ML24250A195

VTS-5.6.5 Updated 8-29-24

ML24250A195
Person / Time
Site:	Vogtle
Issue date:	08/29/2024
From:	Southwest Nuclear Co
To:	Office of Nuclear Reactor Regulation
References
Download: ML24250A195 (1)

Text

Reporting Requirements 5.6

5.6 Reporting Requirements

5.6.5 Core Operating Limits Report (COLR)

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

SL 2.1.1 Reactor Core Safety Limits LCO 3.1.1 "SHUTDOWN MARGIN" LCO 3.1.3 "Moderator Temperature Coefficient" LCO 3.1.5 "Shutdown Bank Insertion Limits" LCO 3.1.6 "Control Bank Insertion Limits" LCO 3.2.1 "Heat Flux Hot Channel Factor" LCO 3.2.2 "Nuclear Enthalpy Rise Hot Channel Factor" LCO 3.2.3 "Axial Flux Difference" LCO 3.3.1 Reactor Trip System (RTS) Instrumentation LCO 3.4.1 Reactor Coolant System ( RCS) Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits LCO 3.9.1 "Boron Concentration"

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

WCAP-9272-P-A, "WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY," July 1985 (W Proprietary). (Methodology for Moderator Temperature Coefficient, Shutdown Bank Insertion Limit, Control Bank Insertion Limits, and Nuclear Enthalpy Rise Hot Channel Factor, Reactor Trip System Instrumentation, and Reactor Coolant System Pressure, Temperature, and Flow Departure from Nucleate Boiling Limits.)

WCAP-10216-P-A, Revision 1A, " RELAXATION OF CONSTANT AXIAL OFFSET CONTROL FQ SURVEILLANCE TECHNICAL SPECIFICATION," February, 1994 (W Proprietary). (Methodology for Axial Flux Difference (Relaxed Axial Offset Control) and Heat Flux Hot Channel Factor.)

WCAP-10266-P-A, Revision 2, "The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code," March 1987.

(W Proprietary) (Methodology for Axial Flux Difference (Relaxed Axial Offset Control) and Heat Flux Hot Channel Factor.)

WCAP-13749-P-A, Safety Evaluation Supporting the Conditional Exemption of the Most Negative EOL Moderator Temperature Coefficient Measurement, March 1997.

(continued)

Vogtle Units 1 and 2 5.6-3 Amendment No. 223 (Unit 1)

Amendment No. 206 (Unit 2)

Reporting Requirements 5.6

5.6 Reporting Requirements

5.6.5 Core Operating Limits Report (COLR) (continued)

WCAP-16045-P-A, Qualification of the Two-Dimensional Transport Code PARAGON, August 2004 (Methodology for Moderator Temperature Coefficient.)

WCAP-16045-P-A, Addendum 1-A, Qualification of the NEXUS Nuclear Data Methodology, August 2007 (Methodology for Moderator Temperature Coefficient.)

WCAP-12610-P-A, VANTAGE+ Fuel A ssembly Reference Core Report, April 1995 (Westinghouse Proprietary). (Methodology for Axial Flux Difference (Relaxed Axial Offset Control) and Heat Flux Hot Channel Factor.)

WCAP-12610-P-A & CENP D-404-P-A, Addendum 1-A, Optimized ZIRLO TM, July 2006 (Westinghouse Pro prietary). (Meth odology for Axial Flux Difference (Relaxed Axial Offset Control) and Heat Flux Hot Channel Factor.)

WCAP-17661-P-A Revision 1, Improved RAOC and CAOC F Q Surveillance Technical Specifications, February 2019 (W Proprietary).

(Methodology for Control Bank Insertion Limits, Heat Flux Hot Channel Factor (W(Z) surveillance requirements for F Q Methodology), and Axial Flux Difference (Relaxed Axial Offset Control).)

WCAP-8745-P-A, Design Bases for the Thermal Overpower T and Thermal Overtemperature T Trip Functions, September 1986 (W Proprietary). (Methodology for Reactor Trip System Instrumentation.)

WCAP-11397-P-A, "Revised Thermal Design Procedure," April 1989 (W Proprietary). (Methodology for Reactor Core Safety Limits and RCS Pressure, Temperature and Flow Departure from Nucleate Boiling Limits.)

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

(continued)

Vogtle Units 1 and 2 5.6-4 Amendment No. 223 (Unit 1)

Amendment No. 206 (Unit 2)

Reporting Requirements 5.6

5.6 Reporting Requirements

5.6.6 Reactor Coolant System (RCS) PR ESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a. RCS pressure and temperature limits for heatup, cooldown, operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

LCO 3.4.3 "RCS Pressure and Temperature (P/T) Limits"

b. The power operated relief valve lift settings required to support the Cold Overpressure Protection Systems (COPS) and the COPS arming temperature shall be established and documented in the PTLR for the following:

LCO 3.4.12 "Cold Overpressure Protection Systems"

c. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1. WCAP-14040-A, Rev. 4, Me thodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves.

2. WCAP-16142-P, Rev. 1, Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Vogtle Units 1 and 2.

3. The PTLR will contain the complete identification for each of the TS reference Topical Reports used to prepare the PTLR (i.e.,

report number, title, revision, date, and any supplements).

d. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

5.6.7 EDG Failure Report

If an individual emergency diesel generator (EDG) experiences four or more valid failures in the last 25 demands, these failures and any nonvalid failures experienced by that EDG in that time period shall be reported within 30 days.

Reports on EDG failures shall include the information recommended in Regulatory Guide 1.9, Revision 3, Regulatory Position C.4, or existing Regulatory Guide 1.108 reporting requirement.

(continued)

Vogtle Units 1 and 2 5.6-5 Amendment No. 223 (Unit 1)

Amendment No. 206 (Unit 2)

Reporting Requirements 5.6

5.6 Reporting Requirements

5.6.8 PAM Report

When a Report is required by Condition G or J of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.9 Deleted.

5.6.10 Steam Generator Tube Inspection Report

A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.9, Steam Generato r (SG) Program. The report shall include:

a. The scope of inspections performed on each SG:

b. The nondestructive examination techniques utilized for tubes with increased degradation susceptibility;

c. For each degradation mechanism found:

1. The nondestructive examination techniques utilized;

2. The location, orientation (if linear), measured size (if available), and voltage response for each indication. For tube wear at support structures less than 20 percent through-wall, only the total number of indications needs to be reported;

3. A description of the condition monitoring assessment and results, including the margin to the tube integrity performance criteria and comparison with the margin predicted to exist at the inspection by the previous forward-looking tube integrity assessment; and

4. The number of tubes plugged during the inspection outage.

d. An analysis summary of the tube integrity conditions predicted to exist at the next scheduled inspection (the forward-looking tube integrity assessment) relative to the applicable performance criteria, including the analysis methodology, inputs, and results;

(continued)

Vogtle Units 1 and 2 5.6-6 Amendment No. 223 (Unit 1)

Amendment No. 206 (Unit 2)

Reporting Requirements 5.6

5.6 Reporting Requirements

5.6.10 Steam Generator Tube Inspection Report (continued)

e. The number and percentage of tubes plugged to date, and the effective plugging percentage in each SG;

f. The results of any SG secondary side inspections;

g. The primary to secondary LEAKAGE rate observed in each SG (if it is not practical to assign the LEAKAGE to an individual SG, the entire primary to secondary LEAKAGE should be conservatively assumed to be from one SG) during the cycle preceding the inspection which is the subject of the report;

h. The calculated accident induced leakage rate from the portion of the tubes below 15.2 inches from the top of the tubesheet for the most limiting SG. In addition, if the calculated accident induced leakage rate from the most limiting accident is less than 2.48 times the maximum operational primary to secondary leakage rate, the report should describe how it was determined; and

i. The results of monitoring for tube axial displacement (slippage). If slippage is discovered, the implications of the discover and corrective action shall be provided.

Vogtle Units 1 and 2 5.6-7 Amendment No. 223 (Unit 1)

Amendment No. 206 (Unit 2)