that the new setpoints were established using the instrument setpoint methodology identified in WCAP-12745 Revision 1,

"Westinghouse Setpoint Methodology for Protection System -- Turkey Point Units 3 5 4," dated December 1995.

In August 1991, the staff, had previously reviewed and approved the setpoint methodo'gy in Revision 0 of WCAP-12745 for use at Turkey Point Units 3 5 4.

Therefore, the staff asked the licensee to identify the changes in WCAP-12745 between Revision 0 and Revision l.

In

response, by letter dated June ll, 1996, the licensee stated that the instrument setpoint methodology is defined in Revision 0 and that Revision 1

documents the calculations conducted based on use of the methodology.

30 Similarly, for determining the OPBT and OTBT setpoints, the licensee used the methodology documented in WCAP-13719 Revision 1,

"Westinghouse Revised Design Procedure Instrument Uncertainty Methodology Florida Power 8 Light'ompany, Turkey Point Units 3 5 4," dated January

1995, and the associated calculations are documented in Revision 2 of WCAP-13719, dated September 1995.

The staff has previously reviewed and approved the setpoint methodology documented in Revision 1 of WCAP-13719.

The licensee stated that the proposed setpoint changes a} e intended to maintain the existing margins between operating conditions and the reactor trip setpoints.

Thus, these new setpoints do not significantly increase the likelihood of a false trip nor failure to trip (actuate the'p'rotection system) upon demand.

Therefore, the existing licensing basis is not affected by the TS setpoint changes.

Based on this the staff finds the proposed setpoint

'hanges acceptable.

6.4 TS Table 3.3-3 Plant specific calculations resulted in changes to various engineered safety features values of TS Table 3.3-3, "Engineered Safety Features Actuation System Instrumentation Trip Setpoints" Functional Unit 1, Safety Injection, Functional 'Unit 4, Steamline Isolation, and Functional Unit 6, Auxiliary Feedwater.

The licensee has modified the setpoints associated with safety injection, steamline isolation, and auxiliary feedwater actuation using the methodology of WCAP-12745.

The revised setpoints are used in the transient and accident analysis with acceptable results which are documented in WCAP-14276.

As stated in section 6.3, the existing licensing basis is not affected by the TS setpoint changes.

Based on this the staff finds the proposed setpoint changes acceptable.

6.5 TS 3.2.5 "DNB Parameters" and Associated BASES The departure from nucleate boiling (DNB) parameters were modified to reflect the plant specific instrument uncertainties associated with the uprate.

The revised values of T (581.2'F) and pressurizer pressure (2200 psig) correspond to analytical limits of 583.2'F and 2175 psig with allowance for measurement uncertainty.

The measured RCS flow value of 264,000 gpm corresponds to an analytical limit of 255,000 gpm (85,000 gpm per loop), which assumes a steam generator tube plugging level of 20 percent and includes a 3.5 percent calorimetric measurement uncertainty.

These values are consistent with the values used in the safety analyses, which gave acceptable

results, and their effects have been included in the revised core thermal limits of TS Figure 2. 1-1.

The changes are, therefore, acceptable.

6.6 TS BAS S

Pa e

B 2-7 Reactor Coolant Pum Breaker Position Tri This section was changed to indicate that no credit was taken in the accident analyses for operation of these trips.

The underfrequency signal does not directly result in a reactor trip, but rather it trips the RCP breakers which

31 in turn trip the reactor.

The staff agrees with the licensee proposed change which makes its TS more accurate.

6.7 TS d

.6 The licensee proposed changes to TS 3.7. 1.3, "Condensate Storage Tank" and Associated

BASES, and TS 3.7.1.6, "Standby Steam Generator Feedwater System" and Associated BASES, to reflect the required water volumes for the uprated condition.

These changes are acceptable to the staff as discussed in section 4.9.

6.8 TS

.5.

" mer enc Core Coo 'in S stem

" and Associated BASES The licensee proposed a reduction in the safety injection pump discharge head in surveillance tests.

The changes are from 1126 psid to 1083 psid for normal alignment for Unit 4 SI pumps aligned to Unit 3

RWST, and from 1156 psid to 1113 psid for Unit 3 SI pumps aligned to Unit 4 RWST.

The reduced pump discharge heads have been incorporated in the safety analyses with acceptable results which are documented in WCAP-14276. ~ The staff finds the proposed changes acceptable since the safety analyses meet the acceptance criteria.

FPL proposed changes to TS 3.4.2. 1 - "Safety Valves,"

TS 3.4.2.2 - "Safety Valves",

TS Table 3.7 "Steam Line Safety Valves Per Loop," and Associated BASES for TS 3/4.4.2 and 3/4.7. 1.1.

The licensee proposed changes to increase the pressurizer safety valve tolerances from

+/-1 percent to +2 percent,-3 percent and increase the main steam safety valve tolerances from +/-1 percent to +/-3 percent and add the footnote "All valves tested must have 'as-left'ift setpoints that are within 1 percent of the lift setting value."

The proposed safety valve tolerances are assumed in the transient and accident analyses with acceptable results which are documented in WCAP-14276.

The requirement of making "as-left" -lift setpoints within 1 percent of the lift setting value following testing would ensure that the results of any transient and accident would be bounded by safety analyses.

The licensee indicated that peak pressure remains below the ASIDE allowable of 110 percent of design pressure and that valve operability is not affected by the proposed change.

The staff finds the proposed changes acceptable since the indicated tolerances have been assumed in the analyses with acceptable results and peak pressure remains below the allowable pressure.

6. 10 TS Table 3.7 "Steam Line Safet Valves Per Loo "

The licensee proposed changing the maximum allowable power level with inoperable main steam line safety valves (HSSV) to reflect the revised power level.

Since the maXimum allowable power range neutron flux high setpoint is based on the nominal Nuclear Steam Supply System (NSSS) power rating of the plant, the licensee has performed a reanalysis to establish the revised values consistent with uprated power level.

The licensee used the method consistent

~P f

~ ~

~ ~

s ~

~

~ ~

. 32 with the current licensing bases to develop the revised values.

The staff finds the proposed changes acceptable.

6. 11 e t C

dow ves FPL proposed changes to TS Figure 3.4 "RCS Meatup Limitations (60 'F/Hr)",

TS Figure 3;4 "RCS Meatup Limitations (100 F/Hr)", and TS Figure 3.4 "RCS Cooldown Limitations (100'F/Hr)."

The licensee proposed changing the applicability of the curves from up to 20 effective full power years (EFPY) to 19 EFPY due to increased 'fluence projections on the vessel for the uprated power level.

The staff found this acceptable, as discussed in section 4. 1 of this SE.

6. 12 e

e Conta' t

C

'n S stem" and Assoc at d

BASES The licensee proposed revising TS to require that two emergency containment cooling units start automatically on a safety injection (SI) signal since analysis has shown that auto-start of all three units on an SI signal is not required.

The staff finds this acceptable, as discussed in Section 3.4.

6. 13 TS 4.7 5c "Control.Ro'om Emer enc Ventilation S stem" The licensee proposed revising the methyl iodide removal efficiency from "90 percent" to "99 percent" to provide consistency between testing efficiency and analysis assumptions for post-accident control room doses.

This is acceptable, as discussed in section 5,8 of this SE.

6.14 S 3.

- "Heat Flux Hot Channel Factor" FPL proposed relocating the heat flux hot channel

factor, F~,

and the nuclear enthalpy rise hot channel factor, F~, to the Turkey Point Core Operating Limits Report (COLR).

The TS will continue to require operation within the COLR parameters and appropriate actions are incorporated if the Fz or F~

limits are exceeded.

The determination of the F, and F" limits we'll be performed using NRC-approved methodology as defined in TS 6.9. 1.7.

Therefore, the staff finds the proposed relocation to the COLR acceptable.

6. 15 TS 6.9 "Co 0 eratin Limits Re ort" COLR The licensee proposed revising TS to (1) add the appropriate wording to reflect the inclusion of F,(Z) and F, in the COLR, (2) add the following statement - "4.

Nuclear Enthalpy Rise Hot Channel Factor for Specification 3/4.2.3" and (3) update references to be consistent with the current analyses.

In addition to the relocation of F, and F~> to the COLR, updated references to the Mestinghouse ECCS evaluation model using the BASH code have been included in the COLR list of analytical methods used to determine F, and F,".

This is acceptable since these references have been approved by the NRC.

P 1

33 6.16 TS S

4

.4 The licensee proposed a change to correct the abbreviation for "Dose Conversion Factor" to read "DCF" to ensure consistency within the TS.

This change is editorial, has no effect on the technical

content, and is therefore acceptable to the staff.

6.17 TS B

S S

a e

B 3 4 2-4 The licensee proposed deleting the reference to steam generator plugging limit of 5 percent to support anticipated future requests for higher plugging limits.

The current limit remains at 5 percent.

The licensee stated that the analysis in WCAP-14276 assumed up to 20 percent steam generator tube plugging level for the Small Break LOCA and non-LOCA analyses, while the LBLOCA is currently analyzed assuming a 5 percent steam generator tube plugging level.

After the NRC approval of the Westinghouse Best Estimate LBLOCA methodology.

(BELOCA), the licensee intends to reanalyze the LBLOCA event using BELOCA methodology and assuming' 20 percent tube plugging level.

The proposed change would avoid future inconsistency in the TS bases.

Since TS bases are used as a matter of reference and the 5 percent value will soon be invalidated and because the TS bases are not enforceable, the staff finds the licensee proposed change acceptable.

7.0 STAT CONSU TATION In accordance with its stated policy, on September 12, 1996 the NRC-staff consulted with the Florida State official, Mr. Harland Keaton of the State Office of Radiation Control, regarding the environmental impact of the proposed action.

The State official had no comments.

8.0 ENVIRONMENTAL CONSIDERATION

Pursuant to 10 CFR 51.21, 51.32, and 51.35, an environmental assessment and finding of no significant impact was published in the Federal

~Re ister on September 18, 1996 (61 FR 49176).

In this finding, the Commission determined that issuance of these amendments would not have a significant effect on the quality of the human environment.

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

Principal Contributors:

C. Liang, L. Kopp, J. Minns, R. Goel, D.

Shum, C.

Wu, J. Medoff, H. Garg

Attachment:

Tables

2. 1-1 through 2.4-2 Date: September 26,. 1996

Table 2.1-1 Change in RT>>

Values of Limiting Weld Material SA-1101 in the Turkey Point Unit 3 Reactor Pressure Vessel at End-of-License RT>><<

IO Neut.

Fluence Chemistry (Unirr~ah.)

Fluence Factor Factor

'F E19 n cm

'F Margin

('F)

RTpp

('

10 2.64 10 2.74 1.260 1.268 180 180 226.8 228.2 56 56 293 2952'3

~Foo n2~>~~

1.

Values Under Current Haxiaam Licensed Power Levels 2.

Values Under Proposed Uprated Pouer Levels 3.

Value uas conservatively rounded up from 294.2'F Table 2.1-2 Change in RT>>> Values of Limiting Weld Material SA-1101 in the Turkey Point Unit 4 Reactor Pressure Vessel at End-of-License RTwv<U ID Neut.

(Unirrad.)

Fluence

'F)

E19 n/cm Fluence Factor Chemistry Factor F)

MT,

Margin

(

~J (F)

RTpp s

('

10 2.53 10 2.68 1.249'.2632 180 180 224.8 227.32

'56 56 291 294

'ee tInntet Velues Under current nesistn Ltestued Fever Levels 2.

Values Under Proposed Uprated Pater Levels 3.

Value uas conservatively rounded up from 293.3'F

Table 2.3-1.

Turke Point Reactor Vessel Material Surveillance Pro ram Data Unit No.

Turke Pt.

3 TuT ke Pt.

3 Turkey Pt.

4 Capsule Id.

Ca sule T

Ca sule V

Capsule T

Fluence (n/cm )

5.68 x 10'.229 x 10'9 6.05 x 10'easured bRTPT

(

)

155 180 225 Table 2.4-1 Adjusted Reference Temperatures at the 1/4T RPV Location:

Turkey Point Units 3 and 4

Cheaistry 1/4T Nethod Factor Fluence ())

('F)

(E19 n/cm )

Surv.

Cap.

200.2 1.26 Data 11RT~F IRT~F 214.5 10 ART Nargin 9 1/4T

( F)

('F) 28 252.5 Table 1

180 1.26 191.6 10 56 257.6 3.

/g.

Table 2.4-2 Adjusted Reference Temperatures at the 3/4T RPV Location:

Turkey Point Units 3 and 4

~yfeNeI 8$

1.

Surveillance Capsule Data and Part 2 of Regulatory Position 2.1 in RG 1.99, Revision 2 used to establish chemistry factors and margins values used in calculation of adjusted reference temperatures.

2.

Table 1 and Regulatory Position 1.1 in RG 1.99, Revision 2 used to establish chemistry factors and margins values used in calculation of adjusted reference temperatures.

hRT aa (Chemistry factor) (f '

~

)

Adjusted Reference Temperature (ART) ~

Unirradiated Value (iRT,) + Shift (hRT,) + Hargin ("H")

Chemistry 3/4T Rethod Factor Fluence ([)

BtT 1

1RT,

('F)

(E19 n/cm )

(urv.

Cap.

200.2 0.48 160.0 10 Data'argin('F) 28 ART 9 3/4T

('F) 200.4 Table 1

180 0.48 143.9 10 56 209.9 Feat not os 1.

Rurvelllaneo capsule Date ang part 2 of Regulatory position 2.1 ln RG 1.99, Revision 2 used to establish chemistry factors and margins values used in calculation of adjusted reference tenperatul es ~

2.

Table 1 and Regulatory Position 1.1 in RG 1.99, Revision 2 used to establish chemistry factors and margins values used in calculation of adjusted reference temperatures.

3.

hRT (Chemistry Factor)n(F"'" '")

/g.

Adjusted Reference Temperature (ARl') n Unirradiated Value (IRT ;) + Shift (hRT,) + Hargin

4 0