ML17353A645

Summary of 960404 Meeting W/Util in Rockville,Md to Discuss Thermal Uprate Submittal

ML17353A645
Person / Time
Site:	Turkey Point
Issue date:	04/17/1996
From:	Croteau R NRC (Affiliation Not Assigned)
To:	NRC (Affiliation Not Assigned)
References
NUDOCS 9604230371
Download: ML17353A645 (76)

Text

April 17, 1996 LICENSEE:

FACILITY:

SUBJECT'lorida Power tm Light Company Turkey Point Units 3 and 4

SUMMARY

OF MEETING ON APRIL 4,

1996, REGARDING THE THERMAL UPRATE PROJECT Licensee representatives met with members of the staff on April 4, 1996, in Rockville, Maryland, to discuss the thermal uprate submittal dated December 18, 1995.

Enclosure 1 is a list of attendees and Enclosure 2

consists of the licensee's handout.

The meeting was a followup to the staff's request for additional information (RAI) dated March 26,

1996, on the above subject.

The licensee presented an overview of the uprate submittal and addressed'the questions presented in the March 26,

1996, RAI.

The licensee agreed to formally provide the answers to the staff's questions in a letter, which is expected in approximately 30 days.

The meeting was helpful in providing a better understanding of both the staff and licensee positions on the issues associated with the thermal uprate project.

(Original Si,gned By)

Richard P. Croteau, Project Manager Project Directorate II-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation Docket Nos.

50-250 and 50-251

Enclosures:

As stated cc w/enclosures:

See next page FILENAME G:

TURKEY MTG UPR.SUH

'FFICE LA:PDI I-1 EDunnin on PH:PDII-1 RCroteaug:

D:PDII-1 EImbro~

DATE 04 /4 96 04 b

96 04 /

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Yes No FICIAL RECORD COPY 220'018 960423037i 9604i7 PDR ADOCK 05000250 P

PDR XIIII;HLKC~~ ~P>

C il LI 0

IN MEMORANDUM DATED APPil i 7-,

1996 I

I Distributi on,,

Docket File*~

PUBLIC*

PD2-I RF*

WRussell/FMiraglia RZimmerman Ii SVarga JZwolinski OGC EJordan RButcher ACRS

BBurton, EDO RII KLandis, RII

• Handout only

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Florida. Power and Light Company Turkey Point Plant Units 3 and 4

J.

R. Newest, Esquire

Morgan, Lewis 8 Bockius 1800 H Street, NW Washington, DC 20036 Jack Shreve, Public Counsel Office of the Public Counsel c/o The Florida Legislature 111 West Hadison Avenue, Room 812 Tallahassee, Florida 32399-1400 John T. Butler, Esquire

Steel, Hector and Davis 4000 Southeast Financial Center Miami, Florida 33131-2398 Mr. Robert J.

Hovey, Site Vice President Turkey Point Nuclear Plant Florida Power and Light Company P.O.

Box 029100 Miami, Florida 33102 Armando Vidal County Manager Metropolitan Dade County 111 NW 1 Street, 29th Floor Miami, Florida 33128 IC Hr. Joe Hyers, Director Division of Emergency Preparedness Department of Community Affairs 2740 Centerview Drive Tallahassee, Florida 32399-2100 Regional Administrator, Region II U.S. Nuclear Regulatory Commission 101 Marietta Street, N.W. Suite 2900 Atlanta, Georgia 30323 Attorney General Department of Legal Affairs The Capitol Tallahassee, Florida 32304 Plant Hanager Turkey Point Nuclear Plant Florida Power and Light Company P.O.

Box 029100 Miami, Florida 33102 Hr. H. N. Paduano, Hanager Licensing

& Special Programs Florida Power and Light Company P.O.

Box 14000 Juno

Beach, Florida 33408-0420 Senior Resident Inspector Turkey Point Nuclear Generating Station U.S. Nuclear Regulatory Comission P.O.

Box 1448 Homestead, Florida 33090 Mr. Bill Passetti.'ffice of IOSwtion Control Department Of. Health and'ehabi1 ititive Services 1317 Winewood Blvd.,

Tall ahassee, Flor ida 32399-0700 Mr. Kerry Landis U.S. Nuclear Regulatory Commission 101 Marietta Street, NW Suite 2900 Atlanta, Georgia 30323-0199 Hr. Gary E. Hollinger Licensing Manager Turkey Point Nuclear Plant P.O.

Box 4332 Princeton, Florida 33032-4332 Mr. T, F. Plunkett President - Nuclear Division Florida Power and Light Company P.O.

Box 14000 Juno

Beach, Florida 33408-0420

ENCLOSURE 1

FPSL/NRC Meeting April4, 'l996 List of Attendees Name Office E.

Thompson D. Baker J.

Luke R. Grendys J.

Hoffman L. Hartin J. Hickey H.

Fagan M. Schoppman J. Deblasio G. Sharp E.

Imbro R. Croteau H. Garg D.

Shum C. Liang L. Kopp C.

Wu L. Brown J. Hinns J. Medoff R.

Goe1 C. Craig J.

Tsao FPL FPL FPL Westinghouse FPL FPL FPL Westinghouse FPL Westinghouse Westinghouse NRC NRC NRC NRC NRC NRC NRC NRC NRC NRC NRC NRC NRC

ENCLOSURE 2 FLORIDA POWER 8c LIGHT COMPANY

+ lear +

THERMALPOWER UPRATE PROJECT NRC / FPL MEETING APRIL 4, 1996

TURKEY POINT NUCLEAR PLANT THERMALPOMfER UPRATE PROJECT I.

Introduction II.

Uprate Program Tom Luke - Engineering Mgr.

Liz Thompson - Project Engineer Overview Operational Benefits Schedule III.

Proposed License Amendment Jim Hickey - Licensing Engineer Contents Scope IV. Accident Analyses Input Parameters Large Break LOCA Small Break LOCA Non-LOCA Transients Containment Dose Analyses Combustible Control Gases Leo Martin - Nuc. Fuels Supv.

Dean Baker - Eng. Lic. Supv:

V.

Selected Systems and Components Environmental Considerations Reactor Vessel Integrity RPS/ESFAS Instrumentation Spent Fuel Pool Cooling Dean Baker - Eng. Lic. Supv.

Jack Hoffman - Mech. Eng.

Supv.

Vl.

Questions / Responses

/ Open Discussion Liz Thompson - Project Engineer Attachments 1) 2)

3)

Tentative FPL Participant List Tentative NRC Participant List Letter to T. F. Plunkett, FPL, from R. P. Croteau, NRC, dated March 26, 1996

0

TURKEY POINT NUCLEAR PLANT THERINALPOWER UPRATE PROJECT Introduction

+

Purpose

~

Answer staff questions Facilitate staff review

~

~

gq IA

~

~ e ~ ga,t a

+

Power Increase

~

2200 IVIW~to 2300 MW,

~

4.5% increase

~

Insignificant effect on risk to public

+

Presentation Format

~

Overview Incorporates many question responses

~

Responses

TURKEY POINT NUCLEAR PLANT THERMALPOWER UPRATE PROJECT II.

Uprate Overview

+

Current licensed core power level - 2200 MW~/unit 4

+

Request to re-license Turkey Point Units 3 5 4 Uprated core power level - 2300 MW~/unit 31 MW./unit output gain projected, 62 MW. total Previous studies indicated prudency IA'300 MW~

Original plant design power level H.B. Robinson, a sister plant, is licensed to 2300 MW, Major re-analysis Large Break LOCA Small Break LOCA Steam Generator Tube Rupture Non-LOCA Transients Radiological Dose Calculations Containment Analyses Structures, Systems and Components Limited physical modifications No major equipment replacements

+

Benefits Design Basis Documentation Improved and Updated Greater In-Mouse Experience and Knowledge Benefit to the Rate Payer Reduces system fuel expense j owest cost capacity management option Benefit to the FPL Shareholder Deferral of cash requirements for new generation Better utilization of existing assets 3% Reduction in Overall Cost per MWH

TURKEY POINT NUCLEAR PLANT THERMALPOWER UPRATE PROJECT

+

Operational Benefits

~

MSSV/PSV tolerance increase

+ 3% justified for MSSVs

+ 2% - 3% justified for PSVs

~

HHSI Pumps Large Break LOCA Small Break LOCA One degraded pump assumed

~

Component Cooling Water (CCW)

CCW heat exchanger allowed increased fouling Cleaning needed less often Increased CCW heat exchanger availability Analyzed post-LOCA system temperatures increased

~

Emergency Containment Coolers (ECC)

CCW and containment analyses Containment temperature and pressure profiles show 1 ECC needed ( 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, 2 ECCs ) 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Swing ECC will no longer auto start; may be manually started within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> OPbT, OTBT Maintain increased operating margin K, - 1.24

~

Steam Generator Lo-Lo Level Reactor Trip Nominal trip setpoint of 10% supported Improves ability to withstand feedwater pump trip

0

TURKEY POINT NUCLEAR PLANT THERMALPOWER UPRATE PROJECT Pro'ec S he uie Engineering Analp8I8

~m 3've eve res see ebs woe tnt asa aae eve sm ao1 em t

I

~

i ucensing Submittal NRC Review Implementation Preparation i

II Implementation Of Modifications Implement Uprating I

4 NO.

TOPIC 1

LEAK BEFORE BREAK LICENSING SUBMIT ALS SUBMITTAL DATE 2/2/95 REQUESTED APPROVAL DATE 6/23/95A REVIEW STATUS APPROVED REVISED THERMALDESIGN PROCEDURE (RTDP)

INSTRUMENTATIONTECH SPEC FORMAT 4

SMALLBREAK LOCA UPRATING 5/5/95 5/23/95 7/26/95 12/18/95 2/20/96A

'/24/95A 5/31/96 9/30/96 APPROVED APPROVED IN REVIEW IN REVIEW

TU.RKEY POINT NUCLEAR PLANT THERMALPOWER UPRATE PROJECT III.

Proposed License Amendment (PLA)

+

PLA Content

~

Cover Letter Description of Amendments requested Determination of No Significant Hazards Consideration Proposed Technical Specifications Marked up Technical Specification pages Core Operating Limits Report Uprating Licensing Report WCAP-14276, Revision 1

Integrated report Environmental Considerations Westinghouse Revised Thermal Design Procedure Instruments Uncertainty WCAPs-13718, Revision 2 and 13719, Revision 2 Non-Proprietary and Proprietary Updated to reflect RPS/ESFAS setpoint changes due to uprating

~

Proprietary Information Notice (WCAP-13719)

TURKEY POlNT NUCLEAR PLANT THERMALPOWER'PRATE PROJECT PLA SCOPE 1.

License Condition, Rated Thermal Power, Core Safety Limits, Reactor Trip System Instrumentation Trip Setpoints, Engineered Safety Features Actuation System (ESFAS) Instrumentation Trip Setpoints, Departure from Nucleate Boiling (DNB) Parameters and Reactor Coolant Pump (RCP)

Breaker Position Trip Increase in RATED THERMAL POWER from 2200 MW,to 2300 MW,.

~

New Core Safety Limits curve which reflects the increased power level, revised flowrate and increased peaking factors.

~

Revisions to Overtemperature

~T and Overpower ~T based on the use of the Revised Thermal Design Procedure (RTDP) methodology at the higher power level.

~

Reduced TS value for loop flowrate to account for the increase in the percentage of steam generator tube plugging margin.

~

Changes to T,Pressure and Flow associated with changes to the specific instrument uncertainties and increased power level.

2.

Available Volume Change for Condensate Storage Tank and the Demineralized Water Storage Tank and Reduced Safety Injection (Sl) Pump Discharge Head Requirement Revise the minimum volume available for the condensate storage tank (CST) and the demineralized water storage tank (DWST) so that their design safe shutdown functions can be met.

Reduce the Sl pump discharge head.

This reduction in required head will provide margin for meeting Sl pump test acceptance criteria and also provide available margin should pump degradation occur.

TURKEY POINT-NUCLEAR PLANT.

THERMALPOWER UPRATE PROJECT 3.

Pressurizer and Main Steam Safety Valve Setpoint Tolerances

~

Increase PSV tolerance from "~1%" to "+2%, -3%".

~

Increase MSSV tolerance from "~ 1%" to "~ 3%".

~

Reset the PSVs and MSSVs following testing to "within~ 1%".

4.

Operation at Reduced Power with Inoperable Main Steam Safety Valves

~

For one inoperable MSSV (per steam line), reduce the maximum allowable power level from "56%" to "53%".

For two inoperable MSSV (per steam line), reduce the maximum allowable power level from "35%" to "33%".

5.

Service Period for Heatup and Cooldown Pressure-Temperature Limit Curves

~

Change the service period for the Heatup and Cooldown Pressure-.

Temperature limit curves from "20 EFPY" to "19 EFPY".

6.

Modification to Surveillance Requirement for Emergency Containment Cooling (ECC) System

~

Revise the ECC start logic to support both the containment integrity safety analyses and component cooling water system thermal

analyses, by requiring a maximum of two ECCs receive an automatic start signal following generation of an Sl signal to ensure component cooling water system (CCWs) temperature limits during injection and/or recirculation phases of a LOCA are not exceeded.

7.

Control Room Emergency Ventilation System

~

Revise the methyl iodide removal efficiency from "90%" to "99%."

8.

Relocate F~ and F~ Limits from Technical Specifications to the Core Operating Limits Report (COLR) (also includes a Power Factor Multiplier) and Editorial Corrections.

TURKEY POINT NUCI EAR PLANT Pyr THERIIIIALPOWER UPRATE PROJECT

+ice, IV.

Accident Analysis

+ Input Parameters Selected Parameter Comparison Parameter NSSS Power (MWN)

NSSS TDF (per loop)

Full Power Tavg Max SGTP Max FhH Max FQ DNB Methodology Max MDC Max MTC PSV Setpoint Tolerance SG Safety Valve Tolerance Current Values 2208 MWN 89500 gpm 574.2'F 5

1.62 2.32 STDP 0.43 b,k/g/cc

+5.0 pcm/'F at 70% RTP ramp to Opcm/'F at 100% RTP

+/- 1%

Design + 3%

tolerance and accumulation "RTDP" Values (SER issued 2/20/96) 2208 MWN 85000 gpm 574.2'F 20%

1.70 2.50 RTDP 0.50 hk/g/cc

+7.0 pcm/'F at 70% RTP ramp to Opcm/'F at 100%

RTP

+2%

-3%

3% tolerance and 3% accumulation Uprating Values (WCAP-1 4276, Rev. 1) 2308 MWN 85000 gpm 574.2'F

+/-

3'F 20%

• 1.70

'.50*

RTDP 0.50 b,k/g/cc

+7.0 pcm/'F at 70% RTP ramp to Opcm/'F at 100% RTP

+2%, -3%

3% tolerance and 3%

accumulation

+ Only safety grade components assumed available for mitigation

+ Revised instrumentation safety analysis limit values used

+ Worst case parameter value sets utilized for each analysis

+

• Appendix K LBLOCA limits are 5% SGTP, 2.35 FQ and 1.64 Fb,H

~O Op TURKEY POINT NUCLEAR PLANT THERIIIIALPOWER UPRATE PROJECT

+eO(gag Q4 LOCA Analyses

+ NRC approved evaluation models (EMs) used for LOCA analyses

~

Exception of COSI/Safety Injection in Broken Loop for SBLOCA Currently being reviewed by NRC Comparisons of calculated results made during EM licensing Compared to single effect and integral test data EMs meet 10CFR50, Appendix K requirements

+ Large Break LOCA Originally - 4 codes (each phase of transient and containment response)

SATAN - blowdown WREFLOOD - refill and reflood LOCTA - cladding heatup COCO - containment pressure response BART developed Calculated detailed thermal hydraulic conditions in core during reflood Generically reviewed and approved by NRC (1984)

WCAP-9561-P-A

~

BASH developed Included a more detailed loop model Generically reviewed and approved by NRC (1986)

WCAP-10266-P-A, Revision 2

TURKEY POINT NUCLEAR PLANT THERMALPOWER UPRATE PROJECT

~

Current LBLOCA Analysis in UFSAR Section 14.3.2.1 Used 1981 Evaluation Model (EM) with BART

~

Uprate LBLOCA Analysis in WCAP-14276, Revision 1, Section 3.3.1 Used 1981 Evaluation Model with BASH EM shown in Figure 3.3.1-2 of Uprate Licensing Report EM described in Section 3.3.1.3 of Uprate Licensing Report Status of Evaluation Model Changes Used for Uprating No models used which are currently on NRC docket for review ESHAPE methodology for explicit modeling of Power Shapes Approved in Addendum 1-A of WCAP-10266-P-A, Revision 2

+ Limiting Single Failure for Large Break LOCA

~

Limiting Single Failure is loss of one Low Head Safety Injection (LHSI) pump Limiting single failure per Westinghouse sensitivity studies Documented for 1981 EM with BASH in Section 11.0 of WCAP-10266-P-A, Revision 2

~

Loss of a Diesel Generator is not the limiting single failure for LBLOCA Failure results in loss of one train of containment pressure reducing equipment All pressure reducing equipment operating at maximum heat removal capacity is required by 10CFR50, Appendix K, Section I.D.2 and Branch Technical Position CSB 6-1

TURKEY POINT NUCLEAR PLANT THERMALPOWER UPRATE PROJECT a

+ Small Break LOCA

~

NOTRUMP EM utilized Calculates thermal-hydraulic response of SBLOCA LOCTA calculates the fuel rod cladding heatup transient COSI/Safety Injection in the Broken Loop-.improved condensation model

~

Current SBLOCA Analysis in UFSAR Section 14.3.2.2 Used 1985 NOTRUMP Evaluation Model Reviewed and approved by NRC (1985)

WCAP-10054-P-A Uprate SBLOCA Analysis in WCAP-14276, Revision 1, Section 3.3.2 Used 1985 NOTRUMP Evaluation Model Reviewed and approved by NRC (1985)

WCAP-10054-P-A, Addendum 2 EM shown in Figure 3.3.2-3 of Uprate Licensing Report EM described in Section 3.3.2.3 of Uprate Licensing Report

~

Status of Evaluation Model Changes used for Uprating COSI/Sl in Broken Loop Model currently on docket for NRC review Documented in Addendum 2 of WCAP-10054-P-A Uprate SBLOCA Report submitted in advance to allow sufficient review time Evaluation Model Change used in analyses submitted by 2 other utilities

TURKEY POINT NUCLEAR PLANT'HERMAk POWER'PRAISE PROJECT' Non-LOCA Transients

~

Analysis margin used to improve operational margin and power increase

~

RTDP and LOFTRAN Both NRC approved methodologies

~

Revised Core Thermal Limits

~

New OT/OPBT Setpoints

~

Uprate analyses analyzed with same inputs as RTDP program for:

Reduced RCS flow Increased MSSV/PSV setpoint tolerances F~ = 1.7; F~=2.5 (limited to lower values by LBLOCA)

~

All acceptance criteria met

s

TURKEY POINT NUCLEAR PLANT THERMALPOWER UPRATE PROJECT o

Summary of Turkey Point Non-LOCA Transients'nalyses FSAR Section 1 4.1.1 1 4.1.2 14.1.4 14.1.5 14.1.6 Accident Uncontrolled Rod Cluster Control Assembly Bank Withdrawal from a Subcritical Condition Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power Rod Cluster Control Assembly Misoperation (Dropped RCCA(s))

Chemical and'Volume Control System (CVCS)

Malfunction Start-up of an Inactive Reactor Coolant Loop Changes from Current FSAR to "RTDP" analyses (SER

~ Issued 2/20/96)

Same Methodology.

Incorporated RTDP program parameters

- Reanalyzed RTDP Methodology utilized.

Incorporated RTDP program parameters

- Reanalyzed RTDP Methodology utilized.

Incorporated RTDP program parameters

- Reanalyzed Same Methodology.

Incorporated RTDP program parameters

- Reanalyzed Because the Turkey Point Technical Specifications preclude operation with N-1 loops out of operation, this event was deleted from the Turkey Point non-LOCA licensing basis Changes from RTDP analyses to Uprated CondMons (WCAP-14276, Rev. 1)

Same Methodology as RTDP Program.

Incorporated Uprating Program Parameters Although a HZP event, reanalyzed for completeness.

Same Methodology as RTDP Program.

Incorporated Uprating Program Parameters-Reanalyzed Westinghouse Dropped Rod Methodology detailed in WCAP-11394-P-A (Ref 3) was utilized - Reanalyzed Same Metliodology as RTDP Program.

Incorporated Uprating Program Parameters-Reanalyzed Not Applicable

-~

0

TURKEY POINT NUCLEAR PLANT THERNIALPOWER UPRATE PROJECT FSAR Section Accident Changes from Current FSAR to "RTDP" enelyses

{SER Issued 2/20/96)

Changes from RTDP enelyses to Upreted Conditions IWCAP-14276, Rev. 1) 14.1.7 Feedwater System Malfunctions Causing an Increase in Feedwater Flow Feedwater Malfunction (FWM) event that results in excessive feedwater flow is known to be more limiting than the Feedwater Enthalpy Reduction analysis.

The FWM which results in excessive feedwater flow was analyzed.

RTDP

'ethodology was employed.

Incorporated RTDP Program parameters

- Reanalyzed Same Methodology as RTDP Program.

Incorporated Uprating Program Parameters.

Reanalyzed HFP case, Evaluated HZP case 14.1.8 'xcessive Load Increase RTDP Methodology utilized.

Incorporated RTDP program parameters

- Reanalyzed Same Methodology as RTDP Program.

Incorporated Uprating Program Parameters-Reanalyzed 14.1.9 14.1.9 14.1.10 Partial/Complete Loss of Forced Reactor Coolant Flow Reactor Coolant Pump Shaft Seizure ILocked Rotor)

Loss of External Electrical Load and/or Turbine Trip RTDP Methodology utilized.

Incorporated RTDP program parameters

- Reanalyzed RTDP Methodology utilized for rods in - DNB case.

STDP methodology utilized for peak pressure / peak clad temperature cases-Reanalyzed RTDP Methodology utilized.

Incorporated RTDP program parameters

- Reanalyzed Same Methodology as RTDP Program.

Incorporated Uprating Program Parameters-Reanalyzed Same Methodology as RTDP Program.

Incorporated Uprating Program Parameters-Reanalyzed Same Methodology as RTDP Program.

Incorporated Uprating Program Parameters-Reanalyzed 14.1.11 Loss of Normal Feedwater RTDP Methodology utilized.

Incorporated RTDP program parameters

- Reanalyzed Same Methodology as RTDP Program.

Incorporated Uprating Program Parameters-Reanalyzed

TURKEY POINT NUCLEAR PLANT THERMALPOWER UPRATE PROJECT a

FSAR Section 14.1.12 14.'I.13 14.2.4 1 4.2.5 1 4.2.6 Accident Loss of Non-Emergency AC Power Turbine Generator Design Analysis Steam Generator Tube Rupture Rupture of a Steam Pipe at HZP Rupture of a Control Rod Mechanism (CRDM)

Housing - RCCA Ejections Chenges from Current FSAR to "RTDP" eneiyses (SER issued 2/20/96)

RTDP Methodology utilized.

Incorporated RTDP program parameters

- Reanalyzed No Changes No Changes Same Methodology.

Incorporated RTDP 'program parameters

- Reanalyzed Same Methodology.

Incorporated RTDP program parameters

- Reanalyzed Chmges from RTDP melyses to Upreted Condlt(ons (WCAP-14276, Rev. 1)

Same Methodology as RTDP Program.

Incorporated Uprating Program Parameters-Reanalyzed Existing evaluation applicable to uprated conditions.

Speed control unchanged.

LP turbine ratings exceed uprating conditions.

Same Methodology as current analyses.

Calculations updated for uprate conditions.

RTDP program non-LOCA statepoints were evaluated for applicability and utilized in the DNB analysis-Evaluated Same Methodology as RTDP Program.

Incorporated Uprate Program parameters'-

Reanalyzed

TURKEY POINT NUCLEAR PLANT THERMALPOWER UPRATE PROJECT Containment Analyses

+ Events performed at uprated conditions for LOCA and Steam Line Break

~

MSLB mass 5 energy releases-LOFTRAN, WCAP-7907-P-A

~

SATAN, REFLOOD and FROTH-LOCA mass 5 energy releases WCAP-10325-P-A

~

Containment response-COCO, WCAP-8327

+ Inputs Conservatively low CCW Heat Exchanger UA values used

~

Maximum of 2 ECCs assumed

~

Containment Spray pump in operation for 31 days for LOCA

TURKEY POINT NUCLEAR PLANT THERIMIALPOWER UPRATE PROJECT

+ Results

~

Peak LOCA (DEHL) containment pressure of 48.1 psig and, temperature of 273.9oF

~

Peak SLB containment pressure of 48.1 psig

~

Peak pressure/temperature is less than current analyses

~

IVlargin gained using new methodology and more rigorous accounting of heat sink input data from original analyses

+ Procedural Changes

~

Limit of one Sl pump operating during recirculation

~

Containment spray pump operated for duration (31 days) of LOCA event

~

ECC operation Only 1 ECC needed ( 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 2 ECCs credited ) 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Swing ECC will no longer auto start Swing ECC may be manually started within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />

-0

TURKEY POINT NUCLEAR PLANT THERMALPOWER UPRATE PROJECT Dose Analyses

+

Original dose analyses performed at 2300 MW~

+

Uprate dose analyses performed at 2346 MW~

+

All accidents/events reanalyzed

~

Fuel Handling

~

Cask Drop

~

Waste Gas Release

~

Steam Generator Tube Rupture Steam Line Break Rod Ejection

~

Locked Rotor 0

LOCA

+

Results clearly acceptable

TURKEY POINT NUCLEAR PLANT THERMALPOWER UPRATE PROJECT lntpact of Uprating on Postulated Accident Doses Event Current Dose Uprete Dose Fuel Handling Accident Cask Drop Waste Gas VCT GDT Steam Generator Tube Rupture Steam Line Break Rod Ejection 34.5 Rem Thy.

(1 assembly) 2,54 Rem Thy.

(1 Row) 27 Rem (157 assemblies at 1525 hours0.0177 days <br />0.424 hours <br />0.00252 weeks <br />5.802625e-4 months <br />) 1 Rem 14K Ci Xe-133 70K Ci Xe-133

(

1 Rem Thyroid

<0.1 Rem Whole body Not Calculated Not Calculated 33 Rem Thy.

(1 assembly) 2 4 Rem Thy. (1 Row) 17.7 Rem (157 assemblies at 1525 hours0.0177 days <br />0.424 hours <br />0.00252 weeks <br />5.802625e-4 months <br />)

(0.5 Rem 32K Ci Xe-133 55K Ci Xe-133 Acc Spike: 0.068 thy,0.02 whole Pre-Acc spike: 0.41 Thy.

Acc, Spike: 0.42 rem Thyroid 0.00019 rem whole Pre-acc spike: 0.52 rem thyroid 0.59 Rem Thyroid, 0.016 whole body Doses at Exclusion Boundary (EB), ICRP-30 DCFs used Thyroid doses at EB, ICRP-30 DCFs used Whole Body doses Doses at EB.

Methodology changed to SRP Doses at EB.

Methodology per SRP Doses at EB.

Methodology per SRP Locked Rotor LOCA Not Calculated 62 rem thyroid at EB, 18 rem thyroid at control room 1 Rem Thyroid, 0.099 Rem Whole Body 24 rem thyroid at EB, 15 rem thyroid at control room Doses at EB.

Methodology per SRP Iodine DCFs changed to ICRP-30 values, with lower. doses

TURKEY PO1NT NUCLEAR PLANT THERWIALPOWER UPRATE PROJECT

+

Combustible Gases Control

~

Original hydrogen generation calculation performed by hand

~

Hydrogen generation based on 2300 MW,originally

~

Post-LOCA hydrogen generation analysis performed for uprating

~

102% uprated power, 2346 MW~

~

Hydrogen concentration < 4% at 17 days

~

Hydrogen recombiner inservice thereafter to maintain < 4% hydrogen concentration

~

Recombiner capacity acceptable

TURKEY POINT NUCLEAR PLANT THERIMIALPOWER UPRATE PROJECT V. Selected Systems and Components

+ Environmental Considerations

~

Radiological and non-radiological environmental effects evaluated Operating License including Technical Specifications NPDES permit Final Environmental Statement Recapture Amendments'valuations

~

Only slight increases in discharge amounts expected

~

Radiological Normal operational exposure Minimal radiological impact Liquid effluent discharged to closed cooling canal system'ithin limits of Radiological Effluent Technical Specification Within requirements of 10CFR50 Appendix I

Accident exposure Within 10CFR100 limits

~

Non-radiological Environmental impacts evaluated in FES and Recapture Amendments remain valid NPDES permit No discharges to Biscayne Bay or Card Sound No limits on flow or temperature Estimated 0.7'F increase in outlet temperature Small heat load increase compared to 4 units and solar heating

0

TURKEY POINT NUCLEAR PLANT Z

THERIMWAL:POWER UPRATE PROJECT

+ Reactor Vessel Integrity

~

Surveillance capsule removal schedule

~

PTS Rule

~

Applicability of heatup and cooldown curves

~

Analysis Neutron fluence projections recalculated

~

Results Surveillance capsule removal schedule to be revised Will meet intent of ASTM E185-82 RT>>s values remain below PTS Rule Screening Criteria for licensed plant life Applicability of heatup and cooldown curves is 19 EFPY No new data used No new methodology used No new curves calculated Applicability scaled based on change in fluence at limiting location Generic Letter 92-01 responses unaffected by uprating

TURKEY POINT NUCLEAR PLANT THERMALPOWER UPRATE PROJECT

+ RPS/ESFAS Instrumentation

~

Setpoint uncertainties calculated Existing Westinghouse 5-column methodology used (WCAP-12745)

NRC approved for Turkey Point (8/91)

Presented in a 2 column format Plant specific calibration practices incorporated Used current plant maintenance procedures as input

~

Uncertainties used in the selection of setpoints Existing setpoints retained where possible OPbT, OTIT, Main Steam High Flow exceptions Verified adequate operating margin Retained margin to safety analysis limits

+ RTDP Parameter 'Uncertainties

~

Parameters

- Power, RCS Flow, Temperature and Pressure

~

Uncertainties calculated using plant specific practices as input Safety analyses performed

TURKEY POINT NUCLEAR PLANT THERMALPOWER UPRATE PROJECT a

+ Spent Fuel Pool Cooling Original design basis configuration One cooling loop consisting of one pump and one heat exchanger Non-safety grade

~

Installed configuration Two 100% capacity pumps powered from vital buses One heat exchanger Backup non-design basis "Emergency Pump" Current design classification Quality Group C Consistent with Regulatory Guide 1.26 Seismically qualified Two pumps are tested per ASME Section XI Current design of SFP cooling system consistent with that reviewed and approved by the NRC in 1984 Rerack Safety Evaluation.

PLA analyses consistent with SRP and UFSAR Section 14D Conservative inputs CCW supply temperature increased to 105'F Maximum allowable CCW heat exchanger fouling Decay heat loads consistent with BTP ASB 9-2 Results acceptable Results extremely conservative when compared to field data Additional analyses performed for full core offload refueling condition Analyses conclude that current administrative limit of 140'F will maintain SFP peak temperature below 150'F

TURKEY POINT NUCLEAR PLANT THERMALPOWER. UPRATE PROJECT

~

Design basis considerations Redundancy not required per UFSAR Large pool capacity Slow heatup rate Pool can accommodate boiling Four makeup water sources available Pro ca dura!ized Refueling Water Storage Tank Primary Water Storage Tank Demineralized Water Storage Tank CVCS Holdup Tank Spent fuel pit ventilation system Boiloff monitored and exhausted Not credited in dose analysis Dose analysis Assumes both spent fuel pools boiling 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> after shutdown Assumes 1% failed fuel No treatment of boiloff Unrestrained release of boiloff Results acceptable

0

TURKEY POINT NUCLEAR PLANT THERMALPOWER UPRATE PROJECT u

g2 SPENT FUEL POOL

0

TURKEY POINT NUCLEAR PLANT THERINALPOWER UPRATE PROJECT TURKEY POINT UNITS 5 5 4

SPENT FUEL POOL COOLING PUMP DIAGRAM 3A T),)

EOG 3A 480V L.C, 3C 4B I

I

,),)

EOG 48 4p STATION BLACKOUT TIE

.31 I

I I

,)

),)

I I

.gp EOG 38 4A l

,),)

I EOG 4A 480V LC. 4C Qioo XFER SWITCH (Ioo

~PENT FUEI. POOL SPENT FUEL POOL COOLING PUMP 3A COOLING PUMP 38 XFER SWITCH l00 IOO HP HP SPENT FUEL POOL SPENT FUEL POOL COOLING PUMP 4A COOLING PUMP 48

0 0=

TURKEY POINT NUCLEAR PLANT THERMALPOWER UPRATE PROJECT TURKEY POINT UNITS 3 8(

4 SPENT FUEL POOL VENTILATION AUK. T)LOO VENT STACK UNTT A

RO TA OUTSIDE AIR (2000 CT4)

SUPPLY tANS p

HEPA R/r EXHAUST FAIT RO lg SPEIIT tUEL POOL STACK UNIT S OUTSIOE AIR (1000 Cr4) rau 1

4 RP 20.000 Cf4 R/r Rjt 4

OUTSIOE AIR INLET 4

OUTSIDE AIR WLET EXIIAUST

TURKEY POINT NUCLEAR PLANT THERMALPOWER UPRATE PROJECT a

Vl. Questions / Responses A.

Emergency Preparedness and Radiation Protection Branch 1.

Pages 3 and 14 of 17 Attachment 1: Reference is made to RG 1.52 (Revision 2 issued in 1978).

Please address the use of more current standards or methods specified in more recent industry standards, such as ASME N510-1989 and ASTM D3803-1989, as opposed to the standards referenced in RG 1.52, Revision 2 for filter testing.

+ Control Room Emergency Ventilation System in Tech Specs Section 3/4.7.5

+ Section 4.7.5.c.2 identifies testing methods

~

Sample obtained per Regulatory Position C.6.b of RG 1.52, Revision 2, March 1978

~

Sample analyzed per ANSI N510-1975

+ PLA requests change in acceptance criteria (testing results)

~

Based on sample testing efficiency

~

Analytical value for removal of methyl iodide is 95% filter efficiency Tested filter efficiency proposed in PLA is 99%

~

PLA value consistent with Regulatory Guide 1.52, Revision 2

TURKEY POINT NUCLEAR PLANT THERMALPOWER UPRATE PROJECT 2.

Provide a detailed description of the fission product removal of the containment spray system and the extent to which credit is taken for the cleanup function in the analysis of the large break LOCA accident analysis (WCAP '14276, Rev. 1, pg 3-148). Listthe containment volumes not covered by the spray and the estimated forced or convective postaccident ventilation of these unsprayed volume.

+ LOCA dose analysis performed consistent with SRP

+ Single well mixed containment volume modeled

~

Blowdown

~

Containment sprays

~

Emergency containment coolers

~

Emergency containment filters

+ Sprays not credited for fission product removal

+ Elemental iodine removal up to deposition factor of 100

+ 2 ECFs operating for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after a 90 second delay

+ ECF efficiencies assumed

~'0% elemental iodine

~

95% particulate iodine

~

30% organic iodine 3.

For the LOCA analysis, at least ten or more computer codes were used in your evaluation.

Discuss briefly why so may codes were used and discuss the accuracy and veracity of the final results.

+ LOCA computer codes model separate phases of the LOCA transients as discussed on pages 11-13.

+ AllLOCAmodels applied to the uprate analyses have been previously approved for generic use on Westinghouse PWRs, except for NOTRUMP with the COSI model (SBLOCA).

TURKEY POINT NUCLEAR PLANT THERMALPOWER UPRATE PROJECT 4.

Discuss the impact of power uprate on radiolysis.

Our experience indicates that, as a result of a power uprate, the production of oxygen by radiolysis after a LOCA will increase proportionally with the power level.

Does sufficient capacity exist in the licensee combustible gas control system to accommodate this increase in oxygen production.

+ Impact of increased radiolysis addressed by uprating analysis

+ Oxygen and Hydrogen production due to radiolysis increases proportionately with power level

+ Hydrogen production specifically addressed

+ Less than 4% hydrogen concentration for 17 days

+ Hydrogen recombiner to be inservice thereafter

+ Recombiner capacity capable of maintaining Hydrogen ( 4%

5.

Briefly discuss how the higher power level effects the source terms, onsite and offsite doses, and control room habitability during normal and accident conditions.

+ Original dose analyses performed at 2300 MW,

+ Uprate dose analyses performed at 2346 MW,

+ Results presented on page 21

+ ORIGEN2 code determined uprated source terms

+ Original and uprated source terms compared

+ Total beta and gamma source strengths unchanged

+ Spectrum hardness unchan'ged

TURKEY POINT NUCLEAR PLANT THERMALPOWER UPRATE PROJECT 6.

Discuss how the radiation levels from both accident and normal operations are affected by the uprated power level.

+ Original and uprated source terms compared

+ Total beta and gamma source strengths unchanged

+ Spectrum hardness unchanged 7.

Discuss the effects of the power uprate on coolant activation products, activated corrosion products, and fissions products.

+ Technical Specification 3/4.4.8 addresses coolant specific activity limits

+ Specific activity limits unchanged

~ ~

+ More stringent than 1% failed fuel

+ Operation to continue within Technical Specification limits

TURKEY POINT NUCLEAR PLANT THERMALPOWER UPRATE PROJECT OP B.

Reactor Systems Branch 1.

Please confirm that the methodology used in the transient and accident analyses documents in WCAP-14276, Revision 1, is consistent with that used in the UFSAR.

Identify any differences and discuss their acceptability.

+ Transient and accident analyses addressed on pages 14-17 of this presentation

+ UFSAR updates due to RTDP license amendments pending

+ Current UFSAR shows pre-RTDP analyses 2.

Please confirm that only safety grade systems and components are assumed in mitigating design basis events.

+ Safety grade systems assumed to operate

+ Non-safety grade systems operate if it results in more adverse consequences

TURKEY POINT NUCLEAR PLANT THERMALPOWER UPRATE PROJECT 3.

Provide the results of the analyses for Locked Rotor/Shaft Break accident assuming a loss of off-site power coincident with the calculated radiological consequences.

Are all fuel pins with DNBR below the MDNBR assumed to fail7

+ Locked rotor analysis in Section 3.2.8.2 of Uprating Licensing Report

+ Assumes offsite power available

~

Uprating analysis consistent with original licensing basis

+ Less than 10% of the fuel pins calculated to be rods-in-DNB

+ Postulated offsite doses less than 10% of 10CFR100 guideline values

+ Rods-in-DNB assumed to fail for dose calculations 4.

Provide the results of an analysis for a postulated main feedwater line break.

+ Non-LOCA transients analyzed consistent with plant's licensing basis

+ Turkey Point licensed prior to Standard Review Plan

+ IVlain feedwater line break not a part of Turkey Point licensing basis

+ Consistent with other utilities of same vintage submittals gaining NRC approval 5.

Provide major transient curves for the reanalysis of the postulated main steam line break.

+ Event reanalyzed under RTDP project

+ Limiting initial conditions are from hot, zero power (HZP) conditions

+ RTDP NRC approved in 2/96

+ Transient curves to be included in UFSAR update after implementation

+ Analysis not changed by uprating parameters

44<~

TURKEY POINT NUCLEAR PL'ANT A

0 THERMALPOWER UPRATE PROJECT

+q i~

6.

In your analysis of a large break LOCA, for the case of minimum ECCS case, the loss of the LHSI pump is assumed as the most limitingsingle failure. Please discuss the potential loss of a diesel affecting ECCS.

+ Limiting single failure is addressed on page 12 of this presentation 7.

Discuss the most limiting single failure assumed in the SGTR analysis in light of the maximum dose release.

+ PLA analysis consistent with Turkey Point UFSAR

+ No single failures assumed

+ Calculational methods provide conservative postulated offsite doses

+ Consistent approach to that used by other utilities 8.

Please confirm that the new proposed loop design flow rate of 85,000 gpm is incorporated in the analyses documented in WCAP-14276, Rev. 1, including the loss of RCS flow and Locked Rotor/Shaft Break.

+ Thermal design flow for uprating analyses is 85,000 gpm/loop incorporated into LOCA and Non-LOCA analyses

+ Locked rotor/shaft break rods-in-DNB analysis

~

RTDP methodology event Power, pressure, temperature and flowuncertainties statistically combinedinto DNBR limit-value Minimum measured flow is direct input - 88,000 gpm/loop

+ Locked rotor/shaft break peak pressure/peak clad temperature analysis STDP methodology event

~

Power, pressure, temperature and flow uncertainties explicitly assumed

~

Thermal design'low is direct input - 85,000 gpm/loop

TURKEY POINT NUCLEAR PLANT THERMALPOWER UPRATE PROJECT 9.

Please confirm that the proposed setpoints for ESFAS are incorporated in the transient and accident analyses in WCAP-14276, Rev. 1.

+ Proposed RPS/ESFAS setpoints used in uprated accident analyses

~

Safety analysis limit values input

~

Uncertainties calculated as described on page 25 of this presentation

~

Nominal trip setpoints consistent 1

10.

Describe the steam generator tube plug level assumed in non-LOCA'vent analyses.

+ Operation up to 20% SGTP bounded

+ Worst case parameter value sets utilized for each non-LOCA event analysis

TURKEY POINT NUCLEAR PLANT THERMALPOWER UPRATE PROJECT C.

Plant Systems Branch 1.

Please address the increase in the probability of turbine overspeed and associated turbine missile production due to plant operations at the proposed uprated power level.

+ Probability of turbine overspeed and missile production is not increased Probability of overspeed based on capacity factor

~

Probability of missile production dominated by destructive overspeed failure probability Destructive overspeed failure probability independent of rating WCAP-11525

+ Turbine operating speed unchanged

- 1800 rpm

+ Redundancy in turbine valving unchanged

+ Speed control unchanged

~

Electrical tie to grid Turbine governor

~

Auxiliary governor

~

Overspeed protection controller

~

Overspeed trip mechanism

+ Uprated conditions within turbine ratings

~

Turbines periodically inspected

TURKEY POINT NUCLEAR PLANT e~

THERMALPOWER UPRATE PROJECT

+tOg+~ +4 2.

In page 5-33 of WCAP-14276, Rev.

1, Westinghouse indicated that for normal

, refueling the maximum expected SFP heat load and temperature for a 1/2 core offload at 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> after shutdown are 16.6 x 10'tu/Hr and 147'F, respectively.

In page 14D-16 of the FSAR Appendix 14D, Florida Power 8 Light Company stated that as the result of the expansion of spent fuel storage in the pool, the decay heat load for each pool increases to 16.98 x 10'tu/Hr and the corresponding pool peak transient temperature after refueling increases to less than 141'F. It is not clear why the pool with.a higher heat load (1'6.98 x 10'tu/Hr vs. 16.6 x 16'tu/Hr) would have a lower peak temperature (141'F vs. 147'F).'lease provide detailed discussion for the above discrepancy.

+ Existing UFSAR analysis, Appendix 14D

~

CCW supply temperature of 100'F Burnup and batch sizes conservatively assumed for past cycles

+ Uprate PLA analysis CCW supply temperature of 105'F 60,000 MWD/MTUburnup for future cycles'xposure Actual burnups, batch sizes and decay times used for past cycles

+ Both analyses conservative and consistent with BTP ASB 9-2

TURKEY POINT NUCLEAR PLANT THERMAL POWER UPRATE PROJECT 4%~

+

4)4~@ f 3.

For abnormal operation without SFP cooling, the time to reach boiling is 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> and the maximum boil off rate is 76.3 gpm. Assuming a loss, of SFP cooling,.provide the following information *,.

How long abnormal operation without SFP cooling, is expected to be?

What scenarios that lead to a loss of SFP cooling were considered?

What actions would be required to restore SFP cooling?"',.

1, W I I Willthere be, sufficient make-up water for',the SFP?

. Provide detail description of the make-up water sources.

How will the pool boil-off be collected/treated?

+ Active component failures considered (A or B pump)

+ Restoration completed within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> by:

Manual valve alignment

~

Powering second pump

+ Makeup water sources available

+ Pool boiloff monitored by SFP ventilation system 4.

It appears that EQ outside containment has not been addressed.

Please demonstrate what impact plant operations at the proposed uprated power level will have on EQ outside containment.

+ EQ outside containment analyzed and not impacted

~

WCAP-14276, Section 6.3.2 - radiological WCAP-14276, Section 6.4.1 - temperature, pressure and humidity 5.

It is stated on page 9.3-7 of the FSAR that the SFP cooling loop consists of a pump.

heat exchanger and associated components (i.e., filters, demineralizer, etc.) and that in the event of a failure of the SFP cooling pump, a 100% capacity spare pump is permanently piped into the SFP cooling system and is available as a standby pump.

However, in Figure 9.3-11 of the FSAR, three SFP cooling pumps are shown as part of the SFP cooling system.

Please provide a clarification for this discrepancy.

In addition, please identify the safety class and capacity for each of these pumps.

+ Addressed on pages 26-27 of this presentation

TURKEY POINT NUCLEAR PLANT THERMALPOWER UPRATE PROJECT D.

Miscellaneous'.

Section 7.4, Non-Radiological Effects, of WCAP-14276, Rev.

1, states that "Protection of:the environment is assured by compliance with permits issued by federal, state, and local agencies."

Please confirm that none 'of these permits are affected by the proposed thermal uprate and no changes to the permits are necessary fo'r the uprate.

+ Environmental permits reviewed by FPL

+ No changes to environmental permits required'

~

National Pollutant Discharge Elimination System (NPDES) ll 2.

Please discuss the maximum anticipated discharge temperature from the circulating water system during normal operation forthe uprate condition and any limits which exist on the discharge temperature..

+ Circulating water system discharges to closed cooling water canal system

~

No discharge temperature limit

~

Technical Specifications limit intake temperature

- 100'F

+ 0.7'F increase in discharge temperatures over existing plant operation Lesser impact on intake temperatures

- 0.2'F

~

Seasonal effects and solar radiation heat gain dominant

g'1

TURKEY POINT NUCLEAR PLANT THERMALPOWER UPRATE PROJECT 3.

Table ill-3,Anticipated Annual Release of Radioactive Materials in Liquid Effluents from Turkey Point Plant Units 3 &4-Reconcentration Factors for Cooling Canal System, of the FES indicates that the table values were calculated for a power level of 2200 MWt. Following a thermal uprate, will operation continue to be bounded by the values of FES Table ill-3 and all other parameters, in the., FEST

. 4 ~

+ Annual discharges will be small percentage of allowable limits"and FES estimates following uprating

+ Turkey. Point operates well within.the limits of..1.0CFR50 Appendix I for all types of releases

,pl

+ Projected doses and anticipated annual release'of '10CFR50 Appendix I are lower than projections assumed in FES

+ Operation willcontinue to be bounded by the results in FES at uprated power

e~

TURKEY POINT NUCLEAR PLANT Q

THERMALPOWER UPRATE PROJECT

+a,

~a+

ATTACHMENT1 TENTATIVEFPL PARTICIPANT LIST Tom Luke Liz Thompson Jim Hickey Leo Martin Dean Baker Jack Hoffman John DeBlasio Rick Grendys Maria Fagan Gregg Sharp lf I Q w%

~ ~,u.

Licensing Engineer (Westinghouse)

LOCA Analyst (Westinghouse)

Non-LOCA Analyst (Westinghouse)

Non-LOCA Analyst (Westinghouse)

Engineering Manager Project Engineer

~ ~

Licensing* Engineer

~I 1

Nuclear Fuels Supervisor Engineering Licensing Supervisor Mechanical Design Supervisor

TURKEY POINT NUCLEAR PLANT

~pe Op THERMAL POWER UPRATE PROJECT

+or~

ATTACHMENT2 TENTATIVENRC PARTICIPANT LIST

~Pufhiuant Eugene Imbro David Shum Duc Nguyen Hukam Garg John Minns Cheng-Ih Wu Chu-Yu Liang Larry Kopp Jim Medoff Raj Goel Rick Croteau Lambrose Lois Steve Reynolds Frank Akstulewicz

~B Project Directorate Plant Systems Electrical I &C Radiological Protection Mechanical Reactor Systems Reactor Systems Materials Containment Project Manager Reactor Systems Environmental Environmental

0

~ Rgb,

~4 C

P0 r4 P

O

$ VSce~nFalr Z NVCLEAR RMULATORYCOMMISSION March 26'996 Nr. T. F. Plunkett President - Nuclear Division Florida Po<<er and Light Co<<pany P.O.

Box 14000 Juno Beach, Florida 33408-0420

SUBJECT:

TNKEY POINT NITS 3 AN 4 - NEEYI88 ON PROPOSED LICENSE QKgyIENTS

- THERNAL POfER lJPRATE (TAC NOS N94314 AND N94315)

Dear Nr. Plunkett:

Ne<<hers of your staff plan to <<eet <<ith the NRC staff on April 4, 1996, to discuss informtion regarding technical specifications (TS) for the proposed thermal upr ate subaitted by [[letter::L-95-245, Application for Amends to Licenses DPR-31 & DPR-41,revising Definition of RTP from 2,200 Mwt to 2,300 Mwt.Proprietary TR WCAP-13719,Rev 2 & Nonproprietary TRs WCAP-13718,Rev 2 & WCAP-14276,Rev 1 Encl.Proprietary TR WCAP-13719 Withheld|letter dated December 18, 1995]].

In the <<ecting, please be prepared to address the issues described in the Enclosure to this letter.

If you have any questions, please call <<e at (301) 415-1475.

Since C.

Richard P. Croteau, Project Nanager Project Directorate II-1 Division of Reactor Projects - I/II OfFice of Nuclear Reactor Regulation Docket Nos.

50-250 and 50-251

Enclosure:

As Stated cc <</enclosure:

See next page

THERNL UPRATE QUESTION TO BE DISCUSSED A.

Emergency Preparedness and Radiation Protection Branch 2.

3.

5.

7.

Pages 3 and ll of 17 attachment 1: Reference is made to RG 1.52 (rev)sion 2 issued fn 1978).

Please address the use of sore current standards or methods specified in more recent industry standards, such as ASNE N510-1989 and ASSN 03BO3-1989, as opposed to the standards referenced.in RS 1.5Z, revision 2 for filter testing.

Provide a detailed description of the fission product removal of the containment spray system and the extent to which credit is taken for the cleanup function in the analysis of the large break LOCA accident analysis (MCAP 14276, Rev.1, pg.3-148).

List the containment volumes not covered by'he sprajy and the estimated forced or convective postaccident ventilation of these unsprayed voluaa.

For the LOCA analysis, at least ten or more computer codes were used in your evaluations.

Discuss briefly why so many codes were used and discuss the accuracy and veracity of the final results.'iscuss the impact of power uprate on radiolysis.

Our experience indicates that, as a result of a power uprate, the production of oxygen by radiolysis after a

LOCA will increase proportionally with the power level.

Does sufficient capacity exists in the licensee combustible gas control system to accommodate this increase in oxygen production.

Briefly discuss how the higher power level effects the source

terms, onsite and offsite doses, and control room habitability during normal and accident conditions.

Discuss how the radiation levels from both accident and normal, operations are affected by the uprated power level.

Discuss the effects of the power uprate on coolant activation

products, activated corrosion products, and fission products.

Reactor Systems Branch 1.

2.

Please confiril that the methodology used in the transient and accident analyses documented in RECAP-14276, Rev 1, is consistent with that used in the UFSAR.

Identify any differences and discuss their acceptability.

Please confirm that only safety grade systems and components are assumed in mitigating design basis events.

0'

3.

4.

5.

6.

7, 8.

9.

10.

Provide the results of the analyses for Locked Rotor/Shaft Break accident ass[ming a loss of off-site power coincident with the event.

Discuss the a<<ount of fuel failure during the event and the calculated radiological consequences.

Are all fuel pins with NBR below the NNBR assuned to fail.

Provide the results of an analysis for a postulated aain feed water line break.

Provide major transient curves for the reanalysis of the postulated

<<ain stea<< line break.

In your analysis of a large break LOCA, for the case of <<ini<<u<<ECCS

case, the loss of the LHSI puny is assu<<ed as the aest li<<iting single failure.

Please discuss the potential loss of a diesel affecting ECCS.

Discuss the, <<ost liaiting single failure assu<<ed fn the S6TR analysis in light of the <<aximu<<dose release.

Please confir<< that the new proposed loop design flow rate of 85000 gp<< is incorporated in the analyses documented in iKAP-14276, Rev 1, including the loss of RCS flow and Locked Rotor/Shaft Break accidents.

Please confir<< that the proposed setpoints for ESFAS are incorporated in the transient and accident analyses in MCAP-14276, Rev 1.

Describe the steam generator tube plug level assumed in Non-LOCA event analyses.

Plant Systeas Branch Please address the increase in the probability of turbine overspeed and associated turbine <<issile production due to plant operations at the proposed uprated power level.

2 ~

In paae 5-33 of itCAP-14276, Rev.

1, Mestinghouse indicated that for nor<<ai refueling the <<aximum expected SFP heat load and tenperatyre for a 1/2 core offloa at 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> after shutdown are 16.6 x 10 Btu/Hr and 147 F, respectively.

In page 140-16 of the FSAR Appendix 14 D, Florida Power and light Company stated that as the result of the expansion of spent fuel storage in the pool, the decay heat load for each pool increases to 16.98 x 10 Btu/Hr and the corresponding pool peak transient temperature after refueling increases to less than 141'F.

It is not clear why the pool with a higher heat load (16.98 x 10 Btu/Hr vs 16.$ x 10'tu/Hr) would have a lower peak tenperature (141 F vs 147"F).

Please provide detailed discussion for the above discrepancy.

3 5.

I Fer. abnormal operation without SFP cooling, the time to reach boAing is 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> and the maximum boil off rate is 76.3

GPN, Assmfng a loss of SFP cooling, provide the following information:

How long abnormal operation without SFP cooling is expected to bc?

What scenarios that load to a loss of SFP cooling were considered?

What actions would be required to restore SFP cooling?

Nll there be sufficient Nake-up water for the SFP?

Provided detail description of the make-up water sources.

How the pool boil-off will be collected/treated?

It appears that Eg outside containment has not been addressed.

Please demonstrate what impact plant operations at the proposed uprated peer level will have on Eg outside the containoent.

It is stated on page 9.3-7 of the FSAR that the SFP cooling loop consists of a pump, heat exchanger and associated components {i.e.

filters, deeincralizer, ctc.),

and that in the event of a failure of the SFP cooling pump, a lON capacity spare puwp is permanently piped into the SFP cooling system and is available as a standby pump.

However, in Figure 9.3-1I of thc FSAR, three SFP cooling pumps are shown as part of the SFP cooling system.

Please provide clarification for this discrepancy.

In addition, please identify the safety class and capacity each of these pusps.

D.

Niscellaneous Section T.l, NON-RADIOLOGICAL EFFECTS, of WCAP-l4276, Rev I, states that 'Protection of the environment is assured by compliance with permits issued by federal, state, and local agencies.'lease confirm that none of these permits are affected by the proposed thermal upratc and no changes to the permits are necessary for the uprate.

2.

3.

Please discuss the maximum anticipated discharge temperature from the circulating water system during normal operation for the uprate condition and any limits which exist on the discharge temperature.

Table III<<3, ANTICIPATED ANNUAL RELEASE OF RADIOACTIVE MATERIALS IN LIQUID EFFLUENTS FROM TURKEY POINT PLANT UNITS 3 AND 4--

RECONCENTRATION FACTORS FOR COOLING CANAL SYSTN, of the FES indicates that the table values were calculated for a power level of 2200 Stt.

Following a thermal upratc, will operation continue to be bounded by the values of FES Table III-3 <<nd all other parameters in the FES?

a 4