ML20212G113
| ML20212G113 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 09/22/1999 |
| From: | Emch R NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20212G118 | List: |
| References | |
| NUDOCS 9909290094 | |
| Download: ML20212G113 (16) | |
Text
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UNITED STATES
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j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20666-0001
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f DUKE ENERGY CORPORATION DOCKET NO. 50-369 McGUIRE NtJOl EAR STATION. UNIT 1 4
. AMENDMENT YnJALITY OPERATING LICENSE Amendment No.188 License No. NPF-0
- 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 !.icense No. NPF-9 filed by the Duke Energy Corporation (licensee) dated July 22,1996, and supplemented by letters dated October 22,1998, and January 28, May 6, June 24 and August 17,1999, 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; 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 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 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 of the Commission's regulations and all applicable requirements have been =?tisf:ed.
9909290094 990922 PDR ADOCK 05000369 P
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. 2. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraphs 2.C.(2) and 2.c.(13) of Facility Operating License No. NPF-9 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 188, are hereby incorporated into this license. The licensee shall
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operate the facility in accordance with the Technical Specifications.
(13) AdditionalConditions The Additional Conditions contained in Appendix C, as revised through Amendment No.188, are hereby incorporated into this license. Duke Energy Corporation shall operate the facility in accordance with the Additional Conditions.
- 3. This license amendment is effectivn as of its date of issuance and shall be implemented prior to beginning the instaNation at he Westinghouse fuel, currently projected to be Fuel Cycle 15.
FOR THE NUCLEAR REGULATORY COMMISSION fuAs//Cas/,2-Richard L. Emch, Jr., Chief, Section 1 Project Directorate 11 Division of Licensing Project Management Office of Nuclear Reactor Regulation l
Attachment:
Operating License and Technical Specification Changes l
Date of issuance: September 22, 1999 l
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k UNITED STATES
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NUCLEAR REGULATORY COMMISSION i
4 WASHINGTON, D.C. 20665 0001 1
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l DUKE ENERGY CORPORATION DOCKET NO. 50-370 McGUIRE NUCLEAR STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE i
Amendment No.169 License No. NPF-17
- 1. The Nuclear Regulatory Commission (the Commission) 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 Energy Corporation (licensee) dated July 22,1998, and supplemented by letters dated October 22,1998, and January 28, May 6, June 24, August 17 and September 15,1999, 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; B. The '. Omty 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 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 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 of the Commission's regulations and all applicable requirements have been satisfied.
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' 2. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraphs 2.C.(2) and 2.C.(13) of Facility Operating License No. NPF-17 is hereby amended :o read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 169, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications.
(13) Additional Conditions The Additional Conditions contained in Appendix D, as revised through Amendment No.169, are hereby incorporated into this license. Duke Energy Corporation shall operate the facility in accordance with the Additional Conditions.
- 3. This license amendment is effective as of its date of issuance and shall be implemented prior to beginning the installation of the Westinghouse fuel, currently projected to be Fuel Cycle 14.
FOR THE NUCLEAR REGULATORY COMMISSION Richard L. Emch, Jr., Chief, Section 1 Project Directorate 11 Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Operating License and Technical Specification Changes Date of Issuance: September 22, 1999
ATTACHMENT TO LICENSE AMENDMENT NO. 188 FACILITY OPERATING LICENSE NO. NPF-9 DOCKET NO.' 50-369 ATTACHMENTTO LICENSE AMENDMENT NO.169 FACILITY OPERATING LICENSE NO. NPF-17 DOCKET NO 50-370 Replace the following pages of Appendices C (for Unit 1) and D (for Unit 2), Additional Conditions, of the operating licenses with the attached revised pages. The revised pages are identified by amendment number and contain marginallines indicating the areas of change.
Remove Insert 2
Replace the following pages of 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 2.0-2 2.0-2 2.0-3 3.2.1-4 3.2.1-4 3.2.1-5 3.2.1-5 3.2.2-4 3.2.2-4 4.0-1 4.0-1 5.6-3 5.6-3 5.6-4 5.6-4 Replace the following pages of the Technical Specifications Bases document with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
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Remove n_
l B 3.2.1-11 B 3.2.1-11 B 3.2.2-9 8 3.2.2-9 l
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APPENDIX C ADDITIONAL CONDITIONS FACILITY OPERATING LICENSE NO. NPF-9 (Continued) l Amendment implementation Number
' Additional Condition Date
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l The maximum rod average burnup for any rod shall Within 30 days of l
be limited to 60 GWd/mtU until the completion of an date of amendment l
NRC environmental assessment supporting an increased limit.
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i i Amendment No.188
APPENDIX D ADDITIONAL CONDITIONS FACILtTY OPERATING LICENSE NO. NPF-17 (Continued)
Amendment implementation Number Additional Condition Date The maximum rod average burnup for any rod shall Within 30 days of be limited to 60 GWd/mtU until the completion of an date of amendment NRC environmental assessment supporting an increased limit.
. Amendment No.169 l
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I SLs 2.0 i
670=
1 DO NOT OPERATE IN THIS AREA l
660 650 640..
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2400 psia V 630.
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2280 psia
- C/)0 620' 2100 psia 610 P
'600 1945 psia 590 ACCEPTABLE OPERATION 580 0.0 -
0.2 0.4 -'
O.6
.0.8-1.0 1.2 Fraction of Rated Thermal Power Figure 2.1.1-1 Reactor Core Safety Limits -
Four Loops in Operation McGuire Units 1 an'd 2 2.0-2 Amendment Nos: 188 (Unit 1) 4 169 (Unit 2)
o Fo(X,Y,Z) 3.2.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY l
N OTE------
1.
Extrapolate F"o(X,Y,Z) using at least two measurements to 31 EFPD beyond the most recent measurement. If F"o(X,Y,Z)is within limits and the 31 EFPD extrapolation indicates:
F"o(X,Y,Z)extaneoarco Z Fb(X,Y,Z) "ampum and E"(X.Y.Zhxmpourto > E"o(X.Y.Z)
Fh(X,Y,Z) "exta4 pot 41eo Fh(X,Y,Z)"
then:
a.
Increase F"o(X,Y,Z) by the appropriate factor specified in the COLR and reverify F"o(X,Y,Z) s Fh(X,Y,Z) " a b.
Repeat SR 3.2.1.2 prior to the time at which F"o(X,Y,Z) $ Fb(X,Y,Z) " is extrapolated to not be met.
2.
Extrapolation of F"o(X,Y,Z)is not required for the initial flux map taken after reaching equilibrium conditions.
Verify F"o(X,Y,Z)s Fh(X,Y,Z) P Once within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after achieving equilibrium conditions after exceeding, by Z 10% RTP, the THERMAL POWER at which F"o(X,Y,Z) was last verified A__R 31 EFPD thereafter (contir ued)
McGuire Units 1 and 2 3.2.1-4 Amendment Nos.188 (Unit 1) 169 (Unit 2)
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l Fo(X,Y,Z)-
l 3.2.1
$URVEILLANCE REQUIREMENTS (continued) l SURVEILLANCE FREQUENCY l
SR 3.2.1.3 NOTES---
1.
Extrapolate F"(X,Y,Z) using at least two measurements to 31 EFPD beyond the most recent measurement. If F"o(X,Y,Z) is within limits
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and the 31 EFPD extrapolation indicates:
F*o(X,Y,Z) EXTRAPOLATED 2 Fh(X,Y,Z)*"SEXTRAPOtATED, 1
and E"a(X.Y.Ziex1R4Potarro > E"o(x.Y.Z)
Fb(X,Y,Z)""8 Fh(X,Y,Z)""8 ExTR,Potarso then:
a.
increase F"o(X,Y,Z) by the appropriate factor specified in the COLR and reverify F"(X,Y,Z) s Fb(X,Y,Z)""8; or b.
Repeat SR 3.2.1.3 prior to the time at which F"o(X,Y,Z) s Fh(X,Y,Z)""8 is extrapolated to not be met.
2.
Extrapolation of F"a(X,Y,Z)is not required for the initial flux map taken after reaching equilibrium conditions.
Verify c"o(X,Y,Z) s Fb(X,Y,Z)RPS.
Once within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after achieving equilibrium conditions after exceeding, by 2 i
10% RTP, the THERMAL POWER at which F"o(X,Y,Z) was last verified i
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31 EFPD thereaner i
McGuire Units 1 and 2 3.2.1-5 Amendment Nos. 188 Unit 1 169 Unit 2
F s@,W 3
3.2.2 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY l
SR 3.2.2.2 NOTES l
1.
Extrapolate F"3s(X,Y) using at least two j
measurements to 31 EFPD beyond the most recent measurement. If F"3s(X,Y) is within limits I
and the 31 EFPD extrapolation indicates:
F"3s(X,Y)exTa4potarco 2 Fin (X,Y)"exTnapot4Tro and E"3sMExTRAPOLATED
> E"3sn F'3s(X,Y)" EXTRAPOLATED Fin @,Y)"
then:
a.
Increase F",,(X,Y) by the appropriate factor specified in the COLR and reverify F"3s @,Y) $ P s @,W", w 3
b.
Repeat SR 3.2.2.2 prior to the time at which F"3s @,Y) 5 P s (X,Y)suav is 3
extrapolated to not be met.
2.
Extrapolation of F"3s (X,Y) is not required for the initial flux map taken after reaching equilibrium conditions.
Verify F"3s @,Y) 5 P s (X,Y)suav.
Once within 12 3
hours after achieving equilibrium conditions after exceeding, by 2 10% RTP, the THERMAL POWER at which F"3s (X,Y) was last verified AND 31 EFPD thereafter McGuire Units 1 and 2 3.2.2-4 Amendment Nos.188 (Unit 1) 169 (Unit 2) 1 L_
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Design Features 4.0 i
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4.0 DESIGN FEATURES 4.1 Site Location The McGuire Nuclear Station site is located at latitude 35 degrees,25 minutes,59 seconds north and longitude 80 degrees,56 minutes,55 seconds west. The Universal Transverse Mercator Grid Coordinates are E 504,669,256, and N 3,920,870,471. The
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site is in northwestern Mecklenburg County, North Carolina,17 miles north-northwest of Charlotte, North Carolina.
4.2 Reactor Core l
4.2.1 Fuel Assemblies l
l The reactor shall contain 193 fuel assemblies. Each assembly shall consist of a matrix of either ZlRLO" or Zircalloy fuel rods with an initial composition of l
natural or slightly enriched uranium dioxide (lJO ) as fuel material. Limited l
2 substitutions of ZlRLO*, zirconium alloy, or stainless steel filler rods for fuel l
l rods, in accordance with 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 or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be t
placed in nonlimiting core regions.
4.2.2 Control Rod Assemblies The reactor core shall contain 53 control rod assemblies. The control material shall be silver indium cadmium (Unit 1) silver indium cadmium and boron carbide (Unit 2) as approved by the NRC.
4.3 Fuel Storage 4.3.1 Criticality 1
4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:
a.
Fuel assemblies having a maximum nominal U-235 enrichment of 4.75 weight percent; b.
k,, s 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.1 of the UFSAR; (continued) l McGuire Units 1 and 2 4.0-1 Amendment Nos.188 (Unit 1) 169 (Unit 2) l l
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Reporting Requirements 5.6 5.6' Reporting Requirements 5.6.5 '
CORE OPERATING LIMITS REPORT (COLR) (continued)
' 2.
' Shutdown Bank Insertion Limit for Specification 3.1.5, 3.
Control Bank Insertion Limits for Specification 3.1.6, 4.
Axial Flux Difference limits for Specification 3.2.3, 5.
Heat Flux Hot Channel Factor for Specification 3.2.1, 6.
Nuclear Enthalpy Rise Hot Channel Factor limits for Specification 3.2.2, 7.
Overtemperature and Overpower Delta T setpoint parameter j
values for Specification 3.3.1, 8.
Accumulator and Refueling Water Storage Tank boron concentration limits for Specification 3.5.1 and 3.5.4, 9.
Reactor Coolant System and refueling canal boron concentration limits foi Specification 3.9.1, 1
10.
Spent fuel pool boron concentration limits for Specification 3.7.14, 11.
SHUTDOWN MARGIN for Specification 3.1.1.
b.
The snalytical 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).
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WCAP-10266-P-A Rev. 2. "THE 1981 VERSION OF 1
WESTINGHOUSE EVALUATION MODEL USING BASH CODE",
March 1987, (W Proprietary).
3.
BAW-10168P-A,"B&W Loss-of-Coolant Accident Evaluation l
Model for Recirculating Steam Generator Plants," Rev.1 SER dated January 22,1991; Rev. 2, SERs dated August 22,1996 and November 26,1996; Rev. 3, SER dated June 15,1994 (B&W Proprietary).
i (continued)
McGuire Units 1 and 2 5.6-3 Amendment Nos.188 (Unit 1) 169 (Unit 2)
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued) 4.
DPC-NE-2011PA, " Duke Power Company Nuclear Design l
Methodology for Core Operating Limits of Westinghouse Reactors," March,1990 (DPC Proprietary).
5.
DPC-NE-3001PA, " Multidimensional Reactor Transients and l
Safety Analysis Physics Parameter Methodology," November, 1991 (DPC Proprietary).
6.
DPC-NF-2010A, " Duke Power Company McGuire Nuclear Station l Catawba Nuclear Station Nuclear Physics Methodology for Reload Design," June,1985.
7.
DPC-NE-3002A, Rev. 3 "FSAR Chapter 15 System Transient Analysis Methodology," SER dated February 5,1999.
8.
DPC-NE-3000PA, Rev. 2 " Thermal-Hydraulic Transient Analysis Methodology," SER dated October 14,1998. (DPC Proprietary).
9.
DPC-NE-1004A, Rev.1, " Nuclear Design Methodology Using l
CASMO-3/ SIMULATE-3P," SER dated April 26,1996.
10.
DPC-NE-2004P-A, Rev.1, " Duke Power Company McGuire and l
Catawba Nuclear Stations Core Thermal-Hydraulic Methodology using VIPRE-01," SER dated February 20,1997 (DPC Proprietary).
11.
DPC-NE-2005P-A, Rev.1, " Thermal Hydraulic Statistical Core Design Methodology," SER dated November 7,1996 (DPC Proprietary).
12.
DPC-NE-2008P-A," Fuel Mechanical Reload Analysis l
Methodology Using TACO 3," SER dated April 3,1995 (DPC Proprietary).
13.
WCAP-10054-P-A, " Westinghouse Small Break ECCS Evaluation Model using the NOTRUMP Code," August 1985 (W Proprietary).
14.
DPC-NE-2009-P-A, " Westinghouse Fuel Transition Report, "SER dated September 22, 1999 (DPC Proprietary).
(continued)
McGuire Units 1 and 2 5.6-4 Amendment Nos. 188 (Unit 1) 169 (Unit 2)
Fm(X,Y)
B 3.2.2 BASES SURVEILLANCE REQUIREMENTS (continued) channel factor to the surveillance limit is likely to decrease below the value of that ratio when the measurement was taken.
Each of these extrapolations is applied separately to the enthalpy nse hot channel factor surveillance limit. If both of the extrapolations are unfavorable, i.e., if the extrapolated factor is expected to exceed the extrapolated limit and the extrapolated factor is expected to become a larger fraction of the extrapolated limit than the measured factor is of'the current Fmit, additional actions must be taken. These actions are to meet the F"m(X,Y) limit with the last F*m(X,Y) increased by the appropriate factor specified in the COLR or to ovaluate F"m(X,Y) prior to the point in time when the extrapolated values are expected to exceed the extrapolated limits. These alternative requirements attempt to prevent F"3n(X,Y) from exceeding its limit for any significant period of time without detection using the best available data. F $(X,Y) is not reouired to be extrapolated for the initial flux map taken after reaching equilibrium conditions since the initial flux map establishes the baselina measurement for future trending.
F"m(X,Y) is verified at power levels 10% RTP above the THERMAL POWER of its last verification,12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after achieving equilibrium conditions to ensure that F"m(X,Y) is within its limit at high power levels.
The Surveillance Frequency of 31 EFPD is adequate to monitor the change of power distribution with core burnup. The Surveillance may be done more frequently if required by the results of F"m(X,Y) evaluatiorts.
The Frequency of 31 EFPD is adequate to monitor the change of power distribution because such a change is sufficiently slow, when the plant is operated in accordance with the TS, to preclude adverse peaking factors between 31 day surveillances.
REFERENCES 1.
UFSAR Section 15.4.8 2.
10 CFR 50, Appendix A, GDC 26.
3.
4.
10 CFR 50.36, Technical Specifications, (c)(2)(ii).
5.
DPC-NE 2005P, " Duke Power Company Thermal Hydraulic Statistical Core Design Methodology", September 1992.
McGuire Units 1 and 2 B 3.2.2-9 Revision No.1
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Fo(X,Y,Z)
B 3.2.1 BASES I
SURVEILLANCE REQUIREMENTS (continued) than the measured factor is of :he current limit, additional actions must be taken. These actions are to meet the Fo(X,Y,Z) limit $/ith the last F"o(X,Y,Z) increased by the appropriate factor specLed in the COLR or l
to evaluate Fo(X,Y,Z) prior to the projected point in time when the extrapolated values are expected to exceed the extrapolated limits.
i These alternative requirements attempt to prevent Fo(X,Y,Z) from exceeding its limit for any significant period of time without detection using the best available data. F"o(X,Y,Z) is not required to be extrapolated for the initial flux map taken after reaching equilibrium conditions since the initial flux map establishes the baseline measurement for future trending. Also, extrapolation of F"o(X,Y,Z) limits are not valid for core locations that were previously rodded, or for core locations that were previously within *2% of the core height about the demand position of the rod tip.
Fo(X,Y,Z) is verified at power levels 210% RTP above the THERMAL POWER of its last verification,12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after achieving equilibrium conditions to ensure that Fo(X,Y,Z) is within its limit at higher power
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levels.
J The Surveillance Frequency of 31 EFPD is adequate to monitor the change of power distribution with core burnup. The Surveillance may be done more frequently if required by the results of Fo(X,Y,Z) evaluations.
The Frequency of 31 EFPD is adequate to monitor the change of power distribution because such a change is sufficiently slow, y Aan the plant is operated in accordance with the TS, to preclude adverse peaking factors between 31 day surveillances.
REFERENCES 1.
10 CFR 50.46,1974.
2.
UFSAR Section 15.4.8.
3.
10 CFR 50, Appendix A, GDC 26.
4.
10 CFR 50.36, Technical Specifications, (c)(2)(ii).
5.
DPC-NE-2011PA ' Duke Power Company Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors", March 1990.
McGuire Units 1 and 2 B 3.2.1-11 Revision No.1 1