LR-N03-0448, Core Operating Limits Report - Cycle 14

Core Operating Limits Report - Cycle 14

ML033180271
Person / Time
Site:	Salem
Issue date:	11/03/2003
From:	Salamon G Public Service Enterprise Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LR-N03-0448, NFS-0231, Rev 0
Download: ML033180271 (14)

Text

PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 NOV 0 3 2003 PSEG 0 ANuclearLLC LR-N03-0448 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 CORE OPERATING LIMITS REPORT - CYCLE 14 SALEM GENERATING STATION UNIT NO. 2 FACILITY OPERATING LICENSE DPR-75 DOCKET NO. 50-311 In accordance with section 6.9.1.9 of the Salem Unit 2 Technical Specifications, PSEG Nuclear LLC submits Revision 0 of the Core Operating Limits Report (COLR) for Salem Unit 2 Cycle 14 (NFS-0231, Rev. 0) in Attachment 1 to this letter.

Should you have any questions, please contact Mr. Paul Duke at (856) 339-1466.

Sincerely, G. Salamon Manager - Nuclear Safety & Licensing Attachment kDo(

95-2168 REV. 7/99

Document Control Desk NOV 0 3 2003 LR-N03-0448 C Mr. H. J. Miller, Administrator - Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. R. Fretz, Licensing Project Manager - Salem U. S. Nuclear Regulatory Commission Mail Stop 8B2 Washington, DC 20555 USNRC Senior Resident Inspector - Salem (X24)

Mr. K. Tosch, Manager IV Bureau of Nuclear Engineering PO Box 415 Trenton, NJ 08625

Document Control Desk LR-N03-0448 Attachment I SALEM GENERATING STATION UNIT NO. 2 FACILITY OPERATING LICENSE DPR-75 DOCKET NO. 50-311 CORE OPERATING LIMITS REPORT - CYCLE 14 REVISION 0

PSEG Nuclear LLC NFS-0231 Revision 0 June 2003 Core Operating Limits Report for Salem Unit 2, Cycle 14 k I <

PREPARED: DATE: sl27/ioO3 Michael R. Merholz Senior Engineer PREPARED: (2oLZ &. 'Ri-*I DATE: ilJ a-i Corie A. Reeves Engineer REVIEWED: DATE: /a w/x03 Bradley A. Beck Lead Engineer CONCURRANCE: 7/c$a.O . DATE: 6Z71ZO03 Thomas K. Ross Supervisor, SR&SA APPROVED: 7t//*A 7CAAm M/,Un1 DATE: 0203 Michael M. Mannion Manager, Nuclear Fuel Section Page 1 of 11

NFS-023 I PSEG Nuclear LLC Page 2 of I I Revision 0 SALEM UNIT 2 CYCLE 14 COLR June 2003 TABLE OF CONTENTS Section Section Title Page Number Number Table of Contents 2 List of Figures 3 1.0 Core Operating Limits Report 4 2.0 Operating Limits 5 2.1 Moderator Temperature Coefficient (Specification 3/4.1.1.3) 5 2.2 Control Rod Insertion Limits (Specification 3/4.1.3.5) 6 2.3 Axial Flux Difference (Specification 3/4.2.1) 6 2.4 Heat Flux Hot Channel Factor - Fq(z) (Specification 3/4.2.2) 6 2.5 Nuclear Enthaply Rise Hot Channel Factor FNai (Specification 3/4.2.3) 8 3.0 References 8

NFS-023 1 PSEG Nuclear LLC Page 3 of II Revision 0 SALEM UNIT 2 CYCLE 14 COLR June 2003 LIST OF FIGURES Figure Fh'ure Title Page Number Number 2 Rod Bank Insertion Limits vs. Thermal Power 9 2 Axial Flux Difference Limits as a Function of Rated Thermal Power 10

.3 K(z) - Normalized Fq(z) as a Function of Core Height 11

NFS-0231 PSEG Nuclear LLC Page 4 of II Revision 0 SALEM UNIT 2 CYCLE 14 COLR June 2003 1.0 CORE OPERATING LIMITS REPORT This Core Operating Limits Report (COLR) for Salem Unit 2 Cycle 14 has been prepared in accordance with the requirements of Technical Specification 6.9.1.9.

The Technical Specifications affected by this report are listed below:

3/4.1.1.3 Moderator Temperature Coefficient 314.1.3.5 Control Rod Insertion Limits 3/4.2.1 Axial Flux Difference 314.2.2 Heat Flux Hot Channel Factor - FQ(Z) 3/4.2.3 Nuclear Enthalpy Rise Hot Channel Factor - Fob

NFS-0231 PSEG Nuclear LLC Page 5 of l l Revision 0 SALEM UNIT 2 CYCLE 14 COLR June 2003 2.0 OPERATING LIMITS The cycle-specific parameter limits for the specifications listed in Section 1.0 are presented in the following subsections. These limits have been developed using the NRC-approved methodologies specified in Technical Specification 6.9.1.9.

2.1. Moderator Temnerature Coefficient (Specification 3/4.1.1.3) 2.1.1 The Moderator Temperature Coefficient (MTC) limits are:

The BOIJARO/HZP-MTC shall be less positive than or equal to 0 Akfk/ 0 F.

The EOIUAROIRTP-MTC shall be less negative than or equal to -4.4xl04 kIc/k/F.

2.1.2 The MTC Surveillance limit is:

The 300 ppm/ARO/RTP-MTC should be less negative than or equal to -3.7x10-4 Ak/kr/0 F.

where: BOL stands for Beginning of Cycle Life ARO stands for All Rods Out HZP stands for Hot Zero THERMAL POWER EOL stands for End of Cycle Life RTP stands for Rated THERMAL POWER

NFS-0231 PSEG Nuclear LLC Page 6 of II Revision 0 SALEM UNIT 2 CYCLE 14 COLR June 2003 2.2 Control Rod Insertion Limits (Specification 3/4.1.3.5) 2.2.1 The control rod banks shall be limited in physical insertion as shown in Figure 1.

2.3 Axial Flux Difference (Specification 3/4.2.1)

[Constant Axial Offset Control (CAOC) Methodology]

2.3.1 The Axial Flux Difference (AFD) target band shall be the more restrictive of

(+6%, -9%) or the target band as defined in Reference 2.

2.3.2 The AFD Acceptable Operation Limits are provided in Figure 2.

2.4 Heat Flux Hot Channel Factor - FQ(Z) (Specification 3/4.2.2)

[F1 yMethodology]

FQ(Z) <

• K(Z) forP>Oa5 FQ~

FQ(Z) S 05 K(Z for P <05a THERMAL POWER where: P = RATED THERMAL POWER 2.4.1 FQR = 2.40 2.4.2 K(Z) is provided in Figure 3.

2.4.3 F1 L = F,m"(1.0 + PF, (1.0 - P)J where: FF,,RT = 1.76 for unrodded upper core planes I through21 1.82 for unrodded lower core planes 22 through 61 2.13 for the core planes containing Bank D control rods PF,= 0.3

NFS-023 I PSEG Nuclear LLC Page 7 of I I Revision 0 SALEM UNIT 2 CYCLE 14 COLR June 2003 2.4.4 If the Power Distribution Monitoring System (PDMS) is used for core power distribution surveillance and is OPERABLE, as defined in Technical Specification 3.3.3.14, the uncertainty, UFQ, to be applied to the Heat Flux Hot Channel Factor FQ(z) shall be calculated by the following formula:

UQ)

U UF (10 1.00.0 where:

UQ = Uncertainty for power peaking factor as defined in equation 5-19 of Reference 1.

• Ue = Engineering uncertainty factor.

= 1.03 Note: UFQ= PDMS Surveillance Report Core Monitor Fxy Uncertainty in %.

2.4.5 If the INCORE movable detectors are used for core power distribution surveillance, the uncertainty, UFQ, to be applied to the Heat Flux Hot Channel Factor FQ(z) shall be calculated by the following formula:

UFQ=Us,, *U, where:

Uqu = Base FQ measurement uncertainty.

= 1.05 Ue = Engineering uncertainty factor.

= 1.03

NFS-0231 PSEG Nuclear LLC Page 8 of II Revision 0 SALEM UNIT 2 CYCLE 14 COLR June 2003 2.5 Nuclear Enthalpy Rise Hot Channel Factor - FNAH (Specification 3/4.2.3)

Fvj_,RTPfI.O+pF9,(,.O-P) .

THERMAL POWER where: P = RATED THERMAL POWER 2.5.1 Ftj1 RTP(RFA with IFM) = 1.65 and Ff'(VSH without JFM) = 1.57 2.5.2 PFH = 0.3 2.5.3 If the Power Distribution Monitoring System (PDMS) is used for core power distribution surveillance and is OPERABLE, as defined in Technical Specification 3.3.3.14, the uncertainty, UFm, to be applied to the Nuclear Enthalpy Rise Hot Channel Factor, PFN, shall be calculated by the following formula:

UFAH = 1-0+

100.0 where:

UAH = Uncertainty for enthalpy rise as defined in equation 5-19 of Reference 1.

2.5.4 If the INCORE movable detectors are used for core power distribution surveillance, the uncertainty, UFH, to be applied to the Nuclear Enthalpy Rise Hot Channel Factor.FitHl shall be calculated by the following formula:

UFAH = UFfflm where:

-UFw =Base FI measurement uncertainty.

-1.04

3.0 REFERENCES

1. WCAP-12472-P-A, BEACON Core Monitoring and Operations Support System, August 1994.

2. S2.RE-RA.ZZ-001 1(Q), Tables.

NFS-0231 PSEG Nuclear LLC Pagc9of II Revision 0 SALEM UNIT 2 CYCLE 14 COLR June 2003 FIGURE 1 ROD BANK INSERTION LIMITS vs. THERMAL POWER 240 0111 IIIIImII1111111 mmII1 220 E,2~~.5228 1 1__liTll1 70.,'

200 BAN B + 1 180 _ _iC_II l ,KI_ _ I II 1 11 11 1 I Z Iz 7~~~~~~~'

_ / _M= == _ _ ___

_AI

_717Of TT T_

1 71 r__l 2_____llllL

%O i 160

°. 140

/ 11 11114100,171 M~~~~~~

a) zZ 0

<e1-e E 120 0

a-larr~~~~~~~~~~~~N i DF -lI111 100

-j 0 I I II I I _ _ _ I Z __ _ _ __ _ _ _ I I I I T 1g-1 cc 880 60 40 20 0

0 10 20 30. 40 50 .60 70 80 90 100 PERCENT OF RATED THERMAL POWER (%)

NFS-0231 PSEG Nuclear LLC Page 10 of 11 Revision 0 SALEM UNIT 2 CYCLE 14 COLR June 2003 FIGURE 2 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER 100 80

-0 a) 60 0R 0

a)

E H

co EC 0

a, 40 a,

20 -

0-L

-50 -40 -30 -20 -10 0 10 20 30 40 50 Flux Difference (% Delta I)

NFS-023 I PSEG Nuclear LLC Page I I of I I Revision 0 SALEM UNIT 2 CYCLE 14 COLR June 2003 FIGURE 3 K(Z) - NORMALIZED FQ(Z) AS A FUNCTION OF CORE HEIGHT 1.2 Y I 1.0 0.8

- .EQ. ~K(Z) Height (FIT)

C:

I-1 2.40 1.0 0.0 2.40 1.0. 6.0 z 2.22 0.925 12.0

<s 0.6 Lu a

N cc:

0.4 0.2 0.0 0 2 4 6 8 10 12 CORE HEIGHT (FEET)