ML20211J946

From kanterella
Jump to navigation Jump to search
Amends 67 & 67 to Licenses NPF-87 & NPF-89,respectively, Removing Several cycle-specific Parameter Limits from TS & Adding Parameter Limits to COLR
ML20211J946
Person / Time
Site: Comanche Peak  Luminant icon.png
Issue date: 08/30/1999
From: Gramm R
NRC (Affiliation Not Assigned)
To:
Shared Package
ML20211J951 List:
References
NUDOCS 9909070070
Download: ML20211J946 (12)
v • d • e


Text

{{#Wiki_filter:3 W "4 ,g g ,o UNITED STATES

5

) NUCLEAR REGULATORY COMMISSION g WASHINGTON, D.C. 20666-0001 %,*****y \\ TEXAS UTILITIES ELECTRIC COMPANY COMANCHE PEAK STEAM ELECTRIC STATION. UNIT 1 DOCKET NO. 50-445 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 67 License No. NPF-87 1. The Nuclear Regulatory Commission (the Commission) has found that: I A. The application for amendment by Texas Utilities Electric Company (TU Electric, the licensee) dated May 24,1999, as supplemented by letter dated July 9,1999, complies with the standards and requirements of the Atomic Energy Act of i 1954, as amended (the Act), and the Commission's rules and regulations set l forth in 10 CFR Chapter I; l B. The facility will operate in conformity with the application, as amended, the I provisions of the Act, and the rules and regulations of the Commission; l C. There is reasonable assurance (i) that the activities authorized by this l amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; art." E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied. 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license ameridment and Paragraph 2.C.(2) of Facility Operating License No. NPF-87 is hereby amended to read as follows: 9909070070 990830 PDR ADOCK 05000443 P PDR

v - 1 2. Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 67 , and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan. 3. The license amendment is effective as of its date of issuance and shall be implemented within 30 days from the date of issuance. . FOR THE NUCLEAR REGULATORY COMMISSION Mk h Robert A. Gramm, Chief, Section 1 Project Directorate IV & Decommissioning Division of Licensing Project Management Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: August 30, 1999

8 O nt: p UNITED STATES g j NUCLEAR REGULATORY COMMISSION e f WASHINGTON, D.C. 20666 4001. I TEXAS UTILITIES ELECTRIC COMPANY ^ COMANCHE PEAK STEAM ELECTRIC STATION. UNIT 2 j DOCKET NO. 50-446 i AMENDMENT TO FACILITY OPERATING LICENSE j Amendment No. 67 . License No. NPF-89 1. The Nt. clear Regulatory Commission (the Commission) has found that: A. The application for amendment by Texas Utilities Electric Company (TU Electric, the licensee) dated May 24,1999. as supplemented by letter dated July 9,1999, complies with the standards and sequiiements of the Atomic Energy Act of i 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 cf the Commission's regulations and all applicable recuirements have been satisfied. 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and Paragraph 2.C.(2) of Facility Operating Licenso No. NPF-89 is hereby amended to read as follows:

l 2-(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 67, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this license. TU Electric shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan. 3. This license amendment is effective as of its date of issuance and shall be implemented within 30 days from the date of issuance. FOR THE NUCLEAR REGULATORY COMMISSION k h Robert A. Gramm, Chief, Section 1 Project Directorate IV & Decommissioning Division of Licensing Project Management Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: August 30, 19 %

ATTACHMENT TO LICENSE AMENDMENT NO. 67 TO FACILITY OPERATING LICENSE NO. NPF-87 AND AMENDMENT NO. 67 TO FACILITY OPERATING LICENSE NO. NPF-89 DOCKET NOS. 50-445 AND 50-446 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change. Remove insert 2.0-1 2.0-1 2.0-2 2.0-2 2.0-3 2.0-3 3.3-21 3.3-21 3.4-1 3.4-1 3.4-2 3.4-2 3.4-3 3.4-3 5.0-32 5.0-32 5.0 33 5.0-33 5.0-34 5.0-34

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 In MODES 1 and 2, the departure from nucleate boiling ratio (DNBR) shall be maintained 2 the 95/95 DNB criterion for the DNB correlation (s) specified in Section 5.6.5. 2.1.1.2 in MODES 1 and 2, the peak fuel centerline temperature shall be maintained < 4700*F. 2.1.2 RCS Pressure SL in MODES 1,2,3,4, and 5, the RCS pressure shall be maintained s 2735 psig. 2.2 SL Violations i 2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour. 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 hour. 2.2.2.2 in MODE 3,4, or 5, restore compliance within 5 minutes. l COMANCHE PEAX - UNITS 1 AND 2 2.0-1 A = ar= int No. H, 67 1

SLs 2.0 Figure 2.1.1-1 (page 1 of 2) Reactor Core Safety Limits (Unit 1) [THIS FIGURE AND PAGE HAVE BEEN DELETED.) l i i 1 COMANCHE PEA,K - UNITS 1 AND 2 2.0-2 Amendment tb. H, 67

SLs 2.0 Figure 2.1.1-1 (page 2 of 2) Reactor Core Safety Limits (Unit 2) my ,ps. [THIS FIGURE AND PAGE HAVE BEEN DELETED.] l t l l COMANCHE PEAK-UNITS 1 AND 2 2.0-3 k endment No. #, 67

RTS Instrumentation 3.3.1 Note 1: Overtemoerature N-16 The Overtemperature N-16 Function Allowable Value shall not exceed the following setpoint by more than 1.72% of span for Unit 1, or 2.82% of span for Unit 2. Qutpoing = K,- K (1 + T,s) T -T + K,(P-P')- f,(A q) 2 c c (1 + T s) 2 Where: Q,= Overtemperature N-16 trip setpoint, j K l = i K = YF l 2

  • /psig l

j K = 3 Tc Cold leg temperature = \\ T,* Reference Tc at RATED THERMAL POWER, **F l ) = P = Measured pressurizer pressure, psig P' a

  • psig (Nominal RCS operating pressure)

{ the Laplace transform operator, sec '. s = Time constants utilized in lead-lag controller for T, T,,T2 = T,2

  • sec, and T s
  • sec l

2 f (Aq) = i

  • -((q, - q ) + *%} when (q,- q.) s *% RTP 0%

when *% RTP < (q,- q,) < *% RTP

  • -{(q, - q,) *%}

when (q,- q,) >*% RTP Note 2: Not Used.

  • as specified in the COLR l

COMANCHE PEAK - UNITS 1 AND 2 3.3-21 k = a =it N3. N, 67

RCS Przssura, Temperatura, cnd 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 a the limit specified in the COLR; l
b. RCS average temperature s the limit specified in the COLR; and I
c. RCS total flow rate 2 389,700 gpm and a the limit specified in the COLR.

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 CONDITION REQUIRED ACTION COMPLETION TIME A. One or more RCS DNB A.1 Restore RCS DNB 2 hours parameters not within parameter (s) to within limits. limit. ~ (continued) COMANCHE PEAK - UNITS 1 AND 2 3.4-1 Awa=.t No. 64, 67 ~

d e RCS Pressura, Temp $ratura, and Flow DNB Limits 3.4.1 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B. NOTE B.1 Maintain THERMAL Immediately Only applicable prior to POWER less than 85% exceeding 85% RTP after a RTP. refueling outage. Measured RCS Flow not within limits. C. Required Action and C.1 Be in MODE 2. 6 hours associated Completion Time not met. l SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY G R 3.4.1.1 Verify pressurizer pressure is 2 the limit specified in the 12 hours COLR. SR 3.4.1.2 Verify RCS average temperature is s the limit specified 12 hours in the COLR. (continued) COMANCHE PEAK - UNITS 1 AND 2 3.4-2 kendment No. 64, 67

RCS Pr:ssura, Temperatura, and Flow DNB Limits 3.4.1 ACTIONS (continued) SURVEILLANCE FREQUENCY ^ i SR 3.4.1.3 Verify RCS total flow rate is a 389,700 and a the limit 12 hours specified in the COLR. 1 i SR 3.4.1.4 NOTE-- i Not required to be performed until after exceeding 85% RTP after each refueling outage. 1 Verify by precision heat balance that RCS total flow rate 18 months l is 2 389,700 and 2 the limit specified in the COLR. l l i COMANCHE PEAK - UN'TS 1 AND 2 3.4-3 Amendment No. M, 67

Reporting Requiremtnts 1 5.6 5.6 Reporting Requirements (continued) 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 for Specification 3.1.3, 2) Shutdown Rod Insertion Limit for Specification 3,1.5, 3) Control Rod Insertion Limits for Specification 3.1.6, '4) AXIAL FLUX DIFFERENCE Limits and target band for Specification 3.2.3, 5) Heat Flux Hot Channel Factor, K(Z), W(Z), Fo"TP, and the Fo (Z) C allowances for Specification 3.2.1, 6) Nuclear Enthalpy Rise Hot Channel Factor Limit and the Power Factor Multiplier for Specification 3.2.2.

7).

SHUTDOWN MARGIN values in Specifications 3.1.1, 3.1.4, 3.1.5, 3.1.6 and 3.1.8. 8) Refueling Boron Concentration limits in Specification 3.9.1. 9) Overtemperature N-16 Trip Setpoint in Specification 3.3.1. l l

10) Reactor Coolant System pressure, temperature, and flow in l

Specification 3.4.1. l l

11) Reactor Core Safety Limit figures (Safety Limit 2.1.1) l 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," July 1985 (W Proprietary). 2) WCAP-8385," POWER DISTRIBUTION CONTROL AND LOAD FOLLOWING PROCEDURES - TOPICAL REPORT," September 1974 (W Proprietary). (continued) J C'OMANCHE PEAK - UNITS 1 AND 2 5.0-32 h e d ent !*>. W, 67

Reporting Raquircments 5.6 5.6. Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (continued) 3) T. M. Anderson To K. Kniel(Chief of Core Performance Branch, NRC) January 31,1980--

Attachment:

Operation and Safety Analysis-Aspects of an Improved Load Follow Package. 4) 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. 5) WCAP-10216-P-A, Revision 1 A, " RELAXATION OF CONSTANT AXIAL OFFSET CONTROL Fa SURVEILLANCE TECHNICAL SPECIFICATION," February 1994 (W Proprietary). 6) WCAP-10079-P-A, "NOTRUMP, A NODAL TRANSIENT SMALL BREAK AND GENERAL NETWORK CODE," August 1985, (W Proprietary). 7) WCAP-10054-P-A, " WESTINGHOUSE SMALL BREAK ECCS EVALUATION MODEL USING THE NOTRUMP CODE", August 1965, (W Proprietary). 8) WCAP-11145-P-A, " WESTINGHOUSE SMALL BREAK LOCA ECCS EVALUATION MODEL GENERIC STUD ( WITH THE NOTRUMP CODE", October 1986, (W Proprietary). 9) RXE-90-006-P-A," Power Distribution Control Analysis and l Overtemperature N-16 and Overpower N-16 Trip Setpoint i Methodology, " June 1994. l

10) RXE-88-102-P-A,"TUE-1 Departure from Nucleate Boiling l

Correlation", July 1992. l

11) RXE-88-102-P, Sup.1, "TUE-1 DNB Correlation - Supplement 1",

December 1990. (continued) COMANCHE PEAK - UNITS 1 AND 2 5.0-33 Anmhnt No. 64, 67 i g gi. _t, .=4 am._< wq met..-N_ aps.s4t' q.

ygpp, er, ar.

_m,esa.are-'*w-

Rrporting Rsquirtm:nts 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (continued)

12) RXE-89-002-A,"VIPRE-01 Core Thermal-Hydraulic Analysis Methods l

for Comanche Peak Steam Electric Station Licensing Applications", September 1993. l

13) RXE-91-001-A," Transient Analysis Methods for Comanche Peak l

Steam Electric Station Licensing Applications", October 1993. l

14) RXE-91-002-A, " Reactivity Anomaly Events Methodology", October l

1993. l

15) RXE-90-007-A,"Large Break Loss of Coolant Accident Analysis l

Methodology", April 1993. l

16) TXX-88306, " Steam Generator Tube Rupture Analysis", March 15, 1988.
17) RXE-91-005-A," Methodology for Reactor Core Response to '

l Steamline Break Events," February 1994. l

18) RXE-94-001-A," Safety Analysis of Postulated inadvertent Boron Dilution Event in Modes 3,4, and 5," February 1994.
19) RXE-95-001-P-A, "Small Break Loss of Coolant Accident Analysis l

Methodology," September 1996. l c. The core operating limits shall be determined nach that all applicable limits (e.g., fuel thermal mechanical limits, core therr,ial 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 upoa issuance for each reload cycle to the NRC. (continued) COMANCHE PEAK-UNITS 1 AND 2 5.0-34 Anendnumt No. #, 67 _ _ - _. _._}}

Retrieved from "https://kanterella.com/w/index.php?title=ML20211J946&oldid=4007084"