Decrease By the Secondary · 2018-07-31 · NTH-TR 01 FLORIDA POWER & LIGHT COMPANY THERMAL...

FPL

NTH-T8-01

Decrease in Heat Removal

By the Secondary System

Issued By:Fuel Resources

Date: July 1989

89i0060230 891002PDR AD(ICK 05000250P PDC

NTH-TR 01

FLORIDA POWER & LIGHT COMPANY

THERMAL HYDRAULIC MODEL QUALIFICATION

DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM EVENTS

JULY 1989

Jorge ArpaWilliam BryanRandolph DeckerJoel HandschuhOlga HanekSundershan MathavanStavroula MihalakeaJayaram PolavarapuJose Ramos

APPROVED BYC

J. A. HandschuhSupervisor of ThermalHydraulic.cs and Safety

APPROVED BY! o

D. C. PoteralskiManager of Nuclear Fuel

DISCLAIMER OF RESPONSXBXLITY

This document was prepared by the Fuel Resources Department ofFlorida Power & Light, Company. The material therein is believedto be true and accurate to the best of our knowledge andinformation. However, ~ it is authorized and intended for use andapplication only by Florida Power & Light Company.

FLORIDA POWER & LIGHT COMPANY~ ITS OFFXCERS ~ DXRECTORS, AGENTAND EMPLOYEES SHALL NOT BE RESPONSIBLE OR LIABLE FOR ANY CLAIMS,LOSSES, DAMAGES OR LIABILITIES, WHETHER OR NOT DUE TO OR CAUSEDBY THE NEGLIGENCE OF FLORXDA POWER & LIGHT COMPANY, RESULTINGFROM THE USE OR MISUSE OF THXS DOCUMENT OR ANY INFORMATIONCONTAXNED HEREIN

ACKNOWLEDGEMENTB

The authors would like to thank A. Mirza for has dedicatede orff ts in the development of the graphics used in the topical.The assistance of M. Varela in the typing and preparat'\ ation of thisreport is gratefully acknowledged.

The authors would also like to acknowledge the many3n ivld'viduals who have contributed to previous RETRAN applicationsat FPL. This report would not have been possible withefforts.

ty TABLE OF CONTENTS

PAGE

Acknowledgements

List of Tables

List of Figures

List of Acronyms

1.0 INTRODUCTION

1.1 Background1.2 FPL RETRAN Models

v3.

vi1i

1.2.1 Steam By-pass Contgol Systems1.2.2 Multi-node Steam Generator Models1.2.3 Modelling Approaches

1.2 ~ 3 ~ 11 ' '.21 ' ' '1 2 ' '1.2 ' '

Pressurizer ModellingNodalizationSafety Valve ModellingSteam Generator Tube PluggingControl Systems

2 ~ 0 T&UCEY POINT LOSS OF INVERTER EVENT 15

2.1 Summary of Events2.2 Analysis2.3 Results

2.3.1 RETRAN Analysis

2.3.1.1 Secondary System Response2.3.1.2 Primary System Response

151619

1921

2.4

2.3.2 RETRAN Comparison to Plant Data

Sensitivity Studies with RETRAN Models

22

25

2.5 Conclusions 27

3.0 ST. LUCIE UNIT 1 PARTIAL LOSS OF FEEDWATER 35

3 ~ 13.2

Summary of EventsAnalysis

3535

~ 1

0

TABLE OF CONTENTS (Cont9.nuect)

PAGE

3.3 Results

3.3.1 RETRAN Analysis

3.3.1.1 Secondary System Response3.3.1.2 Primary System Response

3.3.2 RETRAN Comparison to Plant Data

3.4 Conclusions

38

38

3841

41

43

4.0 SAFETY ANALYSIS METHODS

4.1 Introduction4.1.1 RETRAN Modelling

54

54

56

4.1 ~ 1.14.1.1.24 ~ 1 ~ 1.34.1.1.4

Safety ValvesPressurizer ModelTemperature Transport DelayMulti-node Steam Generator Model

56565757

4.2 Limiting Transient Determination

4.2.1 St. Lucie Unit 1

60

60

4.2.1.1 Loss of External Load4.2.1.2 Turbine Trip4.2.1.3 Loss of Condenser Vacuum

4.2.1.3.1 Limiting Event Development4.2.1.3.2 Results

606162

6265

4.2 1.44.2 ' 54.2 '.64.2 1.74.2. 1.8

.Main Steam Isolation Valve ClosureSteam Pressure Regulator Failure .

Loss of Non-emergency AC Powerto the Station AuxiliariesLoss of Normal FeedwaterFeedwater Line Break

737374

7576

TABLE OF CONTENTS (Continued)

4.2.2 St. Lucie Unit 2

PAGE

77


4.2.2.3.1 Limiting Event Development4.2.2.3.2 Results

777879

7982

4.2.2.44.2.2.54.2.2.6

4.2.2.74.2.2.8

Main Steam Isolation Valve ClosureSteam Pressure Regulator FailureLoss of Non-emergency AC Powerto the Station AuxiliariesLoss of Normal FeedwaterFeedwater Line Break

909091

9293

i4.2.2.8.1 Limiting Event Development4.2.2.8.2 Results

4.2.3 Turkey Point Units 3 and 4


9396

105

105106107

4.2.3.3.1 Limiting Event Development 1074.2.3.3.2 Results 110

4.2.3.44.2.3.54.2.3.6

Main Steam Isolation Valve ClosureSteam Pressure Regulator FailureLoss of Non-emergency AC Powerto the Station Auxiliaries

118118119

4.2.3.6.1 Limiting Event, Development4.2.3.6.2 Results

119122

4.2.3.7 Loss of Normal Feedwater4.2.3.8 Feedwater Line Break

129129

5.0 Conclusions 130

6.0 References 131

LIST OF TABLES

TABIE TITLE PAGE

1 '-12 ~ 2 1

FPL RETRAN Base Models Summary

Initial Conditions forLoss of Inverter Event

2 ~ 3 1 Sequence of Events, RETRAN AnalysisLoss of Inverter Event

20

2 ~ 3 ~ 2 1 Sequence of Events, RETRAN Comparisonto Plant Data

24

2.4-1 RETRAN Sensitivity Studies, Loss ofInverter Event

28

3 ~ 2 1 Initial Conditions for PartialLoss of Feedwater Event

37

3 ~ 3 ~ 1 1 Sequence of Events, RETRAN AnalysisPartial Loss of Feedwater Event,

39

3 ' '-1~ ~ Sequence of Events, RETRAN Comparisonto Plant Data

4 '-14.1-2

4.2.1.3-1

Standard Review Plan Events

FPL RETRAN Methodology

Key Parameters Assumed for the Loss ofCondenser Vacuum Event, St. Lucie Unit 1

55

58

64

4.2.1 ~ 3-2 Sequence of Events, Loss of CondenserVacuum for St. Lucie Unit 1

66

4 ' '.3-1 Key Parameter Assumed for the Loss ofCondenser Vacuum Event, St. Lucie Unit 2

81

4 '.2.3-2 Sequence of Events, Loss of CondenserVacuum for St. Lucie Unit 2

83

4 '.2.8-1

4 ~ 2. 2. 8-2

Key Parameters Assumed for the FeedwaterLine Break Event, St. Lucie Unit 2

Sequence of Events, Feedwater LineBreak for St. Lucie Unit 2

95

97

LIST OF TABLES (Continued)

TABLE TITLE PAGE

4.2.3.3-1 Key Parameters Assumed for the Loss ofCondenser Vacuum Event, Turkey PointUnits 3 & 4

109

4.2.3.3-2

4.2.3.6-1

Sequence of Events, Loss of CondenserVacuum for Turkey Point Units 3 & 4

Key Parameters Assumed for the Loss ofNon-emergency AC Power, Turkey PointUnits 3 & 4

111

121

4.2. 3. 6-2 Sequence of Events, Loss of Non-emergencyAC Power, Turkey Point Units 3 & 4

123

LIST OF FIGURES

F1GURE

1 ~ 2 1

1 ~ 2 2

1 % 2 3

1.2-4

1.2-5

TITLE

St. Lucie Unit 1 RETRAN Model

St. Lucie Unit 2 RETRAN Model

Turkey Point Units 3 & 4 RETRAN Model

St. Lucie Multi-node SG Model

Turkey Point Multi-node SG Model

PAGE

10

12

13

14

2 ~ 3 1 Steam Generator Pressure "B" and "C",Turkey Point Loss of Inverter Event

2%32

2%3 3

Steam Generator "A" Pressure,Turkey Point, Loss of Inverter Event

RCS Average Temperature,Turkey Point Loss of Inverter Event

30

2.3-4

E.E-E

2. 3-6

Reactor Power,Turkey Point Loss of Inverter Event

Pressurizer Pressure,Turkey Point Loss of Inverter Event

Pressurizer Level,Turkey Point Loss of Inverter Event

32

33

3 ~ 3 1 Steam Generator "B" LeveliSt. Lucie 1 Partial Loss of Feedwater

3 ~ 3 2 Steam Generator "B" PressureiSt. Lucie 1 Partial Loss of Feedwater

45

3 ~ 3 3 Steam Generator "A" Pressure,St. Lucie 1 Partial Loss of Feedwater

3. 3-4 Steam Generator "A" Level/St. Lucie 1 Partial Loss of Feedwater

47

3.3-5 Pressurizer Pressure,St. Lucie 1 Partial Loss of Feedwater

48

3.3-6 Pressurizer Level,St. Lucie 1 Partial Loss of Feedwater

E.E-7 Loop "A" Average Temperature/St. Lucie 1 Partial Loss of Feedwater

50

LIST OF FIGURES (Continued)

FIGURE TITLE PAGE

3.3-8 Loop "B" Average Temperature,St. Lucie 1 Partial Loss of Feedwater

51

3 '-9 Steam Generator "B" Pressure, MSSV FlowSensitivity, St. Lucie 1 Partial Loss ofFeedwater

52

3.3-10 Steam Generator "A" Pressure, MSSV FlowSensitivity, St. Lucie 1 Partial Loss. ofFeedwater

53

4-1

4.2.1.3-1

Primary and Secondary Safety ValveModels, RETRAN Licensing Methodology

Reactor Power, St. Lucie Unit 1Loss of Condenser Vacuum

59

67

4.2 '.3-2 Heat Flux, St. Lucie Unit 1Loss of Condenser Vacuum

68

4.2.1.3-3

4.2.1.3-4

RCS Pressure, St. Lucie Unit 1Loss of Condenser Vacuum

RCS Temperatures, St. Lucie Unit 1Loss of Condenser Vacuum

69

70

4.2.1.3-5 SG Pressure, St.'ucie Unit 1Loss of Condenser Vacuum

71

4.2. 1. 3-6 Pressurizer Level, St. Lucie Unit 1Loss of Condenser Vacuum

72

4.2.2 '-1 Reactor Power, St. Lucie Unit 2Loss of Condenser Vacuum

84

4.2.2.3-2 Heat Flux, St. Lucie Unit 2Loss of Condenser Vacuum

85

4.2.2.3-3 RCS Pressure, St. Lucie Unit 2Loss of Condenser Vacuum

86

4.2.2.3-4 RCS Temperatures, St. Lucie Unit 2Loss of Condenser Vacuum

87

4.2.2.3-5 SG Pressure, St. Lucie Unit 2Loss of Condenser Vacuum

88

~~ ~ ~4.2.2.3-6 Pressurizer Level, St. Lucie Unit 2

Loss of Condenser Vacuum89

LXST OF FIGURES (Continued)

FIGURE TITLE PAGE

4.2.2.8-1

4.2.2.8-2

Pressurizer Pressure, St. LucieUnit 2 Feedwater Line Break

RCS Average Temperature, St. LucieUnit 2 Feedwater Line Break

98

99

4.2.2.8-3 Reactor Power, St. Lucie Unit 2Feedwater Line Break

100

4.2.2.8-4 Pressurizer Level, St. Lucie Unit 2Feedwater Line Break

101

4.2.2.8-5 Unaffected Loop Temperatures,St. Lucie Unit 2, Feedwater Line Break

102

4.2.2.8-6

4.2.2.8-7

Affected Loop Temperatures,St. Lucie Unit 2, Feedwater Line Break

Steam Generator Pressure, St. LucieUnit 2 Feedwater Line Break

103

104

4.2.3.3-1~ ~ ~ Reactor Power, Turkey Point Units 3 & 4Loss of Condenser Vacuum

112

4.2.3.3-2 Heat Flux, Turkey Point Units 3 & 4Loss of Condenser Vacuum

113

4 '.3.3-3 RCS Pressure, Turkey Point Units 3 & 4Loss of Condenser Vacuum

114

4.2.3.3-4 RCS Temperatures, Turkey PointUnits 3 & 4 Loss of Condenser Vacuum

115

4.2.3.3-5 SG Pressure, Turkey Point Units 3 & 4Loss of Condenser Vacuum

116

4 '.3.3-6 Pressurizer Level, Turkey PointUnits 3 & 4 Loss of Condenser Vacuum

117

4.2.3 '-1

4.2.3.6-2

4 ' ' '-3

Reactor Power, Turkey Point Units 3 & 4Loss of Non-emergency AC Power

Heat Flux, Turkey Point Units 3 & 4Loss of Non-emergency AC Power

Pressurizer Pressure, Turkey Point Units3 & 4 Loss of Non-emergency AC Power

124

125

126

LIST OF FIGURES (Continued)

F1GURE TITLE PAGE

4 '.3.6-4

4.2.3.6-5

Pressurizer Level, Turkey Point Units 3& 4 Loss of Non-emergency AC Power

RCS Temperatures, Turkey Point Units 3& 4 Loss of Non-emergency AC Power

127

128

LIST OF ACRONYMS

AFAS

CE

CEA

CEDM

DEH

DNBR

EPRI

FLB

FPL

FRV

IHTC

LOAC

Auxiliary Feedwater Automatic Signal

Auxiliary Feedwater

Combustion Engineering

Control Element Assembly

Control Element Drive Mechanism

Digital Electronic Hydraulic

Departure from Nucleate Boiling Ratio

Electric Power Research InstituteFeedwater Line Break

Florida Power & Light Company

Feedwater Regulating Valve

Final Safety Analysis Report

Gallons Per Minute

Interface Heat Transfer Coefficient

Loss of Non-Emergency Power to the StationAuxiliaries

MFRV

MSIV

Loss of Forced Coolant Flow

Main Feedwater Regulating Valve

Main Feedwater4

Main Steam Isolation Valve

NSSS

PORV

RCP

RCS

Megawatt

Nuclear Regulatory Commission

Nuclear Steam Supply System

Power Operated Relief Valve

Reactor Coolant Pump

Reactor Coolant System

X3.3.

LIST OF ACRONYHS (Continued)

RPS

SER

SDBS

SG

SI

SLB

TBS

TBV

TSV

Reactor Protective System

Safety Evaluation Report

Steam Dump and By-pass System

Steam Generator

Safety InjectionSteam Line Break

Turbine By-pass System

Turbine By-pass Valve

Turbine Stop Valve

Westinghouse

>3,

V

l4

1

1.0 Introduction

In 1986, Florida Power and Light Company (FPL) submitted atopical report to the NRC (Reference 1) which provided RETRANanalyses ranging over a broad spectrum of transients. FPL's goalwith this submittal was to meet the requirements of the NRC asdelineated in Generic Letter 83-11 (Reference 2) to demonstratestaff technical proficiency. This was a first step in the processof obtaining qualification of FPL to perform safety analyses usingthe RETRAN code to support licensing actions.

In 1988, the NRC safety evaluation (Reference 3) was obtainedon the RETRAN topical report. The conclusion of this evaluationwas that "...the topical report does demonstrate that FPL has thecapability to use RETRAN computer code to perform systems transientcalculations for the Turkey Point and St. Lucie plants and,therefore, fulfillsthe requirements of Generic Letter 83-11." TheNRC safety evaluation also stated, "However, additional comparisonsbetween the RETRAN computed results and plant operating datatogether with appropriate nodalization, sensitivity studies andlicensing assumptions will be necessary in future reports beforethese models are acceptable for licensing submittals." The purposeof this document is to provide the additional information requiredby the NRC in one specific class of transient events, Decrease inHeat Removal by the Secondary System and obtain NRC approval forthe use of RETRAN in licensing actions associated with transientswithin this category.

The sections that follow address the NRC requirement foradditional information. Basic information related to the RETRANmodels currently used at FPL will be discussed in Section 1.1.Models for St. Lucie Unit 1 and Unit 2 as well as that for TurkeyPoint Units 3 & 4 are provided.

Section 2.0 is a detailed description of a reactor transientat Turkey Point which produced a reduction in heat removal of thesecondary side because of a turbine runback. The mismatch betweenprimary power and secondary heat removal resulted in a reactorscram on high pressure. This section also provides sensitivitystudies which will provide justification for the RETRAN modellingused in transients within .the Decrease in Heat Removal by theSecondary System category. Prediction of this transientdemonstrates the accuracy of two key RETRAN models, i.e.,calculation of primary to secondary heat transfer and thepressurizer model, which are needed for the analysis of this classof event.

I

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V

Section 3.0 presents a benchmark analysis for a partial lossof feedwater event that occurred at St. Lucie Unit 1. The RETRANmodel used in this analysis is consistent with the Turkey Pointmodel discussed in Section 2.0 with the exception of geometric andsystems differences representative of Westinghouse versusCombustion Engineering Nuclear Steam Supply Systems (NSSS). Thiscalculation demonstrates the use of the multi-node steam generatormodel and presents qualification of this model for futureapplications when accurate tracking of the steam generator levelis important.

Section 4.0 willpresent the assumptions that willbe utilizedin the future for licensing actions within the category of Decreasein Heat Removal by the Secondary System events. A discussion ofeach event within this category will be provided along with thedetermination of the limiting event within the category throughcomparison of the key physical phenomena that impact thetransients. The limiting transients are executed and presented inSection 4.0 using input parameters chosen to produce boundingresults for those transients in RETRAN. Section 4.0. also includesthe RETRAN modelling assumptions which will be used in the futurefor licensing actions. The modelling used is derived from theresults of sensitivities performed in Section 2.0 using the basicplant models which will be discussed in Section 1.2.

1.2 FPL RETRAN Models

All the analyses presented in this report have been performedwith RETRAN02 MOD004 (Reference 4). Approval of this code for usein licensing applications was obtained from the NRC in October 1988(Reference 5).

The noding diagrams for the St. Lucie Unit 1, Unit 2 and theTurkey Point Units 3 & 4 RETRAN base models are shown in Figures1.2-1, 1.2-2 and 1.2-3, respectively. A general description ofthese models is provided in Reference 1. A more detaileddescription of the Steam Bypass Control Systems and the Multi-nodeSteam Generator Models available for use in the respective basedecks is provided in Sections 1.2.1 and 1.2.2. These two componentmodels are described here because of their important roles in theprediction of several of the transients presented in this report.A summary matrix of the models used in the respective plant basedecks is shown on Table 1.2-1. A discussion on the modellingapproaches followed in the development of the plant base models ispresented in Subsection 1.2.3.

I

~)

li

1.2.1 Steam B ass Control S stems~ ~

The discharge flows of the respective Steam Dump BypassSystems (SDBS) for the St. Lucie Units and the Turbine By-passSystem (TBS) for the Turkey Point Units are represented with filltables o'f flow versus demand. Except for the cases where thedemand is a function of pressure, allowance for pressuredependence of the discharge flow is ignored. The demand iscomputed by the respective Steam Bypass Control Systems (SBCS).

ST. LUCIE STEAM DUMP BYPASS SYSTEM

The SDBS and their associated SBCS at both St. Lucie Unitsare identical. The total steam bypass flow capacity of theSDBS is 454 of the nominal steam flow that corresponds to2560 Mwth (original plant rating). The system consists ofthree sequentially operated valve groups with one 5%, one104 and three 104 capacity valves respectively. The systemutilizes two modes of operation, the modulation and theQuick Opening (QO) modes. During. load reductions and othertransients, all groups are modulated to maintain steamheader pressure at 910 psia. If a reactor trip occurs,valve groups 2 and 3 (40% total capacity) are switched frompressure to temperature modulation to reduce primary coolantaverage temperature to a value of 535.1 Degree F while valvegroup 1 continues to modulate based on the steam generatorheader pressure.

The pressure control mode of operation is accomplished bymeans of a proportional-plus-integral-plus-derivative (PID)controller operating on the difference between the steamheader pressure and a constant setpoint of 910 psia. Thismode of operation regulates the valves at a relatively lowspeed, which limits the maximum load changes that can beaccommodated by this type of control alone.

For large load reductions and for unit trips from high powerlevels, the energy accumulated in the system would typicallybe large enough that the secondary safety valves would berequired to relief secondary system pressure before thebypass valves would reach a fully open position under thepressure control mode of operation. To avoid this, the QO

control mode is actuated in these situations to open allvalves at a much faster rate. A QO signal is generated whena load reduction rate is greater than a given rate or if,after reactor trip, the primary coolant average temperatureexceeds a preset threshold.

I

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1

I4 '

— TlJRKEY POINT TURBINE BY-PASS SYSTEM

At the Turkey Point Units, the relief capacity of theTBS is 40% of nominal steam flow. The system consists oftwo valve groups with two valves each and a total dischargecapacity of 20% per group. The TBS at Turkey Point differsfrom that at St. Lucie in the manner in which the bypassflow is controlled. At Turkey Point, the only automaticcontrol variable is primary coolant average temperature.The magnitude of the flow is based on the difference betweenactual coolant average temperature and a referencetemperature. The reference temperature can be based eitheron a load reduction program or a temperature setpoint of 547F. The first, is used for transients not involving turbinetrip and the second for situations where turbine trip hasoccurred. A QO logic is also available to open the valvesat a faster rate than during temperature control. Itsopening logic varies depending on whether or not a turbinetrip has occurred.

1.2.2 Multi-node Steam Generator Models

Detailed steam generator models are available for caseswhere predictions of level, or inventory are important. The

~~ ~ ~

~

~

~

~

~

~

~

models for St. Lucie Units and Turkey Point are shown in Figures1.2-4 and 1.2-5 respectively. Both models use the Non-Equilibrium code option to better represent the phenomena in theupper downcomer (outside the separators) region. The separatorsare represented with the RETRAN bubble rise model. Modelparameters such as mixture levels and enthalpy of the liquidregion in the non-equilibrium volumes are adjusted to yielddesign liquid mass inventories and good predictions of planttransient level responses.

1.2.3 Modellin A roaches

1.2.3.1 Pressurizer Modellin

The primary system pressure response in a PWR is largelydetermined by the pressurizer insurge or outsurge flows, theprocesses that take place within the pressurizer and by theresponse of the pressure control systems. It is important thatall these effects be correctly modelled to ensure that the systempressure response is calculated adequately.

For heatup type transients, the pressurization or insurgephase of the transient is crucial because it affects pressurepeak magnitude and timing and hence the time of the reactorprotection system actuation.

The modelling approach for the pressurizer in the FPL RETRANmodels has the following features:

SINGLE NODE, NON-EQUILIBRIUMVOLUME

The non-equilibrium option is used to model the pressurizervolume in the FPL RETRAN models. This option allows thevapor and liquid regions in the pressurizer to havedifferent temperatures while at the same pressure. This isa needed feature to model the phenomena taking place in thepressurizer especially during insurges. When an insurgeoccurs, the liquid region in the pressurizer becomessubcooled with the addition of cooler fluid from the hot. legwhile the vapor region, compressed by the addition of newfluid into the pressurizer, becomes superheated.

Another important process taking place in the pressurizer isthermal stratification within the liquid region. Duringfluid insurges into the pressurizer thermal stratificationtends to delay the cooling of the vapor region and thereforeresults in higher pressurization rates. During outsurges,thermal stratification causes the upper layers of hotterliquid to stay hotter for a longer time thus causing thepressure to decrease less rapidly than in the cases whereperfect mixing is allowed. The effects of the perfectmixing assumption used in the one node pressurizer model canbe balanced with the use of a very low inter-region heattransfer coefficient (IHTC) value to better approximate theeffects of thermal stratification. This is the approachtaken in the FPL RETRAN base models.

INTER-REGION HEAT TRANSFER COEFFICIENT

The IHTC has been shown to have little effect on thepressure response unless high values in the order of 30,000to 50~000 Btu / hr-ft -F are used. A low value of 50 Btu /hr-ft -F is used in the FPL RETRAN models to compensate forthe instantaneous mixing assumption in the liquid region.

BUBBLE RISE AND RAINOUT VELOCITIES

These two velocities are important in de-pressurizationsituations where the saturation temperature in thepressurizer decreases. As the temperature decreases,liquid droplets start forming in the superheated vaporregion and vapor bubbles form in the subcooled liquidregion. The water droplets fall onto the interface at acertain velocity called rainout velocity. Vapor bubbles in

the liquid rise to the surface or interface between theliquid and the vapor at a certain velocity called the bubblerise velocity.The rainout velocity is specified by the user and is keptconstant by the code throughout the calculation. Typically,values between 1 and 5 ft/sec are recommended. Thisparameter has no impact during insurges into thepressurizer. The value of 5 ft/sec is used in the FPLRETRAN models.

The user has several options for the bubble rise model. Inaddition to the bubble velocity, the user can also select abubble gradient to represent the increase in bubbleconcentration with elevation . in the mixture region. Therange of gradients is between 0 and 1 where 0 corresponds tothe case with a homogeneous distribution of bubbles in themixture. Bubble velocity and gradient define how quicklyvapor in the liquid region moves into the vapor region. Forregions that require the existence of a steam dome (e.g.,pressurizer and steam generators) and well defined phaseseparation in the mixture region, values of 0.8 and 3.0ft/sec for the gradient and initial bubble velocity arerecommended (Reference 4) and have been incorporated intothe RETRAN models.

HEAT TRANSFER TO AND FROM PRESSURIZER WALL

During insurge transients, the colder metal in . thepressurizer wall tends to absorb some of the heat from thesuperheated vapor region thus reducing the pressurizationrate from the values that could be observed if wall heattransfer did not occur. Similarly during outsurges thepresence of heat transfer from the hotter wall tends todecrease the de-pressurization rate.

Heat transfer to or from the pressurizer wall is notmodelled in the FPL RETRAN models. In this case, it wasdetermined that the conservatism in not modelling theeffects of the pressurizer wall 'heat transfer (both forinsurge and outsurge events) would be the appropriate choicefor the base models.

— SPRAY OPTION

The RETRAN code offers two options to model the spray fluidinto the pressurizer, one de-superheats the vapor regionwhile the other does not. The second option tends to yieldhigher peak pressures on insurge transients and has beenselected for the FPL RETRAN models.

H—»'~ ~

OVERALL NODALIZATXON

The nodalization approach for the three FFL RETRAN basemodels is shown in Figures 1.2-1, 1.2-2 and 1.2-3. The threemodels are very similar in nodalization with the onlydifferences being due to the geometric differences betweenCE and W NSSS. The nodalization for these models isdiscussed in some detail in Reference 1.

For analysis of transients within the Decrease in HeatRemoval by the Secondary System category, the nodalizationdetail in regions such as the vessel and hot and cold legsis not important. The level of detail found in the threeFPL RETRAN base models is more explicit than required forthe analysis of Decrease in Heat Removal by the SecondarySystem events presented in this report. Simplified basemodels with combined volumes in the reactor vessel and otherparts of the system could have been used with practicallythe same results. For consistency reasons, however, a singlenodalization approach that can be expanded to include amulti-node steam generator when needed has been preferredfor the three FPL RETRAN base models.

- STEAM GENERATOR NODALXZATXON

The most important consideration in the modelling of thesteam generators is whether or not a single node approach isadequate in the simulation of certain events. Based onindustry experience with this issue, . FPL has resolved toretain the single node approach for its base models. Thisapproach is valid for most applications of the models whereprediction of steam generator level response is not crucial.For situations where level is important (e.g., situationswhere reactor trip is on low level), the multi-node modelsfor'ach plant (Figures 1.2 4 and 1.2-5) will be attached totheir respective RETRAN base models.

1.2.3.3 Safet Valve Modellin

The safety valves in the existing RETRAN base decks are.modelled in two parts: a fill junction to model the dischargeflow and a valve to ensure initiation and termination of the fillflow at the selected setpoints. The discharge flow is entered asa table of pressure dependent values based on the Moody criticalflow correlation. Design flow is assumed at the opening setpointpressure and it is allowed to= increase if the system pressureincreases. Accumulation and hysteresis are not assumed. In theTurkey point model one-second ramps are used to model the openingand closing of the pressurizer safety valve at the respective

f

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I

setpoints and half-second ramps are used for the secondarysafeties. In the St. Lucie base models the opening and closingof all safety valves is, assumed instantaneous at the respectivesetpoints.

1.2.3.4 Steam Generator Tube Plu inCurrent plant levels for steam generator tube plugging have

been approximately incorporated into the respective RETRAN basemodels. Values of 04 for Turkey Point and 54 for the St. LucieUnits, are used for the SG tube plugging in the base models.

1.2.3.5 Control S stems

Control systems are modeled in all the base models torepresent as closely as possible the operation of thecorresponding plant systems. Table 1.2-1 summarizes the controlsystems available in the respective FPL RETRAN models. The mostelaborate controller in the base decks is that of the Steam DumpBypass System developed for both the St. Lucie and the TurkeyPoint models. These systems have been described in detail inSection 1.2.1. A less complex system is the Feedwater ControlSystem which currently is only available for the Turkey PointRETRAN model. Such a system facilitates the analyses of planttransients and will be developed for the St. Lucie models in thefuture.

In addition to that of the the feedwater controller, twoadditional differences can be noted between the control systemsused in the different FPL RETRAN base models. One is in themodelling of the PORVs which at Turkey Point require the actionof a controller to simulate the operation of one of the twoavailable PORVs to generate an anticipatory opening signalearlier than at the pressure setpoint. The other difference isin the need for a controller for the Atmospheric Dump Valves atTurkey Point." Neither the PORV's or the ADV's operate in amanner which requires a control system at the St. Lucie Units.

TABLE 1 ~ 2-1

FPL RETRAN BASE MODELS SUMMARY

TURKEY POINT ST LUCIE 1 ST LUCIE 2

PRESSURIZER

-Non-Equilibrium Volume YES YES YES

-Metal Heat

-Spray Option

NO

YES

NO

YES

NO

YES

ENTHALPY TRANSPORT YES YES YES

TEMPERATURE TRANSPORT NO NO NO

STEAM GENERATOR

-Multi-node

-Tube Plugging

NO

NO

NO

YES

NO

YES

CONTROL SYSTEMS

-Pressurizer Heaters

-PORV Opening

YES

YES(1 out of 2)

YES

NO

YES

NO

-Feedwater Flow YES NO NO

-Steam Bypass YES YES YES

-Atmospheric Steam Dump YES NO NO

Setety85 Votree Sd

CONTAINMENT(SINKVOLUME)

99To Torbtoe

Sotety88 Votreo by

Q)

Qss

Hoor)or

94Qs)

91yteyet SetetyVotoee Veloce

Steeo)ON)V)11 01st

Steer)ISypeso

I5'4)

Qs)

98 105 97

QH

Oo

bt 89MFVy

Heetere

Q»

3dQ»

oo

3 Qs

Q»

22Qe II 0

~ IO

po10

100 103

» o)o

p.

26

25

OI Charb)oy

101 tde

21 A1

Qo

1PP

19

22

O voLVMZS 105

Letdown

OI

A2(n)

1PP

I)SAT CONOUCTOR

WIIOttSFIGURE 1.2-1

St. Luoie 1 RETRAN Base Model

-lQss

Cvh CONTAINMENTISINKVOLUME)

99To Turbine

Qss

Sar cry88Valeea 8y

Qs<

91Aeoel SalaryValsee Valrree

SsennDump(1oss)

SSennDumpfXss)

Sse am

Bypaaa15'%)

Qsl 0$ 2

41 89

3s

s

.35sots

Qs

Loop8

~ess

Qn

ss n 2213

a" nsasa

sa sr

Qs

2PPa

10

I Ctwyty

100 103

nn sa Qn

—~Oas Charoln9

2$ 10'I S 0l

Al0»

1PP

19

O VCL~

HEAT CNNUCTOR~ JUSICTlOSSS

FIGURE 1.2-2St. Lucie 2 RETRAN Base Model

2e

1PP23

s lsrsl

~ ~ ~ ~ ~ ~ ~ ~

0

P

E

'r

91 92k

051

151 152

132

141

12

012 l

142112

FW 61

221 211 4 212 222

16

10 7 Q5 Q4 5(y 9 9 Q1

17

214 15 215

e Qe Q3 4(y 8 Q1 Q1 O116

18

T14

21B

'OOP B LOOP A

PIGURE 1.2-4St. Lucie Multi-node Steam Generator Model

-13-

0

Ill

42

241 141 131

112

12

12

14

110 15

10Gj) Cm)

25 31

16

10

TWO LOOP

(LOOP A&G)

SNGLE LOOP(LOOP B)

FIGURE 1.2-5Turkey Point Multi-node Steam Generator Model

-14-

2 ' Turke Point Unit 4 Loss of Inverter

2.1 Summa of Events

On June 20, 1985, Turkey Point Unit 4 experienced a reactortrip from 1004 power. The initiating event was the tripping ofthe 4C inverter which was supplying power to the 120 volt vitalinstrument panel 4P06. The loss of the inverter initiated aturbine runback due to the loss of power to a 'nucleari'nstrumentation system channel. Xn addition, Loss of 4P06 de-energized the pressurizer level and spray valve controller s(causing the spray valve to remain at its last demand position).De-energizing of the level controller caused a false indicationof low pressurizer level (less than 144) which in turn de-energized the pressurizer heaters {control and backup) andinitiated letdown isolation.

Loss of the 4C inverter also resulted in the loss ofautomatic operation of one of the two Power Operated ReliefValves {PORV), the other PORV was available but had its blockvalve closed due to leakage problems. These conditions resultedin the reactor coolant system pressure increasing until itreached the pressurizer high pressure reactor trip setpoint of2385 psia which initiated an automatic reactor trip. Zt shouldbe noted that the Technical Specification setpoint is 2385 psigfor the high pressure trip. Plant procedures provide a 15 psiauncertainty allowance for instrument. drift, therefore thesetpoint applicable for this event is 2385 psia.

Pressurizer pressure decreased after the trip and continuedto decrease because of the de-energized pressurizer spray valvecontrollers which maintained spray flow even as the pressuredecreased.

Loss of the 4P06 panel also caused the "A" Steam Generator(SG) feedwater level controller to transfer from automatic tomanual. Feedwater to SG "A" remained at the 100% power flowrate during the early stages of the transient. Loss of automaticlevel control along with continuous supply of 1004 feedwater flowresulted in the "A" SG level increasing until it reached the Hi-Hi level setpoint {804). Both SG feedwater pumps are tripped dueto reaching the high SG level setpoint about one minute after thereactor trip. This resulted in a feedwater isolation signal andan automatic start of the auxiliary feedwater pumps.

A varxe y o ot f countermeasures were taken by the operators tobout't' th ooldown caused by the feedwater trans'.ent. At a ou

seventeen minutes into the event, power to the 4P06 panel wasrestored and the lost instrumentation on Unit 4 was regained.

-15-

'E y,t~l

2

I

-fl

2.2 A~nal sisAnalysis of plant events can be performed in order to

validate the ability of the simulation model being used toaccurately represent actual plant performance. The plant eventdescribed above in Section 2.1 has been chosen as a benchmark dueto the many characteristics which are similar to the licensingtype transients found in the Decrease in Secondary Heat Removalby the Secondary System category of events.

Specifically, this plant event demonstrates a primary tosecondary heat generation/removal mismatch sufficient to resultin a primary pressurization to the reactor trip setpoint.Another similarity was that the PORVs were not, available tomitigate the primary pressurization, as in most. licensingtransients. Even though the over-filling of SG "A" and thepossibility of operator action are not consistent with alicensing transient, this plant event, within the first 60seconds, clearly offers a valid means to compare the predictivecapability of the component models within RETRAN to the actualplant response to a decrease in the secondary system heatremoval.

The plant event was analyzed with the initial conditionssummarized in Table 2.2-1. A discussion of the assumptions andinitial conditions used in the RETRAN analysis follows. Thebasis for this information is found in References 6-8.

1) Simulation Time

Only the first 60 seconds of the event are provided forcomparison of the plant data to the RETRAN simulation.After the initial 60 seconds, possible operator actionsto try to reduce the cooldown and de-pressurizationcaused by the excess feedwater flow and the continuedspray flow are not well documented.

2) Feedwater and Auxilia Feedwater Flows

Feedwater flow to the "A" SG switches to manual at theinitiation of the event and the MFRV to SG "A" remainsat the 100% power position throughout the first 100seconds. Feedwater flow to SG's "B" and "C" remainedin automatic for the duration of the event. Flow tothe "B" SG was the only feedwater flow recorded on theSystem Parameter Display System (SPDS) and thatinformation is provided at 10 second intervals.

-16-

For the RETRAN analysis, flow to the "C" SG is assumedto be the same as that to the "B» SG and is used as aboundary condition for the analysis. The flow'or the"A" SG is kept constant at the initial value. Afterreactor trip, feedwater temperature is reduced based onthe available measured data. These feedwater flowanalysis assumptions have some uncertainty since noinformation is available for SG "A" and "C" feedwaterflow and the potential exists for the operator to havetaken action to reduce feedwater flow in order tomitigate the primary cooldown that was occurring duringthe event.

One auxiliary feedwater pump was undergoing routinetesting at the time of the event. The total pump flowof 375 gpm. as indicated in the Test Procedure, isdivided into 125 gpm per SG and kept constant for theduration of the RETRAN simulation.

Pressurizer S ra and Heaters

All pressurizer heaters were on at the time of theevent initiation trying to compensate a faultypressurizer low level indication. The pressurizerspray valve was 10$ open to compensate for the heaters.At initiation of the simulation the heaters are lostwhile the spray valve is kept at the 10% open position.

Pressurizer PORVs

Of the two PORVs available at the plant, one wasisolated for leakage problems at the time of the event,while the other was lost with the inverter failure. Enthe RETRAN simulation the two PORVs have been assumedunavailable.

Turbine Runback

As a result of the loss of inverter, the 'turbineexperienced a runback to 70% power. This is modeled inRETRAN by decreasing the steam flow at the designrunback rate of 200% /minute.

-17-

TABLE 2.2-1

KJRKEY POINT UNXT 4

XNXTIAL CONDITXONS FOR LOSS OF XNVERTER W/ENT

PLANT CONDITIONS VALUE

POWER LEVEL (4 OF NOMINAL)

TIME IN CYCLE 11

PRESSURIZER PRESSURE (PSIA)

PRESSURIZER LEVEL (0 NR)

COLD LEG TEMPERATURE (DEG.F)

PRIMARY COOLANT AVERAGE TEMPERATURE (DEG.F)

CHARGING FLOW (GPM)

STEAM GENERATOR PRESSURE (PSIA)

STEAM GENERATOR LEVEL (4 NR)

FEEDWATER TEMPERATURE (DEG.F)

AUXILIARYFEEDWATER FLOW / SG (GPM)

100

MOC

2255

51.7

553

572

72

825

60.6

427

125

6) Plant Initial Conditions

Initial Pressurizer and SG pressures, cold . leg andaverage temperatures, pressurizer level and chargingflow for the .RETRAN model have been obtained fromReference 6 and are shown in Table 2.2-1.

7) Turbine B -Pass

With the loss of the inverter, automatic operation ofthe Turbine By-Pass system is not available until afterthe reactor trip occurs.

2.3 Results

The results of the RETRAN analysis are presented in alisting of the Sequence of Events shown in Table 2.3-1. Adetailed discussion of the RETRAN results is found in Section2.3.1. Section 2.3.2 provides a discussion of the maindifferences between the results predicted by RETRAN and the plantresponse. Section 2.4 describes the results of a variety ofparametric studies performed with RETRAN to better understand thecode capabilities and limitations in modelling these type ofevents.

2.3.1 ETRAN Anal sis

2.3.1.1 Seconda S stem Res onse

Following the loss of the 4C inverter, at the initiation ofthe event, the turbine admission valve closes to reduce steamflow from 1004 to 704 in 9 seconds (turbine runback). Turbinerunback is occurring since the loss of the inverter initiatesactions as if there was a control rod drop event. The turbinerunback is designed to reduce secondary heat removal in order tobetter match the reduced core power expected due to the insertionof a control rod. Since no control rod action actually happened,this produced a mismatch between the power generated by the coreand the power removed by the secondary system which results in aheatup of the secondary system with 'the subsequent heatup of theprimary. Figure 2.3-1. shows the ensuing secondary pressureresponse for Sois »B» and »C»

-19-

!

~ I

'TABLE 2-3-1

SEQUENCE OP EVENTS

RETRAN ANALYSIS LOSS OF INVERTER EVENT

EVENT TIME seconds SETPOINT OR VALUE

Loss of Inverter 0.0

Turbine Runback 0.1

High Pressurizer Trip 17 ' 2385 psia

Rods Begin to Drop 19.7

Peak Pressurizer Pressure 20. 6 2428 psia

Turbine By-pass Actuates 20.8

Turbine Stop Valves Closed 21.2

Rods Fully Inserted 22.1

Peak SG Pressure 28 ' 982 psia

-20-

0

I

4 '~

The secondary pressure response in RETRAN shows a smooth rateof increase until the reactor/turbine trip. After the turbine stopvalves are fully closed, 21.2 seconds after the initiation of theevent, the secondary pressurization increases more rapidly untilthe Turbine By-pass system reaches full flow at 24 seconds. TheTurbine By-pass system monitors the difference between primarytemperature and the reference No-Load temperature. As primarytemperature decreases after the reactor trip, the Turbine By-passvalves close (prior to the actual attainment of the No-Load Tavgdue to the anticipatory action of the control system), with bothg or ups fully closed at approximately 45 seconds into the transient.With the closure of the Turbine By-pass system, the raterate ofsecondary system pressure decrease is reduced as shown on Figure2.3-1 at 45 seconds.

The secondary pressure decrease seen from about 50 seconds to60 seconds reflects the analysis assumptions made relating to themain ee wa er af d t ddition to SG "A" and the reduction of feedwater

to SG "A"temperature over time. Maintaining 100% feedwater flow in oat a reduced temperature along with normal main feedwater flow toSG's "B" and "C" is more than sufficient to remove decay heat fromthe primary without pressurizing the secondary system. Thisexpected behavior of the secondary system pressure is shown onFigure 2.3-1- .

The secondary system pressure response for SG "A" is shown on2 3-2. No plant data is available for comparison with the

RETRAN prediction of the SG "A" behavior. During the ini 'Figure . - . o

of the transient, the SG response is very similar to that shown onFigure 2.3-1. This is as expected since no real deviation betweenthe three SG's was acknowledged at the plant and the RETRANanalysis assumes no differences until after the reactor trip whenlow to SG "A" is assumed to be maintained at 1004 flow. In fact,

SG pressure response for the three SG's is basically the same untilthe Turbine By-pass valves close. At that point, the addition ofmain feedwater to SG "A" produces a faster de-pressurization of theSG. At the 60 second end point for this comparison, there iscalculated to be a 12 psia reduction in pressure in SG "A" relativeto that shown in Figure 2.3-1.

2.3.1.2 Prima S stem Res onse

As the secondary temperature and pressure increases, theprimary coo an el t temperature also increases as shown in Figure 2.3-3. At approximately 23 seconds the primary hea upt d b th reactor trip on high pressurizer pressure of 2385sia and the action of the Turbine By-pass system. The RETRANRAN data

shown in Figure 2.3-3 includes a time delay corresponding to theeffect found in the Resistance Temperature Detectors (RTD's)

th 1 t. The reactor trip causes the turbine to tripe the enerand the Turbine By-pass system to open and relieve e gy

accumulated in the system. The Turbine Bypass valves operate basedon the difference between actual primary temperature and areference No-Load temperature.

-2l-

~ ~

1

F

,'El%

0,

As shown on Figure 2.3-3, the primary temperature decreasessmoothly until the Turbine By-pass system closes. It can be seenthat the rate of temperature decrease changes after the TurbineBy-pass system closes. Primary temperature continues todecrease, however, due to the combination of continued additionof 100% feed flow into SG "A" and continued addition of chargingwith letdown isolation.

The primary system power is shown in Figure 2.3-4. Theinitial response to the decrease in heat removal by the secondarysystem is a slight power decrease due to the action of thenegative moderator temperature coefficient. Power decreasesrapidly after the reactor scram as the rods begin to be insertedat 19.7 seconds after the initiation of the event.

i

The pressurizer pressure as calculated by RETRAN is shown inFigure 2.3-5. As shown, the primary system heatup caused by thedecrease in secondary system heat removal results in an insurgeinto the pressurizer. The pressurizer pressure increases andreaches a maximum of 2428 psia at 20.6 seconds into the event.Pressurizer pressure after that point decreases steadilythroughout the rest of the simulation as primary temperaturedecreases and as the pressurizer sprays continue to operate atapproximately 10% of full capacity.

The pressurizer level response calculated by RETRAN is shownin Figure 2.3-6. As the heatup progresses, the level increasesdue to the insurge into the pressurizer. A maximum level isreached at 20.8 seconds into the simulation. Similar to thepressurizer pressure, level thereafter decreases throughout thesimulation. A change in slope of the decrease in pressurizerlevel occurs at 26.6 seconds. This change in slope in the leveldecrease corresponds to a reduction in the rate of de-pressurization calculated by RETRAN as shown in Figure 2.3-5.

2.3.2 RETRAN Com arison to Plant Data

The data calculated by RETRAN for the Loss of Inverter eventhas been described in Section 2.3. The sequence of events ascalculated by RETRAN and measured at the plant is compared inTable 2.3-2. As previously mentioned, the plant data was onlyavailable at 10 second intervals in most cases. Therefore datapresented as maximum only represents the maximum data pointrecorded.

In general, the RETRAN calculations show responses similarto the measured plant data. Specific comparisons show that theRETRAN model reacts slightly slower to the event than the actualplant. This can be observed most clearly by examining theresponse of SG "B" and "C".

)f„', Oi

As shown in Figure 2.3-1, the SG model in our base RETRANcalculation shows a slower pressurization during the period ofturbine runback (0 - 21 se'conds). Aftex turbine trip however,the SG pressurization in RETRAN is faster and results in a higherpeak SG pressure. The model then starts to de-pressurize due tothe action of the =Turbine By-pass system and the action of thefeedwater, both the reduction in feedwater tempexature as well asthe continuation of full feedwater flow to SG "A".

The plant data demonstrates a de-pressurization afterturbine by-pass system operation, however, the plant data after40 seconds fox SG "B" shows a stabilization of pressure not seenin RETRAN. The effect of a reduced feedwater temperature in theRETRAN calculation appears to be the cause of the difference.Instantaneous mixing of the fluid within the single node SGresults in an overprediction of the impact the colder feedwaterhas in reducing secondary temperature and pressure.

Comparisons between the other key parameters and the RETRANresults show that RETRAN predicts the same general trends as theplant data. The RETRAN primary pressuxization is larger andtakes longer to be reduced than the plant data. This differencemay be due to the fact that the RETRAN model ignores pressurizerwall heat transfer.

!

Primary temperature differences between RETRAN and the plantdata show that the RCS temperature trends are very'imilar.RETRAN predicts a peak temperature somewhat later than the plantdata, however the cooldown rate after the peak matches wellbetween the plant and RETRAN with a relatively constantdifference after 38 seconds.

Pressurizer level comparisons show a slightly larger levelincrease at the plant during the turbine runback with the longterm level response showing almost exact agreement after about 38seconds.

One key conclusion that can be drawn from these comparisonsis that the single node SG 'model produces peak primary andsecondary pressures which are conservative relative to the plantdata. This is a key item when considering modelling applicationsfor the licensing events.

-23-

TABLE 2 3.2-1

SEQUENCE OF EVENTS

TUEUG"Y POINT UNIT 4 LOSS OF INVERTER EVENT

RETRAN COMPARISON TO PLANT DATA

EVENT Plant DatTIME (seconds)

gETRAN

Loss of Inverter 0.0 0.0

Turbine Runback 0.1 0.1

High Pressurizer Trip

Rods Begin to Drop 19 '

17.7

19.7

Peak Pressurizer Pressure 20 ' 20.6

Turbine By-pass Actuates 20.8

Turbine Stop Valves Closed 21.2

Rods Fully Inserted 21 8 22. 1

Peak SG Pressure 28.7 28.5

-24-

Jt„

~

~

I

l4+

E

2.4 SENSITIVITY STUDIES WITH RETRAN MODELS

The Loss of Inverter Event at Turkey Point has been selectedto assess the impact of various RETRAN modelling techniques andoptions on the prediction of heatup events. These effectsalthough investigated with the Turkey Point RETRAN model are alsoapplicable to the St. Lucie models in the analysis of heatupevents. Since, as described in Section 1.2, the three FPL RETRANbase models have a very similar nodalization approach and thephysical phenomena involved in heatup events is the same, it isreasonable to expect applicability of relative effects from onemodel to the others. This is confirmed by the similarity of theresults of the Loss of Condenser Vacuum analyses performed withthe three FPL RETRAN models (see Section 4.0).

The relative impact of the modelling techniques and codeoptions has been evaluated against the two following criteria:

— Fidelity of the RETRAN prediction to the plant data.

Conservatism of the RETRAN prediction with respect toplant data.

The results of these sensitivity studies support the choicesin modeling techniques made for the licensing methodology toanalyze Decreases in Heat Removal by the Secondary Systempresented in Section 4.0.

The various sensitivity studies performed with the Loss ofInverter analysis are described below. A summary of thesestudies is presented in Table 2.4-1.

Pressurizer Inter-Re ion Heat Transfer Coefficient

The Inter-Region Heat Transfer Coefficient (IHTC) waschanged from the value of 50. Btu/hr-ft -F used in the basemodel to a value of 20,000. Btu/hr-ft -F with no noticeableeffect in the predicted peak pressure or its timing.Therefore it can be concluded that changes in the IHTCwithin the above range of values do not impact the keyparameters of this event and that the value of 50. Btu/hr-ft -F used in the FPL RETRAN base models is adequate inpreventing the two regions in the pressurizer from reachingequal temperatures.

S ra 0 tionThe effects of having the spray option activated wereinvestigated and found to cause an increase in predictedpeak primary pressure as expected. The option, however, wasnot kept for the base case analysis of the Loss of Inverterevent on the basis of fidelity to plant data. The effects

-25-

I

I

'g ~

i

of the spray option are not applicable for analysis oflicensing events involving Decrease in Heat Removal by theSecondary System because no credit for the use of the spraysystem is taken in such methodology.

H draulic Resistance

Higher hydraulic resistances in the surge line or at thepressurizer inlet will tend to yield higher peak pressuresin insurge transients. This is because less cold fluid fromthe surge line can flow into the pressurizer. The impact ofthis effect has been investigated with the Loss of InverterEvent. The results show a very small sensitivity withhydraulic xesistances. The increase of the hydraulicresistance at the entrance of the pressurizer from a valueof 0 to a value of 10 resulted only in an increase of 0.9.psia in peak pressure. Based on this, it was determinedthat the FPL RETRAN models will use the nominal hydraulicresistances for both best estimate and licensingcalculations.'em

erature Trans ort Dela

The temperature transport delay model is intended tosimulate the displacement of a temperature front through achannel with little mixing such as in straight pipes. Theeffects of this option have been investigated with the Lossof Inverter Event by subdividing the hot and cold legs inthe RETM model into 10 sections to more accuratelyrepresent the temperature variation throughout the system.The effects on predicted peak pressure and time of the peakare negligible. Therefore this option will not be utilized.

Courant LimitThe version of the code (MOD004) used in this report has adefault value of 0.3 fox the Courant time step controlcoefficient in the iterative numerics solution. This value.can be changed by the user to try to improve running timesas long as the accuracy ox stability of the solution is notaffected. Sensitivity studies have been performed bychanging the base case value of 0.3 to 0.6 and 1.0. Table2.4-1 shows that the results are insensitive to reasonablechanges in the value of the Courant coefficient. Values ofthe coefficient above 1.0 are not considered realistic andcould result in unstable solutions or convergence errors.

-26-

2.5 Conclusions~~

The Loss of Inverter Event has been analyzed with RETRANusing the known plant inital conditions and equipmentactuations during the transient as boundary conditions. Theresults of the comparison shows the same trends for theparameters and general agreement in timing and magnitude.The key parameter for this type of event, primary systempressure, was calculated conservatively relative to theavailable data.

The sensitivity studies have reviewed the impact of varyinginputs and the conclusions derived will be used indevelopment of the licensing methodology discussed inSection 4.0

-27-

TABLE 2-4-1

RETRAN SENSITIVITY STUDIES

TURKEY POINT LOSS OF INVERTER EVENT

REACTORTRIPTIME

PEAKPRIMARYPRESSURE TIME

PEAKSECONDARYPRESSURE TIME

Base Case 17. 7 2428 20.6 981 28.5

IHTC Variation 17.720,000 BTU/hr-ft -F

2428 20.6 981 28 ~ 5

PressurizerSpray Option

16.3 2430 19-1 975 27.0

Surge LineHydraulic Resistance:

K~= 0.0

K)= 10.0

17 '

17 '

2428

2429

20.6

20.6

981

982

28.5

28.5

TemperatureTransport Delay

17 F 7 2428 20.6 982 28 '

CourantCoefficient (C4)

C~=O ~ 6

C( = 1.0

17 '

17 '

2428

2429

20 '20.6

981

981

28 '

28.5

-28-

FIGURE 2.3-1

1100

4

Turkey Point Loss of InverterRETRAN Benchmark Analysis

STEAM GENERATOR PRESSURE "B"AND "C" vs. TIME

1080

1060

1040

1020

1000

980

960

940

920

900

880

860

840

820

800

40

RETRAN Data

TlME (SECONDS)

Plant Data

FIGURE 2.3-2

1100

Turkey Point Loss of InverterRETRAN Benchinark AnalysisSTEAM GENERATOR "A"PRESSURE vs. TIME

1080

1060

1040

1020

980

960

940

920

900

880

860

840

820

40

71MB (SECONDS)

-30-

FIGURE 2.3-3

590


RCS TEMPERATURE vs. TIME

580

560

SS0

540

530

40

REIRAN Data

llME{SECONDS)

Hant Data

-31-

FIG&&2.3-4

1.2


REACTOR POWER vs. TIME

0.9

0.8

0.7

0.6

0.5

0.4

0.3

0.2

0.1

40

TIME (SECONDS)

-32-

FIGURE 2.3-5


PRESSURIZER PRESSURE vs. TIME

R

fomI 2100

1900

1800

40

RETRAN Data

TIME (SECONDS)

Plant Data

-33-

FIGURE 2.3-6


PRESSURIZER LEVEL vs. TIME

50

40

30

10

40

RETRAN Data

TIME(SECONDS)

Plant Data

-34-

3.0 St. Lucie Unit 1 Partial Loss of Feedwater Flow~~ ~

3.1 Summa of Events

On September 20, 1988 a plant trip occurred at St. Lucie Unit1. The initiating event was a loss of power to the "B" Main FeedRegulating Valve (MFRV) controller. The loss of power caused the"B" MFRV to shut and a loss of feedwater to the "B" Steam Generatorto occur. The loss of feedwater resulted in a reactor trip due toreaching the low SG level setpoint, set at the plant atapproximately 39.25% of narrow range. Following the trip, anAuxiliary Feedwater Actuation Signal (AFAS) was generated.

In this transient, the primary pressure did not. reach thePower Operated Relief Valve (PORV) setpoint of 2400 psia. The MainSteam Saf ety Valves (MSSV) opened to relieve secondary systempressure in conjunction with the operation of the Steam Dump andBypass System (SDBS). After reset of the MSSV's, decay heat isremoved through action of the SDBS. No unusual actuations oroperator actions occurred.

3.2 '~nal sis~~

The RETRAN analysis of a plant transient can be performed tovalidate the ability of the model to accurately predict actualplant performance. The key RETRAN component model which is beingexamined with this transient is the multi-node model used fortracking the behavior of the water level in the steam generators.A noding diagram for the multi-node steam generator model is shownin Section 1.2. This transient was chosen since it demonstratesa loss of secondary system heat removal which results in a lowsteam generator trip rather that a high pressurizer pressure trip.The characteristics of the Loss of Normal Feedwater or Loss of ACtype events within the category of Decrease in Heat Removal by theSecondary System are similar to the event that occurred at St.Lucie Unit 1.

One of the difficulties in examining the comparison of RETRANto the plant results in this case is the lack of data from theplant instrumentation. Most of the parameters'f interest wereonly recorded every 10 seconds. The data available is plotted aspoints on the comparison graphs rather than as lines since linearinterpolation over such a large time interval is not representativeof the way the plant responded.

-35-

leap

T'i,

WI 0

This plant event was analysed using the initial conditions~ ~ ~

~

~

~

~ ~

~ ~~

~ ~ ~

~

~

summarized in Table 3.2-1. A discussion of the assumptions andinitial conditions follows. The basis for the plant data is foundin References 9 and 10.

Simulation Time

Only the first 100 seconds of the event are provided forcomparison of plant data to the RETRAN simulation. Thisevent is an uncomplicated reactor trip on low level anddata beyond 100 seconds provides no insights on theability of the calculational models to accurately predictplant response during the time period important tolicensing analysis.

2) Feedwater Assum tionsThe main feedwater flow to SG "B" is ramped to zero in10 seconds after initiation of the event. Main feedwaterto SG "A" is assumed to be in automatic mode of control.After the reactor trip signal, main feedwater to SG "A"was also lost with flow being reduced to zero in 10seconds. While data on feedwater flow to SG "A" wasavailable for each second of the event, data for SG "B"was not available and the assumption of a linear rampdownwas chosen.

3) Pressurizer Pressure Control S stem

The pressurizer pressure control system is assumed to bein automatic mode and is available when needed.

4) Plant Initial Conditions

Plant initial conditions are shown in Table 3.2-1. Theinitial conditions for the key parameters of interest inthis event, i.e., initial SG pressure and level, wereavailable from the plant data.

5) eactor Protection Dela Times

The time delays associated with the action of the ReactorProtection System (RPS) is taken from the plant sequenceof events recorder. The time delays recorded for openingreactor trip breakers and closing of the turbine stopvalves are significantly shorter than the valuestypically assumed. For purposes of evaluating thisevent, the time delays as recorded at the plant will beassumed in the RETRAN calculation.

-36-

TABLE 3.2-1

,ST. LUCIE UNIT 1

INITIALCONDITIONS FOR PARTIAL LOSS OF FEEDWATER EVENT

PLANT CONDITIONS VALUE

POWER LEVEL (4 OF NOMINAL)

TIME IN CYCLE 9

PRESSURIZER PRESSURE (PSIA)

SG "A" LEVEL (4'R)SG "B" LEVEL (4 NR)

SG PRESSURE (PSIA)

COLD LEG TEMPERATURE (DEG. F)

RCS AVERAGE TEMPERATURE (DEG. F)

FEEDWATER TEMPERATURE (DEG. F)

100

BOC

2250

65.9

69.9

888

549

574

434

-37-

3.3 Results

The results of the RETRAN analysis are provided in Section3.3.1. Section 3.3.1 discusses the RETRAN analysis resultsrelative to the analysis assumptions and RETRAN modelling used forthis event. Section 3.3.2 provides a discussion of the maindifferences between the results predicted by RETRAN and the plantdata.

3.3.1 RETRAN Anal sisA sequence of events as calculated by RETRAN is provided in

Table 3.3-1. As shown, the key actions during this transient arefound in the secondary system, which is discussed in Section3.3.1.1. The response of the secondary system is described overtwo time intervals due to the different actions occurring. beforeand after reactor/turbine trip.

3.3.1.1 Seconda S stem Res onse

1) Time Interval 0 to 21 seconds

Following the loss of power to MFRV "B", main feedwater to SG"B" is assumed to ramp down to zero flow in 10 seconds. Asthe feedwater flow is reduced, the SG level for SG "B", Figure3.3-1, shows a gradual reduction. When all feedwater isisolated, the level begins to reduce much more rapidly withthe low level signal occurring at 21.6 seconds after theinitiation of the event. The SG "B" pressure response isshown in Figure 3.3-2. The use of the multi-node SG modelprovides a much more sensitive response to small steam flowvariations which are occurring between the two SG's than theuse of a single node model. As shown in Figure 3.3-2, the SG

pressure. increases after feedwater flow has stopped. Thepressure increases until action of the SDBS at 20 secondsbegins to mitigate any further pressurization.

SG "A" behavior is shown in Figures 3.3-3 and Figures 3.3-4.For the first 21 seconds of the event, RETRAN calculates onlya small increase in SG level. This increase is due to aslightly increased MFW flow to SG "A" that occurs when the"B" MFRV closes as both MFW pumps continue to operate. Thedrop in pressure that begins at approximately 20 seconds isdue to the opening of the SDBS which has reacted to thepressurization of SG "B".

-38-

TABLE 3.3.1-1

SEQUENCE OF EVENTS

RETRAN ANALYSIS PARTIAL LOSS OF FEEDWATER

ST. LUCIE UNIT 1


Loss of Power to MFRV

MFRV SG "B" Fully Closed

SDBS Begins to Open

Low SG "B" Level TripReactor Trip Breakers Open

Turbine Stop Valves Closed

SDBS Full Capacity

Rampdown of Feedwater to SG "A"

CEA's Fully Inserted

SG "B" MSSV's Open

SG "A" MSSV's Open

SDBS Begins to Modulate Flow

MSSV's Close

0.0

10.0

20.0

21.63

21.67

22 '723.0

25.0

25.1

25.5

26.0

41. 0

44.0

39.25 4 NR

AFAS Signal

1000 psia

1000 psia

920 psia

-39-

~ 1

h

~ I

2) Time Interval 21 to 100 seconds

After the reactor trip and turbine trip, RETRAN calculates aswift increase in secondary pressure until the MSSV firstbank setpoint is reached at 25.5 seconds for SG "B". Theaction of the MSSV's and the SDBS reduce the pressure untilthe MSSV's close. Plant data showed that the pressure forfully closing the MSSV's was at 920 psia. After the closureof the MSSV's, RETRAN predicts a slight re-pressurizationwhich is turned around by the continued action of the SDBS.The changes in de-pressurization rates seen at 32.0 and 37.0seconds are related to the sensitivity of the multi-nodemodel. The changes in de-pressurization rates are tieddirectly to observed steam flow changes as both SG's competeto provide flow to the SDBS and the MSSV's. The increase in"B" SG level starting at approximately 27 secondscorresponds to the initiation of the secondary systempressure turnaround through action of the MSSV's and SDBS.This represents only a short term- effect since there is acontinuation of steam flow with no feedwater additionthroughout this time interval. The level decrease is slowedwhen the MSSV's close and the SDBS begins to modulate andshows only a small rate of decrease until the end of thesimulation.

The response of SG "A" is very similar in this time frame toSG "B" discussed above. The changes in SG "A" pressureafter the turbine trip are associated with the changes inthe steam flow. The SDBS acts as a constant demand in theRETRAN model which combined with the higher and earlierpressurization of SG "B" results in the behavior seen onFigure 3.3-3. In addition, the MSSV model used in theevaluation of this event is the simple flow versus pressuremodel discussed in . Section 1.1. This model calculatesoscillatory flows, that is, when pressure increases the flowwill increase which in turn reduces the pressure whichreduces the flow and so on. This results in the pressurespike behavior seen in the time interval of 23 to 30seconds.

The level response shown in Figure 3.3-4 shows a decrease inlevel which occurs when the feedwater flow is lost to SG "A"and the pressure increases due to the turbine trip. Thechange in slope in the level decrease which is seen atapproximately 32 seconds is related to the reduction inpressure and correspond to the changes in slope shown inFigure 3.3-3.

-40-

l

0

'\h '

3.3.1.2 Prima S stem Res onse

The primary system response as calculated by RETRAN followsthe effects seen in the secondary system. As the feedwater to SG"B" ramps to zero, RETRAN calculates a small insurge to thepressurizer and a consequent increase in primary system pressure.This behavior is shown in Figure 3.3-5. After the reactor trip,the steam relief from the SDBS and the MSSV's are sufficient toproduce a de-pressurization. The rate of de-pressurization isslowed when the MSSV's close at 44 seconds into the event.

The pressurizer level is shown in Figure 3.3-6. Thebehavior of the level is consistent with that calculated for thepressurizer pressure.

The responses calculated by RETRAN for Tavg for loops "A"and "B" are shown in Figures 3.3-7 and 3.3-7 respectively. Afterthe reactor trip, Tavg is calculated to decrease through theaction of the MSSV's and the SDBS. A RTD delay time of 3.0seconds is included in the values plotted for the RETRANsimulation.

3.3.2 RETRAN Com arison to Plant Data

'A sequence of events comparing the RETRAN simulationdescribed in Section 3.3.1 to the plant data is shown in Table3.3.2-1. As shown the calculated time for reactor trip isearlier than the time inferred from the plant data. The basictrends of the data show excellent agreement. In particular, asshown on Figures 3.3-1 and 3.3-4, the SG level response ascalculated by the multi-node SG model in RETRAN shows a closeagreement to the plant data.

8

One source of the differences between RETRAN and the plantcomparisons that was investigated was the short delay timesmeasured by the plant and used in the RETRAN simulation. Theresponse of the primary system indicates that the use of a longerdelay time between reactor trip signal and rod motion wouldimprove the timing of the primary system responses considerably.That is, the delay of the reactor scram would result in theprimary cooldown being delayed in the RETRAN simulation andtherefore the comparison would be better to the plant data. TheTechnical Specifications for low SG trip signal delay time is1.15 seconds. Plant sequence of events recorder show completionof the low level trip logic and reactor trip breaker opening totake only 0.04 seconds. Discussion with Plant Staff after thisbenchmark was completed indicates that some additional timeshould have been credited for sensor delay times, however, thisadditional time would add only about 0.1 seconds to the, reactortrip delay. This small additionally delay would not make asignificant difference in the RETRAN results and the comparisonto plant data.

-41-

TABLE 3. 3 2-1

SEQUENCE OF EVENTS

ST. LUCIE UNIT 1 PARTIAL LOSS OF FEEDWATER

RETRAN COMPARISON TO PLANT DATA

VENTTIME (seconds)

Plant Data RETRAN

Loss of Power to MFRV

"B" MFRV .Fully Closed

SDBS Begins to Open

Low SG "B" Level TripReactor Trip Breakers Open

Turbine Stop Valves Closed

SDBS Full Capacity

Reduction Feedwater SG "A"

CEA's Fully Inserted

SG "B" MSSV's Open

SG "A" MSSV's Open

SDBS Begins to Modulate

MSSV's Close

0.0

25.4

25.44

25.94

28.7

29 '

29.0

53.0

0.0

10.0

20.0

21.63

21.67

22.17

23.0

25.0

25 '

25.5

26. 0 ~

41. 0

44.0

-42-

~

~ I

As shown, it is clear that RETRAN is calculating a faster de-pressurization than the plant data indicates. Steam flow was notavailable in enough detail to determine the root cause of thedifferences, however, one possibility is the MSSV model used inthe RETRAN simulation. The noding used in RETRAN places all theMSSV's in the same volume which is then represented by onecomposite valve for each of the two banks. Therefore, our RETRANmodel opens all of the four MSSV's/SG that compose the first bankwhen the pressure setpoint is reached. This modelling would notbe expected to simulate exactly the actual behavior at the plantsince the valves are sure to have differences in opening setpointsas well as being physically separated. Xn addition, it would notbe expected that. all the MSSV's would close at the same setpointas assumed in the RETRAN analysis.

Calculations, therefore, were made which varied the flowcapacity allowed through the MSSV's in RETRAN. The secondarysystem response was closest to the plant data when it, was assumedthat only 75% of the MSSV capacity operated, i.e., only three ofthe four valves operated per SG. The results of that calculationare shown for SG pressure response in Figures 3.3-9 and 3.3-10.These results demonstrate a closer match to the measured datathroughout the event for this parameter.

3.4 Conclusions

A partial Loss of Feedwater Flow event has been examined usingthe FPL RETRAN model. The results of the comparisons to plant datashows good agreement in trends and demonstrates the ability of theSG multi-node model to accurately predict level response.

-43-

~ I

0

FIGURE 3.3-1

St. Lucie 1 Partial Loss ofFeedwaterRETRAN Benchmark Analysis

STEAM GENERATOR "B"LEVEL vs. TIME

80

70

40

30

10

40 80

RETRAN Data

TIME(SECONDS)

Plant Data

-44

FIGUIM3.3-2

1040

1030

1020

1010

1000

990

980

970

960

950

940

St. Lucie 1 Partial Loss of FeedwaterRETRAN Benchmark AnalysisSTEAM GENERATOR "B"PRESSURE vs. TIME

920

910

900

890

880

870

860

850

840

80

TIME(SECONDS)

Plant Data

-45-

FIGURE 3.3-3

1040

1030

1020

1010

1000

St. Lucie 1 Partial Loss of FeedrvaterRETRAN Benchmark AnalysisSTEAM GENERATOR "A"PRESSURE vs. TIME

980

970

960

950

g 930

920

910

880

870

850

840

80 100

REIRAN Data

YlME(SECONDS)

Plant Data

-46-

0

5„

FIGURE 3.3-4

St. Lucie 1 Partial Loss of FeedpumpRETRAN Benchmark Analysis

STEAM GENERATOR "A"LEVEL vs. TIME

80

70

40

10

80 I00

RETRAN Data

TIME(SECONDS)

Plant Data

-47-

FIGURE 3.3-5


PRESSURIZER PRESSURE vs. TIME

N2100

CA

M

1900

1800

40 80 100

TIME (SECONDS)

RElRAN Data Plant Data

-48-

FIGURE 3.3-6

80


PRESSURIZER LEVEL vs. TIME

70

50

40

80 100

REIRAN Data

TlME (SECONDS)

Plant Data

-49-

FIGUfM3.3-7

590

St. Lucie 1 Partial Loss ofFeedwaterRETRAN Benchmark AnalysisLOOP A AVERAGETEMPERATUI&vs. TIME

580

570

560

550

540

530

80 100

TIME(SECONDS)

RIHRANData PIant Data

-50-

FIG&&3.3-8

590

St. Lucie I Partial Loss ofFeedwaterRETRAN Benchmark AnalysisLOOP 8 AVERAGETEMPERATURE vs. TIME

580

570

560

550

540

530

80 IOO

TIME(SECONDS)

RE1RAN Data Plant Data

-51-

FIGUIM3.3-9

1040

1030

1020

1010

1000

990

980

970

960

950

940

930

920

910

St. Lucie 1 Partial Loss ofFeedwaterMSSV Reduced Flow Sensitivity

STEAM GENERATOR "B"PRESSURE vs. TME

890

880

870

860

850

840

80 100

'IIME(SECONDS)

REIRAN Data Plant Data

-52-

FIGURE 3.3-10

1040

1030

1020

1010

St. Lucie 1 Partial Loss ofFeedwaterMSSV Reduced Flow Sensitivity

STEAM GEMS TOR "A"PRESSURE vs. TIME

990

980

970

960

v) ~ 950

940

I 930

920

910

890

880

870

850

840

80 100

TIME(SECONDS)

REIRAN Data Plant Data

-53-

4.0 Safet Anal sis Methods

4.1 IntroductionThe events within the Decrease in Heat Removal from the

Secondary System category are discussed in Standard 'Review Plan(SRP) Section 15.2 (Reference 11). Events in this category arelisted in Table 4.1-1 with the numerical designation found in theSRP. The events in this category typically produce a reductionin secondary side heat removal which causes a heatup andpressurization of the primary side. The event and the associatedpressurization is eventually terminated by a reactor trip.=

The first task in developing the safety analysis methodologyto be used with this category of events is to discuss each eventand determine the limiting transients for each of FPL's plants.After the limiting transients have been identified, the key inputparameters that will be used in the licensing analysis of theevent will be discussed and justified. The analysis of the eventwill then be performed to provide the results of the limitingtransient. This process is repeated for all three units in orderto provide full and complete information in a format designed tofacilitate FSAR updates when the FPL methods are utilized in thefuture. This approach has generated sections which are verysimilar from unit to unit since the parameters which impact theresults of the events are in general the same, irrespective ofthe vendor;

It should be noted that the choice of the limiting event aswell as the limiting set of input parameters is a straight-forward process.'hat is, these events are evaluated to ensurethat under limiting conditions the appropriate acceptancecriteria, as defined in the SRP will be met. For the events inthe Decrease in Heat Removal from the Secondary System category,the key acceptance criteria is maintaining peak pressures, bothprimary and secondary, below 1104 of design. Pressurizer levelmust also be examined since there is a potential for going watersolid in the pressurizer during these events. In addition, forthe Loss of Normal Feedwater type events an additional safetyconcern is maintaining an adequate water level in the SteamGenerators to remove decay heat.

these criteria will form the basis of the determination oflimiting transient and limiting input for the FPL safetymethodology in the Decrease in Heat Removal by the SecondarySystem category of events. While transients within this

category're

analyzed primarily because of the potential to exceed the1104 design pressure limits on the primary and secondary system,other safety criteria (such as Departure From Nucleate BoilingRatio (DNBR) behavior) will be evaluated in the detaileddiscussions for each transient. The major focus, however, in thelimiting transient determination will be on assuring that thetransient (with its associated input) will produce the limitingpressure excursion.

54»

P

'i

7.<

Table 4.1-1

Standard Review Plan Events

Decrease in Secondary Side Heat Removal

SRPEvent Number Event Title

15.2.1 Loss of External Load

15.2.2 Turbine Trip

15.2.3 Loss of Condenser Vacuum

15.2.4 Main Steam Isolation Valve Closure (BWR)

152 5~ ~ Steam Pressure Regulator Failure

15.2.6 Loss of Non-emergency AC Power to theStation Auxiliaries

15.2.7 Loss of Normal Feedwater Flow

15.2.8 Feedwater System Pipe Breaks Inside andOutside Containment (PWR)

-55-

4 1 1 ETRAN MODELLING

Based on the results of the modelling sensitivity studiesperformed for the Loss of Inverter analysis (Section 2.0) andother licensing considerations, several changes to the modellingapproaches used in the base models (see Section 1.2.3) have beenevaluated for the methodology used to examine Decrease in HeatRemoval by the Secondary System events. These changes to thebase models along with the basis for either incorporating thechange or maintaining the current models are described below.The main modelling approaches adopted in the FPL RETRAN modelsfor the analysis of this category of events is summarized inTable 4.1-2.

4.1.1.1 Safet Valves

The safety valves model in the current RETRAN .base decksused for best estimate evaluations, (Section 1.2.3 ') open andclose at the respective setpoint pressures with small delays andstart discharging full design flow at, the opening setpoint,pressure. This flow is allowed to increase as the pressurecontinues to increase. This model has been upgraded forlicensing applications to include accumulation and hysteresis andalso to limit the discharge flow at the design value. Thechanges affect only the fillpart of the model where the criticalflow table is now multiplied by a valve flow area fraction whichis based on the hysteresis curves shown in Figure 4-1. Based onexperimental data (Reference 13) on safety valves of very similardesign that shows that the valves open and close nearlyinstantaneously, the time dependence for the valve opening in themodels has been removed. The opening and closing of the valvesis now modelled as pressure dependent (Figure 4-1) instead oftime dependent. The discharge flow, in the new model, does notreach the design value until the accumulation pressure (openingsetpoint pressure +,3%) is reached.

4.1.1.2 Pressurizer Model

The results of the sensitivity studies performed on some ofthe pressurizer parameters (Section 2..4) support the use of themodelling used in the current RETRAN base models for the eventsof interest. These studies also confirm that peak pressurescould be slightly increased if the surge line hydraulicresistances were increased. However the calculated effect wasvery small and since the current method predicts a conservativelyhigh pressure compared to plant data, changes to the resistanceswill not be included. Therefore no changes to the pressurizermodelling used in the current base decks as discussed in Section1.2 ', are recommended. The modelling to be utilized issummarized in Table 4.1-2.

"56-

)1

'tk

4.1.1.3 Tem erature Trans ort Dela~ ~

The observed sensitivity of the peak pressure to the use ofthe temperature transport delay is small (Section 2.4) and willnot be used for Decrease in Heat Removal by the Secondary SystemEvents.

4.1.1.4 Multi-node Steam Generator Model

The results of the Loss of Inverter benchmark analysispresented in Section 2.0 confirm the use of the single node steamgenerator as a valid nodalization approach for situations wherethe prediction of level is not crucial. The single node steamgenerator model is justified because, in addition to itssimplicity of use, it tends to over-predict the primary peakpressure. Based on this, the single node model will be used inanalyses of Decrease in Heat Removal by the Secondary Systemevents that do not require accurate level predictions or do notinvolve severe losses of steam generator inventories. The use ofthe single node SG model for evaluation of peak pressure for thiscategory of event is consistent with the methods used by otherRETRAN users (Reference 12) as well as that used by the fuelvendors (References 14-16).

For situations in which the evolution of the transientdepends on the level prediction, such as in the Loss of Non-Emergency AC Power for Turkey Point or the Feedwater Line Breakfor St. Lucie Unit 2, the multi-node steam generator model willbe used.

-57-

i<

Y1

TABLE 4.1-2

FPL RETRAN METHODOLOGY

DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM EVENTS

PRESSURIZER MODEL

-Single Node-Non-Equilibrium Option-Inter-Region HTC = 50.0 Btu/hr-ft -F-No Metal Heat-Nominal Hydraulic Resistances

0 R PRIMARY MODELLING OPTIONS

-No Temperature Transport Delay

STEAM GENERATOR

-Single Node if Level is not Required-Multi Node if Level is Required

SAFETY VALVES

-3% Accumulation and Hysteresis Included-Design Flow Assumed at Accumulation-Flow Limited to Design Value

-58-

*

0

)

Secondar Safet 8 steresis

1.0

VALVEFRACTIONALAREA

0.82

0.7

PRESSURE

ClosingSctpoint

OpeningSctpoint

OpeningSctpoint+ 3.0%

Primar Safet 8 steresis

1.0

VALVEFRACTIOliiALAREA

0.7

PRESSURE

ClosingSetpoint

OpeningSetpoint-25 psia

OpeningSctpoint

OpeningSctpoint+ 3.0%

FIGURE 4-1Primary and Secondary Safety Valves Hysteresis in the RETRANlicensing models.

-59-

'k

f'

(

4.2 LIMITING TRANSIENT DETERMINATION8

4.2.1 ST. LUCIE UNIT 1

4.2.1.1 Loss of External Load

A loss of external load event is caused by abnormal, eventsin the electrical distribution network. The turbine is protectedfrom a complete loss of external load by two overspeed protectionsystems which act to trip the turbine if a complete loss of loadoccurs. The Overspeed Protection Controller and the mechanicaloverspeed protection system are both designed to trip the turbineat approximately 111 percent overspeed condition. More detailsof these two equipment protection systems are found in Section10.2.2 of the St. Lucie Unit 1 Final Safety Analysis Report(FSAR). Upon occurrence of a turbine trip above 154 of fullpower at St. Lucie Unit 1 , a signal" would be supplied to thereactor protective system to trip the reactor. Subsequent to theturbine trip, the main feedwater regulating valves would closeand feedwater would be supplied to the steam generator throughthe feedwater bypass valve by the main feedwater pumps.

A fast pressurization transient will result from the closureof the turbine stop valves due to their fast response time ofapproximately 0.25 seconds. The primary side pressurizationresults from the reduced primary to secondary heat transfer whichoccurs with the stoppage of steam flow from the secondary side.However, the reactor trip on turbine trip that would occur wouldquickly reverse the pressurization event and bring the reactor toa safe shutdown condition before a reactor trip on high pressurecould occur. In addition, action of the Steam Dump and Bypass,System (SDBS) would mitigate the pressurization effects of anactual Loss of External Load event.

A bounding Loss of Load event is postulated if credit is nottaken for the reactor trip on turbine trip or action of the SDBS.However, this transient is bounded by the Loss of CondenserVacuum event discussed in Section .4.2.1.3 because of theassumption in that event of an instantaneous loss of mainfeedwater at the initiation of the event. The loss of mainfeedwater increases slightly the heatup of the primary andtherefore increases the pressurization rate of the transient.

Evaluation of the DNBR response to a Loss of External LoadEvent shows qualitatively that DNBR will vary only slightlyduring this transient. DNBR is negatively effected by decreasesin flow and pressure and increases in heat flux and temperature.Transients which can challenge the DNBR limit result from eithersignificant decreases in flow, i.e., Loss of Flow events orsignificant increases in heat flux, i.e, Control ElementAssembly (CEA) Drop. During the Loss of External Load, the,

-60-

4

~ 4P

increase in temperature will be offset by the increase inpressure and only slight variations in DNBR would be expected.At all times in this transient, the Loss of Flow event asdescribed in Section 15.2.5 of the FSAR would remain limitingwith regard to DNBR results.

4.2.1.2 Turbine TriA turbine trip is an event in which the steam flow from the

steam generators is stopped through closure of the turbinecontrol and/or. stop valves. The turbine is equipped with anautomatic stop and emergency trip system which trips the stop andcontrol valves to a closed position in the event of turbineoverspeed, low bearing oil pressure., low condenser vacuum, orthrust bearing failure. Turbine protection devices which couldcause a turbine trip are described in depth in Section 10.2.2 ofthe FSAR. Upon occurrence of a turbine trip above 15% of fullpower, a signal is supplied to the reactor protective system totrip the reactor. Subsequent to the turbine trip, the mainfeedwater regulating valves close and feedwater is supplied tothe steam generator through the feedwater bypass valve by themain feedwater pumps.

A fast pressurization transient will result from the closureof the turbine stop valves due to their fast response time ofapproximately 0.25 seconds. The primary side pressurizationresults from the reduced primary to secondary heat transfer whichoccurs with the stoppage of steam flow from the secondary side.However, the reactor trip on turbine trip would quickly reversethe pressurization event and bring the reactor to a safe shutdowncondition before a reactor trip on high pressure could occur. Inaddition, action of the SDBS would mitigate the pressurizationeffects of a Turbine Trip.

A bounding Turbine Trip could be postulated if credit is nottaken for the reactor trip on turbine trip or action of the SDBS.However, this transient is bounded by the Loss of CondenserVacuum event discussed in Section 4.2.1.3 because of. theassumption in that event of an instantaneous loss of mainfeedwater. The loss of main feedwater increases the heatup ofthe primary and therefore increases the pressurization rate ofthe transient.

Evaluation of the DNBR response to a Turbine Trip Eventshows qualitatively that DNBR will vary only slightly during thistransient. DNBR is negatively effected by decreases in flow andpressure and increases in heat flux and temperature. Asdiscussed in Section 4.2.1.1 transients such as Loss of Load andTurbine trip typically show no decrease in DNBR. Clearly, atall times in the transient, the results from the Loss of Flowevent as described in Section 15.2.5 of the FSAR remain limitingwith regard to DNBR results.

-61-

5'c

0

~)

4.2.1.3 Loss of Condenser Vacuum~ ~ ~

4.2.1.3.1 Limitin Event Develo ment

A Loss of Condenser Vacuum event may occur due to thefailure of the Circulation Water System, the failure of the MainCondenser Evacuation System to remove non-condensible gases, orthe in-leakage of an excessive amount of air through a turbinegland. Low condenser pressure generates a turbine trip signal.The turbine trip signal causes the turbine stop and control valveto close. Upon occurrence of a turbine trip above 154 of fullpower, a signal is supplied to the reactor protective system totrip the reactor. The Loss of Condenser Vacuum event willdisable the SDBS and will result in a gr'adual rampdown of mainfeedwater.

For purposes of developing an event that bounds thepotential. for primary pressurization, several conservativeassumptions will be made relative to the Loss of Condenser Vacuumevent. The following assumptions will be utilized in thisanalysis:

1) For this bounding event, coincident with the Loss ofCondenser Vacuum, the turbine stop valves are assumed toinstantly close on high condenser back pressure. Thisassumption assures the fastest, possible pressurization rate.

2) A reactor trip on turbine trip signal will not becredited. Not allowing the reactor trip on turbine tripsignal to operate will allow the high pressurizer pressuretrip to activate to initiate reactor scram. Use of the highpressurizer pressure trip rather than any other system trip(such as steam generator low level) will insure that thehighest possible primary pressure will be reached during thetransient.

3) Main feedwater flow rate will be instantaneously set tozero to minimize the secondary side heat removal capacity.

4) Action of the Power Operated Relief Valves (PORV) andpressurizer spray will not be credited since it wouldmitigate the primary pressurization.

-62-

The choice for other key initial conditions are listed in Table~ ~

~

~~

~ ~

~

~

~ ~ ~

~

~

~ ~

~4.2.1.3-1. For evaluation of the peak primary pressure, initialconditions are chosen such that the time to reactor trip on highpressure is maximized and that a high rate of change of .pressureis achieved. The basis of the key input parameters is discussedbriefly in the following:

5) The maximum possible power (with uncertainties) willincrease the heatup rate during the transient and result ina higher peak pressure.

6) A minimum initial primary pressure value is utilized toallow for a larger heatup and power increase prior toreaching the reactor trip setpoint.

7) In order to maximize the length of time prior to openingthe secondary side safety valves, a minimum initial reactortemperature and a maximum steam generator tube plugginglevel are chosen. Opening of the secondary side safetyvalves will decrease secondary side pressure causing anincrease in primary to secondary heat transfer which thenproduces a reduction in the primary system temperature.Delay in opening the secondary safety 'alves will,therefore, act to maximize the primary pressurization. Inaddition, the use of the maximum steam generator tubeplugging value acts to reduce initial primary to secondaryheat transfer and will cause a slightly faster primarypressurization.

8) The minimum flow allowed by Technical Specifications ischosen to reduce primary to secondary heat transfer duringthe transient.

9) A positive Moderator Temperature Coefficient (MTC)consistent with the limits in Technical Specifications ischosen in order to produce a power increase in conjunctionwith the primary coolant temperature increase. The leastnegative Doppler Coefficient is also chosen to allow themaximum possible power increase.

10) A bottom peaked axial shape corresponding to the limitsof the full power operating " band is used as the initialcondition for the scram reactivity data. The use of abottom peaked axial shape increases the power increaseduring the transient slightly since the negative reactivitydue to the insertion of the control rods will be delayedcompared to the reactivity insertion associated with a toppeaked or cosine axial shape.

-63-

TABLE 4.2.1.3-1

KEY PARAMETERS ASSUMED FOR THE LOSS OF CONDENSER VACUUM EVENT

ST LUCIE UNIT 1

Parameter unite Value

Total RCS Power(Core Thermal Power x 1.02)

MWth 2754

Initial Reactor Coolant SystemPressure

psia 2203

Initial Core Coolant InletTemperature

oF 547

Tube Plugging 15

Initial RCS Vessel Flow Rate 370,000

Moderator Temperature Coefficient pcm/ F 2.0

Doppler Coefficient pcm/'F -Oe8

CEA Worth at Trip -5.3

-64-

4.2.1.3.2 RESULTS~ ~

~

~

~

~

~ ~The Loss of Condenser Vacuum event initiated from the

conditions given in Table 4.2.1.3-1, results in a highpressurizer pressure trip condition at 6.50 seconds. At 9.60seconds, the primary pressure reaches its maximum value of 2603psia. The maximum pressure point is calculated by RETRAN at thepump discharge. The maximum pressure value is shown to be wellbelow the 2750 psia acceptance criteria.

The peak secondary side pressure occurs at 9.70 seconds andreaches a value of 1054 psia. This value is well below the 1100psia acceptance criteria for the secondary side system.

Table 4.2.1.3-2 presents the sequence of events for thistransient. Figures 4.2.1.3-1 to 4.2.1.3-6 show the power, heatflux, RCS pressure, RCS coolant temperatures, steam generatorpressure, and pressurizer level response to this bounding Loss ofCondenser Vacuum event.

Evaluation of the DNBR response to a Loss of CondenserVacuum Event shows qualitatively that DNBR will vary onlyslightly during this transient. Previous discussion for the Lossof Load transient, Section 4.2.1.1 of DNBR response is alsoapplicable to this event. The Loss of Flow event as described inSection 15.2.5 of the FSAR will remain the bounding event withregard to DNBR response.

-65-

TABLE 4.2.3..3-2

SEQUENCE OF EVENTS

LOSS OF CONDENSER VACUUM

ST. LUCIE UNIT 1


Loss of Condenser Vacuum

High Pressurizer Pressure TripTrip Breakers Open

Pressurizer Safeties Open

CEAs Begin to Drop Into Core

Steam Generator Safeties Open

Peak RCS Pressure

Maximum Steam Generator Pressure

Pressurizer Safeties Close

0.0

6.5

7.65

7.8

8. 15

8.2

9.6

9.7

12.7

2422 psia

2525 psia

1010 psia

2603 psia

1054 psia

2424 psia

-66-

FIGURE 4.2.1.3-1

St. Lucie Unit 1 Loss of Condenser Vacuum

Power vs. Time

10 1520'INE

(SECONDS)

-67-

HGURE 4.2.1.3-2


Heat Flux vs. Time

10 15 20TIME (SECONDS)

30

-68-

'

!

!

~

Oi1

j

!

!

~

FIGURE 4.2.1.3-3


CODCaC4

RCS Pressure vs. Time

oCV

Ch0-C

O

~ CV

Q

V)V)~ O

O0 cv

CV

Cf)

QCI

X~ CV

C)CI40

10 'S 20

TINE (SECONDS)25

-69-

0

d"f

FlGURE 4.2.1.3-4


RCS Coolant Temperatures vs. Time

HOT

AV

COL

1D 15 20

TINE (SECONDS)25

-70-

HGURE 4.2.1.3-5


ooCV

Steam Generator Pressure vs. Time

-o I

CL—

Cf)UJKo0-o

CYDo~ O

LU

LUQoo~ CQ

Cflooh

oo~ l0 15 .20

TINE (SECONDS)25

-71-

FIGURE 4.2.1.3-6


Paasurizer Level vs. Time

10 15 20

TINE. (SECONDS)

-72-

h,

ly

fp,'$c

4.2.1.4 Main Steam Isolation Valve Closure~ ~ ~~

The Main Steam Isolation Valve (MSIV) Closure event isevaluated by assuming that one or both of the MSIV's clo'se andstop steam flow from either one or both SG's. The MSIV closuretime is much greater than that of the turbine stop valves andtherefore the resultant heatup and pressurization of thesecondary and primary systems would be less than that produced bythe Loss of Condenser Vacuum event. The closure of only a singleMSIV would produce a asymmetric pressure difference between steamlines. This asymmetric event would be quickly terminated by theasymmetric steam generator pressure trip.

DNBR response to this event is less limiting than thatexperienced in the Loss of Flow event described in Section 15.2.5of the FSAR.

4.2.1.5 Steam Pressure Re ulator Failure

A Steam Pressure Regulator Failure for St. Lucie Unit 1

would be similar to a failure of the turbine control system. Theturbine control system is a digital electronic hydraulic (DEH)system which controls the turbine automatically using a processcontrol computer, servo-mechanism and hydraulic valve actuators.The computer represents the digital portion of the system, theservo hardware represents the electronic portion of the systemand the valve actuators represent the hydraulic part of thesystem. During automatic operation the DEH control system sendsoutput signals to the servo system which in turn positions thehydraulic valve actuators and controls turbine speed or load.

A failure in this system which would close the turbinevalves would be bounded by the Loss of Condenser Vacuum event dueto the much more rapid action of the turbine stop valves. Thisrapid action of the stop valves increases the primary heatup andmaximizes primary pressurization.

DNBR response to this event is less limiting than thatexperienced in the Loss of Flow event described in Section 15.2.5of the FSAR.

-73-

1

'i

A'C

4.2.1.6 Loss of Non-emer enc AC Power to the Station Auxiliaries

The Loss of Non-emergency AC Power to the Station Auxiliaries(LOAC) is defined as a complete loss of offsite electrical powerand concurrent. turbine trip. As a result of this event, electricalpower would be unavailable for the reactor coolant pumps and mainfeedwater pumps. The plant would therefore experience asimultaneous loss of load, a loss of feedwater and a loss of forcedreactor coolant flow.

The LOAC is followed by automatic startup of the emergencydiesel generators. The power output of each is sufficient tosupply electrical power to all engineered safety features and toensure the capability of establishing and maintaining the plant ina safe shutdown condition. Since power is not available for thecontrol element assembly drive mechanisms (CEDM'), the de-energization of the CEDM magnetic holding coils would release theCEA's and initiate reactor shutdown.

iThe early part of this transient (0-10 seconds) would be

'oundedby the Loss of Forced Reactor Coolant Flow (LOF) eventsince any change in primary to secondary heat transfer resultingfrom the loss of load would not be experienced by the core duringthis time period of the transient due to loop transit time effects.Even if reactor scram on LOAC is not credited, reactor trip on lowreactor coolant flow would quickly occur and the transient DNBRresponse would be the same as the LOF.

For the remainder of the transient, the ability to maintaina secondary side heat sink is assured by the action of theauxiliary feedwater system. Evaluation of the auxiliary feedwatersystem ability to maintain the secondary side heat sink duringtransient conditions is presented in Chapter 10 of the FSAR andbounds the results of the LOAC.

Radiological consequences of this event are bounded by theresults 'of the Inadvertent Opening of a Steam Generator SafetyValve presented in Section 15.2.11.3.2 of the FSAR. The Openingof the Safety Valve is bounding since .that postulated event willresult in the complete blowdown of one steam generator and partialblowdown of the other. Since the mass of steam released is muchlarger, the radiological releases are also much larger than thosethat result from the LOAC. It should be noted that the calculationof radiological doses from a new analysis, if needed, would beperformed by using a simple ratio of the steam masses released (NewAnalysis/FSAR Analysis) to multiply the FSAR calculated doses,since the basic assumptions related to initial concentrations ofradioisotopes, steam generator partition factor, atmosphericdispersion coefficient,'breathing rate and dose conversion factorwould all remain constant.

-74-

~ )

'yS

4.2.1.7 Loss of Normal Feedwater Flow~ ~

1

The Loss of Normal Feedwater Flow event is defined as a .

reduction in feedwater flow to the steam generators whenoperating at power, without a corresponding reduction in steamflow from the steam generators. The result of this mismatch is areduction in the water inventory in the steam generators, whichwill cause a reduction in primary to secondary heat transfer andsubsequent pressurization of the primary system.

A complete loss of feedwater flow to the steam generatorscan occur when;

1) a malfunction in the feedwater regulating system forboth steam generators causes all feedwater regulating valvesto close, or

2) a loss of all feedwater or condensate pumps occursgor

3) in manual feedwater control, the operator either closesthe feedwater regulating valves or closes the feedwater stopvalves, or

4) a main feedwater header ruptures.

Action of the low steam generator level trip will initiatereactor scram and end the primary pressurization. Auxiliaryfeedwater will actuate to provide sufficient feedwater flow toremove decay heat from the RCS following the reactor trip. Theauxiliary feedwater system consists of one non-condensing steamturbine driven auxiliary feed water pump and two motor drivenauxiliary feedwater pumps. The steam generators are designed towithstand the thermal shock and loading imposed by a total lossof feedwater and subsequent refill transient. Evaluation of theauxiliary feedwater system performance to provide decay heatremoval capability is discussed in detail in Chapter 10 of theFSAR.

Primary pressure will increase during this transient as thewater level in the steam generator is reduced from nominal valuesto the low level setpoint. However, the pressurization rate willbe significantly lower than that experienced during the Loss ofCondenser Vacuum transient described in Section 4.2.1.3 of thisdocument. The reason the pressurization rate is much lower forthe Loss of Normal Feedwater Flow event is that the continuationof the steam flow until reactor trip will maintain the primarytemperatures at near constant conditions. Therefore, thebounding event for this category remains the Loss of CondenserVacuum event.

DNBR response during this transient would remain bounded bythe Loss of Flow event described in Section 15.2.5 of the FSAR.

4.2.1.8 Feedwater Line Break~ ~ ~

This event is a cooldown event in the licensing basis forSt. Lucie Unit. l. Zt should be noted that while this event canbe made to be a heatup type event by artificially placing thefeedwater entry into the steam generator at the bottom of thesteam generator (as done for St. Lucie Unit 2), this does notresult in a realistic representation of the physical effects ofthis type of transient. Analysis in this artificial way resultsin potential misunderstanding of the transient response byoperation personnel and may also result in plant changes beingexamined incorrectly when they are evaluated relative to theresults in the FSAR.

When the actual physical configuration of the steam.generators and feedwater system is examined, it can be seenclearly that steam, not liquid would be released from the steamgenerators during the major portion of a'eedwater line breakevent. As such, the feedwater pipe break event is bounded by thesteamline break event since the area for flow in a brokenfeedwater pipe is less than that of a severed steamline. Thesmaller area for flow results in lower steam relief rate whichproduces a more benign event.

I

-76-

~ I

II

1"

~ i

, 'fig ~,'I

P

lt

4.2.2.1 oss of External Load

A loss of external load event is caused by abnormal eventsin the electrical distribution network. The turbine is protectedfrom a complete loss of external load by two overspeed protectionsystems which act to trip the turbine if a complete loss of loadoccurs. The Overspeed Protection Controller and the MechanicalOverspeed Protection system are both designed to trip the turbineat approximately ill percent overspeed condition. More detailsof these two equipment protection systems are found in Section10.2.2.2 of the St. Lucie Unit 2 FSAR. In addition to theoverspeed trip mechanism, the turbine is provided protection alsofor the following:

1) Low Condenser Vacuum. This tripping device is designedto trip the turbine in case of a serious rise inexhaust pressure. The turbine will trip when thevacuum decreases to 19-22 inches of mercury.

!2) Low Turbine Bearing Oil Pressure. The bearing oil

pressure setting at normal speed is approximately 14-18psig. However, should this pressure reach 6 psig thelow bearing oil pressure protective device will tripthe turbine.

3) Thrust Bearing Trip. The thrust bearing trip device isdesigned to shut down the turbine when the thrustbearing trip control pressure rises to above 75-80pslgo

Any turbine trip causes the hydraulic trip fluid headerpressure to decrease and close steam to the turbine. Uponoccurrence of the turbine trip above 154 of full power, a signalis supplied to the reactor protective system to trip the reactor.Subsequent to the turbine trip, the main feedwater regulatingvalves close and feedwater is supplied to the steam generatorthrough the feedwater bypass valve by the main feedwater pumps.

A fast pressurization transient will result from the closureof the turbine stop valves due to their fast response time,approximately 0.25 seconds. The primary side pressurizationresults from the reduction in primary to secondary heat transferafter event initiation. The increased secondary temperature andpressure which occurs with the stoppage of steam flow from thesecondary side is the driving force in the heat transfer,degradation. However, the reactor trip on turbine trip wouldquickly reverse the pressurization event and bring the reactor toa safe shutdown condition before a reactor trip on high pressurecould "occur. 'n addition, action of the SDBS will mitigate thepressurization effects of a Loss of External Load event.

-77-

~ I

I'

t 4

Qt'IC Ol~1r

'

A bounding Loss of Load event is postulated i'f cxedit is nottaken for the reactor trip on turbine trip or action of the SDBS,this transient is bounded by the Loss of Condenser Vacuum eventdiscussed in Section 4.2.2.3 because of the assumption in theLoss of Condenser Vacuum of loss of main feedwater. The loss ofmain feedwater increases the heatup of the primary and thereforeincxeases the pressurization rate of the transient.

Evaluation of the DNBR response to a Loss of External LoadEvent shows qualitatively that DNBR will vary only slightlyduring this transient as discussed in Section 4.2.1.1 above. TheLoss of Flow event as described in Section 15.3.2-2.6 of the FSARwill remain limiting relative to DNBR results.

4.2.2.2 Turb'ne TriA turbine trip is an event in which the steam flow from the

steam generators is stopped through closure of the turbinecontrol and/or stop valves. The turbine is equipped with anautomatic stop and emergency trip system as discussed in Section4.2.2.1 above. Turbine protection devices which could cause aturbine trip are described in depth in section 10.2.2 of theFSAR. Upon occurrence of a turbine trip above 15% of full power,a signal is supplied to the reactor protective system to trip thereactor. Subsequent to the turbine trip, the main feedwaterregulating valves close and feedwater is supplied to the steamgenerator through the feedwater bypass valve by the mainfeedwater pumps.

A fast pressurization transient will result from the closureof the turbine stop valves due to their fast response time,approximately 0.25 seconds. The primary side pressurizationresults from the reduced primary to secondary heat transfer whichoccurs with the stoppage of steam flow from the secondary side.However, the reactor trip on turbine trip would quickly reversethe pressurization event and bring the reactor to a safe shutdowncondition before a reactor trip on high pressure could occur. Inaddition, action of the SDBS will mitigate the pressurizationeffects of a Turbine Trip.

A bounding Turbine Trip event is postulated if credit is nottaken for the reactor trip on turbine trip or action of the SDBS,this transient is bounded by the Loss of Condenser Vacuum eventdiscussed in Section 4.2.2.3 because of the assumption, in thatevent, of an instantaneous loss of main feedwater. The loss ofmain feedwater increases the heatup of the primary and thereforeincreases the pressurization xate of the transient.

Evaluation of the DNBR response to a Turbine Trip Eventshows qualitatively that DNBR will vary only slightly during thistransient as discussed in Section 4.2.1.1 above. The Loss ofFlow event as 'described in Section 15.3.2.2.6 of the FSAR willremain the limiting event with regard to DNBR xesults.

k

-78-

IW.

Z

% ~1



A Loss of Condenser Vacuum event may occur due to thefailure of the Circulation Water System, the failure of the MainCondenser Evacuation System to remove non-condensible gases, orthe in-leakage of an excessive amount of air through a turbinegland. Low condenser pressure generates a turbine trip signal.The turbine trip signal causes the turbine stop and control valveto close. Upon occurrence of a turbine trip above 154 of fullpower, a signal is supplied to the reactor protective system totrip the reactor. The Loss of Condenser Vacuum event willdisable the SDBS and will result in a gradual rampdown of mainfeedwater.

'or

purposes of developing an event that bounds thepotential for primary pressurization, several conservativeassumptions will be made relative to the Loss of Condenser Vacuumevent. The following assumptions will be utilized in thisanalysis:

1) For this bounding event, coincident with the Loss ofCondenser Vacuum, the turbine stop valves are assumed toinstantly close on high condenser back pressure. Thisassumption assures the fastest possible pressurization rate.

2) A reactor trip on turbine trip signal will not becredited. Not allowing the reactor trip on turbine tripsignal to operate will allow the high pressurizer pressuretrip to activate to initiate reactor scram. Use of the highpressurizer pressure trip rather than any other system trip(such as steam generator low level) will insure that thehighest, possible primary pressure will be reached during thetransient.

3) Main feedwater flow rate will be instantaneously set tozero to maximize the decrease in primary to secondary heattransfer.

4) Action of the Power Operated Relief Valves (PORV)and thepressurizer sprays will not be credited since this wouldmitigate the pressurization of the primary.

5) A Loss of AC will be initiated such that the trips onHigh Pressurizer Pressure and Low RCS Flow will occurapproximately coincident. The loss of forced coolant flowwill decrease the primary to secondary heat transfer andproduce a higher peak primary pressure. Coincident tripswill produce the maximum pressurization rate since theduration of the loss of flow is maximized while maintainingthe requirement for tripping on high pressure.

-79-

The choice for other key initial conditions are listed in Table~ ~

~~

~~4.2.2.3-1. For evaluation of the peak primary pressure, initial

conditions are chosen such that the time to reactor trip on highpressure is maximized and that a high rate of change of pressureis achieved. The basis of the key initial conditions isdiscussed briefly in the following:

6) The maximum possible power (with uncertainties) willincrease the heatup rate during the transient and result ina higher peak pressure.

7) A minimum initial primary pressure value is utilized toallow for a larger heatup and power increase prior toreaching the reactor trip setpoint.8) In order to maximize the length of time prior to openingthe secondary side safety valves, a minimum initial reactortemperature value and a maximum steam generator tubeplugging level are chosen. A minimum primary temperatureinitiates the transient with the minimum secondary sidepressure which maximizes the pressure increase requiredbefore the secondary safeties open. In addition, the use ofthe maximum steam generator tube plugging value acts toreduce initial primary to secondary heat transfer and willcause a slightly faster primary pressurization.

9) The minimum flow value consistent with TechnicalSpecifications is chosen to minimize primary to secondaryheat transfer during the event.

10) A positive Moderator Temperature Coefficient (MTC)consistent with the limit@ in Technical Specifications ischosen in order to produce a power increase in conjunctionwith the primary coolant temperature increase. The leastnegative Doppler Coefficient is also chosen to allow themaximum possible power increase.

11) A bottom peaked axial shape corresponding to the limitsof the full power operating band is used as the initialcondition for the scram reactivity data. The use of abottom peaked axial shape maximizes the power increase,during the transient, slightly because the effects of thenegative control reactivity are delayed with respect tothose from top peaked or cosine axial shapes.

-80-

«ky

~ ls,

TABLE 4 2 2 3-1


ST LUCIE UNIT 2

Parameter Units Value


MWth 2754


psia 2167


oF 532

Steam Generator Tube Plugging 15

Initial RCS Vessel Flow Rate gpm 363,000

Moderator Temperature Coefficient pcm/'F 3.0

Doppler Coefficient pcm/'F -0.8

CEA Worth at Trip 4hp -5.5

-81-

1 a$

'I'I

f 'j,

4.2.2.3.2 Results~ ~ ~

The Loss of Condenser Vacuum event initiated from theconditions given in Table 4.2.2.3-1, results in a highpressurizer pressure trip condition at 7.14 seconds. A loss ofAC is chosen such that the reactor trip on low flow occurscoincident with the reactor trip on high pressure. At 11.0seconds, the maximum pressure point is calculated by RETRAN, as2675 psia, located in the lower plenum. The maximum pressurevalue is shown to be well below the 2750 psia acceptancecriteria.

The peak secondary side pressure occurs at. 16.0 seconds andreaches a value of 1041 psia. This value is well below the 1100psia acceptance criteria for the secondary side system.

Table 4.2.2.3-2 presents the sequence of events for thistransient. Figures 4.2.2.3-1 to 4.2.2.3-6 present the power,heat flux, RCS pressure, RCS coolant temperatures, steamgenerator pressure, and pressurizer level response to thisbounding Loss of Condenser Vacuum event.

A Loss of Condenser Vacuum event was also evaluated usingbest-estimate parameters. In this case, all initial conditionswere nominal in an attempt to represent the actual plant responseif a real Loss of Condenser Vacuum event were to occur. In order~

~ ~

~

~

~

~

~

~

~

~

~

~

to obtain some pressurization in the primary during the event,even this case did not take credit for action of the reactorprotective system reactor trip on turbine trip. For this case,the peak pressure reached a maximum of 2441 psia. The differencebetween the peak pressure calculated using our licensing methodsand the best-estimate results provides a measure of theconservatism applied in the design inputs.

Evaluation of the DNBR response to a Loss of CondenserVacuum Event shows qualitatively that DNBR will vary onlyslightly during this transient. Previous discussion for the Lossof Load. transient, Section 4.2.1.1 of DNBR response is alsoapplicable to this event. The Loss of Flow event as described inSection 15.3.2.2.6 of the FSAR will remain the bounding eventwith regard to DNBR response.

-82-

TABLE 4.2 2 3-2

SEQUENCE OF EVENTS


ST. LUCIE UNIT 2



Loss of Offsite Power

High Pressurizer Pressure TripLow Flow TripTrip Breakers Open


CEAs Begin to Drop Into Core

Peak RCS Pressure

Steam Generator Safeties Open

Pressurizer Safeties Close

0.0

6.7

7.14

7.57

8.22

8.26

8.96

11.0

11.05

14.5

Maximum Steam Generator Pressure 16.0

2428 psia

337,590 gpm

2525 psia

2675 psia

1010 psia

2424 psia

1041 psia

-83-

FIGURE 4.2.2.3-1


,Power vs. Time

OO

Ql~

C)CL o

CO

C)CD

C5QJ O

ClY.Oo~e

oC4

10 « 1S 20 . 25

TIME (SECONDS)

-84-

HGURE 4.2.2.3-2


Heat Hux vs. Time

10 1S 20

T I NE (SECONDS)'25

-85-

FlGURE 4.2.2.3-3


RCS Pressure vs. Time

10 . 15 20

T ICE. (SECONDS)

-86-

HGURE 4.2.2.3Q


RCS Coolant Tempcrattrres vs. Time

HOT

AV

COLD

10 15 20

TINE (SECONDS)25

-87-

HGURE 4.22.3-5


ooCV

Steam Generator Pressure vs. Time

~ ocnQ

MI~Q o~o

fX0 o~ o

I~~l

CD oo~ CO

I~

U)oo

oo~ l0 15 20

TIME (SECONDS)25

-88-

HGURE 4.2.2.3-6


Pressurizer Level vs. Time

IO 15 „20TiNE (SECONDS)

25

-89-

IA

S

II~

i

j

4.2.2.4 ain Steam Isolation Valve Closure~ ~ ~~

The MSIV Closure event is evaluated by assuming that one orboth of the MSIV's close and stop steam flow from one or bothSG's. The MSIV closure time is much greater than that of theturbine stop valves and therefore the resultant heatup andpressurization of the secondary and primary systems would be lessthan that produced by the Loss of Condenser Vacuum event. Theclosure of only a single MSIV would produce a asymmetric pressuredifference between steam lines. This asymmetric event would bequickly terminated by the asymmetric steam generator pressuretrip.

DNBR response during this event is bounded by the Loss ofFlow event described in Section 15.3.2.2.6 of the FSAR.

4.2.2.5 Steam Pressuxe Re ulator Failure

A Steam Pressure Regulator Failure for St. Lucie Unit 2would be similar to a failure of the turbine control system. Theturbine control system is a DEH system which controls the turbineautomatically using a process control computer, servo-mechanismand hydraulic valve actuators. The computer represents thedigital portion of the system, the servo hardware represents theelectronic portion of the system and the valve actuatorsre resent the h draulic art of the s stem. The turbine control

i

p y psystem is designed to:

a) control automatically the turbine-generator outputpower during all phases of normal opexation,

b) trip the turbine to guard the equipment from exposureto hazardous conditions,

c) provide an automatic reactor trip signal when theturbine is tripped,

d) provide a turbine runback signal upon privation of onemain feedwater pump or upon loss of service of twoheater drain pumps.

A failure in this system which would close the turbinevalves would be bounded by the Loss of Condenser Vacuum event dueto the much more xapid action of the turbine stop valves. Thisrapid action of the stop valves increases the primary heatup andmaximizes primary pressurization.

DNBR response to this event is bounded by the Loss of Flowevent described in Section 15.3.2.2.6 of the FSAR.

-90-

yiGr

0

4,2.2.6 Loss of Non-emer enc Ac power to the Station Auxiliaries~ ~ ~

The Loss of Non-emergency AC Power to the Station Auxiliaries(LOAC) is defined as a complete loss of offsite electrical powerand concurrent turbine trip. As a result of this event, electricalpower would be unavailable for the reactor coolant pumps and mainfeedwater pumps. The plant would therefore experience asimultaneous loss of load, a loss of feedwater and a loss of forcedreactor coolant flow.

The LOAC is followed by automatic startup of the emergencydiesel generators. The power output of each is sufficient tosupply electrical power to all engineered safety features and toensure the capability of establishing and maintaining the plant ina safe shutdown condition. Since power is not available for thecontrol element assembly drive mechanisms (CEDM's), the de-energization of the CEDM magnetic holding coils would release theCEA's and initiate reactor shutdown.

i'

The early part of this transient (0-10 seconds) would bebounded by the Loss of Forced Reactor Coolant Flow (LOF) eventsince any change in primary to secondary heat transfer resultingfrom the loss of load would not be experienced by the core duringthis time period of the transient due to loop transit time effects.Even if reactor scram on LOAC is not credited, reactor trip on lowreactor coolant flow would quickly occur and the transient DNBRresponse would be the same as the LOF.

For the remainder of the transient, the ability to maintaina secondary side heat sink is assured by the action of theauxiliary feedwater system. Evaluation of the auxiliary feedwatersystem ability to maintain the secondary side heat sink duringtransient conditions is presented in Chapter 10 of the FSAR andbounds the results of the LOAC.

Radiological consequences of this event are bounded by theresults of the Inadvertent Opening of a Steam Generator SafetyValve presented in Section 15.1.3.1.1 of the FSAR. The Opening ofthe Safety Valve is bounding since .that postulated event willresult in the complete blowdown of one steam generator and partialblowdown of the other. Since the mass of steam released is muchlarger, the radiological releases are also much larger than thosethat result from the LOAC. It should be noted that the calculationof radiological doses from a new analysis, if needed, would beperformed by using a simple ratio of the steam masses released (NewAnalysis/FSAR Analysis) to multiply the FSAR calculated doses,since the basic assumptions related to initial concentrations ofradioisotopes, steam generator partition factor, atmosphericdispersion coefficient, breathing rate and dose conversion factorwould all remain constant.

-91-

%t

9, 0

4 2 2.7 Loss of Normal Feedwater Flow

The Loss of Normal Feedwater Flow event is defined as areduction in feedwater flow to the steam'enerators whenoperating at power, without a corresponding reduction in steamflow from the steam generators. The result of this mismatch is areduction in the water inventory in the steam generators, whichwill cause a reduction in primary to secondary heat transfer andsubsequent pressurization of th'e primary system.

A complete loss of feedwater flow to the steam generatorscan occur when:

1) a malfunction in the feedwater regulating system forboth steam generators causes all feedwater regulating valvesto close, or

2) a loss of all feedwater or condensate pumps occurs, or

3) in manual feedwater control, the operator either closesthe feedwater regulating valves or c1oses the feedwater stopvalves, or

4) a main feedwater header rupture.

~~ ~~

~~

~

Action of the low steam generator level trip will initiatereactor scram and end the primary pressurization. Auxiliaryfeedwater will actuate to provide sufficient feedwater flow toremove decay heat from the RCS following the reactor trip. Theauxiliary feedwater system consists of one non-condensing steamturbine driven auxiliary feedwater pump and two motor drivenauxiliary feedwater pumps. The steam generators are designed towithstand the thermal shock and loading imposed by a total lossof feedwater and subsequent refill transient. Evaluation of theauxiliary feedwater system performance to provide decay heatremoval capability is discussed in detail in Chapter 10 of theFSAR.

Primary pressure will increase during this transient as thewater level in the steam generator is reduced from nominal valuesto the low level setpoint. However, the pressurization rate willbe significantly lower than that'xperienced during the Loss ofCondenser Vacuum transient described in Section 4.2.2.3. Thereason the pressurization rate is much lower for the Loss ofNormal Feedwater Flow event is due to the continuation of thesteam flow until reactor trip which maintains the primarytemperatures at near constant conditions prior to reactor trip.Therefore, the bounding event for this category remains the Lossof Condenser Vacuum event.

DNBR response during this transient would remain bounded bythe Loss of Flow event described in Section 15.3.2.2.6 of theFSAR.

-92-

4

AJ.

4.2.2.8 Feedwater Li e Brea

4.2.2.8.1 Li itin Event Develo ment

A Feedwater Line Break (FLB) is defined as the failure of amain feedwater system pipe during plant operation. A rupture inthe main feedwater system rapidly reduces the steam generatorsecondary inventory as first a two phase mixture and then steamexits the break. The limiting location for a feedwater linebreak occurs between the steam generator and the feedwater linecheck valves, since blowdown of the affected steam generatorwould continue until the steam generator loses all inventory. Anadequate secondary heat sink is provided by auxiliary feedwateractuation on steam generator low level and by initiation of theauxiliary feedwater to the unaffected SG after the expiration ofthe AFAS time delay.

Another potential FLB event could assume that the breakoccurs between the main feed header and the check valve., Thisbreak would essentially be a loss of main feedwater event withreactor trip occurring on low steam generator level, therefore,all conclusions reached in Section 4.2.2.7 would be valid forthis type of FLB event.

Previous analysis of the FLB events for St. Lucie Unit 2have been performed assuming that the ruptured feedline wasphysically located at the bottom of the Steam Generators. Thisassumption results in a FLB event that produces a primary systemheatup with a large primary pressurization. Even for thisphysically non-realistic event, previous analyses by the reactorvendor demonstrated that the Loss of Condenser Vacuum event wasstill bounding relative to peak primary pressure.

The analysis presented here evaluates the FLB event as itwould occur at St. Lucie Unit 2, i.e., as a cooldown event whenthe break occurs at the limiting location between the SG and thefeedwater line check valves. The purpose of this analysis is todemonstrate that the FLB meets all design criteria usingconservative analysis inputs. It. will also demonstrate that thelimiting cooldown event for St. Lucie Unit 2 remains the SteamLine Break (SLB) presented in Section 15.1.4.3.5.3 of the FSAR.The SLB event with a break size of 6.358 ft produces a muchquicker cooldown since the break size is so much larger than thelargest possible FLB (1.44 ft ) at the limiting location.

The following assumptions will be utilized in this analysis:

1) The largest feedline break size (1.44 ft ) is assumed inorder to maximize the cooldown in the primary system.

2) Maximum time delays and worst case setpoints for MainSteam Isolation'nd Auxiliary Feedwater Isolation are chosento account for the potential environmental factors which

could occur with a FLB within containment. AuxiliaryFeedwater actuation is assumed to act with the minimum timedelay in order to maximize the amount of feedwater to theaffected SG.

3) Main feedwater will continue to the unaffected steamgenerator until a Main Feedwater Isolation signal isgenerated and the. valve has fully closed.

4) Consistent with the SLB methodology to maximize thepotential to exceed the DNBR limits following a return topower, a Loss of AC is assumed at time of reactor trip. TheLoss of AC reduces the reactor coolant flow which mapnegatively impact the minimum DNBR value. Of course, if thereactor never returns to power following a FLB there will beno approach to the DNBR limit as long as natural circulationflow is maintained.

The choice for other key initial conditions are listed in Table4.2.2.8-1. The basis of the key initial conditions is discussedbriefly in the following:

5) Reactor power is assumed to be at 1024 of rated power.Having the initial power high increases the reactivityincrease associated with the cooldown.

6) Initial coolant temperature and pressure assumptions aremaximized, consistent with the values used in the limitingSLB event. High initial temperature and pressure increasesthe reactivity feedback associated with the primary systemcooldown during the feedwater line break event. The initialcoolant average temperature utilized is slightly less thanthat used in the previous evaluation of the SLB event. Useof the maximum coolant flow to maximize primary to secondaryheat transfer prior to reactor coolant pump trip results inthe FLB calculation being initialized at the lower value.

7) Initial reactor coolant flow is assumed to be at themaximum value as described above.

8) The MTC and Doppler reactivity "coefficients are assumedto be representative of End of Cycle (EOC) conditions andare the most negative possible. A curve of moderatorreactivity versus temperature is used assuming all rods inwith the exception of the worst stuck rod.

9) The control rod worth at trip is the minimum rod worthwith the worst stuck rod.

10) The multi-node SG model will be utilized in analyzingthis event. This model provides the more accurateassessment of SG level needed for this evaluation.

-94-

TABLE 4 2 2 8-3.

KEY PARAMETERS ASSUMED FOR THE FEEDMATER LINE BREAK EVENT

ST LUCIE UNIT 2

Parameter


Units

MWth

Value

2754


psia 2408


oF 552

Initial RCS Vessel Flow Rate gpm 400, 000

Moderator Temperature Coefficient pcm/ F 27 ~


CEA Worth at Trip 4'hp -7.52

I„

J

gf

7

Cf

~ ~ ~4.2.2.8.2 Results

The Feedwater Line Break event initiated from the conditionsgiven in Table 4.2.2.8-1 with the assumptions listed aboveresults in a low SG level trip at 3.7 seconds. A Main FeedwaterIsolation Signal is generated from the AFAS initiation at 3.8seconds. The loss of AC occurs at 4.9 seconds. A Main SteamIsolation signal is generated on low SG pressure at 41.7 secondsinto the event. The minimum primary average temperature prior tothe actuation of Safety Injection (SI) occurs at 120 seconds witha value of 472 Degrees F. SI flow occurs at 130 seconds afterthe initiation of the event. After the affected SG goes dry, theunaffected SG is available to remove decay heat and bring theplant to a safe shutdown condition.

Table 4.2.2.8-2 presents the sequence of events for thistransient. Figures 4.2.2.8-1 and 4.2.2.8-2 present thecomparison between the results of the Feedwater Line Breakperformed using RETRAN and the results obtained by CE for thelimiting SLB in the FSAR for pressurizer pressure and average RQStemperatures, respectively. It can be seen that the SLB producesa more limiting cooldown transient as expected. The differencebetween the SLB and the FLB is the more rapid cooldown in the SLBdue to the larger break area associated with the SLB. The FLBres ponse is representative of the results which would occur by

1.44 ft~~

~~

~~

~~

~~

~

~~

~~

examining a non-limiting SLB with the same break area (1. ) ~

Figures 4.2.2.8-3 to 4.2.2.8-7 present the response ofprimary power, pressurizer level, RCS coolant temperatures forthe unaffected loop and the affected loop, and the calculated SGpressures for the two loops for this Feedwater Line Break event.

The results clearly show that the Feedwater Line Break eventis a cooldown which is bounded by the results of the SteamlineBreak event found in Section 15.1.4.3.5.3 of the FSAR.

The DNBR response to the FLB would be bounded by the Loss ofFlow event described in Section 15.3.2.2.6 of the FSAR. Sincethere is no return to power for the FLB, and natural circulationflow would be maintained, this event would not approach the DNBRlimit.

-96-

TABLE 4 ~ 2 ~ 2 ~ 8-2

SEQUENCE OF EVENTS

FEED'RATER LINE BREAK

ST LUCIE UNIT 2

EVENT T1ME seconds SETPOINT OR VALUE

Feedwater Line Break

Low SG Level TripAFAS Signal GeneratedAffected SG

MFIS Generated

Trip Breakers Open

Loss of AC Power

Turbine Trip Signal

MSIS Signal

AFAS Signal GeneratedUnaffected SG

Auxiliary FeedwaterIsolation Signal, Affected SGon High Differential Pressure

SI Signal InitiatedMinimum Average TemperaturePrior to SI

SI Flow Begins

0 ~ 0

3 '

3.8

3.8

4.8

4.9

5.8

42.0

48.0

81.0

100.0

120.0

130.0

1.44 ft20.5 4 NR

19.0 O'R

AFAS signal

460 psia

19 ' 4 NR

281 psia

1578 psia

472 Degree F

-97-

HGURE 4.2.2.8-1

St. Lucie Unit 2 Feedline Break

Presmuizer Pressure vs. Tiae

t

I

I

lI

I

FLB

MSLB

50 100 150 200T I NE (SEC)

250 300

FIGURE 4.2.2.8-2


oRCS Average Teagerature vs. Time

~ oeJ iA

uJlY

~o<o~ ED

LU0

UJ

o

QI~'

ooCf) FLB

oIA MSLB

o%0 SQ 300 )$0 200

T I NE (SEC)2SO 300

-99-

FIG&M4.2.2.8-3


Coxe Power vs. Time

I

I

50 I00'50 20.0

T I NE (SEC).250 300

-100-

aGURF.4.2.2.8-4

St. Lucie Unit 2 Feed1ine Break

Pressurizer Level vs. Tine

50 100 150 200 '

I NE (SEC)~ 250 300

-101-

HGURE 4.2.2.8-5


oo

Unaffected Loop Temperatttzes vs. Time

oill

I~

V)o

Q oI

0~l

<o~ l/7

HOT

AV

0-oQgQ D

COLD

oEA

o%0 50 100 150 200

T I NE (SEC)250 300

-102-

HG&&4.2.2.8

St. Lucie Unit 2 Feed1ine Break

Affected Loop Tempet.tttttes vs. Time

HOT

AV

COLD

50 100 1SO 250.. T I f1E (SEC)

300

-103-

St. Lucie Unit2 Feedline Break

ooCV

hSeeted and Unafhcted Steam Generate Pxessuzes vs. Time

oooQJ ~

DMCOUJ~o0 cu

o UNAFFECTED

C)

+ oLLl ~o

UJ

<o~s o

V)

AFFECTED

OoOl

50 100 150 200T I NE (SEC)

250 300

-104-

C

~P

I

"~k

4.2.3 Turke Point Units 3 and 4~ ~

4.2.3.1 Loss of External Load

A loss of external load event is caused by abnormal eventsin the electrical distribution network. The turbine is protectedfrom a complete loss of external load by a mechanical overspeedtrip mechanism as well as a electro-magnetic pickup on thegoverning channel which acts to trip the turbine if a completeloss of load occurs. More details of these equipment protectionsystems are found in Section 10.2.2 and Section 14.1.13 of theTurkey Point FSAR. Upon occurrence of a turbine trip above 104of full power, a signal is supplied to the reactor protectivesystem to trip the reactor.

A fast pressurization of the primary will result from theclosure of the turbine stop valves due to their fast responsetime of approximately 0.1 seconds. The primary sidepressurization results from the reduced primary to secondary heattransfer which is produced by the increase in secondarytemperature and pressure that occurs with the stoppage of steamflow from the secondary side. However, the expected reactor tripon turbine trip would quickly reverse the pressurization eventand bring the reactor to a safe shutdown condition prior to theprimary pressure reaching the high pressure trip setpoint. Inaddition, action of the turbine bypass system will mitigate thepressurization effects of a Loss of External Load. Even ifcredit is not taken for the reactor trip on turbine trip oraction of the turbine bypass system, this transient is boundedrelative to primary pressurization by'he Loss of CondenserVacuum event discussed in Section 4.2.3.3 because of theassumption in the Loss of Condenser Vacuum that an instantaneousloss of feedwater will occur. The loss of main feedwaterincreases the heatup of the primary and therefore increases thepressurization rate of the transient.

Evaluation of the DNBR response to a Loss of External LoadEvent shows qualitatively that DNBR .will vary only slightlyduring this transient. DNBR is negatively affected by decreasesin flow and pressure and increases in heat flux and temperature.Transients which can challenge the DNBR limit result from eithersignificant decreases in flow, i.e., Loss of Flow events orsignificant increases in heat flux, i.e., Rod Withdrawal atPower. During the Loss of External Load, the increase intemperature will be offset by the increase in pressure and onlyslight variations in DNBR would be expected. For all reasonablecombinations of input parameters for the Loss of External Load,other events reported in Chapter 14 of the FSAR would be limitingwith respect to DNBR response.

-105-

Jql Jl

4.2.3.2 Turbine Tri~ ~ ~~ ~

A turbine trip is an event in which the steam flow from thesteam generators is stopped through closure of the turbine stopvalves and governing control valves. Each stop valve is an oiloperated, spring closing valve controlled primarily by theturbine over speed trip devices. The turbine overspeed trippilot is actuated by one of the following signals to close thestop valves:

Turbine thrust bearing tripLow bearing oil pressure tripLow condenser vacuumSolenoid tripOverspeed tripManual trip

Any generator faultGenerator lockoutHi-hi SG levelSafety Injection signalReactor Trip (above P-7)

Upon occurrence of a turbine trip above 10% of full power, asignal is supplied to the reactor protective system to trip thereactor.

A fast pressurization of the primary will result from theclosure of the turbine stop valves due to their fast responsetime of approximately 0.1 seconds. The primary sidepressurization results from the reduced primary to secondary heattransfer which occurs with the stoppage of steam flow from thesecondary side. However, the expected reactor trip on turbinetrip would quickly reverse the pressurization event and bring thereactor to a safe shutdown condition prior to the primarypressure reaching the high pressure trip setpoint. In addition,action of the turbine bypass system will mitigate thepressurization effects of a Turbine Trip. Even if credit is nottaken for the reactor trip on turbine trip or action of theturbine bypass system, this transient is bounded relative to

- primary pressurization by the Loss of Condenser Vacuum eventdiscussed in Section 4.2.3.3 because of the assumption in theLoss of Condenser Vacuum that a instantaneous loss of feedwaterwill occur. The loss of main feedwater increases the heatup ofthe primary and therefore increases the pressurization rate ofthe transient.

Evaluation of the DNBR response to a Turbine Trip showsqualitatively that DNBR will vary only slightly during thistransient. As discussed in Section 4.2.3.1, the bounding eventsfor DNBR response would remain the "Loss of Flow transients asdescribed in Chapter 14.1.9 of the FSAR.

-106-

4

4 ~

4"

\

IP~



A Loss of Condenser Vacuum may occur due to the failure ofthe Circulation Water System, the failuxe of the Main CondenserEvacuation System to remove non-condensible gases, or the in-leakage of an excessive amount of air thxough a turbine gland orcondenser seal. Low condenser pressure will generate a turbinetrip signal. The turbine trip signal causes the turbine stopvalves to close and stops steam flow out of the steam generators.Upon occurrence of a turbine trip above 10% of full power, asignal is supplied to the reactor protective system to trip thereactox. The Loss of Condenser Vacuum will disable the turbinebypass system and will result in a gradual rampdown of mainfeedwater.

For purposes of developing an event that bounds thepotential for primary pressurization, several conservativeassumptions will be made relative to the Loss of Condenser Vacuumevent. The following assumptions will be utilized in thisanalysis:

1) For this bounding event, coincident, with the Loss ofCondenser Vacuum, the turbine stop valves are assumed toinstantly close on high condenser back pressure. Thisassumption assures the fastest possible pressurization xate.

2) A reactor trip on, turbine trip signal will not becredited. Not allowing the reactor trip on turbine tripsignal to operate will allow the high pressurizer pressuretrip to activate to initiate reactor scram. Use of the highpressurizer pressure trip rather than any other system trip(such as steam generator low level) will insure that thehighest possible primary pressure will be reached during, thetransient.

3) Main feedwater flow rate will be instantaneously set tozero to maximize the decrease in primary to secondary heattransfer.

4) Action of the PORV's and .pressurizer sprays will not becredited since this would mitigate the pressurization of theprimary system.

-107-

s 'f.

.I' %j

Ig

l

I

1

1

~ ~ ~

~ ~ ~~

~

The choices for other key initial conditions are listed inTable 4.2.3.3-1. For evaluation of the peak primary pressure,initial conditions are chosen such that the time to reactor tripon high pressure is maximized and that a high rate of change ofpressure is achieved. The basis of the key initial conditions isprovided briefly in the following discussion:

5) The maximum possible power {with uncertainties) willincrease the heatup rate during the transient and result ina higher peak pressure.

6) A minimum initial primary pressure value is utilized toallow for a larger heatup and power increase prior toreaching the reactor trip setpoint.

7) In order to maximize the length of time prior to openingthe secondary side safety valves, a minimum initial reactortemperature value and a maximum steam generator tubeplugging level are chosen. Opening of the secondary sidesafety valves will increase primary to secondary heattransfer and result in a reduction in the primarytemperature. Delay in opening the secondary safety valveswill, therefore, act to maximize the primary pressurization.Xn addition, the use of the maximum steam generator tubeplugging value acts to reduce initial primary to secondaryheat transfer and will cause a slightly faster primarypressurization and a higher final peak pressure.

.8) A positive MTC, conservative relative to the limits inTechnical Specifications is chosen in order to produce apower increase in conjunction with the primary coolanttemperature increase. The least negative DopplerCoefficient is also chosen to allow the maximum possiblepower increase.

9) A bottom peaked axial shape corresponding'o the limitsof the full power operating, band is used as the initialcondition for the scram reactivity data. The use of abottom peaked axial shape maximizes the power increase,during the transient, slightly because the effects of thenegative control reactivity are delayed with respect tothose from top peaked or cosine axial shapes.

-108-

TABLE 4 2 ~ 3 ~ 3-1


TURKEY POINT UNITS 3 & 4



MWth 2244


psia 2200


oF 542

Steam Generator Tube Plugging

Initial RCS Vessel Flow Rate gpm 268,500

Moderator Temperature Coefficient pcm/ F +2.0


Rod Worth at Trip 4.0

-109-

Si

raQ

4.2.3.3.2 ~esulus~ ~ ~ ~

The Loss of Condenserconditions given in Tablepressurizer pressure tripinitiation of the event. Atreaches its maximum value ofpoint is calculated by RETRANpressure value is shown to becriteria.

Vacuum event initiated from the4.2.3.3-1, results in a high

condition at. 7.1 seconds after8.3 seconds, the primary pressure2601 psia. The maximum pressure

at the pump discharge. The maximumwell below the 2750 psia acceptance

The peak secondary side pressure occurs at 15.7 seconds andreaches a value of 1151 psia. This value is well below the 1210psia acceptance criteria for the secondary side system.

Table 4.2.3.3-2 presents the sequence of events for thistransient. Figures 4.2.3.3-1 to 4.2.3.3-6 show the power, heatflux, RCS pressure, RCS coolant temperatures, steam generatorpressure, and pressurizer level response to this bounding Loss ofCondenser Vacuum event.

A Loss of Condenser Vacuum event was also evaluated usingbest-estimate parameters. In this case, all initial conditionswere nominal in an attempt to represent the actual plant responseif a real Loss of Condenser Vacuum event were to occur. In orderto obtain some pressurization of the primary system, this casedid not take credit for the action of the reactor protectivesystem to initiate a reactor trip on the turbine trip signal.For this case, the peak pressure reached a maximum of 2488 psia.The difference between the peak pressure calculated using ourlicensing methods and the best-estimate results provides ameasure of the conservatism applied in the design inputs.

Evaluation of the DNBR response to a Loss of CondenserVacuum Event show qualitatively that DNBR will vary only slightlyduring this transient. The Loss of Flow events as described inSection 14.149 of the FSAR will remain bounding relative to DNBRresponse.

-110-

TABLE 4 2.3.3-2

SEQUENCE OF EVENTS


T(JHKEY POINT UNITS 3 & 4



High Pressurizer Pressure


Peak RCS Pressure

0 0

7.10

7.98

8.3

2455 psia

2525 psia

2601 psia

Rods Begin to Drop Into Core

Steam Generatox Safeties Open

Pressuxizer Safeties Close

F 1

10.15

12.8

Maximum Steam Generator Pressure 15.7

1111 psia

2400 psia

1151 psia

FIGUI&4.2.3.3-1

Turkey Point Units 3 &4 Loss of Condenser Vacuum

Power vs. Time

C)

ol~

C)CL

LUo 0~ EO

or

10 15 20TINE (SEC)

25

—.112-

HGURE 4.2.3.3-2

Turkey Point Units 3 4 4 Loss of Condenser Vacuum

Heat Flux vs. Time

Ox<4

UJKo0+

10'5 20T I ME (SEC)

FIGURE 4.2.3.3-3

Turkey Point Units 3 Ec 4 Loss of Condenser Vacuum

oCO

RCS pressure vs. Time

CIC)EDAl

V)0 o

O

QJ NQ

CA0)QJ O~ cv

C)

CL cv

I~IMoM cv

C7oCO

10 15 20

T I NE (SEC)

-114- '

FIGURE 4.2.3.3-4

Turkey Point Units 3 Ez 4 Loss of Condenser Vacuum

RCS Coolant Temperatures vs. Time

HOT

AV

COLD

10 15 . 20

T 1ME (SEC)

-115-

FIGURE 4.2.3.3-5

Turkey Point Units 3 Ec 4 Loss of Condenser Vacuum

OCOPV

Steam Gerrerator Pressure vs. Time

OO

"sL~I

V)UJ o0 o

GoM

oO

O10 15 20

T I ME (SEC)25

-116-

FIGURE 4.2.3.3-6

Turkey Point Units 3 K4 Loss of Condenser Vacuum

Ressurixer Level vs. Tame

)0 , 15 20T I NE (SEC)

"25

-117-

el„

1

4.2.3.4 ain Steam Isolation Valve Closure~ ~ ~~

A MSIV. Closure event which resulted in the closure of thethree MSIV's for the Turkey Point Units would be bounded by theLoss of Condenser Vacuum event due to the slower response time ofthe MSIV compared to the turbine stop valves. Closure of lessthan all three MSIV's would produce only a limited primary heatproduction to secondary heat removal mismatch which would also bebounded by the Loss of Condenser Vacuum results.

DNBR response to this event is less limiting than the Lossof Flow events described in Section 14.1.9 of the FSAR.

4.2.3.5 Steam Pressure Re ulator Failure

A Steam Pressure Regulator Failure for Turkey Point Units 3& 4 would be similar to a failure of the turbine control system.A failure in this system which would close the turbine valveswould be bounded by the Loss of Condenser Vacuum event due to themuch more rapid action of the turbine stop valves.. The rapidaction of the stop valves relative to the control valvesincreases the primary heatup and maximizes primarypressurization.

DNBR response to this event is less limiting than the Lossof Flow events described in Section 14.1.9 of the FSAR.

-118-

l'

~

r

k+

1

4.2.3.6 Loss of Non-emer enc AC Power to the Station Auxiliaries~ ~ ~

4.2.'3.6.1 Limitin Event Develo ment

The Loss of Non-emergency AC Power to the Station Auxiliaries(LOAC) is defined as a complete loss of offsite electrical powerand concurrent turbine trip. As a result of this event, electricalpower would be unavailable for the reactor coolant pumps and mainfeedwater pumps. The units would therefore experience asimultaneous loss of load, a loss of feedwater and a loss of forcedreactor coolant flow.

The LOAC is followed by automatic startup of the emergencydiesel generators. The power output of each is sufficient tosupply electrical power to all engineered safety features and toensure the capability of establishing and maintaining the units ina safe shutdown condition. Since power is not'available for thecontrol rod drive mechanisms, an immediate reactor shutdown willoccur.

The three turbine driven auxiliary feedwater pumps would bestarted on the loss of offsite power and would begin providingauxiliary feedwater to the two units. The turbines utilize steamfrom the main steam line to drive the auxiliary feedwater pumps to

~

~deliver feedwater to the steam generators. The turbine driverexhausts the steam to the atmosphere. The pumps take suctiondirectly from the condensate storage tanks for delivery to thesteam generators.

Following the reactor trip, decay heat. removal would beprovided by the action of the auxiliary feedwater pumps and thesteam generator safety valves.

The evaluation of this transient is performed in order toverify that sufficient heat removal capacity is available to ensurethat the pressurizer does not become water solid.

In order to develop a bounding analysis of the response ofthe Turkey Point units to a LOAC, several conservative assumptionswill be made relative to the LOAC event. The following assumptionswill be utilized in this analysis:

1) Reactor trip due to loss of power to the control rod drivemechanisms at the initiation of the LOAC is not credited.

2) Reactor coolant pump coastdown will be delayed until thereactor trip signal is generated. This assumption ensuresthat a quick reactor trip on low flow will not be generated.An early reactor trip- would reduce the severity of the eventsince the Steam Generator water mass inventory after reactortrip would be larger and consequentially the primary systemlong term heatup would be reduced..

-119-

3) A single failure will occur in the auxiliary feedwatersystem. This results in the availability of only oneauxiliary feedwater pump supplying water to the two units.Auxiliary feedwater for this analysis will be assumed to beonly 110 gpm until operator action, 10 minutes after reactortrip, raises the flow to 230 gpm. It is also assumed that afeedwater piping volume of 110 cubic feet contains liquid at.normal feedwater temperatures which must be emptied into theSG's prior to cold auxiliary feedwater reaching the SG's.

4) The pressurizer PORV's are available and assumed tooperate. The action of the PORV's will maximize thepressurizer level when they open to relief primary systempressure. Pressurizer sprays are also assumed available toaid in the maximization of the pressurizer level.5) Core decay heat generation is based on the 1971 versionANS-5.1. This standard is a very conservative representationof the decay heat release rates.

The choice for the key initial conditions are listed in Table4.2.3.6-1. For evaluation of the heat removal capacity, initialconditions are chosen to maximize the primary heat content. Thebasis of the key input parameters is discussed briefly in thefollowing:

6) The maximum possible power (with uncertainties) willincrease the primary heat source during the transient andmaximize the primary heatup after trip.

7) The maximum initial primary pressure and temperature willbe assumed to maximize the initial water density within theprimary system.

8) Minimum RCS flow and maximum steam generator tube pluggingare assumed to minimize the primary to secondary heat transferduring the transient.

9) Conservative reactivity parameters are assumed in orderto increase primary power prior to the reactor trip.

N

10) The multi-node SG model will be used in this analysis.This model provides a more accurate assessment of SG levelthan our standard single node model and is needed for thisevaluation"since the reactor trip occurs due to SG low level.

-120-

TABLE 4. 2 ~ 3 6-1

KEY PARAMETERS ASSUMED FOR THE LOSS OF NON-EMERGENCY AC POWER

TURKEY POINT UNITS 3 8 4



2244


psia 2300


oF 550.2

Steam Generator Tube Plugging

Initial RCS Vessel Flow Rate~ ~

gpm 268,500

Moderator Temperature Coefficient pcm/'F +2.0


Control Rod Worth at Trip 4hp -4.0

-121-

$f,f

4

f

VI

ff4

~i

4.2.3.6.2 Results~

~The LOAC event, initiated from the conditions given in Table

4.2.3.6-1, is presented in Figure 4.2.3.6-1 to Figure 4.2.3.6-5.The sequence of events is shown in Table 4.2.3.6-2.

After the reactor trip and the subsequent coastdown of theRCP's, energy removal from the reactor core is through naturalcirculation to the steam generators and is ultimately removedthrough the secondary safety valves. With an auxiliary feedwaterflow of only 110 gpm to feed all three SG's, the primary systemwill not initially cooldown after the reactor trip due to thehigh level of decay heat. As shown in Figure 4.2.3.6-3, primarypressure is maintained at a high value and reaches the PORVsetpoint approximately 400 seconds into the transient. ThePORV's cycle until the auxiliary feedwater begins to cool theprimary system. Pressurizer liquid level is shown on Figure4.2.3.6-4. As shown, the maximum level occurs at 720 secondsafter the initiation of the eveht. After that peak, liquid leveldecreases as secondary cooling becomes more effective as theoperator acts to increase auxiliary feedwater flow.

The results of this transient show that for this boundingevent, liquid is never released from the primary relief valvesand that the primary system can be cooled through use of oneauxiliary feedwater pump.

-122-

TABLE 4.2.3.6-2

SEQUENCE OF EVENTS

LOSS OF NON-EMERGENCY AC POWER

'JKJRKEY POINT UNITS 3 & 4


Loss of Non-emergency AC

Low SG Level TripRods Begin to Drop

RCP's Coastdown InitiationFlow From one AuxiliaryFeedwater pump started

Operator Realigns Systemto Increase AuxiliaryFeedwater Flow

0.0

52.0

54.0

58.0

232.0

652.0

10% NR

110 gpm

230 gpm

Peak Water Level inPressurizer

720.0 1020 cubic feet

Feedwater Lines Purgedand Cold Auxiliary Feedwateris Delivered

1095.0

-123-

FIGURE 4.2.3.6-1

Turkey Point Units 3 2 4 Loss ofNon-Emergency AC Power

OPower vs. Time

OO

Q OI~

oCL

l~0 O~ 4

O

o 10 10 1'I NE (SEC)

-124-

FIGURE 4.2.3.6-2

Turkey Point Units 3 2 4 Loss of Non-Emergency AC Power

oCV

Has Flux vs. Time

o

ox<I

I~KoO<O

o 100 101 10

T I NE (SEC)104

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FIGURE 4.2.3.6-3

Turkey Point Units 3 R, 4 Loss ofNon-Emergency AC Power

oCV

Pressurizer Pressure vs. Time

oCVCV

QhJ

LUCV

V)Rgl~~ CV

CL

io'~ ~

10TINE (SEC)

~ ~

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FIGURE 4.2.3.6<

Turkey Point Units 3 Ez 4 Loss ofNon-Emergency AC Power

Phmarizer Level vs. Tine

I

4

CD umCD

FV

O

OC7h

C710 10'07

7 I NE (SEC)

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FIGURE 4.2.3.6-5

Turkey Point Units 3 k 4 Loss ofNon-Emergency AC Power

RCS Temperatures vs. Time

l~

~ OLU ~CI

CD

CL

l~0

l~

Ch

Mo

HOT

COLD

l0 10 1''IME (SEC)

128-

t

Y 'g

yJ;

g

0'

4.2.3.7 Loss of Normal Feedwater Flow~ ~ ~

The Loss of Normal Feedwater Flow event is defined as areduction in feedwater flow to the steam generators whenoperating at power without a corresponding reduction in steamflow from the steam generators. The result of this mismatch is areduction in the water inventory in the steam generators, whichwill cause a reduction in primary to secondary heat transfer andsubsequent pressurization of the primary system.

The loss of normal feedwater flow event would be similar tothe LOAC Power event described in the proceeding section,however, the auxiliary feedwater flow would be larger in thisevent since all the auxiliary feedwater would go to the affectedunit rather than being split between the two units. Assuming asingle failure of an auxiliary feedwater train, the minimum flowto the secondary side would be assumed as 315 gpm. Theadditional flow is sufficient to remove decay heat as well as theheat generated in the reactor coolant pumps without producing theamount of pressurizer level increase seen in the LOAC powerevent. Therefore, the results of the LOAC power will bound thisevent.

DNBR response during this transient would remain bounded bythe Loss of Flow events described in Section 14.1.9 of the FSAR.

~ ~ ~4.2.3.8 Feedwater Line Break

This event is a cooldown event in the licensing basis forTurkey Point Units 3 & 4. As such, the feedwater pipe breakevent is bounded by the steamline break event since the area forflow in a broken feedwater pipe is less than that of a severedsteamline. The smaller area for flow results in lower steamrelief rate which produces a more benign event.

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)V

~x

5.0 Conclusions

FPL has presented in the preceding sections information oncomparisons between the RETRAN computed results and plant operatingdata together with sensitivity studies for key parameters for twoplant events. In addition, the licensing assumptions which willbe utilized for events within the category of Decrease in HeatRemoval by the Secondary System have been discussed in depth andthe limiting transients for each of the units has been presented.

FPL believes that the information that the NRC requested inReference 3 has been supplied for this one category of events. Thework presented here has firmly tied the RETRAN model to actualplant operation through the comparison to two plant transients.It should also be noted that in the 1986 Topical Report, threeother plant transients were presented in this category of event.This extensive plant benchmarking has demonstrated the adequacy ofthe FPL RETRAN modelling.

The methodology that willbe used in future licensing analyseshas been described in detail in Section 4.0. It is FPL's beliefthat the necessary conservatisms required for licensingapplications have been utilized in the analysis of these limitingevents.

~

~~

~~

~

~~

~

FPL believes that this combination of a benchmarked, accurateRETRAN model, with conservative licensing assumptions providessufficient basis for a determination that FPL can support licensingactions within the Decrease in Heat Removal from the SecondarySystem event category in a manner consistent with maintaining thehealth and safety of the public.

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6.0 References

1 ~

2 ~

3 ~

4 ~

5.

7 ~

8.

9 ~

NTH-G-06,«Topical Report, RETRAN Code Transient AnalysisModel Qualification," Letter from C.O. Woody to H.N. Berkow{USNRC), January 7, 1986, L-86-4.

D.G. Eisenhut, Licensee Qualification for Performing SafetyAnalyses in Support of Licensing Activities, NRC GenericLetter Number 83-11, February 8, 1983.

Letter from G.C. Lainas (USNRC) to W.F. Conway, "FloridaPower and Light Company — Topical Report on RETRAN (TAC No.60550) and Topical Report on PWR Physics Methodology (TACNo. 60549),« April 19, 1988.

«RETRAN02 — A Program for Transient, Thermal-HydraulicAnalysis of Complex Fluid Flow System," EPRI NP-1850 CCM-A,Rev. 4 (MOD004), November 1988.

Letter from A.C. Thadani (USNRC) to R. Furia, "Acceptancefor Referencing Topical Report EPRI-NP-1850 CCM-A, Revisions2 and 3 Regarding RETRAN02/MOD003 and MOD004, October 19,1988.

"Turkey Point 4 Reactor Trip Due to Loss of 4C Inverter{Vital AC Bus 4P06),« Shift Technical Advisor Report, June20, 1985.

Licensee Event Report 85-017-00, "Engineered Safety FeatureActuation — Reactor Trip, July 22, 1985.

Florida Power and Light Company, Operating Procedure 4-ONOP-003.6, "Loss of 120 V Vital Instrument Panel 4P06,«September 20, 1985.

"Reactor Trip on Low S/G Level due to MFRV Closure," ShaftTechnical Advisor Report, September 20, 1988.

10. "Post Trip Review Check List," M. Snyder,1988.

September 20,

12 ~

13

«U.S. N clear Regulatory Commission, STANDARD REVIEW PLAN,Office of Nuclear Reactor Regulation"g NUREG 0800.

«NUSCo Thermal Hydraulic Model Qualification", Volume 1(RETRAN), NUSCo 140-1, August 1, 1984.

".EPRI PWR Safety and Relief Valve Test Program, Safety andRelief Valve Test Report", EPRI NP-2628-SR, December 1982.

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"Description of the Exxon Nuclear Plant Transient SimulationModel for Pressurized Water Reactors (PTS-'WR)", XN-74-5(P)(A), Rev.2 and Supplements. 1-6, Exxon Nuclear Co.,October 83.

15. "CESEC Topical Report", CENPD-107, July 1974.

16. "LOFTRAN Code Description" WCAP-7907-P-A, April 1984.

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