ML20203G509
| ML20203G509 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 02/16/1999 |
| From: | Rossbach L NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML17191B241 | List: |
| References | |
| NUDOCS 9902190331 | |
| Download: ML20203G509 (4) | |
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION l
2' WASHINGTON, D.C. 30ee6-0001
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COMMONWEALTH EDISON COMPANY DOCKET NO. 50-237
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DRESDEN NUCLEAR POWER STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 171 License No. DPR-19 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Commonwealth Edison Company (the licensee) dated August 14,1998, as supplemented on October 13,1998, and December 23,1998, complies with the standards 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 j
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 wedment will not be inimical to the common defense and security or to the hadei a:4 safety of the public; and E.
The issuance of this a mer.dment is in accordance with 10 CFR Part 51 of the Commission's regulatiem and all applicable requirements have been satisfed.
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 License No. DPR-19 is hereby amended to read as follows:
l 9902190331 990216 PDR ADOCK 05000237 P
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(2) Technical Soecifications i
The Technical Specifications contained in Appendix A, as revised through Amendment No. 171, are hereby incorporated in the license. The licensae 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.
FOR THE NUCLEAR REGULATORY COMMISSION i
m k Lawrence W. Rossbach, Project Manager 1
Project Directorate lll-2 Division of Reactor Projects - lil/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of lasuance: February,16,' 1999 I
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UNITED STATES g
j NUCLEAR REGULATORY COMMISSION e
WASHINGTON, D.C. 30866 0001 o<n f
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l COMMONWEALTH EDISON COMPANY DOCKET NO. 50-249 DRESDEN NUCLEAR POWER STATION. UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.166 License No. DPR-25 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Commonwealth Edison Company (the licensee) dated August 14,1998, as supplemented on October 13,1998, and December 23,1998, complies with the standards 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;
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B.
The facility will operate in conformity with the application, the provisions of the Act and the rules and regulations of the Commission; i
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 amendment will not be inimical to the common defense and secunty or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the i
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-25 is hereby amended to read as follows:
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. - L, 2-B.
Technical Soecifications The Technical Specifications contained in Appendix A, as revised through i
Amendment No. 166, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
i 3.
This license amendment is effective as of the date of its issuance and shall be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION k
e Lawrence W. Rossbach, Project Manager Project Directorate lil-2 Division of Reactor Projects - lil/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date ofissuance: February 16, 1999 1
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s ATTACHMENT TO LICENSE AMENDMENT NOS.171 AND 166 FACILITY OPERATING LICENSE NOS. DPR-19 AND DPR-25 DOCKET NOS. 50-237 AND 50-249 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 I
I 1-1 1-1 2-1 2-1 i
3/4.11-1 3/4.11-1 3/4.11-4 3/4.11-4 5-5 5-5 6-16 6-16 l
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TABLE OF CONTENTS TOC DEFINITIONS SECTION PAGE Section 1 DEFINITIONS ACTION....................................................
1-1 l
AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)....
1-1.'
' C HAN N E L...................................................
1-1 i
CHANNEL CALIBRATION.......................................
1-1 C HAN N EL C H E C K.............................................
1-1 CHANNEL FUNCTIONAL TEST..................................
1-2 CORE ALTE RATION...............................
1-2 CORE OPERATING LIMITS REPORT (COLR).....................
1-2 CRITICAL POWER RATIO (CPR)................................
1-2 DOSE EQUIVALENT l-131....................................
1-2 FRACTION OF RATED THERMAL POWER (FRTP)..................
1-3 FR EQ U ENCY NOTATION......................................
1-3 FUEL DESIGN LIMITING RATIO (FDLRX)........................
1-3 FUEL DESIGN LIMITING RATIO for CENTERLINE MELT (FDLRC)......
1-3 I DE NTI FI E D LEAKAG E.........................................
1-3 LIMITING CONTROL ROD PATTERN (LCRP).......................
1-3 LINEAR HEAT GENERATION RATE (LHGR).......................
1-3 LOGIC SYSTEM FUNCTION L TEST (LSFT).......................
1-3 MINIMUM CRITICAL POWER RATIO (MCPR).......................
1-4 OFFSITE DOSE CALCULATION MANUAL (ODCM).................
1-4 c
DRESDEN - UNITS 2 & 3 i
Amendment Nos.171; 166
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Dsfinitions 1.0
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1.0 DEFINITIONS
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The following terms are defined so that uniform interpretation of these specifications may be l
l 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.
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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 i
all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle.
CHANNEL A CHANNEL shall be an arrangement of a sensor and associated components used to evaluate l
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 system.
CHANNEL CAllBRATION l
A CHANNEL CAllBRATION shall be the adjustment, as necessary, of the CHANNEL output l
such that it responds with the necessary range and accuracy to known values of the parameter i
i which the CHANNEL monitors. The CHANNEL CAllBRATION shall encompass the entire
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CHANNEL including the required sensor and alarm and/or trip functions, and shall include the 1
l CHANNEL FUNCTIONAL TEST. The CHANNEL CALIBRATION may be performed by any series i
of sequential, overlapping or total CHANNEL steps such that the entire CHANNEL is calibrated.
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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.
DRESDEN - UNITS 2 & 3 11 Amendment Nos.171; 166 4
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SAFETY LIMITS 2.1 2.O SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS __.
2d SAFETY LIMITS THERMAL POWER Low Pressure or Low Flow 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.
APPLICABILITY: OPERATIONAL MODE (s) 1 and 2.
ACTION:
With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 785 psig or core flow less than 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.
c THERMAL POWER Hioh Pressure and Hioh Flow 2.1.B The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.10 for Unit 3 and 1.09 fo Unit 2 with 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. During single recirculation loop operation, this MCPR limit shall be increased by 0.01.
APPLICABILITY: OPERATIONAL MODE (s) 1 and 2.
ACTION:
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.
DRESDEN - UNITS 2 & 3 2-1 Amendment Nos.171; 166
4 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 q
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 in 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 />, 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 LIMITING CONTROL ROD PATTERN for ACTION:
With an APLHGR exceeding the limits 4.
The provisions of Specification 4.0.D specified in the CORE OPERATING LIMITS are not applicable.
REPORT:
1.
Initiate corrective action within 15 minutes, and 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 within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
DRESDEN - UNITS 2 & 3 3/4.11-1 Amendment Nos. 171; 166
4 POWER DISTRIBUTION LIMITS SLHGR 3/4.11.D 3.11 - LIMITING CONDITIONS FOR OPERATION 4.11 - SURVEILLANCE REQUIREMENTS D.
STEADY STATE LINEAR HEAT D.
STEADY STATE LINEAR HEAT GENERATION RATE GENERATION RATE The LINEAR HEAT GENERATION RATE The SLHGR shall be determined to be equal l
(LHGR) shall not exceed the STEADY STATE to orless than the limit:
LINEAR HEAT GENERATION RATE (SLHGR) limits specified in the CORE 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 />, OPERATING LIMITS REPORT.
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 APPLICABILITY:
15% of RATED THERMAL POWER, and OPERATIONAL MODE 1, when THERMAL 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 /> POWER is greater than or equal to 25% of when the reactor is operating with a RATED THERMAL POWER.
LIMITING CONTROL ROD PATTERN for SLHGR.
ACTION:
4.
The provisions of Specification 4.0.D are not applicable.
With an LHGR exceeding the SLHGR limits specified in the CORE OPERATING LIMITS REPORT:
1.
Initiate corrective ACTION within 15 minutes, and 2.
Restore the LHGR to within the SLHGR 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 within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
I DRESDEN - UNITS 2 & 3 3/4.11-4 Amendment Nos.171; 166 l
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t REACTOR CORE 5.3 5.0 DESIGN FEATURES l
5.2 REACTOR CORE Fuel Assemblies l
5.3.A The reactor core shall contain 724 fuel assemblies. Each assembly consists of a l
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matrix of Zircaloy clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide as fuel material. The assemblies may contain water rods or a i
3 water box. Limited substitutions of Zircaloy or ZlRLO or stainlest steel filler rods for fuel rods, in accordance with NRC-approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods, and shown by tests j
or analyses to comply with all fuel safety design bases. A limited number of lead test i
assemblies that have not completed representative testing may be placed in non-limiting core regions.
P Control Rod Assemblies 5'3.B The reactor core shall contain 177 cruciform shaped control rod assemblies. The control material shall be boron carbide powder (B C) and/or hafnium metal. The control rod assembly shall have a nominal axial absorber length of 143 inches.
DRESDEN - UNITS 2 & 3 5-5 Amendment Nos.171; 166 A
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R: porting R quirements 6.9 ADMINISTRATIVE CONTROLS
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(12)
ANF-1125 (P)(A), ANFB Critical Power Correlation Determination of ATRIUM-9B Additive Constant Uncertainties, Supplement 1, Appendix E, Siemens Power Corporation, September 1998.
c.
The core operating limits report 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 analysi: llmits) i of the safety analysis are met. The CORE OPERATING LIMITS REPORT, including
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any mid-cycle revisions of supplements thereto shall be provided on issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident inspector.
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6.9.8 Special Reports i
Special reports shall be submitted to the Regional Administrator of the NRC Regional Office within the time period specified for each report.
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4 DRESDEN - UNITS 2 & 3 6-16 Amendment Nos.171; 166
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