ML021060236

Technical Specifications, Amendment Nos. 138 and 127 Approving Uprated Core Thermal Power and Revising the Associated Technical Specifications (Tac Nos. MB2899 and MB2903)

ML021060236
Person / Time
Site:	South Texas
Issue date:	04/12/2002
From:	NRC/NRR/DLPM
To:
References
TAC MB2899, TAC MB2903
Download: ML021060236 (6)

Text

DEFINITIONS PROCESS CONTROL PROGRAM 1.24 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, tests, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71, State regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.

PURGE - PURGING 1.25 PURGE or PURGING shall be any controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

QUADRANT POWER TILT RATIO 1.26 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average.

RATED THERMAL POWER 1.27 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3,853 Mwt (Model A94 steam generators installed) or 3,800 Mwt (Model E steam generators installed).

REACTOR TRIP SYSTEM RESPONSE TIME 1.28 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its Trip Setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components or methodology for verification have been previously reviewed and approved by the NRC.

REPORTABLE EVENT 1.29 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 of 10 CFR Part 50.

SHUTDOWN MARGIN 1.30 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming all full-length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.

SOUTH TEXAS - UNITS 1 & 2 1-5 Unit 1 - Amendment No. 138 Unit 2 - Amendment No. 127

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (Tayg) shall not exceed the limits shown in the Core Operating Limits Report.

2.1.1.1 In MODES 1 and 2, the departure from nucleate boiling ratio (DNBR) shall be maintained > 1.17 for the WRB-1 DNB correlation and > 1.14 for the WRB-2M DNB correlation.

2.1.1.2 In MODES 1 and 2, the peak fuel centerline temperature shall be maintained

< 50800 F, decreasing by 58 OF per 10,000 MWD/MTU of bumup.

APPLICABILITY: MODES 1 and 2.

ACTION:

Whenever the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and comply with the requirements of Specification 6.7.1.

REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.

APPLICABILITY: MODES 1, 2,3,4, AND 5.

ACTION:

MODES 1 and 2:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and comply with the requirements of Specification 6.7.1.

MODES 3, 4 and 5:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes, and comply with the requirements of Specification 6.7.1.

2-1 Unit 1 - Amendment No. 138 SOUTH TEXAS - UNITS 1 & 2 Unit 2- Amendment No. 127

TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING 4 LOOP OPERATION MODEL A94 STEAM GENERATORS ONLY MAXIMUM NUMBER OF INOPERABLE MAXIMUM ALLOWABLE POWER RANGE SAFETY VALVES ON ANY NEUTRON FLUX HIGH SETPOINT OPERATING STEAM GENERATOR (PERCENT OF RATED THERMAL POWER) 1 61 2 43 3 26 MODEL E STEAM GENERATORS ONLY MAXIMUM NUMBER OF INOPERABLE MAXIMUM ALLOWABLE POWER RANGE SAFETY VALVES ON ANY NEUTRON FLUX HIGH SETPOINT OPERATING STEAM GENERATOR (PERCENT OF RATED THERMAL POWER) 1 63 2 45 3 27 Unit 1- Amendment N1.38 SOUTH TEXAS - UNITS 1 & 2 3/4 7-2 Unit 2 - Amendment No127

VO CHANGES ON THIS PAGE Attachment 4 NOC-AE-01001162 ADMINISTRATIVE CONTROLS MONTHLY OPERATING REPORTS 6.9.1.5 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the PORVs or safety valves, shall be submitted on a monthly basis to the Director, Office of Resource Management, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to the Regional Administrator of the Regional Office of the NRC, no later than the 15th of each month following the calendar month covered by the report.

CORE OPERATING LIMITS REPORT 6.9.1.6.a Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle, or any part of a reload cycle for the following:

1. Safety limits for thermal power, pressurizer pressure, and the highest operating loop coolani temperature (Tavg) for Specification 2.1,

2. Limiting Safety System Settings for Reactor Coolant Flow-Low Loop design flow, Overtemperature AT, and Overpower AT setpoint parameter values for Specification 2.2,

3. Moderator Temperature Coefficient BOL and EOL limits, and 300 ppm surveillance limit for I Specification 3/4.1.1.3,

4. Shutdown Bank Insertion Limit for Specification 3/4.1.3.5,

5. Control Bank Insertion Limits for Specification 3/4.1.3.6,

6. Axial Flux Difference limits and target band for Specification 3/4.2.1,

7. Heat Flux Hot Channel Factor, K(Z), Power Factor Multiplier, and (F,* RTP) for Specification I 3/4.2.2,

8. Nuclear Enthalpy Rise Hot Channel Factor, and Power Factor Multiplier for Specification 3/4.2.3, and

9. DNB related parameters for Reactor Coolant System Tavg Pressurizer Pressure, and the Minimum Measured Reactor Coolant System Flow for Specification 3/4.2.5.

The CORE OPERATING LIMITS REPORT shall be maintained available in the Control Room.

6.9.1.6.b The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in:

1. WCAP 9272-P-A, "WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY,"

July, 1985 FN Proprietary).

(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5 Shutdown Rod Insertion Limit, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 Heat Flux Hot Channel Factor, 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor, and 3.2.5 - I DNB Parameters.)

SOUTH TEXAS - UNITS 1 & 2 6-21 Unit 1 - Amendment No. 9,27,.35A* , 15.

Unit 2 - Amendment No. 4,17,,26,36 103

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Continued)

2. WCAP12942-P-A "SAFETY EVALUATION SUPPORTING A MORE NEGATIVE EOL MODERATOR TEMPERATURE COEFFICIENT TECHNICAL SPECIFICATION FOR THE SOUTH TEXAS PROJECT ELECTRIC GENERATING STATION UNITS 1 AND 2."

(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient)

3. WCAP-8745-P-A, "Design Bases for the Thermal Overpower AT and Thermal Overtemperature, AT Trip Functions," September 1986 (Westinghouse ProprietaryClass 2)

(Methodology for Specification 2.1 - Safety Limits, and 2.2 - Limiting Safety System Settings)

4. WCAP 8385, "POWER DISTRIBUTION AND LOAD FOLLOWING PROCEDURES TOPICAL REPORT', September, 1974 (W Proprietary).

(Methodology for Specification 3.2.1 - Axial Flux Difference (Constant Axial Offset Control.)

5. Westinghouse letter NS-TMA-2198, T.M. Anderson (Westinghouse) to K. Kniel (Chief of Core Performance Branch, NRC) January 31, 1980 -

Attachment:

Operation and Safety Analysis Aspects of an Improved Load Follow Package.

(Methodology for Specification 3.2.1 - Axial Flux Difference (Constant Axial Offset Control).

Approved by NRC Supplement No. 4 to NUREG-0422, January, 1981 Docket Nos. 50-369 and 50-370.)

6. NUREG-0800, Standard Review Plan, U. S. Nuclear Regulatory Commission, Section 4.3, Nuclear Design, July, 1981. Branch Technical Position CPB 4.3-1, Westinghouse Constant Axial Offset Control (CAOC), Rev. 2, July 1981.

(Methodology for Specification 3.2.1 - Axial Flux Difference (Constant Axial Offset Control).)

7. WCAP-1 0266-P-A, Rev.2, WCAP-1 1524-NP-A, Rev.2, "The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code", Kabadi, J.N., et al., March 1987; including Addendum 1-A, "Power Shape Sensitivity Studies," December 1987 and Addendum 2-A, "BASH Methodology Improvements and Reliability Enhancements" May 1988.

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)

8. WCAP-12610-P-A, "VANTAGE + Fuel Assembly Reference Core Report," April, 1995 (W Proprietary) for Loss of Coolant Accident (LOCA) Evaluation models with ZIRLO clad fuel for rod heatup calculation.

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)

SOUTH TEXAS - UNITS 1 & 2 6-22 Unit 1 - Amendment No. 115 Unit 2 - Amendment No. 103

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Continued)

9. CENPD-397-P-A, Revision 01, "Improved Flow Measurement Accuracy Using Crossflow Ultrasonic Flow Measurement Technology," May 2000.

(Methodology for operating at a RATED THERMAL POWER of 3,853 Mwt) 6.9.1.6.c 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.

138 SOUTH TEXAS - UNITS 1 & 2 6-22a Unit 1 - Amendment No.

Unit 2 - Amendment No. 127