ML072080462
| ML072080462 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 07/26/2007 |
| From: | NRC/NRR/ADRO/DORL/LPLII-1 |
| To: | |
| Stang J, NRR/DORL, 415-1345 | |
| Shared Package | |
| ML071830539 | List: |
| References | |
| TAC MD3328, TAC MD3329 | |
| Download: ML072080462 (6) | |
Text
.3-(4) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; (5) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproducts and special nuclear materials as may be produced by the operation of McGuire Nuclear Station, Units I and 2, and; (6) Pursuant to the Act and 10 CFR Parts 30 and 40, to receive, possess and process for release or transfer such byproduct material as may be produced by the Duke Training and Technology Center.
C. This renewed operating 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 The licensee is authorized to operate the facility at a reactor core full steady state power level of 3411 megawatts thermal (100%).
(2) Technical Specifications The Technical Specifications contained InAppendix A, as revised through Amendment No. 243are hereby Incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.
(3) Updated Final Safety Analysis Report The Updated Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21 (d), as revised on December 16, 2002, describes certain future activities to be completed before the period of extended operation.
Duke shall complete these activities no later than June 12, 2021, and shall notify the NRC in writing when implementation of these activities Is complete and can be verified by NRC Inspection.
The Updated Final Safety Analysis Report supplement as revised on December 16, 2002, described above, shall be Included In the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71 (e)(4), following Issuance of this renewed operating license.
Until that update is complete, Duke may make changes to the programs described in .such supplement without prior Commission approval, provided that Duke evaluates each.such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.
Renewed License No. NPF-9 Amendment No. 243
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(4) Pursuant to the Act and 10 CFR Parts 30, 40.and 70, to receive, posiess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; (5) Pursuant to the Act and 10 CFR Parts 30,40 and 70, to possess, but not separate, such byproducts. and special nuclear materials as may be produced by the operation of McGuire Nuclear Station, Units I and 2; and, (6). Pursuant to the Act and 10 CFR Parts 30 and 40, to receive, possess and process for release or transfer such byproduct material as may be produced by the Duke Training andTechnology Center.
C. This renewed operating 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 The licensee is authorized to operate the facility at a reactor core full steady state power level of 341.1 megawatts thermal (100%).
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 224 are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.
(3) Updated Final Safety Analysis Report The Updated Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21(d), as revised on December 16, 2002, describes certain future activities to be completed before the period of extended operation.
Duke shall complete these actvitles no later than March 3, 2023, and shall notify the NRC in writing when implementation of these activities Is complete and can be verified by NRC Inspection.
The Updated Final Safety Analysis Report supklement as revised on December 16,2002, described above, shall be Included In the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71(e)(4), following issuance of this renewed operating license.
Until that update is complete, Duke may make changes to the programs described in such supplement without prior Commission approval, provided that Duke evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59, and otherwise complies with the requirements in that section.
Renewed License No. NPF-17 Amendment No. 224
TABLE OF CONTENTS B 2.0 SAFETY LIMITS (SLs)
B 2.1.1 Reactor Core SLs ................................... B 2.1.1-1.
B 2.1.2 Reactor Coolant System (RCS) Pressure SL ................................... B 2.1.2-1 B 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY ........ B 3.0-1 B 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY ....................... B 3.0-9 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.1 SHUTDOWN MARGIN (SDM) ......................................................... B 3.1.1-1 B 3.1.2 Core Reactivity ................................................................................ B 3.1.2-i B 3.1.3 Moderator Temperature Coefficient (MTC) ...................................... B 3.1.3-1 B 3.1.4 Rod Group Alignment Limits ............................................................ B 3.1.4-1 B 3.1.5 Shutdown Bank Insertion Limits ....................................................... B 3.1.5-1 B 3.1.6 Control Bank Insertion Limits ........................................................... B 3.1.6-1 B 3.1.7 Rod Position Indication ................................................................... B 3.1.7-1 B 3.1.8 PHYSICS TESTS Exceptions ................................................... ...... B 3.1.8-1 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1 Heat Flux Hot Channel Factor (FQ(X.Y,Z)) ......... ......... B 3.2.1-1 B 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (FAH(X,Y)) ................... B 3.2.2-1 B 3.2.3 AXIAL FLUX DIFFERENCE (AFD) .................................... B 3.2.3-1 B 3.2.4 QUADRANT POWER TILT RATIO (QPTR) .................................... B 3.2.4-1 B 3.3 INSTRUMENTATION B 3.3.1 Reactor Trip System (RTS) Instrumentation .................................... B 3.3.1-1 B 3.3.2 Engineered Safety Feature Actuation System (ESFAS)
Instrumentation ................................. B 3.3.2-1 B 3.3.3 Post Accident Monitoring (PAM) Instrumentation ............................. B 3.3.3-1 B 3.3.4 Remote Shutdown System ............................. B 3.3.4-1 B 3.3.5 Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation .B 3.3.5-1 B 3.3.6 Not Used ........................................... ..................................... B 3.3.6-1 B 3.4 REACTOR COOLANT SYSTEM (RCS)
B 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits .... ............................ B 3.4.1-1 B 3.4.2 RCS Minimum Temperature for Criticality .................. B 3.4.2-1 B 3.4.3 RCS Pressure and Temperature (P/T) Limits ................ B 3.4.3-1 B 3.4.4 RCS Loops-MODES I and 2 .......................... B 3.4.4-1 B 3.4.5 RCS Loops-MODE 3 ............................................. 1.B 3.4.5-1 B 3.4.6 RCS Loops-MODE 4. ........ ................. ...... B 3.4.6-1 B 3.4.7 RCS Loops-MODE 5, Loops Filled ...................... B 3.4.7-1 B 3.4.8 RCS Loops-MODE 5, Loops Not Filled ............................ .............. B 3.4.8-1 B 3.4.9 Pressurizer ...................................................................................... B 3.4.9-1 B 3.4.10 Pressurizer Safety Valves............. :.................................................. B 3.4.10-1 B 3.4.11 Pressurizer Power Operated Relief Valves (PORVs) ....................... B 3.4.11-1 B 3.4.12 Low Temperature Overpressure Protection (LTOP) System ............ B 3.4.12-1 B 3.4.13 RCS Operational LEAKAGE ............................ B 3.4.13-1 McGuire Units 1 and 2 1 ,, i No.
Amendment Nos. 243, 224
Containment Isolation Valves 3.6.3 3.6 CONTAINMENT SYSTEMS 3.6.3 Containment Isolation Valves LCO 3.6.3 Each containment isolation valve shall be OPERABLE.
APPLICABILITY: MODES 1, 2,3, and 4.
ACTIONS NOTES
- 1. Penetration flow path(s) except for containment purge supply and/or exhaust isolation valves for the lower compartment, upper compartment, and incore instrument room may be unisolated intermittently under administrative controls.
- 2. Separate Condition entry is allowed for each penetration flow path.
- 3. Enter applicable Conditions and Required Actions for systems made inoperable by containment isolation valves.
- 4. Enter applicable Conditions and Required Actions of LCO 3.6.1, "Containment," when isolation valve leakage results in exceeding the overall containment leakage rate acceptance criteria.
CONDITION REQUIRED ACTION COMPLETION TIME A. NOTE A.1 Isolate the affected 4 hours0.167 days <br />0.0238 weeks <br />0.00548 months <br /> Only applicable to penetration flow path by penetration flow paths use of at least one closed with two containment and de-activated automatic isolation valves, valve, closed manual valve, blind flange, or check valve inside One or more penetration containment with flow flow paths with one through the valve secured.
containment isolation valve inoperable except AND for purge valve or reactor building bypass leakage not within limit.
(continued)
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Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.3.1 Verify each containment purge supply and exhaust valve 31 days for the lower compartment, upper compartment, and incore instrument room is sealed closed, except for one purge valve in a penetration flow path while in Condition E of this LCO.
SR 3.6.3.2 Not Used. I SR 3.6.3.3 NOTE-Valves and blind flanges in high radiation areas may be verified by use of administrative controls.
Verify each containment isolation manual valve and blind 31 days flange that is located outside containment or annulus and not locked, sealed, or otherwise secured and required to be closed during accident conditions is closed, except for containment isolation valves that are open under administrative controls.
(continued)
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Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.6.3.4 NOTE-Valves and blind flanges in high radiation areas may be verified by use of administrative controls.
Verify each containment isolation manual valve and blind Prior to entering flange that is located inside containment or annulus and MODE 4 from not locked, sealed, or otherwise secured and required to MODE 5 if not be closed during accident conditions is closed, except for performed Within containment isolation valves that are open under the previous administrative controls. 92 days SR 3.6.3.5 Verify the isolation time of automatic power operated In accordance with containment isolation valve is within limits. the Inservice Testing Program In accordance with SR 3.6.3.6 Perform leakage rate testing for containment purge lower the Containment and upper compartment and incore Instrument room Leakage Rate I valves with resilient seals. Testing Program SR 3.6.3.7 Verify each automatic containment isolation valve that is 18 months not locked, sealed or otherwise secured in position, actuates to the isolation position on an actual or simulated actuation signal.
(continued)
Initc I and'2 34 .6.3-6 Uina6Amendmen' KNos 2431, 22"04,