ML20197G439
| ML20197G439 | |
| Person / Time | |
|---|---|
| Site: | Catawba, McGuire, Mcguire  |
| Issue date: | 12/17/1997 |
| From: | DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML20197G431 | List: |
| References | |
| NUDOCS 9712310017 | |
| Download: ML20197G439 (53) | |
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i , i Attachment I l Marked-up Changes to Technical Specifications A. Technical Specification Markups - CTS !
- 1. McGuire Unit 1
- 2. McGuire Unit 2
- 3. Catawba Unit 1
- 4. Catawba Unit 2 B. Amended ITS Submittal Pages - McGuire C. Amended ITS Submittal Pages - Catawba 9712310017 971217 PDR ADOCK 05000369 P PDR
1
- ADMINISTRATIVE CONTROLS
, J CORE OPERATING tlHITS 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 (W Proprietary). (Methodology for Specifications 3.1.1.3 - Moderator Tempe ature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limit 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.)
- 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 F o Methodology.)
- 3. WCAP-10266-P-A Rev. 2 "THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE", March 1987 (W Proprietary). -
(Methodology fo* Specification 3.2.2 - Heat Flux Hot Channel Factor.) l
- 4. BAW-10168P, W Loss-of-Coolant Accident Evaluation Model for Recirculating Steam Generator Plants,% SER dated January 1991-(B&W l Proprietary g h-gg l (Methodology for Specification 3.2.2 - Heat tiux" Hot Channel Factor.)
- 5. DPC-NE-2011PA, " Duke Power Company Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors," March 1990 (DPC Proprietary).
(Methodology for Specification 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-3001PA, Physics Parameter Methodology," "Hultidimensional November 1991 Reactor Transients (DPC Proprietary . and Safety) An (Methodology for S)ecification 3.1.1.3 - Moderator Temperature Coeffi-cient, 3.1.3.5 - Slutdown Rod Insertion Limits, 3.1.3.6 - Control Bank l 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-2010A, " Duke Power Company McGuire Nuclear Station Catawba Nuclear Station Nuclear Physics Methodology for Reload Design," June 1985 (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 - UNIT 1 6-21 Amendment No. MQ - l_ . . _ _ _ . - . - . _ _ . __ ._. . _ _ _ . . . _ _ - - _ _ . - -
mmmmmmmummu l . m
; ADMINISTRATIVE CONTR01S CORE OPERATING LIMITS REPORT (Continued) jfy/W4 # .C6K' 2.
- 8. DPC-NE-3D02,-itenr-1,#'FSAR Chapter 15 System Transient Analysis Methodology," SER dated ha D%. APR/t_ IG; 199 6 l p
(Methodology used in the system thermal-hydraulic analyses which determine the core operating limits)
- 9. DPC-NE-3000P, Rev. 1, " Thermal-Kydraulic Transient Analysis Methodology,"
SER dated De anbere1995 9b ] (Modeline ted in the system thersal-hydraulic analyses)
- 10. DPC-MF a,6 hcNakDesign N Methodology Using CASMO-3/ SIMULATE-3P,"
Novek.. -199t. S G g MrC h APRIt. 2.Gs d96 , (Methodo) ty for Specification 3.1.1.3 - Moderator Temperature I Coefficient.) g ,,
- 11. DPC-NE-2004P-A, " Duke Power Ccapany McGuire and Catawba Nuclear Stations -
Dec- t , lii-r (DPC Core Thermal-Hydraulic Methodology using Proprietary). G$~pVIPRE-01 "16krGb F66RdAP.Y' 78 I999 (Methodology for Specifications L2.1 - Ret.ctor Trip System Instrumenta-tion Setpoints, 3.2.1 - Axial Fiv Difference (AFD), and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor rNi(X,Y).)
- 12. DPC-NE-2001P-A, Rev.1 ' Fuel Mechanical Reload Analysis Methodology for -
Mark-BW fuel," October 1990 (DPC Proprietary). (Methodology for S
} g[Setpoints.)_ g6 pecification gf 2.2.1 - Reactor Trip System Instrumentation (f g E(f- 13.
DPC-1005P-A,6 mal Hydraulic Statistical Core Design Methodology,*. u M a ui 199& (DPC Proprietary).
$ (Methodology for Specification 2.2.1 - Reactor Trip System Instrumentation Setpoints, Specification 3.2.1 - Axial Fluz Difference, and 3.2.3 -
Nuclear Enthalpy Rise Hot Channel Factor).
- 14. RAW 40162PrA;-TAC 03-fuel-Pin Th:r.& Analysts Ev.per Code, Ei Fueb
-L ,,. m Xv - -1989s _
(Methodology used for Specification 2.2.1 - Reactor Trip System Instru- r mentation setpoints). ;
- 15. BAW-101B3P, Fuel Rod Gas Pressure Criterion, B&W Fuel Company, as approved by SER dated Fi.;rj-19% avty,' fyt.5' ;
(Used fa,r Specification 2.2.1, Reactor Trip System Instrumentation 1 Setpoints). '} The core operating limits shall be determined so that all applicable limits 3 (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear . sits 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 theretcr, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and T Resident Inspector.
]
J pc.-ac aog "Fucc Mscunwc4L 40mb A^'"YS'S Il
'Mo M 1
()SIA)G 1/b() $6 A hTTE6 AMIL 3, Ft95 O)PC PRofRicTMY} McGUIRE - UNIT 1 6-22 Amendment No. 1 M
!% h -- .u-m- - mm--w.-o.o.h ! ,1 ' ADMINISTRATIVE CONTROLS
) CORE OPERATING LIMITS REPORT The analytical methods used to detertine 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 (W Proprietary). (Methodology for Specificntions 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limit. 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat flux Hot Channel Factor, and L2.3 - Nuclear Enthalpy Rise Hot Channel Factor.) i
- 2. WCAP-10216-P-A, 'RELAXA110N OF CONSTANT AXIAL OFFSET CONTROL FQ SURVEILLANCE TECHNICAL SPECIFICATION", June 1983 (W Proprietary).
! (Methodology for Specifications 3.2.1 - Axial Flux Difference (Relaxed Axial Offset Control) snd 3.2.2 - Heat Flux Hot Channel Factor (W(Z) surveillance requirements for Fo Methodology.)
1 3. WCAP-10266-P-A Rev. 2, "THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL g USING BASH CODE", March 1987 (W Proprietary). 1 (Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.) a k 4. BAW-10168P,b &W Loss-of-Coolant Accident Evaluation Model for Recirculating Proprietary) . Meam G rnater'!n4Z4SER dated January 1991g f) W M (Methodology for pec Ticauon 3.2.2 - pleat Fliax Ho fisiinel Factor.) MbMWAQ L.AvGv5T % Aq (, j 5. DPC-NE-2011PA, " Duke Power Company Nuclear Design Methodology for Core
~
Operating Limits of Westinghouse Reactors," March 1990 (DPC Proprietary). (Methodology for Speci fication 2.2.1 - Reactor Trip System Instrumentation . i Setpoints, 3.1.3.5 - Shutdown Rod Insertion Limits, 3.1.3.6 - Control Bank i ; Insertion Limits, 3.2.; - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.P.3 - Nuclear Enthalpy Rise Hot Channel Factor.) 6. DPC-NE-3001PA, " Multidimensional Physics Parameter Methodology,' Novembcr 1991Reactor Transients (DPC Proprietary , and Safety) A f (Methodology for Saecification 3.1.1.3 - Moderator Temperature Coeffi-i cient, 3.1.3.5 - Sautdom Rod Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot j Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor.) l 7. DPC-NE-2010A, ' Duke Power Company McGuire Nuclear Station Catawba Nuclear ji Station Nuclear Physics Methodology for Reload Design," June 1985 (Methodology for Specif cation 3.1.1.3 - Moderator Temperature
- ] Coefficient, Specificat'on 3.9.1 - RCS and Refueling Canal Boron u Concentration, and Spec'fication 3/4.9.12 - Spent Fuel Pool Boron lg Concentration.) :
l l d McGUIRE - UNIT 2 6-21 Amendment No. 148 ,
ADMINISTRATIVE CONTR0!S l CORE OPE!LATING LIMITS REPORT (Continued)jrH ft.oWfl M 2 3
- 8. DPC-NE-3002, 4ev4,GFSAR Chapter 15 System Transient Analysis ,
Methodology," SER dated Mrkr 10;";. APAu. 2/9 't9 t 6 I (Methodology used in the rystem thermal-hydraulic analyses which determine i the core operating limits)
- 9. DPC-NE-3000P, Rev.1, 'Tnermal-Hydraulic Transient Analysis Methodology," l SER dated December g 5y2My (Modeling used igtpe 7,ystem thermal-hydraulic analyses)
- 10. DPC-NE-1004A,dN uc'leah Design Methodology Using CASMO-3/ SIMULATE-3P,"
kve&r, 222. Sg(f bgrsb APRit. %, ITyd (Methodology for Sper;ification 3.1.1.3 - Moderator Temperature Coefficient.) j
- 11. DPC-NE-2004P-A,6Duxe Power Company McGuire and Catawba Nuclear Stations Core Thermal-Hydrat,lic Methodology using VIPRE-01,"7--"-- Str (DPC Proprietary). Cgga y rg6 ftdA M 24,/ m (Methodology for Specifications 2.2.1 - Reactor Trip System Instrumenta-tion Setpoints, 3.2.1 - Axial Flux Difference (AFD), and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor foH(X,Y).)
- 12. DPC-NE-200lP-A, Rav.1, " Fuel Mechanical Reload Analysis Methodology for Hark-BW fuel," October 1990 (DPC Proprietary).
MMethodology for Specification 2.2.1 - Reactor Trip System Instrumentation f y b' f Setp )oints. pg ,,,,_ggy,g,
- 13. DPC4 005P-A,("Tharmal Hydraulic Statistical Core Design Methodology,"
[ F e,very 1995"(DPC Proprietary). (Methodology for Specification 2.2.1 - Reactor Trip System Instrumentation Setpoints, Specification 3.2.1 - Axial Flux Difference, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).
- 14. iWH0162P-A, TAG 034ueLAin.Theml_ Analysis Cc=put:r Cede, otu eg C..gany, Wu m die r1999, (Methodology usest for Specification 2.2.1 - Reactor Trip System Instru-mentation setpoints).
- 15. BAW-10183P, Fuel Rod Gas Pressure Criterion, B&W Fuel Company, as approved by SER dated felnMFy i;;4. 7W.9, q9y (Used for Specification 2.2.1, Reactor Trip System Instrumentation Setpoints). -
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 1 NRC De-" ment Control Des ( with copies to the Regional Administrator and Resident Inspector. Q bpc-NG-zrog dfijSt. 86CMAhCAL. Rswnb AlJALWLS NCG'ckAbbY U$1 A) G, -ff]C
"$6tZ 6RT&O ANEILbI9$5"(O W '
McGUIRE - UNIT 2 6-22 Amendment No. 157
ADMINISTRATIVE CONTROLS CORE OPEp,A_ LING LIMITS REPORT (Continued) (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-NF-2010A, " Duke Power Company McGuire Nuclear Station t uk.ua Huclear Station Nuclear Physics Methodology for Reload Design," June 1985 (Methodology fo- Specification 3.1.1.3 - Moderator Temperature Coefficient, Specification 4.7.13.3 - Standby Makeup Pump Water Supply Boron Concentration, and Specification 3.9.1 - RCS and Refueling Canal Boron Concentration, and Specification 3.9.12 - Spent fuel Pool Boron Concentration.)
- 8. DPC-NE-3002A, Through Rev. 2, "FSAR Chapter 15 System Transient Analysis Methodology," SER Dated April 26, 1996.
(Methodology used in the system thermal-hydraulic analyses which determine the core operating limits)
- 9. DPC-NE-3000P-A, Rev. 1, " Thermal-Hydraulic Transient Analysis i Methodology," SER Dated December 27, 1995. I i
(Modeling used in the systen thermal-hydraulic analyses) (
- 10. DPC-NE-1004A, Rev.1, " Design Methodology Using CASMG-3/ Simulate-3P," SER Dated April 26, 1996.
(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient.)
- 11. DPC-NE-2004P-A,l" Duke Power Company McGuire and Catawba Nuclear Stations Core Thermal-Hydraulic Methodology using VIPRE-01," De-MM99t (DPC Proprietary).
- E- M gb (Edf('>Ak (Methodology for Specifications 2.2.1 - Reactor Trip System J Instrumentation Setroints, 3.2.1 - Axial Flux Difference (AFD), and 3.2.3
- Nuclear Enthalpy Rise Hot Chanr.el Factor Fu (X,Y).)
- 12. DPC-NE-200lP-A, Rev.1, " Fuel Mechanical Reload Analysis Methodology for Mark-BW Fuel," October 1990 (DPC Proprietary).
h gd' r (Methodology for Specification 2.2.1 - R ,ctor Trip Sjstem Instrumentation Setpoints.) b f cp 13. DPC-NE-2005P-A,/"REV lThermO Hydraulic Statistical Core Design Method 6 Fiwi 1993 (DPC Proprietary). (Methodology for Specification 2.2.1 - Reactor Trip System Instrumentation Set, points, Specification 3.2.1 - Axial Flux Difference, and 3.2.3 - Nucleat Enthalpy Rise Hot Channel Factor) i CATAWBA - UNIT 1 6-22 Amendment No.154 l
Q-fJG-2.00$ UR)6 f [16 CHAM C4L RCLDAD lWALy$y
, Q{ CI H0boLc6f US106 BT 03, " S6il b9TEb ADMINISTRATIVE CONTROLS APRJIJ, s995 [6fC pgoeggp;M
) CORE OPERATING llMITS REPORT (Continued)
- 14. bat #10102NA, TACO 3 Tscl fin-Thefm! %1ych Compttte Cedt, O'." i= L Q Empany,- Navc;icr 19897 (Methedology used for Specification 2.2.1 - Reactor Trip System InstrumentationSetpoints)
- 15. BAW-10183P, Fuel Rod Gas Pressure Criterion, B&W Fuel Company,qJtJf 1 riey 1994.
(Used for Specification 2.2.1, Reactor Trip System Instrumentat.on Setpoints) The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, cora thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met. The COM OPERATING LIMITS REPORT, including any mid-cycle revisions or supple-ments thereto, shall be provided upon issuance, for each reload cycle, to the NRC in accordance with 10 CFR 50.4. SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the NRC in accordance with 10 CFR 50.4 within the time period specified for each report.
#_,_10 RECORD RETENTION 6.10.1 In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum oeriod indicated.
The following records shall be retained for at least 5 years:
- a. Records and logs of unit operation covering time interval at each power level;
- b. Records and logs of principal maintenance activities, inspections, repair, and replacement of principal items of equipment ralated to nuclear safety;
- c. All REPORTABLE EVENTS;
- d. Records of surveillance activities, inspections, and calibrations required by these Technical Specifications; ,
- e. Records of changes made to the procedures required by Specification 6.8.1;
- f. Records of radioactive shipments; Records of sealed source and fission detector leak tests and results;
-) 9 and CATAWBA - UNIT 1 6-23 Amendment No. 148
.y ,, . . -
t L ADMINISTRATIVF CONTROLS CDRE OPERATING LIMITS REPORT (Continued) (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-NF-2010A, " Duke Power Company McGuire Nuclear Station Catawba Nuclear Station Nuclear Physics Methodology for Reload Design," June 1985 (Methodology for Specification 3.1.1.3 - Moderator Temperacure Coefficient, Specification 4.7.13.3 - Standby Makeup Fump Water Supply Boron Concentration, and Specification 3.9.1 - RCS and Refueling Canal Boron Concentration, and Specification 3.9.12 - Spent Fuel Pool Boron Concentration.)
- 8. DPC-NE-3002A, Through Rev. 2, "FSt.R Chapter 15 System Transient Analysis Methodology," SER Dated April 26, 1996.
(Methodology used in the system thermal-hydraulic analyses which determine the core operating limits)
?E 9. DPC-NE-3000P-A. Rev.1," Thermal-Hydraulic Transient Analysis 3 Methodology," SER Dated December 27, 1995. ,
g: J 4 r (Modeling used in the system therral-hydraulic analyses)
- 10. DPC-NE-1004A, Rev. 1, " Design Methodology Using CASMO-3/ Simulate-3P," SER Dated April 26, 1996. l
(( J (Methodology for Specification 3.1.1.2 - Moderator Temperature Coefficient.) e A6V la fl
}d d
- 11. DPC-NE-2004P-A,[ Duke Power Company McGuire and Catawba Nuclear Stations i
Core Thermal-Hydraulic Methodology using VIPRE-01," g t c at:r 1991--(DPC [ Proprietary). ,56A 0 1 (Methodology for Specifications 2.2.1 - Reactor Trip System Instrumentation Setpoints, 3.2.1 - Axial Flux Difference (AFD), and 3.2.3 ND 1
- Nuclear Enthalpy Rise Hot Channel Factor Fu (X,Y).)
- 12. DPC-NE-200lP-A, Rev.1, " Fuel Mechanical Reload Analysis Methodology for 3 Mark-BW Fuel," Or.tober 1990 (DPC Proprietary).
) (Methodology for Specification 2.2.1 - Reactor Trip System nstrumentationStgits.)
13 g [ g . DPC-HE-2005P-A,I"
-febrseri M%(DPC Proprietary). Thermal Hydraulic Statistical Core Design Metho (Methodology for Specification 2.2.1 - Reactor Trip System
' Instrumentation Setpoints, Specification 3.2.1 - Axial Flux Difference, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor) j (
a I -CATAWBA - UNIT 2 6-22 Amendment No.
- 46 i f
Ck
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wwmm w -ww~mn,. 1
bpc,-HG 2% "F'vCL- MWN M Md AA>ALY3LS
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.. bM APlatif l9W (f.)pc PR.oPAM742& L\
ADMINISTRATIVE CONTROLS .
) CORE OPERATING LIMITS REPORT (Continued) y( 3 %
- 14. BAW:1Mfr2M;-TAC 43-F=1 Pir, Ther el Arielysis Ce..Wicr C de, ~&W Fue+
Ce.veny, iicher 1000. q I (Methodology used for Specification 2.2.1 - Reactor Trip System F; N V InstrumentationSetpoints) soul,tenf k
- 15. BAW-10183P, Fuel Rod Gas Pressure Criterion, B&W Fuel Company, May 100?.
h {}!
! (Used for S ik
[ Setpoints) pecification 2.2.1, Reactor Trip System Instrumentation r The core operating limits shall be decermined so that all applicable limits E {h. g (e ., fuel thennal-mechanical limits, core thermal-hydraulic limits, ECCS lim ts, nuclear limits such as shutdown margin, and transient and accident I l r p analysis limits) of the safety analysis are met. g [ s The CORE OPERATING LIMITS REPORT, including any mid-cycle revisiona or supple- kU ments thereto, shall be provided upon issuance, for each reload cycle, to the j NRC in accordance with 10 CFR 50.4. fp"t jt; SPECIAL REPORTS (. m F
' 6.9.2' S)ecia) reports shall be submitted to the NRC in accordance with 10 CFR ki i 50.4 wit 11n the time period specified for each report. M
)
6.10 RECORD RETENTION h@
} 6.10.1 In addition to the applicable record retention requirements of Title 10, } Code of Federal Regulations, the following records shall be retained for at 4 f
least the minimum period indicated. i j The following records shall be retained for at least 5 years: e
- a. Records and logs of unit operation covering time interval at each power f 9 level; d
y e
- b. Records and logs of principal maintenance activities, inspections.
l repair, and replacemmt of principal items of equipment related to 1 nuclear safety; f I' a
- c. All REPORTABLE EVENTS;
- d. Records of surveillance activities, inspections, and calibrations t required by these Technical Specifications; '
- e. Records of changes made to the procedures required by Specification 6.8.1; ;
- f. Records of radioactive shipments;
> g. Records of sealed source and fission detector leak tests and results; d and u
I k CATAWBA - UNIT 2 6-23 Amendment No. 142
)
1
- ~ f .WMWWWW@[rSTN . "
*MiR .. . . . .
Attachment I.B Amended ITS Submittal Pages - McGuire f 1 f i f I l ( I r i-f.
. p.s*
'1
l I . Meporting Requirements i 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)
- 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 cote operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
- 1. WCAP-9272-P-A, "WESTIrlGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY," July 1985 (H Proprietary).
- 2. WCAP-10216-P-A, " RELAXATION OF CONSTANT AXIAL OFFSET CONTROL FQ SURVEILLANCE TECHNICAL SPECIFICATION",
June 1983 (H Proprietary).
- 3. WCAP-10266-P-A Rev. 2, "THE 1981 VERSION OF
['Q/z,hRT6b g WESTINGHOUSE EVALUA1 ION MODEL USING BASH CODE", y March 1907, (H Proprietary). [2h8%'M 4. BAW-10168P, Rev.1, "B&W Loss-of-Coolant Accident Evaluation Mo or Recirculating Steam Generator (jdM6 6 IU4 Plants,ySER dated January 1991 (B&W Proprietary).
- 5. DP NE-20.1P-A, " Duke Power Company Nuclear Design Methodology for Core Operating Limits of Westinghouse P.eactors," March,1990 (DPC Proprietary).
- 6. DPC-NE-3001P-A, " Multidimensional Reactor Transients and Safety Analysis Physics Parameter Methodology,"
November,1991 (DPC Proprietary).
- 7. DPC-NF-2010A, " Duke Power Company McGuire Nuclear Station Catawba Nuclear Station Nuclear Physics Methodology for Reload Design," June,1985.
(continued) McGuire Unit 1 5.0-28 5/20/97
L . Reporting Requirements 5.6 1 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR)_ (continued)
- 8. DPC-NE-3002,fMROJkil Rev. "L"FSAR Chapter 15 System Transient AnalysO Methodology," SER dated n.c:ry,10;rt. 4pgit.
- 9. DPC-NE-3000P, Rev.1 " Thermal-Hydraulic Transient Analy31s Methodolo ." SER dated Decembe d
- 10. OPC-NE-1004A,4" Nuclear Design Methodology Using CASMO-3/ SIMULATE-3P," W/etr,1991. J6A M AFAst.20 l996 3
- 11. DPC-NE-2004P-A.I"Du P$werCompanyMcGuireand Catawba Nuclear Stations Core Thermal-Hydraulic Methodology using VIPRE-01," tcc;...LJ. 1074 (DPC Proprietary).
See bRTCb F4BauARY 29 R9')
- 12. DPC-NE-2001P-A, Rev.1, . " Fuel Mechanical Reload Analysis Methodology for Mark-BW fuel," October 1990 (DPC Proprietary).
13.
/W 13 DPC-NE-2005P-A,f" Thermal Hydraulic Statistical Core Design Methodology," Februay 2005 (DPC Proprietarv).
ssgspr7g A.tonnecA qj996
) 14. aAw.in162P ^., TACO Fed 94n Thermal Analysj C;7,putep
% pgu r g- u 3,, g y, % ;e3_ m
- 15. BAW-10183P, Fuel Rod Gas Pressure Criterion, B&W Fuel Company, tesy=199+. Jdt-( I495
- c. The core operating limits shall be deteiinined such that all applicable limits (e.g., fuel thermal rechanical limits, core thermul 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.
bPC -Nc-zco8,
- rust fiscHtWtLAL M& AAMWS45 M6THobolo6y us!UG TACO 3," SEld bar66 AMIL [Mgr (bPC MoPRI67ARY}
} (continued)
McGuire Unit 1 5.0-29 5/20/97
- - . . - - - . - . . - .. .- - . ~. - ... .
1 o ,- -i r-.> : Reporting Requirements ; . '5.6 i . ; ;5.6' Reporting Requirements i 5.6.5- ' a CORE OPERATING LIMITS REh0RT- (COLRl- (continued) 9. f
.-Reactor Coolant System and refueling canal boron
-- concentration limits for Specification 3.9.1, i
- 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:
- j. ,
- 1. WCAP-9272-P-A, " WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY," July 1965 (H Proprietary).
1
.2. WCAP-10216-P-A, " RELAXATION OF CONSTANT AXIAL OFFSET CONTROL FQ SURVEILLANCE TECHNICAL SPECIFICATION",
4
~ June 1983 (M Proprietary).
- 3. WCAP-10266-P-A Rev. 2, "THE 1981 VERSIGN OF
) WESTINGHOUSE EVALUATION MODEL USING BASH CODE",
[ [s March 1987,- (H Proprietary). - o QT8 4 4. .BAW-10168 M "B&W Loss-of-Coolant Accident
- Evaluation Model for Recirculating Steam Generator ,
7 - q % [ W b g7gh PlantspSERdatedJanuary199 B&W Proprietary). 3dA>6 .) 4. .DPC-NE-20)1P-A, " Duke rower Company Nuclear Design
- _ ~( Methodology for Core Operating Limits of Westinghouse Reactors " March 1990 (DPC Proprietary).
- 6. DPC-NE-3001P-A, "Mult1 dimensional Reactor Transients i
and Safety Analysis Physics Parameter Methodology," . November,1991 (DPC Proprietary).
- 7. DPC-NF-2010A, " Duke Power Company McGuire Nuclear l
Station Catawba Nuclear Station Nuclear Physics. > a_ Methodology for' Reload Dt.;tgn," June,1985. a (centinued) , -McGuire Unit 2- 5.0-28 5/20/97 z__ . . _ . _ _. . . . ~ . ,_- . __ _ _ . _ _ _ __ .~. ,
L. - Reportir.g Requirements 5.6 i 5.6 Reporting Requirements 5.6.5 CORE OPERATING 1.IMITS REPORT (COLR) (continued)
- 8. DPC-NE-3002,I-fptpJf>H Rev. 't. "FSAR Chapter 15 System 1ransient Analysis Methodology," SER dated "Econier,1;;L. Afst/c zg
- 9. DPC-NE-3000P, Rev.1 "Thennal-Hydraulic Transient Analysis Methodology," SER dated December 1995. M ACVI
- 10. DPC-NE-1004A,0N uclear Design Methodology Using CASMO-3/ SIMULATE-3P," L,suter,1;;2. 56A 6473 6 Ar%ft v.jt19f,,
- 11. DPC-NE-2004P-A,f"MV i Duke Power Company McGuire and Catawba Nuclear Stations Core Thermal-Hydraulic Methodology)using Proprietary . VIPRE-01," -D:::-ter 1991 (DPC36fL OAT
- 12. DPC-NE-2001P-A, Rev.1, " Fuel Mechanical Reload Analysis Methodology for Mark-BW fue'," October 1990 (DPCProprietary).
- 13. OPC-NE-2005P-A.I"p.gs!I Thermal Hydraulic Statistical Core Design Methodology," Fn.wiy 1995 AtoVF/1 S< bkfGb (DPC Prop ietary).&
9, IQ
) 14. BAW4016?p A. TACO 3 Fuel P!n T h a r'n> 1 AMys-i: Cc ;ute C6 Fuel Ca pany, heReer-1989.
- 15. BAW-1018P, Fuel Rod Gas Pressure Criterion, B&W Fuel Company, ty ICL dinJ/ /cg4X
- c. The core operating limits shall be determinid 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.
yg gg ogL. f1ECHA NICPL N # b A^' N M t4c-rHoaxos9 OsW6 %3a " Sga onw6 WRua mr (bPc PaonvereAV).
} (continued)
McGuire Unit 2 5.0-29 5/20/97
$$$.ntp 85S Y $ 0 h{*' ff O&h
[.- a,ea;ny( ADMINISTRATIVE CONTROLS w ssaua su CORE OPERATING LIMITS REPORT- I L The analytical methods used to deteneine the core operating limits shall be
' those previously reviewed and approved by NRC int -
-i 1. WCAP-9272-P-A,
- WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY,"
July 1985 (M _ Proprietary). f(Methocolo ror spectftcation 3.1.1.3 - Modera 5 'Tepoerature C0efficien 3 Bank Insertion Li t. 3.1.3.6 - C rol Bank Inse Ion.1.3.5-Shut Limits, 3.2.1 - Axial Flux Diffe e, 3.2.2 - He flux ' i ot Ch H 1 Factor, and 3. - Nuclear Enthalp Rise Hot Chann Factor))
- 2. WCAP-10216-P-A, ' RELAXATION OF CONSTANT AXIAL OFFSET CONTROL FQ SURVEILUWICE TECHNICAL SPECIFICATION *, June 1983 (M Proprietary).
~
t fiftet logy for Speciff tions 3.2.1 - 1 Flux Difference /(RelM Axial ffset Control)-a 3.2.2 - Heat F x Hot Channel Facpr (W(Z) _surv lance requir s fi,r Fn Method oov.1 y _J 3. WCAP-10266-P-A Rev. 2 'THE 1981 VERSION OF WESTIIIGIOJSE EVALUATION IN00EL USING BASH CODE *, March 1987 (M Proprietary). , j (( Method /logyforSagIcification3.2.j/-HeatFluxHotCh/nelFactor.D
- 4. BAW-10168P,fR W. I1 'A&W Loss-of-Coolant Accident Evaluation ppdal far i Recirculating steam Generator Plants,75ER dated January 1991 '(ban y'L Proprietary) j
~
(Ethodsiaa= rne saa'1fication 3.2.7 - Heat Flux ljpt Channel Factor.)
- 5. DPC-IIE-2011PA, " Duke P2wer Company Kuclear Design Methodology for Core Operating Limits of Westinghouse Reactors.' March 1990 (DPC Proprietary).
f(Metr-"ogyforspecificati 2.2.1 - Reactor Tr SystemInstrumphtation) Setpo71s. 3.1.3.5 - Shutd Rod Insertion Limi ,.3.1.3.6 - Con 1 Bank'- Insert /o Limits, 3.2.1 - .ial Flux Differenc 3.2.2 - Heat F1 x Hot - t1 Factor and 3.2.3 Nuclear Enthalpy se Hot Channel ctor. b 6. DPC-NE-3001PA, Physics Parameter Methodology," ' Multidimensional November 1991 Reactor Transients (DPC Proprietary . and Safety) Analy
~
FIMethoeol for Specification .1.T.3 - Moderator Temperat Coeffi cient,i 3 .3.5 - Shutdown Rod nsertion Limits, 3.1.3.6 - ntrolBan]k Insert Limits, 3.2.1 - 1 Flux Difference, 3.2.2 - at Flux Hot Chanc Factor, and 3.2.3 , leuclear Entkainv n h a Hot , nnel Factar.l__
- 7. - DPC-NE-2010A, ' Duke Power Company McGuire Nuclear Station Catawba Nuclear Station leuclear Physics Methodology for Reload Design," June 1985 iMetneasiogy r specification .1.3 - Moderator T ature coefficien Specification 3. - RCS and Refueling nal Baron Loncent e , .i <on.-and on 3/4.9.12 - Spent F i Pool-Boron q Specif _
McGUIRE - UNIT 1 6-21 Amendment No. 166 5 NCtJ156 kN Netf M6$h
- w. .w
I-
- Eft $$iM b 5% i C.6 ( -
ADMINISTRATIVE CONTR0tS ~ jr 8. CORE OPERATlhC LIMITS REPORT (Continued) M/$df DPC-NE-3002, 'FSAt ru nt.c. E syst h, g,)
~ --
z'" tis
/. Methodology, ated(Iposeer M5JM// 24)ff[] ['
Me ed ystem ther d -hydraulic ana M ich determine]
. 9. 'Th 1-Hydraulic Transtant Analysis Methodology,"
DPC-NE-3000P,Rev./ SER _ Cated December
~
((Modeling used in the systok thermal-hydrau)4(analysesh ES hdf Q ""q r :.mi
- 10. DPC-NE-1006A #*h sign Methodology Using CASMD-3/ SIMULATE-3P,'
'p
// ph gfh i" don 3.1.1.3 - Mr TempMe
'(Methodefegy for S Coefficient 4 -- p, /, ,
- 11. DPC-NE-2004P-A i,. 7 - . Company McGuire and Catawha Nuclear Stations Core Thermal-Hydraulic Methodology using VIPRE-Ol_
.$(4 4I Qqg gf1
\ Proprietary).
f(Methodolony f pecifications 2.2.1 - ctor Trip System (DPC Jtion Setpo trumenta '
. 3.2.1 - Axial Flux erence (AFD), and . 3 - Nucle
[Enthalp M se Hot Channel Fact (X,Y).) -
- 12. OPC-NE-200lP-A, Rev.1, " Fuel Mechanical Reined Analysis Methodology for' Mark-BW fuel," October 1990 (DPC Proprietary).
A.37 ggs,ma 3,,cte._ , , , . ..g. 4.t4s2 v./,)
,,, s.,t , insi,,,,1,1,ti o,7 (QMk
- 13. M 005P-(i @ l H B (DPC Proprietary).
for Specificattor. 2.
.voi..oc Statistical Core Design Methodology,'
Ih TMethodel - Reactor Trip Syst ( [#MMb hf k nstrumentat n I Set i pecification 3.2.1 1al Flux Differene and 3.2.3 Nuc ear Enthalpy Rise Hot nel Futari- ~ * " i
- 14. 2P-A alysis C Cod Fue % Thersa
, W f\.. .--
Me u cation 2.2 actor Trip Sypeer Instru-) S da x0Y4 f pecificat p I17 . m . .. % stem Instrum peon] The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-sechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and acetdent analysis limits) of the safety analysis are met. The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supple-meats thereto, shall be provided upon ussuance, for eaG reload cycle, _to the I with coAlef to the Reg 1 swat mainisyrmor and
//]stit IO Q g InStrtMcGUIRE '//&- UNIT 1 6-22 Amendment No.175
%pc-we-ame, %et nedenial Rehw/Aiw/pis Mehhbf y,j x,
giy TAC 03,? seRdefed api /4iHSOKpm&&'7J/ pq 6io- W Ylicense AMnd/hMS Reguest
Sab6 c;. 6. f ADMlHISTRATIVE CONTROLS [Q R OPERATING LIMITS REPORT .' The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by HRC in:
- 1. WCAP-9272-P-A. ' WESTINGHOUSE RELOAD SAFETY EVALUATION SETH000 LOGY,"
July 1985 (W Proprietary). , f(Methodologyf Specffications .1.1.3 - Moderator erature Coefficient, .1.3.5 - Shutd Bank Insertion Limi , 3.1.3.6 - Co rol Bank Insert' n Limits, 3.2.1 Axial Flux Differe e, 3.2.2 - les Flux gtChann Factor, and 3.7 - Nuclear Enthalpy ise Hot Channe Factor.)
- 2. WCAP-10216-P-A 'RELAXATI0t: OF CONSTANT AXIAL OFFSET CONTROL FQ SURVEILLANCE TECHNICAL SPECIFICATION" June 1983 (M Proprietary).
(Method ogy for Speciff Lions 3.2.1 - Axial lux Differene (RelaxM Axial fsetContrel) d 3.2.2 - Heat Flux ot Channel Fa or (H(Z) . I surv llance requir ts for Fo Methodel y.) > h 3. WCAP-10266-P-A Rev. '2, 'THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE", March 1987 (M Proprietary). .. QMethodologpfor Specificatipd 3.2.2 - Heat Flupi{ot Channel [ac
- 4. bah-10168P,diew. 4'm loss-of-Coolant Accident Evaluation Moce; for I Recirculating neam Generator Flants,'ASER dated January 1991*lB&W Proprietary). #4 h[.
[ @ethodology/ror specification /2.2 - Heat bux Ho/ Channel Factor.D
/q 5. DPC-NE-2011PA, " Duke Power Company Nuclear Design Methodology for Core h.3 Operating Limits of Westinghouse Reactors," March 1990 (DPC Prontietary).
! (Mett.odology Mr Specification .2.1 - Reactor T ip System Inst ' ntatio d JSetpoints, 3 . 3.5 - Shutdown Insertion Lim ts, 3.1.3.6 - ntrol Bank lInsertion L its, 3.2.1 - Axi Fm Differenc , 3.2.2 - Htat lux Hot )
(ChannelF4 or, and 3.2.3 - clear Enthalpy } se Hot Channe f Factor.) _)
- 6. OPC-NE-3001PA, " Multidimensional Reactor Transients and Safety Analysis Physics Parameter Methodology." November 1991 (DPC Proprietary).
f f[iietho:lology fo Specification 3.1 3 - Moderator T erature effi-cient, 3.1.3.5 Shutdown Rod ins ion Limits, 3.1. 6 - Cont 1 Bank]
. Insertion Lie s. 3.2.1 - Axial x Olfference, 3. 2 - Heat lux Hot Thannel Fact , and 3.2.3 - Nuc r Enthalpy Rise ot Channe Factor.l>i
- 7. DPC-NE-2010A, " Duke Power Company McGuire Nuclear Station Catawba Nuclear Station Nuclear Physics Methodology for Reload Design,' June 1985 f(Methodology or Specificati 3.1.1.3 - Mod retor Temperatu Coefficient Specification .9.1 - RCS and efueling Canal n .
Concentratd n, and Specift ation 3/4.9.12 Spent Fuel Poo oron i
@centrafon.) _ _
(Rev. .% SER defed Awytsh % M6;M5 SER dtk6Mtdl lmb McGUIRE - UNIT 2 ' 6-21 Amendment No. 148
$l0.04)5& & N
_ t ( (, o e r LA
f iv.' 5 .:.t i L E 4. f l ADMINISTRATIVE CONTROLS ,__ C N CORE OPERATING TIN
- 8. DPC-NE-3002',
TS REPORT (Continued)Drp / h,4 ~ TSAR rh*atar 15 Sy sisj
. Methodology,( s'!edcues*%sF)M O_ p )/J4f /d f clog u ystp4hermal'y I... c ana' pts whichjet1l imine]
- 9. DPC-NE-3000P, Rev. m 'The Hydraulle Transient Analysis Methodology,*/S '
SER edDecembergyn.12
'( ine used in Dr. avstem tha=dil-hwdranMianalM * //>
p 40 E-gn Methodology Using CASMO-3/ SIMULATE-3P "
! / k Netho gogy for ton 3.1.1.3Aerater Tese6aturd
.CpefKctent.) -
- 11. DPC-NE-2004P-A CompanyMcduireandCat are tations raulic Methodology using VIPRE-0g r (DPC
(/*drygpg g .TMeth
- tion for speciff points, 3.2.1 tons 2.2.1 -
Axial Flux Dif tor Trip System nce (AFD), tnd . 3 - Nuclear; strumenta-) Jn Ipy Rise Hot nel Factor FAH ).) d87 12. DPC-NE-200lP-A, Rev.1 " Fuel Mechanical Reload Analysis Methodology for ' Mark-W fuel," October 1990 (DPC Proprietary). for ton z.Z.1 -Jetctor Trip Systes4nstrumentatiof e
- 13. DPMY .Wulic'StatisticalCortDesignMethodology,'"
JM (glirtrapy"lAFJMDPC Proprietary). Ia
/& 4 (I OMfMI I// er (Methodo Setpo s, Juc ar Enthaley Rise y for Specificat Specification
.2.1 - Reactor T
.1 - Axial Flux ch=aami raetor) -
System In erence,and_st/rntation u .3 - h'[' aba ' f f(Meth ogy used for S
. men
. ion setpoints). /pteffication 2.2 % Reactor Trin p == fnstru-J
)
- 15. BAW-10183P, Fuel DM c2e Pressu*ftriterion KFusl Company, as approved by SER dated (FgbimrfrN.)y/g fyQJfJ j e Specification
.1, 1 ..J ;)7 % fes Instrue g atto g The core operating limits shall be determined so that all applicable limits (e.g., fuel themal-sechanical 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 RZPORT, including any mid-cycle revisions or supple-ments thereto, shall be provided upon issuance, for each reload _ cycle,_to the {3 {s* Pius 53 ens newspo -imin *5crjad
//WJC top 4dl - UNIT 2 6-22 n a .., um 157 ws-ace,%c/ MeJaiaIbludbo/pn Mhdoby
( [was Tacos " .sena a' ted 4ed 5 /Ws(Mr Md"r] M l.acense AmendiwtRy<nt ,,, s, ,: ; y
[' . Reporting Requirements 5.6 5.6 Reporting Requirements (continued) M 5.6.4 Monthly Ooeratino Reoorts ONLY Routine reports of operating statistics and shutdown experience , including documentation of all challenges to the pressurife I power operated relief valves or pressurizer 56fety valves. F jr shall be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report. 5.6.5 CORE OPERATING LIMITS REPORT (CO' R)
- a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining rtion of a reload cycle, and shall be documented in the CO R for the following:
(IM 8 ndiv 'el specific st be referen sthataddressloreoperating
~
yisits nere.
/ _
- b. The analytical methods used to determine the core opera' ting 11 Lits shall be those prevloesly reviewed and approved by the NRC. specifically those described in the following documents:
n -, Identif/theTopicalReport(s) y number, title, da 'e. and
<IU9b l NRC sttff approval document.
Eval tion Report for a pla specific methodol identify the staf Safety by NRC [ t let r and date.
- 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 SDH. 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.
6
% / /leactorCoolant/vstem(RCS)FTISShEANDTEMPERATbRELIMk REMRTDi TLn / /
(continued) WOffTS 5.0 20 Rev 1. 04/07/95 Ncc e i L_ :........... - - -
. _ _ _ ._ _ _ _ _ _ _ _ _ - - . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ______m
. - - - .. . - ~ .. - . -. . - - . - . - - - - . - - . _ ~ - - - - _ - _ ..
ftV.h)E 4Y AGU3bAA j kh) $tb.3,3fRldk l
- JMnt /S l$$( i INSERT 9
( l.L 3 WCAP-9272-P-A, " WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY," July 1985 M Proprietary). t ! 2. ' WCAP-10216-P-A, " RELAXATION OF CONSTANT AXIAL OFFSET CONTROL i FQ SURVEILLANCE TECHNICAL SPECIFICATION", June 1983 M Proprietary).
- 3. WCAP-10266-P-A Rev. 2, "THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE", March 1987, M Proprietary). -
- 4.- BAW-10168P@ev 1)B&W Loss-o tant Accident Evaluation Model for Recirculating Steam Generator Plants, 'ER dated January 1991; &W Proprietary). ,
- 5. ' DPC-NE-201 IP-A, " Duke Power Company Nuclear Design Methodology for Core -
Operating Eimits of Westinghouse Reactors," March,1990 (DPC Proprietary).
- 6. DPC-NE-300lP-A, " Multidimensional Reactor Transients and Safety Analysis Physics Parameter Methodology," November,1991 (DPC Proprietary).
- 7. DPC-NF-2010A, " Duke Power Company McGuire Nuclear Station Catawba Nuclear
- Station Nuclear PhysicsEethodology for Reload Design," June,1985.
Thnpugh M;Q ! 8. DPC-NE-3002 SERant* s w
> "FSAR Chapter Inystem Transient Analysis Methodology,"
Af rifM,/ffh 4 h 9. DPC-NE-3000P, Rev.1, "'Ihermal-Hydraulic Transient Analysis Method t 9, 4
- 10. Dm-N t:-1004A. " Nuclear Design Methodology Usina CASMO-3/ SIMULATE-3P "
} ? ^ " "' " M g M R d e f ttf ,4 f t / S j , / M h E*V*I) 11. .DPC-NE-2004P-Ah" Duke Power Company McGuire and Catawba Nuclear Stations Core Thermal-Hydraulic Methodolorv usine VIPRF-01."dD32-6PC Proprietary). jk (SGR gfq M pejynggy,y go,jg f
- 12. DPC-NE-2001P-A, Rev.1. " Fuel Mechanical Reload Analysis Methodology for Mark-BW fuel," October 1990 (DPC Proprietary).
? e d. ),
i
- 13. DPC-NE-2005P-A, ermur ydrnnlic c'-'" 1 eaa Nion u,rknantogy " ;,
Warwt99RPCProprietary).@fMitYt,/ AfoM6en[er 9)M9D , T li
- 14. f% . - COMI PidrmadalvMa"'#N
%n l
1
,oec.-NE-3DDF, "Fnei Medteteni krAMd Andby5M
@e%eMoey hssen TMet."SERdetenfARW4 Mtf(DM 1hyv'ietot}h $ i
- 15. - BAW-10183P, Fuel kod Gas Pressure Criterion, B&W Fuel Company,24 D^,4.
, ' INSERT Page 5.0-20 v
f license. Amebuf egh
. Attachment I.C Amended ITS Submittal Pages-- Cat'awba i
4 f a E
- l l
t e i i 1 1 I I
]
er a w + ~ , y .M - g - gr - m--
l-Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)
- 4. BAW-10168P, "B&W Loss-of-Coolant Accident Evaluation Model for Recirculating Steam Generator Plants,"
Rev.1, SER dated January 1991; Rev. 2 SER Dated August 22, 1996; Rev. 3, SER Dated June 15, 1994 (B&W Proprietary).
- 5. DPC-NE-2011P-A, " Duke Power Company Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors," March,1990 (DPC Proprietary).
- 6. DPC-NE-3001M, " Multidimensional Reactor Transients and Safety Analysis Physics Parameter Methodology,"
November,1991 (DPC Proprietary).
- 7. DPC-NF-2010A, " Duke Power Company McGuire Nuclear Station Catawba }.uclear Station Nuclear Phys ~es Methodology for Reload Design," June,1985.
- 8. DPC-NE-3002A, Through Rev. 2 "FSAR Chaptt- 15 System Transient Analysis Methodology," SER dated April 26, 1996.
I
- 9. DPC-NE-3000P-A, Rev. 1 " Thermal-Hydraulic Transient Analysis Methodology," SER Dated December 27, 1995.
- 10. DPC-NE-1004A, Rev.1, " Design Methodology Using CASM0-3/ SIMULATE-3P " SER Dated April 26, 1996.
11. 86Vl DPC-NE-2004P-A I" Duke Power Company McGuire and Catawba Nuclear Stations Core Thermal-Hydraulic Methcdology using VIPRE-01," Deccd ,1991 (DPC Proprieta ry) . .sGR bRTGb faagy g lo lff')
- 12. DPC-NE-2001P-A, Rev.1, " Fuel Mechanical Reload '
Analysis Methodology for Mark-BW fuel," October 1990 (DPCProprietary).
/ k&L/ L
- 13. DPC-t!E-2005P-A,V" Thermal flydraulic Statistical Core 993 DPC Proprietary).
Design Methodology," ;5 LR TdraeNf_h s(foygf3ggf r) 1996
- 14. BAW-10102P A, TAC 03 TLd Mr The-el halvsu e g eter
- 2, g 5 IlClOf0 bNAIN5lb M6Tilobg'05Y c)31 AIG TAcpx
- SFA syf pyopid, Wk Ppfff$tNued)
Catawba Unit 1 5.0-29 5/20/97
Reporting Requirements 5.6
\ 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)
- 15. BAW-10183P, Fuel Rod Gas Pressure Criterion, B&W Fuel
^
Company, .
- c. The core operating limits shall be detemined stch that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core taoling 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 Reoort When a report is required by LC0 3.6.10, " Annulus Ventilation System (AVS) " LC0 3.7.10. " Control Room Area Ventilation System (CRAVS)," LC0 3.7.12, Auxiliary Building Filtered Ventilation Exhaust System (ABFVES)," LC0 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 Reoort 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 aliernate 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 Insoection Reoort .
- a. The numoer of tubes plugged in each steam generator shall be reported to the NRC within 15 days following completion of the program; -
(continuedi
)
Catawba Unit 1 5.0-30 5/20/97 1 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ . - _ _ _ - _ _ _ _. ______s
l Reporting Requirements 5.6
) 5.6 Reporting Requiremer.ts 5.5.5 CORE OPERATING LIMITS REPORT (COLR) (con +.inued) -
- 4. BAW-10168P, "B&W Loss-of-Coolant Accident Evaluation Model for Recirculating Steam Generator Plants,"
Rev.1 SER dated January 1991; Rev. 2, SER Dated August 22, 1995; Rev. 3, SER Dated June 15,1994(B&W Proprietary) . -
- 5. DPC-NE-2011P-A, ' Duke Power Company Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors," March,1990 (DPC Proprietary).
- 6. DPC-NE-3001P-/. " Multidimensional Reactor Transients and Safety Analysis Physics Parameter Methodology,"
November,1991 (DPC Proprietary).
- 7. DPC-NF-2010A, " Duke Power Ctapany McGuire Nuclear Station Catawba Nuclear Station Nuclear Physics Methodology for Reload Design," June, 1985.
- b. DPC-NE-3002A, Through Rev. 2 "FSAR Chapter 15 System Transient Analysis Methodology," SER dated April 26, 1996.
) 9. DPC-NE-3000P-A, Rev.1 "Themal-Hydraulic Transient Analysis Methodology," SER Dated December 27, 1995.
- 10. DPC-NE-1004A, Rev. 1, " Design Methodology using CASM0-3/ SIMULATE-3P,"SERDatedApril 26, 1996.
- 11. DPC-NE-2004P-A,I"RE Duke PowerV Company I McGuire and Catawba Nuclear Stations Core Themal-Hydraulic Methodology)using Proprietary . VIPRE-01." 1001Ou (DPCSER DAT4b f'E a
- 12. DPC-NE-2001P-A, Rev.1, " Fuel Mechanical Reload Analysis Methodology for Mark-BW fuel," October 1990 (DPCProprietary)
- 13. DPC-NE-2005P-A.I"NCV i Core Thermal Hydraulic Statistical 1^
Design Methodology," F:k;.$ocb;"i,(DPC Sen Proprietary). NDLKA186e Q gyg
- 14. BAW40MEP A. TACO 3 Fel-Yln Th:; 1.^raly:S Cui.pder
% -u rimi c - _ uma m ,aan 2bD $ k&cHNuIc $#W WV
~
b hgTl4bbOWf WIN A @c3,u'Sgg o bgff gg;qIN (AR 44d41 (continued)
.)
Catawba Unit 2 5.0-29 5/20/97
Reporting Requirements 5.6 i 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)
- 15. BAW-10183P, Fuel Rod Gas Pressure Criterion B&W Fuel Company, ;, I
- c. li The core operating applicable limits (e.mits g., fuel shall thermal be mechanical detemined limits, such that all core themal h Systems (ECCS)ydraulic limits, nuclear limits, limits Emergency such as Core SDM,Cooling 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 z- 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)," LC0 3.7.12. Auxiliary Building Filtered Ventilation Exhaust System (ABFVES)," LCO 3.T.13. "Fual 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 ReDort 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. 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; ,
(continued)
)
Catawba Unit 2 5.0-30 5/20/97 i
3(. t'Ct iC3 bien $. (o.5" Af*INISTRATIVF CONTk0tS ,_ CORE OPERATING 11MITS REPORT (Continued) _ [(Method / logy for SpecAfication 3.1 .3 - Mocerat Temperatur Coef
- J cient, .1.3.5 - Sh down Rod Ins ion limits, .1.3.6 - Co rol Bank 3 Inser ton Limits. .2.1 - Axial ux Differenc , 3.2.2 - He F1.sx Hot
. Qha 1 Factor, a 3.2.3 - Nuc ear Enthalpy 1se Hot Chan 1 Factor.))
- 7. DPC-WF-2010A, " Duke Power Company McGuire Nuclear Station Catawba Nuclear Station Nuclear Physics Methodology for Reload Design,' June 1985 iMeth logy for S ification .1.1.3 - ator Temper ure 7 Coeffi lent, Spect 1 cation 4. 13.3 - Sitnd Makeup P ater Supply Boron oncentrati n, and S fication 3. - RCS and fceling Canal ,
Boro Concentrat n, and S ification 3 .12 - Spent elPoolBoronj Con, ntration.) _.
- 8. DPC-NE-3002A, Through Rev. 2, 'FSAR Chapter 15 System Transient Analysis Methodology," SER Dated April 26, 1996.
hodolog 'sedinthe/ystemthe -hydraulig[ analyses whi_c0 - qtermine t core opernino limi%
- 9. DPC-NE-3000P-A, Rev.1, " Thermal-Hydraulic Transient Analysis Methodology," SER Dated Dec*r 27, 1795. l
((Modeyng used in the systepithermal-hydraulic /nalysesh
- 10. DPC-NE-1004A Rev.1 " Design Methodology Usik CASMD-3/ Simulate-3P,' SER g7 Dated April 26, 1996. l
-(Sethodolog for Specification .1.1.3 - Moderat Tamper Goefficien oe .)
( h, /
- 11. Core DPC-NE-2004P-A)L*
Thermal-Hydraulic Duke Proprietary). M&hodoloovPowerudna Company vincr_01 McGuire V
*e^p and Catawba Nuclear Station DPC l
($gg gyp) p&fm /
' ' ' '/' W (Methodolqgy for Specific t ons L g. a n..a si ip m-Instrumes ation Setpoint 3.2.1 - Axial F1 ifference ( , and 3.2.3
- Nuclea/ Enthalpy Rise t Channel Factor F, (X.Y).) _
- 12. DPC-NE-200lP-A, Rev.1 " Fuel Mechanical Reload Analysis Methodology for Mark-BW Fuel,' Dctober 1990 (DPC Proprietary).
Method cation 2.2.1 - actor Trip Sysf (nstrup/logyforSpeci entation Setpot ts.) McF,h 13. DPC-NE-wo-AA' Thermal Hydrauli 4 . ." -
- g, Q W 9 h JDPC Procrietarv) ER ?"$; '5m"_=h
'/M ology,"Q)i (Methodo ogy for 5 ification 2.2.1 - Reactor Trip stem Inst tation Se ints, Specific ion 3.2.1 - Ax Flux Difference, (and 3. 3 - Nuclea Enthalpy Rise t Channel Facto l
I' CATAWBA - UNIT 1 6-22 Amendment No.154 N N lllf D q G C .* $b
- oNf*e900 W W i -
wyrg o's.sakref an
)S " f l/pt/.x
- r.1 - s H l & Mer fr! Y t'm NPnpie%I fh0M h N f N b. %-f Y CORE OPERATING LIMITS REPORT (Continued)
~
14 A 2 al An terDdeT-1Speg)7 h O @n",[8tE'!e%@f"""'" 2.2.1 - Rpct w Syg
- 13. BAW-10183P, Fuel Rod Gas Pressure Criterion B&W Fuel Company, b~
4 fJ f ecification 2. , Reactor Trip Sptem Instrumenpti . The core operating limits shall be determined so that all applicable limits (e.g., fuel thennal-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. M The CORE OPERATING LINITS REPORT, including any mid-cycle revisions or supple-INWtT top ments thereto, shall be provided upon issuance, for each reload cycle, to the NRC in accordance with 10 CFR 50.4. TaytuALREPbRTS MO . ecial reports sha be submitted to e NRC in accordar.ce /th 10 CFR 50.4 it in the time peri (6.9.2 specified for e report. / f5$10 RECORD RETENTt0h 6.10.1 In additt to the applicable reco retention requirements of Title A Code least the of Federal egulations, min um period indicated. the followin records shall be retair.ed for a[ , (- he' folio ng records shall be reta ed for at least 5 years: '
. Records and logs of un operation covering time interva at each power level;
- b. Records and logs f principal maintenance activitie , inspections, repair, and rep cement of principal ites.s of equi . nt related to nuclear safet ,
C4.19
- c. All REPOR BLE EVENTS;
- d. Recor of surveillance activities, inspe ions, and calibrations req ed by these Technical Specificati -s;
- e. cords of changes made to the proce res required by Specification 6.8.1; !
/ ,
/
- f. Records of radioactive shipmen ; /
- g. Records of sealed source a fission detector leak tests and suits;
( and i CATAWBA - UNIT 1 6-23 Amendment No. 148 l0 W N N & O t pay TV O Sh
. hh*/ Ibe Nhiu- 5D ADMINISTRATIVF CONTROLS
( COREOPERATING'.tMITSREPORT(Continu4 l for Spect atton 3.1.1.3 - rator Temperat Coefft
- [c(Methodel ient, 3 .3.5 - Shut Rod inserti inits, 3.1.3.6 - trol Bank l
Inserti Limits, 3.2 - Axial Flux fference, 3.2.2 - at Flux Hot Qhanne Factor, and .3 - Muclear thalpy Rise Hot anel Factor.)
- 7. DPC-NF-2010A, " Duke Power Company McGuire Nuclear Station Catawba Nuclear Station Nuclear Physics Methodology for Reload Design," June 1985 (Met ology for Spec leation 3.1.1.3 - rator Temperdre Coeff cient, specif t ation 4.7.13.3 - andby Makeup dater Supply Bo Concentration and Specificati 3.9.1 - RCS and efueling Canal Bo Concentrati , and Specificatt' 3.9.12 - Spent uel Pool Boron C centration.)
- 8. DPC-NE-3002A, Through Rev. 2 'FSAR Chapter 15 System Transtent Analysis' Methodology,' SER Dated April 26, 1996. .
l (Methodo1 used in/he system theM-hydraulic analy[es which Cdeterut thecorepratinglimits)/ /
- 9. DPC-ME-3000P-A, Rev.1,' Thermal-Hydraulic Transient Analysts c Methodology,' SER Dated December 27, 1995.
MN (MEle11[used in th[ system thermal-h raulic analy
- 10. DPC-ME-1004A, Rev.1, ' Design Methodology Ustag CASMD-3/Steulate-3P
- SER DatedAprt] 26, 1996; l for Specif tion 3.1.1.3-peratorTyperatur)e d
-( kfYe h} H. DPC-E-2004P-A 'hr Compuy McGuire and Cathm Statius Core Thermal raultr Methodoloav usina VIPRE-01.*/U ~ MD(DPC f
4 Proprietary). {ggg,fpfgg fe yel3; fppyy i Methodo1 for Specift tons z.z.I - Re tor trip sy W Inst ation Setpoin 3.2.1 Axial um Otfference AFD), and 3.2.3
- Nucle Enthalpy Rise t Channel Fact F (X,Y).)
- 12. DPC-ME-2001P-A, Rev.1. " Fuel Mecharitcal Reload Analysis Methodology for Mark-BW Fuel,* October 1990 (DPC Prcprietary),
h for f ton 2.2.1 - Re tor Trip $[t
- 13. opc.ar ?nasp-A3' Thermal Hydraulk ":1 m *Ij Cg _ ArynIDPC Proprietary 1f.fgA g Mm 6kWrd1. %/ 9 j/
(Methodel "for Specif cation 2.2.1 - ctor Trth System Inst ation Setpot ts, Specificatt .2.1 - Axial F1 Difference, and 3.2 - Nuclear thalpy Rise Hot annel Factor) CATAWBA - UNIT 2 6-22 Amendment No.146 s
% License Anrew/med Qued paje Va
'Y' Yr NhA/ fe9hf h . esm:maonasa/$ fg'fn ,,.m p,, -
ADMiKISTRATIVE CONTR0tS ^ CORE OPERATING 8.!MITS REPORT (Continued) 2 A C0 lysis er The p lon in ' Oa/yi AGS c lb. BAW-10183P, fuel Rod Gas Pressure Criterion, B&W Fuel L.v=pany, gay 199M f, r Sp icatior 2.2.1, eactor Trip tem instrume atio - The core operating limits shall be determined so that all applicable limits (e.g., fuel themal-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. h b # gop! The CORE OFERATING LIMITS REPORT, includin? any mid-cycle revisions or supple-ments thereto, shall be per>vided upon issuance, for each reload cycle, to the NRC in accordance with 10 CFR 50.4. h#I D g*7D SPECIAL dPORTS 6.9. Special reports 11 be su i ted to the NRC accordance w h 10 CFR
- 50. ithin the time iod specif d for each repo .
; 6.10 RECORDRETFhTION ,
6.10.1 In ad tion to the applicable re rd retention requir ts of Title 10, Code of Fed al Regulations, the foll ng records snall be re ained for at least the nimum period indicated. The fo owing records shall be re ned for at least 5 yea :
- a. Records and logs of un' operation covering ti level; interval at each power f
- b. Records and logs principal maintenance a ivities, inspections, repair, and repl cement of principal items f equipment related to
,(p nuclear safety
- c. Als REPORT E EVENTS;
- d. Records f surveillance activities, spections, and calibration '
requi d by these Technical Specif ations; i
- e. Rec s of changes made to the cedures required by Speciff ation 6 .1; i I
- f. Records of radioactive shi nts; Recoros and of sealed source nd fission detector leak tes and results; 4
CATAWBA - UM1T 2 6-23 Amendment ho. 142 u se n a u. . 1
s Reporting Requirements. 5.6 5.6 Reporting Requirements (continued) I D 5.6.4 Honthly Doeratina Reoorts Routine reports of operating statistics and shutdown experience including documentation of all challenges to the pressurizer l power operated relief valves or pressurizer safety valves. - shall be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report. 5.6.5 CORE ODERATING LIMITS REPORT (COLR)
- a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:
h-
/INSEC8 The indi ual specificat sthataddressloreoperating -
limit st be refere here.
/
- b. The analytical methods used to determine the core opera'ing t limits shall be those previously reviewed and approved by
' the NRC. specifically those described in the following documents
< bD 9b
~~
ntif the Topical Report (s) y number, title, d 'e. and [- NRC st ff approval document. identify the staf Safety Eval tion Report for a pla specific methodol by NRC jh' let er and date. t c. b 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 SOH. transient anal sis limits, and accident analysis limits) of the safety anal sis 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 Reactd Coolant System I S) PRESSURE AND T[MPERATURE Q MIS EP9RT (PTIR)
/
- (continueo)
WOG STS' 5.0 20 Rev 1. 04/07/95 CJhk C)
~
i' . L INSERT 9 i
- i. WCAP-9272-P-A, " WESTINGHOUSE RELOAD SAFETY EVALUATION
-{- METHODOLOGY," July 1985 M Proprietary).
- 2. WLAP-10216-P-A, " RELAXATION OF CONSTANT AXIAL OFFSET CONTROL-FQ SUFVRTI. LANCE TECHNICAL SPECIFICATION", June 1983 M Proprietary).
- 3. WCAP-10266-P-A Rev. 2, "THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE", March 1987, M Proprietary). '
4. BAW-10168P, "B&W Loss-of-Coolant Accident Evaluation Model for Recirculating Steam Generator Plants," Rev.1, SER dated January 1991; Rev. 2, SER Dated August 22,1996; Rev. 3, SER Dated June 15,1994 (B&W Proprietary) ' 5. DPC-NE-201 IP-A, " Duke Power Company Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors," March,1990 (DPC Proprietary).
- 6. '
DPC-NE-300lP-A, " Multidimensional Reactor Transients and Safety Analysis Physics Parameter Methodology," November,1991 (DPC Proprietary).
- 7. DPC-NF-2010A, " Duke Fower Company McGuire Nuclear Station Catawba Nuclear Station Nuclear Physics Methodology for Reload Design," June,1985.
8. DPC-NE-3002A, Through Rev. 2 "FSAR Chapter 15 System Transient Analysis 1
' Methodology," SER dated April 26,1996.
9. DPC-NE-3000P-A, Rev.1 " Thermal-Hydraulic Transient Analysis Methodology," SER Dated December 27,1995. 10. DPC-NE-1004A, Rev.1, " Design Methodology Using CASH 7-3/ SIMULATE-3r," SER Dated Apw 26,1996. D' a. DPC-NE " Duke Power Company McGuire and Catawba Nuclear Stations Core Therma!-Hydraulic Methodology using VIPRE-Olyd2peernberTMPOI)eC Proprietary). Q R defgg gekupyap,jgq[ 12. DPC-NE-200lP-A, Rev.1, " Fuel Mechanical Reload Analysis Methodology for Mark-BW fuel," October 1990 (DPC Proprietary). Rev.ls) 13.
@,@P-A,("Thermil DPC-NE-2005 Hydraulic Statistical Core D LDPCProprietary)f.5(Jt 4fM Ahve4er 7,
- 14. BA 0162P- TACO 3 el N dnn1 AnWEs Comadfer COM thWt# I l nv. her of DPC.-NE-2009, " fuel M&h(MA)Mh (Ms%dehf ( Msky TACOR"3 erd &Af /4MfSTON hsfI'lshM 15.
BAW-lutar, ruci koo uas Pressure Untenon, umw ruel company (N Catawba Unit 1 INSERT Page 5.0-20
+ocene kwedrest Ryd
Attachment II New Originals A. New Original Pages - Current TS
- 1. McGuire Unit 1
- 2. McGuire Unit 2
- 3. Catawba Unit 1
- 4. Catawba Unit 2 B. New Original Pages - Improved Standard TS
- 1. McGuire Unit 1
- 2. McGuire Unit 2
- 3. Catawba Unit 1
- 4. Catawba Unit 2
Attachment II.A' e New Original Pages - Current TS A i
1 l
. l 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 (W Proprietary). (Methodology for Specifications 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limit, 3.1.3.6 - Control Bank Insert ion 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 (W 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 Fg Methodology.)
- 3. WCAP-10266-P-A Rev. 2, "THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE", March 1987 (W Proprietary).
(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)
- 4. BAW-10168P, "B&W Loss-of-Coolant Accident Evaluation Model for Recirculating Steam Generator Plants," Rev.1, SER dated January 1991; Re/. 2, SER dated August 22, 1996; Rev. 3, SER dated June 15, 1994. (B&W Proprietary) .
(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)
- 5. DPC-NF-2011PA, " Duke Power Company Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors," March 1990 (DPC Proprietary).
(Methodology for Specification 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-3001PA, " Multidimensional Reactor Transients and Safety Analysis Physics Parameter Methodology," November 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 Fac' r, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor.)
- 7. DPC-NE-2010A, " Duke Power Company McGuire Nuclear Station Catawba Nuclear Station Nuclear Physics Methodology for Reload Design," June 1985 (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 - UNIT 1 6-21 Amendment No.
} ADMINISfRATIVE CONTROLS CORE OPERATING LIMITS REPORT 4
- 8. DPC-NE-3002, through Rev 2, "FSAR Chapter 13 System Transient Analysis Methodology," SER dated April 26, 1996.
(Methodology used in the system thermal-hydraulic analyses which determine the core operating limits)
- 9. DPC-NE-3000P, Rev.1 " Thermal-Hydraulic Transient Analysis Methodology,"
SER dated December 27, 1995. (Modeling used in the system thermal-hydraulic analyses)
- 10. DPC-NE-1004A, Rev.1, " Nuclear Design Methodology using CASM0-3/ SIMULATE-3P," SCR dated April 26, 1996.
(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient.)
- 11. DPC-NE-2004P-A, Rev.1. " Duke Power Company McGuire and Catawba Nuclear Stations Core Thennal-Hydraulic Methodology using VIPRE-01," SER dated February 20, 1997 (DPC Proprietary).
(Methodology for Specifications 2.2.1 - Reactor Tri tion Setpoints, 3.2.1 - Axial Flux Difference (AFD)p System
, and 3.2.3 -Instrumenta-Nuclear Enthalpy Rise Hot Channel Factor FAH(X,Y).) s
- 12. DPC-NE-2001P-A, Rev.1, " Fuel Mechanical Reload Analysis Methodology for Mark-BW fuel," October 1990 (DPC Proprietary).
(Methodology for Specification 2.2.1 - Reactor Trip System Instrumentation Setpoints.)
- 13. DPC-hE-2005P-A, Rev.1, " Thermal Hydraulic Statistical Core Design Methodolo3y, " SER dated November 7,1996 (DPC Proprietary).
(Methodology for Specification 2.2.1 - Reactor Trip System Instrumentation Setpoints, Specification 3.*e .1 - Axial Flux Difference, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).
- 14. DPC-NE-2008, " Fuel Mechanical Reload Analysis Methodology Using TAC 03,"
SER dated April 3,1995 (DPC Proprietary). (Methodology used for Specification 2.2.1 - Reactor Trip System Instru-mentation setpoints).
- 15. BAW-10183P, Fuel Rod Gas Pressure Criterion, B&W Fuel Company, as approved by SER dated July, 1995.
(Used for Specification 2.2.1, Reactor Trip System Instrumentation Setpoints). 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. 3 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. McGUIRE - UNIT 1 6-22 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 (W Proprietary). (Methodology for Specifications 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limit, 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.)
- 2. WCAP-10216-P-A, " RELAXATION OF CONSTANT AXIAL OFFSET CONTROL FQ SURVEILLANCE TECHNICAL SPECIFICATION", June 1983 (W 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 Fg Methodology.)
- 3. WCAP-10266-P-A Rev. 2, "THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE", March 1987 (W Proprietary).
(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)
- 4. BAW-10168P, "B&W Loss-of-Coolant Accident Evaluation Model for Recirculating Steam Generator Plants," Rev. 1 SER dated January 1991; 1 Rev. 2, SER dated August 22, 1996; Rev. 3, SER dated June 15, 1994 (B&W Proprietary).
(Methodology for Specification 3.2.2 - Heat flux Hot Channel Factor.)
- 5. DPC-NE-2011PA, " Duke Power Company Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors," March 1990 (DPC Proprietary).
(Methodology for Specification 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-3001PA, " Multidimensional Reactor Transients and Safety Analysis Physics Parameter Methodology," November 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-2010A, " Duke Power Company McGuire Nuclear Station Catawba Nuclear Station Nuclear Physics Methodology for Reload Design," June 1985 (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.)
McGU ME - UNIT 2 6-21 Amendment No.
ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT
- 8. DPC-NE-3002, Through Rev. 2, "FSAR Chapter 15 System Transient Analysis l Methodology," SER dated April 26, 1996.
l (Methodology used in the system thermal-hydraulic analyses which determine the core operating limits)
- 9. DPC-NE-3000P, Rev.1, " Thermal-Hydraulic Transient Analysis Methodology,"
SER dated December 27, 1995. (Modeling used in the system thermal-hydraulic analyses)
- 10. DPC-NE-1004A, Rev.1, " Nuclear Design Methodology Using CASM0-3/ SIMULATE-3P " SER dated April 26, 1996.
(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient.)
- 11. 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).
(Methodology for Specifications 2.2.1 - Reactor Trip System Instrumenta-tion Setpoints, 3.2.1 - Axial Flux Difference (AFD), and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor FAH(X,Y).)
- 12. DPC-NE-2001P-A, Rev.1 " Fuel Mechanical Reload Analysis Methodology for Mark-BW fuel," October 1990 (DPC Proprietary).
(Methodology for Specification 2.2.1 - Reactor Trip System Instrumentation Setpoints.)
- 13. DPC-NE-2005P-A, Rev.1, " Thermal Hydraulic Statistical Core Design Methodology," SER dated November 7,1996 (DPC Proprietary).
(Methodology for Specification 2.2.1 - Reactor Trip System Instrumentation Setpoints, Specification 3.2.1 - Axial Flux Difference, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).
- 14. DPC-NE-2008, " Fuel Mechanical Reload Analysis Methodology Using TAC 03,"
SER dated April 3,1995 (DPC Proprietary). (Methodology used for Specification 2.2.1 - Reactor Trip System Instru-mentation setpoints).
- 15. BAW-10183P, Fuel Rod Gas Pressure Criterion, B&W Fuel Company, as approved by SER dated July,1995. l (Used for Specification 2.2.1, Reactor Trip System Instrumentation Setpoints).
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 anclysis limits) of the safety analysis are met. The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supple-ments therete, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector. McGUIRE - UNIT'2 6-22 Amendment No.
ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Continued) (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-NF-2010A, " Duke Power Company McGuire Nuclear Station Catawba Nuclear Station Nuclear Physics Methodology for Reload Design," June 1985 (Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient, Specification 4.7.13.3 - Standby Makeup Pump Water Supply Boron Concentration, and Specification 3.9.1 - RCS and Refueling Canal Boron Concentration, and Specification 3.9.12 - Spent Fuel Pool Boron Concentration.)
- 8. DPC-NE-3002A, Through Rev. 2, "FSAR Chapter 15 System Transient Analysis Methodology," SER Dated April 26, 1996.
(Methodology used in the system thermal-hydraulic analyses which determine the core operating limits)
- 3. DPC-NE-3000P-A, Rev.1," Thermal-Hydraulic Transient Analysis Methodology," SER Dated December 27, 1995.
(Modeling used in the system thermal-hydraulic analyses)
- 10. DPC-NE-1004A, Rev.1, " Design Methodology Using CASM0-3/ Simulate-3P," SER Dated April 26, 1996.
(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient.)
- 11. DPC-NE-2004P-A, Rev.1, " Duke Power Company McGuire and Catawba Nuclear Stations Core Thennal-Hydraulic Methodology using VIPRE-01," SER dated February 20, 1997 (DPC Proprietary).
(Methodology for Specifications 2.2.1 - Reactor Trip System Instrumentation Setpoints, 3.2.1 - Axial Flux Difference (AFD), and 3.2.3
- Nuclear Enthalpy Rise Hot Channel Factor Fa (X,Y).)
- 12. DPC-NE-2001P-A, Rev.1, " Fuel Mechanical Reload Analysis Methodology for Mark-BW Fuel," October 1990 (DPC Proprietary).
(Methodology for Specification 2.2.1 - Reactor Trip System Instrumentation Setpoints.)
- 13. DPC-NE-2005P-A, SER Rev 1, " Therma; Hydraulic Statistical Core Design Methodology," SER dated November 7,1996 (DPC Proprietary).
(Methodology for Specification 2.2.1 - Reactor Trip System Instrumentation Setpoints, Specification 3.2.1 - Axial Flux Difference, and 3.2.3 - Nuclear Enthalpy Rise Hof. Channel Factor) CATAWBA - UNIT 1 6-22 Amendinent No.
ADMIMSTRbTIVE CONTROLS CORE OPERATIpG LIMITS REPORT (Continued)
- 14. DPC-NL-2008, " Fuel Mechnical Reload Analysis Methodology Using TAC 03, " SER dated April 3,1995 (DPC Proprietary).
(Methodology used for Specification 2.2.1 - Reactor Trip System Instrumentation Setpoints)
- 15. BAW-10185P, Fuel Rod Gas Pressure Criterion, B&W Fuel Capany, July,1995.
(Used for S Setpoints) pecification 2.2.1, Reactor Trip System Instrumentation 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 MC in tccordance with 10 CFR 50.4. SEECIAL REPORTS 6.9.2 Special reports shall be submitted to the NRC in accordance with 10 CFR 50.4 within the time period specified for eat.h report. 6.10 RECORD RETENTION 6.10.1 In addition to the applicable record retention requirements of Title 10, Code of Federal Xegulations, the following records shall be retained for at least the minimum period indicated. The following records shall be retained for at least 5 years:
- a. Records and logs of unit operation covering time interval at each power level;
- b. Records and logs of principal maintenance activities, inspections, repair, and replacement of principal items of equipment related to nucleer safety;
- c. All REPORTABLE EVENTS;
- d. Records of surveillance activities, inspections, and calibrations required by these Technical Specifications;
- e. Records of changes made to the procedures required by Specification 6.8.1;
- f. Records of radioactive shipments;
- g. Records of sealed source and fission detector leak tests and results; and CATAWBA - UNIT 1 6-23 Amendment No.
i ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Contillued) (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 l 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-NF-2010A, " Duke Power Company McGuire Nuclear Station Catawba Nuclear 1 Station Nuclear Physics Methodology for Reload Design," June 1985 l (Methodology for Specification 3.1.1.3 - Moderator Temperature l Coefficient, Specification 4.7.13.3 - Standby Makeup Pump Water Supply i Boron Concentration, and Specification 3.9.1 - RCS and Refueling Canal Boron Concentration, and Specification 3.9.12 - Spent Fuel Pool Boron i Concentration.)
- 8. DPC-NE-3002A, Through Rev. 2, "FSAR Chapter 15 System Transient Analysis Methodology," SER Dated April 26. 1996.
(Methodology used in the system thermal-hydraulic analyses wnich determine the core operating limits)
- 9. DPC-NE-3000P-A, Rev.1," Thermal-Hydraulic Transient Analysis Methodology," SER Dated December 27, 1995.
(Modeling used in the system thermal-hydraulic analyses)
- 10. DPC-NE-1004A, Rev.1 " Design Methodology Using CASM0-3/ Simulate-3P " SER Dated April 26, 1996.
(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient.)
- 11. 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).
(Methodology for Specifications 2.2.1 - Reactor Trip System Instrumentation Setpoints, 3.2.1 - Axial Flux Dif ference (AFD), and 3.2.3
- Nuclear Enthalpy Rise Hot Channel Factor F3 g Q,Y).)
- 12. DPC-NE-2001P-A, Rev.1, " Fuel Mechanical Reload Analysis Methodology for Mark-BW Fuel," October 1990 (DPC Proprietary).
(Methodology for Specification 2.2.1 - Reactor Trip System Instrumentation Setpoints.)
- 13. OPC-NE-2005P-A, Rev.1. " Thermal hydraulic Statistical Core Design Methodology," SER dated November L 1996 (DPC Proprietary).
(Methodology for Specification 2.2.1 - Reactor Trip System Instrumentation Setpoints, Specification 3.2.1 - Axial Flux Difference, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor) CATAWBA - UNIT 2 6-22 Amendment No.
ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Continued)
- 14. DPC-NE-2008, " Fuel Hechanical Reload Analysis Methodology Using TAC 03, "
SER dated April 3,1995 (DPC Proprietary). (Methodology used for Specification 4.2.1 - Reactor Trip System , InstrumentationSetpoints)
)
- 15. BAW-10183P, Fuel Rod Gas Pressure Criterion, B&W Fuel Company, July,1995. l (Used for Specification 2.2.1, Reactor Trip System Instrumentation '
Setpoints) 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 LINITS REPORT, including any mid-cycle revisions or supple-ments thereto, shall be provided upon issuance, for each reloao cycle, to the NRC in accordance with 10 CFR 50.4. SPECIAL REPORTS 6.9.2 Saecial reports shall be submitted to the NRC in accordance with 10 CFR 50.4 wit 11n the time period specified for each report. 6.10 . RECORD RETENTION 6.10.1 In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated. The following records shall be retained for at least 5 years:
- a. Records and logs of unit operation covering time interval at each power level;
- b. Records and logs of principal maintenance activities, inspections, repair, and replacement of principal items of equipment related to nuclear safety;
- c. All REPORTABLE EVENTS;
- d. Records of surveillance activities, inspections, and calibration,s required by these Technical Specifications;
- e. Records of changes made to the procedures required by Specification
- 6.8.1;
- f. Records of radioactive shipments;
- g. Records of sealed source and fission detect? leak tests and results; and CATAWBA - UNIT 2 6-23 Amendment No.
._... . . _~ _ .. .._.. . . . _ _. . _ .. .. . _ . . _ _ _ . _ . _ _ _ _ . - . . . . _ _ _ _ _ . . _ . - _ - . _ _ . _ . . . _ _ _ _ _ . . . _ . . . . . _ . . . __.
.i Attachment II.B
.; New Originni Pages - Improved Standard TS i [
?
I t 1 I e n I 1 i T
?
e 5
)
b Y ? i .! x i t i E 4 9 9 h 5 a P k E i 6 5 3 i
,v.e. . - , . . . .. ..e
Reporting Requirements
. 5.6
_ 5.6 Reporting Requirements 5.6.5 [Qg _0PERATING LIMITS REPORT (COLR) (continued)
- 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 SpeHfication 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-10216-P-A, " RELAXATION OF CONSTANT AXIAL OFFSET CONTROL FQ SURVEILLANCE TECHNICAL SPECIFICATION",
June 1983 (W Proprietary).
- 3. WCAP-10266-P-A Rev. 2. "THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE",
March 1987, (W Proprietary).
- 4. BAW-10168P, "B&W Loss-of-Coolant Accident Evaluation Model for Recirculating Steam Generator Plants," Rev.
- 1. SER dated January 1991; Rev. 2. SER dated August 22, 1996; Rev. 3, SER dated June 15, 1994 (B&W Proprietary) .
- 5. DPC-NE-2011P-A, " Duke Power Company Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors," March, 1990 (DPC Proprietary).
- 6. DPC-NE-3001P-A, " Multidimensional Reactor Transients and Safety Analysis Physics Parameter Methodology,"
November,1991 (DPC Proprietary).
- 7. DPC-NF-2010A, " Duke Power Company McGuire Nuclear Station- Catawba Nuclear -Station Nuclear Physics Methodology for Reload Design," June,1985.
(continued) McGuire Unit 1 5.0-28 11/4/97 a
Reporting Requirements ] 5.6 5.6 Reporting Requirements 1 5.6.5 CORE OPERATING LIMITS REPORT (COLR1 (continued)
- 8. DPC-NE-3002, Through Rev. 2 "FSAR Chapter 15 System ;
Transient Analysis Methodology," SER datei April 26, 1996.
- 9. DPC-HE-3000P, Rev. 1 " Thermal-Hydraulic Transient 3 Analysis Methodology," SER dated December 27, 1995. !
- 10. DPC-NE-1004A, Rev.1, " Nuclear Design Methodology }
UsingCASMO-3/ SIMULATE-3P,"SERdatedApril 26, 1996.
- 11. 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). -
- 12. DPC-NE-2001P-A, Rev. 1. " Fuel Mechanical Reload ,
Analysis Methodology for Mark-BW fuel," October 1990 , (DPC Proprietary).
- 13. DPC-NE-2005P-A, Rev. 1, " Thermal Hydraulic Statistical Core Design Methodology," SER dated November 7, 1996 (DPC Proprietary).
i
- 14. DPC-NE-2008, " Fuel Mechanical Reload Analysis Mathodology Using TAC 03," SER dated April 3, 1995 (DPC Proprietary) .
- 15. BAW-10183P, Fuel Rod Gas Pressure Criterion, B&W Fuel Company, July, 1995. l
- c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal me'.nanical 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.
(continued) ; McGuire Unit 1 5.0-29 11/4/97 e=i - m .m- +- W o a p- 4 m- w ._---s- w- m iv ng w --w - - -- e
Reporting Requirements ,
*~
, 506 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued) i
- 9. Reactor Coolant System and refueling canal boron concentration limits for Specification 3.9.1, i
- 10. Spent fuel pool boron concentration limits for Specification 3.7.14, ,
- 11. SHUTDOWN MARGIN for Specification 3.1.1.
- b. The analytical nethods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those de;cribed in the following documents:-
- 1. WCAP-9272-P-A, " WESTINGHOUSE RELOAD SAFETY EVALUATICN ME1HODOLOGY," July 1985 (W Proprietary).
i . 2. WCAP-10216-P-A, "RELAXA110N OF CONSTANT AXIAL OFFSET CONTROL FQ SURVEILLANCE TECHNICAL SPECIFICATION", June.1983 (W Proprietary). >
- 3. WCAP-10266-P-A Rev. 2 "THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE",
March 1987, (W Proprietary). , BAW-1016SP, "B&W Loss-of-Coolant Accident Evaluation
- 4. .
Model for Recirculating Steam Generator Plants," Rev. 1, SER dated January 1991; RE:. 2. SER dated August 22, 1996; Rev. 3 SER dated June 15, 1994 (B&W , Proprietary).
- 5. DPC-NE-2011P-A, " Duke Power Company Nuclear Design
, Methodology for Core Operating Limits of Westinghouse Reactors," March,1990 (DPC Proprietary).
- 6. DPC-NE-3001P-A, " Multidimensional Reactor Transients and Safety Analysis Physics Parameter Methodology,"
November,1991 (DPC Proprietary).
- 7. DPC-NF-2010A, " Duke Power Company McGuire Nuclear i Station' Catawba Nuclear Station Nuclear Physics Methodology for Reload Design," June,1985.
(continued)
'McGuire Unit'2 5.0-?9 11/4/97 . , , , ,..---n, - - . . n--- .,, ,,,, , - . - . - , , , ,- , , ,- -- - - - - , ,
Reporting Requirements
, 5,6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)
- 8. DPC-NE-3002, Through Rev. 2 "FSAR Chapter 15 System Transient Analysis Methodology," SER dated April 26, 1996.
- 9. DPC-NE-3000P, Rev. 1 " Thermal-Hydraulic Transient Analysis Methodology," SER dated December 27, 1995.
- 10. DPC-NE-1004A, Rev. 1. " Nuclear Design Methodology Using CASMO-3/ SIMULATE-3P," SER dated April 26, 1996.
- 11. DPC-NE-2004P-A, Rev. 1. " Duke Power Ccmpany McGuire and Catawba Nuclear Stations Core Thermal-Hydraulic Methodology using VIPRE-01," SER dated February 20, 1997 (DPC Proprietary).
- 12. DPC-NE-2001P-A, Rev. 1, " Fuel Mechanical Reload Analysis Methodology for Mark-BW fuel," October 1990 (DPC Proprietary).
- 13. DPC-NE-2005P-A, Rev.1. " Thermal Hydraulic Statistical Core Design Methodology," SER dated November 7, 1996 (DPC Proprietary).
- 14. DPC-NE-2008, " Fuel Mechanical Reload Analysis Methodology)Using Proprietary . TAC 03,' SER dated April 3,1995 (DPC
- 15. BAW-10183P, Fuel Rod Gas Pressure Criterion, B&W Fuel Company, July, 1995. l
- c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thennal 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.
(continued) McGuire Unit 2 5.0-29 11/4/97
l Reporting Requirements
. . 5.6 '!
5.6 Reporting Requirements S.6.5 CORE OPERATING LIMITS REPORT (COLR) ! (continued)
- 4. BAW-10168P, "B&W Loss-of-Coolant Accident Evaluttion !
Model for Recirculating Steam Generator Plants," ; Rev.-1 SER dated January 1991; Rev. 2. SER Dated ! August 22, 1996; Rev. 3. SER Dated June 15, 1994 (B&W Proprietary) .
- 5. DPC-NE-2011P-A, " Duke Power Corepany Nuclear Design .
Methodology for Core Operating Limits of Westinghouse j Reactors," March,1990 (DPC Proprietary).
- 6. DPC-NE-3001P-A, "Mult1 dimensional Reactor Transients ;
and Safety Analysis Physics Parameter Methodology," . November,1991(DPCProprietary).
- 7. DPC-NF-2010A, " Duke Power Company McGuire Nuclear Station Catawba Nuclear Station Nuclear Physics Methodology for Reload Design," June, 1985.
- 8. DPC-NE-3002A, Through Rev. 2 "FSAR Chapter 15 System Transient Analysis Methodology," SER dated April 26, 1996.
- 9. DPC-NE-3000P-A, Rev. 1 " Thermal-Hydraulic Transient ,
Analysis Methodology," SER Dated December 27, 1995.
- 10. DPC-NE-1004A, Rev.1, " Design Methodology Using CASMO-3/ SIMULATE-3P"SERDatedApril 26, 1996. -
- 11. 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).
- 12. DPC-NE-2001P-A, Rev.1, " Fuel Mechanical Reload Analysis Methodology for Mark-BW fuel," October 1990 (DPC Proprietary).
- 13. DPC-NE-2005P-A, Rev. 1. " Thermal Hydraulic Statistical Core Design Methodology," SER dated November 7, 1996 (DPC Proprietary).
- 14. DPC-NE-2008, " Fuel Mechanical Reload Analysis Methodology Using TAC 03," SER dated April 3,1995 (DPC Propietary).
(continued)- Catawba Unit 1 5.0-29 11/4/97 3 - y 1- v- -wq9 -
-cw p y-- e e3 _..y_ w w m w m. . . . ww,em- gr-
Reporting Requirements
. 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR1 (continued)
- 15. BAW-10183P, Fuel Rod Gas Pressure Criterion, B&W Fuel Company, July,1995.
- c. li The core operating applicable limits (e.mits g., shall bemechanical fuel thermal determined such that all 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 3.6.10. " Annulus Ventilation System-(AVS) " LC0 3.7.10. " Control Room Area' Ventilation System (CRAVS)," LC0 3.7.12. Auxiliary Building Filtered Ventilation Exhaust System (ABFVES) " LCO 3.7.13. " Fuel Handling Ventilation Exhaust System (FHVES)," or LC0 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 Reoort When a report is required by LC0 3.3.3, " Post Accident Monitoring , (PAM) Instrumertation," 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. 5.6.8 Steam Generator Tube Insoection Reoort
- a. The' number of tubes plugged in each steam generator shall be reported to the NRC within 15 days following completion of the program; 1 (continued)
? Catawba Unit 1 5.0-30 11/4/97-l
Reporting Requircments l
. - 5.6 l 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued) [
4.- BAW-10168P, "B&W Loss-of-Coolant. Accident Evaluation l Model for Recirculating Steam Generator Plants," ' Rev. 1. SER dated January 1991; Rev. 2, SER Dated i August 22, 1996; Rev. 3 SER Dated June 15, 1994 (B&W i Proprietary) .
- 5. DPC-NE-2011P-A, " Duke Power Company Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors," March, 1990 (DPC Proprietary).
- 6. DPC-NE-3001P-A, " Multidimensional Reactor Transients '
and Safety Analysis Physics Parameter Methodology," 4 November,1991 (DPC Proprietary).
- 7. DPC-NF-2010A, " Duke Power Company McGuire Nuclear Station Catawba Nuclear Station Nuclear Physics Methodology for Reload Design," June, 1985.
- 8. DPC-NE-3002A, Through Rev. 2 "FSAR Chapter 15 System Transient Analysis Methodology," SER dated April 26, 1996.
- 9. DPC-NE-3000P-A, Rev.1 " Thermal-Hydraulic Transient ;
Analysis Methodology," SER Dated December 27, 1995.
- 10. DPC-NE-1004A, Rev. 1, " Design Methodology Using CASMO-3/ SIMULATE-3P " SER Dated April 26, 1996. .
- 11. 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).
- 12. DPC-NE-2001P-A, Rev. 1, " Fuel Mechanical Reload Analysis Methodology for Mark-BW fuel," October 1990 (DPC Proprietary).
- 13. DPC-NE-2005P-A, Rev.1, " Thermal Hydraulic Statistical Core Design Methodology," SER dated November 7, 1996 (DPC Proprietary).
- 14. DPC-NE-2008, " Fuel Mechanical Reload Analysis Methodology Using TAC 03." SER dated April 3, 1995 (DPC ;
Propietary)'. , (continued) ^
~
Catawba Unit'2 '5.0-29 11/4/97
1 l Reporting Requirements
. - 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)
- 15. BAW-10183P, Fuel Rod Gas Pressure Criterion, B&W Fuel Company, July, 1995.
- c. li The core operating applicable limits (e.mits g., fuelshall thernal be determined such that all 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 net.
- 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 Reoort When a report is required by LC0 3.6.10. " Annulus Ventilation System (AVS)," LCO 3.7.10. " Control Room Area Ventilation System (CRAVS)," LC0 3.7.12, Auxiliary Building Filtered Ventilation ExhaustSystem(ABFVES),"LCO3.7.13,"FuelHandlingVentilation Exhaust System (FHVES)." or LC0 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 Reoort When a report is required by LC0 3.3.3, " Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline taa preplanned alternate nethod of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentaticn channels of the Function to OPERABLE status.
)
5.6.8 Steam Generator Tube Insoection Reoort
- a. The number of tubes plugged in each steam generator shall be reported to the NRC within 15 days following completio7 of the program; (continued)
Catawba Unit 2 5.0-30 11/4/97
. - l l Attachment III Reason for Change I and Technical Justification Generic Letter 88-16 provided guidance on removing cycle-specific parameters which are calculated using NRC-approved methodologies which are listed in Technical Specifications. The parameters are replaced in Tech Specs with a reference to a named report which contains the parameters, and a requirement that the parameters remain within the 12mits specified in the report. The report, unlike the Tech Specs, may be changed by the licensee without prior Commissi(n approval. The proposed changes to TS 6.9.1.9 incorporate NRC-approved revisions to previously-approved methodologies. Since the proposed changes only incorporate NRC-approved methodologies into Technical Specifications, the changes are administrative in nature and can be assumed to have no impact, cr potential impact, on the health and safety of the public or Duke employees. Note that in the case of BAW-10168 (item 4 in each 13: , Revisions 2 and 3 (revisions to different portions or the Topical Report) were pursued simultaneously. This resulted in Revision 3 being approved by the NRC before Pevision 2. For completeness, the SER date for each revision is listed.
Attachment IV No Significant Hazards Analysis The following analysis is presented, pursuant to 10 CFR 50.91, to demonstrate that the proposed change will not create a Significant Hazard Consideration.
- 1. The proposed change will not involve a significant increase in the probability or consequences of an accident previously :
evaluated. The proposed changes to TS 6.9.1.9 are administrative in nature, and do not affect any system, procedure, or manipulation of any equipment which could affect the probability or consequences of any accident..
- 2. The proposed change will not create the possibility of any new accident not previously evaluated.
The proposed changes to TS 6.9.1.5 are administrative in nature, and cannot introduce any new failure mcde or transient which could create any new accident.
- 3. There is no significant reduction in a margin of safety.
The proposed changes to TS 6.9.1.9 are administratiN; in nature, and cannot introduce any new failure mode or transient which could create any accident or reduce any margins. Based on the foregoing analysis, it is concluded that the proposed amendments will not create a significant hazards consideration. r + --. . - . . . .. .- . - - , _ . . , , . - . _ . , - .--.r,, ,m..
_ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ . _ _. _ _ - _ . . . . . __. - _-__= . _ = _ . . . . -- e e< Attachment V Statement of Environmental Impact Pursuant to 10 CF951.22(b), an evaluation has been performed to see if this license amendment request meets the criteria for categorical exclusion set forth in 10CFR51.22 (c) (9) . This proposed amendment to the Technical Specifications of McGuire and Catawba would update the list of approved methodologies by which cycle-specific parameters are calculated. The change will not:
- 1) Create a Significant Hazards Consideration i See Attachment IV.
- 2) Creat< y new effluents or significantly increase any previously identified effluents that may be released offsite ,
This change is administrative, and will not change any station practices or procedures which could result in the creation of new effluents or increase existing releases.
- 3) Will not result in an increase in either personal or .
cumulative radiation exposure. I This change is administrative, and will not change any station practices or procedures which could result in the increase in , exposure. Therefore, this amendment meets the criteria of 10CFR51.22 (c) (9) for categorical exclusion from the requirement for an s environmental impact statement. i h e s - ,m., s w_ -,y,., ,- . - , - , - , - + ..,.nn- ,}}