ML20066L050

Nonproprietary Rev 2 to WCAP 12614, RTD Bypass Elimination Licensing Rept for Jm Farley Nuclear Plant Units 1 & 2

ML20066L050
Person / Time
Site:	Farley
Issue date:	01/31/1991
From:	Morrison R WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML19310E557	List: ML20066L017 ML20066L034 ML20066L050
References
WCAP-12614, WCAP-12614-R02, WCAP-12614-R2, NUDOCS 9102060230
Download: ML20066L050 (76)

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HESTINGHOUSE CLASS 3 HCAP 12614 Rev. 2 1

RTD BYPASS ELIMINATION LICENSING REP 0Ff FOR J. M. FARLfY NUCLEAR PLANT UNITS 1 and 2 repared by R. J. Morrison o

January 1991 Westir.ghouse Electric Corporation Pittsburgh, PA o 1990, Westinghouse Eiectric Corporation, all Rights Reserved 08020:10/012391

WESTINGHOUSE PROPRIETARY CLASS'3 e

This ner. proprietary report bears a Westinghouse copyright notice.

The NRC is permitted to make the number of copies of this report necessary for its internal use and such additional copies which are necessary in order to have one copy'available for public viewing in the appropriate docket files in the.public document room in Washington, D.C. and-in local public document rooms as may be required by NRC regulations if the number of copies submitted

]

is' insufficient for this purpose.

The NRC is not authorized to make copies for the personal use of members of the public who make use of the NRC public document rooms.

Copies of this report or

' portions thereof made by the the NRC must include the copyright notice.

i e

mkh$

'I

-$ 8 ]' $'U,

'?

FOREHARD s

Extensive studies were performed for farley Units 1 and 2 for the effects of f..

increased SG Tube Plugging and Reduced Thermal Design Flow (HCAP-12694 for Unit 1, HCAP-12659 for Unit 2).

The purpose of this report is to show the effects of RTD bypass elimination on the Farley Units while also considering the effects of the increased SG Tube Plugging and Reduced Thermal Design Flow.

08020:10/082290

4 ACKNOWLEDGEMENT The authors wish to recognize contribution by the following individuals:

Mike Emery Steve Zawalick Halt Tauche Fred Baskerville Marvin Hengerd Hally Moomau Dick Haessler Rick Tuley Phil Rosenthal Glen Lang Jim Mermigos Mimi Weaver Pete Morris Ron Carlson i

i I

08020:1D/082290

4 lABLE OF CONTENTS StLt19D

.P1g9 List of Tables lii

-List of Figures iv 1.0 Introduction 1.1 Historical Background 1

1.2 Mechanical Modifications 2

1.3 Electrical Modifications 4

2.0 Testing 2.1 Response Time Test 13 2.2 Streaming Test 13 3.0 Uncertainty Considerations 3.1 Calorimetric Flow Measurement Uncertainty 16 3.2 Hot Leg Temperature Streaming Uncertainty 16 3.3 Control and Protection Function Uncertainties 19 4.0 Safety Evaluation 4.1 Response Time 31 4.2 RTD Uncertainty 31 4.3 Non-LOCA Evaluation 32 4.4-LOCA Evaluation 41 4.5 Instrumentation and Control Safety Evaluation 41 4.6 Hechanical Safety Evaluation 45 4.7 Technical Specification Evaluation 46 08020:10/082290 i

i

i IABLE OF CONTENTS (Cont)

SEliQD East 5.0 Control System Evaluation 47 6.0- Conclusions 48 7.0 References 49 Appendix A - Definition of An Operable Channel And 50 Hot Leg RTD Failure Compensation Procedure 1

Appendix B - Definition of Acronyms Used in Uncertainty 59 Calculations 0802D:1D/082290 ii

LIST OF TABLES IAble Title Eage 2.1-1 Response Time Parameters for RCS Temperature Measurement 15 3.1 Rod Control System Accuracy 20 3.1-2 Flow Calorimetric Instrumentation Uncertainties 21 3.1-3 Flow Calorimetric Sensitivities 22 3.1-4 Calorimetric RCS Flow Measurement Uncertainties 23 3.1-5 Overtemperature Delta-T Trip 25 3.1-6 Overpower Delta-T Trip 26 3.1-7 T

low-tow Trip 27 ayg 3.1-8 Cold Leg Elbow Tap Flow Uncertainty 28 3.1-9 Low Flow Reactor Trip 29 3.1-10 Technical Specification Modification 30 I

08020:1D/082290 lii

LL11_OF FIGURES Ficure' 11tle P_ase 1.2-1 Hot Leg RTO Scoop Modification for Fast-Response 6

RTD Installation 1.2-2 Hot Leg RTO Boss Installation for Fast-Response 7

RTO Installation 1.2-3 Cold Leg Pipe Nozzle Modification Fast-Response 8

I RTD Installation 1.2-4 Crossover Leg Cap Installation 9

1.3-1 RTD Averaging Block Diagram, Typical for Each of 3 10 Channels 1.3-2

-Median Signal Selector 11 1.3-3 Control System Schematic 12 08020:10/082290 iv

?

1.0. INTRODUCTION l

Hestinghouse. Electric Corporation has been contracted by Alabama Power Company

.(APCO) to remove the existing Retistance Temperature Detector (RTD) 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 the J. H. Farley Nuclear Plant Units 1 & 2 utilizing the new thermowell mounted RTDs.

l.-l HISTORICAL BACKGROUND Prior to 1968, PHR designs had been based on the assumption that'the hot leg temperature was uniform across the. pipe. Therefore, placement of the temperature instruments was not considered to be a factor affecting the accuracy of the' measurement.

The hot leg temperature was measured with direct immersion RTDs extending a short distance into the pipe at one location.

By the late'1960s as a result of accumulated operating experience at several plants,-the following-problems associated with direct immersion RTDs were

. identified:

Temperature streaming conditions-(the incomplete mixing of'the coolant o

leaving regions of the reactor. core at different temperatures which produces significant temperature gradients within the pipe),

o The reactor coolant 11 oops required cooling and-draining before the

RTDs could be replaced, s

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 ownLobstacles 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.

(

0802D:10/082290 1

't c.

o - ' Increased radiation exposure due to maintenance on the bypas; line and to crud traps which increase radiation exposure throughout the loop l

compartments.

ij The proposed' temperature measurement modification has been deve. loped in-

- response..to_ both sets of. problems encountered in:the past._ Specifically:

._7

- o- -Removal of the' bypass lines eliminates the compo'nents which have'been a major source of plant outages as well as Occupational Radiation Exposure (ORE)..

i o

Three thermoweil mounted' hot leg RTDs provide an average _ measurement (equivalent _to.the temperature measured by the bypass system) to j

account for temperature streaming.

(

oz Use of thermowells permits RTD replacement without draining the reactorocoolant loops.

FollowingIis a' detailed description:of the offort required..to perform this modi fication. -

j 1.2 MECHANICAL HODIFIC TIONS-l a

i The,individualiloop temperature signals' required for_. input to the Reactor

~

- Control and: Protection System wi_11 be obtained-using',RTDs installed in each reactor' coolant loop, j

i

-1.2.1 Hot-Lea y

a) The hot leg-temperature measurement on-each' loop will be accomplished with i

three fast response, narrow range. dual element RTDs mounted in'

-thermowells. One element of the RTD will be considered active and the 3

.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 08020:1D/082290 2

i

-.,,4

,7 located within the three existing RTD bypass manifold scoops wherever possible.

A hole will be made through the end of each scoop so that water will-. flow in through the existing holes in the leading edge of the scoop, past,the RTD and_out through the new hole (Figure 1.2-1).

If plant

-interferences preclude-the placement of a thermowell in a scoop,-then the scoop will be capped and a new penetration made to accommodate the thermowell (Figure 1.2-2).

These three RTDs will measure the hot leg temperature which is used to calculate the reactor coolant loop differential temperature (AT) and average temperature (Tavg)*

b) This modification will not affect the single wide range RTD currently

-installed near-the entrance of each steam generator.

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

.1 ^. 2. 2 Cold leg a) One fast response, narrow range, dual-element RTD will be located.in each cold leg at the discharge of the reactor coolant pump (as replacements for j

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

The existing cold leg RTD bypass. penetration nozzle-will be avg.

modified (Figure 1.2-3) 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 acci at conditions.

08020:10/082290 3

c

~ 1.2.3 Crossover Lea LThe'RTD bypass manifold return line will _ be capped At the nozzle on the crossover-leg as shown on Figure 1.2-4..

1.3? ELECTRICAL H0DIFICATIONS.

1.3.1-Control & 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 Thot signal to calculate reactor coolant signal will'then be used with the Tcold loop AT and T which are used in the reactor control and protection avg system. 'This.will be accomplished by. additions to the: existing process protection system equipment.

It is planned to wire the T and T hot cold spare RTO elements to the control room and terminate them at the 7300 rack

input terminals.- This arrangement will allow on-line accessibility to the spare elements for RTO cross calibrations and to facilitate connection of'the spare RTD element in-the event of an RTO element failure.

The presentcRCS loop temperature measurement system uses dedicated direct immersion RTDs for the' control 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.will be: Used for both control and protection.

In order to continue to satisfy the 1

requirements of'IEEE-Standard 279-1971,'the T and.AT signals generated avg in the protection system will.be electrically isolated and transmitted to the control-system into Hedian Signal Selectors forfTavg _and AT which -will =

select the signal which-islin between the highest and lowest values of the

three loop: inputs'- This will preclude ~an unwarranted control-. system response-Lthat could be caused by a single signal failure.

08020:10/082290' 4

~

L1.3.2 OualificatiaD The 7300_ Process System Electronics modifications will be qualified-_to the same level-as-the existing 7300 electronics.

RTD qualification will be-verified to support APCO's compliance to 10CFR50.49.

The Westinghouse qualification program entailed a review of the HEED Instrument: Company's qualification documentation for testing performed on these RTDs.

It was concluded that the equipment's qualification was-in compliance with IEEE Standards 344-1975 and 323-1974 with one exception.-

Specifically, 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.v_ibration aging,and a simulated seismic event.

These tests confirmed that'the HEED-RTDs do comply with the above IEEE standards.

' l.3.'3 _RTD:00erabilitv= Indication Existing control-board AT and Tavg _ indicators and alarms will provide the means of: identifying-RTD failures, although the now redundant indication for the T and-AT control signals will be removed.

The spare cold leg RTD avg

~

element provides sufficient spare capacity.to accommodate a single cold leg RTD failure per loop.

Failure of a hot leg RTO can be handled in two_ ways.

The first method-disconnects the failed: element and-utilizes the second element of the same RTD.

In the second, manual action' initiated by the operator. defeats the failed signal and rescales the electronics to average the remaining. signals.

H j-i r

08020:10/082290 5

L

0-

a. C.

Figure 1.2-1 Hot Leg RTD Scoop Modification for Fast Response RTO Installation 0802Di10/082290 6

7 1

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- Figure l.2-2_ Cold Leg Pipe Nozzle Modification for Fat Response RTD Installation

/

08020:10 082290 7

O. C.

Figure 1.2-3 Cold Leg Pipe Nozzle Hodification Fast Response RTD Installation 08020:10/082290 8

.. n.

0,g C Figure 1.2-4 Crossover Leg Cap Installation 08020:10/082290 9

4 A,c.

9 Figure-l.3 RTD Averaging Block Diagram lypical for Each of 3 Protection Channels 08020:10/082290

-10

4 a.ic.

Figure 1.3-2 Median Signal Selector Block Diagram 08020:10/082290 11

cLut 1

l Figure 1.3-3 RTD Bypass Elimination Control System Schematic i

0802D:10/082290 12

_. _ _. _ _ _ _.. _.. _. _. ~

2.0 IESTING 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 Farley Units 1 & 2 application will be verified by testing at the RTD. manufacturer and by in-situ testing.

Data from-l thermowell/RTD performance tests'at operating plants provide additional

. support for the -system by confirmation of RTD/thermowell response times and by l

confirmation of the magnitude of temperature streaming.

c 2.1 RESPONSE TIME TEST

'The RTD manufacturer, HEED Instruments Inc., will perform time response Ltesting of each RTD and.thermowell prior to installation at the Farley Units 1

& 2.

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 j

transient analyses discussed in Section 4.0.

1 In addition, response time testing of the HEED RTDs will be performed in-situ.

This testing will demonstrate that the HEED RTDs can satisfy.the response. time requirement -when installed in the plant.

2.2 STREAHING TEST iPast testing at Westinghouse PHRs has established that temperature

-stratification exists in the _ hot leg pipe with a temperature gradient from j

minimum

• to maximum-of ['

]b,c.e A test program was implemented at

~

4 an. operating. plant to confirm the temperature streaming magnitude and stability with' measurements of the.RTD bypass branch 11ne temperatures on two

' adjacent hotlleg pipes.

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 strenming patterns and i

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 te:nperature error for use in the i

0802D:10/082290 13 m

~.

u safety analysis and calorimetric' flow calculations.

Section 3 will discuss the1 specifics of these uncertainty considerations.

The-test-data was reduced and characterized to answer the three objectives of i

'the test program.

First,_it is conservative to state'that the streaming-pattern'(

]b,c.e.

Steady state data taken at

=

100% power for aLperiod of four months indicated that the. streaming pattern bl,c.e In other words, the temperature Li-bl,c.e This gradient-[

bl,c.e is. inferred by [

observed between branch lines.

Since the b3,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 L (- -

)b.c.e into the "two RTD" average to obtain the "three RTD" expected reading.

A generic procedure has-been provided _to-APC0 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 Hestinghouse. plants.

The itemperature gradients defined by the.recent 71 ant operating _ data are well within.the upper bound temperature gradients that characterize the previous

-test data. ' Differences observed in the operating data compared with the

~

previous!testidata indicate that the temperature gradients _are smaller, so.-the

measurement. uncertainties are conservative. -The measurements at the operating _

plants.'obtained from thermowellTRTDs installed inside the.' bypass scoops, were expected to be, and were found1to'be, consistent with.the measurements

obtained previously_from the_ bypass-loop RTDs.

I 08020:10/082290 14

TABLE 2.1-1 RESPONSE TIME PARAMETERS FOR RCS TEMPERATURE MEASUREMENT RTD Fast Response Byoass System Thermowell RTD System a,c a,c y

RTD Bypass Piping and Thermal Lag (sec) 3 RTD Response Time (sec)

Electronics Delay (sec)

Total Response Time (sec) 6.0 sec 6.0 sec 0802D:1D/082290 15

3.0 UNCERTAINTY CONSIDERATIONS This method of hot leg temperature measurement has been analyzed to determine the magnitude of the two uncertainties included in the Safety Analysis:

Calorimetric Flow Heasurement Uncertainty and Hot Leg Temperature Streaming Uncertainty.

Tables 3.1-1 through 3.1-10 were generated specifically for APC0 and reflect plant specific measurement uncertainties and operating conditions.

3.1 CALORIMETRIC FLOW HEASUREMENT UNCERTAINTY

]

Reactor coolant flow is verified with a calorimetric measurement performed after the return to power operation following a refueling shutdown.

The two most important instrument parameters for the calorimetric measurement of RCS E

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 RTDs (resulting from the elimination of the RTD hot bypass lines) and the recommendations of the Westinghouse RTD cross-calibration procedure (resulting in low RTD calibration uncertainties at the beginning of a fuel cycle), the Farley Units 1 & 2 RCS Flow Calorimetric l,c including use of cold leg a

uncertainty is estimated to be (

elbow tape (see Tables 3.1-2, 3, 4 and 8).

This estimate is based on the standard Westinghouse methodology previously approved on earlier submittals of other plants associated with RTO Bypass Elimination or the use of the Westinghouse Improved Thermal Design Procedure.

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 08020:10/082290 16

4 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

[

3,c.e The calculations for numerous temperature distributions have shown that, even with margins applied to the observed temperature gradients, the three-point temperature measurement (scoops or thermowell RTDs) is very effective in determining the average hot leg temperature.

The most recent calculations for the thermowell RTD system have established an overall streaming uncertainty of [

]b,c.e for a hot leg measurement.

Of this total,

(

)b,c.e This overall temperature streaming uncertainty determined for plants with similar or symmetrical temperature distributions is conservative when applied to 3 loop plants such as Farley Units 1 IL 2 since the 3 loop temperature distributi<ons are not symmetrical, This non-symmetric distribution results 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,

(

J,c Although the new method measures temperature at one point at a

the RTD/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 thermow' ell measurement may have a small error relative to the scoop measurement if the temperature gradient over the 5-inch scoop span is nonlinear.

Assuming that the maximum inferred temperature gradient of C

]b,c.e exists from the center to the end of the scoop, the difference between the thermowell and scoop measurement is limited to [

bl,c,e Since three RTD measurements are averaged, and the nonlinearities at each scoop are random, the effect of this error on the hot 08020:1D/082290 17

)

leg temperature measurement is. limited to (

]b.c.ei 'On the-other

' hand, imbalanced scoop flows can introduce temperature measurement uncertainties of up to

(-

J,c,.In all cases, the scoop a

flow imbalance uncertainty will equal or exceed the [

]bec,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 loop plants and from thermowell l1TO installations at 4-loop plants.

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

[

b3,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. twoliot' leg RTDs.inLservice.

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-Lcomparisons show that the magnitud'e of this bias varied -less than [

)b.c.e~over the test: period.. In addition, the uncertainty calculaticns=

assumed that two T RTD's were utilized to determina T APP 0"U'* A hot hot' 5provides at procedure.for utilizing the actual plant bias data.

N0te that this procedure only allowsLthe~usetof-positive'(or zero) bias values.

f 3i

-08020:1D/082290 18

y m

i m

s.

3.3. CONTROLLAND PROTECTION FUNCTION UNCERTAINTIES

}

X,,

N' Calculations were performed to determine or verify the instrument

? uncertainties:for the control and protection functions affected by the RTD Bypass; Elimination.

Methodology for these calculations has been accepted in Reference 6.

Table 3.1-1 (Rod Control System Accuracy) notes that an

[

t 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. Table.3si-5 yrovides the uncertainty breakdown for Overtemperature AT.. As noted.on this table, TA is greater than CSA, thus acceptable results are. calculated for this function.

Teois. 3.1-6 provides the breakdown for 0verpower AT, with the same conclusions as for Overtemperature

AT.

Table 3.1-7 notes the uncertainty breakdown for Tavg: Low-1.ow.

Again acceptable:results are calculated. Table 3.1-9 is concerned with the RCS Low c

. 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 q

3.1H.- Farley Nuclear Plant Units 1 & '2 Technical Specifications. As noted, j

relatively minor changes' a're necessary to reflect the modified calculathn D

results',- primarily the Allowable Values. Appendix B'contains a listing of

' acronyms for:thoso:usedlin the uncertainty calculations.

n

-i i:

g 1

l; u

0802D:'1D/082290 19 1

l

4 TABLE 3.1-1

/00 CONTROL SYSTEh SCGDAC'!

Tug RlRB PMC"

%C p

1 SCA -

SMTE.

STE -

SD L

BIAS-I RCA =

l RHTE-RMTE-c t

f RTE -

RD

(

CA

=

DIAS-(

NO. RTDs USED TH =.2 TC = i

~

ELECTR0NIC3 CSh

\\

ELECTRONICS SIGMA -

D CONTROL 1ER SIGMA

(

CONTROLLER BIAS CONTROLLER CSA i

F A-(

08MD:30/082290 20 i

7 %

m--

, e,

.m...

~_

TABLE 3.1-2

)

FLOH CALORIMETRIC INSTRUMENTATION UNCERTAINTIES (4, SPAN)

FH TEMP FH FRES FH DP STM PRESS TH TC PRZ FRESS

• a.C SCA -

SMTE.

SPE -

STE -

SD R/E -

RDOT.

BIAS.

CSA -

NO. OF INST USED 3

1 2

DEG F PSIA

% DP PSIA DEG F DEG F PSIA INST SPAN 500 2000 120 1200 120 120 800

+a.C INST UNC.

(RANDOH) =

INST UNC.

(BIAS)

NOMINAL These calculations were performed assuming that ct,C.

08020:10/082290 21

~-

= _-

i TABli 3.1-3 FLOW CALORIMElRIC SENSITIVITIES FEEDWATER FLOW Fa

._ +a.c i

TEMPERATURE MATERIAL DENSITY TEMPERATURE PRESSURE DELTA P FEEDWATER ENTHALPY TEMPERATURE PRESSURE h5

- 1199.9 BTU /LBM 416.4 BTU /LBM hF 783.5 BTU /LBM Dh(SG)

STEAM ENTHALPY

+a,C PRESSURE MOISTURE HOT LEG ENTHALPY TEMPERATURE PRESSURE

(,

hH

- 629.7 BTV/LBM hC

- 538.6 B1U/LBM 91.1 BTV/LBM Dh(VESS)

Cp(TI.)

- 1.495 BTU /LBM-DEGF COLD LEG ENTHALPY

+a,C TEMPERATURE PRESSURE Cp(TC)

- 1.227 BTV/LBM-DEGF COLD LEG SPECIFIC VOLUME

+a,c TEMPERATURE PRESSURE 08020:1D/082290 22

4 8

TABLE 3.1-4 4

CALORIMETRIC RCS FLOW MEASUREMENT UNCERTAINTIES COMPONENT INSTRUMENT ERROR FLOW UNCERTAINTY

(% FLOW) a

- +4.C FEEDWATER FLOW VENTURI THERMAL EXPANSION COEFFICIENT TEMPERATURE MATERIAL DENSITY TEMPERATURE PRESSURE DELTA P FEEDi4ATER ENTHALPY TEMPERATURE PRESSURE STEAM ENTHALPY PRESSURE MOISTURE NET PUMP HEAT ADDITION HOT LEG ENTHALPY TEMPERATURE STREAMING, RANDOH STREAMING, SYSTEMATIC PRESSURE COLD LEG ENTHALPY TEMPERATURE.

PRESSURE COLD LEG SPECIFIC VOLUME TEMPERATURE PRESSURE l

t

(

0802D:10/082290 23 L

=. - -.

TABLE 3.1-4 (continued)

CALORIMETRIC RCS FLOW HEASUREMENT UNCERTAINTIES BIAS VALUES

__4a,c 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 der ~.NDENT PARAMETERS SINGLE LOOP UNCERTAINTY (HITHOUT BIAS VALUES)

N LOOP UNCERTAINTY (HITHOUT BIAS VALUES)

N LOOP UNCERTAINTY (HITH DIAS VALUES) 08020:10/082290 24

1 di i

TABLE 3.1-5 1

OVERTEMPERATURE DELTA-T TRIP DELTA-T Tavg PRESS DELTA-1 e

+B C PHA SCA ~ =

J SMTE -

i STE j

-SD-1 4

BIAS -

RCA RHTE RH'TE.

RCSA =

4

(

' RTE i

t 1

RD i

SA-o NO. OF RTD USED TH - 2 TC - 1 102.3 DEGF

. INSTRUMENT-SPAN-

-+a.c SAFETY ANALYSIS LIMIT (SAL)

I,

.. 2.60% DELTA-T SPAN-

.ALLOHABLE VALUE__

1.1800

.K3 0.000635' NOMINAL SETPOINTS K1 VESSEL DELTA-T-

- 68.2 DEGF DELTA-1 GAIN -

1.75

i L

+a.c

'I PRESSURE GAIN

}:

l

,,__ B,C

_. a.C

+

+a,C-

+

T Z_-

S CSA =

MAR L

TA

=

l

?

1 r

08020:10/012391:

25 i

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

+a c PHA

=

SCA

$0 BIAS -

RCA RHTE -

RHTE -

RCSA -

RTE RD NO. OF RTD USED TH - 2 TC - 1 INSTRUMENT SPAN 102.3 GE0"

+a.C SAFETY ANALYSIS LIMIT 2.93% uELTA-T SPAN ALLOHABLE VALUE 1.0800 NOMINAL SETPOINT VESSEL DELTA-T

- 68.2 DEGF

+B.C

-. + B. C

+ &, C T

S Z

CSA -

HAR TA 1

1 0802D:10/082290 26

TABLE 3,1-7 Tavg Low-Low TRIP

.N

+a.C

~

PHA SCA SD

=

BIAS -

RCA RMTE -

RCSA -

RTE RD NO. OF RTD USED TH 2

TC 1

100,0 DEGF INSTRUMENT SPAN

+a.C SAFETY ANALYSIS LIMIT

-(

)

540,2 DEGF ALL0HABLE VALUE 543.0 DEGF NOMINAL TRIP SETPOINT

+a.C

-+a.C

+a.C

~

T S

Z

=

CSA MAR TA 0802D:10/082290 27

._=

4 TABLE 3.1-8 COLD LEG ELBOW TAP FLOW UNCERTAINTY INSTRUMENT UNCERTAINTIES

+5,C

% DP SPAN

% FLOW

)

PHA PEA SCA SPE

=

l STE SD RCA RMTE -

RTE RD.

ID A/D RDOT -

FLOW CALORIM. BIAS FLOW CALORIMETRIC INSTRUMENT SPAN

+R.C SINGLE LOOP ELB0H TAP FLOW UNC -

% FLOW N LOOP ELB0H TAP FLOW UNC N LOOP RCS FLOH UNCERTAINTY

+a c (HITH BIAS VALUES) 08020:i0/082290 28

TABLE 3.1-9 LOH FLOH REACTOR TRIP INSTRUMENT UNCERTAINTIES l

4B,C

% DP SPAN

% FLOW SPAN PHA1 -

PMA2 -

PEA

=

SCA

=

SPE STE SD BIAST.

DIAS 1-BIAS 2 RCA RHTE -

RCSA -

RTE RD BIAS -

120.0 % FLOW FLOH SPAN

+B.C SAFETY ANALYSIS LIMIT

-[

]

I 88.5 % FLOW

'ALLOHABli VALUE NOMINAL TRIP SETPOINT

- 90.0 % FLOW

+B.C

+a.C

~

~

+B,C T

S Z

L CSA -

MAR TA 4

i l

08020:1D/082290 29 i

m.,

-_-._._.__.._,.m

TABLE 3.1-10 TECHNICAL SPECIFICATION HODIFICATIONS Overtemperature AT K

- 1.18 g

Z

- 4.76 S

= 1.47(AT) + 0.64(pressure)

Allowable Value 1 2.6% AT span Response Time 1 6 sec al penalty

- 1.75%

Overpower AT Z

- 1.10 S

- 1.47 Allowable Value 1 2.9% AT span Loss of Flow Z

- 1.71 S

0.60 Allowable Value 1 88.5% of Loop Design Flow DNB Parameters RCS Tavg - 581.5'F RCS Total Flow Rate 1 267,400 gpm

-Tavg Low-Low Allowable Valve - 540'F P-12 1 547'F (Increasing) 1 540'F (Decreasing)

• Includes approximately 1.5% TDF Reduction and 2.P% Increase for Uncertainty.

t 08020:10/012391 30 I

.. _ _ _ _. __ _. _. ~. _ _ _ _. _ _ _.. _ _ _ _ _ _. _.. -

4 a

4.0 SAFETY EVALUAII.0N The primary impact of the RTO Bypass Elimination on the FSAR Chapter 15 (Reference 1) safety analyses are the differences in response time characteristics and instrumentation uncertainties associated with the fast response thermowell RTO system.

The effects of these differences are discussed in the following sections.

4.1 RESPONSE TIME The response time parameters of the J. H. Farley Nuclear Plant Units 1 & 2 RTO 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 (

J.c (as' presented in Section 2.1 and as given in Table 2.1-1).

a The new thermowell mounted RTOs have a response time equal to or faster than the maximum allowed time for the old bypass piping transport, thermal lag and direct immersion RTO.

This response time is factored into the Overtemoerature 4T trip performance.

Therefore, those transients that rely on the above mentioned trip must be evaluated for the modified response characteristics.

Section 4.3 includes a discussion of the evaluations performed for these events.

^

4.2 RTD UNCERTAINTY

-The proposed fast response thermowell RTD system will-make use of RTDs, manufactured by Heed Instruments Inc., with a total uncertainty of [

a 3,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 nade for the average-reactor coolant system (RCS) temperature, pressure and poter.

G These ellowances are_ explicitly. applied to the initial conditions for the transients.

08020:lD/100990 31 y

,...-,w,.-~,...,,

m,,,,_,

,,-,_.,,,,,,.4

-,. ~.. _.

,,,.-,_m..

e The following protection and control system parameters were evaluated (with respect to accident analysis assumptions) for the change from one hot leg RTD to three hot leg RTDs:

the Overtemperature AT (OTAT), Overpower AT (OPAT), and Low RCS Flow reactor trip functions; RCS loop Tavg measurements used for input to the rod control system, steam dump system, feedwater isolation, steam line isolation, safety injection; and the calculated value of the RCS flow uncertainty.

System uncertainty calculations were performed for these parameters to determine the impact of the change in the nember of hot leg RTDs.

The results of these calculations, noted in 3.3, indicate sufficient margin exists to account for known instrument uncertainties for all of the above except the rod control system accuracies and the low RCS flow reactor trip.

Therefore, these items are addressed in Section 4.3.1 and 4.3.2.

4.3 NON-LOCA EVALUATION As discussed in HCAP-12659 (Reference 7) and NCAP-12694 (Reference 8), the evaluations presented in this section have conservatively considered an operating configuration of 15% average steam generator tube pluggir.g. With a maximum plugging level in one steam generator of 20% and with an analyzed minimum average thermal design flow of 87200 gpm/ loop.

The evaluation results are applicable to this level of tube plugging or any lower level.

The RTD response time discussed in Section 2.1 and the instrumentation uncertainties calculated in Section 3.3 have been considered for the J. M.

Farley Nuclear Plant non-LOCA safety analysis design basis.

Only those i

transients ;hich assume OToi protection are potentially affected by changes in the RTD respon e time.

As noted in Section 4.1, the new thermowell mounted RTDs have a restoise time equal to or better than the old bypass piping

[

transport, thermal lag and direct immersion RTD. On the basis of the information documented in Table 2.1-1, it is concluded that the safety analysis assumption for total OTAT channel response time of 6.0 seconds remains valid.

Evaluation of the effects of the RTD Bypass Elimination on the uncertainties associated with these setpoints supports the continuing validity of the non-LOCA safety analyses (References 1, 7 and 8).

l 08020:1D/100990 32

Instrumentation uncertainties can affect the non-LOCA transient initial condition assumptions and those transients which assume protection from low primary coolant flow reactor trip. These effects are discussed in the following sections.

4.3.1 EFFECTS OF ROD CONTROL SYSTEM ERRORS As noted in Section 3.0, the RTD Bypass Elimination affects the Rod Control System accuracies.

These accuracies affect the initial RCS Tavg assumed in the non-LOCA safety analyses.

The current analysis assumptions are based on a 1 'F allowance as discussed in the J. M. Farley Nuclear Plant Units 1 and 2 4

FSAR, Section 15.1.2.

For the RTD Bypass Elimination, the allowance is increased to 1 3'F.

4 The initial Tavg assumed for the non-LOCA transients initiated from full or partial power includes an error allowance for the rod control system. The allowance is not assumed for transients initiated from zero power conditions.

Therefore, the following zero power transients are not affected by the increase in the Tavg allowance from 1 'F to 1 3*F:

4 4

RCCA Bank Hithdrawal from a Subtritical Condition ((15.2.1) and the new analysis presented in References 7 and 8).

Excessive Heat Removal due to Feedwater System Malfunctions ('ero power case) (15.2.10);

tecidental Depressurization of the Main Steam System (15.2.13) 1 RCCA Ejection (zero power cases) (15.4.6);

Rupture of a Main Steam Line (15.4.2.1)

The conclusions of these analyses as well as the conclusions in Reference 7 and Reference 8 remain valid.

0802D:10/082290 33

For transients analyzed to confirm that the DNB design basis is met, generic plant DNB margin has been allocated to offset the DNB penalty of the additional 0.3*F in the initial Tavg.

Therefore, the conclusions of the following DNB transients remain valid:

Uncontrolled RCCA Bank Hithdrawal at Power (FSAR Section 15.2.2)

RCCA Misalignment (15.2.3)

Partia! Loss of forced Reactor Coolant Flow (15.2.5, and the new analysis presented in References 7 and 8)

Startup of an Inactive Reactor Coolant Loop (15.2.6)

Loss of External Electrical Load (15.2.7)

Excessive Heat Removal Due to Feedwater System Halfunctions (15.2.10)

Excessive Load Increase Incident (15.2.11)

Accidental Depressurization of the RCS (15.2.12)

Inadvertent Operation of ECCS During Power Operation (15.2.14)

Complete Loss of Forced Reactor Coolant Flow (15.3.4)

A number of non-LOCA transients are analyzed to demonstrate acceptability for criteria other than DNB. A discussion of the effects of increasing the Tavg uncertainty by 0.3'F for these transients follows.

Uncontrolled Boron Dilution (15.2.4)

The boron dilution event is an uncontrolled addition of unborated reactor makeup water into the RCS via the Chemical and Volume Control System.

The boron dilution event is analyzed to demonstrate that, prior to total loss of shutdown margin, there is sufficient operator action time available to 08020:10/082290 34

. - - - - = -.. - -. _ - - -. - - - _

recognize the event and terminate the dilution.

The increased temperature uncertainty does not change the critical parameters assumed in the analysis:

the maximum dilution rate, RC3 boron concentrations, or the dilution volume for any of the operational modes.

Therefore, the current boron dilution analysis and the evaluation of this event in References 7 and 8 remain valid.

Loss of External Electrical Load (15.2.7)

The loss of external electrical load is a complete loss of steam load from full power without a direct reactor trip.

Four cases are analyzed which are based on two different primary side pressure control strategies (automatic and no mitigating control) and two sets of core physics characteristics (minimum and maximum reactivity feedback).

The key acceptance criteria for this transient besides ONB are primary and secondary pressures remaining below 110%

of design.

The RCS design pressure is 2485 psig (2500 psia) and the steam generator design pressure is 1085 psig (1100 psia).

As shown in FSAR Figures 15.2-19 through 15.2-26, the peak pressurizer pressure for all of the cases all remains below 110% of RCS Design Pressure.

The peak calculated RCS and secondary side pressures are not sensitive to the initial temperature assumed. Considering the temperature increase is only 0.3'F and given the margin in the current analysis, it is concluded that the increase in temperature uncertainty will not change the conclusions of the FSAR or the conclusions of References 7 and 8.

Loss of Normal Feedwater (15.2.8)

The loss of normal feedwater is the simultaneous loss of feedwater flow to all three steam generators.

The FSAR analysis assumes that the reactor coolant pumps coastdown due to an assumed loss of offsite power. As stated in the FSAR, this event is analyzed to demonstrate that the pressurizer does not become water solid during the transient.

The analysis assumes a power level corresponding to the 102% of the engineered safeguards power rating (a conservative assumption because Farley Units 1 and 2 are not licensed to operate at engineered safeguard power).

The initial Tavg is assumed to be the engineered safeguards Tavg minus 4'F.

The temperature uncertainty is subtracted from the nominal Tavg for this analysis because a lower temperature 0802D:lD/082290 35

J results in more initial RCS mass The larger RCS mass is more conservative when verifying that the pressurizer does not fill.

For a change of 0.3*F, however, this effect is small.

-FSAR Figure 15.2-27 shows that the peak pressurizer water level is less than 1200 cubic feet.

The pressurizer has an internal volume of 1400 cubic feet.

The resulting margin to pressurizer filling has been compared to the effects of a 0.3'F change in Tavg and increased steam generator tube plugging

-(Reference 7).

The analysis margin is sufficient to accommodate the effects of both changes and still maintain margin for filling the pressurizer.

It should be noted that the decay heat model used in the current analysis is based on the ANS-1971 decay-heat model.

Additional analysis margin would be gained if the ANS-1979 model was used because the total energy released into the RCS is lower.

Therefore, the conclusion that the pressurizer does not fill for~this event remains valid.

4 Loss of All AC Power to the Station Auxiliaries (15.2.9) u This event represents a complete loss of power to the plant auxiliaries (i.e.,

the reactor coolant pumps, feedwater pumps, condensate pumps, etc.).

The conclusions section in the FSAR states that the loss of forced flow and loss

-l

.of normal feedwater results show that the acceptance criteria.will be met for this. transient.

Both of these transients have been evaluated and found acteptable; therefore, the conclusions ofEthe FSAR and References 7 and 8 remain valid.

i Single RCCA Hithdrawal at Full Power (15.3.6) i This event represents the accidental withdrawal of a single RCCA from the inserted bank at full power operation. An evaluation for this transientLwas performed for the increased Tavg uncertainty and it was' demonstrated that the-conclusions' in the FSAR (i.e'., less than 5% of the fuel rods are below the DNB

-limit value) and References 7 and 8 remain valid, t

08020:10/082290-36

Major Rupture of a Main feedwater Pipe (15.4.2.2)

The rupture of a main feedwater pipe is a break in the feedwater pipe large enough to prevent the addition of feedwater to the steam generators.

There are two cases for feedline break presented in the FSAR.

The primary acceptance criterion in the FSAR is that the core remains covered. Case A was analyzed assuming the initial temperature was 6.5'F above the engineered safeguards Tavg.

Therefore, the current analysis bounds the increase in Tavg uncertainty.

Case B assumed that the initial temperature was 4'F above the engineered safeguards Tavg.

This analysis is performed at 102% engineered safeguards power with an ANS 1971 decay heat model'. The plant is not licensed to operate at engineered safeguard power and the results would be less limiting using an ANS 1979 decay heat model. Taking credit for the conservative power level h

-assumption and the 1979 decay heat model, it is concluded that the FSAR

[

conclusions remain-valid (i.e. a feedline break at the licensed plant power

)

' level will have acceptable results) for_the increase of 0.3'F in Tavg l

uncertainty.

In addition, this transient was analyzed for the steam generator tube plugging program. 'As discussed in Reference 7 and Reference 8, the analysis assumed a Tavg uncertainty of 6*F.

This provides a bounding analysis for a Tavg l

uncertainty of 4.3*F.

Therefore, the analysis described in References 7 and 8 i

_is.not affected by this-increase in Tavg uncertainty.

i

- Single Reactor Coolant Pump Locked Rotor (15.4.4)

This transient was reanalyzed to demonstrate acceptable results with a. lower l

analysis value for the low reactor coolant loop flow-setpoint. The increased uncertainty in Tavg was explicitly modeled in the analysis.

Refer to L

subsection 4.3.2.

i

-RCCA Ejection (15.4.6)

The RCCA ejection event is defined as the mechanical failure of a control rod L

drive-mechanism pressure housing resulting in-the ejection of a RCCA and-drive l

L 08020:1D/082290 37

O shaft.

Beginning and end of cycle conditions are analyzed at full and hot zero power levels.

As previously discussed, the hot zero power cases are not affected by the increase in Tavg uncertainty.

The hot full power cases are analyzed assuming DNB conditions immediately following the RCCA ejection.

This minimizes the heat transfer from the fuel to the coolant which maximizes the fuel rod temperature transient.

Therefore a small change in the coolant temperature does not have a significant effect on the results.

There is sufficient margin in the current results to conclude that, with the increased temperature uncertainty and the steam generator tube plugging effects (Reference 7 and 8), the acceptance criteria will continue to be met.

Therefore, the conclusions of the FSAR analyses and Reference 7 and 8 evaluations remain valid.

Steamline Break Mass and Energy Releases Outside Containment (15.4.2.1).

Steamline break mass and energy release data are calculated for several break sizes at different power levels for the purposes of equipment environmental qualification outside containment.

Reference 2 contains the results of the outside conthinment steamline break mass and energy releases.

Farley Units 1 and 2 were included in Reference 2 as part of Category 4.

For this analysis, the Tavg uncertainty assumed was 6.5'F.

Therefore, the new Tavg uncertainty of 4.3*F is bounded by the current analysis assumptions and the mass and energy release data remains applicable.

Because the analysis assumptions do not change, the conclusions of References 7 and 8 remain valid.

Steamline Break Mass and Energy Releases Inside Containment (15.4.2.1).

Steamline break mass and energy release data are calculated for several break sizes at different power levels for the purposes of calculating the containment pressure and temperature response.

The 1".:,'y:i: ta

• ulate mass and energy releases inside containment assumed a Tavg uncertainty of 4'F.

The analysis also used a conservative ANS 1971 decay heat model.

Raising the Tavg uncertainty to 4.3*F, and using the less limiting ANS 1979 decay heat model, results in mass and energy release data which is negligibly different from the current analysis.

It is therefore concluded that the current mass and energy release data remains applicable for Farley Units 1 and 2, and the conclusions

~

of Reference 7 remain valid.

l 08020:1D/082290 38 l

1 4

i 4.

1.2 CONCLUSION

1 Th'e effects of the increase in Tavg uncertainty have been evaluated for all of I

the non-LOCA transients.

The zero power transients are not affected by the change. The DNB related transients have been shown to be acceptable by using existing DNB margin. The preceding discussions for the remaining transients i

demonstrate that the conclusions of the FSAR and References 7 and 8 remain b

valid, and the steamline break mass and energy release data remains applicable l'

for the Farley units, l

4.3.3 EVALVATION OF THE LOS$ OF FLOW REACTOR TRIP SETPOINT The uncertainty for the loss of flow trip has increased with the RTD bypass elimination.

In order to maintain the same Technical Specification trip l

setpoint, a lower analysis value was required. The current analysis value is

~

87% of nominal loop flow.

The revised analysis value is 85% of nominal 1oop f~

flow. Two transients _' rely on the'10w 100p flow reactor trip; partial loss of flow and locked rotor.

A discussion of the analysis performed for these transients follows, f

partial Loss of Forced Reactor Coolant flow (15.2.5)

L-l 1

This analysis was performed similar to the analysis presented in the FSAR.

Because Farley Units.1 'and-2 are not licensed to operate at power with only 1

two reactor. coolant pumps in operation, a partial' loss of flow with three reactor coolant pumps initially operating was analyzed.

Three digital computer codes were used in the analysis.

The LOFTRAN code (Reference 3) was used to calculate the flow coastdown, RCS transient conditions, the nuclear power transient, and th'e reactor trip on low loop flow.

The FACTRAN code (Reference 4) was used to calculate the heat flux transient _ based on the-L

_ nuclear power and flow data from LOFTRAN.

Finally, the THINC code (described

'in.Section-4.4 of.the FSAR) was used.to calculate'the minimum DNBR during-the transient based _on the_ heat-flux from FACTRAN and the flow from LOFTRAN.

~

Conservative initial conditions were assumed _which--included a 5.5'F uncertainty for lavg._ The-low flow trip setpoint was assumed to be 85% of

. nominal _ flow.

The effects' of increased steam generator tube plugging were

.also modeled in the analysis (References 7'and 8).

[

l08020:10/082290 39

--_,,_.,_.,,.....,,.,m..v..-.

,,.,,,_,,..._,_m.,.m.,.,_.,,,,,_,,..,._,_,J._.___.

.,m...,_,,..

,.,.,_-.~,4._,_

The results of the analysis confirmed that the minimum DNBR during the transient remained above the limit value.

Therefore, the revised low flow trip setpoint and the increased Tavg uncertainty and the effects of increased SG tube plugging have been shown to be acceptable for this transient.

Refer to References 7 and 8 for the transient plots for this analysis.

Single Reactor Coolant Pump Locked Rotor (15.4.4) s This analysis was performed similar to the analysis presented in the FSAR.

Because Farley Units 1 and 2 are not licensed to operate with only two reactor coolant pumps in operation, the analysis modeled three reactor coolant pumps initially operating.

Two digital computer codes were used in the analysis.

The LOFTRAN code (Reference 3) was used to calculate the flow coastdown, RCS transient conditions, the nuclear power transient, the reactor trip on low loop flow, and the peak RCS pressure.

The FACTRAN code (Reference 4) was used to calculate the thermal behavior of the fuel at the core hot spot based on the nuclear power and flow data from LOFTRAN.

Conservative initial conditions were assumed which included a 5.5'F uncertainty for Tavg.

The low flow trip setpoint was assumed to be 851. of nominal flow.

The effects of increased steam generator tube plugging were also modeled in the analysis (References 7 and 8).

The results of the analysis confirmed that the peak RCS pressure remained below that which would ceuse stresses to exceed the faulted condition stress limits.

In addition, the calculated zirconium-water reaction remained a small fraction, and the peak clad surface temperature was less than 2700'F.

Therefore, the revised low flow trip setpoint and t'ne increased Tavg uncertainty and the effects of increased SG tube plugging have been shown to be acceptable for this transient.

Refer to References 7 and 8 for the transient plots for this analysis.

4.3.4

SUMMARY

In summary, non-LOCA safety analyses applicable to the Farley Units 1 and 2 have been evaluated for the replacement of the existing RTD Bypass System with fast response thermowell mounted RTDs installed in the reactor coolant loop piping.

It is concluded that an increase in RCS temperature uncertainty 08020:10/082290 40

9 can be accommodated by the margins in the safety analyses and allocation of gtneric DNB margin.

In addition, it has been demonstrated by analysis that the revised analysis value for the loss of flow reactor trip setpoint is acceptable. All other safety analysis assumptions remain valid.

The evaluations have also considered the tube plugging effects of References 7 and B.

The FSAR and References 7 and 8 conclusions applicable to the J. M. Farley Units 1 L 2 are unchanged and all applicable non-LOCA safety analysis acceptance criteria continue to be met.

4.4 LOCA Evaluation l

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 out'et temperatures, thermal design flow rate and g

l the steam generator performance data used in the LOCA analyses will be slightly affected.

The evaluation of the slight increase in the Tavg uncertainty has resulted in an estimated increase of 3'F for the Large Break LOCA Peak Cladding Temperature (PCT) and a 2'F increase for the Small Break LOCA PCT.

There is sufficient margin to 2200'F for both the Large and Small Break LOCA analyses to offset the estimated increases due to RTD bypass elimination at the Farley Units.

The analytical results represented in References 7 and B include the effect of these PCT increases.

4.5 INSTRUMENTATION AND CONTROL (ILC) SAFETY EVALUATION The RTD Bypass Elimination modification for the J. M. Farley Units 1 & 2 does not functionally change the oT/T pr tection channels. The implementation avg of the fast response RTDs in the reactor coolant piping will change the inputs

[

to the AT/T Protection Set I, II, and III, circuitry as follows:

avg 1

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 I

T signal. One element of the RTD will be held in reserve as a spare.

cold i

08020:1D/082290 41

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 similar means as before the modifications, i.e., existing control board alarms and protection A

channel indicators, except that the control sysums will not be sensitive to RTD failures or protection channel failures due to HSS.

4.

The NR cold leg RTD signals and the NR hot leg RTD signals are electronically processed in the plant 7300 series process protection racks to generate loop T and delta T signals.

These signals (one per loop) areelectronicallyi!olatedandtransmittedtotheplant7300 series av process control racks.

The T and delta T signals are input to a av Median Signal Selector, respect vely, which selects the median signal for use in the plant control systems.

By rejecting the high and low signals, the control system will not act on any single failed input channel.

Since no adverse control system action therefore results from a single failed instrument channel, a second random failure is not required per IEEE 279-1971, section 4.7.

The existing protection channel control board T,yg and delta T indicators and alarms will provide the means of identifying RTD failures. As part of the RTD Bypass Elimination modification, the electronically isolated T and avg delta T signals will be utilized for control grade signals and alarms which can also be utilized to detect failed RTD or a protection channel input signal.

Upon identification of a failed hot leg or cold leg RTD, the operator would request that I&C personnel place the failed protection channel in a tripped condition, identify the failed RTD, disconnect the failed RTD, connect the other RTD in the dual element device and rescale the applicable RTD amplifier.

After this process, the channel would be returned to service.

If both RTDs in a dual element device are bad, the RTD input is removed from the averaging process and a bias is 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 prior to the failure of the dual element RTD.

l 0802D:1D/082290 42

The conversion to thermowell mounted RTDs will result in elimination of the control grade RTOs and their associated control board indicators.

The i

protection grade channels will now be used to provide inpuis to the control system through electrical 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 or AT signal avg (there will be a separate MSS for each function).

The Median Signal Selector will use as inputs the isolated protection grade T or AT signals from avg 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 control grade systems will still use a valid RCS temperature in the case of a single signal failure.

To ensure proper action by the Median Signal Selector, the present manual switches that allow for defeating of a T or AT signal from a single avg 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 and 6T deviation alarms.

avg Other than the above changes, the Reactor Protection System and Control System will remain the same, as that previously utiliz;d.

For example, two out of three voting logic continues to be utilized for the thermal overtemperature and overpower protection functions, with the model 7300 process control bistables continuing to operate on a "de-energize to trip" principle.

Nonsafety-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 asec;iated 10CFR50 General Design Criteria (GDC), Regulatory Guides, and other applicable industry standards.

IEEE Standard 279-1971 requires documentation of a design basis.

Following is a discussion of design basis requirements in conformance to pertinent I&C criteria, s

08020:10/082290 43

a.

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

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 Appendix 17C of the FSAR.

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.

d.

Channel independence and electrical separation is maintained because the Protection Set circuit assignments continue to be RCS Loop 1 circuits input to Protection Set I; RCS Loop 2 to Protection Set II; and RCS Loop 3 to Protection Set III, with appropriate observance of field wiring interface criteria to assure the independence, 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 channels.

The details of the signal selector implementation are contained in Section 1.3.1 and Section 4.5.

On the basis of the foregoing evaluation, it is concluded that the compliance of the Farley units to IEEE Standard 279-1971, applicable GDCs, and industry sta.ndards and regulatory guides has not been changed with the I&C modifications required for RTD bypass removal.

08020:1D/082290 44

9 4.6 HECHANICAL SAFET1' EVALUATION The presently insts.11ed RTD bypass system is to be replaced with fast acting narrow range RTD thernowells.

This change requires modifications to the hot leg scoopt, 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.

The hot leg temperature measurement on each loop will be accomplished using thren (3) fe.st response, narrow range single element RTDs mounted in thermowells. To accomplish the sampling function of the RTD bypass manifold system and minimize the need for additional hot leg piping penetrations, the RTD thermcwell assemblies will be located within the existing RTD Bypass Manifold Scoops wherever possible.

aJ.c to provide

[

the Droper flow path.

If structural interferences preclude the placement of a thernowell in a given scoop, then the scoop will be capped and a new RCS penetration made to accommodate the relocated thermowell.

The relocated thermowell will be located in an installation boss. A thermowell design will be used such that the thermowell will be positioned to provide an average temperature reading.

The thermowell and installation boss will be fabricated in accordance with Section III (Class 1) of the ASME Code.

The installation of the thermowell into the scoop or boss will be performed using Gas Tungsten Are Held (GTAH) for the root pass and finished out with either GTAH or Shielded Metal Art Held (SMAH).

The welding will be examined by penetrant test (PT) per the ASME Code Section XI.

Prior to welding, the surface of the scoop or boss onto which welding will be pir'ormed will be examined as required by Section XI.

The cold leg RTD 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 GTAH for the root pass and finished with either GTAH or SHAH, Held inspection by PT will be performed as a3c required by Section XI.

The thermowells will extend approximately (

inches into the flow stream.

Tnis depth has been justified based on 3,c analysis.

The root weld joining the thernowells to a

[

0802D:10/082290 45

the modified nozzles will be deposited with GTAW and the remainder of the i; eld may be deposited with GTAW or SMAN.

Peneient testing will be performed in accordance with the ASME Code Section Ta.

The thermowells will be fabricated in accordance with the ASME Section II (Class 1).

The cross-over leg bypass return nozzle will be modified and capped or the existing piping connection will be severed to leave a stub of pipe protruding from the nozzle and the stub will be capped.

The cap design, including materials, will mee 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 GTAH or :> MAW.

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

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 ( J c inches, a system hydrostatic test is a

required after the bypass elimination modification is complete.

Paragraph 108-5222 of Section XI defines this test pressure to be 1.02 times the normal operating pressure at a temperature of GOO'F or greater.

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.

4.7 TECHNICAL SPECIFICATION EVALUATION As a result of the calculttions 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.

08020:10/082290 46

5.0 CQETROL SYSIEH EVALUA110![

A prime input to the various KSSS control systems is the RCS averatr temperature, T(avg).

This is calculated electronically as the average of N measured hot and cold leg temperatures in each loop.

The effect of the new RTD temperature measurement sistem it to potentially change the time response of the T(avg) channels in ne varicus loops.

This in turn could impact the response of (

'J"'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 (HSS) is small in comparison with the RTD time response

((

)]"'C.

Therefore, there will be no significant impact on the T(avg) channel response and no 4

need, as a result of implementing the new system, to revise any of the control system setpoints.

However APC0 always has the option of K, king setpoint adjustnots.

If desired, system performance can be 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 1

accident analysis.

Hith respect to accident analyses, control systems are assumed operative only in cases in which their action aggravates the consequences of an event, and/or as required to establish inhial plant conditions 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.

0802D:10/082290 47

c.0 COMChuS10NS The method of utilizing fast-rpsponse RTDs installed in the reactor cealant loop piping as a means for RC3 temperature indication has undergone exter.tive at,alyses, evaluation and testing as described in this report.

The 6

incorpcrAtton of this system into the J. H. Farley Nuclear F' ants Units 1 and 2 design meets all safety, licensing and control requirements necessary for safe operation of these units.

The analytical evaluation has been supplemented with in-plant and laboratory testing to further verify system perfort;ance.

The fast response RTOs installed in the reactor coolant loop piping adequately replace the present hot and cold leg temperature measurement system and enhance ALARA efforts as well as improve plant reliability.

In addition to the effects of the RTD Bypass Elimination, this evaluation also consider the effects of increased SG tube plugging and reduced RCS flowrate as described in References 7 and 8.

1 i

N 08020:10/082290 48

h 1.

7.0 REFERENCES

11. FNP Final Safety Analysis-Report L

2.-Butler, J. C.,Elove D. S., "Steamline Break Mass / Energy R' leases for Equipment Environmental Qualification Outside Containment," WCAP-10961, Revision 1 (Proprietary), October 1985.

3. Burnett,= T. H. T., et al., "LOFTRAN Code Description," HCAP-7907-P-A

-(Proprietary),- HCAP-7907-A (Non-Proprietary), April 1984.

4.~Hargrove, H.G., "FACTRAN - A FORTRAN IV CODE FOR THERMAL-TRANSIENTS IN A UO FUEL '.00,".HCAP.7908-A (Non-Proprietary), Decembre 1989.

2

5. - ANSI / ANSI 5.1-1979, ~American National Standard for Decay Heat Pcwer in Light Hater Reactors, Augu:t 29, 1979.

6. Nureg-0717, Supplement No. 4. " Safety Evaluation Report Relatet! to the Operation of-Virgi' C. Summer Nuclear Station, Unit No.1," Docket No.

50-395, August,z1982.-

i

7. Morrison,- R. J.,:" Alabama Power-Joseph M. Farley: Unit 2 Increased Steam.

Generator Tube Plugging and Reduced Thermal Design Flow Licensing Report,"

HCAP-12659. (Non-Pioprietary), July 1990.

.8. Morrison, R.:R.

" Alabama Power Josepn M. Farely Unit 1 Increased Steam Generator Tube.Pluggin; and Reduce Thermal _ Design Flow Licensing Report,"

HCAP-12694 -(Non-Proprietary), Auguet 1990.-

9. FNP.-Precautions, Limitations and Setpoints.

10.'Ciocca, t.

F., et-al., " Evaluation-of the Impact of: Cable Splices and E

Penetration Leakage On RPS/ESFAS and _ERP -Setpoints, Farley Units.1 -and 2."

hCAP-11658. Pev L1, August,- 1989.

4

- 0802DilD/082290 49

' =- ------ ----.

1

• m

i APPENDIX A DEFINITION OF AN OPERABLE CHANNEL AND HOT LEG RTD FAILURE COMPENSATION PROCEDURE i

08020:10/082290 50

RTD BYPASS ELIMINATION FOR J. M. FARLEY NUCLEAR PLANT UNITS 1 AND 2 DEFINITION OF AN OPERABLE 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 in whole or in part without the written authorization of Westinghouse Electric Corporation.

Westinghouse Electric Corporation 4

Pittsburgh, PA

-38020:lD/082290 51

DEFINIT 10N13F'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 thati RTD's: input to the averaging function.

J. H. Farley Nuclear Plant-(FNP) will

have dual element RTDs installed.in each hot leg thermowell location.

The 1

"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 invoked.

Sinale RTO 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 [:

c

.]a,e : Typically this data is recorded at initial 100% power and, ther.eafterp during-the normal protection channel surveillance interval.

]a,c any hot leg can then Once

[-

tolerate' failure of both elements of a single dual element RTD and still f

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.

[

l' l

j

.]a c-L a

08020:10/082290 52 m

, z.

x The plant may operate with asfailed hot leg RTD at any power level during that q

tsame fuel cycle.

It is permissible to shutdown _and startup during the cycle

--/

'without requiring.that the failed RTD be replaced.

1 4

ga 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 the

contsl. systems. -If another hot leg RTD fails in a different loop the FNP

.'should operate using manual-control.

Manual Rod Control-is recommended so

'that;the-operator can control the plant based on the best measurement

available.-_If automatar operation is continued the control system may choose j

the biased channel due to.the positive-(or zero) biat application. 'This means the control system'wi_11. perceive a higher Tavg than actually exists at reduced power.'and the pla'nt-wil1~ operate at reduced temperatures. While this is not

]

'necessarily' undesirable it does reduce the total plant megawatt output.- The

?use'of automatic rod control should-be considered based on utility power j

i requirementst

- Cold Leg:

Ifithe active cold-leg RTD fails, then that RTD should be t

disconnected from'the 7300 cabinets.

The dual element spare RTD should'then

_be connected in the failed RTD's place.

l q

Double RTO Failure:

Inocerable Channel i

Hott leg or

Cold Leg:

If_ both elements.of two or more of the three hot leg 3

' dual: element RTDs.or the cold' leg dual elements 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 hot. leg-RTD.is not: presently

.' analyzed as'.part,of the licensing basis.

.f 1

1 5

0'B'02DilD/082290 53

4 P_ROCEDURE FOR OPERATION WITH A HOT LEG DUAL ELEMENT RTD QULQF SERVICE The hot leg temperature measurement is obtained by averaging the measurements from the three thermowell RTDs installed on the hot leg of each loop.

[

2

,)a,c In the event that one of the three dual element RTDs fails, the failed RTD will be disconnected and the hot 100 temperature measurement will be obtained by averaging the remaining two RTO me,'.surements.

[

,]a.C r

The bias adjustment corrects for

(

[

.)a 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,

[

,3

.C a

?3 l

1 08020:'0/082290 54 I

An RTO failure will most likely result in an offstale high or low indication and will be detected through the normal means in use today (i.e., T and avg 6T deviation alarms and indicators).

Although unlikely, the RTD (or its electronics channel) can fail gradually, causing a gradual change in the loop temperature measurements.

[

,)a.C The detailed procedure for correcting for a f ailed hot leg RTD is presented below:

a,c 0802D:10/082290 55

~aC 08020:10/082290 56 1

I

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. j 1

I f

t 3

+

l l

..08020:1D/082290 57 A

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4 APPENDIX

. CALCULATION OF HOT-LEG TEMPERATURE BIAS a.C 4

i l

l l

08020:10/082290 58

4 APPENDIX B ACRONYMS FOR UNCERTAINTY CALCULATIONS l-l-

08020:10/082290-59

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