LR-N04-0167, Core Operating Limits Report - Cycle 14, Revision 1

Core Operating Limits Report - Cycle 14, Revision 1

ML041120365
Person / Time
Site:	Salem
Issue date:	04/14/2004
From:	Mannon S Public Service Enterprise Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LR-N04-0167
Download: ML041120365 (16)

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PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge. New Jersey 08038-0236 APR 1 4 2004 PSEG NuclearILLC LR-N04-0167 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 CORE OPERATING LIMITS REPORT - CYCLE 14, REVISION I 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 1 of the Core Operating Limits Report (COLR) for Salem Unit 2 Cycle 14 (NFS-0231, Rev. 1) in Attachment 1 to this letter.

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

Sincerely, S. Mannon Manager - Nuclear Safety & Licensing Attachment A40D 95-2168 REV. 7/99

Document Control Desk LR-N04-0167 APR 14 2004 C Mr. H.J. Miller, Administrator- Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. D. Collins, Licensing Project Manager - Salem U. S. Nuclear Regulatory Commission Mail Stop 8C2 Washington, DC 20555 USNRC Senior Resident Inspector - Salem (X24)

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

Attachment I LR-N04-01 67 SALEM GENERATING STATION UNIT NO. 2 FACILITY OPERATING LICENSE DPR-76 DOCKET NO. 50-311 CORE OPERATING LIMITS REPORT - CYCLE 14 REVISION 1

PSEG Nuclear LLC NFS-0231 Revision 1 March 2004 Core Operating Limits Report for Salem Unit 2, Cycle 14 r'Al U

PREPARED: ."a nyd 1it DATE: 2/Z/1/2iS Frances Pimentel Staff Engineer REVIEWED: /-vt DATE: 3/ql/ZCoq Michael R. Merholz C)

Senior Engineer CONCURRANCE: DATE: 3 2 Thomas K. Ross Supervisor, SR&SA APPROVED: DATE: 3 (4 [ o014 Michael M. Mannion Manager, Nuclear Fuel Section Page 1 of 12

NFS-023 1 PSEG Nuclear LLC Page Iaof 12 Revision I SALEM UNIT 2 CYCLE 14 COLR March 2004 LIST OF EFFECTIVE PAGES Page Revision Level Pages 1-12 I

SUMMARY

OF CHANGES Revision 1: The purpose of Revision I is to incorporate the provisions of Amendment No. 244 regarding Technical Specification 3.9.1 Boron Concentration.

NFS-0231 PSEG Nuclear LLC Page 2 of 12 Revision I SALEM UNIT 2 CYCLE 14 COLR March 2004 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 3.1.1.3) 5 2.2 Control Rod Insertion Limits (Specification 3.1.3.5) 6 2.3 Axial Flux Difference (Specification 3.2.1) 6 2.4 Heat Flux Hot Channel Factor - Fq(z) (Specification 3.2.2) 6 2.5 Nuclear Enthaply Rise Hot Channel Factor FNA1I (Specification 3.2.3) 8 2.6 Boron Concentration (Specification 3.9.1) 8 3.0 References 9

NFS-023 1 PSEG Nuclear LLC Page 3 of 12 Revision I SALEM UNIT 2 CYCLE 14 COLR March 2004 LIST OF FIGURES Figure FigureTitle Page Nember Number 1 Rod Bank Insertion Limits vs. Thermal Power 10 2 Axial Flux Difference Limits as a Function of Rated Thermal Power 11 3 K(z) - Normalized Fq(z) as a Function of Core Height 12

NFS-0231 PSEG Nuclear LLC Page 4 of 12 Revision I SALEM UNIT 2 CYCLE 14 COLR Mtarch 2004 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:

3.1.1.3 Moderator Temperature Coefficient 3.1.3.5 Control Rod Insertion Limits 3.2.1 Axial Flux Difference 3.2.2 Heat Flux Hot Channel Factor - FQ(Z) 3.2.3 Nuclear Enthalpy Rise Hot Channel Factor - FNA, 3.9.1 Boron Concentration

NFS-0231 PSEG Nuclear LLC Page 5 of 12 Revision I SALEM UNIT 2 CYCLE 14 COLR March 2004 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 Temperature Coefficient (Specification 3.1.1.3) 2.1.1 The Moderator Temperature Coefficient (MTC) limits are:

The BOL/AROIHZP-MTC shall be less positive than or equal to 0 Ak/k/ 0 F.

The EOL/ARO/RTP-MTC shall be less negative than or equal to -4.4xl04 Ak/k/0F.

2.1.2 The MTC Surveillance limit is:

The 300 ppm/ARO/RTP-MTC should be less negative than or equal to -3.7x104 Ak/k/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 12 Revision I SALEAI UNIT 2 CYCLE 14 COLR March 2004 2.2 Control Rod Insertion Limits (Specification 3.1.3.5) 2.2.1 The control rod banks shall be limited in physical insertion as shown in Figure 1.

2.3 Axial Flux Difference (Specification 3.2.1)

[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 3.2.2)

[F,,y Methodology]

FOQZ) S Fo

• K(Z)forP>0.5 FRP FQ(Z) -

• K(Z)forP*<0.5 where: P = THERMAL POWER RATED THERMAL POW'ER 2.4.1 FQRTP = 2.40 2.4.2 K(Z) is provided in Figure 3.

2.4.3 F, = FRhp[J.0 + PF,(I.0-P)J where: FIRTP = 1.76 for unrodded upper core planes I through 21 1.82 for unrodded lower core planes 22 through 61 2.13 for the core planes containing Bank D control rods PFv, = 0.3

NFS-023 I PSEG Nuclear LLC Page 7 of 12 Revision I SALEMNI UNIT 2 CYCLE 14 COLR Mlarch 2004 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:

UF (1.0

+~-)0.0)

UQ where:

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

Ur = Engineering uncertainty factor.

= 1.03 Note: UFQe 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 = Uq

• U.

where:

Uqu = Base FQ measurement uncertainty.

= 1.05 U, = Engineering uncertainty factor.

= 1.03

NFS-023 I PSEG Nuclear LLC Page 8 of 12 Revision I SALEM UNIT 2 CYCLE 14 COLR March 2004 2.5 Nuclear Enthalpy Rise Hot Channel Factor - FNAH (Specification 3.2.3)

IFra = FujRTP [1.0 + PF4,, (1.0 - P)]

where: P THERMAL POWIVER RA TED THERMAL POWER 2.5.1 FdJ7hRTP(RFA with IFM) = 1.65 and F&jaRTP(V5H witiou!tIFMf) 1.57 2.5.2 PFAH = 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, UFII, to be applied to the Nuclear Enthalpy Rise Hot Channel Factor, FAIIN, shall be calculated by the following formula:

UFAH = 1.-0+ UAjH 100.0 where:

UA1l = 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, UF&H, to be applied to the Nuclear Enthalpy Rise Hot Channel Factor FiMN shall be calculated by the following formula:

UFA,1 = UFAIm where:

UFWIm =Base FM measurement uncertainty.

= 1.04 2.6 Boron Concentration (Specification 3.9.1)

A Mode 6 boron concentration, maintained at or above 2076 ppm, in the Reactor Coolant System, the fuel storage pool, the refueling canal, and the refueling cavity ensures the most restrictive of the following reactivity conditions is met:

a) A K-effective (Keff) of 0.95 or less at All Rods In (ARI), Cold Zero Power (CZP) conditions with a 1% Ak/k uncertainty added.

b) A Kfr of 0.99 or less at All Rods Out (ARO), CZP conditions with a 1%

Ak/k uncertainty added.

c) A boron concentration of greater than or equal to 2000 ppm, which includes a 50 ppm conservative allowance for uncertainties.

NFS-023 I PSEG Nuclear LLC Page 9 of 12 Revision I SALEM UNIT 2 CYCLE 14 COLR March 2004

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.

3. Salem Nuclear Generating Station Unit No. 2, Amendment No. 244, License No.

DPR-75, Docket No. 50-311.

LLC PSEG Nuclear NFS-0231 PSEG Nuclear LLC Page 10 of 12 Revision I SALEM UNIT 2 CYCLE 14 COLR March 2004 FIGURE 1 ROD BANK INSERTION LIMITS vs. THERMAL POWER 240

_ ___ 117.5;,228 II IIIIII l Z 70.8,22 220 200 7 ___ I__ I I _ I I A 1I I IIIIIIIIII 180

{ D, 18t3 _ _t @ X01 i 160 t-1 II __ - I 71 IA II I I I I I II II 11Of m 140 C -

zn2L 120~

00 120 (I) 0 I II IH1 I II

+I e IW1 I5 I 100

-j 0

z 8 80 I 1l IH +I THTT1,X 60 40 1 L II __ I II l II II II II 20 n

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

NFS-023 I PSEG Nuclear LLC Page IIof 12 Revision I SALEM UNIT 2 CYCLE 14 COLR lMarch 2004 FIGURE 2 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER 100 80

-0 0) 3- 60

• 0 EL a) 0 0) a)

E m

2!

40 ca)

AL 20 -

0 .

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

lI I l NFS-023 1 PSEG Nuclear LLC Page 12 of 12 Revision I SALEM UNIT 2 CYCLE 14 COLR

- March 2004 FIGURE 3 K(Z) - NORMALIZED FQ(Z) AS A FUNCTION OF CORE HEIGHT 1.2 I I Y 1.0 0.8 FQ K(Z)i Height (FTr) a: 2.40 1.0 0.0 0 2.40 1.0 6.0 LL 2.22 0.925 12.0 z

y 0.6 .1 Ui 0

N il CE 0

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