ML081830638

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Tech Spec Pages for Amendment 278 Regarding Measurement Uncertainty Recapture Power Uprate
ML081830638
Person / Time
Site: Davis Besse Cleveland Electric icon.png
Issue date: 06/30/2008
From:
NRC/NRR/ADRO/DORL/LPLIII-2
To:
Wengert, Thomas J, NRR/DORL, 415-4037
Shared Package
ML081420569 List:
References
TAC MD8326
Download: ML081830638 (9)
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1.0 DEFINITIONS DEFINED TERMS 1.1 The DEFINED TERMS of this section appear in capitalized type and are applicable throughout these Technical Specifications.

THERMAL POWER 1.2 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

RATED THERMAL POWER 1.3 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2817 MWt.

OPERATIONAL MODE 1.4 An OPERATIONAL MODE shall correspond to any one inclusive combination of core reactivity condition, power level and average reactor coolant temperature specified in Table 1.1.

ACTION 1.5 ACTION shall be those additional requirements specified as corollary statements to each principal specification and shall be part of the specifications.

OPERABLE - OPERABILITY 1.6 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function(s). Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubrication or other auxiliary equipment, that are required for the system, subsystem, train, component or device to perform its furiction(s), are also capable of performing their related support function(s).

DAVIS-BESSE, UNIT I 1-1 DAmendment 82,135, 278

Table 2.2-1 Reactor Protection System Instrumentation Trip Setpoints Functional uni it Allowable values

1. Manual reactor trip Not applicable.

~11 (A

2. High flux <104.9% of RATED THERMAL POWER with four pumps operating with m secondary heat balance based on ultrasonic flow meter instrumentation*

z

< 103.3 % of RATED THERMAL POWER with four pumps operating with secondary heat balance not based on ultrasonic flow meter instrumentation*

  • 80.6% of RATED THERMAL POWER with three pumps operating*
3. RC high temperature <618oF*
4. Flux -- Aflux/flow(1) Pump allowable values not to exceed the limit lines shown in the CORE OPERATING LIMITS REPORT for four and three pump operation.*
5. RC low pressure(1) >1900.0 psig*
6. RC high pressure <2355.0 psig*

-0 >

7. RC pressure-temperature ,(1) (16.25 Tou,°F - 7899.0) psig*

zz 8. High flux/number of RC pumps on(l) <55.1% of RATED THERMAL POWER with one pump operating in each loop*

.3CD

<0.0% of RATED THERMAL POWER with two pumps operating in one loop and no pumps operating in the other loop*

  • 0.0% of RATED THERMAL POWER with no pumps operating or only one pump 00 operating*
9. Containment pressure hi ,h _<4 psig*

TABLE 3.3-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

1. Manual Reactor Trip 2 1 2 1, 2 and
  • 1
2. High Flux 4 2 3 1,2 2#,10,11#
3. RC High Temperature 4 2 3 1,2 3#, 10
4. Flux - AFlux - Flow 4 2(a)(b) 3 1,2 2#, 10
5. RC Low Pressure 4 2(a) 3 1,2 3#, 10
6. RC High Pressure 4 2 3 1,2 3#, 10
7. RC Pressure-Temperature 4 2(a) 3 1,2 3#, 10
8. High Flux/Number of Reactor Coolant Pumps On 4 2(a)(b) 3 1,2 3#, 10
9. Containment High Pressure 4 2 3 1,2 3#, 10
10. Intermediate Range, Neutron Flux and Rate 2 N/A 2(c) 1.,2 and* 4
11. Source Range, Neutron Flux and Rate I

A. Startup 2 N/A 2 2 ## and

  • 5 B. Shutdown 2 N/A 1 3, 4 and 5 6
12. Control Rod Drive Trip Breakers 2 per trip 1 per trip 2 per 1, 2 and
  • 7#, 8#

system system trip system z 13. Reactor Trip Module 2 per trip I per trip 2 per 1,2 and

  • 7#

I-system system trip system

14. Shutdown Bypass High Pressure 4 2 3 2"*, 3"* 6#

4**, 5**

co 15. CR Relays 2 2 2 1, 2and* 9

TABLE 3.3-1 (Continued)

ACTION STATEMENTS (Continued)

ACTION 8 With one of the Reactor Trip Breaker diverse trip features (undervoltage or shunt trip devices) inoperable, restore it to.

OPERABLE status in 48 hours2 days <br />0.286 weeks <br />0.0658 months <br /> or place the breaker in trip in the next hour.

ACTION 9 With one or both channels of SCR Relays inoperable, restore the channels to OPERABLE status during the next COLD SHUTDOWN exceeding 24 hours1 days <br />0.143 weeks <br />0.0329 months <br />.

ACTION 10 - With the number of channels OPERABLE one less than the Minimum Channels OPERABLE requirement, within one hour, place one inoperable channel in trip and the second inoperable channel in bypass, and restore one of the inoperable channels to OPERABLE status within 48 hours2 days <br />0.286 weeks <br />0.0658 months <br /> or be in HOT STANDBY within the next 6 hours0.25 days <br />0.0357 weeks <br />0.00822 months <br /> and open the reactor trip-breakers.

ACTION 11 - In MODE I above 50% RATED THERMAL POWER, when the calculated required secondary heat balance is no longer based on ultrasonic flow meter instrumentation,

a. Immediately reduce THERMAL POWER to < 98.4% of RATED THERMAL POWER with four reactor coolant pumps operating or to _ 73.8% of RATED THERMAL POWER with three reactor coolant pumps operating, and
b. Within 10 hours0.417 days <br />0.0595 weeks <br />0.0137 months <br />, reduce the High Flux trip setpoint to < 103.3%

of RATED THERMAL POWER with four reactor coolant pumps operating.

DAVIS-BESSE, UNIT I 3/4 3-5a Amendment No. 108, 135, 185,278 (Next page is 3/4 3-6.)

TABLE 4.3-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS cj~

CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE cj~

FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED z 1. Manual Reactor Trip N.A. N.A. S/U(1) N.A.

H 2. High Flux S D(2), and Q(6,9,10) N.A. 1,2

3. RC High Temperature S R SA(9) 1,2
4. Flux - AFlux - Flow S(4) M(3) and Q(6,7,9) N.A. 1,2
5. RC Low Pressure S R SA(9) 1,2
6. RC High Pressure S R SA(9) 1,2
7. RC Pressure-Temperature S R(10) SA(9, 10) 1,2
8. High Flux/Number of Reactor S Q(6,9) N.A. 1,2 Coolant Pumps On
9. Containment High Pressure S E SA(9) 1,2
10. Intermediate Range, Neutron S E(6) N.A.(5) 1, 2 and
  • C Flux and Rate
11. Source Range, Neutron Flux S E(6) N.A.(5) 2, 3, 4 and 5 z and Rate 0c 12. Control Rod Drive Trip Breakers N.A. N.A. Q(8,9) and 1, 2 and
  • I'. 0 SIU(1)(8)
13. Reactor Trip Module Logic N.A. N.A. Q(9) 1, 2 and
  • 0o0 14. Shutdown Bypass High Pressure S R SA(9) 2**, 3**, 4**, 5**

0 0 t,

15. SCR Relays N.A. N.A. R 1,2 and *

-1

TABLE 4.3-1 (Continued)

Notation (1) - If not performed in previous 7 days.

(2) - Heat balance only, above 15% of RATED THERMAL POWER. When > 50% RATED THERMAL POWER,. ultrasonic flow meter instrumentation is required to be utilized.when performing secondary calorimetric heat balance unless ACTION 11 of Table 3.3-1 is entered.

Adjust power range channel output if calorimetric heat balance calculation results exceed power range channel output by greater than 2% RATED THERMAL POWER.

(3) - When THERMAL POWER [TP] is above 50% of RATED THERMAL POWER [RTP],

and at a steady state, compare out-of-core measured AXIAL POWER IMBALANCE

[APIo] to incore measured AXIAL POWER IMBALANCE [API1] as follows:

RTP [APIo - APII] = Offset Error TP Recalibrate if the absolute value of the Offset Error is > 2.5%

(4) - AXIAL POWER IMBALANCE and loop flow indications only.

(5) - CHANNEL FUNCTIONAL TEST is not applicable. Verify at least one decade overlap prior to each reactor startup if not verified in previous 7 days.

(6) - Neutron detectors may be excluded from CHANNEL CALIBRATION.

(7) - Flow rate measurement sensors may be excluded from CHANNEL CALIBRATION.

However, each flow measurement sensor shall be calibrated at least once each REFUELING INTERVAL.

(8) - The CHANNEL FUNCTIONAL TEST shall independently verify the OPERABILITY of both the undervoltage and shunt trip devices of the Reactor Trip Breakers.

(9) - Performed on a STAGGERED TEST BASIS.

(10) - If the as-found channel setpoint is conservative with respect to the Allowable Value but outside its predefined as-found acceptance criteria band, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service. If the as-found instrument channel setpoint is not conservative with respect to the Allowable Value, the channel shall be declared inoperable.

The instrument channel setpoint shall be reset to a value that is within the as-left tolerance of the Limiting Trip Setpoint, or a value that is more conservative than the Limiting Trip Setpoint; otherwise, the channel shall be declared inoperable. The Limiting Trip Setpoint and the methodology used to determine the Limiting Trip Setpoint, the predefined as-found acceptance criteria band, and the as-left setpoint tolerance band are specified in a document incorporated by reference into the Updated Safety Analysis Report.

    • - When Shutdown Bypass is actuated.

DAVIS-BESSE, UNIT 1 3/4 3-8 Amendment No. 43, 108, 123, 135, 185, 218, 274 ,278

PLANT SYSTEMS CONDENSATE STORAGE TANKS LIMITING CONDITION FOR OPERATION 3.7.1.3 The condensate storage tanks shall be OPERABLE with a minimum usable volume bf 270,300 gallons of water.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

With the condensate storage tanks inoperable, within 4 hours0.167 days <br />0.0238 weeks <br />0.00548 months <br /> either:

a. Restore the condensate storage tanks to OPERABLE-status or be in-HOT SHUTDOWN within the next 12 hours0.5 days <br />0.0714 weeks <br />0.0164 months <br />, or
b. Verify by administrative means the OPERABILITY of the service water system as a backup supply to the auxiliary feedwater system, verify once per 12 hours0.5 days <br />0.0714 weeks <br />0.0164 months <br /> thereafter, and restore the condensate storage tanks to OPERABLE status within 7 days or be in HOT SHUTDOWN within the following 12 hours0.5 days <br />0.0714 weeks <br />0.0164 months <br />.

SURVEILLANCE REQUIREMENTS 4.7.1.3.1 The condensate storage tanks shall be demonstrated OPERABLE at least once per 12 hours0.5 days <br />0.0714 weeks <br />0.0164 months <br /> by verifying the usable water volume to be within its limits when the tanks are the supply source for the auxiliary feedwater pumps.

DAVIS-BESSE, UNIT I ý 3/4.7-6 Amendment No. 164, 200, 278

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT 6.9.1.7 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle and any remaining part of a reload cycle for the following:

2.1.2 AXIAL POWER IMBALANCE Protective Limits for Reactor Core Specification 2.1.2 2.2.1 Trip Setpoint for Flux -- AFlux/Flow for Reactor Protection System Setpoints Specification 2.2.1 3.1.1.3c Negative Moderator Temperature Coefficient Limit 3.1.3.6 Regulating Rod Insertion Limits 3*13.7 Rod-P0r6ora.m 3.1.3.8 Xenon Reactivity 3.1.3.9 Axial Power Shaping Rod Insertion Limits 3*2.1 AXIAL POWER IMBALANCE 3.2.2 Nuclear Heat Flux Hot Channel Factor, FQ 3.2.3 Nuclear Enthalpy Rise Hot Channel Factor, FN AH 3.2.4 QUADRANT POWER TILT The analytical methods used to determine the core operating limits addressed by the individual Technical Specifications shall be: those previously reviewed and approved by the NRC, as described in BAW-10179P-A, "Safety Criteria and Methodology for Acceptable Cycle Reload Analyses", or any other new NRC-approved analytical methods used to determine core operating limits that are not yet referenced in the applicable approved revision of BAW-10179P-A. The applicable approved revision number for BAW- 101 79P-A at the time the reload analyses are performed shall be identified in the CORE OPERATING LIMITS REPORT. The CORE OPERATING LIMITS REPORT shall also list any new NRC-approved analytical methods used to determine core operating limits that are not yet referenced in the applicable approved revision of BAW- 10179P-A.

DAVIS-BESSE, UNIT I 6-14 Amendment No. 144, 154, 189, 267, 278

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Continued)

As described in reference documents listed in accordance with the instructions given above, when an initial assumed power level of 102% of RATED THERMAL POWER is specified in a previously approved method, an actual value of 100.37% of RATED THERMAL POWER may be used when the input for reactor thermal power measurement of feedwater mass flow and temperature is from the Ultrasonic Flow Meter. The following NRC approved documents are applicable to the use of the Ultrasonic Flow Meter with a 0.37% measurement uncertainty:

Caldon Inc. Engineering Report-80P, "Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level Using the LEFMITM System,"

Revision 0, dated March, 1997.

Caldon Inc. Engineering Report-i 57P, "Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFMV'TM or LEFM CheckPlusTM System," Revision 5, dated October, 2001.

The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

The CORE OPERATING LIMITS REPORT, including any mid-cycle revision or supplements thereto, shall be provided upon issuance for each reload cycle to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

DAVIS-BESSE, UNIT I 6-14a Amendment No. 144, 154, 189, 276,278

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