ML20247M850

Resistance Temp Detector Bypass Elimination Licensing Rept for VC Summer Nuclear Station

ML20247M850
Person / Time
Site:	Summer
Issue date:	04/30/1989
From:	Proviano M, Waters R WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML19292J364	List: ML20247M808 ML20247M814 ML20247M828 ML20247M850
References
WCAP-12190, NUDOCS 8908020239
Download: ML20247M850 (55)

Text

- - _ - . _ - _ - - - - _ _ _ - _ - .- _ ___ . _ . - _ _ _ _ - _ - _ _ _ _ _ .

WESTINGHOUSE PROPRIETARY CLASS 3 r

WCAP-121'90 j .

f E

ll RTD BYPASS ELIMINATION LICENSING REPORT FOR o , V. C. SUMMER NUCLEAR STATION o

M. J. PROVIAN0' R. M. WATERS APRIL, 1989 a

Westinghouse Electric Corporation Pittsburgh, PA

~ s908020239 890721 PDR ADOCK 05000395 P PDC ,

1789v:1D/C22BB9

4 p.

l L, ACKNOWLEDGEMENT-u The authors wish to recognize ' contribution by the following individuals:

W.~G. Lyman C.'R. Tuley-

! G. E..Lang i, P. W. Robertson R. A. Calvo l.. D. J. Shimeck J. S. Galembush J. -

- l 1789v:1D/050389

l

,i' TABLE OF CONTENTS Section Pace List of Tables iii p

List of figures iv

' 1.0 Introduction 1.1 Historical Background 1 1.2 Mechanical Modifications 2

.1.3' Electrical Modifications 4 L

2.0 Testing 2.1 Response Time Test 9 2.2 Streaming Test 9 3.0 Uncertainty Considerations h 3.1 Calorimetric F1rw Measurement Uncertainty 12

.- 3.2 Hot Leg Temperature Streaming Uncertainty 12 3.3 Control and Protection Function Uncertainties 15 E 4.0 Safety Evaluation 4.1 Response Time 27 4.2 RTD Uncertainty 27 4.3 Non-LOCA Evaluation 28 4.4 LOCA Evaluation 30 4.5 Instrumentation and Control Safety Evaluation 31 4.6 Mechanical Safety Evaluation 34 4.7 Technical Specification Evaluation 36

- \

l 1789v:1D/022 TBS i

TABLE OF CONTENTS (Cont)

Section Page 5.0 Control System Evaluation 37 6.0 Cor.clusions 38 7.0 References 39 Appendix A - Definition of An Operable Channel And 40 Hot Le; RTD Failure Compensation Procedure 1789:1D/022889 ii

['

.7 LIST OF TABLES

' Table- Titla Pace 2.1-1 . Response Time Parameters for RCS Temperature Measurement 11

- 3.1-1 Rod Control. System' Accuracy 16 3.1-2 Flow Calorimetric Instrumentation Uncertainties 17 3.1-3 Flow Calorimetric Sensitivities 18 3.'l Calorimetric RCS Flow Measurement U:. certainties 19

' 3.1-5 Overtemperature Delta-T Trip 21 3.1-6 Overpower Delta-T Trip 22

. 3.1-7 T,yg-Low-Low-irip 23

- 3.1-8 Cold Leg Elbow Tap Flow Uncertainty 24 3.1-9 Low Flow Reactor Trip 25 3.1-10 Technical Specification Modification 26 4 '

O 1789v;1o/022BB9 iii

LIST OF' FIGURES i

Figure Title Page.

1.2 . Hot Leg RTO Scoop Modification for Fast-Response 6 RTD Intta11ation

. 7.,

1.2-2 Cold Leg Pipe Nozzle Modification Fast-Response 7 RTD Installation i

1.3-1 RTD Averaging Block Diagram,

• Typical for Each of 3 8 Channels i

e

! t 1789v.1o/022BB9 iV

1:

! I

1.0 INTRODUCTION

Westinghouse Electric Corporation has been contracted by South Carolina Electric and Gas to remove the existing Resistance Temperature Detector (RTD) l Bypass System and replace this hot leg and cold leg temperature measurement method with fast response thermowell mounted RTDs installed in the reactor coolant loop piping. This report is submitted for the purpose of supporting operation of V. C. Summer Nuclear Station utilizing the new thermowell mounted RTDs.

1.1 HISTORICAL BACKGROUND 1

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

O 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 The reactor coolant loops required cooling and draining 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 created its own obstacles such as:

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

1789v;1o/042889 1

4 o Increased radiation exposure due to maintenance on the bypass line and i

~

to crud traps which increase radiation exposure 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).

e Three thermowell mounted hot leg RTDs provide an average measurement (equivalent to the temperature measured by the bypass system) to account for temperature streaming.

o Use of thermowells permits RTD replacement without draining the reactor coolant loops.

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

1.2 MECHANICAL MODIFICATIONS The individual loop temperature signals required for 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 i.easurement on each loop will be accomplished with three fast response, narrow range, dual element RTDs mounted in thermowells. One element of the RTD will be considered active and the other element will be held in reserve as a spare. To accomplish the sampling function of the RTD bypass manifold system and minimize the need for additional hot leg piping penetrations, the thermowells will be 1789v:1o/042BE9 2

i 4.-

located within the three existing RTD bypass manifold scoops. A hole will-be made through the'and 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 Sie (Figure 1.2-1). These three RTDs will measure the l hot. leg temperature which is used to calculate the reactor-coolant loop differential temperature (AT) and average temperature (T,yg). l 1

b) This modification will not affect the single wide range RTD currently inctalled 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, dual-element RTD will be located in each cold leg at the discharge of the reactor coolant pump (as replacements for the cold leg RTDs 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 1.2-2) to accept the RTD thermowell. One element of the RTD will be considered active and the other element will be held in reserve as a spare.

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.

1789v:1o/0428B9 3 m

R g . 3 i.

[.

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 l

1.3.1 C_nntrol & Protection System figure 1.3-1 shows a block diagram of the modified protection system electronics. The hot leg RTD measurements (three per loop) will be

, electronically averaged in the process protection system. The averaged T hot signal will then be used with the Tcold signal to calculate reactor coolant loop AT and T,yg which are used in the reactor control and protection system. This will be accomplished by additions to the existing process protection system equipment.

The present RCS loop temperature measurement system uses dedicated direct

- immersion RTDs for the control and protection systems. This was done largely to satisfy the IEEE Standard 279-1971 which applied single failure criteria to control and protection system interaction. The new thermowell mounted RTDs

~w ill be used for both control and protection. In order to continue to satisfy the requirements of IEEE Standard 279-1971, the T,yg and AT signals used in the control grade. logic will be input into a median signal selector, which will seleci the signal which is in between the highest and lowest values of the three loop inputs. This will avoid any adverse plant response that could be caused by a single signal failure.

1.3.2 Ouelification The 7300 Process Electronics modifications will be qualified to the same level as the existing 7300 electronics. RTD qualification will be verified to support South Carolina Electric and Gas's compliance to 10CFR50.49.

l j

i 1789v:1o/o42E89 4

The Westinghouse qualification program entailed a review of the WEED Instrument Company's qualification documentation for testing performed on 1

these RTDs. It was concluded that the equipment's qualification was in

, compliance with IEEE Standards 344-1975 and 323-1974 with one exception.

SMeifically, requirements relative to flow induced vibration were not addressed. To demonstrate that flow induced vibration would not result in significant aging mechanisms that could cause common mode concerns during a seismic event, Westinghouse performed flow induced vibration tests followed by pipe vibration aging and a simulated seismic event. These tests confirmed that the WEED RTDs do comply with the above IEEE standards.

1.3.3 RTD Doerability Indication Existing control board AT and T,yg indicators and alarms will provide the means of identifying RTD failures, although the now redundant indication for the T,yg and AT signals will be removed. The spare cold leg RTD element provides sufficient spare capacity to accommodate a single cold leg RTD failure per-loop. Failure of a hot leg RTD can be handled in two ways. In the first, manual action by the operator' defeats the failed signal and rescales the electronics to average the remaining signals (see Figure 1.3-1 and Section 4.5). The second method disconnects the failed element and utilizes the second element of that same RTD.

1789v:1o/042889 5

p 5-, ,

a., c.

[> . ..

r .p

}

Figure 1.2-1 Hot Leg RTD Scoop Modification for Fast Response RTD Installation

' 1789v:1D/042BE9 6 L.

n .

i- g

^

(,

r ., x i:

H t :-

a

~

Figure 1.2-2 Cold Leg Pipe Nozzle Modification for Fast Response RTD Installation 1789v:10/042BB9 7 a-

[.

.t

(

%c

_F

/.

'A Figure 1.3-1 PTD Averaging Block Diagram.

Typical for Each of 3 Channels 1789c1D/042889 8

1 t

I 2.0 TESTING There are,two specific types of tests which are performed to support the installation of the thermowell mounted fast-response RTDs in the reactor

~

coolant piping: RTD response time tests and a hot leg temperature streaming test. The response time for the V. C. Summer Nuclear Station application will be verified by testing at the RTD manufacturer and by in-situ testing. Data from thermowell/RTD performance at operating plants provide additional support for the system.

2.1 RESPONSE TIME TEST The RTD manufacturer, WEED Instruments Inc., will perform time response -

testing of each RTD and thermowell prior to installation at V. C. Summer Nuclear Station. These RTD/thermowells must exhibit a response time bounded by the values shown in Table 2.1-1. The revised response time has been factored'into the transient analyses discussed in Section 4.0.

. .In addition, response time testing of the WEED RTDs will be performed in-situ. This testing will demonstrate that the WEED RTDs can satisfy the

- response tine requirement when installed in the plant.

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 maximum to minimum of [^ )b,c.e A test program was implemented at an operating plant to confirm the temperature streaming magnitude and stability with measurements of the RTD bypass branch line temperatures on two adjacent het leg pipe:,, 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 is used in conjunction with data taken from other Westinghouse designed plants to determine an appropriate temperature error for use in the 1789v;1o/o42ses 9

l I

safety analysis and calorimetric flow calculations. Section 3 will discuss ,

the specifics of these uncertainty considerations.

, The test data was reduced and characterized to answer the three objectives 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 months indicated that the streaming pattern

[ )b,c.e In other words, the temperature gradient [. )b,c.e This is inferred by [- 3b ,c.e observed between branch lines. Since the [ , .

)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 recorded prior to the RTD failure and determine an [

)b,c.e into the "two RTD" average to obtain the "three RTD" expected reading. A generic procedure has been provided to South Carolina

. Electric & Gas which specifies how these [ )b,c.e are to be determined (Appendix A). This significantly reduces the error introduced by a failed RTD.

Both the test data and the operating data support previous calculations of streaming errors determined from tests at other Westinghouse plants. The temperature gradients defined by the recent plant operating data are well ,

within the upper bound temperature gradients that characterize the previous data. Differences observed in the operating data compared with the previous data indicate that the temperature gradients are smaller, so the measurement uncertainties are conservative. The measurements at the operating plants, obtained from thermowell RTDs installed inside the bypass scoops, were expected to be, and were found to be, consistent with the measurements obtained previously from the bypass loop RTDs.

1789clo/042BE9 10

{ u l

TABLE 2.1-1 RESPONSE TIME PARAMETERS FOR' RCS TEMPERATURE MEASUREMENT RTD Fast Response'-

Bypass System Thermowell RTD System

~ ~

RTD Bypass Piping.and Thermal Lag (sec)

A70 Response Time (sec)

ElectronicsDelay(sec)

Margin (sec) .

Total Response Time (sec) 6.0 sec 8.5 see 4

1789v:1D/042889 11 w_--____-_-__----__________

a 3.0 UNCERTAINTY CONSIDERATIONS L- This method of hot leg temperature measurement has been analyzed to determine l

the magnitude of the two uncertainties included in the Safety Analysis:

l Calorimetric Flow Measurement. Uncertainty and Hot Leg Temperature Streaming

' Uncertainty.

3.1 CALOR! METRIC FLOW MEASUREMENT UNCERTAINTY Reactor coolant flow is verified with a calorinckic measurement performed after the return to power operation following a refueling shutdown. The two most important instrument parameters for the calorimetric measurement of RCS flow 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.

With the use of three T hot RTDs (resulting from the elimination of the RTD Bypass lines) and the latest Westinghouse RTD cross-calibration procedure (resulting in low RTD calibration uncertainties at the beginning of a fuel

~

cycle), the V. C. Summer Nuclear Station RCS Flow Calorimetric uncertainty is estimated to be [. Ja,c including use of cold leg Elbow Taps

~

(see Tables 3.1-2, 3, 4 and 5). This estimate is based on the standard Westinghouse methodology previously approved on earlier submittals of other plants associated with RTD Bypass Elimination or the use of the Westinghouse Improved Thermal Design Procedure and SCE&G 1etter to H. R. Denton (NRC) from D. A. Nauman (SCE&G), 6/27/86. Tables 3.1-1 through 3.1-10 were generated specifically for V. C. Summer Nuclear Station and reflect plant specific measurement uncertainties and operating conditions.

3.2 . 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 temperatures. This temperature streaming uncertainty is based on an analysis of test data from other Westinghouse plants, and on calculations 17a9v:1D/o42889 12

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 establisNd an overall streaming uncertainty of [ ]b c.e for a hot leg measurement. Of this total, [

.)b,c.e This overall temperature streaming uncertainty determined for plants with similar or symmetrical temperature distributions is conservative when applied to 3 loop plants such as V.C. Summer since the 3 loop temperature distributions are not similar

. resulting in a smaller systematic uncertainty for 3 loop plants.

. The new method of measuring hot leg temperatures, with the three hot leg thermowell RTDs, is at least as effective as the existing RTD bypass system, '

l

]C . Although the new method measures temperature at one point at the RlD/thermowell tip, 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 the temperature gradient over the 5-inch sco5 span is nonlinear. Assuming that the maximum inferred tempe ature 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 1789v;1o/o42889 13

leg temperature meas'urement is limited to [ Jb,c e . On the other hand, imbalanced scoop flows ca.n introduce temperature measurement

. uncertainties of up to [.

)a c ,

. In all cases . the flow imbalance uncertainty will equal or exceed the

[ ']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 have been obtained from tests at 2, 3 and

'4-loop plants and from thermowell RTD installations at 4-loop plants.

Although there have been some differences. observed in the orientation of the individual loop temperature distributions from plant to plant, the magnitude of the differences have been (

)b,c.e. ,

Over the testing and operating periods, there were 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 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 test data, the

~ bias value would be expected to range between [

)b,c.e Data comparisons show that the magnitude of this bias varied less than

[ )b,c.e over the test period. Appendix A provides a procedure for utilizing the actual plant bias data. Note that this procedure *only allows the use of positive (or zero) bias values.

9 1789v:1D/0428B9 14

i 3.3 CONTROL AND PROTECTION FUNCTION UNCERTAINTIES Calculations were performed to determine or verify the instrument

- uncertainties for the control and protection functions affected by the RTD Bypass Elimination. Table 3.1-1 (Rod Control System Accuracy) notes that an acceptable value for control is calculated. Table 3.1-2, 3.1-3 and 3.1-4 provide the uncertainties, sensitivities and final result of the Precision RCS Flow Calorimetric. The total uncertainty for the measurement of RCS Flow, as noted on Table 3.1-8, is less than the value noted in the Virgil C. Summer Technical Specifications. Table 3.1-5 provides the uncertainty breakdown for Overtemperature AT. As noted on this table TA is greater than CSA, thus acceptable results are calculated for this function. Table 3.1-6 provides the breakdown for Overpower AT, with the same conclusions as for Overtemperature AT. Table 3.1-7 notes the uncertainty breakdown for Tavg Low-Low. Again acceptable results are calculated. Table 3.1-9 is concerned with the RCS Low Flow reactor trip. Based on the earlier calculations for the RCS Flow Calorimetric and the Rod Control System Accuracy, acceptable results are determined. Finally, Table 3.1-10 notes the changes necessary to the

. Virgil C. Summer Technical Specifications. As noted, relatively minor changes are necessary to reflect the modified calculation results, primarily the

- Allowable Values.

l

~

1789v:1D/042EB9 15 m . . . .. .

f

y gp % , "' -

3, f

TABLE 3.1-1 4 ROD CONTROL-SYSTEM ACCURACY

~ T8vg ; TURB' PRES

- .+a,c PMA = ~

SCA =

M&TE=

-STE'=

SD . =

BIAS ='

RCA =

M&TE=-

M&TE=

RTE =

RD =

,  ;- CA =

BIAS =

1. RTDs USED - TH = 2 TC = 1

+a c ELECTRONICS CSA = 'l ELECTRONICS SIGMA =

CONTROLLER SIGMA =

- CONTROLLER BIAS =

CONTROLLER CSA =

_9 .

1786;1D/042889 16

t.:

TABLE 3.1-2 l

FLOW CALORIMETRIC INSTRUMENTATION UNCERTAINTIES

'(% SPAN) FW TEMP FW. PRES FW DP STM PRESS TH TC PRZ PRESS

+a,e

~

. ' SCA =

..M&TE=

g SPE =

'STE =

SD =

R/E =

RDDT=

BIAS =

CSA = ,

1. OF' INST USED 3 1 1 DEG F PSIA  % DP PSIA DEG F DEG F PSIA INST SPAN = 100. 1500. 125. 1000. 100. 100. 800.

+a,c

. INST UNC. __ __

(RANDOM)=

INST UNC.

=

(BIAS) ,

. NOMINAL = 437. 1064. 964. 618.7 556.0 2250.

1789v:1D/042889 17

i"

~  ;

TABLE 3.1-3 FLOW CALORIMETRIC SENSITIVITIES-FEEDWATER FLOW.

FA __ +a,e TEMPERATURE =

MATERIAL- =

DENSITY TEMPERATURE =

PRESSURE =

=

-DELTA.P-FEEDWATER ENTHALPY  !

TEMPERATURE =

PRESSURE = -

.h5 = 1194.2 BTU /LBM hF = 414.0 BTU /LBM

= 780.2 BTU /LBM Dh(SG)

STEAM ENTHALPY l __ +a c PRESSURE =

MOISTURE =

' HOT LEG ENTHALPY.

TEMPERATURE =

PRESSURE =

hH = 641.0 BTU /LBM -

hC = 554.7 BTU /LBM .

= 86.3 BTU /LBM Dh(VESS)

Cp(TH) =- 1.552 BTU /LBM-DEGF. 1 COLD LEG ENTHALPY l

, - +a c .

. TEMPERATURE =

l PRESSURE =

l J

Cp(TC) = 1.258 BTU /LBM-DEGF

^

COLD LEG SPECIFIC VOLUME

_ -. +a,c TEMPERATURE =

PRESSURE =

1789v:1D/042889 1B i i

m_________1_ _ _

l

L V

f TABLE 3.1-4

\. ,

'~

CALORIMETRIC RCS FLOW MEASUREMENT UNCERTAINTIES

' COMPONENT INSTRUMENT ERROR FLOW UNCERTAINTY

(% FLOW)

+a , c -

l FEEDWATER FLOW --

. VENTURI

THERMAL EXPANSION COEFFICIENT l TEMPERATURE l MATERIAL DENSITY TEMPERATURE' PRESSURE l DELTA P FEEDWATER ENTHALPY TEMPERATURE PRESSURE STEAM ENTHALPY PRESSURE-MOISTURE NET PUMP HEAT ADDITION HOT LEG ENTHALPY.

TEMPERATURE STREAMING, RANDOM STREAMING, SYSTEMATIC i PRESSURE COLD LEG ENTHALPY TEMPERATURE PRESSURE COLD LEG SPECIFIC VOLUME TEMPERATURE PRESSURE 1789v;1D/042 ass 19 i

TABLE 3.1-4 (continued)

' CALORIMETRIC RCS FLOW MEASUREMENT UNCERTAINTIES BIAS VALUES -

+a, e 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 N LOOP UNCERTAINTY (WITHBIASVALUES) l

.,i 1789v;1D/042BB9 20

c . .a TABLE 3.1-5

.. OVERTEMPERATURE DELTA-T TRIP DELTA-T Tavg PRESS' DELTA-I

~

~ PMA = -

SCA =

M&TE =

STE =

SD =

BIAS =

RCA =

M&TE =

M&TE =

RCSA =

. RTE =

RD =

SA =

( OF RTD USED- TH = 2 TC = 1 INSTRUMENT SPAN = 92.7 DE3F SAFETY ANALYSIS LIMIT =[ ]

i ALLOWABLE VALUE = 2.18% DELTA-T SPAN 1

~

lb NOMINAL SETPOINTS K1 = 1.2030 K3 = 0.001470 VESSEL DELTA-T = 61.8 DEGF DELTA-I GAIN = 2.14

+a,e PRESSURE GAIN =[ .)

i 4' - _. +a,c - - +a,c -

- +a c 2 =- T =

l TA = MAR =

l-i

.1789v:1D/042889 21

)

., TABLE 3.1-6 OVERPOWER DELTA-T TRIP DELTA-T Tavg

- - +a,e p ,

SCA =

SD =

BIAS =

RCA =

N&TE =

M&TE =

RCSA =

RTE =

RD =

f 0F RTD USED TH = 2 - TC = 1 INSTRUMENT SPAN = 92.7 DEGF SAFETY ANALYSIS LIMIT =[ ]

ALLOWABLE VALUE = 2.39% DELTA-T-SPAN NOMINAL SETPOINTS 1.D875 VESSEL DELTA-T = 61.8 DEGF

.- - +a,e - - +a,c - - +a,c 2 = 5 = T =

TA = CSA = NAR =

I 1789v:1D/050389 22

l l-l"e l

L TABLE 3.1-7 l.

r .'

l; ~ .Tavg-LOW-LOW TRIP:

+a,c PMA =

SCA =

SD - =

BIAS =

RCA =

4 l N&TE =

l RCSA =

RTE. =:

RD. =

l l

l # OF RTD USED TH = 2 TC = 1 I: e INSTRUMENT SPAN = 100.0 DEGF

+a,c SAFETY ANALYSIS LIMIT =[. )

ALLOWABLE VALUE = 548.4 DEGF NOMINAL TRIP SETPOINT = 552.0 DEGF l

- -- +a,e - - +a,c -

+a,c Z = 5 = =

T TA = CSA = NAR =

1789v:1D/042889 23

p i

lI TABLE 3.1-8 h

COLD LEG ELBOW TAP FLOW UNCERTAINTY L~ . INSTRUMENT UNCERTAINTIES

% DP SPAN  % FLOW

~

'~

PMA =

PEA =

SCA =

L-SPE

• STE =

SD =

1 RCA =

MLTE =

RTE =

l. . RD =

JD =

1 l- A/D =

RDOT =

FLOW CALORIN. BIAS =

FLOW CALORIMETRIC =

INSTRUMENT SPAN =

- - +a,e .

SINGLE LOOP ELBOW TAP FLOW UNC *  % FLOW

- N LO3P ELBOW TAP FLOW UNC =

I N LOOP RCS FLOW UNCERTAINTY _ +a,c (WITH BIAS VALUES) 1 1789v:1D/042889 24 E- ______ _

I .-

I TABLE 3.1-9

~

LOW FLOW REACTOR TRIP

.f INSTRUMENT UNCERTAINTIES.

)' DP SPAN  % FLOW SPAN PMA1 *-

I' PMA2 =

l l- PEA =

l l SCA =

SPE =

~STE =

SD =

BIASF=

BIASl=

. BIA52=

RCA =

M&TE =

RCSA =

RTE =

RD =

BIAS =

. FLOW SPAN = 120.0 % FLOW

+a.c SAFETY ANALYSIS LIMIT =[ ]

ALLOWABLE VALUE = 88.9 % FLOW NOMINAL TRIP SETPOINT = 90.0% FLOW

- +a,e ._ _

+a,e - +a,c

-. 2 = S = T =

TA = CSA = MAR =

1789v;1D/D42sts 25

/

j

. TABLE 3.1-10 L g. TECHNICAL SPECIFICATION MODIFICATIONS Overtemperature AT 2 = 7.21 4 S' 1.6 (AT) + 1.2 (pressure)

A11owa n Value 5 2.2-% AT span

. Overpower AT l

i 2 = 1'.96

.S = 1.6 Allowable Value's 2.4 %'AT span Loss of Flow

'I = 1.48 S = 0. 6 '

Allowable '! slue 3 88.9 % of Loop Design Flow

.Tavg Low-Low 2 = 0.71 S = 0.8 .

Nominal Trip Setpoint g 552.0*F Allowable Value 1 548.4*F P-12 Nominal Trip Sr:tpoint = 552.0 'F i Allowable Value 5 555.6'F Allowable Value 3 548.4*F O

i 1789v:1o/D42869 26 )

4.0 SAFETY EVALUATION-The primary impact of the RTD Bypass ElimInatir1 on the FSAR Chapter 15

~

-(Reference 1) safety' analyses are the differenos in response time characteristics and instrumentation uareitaiMies associated with the fast response thermowell_RTD system. The affects'of these differences are discussed in the following sections.

4.1 RESPONSE TIME-The response time parameters of the V. C. Summer Nuclear Station P.TD bypass system assumed in the safety analyses are shown in Table 2.1-1. For the fast response thermowell RTD system, the overall response time will consist of [.

Ja,c (as presented in Section 2.1 and as given in Table 2.1-1).

This allows the total RCS temperature measurement response time to be increased to 8,5 seconds (Reference Table 2.1-1). This response time is

. factored into the Overtemperature AT trip and Overs.ower AT trip performance. Therefore, those transients that rely on the above mentioned trips must be evaluated for the modified response characteristics. Section 4.3 includes a discussion of the analyses and/or evaluations performed for these events.

4.2 RTD UNCERTAINTY

' f

_The proposed fast response thermowell RTD system will make use of RTDs, j

_ manufactured by Weed Instruments Inc., with a total uncertainty of l

[. Ja c assumed for the analyses.

The FSAR analyses make explicit allowances for instrumentation errors for some of the reactor protection system setpoints. In addition, allowances are made for the average reactor coolant system (RCS) temperature, pressure and power.

~

These allowances are made explicitly to the initial conditions.

1ne<:10/o42ses 27 ai_---___---_------

The following protection and control system parameters were evaluated and determined to be unaffected (with respect to accident analysis assumptions) by the change from one hot leg RTD to three hot leg RTDs; the Overtemperature AT (OTDT), Overpower AT (0PDT), and Low RCS Flow reactor trip functions, RCS loop T,yg measurements used for input to the rod control system and safety injection, and the calculated value of the RCS flow uncertainty.

System uncertainty calculations were performed for tSese parameters to determine the impact of the change in the number of hot leg RTDs. The results of these calculations, noted in 3.3, indicate sufficient margin exists to account for known instrument uncertainties.

In summary, changes have been made in the Reactor Protection System response times only to account for the new thermowell mounted RTDs.

4.3 NON-LOCA EVALUATION The changes in the RTD response time discussed in Section 2.1 and the instrumentation uncertainties discussed in Section 3.3 have been considered

, for the V. C. Summer Nuclear Station non-LOCA safety analysis design basis.

Only those transients which assume OTAT/0 PAT protection are potentially

. affected by changes in the RTD response time. Instrumentation uncertainties can affect the non-LOCA transient initial condition assumptions and those transients which assume protection from low primary coolant flow reactor trip.

As noted in Section 3.0, the RTD bypass elimination can potentially affect the rod control system accuracies and flow calorimetric instrumentation uncertainties. The calculations documented in Section 3.0 support the validity of the non-LOCA safety analysis initial condition RCS temperature and flow assumptions used in the VANTAGE 5 fuel analyses (Reference 3). On this basis, the non-LOCA safety analysis initial condition assumption's are appropriate and conservative for the proposed RTD Bypass Elimination and no revision to the accident analysis initial condition assumptions used in the VANTAGE 5 fuel analyses (Reference 3) is required.

1789v;1D/o42889 28

1' The RTD response time and instrumentation uncertainties associated with the i RTD Bypass Elimination can potentially affect protection systems assumed to be available for mitigation of design basis non-LOCA transients. ihe protection system setpoints evaluated in Section 3.0 are OTAT, OPAT and Low Primary Coolant Loop Flow (Loss of Flow) reactor trip and T,y,-Low-Low coincidtace for SI. The transients which assume protection from these functions are listed below, note that the T,yg-Low-Low coincidence setpoints were not utilized in the non-LOCA transients.

1 Assumed Protection l

Reference Accident Function RTSR Section Uncontrolled RCCA Bank With 15.2.2 drawal at Power OTAT RTSR Section Uncontrolled Boron Dilution 15.2.4 OTAT

. RTSR Section Partial Loss of Forced l 15.2.5 Reactor Coolant Flow Loss of Flow

!- RTSR Section Startup of an Inactive 15.2.6 Reactor Coolant flow Loss of Flow RTSR Section Loss of External Electrical l 15.2.7 Load / Turbine Trip OTAT RTSR Section - Accidental Depressurization 15.2.12 of the RCS OTAT WCAP-10961-P Steamline Break for E0 (Revision 1) Outside Containment OPAT (Reference 2) l l

1789v:1o/o42BBB 29

/

On the basis of the information documented in Table 2.1-1, it is concluded that the V. C. Summer VANTAGE 5 fuel RTSR non-LOCA safety analysis assumptions l for the total DTAT/0 PAT trip function response times remain valid. As documented in Table 15.1-2 of. Reference 3, a total response time of 8.5  !

seconds has been assumed. Based on an examination of the assumptions and results discussed in WCAP-10961-P, Revision 1 (Reference 2) it is determined that an 8.5 second response time for the OTAT and OPAT trip functions will have a negligible impact on the mass / energy releases. Thus, the tables of I mass / energy release documented in Reference 2 reesin valid. In addition, evaluation of the effects of the RTD Bypass Elimination on the uncertainties associated with the OTAT/0 PAT setpoints, as well as the Loss of Flow 4

setpoint, supports the continuing validity of the current non-LOCA safety analysis assumptions for the transients listed above. ,

1 In conclusion, the Reference 3 non-LOCA safety analysis assumptions applicable to V.C. Summer remain valid for the replacement of the existing RTD bypass system with fast response thermowell mounted RTDs installed in the reactor coolant loop piping. In addition, an 8.5 second response time for the OTAT

. and OPAT trip functions will have a negligible impact on the mass / energy '

releases documented in Reference 2. Therefore, the conclusions in the RTSR remain valid and all applicable non-LOCA safety analysis acceptance criteria c.ontinue to be met.  !

4.4 LOCA Evaluation The elimination of the RTD bypass system impacts the uncertainties associated with RCS temperature and flow measurement. The magnitude of the uncertainties are such that RCS inlet and outlet temperatures, thermal design flow rate and i the steam generator pc.innance data used in the LOCA analyses will not be affected. Past sensitivit/ studies have shown that the variation of the core

~

inlet temperature (Tin) used in the LOCA analyses affects the predicted core flow during ths blowdown period of the transient. The amount of flow into the core is influenced by the two phase vessel-side break flow, and the core cooling is affected by the quality of the fluid. These sensitivity studies concluded that the inlet temperature effect on peak clad temperature is 1789v;1D/050389 30

dependent on break size. As a result of these studies, the LOCA analyses are, performed at a nominal value of Tin.without consideration of small uncertainties. The RCS flow rate and steam generator secondary side temperature and pressure are also determined using the loop average

- temperature (T,yg) output. ' These neminal values used as inputs to the analyses are not affected due to the RTD bypass elimination. It is concluded

'that the elimination of the RTD bypass piping will not affect the LOCA analyses input and hence, the results of the analyses for V. C. Summer Nucisar Station remain unaffected. Therefore, the plant design changes due to the RTD bypass elimination are acceptable from a LOCA analysis standpoint without requiring any reanalysis.

4.5 INSTRUMENTATION AND CONTROL (I&C) SAFETY EVALUATION

, The RTD Bypass Elimination modification for V. C. Summer Nuclear Station does not functionally change the AT/T,yg protection channels. The implementation of the fast response RTDs in the reactor coolant piping will change the inputs into the AT/T,yg Protection Sets I, II, and III, as

.. follows:

1. .The Narrow Range (NR) cold leg RTD (used in the protection system) in the cold leg manifold will be replaced with a fast response NR dual element well mounted RTD in the RCP pump discharge pipe. The signal from this fast response NR RTD will perform the same function as the existing RTD Teold signal. One element of the RTD will be held in reserve as a spare.

2. The NR hot leg RTD in the bypass manifold will be replaced with 3 fast response NR dual element well mounted RTDs in the hot leg that are electronically averaged in the process protection system.

3. Identification of failed signals will be by the same means as before the modifications, i.e., existing control board alarms and indications.

4. The RTD Bypass Elimination conditioning is accomplished via additional circuitry in the Protection Set racks. Additions to the precese centrcl 1789v:1D/042889 31 w-_________-________ _ _ _ _ . i

l l

cabinets are in the form of the Mediari Signal Selector. These changes are done using 7300 technology. J Existing control board AT and Tavg indicators and alarms will provide the means of identifying RTD failures. Upon identification of a failed hot leg RTD, the operator would place that protection channel in trip (consistent with the time requirements specified in the Technical Specifications), identify and disconnect the failed RTD, and rescale the summing amplifier for a two RTD input condition. Specifically if one hot signal is. removed from the averaging process, the electronics will T

allow a bias to be manually added to a 2-RTD average Thot (as opposed to a 3-RTD average Thot) in order to obtain a value comparable with the 3-RTD average Thot pri r to the failed RTD. An alternative procedure would be to utilize the spare hot leg element within the dual element RTD by manually connecting the spare element to the 7300 circuitry in place of the failed element. In the event of a cold leg RTD failure, the spare cold leg RTD element will be manually connected to the 7300 circuitry in place of the failed RTD. After this process, the channel would then be

. returned to service. As noted, during this rescale process the plant will >

be in a partial trip mode and will therefore be in a safe condition.

Ihe conversion to thermowell mounted RTDs will result in elimination of the control grade RTDs and their associated control board indicators. The protection grade channels will now be used to provide inputs to the control system through isolators to prohibit faults in the control rack from propagating into the protection racks.

In order to satisfy the control and protection interaction requirements of IEEE Standard 279-1971, a Median Signal Selector (MSS) will be used in the control channels presently utilizing a high auctioneered T,yg or AT signal (there will be a separrJ.a MSS for each function). The Median Signal Selector

-will use as inputs the isolated protection grade T,yg or AT signals from all three loops, and will supply as an output the channel signal which is the median of the three signals. The effect will be that the various centrol grade systems will still use a valid RCS temperature in the case of a single signal failure.

1789v:1D/050389 32 1

I

I 1

To ensure proper action by the Median Signal Selector, the present manual

. switches that allow for' defeating of a T,yg or AT signal from a single loop will be eliminated. The MSS will automatically select a valid signal'in

~the case of a signal. failure.- Warnings that a failure has occurred will be provided by loop to median T,yg and AT deviation' alarms.

Other than the above changes, the Reactor Protection ' System will reimain the same, as that previously utilized. For example, two out of. three voting 11ogic continues to be utilized for the thermal overpower protection functions, with the model 7300 process' control bistables continuing to cperate on a "de-energize to actuate" principle. Non-safety related control signals will now be derived from isolated protection channels. )

The above principles of the modification have been reviewed to evaluate conformance to the requirements of IEEE Standard 279-1971 criteria and associated 10CFR 50 General Design Criteria (GDC), Regulatory Guides, and other applicable industry standards. IEEE Standard 279-1971 require, documentation of a design basis. Following is a discussion of design basis

.. requirements in conformance to pertinent I&C criteria:

1. a. The single failure criterion continues to be satisfied by this change because the independence of redundant protection sets is maintair.ed.

i

- b. The quality of the components and modules being added is consistent with ,

use in a Nuclear Generating Station Protection System. For the

, Westinghouse Quality Assurance program, refer to Chapter 17 of the FSAR. ,

4

c. '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. J d.. Channel independence and electrical separation is maintained because the '

Protection Set circuit assignments continue to be Loop 1 circuits input to o Protection Set I; Loop 2 to Protection Set II; and Loop 3 to Protection Set III, with appropriate observance of field wiring interface criteria to assure the independence.

1789v;1D/o4288s 33

_ }

i

e. Due to the elimination of the dedicated control system RTD elements, temperature signals for use in the plant control systems must now be

~

derived from the protection system RTDs. To eliminate any degrading control and protection system interaction mechanisms introduced as a consequence of the.RTD Bypass Elimination modification, a Median Signal-Selector has been introduced into the control system. The Median Signal Selector preserves the functional isolation of interfacing control and protection systems that share common instrument .:hannels. The details of the signal selector impleni.' ation are contained in Section 1.3.1.

On the basis of the foregoing evaluation, it is concluded that the compliance a of V. C. Summer Nuclear Station to IEEE Standard 279-1971, applicable GDCs, and industry standards and regulatory guides has not been changed with the I&C modifications required for RTD bypass removal.

4.6 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 hot leg piping, the crossover leg bypass return nozzle, and the cold leg bypass manifold connection. All welding and NDE will be performed per ASME Code Section XI requirements. Each of these modifications is evaluated below.

{

\

i The original three scoops in the loop A, B and C hot legs, which feed the I bypass manifold, and the bypass manifold connection must be removed and all scoops modified to accept three fast response RTD thermowells. [

]"'C to provide the proper flow l path. A thermowell design will be used such that the thermowell will be f positioned to provide an average temperature reading. The therm'owell will be fabricated in accordance with Section III (Class 1) of the ASME Code. The installation of the thermowell into the scoop will be performed using Gas

{

Tungsten Arc Weld (GTAW) for the root pass and finished out with either GTAW

~

or Shielded Metal Arc Weld (SMAW). The welding will be examined by penetrant test (PT) per the ASME Code Section XI. Prior to welding, the surface of the scoop onto which welding will be performed will be examined as required by Section XI.

1789v:13/o42889 34 l

, _ . ~ . ~

The cold leg R1D bypass line must also be removed. The nozzle must then be modified to accept the fast response RTD thermowell. The installation of the thermowell into the nozzle will be performed using GTAW for the root pass and 1

, finished with either GTAW or SMAW. Weld inspection by PT will be performed as required by Section XI. The thermowells will extend approximately [ ja,e inches into the flow stream. This depth has been justified based on [

]C analysis. The root weld joining the thermowells to the modified nozzles will be deposited with GTAW and the remainder of the weld may be. deposited with GTAW or SMAW. Penetrant testing will be performed in accordance with the ASME Code Section XI. The thermowells will be fabricated in accordance with the ASME Section III (Class 1).

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 of ASME Section III (Class 1). The cap will be root welded to the nozzles by GTAW and fill welded by either GTAW or SMAW.

Non-destructive examinations (PT and radiographs) will be performed per ASME Section XI. Machining of the bypass return nozzle, 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.

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 [ Ja,c inches and )

the cold leg RTD connections are [ Ja,c inches, a system hydrostatic test is i required after the bypass elimination modification is complete. Paragraph IWB-5222 of Section XI defines this test pressure to be 1.02 times the normal operating pressure at a temperature of 500*F or great'er.

• l

, 4

. In summary, 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. Further, the pressure retaining capability and fracture prevention characteristics of the piping is not compromised by these modifications.

1789v:1o/042889 35

c- . . . . . . .

.i

~ .

F,.

l = 4.7 TECHilICAL SPECIFICATION EVALUATION 1

As a result of the calculations summarized in Section 3.0, several protection

- functions' Technical Specifications must be modified. The affected functions and their associated Trip Setpoint' information, are noted on Table 3.1-10.

i e

e 1789v:10/042889 36

e 5.0 CONTROL SYSTEM EVALUATION 1:

~

! A prime input to the various NSSS control systems is the RCS average temperature,T(avg). This is calculated electronically as the average of the measured hot and cold leg temperatures in each loop.

The effect of the new RTD temperature measurement system is to potentially change the time response of the T(avg) channels in the various 1 cops. This in-turn could impact the response of (

Ja c As previously noted, the new RTD system (RTD + thermowell) will have a time response slightly longer than that of the current system (RTD

+ bypass line). The additional delay resulting from the Median Signal Selector (MSS) is small in comparison with the RTD time response [ i

]a,c Therefore, there will be no significant impact on the T(avg) channel response and no need, as a result of implementing the new system, to revise any of the control system setpoints. However, V. C. Summer always has the option of making setpoint adjustments. If desired, system performance can he verified by performing a

. series of plant tests (e.g., step load changes, load rejections, etc.)

following installation of the new RTD system. Control system setpoints can then be adjusted based on the results of the tests. It should be recognized that control systems do not perform any protective function in the FSAR

]

accident analysis. With respect to accident analyses, control systems are )

ass w.ed operative only in cases in which their action Eggravates the l consequences of an event, and/or as required to establish initial plant I i

ec.1ditions for an analysis. The modeling of control systems for accident analyses is based on nominal system parameters as presented in the Precautions, Limitations, and Setpoint document.  !

. . I

.n i

1789v:1o/042889 37

-_--______-__ Q

6.0 CONCLUSION

S The method of utilizing fast-response RTDs installed in the reactor coolant loop piping as a means for RCS temperature indication has undergone extensive

~

analyses, evaluation and testing as described in this report. The incorporation of this system into the V. C. Summer Nuclear Station design meets all safety, licensing and control requirements necessary for safe operation of this unit. 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 replace the present hot and cold leg temperature measurement system and anhances ALARA efforts as well as improve plant reliability.

l s

i l

1789v:1D/042889 38

[

I-L.

k

7.0 REFERENCES

l'~

2. Virgil C. Summer Final Safety Analysis Report, January 20, 1989.

2. WCAP-10961-P, Steamline Break Mass / Energy Releases for Equipment Qualification Outside Containment, October 1985.

l

3. VANTAGE 5 Reload Transition Safety Report for the Virgil C. Summer Nuclear Station, September 1988.

- Q, e

9 J

)

1789v;1D/042889 39 I

__ _ _ _ _ _ _ _- -)

i

! i

]

1

, . {

1 I

l

{

I APPENDIX A DEFINITION OF AN OPERABLE CHANNEL AND HOT LEG RTD FAILURE COMPENSATION FROCEDURE 4

e 1789v.1D/042BB9 40

(

'd RTD BYPASS ELIMINATION FOR 4

V. C. SUMMER DEFINITION OF AN OPEP.ABLE CHANNEL AND HOT LEG RTD FAILURE COMPENSATION PROCEDURE This document contains information proprietary to Westinghouse Electric Corporation; it is submitted in confidence and is to be used solely for the purpose for which it is furnished and returned upon request.

This document and such information is not to be reproduced, transmitted, disclosed or used otherwise i in whole or in part without the written authorization l

~

of Westinghouse Electric Corporation. ,

i i

Westinghouse' Electric Corporation Pittsburgh, PA  !

I 1789v:1o/o42BB9 41 i I

I

.I

DEFINITION OF AN OPERABLE CHANNEL The RTD. Bypass Elimination modification uses the average of 3 RTDs in each hot leg to provide a representative temperature measurement. In the event one or more of the RTDs fails, steps must be taken to compensate for the loss of that

'RTD's input to the averaging function. V.C. Summer will have dual element RTDs installed in each hot leg thermowell location. The second element may be used when the first element fails and the three RTD average maintained. In the event of the second element failing in the same RTD, then this procedure could be envoked.

Single RTD Failure Hot Leg: All three hot leg RTDs must be operable during the period following refueling from cold to hot zero power and from hot zero power to full power.

During the heat up period the plant operators will be [.

cross-calibration procedure. Once full power is reached bias data should be taken as outlined in the following procedure for Operation With A Hot Leg RTD 8

Ja.c Typically this data is recorded at initial 100% power i

and, thereafter, during the normal protection channel surveillance interval.

Once [. Ja,c any hot leg can then l tolerate failure of both elements of a single dual element RTD and still remain operable. If the situation arises where such a failure occurs a bias  !

value must be applied to the average of the remaining two valid RTDs. ['

)

. Ja,e 1789v;1o/050389 t,2

s l

l The plant may operate with a failed hot leg RTD at any power level during that l

same fuel cycle. It is permissible to shutdown and startup during the cycle l

without requiring that the failed RTD be replaced. [

s ya,c The Median Signal Selector will eliminate any control system concerns, the Tavg and AT signal associated with the loop containing the failed hot leg RTD will most likely not be the Median Signal chosen as the input to tho control systems. If another hot leg RTD fails in a different loop the utility should operate using manual control. Manual control is recoarnended so that the operator can control the plant based on the best measurement available.

If automatic operation is continued the control system may choose the biased channel due to the positive (or zero) bias application. This means the control system will perceive a higher Tavg than actually exists at reduced power and the plant will operate at reduced temperatures. While this is not necessarily undesirab1'e it does reduce the total plant megawatt output. The use of automatic control can be considered based on utility power requirements.

Cold Leg: If the active cold leg RTD fails, then that RTD should be disconnected from the 7300 cabinets. The installed spare RTD should then be connected in the failed RTD's place.

Double RTD Failure: Inoperable Channel not Leg or Cold Leg: If two or more of the three hot leg RTDs or both cold leg RTD elements fail in the same protection channel then that channel is considered inoperable and should be placed in trip. Operation with only one valid het leg RTD is not presently analyzed as part of the licensing basis.

s 1789v:1o/042889 43

h PROCEDURE FOR OPERATION WITH A HOT LEG RTD OUT OF SERVICE l ..

The hot' leg temperature measurement is obtained by averaging the measurements from the. three thermowell RTDs installed on the hot leg of each loop. [

~

.ja,c -

l-l: '

In.the event that one of the three RTDs fails, the failed RTD will be disconnected and the hot leg temperature measurement will be obtained by averaging the remaining two RTD measurements. ['

I

.)a,c

-The bias adjustment corrects for [

~

~

Ja,c To assure that the measured hot leg temperature is' maintained at or above the true hot leg temperature, and thereby avoid a reduction in safety margin at reduced power, [<

q Ja,c 1789v:10/042889 44

An'.RTD failure will most likely result in.an offseale high or low indication and will be detected through the normal means in use today (i.e., TAVG and

. - AT. deviation alarms). - Although unlikely, the RTD (or its electronics channel) can fail gradually,. causing a gradual change in the loop temperature i .. measurements. [

Ja,c The detailed procedure for correcting for a failed hot leg RTD is presented below:-

a,e o

1 1789v:1o/042889 45 u__u __ __ . _ __ i

3i ,, 3 r 1-p

[ 't.

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f-

. '1 l

6

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1789v:1D/042889 46 is

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1 1789v:1D/042889 47 l

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t - ...:3 i i

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APPENDIX

~

CALCULATION OF HOT LEG TEMPERATURE BIAS

• C'

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- _ 8,C i

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4.

1789v:1D/042889 48

- - - . . - _ - - - - - - _ - - - _ _ _ _ - - _