ML20076G411
| ML20076G411 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 10/12/1994 |
| From: | Berkow H Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20076G412 | List: |
| References | |
| NUDOCS 9410190313 | |
| Download: ML20076G411 (13) | |
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NUCLEAR REGULATORY COMMISSION Mf WASHINGTON, D.C. 20555 0001
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DUKE POWER COMPANY DOCKET NO. 50-369 McGUIRE NUCLEAR STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 149 License No. NPF-9 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the McGuire Nuclear Station, Unit 1 (the facility), Facility Operating License No. NPF-9 filed by the Duke Power Company (11censee) dated May 24, 1994, as supplemented August 4 and September 8, 1994 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as-set forth in 10 CFR Chapter I; 8.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulaticr.s 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 set forth in 10 CFR Chapter I; i
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; i
and 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.
9410190313 941012 PDR ADOCK 05000369 P
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Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Facility Operating License No. NPF-9 is hereby amended to read as follows:
Technical Specifications The_ Technical Specifications contained in Appendix A, as revised through Amendment No. 149, are hereby incorporated into this license.
The licensee shall operate the facility in accordance t
with the Technical Specifications and the Environmental Protection Pl an.
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 He bert N. Berkow, Director Project Directorate II-3 Division of Reactor Projects - I/II' Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuar.ce:
October 12, 1994
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21 UNITED STATES p
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NUCLEAR REGULATORY COMMISSION 2
WASHINGTON, D.C. 2055!H)D01 l
' DUKE POWER COMPANY DOCKET NO. 50-370 McGUIRE NUCLEAR STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 131 License No. NPF-17 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment to the McGuire Nuclear Station, Unit 2 (the facility), Facility Operating License No. NPF-17 filed by the Duke Power Company (licensee) dated May 24, 1994, as supplemented August 4 and September 8,1994 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations as 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 Comission; C.
There is reasonable assurance (1) 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 set forth in 10 CFR Chapter I; D.
The issuance of this amendment will not be inimical to the comon 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 of the Comission's regulations and all applicable requirements have been satisfied.
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Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated.in the attachment to this license amendment, and Paragraph 2.C.(2) of Facility Operating License No.
.i NPF-17 is hereby amended to read as follows:
Technical Specifications The Technical Specifications contained in Appendix A, as revised i
through Amendment No. 131
, are hereby incorporated into this license.
The licensee shall operate.the facility in accordance with the Technical Specifications and the Environmental Protection l
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 Her ert N. Berkow, Director Project Directorate II-3 Division of Reactor Projects - I/II Office of Nuclear _ Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance:
October 12, 1994 l
I l
i
ATTACHMENT TO LICENSE AMENDMENT NO. 149 FACILITY OPERATING LICENSE NO. NPF-9 DOCKET N0. 50-369 AN_Q TO LICENSE AMENDMENT NO. 1 31 FACILITY OPERATING LICENSE NO. NPF-17 DOCKET N0. 50-370 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the areas of change.
Remove Paaes Insert Paaes 3/4 9 - 1 3/4 9 - 1 3/4 9 - la 3/4 9 - 16 3/4 9 - 16 3/4 9 - 16a 6 - 21 6 - 21 6 - 21a 6 - 21a 6 - 21b 8 3/4 9 - 1 B 3/4 9 - 1 B 3/4 9 - la B 3/4 9 - 3 B 3/4 9 - 3 8 3/4 9 - 3a
I 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION 1
LIMITING CONDITION FOR OPERATION 3.9.1 The boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained uniform and sufficient to ensure that the more restrictive of the following reactivity conditions is met:
j a.
Either a K,, of 0.95 or less, or b.
A boron concentration of greater than or equal to the minimum boron concentration specified in the Core Operating Limits Report.
APPLICABILITY: MODE 6*, with the reactor vessel head closure bolts less than fully tensioned or with the head removed.
j ACTION:
With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity 1
changes and initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm baron or its equiva-lent until K,, d to greater than or equal to the minimum boron concentration is reduced to less than or equal to 0.95 or the boron concentra-tion is restore specified in the Core Operating Limits Report, whichever is the more restrictive.
SURVEILLANCE REQUIREMENTS 4.9.1.1 The more restrictive of the above two reactivity conditions shall be determined prior to:
a.
Removing or unbolting the reactor vessel head, and b.
Withdrawal of any full length control rod in excess of 3 feet from its fully inserted position within the reactor vessel.
4.9.1.2 The boron concentration of the Reactor Coolant System and the refuel-ing canal shall be determined by chemical analysis at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
4.9.1.3 NV-250 shall be verified closed under administrative control at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; or, NV-131, NV-140, NY-176, NV-468, NV-808, and either NV-132 or NV-1026 shall be verified closed under administrative control 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 necessary to makeup to the RWST during refueling operations.
- The reactor shall be maintained in MODE 6 whenever fuel is in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.
McGUIRE - UNITS 1 AND 2 3/4 9-1 Amendment No. 149 (Unit 1)
Amendment No.131 (Unit 2)
REFUELING OPERATIONS 3f4.9.12 FUEL STORAGE - SPENT FUEL STORAGE POOL LIMITING CONDITION FOR OPERATION 3.9.12 Fuel is to be stored in the spent storage pool with:
a.
The boron concentration in the spent fuel pool maintained at greater than or equal to the minimum boron concentration specified in the Core Operating Limits Report; and b.
Storage in Region 2 restricted to irradiated fuel which has decayed at least 16 days and one of the following:
1) fuel whi:h has been qualified in accordance with Table 3.9-1; or 2)
Fuel which has been qualified by means of an analysis using NRC approved methodology to assure with a 95 percent probability at a 95 percent confidence level that k,,, is no greater than 0.95 including all uncertainties; or 3)
Unqualified fuel stored in a checkerboard configuration.
In the event checkerboard storage is used, one row between normal stor-age locations and checkerboard storage locations will be vacant.
APPLICABILITY:
During storage of fuel in the spent fuel pool.
ACTION:
a.
Suspend all actions involving the movement of fuel in the spent fuel pool if it is determined a fuel assembly has been placed in the incorrect Region until such time as the correct storage location is determined. Move the assembly to its correct location before resumption of any other fuel movement.
b.
Suspend all actions involving the movement of fuel in the spent fuel pool if it is determined the pool boron concentration is less than 2000 ppm, until such time as the boron concentration is increased to the minimum boron concentration specified in the Core Operating Limits Report or greater.
c.
The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.9.12a.
Verify all fuel assemblies to be placed in Region 2 of the spent fuel pool are within the enrichment and burnup limits of Table 3.9-1 or mentatl,o,n ;or.95 by checking the assemblies' design and burnup docu-that k s0 the assemblies' qualifying analysis documentation respectively.
b.
Verify at least once per 31 days that the spent fuel pool boron con-centration is greater than the minimum boron concentration specified in the Core Operating Limits Report.
McGUIRE - UNITS I and 2 3/4 9-16 Amendment No. 149 (Unit 1)
Amendment No. 131 (Unit 2)
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ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT 6.9.1.9 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a re%ad cycle for the following:
1.
Moderator Temperature Coefficient BOL and E0L limits and 300 ppm surveillance limit for Specification 3/4.1.1.3, 2.
Shutdown Bank Insertion Limit for Specification 3/4.1.3.5, 3.
Control Bank Insertion Limits for Specification 3/4.1.3.6, i
4.
Axial Flux Difference limits, target band *, and APL** for Specification 3/4.2.1, i
l 5.
Heat Flux Hot Channel Factor, F"", K(Z), W(Z)**, APL*** and W(Z),t** for j
Specification 3/4.2.2, and Nuclear Enthalpy Rise Hot Channel Fa~ctor, FI,*ification Y/4.2.3., o 6.
Factor Multiplier, MF,,****,
limits for Spec 7.
Overtemperature and Overpower Delta T setpoint parameter values for Specification 2.2.1.
8.
Boric Acid Storage System and Refueling Water Storage Tank volume and i
boron concentration limits for Specifications 3/4.1.2.5 and 3/4.1.2.6.
)
9.
Accumulator and Refueling Water Storage Tank boron concentration limits for Specification 3/4.5.1 and 3/4.5.5.
10.
Reactor Coolant System and refueling canal boron concentration limits for Specification 3/4.9.1.
11.
Spent fuel pool boron concentration limits for Specification 3/4.9.12.
Reference 5 is not applicable to target band and APL".
References 4 and 5 are not applicable to W(Z), APL*, and W(Z),t.
]
Reference 1isnotapplicabletoFI,.
Reference 5 is not applicable to F[and MF 33 g
((Unit 1)
McGUIRE - UNITS 1 and 2 6-21 Amendment No.
Unit 2)
Amendment No.
i
ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by NRC in:
1.
WCAP-9272-P-A, " WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY,"
July 1985 (H Proprietary).
(Methodology for Specifications 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limit, 3.1.3.6 - Control l
Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor.)
2.
WCAP-10216-P-A, " RELAXATION OF CONSTANT AXIAL OFFSET CONTROL FQ SURVEILLANCE TECHNICAL SPECIFICATION," June 1983 (H Proprietary).
(Methodology for Specifications 3.2.1 - Axial Flux Difference (Relaxed Axial Offset Control) and 3.2.2 - Heat Flux Hot Channel Factor (W(Z) surveillance requirements for FQ Methodology.)
3.
WCAP-10266-P-A Rev. 2, "THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE," Harch 1987 (H Proprietary).
)
(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)
i 4.
BAW-10168P, Rev.1, "B&W Loss-of-Coolant Accident Evaluation Model for Recirculating Steam Generator Plants," September 1989 (B&W Proprietary).
(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)
5.
DPC-NE-2011P, " Duke Power Company Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors," March 1990 (DPC Proprietary).
(Methodology for Specifications 2.2.1 - Reactor Trip System Instrumentation Setpoints, 3.1.3.5 - Shutdown Rod Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor.)
6.
DPC-NE-300lP, " Multidimensional Reactor Transients and Safety Analysis Physics Parameter Methodology," March 1991 (DPC Proprietary).
(Methodology for Specification 3.1.1.3 - Moderator Temperature Coeffi-cient, 3.1.3.5 - Shutdown Rod Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor.)
7.
DPC-NE-2010P, " Duke Power Company McGuire Nuclear Station Catawba Nuclear Station Nuclear Physics Methodology for Reload Design," April 1984 (DPC l
Proprietary).
(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient, Specification 3.9.1 - RCS and Refueling Canal Boron Concentration, and Specification 3/4.9.12 - Spent Fuel Pool Boron Concentration.)
McGUIRE - UNITS I and 2 6-21 a Amendment No. 149 (Unit 1)
Amendment No. 131 (Unit 2)
ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT 8.
DPC-NE-3002, "FSAR Chapter 15 System Transient Analysis Methodology,"
August 1991.
(Methodology used in the system thermal-hydraulic analyses which determine the core operating limits) 9.
DPC-NE-3000, Rev.1, " Thermal-Hydraulic Transient Analysis Methodology,"
May 1989.
(Modeling used in the system thermal-hydraulic analyses) 10.
DPC-NE-1004A, " Nuclear Design Methodology Using CASMO-3/ SIMULATE-3P,"
November 1992.
(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient.)
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 supple-ments 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.
SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the NRC Regional Office within the time period specified for each report.
McGUIRE - UNITS I and 2 6-21b Amendment No.149 (Unit 1)
Amendment No. 131 (Unit 2)
3/4.9 REFUELING OPERATIONS BASES 3/4.9.1 BORON CONCENTRATION The limitations on reactivity conditions during REFUELING ensure that:
l (1) the reactor will remain subcritical during CORE ALTERATIONS, and (2) a i
uniform baron concentration is maintained for reactivity control in the water i
volume having direct access to the reactor vessel. These limitations are con-sistent with the initial conditions assumed for the boron dilution incident in i
the accident analyses.
k/k conservative allowance for uncertainties.The value of 0.95 or less for K,, the minimu includes a 1% delta Similarly, concentration value specified in the Core Operating Limits Report or greater includes a conservative uncertainty allowance of 50 ppm boron.
i The Reactor Makeup Water Supply to the Chemical and Volume Control (NV) System is normally isolated during refueling to prevent diluting the Reactor Coolant System boron concentration.
Isolation is normally accomplished by closing valve NV-250. However, isolation may be accomplished by closing valves NY-131, NV-140, NV-176, NV-468, NV-808, and either NV-132 or NV-1026, when it is neces-sary to makeup water to the Refueling Water Storage Tank during refueling i
operations.
3/4.9.2 INSTRUMENTATION The OPERABILITY of the Source Range Neutron Flux Monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core.
3/4.9.3 DECAY TIME The minimum requirement for reactor subcriticality prior to movement of irradiated fuel assemblies in the reactor vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short-lived fission products.
l This decay time is consistent with the assumptions used in the accident analyses.
3/4.9.4 CONTAINMENT BUILDING PENETRATIONS i
The requirements on containment building penetration closure and OPERABILITY of the Reactor Building Containment Purge Exhaust System HEPA filters and charcoal adsorbers ensure that a release of radioactive material within containment will be restricted from leakage to the environment or fil-tered through the HEPA filters aad charcoal adsorbers prior to discharge to the atmosphere. The OPERABILITY and closure restrictions are sufficient to restrict i
radioactive material release from a fuel element rupture based upon the lack of containment pressurization potential while in the REFUELING MODE. Operation of the Reactor Building Containment Purge Exhaust System HEPA filters and charcoal adsorbers and the resulting iodine removal capacity are consistent with the assumptions of the accident analysis. The methyl iodide penetration test cri-teria for the carbon samples have been made more restrictive than required for the assumed iodine removal in the accident analysis because the humidity to be j
seen by the charcoal adsorbers may be greater than 70% under normal operating conditions.
McGUIRE - UNITS 1 AND 2 B 3/4 9-1 Amendment No. 149 (Unit 1)
Amendment No. 131 (Unit 2) t y
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BASES i
3/4.9.9 and 3/4.9.10 WATER LEVEL - REACTOR VESSEL and STORAGE P0OL The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly. The minimum water depth is consistent with the assumptions of the accident analysis.
3/4.9.11 FUEL HANDLING VENTILATION EXHAUST SYSTEM i
The limitations on the Fuel Handling Ventilation Exhaust System ensure that all radioactive material released from an irradiated fuel assembly will be filtered through the HEPA filters and charcoal adsorbers prior to discharge to the atmosphere. The OPERABILITY of this system and the resulting iodine removal capacity are consistent with the assumptions of the accident analyses.
ANSI N510-1975 will be used as a procedural guide for surveillance testing.
The methyl iodide penetration test criteria for the carbon samples have been made more restrictive than required for the assumed iodine removal in the acci-dent analysis because the humidity to be seen by the charcoal adsorbers may be greater than 70% under normal operating conditions.
3/4.9.12 FUEL STORAGE - SPENT FUEL STORAGE POOL The requirements for fuel storage in the spent fuel pool on 3.9.12 (a) and (b) ensure that:
(1) the spent fuel pool will remain subcritical during fuel stor-age; and (2) a uniform boron concentration is maintained in the water volume in the spent fuel pool for reactivity control. The value of 0.95 or less for K which includes all uncertainties at the 95/95 probability / confidence level,,
as described in Section 9.1.2.3.1 of the FSAR is the acceptance criteria for fuel storage in the spent fuel pool. Table 3.9-1 is conservatively developed in accordance with the acceptance criteria and methodology referenced in Section 5.6 of the Technical Specifications. Storage in a checkerboard configuration in Region 2 meets all the acceptance criteria referenced in Section 5.6 of the Technical Specifications and is verified.in a semi-annual basis after initial verification through administrative controls.
The Action Statement applicable to fuel storage in the spent fuel pool ensures that:
(1) the spent fuel pool is protected from distortion in the fuel storage pattern that could result in a critical array during the movement of fuel; and (2) the minimum boron concentration is maintained at the limit specified in the Core Operating Limits Repcrt during all actions involving movement of fuel in the spent fuel pool.
.he Surveillance Requirements applicable to fuel storage in the spent fuel pool l
ensure that:
(1) fuel stored in Region 2 meets the enrichment and burnup limits of Table 3.9-1 or the K
<; 0.95 acceptance criteria of an analysis using NRC approved methodology,;,,and (2) the minimum boron concentration meets the limit specified in the Core Operating Limits Report.
McGUIRE - UNITS I and 2 B 3/4 9-3 Amendment No. 149 (Unit 1)
Amendment No. 131 (Unit 2)