ML20059D437

Alabama Power Jm Farley Unit 2 Increased Steam Generator Tube Plugging & Reduced Thermal Design Flow Licensing Rept

ML20059D437
Person / Time
Site:	Farley
Issue date:	07/31/1990
From:	Morrison R WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20059D432	List: ML20059D427 ML20059D434 ML20059D437
References
WCAP-12659, NUDOCS 9009070040
Download: ML20059D437 (99)

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i WESTINGHOUSE 'PROPRIETAR,Y CLASS 3 L

WCAP-12659 4

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- ALABAMA POWER 1

JOSEPH M. FARI.EY UNIT 2 INCREASED STEAM GENERATOR TUBE PLUGGING AND REDUCED THERMAL DESIGN FLOW LICENSING REPORT i

PREPARED BY:

o I

. R. J. MORRISON ?

MANAGER,OPEydTiNGPLANTLICENSINGI JULY, 1990 l

WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Division P.O. Box 355 Pittsburgh, PA 15230 RFVISION 0 l

Copyright 1990, Westinghouse Electric Corporation, all rights reserved.

i.

I WESTINGHOUSE PROPRIETARY CLASS 3 l

l l

This proprietary report bears a Wes'inghousa copyright notice. The NRC is permitted to make the number of cop as of the information contained in this report which are necessary for its b _ 4*nal use in connection with generic and plant specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.790 regarding restrictions on public disclosure.

Copies of this report or portions thereof made by the NRC must include the copyright notice and the proprietary notice.

1 l

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We WESTINGHOUSE PRORRIETARY CLASS 3 "0NTRIBUTORS TO WCAP-12659 M. Asztalos J. King F. Baskerville P. Kotwicki D. Bird R. Magee R. Blaushild W.' Middlebrooks l

M. Bonfiglio M. Osborne D. Boyd R. Patel C. Boyd R. Pfeifer A. Cheung K. Rubin A. Fakhri L. Schaub B. Gowda L. Smith J. Gresham G. Springer-R. Haessler J. Stackhouse C. S. Hauser T. Timmons J. Hall J. Visaria D. Holderbaum M. Wengerd G. Hopkins R. Wilson J. Houtni;n S. Zawalick l

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WESTINGHOUSE PROPRIETARY CLASS 3 TABLE OF CONTENTS 1

SECTION TITLE PAGE j

1.0 INTRODUCTION

11 2.0 OPERATING PARAMETERS 21 3.0 NSSS SYSTEMS AND COMPONENTS EVALUATIONS 31

[

3.1 NSSS Systems 31 3.1.1 NSSS Fluid Systems 31 3.1.2 Reactor Control System and Margin To Trip Evaluation 31 3.2 Component Design Transients 35 3.3 NSSS Primary Components 33 3.3.1 Reactor Vessel 33 3.3.2 Reactor Pressure Vessel System 34 3.3.3 Reactor Coolant Pump /RCP Motor 35 3.3.4 Control Rod Drive Mechanisms 35 l

3.3.5 Pressurizer 35 l

3.3.6 Reactor Coolant Loop Piping and Primary Equipment Supports 36 3.3.7 Steam Ger1rators 3-6 3.4 Auxiliary Equipm' nt 39 4.0 ACCIDENT ANALYSES AND EVALUATIONS 41 4.1 Loss of Coolant Events 41 4.1,1 Small Break LOCA Evaluation 41 4.1.2 Large Break LOCA Analysis 43 4.1.3 Miscellaneous LOCA Considerations 4 12 4.1.4 Steam Generator Tube Rupture 4 15 4.1.5 Conclusion 4 16 4.2 Non LOCA Events 4 26 4.2.1 Introduction and Purpose 4 26 4.2.2 Uncontrolled RCCA Bank Withdrawal from Suberitical 4 27 4.2.3 Uncontrolled RCCA Bank Withdrawal at Power 4 28 4.2.4 RCCA Misalignment 4 30 4.2.5 Uncontrolled Boron Dilution 4 30 4.2.6 Partial Loss of Forced Reactor Coolant Flow 4 31 4.2.7 Start up of an inactive Reactor Coolant Flow 4 33

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WESTINGHOUSE PROPRIETARY CLASS 3 l

TABLE OF CONTENTS (cont'd) l l

SEC'!ON TITLE PAGE 4.2.8 Loss of External Electrical Load and'or Turbine Trip 4 34 4.2.9 Loss of Normal Feedwater 4 05 4.2.10 Loss of Offsite Power to the Station Auxiliaries 4 37 4.2.11 Excessive Heat Removal Due to Feedwater Malfunction 4 38 4.2.12 Excessive Load increase incident 4 39 4.2.13 Accidental Depressurization of the RCS 4 40 4.2.14 Accidental Depressurization of the Main Steam System 4 40' 4.2.15 inadvertent Operation of ECCS During Power Operation 4 41 4.2.16 Complete Loss of Forced Reactor Coolant Flow 4 42 4.2.17 Single RCCA Withdrawal at Full Power 4 43 4.2.15 Rupture of a Main Steam Line 4-43 4.2.19 Major Rupture of a Main Feedwater Pipe 4 44 4.2.20 Single RCP Locked Rotor 4 46 4.2.21 Rupture of a CRDM Housing 4-47 4.2.22 Steam Line Break Mass Energy Releases 4 50 4.2.23 Setpoint impact 4 50 4.2.24 Conclusion 4 52 References 4 65 5.0 NUCLEAR FUEL EVALUATION 51 5.1 Core Design 51 5.2 Thermal Hydraulic Design 51 5.3 Fuel Rod Performance 52 6.0 NSSS/ BALANCE OF PLANT INTERFACE EVALUATION 61 6.1 Main Steam System 61 6.2 Main Feedwater System 61 6.3 Auxiliary Feedwater System 61 APPENDIX A TECHNICAL SPECIFICATIONS REVISIONS l

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WESTINGHOUSE PROPRIETARY CLASS 3 LIST OF iABLES TABLE TITLE PAGE 2-1 NSSS Performance Parameters 2-2 4.1-1 Input Parameters Used in LOCA Analyses 4-17 4.1-2 Large Break LOCA Results - Fuel Cladding Data - 4inima: Safeguards 4-18 4.1-3 Large Break LOCA Time Sequence of Events -

Min'. mum Safeguards 4-19 i

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WESTINGHOUSE PROPRIETARY CLASS 3 LIST OF FIGURES FIGURE TITLE PAGE l

4.1-1 LBLOCA - Peak Clad Temperature -

DECLG (CD = 0.4) 4-20 4.1-2 LBLOCA - (sre Pressure During Blowdown -

DECLG (CD = 0.4) 4-21 4.1-3 LBLOCA - Fluid Temperature -

DECLG (CD = 0.4) 4-22 4.1-4 LBLOCA - Core and Downcomer Mixture Levels During Reflood - DECLG (CD = 0.4) 4-23 4.1-5 LBLOCA - Inlet Velocity During Reflood -

DECLG(CD=0.4) 4-24 4.1-6 LBLOCA - Core Mass Flow Rate -

DECLG (CD = 0.4) 4-25 4.2-1 Startup from Subcritical - Nuclear Power vs. Time 4-53 4.2-2 Startup from Subcritical - Heat Flux vs.

Time 4-54 4.2-3 Startup from Suberitical - Fuel Temperature vs. fime 4-55 4.2-4 All Loops Operating, One Loop Coasting Down -

Flow Coastdowns vs. Time 4-56 4.2-5 All Loops Operating, One Loop Coasting Down -

Flux Transients vs. Time 4-57 4.2-6 All Loops Operating, One Loop Coasting Down -

DNBR vs. Time 4-58 kV 0035) 1D 972793 l

WESTINGHOUSE PROPRIETARY CLASS 3 l

LIST OF FIGURES (cont'd) l FIGURE TITLE PAGE 4.2-7 Major Rupture of a Main Feedwater Pipe, Without Offsite Power - Pressurizer Pressure and Volume vs, Time 4-59 4.2-8 Major Rupture of a Main feedwater Pipe, Without Offsite Power - Loop Temperature vs. Time 4-60 4.2-9 All Loops Operating, One Locked Rotor -

Pressure vs. Time 4-61 4.2-10 All Loops Operating, One Locked Rotor -

Core Flow vs. Time 4-62 4.2-11 All Loops Operating, One Locked Rotor -

Flux Transients vs. Time 4-63 4.2-12 All Loops Operating One Locked Rotor -

Clad Temperature vs. Time 4-64 I

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l WESTINGHOUSE PROPRIETARY CLASS 3 OBJECTIVES The Joseph M. Farley Nuclear Unit 2 is currently licensed to operate with a Reactor Coolant System (RCS) Thermal Design Flow (TDF) of 88,500 gpm per loop (265.500 gpm total flow) and up to 10% steam generator tube plugging.

This report supports the Alabama Power application to the Nuclear Regulatory Commission for approval to operate farley Unit 2 at increased steam generator tube plugging and reduced RCS Thermal Design Flow. A safety evaluation of NSSS design, operations, and analyses has been performed to provide the following information relevant to that application:

1.

A description of the proposed change in the NSSS design parameters of Joseph M. Farley Unit 2.

2.

An assessement of the impact of that change on NSSS equipment designs, safety analyset, and systems operations.

3.

A technical basis for establishing that the proposed increase in steam generator tube plugging and reduction in RCS Thermal Design Flow do not involve an unreviewed safety question in accordance with requirements of 10 CFR 50.59.

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WESTINGHOUSE OROPRIETARY CLASS 3 1

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l SECTION

1.0 INTRODUCTION

The current bases of the Joseph M. Farley Nuclear Unit 2 accident analyses and Nuclear Steam Supply System (NSSS) systems and components design assume a j

Reactor Coolant System (RCS) Thermal Design Flow (TDF) of 88,500 gpm per loop (265,500 gpmtotalflow). The Technical Specifications for the plant allow up to 10% steam generator tube plugging. Recently Alabama Power Company initiated a program to increase the allowable steam generator tube plugging limit and to redu:e the required RCS Thermal Design Flow for this unit.

As part of this prrgram Westinghouse has performed the analyses and evaluations necessary to verify that the Joseph M. Farley Unit 2 NSSS is structurally and functionally capable of safe and reliable operation at the new design parameters.

The analyses and evaluations were performed in accordance with the following criteria:

1.

The program encompassed all aspects of the NSSS design and operation which are impacted by the increased steam generator tube plugging and reduced Thermal Design Flow. The scope of the program included the NSSS safety analyses, the functional capability of the systems for normal and abnormal plant operations, and the mechanical design of the NSSS components and structures.

2.

Safety analyses performed as part of the program were executed to FSAR quality standards, using current Nuclear Regulatory Commission (NRC) approved analytical techniques, and were evaluated in accordance with the criteria and standards that apply to the current Farley Unit 2 operating license.

3.

NSSS system and component designs were evaluated in accordance with the regulatory requirements, codes, and standards which were applicable to Joseph M. Farley Unit 2 when it was originally oom an.*o 11 m-,

t WESTINGHOUSE PROPRIETARY CLASS 3 r

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licensed plus any subsequent criteria specifically applied to Farley Unit 2 by the NRC.

The following assumptions and design inputs formed the basis for the analyses and evaluations:

r 1.

An average steam generator tube plugging level of 15%, with a peak value in any one generator of 20%.

2.

A Thermal Design Flow reduced by approximately 1.5% to 87,200 gpm per loop (261,600 gpmtotal) 3.

A full core of Westinghouse 17x17 standard fuel.

4.

The thimble plugs in the reactor internals are installed.

5.

The core bypass flow is 4.5%.

The revised design parameters and the results of the evaluations and analyses performed during the program are presented in the following chapters.

l 1-2 com mom,o

WESTINGHOUSE PROPRIETARY CLASS S SECTION 2.0 OPERATING PARAMETERS 4

2.1 CURRENT LICENSED PARAMETERS l

i Joseph M. Farley Unit 2 is currently licensed to operate at an NSSS power level of 2660 MWt, with a Reactor Coolant System (RCS) Thermal Design Flow of 88,500 gpm per loop (265,500 gpm total flow), and up to 10% steam generator tube plugging.

The NSSS parameters associated with these conditions are shown in Table 2-1.

2.2 REVISED OPERATING PARAMETERS p

At the revised conditions, Farley Unit 2 will continue to operate at 2660 MWt (NSSS), but with a reduced RCS Therual Design Flow of 87,200 gpm per loop (261,600 gpm total) and an increased steam generator tube plugging level up to 15% average and 20% peak. The reduced Thermal Design Flow will cause a slight variation in the temperature rise across the reactor vessel (delta T),

or the difference between the reactor vessel inlet (cold leg) and outlet (hot leg) temperatures.

For a Thermal Design Flow of 87,200 gpm per loop, the hot leg temperature increases from 610.9 to 611.3'F and the cold leg temperature decreases from 543.5 to 543.1'F.

l l

The revised design parameters were calculated for two different levels of l

steam generator tube plugging: a 15% average tube plugging level; and a 20%

peak level. Performing the analyses and evaluations using this combination l

allows some flexibility for asymmetric plugging, if necessary.

Both cases result in a decrease in the steam generator' secondary side steam pressure and temperature.

For the 15% plugging conditions the steam would drop to a pressure of approximately 780 psia at a temperature of 515.5'F. To analyze for a peak plugging level of 20%, bounding parameters were calculated assuming a uniform 20% tube plugging but maintaining the Thermal Design Flow i

of 87,200 gpm/ loop. For_this case, the steam would drop to a pressure of approximately 765 psia at a temperature of 513.2*F.

The parameters for the revised conditions are shown in Table 2-1, 2-1 mu unm

CESTINGHOUSE PROPRIETARY CLASS 3 i

l TABLE 2.1 i

I ALABAMA POWER COMPANY Joseph M. Farley Unit 2 NSSS Performance Parameters for Increased S/G Tube Plugging and Reduced TOF 15% Avg.

20% Peak

• S/G Tube S/G Tube

Plugging, Plugging, Original TO Flow TO Flow Parameter Conditions Reduced Reduced NSSS Power, MWt 2660 2660 2660 Reactor Power, MWt 2652 2652 2652 Thermal Design Flow, gpm/ loop 88,500 87,200 87,200 Reactor Thermal Design Flow, Total, 106 lb/hr 100.67 99.25 99.25 Reactor Coolant Pressure, psia 2250 2250 2250 Reactor Coolant Temperature. 'F Core Outlet 613.7 614.2 614.2 Vessel Outlet 610.9 611.3 611.3 Core Average 580.3 580.5 580.5 Vessel Average 577.2 577.2 577.2 Vessel / Core Inlet 543.5 543.1 543.1 Steam Generator Outlet 543.3 542.9 542.9 Steam Generator Steam Temperature, 'F 517.2 515.5 513.2 Steam Pressure 793.0 781.2 765.7 SteamFlow,10Epsia lb/hr tot.

11.61 11.61 11.60 Feedwater Temperature, 'F 437.3 437.3 437.3

)

Zero Load Temperature, 'F 5^7.0 547.0 547.0 Core Bypass Flow, %

4.5 4.5 4.5 l

l This set of parameters were calculated to determine the steam generator secondary side temperatures and pressures associated with 20% plugging. The RCS primary parameters maintain the Thermal Design Flow associated with 15% average S/G tube plugging. These parameters provide conservative analytical inputs to bound a peak plugging level of 20% in any one5/G.

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I WESTINGHOUSE PROPR!ETARY CLASS 3 l

SECTION 3.0 NSSS SYSTEMS AND COMPONENTS EVALUATIONS 3.1 NSSS SYSTEMS 3.1.1 NSSS Fluid Svitems The Reactor Coolant System (RCS) and certain auxiliary NSSS systems provided by Westinghouse were evaluated to verify their continued adequacy for plant operation with the itcreased steam generator tube plugging and reduced Thermal Design Flow. For the proposed design parameters, the RCS temperatures remain within 0.5'F of the original design values.

Therefore the impact on the fluid systems is negligible and the RCS and NSSS auxiliary systems continue to comply with the origsna11y applicable design codes, standards, and criteria.

3.1.2 Reactor Control 3ystem and Marain to Trio Evaluation 3.1.2.1 Control System Response An evaluation of the RCS control systems was performed, including: rod control, pressurizer pressure and level control, steam dump control, and steam generator water level control.

The evaluation showed that the proposed design parameters will not have a major impact on the systems. The RCS and cold leg temperatures change by less than 0.5'F and the full load steam pressure decreases by 11 psi for the 15% average tube plugging case.

. hile W

the lower steam pressure will decrease the flow capacity of the steam dump system by about 1.5% at full power, this reduction will be less at 50% power, and will not affect the load rejection capabilities.

Both the RCS l

temperature and steam pressure changes are therefore insignificant to the control system response to normal operating transients.

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00353 10 072190 31

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WESTINGHC"SE PROPRIETARY CLASS 3 i

3.1.2.2 Margin to Trip The operating margin to reactor trips and ESS actuations is not significantly reduced as a result of the proposed design parameters.

As mentioned above, the changes in the NSSS parameters caused by the additional tube plugging and Thermal Design Flow reduction are small. This will cause little or no change in control system responses to normal operating transients.

Therefore, in most cases, the margin to the various reactor trips is not changed.

The two exceptions are the overtemperature delta-T (OTAT) trip and the low steam pressure Safety Injection actuation.

The margin to the OTAT trip decreased due to a change in the values used in the setpcint calculation.

The K1 value decreases from 1.22 to 1.18 and the positive gain in the f(AI) penalty function went from 1.60 to 1.75% OTAT/% al. These combine to produce a reduction of about 5% to the OTAT trip.

At least 15% in AT will be available for most operations beginning at full power.

This is more than sufficient for a 50% load rejection which creates the biggest challenge to the OTAT trip.

This assessment is based on the assumption that the core axial offset (or al) is not in excess of about 11%

at the beginning of a transient or greater than 13% during steady-state full power operation.

Above these values, insufficient margin to the OTAT trip may exist. The nuclear design documentation of the latest Unit 2 core cycle (WCAP-12193) predicts that the peak axial offset is 8.2% at the beginning of core life and will be less than 0% after about 30 effective full power days.

Thus, with a Constant Axial Offset Control (CAOC) operating band of 15%, the limits on al are maintained and adequate margin to the OTAT trip is preserved.

The margin to the low steam pressure Si actuation is reduced by 11 psi due to the reduction in the full power steam pressure. With a SI setpoint of 600 psia, this represents about a 5% reduction in margirl to about 180 psi.

The transients that create the largest challenge to this SI actuation are load increases to full power. The most severe of these for which the Farley Nuclear Unit 2 is designed is the 10% step load inc* ease to full power.

32 cem ioonm

WESTINGHOUSE PROPRIETARY CLASS 3 During this transient, steam pressure can decrease about 60 psi below the full power pressure and this can be che equivalent of about 150 psi when the lead / lag compensation in the steam prissure ciernel is considered.

This will itill leave about 30 psi to the SI setpoint.

Considering that farley Unit 2 will normally increase load in much smaller increments at much slower rates, this is considered adequate margin.

3.2 COMPONENT DESIGN TRANSIENTS Systems Standard Design Criteria (5500) 1.3, "NSSS Design Transients",

contain the postulated number of occurrences and the pressure, temperature and flowrate transients for which the primary components and certain auxiliary system components are designed to withstand.

The design transients specified for the design and analysis of Westinghouse-supplied equipment include operational transients which may result from either normal, upset, emergency, faulted, or test conditions, as defined by Section Ill of the ASME Code.

The design transients originally used for Joseph M. Farley Unit 2 NSSS components are contained in Westinghouse System Standard Design Criteria 1.3 Rev. 1, Nuclear Steam Supply System Design Transients (SSDC 1.3 Rev. 1).

An evaluation was conducted to determine the continued applicability of the design transients in SSDC 1.3 Rev. 1 for the revised parameters shown in Table 2-1.

The revised conditions result in RCS temperature changes of less than 0.5'F, and a steam temperature decrease of 4'F.

Both of these design parameter changes are conservatively bounded by the original Farley Unit 2 design transients. Therefore the Design Transients contained in SSDC 1.3 Rev. 1 remain in effect.

3.3 NSSS PRIMARY COMPONENTS 3.3.1 Reactor Vessel While the proposed change in the steam generatc* tube plugging level alone does not impact the reactor vessel, a change in the Thermal Design Flow does affect the temperature rise across the component. Accordingly, the Farley Unit 2 reactor vessel was evaluated for the impact of a slightly lower inlet ociss tocmso 33

WESTINGHOUSE PROPRIETARY CLASS 3 temperature (543.1 instead of 543.5'F) and the impact of a slightly higher outlet temperature (611.3 instead of 610.9'F).

The evaluation confirmed that the stress intensities and fatigue usage factors for the critical parts of the Farley Unit 2 reactor vessel are still within the limits of the stress analysis, so the reactor vessel remains in compliance with the codes and standards in effect when the unit was originally licensed.

3.3.2 Reactor pressure Vessel Svstem The reactor pressure vessel system consists of the reactor vessel, the reactor upper and lower internals assemblies and the reactor core.

Since these components are interdependent from a thermal-hydraulic and structural viewpoint, they were evaluated as a system.

The reactor pressure vessel system is not directly impacted as a result of steam generator tube plugging, however, it is sensitive to variations in the reactor coolant system flowrate. Therefore the reactor pressure vessel system was evaluated with respect to the reduction in the Thermal Design Flow.

New flows and pressure drops were calculated for the various flow paths within the reactor pressure vessel system.

The results showed that the changes in pressure drops associated with the new design parameters are evenly distributed throughout the reactor internals, and that the total pressure drop across the internals would decrease an insignificant amount.

Since the internals flow and pressure drop changes are not changed significantly by the new design parameters, detailed calculations of the

'ffect on core bypass flow, hydraulic lift forces, flow induced vibration, and Rod Control Cluster Assembly (RCCA) rod drop times were not necessary.

The second result of a Thermal Design Flow reduction is to increase the temperature rise across the reactor vessel (i.e., hot leg temperature - cold leg temperature). For the approximate 1.5% Thermal Design Flow reduction, l

included in the revised Farley Unit 2 design parameters, the delta-T increases by 0.8'F.

This r.orresponth to a change in reactor inlet temperature of -0.4'F an in the reactor outlet temperature of +0.4*F.

Temperature variations of.nis magnitude are bounded by the original structural analyses of the Farley Unit 2 internals.

l mu wve 34

WESTINGHOUSE PROPRIETARY CLASS 3 l

l The evaluation of the reactor pressure vessel system demonstrated that there l

would be no adverse impact on the performance of the system by the proposed increased steam generator tube plugging nor the reduced Thermal Design Flow.

l 3.3.3 Reactor Coolant Pump and RCP Motor The Model 93A reactor coolant pumps (RCPs) and RCP motors used in Farley Unit 2 were evaluated for the proposed design parameters shown in Table 2-1.

At these conditions the RCP is exposed to a new steam generator outlet temperature of 542.9'F, which represents a 0.4*F drop.

The pump is also exposed to higher flow resistance due to the increased plugging level.

The evaluation of the pump showed that the structural analysis is not impacted by the temperature change, and that the new hydraulic loads do not have any adverse safety impact on the pump internals or motor.

3.3.4 Control Rod Drive Mechanisms Each Control Rod Drive Mechanism (CRDM) consists of four basic subassemblies:

the pressure housing, the latch assembly, the drive rod, and the operating coil stack. The pressure housing is threaded and seal-welded to the reactor vessel head, and is therefore exposed to the core outlet temperature.

The core outlet temperature for the proposed design parameters will be 614.2'F compared to a value of 613.7'F for the current design conditions.

This temperature increase was reviewed and determined to have no significant effect on thermal and structural analyses performed for the control rod drive mechanisms.

Review of the original design analyses verified continued compliance with the codes and standards in effect when Farley Uni.t 2 was originally licensed.

3.3.5 Pressurizer The pressurizer is interconnected with the Reactor Coolant System through the hot leg (surge line) and the cold leg (spray line).

During an insurge, the spray system, condenses steam in the pressurizer vessel to prevent the pressurizer pressure from reaching the setpoint of the power-operated relief oem im ma 35

WESTINGHOUSE PROPRIETARY CLASS 3 valves. During an outsurge, flashing of water to steam and generation of steam by automatic actuation of the pressurizer heaters maintains the pressure above the low pressure reactor trip setpoint. While the proposed change in the steam generator tube plugging level alone does not impact the pressurizer, a change in the Thermal Design Flow does affect the temperatures to which the pressurizer is exposed.

Therefore the effect of the proposed operating temperatures on the pressurizer was evaluated.

The results indicated that there will be no significant changes to the pressurizer j

stresses or fatigue usage factors determined by the original stress analyses.

Therefore, the pressuri2er will remain in compliance with the codes and standards in effect when Farley Unit 2 was originally licensed.

3.3.6 Reactor Coolant loop Pioina and Primary Eauipment Supports The design basis analysis for the reactor coolant loop piping, the primary equipment supports, and the primary equipment nozzles was evaluated for possible impact by the proposed design parameters shown in Table 2-1.

The inputs significant to the evaluation were the RCS operating temperatures, the NSSS component design transients, the LOCA hydraulic forcing functions, and the reactor pressure vessel dynamic LOCA displacements. The design transients did not change, and the changes to the other inputs are small.

The temperatures change by less than 0.5'F, the forcing functions change by less than 0.3%, and the displacements do not change by any appreciable amour.t.

Therefore the design basis analysis and evaluation remain applicable

(

for the Farley Unit 2 reactor coolant loop piping, primary equipment supports, and primary equipment nozzles for the proposed design parameters.

l 3.3.7 Steam Generators The steam generator evaluation for the revised conditions was divided into three sections - a structural evaluation, a thermal-hydraulic evaluation, and a U-bend tube fatigue evaluation.

wummo 36

WESTINGHOUSE PROPRIETARY CLASS 3 3.3.7.1 Structural Evaluation A structural evaluation of the Model 51 steam generators was performed to determine the impact of operation at 15% (20% peak) steam generator tube plugging with reduced Thermal Design Flow conditions.

The evaluation was based on Design Transients in System Standard Design Criteria 1.3, Rev. 1, l

and was performed according to the code requirements in effect when the unit was originally licensed (ASME B&PV Code,Section III, 1968 and 1971 Editions).

The limiting components including: the tubesheet and shell junction; the divider plate; the steam generator tubes; the tube /tubesheet weld; the nozzles, and the shell including the upper shell penetrations were considered.

The evaluation demonstrated that the RCS temperature changes (1 0.4'F) were insignificant, and would have no effect on the primary side components of the steam generator. The increased pressure differential across the primary to secondary L.Jndary due to lower steam pressures, caused only slight increases in the component stress ranges.

The resulting increases in fatigue usages were small and within acceptable limits.

The minimum acceptable tube wall is not significantly affected.

in summary, the structural evaluation of the limiting components of the steam generators verified that at the proposed conditions the steam generators will remain in compliance with the codes and standards in effect when Farley Unit 2 was originally licensed.

3.3.7.2 Thermal-Hydraulic Evaluation The thermal-hydraulic evaluations were performed to determine the effect of the proposed design conditions shown in Table 2-1. The evaluations included the expected performance of the currently installed steso generator moisture separator package, and other operating characteristics h as: the circulation ratio; stability; and secondary side mass. ihe moisture carryover at the proposed conditions will be less than the 0.1% design limit and is therefore acceptable.

The circulation ratio, which is the total bundle flow divided by steam flow, will remain unchanged at the new 37 oom wn.o

WESTINGHOUSE PROPRIETARY CLASS 3 conditions, and the proposed design parameters will have an insignificant effect on the thermal-hydraulic stability of the units, and the predicted secondary mass.

The thermal-hydraulic evaluation demonstrated that the steam generator performance at the proposed conditions will continue to be acceptable.

3.3.7.3 U-bend Tube Fatioue Evaluation An evaluation was performed to assess the impact of the proposed design parametersonthepotentialforflowinducedvibration(FIV)ofinnerrowU-bend tubes to cause a tube rupture in a manner similar to the tube that ruptured at North Anna Unit 1 R9C51.

A complete evaluation for potential U-b0nd vibration and fatigue at current ratings for Farley Units 1 & 2 was comp h -d and reported in WCAP-11875

" Joseph M. Farley Unit 1 and 2 Evaluation for Tube Vibration Induced Fatigue" July 1988.

That evaluation determined that no tubes required preventive action.

For this evaluation, the previous calculations were modified with a one dimensional technique, described in the above WCAP, to determine the effect of changes in U-bend fluid conditions. The calculation method evaluated new U-bend flow conditions based on the changes in steam pressure and bundle flow which result from the proposed operation with increased tube plugging / reduced RCS flow.

The vibration potential and fatigue usage for each susceptible tube was then determined.

The results of the calculations showed that the fatigue usages for each susceptible tube remained within acceptable ranges, and therefore no tubes will require preventive actions for operation at the revised conditions.

l ocm io om,o 38

1 WESTINGHOUSE PROPRIETARY CLASS 3 l

3.3.7.4 Conclusion A detailed multi-faceted evaluation of the steam generator was performed assuming operation at the increased tube plugging and reduced Thermal Design Flow conditions.

The evaluation confirmed that the change in operating parameters will have an small effect on the structural, thermal-hydrellic, and U-Bend tube vibration analyses for the steam generator, however the generator will continue to meet the industry codes and standards in effect

^

when the unit was originally licensed and the U-bend tube vibration acceptance criteria.

3.4 NSSS AUXILIARY EQUIPMENT The impact of the increased steam generator tube 'alugging and reduced Thermal Design Flow on the NSSS auxiliary equipment was tvaluated.

It was determined that the small change in the RCS temperatures will not significantly impact the operation of these componente,.

i 1

oom toctuso 39

WESTINGHOUSE PROPRIETARY CLASS 3 i

SECTION 4.0 ACCIDENT ANALYSES AND EVALUATION l

4.1 LOCA EVALUATIONS 4.1.1 Small Break LOCA The licensed FSAR small break LOCA analysis for Farley Unit 2 was performed using the Westinghouse WFLASH Evaluation Model (Ref. 4.1) assuming 0% steam generator tube plugging (SGTP) and a loop thermal design flow of 88,500 gpm.

The 6-inch equivalent diameter limiting break Peak Cladding Temperature (PCT) including all safety evaluations for Farley Unit 2 is 1797'F. The Westinghouse Small Break LOCA analysis consists of a Reactor Coolant System (RCS) thermal hydraulic model and a hot rod heatup model.

For the Farley Unit 2 these analyses were performed with the WFLASH and LOCTA computer codes, respectively.

Analyses (Ref. 4.4) were performed to demonstrate, in general, that the NOTRUMP small break LOCA evaluation model (Ref. 4.5) calculates lower PCTs than the WFLASH evaluation model. As such, the Farley WFLASH small' break LOCA analysis remains the analysis of record provided that plant changes do not adversely affect the results.

The effects of the increase in SGTP and the reduction in Thermal Design Flow (TOF) on the transient characteristics are discussed below.

l Steam Generator Tube Plugging can potentially affect the performance of the calculated small break LOCA analysis in three important aspects; namely:

1.

the reduction in the steam generator heat transfer area between the primary and secondary, i

2.

the increase in temperature difference between the primary and secondary j

initial operating temperatures, and i

3.

the potential effect on the countercurrent flow limit due to the reduction in flow area.

oom io omeo 4-1

l WESTINGHOUSE PROPRIETARY CLASS 3 l

Direct model sensitivities to SGTP are not available for the WFLASH model.

However, References 4.2 and 4.3 document the small break LOCA cases which were performed with NOTRUMP to assess the effect of these parameters on a small break LOCA transient.

Based on these analyses, the effects of tube plugging on small break LOCA transients are discussed in the following:

1.

During Small Break LOCA only a small portion of the steam generator heat transfer area is required to provide an effective heat sink. With up to 20% of the tubes plugged, there would still be sufficient heat transfer area to provide an effective heat sink to the primary side.

2.

As a result of the increased SGTP and the T0F reduction, the temperature difference between the primary and the secondary will increase slightly during steady state plant operation.

The effect of this temperature difference on the heat transfer capability betseen the primary and the secondary was studied in Reference 4.3.

It was determined that the increase in temperature difference between the primary and the secondary side will disappear right after the break, as the secondary side pressure stabilizes at the steam generator safety valve set point and acts as a governing influence on the continued primary blowdown.

Considering the rapid rate at which this occurs, it can be concluded that there would be a negligible effect due to the temperature difference.

3.

For a small break in the cold leg. the steam generated in the core flows to the steam generators. A part of this steam is condensed in the steam generator tubes and flows back to the core along with the drained liquid. This counicrcurrent flow affects the draining rate from the steam generators and directly affects core uncovery. A decrease in the countercurrent flow rate could lead to an increase in the PCT, depending on the level of core uncovery.

Reference 4.3 documents a study performed to assess the countercurrent flow effect under the assumption that there would be some lootion between the vessel upper plenum and the steam generator which would 42 ocm imv,o

WESTINGHOUSE PROPRIETARY CLASS 3 restrict the countercurrent flow to the greatest extent.

Restrictions at several locations were studied, and it was determined that the limitations to countercurrent flow were the greatest in the inclined piping between the hot leg and the steam generator inlet plenum.

It was concluded from the Reference 4.3 study that at low SGTP levels (up to 20%), the limiting countercurrent flow resistance in the inclined pipe would govern the draining of the tubes,and the tube plugging would have no effect on this phenomenon.

f Further consideration of the reduced TDF shows that there would be no

,significant difference in the RCS depressurization, reactor trip, the RCS thermal hydraulic respense, and the fuel rod initial parameters due to the small change in primary temperatures and pressures.

Based on the previous discussions, the reduction in loop TDF to 87200 gpm and up to 20% SGTP will have no significant effect on the small break LOCA analysis results.

Thus the changes will have no adverse effect on the Farley small br,eak margin to the PCT limit of 2200*F. Furthermore, there would be no effect on the conclusions of WCAP-11145-P-A (Ref. 4.4), which demonstrated that the previous NRC approved WFLASH Small Break LOCA Evaluation Model results were conservative when compared with the NRC approved NOTRUMP Small Break LOCA Evaluation Model.

4.1.2 Large Break Loss-of-Coolant Accident This analys i considers the effect of the increased Steam Generator Tube Plugging (SGTP), reduced Thermal Design Flow (TDF) and associated changes to the NSSS design parameters on the large Break LOCA ECCS analycis.

Full power, steady state reactor coolant pressure, and vessel average temperature are maintained for the increased SGTP and reduced T0F. However, the increased SGTP and reduced TDF result in a decrease in vessel inlet temperature of 0.4*F, an increase in vessel outlet temperature of 0.4*F, a 27 psi reduction in steam pressure, and a 4.0*F reduction in steam temperature.

com iew,o 43

WESTINGHOUSE PROPRIETARY CLASS 3 i

l In support of the increased SGTP and reduced TDF at Farley Unit 2. the limiting f.ase large break LOCA ECCS analysis has been performed. The results of this t.nalysis is presented in this section.

Identification of Causes and frequency Classification A Lots-of-Coolhnt Accident (LOCA) is the result of a pipe rupture of the RCS pref,sure boundary.

For the analysis reported here, a major pipe break (large break) is defined as a rupture with a total cross-sectional area equal to or greater than 1.0 square foot (ft2).

This event is considered an ANS Condition IV event, a limiting fault, in that it is not expected to occur during the lifetime of the plant but is postulated as a conservative design basis.

The Acceptance Criteria for the LOCA are described in 10CFR50.46 as follows:

1.

The calculated PCT is below the requirement of 2200'F.

2.

The amount of fuel element cladding that reacts chemically with water or steam does not exceed 1 percent of the total amount of Zircaloy in the reactor.

3.

The clad temperature transient is terminated at a time when the core geometry is still amenable to cooling.

The localized cladding oxidation limit of 17% is not exceeded during or after quenching.

4.

The core emains amenable to cooling during and after the break.

5.

The core temperature is reduced and decay heat is removed for an extended period of time. This is required to remove the heat from the long lived radioactivity remaining in the core.

I These criteria were established to provide significant margin in Emergency Core Cooling System (ECCS) performance following a LOCA.

l oom toomeo 44 i

l WESTINGHOUSE PROPRIETARY CLASS 3 Sequence of Events and Systems Operations Should a major break occur, depressurization of the RCS results in a pressure decrease in the pressurizer. The reactor trip signal subsequently occurs when the pressurizer low pressure trip setpoint is reached.

A safety injection signal is generated when the appropriate setpoint is reached.

These countermeasures will limit the consequences of the accident in two ways:

t 1.

Reactor trip and borated water injection complement void formation in the core and cause a rapid reduction of nuclear power to a residual level corresponding to the delayed fission and fission product decay heat.

However, no credit is taken during the LOCA blowdown for negative reactivity due to boron content of the injection water.

In addition, the insertion of control rods to shut down the reactor is neglected in the large break analysis.

2.

Injection of borated water provides the fluid medium for heat transfer from the core and prevents excessive clad temperatures.

Description of a large Break LOCA Transient Before the break occurs, the unit is in an equilibrium condition, i.e., the heat generated in the core is being removed via the secondary system.

During blowdown, heat from fission product decay, hot internals, and the vessel continues to be transferred to the reactor coolant. At the beginning of the blowdown phase, the entire RCS contains subcooled liquid which transfers heat from the core by forced convection with some fully developed nucleate boiling. Thereafter, the core heat transfer is based on local conditions with transition boiling, film boiling, and forced convection to steam as the major heat transfer mechanisms.

The heat transfer between the RCS and the secondary system may be in either direction depending on the relative temperatures.

In the case of continued heat addition to the secondary, secondary system pressure increases, and the atmospheric relief and/or main steam safety valves may actuate to limit the oom w mo 4-5

t WESTINGHOUSE PROPRIETARY CLASS 3 i

pressure. Makeup water to the secondary side is automatically provided by the Auxiliary Feedwater System.

The safety injection signal actuates a feedwater isolation signal which isolates normal feedwater flow and also initiates auxiliary feedwater flow by starting the auxiliary feedwater pumps.

The secondary flow aids in the reduction of RCS pressure. When the RCS depressurizes to approximately 600 psia, the cold leg accumulators begin to t

inject borated water into the reactor coolant loops.

Since the loss of offsite power is assumed, the reactor coolant pumps are assumed to trip at the time of reactor trip during the accident.

The effects of pump coastdown are included in the blowdown analyses.

The blowdown phase of the transient ends after the RCS pressure (initially e

assumed at a nominal 2280 psia) falls to a value approaching that of the containment atmosphere.

Near the end of the blowdown, the mechanisms that are responsible for the bypassing of emergency core cooling water injected into the RCS are calculated not to be effective.

At this time (called end-of-bypass), refill of the reactor vessel lower plenum begins.

Refill is complete when emergency core cooling water has filled the lower plenum of the reactor vessel which is defined by the bottom of the fuel rods (called bottom of core recovery time).

The reflood phase of the transient is defined as the time period lasting from the end-of-refill until the reactor vessel has been filled with water to the extent that the core temperature rise has been terminated.

Fr" the later stage of blowdown and then the beginning-of-reflood, the safety injection

]

accumulator tanks rapidly discharge borated cooling water into the RCS, contributing to the filling of the reactor vessel downcomer.

The downcomer water elevation head provides the driving force required for the reflooding

]

of the reactor core. The ECCS safety injection aids in the filling of the downcomer and subsequently supplies water to maintain a full downcomer and i

l complete the reflooding process.

l Continued operation of the ECCS pumps supplies water during long term cooling. Core temperatures have been reduced to long-term steady state co m ico m eo 4-6 I

,_,a.

WESTINGHOUSE PROPRIETARY CLASS 3 levels associated with dissipation of residual hect generation. After the water level of the refueling wate-storage tank reaches a minimum allowable value, coolant for long-term cooling of the core is obtained by switching to the cold leg recirculation phase of operation in which spilled borated water is drawn from the containment sump and returned to tLe RCS cold legs.

The Containment Spray System continues to operate to further reduce containment pressure. Approximately 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> after initiation of the LOCA, the ECCS is realigned to supply water to the RCS hot legs in order to control the boric acid concentration in the reactor vessel.

Method of Analysis The requirements of an acceptable ECCS evaluation model are presented in Appendix K of 10 CFR 50. The requirements of Appendix K regarding specific model features were met by selecting models which provide a significant overall conservatism in the analysis. The assumptions made pertain to the conditions of the reactor and associated safety system equipment at the time that the LOCA occurs and include such items as the core peaking factors, the containment pressure, and the performance of the ECCS. Decay heat generated throughout the transient is also conservatively calculated as required by Appendix K of 10 CFR 50.

Large Break LOCA Evaluation Model The analysis of a large break LOCA transient is divided into three phases:

(1) blowdown. (2) refill, and (3) reflood. There are three distinct transients analyzed in each phase:

(1) the thermal-hydraulic transient in the RCS, (2) the pressure and temperature transient within the containment, and (3) the fuel and clad temperature transient of the hottest fuel rod in the core.

Based on these considerations, a system of interrelated computer codes has been developed for the analysis of the LOCA.

The description of the various aspects of the LOCA analysis methodology is given in References 4.6 through 4.15.

These document 5 describe the major phenomena modeled, the interfaces among the computer codes, and the features of the codes which ensure compliance d th the Acceptance Criteria.

The mu iwnw 47

I WESTINGHOUSE PROPRIETARY CLASS 3 SATAN-VI, WREFLOOD, COCO, and LOCTA-IV codes which are used in the LOCA analysis, are des.:ribed in detail in References 4.7 through 4.10.

Modifications to these codes are specified in References 4.11 through 4.13.

The BART code is described in Reference 4.14.

The BASH and LOCBART codes are described in Reference 4.15.

These codes are used to assess the core heat transfer geometry and to determine if the core remains amenable to cooling throughout the blowdown, refill, and reflood phases of the LOCA. The SATAN-VI computer code analyzes the thermal-hydrsulic transient in the RCS during blowdown, and the WREFLOOD and BASH computer codes are used to calculate this transient during the refill and reflood phases of the accident.

The BART computer code is used to calculate the fluid and heat transfer conditions in the core duriig reflood. The C0CO conputer code is used to calculate the containmers in.ssure transient during all three phases of the LOCA analysis.

Similarly, thi LOCBART computer code is used to compute the thermal transient of the hottes' fuel rod during the three phases.

The large break analysis was performed with the approved December, 1981 version of the Evaluation Model (Ref. 4.12), with the approved version of BASH (Ref. 4.15).

SATAN-VI is used to calculate the RCS pressure. enthalpy, density, and the mass and energy flow rates in the RCS, as well as steam generator energy transfer between the primary and secondary systems as a function of time during the blowdown phase o' the LOCA. SATAN-VI also calculates the accumulator water mass and internal pressure and the pipe break mass and energy flow rates that are assumed to be vented to the contain9ent during blowdown. At the end of the blowdown and refill phases, these data are transferred to the WREFLO0E code. Also at the end-of-blowdown, the mass and energy release rates during blowdown are transferred to the C0C0 code'for use in the determination of the containmerit pressure response during these phases of the LOCA. Additional SATAN-VI output data from the end-of-blowdown, including the core pressure and the core power decay transient, are input to the LOCBART coce.

With input from the SATAN-VI Code, WREFLOOD uses a system thermal-hydraulic model to determine the core flooding rate (i.e., the rate at which coolant mumum 48

l L

WESTINGHOUSE PROPRIETARY CLASS 3 l

enters the bottom of the core), the coolant pressure and temparature, anJ the quench front height during the refill phase of the LOCA. WREFLOOD also calculates the mass and energy flow addition to the containment through the break. Since the mass flow rate to the containment depends upon the core flooding rate and the local core pressure, which is a function of the containment backpressure, the WREFLOOD and C0C0 codes are interactively linked.

BASH is an integral part of the ECCS evaluation model which provides a realistic thermal-hydraulic simulation of the reactor core and RCS during the reflood phase of a LOCA.

Instantaneous values of accumletor conditions and safety injection flow at the time of completion of iower plenum refill are provided to BASH by WREFLOOD.

BASH has been substituted for WREFLOOD in calculating transient values of core inlet flow, en'.halpy, and pressure for the detailed fuel rod model, LOCBART. A detailed de:.cription of the BASH code is available in Reference 4.15.

The BASH code provides a sophisticated treatment of steam / water flow phenomena in the reactor coolant system during core reflood. A dyr.amic interaction between the core thermal-hydraulics and system behavior is expected, and recent experiments have borne this out.

In the BASH code reflood model, BART provides the entrainment rate for a given flooding rate, and then a system model determines loop flows and pressure drops in response to the calculated core exit flow.

An updctcf inlet flow is used to calculate a new entrainment rate.

This system produces a dynamic flooding rate, which r3flects the close coupling between core thermal-hydraulics and loop behavior.

The r0C0 code is a mathematical model of the containment. COCO is run using mass and energy releases to the containment provided by SATAN and WREFLOOD.

COC0 is described in detail in Reference 4.9.

The LOCBART ude is a coupling of LOCTA-IV and BART.

The LOCTA-IV code is a computer program that evaluates fuel, cladding, and coolant temperatures during a LOCA. A more complete description of LOCTA-IV can be found in i

Reference 4.10.

In the LOCTA detailed fuel rod model, for the calculation of local heat transfer coefficients, the empirical FLECHT correlation is l

replaced by the BART code.

BART employs rigorous mechanistic models to 1

com suonm 49 1

k WESTINGHOUSE PROPRIETARY. CLASS 3 generate heat transfer coefficients appropriate to the actual flow and heat transfer regimes experienced by the LOCTA fuel rods.

This is considered a more dynamic realistic' approach than relying on'a static empirical correlation.

Large Break input Parameters and Initial Conditions Table 4.1-1 lists important input parameters and initial conditions used in the large break analyses. To account for the increased steam generator tube plugging and the reduced thermal design flow in the Large Break LOCA analysis at Farley Unit 2, the limiting discharge coefficient (CD = 0.4) for minimum safeguards ECCS capability and operability has been analyzed.

Large Break Results Based on the results of the LOCA sensitivity studies, the limiting large break was found to be the dot'51e ended cold leg guillotine (DECLG).

Therefore, only the DECLG break is considered in the large break ECCS performance analysis. Calculations were perform d for the limiting Moody break discharge coefficient (CD = 0.4) under m himum safesuarjs conditions and the results of this calculation are summarized in Tab 1'e', 4.1-2 and 4.1-3.

Graphical results are presented in Figures 4.1-1 through 4.1-6.

The PCT calculated for the large break LOCA to account for increased steam l

generstor tube plugging and reduced thermal design finw is 2069'F.

Added to the calculated PCT is a 4*F increase due to-delayed isolation of-the containment mini-purge valves. This brings-the restritant-PCT to 2073'F at L

the Farley Unit 2.

In addition, the impact of steam generator flow area l

reduction due to seismic effects has been considered, and a PCT penalty'of 50'F has conservatively been assessed.

The resulting PCT for Farley Unit 2 is 2123*F, which is below the 10 CFR 50.46 limit of 2200*F. See below for a l

complete description of the flow area reduction issue. The maximum local.

metal-water reaction is 5.76' percent, which is well below the embrittlement limit of 17 percent as required by 10 CFR 50.46' The total core metal-water reaction is less than 0.3 percent for all breaks, as compared with the 1%

criterion of 10 CFR 50.46, and the clad tem]erature transient is terminated l

com vomo 4 10

WESTINGHOUSE PROPRIETARY CLASS 3-at a time when the core geometry is still. amenable to cooling.

As a result, the core temperature will continue to drop, and the' ability to remove decay heat generated in the fuel for an extended period of time will be provided.

Description of SG Flow Area Reduction Issue

Background

Licensees are normally required to provide assurance that there exists only i

an extremely low probability of abnormal leakage or gross rupture of any part of the reactor coolant pressure boundary (General Design Criteria 14 and 31).

l The NRC issued a Regulatory Guide (RG 1.121) which addressed this requirement f

specifically for steam generator tubes in pressurized water reactors.

In that guide, the staff required analytical and experimental evidence that steam generator tube integrity will be maintained under design basis conditions such as a loss of Coolant Accident (LOCA) in combination with a Safety Shutdown Earthquake (SSE). This analysis would provide the basis for establishing criteria for removing from service tubes which had experienced significant degradation.

Analyses performed by Westinghouse in support of the above requirement for various utilities combined the most severe LOCA loads with the bounding SSE.

~

Generally, these analyses showed that the combined tube collapse reduces the flow area through the steam generator. The reduced flow area increases the resistance to the flow of steam from the core during a LOCA, which potentially could increase the calculated PCT.

It was concluded at the time the initial analyses were performed that this level of tube collapse did not significantly affect the consequences, in terms of PCT, of the most severe load LOCA, which was the double-ended break at the steam generator outlet.

This conclusion was reached because the LOCA at the steam generat7r outlet results in substar,tially lower PCT than the LOCA at the cold leg.

Additional steam generator design evaluations have shown a small level of tube collapse due to SSE alone.

This tube collapse and resulting flow area reduction would then be present in all LOCAs, including the cold leg I.0CA,,f the LOCA was assumed to occur in combination with the SSE.

com io enm 4 11

WESTINGHOUSE PROPRIETARY CLASS 3 While nc specific regulatory position exists which clearly addresses the requirements in this area, Westinghouse has concluded that it must be assumed, on the basis'of the regulations, previous communications with the

[

staff on this issue, and regulatory positions in related areas,'that this L

effect should be accounted for in the LOCA analysis.

It should be noted, however, that plants'are currently required by their Tech Specs to inspect their steam generators following a seismic event which exceeds an Operating Basis Earthquake.

Since'ovalized or collapsed tubes would be removed from service as a result of this-inspection, it would appear that these steps would preclude the presence of collapsed tubes in the unlikely event of a LOCA.

In addition, the FNP plant specific SSE is much less than the bounding SSE.

A bounding flow area-reduction of five percent has been' established for the Farley Model 51 steam generators. This flow area reduction, if included in the LOCA analysis, has been conservatively estimated to result ~in a 50'F increase in peak cladding temperature. This results in.a PCT of 2123*F, which is below the limit of-2200'F, and therefore remains-in compliance-with the regulations while this issue is being resolved.

4.1.3 Miscellaneous LOCA Considerations Blowdown Reactor Vessel and Loop Forces The blowdown vessel and loop forcing functions are used as input to the design basis analysis for the effects of a-postulated LOCA on.the mechanical' and structural integrity of the reactor vessel internals and loops. 'A sensitivity study has shown that the primary influence of steam generator tube plugging and thermal design flow reduction on the LOCA hydraulic forcing functions is the change in the plant's operating temperatures. With a reduction in TOF, minor changes could be expected in LOCA reactor vessel forces, primarily as a re3 ult of a slight decrease in cold leg temperature.

The reduction in TDF to 87,200 gpm per loop results in less than a l' cold leg temperature change, the magnitude of the vessel and core barrel mumnm 4-12

WESTINGHOUSE PROPRIETARY CLASS 3.

horizontal forces'has been estimated to increase less than 0.3%.

This is considered a negligible effect for the LOCA forcing. functions at Farley Unit 2.

Post-LOCA Long-term Core Cooling for post-LOCA cooling considerations, steam generator tube pluggiaq will reduce the primary water volume.(RCS' volume), and since the RCS boren concentration is less than the RWST, the RCS is a source of diluticn in e

1st-LOCA-containment sump' concentration..Thus, a reduction determinir in the RC

, a result in steam generator tube plugging will_resultLin a slight"

.en.-

the p..t-LOCA sump boron concentration given that the pre-LOCA RCS boron concentration is '

' than the RWST boron concentration.

On this basis, the anticipated.increas.

.n the steam generator tube plugging up to 20% will not reduce the containment sump post-LOCA boron calculatior.

cirrently being used to satisfy this requirement.

The post-LOCA containment sump boron concentration calculation is dependent on the RCS, RWST, and accumulator water volumes and boron concentrations.

Since the reduction of thermal design flow does not affect the boron concentrations or volumes assumed for the RCS, RWST, and accumulators, the reduced TDF would not affect the post-LOCA sump boron calculation.. The increased tube plugging has an insignificant'effect on' sump pH and the.pH l

will be maintained within the proper range.

Hot Leg Switchover to Prevent Potential Boron Precipitation Post-LOCA hot leg recirculation switchover time is determined for inclusion in emergency procedures to ensure no boron precipitation in the reactor vessel following boiling in the core. The current hot leg switchover time for Farley Unit 2 is 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br />. This time is dependent on power: level and the RCS, RWST and accumulator water volumes and bcron concentrations. Since the reduction of thermal design flow does not affect the power level or the maximum boron concentrations or volemes assumed for the RCS, RWST, and i

accumulators, the reduced TDF would not affect the post-LOCA hot leg switchover time.

4 13 com mamu

L WESTINGHOUSE PROPRIETARY CLASS.3 j.

However, the RCS volume will be reduced slightly as the SGTP level is increased. Moreover, since the RCS is a source of dilution, the computed switchover time will be slightly earlier.

Plugging up to 20% of the steam generator tubes reduces the RCS volume by approximately 5% which subsequently results in an 0.65% reduction in the total available post-LOCA water volume, A conservative estimate of this effect would be to shorten the time to hot leg switchover by 0.65% of 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> or 0.072 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

The actual calculated switchover time was 11.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. Therefore, there is adequate margin present in the analysis to account for up to 20% steam generator tube plugging without invalidating the 11 hour1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> switchover time.

LOCA Mass and Energy Releases The impact of increasing the tube plugging _ level from a uniform 10% to an average level of ;51, with a peak'in any generator of 20%, and of reducing the Thermal Design Flow (T0F) by approximately 1.5% has been assessed.

For the short-term and long-term releases, utilized in the subcompartment and the containment integrity analyses, modeling of steam generator plugging is a l-benefit. The reactor coolant system volume, and therefore-the amount of energy available to be released to containment, is reduced.

Minor reductions in TOF, on the order of 5% or less, have a negligible impact on the short-term and long-term mass and energy releases. The reduction in the TOF by approximately 1.5% is therefore insignificant.

l l

The small changes in reactor coolant system temperatures associated with the l

TOF reduction cnd the increased tube plugging were evaluated, and'it was determined that there was an insignificant effect to'LOCA mass and energy l

release calculations.

Therefore, the impact of increasing the tube plugging level from a uniform 10% to an average level of 15% with a peak in any l

generator of 20%, and of reducing the T0F by approximately 1.5% has been assessed. The impact was determined to be insignificant. Since, the current LOCA short-and long-term mass and energy releases remain applicable, the containment responses remain valid.

oom io onm 4 14 4

WESTINGHOUSE PROPRIETARY CLASS 3

4. l' 4 Steam Generator Tube Rupture l

For the' Steam Generator Tube Rupture (SGTR) event, the Farley Unit 2 FSAR analysis was performed to evaluate the radiological consequences of an SGTR accident. The major factors that affect'the resultant offsite doses are'the amount of fuel defects (level of reactor coolant contamination), the primary to secondary mass transfer through the ruptured. tube, and the steam released from the ruptured steam generater to the atmosphere.

An m luation has been performed to determine, with respect to these mal,or factors, the.

effect of the Farley Unit 2 Steam Generator Tube Plugging (SGTP) increase and 2

Thermal Design Flow-(TOF) reduction program.

The SGTP increase and T0F reduction program at Farley Un'it 2 includes an increase in the tube plugging level to an average of 15% with a peak in any generator of 20%, and a reduction in thermal design flow of approximately 1.5%. A sensitivity study was performed to assess the impact of this program on the Farley Unit 2 FSAR SGTR analysis.

Since the conservative fuel failure assumption of 1% defective fuel will not change due to the SGTP increase and T0F reduction program, the sensitivity' study determined changes to the primary to secondary break flow through the ruptured tube and the steam released to the atmosphere via the ruptured steam generator. -To perform this sensitivity study, the Farley Unit 2 FSAR analysis was reevaluated in order to provide an accurate basis for performing a sensitivity study. -A sensitivity analysis was then completed to determine the-impact of the SGTP increase and TOF reduction program.

The results of the SGTR analyses' indicate that the primary to secondary break flow and atmospheric steam release via the ruptured steam generator increased as compared to the results of tne Farley Unit 2 FSAR SGTR analysis.

The increases are due to the T0F reduction and steam. generator tube plugging increase, and the reevaluation of the FSAR events using Farley licensing-basis methodology and current inputs.

j These increased mass releases were subsequently utilized in a radiological analysis to determine the effect of the SGTP increase and T0F reduction 4-15 cam io.enm i

j-

WESTINGHOUSE PROPRIETARY CLASS 3 program on the offsite doses. The assumptions used for th'e radiological analysis are consistent with those used in the FSAR analysis, with.the exception of the initial iodine activity in the secondary coolant.

The Technical Specification limit of 0.1 uCi/gm of dose equivalent I-131 was used. The results of the radiological analysis indicate that the site boundary thyroid and whole-body gamma doses are 3.3 and 0.14 REM respectively. Additionally, the low population zone thyroid and whole-body gamma doses are 1.4 and 0.05 REM respectively.

Althcugh these.results show a slight increase in the offsite doses over those presented in the FSAR, the dose increases are small and the total dose is well within the~NRC definition of a "small-fraction" of the 10 CFR 100 exposure guidelines.

4.

1.5 CONCLUSION

The effects rf increased tube plugging and reduced Thermal Design Flow have been evalua.ted or reanalyzed for Small Break LOCA, large Break LOCA, LOCA forces, long-term cooling, potential boron precipitation, LOCA mass energy releases, SGTR, and all licensing basis conclusions remain valid.

Therefore, safe operation with increased tube plugging and reduced T0F is assured.

l l-xm unm 4 16 l

\\

WESTINGHOUSE. PROPRIETARY CLASS.'31 s

TABL'E 4,1-1 Input Parameters.Used in the LOCA' Analyses Parameter Large Break s

Reactor' Core Design Thermal Power * ( Nt).

2705*

~

Peak Linear Power (kw/ft)

-12.314 at 6.0 ft Total Peaking Factor (FQT) 2.32 Hot Channel' Peaking Fr.ctor (FAH)

-1.55; Power Shape Chopped Cosine, Fz = 1.4967-Fuel Assembly Array 17 X'17.STD Nominal Cold-Leg Act:umulator:

1025 L

' Water Volume (ft3/ accumulator)

N: :inal Cold Leg Accumulatorf 1450

.nk Volume (ft3/ accumulator)

Hinimum Cold Leg Accumulator 600 3

Gas Pressure (psia)

Steam Generator Initial Pressure (psia) 751.78 i

Steam Generator Tube

- Pitgging Level (%)

20 l

Initial Flow-In Each Loop (lb/sec) 9216.1 Vessel Inlet Temperature (*F) 541.23**

l Vessel Outlet Temperature (*F) 610.97**

b Reactor Coolant Pressure (psia) 2280.0 i

2

• Two percent has been added to account for calorimetric error.

• • Temperatures consistent with power level of 2705 Nt.

co u inen m 4 17

WESTINGHOUSE PROPRIETARY CLASS'3-TABLE 4.1-2 large Break LOCA Results Fuel Cladding Data Minimum Safeguards i

4 Unit 2-CD = 0.4 RESULTS DECLG Peak Clad lemperature '(*F) 2123.0*

i Peak Clad Temperature Locaticn (ft)

', ' S Local Zr/H2O Reaction-(maximum %)

7.23 Local Zr/H2O Reaction Location for maximum 5.= 76 reaction (ft)

Total Zr/H2O Reaction, (%)

<0. 3 '

Hot Rod Burst Location, (ft) 6.0

• Includes:

1) a 4*F increase due to containment mini-purge isolation

2) a 50*F increase due.to SG reduced flow area.

ocu anm 4-18

WESTINGHOUSE PROPRIETARY CLASS 3

.l Table 4.1-3 Large Break LOCA Time Sequence of Events Minimum Safeguards CD = 0.,4 DECLG Time-(sec)-

Break 0.0 Reactor Trip Signal 0.505 SI-Signal 0.950 Intact Loop Accumulator' Injection 114.20 Pump Injection 27.950L End of Bypass 30.206 t

i End of Blowdown 30.206 i

BOC Time 43.224 Intact Loop Accumulator Empty 51.485 Hot Rod Burst Time 40.50 peak Clad Temperature (PCT) Time 235.32 l

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,nm,m:xm

WESTINGHOUSE PROPRIETARY CLASS 3 4.2 NON-LOCA EVALUATION 4.

2.1 INTRODUCTION

AND PURPOSE The current non-LOCA analysis bases for Farley Unit 2 assume a total RCS thermal design.flowrate of 265,500 gpm (88500 per loop).

The Farley.

Technical Specifications allow up to 10% steam generator tube plugging. Tube plugging increases the system hydraulic resistance of the RCS and may result in lower loop flowrates'and greater reactor vessel temperature differentials (Thot-Tcold).

This evaluation supports an increase in the steam generator tube plugging level.to a maximum of 20% in one steam generator with an average plugging R

level-of 15%.

This evaluation also considers a T0F of 261,600 gpm (87,200 gpm/ loop).

A reduction in Thermal Design Flow has an adverse effect on the core thermal limits (consequently the Overtemperature and-Overpower AT analysis setpoint equations) and the initial conditions assumed for all of the non-LOCA transients. The core thermal limits were reviewed and revised to account for the reduced flow. The current OTAT and OPAT analysis setpoint equations were revised to be more restrictive.

Because the analysis setpoint equations are more restrictive, a transient which approaches a trip condition will. produce an earlier reactor trip. Therefore, analyses which use_the Overtemperature and Overpower AT reactor trip functions-were not revised because of-the setpoint change.

l All non-LOCA transients were examined to determine-the impact of the reduced TDF. Generic DNB margin was allocated to cover the DNB penalty associated with the reduced flow. Allocation of this margin ensures that the DNB design basis continues to be met with the lower flow.

Because the T0F reduction is limited to approximately 1.5%, existing flow sensitivities were used to demonstrate that non-DNB safety criteria (e.g. peak clad temperature, RCS pressure) will also continue to be met.

eem unm 4-26 l

WESTINGHOUSE PROPRIETARY CLASS 3 The non-LOCA transients which were analyzed explicitly to account for the reduced T0F are Major Rupture of a Main Feedwater Pipe and' Uncontrolled Bank j

Withdrawal From Subcritical. The aralyses for these events currently in the Farley FSAR use a previous methodology and computer codes.

These events were reanalyzed using current (and NRC accepted) Westinghouse methodology and j

computer codes.

l Steam generator tube plugging asymmetries lead to flow asymmetries among the reactor coolant loops. The loop with the largest amount of tube plugging will have the lowest reactor coolant flow.

Because of the mixing in the reactor vessel lower plenum, temperature asymmetries resulting from the flow asymmetries are minimized. The greatest concern which must be addressed is the effect on transients which are sensitive to flow asymmetries.

The following events were analyzed to account for the loop flow differences and a reduced TOF:

Partial Loss of Forced Reactor Coolant flow Single Reactor Coolant Pump Locked Rotor The following discussion addresses the impact cn the non-LOCA safety analyses-for both 1) an approximate 1.5% reduction in thermal design flow and 2) an-average tube plugging level of 15% with a peak tube plugging level of 20% in one steam generator.

Each non-LOCA licensing basis event is discussed in the order which a description of it appears in the FSAR.

l 4.2.2 Uncontrolled Rod Cluster Control Assembly (RCCA) Bank Withdrawal from a Subcritical Condition (FSAR 15.2.1)

For this Condition II event, rod withdrawal results in a rapid reactivity insertion and increase in core power potentially leading to high local fuel temperatures and heat fluxes and a reduction in the minimum DNBR.

The power excursion is terminated by Doppler feedback and then the transient is promptly terminated by a reactor trip on the Power Range High Neutron Flux ~-

low setpoint.

Due to the inherent thermal lag in the fuel pellet, heat transfer to the RCS is relatively slow.

com mum 4-27

WESTINGHOUSE PROPRIETARY CLASS 3 In order to assess the affect of the reduced TOF, the transient was analyzed with the current Westinghouse methodology and computer codes.

The updated analysis used the TWINKLE (Ref. 4.16) neutron kinetics code to calculate the reactivity transient.

The FACTRAN (Ref. 4.17) code was used to calculate the thermal heat flux transient based on the nuclear power transient calculated by TWINKLE.

FACTRAN also calculates the fuel, clad, and coolant temperatures.

Key assumptions used in this analysis include:

1)

Two reactor coolant pumps are operating.

2)

The maximum steam generator tube plugging level as a plant average is 15% with a peak in one steam generator of 20%.

A thermal design-flow of 87200 gpm/ loop was used.

3)

.A maximum reactivity insertion rate of 72 pcm/sec was assumed.

This bounds the simultaneous withdrawal of the combination of the two control banks having the greatest combined worth at the maximum withdrawal speed.

In the event of a rod cluster control assembly withdrawal accident from the subcritical condition, the core and the Reactor Coolant System are not adversely affected since the combination of thermal power and coolant temperature result in a ONBR above the limiting value. The limiting DNBR was calculated by a detailed thermal hydraulics code (THINC). The increased average steam generator tube plugging and the resultant reduction in T0F are acceptable with respect to the RCCA Bank Withdrawal from Subcritical event.

See Figures 4.2-1 through 4.2-3 for the transient results.

4.2.3 Uncontrolled Rod Cluster Control Assembly (RCCA) Bank Withdrawal at Power (FSArt Section 15.2.2)

For this Condition 11 event, three initial power levels (i.e., 100%, 60%, and 10%) and a range of reactivity insertion rates assuming two (minimum and maximum) reactivity feedback conditions are analyzed. The resulting power excursion produced by a RCCA withdrawal at power results in high local fuel temperatures and an increase in the core heat flux.

Since the heat cannomme 4 28 s

WESTINGHOUSE' PROPRIETARY CLASS 3 extraction capability of the steam generator lags behind the core power generation, a net increase in the moderator temperature occurs..The resulting power mismatch and increase in reactor coolant temperature can result in DNB unless the transient is terminated by either manual or automatic means. Automatic reactor protection is provided via the High Neutron Flux and the Overtemperature AT trip functions.

The licensing-basis j

analysis presented in Reference 4-18 ensures that fuel damage will not occur

]

by demonstrating that the minimum DNBR remains above the safety analysis l

limit value.

Reductions in the RCS thermal design flow serve to reduce the' margin to the DNB licensing-basis limit. Therefore, generic DNBR margin was allocated to offset the effect of the reduced TOF.

The Overtemperature AT reactor trip analysis setpoint equation was impacted by the change in TDF. As previously mentioned, this reactor trip function is used in this analysis.

Because the revised analysis setpoint equation is more restrictive, the transient conditions.would cause an earlier reactor trip. Allocation of genaric DNB' margin precluded changes to the DNB-segment of the core thermal limits. Therefore, an earlier reactor trip results in greater margin to the core limits and thus DNB cor.ditions. Consequently, it was not necessary to explicitly analyze the RCCA Bank. Withdrawal at Power event with the revised OTAT analysis setpoints.

l The asymmetric steam generator tube plugging levels discussed in the l

introduction will result in flow and inlet temperature asymmetries between i

the RCS loops. However, these asymmetries will be attenuated by coolant mixing in the reactor vessel lower plenum resulting in a negligible effect on the transient, i

l

'1 1

l With the allocation of generic DNBR margin and the conservative change in the l

OTAT reactor trip setpoint, the conclusion of this event as presented in the FSAR (Ref. 4.18) remains valid.

l ocu inenm 4-29

WEST!NGHOUSE PROPRIETARY CLASS 3 4.2.4 Rod Cluster Control Assembly Misalignment (FSAR Section 15.2.3)

This Condition II event is analyzed to demonstrate that following various RCCA misoperation events, such as dropped red (s)/ bank or statically misaligned rods, the minimum DNBR remains above the safety analysis limit value. The reduction in RCS flow potentially impacts the RCCA misoperation events by changing the initial condition assumptions used in this analysis.

Reductions in the RCS flow and/or incre ues in the RCS temperature serve to reduce the margin to the DNB licensing-basis-limit. Therefore, generic DNBR margin was allocated to offset the effect of the reduced. thermal design flow.

The asymmetric steam generator tube plugging levels discussed in the introduction will result in flow and inlet temperature asymmetries between the RCS loops. Attenuation of these asymmetries will occur, however, as a result of coolant mixing in the reactor vessel lower plenum.

Furthermore, the nominal RCS average temperature will continue to be controlled to the same value which was previously assumed in the Reference 4.18 analysis.

The allocation of the generic DNBR margin ensures that the DNB licensing-besis criteria will continue to be met and the conclusions from Reference 4.18 for this event remain valid.

4.2.5 Uncontrolled Boron Dilution (FSAR Section 15.2.4) l This Condition II event is analyzed for i,hree modes of plant operation. The analysis demonstrates that sufficient shutdown margin exists, such that should a dilution event occur, there is sufficient time following the start of dilution to allow operator detection and termination of the event prior to a complete loss of shutdown margin. This event is analyzed for operating modes 1, 2, and 6.

l An input to the boron dilution analysis for Modes 1 and 2 is the RCS active l

volume, i.e., the total RCS volume minus the volumes of the pressurizer, the l

pressurizer surge line, the dead volume of the reactor vessel head, and plugged steam generator tubes.

Reduction of the RCS active volume is l

directly proportional to the reduction in operator response time for the boron dilution event described in the Farley FSAR.

It has been estimated l

4-30 xm iwv*

4 e

-e

~

~,

e

-n-e

WESTINGHOUSE PROPRIETARY CLASS 3 that a tube plugging level of 10% would reduce the the active RCS volume by 4%. The increase in tube plugging to an average of 15% will reduce the volume even further.

The current Farley analyses for the Modes 1 and 2 Boron Dilution could accommodate mota than a 50% reduction in RCS volume without violating the acceptance critr.ria.

Therefore, the increased tut,' plugging levels (and subsequent reduction in thermal design flow) will not impact the conclusions that the operator has-more than 15 minutes between initiation of event and loss of shutdown margin.

Asymmetric tube plugging would need to be addressed if the boron dilution analysis considered modes of operation in which less than 3 reactor coolant pumps were assumed and the average tube plugging could be greater than 15%.

However, the active volume assumptions for Modes 1 and 2 will temain unaltered as a result of asymmetric steam generator tube plugging since all RCS loops are in operation and since the average steam generator tube plugging level remains at 15%. Mode 6 operation also is unaffected since,the active volume assumed comes from the reactor vessel and one RHR loop (i.e.,

the active volume of the steam generator is not assumed in Mode 6).

As a result, the conclusions presented in the FSAR for the boron dilution event in Modes 1, 2, and 6 remain unaltered with the reduced RCS flow and the asymmetric tube plugging levels.

4.2.6 Partial Loss of Forced Reactor Coolant Flow (FSAR Section 15.2.5) l.

The Partial Loss Of forced reactor coolant Flow (PLOF) transient is a l

Condition II event which is analyzed under full power conditions assuming l

that 1 of 3 operating reactor coolant pumps coasts down. The reactor is promptly tripped on low reactor coolant loop flow. The current FSAR analysis demonstrates that the minimum DNBR remains above the safety analysis limit value. The case with 3 loops operating prior to 1 loop coastdown was analyzed to incorporate the RCS flow reduction as well as the asymmetric tube plugging scenario.

(It should be noted that a second PLOF case is presented in the FSAR which addresses N-1 loop operation. As Farley is not licensed to.

com monm 4 31

\\

L WESTINGHOUSE PROPRIETARY CLASS 3 operate in an N-1 loop condition, this case is not included in this

. evaluation.)

The reduced RCS thermal design flow will have a neg3tive impact on the L

calculated DNBR because it is a critical parameter in DNBR determination, l

l s

(-

Asymmetric steam generator tube plugging may adversely impact the PLOF

{

results from Reference 4.18 since the reactor coolant pump could be lost in L

the loop with the lowest level of tube plugging (and thus highest flow).

In this case, an additional reduction in RCS flow occurs since forced flow is maintained'through the two loops having higher steam generator tube plugging I

l levels. lne reduction in RCS thermal design flow may also produce.an l

increase in.the RCS moderator temperature, both of which tend to reduce thef margin to the licensing-basis DNB limit for this event.

l t

L The event was analyzed to determine the effects of the reduced TDF and l

incorporate the asymmetric scenario using the LOFTRAN (Ref. 4.19) and FACTRAN j

(Ref. 4.17) computer codes.

The THINC code was used to calculate the minimum l

DNBR during the transient.

Some of the key assumptions used in this analysis l

include:

a.

Three reactor coolant loops were operating prior to the loss of flow in one loop.

i b.

To account for asymmetric tube pluggiis effects, the loss of flow occurred in the loop where the reactor coolant flow rate was 105% of the average loop flowrate.

l c.

The initial power level was 102% of-the rated thermal power.

j d.

The system pressure was assumed to be 60 psi lower.than the nominal RCS pressure.

1 e.

The low flow reactor trip setpoint was 85% of nominal.

The results of this analysis show that the minimum DNBR is still bounded by the complete loss of forced reactor coolant flow analysis. As a result, the i

. com iocnno 4 32

o I

WESTINGHOUSE PROPRIETARY CLASS 3 j

increased tube plugging with reduced thermal design flow as well as the l

~

asymmetrical steam generator tube plugging levels do not alter the conclusions presented in the FSAR for the PLOF event. Figure 4.2-4 through 4.2-6 depict the transient results for the partial loss of flow event.

'I 4.2.7 Startup of an Inactive Reactor Coolant Loop (FSAR Section 15.2.6)

This Condition II event is analyzed assuming a maximum initial power level consistent with 2 loop operation.

The startup of an inactive loop results in i

a reactivity insertion since the inactive-loop fluid being injected into the i

core is at a lower temperature relative to the remaining fluid in the core.

4 The analysis demonstrates that the minimum DNBR remains above the safety analysis limit value.

The automatic reactor protection system terminates this transient on power range High Neutron Flux.- The licensing-basis.

analysis presented in Reference 4.18 demonstrated that the minimum DNBR was

)

greater than the safety analysis limit value.

The reduction in RCS thermal design flow does not affect the conclusion that the DNBR limit is met because generic margin has been allocated to of.fset any penalty associated with the reduced flow.

However, asymmetric steam generator tube plugging potentially impacts these results since the mass of cold water from the inactive-loop can hypothetically come from the loop with the lowest steam generator tube plugging level.

In this situation, a larger mass of cold fluid at a higher rate enters the core and produces a larger reactivity insertion (due to the negative moderator temperature coefficient) relative to the case where symmetric steam generator tube plugging was assumed. However, with the reduction in TDF, the mass and flow rate of the inactive loop, as well as the active loop, are actually smaller than those used in the FSAR.

For conservatism, this evaluation assumes that the inactive loop undergoing startup contains the lowest percentage of tube plugging. This will produce the largest insurge of cold water into the core and hence the largest positive reactivity excursion.

Initial RCS thermal design flow from the two active loops would be reduced somewhat, in comparison to the Reference 4.18 com eenm 4-33 I

WESTINGHOUSE PROPRIETARY CLASS 3 analysis, since this flow woulo be based on the two-loops containing the highest levels of steam generator tube plugging.

During the transient, the reactor coolant ~ flow increases due to the-startup of the inactive pump, thereby producing a ONB. benefit.

However, a ONB penalty may be introduced as a result of a larger positive reactivity insertion in the core produced by this increased flow.

Also note that since the reactivity feedback will be larger, the power will also reach the high neutron flux trip setpoint more quickly relative to the analysis documented in Reference 4.18.

Since the increases in thermal-design flow and power are competing DNB effects, the changes in minimum DNBR due to the asymmetric tube plugging scenario are estimated to be small. The results from the analysis documented in Reference 4.18 indicate that sufficient margin exists to the safety analysis DNB limit (i.e., to accommodate minor perturbations in the minimumDNBR).

Therefore, since the asymmetric tube elugging will have an insignificant impact on the transient and the RCS ti,ecmal design flow decrease is I

accommodated by the allocation of generic DNB margin, the conclusions in-l Reference 4.18 will remain valid for this event.

l 4.2.8 Loss of External Electrical Load and/or Turbine Trip (FSAR Section 15.2.7) l The analysis presented in the FSAR represents a complete loss of steam load l

from full power without a direct reactor trip.

Four cases are analyzed which are based on two different primary side pressure control strategies (automatic versus none) and two sets of core physics characteristics (maximum tersus minimum reactivity feedback). The analysis demonstrates that, with the power mismatch between the core and turbine, the primary and secondary system pressures remain below 110% o: the design values and the minimum DNBR remains above the safety analysis limit'value. Automatic reactor trip signals which may be generated during this event include high pressurizer pressure and overtemperature AT.

The reduction in RCS flow can potentially impact the results of this analysis with respect to the minimum DNBR com io crim 4 34

1 WESTINGHOUSE PROPRIETARY CLASS 3 I

calculated. The allocation of generic DNBR margin ensures that the DNB safety analysis limit continues to be met.

In all four case', there is substantial margin to the primary / secondary

• te pressure limits..This transient is insensitive, with respect to the prc

.re limits, to a change in reactor coolant flow of the magnitude specified in this evaluation.

As discussed for the Rod Withdrawal at Power event, the Overtemperature AT reactor trip analysis setpoint equation was impacted by the change in theini design flow (see Setpoint Calculation Impact). As previously mentios.cd, this reactor trip function is used in this analysis.

Because the revised setpoint equation is more restrictive, the transient conditions.would cause an earlier reactor trip. Allocation of generic DNB margin precluded changes to the DNB

~

segment of the core thermal limits. Therefore, an earlier reactor trip results in greater margin to the core limits and thus, DNB conditions.

Consequently, it was not necessary to explicitly analyze the Loss of Load / Turbine Trip event with the revised OTAT. coefficients.

l Thus, an approximate 1.5% reduction in RCS flow would not result in the violation of the licensing basis criteria following a loss of load event.

The asymmetric steam generator tube plugging. levels' discussed'in the introduction will result in flow and inlet temperature asymmetries between-i the RCS loops. These asymmetries will be attenuated, however, by coc. ant mixing in the reactor vessel lower plenJm. Also note that operability of the Overtemperature AT reactor trip function in this event is still assured under asymmetric flow and temperature conditions since each channel (ichannel/ loop) is calibrated for the specific loop inlet and outlet conditions.

4.2.9 Loss of Normal Feedwater (FSAR Section 15.2.8) l The loss of normal feedwater analysis in Section 15.2.8 of the FSAR presents the consequences of a complete loss of normal feedwater flow simultaneous to all three steam generators.

The loss of AC power event is similar except l

00)13100 F2790 4 35

WESTINGHOUSE PROPRIETARY CLASS 3 i

i that the loss of offsite power also results in all three reactor coolant pumps (RCPs)coastingdown.

These transients are analyzed to demonstrate that neither the primary nor secondary sides are overpressurized, that the core is not adversely affected, and the pressurizer does not fill.

The Loss of Normal Feedwater event is sensitive to initial steam generator mass as well as the.. lass in the steam generators at the time of reactor trip.

An increased tube plugging level causes a slight decrease (< 0.5%) in the initial steam generator mass used in the Loss of Normal Feedwater event.

Following the loss of normal feedwater, the reactor continues to operate until, due to the rapid loss of steam generator inventory and the continued heat transfer to the secondary side, it is tripped on a low-low steam generator level signal.

The increased tube plugging level has less than a 0.5% decrease in initial mass and mass at time of reactor trip.

The effect of reducing the RCS flow would be an increase in the heatup of the RCS during the initial phase of the transient.

The increased heatup results in a decrease in the coolant density which in turn would increase the pressurizer insurge during this heatup.

During the lorg-term portion of the transient, the 5,. 3 RCS temperature (and resultant peak pressurizer water volume) is reached when the heat removal capability of the auxiliary feedwater system matches the core decay heat generation.

If the assumed RCS flow reduction is due to higher than anticipated loop flow resistances, the natural circulation flow will be reduced by an amount proportional to the approximate 1.5% thermal design flow reduction. This slight reduction in natural circulation flow at the peak RCS temperature condition would not significantly impair the heat transfer across the steam generator tubes, thus, resulting in a similar hot leg temperature and peak pressurizer water volume.

The FSAR Loss of Normal Feedwater analysis has etnough margin to accomodate at least a 5% reduction in RCS flow and a 50% reduction in steam generator tube heat transfer area due to tube l

plugging. Therefore, the approximate 1.5% reduction in RCS flow and an

(

average steam generator tube plugging level of 15% does not cause the safety l

analysis acceptance criteria to be exceeded.

It should be noted that the l

l coiss ioon,,e 4 36

WESTINGHOUSE PROPRIETARY CLASS 3 decay heat model used in the loss of normal feedwater analysis is based on the ANS-1971 decay heat model.

Significant margin would be gained if the ANS-1979 model was used in this analysis since the total energy released into the RCS is lower.

The asymmetric steati: generator tube plugging scenaric will result in flow and inlet temperature asymmetries between the RCS loops.

However, these asymmetries will be attenuated by coolant mixing in the reactor vessel lower plenum.

Thus, the primary and secodary side peak pressure licensing-basis design criteria will continue to be met, the pressurizer will not go solid and the conclusions made in Reference 4.18 for the loss of normal feedwater event remain valid.

4.2.10 Loss of Offsite Power to the Station Auxiliaries (St&tica Blackout)

(FSARSection15.2.9)

The analysis prasented in the FSAR represents a complete loss of power to the plant auxiliaries (i.e., the reactor coolant pumps, t.ondensate pumps, etc.)

from full power.

The loss of power results in a heatup and pressurization of the primary and secondary systems. The analysis demonstrates that adequate auxiliary feedwater flov is delivered to the steam generators to remove decay heat such that DNB dili not occur, overpressurization of the primary and secondary systems will not occur, and the pressurizer will not become water solid.

Steam generator tube plugging potentially impacts this ANS Condition 11 event by reducing the RCS thermal design flow.

Reductions in the RCS thermal design flow decreas the minimum DNBa, increase the peak RCS pressure and potentially lead to increased coolant expansion and a reduction in the margin to pressurizer filling.

When offsite power is lost, the reactor coolant pumps coastdown and the RCS therral design flow will eventually reduce to natural circulation flow.

With asymmetric steam generator tube plugging levels, natural circulation flow rates will be slightly different be. tween the loops; however, the change in flow resistance is expected to be proportional under full and natural oem ioom o 4 37 M

WESTINGHOUSE PROPRIETARY CLASS 3 circulation flow. The dominent driving force for natural cirrulation is the density difference between the fluid in the reactor vessel downcomer and the fluid within the core barrel (in the core and upper core plenum). This driving force will act to force ficw thre.agh all of the reactor coolant loops. The reductio's in TDF has an insignificant impact on this interaction.

With asymmetric steam generator tube plugging levels, natural circulation flow rates will be slightly different between the loops; however, the change in flow resistance is expected to be proportional under full and natural circulation flow conditions.

Inlet temperature asymmetries may also exist between the RCS loops; however, coolant mixing in the reactor vessel lower plenum will minimize this effect.

The results of the complete loss of forced reactor coolant flow analysis (FSAR section 15.3.4) and the loss of normal feedwatar analysis (FSAR section 15.2.8) continue to show that for a loss of all non-emergency AC power, no adverse conditions occur in the reactor core.

As a result. the DNBR remains above the safety analysis limit value and the primary and sr idary side peak pressure licensing-basis design criteria continues '

be met.

Pressurizer filling also will not occur; therefore, the conclusions for the station blackout event which are documented in Reference 4.18 remain valid.

4.2.11 Excessive Heat Ren.Jval Due to Feedwater System Malfunctions (FSAR15.2.10)

Two cases are analyzed and described for this ANS Condition 11 event in the FSAR. A full power case is used to determine the plant response to a 184%

step increase in the feedwater flow to one steam generator; a zero power case examines a step increase in feedwater flow from zero to nominal full-load flow in one steam generator.

For the full power case, the minimum DNBR is l

shown to remain above the safety analysis limit value. The zero power case demonstrates that the reactivity transient, and hence the minimum DNBR, is bounded by the rod withdrawal from subcritical event. The reduction in RCS flow would impact the results of this analysis however, the allocation of generic DNBR margin ensures that the DNB design basis as presented in the FSAR continues to be met.

l 4 38 mu ocm :

WESTINGHOUSE PROPRIETARY CLASS 3 Asymmetric steam generator tube plugging levels (as discussed in the introduction) will result in flow and inlet temperature asymmetries between the RCS loops.

However, these asymmetries will be attenuated by coolant mixing in the reactor vessel lower plenum.

The reactivit.Y insertion rate of the feedwater malfunction event, driven by the RCS cooldown, will decrease if the thermal design flow decreases and the steam generator tube plugging increases because the primary to secondary heat transfer capacity will decrease. However, as previously discussed, the change in her transfer capability is small and these effects are minimal.

Therefore..he current calculated reactivity insertion rate remains valid.

The reactivity insertion rate calculated for the zero power case is bounded by the RCCA withdrawal from suberitical analysis discussed previously.

Therefore, She conclusions in Reference 4.18, pertaining to the excessive ftedwater flow event, remain valid.

4.2.12 hesssiveLoadIncreaseIncident(FSARSection15.2.11) d The analysis presented in the FSAR describes the plant response to a 10% step increase in load from nominal full power conditions.

Four cases are analyzed for this ANS Condition II event based on automatic versus manual rod control and minimum versus maximum reactivity feedback parameters.

Reactor protection against an excessive load increase transient is provided by the l

power range high neutron flux, overpower AT, and overtemperature AT reactor protection system signals. However, since the plant is capable of removing sufficient heat from the core while reaching a new equilibrium condition (at a higher power level cor-esponding tc the increase in steam flow) for each case which was analyzed, no reactor protectio 5 system trip signals were generated.

In addition, each case showed that the minimum DNBR remained above the safety analysis limit value. The minimum DNBR would be impacted by the RCS flow reduction however, allocation of generic DNBR margin ensures that the DNB design basis limit continues to be met.

Asymetric steam generator tube plugging levels cause flow and inlet temperature asymmetries between the RCS loops.

Coolant mixing in the reactor mu umu 4 39

WESTINGHOUSE PROPRIETARY CLASS 3 vessel lower plenum, however, will act to reduce these affects.

Thus, the DNBR will remain above the safety analysis limit value and the conclusions in Reference 4.18 pertaining to the excessive load increase event, remain valid.

4.2.13 Accidental Depressurization of the Reactor Coolant System (FSAR15.2.12)

For this ANS Condition 11 event, the transient is initiated by the opening of a single pressurizer relief or safety valve while the reactor is at full power.

Initially, the RCS pressure drops rapidly until a reactor trip occurs on either the pressurizer low pressure or Overtemperature AT reactor protection signals.

At this time, the pressure decrease continues, but at a much slower rate. The analysis demonstrates that the minimum DNBR remains above the safety ana1ysis limit value.

The RCS flow reduction will impact the minimum DNBR however, generic DNBR margin nas been allocated to ensure that the DNB design basis safety limit continues te be met, i

The asymmetric steam generator tube plugging levels previously discussed in the introduction will produce flow and inlet temperature asymmetries between the RCS loops. These asymmetries, however, will be m D 512ed by coolant mixing in the reactor vessel lower plenum. Thus, the DNBR remains above the safety analysis limit value, and the conclusions in Reference 4.18, pertaining to this event, remain valid.

4.2.14 Accidental Depressurization of the Main Steam System (FSAR 15.2.13) l This ANS Condition II event is initiated by the full opening of a single steam dump, relief, or safety valve from zero power conditions.

The analysis confirms that the minimum DNBR remains above the safety analysis limit value.

Steam generator tube plugging potentially impacts the Main Steam System depressurization event by reducing the RCS thermal design flow assumed in the analysis. Reductions in the RCS thermal design flow potentially decrease the minimum DNBR calculated during the event. However, reduced flowr m s result onu io em,o 4-40

. ~......

i WESTINGHOUSE PROPRIETARY CLASS 3 in less primary to secondary heat transfer and consequently, less of a power increase.

Generic DNBR margin has been allocated to ensure that the DNB design basis continues to be met with the increased tube plugging and subsequent reduction l

in thermal design flow.

The asymmetric steam generator tube plugging levels discussed in the introduction produce flow and temperature asymmetries betwtun the RCS loops.

This event is analyzed at no-load conditions where the temperature j

asymmetries would be minor. The flow asymetries will be attenuated by l

coolant mi ing in the reactor vessel lower plenum.

The DNBR remains above the safety analysis limit value and the conclusions presented in

]

Reference 4.18 for the Main Steam Supply depressurization event, remain 1

valid.

4.2.15 Inadvertent Operation of ECCS During Power Operation (FSAR 15.2.14) l A spuriou. Safety injection System (SIS) signal is an ANS Condition Il event which is assumed to be initiated at full power. The injection of highly concentrated (2000 ppm) borated water into the RCS reduces core power, temperature and pressure until the reactor trips on low pressurizer pressure.

The RCS power and temperature reductions produce a similar reduction in j

pressure on the secondary side of the plant. The analysis demonstrates that the minimum DNBR remains above the safety analysis limit value.

Steam generator tube plugging potentially impacts the spurious SIS actuation event by reducing the RCS thermal design flow. Reductions in the RCS thermal design flow potentially decrease the minimum DNBR calculated during the event.

Asymmetric steam generator tube plugging levels will create flow and inlet temperature asymmetries betwcen the RCS loops.

The affect of these asymmetries, however, will be minimal due to coolant mixing in the reactor vessel lower plenum. Additionally, the plant average steam generator plugging level will not exceed 15% and the nominal RCS average temperature will continue to be controlled to the same value previously assumed in the muicomu 4 41

f WESTINGHOUSE PROPRIETARY CLASS 3 Reference 4.18 analysis. The results from Reference 4.18 show that the DNBR is never less than the initial value.

Furthermore, the effects of asymmetric tube plugging will not impact the behavior of this event. Therefore, the conclusions presented in the FSAR remain valid.

4.2.16 Complete Loss of Forced Reactor Coolant Flow (FSAR 15.3.4)

This Condition 111 event is analyztd under full power conditions assuming that 3-out-of-3 operating reactor coolant pumps coastdown. The reactor is assumed to trip on an undervoltage signal. The analysis demonstrates that the minimum DNBR remains above the safety analysis limit value.

Steam generator tube plugging potentially affects the complete loss of flow event by decreasing the RCS flow rate assumed in the analysis. A decrea:e in the RCS flow rate potentially decreases the minimum DNBR calculated during the event. While the reduced TDF is a DNB penalty, thr resulting slower flow coastdown would a a small offsetting benefit.

The flow coastdown of an underfrequency event is not impacted by the thermal design flow change because the event is driven by the RCP response to the frequericy decay.

The asymmetric steam generator tube plugging levels discussed in the introduction cause flow and inlet temperature asymmetries between the RCS loops.

RCS inlet temperature asymmetries will be attenuated by coolant mixing in the reactor vessel lower plenum and the nominal RCS average l

temperature will be controlled to the same value which was previously assumed 1

I in Reference 4.18. With asymmetric loop flow, the core flow basically remains unchanged relative to a uniform plugging case. Generic DNBR margin has been allocated to account for the reduced RCS flow without considering any benefit from the slower coastdown rate.

Thus, the ONBR remains above the safety analysis limit value and the conclusions from Reference 4.18 also remain valid.

mummm 4 42

WESTINGHOUSE PROPRIETARY CLASS 3 4.2.17 Single Rod Clust'r Control Assembly (RCCA) Withdrawal at Full Power (FSAR15.3.6)

TwocasesareanalyzedandpresentedintheFSARforthisConditionIli event:

automatic and manual reactor control.

In both cases, an increase in core power, reactor coolant temperature and hot channel factor produce a reduction in the minimum DNBR.

The analysis demonstrates that, although it is not possible in all cases to ensure that DNB will not occur, an upper bound on the number of fuel rods experiencing DNB is less than or equal to 5%. Steam generator tube plugging potentially impacts the single RCCA withdrawal event by reducing the RCS flow which is assumed in the analysis.

A reduction in the RCS flow potentially decreases the minimum DNBR calculated during the event. Generic DNBR margin has been allocated to offset the effect of the reduced thermal design f Mw and ensure that less than 5% of the fuel rods experience DNB during this transient.

Asymmetric steam generator tube plugging levels will result in flow and inlet temperature asymmetries between the RCS loops.

Coolant mixing in the reactor vessel lower plenum, however, will minimize these affects.

Furthermore, the nominal RCS average temperature will continue to be controlled to the same value previously assumed in Reference 4.18. Therefore, the conclusions of the analysis discussed in Reference 4.18 remain valid for the increased tube plugging.

4.2.18 Rupture of a Main Steam Line (FSAR 15.4.2.1)

For this ANS Condition IV event, the transient is assumed to be initiated by the instantaneous double-ended rupture of a main steam line while at hot zero power conditions. Two cases (with and without offsite power) are' considered.

The analysis demonstrates that the minimum DNBR remains.above the licensing limit value in each case.

Steam generator tube plugging potentially impacts the main steam line break event by reducing the RCS thermal design flow assumed in the analysis.

Reductions in the RCS thermal design flow potentially decrease the minimum DNBR calculated for the. event. This DNB penalty would be partially offset because the lower flow would lead to reduced primary to secondary heat transfer and subsequently less of a power com+cmu 4-43

WESTINGHOUSE PROPRIETARY CLASS 3 1

increase. However, generic DNBR margin was allocated to ensure that the DNB design basis for this event continues to be met with the increased tube plugging.

The asymmetric steam generator tube plugging levels discussed in the introduction create flow asymmetries between the RCS loops.

These flow asymmetries, however, will be attenuated as the coolant is mixed in the reactor vessel lower plenum.

Since this analysis is initiated at hot zero 1

power, temperature asymmetries will be minimal.

Thus, the calculated minimuni DNBR will stay above the licensing limit value, and the conclusions reported in Reference 4.18 for this event remain valid.

4.2.19 Major Rupture of a Main Feedwater Pipe (FSAR Section 15.4.2.2)

For this ANS Condition IV event, the double-ended rupture of a main feedwater pipe initially results in a cooldown of the RCS due to the heat removal of the steam generator blowdown. This cooldown period is followed by a heatup as the high levels of decay heat and the lack of inventory on the secondary side result in inadequate heat transfer. The event is analyzed to show that adequate auxiliary feedwater flow exists to remove core decay heat and stored energy following a reactor trip from full power and that the core remains in a coolable geometry, and covered with water.

Steam generator tube plugging potentially impacts the main feedwater line break event by reducing the RCS j

thermal design flow, and therefore, the primary to secondary heat transfer capability, assumed in the analysis.

l l

The analysis currently in the FSAR was performed using the MARVEL digital computer code. To evaluate the effects of the reduced thermal design flow, the feedline break event was analyzed with an improved transient computer code and analytical methods.

The new analysis was completed using the LOFTRAN (Ref. 4.19) code. The same two auxiliary feedwater scenario cases presented in the FSAR for this event were analyzed. The safety analysis criterion for this event which is used with the improved modeling capability is that hot leg boiling does not occur in the reactor coolant system prior to the auxiliary feedwater heat removal capacity exceeding decay heat build up, com io cum 4 44

WESTINGHOUSE PROPRIETARY CLASS 3 Some of the key analysis assumptions which are common to both D M s follow:

a.

The thermal design fic.v is 87200 gpm.

b.

The plant is initially operating at 102 percent of the engineered safeguards design rating, i

1 c.

Loss of offsite power occurred at the time of reactor trip.

d.

Main feed to all steam generators is assumed to stop at the time the break occurs.

e.

Initial reactor coolant average temperature is 6 deg-F above the nominal value, and the initial pressurizer pressure is 60 psi above its nominal value.

f.

ANS-1979 standard residual heat generation (Reference 4.20) is assumed based upon long-term operation at the initial power level preceding the trip.

The first case modeled auxiliary feedwater operation 10 minutes after the reactor trip on low-low steam generator level at a capacity of 350 gpm to the intact stea.6 generators. The analysis showed that there was no bulk boiling l-prior to the auxiliary feedwater heat removal capacity exceeding the decay heat build up.

l The second case modeled auxiliary feedwater operation 1 minute after reactor trip on steam generator low-low level at a capacity of 150 gpm to the intact steam generators. This analysis demonstrated that the operator has at least 30 minutes to increase the flow to the intact steam generators to 350 gpm.

During the 30 minute interval, bulk bolling does not occur in the RCS and subsequently, the auxiliary feedwater system heat removal capacity is able to exceed the decay heat build up.

com wvw 4-45

I WEST!NGHOUSE PROPRIETARY CLASS 3 RCS thermal design flow and inlet temperature variations will be produced in the loops having asymmetric steam generator tube plugging levels however, coolant mixing in the reactor vessel lower plenum attenuates these conditions. Therefore, it was not necessary to consider asymmetric conditions in the previously discussed analysis.

Although the results of the analysis have changed, the conclusions presented in the FSAR remain valid for the new analysis.

The transient results are shown in Figures 4.2-7 and 4.2-8.

4.2.20 Single Reactor Coolant Pump Locked Rotor (FSAR Section 15.4.4)

This Condition IV event is analyzed unde" full power conditions assuming the instantaneous seizure of one RCP rotor using the LOFTRAN and FACTRAN computer codes.

This results in a rapid RCS flow reduction and pressure rise which may lead to DNB.

The reactor is promptly tripped on a low flow signal.

The analysis demonstrates that the maximum reactor coolant system pressure is less than 110% of design pressure, the maximum fuel clad temperature is less than 2700'F and the amount of zirconium-water reaction is small.

The lower RCS flow will result in slightly higher system pressures than those calculated in the current FSAR analysis.

The PCT analysis performed for the locked rotor event conservatively assumed tnat DNB occurs upon the initiation of the event. DNB significantly decreases fuel-to-clad heat transfer.

This assumption maximizes the calculated peak clad temperature and minimizes the impact of a flow reduction since fuel-to-clad heat transfer is already substantially degraded. Additionally, the asymmetric tube plugging scenario may impact the analysis because the locked rotor could occur in the loop with the lowest number of plugged tubes (thus the greatest fraction of RCS total flow).

Therefore, this event was analyzed to incorporate the lower RCS flow and the asymmetric tube plugging conditions.

Some of the key assumptions used in this analysis include:

Three reacter coolant loops were operating prior to the pump seizure in a.

one loop.

(The case with two loops operating prior to the pump seizure mu toemx 4 46

WESTINGHOUSE PROPRIETARY CLASS 3 was not analyzed because Farley is not licensed to operate in the N-1 loop configuration.)

l b.

To account for asymmetric tube plugging effects, the flow rate in the loop where the pump seizure occurred was 105% of the average loop flow rate.

c.

The initial power level was 102% of the rated thermal power.

d.

In the peak pressure analysis, the initial system pressure was assumed to be 60 psi greater than the nominal RCS pressure, and the initial temperature was 5.5'F greater than the nominal RCS average temperature.

e.

The low flow reactor trip setpoint was 85% of nominal.

The results of this analysis met the criteria stated above. The conclusions of the FSAR with respect to the locked rotor event are met for the increased steam generator tube plugging as well the asymmetry discussed previously.

Figures 4.2-9 through 4.2-12 depict the transient response

.J the locked rotor event.

4.2.21 Rupture of a Control Rod Drive Mechanism Housing (FSARSection15.4.6)

For this Condition IV event, a rapid reactivity insertion and increase in core power leads to high local fuel and clad temperatures and possible fuel l

and/or clad damage. The RCCA ejection analysis is analyzed at four conditions: beginning and end-of-life core physics characteristics, at hot f

zero power and full power. The analysis demonstrates that gross fuel damage will not occur, that the core will remain in a coolable geometry, and that the RCS will remain intact.

In order to demonstrate that these criteria are met Westinghouse applies the following, more restrictive, criteria:

1)

The average fuel pellet enthalpy at the hot spot is less than 200 cal /gm (360 Btu /lbm).

4 47 ocm mom,o l

WESTINGHOUSE PROPRIETARY CLASS 3 2)

Fuel melt at the 'ot spot is lim 4ed to less than the innermost 10% of the fuel pellet.

3)

Peak RCS pressure is less than that which would cause stresses to exceed the Faulted Condition Stress Limits.

The -od ejection event is characterized by a rapid power excursion terminated by Doppler feedback. The reactor is tripped on high neutron flux (low setting for the zero power cases, high setting for the full power cases).

A reduction in RCS flew will result in a reduction in the fuel rod-to-coolant heat transfer. This may result in an increase in the calculated fuel and clad temperatures as well as the fuel stored energy during an RCCA ejection.

As shown in the FSAR, the f.11 power cases result in the highest fuel pellet temperatures and are the most limiting with respect to criteria 1 and 2.

Examination of these cases reveals that, due to the rapid power and fuel temperature rise coup' d with the thermal lag in the fuel pellet itself, the time at which the maximum pellet enthalpy and fuel melt are calculated to occur is before any significant amount of heat has reached the coolant. A sensitivity analysis, which used methods consistent with WCAP-7588, Rev. 1 (Raf. 4.23), demonstrated that for a 2% reduction in thermal design flow, there was only a minor change to the maximum pellet enthalpy and fuel melt l

results for the full power rod ejection cases. There exists enough margin in the Farley analyses to absorb the differences in results from the sensitivity. Therefore, the apprcximate 1.5% reduction in thermal design flow caused by the increased steam generator tube plugging would not cause the safety criteria to be violated for.he full power cases.

RCS thermal design flow and inlet temperature variations will be produced in the loops having asymmetric steam generator tube plugging levels however, coolant l

mixing in the reactor vessel lower plenum attenuates these conditions.

The zero power rod ejection cases are characterized by a sharp increase in the clad average temperature. Asymetric considerations should be addressed for the hot zero power (HZP) rod ejection cases since these cases are performed using RCS thermal design flowrates which are based on two-out-of-three RCPs operating. Asymmetric steam generator tube plugging levels could oom um o 4 48

WESTINGHOUSE PROPRIETARY CLASS 3 adversely impact the corresponding Reference 4.18 results if the loop having the highest thermal design flow (i.e., the loop with the lowest steam generator tube plugging) is not assumed to be in operation.

RCS thermal design flow would be further reduced in this scenario since flow is being maintained in the two loops having more steam generator tube plugging.

However, the sensiti 'ty which addressed a 2% reduction in thermal design flow showed only a slight increase (< 1%) in the maximum clad average temperature. There exists enough margin in the Farley analyses to absorb the differences created from the 2% reduction in thermal design flow.

Therefore, the approximate 1.5% reduction in thermal design flow caused by the increased steam generator tube plugging as well as the possible additional reduction caused by the asymmetric tube plugging would not produce a significant increase in the Farley maximum PCT.

i The analysis of the peak pressure transient for the RCCA ejection event is discussed in WCAP-7588, Rev. 1 (Ref. 4.23).

A reduction in RCS flow could increase the primary side pressurization by reducing the primary-to-secondary side heat transfer.. However, due to the rapid nature of this event any secondary side heat removal will lag well behind the heat addition to the primary side. Thus, an approximate 1.5% flow reduction will have a minimal impact on the primary side peak pressure. However, in WCAP-7588, several cases are presented which calculate the peak RCS pressure. The most detailed of these cases calculates a peak pressurizer pressure of 2600 psia. This is more than sufficient margin to the Faulted Condition Stress Limits to accommodate an approximate 1.5% reduction in the RCS flow.

RCS thermal design flow and inlet temperature variations will be produced in the loops having asymmetric steam generator tube plugging levels; however, since the analysis was done at full power, coolant mixing in the reactor vessel lower plenum attenuates these conditions.

Based upon the preceding discussions, an approximate 1.5% reduction in RCS flow and the asymmetric tube plugging effects do not result in the violation of the licensing basis cr' aria following a RCCA ejection event, and the conclusions of the FSAR remain valid.

nm umu 4 49 1

I WESTINGHOUSE PROPRIETARY CLASS 3 4.2.22 Steam Line Break Mass / Energy Release - Inside/Outside Containment i

Various steam line break cases

'e analyzed for the purposes of generating mass and energy release rates w lch are then opD11ed to containment response or compartment environmental analyses.

Cases are performed assuming various break sizes and initial power levels.

Four major factors influence the release of mass and energy following a steam line break. These are steam generator fluid inventory, protection system operation, state of the secondary fluid blowdown, and primary to secondary heat transfer.

Tae RCS flow reduction due to the increased steam generator tube plugging levels would not affect the first factor and will have an insignificant impt.ct on the last three factors. The initial S/G mass will decrease by < 0.51.

Steam generator tube uncovery would occur slightly sooner. A dec ase in RCS flow would tend to reduce the primary to secondary heat transfer, theret reducing the steam pressure and temperature during normal operation. Any reduction in the secondary side temperature and pressure would tend to lessen the mass and energy released during a steamline break event. As a result, an approximate 1.5% reduction in RCS flow would not adversely affect the steamline break mass / energy releases. These statements are supported by the discussion in WCAP-10961, Rev. 1.

Asymmetric steam generator tube plugging levels would have the same effect described above (if the faulted S/G had a large amount of plugged tubes) or be bounded by the assumptions in the current analysis (for the S/G with less than average tube plugging).

Therefore, the conclusions of the current steamline break mass / energy release calculations are considered to be applicable for the reduced RCS flow and the

}

asymmetric tubt plugging scenario.

4.2.23 Setpoint Impact The impact of asymmetrical steam generator tube plugging levels on the J. M. Farley non-LOCA accident analyses have been presented; however, consideration of the affect of this increased steam generator tube plugging mu memo 4 50

l WESTINGHOUSE PROPRIETARY CLASS 3 and the asymmetry on the Overpower and Overtemperature reactor protection functions still needs to be addressed.

The coefficients for the setpoint equations are calculated using (among other inputs) the core thermal limits, RCS thermal design flow and the expected steam pressure at full power, nominal conditions.

Each of these parameters changed as a result of the increase in average steam generator tube plugging to a level of 15%.

It was determined that the setpoints currently in the safety analyses, and used as input to the Technical Specifications limits, l

need to be slightly revised to accommodate the increased steam generator tube plugging. Setpoint calculations have.shown that the Technical Specifications j

value for the K1 coefficient of the OTAT equation will change from 1.22 to l

1.18. There will be an effect on the K4 analysis value; however, the effect is so small that no choge is required for the Technical Specifications, since adequate margin exists in the Technical Specification value.

Additionally, the positive slope of the OTAT F(41) penalty function will be increased from 1.60 to 1.75 in the Technical Specifications.

It should be restated that the changes to the Overtv.perature and Overpower setpoint equations are such that a reactor trip will occur earlier in the analyses in which they actuate, when these setpoints are used. Allocation of generic ONB margin precluded changes to the DNS segment of the core thermal limits.

Therefore, an earlier reactor trip results in greater margin to the core limits and thus, DNB conditions. Therefore, it was not necessary to reanalyze the transients which credit OTAT or OPAT for reactor protection.

The asymmetric steam generator plugging levels create flow and inlet temperature asymmetries between the RCS i wps. However, each char.nel is being used to determine the AT in individual coolant loops under spe: _ric loop inlet and outlet conditions. Since only one channel exists in each loop, AT's may vary from loop-to-loop; however, the K-terms in the overtemperature AT setpoint equation (i.e., K1, K2 and K3) and in the overpower aT setpoint equation (i.e., K4, K5 and K6) will remain constant for all three loops.

Since the AT setpoints are based upon a fraction of the individual loop AT's and the loop channels are individually calibrated based i

upon the loop temperatures, the OTAT/0 PAT setpoints discussed above will am io emio 4 51

WESTINGHOUSE PROPRIETARY CLASS 3 i

continue to be valid under the asymmetric conditions considered in this evaluation.

4.2.24 CONCLUSION The impact on the non-LOCA licensing-basis analyses of plant operation with an increased average tube plugging level of 15% and asymmetric steam generator tube plugging levels (a maximum of 20% in one loop) has been examined. To support this evaluation, new analyses, as well as the evaluation of existing analyses have been performed.

Since the plant average steam generator tube plugging level remains at 15%, it was necessary to decrease the RCS thermal design flow assumption to 87200 gpm/ loop.

It has been concluded that this operation will not have a significant adverse impact upon the non-LOCA licensing-basis analyses.

In addition, the steamline' creak mass & entrgy release rates insioe and outside of containment also remain valid.

The impact of asymmetric steam generator tube plugging on the overpower and overtemperature reactor protection functions has also been considered and it has been determined that with minimal changes, the current setpoints for these functions provide adequate protection to the core limit' lines.

Therefore, all licensing-basis criteria continue to be met and the conclusions in the FSAR remain valid.

l i

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i Ffgure 4,g.g All Loops Operating. One Loop Coasting Down - DNBR Versus Time oo2ssacc73w 4 58

- _... -......... ~... -. - -. - -.

l l

)

1 h 1600. '

i i

1600.

1200.

4 I

e E 1000.

l I

=

see.

l g

r aoe.

~

0 100 tg1 102 to3 10' TIM (SEC) 1 i

4 2800.

2 00.

b E :...

It i2009.

l tett.

I l

l 1660. iet tot 108 tel to' Tim test) l Figure 4.2-7 NWOR RUPTURE OF A 114!11 FEEDWATER P!Pt.

WITHOUT OFF8ITE F0WER Pressuriser Pressure and Water Volume as a Function of Time conmomosso 4 59

700.

6St.

Tsat

-w pt.

660.

Tm 630.

r....

180.

oW Mo.

Teold b

~

w $49.

Sto.

SH'go got iet te3 it' tim (set) 700.

640. '

Tsat t Mt.

I660.

Thot 640.

W 600.'

t$...

Teold Me.

166.

Ste.

'# ',o ist w4 108 M'

3 fle test)

Figure 4.2-8 l

a apptgag w A 9818 FREDWATER P!Pt.

l

.ggggwt W F31Tt pcWER Loop Temperat e as a resttoa of Time 4 60-00353:10 470690

g m4 F' 8

'4 g

I 2

_i 9

a..

w l

u.a.e m

E v

8 g

i..

W

.m...

a' 5..

3..

TIME (SECONDS)

Figure 4.2-9 All Loops Operating, One Locked Rotor - Pressure Versus Time 4 61 oossanwosso

1.4 1.2 1_

i.

I a

5%

.6 t.

.4 g

d l

.2 v

0.

0 1

2 3

4 5

6 7

9 10 TIN 8 (SED Figure 4.2-10 All Loops Operating One Locked Rotor - Core Flow Versus Time 4 62 oos53:1o m osso

I 1.6 1.2 -

1.

f

.s I

g

.6 1

.2

e 1

2 3

6 9

6 7

8 9

it itu

<sse 1.6 1.2 1

t

.8

.I

=

.6 5

• 6 3

W

.4 e

i a

s 6

s 6

7 e

e to 71 4 (908)

Figure 4.2-11 411 Loops Operating, one '"cked Rotor - Flux T*ansients Versus Time oossa.io c os,o 4 63

I-i l

l l

2000.

1

1800, i

f-1600.

C

- 1400. <

w w

1200.

1 ak' w

1000..

%d 800.

l-600.

o.

1.

2.

3.

6 5.

4.

.7.

4.

9.

10.

l TIME (SEC) l l

L I'

f Figure 4.2-12 4

[:

).

All i. oops Operating, One Locked Rotor - Clad Temptr?ure Versus Time 4 64 c m s3.to e te:

WESTINGHOUSE PROPRIETARY CLASS 3 REFERENCES 4.1 WCAP-8970 (Proprietary) and WCAP-8971 (Non-Proprietary), " Westinghouse r

Emergency Core Cooling System Small Break October 1975 " April 1977.

4.2 Ciani, S., et al, " Simulation of Small Break Type Behavior of PUN and SPES Using the NOTRUMP Code,' Droceedings of the Specialist Meeting on l

Small Break LOCA Analyses in LWR's, Pisa, Italy, June 1985.

4.3 Lee, " Limiting Countercurrent Flow Phenomena in Small Break LOCA Transients," Proceedings of the Specialists Meetin; on Small Break LOCA Ar43yses in LWR's, Pisa, Italy, June 1985.

a 4.4 Rupprecht, S. D., et al, " Westinghouse Small Break LOCA ECCS Evaluation Model Generic Study with the NOTRUMP Ccde," WCAP-11145-P-A (Proprietary),WCAP-11372-A(Non-Proprietary), October 1986.

l 4.5 Lee, N., S. D. Rupprecht, W. R. Schwarz and W. D. Tauche, t

" Westinghouse Str.all Break ECCS Evaluation Model Using the NOTRUMP Code," WCAP-10054-P-A, (Proprietary) and WCAP-10081-A (Non-i Proprietary), August 1985.

i 4.6 Bordelon, F. M., Massie H. W. and Borden, T. A.

" Westinghouse ECCS Evaluation Model-Summary". WCAP-8339, (Non-Proprietary), July 1974.

l 4.7 Bordelon, F. M., et al, " SATAN-VI Program:

Comprehensive Space Time L

Dependent Analysis of Loss of Coolant", WCAP 8302, (Proprietary) June l'

1974, and WCAP-8306, (Non-Proprietary), June 1974.

4.8 Kelly, R. D., et al, " Calculated Model for Core Reflooding After a i.

Loss of Coolant Accident (WREFLOOC, Code)", WCAP-8170 (Proprietary) and l

WCAP-8171 (Non-Proprietary), June 1974.

4.9 Bordelon, F. M..and Murphy, E.

T., " Containment Pressure Analysis Code (C0CO)," WCAP-8327 (Proprietary) and WCAP-8326 (Non-Proprietary), June 1974.

l com,c onm 4 65

WESTINGHOUSE PROPRIETARY CLASS 3 4.10 Bordelon, F. M., et al, "LOCTA-IV Program:

Loss of Coolant Transient Analysis", WCAP-8301, (Proprietary).and WCAP-8305, (Non-Proprietary),

June 1974.

4.11 Bordelon, F. M., et al, " Westinghouse ECCS Evaluction Model -

Supplementary Information,' WCAP-8471-P-A, April ~1975 (Proprietary)-

and WCAP-8472-A. April 1975 (Non-Proprietary).

4.12

" Westinghouse ECCS Evaluation Model, 1981 Version," WCAP-9220-P-A, Rev.1 (Proprietary), WCAP-9221-A, Rev.1 (Non-Proprietary), February 1982.

4.13-Salvatcri, R., " Westinghouse Emergency Core Cooling System - Plant SensitivityStudies,"WCAP-8340,(Proprietary) July-1974.

l 4.14 Young, M., et al, "BART-1A: A Computer Code for the Best Estimate Analyzed Reflood Transients," WCAP-9561-P-A, with Addendum 2, 1984, (WestinghouseProprietary).

4.15 Kabadi, J. N., et-al, "The 1981 Version of the Westinghouse ECCS l

Evaluation Model using the BASH Code,' WCAP-10266, Revision,2, with-I Addenda, August '.986, (Westinghouse Proprietary).

l 4.16

9. H. Risher, J. and R. F. Barry, " TWINKLE - A Multi-Dimensional-Neutr n Kinetics Computer Code," WCAP-7979-P-A (Proprietary), WCAP-l 8028-A (Non-Proprietary), January 1975.

4.17 H. G. tiargrove, "FACTRAN - A Fortran IV Code for Thermal Transients in a UO2 Fue", Rod," WCAP-7908, June 1972.

p l

4.18 J. M. Farley Nuclear Plant Unit 1 :nd Unit 2 Final Safety Analysis Report Update.

l 4.19 T. W. T. Burnett, et al, "LOFTRAN Code Description," WCAP-7907-P-A 1

l (Proprietary), WCAP-7907-A (Non-Proprietary), April 1984.

i i

- :om ictm,a 4 66

~~

i WESTINGHOUSE PROPRIETARY CLASS 3 l

4.20 ANSI /ANS-5.1-1979, American National Standard for Decay Heat Power in Light Water Reactors, August 29, 1979.

4.21 NS-NRC-89-3466, "Use of 2700'F PCT Acceptance Limit in Non-LOCA Accidents," W. J. Johnson (Westinghouse) to Mr. Robert C. Jones (NRC),

October 23, 1989.-

4.22 Butler,-J. C., Love, D. S., et al, "Steamline Break Mtss/ Energy Releases for Equipment Qualification Outside Containme7t - Report to-the_ Westinghouse Owners Group High Energy-Line B eak/Superheated

-j Blowdowns Outside Containment Subgroup." WCAP-19061, Rev. 1,-October 1985.

4.23 Risher, D. H.,'Jr.; "An Evaluation of the Rod Ejection Accident ir, Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods," WCAP-7588, Revision I-A, January 1975.

I

. com io.onm 4-67 l

I WESTINGHOUSE PROPRIETARY CLASS 3 SECTION 5.0 NUCLEAR FUEL EVALUATION An evaluation of the effects of the increased steam generator tube plugging-l and reduced Thermal Design Flow on the fuel design was conducted.

The fuel:

l was reviewed with respect to the nuclear design, the thermal-hydraulic

.l I

design, and the fuel rod performance.

- 5.1 CORE DrGN The resulw d the core design evaluation indicated that the increased steam-generator tuue plugging level and reduced Thermal Design Flow result in no impacts to the core design except for the values for the maximum rod worth j

limit and the least negative Doppler defect assumed for the Rod Withdrawal from Subcritical analysis.

See Section 4.2.2 for results of the Rod Withdrawal from Subcritical analysis.

5.2 THERMAL-HYDRAULIC DESIGN The evaluation of the thermal-hydraulic design aspects of the core was performed and revealed that the proposed change in design parameters would result in a ONBR penalty.

For all DNBR safety analyses based on the W-3, R-grid correlation, generic DNBR margin was used to offset this. penalty.

For the steamline break analysis, which is based on the W-3 correlation, sufficient margin is available in the current analysis to accomodate the reduced flow.

1 For the core thermal limits, the exit boiling and exit quality limited portions of the limit lines were lowered to incorporate the. lower flow rate.

The DNB limitea portions of the current limit lines were unchanged due to the allocation of generic margin.

s The DNB design basis is met for Farley Unit 2 with the reduced Thermal Design Flow associated with the inc, eased steam generator tube pluggino level.

xm ie.mm 5-1.

4

WESTINGHOUSE PROPRIETARY CLASS 3 f

5.3 FUEL ROD PERFORMANCE Fuel performance evaluations are completed for_each fuel region to comonstrate that the design criteria will be satisfied for all fuel in the core under the planned operating conditions.

Evaluations of the effect of the proposed design parameters on meeting the fuel rod design criteria were performed for the Farley Unit 2 Cycle 8 design, which is the first anticipated occurrence of such conditions.

Based on this Unit 2 Cycle 8 evaluation, the effect of the reduced Thermal Design Flow and increased tube plugging has a negligible impact on meeting the the fuel rod design criteria for the Farley Units.

This will be confirmed _for_all future fuel regions as part of the cycle specific reload safety evaluation process.

52 com e anm l

WESTINGHOUSE PROPRIETARY CLASS 3 1

SECTION 6.0 NSSS/ BALANCE OF PLANT INTERFACE EVALUATION j

(-

l 6.1 MAIN STEAM SYSTEM l

An evaluation was performed to determine the effect of the proposed design l~

parameters on the main steam system.

The proposed' design parameters in Table l

2-1 show.that the full power steam flowrate will remain essentially l_

unchanged. Therefore the main steam system will not be significantly impacted by the increased tube plue@g and reduce 3 Thermal. Design Flow.

S/G Safety and Power Operated Relier u lves, MSIV's Since there is no increase in the full power steam flow rate, the installed Steam Generator Safety Relief Valve and Power Operated Relief Valve capacities remain adequate.

In additon, since there is no change to the zero load steam generator pressure due to the proposed design parameters,-there-will be no change to the steam. generator safety relief valve setpoints or the-steam generato power operated relief valve'setpoints.

Since there'will be

.no increase in the pressure drop across the Main Steam Isolation. Valves (MSIV's), the proposed conditions will not adversely impact the closure time of these valves.

4 6.2 MAIN FEE 0 WATER SYSTEM L

Since there is no increase in the full power feedwater flow rate nor in the i

feedwater temperatures, there is no significant impact on the main feedwater I

system.

6.3 AUXILIARY FEE 0 WATER SYSTEM

. increased steam generator tube plugging and reduced Thermal Design Flow do not impact the core decay hea.t nor the plant sensible heat, therefore there is'no change to the functional requirements or performance of the auxiliary feedwater system, f

b*l

'- C031310 012390

WESTINGHOUSE PROPRIETARY APPENDIX A l

l j

l l

1 l

l.-

4*

._ _ _. ________ ~ _ ___ _

(.

\\

l

(

l 670 "

Unacceptable Operation ese-2400 psia "8"

2250 psia s4e "

2000 psia sse 187g,4,

g. sas "

?

," s i s "

l see "

Acceptable Operation

(

gqq,,

See "

'~

570 "

Ss0e.

.1

.2

.5 4

.5

.6

.7

.e

.9 1.

L.1 1.2 POWCe Ifrect. ion'of nominalI l

Figun 2.11* Resster con safety Limit Three Leops in Opentien N

N As t

ty es t 6

n r.

1 NU N k l

Amen e ent lee. 21 is t

t.t

I lilg7lg7,lg}g 3

[

t 5,

si a1 1

1 L

8i1l3leleI31., ijl g],,,,

a s

~I E

s l

=3 I

s s.,

, aayy ge-g*,

Ig 3

sssgR..

))Ii5 5

s 3

3 ' a s,

'v 1 g3 E

vil vil vi vi vi as 5

g-s

! l5

~)

1] 5 s

g 1I]Ii_'i_il:

gi zi i

..i g

li n

i j' 8 3

i l.T!!!}ll;illIjis18nn tij !

{

l

)s-l,,.,n!as e

n I3w vi %i as 5 $$

E vie via vi vi g

/n g

g I

I j, !

4 1, #

i f l l1 i j !.5 )

Ty 1

1))

1 flip}!!'[hlIlili li 94 i

li it t il g

..,4as na FAALD-WIT 2 2-5

4

~

Table 2.2-1 (Continued)'

~

h REACIOR 1 RIP SYST[M INSTRUElliAll001 1 RIP 5(IP0f tlIS

.. g.4

. 1 7

BIDI AI1006 n

T I'8 S A ak g -K2 3

(M

• M &-F ) 4g (all)

U NOIE I: Overtemperature al <

3 les s u

2

~

where: aI, = Indicated af at RAi[8 Ile[RetAL POWER T = Average temperature, *f I' < 571.2*F (Itaximus Reference Tay, at RAi[0 TH[RptAL POWER)

P = Pressurizer pressure, psig P' = 2235 psig (Isominal RCS operating pressure) u 1

1+: S

• • 9 It s. S 2

compensation 3 g &

• 2 = Time constants attilred in the lead-lag controller for T,,,g = 30 secs,

• 2

• 4 58C5-S = Laplace transform operator, sec-I.

Operation with 3 Loops Operation with 2 Loops 9%

Kg =g

1. 8 Kg = (values blantt pending l

Eb gK2 = 0.0154 K2 = alRC approval of l

mio

{j K3 = 0.000635

.K3 = 2 Ioop operation) l ll15 and fg (AI) is a fenCtion of the Indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:

w o+--

_-w-,

-,w-,,,'-m+-=--er---+,

-eN-r~+..m s.-, --

c=m---

- - - ~ ' - - - - - -

m---

.s-~m-.<,v

+m

TABLE 2.2-1 (continued) m

)

REACTOR TRIP SYSTEM INSTRUENTATION TRIP SETPOINTS tu NOTATION continued l

6

[1) for qt -

between -35. percent and +9 percent. fg (al) = 0 (where qkvely and 9b are percent

\\

h RATED TE POK R in the top and bottom halvet of the core respect

, and qt

• 9b 15

e total T E RMAL POWER in percent of RATED THERMAL POWER).

(11) for each percent that the magnitude of (qt - qb) exceeds -35 percent, the as trip setootnt i

u l

l shall be automatically reduced by 1.37 percene of its value at RATED THERMAL. POWER.

j (111) for each percent that the magnitude of (qt - gh) Cacceds :-9 percent, the AT trip setpoint shall be setematically reduced by ercent of 1ts value at RATED THERMAL POWER.

Note 2:

Overpower AT < AT,[K -K5 1+ 35 T-y W W2

^

4 i

where:.

4, = Indicated AT at RATED THERMAL POWER

P T = Average tesperature. *F T* = Reference Tl bat RATED THERMAL POWER (Calliwatton tesperature for al l

testrumenta

, < 577.2*F)

K ~= 1.08 l

4 l

K5 = 0.0?/*F for increasing average temperature and 0 for decreasing average ' temperature i

]

K6 = 0.00iO9/*F for T > T"; K6 = 0 for T < T*

l

!95 lQ$

• 35

= The function generated by the rate lag controller for T.,9 dynande compensation jg 1+r 5 i

3

! =. $,

o

?e '

i j"

I i

l a

-l

.{

3 s

i e

s u

~

I

. e mi

=s 32 k

ml a

e r

sg h

W

• 4-g 3

• g-4 z.

~

g E.

~

-d 7

4 6 e wgg ea gis e

g a

f a.I E

S 2

a y

- g g~,

C vi ai

.2 6

4 8

g

.i s.

rt I

il*l IS s'

Y [ 13 3*'l

-a l-3 1

3e gly.

ain 5p a

g sa

..,1 a

a

, Es;

/

_s.....

,a e.se