ML20196F887
| ML20196F887 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 12/03/1998 |
| From: | Pulsifer R NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20196F891 | List: |
| References | |
| NUDOCS 9812070150 | |
| Download: ML20196F887 (8) | |
Text
I p ur g
UNITED STATES s
j NUCLEAR RE2ULATORY COMMISSION 2
WASHINGTON, D.C. 2006H001 COMMONWEALTH EDISON COMPANY AND MIDAMERICAN ENERGY COMPANY DOCKET NO. 50-254 QUAD CITIES NUCLEAR POWER STATION. UNIT 1 AMENDMENT TO FACILITY OPERATINQ LICENSE Amendment No.182 License No. DPR-29 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Commonwealth Edison Company (the licensee) dated August 14,1998, as supplemented on October 13 and Nove.mber 23,1998, complies with the standarus and requirements of the Atomic Energy Act of 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, 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 actitities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and l
E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the l
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 amendment, and paragraph 3.B. of Facility Operating License No. DPR-29 is hereby amended to read as follows:
9812070150 981203 PDR ADOCK 05000254 P
. /
i B.
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.182 are hereby incorporated in the license. The i
licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance and shall be implemented within 30 days.
l FOR THE NUCLEAR REGULATORY COMMISSION i
\\
l'
/
V Robert M. Pulsifer, Project Manager I
Project Directorate lil-2 Division of Reactor Projects - lil/IV Office of Nuclear Reactor Regulation
Attachment:
l Changes to the Technical Specifications Date of Issuance: December 3,1998 l
]
1
1 ATTACHMENT TO LICENSE AMENDMENT NO.182 FEILITY OPERATING LICENSE NO. DPR-29 DOCKET NO. fiO-254 i
)
Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by the captioned amendment number and contain marginallines indicating the area of change.
REMOVE INSERT la l-1-1a 2-1 2-1b 3/4.11-1a 6-16b I
i i
L l
l I
l i
j I
4 k
4
__ _. _ _. _i TABLE OF CONTENTS TOC DEFINITIONS
~
SECTION PAGE 9
Section 1 DEFINITIONS ACTIO N ;...........................................................
1 - 1 l
AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)............
1-1 C H AN N E L..........................................................
1 - 1 i
CHANNEL CALIBRATION.............................................
1-1 C HAN N E L C H E CK...................................................
1 -1 CHANNEL FUNCTIONAL TEST.........................................
1-2 CORE ALTERATION................................................
1 -2 CORE OPERATING LIMITS REPORT (COLR).............................
1-2 CRITICAL POWER RATIO (CPR).......................................
1-2 DO SE EQU IVALENT l-131..............................................
1 -2 j
i FRACTION OF LIMITING POWER DENSITY (FLPD) -
- (applicable to GE fuel).................................................
1 -3 FRACTION OF RATED THERMAL POWER (FRTP).........................
1-3 FREQUENCY NOTATION.............................................
1-3 FUEL DESIGN LIMITING RATIO (FDLRX)................................
1-3 FUEL DESIGN LIMITING RATIO FOR CENTERLINE MELT (FDLRC)............
1-3 IDENTIFIED LEAKAGE...............................................
1 -3 LIMITING CONTROL ROD PATTERN (LCRP)..............................
1-3
. LINEAR HEAT GENERATION RATE (LHGR).............................
1-3 LOGIC SYSTEM FUNCTIONAL TEST (LSFT)...............................
1-4 a
QUAD CITIES - UNIT 1 la Amendment No. 182 y
o
Dsfinitions 1.0 1.0 DEFINITIONS r
l
[
The following terms are defined so that uniform interpretation of these specifications may be achieved. The defined terms appear in capitalized type and shall be applicable throughout these
- Technical Specifications.
ACTION ACTION shall be that part of a Specification which prescribes remedial measures required under l
designated conditions.
j AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)
The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) shall be applicable to a specific planar height and is equal to the sum of the LINEAR HEAT GENERATION RATE (s) for all the fuel rods in the specified bundle at the specified height divided by the number of fuel j
rods in the fuel bundle.
CHANNEL A CHANNEL shall be an arrangement of a sensor and associated components used to evaluate plant variables and generate a single protective action signal. A CHANNEL terminates and loses its identity where single action signals are combined in a TRIP SYSTEM or logic systen..
i CHANNEL CAllBRATION I
A CHANNEL CAllBRATION shall be the adjustment, as necessary, of the CHANNEL output such that it responds with the necessary range and accuracy to known values of the parameter which the CHANNEL monitors. The CHANNEL CAllBRATION shall encompass the entire CHANNEL including the required sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST. The CHANNEL CAllBRATION may be performed by any series l
of sequential, overlapping or total CHANNEL steps such that the entire CHANNEL is calibrated.
CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment of CHANNEL behavior during operation by observation. This determination shallinclude, where possible, comparison of the CHANNEL indication and/or status with other indications and/or status derived from independent instrument CHANNEL (s) measuring the same parameter.
I l
l l
l I
h l
L QUAD CITIES - UNIT 1 1-la Amendment No.182 1
(.
l SAFETY LIMITS 2.1
~
2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 1
l
'2d SAFETY LIMITS THERMAL POWER. Low Pressure of Low Flow h
2.1.A THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow.
l l
l APPLICABILITY: OPERATIONAL MODE (s) 1 and 2.
ACTION:
With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam l
dome pressure less than 785 psig or core flow less than 10% of rated flow, be in at least HOT l
SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.
I
. THERMAL POWER'. Hiah Pressure and Hiah Flow 2.1.B The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.11 with the l
reactor vessel steam dome pressure greater than or equal to 785 psig and core flow greater than or equal to 10% of rated flow. During single recirculation loop operation, this MCPR limit shall he increased by 0.01.
APPLICABILITY: OPERATIONAL MODE (s) 1 and 2.
I-ACTION:
l With MCPR less than the above applicable limit and the reactor vessel steam dome pressure greater than or equal to 785 psig and core flow greater than or equal to 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.
QUAD CITIES - UNIT 1 2-1 b Amendment No. 182
POWER DISTRIBUTION LIMITS APLHGR 3/4.11.A 3.11 - LIMITING CONDITIONS FOR OPERATION 4.11 - SURVEILLANCE REQUIREMENTS A. AVERAGE PLANAR LINEAR HEAT A. AVERAGE PLANAR LINEAR HEAT GENERATION RATE GENERATION RATE l
All AVERAGE PLANAR LINEAR HEAT The APLHGRs shall be verified to be equal l
GENERATION RATES (APLHGR) shall not to or less than the limits specified in the exceed the limits specified t.: the CORE CORE OPERATING LIMITS REPORT.
OPERATING LIMITS REPORT.
1.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, i
APPLICABILITY:
2.
Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at least OPERATIONAL MODE 1, when THERMAL 15% of RATED THERMAL POWER, and POWER is greater than or equal to 25% of RATED THERMAL POWER.
3.
Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with a j
LIMITING CONTROL ROD PATTERN for ACTION:
With an APLHGR exceeding the limits 4.
The provisions of Specification 4.0.D l
specified in the CORE OPERATING LIMITS are not applicable.
REPORT:
!~
1.
Initiate corrective ACTION within 15 l
minutes, and I
2.
Restore APLHGR to within the required limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
With the provisions of the ACTION above not met, reduce THERMAL POWER to less than 25% of RATED THERMAL POWER withia the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
i i
I QUAD CITIES - UNIT 1 3/4.11-la Amendment No.182
(
. ~
,c Reportirqiflaquirements 6.9 l
ADMINISTRATIVE CONTROLS (14) ANFB Critical Power Correlation, ANF-1125(P)(A) and Supplements 1 and 2, Advanced Nuclear Fuels Corporation, April 1990.
l (15) Advanced Nuclear Fuels Corporation Critical Power Methodology for Boiling Water Reactors / Advanced Nuclear Fuels Corporation Critical Power Methodology for Boiling Water Reactors: Methodology for Analysis of Assembly Channel l
Bowing Effects /NRC Correspondence, ANF-524(P)(A), Revision 2, Supplement 1 Revision 2, Supplement 2, Advanced Nuclear Fuels Corporation, November 1990.
(.
(16) COTRANSA 2: A Computer Program for Boiling Water Reactor Transient Analyses, ANF 913(P)(A) Volume 1 Revision 1 and Volume 1 Supplements 2,3, and 4, Advanced Nuclear Fuels Corporation, August 1990.
(17) Advanced Nuclear. Fuels Corporation Methodology for Boiling Water Reactors EXEM BWR Evaluation Model, ANF-91-048(P)(A), Advanced Nuclear Fuels Corporation, January 1993.
(18) Commonwealth Edison Topical Report NFSR-OO91, " Benchmark of CASMO/MICROBURN BWR Nuclear Design Methods," Revision 0, Supplements 1 and 2, December 1991, March 1992, and May 1992, respectively; SER letter dated March 22,1993.
(19) ANFB Critical Power Correlation Application for Coresident Fuel, EMF-L 1125(P)(A), Supplement 1, Appendix C, Siemens Power Corporation, August l
1997.
i 1
(20) ANFB Critical Power Correlation Determination of ATRIUM-9B Additive ConstantUncertainties, ANF-1125(P)(A), Supplement 1 Appendix E, Siemens Power Corporation, September 1998, i
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. The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions 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.
6.9.B Special Reports Special reports shall be submitted to the Regional Administrator of the NRC Regional Office within the time period specified for each report.
i
' OUAD CITIES - UNIT 1 6-16b Amendment No.182 1
l
.. -