ML20077E795

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Proposed Tech Specs,Revising Plant Protection Sys Trip Setpoint & Allowable Values for Consistency W/Current Setpoint/Uncertainty Methodology Being Implemented at Facility
ML20077E795
Person / Time
Site: Waterford Entergy icon.png
Issue date: 12/09/1994
From:
ENTERGY OPERATIONS, INC.
To:
Shared Package
ML20077E783 List:
References
NUDOCS 9412130131
Download: ML20077E795 (13)
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TABLE 2.2-1 U- .i k REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS E'

3 8 FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES -

1. Manual Reactor Trip Not Applicable Not Applicable 5 2. Linear Power Level - High w

Four Reactor Coolant Pumps i 110.1% of RATED THERMAL POWER 1 110.7% of RATED THERMAL POWER Operating

3. Logarithmic Power Level - High (1) 1 0.257% of RATED THERMAL POWER 1 0.275% of RATED THERMAL POWER
4. Pressurizer Pressure - High 1 2365 psia 1 2372 psia 1
5. Pressurizer Pressure - Low > 1684 psia (2)

_ > 1644 psia (2) m 6. Containment Pressure - High i 17.1 psia i 17.3 psia

7. Steam Generator Pressure - Low > 764 psia (3) > 748 psia (3)
8. Steam Generator Level - Low-

> 27.4% (4)~

_ > 26.7% (4)

9. Local Power Density - High 5 21.0 kW/ft (5) 1 21.0 kW/ft (5) rTl
10. DNBR - Low > 1.26 (5) > 1.26 (5)

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11. Steam Generator Level - High i 87.7% (4) 1 88.4% (4)  :
12. . Reactor Protection System Logic Not Applicable Not Applicable Z

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13. Reactor Trip Breakers Not Applicable Not Applicable 5 14. Core Protection Calculators Not Applicable Not Applicable f 15. CEA Calculators Not Applicable Not Applicable j
16. Reactor Coolant Flow - Low > 23.8 psid (7) > 23.6 psid (7) 9412130131 941209 PDR ADOCK 05000382 P PDR

g TABLE 3.3-4 (Continued) -

lN ENGIE ERED SAFETY FEATURES ACTUATION SYSTEN INSTRUMENTATION TRIP VALUES

o
5 ALLOWABLE i

, FUNCTIONAL UNIT TRIP VALUE VALUES 5 5. SAFETY INJECTION SYSTEM Sist RECIRCULATION (RAS)

- l l -<

lw a. Manual BAS (Trip Buttons) Not Applicable Not Applicable

b. Refueling Water Storage Pool - Low 10.0E (57,967 gallons) 9.3% (53,910 gallons) ,.

i c. Automatic Actuation Logic Not Applicable Not Applicable ,

6. LOSS OF PO ER r
a. 4.16 kV Emergency Bus Undervoltage > 3245 volts > 3245 volts '

(Loss of Voltage)

.. b. 480 V Emergency Bus Undervoltage > 372 volts > 354 volts D c. 4.16 kW Emergency Bus Undervoltage

y (Begraded Voltage) > 3875 volts > 3860 volts l

!U 7. EERGENCY FEEDWATER (EFAS)

a. Manual (Trip Buttons) Not Applicable Not Applicable
b. Steam Generator (1&2) Level - Low > 27.4X I3) I4) 1E7E I3) I4)
c. ' Steam Generator AP - High (SG-1 > SG-2) i 127.6 psid i 136.6 psid @
d. < 127.6 psid < 136.6 psid
e. SteamGeneratorAP'-High(SG-2>SG-1)k764 Steam Generator (1&2) Pressure - Low I2) psia k-748 psia (2) @j f.

g.

Automatic Actuation Logic Control Valve Logic (Wide Range Not Appilcable Not Applicable (

g SG Level - Low) > 36.3%(3) (5) > 35.3%(3) (5)

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EX SUG BASES 2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

The Reactor Coolant System components are designed to Section III, IS74 Edition, of the ASME Code for Nuclear Power Plant Components which persi'.s a maximum transient pressure of 110% (2750 psia) of design pressure.

'ihe Safety Limit of 2750 psia is therefore consistent with the design criteria i ed associated code requirements.

The entire Reactor Coolant System is hydrotested .at 3125 psia to demonstrate integrity prior to initial operation.

2.2.1 REACTOR TRIP SETPOINTS The Reactor Trip Setpoints specified in Table 2.2-1 are the values at which the Reactor Trips are set for each functional unit. The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their Safety Limits during normal operation and design basis anticipated operational occurrences and to assist the Engineered  !

Safety Features Actuation System in mitigating the consequences of accidents.

Operation with a trip set less conservative than its Trip 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 drift allowance assumed for each trip in the safety analyses.

The DN8R - Low and Local Power Density - High are digitally generated trip setpoints based on Limiting Safety System Settings of 1.26 and 21.0 kW/ft, respectively. Since these trips are digitally generated by the Core Protection Calculators, the trip values are not subject to drifts common to trips generated by analog type equipment. The Allowable Values for these trips are therefore the same as the Trip Setpoints.

To maintain the margins of safety assumed in the safety analyses, the calculations of the trip variables for the DN8R - Low and Local Power Density -

High trips include the measurement, calculational and processor uncertainties and dynamic allowances as defined in the latest applicable revision of CEN-305-P, " Functional Design Requirements for a Core Protection Calculator" and CEN-304-P, " Functional Design Requirements for a Control Element Assembly Calculator."

I WATERFORD - UNIT 3 B 2-2 AMENOMENT NO. 12 i

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EXISTI%

3/4.3 TNS f dENTATION BASES ,

. 3/4.3.1 and 3/4.3.2 ROCTOR PROTECTIVE AND ENGINEERED SAFETY FEATU ACTUATION SY5 TEM 5 IN5TRJMENTATION The OPERABILITY of the Reactor Protective and Engineered Safety Features Actuation Systems instrumentation and bypasses ensures that (1) the associated Engineered Safety Features Actuation action and/or reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its setpoint,-(2) the specified coincidence logic.is maintained (3) sufficient redundancy is maintained to permit a channel to be out.of service for testing or maintenance, and (4) sufficient system functional capability is available from diverse parameters.

The OPERABILITY of these systems is required to provide the overall reliability, redundancy, and diversity assumed available in the facility design l for the protection and mitigation of accident and transient conditions. The integrated used in theoperation of each of these systems is consistent with the assumptions safety analyses.

The redundancy design of the Control Element Assembly calculators (CEAC) provides reactor protection in the event one or both CEACs become inoperable.

If one at leastCEACeveryis4in test or inoperable, verification of CEA position is performed houri.- If the second CEAC fails, the CPCs will use DN8R and LPD penalty factors to restrict reactor operation to some maximum fraction of RATED THERMAL POWER.

will occur. If this maximum fraction is exceeded, a reactor trip The Surveillance Requirements specified for the's~e systems ensure that the overall system functional capability ~is maintained comparable to the original design standards. i The periodic surveillance tests performed at the minimum  ;

" equencies are sufficient to demonstrate this capability. The quarterly fre-jency for the channel functional tests for these systems comes from the analy-

.es presented in topical report CEN-327: RPS/ESFAS Extended Test Interval Evaluation, as supplemented.

The measurement of response time at the specified frequencies provides assurance that the protective and ESF action function associated with each channel is 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 asasurements provided that such tests demonstrate the 1otal channel response time as defined. Sensor response time verification may be demonstrated by either (1) in place, onsite, or offsite test seasurements or (2) utilizing replacement sensors with certified response times.

3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION The OPERASILITY of the radiation monitoring channels ensures that: 1 (1) the radiation levels are continually measured in the areas served by the WATERFGRD - UNIT 3 B 3/4 3-1 AMENDMENT NO.69

INSTRUMENTATION. .:XLSE\ G BASES individual channels; (2) the alare or automatic action is initiated when the radiation level trip setpoint is exceeded; and (3) sufficient information is available following anon selected plant parameters to monitor and assess these variables accident.

This capability is consistent with the recommendations '

of Regulatory Guide 1.97, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," December 1980 'and NUREG-0737, " Clarification of TMI Action Plan Requirements," November 1980.

3/4.3.3.2 INCORE DETECTORS The OPERABILITY of the incore detectors with the srecified minimum complement of equipment ensures that the measurements 60tained from use of this reactor system core. accurately represent the spatial neutron flux distribution of the 1

3/4.3.3.3 SEISMIC INSTRUMENTATION  !

. The OPERA 8ILITY of the seismic instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety.

is required to permit comparison'of the measured response to that n ed in theThis capability design pursuantbasis for the"A" to Appendix facility to determine if plant shutdown is required

! of 10 CFR Part 100. The instrumentation is consistent with the recommendations of Regulatory Guide 1.12, " Instrumentation for Earthquakes," April 1974.

3/4.3.3.4. METEOROLOGICAL INSTRUMENTATION The OPERABILITY of the meteorological instrumentation ensures that sufficient meteorological data are available for estimating potential radiation doses to thethe to public atmosphere. as a result of routine or accidental release of radioactive materials This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public i and is consistent with the recommendations of Regulatory Guide 1.23 "Onsite Meteorological Programs," February 1972.

3/4.3.3.5 REMOTE SHUTDOWN INSTRUMENTATION The OPERABILITY of the remote shutdown instrumentation ensures that sufficient capability is available to permit shutdown and maintenance of HOT STANDBY of the facility from locations outside of the control room. This espability is required in the event control room habitability is lost and is consistent with General Design Criterion 19 of 10 CFR Part 50.

WATERFORD - UNIT 3 8 3/4 3-2

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TABLE 2,2-1 REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES

1. Manual Reactor Trip Not Applicable Not Applicable
2. Linear Power Level - High Four Reactor Coolant Pumps Operating s 110.1% of RATED THERMAL POWER s 110 ~ r FA fHERMAL POWER
3. Logarithmic Power Level - High (1) s 0.257% of RATED THERMAL POWER s0'a " . RATED THERMAL POWER l
4. Pressurizer Pressure - High s 2365 psia s 2J
5. Pressurizer Pressure - Low 2 1684 psia (2) 2 M441649.7 psia (2) l~U T
6. Containment f. essure - High s 17.1 psia s U--317.4 psia lC T
7. Steam Generator Pressure - Low 2 764 psia (3) 2 748 749.9 psia (3) lO
8. Steam Generator Level - Low 2 27.4% (4) 2 26-7%26.48% (4) lCD

[yg

9. Local Power Density - High s 21.0 kW/ft (5) s 21.0 kW/ft (5)
10. DNBR - Low 2 1.26 (5) 2 1.26 (5)
11. Steam Generator Level - High s 87.7% (4) s 88-4% 88.62% (4) l
12. Reactor Protection System logic Not Applicable Not Applicable
13. Reactor Trip Breakers Not Applicable Not Applicable
14. Core Protection Calculators Not Applicable Not Applicable
15. CEA Calculators Not Applicable Not Applicable
16. Reactor Coolant Flow - Low 2 24419.00 psid (7) 2 2& 6 18.47 psid (7) l WATERFORD - UNIT 3 2-3 AMENDMENT NO. 12

TABLE 3.3-4 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES ALLOWABLE FUNCTIONAL UNIT TRIP VALUE VALUES-

5. SAFETY INJECTION SYSTEM SUMP RECIRCULATION (RAS)
a. Manual RAS (Trip Buttons) Not Applicable Not Applicable
b. Refueling Water Storage Pool - Low 10.0% (57.967 gallons) 9-e% 9.08% (63-9M 52.634 gallons) l C. Automatic Actuation Logic Not Applicable Not Applicable
6. LOSS OF POWER
a. 4.16 kV Emergency Bus Undervoltage 2 3245 volts 2 3245 volts (Loss of Voltage)
b. 480 V Emergency Bus Undervoltage 2 372 volts 2 354 volts
c. 4.16 kV Emergency Bus Undervoltage (Degraded Voltage) 2 3875 volts 2 3860 volts T
7. EMERGENCY FEEDWATER (EFAS) y
a. Manual (Trip Buttons) Not Applicable Not Applicable
b. Steam Generator (1&2) Level - Low 2 27.4%C3) (4) 2 26.7%C3) (4)
c. Steam Generator AP - High (SG-1 > SG-2) s 127.6 psid s 136.6 psid g
d. Steam Generator AP - High (SG-2 > SG-1) s 127.6 psid s 136.6 psid
e. Steam Generator (1&2) Pressure - Low 2 764 psia (2) 2 748 psia (2) f, Automatic Actuation Logic Not Applicable Not Applicable 9 Control Valve Logic (Wide Range 2 36.3%(3) (5) 2 35.3%C3) (5)

SG Level - Low) l WATERFORD - UNIT 3 3/4 3-20 AMENDMENT NO. 74

9030S D BASES 2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant ,

System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

The Reactor Coolant System components are designed to Section III,1974 Edition, of the ASME Code for Nuclear Power Plant Components which permits a maximum transient pressure of 110% (2750 psia) of design pressure. The Safety Limit of 2750 psia is therefore consistent with the design criteria and associated code requirements.

The entire Reactor Coolant System is hydrotested at 3125 psia to demonstrate integrity prior to initial operation.

2.2.1 REACTOR TRIP SETPOINTS The Reactor Trip Setpoints specified in Table 2.2-1 are the values at which the.

Reactor Trips are set for each functional unit. The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their Safety Limits during normal operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuation System in mitigating the consequences of accidents. RPS Trio Setpoints values are determined by means of an explicit setpoint calculation analysis. A Total loop Uncertainty (TLU) is calculated for each RPS instrument channel. The Trio setpoint is 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 Trio Setpoint and the Analytical Limit to account for RPS cabinet Periodic Test Errors (PTE) which are present during a CHANNEL FUNCTIONAL TEST. PTE combines RPS cabinet reference accuracy, calibration equipment errors (M&TE), and RPS cabinet bistable dri f t. Periodic testing assures that actual setpoints are within their Allowable Values. -I 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 1 Trip 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 d4 4t PTE allowance assumed for each trip in the safety analyses.

The DNBR - Low and Local Power Density - High are digitally generated trip setpoints based on Limiting Safety System Settings of 1.26 and 21.0 kW/ft. l respectively. Since these trips are digitally generated by the Core Protection Calculators, the trip values are not subject to drifts common to trips generated by analog type equipment. The Allowable Values for these trips are therefore the same as the Trip Setpoints.

To maintain the margins of safety assumed in the safety analyses, the calculations l of the trip variables for the DNBR - Low and Local Power Density -High trips include the  ;

measurement, calculational and processor uncertainties and dynamic allowances as defined i in the latest applicable revision of CEN-305-P, " Functional Design Requirements for a Core Protection Calculator" and CEN-304-P. " Functional Design Requirements for a Control Element Assembly Calculator."  ;

1 WATERFORD - UNIT 3 B 2-2

~

PROPOSED 3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR PROTECTIVE AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEMS INSTRUMENTATION The OPERABILITY of the Reactor Protective and Engineered Safety Features Actuation Systems instrumentation and bypasses ensures that (1) the associated Engineered Safety Features Actuation action and/or reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its setpoint. (2) the specified coincidence logic is maintained. (3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and (4) sufficient system functional capability is available from diverse parameters.

The OPERABILITY of these systems is required to provide the overall reliability, redundancy, and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions. The integrated operation of each of these systems is consistent with the assumptions used in the safety analyses.

The redundancy design of the Control Element Assembly Calculators (CEAC) provides reactor protection in the event one or both CEACs become inoperable. If one CEAC is in test or inoperable, verification of CEA position is performed at least every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. If the second CEAC fails, the CPCs will use DNBR and LPD penalty factors to restrict reactor operation to some maximum fraction of RATED THERMAL POWER. If this maximum fraction is exceeded, a reactor trip will occur.

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 Setpoint 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 Trio 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 setpoints are within their Allowable 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.

WATERFORD - UNIT 3 8 3/4 3-1 AMENDMENT NO. 69 o

. ?30 POSED 3/4.3 INSTRUMENTATION BASES (Cont'd) +

3/4.3.1 and 3/4.3.2 REACTOR PROTECTIVE AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEMS INSTRUMENTATION The measurement of response time at the specified frequencies provides assurance that the protective and ESF action function associated with each channel is 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 demonstrated by either (1) in place, onsite. or offsite test measurements or (2) utilizing replacement sensors with certified response times, I

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WATERFORD - UNIT 3 8 3/4 3-la l

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?30?OSD l INSTRUMENTATION BASES 3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring channels ensures that:

(1) the radiation levels are continually measured in the areas served by the individual channels; (2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded; and (3) sufficient information is available on selected plant parameters to monitor and assess these variables following an accident. 'This capability is consistent with the recommendations of Regulatory Guide 1.97 " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," December 1980 and NUREG-0737, " Clarification of TMI Action Plan Requirements," November 1980, 3/4.3.3.2 INCORE DETECTORS The OPERABILITY of the incore detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the reactor core.

3/4.3.3.3 SEISMIC INSTRUMENTATION The OPERABILITY of the seismic instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety. This capability is required to permit comparison of the measured response to that used in the design basis for the facility to determine if plant shutdown is required pursuant to Appendix "A" of 10 CFR Part 100. The instrumentation is consistent with the recommendations of Regulatory Guide 1.12

" Instrumentation for Earthquakes," April 1974.

3/4.3.3.4 METEOROLOGICAL INSTRUMENTATION The OPERABILITY of the meteorological instrumentation ensures that sufficient meteorological data are available for estimating potential radiation doses to the public as a result of routine or accidental release of radioactive materials to the atmosphere.

This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public and is consistent with the recommendations of Regulatory Guide 1.23 "Onsite Meteorological Programs," February 1972.

3/4.3.3.5 REMOTE SHUTDOWN INSTRUMENTATION The OPERABILITY of the remote shutdown instrumentation ensures that sufficient capability is available to permit shutdown and maintenance of HOT STANDBY of the facility from locations outside of the control room. This capability is required in the event control room habitability is lost and is consistent with General Design Criterion 19 of 10 CFR Part 50.

WATERFORD - UNIT 3 B 3/4 3-2

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