ML20112J703

Nonproprietary Safety Evaluation for Indian Point Unit 3 W/Asymmetric Tube Plugging Among Steam Generators

ML20112J703
Person / Time
Site:	Indian Point
Issue date:	10/31/1984
From:	Forcht K, Kemper R, Segletes J WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
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ML093410203	List: ML20112J703
References
WCAP-10705, NUDOCS 8504050275
Download: ML20112J703 (72)

Text

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-10705 l

4 SAFETY EVALUATION FOR INDIAN POINT UNIT 3 WITH ASYMMETRIC TUBE PLUGGING AMONG STEAM GENERATORS OCTOBER 19B4 WORK PERFORMED FOR THE NEW YORK POWER AUTHORITY CONTRIBUTORS:

J. A. Segletes R. M. Kemper K. A. Forcht.

T. W. T. Burnett D. P. Dominicis Westinghouse Electric Corporation Nuclear Technology Division P. O. Box 355 Pittsburgh, Pennsylvenia 15230 gghk 6

4050 P

7583Q:10/100884

WESTINGHOUSE NON-PROPRIETARY CLASS 3 This document contains material that is proprietary to the Westinghouse Electric Corporation. The proprietary inf ornation has been marked by brackets. The basis for marking the information proprietary and the basis on which the information may be withheld from public disclosure is set forth in the affidavit of R. A. Wiesemann. Pursuant to the provisions of Section 2.790 of the Commission's regulations, this affidavit is attached to the application for withholding from public disclosure which accompanied this document.

This information is for your internal use only and should not be released to any persons or organizations outside the Office of Nuclear Reactor Regulation and the ACRS without the prior approval of Westinghouse Electric Corporation. Should it become necessary to obtain such approval, please contact R. A. Wiesemann, Manager, Regulatory and Legislative Affairs. Wastinghouse Electric Corporation, P. O. Box 355, Pittsburgh, Pennsylvania 15230.

4 5226Q:lD/112983

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 TABLE OF CONTENTS PBSe 1

INTRODUCTION I-1 II SAFETY EVALUATION - NON-LOCA II-l A. Asymmetric Tube Plugging Assumptions B. Computer Code Capabilities / Asymmetric Tube Plugging C. Initial Temperatures and Flow Distributions D. Transients Reanalyzed 1.

Rod Withdrawal at Power 2.

Steamline Break 3.

Loss of Flow 4.

Locked Rotor 5.

Dropped Rod 6.

Loss of Normal Feedwater E. Transients Not Reanalyzed F. Delta-T Response Evaluation III SAFETY EVALUATION - LOCA III-1 IV CONTROL AND PROTECTION SYSTEM SETPOINTS IV-1 V

CONCLUSIONS V-1 VI REFERENCES VI-1 7583Q:10/102384

WESTINGHOUSE NON DROPRIETARY CLASS 3 LIST Of TABLES Sunenary of Reactor Coolant Loop Flows and Temperatures at Normal Design 11-1 Full Power Sununary of Reactor Coolant Loop Flows and Temperatures at Worst Steady 11-2 State Operation Reactor Coolant Temperature at Reference Overpower Core Limit 11-3 II-4 Time Sequence of Events I

I 7583Q:1D/100584

WESTINGHOUSE NON-PROPRIETARY CLASS 3 LIST OF FIGURES 11-1 Rod Withdrawal at Power at 80 pcm/sec

- Nuclear Power

- Core Heat Flux II-2 Rod Withdrawal at Power at 80 pcm/sec

- Pressurizer Pressure

- Pressurizer Water Volume II-3 Rod Withdrawal at Power at 80 pcm/sec

- Loop 2 T,yg

- DNBR II-4 Rod Withdrawal at Power at 1 pcm/sec

- Nuclear Power

- Core Heat Flux 11-5 Rod Withdrawal at Power at 1 pcm/sec

- Pressurizer Pressure

- Pressurizer Water Volume II-6 Rod Withdrawal at Power at 1 pcm/sec

- Loop 2 T,yg

- ONBR II-7 Rod Withdrawal at Power

- Minimum DNBR vs. Reactivity Insertion Rate 11-8 Steam Line Break

- Nuclear Power

- Core Heat Flux

- RCS Pressure

- Pressurizer Water Volune II-9 Steam Line Break

- Feedwater Flow

- Steam Flow

- Core Flow 75830:10/101884

WESTINGHOUSE NON-PROPRIETARY CLASS 3 LIST OF FIGURES (Cont.)

l 11-10 Steam Line Break

- Core Average Temperature

- Reactivity

- Core Boron

- Steam Pressure 11-11 Partial Loss of Flow

- Nuclear Power

- Core Heat Flux 11-12 Partial Loss of Flow

- Pressurizer Pressure

- Core Flow II-13 Partial Loss of Flow

- DNBR II-14 Complete Loss of Flow

- Nuclear Power

- Core Heat Flux II-15 Complete Loss of Flow

- Pressurizer Pressure

- Core Flow 11-16 Complete Loss of Flow

- DNBR II-17 Locked Rotor

- Nuclear Power

- Core Heat Flux II-18 Locked Rotor

- RCS Pressure

- Core Flow 11-19 Locked Rotor

- Clad Inner Temperature 11-20 Dropped Rod

- Nuclear Power

- Core Heat Flux 11-21 Dropped Rod

- Average Coolant Temperature

- Pressurizer Pressure 7583Q:10/101884

WESTINGHOUSE NON-PROPRIETARY CLASS 3 LIST OF FIGURES (Cont.)

11-22 Dropped Rod

- Inlet Temperature

- Steam Load 11-23 Loss of Normal Feedwater

- Pressurizer Pressure

- Pressurizer Water Volume 11-24 Loss of Normal Feedwater

- Loop 1 Temperature

- Steam Generator Pressure 111-1 Schematic of WREFLOOD Model of Westinghouse PWR III-2 WREFLOOD Resistance Network Representation of a PWR 7583Q:10/102384

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 SAFETY EVALUATION FOR INDIAN POINT UNIT #3 WITH ASYMMETRiJ TU8E PLUGGIN3 AMONG STEAM GENERATORS i

I.

INTRODUCTION i

Indian Point Unit 3 (IP3)is currently operating under a Technical Specification (Tech Spec) that limits the amount of tube plugging in any steam generator to 24%. Approval of this Tech Spec was granted by the NRC on January 13, 1984. Safety analyses (Reference I-1) were performed for a 24%

uniform tube plugging level in support of this Tech Spoc change. The results of the uniform 24% tube plugging evaluation for the FSAR (Reference I-2)

Chapter 14 transients, other than LOCA, showed that the reactor coolant ficw rate and reactor vessel average temperature are the only safety related parameters that are significantly affected by tube plugging. The implication of the 24% uniform tube plugging analysis was that or.e or more of the four steam generators (SG) could have greater than 24% tube plugging if the others had less than 24% tube plugging provided that the reactor coolant flow rate remained greater than that predicted for 24% uniform tube plugging and the hottest cold leg temperature remained less than the T, predicted for 24%

g uniform tube plugging.

1 1

i For the'ECCS performance analysis, the overall loop resistance to flow, rather than the relative resistance among loops, is of primary importance.

i Currently the equivalent tube plugging levels

• in the four steam generaturs are 16.0%, 9.8%, 8.4% and 8.0%.

While these levels are significantly less than 24%, they represent a degree of asymmetry. The present level of plugging The equivalent tube plugging level accounts both for tubes plugged and tubes sleeved. Twenty sleeved tubes are assumed to be the hydrodynamic equivalent of one plugged tube. The current plugging, sleeving, and normalized equivalent plugging level based on 3260 tubes /SG are as follows:

Plugged Tube SG Tubes Plugged Tubes Sleeved Equivalent (%)

31 483 768 16.0 32 286 651 9.8 33 231 850 8.4 34 227 701 8.0 r

7583Q:1D/100984 I-l

WESTINGHOUSE NON-PROPRIETARY CLASS 3 in one of the steam generators could, with continued plugging, cause it to exceed the 24% level in the future.

To demonstrate that asymmetry will not jeopardize safe operation, a study has been performed in which the plugging level in one steam generator is 30%, with two steam generators at 245 and the fourth steam generator at 85. This was chosen to encompass the highest postulated asymmetry level assumed in the study, while still maintaining a high overall plugging level. Steady-state DNBR and peak linear power remain equal to the value previously assessed for 24% uniform tube plugging,.

The most apparent effects of asymmetry are the resulting differences in loop flow rates and, to a lesser extent, reactor coolant system (RCS) temperatures. The asymmetry in temperature, however, is attenuated because of mixing that occurs in the lower and upper reactor vessel plenums.

The core safety limits are preserved, in part, by the overtemperature delta-T (OTAT) and overpower delta-T (OPAT) reactor trips. Asymmetric tube plugging may cause asymmetries in measured loop temperatures, T,yg, and l

ATs. These temperature asymmetries, however, will not cause significant loop-to-loop variations in the delta-T protection system because each channel is calibrated at power based on measured temperatures in that loop and the effect of loop average temperature asymmetry is factored into the overtemperature delta-T equation. Operating data to date from Indian Point Unit 3 indicates that differences in core flow and inlet temperature between loops is less than predicted.

1 i

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75830:10/101884 I-2

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 II. SAFETY EVALUATION - NON LOCA A.

Asymmetric Tube Plugging Assumptions The following assumptions were made in the reanalysis of the non-LOCA transients:

(1) Tube Plugging Distribution The tube plugging distribution analyzed is 30%, 24%, 24% and 8%. This distribution encompasses the upper plugging level assumed in the study (30%), the lowest possible level based on the current plugging levels (8.0%) and two equal intermediate levels that play a role in simulating the overtemperature Delta-T reactor trip setpoint equation.

(2) Loop Flows Although loop flows will be asymmetric, the sum of all loop flows are assumed to equal the reactor vessel flow of 323,600 gpm. This reactor vessel flow equals the therwel design flow established for 24% uniform tube plugging analysis in Reference I-1.

The flow for each icop was calculated based on the reactor coolant pump head curve and the loop flow impedance, including the effect of plugged tubes in the steam generator.

(3) Inlet Temperature Inlet temperatures will vary from loop to loop; however, the highest inlet temperature is assumed to be 542.9'F at nominal full power steady state conditions. This temperature equals the inlet temperature predicted for 24% uniform tube plugging in Reference I-1.

This assumption is consistent with the operating constraints of the plant, i.e., the hottest inlet temperature is limiting. The safety analysis uses a value of 546.9'F to account for 4*F control and measurement uncertainty.

7583Q:10/100584 11-1

s WESTINGHOUSE NON-PROPRIETARY CLASS 3 (4) DNB/ Inlet and Core Temperatures The departure from nucleate boiling ratio (DNBR) is romputed conservatively for the core hot channel assuming that fluid from the hottest cold leg ficws directly into that channel.without benefit of mixing with fluid from cooler cold legs.

(5) Steam Generator Heat Transfer For each steam generator, the overall heat transfer coef ficient was taken as proportional to the number of unplugged tubes.

s (6) Reactor Vessel Fluid Mixing Loop-to-loop temperature asymmetries depend-upon the degree of mixing in the reactor vessel. Three different mixing assumptions were assessed:

Perfect mixing (all hot leg temperatures equal); minimum mixing within the constraints of the model (see Section II.C); and an intermediate mixing case. [

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B.

Computer Code Capabilities / Asymmetric Tube Flugging The asymmetric effects for non-LOCA accident analyses were computed using the LOFTRAN computer code (Raference II-1). This code has the capability to

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permit the input of pressure drop ratios (loop / loop average) and SG heat transfer area ratios (loop / loop average) for each loop. This information, in conjunction with the plant data, allows the calculation'of initia'l (steady state) conditions end subsequent transient responses to imposed accident conditions.

C.

Initialization of Flows and Temperatures and Effect of Mixing Initial temperatures and loop flow distributions were calculated using the LOFTRAN code. These calculations resulted in a set of initial conditions in which the vessel flow agreed with the desired vessel flow (323,600 gpm) and the hottest inlet temperature agreed with the desired value (542.9 without uncertainty or 546.9*F with uncertainty).

7583Q:10/100884 II-2

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MESTINGHOUSE N0'i-PROPRIETARY CLASS 3 The calculations were made for three reactor vessel inlet plenum chamber

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< mixing assumptions: (1) perfect (i.e. complete'. mixing (2) minimum mixing, and (3) intermediate mixing.

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The results of the initiaTization computations are pret,erted in Table IT-1 for nem?ng1 conditions (3025 MWt, 323600 gpm, 2250 psia, 542.9'F Tinlet) ""d i"

Toble 11-2 for worst steady-state operation (102% x 3025 Nwt, 323600 gpm, 2220 psia, 546.9'F Tinlet). The conditions sh6wn in Table 11-1 are used to set the overtemperature delta-T constants. Table II-2 values are used as the initial conditions for the accident aralyses.

D.

Transients Reanalyzed l'

1.

12od Withdrawal At Power (IP3 FSAR Section 14.1.2)

- For the majority of plant transients (those for which the core DNB limits are applicable, have peaking factors no worse than design, have core flow noles>thandesign.,andproceed$10wlyenoughforresponseofloop temperatures), core DNB protection is provided, if necessary, by the overtemperature delta-T trip.

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The nominal overtemperature delta-T reactor trip setpoint egaation inscluding dynamic compstsatior, and setpoint reductio: for adverse core -

^

sxtil power distributibn) is given by the formula below:

te

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7583Q:1D/102384 II-3

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3p = AT, [1.135

.0114 (T,yg - T,0yg) +.00066 (P - 2250)],

AT where AT, is indicated AT, "F, at nominal full power for the channel being calibrated:

T is indicated average temperature. *F, for the channel being g

calibrated; T*yg is indicated average temperature, *F, at nominal full power for the channel being calibrated; and P is pressurizer pressure.

Thus, each protection channel is calibrated in terms of the individual temperatures, AT,and that exist at nominal full power in yg each individual loop. This effectively makes the margin to trip equal for all channels, and independent of temperature asymmetries.

Full power initial conditions were assumed in this analysis to best demonstrate the dynamic response of the loop temperatures and the overtemperature delta-T protection system with asymmetries, and to illustrate the way in which the protection system prevents the core f rom exceeding the DNBR limit.

Method of Analvsis The rod withdrawal at power accident analysis was simulated using the LOFTRAN computer code.

The recomended AT calibration is accounted for in this analysis. This includes the effect of each channel being calibrated at power based on the indicated AT and T,yg for that channel at normal full power. (Refer to Table 11-1 for values of AT and T,yg for different loops at full power with different mixing assumptions.)

Except as noted in Section II.C and above, analysis methods and assumptions are consistent with the original FSAR assumptions:

75830:10/101884 II-4

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Initial power = 1.02 x nominal Initial pressure = nominal - 30 psi Nuclear flux trip setpoint = 118% of nominal Nuclear flux trip delay = 0.5 sec.

OTAT trip delay = 6.0 sec.

Reactivity Insertion Rates = variable (0.6 pcm/sec. to 80 pcm/sec)

Conservative feedback conditions This analysis was first performed for the three mixing assumptions described in Section II.C, for both steady state and traasient conditions,

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Steady State Analysis and Results

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7583Q:lD/100984 I -5

WESTINGHOUSE NON-PROPRIETARY CLASS 3 L

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Transient Analvsis and Results For transient verification of the adequacy of the overtemperature delta-T trip, a rod withdrawal analysis for 3 pcm/sec reactivity insertion rate was performed for each of the three mixing assumptions. A rate of 3 pcm/ set was selected as approximately the worst rate (lowest minimum DNB ratio) since it approaches both the nuclear overpower and the overtemperature delta-T trip.

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The time sequence of events for the rod withdrawal is shown in Table 11-4.

Nuclear power, core heat flux, pressurizer pressure, 75830:10/101884 II-6

UEST!NGHOUSE NON-PROPRIETARY CLASS 3 pressurizer water volume, channel 2 vessel average temperature and DNBR are presented in Figures 11-1. II-2, and II-3 for the 80 pcm/sec reactivity insertion and in Figures II-4, II-5, and II-6 for the 1 pcm/sec reactivity insertion.

The DNBR as a function of reactivity insertion rate is presented in Figure II-7 for asymetric tube plugging over a wide range of insertion rates.

Conclusions

1) [

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2) Because of the way in which the delta-T reactor trips are calibrated at power, loop-to-loop temperature and flow asymmetries do not reduce core safety margins during the rod withdrawal at power transient.

3) Minimum DNBR during the rod withdrawal at power transient is in excess of 1.30, i.e., the core safety limits are preserved, for all reactivity insertion rates.

2.

Main Steamline Break (IP3 FSAR Section 14.2.5)

In general, steam generator tube plugging will make the steamline break accident less severe since the cooldown rate of the reactor coolant system following a steamline break will be retarded as a result of tube plugging. It is conceivable, however, that the results could be affected if the postulated break is in the steamline associated with the least plugged steam generatcr. Therefore this was the configuration analyzed.

Method of Analysis An analysis was performed for the most severe case in the IP3 FSAR, which is an iaside containment break at zero power with offsite power l

7583Q:10/101884 II-7 l

WESTINGHOUSE NON-PROPRfETARY CLASS 3 available. A double ended rupture was assumed to occur in Loop 4 (plugging level = 8%) between the steam generator and the flow restrictor. A safety injection signal was assumed to occur at 0.6 seconds into the accident, resulting from a high steamline differential pressure signal. Steamline isolation was assumed to occur later on a lo-lo Tavg coincident with high steam flow signal. Asymmetric plugging parameters were input to LOFTRAN to provide a 30%, 24%, 24%, 8% plugging distribution in Loops 1 through 4 respectively.

Except as noted above and accounting for asymetries, the assumptions are consistent with the FSAR assumptions, i.e., shutdown margin is equal to 1.72%, initial temperature is equal to no load temperature and initial pressure is equal to 2250 psia.

Results The time sequence of events for the steamline break transient is given in Table II-4.

Nuclear power, core heat flux, RCS pressure, pressurizer water volume, feedwater flow, steam flow, core flow, core average temperature, reactivity, core boron and steam pressure are presented in Figures II-8 through 11-10 for the steamline break analysis. A detailed independent DNBR analysis was performed based on state points computed by LOFTRAN. The DNBR for this case was determined to be greater than the limit value of 1.30.

Conclusions A 30%, 24%, 24%, 8% plugging distribution during a worst case steamline break will not result in a DNBR that is lower than the allowable limit.

3.

Loss of Flow (IP3 FSAR Section 14.1.6)

The 1/4 and 4/4 loss of flow transients were recalculated to demonstrate the DNBR criteria (DNBR >1.30) can be met. These analyses were based on revised reactor coolant pump data that provide more conservatism than in the original FSAR.

J l

75830:10/102384 II-8

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Method of Analysis An analysis was performed for a 1/4 and 4/4 loss of flow accident assuming a 30%, 24%, 24%, 8% plugging distribution. For the 1/4 case, the coastdown was assumed to occur in Loop 4 (8% plugging) because a coastdown in the least plugged loop will result in the lowest core flow. Asymmetric plugging parameters were input to LOFTRAN to provide a 30%, 24%, 24%, 8%

tube plugging distribution in Loops 1 through 4 respectively. The reactor coolant pump dynamic response was computed by using pump data which simulated 4/4 loop coastdown measurements.

Conservatism in the flow coastdown analysis was assured as follows:

First, a "best-estimate" 4/4 flow coastdown analysis was done for comparison with the actual flow coastdown measurements taken during initial plant startup tests conducted in 1976. This best-estimate analysis assumed as-built reactor coolant pump performance, best-estimate flows and pressure drops, and design pump inertia. The best-estimate calculated flow coastdown was found to be in excellent agreement with actual plant startup measurements, as indicated below:

Core Flow. Fraction of Initial Time af ter loss Best-Estimate Plant of DumD Dower. see Calculation Mea %urement*

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1.5 3.5 5.5 7.5 9.5

• corrected for 0.50-second flow sensor delay.

Then, conservatism was added to provide margin in the safety analyses by a) assuming only 90% of the pump design inertia; and b) increasing the 7583Q:10/100884 11-9

WESTINGHOUSE NON-PROPRIETARY CLASS 3 loop pressure drops such that total vessel flow was reduced to its thermal design value of 323,600 gpm, with the loop-to-loop asymmetry shown in Table 11-2.

The flow coastdown at full power is shown in Figure II-15.

Except as noted in Section II.C and above, the analysis assumptions are consistent with the FSAR assumptions, i.e., initial power = 1.02 x nominal, initial pressure includes a -30 psi uncertcinty, and beginning of core 1tfe reactivity coefficients.

The loss of flow transients were computed using the LOFTRAN, FACTRAN and THINC conputer codes.

Results Nuclear power, core heat flux, pressurizer pressure, core, flow and DNBR are presented in Figures 11-11 through 11-13 for the 1/4 loss of flow case and in Figures 11-14 through 11-16 for the 4/4 loss of flow case. The DNBR for these cases was determine'd to have met the allowable value. of 1.30.

1 Conclusions A 30%, 245, 24%, 8% tube plugging distribution during a worst case.1/4 loss of flow or 4/4 loss of flow will not result in a DNBR that is lower than the allowable limit. Asynnetric tube plugging will have essentially l

no effect on a 4/4 LOF transient and only a slightly adverse effect for a loss of flow in the loop containing 85 SG tube plugging. In the complete loss of flow case the core flow will remain essentially the same as with uniform tube plugging and in the partial loss of flow case the core flow will be slightly reduced when compared with uniform plugging.

4.

Locked Rotor (IP3 FSAR Section 14.1.6)

The 24% uniform tube plugging analysis will bound the asymmetric case if the average flow in the three lowest flow loops meets the thermal design flow for three 24% uniform tube plugged loops (3 x 80.900 = 242,700 gpm).

To demonstrate, however, that this criterion is not necessary, the locked rotor accident was reanalyzed.

7583Q:10/101884 11-10

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Method of Analysis An analysis was performed assuming the locked rotor is in Loop 4 (plugging level = 8%) because a coastdown in the least plugged loop will result in the lowest core flow. Tube plugging parameters were input to LOFTRAN to provide a 30%, 24%, 24%, 8% tube plugging distribution in Loops 1 through 4 respectively. The reactor coolant pump dynamic responses are based on the pump homologous curves developed from test data.

Plant responses were computed using the LOFTRAN code. Fuel and clad temperatures at the hot spot were computed using the FACTRAN code.

Current analytical methods were used for this transient. Unlike previous IP3 locked rotor analyses, the three unfaulted pumps were assumed to coastdown as a result of a loss of offsite power. A hot spot analysis was performed to demonstrate a coolable geometry.

A DNBR analysis was nbt necessary since rods in DN8 are predicted to not fail based on the temperature-time criteria presented in Reference II-2.

Results The time sequence of events for the locked rotor transient is given in Table 11-4.

Nuclear power, core beat flux, RCS pressure, core flow and clad inner temperature are shown in Figures II-17 throt;h II ~i3.

A peak clad temperature of 2006*F and a peak reactor coolant system pressure of 2565 psia were conservatively calculated to be reached during the transient.

Conclusions A 30%, 24%, 24%, 8% tube plugging distribution during a locked rotor event will not result in exceeding any safety limits.

75830:10/100984 I1-11

WESTINGHOUSE NON-PROPRIETARY CLASS 3 5.

Oropped Rod (IP3 FSAR Section 14.1.3)

The dropped rod accident may result in a reactor trip on OTAT. As noted previously in the uncontrolled bank withdrawal at power accident analysis, the response of the overtemperature delta-T reactor trip may be slightly affected by asymmetric tube plugging. For this reason the dropped rod accident was reanalyzed for asymmetric tube plugging.

I l

l Method of Analysis I

The analysis was performed for a 30%, 245, 24%, 8% plugging distribution by providing the asymmetric plugging parameters to LOFTRAN. In this analysis, the response of the plant is computed for a series of dropped rod worths. Statepoints (thermal flux, reactor coolant pressure, inlet temperature) at the limiting point in the transient are used to compute the nuclear enthalpy rise hot channel factor (FAH). Rod drop is assumed to occur from 102% power with conservative feedback properties.

Results The time sequence of events for the dropped rod transient is shown in Table 11-4.

Figures II-20 through 11-22 illustrate the transient response following a dropped rod of worth 100 pcm. The reactor coolant average temperature decreases initially, due to the decrease in reactor core power. Since the drop in power is less than the drop in load, with no reactivity feedback, coolant temperature then increases. As a result of the vessel average temperature increase, an overtemperature delta-T reactor trip occurs to terminate the transient. A DNS evaluation at the limiting condition in the transient shows that the DNBR remains above 1.30.

Conclusions

[

A 30%, 24%, 24%, 8% tube plugging distribution during a dropped rod event will not result in a minimum DNBR lower than the allowable limit.

7583Q:1D/102384 II-1?

I

WESTfNGHOUSE NON-PROPRIETARY CLASS 3 6.

Loss of Normal Feedwater Flow (IP3 FSAR Section 14.1.9)

The primary concern in the Loss of Normal Feedwater analysis is the ability of the auxiliary feedwater system to remove decay heat with the reduced heat transfer area due to the asymmetric tube plugging. To address this concern, a uniform 30% steam generator tube plugging is assumed. This represents the most conservative assumption in order to depict the minimum amount of primary to secondary heat transfer capability f rom any one steam generator.

A loss of normal feedwater results in a reduction in capability of the secondary system to remove the heat generated in the reactor core.

If an alternative supply of feadwater is not supplied to the plant, core residual heat following reactor trip will heat the primary system water to the point where water relief from the pressurizer will occur, resulting in a substantial loss of water from the RCS. Since the plant is tripped well before the steam generator heat transfer capability is reduced, the primary system variables never approach a DN8 condition.

The reactor trip and auxiliary feedwater initiation on low-low water level in any steam generator provides the necessary protection against a loss of normal feedwater.

An analysis of the system transient is performed to show that following a loss of normal feeduater and with rad'sced primary to secondary heat transfer due to steam generator '.obe plugging, the auxiliary feedwater system is capable of removiri the stored and residual heat, thus preventing either overpressurization of the RCS or loss of water from the reactor core, and returning the plant to a safe condition.

' Method of Analysis A detailed analysis using the LOFTP.AN code is performed in order to obtain the plant transient following a loss of normal feedwater.

Assumptions made in the analysis are:

7583Q:10/100984 II-13

UEST!NGHOUSE NON-PROPRIETARY CLASS 3 A.

The plant is initially operating at 102 percent of 3216 MWt (the maximum calculated turbine rating).

B.

A conservative core residual heat generation based upon long term operation at the initial power level preceding the trip.

C.

Reactor trip occurs on steam generator low-low level.

D.

Only one auxiliary feedwater pump with a capacity of 400 gpm is available one minute after the low-low level setpoint is reached.

E.

Auxiliary feedwater is delivered to only two steam generators, both of which have 30% of their tubes plugged.

F.

Secondary system steam relief is achieved through the steam generator power-operated relief valves and/or safety valves.

G.

The initial reactor coolant average temperature is 4*F lower than the nominal value, and initial pressurizer pressure is 30 psi higher than nominal.

H.

Reactor coolant pump coastdown was assumed to occur after reactor trip and a steam generator heat transfer coefficient consistent with natural circulation was assumed.

The assumptions used in this analysis are designed to minimize the energy removal capability of the system and to maximize the possibility of water relief from the coolant system by maximizing the coolant system expansion, as noted in the assumptions listed above.

Results Figures II-23 and 11-24 show the significant plant parameters following a loss of norum1 feedwater without offsite power available.

The capacity of one auxiliary feedwater pump is such that the water level in the steam generators being fed does not recede below the lowest level 7583Q:10/102284 II-14

WESTINGHOUSE NON-PROPhiETARY CLASS 3 at which sufficient heat cransfer area is available to dissipate core residual heat without water relief from the pressurizer safety vcives.

Figure II-23 shows that at no time is there water relief from the pressurizer.

l The calculated sequence of events for this accident is listed in Table

)

II-4.

As shown in Figures II-23 and II-24, the plant approaches a stabilized condition following reactor trip and auxiliary feedwater initiation.

Conclusions Results of the analysis show that a loss of normal feedwater does not adversely affect the core, the RCS, or the steam system since the auxiliary feedwater capacity is such that reactor coolant water is not relieved from the pressurizer relief or safety valves.

E.

Transients Not Reanalyzed i

1.

Rod Withdrawal From Subcritical (IP3 FSAR Section 14.1.1)

The asymmetric plugging case is bounded by the 24% uniform pluggir.g case provided that the DNB margins and operating limits are paintained. These" requirements are ensured by maintaining the same reactor flow as in the 24% uniform plugging analysis (323,600 gpm) and maintaining the same reactor coolant temperature at zero power.

2.

Boron Dilution (IP3 FSAR Section 14.1.5)

Boron dilution during shutdown will not be affected by plugging asymmetries. Total RCS volume, which is the plant parameter of concern in this event, will Hbe no less for asymmetric tube plugging than it is for the 245 uniform plugging case.

Boron dilution at power with asymmetric tube plugging is no worse than for 245 uniform plugging since the RCS volume will be no less for asymmetric 75830:10/100884 I1-15

WESTINGHOUSE NON-PROPRIETARY CLASS 3 plugging than it was for 24% uniform plugging. A boron dilution incident with the reactor in manual control is bounded by the rod withdrawal at power event which was already analyzed in Section II.D.l.

3.

Startup of an Inactive Loop (IP3 FSAR Section 14.1.7)

Startup of an Inactive Loop could be slightly affected by asymmetric tube plugging. Since N-1 operation at Indian Point Unit 3 is not licensed, this event was not reanalyzed.

4.

Loss of External Electrical Load (IP3 FSAR Section 14.1.8)

The asymmetric plugging case is bounded by the 24% uniform plugging case provided DNB margins and operating limits are maintained.

5.

Excessive Heat Removal Due to Feedwater System Malfunctions (IP3 FSAR j

I Section 14.1.10)

This event will not be adversely affected by asymmetric tube plugging. No reactor trip on over temperature delta-T is needed because, as noted in the FSAR, the DNBR increases following initiation of the event.

l 6.

Excessive Load Increase (IP3 FSAR Section 14.1.11)

This event will not be adversely affected by asymmetric tube plugging. No reactor trip on overtemperature delta-T is needed because, as noted in the FSAR, the DNBR increases following initiation of this event.

7.

Rod Ejection (IP3 FSAR Section 14.2.6)

The asymmetric tube plugging case is bounded by the 24% uniform plugged case provided the operating limits are maintained. These requirements are ensured by maintaining conservative reactor coolant flows and vessel inlet temperatures relative to the 24% uniform tube plugging analysis, i

l 7583Q:1D/100884 II-16 l

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8.

Blackout / Natural Circulation (FSAR Section 14.1.12)

In the event offsite power is lost, the reactor coolant pumps will coast down and the RCS flow will eventually reduce to natural circulation flow.

With asymmetric tube plugging, natural circulation flow rates will be slightly different between the loops. However, this is neither a large nor a significant effact. The dominant driving force for natural circulation 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 acts to force flow through all reactor coolant loops. The largest impedance to flow through each loop is the reactor coolant pump with its stationary rotor. This locked-rotor pump impedance is much greater than the impedance through steam generator tubes, even with 30% tube plugging. Thus, the difference between two loops, one with 8% plugged tubes and one with 305 plugged tubes, is less than 10% in relative flow.

In any event, natural circulation flow is low enough that all cold leg temperatures are essentially equal to the temperature on the secondary side of the steam generator. Therefore, no significant loop temperature assymmetry is expected during natural 1

circulation as a result of asymmetric tube plugging.

I 9.

Dropped Bank (IP 3 FSAR Section 14.1.4)

The asymmetric plugging case is bounded by the 24% uniform plugging case provided the ONB margins and operating limits are maintained. These requirements are ensured by maintaining conservative reactor coolant flows and vessel inlet temperatures relative to the 24% uniform tube plugging analysis. Unlike the dropped rod case, a reactor trip on overtemperature delta T will not occur and the rod worth inserted will be uniformly distributed within the core for the dropped bank case. Therefore, the-concerns due to tube plugging asymmetries that existed for the dropped rod case do not exist for the dropped bank case.

7583Q:10/102384 II-17

WESTINGHOUSE NON-PROPRIETARY CLASS 3 F.

Delta-T Response Evaluation The hot leg temperature response time is important to the operation of the delta-T protection system, and has been assessed for the effects of asymmetric tube plugging. The results show that hot leg temperature response is relatively insensitive to tube plugging.

Fluid transit time from the reactor vessel outlet plenum to the hot leg temperature sensor is composed of two components:

(1) transit time from the vessel outlet plenum to the hot leg scoops and (2) transit time from the hot leg scoops to the RTD manifold. Fluid transit time from the vessel outlet plenum to the hot leg scoops (less than 0.5 seconds at design flow) is inversely proportional to flow, and will increase as the number of plugged tubes increases. However, flow through the hot leg RTD bypass (from hot leg scoop to RTD manifold) increases as the pressure drop across the steam generator increases. Therefore, fluid transit time from the hot leg scoops to the RTD manifold (calculated as less than 0.4 seconds at design flow) increases as the number of plugged tubes increases. The total of these two components is essentially constant for tube plugging levels less than 50%.

Based on FSAR assumptions on time delays, a 1.5 second delay time from vessel to hot leg manifold would be acceptable. This delay time would not.

be reachef unless the plugging level in a steam generator substantially exceeded 50%.

Based on the above evaluation, it is concluded that:

a.

Tube plugging up to a 50% level has virtually no ef fect on the AT response and would be acceptable up to approximately 75% plugging.

b.

T,yg and AT filters and rate compensation need not be reset as a result of tube plugging.

l l

l 75830:10/102384 11-18

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Table 11-1 Summary of Reactor Coolant Loop Finws and Temperatures at Nominal Design Full Power (3025 MWt, 323600 gpm, 2250 psia, 542.9'F T inlet Pct Perfect Intermed.

Minimum Pa rameter Loog Pluaaina Mixina Mixina Mixina Normalized Flow, 1

30

"~

"~

Fraction of 80,900 gpm 2,3 24 4

8 Vessel Inlet 1

30 Temp,(*F) 2,3 24 (542.9* for uniform 4

8 24% plugging)

Vessel Average 1

30 Temp (*F) 2,3 24 (574.7* for uniform 4

8 24% plugging)

Vessel Delta-T (*F) 1 30 (63.7 for uniform 2,3 24 24% plugging) 4 8

75830:10/101884 II-19

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Table II-2 Sununary of Reactor Coolant Loop Flows and Temperatures at Worst Steady-State Operation (1025 x 3025 Mwt, 323600 gpm, 2220 psia, 546.9'F Tinlet}

Pct Perfect Intermed.

Minimum Parameter M

Pluaaina Mixina Mixing Mixina Normalized Flow, 1

30 a,c Fraction of 80,900 gpm 2,3 24 4

8 Vessel Inlet 1

30 Temp, including 2,3 24 4*F uncertainty (*F) 4 8

Vessel Average 1

30 Temp (*F) 2,3 24 4

8 Vessel Delta-T (*F) 1 30 2,3 24 4

8 l

l l

l 7583Q:10/102384 II-20

l l

WESTINGHOUSE NON-PROPRIETARY CLASS 3 l

TABLE II-3 REACTOR COOLANT TEMPERATURE AT REFERENCE OVERPOWER CORE LIMIT 3c a

Non-Uniform (30-24-24-8%) Dluacina Plugging Uniform Perfect Intermediate Minimum L2gg Level (5) 245 Pluacina Mixina Mixina Mixina Vessel Inlet Temp, *F 1

30 a,c 2,3 24 4

8

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'F (and change from design full power) 2,3 24 4

8 i

Vessel Delta-T, 'F 1

30 (and, as percent of j

indicated AT at design full power) 2,3 24 4

8 Protection Channel 1

30 Safety Margin, *F (1)

(and as percent of indicated AT at 2,3 24 design full power) 4 8

(1) Protection Channel Safety Margin = AT - AT 3p, AT3p = AT, [1.135

.0114 (T,yg - T yg) +.00066 (2400 - 2250)]

3

= AT, [1.234

.0114 (T,,, - T,,9)]

7583Q:10/102384 II-21

WESTINGHOUSE NON-PROPRIETARY CLASS 3 TABLE II-4 TIME SEQUENCE OF EVENTS Time Accident Event (sec.)

Rod Withdrawal at Power 1.

Case A Initiation of uncontrolled RCCA withdrawal 0

at a high reactivity insertion rate (80 pcm/sec)

Power range high neutron flux high trip 2.0 point reached Rods begin to drop 2.5 Minimum DN8R occurs 3.5 2.

Case 8 Initiation of uncontrolled RCCA withdrawal 0

at a small reactivity insertion rate (1 pcm/sec)

Overtemperature al reactor trip signal 81.6 initiated Rods begin to drop 83.6 Minimum DN8R occurs 83.7 Steamline Break Steamline Ruptures 0.0 i

Pressurizer Empties 14.4 Criticality Attained 20.4 20,000 ppm boron reaches core 24.0 i

7583Q:10/100984 11-22

WESTINGH0USE NON-PROPRIETARY CLASS 3 TABLE II-4 (Continued)

TIME SEQUENCE OF EVENTS Time Accident Event (sec.)

Partial Loss of Forced Coastdown begins 0.0 Reactor Coolant Flow Low flow reactor trip 2.0 Rods begin to drop 3.0 Minimum DNBR occurs 3.6 Complete Loss of All operating pumps lose power 0.0 Forced Reactor and begin coasting down j

Coolant Flow Reactor coolant pump under-0.0 l

voltage trip point reached Rods begin to drop 1.5 Minimum DNBR occurs 3.0 Reactor Coolant Rotor on one pump locks 0.0 Pump Shaft Seizure Low flow trip point reached 0.1 (Locked Rotor)

Rods begin to drop 1.5 i

Maximum clad temperature occurs 2.9

• Maximum RCS pressure occurs 3.5 Dropped Rod Initiation of a rod drop (100 pcm) 0.0 Overtemperature delta-T 62.2 reactor trip signal initiated Rods begin to drop 64.2 l

7583Q:10/102284 I'.-23

WESTINGHOUSE NON-PROPRIETARY CLASS 3 TABLE II-4 (Continued)

TIME SEQUENCE OF EVENTS 8

Time Accident Event (sec.)

Loss of Normal Main feedwater flow stops 10.0 Feedwater Flow Low-low steam generator water level trip 61.3 Rods begin to drop 63.3 2

Auxiliary feedwater pump starts 121.3 Two steam generators begin 550.0 to receive auxiliary feedwater Peak water level in pressurizer occurs 2032.0 Core decay heat decreases to auxiliary 2124.0 feedwater heat removal capacity 1

75830:10/102284 II-24

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l WESTINGHOUSE NON-PROPRIETARY CLASS 3 III. SAFETY EVALUATION - LOCA A 1981 Westinghouse Evaluation Model ECCS Performance analysis for Indian Point 3 has demonstrated acceptable results for 24% uniform steam generator tube plugging. This docketed, NRC-approved analysis demonstrates a calculated 1

peak clad temperature (PCT) of 1995'F at a peaking f actor (FQ) value of J.14.

The amount of steam generator tube plugging modeled in this LOCA analysis far exceeds the actual tube plugging level which currently exists in Indian Point Unit 3 steam generators.

I Sensitivity studies performed in the past and documented in WCAP-8986 (Reference III-1) have demonstrated that the increase in calculated peak clad temperature with uniform steam generator tube plugging is linear for many different Westinghouse PWR designs (2, 3 and 4-loop plants). These sensitivities to tube plugging are for an equal amount of plugging in each steam generator, hence the term uniform plugging.

The increase in PCT observed with increasing steam generator tube plugging is j

primarily a consequence of the added resistance to fluid flow through the coolant loops during core reflood. Because the added resistance represents j

the predon.inant phenomenon associated with tube plugging and because of the linear nature of the PCT relationship, deviations in plugging from one steam generator to another do not significantly affect LOCA analysis results.

I The impact of asynenetric tube plugging upon calculated ECCS performance may be determined by a review of the equations which describe the system behavior during core reflooding. The WREFLOOD computer code is described in Reference 4

111-2, WCAP-8170. The WREFLOOD model, as shown in Figure !!!-1, represents the loops (lumped intact and broken) and the reactor vessel. As Figure III-2 indicates, the intact loops constitute a resistance network which connects core and downcomer regions. Resistance networks also model the broken loop piping. Nomenclature of Figure !!!-2 is as follows:

P is downcomer static pressure D

i P is core static pressure C

7583Q:10/100534 111-1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 P is c ntainment pressure X

K is the resistance loss coefficient Subscripts to K refer to loop (intact loop or broken loop) and location (hot leg, steam generator, etc.)

WREFLOOD is a quasi-steady-state code which models the venting of a core-generated steam-water mixture through the loops. The pertinent equations are presented below using the following additional nomenclature:

AP STUB is the pressure difference between vessel downcomer and containment p is the liquid density in the downcomer i

g is the gravitational constant AZ is the difference in water level between downcomer and core j

w is the mass flow rate through a reactor coolant loop p is the gas density through the loops G

A is the loop flow area; A is the total flow area in all loops V, is the core inlet velocity F

is the mass effluent fraction, the fraction of mass entering the out core which is expelled G

is the mass velocity at the core exit eore Consider loop behavior during the core reflood transient:

7583Q:10/092184

!!!-2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WREFLOOD equations state pressure relationships are

'O "

X+

' STUB P =PD + # EA C

L The driving force for intact loop flow is l

AP = P ~ 'O " # 80 C

L Sinoplify the loop equation of WCAP-8170 p. 2-2 by eliminating small magnitude terms:

1 i

this gives AP = it (Wit)2 I

K I

f r the intact loop 2

2 pG AIL

.Kyt (Wyt)2 then p(gAZ =

2 2 p6 IL 2# # g AZ

]1/2 G L and wgg = Agt (

g IL i

Thus at any particular point AZ during the core reflood process 4

"!L

• AIL I_K1_I1/2 IL I

a (frictional resistance) 1/2 wgt Apply the simplified equation to broken loop:

4

{

Kgg (wgL)2 l

A'8L " 2p A 2 g BL i

1 a

1 7583Q:10/100584 111-3

WESTINGHOUSE NON-PROPRIETARY CLASS 3 for the broken loop AP BL " C X"#9 STUB

~

L I"BL)2 K8L then ptg&Z + APSTUB "

2 2p A g gg 2p # gaZ + 2p 8P GL G STUB}1/2 gt = Agt [

g I

50 w BL A review of the IP3 limiting break (CD = 0.4 DECLG) reveals that APSTUB is small compared to p(gAZ until calculated clad temperature has increased to a value near PCT. Therefore, the AP term may be ignored STUB to obtain 2A # IA G L 1/2 E

3 I

"8L " ABL KBL a

"8L * ^BL I 1 )1/2 K BL From p. 2-6 of WCAP-8170, the loop flow boundary condition at the core is

• F Geore " YC L

out which may be written as i

"8L * "IL i

A C*#L+Fout "Y

C "8L * "Il Therefore VC"A # F C L out

.s 1

'7583Q:10/100584 III-4

=

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Core flooding rate V is determined by the ability to vent core-generated g

steam through the loops and is directly proportional to the sum (wgL +

wgL). IP3 exhibits its calculated PCT well into the core reflood portion of the LOCA transient: PCT is directly related to the magnitude of the flooding rate. The higher the value of V E*"d I"8L * "ILM er W C

calculated PCT in the IP31981 Model analysis.

The effect of tube plugging configurations upon total flow exiting the core (w,L + wgL) can now be assessed from the proportionality relationships.

For a 4-loop plant the total flow through the loops, w, is given as i

1 2

2 w,=wgg- + wSL

• I

)

+I I

i IL BL

, a.75 A, KgL

+. 2 5 A, K8L or w I

i in an original, unplugged state Kit,= Kat,-

2

, a 1.0 A, Kgg g,g w

When SG tube plugging is introduced, w, will be diminished due to an increase in frictional resistance. In the following presentation changes in I

resistance caused by SG tube plugging will be applied to the loss coefficient (K) term of the [A*K-1/2] expressions while A is held constant for ease of computation. Since no critical flow effects are involved the flow impact of SG tube plugging can be properly represented in this fashion. The uniform

)

plugging case and two bounding asy metric plugging cases are considered.

i j

1.

Uniform SG Tube Plugging Case 2

An added resistance (considered to be due to SG tube plugging) is introduced into each loop at IP3. Assume conservatively that the magnitude of the added resistance to flow is 105 of the original j

total loop resistance.

In the 24% uniform SG tube plugging WREFLOOD cases, the steam generator accounts for slightly more than 30% of the total IP3 loop resistance to flow.

4 75830:10/102384

!!!-5 i

WESTINGHOUSE NON-PROPRIETARY CLASS 3

, a.75A (K () 1 2 +.25A (KBL) w g

• 1.1) 1/2 +.25A, (K

• 1.1) 1/2 a.75A, (Kg w

gg r

, a 1.0A, (Kgg *1.1) a w,*1.1 w

i

, a.9535 w w

Total flow exiting the core is reduced by.0465 in this uniform resistance case.

II. All Plugging in Broken Loop The [4*(10% of individual loop resistance)] added resistance is placed totally into the broken loop in WREFLOOD. Then

~

w a.75A K

+.25A K gg 8L

, a.75A, (K ( ) 1/2 +.25A (1.4 KgL )

w g

g

, a (.75 +

4) A K 4

w T gg g w a (.75 +.2113) w,=

.9613 w Total flow exiting the core is reduced by.0387 in this case. The reduction in w and VC (aid the subsequent rise in PCT) is notably less for this configuration.

It is therefore less limiting than the uniform plugging configuration.

7583Q:1D/092184 III-6

WESTINGHOUSE NON-PROPRIETARY CLASS 3 III. No Plugging in Broken Loop j

None of the added resistance is placed into the broken loop. Thus 4(0.1) = 0.4 of the base loop resistance is added to the lumped intact loop, so its K value becomes 0.4/3 = 1.133 of its original value on a lumped basis.

j

, a.75A, Kgg T BL

+.25A K w

l

• 1.133) 1/2 +.25 A, K a.75 A, (Kg yto

, a [.75 (1.133) 1/2 +.25] w, w

w a.9545 w,

Total flow exiting the core is reduced by.0455. The reduction in

, and V is a bit less for this configuration so it is bounded w

by the uniform plugging configuration. Asymmetry presumed in steam generator tube plugging causes no adverse effects based on the WREFLOOD equations. However, the arguments presented here should only be applied to the established range of applicability in which WREFLOOD has been employed in Evaluation Model ECCS computations.

The indicated upper bound is a 30% steam generator tube plugging level in any one SG unit.

l The above discussion has demonstrated based upon the pertinent WREFLOOD equations that presumed asymmetry in steam generator tube plugging does not adversely impact calculated ECCS performance at a given plugging level.

Therefore, the existing 24% uniform tube plugging ECCS performance analyses for Indian Point 3 will support continued plant operation as long as:

1.

The number of tubes plugged in all four steam generators remains less than l

24% of the total number of SG tubes present in the plant.

i 2.

The number of tubes plugged ir. any one steam generator is less than 30% of the 3260 tubes present.

7583Q:10/100584

!!I-7

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 O

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 9

UESTINGHOUSE NON-PROPRIETARY Cl. ASS 3 St u

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)

IV. CONTROL AND PROTECTION SYSTEM SET POINTS I

I l

Only two changes are needed in the control and protection system instrumentation to account for asynnetric tube plugging:

1.

The T program in the reactor control system must be adjusted, if t

necessary, such that no cold leg temperature exceeds its allowable value. This adjustment preserves core DNB margin during steady-state operation.

2.

The overtemperature delta-T reactor trip channels must be calibrated during power operation in terms of both the delta-T and T,yg indicated by each channel at nominal full power. This calibration preserves the ability of the reactor protection system to prevent exceeding the core safety limits in the presence of asymmetry in loop temperatures.

These adjustments, plus total reactor coolant flow of no less than 323,600 gpm, are consistent with the assumptions for this analysis. The appropriate control system adjustment is discussed below.

With normal (symmetric) plant operation, plant control and protection limits are based on average vessel temperature, T,yg.

With asymmetric tube plugging; however, inlet temperature, Tin, is a m re appropriate limit, and must not exceed the value assumed in the analyses. Since fluid from any loop cold leg is conservatively assumed to enter the core hot channel without benefit of mixing with cooler fluid from other loops, the temperature limits apply to each loop.

The analyses in this report assumed an inlet temperature of 546.9'F, 4*F above the nominal design value of 542.9'F.

This 4*F allowance includes 2'F for control deadband and 2*F for temperature error. Thus, during steady-state operation with loop asynsnetries, measured T,yg should not exceed 576.7'F, and T should not exceed 544.9'F, for any loop. The precaution below in should be included with appropriate station procedures.

75830:10/100984 IV-1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 "During steady-state operation with either automatic or manual control, the measured average coolant temperature (T,yg) in each and every loop must be no greater than the programmed T,yg at full power (574.7'F) plus 2*F for control deadband. In addition, if steady-state loop-to-loop asymmetries exist in both reactor coolant flow and temperature (e.g., because of non-uniform steam generator tube plugging), then during steady-state full power operation, the measured cold leg temperature in each and every loop must not exceed the j

thermal design value (542.9'F) plus 2*F for control deadband. Reduction in limit."

the T,yg program must be made if necessary to preserve this Tinlet l

7583Q:10/100584 IV-2

WESTINGHOUSE NON-PROPRIETARY CLASS 3 V.

CONCLUSIONS The impact of the tube plugging asymetry has been evaluated for the Indian Point 3 FSAR Chapter 14 analyses, and has been shown to satisfy all safety criteria.

The steps that must be taken to ensure there will be no safety criteria violations are the following:

1.

The reactor vessel flow must be equal to or greater than 323,600 gpm.

2.

During steady state operation at full power, the hottest cold leg inlet temperature must not exceed 542.9 'F plus 2*F for control deadband.

3.

The overtemperature delta-T reactor trip channels must be calibrated during power operation in terms of both delta-T and T,,g indicated by each channel at nominal full power.

With the above restrictions applied, most transients will be bounded by the existing 24% uniform tube plugging analysis. There are several transients such as locked rotor and partial loss of flow for which an asymetric flow distribution is coupled with an asymetric tube plugging distribution.

Reanalyses of these transients, however, have demonstrated satisfactory results, even when the most adverse plugging / flow combination was assumed, i.e., when the faulted loop was assumed to be the loop with the least tubes plugged. Thus, it has been demonstrated by this study and the previous 24%

uniform tube plugging study (Reference I-1) that all Indian Point 3 FSAR Chapter 14 safety criteria have been satisfied.

With respect to the effects on the Indian Point 3 LOCA analysis, loop to loop asymmetry in steam generator tube plugging does not adversely impact calculated ECCS performance. The approved 24% uniform steam genera' tor tube plugging ECCS performance analysis for Indian Point 3 remains a valid basis for plant operation as long as the plugging level in any one steam generator does not exceed 30% and the total number of tubes plugged in all four steam generators remains less than 24% of the total number of tubes present in the plant.

7583Q:10/101884 V-1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 REFERENCES I-1 Indian Point Unit 3, 24 Percent Uniform Tube Plugging Analysis Letters J. P. Bayne (NYPA) to S. A. Varga (NRC)

IPN-83-5, January 13, 1983 i

IPN-83-37, May 5, 1983 IPN-83-101, December 14, 1983 I-2 Final Safety Analysis Report, Indian Point Unit 3. Docket Number 50-286.

11-1 Burnett, T.W.T., et al, 'LOFTRAN Code Description," WCAP-7907-P-A (Proprietary Class 2), WCAP-7907-A (Proprietary Class 3) April,1984.

II-2 Van Houten, R., " Fuel Rod Failures as a Consequence of Departure from Nucleate Boiling or Dryout," NUREG-0562, June 1979.

111-1 Thompson, C. M., and Esposito, V.

J., " Perturbation Technique for Calculating ECCS Cooling Performance," WCAP-8986, February, 1977.

III-2 Collier, G. et al., " Calculational Model for Core Reflooding After a i

Loss of Coolant Accident (M REFLOOD Code)," WCAP-8170 (Proprietary Class 2), WCAP-8171 (Proprietary Class 3), June,1974.

e 4

4 7583Q:10/101884 VI-1

_ _, _