ML20206M939

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Proposed Tech Specs,Revising List of Referenced Documents in TS 5.6.5b Re COLR Requirements
ML20206M939
Person / Time
Site: Mcguire, Catawba, McGuire  Duke Energy icon.png
Issue date: 05/06/1999
From:
DUKE POWER CO.
To:
Shared Package
ML20206M928 List:
References
NUDOCS 9905170085
Download: ML20206M939 (19)
v • d • e


Text

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Attachment la Technical specifications for McGuire Units 1 and 2 Marked copy

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1 4 9905170085 990506 PDR ADOCK 05000369 P _

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  • t .{ Reporting Requirements 5.6 i l

1 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 Specif'cation 3.2.2, 4
7. Overtemperature and Overpower Delta T setpoint parameter 1 values for Specification 3.3.1,
8. Accumulator and Refueling Water Storage Tank boron l concentration limits for Specification 3.5.1 and 3.5.4,
9. Reactor Coolant Systern and refueling canal boron concentration limits for Specification 3.9.1,
10. Spent fuel pool boron concentration limits for Specification 3.7.14,
11. SHUTDOWN MARGIN for Specification 3.1.1.

I

b. The analytical 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).

1

![ CA 10216 OFF ET C O

,R" QS ON vel CON C EC AXl ICA J SP CIFl ION" une Prop ' tary f I

f/. WCAP-10266-P-A Rev. 2, "THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE *,

March 1987, Q$(, Proprietary).

BAW-10168P-A, "B&W Loss-of-Coolant Accident Evaluation 3 [. Model for Recirculating Steam Generator Plants," Rev.1, SER dated Janua 1991; Rev. 2, SE@ted August 22,1  : Rev. ,

, SER dated J 15,1994 (B& 3roprietary). a.4 22,. N*V*** bet?&1994 (continued)

McGuire Units 1 and 2 5.6-3 Amendment Nos. g

  • t -( Reporting Requirements 5.6 t

5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued) y[ DPC-NE-2011PA, " Duke Power Company Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors," March,1990 (DPC Proprietary).

.E p $[ DPC-NE-3001 PA, " Multidimensional Reactor Transients and Safety Analysis Physics Parameter Methodology," November,

]4 g 1991 (DPC Proprietary).

i (( DPC-NF-2010A, " Duke Power Company McGuire Nuclear Station Catawba Nuclear Station Nuclear Physics Methodology for Reload

]^y q.p Design," June,1985.

kg 3 yq DPC-NE-3002A,-T;..wp. Rev SAR Chapjer 15 System yO Transient Analysis Me , SER dates. y ~=, d.

V pier ==vy 5,1997.

Jg d DPC NE-g ermal-Hydraulic Transient ysis  !

j.

Methodolog dat CD/C. A*pnitlop, 3

7J.d/

DPC-NE-1004A, Rev.1, " Nuclear Design Methodology Using CASMO-3/ SIMULATE-3P," SER dated April 26,1996.

D 4g p g

/8[. DPC-NE-2004P-A, Rev.1, " Duke Power Company McGuire and Catawba Nuclear Stations Core Thermal-Hydraulic Methodology 1 O

og using VIPRE 01," SER dated February 20,1997 (DPC Proprietary).

Nq sW 1. hP NE- A, Alv.1, "

.]

i lij

  • 1 Fuel Necharhcal eloa Anaftsis i IMe yf Mar (-BW fpel," ()ctobe[1 (DP Pr@rieta ). )

St

.p //1d. j DPC-NE-2005P-A, Rev.1, " Thermal Hydraulic Statistical Core da Design Methodology," SER dated November 7,1996 (DPC

%h Ok Proprietary).

, /2, f. DPC-NE-2008P-A, " Fuel Mechanical Reload Analysis y Methodology Using TACO 3," SER dated April 3,1995 (DPC i s , Proprietary).

q l 1

A5. B -10 3P- Fuel Ga re C ' rion, W el/

pa ,Ju ,1995.

/ ]

A a(pts fa l,eu.9e Sms// d<eek EccS i gg +, e } ((Ss'Q t. NOTNY 0 4,+ ,m w n . +e. .

(continued)

McGuire Units 1 and 2 5.6-4 Amendment Nos. k

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  • i .:  ;

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-I Attachment 1b Technical Specifications for Catawba Units.1 and 2 ^

Marked Copy i

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Reporting Requirrments-5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

1. Moderator Temperature Coefficient BOL and EOL limits and 300 ppm surveillance limit for Specification 3.1.3,
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 for Specification 3.2.2,
7. Overtemperature and Overpower Delta T setpoint parameter 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 for Specification 3.9.1,
10. Spent fuel pool boron concentration limits for Specification 3.7.15,
11. SHUTDOWN MARGIN for Specification 3.1.1.
b. The analytical 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 M Proprietary).
2. CAF1-10 16- RELAKATI )N O COfISTAfE AL FF ET OL FO UR I CETE ilCA PE IFl TI *

, ne 83 Pr rietpry).

2f. WCAP-10266-P-A Rev. 2. "THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE",

March 1987, M Proprietary).

(continued)

Catawba Units 1 and 2 5.6 Amendment Nos. 5

... ....4 . . .

l

- Reporting Requirements 5.6 s

5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued) 3[. BAW-10168P A,"B&W Loss-of Model for Recirculating Steam G nt Accident Evaluation rator Plants," Rev.1, SER dated January,1991; Rev. 2, SE Dated August 22,1 Rev. 3, SER Dated 46 -

L. z2j 1994 (B&W Proprietary)'Mmber zC,If 9C ,

//g. DPC-NE-2011P- uke Power Company Nuclear Desigri I Methodology for Core Operating Umits of Westinghouse Reactors," March,1990 (DPC Proprietary).

.5'/. DPC-NE-3001 P-A, " Multidimensional Reactor Transients and Safety Analysis Physics Parameter Methodology," November, 1991 (DPC Proprietary). '

[g /. DPC-NF-2010A, " Duke Power Company McGuire Nuclear Station Catawba Nuclear Station Nuclear Physics Methodology for Reload Design," June,1985.

S 7A. DPC-NE-3002A, T.'.=;;h Re / SAR Chapter 5 Transient Analysis Meth -i 20 1^00.

2. , SER dated fk' 5 :esw, e4 S.,/199'  %,

8 /. DPC-NE- P) a ./* ermal-Hydraulic

~

Transient Analysis Methodolog , RD

$b,e,:~r l'G, lit} (D#C. feb/dO '

s27 10^"

7 1d.

j DPC-NE-1004A, Rev.1, " Design Methodology Using CASMO-3/ SIMULATE-3P," SER Dated April 26,1996.

/o p. DPC-NE-2004P-A, Rev.1, " Duke Power Company McGuire and Catawba Nuclear Stations Core Thermal-Hydraulic Methodology using VIPRE-01," SER dated February 20,1997 (DPC Proprietary).

Re .1, " uel

[2. _ pPC- E-2 1P i Ma BW et,"

cha cal eloa An ysis oprie ry

/VIett r19 (D C

// [. DPC-NE-2005P-A, Rev.1, " Thermal Hydraulic Statistical Core Design Methodology," SER dated November 7,1996 (DPC

' Proprietary).

/2 . DPC-NE-2008P-A, ' Fuel Mechanical Reload Analysis Methodology Using TACO 3," SER dated April 3,1995 (DPC Proprietary).

(continued) ..

Catawba Units 1 and 2 5.6-4 Amendment Nos.

  • ~

Neporting Requirrments

, (A)C/\ f- / 00 SQ ~ f. k

  • LA}t.tk in \seg3 g, Sm I t 5s I

15vesk ECCS E AkItIS" 41. e l 4 s / 1 Laeme.uw c..te., % 3ast mes CD e4<m ).

5.6 Reg rting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

5. B W-1 183P A, Fu I Rof Ga Pre ure rite on, Fu I mp y, J' y,19 5.

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The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as I

SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 Ventilation Systems Heater Report When a report is required by LCO 3.6.10, " Annulus Ventilation System (AVS),"

LCO 3.7.10, " Control Room Area Ventilation System (CRAVS)," LCO 3.7.12, Auxiliary Building Filtered Ventilation Exhaust System (ABFVES)," LCO 3.7.13,

" Fuel Handling Ventilation Exhaust System (FHVES)," or LCO 3.9.3,

' Containment Penetrations," a report shall be submitted within the following 30 days. The report shall outline the reason for the inoperability and the planned actions to retum the systems to OPERABLE status.

5.6.7 PAM Report When a report is required by LCO 3.3.3, " Post Accident Monitoring (PAM)

Instrumentation,* a report shall be submitted within the following 14 days. The report shall outline the preplanned attemate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.8 Steam Generator Tube inspection Report

a. The number of tubes plugged in each steam generator shall be reported to the NRC within 15 days following completion of the program, 7

/4, DK - N& - 200W-A, "We.s+64onse Fue I Tr~wr/h D

Weport, " SCR. 6+e[ (D PC Pnp'e+avy).

(continued)

Catawba Units 1 and 2 5.e.5 Amendment Nos.

Attachment 2a Reprinted Technical-specifications for McGuire Units 1 and 2 Remove Pages Insert Pages-5.6-3 -5.6-3 5.6-4 5.6-4 w

-i-4

R: porting Requiram:nts

' 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 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 for Specification 3.9.1,
10. Spent fuel pool boron concentration limits for Specification 3.7.14,
11. SHUTDOWN MARGIN for Specification 3.1.1.
b. The analytical 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).
2. WCAP-10266-P-A Rev. 2, "THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE",

March 1987, (W Proprietary). _

3. BAW-10168P-A,"B&W Loss-of-Coolant Accident Evaluation 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).

(continued)

McGuire Units 1 and 2 5.6-3 Amendment Nos:

b ,r2-- - i -- o- - ... . . . . .. . . . . . . . . . , . . . . . . .. . . . . . . . .

..m.._ _m..

I

~

Reporting R:quirem:nts q 5.6 1 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

4. DPC-NE-2011PA, " Duke Power Company Nuclear Design ]

Methodology for Core Operating Limits of Westinghouse ]'

Reactors," March,1990 (DPC Proprietary).

5. DPC-NE-3001 PA, " Multidimensional Reactor Transients and Safety Analysis Physics Parameter Methodology," November, 1991 (DPC Proprietary).
6. DPC-NF-2010A, " Duke Power Company McGuire Nuclear Station -

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 l Methodology," SER dated October 14,1998 (DPC Proprietary).. j
9. DPC-NE-1004A, Rev.1, " Nuclear Design Methodology Using CASMO-3/ SIMULATE-3P," SER dated April 26,1996.
10. DPC-NE-2004P-A, Rev.1, " Duke Power Company McGuire and 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). l
12. DPC-NE-2008P-A, " Fuel Mechanical Reload Analysis 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 (DPC Proprietary).

(continued)

McGuire Units 1 and 2. 5.6-4 Amendment Nos:

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Attac h nt 2b Reprinted Technical Specifications; for Catawba Units 1 and 2 Remove Pages Insert Pages 5.6-3 5.6-3 5.6-4 5.6-4 5.6-5 5.6-5 l

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'3 Reporting RDauirim:nts-5.6 -

5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

1. Moderator Temperature Coefficient BOL and EOL limits and 300 ppm surveillance limit for Specification 3.1.3,-
2. Shutdown Bank Insertion Limit for Specification 3.1.5,
3. Control Bank Insertion Limits for Specification 3.1.6,
4. Axlal 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 for Specification 3.2.2,
7. Overtemperature and Overpower Delta T setpoint parameter 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 for Specification 3.9.1,
10. Spent fuel pool boron concentration limits for Specification 3.7.15,
11. SHUTDOWN MARGIN for Specification 3.1.1.
b. The analytical 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).
2. WCAP-10266-P-A Rev. 2, "THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE',

March 1987, (W Proprietary).

(continued)

Catawba Units 1 and 2 5.6-3 Amendment Nos.

.1.. ..

3 R porting R:quir:m:nts l 5.6 l

l 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

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).
4. DPC-NE 2011P-A, " Duke Power Company Nuclear Design l Methodology for Core Operating Limits of Westinghouse Reactors," March,1990 (DPC Proprietary).
5. DPC-NE-3001 P-A, " Multidimensional Reactor Transients and l l Safety Analysis Physics Parameter Methodology," November, 1991 (DPC Proprietary).
6. DPC-NF-2010A, " Duke Power Company McGuire Nuclear Station l l

Catawba Nuclear Station Nuclear Physics Methodology for Reload

  • Design," June,1985.
7. DPC-NE-3002-A, Rev. 3 "FSAR Chapter 15 System Transient Analysis Methodology," SER dated February 5,1999.

l l 8. DPC-NE-3000PA, Rev. 2 " Thermal-Hydraulic Transient Analysis Methodology," SER Dated October 14,1998 (DPC Proprietary). i

! 9. DPC-NE-1004A, Rev.1, " Design Methodology Using CASMO- l 3/ SIMULATE-3P," SER Dated April 26,1996. j

10. DPC-NE-2004P-A, Rev.1, " Duke Power Company McGuire and l l l Catawba Nuclear Stations Core Thermal-Hydraulic Methodology 1 l 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 i Proprietary).
12. DPC-NE-2008P A," Fuel Mechanical Reload Analysis l 1 Methodology Using TACO 3," SER dated April 3,1995 (DPC i Proprietary). l

.)

1, I

(continued)

Catawba Units 1 and 2 5.6-4 Amendment Nos.

. - - . - . . - - m ______._ .__ _._ _ _ _ _ ___ _ _ . . . . ._ _ _. _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _

c Rtporting Requiram::nts 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

13. WCAP-10054-P A," Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," August 1985 (W Proprietary).
14. DPC-NE-2009P-A, " Westinghouse Fuel Transition Report," SER dated (DPC Proprietary).
c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, - .

Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the. safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 Ventilation Systems Heater Report When a report is required by LCO 3.6.1_0, " Annulus Ventilation System (AVS),"

LCO 3.7.10, " Control Room Area Ventilation System (CRAVS)," LCO 3.7.12, Auxiliary Building Filtered Ventilation Exhaust System (ABFVES)," LCO 3.7.13,

" Fuel Handling Ventilation Exhaust System (FHVES)," or LCO 3.9.3,

" Containment Penetrations," a report shall be submitted within the following 30 days. The report shall outline the reason for the inoperability and the planned actions to return the systems to OPERABLE status.

5.6.7 PAM Report When a report is required by LCO 3.3.3, " Post Accident Monitoring (PAM)

Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

=1 5.6.8 Steam Generator Tube Inspection Report i J

a. The number of tubes plugged in each steam generator shall be reported '

to the NRC within 15 days following completion of the program; (continued)

Catawba' Units 1 and 2 5.6-5 Amendment Nos.

_ _ . . - - - - - . - - - - - _ J

c ,.

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Attachment 3 Description of Proposed Changes and Technical Justification. i

Background

In letters to the NRC dated July 22, 1998 and. October 22, 1998'  ;

Duke Energy Corporation proposed changes to the McGuire and l Catawba Nuclear Stations' Technical Specifications (TS) to 1 permit transition to Westinghouse fuel. During the time that has j I

transpired since the dates of these referenced letters,-Duke has identified the need to supplement these-previous submittals.

'This supplement updates the list of reference documents contained in TS 5.6.5b for both McGuire and Catawba. The changes proposed in this supplement are administrative in nature. The l

-details of the proposed changes are discussed in the subsequent paragraphs. ]

Description and Technical Justification TS 5.6.5b:

1. Current Item 2, WCAP-10216-P-A, " Relaxation of Constant Axial Offset Control FQ Surveillance Technical Specification," June 1983 (!f Proprietary), is being deleted. This topical report is replaced by Item 5 (proposed Item 4), DPC-NE-2011P-A, j

" Duke Power Company Nuclear Design Methodology for Core ')

Operating Limits of Westinghouse Reactors," March, 1990 (DPC l Proprietary).

2. For Current Item 4 (Proposed Item 3), BAW-10168P-A, "B&W Loss-of-Coolant Accident Evaluation Model for Recirculating Steam Generator Plants," SER dates are being added. This is an administrative addition to this reference document.
3. For Current Item 8 (Proposed Item _7), DPC-NE-3002A, "FSAR Chapter 15 System Transient Analysis Methodology," Rev.2, the revision number is being changed to Rev. 3. Also,-the SER date is being changed to February'5, 1999. This administrative change updates this reference document to current status.

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i Attachment 3 Description of Proposed Changes and Technical Justification l

4. For Current Item 9 (Proposed Item 8), DPC-NE-3000PA,

" Thermal-Hydraulic Transient Analysis Methodology,"'Rev. 1, the revision number is being changed to 2. Also, the SER date is being changed to October 14, 1998 and DPC Proprietary is being added. This administrative change. updates this reference. document to current status.

5. Current Item 12, DPC-NE-2001P-A, Rev. 1, " Fuel Mechanical Reload Analysis Methodology for Mark-BW' fuel," October.1990 (DPC proprietary) is being deleted. This topical report is being replaced by Item 14 (proposed Item 12), DPC-NE-2008P-A, Fuel Mechanical Reload Analysis Methodology Using TACO 3."
6. Current Item 15, BAW-10183P-A, " Fuel Rod Gas Pressure Criterion," B&W Fuel Company, July, 1995 is being deleted.

This document is referenced in Item 14 (proposed Item 12),

DPC-NE-2008P-A, " Fuel Mechanical Reload Analysis Methodology.

Using TACO 3", and therefore does not require a separate listing in TS 5.6.5b.

7. Proposed Item 13, WCAP-10054-P-A, " Westinghouse Small Break ECCS Evaluation Model using the NOTRUMP Code," August 1985 (W Proprietary) is being added to the list of reference documents contained in TS 5.6.5b. Since McGuire and Catawba are transitioning back to Westinghouse fuel, the Westinghouse Small Break LOCA Evaluation Model is now again applicable to these two stations. Consequently, this vendor topical report is being returned to the list of reference documents contained in Ts 5.6.5b.
8. Proposed Item 14, DPC-NE-2009P-A, " Westinghouse Fuel Transition Report," SER dated (DPC Proprietary), was included and discussed in the original July 22, 1998 Duke submittal.

2 a

Attachment 4 No Significant Hazards Consideration Determination The following discussion is a summary of the evaluation of the changes' contained in this proposed amendment, as supplemented, against the 10CFR50.92(c) requirements to demanstrate that all three standards are satisfied. This evaluation has been revised since originally being submitted to the NRC by Duke Energy Corporation letter dated July 22, 1999. This revision-includes discussion of additional changes proposed to McGuire and Catawba TS 5.6-5b. These additional. changes are contained in a supplemental submittal made by Duke Energy Corporation letter dated May 6, 1999. These additional changes update the list of reference documents used for each station's COLR.

A no significant hazards consideration is indicated if operation of the facility in accordance with the proposed amendment would not:

1. Involve a significant increase in _the probability or consequences of an accident previously evaluated, or
2. Create the possibility of a new or different kind of accident from any accident previously evaluated, or
3. Involve a significant reduction in a margin of safety.

First Standard No. Implementation of this LAR, as supplemented, would not involve a significant increase in the probability or consequences of an accident previously evaluated. The_ revised Reactor Core Safety Limits Figure further restricts acceptable operation, and moving an uncertainty factor from the Technical Specifications to the COLR_does not exempt this factor from regulatory restrictions. COLR parameters are generated by NRC approved methods.and use NRC approved reference documents to ensure that previously evaluated accidents remain bounding. Therefore, no accident probabilities or consequences will be impacted.

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Attachment 4 No Significant Hazards Consideration Determination-Second Standard No. Implementation of this LAR, as supplemented, would not' create the possibility of a new or different kind of accident from any previously evaluated. The revised Reactor Core Safety Limits Figure further restricts acceptable operation, and moving an uncertainty factor from the Technical Specifications to the COLR does not exempt this factor from regulatory restrictions.

Since the parameter in question is not being-deleted, the possibility of a new or different kind of accident from'any previously evaluated does not exist. The revision to the list of NRC approved COLR reference documents, as included in the supplement to this LAR, only contains administrative changes which have no impact on the possibility of any accident.

Third Standard No. Implementation of this LAR, as supplemented, would not ,

involve a significant reduction in a margin of safety. Margin of '

safety is related to the confidence in the ability of the fission product barriers to perform their design functions during and following an accident situation. These barriers include the fuel cladding, the reactor coolant system, and the containment system.

Use of ZIRLO" cladding material has been reviewed and approved in Reference 1 (as listed in chapter 2.1 of Topical Report DPC-NE-2009/DPC-NE-2009P, Duke Power Company W'stinghouse e Fuel Transition Report) . ZIRLO" cladding has been extensively used in Westinghouse nuclear reactors. The changes proposed in this LAR, as supplemented, are necessary to ensure that the performance of the fission product barriers (cladding) will not be impacted following the replacement of one fuel design for another.

Additionally, the proposed changes to the list of NRC approved.

COLR reference documents contained in the supplement to this LAR are administrative in nature and have no significant impact on any safety margin.

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Attachment 4 No Significant Hazards Consideration Detemination Based upon the preceding evaluation, Duka has concluded that implementation of this LAR, as supplemented, at McGuire and Catawba Nuclear Stations will not involve a significant hazards consideration.

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