ML20141N888
| ML20141N888 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 01/31/1986 |
| From: | Forcht K, Kemper R, Segletes J WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML093440582 | List: |
| References | |
| IPN-86-14, WCAP-10705, WCAP-10705-R01, WCAP-10705-R1, NUDOCS 8603180248 | |
| Download: ML20141N888 (76) | |
Text
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WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-10705 REVISION 1
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SAFETY EVALUATION FOR INDIAN POINT UNIT 3 WITH ASYMMETRIC TUBE PLUGGING AMONG STEAM GENERATORS JANUARY, 1986 o
WORK PERFORMED FOR THE NEW YORK POWER AUTHORITY CONTR18UTORS:
J. A. Segletes R. M. Kemper K. A. Forcht T. W. T. Burnett O. P. Dominicis Westinghouse Electric Corporation Nuclear Technology Division P. O. Box 355 Pittsburgh, Pennsylvania 15230 860318024e g60314 PDR ADOCM 05000286 15830:10/100884
WESTINGHOUSE NON-PROPRIETARY CLASS 3
- ?
TABLE OF CONTENTS
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Pggi I
INTRODUCTION I-1
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11 SAFETY EVALUATION - NON-LOCA 11-1 A. Asymmetric Tube Plugging Assumptions
- 8. Computer Code Capabilities / Asymmetric Tube Plugging C. Initial Temperatures and Flow Distributions O. Transients Reanalyzed l.
Rod Withdrawal at Power 2.
Steamline Break 3.
Loss of Flow 4.
Locked Rotor 5.
Oropped Rod
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6.
Loss of Normal Feedwater 7
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E. Transients Not Reanalyzed l
F. Delta-T Response Evaluation III SAFETY EVALUATION - LOCA 111-1 IV CONTROL AND PROTECTION SYSTEN SETPOINTS IV-1 V
CONCLUSIONS V-1 VI REFERENCES VI-1 1
APPENDIX A: Additional Locked Rotor Analysis A-1
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1583Q:10/010686
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WESTINGHOUSE NON-PROPRIETARY CLASS 3 LIST OF TABLES 11-1 Sununary of Reactor Coolant Loop Flows and Temperatures at Normal Design Full Power 11-2 Sunenary of Reactor Coolant Loop Flows and Temperatures at Worst Steady State Operation 11-3 Reactor Coolant Temperature at Reference Overpower Core Limit 11-4 Time Sequence of Events A-1 Time Sequence of Events e
1583Q:10/010686
r WESTINGHOUSE MON-PROPRIETARY CLASS 3 UST OF FIGutES 4
11-1 Red WitWrawal at Power at 80 pcm/sec
- Nuclear Power
- Core Meat Flus C
11-2 Rod Withdrawal'at Power at 90 pcm/sec
- Pressurite? Pressure
- Pressurizer Water Volume 1
11-3 Rod withdrawal at Power,at 80 pcm/she '
.\\
- Loop.7 T,
- DNSA
!!-4 Red Withdrawal at Power at 1 pcm/sec
- Nuclear Power
- Core Meat Flux
!!-5 Red Withdrawal at Powr at 1 pcm/sec
- Pressurizer Pressure i
- Pressurizer Water Volume
!!-6 Rod Withdrawal at Powr at 1 pcm/see t
t
- Loop 2 1,,,
3
- D'#SR
.~N
!!-7 Rod Withdrawal at Power
- Minimum DNSR vs. Reactivity Insertion Rate 11-4 5 team Line treak
- Nuclear Power
- Core Heat Flux N
- RC5 Pressure
- Pressurizer Water Volume
- l
!!-9 Steam Line treak I
- Feedwater Flow j
- Steam Flow
- Core Flow i
15830:10/101884
WESTINGHOUSE NON-PROPRIETARY CLASS 3 LIST OF FIGURES (Cont.)
11-10 Steam Line Sreak
- Core Average Temperature
- Reactivity
- Core boron
- Steam Pressure 11-11 Partial Loss of Flow
- Nuclear Power
- Core Meat Flux II-12 Partial Loss of Flow
- Pressurizer Pressure
- Core Flow II-13 Partial Loss of Flow
- DNSR 11-14 Complete Loss of Flow
- Nuclear Power
- Core Heat Flux 11-15 Complete Loss of Flow
- Pressurizer Pressure
- Core Flow 11-16 Complete Loss of Flow s
j
- DNSR
!!-17 Locked Rotor
- Nuclear Power
- Core Heat Flur 11-18 Locked Rotor
- RCS Pressure
- Core Flow II-19 Locked Rotor l
- Clad Inner Temperature 11-20 tropped Rod
- Nuclear Power
- Core Neat Flux II-21 Dropped Rod f-
- Average Coolant Temperature
- Pressurizer Pressure i
j 75430:10/101884
WESTINGHOUSE NON-PROPRIETARY CLASS 3 LIST OF FIGURES (Cont.)
11-22 Oropped 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 A-1 Locked Rotor
- Nuclear Power
- Core Heat Flux A-2 Locked Rotor
- RCS Pressure
- Core Flow A-3 Locked Rotor
- Max Clad Inner Temperature e
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i 7583Q:10/010686 f
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WESTINGHOUSE NON-PROPRIETARY CLASS 3
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SAFETY EVALUATION FOR INDIAN POINT UNIT #3 WITH A5YIU4ETRIC Tutt PLUGGING AMONG STEAM GENERATOR $
I.
INTRODUCTION 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 245. Approval of this Tech Spec was granted by the NAC on January 13, 1994. Safety analyses (Reference 1-1) were performed for a 245 uniform tube plugging level in support of this Tech Spec change. The results of the uniform 245 tube plugging evaluation for the FSAR (Reference I-2)
Chapter 14 transients, other than LOCA, showed that the reactor coolant flow rate and reactor vessel average temperature are the only safety related parameters that are significantly affacted by tube plugging. The implication of the 245 uniform tube plugging ana'1ysis was that one or more of the four 4
steam generators (SG) could have greater than 245 tube plugging if the others had less than 245 tube plugging provided that the reactor coolant flow rate remained greater than that predicted for 245 uniform tube plugging and the hottest cold leg temperature remained less than the T, predicted for 245 g
uniform tube plugging.
For the [CCS performance analysis, the overall loop resistance to flow, rather than the relative resistance among loops, is of primary importance.
Currently the equivalent tube plugging levels
- in the four steam generators are 16.05, g.85, 8.45 and 8.01.
While these levels are significantly less than 245, they represent a degree of asyuumetry. 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 /54 are as follows:
Plugged Tube 54 Tubes Plugged Tubes sleeved Equivalent (5) 31 483 768 16.0 k
32 286 651 g.t l*
33 231 850 8,4 34 227 701 0.0 75830:10/100ge4 t-1 l
WESTINGHOUSE NON-PROPRIETARY CLASS 3 in any of the steam g:nsrat:rs 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 ef fects of asymmetry are the resulting dif ferences 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,,g, and STs. 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 l
1583Q:10/010686 1-2
WESTINGHOUSE NON-PROPRIETARY CLASS 3 j
!!. SAFETY EVALUATION - NON LOCA A.
Asysumetric Tube Plugging Assumptions The following assumptions were made in the reanalysis of the non-LOCA transients:
(1) Tube Plugging 01stribution The tube plugging distribution analyzed is 305, 245, 245 and 85.
This distribution encompasses the upper plugging level assumed in the study (305), the lowest possible level based on the current plugging levels
-(0.05) and two equal intermediate levels that play a role in simulating the overtemperature Delta-T reactor trip setpoint equation.
(2) Loop F1cws Although loop flows will be asynnetric, the sum of all loop flows are assumed to equal the reactor vessel flow of 323.600 gpm. This reactor vessel flow equals the thermal design flow established for 245 uniform tube plugging analysis in Reference I-1.
The flow for each loop was calculated based on the reactor coolant pamp 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.g'F at nominal full power steady state conditions. This temperature equals the inlet temperature predicted for l
245 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.g*F to -
account for 4*F control and measurement uncertainty.
l 75830:10/100584
!!-1
~_ -. -.
WESTINGHOUSE NON-PROPRIETARY CLASS 3
(, 4) DN8/ Inlet and Core Temperatures The departure from nucleate boiling ratio (DNSR) is computed conservatively for the core hot channel assuming that fluid f rom the hottest cold leg flows directly into that chanr.el 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.
i I
(6) Reactor Vessel Fluid Mixing Loop-to-loop temperature asynnetries depend upon the degree of mixing in the reactor vesselt 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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8.
" Computer Code Capabilities /Asynnetric Tube Plugging The asynnetric of fects for non-LOCA accident analyses were computed using the i
LOFTRAN computer code (Reference Il-1). This code has the capability to permit the input of pressure drop ratios (loop / loop average) and SG heat transfer area ratios (loop / loop average) for eaa.h loop. This information, in conjunction with the plant data, allows the calculation of initial (steady state) conditions and subsequent transient responses to imposed accident conditions.
l C.
Initialization of Flows and Temperatures and Effact 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 spa) and the hottest inlet temperature agreed with the desired value (542.9 without P
l uncertainty or 546.9'F with' uncertainty).
75830:10/100884 II-2 i
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---,,n n,_
,-.,-,.w-,-,-..
w--.--r---,mn-e->mw---nc-
---m-~~---
WESTINGHOUSE NON-PROPRIETARY CLASS 3 The calculations were made for three reactor vessel 1.1et plenum chamber mixing assumptions: (1) perfect (i.e. complete) mixing (2) minimum mixing, and (3) intermediate mixing.
[
aj.c The results of the initiaTization computations are presented in Table !!-1 for nominal conditions (3025 MWt, 323600 spa, 2250 psia, 542.9'F Tinlet) and in Table 11-2 for worst steady-state operation (102% x 3025 phet, 323600 spa, 2220 psia, 546.9'F Tinlet). The conditions shown in Table 11-1 are used to set the overtemperature delta-T constants. Table !!-2 values are used as the initial conditions for the accident analyses.
D.
Transients Reanalyzed 1.
Rod Withdrawal At power (IP3 FSAR Section 14.1.2)
For the majority of plant transients (those for which the core DNS limits are applicable, have pecking factors no worse than design, have core flow no less than design, and proceed slowly enough for response of loop temperatures), core DNO protection is provided, if necessary, by the overtemperature delta-T trip.
i The nominal overtemperature delta-T reactor trip setpoint equation (excluding dynamic compensation and setpoint reduction for adverse core axial power distribution) is given by the formula below:
Y l
75g30:10/102384 II-3
WESTINGHOUSE NON-PROPRIETARY CLASS 3 SP = AT, [1.135
.0114 (T,,, - T' g) +.00066 (P - 2250)],
AT Where AT, is indicated AT
'F. at nominal full power for the channel being calibrated:
l
,1 is indicated average temperature. *F, for the channel.being calibrated; T*,, is indicated average temperature,
'F, at nominal full power for the channel being calibrated; and P is pressurizer pressure.
l Thus, each protection channel is calibrat2d in terms of the individual temperatures AT, and T*,, that exist at nominal full power in 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 asyumetries, and to illustrate the way in which the protection system prevents the core from exceeding the DNSR limit.
d Method of Analysis The' rod withdrawal at power accident analysis was simulated using the LOFTRAN computer code.
The recommended AT calibration is accounted for in this analysis. This includes the offect of each channel being calibrated at power based on the indicated AT and T,,
for that channel at normal fall power. (Refer to Table !!-1 for va ues of AT and T,,, for different loops at full power with different mixing assumptions.)
i I
Except as noted in Section !!.C and above, analysis methods and i
assumptions are consistent with the original FSAR assumptions:
b 15430:10/101484
!!-4 i
O
__m__--,
m
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 !!.C. for both steady state and transient conditions,
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11ggdv State Analysis and Results t
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!!-5
WESTINGHOUSE NON-PROPRIETARY CLASS 3
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Transient Analysis and Results For transient verification of the adequacy of the overtemperature delta-T trip. a rod withdrawl analysis for 3 pcm/sec reactivity insertion rate was performed for each of the three mixing assumptions. A rate of 3 pcm/see was selected as approximately the worst rate (lowest minimum DNS I
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
!!-4. Nuclear power, core heat ficx, pressurizer pressure, c
75830:10/101884
!!-6
l WESTINGHOUSE NON-PROPRIETARY CLASS 3 l
pressurizer water volume, channel 2 vessel average temperature and DN8R are presented in Figures 11-1. II-2, and !!-3 for the 80 pcm/sec reactivity insertion and in Figures !!-4, !!-5, and !!-6 for the 1 pcm/sec reactivity insertion.
The SNSR as a function of reactivity insertion rate is presented in Figure 11-1 for asynunetric tube plugging over a wide range of insertion rates.
Conclusions
- 1) [
aj.c
- 2) Because of the way in which the delta-T reactor trips are calibrated at power, loop-to-loop temperature and flow asynenetries do not reduce core safety margins during the rod withdrawal at power transient.
- 3) Minimum DNSR 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 l
if the postulated break is in the steamline associated with the least plugged steam generator. 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 inside containment break at zero power with of fsite power 75830:10/101844
!!-7
.. -. ~
WESTINGHOUSE NON-PROPRIETARY CLASS 3 available. A double ended rupture was assumed to occur in Loop 4 (plugging level = 05) 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 differentia 1' 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 305, 245, 245, 85 plugging dist'ribution in Loops 1 through 4 respectively.
Except as noted above and accounting for asynnetries, the assumptions are consistent with the FSAR assumptions, i.e., shutdown margin is equal to 1.725, initial temperature is equal to no load temperature and initial pressure is equal to 2250 psia.
4 Results The time sequence of events for the steamline break transient is given in Table !!-4.
Nuclear power, core heat flux, RC5 pressure, pressurizer water volume, feedwater flow, steam flow, core flow, core average temperature, reactivity, core boren and steam pressure are presented in Figures 11-0 through II-10 for the steamline break analysis. A detailed independent DN8R analysis was performed based on state points computed by LOFTRA4. The ONGR for this case was determined to be greater than the 11m'it value of 1.30.,,
Conclusions A 305, 245, 245, 85 plugging distribution during a worst case steamline break Will not result in a ONGR that is lower than the allowable limit.
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l 3.
Loss of Flow (IP3 FSAR Section 14.1.6) l i
The 1/4 and 4/4 loss of flow transients were recalculated to demonstrate the DNGR criteria (DNBR 11.30) can be met. These analyses were based on revised reactor coolant pump data that provide more conservatism than in the original FSAR.
4: :
75830:10/102384
!!-8 i
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 305, 245, 245, 85 plugging distribution. For the 1/4 case, the coastdown was assumed to occur in Loop 4 (85 plugging) because a coastdown i
in the least plugged loop will result in the lowest core flev.
syneetric plugging parameters were input to LOFTRAN to provide a 305, 245, 245, et tube plugging distribution in Loops 1 through 4 respectively. The reactor coolant pump dynamic rerponse was computed by using pump data which simulated 4/4 loop corstdown 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 1g76. This best-estimate analysis assumed as-built reactor coolant pump performance, best-estimate f1ws and pressure drops, and design pump Inertle. The best-estimate calculated flow coastdown was found to be in excellent agreement with actual plant startup measurements, as indicated below:
I Core Flow. Fraction of Initial Time after loss test-Estimate Plant of ouac nower see Calculation M.easurement*
a.c 1.5 3.5 5.5 7.5 g.5
- corrected for 0.50-second flow sensor delay.
then, conservatism was added to provide margin in the safety analyses by t
a) assuming only g05 of the pump design inertia; and b) increasing the l
75830:10/100884
!!-g
. _ _ - - _ = - _
E STINGHOUSE NON-PROPRIETARY CLASS 3 loop pressure drops such that total vessel flow was reduced to its thermal design value of 323,600 gem, with the loop-to-loop asyuumetry shown in I
Table !!-2. The flow coastdown at full power is shown in Figure !!-15.
i Except as noted in Section !!.C and above, the analysis assumptions are consistent with the FSAA assumptions, i.e., initial power = 1.02 x.
nominal,' initial pressure includes a -30 pst uncertainty, and beginning of core life reactivity coefficients.
The less of flow transients were computed using the LOFTRAN, FACtRAN and THINC computer codes, tasults Nuclear power, core heat flux, pressurizer pressure, core flow and DNGR are presented in Figures !!-11 through !!-13 for the 1/4 loss of flow case and in Figures !!-14 through !!-16 for the 4/4 less of flow case. The Dutt for these cases was determined to have met the alloweble value of 1,30.
Conclusions i
A 305, 245, 245, 85 tube plugging distribution during a worst case 1/4 I
loss of flow or 4/4 less of f1w will not result in a DNOR that is lower
~
thanthealloweilelimit. Asponetric tube plugging will have essentially no effect on a 4/4 LOF transient and only a slightly adverse effect for a loss of flow in the loop containing SE S4 tube plugging. In the complete less of flow case the core flow will remain essentia11y the same as with uniform tube plugging and in the partist loss of flow case the core flow I
will be slightly reduced when compared with uniform plugging.
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i 4.
Locked Rotor (!P3 FSAR 5ection 14.1.6)
The 245 uniform tube alugging analysis will bound the aspunetric case if the average flow in the three lowest flow loops meets the thermal design L
i flow for three 245 uniform tube plugged loops (3 x 80.g00 = 242,700 gon).
To demonstrate, however, that this ' criterion is not necessary, the locked rotor accident was reenalyzed.
15630:10/101864
!!-10
WESTINGHOUSE NON-PROPRIETARY Ct. ASS 3 Method of Analysis An analysis was performed assuming the locked rotor is in Loop a (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 305, 245, 245, 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 aialytical 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 ONtt analysis was not necessary since rods in DNS are predicted to not fail based on the temperature-time criteria presented in Reference !!-2.
An additional Locked Rotor Analysis is presented in Appendix A.
Results I
The time sequence of events for the locked rotor transient is given in Table !!-4.
Nuclear power, core heat flux, RCS pressure, core flow and clad inner temperature are shown in Figures !!-17 through !!-19. 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%, 244, 85 tube plugging distribution during a locked rotor event will not result in exceeding any safety limits.
7 75e30:lo/0106e6 t!-11
WESTINGHOUSE NON-PROPRIETARY CLASS 3 5.
Dropped Rod (!p3 FSAR 5ee. tion 14.1.3)
The dropped rod accident may result in a reactor trip on OT&T. 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 reenalyzed for asymmetric tube plugging.
Method of Analysis The analysis was performed for a 305, 245, 245, es 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 het channel factor (F N).
drop is assumed to occur from 1025 power with conservative feedback properties.
Results The time sequence of events for the dropped rod transient is shown in Table !!-4.
Figures !!-20 through !!-22 illustrate the transient response I
following a dropped rod of worth 100 pcm. The reactor coolant average temperature decreases initially, $we 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-1 reactor trip occurs to terminate the transient. A DNS evaluation at the limiting condition in the transient shows that the DNSR remains above 1.30.
j Conclusions A 305, 245, 245, 35 tube plugging distribution during a dropped rod event will not result in a minimum DNeR lower than the allowable limit.
)
i 75430:10/1023e4
!!-12
-,,w-,-
-nen,-,,,--~~m
w
,--,,-,-e--
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WESTINGHOUSE NON-PROPRIETARY CLASS 3 6.
Loss of Normal Feedwater Flow (IP3 FSAR Section 14.1.g)
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 305 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 feedwater 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 wil before the steam generator heat transfer capability is reduced, the primary system variables never approach a DNS 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.
i*
' Art analysis of the system transient is performed to show that following a loss of normal feedwater and with reduced primary to secondary heat transfer due to steam generator tube plugging, the auxiliary feedwater system is capable of removing the stored and residual heat, thus preventing either overpressurization of the RCS or loss of water f rom the reactor core, and returning the plant to a safe condition.
'M th d of Analysis e o A detailed analysis using the LOFTRAM code is performed in order to obtain the plant transient following a loss of normal feedwater.
Assumptions made in the analysis are:
7583Q:10/100g84
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,-,-..n---,-..,-,--n,
.,.,--a-
-, - e
~w.--.,,-.--.,-n,_,
.--,-,en.,-----w_--nnn-,
WESTINGHOUSE NON-PROPRIETARY CLASS 3 A.
The plant is initially operating at 102 percent of 3216 feft (the maximum calculated turbine rating).
.l S.
A conservative core residual heat generation based upon'.long t'ers
~
operation at the initial power level preceding the trip.
a' C.
Reactor trip occurs on steam generator low-low level.
D.
Only one auxiliary feedwater pump with a capacity of 400 spa 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 305 of their tubes plugged.
F.
Secondary system steam relief is achieved through the steam generator power-operated r'elief 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 pst higher than l
nominal.
/
H.
Reactor coclant pump coastdown was assumed to occur af ter react'or 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.
i Results L
i Figures 11-23 and !!-24 show the significant plant parameters following a loss of t.ormal feedwater without of f site power available.
The capacity of one auxiliary feedwater pump is such that the water level
~
- n the steam generators being fed does not recede below the lowest level 75830:10/102284 11-14 e
---,-n.,-----
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r--r-w,-----,
---,-,-n-,,--
--._-r,-
WESTINGHOUSE NON-PROPRIETARY CLASS 3 l
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at which sufficient heat transfer area is available to dissipate core residual heat without water relief from the pressurizer safety valves.
Figure II-23 shows that at no time is there water relief from the pressurizer.
The cakulated sequence of events for this accident is listed in Table
!!-4.
As shown in Figures !!-23 and II-24, the plant approaches a stabilized condition following r'esctor trip and auxiliary feedwater initiation.
2 Conclusions Results of the analysis show that a loss of normal feedwater does not-adversely affact 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 1.
Rod Withdrawal From Suberitical (IP3 FSAR Section 14.1.1)
The asynnetric plugging case is bounded by the 245 uniform plugging case provided that the DNS margins and operating limits are maintained. These' requirements are ensured by maintaining the same reactor flow as in the 24% uniform plugging analysis (323,600 spm) and maintaining the same reactor coolant temperature at zero power.
2.
Soron 011ution (IP3 F5AR 5ection 14.1.5)
Boron dilution during shutdown will not be affected by plugging asymmetries. Teid1 RC5 volume, which is the plant parameter of concern in I
this event, will be no less for asymmetric tube plugging than it is for t
the 245 unifora plugging case.
l toren dilution at power with asymmetric tube pluggins is no worse than for l,.
245 uniform plugging since the RC5 volume will be no less for asynnetric f
15830:10/100884
!!-15
WESTINGHOUSE NON-PROPRIETARY CLASS 3 plugging than it was for 245 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.O.1.
3.
Startup of an Inactive Loop (IP3 FSAR 5ection 14.1.7)
Startup of' an Inactive Loop could be slightly affected by asynnetric tube plugging. Since N-1 operation at Indian Point Unit 3 is not licensed, this event was not reenalyzed.
4.
Loss of External Electrical Load (!P3 FSAR Section 14.1.8)
The asymmetric plugging, case is bounded by the 245 uniform plugging case provided DNS margins and operating limits are maintained.
5.
Excessive Heat Removal Oue to Feedwater System Malfunctions (IP3 FSAR 5ection 14.1.10)
This event will not be adversely affacted by asymmetric tube plugging. No reactor trip on overtemperature delta-T is needed because, as noted in the FSAR, the DNSR increases following initiaticn of the event.
- 6., Excessive Load increase (IP3 FSAR 5ection 14.1.11)
This event wil1 not be adversely affected by asymmetric tube plugging. No reactor trip on overtemperature delta-T is needed because, as noted in the FSAR the DNOR increases following initiation of this event.
7.
Rod Ejection (IP3 FSAR Section 14.2.6)
The asymmetric tube plugging case is bounded by the 245 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 245 uniform tube plugging analysis.
7503Q:10/100884
!1-16
2mi 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.
'j With asyuunetric tube plugging, natural circulation flow rates will be slightly different between the loops. However, this is neither a. large
(
nor a s'ignificant effect. 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 y
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 5
pump impedance is much greater than the impedance through steam generater tubes, even with 305 tube plugging. Thus, the difference between two 3
loops, one with 85 plugged tubes and one with 305 plugged tubes, is less than 105 in relative flow. In any event, natural circulation flow is low l
enough that all cold leg temperatures are essentially equal to the temperature on the stcondary side of the steam generator. Therefore, no significant loop temperature assymmetry is expected during natural F
circulation as a result of asynenetric tube plugging.
9.
Oropped Bank (IP 3 FSIR Section 14.1.4) 5 The asynenetric plugging case is bounded by the 245 uniform plugging case provided the DN8 margins and operating limits are maintained. These 2
requirements are ensured by maintaining conservative reactor coolant flows and vessel inlet temperatures relative to the 245 uniform tube plugging
]
analysis. Unlike the dropped rod case, a reactor trip on overtemperature y
delta T will not occur and the rod worth inserted will be uniformiy 5
distributed within the core for the dropped bank case. Therefore, the concerns due to tube plugging asynenetries that existed for the dropped rod d
case do not exist for the dropped bank case.
d O
75830:10/102384
!!-17 i-,-..-
ii=
. i.
l 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 offacts of asymmetric tube plugging. The results show that het leg temperature response is relatively insensitive to tube plugging.
Fluid transit time from the reactor vessel outlet plenum to the het leg temperature sensor is composed of two components (1) transit time from the vessel outlet plenum to the het leg scoops and (2) transit time from the het les scoops to the RTO manifold. Fluid transit time from the vessel outlet plenum to the het leg scoops (less than 0.5 seconds at design flow) is inversely proportint.al to flow, and will increase as the number of plugged tubes increases. However, flow thrnugh the het leg RTO bypass (from hot leg scoop to RTO manifold) increases as the pressure drop across the steam generator increases. Therefore, fluid transit time from the het leg scoops to the RTO 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 505.
tened on FSAR assumptions on time delays, a 1.5 second delay time from vessel to hot les mentfold would be acceptable. This delay time would not be teached unless the plugging level in a steam generator substantially exceeded 505.
Based on the above evaluation, it is concluded thatt a.
Tube plugging up to a 505 level has virtually no offect on the AT response and would be acceptable up to approximately 755 plugging, b.
Y,,, and AT filters and rate compensation need not be reset as a result of tube plugging.
7583010/1023g4
!!-It
WESTINGHOUSE NON-PROPRIETARY CLASS 3 Table !!-1 Summary of Reactor Coolant Loop Flows and Temperatures at Nominal Design Full Power (3025 left. 323 H O gem. 2250 psia. 542.g*F T I
inlet v
Pct Perfect Intermed.
Minimum Parameter Lggg PlumnME B.11.1AR 51118g minina
~
~
Normalized Flow.
1 30 Fraction of 80.g00 gen 2.3 24 4
8 Vessel Inlet 1
30 Temp. (*F) 2.3 24 (542.g* for uniform 4
8 245 plugging)
Vessel Average 1
30 Temp (*F) 2.3 24 (574.7* for uniform 4
e 245 plugging)
Vessel Delta-T (*F) 1 30 (63.7 for unifom 2.3 24 245 plugging) 4 8
75030110/101884 13.)g
idESTINGHOUSr. NON-PROPA!ETARY CLASS 3 Table !!-2 Swanary of Reacter Coolant Loop Flows and Temperatures at tierst Steady-State Operetten (1025 a 3025 phet 323600 gem. 2220 psia, 546.t*F Tinlet)
Pct Perfect Interised.
Minimum Paramatar Legg Pluatina 311133 gigig Minina
~
~
periaaltmed Flow, 1
30 a.c Fraction of 90.900 gen 2.3 24 4
e vessel Inlet 1
30 Temp, including 2.3 24 4'F uncertainty (*F) 4 8
Vessel Average 1
30 Tese ('F) 2.3 24 4
8 Vessel Delta-T (*F) 1 30 2,3 24 4
8 O
l 4
s e
i 3
e 15030:10/102364
!!-20 9
WESTINGHOUSE NON-PROPRIETARY CLASS 3 TASLE !!-3 REACTOR COOLANT TEMPERATURE AT REFERENCE OVERPOWER CORE LIMIT
[(1105 power, 2400 psia, 555.2*F T II inlet Man-uniform (30-24-H4-85) aluasine Plugging uniform Perfect Intermed ate Minimum Lega Lays.]_,11), 245 Pluatina ghing.
Minine 5,13,133.
~
Vessel Inlet Temp, *F 1
30 ac 2,3 24 4
8 l
Vessel Average Temp, 1
30 CF (and change f rom design full power) 2,3 24 4
O Vessel Delta-T, 'F 1
30 (and, as percent of indicated AT at design full power) 2,3 24 i*
4 8
Protection Channel 1
30 Safety Margin, 'F (1)
(and as percent of indicated AT at 2,3 24 design full power) 4 e
(1) Protection Channel Safety Margin = AT - AT,,
3 AT, = AT, [1.135
.0114 (T,,, - T',,) +.00066 (2400 - 2250)]
3
= AT,[1.234
.0114 (7,,, - T,',,)]
75030:10/102384
!!-21
idESTINGH00SE NON-PROPRIETARY CLASS 3 TASLt 11-4 TIME $tQUENCE OF EVENTS O
Time Actidan.t LYtal 113Ld Red withdrawal at Power 1.
Case A Initiatten of uncontrolled RCCA withdrawel
0 at a high reactivity insertion rate (SG pcm/sec)
Power range high neutron flus high trip 2.0 point reached Reds begin to drop r
2.5 Minimum 0#9R occurs 3.5 2.
Case 8 Initiation of uncontrolled eCCA with$rawal 0
at a small reactivity insertion rat'e (1 pce/sec)
Overtemperature AT reactor trip signal 81.6 initiated Rods begin to drop 83.6 Minimum DWOR occurs 83.1 Steseline treak Steseline Ruptures 0.0 Pressuriser taptles 14.4 Critica11ty Attained 20.4
~1 20,000 ppe boron reaches core 24.0 f
l t
l I
1 75430:10/100904
!!-22
(
WESTINGHOUSE NON-PROPRIETARY CLASS 3 TA8LE 11-4 (Continued)
TIME SEQUENCE OF EVENTS s
Time Accident
[vg.n1,
($gc.,)
n Partial Loss of Forced Coastdown begins 0.0 Reactor Cociant Flow Low flow reactor trip 2.0 Rods begin to drop 3.0 Minimum DN8R occurs 3.6 Complete Loss of All operating pumps lose power 0.0 Forced Reactor and begin coasting down Coolant Flow Reactor coolant pump under-0.0 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 1.1 i
Maximum RCS pressure occurs 3.5 Maximum clad temperature occurs 3.8 Dropped Rod g
Initiation of a rod drop (100 pcm) 0.0 Overtemperature delta-T 83.1 Reactor trip' signal initiated Rods begin to drop 85.1 1
8893Q:10/010986
!!-23
l
.\\
WESTINGHOUSE NON-PROPRIETARY CLASS 3 l
TA4LE 11-4 (Continued) l i
i*
TIME SEQUENCE OF EVENTS Time Accident (g33,1 M
t I
Loss of Norinal Main feedwater flow stops 10.0 Feeduster Flow Low-low steam generator water level trip 61.3
- Rods begin to drop 63.3
-s Auxiliary feedwater pump starts 121.3 Two steam generators begin 550.0 to receive auxiliary feedwater 9
Peak water level in pressuciter occurs 2032.0 Core decay heat decreases to auxiliary 2124.0.
feedwater heat removal capacity I
9 i
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l 1-1 l
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l 75430:10/102284 11-24
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Figure II-2
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(
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2 Figure 11-8
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Asymmetric Tube Plugging Analysts Steam Line Rupture t
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Figure
!!-10
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M TIMC (SECS Indian Point Unit 3 Asymmetric Tube Plugging Analysis Partial Loss of Flow (1/4)
Figure
!!-11
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!!.13
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Figure.!!-14
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! !.15 O
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Indian Point Unit 3 Asymmetric Tube Plugging Analysis Locked Rotor I
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!!-18
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!!-19 l
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DROPPED RCCA 0F WORTH 100 PCM FIGURE !!.20
WESTINGHOUSE NON-PROPRIETARY CLASS 3 620.00 610.00 -
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{N0!AN POINT UNIT 3 A5YmtTRIC TUBE PLUGGING ANALY315 DROPPED RCCA 0F WORTH 100 PCM FIGURE !!-21
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OROPPED RCCA 0F WORTH 100 PIM FIGURE !!-22
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INDIAN POINT UNIT 3 L
i AsYmETn!C TUBE PLUGGING ANALY315 Loss or nonn4L FttoWAtta l
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t
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INDIAN POINT UNIT 3 ASY METRIC TUBE PLUGGING ANALYSIS L055 0F NORMAL FEEDWATER FIGURE-!!-24 e
WESTINGHOUSE NON-PROPRIETARY CLASS 3
!!!. SAFETY EVALUATION - LOCA 1
A 1981 Westinghouse Evaluation Model ECCS Performance analysis for Indian j
Point 3 has demonstrated acceptable results for 24% uniform steam generator tube plugging. This docketed, NRC-approved analysis demonstrates a calculated peak clad temperature (PCT) of 1395'F at a peaking factor (FQ) value of 2.14.
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.
Sensitivity studies performed in the past and documented in WCAP-0906 (Reference !!!-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 primarily a consequence of the added resistance to fluid flow through the coolant loops during core reflood. Because the added resistance represents the predominant phenomenon associated with tube plugging and because of the linear nature of the. PCT relationship, deviations in plugging from one steam gene,rator to another do not significantly affect LOCA analysis results.
The impact of asynestric 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
!!!-2, WCAP-gl70. 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 III-2 is as follows:
P, is downcomer static pressure P
I**
- E "'
C 75830:10/100584 111-1 O
_= _
j',
WESTINGHOUSE NON-PROPRIETARY CLASS 3
/
P is containment pressure X
K is the resistance loss coefficient
'l 4
Subscripts to K refer to loop (intact loop or broken loop) and location (hot
)
leg, steam generator, etc.)
)
WREFLOOO 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 STUS is the pressure difference between vessel downconer and.
containment p is the liquid density in the downconer g
g is the gravitational constant AZ is the difference in water level between downconer and cors 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 mess effluent fraction, the fraction of mass entering the out core which is expelled S
is the unss velocity at the core exit core Consider loop behavior during the core reflood transient:
75830:10/0g2184
!!!-2
l WESTINGHOUSE NON-PROPRIETARY CLASS 3 tnt (FLOOO equations state p'ressure relationships are
~
,,=,,...,,,
PC " 'O * # NI L
The driving force for intact loop flow is
.P=P
~PO " # 'AI C
L Simplify the loop equation of WCAP-0170 p. 2-2 by eliminating small magnitude terms:,
Kgt (wgt) tais gives &P =
for the intact loop 2
2 p6 Agg Ett (Wit) ther p(g&Z =
2 2p A g gg 2p At g AZ g
and wgg = Agg (
)1/2 gIL Thus at any particular point &Z during the core reflood process gL(h)
W eA gt IL
/2 w
a (frictional resistance) gg Apply the simplified equation to broken loop:
Kat (Wat)
'BL " 2p A 2 g gg 75830:10/100584 III-3
WESTINGHOUSE NON-PROPRIETARY CLASS 3
-t for the broken'. loop
\\
m 3
-P
- # g I + AP APgg = PC X
L 5TUS KgL (wgt)2 then p g&Z + APSTUS "
2 g
2pG A BL
)
2p a g&2 + 2p &P gt g STUS 1/2 E
3 sc wSL " BL K
SL i
~
A review of the IP3 limiting break (Cg = 0.4 DECLG) reveals that AP5TUS is small compared to p g&Z until calculated clad temperature has 4
g increased to a value near PCT. Therefore, the.lP term may be ignored STUS to obtain t
2P P 9 AI
?
g L 1/2 E
3 "8L
- ABL KBL z
L,1/2 "8L
- SL I_Kg from p. 2-6 of WCAP-8170, the loop flow boundary condition at the core t's 0core " C*#L out i
which may be written as s
"BL * "IL "Y
- F A
C*#L out C
Therefore VC" C L V.1 15830:10/100584 III-4
WESTINGHOUSE NON-PROPRIETARY CLASS 3 Core flooding rate V is detemined by the ability to vent core-generated C
steam through the loops and is directly proportional to the sum (wgL +
l gg). IP3 exhibits its calculated PCT well into the core reflood portion of w
the LOCA transient: PCT is directly related to the magnitude of th'e flooding rate. The higher the value of V [and (wgL + w!LN C
calculated PCT in the IP3 1981 Model analysis.
The effect of tube plugging configurations upon total flow exiting the core (wgL + wgL) can now be assessed from the proportionality relationships.
l For a 4-loop plant the total flow through the loops, w, is given as j
2 2
a(
)M + (
)M w, = wgg + wgg IL BL
, a.75 A, Kgg
+.25 A, Egg or w in an original, unplugged state Egt, = Kgt,:
w, a 1.0 A, Kgg.
E w, When SG tube plugging is introduced, w, will be diminished due to an increase in frictional resistance. In the following presentation changes in resistance caused by.56 tube plugging will be applied to the loss coefficient (K) tem of the (A*K-1/2] expressions while A is held constant for ease of computation. Since no critical flow offacts are involved the flow impact of SG tube plugging can be properly represented in this fash.on. The uniform plugging case and two bounding asynnetric plugging cases are considered.
I.
Unifom SG Tube Plugging Case 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 10% of the original total loop resistance. In the 245 unifom SG tube plugging WREFL000 cases, the steam generator accounts for slightly more than 30% of the total IP3 loop resistance to flow.
7583Q:10/102384
!!1-5
WESTINGHOUSE NON-PROPRIETARY CLASS 3
, a.75A,(K () 1/2 +.25A,(K I w
g BL
, =.75A, (Kgg *1.1)
+.25A, (Kgt *1.1)
W g
o M
M
, a 1.0A, (Kgg 9W
= w,9 J w
o
, e.9535 w, w
Total flow exiting the core is reduced by.0465 in this uniform resistance case.
II. A11 Plugging in Broken Loop
,.).)3.
)! s; 7 The (4*(10% of individual loop resistance)h ' i. red delstance is plaged totally into the broken loop in WREFL000. Then
, e.75A K
+.25A,Kgg w
IL
/
) /2 +.25A, (1.4 KgL )
, =.75A, (Kg w
g
, a (.75 +
- 4) A, Kgg w
o
, = (.75 +.2113) w, =.9613 w, w
Total flow exiting the core is reduced by.0387 in this case. The reduction in w and VC (and the subsequent rise in PCT) is predicted to be less for this configuration. Therefore, there is no difference of significance relative to the uniform plugging j
configuration.
7583Q:10/011686 III-6
WESTINGHOUSE NON-PROPRIETARY CLASS 3 III. No Plugging in Brcken Loop 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.
75A K
+.25A, K8L w
gg H.133)
+
e.75 A, (Kgg 25 A, Kgg g
g
, = [.75 (1.133) 1/2 +.25] w, w
, e.9545 w, w
Total flow exiting the core is reduced by.0455. The reduction in w and V is a bit less for this configuration. The above g
discussion has shown that asymmetry presumed in steam generator tube 1
plugging causes no adverse offects based on the WREFLOOD equations.
However, the arguments presented here should only be applied to the established range of applicability in which WREFLOOO has been
. employed in Evaluation Model ECCS computations. The indicated upper bound is a 30% steam generator tube plugging level in any SG unit.
The above discussion has demonstrated based upon the pertinent WREFLOOO 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 24% of the total number of SG tubes present in the plant.
2.
The number of equivalent tubes plugged does not exceed 30% of the 3260 tubes present in any steam generator.
75830:10/011586 III-7 1
.m m,_, _..
.,-,.,r___-___,-m.--m,_,_.~r
WESTIN6 HOUSE NON-PROPRIETARY CLASS 3 6
e
.(
C j
J g
il 1
I E
E j
W
=
y a
j i
t i
4A; al N
.a l
t y.
=E P :I E
g-V.i g
g x
I I
w
- b ln 5
[
l' r
e
.g W
G
--_-.------..------------,-,--n,._,.,
..-.----...,.n-,
,,.,,,,, _ _ _ _ - ~ -,, -,. - - -. _.., - - _.,
WESTINGHOUSE NON-PROPRIETARY CLASS 3 e
0 3
-., e.
Y *8 u,
E W1 A
d$E 5
- =4 e
1 z
r-W k
>< I a
e m
~
]
3 4
A 1 '.
=
+.. +
- d. d e
a?
g it 11 o*
c2 t
..7 84 it) :
5 w
[d t
=
x, fe 4
e
WESTINGHOUSE NON-PROPRIETARY CLASS 3 IV. CONTROL AND PROTECTION SYSTEM SET POINTS Only two changes are needed in the control and protection system instrumentation to account for asymmetric tube plugging:
1.
The T, program in the reactor control system must be adjusted, if necessary, such that no cold leg temperature exceeds its allowable value. This adjustment preserves core DN8 margin during steady-state l
operation.
2.
The overtemperature delta-T reactor trip channels must be calibrated during power operation in terms of both the delta-T and T,,,
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 gpe, 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,,g.
With asymmetric tube plugging; however. inlet temperature. T, is a more appropriate limit, and g
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 cortrol deadband and 2*F for temperature error. Thus, during steady-state operation with loop asynnetries, measured T,,, should not exceed 576.7'F.
and T, should not exceed 544.9'F, for any loop. The precaution below g
should be included with appropriate station procedures.
15830:1D/100984 IV-1
WESTINGHOUSE NON-PROPRIETARY CLASS 3 "Ouring steady-state operation with either automatic or manual control, the measured average coolant temperature (T,, ) in each and every loop must be no greater than the programmed T,,, at fu$1 power (574.7'F) plus 2*F for control deadband. In addition, if steady-state loop-to-loop asyuumetries 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 thermal design value.(542.g'F) plus 2*F for control deadband. Reduction in the T,,, program must be made if necessary to preserve this T limit.'
inlet e
9 i
s e
15830:10/100584 IV-2 1
~ -
WESTINGHOUSE NON-PROPRIETARY CLASS 3 V.
CONCLUSIONS c
The impact of the tube plugging asymmetry 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:
(
l.
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 rat exceed 542.9 *F plus 2*F for control deadband.
I t
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 t
such as locked rotor and partial loss of flow for which an asyneetric flow t
distribution is coupled with an asyneetric 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 f
plugged. Thus, it has been demonstrated by this study and the previous 24%
i 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 generator tube plugging ECCS performance analysis for Indian Point 3 remains a valid basis for plant operation as long as the plugging level does not exceed 50% in any steam generator 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:10/010686 V-1 l
- _ _ _ _ _,. _ - -., _ _. -.. -..... _. - _ _ _ _, _ _, _ _ _ _ _ _. _.. ~ -. _,. _.. _. _. _ _ _ _ _ _, _
WESTINGHOUSE NON-PROPRIETARY CLASS 3 REFERENCES I-1 Indian Point Unit 3, 24 Percent Uniform Tube Plugging Analysis Letters J. P. Sayne (NYPA) to S. A. Varga (NRC)
IPN-83-5, January 13, 1983 IPN-83-37 May 5,1983 IPN-83-101. December 14, 1983 1-2 Final Safety Analysis Report, Indian Point Unit 3. Docket Number 50-286.
II-1 Surnett, T.W.T., et al, 'LOFTRAN Code Description," WCAP-7907-P-A (Proprietary Class 2) WCAP-7907-A (Proprietary Class 3) April,1984.
11-2 Van Houten R., " Fuel Rod Failures as a Consequence of Departure from Nucleate Soiling or Dryout " NUREG-0562. June 1979.
!!!-1 Thompson, C. N., and Esposito, V.
J., " Perturbation Technique for Calculating ECCS Cooling Performance " WCAP-8986. February,1977.
III-2 Collier, 6. et al., " Calculational Model for Core Reflooding Af ter a Loss of Coolant Accident (W REFL000 Code)," WCAP-8170 (Proprietary Class 2) WCAP-8171 (Proprietary Class 3), June,1974.
7583Q:10/101884 V!-1
WESTINGHOUSE NON-PROPRIETARY CLASS 3 APPENDIX A As a source of additional information and in response to NRC questions, an additional Locked Rotor analysis was performed. This particular analysis used several different assumptions.
Method of Analysis An analysis was performed consistent with the analysis presented in Section II.D.4 with the following exceptions:
1.
Loss of of fsite power was not assumed. Based on the grid stability study conducted by New York Power Authority, it was determined that for the temperature and DNB criteria power would be maintained to the three l
unfaulted pumps past the crucial period of minimum DNBR conditions (< 2.5 seconds after rod motion).
2.
For offsite dose calculations, fuel rods that experience DNBR values less than the 95/95 limit are assumed to fail.
3.
Since IP3 was planning to operate with Optimized Fuel Assemblies (OFA), an evaluation was done to determine the limiting fuel type and configuration and the analysis was done consistent with that information.
Plant responses were computed using the LOFTRAN code. Fuel and clad temperatures at the hot spot were computed using the FACTRAN code. The DNBR
~
was talculated using the THINC code.
Three effects were evaluated for the Locked Rotor transient:
A) Primary pressure transient.
- 8) Fuel clad temperature transient (calculated assuming film boiling in order to produce the worst possible results).
C) DN8 transient (for determining the percentage of rods in DN8 for of fsite dose calculations).
7583Q:10/010986 A-1
WESTINGHOUSE NON-PROPRIETARY CLASS 3 Results The time sequence of events for the Locked Rotor transient is presented in Table A-1.
Nuclear power, core heat flux, RCS pressure, core flow, and clad inner temperaturs are shown in Figures A-1 through A-3.
For this analysis assuming offsite power available, a peak clad conservatively calculated for this event. For the DNB calculation, the full 0FA core (most limiting case) yielded 1.9% rods in DNS. This percentage of rods in DNS resulted in a dose release less than the 10CFR100 limit.
Conclusions A tube plugging distribution of not greater than 30% in any steam generator during a Locked Rotor event will not result in exceeding any safety limits.
l
- 0 1
7583Q:10/010686 A-2
E STINGHOUSE NON-PROPRIETARY CLASS 3 TA8LE A-I TIME SEQUENCE OF EVENTS Time Accident
[ygni M
n 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 1.1 i
Maximum clad temperature occurs 3.5 Maximum RCS pressure occurs 3.5 l
.?
75830:10/010986 A-3
WESTINGHOUSE NON-PROPRIETARY CLASS 3 1.2000 1.0000 I
8.80000- -
~
T E
et
.60000 -
1 E. 40000 - -
M V
.20000 -
0.0
~
1.2000
^
.~
1.0000
~
I
%.30000 -
y E
.60000 "
m T
d e-
. y.40000 -
E
.20000 "
I I
i 1
~
TIMC (SEC)
Indian Point Unit 3 Asymetric Tube Plugging Analysis Locked Rotor with Offsite Power Figure A.1
-,e
-p, e-egee-,,-ye
-,--+m-e e
%e--
---ve-yw-e-wr-~-----------e+------e-
-v'
-'- ='
WESTINGHOUSE NON-PROPRIETARY CLASS 3 2600.0 2500.0 E
j 2400.0 -
[
2300.0 -
e 2200.0- -
mM 2100.0 -
E 2000.0 s.
1900.0 -
1900.0
~
1.2000 1.0000< -
- 3S j
.50000 -
m E.50000 -
E g
st.40000 -
E w
.20000 -
i.
A i
1.
,4 e
a a
J E
e TlHE (SECl Indian Point Unit 3 Asymetric Tube Plugging Analysis Locked Rotor with Offsite Power Figure A-2
WESTINGHOUSE NON-PROPRIETARY CLASS 3 2000.0 c
1750.0 H u
5 l1500.0*
=
gWE 1750.O <
5*
I g 1000.00 +
4.
d 750.00 <
I I
I I
s a
d J
i E
TIME 15tc) i Indian Point Unit 3 Asymetric Tube Plugging Analysis Locked Rotor (Hot Spot)
Figure A-3
-