ML17227A786

Proposed Tech Specs for Reduction in RCS Design Flowrate from Current Value of 370,000 Gpm to 355,000 Gpm

ML17227A786
Person / Time
Site:	Saint Lucie
Issue date:	03/19/1993
From:	FLORIDA POWER & LIGHT CO.
To:
Shared Package
ML17227A785	List: ML17227A784 ML17227A786 ML17227A787 ML17227A788 ML17227A789
References
NUDOCS 9303310052
Download: ML17227A786 (41)

Text

St. Lucie Unit 1 Docket No. 50-335 Proposed License Amendment St. Lucie Unit 1 Reduction of Reactor Coolant S stem Desi n Flow ATTACHMENT 1 ST LUCIE UNIT 1 MARKED-UP TECHNICAL SPECIFICATION PAGES Page 2-2 Revised Figure 2.1-1 Page 2-4 Page B 2-1 Page B 2-2 Revised Figure B2.1-1 Page 3/4 2-14 P,'DR 93033i0052 9303i9 PDR ,ADOCK 05000335

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) UNACCEPTABLE OPERATION Q 1 Ce n 580 REACT OPERA'I'ION 1.1MITED TO I.ESS U CCL'PTA DLE P~ I TIIAH 58 MAIN STEAL DY AC'I'UAT10N OF Till:.

IHE SAFETY VALVES.. OPERATION 560 VESSEL FLOW 1. S MEASUREMEN'f I ~ .9NCERTAIHTIES 0,000 GPM FOR PRE-CLAD COI.IMPS .

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)c 460 0 0.20 0.40 0.60 0.80 1.00 1.20 1.40 '.6 1.80 FRACTION OF RATED TIIERMAL POWER Figure 2 l-l REACTOR CORE TIIERMAL MARGIN SAFETY LIMIT FOUR REACTOR COOLING PUMPS OPERA'fINO

600 UNACCEPTABLE OPERATION 580 UNACCEPTABLE OPERATION I

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REACTOR OPERATION LIMITED TO LESS 540 THAN 580 oF BY ACTUATION OF THE MAIN STEAM LINE SAFETY VALVES (3 I bl VESSEL FLOW LESS MEASUREMENT I UNCERTAINTIES 355,000 GPM I I

Cl 520 I LIMITS CONTAIN NO ALLOWANCE FOR I O I O INSTRUMENT ERROR OR FLUCTUATIONS I I

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2400 I 0 460 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 FRACTION OF RATED THERMAL POWER FIGURE 2.1 1 REACTOR CORE THERMAL MARGIN SAFETY LIMIT FOUR REACTOR COOLING PUMPS OPERATING-

TABLE 2.2-1 REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES

1. Manual Reactor Trip Not Applicable Not Applicable

2. Power Level - High (1)

Four Reactor Coolant Pumps < 9.61% above THERMAL POWER, < 9.61$ above THERMAL POWER, and Operating with a minimum setpoint of 15K a minimum setpoint of 15K of RA~

of RATED THERMAL POWER, and a THERMAL POWER and a maximum of maximum of < 107. OX of RATED < 107.0X of RATED THERMAL POWER.

THERMAL POWER.

3. Reactor Coolant Flow - Low (1)

Four Reactor Coolant Pumps > 95$ of design reactor coolant > 95K of design reactor coolant Operating flow with 4 pumps operating* flow with 4 pumps operating*

4. Pressurizer Pressure - High < 2400 psia < 2400 psia

5. Containment Pressure - High <33pslg 3. 3 ps 1g

6. Steam Generator Pressure - Low (2) > 600 psia > 600 psia

7. Steam Generator Water Level -Low > 20.54 Water Level - each > 19.5X Water Level - each steam generator steam generator

8. Local Power Density - High (3) Trip setpoint adjusted to not Trip set point adjusted to not exceed the limit lines of exceed the limit lines of Figures 2.2-1 and 2.2-2 Figures 2.2-1 and 2.2-2.

• Design reactor coolant flow with 4 pumps operating is 9~680 gpm.

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Sierne ns Power Corporakt'on (SPC) 2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel cladding and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel is prevented by maintaining the steady state peak linear heat rate below the level at which centerline fuel melting will occur. 'Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reacto~ Coolant Temperature and Pressure have been related to ONB throu h the XNB correlation. The XNB DNB correlation i as een deve oped to pre sct the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local ONB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to ONB.

.The minimum value of the DNBR during steady state operation, transients, and anticipated transients is limited to 1.22 using normal'perational the XNB ONBR correlation. This value corresponds to a 95 percent probability at a 95 percent confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.

The curves of Figure 2.1-1 show the loci of points of THERMAL'OWER, Reactor Coolant System pressure and maximum cold leg temperature with four Reactor Coolant Pumps operating for which the minimum ONBR is no less than the ONBR limit for the 0 shown in Figure 8 2.1-1. The limits in Figure . -1 were calculated for reactor coolant inlet temperatures less than or equal to 580'F. The dashed line at 580'F coolant inlet temperature is not a safety limit; however, operation above 580'F is not possible because of the actuation of the main steam line safety valves which limit the maximum value of reactor inlet temperature. Reactor operation at THERMAL POWER levels higher than 112" of RATED THERMAL POWER is prohibited by the high power level trip setpoint specified in Table 2.1-1. The area of safe operation is below and to the left of these lines.

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ST. LUCIE UNIT 1 8 2-1 Amendment No. 27, 8B,

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TABLE 3.2-1 DNB MARGIN L I I I ITS Four Reactor Coolant Pumps Parameter ~0eratin Cold Leg Temperature ~ 549 Pressurizer Pressure . > 2225, psia*

Reactor Coolant Flow Rate ~ gPm RGl'LAcp AXIAL SHAPE INDEX Figure 3.2-4 355,00o

Limit not applicable during either a THERMAL POWER ramp increase in excess of 5X of RATED THERMAL POWER or a THERMAL POWER step increase of greater than 10K of RATED THERMAL POWER.

ST. LUCIE - UNIT 1 '3/4 2-14 Amendment No. pe,g ~

8

St. Lucie Unit 1 Docket No. 50-335 Proposed License Amendment St. Lucie Unit 1 Reduction of Reactor Coolant S stem Desi n Flow ATTACHMENT 2 SAFETY ANALYSIS 1 of 20

St. Lucie Unit 1 Docket. No. 50-335 Proposed License Amendment St. Lucie Unit 1 Reduction of Reactor Coolant S stem Desi n Flow TABLE OF CONTENTS DESCRIPTION PAGE

1. Introduction

2. Description of Proposed Changes

3. Impact on Plant Safety Analysis

4. Transient Events Requiring Re-Analysis

5. Transient Events Not Requiring Re-Analysis

6. Impact of the License Amendment on 16 Other Analyses

7. Conclusions 19

8. References 20 2 of 20

St. Lucie Unit 1 Docket No. 50-335 Proposed License Amendment St. Lucie Unit 1 Reduction of Reactor Coolant S stem Desi n Flow

1. Introduction Florida Power and Light. Company (FPL) proposes to modify St.

Lucie Unit 1 Technical Specification (TS) Figure 2.1-1, and Tables 2.2-1 and 3.2-1 by reducing the design Reactor Coolant System (RCS) flowrate from 370,000 gpm to 355,000 gpm. The associated TS Bases pages B2-1 and B2-2 are also revised as part of this proposal. The reduced flowrate will accommodate an increase in the number of plugged steam generator (SG) tubes beyond the 15% value which is used in existing safety analysis assumptions and is commensurate with up to 254 (average) of all SG tubes plugged.

The analysis performed by FPL to evaluate the impact of the proposed change has demonstrated that this flow reduction will not have an adverse impact on safety analysis conclusions or full power plant operation. Details of this analytical effort are discussed in the following sections.

2 ~ Descri tion of Pro osed Chan es 2-1 Technical S ecification 2.1.1 Reactor Core Figure 2.1-1 is revised to include the effects of an RCS flowrate of 355,000 gpm with four reactor coolant pumps operating. Consistent with the existing specification, implementation of the safety limits shown by the thermal limit lines of Figure 2.1-1 requires allowances for instrument error and other uncertainties. Such considerations are included in the generation of the Thermal Margin/Low Pressure (TM/LP) Trip and Variable High Power Trip (VHPT) protective system functions and in the pressure values selected for actuation of the Main Steam Safety Valves.

Updating Figure 2.1-1 and its associated bases pages is required to maintain configuration documentation consistent with safety analysis assumptions. The new limits on plant operating space shown in the updated figure were obtained utilizing Siemens Power Corporation (SPC) methodology which includes the most limiting axial power distribution shown in revised Figure B2.1-1.

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St. Lucie Unit 1 Docket No. 50-335 Proposed License Amendment St. Lucie Unit 1 Reduction of Reactor Coolant S stem Desi n Flow 2-2 Technical S ecification 2.2.1 Reactor Tri Set pints The footnote at the bottom of Table 2.2-1 is revised to change the design reactor coolant flow with 4 pumps operating from 370,000 gpm to 355,000 gpm. The proposed flowrate is consistent with the assumptions used in this safety analysis.

2-3 Technical S ecification 3.2.5 DNB Parameters Table 3.2-1 of this Limiting Condition for Operation (LCO) is revised to reduce the reactor coolant system total flow rate from > 370,000 gpm to > 355,000 gpm. The proposed flowrate is consistent with the assumptions used in this safety analysis.

Im act on Plant Safet Anal sis The changes proposed by this license amendment affect the plant safety analysis in the following manner.

A reduction in reactor coolant flow rate can impact the calculated Departure from Nucleate Boiling Ratio (DNBR) for certain transients. This parameter is a direct indication of available thermal margin. A reduction in the calculated minimum DNBR indicates that thermal margin for the corresponding transient has been reduced.

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~ The removal of additional SG tubes from service (plugging) reduces the primary to secondary heat transfer area in the steam generators. This effect is most relevant to transients involving a sudden reduction in the heat removal capability of the secondary plant.

A reduction in RCS flowrate results in a corresponding increase in RCS average coolant temperature (Tave). An increase in Tave can impact both DNBR-related and loss of primary inventory types of events.

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St. Lucie Unit 1 Docket. No. 50-335 Proposed License Amendment.

St. Lucie Unit 1 Reduction of Reactor Coolant S stem Desi n Flow 4 ~ Transient Events Re irin Regnal sis A review of events described in the St. Lucie Unit 1 Updated Final Safety Analysis Report (UFSAR) was performed to assess the impact of an increase in average steam generator tube plugging to 254, with an asymmetry in plugging .of + 7~o and a reduced RCS flowrate of 355,000 gpm. Based on this review, FPL determined that the following events required reanalysis:

Loss of External Load 4-Pump Loss of Reactor Coolant Flow Seized RCP Rotor Dropped CEA Large Break LOCA Small Break LOCA In addition to the above listed transients, this evaluation verified the adequacy of the Reactor Protection System (RPS)

Thermal Margin/Low Pressure (TM/LP) trip function as described in TS Figures 2.2-3 and 2.2-4, and demonstrated that the existing Variable High Power Trip (VHPT) and RCS Low Flow Trip functions continue to provide adequate protection after accounting for the proposed changes.

Event. reanalyses were performed by the fuel vendor (SPC) after the determination was made that: (a) these events would be affected most significantly by the proposed changes, and (b) the impact of the proposed changes on these events must be quantified to ensure that available margin to the limiting acceptance criteria exists. References 3, 4,'and 5 describe the SPC analyses. Results of these reanalyses are discussed in the following sub-sections of part 4 to this attachment.

4-1 Decrease in Heat Removal b the Secondar S stem 4-1.1 Loss of External Load LOEL The LOEL transient is the limiting event in the "Decrease in Heat Removal by the Secondary System Class" because the most rapid reduction of secondary heat removal is achieved via closure of the turbine stop valves. This transient bounds (is more severe than) other transients in this category such as Turbine Trip, Loss of Condenser 5 of 20

St. Lucie Unit 1 Docket No. 50-335 Proposed License Amendment St. Lucie Unit 1 Reduction of Reactor Coolant S stem Desi n Flow Vacuum, and Main Steam Isolation Valve Closure. The LOEL event is typically analyzed for both RCS pressure and minimum DNBR (MDNBR). However, the challenge to DNBR acceptance criteria for this event is shown by the existing UFSAR analyses to be bounded by (less severe than) the RCS Loss of Flow transient. Since increasing steam generator tube plugging (SGTP) and reducing initial RCS flow will tend to reduce the calculated MDNBR for DNB-related events in a similar manner, FPL determined that the LOEL event will remain bounded (relative to MDNBR) by the Loss of Flow transient. Therefore, the primary focus of the LOEL reanalyses is on the impact, of the proposed changes on the calculated maximum RCS pressure.

Important assumptions used to maximize RCS pressure during this transient. included: (a) positive Moderator Temperature Coefficient consistent with the maximum allowed by the existing LCO, (b) reduced SG heat transfer area consistent with the proposed tube plugging value, (c) inoperable steam dump and bypass system, (d) event initiation by closure of fast. acting (turbine stop) valve, and (e) reactor trip by turbine trip disabled.

The increased SGTP has the effect of increasing the primary coolant insurge into the pressurizer. This is. a result of the reduced primary-to-secondary heat transfer area which causes a lower initial secondary side pressure. The lower secondary pressure delays Main Steam Safety Valve (MSSV) actuation and leads to a greater expansion of the RCS fluid. Details of this reanalysis and applied methodologies are contained in Reference 3.

Results of the reanalysis show the calculated peak primary system pressure to be 2701 psia. This is below the limiting acceptance criteria of 1104 of design pressure (2750 psia). Secondary system pressure was also determined to be substantially less than the 1100 psia secondary side acceptance criteria. Therefore, it. is concluded that the proposed SGTP,level and associated reduction in RCS flow are acceptable relative to the decrease in heat removal by secondary system class of transients.

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St. Lucie Unit 1 Docket No. 50-335 Proposed License Amendment St. Lucie Unit 1 Reduction of Reactor Coolant S stem Desi n Flow 4-2 Decrease in Reactor Coolant S stem Flow Events within this category of transients are initiated by a malfunction of the Reactor Coolant Pumps (RCP) with the resultant decrease in coolant flow causing a degradation in the calculated MDNBR. Two events in this category were determined to be impacted by a reduction in design RCS flow: 4 Pump Loss of Reactor Coolant Flow and Seized RCP Rotor.

4-2.1 Loss of Reactor Coolant Flow 4 Pum Coastdown The Loss of Flow (LOF) event is the limiting Anticipated Operational Occurrence (AOO) with respect to MDNBR for this category of transients. The evaluation of this event included performance of a verification analysis which ensures that the existing DNB-LCO setpoints (TS Figure 3.2-4), in conjunction with the RPS Low Flow Trip, prevents the MDNBR limit of 1.22 from being violated.

The validity of the DNB-LCO for the LOF transient was verified using the same NRC approved methodology that is applied to reanalysis of this event for each fuel cycle.

System calculations were not performed. Rather, simulations of the LOF transient were conducted with initial conditions modified to include the proposed RCS design flow and allowable core power as a function of Axial Shape Index (ASI). Axial power profiles and scram curves for fuel cycle 11 were included in the analysis.

, Details of this reanalysis and applied methodologies are contained in Reference 3.

The existing Cycle 11 reload analysis for this event shows a conservative margin of more than 12~ to the limiting acceptance criterion of MDNBR 1.22. Results of the reanalysis show a reduction in this cycle 11 margin of less than 0.5>. Therefore, FPL has concluded that the 4 Pump LOF event initiated with an initial RCS flow of 355,000 gpm and from within the DNB-LCO constraints for allowable core power as a function of ASI will not result in violation of the Specified Acceptable Fuel Design Limit (SAFDL) for DNBR.

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St. Lucie Unit 1 Docket No. 50-335 Proposed License Amendment St. Lucie Unit 1 Reduction of Reactor Coolant S stem Desi n Flow 4-2.2 Seized RCP Rotor The seized rotor event is assumed to be initiated by an instantaneous seizure of a reactor coolant pump shaft.

The event is terminated by the RPS Low Flow Trip.

Because of the very low probability associated with this event, a limited number of fuel rod failures are permitted to occur.

The transient was reanalyzed to quantify the number of fuel rods expected to fail as a result of the lower proposed value of initial RCS flow. A new system calculation was not performed. Rather, a set of conservative boundary conditions was assumed utilizing the original system calculation as a basis to perform the reanalysis. Details of this reanalysis and applied methodologies are contained in Reference 3.

The percentage of fuel rods predicted to experience DNB is 14 for a seized rotor event that occurs with an initial RCS flow of 355,000 gpm. This failed fuel fraction is bounded by (less severe than) postulated values used in the existing analyses of record for assessment of radiological consequences at the site boundary and demonstrated compliance with 10CFR100 criteria. Therefore, FPL has concluded that the impact of the proposed SGTP level and RCS design flowrate on the fraction of failed fuel resulting from the Seized RCP Rotor accident is acceptable.

4-3 Reactivit and Power Distribution Anomalies This event category contains several types of events that were considered when evaluating the impact of reduced RCS flow and increased steam generator tube plugging. It was determined that only the transient from a dropped Control Element Assembly (CEA) required specific reanalysis.

Brief discussions of Uncontrolled CEA Withdrawal and the CEA Ejection Accident are presented in section 5 of this attachment.

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St. Lucie Unit 1 Docket No. 50-335 Proposed License Amendment St. Lucie Unit 1 Reduction of Reactor Coolant S stem Desi n Flow The dropped CEA is an AOO that creates changes in the radial core power distribution which, in turn, impact the DNBR. The magnitude of these changes depends upon the characteristics of cycle to cycle fuel loading patterns and, similar to the LOF transient discussed in section 4-2.1, the Dropped CEA event is included as part of the DNB-LCO setpoint verification for each fuel cycle.

Variations of 1-2% in the calculated MDNBR are typically observed between fuel cycles for this event.

The validity of the DNB-LCO for the Dropped CEA transient was verified using the same NRC approved methodology that is applied to reanalysis of this event for each fuel cycle. Simulations of the event were conducted with initial conditions adjusted to include the proposed RCS design flow and allowable values of core power as a function of ASI. Details of this reanalysis and applied methodologies are contained in Reference 3.

Results of the reanalysis show a conservative margin of more than 13< to the limiting acceptance criterion of MDNBR 1.22. Therefore, FPL has concluded that the occurrence of a dropped CEA event initiated with an initial RCS flow of 355,000 gpm and from within the DNB-LCO constraints for allowable core power as a function of ASI will not result in violation of the SAFDL for DNBR.

4-4 Decrease in Reactor Coolant Inventor Events 4-4.1 Lar e Break LOCA LBLOCA The LBLOCA was reanalyzed to evaluate the impact of the proposed changes on Emergency Core Cooling System (ECCS) acceptance criteria. System and hot channel blowdown calculations confirmed that the limiting event is the'.8 Double Ended Cold Leg Guillotine (DECLG) break which is consistent with the existing LBLOCA analysis. The break was conservatively assumed to occur in an RCS cold leg having the maximum SGTP proposed by this amendment (32%) .

This location results in more steam binding during the 9 of 20

St. Lucie Unit 1 Docket No. 50-335 Proposed License Amendment St. Lucie Unit 1 Reduction of Reactor Coolant S stem Desi n Flow reflood stage of the transient and maximizes the fuel element peak cladding temperature (PCT).

Consistent, with previous analyses, loss of offsite electrical power and a single failure of one Low Pressure Safety Injection (LPSI) pump were assumed to occur.

Conservative values of Linear Heat Rate (LHR) and core power radial peaking factors (F) were also assumed for the reanalysis. Details of the LBLOCA reanalysis and applied methodologies are contained in Reference 4.

The maximum PCT achieved in the several LBLOCA cases that were analyzed is 1912'F, using Middle of Cycle (MOC) fuel stored energy and an End of Cycle (EOC) axial power shape peaked at 854 of core height. The calculated maximum local cladding oxidation is less than 34 and core wide oxidation is much less than 14. The reanalysis demonstrates that acceptance criteria of 10CFR50.46(b) are satisfied for the limiting LBLOCA. In addition, the results of this reanalysis are bounded by (less severe than) the postulated LBLOCA used in the existing analyses of record performed for the assessment of potential radiological conseguences and demonstrated compliance with 10CFR100 criteria.

Therefore, FPL has concluded that operation of St. Lucie Unit 1 in accordance with the proposed amendment is acceptable with regard to the cooling performance of the existing ECCS and the potential radiological consequences associated with a LBLOCA.

4-4.2 Small Break LOCA SBLOCA The Small Break LOCA was reanalyzed for St. Lucie Unit 1 to evaluate the impact of up to 254 (average) SGTP on the ECCS acceptance criteria. Break spectrum calculations and sensitivity calculations were performed as part of the analyses. The sensitivity calculations considered a delayed RCP tripping scheme that is consistent with plant Emergency Operating Procedures (EOP).

The break is conservatively assumed to occur in the cold leg discharge of a RCP and conservative values for LHR and F were used in the reanalyses. A limiting break 10 of 20

St. Lucie Unit 1 Docket No. 50-335 Proposed License Amendment St. Lucie Unit 1 Reduction of Reactor Coolant S stem Desi n Flow size of 0.1 ft~ was determined from the break spectrum calculations. The SGTP asymmetry was not specifically evaluated in the reanalyses since prior SBLOCA studies indicate that asymmetric tube plugging conditions would be bounded by the results from symmetric SGTP conditions.

Consistent with previous analyses, single failure criterion was satisfied by assuming loss of one emergency diesel generator. Details of the SBLOCA reanalyses and applied methodologies are contained in Reference 5.

The maximum PCT achieved in the several SBLOCA cases that were analyzed is 1672'F. The calculated maximum local cladding oxidation is less than 24 with core wide oxidation much less than 1:. The reanalyses demonstrate that acceptance criteria of 10CFR50.46(b) are satisfied for the limiting SBLOCA.

The SBLOCA reanalysis was performed utilizing updated vendor methodology described in Reference 1. A topical report describing this revised EXEM PWR Small Break Model was submitted to NRC via SPC letter dated May 1, 1992, for review. Contingent upon approval of this methodology, FPL considers that operation of St. Lucie Unit 1 in accordance with the proposed amendment is acceptable with regard to the cooling performance of the existing ECCS in response to the SBLOCA.

5 Transient Events Not Re irin Regnal sis In addition to the reanalyzed transients discussed in the previous section, a number of events were evaluated and determined not to require reanalysis. A discussion of the expected impact of the proposed changes on these events follows.

5-1 Uncontrolled CEA Withdrawal The uncontrolled CEA withdrawal from low power and the CEA withdrawal initiated from high power conditions are events that are analyzed for compliance with MDNBR criteria. Existing analyses demonstrate that these ll of 20

St. Lucie Unit 1 Docket No. 50-335 Proposed License Amendment St. Lucie Unit 1 Reduction of Reactor Coolant S stem Desi n Flow events are bounded by the 4 pump LOF transient. Since the proposed reduction in RCS flow does not significantly change the relative MDNBR between DNB-related events, CEA withdrawal transients will remain bounded by the LOF event that is discussed in section 4-2.1. Therefore, FPL has concluded that Uncontrolled CEA withdrawal events will not result in violation of the DNB SAFDL when initiated from within the constraints of the DNB-LCO.

5-2 Boron Dilution Event A small change in RCS fluid volume is associated with increased SGTP. The reduction in fluid volume, in turn, can impact the amount, of injection required for a boron dilution event. A boron dilution event is possible during any mode of plant operations.

The existing TM/LP trip, the Variable High Power Trip (VHPT), and the Local Power Density (LPD) Trip Limiting Safety System Settings (LSSS) provide protection against boron dilution events initiated, at power and ensure that SAFDL's are not, exceeded.

For the lower operating modes, the acceptance criterion for analysis of this transient is that the time to criticality allows operator action to terminate the event. The changes proposed by this amendment will have no impact on Mode 6 results. The existing analyses show that the times calculated to lose the required shutdown margin for dilution events initiated during modes 2 through 5 were all significantly greater than the acceptance criteria. FPL has concluded that. the calculated times to lose the required shutdown margin for dilution events initiated during these operating modes, have sufficient, margin to the acceptance criteria to accommodate the impact from the small change in RCS fluid volume associated with the proposed level of SGTP.

5-3 CEA E'ection Accidents A CEA ejection accident is defined as the mechanical failure of a control rod drive mechanism pressure housing resulting in ejection of a CEA and its drive shaft. The 12 of 20

St. Lucie Unit 1 Docket No. 50-335 Proposed License Amendment St. Lucie Unit 1 Reduction of Reactor Coolant S stem Desi n Flow resultant rapid reactivity insertion and adverse core power distribution are analyzed to determine the extent of localized core damage that may occur.

Predictions of fuel failure for this event are based on fuel centerline melt criteria (deposited energy in the fuel rod) rather than DNBR criteria. Since the impact of SGTP and reduced RCS flow do not change the nuclear characteristics of the reactor core, FPL has concluded that operation of the facility in accordance with this proposed amendment will not impact, the results of existing analyses of this event with respect to core damage or offsite radiological dose consequences.

5-4 Inadvertent 0 enin of Pressurizer Pressure Relief Valves This RCS depressurization event is the limiting DNB event within the "Decrease in Reactor Coolant Inventory" category of transients. The existing analyses show this event to be bounded by the 4 pump LOF transient. Since the changes proposed by this amendment will not significantly affect the relative MDNBR between DNB-related events, FPL has concluded that the Inadvertent Opening of Pressurizer Pressure Relief Valves event will remain bounded by (less severe than) the 4 pump LOF transient discussed in section 4-2.1.

This event is one of the considerations in setting the limiting pressure bias term in the TM/LP equation. The bias term is dependent on the maximum rate of change of DNBR experienced during the event, which for this case, is directly dependent on the rate of depressurization.

Since the proposed changes do not affect the rate of depressurization in this transient, FPL has concluded that there is no impact on the existing TM/LP pressure bias.

5-5 Steam Generator Tube Ru ture SGTR Existing analyses demonstrate the potential radiological consequences of the SGTR event. These analyses show that the radiological release associated with this transient is dependent upon the break flow rate and corresponding 13 of 20

I t,

St. Lucie Unit 1 Docket No. 50-335 Proposed License Amendment St. Lucie Unit 1 Reduction of Reactor Coolant S stem Desi n Flow primary to secondary mass transfer that occurs. The maximum mass flow rate for a given RCS pressure will occur for the condition of choked-flow which is a function of the differential pressure (primary-to-secondary) across the SG tubes. The existing analysis of record examined the bounding case where break flow was assumed to be choked (maximized) throughout the event.

For the bounding case, radiological consequences were determined to be a small fraction of 10CFR100 limits.

The proposed SGTP level will result in reduced secondary side operating pressure at St. Lucie Unit 1 which, for a given RCS pressure, will enhance the conditions necessary for choked-flow to occur. However, since the analysis of record considers choked-flow conditions throughout the postulated event, that analysis will remain bounding.

Therefore, FPL has concluded that operation of the facility in accordance with this proposed amendment will not alter the system response and potential offsite dose consequences predicted for the SGTR event.

5-6 Increase in Heat Removal b the Secondar S stem Events in this category are evaluated by calculating the increase .in primary system cooling due to the particular event initiator. A discussion of the individual transient evaluations follows.

5-6.1 Excess Load Three events with different initiators are postulated in the existing analyses with the limiting sub-event being the inadvertent opening of all the steam dump and bypass control system (SBCS) valves at full power. This scenario would result in an increase of approximately 43.44 of the total steam mass flow rate. This event has previously been determined to be bounded by the 4 pump LOF event for DNB considerations and the proposed changes will not significantly impact the relative DNBR behavior between these two transients. Therefore, FPL concluded that reanalysis of this event is not required.

14 of 20

7 St. Lucie Unit 1 Docket No. 50-335 Proposed License Amendment St. Lucie Unit 1 Reduction of Reactor Coolant S stem Desi n Flow 5-6.2 Steam S stem Pi in Failures Inside & Outside Steam System Piping Failures are analyzed to ensure that any fuel failures which might occur are limited to a small percentage of the fuel in the core. These analyses are used to determine whether fuel failures would result from violation of the SAFDL for either DNBR (initiated from hot zero power conditions) or fuel centerline melt (initiated from hot full power conditions).

The proposed changes do not alter the initial core power distribution, and since the time of minimum DNBR occurs during a period of natural circulation following a loss of offsite power, the key parameter affecting fuel failures is the return to power caused by reactivity feedback following the break.

The reduced primary to secondary heat transfer rate across the steam generators and the lower initial secondary pressure expected from the changes proposed in this amendment both contribute to making this a more benign event by reducing the rate of RCS cooldown. The total energy available for release from the SG secondary side during the event is also reduced.

Therefore, FPL has concluded that a primary system cooldown initiated with the proposed SGTP level and RCS design flow and caused by a limiting steam system piping failure will be bounded by (no more severe than) the existing analyses of record. The predicted off-site radiological dose consequences will also remain unchanged from the existing analyses.

5-6.3 Inadvertent 0 enin of a Steam Generator Relief or Safet Valve The existing analysis of record for this event evaluates the potential radiological consequences from the opening of a power operated atmospheric dump valve (ADV). The radiological consequences are based on TS limits for primary to secondary leak rate which are unchanged by this proposal. At the SGTP level proposed by this amendment, inadvertent opening of an ADV will be less 15 of 20

St. Lucie Unit 1 Docket. No. 50-335 Proposed License Amendment St. Lucie Unit 1 Reduction of Reactor Coolant S stem Desi n Flow likely to occur and the radiological release less severe because of the expected lower initial secondary side pressure. Therefore, FPL has concluded that the existing analysis of record will remain bounding for this event.

6. Im act of the License Amendment on Other Anal ses While the discussions in Sections 4 and 5 focus on the Unit 1 UFSAR Chapter 15 Safety Analysis, FPL determined that several additional analyses should be addressed to complete the safety assessment. These analyses are discussed in the following sub-parts of section 6.

6-1 Plant Natural Circulation Ca abilit FPL examined the impact of the proposed level of SGTP and asymmetry on the ability to maintain adequate decay heat removal by natural circulation in the RCS. Using the FPL St. Lucie Unit 1 RETRAN Model previously reviewed by the NRC (Reference 2), the sequence of events following a coastdown of all 4 RCPs was simulated. No operator actions were credited during the 4500 second simulation.

Temperatures were controlled by automatic SBCS action with 150 gpm auxiliary feedwater flow per operable SG.

The simulation demonstrated that adequate RCS subcooled natural circulation decay heat removal will be maintained with the proposed SGTP, level. Criteria established in plant off-normal and emergency procedures for operator identification and assurance that subcooled natural circulation is occurring were verified. In addition, a simulation was performed to verify that natural circulation cooldown could be maintained with one SG isolated. Therefore, FPL has concluded that the natural circulation decay heat removal capability of the plant will not be adversely impacted by the SGTP level consistent with this proposed amendment.

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St. Lucie Unit. 1 Docket No. 50-335 Proposed License Amendment St. Lucie Unit 1 Reduction of Reactor Coolant S stem Desi n Flow 6-2 Peak Containment Pressurization Followin LBLOCA or Steam S stem Pi in Failure Analyses of record for Large Break LOCA and Steam Pipe Break inside containment were evaluated to determine the increased SGTP level and corresponding reduced RCS if flow would cause the containment design pressure to be exceeded.

For the LBLOCA event. inside containment, the reduction in primary system fluid volume available for blowdown almost completely offsets the effects from a slight increase in system energy due to the, expected higher initial RCS Tave. Conservative estimates show a slight increase in containment pressure < 0.2 psia. Since the existing analyses show a margin of more than 4 psia to the acceptance limit, FPL has concluded that there is no significant impact from this proposed change on the capability of the Unit 1 containment to withstand a LOCA.

Steam Piping Failures inside containment were also examined and FPL concluded that, after allowing for the proposed changes, no compromise of the pressure limits on containment analysis would result. Initial FPL estimates indicate that. increased tube plugging results in a small increase ( 2000 ibm) in the total secondary side mass inventory due to an expected lower operating pressure, but that the overall energy stored in the fluid (and eventually released to containment. during this event) is not increased. In addition, the lower initial secondary pressure will allow less blowdown from the intact SG prior to Main Steam Isolation Signal (MSIS) actuation.

6-3 Auxiliar Feedwater S stem AFW Hi h Ener Line Break The existing system evaluation for the AFW system identified this event in conjunction with loss of off-site power as the limiting condition for plant operators to be able to initiate auxiliary feedwater flow prior to steam generator dryout. Since increased SGTP will result in a small increase in total secondary mass, FPL has concluded that the steam generator dryout time is not adversely impacted by increased steam generator tube plugging or reduced RCS flow.

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St. Lucie Unit 1 Docket No. 50-335 Proposed License Amendment St. Lucie Unit 1 Reduction of Reactor Coolant S stem Desi n Flow 6-4 Low Tem erature Over ressure Protection LTOP Anal sis The existing LTOP analysis was evaluated to determine whether the postulated increase in SGTP would impact the consequences of starting a RCP with the plant secondary side at a higher temperature than the primary side. It was determined that plugging an increased number of SG tubes would result in a slight increase of the thermal time constant of the system. This leads to a slower rate of energy addition to the primary side for this event.

Only a change in the RCP heat output or in the initial condition primary to seconday hT could change the energy deposited in the primary system, and hence, the peak pressure. Therefore, FPL has concluded that increasing the SGTP level to 254 (average) has no adverse impact on the pressure transient caused by starting a RCP under low temperature conditions.

6-4 Over ressure Protection Anal sis In Appendix 5A of the St. Lucie Unit 1'UFSAR, an analysis documenting the adequacy of overpressure protection provided for the SG and RCS is presented. The intent of this analysis, which is based on a "worst case" loss of load event, is to demonstrate adequate capacity of the safety valves such that peak system pressures (primary and secondary) will not exceed 110% of design.

The impact of the proposed increase in SGTP and reduced RCS flow on the analysis for the limiting loss of load event was previously discussed in section 4-1.1. Since that analysis confirmed compliance with the acceptance criteria for overpressure events, it. indirectly verifies the continued validity of the overpressure protection analysis in Appendix 5A of the UFSAR. Therefore, concluded that the proposed changes do not require an it is increase in main steam safety valve capacity to satisfy the overpressurization criteria.

6-6 Im act on Steam Generator Mechanical Loads As part of our review, FPL examined the calculations of static and dynamic loading for the St. Lucie Unit 1 18 of 20

4 St. Lucie Unit 1 Docket No. 50-335 Proposed License Amendment St. Lucie Unit 1 Reduction of Reactor Coolant S stem Desi n Flow Steam Generators to determine if the proposed change will result in any adverse impact on acceptance criteria for the tube sheet and steam generator tube bundle. Through our review of existing calculations and independent analysis, FPL has concluded that substantial margin to stress limits remains available and that this change will not adversely impact Steam Generator operability.

7. Conclusion Preceding discussions in this safety evaluation demonstrate that the results of relevant analyses for St. Lucie Unit 1 satisfy established acceptance criteria for plant performance.

In addition, radiological consequences determined in the analyses of record for St. Lucie Unit 1 and which demonstrate compliance with 10CFR100 acceptance criteria remain bounding for the proposed RCS design flow and SGTP level.

Continued validity of existing RPS settings and trip functions in conjunction with existing, relevant LCO s, which are designed to provide assurance that reactor core design limits are not exceeded during facility operation has been verified (Reference 3) for the proposed change in RCS design flow.

The reactor core thermal margin safety limits illustrated in TS Figure 2.1-1 have been adjusted to account for the proposed value of RCS design flow and define the areas of acceptable operation in terms of thermal power, RCS pressure, and maximum cold leg temperature with four RCP's operating and for which the MDNBR is no less than the MDNBR limit. The MDNBR limit during steady state operation, normal operational transients, and anticipated transients remains unchanged from the existing, approved value of 1.22 and the redefined areas of acceptable operation do not result in restrictions on full-power operation of the facility.

Contingent upon NRC approval of Reference 1, FPL therefore concludes that operation of the facility in accordance with this proposed amendment is acceptable.

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1 t

A

St. Lucie Unit 1 Docket No. 50-335 Proposed License Amendment St. Lucie Unit 1 Reduction of Reactor Coolant S stem Desi n Flow

8. References XN-NF-82-49(P), "Exxon Nuclear Company Evaluation Model Revised EXEM PWR Small Break Model", Rev 1, Supplement 1, May 1992.

2. NRC Letter from Gus C. Lainas to W.F. Conway "Florida Power and Light Company Topical Report on RETRAN (TAC No. 60550) and Topical Report on PWR Physics Methodology (TAC No. 60549), April 19, 1988; Docket . Nos. 50-250 and 50-251, 50-335 and 50-389.

3. EMF-92-165, St. Lucie Unit 1 Chapter 15 Event Review and Analysis for 25> Steam Generator Tube Plugging; Siemens Power Corporation, February 9, 1993.

4 ~ EMF-92-176, St. Lucie Unit 1 Large Break LOCA/ECCS Analysis with 25% SGTP; Siemens Power Corporation, February 5, 1993.

5. EMF-92-148, St. Lucie Unit 1 Small Break LOCA Analysis; Siemens Power Corporation, February 5, 1993.

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St. Lucie Unit 1 Docket No. 50-335 Proposed License Amendment St. Lucie Unit 1 Reduction of Reactor Coolant S stem Desi n Flow ATTACHMENT 3 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION Pursuant to 10CFR50.92, a determination may be made that a proposed license amendment involves no significant hazards consideration operation of the facility in accordance with the proposed amendment if would not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. Each standard is discussed as follows:

(1) Operation of the facility in accordance with the proposed, amendment would not involve a significant increase in the probability or consetxuences of an accident previously evaluated.

The proposed amendment would permit full power operation of St.

Lucie Unit 1 with a decreased value of design reactor coolant flow.

It is postulated that this steady state RCS flow decrease would occur as a result of an increased level of steam generator tube plugging. No increase in the probability of any accident previously analyzed is likely to occur because no changes are being made to the plant's required mode of operation or to any active plant component. By their nature, steam generator tube plugs are passive components and this amendment does not change the nature or type of plugs which may be used.

Florida Power and Light Company (FPL) has submitted the results of a safety evaluation which concludes that potential radiological consequences of previously analyzed accidents remain within their established acceptance criteria when including the values of increased steam generator tube plugging or reduced RCS flow consistent with this proposed amendment. This conclusion is supported by reanalysis of relevant limiting events in several accident categories, as appropriate.

Therefore, operation of the facility in accordance with this amendment would not involve a significant increase in the probability or consequences of any accident previously evaluated.

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V St. Lucie Unit 1 Docket No. 50-335 Proposed License Amendment St. Lucie Unit 1 Reduction of Reactor Coolant S stem Desi n Flow (2) Operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated.

This amendment does not result in a change to any active plant component or in the mode of operation of the plant. The "design function of the Reactor Coolant System, Steam Generators, Emergency Core Cooling System and other systems will not change. The designed excess capacity of the steam generators can accommodate the potential reduction in primary to secondary heat transfer area caused by increased tube plugging without significantly impacting the plant's dynamic behavior. The presence of plugged steam generator tubes is not changed by this proposed amendment.

Therefore, operation of the facility in accordance with the proposed amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated.

(3) Operation of the facility in accordance with the proposed amendment would not involve a significant reduction in a margin of safety.

The impact of the proposed change on available margin to the acceptance criteria for Specified Acceptable Fuel Design Limits (SAFDL), primary and secondary overpressurization transients, 10CFR50.46(b) criteria, peak containment pressure, potential radioactive dose releases and to existing Limiting Conditions for Operation has been examined as part of the safety evaluation submitted by FPL in support of this amendment. After considering the proposed steam generator tube plugging level and resultant RCS flow, a comparison of limiting events to the acceptance criteria listed above shows that a conservative safety margin to the acceptable limits remains available. For this reason, FPL has concluded that there is no significant reduction in a margin of safety as a result of this proposed license amendment.

Based on the discussions presented above and on the supporting.

technical justifications, FPL has concluded that this proposed license amendment involves no significant hazards consideration.

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St. Lucie Unit 1 Docket No. 50-335 Proposed License Amendment St. Lucie Unit 1 Reduction of Reactor Coolant S stem Desi n Flow ENCLOSURE 1 EMF-92-165, "ST. LUCIE UNIT 1 CHAPTER 15 EVENT REVIEW AND ANALYSIS FOR 254 STEAM GENERATOR TUBE PLUGGING": Siemens Power Corporation, February 9, 1993.