ML20207H976

Nonproprietary Resistance Temp Detector Bypass Elimination Licensing Rept for Byron 1 & 2 & Braidwood 1 & 2

ML20207H976
Person / Time
Site:	Byron, Braidwood, 05000000
Issue date:	11/30/1986
From:	Ditommaso S, Rice W WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML19292G540	List: ML20207H873 ML20207H901 ML20207H947 ML20207H976
References
WCAP-11324, NUDOCS 8701080243
Download: ML20207H976 (71)

Text

WESTZNGHOUSE CLASS 3 WCAP-11324 RTD BYPASS ELIMINATION LICENSING REPORT l FOR BYRON UNITS 1 & 2 AND BRAIDWOOD UNITS 1 & 2 NOVEMBER, 1986 W. R. RICE S. M. DITORMSO WESTINGHOUSE ELECTRIC CORPORATION Pittsburgh, Pennsylvania NOIOfob$k Sbb j4 P

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

E. Manz R. A. Holmes E. K. Hackmann J. C. Bass J. C. Mesmeringer M. Weaver 97870 1D/110586

TABLE OF CONTENTS .

Section g List of Tables i List of Figures 11 1.0 Introduction 1 1.1 Historical Background 1 ,

1.2 Mechanical Modifications 2 1.3 Electrical Modifications 4 2.0 Testing 10 2.1 Response Time Test 10 I 2.2 Streaming Test 10 3.0 Uncertainty Considerations 13 3.1 Calorimetric Flow Measurement Uncertainty 13 3.2 Hot Leg Temperature Streaming Uncertainty 13 4.0 Safety Evaluation 20 4.1 Response Time 20 4.2 RTD Uncertainty 20 4.3 Non-LOCA Transients Reanalyzed 21 4.4 LOCA Evaluation 24 4.5 Instrumentation and Control Safoty Evaluation 25 4.6 Mechanical Safety Evaluation 28 4.7 Technical Specification Evaluation 30 5.0 Control System Evaluation 54 6.0 Conclusions 55 -

7.0 References 56 Appendix A- Technical Specification Modifications

• 97870:10/112186 l . . . . . . . _ . . . . . . . . . . . _ . . . . . . . . . _ _ _

LIST OF TABLES Table Title Pace 2.1-1 Response Time Parameters for RCS Temperature Measurement 12 3.1-1 Flow Calorimetric Instrumentation Uncertainties 16 3.1-2 Flow Calorimetric Sensitivities 17 3.1-3 Calorimetric RCS Flow Measurement Uncertainties 18 3.1-4 Cold Leg Elbow Tap Flow Uncertainty 19 4.3-1 Time Sequence of Events for a RCCA Bank Withdrawal 31 at Power 4.3-2 Time Sequence of Events for a Turbine Trip 33 O

9787Q:10/110586

LIST OF FIGURES Figure Title Page 1.2-1 Hot Leg RTD Scoop Modification for Fast-Response 6 RTD Installation 1.2-2 Cold Leg Pipe Nozzle Modification Fast-Response 7 RTD Installation 1.2-3 Additional Boss for Cold Leg Fast-Response RTD 8 Installation 1.3-1 RTD Averaging Block Diagram, Typical for Each of 4 9 Channels 4.3-1 Nuclear Power, Core Heat Flux, and Core Average 35 Temperature for a RCCA Bank Withdrawal at Full Power with Minimum Reactivity Feedback (75 PCM/SEC Rate) 4.3-2 Pressurizer Pressure, Water Volume, and DNBR for a RCCA 36 Bank Withdrawal at Full Power with Minimum Reactivity Feedback (75 PCM/SEC Rate) 4.3-3 Core Average Temperature, Heat Flux, and Nuclear Power 37 for a RCCA Bank Withdrawal at Full Power with Minimum Reactivity Feedback (3 PCM/SEC Rate) 4.3-4 Pressurizer Pressure, Water Volume, and DNBR for a 38 RCCA Bank Withdrawal at Full Power with Minimum Reactivity Feedback (3 PCM/SEC Rate) 4.3-5 Nuclear Power, Heat Flux and Core Average Temperature , 39 for a RCCA Bank Withdrawal at Full Power with Maximum Reactivity Feedback (75 PCM/SEC Rate) ii 97870:1D/110586

LIST OF FIGURES (Cont.)

Figure Title Page 4.3-6 Pressurizer Pressure, Water Volume, and DNBR for a 40 RCCA Sank Withdrawal at Full Power with Maximum '

Reactivity Feedback (75 PCM/SEC Rate) 4.3-7 Nuclear Power, Heat Flux and Core Average Temperature 41 l for a RCCA Bank Withdrawal at Full Power with Maximum l Reactivity Feedback (3 PCM/SEC Rate) 4.3-8 Pressurizer Pressure, Water Volume and DNBR for a RCCA 42 Bank Withdrawal at Full Power with Maximum Reactivity p I

Feedback (3 PCM/SEC Rate) r 4.3-9 Minimum DNBR vs. Reactivity Insertion Rate for a RCCA 43 Bank Withdrawal at Full Power 4.3-10 Minimum DNBR vs. Reactivity Insertion Rate for a RCCA 44 Bank Withdrawal at 10% Power 4.3-11 Minimum DNBR vs. Reactivity Insertion Rate for a RCCA 45 Bank Withdrawal at 10% Power 4.3-12 Pressurizer Pressure, Water Volume and Nuclear Power 46 for Turbine Trip With Pressure Control and Minimum Reactivity Feedback 4.3-13 Core Inlet Temperature, Coolant Average Temperature 47 and DNBR for Turbine Trip With Prossure Control and Minimum Reactivity Feedback iii 97870:1D/111386 -

' ~

LIST OF FIGURES (Cont.)

F,igu_re Title Page 4.3-14 Pressurizer Pressure, Water Volume and Nuclear Power 48 for Turbine Trip With Pressure Control and Maximum Reactivity Feedback 4.3-15 Core Inlet Temperature, Coolant Average Temperature 49 and DNBR for Turbine Trip With Pressure Control and Maximum Reactivity Feedback 4.3-16 Pressurizer Pressure, Water Volume and Nuclear Power 50 for Turbine Trip Without Pressure Control and Minimum _

Reactivity Feedback 4.3-17 Core Inlet Temperature, Coolant Average Temperature 51 and DNBR for Turbine Trip Without Pressure Control and Minimum Reactivity F.eedback 4.3-18 Pressurizer Pressure, Water Volume and Nuclear Power 52 for Turbine Trip Without Pressure Control and Maximum Reactivity Feedback 4.3-19 Core Inlet Temperature, Coolant Average Temperature 53 and DNBR for Turbine Trip Without Pressure Control and Maximum Reactivity Feedback iv 97870.1D/110586

l

1.0 INTRODUCTION

{

)

Westinghouse Electric Corporation has been contracted by Connonwealth Edison l Co. (CECO) to remove the existing Resistance Temperature Detector (RTD) Bypass l System and replace this hot leg and cold leg temperature measurement method

! with fast-response RTDs installed in the reactor coolant loop piping. This report is submitted for the purpose of supporting operation of the Byron Units 1 and 2 and Braidwood Units 1 and 2 (Byron /Braidwood) with the new RTD System i installed. I 1.1 HISTORICAL BACKGROUND

(

Prior to 1968, PWR designs had been based on the assumption that the hot leg temperature was uniform across the pipe. Therefore, placement of the temperature instruments was not considered to be a factor affecting the l 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 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 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 it's 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.

9787o:1D/111386 1 -

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

The proposed temperature measurement modification has been develop d in response to both sets of problems encountered in the past. Specifically:

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

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

o Use of thermowells permits RTD replacement without draining the 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 C'ontrol and Protection System will be obtained using RTDs installed in each reactor coolant loop. .

1.2.1 Hot Leg a) Byron 1 and Braidwood 1 The hot leg temperature measurement on each loop will be accomplished with three fast response. narrow range RTDs mounted in thermowolls. To accomplish the sampling function of the RTD bypass manifold system and eliminate the need for additional hot leg piping penetrations, the thermowcils will be located within. the three existing RTD bypass manifold scoops. A hole will be drilled through the end of each scoop so that 97870:1o/110586 2

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). These three RTDs will measure the hot leg temperature which is used to calculate the reactor coo'lant loop differential temperature (AT) and average temperature (T,yg).

b) Byron 2 and Braidwood 2 In order to take advantage of a non radioactive environment inside containment prior to plant operation, independent bosses and RTD thermowells (without scoops) have been installed. The RTD thermowells (Figure 1-2.3) are located 120 degress apart in the same plane, thereby providing the same averaging function as if the RTDs had been mounted in the existing scoops. These three RTDs are used in the same manner as described .in paragraph 1.2.1(a).

c) This modification will not affect the single wide range RTD currently installed near the entrance of each steam generator. This RTD will continue to provide the hot leg temperature used to monitor reactor coolant temperature during startup, shutdown, and post accident conditions.

1.2.2 Cold Leg a) One fast response, narrow range, RTD will be located in each cold leg at the discharge of the reactor coolant pump (as replacements for the cold leg 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. For Byron 1 and Braidwood 1, the existing cold leg RTD bypass penetration nozzle will be modified (Figure 1.2-2) to accept the RTD thermowell. For Byron 2 and Braidwood 2, a new penetration will be made and an RTD thermowell installed as described in paragraph 1.2.2(c).

s7s7o.to m osas 3

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

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

c) A new penetration will also be made to each cold leg to accept an additional therrowell mounted narrow range RTD, for use as an installed spare. This will give the new modification a tolerance for RTD failures equivalent to the bypass loops. A new cold leg boss will be added (Figure 1.2-3) to accept the RTD thermowell.

1.2.3 Crossover Leo When RTD bypass elimination is implemented for each unit, the RTD bypass manifold return line will be capped at the nozzle on the crossover leg.

1.3 ELECTRICAL MODIFICATIONS 1.3.1 Function

~

Figure 1.3-1 shows a block diagram of the modified electronics. The hot leg R.TD measurements (three per loop) will be electronically averaged in the process protection system. The averaged T hat signal will then be input to the appropriate protection function. This will be accomplished by additions to the existing process control equipment.

1.3.2 Qualification Equipment seismic and environmental qualification will be to IEEE standards 344-1975 and 323-1974, respectively, as described in WCAP-8587, Rev. 5,

" Methodology for Qualifying Westinghouse WRD Supplied NSSS Safety Related Electrical Equipment".

978701D/110586 4

1.3.3 RTD Operability Indication .

Existing control board AT and T,yg indicators and alarms will provide the \

l means of identifying RTD failures. The spare cold leg RTD provides, sufficient spare capacity to accommodate a single cold leg RTD failure per lo6p. Failure of a hot leg RTD will require manual action to defeat the failed signal, and a manual rescaling of the electronics to average the remaining signals (see i

Figure 1.3-1).

e 97870.)o/11058s 5

a, c Figure 1.2-1 Hot Leg RTD Scoop Ndification fcr Fast Response R11) Installation 6

a,c l

Figure 1.2-2 Cold Leg Pipe Nozzle Modification for Fast Response RTD Installation

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S 7

1 1

Figure 1.2-3 Additional Bosses for Hot and Cold Leg Fast-Response RTD Installation ,

8 j

n,c

\

t A

F Figure 1.3-1 KTD Averaging Block Diagem. Typical fcr Each of 4 Channels 97870.10/1105sa 9

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l l 2.0 TESTING There are two specific tests which have been performed to support the installation of the fast-response RTDs in the reactor coolant pipirig: a response time test and a hot leg temperature streaming test.

2.1 RESPONSE TIME TEST Westinghouse has performed an RTD response time test at its Forest Hills Test Facility. This test placed a fast response RTD, manufactured by RdF Corporation, inside a scoop, within a thermowell, which modelled the actual I in plant installation. The flow conditions were adjusted to equal the high l velocity reactor coolant system flows of approximately [ Ja,b,c ,

~

The RTD's response time is determined based on a comparison of the RTD with {

thermocouples which had been previously calibrated and response time I characterized. Sixty-five test runs were made at various flow rates while l gathering data on 2 RTDs. The test results demonstrated a mean response time I for the RTD, thermowell and scoop of less than [ Ja,b,c seconds. Table 2.1-1 provides a comparison of the original RTD Bypass System response time

(

and how it differs from the new fast response thermowell system.

. 2.2 STREAMING TEST Past testing at Westinghouse PWRs has established that temperature stratification exists in the hot leg pipe with a temperature gradient from top to bottom of [ ]b,c.e A test program was implemented at an operating plant to confirm the temperature streaming magnitude and stability with measurements of the RTD bypass branch line temperatures on two adjacent (

reactor coolant loops. Specifically, it was intended to determine the magnitude of the differences betv en branch line temperatures, confirm the short-term and long-term stability o. the temperature streaming patterns and

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, 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 i

~

specific data is used in conjunction with data taken from other Westinghouse 0787o;1D/110586 10 l

designed plants to determine an appropriate temperature error for use in the safety analysis and calorimetric flow calculations. Section 3 will discuss the specifics of these uncertainty considerations.

~

The test data has been reduced and characterized to answer the three objectives of the test program. First, it is conservative to state that the streaming pattern [ ]D'C. Steady state data taken at l'00% power for a period of four weeks indicates that the streaming pattern [ ]b,c.e In other words, the temperature gradient [

]b,c.e This is inferred by [.

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

]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. This significantly reduces the error introduced by a failed RTD.

The test data also supports previous calculations of streaming errors determined from tests at other Westinghouse plants. The recent data is consistent with the upper bound temperature gradients that characterize the previous data. There were no new discoveries, but the data did add a dimension previous tests did not have. The test sampled temperatures from the pipe interior while all previous tests investigated temperature gradients at the pipe surface. The pipe internal temperature data has greatly strengthened the assumptions and inferences made with previous test data.

The streaming test and response time test have both provided valuable information needed to support the design of the fast response RTDs installed in the reactor coolant piping.

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97s7Q:10/110586 11.

1. . . --. .

TABLE 2.1-1 RESPONSE TIME PARAMETERS FOR RCS TEMPERATURE MEASUREMENT Fast Response RTD Bypass System Thermowell RTD System

_ _ a,c _

_ a,c RTD Bypass Piping and Thermal Lag (sec)

RTD Response Time (sec)

RTD Filter Time Constant (sec)

Electronics Delay (sec)

Total Response Time (sec) 6.0 sec 7.0 sec l

l l

l B

97870:1D/1105ss 12

3.0 UNCERTAINTY CONSIDERATIONS This new method of hot leg temperature measurement has been analyzed to determine if it will have an impact on two uncertainties included in the Safety Analysis: Calorimetric Flow Measurement Uncertainty and Hoi Leg Temperature Streaming Uncertainty.

3.1 CALORIMETRIC FLOW MEASUREMENT 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 are the narrow range hot leg and cold leg coolant temperatures. The accuracy of the RTDs has, therefore, a major impact on the accuracy of the flow measurement.

The current licensed flow measurement uncertainty for Byron /Braidwood for the sum of the four loop flows including albow taps, is about + 2.1% flow (not including 0.1% flow for feedwater venturi fouling allowance). However, with ,

the use of three T hot RTDs (resulting from the elimination of the RTD Bypass lines) and tha latest Westinghouse RTD cross-calibration procedure (resulting in lower RTD calibration uncertainties at the beginning of a fuel cycle), it is possible to reduce the RCS flow measurement uncertainty to approximately 3 1.8% flow (including the cold leg elbow taps and excluding feedwater venturi fouling). Utilizing the uncertainty calculational methodology explicitly described in WCAP-11168-R1 (Reference 1), Tables 3.1-1 through 3.1-4 were generated to provide the Byron specific instrument uncertainties, calorimetric sensitivities, and flow uncertainties. Prior to bypass elimination implementation at the Braidwood units, these values must be reviewed for applicability to Braidwood.

3.2 HOT LEG TEMPERATURE STREAMING UNCERTAINTY 4

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 s7snkiomates ,

13

an analysis of test data from other Westinghouse plants, and on calculations to evaluate the impact on temperature measurement accuracy of numerous possible temperature distributions within the hot leg pipe. The test data has shown that the circumferential temperature variation is no more than (

)b,c.e. and that the inferred temperature gradient within the pipe is limited to about

[ ]b,c.e The calculations for numerous temperature distributions have shown that, even with margins applied to the observed temperature gradients, the three point temperature measurement (scoops or thermowell RTDs) is very effective in determining the average hot leg temperature. The most recent calculations for the thermowell RTD system have established an overall streaming uncertainty of [ ]b,c.e for a hot leg measurement. Of this total. [

]b , c', e. The overall temperature streaming uncertainty applied to the calorimetric flow measurement is only slightly larger than the uncertainty used in previous analyses. 1 The new method of measuring hot leg temperatures, witii the thermowell RTDs located within the three scoops, is at least as effective as the existing RTD

~

bypass system, [

]. Although the new method measures temperature at one point within the thermowell, compared to the five sample points in a 5-inch span of the scoop measurement, the thermowell measurement point is opposite the center hole of the scoop and therefore measures the j equivalent of the average scoop sample if a linear radial temperature gradient exists in the pipe. The thermowell measurement may have a small error i

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 ! ]b,c.e exists from the center to the end of the scoop, the differenes between the thermowel'1 and scoop measurement is limited to [ l b,c.e . Since three RTD measurements are averaged, and the nonlinearities at each scoop are random, the effect of this error on the hot leg temperature measurement is limited to [ ]b,c.e On the other l

87870:10/112186 14

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

3a.c ,

In all casas, the flow imbalance uncertainty will equal or exceed the

[ ']b,c.e sampling uncertainty for the thermowell RTDs, so tho' new measurement system tends to be a more accurate measurement with respect to streaming uncertainties.

Temperature streaming measurements from testing at an operating plant have been obtained. The measurements confirm the [

)b,c.e ,

Over the 4-week testing period, 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 ,

l Provisions were made in the RTD electronics for operation with only two hot t

l 1eg RTDs in service. The two-RTD measurement will be biased to correct for the difference compared with the three-RTD average. Based on recent test data, the bias would be limited to between [ ]b,c.e Data comparisons show that the magnitude of this bias varied less than

[ ]b,c.e over the test period.

l I

i 97870:1D/112186 15

TABLE 3.1-1 FLOW CALORIMETRIC INSTRUMENTATION UNCERTAINTIES FW TEMP FW PRES FW d/p STM PRESS T T, PRZ PRESS H

1. OF INST USED 3 1 1 **

'F psia  % d/p psia 'F 'F psia INST SPAN = 618. 2000. 120. 1500. 100. 100. 3000.

INST UNC.

(RANDON) = a,e INST UNC. ,

=

(BIAS) ___

NOMINAL = 440. 990. 990. 618.4 558.4 2250..

• • Number of Hot Leg and Cold Leg RTDs used for measurement in each loop and the number of Pressurizer Pressure transmitters used overall, i.e., one per loop.

l l

l 97870;1D/112186 16

TABLE 3.1-2 FLOW CALORIMETRIC SENSITIVITIES FEEDWATER FLOW -

F, TEMPERATURE = a,c MATERIA'l =

DENSITY TEMPERATURE =

PRESSURE =

DELTA P =

FEEDWATER ENTHALPY TEMPERATURE =

PRESSURE =

= 1193.3 BTU /LBM h, -

h = 419.4 BTU /LBM f

= 773.8 BTU /LBM Dh(SG)

STEAM ENTHALPY _ _

PRESSURE = a,c MOISTURE =

HOT LEG ENTHALPY TEMPERATURE =

PRESSURE =

h = 640.5 BTU /LBM H

h = 557.7 BTU /LBM c

i = 82.8 BTU /LBM Dh(VESS)

Cp(T ) = 1.550 BTU /LBM *F H

COLD LEG ENTHALPY .

TEMPERATURE = a,c PRESSURE =

=

Cp(Tc ) 1.264 BTU /LBM *F COLD LEG SPECIFIC VOLUME TEMPERATURE = a,c PRESSURE =

l ora 7o:io/iiziss 17

TABLE 3.1-3 CALORIMETRIC RCS FLOW MEASUREMENT UNCERTAINTIES COMPONENT INSTRUMENT ERROR FLOW UNCERTAINTY FEEDWATER FLOW -

a,c VENTURI THERMAL EXPANSION COEFFICIENT TEMPERATURE MATERIAL DENSITY TEMPERATURE PRESSURE DELTA P FEEDWATER ENTHALPY TEMPERATURE PRESSURE STEAM ENTHALPY PRESSURE MOISTURE NET PUMP HEAT ADDITION HOT LEG ENTHALPY TEMPERATURE STREAMING, RANDOM -

STREAMING, SYSTEMATIC PRESSURE COLD LEG ENTHALPY TEMPERATURE PRESSURE COLD LEG SPECIFIC VOLUME TEMPERATURE PRESSURE ATD CROSS-CAL SYSTEMATIC ALLOWANCE BIAS VALUES l FEEDWATER PRESSURE DENSITY 1 ENTHALPY l STEAM PRESSURE ENTHALPY l PRESSURIZER PRESSURE ENTHALPY - HOT LEG ENTHALPY - COLD LEG SPECIFIC VOLUME - COLD LEG FLOW BIAS TOTAL VALUE __

• , ** , +, ++ INDICATE SETS OF DEPENDENT PARAMETERS a,c l SINGLE LOOP UNCERTAINTY (WITHDUT BIAS VALUES) l N LOOP UNCERTAINTY (WITHOUT BIAS VALUES)

N LOOP UNCERTAINTY (WITH BIAS VALUES) ._

I 97870:1D/112186 18

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

% d/p SPAN  % FLOW  ;

PMA = -

a,c l PEA =

SCA = 1 SPE =

STE =

SD =

RCA =

M&TE =

RTE =

RD =

ID =

A/D =

RDOT =

BIAS =

FLOW CALORIM. BIAS =

FLOW CALORIMETRIC =

INSTRUMENT SPAN =

~ ~

I SINGLE LOOP ELBOW TAP FLOW UNC =  % FLOW a,c N LOOP ELBOW TAP FLOW UNC =

l N LOOP RCS FLOW UNCERTAINTY

=

(WITHOUT BIAS VALUES)

N LOOP RCS FLOW UNCERTAINTY (WITH BIAS VALUES) = 1.8 j 97870.1D/112186 19

4.0 NON-LOCA SAFETY EVALUATION 4.1 RESPONSE TIME The primary impact of the RTD bypass elimination on the FSAR Chapte'r 15 non-LOCA safety analyses (Reference 2) is the increased response time associated with the fast response thernowell RTD system. The secondary impact is the possible increase in instrument uncertainties. Currently, the overall response time of the Byron /Braidwood RTD bypass system assumed in the safety analyses is approximately 6.0 seconds (see Table 2.1-1). For the fast response thermowell RTD system, the overall response time will be approximately 7.0 seconds as described in Section 2.1 and as given in Table 2.1-1.

This increased RTD response time results in longer delays from the time when the fluid conditions in the RCS require an Overtemperature Delta-T or Overpower Delta-T reactor trip until a trip signal is actually generated.

Therefore, those transients that rely on the above mentioned trips must be evaluated for the longer response time. The affected transients include the Uncontrolled RCCA Withdrawal at Power, the Loss of Load / Turbine Trip, and the Steamline Rupture at Power events and are discussed in Section 4.3.

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1. ~

4.2 RTD UNCERTAINTY The proposed fast response thermowell RTD system will make use of RTDs, manufactured by the RdF Corporation, with a total uncertainty of [ ]a,c assumed for the analyses.

The FSAR analyses make explicit allowances for instrumentation errors for some of'the reactor protection system setpoints. In ;ddit!on, allowances are made for the initial average reactor coolant system (RCS) tercerature, pressure and power as described in FSAR Section 15.0. These allowances are made explicitly to the initial conditions for the non-DNB events; for the DNB events, these allowances are statistically combined into the design limit DNBR value, consistent with the Improved Thermal Design Procedure (Reference 3).

97870:1o/1121ss 20

l l

l The following protection and control system parameters were affected by the i change from one hot leg RTD to three hot leg RTDs; the Overtemperature Delta-T (OTDT), Overpower Delta-T (OPDT), and Low RCS Flow reactor trip functions, RCS average temperature measurements used for control board indication-and input to the rod control system, 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 number of hot leg RTDs. The results of these calculations indicate sufficient margin exists to account for all known instrument uncertainties.

The results of the system uncertainty calculations verify that sufficient allowance has been made in the reactor protection system setpoints to account for the increased RTD error. Therefore, the current values of the nominal setpoints 'noted above as defined by the Byron Technical Specifications remain valid.

. 4 :3 NON-LOCA TRANSIENTS REANALYZED All the events reanalyzed in this section used the LOFTRAN computer code.

LOFTRAN (Reference 4) is a digital computer code, developed to simulate transient behavior in a multi-loop pressurized water reactor system. The

_- program simulates the neutron kinetics, thermal-hydraulic conditions, .

pressurizer, stean generators, reactor coolant pumps, and control and p'rotection system operation. The secondary side of each steam generator utilizes a homogeneous saturated mixture for the thermal transients.

4.3.1 Uncontrolled RCCA Bank Withdrawal at Power The Uncontrolled RCCA bank withdrawal at power event is described in Section 15.4.2 of the FSAR. An uncontrolled RCCA bank withdrawal at power causes a positive reactivity insertion which results in an increase in the core heat l flux. Since the steam generator lags behind the core power generation, there i is a net increase in the reactor coolant temperature. Unless terminated by manual or automatic action, the increase in coolant temperature end power could result in DNB. For this event, the Power Range High Neutron Flux and e7s7o;1omossa 21

Overtemperature Delta-T reactor trips are assumed to provide protection against DNB. Therefore, this event was reanalyzed with increased time j constants to show that the DNBR limit is met.

Methods The assumptions used are consistent with the FSAR for the ITDP methodology in that initial power, pressure, and RCS average temperature are assumed to be at the nominal values corresponding to 10%, 60%, and 100% power. Both minimum and maximum reactivity feedback cases were reanalyzed with the increased time response value. The analysis was done using the LOFTRAN Computer Code.

Results For both minimum and maximum reactivity insertions, at the various power levels analyzed, the DNBR limit is met. A calculated sequence of events for a

, fast and s. low insertion rate for each reactivity feedback assumption is presented on Table 4.3-1 for full power. Figures 4.3-1 through 4.3-8 show results for a fast insertion case and a slow insertion case corresponding to i the 100% power case and both reactivity assumptions. The plots of minimum DNBR versus reactivity insertion rate at all three power levels are shown as

._~ Figures 4.3-9 through 4.3-11. -

~

Conclusions t

The limit DNBR is met, and therefore, the conclusions presented in the FSAR remain valid.

4.3.2 Loss of Load / Turbine Trip The Byron /Braidwood FSAR only explicitly analyzes the Turbine Trip Event which is presented in Section 15.2.3. This event relies on any of three reactor trips for primary protection: High Pressurizer Pressure, Low Low Steam Generator Water Level, and Overtemperature Delta-T. Thus, the increase in RTD response time may have an effect on the results of this transient.

O e7s7onomossa 22

4 Nothods The assumptions used are consistent with the FSAR for the ITDP methodology in that initial power,' pressure, and RCS average temperature are assumed to be at the nominal values corresponding to 100% power. All four cases presented in the FSAR were reanalyzed incorporating the assumptions of the RTD Bypass Elimination. These are Beginning of Life and End of Life, with and without pressure control (pressurizer spray and PORVs). The analysis was done using the LOFTRAN Computer Code.

Results For all combinations of reactivity feedback and pressure control, the DNBR limit is met. The results of these four cases are presented as Figures 4.3-12 through 4.3-19. A calculated sequence of events is shown in Table 4.3-2.

Figures 4.3-12 and 4.3-13 show the responses for a turbine trip event with

minimum reactivity feedback (Beginning of Life) assuming operability of l pressurizer sprays and PORV.'s. The reactor is tripped by the High Pressurizer j Pressure trip function. The DNBR increases throughout the transient and never drops below the initial value. The pressurizer safety valves are actuated and primary system pressure remains below the 110% design value.

Figures 4.3-14 and 4.3-15 show the responses for a turbine trip with maximum reactivity feedback (End of Life) and pressure control. The reactor is tripped by the Overtemperature Delta-T trip function, and the DNBR never drops below the initial value. The pressurizer safety valve lift set pressure is not reached.

Figures 4.3-16 and 4.3-17 show the responses for a turbine trip with minimum reactivity feedback (BOL) and without pressure control. The reactor is tripped by the High Pressurizer Pressure trip function, and the DNBR never drops below the initial value. The pressurizer safety valves are actuated and maintain system pressure below 110% of the design value.

l l

l 87s7tL1on tossa 23

Figures 4.3-18 and 4.3-19 show the responses for a turbine trip with maximum reactivity feedback (End of Life) and without pressure control. The reactor is tripped by the High Pressurizer Pressure trip function, and the DNBR never drops below the initial value. The pressurizer safety valves are aictuated and maintain system pressure below 110% of the design value.

Conclusions The DNBR limit value is met in all four cases, and therefore, the conclusions i

presented in the FSAR remain valid. The Overpressure Protection Report is also not impacted by the RTD bypass elimination effort, and thus, the conclusions presented in that document remain unchanged.

4.3.3 Steamline Rupture at Power The Steamline Rupture at Power transient was analyzed consistent with WCAP-9226-R1. The analysis included the increased time constants and the i increased temperature uncertainties mentioned in Section 4.2. For this event the design. basis as described in WCAP-9226-R1 was met.

i 4.3.4 Conclusion The impact of the RTD bypass elimination for Byron and Braidwood Units 1 and 2 o'n the FSAR Chapter 15 non-LOCA accident analyses has been evaluated. For the events impacted, it was demonstrated that the conclusions presented in the FSAR remain valid.

4.4 LOCA Evaluation t

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

! the steam generator performance data used in the LOCA analyses wil4 not be affected. Past sensitivity studies have shown that the variation of the core

inlet temperature (Tin) used in the LOCA anlayses affects the predicted core flow during the blowdown period of the transient. The amount of flow into the ers m io m ossa 24

[. .. I 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 temerature effect on peak clad temeprature is dependent on break size. As a result of these studies, the LOCA analyses are performed at a nominal value of Tin withoutconsiderationofsmall[ l uncertainties. The RCS flow rate and steam generator secondary side temperature and pressure are also determined using the loop average 1

temperature (T,yg) output. These nominal 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 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 i

The RTD Bypass Elimination modification for Byron /Braidwood 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, III, and IV as follows:

1. The Narrow Range (NR) cold leg RTD in the cold leg manifold will be

~

replaced with a fast response NR RTD well mounted in the RCP pump

. discharge pipe. The signal from this fast response NR RTD will perform the same function as the existing RTD Teold signal. A second narrow range RTD will be installed as a spare.

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

hot circuit obtained from these 3 NR T hot function as the existing RTD T hot signal.

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

97870:1D/111386 25 -

~

4. Signal process and the added circuitry to the Protection Set racks will be accomplished by additions to the process control (Westinghouse Model 7300) racks using 7300 technology. When one T hot signal is removed from the averaging process, the electronics will allow a bias to be manually added to a 2-RTD average Thot (as pp sed to a 3-RTO average Thot) in order to obtain a value comparable with the 3-RTD average Thot pri r to the failed RTD. In the event of a cold leg RTD failure, the spare cold leg RTD will be manually connected to the 7300 circuitry in place of the failed RTD.

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

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

_- Followingisadiscussionofldesignbasisrequirementsinconformanceto pertinent I&C criteria:

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

b. Quality components and modules being added is consistent with use in a Nuclear Generating' Station Protection System. For the Westinghouse Quality Assurance program, refer to Chapter 17 of the FSAR.

l i

j c. Equipment seismic and environmental qualification will be to IEEE l standards 344-1975 and 323-1974, respectively, as described in.WCAP 8587, Rev. 5 " Methodology for Qualifying Westinghouse WRD Supplied NSSS Safety Related Electrical Equipment".

l l

97870;1o/110586 26 1

d. The changes will continue to maintain the capability of the protection system to initiate a reactor trip during and following natural phenomena credible to the plant site to the same extent as the existing system,

e. Channel independence and electrical separation is maintained because the Protection Set circuit assignments continue to be Loop 1 circuits input to Protection Set I; Loop 2, to Protection Set II; Loop 3, to Protection Set III; and Loop 4 to Protection Set IV, with~ appropriate observance of field wiring interface criteria to assure the independence. Output circuits are the same as before except that there will be one T cold and 3 T hot outputs to the computer sent through Class IE isolators in each Protection Set.

f. The IEEE 279-1971 Section 4.7 and GDC 24 requirements concerning Control and Protection System interaction are satisfied because, even though control signals are derived from Protection Sets, the 2/4 voting

~

coincidence logic of the Protection Sets is maintained.

Where a single random failure can cause a control system action that results in a condition requiring protective action and can also prevent proper action of a protection system channel designed to protect against the condition, the j-

~

remaining three redundant protection channels will be capable of providing the

! protective action even when degraded by a second random failure. This is i because even though 1/4 channels failed without partially tripping, only 2 of the remaining 3 channels are necessary for a plant trip.

On the basis of the foregoing evaluation, it is concluded that these I&C modifications required for RTO bypass removal for the Byron /Braidwood units will meet IEEE 279-1971, applicable GDCs, and industry standards and regulatory guides. _

l 4

6 9757Q:1D/110586 27

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 crossover leg bypass return nozzle, the cold leg piping 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.

Byron 1 and Braidwood 1 The original three scoops in each hot leg, which feed the bypass manifold, and the bypass manifold connection must be removed and the scoops modified to accept three fast response RTD thermowells. [.

Ja.c to provide the proper flow path. A thermowell design will be used such that the tip of the thermowell will be positioned to provide an average temperature reading ~ . The thermowell will be

, fabricated in accordance with Section III Class 1 of the ASME Ccde. The j installation of the thermcwell into the scoop will be performed using GTAW for the root pass and finished out with either GTAW or 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.

l The cold leg RTD bypass line must also be removed. The nozzle must then be modified to accept the fast response RTD thermowell. Additionally, a spare fast response thermowell will be added to the cold leg in the length between the reactor coolant pump discharge and the loop isolation valve. This necessitates the creation of a new penetration into the piping. The boss for

the new connection will be root welded by GTAW. Finish welding can be either j GTAW or SMAW. Weld inspection by PT will be performed as required by Section l

XI. The thermowells will extend approximately [ la,c inches into the flow stream. This depth has been justified based on [ . Ja,e analysis.

I I

e7s7o:io m ossa 28

.- a w pe. - u--,,w - -

-- w w -, - .

The root wid joining the thermowells to the modified nozzles or bosses will be deposited with GTAW and the remainder of the weld may be deposited with GTAW or SMAW. Pe.netrant testing will be performed in accordance with the ASME Code Section XI. The thermowells and installation boss.es will be fabricated in accordance with the ASME Section III (Class 1). These two thermowells will be installed in the upper half of the piping.

Byron 2 and Braidwood 2 Threo hot leg fast response RTDs will be installed into new penetrations between the reactor vessel and the loop isolation valves. This is done in order to facilitate initially operating the units with the bypass system before implementing the fast response RTDs. The design will be such that the thermowell will extend [ Ja,c inches into the flow stream from the ID of the pipe. The. installation boss for the new penetrations and the thermowell will be root welded by GTAW. Finish welding can be either GTAW or SMAW. Weld inspection by PT will be performed per Section XI. The installation bosses and thermowells are fabricated in accordance with Section III Class 1 of the ASME Code.

~

In the cold leg between the reactor coolant pump discharge and the loop isolation valves two fast response RTDs will be installed into new penetrations. The design will be such that the-tip of the thermowells will extend approximately [ ]a,c inches into the flow stream.

l The installation bosses and thermowells will be fabricated in accordance with I i

Section III Class 1 of the ASME Code. Root welding will be performed by GTAW. Finish welding can be either GTAW or SMAW. Weld inspection by PT will

, be performed per the ASME Code Section XI.

i Upon removal of the RTD bypass piping, the hot leg scoops and the cold leg

( nozzles will be capped. The caps will be fabricated in accordance with Section III Class 1 of the ASME Code. The root weld joining the cups to either the scoops or cold leg nozzle will be done by GTAW. Finish welding will be done by either GTAW or SMAW. The welds will be' inspected by PT per the ASME Code Section XI.

s7:7o:ioniosas 29

1. - ~. -_

.-y - - -

~

With the three thernowells in the hot leg and the two thernowells in the cold leg, a total of 20 thernowells will be utilized at each of the four-loop Byron /Braidwood units and they will perform the same function as the original bypass T hat and Teold signals. ,

The cross-over leg bypass return piping connection must be removed and the nczzles capped. The cap design, including materials, will meet the' pressure boundary criteria and ASME Section III (Class 1). The cap will be root velded to the nozzles by GTAW and fill welded by either GTAW or SNAW. j Non-destructive examinations (PT and radiographs) will be performed per ASME Section XI. Nachining 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 [ la,c of the ASME Code, a hydrostatic test of new pressure boundary welds is required when the connection to the pressure boundary is [ la.c 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 required after bypass elimination. Paragraph [

Ja.c defines this test pressure to be [ ]a,c times the normal

{ operating pressure at a temperature of [ ).a.c The integrity of the reactor coolant piping as a pressure boundary component, is maintained by adhering to the applicable ASME Code sections and Nuclear Ryulatory Comission General Design Criteria. 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 calculations sununarized in Section 3.1 on the impact of the

! fast response RTDs on flow measurement uncertainty, the Technical Specification modifications identified in Appendix A are necessary to achieve proper reactor trip and engineered safety features system operability.

l s7s m w iiri ,s 30

TABLE 4.3-1 (page 1 of 2)

• TIME SEQUENCE OF EVENTS FOR A RCCA_ BANK WITHDRAWAL AT POWER ACCIDENT EVENT TIME (SECS)

Case A Initiation of uncontrolled RCCA 0.0 withdrawal at a fast reactivity insertien rate (75 pcm/sec) with minimum reactivity feedback at full power Power range high neutron flux 1.6 reactor trip signal generated Rods begin to drop 2.1 l

Minimum DNBR occurs 3.2 Peak water level in the 4.9 pressurizer occurs Case B Initiation of uncontrolled RCCA 0.0 withdrawal at a low reactivity insertion rate (3 pcm/sec) with minimum reactivity feedback at full power Overtemperature Delta-T reactor 31.5

, trip signal initiated Rods begin to drop 33.5 Minimum DNBR occurs 33.7

, Peak water level in the 35.7 pressurizer occurs 31 97870:1D/110586

TABLE 4.3-1 (page 2 of 2)

TIME SEQUENCE OF EVENTS FOR A -

RCCA BANK WITHDRAWAL AT POWER ACCIDENT EVENT TIME (SECS)

Case C Initiation of uncontrolled RCCA 0.0 withdrawal at a fast reactivity insertion rate (75 pcm/sec) with maximum reactivity feedback at full power Power range high neutron flux 5.3

, reactor trip signal initiated Rods begin to drop 5.8 Minimum DNBR occurs 6.4 Peak water level in the 8.6 pressurizer occurs Case D Initiation of uncontrolled RCCA 0.0 withdrawal at a low reactivity insertion rate (3 pcm/sec) with maximum reactivity feedback at full power Overtemperature Delta-T reactor 203 i trip signal initiated l Rods begin to drop 205 Minimum DNBR occurs 203 Peak water level in the 207 pressurizer occurs a

97870i1D/1105a6

TABLE 4.3-2 (page 1 of 2)

TIME SE0VENCE OF EVENTS FOR A -

TURBINE TRIP ACCIDENT EVENT TIME (SECS)

Case A Initiation of turbine trip, 0.0 loss of main feedwater flow, minimum reactivity feedback with pressure control High pressurizer pressure 6.9 reactor trip signal generated Initiation of steam release 8.5 from S/G safety valves Rods begin to drop 8.9 Peak pressurizer pressure occurs 10.5 Minimum DNBR occurs (1)

Case B Initiation of turbine trip, 0.0

~

loss of main feedwater flow. '

- maximum reactivity feedback with pressure control Initiation of steam re19ase 8.1 from S/G safety valves Overtemperature Delta-T ' 8.5 reactor trip signal generated Peak pressurizer pressure occurs 9.4 Rods begin to drop 10.5 Minimum DNBR occurs (1)

~

(1) DNBR does not decrease below its initial value.

l 1

l 33 e7s70:to/11 ossa  ;

TABLE 4.3-2 (page 2 of 2)

TIME SEQUENCE OF EVENTS FOR A . .

TURBINE TRIP ACCIDENT EVENT TIME (SECS)

Case C Initiation of turbine trip, 0.0 loss of main feedwater flow, minimum reactivity feedback without pressure control High pressurizer pressure 4.1 reactor trip signal generated Rods begin to drop 6.1 Initiation of steam release 8.0 from S/G safety valves Peak pressurizer pressure occurs 8.0 Minimum DNBR occurs (1)

Case D Initiation of turbine trip, 0.0 loss of main feedwater flow,

_~ maximum reactivity feedback -

without pressure control  !

High pressurizer pressure 4.1 reactor trip signal generated i Rods begin to drop 6.1 l Peak pressurizer pressure occurs 7.5 l Initiation of steam release 8.0 from S/G safety valves Minimum DNBR occurs (1)

(1) DNBR does not decrease below its initial value. -

l 97870.1D/110586 1

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reco.e < -

E asoc.e < <

G troo.e - - -

5. eseo.e - - -

E

. acco.o < -

M j isoo.e - - -

E seco.e < - < -

im.. . - -

ison.o -

^ ^ ^

'g@l . .

E ison.s - - < -

C I

, usa... .

W 5 8000.00 < -

E l

[ pse.oo . - --

E soc.co . --

eso.co "

0 25 50 75 100 125 150 175 200 FIGURE 4.3-14 TURBINE TRIP EVENT TIME (Seconds) WITH PRESSURE CONTROL NAXIMUM REACTIVITY FEEDBACK 48

. W W W W W see.es < -

e ties.es - < -

u I sto.se < <.

sas.so < <

,E 33.. . . .

h seo.co - -

sas.so < -

m.= .

seo.co .  : . . . .  :

c w

g sao.so - -

I

. = son.oo - - -

I w

seo.go < < *

.V sa.oo -

W

=

see.se -

^

s.coes  : .  : .  : ^

e.cono . -

g 3.0ooo < < -

3 I . ooo . .

3.sooo . -

e.e - '

O 25 50 75 100 125 150 175 200 TIME (Seconds) FIGURE 4.3-15 TURBINE TR!P EVENT 49 WITH PRESSURE CONTROL MAXIMUM REACTIVITY FEEDBACK

b e . . * . .

1.0000 < % < -

i

=. . .

W

• E

. son. - - - -

1. .

k e.e . .

n00.8 . . . .

^

2500.0 -

E rico.s - - -

5. cm.e -

r :co. < -

E naa.e- - -

l seo.e < < .

r ,, . . . .

irm. . .

1800.8 . ,

13.8. .

a E

soc. - < -

C l irso.e - < .

.i 1000.00 <- *-

e 1

g rs . . .

r

^

400.00 ^

0 25 50 75 100 125 150 175 200 FIGURE 4.3-16 TIME (SECONDS) TURBINE TRIP EVENT WITHOUT PRESSURE CONTROL 50 MINIMUM REACTIVITY FEEDBACK

l seo.es  : . . .

seo.es - -

c seo.m - -

u -

I ses.so . -

! sm.m . < .

sno.co . .

! seo.oo - -

sm.oo . . -

seo.es -

sno.m . . . . .

c o

y sao.es < -

! Y-a seo.co - - -

I .

w seo.ee - - -

tr i sac.co - -

t

=

seo.co .

s. esse .  : .  :  : . .

e.eoco - - -

g 3.eoco - -

I.. -)a .

l l

l s.e000 - -

e.s - -

O. 25 50 75 100 125 150 175 200 FIGURE 4.3-17 TIME (Seconds) TURBINE TRIP EVENT WITHOUT PRESSURE CONTROL 51 MINIMUM REACTIVITY FEEDBACK

S 3.seos  : . .  :  :  : -

l. .so . *

• E m ..o.o. -

\

y .

.1 .,,oco .

s ..co . - -

I .n .o . .

..e  : .

asco.e .  : . . . .

eson.e . -

g reso.s . -

N eson.e . f - -

l stoo.e . - -

E em.. . .

i soon.e - -

l son.e - -

r .. . . . .

i=. e . -

soon.e ^

lg,[ . . - - - - - - -

E stoo.s . < -

C l irm.s . -

3 seco.es . - -

G

[ m.m .

r no.. . .

0.0 25 50 75 100 125 150 175 200 FIGURE 4.3-18 TIME (Seconds) TUR8INE TRIP EVENT WITHOUT PRESSURE CONTROL 52 MAXIMUN REACTIVITY FEEDBACK

5 seo.es . . .  :  : . .

seo.se < -

e ses.se < -

i I ..se -

y 1

2 ses.se < . .

f sso.co . -

k seo.es . .

sas.es < . .

seo.so -

seo.se . . .  :  : . .

c w

g seo.as - -

I.

• eso.es . -

I w -

seo.co - -

.V sa.se . .

W

=

seo.co .

s. sees . .  :  :  : .  :

e.como . .

p 3.sooo . .

I u . sono . -

l 1.o000 ..

e.e - -

0.0 25 50 75 100 125 150 175 200 l FIGURE 4.3-19 i TIME (Seconds) TURBINE TRIP EVENT WITHOUT' PRESSURE CONTROL 1 53 MAXIMUM REACTIVITY FEEDBACK '

l

5.0 CONTROL SYSTEM EVALUATION A prime input signal to the various NSSS control systems is the RCS average temperature (T,yg). This is calculated electronically as the average of the measured hot leg and cold leg temperatures in each loop. -

The major control systems affected are [

Ja,c The effect of the new RTD is to potentially change the time response of the T,y However, as noted in Section 2.1. Table  !. channels 1-1, the new RTD system in the various will have a timeloops.

response close to that of the present system. There will therefore be no significant effect on the T,yg channel response, and no apparent need to revise any of the control system setpoints from those presently installed in the plant. The need to modify control system setpoints will be determined during the plant startup following the installation of the new RTD system by observing control system behavior. In addition, the Cold Overpressurization. Mitigation System i (CONS) will be unaffected by the RTD bypass elimination since the CONS utilizes the wide range RTDs which are unaffected by this program.

l l _- -

W 97870;1o/111386 .

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

6.0 CONCLUSION

S The method of util.izing fast-response RTDs installed in the reactor coolant loop piping as a me'ans for RCS temperature indication has undergone extensive analyses, evaluation and testing as described in this report. The incorporation of this system into the Byron /Braidwood 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 performance. The fast response RTDs installed in the reactor coolant loop piping adequately replace the present hot and cold leg temperature measurement system and enhances ALARA efforts and improved plant reliability.

l l

l l

I e7s70:10/11osas

7.0 REFERENCES

1. Tuley, C. R., Moomau, W. H., "RCS Flow Uncertainty for Shearon Harris Unit 1", WCAP-11168 Rev. 1 Proprietary, WCAP-11169 Rev. 1

~

Non-Proprietary, October, 1986. -

2. Byron /Braidwood Final Safety Analysis Report, Amendment #47, April 1986.

3. Chelemer, H., et al., " Improved Thermal Design Precedure," WCAP-8567-P (Proprietary, WCAP-8568 (Non-Proprietary), July 1975.

4. Burnett, T.W.T., e't al., "LOFTRAN Code Description," WCAP-7907-P-A (Proprietary), WCAP-7907-A (Non-Proprietary), April 1984.

i D

e l

l l

~

I 97870:1o/1113ss 56 ,

l

_- - - - - - - - . - - - . - - - - - - - . - - . . . - - -- . - - - - - - -- - l

s . - -

6 APPENDIX A TECHNICAL SPECIFICATION NODIFICATIONS O

m W I

=

a

f

'y a '

g9M _T_ABLE 2.2-1 ia g REACTOR TRIP SYSTEN INSTRUNENTATION TRIP SETPOINTS y

iE$ SENSOR

)3 TOTAL j_. h FUNCTIONAL UNIT ERROR ALLOWANCE (TA) Z (SE) TRIP SETPOINT

'* ALLOWAOLE VALUE

1. Manual Reactor Trip N.A. N.A. N.A. N.A. N.A.

M 2. Power Range, Neutron Flux

a. High Setpoint 7.5 4.56 0 $109E of RTP* $111.1% of RTP* ,

b. Low Setpoint 8.3 4.56 0 $25% of RTP* $27.1% of RTP*

3. Power Range, Neutron Flux, 1. 6 0.5 0

High Positive Rate $5% of RTP* wtth $6.3E of RTP* with

,y a time constant a time constant

!* 12 seconds 12 seconds 4 Power Range, Neutron Flux, 1.6 0.5 0 $55 of RTP* wIth <6.35 of RTP* vfth i High Negative Rate

) a time constant a time constant 12 seconds 12 seconds j 5. Intermediate Range, 17.0 8.4 0

Neutron Flux $25% of RTP* $30.9E of RTP*

i

6. Source Range, Neutron F1un 17.0 10.0 0 $108 cps $1.4 x 10s ep,

! 7. Overtemperature AT T.7 M

! F 86 .5M See 5 e Note 1 See Note 2 Note 5 '

i

8. Overpose,r AT 4.8 .4 # /.22 J # 1.2 See Note 3 See Note 4 , -

So Pressurizer Pressure-Low 5.0 2.21 1.5 11885 psig 11871 psfg

10. Pressurizer Pressure-High 3.1 0.71 1. 5 $2385 psig $2396 psig l 11. Pressurizer Water level-High 5.0 2.18 1.5 $92% of instrument

$93.8E of instrument span span "RTP = RATED THERMAL POWER l

f masy E0 TABLE 2.2-1 (Continued)

GUM biE REACTOR TRIP SYSTEN INSTRUMENTATION TRIP SETPOINTS i

SENSOR g g,g gc. TOTAL ERROR w FUNCTIONAL UNIT ALLOWANCE (TA) Z (SE) TRIP SETPOINT ALLOWABLE VALUE e -

25 22- " ct r caaat -'a" 25 /.4 2. p 0.6 > ,0x or io , .ini-

, or ioo, .ini-mun measured flow" mum measured flowa

13. Steam Generator Water Level Low-Low '

a. Unit 1 27.1 18.28 1.5 >40.8x of narrow $39.1% of narrow Vange instrument range instrument

span span -

b. Unit 2 17.0 14.78 1.5 >17% of narrow >15.3% of narrow i range instrument range instrument SP span span

, =

14. Undervoltage - Reactor 12.0 0.7 0 >5268 volts - 14728 volts -

Coolant Pumps each bus each bus

15. Underfrequency - Reactor 14.4 13.3 0 157.0 Hz >56.5 Hz Coolant Pumps i

16. Turbine Trip  ;

a. Emergency Trip Header N.A. M.A. M.A. >S40 psig. >520 psig

• Pressure

b. Tu,rbine Throttle Valve M.A. M.A. N.A. 11% open 11% open ,,

Closure

17. Safety Injection Input N.A. N.A. N.A. M.A. M.A.

i from ESF -

18. Reactor Coolant Pump N~. A. . N.A. N.A. N.A. N.A.

Breaker Position Trip

• Minimum measured flow = 97,600 gpa

l

< [. .' ;*

' e ::,, . 3 -

[.. -i ggM TABLE 2.2-1 (Continued) ,

TABLE NOTATIONS (Continued) 6 NOTE 1: (Continued) 3'5 re =

Time constant utt11 red in the measured T,,,las compensator, re=0s,

, T' <

588.4*F (Nominal T,,,at RATED THERMAL POWER),

, K3 = 0.00134, l

l P = Pressurfrer pressure, psig -

l P' =

2235 psig (Nominal RCS operating pressure),

J

! S = Laplace transform operator, s 8,

! 7

• • and f (al) is a function of the indicated difference between top and bottom detectors of the Power-range neutron ton chambers; with gains to be selected based on measured instrument ,

response during plant STARTUP tests such that: ,  !

j (1) for q g g between -=K and +10X, f (AI) = 0, where g g W g a m p m ent '

j RATED THERMAL POWER fn the top and bottom halves of the core respectively, and gg+qb I' '

) total THERMAL POWER in percent of RATED THERMAL POWER; (11) for each percent that the magnitude of q g g exceeds +105, the AT Trip Setpoint i shall be automatically reduced by 2.0% of 1ts value at RATED THERMAL POWER.

i NOTE 2: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more theti Je9E of AT span.

3.2 */o 1

l 1

• 1 i

I I

i .

' f TABLE 2e2-1 (Continued) ,,' *

$$h san e >

TABLE NOTATIONS (Continued) 8'" NOTE 3: (Continued)

• E Q K. =

0.00170/'F for T > T" and K. = 0 for T i T",

3' T = As defined in Note 1, 5 i T" =

Indicated T, 5, at RATED THEllMAL POWER (Calibration temperature for AT instro.entation, 3 so.. r),

5 = As deffned in Ncte 1, and

• f 2(al) = 0 for all AI. F i NOTE 4:

The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than '

7 2.7 M of AT span. l g I. 7 l.1 I

NOTE 5: The sensor error for temperature is J4 and for pressure is 3 #. ,

9 I

4

r

(.,

. u, =5.:.1 1

- r taste 3.3-4 (Continued)

NEh .

i 8'" ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTMSIENTAT "E ,!

i *4 TOTAL FUNCTIONAL UNIT tu SENSOR TRIP

.E" ALLOWANCE (TA) Z ERROR (SE) SETPOINT ALLOIMSLE '

VALUE

! Q ~~ 8. Less ef Power I i~~ a. ESF Sus Underveltage N.A. a ig" M.A. M.A. 2870 welts

b. >2730 volts

" Grid Degraded w/1.8s delay w/il.9s delay Voltage M.A.

i N.A. M.A. 3004 velts 13728 volts

. w/310s delay w/310 1 30s delay

9. Engineered Safety i Feature Actuation -

!g* System Interlocks 1

e;e a.

g Pressurizer Pressure.

P-11 N.A. N. A. N.A. $1930 psIg 11936 psfg [

b. Reactor Trip, P-4 N.A.

i i N.A. N.A. N.A.

N.A. i c.

Low-Low T ,,, F-12 N.A. N.A. N.A. E47.0 I 1550*F >3454*F '

i i

} d. Steam Generator Water t a

Level, P-14 See Item 5.b. above for all Steam Generator Water Level Trip i i (Hfgh-Nigh) Setpoints and Allowable Values. '

i  !

i l

! t ..  !

l i l

I o

POWER DISTRIBUTION LIMITS SASES HEAT FLUX HOT CHANNEL FACTOR. and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE NOT CHANNEL FACTOR (Continued)

c. The control rod insertion limits of Specification 3.1.3.6 are maintained, and

d. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.

Fh will be maintained within its limits provided the Conditions a. through

d. above are maintained.y The combination of the RCS flow requirement (390,400 gpm) and the requirement on Fg guarantee that the DNBR used in the safety analysis will be met.

A rod following bow penalty is not applied to the final value of Fh for the reason: ,

Fuel rod bowing does reduce the value of the DNBR. However, predictions with the methods described in WCAP-8691, Revision 1 " Fuel Rod Bow Evaluation," July )

1979 for the 17x17 Optimized Fuel Assemblies indicate that the fuel rod bow

.' reduction on DNBR'will be less than 3% at 33,000 MWD /MTU assembly average burnup.

At higher burnups, the decrease in fissionable isotopes and the buildup of fission product inventory more than compensate for the rod bow reduction in DNBR.

There is a 11% margin available between the 1.32 and 1.34 design DNBR limits and the 1.47 and 1.49 safety analysis DNBR limit. Use of the 3% fuel rod bow i DNBR margin reduction still leaves a 8% margin in DNBR between design limits and safety analysis limits. ,

~

The RCS flow requirement is based on the loop minimum measured flow rate of 97,600 gpm which is used in the Improved Thermal Design Procedure described l 4n FSAR 4.4.1 and 15.0.3. A precision heat balance is performed once each cycle and is used to calibrate the RCS flow rate indicators. Potential fouling of the feedwater venturi, which might not be detected, could bias the results from the precision heat balance in a non' conservative manner. Therefore, a penalty of 0.1% is assessed for potential feedwater venturi fouling. A maximum measurement I.9 % 1 uncertainty ofM has been included in the loop minimum measured flow rate to account for potential undetected feedwater venturi fouling and the use of the RCS flow indicators for flow rate verification. Any fouling which might bias the RCS flow rate measurement greater than 0.1% can be detected by monitoring and trending various plant performance parameters. If detected, action shall be taken, before performing subsequent precision heat balance measurements, i.e.,

either the effect of fouling shall be quantified and compensated for in the RCS flow rate measurement, or the venturi shall be cleaned to eliminate the fouling. }

Surveillance Requirement.4.2.3.4 provides adequate monit'oring to detect l possible flow reductions due to any rapid core crud buildup. -

l Surveillance Requirement 4.2.3.5 specifies that the measurement instrumen-  !

tation shall be calibrated within seven days prior to the perfomance of the  ;

calorimetric flow measurement. This requirement is due to the fact that the' i drift effects of this instrumentation are not included in the flow measurement l uncertainty analysis. This requirement does not apply for the instrumentation l whose drift effects have been included in the uncertainty analysis, i TYPICAL FYRCN - UNITS 1 AND 2

• • • 8 3/4 2-4

_ -_ ?_Y "? L A** * - - - - - - - -

l