ML20235T999

Rev 2 to RTD Bypass Elimination Licensing Rept for Catawba Units 1 & 2

ML20235T999
Person / Time
Site:	Catawba
Issue date:	09/30/1987
From:	Etling R, Owoc R, Rice W WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML19304B568	List: ML20235T930 ML20235T974 ML20235T986 ML20235T999
References
WCAP-11309, WCAP-11309-R02, WCAP-11309-R2, NUDOCS 8710130317
Download: ML20235T999 (44)

Text

WESTINGHOUSE CLASS 3 WCAP-11309 Rev 2 RTD BYPASS ELIMINATION LICENSING REPORT FOR CATAWBA UNITS 1 & 2 September, 1987-

. W. R. RICE R. H. 0 WOC

~

1 APPROVED: n9 R[R.Etlin[-Manager APPROVED: [ mM D. P. Dominic.is, Manager BOP & Turbine Island Systems Operating Plant Licensing-II i

l Westinghouse Electric Corporation Pittsburgh, PA Fi*'iB8M glasgi ,

0921v:10/092187 PD

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ACKNOWLEDGEMENT The authors wish to recognize contributions by the following individuals:

R. CALVO R. A. HOLMES-l 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 o

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TABLE OF CONTENTS Section Page .

1.0 Introduction' 1.

1.1 Historical Background 1 1.2 Mechanical Modifications 2 13 Electrical Modifications 4 2.0 Testing 5 2.1 Response Time Test 5 ,

2.2 Streaming Test 6 3.0 Uncertainty Considerations 7 3.1 Calorimetric Flow Measurement Uncertainty 8

^

3.2 Hot Leg Temperature Streaming Uncertainty 13 4.0 Safety Evaluation 15 4.1 Response Time 15

4.2 RTD Uncertainty 15 4.3 Accidents. Reanalyzed / Evaluated 16 4.4 Instrumentation and Control Safety Evaluation 20 l

, 4.5 Mechanical Safety Evaluation 22 l l

I 5.0 Control System Evaluation 25' 6.0 Conclusions 25 7.0 References 26 i e

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LIST OF TABLES 4

Table Title Page

, 2.1 Response Time Parameter for RCE Temperature 6 Measurement 3.1 Flow Calorimetric Instrumentation Uncertainties 9

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3.2 Flow Calorimetric Sensitivities 10 3.3 Calorimetric RCS Flow Measurement Uncertainties 11 3.4 Cold Leg Elbow Tap Flow Uncertainty 12 4.1 . Time Sequence of Events for a RCCA Bank Withdrawal 19 at 60% of Full Power j j

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LIST OF FIGURES

, Figure Title P_ age 1 Hot Leg RTD Scoop Modification for Fast-Response 27 l

- RTD Installation 2 Cold Leg Pipe Nozzle Modification Fast-Response 28 J RTD Installation '

3 Additional Boss for Cold Leg Fast-Response RTD 29 Installation 4 RTD Averaging Block Diagram, Typical for Each of 4 30 Channels l l

5 Nuclear Power and Core Heat Flux for a RCCA Bank Withdrawal 31 at 60% of Full Power with Minimum Reactivity Feedback l (75 PCM/SEC Rate) 6 Pressurizer Pressure and Water Volume for a RCCA Bank 32 Withdrawal at 60% of Full Power with Minimum Reactivity Feedback (75 PCM/SEC Rate) 7 Core Average Temperature and DNBR RCCA Bank Withdrawal 33 at 60% of Full Power with Minimum Reactivity Feedback (75 PCM/SEC Rate) 8 Nuclear Power and Core Heat Flux for a RCCA Bank Withdrawal 34 at 60% of Full Power with Minimum Reactivity Feedback (3 PCM/SEC Rate) ,

9 Pressurizer Pressure and Water Volume for a RCCA Bank 35 Withdrawal at 60% of Full Power with Minimum Reactivity Feedback (3 PCM/SEC Rate) 10 Core Average Temperature and DNBR RCCA Bank Withdrawal 36 at 60% of Full Power with Minimum Reactivity Feedback (3 PCM/SEC Rate) 11 Minimum DNBR vs. Reactivity Insertion Rate; RCCA Bank 37 Withdrawal From 100% Power 12 Minimum DNBR vs. Reactivity Insertion Rate; RCCA Bank 38 Withdrawal From 60% Power 13 Minimum CNBR vs. Reactivity Insertion Rate; RCCA Bank 39 ,

Withdrawal From 10% Power r

0921v:1o/092287

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1.0 INTRODUCTION

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,, 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 1 installed in the reactor coolant loop piping. This report is submitted in l support of continued operation of the Catawba Units with the new RTD System installed.

1.1 HISTORICAL BACKGROUND l 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 l 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.

o Temperature streaming conditions; the incomplete mixing of the coolant leaving regions of the reactor core at different temperatures produces significant temperature gradients within the pipe.  ;

o Cooling and draining of the loops before the RTDs could be replaced.

The RTD bypass system was designed to resolve these problems; however, operating plant experience has now shown that operation with the RTD bypass loops has caused some new problems:

o Plant shutdowns caused by excessive primary leakage through valves, flanges, etc., or by interruptions of bypass flow due to valve stem failure. '

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o Increased rediation exposure due to maintenance on the bypass line and to crud traps which increase radiation ay.posure throughout the loop

, compartments.

. 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 plant draindown.

Following is a detailed description of the effort required to perform this modification.

. 1 1.2 MECHANICAL MODIFICATIONS The individual loop temperature signals required for input to the Reactor I Control and Protection System will be obtained using RTDs installed in each reactor coolant loop.

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1.2.1 Hot Leg a) Loops A, C, and D 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 r

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l

)

I l thermowells will be located within the three existing RTD bypass manifold j scoops. A hole will be drilled through the end of each scoop so that l

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. water will flow in through the existing holes in the leading edge of the scoop, past the RID, and out through the new hole (Figure 1).

b) Loop B i

Two of the three thermowells will be mounted as described in (a) above.

The third thermowell cannot be installed in the existing scoop 1ccation due to structural interference. Therefore, this thermowell will be i relocated downstream of the existing scoop. This thermowell will be ]

mounted in an independent boss (Figure 3) and the resulting unused hot leg scoop will be capped. These three RTDs in each loop will measure the hot  !

leg temperature which is used to calculate.the reactor coolant loop j differential temperature (AT) and average temperature (1,yg).

c) '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 I coolant temper.ature during startup, shutdown, and post accident conditions.

1.2.2 Cold Leg i

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 (Figure 2) to, accept the RTD thermowell.

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b) This modification will not affect the single wide range RTD in each cold leg. currently installed at the discharge of the reacter coolant pump.

This RTD will continue to provide the cold leg temperature used to monitor reactor coolant temperature during startup, shutdown, and post accident conditiens. 4 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.

l This will give the new modification a tolerance for RTD failures equivalent to the bypass loops. A new cold leg boss will be added (Figure i i 3) to 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.

)

1.3 ELECTRICAL MODIFICATIONS 1.3.1 Function 1 .

l l Figure 4 shows a block diagram of the modified electronics. The hot leg RTD 1

measurements (three per loop) will be electronically averaged in the process hot signal will then be input to the protection system. The averaged T 1 appropriate protection function. This will be accomplished by additions to the existing 7300 equipment.

1.3.2 Qualification l

l 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".

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1 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 l manual rescaling of the electronics to average the remaining signals l (Figure 4).

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.

2.1 RESPONSE TIME TEST Westinghouse has performed an RTD Response Time Test at its Forest Hills 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 equal the high velocity Reactor Coolant System flows of approximately [ J+". The RTD's response time is determined based on a co,mparison 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.

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TABLE 2.1 1

, RESPONSE TIME PARAMETERS FOR RCS TEMPERATURE MEASUREMENT l Fast Response RTD Bypass System Thermowell RTD System i

_ _ +a,c _

_ +a,c RTD Bypass Piping and Thermal Lag (sec) l RTD Response Time (sec)  ;

RTD Filter Time Constant (sec)

Electronics Delay (sec)

Total Response Time (sec) 6.0 sec 7.0 sec

. i 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. '

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2.2 STREAMING TEST 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 [ )+b,c e A test program was implemented at McGuire Unit 1 to confirm the temperature streaming magnitude and stability with measurements of the RTD bypass branch line temperatures on two 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.

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The McGuire Unit 1 test data has been reduced and characterized to answer the three cbjectives of the test program. First, it is conservative to state that

. the streaming pattern [ ]+b,c.e Steady state data taken at 100% power for a period of four weeks indicates that the streaming pattern [ ]+b,c.e In other words, the temperature gradient [

)+b,c.e This is inferred by [

]+b,c.e observed between branch lines. Since the [

3+b,c e 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 l

recorded prior to the RTD failure and determine an [

]+b,c,e into the "two RTD" average to obtain the "three RTD" expected reading. This significantly reduces the error introduced by a failed RTD.

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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 did add a dimension previous tests did not have. The b ?g re Unit I test l sampled temperatures from the pipe interior while all previous tests investigated temperature gradients at the pipe surface. The pipe internal 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 reactor coolant piping.

3.0 UNCERTAINTY CONSIDERATIONS This new method of hot 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

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Temperature Streaming Uncertainty.

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i 3.1 cal.0RIMETRIC FLOW MEASUREMENT UNCERTAINTY Reactor coolant flow is verified with a calorimetric measurement performed after the return to full power operation following a refueling shutdown. The .

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.. two most important instrument parameters for the calorimetric measurement are  ;

1 the narrow range hot leg and cold leg coolant temperatures. The accuracy of j the RTDs has, therefore, a major impact on the accuracy of the flow j l

measurement. i

-The current licensed flow measurement uncertainty for Catawba for the sum of j the four loop flows including elbow taps, is about + 2.1% flow (not including ]

0.1% flow for feedwater venturi fouling allowance). However, with the use of j three T Hot RTDs (resulting from the elimination of the RTD Bypass lines) and i the latest Westinghouse RTD cross-calibration procedure (resulting in lower RTD calibration uncertainties at the beginning of a fuel cycle), it is

)

possible to reduce the RCS flow measurement uncertainty to approximately  !

[ ]+a,e flow (including the cold leg elbow taps and excluding feedwater venturi fouling). 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 instrumeat uncertainties, calorimetric sensitivities, and flow uncertainties.

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i TABLE 3.1'-

FLOW CALORIMETRIC INSTRUMENTATION UNCERTAINTIES ,

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(% SPAN) FW TEMP FW PRES FW d/p STM PRESS T T PRZ PRESS H c SCA = +a c j

.M&TE =  !

SPE =

STE = '

SD =

R/E =

i RDOT =

BIAS = i CSA =

)

, _ i i

1. OF INST USED 3 1 4 ** i I

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• F psia  % d/p psia 'F 'F psia i l

l l INST SPAN = 692. 2000. 120. 1500. 100. 100. 800.  !

l INST UNC. -

(RAND 0M) = +a,e INST UNC.

=

(BIAS) ,

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= 440. 1100. 1000.

NOMINAL 620.0 561.6 2250. l

[

)+a,c Number of Hot Leg and Celd Leg RTDs used for measurement in each loop I

and the number of Pressurizer Pressure transmitters used overall, i.e., one per loop. -

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l TABLE 3.2

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FLOW CALORIMETRIC SENSITIVITIES FEEDWATER FLOW F, -

-TEMPERATURE = +a,c MATERIAL =-

DENSITY TEMPERATURE =

PRESSURE =

DELTA P =

FEEDWATER ENTHALPY TEMPERATURE =

PRESSURE = '

h = 1192.9 BTV/LBM s

h = 419.5 BTU /LBM f

=

Dh(SG) 773.4 BTU /LBM STEAM ENTHALPY -

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PRESSURE = +a,e MOISTURE =

HOT LEG ENTHALPY TEMPERATURE =

PRESSURE =

l h =

643.0 BTU /LBM H

b =

c 561.7 BTU /LBM

=

Dh(VESS) 81.3 BTV/LBM 1 Cp(T ) =

1.553 BTU /LBM *F H

COLD LEG ENTHALPY l

TEMPERATURE = +a,c PRESSURE = i Cp(Te ) =

1.273 BTU /LBM 'F COLO LEG SPECIFIC VOLUME ,

! =

TEMPERATURE +a,c  !

PRESSURE =

l l

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TABLE 3.3

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CALORIMETRIC RCS FLOW MEASUREMENT UNCERTAINTIES l

COMPONENT INSTRUMENT ERROR FLOW UNCERTAINTY i FEEDWATER FLOW - -

+a,c VENTURI )

THERMAL EXPANSION COEFFICIENT l TEMPERATURE I MATERIAL j DENSITY '

i TEMPERATURE l

PRESSURE ,

l DELTA P FEEDWATER ENTHALPY TEMPERATURE PRESSURE STEAM ENTHALPY PRESSURE MOISTURE ,

NET PUMP HEAT ADDITION

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HDT LEG ENTHALPY TEMPERATURE STREAMING, RANDOM ,

STREAMING, SYSTEMATIC PRESSURE

)

COLD LEG ENTHALP'Y TEMPERATURE PRESSURE COLD LEG SPECIFIC VOLUME TEMPERATURE PRESSURE >

l RTD CROSS-CAL SYSTEMATIC ALLOWANCE 4 l

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TABLE'3.4-COLD LEG ELBOW TAP FLOW UNCERTAINTY INSTRUMENT UNCERTAINTIES

% d/p SPAN  % FLOW PMA = +a,c PEA =

SCA =

SPE =

STE =

SD =

RCA =

M&TE =

RTE =

RD =

ID =

A/D =

RDOT =

BIAS =

FLOW CALORIM. BIAS =

FLOW CALORIMETRIC =

INSTRUMENT SPAN =

SINGLE LOOP ELBOW TAP FLOW UNC = +a;c N LOOP ELBOW TAP FLOW UNC =

N LOOP RCS FLOW UNCERTAINTY -,

=

(WITHOUT BIAS VALUES)

N LOOP RCS FLOW UNCERTAINTY

- =

(WITHBIAS. VALUES) 0921v:1D/092187 12 1

'3.2 HOT LEG TEMPERATURE STREAMING UNCERTAINTY l

3 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 .

l at different temperatures. 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 l

i possible temperature distributions within the hot leg pipe. The test data has l

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 l i

temperature. The most recent calculations for the thermowell RTD system have j established an overall streaming uncertainty of [ )+b,c.e for a hot leg measurement. Of this total, [

]+b,c.e This overall temperature-streaming uncertainty is applicable to loops A, C and D (thermowells in 3 )

existing scoops) as well as loop B with the relocation of the third thermowell downstream of the existing scoops.

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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, [ ,

]**'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

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relative to the scoop measurement if the temperature gradient!6ver the 5 inch scoop span is nonlinear. Assuming that the maximum inferred temperature

.. gradient of [ ]+b,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 nonlinearities'at each scoop are random, the effect of this error on the hot leg temperature measurement is limited to [ ]+b,c.e.j.0n the other hand, imbalanced scoop flows can introduce temperature measurement uncertainties of up to [

3+a,c ,

In all cases, the flow' imbalance uncertainty will equal or exceed the i [ ]+b,c.e sampling uncertainty for the thermowell RTDs, so the:new measurement system tends to be a more accurate measurement with respect to stre wing uncertainties.

Temperature streaming measurements from the test at McGuire Unit I have been obtained. The me2surements conf.irm the [

.o

)+b.c.e ,

Over the 4-week testing period, there have been only mEer variations of.less thar/ (' ]+b,c e in the temperature differentials between scoops, and smaller variations in the average value of the temperature t

differentials. [

)+b,c.e ,

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Provisions were made 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-Unit l' test data,jthe bias would be limited to between [ 3+b,c.e ,

. Data comparisons show that the magnitude of this bias varied less than 1

,[ J+b,c.e over' the test period.

4.0 SAFETYtiVALUATION 4.1 RESPONSE TIME

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The primary impact of the'RTD bypass elimination on the FSAR Chapter 15 non-LOCA safety analyses (Reference 2) is the increased response time l 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 i 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. .

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This increased RTD response time results in longer delays from 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 a  !

sThe proposed fast response thermowell RTD system will make itse of RTDs

manufactured by the RdF Corporation with a total uncertainty of [ J+a,c assumed for the analyse's. These are the same RTDs as currently installed in the plant. '

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.- l

-]

rd' ,

1 I i 7he FSAR, analyses makeNxp' licit allowances for instrumentation errors for some of be rehter protection system setpointsi 'In addition, allowanct.a are made

- for the initial average reactor coolant sydtem (RCS) temperature 3 pressure and power as d(s:ribed in FSAR Section 15.0. Th'ere allowances are made explicitly to the initial conditions for ndn-DNB m$$tI; for DNB events these alloivances are statistically conbined'into the design liMt DNBR value, consistent with the Improved Thermal Onsign Procedure, gTDP) trethodology (Reference 4).

The folloiing protecticn and control ' system, parameters were affected by the changefro$oneT RTD, to three T RTDs; the Overtemperature AT hot hot (OTATP, Overpqwdr AT (OPAT), and Low RCS Finw reactor trips, the RCS a <

average tempr;ature measurements used for control board indiction and input to l

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the rod control system, and the calculated value of the RCS flow uncertainty.

SEstemuncertaintycalculationswereperformecfcftheseparametersto det' ermine the impact of the change in number cf T RTDs. The results of hot tNse calculations f show stdficVent nrgin saists to account for all kr.own instrumentuneprtainties,l after 'tne ' adjustment of the OTDT K1 nom gak, from l its current value of 1.411 to 1.38. As a result, to ensure adequato mpg in to an Overtymporatw e AT reactor trip" exists f:;r a large load rejection, the lead / lag'of the Lee.sured AT of the Overtemperaturryand Overpower AT

, reactor tr os was changed from 8/3 to 12/3. %ddittonally, the lead / lag of the

, 1 measured Tavg of the Overtemperature AT react W tA ) was changed from 28/4 to 22/4.'

,- 3 $j

) h This che ge to the lead / lags of the Overtemperature and Overpower AT reactor trip setCoira only impacts those transients whi[h assume these trip l

5 functions,s i.e., Uncontrolled RCCA Withdrawal at Per,ir, the Uncontrolled Boron DilutionatPower,arpteamlineReptureatPssrevents. These transients are addressed in the iallcwing Section 4.3.

) -

4.3 ACCIDENTS REINALYZED/ EVALUATED 1

a. t 3

1 6 ' l. , ,

, All the even'ts reanalymj in this sectice u:a thq LCETRAN computer code.

LOFTRAN (Refernce 3) is 'a digital compbt9r* code, developed to simulate transien+. behavior in a multi-loop pressurized ,

water reactor system. The

,f 6- g as is

+

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s ki f 9 1 ' /1

l 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.

4.3.1 Uncontrolled RCCA Bank Withdrawal at Power 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 net increase in the reactor coolant temperature. Unless terminated by j manual or 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, this event was analyzed with increased time constants and the lead / lag changes to show that the DNBR limit is met.

1 Methods P

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 I 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 cialculated 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 for a large reactivity insertion rate and a small reactivity insertion rate for a 60%

~

power level. Figures 11 through 13 illustrate minimum DNBR calcuated for

, minimum and maximum reactivity feedback, for power levels of 100%, 60%, and 10% power.'

0921v:1D/092187 17

l Conclusions

, The limit DNBR continues to be met, therefore, the conclusions presented in the FSAR remain valid. j 4.3.2 Uncontrolled Boron Dilution at Power 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.

Based upon the results of the Uncontrolled RCCA Withdrawal at Power analysis,  ;

the conclusions presented in the FSAR for the Boron Dilution at Power event, I manual operation, remain valid, i.e., there is greater than 15 minutes from the time of an alarm until the total loss of shutdown margin occurs. 1 4.3.3 Steamline Ruoture at Power l The Steamline Rupture at Power transient was analyzed consistent with WCAP-9226-RI. 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. ,

1 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.

0921v;1o/092187 IS l

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TABLE 4.1 TIME SEQUENCE OF EVENTS FOR A RCCA BANK WITHDRAWAL AT 60% OF FULL POWER ,

Accident Event Time (Secs) .

Case A Initiation of uncontrolled RCCA 0.0 withdrawal at a high reactivity-insertion rate (75 pcm/sec) with-minimum reacth'.ty feedback.

i l'

Power. range high' neutron flux 6.4 Rods begin to drop 6.9 Minimum DNBR occurs 8.4 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 1 l

Overtemperature Delta-T reactor 80.7 trip signal initiated j i

i Rods begin to drop 82.7 l

Minimum DNBR occurs 84.2 )

~

l l 1

l Peak water level in the 85.2 pressurizer occurs i i

DB21v:1o/092187 19

4.4 INSTRUMENTATION AND CONTROL (I&C) SAFETY EVALUATIGN

. The RTD Bypass Elimination modification for Catawba Units 1 and 2 does not functionally change the AT/T,yg protection channels. The implementation

. of the fast response RTDs in the reactor coolant piping will chenge the inputs into the AT/T,yg Protection Sets I, II, III, and IV as follows:

1

1. The Narrow Range (NR) cold leg RTD in the cold leg manifold will be t replaced with a fast response NR RTD well mounted in the RCP pump discharge pipe. The signal from this fast response NR RTD will perform the same function as.the existing RTD Tcold signal.

l

2. The NR hot leg RTO in the bypass manifold will be replaced with 3 fast response NR RTDs well mounted in hot leg scoops that are electronically averaged in the process protection system. The signal from this average q hot circuit obtained from these 3 NR T hot will perform the same T

function as the existing RTD T hot signal.

j l

3. Identification of failed signals will be by the same means as before the modifications, i.e., existing control board 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 hot signal is removed from the averaging process, the electronics one T will allow a bias to be manually 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 pri r to.the failed RTD.

Other than the above change.s, the instrumentation and control will remain the same and unchanged from what has previously been reviewed by the Staff. For example, two out of four voting logic continues to be utilized for protection functions, with the 7300 process centrol bistables continuing to operate on a "de-energize to actuate" principle. Non-safety related control signals continue to be derived from protection channels.

0921v:1o/092187 20

The above principles of the modification, including information presented in this report, and Figure 4, have been reviewed to evaluate conformance 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 a design basis. The information 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 a design basis. Following is a discussion of conformance to pertinent I&C criteria:

a. Sinole failure criterion continues to be satisfied by this change because the independence of redundant protection sets is maintained.

b. Quality components and modules being added are consistent with use in a Nuclear Gen .ating 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 electrical 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 0921v:10/092187 21

i 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

. Set. j

f. The Section 4.7 of IEEE 279-1971 and GDC 24 requirements concerning Control and Protection System interaction are satisfied because, even ,

1 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 partially 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 meet 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 piping, the hot leg scoops, the crossover leg bypass return nozzle, the l cold leg piping and the cold leg bypass manifold connection. Each of these modifications is evaluated below.

The original three branch line connections in loops A, C, and D hot legs, which feed the bypass manifold must be removed and the scoops modified to accept three fast response RTD thermowells. A[ )+a,c hole will be

, machined through the tip of the scoop to provide the proper flow path. A I

0921v:10/092187 22 l

3 I

thermowell design will be used such that the tip of the thermowell (

)+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 GTAW for the root pass and finished out with either GTAW or SMAW. The root and final weld passes will be 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.

In loop B, one hot leg fast response RTD will be installed into a new penetration downstream of the existing scoops. To accomplish this, a new boss will be installed approximately [ ]+a,c inches downstream from the existing hot leg scoops. The remaining two hot leg scoops on each loop will be utilized for thermowell mounting as was done for loops A, C, and D. The  ;

installation boss for the new penetrations and the thermowells will be root l welded by GTAW. Finish welding can be either GTAW or SMAW. Weld inspection by PT will be performed per Section XI. The installation bosses and {

thermowels are fabricated in accordance with Section III Class 1 of the ASME j Code.

Upon removal of the RTD bypass piping, the hot leg scoop not utilizied in loop

~

B will be capped. The cap will be fabricated in accordance with Section III Class 1 of the ASME Code. The root weld joining the cap to the scoop will be done by GTAW. Finish welding will be done by either GTAW or SMAW. The weld )

will be inspected by PT per the ASME Code 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 '

SHAW. Weld inspection by PT will be performed after the root pass and the i 0921v:1o/092187 23 4

.._._-_____-_.___.__._a

final pass. The thermowells will extend [ ]+a,c into the flow stream from the ID of the pipe. This depth has been justified based

, on [ ]+a,c analysis.

. These two thermocouple on each cold leg will be installed in the upper half of the piping. The root weld joining the thermowells to the modified nozzles or bosses will be deposited with GTAW and the re.mainder of the weld may be deposited with uTAW or SMAW. Penetrant testing will be performed in accordance with the ASME Code Section XI. The thermowells and installation bosses will be fabricated in accordance with the ASME Section III (Class 1).

With the three thermowells in the hot leg and the two thermowells in the cold leg, a total of 20 thermowells will be utilized at each of the four-loop CATAWBA units ahd they will perform the same function as the original bypass hot and Tcold signals.

T The cross-over leg bypass return piping connection must be removed and 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 l to the nozzles by GTAW and fill welded by either GTAW or SMAW. Penetrant tests will be performed per ASME Section XI. The completed weld will be radiographically examined. Machining of the bypass return piping, as well as any machining performed during modification of the penetrations in the hot and cold legs, shall be performed such as to minimize debris escaping into the reactor coolant system.

l In accordance with Article IWA-4000 of Section XI of the ASME Code, a hydrostatic test of new pressure boundary welds is required when the connection to the pressure boundary is larger than one inch in diameter.

Since the cap for the crossover leg bypass return pipe is [']+"'c inches and the cold leg RTD connections are [ ]+a,c inches, a system hydrostatic test is required after bypass elimination at CATAWBA. Paragraph IWB-5222 of Section XI defines this test pressure to be [ ]+a c times the normal operating pressure at a temperature of [ ].+a,c 0921v:1o/092187 24

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. The pressure retaining capability and fracture prevention characteristics of the piping is not

. compromised by these modifications. Therefore, no unresolved safety issue is i involved as defined in 10CFR 50.59.

5.0 CONTROL SYSTEM EVALUATION l

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.

The major control systems affected are [

fa.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 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 j 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 l

operation of the Catawba station. The analytical evaluation has been '

supplemented with in plant and laboratory testing to further verify system i

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.

l 0921v:1o/092187 25 l 1

l 1

- 7.0 ' REFERENCES

. .- l. Tuley, C. R., Moomau,'W. H., "RCS Flow Uncertainty for Shearon Harris Unit.1", WCAP-11168 Rev. 1, 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.

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0921v:10/092187

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. Figure 1.. Hot Leg RTD Scoop Modification for Fast-Response RTD Installation 0921v:1o/091887 27

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. Figure 2. Cold Leg Pipe Nozzle Modification Fast-Response RTD Installation 0921v:10/091gg7 28

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0921v:1D/091887 29

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! Minimum Reactivity Feedback.(75 PCM/SEC RATE)  ;

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0921v;1D/091887 31

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.i 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)

' 0921v:1o/091887 32

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, Figure 7. Cote Average Temperature and DNBR RCCA Bank Withdrawal at 60% of Full Power With Minimum Reactivity Feedback (75 PCM/SEC RATE) 0921v:10/091887 33

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0921v:10/091BB7 34 a

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, 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) 0921v:1D/091887 35

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, Figure 10, Core Average Temperature and DNBR RCCA Bank Withdrawal at'60% of Full Power With Minimum Reactivity Feedback.(3 PCM/SEC RATE) 0921v:10/091887 36 i

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1 Figure 11. Minimum DNBRE vs. Reactivity insertion Rate; RCCA Bank Withdrawal From 100% Power 0921v:10/091887 37

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l Figure 12. Minimum DNBRE vs. Reactivity Insertion Rate; RCCA Bank Withdrawal From 60% Power 0921v:10/091887 38

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. Figure 13. Minimum DNBRE vs. Reactivity Insertion Rate; RCCA Bank Withdrawal From 10% Power 0921v:1D/091887 39

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