ML20238F033
| ML20238F033 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 06/30/1987 |
| From: | Rice W, Sterdis R WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML19292H758 | List: |
| References | |
| WCAP-11497, NUDOCS 8709150349 | |
| Download: ML20238F033 (75) | |
Text
.____ _____-___ _ - _- _ _ - _ -
/
[' L,, WE3TINGHOUSE CLASS 3 L WCAP-11497 M ..
v.
l l
p, RTD BYPASS ELIMINATION LICENSING REPORT o.- FOR NILLSTONE UNIT 3 1
W. R. RICE l l
!r R. J. STERDIS iw' JUNE, 1987 si '
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'* Westinghouse Electric brporation Pittsburgh, PA ar 8709150349 870909 r
PDR ADOCK 05000423' li p. PDR >
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f ACKNOWLEDGEMENT i
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-The authors wish to recognize contributions by the following individuals:
l P. Rosenthal i ..
G. Lang I g M. Weaver W. Moomau R. Carlson i
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-P. Huang 'l W. Lyman
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D. Altman l-I
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l 0523v:1D/082287 i
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TABLE OF CONTENTS Seetion-Paste List of Tables iii List of Figures iv 1.0 Introduction 1
1.1 Historical Background 1
1.2 Mechanical Modifications '2
.1.3 Electrical Modifications 4
h 2.0 Testing 10 2.1 Responce Time Test 10 2.2 Streaming Test y 10 3.0 Uncertainty Considerations 13 3.1 Calorimetric Flow Measurement Uncertain'v 13 3.2 Hot Leg Temperature Streaming Uncertainty 13 4.0 Safety' Evaluation 24 4.1 Response Time 24 4.2 RTO Uncertainty 24 4.3 Non-LOCA Transients Reanaly nd
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25 '
4.4 LOCA Evaluation 30 4.5 SGTR Evaluation 31 4.6 Instrumentation and Control Safety Evaluation 32 4.7 Mechanical Safety Evaluation 35 4.8 Technical Specification Evaluation 37
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1 I-oS23v:10/062287 i
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TABLE OF. CONTENTS-(Cont)
Section Page 5.0: Co'ntrol System Evaluation l 84 q
6.0 Conclusions-85 7.0 References' 1 B6 '
f
' Appendix A - Tech'nical Specification Modifications 87 Appendix B.- Hot Leg RTD Failure Compensation Procedure 0523v:1D/062287 '
ii J
. LIST 0F TABLES
' Table Tit 1e .Page
-2.1 -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 Co'id Leg Elbow Tap Flow Uncertainty 19 3.1-5 Flow Calorimetric Instrumentation Uncertainties' 20 (N-1 Loop Operation) 3.1-6' Flow Calorimetric Sensitivities (N-1 Loop Operation) 21 3.1-7 Calorimetric RCS Flow Measurement Uncertainties 22 (N-1 Loop Operation) 3.1-8 Cold Leg Elbow Tap Flow Uncertainty (N-1 Loop Operation) 23 {
. 4.3-1 Time Sequence of Events for a RCCA Bank Withdrawal 38 l at Power -
4.3-2 Time Sequence of Events for a Turbine Trip i
'40 4.3-3 Time Sequence of Events for a Inadvertent Opening 42 j of a Pressurizer Safety or Relief Valve i
j 1
' 0523v:f o/062287 iii l
L .-
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 3oss for Cold leg Fast-Response RTD 8 Installation i
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 43 Temperature for a RCCA Bank Withdrawal at Full Power with Minimum Reactivity Feedback (70 PCM/SEC Rate) 4.3-2 Pressurizer Pressure, Water Volume, and DNBR for a RCCA 44 Bank Withdrawal at Full Power with Minimum" Reactivity Feedback (70 PCM/SEC Rate) 4.3-3 Core Average Temperature, Heat Flux, and Nuclear Power 45 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 46 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 47 3
for a RCCA Bank Withdrawal at Full Power with Maximum j
Reactivity Feedback (70 PCM/SEC Rate) p l
f 0523v:1D/062287 iy I. _ _
a
I LIST OF FIGURES (Cont.)
-Figure. Ti tl'e Page 4.3-6 Pr'essurizer Pressure, Water Volume, and DNBR for a 48 RCCA Bank Withdrawal at Full Power with Maximum Reactivity Feedback (70 PCM/SEC Rate) 4.3-7 Nuclear Power, Heat Flux and Core Average Temperature 49 for a RCCA Bank Withdrawal at Full Power with Maximum Reactivity Feedback (3 PCM/SEC Rate)
-4.3-8 Pressurizer Pressure, Water Volume and DNBR for a RCCA- '50 Bank Withdrawal at Full Fower with Maximum Reactivity !
Feedback (3PCM/SECRate) 4.3-9 Minimum DNBR s.. Reactivity Insertion Rate for a RCCA 51 Bank Withdrawal at Full Power 4.3 Minimum DNBR vs. Reactivity Insertion Rate for a RCCA 52 Bank Withdrawal at 60% Power 4.3-11= Minimum DNBR vs. Reactivity Insertion Rate for a RCCA 53 Bank Withdrawal at 10% Power
~
4 4.3-12 Pressurizer Pressure, Water Volume and Nuclear Power 54
,for Turbine Trip With Pressure Control and Minimum Reactivity Feedback 4.3-13 Core Inlet Temperature, Coolant Average Temperature 55 and DNBR for Turbine Trip With Pressure Control and Minimum Reactivity Feedback D523v:10/062287 Y
LIST OF F1GURES (Cont.):
Figure Title P_ age .
4.3-14 Pressurizer Pressure, Water Volume and Nuclear Power 56 for Turbine Trip With Pressure Control and Maximum Reactivity Feedback 4.3-15' Core Inlet Temperature, Coolant' Average Temperature 57 and DNBR for Turbine Trip With Pressure Control and Maximum Reactivity Feedback 4.3-16 Pressurizer Pressure, Water Volume and Nuclear Power 58 for Turbine Trip Without Pressure Control and Minimum Reactivity Feedback 4.3-17 Core Inlet Temperature, Coolant Average Temperature 59 and DNBR for Turbine Trip Without Pressure Control and Minimum Reactivity Feedback 4.3-18 Pressurizer Pressure, Water Volume and Nuclear Power 60 for Turbine Trip Without Pressure Control and Maximum Roactivity Feedback 4.3-19
' Core Inlet Temperature, Coolant Average Temperature 61 and DNBR for Turbine Trip Without Pressure Control and Maximum Reactivity Feedback 4.3-20 Nuclear Power and Coolant Average Temperature for 62 Inadvertent Opening of a Pressurizer Safety or Relief Valve 4.3-21 Pressurizer Pressure and DNBR for Inadvertent Opening 63 of a Pressurizer Safety or Relief Valve l.
}
0513v;1o/062287 vi I
J
LIST OF FIGURES (Cont.)
Figure Title Page 4.3-22 Nuclear Power, Core Heat Flux, and Core Average 64 Temperature for a RCCA Bank Withdrawal at Full Power with Minimum Reactivity Feedback (70 PCM/SEC Rate)
(N-1 Loop Operation) 4.3-23 Pressurizer Pressure, Water Volume, and DNBR for a RCCA 65 Bank Withdrawal at Full Power with Minimum Reactivity Feedback (70 PCH/5EC Rate)
(N-1 Loop Operation) 4.3-24 Core Average Temperature, Heat Flux, and Nuclear Power 66 for a RCCA Bank Withdrawal at Full Power with Minimum Reactivity Feedback (3 PCM/SEC Rate)
(N-1 Loop Operation) 4.3-25 Pressurizer Pressure, Water Volume, and DNBR for a 67 RCCA Bank Withdrawal at Full Power with Minimum
~
Reactivity Feedback (3 PCM/SEC Rate)
(N-1LoopOperation) 4.3-26 Nuclear Power, Heat Flux and Core Average Temperature 68 for a RCCA Bank Withdrawal at Full Power with Maximum Reactivity Feedback (70 PCM/SEC Rate)
(N-1 Loop Operation) 4.3-27 Pressurizer Pressure, Water Volume, and DNBR for a 69 RCCA Bank Withdrawal at Full Power with Maximum Reactivity Feedback (70 PCH/SEC Rate)
(N-1 Loop Operation)
I 0523v:10/062287 Vii L__1__________ -
_ LIST OF FIGURES (Cont.)
i Figure Title Pa_ge a.
4.3-28 Nuclear Power, Heat Flux and Core Average Temperature 70 a.
for a RCCA Bank Withdrawal at Full Power with Maximum Reactivity Feedback (3 PCM/SEC Rate)
(N-1 Loop Operation)
, 4.3-29 Pressurizer Pressure, Water Volume and DNBR for a RCCA 71 Bank Withdrawal at Full Power with Maximum Reactivity o-Feedback (3 PCM/SEC Rate)
(N-1 Loop Operation) 4.3-30 Minimum DNBR vs. Reactivity Insertion Rate for a RCCA 72 Bank Withdrawal From 75% Power (N-1 Loop Operation) 4.3-31 Minimum DNBR vs Reactivity Insertion Rate for a RCCA 73 Bank Withdrawal From 10% Power (N-1 Loop Operation) 4.3-32 Pressurizer Pressure, Water Volume and Nuclear Power 74 for Turbine Trip With Pressure Control and Minimum
, Reactivity Feedback (N-1 Loop Operation) 4.3-33 Core Inlet Temperature, Coolant Average Temperature 75 e
and DNBR for Turbine Trip With Pressure Control and Minimum Reactivity Feedback (N-1 Loop Operation) 4.3-34 1 Pressurizer Pressure, Water Volume and Nuclear Power 76 for Turbine Trip With Pressure Control and Mar,imum Reactivity Feedback )
(N-1 Loop Operation) os23v:10/os2287 viii '
L
~g -
d '
LIST OF FIGURES (Cont.).
Figure Title Page 4.3-35 Core Inlet Temperature, Coolant. Average Temperature 77 and DNBR for Turbine Trip With Pressure Control and Maximum Reactivity feedback (N-1 Loop Operation) 4.3-36 Pressurizer Pressure, Water Volume and Nuclear Power 78 f for Turbine' Trip Without Pressure Control and Minimum Reactivity Feedback (N-1 Loop Operation) 4.3-37 Core Inlet Temperature, Coolant Average Temperature 79 and DNBR for Turbine Trip Without Pressure Control and Minimum Reactivity Feedback (N-1 Loop Operation) 4.3-38 Pressurizer Pressure, Water Voluma and Nuclear Power 80 for Turbine Trip Without Pressure Control and May.imum
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Reactivity Feedback (N-1LoopOperation) 4.3-39 Core Inlet Temperature, Coolant Average Temperature 81 and DNBR for Turbine Trip Without Pressure Control and Maximum Reactivity Feedback (N-1 Loop Operation)
- 4.3-40 Nuclear Power and Coolant Average Temperature for 82 Inadvertent Openine, of a Pressurizer Safety or Relief I
Valve 1
- g. 4.3-41 Pressurizer Pressure and DNBR for Inadvertent Opening 83 of a Pressurizer Safety or Relief Valve
- 0523v:1o/062287 ix
1.0 INTR') DUCTION Westinghouse Electric Corporation has been contracted by Northeast Utilities to remove the existing Resistance Temperature Detector (RTD) Bypass System
-(described in FSAR Section 5.4.3.2) and replace this hot leg and cold leg temperature measurement method witn fast response thermowell mounted RTDs installed in the reactor coolent loop piping. This report is submitted for the purpose of supporting the N and N-1 loop operation of Millstone Unit 3 utilizing the new thermowell mounted RTDs.
1.1 HISTORICAL BACKGROUND Piier to 1968, PWR designs had been based on the assumption that the hot leg i temperature was uniform across the pipe. Therefore, placement of the temperature instruments was not considered to be a factor affecting the 4 accurecy 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 regior.s of the reactor core at different temperatures produces significant temperature gradients within the pipe.
o
~
The reactor coolant loops required cooling and draining before the j RTDs could be replaced.
The RTD bypass system was designed to resolve these problems; however, operating plant experisnce has now shown that operation with the RTD bypass j loops has created it's own obstacles such as:
l i
o Plant shutdowns caused by excessive primary leakage through valves, l flanges, etc., or by interruptions of bypass flow due to valve stem failure.
l Os23v:1D/062287 1
o Increased radiation exposure due to maintenance on the bypass line and to crud traps which increase radiation exposure throughout the loop l compartments.
l The proposed temperature measurement modification has been developed in response to both sets of problems encountered in the past. Specifically:
o Removal of the bypass lines eliminates the components which have been a major source of plant outages as well as Occupational Radiation Exposure (ORE).
o Three therm'owell-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 thermowcils permits RTD replacement without draining the reactor coolant loops.
Following is a detailed description of the effort required to perform this modification.
1.2 MECHANICAL MODIFICATIONS The individual loop temperature signals required for input to the Reactor
, Control and Protection System will be obtained using RTDs installed in each
. reactor coolant loop.
1.2.1 Hot Leg a) Loops B and C The hot leg temperature measurement on each loop will De accomplished with three fast response, narrow range, single-element RTDs mounted in j thermowells. To accomplish the sampling function of the RTD bypass i
manifold system and eliminate the need for additional hot leg piping penetrations, the thermowells will be located within the three existing Os23v;1D/062287 2
[l 1
\
RTD bypass manifold scoops. A hole will be drilled through the end of. I 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 j (Figure-1.2-1).- These three RTDs will measure the hot leg temperature j' which is used to calculate the reactor coolant loop differentia 1 {
L temperature (AT)'and average temperature (T,yg).
]
{
[ b) Loops A and D l
l Two of the three thermowells will be mounted as described in (a) above.
The third thermowell cannot be installed in the existing scoop location due to structural interference. Therefore, this thermowell will be i relocated downstream of the existing scoop on the outside radius of the steam generator inlet elbow.
These thermowells will be mounted in i
independent bosses (Figure 1.2-3) and the resulting unused hot leg scoops will be capped.
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 1. 'd Leg 3
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 {
4 l' the cold leg RTDs located in the bypass manifold). Temperature streaming in the cold leg is not a concern due to the mixing action of the RCP. For this reason, only one RTD is required. This RTD will measure the cold leg temperature which is used to calculate reactor coolant loop AT and T,yg. The existing cold leg RTD bypass penetration nozzle will be modified (Figure 1.2-2) to accept the RTD thermowell. One element of the i
RTD will be considered active and the other element will be held in reserve as a spare.
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- 10/062287 3 l
b) Thit 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.
1.2.3 Crossover Leg The RTD bypass manifold return line will be capped at the nozzle on the crossover leg.
1.3 ELECTRICAL HDDIFICATIONS l 1.3.1 Function Figure 1.3-1 shows a block diagram of the modified electronics. The hot leg i
RTD measurements (three per loop) will be electronically averaged in the process protection system. The averaged T signal will then be used with hot the T cold signal to calculate reactor coo' ant loop AT and I,yg which are used in the reactor control and protection system. This will be accomplished by additions to the existing process control equipment.
1.3.2 Qualification The 7300 Process Electronics and RTD qualification will be verified to support
. Northeast Utilities compliance to 10CFR50.49. 7300 Process Electronic Equipment seismic and environmental qualification will be to IEEE standards 344-1975 and 323-1974, respectively, as described in WCAP-8587, Rev. 6-A,
" Methodology for Qualifying Westinghouse WRD Supplied NSSS Safety Related Electrical Equipment". The RTD qualification testing, to address the requirements of IEEE Standards 344-1975 and 323-1974, will be performed by the RTD manufacturer, WEED Instruments Inc. The WEED qualification documentation will be reviewed to verify compliance to the IEEE standards.
o523v:1o/062287 4
1.3.3 RTD Operability Indication Existing control board AT and T,yg indicators and alarms will provide the means of identifying RTD failures. The spare cold leg RTD element provides sufficient spare capacity to accommodate a single cold leg RTD fail.ure per loop. Failure of a hot leg RTD will require manual action to defeat the failed signal, and a manual rescaling of the electronics to average the remaining signals (see Figure 1.3-1 and Section 4.6).
0523v:1D/062287 5
a, c l
I 1
t-9 c.
t 3
Figure 1.2-1 Hot Leg RTD Scoop Modification for Fcst Response RTD Installation 4
6 e
a, c 1
i Figure 1.2-2 Cold Leg Pipe Nozzle Modification for Dual Element Fast Response RTD Installation 1
e 7
. a,c I
r Figure 1.2-3 Additional Bosses for Hot Leg Fast-Response RTD Installation
\ r O
e 8
_ _ - - - _ _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ~ ~ ~ ~ ~~
f' -
- o,e
+
'i i
l I
'I h,
1 f
Figure 1.3-1 RTD Averaging Block Diagram. Typical for Each of 4 Channels 9
l 2.0 TESTING i
There are two specific tests which have been performed to support the installationofthefastresponseRTDsinthereactorcoolantpipftg: a response time test and a hot leg temperature streaming test.
2.1 RESPONSE TIME TEST !
The RTD manufacturer, WEED Instruments Inc., will perform time response testing of each RTD and thermowell prior to installation at Millstone Unit 3.
- These RTD/thermowells must exhibit a response time bounded by the values shown in Table 2.1-1.
This table provides a comparison of the existing RTD Bypass System overall response time with that of the new fast response Thermowell RTO System. The overall response time of the new system is 1.0 second slower.
This slower response time has been factored into the transient analyses discussed in Section 4.0 Northeast Utilities has performed response time testing of similar WEED RTDs installed at their Haddam Neck and Millstone Unit 2 plants. This in-situ testing has demonstrated that the WEED RTDs can satisfy the response times requirements when installed in the plant.
2.2 STREAMING TEST Past testing at Westinghouse PWRs has established that temperature stratification exists in the hot leg pipe with a temperature gradient from 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 betwean branch line temperatures, confirm the short-term and long-term stability of the temperature streaming patterns and evaluate the impact on the indicated temperature if only 2 of the 3 branch line temperatures are used te determine an average temperature. This plant spec 5fic data is used in conjunction with data taken from other Westinghouse os23v:1D/062287 10
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 orocram. First. it is conservative to state that the streaming pattern [: Jb ,c.e Steady state data taken at 100% power for a period of fcur weeks indicates that the streaming pattern [ )b,c.e In other words, the .
temperature gradient [
Jb ,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 J ,c.e into the "two RTD" average to obtain the "three RTD" expected reading. A generic procedure (see Appendix B) has been provided to Northeast Utilities which specifies how these [ )b,c.e are to be determined. 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.
c523v;1o/o62287 11
TABLE 2.1-1 RESPONSE TIME PARAMETERS FOR RCS TEMPERATURE MEASUREMENT e Fast Response RTD Bypass System Thermowell RTD System 3
~
ac _ a,c
' ~ ~
RTD Bypass Piping and Thermal Lag (sec)
RTD Response Time (sec) i-ElectronicsDelay(sec)
TotalResponseTime(sec) 6.0 sec 7.0 see
- f. .
r O
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.e e
.o 1
l Os23v:1D/062287 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 includs4 in the Safety Analysis: Calorimetric Flow Measurement Uncertainty and Hot 'eg Temperature Streaming Uncertainty.
p 3.1 CALORIMETRIC FLOW MEASUREMENT LAERTAINTY 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, e major impact on the accuracy of the flow measurement.
The current licensed flow measurement uncertainty for Millstone Unit 3 for the sum of the four loop flows including elbow taps, is about i 2.4% flow (not including 0.1% flow for feedwater venturi fouling allowance). For N-1 loop operation the current flow measurement uncertainty is 12.76% flow. However, with the use of three T hot RTDs (resulting from the elimination of the RTD Bypass lines) and the latest Westinghouse RTD cross calibration procedure (resulting in lower RTD calibration uncertainties at the beginning of a fuel cycle), it is possible to reduce the RCS flow measurement uncertainty to approximately 1 1.8% flow (including the cold leg elbow taps and excluding feedwater venturi fouling) for N-loop operation and 12.0% flow for N-1 loop operation. 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 Millstone Unit 3 specific instrument uncertainties, calorimetric sensitivities, and flow uncertainties.
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 0523v;1 D/062287 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 E
, Jb.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 [ l b,c,e for a hot leg measurement. Of this total, [
)b,c.e. This overall temperature
~
streaming uncertainty is applicable to Loops B and C (thermowells in 3 existing scoops) and Loops A and D with the relocation of the third thermowell downstream of the existing scoops.
~
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,
[
]. Although the new method measures temperature at one point
.- at 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 thermowell 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 [
i )b,c.e exists from the center to the end of the scoop, the i
difference between the thermowell and scoop measurement is limited to
[ )b,c e .
Since three RTD measurements are averaged, and the nonlinearities at each scoop are random, the effect of this error bn the hot leg temperature measurement is limited to [ ]b,c e On the other 0523v:10/062287 14
h'and, imbalanced scoop flows can introduce temperature measurement
-uncertainties of up to [
,3'.
t . In all cases, the flow'inbalance uncertainty will equal or exceed the b
[ J c.e sampling uncertainty for the thermowell RTDs, so the'new measurement system tends to be a more accurate measurement with respect to streaming uncertainties.
Temperature streaming measurements from. 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, Provisions were made in the RTD electronics for operation with only two hot leg RTDs in service. The two-RTD measurement will be biased to correct for l* the difference compared with the three-RTD average. Based on test data, the bias value would be expected to range between [ ^]bc.e. Data comparisons show that the magnitude of this bias varied less than
[ ']b,c.e over the test period.
l 0523v:1o/062287 15
TABLE 3.1-1 FLOW CALORIMETRIC INSTRUMENTATION UNCERTAINTIES
(% SPAN) FW TEMP FW PRES FW d/p STM PRESS T T RdPRESS a,c H c SCA =
N&TE =
SPE =
STE =
SD =
R/E =
RDOT =
BIAS =
i; CSA =
[ $W i.
f 0F INST USED 1 3 1 1 **
1
- F psia % d/p psia *F *F psia INST SPAN =
INST UNC.
(RANDOM) = a,c INST UNC.
=
(BIAS) i
~~
NOMINAL =
Number of Feedwater, Hot Leg and Cold Leg RTDs used for measurement in each loop and the number of narrow range RCS pressure channels used overall, i.e. one per loop.
0523v:1o/062287 16
i TABLE 3.1 ,
FLOW CALORIMETRIC SENSITIVITIES f,
I.
I'
]
FEEDVATER FLOW
- 1 F, ._
TEMPERATURE =
a,e MATERIAL =
DENSITY TEMPERATURE =
PRESSURE =
DELTA P- =
FEEDWATER ENTHALPY TEMPERATURE =
PRESSURE =
h~
, h, h
=
=
1193.3 BTU /LBM f 415.3 BTU /LBM
=
r Dh(SG) 778.0 BTU /LBM STEAM ENTHALPY PRESSURE =
a,e MOISTURE =
HOT LEG ENTHALPY
!'- =
TEMPERATURE 0 .
PRESSURE =
hg =
638.7 BTU /LBM h =
555.9 BTU /LBM
- e j Dh(VESS) =
' 82.7 BTU /LBM Cp(T ) =
1.540 BTU /LBM *F H
)
COLD LEG ENTHALPY c TEMPERATURE =
l a , c.
PRESSURE = (
l Cp(Tc ) =
1.261 BTU /LBM 'F COLD LEG SPECIFIC VOLUME TEMPERATURE =
ae,
. PRESSURE =
m
'o e 0523v:10/062287 17 i
TABLE 3.1-3 CALOR! METR!C RCS FLOW MEASUREMENT UNCERTAINTIES COMPONENT INSTRUMENT ERROR FLOW UNCERTAINTY FEEDWATER FLOW - -
VENTURI **C THERMAL EXPANSION COEFFICIENT TEMPERATURE NATERIAL DENSITY TEMPERATURE PRESSURE DELTA P FEEDWATER ENTHALPY TEMPERATURE PRESSURE STEAM ENTHALPY PRESSURE -
i 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
, RTD CROSS-CAL SYSTEMATIC ALLOWANCE BIAS VALUES
.- FEEDWATER PRESSURE DENSITY ENTHALPY STEAM PRESSURE ENTHALPY PRESSURIZER PRESSURE ENTHALPY - HOT LEG ENTHALPY - COLD LEG
' SPECIFIC VOLUME - COLD LEG FLOW BIAS TOTAL VALUE
- +,
, ++ INDICATE SETS OF DEPENDENT PARAMETERS SINGLE LOOP UNCERTAINTY (WITHOUT BIAS VALUES) a,e N LOOP UNCERTAINTY N (WITHOUTBIASVALUES)
LOOP UNCERTAINTY (WITHBIASVALUES) -
esn<:10/osus7 18
TABLE 3.1-4 COLD LEG ELBOW TAP FLOW UNCERTA!NTY f
INSTRUMENT UNCERTAINTIES t
, % d/p SPAN % FLOW .
PMA
- a,c PEA =
SCA =
SPE =
STE =
SD =
RCA =
i N&TE =
I RTE =
lr RD =
4: ID = .
- A/D =
.- RDOT = ,
- . BIAS =
- ~
FLOW CALORIM. BIAS =
FLOW CALORIMETRIC =
INSTRUMENT SPAN =
SINGLE LOOP ELBOW TAP FLOW UNC = -
% FLOW a*e N ' LOOP ELBOW TAP FLOW UNC =
N- LOOP RCS FLOW UNCERTAINTY i (WITHOUT BIAS VALUES) =
N - -
. LOOP RCS FLOW UNCERTAINTY (WITH BIAS VALUES) = 1.8 0
t o
DE23v:1D/DC2287 19
TABLE 3.1-5 FLOW CALORIMETRIC INSTRUMENTATION UNCERTAINTIES N-1 LOOP OPERATION l
(% SPAN) FW TEMP FW PRES FW d/p STM PRESS T T, RCS PRESS a,e H
SCA =
N&TE =
SD =
R/E =
~
RDOT =
- . BIAS =
CSA =
~
? -
~
f 0F INST USED 1 3 1 3 we
- F psia % d/p psia *F 'F psia
~'
- INST SPAN =
[' INST UNC.
,... (RANDOM) = ,,e
.- INST UNC.
. (BIAS) =
- =
NOMINAL
~
~
Number of Feedwater, Hot Leg e.nd Cold Leg RTDs used for measurement in each loop and the number of narrow range RCS pressure channels used 1 overall i.e. one per loop.
l
- l 0523v:1D/062267 20 )
i TABLE 3.1-6 4
FLOW CALORIMETRIC SENSITIVITIES N-1 LOOP OPERATION FEEDWATER FLOW F
A _
TEMPERATURE =
,,e NATERIAL =
DENSITY TEMPERATURE =
PRESSURE =
l DELTA P =
FEEDWATER ENTHALPY p TEMPERATURE =
f PRESSURE =
' ~
h, =
1195.2 BTU /LBM h =
380.5 BTU /LBM f
.- Dh(SG) =
814.7 BTU /LBM STEAM ENTHALPY
~
= --
PRESSURE a,c MOISTURE =
HOT LEG ENTHALPY TEMPERATURE =
PRESSURE =
~
~
hg =
625.7 BTU /LBM h =
.. c 547.9 BTU /LBM Dh(VESS) =
_ 77.8 BTU /LBM Cp(T )
- 1.476 BTU /LBM 'F H
COLD LEG ENTHALPY
~
~
TEMPERATURE = -
,,e PRESSURE' =
Cp(Tc ) =
1.244 BTU /LBM *F
~
COLD LEG SPECIFIC VOLUME
~
~
TEMPERATURE =
,,e PRESSURE =
ceavaomcus, g3
______-_-______._____.______._m___ -_
-1 J TABLE 3.1-7 i CALORIMETRIC RCS FLOW MEASUREMENT UNCERTAINTIES I
N-1 LOOP OPERATION COMPONENT INSTRUMENT ERROR FLOW UNCERTAINTY
. FEEDWATER FLOW - -
VENTURI a,c THERMAL EXPANSION COEFFICIENT '
TEMPERATURE MATERIAL DENSITY TEMPERATURE PRESSURE DELTA P
. FEEDWATER ENTHALPY TEMPERATURE PRESSURE i STEAM ENTHALPY 9: PRESSURE ~
i ' ~
MOISTURE
'r NET PUMP HEAT ADDITION E.
HOT LEG ENTHALPY
. TEMPERATURE
' STREAMING. RANDOM
, STREAMING, SYSTEMATIC i
PRESSURE ,
. COLD LEG ENTHALPY TEMPERATURE PRESSURE COLD LEG SPECIFIC YOLUME TEMPERATURE
?, PRESSURE RTD CROSS-CAL SYSTEMATIC ALLOWANCE
. BIAS VALUES FEEDWATER PRESSURE DENSITY
~ ENTHALPY STEAM PRESSURE ENTHALPY PRESSURIZER PRESSURE ENTHALPY - HOT LEG '
ENTHALPY - COLD LEG SPECIFIC VOLUME - COLD LEG FLOW BIAS TOTAL VALUE
, +, ++ INDICATE SETS OF DEPENDENT PARAMETERS
~
SINGLE LOOP UNCERTAINTY (WITHOUT BIAS VALUES a,e N-1 LOOP UNCERTAINTY (WITHOUT BIAS VALUES
, . N-1 LOOP UNCERTAINTY (WITH BIAS VALUES) -
l 0523w1D/062287 22 o ____ _ __ _ _-
l TABLE 3,1-8 COLD LEG ELBOW TAP FLOW UNCERTAINTY N-1 LOOP OPERATION INSTRUMENT UNCERTAINTIES
% d/p SPAN % FLOW PMA =
=
a,c PEA SCA =
SPE =
STE =
l SD =
RCA =
N&TE =
l ;. RTE =
l- RD =
c 10 =
6 A/D =
RDOT =
BIAS =
+
FLOW CALORIM. BIAS =
FLOW CALORIMETRIC =
INSTRUMENT SPAN =
SINGLE LOOP ELBOW TAP FLOW UNC = % FLOW a,e N-1 LOOP ELBOW TAP FLOW UNC =
N-1 LOOP RCS FLOW UNCERTAINTY (WITHOUT BIAS VALUES) =
N-1 LOOP RCS FLOW UNCERTAINTY -
(WITH BIAS VALUES) = 2.0 f
0523v:10/D62287 23
l 4.0 SAFETY EVALUATION 4.1 RESPONSE TIME The primary impact of the RTD bypass elimination on the FSAR Chapter 15 non-LOCA safety analyses (Reference 2) is the increased response time '
associated with the fast response thermowell RTD system. The secondary impact is the possible increase in instrument uncertainties. Currently, the overall response time of the Millstone Unit 3 RTD bypass system assumed in the safety
,. analyses is approximately 6.0 seconds (see Table 2.1-1). For the fast i
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 AT or Overpower -- -~
AT reactor trip until a trip signal is actually generated. Therefore, those transients that rely on the above mentioned trips must be evaluated for the longer response time. The affected transients include Uncontrolled RCCA Bank Withdrawal at Power, Loss of External Electrical Load / Turbine Trip, Steamline Rupture at Power, and Inadvertent Opening of a Pressurizer Safety or Relief Valve. These events are discussed in Section 4.3.
4.2 RTD UNCERTAINTY The proposed fast response Germowell RTD system will make use of RTDs, manufactured by Weed Instruments Inc., with a total uncertainty of
[ ]"'C assumed for the analyses, b
The FSAR analyses make explicit allowances for instrumentation errors for some of the reactor protection system setpoints. In addition, allowances are made for the average reactor coolant system (RCS) temperature, pressure and power as described in FSAR Section 15.0. These allowances are made explicitly to the initial conditions.
oG23v;1o/D62287 24 s -
The following protection and control system parameters were affected by the change from one hot leg RTD to three het leg RTDs; the Overtemperature AT (OTDT), Overpower AT (0PDT), 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.
Changes have been made in the reactor protection system setpoints to acccunt L
for the new thermowell mounted RTDs. With one exception, the current values of the nominal setpoints as defined by the Millstone Unit 3 Technical Specifications remain valid, with a change ir the corresponding Allowable Values. The only nominal setpoint that requires modification is ty, a time constant for lead-lag compensation of AT, which will increase from 8.0 to 12.0 sec. This change was made to support the non-LOCA transient analysis. (See Section 4.8 and Appendix A) 4.3 NON-LOCA TRANSIENTS REANALY2ED 0
All the events reanalyzed in this section used the LOFTRAN computer code.
LOFTRAN (Reference 3) is a digital computer code, developed to simulate transient behavior in a multi-loop pressurized water reactor system. The progran simulates the neutron kinetics, thermal-hydraulic conditions, pressurizer, steam generators, reactor coolant pumps, and control and protection system operation. The secondary side of each steam generator utilizes a homogeneous saturated mixture for the thermal transients.
For each event reanalyzed the basic assumptions regarding initial conditions, instrumentation errors, and setpoint errors that are not directly related to the RTD Bypass Elimination remain largely the same as those found in Chapter 15 of the FSAR. However, the current analysis does incorporate certain
, additional changes that should be noted. FSAR Section 15.0.3.2 specifies a 1 30 psi allowance on pressurizer pressure for steady state fluctuations and
. measurement penalty. The reanalyzed events include a more conservative 1 45 psi allowance for pressurizer pressure. Additionally, increased uncertainties 0523v;1o/062287 25
have been applied to pressurizer and steam generator water levels. These uncertainties have been increased from 5% to 5.73% and 5.53%, respectively, for the pressurizer and the steam generator. These increased uncertainties have been incorporated to bound calculated increases in the associated transmitter uncertainties. Finally, the revised analyses presented here bound the increase in the lead-lag compensation of AT noted in Section 4.2.
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 re;11ts in an increase in the core heat flux. Since the heat extraction from the steam generator lags behind the core power generation, there is a net increase in the reactor coolant temperature.
Unless terminated by manual or automatic action, the increase in coolant temperature and power could result in DNB. For this event, the Power Range High Neutron Flux and Overtemperature AT reactor trips are assumed to provide protection against DNB. Therefore, this event was reanalyzed with increased time constants to show that the DNSR limit is met.
Methods With the exception of the items noted here and in Section 4.3, the assumptions used are consistent with the FSAR. The transient is analyzed at 10%, 60%, and 100% power for N-loop operation and at 10% and 75% power for N-1 loop operation. Both minimum and maximum reactivity feedback cases were reanalyzed with the increased time response value. A constant moderator coefficient of
+ 5 pcm/*F was used for the cases based on minimum feedback. The assumption that a positive moderator coefficient exists at full power is conservative since at full power the moderator coefficient will actually be zero or negative. 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 slow insertion rate from full power (N and N-1 Loop Operation) is I
os2mo/oc22e7 26
)
i I
t I
presented on Table 4.3-1. The transient response for a fast insertion case and a slow insertion case from full power is shown in Figures 4.3-1 through 4.3-8 (N-loop) and Figures 4.3-22 through 4.3-29 (N-1 loop operation). The plots of minimum DNBR versus reactivity insertion rate at the analyzed power levels are shown as Figures 4.3-9 through 4.3-11 (N-loop) and Figures 4.3-30 and 4.3-31 (N-1 loop operation).
Conclusions The limit DNBR is met, and therefore, the conclusions presented in the FSAR remain valid for both N and N-1 loop operation.
4.3.2 Loss of Load / Turbine Trip The Millstone Unit 3 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 AT. Thus, the increase in RTD response time may have an effect on the results of this transient, since the RTDs provide input to the Overtemperature AT trip.
Methods With the exception of items noted here and in Section 4.3, the assumptions
, used are consistent with the FSAR. All fcer cases presented in the FSAR were reanalyzed for both N and N-1 loop operation wcorporating the assumptions of the RTD Bypass Elimination. These are minimum recctivity feedback (Beginning of Life) and maximum reactivity feedback (End of Life), with and without pressure control (pressurizer spray and PORVs). The minimum feedback cases used a constant moderator coefficient of +5 pcm/*F, which is conservative at full power, as discussed in Section 4.3.1. 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 0523v;10/0b2287
'27
through 4.3-19 (N-loop) and Figures 4.3-32 through 4.3-39 (N-1 loop operation). A calculated sequence of events is shown in Table 4.3-2. Figures 4.3-12 and 4.3-13 (N-loop) and Figures 4.3-32 and 4.3-33 (N-1 loop operation) show the responses for a turbine trip event with minimum reactivity feedback assuming operability of pressurizer sprays and PORV's. The reactor is tripped by the High Pressurizer Pressure trip function. The DNBR increases throuchout most of the transient and never drops below the design limit. The primary system pressure remains below the 110% design value.
Figures 4.3-14 and 4.3-15 (N-loop) and Figures 4.3-34 and 4.3-35 (N-1 loop j
operation) show the responses for a turbine trip with maximum reactivity feedback and pressure control. The reactor is tripped by the Low-low Steam Generator Level trip function, and the DNBR never drops below the initial value.
The primary system pressure remains below the 110% dusign value.
Figures 4.3-16 and 4.3-17 (N-loop) and Figures 4.3-36 and 4.3-37 (N-1 loop operation) show the responses for a turbine trip with minimum reactivity p
feedback 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 primary system pressure remains below the 110% design value.
Figures 4.3-18 and 4.3-19 (N-loop) and Figures 4.3-38 and 4.3-39 (N-1 loop operation) show the responses for a turbine trip with maximum reactivity feedback 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 primary system pressure remains the below the 110% design value.
Conclusions The DNBR design basis is met and the system pressure remains below 110% of the design value in all four cases for both N and N-1 loop operation, and therefore, the conclusions presented in the FSAR remain valid. The Overpressure Protection Report is also not impacted by the RTO bypass elimination effort, and thus, the conclusions presented in that document remain unchanged.
0523v:1o/062287 28
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 was performed for both N and N-1 loop operation. A range of break sizes was considered and the results demonstrated that the design basis as described in WCAP-9226-R1 was met.
4.3.4 Inadvertent Opening of a Pressuriree Safety or Relief Valve This event is described in Section 15.6.1 of the FSAR. Initially, the event '
results in rapidly decreasing RCS pressure until the pressure reaches a value corresponding to the hot leg saturation pressure. At that time, the pressure decrease is slowed considerably. The reduction in pressure could result in DNB unless terminated by manual or automatic action. For this event, the Overtemperature AT function is assumed to provide protection against DNB.
Therefore, the RCS depressurization incident is analyzed with increased RTD time constants to show the DNBR limit is met.
Methods Assumptions made in the RCS Depressurization analysis include a constant moderator temperature coefficient (+5 pcm/*F) and 2 small -(absolute value)
Doppler coefficient of reactivity such that the resultant amount of positive feedback is conservatively high. The rod control system is assumed to be in the manual mode in order to prevent rod insertion due to ar. RCS temperature and power mismatch prior to reactor trip. With the exception of the items noted here and in 4.3, the method of analysis and assumptions used were otherwise in accordance with those presented in the FSAR. As m the FSAR, the reanalysis considers both N and N-1 loop operation.
Results A calculated sequence of events in presented in Table 4.3-3 and Figures 4.3-20 and 4.3-21 (N-loop) and Figures 4.3-40 and 4.3-41 (N-1 loop operation) show the nuclear power, average temperature, pressurizer pressure, and DNBR vs.
time for the accidental depressurization of the RCS. The positive moderator os23v:1o/062287 29
coefficient causes nuclear power to increase as pressure decreases until reactor trip occurs on Overtemperature AT.
The DNBR decreases initially, but increases rapidly following the trip. The DNBR remains above 1.30 throughout the transient.
Conclusions The analysis demonstrates that the DNBR remains above 1.30 and therefore, the conclusions presented in the FSAR remain valid for both N and N-1 loop operation.
4.3.5 Conclusion The impact of the RTD bypass elimination for Millstone Unit 3 on 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 for both h and N-1 loop operation.
4.4 LOCA Evaluation The elimination of the RTD bypass system impacts the uncertainties associated with RCS temperature and flow measurement.
The magnitude of the uncertainties are such that RCS inlet and outlet temperatures, thermal design flow rate and the steam generator performance data used in the LOCA analyses will not be affected.
~
Past sensitivity studies have shown that the variation of the core inlet temperature (Tin) used in the LOCA analyses affects the predicted core flow during the blowdown period of the transient.
The amount of flow into the core is influer.:ed by the two phase vessel side break flow, and the core cooling is affected by the quality of the fluid. These sensitivity studies concluded that the inlet temperature effect on peak clad temperature is dependent on break size.
As a result of these studies, the LOCA analyses are performed at a nominal value of T in without consideration of small uncertainties.
The RCS flow rate and steam generator secondary side temperature and pressure are also determined using the loop average temperature (T analyses are n,yg) output. These nominal values used as inputs to the ot affected due to the RTD bypass elimination. It is concluded oS23v:1o/062287 30
that the elimination of the RTD bypass piping will not affect the LOCA analyses input and hence, the results of the analyses for both N and N-1 loop 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 SGTR EVALUATION For the Steam Generator Tube Rupture (SGTR) event, the FSAR SGTR analysis (FSAR Section 15.6.3) was performed using the LOFTRAN Code. The primary to secondary break flow was assumed terminated at 30 minutes after initiation of the SGTR event.
The major factors that affect the radiological doses of an SGTR event are the amount of fuel failure, the amount of primary coolant transferred to the secondary side of the ruptured steam generator through the ruptured tube after reactor trip, and the steam released from the ruptured steam generator to the atmosphere.
The modifications for the RTD Bypass Elimination will result in a small change in the time of reactor trip on Overtemperature AT.
The assumption of an earlier trip time is conservative for SGTR analysis since a reduction in time of reactor trip will increase the calculated offsite radiation doses. Prior to reactor trip with the condenser in service, the radioactivity released to the atmosphere is through the condenser air ejector and is unimportant for offsite doses. After reactor trip and with an assumed loss of the offsite power, the steam and radioactivity are released directly to the atmosphere through the ruptured SG relief valve. An earlier reactor trip will result in a longer time for the radioactivity released from the ruptured SG relief valve, thereby increasing the calculated offsite radiation doses. Conversely, the assumption of an increase in RTD response time resulting in a delay in reactor trip time will result in a decrease in the calculated radioactivity release to the atmosphere.
The impact on the FSAR SGTR analysis of the change in RTD response time and lead / lag compensation on the measured AT was evaluated.
The results of the evaluation indicated that the minimum DNBR for the SGTR event remains above the DNBR limits; therefore, fuel O ilure will not occur for the SGTR event.
The results also indicated that the changes in RTD response time and lead / lag compensation for the measured AT result in an insignificant change in the os23v:1D/062287 31
time of actuation of the reactor trip signal, and therefore, an insignifant p
change in the primary to secondary leakage and steam released from the i
j ruptured SG after trip. Since the fuel thermal limits are not violated for i
the SGTR event and the impact on the primary to secondary leakage and steam released from the ruptured SG after trip is insignificant, the impact on the i
- j. offsite radiation doses is insignificant.
i Based on the above evaluation, it is concluded that the RTD Bypass Elimination t
will not change the conclusions reported in the FSAR SGTR analysis. Since the offsite doses with the RTD Bypass Elimination are much less than a small fraction of ^he 10CFR100 limits and the DNB limits are met, no reanalysis or FSAR changes are required.
4.6 INSTRUMENTATION AND CONTROL (I&C) SAFETY EVALUATION The RTD Bypass Elimination modification for Hillstone Unit 3 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 dual element 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 T One element cf the RTD will be held in reserve as a spare. eoid signal.
- 2. !
The NR hot leg RTD in the bypass manifold will be replaced with 3 fast response NR RTDs well mounted in the hot leg that are electronically averaged in the process protection system. The signal from this average T
hot circuit obtained.from these 3 NR T hot RTDs will perform the same function as the existing RTD T hot signal.
3.
Identification of failed signals will be by the same means as before the modifications, i.e., existing control board alarms and indications.
I 1
0523v:10/062287 32
4.
Signal process and the add:d circuitry to the Protection Set racks will be accomplished by additions to the process control (Westinghouse Model 7300 racks using 7300 technology. When one Thot signal is removed from the averaging process, the electronics will allow a bias to be manually added to a 2-RTD average Thot (as opposed to a 3-RTD average Thot) in rder to obtain a value comparable with the 3-RTD average T hot prior,to the failed RTD.
In the event of a cold leg RTD failure, the spare cold leg RTD element will be manually connected to the 7300 circuitry in place of the failed RTD.
Existing contrc1 board AT and Tavg indicators and alarms will provide
. the means of identifying RTD failures. {
Upon identification of a failed RTD, the operator would place that protection channel in trip (consistent with the time requirements specified in the Technical Specification),
identify and disconnect the failed RTD, and rescale the summing amplifier averaging card for a two RTD input condition. The channel would then be returned to service.
During this process the plant will be in a partial trip mode and will therefore be in a safe condition.
For more detail on this procedure, see Appendix B.
~
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 four 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.
Following is a discussion of design basis requirements in conformance to portinent I&C criteria:
i i
Os23v;10/062iB7 33
a.
Single failurs 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.
c.
Equipment seismic and environmental qualification will be to IEEE standards 344-1975 and 323-1974, respectively, as described in WCAP 8587, Rev. 6-A " Methodology for Qualifying Westinghouse WRD Supplied NSSS Saf Related Electrical Equipment".
d.
The changes will continue to maintain the capability of the protection system to initiate a reactor trip during and following natural phenomena credible to the plant site to the same extent as the existing system.
e.
Channel independence and electrical separation is maintained because the Protection Set circuit assignments continue to be Loop 1 circuits input to Protection Set I; Loop 2, to Protection Set II; Loop 3, to Protection Set l 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 That outputs to the computer sent through Class IE isolators in each Protection Set.
,f.
- The compliance of the hardware to IEEE 279-1971 Section 4.7 and GDC
~
requirements concerning Control and Protection interaction has not been changed for N and N-1 loop operation.
On the basis of the foregoing evaluation, it is concluded that the compliance of Millstone Unit 3 to IEEE 279-1971, applicable GDCs, and industry standards and regulatory guides has not been changed with the I&C modifications required for RTD bypass removal.
\
i c523v:1o/o62287 !
34 1
1 l
4.7 MECHANICAL SAFETY EVALUATION
'The presently instilled RTD bypass system is to be replaced with fast acting c- narrow range RTD thermowells. This change requires modifications t'b the hot
^ leg scoops, the hot leg piping, the crossover leg bypass return nozzle, and
-the cold leg bypass manifold connection. All welding and NDE will!be performed per ASME Code Section XI requirements. Each of these modifications is evaluated below.
The original three scoops in the loop B and C hot legs, which feed the bypass
~
manifold, and the bypass manifold connection must be removed and the scoops modified to accept three fast response RTD thermowells. C.
ja.: to provide the proper flow path. A thermowell design will be used such that 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 Code. The installation of the thermowell into the scoop will be performed using GTAW for the root pass and finished out with either GTAW or SMAW. The 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.
~
One het leg fast response RTD will be installed into a new penetration in the reducing elbow leading to the steam generators in loops A and D. To accomplish this, a new boss will be installed approximately one foot downstream from the existing hot leg scoops. The remaining two hot leg scoops on each loop will be utilized for thermowell mounting as was done for loops B and C. The installation boss for the new penetrations and the thermowells will be root welded by GTAW. Finish welding can be either GTAW or SMAW. Weld inspection by Pi will be performed per Section XI. The installation bosses and thermowells are fabricated in accordance with Section III Class 2 of the ASME Code.
Os23v:10/062287 35
. _ _ _ . , , _ ..._ ~ ... _ _ . ..._ _ . _ . . _ . . .._ _ _.u .__, -_-
The cold leg RTD bypass line must also be removed. The nozzle must then be modified to accept the fast response RTD thermowell. TheinstallationofthE thermowell into the nozzle will be performed using GTAW for the root pass and ;
finished with either GTAW or SMAW. Weld inspection by PT will be performed as !
~
required by Section XI. The thermowells will extend approximate'ly T2.5)a,c inches into the flow stream. This depth has been justified based on (flow induced vibration]a.c analysis. The root weld' joining the thermowells to the modified nozzles will be deposited with GTAW and the remainder of the weld may be deposited with GTAW or SMAW. Penetrant testing will be performed in accordance with the ASME Code Section XI. The thermowells will be fabricated in accordance with the ASME Section III (Class 1).
~
Upon removal of the RTD bypass piping, the two hot leg scoops not utilized in Loops A and D will be capped. The caps will be fabricated in accordance with Section III Class 1 of the ASME Code. The root weld joining the caps to the '
A scoops 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.
The cross-over leg bypass return piping connection must be removed and the nozzles capped. The cap design, including materials, will meet the pressure boundary criteria and ASME Section III (Class 1). The cap will be root welded
~
to the nozzles by GTAW and fill welded by either GTAW or SMAW.
~
Non-destructive examinations (PT and radiographs) will be performed per ASME '
Section XI. Machining of the bypass return nozzle, as well as any machining performed during modification of the penetrations in the hot and cold legs, shall be performed such as to minimize debris escaping into the reactor coolant system.
In accordance with Article IWA-4000 of Section XI of the ASME Code, a
. hydrostatic test of new pressure boundary welds is required when the, connection to the pressure boundary is larger than one inch in diameter.
Since the cap for the crossover leg bypass return pipe is [ Ja,c inphes and
- the cold leg RTD connections are [ la,c inches, a system hydrostatic test is required after bypass elimination. Paragraph IWB-5222 of Section XI defines this test pressure to be 1.02 times the normal operating pressure af, a ,
temperature of [ ).a.c os23w1o/o62687 36
The integrity of the reactor coolant piping as a pressure boundary component, is maintained by adhering to the applicable ASME Code sections and Nuclear !
Regulatory Commission General Design Criteria. The pressure retaining !
capability and fracture prevention characteristics of the piping is not I compromised by these modifications.
~
~
4.8 TECHNICAL SPECIFICATION EVALUATION As a result of the calculations summarized 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 0523v:10/062287 37
- r. _
)
TABLE 4.3-1 p -(page 1 of 2)
TIME SE0VENCE OF EVENTS FOR A' t
RCCA BANK WITHDRAWAL AT POWER N-LOOP N-1 LOOP' s
. ACCIDENT- EVENT TIME (SECS) ' TIME (SEC)
'l Case-A Initiation of uncontrolled RCCA 0. 0 ~ 0.0 withdrawal at a fast reactivity
.. insertion rate (70 pcm/sec) with minimum reactivity feedback at full power-Power range high neutron flux 1.5 1.5 reactor. trip signal initiated Rods begin to' drop 2.0 2.0 ,
Minimum DNBR occurs 3.0 2.9 Peak water level in the 4.3 4.3 pressurizer occurs Case B Initiation of uncontrolled RCCA 0.0 0.0 withdrawal at a low reactivity insertion rate (3 pcm/sec) with
' minimum reactivity feedback at full power Overtemperature AT reactor 21.9 ---
trip signal initiated
. Power range high neutron flux ---
29.8 reactor trip signal initiated Rods begin to drop 23.4 30.3 Minimum DNBR occurs 23.9 30.7 Peak water level in the 25.5 32.3 pressurizer occurs
^4 Os230:1o/062287 38
1 1
TABLE 4.3-1 . !
w
. (page2of2)- i
-TIME SEQUENCE OF EVENTS FOR A RCCA BANK WITHDRAWAL AT POWER
. N-LOOP. N-1 LOOP-
' ACCIDENT EVENT TIME (SECS) TIME (SEC) e Case C Initiation'of: uncontrolled RCCA 0.0 0.0 withdrawal at a fast reactivity insertion rate.(70 pcm/sec) with maximum reactivity feedback at full power L. '
Power range high neutron flux 4.6' 4.5 reactor trip signal. initiated Rods begin to drop 5.3 5.0 Minimum DNBR occurs 5.7 5.2 Peak water. level in the 7.5- 7.1 pressurizer occurs Case'D Initiation of uncontrolled RCCA 0.0 0.0
- withdrawal at a low reactivity insertion rate (3 pcm/sec) with maximum' reactivity feedback at
, full power Overtemperature AT reactor 212.7 427.6 trip signal initiated Rods begin to drop 214.2 429'.1 Minimum DNBR occurs 213.1 429.1
- - Peak water level in the 216.1 431.1 pressurizer occurs L-0523v:1D/062287 39
TABLE 4.3 (page 1 of 2)'
..- TIME SEQUENCE OF EVENTS FOR A TURBINE TRIP
~
N-LOOP N-1 LOOP ACCIDENT EVENT TIME (SECS) ' TIME-(SEC)
Case A . Initiation of turbine trip, 0.0 0.0 loss of main feedwater flow, minimum reactivity feedback with pressure control Initiation of steam release 5.5 -8.5 from S/G safety valves High. pressurizer pressure 9.2 12.5 reactor trip signal generated Rods begin to drop 11.2 14.5 Peak pressurizer pressure occurs 12.5 16.0 Minimum DNBR occurs 12.5 15.5 Case B Initiation of turbine trip, 0.0 0.0 loss of main feedwater flow, maximum reactivity feedback with pressure control
- Initiation of steam release 5.5 8.5 from S/G safety valves
, Peak pressurizer pressure occurs 6.5 6.0 Low-low steam generator water 50.8 77.6
- level reactor trip signal generated Rods begin to drop 52.8 79.6 Minimum DNBR occurs (1) (1)
(1) DNBR does not decrease below its initial value.
oszav:1o/os22s7 40
i TABLE 4.3-2 (page 2 of 2)
TIME SEQUENCE OF EVENTS FOR A TURBINE TRIP i
l ..
N-LOOP N-1 LOOP ACCIDENT EVENT.
TIME (SECS) TIME (SEC)
Case C Initiation of turbine trip, 0.0 0.0 loss of main feedwater flow, minimum reactivity feedback i without pressure control High pressurizer pressure 5.1 6.7 i
reactor trip signal generated Initiation of steam release 5.5 8.5 from S/G safety valves l Rods begin to drop 7.1 8.7 Peak pressurizer pressure occurs 8.5 10.0 Minimum DNBR occurs (1) (1)
Case D Initiation of turbine trip, i
~
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i maximum reactivity feedback !
without pressure control High pressurizer pressure 5.0 6.7 reactor trip signal generated i
. Initiation of steam release 5.5 8*5
, from S/G safety valves Rods begin to drop 7.0 8.7 ,
Peak pressurizer pressure occurs 7.5 9.5 '
Minimum DNBR occurs (1) (1) 1 (1) DNBR does not decrease below its initial value. 4
. I l
0523v:1o/062287 41
___ _ _________J
,. TABLE 4.3 TIME SEQUENCE'0F EVENTS FOR AN~
INADVERTENT OPENING OF A PRESSURIZER SAFETY OR RELIEF VALVE-
~
N-LOOP N-1. LOOP EVENT TIME (SECS)' TIME (SEC)
Safety valve 0.0 0.0 fully opens Overtemperature AT reactor trip 15.5 31.1 signal.. initiated Rods begin to drop 17.0 32.6-Minimum DNBR occurs 17.4 33.2 e
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5.,0. CONTROL _SJSTEMEVALUATICS
- L u s
r 4 price ipput"tignal to the various NSSS control systems is the RCS. average :
hmperatwe (T,yg). This is calculated electronically as the averrtje of the
' nr.sured hot leg and cold leg temperatures in each loop.
T 1
The effect of thd riew RTD temperature measurement system is to potentially change the time response of the T,yg channels in the various loops. This in turri could impact the response of (
~
Ja.c liorever, as noted in Section 2.1, Table 2.1-1, the new RTD system will have e time ror.ponso close to that of the present system.
, Therefore, thers should }e no Significant effect in the T,yg channel 2~ restanse, add no apparent need to revise any of the control system setpoints.
g The need to modify control systou setpoints will be determined during the plant starthp fs11owing the installation of the new RTD system by observing the response'of the control systems. If necessary, signal compensators and function generators in the control systems could be adjusted to obtain a more e@imum dystem response.
In any case, the parameters listed in Table 2.1-1 would not require modification. Also, centrol system responses are not assumed in the FSAR transient analyses where reactor protection is provided by
- the Overtemperature and Overpowar AT trips, hence changing rod contro1
' system setpoints will not impact the results of these analyses where the values of 1'able 2.1-1 are assumed.
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__ -__ ._-:------ ~
6.0' CONCLUSIONS 1
The method of utilizing fast-response RTDs installed in the reactor coolant loop piping as a means for RCS. temperature indication has undergone extensive !
analyses, evaluation and testing as described in this report. The.
incorporation'of this system into the Millstone Unit 3 design meet,s all I
safety, ifcensing and control requirements necessary for safe'opeIration 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 as well as improve plant reliability.
. i l
oS23v:1D/o62287 g
E
7.0 REFERENCES
1.- Tuley, C.' R., Moomau,'W. H., "RCS Flow 1 Uncertainty for Shearon Harris
' Unit- 1", WCAP-11168 Rev.1, Proprietary,. WCAP-11169 Rev.1,-
Hon-Proprietary, October, 1986.
- 2. ..
Millstone Unit 3 Final Safety Analysis Report, Amendment 18, March 1986.
- 3. Burnett, T.W.T., et al. "LOFTRAN Code Description," WCAP-7907-P-A s .(Proprietary), WCAP-7907-A (Non-Proprietary), April 1984.
~
4523v:1D/062287 86
n-.__. --.
l i
- APPENDIX A TECHNICAL SPECIFICATION MODIFICATIONS 1
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a e m= g a m i
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$ M l i 2 T1 i n hy i s e
v g
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) Tt A L T 3 = =
s = = = = = = = = =
t = =
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R
(
++
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l uL m
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)
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) r m d E e o W o
)
u f O c
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t u t =
L A i n n i o . M d t C R ,
n ( K E ) e o H F e S d T
- d c N n 1 n x O a D . a e R.(C .
2 1
A T
I T "
T 1 E 7 T 8 A 5 R
1 t
o n
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o a L E r N n N h B L o n o s
. A B f i , t i n .
T A i ! t a
T F d T a a n t d
- e d n e i
o
/
9 i n e e n l
l p t
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e
.. T 1 0
0 d
e c r i t d e r o S 4 s d s n n s f p
,- 0 A I i i
A 0 r 4 T B
A===
m
=
)
=
i u
m x
m e a
, u I m O A
. ( s .
)
K T T S f
' n l a d ep e ns u n n aT i
t hA c n K o e# 8 C
(
h)1 T3 3 : ,
4 E .
T E O T M O N
2PG ] mg -
POWER DISTRIBUTION LIMITS y
3/4.2.3 ' RCS FLOW RATE AND NUCLEAR ENTHALPY RISE FOUR LOOPS OPERATING LIMITING CONDITION'FOR OPERATION '
3.2.3.1
- shall be maintained as follows:The indicated Reactor Coolant System ~
N (RC
_- AN a.- 3I52l0 RCS total flow rate 3 3 % gpm, and-N
- b. F 3H $ 1.49 [1.0 + 0.3 (1.0 - P))
Where:
y) p., .
THERMAL POWER ,
- RATED THERMAL POWER
- 2) Ffg= Measured values of FN
. - 3g obtained by using the movable-in '
core detectors to obtain a power distribution map. The measured value of F H should be used since Specification 3.2.3.Ib. "
.- takes into consideration a measurement uncertainty of 4% for incore measurement, and .
l . S 7. .. ". . '. "
3)
- The measured value[of RCS total flow rate shall be u Specification 3.2.3.la. uncertainties of SM% for flow measuremen
- ~
APPLICABILITY: NODE 1.
~
, ACTION: .
WiththeRCStotalflowrateorFfg outside the region of acceptable operation:
- s. Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> either:
1.
Restore limits, or the RCS total flow. rate and Fh to within the abo -
- 2.
Reduce THERMAL POWER to less than 50% of RATED THERM and reduce the Power Range Neutron Flux - High Trip Setpoint to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
e .
e 4
MILLSTONE - UNIT 3 3/4 2-15
POWER DISTRIBUTION LIMITS RCS FLOW RATE AND NUCLEAR ENTHALPY RISE' H07 CHANNEL FAC THREE LOOPS OPERATING '
l '
.J.!MITING CONDITION FOR OPERATION _
3.2.3.2 shall The indicated be maintained as follows: Reactor Coolant System (RCS) total flow
- a. Sopro RCS total flow rate 1-3 Qgpm,and b.'
. Fh51.351[1.0+0.43(1.0-P)]
Where:
gy THERMAL POWER ,
p , _ RATED THERMAL POWER
- 2) F =
Measured values of Ffg obtained by using the movable incere detectors to obtain a power distribution map.
ThemeasuredvalueofFhshouldbeusedsinceSpeci-fication 3.2.3.2b. takes inte consideration a measure-ment uncertainty of 4% for incore measurement, and c
- 3) ; ,
The measured since uncertainties value of of RCS total flow rate shall be used _
been included in Speciff ation 3.2.3.2a.for flow measurement have APPLICABILITY: MODE 1.
' 2,07,
. ACSTION:
i With the RCS total. flow rate or Fh outside the region of acceptable operation a.
Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> either:
1.
limits, or the ACS total flow rate and Fh to within the above Restore 2.
Reduce THERMAL POWER to less than 32% of RATED THER and less reduce than or the e Power Range Neutron Flux - High Trip Setpoint to next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. qual to 37% of RATED THERMAL POWER within the MILLSTONE - UNIT 3 3/4 2-18
i
_ LIMITING SAFETY SYSTEM SETTINGS BASES o
Intermediate and Source Rance. Neutron Flux *
- The Intermediate and Source Range, Neutron Flux trips provide fore i protection during reactor startup to mitigate the consequences of an uncon-trolled rod cluster control assembly bank withdrawal from a suberitical condition. l l
of the Power Range, Neutron Flux channels.These trips provide redundant p initiate a Reactor trip at about 105 The Source Range channels will
' when P-6 becomes active. The Intermediate Range counts per second unless manually blocked 3 channels will initiate a l Reactor. trip at a current level equivalent to approximately 25% of RATED
~
THERMAL POWER unless manually blocked when P-10 becomes active.
taken for operation of the trips associated with either the Intermediate orNo credit was
.. Source Range Channels in the accident analyses; however, their functional capability at the specified trip settings is required by this specification ;
to enhance the overall reliability of the Reactor Trip System.
Overtemperature AT {
combinations. of pressure, power, coolant temperature, and
{g tion, provided that the transient is slow with res ,
from the core to the temperature detectors 7 *p y -t -tot piping transit delays
( , and pressure is within the range between the Pressurizer High and point is automatically varied with: (1) coolant temperature to correct for Set- Low Pressure trips. The
= -
temperature induced changes in density and heat capacity of water and includes dynamic compensation for piping dela l
- detectors, (2) pressurizer pressure,ysand from (3)the axial corepower to thedistribution.loop temperature With d
normal . axial power distribution, this Reactor trip limit is always below the 1 core Safety Limit as shown in Figure 2.1-1. If axial peaks are greater than I design, as indicated by the difference between top and bottom power range nuclear }
notationsdetectors, the Reactor trip is automatically reduced according to the in Table 2.2-1. !
~. I Operation Trip System with a reactor coolant loop out of service requires Reactor modification. I Three loop operation is permissible after resettir.g the K1 input to the Overtemperature AT channels, reducing the
. Power Range Neutron Flux High setpoint to a value just above the three loop maximum loop value.permissible power level, and resetting the P-8 setpoint to its three These modifications have been chosen so that, in three loop operation, each component of the Reactor Trip System performs its normal ,
four loop function, prevents operation outside the safety limit curves, and l prevents the ONBR from going below 1.30 during normal operational and antici-pated transients.
Overpower AT !
(.
\
The Overpower AT trip provides assurance of fuel integrity (e.g. , no
~
fuel pellet melting and less than 1% cladding strain) under all possible 4 overpower conditions, ifmits the required range for Overtemperature AT j MILLSTONE - UNIT 3 82-5
)
1
a r .
ep *
- pod a
- Ot e l Ss F s n o p iso poCl 7.'t 70 7
pLsT o2 o e T0 2 L
o&vI5 1.2 2 na l
<~ 1 eov ei R
E rt H h T T
S D
T E
T A
I M
I R
L f o
5
- f i
n a o i
n .F s s so2 p s
., pi e S c 1 R oa 0 x 2
- E T
E M
ot3Mr Lr9 r e1 n#
o2 e n
3 E
L B
A R
A F u
oJ<~1 i%
p1 m
- P af A r
- T B N Rs D_ Es W
Oc
. Px
- L
. An Mi R
E p
, He
, 9 Tt
' a
- R
.. T rE m eW e hO
- t e tP s r i y u eL S s A s gM t e nR
. n r iE rH l
a P uT o r d o e a R C z e E i l r T r r b o a
E o u .
M t s ce A c s i t R a e l u pn A e r pi P R P f am (d
, m .
< . t r i
f'
- ; oe np
- c. ,
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/ pJ iO LP x%4Em9 g*"
"s 4 P %
tl re on pn
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'
- t xn ue l n fo p nm oo E
- rc M d
- t I n d *
- uc
. T o n s s s s ei S c o d d d d nn E E e c n n n n o M S s e o o o o er I N s c c T O e e c c ht P
5 5 e e t c E S
. . s s s e A.0.7 sy 7. A S 0 0 0 fl E A. A. A. 2 2 oe N R N i O N 1 N N A.
P S
i i i 1 N et ms E ir R ti f
N
- e sf O no I
T o
-, pt 2 A su
- T 3 N ep E Rn i
3 M U .r E R go L T B S n A it N t u T I sp M et E t u T
S
- e o Y mr io S tt eec
. P I
st R ne T od p
R x sm O u eo T l rr C F f
,, A h m E n g R , ,
o od x x x r x h i
H re
- u u u u w g f r l l l t
u l
F o i u
F F L H ts F e -
l e pa n n n N n -
- v me p o o o ,
o r e e e em
. i r re re e r r L x r t tt g t T u u ee T u ua tt ua n u A s s r b e eR eR a N e s s e e r N N N R e e e t rl o e e r r r a al t . ,v ,v e e u P P W a c e ei ei t g t T sh T a g gt gt a n a A r r r rs I
N e n ni na a r e e e o R a as ag i
d e r z z z tl U R Ro Re R p e ce l
P e m w i
r i
r i
en L a r r N m e e o u r t n A u e r r c t p s u u ea N n eh eh e r r s s ,
. w wg wg u r s s s dh O a o oi t e e e e e .
c I M P PH PH oi n o v v r r r n T I S O O P P oe C P ,
rh N .
tt U . . . . . u F 1 2 3 0
ef 4 5 6 7 1
8 9 1 1 No 5 rGdE [ g
- y **
DI R * *
- 2 OHUS , ,
MWSI 1 '
1 1 1 1 1 2 1 1
1 1 T
AC UI . . . . .
TG . . . . . .
CO A. A. A. A. A. A. A. A.
AL N N N N N A. A. A.
N N N N N N L
A G N N
' O I I TET ACA )
PUIRT 4 . . . . .
ITVES 1 . . . . .
RCEPE A. A.
S
( A. A. A. A. A. A. A. A.
TADOT R N N N N N T M N N N N N
E M L E A
. R N I O U LI ,)
) )
Q GET ) ) 8 8 E ONA )7 1,
R LNRT .
)
7 1
(
)
7
) 1 11 ) ) 1, AAES 7 ( ( 7 7 7 7 E A.
1 U 1 1 U U9 NHPE ( / ( ( / /(
1
(
1 1 1
., C ACOT M Q S Q
( ( (
N Q S SQ Q Q Q Q Q Q A
L L N I
E O
. I V T , , ,
R LA ))))
U ER ) )
S 4465 5 5 1
NB NI . , , , ,) ) )
. - N AL 23444 4 4
- 3 O HA A. ((((( 4 4
( ( ( (
. I C C_ N OMQRR R R R R R 4 T R R R A
. . E T L N L B E E
. A M NK T U NC .
R AE
- T HH A. A. A.
S CC N S
, N S N M S 5 S S S 5 I
,, M E h T '
. S i Y x x x
x h H S u u u u w g P
l l l l
F o i I
F F F L H l R n n n n - -
T p o o o o e e r re r r r L R
i r tt tt re e g
t T u u O T un t ua tt ua u A s s T ei n n e s s C r eR eR a N e e e A No i N N R r r t
E o p o e e u r a R
t c ee
,t t
p ,v ,v ex e t T P P W ei ei t u g a r r r T a gS e gt gt al n r A I e n S ni na a e e e N R ah as ag d iF R
e p
r z z z U R g w Ro R e en e i i i l i o P N mo e m w r r r
, L a rH L r r rr e o u u u A u e eh c t p s s s N n w wg eh wg et r r r s s s O a o t u u e e e e e I M P a b
. oi Pi oi PH ne o v v r r r
, T l IN S O O P P P C
N U . . . . . .
F 1 2 3 4 5
. . . 0 1 6 7 8 9 1 1 x ? 0 8 F. e c} " R* , 0
. TABLE 4.3-1 (Continued)
TABLE NOTATIONS (Continued)
(10) Setpo' int verification is not applicable.
[ .
-. -(11) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify OPERABILITY Trip Breakers. of the undervoltage and shunt trip attachments cf.the Reactor f'?} ""'" ::L Cationniiva sna r : ,;P;d- +"- etn hunacc Inaa- <!r.. --+-
(13) Reactor Coolant Pump Shaft Speed Sensor may be excluded from CHANNEL CALIBRATION.
(14) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently ve OPERABILITY Reactor of the undervoltage and shunt trip circuits for the Manual Trip Function.
Bypass Breaker trip circuit (s).The test shall also verify the OPERABILITY of the
'(15) Local manual shunt trip prior to placing breaker in service.
(16) Automatic undervoltage trip.
- ~
(17) Each BASIS.channel shall be tested at least every 92 days on a STAGGERED TEST ,
(18) The surveillance frequency and/or MODES specified for these channels in ^'
Table 4.3-2 are more restrictive and, therefore, applicable.
. s 9
e
- i. .
~
O e
O
~%,
r
. MILLSTONE - UNIT 3 3/4 3-14 f
1{ll\\i I '
w ot w E
U rn re *F 'F ot L
am 6 (
, rn re A nu am V r ft d n e (oc nu r E os l L
n a ff ft B %i s ,- ns A t F F n W 8. e n * * %i O .
e t L
. 2gn 8na i
o p 6 .
- 5. e .
L A. ap p S $ . 2gn A N <rs t A. A. ' . 2naap S e 21 N N 1rs T S N T I N p O I e g
i P O f nt r e T P T g E T oan fnt S E re n oan m
- P I
S P
%wu
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o F F
- * %wu rem 4 4 R I 2rt n c 8rsa 6 6 5. or .
T anpej R A. 5 5 3rt n T N i A. A. 2rsa N <i is n anp e
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. 5 7
f .
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5
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d R SE l N N 1 e T l
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r t
n o 8 o N A. 3 f 9
E Z- A. A.
C T N 2 e A. A. 8
( S vs N N N N 1 Y ) oe 4 S A bu
- T al 3 N ( a l O .V 3 I E 1 T C e E A N ml (
.
- L U A eb B T LW ta A C AO I w T A TL o
OL 7 el S A. .
5 E T A- N 3
. el A. A. A. A.
R SA N N N N 0 U 2 T
A E
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i t
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E R d w i ts W(
a h
A n
n )
e 4 d -
e p O n nyt oa a il W i
D r
E e ay rg o O o te E e ua oi i P s -
F tl i c( s p r i aR r r N
I ce tH a- t c n p o e t u a t n t o t ow o G d n AR rh eg e i p o i o o o L t
a i co aL r- o N a j C r L t Ai M E cn ni n T w i t ew i o eH e d n ca no t p ti G-I w o r r u re e I iu eL r i
r at - y e tt a I
T_
Tn ma al ou ce tc L o
,tc o h F T F l ac G- - t N
U ei o tt uc t e ev fg ei ya y r u o mlaeS a mA nt AA SL ao T, R e ) ) a n td ev L i a SL 1 2 i a un t e )
- A bl l N Aa SL 1 N
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T a b c d
u . . ,
C A a b c
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F 5 .
6 5{8x* ' E w ,1 we@
,1 lllj!\jlI !j lilill l!tl{ ! l! {lllllI\ll
l Y s E 6 t U - n L F i A s S M o p
s V t l2m i e t e Erdon F g E t E o l8mh o yS e L v<t o it ncaEn . i s F S .
B v$tioel o p p A 0ad wiseti y W
O 2 ny 0 6ad nya0 t dut oa 5 C. i r
7hoa 9 T L
L 2tcl 7hoau0ehu 3tclt3mtt 9 9) . .
iee 1 9 A ieec l S
>Gsd >ssda<Iwali c 5 >
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T e N T F L I N S O d n . ts r P
I O hoy l8mh e Erdon F g t e
T P oyS E T tca o itncaEn . i s a S E iel wsed v5tioel o p W P
S 0 0ad wiseti t F r
. 0s2 dut 5
- o I
R P
I 8t e 1 nya0 7hoau0ehu oa 8 3 t 2l <~mi 9 5 a T R o 3tcl ieec t3mtt 1 5 r .
T > vat li c A. e A.
N O
>ssda<Iwa $ > N n N I e T G
_ A T R m
. N OR
. . a E SO) A. A.
. . . e
, N NRS M A. A. A.
t
) U ER( N N S A.
d R SE _ N N N e T l u S l i
n N a I
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t . . r . -
n M A. A.
..os .
o E Z_ M M A. A. A. fe C T N N N u A.
( S el va M Y )
4 S A oV
- T b 3 N ( ael O _
3 I E .b T C 5a E
- L B T A
U N
A mow el A C LW tl AO .
T_ A TL .
. OL A. A. .
S TA N A. A. A. ed
- E N N H N en A.
. R Sa N U
T A
~
_ , E ,
_ , F l
Y 1 e
, T 1 v E ) - e F e P L A e g eg sk s
S ga ,
r t
a at ec ro r e e D l) tl ul u 2 t d E oe l o a a R
E vg oV v
tr ae s
s 1
4 W L o
E ra rd et Fn e
r P P r N et dl ee I P o r I
G no dd na y g p
, t a t o
N UV Ur t m r , i r a E r g ee e , r e r sf se ft z T T n e e uo uD as i e n w B Sy r w r G I
T P
o B Vs
's d d S u s
o o G e
Vi L t m N ko kr en s -
c a4 yr U f L G ro e w a o 4( ei r o e e1 ce L 4( et P L R t -
SP nc en A
N s s na iu ge O o . .
gt ru eq T
I L a b nc . . . . me C EA a b c d *- ES
, N U .
F 8 . .
9 0 1 .,
5-irGs5 i ij e
C} " R* Y8
TABLE 3.3-5 (Continued)
ENGINEERED SAFETY FEATURES RESPONSE TIMES 1 INITIATING SIGNAL AND FUNCTION RESPONSE 11HE IN SECONDS l
- 11. Loss of Power
- a. 4 kV Bus Undervoltage ~
< 12 (Loss of Voltage)
- b. 4 kV Emergency Bus 5 18I )/310(8)
Undervoltage (Grid Degraded Voltage)
- 12. T,y, Low Coincident With Reactor Trip (P-4) jg
- a. Feedwater Isolation 5 h3) ;
- 13. Control Building Inlet Ventilation Radiation
- a. Control Building Isolation 5 3.7
- 14. Outside Chlorine High r %
- a. Control Building Isolation 5, 7 4 )
o b
n MILLSTONE - UNIT 3 3/4 3-34
JAN 311986 1 POWER DISTRIBUTION LIMITS BASES l ,' __
HEAT FLUX HOT CHANNEL FACTOR and RCS FLOW RAT HOT CHANNEL FACTOR (Continued) -
c.
The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained; and d.
The axial power distribution, expressed in tems of AXIAL FLUX ;
i DIFFERENCE, is maintained within the limits.
F N will be maintained within its limits provided Conditions a. through 3H
- d. above are maintained. The relaxation of Fh as a function of THERMAL P allows changes in the radial power shape for all permissible rod insertion limits.
The Fh as calculated in Specifi ations 3.2.3.1 influences parameters other than and DNBR,3.2.3.2 a the various accident analyses where Fg (e /h e.g. , peak clad temperature, and thusCredit is theis maximum available to "as mea Fuel rod bowing reduces the value of DNB ratio.The generic margins, totaling ,
this reduction in the generic margin. This margin includes the f
}". NBR completely offset any rod bow penalties.
wing:
- a. Design limit DNBR of 1.30 vs 1.28,
- b. Grid Spacing (X,) of 0.046 vs 0.059,
- c. Thermal Diffusion Coefficient of 0.038 vs 0.059,
- d. DNBR Multiplier of 0.86 vs 0.88, and i
- 04J. .t _ #h at '
- The1applicable /drAvalues W of rod bow penalties are referenced in the FSA ,
e e
em-8 3/4 2-5 MILLSTONE - UNIT 3 .
POWER DISTRIBUTION LIMITS BASES .
_ CHANNEL FACTOR (Continued) HEAT FLUX HOT CHA '
, When an F m q easurement is taken, an allowance for both experiment 1 error and manufacturing tolerance must be made.
3% allowance is appropriate for manufacturing , andtolerance a
, The Radial Peaking Factor, F,y(Z), is measured periodically to provide assurancethattheHotChannelFactor,F(Z),remainswithinitshimit. The F,y i
limit for RATED THERMAL POWER (F TP)9 as provided in the Radial Peaking r Factor Limit Report per Specification 6.9.1.6 was determined from expe power control manuevers over the full range of burnup conditions in the core When RCS flow rate and F
! necessary g arj es are
. Operation. prior to comparison with limits t. easured, of the Limiting Con no additional allog Measurement errors of for four loop flow and on for i .- for three determination of the design DNBR value. loop flow for RCS total flow
.i t
precision indicators. heat balance and using the result to calibrate
- ~
detected could bias the result from the precision heat balanc conservative manner.
the feedwater venturi will be added if venturis are not verifie 18 months.
i -
- than 0.1%
parameters. can be detected by monitoring and trending va If detected precision heat balance me,asurements, i.e. , either the effect o shall be quantified and compensated for in the JtCS flow rate measurement or the venturf shall be cleaned to eliminate the fouling.
The 12-hour periodic surveillance of indicated RCS flow is sufficient to detect only flow degradation which could lead to operation outside the acce able region of operation defined in Specifications 3.2.3.1 and 3.2.3.2.
3/4.2.4 OVADRANT POWER TILT RATIO tion satisfies the design values used in the power capabi Radial power periodically distribution during measurements power operation. are made during STARTUP testing a The limit of 1.02, at which corrective action is required '
and linear heat generation rate protection with x y plane power, ti l provides ts'. A DNB limiting F i tilt of 1.025 can be tolerated before the margin for uncertainty in q s depleted. A limit of 1.02 was selected to provide an allowance for the (i uncertainty associated with the indicated power tilt. ,
MILLSTONE - UNIT 3 8 3/4 2-6
POWERDISTRIBUTIONLIMI]
(
b BASES DUADRANT POWER TILT RATIO (Continued) "
~
The 2-hour time allowance for operation with a tilt condition greater than of a dropped1.02 but less than 1.09 or misa11gned is rod.
control provided to allow identification and correction not correct the tilt, the margin for uncertainty on fin the event such action action does q is reinstated by reducing the maximum a110wed power by 3% for each percent of tilt in excess of 1 . ,
For purposes of monitoring QUADRANT POWER TILT RATIO when one excore the normalized symetric power distribution is consistent POWER TILT RATIO.
flux map or two sets of four symetric thimbles.The incore detector monitorin thimbles is a unique set of eight detector locations.The two sets of four symmetric C-8, E-5, E-11, H-3, H-13, L-5, L-11, N-8. These locations are ,
3/4. 2. 5 DNB PARAMETERS i
The limits on the DNB-related parameters assure that each of the parameters the transient and accident analyses.are maintained within the norinal stea s The limits are consistent with the i
maintain a minimum DNBR of 1.30 throughoutTheeach analyz dr'f ^
an Ts T values v :>:.T:4avg .. a.
tF
'F (three loops operating) and the i loop or three loops"8$ ,2 (four pressure pressurizer ing),6vrru W t loops operating) v;alhgg y or %yheQS N
% [. $ 7;;p.;;i.;!y, rith : 1? "r :: im aelytter' 'frit; ef W.4 . 4 m.. ; :nt r : -* = f-tg .
The 12-hour periodic surveillance of these parameters through instrument I I. readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation. Measure- 1 l
ment uncertainties have been accounted for in determining the parameter limits. (
l 9
- i<
- I 4
MILLSTONE - UNIT 3 8 3/4 2-7
e e
\
1
. APPENDIX B 1
HDT LEG RTD FAILURE COMPENSATION PROCEDURE e
6522e:1d/061887 l
L__ _ _ - _ - - _ _ _ - - - _ - _ - _ - - _ - - - _ - _ - - _ - - - - - - - - - - - - - - - _ - - - _ - - - - - - - - - - - - - - - - _ - - - - - - - - - - - - - - - - - - - - - _ - - - - - - - _ - - - - - - - - - - - - - - . - - - - - - - - - _ - - - - - _ _ ..
DEFINITION OF AN OPERABLE CHANNEL The RTD Bypass Elimination modification uses the average of 3 RTDs in each hot leg to provide a representative temperature measurement. In the event one or more of the RTDs fails steps must be taken to compensate for the Jess of that i- RTD's input to the averaging function. I Single RTD Failure l 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 [
3a,c Once [ Ja,c any hot leg can then tolerate a sin.gle RTD failure and still remain operable. If the situation arises where a single hot leg RTD failure occurs a bias value must be applied to the averaging of the remaining two valid RTDs. [
l t
.. Ja,c No reanalysis will be necessary to
. evaluate this situation. The plant will be allowed to operate for the balance-of the fuel cycle with this single RTD failure in one of the hot legs. If another single RTD subsequently fails in a different hot leg the same bias application methodology will apply.
The plant may operate with a failed hot leg RTD at any power level during that
. same fuel cycle. It is permissible to shutdown and startup during the cycle without requiring that the failed RTD be replaced. [
Ja,C l
6522e:1d/061887 1 l
_ __ - - . - - - - - - - - - - - - - - - - - - -- --- - - - - - ~ - - -
h In order to eliminate any control system concerns, the Tavg and AT signal associated with the loop containing the failed hot leg RTD will be defeated as an input to the control system. This will prevent the control system from ,
using a Tavg or AT at power levels less than 100% which may be offset due to the fixed bias.
If another hot leg RTD fails in a different loop the utility ~
should operate using manual control. Manual control is recommended because only one control channel at a time can be defeated. If automatic operation is continued the control system will most likely auctioneer the biased channel because it will be the highest Tavg due to the positive (or zero) bias application.
This means the control system will perceive a higher Tavg than is real at reduced power and the plant will operate at depresseo tempera-tures.
While this is not necessarily undesirable it does reduce the total plant megawatt output.
on utility power requirements.
The use of automatic control can be considered based Cold Leg:
If the active cold leg RTD fails that RTD should be disconnected
, Double RTD Failure: Inoperable Channel
-. Hot leg or Cold Leg:
If two or more of the three hot leg RTDs or both cold leg RTDs fail in the same protection channel then that channel is considered inoperable and should be placed in trip. Operation with a single valid hot leg RTD is not presently analyzed as part of the licensing basis.
s 6522e:1d/061887 2
_ _ _ - _ _ , - - - - - - - - - ~~~~~~~~ __ _ _____ -- ---- - --- - - - - - - - - - - ~ - - - - - - ~ ~ - ~ - ~ ~
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, 'V PROCEDURE'FOROPERATIONWITHAHOTLEGRTDOUTOF.$gVICE /
.g. _(
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i i, . t ' DThe' hot leg temperature tsasurement is obtained by av(eraginI; the measurements
-a from the three thermowell RTDs installed;N[the hot leg of each lo.op. [-
.,1 :
Ja,c
.a f
?
9 '1 7 sIntheeventthatoneof't'hethreeRTDsfailsl'thefailedRTDwillbediscon-a
.. j nected and the hot' leg, temperature measurement will he obtained by averaging the remaining two RTD measurements plus the bias, [
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/
f p %'
_ -) a . c r.
The bias adjustment' corrects for [ ., .at fu. I r t iced, iduced l \ '9 >
r
/ 4 e is I
,\
ro a
- i. Ja.c To assure that the measured ict leg, temperature j ,
'is maintained at or above the true hot leg temperature, and therefore, to i avoid a reduction in safety margin at reduced power, [ -
ihy , ja.c y
'F An RTD failure will most likely result in an offscale high or low indication ej[ . and will be detected through the normal means in use today (i.e., T andy' AVG e 4 AT deviation alarms).. Althougl unlikely, the RTD (or its electronics e channel) can fail grjdually, dausing a gradual change in the loop temperature /
6 y! d 6522e:1d/061887 3
y i-l3
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r, a'r measurements. [
V 1
. Ja,e The detailed procedure for correcting for a failed hot leg RTD is presented below: ..
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