LR-N15-0233, Core Operating Limits Report - Cycle 22

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Core Operating Limits Report - Cycle 22
ML15322A198
Person / Time
Site: Salem PSEG icon.png
Issue date: 11/18/2015
From: Jamila Perry
Public Service Enterprise Group
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
LR-N15-0233
Download: ML15322A198 (16)
v • d • e


Text

PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, NJ 08038-0236 NOV 18 2015 LR-N15-0233 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-001 Salem Nuclear Generating Station Renewed Facility Operating License No. DPR-75 NRC Docket No. 50-311

Subject:

Salem Unit 2 Core Operating Limits Report - Cycle 22 Technical Specification 6.9.1.9 In accordance with the requirements of section 6.9.1.9 of the Salem Unit 2 Technical Specifications, PSEG Nuclear LLC is submitting the enclosed Core Operating Limits Report (COLR) for Salem Unit 2, Cycle 22.

If you have any questions or require additional information, please contact Mr. David Lafleur at (856) 339-1754.

There are no regulatory commitments contained in this letter.

Sincerely,

<1Lr

z:n 1

F. Perry

_, U Site Vice President Salem Generating Station Enclosure - Core Operating Limits Report for Salem Unit 2, Cycle 22

NOV 18 2015 Technical Specification 6.9.1.9 Document Control Desk Page 2 LR-N15-0233 cc:

Mr. D. Dorman, Administrator - Region 1, NRC Mr. T. Wengert, Licensing Project Manager (acting) - Salem, NRC Mr. P. Finney, USNRC Senior Resident Inspector, Salem (X24)

Mr. P. Mulligan, Manager IV, NJBNE Mr. R. Braun, President and Chief Nuclear Officer - Nuclear Mr. T. Cachaza, Salem Commitment Tracking Coordinator Mr. L. Marabella, Corporate Commitment Tracking Coordinator Mr. D. Lafleur, Salem Regulatory Assurance

LR-N 15-0233 Enclosure Core Operating Limits Report for Salem Unit 2, Cycle 22

COLR SALEM 2 Revision 6 July 2015 Core Operating Limits Report for Salem Unit 2, Cycle 22 Page 1of13

COLRSALEM2 Revision 6 July 2015 Section Number 1.0 2.0 2.1 2.2 2.3 2.4 2.5 2.6 3.0 4.0 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 22 COLR TABLE OF CONTENTS Section Title Table of Contents List of Figures Core Operating Limits Report Operating Limits Moderator Temperature Coefficient (Specification 3.1.1.3)

Control Rod Insertion Limits (Specification 3.1.3.5)

Axial Flux Difference (Specification 3.2.1)

Heat Flux Hot Channel Factor - F 0(z) (Specification 3.2.2 )

Nuclear Enthalpy Rise Hot Channel Factor FN,,.H (Specification 3.2.3)

Boron Concentration (Specification 3.9.1)

Analytical Methods References Page Number 2

3 4

5 5

6 6

6 8

9 9

10 Page 2 ofl3

COLRSALEM2 Revision 6 July 2015 Figure Number 2

3 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 22 COLR LIST OF FIGURES Figure Title Rod Bank Insertion Limits vs. Thermal Power Axial Flux Difference Limits as a Function of Rated Thermal Power K(z)- Normalized FQ(z) as a Function of Core Height Page Number 11 12 13 Page 3 of13

TS Section 3.1.1.3 3.1.3.5 3.2.1 3.2.2 3.2.3 3.9.1 COLRSALEM2 Revision 6 July 2015 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 22 COLR 1.0 CORE OPERA TING LIMITS REPORT Page 4of13 This Core Operating Limits Report (COLR) for Salem Unit 2 Cycle 22 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 along with the NRC-approved methodologies used to develop and/or determine COLR parameters identified in technical Specifications.

NRC Approved COLR Methodology Technical Specifications COLR Parameter Section (Section 3.0 Number)

Moderator Temperature Coefficient MTC 2.1 3.1, 3.6 Control Rod Insertion Limits Control Rod Insertion Limits 2.2 3.1, 3.6 Axial Flux Difference AFD 2.3 3.1, 3.2, 3.6 3.1, 3.3, 3.4, 3.5, Heat Flux Hot Channel Factor - F0(Z)

Fo(Z) 2.4 3.6, 3.7, 3.8 Nuclear Enthalpy Rise Hot Channel Factor - FN m

FN m

2.5 3.1, 3.5, 3.6, 3.8 Boron Concentration Boron Concentration 2.6 3.1, 3.6

COLRSALEM2 Revision 6 July 2015 2.0 OPERA TING LIMITS PSEG Nuclear LLC SALEM UNIT 2 CYCLE 22 COLR 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 i1k/k/°F.

Page 5 of13 The EOL/ARO/RTP-MTC shall be less negative than or equal to -4.3x10*4 i1k/k/°F.

2.1.2 The MTC Surveillance limit is:

The 300 ppm/ARO/RTP-MTC should be less negative than or equal to -3.6x10*4 i1k/k/°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 RA TED THERMAL POWER

COLRSALEM2 Revision 6 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 22 COLR July 2015 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 (+6%, -9%).

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)

[Fxy Methodology]

FQRTP FQ(Z) ::;;

p

  • K(Z) for P > 0.5 FlTP FQ(Z) ::;;
  • K(Z) for P ::::0.5

0.5 where

P THERMAL POWER RATED THERMAL POWER 2.4.1 F/TP

= 2.40 2.4.2 K(Z) is provided in Figure 3.

2.4.3 Fr/

= Fx/TP [1.0 + PFxy (1.0 - P)]

where: from BOL to 12000 MWD/MTU F RTP

=

xy 2.10 for unrodded upper core planes 1 through 6 1.89 for unrodded upper core planes 7 through 8 1.76 for unrodded upper core planes 9 through 31 1.86 for unrodded lower core planes 32 through 53 1.96 for unrodded lower core planes 54 through 55 2.10 for unrodded lower core planes 56 through 61 2.13 for the core planes containing Bank D control rods PFxy 0.3 Page 6 of13

COLRSALEM2 Revision 6 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 22 COLR July 2015 where: from 12000 MWD/MTU to EOL F RTP

=

xy 2.03 for unrodded upper core planes 1 through 6 1.84 for unrodded upper core planes 7 through 8 1.77 for unrodded upper core planes 9 through 31 1.81 for unrodded lower core planes 32 through 53 1.84 for unrodded lower core planes 54 through 55 1.97 for unrodded lower core planes 56 through 61 2.13 for the core planes containing Bank D control rods PFxy

= 0.3 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:

where:

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

U0 =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=Uq11*Ue where:

Uqu =Base Fq measurement uncertainty.

= 1.05 U0 = Engineering uncertainty factor.

= 1.03 Page 7of13

COLRSALEM2 Revision 6 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 22 COLR July 2015 2.5 Nuclear Enthalpy Rise Hot Channel Factor - FN liH (Specification 3.2.3) where:

P THERMAL POWER RATED THERMAL POWER 1.65 2.5.2 PF Lili

= 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 tmcertainty, UF.iH, to be applied to the Nuclear Enthalpy Rise Hot Channel Factor, FN AH, shall be the greater of 1.04 or as calculated by the following formula:

UFAff = 1.0 + u Af{

100.0 where:

UAfl = Uncertainty for enthalpy rise hot channel factor as defined in equation 5-19 of Analy!ical Method 3.5.

2.5.4 If the INCORE movable detectors are used for core power distribution surveillance, the tmcertainty, UF&H, to be applied to the Nuclear Enthalpy Rise Hot Channel Factor FN AfI shall be calculated by the following formula:

where:

UFAflm =Base F AfI measurement uncertainty.

= 1.04 Page 8of13

COLRSALEM2 Revision 6 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 22 COLR July 2015 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 (K0tr) of 0.95 or less at All Rods In (ARI), Cold Zero Power (CZP) conditions with a 1 % ik/k uncertainty added.

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

ik/k uncertainty added.

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 (Westinghouse proprietary). Methodology for Specifications listed in 6.9.1.9.a.

Approved by Safety Evaluation dated May 28, 1985.

3.2 WCAP-8385, Power Distribution Control and Load Following Procedures - Topical Report, September 1974 (Westinghouse 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, Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code, August 1985 (Westinghouse 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, Revision 2, The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code, March 1987 (Westinghouse proprietary).

Methodology for Specification 3/4.2.2 Heat Flux Hot Channel Factor. Approved by Safety Evaluation dated November 13, 1986.

Page 9 of13

COLRSALEM2 Revision 6 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 22 COLR July 2015 3.5 WCAP-12472-P-A, BEACON - Core Monitoring and Operations Support System, August 1994 (Westinghouse proprietary). Approved by Safety Evaluation dated February 16, 1994.

3.6 CENPD-397-P-A, Revision 01, Improved Flow Measurement Accuracy Using Crossflow Ultrasonic Flow Measurement Technology, May 2000. Approved by Safety Evaluation dated March 20, 2000.

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

Approved by Safety Evaluation dated August 12, 1996.

3.8 WCAP-12472-P-A, Addendum 1-A, BEACON Core Monitoring and Operations Support System, January 2000 (Westinghouse proprietary). Approved by Safety Evaluation dated September 30, 1999.

4.0 REFERENCES

1. Salem Nuclear Generating Station Unit No. 2, Amendment 290, Renewed License No.

DPR-75, Docket No. 50-311.

Page 10of13

COLRSALEM2 Revision 6 July 2015 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 22 COLR FIGURE 1 ROD BANK INSERTION LIMITS VS. THERMAL POWER 240 220 200 180 c

... 160

'ti

.c

"' 140

c.

.Sl o

z 0

j:: 120 in 0

ll..

p z

<C 100 Ill

...I 0

ct:

I-z 0

80 u

60 40 20 0

/I 11.1, 221 I

, v

/I BANK B I I

0, 186 I

J v j v v

v v

J 1 0,58 I

 v 0

10 20 30 v

j I

v VI BANK c I

/

I I

/

I/

i/

40 50 I 170.4, 227

/

I v

v v

I/ v I

/ I BANK D I

I v

60 70 80 PERCENT OF RATED THERMAL POWER(%)

1 100, 110 I v

I j

v 90 100 Page 11 of13

COLRSALEM2 Revision 6 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 22 COLR July 2015 100 80

R 0

60 1i) h 0

Po<

ro 5

n E-<

"O

(!)

+

40 0

1:l

(!)

u

(!)

Po<

20 0

FIGURE 2 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER k-11,90%

I c11,9o)I I

\\

UNACCEPTABLE UNACCEPTABLE OPERATION OPERATION

!ACCEPTABLE OPERATION I

\\

j

\\

lc-31,50)1 I (31,5o]

-50

-40

-30

-20

-10 0

10 20 30 40 Flux Difference (% Delta I)

Page 12 of13 50

COLRSALEM2 Revision 6 July 2015 1.2 1.0 g

g 0.8



E-<



µ...  06



0 0.4 z

0.2 0.0 0

PSEG Nuclear LLC SALEM UNIT 2 CYCLE 22 COLR FIGURE 3 K(Z) - NORMALIZED FQ(Z) AS A FUNCTION OF CORE HEIGHT 2

FQ K(Z) 2.40 1.0 2.40 1.0 2.22 0.925 4

6 8

CORE HEIGHT (FEET)

Height (FT) 0.0 6.0 12.0 10 12 Page 13 ofl3

PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, NJ 08038-0236 NOV 18 2015 LR-N15-0233 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-001 Salem Nuclear Generating Station Renewed Facility Operating License No. DPR-75 NRC Docket No. 50-311

Subject:

Salem Unit 2 Core Operating Limits Report - Cycle 22 Technical Specification 6.9.1.9 In accordance with the requirements of section 6.9.1.9 of the Salem Unit 2 Technical Specifications, PSEG Nuclear LLC is submitting the enclosed Core Operating Limits Report (COLR) for Salem Unit 2, Cycle 22.

If you have any questions or require additional information, please contact Mr. David Lafleur at (856) 339-1754.

There are no regulatory commitments contained in this letter.

Sincerely,

<1Lr

z:n 1

F. Perry

_, U Site Vice President Salem Generating Station Enclosure - Core Operating Limits Report for Salem Unit 2, Cycle 22

NOV 18 2015 Technical Specification 6.9.1.9 Document Control Desk Page 2 LR-N15-0233 cc:

Mr. D. Dorman, Administrator - Region 1, NRC Mr. T. Wengert, Licensing Project Manager (acting) - Salem, NRC Mr. P. Finney, USNRC Senior Resident Inspector, Salem (X24)

Mr. P. Mulligan, Manager IV, NJBNE Mr. R. Braun, President and Chief Nuclear Officer - Nuclear Mr. T. Cachaza, Salem Commitment Tracking Coordinator Mr. L. Marabella, Corporate Commitment Tracking Coordinator Mr. D. Lafleur, Salem Regulatory Assurance

LR-N 15-0233 Enclosure Core Operating Limits Report for Salem Unit 2, Cycle 22

COLR SALEM 2 Revision 6 July 2015 Core Operating Limits Report for Salem Unit 2, Cycle 22 Page 1of13

COLRSALEM2 Revision 6 July 2015 Section Number 1.0 2.0 2.1 2.2 2.3 2.4 2.5 2.6 3.0 4.0 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 22 COLR TABLE OF CONTENTS Section Title Table of Contents List of Figures Core Operating Limits Report Operating Limits Moderator Temperature Coefficient (Specification 3.1.1.3)

Control Rod Insertion Limits (Specification 3.1.3.5)

Axial Flux Difference (Specification 3.2.1)

Heat Flux Hot Channel Factor - F 0(z) (Specification 3.2.2 )

Nuclear Enthalpy Rise Hot Channel Factor FN,,.H (Specification 3.2.3)

Boron Concentration (Specification 3.9.1)

Analytical Methods References Page Number 2

3 4

5 5

6 6

6 8

9 9

10 Page 2 ofl3

COLRSALEM2 Revision 6 July 2015 Figure Number 2

3 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 22 COLR LIST OF FIGURES Figure Title Rod Bank Insertion Limits vs. Thermal Power Axial Flux Difference Limits as a Function of Rated Thermal Power K(z)- Normalized FQ(z) as a Function of Core Height Page Number 11 12 13 Page 3 of13

TS Section 3.1.1.3 3.1.3.5 3.2.1 3.2.2 3.2.3 3.9.1 COLRSALEM2 Revision 6 July 2015 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 22 COLR 1.0 CORE OPERA TING LIMITS REPORT Page 4of13 This Core Operating Limits Report (COLR) for Salem Unit 2 Cycle 22 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 along with the NRC-approved methodologies used to develop and/or determine COLR parameters identified in technical Specifications.

NRC Approved COLR Methodology Technical Specifications COLR Parameter Section (Section 3.0 Number)

Moderator Temperature Coefficient MTC 2.1 3.1, 3.6 Control Rod Insertion Limits Control Rod Insertion Limits 2.2 3.1, 3.6 Axial Flux Difference AFD 2.3 3.1, 3.2, 3.6 3.1, 3.3, 3.4, 3.5, Heat Flux Hot Channel Factor - F0(Z)

Fo(Z) 2.4 3.6, 3.7, 3.8 Nuclear Enthalpy Rise Hot Channel Factor - FN m

FN m

2.5 3.1, 3.5, 3.6, 3.8 Boron Concentration Boron Concentration 2.6 3.1, 3.6

COLRSALEM2 Revision 6 July 2015 2.0 OPERA TING LIMITS PSEG Nuclear LLC SALEM UNIT 2 CYCLE 22 COLR 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 i1k/k/°F.

Page 5 of13 The EOL/ARO/RTP-MTC shall be less negative than or equal to -4.3x10*4 i1k/k/°F.

2.1.2 The MTC Surveillance limit is:

The 300 ppm/ARO/RTP-MTC should be less negative than or equal to -3.6x10*4 i1k/k/°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 RA TED THERMAL POWER

COLRSALEM2 Revision 6 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 22 COLR July 2015 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 (+6%, -9%).

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)

[Fxy Methodology]

FQRTP FQ(Z) ::;;

p

  • K(Z) for P > 0.5 FlTP FQ(Z) ::;;
  • K(Z) for P ::::0.5

0.5 where

P THERMAL POWER RATED THERMAL POWER 2.4.1 F/TP

= 2.40 2.4.2 K(Z) is provided in Figure 3.

2.4.3 Fr/

= Fx/TP [1.0 + PFxy (1.0 - P)]

where: from BOL to 12000 MWD/MTU F RTP

=

xy 2.10 for unrodded upper core planes 1 through 6 1.89 for unrodded upper core planes 7 through 8 1.76 for unrodded upper core planes 9 through 31 1.86 for unrodded lower core planes 32 through 53 1.96 for unrodded lower core planes 54 through 55 2.10 for unrodded lower core planes 56 through 61 2.13 for the core planes containing Bank D control rods PFxy 0.3 Page 6 of13

COLRSALEM2 Revision 6 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 22 COLR July 2015 where: from 12000 MWD/MTU to EOL F RTP

=

xy 2.03 for unrodded upper core planes 1 through 6 1.84 for unrodded upper core planes 7 through 8 1.77 for unrodded upper core planes 9 through 31 1.81 for unrodded lower core planes 32 through 53 1.84 for unrodded lower core planes 54 through 55 1.97 for unrodded lower core planes 56 through 61 2.13 for the core planes containing Bank D control rods PFxy

= 0.3 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:

where:

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

U0 =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=Uq11*Ue where:

Uqu =Base Fq measurement uncertainty.

= 1.05 U0 = Engineering uncertainty factor.

= 1.03 Page 7of13

COLRSALEM2 Revision 6 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 22 COLR July 2015 2.5 Nuclear Enthalpy Rise Hot Channel Factor - FN liH (Specification 3.2.3) where:

P THERMAL POWER RATED THERMAL POWER 1.65 2.5.2 PF Lili

= 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 tmcertainty, UF.iH, to be applied to the Nuclear Enthalpy Rise Hot Channel Factor, FN AH, shall be the greater of 1.04 or as calculated by the following formula:

UFAff = 1.0 + u Af{

100.0 where:

UAfl = Uncertainty for enthalpy rise hot channel factor as defined in equation 5-19 of Analy!ical Method 3.5.

2.5.4 If the INCORE movable detectors are used for core power distribution surveillance, the tmcertainty, UF&H, to be applied to the Nuclear Enthalpy Rise Hot Channel Factor FN AfI shall be calculated by the following formula:

where:

UFAflm =Base F AfI measurement uncertainty.

= 1.04 Page 8of13

COLRSALEM2 Revision 6 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 22 COLR July 2015 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 (K0tr) of 0.95 or less at All Rods In (ARI), Cold Zero Power (CZP) conditions with a 1 % ik/k uncertainty added.

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

ik/k uncertainty added.

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 (Westinghouse proprietary). Methodology for Specifications listed in 6.9.1.9.a.

Approved by Safety Evaluation dated May 28, 1985.

3.2 WCAP-8385, Power Distribution Control and Load Following Procedures - Topical Report, September 1974 (Westinghouse 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, Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code, August 1985 (Westinghouse 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, Revision 2, The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code, March 1987 (Westinghouse proprietary).

Methodology for Specification 3/4.2.2 Heat Flux Hot Channel Factor. Approved by Safety Evaluation dated November 13, 1986.

Page 9 of13

COLRSALEM2 Revision 6 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 22 COLR July 2015 3.5 WCAP-12472-P-A, BEACON - Core Monitoring and Operations Support System, August 1994 (Westinghouse proprietary). Approved by Safety Evaluation dated February 16, 1994.

3.6 CENPD-397-P-A, Revision 01, Improved Flow Measurement Accuracy Using Crossflow Ultrasonic Flow Measurement Technology, May 2000. Approved by Safety Evaluation dated March 20, 2000.

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

Approved by Safety Evaluation dated August 12, 1996.

3.8 WCAP-12472-P-A, Addendum 1-A, BEACON Core Monitoring and Operations Support System, January 2000 (Westinghouse proprietary). Approved by Safety Evaluation dated September 30, 1999.

4.0 REFERENCES

1. Salem Nuclear Generating Station Unit No. 2, Amendment 290, Renewed License No.

DPR-75, Docket No. 50-311.

Page 10of13

COLRSALEM2 Revision 6 July 2015 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 22 COLR FIGURE 1 ROD BANK INSERTION LIMITS VS. THERMAL POWER 240 220 200 180 c

... 160

'ti

.c

"' 140

c.

.Sl o

z 0

j:: 120 in 0

ll..

p z

<C 100 Ill

...I 0

ct:

I-z 0

80 u

60 40 20 0

/I 11.1, 221 I

, v

/I BANK B I I

0, 186 I

J v j v v

v v

J 1 0,58 I

 v 0

10 20 30 v

j I

v VI BANK c I

/

I I

/

I/

i/

40 50 I 170.4, 227

/

I v

v v

I/ v I

/ I BANK D I

I v

60 70 80 PERCENT OF RATED THERMAL POWER(%)

1 100, 110 I v

I j

v 90 100 Page 11 of13

COLRSALEM2 Revision 6 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 22 COLR July 2015 100 80

R 0

60 1i) h 0

Po<

ro 5

n E-<

"O

(!)

+

40 0

1:l

(!)

u

(!)

Po<

20 0

FIGURE 2 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER k-11,90%

I c11,9o)I I

\\

UNACCEPTABLE UNACCEPTABLE OPERATION OPERATION

!ACCEPTABLE OPERATION I

\\

j

\\

lc-31,50)1 I (31,5o]

-50

-40

-30

-20

-10 0

10 20 30 40 Flux Difference (% Delta I)

Page 12 of13 50

COLRSALEM2 Revision 6 July 2015 1.2 1.0 g

g 0.8



E-<



µ...  06



0 0.4 z

0.2 0.0 0

PSEG Nuclear LLC SALEM UNIT 2 CYCLE 22 COLR FIGURE 3 K(Z) - NORMALIZED FQ(Z) AS A FUNCTION OF CORE HEIGHT 2

FQ K(Z) 2.40 1.0 2.40 1.0 2.22 0.925 4

6 8

CORE HEIGHT (FEET)

Height (FT) 0.0 6.0 12.0 10 12 Page 13 ofl3

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