ML20235Z584
| ML20235Z584 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 05/31/1987 |
| From: | Dominicis D, Etling R WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML19292H566 | List: |
| References | |
| TAC-65524, TAC-65525, WCAP-11309, WCAP-11309-R01, WCAP-11309-R1, NUDOCS 8707270337 | |
| Download: ML20235Z584 (57) | |
Text
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WESTINGHOUSE CLASS 3 WC,AP-11309 REV. 1
}
i RTD BYPASS ELIMINATION LICENSING REPORT FOR CATAWBA UNITS 1 & 2 May, 1987 W. R. RICE R. H. 0 WOC
- )
,6/
APPROVED:
d ch APPROVED:
R.I.Etling(dnager
'D. P. Dominicis, Manager BOP & Turbine'-Island Systems Operating Plant Licensing-II Westinghouse Electric Corporation
~
Pittsburgh, PA B707270337 9707pg DR ADOCK 05000413 p
PDR 0514v;1D/021387
i o
ACKNOWLEDGEMENT The authors wish to recognize contributions by the following individuals:
R. CALVO R. A. HOLMES E. K. HACKMEN D. S. HUEGEL W. G. LYMAN J. C. MESMERINGER P. SCHUEREN C. R. TULEY R. T. WASIL M. WEAVER G. E. LANG
.. en
'0514v:1o/051387
_ _ _ _ _ _ _ _ _ - _ _ _ _ _ = _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
TABLE OF CONTENTS a
Section Pm 1.0 Introduction 1
1.1 Historical Background 1
1.2 Nechanical Modifications 2
1.3 Electrical Modifications 4
2.0 Testing 4
2.1 Response Time Test 5
2.2 Streaming Test 6
3.0 Uncertainty Considerations 7
3.1 Calorimetric Flow Measurement Uncer.tainty 7
3.2 Hot Leg Temperature Streaming Uncertainty 12 4.0 Safety Evaluation 14 4.1 Response Time 14 4.2 RTD Uncertainty 14 4.3 Accidents Reanalyzed / Evaluated 15 4.4 Instrumentation and Control Safety Evaluation 19
.4.5 Nechanical Safety Evaluation 21 5.0 Control System Evaluation 23
)
i 6.0 Conclusions 24 7.0 References 24 D514v:1D/051387 l
LIST OF TABLES Table Title Page a
2.1 Response Time Parameter fer RCE Temperature 5
Measurement 3.1 Flow Calorimetric Instrumentation Uncertainties 8
3.2 Flow Calorimetric Sensitivities 9
3.3 Calorimetric RCS Flow Measurement Uncertainties 10 3.4 '
Cold Leg Elbow Tap Flow Uncertainty 11 4.1 Time Sequence of Events for a RCCA Bank Withdrawal 18 at 60% of Full Power
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l e
0514v;1o/051387
LIST OF FIGURES
.~
Figure Title Page a
1 Hot Leg RTD Scoop Modification for Fast-Response 25 RTD Installation
-2 Cold Leg Pipe Nozzle Modification Fast-Responsa 26 RTD Installation
[
3 Additional Boss for Cold Leg Fast-Response RTD 27 Installation 4
RTD Averaging Block Diagram, Typical for Each of 4 28 Channels 5
Nuclear Power and Core Heat Flux for a RCCA Bank Withdrawal 29 at 60% of Full Power with Minimum Reactivity Feedback (75PCWSECRate) 6 Pressurizer Pressure and Water Volume for a RCCA Bank 30 Withdrawal at 60% of Full Power with Minimum Reactivity Feedback (75 PCWSEC Rate)
-7 Core Average Temperature and DNBR RCCA Bank Withdrawal 31 at 60% oflull Power with Minimum Reactivity Feedback (75PCWSECRate) 8 Nuclear Power and Core Heat Flux for a RCCA Bank Withdrawal 32 at 60% of Full Power with Minimum Reactivity Feedback (3 PCWSEC Rate) 9 Pressurizer Pressure and Water Volume for a RCCA Bank 33 Withdrawal at 60% of Full Power with Minimum Reactivity Feedback (3 PCUSEC Rate) 10 Core Avera!]e Temperature and DNBR RCCA Bank Withdrawal 34 at 60% of iull Power with Minimum Reactivity Feedback (3 PCWSEC Rate) 11 Minimum DNBR vs. Reactivity Insertion Rate; RCCA Bank 35 Withdrawal From 100% Power i
12 Minimum DNBR vs. Reactivity Insertion Rate; RCCA Bank 36 Withdrawal From 60% Power 13 Minimum DNBR vs. Reactivity Insertion Rate; RCCA Bank 37 Withdrawal from 10% Power i
L 0514v:1D/D$1387
[
r
1.0 INTRODUCTION
Westinghouse Electric Corporation has been contracted by Duke Power to remove the existing RTD (Resistance Temperature Detector) Bypass System and replace the hot leg and cold leg temperature measurement with fast response RTDs installed in the reactor coolant loop piping. This report is submitted in support of continued operation of the Catawba Units with the new RTD System installed.
1.1 HISTORICAL BACKGROUND Prior to 1968, PWR designs had been based on the assumption that the hot leg temperature was uniform across the pipe. Therefore, placement of the temperature instruments was not considered to be a factor affecting the accuracy of the measurement. The hot leg temperature was measured with direct-immersion RTDs extending a short distance into the pipe at one location. By the late 1960s, as a result of accumulated operating experience at several plants, the following problems associated with direct immersion RTDs were identified.
Temperature streaming conditions; the incomplete mixing of the coolant o
leaving regions of the reactor core at different temperatures produces significant temperature gradients within the pipe.
Cooling and draining of the loops before the RTDs could be replaced.
o The RTD bypass system was designed to resolve these problems; however, operating plant experience has now shown that operation with the RTD bypass j
loops has caused some new problems:
)
J Plant shutdowns caused by excessive primary leakage through valves, o
flanges, etc., or by interruptions of bypass flow due to valve stem failure.
o514v:1o/051387 1
.o
j o. Increased radiation exposure due to maintenance on the bypass line and to crud traps which increase radiation exposure throughout the loop compartments.
j s
The proposed temperature measurement modification has been developed in response to both sets of problems encountered in the past. Specifically:
o Removal of the bypass lines eliminates the components which have been a major source of plant outages as well as Occupational Radiation Exposure (ORE).
o Three thermowell-mounted hot leg RTDs provide an average measurement (equivalent to the temperature measured by the bypass system) to account for the temperature streaming phenomenon.
o Use ~of thermowells permits RTD replacement without punt draindown.
Following is a datailed description of the effort required to perform this modification.
1.2 MECHANICAL MODIFICATIONS The individual loop temperature signals required S e input to the Reactor Control and Protection System will be obtained using RTDs installed in each reactor coolant loop.
1.2.1 Hot Leg a) The hot leg temperature measurement on each loop will be accomplished with three fast response narrow range RTDs mounted in thermowells. To accomplish the sampling function of the RTD bypass manifold system and eliminate the need for additional hot leg piping penetrations, the thermowells will be located within the three existing RTD bypass manifold scoops. A hole will be drilled through the end of each scoop so that water will flow in through the existing holes in the leading edge of the scoop, past the RTD, and out through the new hole [
]+a,c,
o514v:1D/oS1387 2
A
p 1
These threrg RTDs will' measure the. hot leg temperature which is used to calculate the reactor coolant loop differential temperature (AT) and average temperature (T,yg).
I b) This modification will not affect the single wide range RTD currently installed near the entrance of each steam generator. This RTD will continue to provide the hot leg temperature' used to monitor reactor coolant temperature during startup, shutdown, and post accident conditions.
1.2.2 Cold Leg a)
One fast response, narrow range, RTD will be located in each cold leg at the discharge of the reactor coolant pump (as replacements for the cold leg RTD's located in the bypass manifold). Temperature streaming in the-cold leg is not a concern due to the mixing action of the RCP.
For this reason, only one RTD is required. This RTD will measure the cold leg temperature which is used to calculate reactor coolant loop AT and T,yg.
The existing cold leg RTD bypass penetration nozzle will be modified [.
]+a,c to accept the RTD thermowell.
b) This modification will not affect the single wide range RTD in each cold leg currently installed at the discharge of the reactor coolant pump.
This RTD will continue to provide the cold leg temperature used to monitor reactor coolant temperature during startup, shutdown, and post accident conditions.
c) A new penetration will also be made to each cold leg to accept an additional well mounted narrow range RTD, for use as an installed spare.
This will give the new modification a tolerance for RTD failures equivalent to the bypass loops. A new cold leg boss will be added
[
] $ accept the RTD thermowell.
1.2.3 Crossover Leg The RTD bypass manifold return line will be capped at the nozzle on the crossover leg.
os14v;1o/osiss7 3
i.
1.3 ELECTRICAL MODIFICATIONS 1.3.1 Function
[
]**'C shows a block diagram of the modified electronics. The hot leg RTD measurements (three per loop) will be electronically averaged in the process protection system. The averaged Thot signal will then be input to the appropriate protection function. This will be accomplished by additions to the existing 7300 equipment.
1.3.2 Qualification Equipment seismic and environmental qualification will be to IEEE. standards
~--
344-1975.and 323-1974, respectively, as described in WCAP-8587, Rev. 5,
" Methodology for Qualifying Westinghouse WRD Supplied NSSS Safety Related Electrical Equipment".
1.3.3 RTD Failures Existing control board AT and T,yg indicators and alarms will provide the means of identifying RTD failures. The spare cold leg RTD provides sufficient spare capacity to accommodate a single cold leg RTD failure per loop.
Failure of a hot leg RTD will require manual action to defeat the failed signal, and a manual rescaling of the electronics to average the remaining signals [
)+a,c 2.0 TESTING There are two specific tests which have been performed to support the installation of the fast response RTDs in the reactor coolant piping:
a
^
response time test and a hot leg temperature streaming test.
l C614v:1o/051387 4
I
L 2.1 RESPONSE TIME TEST Westinghouse has performed an RTD Response Time Test at its Forest liills Test facility. This test placed a fast response RTD, manufactured by RdF Corporation, inside a scoop, within a thermowell, which modelled the actual in plant installation. The flow conditions were adjusted to aque.1 the high velocity Reactor Coolant System flows of approximately [.
]+a The RTD's response time is determined based on a comparison of the RTD with thermocouple which had been previously calibrated and response time characterized. Sixty-five test runs were made at various flow rates while gathering data on 2 RTDs. The test results demonstrated a mean response time for the RTD, thermowell and scoop of less than [
]+a,b,c seconds. Table 2.1 provides a comparison of the present RTD Bypass System response time and how it would differ _with the new system in place.
TABLE 2.1 RESPONSE TIME PARAMETERS FOR RCS TEMPERATURE MEASUREMENT Fast Response RTD Bypass System Thermowell RTD System
+
+
_ a,c
_ a,e RTD Bypass Piping and Thermal Lag (sec)
RTD Response Time (sec)
RTD Filter Time Constant (sec)
Electronics Delay (sec)
Total Response Time (see) 6.0 sec 7.0 see h
Based upon the response time parameters in Table 2.1, it becomes evident that the Catawba Units can accommodate the new response time with no further plant testing required.
o614v;1o/051387 5
l
l 2.2 STREAMING TEST 4 -
Past. testing at Westinghouse PWRs has established that temperature stratification exists in the hot leg pipe with a temperature gradient from top to bottom of ['
]4,c.e A test program was implemented at McGuire
- Unit I to confirm the temperature streaming magnitude and stability with measurements of the RTD bypass branch line temperatures on twc adjacent reactor coolant loops. Specifically, it was intended to determine the magnitude of the differences between branch line temperatures, confirm the short-term and long-term stability of the temperature streaming patterns and evaluate the impact on the indicated temperature if only 2 of the 3 branch line temperatures are used to determine an average temperature. This plant specific data will be used in conjunction with data taken from other Westinghouse designed plants to determine an appropriate temperature error for use in the safety analysis and calorimetric flow calculations. Section 3 will discuss the specifics of these uncertainty considerations.
9 The McGuire Unit 1 test data has been reduced and characterized to answer the three objectives of the test program. First, it is conservative to state that the streaming pattern [
]%,c,e Steady state data taken at 100% power for a period of four weeks indicates that the streaming pattern [
]%,c.e.
In other words, the temperature gradient [
]4,c.e This is inferred by [
]%,c.e observed between branch lines. Since the [
]*'C'* into the RTD averaging circuit if a hot leg RTD fails and only 2 RTDs are used to obtain an average hot leg temperature. The operator can review temperatures recorded prior to the RTD fail.ure and determine an [.
)%,c.e into the "two RTD" average to obtain the "three RTD" expected reading. This significantly reduces the error introduced by a failed RTD.
The McGuire Unit 1 data also supports previous calculations of streaming errors determined from tests at other Westinghouse plants, The McGuire Unit 1 data is consistent with the upper bound temperature gradients that characterize the previous data. There were no new discoveries, but the data
. 0514v:1o/051387 6
1 did add a' dimension previous tests did not have. The McGuire Unit I test I
sampled temperatures from the pipe interior while all previous tests investigated temperature gradients at the pipe surface. The pipe internal i
. temperature data has greatly strengthened the assumptions and. inferences made with previous test data.
The streaming test and response time test have both provided valuable information needed to support the design of the fast-response RTDs installed in the reactu coolant piping.
3.0 bNCERTAP!TYCONSIDERATIONS This new method of tot leg temperature measurement has been analyzed to-determine if it will have an impact on two uncertainties included in the Safety Analysis:
Calorimetric Flow Measurement Uncertainty and Hot Leg Temperature Streaming Uncertainty.
3.1 CALORIMETRIC FLOW MEASUREMENT UNCERTAINTY Reactor coolant flow is verified with a calorimetric measurement performed
.af ter the return to full power operation following a refueling shutdown. The two most important instrument parameters for the calorimetric measurement are the narrow range hot leg and cold leg coolant temperatures. The accuracy of the RTDs has, therefore, a major impact on the accuracy of the flow measurement.
The current licensed flow measurement uncertainty for Catawba for the sum of the four loop flows including elbow taps, is about i 2.1% flow (not including 0.1% flow for feedwater venturi fouling allowance). However, with the use of three T RTDs (resulting from the elimination of the RTD Bypass lines) and Hot the latest Westinghouse RTD cross-calibration procedure (resulting in lower RTD calibration uncertainties at the beginning of a fuel cycle), it is possibigoreducetheRCSflowmeasurementuncertaintytoapproximately flow (including the cold leg elbow taps and excluding feedwater venturi
~
louling). Utilizing the uncertainty calculational methodology explicitly described in WCAP-11168-R1 (Reference 1), Tables 3.1 through 3.4 were generated to provide the Catawba specific instrument uncertainties, calorimetric sensitivities, and flow uncertainties.
s7
o TABLE 3.1 FLOW CALORIMETRIC INSTRUMENTATION UNCERTAINTIES
(% SPAN)
FW TEMP FW PRES FW d/p S1H PRESS T
Ic PR2 PRESS H
SCA
+a,c
=
M&TE-=
=
=
SD '=
R/E
=
RDOT =
BIAS =
CSA
=
i 0F INST USED 3
1 4 **
'F psia
% d/p psia
- F
'F psia INST SPAN =
692.
2000.
120.
1500.
100.
100.
800.
INST UNC.
(RANDOM)=
+a,e INST UNC.
(BIAS)
=
440.
1100.
1000.
620.0 561.6 2250.
NOMINAL
=
[
3+a,e Number of Hot Leg and Cold Leg RTDs used for measurement in each loop and the number of Pressurizer Pressure transmitters used overall, i.e., one per loop.
0514v:1D/051387 8
l TABLE 3.2-FLOW CALORIMETRIC SENSITIVITIES FEEDWATER FLOW F,
TEMPERATURE
=
+a,e MATERIAL
=
L DENSITY TEMPERATURE
=
PRESSURE
=
DELTA P
=
FEEDWATER ENTHALPY TEMPERATURE
=
l PRESSURE
=
h, 1192.9 BTU /LBM
=
h 419.5 BTU /LBM
=
f Dh(SG) 773.4 BTU /LBM
=
STEAM ENTHALPY PRESSURE
+a,c
=
NOISTURE
=
HOT LEG ENTHALPY TEMPERATURE
=
PRESSURE
=
T.0 BTU /LBM
~
h
=
g 561.7 BTU /LBM h
=
c Dh(VESS) 81.3 BTU /LBM
=
Cp(T )
1.563 B W/LBM 'F
=
H COLD LEG ENTHALPY TEMPERATURE
+a,c
=
PRESSURE
=
1.273BW/LBM*7 C (T )
=
p c COLD LEG SPECIFIC VOLUME TEMPERATURE
+a,c
=
PRESSURE
=
- 0514v:1D/051387 9
TABLE 3.3 CALORIMETRIC RCS FLOW MEASUREMENT UNCERTAINTIES COMPONENT INSTRUMENT ERROR FLOW UNCERTAINTY FEEDWATER FLOW f
VENTURI a,c THERMAL EXPANSION COEFFICIENT TEMPERATURE MATERIAL DENSITY TEMPERATURE PRESSURE DELTA P FEEDWATER ENTHALPY TEMPERATURE PRESSURE
- STEAM ENTHALPY PRESSURE MOISTURE NET C' MP HEAT ADDITION J
HOT LEG ENTHALPY TEMPERATURE
~
STREAMING, RANDOM STREAMING,. SYSTEMATIC PRESSURE ~
COLD LEG ENTHALPY TEMPERATURE
-PRESSURE COLD LEG SPECIFIC VOLUME TEMPERATURE PRESSURE RTD CROSS-CAL SYSTEMATIC ALLOWANCE t
0514v:1D,151387 10
TABLE 3.4 COLD LEG ELBOW TAP FLOW UNCERTAINTY
~
INSTRUMENT UNCERTAINTIES
% d/p SPAN
% FLOW PMA
+a,e
=
=
SCA
=
=
=
=
=
M&TE =
RTE =
RD
=
ID
=
A/D
=
RDOT =
BIAS =
FLOW CALORIM. BIAS
=
FLOW CALORIMETRIC
=
INSTRUMENT SPAN 120.
=
SINGLE LOOP ELBOW TAP FLOW UNC =
+a,c
~
N LOOP ELBOW TAP FLOW UNC
=
N LOOP RCS FLOW UNCERTAINTY (WITHOUT SIAS VALUES)
=
N LOOP RCS FLOW UNCERTAINTY (WITH BIAS VALUES) 1.7
=
0514v.1D/051387 11
- 3. P.
HOT LEG TEMPERATURE STREAMING UNCERTAINTY The safety analyses incorporate an uncertainty to account for the difference between'the actual hot leg temperature and the measured hot leg temperature caused by the incomplete mixing of coolant leaving regions of the reactor core at different teniperatures. This temperature streaming uncertainty is based on an analysis of test data from other Westinghouse plants, and on calculations to evaluate the impact on temperature measurement accuracy of numerous possible temperature distributions within the hot leg pipe. The test data has shown that the circumferential temperature variation is no more than [
)+b,c.e,
and that the inferred temperature gradient within the pipe is limited to about
[
]+b,c.e The calculations for numerous temperature distributions have shown that, even with margins applied to the observed temperature gradients, the three point temperature measurement (scoops or thermowell RTDs) is very effective in determining the average hot leg temperature. The most recent calculations for the thermowell RTD system have established an overall streaming uncertainty of [
]+b,c.e for a hot leg measurement. Of this total, [
)+b,c.e The overall temperature stenaming uncertainty applied to the calorimetric flow measurement is only slightly larger than the uncertainty used in previous analyses.
The new method of measuring hot leg temperatures, with the thermowell RTDs located within the three scoops, is at least as effective as the existing RTD bypass system, [
1+a,c Although the new method measures temperature at one point within the thermowell, compared to the five sample points in a 5-inch span of the scoop measurement, the thermowell measurement point is opposite the center hole of the scoop and therefore measures the equivalent of the average scoop sample if a linear radial temperature gradient exists in the pipe. The thermowell measurement may have a small error relative to the scoop measurement if tha temperature gradient over the 5-inch scoop span is nonlinear. Assuming that the maximum inferred temperature oS14v:1o/051387 12
)
gradient of [
]4,c.e exists from the center to the end of'the scoop, the difference between the thermowell and scoop measurement is limited to [
]+b,c.e Since three RTD measurements are averaged, and the nonlinear'. ties at each scoop are random, the effect of this error on the hot
%,c.e On the other leg temperature measurement is limited to [
3 hand, imbalanced scoop flows can introduce temperature measurement uncertainties of up to [
3+a,c,
In all cases, the flow irbalance uncertainty will equal or exceed tne
[
~)+b,c.e sampling uncertainty for the thermowell RTDs, so the new measurement system tends to be a more accurate measurement with respect to streaming uncertainties.
Temperature streaming measurements from the test at McGuire Unit 1 have been obtained. The measurements ecnfirm the [
.)+b,c.e, Over the 4-week testing period, there have been only minor variations of less than [.
)+b,c.e in the temperature differentials between scoops, and smaller variations in the average value of the temperature differentials.
[
)+b,c.e,
Provisions were mede in the RTD electronics for operation with only two hot leg RTDs in service.
The two-RTD measurement will be biased to correct for
~
the difference compared with the three-RTD average. Based on the McGuire 4,c,o
~
Unit 1 test data, the bias would be limited to between [
3 Data comparisons show that the magnitude of this bias varied less than
[
']4,c.e over the test period.
0514v:1o/c51387 13
m 4.0 SAFETY EVALUATION 4.1 RESPONSE TIME
~-
The primary impact of the RTD bypass elimination on the FSAR Chapter 15 non-LOCA safety analyses (Reference 2) is the increased response time associated with the fast. response thermowell RTD system.. Currently, the overall response time of the Catawba RTD bypass-system assumed in the' safety analyses is approximately 6.0 seconds (see Table 2.1). For the fast response thermowell RTD system the overall response time is approximately 7.0 seconds as described in'Section 2.1 and as given in Table 2.1.
This increased RTD response time results in longer delays fro'n the time when the fluid conditions in the RCS require an Overtemperature AT or Overpower AT reactor trip _until a trip signal is actually generated. Therefore, those transients that rely on the above mentioned trips must be evaluated for the longer response time. The affected transients include the Uncontrolled RCCA Withdrawal at-Power, the' Uncontrolled Boron Dilution at Power, and the Steamline Rupture at Power events and are discussed in Section 4.3.
4.2 RTD UNCERTAINTY The proposed fast response thermowell RTD system will make use of RTDs
'" ' C manufactured by the RdF Corporation with a total uncertainty of
~
~
assumed for the analyses. These are the same RTDs as currently installed in the plant.
The FSAR analyses make explicit allowances for instrumentation errors for some of the reactor protection system setpoints.
In addition, allowances are made
.for the initial. average reactor coolant system (RCS) temperature, pressure and power as described in FSAR Section 15.0. These allowances are made explicitly to the initial conditions for non-DNB events; for DNB events these allowances are statistically combined into the design limit DNBR value, consistent with the Improved Thermal Design Procedure (ITDP) methodology (Reference 4).
1 osier.to/osiss7 14 l
y J
~-
2 A..
q W
3 L
The following protection and contro? system parameter: were affected by the-chastefromo,neThot RTD, to three T RTDs: theOvertempeNatureAT.
hot (01kNE OpilpowerjaT?(OPAT), and Low RCS Flow reactor trips, the RCS O
averahetomhdraturemedsurementsusedforcontrolboardindictionandinputto y
the rod control system, and the calculated value of the RCS flow uncertainty,-
,b System uncertainty calculations were performed for these parameters to w
deterrine the impact cf the change'in number of T RTDs. The results of hot these esiculatfo#s show sufficient margin exists to account for all known instrument uncertainties, after the adjustment of the OTDT Ki gain from nom its current value of 1.411 to 1.38. As a result, to ensure adequate margin to an Overtemperatura AT reactor trip exists for a large load rejection, the.
lead / lag of the measured AT of-the Overtemperature and Overpower AT reactor. trips was changed from 8/3 to 12/3. Additionally, the lead / lag of the measured Tavg of the Overtemperature AT reactor trip was changed from 26/4 1
to 2.2/4.
This change to the lead / lags of the Overtemperature and Overpower AT reactor trip setpoints only impacts those transients which assume these trip functions, i.e., Uncontrolled RCCA Withdrawal at Power, the Uncontrolled Boron Dilution at Power, and Steamline Rupture at Power events. These transients are addressed in the following Section 4.3.
4.3 ACCIDENTS REANALYZED / EVALUATED All the events reanalyzed in this section use the LOFTRAN computer code.
LOFTRAN (Reference 3) is a digital computer code, devaioped to simulate transient behavior in a multi-loop pressurized water reactor system.
The program simulates the neutron kinetics, thermal-hydraulic conditions, pressurizer, steam generators, reactor coolant pumps, and control and protection systems operation. The secondary side of each steam generator utilizes a homogeneous saturated mixture for the thermal transients.
I I,
0514rlo/051387 15
L.
4.3.1 Uncontrolled RCCA Bank Withdrawal at Power m
The Uncontrolled RCCA Bank Withdrawal at Power event is described in Section 15.4.2 of the FSAR. An uncontrolled RCCA bank withdrawal at power causes a positive reactivity insertion which results in an increase'in the core heat flux. 'Since the steam generator lags behind the core power generation, there is a not increase in the reactor coolant temperature. Unless terminated by manual se automatic action, the increase in coolant temperature and power could result in DNB. For this event, the High Neutron Flux and.
Overtemperature AT reactor trips are assumed to provide protection against DNB. Therefore, thi's event was analyzed with increased time constants and the lead / lag changes to show that the DNBR limit is met.
Methods The assumptions are consistent with the FSAR for the ITDP methodology in that initial power, pressure, and RCS average temperature are assumed to be at the nominal values corresponding to 10%, 60% and 100% power.
Both minimum and
-maximum reactivity feedback cases were reanalyzed. The analysis was done using the LOFTRAN Computer Code.
Results For both minimum and maximum reactivity insertions, at the various power
~ levels analyzed, the DNBR limit is met for this event. A calculated sequence of events for a small and large insertion rate is presented on Table 4.1 for a power level of 60% of RTP.
Figures 5 through 10 show results fr;r a large reactivity insertion rate ar.d a small reactivity insertion rats for a 60%
power level. Figures 11 through 13 illustrate minimuu DNBR calcuated for minimum and maximum reactivity feedback, for power levels of l'00%, 60%, and 10% power.
Conclusions The limit DNBR continues to be met, therefore, the conclur, ions presented in the FSAR remain valid.
oS14v:1o/o513a7 16
4.3.2 Uncontrolled Boron Dilution et Poo r For the Boron Dilution at Power event, manual operation, as described in
~
Section 15.4.6 of the FSAR, the time from initiation of the event to reactor trip is determined from the Uncontrolled RCCA Withdrawal at Power analysis.
l Based upon the results of the Uncontrolled RCCA Withdrawal at Power analysis, l
the conclusions presented in the FSAR for the Boron Dilution at Power event, manual operation, remain valid, i.e., there is greater titan 15 minutes from the time of an alarm until the total loss of shutdown mcegin occurs.
4.3.3 Steamline Rupture at Power The Steamline Rupture at Power transient was analyzed consistent with WCAP-9226-R1. The analysis included the increased time constants mentioned in Section 4.1 and the lead / lag changes mentioned in Section 4.2.
For this event the design basis as described in WCAP-9226-R1 was met.
4.
3.4 CONCLUSION
The impact of the RTD bypass elimination for Catawba Units 1&2 on the FSAR Chapter 15 non-LOCA accident analyses has been evaluated.
For the events impacted, it was demonstrated that the conclusions of the FSAR remain valid.
4 oS14v:10/051387 17
1 TABLE 4.1 TIME SEQUENCE OF EVENTS FOR A RCCA BANK WITHDRAWAL AT 60% OF FULL POWER Accident Event Time (Sees)
Case A Initiation of uncontrolled PCCA 0.0 withdrawal at a high reactivity insertion rate (75 pcm/see) with I
minimum reactivity feedback Power range high neutron flux 6.4 Rods begin to drop 6.9 Minimum DNBR occurs 8.4 i
l Peak water level in the 10.4 pressurizer occurs Case B Initiation of uncontrolled RCCA 0.0 withdrawal at a low reactivity insertion rate (3 pcm/sec) with minimum reactivity feedback Overtemperature Delta-T reactor 80.7 trip signal initiated Rods begin to drop 82.7 Minimum DNBR occurs 84.2 Peak water level in the 85.2 pressurizer occurs os14v.1o/os13s7 18
4.4 INSTRUMENTATION AND CONTROL (I&C) SAFETY EVALUATION The RTD Bypass Elimination modification foi Catawba Units 1 and 2 does not functionally change the AT/T,yg protection channels. The implementation of the fast respon:c RTDs in the reactor coolant piping will change the inputs into the AT/T,yg Protection Sets 1, II, III, and IV as follows:
l 1.
The Narrow Range (NR) cold leg RTD in the cold leg manifold will be replaced with a fast response NR RfD w211 mounted in the RCP pump l
discharge pipe. The signal from tnis fast response NR RTD will perform the same function as the existing RTD Teold signal.
2.
The NR hot leg RTD in the bypass manifold will be replaced with 3 fast response NR RTDs wall mounted in hot leg scoops that are electronically averaged in the process protection system.
The signal from this average T
circuit obtained from these 3 NR T will parform the same hot hot function as the existing RTD T signal.
hot 3.
Identification of failed signals will be.by the same means as before the modifications, i.e., existing control bocrd alarms and indications.
4.
Signal process and the added circuitry to the Protection Set racks will be accomplished by additions to the 7300 racks using 7300 technology. When one T signal is removed from the averaging process, the electronics hot will allow a bias to be marually added to a 2-RTD average Thot(as opposed to a 3-RTD average Thot) in rder to obtain a value comparable with the 3-RTD average Thot prior to the failed RTD.
Other than the above changes, the instrumentation and control will remain the same and unchanged froa what has previously been reviewed by the Staff.
For exa:nple, two out of four voting logic centinues to be utilized for protection functions, with the 7300 process control bistables continuing to operate on a "de-energize to actuate" principle. Non-safety related control signals continue to be derived from protection channels.
o514v:1o/051387 19
l l
The above principles of the modification, including information presented in this report, and Figure 4, have been reviewed to evaluate conforstance to the I
Section 4 requirements of IEEE-279-1971 criteria and associated 10CFR 50 General Design Criteria (GDC), Regulatory Guide, and other applicable industry j
standards. Section 3 of the IEEE 279-1971 standard requires documentation of
- a. design basis. The infonnation presented in this report, including Figure 4, provide the documentation for the proposed design change and conform to the Section 4 requirements of IEEE-279-1971 criteria and associated 10CFR 50 General Design Criteria (GDC), Regulatory Guide..and other applicable industry standards. Section 3 of the IEEE 279-1971 standard requires documentation of i
.a design basis. Following is a discussion of conformance to pertinent I&C criteria:
Single failure criterion continues to be satisfied by this change because a.
the independence of redundant protection sets is maintained.
b.
Quality components and modules'being added are consistent with use in a Nuclear Generating Station Protection System.
c.
Equipment seismic and environmental qualification will be to IEEE standards 344-1975 and 323-1974, respectively, as described in WCAP 8587, Rev. 5 " Methodology for Qualifying Westinghouse WRD Supplied NSSS Safety Related Electrical Equipment".
d.
The changes will continue to maintain the capability of the Protection System to initiate a reactor trip during and following natural phenomena credible to the plant site to the same extent as the existing system.
e.
Channel independence and elactrical separation is maintained because the Protection Set circuit assignments continue to be Loop 1 circuits input to Protection Set I; Loop 2, to Protection Set II; Loop 3, to Protection Set III; and Loop 4 to Protection Set IV, with appropriate observance of field wiring interface criteria to assure the independence.
Output circuits are 0514v:1D/051387 20
l' the same as before except that there will be one Tcold and 3 Thot outputs to the computer sent through Class 1E isolators in each Protection L.
Set.
f.
The Section 4.7 of IEEE 279-1971 and GDC 24 requirements concerning Centrol and Protection System interaction are satisfied because, sven though control signals are derived from Protection Sets, the 2/4 voting coincidence logic of the Protection Sets is maintained.
Where a single random failure can cause a control system action that results in a generating station condition requiring protective action and can also prevent proper action of a protection system channel designed to protect against the condition, the remaining three redundant protection channels will be capable of providing the protective action even when degraded by a second random failure.
This is because even though 1/4 channels failed without p utially tripping, only 2 of the remaining 3 channels are necessary for a plant trip.
On the basis of the foregoing evaluation, it is concluded that these I&C modifications required for RTD bypass removal for the Catawba units will uset IEEE 279-1971, applicable GDC's, and industry standards and regulatory guides.
4.5 MECHANICAL SAFETY EVALUATION The presently installed RTD bypass system is to be replaced with fast acting narrow range RTD thermowells. This change requires modifications to the hot leg scoops, the crossover leg bypass return nozzle, the cold leg piping and the cold leg bypass manifcid connection. All welding and NDE will be performed per ASME Code Section XI requirements. Each of these modifications is evaluated below.
The original bypass piping which connects each hot leg to a separate bypass manifold, and the bypass manifold itself must be removed and the scoops, which are left intact inside the RCS piping, are modified to accept three fast response RTD thermowells.
[
[
0514v:1D/051387 21
C
)+*** A thermowell design will be used such that the tip of the thermowell [
J+a,c The thermowell will be fabricated in accordance with Section III of the ASME code (Class 1). The installation of the thermowell into the scoop will be performed using [.~
. Ja,e The root and final weld passes will ha examined by penetrant test (PT). Prior to welding, the surface of the scoop onto which welding will be performed will also be examined by PT per Section XI.
The cold leg RTD bypass nozzle must also be modified to accept a fast response thermowell and the bypass line removed. The nozzle must be modified to accept the fast response RTD thermowell. Additionally, a spare fast response thermowell will be added to the cold leg in the length between the reactor coolant pump discharge and the accumulator nozzle. This necessitates the creation of a new penetration into the piping. The boss for the new connection will be root welded by GTAW. Finish welding can be either GTAW or SMW. Weld inspection by PT will be performed after the root pass and the final pass.
The thermowells will extend [.
J+*** into the flow stream from the ID of the pipe. This depth has been justified based on [
]+a,c analysis.
The root weld joining the thermowells to the modified nozzles or bosses will be deposited with GTAW and the remainder of the weld may be deposited with GTAW or SMAW. Penetrant testing will be performed on the root and final passes. The thermowells and installation bosses will be fabricated in accordance with the ASME Section III (Class 1).
The three thermowells in the hot leg and the two thermowells in the cold leg provide a total of 20. Thermowells will be utilized at each of the four-loop CATAWBA units and they will parform the same function as the original bypass T
and Teold signals.
hot Os14v:1D/051387 22
The cross-over leg bypass return piping connection must be removed e.nd the nozzles capped. The cap design, including materials, will meet the pressure boundary criteria and ASME Section III (Class 1). The cap will be root welded
~
to the nozzles by GTAW and fill welded by either GTAW or SMAW. Penetrant testing will be performed on the root and final weld passes. The completed weld will be radiographically examined. Nachining of the bypass return piping, as well as any machining performed during modification of the penetrations ir, the hot and cold legs, shall be performed such as to minimize debris escaping into the reactor coolant system.
In accordance with (
)+a,c of the ASME Code, a hydrostatic test of new pressure boundary welds is required when the connection to the pressure boundary is [
]+a,c in diameter. Since the cap for the crossover leg bypass return pipe is [ )+a,e inches and the cold leg RTD connections are-[ )+8'C inches, a system hydrostatic test is required after bypass elimination at CATAWBA. Paragraph
[
1+e,c defines this test pressure to be (
)+a,c times the normal operating pressure at a temperature of [
).+a,c The integrity of the reactor coolant piping as a pressure boundary component, is maintained by adhering to the applicable ASME Code sections and Nuclear Regulatory Commission General Design Criteria 1, 14, 15, 30, 31 and 32. The pressure retaining capability and fracture prevention characteristics of the piping is not compromised by these modifications. Therefore, no unresolved safety issue is involved as defined in 10CFR 50.59.
5.0 CONTROL SYSTEM EVALUATION A prime input signal to the various NSSS control systems is the RCS average temperature (T,yg). This is calculated electronically as the average of the measured hot leg and cold leg temperatures in each loop.
i The major control systems affected are [
)+a,c The effect of the new RTD is to potentially change the time response of the T,yg channels in the various loops. However, as noted in Section 2.1, Table 2.1, the new RTD system will have a time response close to 0514v:1o/051387 23
that of the present system. There will therefore be no significant effect on the T,yg channel response, and no apparent need to revise any of the control system setpoints from those presently installed in the plant. The need to modify control system setpoints will be determined during the plant startup following the installation of the new RTD system by observing control system behavior.
6.0 CONCLUSION
S The fast response RTDs installed in the reactor coolant loop piping has undergone extensive analyses, evaluation and testing as' described in this report. The incorporation of this system into the Catawba design meets all Safety, Licensing and Control requirements necessary for continued licensed operation of the Catawba station. The analytical evaluation has been supplemented with in plant and laboratory testing to further verify system performance.
The fast-response RTDs installed in the reactor coolant loop piping adequately replaces the present hot and cold leg temperature measurement system and enhances ALARA efforts and improved plant reliability.
7.0 REFERENCES
1.
Tuley, C. R., Moomac, W. H., "RCS Flow Uncertainty for Shearon Harris Unit 1", WCAP-11168 Re"
., Proprietary, WCAP-11169 Rev. 1, Non-Proprietary, October, 1986.
2.
Catawba Final Safety Analysis Report, Amendment #33, 1986.
3.
Burnett, T.W.T., et al., "LOFTRAN Code Description," WCAP-7907-P-A (Proprietary), WCAP-7907-A (Non-Proprietary), April 1984.
4.
Chelemer, H., et al., " Improved Thermal Design Procedure," WCAP-8567-P (Proprietary,WCAP-8568(Non-Proprietary), July 1975.
oS14v:1o/0513s7 24
[-
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- a..c Figure 1.
Hot Leg RTD Scoop Modification for Fast-Response RTD Installation os14r.to/osiss7 -
25
r
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Cold Leg Pipe Nozzle Modification Fart-Response RTD Installation l
0514v:10/05 387 26 r
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Additional Boss for Cold Leg Fast-Response RTD Installation
. ost4v:1o/osisa7 27
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Figure 4.
RTD Averaging Block Diagram Typical for Each of 4 Channels 0514v:1D/051387 '
28
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Nuclear Power and Core Heat Flux for a RCCA Bank Withdrawal at 60% of Full Power with Minimum Rear.tivity Feedback (75 PCM/SEC RATE) 0514v:10/051387 20
l
- a.,c.
a.,c
=
Figure 6.
Pressurizer Pressure and Water Volume for a RCCA Bank Withdrawal at 60% of Full Power With Minimum Reactivity Feedback (75 PCM/SEC RATE) 30 0514v:10/051387
i e
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Figure 7.
Core Average Temperature and DNBR RCCA Bank Withdrawal at 60% of Full Power With Minimum Reactivity Feedback (75 PCM/SEC RATE) 0514v:1D/051387 31
)
Ast i
(
A > C-
?
s Figure 8.
Nuclear Power and Core Heat Flux for a RCCA Bank Withdrawal at 60% of Full Power With Minimum Reactivity Feedback (3 PCM/SEC RATE) 0514v:1D/051387 32
1 I
- a., c.
ex-i c.
k.
'4 Figure 9.
Pressurizer Pressure and Water Volume for a RCCA Bank Withdrawal at 60% of Full Power With Minimum Reactivity Feedback (3 PCM/SEC RATE) 0514v:1D/051387 33 i
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A.3 c I
n.
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Figure 10. Core Average Temperature and DNBR RCCA Bank l
Withdrawal at 60% of Full Power With Minimum Reactivity Feedback (3 PCM/SEC RATE) j 0514v:1D/051387 34
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f I
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Figure 11. Minimum DNBRE vs. Reactivity Insertion Rate; RCCA Bank Withdrawal From 100% Power 0514v:10/051387 35
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G Figure 12. Minimum DNBRE vs. Reactivity Insertion Rate; RCCA Bank Withdrawal From 60% Power 0514v:1D/051387 36
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Figure 13. Minimum DNBRE vs. Reactivity Insertion Rate; RCCA Bank Withdrawal From 10% Power
^
F 0514v:1D/051387 37
WESTINGHOUSE CLASS 3 To WCAP 11309, Rev. 1 RCS N0ZZLE AND THERM 0WELL LOCATIONS i
l l
l l
~...
This attachment consists of a copy of Duke Power Company Draving Number CN-2680-1. The drawing shows the RCS penetrations (nozzles and thermowells) for Catawba L' nit 2.
The drawing revisions associated with the installation of the RTD/thermovell system are marked 8A.
The locations of the thermovells at the McGuire Nuclear Station will be similar.
l The identification numbers for the thernovells are:
Loop A Loop B Loop C Loop D Hot Leg RTDs 1-7 2-6 3-6 4-6 (Narrow Range) 1-8 2-8 3-7 4-7 1-9 2-9 3-8 4-8 Cold Leg RTDs 1-6 2-5 3-5 4-5 (Narrow Range) 1-20 2-18 3-19 4-9 Wide Range HL 1-11 2-7 3-9 4-10 Wide Range CL 1-5 2-4 3-4 4-4 Note: Wide t'ange RTDs used for indication only. Protection System and calorimetric procedure utilize narrow range RTDs.
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WESTINGHOUSE CLASS 3 To WCAP 11309, Rev. 1 CATAWBA RTC BYPASS ELIMINATION UNCERTAINTY CALCULATIONS l
l.
i 1
1 ROD CONTROL SYSTEM ACCURACY Tavo TURB PRES
+a.c PMA =
SCA n M&TE=
STE =
=
BIAS =
RCA =
M&TE=
M&TE=
RTE =
RD
=
CA
= '
BIAS =
w RTDs USED -
TH = 2 TC = 1 ELECTRONICS CSA
=
ELECTRONICS SIGMA
=
CONTROLLER SIGMA
=
CONTROLLER BIAS
=
CONTROLLER CSA
=
l
s.
l LOW FLOW REACTOht TRIP
% DP SPAN
% FLOW SPAN
+a,c PMA1 =
PMA2 =
PEA -. =
SCA
=
l SPE
=
.STE
=
=
BIASF=
BIAS 1=.
BIAS 2=
=
M&TE =
RCSA =
RTE
=
RD
=
BIAS =
= 0 TO 120.0 % FLOW INSTRUMENT RANGE 120.0 % FLO'A FLOW SPAN
=
+a.c t
3% FLCW SAFETY ANALYSIS LIMIT
=
88.8 % FLOW ALLOWABLE VALUE
=
' NOMINAL TRIP SETPOINT =
90.0-% FLOW
+a,c 1.41 C
3 0.60 2
S
=
=
+a,c TA
=
2.50 C
3
- p i
!~
OVERTEMPERATURE; DELTA-T TRIP
.)
DELTA-T.
Tavg PRESS DELTA-I
- a,c PMA =
i SCA =
'M&TE=
1
~
STE =
=
BIAS =
=
M&TE=
'M&TE=
RCSA=
]
l RTE =
RD
=
=
m'OF RTDs USED TH = 2 TC = 1 88.5 DEGF INSTRUMENT SPAN
=
+a,c SAFETY ANALYSIS LIMIT
=t 3
3.05 % DELTA-T SPAN ALLOWABLE VALUE
=
+a,c MAXIMUM VALUE
0 3% DELTA-T SPAN 1.3800 K3
0.001189 NOMINAL SETPOINTS K1 =
59.0 DEGF DELTA-I GAIN =
1.64 j
VESSEL DELTA-T
=
1
+a,c PRESSURE GAIN
=t 3
0.95 S1
=
1.70 S2
=
+a,c Z
=
5.41 T
= t 3
+a,c 8.93 C
3
.T A
=
l l
. OVERPOWER DELTA T TRIP DELTA-T Tave
+ a. c..
PMA =
.SCA =
=
BIAS =
RCA =
M&TE=-
M&TE=
RCSA=
RTE =
RD ' =
= OF RTDs USED
~TH = 2 TC = 1 88.5 DEGF INSTRUMENT SPAN
=
+a,c SAFETY ANALYSIS LIMIT
=C 3
2.82 % DELTA-T SPAN ALLOWABLE VALUE
=
+a,c MAXIMUM VALUE
=C 3
1.0704 NOMINAL SETPOINT
=
59.0 DEGF VESSEL DELTA-T
=
+a,c S
as 1.70
-2
=
1.24 T
=C 3
+a,c
.TA
=
4.92 CSA = C 3
? -
CALORIMETRIC RCS FLOW MEASUREMENT UNCERTAINTIES
+a,e BIAS VALUES; FEEDWATER PRESSURE DENSITY ENTHALPY STEAM PRESSURE ENTHALPY PRESSURIZER PRESSURE ENTHALPY - HOT LEG ENTHALPY - COLD LEG SPECIFIC VOLUME - COLD LEG FLOW BIAS. TOTAL VALUE
- ,**,+,++ INDICATE SETS OF DEPENDENT PARAMETERS
+a,c SINGLE LOOP UNCERTAINTY - ( WITHOUT BI AS VALUES )
N LOOP. UNCERTAINTY (WITHOUT BIAS VALUES)
N LOOP UNCERTAINTY (WITH BIAS VALUES) e f
-._____m.__
__m_..____,m