ML20205H377

Proposed Tech Spec Changes,Supporting Station Mod Re Removal of Resistance Temp Detector Bypass Manifold

ML20205H377
Person / Time
Site:	Mcguire, McGuire
Issue date:	10/29/1985
From:	DUKE POWER CO.
To:
Shared Package
ML19344C184	List: ML20205H332 ML20205H357 ML20205H377 ML20205H390
References
NUDOCS 8511150119
Download: ML20205H377 (14)

Text

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/ ATTACHMENT II Technical. Specification.

Changes

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l x TABLE 2.2-1 S

g REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS El FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES

1. Manual Reactor Trip N.A.

N.A.

d 2. Power Range, Neutron Flux w Low Setpoint -< 25% of RATED

, THERNAL POWER -

. Low Setpoint 1 26% of RATED

=

THERMAL POWER E

m 'High Setpoint - < 109% of RATED

~

High Setpoint THERMAL POWER 1110% of RATED THERMAL POWER

3. Power Range, Neutron Flux, High Positive Rate 5 5% of RATED THERMAL POWER with 5 5.5% of RATED THERMAL POWER a time constant 1 2 seconds with a time constant 1 2 seconds

4. Power Range, Neutron Flux, High Negative Rate i 5% of RATED THERMAL POWER with

< 5.5% of RATED THERMAL POWER a time constant 1 2 seconds 7 5. Intermediate Range, Neutron with a time constant 1 2 seconds Flux i 25% of RATED THERMAL POWER S 30% of RATED THERMAL POWER

6. Source Range, Neutron Flux 1 105 counts per second i 1.3 x 105 counts per second

7. Overtemperature at See Note 1 See Note 3 N

8. Overpower AT See Note 2 SeeNote/ l

(( 9. Pressurizer Pressure--Low 1 1945 psig'

> 1935 psig

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10. Pressurizer Pressure--High 1 2385 psig i 2395 psig OO

11. Pressurizer Water Level--High 192% of instrument span i 93% of instrument span

12. Low Reactor Coolant Flow 190% of design flow per loop

• 22 I 1pf%ofdesignflowperloop*

EE r* <* SS.S%

• Design flow is 97,220 gpm per loop. '

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TABLE 2.2-1 (Continued) b 5x REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS m

NOTATION s

e i NOTE 1: OVERTEMPEFATURE AT w

1+rS 1 AT(f ) (y , Iga 3) $ AT, (Ky -Kp (y . 3)[T(3 , 3)-T'] + N3 (P-P') - f y(AI))

Where: AT. =

Heasured AT by RTD Manifold Instrumentation, f =

lead-lag compensator on measured AT, t,12 i = Time constants utilized in the lead-lag controller for AT, T > 8 sec., 12 1 3 sec.,

E 1

=

y, Lag compensator on measured AT, l T3 =

Time constants utilized in the lag compensator for AT,13$

3 /rsec. l AT, =

Indicated AT at RATED TiiERMAL POWER, Kg 5 1.200, li gg K 2.

= 0.0222

.&2" 1+TS4 55 =

1 + TsS The function generated by the lead-lag controller for T,yg dynamic compensation, EE T4* ts =

gg Time constants utilized in the lead-lag controller for T,,g, 14 > 28 sec, is I 4 sec.,

oa 33 T .= Average temperature, 'F, nw

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=

3, 3 Lag compensator on measured T,yg,

• f TABLE 2.2-1 (Continued) k . REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS S

N NOTATION (Continued) e NOTE 1: (Continued) 3 d r, =

- 3 / secTime constant utilized in the measured T""8 lag compensator, rs< l R T' =

5 588.2*F Reference T,yg at RATED THERMAL POWER, K

3

= 0.001095, P = Pressurizer pressure, psig, P' = 2235 psig (Nominal RCS operating pressure),

u 5 = Laplace transform operator, sec 8, d

and f (AI) is a function of the indicated difference between top and bottom detectors of the power-range nuclear fon chambers; with gains to be selected based on measured inst.rument response during plant startup tests such that:

(i) for q ~9 b between -29% and +9.0%;y f (AI) = D, where gt "" 9 are percent RATED t b [

THERMAL POWER in the top and bottom halves of the core respectively, and qt+qb is total THERMAL POWER in percent of RATED THERMAL POWER; g (ii) for each percent that the magnitude of qt"9b exceeds -29%, the AT Trip Setpoint

,3 c+ e+ shall be automatically reduced by 3.151% of its value at RATED THERMAL POWER; and yy (iii) for each percent that the magnitude of qf gbexceeds +9.0%, the AT Trip setpoint '

shall be automatica11y reduced by 1.50% of its value at RATED THERMAL POWER.

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TABLE 2.2-1 (Continued)

.c y REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 1 k NOTATION (Continued) e C

%, T = .As defined in Note 1, v *

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~ T" =

< 588.2*F Reference T ,g at RATED THERMAL POWER,

=

E S = As defined in Note 1, and f 2(AI) = 0 for all AI.

Note 3: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 3,(, g cc 12ints $'EKf9-. Vo:isid.

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LIMITING SAFETY SYSTEM SETTINGS BASES (wrrH RT3 69PA55 Syyrm /N6fhG)

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 i 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 I power range nuclear detectors, the Reactor trip is automatically reduced i according to the notations in Table 2.2-1.

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 overtemperature delta T protection, and provides a backup to the High Neutron Flux trip. The Setpoint i 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 temperature for dynamic compensation for piping delays from the core to the loop temperature detectors, and (3) axial power distribution,The to ensure that Overpower AT the allowable heat generation rate (kW/ft) is not exceeded.

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 2-3 4cL

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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 p4p4eg territ d:1:y: #m- th: : r: te the t: per:tur: det;;ters (ebeut 4 se;;r,d;),

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 I 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, t!is 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 accordina to the notations in Table 2.2-1.

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Overpower AT 3ccatcf.h ,

I The Overpower Delta T trip provides assurance of fuel integrity (e.g., no I fuel pellet melting and less than 1% cladding strain) under all possible l overpower conditions, limits the required range for oyertemperature delta T l 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 tempe ature t for dynamic compensatio % r.r. , ,- - ..

cop temperature detectors, and (3) axial power distribution, to ensure that the allowable heat generation rate (kW/ft) is not exceeded. The Overpower AT 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 B 2-5

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C REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES I 5

) d FUNCTIONAL UNIT 1 ** RESPONSE TIME e, 1. Manual Reactor Trip

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N.A.

ro 2. Power Range, Neutron Flux 4

10.5 second4 (1)

! 3. Power Range, Neutron Flux, j High Positive Rate N. A.

4. Power Range, Neutron Flux, High Hegative Rate 10.5 second*(s')

5. Intermediate Range, Neutron Flux N.A.

y 6. Source Range, Neutron Flux t e N.A.

7. Overtemperature AT 1 k secondsE(.8)(.C(3}

8. Overpower AT

< NsecondsM002.)(3) -

9. Pressurizer Pressure--Low FF < 2.0 seconds

10. Pressurizer Pressure--High l

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33 1 2.0 seconds j gg 11. Pressurizer Water Cevel--High N.A.

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i g) Neutron detectors are exempt from response time testing. Response time of the neutron flux signal portion E2 of the channel shall be measured from detector output or input of first electronic component in channel.

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ATTACHMENT III Safety Analysis and Technical Justification

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e Safety Analysis and Technical Justification Removal of the RTD Bypass System impacts the Technical Specifications in two areas. First, the new RdF RTDs have an uncertainty which is greater than the uncertainty associated with the Rosemount RTDs cur-rently in use at McGuire to measure hot leg and cold leg temperatures.

A change in temperature uncertainty affects the overtemperature AT, overpower AT, and loss of flow setpoints. It should be noted that the dependence of the loss of flow setpoint on temperature error is due to 1the fact that the elbow taps are calibrated via the flow calculated from the precision heat balance. Westinghouse concluded that sufficient 4'

margin was available to justify the current values for the overtemper-ature AT, overpower AT, and loss of flow setpoints. However, the allowable values for these setpoints were recalculated to reflect the redistribution of setpoint uncertainties and available margin.

Second, removal of the RTD Bypass System impacts the response time associated with the overtemperature AT and overpower AT trip functions.

, Originally, the total response time of the McGuire RTD Bypass System assumed in the accident analyses was 6 seconds. Recent analyses were performed to increase the lag time constant in the Tavg and AT channels from 2 to 6 seconds, resulting in an overall response time used in the accident analyses of 10 seconds. Of these 10 seconds,2 seconds were

included to account for the piping delay between the core and the RTDs 4

in the bypass loops. The response time given in the Technical Specifi-cations accounts for the delay from the time the parameter exceeds its setpoint at the sensor untilloss of stationary gripper coil voltage.

, Thus, the 2 second piping delay is not included in this response time resulting in the 8 second response time currently in the Technical

, Specifications. Due to the inherent filtering effect present in the proposed thermowell RTD combination, the filter time constant in the Tavg and AT channels has been decreased from 6 seconds to 3 seconds.

The result of the increased delay associated with the thernovell RTD combination (5.5 seconds vs. 0.5 seconds) and the decreased filter time constant is a total response time equivalent to the 10 second

_ delay assumed in the accident analyses. Since the proposed fast j_

' response thermowell RTD system does not have a piping delay associated with it, the 10 second response time is incorporated in the Technical '

p. Specifications. The above described Technical Specification changes completely reflect the assumptions and methods described in the RTD

, Bypass Elimination Licensing Report in order to justify installation of the fast-response RTD thermowell system at McGuire, i

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ATTACHMENT IV Significant Hazards Consideration Analysis D

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In accordance with the requirements of 10 CFR 50.91, folloting is an analysis which supports the conclusion that the Technical Specification changes contained in Attachment II do not represent a significant hazard, as defined in 10 CFR 50.92; 1.e:

The proposed amendments would not:

Involve a significant increase in the probability or consequences of and accident previously evaluated; or 1

Create the possibility of a new or different kind of accident from any accident previously evaluated; or Involve a significant reduction in a margin of safety.

j 1. The proposed amendments will not involve a significant increase in the probability or consequences of an accident previously evaluated. The 1

modification and Tech Spec changes will provida equivalent temperature detection capability with RIDS installed in each loop, in place of the existing bypass system, for input to the Reactor Control and Protection System (RCPS). Three narrow-range RTDs in each Hot Leg will provide input for reactor coolant loop differential temperature and average coolant temperature. One narrow-range RID will be installed in the cold leg (at the disc'harge of the Reactor Coolant Pump), as well as an additional narrow-range RTD installed as a spare. The modification will not affect the wide-range RTDs currently installed in the hot and cold legs of each loop.

These wide-range RTDs are used to monitor reactor coolant temperatura during startup, shutdown, and post-accident conditions.

The RPCS parameters which are affected by narrow-range RTD accuracy have been analyzed to assure that sufficient allowance is availabla in the RPCS setpoints to accomodate RTD error. It has been determined that Departure-from Nucleate-Boiling (DNB) transients which are analyzed in Chapter 15 of the Final Safety Analysis Report (FSAR) remain conservative and need not ba reanalyzed. Transients for which DNB is not a concern (or not the only concern) were reanalyzed and found acceptable. The Instrumentation and Control portion of the modification has been evaluated and remains function-ally unchanged and physically equivalent to the existing hardware, and

meets applicable criteria, including (but not limited to) 10 CFR 50 General 4

Design Criteria, IEEE-279-1971 and other industry standards.

2. The proposed amendments will not create the possibility of's new or different kind of accident from any accident previously evaluated. The modification involves the removal of a section of reactor coolant loop bypass piping, and the replacement of certain associated instrumentation and circuitry.

The only credible occurences which could result are a small-break Loss-of-Coolant-Accident (LOCA), or a failure of the instrumentation to provide proper input to the RPCS. The possibility and consequences of a LOCA have been exhaustively analyzed, and therefore are not relevant to this item (LOCAs are implicitly included in Item 1, above). The failure of the instrumentation to provida proper input to the RPCS is not indicative of or a precursor to an unanalyzed accident; at worst such a failure would l lead to a reactor trip. '

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3. The proposed amendments will not involve a significant reduction in a margin of safety. Changes to instrument response times and uncertainties have been determined, through test and analysis, to be consistent with, or not significantly different from, current values. The increased response time of the RTDs is partially offset by the elimination of the delay associated with the bypass manifold piping..and partly by the reduction of the RID electronic filter time constant. Evaluations of uncertainties associated with the modification confirm that the setpoints defined in the McGuire Technical Specifications remain valid.

Based on the above, Duke Power Company concludes that the proposed amend-ments do not involve a Significant Hazards consideration.

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ATTACHMENT V Non-Proprietary RTD Bypass Elimination Licensing Report