ML17228B502

Application for Exigent Amend to License DPR-67,revising TS Re Thermal Margin & RCS Flow Limits.Siemens Power Corp Rept EMF-96-135, St Lucie Unit 1 Chapter 15 Event Review & Analysis for 30% SG Tube Plugging Encl

ML17228B502
Person / Time
Site:	Saint Lucie
Issue date:	06/01/1996
From:	Bohlke W FLORIDA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML17228B503	List: ML17228B502 ML17228B504 ML17228B505
References
L-96-141, NUDOCS 9606030191
Download: ML17228B502 (32)

Text

CATEGORY 1, REGULPT% INFORMATION DZSTRIBUTIOlZSTEM (RIDE)

I ACCESSION NBR:9606030191 DOC.DATE: 96/06/01 NOTARIZED: YES DOCKET FACIL:50-335 St. Lucie Plant, Unit 1, Florida Power 6 Light Co. 05000335 AUTH. NAME AUTHOR AFFILIATION BOHLKEFW.H. Florida Power & Light Co.

RECIP.NAME RECIPIENT AFFILIATION 5~

Document Control Branch (Document Control Desk) p~

Application for exigent amend to license DPR-67,revising TS ~pge.

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SUBJECT:

re thermal margin a RCS flow limits. Siemens Power Corp Rept EMF-96-135, "St Lucie Unit 1 Chapter 15 Event Rview 6 Analysis for 30% SG Tube Plugging" encl.

DISTRIBUTION CODE: AOOZD TITLE: OR COPIES RECEIVED:LTR l Submittal: General Distribution ENCL j SIZE: Zk NOTES: E RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL 0 PD2-3 LA 1 ~ 1 PD2-3 PD 1 1 WIENS,L. 1 1 INTERNAL: ACRS 1 1 1 1 NRR/DE/EMCB 1 1 NRR/DRCH/HICB 1 1 NRR/DSSA/SPLB 1 1 NRR/DSSA/SRXB 1 1 NUDOCS-ABSTRACT 1 1 OGC/HDS3 1 0 EXTERNAL: NOAC 1 1 NRC PDR 1 1 D

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T NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOii! OWFN 5D-5(EXT. 415-2083) TO ELIMINATE YOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!

TOTAL NUMBER OF COPIES REQUIRED: LTTR 13 ENCL 12

Florida Power 5 Light Company, P.O. Box128, Fort Pierce, FL34954-0128 June 1, 1996'-96-141 10 CFR 50.90 10 CFR 50.91 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555 Re: St. Lucie Unit 1 Docket No. 50-335 Proposed Exigent License Amendment Thermal Mar in and RCS Flow Limits Pursuant to 10 CFR 50.90, Florida Power 6 Light Company (FPL) requests to amend Facility Operating License DPR-67 for St. Lucie Unit 1 by incorporating the attached Technical Specifications (TS) revisions. Based on safety analyses assumptions of 30'average) of all steam generator tubes removed from service, the amendment.

reduces the stated value of design reactor coolant flow from 355,000 gpm'to 345,000 gpm, revises the reactor core thermal margin safety limits shown in FIGURE 2.1-1, and modifies the reactor coolant system total water and steam volume described in the design features. The amendment also reduces the Limiting Safety System Setting for the reactor coolant low flow trip function from > 95%

to 2 934 of design reactor coolant flow. Finally, TS 2.1.1 is modified to limit reactor power to < 90'ated thermal power for Cycle 14 operation exceeding mid-cycle fuel burn up conditions.

It is requested by June that 20, 1996, the proposed amendment, if to facilitate timely resumption of power approved, be issued operations. FPL believes that the present situation at St. Lucie Unit 1 satisfies the requisite conditions for issuance of an exigent amendment and hereby requests consideration of this submittal pursuant to 10 CFR 50.91(a)(6).

Attachment 1 provides the basis for consideration of the request. as an exigent amendment. Attachment 2 is an evaluation of the proposed TS change. Attachment 3 is the "Determination of No Significant Hazards Consideration." Attachment 4 contains a copy of the appropriate TS pages marked-up to show the proposed changes.

Enclosed with this submittal is a copy of "Siemens Power Corporation-Nuclear Division, St. Lucie Unit 1 Chapter 15 Event Review and Analysis for 304 Steam Generator Tube Plugging, EMF 135; May, 1996" bOb030 qbObOi P

PDR ADQCK 0 000gg5 PDR P

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St. Lucie Unit 1 L-96-141 Docket No. 50-335 Page 2 Proposed Exigent License Amendment Thermal Mar in and RCS Flow Limits The proposed amendment has been reviewed by the St. Lucie Facility Review Group and the Florida Power 6 Light Company Nuclear Review Board. In accordance with 10 CFR 50.91 (b)(1), a copy of the proposed amendment is being forwarded to the State Designee for the State of Florida.

Please contact us if there are any questions about this submittal.

Very truly yours, W. H. Bohlke Vice President Nuclear Engineering and Licensing WHB/RLD Attachments Enclosure cc: Stewart D. Ebneter, Regiona'l Administrator, Region II, USNRC.

Senior Resident Inspector, USNRC, St. Lucie Plant.

Mr. W.A. Passetti, Florida Department of Health and Rehabilitative Services.

St. Lucie Unit 1= L-96-141 Docket No. 50-335 Page 3 Proposed Emergency License Amendment Thermal Mar in and RCS Flow Limits

)

) SS.

COUNTY OF ST. LUCIE )

W. H. Bohlke being first duly sworn, deposes and says:

That he is Vice President, Nuclear Engineering and Licensing, for the Nuclear Division of Florida Power 6 Light Company, the Licensee herein; That he has executed the foregoing document; .that the statements made in this document are true and correct to the best of his knowledge, information and belief, and that he is authorized to execute the document on behalf of said Licensee.

W. H. Bohlke STATE OF FLOR DA COUNTY OF, ~CA~

The foregoing instrument was acknowledged before me this day of , 19'7(

by W. Bohlke, o 'ersonally known to me and wh did tak a h.

OFHCIAL NOTARY S EAL KAREN E GUTOWSKI FLORIDA

. +MYou)z(~ NOI'ARY PUBLIC STATE OF COMMISSION NO. CC387743 Name of Notary Public MYCOMMISSION EXP.

LY30,1998 My Commission expires No.

RRE'ommission

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St. Lucie Unit 1 Docket No. 50-335 Proposed Exigent License Amendment Thermal Mar in and RCS Flow Limits ATTACHMENT 1 BASIS FOR EXIGENT AMENDMENT CONSIDERATION

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z St. Lucie Unit 1 L-96-141 Docket No. 50-335 Attachment 1 Proposed Exigent License Amendment Page 1 of 2 Therma Mar in and RCS Flow Limits BASIS FOR EXIGENT AMENDMENT CONSIDERATION Pursuant to 10 CFR 50.91(a)(5) and (6), the NRC may adjudge that an emergency or exigent situation exists, in that failure of the NRC to act in a timely way on a request for license amendment would result in prevention of resumption of operation of a nuclear power plant. For consideration of this provision, the licensee is required to explain why the emergency or exigent circumstance occurred and why it could not be avoided. The NRC will then assess the licensee's reasons for failing to file the application it sufficiently in advance of the event, and if determines that the licensee has not abused the emergency/exigent provision, it may issue a license amendment involving no significant hazards consideration without the 30 days normally allowed for prior notice and opportunity for a hearing or for public comment. A discussion of why the present situation at St. Lucie Unit 1 satisfies the requisite conditions for issuance of an exigent amendment follows.

1. Wh the Exi enc Occurred and Could Not be Avoided.

On April 29, 1996, St. Lucie Unit 1 entered a scheduled refueling outage. A margin of approximately 14% existed between the average number of steam generator (SG) tubes that had been previously removed from service and the number of plugged tubes assumed in the safety analyses. Based on a 10 year history of 1004 Eddy Current Testing (ECT), and including additional inspection commitments pursuant to generic letter (GL) 95-03, "Circumferential Cracking of Steam Generator Tubes," the number of tubes conservatively estimated to be removed from service during this outage was far less than the remaining analytical margin.

Based on concerns involving the qualification of techniques for sizing SG tube crack-like indications that were communicated to industry by the NRC staff at the NRC Regulatory Information Conference held April 9-10, 1996, FPL made a pro-active presentation to the staff on April 22, 1996, outlining the status of SG condition, ECT technique qualification, and scheduled refueling outage plans for the St. Lucie Unit 1 SGs. In a follow up telephone conversation with the staff on May 7, 1996, the staff questioned the repair criteria which have been in place at FPL since 1985. FPL subsequently agreed to implement a more conservative criteria for the Cycle 14 inspection. FPL documented this commitment as outlined in FPL letter L-96-129 dated May 14, 1996. Our assessment, of the impact of implementing this criteria indicates that the number of SG tubes to be plugged may exceed the existing 254 (average) analyses limit.

St. Lucie Unit 1 L-96-141 Docket No. 50-335 Attachment 1 Proposed Exigent License Amendment ~

Page 2 of 2 Thermal Mar in and RCS Flow Limits

2. Basis for the Exi ent Amendment Re uest.

Steam generator tube inspections at St. Lucie meet or exceed criteria contained in the St. Lucie Unit 1 Technical Specifications, and PWR Steam Generator Examination Guidelines, Revision 3, EPRI Report NP-6201, November, 1992. Upon completion of review and evaluation of data by industry groups and FPL, new technology has been routinely implemented at St. Lucie in a manner to provide a link to previous examination data.

The change in repair criteria and the magnitude of resultant SG tube plugging could not have been reasonably anticipated prior to NRC staff concerns having been communicated to FPL during the recent meeting and discussions. The need for an amendment to implement revised St. Lucie Unit 1 power and RCS flow limits could not have been anticipated prior to assessing the impact of the change in repair criteria following FPL's meeting and discussions with the NRC staff. The necessary evaluations and preparation of the proposed license amendment were initiated without delay and at the earliest practical time. Analyses and quality assurance verifications to support the proposed license amendment were completed in an expeditious manner, and were performed in parallel with the ongoing tube examinations.

FPL expects to complete the refueling overhaul and the required startup preparations by June 20, 1996. Until a license amendment is issued to authorize operation with the proposed changes, resumption of St. Lucie Unit 1 power operations will be prevented by the current Technical Specifications.

Based on the preceding discussions, FPL believes that the present situation could not have been anticipated to the extent that a request for amendment could have been filed sufficiently in advance; that the emergency/exigency provision has not been abused by not making a timely application and thus itself creating exigent circumstances; and that this license amendment request satisfies criteria for consideration pursuant to 10 CFR 50.91(a)(6).

St. Lucie Unit 1 Docket No.'0-335 Proposed Exigent License Amendment Thermal Mar in and RCS Flow Limits ATTACHMENT 2 EVALUATION OF PROPOSED TS CHANGES

St. Lucie Unit 1 L-96-141 Docket No. 50-335 Attachment 2 Proposed Exigent License Amendment Page 1 of 16 Thermal Mar in and RCS Flow Limits TABLE OF CONTENTS

1. Introduction

2. Proposed TS Changes and Bases

3. General Effect of the Proposed Changes on Event Analyses

4. ,Evaluation of UFSAR Chapter 15 Events

a. Decrease in Secondary Side Heat Removal

b. Decrease in Reactor Coolant System Flow Rate

c. Reactivity and Power Distribution Anomalies

d. Decrease in Reactor Coolant Inventory Events

e. Uncontrolled CEA Withdrawal

f. Boron Dilution Event

g. CEA Ejection Accidents

h. Inadvertent Opening of Pressurizer Pressure Relief Valves

i. Steam Generator Tube Rupture (SGTR) j.

k.

Increase in Heat Removal by the Secondary System Reactor Protection Setpoints

5. Evaluation of Other Selected UFSAR Analyses

a. Plant Natural Circulation Capability

b. Peak Containment Pressurization Following LBLOCA or Steam System Piping Failure

c. Auxiliary Feedwater (AFW) High Energy Line Break

d. Low Temperature Overpressure Protection (LTOP)

e. Overpressure Protection Analysis

f. Impact on Steam Generator Mechanical Loads

6. Conclusion 16

St. Lucie Unit 1 L-96-141 Docket No. 50-335 Attachment 2 Proposed Exigent License Amendment Page 2 of 16 Thermal Mar in and RCS Flow Limits EVALUATION OF PROPOSED TS CHANGES 1., Introduction Safety analyses approved for St. Lucie Unit 1 (PSL1) assume a minimum design Reactor Coolant System (RCS) flow rate of 355,000 gpm and an average 25~ (+ 7%) of all steam generator tubes plugged (SGTP). During the Spring 1996 refueling overhaul, the estimated number of additional steam generator tubes that will be removed from service (currently in excess of 2000) will likely exceed the 254 (average) analyses limit. To conservatively accommodate the larger number of plugged SG tubes, Florida Power and Light Company (FPL) proposes to change the PSL1 Technical Specifications (TS) to reflect safety analysis assumptions of 345,000 gpm minimum RCS design flow rate (based on 30% average of all steam generator tubes plugged), and proposes a change in the Reactor Protective System RCS Low Flow Limiting Safety System Setting from >954 to >93~~ of design reactor coolant flow.

Evaluations to assess the impact of the proposed changes were performed by FPL and Siemens. Power Corporation-Nuclear Division (SPC). It has been determined that the results of the current Small Break Loss of Coolant Accident (SBLOCA) analysis, which assumes 25~ average SGTP (+ 74) and 355,000 gpm reactor coolant flow, will continue to bound full power operation with the proposed amendment for fuel batch average burn up conditions not exceeding 9135 Mwd/MTU (7000 Effective Full Power Hours (EFPH) in Cycle 14).

To assure acceptable margins for full power operation beyond this point, the SBLOCA event analysis must be performed using the values of higher SGTP and reduced flow. Accordingly, the proposed amendment modifies TS 2.1.1 to limit reactor power to < 904 rated thermal power for Cycle 14 operation beyond 7000 EFPH.

2. Pro osed TS Chan es and Bases Copies of the affected TS pages, marked-up to show the changes, are contained in Attachment 4 to this submittal.

a. Pa e 2-1 S ecif ication 2. 1.1 REACTOR CORE: Insert an asterisk following THERMAL POWER, and add the following footnote:

• For Cycle 14 operation beyoml 7000 LPPH, THERMAL POWER shall not erceed 90% of 2700 Megawatts (thermal).

St. Lucie Unit.'1 L-96-141 Docket No. 50-335 Attachment 2 Proposed Exigent License Amendment Page 3 of 16 Thermal Mar in and RCS Flow Limits The limit on thermal power for Cycle 14 operation beyond 7000 EFPH assures that calculated peak fuel cladding temperatures during SBLOCA remain within 10 CFR 50.46 acceptance criteria for the entire operating cycle.

b. Pa e 2-2 FIGURE 2.1-1 Reactor Core Thermal Mar in Safet Limit-Four Reactor .Coolin Pum s O eratin: Replace this figure in its entirety with the revised FIGURE 2.1-1 shown in Attachment 4 of this submittal.

The "vessel flow less measurement uncertainties" is changed from 355,000 gpm to 345,000 gpm based on 30~ average SGTP.

The thermal limit lines shown in this figure have been revised to reflect the reduced flow using SPC methodology. The lines represent the loci of points of thermal power, Reactor Coolant System (RCS) pressure, and maximum cold leg temperature with .

four reactor coolant pumps operating for which the Departure from Nucleate Boiling Ratio (DNBR) is no less than the DNBR limit for the most limiting axial power distribution shown in TS Figure B2.1-1. The minimum DNBR 'limit for steady state operation, normal operational transients, and anticipated tr'ansients remains unchanged from the existing, approved value of 1.22. This value corresponds to 95% probability at a 954 confidence level that DNB will not occur and represents an acceptable margin to DNB for all operating conditions.

c Pa e 2-4 TABLE 2.2-1 Reactor Protective Instrumentation Tri

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Set oint Limits:

(1) Change the TRIP SETPOXNT and ALLOWABLE VALUES for FUNCTIONAL UNIT 3, Reactor Coolant Flow-Low, from > 954 of design reactor coolant flow with 4 pumps operating* to 2 93%

of design reactor coolant fio>v with 4 pumps operating*.

RCS flow is determined by calorimetric methods during power ascension testing after a refueling outage. The actual low flow trip setpoint is based on this inferred flow measurement and is procedurally controlled to consider calorimetric uncertainties, instrument loop uncertainties, instrument signal noise, and the specified design RCS flow. The proposed LSSS is consistent with Cycle 14 safety analysis assumptions and will provide additional operating margin to protect against unwarranted, spurious low flow trips and/or pre-trip alarms.

(2) In Footnote *, change the design reactor coolant flow with 4 pumps operating from 355,000 gpm to 345,000 gpm. The

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St. Lucie Unit 1 L-96-141 Docket No. 50-335 Attachment 2 Proposed Exigent License Amendment Page 4 of 16 Thermal Mar in and RCS Flow Limits proposed value is commensurate with the minimum design flow expected with 30~ SGTP. I

d. Pa e 3 4 2-14 TABLE 3.2-1 DNB MARGIN: Change the Reactor Coolant Flow Rate from 355,000 gpm to 345,000gpm. The proposed value is commensurate with the minimum design flow expected with 30>o SGTP ~

e. Pa e 5-5 DESIGN FEATURES S ecification 5.4.2: Modify the description of the reactor coolant syst: em VOLUME to read:

The total water and steam volume of the reactor coolant system is 11,100 + 180 cubic feet at a nominal T,, of 567 'F, when not accounting for steam generator tube plugging.

This is an administrative change to clarify the condition for which the stated value of RCS volume is applicable.

3. General Effect of the Pro osed Chan es on Event Anal sis The changes proposed by this license amendment can affect the plant safety analyses in the following manner.

a. A reduction in RCS flow rate has an adverse effect on the calculated Departure from Nucleate Boiling Ratio (DNBR). DNBR is a direct indication of available thermal margin, and a reduction in the calculated minimum DNBR indicates that thermal margin for the corresponding transient has been reduced.

b. A reduction in the value of the low flow trip setpoint will result in a lower reactor core flow rate at the time of reactor trip, and can thereby impact the calculated minimum DNBR for certain transients.

c. A reduction in RCS flow rate results in a corresponding increase in RCS average coolant temperature (Tave). A higher Tave can impact both DNBR-related and loss of primary inventory types of transients.

d. The removal of additional steam generator 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. In addition, a reduction of initial RCS inventory due to significant SGTP can affect

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St. Lucie Unit 1 L-96-141 Docket No. 50-335 Attachment 2 Proposed Exigent License Amendment Page 5 of 16 Thermal Mar in and RCS Flow Limits the results of boron dilution events, as well as the depth of core uncovery and calculated peak containment pressure resulting from loss of coolant accidents.

4. Evaluation of UFSAR'ha ter 15 Events A review of events in the St. Lucie Unit 1 Updated Safety Analysis Report (UFSAR) to assess the impact from plant operation with an increase in average SGTP to 30% (+ 74), a reduction in RCS design flow to 345,000 gpm, and a low flow LSSS of 93% of design flow was performed'y both FPL and SPC. SPC's evaluation, including reanalysis data, is reported in the Enclosure (EMF-96-135) with this submittal, and includes a summary disposition with Standard Review Plan/UFSAR event cross-reference in Table 3.1.

Descriptions of how the increased SGTP and reduced RCS flow has been evaluated to impact selected events follow.

a. Decrease in Secondar Side Heat. Removal Loss of External Load LOEL The Loss of External Load (LOEL) was reanalyzed to examine the impact of the proposed changes on the calculated maximum primary and secondary pressures, and to determine whether the existing pressure limit of 2750 psia for the primary, and 1100 psia for the secondary would be exceeded.

The LOEL transient is the limiting event in the "Decrease in Heat Removal by the Secondary System" class because of most rapid reduction of secondary heat removal capability through closure of the turbine stop valves. The assumptions used in this analysis result in this event being more severe than the Turbine Trip, Loss of Condenser Vacuum, and Main Steam Isolation Valve Closure events from a pressurization viewpoint. From DNBR considerations, this event is bounded by the Loss of Flow (LOF) transient.

initial RCS flow is expected to impact DNB-related events in Since a reduction in a similar manner, it is concluded that this event will continue to remain bounded for DNBR by the Loss of Flow event.

Important assumptions used to maximize RCS pressure in this transient are: (a) positive Moderator Temperature Coefficient

(+7 pcm/ F) consistent with the maximum allowed by TS 3.1.1.4, (b) reduced S/G heat transfer area consistent with the proposed tube plugging value, (c) inoperable steam dump and

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St. Lucie Unit 1 L-96-141 Docket No. 50-335 Attachment 2 Proposed Exigent License Amendment Page 6 of 16 Thermal Mar in and RCS Flow Limits bypass system, (d) transient initiated by closure of fast acting (turbine stop) valve, and (e) reactor trip by turbine trip disabled.

A reduction in the RCS flow has no direct impact on the system pressurization. An increased S/G tube plugging has the effect of increasing the primary coolant insurge into the pressurizer. This is because the reduced primary-to-secondary heat transfer area results in a lower initial secondary side pressure, thus delaying Main Steam Safety Valve (MSSV) actuation and leading to a greater expansion of the RCS fluid.

This effect will tend to increase primary syst: em pressure.

Results of the re-analysis for this event indicate the calculated peak primary pressure to be 2714 psia, below the.

limiting criteria of 110'o of design pressure (2750 psia).

Secondary system pressure was determined to be 1031 psia, which is less than the 1100 psia secondary side acceptance criteria. Therefore, it is concluded that increased steam generator tube plugging and the associated reduction in RCS flow, has no adverse impact on compliance with over-pressurization criteria for the decrease in heat removal by the secondary system class of transients.

b. Decrease in Reactor Coolant S stem Flow Rate 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 DNBR (closer to the limit of 1.22). Two events in this category are impacted by the proposed reduction in design RCS flow and low flow trip setpoint: Loss of Reactor Coolant Flow (LOF) and Seized RCP Rotor.

(1) Loss of Reactor Coolant Flow The Loss of Flow event was evaluated with the initial conditions modified to include the proposed changes. The objective of this evaluation was to determine whether the existing DNB-LCO (TS 3/4.2.5), in conjunction with the RPS Low Flow Trip, will prevent the DNBR limit of 1.22 from being violated.

This event is analyzed on a cycle specific basis as part of the Limiting Conditions for Operation (LCO) setpoint verification, because Operational Occurrence it(AOO) is the most limiting Anticipated with respect to DNBR. After

St. Lucie Unit 1 L-96-141 Docket No. 50-335 Attachment 2 Proposed Exigent, License Amendment Page 7 of 16 Thermal Mar in and RCS Flow Limits accounting for the proposed RCS flow reduction, reduced low flow trip setpoint, and increased tube plugging, the transient was evaluated for Cycle 14 reload by applying deterministic penalties to the available power margin. Results of this evaluation show a reduction in the minimum power margin from 6.8> to 1.9% of rated power. The available margin confirms that the minimum DNBR is greater than its limit value of 1.22.

Or, equivalently, the LOF event initiated within the existing DNB LCO constraints will not result in violation of the Specified Acceptable Fuel Design Limit (SAFDL) for DNBR.

(2) Seized RCP Rotor The seized rotor event, is assumed to be initiated by an instantaneous seizure of one of the reactor coolant pump shafts. Because of the very low probability associated with this event, a limited number of fuel rod failures are permitted to occur. This event was evaluated to assess the number of fuel rods expected to fail as a result of the proposed changes to the RCS flow and the low flow trip setpoint.

A decrease in the RCS flow and a reduction in the low flow trip setpoint will result in lower DNBRs and a higher percentage of fuel rod failures for this event. The results of the analysis of record show that, 1'. of the fuel rods will fail. The analysis of record for radiological consequences conservatively assumed fuel failures of 2.54. The margin available in this analysis, due to excess conservatism in the reactor power and Radial Peaking Factor (Fr), has been .

determined to nearly offset the effects of the decreased coolant flow and reduced low flow trip setpoint, resulting in a net power penalty of 0.57:. The small decrease in DNBR associated with the 0.57% power penalty will not cause the fuel rod failures to increase from the present value of 14 to more than 2.5% value used in the radiological analysis, particularly with the statistical treatment of power used in the seized rotor analysis. Therefore, FPL has concluded that the impact of increased SGTP, reduced RCS flow, and reduced low flow trip setpoint on the fuel failure rate resulting from the Seized RCP Rotor accident is acceptable.

c. Reactivit and Power Distribution Anomalies The events in this category are not impacted by the change in low flow trip setpoint. Only the dropped CEA transient required evaluation due to the reduced RCS flow.

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St. Lucie Unit 1 L-96-141 Docket No. 50-335 Attachment 2 Proposed Exigent License Amendment Page 8 of 16 Thermal Mar in and RCS Flow Limits Dro ed CEA This transient is explicitly analyzed each cycle as part of the DNB-LCO setpoint verification. This event causes an asymmetry in the core power peaking distribution and its consequences on DNBR depend on the cycle to cycle fuel loading pattern characteristics. Results of the evaluation performed, after accounting for the proposed changes, show a reduction in the minimum power margin from 8.04 to 4.6% of rated power.

Based on the available margin, it is concluded that the occurrence of a CEA drop event, after implementation of the proposed changes, will not result in violation of the DNBR SAFDL, provided the transient is initiated within the constraints of the DNB LCO.

d. Decrease in Reactor Coolant Inventor Events (1) Lar e Break LOCA LBLOCA This event was evaluated to account for the impact of the proposed changes on the peak cladding temperature. The effects on the blowdown and re-flood phases of the transient due to a lower initial secondary pressure (result of higher tube plugging) and an increased core average temperature (result of lower RCS flow) were estimated to result in a minor impact on the calculated peak cladding temperature (PCT), well within the conservatism in the analysis as described below.

The analysis of record uses a conservative combination of initial fuel pellet stored energy and axial power profiles to bound the entire cycle of operation. Additional conservatism lies in the assumption of a maximum resinter density of 1.14 compared to 0.81% as-built value for Cycle 14. The use of an as-built resinter density is estimated to result in a reduction in initial fuel average temperature of 34 F for the case of fuel stored energy near BOC and at least 16 F for the case of fuel stored energy representing MOC. This amount of conservatism in the fuel stored energy represents a significant conservatism in PCT, and the analysis of record would continue to remain bounding. Additional discussion involving the LBLOCA evaluation is provided in the enclosed report, EMF-96-135.

It is concluded that, a LBLOCA initiated from plant conditions consistent with the proposed changes, will have consequences which satisfy the requirements of 10 CFR 50.46(b).

St. Lucie Unit 1 L-96-141 Docket No. 50-335 Attachment 2 Proposed Exigent License Amendment Page 9 of 16 Thermal Mar in and RCS Flow Limits (2) Small Break LOCA SBLOCA The Small Break LOCA event for St. Lucie Unit 1 was evaluated for the impact of the reduced primary system flow and the increased S/G tube plugging level (30: average).

Increased S/G tube plugging will result in a reduction in the initial RCS inventory which could cause a deeper core uncovery. Also, a small increase in the primary system average temperature due to the lower RCS flow will tend to slow the depressurization, delaying safety injection. These effects, however, are small. The results of this event are influenced more by changes to the top-peaked axial profiles.

The analysis of record uses a top-peaked axial profile from the end-of-cycle (EOC) conditions. A review of Cycle 14 axial profiles showed that the maximum peak power elevation corresponding to the middle-of-cycle (MOC) was lower than that used in the analysis of record. The average burn up of the MOC axial profiles was 7000 EFPH. The conservatism in the analysis due to this axial profile, up to this burn up, will offset any adverse effects due to the increased tube plugging and decreased RCS flow.

A determination of the offsetting effects beyond MOC was not possible without quantifying the effects by re-analysis.

However, since SBLOCA is very sensitive to core power, a reduction in the reactor thermal power to 90: of rated power beyond 7000 EFPH of operation is estimated to provide sufficient margin to offset any adverse effects of the proposed changes. This reduction in core thermal power is estimated to reduce the mass inventory lost out of the break, which is approximately twice as large as the reduction in initial inventory caused by the increased tube plugging.

Also, the primary system pressure will decrease more rapidly because of the lower steam generation rate in the primary system. The analysis of record, therefore, would remain bounding. Additional discussion involving the 904 power constraint is provided in Section 3.2.6.4 of the enclosed report, EMF-96-135.

It is concluded that the current SBLOCA analysis will bound Cycle 14 operation only up to 7000 EFPH.

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St. Lucie Unit 1 L-96-141 Docket No. 50-335 Attachment 2 Proposed Exigent License Amendment Page 10 of 16 Thermal Mar in and RCS Flow Limits

e. Uncontrolled CEA Withdrawal Both the uncontrolled CEA withdrawal from low power and the CEA withdrawal initiated from high power conditions are events analyzed against DNBR criteria. The most limiting anticipated operational occurrence from the DNB considerations is the Loss of Flow event which was evaluated and previously discussed in Section 4.b(1). Since the proposed reduction in RCS flow is expected to affect the DNB-related events in a similar manner, the CEA withdrawal event will continue to remain bounded by the Loss of Flow transient. Since the LOF analysis results are acceptable, it. is concluded that the uncontrolled CEA withdrawal will not result in violation of the DNBR SAFDL, when initiated from within the DNB-LCO.

Boron Dilution Event A boron dilution event can occur during any mode of plant operations. Protection against violation of SAFDL's for boron dilution events initiated at power is provided by the existing TM/LP trip, the Variable High Power Trip (VHPT) and the LPD LSSS.

Increased S/G tube plugging will result in a small change in RCS fluid volume (-1.28%). This in turn will impact the time to criticality determined in the boron dilution event analyses. The reference analyses for dilution events initiated from hot standby or hot/cold shutdown conditions at St. Lucie Unit 1, show that margin exists to the acceptance criteria in the time to criticality. Since Mode 6 only considers the mass inventory in the, reactor vessel, the increase in S/G tube plugging does not affect Mode 6.

For Modes 2 to 4, the decrease in the RCS inventory was calculated to reduce the time to criticality from 72.02 minutes to 71.1 minutes. This time is greater than the acceptance criteria of 15 minutes. The time to criticality for Mode 5 reduced from 20.54 minutes to 20.3 minutes, relative to the criteria of 15 minutes. The boron dilution event results are, therefore, acceptable for the proposed changes.

g, CEA E'ection Accidents A control rod ejection accident is defined as the mechanical failure of a control rod mechanism pressure housing resulting

St. Lucie Unit 1 L-96-141 Docket No. 50-335 Attachment 2 Proposed Exigent License Amendment Page 11 of 16 Thermal Mar in and RCS Flow Limits in the ejection of a CEA and its drive shaft. The consequence of this mechanical failure is a rapid reactivity insertion and an adverse core power distribution, which may result in localized fuel damage. 4 Increased S/G tube plugging will have no significant effect on the results of this transient because predictions of fuel failure are based on fuel centerline melt criteria (deposited energy in the fuel rod), not on DNBR criteria. Therefore,,a reduction in RCS flow proposed here, will not impact the results of this event with respect to core damage or offsite radiological dose consequences.

Inadvertent 0 enin of Pressurizer Pressure Relief Valves Although this event is the most DNB-limiting within the "Decrease in Reactor Coolant Inventory" category of transients, it is bounded by the Loss of Flow event. Since the proposed changes will not affect the relative behavior of DNBR between the two transients, this event will continue to remain bounded by the Loss of Flow event.

The inadvertent PORV opening is also one of the transients used in the determination of the limiting pressure bias term in the TM/LP equation. This 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 depressurization rate in this transient, it is concluded that there is no impact on the existing TM/LP pressure bias.

Steam Generator Tube Ru ture SGTR Existing analyses of the SGTR event demonstrate the potential radiological consequences of this transient. These analyses have concluded that the associated radiological release is primarily dependent on the break flow rate and the corresponding primary-to-secondary mass transfer during the event. The radiological releases were determined to be a small fraction of 10 CFR 100 limits. The differential pressure (primary-to-secondary) across the steam generator tubes determines if the flow through the break is choked or not. The existing analysis of record examined the bounding case where break flow was choked before reactor scram. After reactor scram, the transient response is governed by the opening of steam dump and bypass valves.

St. Lucie Unit 1 L-96-141 Docket No. 50-335 Attachment 2 Proposed Exigent License Amendment Page 12 of 16 Thermal Mar in and RCS Flow Limits The postulated increase in S/G tube plugging level will result in reduced secondary side operating pressure at St. Lucie Unit

1. This change could result in slightly longer times of choked flow for an actual SGTR. However, the analysis of record assumes choked flow conditions during the period of interest before reactor scram, and therefore will remain bounding. It is, therefore, concluded that the proposed changes will not alter the system response and the resultant potential offsite dose consequences for the SGTR event.

j. 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. None of the events in this category are adversely impacted by the proposed increase in steam generator tube plugging or reduced RCS flow and low flow trip setpoint. A discussion of the individual transient evaluations follows.

(1) Excess Load Three events with different initiators are postulated with the limiting sub-event being the inadvertent opening of all the steam dump and bypass system valves at full power. This scenario would result, in an approximate 43.4% increase in steam mass flow rate. This event has been determined to be bounded by the Loss of Flow event for DNB considerations and none of the proposed changes will significantly impact the relative DNBR behavior between these two transients.

Therefore, no reanalysis of this event was required.

(2) Steam S stem Pi in Failures Inside Outside Containment Steam System Piping Failure events 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 either the DNBR or fuel centerline melt SAFDL's. Because the initial core power distribution has not been changed and since the time of minimum DNBR, in the limiting case, occurs during a period of natural circulation flow, the key parameter affecting fuel failures is the return to power level caused by reactivity feedback following the break.

The primary system cooldown following a limiting steam system piping failure initiated with increased steam generator tube plugging and reduced RCS flow will be bounded by (no more

St. Lucie Unit 1 L-96-141 Docket No. 50-335 Attachment 2 Proposed Exigent License Amendment. Page 13 of 16 Thermal Mar in and RCS Flow Limits severe than) the existing analysis. The reduced primary to secondary heat transfer rate across the steam generator and the lower initial secondary pressure both contribute to make this a more benign event. These effects ensure that the existing analysis of record for steam system. piping failures will remain bounding and potential off-site dose consequences remain unchanged.

(3) Inadvertent 0 enin of a Steam Generator Relief Atmos heric Dum Valve This event is normally evaluated to assess radiological consequences. Radiological releases caused by this event will be less severe and less likely to occur after implementation of the proposed changes because of the lower initial secondary side pressure resulting from the increased steam generator tube plugging level. The analysis of .record assumes conservative Technical Specification limits for the primary to secondary leak rate which remains unchanged. Therefore, the existing analysis of record will remain bounding for this event.

k. The impact of the proposed changes on relevant setpoint analyses was also evaluated. The setpoint analyses include the Reactor Protection System (RPS) Local Power Density (LPD)

LSSS, LPD Limiting Condition for Operation (LCO), Thermal Margin/Low Pressure (TM/LP) LSSS, and the DNB LCO for allowable core power as a function of Axial Shape Index (ASI).

Verification of the validity of these setpoints is discussed in Section 4.0 of EMF-96-135.

5. Evaluation of Other Selected UFSAR Anal ses While the discussion in Sections 3 and 4 focused on the impact of the proposed RCS flow reduction and increased steam generator tube plugging on the results of the plant UFSAR Chapter 15 Safety Analysis, a complete evaluation of the impact of the proposed changes requires the evaluation of several additional issues.

These are summarized below.

a. Plant Natural Circulation Ca abilit FPL has examined the increased tube plugging to determine any adverse impact on natural circulation cooling capability if would result.

St. Lucie Unit 1 L-96-141 Docket No. 50-335 Attachment 2 Proposed Exigent License Amendment Page 14 of 16 Thermal Mar in and RCS Flow Limits The calculations performed by FPL in support of the current analysis limit of 25: (+ 7:) tube plugging showed that natural circulation capabilities with 18% and 324 SGTP were very similar, with cold leg temperatures differing by less than 1 F (FPL letter L-93-035; 3/19/93 in support of License Amendment 130). The simulation was performed assuming no operator actions, and with heat removal by the SG safety valves. FPL determined that the cooldown rate was dominated by operation of the secondary safety valves, and that increased SGTP had no adverse impact. Therefore, FPL has concluded that the proposed changes will not prevent the occurrence of natural circulation.

b. Peak Containment Pressurization Followin LBLOCA or Steam S stem Pi in Failure Large Break LOCA and Steam Pipe Break Inside Containment analyses of record were evaluated to determine if the reduced RCS flow and/or increased tube plugging level would cause the containment design pressure value to be exceeded.

For the LBLOCA event inside containment, the reduction in primary system fluid volume available for blowdown, a higher resistance to blowdown, and less secondary to primary heat.

transfer completely offset the effects from a slight increase in system energy due to the higher initial RCS Tave. The peak pressure in the analysis of record will remain bounding.

Steam Piping Failures inside containment were also examined and it was concluded that, after allowing for the proposed changes, no compromise of the pressure limits on containment analysis would result. Increased tube plugging will result in a small increase in the total secondary side mass inventory, 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 S/G) prior to Main Steam Isolation Signal (MSIS).

co 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 offsite power as the limiting condition for plant operators to be able to initiate auxiliary feedwater flow. The analysis for this event was evaluated with respect to the increased average primary coolant temperature. It was determined that an additional 637

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St. Lucie Unit. 1 L-96-141 Docket No. 50-335 Attachment 2 Proposed Exigent License Amendment Page 15 of 16 Thermal Mar in and RCS Flow Limits ibm of inventory would be boiled off from the secondary side reducing the dryout time from 650 seconds to 611.7 seconds.

No credit, was taken for the increased initial secondary side mass inventory. It is concluded that will be satisfied with increased all acceptance criteria steam generator tube plugging and reduced RCS flow.

d. Low Tem erature Over ressure Protection LTOP Anal sis The existing LTOP analysis was evaluated to determine whether the postulated increase in steam generator tube plugging would impact the consequences of starting a RCP with the plant secondary side at a higher temperature than the primary side.

Plugging an increased number of S/G tubes was determined to result in a slight increase of the thermal time constant of the system, which for this event leads to a slower rate of energy addition to the primary side.

Only a change in the RCP heat output or in the initial condition of primary to secondary hT could change the energy deposited in the primary system, and hence, the peak pressure.

Therefore, increasing the steam generator average tube plugging to 30: has no adverse impact on the pressure spike caused by starting a RCP pump under low temperature conditions.

e. Over ressure Protection Anal sis In Appendix 5A of the St. Lucie 1 UFSAR, an analysis documenting the sizing of the primary and secondary safety valves is presented. The intent of this analysis, which is based on a "worst case" loss of load event, is to size the safety valves. The safety valve flow rates are chosen such that the peak system pressure is less than 110% of design.

The impact of the proposed increase in steam generator tube plugging and reduced RCS flow on the licensing analysis for the Loss of External Load event was previously discussed (see Section 4.a) . Since that analysis confirmed compliance with the pressurization criteria, it indirectly verified the continued validity of the main steam safety valve sizing analysis of reference. Therefore, it is concluded that the proposed changes do not require an increase in main steam safety valve capacity to satisfy the overpressurization criteria.

St. Lucie Unit 1 L-96-141 Docket No. 50-335 Attachment 2 Proposed Exigent License Amendment Page 16 of 16 Thermal Mar in and RCS Flow Limits Im act on Steam Generator Mechanical Loads The steam generator inlet temperature corresponding to 345,000 gpm RCS design flow is calculated to be less than 604 F. The temperature value of 604 F is supported by the S/G mechanical load calculations performed for 25> + 7> asymmetry tube plugging case. Therefore, there is no adverse impact on any acceptance criteria for the tube sheet and steam generator tube bundle, and sufficient margin to stress limits will remain available.

6 ~ Conclusion Reactor core thermal margin safety limits illustrated in TS Figure 2.1-1 have been adjusted to account for the proposed value of design RCS flow, and define the areas of safe operation in terms of thermal power, RCS pressure, and cold leg temperature for which the DNBR is no less than the MDNBR limit. The minimum DNBR limit for - steady state operation, normal operational transients, and anticipated transients remains unchanged from the existing, approved value of 1.22.

The validity of Reactor Protective Instrumentation settings and trip functions in conjunction with related Limiting Conditions for Operation has been verified to provide assurance that reactor core design limits are not exceeded for the proposed change in RCS design flow.

The assessments performed of relevant safety analyses for St.

Lucie Unit 1 demonstrate that established acceptance criteria for plant performance will remain satisfied during operation with the proposed amendment, which includes a derate to g 90%

rated thermal power for operation beyond 7000 EFPH in Cycle

14. In addition, potential radiological consequences determined in the analyses of record, and which demonstrate compliance with 10 CFR 100 acceptance criteria will remain bounding for operation with the reduced RCS flow and increased SGTP. Therefore, FPL concludes that operation of the facility in accordance with this proposed amendment is acceptable.

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St. Lucie Unit I Docket No. 50-335 Proposed Exigent License Amendment Thermal Mar in and RCS Flow Limits ATTACHMENT 3 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION

St. Lucie Unit 1 L-96-141 Docket No. 50-335 Attachment 3 Proposed Exigent License Amendment Page 1 of 2 Thermal Mar in and RCS Flow Limits DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION Descri tion of'mendment: The proposed license amendment changes the Technical Specifications (TS) by reducing the design reactor coolant flow with four reactor coolant pumps operating from 355,000 gpm to 345,000 gpm (TABLE 2.2-1 and TABLE 3.2-1), revising the reactor core thermal margin safety limits shown in FIGURE 2.1-1, and modifying the reactor coolant system total water and steam volume described in design features section 5.4.2. The amendment also reduces the reactor coolant low flow trip function Limiting Safety System Setting (LSSS) from > 95% to h 93% of design reactor coolant flow. Finally, TS 2.1.1 is modified to restrict reactor power to S 904 rated thermal power for operation beyond'iddle of Cycle 14. The revisions are made to support changes in the safety analyses which accommodate a larger number of plugged steam generator tubes for Operating Cycle 14.

Pursuant to 10 CFR 50.92, a determination may be made that a proposed license amendment involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment 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 consequences of an accident previously evaluated.

The proposed amendment defines reactor core thermal margin safety limits for a reduced value of design reactor coolant flow, and establishes a revised Limiting Safety System Setting (LSSS) for the protective system low flow trip. As core protection variables, these limiting parameters are not accident initiators and do not affect the frequency of occurrence of previously analyzed transients. The design features'otal water and steam volume revision accounts for steam generator tube plugging and is simply administrative in nature. Evaluations performed to assess the impact of the proposed amendment conclude that, when considering a unit derate to 90> rated thermal power for operation beyond 7000 EFPH in Cycle 14 as required by the proposed change to TS 2.1.1, the potential radiological consequences of previously analyzed

St. Lucie Unit 1 L-96-141 Docket, No. 50-335 Attachment 3 Proposed Exigent License Amendment Page 2 of 2 Mar in and RCS Flow Limits 'hermal transients will conservatively remain within established acceptance criteria. Therefore, operation of the facility in accordance with this amendment would not involve a significant. increase in the probability or the consequences of any accident previously evaluated.

(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.

The proposed amendment revises limiting parameters to assure safe operation commensurate with the impact of steam generator tube plugging, and will not change the modes of operation defined in the facility license. The analysis of transients associated with steam generator failures are part of the design and licensing bases.

Therefore, 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.

t3) Operation of the facility in accordance with the proposed amendment would not involve a significant reduction in a margin of safety.

The proposed amendment allows full power operation at an RCS flow commensurate with 30% (average) steam generator tube plugging for Cycle 14 fuel batch average burn up conditions corresponding to mid-cycle. For operation beyond mid-cycle, reactor power will be restricted to < 90% rated thermal power. An evaluation of limiting events to established acceptance criteria for Specified Acceptable Fuel Design Limits (SAFDL), primary and secondary over pressurization transients, 10 CFR 50.46(b) emergency core cooling systems acceptance criteria, peak containment pressure, potential radiation dose during accidents, and to TS Limiting Conditions for Operation has been completed in support of this amendment request.

The evaluation concludes, when considering the proposed LSSS for the Low Flow trip, that a conservative margin to acceptable limits remains available. Therefore, operation of the facility in accordance with this proposed amendment would not. involve a significant reduction in a margin of safety.

Based on the above discussion and the supporting evaluation of technical specification changes, FPL has determined that the proposed license amendment involves no significant hazards consideration.