LR-N08-0062, Submittal of Core Operating Limits Report - Cycle 17

Submittal of Core Operating Limits Report - Cycle 17

ML081190086
Person / Time
Site:	Salem
Issue date:	04/23/2008
From:	Braun R Public Service Enterprise Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LR-N08-0062
Download: ML081190086 (17)

Text

PSEG Nuclear LLC P.O. Box 236,, Hancocks Bridge, NJ 08038-0236 0 PSEG Nuclear L.L. C.

Technical Specification 6.9.1.9 APR 2 3 2008 LR-N08-0062 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington DC 20555-001 Salem Nuclear Generating Station Unit 2 Facility Operating License No. DPR-75 NRC Docket No. 50-311

SUBJECT:

CORE OPERATING LIMITS REPORT- CYCLE 17 In accordance with section 6.9.1.9 of the Salem Unit 2 Technical Specifications, PSEG Nuclear LLC submits the Core Operating Limits Report (COLR) for Salem Unit 2 Cycle 17 in Attachment 1 to this letter.

Should you have any questions regarding this submittal, please contact Mr. E. Villar at (856) 339-5456.

Sincerely, Robert C. Braun Site Vice President - Salem Attachments (1) 7400/

Aii4~

APR 2 3 2008 Document Control Desk Page 2 LR-N08-0062 C Mr. Samuel Collins, Administrator - Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 U. S. Nuclear Regulatory Commission Attn: Mr. R. Ennis, Licensing Project Manager- Salem Mail Stop 08B1 Washington, DC 20555-0001 USNRC Senior Resident Inspector - Salem (X24)

Mr. P. Mulligan, Manager IV Bureau of Nuclear Engineering P.O. Box 415 Trenton, NJ 08625 LR-N08-0062 Core Operating Limit Report For Salem Unit 2, Cycle 17

COLR SALEM 2 Revision 1 February 2008 Core Operating Limits Report for Salem Unit 2, Cycle 17 InI

COLR SALEM 2 PSEG Nuclear LLC Page 2 of 13 Revision I SALEM UNIT 2 CYCLE 17 COLR February 2008 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 Enthalpy Rise Hot Channel Factor FNAH (Specification 3.2.3) 8 2.6 Boron Concentration (Specification 3.9. ) 9 3.0 Analytical Methods 9 4.0 References 10

COLR SALEM 2 PSEG Nuclear LLC Revision I SALEM UNIT 2 CYCLE 17 COLR Page 3 of 13 February 2008 LIST OF FIGURES Figure Fizure Title Page Number Number Rod Bank Insertion Limits vs. Thermal Power 11 2 Axial Flux Difference Limits as a Function of Rated Thermal Power 12 3 K(z) - Normalized FQ(z) as a Function of Core Height 13

COLR SALEM 2 PSEG Nuclear LLC Revision I SALEM UNIT 2 CYCLE 17 COLR Page 4 of 13 February 2008 1.0 CORE OPERATING LIMITS REPORT This Core Operating Limits Report (COLR) for Salem Unit 2 Cycle 17 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 - FAH N 3.9.1 Boron Concentration

COLR SALEM 2 PSEG Nuclear LLC Revision I SALEM UNIT2 CYCLE 17 COLR Page 5 of 13 February 2008 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 and in Section 3.0 of this report.

2.1 Moderator Temperature Coefficient (Specification 3.1.1.3) 2.1.1 The Moderator Temperature Coefficient (MTC) limits are:

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

The EOL/ARO/RTP-MTC shall be less negative than or equal to -4.3xl0"4 Ak/k/ 0 F.

2.1.2 The MTC Surveillance limit is:

The 300 ppm/ARO/RTP-MTC should be less negative than or equal to -3.6x 10-4 Ak/k/ 0F.

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

COLR SALEM 2 PSEG Nuclear LLC Revision I SALEM UNIT 2 CYCLE 17 COLR Page 6 of 13 February 2008 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 1.

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,,, Methodology]

FQ(Z) <FQ* p

• K(Z) for P > 0.5 FRP F6(Z) ' 0 *K(Z)for PO0.5 THERMAL HRIALP0TE POWER where: P = RATED THERMAL POWER 2.4.1 FORTr = 2.40 2.4.2 K(Z) is provided in Figure 3.

2.4.3 FjP= Ff [1.up+ (1.OunRrod pple thgp)]

where: F, Rp = 2.00 for unrodded upper core planes 7 through 6 1.71 for unrodded upper core planes 7 through 31 1.78 for unrodded lower core planes 32 through 55 2.00 for unrodded lower core planes 56 through 61 2.13 for the core planes containing Bank D control rods P = 0.3

COLR SALEM 2 PSEG Nuclear LLC Revision I SALEM UNIT 2 CYCLE 17 COLR Page 7 of 13 February 2008 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:

UFO = ( 1.0+ - U, where:

UQ = Uncertainty for power peaking factor as defined in equation 5-19 of Analytical Method 3.5.

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:

UFO= Uqp U, where:

Uqu = Base FQ measurement uncertainty.

- 1.05 Ue = Engineering uncertainty factor.

= 1.03

COLR SALEM 2 PSEG Nuclear LLC Revision 1 SALEM UNIT2 CYCLE 17 COLR Page 8 of 13 February, 2008 2.5 Nuclear Enthalpy Rise Hot Channel Factor - FNAH (Specification 3.2.3)

FW',H = F/R77' [1.0 + PF'1 (1.0 _P)]

THERA1AL POWER RATED THERMAL POWER 2.5.1 FAJ'dHR 1.65 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, UFAH, to be applied to the Nuclear Enthalpy Rise Hot Channel Factor, FAHn, shall be calculated by the following formula:

UFA = 1.0 + UAH 100.0 where:

UAH = Uncertainty for nuclear-enthalpy rise hot channel factor as defined in equation 5-19 of Analytical Method 3.5.

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

UFAAH = UFA.m where:

UFAHm =Base FAH measurement uncertainty.

= 1.04

COLR SALEM 2 PSEG Nuclear LLC Revision I SALEM UNIT 2 CYCLE 17 COLR Page 9 of 13 February 2008 2.6 Boron Concentration (Specification 3.9.1)

A Mode 6 boron concentration, maintained at or above 2124 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 I% Ak/k uncertainty added.

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

Ak/k uncertainty added.

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

3.0 ANALYTICAL METHODS 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.

3.1 WCAP-9272-P-A, Westinghouse Reload Safety Evaluation Methodology, July 1985 (W proprietary), Methodology for Specifications listed in 6.9.1.9.a., Approved by

-Safety-EvaluationDated-May 28, 1985 .

3.2 WCAP-8385, Power Distribution Control and Load Following Procedures - Topical Report, September 1974 (_W proprietary) Methodology for Specification 3 / 4.2.1 Axial Flux Difference, Approved by Safety Evaluation dated January 31, 1978.

3.3 WCAP-10054-P-A, Rev. 1, Westinghouse Small Break ECCS Evaluation Model Using NOTRUMP Code, August 1985 (W proprietary), Methodology for Specification 3 / 4.2.2 Heat Flux Hot Channel Factor. Approved for Salem by NRC letter dated August 25, 1993.

3.4 WCAP-10266-P-A, Rev. 2, The 1981 Version of Westinghouse Evaluation Model Using the BASH Code, Rev. 2 March 1987 (W proprietary) Methodology for Specification 3 / 4.2.2 Heat Flux Hot Channel Factor, Approved by Safety Evaluation dated November 13, 1986.

3.5 WCAP-12472-P-A, BEACON Core Monitoring and Operations Support System.

Revision 0, (W proprietary). Approved February 1994.

3.6 CENPD-397-P-A, Rev. 1, Improved Flow Measurement Accuracy Using Crossflow Ultrasonic Flow Measurement Technology, May 2000.

3.7 WCAP-10054-P-A, Addendum 2, Rev. 1, Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection into the Broken Loop and COSI Condensation Model. July 1997

COLR SALEM 2 PSEG Nuclear LLC Revision I SALEM UNIT 2 CYCLE 17 COLR Page 10 of 13 February 2008

4.0 REFERENCES

1. Salem Unit 2 Reactor Engineering Procedure, S2.RE-RA.ZZ-00 11 (Q), Tables.

2. Salem Nuclear Generating Station Unit No. 2, Amendment No. 262, License No.

DPR-75. Docket No. 50-311.

COLR SALEM 2 PSEG Nuclear LLC Revision I SALEM UNIT 2 CYCLE 17 COLR Page 11 of 13 Februar' 2008 FIGURE 1 ROD BANK INSERTION LIMITS VS. THERMAL POWER 240 220 200 180 160 140 z

0 120 0 -100 C4 0 80 0

u 60 40 20 0

0 20 40 60 80 100 PERCENT OF RATED THERMAL POWER (%)

COLR SALEM 2 PSEG Nuclear LLC Page 12 of 13 Revision I SALEM UNIT 2 CYCLE 17 COLR February 2008 FIGURE 2 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER 100 80 60 0

H

-e C-0 40 U

20 0 111 1 I . , ý .I .

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

COLR SALEM 2 PSEG Nuclear LLC Page 13 of 13 Revision I SALEM UNIT 2 CYCLE 17 COLR February 2008 FIGURE 3 K(Z) - NORMALIZED FQ(Z) AS A FUNCTION OF CORE HEIGHT 1.2 1.0 0.8 FQ K(Z) Height (FT) u H

Q

<0- 2.40 1.0 0.0 2.40 1.0 6.0 0 0.6 -2.22 -0;925 . 12.0 z

o S0.4 0.2 0.0 0 2 4 6 8 10 12 CORE HEIGHT (FEET)

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