ML17346B171

Retran Code Transient Analysis Model Qualification.

ML17346B171
Person / Time
Site:	Saint Lucie, Turkey Point, 05000000
Issue date:	07/31/1985
From:	Hankel R, Poteralski D, Siebe A FLORIDA POWER & LIGHT CO.
To:
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ML17346B170	List: ML17346B169 ML17346B171
References
NTH-G-6, NUDOCS 8601140330
Download: ML17346B171 (228)

Text

TOPICAL REPORT RETRAN CODE TRANSIENT ANALYSIS ilODEL QUAL IF ICATION FLoRIDA POMER R LIGHT COMPANY JULY, 1985 REVIEMED BY:

A KKA ASSISTANT l'lANAGERi THERMAL HYDRAUL ICS SYSTEM ANALYSIS APPROVED BY: gC K I gCI E;/ei 0, C, POTERALS KI NANAGERi iNUCLEAR FUEL TECHNOLOGY APPROVED BY:

c'Kiack C .'Xc~.C~~

IKKK

'.IANAGERi ihiUCI EAR FUEL 860ji40330 8~0~>07>50 pDR ADOCK 050 @DR P

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DISCLAIMER OF RESPONSIBILITY This document was prepared by the Fuel Resources Department of Florida Power dc Light Company. The material therein is believed to be true and accurate to the best of our knowledge and information. However, it is authorized and intended for use and application only by Florida Power 2 Light Company.

FLORIDA POPOVER

• LIGHT COMPANY, ITS OFFICERS, DIRECTORS, AGENTS AND EMPLOYEES SHALL NOT BE RESPONSIBLE OR LIABLE FOR ANY CLAIMS, LOSSES, DAMAGES OR LIABILITIES, ~VHETHER OR NOT DUE TO OR CAUSED BY THE NEGLIGENCE OF FLORIDA POPOVER dc LIGHT COMPANY, RESULTING FROM THE USE OR MISUSE OF THIS DOCUMENT OR ANY INFORMATION CONTAINED HEREIN.

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ABSTRACT Qualification results are presented to the NRC for the purpose of gaining NRC approval for Florida Power 4 Light's plant-specific models based on the RETRAiN code for non-LOCA licensing support analysis. RETRAN analysis results, benchmarked to actual plant data, generic vendor analyses, and plant-specific FSAR analyses demonstrate the adequacy of plant modeling techniques and FPL staff proficiency as required by the NRC (NRC Generic Letter 83-11, Licensee Qualification for Performing Safety

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Analyses in Support of Licensing Activities).

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ACKNOWLEDGEMENTS The following contributed significantly to the material in this report:

3. Arpa R. Decker A. Fatemi S. Mathavan

3. Perryman

3. Ramos K. Schnoebelen R. Taboas The assistance of C. Batista, D. Beatty, 3. Collins, and L. Tymms in the typing and preparation of this report is gratefully acknowledged.

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SUMMARY

Current NRC practice requires licensees to validate the computer code models utilized for safety analysis of their nuclear plants and to demonstrate that the Licensee's Staff has the ability to set up the input, execute the code and properly interpret the results (Ref. 2). This report presents the validation of the models and methods for performing non-LOCA system safety analyses in support of licensing submittals for Florida Power 4 Light Company's four nuclear plants, Turkey Point Unit 3, Turkey Point Unit 0, St. Lucie Unit l and St. Lucie Unit 2.

FPL's system simulation is performed with the RETRAN code (Ref. 9) developed by the Electric Power Research Institute (EPRI) and approved for non-LOCA analyses by the NRC (Ref. 10). Verification of RETRAN models has been performed by benchmarking RETRAN calculations to actual plant test data, and to analyses of anticipated operational occurrences (AOOs) documented in plant-specific FSARs. Analyses are presented encompassing all four of FPL's plants as well as a range of events which include. a spectrum of non-LOCA design basis events, unusual occurrences at the plants and startup tests. The design basis events which include Main Steam Line Break, Loss of Load, Inadvertent Power Operated Relief Valve Opening, Loss of Forced Flow, Uncontrolled Rod Control Cluster Assembly Withdrawal and Control Element Assembly Drop are benchmarked against FSAR transients. Plant occurrences and tests are benchmarked against available plant data. The validation effort thus covers both licensing and

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best estimate type of analyses. As shown by a comparison of results, there is good agreement between the RETRAN analyses and the benchmarks.

Also included are Turkey Point RETRAN analysis results of three overcooling transients performed in support of the pressurized thermal shock issue resolution. These transients demonstrate FPL's capability of performing analyses to meet specific regulatory or licensing requirements.

The spectrum of analyses and benchmarks, ranging over four plants and a variety of transient events, validates the FPL, RETRAN based, system analysis models and also demonstrates the ability of FPL's staff to perform non-LOCA system safety analyses in accordance with NRC requirements and industry practices.

I CONTENTS PAGE ABSTRACT ACKNOWLEDGEMENTS

SUMMARY

iv List of Acronyms Xix 1.0 'INTRODUCTION

1.1 Background

1.2 Objective 1-2 1.3 Florida Power R Light Company's Nuclear Plants 1-3 1.0 FPL Modeling and Analysis Experience 1.5 Quality Assurance 1-6

'1.6 RETRAN Base Model Formation and Verification 1-8

'.0 INCREASE IN SECONDARY COOLANT SYSTEM HEAT REMOVAL 2-1 2.1 St. Lucie Unit 1 Natural Circulation Cooldown Event 2-2 2.1.1 Transient Description 2-2 2.1.2 RETRAN Analysis Description 2-3 2.1.3 Results 2-0 vl

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2.2 St. Lucie Unit 1 Main Steam Line Break 2-6 2.2.1 Transient Description 2-6 2.2.2 RETRAN Analysis Description 2-6 2.2.3 Results 2-7 3.0 DECREASE IN SECONDARY COOLANT SYSTEM HEAT REMOVAL 3-1 3.1 Turkey Point Unit 0 Loss of One Main Feedwater Pump Event 3-1 3.1.1 Transient Description 3-1 3.1.2 RETRAN Analysis Description 3-3 3.1.3 Results 3.2 St. Lucie Unit 1 Loss of Load 3.2.1 Transient Description 3.2.2 RETRAN Analysis Description 3.2.3 Results 3-6 3.3 St. Lucie Unit 2 Generator Trip Test 3-7 3.3.1 Transient Description 3-7 3.3.2 RETRAN Analysis Description 3-7 3.3.3 Results 3-8 3.0 St. Lucie Unit 1 MSIV Closure Event 3-9 3.0.1 Transient Description 3-9 3.0.2 RETRAN Analysis Description 3-10 3.0.3 Results 3-11

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0.0 CHANGE IN PRIMARY COOLANT SYSTEM INVENTORY-0.1 St. Lucie Unit 2 Inadvertent Opening of the PORVs O.l.l Transient Description 0.1.2 RETRAN Analysis Description 0.1.3 Results 5.0 LOSS OF REACTOR COOLANT SYSTEM FLOW 5-1 5.1 Turkey Point Pump Coastdown Test 5-1 5.1.1 Transient Description 5-1 5.1.2 RETRAN Analysis Description 5-2 5.1.3 Results 5-2 5.2 St. Lucie Unit I Loss of Forced Flow 5-3 5.2.1 Transient Description 5.2.2 RETRAN Analysis Description 5-3 5.2.3 Results 6.0 REACTIVITY INSERTION 6-1 6.1 Turkey Point Uncontrolled RCCA Withdrawal 6-1 6.1.1 Transient Descr iption 6-1 6.1.2 RETRAN Analysis Description 6-2 6.1.3 Results 6-2

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6.2 St. Lucie Unit 2 CEA Drop 6-3 6.2.1 Transient Description 6-3 6.2.2 RETRAN Analysis Description 6.2.3 Results 7.0 TURKEY POINT PRESSURIZED THERMAL SHOCK TRANSIENTS 7-1 7.1 Small Break Loss of Coolant Accident 7-2 7.1.1 Transient Description 7-2 7.1.2 RETRAN Analysis Description 7-3 7.1.3 Results 7-3 7.2'tuck Open Steam Generator Relief Valve 7-5 7.2.1 Transient Description 7-5 7.2.2 RETRAN Analysis Description 7-6 7.2.3 Results 7-6 7.3 Steam Generator Tube Rupture 7-7 7.3.1 Transient Description 7-7 7.3.2 RETRAN Analysis Description 7-8 7.3.3 Results 7-9

8.0 CONCLUSION

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9.0 REFERENCES

9-1 APPENDIX A ............ RETRAN COMPUTER CODE DESCRIPTION A-1 APPENDIX B .........-. RETRAN MODEL DESCRIPTIONS B-1 APPENDIX C ............ STAFF EXPERIENCE C-1 IX

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LIST OF TABLES TABLE DESCRIPTION PACiE 1.3.1 CHARACTERISTICS OF FPL's NUCLEAR PLANTS 1-13 1.6.1 ACTUATION OF RETRAN MODEL COMPONENTS 2.1.1 ST. LUCIE 1 RETRAN01 MODEL VOLUME DESCRIPTION 2-9 2.1.2 ST. LUCIE 1 RETRAN01 MODEL 3UNCTION DESCRIPTION 2-10 2.1.3 ST. LUCIE 1 RETRAN01 MODEL HEAT CONDUCTOR DESCRIPTION 2-12 2.1.0 INITIALAND BOUNDARY CONDITIONS, NATURAL CIRCULATION COOLDOWN EVENT 2-13 2.1.5 SAFETY SYSTEMS STATUS ASSUMED IN MODEL, NATURAL CIRCULATION COOLDOWN EVENT 2-10 INITIALCONDITIONS AND KEY PARAMETERS, MAIN STEAM LINE BREAK 2-15 2.2.2 SAFETY SYSTEMS STATUS ASSUMED IN MODEL, MAIN STEAM LINE BREAK 2-16 2.2.3 SEQUENCE OF EVENTS, MAIN STEAM LINE BREAK 2-17 3.1.1 INITIALAND BOUNDARY CONDITIONS, 3-12 LOSS OF ONE FEEDWATER PUMP 3.1.2 SAFETY SYSTEMS STATUS ASSUMED IN MODEL, 3-13 LOSS OF ONE FEEDWATER PUMP

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3.1.3 SEQUENCE OF EVENTS, LOSS OF ONE FEEDV/ATER PUMP 3.2.1 INITIALCONDITIONS AND KEY PARAMETERS, LOSS OF LOAD 3.2.2 SAFETY SYSTEMS STATUS ASSUMED IN MODEL, LOSS OF LOAD 3.2.3 SEQUENCE OF EVENTS, LOSS OF LOAD 3.3.1 INITIALCONDITIONS, GENERATOR TRIP TEST 3.3.2 SAFETY SYSTEMS STATUS ASSUMED IN MODEL, GENERATOR TRIP TEST 3.3 3 . SEQUENCE OF EVENTS, GENERATOR TRIP TEST 3.0.1 INITIALCONDITIONS, MSIV CLOSURE EVENT 3.V.2 SAFETY SYSTEMS STATUS ASSUMED IN MODEL, MSIV CLOSURE EVENT 3.0.3 SEQUENCE OF EVENTS, MSIV CLOSURE EVENT O.l.l INITIALCONDITIONS AND KEY PARAMETERS, INADVERTENT OPENING OF THE PORVS 0.1.2 SAFETY SYSTEMS STATUS ASSUMED IN MODEL, INADVERTENT OPENING OF THE PORVS

I 0.1.3 SEQUENCE OF EVENTS, INADVERTENT OPENING OF THE PORVS 5.1.1 INITIALCONDITIONS, 5-6 PUMP COASTDOWN TEST 5.1.2 SAFETY SYSTEMS STATUS ASSUMED IN MODEL, 5-7 PUMP COASTDOWN TEST 5.2.1 INITIALCONDITIONS AND KEY PARAMETERS, LOSS OF FORCED FLOW 5.2.2 SAFETY SYSTEMS STATUS ASSUMED IN MODEL, 5-9 LOSS OF FORCED FLOW 5.2.3 SEQUENCE OF EVENTS, 5-10.

LOSS OF FORCED FLOW 6.1.1 INITIALCONDITIONS AND KEY PARAMETERS, 6-5 UNCONTROLLED RCCA WITHDRAWAL 6.1.2 SAFETY SYSTEMS STATUS ASSUMED IN MODEL, 6-6 UNCONTROLLED RCCA WITHDRAWAL 6.1.3 SEQUENCE OF EVENTS, 6-7 UNCONTROLLED RCCA WITHDRAWAL 6.2.1 INITIALCONDITIONS AND KEY PARAMETERS, CEA DROP 6.2.2 SAFETY SYSTEMS STATUS ASSUMED IN MODEL~ 6-9 CEA DROP 6.2.3 COMPARISON OF KEY PARAMETERS, 6-10 CEA DROP

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7.1.1 INITIALAND BOUNDARY CONDITIONS, 7-11 SMALL BREAK LOCA 7.1.2 SAFETY SYSTEMS STATUS ASSUMED IN MODEL, 7-12 SMALL BREAK LOCA 7.1.3 SEQUENCE OF EVENTS, 7-13 SMALL BREAK LOCA 7.2.1 INITIALAND BOUNDARY CONDITIONS, 7-10 OPEN S.G. RELIEF VALVE 4

'.2.2 SAFETY SYSTEMS STATUS ASSUMED IN MODEL, 7-15 OPEN S.G. RELIEF VALVE 7.2.3 SEQUENCE OF EVENTS, 7-16 OPEN S.G. RELIEF VALVE 7.3.1 INITIALAND BOUNDARY CONDITIONS 7-17 S.G. TUBE RUPTURE 7.3.2 SAFETY SYSTEMS STA'TUS" ASSUMED IN MODEL, 7-18 S.G. TUBE RUPTURE 7.3.3 SEQUENCE OF EVENTS, 7-19 S.G. TUBE RUPTURE Bl.l TURKEY POINT RETRAN MODEL VOLUME B-15 DESCRIPTION Bl.2 TURKEY POINT RETRAN MODEL JUNCTION B-17 DESCRIPTION B2.1 ST. LUCIE UNIT 1 RETRAN MODEL VOLUME B20 DESCRIPTION B2.2 ST. LUCIE UNIT 1 RETRAN MODEL JUNCTION B-22 DESCRIPTION X111

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LIST OF FIGURES FIGURE DESCRIPTION PAGE 2.1.1 ST. LUCIE 1 RETRAN NODALIZATIONDIAGRAM 2-18 2.1.2 PRESSURIZER LEVEL, 2-19 NATURAL CIRCULATION COOLDOWN EVENT 2.1.3 PRESSURIZER PRESSURE, 2-20 NATURAL CIRCULATION COOLDOWN EVENT 2.2.1 STEAM GENERATOR PRESSURE IN AFFECTED LOOP, 2-21 MAIN STEAM LINE BREAK 2.2.2 PRESSURIZER PRESSURE, 2-22 MAIN STEAM LINE BREAK 2.2.3 'NLET CORE"COOL'ANT'TEMPER'ATURE, 2-23 MAIN STEAM LINE BREAK 2.2.0 TOTAL REACTIVITY, 2-20 MAIN STEAM LINE BREAK 2.2.5 CORE POWER, 2-25 MAIN STEAM LINE BREAK 3.1.1 AVERAGE RCS TEMPERATURE, 3-25 LOSS OF ONE FEEDWATER PUMP I

3.1.2 PRESSURIZER PRESSURE, 3-26 LOSS OF ONE FEEDWATER PUMP 3.1.3 PRESSURIZER LEVEL, 3-27 LOSS OF ONE FEEDWATER PUMP XIV

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STEAM GENERATOR PRESSURE, 3-28 LOSS OF ONE FEEDWATER PUMP 3.1.5 STEAM HEADER PRESSURE, 3-29 LOSS OF ONE FEEDWATER PUMP 3.1.6 REHEAT STEAM FLOW, 3-30 LOSS OF ONE FEEDWATER PUMP 3.2.1 PERCENT CORE POWER, 3-31 LOSS OF LOAD 3.2.2 PERCENT CORE HEAT FLUX, 3-32 LOSS OF LOAD 3.2.3 AVERAGE RCS TEMPERATURE, 3-33 LOSS OF LOAD 3.2A PRESSURIZER PRESSURE, 3-30 LOSS OF LOAD 3.3.1 STEAM GENERATOR PRESSURE, 3-35 GENERATOR TRIP TEST 3.3.2 PRESSURIZER PRESSURE, GENERATOR TRIP TEST 3.3.3 EXIT CORE COOLANT TEMPERATURE, GENERATOR TRIP TEST 3.0.1 PERCENT PRESSURIZER LEVEL, MSIV CLOSURE EVENT 3.0.2 PRESSURIZER PRESSURE, MSIV CLOSURE EVENT XV

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0.1.1 INLET CORE COOLANT TEMPERATURE, INADVERTENT OPENING OF THE PORVS 0.1.2 PRESSURIZER PRESSURE, INADVERTENT OPENING OF THE PORVS 0.1.3 STEAM GENERATOR PRESSURE, INADVERTENT OPENING OF THE PORVS 5.1.1 NORMALIZED RCS COOLANT FLOW RATE, 5-11 THREE PUMP COASTDOWN 5.1.2 NORMALIZED FLOW RATES FOR LOOPS AND CORE, 5-12 TWO PUMP COASTDOWN 5.1.3 NORMALIZED FLOW RATES FOR LOOPS AND CORE, 5-13 ONE PUMP COASTDOWN 5'.2.1 PERCENT CORE'LOW, 5-10 LOSS OF FORCED FLOW 5.2.2 PERCENT CORE POWER, 5-15 LOSS OF FORCED FLOW 5.2.3 PERCENT CORE HEAT FLUX, 5-16 LOSS OF FORCED FLOW 5.2.0 PRESSURIZER PRESSURE, 5-17 LOSS OF FORCED FLOW 6.1.1 PRESSURIZER PRESSURE, 6-11 UNCONTROLLED RCCA WITHDRAWAL" 6.1.2 PERCENT CORE POWER, 6-12 UNCONTROLLED RCCA WITHDRAWAL XVI

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6.1.3 AVERAGE CORE COOLANT TEMPERATURES 6-13 UNCONTROLLED RCCA WITHDRAWAL 6.2.1 PRESSURIZER PRESSURE, 6-10 CEA DROP 6.2.2 AVERAGE CORE COOLANT TEMPERATURE. 6-15 CEA DROP 6.2.3 PERCENT CORE POWER, 6-16 CEA DROP 6.2.0 PERCENT CORE HEAT FLUX, 6-17 CEA DROP 7.1.1 DOWNCOMER PRESSURE, 7-20 SMALL BREAK LOCA 7.1.2 DOWNCOMER COOLANT TEMPERATURE, 7-21 SMALL BREAK LOCA 7.1.3 RETRAN BREAK AND SAFETY INJECTION FLOW RATES,7-22 SMALL BREAK LOCA 7.2.1 PRESSURIZER PRESSURE, 7-23 OPEN S.G. RELIEF VALVE 7.2.2 INLET CORE COOLANT TEMPERATURE, 7-20 OPEN S.G. RELIEF VALVE XV11

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7.3.1 PRESSURIZER PRESSURE, 7-25 S.G. TUBE RUPTURE 7.3.2 INLET CORE COOLANT TEMPERATURE, 7-26 S.G. TUBE RUPTURE B.l.l TURKEY POINT RETRAN MODEL NODAL'IZATIONDIAGRAM 8.2.1 ST. LUCIE 1 RETRAN MODEL NODALIZATIONDIAGRAM XV111

LIST OF ACRONYMS AFW. AUXILIARYFEEDWATER AOO ANTICIPATED OPERATIONAL OCCURRENCE CEA CONTROL ELEMENT ASSEMBLY DNB DEPARTURE FROM NUCLEATE BOILING ECCS EMERGENCY CORE COOLING SYSTEM EPRI ELECTRIC POWER RESEARCH INSTITUTE FPL FLORIDA POWER R LIGHT COMPANY FSAR FINAL SAFETY ANALYSIS REPORT HPSI HIGH PRESSURE SAFETY IN3ECTION INPO INSTITUTE OF NUCLEAR POWER OPERATION LCO LIMITINGCONDITIONS FOR OPERATION LOAC LOSS OF OFFSITE POWER LOCA LOSS OF COOLANT ACCIDENT MFW MAIN FEEDWATER MSIV MAIN STEAM ISOLATION VALVE ~

NSSS NUCLEAR STEAM SUPPLY SYSTEM NRC NUCLEAR REGULATORY COMMISSION PORV POWER OPERATED RELIEF VALVE PTS PRESSURIZED THERMAL SHOCK QA QUALITY ASSURANCE RCCA ROD CLUSTER CONTROL ASSEMBLY RCS REACTOR COOLANT SYSTEM RCP REACTOR COOLANT PUMPS, SAFETY EVALUATIONREPORT XIX

SG STEAM GENERATOR SGTR STEAM GENERATOR TUBE RUPTURE SI SAFETY IN3ECTION SIAS SAFETY INJECTION ACTUATION SIGNAL SRV SAFETY RELIEF VALVE TM/LP THERMAL MARGIN/LOKVPRESSURE RWST REFUELING QUATER STORAGE TANK XX

1.0 INTRODUCTION

1.1 Background

Ever since FPL's first nuclear plant, Turkey Point Unit 3, came on line in 1972, and more so now with four nuclear plants in operation, there has been a requirement for reanalysis of those accidents that form the design basis of the operating plants.

Among the major sources of analytical demand have been the issue of upper head voiding for the St. Lucie Unit 1 plant, the steam generator degradation of the Turkey Point units and the pressurized thermal shock (PTS) issue that arose in connection with the Turkey Point'reactor vessels.

The increasing requirement for plant analytical support is further complicated when the fuel supplier is not the orig.nal designer of the plant, as is the case with the current fuel loadings of St. Lucie Unit 1. The models of two suppliers may differ substantially. In this case it.is desirable for the utility to have their model in place, so that they can evaluate the conservatism of the vendor's model. The plant specific model allows timely evaluation of plant modifications, design changes, and special concerns.

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I The ultimate objective for developing and maintaining the plant models in accordance with current requirements and state-of-the-art techniques is to develop full reload capability. By developing full reload capability, the analytical basis supporting plant operation, including physics parameters, fuel management, and safety analysis is contained in a unified 'methodology package.

This volume addresses FPL's qualifications for performing plant system simulation by means of the RETRAN computer code (Appendix A) which has been reviewed and approved by. the NRC for non-LOCA analysis. A subsequent report. will cover -FPL's capabilities in the area of core thermal-hydraulic analysis. The FPL lattice physics methodology has been described in a previous report (Ref. 1).'Additional'ph'ysics rep'orts-are to "be submitted at a later date.

1.2 Objective The objective of this topical report is to present, as qualifica ion basis, RETRAN base model verification results for each of FPL's nuclear plants. It is requested that these models be approved by the NRC for non-LOCA licensing support analysis. RETRAN analyses ranging over a broad spectrum of transients are benchmarked to plant data, FSAR analyses, and generic vendor il analyses to demonstrate the adequacy of these RETRAN base models for non-LOCA licensing analyses.

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These: analyses also demonstrate the proficiency of FPL personnel to perform credible system safety analyses, thereby satisfying NRC requirements documented in their letter on Qualification For Performing Safety Analyses in Support of Licensing Actions (Ref. 2).

1.3 Florida Power 8( Light Company's Nuclear Plants Florida Power 8 Light Co. operates four nuclear power plants:

Turkey Point Units 3 and 0, and St. Lucie Units 1 and 2. The Turkey Point units are geometrically similar, three-loop, Westinghouse built pressurized water reactors. Both units are licensed to operate at 2200 MW (thermal). Each primary coolant loop consists of a reactor coolant pump and a steam generator.

The reactor-core'is composed~of "157,-15XI5 fuel assemblies with- --.-

an active core height of 100 in.

The St. Lucie units are Combustion Engineering two-loop pressurized water reactors and, except for the fuel, are geometrically similar"to-each-other. Each-loop-contains.a-st~wm-generator and two reactor coolant pumps. St. Lucie Unit 1 is licensed to operate at 2700 MW (thermal). St. Lucie Unit 2 had been licensed at 2560 MW (thermal) until early 1985 when it was uprated to 2700 MW (thermal). Both reactor cores consist of 217 fuel assemblies with St. Lucie Unit 1 fuel rods arranged in a 10X10 matrix and St. Lucie Unit 2 fuel in a 16X16 matrix.

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The principal characterics of these plants are summarized in Table 1.3.1.

1.0 FPL Modeling and Analysis Experience FPL has gained considerable experience since the Mid1970's performing plant system simulations. Some of this early experience, principally based on the DYNODE code, (Ref. 12) has been documented in a topical report submitted to the NRC in 1978 (Ref. 3). The following reflects the breadth of FPL application experience:

1. RETRAN CODE VERIFICATION. Starting in 1977, FPL

'participated in the-RETRAN.01 code verification effort-at the request of the Electric Power Research Institute (EPRI). This work culminated with the documentation of the following transients benchmarked to Turkey Point data in the RETRAN Code Manual (Ref. 0): uncontrolled rod withdrawal, loss of flow, and RCS pump coastdown.

2. RESOLUTION OF SAFETY CONCERNS. Certain analyses required to meet specific regulatory or licensing requirements, such as those for the pressurized thermal shock (PTS) issue resolution, have been performed at FPL.

Overcooling transients were analyzed with RETRAN to determine thermal-hydraulic boundaryII conditions for

thermal stress and fracture mechanics computer codes as a part of FPL's effort to address the PTS concern for the Turkey Point reactor vessels. Analysis results for some of these overcooling transients, compared to similar results for generic Westinghouse plants, 'are presented in section 7.0 of this report.

Another example of this type of analysis is the development at FPL of natural circulation cooldown curves for St. Lucie Unit 1 in response to NRC requirements to prevent boiling in the upper head of the reactor vessel following the St. Lucie Unit 1 Natural Circulation Cooldown incident of June 11, 1980. The analysis was found acceptable by the NRC (Ref. 8). Validation of the RETRAN model for this application was accomplished by simulating the. incident and comparing results. to plant-recordings..The results of .this simulation are presented in Section 2.1 of this report.

3. OPERATIONAL FLEXIBILITYIMPROVEMENTS. In 1983 a study was performed at FPL to determine the optin.am pressurizer level setpoints for St. Lucie Unit 1 which would prevent uncovering the pressurizer heaters during transients.

Another example is the development of an algorithm using data calculated with RETRAN for converting measured 1-5

steam generator pressure drop into loop flow rate for St.

Lucie Unit l.

0. OPERATOR TRAINING. Best-estimate predictions of recorded plant parameters during transients have been performed to support operator training at the Turkey Point plant. An example of this type of work is a recent study to determine the time available for an operator to reduce the feedwater flow to match a turbine runback on a spurious control rod drop signal. Considered in this study were various combinations of control rod motion, availability of PORVs and steam dumps and moderator temperature coefficients.

5. VENDOR ANALYSIS 'VALUATION. Check-type-calculations have been made to assist. in. evaluating. analysis.

results generated by vendors. FPL performed analyses of limiting transients such as Loss of Load, Main Steam Line Break and Loss of Coolant Flow for St. Lucie Unit I Cycle 6 to verify vendor (Exxon Nuclear Company) calculation..

1.5 Quality Assurance The FPL Quality Assurance Program encompasses the full spectrum of activities involved in performing safety analyses.

Requirements and responsibilities for the quality program have been delineated in the FPL Quality Assurance Report (Ref. 11) 1-6

submitted to the NRC and in internal FPL QA Manuals and Procedures. Implementation of these requirements are provided in detailed Quality Instructions. The following highlights some aspects of the quality program as it relates to safety analyses.

CONFIGURATION CONTROL - FPL analyses are performed with EPRI released versions of RETRAN. Changes to the code are precluded because only the "read only" compiled listing of the code can be accessed by a user. Desired changes to the code (error corrections, etc.) are currently made by EPRI.

Proper installation of the RETRAN code on the FPL IBM Computer System is verified by running the EPRI supplied test problems. The results of the test problems are compared to validated solutions'o assure that the code. has been properly- "--"."" -"

installed.

RETRAN BASE MODEL - QA procedures were followed during the construction and verification of each plant's base model.

Changes to the base models are controlled and verified. "R .ad only" copies of the base models are available for general use.

Transient-specific changes to a base model are made by appending them to a copy of the base model.

CALCULATIONS - A calculation notebook is maintained for all safety related analyses in accordance with written instructions

, approved by the FPL QA Department. Calculations are checked 1-7

and verified by an independent reviewer and the records are maintained in accordance with established procedures. All QA related activities are audited periodically by the FPL QA Department.

1.6 RETRAN Base Model Formation and Verification Three RETRAN base models (described in Appendix B), one for Turkey Point Units 3 and 0, one for St. Lucie Unit 1, and one for St. Lucie Unit 2 have been developed and are maintained by FPL.

The development, modification, verification and maintenance of each model is in accordance with the .FPL.Quality Assurance program described in the previous section.

The RETRAN base 'models described'in-Appendix B'have'een-"" ""-"

developed as 'best estimate" models with component configuration, system interaction and operating parameters simulating actual design and operating conditions as closely as possible. The models simulate the action of safety and non-safety grade equipment and do not account for effects of instrument uncertainties or measurement errors. Best estimate values of neutronics parameters, such as moderator temperature coefficient, Doppler coefficient and scram reactivity worths and best estimate values of .heat transfer parameters, such as primary to secondary side heat transfer coefficients and clad to fuel'gap heat transfer coefficients were used. The best estimate models have been utilized for the simulation of plant tests and

plant events where RETRAN analysis results have been compared with available plant data.

For the analysis of design basis events, where comparisons have been made with FSAR accident analysis results, the best estimate models were converted into licensing models.

Conservatively bounding values were assigned to the operating parameters, such as core power, reactor coolant flow, temperatures, pressures, moderator temperature coefficients and scram reactivity worths. Only safety grade engineered safeguard systems were assumed to actuate during these events..-

For example, in= the case of= a -Loss of Load transient,-it was assumed that the turbine bypass valves do not open on demand and that the steam generator pressure builds up until the safety valves lift. The initial conditions,-key parameters and safety systems that were assumed to operate during each of the transients are tabulated in the appropriate sections.

I Based on results obtained by Northeast Utilities with their RETRAN models (Ref. 18), the effect of varying such m~del parameters as number of secondary side volumes, primary side steam generator tube noding, bubble rise velocity in the steam generator, blowdown back pressure, pump inertia and junction inertia on results mostly is minimal. Therefore, only a few sensitivity studies were conducted for the benchmark investigations reported here. Reasonable noding and parameter values were utilized. For analyses leading to actual licensing 1-9

submittals extensive sensitivity studies would be conducted to assure that results obtained are in the conservative direction.

Benchmarked RETRAN analyses of transients, divided into five major categories according to the initiating system response, are presented in the following sections.

1. INCREASE IN SECONDARY COOLANT SYSTEM HEAT REMOVAL St. Lucie Unit 1 Natural Circulation Cooldown Event St. Lucie.Unit 1 Main Steam Line Break

2. DECREASE IN"SECONDARY COOLANT 'SYSTEM HEAT REMOVAL Turkey Point Loss of One Main Feedwater Pump Event St. Lucie Unit 1 Loss of Load St. Lucie Unit 2 Generator Trip Test St. Lucie Unit 1 Main Steam Isolation Valve (MSIV)

Closure Event 1-10

3. DECREASE IN PRIMARY COOLANT SYSTEM INVENTORY St. Lucie Unit 2 Inadvertent Power Operated Relief Valve (PORV) Opening

0. LOSS OF REACTOR COOLANT SYSTEM (RCS) FLOW Turkey Point Pump Coastdown Test St. Lucie Unit 1 Loss of Forced Flow REACTIVITY INSERTION Turkey Point Uncontrolled Rod Control Cluster Assembly (RCCA) Withdrawal St. Lucie Unit 1 Control Element Assembly (CEA) Drop Also included are RETRAN analyses of three overcooling transi nts performed in support of the resolution of the pressurized thermal shock (PTS) issue for Turkey Point.

1. Turkey Point Small Break Loss of Coolant Accident (LOCA)

2. Turkey Point Stuck Open Steam Generator Relief Valve 1-11

3. Turkey Point Steam Generator Tube Rupture IVhere possible the results of these analyses are compared to those of generic analyses performed by the westinghouse Owners Group.+ These transients demonstrate FPL's ability to perform analyses to meet specific regulatory or licensing requirements.

As demonstrated by the matrix shown in Table 1.6.1, the above RETRAN analyses collectively exercise all major components of the RETRAN models.

Presented in the appendices of this report are brief descriptions of the RETRAN computer code and the RETRAN base models.

+Generic analyses are performed for a particular plant, but results are generally applicable to other plants of similar design and operational characteristics.

1-12

TABLE 1.3.1 CHARACTERISTICS OF FPL'S NUCLEAR PLANTS

'I TURKEY POINT ST. LUCIE ST. LUCIE UNITS 3 6 4 UNIT 1 UNIT 2 Core Power, MW (Thermal) 2200 2700 2560*

Nominal RCS Pressure, psia 2250 2250 2250 Nominal RCS Flowrate, Mlbm/hr 101.5 139.4 136.6 Core Inlet Temperature, 'F 547 549 549 Steam Generator Pressure, psia 840 900 815 Number of Fuel Assemblies 157 217 217 Fuel Rod Matrix 15 X 15 14 X 14 16 X 16

• CORE POWER HAS BEEN INCREASED TO 2700 MW IN 1985, BUT ANALYSES PRESENTED HERE WERE PERFORMED PREVIOUSLY, AT LOWER POWER LEVEL.

Table 1.6.1 Actuation of REIRAM Ihdel Orxponents 2.1 2.2 3.1 3.2 33 3II II 1 5o1 5o2 61 62 7+I 72 703 IIatural Ndn Steea Loss of One Loss of Generator HEIV Closurvr Opadsg of Pixp Loss of Unoontrxriled CFA Dmp Sall Ihvrak Open S.G. S.G. Turn CimQaticn Line Break Ferxhrater Prrrp Load Tlap Test Event IORVs Coastdcra Favxd Flar HXA iles& Relief Rupture Cooldown Event Test IIitIdraral Valve Prvamrher. Mel X X Pressurizer Level X X Prosaudxer Relief X Mmary Pxxssm & X X Tecpemturvr Response IIeactcr Cavr Kinetics Reactivity Feedback Scours Ikrth & Insertion Rate RCS Prep Qarecteristlcs X

~

& Flar Resistarae Reactor hetective Systra Seoordary Side Pvmsuuw

& Trxperaturvr Seocrdary SIde Presa' Rrss Relief'terxu Garcrator Level May to kcordary X Side Heat 7lvmsfer AuxQiary Feeduater Actuation X Safety Injection Decay Ihat Ihvik Flar Phase Flaw X Ikod VoidirVA X Bml Rcd Ibat Flux IISIV & Ferxhrater Lolation X 1-14

2.0 INCREASE IN SECONDARY COOLANT SYSTEM HEAT REMOVAL Events in this category result in a core coolant temperature decrease in response to an increase in heat removal by the secondary coolant system. Due to a negative moderator temperature coefficient, .

reactivity and therefore core power increases as the core inlet temperature decreases. The important modeling consideration is core reactivity change due to moderator and Doppler reactivity coefficients, and the secondary coolant system energy removal rate.

The St. Lucie Unit I Main Steam Line Break Accident (Section 2.2).

benchmarked to St. Lucie Unit I FSAR analysis results is, presented in this category.

Cooldown of the RCS-due 'to an "increase" i':secondary-coolant-system ~

heat removal may cause voiding of the fluid in the upper head of the reactor vessel if the fluid pressure is reduced below the saturation pressure. This condition occurred at St.'ucie Unit 1 on August 11, 1980 when the loss of component cooling water to the seals of the reactor coolant pumps (RCPs) necessitated tripping" the reac'or; stopping the RCPs, and attempting to bring the plant to cold shutdown by removing heat through the atmospheric dump valves (ADVs).

Stored heat in the reactor vessel structures coupled with stagnation of fluid in the upper head during the natural circulation cooldown caused the upper head fluid to be hotter than the saturatio'n temperature corresponding to the pressure in the region.

2-1

This resulted in void formation in the upper head. This event was simulated- and benchmarked to measured plant parameters (Section 2.1) to assess modeling of the reactor vessel structures, heat transfer and flow circulation in the upper head region for the St. Lucie Unit 1 RETRAN base model. This benchmark served to demonstrate the adequacy of modeling such structures should they-be. needed in any of- *- =-

FPL's RETRAN base models.

2.1 St. Lucie Unit 1 Natural Circulation Cooldown Event 2.1.1 Transient Descri tion The St. Lucie Unit 1 Natural Circulation- Cooldown Event occurred on August 11, 1980 (Ref. 6); An electrical failure caused a component cooling water isolation valve to shut off cooling water flow to the seals on all four reactor coolant pumps necessitating shutdown of the pumps to protect the pump seals.

The reactor was cooled down with the steam generator atmospheric dump system with natural circulation providing reactor coolant flow. As the reactor was cooled down, steam formation in the upper head region coupled with charging system flow injection alternating between the normal cold leg connection and pressurizer spray, caused the pressurizer level and pressure to fluctuate. This procedure was continued until the reactor vessel upper head was cooled below the saturation 2-2

temperature of the fluid in this region.

Simulation of this event challenges, in particular, RETRAN's to calculate transient responses of long duration.

V'bility This transient lasted approximately 0 hours before void formation occurred in the upper. head and 9.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> before the anomolous-.

pressurizer behavior ceased. Other. noteworthy areas of code validation that are provided by simulating this event are the code's ability to calculate upper head flow stagnation under natural circulation conditions and the onset of void formation in the upper head during-a cooldown.Accurate calculation of void- =-

formation in the upper head has led to -the development-of operating procedures which would prevent this occurrence in the future.

2.1.2 RETRAN Anal sis Descri tion FPL simulated the transient with RETRANOI to assess the modeling of the vessel upper head before using it to develop procedures for cooling down on natural circulation without upper head voiding (Ref. 8). (RETRANO1 was the only version of the code available when this analysis was performed). A simplified RETRANOI 'model was developed which was fast-running, yet detailed enough to calculate upper head fluid temperature during natural circulation, and heat transfer from reactor vessel structures. A single reactor coolant system loop was adequate to simulate the symmetric'cooldown of the RCS with the steam 2-3

generators. All manual operator actions, such as the operation of the atmospheric dump system to .cool the reactor coolant system, were modeled. The RETRAN01 model nodalization diagram is shown in Figure 2.1.1 and a description of the control volumes, junctions, and heat slabs are presented, respectively, in Tables 2.1.1, 2.1.2 and 2.1.3.

Table 2.1.0 presents the initial conditions at the time of reactor trip. Table 2.1.5 presents the status of safety systems in the RETRAN model assumed for this analysis. The analysis is a prime example of a best estimate type of analysis where. system ..

action and system parameters were modeled,to,represent actual.

operating conditions as closely as possible.

2.l.3 Results Figure 2.1.2 presents the measured and calculated pressurizer level (percent of indicated range). The calculated level agreed well with measured data for 3.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after scram. This is an indication that changes in the volumetrically averaged RCS f'uid .-

density due to fluid temperature changes were reasonably calculated. The calculated and measured pressurizer pressures are presented in Figure 2.1.3. Cooling of the'luid in the pressurizer, coupled with level changes due to shrinkage of RCS fluid as it is cooled, cause the pressure to decrease. Good agreement was obtained between the measured and calculated data.

Flow stagnation and heat transfer from structures in the reactor vessel upper head caused the cooldown of fluid in this region to be less than that in the hot legs. As a result of cooling the fluid in the pressurizer substantially below that in the upper head, void formation was calculated to occur at approximately 3.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after scram. The calculated onset. of void formation occurred approximately I/2 hour before this was indicated by data. Voiding in the upper head caused fluid in this region to be displaced and consequently the pressurizer level to increase.

The onset of voiding in the-upper head"was.calculated"to-occur--

sooner than indicated by the'data'because coolant in the upper-head remained hotter. This is affected primarily by the amount of coolant circulation in the upper head during natural circulation assumed in the model.

'ood overall agreement of trends and magnitudes of RETRAN calculation with plant data indicates that relevant phenomena leading to upper head voiding during a natural circulation cooldown were simulated. In addition, because the mendel assumptions result in calculating voiding in the upper head sooner than would actually occur in the plant, it is concluded that this model would conservatively establish operational limitations to prevent upper head voiding.

2-5

2.2 St. Lucie Unit I Main Steamline Break 2.2.1 Transient Descri tion The main steam line break (MSLB) event has been simulated for St. Lucie Unit 1 with the RETRAN02 computer code. The, results have been benchmarked to a similar calculation described in the St. Lucie Unit 1 FSAR (Ref. 5) performed with the Combustion Engineering CESEC computer code. This calculation serves to check the modeling of the reactor kinetics, the break energy removal rate'and the primary to secondary heat. transfer--

mechanism.

2.2.2 RETRAN Anal sis Descri tion This analysis was performed with the St. Lucie Unit 1 RETRAN Base Model (Appendix B). Modifications to the model were made to incorporate the break, and the end-of-cycle 0 reactor physics parameters and to achieve a conservative, licensing type, model.

Table 2.2.1 delineates the initial conditions and key parameters incorporated in the RETRAN model. The'se parameters were taken to be identical to the values given in the FSAR. The status of the plant safety systems in'he RETRAN model is summarized in Table 2.2.2. The table shows which of the modeled safety systems actuated during the transient and which did not. Systems that were not modeled for this transient are also identif ied.

2-6

The scram worth assumed that the most reactive control element assembly (CEA) was stuck fully withdrawn. The reactivity insertion due to accumulation of boron, from the FSAR analysis was input as a function of time. The failure of one high pressure safety injection (HPSI) pump was conservatively assumed in this analysis. Safety injection (Sl) flow was delayed an additional 30 seconds after receipt of the safety injection actuation signal (SIAS) to account for the time required to bring the pumps to full speed. The RCPs were manually tripped after receipt of the SIAS.

The break was modeled as a double ended guillotine rupture of -- - ~

the main steam line, located between the steam generator exit nozzle and an MSIV. This resulted in the greatest rate and magnitude of temperature reduction in the reactor core region.

The break flow was based on the Henry-Fauske/isoenthalpic expansion critical flow models with a discharge coefficient of 1.0.

2.2.3 Results Figure 2.2.1 shows the steam generator secondary pressure response. Good agreement with the steam generator pressure response presented in the FSAR was obtained. This indicated that the break mass and energy discharge rates were adequately modeled.

2-7

The pressurizer. pressure (Figure 2.2.2) was in good agreement with the FSAR results.

The core inlet coolant temperature (Figure 2.2.3) was in fairly good agreement with the FSAR results, though the cooldown was slightly underpredicted for most of the transient. Because the cooldown of the RCS for RETRAN was slightly less than that for the FSAR results, the positive reactivity insertion due to moderator temperature feedback -

effects was less and total reactivity was somewhat less (Figure 2.2A) than for the FSAR calculation. As a result, the core power (Figure 2.2.5) calculated by RETRAN was less than that in the FSAR.

The RETRAN analysis showed no return to power while the FSAR analysis showed a peak power of.about 1% at 500 seconds.

The agreement of the timing of the sequence of events between RETRAN analysis and FSAR results (Table 2.2.3) is excellent.

The RETRAN analysis was continued for 600 seconds wi,ich corresponds to the range of data presented in the FSAR.

TABLE"2.1.1 .

ST. LUCIE 1 RETRANOl MODEL VOLUME DESCRIPTION FLUID VOLUME DESCRIPTION Combined hot leg volume.

Combined steam generator inlet plenum volume.

Combined steam generator tube volume from tube sheet to top of tube bundle.

Combined steam generator tube volume from top of tube bundle to tube sheet.

Combined steam generator outlet plenum volume .

Combined cold leg- volume -upstream of reactor coolant pump.

Combined reactor coolant pump volume.

Combined cold leg volume downstream of reactor coolant=p.ump:.= ~~

Reactor vessel downcomer volume.

10 Reactor vessel inlet plenum volume.

Core volume.

16 Volume from top of active core to fuel alignment plate.

Outlet plenum volume from fuel alignment plate to upper guide structure support plate.

13 CEA shroud volume 17 Upper head volume from upper guide structure support plate to top of CEA shrouds 14 Upper head volume above top of CEA shrouds.

32 Surge line volume.

34 Pressurizer volume.

Steam generator shell side volume.

2-q

TABLE 2.1.2 ST. LUCIE 1 RETRAN01 MODEL JUNCTION DESCRIPTION PLOW JUNCTION DESCRIPTION Flow from volume 12 to volume 1.

Plow from volume 1 to volume 2.

Flow from volume 2 to volume 3.

Flow from volume 3 to volume 4.

Flow from volume 4 to volume 5.

Flow from volume 5 to volume 6.

Flow from, volume 6 to volume 7.

Flow from volume 7 to volume 8-Flow from volume 8 to volume 9.

10 Flow from-volume 9 to volume 10.

Flow from volume 10 to volume 11.

17 Flow from volume 11 to volume 16.

12 Flow from volume 16 to volume 12.

13 Flow, from volume 16 to volume 13.

Flow from volume 13 to volume 17.

15 Flow from volume 17 to volume 12.

18 Flow from volume 14 to volume 17.

35 Flow from volume 32 to volume l.

36 Flow from volume 34 to volume 32.

37 Spray flow to volume 34.

38 Charging flow to volume B.

39 Letdown flow from volume 6.

2-10

TABLE 2.1.2 (CONTINUED)

FLON JUNCTION DESCRIPTION

,81 Feedwater flow to volume 51.

82 Atmospheric relief valve flow from volume 51.

83 Steam bypass valve flow to condenser from volume 51.

84 Steam dump valve flow to condenser from volume 51.

91 Steam flow to turbine from volume 51

'-11

TABLE 2.1.3 ST. LUCIE 1 RETRANOl MODEL HEAT CONDUCTOR DESCRIPTION HEAT CONDUCTOR DESCRIPTION Fuel conductor connecting fuel to volume 11.

2 Steam generator tubes connecting volume 3 and volume 51.

Steam generator tubes connecting volume 4 and volume 51.

4 Metal in reactor vessel walls adjacent to volume 14.

Metal associated with upperhead drive shafts in volume 14.

Metal associated with CEA shrouds connecting volume 13 and volume 17.

Metal associated with CEA shrouds connecting volume 13 and volume 12.

Metal associated with CEA shrouds connecting, volume 13 and volume 12.

Upper guide structure support plate connecting volume 17 and volume 12.

10 Metal associated with upper guide structure adjacent to volume 12.

Metal in reactor vessel wall adjacent .to volume 17.

12 An effective conductor to allow axial heat conduction between volume 14 and volume 17.

2-12

TABLE 2.1.4 INITIAL AND BOUNDARY CONDITIONS NATURAL CIRCULATION COOLDOWN EVENT PARAMETER Core Power, .MW (Thermal) 2560 Core Inlet Coolant Temperature, 'F 539 Core .Outlet Coolant Temperature, 'F 588 Vessel Mass Flow Rate, 106 ibm/hr 136.8 Pressurizer Pressure, psia 2237 Steam Generator Pressure, psia 800 2-13

TABLE 2. l,. 5 SAFETY SYSTEMS STATUS ASSUMED IN MODEL NATURAL CIRCULATION- COOLDOWN EVENT'YSTEM AVAILABLE BUT NOT NOT ACTUATED ACTUATED SIMULATED Reactor Protection System (SCRAM)

Pressurizer Pressure Control System Main Steam Safety Valves Pressurizer Safety Valves Main Steam Isolation Main Feedwater Isolation Auxiliary Feedwater System Safety Injection System HPS.I Accumulators X LPSI X Atmospheric Dump Valve Systems Steam Dump and Bypass System Pressurizer Level Control System Pressurizer Power Operated Relief Valves (PORV)

Chemical and Volume Control System S. G. Level Control System X Automatic Rod Motion X 2-14

TABLE 2. 2. 1 INITIAL CONDITIONS AND KEY PARAMETERS MAIN STEAM LINE BREAK P2GD,METER VALUE Core Power, MW (Thermal)

Core Inlet Coolant Temperature, 'F 532 Pressurizer Pressure, psia 2300 Steam Generator Pressure, psia 900 Minimum CEA Worth Available at Trip, 2-1'5

TABLE 2.2.2 SAFETY SYSTEMS STATUS ASSUMED IN MODEL MAIN STEAM LINE BREAK AVAILABLE BUT NOT SYSTEM ACTUATED ACTUATED='OT SIMULATED Reactor Protection System (SCRAM)

Pressurizer Pressure Control System Main Steam Safety Valves X Pressurizer Safety Valves Main Steam Isolation Main Feedwater Isolation Auxiliary Feedwater System Safety Injection System HPSI Accumulators LPSI Atmospheric Dump Valve Systems Steam Dump and Bypass System Pressurizer Level Control ..

System Pressurizer Power Operated Relief Valves (PORV) X Chemical and Volume Control System S. G. Level Control System 2-16

TABLE 2.2.3 SEQUENCE OF EVENTS MAIN STEAM LINE BREAK TIME(S)

FSAR SETPOINT Steam Line Rupture 0.0 0.0 Low Steam Generator Press, Signal 3.8 2.8 578 psia Main Steam Isolation Signal 3.8 2.8 578 psia MSIV's Begin to Close 4.7 3.7 Trip Breakers Open 4.7 3.7 CEA's Begin to Drop 5.2 4.2 Pressurizer Empties 10.2 12. 0 MSIV's Closed 10.7 9.7 SIAS on Low RCS Press. 12.0 14.3 1578 psia RCP's Tripped 12.0 14.3 SI Actuation 42.0 44.3 MFW Isolation 86.0 86.0 AFW Initiated 180.1 ..180.1 . 253.6 ibm/sec Transient Terminated 600.0 600.0 2-17

TUR 8 I HE.

ATMOSPHERIC'RELI EF STEAM OUMP 82 TO COHOEHSER STEAM SYPASS SPRAY TO COHOENSER 051 3

36 FEEOWATER 18 81 35 05 15 O

8 12 ly 9 l~s CHARGING 09 39 LETOOWN 10 Go 0 VOLUMES

~

~ ) UNCT I ONS f7'EAT CONDUCTORS FIGURE 2.1.1 ST. LUCIE 1 RETRAN NODALIZATION DIAGRAM 2-18

i 00 P3 OAT A Rf TRANOi 0.

T I '.1'.: AFT f'R TR I P (HOURS)

FIGURE 2.1.2 PRESSURIZER LEVEL NATURAL CIRCULATION COOLDONN EVENT

2400 P3 OAT A RETRANOI 2GGO i600 Lr')

0 aJ

~r V) i200 0

8GG 40G T I."IE AFT':.R TRIP (HOURS)

FIGURE 2.1.3 PRESSURIZER PRESSURE NATURAL CIRCULATION COOLDOWM EVENT

oo CV oo o FSAR o

> RETRAN

~ 000

~l M

GO r C/}

CO oo oo A

60 120 240 300 360 420 480 540 600 TIME ( SECONDS )

FIGURE 2.2.1 STEAM GENERATOR PRESSURE IN AFFECTED LOOP MAIN STEAM LINE BREAK

O O

Ol O

O O

0 FSAR EV 0 RETRAN o

lO CQ C4 O

0 Ol C4 CQ CQ Ck O

O O

O 60 120 180 240 300 360 420 480 540 600 TIME ( SECONDS )

PIGURE 2.2.2 PRESSURIZER PRESSURE MAIN STEAM LINE BREAK

o o

oo FSAR I/l RETRAN oo o

oY) oo CV o

o 60 120 180 240 300 360 420 480 540 600 TIME { SECONDS )

PIGURE 2.2.3 INLET CORE COOLANT TEMPERATURE'AIN STEAM LINE BREAK

0 FSAR RETRAN 60 t 120 180 240 300 360 420 480 540 600 TINE ( SECONDS )

FIGURE 2.2.4 TOTAL REACTIVITY MAIN STEAM LINE BREAK

ST - LUG I E 1, CYCLE 5, MSLB BENCHMARK Q FSAR D RETRAN 60 l 20 l80 240 300 360 4 )0 540 600 TIME ( SECONDS )

FIGURE 2.2.5 CORE POWER, bRIN STEAN LINE. BREAK

3.0 DECREASE IN SECONDARY COOLANT SYSTEM HEAT REMOVAL Events in this category result in significant RCS pressure increases due to a sudden reduction in heat removal by the secondary coolant system. These pressure increases may pose a direct challenge to the RCS boundary limits. A reduction in heat removal may be a result of either an increase in steam generator secondary side pressure or a decrease in steam generator secondary side fluid inventory. Important modeling considerations are the secondary coolant system modeling, reactivity coefficients, pressurizer level and pressure control systems, and reactor protection system functions.

Four benchmark analyses are presented in this category. The Turkey Point Loss of One Main Feedwater Pump Event (Section 3.1) benchmarked to plant-.data involves, a. decrease-in";the dluid;inventory. ~ . ~~<<~-.--

of the steam generators. The St. Lucie Unit 1 Loss of Load transient (Section 3.2) benchmarked to FSAR results, St. Lucie Unit 2 Generator Trip Test (Section 3.3) benchmarked to test data, and the St. Lucie Unit 1 MSIV Closure Event (Section 3.0) benchmarked to plant data involve increased steam -generator--pressures. as-the" cause-"of'-the reduction in secondary coolant system heat removal.

3.1 Turkey Point Unit 0 Loss of One Main Feedwater Pump Event 3.1.1 Transient.Descri tion On 3une 0, 1980, while operating at full power, Turkey Point 3-1

Unit 0 lost one of two main feedwater pumps due to a maintenance problem. As per design, the turbine ran back to 70%. The core power was reduced by inserting control rods manually at 72 steps per minute. In spite of this, the reactor tripped at 52 seconds into the transient. A peak primary pressure of 2305 psia, a peak secondary pressure of 1126 psia and- -*-

a peak average primary temperature of 585 F were recorded.

After the trip, the RCS experienced excessive cooldown with the average coolant temperature reaching 525 F.

The plant Technical-Staff observed. that the steam dump to.the-condenser failed to open; -however, the atmospheric steam dump "

valves did open and one of the four steam generator safety relief valve groups also lifted. One of the four reheater steam isolation valves failed to close and continued to supply steam.to. one of ---= -*

the two high pressure feedwater heaters. The feedwater heater I

safety valve lifted at the setpoint of 065 psia and stayed open.

The auxiliary feedwater initiated on a low-low level signal after=

reactor trip and was isolated manually by the plant operator.

The plant monitoring computer failed to record the exact sequence of events and the cause of the reactor trip. In order to reconstruct the sequence of events, determine the cause of the reactor trip, confirm the safety systems actuated, and to determine the cause of the post-trip cooldown, this event was simulated with the Turkey Point RETRAN base model.

3-2

Simulation of this event and comparing the results to the available plant data serves as a basis for assessing the adequacy of modeling of the plant protective systems, and reactivity feedback.

3.1.2 RETRAN Anal sis Descri tion Simulation of this event was performed with the Turkey Point RETRAN base model (Appendix B) with modifications to explicitly account for the discharge of. reheat steam to one of the feedwater heaters-after-turbine..trip.-This. occurrence.'was - =--

simulated by- opening -a "valve on - the main. steam line,.header .

which discharges steam into a constant pressure volume at 065 psia, the setpoint pressure for the heater relief valve.

The event was intiated by tripping one of the two MFW pumps.

Steam dump to the condenser was deactivated. Reactivity due to insertion of control rods were- input via tables. The auxiliary feedwater was initiated on reactor trip and isolated after two minutes. Feedwater enthalpy was varied as recorded by 'lhe plant computer.

Initial conditions for the analysis are given in Table 3.1.1. 250 seconds of the event were simulated which corresponds to the range covered by the available plant data. The safety systems 3-3

status assumed in the RETRAN model is presented in Table 3.1.2.

3.1.3 Results Results of the RETRAN-simulation and their comparison with the plant data are presented in Figures 3.1.1 through 3.1.6. The sequence of events for the data and the RETRAN analysis is presented in Table 3.1.3. The analysis shows that the reactor tripped on low steam generator level coincident with steam/feed mismatch.. The..analysis -confirms . actuation. of..atmospheric....

steam dump valves,,one of. the four safety relief. valve groups, pressurizer spray valve, and both pressurizer relief valves. The pressurizer pressure and level, the steam generator pressure and primary loop *average-~temperature-; calculated by .RETRAN -

-~=. ~'v~

compare well with the plant data.

3.2 St. Lucie Unit 1 Loss of Load 3.2.1 Transient Descri tion The St. Lucie Unit 1 Loss of Load event has been simulated with the RETRAN02 computer code. The results have been

benchmarked to a similar calculation described in the St. Lucie Unit l FSAR (Ref. 5) performed with the CESEC computer code.

This benchmark demonstrates the technique of turning the RETRAN base model into a licensing type model and to simulate the system response to a decrease in secondary coolant system heat removal based on FSARassumptions.

The transient is initiated from full power by the sudden closure of the turbine stop valves without a simultaneous reactor trip.

After losing the secondary heat sink, the pressure in the primary and secondary side increases rapidly. A reactor trip signal-is generated on a high pressurizeryressure; <<-"

Following reactor trip, the RCS pressure continues to increase, actuating the pressurizer safety valves. The rapid RCS cooldown associated with'the reactor trip-in conjunction-with-the mass and=--

energy discharge through the safety valves reduces the RCS pressure. Likewise, the steam generator pressure is controlled by actuation of the main steam safety valves.

The average RCS coolant temperature and core inlet coolant temperature increase initially as the RCS pressure and core power increase. Following scram, the coolant temperatures decrease to hot zero power conditions, with the only energy sources the reactor coolant pumps and the core decay heat.

3.2.2 RETRAN Anal sis Descri tion The St. Lucie Unit 1 RETRAN Base Model (Appendix B) was turned into a licensing type model for this simulation. Table 3.2.1, delineates the salient initial conditions incorporated in the RETRAN model;-'The. status of the safety" systems integrated-in- - =-- -=

the RETRAN model are summarized in Table 3.2.2. The reactor physics parameters represent beginning of cycle conditions.

3.2.3 Results

. The RETRAN02= results-compares quite well with- the -FSAR.* . ~

predictions, as illustrated in Figures 3.2.1 - 3.2.0. The sequence of events is given in Table 3.2.3. The peak pressurizer pressure predicted by RETRAN is ess'entially 'the.'sa'me sas -that'of- the FSAR.

The results substantiate the modeling techniques- incorporated for predicting RCS pressure during over-pressure transients.

3-6

3.3 St. Lucie Unit 2 Generator Trip Test 3.3.1 Transient Descri tion The St. Lucie Unit 2 Generator Trip Test was performed on 3uly the:St..Lucie Unit.'2"Power-Ascension Test- ---<<

25, 1983 as part 'of Program. During the ten minutes prior to initiating the test, initial conditions were established and maintained. The test was initiated by manually tripping the generator output oil circuit breakers which caused the turbine stop valve to close while operating at 100-%-of rated power;-- There-was no operator- *- ---

intervention for two minutes. The- RETRAN02.simulation of this.....~ .,

transient served to assess the adequacy of the steam bypass and dump system, the pressurizer and the steam generator modeling.

3.3.2 RETRAN Anal sis Descri tion Simulation of the Generator Trip Test was performed with the St. Lucie Unit 2 RETRAN base model. The initial conditions are presented in Table 3.3.1. In accordance with observed beha.'ior dulling the test, the auxiliary feedwater system was forced to start immediately following reactor trip. The benchmark effort in this analysis was limited to the first 120 seconds of the transient during which no operator action was taken.

3-7

3.3.3 Results The sequence of events corresponding to the test and the RETRAN calculation appears in Table 3.3.3 and the status of the safety systems in Table 3.3.2. As a result of the turbine stop'alve closure there was -a =.rapid:=:increase..",:in- the 'secondary;-...-..--. - ~ -"-=-

pressure (Figure 3.3.1). Generally, good agreement between measured and calculated pressure responses in both steam generators was obtained. The inflection in the measured steam generator pressure response at approximately 10 seconds was not calculated because" of-small"differences-between-the actual-and'----

calculated energy removal-rates. of-.the'steam-dump and".bypass system. A maximum steam generator secondary pressure of 972 psia at 70 seconds was calculated. This compares well to the measured peak steam'generator pressure. of-953 psia:

Figure 3.3.2 depicts the calculated and measured pressurizer pressure. The initial heatup of the RCS caused by. the loss of steam generator heat removal is reflected in the momentary pressure increase following the trip. RETRAN calculated a p ak RCS pressure of 2263 psia at 2.0 seconds which, again, compares well to the measured pressure of 2262 psia at 2.0 seconds. Good agreement with data was achieved during the subsequent depressurization. A minimum pressure of 2057 psia was calculated which is in good agreement with the recorded value of 2078 psia. Following the closure of the .steam bypass valves together with the activation of the pressurizer 3-8

backup heaters, the RCS started to slowly repressurize. This trend was calculated, but the amount of repressurization calculated was less than measured.

The calculated and measured hot leg temperatures are presented in Figure 3.3.3. The hot leg temperature remains essentially constant throughout the.firstdive seconds-of..the..transient dueco;=-

the competing effects of the reactor trip and loss of steam generator load. Corresponding to the decrease in core power, together with the steam bypass action, the hot leg temperature then decreased rapidly. A maximum difference of 0 F between measured and calculate'd~emperatures-exists;-"- -"------

As good agreement with"data-was calculated-with the St. Lucie- . - ' --

Unit 2 RETRAN Base Model, this model is judged to be adequate to simulate transients, involving a decrease in secondary coolant system heat removal; 3.0 St. Lucie Unit 1 MSIV Closure Event 3.0.1 Transient Descri tion In December 1981, St. Lucie Unit 1 was operating at 2606 MEV (thermal) when the MSIV closed on steam generator "A" due to a malfunction. A few seconds later the steam generator "B" MSIY closed due to increasing steam flow. The reactor tripped on high pressurizer pressure. The pressurizer PORVs responded to the 3-9

high pressure when it reached their setpoint, by opening automatically. At approximately the time of the PORV opening, a pressurizer safety relief valve (SRV) also lifted. The RCS pressure decreased as a result of the opening of the PORVs and safety valve.

As the pressure continued to decrease below the PORV reset setpoint, the operators shut the PORV block valves and the pressure stablized at 1650 psia (just above Sl setpoint).

3.0.2 RETRAN Anal sis Descri tion The MSIV closure transient has been simulated with the RETRAN02 computer code in conjunction with'he St. Lucie 1 RETRAN base deck. The analysis was performed based upon available plant data and an -INPO report -

(Ref; 7). 'he initial conditions are presented in Table 3.0.1. It was concluded from plant data that the PORVs and pressurizer SRV opened virtually at the same time. ~Vhen the PORVs did not reset at their reset pressures, the operator closed the block valves upstream of the PORVs at 1650 psia. In order to benchmark the RETRAN results with actual transient results, the PORVs reset pressure was adjusted in the RETRAN model to correspond to the pressure at which the block valves were closed.

3-10

3.0.3 Results The results of the RETRAN calculation and plant data for pressurizer pressure and level are compared in Figures 3.0.1 and 3.0.2. The status of the safety systems is given in Table 3.0.2.

The comparison between the sequence of events is given in Table 3.0.3. The 'A'SIV closed at the start of the transient and

'B'SIV closed at 2.2 seconds because of increasing steam flow in steam generator B. The reactor tripped on high RCS pressure and PORVs reached their setpoint and opened. A pressurizer safety valve had lifted at the same time and closed when the pressure dropped to its reset value. The charging pumps were initiated according to the pressurizer level control system'and the pressurizer level rose and stabilized at 600 seconds (Figure 3.0.1). The RETRAN02 predictions have been compared to available plant transient data and agree reasonably well.

3-11

TABLE 3.1.1 INITIAL AND BOUNDARY CONDITIONS" LOSS OF ONE FEEDWATER PUMP PARAMETER VALUE Core Power, MW (Thermal) 2200 Core Inlet Coolant Temperature, 'F 546. 2 Core Outlet Coolant Temperature, 'F 602.2 Vessel Mass Flow Rate, 106 ibm/hr 101.

Pressurizer Pressure, psia 2250.

Steam Generator Pressure, psia 843 3-12

TABLE 3.1.2 SAFETY SYSTEMS STATUS ASSUMED IN MODEL LOSS OF ONE FEEDWATER PUMP AVAILABLE BUT NOT NOT SYSTEM ACTUATED ACTUATED SIMULATED Reactor Protection System (SCRAM)

Pressurizer Pressure Control System Main Steam Safety Valves Pressurizer Safety Valves Main Steam Isolation Valves Main Feedwater Isolation Valves X Auxiliary Feedwater System Safety In)'ection-System="'<'~'~

HPSI Accumulators X LPSI X Atmospheric Dump Valve Systems Steam Dump and Bypass System Pressurizer Level Control System Pressurizer Power Operated Relief Valves (PORV)

Chemical and, Volume Control System S. G. Level Control System Manual Rod Motion 3-13

TABLE 3.1.3 SEQUENCE OF EVENTS LOSS OP ONE FEEDWATER PUMP TIME(S) PAEUQCETER EVENT PLANT RETRAN PLANT RE TRAN Loss of One MFW Pump 0.0 0.0 Pressurizer Spray 13.0 2275. psia Valve Opened Pressurizer Relief 21.0 2350 psia Valve-1 Opened Pressurizer Relief 21.1 2350 psia Valve-2 Opened Pressurizer Relief 26.8 2330 psia Valve-2 Closed Pressurizer Relief 43.3 2330 psia Valve-1 Closed Reactor Tripped "

51.5 -'8.9 - "4 SG Low-"-L'evel - =-.--

of 30% With Steam/Feed Mismatch Turbine Tripped 51.9 49.9 Reactor Trip MSR/MOV Stuck Open 49.9 Valve Failure S.G. Relief Valve 52.0 1050 psia Opened

'275 Pressurizer Spray 52.0 psia.

valve closed S.G. SRV Lifted 54.2 1100 psia AFW Initiated 58.0 S.G. Low G.S. Low Level-15% Level-15%

Trip Trip S.G. SRV Reseated 68.6 1089 psia 3-14

TABLE 3.1.3 (CONTINUED)

LOSS OF ONE FEEDWATER Pl&1P TIME(S) PARAMETER PLANT RETRAN PLANT RETRAN S.G. Relief Valve 103.0 1100 psia Closed AFW Manually Termin- 105.0 Manual Manual ated MFW Isolated 135.5 124.0 Low Tave of 554'F with Reactor Trip Peak Pressurizer 34. 21.-42.. . 2345. 2350.

Pressure, psia Peak S.G. Pressure, 75. 58. 1126. 1124.

psia 3-15

TABLE 3.2.1 INITIAL CONDITIONS AND KEY P2GUQIETERS LOSS OF LOAD PARAMETER VALUE Core Power MW (Thermal) 2754 Core Inlet Coolant Temperature ('F) 551 Core Coolant Flow (106 ibm/hr) 138.3 Pressurizer Pressure, psia 2200 Steam Generator Pressure, psia 820 Moderator Temperature Coefficient (10 46k/'F) +0.5 Fuel Temperature Coeffi'cient"(10 56k/'F) -1.07 CEA Worth at Trip '(SQQ) -4.7 3-16

TABLE 3.2.2 SAFETY SYSTEMS STATUS ASSUMED IN MODEL LOSS OP LOAD AVAILABLE BUT NOT NOT SYSTEM ACTUATED ACTUATED SIMULATED Reactor Protection System (SCRAM)

Pressurizer Pressure Control System Main Steam Safety Valves Pressurizer Safety Valves Main Steam Isolation Valves Main Feedwater Isolation Valves Auxiliary Feedwater System Safety Injection..Syatem;,, -, .:.

HPSI Accumulators LPSI Atmospheric Dump Valve Systems Steam Dump and Bypass System Pressurizer Level Control System Pressurizer Power Operated Relief. Valves (PORV)

Chemical and Volume Control System S. G. Level Control System Automatic Rod Motion 3-17

TABLE 3.2.3 SEQUENCE OF EVENTS LOSS OF LOAD TIME P2GVQQ~ TER RETRAN F SAR RETRAN F SAR Loss of Secondary Load 0.0 0.0 Steam Generator SRVs Open 8.17 5.1 1010 psia High Pressurizer Pressure 8.00 7.8 2422 psia Trip Signal Pressurizer SRVs Open 9.10 9.0 2500 psia CEA's Begin to Drop Into 9.40 9.2 Core Maximum 'Pressure..zer..'u-.;

Pressure

-... 9.68 ll ~0 257'4 ps';a- =-2572 <<psia>~ --

Maximum Steam Generator 12.4 11.2 1040 psia 1057 psia Pressure Pressurizer SRVs Fully 14.8 13.5 Close 3-18

TABLE 3.3.1 INITIAL CONDITIONS GENERATOR TRIP TEST PARAMETER VALUE Core Power, MW (Thermal) 2560 Pressurizer Pressure, psia 2257.5 Steam Generator Pressure, psia 809.7 Core Inlet Coolant Temperature, 'F 539.0 Core Outlet Coolant Temperature, 'F 588.0 Pressurizer Level, 55

TABLE 3.3.2 SAFETY SYSTEMS STATUS ASSUMED IN MODEL GENERATOR TRIP TEST AVAILABLE BUT NOT NOT SYSTEM ACTUATED ACTUATED SIHULATED Reactor Protection System (SCRAM)

Pressurizer Pressure Control System Main Steam Safety Valves X Pressurizer Safety Valves Main Steam I

Isolation Valves Main Feedwater Isolation Valves Auxiliary Feedwater System Safety- Injection -Syztem~~i:.-

HPSI Accumulators X LPSI X Atmospheric Dump Valve Systems Steam Dump and Bypass System Pressurizer Level Control System Pressurizer Power Operated Relief Valves (PORV)

Chemical and Volume Control System S. G. Level Control System Automatic Rod Motion 3-20

TABLE 3.3.3 SEQUENCE OF EVENTS GENERATOR TRIP TEST EVENT TIME(S) P2GVLMETER TIME RETRAN TEST Turbine Trip 0. 0.

Turbine Stop Valves 0.22 0.22 Closed Reactor Trip Steam Dump and 0.4 0.5 On High Average RCS Bypass Activated Temperature After Reactor Trip Peak Pressurizer 2.0 2.0 2262 psia 2263 psia Pressure Peak Steam Generator 65.0 74.0 953 psia 972 psia Pressure 3-21

TABLE 3.4.1 INITIAL CONDITIONS MSIV CLOSURE -EVENT P2GQLMETER VALUE Core Power, MW (Thermal) 2646 Core Inlet Coolant Temperature, 'F -... 535 Pressurizer Pressure, psia 2250 Pressurizer Liquid Volume, ft3 743 Pressurizer Steam Volume, ft3 757 3-22

TABLE 3.4.2 SAFETY SYSTEHS STATUS ASSUHED IN HODEL MSIV CLOSURE EVENT AVAILABLE BUT NOT NOT SYSTEH ACTUATED SIHULAT Reactor Protection System (SCRAM)

Pressurizer Pressure Control System humain Steam Safety Valves Pressurizer Safety Valves Main Steam Isolation Valves h1ain Feedwater Isolation Valves Auxiliary Feedwater System Safety Injection*=System -;--" =-

HPSI Accumulators. X LPSI Atmospheric Dump Valve Systems Steam Dump and Bypass System Pressurizer Level Control System Pressurizer Power Operated Relief Valves (PORV)

Chemical and Volume Control System S. G. Level Control System Automatic Rod Motion 3-23

TABLE 3.4.3 SEQUENCE OP EVENTS MSIV CLOSURE EVENT PLANT DATA . RETRAN (SECONDS) (SECONDS)

MSIV'(A) Valve Closed 0.0 0.0 MSIV (B) Valve Closed 2.6 2.6 PORV s Opened 6.7 5.8 Pressurized Safety Valve Opened 6.7 5.8 Reactor Tripped on HI Pressurizer Press. 6.8 6.7 Turbine Tripped on Reactor Trip. 7.0 6.9 Minimum Pressurizer Pressure 112 106 (1650 psia) (1640 psia) 3-24

nn 0 DATA RETRAN n

00 0

0 LA

~o

~ v)

O 0

O O

.CU I )

n U6 40 l20 l60 200 240 280 TINE, SEC FIGURE 3.1.1 AVERAGE RCS TEMPERATURE LOSS OP ONE FEEDWATER PUMP

oo ED C1 DATA oo A RETRAN CU

(/)

oo CU a hl CL V) v) o 0 QJ Q Pu CL oo 00 40 SO '120 160 200 240 2SO I I NL 1

SLC FIGURE 3.l.2 PRESSURIZER PRESSURE LOSS OF ONE FEEDWATER PUMP

0 DATA RETRAN 40 80 I20 160 - 200 240 T I MF, SEC FIGURE 3.1.3 PRESSURIZER LEVEL LOSS OP ONE PEEDNATER PUMP

oo CU 0 DATA o

CU cl RETRAN 40 fsG I20 i 60 200 240 280 T I NL, SLC FIGURE 3.1.4 STEAM GENERATOR PRESSURE LOSS OF ONE FEEDWATER PUMP

C) o CU DATA RETHAN 40 120 l60 200 240 2SO TIME. SEC FIGURE 3.l.5 STEM@ HEADER PRESSURE l

LOSS OF ONE FEEDWATER PUMP

o o

RHTRAN o

Rg V)

CQ io

~ Y>

C) oo o

40 80 120 160 200 240. 280 TIME, SEC FIGURE 3.1.6 REHEAT STEAM FLOW LOSS OF ONE FEEDWATER PUMP

G FSAR o

o A RETRAN N

O 4 o 0

D 25 50 75 100 TINE -( SECON.DS.)

125 150 '75 200 PIGURE.3.2.1 PERCENT CORE POSER LOSS OP LOAD

0 FSAR 4 RETRAN 50 75 100 125 150 175 200 TINE- ( SECONDS )

FIGURE 3.2.2 PERCENT CORE HEAT PLUX LOSS OR LOAD

o C4 0 FSAR oo lO RETRAN l/)

CV C4 IP Q LA O

0 0

o oA If) 50 75 l00 i25 150 I75 200 TINE ( SECONDS )

PXGURE 3.2.3 AVERAGE RCS TEMPERATURE LOSS OP LOAD

(

CV o 0 FSAR Cl RETRAN o

C4 o

Cl Cl Cl Ag 25 50 75 I 00 125 50 200 TINE, t SECONDS )

FIGURE 3.2.4 PRESSURIZER PRESSURE' LOSS OF LOAD

I I 50 E DATA R""TRAi"f02 i 050 950

'D Q)

UJ 8 0

?50 6SO.

20 40 60 I v0 TtqE ( SECONDS )

FIGURE 3.3.1 STEAM GENERATOR PRESSURE GENERATOR TRIP TEST 3-35

25GG E DATA RETRAiVC2 2300 2200 O'I 00 Vl UJ Q

isco 1?'0 0 20 30 <0 .'iO 60 ?0 KG 30 1~0 i10 i)C T'N ( SECONO~ )

FIGURE 3.3.2: PRESSURIZER PRESSURE GENERATOR TRIP TEST 3-36

625 E DATA RETRAN02 STS 550 I

525 500

<0 60 so T I,IE ( SL'.CGNDS )

FIGURE 3.3.3 EXIT CORE COOLANT TEMPERATURE GENERATOR TRIP TEST 3-37

a DATA '

RETRAN 0 ~l~G l2 0 1INc. (cf:GOAD )

FIGURE 3.4.l PERCENT PRESSURIZER LEVEL HSIV CLOSURE,~ENT tl

2600 0 DATA 2417

+ RETRAN 2233 2050 1867 1683 1 500.

1 sn0 i500 1840 T 1".1f: (5[:0090".i)

FIGURE 3.4.2 PRESSURIZER PRESSURE MSIV CLOSURE EVENT

0.0 CHANGE IN PRIMARY COOLANT SYSTEM INVENTORY A decrease in RCS inventory results from a failure of the RCS pressure boundary. Reductions in coolant inventory could lead to fuel damage due to a reduction in core heat removal capability. The St.

Lucie Unit 2 Inadvertent PORV Opening transient (Section O. l) --

benchmarked to FSAR data is presented in this category. The important modeling considerations are the rate of change of coolant inventory and plant protection system functions.

O.l St. Lucie Unit 2 Inadvertent Opening of the PORVs O.l.l Transient Descri tion In this transient the two PORVs are assumed to open resulting in the loss of RCS coolant inventory. to containment,.and-.the rapid. = .- =:=

depressurization of the RCS. The RETRAN analysis is benchmarked to the corresponding St. Lucie 2 FSAR analysis (Ref. 10). The reactor trips on a low pressurizer pressure signal.

A loss of offsite power (LOAC) is assumed to occur following the turbine trip which results from the reactor scram. The reactor coolant pumps and main feedwater pumps therefore lose power and begin to coast down.

The steam bypass and steam dump valves are assumed to fail closed, such that the steam generator secondary pressure rises to

the main steam safety valve setpoints.

This transient was chosen to demonstrate the adequacy of the RETRAN non-equilibrium pressurizer model in .the analysis of postulated depressurization events. It also serves to validate the modelling of the PORVs.

0.1.2 RETRAN Anal sis Descri tion The transient was analyzed with the RETRAN02 computer code.

Initial conditions for the RETRAN analysis are presented in Table O.l.l. The status of the plant safety systems assumed in-this transient-.

are shown in Table 0.1.2.

The only significant change made to the RETRAN base model of St.

Lucie Unit 2 was to adjust the discharge coefficients associated with the PORVs in order to obtain the manufacture's maximum flow rate through them at the rated core power. This was done in order to achieve the most rapid depressurization of the RCS.

tt.l.3 Results A comparison of the sequence of events in the FSAR and RETRAN analyses is presented in Table 0.1.3. Figures 0.1.1, 0-2

0.1.2 and 0.1.3 present respectively the RCS temperature, pressurizer pressure, and steam generator secondary pressure variation for the first 120 seconds of the transient. Good agreement is seen throughout. Small differences in the pressurizer pressure and RCS temperature after 65 seconds are due to small differences in the energy removal rate through the steam generator safety valves.

TABLE 4.1.1 INITIAL CONDITIONS AND KEY PARAMETERS INADVERTENT OPENING OF THE PORVs PAR2QCETER Core Power, MW (Thermal) 2754 Core Inlet Coolant Temperature, 'F 552 Pressurizer Water Volume', ft3 61.2 Pressurizer Pressure, psia 2300 Moderator Temperature Coefficient, 10 4At/'F 2~7 CEA Worth for Trip, -5.5 4-4

TABLE 4.1.2 SAEFETY SYSTEMS STATUS ASSUMED. IN MODEL INADVERTENT OPENING OP THE PORVs AVAILABLE BUT NOT NOT SYSTEM ACTUATED ACTUATED SIMULATED Reactor Protection System (SCRAM)

Pressurizer Pressure Control System X Main Steam Safety Valves Pressurizer Safety Valves Main Steam Isolation Main Feedwater Isolation Auxiliary Feedwater System X Safety Injection System HPSI Accumulators X LPSI X Atmospheric Dump Valve Systems X I

Steam Dump and Bypass System Pressurizer Level Control System Pressurizer Power Operated Relief Valves (PORV) X Chemical and Volume Control System S. G. Level Control System Automatic Rod Motion 4-5

TABLE 4.1.3 SEQUENCE 'OF EVENTS INADVERTENT OPENING OF THE PORVs TIME PARAMETER VENDOR RETRAN FSAR RE TRAN Inadvertent. Opening of 0.0 0.0 PORVs Reactor Trip on 21.6 22.2 1799 psia 1799 psia TM/LP Signal Turbine Trip 22.8 22.5 LOAC Assumed 24.8 24.7 Main Steam SRVs Open 26.2 24.9 990 psia 990 psia SI Actuation Setpoint 29.6 28. 2 1648 psia 1648 psia Reached HPSI Pumps at Full 59.6 58.2 Speed

C)

RE TRAN 0 FSAR 20 80 60 80 l00 i 20 T I ME (SEC)

FIGURE 4.1.1 INLET CORE COOLANT TEMPERATURE INADVERTENT OPENING OP THE PORVs 4-7

Q RETRAN 0 FSAR 20 40 60 80 100 T I NE (SEC)

FIGURE 4.l.2 PRESSURIZER PRESSURE INADVERTENT OPENING OF THE PORVs 4-8

o o

CV RETRAN 0 FSAR n

o o

Q O

CQ Oc e

4 o

o o

o u~

n o

n CO 20 40 60 SO IOO i 20 T I NE (SEC)

FIGURE 4.1. 3 STEAM GENERATOR PRESSURE INADVERTENT OPENING OP THE PORVs 4-9

5.0 LOSS OF REACTOR COOLANT SYSTEM FLOW A reduction of core coolant flow rate reduces heat removal from the fuel and can result in exceeding specified acceptable fuel design limits. Transients may involve either a partial or complete loss of forced RCS flow. Important modeling considerations are reactivity coefficients, pump coastdown characteristics, and reactor protective system settings.

Analyses presented in this category are the Turkey Point Pump Coastdown Test benchmarked to plant data (Section 5.1), and the St.

Lucie Unit 1 Loss of Flow Transient benchmarked to FSAR results (Section 5.2).,

5.1 Turkey Point Pump Coastdown Test 5.1.1 Transient Descri tion During preoperational testing for Turkey Point Unit 0, a number of tests were conducted in which one, two and three reactor coolant pumps were simultaneously tripped and coolant flow measurements were made. These pump coastdown tests served to evaluate the response of the system to a total or partial loss of forced RCS flow. The benchmarking of the test results with the RETRAN model serves to validate the flow resistances (frictional and form losses) of the primary loop and the RCS pump characteristics of the model.

5-1

5.1.2 RETRAN Anal sis Descri tion The analysis was performed with RETRAN01 before the 02 version became available. The Turkey Point RETRANOl model is essentially identical to the RETRAN02 model. Best estimate parameters taken from test data and incorporated in the model are presented in Table 5.1.1. The status of safety systems in the RETRAN01 model are summarized in Table 5.1.2.

5.1.3 Results The measured and calculated fraction of total RCS flow following a trip of all three pumps is shown in Figure 5.1.1. The RETRAN values compare well with the test data which shows that the flow resistance and pump characteristics are accurately modeled. In the two-of-three-pumps-tripped analysis, the two pumps that are tripped are represented in the model by Loop B (See Appendix B) and the remaining pump that continues to run is represented by Loop A. The fraction of flow in Loop A (Figure 5.1.2) increases due to the increased loop flow resistance in Loop B. There is good agreement between RETRAN results and the test data for core and loop flow. Figure 5.1.3 shows similar curves for the one of three pumps tripped analysis. The single pump in Loop A is tripped and the pumps in Loop B (representing two pumps) continue to run. Similarly, the flow in Loop B increases due to the increase in flow resistance in Loop A as the pump in this loop coasts down.

5-2

5.2 St. Lucie Unit 1 Loss of Forced Flow 5.2.1 Transient Descri tion The St. Lucie Unit I model was benchmarked against the Loss of Coolant Flow Event described in the St. Lucie Unit I FSAR (Ref.

5). The purpose of the benchmark was to validate pump characteristics, loop hydraulic resistances, trip characteristics and reactor protection features of the St. Lucie Unit 1 RETRAN base deck. Core flow, core heat flux, low-flow trip, and pressurizer pressure are key parameters in this transient which are compared with those of the FSAR analyses. The Loss of Coolant Flow Event assumes that all four reactor coolant pumps lose their power supply and the core is deprived of the required coolant flow to remove the heat generated in the fuel. The pump flywheels provide the necessary inertia to the rotary elements for the pumps to coast down slowly and the flow to decrease gradually over a period of several seconds. As the core flow decreases, the critical heat flux decreases which can result in departure from nucleate boiling (DNB) if the reactor is not tripped. The trip occurs on low coolant flow.

5.2.2 RETRAN Anal sis Descri tion The plant's Reactor Protection System modelled in the 5-3

RETRAN02 base deck generates a reactor trip signal when the total pump flow decreases below 93% of the nominal value. A delay of 0.65 sec. is assumed between the time the signal is generated and the time the trip breakers open. This delay, together with a 0.5 sec. delay before the Control Element Assemblies (CEAs) enter the core, account for delays in signal actuation, opening time of the trip breakers and the release of the rods. The scram reactivity curve for the analysis was generated at FPL with the assumption of the most reactive CEA stuck in the fully withdrawn position. The reactivity worths of CEA banks were taken to be at their minimum allowable value.

The scram reactivity curve was obtained by combining the reactivity worths of shutdown banks minus the most reactive CEA. The fuel gap conductivity was taken at a reasonably low value (approx. 500 BTU/hr ft2 oF) to obtain conservatively high values of core heat flux. The FSAR and RETRAN calculated initial conditions and key parameters are presented in Table 5.2.1.

The charging and letdown systems, and the pressurizer heaters in the RETRAN base deck were assumed inactive for this analysis.

The status of safety systems in the RETRAN model for this transient is summarized in Table 5.2.2.

5.2.3 Results The results of the RETRAN analysis for the loss of flow

transient are shown in Figures 5.2.1 through 5.2A together with FSAR predictions. A comparative sequence of events is presented in Table 5.2.3. The RETRAN results for the key parameters such as core flow, core heat flux, time of reactor trip on low core flow signal and pressurizer pressure are in good agreement with those shown in the FSAR.

5-5

TABLE 5.1.1 INITIAL CONDITIONS PUMP COASTDOWN TEST PZQGLMHTER VALUE Core Power, MW (Thermal)

Average RCS Temperature, 'F Three of Three Pumps 550 Two of Three Pumps 543 One of Three Pumps 547 Vessel Mass Flow Rate, GPM 281000 Pressurizer Pressure, psia 2254 Steam Generator Pressure, psia 800 5-6

TABLE 5.1.2 SAFETY SYSTEMS STATUS ASSUMED IN MODEL PUMP COASTDOWN TEST AVAILABLE BUT NOT NOT SYSTEM ACTUATED ACTUATED SIMULATED Reactor Protection System (SCRAM)

Pressurizer Pressure Control System Main Steam Safety Valves Pressurizer Safety Valves Main Steam Isolation Valves Main Feedwater Isolation Valves Auxiliary Feedwater System Safety Injection System HPSI Accumulators

.LPSI Atmospheric Dump Valve Systems Steam Dump and Bypass System Pressurizer Level Control System Pressurizer Power Operated Relief Valves (PORV)

Chemical and Volume Control System S. G. Level Control System 5-7'

TABLE 5.2.1 INITIAL CONDITIONS AND KEY PARAMETERS LOSS OF FORCED FLOW VALUE Core Power, MW (Thermal) 2700.

Core Inlet Coolant Temperature, 'F 549.

Core Flow (106 ibm/hr) 138.3 Pressurizer Pressure, psia 2225 Steam Generator Pressure, psia 900 Low Flow Signal (% of nominal) 93 CEA Worth at trip, -5.60 Moderator Temperature Coefficient, 10 5Qp/'F +5.0 Doppler Temperature Coefficient, 10 5ik/'F -1.33 5-8

TABLE 5.2.2 SAFE'Y SYSTEMS STATUS ASSUMED IN MODEL.

LOSS OF FORCED FLOW AVAILABLE BUT NOT NOT SYSTEM ACTUATED SHHJLATED Reactor Protection System (SCRAM)

Pressurizer Pressure Control System Main Steam Safety Valves Pressurizer Safety Valves Main Steam Isolation Valves Main Feedwater Isolation Valves Auxiliary Feedwater System X Safety Injection System HPSI Accumulators LPSI Atmospheric Dump Valve Systems Steam Dump and Bypass System Pressurizer Level Control System Pressurizer Power Operated Relief Valves (PORV)

Chemical and Volume Control System S. G. Level Control System Automatic Rod Motion ,X 5-9

TABLE 5.2.3 SEQUENCE OF EVENTS LOSS OF FORCED FLOW EVENT TIME PARAMETER F SAR RETRAN FSAR RETRAN Loss of Power to All 0. 0'.

Four Pumps Low Flow Trip Signal 0.86 0.83 93% 93%

Trip Breakers Open 1.51 1.48 Maximum Core Power 1.91 1.97 102.5% 103.25%

CEA Begin to Drop 2.01 1.98 into Core Maximum Core Outlet 4.35 3.68 610.9 F 608.7 F Temperature Maximum Pressurizer 5.26 5.37 2326 psia 2314 psia Pressure Turbine Trip 1.69 5-10

1 ~ 0 0-8 0'

0 4~

0 DATA

' RETRAN 0 2~

0 0~

10

> I."lE (SECONDS)

FIGURE 5.1.1 NORMALIZED RCS COOLANT FLOW RATE, THREE PUMP COASTDOWN 5-11

1 ~ 2 LOOP A FLOW 1.0 CORE FLOW CD CD LOOP B 0.4 0 DATA FLOW'.2 RETRAN I

20

~ I "1c'SECOND<~)

FIGURE 5.1.2 NORMALIZED FLOW RATES FOR LOOPS AND CORE, TWO PUMP COASTDOWN 5-12

1 ~ 2 LOOP. 8 QLQW.

1.0 CORE FLGh' D 0.8 LL, s 0 6 ~

o D

I Q. 4 LOOP A FLOW 0 DATA

> RETRAN 0 2~

~

Q T la'Sr:CO~~OS)

FIGURE 5.1.3 NORMALIZED FLOW RATES FOR LOOPS AND CORES'NE PUMP COASTDOWN

C)

O 0 RETRAN CV 4 FSAR o

CO O

2 6 lo Tl.NE t SECONDS )

PIGURE 5.2.l PERCENT CORE PLOW LOSS OP PORCED PLOW

0 RETRAN FSAR 6 8 10 12 T1NE ( SEGONDS )

FIGURE 5.2.2 PERCENT CORE POWER LOSS OF FORCED FLOW

o oCV 0 RHTRAN

> FSAR o

& oe W

6 10 l2 T'1.l1E ( SEt:OMDS )

PXGURE 5.2.3 PERCENT CORE HEAT'FLUX LOSS OP PORCED FLOW

2350 2325 .RETRAN FSAR 2300 4J 2275 2250 2225 I l4 T1HE ( SECONDS )

FIGURE 5.2.4 PRESSURIZER PRESSURE LOSS "OF FORCED FLOW

6.0 REACTIVITY INSERTION Events in this category involve localized reactivity additions which cause anomalies in the core power distribution. Important modeling considerations are the reactor protection system, reactor kinetics and reactivity feedback coefficients. Analyses presented in this category are the Turkey Point Uncontroiled RC CA Withdrawal transient benchmarked to FSAR results (Section 6.1), and the St. Lucie Unit 2 CEA Drop transient benchmarked to FSAR results (Section 6.2).

6.1 Turkey Point Uncontrolled RCCA Withdrawal 6.1.1 Transient Descri tion A slow, uncontrolled rod cluster control assembly (RCCA) withdrawal transient from 10096 power was simulated with the RETRAN02 computer code and benchmarked to the analogous transient documented in the Turkey Point FSAR. (Ref. 13). In this transient the rod 'withdrawal causes an increase in core power and heat flux which result in increases in RCS temperature and pressure. Reactor trip can occur on high RCS pressure, high pressurizer level or on exceeding the high power, overpower hT or overtemperature QT setpoints. This transient assesses, the adequacy of the RETRAN reactor kinetics modeling and the modeling of the reactor protection system.'-1

6.1.2 RETRAN Anal sis Descri tion The initial conditions of the benchmark and RETRAN02 analysis, presented in Table 6.1.1, were incorporated into the Turkey Point RETRAN base model (see Appendix 8). These initial conditions represent beginning of cycle conditions as listed in the Turkey Point FSAR. The analysis was performed for a rod withdrawel rate of 2.5x10-5~%/sec. For this case the reactor trips on overtemperature hT. Presented in Table 6.1.2 is the status of safety systems included in the RETRAN simulation of this transient.

6.1.3 Results Results of the RETRAN02 calculation and the FSAR are presented in Figures 6.1.1, 6.1.2, and 6.1.3. A sequence of events for both the RETRAN calculation and the FSAR results is shown in Table 6.1.3. As the core power increases the sensed temperature difference between the hot leg and cold leg reaches the dynamic overtemperature 5T setpoint, when the scram signal It is generated and the reactor trips. The turbine trips on the reactor trip. RETRAN predicts the reactor trip about 3 seconds later than the FSAR analysis but maximum core power, maximum RCS pressure and temperature calculated for the

'RETRAN and FSAR analyses are essentially identical. Overall, the RETRAN simulation shows good agreement with the FSAR analysis results.

6-2

6.2 St. Lucie Unit 2 CEA Drop 6.2.1 Transient Descri tion The CEA drop event performed with the RETRAN02 code has been benchmarked against the CEA drop analyzed in the St.

Lucie 2 FSAR (Ref. 14).

The CEA Drop Event is defined as the inadvertent release of a single or subgroup CEAs causing it to drop into the core. The occurrence of an electrical or mechanical failure in a CEA drive mechanism would result in a CEA drop.

The CEA drop event causes an initial decrease in reactor power.

Since the heat extraction remains relatively constant, the average reactor coolant temperature decreases. The effect of the decrease in temperature in conjunction with a large negative moderator temperature coefficient, is to return the reactor to its initial power level at a slightly reduced core inlet temperature. Additionally, the core power distribution is distorted due to the CEA insertion.

The Limiting Conditions for Operation (LCO) are designed to maintain a DNB ratio sufficiently above the design limit without the necessity for a reactor trip during a CEA drop event.

6-3

Operation of the detection systems which are designed to sense a CEA drop event and to reduce turbine load to a preset value are not assumed in this analysis. The action of the protection system to inhibit CEA withdrawal during a CEA drop event has been credited; 6.2.2 RETRAN Anal sis Descri tion The single full length CEA drop transient was simulated with the RETRAN02 computer code. The assumptions and initial conditions were based on the St. Lucie Unit 2 FSAR (Ref. 10) parameters as shown in Table 6.2.1. The St. Lucie Unit 2 RETRAN base model was initialized at 9096 of rated power. The reactively worth of the dropped rod was taken to be -0.070xl0 in accordance with the FSAR analysis. The safety systems actuated during this transient are shown in Table 6.2.2.

6.2.3 Results The results of the RETRAN calculation compared, to the FSAR results are presented in Figures 6.2.1 to 6.2A. A comparison of results is shown in Table 6.2.3.

The single CEA drop results in a negative reactivity insertion, initially causing a drop in core power. Doppler and moderator feedback eventually bring the core power back to its original level. The RCS pressure and temperature decline slightly as a result of the drop in core power. The RETRAN02 results are reasonable and agree well with FSAR results.

6-0

TABLE 6.1.1 INITIAL CONDITIONS AND KEY PAEVQCETERS UNCONTROLLED RCCA WITHDRAWAL PARCE TER VALUE Core Power, MW (Thermal) 2244 Core Inlet Coolant Temperature, 'F 550.2 Core Mass Flow Rate, 106 ibm/hr 101.5 Pressurizer Pressure, psia 2220 Doppler Coefficient, 10 45k/'F .12 Moderator Temperature Coefficient, 10 45.k/'F ~ 4 Over-Temperature QT Above Nominal 6T Trip setpoint (8)

Rod Withdrawal Rate e.g/sea 2.5X10 6-5

TABLE 6.1.2 SAFETY SYSTEMS STATUS ASSUHED IN HODEL UNCONTROLLED RCCA WITHDRAWAL AVAILABLE BUT NOT NOT SYSTEH ACTUATED ACTUATED SIHULATED Reactor Protection System (SCRAM)

Pressurizer Pressure Control System Main Steam Safety Valves Pressurizer Safety Valves Main Steam Isolation Valves Hain Feedwater Isolation Valves Auxiliary Feedwater System Safety Injection System HPSI Accumulators LPSI Atmospheric Dump Valve Systems Steam Dump and Bypass System Pressurizer Level Control System Pressurizer Power Operated-Relief Valves (PORV)

Chemical and Volume Control System S. G. Level Control System Automatic Rod Motion 6-6

TABLE 6.1.3 SEQUENCE OF EVENTS UNCONTROLLED RCCA WITHDRAWAL EVENT TIME(S) PARAMETER FSAR RETRAN FSAR RE TRAN Rod Withdrawn '0 0 Reactor Tripped on 50.5 55.6 Over-Temp. gT Turbine Tripped on 56.6 Reactor Trip Maximum Core Power 51 55 113.6% 113.4%

Maximum Pressurizer 51 55 2332 psia 2322 psia Pressure Maximum Core Average 52 57 585.8 F 585.6 F Temperature 6-7

TABLE 6 . 2 . 1 INITIAL CONDITIONS AND REX PARAMETERS CEA DROP P2QUQIE TER VALUE Power Level, MW ( Thermal ) 2322 Core Inlet Coolant Temperature, 'F 550 Core Mass Plow Rate, 106 ibm/hr 139 Pressurizer Pressure, psia 2150 Pressurized Water Level, 52 Doppler Coefficient, 10 44k/'P -7.56 Moderator Temperature Coef ficient, 10 4/4/'F -4 Steam Generator Water Level, 8 of Narrow Range Top Span 70

TABLE 6.2.2 SAFETY SYSTEHS STATUS ASSUHED IN HODEL CEA DROP AVAILABLE BUT NOT NOT SYSTEH ACTUATED ACTUATED SIHULATED Reactor Protection System (SCRAM)

Pressurizer Pressure Control System Main Steam Safety Valves Pressurizer Safety Valves blain Steam Isolation Valves Main Feedwater Isolation Valves Auxiliary Feedwater System Safety Injection System HPSI X Accumulators X LPSI X Atmospheric Dump Valve Systems Steam Dump and Bypass System Pressurizer Level Control System Pressurizer Power Operated Relief Valves (PORV)

Chemical and Volume Control System S. G. Level Control System Automatic Rod Motion 6-9

TABLE 6.2.3-COMPARXSON OF KEY PARAMETERS CEA DROP PARAMETER FSAR Minimum Power, 84 Minimum RCS Pressure, psia 2126 2128 Maximum Power, 90.7 93 6-10

C4 M O V)

CC C4 RETRAN C) 0 FSAR C0 Cg o

iV) c>

(3 CJ3 20 30 40 50 T I NL. (SL C j PIGURE 6.l.l PRESSURIZER PRESSURE UNCONTROLLED RCCA WITHDRAWAL

RETRAN 0 FSAR 20 30 40 50 T I NL (SLC)

FIGURE 6.l.2 PERCENT CORE POWER UNCONTROLLED RCCA WITHDRAWAL

RETRAN 0 FSAR 30 40 50 T I ME (SI.: C)

PIGURE 6.1.3 AVERAGE CORE COOLANT TEMPERAggRE UNCONTROLLED RCCA WITHDRAWAL I

oo C4 0 asm o ~ RETRAN GO CU o(0 CV CQ C4 oo nD 20 40 60 80 lpp T I ME (SEC)

FXGURE 6.2.1 PRESSURXZER PRESSURE CEA DROP

0 PBAR Zl RETRAN 0

C3

. O 8O aCI c'a n

O~ 20 40 60 80 100 TIME (SEC)

FIGURE 6.2.2 AVERAGE CORE COOLANT TEHPBRATURB CEA DROP

0 asm h RETRM 20 40 60 .80 IOO I I NE (SEC)

RIGURE 6.2.3 PERCENT CORE POWER CEA DROP

nn 0 FSAR 6 RETRAN g

L n

20 40 60 80 100 T I NE (SEC)

FIGURE 6.2.4 PERCENT CORE HEAT FLUX CEA DROP

7.0 TURKEY POINT PRESSURIZED THERMAL SHOCK TRANSIENTS These are overcooling transients which involve a rapid and severe cooldown of fluid in the RCS with a high RCS pressure. Reduced reactor vessel fracture resistance in the beltline region due to neutron irradiation coupled with the effects of an overcooling transient may induce propagation of a flaw, thereby potentially affecting the integrity of the vessel. Several of these type transients for the Turkey Point units were simulated by FPL with the Turkey Point RETRAN02 base model. The RETRAN calculated results were compared to similar results produced by Westinghouse with four generic type plants. The comparison was to ensure that important phenomena were being represented, and that analysis assumptions and their influence on transient responses were reasonable. Three of these transients are presented in this report:

Small Break Loss of Coolant (Section 7.l)

Stuck Open Steam Generator Relief Valve (Section 7.2)

Steam Generator Tube Rupture (Section 7.3)

These transients demonstrate FPL's capability of performing analyses required to meet specific regulatory or licensing requirements.

7-1

41 7.1 Small Break Loss of Coolant Accident 7.1.1 Transient Descri tion The loss of coolant transient falls into the category of a decrease in primary system coolant inventory. The transient is characterized by a rapid cooldown and depressurization of the RCS.

The transient is initiated by a two inch diameter break, located in the hot leg. Breaks of this size tend to be limiting in terms of pressurized thermal shock. Larger breaks have more rapid depressurization as the SI flow cannot, keep up with the break flow. Smaller breaks cause less severe cooldowns because SI, if at all required, is small. After the break, a loss of subcooled fluid occurs which results in a rapid depressurization until the saturation pressure of the reactor coolant is reached. In accordance with plant emergency procedures, the charging pumps are not activated and the reactor coolant pumps are tripped manually after RCS pressure falls below 1000 psia.

Safety injection flow continues to cool the fluid entering the downcomer, while maintaining the RCS pressure. After pump coastdown, natural circulation develops in the system.

7-2

7.1.2 RETRAN Anal sis Descri tion The RETRAN model utilized the Turkey Point RETRAN Base Model (Appendix 8) to which the initial and boundary conditions (Table 7.1.1) for the small break LOCA analysis were incorporated. The slip option in RETRAN was exercised to If perform this calculation. The status of the safety systems integrated in the RETRAN model are summarized in Table 7.1.2.

The model is initialized at full power. To obtain greater cooldown, stored energy in the metal structures is neglected.

The 2 in. diameter hot leg break is modeled with a critical flow based on the Extended Henry model for subcooled flow and the lsoenthalpic Expansion model in RETRAN for two phase flow, with a discharge coefficent of 1.0.

I The maximum safety injection flow capacity is assumed to be delivered to the RCS at a temperature of 00oF. The auxiliary feedwater flow is modeled to deliver maximum capacity to each steam generator at 00oF. The Steam Dump System in the No-load Tav. control mode is modeled to relieve the pressure on the secondary side. It provides another mechanism for heat removal in addition to the break.

7.l.3 Results The results of the analysis are shown in Figures 7.1.1, 7.1.2 and 7.1.3, together with results of a generic analysis (Ref. 15) 7-3

performed by 'westinghouse for a plant similar to Turkey Point.

A sequence of events for the RETRAN analysis is provided in Table 7.1.3.

Figure 7.1.1, the reactor vessel downcomer pressure response, illustrates that the RETRAN prediction matches the generic analysis quite well. Both analyses predict a minimum pressure of approximately 600 psia at 0000 seconds.

'Figure 7.1.2, the downcomer temperature prediction, shows that RETRAN agrees well with the generic analysis. The 00oF safety injection water tends to accentuate the density gradients throughout the system which produce the fluctuating temperature response observed. At 0000 seconds, both analyses predict downcomer temperatures of approximately 230oF.

Figure 7.1.3, shows the RETRAN calculated break and total safety injection flow rates. The break and safety injection flow rates are nearly equal throughout'the transient. This maintains the RCS pressure, while cooling the fluid entering the downcomer. Only 5% of the initial fluid mass inventory is lost during the first 0000 seconds of the transient.

It can be concluded that the Turkey Point RETRAN model does reasonably well in predicting the trends and magnitudes of complicated phenomena such as a small break LOCA. The good agreement with the results of the vendor's generic analysis provides additional confidence of the validity of the model.

7-0

These RETRAN results also demonstrate that the current slip option, placed in the code to better predict two-phase phenomena, works well. The only drawback experienced with the option is the relatively high computer processing time required.

It was found that a similar calculation without slip ran about six times faster than with slip but exhibited unrealistic oscillations in the RCS pressure and temperature predictions.

7.2 Stuck Open Steam Generator Relief Valve 7.2.1 Transient Descri tion The stuck open steam generator relief valve transient is characterized by a rapid cooldown of the RCS and a rapid depr essurization until charging is actuated and a rapid repressurization occurs. The pressurizer empties but refills shortly after the charging system is actuated. -The transient is initiated by the opening of the secondary relief valve and the concurrent loss of AC power which.entails the loss of reactor coolant pumps and the startup of the auxilliary feedwater system. The RCS cools down due to energy removal through the affected steam generator. The charging system repressurizes the primary. RCS pressure is controlled b'y PORV actuation.

Natural circulation drives flow through the broken steam 7-5

generat'or for the duration of the transient. The intact loop flow stagnates after isolation of the intact steam generators.

7.2.2 RETRAN Anal sis Description The RETRAN02 calculation utilized the Turkey Point RETRAN Base Model (Appendix B) to which the initial and boundary conditions (Table 7.2.1) for the stuck open steam generator relief valve analysis were incorporated. The status of the safety systems integrated in the RETRAN model are summarized in Table 7.2.2.

In order to obtain maximum cooldown the maximum charging flow capacity is assumed to be delivered to the RCS at a temperature of 00oF. One charging pump is started at the time of Sl signal, the other two pumps are started at 10 minutes. The auxiliary feedwater flow is modeled to deliver 00oF water to each steam generator. To maximize break energy removal, no moisture carryover is assumed. The break flow, consisting of all steam, is computed with the Moody correlation. To maximize the cooldown no decay heat and no metal heat storage are assumed.

7.2.3 Results Results of the RETRAN analysis are shown in Figures 7.2.1 and 7.2.2, along with results from a generic analysis (Ref. 16) 7-6

performed by Westinghouse for a plant similar to Turkey Point.

A RETRAN sequence of events is provided in Table 7.2.3.

Figure 7.2.1, the pressurizer pressure response, shows that the RETRAN prediction of the minimum pressure as well as the rates of depressurization and repressurization agree well with the trends predicted by the generic analysis.

Figure 7.2.2, the cold leg temperature response, illustrates that RETRAN predicts the general trend of the cooldown quite well.

At 0000 seconds RETRAN conservatively underpredicts the generic analysis by approximately 30oF.

It can be concluded that the Turkey Point RETRAN model does reasonably well in predicting the trends and magnitudes of the cooldown due to secondary side depressurization. Comparisons with vendor generic results show good agreement.

7.3 Steam Generator Tube Rupture 7.3.1 Transient Descri tion The steam generator tube rupture event falls into the category of a decrease in primary coolant inventory. The transient is initiated by a guillotine rupture of a single steam generator tube slightly above the steam generator tube sheet on the cold leg side. The RCS pressure decreases as, break flow, in excess of 7-7

charging pump capacity, depletes the primary coolant inventory.

A loss of AC is assumed to occur simultaneously with a low pressurizer pressure reactor trip. Main feedwater is automatically terminated on reactor tr ip, actuating the auxiliary feedwater system.

Following the reactor trip, the rate of RCS depressurization increases, as the RCS shrinkage associated with the scram and the loss of primary coolant inventory via the ruptured tube

'continues. As the post-trip RCS cooldown subsides, safety injection and charging flow begin to refill the pressurizer, terminating the RCS depressurization.

7.3.2 RETRAN Anal sis Descri tion The RETRAN02 calculation utilized the Turkey Point RETRAN Base model (Appendix B) to which the initial and boundary conditions (Table 7.3.1) for the steam generator tube rupture analysis were incorporated. The status of the safety systems integrated in the RETRAN model are summarized in Table 7.3.2.

The tube leak was simulated by a junction with flow area

'equivalent to a double ended rupture of a single tube. The Extended-Henry-Fauske Isoenthalpic Expansion critical flow model is incorporated. The break is located slightly above the steam generator tube sheet on the cold side of the tube bundle.

This provides for the maximum rate of RCS depressurization 7-8'

prior to reactor trip.

The maximum safety injection and charging flow capacities are assumed to be delivered to the RCS at a temperature of 70oF.

The auxiliary feedwater flow is modeled to deliver maximum flow capacity to each steam generator at 70oF.

The reactor physics parameters represent beginning-of-cycle conditions.

7.3.3 Results Results of the RETRAN analysis are shown graphically in Figures 7.3.1 and 7.3.2, along with results from a generic analysis (Ref. 17) performed by Westinghouse for a 2770 MWth 3 loop plant. A RETRAN sequence of events is provided in Table 7.3.3.

Figure 7.3.1, the pressurizer pressure response, indicates that the RETRAN analysis predicts the pre-trip depressurization quite well. Both analyses agree well at time of reactor trip.

The minimum RCS pressure calculated by RETRAN is approximately 1590 psia at 500 seconds, compared to 1658 psia at 030 seconds by the generic analysis.

Figure 7.3.2, illustrates the af fected loop cold leg inlet temperature history for both analyses. RETRAN predicts the general trend of the generic analysis quite well. At 600 seconds, 7-9

RETRAN calculates a temperature of 555.6 F, agreeing well with a temperature of 556.3oF for the generic analysis.

It can be concluded that the Turkey Point RETRAN model predicted reasonably well the trends and magnitudes of the steam generator tube rupture. Comparison with generic vendor results show good agreement.

7-10

TABLE 7.1.1 INITIAL AND BOUNDARY CONDITIONS SMALL BREAK LOCA PAKQIETER Core Power, MW (Thermal) 2200 Core Inlet Coolant Temperature, 'F 547 Core Outlet Coolant Temperature, 'F 602 Vessel Mass Flow Rate, 106 ibm/hr 101 Pressurizer Pressure, psia 2250 Feedwater Flow per Steam Generator, 106 ibm/hr 3.16 Steam Generator Pressure, psia 785 Feedwater Temperature, 'F 434 Steam Generator Level,  % NR 52 7-11

TABLE 7.1.2 SAFETY SYSTEHS STATUS ASSUHED IN MODEL SMALL BREAK LOCA.

AVAILABLE BUT NOT NOT SYSTEM ACTUATED ACTUATED SIHULATED Reactor Protection System (SCRAM)

Pressurizer Pressure Control System Main Steam Safety Valves Pressurizer Safety Valves Hain Steam Isolation Valves Main Feedwater Isolation Valves Auxiliary Feedwater System Safety Injection System HPSI Accumulators X LPSI X Atmospheric Dump Valve Systems Steam Dump and Bypass System Pressurizer Level Control System Pressurizer Power Operated Relief Valves (PORV)

Chemical and Volume Control System S. G. Level Control System Automatic Rod Motion 7-12

TABLE 7.1.3 SEQUENCE OF EVENTS POR SMALL BREAK LOCA EVENTS TIME (SEC.) SETPOINT Break Reactor Trip 33 Overtemp. Q T Signal Turbine Trip 34 On Reactor Trip Steam Dump Opens 34 On Turbine Trip Safety Injection (SI) 56 Low Press. (1723 psig)

Signal Auxiliary Feedwater On SI Signal Signal Feedwater Isolation 57 On SI Signal Signal Low Primary Press. 68 1400 psig Signal Reactor Coolant Pumps 98 30 sec after 1400 psig Stopped Low Tave Signal 150 543oF Steamline Isolation 151 See Footnote Steam Dump Closes 151 On Steam Isolation Hi-Hi Steam Gen. Level 231 80% of level span Signal (Intact Loop)

Auxiliary Feedwater Off 231 Indirect on Hi-Hi level (Intact Loop) .

signal Hi-Hi Steam Generator 247 80% of level span Level Signal (Affected Loop)

Auxiliary Peedwater Off 247 Indirect on Hi-Hi level (Affected Loop) signal Footnote The MSIV trip logic in the RETRAN model is based on.a High Steam Plow Signal coincident with either low Tave (543oF) or a low steamline pressure. In this analysis a spurious High Steam Flow Signal at the time of turbine trip combined with a low Tave produced the MSIV closure at 151 seconds.

7-13

TABLE 7.2.1 INITIAL & BOUNDARY CONDITIONS OPEN S.G. RELIEF VALVE PAEULMETER Core Power, MW (Thermal)

Average RCS Coolant Temperature, 'F 547 Vessel Mass Flow Rate, 106 101 Pressurizer Pressure, psia 2250 Steam Generator Pressure, psia 1015 Steam Generator Level,  % NR 39 Charging Flow per Pump, gpm 77 7-14

TABLE 7.2.2 SAFETY SYSTEMS STATUS ASSUMED IN MODEL OPEN S.G. RELIEP VALVE AVAILABLE BUT NOT NOT SYSTEll ACTUATED ACTUATED SIMULATED Reactor. Protection System (SCRAM)

Pressurizer Pressure Control System Main Steam Safety Valves Pressurizer Safety Valves Main Steam Isolation Valves Main Feedwater Isolation Valves Auxiliary Feedwater System Safety Injection System HPSI Accumulators X LPSI X Atmospheric Dump Valve Systems Steam Dump and Bypass System X Pressurizer Level Control System Pressurizer Power Operated Relief Valves (PORV)

Chemical and Volume Control System S. G. Level Control System Automatic Rod Motion 7-15

TABLE 7.2.3 SEQUENCE OF EVENTS t OPEN S. G. RELIEF VALVE SETPOINT OR TIME EVENT VALUE REACHED (SEC.)

Break R.C. Pumps Trip Loss of AC Power Feedwater Pumps Trip Aux. Feedwater Pumps Start Reactor Trip Low Flow Turbine Trip Indirect on Reactor Trip Tav 554 .F SI Signal High Steamline b,P 20 Tav 543 F Aux. Feedwater Off Ten Minutes 600 Low Level in SG Broken SG 890 Low-Low Level in SG 1070 7-16

TABLE 7.3.1 INITIAL AND BOUNDARY CONDITIONS S.G. TUBE RUPTURE VALUE Core Power, MW (Thermal) 2200 Pressurizer Pressure, psia 2250 Core Inlet Coolant Temperature, 'F 547 Reactor Trip Pressure Setpoint, psia 1860 SI Initiation Setpoint, psia 1700 Main Feedwater Isolation, psia 1860 Auxiliary Feedwater Flowrate (3SG's), gpm 1520 RWST Temperature, 'F 70 Condensate Storage Tank Temperature, 'F 70 7-17

TABLE 7 3.2 SAEPETY SYSTELfS STATUS ASSUMED IN MODEL S.G. TUBE RUPTURE AVAILABLE BUT NOT NOT SYSTEH ACTUATED ACTUATED SIMULATED Reactor Protection System (SCRAM)

Pressurizer Pressure Control System Hain Steam Safety Valves Pressurizer Safety Valves Hain Steam Isolation Valves Main Feedwater Isolation Valves Auxiliary Feedwater System Safety Injection System HPSI Accumulators LPSI Atmospheric Dump Valve Systems Steam Dump and Bypass System Pressurizer Level Control System Pressurizer Power Operated Relief Valves (PORV)

Chemical and Volume Control System S. G. Level Control System 7-18

TABLE 7.3.3 SEQUENCE OF EVENTS STEAM GENERATOR TUBE RUPTURE EVENTS TIME (SEC.) SETPOINT Tube rupture 0.0 Reactor Trip on low Primary Pressure 187.0 1860 psia Main Feedwater Isolation Valves Begin to Close 187.0 1860 psia Loss of Offsite Power (RCF's Tripped) 188.0 Turbine Trip Turbine Throttle Valves Close 188.0 Reactor Trip Safety Injection Actuated 201.0 1700 psia Auxiliary Feedwater Delivered to Non-affected SG 367.0 Main Steamline Hi-SG level alarm Isolation 522.0 (Affected SG)

Analysis Terminated 600.0 7-19

O O

IA CV A GENERIC ANALYSIS O

O O G RETRAN CV O

O Ul LA 0

hl D

LA O

O O

O Ill 500 1500 2000 2500 3000 3500 4000 TIME ( SECONDS )

PIGURE 7.l.l DOWNCOMER PRESSURE SMALL BREAK LOCA

oo LO GENERIC ANALYSIS oo tA 0 RETRAN U

4 o o

I-K Ld Id o

oYl I-oo Al o

o~ 500 1000 1500 2000 2500 3000 3500 4000 TINE ( SECONDS )

PIGURE 7.1.2 DOWNCOMER COOLANT TEMPERATURE SMALL BREAK LOCA

o o

Ill oo o0 X

LQ 0

oo Ol BREAK FLOW o

TOTAL SAFETY INJECTION FLOW 500 1000 1500 2000 2500 3000 3500 4000 TINE ( SECONDS )

PIGURE 7.1.3 RETRAN BREAK AND SAPETY INJECTION PLOW RATES SMALL BREAK LOCA

O IA U) ~

4 4J K

Vl Vl hl O L O Q- GENERIC ANALYSIS Q- BETRAN CI CI IA 0 500 1000 1500 2000 2500 3000 3500 4000 TIME (SEC)

FIGURE 7.2.1 PRESSURIZER PRESSURE OPEN S.G. RELlEF VALVE

Cl Cl EO GENERIC ANALYSIS Cl RETRAN IA 0 Cl lK D

I LU LU I-Cl Cl 04 Cl Cl 0 500 1000 1500 2000 2500 3000 3500 4000 TIME (SEC)

PIGURE 7.2.2 INLET CORE COOLANT TEMPERATURE S.G. RELIEF VALVE

'PEN

oo Lh N

0 GENERIC ANALYSIS o RETRAN o

n'o M

9 g M

<.o IP o o

C4 60 120 180 240 300 360 420 480 540 600 T INL < 5LCONDS )

PXGURE 7.3.1 PRESSURIZER PRESSURE S.G. TUBE RUPTURE

< GENERIC ANALYSIS

> RETRAN 60 120 I 80 240 300 360 420 480 540 600 TINE ( SECONDS )

PXGURE 7.3.2 INLET CORE COOLANT TEMPERATURE S.G. TUBE RUPTURE

8.0 CONCLUSION

S Verification results were presented for the Turkey Point Units 3 and 0, St. Lucie Unit I, and St. Lucie Unit 2 RETRAN base models. These results were presented as qualification basis for receiving an SER to use these models for transient and non-LOCA accident analyses in support of licensing actions.

The following are specific conclusions based on the contents of this report:

1. The adequacy of the geometry ~ and system configuration representation in the RETRAN base models was demonstrated by good agreement of RETRAN calculated results with data from preoperational and power ascension plant tests, and off-normal events that have occurred at the plants.

2. Analysis capability for each of FPL's four. nuclear plants was shown by presenting benchmarked transient analyses for each plant.

3. Ability to perform licensing type analyses was demonstrated by benchmarking to results presented 'in FSARs. Best-estimate analysis capability was demonstrated by benchmarking to plant tests and off-normal events.

0. Ability to analyze the full spectrum of transients and non-LOCA

accidents was shown by -benchmarking to transients characterized by: a) decrease in secondary coolant system heat removal, b) increase in secondary coolant system heat removal, c) decrease in reactor coolant system inventory, d) reduction in core flowrate, and e) reactivity insertion.

5. Capability of performing analyses such as those performed in support of the Pressurized Thermal Shock issue resolution, in response to specific regulatory and licensing requirements.

8-2

9.0 REFERENCES

1. Topical Report PWR Lattice Physics Methods at Florida Power and Light Company. Letter from 3. W. Williams, 3r. to S.A.

Varga (USNRC), May 10, 1980, L-80-125

2. D. G. Eisenhut, License Qualification for Pe'rforming Safety Analyses in Support of Licensing Activities, NRC Generic Letter Number 83-11, February 8, 1983.

3. Safety and Fuel Management Analysis Methods, Volume I, Methodology For Analysis of Operational Transients, NAD1083. Letter from Robert E. Uhrig to Victor Stello (USNRC), July 10, 1978, L-78-230.

0. EPRI CCM-5 Vol. 0: Applications, "RETRAN-A Program for One-dimensional Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems" (December 1978).

5. Updated St. Lucie Unit 1 Final Safety Analysis Report, Florida Power R Light Co., Docket no. 50-335.

6. NSAC-16/INPO-2, Analysis and Evaluation of St. Lucie Unit 1 Natural Circulation Cooldown (December 1980).

7. INPO Significant Operating Experience Report 82-6, May 28, 1982.

9-1

8. Safety Evaluation for St. Lucie 1 Regarding Natural Circulation Cooldown, NRC letter from R. A. Clark to R. E.

Uhrig, April 26, 1983.

9. RETRAN-02 - A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems, EPRI NP-1850, May 1981..

10. Cecil 0. Thomas (USNRC) letter to Dr. Thomas W. Schnatz, Acceptance for Referencing of Licensing Topical Reports EPRI CCM-5, RETRAN-A Program for One-Dimensional Transient Thermal Hydraulic Analysis of Complex Fluid Flow Systems, and EPRI NP-1850-CCM, RETRAN-02-A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems, September 0, 1980.

11. Topical Quality Assurance Report, FPL-NQA-lOOA, Rev. 6, 3une 10, 1983.

12. R. C. Kern and D. Hodges, DYNODE-P, Version 2: A Nuclear Steam Supply System Transient Simulator for Pressurized Water Reactors-User Manual, NAI-76-67, Rev. 3, March 25, 1977.

13. Updated Final Safety Analysis Report Turkey Point Plant Units 3 R 0, Docket Nos. 50-250 and 50-251.

10. St. Lucie Plant Unit 2 Final Safety Analysis Report, Docket No. 50-389.

9-2

15. T. A. Meyer, Summary Report on Reactor Vessel Integrity for Westinghouse Operating Plants,,WCAP-10019, December, 1981

16. Westinghouse Owners Group Report on Steamline Break Analysis for Pressurized Thermal Shock Evaluation of Reactor Vessels, February, 1983.

17. A. C. Cheung, et. al., A Generic Assessment of Significant Flow Extension, Including Stagnant Loop Conditions, From Pressurized Thermal Shock of Reactor Vessels on Westinghouse Nuclear Power Plants,. WCAP-10319, December, 1983.

18. NUSCO Thermal Hydraulic Model Qualification, Volume I (RETRAN),

NUSCO 100-1, Northeast Utilities Services Company, August 1, 1980.

9-3

APPENDIX A RETRAN COMPUTER CODE DESCRIPTION

RETRAN Computer Code Description The RETRAN computer code was developed under the Electric Power Research Institute (EPRI) sponsorship by Energy, Inc. The code is an offshoot of the RELAP code and provided the utility industry with a, versatile and reliable thermal-hydraulic code for the analysis of light water reactor systems. An ongoing code development effort since 1975 has produced two major versions of the code, RETRAN01 and RETRAN02. Both.

versions of the code have been extensively validated and qualified. The RETRAN01/MOD003 and RETRAN02/MOD002 codes have been reviewed by the Nuclear Regulatory Commission with the assistance of the Argonne National Laboratory, and approved for non-LOCA transient safety analysis application (Ref. A-1). Both versions of RETRAN have been installed on the FPL computer system. Currently, all analyses at FPL are performed with the RETRAN02 code.

The main features of RETRAN01 are:

1. A one-dimensional, homogeneous equilibrium mixture thermal-hydraulic model for the reactor cooling system.

2. A point neutron kinetics model for the reactor core.

3. Special auxiliary or component models (nonequilibrium pressurizer, temperature transport delay).

A-2

0. Control system models.

5. A consistent steady-state initialization technique.

6. A model for steam separators based on separator efficiency curves.

7. A local conditions heat transfer model.

In order to remove some of the limitations of RETRAN01 and extend its capability the RETRAN02 code was released in May, 1981. New models added to the code include:

l. A dynamic slip equation and an algebraic slip equation.

2. A one-dimensional neutron kinetics model.

3. A set of two-phase natural convection heat transfer correlations.

V/ith the new and improved models in the RETRAN02 version of the code (Ref. A-2), most of the objectives of the development effort are satisfied.

.The one-dimensional kinetics, dynamic slip, vector momentum, separator, and the auxiliary neutron void-fraction models allow the analyses of most PWR transients. The revised pressurizer solution techniques, the local A-3

conditions heat transfer model, and other modifications were used to analyze a number of anticipated transients without scram (ATWS) events.

Both pretest and posttest analyses of losswf-fluid tests and semiscale small-break experiments demonstrated the capability of RETRAN02 for these transients.

The NRC concluded that RETRAN is an acceptable computer program for calculating non-loca transients and can be used in licensing applications.

REFERENCES A-1 Cecil 0. Thomas (USNRC) letter to Dr. Thomas W. Schnatz, Acceptance for Referencing of Licensing Topical Reports EPRI CCM5, RETRAN-A Program for One-Dimensional Transient Thermal Hydraulic Analysis of Complex Fluid Flow Systems, and EPRI NP1850-CCM, RETR AN-02-A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems, September 0, 1980.

A-2 RETRAN02 - A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems, EPRI NP-1850-CCMA, Vol. 1, Rev. 2, 1

Computer Code Manual, November, 1980.

A-5

APPENDIX B RETRAN MODEL DESCRIPTIONS

CONTENTS B 1.0 Turkey Point RETRAN Model Description B 1.1 General B 1.2 Reactor Vessel B 1.3 Reactor Coolant Loops B I.O Steam Generators and Main Steam Piping B 1.5 Safety Systems B 2.0 St. Lucie RETRAN Model Description B 2.1 General B 2.2 Reactor Vessel B 2.3 Reactor Coolant Loops B 2.0 Steam Generators and Main Steam Piping B 2.5 Safety Systems B-2

B 1.0 TURKEY POINT RETRAN MODEL DESCRIPTION B 1.1 General The Turkey Point units are identical Westinghouse 3-loop nuclear steam supply systems (NSSS). As such, the same base model serves for both Turkey Point Unit 3 and Unit 0. The base model consists of 02 control volumes, 66 junctions, and 15 conductors.

The model includes the major components of the NSSS and control systems necessary to simulate most transients and thus should be considered a best estimate type model. For those transients which require a more detailed representation of the system, special modeling considerations or conversion to licensing type models the base model is modified as necessary.

A nodalization diagram for the base model is presented in Figure B 1.1. A description of control volumes and junctions is presented in Tables B 1.1 and B 1.2, respectively. Fission energy is calculated by solution of the point kinetics reactivity model coupled with six delayed neutron precursor groups. Feedback effects due to moderator density and fuel temperature changes are accounted for with reactivity coefficients. CEA scram reactivity is represented by a tabular function of inserted B-3

reactivity worth versus time after reactor trip. The reactivity associated with the presence of soluble boron is incorporated via tabular input or control system model. Reactor physics parameters are input to reflect cycle-specific conditions or FSAR analysis'assumptions. Decay heat is accounted for with the ANS standard decay heat curve inherent in RETRAN coupled with a multiplier to conservatively adjust the decay heat when warranted.

The enthalpy transport option is activated at junctions associated with core volumes in order to represent the axial fluid temperature distribution more accurately. With the exception of

~ the core and the steam generator tubes, II the metal mass associated with NSSS structures are not modeled.

B 1.2 Reactor Vessel Core The Turkey Point reactor cores consists of 157, 15X15 fuel assemblies.

The active core region is modeled as three equal length volumes, Volumes 2, 3 and 0 (Figure B 1.1). Heat slabs 1, 2 and 3 are associated with the core volumes and three radial regions, representing the U02 fuel, the fuel-to-clad gap and the cladding. These regions are subdivided into 0, 1 and 3 mesh intervals, respectively. Thermal conductivity and volumetric heat capacity data tables are provided by the code for the fuel, clad material and helium in the gap.

Downcomer Lower and U er Plenums U er Head Volume 20 models the downcomer region. Flow from the cold legs of both loops enter this volume.

Volume 1 represents the region below the active core. Flow enters this volume from the downcomer and exits to the core and core bypass volumes.

Volumes 6 and 7 represents the upper plenum and upper head volumes, respectively.

B 1.3 Reactor Coolant Loo s The Turkey Point RETRAN model consists of two loops. Loop A, models one of the 3 loops of the plant. The hot leg, the pump suction leg, the reactor coolant pump, and the cold leg are represented by Volumes 8, 17, 18 and 19, respectively.

Loop B, thermal hydraulically lumps the hot legs, the pump suction legs, the reactor coolant pumps, and the cold legs of the remaining two loops. These components are modeled with Volumes 21, 30, 31 and 32, respectively.

The pressurizer (Volume 35) has been modeled utilizing the RETRAN non-equilibrium volume option. The volume occupied by the surge line has been included in the pressurizer volume. The surge line hydraulic B-5

resistance has been incorporated in Junction 37, which connects the pressurizer to the hot leg (Volume 8). Junction 38 represents the hydraulic resistance of the pressurizer spray line connecting the cold leg (Volume 19) to the'pressurizer.

The pressurizer heaters are represented in the model and controlled by pressurizer pressure and level.

The three pressurizer safety valves have been lumped together as Junction 06. The power operated relief valves (PORVs) are modeled as Junctions 00 and 05.

B 1A Steam Generators and Main Steam Pi in The steam generator tube bundles have been modeled as six equal length regions (Volumes 10, 11, 12, 13, IO and 15 for the steam generator in Loop A; and Volumes 23, 20, 25, 26, 27 and 28 for the steam generator in Loop B). The enthalpy transport option is activated at the junctions associated with these volumes.

The tube bundle mass in each steam generator is modeled with four passive conductors. In Loop A, these consist of conductors 0, 5, 6, 7, 8 and 9. Conductors 10, II, 12, 13, 10 and 15 are used in Loop 9. Heat transfer coefficients are calculated by the RETRAN code from correlations that account for local flow conditions inside and outside of the tubes.

B-6

The steam generator inlet plenums and outlet plenums are modeled with Volumes 9 and 22, and Volumes 16 and 29, respectively.

The secondary side consists of a single volume for each of the two steam generators (Volumes 36 and 37). A distinct mixture level exists within each of the steam generators by use of the RETRAN phase separation option. Main feedwater (Junction 108) enters a feedwater header (Volume 106), which regulates flow to the feedwater piping line (Volumes 105 and 100). Auxiliary feedwater is modeled as a fill junction (Junctions 109 and 110) connected to the feedwater piping volumes.. Junctions 00 and 02 provide the feedwater to the loop A and B steam generators, respectively.

The steam line is represented by two control volumes. Volumes 100 and 101 represent that portion of the steam line upstream of the MSIVs represented by Junctions 100 and 101.. Immediatley downstream of the MSIVs, a common main steam header (Volume 102) is connected to a single turbine header volume (Volume 103). Junction 103, acts as the turbine stop valve, providing a flow path to the turbine, modeled with Volume 500.

The code safety valves and the relief valves are located on the steam line upstream of the MSIVs. They are represented by negative fill junctions 51, 52, 53, 50, 55, 56, 57, 58, 59 for the code safety valves; and 50 and 51 for the relief valves. The steam dump system is represented as Junction 100, a negative fill junction off the turbine header (Volume 103).

B-7

Ninety trips are included in the RETRAN base model of Turkey Point.

These trips provide for the complete set of initiating conditions for reactor trip, safety injection, main steam isolation, main feedwater isolation and auxiliary feedwater startup as given in Table 7.2.1 of the updated FSAR for Turkey Point (Ref. 13). Additional trips are included for actuating pressurizer heaters, safety and relief valves and main steam safety and relief valves.

Control system models are included in the base deck for the following control and protection features:

1. Overtemperature Delta-T Trip Setpoint

2. Overpower Delta-T Trip Setpoint

3. Low Pressurizer Pressure Trip Setpoint

0. Pressurizer Pressure Control

5. High Steam Flow Safety Injection Trip Setpoint

6. Steam Generator Water Level Control

7. Pressurizer Water Level Control

8. Main feedwater Flow Control

9. Auxiliary Feedwater, Flow Control

10. Steam Generator Relief Valve Flow Control The status of the modeled safety systems for each one of the analyses described in this report is provided in a separate table in the corresponding section.

B 2.0 ST. LUCIE RETRAN MODEL DESCRIPTION B 2.1 General St. Lucie Unit 1 and Unit 2 are geometrically similar, except for the fuel, which is 10 x 10 for Unit 1 and 16 x 16 for Unit 2. The modeling approach is indentical for each; and as such, only one of the base models, St. Lucie Unit 1, is described here.

Represented in the model are all major components of the NSSS and control systems necessary to simulate most transients. For those transients which require a more detailed representation of the system, special modeling considerations or licensing type parameters, the base model is modified as necessary. The base model consists of 02 control volumes, 60 junctions, and ll heat conductors.

The following sections describe the major areas of the St. Lucie 1 RETRAN base model. A nodalization diagram is presented in Figure B 2.1 and a description of control volumes and junctions is presented in Tables B 2.1 and B 2.2, respectively.

With the exception of the core and the steam generator tubes, the metal mass associated with NSSS structures were not modeled.

B-9

B 2.2 Reactor Vessel Core The St. Lucie Unit 1 reactor core consists of 217, 10X10 fuel assemblies. The active core region is modeled as three equal length volumes, Volumes 26, 27 and 28 (Figure B 2.1). Heat slabs 1, 2 and 3 are associated with the core volumes. These heat slabs have been modeled as three radial regions, representing the U02 fuel, the fuel-to-clad gap and the cladding. These regions are subdivided into 0, 1 and 3 mesh intervals, respectively. Thermal conductivity and volumetric heat capacity data tables are provided for the fuel, clad material and helium in the gap.

Fission energy is calculated by solution of the point kinetics reactivity model coupled with six delayed neutron precursor groups. Feedback effects due to moderator density and fuel temperature changes are taken into account via reactivity coefficients. CEA scram reactivity is represented by a tabular function of reactivity worth inserted versus time after reactor trip. Reactivity associated with the presence of soluble boron is incorporated via tabular input or control system modeling. Reactor physics parameters are input to reflect cycle-specific conditions or FSAR analysis assumptions. Decay heat is accounted for with the ANS standard decay heat curve inherent in RETRAN, coupled with a multiplier to conservatively adjust the decay heat when warranted.

The enthalpy transport option is activated at junctions associated with core volumes in order to represent the axial fluid temperature distribution more accurately.

Flow through the CEA guide tubes and core shroud annulus, which allows coolant to traverse the core region without contacting the heated fuel cladding, is represented in the model by Volume 29. The hydraulic resistance in this path is set to result in the nominal steady state core bypass flow of 2.5% of total loop flow.

Downcomer Lower and U er Plenums U er Head Volume 20 models the downcomer region. Flow from the cold legs in both loops enter this volume.

Volume 25 represents the region below the active core. Flow enters this volume from the downcomer and exits to the core and core bypass volumes.

Volumes 30 and 31 represents the upper plenum and upper head volumes, respectively. The guide tubes in the upper plenum are modeled with Volume 35.

B 2.3 Reactor Coolant Loo s The RETRAN model is divided into two loops. Loop A, a combined loop, lumps the two pump suction legs (Volume 8), the two reactor

coolant pumps (Volume 9), and the two cold legs (Volume 10). The hot leg is modeled with Volume l.

Loop B models each of the NSSS components uniquely. The hot leg is modeled with Volume 11. The pump suction legs are modeled as Volumes 18 and 21. The reactor coolant pumps are modeled as Volumes 19 and 22. The cold legs are modeled as Volumes 20 and 23.

The pressurizer, (Volume 30) has been modeled utilizing the RETRAN non-equilibrium volume option. The surge liny (Volume 32) connects the pressurizer to the hot leg of loop A (Volume 1). Volume 33 represents the pressurizer spray line connecting the lumped cold leg (Volume 10) to the pressurizer.

The pressurizer heaters are represented in the model and controlled by pressurizer pressure and level.

The pressurizer level controller determines the normal charging and letdown flow rates. The charging flow is modeled as positive fill Junctions 103 and 100. The letdown flow is modeled as negative fill Junction 105.

The three pressurizer safety valves have been lumped together as Junction 90. The power operated relief valves (PORVs) are modeled as Junction 89.

B 2A Steam Generators and Main Steam Pi in The steam generator tube bundles have been modeled as four equal length regions (Volumes 3, 0, 5 and 6 for loop A; and Volumes 13, 10, 15 and 16 for loop B). The enthalpy transport option is activated at the junctions associated with these volumes.

The tube bundle metal mass in each steam generator is modeled with four passive conductors consisting in Loop A of conductors 0, 5, 6 and 7 and in Loop B of conductors 8, 9, 10 and 11. Heat transfer coefficients are calculated by the RETRAN code from correlations that account for local flow conditions inside and outside of the tubes.

The steam generator inlet plenums and outlet plenums are modeled with Volumes 2 and 12, and Volumes 7 and 17, respectively, for loops A and B.

The secondary side of the base model has a single volume for each of the two steam generators (Volumes 51 and 52). A distinct mixture level is allowed to exist within each of the steam generators by use of the RETRAN phase separation option. Fill 3unctions 81 and 82 for Volume 51 and 3unctions 83 and 80 for Volume 52 represent the main and auxiliary feedwater.

The steam line piping from the steam generator outlet nozzles to the turbine stop valves comprise the main steam system. Volumes 53 and 50 model the piping upstream of the MSIVs in both loops. The code

safety valves are simulated by Junctions 35, 86, 87 and 33. Volumes 55 and 56 model the piping immediately downstream of the MSIVs.

The main steam header up to the turbine throttle valve is modeled as a single Volume (Volume 57). The steam dump and bypass to condenser is modeled by negative fill Junctions 96 and 97, respectively, from the main steam header.

Sixty two trips are included in the RETRAN base model of ST. Lucie.

These trips provide the required conditions to model initiating signals for reactor trip, safety injection, main steam isolation and main feedwater isolation. Additional trips are included for actuating pressurizer heaters, safety and relief valves, and main steam safety valves.

Control system models are included in the base deck for the pressurizer pressure and level control functions. Additional trips and controls such as for the auxilary feedwater and the steam dump systems are added to the base model on an as needed basis depending on the type of transient being modeled. The status of the modeled safety systems for each one of the analyses described in this report is provided in a separate table in the corresponding section.

TABLE Bl.l TURKEY POINT RERAN K)DEL VOIlME DESCRIPTION 1, Lcwer Plenum 2 Lcwer Core 3 Middle Core 4 Upper Core 5 Core Bypass 6 Upper Plenum 7 Upper Head 20 Dmm~

SINGLE ZDOP ( pod A).-'

Hot Leg Piping 9 Steam Generator Inlet Plenum 10 Steam Generator Tubes, Hotter Side, First Section 11 Steam Generator Tubes, Hotter Side, Second Section 12 Steam Generator Tubes, Hotter Side, Top 13 Steam Generator Tubes, Cooler Side, Top 14 Steam ~erator Tubes, Ccoler Side, First Section 15 Steam Generator Tubes, Cooler Side, Second Section 16 Steam Generator Outlet Plenum 17 Pump Section Piping 18 Pump 19 Cold Leg, Pump Discharge Piping 35 Pressurizer OOMBZmp OR DOUBLE rmP (~<V >)

21 Hot Leg Piping 22 Steam Generator Inlet Plenum 23 Steam Generator Tubes, Hotter Side, First Section 24 Steam Generator Tubes, Hotter Side, Second Section 25 Steam Generator Tubes, Hotter Side, Top 26 Steam Generator Tubes, Cooler Side, Top 27 Steam Generator Tubes, Cooler Side, First Section 28 Steam Generator Tubes, Cooler Side, Second Section 29 Steam Generator Outlet Plenum

TABLE Bl.1 (CONTINUED)

TURKEY'OIHr REGIN MODEL VOMME DESCRIPTION VOLUME NO. EEHCRIP7ION CDMBINED OR DOUBLE LOOP 30 Pump Section Piping 31 Pump 32 Cold Leg, Purp Discharge Piping STEAM GENERATORS (SECONDARY) 36 . Slg+e, Steam Generator 37 . Steam Generator (Combined)

SZEAiKXNEy, 100 Single:. Steamline Upstream of MSIV 101 Double," -Steamline (Canbined) Upstream of MSIV 102 Steamline"'Dmmstream of MSIV' 103 Turbine Header 104 F~ter Piping Fnxn Header to Single S.G.

105 Feedwater Piping Frcm Header to Combined S.G.

106 Feechvater Header

~VZAINMEiVZ 500 Sink Volum (Infinite Volume)

B-16

TABLE B1.2 TURKIC'OOF KTRPH MODEL JVNCTION DESCRIPTION XNiCTION NO.

20 Vessel Inlet from Single Loop 35 Vessel Inlet frcan Double Loop 21 Leakage Path fran Dcumcomer to Upper Head 22 Dmmcaner to Lcurer Plenum 1 Core Inlet 2 Middle Core Inlet 3 Middle Core Outlet 4 Core Outlet 5 Core Bypass Inlet 6 Core Bypass Outlet 7 Upper Plenum to Upper Head 8 Vessel Outlet to Single Loop 23 Vessel Outlet to Ccmbined Loop SIEKZ LOOP 9 Steam Generator Plenum Inlet.,

10 Steam Generator, Inlet Tubes ll Steam Generator Tubes 12 Steam Generator Tubes 13 Steam Generator Tubes 14 Steam Generator Tubm 15 Steam Generator Tubes 16 Steam Generator, Outlet Tubes 17 Steam Generator Plenum Outlet 18 Pump Section 19 Pump Discharge 37 Surge Line 38 Spray Line Inlet 39 Spray Line Outlet 190 Safety Injection 191 Charging COMBINED OR DOUBLE LOOP 24 Steam Generator Plenum Inlet 25 Steam Generator Inlet; Tubes 26 Steam Generator Tubes 27 Steam Generator Tubes 28 Steam Generator Tubes

TABLE Bl. 2 (CONTINUED)

TURKEY POINT RH'RAN MODEL JtPiTCTIG'.7 DESCRIPTION JUNC1'ION NO. DESCRIPTION COMBINED OR DOUBLE ZDOP 29 Steam Generator Tubes 30 Steam Generator Tubes 31 Steam Generator Outlet, Tubes 32 Steam Generator Plenum Outlet 33 Pump Section 34 Pump Discharge 320 Safety Injection 321 Charging STEAM GmmmORS (mmmOARY}

40 Single SG Inlet 41 Single SG Outlet 400 Auxiliary Feechmter Inlet to Single SG 42 Canbined SG Inlet 43 Ccnibined SG Outlet 420 Auxiliary Feedwater Inlet to Ccnhined SG PKXSURI2ZR VALVES 44 Relief Valve (Signal frcm P. Ccntroller) 45 Relief Valve (2350 psia) 46 Safety Valve (2500 psia)

SHRMEaLK 100 ASIV, Single Loop 101 MSIV; Combined Loop 102 SteanGine to Turbine Header 103 Turbine Inlet 50 Atmospheric Steam Relief. Single Loop (1050 psia) 51 Atmospheric Steam Relief. Ccmbined Loop (1050 psia) 52 SG Single Loop Safety Valve 1 (1100., 1089.

psia)

• SG Single Loop Safety Valve 2 (1115., 1104.

psia) 54" SG Single Loop Safety Valve 3 (1130., 1119.

psia); Break (Steanline Break)

TABLE Bl.2 (CONTINUED)

TURKEY POINT REARM K)DEL JUNCVIG% DESCRIPTIOi%

JUMCZIQH iK.

SI'PAMLINE (CDÃZ'D)

SG Single Loop Safety Valve 4 (.1145., 1134.

psia)

SG Double Loop Safety Valve 1 (1100., 1089.

psia)

SG Double Loop Safety Valve 2 (1115., 1104.

psia) 58 S" Double Loop Safety Valve 3 (1130., 1119.

psia) 59 SG Double Loop Safety Valve 4 (1145., 1134.

psia)

SZEM DUMP SYSTEM 104 Steam Dump to Condenser 106 F~ter Header to Single Loop Piping 107 Feedwater Header to Double Locp Piping 108 Feechmter Header Inlet B-19

TABLE B2.1 ST. LUCIE UNIT 1 RETRAN MODEL VOLUME DESCRIPTION VOLUME NO. DESCRIP TION Unaffected Loo (W/Pressurizer) ~LOO A 1 Hot Leg'Piping 2 Steam Generator Inlet Plenum 3 Steam Generator Tubes, Hotter Side, First Section 4 Steam Generator Tubes, Hotter Side, Top 5 Steam Generator Tubes, Cooler Side, Top 6 Steam Generator Tubes, Cooler Side, First Section 7 Steam Generator Outlet Plenum 8 Pump Section Piping 9 Pumps (1-A and 1-B) 10 Cold Leg, Pump Discharge Piping 32 Pressurizer Surge Line 33 Pressurizer Spray Line 34 Pressurizer Affected Loo ~LOO -B ll 12 Hot Leg Piping Steam Generator Inlet Plenum 13 Steam Generator Tubes, Hotter Side, First Section 14 Steam Generator Tubes, Hotter Side, Top 15 Steam Generator Tubes, Cooler Side, Top 16 Steam Generator Tubes, Cooler Side, Second Section 17 Steam Generator Outlet Plenum 18 Pump Suction Piping 19 Pump 2-A 20 Cold Leg, Pump Discharge Piping 21 Pump Suction Piping 22 Pump 2-B 23 Cold Leg, Pump Discharge Piping Vessel 25 Lower Plenum 26 Lower Core 27 Middle Core 28 Upper Core 29 Core Bypass 30 Upper Plenum 31 Upper Head 24 Downcomer 35 CEA Guide Tube B-20

TABLE B2.1 (CONTINUED)

ST. LUCIE UNIT l RETRAN MODEL VOLUME DESCRIPT1ON JUNCTION NO DESCRIPTION Steam Generators (Seconda )

51 Unaffected Loop Steam Generator 52 Affected Loop Steam Generator Steamline 53 Unaffected Loop Steamline Upstream of MSIV 54 Unaffected Loop Steamline Upstream of MSIV 55 Unaffected Loop Steamline Downstream of MSIV 56 Affected Loop Steamline Downstream of MSIV 57 Turbine Header B-21

TABLE B2.2 ST LUCIE UNIT 1 RETRAN MODEL JUNCTION DESCRIPTION JUNCTION NO. DESCRIPTION Vessel Vessel Inlet from Unaffected Loop 25,26 Vessel Inlet from Affected Loop 27 Downcomer to Lower Plenum 28 Core Inlet 29 Middle Core Inlet 30 Middle Core Outlet 31 Core Outlet 32 Core Bypass Inlet 33 Core Bypass Outlet 34 Upper Plenum to Upper Head 1 Vessel Outlet to Unaffected Loop 12 Vessel Outlet to Affected Loop 39 CEA Guide Tube Inlet 40 CEA Guide Tube Outlet Unaffected Loo 2 Steam Generator Plenum Inlet=

3 Steam Generator, Inlet Tubes 4,5g6 Steam Generator Tubes 7 Steam Generator, Outlet Tubes 8 Steam Generator Plenum Outlet 9 Pump Section 10 Pump Discharge 35 Surge Line Inlet 36 Pressurizer Inlet 37 Spray Line Inlet 38 Spray Line Outlet 100 Safety Injection 103 Charging 105 Letdown Affected Loo 13 Steam Generator Plenum Inlet 14 Steam Generator Inlet, Tubes 15,16,17 Steam Genertaor Tubes 18 Steam Generator Outlet, Tubes 19,22 Steam Generator Plenum Outlet 20 Pump Suction 21 Pump Discharge 101 Safety Injection 104 Charging 23 Pump Suction 24 Pump Discharge B-22

TABLE B2. 2 (CONTINUED)

ST LUCIE UNIT 1 RETRAN MODEL JUNCTION DESCRZPTION JUNCTZON NO DESCRIPTION Steam Generators (Second )

81 Affected SG Main Feedwater 91 Single SG Outlet 82 Auxiliary Feedwater Inlet to Affected Loop 83 Unaffected SG Main Feedwater 92 Unaffected SG Outlet 84 Auxiliary Feedwater Inlet to Unaffected Loop SG Pressurizer Valves 89 PORVs (2350 psia) 90 Safety Valve (2500 psia)

Steamline 93 MSIV, Affected Loop 94 MSIV, Unaffected Loop 95,96 Steamline 'to Turbine Header 99 Turbine Inlet 85 SG Unaffected Loop Safety Valve Bank 1 86 SG Unaffected Loop Safety Valve Bank 2 87 SG Affected Loop Safety Valve Bank 1 88 SG Affected Loop Safety Valve Bank 2 Steam Dum S stem 97 Steam Bypass to Condenser 98 Steam Dump to Condenser B-23

STEAM SHIN SAFETY 100 VOLUME 103 IIELIEIVALVE dal ety Vaine o AVAIV 00 il00 00 05 62 dS de dd d0 I

L 102 100 MSIV MSCV MSCV 03 ReNeI VelEEee SAFETY VALVES 40 ee dPAAY ee ee ee 110 AUXILIARY FEEOWATER HEATERS Ac@Elegy Feedmeee Oe ee el el 37 Qe C@ 20 0 Oe CHARGING 30 '0 20 CHARGING le

~O VOLUME JUNCTION HEATCONOUCTOR 101 Sl 320 ee ee 30 TeSF~

100 10 tom EIALILIIL LOOP EE FEEOWATER HEAOER LOOP A IIAAIVIA 100 MAINFEEO FIGURE Bl.l TURKEY POINT RETRAN HODEI NODALIZATION DIAGRAM

Safety iNIPlV as Valves aa oa ~a Valves av To Turbine Bank 2 Bank f

~ .MSIV as ~ MSIV Qsa Qaa Steam Steam Relief Aallaf Dump Bypass Ol Valves Valves 02 Spray ao 07 Qal al ao OR 33

~A AFW ai Qai AFVJ Q31 Ra 13 sa 30 12 le 37 21 33 Qll 2

A'I 30 QIR Qaa Loop 1 Ra Loop 2 Qav lo Ra 2a 21 Qao Ql ~

10 2PPI IPP 10 2O 20 Cbarglng 22 Sl Charging SI Qa 2a 141 loa 100 103 Letdown 24 23

.IPP Ion LOOP LOOP B VOLUMES 102 HEAT Coff DUCTOB FIGURE 82.l ST LUCIE 1 RETR~ HODEL NODALIKATIONDIAGRM F IIIICTI 0 NS 021

APPENDIX C STAFF EXPERIENCE

The Thermal Hydraulics and System Analysis Group is part of the Fuel Resources Department of FPL and has overall responsibility for performing reload safety evaluations, plant transient and safety analyses. All thermal hydraulic system models are maintained by this group. The group currently includes a supervisor, and seven full-time engineers. A number of supporting personnel such as consultants, programmers, technicians, and a co-op student are available. Present nuclear and thermal hydraulic analysis related experience in the group totals approximately 60 engineer-years. Degree levels include PhD, MS, and BS degrees in Engineering and related disciplines. Typically, each member of the group attends two industry related meetings per year, such as user group meetings or topical meetings sponsored by professional societies. Overall, members of the group have presented about 20 papers at industry related meetings. It is expected that the level of experience will be maintained at or above the present level.

The Thermal Hydraulics and System Analysis Group routinely consults with Dr. Joel Veisman of the University of Cincinnati.

C-2