LR-N14-0113, Core Operating Limits Report- Cycle 21

Core Operating Limits Report- Cycle 21

ML14120A106
Person / Time
Site:	Salem
Issue date:	04/28/2014
From:	Jamila Perry Public Service Enterprise Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LR-N14-0113
Download: ML14120A106 (16)

Text

PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, NJ 08038-0236 LR-N14-0113 APR 2 8 2014 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 21 PSf:G Nuclmr ILC Technical Specification 6.9.1.9 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 21 in Attachment 1 to this letter.

There are no commitments contained in this letter. Should you have any questions regarding this submittal, please contact David Lafleur at (856) 339-1754.

Sincerely, cf£r.f1 John F. Perry )

Site Vice Preside(alem Attachments (1)

Document Control Desk Page 2 LR-N14-0113 APR 2 8 Z014 cc:

Mr. W. Dean, Administrator, Region 1, NRC Mr. J. Lamb, Licensing Project Manager - Salem, NRC Mr. P. Finney, USNRC Senior Resident Inspector, Salem Mr. P. McKenna, Resident Inspector, Salem Mr. P. Mulligan, Manager IV, NJBNE Technical Specification 6.9.1.9 Mr. L. Marabella, Corporate Commitment Tracking Coordinator Mr. T. Cachaza, Salem Commitment Tracking Coordinator Mr. D. Lafleur, Senior Regulatory Assurance Engineer Core Operating Limits Report for Salem Unit 2, Cycle 21

COLR SALEM 2 Revision 5 December 2013 Core Operating Limits Report for Salem Unit 2, Cycle 21 Page 1 of 13

COLRSALEM2 Revision S December 2013 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 21 COLR TABLE OF CONTENTS Section Title Table of Contents List ofFigures Core Operating Limits Report Operating Limits Moderator Temperature Coefficient (Specification 3.1.1.3)

Control Rod Insertion Limits (Spec³cation 3.1.3.5)

Axial Flux Difference (Specification 3.2.1)

Heat Flux Hot Channel Factor-FQ(z) (Specification 3.2.2 )

Nuclear Enthalpy Rise Hot Channel Factor FN m (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 Page2 of13

COLRSALEM2 Revision 5 December 2013 Figure Number 1

2 3

PSEG Nuclear LLC SALEM UNIT 2 CYCLE 21 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 ofl3

COLRSALEM2 Revision S December 2013 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 21 COLR 1.0 CORE OPERATING LIMITS REPORT This Core Operating Limits Report (COLR) for Salem Unit 2 Cycle 21 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 - FNH 3.9.1 Boron Concentration Page4 ofl3

COLRSALEM2 Revision 5 December 2013 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 21 COLR 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 0.1.klk/°F.

The EOL/ARO/RTP-MTC shall be less negative than or equal to -4.2xl04

.1-k/k/oF.

2.1.2 The MTC Surveillance limit is:

PageS ofl3 The 300 ppm/ARO/RTP-MTC should be less negative than or equal to -3.5xl0"4.1.klk/°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

COLRSALEM2 Revision 5 December 2013 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 21 COLR 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]

FcfTP Fg(Z) :::;

• K(Z) for P > 0. 5 p

FrfTP Fg(Z) :::;

• K(Z) for P 50.5

0.5 where

P THERMAL POWER RATED THERMAL POWER 2.4.1 FgKI'P

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

2.4.3 F,:f

= FxyRTP {1.0 + PFxy(J.O -P)}

where: from BOL to 12000 MWD/MTU F RTP

• =

xy 2.10 for unrodded upper core planes 1 through 6 1.87 for unrodded upper core planes 7 through 8 1.74 for unrodded upper core planes 9 through 31 1.84 for unrodded lower core planes 32 through 53 1.95 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 ofl3

COLRSALEM 2 Revision 5 December 2013 PSEG Nuclear LLC SALEM UNJT 2 CYCLE 21 COLR where: from 12000 MWD/MTU to EOL F

RTF

=

xy 2.10 for umodded upper core planes 1 through 6 1.88 for umodded upper core planes 7 through 8 1.77 for umodded upper core planes 9 through 31 1.80 for umodded lower core planes 32 through 53 1.82 for umodded lower core planes 54 through 55 1.96 for umodded 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, Upq, to be applied to the Heat Flux Hot Channel Factor Fq(z) shall be calculated by the following formula:

UFQ =(1.0+

UQ l.ue 100.0) where:

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

De =Engineering tmcertainty factor.

= 1.03 Note: Upq= 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 Cha1111el Factor Fq(z) shall be calculated by the following formula:

UFQ=Uqu*Ue where:

Uqu =Base Fq measurement uncertainty.

= 1.05 De = Engineering uncertainty factor.

= 1.03 Page 7 ofl3

COLRSALEM2 Revision 5 December 2013 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 21 COLR 2.5 Nuclear Enthalpy Rise Hot Channel Factor - FN fui (Specification 3.2.3) where:

P THERMAL POWER RATED THERMAL POWER 1.65 2.5.2 PF LJH

=

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, FNm, shall be the greater of 1.04 or as calculated by the following formula:

U o

utili Ftili -1.

+

100.0 where:

VIlli= Uncertainty for enthalpy rise hot channel factor as defmed 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, UFm, to be applied to the Nuclear Enthalpy Rise Hot Channel Factor FN.iH shall be calculated by the following formula:

where:

UFMlin =Base FIll measurement tmcertainty.

= 1.04 Page 8 of13

COLRSALEM2 Revision 5 December 2013 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 21 COLR 2.6 Boron Concentration (Specification 3.9.1)

A Mode 6 boron concentration, maintained at or above 2127 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% k/k uncertainty added.

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

k/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 Cham1el Factor. Approved by Safety Evaluation dated November 13, 1986.

Page 9 of 13

COLRSALEM2 Revision 5 December 2013 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 21 COLR 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: Safety 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, Addendiun 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 No. 285, License No.

DPR-75, Docket No. 50-311.

Page 10 of13

COLRSALEM2 RevisionS December 2013 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 21 COLR FIGURE 1 ROD BANK INSERTION LIMITS VS. THERMAL POWER 240 220 200 180 I 160 "0

K

g_ 140

§_

z 0

i= 120 rn 0

c.

X:

L 100

.J M

z 8 80 60 40 20 0

/ 117.5, 228 I

/

I

'/ 70.8, 228 I

/j BANKB I v

/ v N0, 186 l

v I

j v

Vj BANKC I v I

v

/

I I

v v

I

/ (( BANK D I

v v

I ll v

ro. 58 l I

I v v

129,0 v y

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

11oo,17o 1 I

v I

v 90 100 Page 11 of13

COLRSALEM2 RevisionS December 2013 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 21 COLR 1-<

L 0

p..

Cil J

"i:l

(!)

J 4-l 0

1:1

(!)

0 p..

100 80 60 40 20 0

FIGURE2 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER lc-11,90' 1 (11,90)1 I

\\

UNACCEPTABLE UNACCEPTABLE OPERATION OPERATION

!ACCEPTABLE OPERATION v

\\

j I'

lc-31,50)1 1(31,50)1

-50

-40

-30

-20

-10 0

10 20 30 40 Flux Difference (% Delta I)

Page 12 of 13 50

COLRSALEM2 Revision 5 December 2013 1.2 1.0

á 0.8 I

E-<

d
06 I

0 0.4 z

0.2 0.0 0

PSEG Nuclear LLC SALEM UNIT 2 CYCLE 21 COLR FIGURE3 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 of13