ML20058E385
| ML20058E385 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 10/26/1990 |
| From: | Morrison J WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML19310C816 | List: |
| References | |
| WCAP-12614, WCAP-12614-R01, WCAP-12614-R1, NUDOCS 9011070149 | |
| Download: ML20058E385 (76) | |
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m HESTINGHOUSE CLASS 3 27,
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'RTD BYPASS ELIMINATION LICENSING-REPORT d
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J.;M. FARLEY, NUCLEAR PLANT UNITS I and 2
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'o 1990; Hestinghouse Electric Corporation, all Rights' Reserved t
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10802D:10/082290:
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. WESTINGHOUSE PROPRIETARY CLASS 3-
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4 This nonproprietary-report bears-a Westinghouse copyright notice.
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. The NRC'is permitted to make-the number-of copies of this report-necessary for,itsLinternal use and such additional-copies which-
. are necessary in order to have ~one copy _ ava11abl.e for_ public "y,"
- viewing-in the appropriate docket files in the public; document room in Washington.10.Ce and in-local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. The NRC is not authorized'to
- make copies-for the personal use of members of-the public who make use of the NRC:public document rooms., Copies of..this.reportLor-
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'T FOREHARD s
'N Extensive.stddles were performed for Farley Units 1 and 2 for-the effects of increased SG Tube. Plugging and, Reduced Thermal Design Flow (NCAP-12694 for Unit:1 HCAP-12659 for Unit 2)..The purpose of_this' report is to show the effects-of RTD bypass elimination on the Farley Units while also considering
'the effects of'the: increased SG Tube Plugging and Reduced Thermal-Design Flow, 1
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ACKNOHLEDGEMENT Th'e' authors wish:to recogni;:eicontribution by the following individuals:
Mike. Emery l Steve Zawalick-j Halt:Tauche<
Fred Baskerville Marvin Hengerd-
. Hally Moomau.
- Dick-Haessler:
Rick Tuley Phil Rosenthal Glen Lang fic Jim Hermigos:
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Mimi Heaver Pete Morris:-
Ron Carlson-4 b
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1 TABLE OF CONTENTS e j q
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List of Tables.
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. List.of.Figur'es
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.140 Introduction'
,kg
.1 1-' Historical Background 1
1.2f MechanicaljModifications' 2
L1;3 Electrical Modifications' 4
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2.04 Testing.
2.1LResponse Time Test 13 f
L 77 2.2;1 Streaming Test ~
13 3
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'3.0 ' Uncertainty; Considerations.
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"o 3.1' Calorimetric Flow Measurement Uncertainty:
16.
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-3.2: 1 Hot Leg Temperature Streaming Uncertainty.
16 4
>3.3 DControliand Protection Function Uncertainties -
19' J
4 IW
.;,4.0. Safety Evaluation
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4.1' Respon's.e' Time 1
' 31 m
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'4.2fRTD_Uncertaintyn
,31 u.m y;
'4.'3 Nori-LOCA. Evaluation ~
32
/.4 MLOCA' Evaluation 41 4
,a' E
j 3 y' 14.5? Instrumentation and; Control SafetyL valuatibni 41:
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7-4.6' Mechanical ~SafetyjEvaluatjon:
45.
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7 F4.7 Technical Specification Evaluation
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! 4 TABLE OF CONTENTS ~(Cont).
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5,0 Control System Evaluation 47 F
6.0'-Conclusions!
48 l
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- 7.0? References
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Appendix A'l-Definition of An-Operable Channel And 50
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Hot-Leg RTD Failure' Compensation Procedure' I
P
, - Appendix B - Definition' of Acronyms'Used in Oncertainty 59
.. Cal culations.
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i LIST OF TABLES j
Table Title P_Asa
- 2 '.1-1
' Response Time Parameters for RCS Temperature Measurement IS' D
1 J
3.1-1 Rod; Control ~ System Accuracy 20 I
3.1-2 Flow Calorimetric Instrumentation Uncertainties 21
'1 3.1-3 Flow' Calorimetric Sensitivities 22 l
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. Calorimetric RCS-. Flow Heasureme.it Uncertainties 23-y i
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<Overtemperature Delta-T Trip 25-j 4
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.1-6 Overpower, Delta-T Trip
'26-3 7.1 lT,yg low-Low Trip:
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' Cold Leg Elbow Tap Flow Uncertainty 28 1
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, Low Flow Reactor Trip-
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.i TechnicaltShecification Hodification
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LIST OF-FIGURES-
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.m Fiaure Title Eage L 1. 2-1'
' Hot'LegRTDlScoopModificationforfast-Response 6
,RTD-Installation' 1.2-2:
Hot leg RTD Boss Installation for Fast-Response:
7 RTD Installation i
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1.2-3 Cold: L'eg Pipe Nozzle Modification Fast-Response 8
A, RTD Installation s.,
L1.2-4
' Crossover Leg Cap Installation 9
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.l.3-1L LRTD Averaging Block Diagram, Typical for Each of 3 10
" Channels q
m J1.3-2 Median Signal Selector
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}Hestinghouse'ElectricCorporationhas:beencontractedby.AlabamaPowerCompany
- (APCO) to remove the; existing Resistanct Temperature Detector (RTD) Bypass
-System and replace this hot leg and cold leg temperature measurement method Lwith fast response thermowell. mounted RTDs installed in the reactor coolant 4
loop piping. This report is submitted for the purpose of supporting operation
(, of-the J. M. Farley Nuclear Plant Units 1 & 2 utilizing the new thermowell
[
mounted RTDs.
/11 HISTORICAL BACKGROUND' g:. m, 4
- h Prior to 1968.- PHR designs hsd been based on-the assumption that the hot leg temperature was. uniform across the pipe. Therefore, placement of the jtemperature instruments was not considered to be a factor affecting the jaccuracyofthemeasurement.-.The hot leg temperature was measured with direct-a.
. immersion RTDs extending'a short distance into the pipe at one location.. By, the late 1960s, asia result of accumulated'operat'ng experience at several i
platits, the-following problems; associated with' direct immersion RTDs-were x
identified:-
x.
L ol ' Temperature: streaming = conditions (the incomplete mixing-of the coolant
- g leaving regions.of the reactor
- core at-different temperatures which
-produces significant temperature gradients.within the pipe).-
co The. reactor coolant loops required cooling and draining'before:the
'RTDs could be' replaced.
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cThe.RTD Bypass System was designed.to resolve:these problems; however, operating plant experience,has now shown that operation with the RTD bypass loops has created its own obstacles such as:
o1 ' Plant shutdowns caused by excessive. primary leakage'through valves,
-flanges, etc., or by interruptions of hypass flow due to valve stem-Efailure.
08020ilD/082290 1
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o Increased radiation-exposure due to maintenance.on the bypass line and
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to crud' traps which increase-radiation exposure throughout the loop f &A compartments.
1 Th'e proposed temperature measurement modification has been developed in response to both sets;of' problems encountered in the past.
Specifically:
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.o Removal? offthe' bypass lines eliminates the components which have' been
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N' a major. source'of plant outages as well as'0ccupational Radiation i
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Exposure l(ORE).
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.. Three thermowell: mounted hot leg:RTDs provide.an average measurement
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~(equivalent to the; temperature measured by the bypass system) to
.acco nu t for temperature streaming.
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'o (Oseofthermowe'11spermitsRTD'replacementwithoutdrainingthe j
reactor' coolant loops.
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(F611 ohing uis aidetailed description of the effort required to perform this
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1.2,MECHANICALMODIFICATIONSl o
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?The? individual $1oop temperature signals required for input to;the Reactor 4
Cont'rolfand{ Protection System will. be onained.using RTDs inst'alled in each l
reactor coolantiloop.-
1
-I 2.'1I Hot Lea
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a) The hot: leg: temperature measurement on.each-loop will be accomplished with.
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'three fast response,l narrow range, dual element RTDs mounted in-
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.thermowells. One element of the RTD will be considered active and the
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'otherEelement will?be held in reserve as a spare.
To accomplish the 1 sampling function of;the RTD bypass manifold system and minimize the need 4
for-additional hot-leg piping penetrations, the thermowells will be H
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ildcated withih the-three; existing RTD' bypass manifold scoops wherever :
- possible."'A hole.will-be made
- through the end of each scoop so that water I'
. ill flow in.throughcthe-existing hoiss.in the leading edge of the scoop, w
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- pastithe RTD, and out through the new hole (Figure 1,2-1). -If plant (interferences preclude the placement of a thermowell in a scoop, then the 1
i
. scoop:will be: capped and a new penetration made to' accommodate the j
thermowell (Figure 1.2-2).
These three RTDs will measure the hot leg l
^ temperature which;is~used to-ca kulate the reactor coolant loop differential temperature'(AT) and average temperature (T,yg).
"+E) This modification will not: affect the single wide range RTD currently.
b
- installed near.the entrance of each steam. generator.
This RTD will-
- 4m continue;to provide the. hot' leg temperature used to monitor reactor j
coolant temperature during startup, shutdown,-and post accident conditions.
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- a) O'ne fast response, narrow range, dual-element'RTD will be located in'each
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cold: leg at the discharge of the'. reactor coolant pump.(as replacements for-
, ;the cold' leg'RTDs~ located in_.the bypass'manifol'd)'. 3 Temperature streaming-a o
inthhcoldlegisnotaconcernduetothemixingactionof<theRCP. ~ For-j y ' Jthis reason; only one RTD is required. This RTD wil_1 measure the cold leg j
4 ltemperai$re which..is used' to' calculate reactor coolant loop AT an'd-q 1T 'vg1 The~ existing: cold l leg RTD bypass-penetration. nozzle!will;be a
. modified 4(Figure 1;2-3) to accept the RTD thermowell. :OneLelement of the-
.f
.RTD willibe considered active-and the other element?will be. held in-4 Lt reserve: asi a spare.
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lL b) 'This modification will not affect the~ single wide rangej!.o in each cold
(,
H1egicurrentlyiinstalled at the discharge of th'e reactoricoolant pump.
ThisiRTD will continue to provide the cold: leg temperaturetused to monitor.
y ks reactor.coolantLtemperatureduringstartup,oshutdown,andpostaccident I,
conditions.
I/
?
[.
s i
y 5
=l,
0802DilD/082290 3
so
ha
- 1.2.3 Crossover Leo The RTD bypass manifoid return line will be capped at the nozz1' M the crossover leg as shown on Figure 1.2-4..
1.3 ELECTRICAL MODIFICATIONS l '. 3.1 Control & Prot 1ction System
['
Figure. 3-1 shows a block diagram of the modified protection system electronics.
The hot leg RTD measurements (thra; per loop) will be electronically-averaged in the process protect 16
/ stem.
The averaged Thot p
signal will then be used with the Tcold signal ts alculate reactor coolant i
loop AT and T,y which are used in the reactor et strol and protection tystem.
This w 11 be accomplished by additions to the existing process
-frotection~ system equipment.
It is planned to wire the T and T hot cold spare RTD elements to the control room and terminate them at the 7300 rack input terminals. This arrangement will allow on-line accessibility to the
, spare elements for RTD cross calibrations and to facilitate connection of the spare RTD element in the event of an RTD element failure.
The present RCS loop. temperature measurement system ures dedicated' direct p
immersion RTDs for the control systems. This was done largely to satisfy the IEEE Standard 279-1971 which applied single failure criteria to control and protection system interaction.
The new thermowell mounted RTDs will be used for both control'and protection.
In order to continue to satisfy the
. requirements of IEEE Standard 279-1971, the T,yg and AT signals generated in the protection system will be electrically isolated and transmitted to the E
N control system into Median Signal Selectors for T,yg and AT, which will select the signal-which is in between the highest and lowest values of the-three loop inputs. This will preclude an unwarranted control system response that could be caused by a single signal failure.
7 L
L
.08020:10/082290 4
t.
.1
c 1.3.2 Qualification The 7300 Process System Electronics modifications will be qualified to the f
seme level as the existing 7300 electronics.
RTD qualification will be verified to support APCO's compliance to 10CFR50.49.
l The Westinghouse qualification program entailed a review of th? HEED Instrument Comp:ny's qualification documentation for testing performed on i
these RTDs.
It was concluded that the equipment's qualification was in compliance with IEEE Standards 344-1975 and 323-1974 with one exception.
Specifically, requirements relative to flow induced vibration were not j
I addressed. To demonstrate that flow induced vibration would not result in significant aging mechanisms that could cause common mode concerns during a j
seismic event, Westinghouse performed flow induced vibration tests followed by pipe vibrction aging and a simulated seismic event. These tests confirmed that the HEED RTDs do comply with'the above IEEE standards.
<1.3.3 JLTD Ooerability Indication i
I Existing control' board AT and'T,yg indicators and alarms will provide the means of identifying RTD failures, although the now redundant indication for the~T,yg and AT control signals will be reinoved.
The spare cold leg RTD element provides sufficient spare capacity to accommodate a single; cold leg l
LRTD failure per ' cop.
Failure of a hot leg RTD can be handled in two ways.
The first method disconnects the failed element and utilizes the second
}
element of~the same RTD.
In the second, manual action initiated by the i
operator defeats the failed signal and rescales the electronics'to average the remaining signals.
08020:10/082290 5
f
.,--4
- ~. -.
4
a.,C.
l l
k t
i i
t l
t.
l
1, l
Figure 1.2-1 Hot _ Leg RTD Scoop Modificatien
.for fast Response RTD Installation 08020:10/082290.
6
)
i A,C I i
I i
l
.1 1
i 1
1
'h e
i 1
l l
l Figure 1.2-2 Cold Leg Pipe Nozzle Modification for Fast Response RTD Installation 0802D:10/082290 7
a.C. '
~
i I
ll l
i l
i y
Figure 1.2-3 -Cold Leg Pipe Nozzle Modification fast Response RTO Installation 08020:10/082290 8
i 4
i i
t AgC i
F I
i l
i r
i t
i I.
l:
l 1"
.i l
F figure 1.2-4 Crossover Leg Cap Installation
+
i
?
.0802D:10/082290 9
1:-
f I
i i
A,c.
1 I
i F
i' i
s a
Figure 1,3-1 RTD Averaging Block Diagram Typical for Each of 3 Protection Channels
-i 08020:1D/082290-10
,... =. -.,, _ < ~
1
)
1 NC-J r
t l
1 4
i i
f Figure 1.3-2 Median Signal Selector Block Diagram
- 0802D:1D/082290 11
l 7
i F
0.st.
i I
i
?
)
t Figure 1.3-3 -RTD Bypass Elimination Control System Schematic P
0802D:10/082290 12 9
a w
w s.
wr--
i 2.0 TESTING l
There are two specific types of tests which are performed to support the
)
installation of the thermowell mounted fast-response RTDs in the reactor coolant piping:
RTD response time tests and a hot leg temperature streaming test.
The response time for the Farley Units 1 & 2 application will be verified by testing at the RTD manufacturer and by in-situ testing.
Data from thermowell/RTD performance tests at operating plants provide additional support for the system by confirmation of RTD/thermowell response times and by confirmation of the magnitude of temperature streaming.
2.1 RESPONSE TIME TEST I
i The RTD manufacturer, WEED Instruments Inc., will perform time response testing of each RTD and thermowell prior to installation at the farley Units 1
& 2.
These RTD/thermowells must exhibit a response time bounded by the values shown in Table 2.1-1.
The revised response time has been factored into the transient analyses discussed in Section 4.0.
1 In addition, response time testing of the HEED RTDs will be performed in-situ.
This testing will demonstrate that the HEED RTDs can satisfy the response time requirement when installed in the plant.
2.2 STREAMING TEST-t l
Past testing at Westinghouse.PHRs has established that temperature
~
stratification exists in the hot leg pipe with a temperature gradient from minimum to maximum of (
]b,c.e.
A test program was *mplemented at l
an operating plant to confirm the-temperature streaming magnitude and stability.with measurements of the RTD bypass branch line temperatures on two.
adjacent hot' leg pipes.
Specifically, it was intended.to determine the f
magnitude of the differences between branch line. temperatures, confirm the short-term and long-term stability of the temperature' streaming patterns and evaluate the impact on the indicated temperature.if only 2 of the 3 branch line' temperatures are used to determine an average temperature.
This plant specific data is used-in conjunction with data taken from other Westinghouse designed plants to determine an appropriate temperature error'for use in the 1
.5 1
L i
0802D:10/082290 13
,w.w
~
safety analysis and calorimetric flow calculations.
Section 3 will discuss the specifics of these uncertainty considerations.
I The test data was reduced and characterized to answer the three objectives of l
the test program.
First, it is conservative to state that the streaming 3 'C.
Steady state data taken at D
pattern [
100'f. power for a period of four months indicated that the streaming pattern
[
]b,c.e.
In other words, the temperature bl,c.e.
This 5
gradient (
b3,c.e is inferred by [
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 g
.to obtain an average hot leg temperature.
The operator can review telaperatures recorded prior to the RTD failure and deterniine an
[
]b,c.e into the "two RTD" average to obtain tLe "three RTD" expected reading.
A generic procedure has been provided to A C which specifies how these [
]b,c.e are to be determined (Appendix A).
This significantly reduces the error introduced by a failed RTD.
Both the test data and the operating data support previous calculations of streaming errors determined from tests at other Hestinghouse plants.
The-temperature gradients defined by the recent plant operating data are well within the upper bound temperature gradients that' characterize the previous L
test data.
Differences observed in the operating data compared with the
[
previous test data indicate that the temperature gradients'are smaller, so the teasurementuncethaintiesareconservative.
The measurements at the operating l
[
' plants; obtained from thermowell RTDs. installed inside the bypass scoops, were-l expected to be.-and were found to be, consistent with the measurers.ts cbtained~previously from the bypass loop RTDs.
L 1
l 08020:10/0822903 14 t
m
.. ~.
F TABLE 2.1-1 i
RESPONSE TIME PARAMETERS FOR RCS TEMPERATURE MEASUREMENT t
RTD Fast Response Bvoass System Thermowell RTD System 1
~~
~'
~~
~
RTD Bypass Piping and Thermal Lag (sec)
RTD Response Time (sec)
Electronics Delay (sec)
Total Response Time (sec) 6.0 sec 6.0 set s
T l
t t
i I
0802D:10/082290 15 i
3.0. UNCERTAINTY CONSIDERATIONS I
This method of hot leg temperature measurement has been analyzed to determine i
the magnitude of the two uncertainties included in the Safety Analysis:
J
. Calorimetric Flow Measurement Uncertainty and Hot Leg Temperature Streaming Uncertainty.
Tables 3.1-1 through 3.1.10 were generved specifically for APCO and reflect plant specific measurement uncertainties and operating conditions.
3.1 CALORIMETRIC FLOW MEASUREMENT UNCERTAINTY Reactor coolant flow is verified with a calorimetric measurement performed after.the return to power operation following a refueling shutdown.
The two most important instrument parameters for the calorimetric measurement of RCS flow are the narrow range hot leg and cold leg coolant temperatures.
The accuracy of the RTDs has, therefore, a major impact on the accuracy of the flow measurement.
Hith the use of three T RTDs (resulting from the elimination of the RTD hot bypass lines) and the recommendations of the Hestinghouse RTD cross-cal.ibration procedure (resulting in low RTD calibration uncertainties at
~
the beginning of a fuel cycle), the Farley Units 1 & 2 RCS Flow Calorimetric a
uncertainty is estimated'to_be [
J.c including use of cold leg
= elbow taps'(see Tables 3.1-2, 3, 4 and 8). This estimate is based on the standard Westinghouse methodology previously approved on earlier submittals of other plants associat6d with RTD Bypass Elimination or the use of the.
'Hestinghouse Improved 1hermal: Design Procedure.
l
'3.2, HOT LEG TEMPERATURE STREAMING UNCERTAINTY The safety analyses incorporate an uncertainty to account for.the difference f
between the actual hot-leg' temperature and the measured hot leg temperature
- caused by the incomplete mixing of~ coolant leaving regions of the reactor core at:~different temperatures.
This temperature streaming uncertainty is based on an analysis of1 test data from other Hestinghouse plants, and on~ calculations l
0802Di1D/082290 16
I 3
to evaluate the impact 'on temperature measurement accuracy of numerous possible temperature distributions within the hot leg pipe.
The test data has shown that the circumferential temperature variation is no more than (
)b c.e,
and that the inferred temperature gradient within the pipe is limited to about
(
)b,c.e.
The calculations for numerous temperature l
distributions have shown that, even with margins applied to the observed temperature gradients ~, the three-point temperature measurement (scoops or
[
thermowell RTDs) is very effective in determining the average hot leg temperature.
The most recent calculations for the thermowell RTD system have
' established an overall streaming uncertainty of (
]b c.e for a hot leg measurement. Of this total.
[
E
]b,c.e.
This overall temperature streaming uncertainty determined for plants with similar:or symmetrical temperature distributions is conservative when applied to 3 loop plants such as Farley Units 1 & 2 since the 3 loop temperature distributions are not symmetrical.
This non-symmetric distribution results in a smaller systematic uncertainty for 3 loop plants..
t The new method of measuring hot leg temperatures, with the three hot leg lthermowell RTDs, is. at' least as effective as the' existing RTD bypass system.
[
Ja,c 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 l
' measurement if the tempsrature gradient over the 5-inch ' scoop' span is nonlinear. Assuming 'that the maximum inferred temoerature gradient of-[
i
]b,c.e exists.from the center to the end of the scoop, the i
L difference betweenithe thermowell and scoop measurement is limited.to (
.)b c.e Since three RTD measurements are averaged, and the nonlinearities l
.at each scoop are random, the effect of this error on the hot t
1 08020:10/082290 17 l
i t
bl.c.e.
On the other l
leg temperature measurement is limited to [
hand, imbalanced scoop flows can introduce temperature measurement uncertainties of up to
[
3"'C.
In all cases, the scoop b
J.c.e sampling flow imbalance uncertainty will ' qual or exceed the [
uncertainty for the thermowell RTDs, so the new measurement system tends to be a more accurate measurement with respect to streaming uncertainties.
i Temperature streaming measurements have been obtained from tests at 2, 3 and 4-loop plants and from thermowell RTD installations at 4-loop plants.
Although there have-been some differences observed in the orientation of the individual loop temperature distributions from plant to plant, the magnitude of the differences have been
[
b 3,c.e.,
Over the testing and operating periods, there were only minor variations of b
less than [ l.c.e in the temperature differentials between scoops, and smaller variations in the average value of the temperature differentials.
L b3.c.e, Provisions were made in the RTD electronics for operation with only two hot leg RTDs in service.
The two-RTD measurement will be biased to correct for the diffeience compared with the.three-RTD average. Based on test data, the b
3 c.e Data bias value would be expected to range between [
comparisons show that.the magnitude of this bias varied less than [
b3,c.e over the test period.
In addition, the uncertainty calculations assumed that two T RTD's were utilized to determine T Appendix A
~ hot hot.
provides a procedure for utilizing the actual plant bias data. Note that this L
procedure only allows'the use of positive (or zero) bias values, i
08020:1D/082290 18 l.I
3.3 CONTR0t AND PROTECTION FUNCTION UNCERTAINTIES Calculations were performed to determine or verify the instrument uncertainties for the control and protection functions affected by the RTD Bypass-Elimination. Methodology for these calculations has been accepted in Reference 6.
Table 3.1-1 (Rod Control System Accuracy) notes that an acceptable value for control is calculated.
Table 3.1-2, 3.1-3 and 3.1-4 provide the uncertainties, sensitivities and final result of the Precision RCS Flow Calorimetric.
Table 3.1-5 provides the uncertainty breakdown for Overtemperature AT. As noted on this table, TA is greater than CSA, thus acceptable results are calculated for this function.
Table 3.1-6 provides the breakdown for Overpower AT, with the same conclusions as for Overtemperature AT. Table 3.1-7 notes the uncertainty breakdown for Tavg Low-Low. Again acceptable results are calculated.
Table 3.1-9 is concerned with the RCS Low Flow reactor trip. Based on the earlier calculations for the RCS Flow Calorimetric and the Rod Control System Accuracy, acceptable results are determined.
Finally Table 3.1-10 notes the changes necessary to the J. M. Farley Nuclear Plant Units 1 & 2 Technical Specifications.
As noted, relatively minor changes are necessary to reflect the modified calculation results, primarily the Allowable Values. Appendix B contains a listing of acronyms for those used in the uncertainty calculations.
i 1
h; 1
I f
08020:10/081290 19
TABLE 3.1-1 l
ROD CONTROL SYSTEM ACCURACY i
Tavg TURB PRES *
+a.c p
SCA =
t SMTE.
STE -.
r SD
=
P BIAS.
RCA -
RMTE.
RMTE.
RTE -
RD
=
' CA BIAS.
NO. RTDs USED TH = 2 TC - 1
~"
ELECTRONICS CSA
=
. ELECTRONICS SIGMA -
i CONTROLLER SIGMA'.
CONTROLLER BIAS CONTROLLER CSA
-l 0
sa 0802D:lD/082290 20 i
~
TABLE 3.1-2 i
FLOH CALORIMETRIC INSTRUMENTATION UNCERTAINTIES I
(% SPAN)
FH TEMP FH PRES FH DP STM PRESS TH TC PRZ PRESS
)
+a c l
1 SCA =
SMTE-SPE -
STE -
SD
)
R/E -
RDOT.
' BIAS-CSA =
NO. OF INST USED 3
1 2
DEG F PSIA
% DP PSIA DEG F DEG F PSIA INST SPAN = S00 2000 120 1200 120 120 800
+8,C
' INST UNC.
(RANDOM).
INST UNC.
(BIAS) l NOMINAL l
l These calculations were performed assuming-that l
t l
l-r 08020:1D/082290 21
i i
TABLE 3.1-3 i
i FLOW CALORIMETRIC SENSITIVITIES FEEDWATER FLOW Fa
+a,c TEMPERATURE j
MATERIAL
=
DENSITY l
TEMPERATURE PRESSURE DELTA P i
FEEDWATER ENTHALPY TEMPERATURE PRESSURE h5
- 1199.9 BTU /LBM 416.4 BTU /LBM hF
=
Dh(SG) 783.5 BTU /LBM STEAM ENTHALPY
+R,c
~
PRESSURE
=
M0ISTURE HOT LEG ENTHALPY TEMPERATURE:-
PRESSURE-hH-
- 629.7 BTU /LBM hC
- 538.6 BTU /LBM 91.1 BTU /LBM Dh(VESS)
Cp(TH)-
- 1.495 BTU /LBM-DEGF COLD LEG ENTHALP)
+a c TEMPERATURE PRESSURE Cp(TC)
= 1.227 BTU /LBM-DEGF COLD LEG SPECIFIC VOLUME
~
~
TEMPERATURE
=
PRESSURE
'08020:1D/082290 22
l TABLE 3.1-4
- )
i CALORIMETRIC RCS FLOW MEASUREMENT UNCERTAINTIES l
COMPONENT INSTRUMENT ERROR FLOW UNCERTAINTY
(% FLOH)
- +6,C FEEDWATER FLOW l
VENTURI THERMAL EXPANSION COEFFICIENT TEMPERATURE MATERIAL DENSITY TEMPERATURE PRESSURE DELTA P FEEDHATER ENTHALPY TEMPERATURE PRESSURE STEAM ENTHALPY PRESSURE MOISTURE NET PUMP HEAT ADDITION l
HOT LEG ENTHALPY l
TEMPERATURE I.
STREAMING, RANDOM
..i STREAMING. SYSTEMATIC' PRESSURE
~
COLD LEG ENTHALPY i
-TEMPERATURE l
PRESSURE COLD LEG SPECIFIC VOLUME TEMPERATURE PRESSURE t
.08020:1D/082290 23
TABLE 3.1-4 (continued)
CALOR! METRIC RCS FLOW MEASUREMENT UNCERTAINTIES BIAS VALUES
+a'c
~~
~~'
FEEDWATER PRESSURE DENSITY l
ENTHALPY J
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 (HITHOUT BIAS VALUES)
N LOOP UNCERTAINTY (HITHOUT BIAS VALUES)
N LOOP UNCERTAINTY (HITH BIAS VALUES) n P
l c
E I
l i
1 08020:10/082290
'4
TABLE 3.1-5 OVERTEMPERATURE DELTA-T TRIP DELTA-T Tavg PRESS DELTA-I
+h.c PMA
=
SCA SMTE -
STE SD
=
J BIAS.
j RCA RMTE -
RMTE =
RCSA -
RTE f
RD SA
=
NO. OF RTD USED TH = 2 TC - 1 102.3 DEGF INSTRUMENT SPAN
~
+a,C
~
SAFETY ANALYSIS LIMIT (SAL)
=
3.35% DELTA-T SPAN ALLOWABLE VALUE t
1.1800 K3 0.000635 NOMINAL SETPOINTS K1 VESSEL DELTA-T
= 68.2 DEGF DELTA-I GAIN = 1.75
+a,c PRESSURE GAIN-
-+a.c
+a c
+a,c T
S Z
L TA CSA =
MAR L
t 08020:10/082290 25
TABLE 3.1-6 OVERPOWER DELTA-T TRIP DELTA-T Tavg
+a.C PMA SCA
=
SD BIAS -
RCA RMTE -
RHTE =
t RCSA -
RTE -
l RD i
NO OF'RTD USED TH 2
TC - 1 INSTRUMENT SPAN
- 102.3 DEGF
+8.C SAFETY ANALYSIS LIMIT 2.93% DELTA-T SPAN ALL0HABLE VALUE 1.0800 NOMINAL SETPOINT VESSEL DELTA-T
= 68.2 DEGF l-
+4,C
+R.C
+8,C T
S Z
=
CSA =
MAR TA I
P 08020:1D/082290 26
TABLE 3.1-7 T&vg Low-Low TRIP
+8,C
~
PHA
=
SCA
=
SD BIAS =
RCA RMTE -
RCSA =
RTE RD NO. OF R1D USED TH = 2 TC - 1 100.0 DEGF INSTRUMENT SPAN
=
+8,C SAFETY ANALYSIS LIMIT
=[
]
540.2 DEGF ALLOHABLE VALUE 543.0 DEGF NOMINAL TRIP SETPOINT
=
+8,C
_ &,C
+ 4, C
+
S T-Z CSA =
MAR ' -
TA 0
0802D:1D/082290 27
TABLE 3.1-8 COLD LEG ELB0H TAP FLOW UNCERTAINTY INSTRUMENT UNCERTAINTIES
+8,C "O
SPAN
% FLOW PHA PEA SCA SPE STE SD RCA RHTE -
RTE RD ID A/D RDOT -
FLOH CALORIM. BIAS FLOW CALORIMETRIC INSTRUMENT SPAN
+8 C SINGLE LOOP ELB0H TAP FLOW UNC -
% FLOH N LOOP ELB0H TAP FLOW UNC N LOOP RCS FLOH UNCERTAINTY
+8,C (HITH BIAS VALUES) 08020:10/082290 28
TABLE 3.1-9 LOW FLON REACTOR TRIP INSTRUMENT UNCERTAINTIES
+8.C 1 DP SPAN
% FLOH SPAN 7
PMA1 -
PMA2 -
PEA SCA SPE STE SD BIASF.
BIASI.
BIAS 2-RCA RMTE -
RCSA -
RTE RD BIAS =
120.0 % FLOH FLOW SPAN
+&,C SAFETY ANALYSIS LIMIT
=[
]
88.5 % FLOW ALLOHABLE VALUE NOMINAL TRIP SETPOINT
= 90.0 % FLOH
~
+8.C
+8.C
+8.C T
S Z
=
CSA.
MAR TA 0802D:10/082290 29
TABLE 3.1-10
~
TECHNICAL SPECIFICATION MODIFICATIONS Overtemperature AT K)
- 1.18 Z
- 4.00 S
- 1.47(AT) + 0.64(pressure)
Allo' ble Value 1 3.4% AT span Response Time 1 6 sec A$ penalty
- 1.75%
r Overpower AT Z
- 1.10 S
- 1.47 Allowable Value 1 2.9% AT span Loss of Flow Z
- 1.71 S
- 0.60 Allowable Value 1 88.5% of Loop Design Flow DNB Parameters s,
RCS Tavg - 581.5'F RCS Total Flow Rate 2 267,400 gpm Tavg Low-Low
~
Allowable Valve - 540*F P-12 1 547'F (Increasing) 1 540*F (Decreasing)
- Includes approximately 1.5% TDF Reduction and 2.2% Increase for Uncertainty.
's w
-e 08020:10/100990 30 1
4.0 SAFETY EVALUAl108 The primary impact of the RTD Bypass Elimination on the FSAR Chapter 15 (Reference 1) safety analyses are the differences in response time characteristics and instrumentation uncertainties associated with the fast response thermowell RTD system.
The effects of these difference > are discussed in the following sections.
4.1 RESPONSE TIME The response time parameters of the J. M. Farley Nuclear Plant Units 1 & 2 RTD Bypass System assumed in the safety analyses are shown in Table 2.1-1.
For the fast response thermowell RTD system, the overall response time will consist of [
J c (as presented in Section 2.1 and as given in Table 2.1-1).
a The new thetmowell mounted RTDs have a response time equal to or faster than the maximum allowed time for the old bypass piping transport, thermal lag and direct immersion RTD.
This response time is factored into the Overtemperature AT trip performance.
Therefore, those transients that rely on the above mentioned ': rip must be evaluated for the modified response characteristics.
Section 4.3 includes a disc'ssion of the evaluations performed for these events.
2 4.2 RTD UNCERTAINTY The proposed fast response thermowell RTD system will make use of RTDs, manufactured by Heed Instruments Inc., with a total uncertainty of [
J.c assumed for the analyses.
a The FSAR analyses make explicit allowances for instrumentation errors for some of the reactor protection system setpoints.
In addition, allowances are made for the average reactor coolant system (RCS) temperature, pressure and power.
These allowances are explicitly applied to the initial conditions for the transients.
08020:10/100990 31 iii. iii i
s u
s i
ii ni iiii
..sii
.i.
ii..iv n-i
The following protection and control _ system parameters were evaluated (with
- respect to_ accident' analysis assumptions) for the change from one hot leg RTD to three hot leg RTDs:
the Overtemperature AT (OTAT), Overpower AT
(
(OPAT),'and Low RCS Flow' reactor trip functions; RCS loop T,yg o
- measurements used for input to the rod control system, steam dump system,
[
.feedwater isole. tion, steam line isolation, safety injection; and '.he l
k calculat'ed 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 RIDS.
The results of these calculations, noted in 3.3, indicate _ sufficient margin exists to account for known' instrument l
' uncertainties for all of the above except the rod control system accuracies and the' low RCS flow reactor trip.
Therefore, these item; are addressed in 1
Section 4.3.1 and 4.3.2.
L 4.3-NON-LOCA EVALUATION s
1
? As discussed in HCAP-12659 (Reference 7) and HCAP-12694 (Reference 8), the
[
. evaluations! resented in this section have. conservatively considered an P
I operating' configuration of 151. average steam generator. tube plugging,.with a-maximum plugging level in one steam generator of 201,and with an analyzed 4
- minimum' average thermal design flow of'87200 gpm/ loop.
The evaluation results j
'are' applicable to this level of tube' plugging or any lower level.
l M
lThe;RTD; response time discussed in Section 2.1 and the instrumentation 1
e'
,) uncertainties calculated inL Section 3-~5 have been considered for the J. M.
l un-t n]Farley Ndclear Plan' ' ton-LOCA safety analysis design basis. Only those'
- I g 3 ( transients wh_ich ' assume OTAT, protection'are p o ntially affected byichanges lin{the RTD response time. As'noted in Sectior. 1.l','the'new thermowellumountr'
.[ w IRTDs.hNeTresponseitimeequalto'orbetterthantheoldbypass-piping m-
%- w transport Lthermal_' lag ~and direct immersion RTD. On'the basis offthe j
gj[ihformOlondocumentedinTable2.1-1,'it'isconcludedthatthesafety-A s assumption for total 0 TAT channel response time of 6.0 seconds.
i gg ant i
" remainsivalid.
Evaluation of theleffects of the RTD Bypass Elimination on the j
[p ;Jun'certaint'ies associated;with-these setpoints supports'thetcontinuing valid 1+y fi lof the; non-LOCA safety analyses (References 1, 7 Er.d 8),
a pf p
i
)g I
4 iT.
l b !08020:1D/100990.
32 y
kt 4
Instrumentation uncertainties can affect the non-LOCA transient initial
. condition assumptions and those transients which assume protection from low primary coolant flow reactor trip These effects are discussed in the
_following sections.
4.3 1 EFFECTS OF ROD CONTi... SYSTEM ERRORS As' noted in Section 3.0, the RTD Bypass Elimination affects the Rod Control System accuracies. These accuracies affect the initial RCS Tavg assumed in.
1 i
-the non-LOCA' safety analyses. The current analysis assumptions are based on a
. 14*F allowance as discussed in the J. H. Farley Nuclear Plant Units 1 and 2 q
FSAR, Section'15.1.2.
For the RTD Bypass Elimination, the allowance is
. increased to 14.3'F.
o' The~ initial Tavg assumed for the non-LOCA transients initiated from full or i
~
~
- partial power includes an error' allowance for the' rod control system. ' The 3 allowance is'not assumed for transients initiated from zero power conditions.-
l c
!Therefore, the:following zero power transients:are not affected by the
- incrme-in, th' Tavg= allowance from 14'F to 14.3*F
- -
t RCCA Bank Hithdrawal.from a Suberitical-Condttion ((15,2'.1) and the new
- analysis presented in References 7 and 8).
ExcessiveL Heat Remew1, due' to Feedwater System Halfunctions [zero power
,]>
case) (15.2510)*
t y
1 Accidental!Depressurization of the Main Steam System (15.2.13)
'RCCA Ejection [zero power cases] (15.4.6)-
7 b7 '
Rupture of a~ Main Steam Line (15.4.2.1)
~ '
l I
D iThe-conclusions of these analyses as well as the conclusions in Reference 7
~
f Land Reference 8 remain valid.
1 I,;
y j
L l ?.
- 0802D:10/082290; 33 Ih
-4*..
. 6 a
4,
<.z-- ;
For' transients analyzed to confirm that the DNB design basis is met, generic y
.' plant.DNB margin has been allocated to offset the DNB penalty of the additional 0.3*F in the initial Tavg.
Therefore, the conclusions of the
'following DNB transients remain valid:
Uncontrolled RCCA Bank Hithdrawal at Power (FSAR Section 15.2.2)
-RCCA Misalignment (15.2.3)
. Partial Loss-of Forced' Reactor Coolant Flow (15.2.5, and the raw analysis
. presented in References 7 and 8)
Startup of an Inactive Reactc,r Coolant Loop-(15.2.6)
Loss of External' Electrical Load (15.2.7)
_j
-Excessive Heat Removal Due to Feedwater System Malfunctions (15.2.10)
_Excessiv'e Load' Increase Incident'(15.2.11) l'
-Accidental Depressurization of the RCS (15.2.12)
Inadvertent Operation:of. ECCS During Power Operation (15.2.14)
[
CompleteiLoss of Forced Reactor Coolant flow (15.3.0
?
- 9 fA number of no'n-LOCA transients are analyzed to demonstrate acceptability for f,
criteria other than DNB. ' A discussion of the effects of increasing the:Tavg
- ve
. ~
I, uncertainty ~by 0.3*Fgfor these transients follows.
N b
-Uncon' trol 1edLBoron'D1lution(15~2.4) y, 1The tioronidilution event:is ~an uncontrolled addition of unborated-reactor SL
- makeup! water.idtoithe RCS via the Chemical and Volume Control System.
The f :. -
boron" dilution event is analyzed to demonstrate that, prior to total loss of l
{> :
shutdown margin,'there is sufficient operator action time available to u
!.t f i
+%
1
[
08020:10/082290 34
recognize the~ event and terminate the dilution. The increased temperature uncertainty does not change the critical parameters assumed in the analysis:
~
the maximum dilution rate, RCS boron concentrations, or the dilution volume for any of the operational modes. Therefore, the current boron dilution
=
analysis and the evaluation of this event in References 7 and 8 remain valid.
Loss of External Electrical Load (15.2.7)
The loss of external electrical load is a complete loss of steam load from full power without a direct reactor trip.. Four cases are analyzed which are g
L based on two different primary side pressure control strategies (automatic and no mitigating control) and two sets of core physics characteristics (minimum Landinaximum reactivity feedback). The key acceptance criteria for this trNsient besides DNB are primary and secondary pressures remaining below 110%
of. design. The RCS. design pressure is 2485 psig (2500 psia) and the steat
. generator design. pressure is 1085 psig (1100 psia); As shown in F5AR Figures 15.2-19 throughil5.2-26, the peak pressurizer pressure for all of the cases all. remains belowl110%.of RCS Design Pressure. The peak calculated RCS and secondary side pressures are.not sensitive to the' initial temperature Lassumed.: Considering;the-temperature increase is only 0.3*F and given the
~ margin-in the current analysis, it is concluded that the increase in
~ temperature uncertainty will not change the conclusions of the FSAR or the
- conclusions of References'7 and 8.
'LospofNormal(Feedwater(15.2.8)
U
'The loss' of; normal foedwater is the simultaneousiloss of feedwater flow to all
[
threensteam generators. The FSAR analysis assumes that the reactor coolant pumps coastdown<duelto an assumed loss'of offsite: power.. As stated in the
- FSAR.;this event-is analyzid ts. demonstrate that the pressurizer does not
.become water solid'during the transient. The analysis-assumes a power level o
' corresponding to th'e-102%'of:thv engineered safeguards power rating (a conservative assumption' because Farley-Units 1 Pr.d 2 are not licensed to operate-at engineered. safeguard power). The initial Tavg is assumed to be the engineeredsafeguardsTavgminus)*F.
The temperature uncertainty is subtracted from the nominal Tavg-for this analysis because a lower temperature m,-
0802D:1D/082290 35
L results-in'more initial RCS mass..The larger RCS mass is more conservative when verifying that the pressurizer does not fill.
For a change of 0.3'F,
-however, this'effect is small.
FSAR Figure 15.2-27 shows that the peak pressurizer water level is less than
.1200 cubic. feet.
The pressurizer hts an internal volume of 1400 cubic feet.
The;resulting margin to pressurizer filling has been compared to the effects
, of a 0.3*F change in Tavg and increased steam generator tube plugging (Reference-7).
The analysis margin is sufficient to' accommodate the effects f
.of both changes ~ and still maintain margin for filling the pressurizer.
It
!'4
.should be noted that the decay heat model used in the current analysis is 1
base'd on the ANS-1971 decay heat model. Additional analysis margin would be gained if the ANS-1979 model was used because the total energy released into she RCS is lower. Therefore, the conclusion that the pressurizer does not e
?
fill: for this event remains valid, u e a
Loss of All: AC' Power to the Station Auxiliaries (15.2.9) i This event represents a complete loss of power to the plant au'xiiiaries (i.e.,
1 i
the reactor coolant pumps, feedwater pumps, condensate. pumps, etc.).
T' a 4
concluilons section in-the FSAR states that the loss of. forced flow and loss a
6,
- ofe normal Lfeedwater results show that the acceptance criteria will be met for
?
.this transient.: Both of,these transients have been. evaluated.and found W
Lacceptable;-therefore, the conclusions of'theLFSAR and References 7 and 8 L,
remainivalid; L
L RSingle'RCCA-Hithdrawal at Full Power (15.3.6):
c w
I lllT t
3: This event! represents the accidental withdrawal of a single RCCA from the ye
" inserted bank atLfull power operation. An evaluation for this-transient was.
.]
k '
+ performed for the'_ increased-Tavg uncertainty and it was demonstrated that;the 3
Ea. \\ conclusions in the FSAR.(i.e., lessLthan~ 5% of the fuel ' rods ar.e below the DNB~~
,[
11'mit'value) and-References.7 and 8 remain valid.
4
? n ' 1;
~
a
'Y l{
i7
!!08020:1D/082290:
36-iw$Y L -
}
J
[
Major: Rupture of a Hain Feedwater Pipe (15.4.2.2) n i
The. rupture of ~ a' main feedwater pipe is a break in the feedwater pipe-large' enough to prevent the addition of feedwater to the steam generhtors. 'There.
H are two' cases for feedline break presented in the FSAR.
The primary
- . acceptance criterton in the FSAR is' that-the core remains covered.
Case A was
' analyzed assuming the initial temperature was 6.5'F above the engineered l
f isafeguards Tavg.
Therefore, the current analysis bounds the increase in Tavg uncertainty, n.
[,
' Case B assumed that the initial temperature was.4*F above~ the engineered safeguards Tavg.
This analysis is performed at 102% engineered safeguards C'~
- power with an ANS 1971 decay heat model. The plant is not' licensed to operate.
7 iJat engineered safeguard power and the results would be-less limiting using an 1
lANS 1979 decay heat-model. Taking credit for the conservative power level
~
assumption and;the.1979 decay heat model, it is concluded that the FSAR
-conclusions remain valid (1.e'.-'a feedline break at.the licensed plant-power
, level will-have acceptable.results). for the increase of 0.3*F in Tavg p,
-uncertainty.
IIr addition, this transient.was analyzed for the steam generator-tube plugging
. program.,-As discussed in Reference 7 and Reference 8,.the analytis assumed a' 3
lTavg. uncertainty ofi6*F.' This.provides:aLbounding analysis for a~Tavg funcertaint'y of,4.3*F.
Therefore, the analysis described in-References 17. and 8:
S' lisinotiaffectedibythis-increasein.Tavguncertainty.
N s
'Singli ReactorLCoolant Pump Locked' Rotor-(15.4.4)-
M 4
c Ly-This1 transient was reanalyzed tofdemonstrate acceptable results with a lower NN analysis:value for-the low:. reactor coolant" loop flow setpoint.
The increased-
]
. uncertainty;in Tavg was explicitly:modeled in the analysis.; Refer to f
subsection'4l3.2; m
[,
, L RCCA Ejection (15.4.6);
Ll
~
Lc L, a
lu 7The(RCCA?ejectionevent'isdefinedasthemechanicalfailureofacontrolrod-drive mechanism pressure housing resulting in the ejection of 6 P.CCA'and drive m
'j;; f ^
. 0802'D.:1D/0'82290 37 3,v. )
') f '.
,)
,]
' shaft., Beginning and end'of cycle conditions are analyzed at full and hot
!zero power' levels. As previously discussed, the hot zero power cases are not
- affected by the increase in Tavg uncertainty.
The hot full power cases.are analyzed assuming DNB conditions immediately following the RCCA ejection.
]
This-minimizes.the heat transfer from the fuel to the coolant which maximizes the fuel rod temperature transient.
Therefore a small change in the coolant j
b
. temperature does.not have_a significant effect on the results.
I
.There is sufficient margin in the current results to conclude that, with the f
increased te:nperature uncertainty and the steam generator tube plugging lc m Leffects.(Reference 7 and 8)..the. acceptance criteria will_ continue to be met.
Therefore. :the conclusions'of the FSAR analyses. and Reference 7 and 8
" :evaluktions remain valid.
Steamline Break Mass and Energy Releases Outside Containment (15.4.2.1).
{
b, LSteamlinebreakmassandenergyreleasedata~arecalculatedfor.severalbreak f
sizes at'different-power levels _for'the purposes of equipment environmental
' A 'qqalifv.ation outsideLcontabment.
Reference 2 contains the results of the
~ dtiide-containmentLsteamitnt. break mass and energy releases.
Farley Units 1 o
(
1and_2LwereincludedinLReference2'aspartofCategory4.
For this analysis,
)
3 the.Tavg' uncertainty. assumed was 6.5*F.
Therefore, the new Tavg uncertainty
- of
- 4;3*F is bounded by the current analysis:assumptio~ns~ and the mass and
'i k
energy release; data remains' applicable.
Because the analysis assumptions'do inot changeh theiconclusions of References 7 and 8 remain: valid..
4 1
- m a
$,1 ' :: Steam 1'ine Break Mass and; Energy Releases Inside Containment (15.4.t), '
l L
I i
Steamline break mass:and energy releaseJdata are calculated for several break.
' sizes at different powerElevels_ for-the-purposes of calculating the 1i containmen.t1 pressure anditemperature response.
The' analysis to calculate mass an'd energy (releases: inside ' containment assumed a-Tevg _ uncertainty of.4'F.
The-1 (analysis Lalsojused alconservative. ANS 1971 decay he'd.7.odel.
Raising the_Tavg-y N
UScertainthtoT3*F,Eand:Using 'the less limiting 'ANS 1979 sdecay heat model,
[
-results 'in' mass andienergy release: data.which is negligibly different from the-
. current analysis.
It is therefore concluded that the current mass and energy release data 1 remains! applicable.for Farlm W ts 1 and 2, and the conclusions of Reference >7 remain valid.
y
'i
-.0802D:10/082290-30 x
f (4.
- x g
- i{h '!!
3.2 CONCLUSION
?"The effects of-the increase in Tavg' uncertainty have been evaluated for all of ~ the non-LOCA transients. The zero power transients are not affected by the ' change. The DNB related transients have been shown to be acceptable by using i existing DNB margin. The preceding discussions for the remaining transients ' demonstrate that.the conclusions of the FSAR and References 7 and 8 remain 4 1 - valid, and the steamline break mass and energy release data remains applicable for the Farley: units.. ? 4.3.3 EVALUATION OF THE LOSS OF FLOH REACTOR TRIP SETPOINT A
- The uncertainty-for the loss of flow trip has ' increased with the RTD bypass Ly
- eliinination.
In' order to maintain the same Technical Specification trip setpoini, a lower analysis value was required. The current analysis value is i B7% of nominal loop flow. The~ revised analysis value is 85% of nominal loop s [ flow. Two transients rely on the low loop flow reactor trip; partial loss of-4 L ' flow and locked rotor. A discussion of the analysis performed for these -transients follows. j DPartial Loss of Forced Reactor Coolant Flow (15.2.5) 3-lI This' analysis _was performed similar to the analysis presented in the.FSAR. y[ (Because' Farley Units 1 and 2 are not licensed to operate at power with only - 4 [ Ltworeactorlcoolantpumpsinoperation,apartiallossofflowwiththree- ? reactor coolant < pumps-initially operating was analyzed.. Three digital
- computer codes were used in;the analysis.
The LOFTRAN code (Reference 3) was Lused: toicalcula'te the-flow coastdown. 'S transient conditions, the nuclear -po'wer transient and the reactor trip on low loop. flow. The FACTRAN code k. (Reference'4) was used to calculate the heat flux transient-based on the 'i' [ nuilear power and flow data from LOFTRAN. Finally, the THINC code (described . oinJSection 4.4 of the FSAR)'was used to' calculate the minimum DNBR'during the-3f ? tra'nsient based on the heat flux from FACTRAN and the. flow from LOFTRAN. u Conservative initial conditions were assumed which included a 5.5'F qm .; uncertainty for Tavg.. The. low flow trip setpoint was assumed to be 85% of-m, nominal ' flow. The effects of increased steam generator tube plugging were also mod 61ed in the analysis (References 7 and 8). s y + Y.n ,08020:10/082290-39 ) Y,Y i:
P 'The results of _ the analysis confirmed that the minimum DNBR during the transient rsmained above the limit value. Therefore, the revised low flow - trip setpoint_ and the increased Tavg uncertainty and the effects of increased ^SG tube plugging have been shown to be acceptable for this transient. Refer to References 7 and 8 for the transient plots for this analysis. L 1 . Single Reactor Coolant Pump Locked Rotor (15.4.4) This analysis was performed similar to the analysis presented in the FSAR.
- Because Farley Units 1 and 2 are not licensed to operate with only two reactor
,[ coolant pumps in operation, the analysis modeled three reactor coolant pumps ~ initially operating. Two digital computer codes were used in the analysis. ~ 'The LOFTRAN code (Reference 3) was used to calculate the flow coastdown, RCS E transient conditions, the. nuclear power transient, the reactor trip on low 1 loop flow, and the peak RCS pressure. The FACEAN code (Reference 4) was used
- to calculate the thermal behavior of the' fuel-at the core hot spot based on
. the' nuclear power 'and flow data from LOFTRAN. Conservative initial conditions were as., et' : .h included a 5.5'F ~ tuncertainty for Tavg'. The low flow trip setpoint wcs assumed to be 85% of nominal; flow..The effects of increased steam generator tube plugging were .also modeled in the analysis (References 7.and 8). The results. of_ th'e analysis confirmed that the. peak RCS pressure remained -below,that which would cause stresses to exceed the faulted condition stress. 111mits. ;In addition,.the calculated zirconium-water reaction remained a'small fraction and the peak clad: surface temperature'was less than 2700*F. Therefore, the revisedilow flow trip'setpoint and the increased. Tavg uncertainty ^and the effects of increased SG tube pluggingLhave been shown.to be acceptable for this transient. ' Refer to References 7 and 8 for the transient plots'for;this analysis. ~ R l4.3,4 's w F In< summary, non-LOCALsafety analyses applicable to the.Farley' Units 1 and 2 g have been evaluated for the replacement of the existing RTD Bypass System with fast resp'nse thermowell. mounted RTDs installed in the reactor coolant loop-o piping. It is concluded that an increas'e in RCS temperature uncertainty q r .08020:10/082290 40 s
4 L can be accommodated by. the margins in the safety analyses and allocation of generie DNB margin. In addition, it has been demonstrated by analysis that the revised analysis value for the loss of flow reactor trip setpoint is ' acceptable. Ali other safety analysis assumptions remain valid. The f ' evaluations have also considered the tube plugging effects of References 7 and 8.. The FSAR and References 7 and 8 conclusions applicable to the J. H. Farley JUnits 1 & 2 are unchanged and all applicable non-LOCA safety analysis acceptance criteria continue to be met. 4.4 LOCA Evaluation The elimination of the RTD bypass system impacts the uncertainties associated
- with RCS' temperature and flow measurement.
The magnitude of the uncertainties c 'are such that RCS inlet and outlet temperatures - thermal design flow rate and i the steam generator performance data used in the LOCA analyses will be -t 0 - slightly affected. The avaltation of the slight increase in the Tavg uncertainty has resulted i_n an estimated ~ increase of 3'F for the Large Break LLOCA Peak Cladding Temperature (PCT) and a 2*F increase for the Small Break
- LOCA PCT.
There is sufficient margin to 2200*F for both the large'and Small u f Break LOCA1 analyses to offset the estimated increases due to RTD bypass ~ i elimination at'the Farley Units...The analytical re :lts represented in L l' fReferences17and8includethe.effectof.thesePCTincreases.
- 3
}4.5!lINSTRUMENTATIONANDCONTROL(I&C)SAFETYEVALUATION J 4 i. c.The' RTD Bypass Elimination modification for the J. H. Farley Units 1 & 2 does nottfunctionally change the AT/T,yg protection channels. TheLimplementation of the fast response RTDs in the reactor coolant piping will change the inputs. g to the AT/T,yg Protection Set. I. II, and III, circuitry: as follows: q ~ gl_.,TheNarrowRange.(NR)(coldlegRTD(usedintheprotectionsystem)inthe s, cold l leg:manifol_d will be' replaced with a fast-response NR dual element !well mounted RTD.in the RCP pump discharge pipe. The signal from this i h_# fast response NR..RTD'will perform the same function as the existing RTD [I 'Tcold; signal.lOneelementof-theRTDwill.beheldin.reserveasaspare. j ~ ( X A [k I i0802D!1D/082290 41 ,M 'l.
- j'
~f. I s i l' i' s g.,
i i
- 2.,The NR hot leg RTD in the bypass manifold will be replaced with 3 fast response NR dual element, well mounted RTDs in the hot leg that are electronically averaged in the process protection system.
3. Identification of failed signals will be by the similar means as before the modifications, i.e., existing control board alarms and protection channel indicators, except that the control systems will not be sensitive to RTD failures or protection channel failures due to HSS. 4. The NR cold leg RTD signah and the NR hot leg RTD signals are electronically processed in the plant 7300 series process protection racks to generate loop.T,yg and delta T signals. These signals (one per loop) a are electronically isolated and' transmitted to the plant 7300 series process control racks. The T,yg and delta T signals are input to a Median Signal. Selector, respectively, which selects the median signal for T . use in the plant -control systems. By rejecting the high and low signals, [ the control system will not act on any single failed' input channel. Since no adverse control: system action therefore results from a single failed instrument channel, a second random failure is'not required per IEEE 279-1971,'section 4.7.- E v1 The existing. protection channel control-board T,yg -and delta T indicators j g iand alarms will provide the means of identifying RTD failures. As part of the RTD: Bypass: Elimination modification, the electronically isolated T,yg and delta T signals willibe utilized'for control grade signals.'and alarms which can:also be utilized to detect failed RTD or a protection' channel input signal. i w i L I Uponfidentification of a. failed hot leg or' cold leg RTD, the operator would ~ (request that I&C personnel place the failed protection channel in.a tripped R condition, identify the'faile'd RTD disconnect the iailed RTD', connect the 1 .other RTD in the dual element device and rescale the applicable'RTD m./ amplifier. After this process, the channel would be returned to~ service. + '4lt 1f both RTDs in a dual element device are bad, the RTD input is removed from. y 4 0 t .the, averaging process and a bias is manually added to a:2-RTD average'Thot I J (as opposed to a 3-RTD average Thot) in rder to obtain a value comparable y with the 3-RTD average Thot prior to the failure of the dual element RTD. 3 ij,. ' 08020:10/082290 42 r r
- 1.L I i ',
R The conversion to thermowell mounted RTDs will result in elimination of the control grade RTDs and their associated control board indicators. The protection grade channels will now be'used to provide inputs to the control' l system through electrical isolators to prohibit faults in the control rack from propagating into the protection racks. In order to satisfy the control and protection interaction requirements of IEEE Standard 279-1971, a Median Signal Selector (MSS) will be used in the control-channels presently utilizing a high auctioneered T r AT signal avg (there will be a separate MSS for each function). The Median Signal Selector Lx p .will use as inputs the isolated protection grade T or AT signals from avg alllthree loops, and'will supply as=an output the channel signal which is the
- median of the three signals.
The effect will be that the various control . grade systems will still use a valid RCS temperature in the case of a single -signal failure. p ll To ensure proper action by the Median Signal Selector, the present manual i L
- switches ? th'at allow for defeating of a T or AT signal from a single avg iloop will be eliminated.
The HSS will automatically select a valid signal in 'the case of a. signal failure; Harnings.that a failure has occurred will be 4 provided by loop to median T,yg and AT deviation alarms. l L' Other than the'above changes, the Reactor Protection System and Control System 5 . will remain the same, as that previously-utilized. For example..two out of l . three voting ~ logic continues'.to be utilized for,the. thermal overtemperature ?, vand overpower / protection' functions, with the model 7300. process control -.bistables continuing ;to operate on a'"de-energize to trip"' principle, t JNonsafety-related control signals: will: now be derived from isolated protection , x Y ~ Ehannel s'. t' j The above principles;of the modification have been reviewed to evaluate. conformance to the requirements of IEEE Standard 279-1971 criteria and 4 a$ socia hd110CFR501 Genera 1LDesigniCriteria (GDC), Regulatory Guides, and other. f A
- applicable' industry.' standards.- IEEE Standard 279-1971 requires documentation.
,.of'a. design.-basis. t FolloWing is a discussion of design basis requirements in 'conformance>to pertinent I&'C criteria. 6 .,{- ls i.
- '08020
- 1D/082290 43
1 y ' a.3 cThe single-failure criterion continues to be satisfied by this change =because the independence of redundant protection sets is maintained. l b.. <The quality of the components and modules being added is consistent with use in a Nuclear Generating Station Protection System. For the Westinghouse Quality Assurance program,. refer to Appendix 17C of the FSAR. c. The changes A ll continue to maintain-the capability of the protection system to initiate a reactor trip during and following natural phenomena credible to the plant site to the same extent as the existing system. d..lChannelindependenceandelectricalseparationismaintainedbecausethe Protection Sc; circuit' assignments continue to be RCS Loop 1 circuits ' input to Protection Set I; RCS Loop 2 to Protection Set II; end RCS Loop 3 L1 - to-Protection Set III,.with appropriate observance of field wiring interface criteria to assure the independence. u L
- e.. Due.to' the elimination of
- the. dedicated control system RTD elements, 4
j temperature signals for use in the plant control systems must.now be-p' . derived from-the protection system RTDs.. To eliminate any degrading control >and protection system interaction mechanisms introduced as a [ consequence of the RTD Bypass Elimination modification, a Median Signal V sSelector has been introduced.into the control system.. The Median Signal ~ o u.m Selectorf preserves the1 functional isolation of interfacing control and c protection' systems 4that share common instrument channels. The details of "if f Lthe signal / selector. implementation are contained in Section 1.3.1 a'nd f LSection 4.5. o l' a L a [ J0nl the ba' sis _of the foregoing rialuation, it is concluded that.the' compliance l/h 'ofltheFarleyunitsto.IEEELStandard 279-1971, applicable GDCs,'andl industry o f fstandards and regulatory guides:has nct been changed with the I&C. l[,Nmodifications'requiredforRTD:bypassremoval. y l f y ) i-L 0802D:1D/08229 - 44 i
E 4.6 MECHANICAL SAFETY EVALUATION The. presently installed RTD bypass system is to be replaced with fast acting b narrow range RTD thermowells. This change requires modifications to the hot l: leg scoops', the hot leg' piping, the crossover leg bypass return nozzle, and h tha; cold leg bypass manifold connection. All welding and NDE till be performed per ASME Code Section XI requirements. Each of these modifications R -is evaluated below. The hot leg temperature measurement on each loop will be accomplished using three-(3) fast response, narrow range single element RTDs mounted in thermowells. To accomplish the sampling function of the RTD bypass manifold system and minimize the need for additional hot leg piping penetrations, the RTD thermowell assemblies vill be located within th? existing RTD Bypass Manifold Scoops wherever pussible. aJ.c to provide- [: _the; proper flow path. If structural' interferences preclude the placement of a LthermowellLin a given scoop,Lthen the scoop will be capped and a new RCS penetration'made to accommciate the' relocated thermowell..The relocated-5 M ermowel' vill be~ located in an. installation boss. A-thermowell design will I b'e used scJ that the.thermowell will' be positioned to provide an average 7 , temperature reading. -The thermowell and installation boss will be fabricated lin accordance.with Section III (Class 1) 6f the ASME Code. The installation .of,the:thermow' ell-into the' scoop or boss will be performed using Gas Tungsten ~ Are Weld. (GTAH) for the. root pass-and finished out with either GTAH or lShieldddMetal'ArcHeld(SMAH).-. The welding wlll be examined by penetrant ~ l test (PT)pertheA*ECodeSectionXI. Prior to welding, the surface of the -scoop or bess.onto which welding will be performed will be examined as ~ required by.Section XI.- o The: cold leg. RTD bypass line must also be removed. The nozzle must' then be ' * ' modified to ' accept the fast response RTD thermowell. The installation of the J thermowell.into.the nozzle will b'e' performed using GTAH for the root pass and finished with either GTAH or SMAH. Held inspection-by PT will, be performed'as ' required by Section XI.- The thermowells will extend approximately [ 1]a,c n
- in'ches into the' flow stream.
This depth has been justified based on j J.c analysis. The root weld joining the thermowells to a [ 0802D:lD/082290 45 1
i t 5.. the modified nozzles will be deposited with GTAH and the remainder of the weld may be' deposited with GTAH or SMAN. Penetrant testing will be performed in Laccordance with the.ASME Code Section XI. The thermowells will be fabricated .in'accordance with the ASHE Section III (Class 1). The cross-over leg bypass return nozzle will be modified and capped or the L existing' piping connection will be severed to leave a stub of pipe protruding from the nozzle and the stub will be capped. The cap design, including l ' materials, will meet the pressure boundary criteria of ASME Section III .(Class ll), The. cap will be root welded to the nozzles by GTAH and fill welded by either GTAH or SHAW. Non-destructive examinations (PT and radiographs)
- will be performed per ASME Section XI. Nachining of the bypass return nozzle B
(or-piping).- as well as any_ machining performed during modification of the penetrations in the. hot and cold legs,'shall be performed'such as to minimize debris escaping into the reactor coolant system. p vin,accordance with Articis-IHA-4000 of Section XI of the ASME Code, a hydrostatic tett of new pressure-boundary welds'is. required when the h ? connection to the pressure boundary is larger than one' inch in diameter. a L; Since the cap for the crossover leg bypass return pipe is-[ l,c inches and the cold leg RTD connections are [ l c inches, a system hydrostatic test is a required after the bypass elimination modification is complete. Paragraph I 3 IHB-5222 of Section XI defines this test pressure to be l.02 times the normal operating, pressure at'a temperature of 500*F or greater. ] t I 'In'sutaary,lthe integrity of-the. reactor coolant' piping as a pressure boundary ' Leomponent.-is maintained by adhering to the applicable ASME Code sections and Nuclear RegulatoryLCommission General Design Criteria. Further, the pressure retaini.ng~ capability and fracture prevention characteristics of the piping is-i not' compromised by these modifications. ',n i -4'.7 TECHNICAL SPECIFICATION EVALUATION As'a result of the calculations summarized in Section.3.0, ~several protection a i functions' Technical Specifications must be modified. 'The affected functions .and' their associated Trip Setpoint information, are noted on Table 3.1-10. 1 '08020:1D/082290 '46
S 5.0 CONTROL SYSTEM EVALUATION l A' prime input to the various NSSS ontrol systems is the RCS average-Ltemperature T(avg). This is calculated electronically as the average of the measured hot and cold leg temperatures in each loop. -The effect of the new RTD temperature measurement system is to potentially . change the time response of the T(avg) channels in the various loops. This in turn could impact the response of [ 3.ct As previously noted,'the new RTD system (RTD + thermowell) will have a a Ltime response slightly longer than that of the current system (RTD + bypass ..line). -.The' additional delay resulting from the Median Signal Selector-(HSS) is;small in comparison with the RTD time response .[( )]a.c Therefore,- .there will be no significant impact onithe T(avg) channel response and no "need, as a result of< implementing the new system, to revise any of the control system setpoints. However APC0 always has the option of making setpoint 'adjustme'nts., 'If-desired, system performance can be verified by performing.a g series of plant tests:(e.g... step load changes, load rejections, etc.) ~ ,following installation of the new RTD system.- Control system setpoints can then be adjusted based on the results of'the tests. It should be recognized m ithat; control systemsido not perform 'any protective function in the FSAR Laccident' analysis. Hith respect. to accident analyses, correrol systems are. + 1 assumed operative;only.in cases in which their action aggravates the ,- ' iconsequences ofLan event,'and/or as required.to establish initial plant ' cond1'tions' fori.an' analysis. The modeling of control' systems for accident - d analysestis' based on-nominal system parameters as presented in the L';Preca0tions,. Limitations,and=Setpointdocument. 4f' n3 h 1 a. y< 1 \\ r .[ f-i -4 i i L08020:1D/082290 47
6.0; CMCLUSh4S The method of utilizing fast-response RTDs installed in the reactor coolant loop piping as a means for RCS temperature indication has undergone extensive 1
- analyses, evaluation and testing as described.in this report.
The -incorporation of this system into the J. M. Farley Nuclear Plants Units 1 and 2 design meets all safety, licensing and control requirements necessary for-safe operation of these units. The analytical evaluation has been supplemented with in-plant and laboratory testing to further verify system I perfore;nce. The fast response RTDs. installed in the reactor. coolant loop I piping adequately replace the present hot and cold leg temperature measurement system and enhance ALARA efforts as well as improve plant reliability. In addition to the effects of the RTD Bypass Elimination, this evaluation also consider the effects of increased SG tube plugging and reduced RCS flowrate as described'in References 7 and 8. i +: l a
- i u
d e 'I y a hs, 1 l. k J ,t H 08020:1D/082290 48
7.0 REFERENCES
- 1. FNP Final Safety Analysis Report
- 2. Butler, J. C., Love, D. S., "Stetmline Break M m / Energy Releases for Equipment Environmental Qualification Outside Containment," HCAP-10961, Revision 1 (Proprietary), October 1985.
- 3. Burnett, T. H. T., et al., "LOFTRAN Code Description," HCAP-7907-P-A l
(Proprietary), HCAP-7907-A (Non-Proprietary), April 1984.
- 4. Hargrove, H.G., "FACTRAN-- A FORTRAN IV CODE FOR THERMAL TRANSIENTS IN A.
00 FUEL ROD," HCAP-7908-A (Non-Proprietary), December 1989. 2 R' American National Standard for Decay Heat Power in p _5. ANSI /ANS-5.1-1979, L . Light Water Reactors, August 29, 1979. L: l' L -6.:Nureg-0717, Supplement No 4, " Safety Evaluation Report Related to the-Operation of Virgil..C. Summer Nuclear' Station, Unit No.1," Docket No. L: 50-395 -August, 1982. l l '74 Morrison, R. J..'" Alabama Power Joseph M. Farley. Unit 2 -Increased Steam [ ( Generator Tube ~ Plugging and Reduced Thermal Design Flow Licerting Report," 't HCAP-12659 -(Non-Proprietary), July' 1990. a
- 8. Morrison,lR. R.
" Alabama Power Joseph M. Farely Unit l' Increased Steam m Generator -Tube Plugging and Reduce Thermal Design' Flow Licensing Report," l n. -HCAP-12694-(Non-Proprietary),_ August 1990. j p .ib5
- 9EFNP Precautions,' Limitations and Setpoints.
3 .g m -10.1Ciocca? C. F., et al., " Evaluation of the Impact of Cable Splices and - Penetration Leakage On,RPS/ESFAS and ERP Setpoints, Farley' Units 1 and 2," 4 jj .HCAPfll658 Rev 1, August, 1989. W:;
- / :
e ,o h i
- 0802D:ID/082290 49 j
L i I 't 1,- l 1 APPENDIX A DEFINITION OF AN OPERABLE CHANNEL AND HOT LEG RTD FAILURE COMPENSATION PROCEDURE i i I i i l '.k. [1 ' ,j \\ le ( sl, i a ( f. i ' ' l lt it s' i I I + l 7 1 ,i ' ' ' _'l'. _i \\
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1 f 1 L:n:f, t i l 1 l-50 j ' 0802D:10/082290 s l r
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r-v I l J l l $_ RT3 BYPASS ELIMINATION FOR 1, d' J. M. FARLEY HUCLEAR PLANT-l! UNITS 1 AND 2 lr l l. DEFINITION OF AN OPERABLE CHANNEL AND HOT LEG RTD FAILURE COMPENSATION PROCEDURE L' l: l t This document contains information proprietary to d,, Hestinghouse Electric Corporation; it is submitted in' I ~ confidence and is' to be used solely for: the purpose t 'for'which it is fornished and returned upon request. This: document.and such information is'not-to be reproduced, transmitted, disclosed or.used otherwise' 6-in'whole or in part without the written authorization j .r. of Hestinghouse Electric Corporation. 3 ,t[i i Hestinghouse Electric Corporation 1 , e. e:" ~ Pittsburgh,-PA mui '\\ in ,'i y 'l08020:1D/082290 51 oi ?. w
~ DEFINITION OF AN OPERABLE CHANNEL The RTD Bypass Elimination modification uses the average of 3 RTDs in each hot leg to provide a' representative temperature measurement. In the event one or more of the RTDs fails, steps must be taken to compensate for the loss of that RTD's input to the averaging function. J. H. Farley Nuclear Plant (FNP) will have dual' element RTDs installed in each hot leg thermowell location. The second element may be used chen the first element fails and the three RTD average maintained. In the event of the second element failing in the same 'RTD, then this' procedure could be invoked. Single RTD Failure U 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 [' 4 5'. .]a,c - Typically this data is recorded at initial 100% power and -thereafter, during the normal protection channel surveillance interval,
- ]a.c any hot leg can then Once [
' tolerate. failure of both elements'of a single dual, element'RTD and still remain operable. If'the situation arises where such a. failure occurs a bias value must be-applied:to the average of the remaini'ng two valid RTDs. -- (. s a y> b L]a,c in i,; w . i i / -;08020:1D/082290 52 ~ 't (4 s
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. [- ,)a,c 'The Median Signal Selector will eliminate any control system concerns, the Tavg and'AT signal associated with the loop containing the failed hot leg RTD will most likely not be.the Median Signal chosen as the input to the control ~ systems. If another hot leg RTD fails in a different loop the FNP should operate'using manual control.. Manuil Rott Control is recommended so 1 that.the operator can control the plant based on the best measurement available. If automatic operation is (:,ntinued the control s(stem may choose the biased channel-due to the positive (or zero) bias applica: ion. This means the control system will perceive a higher Tavg than actually axists at reduced . power and the plant will operate at : educed. temperatures. -Hhile this is not necessarily undesirable it aoes rer.uce the total plant megawatt output. The. use of automatic rod control should Se considered based on utility power-requirements.- Cold Leg:: If the active cold leg RTD fails, then that RTD should be
- disconnected from the 7300 cabinets.
The dual element spare RTD should-then be' connected'in the. failed RTD's place.- Double'RTD Failure:- Inocerable Channel Hot,LegohColdLeg: If both elements of two or moretof the three hot leg. t dual element RTDs or the coldgleg dual elements RTD elements fail in the same . protection channel then that channel is considered inoperable and should be -placed in trip. -Operation withsonly one valid hot leg RTD is not presently analyzed'as part.of the licensing. basis. ?I y-l ' 0802D:,10/082290 53 m
i l PROCEDURE FOR OPERATION WITH A HOT LEG DUAL ELEMENT RTD OUT OF SERVICE i The hot leg temperature measurement'is obtained by averaging the measurements from the three thermowell RTDs installed on the hot leg of each loop. { l: ,)a c In the evenc that one of the three dual element RTDs fails, the failed.RTD - will be disconnected and the hot' leg temperature measurement will be obtained by averaging the" remaining two RTD measurements. [- d ,j .C a The bias adjustment corrects for- [L L i A L .]a.c.To assure that the l'" aeasuredhotIlegtemperatureismaintainedatorabovethetrue.hotleg Ll, ' i temperature, and thereby avoid a reductionLin safety margin at reduced power, l L' ,a,c l y. i~ ? 1' sJ + p n, l L 1l5 0802D:1D/082290 54
An_ RTD. failure will most. likely result.in an offscale high or low indication-and will-b'e detected through' the normal means in use today (i.e., T,y and ' AT. deviation alarms and. indicators). Although unlikely, the RTD (or ts electronics channel) can fail gradually, causing a gradual change in the. loop ' temperature measurements. [ l .3 c I a .The detailed procedure for correcting for a failed hot leg RTD is presented 1 below: l .l a,c 1 l \\ n l I l + i I 1 1e l - a i s.ji I 1 1 l i l i 'l I ) 4 !l '08020:10/082290' 55 3! I
&,C + 4 L ji, 9b - m-k k = = 08020:10/082290 56
8.C 1 I0B020:10/082290 57
APPENDIX CALCULATION OF HOT LEG TEMPERATURE BIAS a,c 08020:10/082290 58
M APPENDIX B ACRONYMS FOR UNCERTAINTY CALCULATIONS 0802D:1D/082290 59
M g 9 9 mesan 60
mm -ums-ia.-i-----imiii i i i W m 61
e..-. - +
- n w
w e v b b~ 4 - (j .+4 t' (' r w, .d N sN 4 M t 4 l-h Y 62
A m 2 T 9 = -7
== S 63
W m 3.' 64
mm O
g n E T E' LL M -M mm N mensen 65
N m a. a t e M 66
m t 1 67}}