ML20148H018
| ML20148H018 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 01/19/1988 |
| From: | Tucker H DUKE POWER CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| NUDOCS 8801270091 | |
| Download: ML20148H018 (17) | |
Text
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Duxu POWER GOMPANY P.O. HOX 33180 citARLorrE, N.O. 28242 HALH. TUCKER TE LEPHONE YKE PmEEEDENT
)
stuaan Pacowrton January 19, 1988
)>M
' f.
U.S_. Nuclear. Regulatory Commission
'i f 6cument. Control. Desk Washington, D.C.
20555
Subject:
McGuire Nuclear Station, Units 1 and 2 Docket Nos. 50-369 and 50-370 RID Bypass System Removal Gentlemen:
In response to Dr. K. Jabbour's letter of November 6, 1987, please find attached edditional information concerning the elimination of the Resistance Temperature Detector (RTD) Bypass System at McGuire, originally submitted by letter dated October 29, 1985 and supplemented by letter dated May 26, 1987.
This letter constitutes a response for the McGuire Nuclear Station only.
A response for the Catawba Nuclear Station (as requested in Dr. Jabbour's letter) was transmitted by letter dated November 19, 1987.
In the October 29, 1985 submittal associated with the proposed RTD Bypass Manifold removal, Table 3.3-2 was revised to indicate a 10 second response time for Delta-T trip functions.
That response time included a 3 second electronic filter time constant and a 5.5 second RTD response time (plus a 1.5 second electronics delay).
In recent testing at another plant, following completion of the RTD Bypass modi-fication, response times have been found to be slightly greater than 5.5 seconds.
Therefore, contributions to the 10_second response time, as defined in Table 2.2 1 of the proposed Technical Specification change associated with the modification, have been reallocated to include a 6.5 second RTD response time and a 2 second electronic filter time constant.
The 1.5 second electronics delay remains the same, and the total response time remains at 10 seconds. The changes to Technical Specifications and to Attachment II of the October 29, 1985 submittal are included as Attachment II of this letter.
Because the total response time of 10 seconds has not changed from the original submittal, no additional justification is required.
Very truly yours, M
y Hal B. Tucker l
SAG /102/j ge 0g 8801270091 880119 4
ADOCK0500g9
]
Attachments DR
N ;- -
,y 1 Document, Control Desk
January'19,.1988
.Page_2 p
xc:
Dr. J. Nelson Grace
. Regional Administrator
' U 'S.- _ Nuclear Regulatory ' Commission Region II-101'Marietta-Street' NW - Suite 2900 Atlanta,;GA 30323 Mr. Darl _ Hood, Project Manager Office of. Nuclear Reactor Regulation
,U.S. Nuclear Regulatory Commission Washington,-D.C. 20555 Mr. W.T. Orders NRC Resident Inspector McGuire Nuclear Station
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+
LMCGUIRE NUCLEAR' STATION
.RTD BYPASS ELIMINATION QUESTION.
(1). Reg'arding.the problem of drift of the RTD response time identified in NUREG-0809 (Reference 1), describe (a) the method (s) for checking RTD response time, and (b) the safety allowance or other methods to provide assurance that the response times do not drift outside acceptable limits between'the 18-month checks.
RESPONSE
(1) Duke power has contracted Analysis and Measurement Services (AMS) to perform the response time testing of the RTDs at the McGuire Nuclear Station. The response time verification will utilize the Loop Current Step Response (LCSR) Method as discussed in NUREG-0809. The options presented in NUREG-0809 to address the potential for RTD response time degradation are to increase.the required surveillance frequency or include additional allowances in the safety analysis. Duke has been performing the response time testing of all-RTDs at least once per 18 months as suggested by NUREG-0809 but has not revised the Technical Specifications accordingly.
The increased surveillance requirement will be continued after the RTD
' Bypass System Removal modification. The Technical Specifications revision suggested by NUREG-0809 is provided.
QUESTION.
(2) There is evidence of a slow drift in RTD readings with time (References 2, 3, and 4), possible all in one direction. Discuss how long-term drift would be detected before it reached una.ceptable levels.
The staff has accepted one licensee's commitment to replace two RTDs after each of the next-two refueling outages to check the calibration of the removed RTDs.
RESPONSE
(2) Westinghouse recommends the performance of an RTD cross-calibration during the heatup after each refueling.
This procedure required multiple measurements at three or four different temperatures.
To date, Westinghouse has evaluated the data from over 400 RTDs using this technique, far more than any other recent testing.
Several repeat tests performed one to three years apart have not shown any indication of drift in only one direction.
The results of the tests indicate that the RTDs drift less tnan [+0.4 F, l
which is significantly less than the + or - 0.7 F] a,c, assumed for L
uncertainty calculations for the protection system.
The procedure l
sensitivity is sufficient to discern a random drift of less than 1.0 F by one or several RTDs. At that time, either the calibration of the R/E(s) for the affected RTD(s) would be adjusted to account for the shift, or the RTD(s) would be declared inoperable end would be replaced.
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RTD BYPASS ELIMINATION
! PAGE'2' f
t- :
~ REFERENCES' 4
h f(1) NUREG-0809, Safety Evaluation Report, Review of Resistance Temperature Detector. Time Response Characteristics, August,-1981.
(2);LNUREG/CR-2928, Degradation of Nuclear Plant Temperature Sensors, June, 1987.
(3)l'K. R.._Carr, An1 Evaluation of' Industrial Platinum. Resistance Thermometer.
Temperature - Its-Measurement and Control in Science and Industry, ISA
-publication,' Vol. 4, Part 2, 1972, pages 971-982.
(4)
B. W..Mangum, the Stability.of Small Industrial Platinum Resistance Thermometers, Journal of Research of the NBS, Vol. 89, No. 4, July-August 1984, Pages 305-350.
QUESTION' (3) Failure of a RTD in a particular loop would be signalled and alarmed by the deviation of the Tavg and Delta-T values for the loop compared to the values
.of the other.three loops. What are the deviation values that -will activate the: alarm? What'is the minimum temperature error in a single failed RTD that will cause the deviation alarm to go off? How often will individual RTDs be monitored in normal operations?
RESPONSE
.(3) ;There presently exists as part of the plant electronics both a Tavg and a Delta-T deviation alarm. This alarm compares the four signals; either Tavg or Delta-T, one from each channel, to a pre-set threshold value-This value is' nominally set-to + or - 2 P and is adjusted during startup and subsequent operation such that it is just beyond the range of normal operating variations.
For the present"channel' configuration Tavg is calculated by adding the hot
' leg temperature (Thot) and cold leg temperature (Tcold) and dividing the' sum
.by two.
If we assume the deviation threshold remains at & or - 2 F then the deviation alarm will enunciate.when the Thot value has shifted 4 F.
In the
- proposed new configuration, Thot is first calculated by averaging three (3) independent RTDs. This Thot signal is then used to calculate Tavg in the same manner as before.
Since Thot now relies on three (3) RTDs instead of one (1), the shift in any one RTD could be 12'F.
This represents, however, the same Thot shift of 4 F that presently exists and provides the same level of warning that Thot channel is experiencing problems.
The same logic can be applied to the Delta-T deviation alarm.
Delta-T is calculated by: Thot - Teold.
In this case, the present plant configuration would allow Thot to shift by 2*F before the + or - 2 F threshold value is met.
For the proposed configuration there are 3 RTDs being used to generate Thot and therefore, any one RTD could shift 6*F.
The allowable channel to channel deviation remains the same for the present design and the proposed new configuration. The additional shift that individual RTDs are allowed is a direct result of the averaging nature of the modification. The impact of any one RTD changing is reduced by a factor of 3 and, therefore, its allowable shift can be larger.
RTD BYPASS ELIMINATION PAGE 3
. Current plans are for each RTD indication to be recorded hourly as part of the overall plant performance monitoring. These hourly records will be used to calculate the bias adjustments associated with failure of a hot leg 3TD.
QUESTION (4)
Indicate your plans to check and confirm the accuracy of the new hot leg temperature measurements by comparison against past measurements by RTDs in
~
bypass loops. The staff has accepted one licensee's commitment to obtain confirmatory information by comparing pre-installation and post-installation calorimetric data on RTD temperature measurements in their plant for matching operating conditions. The staff intends to review this confirmatory infornation.
RESPONSE
(4)
The present method of measuring hot leg temperatures utilizing the RTD Bypass Scoops has a basic measurement uncertainty of [1.3 F] +a,c due to the temperature streaming phenomenon. This uncertainty represents the difference between the expected true bulk hot leg temperature and the temperature of the fluid sample taken by the present system.
It is a direct indication of the accuracy of this temperature measurement method.
The new temperature measurement method also has a similar streaming uncertainty factor, but it is somewhat less at [1.0 F] +a,c.
Since both methods have an inherent streaming inaccuracy, accounted.for in the Safety Analyses, it is not appropriate to compare the new method to the old method and declare any differences as errors.
It is possible, however, to make a comparison between the old and new
~
systems using the normalized full power Delta-T measured before and after the modification.
It is expected that these Delta-T readings will be very similar once any secondary side measurement errors, such as feedwater flow, have been factored into the power calculation.
If there were any dramatic differences between the two Delta-T readings it would indicate that a problem existed with one of the measurement methods.
Duke will perform a comparison of the temperature indications after the modification with measurements prior to the modification.
If desired, the NRC will be notified of the results of this comparison including any explanation of variations larger than expected.
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QUESTION (5) provide information on the changed position of one of the RTDs in loop B regarding its distance downstream from the former location and also its circumferential position compared to before.
provide a sketch of this installation. Will the wake from the scoop upstream of this new RTD placement affect the accuracy of the reading? Will the RTD in the new location be placed in a scoop as the others are, or be directly exposed to the flow? If directly exposed to the flow, what offect will this have on the temperature value, and also on the response time, compared to the other RTDs which are insida scoops?
-RTD BYPASS ELIMINAU ON
.PAGE'4.
- RESPONSE -
Not applicable-to McGuire.
~QUESTIONE (6). The RCS flow uncertainty calculated for the new RTD system'1s 11.7%.
Explain why the 12.1% uncertainty calculated for the bypass RTD' system is.
retained in the revised Catawba Technical Specifications for?the new system.
RESPONSE
Not applicable to McGuire.
QUESTION (7) 'The elimination of the RTD bypass system impacts uncertainties associated with the RCS system temperature and flow measurement. This can possibly affect the LOCA analysis. Provide a LOCA analysis or justify the reason for not providing one.
RESPONSE
- (7)- The effect of increased RTD errors would impact the uncertainties associated with RCS temperature and RCS flow measurement..The magnitude of the uncertainties introduced would be less than 1*F and therefore, RCS inlet and outlet temperatures, thermal. design flow' rate and the steam generator
-performance data input, used in the LOCA analyses, will not be affected.
Past studies have shown that the variation of the core inlet temperature T1" used in the LOCA analyses, affects the predicted core flow during the blowdown period of the transient. The level of positive core flow is influenced by the two-phase vessel side break flow, and the core cooling is affected by the quality of the fluid. These sensitivity studies concluded that the inlet temperature effect on Peak Clad Temperature is' dependent on the break size. As a result of these sensitivity studies, the LOCA analyses are performed at a nominal value of T The RCS flow rate and the steam generator secondary side temperature a0d pressure are also determined using the Tavg output. These inputs to the analyses will not be affected due to the RTD bypass climination for reasons stated above.
QUESTION (8) The increase in RTD response time could affect the FSAR Chapter 15 Loss of Load / Turbine Trip non-LOCA transient. Provide an analysis or justify the reason for not providing one.
. RESPONSE Not applicable to McGuire.
l QUESTION (9) Indicate how the RCS Tavg is impacted by the accuracy of the RTD measurements. Will any of the TS setpoints be in need of change because of a change in the accuracy of Tavg with the new system?
RhDBYPASSELYMINATION PAGE 5
RESPONSE
(9) The accuracy of the RTDs and their impact on Tavg, setpoints, and other plant parameters are discussed in the submittals of October 29, 1985 and May 26, 1987.
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ATTACHMENT II Technical Specification Changes i
No C#6#GE p
FRON OcroBEK 29 ngy SuBH TTAL.
f TABLE 2.2-1 x
1 E.S REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS El FUNCTIONAL UNIT
~
_ TRIP SETPOINT ALLOWABLE VALUES
- 1. Manual Reactor Trip N.A.
c i
5 N.A.
d
- 2. Power Range, Neutron Flux Low Setpoint 5 25% of RATED I
THERMAL POWER
. Low Setpoint 5 26% of RATED
~
E THERMAL POWER a
~ High Setpoint
< 109% of RATED THERMAL POWER High Setpoint 1 110% of RATED m
THERMAL POWER
- 3. Power Range, Neutron Flux,
< 5% of RATED THERMAL POWER with High Positive Rate a time constant > 2 seconds
< 5.5% of RATED THERM %L POWER Uith a time constant 1 2 seconds
- 4. Power Range, Neutron Flux,
< 5% of RATED THERMAL POWER with High Negative Rate a time constant > 2 seconds
< 5.5% of RATED THERMAL POWER Gith a time constant 1 2 seconds
'?
- 5. Intermediate Range, Neutron 5 25% of RATED THERMAL POWER Flux S 30% of RATED THERMAL POWER
- 6. Source Range, Neutron Flux 5 10s counts per second 5 1.3 x los counts per second
- 7. Overtemperature AT See Note 1 See Note 3 W
- 8. Overpower AT See Note 2 See Note [ P l
((
- 9. Pressurizer Pressure--Low 1 1945 psig' 2 1935 psig h f
- 10. Pressurizer Pressure--High 1 2385 psig 5 2395 psig ao
- 11. Pressurizer Water Level--High 5 92% of instrument span 5 93% of instrument span OO
- 12. Low Reactor Coolant Flow 1 90% of design flow per loop
- 1 % of design flow per loop
- 22 l
11 88.8 %
sn
- Design flow is 97,220 gpo per loop. *
%t P%oR To REMa/AL Of GACH UAM6, gfg g
jg gg y y A
MPUQABLE
X CHM g f
R I ) f_ {
TABLE 2.2-1 (Continued)
M 8'TTA L, g
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS
- o NOTATION NOTE 1: OVfRTEMPERATURE AT 1
1+T5 1
AT (I + r,S} (1 + IsS} I O 4
{
IK
~K Il + gS)[T(1
)-T'] + K (P-P') - f (a!)}
1+TS o
I 2
2 3
3 Where:
AT Measured AT by RTD Manifold Instrumentation,
=
f{
Lead-lag compensator on measured AT,
=
ti, r2
= Time constants utilized in the lead-lag controller for AT, T: > 8 sec., r2 5 3 sec.,
o'o 1
1 + r3 Lag compensator on measured AT,
=
W ts nstants utilized in the lag compensator for AT,13 $ 2. S ec, l
AT, Indicated AT at RATED THERMAL POWER,
=
K 5 1.200, g
.F.F K
0.0222
=
yE
.2 1+rS SS The function generated by the lead-lag controller for T,9 dynamic compensation,
=
3, 3
EE Time constants utilized in the lead-lag controller for T,yg, T4 ts
=
> 28 sec, rs I 4 sec.,
gg T4 a :2
- 3 T
=
Average temperature,
'F, 3, 7s3 Lag compensator on measured T
=
yg,
Y CHA M 6 (R6H
)
OO6k ? 7> 1975~.~
TABLE 2.2-1 (Continued) v6 Mg xa REACTOR TRIP SYSTEM INSTRtMENTATION TRIP SETPOINTS M
NOTATION (Continued) e NOTE 1:
(Continued) x
~
k d
T.
=
Time constant utilized in the measured T 1ag compensator, r. '<
P__ sec_.
l w
1 s
- 'U k
T' 5 588.2*F Reference T,yg at RATED THERMAL POWEN,
=
m i
K
=
0.00I095, 3
P
=
Pressurizer pressure, psig, P'
= 2235 psig (Nominal RCS operating pressure),
m S
Laplace transform operator, sec 8,
=
E and f (AI) is a function of the indicated difference between top and bottom detectors of the power-range nuclear ton chambers; with gains to be selected based on measured instrument response during plant startup tests such that:
(i) for q
~9 between -25 and +9E; f % = 0, dere q aM g are percent RATED t
b 3
g b
j THERMAL POWER in the top and bottom halves of the core respectively, and q
+o t
b is total THERMAL POWER'in percent of RATED THERMAL POWER; j
g (ii) for each percent that the magnitude of qt "b exceeds -29%, the AT Trip Setpoint shall be automatically reduced by 3.151% of its value at RATED THERMAL POWER; and 3
<+ r+
l yy (iii) for each percent that the magnitude of qf g exceeds 49.0%, the AT Trip Setpoint b
shall be automatically reduced by 1.50% of its value at RATED THERMAL POWER.
CC 1
r* r*
Mo *CH/VIGE RtoM TABLE 2.2-1 (Continued)
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS b
M NOTATION (Continued) m C
6 T
As defined in Note 1,
=
m
~
T" 5 588.2 F Reference T,yg at RATED THERMAL POWER,
=
E S
As defined in Note 1, and
=
m f (AI) 0 for al1 AI.
=
2 Note 3:
The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than
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?7 3/4.3 INSTRUMENTATION O
3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the Reactor Trip System Instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE with RESPONSE TIMES as shown in Table 3.3-2.
APPLICABILITY:
As shown in Table 3.3-1.
ACTION:
As shown in Table 3.3-1.
SURVEILLANCE REQUIREMENTS 4.3.1.1 Each Reactor Trip System Instrumentation channel and interlock shall be demonstrated OPERABLE by the performance of the Reactor Trip System Instrumentation Surveillance' Requirements s,pecified in Table 4.3-1.
4.3.1.2 The REACTOR TRIP SYSTEM RESPONSE TIME of each Reactor trip function shall be demonstrated to be within its limit at least once per 18 months.
Each test shall include at least one train such that both trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific Reactor trip function as shown in the "Total No. of Channels" column of Table 3.3-1.
4.3. l.3 gyg yg0N56 TINS of RTbs Asseumsa kilfH D{g ASAc%t %tIP SYsWM sHAU. 86 b6/10lV$fRATCb *1b 06 WI1'HlAl 1WGiA Litil75
($66 NdT6L ro TAsis s.s-2.)
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McGUIRE - UNITS 1 and 2 3/4 3-1 e
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TABLE 3.3-2 UMQL m
REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES E
-d FUNCTIONAL UNIT RESPONSE TIME w
g 1.
Manual Reactor Trip N.A.
o.
~
2.
Power Range, Neutron Flux s0.5second4Cl) 3.
Power Range, Neutron Flux, High Positive Rate N.A.
4.
Power Range, Neutron Flux, High Negative Rate 10.5 second*(3) 5.
Intermediate Range, Neutron Flux N.A.
6.
Source Range, Neutron Flux N.A.
< k seconds *(.DC2.)(3) 7.
Overtemperature AT
< kseconds C0(.2_)(3) 8.
Overpower AT d
9.
Pressurizer Pressure--tm<
NN
~< 2.0 seconds EE 10.
Pressurizer Pressure--High EE
-< 2.0 seconds EE 11.
Pressurizer Water Cevel--High N.A.
ee E ?..
22 (1) Neutron detectors are exempt from response time testing.
Response time of the neutron flux signal portion of the channel shall be measured from detector output or input of first electronic component in channel.
hh Q & lo.o secml respmse he
- ,iclodes o %ecmd
&% 4 PAe.
RTDs 50 rmounted k ihe-ells.
9 96 U (3)Tg~c gyppf gy,y,cSwJ.5E T/H6 l5 MAL' CASE 5 (dE/4ats.Oj OVTIL TM/] TN6 VAW jogSgpub p
No cHugg_ ppgg R 2% 19ff D L-LIMITING SAFETY SYSTEM SETTINGS BASES (wrrH RTD &/9M5.5VSTEM
/N6fALdd)
Overtemperature AT The Overtemperature Delta T trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to piping transit delays from the core to the temperature detectors (about 4 seconds),
and pressure is within the range between the Pressurizer High and Low Pressure trips. The Setpoint is automatically varied with:
(1) coolant temperature to correct for temperature induced changes in density and heat capacity of water and includes dynamic compensation for piping delays from the core to the loop temperature detectors, (2) pressurizer pressure, and (3) axial power distribu-tion. With normal axial power distribution, this Reactor trip limit is always below the core Safety Limit as shown in Figure'2.1-1. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the Reactor trip is automatically reduced according to the notations in Table 2.2-1.
Overpower aT The Overpower Delta T tr'ip provides assurance of fuel integrity (e.g., no fuel pellet melting and less than 1% cladding strain) under all possible overpower conditions, limits the required range for oyertemperature delta T protection, and provides a backup to the High Neutron Flux trip.
The Setpoint is automatically varied with:
(1) coolant temperature to correct for tempera-ture induced changes in density and heat capacity of water, (2) rate of change i
of temperature for dynamic compensation for piping delays from the core to the loop temperature detectors, and (3) axial power distribution, to ensure that the allowable heat generation rate (kW/ft) is not exceeded.
The Overpower AT t
trip provides protection to mitigate the consequences of various size steam breaks as reported in WCAP 9226, "Reactor Core Response to Excessive Secondary Steam Break."
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McGUIRE - UNITS 1 and 2 8 2-4 %
j tt CHhsifE i Frce; LIMITING SAFETY SYSTEM SETTINGS Oc7, 29; /hf;f sue,1g BASES CalsTH 69F455.59t;7EM /2ff4evEn ; RTA.s /4) TNER/400guL)
Overtemperature AT I
~
The Overtemperature Delta T trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to p&p kg trer.dt thy: < e-the cer: te th: teg er:tur: et ; tera ( ceut 4 se;;. 0 ),
and pressure is within the range between the Pressurizer High and Low Pressure trips.
The Setpoint is automatically varied with:
(1) coolant temperature to correct for temperature induced changes in density and heat capacity of water and includes dynamic compensation for piping delays from the core to the loop,
temperature detectors, (2) pressurizer pressure, and (3) axial power distribu-tion. With normal axial power distribution, this Reactor trip limit is always below the core Safety Limit as shown in Figure'2.1-1. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the Reactor trip is automatically reduced accordino to the notations in Table 2.2-1.
6+Lu l My suoadd wdh RTD.s ~wM h ik-MM Overoower AT
@(,,.5 secondsM 5
The Overpower Delta T tr'ip provides assurance of fuel integrity (e.g., no fuel pellet melting and less than 1% cladding strain) under all possible overpower conditions, limits the required range for oyertemperature delta T protection, and provides a backup to the High Neutron Flux trip. The Setpoint is automatically varied with:
(1) coolant temperature to correct for tempera-ture induced changes in density and heat capacity of water, (2) rate of change of temnerature for dynamic compensatie g r ;. A i
cop temperature detectors, and (3) axial power distribution, to ensure that l
the allowable heat generation rate (kW/ft) is not exceeded.
The Overpower AT l
trip provides protection to mitigate the consequences of various size steam i
breaks as reported in WCAP 9226, "Reactor Core Response to Excessive Secondary l
Steam Break."
for W-frone M M delqb 966ccidh u$
NC
~
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McGUIRE - UNITS 1 and 2 B 2-5
WESTINGHOUSE PROPRIETARY CLASS II TABLE 4.1 RESPONSE TIME PARAMETERS FOR RCS TEMPERATURE MEASUREMENT RTD Bypass System Fast Response Thermowell FSAR Revised RTO System RTD bypass piping and thermal lag 2.0 2.0 RTD response time 0.5 0.5 h 6,5- $
RTD electronic filter time constant 2.0 6.0 he-g,o %
Electronics delay 1.5
.1.5 1.5 Total Response Time 6.0 sec 10.0 sec 10.0 sec 4.2 RTD UNCERTAINTY The proposed fast response thermowell RTO system will make use of RTOs manufactured by the RdF Corporation with a total uncertainty of [il.2'F]+a,c assumed for the analyses. Currently, the McGuire units are supplied with Rosemount RTDs, each of which has a smaller uncertainty value.
Therefore, the increased uncertainty associated with the RdF RTDs must be accounted for in the safety analyses. Since three RTDs are used to measure hot leg temperature, the impact of the larger RdF RTO error is reduced.
The FSAR analyses make explicit allowances for instrumentation errors for some of the reactor protection system setpoints.
In addition, allowances are made j
for the initial 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 for non-ONB events; for DNB event's these allowances are statistically combined into the design limit DNBR value, consistent with l
the Improved Thermal Design Procedure.
The following protection and control system parameters are af fected by the l
assumed narrow range RTO accuracy: Overtemperature Delta-T Reactor Trip (OTOT), Overpower Delta-T Reactor Trip (OPDT), Low RCS Flow Reactor Trip, RCS i
3255e:ld/090485 18
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