ML070610246

Tech Spec Pages for Amendment 183 Relocation of Cycle-Specific Parameter Limits to the Core Operating Limits Report

ML070610246
Person / Time
Site:	Callaway
Issue date:	04/02/2007
From:	NRC/NRR/ADRO/DORL/LPLIV
To:
Donohew J N, NRR/DORL/LP4, 415-1307
Shared Package
ML070600786	List: ML070600768 ML070610246
References
TAC MD2873
Download: ML070610246 (10)

Text

-3-(4) U E, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source of special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5) UE, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.C. This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below: (1) Maximum Power Level UE is authorized to operate the facility at reactor core power levels not in excess of 3565 megawatts thermal (100% power) in accordance with the conditions specified herein.(2) Technical Specifications and Environmental Protection Plan*The Technical Specifications contained in Appendix A, as revised through Amendment No. 183 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license.The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.(3) Environmental Qualification (Section 3.11. SSER #3)**Deleted per Amendment No. 169 Amendments 133, 134, &135 were effective as of April 30, 2000 however these amendments were implemented on April 1, 2000.The parenthetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report and/or its supplements wherein the license condition is discussed.

Amendment 183 SLs 2.0 2.0 SAFETY LIMITS (SLs)2.1 SLs 2.1.1 Reactor Core SLs In MODES 1 and 2, the combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the limits specified in the COLR; and the following SLs shall not be exceeded: 2.1.1.1 The design limit departure from nucleate boiling ratio (DNBR) shall be maintained

> 1.22 for transients analyzed using the revised thermal design procedure (RTDP) methodology and the WRB-2 DNB correlation.

For non-RTDP transients analyzed using the standard thermal design procedure, the DNBR shall be maintained greater than or equal to the applicable DNB correlation limit (> 1.17 for WRB-2, > 1.30 for W-3).2.1.1.2 The peak fuel centerline temperature shall be maintained

< 5080 0 F, decreasing by 58 0 F per 10,000 MWd/MTU of burnup.2.1.2 RCS Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be;maintained

<2735 psig.2.2 SL Violations 2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.2.2.2 If SL 2.1.2 is violated: 2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes.CALLAWAY PLANT 2.0-1 Amendment 183 RTS Instrumentation 3.3.1 TABLE 3.3.1-1 (page 7 of 8)Reactor Trip System Instrumentation Note 1: Overtemoerature AT The Overtemperature AT Function Allowable Value shall notexceed the following setpoint by more than 1.23% of AT span (1.85% RTP).( 1 + 2s) s { '[ S< + ) +K 3 (P -P') fs(AI)Where: AT is measured RCS AT, OF.AT, is the indicated AT at RTP, IF.s is the Laplace transform operator, sec" 1.T is the measured RCS average temperature, IF.T' is the nominal Tavg at RTP, < *IF.P is the measured pressurizer pressure, psig.P' is the nominal RCS operating pressure =

• psig.K 1 =*x, >

• sec t4 >

• sec K 2 = */IoF T 2 -* sec'C 5<

• sec K 3 */psig 3=

• sec t 6= sec fl(AI) * (*% + (qt -qb)}0% of RTP* ((qt"- qb)" }when qt -qb < %RTP when * %RTP qt -qb -* %RTP when qt -qb > %RTP where qt and qb are percent RTP in the upper and lower halves of the core, respectively, and qt + qb is the total THERMAL POWER in percent RTR The values denoted with

• are specified in the COLR.I CALLAWAY PLANT 3.3-23 Amendment 183 RTS Instrumentation 3.3.1 TABLE 3.3.1-1 (page 8 of 8)Reactor Trip System Instrumentation Note 2: Overpower AT The Overpower AT Function Allowable Value shall not exceed the following setpoint by more than 1.21% of AT span (1.82% RTP).AT (V1s(1 _V7)< ATo{ K 4-K ( s)(1 'T-K 6 [T T ]f 2 (AI)}+ T (l rSk---s) _ + (l3 ~+ rs L-T+"'6S Where: AT is measured RCS AT, OF.ATo is the indicated AT at RTP, IF.s is the Laplace transform operator, sec" 1.T is the measured RCS average temperature, 0 F.T" is the nominal Tavg at RTP, _ *IF.I K 4=K6 = */IF for increasing Tavg*/IF for decreasing Tavg T2!* sec T7 -* sec K 6 = */OF when T > T"*/OF when T _ T"'r3 =

• sec 1* sec r6 =* sec f 2 (AI) = *I The values denoted with

• are specified in the COLR.I CALLAWAY PLANT 3.3-24 Amendment 183 RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS)3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits LCO 3.4.1 RCS DNB parameters for pressurizer pressure, RCS average temperature, and RCS total flow rate shall be within the limits specified below: a. Pressurizer pressure is greater than or equal to the limit specified in the COLR', b. RCS average temperature is less than or equal to the limit specified in the COLR; and c. RCS total flow rate >_ 382,630 gpm.APPLICABILITY:

MODE 1.--------------------------------------------

NOTE -------------

Pressurizer pressure limit does not apply during: a. THERMAL POWER ramp > 5% RTP per minute; or b. THERMAL POWER step > 10% RTP.--------------------

ACTIONS COMPLETION CONDITION REQUIRED ACTION TIME TIME A. One or more RCS DNB A.1 Restore RCS DNB 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> parameters not within limits, parameter(s) to within limit.B. Required Action and b B.1 Be in MODE 2. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met.CALLAWAY PLANT 3.4-1 Amendment No. 183 RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1.1 Verify pressurizer pressure is greater than or equal to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> the limit specified in the COLR.SR 3.4.1.2 Verify RCS average temperature is less than or equal 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to the limit specified in the COLR.SR 3.4.1.3 Verify RCS total flow rate is > 382,630 gpm. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SR 3.4.1.4 -----------------------

NOTE ---------------

Calculated rather than verified by precision heat balance when performed prior to THERMAL POWER exceeding 75% RTR Verify by precision heat balance that RCS total flow Once after each rate is __ 382,630 gpm. refueling prior to THERMAL POWER exceeding 75%RTP AND 18 months CALLAWAY PLANT 3.4-2 Amendment No. 183 Reporting Requirements 5.6 5.0 ADMINISTRATIVE CONTROLS 5.6 Reporting Requirements The following reports shall be submitted in accordance with 10 CFR 50.4.5.6.1 Not Used.5.6.2 Annual Radiological Environmental Operating Report The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted by May 1. of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the radiological environmental monitoring program for the reporting period.The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C.The Annual Radiological Environmental Operating Report shall include the results of analyses of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements in a format similar to the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible.5.6.3 Radioactive Effluent Release Report The Radioactive Effluent Release Report covering the operation of the unit during the previous year shall be submitted prior to May 1 of each year. in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR Part 50, Appendix I, Section IV.B.1.5.6.4 Not used.(continued)

CALLAWAY PLANT 5.0-20 Amendment 183 J 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:

1. Moderator Temperature Coefficient limits in Specification 3.1.3, 2. Shutdown Bank Insertion Limit for Specification 3.1.51 3. Control Bank Insertion Limits for Specification 3.1.6, 4. Axial Flux Difference Limits for Specification 3.2.3, 5. Heat Flux Hot Channel Factor, FQ(Z), FQRTP, K(Z), W(Z) and FQ Penalty Factors for Specification 3.2.1, 6. Nuclear Enthalpy Rise Hot Channel Factor FAH, FAH RTP, and Power Factor Multiplier, PFAH, limits for Specification 3.2.2, 7. Shutdown Margin Limits for Specifications 3.1.1, 3.1.4, 3.1.5, 3.1.6, and 3.1.8, 8. Reactor Core Safety Limits Figure for Specification 2.1.1, 9. Overtemperature AT and Overpower AT Setpoint Parameters for Specification 3.3.1, and 10. Reactor Coolant System Pressure and Temperature DNB Limits for Specification 3.4.1.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:

1. WCAP-9272-P-A, "WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY." 2. WCAP-10216-P-A, "RELAXATION OF CONSTANT AXIAL OFFSET CONTROL AND FQ SURVEILLANCE TECHNICAL SPECIFICATION." 3. WCAP-10266-P-A, "THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE." (continued)

CALLAWAY PLANT 5.0-21 Amendment 183 Reporting Requirements 5.6 5.6 Reporting Requirements

4. WCAP-12610-P-A, "VANTAGE + FUEL ASSEMBLY REFERENCE CORE REPORT." 5. WCAP-11397-P-A, "REVISED THERMAL DESIGN PROCEDURE." 6. WCAP-14565-P-A, "VIPRE-01 MODELING AND QUALIFICATION FOR PRESSURIZED WATER REACTOR NON-LOCA THERMAL-HYDRAULIC SAFETY ANALYSIS." 7. WCAP-10851-P-A, "IMPROVED FUEL PERFORMANCE MODELS FOR WESTINGHOUSE FUEL ROD DESIGN AND SAFETY EVALUATIONS." 8. WCAP-15063-P-A, "WESTINGHOUSE IMPROVED PERFORMANCE ANALYSIS AND DESIGN MODEL (PAD 4.0)." 9. WCAP-8745-P-A, "DESIGN BASES FOR THE THERMAL OVERPOWER AT AND THERMAL OVERTEMPERATURE AT TRIP FUNCTIONS." 10. WCAP-10965-P-A, "ANC: A WESTINGHOUSE ADVANCED NODAL COMPUTER CODE." 11. WCAP-11596-P-A, "QUALIFICATION OF THE PHOENIX-P/ANC NUCLEAR DESIGN SYSTEM FOR PRESSURIZED WATER REACTOR CORES." 12. WCAP-13524-P-A, "APOLLO: A ONE DIMENSIONAL NEUTRON DIFFUSION THEORY PROGRAM." 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)

CALLAWAY PLANT 5.0-22 Amendment 183 Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)a. RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, hydrostatic testing and PORV lift setting as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

1. Specification 3.4.3, "RCS Pressure and Temperature (P/T) Limits," and 2. Specification 3.4.12, "Cold Overpressure Mitigation System (COMS)." b. The analytical methods used to determine the RCS pressure and temperature and COMS PORV limits shall be those previously reviewed and approved by the NRC, specifically those described in WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves".c. 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 Not used.5.6.8 PAM Report When a report is required by Condition B or G 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 Not used.(continued)

CALLAWAY PLANT 5.0-23 Amendment 183 1