ML20151W597

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Proposed Tech Specs Re Notes in Table 2.2-1 Re Reactor Protective Instrumentation Trip Setpoints Limits & Table 3.3-1 Re Reactor Protective Instrumentation
ML20151W597
Person / Time
Site: Waterford Entergy icon.png
Issue date: 09/11/1998
From:
ENTERGY OPERATIONS, INC.
To:
Shared Package
ML20151W582 List:
References
NUDOCS 9809160042
Download: ML20151W597 (18)
v • d • e


Text

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f NPF-38-210 l ATTACHMENT A

EXISTING SPECIFICATIONS i

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9809160042 980911 PDR  !"

ADOCM 05000382 PDR a 4 - P -- ' '"

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t TABLE 2.2-1 REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPolNT LINITS ~

FUNCTIONAL UNIT u -TRIP SETPOINT ALLOWABLE VALUES i

1. Manual Reactor Trip Not Applicable Not Applicable
2. Linear Power Level - High
  • Four Reactor Coolant Pumps s 108% of RATED THERMAL POWER Operating S 108.76% of RATED THERMAL POWER i

! 3. Logarithmic Power Level - High (1) 5 0.257% of RATED THERNAL POWER '

s 0.280% of RATED THERMAL- POWER

4. Pressurizer Pressure - High 5 2350 psia 1 2359 psla'
5. Pressurizer Pressure - Low 2 1684 psia (2) 1 1649.7 psia (2) i
6. Containment Pressure - High $ 17.1 psia s 17.4 psia
7. Steam Generator Pressure - Low 2 764 psia (3) 2 749.9 psia (3)
8. Steam Generator level - Low 2 27.4% (4) 2 26.48% (4)  !
9. Local Power Density - High 5 21.0 kW/ft (5) s 21.0 kW/ft (5)
10. DN8R - Low 2 1.26 (5) 2 1.26 (5)
11. Steam Generator Level - High 5 87.7% (4) 5 88.62% (4) l I
12. Reactor Protection System logic Not Applicable Not Applicable
13. Reactor Trip Breakers Not Applicable Not Applicable i
14. Core Protection Calculators Not Applicable Not Applicable t
15. CEA Calculators Not Applicable Not Applicable
16. Reactor Coolant Flow - Low 2 19.00 psid (7) 2 18.47 psid (7) l
WATERFORD - UNIT 3 2-3 -

Amendment No. M ,113-

TABLE 2.2-1 (Continued)

REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS TABLE NOTATIONS (1) Trip may be manually bypassed above 10 4% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is less than or equal to 10 4% of RATED THERMAL POWER.

(2) Value may be decreased manually, to a minimum of 100 psia, as pressurizer pressure is reduced, provided the margin between the pressurizer pressure and this value is maintained at less than or equal to 400 psi; the setpoint shall be increased automatically as pressurizer pressure is increased until the trip setpoint is reached. Trip may be manually bypassed below 400 psia; bypass shall be automatically removed whenever pressurizer pressure is greater than or equal to 500 psia.

(3) Value may be decreased manually as steam generator pressure is reduced, provided the margin between the steam generator pressure and this value is maintained at less than or equal to 200 psi; the setpoirl shall be increased automatically as steam generator pressure is increased until the trip setpoint is reached.

(4) % of the distance between steam generator upper and low level instrument nozzles.

(5) As stored within the Core Protection Calculator (CPC). Calculation of the trip setpoint includes measurement, calculational and processor uncertainties, and dynamic allowances. Trip may be manually bypassed below 10 4% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is greater than or equal to 10 4% of RATED THERMAL POWER. -

(6) Note 6 has been deleted.

(7) The setpoint may be altered to disable trip function during testing j pursuant to Specification 3.10.3.

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WATERFORD - UNIT 3 2-4 AMENDMENT NO. 12 l

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TABLE 3.3-1 5

REACTOR PROTECTIVE INSTRUMENTATION E

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= MINIMUN

. TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION Q 1. Manual Reactor Trip 2 sets of 2 1 set of 2 2. sets of 2 1 2 1 w 2 sets of 2 1 set of 2 2 sets of 2 3A , 4* , 5* 8

2. Linear Power Level - High 4 2 3 1, 2 2#, 3#
3. Logarithmic Power Level-High .
a. Startup and Operating 4 2(a)(d) 3 2** 2#, 3#

4 2 3 3*, 4*, 5* 8

b. Shutdown 4 0 2 3, 4, 5 4
4. Pressurizer Pressure - High 4 2 3 1, 2 2#, 3#

y 5. Pressurizer Pressure - Low 4 2(b) 3 1, 2 2#, 3#

w 6. Containment Pressure - High 4 2 3 1, 2 2#, 3#

7. Steam Generator Pressure - Low 4/SG 2/SG 1, 2 3/SG 2#, 3#
8. Steam Generator Level - Low 4/SG 2/SG 3/SG 1, 2 2#, 3#
9. Local Power Density - High 4 2(c)(d) 3 1, 2 2#, 3#

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10. DNBR - Low 4 2(c)(d) ~3 1, 2 2#, 3#
11. Steam Generator Level - High 4/SG 2/SG(g) 3/SG 1, 2 2#, 3# ~
12. Reactor Protection System Logic 4 2 3 1 2 5  ;

R m 3I , 4* , 5* 8 E 13. Reactor Trip Breakers 4 2(f) 4 1 2 5 i5i 3I , 4* , 5* 8

14. Core Protection Calculators 4 2(c)(d) 3 1, 2 2#, 3# and 7 '

P 15. CEA Calculators 2 2(e) 1 1, 2 6 and 7

,M 16. Reactor Coolant Flow - Low 4/SG 2/SG(c) 3/SG 1, 2 - 2#, 3# l O

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TABLE 3.3-1 (Continued)

TABLE NOTATION j *With the protective system trip breakers in the closed position, the CEA drive system capable of CEA withdrawal, and fuel in the reactor vassel.

j #The provisions of Specification 3.0.4 are not applicable.

    • Not applicable above 10'S RATED THERMAL POWER.

i i (a) Trip may be manually bypassed above 10'S of RATED THERMAL POWER; bypass

! 10'5 of RATED THERMAL POWER.shall be automatically removed whe i

j (b) Trip may be manually bypassed below 400 psia; bypass shall be auto- I j

matica11y to 500 psia. removed whenever pressurizer pressure is greater than or equal (c) Trip may be manually bypassed below 10'S of RATED THERMAL POWER; bypass i to 10'5 of RATED THERMAL POWER.shall be automatically removed w t i During testing pursuant to Special Test  !

' Exception 3.10.3, trip may be manually bypassed below 5% of RATED THERMAL '

POWER; bypass shall be automatically removed when THERNAL POWER is greater than or equal to 5% of RATED THERMAL POWER. '

s j (d) Trip may be bypassed during testing pursuant to Special Test Exception 3.10.3.

(e) See Special Test Exception 3.10.2. l i

(f) Each channel shall be comprised of two trip breakers; actual trip logic j shall be one-out-of-two taken twice.

(g)

High steam generator level trip may be manually bypassed in Modes 1 and 2, at 20% power and below.

j ACTION STATEMENTS

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ACTION 1 -

i With the number of channels OPERABLE one less than required by i the Minimum Channels OPERABLE requirement, restore the 1 inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and/or open the j protective system trip breakers.

I ACTION 2. -

i With the number of channels OPERABLE one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may continue provided the inoperable channel is placed in the bypassed or

tripped condition within I hour. If the inoperable channel is bypassed, the desirability of maintaining this channel in the bypassed condition shall be documented by the Plant Operations
Review Committee in accordance with plant administrative procedures.

j The channel shall be returned to OPERABLE status prior to STARTUP following the next COLD SHUTDOWN.

WATERFORD - UNIT 3 3/4 3-4 AMENDMENT NO. W 44,109 1

-.- -- - - - - ~ . - - . - . - - . - . . ~ - - - . - - -. .- -

a BASES Manual Reactor Trio The Manual Reactor Trip is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability.

Linear Power Level - Hiah The Linear Power Level - High trip provides reactor core protection ~

against rapid reactivity excursions which might occur as the result of an -

ejected CEA. This trip initiates a reactor trip at a linear power level of less than or equal to 108% of RATED THERMAL POWER.

l ';g Loaarithmic Power Level - Hiah The Logarithmic Power Level - High trip is provided to protect the integrity of fuel cladding and the Reactor Coolant System pressure boundary in the event of an unplanned criticality from a shutdown condition. A reactor trip is initiated by the Logarithmic Power Level - High trip at a THERMAL POWER level of less than or equal to 0.257% of RATED THERMAL POWER unless this I

trip is manually bypassed by the operator. The operator may manually bypass this trip when the THERMAL POWER level is above 10-4% of RATED THERMAL POWER; this bypass is automatically removed when the THERMAL POWER level decreases to 10-4% of RATED THERMAL POWER.

Pressurizer Pressure - Hiah l

The Pressurizer Pressure - High trip, in conjunction with the pressurizer safety valves and main steam safety valves, provides Reactor Coolant System protection hyainst overpressurization in the event of loss of load without reactor trip. This trip's setpoint is at less than or equal to 2350 psia which is below the nominal lift setting of 2500 psia for the pressurizer safety valves and its operation avoids the undesirable operation of the pressurizer safety valves.

l Pressurizer Pressure - Low

_ The Pressurizer Pressure - Low trip is provided to trip the reactor and to assist the Engineered Safety Features System in the event of a Loss of Coolant Accident. During normal operation, this trip's setpoint is set at greater than or equal to 1684 psia. This trip's setpoint may be manually l decreased, to a minimum value of 100 psia, as pressurizer pressure is reduced j during plant shutdowns, provided the margin between the pressurizer pressure 4

and this trip's setpoint is maintained at less than or equal to 400 psi; l- this setpoint increases automatically as pressurizer pressure increases until the trip setpoint is reached.

WATERFORD - UNIT 3 B 2-3 Amendment No.113 l

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BASES DNBR - Low (Continued) in actual core DNBR after the trip will not result in a violation of the DNBR Safety Limit of 1.26. CPC uncertainties related to DNBR cover CPC input l measurement uncertainties, algorithm modelling uncertainties, and computer equipment processing uncertainties. Dynamic compensation is provided in the i l

CPC calculations for the effects of coolant transport delays, core heat flux delays (relative to changes in core power), sensor time delays, and protection  ;

system equipment time delays. '

The DNBR algorithm used in the CPC is valid only within the limits indicated below and operation outside of these limits will result in a CPC I

initiated trip.

a. RCS Cold Leg Temperature-Low > 495'F
b. RCS Cold Leg Temperature-High 7 580*F
c. Axial Shape Index-Positive Not more positive than +0.5
d. Axial Shape Index-Negative Not more negative than -0.5
e. Pressurizer Pressure-Low > 1860 psf 4 ,
f. Pressurizer Pressure-High 7 2375 psta
g. Integrated Radial Peaking Factor-Low -> 1.28
h. Integrated Radial Peaking Factor-High < 4.28
1. Quality Margin-Low FO
Steam Generator Level - Hiah The Steam Generator Level - High trip is provided to protect the turbine from excessive moisture carry over. Since the turbine is automatically tripped when the reactor is tripped, this trip provides a reliable means for providing protection to the turbine from excessive moisture carry over. This trip's setpoint does not correspond to a Safety Limit and no credit was taken

! in the safety analyses for operation of this trip. Its functional capability at the specified trip setting is required to enhance the overall reliability of the Reactor Protection System.

Reactor Coolant Flow - Low The Reactor Coolant flow - Low trip provides protection against a reactor '

coolant pump sheared shaft event and a steam line break event with a loss-of-offsite power. A trip is initiated when the pressure differential across the primary side of either steam generator decreases below a nominal setpoint of 23.8 psid. The specified setpoint ensures that a reactor trip occurs to

' prevent violation of local power density or DNBR safety limits under the stated conditions.

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WATERFORD - UNIT 3 B 2-6 AMENDMENT NO. 12

3/4 3 INSTRUMFNTATION l .-

BARFS

} 3/4 31 and 3/4 3 2 REACTOR PROTECTIVE AND ENGINFFRFr) SAFETY FEATURES

[j ACTUATION SYSTEMR INSTRUMFNTATION i=

condition while another RWSP - Low channel is in bypass, the receipt of a valid Safety injection i Actuation Signal Actuation, and a coincident failure of one of the two remaining OPERABLE RWSP - Low channels. These conditions could cause the Emergency Core Cooling System 4

and Containment Spray System suctions to be supplied from the Safety injection System Sump j prematurely due to containment pressure being higher than RWSP outlet pressure and loss of j the Low Pressure Safety injection Systems.

} When one of the inoperable channels is restored to OPERABLE status, subsequent operation in the applicable MODE (S) may continue in accordance with the provisions of l ACTION 19, 1

The Surveillance Requirements specified for these systems ensure that the overall i

system functional capability is maintained comparable to the original design standards. The i periodic surveillance tests performed at the minimum frequencies are ' sufficient to demonstrate 1

this capability The quarterly frequency for the channel functional tests for these systems l comes from the analyses presented in topical report CEN-327: RPS/ESFAS Extended Test j interval Evaluation, as supplemented.-

1 j RPS\ESFAS Trip Setpoints values are determined by means of an explicit setpoint

. calculation analysis. A Total Loop Uncertainty (TLU)is calculated for each RPSESFAS j instrument channel. The Trip Setpoint is then determined by adding or subtracting the TLU j from the Analytical Limit (add TLU for decreasing process value; subtract TLU for increasing j process value).. The Allowable Value is determined by adding an allowance between the Trip

Setpoint and the Analytical Limit to account for RPSESFAS cabinet Periodic Test Errors (PTE) 4 which are present during a CHANNEL FUNCTIONAL TEST. PTE combines the RPSESFAS 3 cabinet reference accuracy, calibration equipment errors (M&TE), and RPSESFAS cabinet i bistable Drift. Periodic testing assures that actual setpoints are withir heir Allowable Values. A l channel is inoperable if its actual setpoint is not within its Mowable Value and corrective action
must be taken. Operation with a trip set less conservativa than its setpoint, but within its j specified ALLOWABLE VALUE is acceptable on the bas.s that the difference between each trip 4 Setpoint and the ALLOWABLE VALUE is equal to or less than the Periodic Test Error allowance assumed for each trip in the safety analyses.

The measurement of response time at the specified frequencies provides assarance that the protective and ESF action function associated with each channel is completed within the i time limit assumed in the safety analyses. No credit was taken in the analyses for those

channels with response times indicated as not applicable.

I Response time < nay be demonstrated by any series of sequential, overlapping, or total channel test measurements provided that such tests demonstrate the total channel responss time as defined. Sensor response time verification may be demonstrated by either (I) in place, onsite, or offsite test measurements or (2) utilizing replacement sensors with certified response times.

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WATERFORD - UNIT 3 B 3/4 3-ia Amendment No. 440,143

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NPF-38-210 ATTACHMENT B PROPOSED SPECIFICATIONS

TABLE 2.2-1 '

REACTOR PROTECTIVE INSTRUMENTATION TRIP S.' A!NT LINITS FUNCTIONAL UNIT , IRIP SETPOINT ALLOWABLE VALULS

1. Manual Reactor Trip Not Appilcable Not Applicable .!
2. Linear Power Level - High ,

.t Four Reactor Coolant P. 7s s 108% of RATED THERMAL POWER Operating 5 108.76% of RATED THERMAL POWER l

3. Logarithmic Power Level - High (1) 5 0.257% of RATED THERNAL 00WER 5 0.280% of RATED THERNAL POWER.
4. Pressurizer Pressure - High 5 2350 psia 1 2359 psia
5. Pressurizer Pressure - Low 2 1684 psia (2) 2 1649.7 psia (2)
6. Containment Pressure - High 5 17.1 psia <; 17.4 psia
7. Steam Generator Pressure - Low 2 764 psia (3) 2 749.9 psia (3) D
8. Steam Generator Level - Low 2 27.4% (4) "

2 26.48% (4) g n

9. Local Power Density - High 5 21.0 kW/ft (5) 5 21.0 kW/ft (5) g% .

g .

10. DNBR - Low tw 2 1.26 (5) 2 1.26 (5)
  • d j
11. Steam Generator Level - High 5 87.7% (4) s 88.62% (4) l
12. Reactor Protection System Logic N L Applicable Not Applicable
13. Reactor Trip Breakers Not Applicable

{ t Not Applicable T

14. Core Protection Calculators Not Applicable Not Applicable )

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15. CEA Calculators Not Applicable Not Applicable E

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16. Reactor Coolant Flow - Low 2 19.00 psid (7) 1 18.47 psid (7) D l

WATERFORD - UNIT 3 2-3 Amendment No. 12,113

TABLE 2.2-1 (Continued)

REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS TABLE NOTATIONS (1) Trip may be manually bypassed above 10 *% of RATED THERMAL POWER; bypass shall b_e automatically removedfhen JMRMAL POWER)(less Enan optquaJ7 fto 1 7 % of RMtU IHERMAl/POWEBC (2) Value may be decreased manually, to a minimum of 100 psia, as pressurizer pressure is reduced, provided the margin between the pressurizer pressure and this value is maintained at less than or equal to 400 psi; the setpoint shall be increased automatically as pressurizer pressure is increased until the trip setpoint is reached. Trip may be manually bypassed below 400 psia; bypass shall be automatically removed whenever pressurizer pressure is greater than or equal to 500 psia.

i (3) Value may be decreased manually as steam generator pressure is reduced, I provided the margin between the steam generator pressure and this value is maintained at less than or equal to 200 psi; the setpoint shall be increased automatically as steam generator pressure is increased until the trip setpoint is reached.

(4) % of the distance between steam generator upper and low level instrument nozzles.

(5) As stored within the Core Protection Calculator (CPC). Calculation of the trip setpoint includes measurement, calculational and processor uncertainties, and dynamic allowances. Trip may be manually bypassed below 10 4% of RATED THERMAL POWER; bypass shall be automatically removed whe VIHERMAL POWFpis greater ttperor equai top-4 of RATED THpMAd (6) Note 6 has been deleted.

(7) The setpoint may be altered to disable trip function during testing pursuant to Specification 3.10.3.

% jO-4 7. 3MC WATERFORD - UNIT 3 2-4 AMEN 0 MENT NO. 12

- . . ,... - - ... - . . - . .. - . - . - . . . . . - . - . . ._....u . ~- ..

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TABLE 3.3-1

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= REACTOR PROTECTIVE INSTRUMENTATION t n -

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. TOTAL NO. CHANNELS MINIMUM CHANNELS APPLICABLE Y$A '

FUNCTIONAL UNIT OF CHANNELS TO TRIP * ,

E OPERABLE MDDES ACit0N g%  ;

M 1. Manual Reactor Trip 2 sets of 2 1 set of 2 2 isets of 2 1 2 S 1 s w 2 sets of 2 1 set of 2 2 sets of 2 3 A , 4*, 5* 8 b 2.

3.

Linear Power Level - High Logarithmic Power Level-High 4 2 3 1, 2 2#, 3# }g ,

a. Startup and Operating 4 4

2(a)(d) 2 3

3 2**

3*, 4*, 5*

2#, 3# k' gS 8  :

b. Shutdown 4 a 3, 4, 5 '

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2 4 g  !

Pressurizer Pressure - High R*

a

5. Pressurizer Pressure - Low 4

4 2

2(b) 2 3

1, 2 1, 2 28, 3#

?#, 3#

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w 6. Containment Pressure - High 4 2 3 1, 2 :f, 3#

0 7.

Steam Generator Pressure - Low 4/SG 2/SG 3/SG 1, 2 2#, 3# '

8. Steam Generator Level - Low 4/5G 2/SG 3/SG 1, 2 2#, 3# i
9. Local Power Density - High 4 2(c)(d) 3 1, 2 2#, 3#
10. DER - Low 4 2(c)(d) 3 1, 2 2#, 3#
11. Steam Generator Level - High 4/SG 2/SG(g) 3/SG 1, 2 2#, 3# j
12. Reactor Protection System Logic 4 2 3 I 2 5 I

m E 3g, 4*, 5* 8 g 13. Reactor Trip Breakers 4 2(f) 4 1 2 5 -

m 3g, 4* , 5* 8

  • 14. Core Protection Calculators t 4 2(c)(d) 3 1, 2 2#, 3# and 7 i h 15. CEA Calculators 2 1 2(e) 1, 2 6 and 7

,M 16. Reactor Coolant Flow - Low 4/SG 2/SG(c) 3/SG 1, 2 2#, 3# l

4 g4 'f 8t.s L ' le. e TABLE 3.3-1 (Continued 1 j TABLE NOTATION i

j

  • With the protective system trip breakers in the closed position, the CEA drive system capable of CEA withdrawal, and fuel in the reactor vessel.
  1. The provisions of Specification 3.0.4 are not applicable.
    • Not applicable above 10'S RATED THERMAL POWER.

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(a) Trip may be manually bypassed above 10'S of RATED THERMAL POWER; bypass j '

. shp%f RAMD THERMA 1,/UWt@ril be automatically removed penge 1

(b) Trip may be manually bypassed below 400 psia; bypass shall be auto- l matica11y removed whenever pressurizer pressure is greater than or equal zl to 500 psia. s-(c) Trip may be manually bypassed below 10'S of RATED THERMAL POWER; bypass l shall_be automatically removed /vnene h unat rouuns areaturtnan or my

' Me 20 4 af nwn m unar mrEnf During testing pursuant to Special rest Exception 3.10.3, trip may be manually bypassed below 5% of RATED THERMAL i

POWER; bypass shall be automatically removed when THERMAL POWER is greater j than or equal to 5% of RATED THERMAL POWER. '

i (d) Trip may be bypassed during testing pursuant to Special Test Exception i 3.10.3. .

(e) See Special Test Exception 3.10.2.

j (f) Each channel shall be comprised of two trip breakers; actual trip logic shall be one-out-of-two taken twice.

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(g) High steam generator level trip may be manually bypassed in Modes 1 and 2,

at 20% power and below.
ACTION STATEMENTS ACTION 1 -

With the number of channels OPERABLE one less than required by 1

the Minimum Channels OPERABLE requirement, restore the i

inoperable channel to OPERA 8LE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least HOT STAND 8Y within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and/or open the protective system trip breakers.

l ACTION 2 -

With the number of channels OPERA 8LE one less than the Total i

Number of Channels, STARTUP and/or POWER OPERATION may continue i

provided the inoperable channel is placed in the bypassed or tripped condition within I hour. If the inoperable channel is bypass,ed, the desirability of maintaining this channel in the 1

bypassed condition shall be documented by the Plant Operations Review Committee in accordance with plant administrative procedures. The channel shall be returned to 0PERABLE status

prior to STARTUP following the next COLD SHUTDOWN.

1 i WATERFORD - UNIT 3 3/4 3-4 AMENDMENT NO. % 40,109 1

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BASES Manual Reactor Trio The Manual Reactor Trip is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability.

Linear Power Level - Hioh The Linear Power Level - High trip provides reactor core protection -

against rapid reactivity excursions which might occur as the result of an ejected CEA. This trip initiates a reactor trip at a linear power level of less than or equal to 108% of RATED THERMAL POWER.

l '; .

Locarithmic Power Level - Hiah ,

The Logarithmic Power Level - High trip is provided to protect the integrity of fuel cladding and the Reactor Coolant System pressure boundary in the event of an unplanned criticality from a shutdown condition. A reactor trip is initiated by the Logarithmic Power Level - High trip at a THERMAL POWER level of less than or equal to 0.257% of RATED THERMAL POWER unless this trip is manually bypassed by the operator. The operator may manually bypass this trip when the THERMAL POWER level is above 10-4% of RATED THERMAL POWER; tMs bvoass is automatically removedfwhenXhe THEKryn. Kmid ipvei Gecreph io)

(10'4% af 88TEVHEP"al DMk.-

Pressurizer Pressure - Hiah k n 10'#'A bi d d 4 1

The Pressurizer Pressure - High trip, in conjunction with the pressurizer safety valves and main steam safety valves, provides Reactor Coolant System

' protection against overpressurization in the event of loss of load without reactor trip. This trip's setpoint is at less than or equal to 2350 psia which is below the nominal lift setting of 2500 psia for the pressurizer >

safety valves and its operation avoids the undesirable operation of the l

pressurizer safety valves.  !

Pressurizer Pressure - Los t . The Pressurizer Pressure - Low trip is provided to trip the reactor and to assist the Engineered Safety Features System in the event of a Loss of ,

Coolant Accident. During normal operation, this trip's setpoint is set at 1 greater than or equal to 1684 psia. This trip's setpoint may be manually decreased, to a minimum value of 100 psia, as pressurizer pressure is reduced during plant shutdowns, provided the margin between the pressurizer pressure and this trip's setpoint is maintained at less than or equal to 400 psi; this setpoint increases automatically as pressurizer pressure increases until I the trip setpoint is reached.

I WATERFORD - UNIT 3 B 2-3 Amendment No.113

_. ._ _ _ _ m . _ _ . _ _ _ . _ - _ _ _ _ _ _ . ~ - _ _ _ _ . _ _ . _ _ . _ . . _ _ . _ . - _ _ . . . _ _ . .

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!, BASES i

l ONBR - Low (Continued) i in actual core DN8R after the trip will not result in a violation of the DN8R i Safety Limit of 1.26. CPC uncertainties related to DNBR cover CPC input measurement uncertainties, algorithm modelling uncertainties, and computer s

equipment processing uncertainties. Dynamic compensation is provided in the CPC calculations for the effects of coolant transport delays, core heat flux delays (relative to changes in core power), sensor time delays, and protection l system equipment time delays.

i i

The DN8R algorithm used in the CPC is valid only within the limits indicated below and operation outside of these limits will result in a CPC initiated trip.

i ,

a. RCS Cold Leg Temperature-Low > 495'F

{ b. RCS Cold Leg Temperature-High < 580*F

c. Axial Shape Index-Positive Not more positive than +0.5 '

i

~i - d. Axial Shape Index-Negative Not more negative than -0.5

e. Pressurizer Pressure-Low > 1860 psia i

f

f. Pressurizer Pressure-High 7 2375 psia
g. Litograted Radial Peaking Factor-Low > 1.28
h. Integrated Radial Peaking Factor-High < 4.28
1. Quality Margin-Low >0 Steam Generator Level - High The Steam Generator Level - High trip is provided to protect the turbine from excessive moisture carry over. Since the turbine is automatically tripped when the reactor is tripped, this trip provides a reliable means for providing protection to the turbine from excessive moisture carry over. This trip's setpoint does not correspond to a Safety Limit and no credit was taken in the safety analyses for operation of this trip. Its functional capability i at the specified trip setting is required to enhance the overall reliability of the Reactor Protection System.

Reactor Coolant Flow - Low The Reactor Coolant Flow - Low trip provides protection against a reactor '

coolant pump sheared shaft event and a steam line break event with a loss-of-offsite power. A trip is initiated when the pressure differential across the primary side of either steam generator decreases below a nominal setpoint of 23.8 psid. The specified setpoint ensures that a reactor trip occurs to prevent violation of local power density or DNSR safety limits under the stated conditions.

JAM GMT p.2.[ 8AS2.5 WATERFORD - UNIT 3 B 2-6 AMENDMENT NO. 12

. 1 INSERT 2.2.1 Bases The 104 % Bistable, described in Notes (1) and (b;, automatically removes the operational bypass at a nominal THERMAL POWER of 104 % RATED THERMAL POWER with an allowance for hysteresh. -

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3/4 3 INSTRbMFNTATION BASFS 3/4 31 and 3/4 3 2 REACTOR PROTECTIVE AND ENGINFFRFD SAFETY FEATURES ACTUATION SYSTEMS INSTRUMFNTATION condition while another RWSP - Low channel is in bypass, the receipt of a valid Safety injection Actuation Signal Actuation, and a coincident failure of one of the two remaining OPERABLE RWSP - Low channels. These conditions could cause the Emergency Core Cooling System and Containment Spray System suctions to be supplied from the Safety injection System Sump prematurely due to containment pressure being higher than RWSP outlet pressure and loss of the Low Pressure Safety injection Systems.

When one of the inoperable channels is restored to OPERABLE status, subsequent operation in the applicable MODE (S) may continue in accordance with the provisions of ACTION 19.

The Surveillance Requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards. The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability. The quarterly frequency for the channel functional tests for these systems comes from the analyses presented in topical report CEN-327: RPS/ESFAS Extended Test Interval Evaluation, as supplemented.

RPS\ESFAS Trip Setpoints values are determined by means of an explicit setpoint calculation analysis. A Total Loop Uncertainty (TLU)is calculated for each RPS/ESFAS instrument channel. The Trip Seipoint is then determined by adding or subtracting the TLU from the Analytical Limit (add TLU for decreasing process value; subtract TLU for increasing process value). The Allowable Value is determined by adding an allowance between the Trip l Setpoint and the Analytical Limit to account for RPS/ESFAS cabinet Periodic Test Errors (PTE) which are present during a CHANNEL FUNCTIONAL TEST. PTE combines the RPS/ESFAS cabinet reference accuracy, calibration equipment errors (M&TE), and RPS/ESFAS cabinet bistable Drift. Periodic testing assures that actual setpo;nts are within their Ahowable Values. A channel is inoperable if its actual setpoint is not within its Allowable Value and corrective action must be taken. Operation with a trip set less conservative than its setpoint, but within its specified ALLOWABLE VALUE is acceptable on the basis that the difference between each trip Setpoint and the ALLOWABLE VALUE is equal to or less than the Periodic Test Error allowance assumed for each trip in the safety analyses.

The measurement of response time at the specified frequencies provides assurance that the protective and ESF action function associated with each channelis completed within the time limit assumed in the safety analyses. No credit was taken in the analyses for those channels with response times indicated as not applicable.

Response time may be demonstrated by any series of sequential, overlapping, or total channel test measurements provided that such tests demonstrate the total channel response time as defined. Sensor response time verification may be dernonstrated by either (i) in place, onsite, or offsite test measurements or (2) utilizing replacement sensors with certified response times.

Z N Ger 3.3./ BAJ 63 WATERFORD - UNIT 3 8 3/4 31a Amendment No. ++3,143

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l INSERT 3.3.1 Bases The 10-4 % Bistable, described in Notes (a) and (c), automatically removes the .

l operational bypass at a nominal THERMAL POWER of 10-4 % RATED THERMAL POWER with an allowance for hysteresis.

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