ML14014A100

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Cycle 24 Core Operating Limits Report, Revision 2
ML14014A100
Person / Time
Site: North Anna Dominion icon.png
Issue date: 01/07/2014
From: Huber T
Dominion, Dominion Resources Services, Virginia Electric & Power Co (VEPCO)
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
13-643
Download: ML14014A100 (25)
v • d • e


Text

j"Dominion Dominion Resources Services, Inc.

Innsbrook Technical Center 5000 Dominion Boulevard, 2SE, Glen Allen, VA 23060 January 7, 2014 U. S. Nuclear Regulatory Commission Attention: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738 Serial No.

NLOS /ETS Docket No.

License No.13-643 50-338 NPF-4 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)

NORTH ANNA POWER STATION UNIT I CYCLE 24 CORE OPERATING LIMITS REPORT. REVISION 2 Pursuant to North Anna Technical Specification 5.6.5.d, attached is a copy of the Dominion Core Operating Limits Report for North Anna Unit 1 Cycle 24 Pattern BUS, Revision 2.

If you have any questions regarding this submittal, please contact Mr. Thomas Shaub at (804) 273-2763.

Sincerely, T. R. Huber, Director Nuclear Licensing and Operations Support Dominion Resources Services, Inc.

for Virginia Electric and Power Company

Attachment:

1. Core Operating Limits Report for North Anna Unit 1 Cycle 24 - Pattern BUS, Revision 2.

Commitments made in this letter: None

Serial No.13-643 Docket No. 50-338 COLR, North Anna 1 Cycle 24, BUS R2 Page 2 of 2 cc:

U.S. Nuclear Regulatory Commission - Region II Marquis One Tower 245 Peachtree Center Ave., NE, Suite 1200 Atlanta, Georgia 30303-1257 Mr. J. E. Reasor, Jr.

Old Dominion Electric Cooperative Innsbrook Corporate Center 4201 Dominion Blvd.

Suite 300 Glen Allen, Virginia 23060 NRC Senior Resident Inspector North Anna Power Station Dr. V. Sreenivas NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 08 G-9A 11555 Rockville Pike Rockville, Maryland 20852-2738

Serial No.13-643 Docket No. 50-338 ATTACHMENT I CORE OPERATING LIMITS REPORT FOR NORTH ANNA UNIT 1 CYCLE 24 PATTERN BUS, REVISION 2 NORTH ANNA POWER STATION VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)

Serial No.13-643 Docket No. 50-338 N 1 C24 CORE OPERATING LIMITS REPORT INTRODUCTION The Core Operating Limits Report (COLR) for North Anna Unit 1 Cycle 24 has been prepared in accordance with North Anna Technical Specification 5.6.5. The technical specifications affected by this report are listed below:

TS 2.1.1 Reactor Core Safety Limits TS 3.1.1 Shutdown Margin (SDM)

TS 3.1.3 Moderator Temperature Coefficient (MTC)

TS 3.1.4 Rod Group Alignment Limits TS 3.1.5 Shutdown Bank Insertion Limit TS 3.1.6 Control Bank Insertion Limits TS 3.1.9 PHYSICS TESTS Exceptions - Mode 2 TS 3.2.1 Heat Flux Hot Channel Factor TS 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (F NA)

TS 3.2.3 Axial Flux Difference (AFD)

TS 3.3.1 Reactor Trip System (RTS) Instrumentation TS 3.4.1 RCS Pressure, Temperature, and Flow DNB Limits TS 3.5.6 Boron Injection Tank (BIT)

TS 3.9.1 Boron Concentration In addition, a technical requirement (TR) in the NAPS Technical Requirements Manual (TRM) refers to the COLR:

TR 3.1.1 Boration Flow Paths - Operating The analytical methods used to determine the core operating limits are those previously approved by the NRC and discussed in the documents listed in the References Section.

Cycle-specific values are presented in bold. Text in italics is provided for information only.

Page 1 of 22

Serial No.13-643 Docket No. 50-338 REFERENCES

1. VEP-FRD-42, Rev. 2.1-A, "Reload Nuclear Design Methodology," August 2003.

Methodology for:

TS 3.1.1 - Shutdown Margin, TS 3.1.3 - Moderator Temperature Coefficient, TS.3.1.4 - Rod Group Alignment Limits TS 3.1.5 - Shutdown Bank Insertion Limit, TS 3.1.6 - Control Bank Insertion Limits, TS 3.1.9 - Physics Tests Exceptions - Mode 2, TS 3.2.1 - Heat Flux Hot Channel Factor, TS 3.2.2 - Nuclear Enthalpy Rise Hot Channel Factor TS 3.5.6 - Boron Injection Tank (BIT) and TS 3.9.1 - Boron Concentration

2. Plant-specific adaptation of WCAP-16009-P-A, "Realistic Large Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM),"

as approved by NRC Safety Evaluation Report dated February 29, 2012.

Methodology for: TS 3.2.1 - Heat Flux Hot Channel Factor

3. WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," August 1985.

Methodology for: TS 3.2.1 - Heat Flux Hot Channel Factor

4. WCAP-10079-P-A, "NOTRUMP, A Nodal Transient Small Break and General Network Code," August 1985.

Methodology for: TS 3.2.1 - Heat Flux Hot Channel Factor

5. WCAP-12610-P-A, "VANTAGE+ FUEL ASSEMBLY - REFERENCE CORE REPORT,"

April 1995.

Methodology for:

TS 2.1.1 - Reactor Core Safety Limits TS 3.2.1 - Heat Flux Hot Channel Factor

6. VEP-NE-2, Rev. 0-A, Statistical DNBR Evaluation Methodology, June 1987.

Methodology for:

TS 3.2.2 - Nuclear Enthalpy Rise Hot Channel Factor and TS 3.4.1 - RCS Pressure, Temperature and Flow DNB Limits Page 2 of 22

Serial No.13-643 Docket No. 50-338

7. VEP-NE-1, Rev. 0.1-A, Relaxed Power Distribution Control Methodology and Associated FQ Surveillance Technical Specifications, August 2003.

Methodology for:

TS 3.2.1 - Heat Flux Hot Channel Factor and TS 3.2.3 - Axial Flux Difference

8.

WCAP-8745-P-A, Design Bases for the Thermal Overpower AT and Thermal Overtemperature AT Trip Functions, September 1986.

Methodology for:

TS 2.1.1 - Reactor Core Safety Limits and TS 3.3.1 - Reactor Trip System Instrumentation

9. WCAP-14483-A, Generic Methodology for Expanded Core Operating Limits Report, January 1999.

Methodology for:

TS 2.1.1 - Reactor Core Safety Limits, TS 3.1.1 - Shutdown Margin, TS 3.1.4 - Rod Group Alignment Limits TS 3.1.9 - Physics Tests Exceptions - Mode 2 TS 3.3.1 - Reactor Trip System Instrumentation, TS 3.4.1 - RCS Pressure, Temperature, and Flow DNB Limits TS 3.5.6 - Boron Injection Tank (BIT) and TS 3.9.1 - Boron Concentration

10. BAW-10227P-A, Rev. 0, "Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel," February 2000.

Methodology for:

TS 2.1.1 - Reactor Core Safety Limits and TS 3.2.1 - Heat Flux Hot Channel Factor

11. EMF-2103 (P) (A), Rev. 0, "Realistic Large Break LOCA Methodology for Pressurized Water Reactors," April 2003.

Methodology for: TS 3.2.1 - Heat Flux Hot Channel Factor

12. EMF-96-029 (P) (A), Rev. 0, "Reactor Analysis System for PWRs," January 1997.

Methodology for: TS 3.2.1 - Heat Flux Hot Channel Factor Page 3 of 22

Serial No.13-643 Docket No. 50-338

13. BAW-10168P-A, Rev. 3, "RSG LOCA - BWNT Loss-of-Coolant Accident Evaluation Model for Recirculating Steam Generator Plants," December 1996. Volume II only (SBLOCA models).

Methodology for: TS 3.2.1 - Heat Flux Hot Channel Factor

14. DOM-NAF-2, Rev. 0.2-P-A, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code," including Appendix A, "Qualification of the F-ANP BWU CHF Correlations in the Dominion VIPRE-D Computer Code," and Appendix C, "Qualification of the Westinghouse WRB-2M CHF Correlation in the Dominion VIPRE-D Computer Code,"

August 2010.

Methodology for:

TS 3.2.2 - Nuclear Enthalpy Rise Hot Channel Factor and TS 3.4.1 - RCS Pressure, Temperature and Flow DNB Limits

15. WCAP-12610-P-A and CENPD-404.-P-A, Addendum 1 -A, "Optimized ZIRLOTM, July 2006.

Methodology for:

TS 2.1.1 - Reactor Core Safety Limits and TS 3.2.1 - Heat Flux Hot Channel Factor Note:

In some instances, the North Anna COLR lists multiple methodologies that are used to verify a single Technical Specification parameter. This is due to the transition from AREVA fuel to Westinghouse fuel which requires the use of different vendor proprietary methodologies to verify the two fuiel products meet the applicable regulatory limits.

Page 4 of 22

Serial No.13-643 Docket No. 50-338 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs In MODES 1 and 2, the combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the limits specified in COLR Figure 2.1-1; and the following SLs shall not be exceeded.

2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained greater than or equal to the 95/95 DNBR criterion for the DNB correlations and methodologies specified in the References Section.

2.1.1.2 The peak fuel centerline temperature shall be maintained < 5080'F, decreasing by 58'F per 10,000 MWD/MTU of burnup, for Westinghouse fuel and < 5173°F, decreasing by 65'F per 10,000 MWD/MTU of burnup, for AREVA fuel.

Page 5 of 22

Serial No.13-643 Docket No. 50-338 COLR Figure 2.1-1 NORTH ANNA REACTOR CORE SAFETY LIMITS 4-E ba (U

w, 665 660 655 650 645 640 635 630 625 620 615 610 605 600 595 590 585 580 575 570 0

10 20 30 40 50 60 70 80 Percent of RATED THERMAL POWER 90 100 110 120 Page 6 of 22

Serial No.13-643 Docket No. 50-338 3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 SHUTDOWN MARGIN (SDM)

LCO 3.1.1 SDM shall be__ 1.77 % Ak/k.

3.1.3 Moderator Temperature Coefficient (MTC)

LCO 3.1.3 The MTC shall be maintained within the limits specified below. The upper limit of MTC is +0.6 x 10-4 Ak/k/IF, when < 70% RTP, and 0.0 Ak/k/OF when > 70%

RTP.

The BOC/ARO-MTC shall be _ +0.6 x 10-4 Ak/k/F (upper limit), when < 70%

RTP, and _0.0 Ak/k/0 F when > 70% RTP.

The EOC/ARO/RTP-MTC shall be less negative than -5.0 x 10-4 Ak/k/OF (lower limit).

The MTC surveillance limits are:

The 300 ppm/ARO/RTP-MTC should be less negative than or equal to

-4.0 x 104 Ak/k/0 F [Note 2].

The 60 ppm/ARO/RTP-MTC should be less negative than or equal to

-4.7 x 104 AkWk/°F [Note 3].

SR 3.1.3.2 Verify MTC is within -5.0 x 10-4 Ak'k/°F (lower limit).

Note 2: If the MTC is more negative than -4.0 x 10-4 Ak/k/°F, SR 3.1.3.2 shall be repeated once per 14 EFPD during the remainder of the fuel cycle.

Note 3: SR 3.1.3.2 need not be repeated if the MTC measured at the equivalent of equilibrium RTP-ARO boron concentration of _ 60 ppm is less negative than -4.7 x 10-4 Ak/k/]F.

3.1.4 Rod Group Alignment Limits Required Action A. 1.1 Verify SDM to be > 1.77 % Ak/k.

Required Action B. 1.1 Verify SDM to be _> 1.77 % Ak/k.

Required Action D. 1.1 Verify SDM to be > 1.77 % Ak/k.

Page 7 of 22

Serial No.13-643 Docket No. 50-338 3.1.5 Shutdown Bank Insertion Limits LCO 3.1.5 Each shutdown bank shall be withdrawn to at least 225 steps.

Required Action A. 1.1 Verify SDM to be > 1.77 % Ak/k.

Required Action B.1 Verify SDM to be > 1.77 % Ak/k.

SR 3.1.5.1 Verify each shutdown bank is withdrawn to at least 225 steps.

3.1.6 Control Bank Insertion Limits LCO 3.1.6 Control banks shall be limited in physical insertion as shown in COLR Figure 3.1-1. Sequence of withdrawal shall be A, B, C and D, in that order; and the overlap limit during withdrawal shall be 97 steps.

Required Action A. 1.1 Verify SDM to be > 1.77 % Ak/k.

Required Action B. 1.1 Verify SDM to be > 1.77 % Ak/k.

Required Action C. 1 Verify SDM to be > 1.77 % Ak/k.

SR 3.1.6.1 Verify estimated critical control bank position is within the insertion limits specified in COLR Figure 3.1-1.

SR 3.1.6.2 Verify each control bank is within the insertion limits specified in COLR Figure 3.1-1.

SR 3.1.6.3 Verify each control bank not fully withdrawn from the core is within the sequence and overlap limits specified in LCO 3.1.6 above.

3.1.9 PHYSICS TESTS Exceptions - MODE 2 LCO 3.1.9.b SDM is _ 1.77 % Ak/k.

SR 3.1.9.4 Verify SDM to be __ 1.77 % Ak/k.

Page 8 of 22

Serial No.13-643 Docket No. 50-338 COLR Figure 3.1-1 North Anna 1 Cycle 24 Control Rod Bank Insertion Limits 0.

1, 4.

In 0

0W-0.

0~

230 220 210 200 190 180 170 160 150 140 130 120 110 100 90 80 70 60 50 40 30 20 10 0

0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 Fraction of Rated Thermal Power Page 9 of 22

Serial No.13-643 Docket No. 50-338 3.2 POWER DISTRIBUTION LIMITS 3.2.1 Heat Flux Hot Channel Factor (FQ(Z))

LCO 3.2.1 FQ(Z), as approximated by FQM(Z), shall be within the limits specified below.

CFQ = 2.32 The Measured Heat Flux Hot Channel Factor, FQM(Z), shall be limited by the following relationships:

CFQ K(Z)

P N(Z))

for e>0.5 P

N(Z)

CFQ K(Z)

F0 (Z) -

for P0.5 Q0.5 N( Z)

THERMAL POWER where:

P = RATED THERMAL POWER ; and K(Z) is provided in COLR Figure 3.2-1 N(Z) is a cycle-specific non-equilibrium multiplier on FQ (Z) to account for power distribution transients during normal operation, provided in COLR Table 3.2-1.

The discussion in the Bases Section B 3.2.1 for this LCO requires the application of a cycle dependent non-equilibrium multiplier, N(Z), to the CFQ limit. N(Z) accounts for power distribution transients encountered during normal operation. As function N(Z) is dependent on the predicted equilibrium FQ(Z) and is sensitive to the axial power distribution, it is typically generated from the actual EOC burnup distribution that can only be obtained after the shutdown of the previous cycle.

The cycle-specific N(Z) function is presented in COLR Table 3.2-1.

Page 10 of 22

Serial No.13-643 Docket No. 50-338 COLR Table 3.2-1 N1C24 Normal Operation N(Z)

NODE HEIGHT (FEET) 10 10.2 11 10.0 12 9.8 13 9.6 14 9.4 15 9.2 16 9.0 17 8.8 18 8.6 19 8.4 20 8.2 21 8.0 22 7.8 23 7.6 24 7.4 25 7.2 26 7.0 27 6.8 28 6.6 29 6.4 30 6.2 31 6.0 32 5.8 33 5.6 34 5.4 35 5.2 36 5.0 37 4.8 38 4.6 39 4.4 40 4.2 41 4.0 42 3.8 43 3.6 44 3.4 45 3.2 46 3.0 47 2.8 48 2.6 49 2.4 50 2.2 51 2.0 52 1.8 0 to 1000 MWD/MTU 1.128 1.128 1.133 1.140 1.143 1.144 1.150 1.155 1.158 1.159 1.162 1.162 1.162 1.160 1.157 1.152 1.147 1.145 1.143 1.134 1.123 1.118 1.113 1.100 1.092 1.092 1.096 1.099 1.101 1.102 1.102 1.104 1.113 1.127 1.137 1.146 1.157 1.170 1.182 1.193 1.203 1.213 1.222 1000 to 3000 MWD/MTU 1.139 1.147 1.155 1.161 1.166 1.168 1.173 1.176 1.177 1.175 1.173 1.169 1.164 1.160 1.157 1.152 1.147 1.145 1.143 1.135 1.123 1.116 1.108 1.091 1.079 1.077 1.080 1.082 1.086 1.091 1.097 1.104 1.112 1.121 1.128 1.135 1.144 1.154 1.163 1.172 1.180 1.187 1.195 3000 to 5000 MWDIMTU 1.143 1.148 1.154 1.161 1.165 1.168 1.173 1.176 1.176 1.176 1.176 1.175 1.175 1.174 1.171 1.167 1.163 1.161 1.158 1.150 1.137 1.131 1.121 1.098 1.081 1.077 1.079 1.079 1.082 1.087 1.091 1.096 1.104 1.114 1.122 1.129 1.135 1.142 1.153 1.164 1.174 1.183 1.192 5000 to 7000 MWD/MTU 1.144 1.148 1.152 1.156 1.156 1.158 1.170 1.181 1.186 1.185 1.186 1.185 1.185 1.180 1.173 1.167 1.163 1.162 1.162 1.158 1.152 1.149 1.142 1.126 1.112 1.107 1.109 1.111 1.114 1.115 1.115 1.115 1.113 1.115 1.124 1.137 1.147 1.155 1.158 1.166 1.183 1.195 1.197 7000 to 9000 MWD/MTU 1.121 1.128 1.137 1.148 1.153 1.157 1.170 1.181 1.186 1.185 1.185 1.184 1.183 1.184 1.187 1.189 1.189 1.190 1.188 1.185 1.178 1.175 1.167 1.148 1.131 1.122 1.120 1.121 1.123 1.121 1.120 1.120 1.116 1.115 1.123 1.137 1.147 1.155 1.159 1.167 1.183 1.195 1.197 These decks are generated for normal operation flux maps that are typically taken at full power ARO. Additional N(z) decks may be generated, if necessary, consistent with the methodology described in the RPDC topical (Reference 7). EOR is defined as Hot Full Power End of Reactivity.

Page 11 of 22

Serial No.13-643 Docket No. 50-338 COLR Table 3.2-1 (continued)

N1C24 Normal Operation N(Z)

NODE HEIGHT (FEET) 10 10.2 11 10.0 12 9.8 13 9.6 14 9.4 15 9.2 16 9.0 17 8.8 18 8.6 19 8.4 20 8.2 21 8.0 22 7.8 23 7.6 24 7.4 25 7.2 26 7.0 27 6.8 28 6.6 29 6.4 30 6.2 31 6.0 32 5.8 33 5.6 34 5.4 35 5.2 36 5.0 37 4.8 38 4.6 39 4.4 40 4.2 41 4.0 42 3.8 43 3.6 44 3.4 45 3.2 46 3.0 47 2.8 48 2.6 49 2.4 50 2.2 51 2.0 52 1.8 9000 to 11000 MWDIMTU 1.119 1.118 1.123 1.133 1.138 1.143 1.156 1.168 1.173 1.175 1.180 1.182 1.182 1.183 1.187 1.189 1.189 1.190 1.188 1.185 1.178 1.175 1.167 1.147 1.132 1.130 1.131 1.130 1.127 1.123 1.119 1.119 1.118 1.122 1.130 1.144 1.156 1.168 1.176 1.185 1.199 1.208 1.209 11000 to 13000 MWD/MTU 1.120 1.118 1.121 1.127 1.128 1.132 1.144 1.158 1.164 1.170 1.182 1.189 1.192 1.193 1.197 1.198 1.198 1.197 1.195 1.196 1.197 1.198 1.194 1.184 1.173 1.168 1.159 1.145 1.131 1.122 1.123 1.133 1.142 1.150 1.156 1.161 1.163 1.168 1.174 1.185 1.199 1.208 1.209 13000 to 15000 MWD/MTU 1.112 1.111 1.108 1.109 1.111 1.120 1.139 1.159 1.165 1.170 1.182 1.189 1.192 1.193 1.197 1.198 1.198 1.197 1.195 1.196 1.197 1.198 1.194 1.184 1.173 1.168 1.163 1.155 1.145 1.134 1.130 1.134 1.141 1.150 1.156 1.161 1.165 1.167 1.169 1.170 1.170 1.171 1.173 15000 to 17000 MWD/MTU 1.119 1.119 1.118 1.117 1.112 1.113 1.123 1.138 1.146 1.157 1.175 1.188 1.193 1.200 1.211 1.217 1.219 1.221 1.222 1.222 1.219 1.218 1.211 1.196 1.181 1.174 1.169 1.161 1.154 1.148 1.144 1.140 1.134 1.128 1.122 1.122 1.130 1.145 1.153 1.163 1.175 1.186 1.194 17000 to EOR MWDIMTU 1.119 1.119 1.118 1.117 1.112 1.113 1.123 1.139 1.148 1.160 1.179 1.194 1.200 1.207 1.218 1.226 1.228 1.231 1.231 1.232 1.229 1.229 1.223 1.209 1.194 1.188 1.182 1.172 1.164 1.159 1.154 1.148 1.135 1.124 1.122 1.133 1.145 1.160 1.170 1.181 1.196 1.209 1.220 These decks are generated for normal operation flux maps that are typically taken at full power ARO. Additional N(z) decks may be generated, if necessary, consistent with the methodology described in the RPDC topical (Reference 7). EOR is defined as Hot Full Power End of Reactivity.

Page 12 of 22

Serial No.13-643 Docket No. 50-338 1.2 1.1 1.0 0.9 0.8 N

0 0.7 M.'

N

-- I 40.6 0Z

- 0.5 N

0.4 0.3 0.2 0.1 0.0 COLR Figure 3.2-1 K(Z) - Normalized FQ as a Function of Core Height 6,1.0 (12.925) 0 1

2 3

4 5

6 7

8 9

10 11 12 13 CORE HEIGHT (FT)

Page 13 of 22

Serial No.13-643 Docket No. 50-338 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (FNAH)

LCO 3.2.2 FNAH shall be within the limits specified below.

FNA

_< 1.587(1 + 0.3(1 - P)}

THERMAL POWER where: P RA TED THERMAL POWER SR 3.2.2.1 Verify FNAH is within limits specified above.

3.2.3 AXIAL FLUX DIFFERENCE (AFD)

LCO 3.2.3 The AFD in % flux difference units shall be maintained within the limits specified in the applicable COLR Figure (3.2-2-A or 3.2-2-B).

Page 14 of 22

Serial No.13-643 Docket No. 50-338 COLR Figure 3.2-2-A Applicable Burnup: BOC* to 5000 MWD/MTU North Anna 1 Cycle 24 Axial Flux Difference Limits 120 110 100 0

C-ra, 0

90 80 70 60 50 40 30 20 10 0

-30

-20

-10 0

10 20 Percent Flux Difference (Delta-I) 30

  • Figure 3.2.-2-A was implemented at a core bumup of approximately 2000 MWD/MTU.

Page 15 of 22

Serial No.13-643 Docket No. 50-338 COLR Figure 3.2-2-B Applicable Burnup: 5000 MWD/MTU to EOC North Anna 1 Cycle 24 Axial Flux Difference Limits 120 110 100 0

E-0 90 80 70 60 50 40 30 20 10 0

(-12,100)

(+6,10))

Unacceptabl/

eUnateiont Unacceptable Operation Acc ptable Operation

(-27,50)

(+20,50)

-30

-20

-10 0

10 20 30 Percent Flux Difference (Delta-I)

Page 16 of 22

Serial No.13-643 Docket No. 50-338 3.3.1 Reactor Trip System (RTS) Instrumentation TS Table 3.3.1-1 Note 1: Overtemperature AT The Overtemperature AT Function Allowable Value shall not exceed the following nominal trip setpoint by more than 2% of AT span, with the numerical values of the parameters as specified below.

AT<~C

+~

{K K)[TT,]+K3 (P-p,)_fl(A)}

where: AT AT 0 S

T T'

P P

1 is measured RCS AT, OF is the indicated AT at RTP, OF is the Laplace transform operator, sec1 is the measured RCS average temperature, OF is the nominal Tavg at RTP, < 586.8 OF is the measured pressurizer pressure, psig is the nominal RCS operating pressure, _> 2235 psig Ki < 1.2715 K 2 >- 0.02174 /OF K3 -> 0.001145 /psig r/, r2 = time constants utilized in the lead-lag controller for Tavg Ti ! 23.75 sec T2 - 4.4 sec (1 + rls)/(J + r2s) = function generated by the lead-lag controller for Tag dynamic compensation f1 (AI) > 0.0291 {- 13.0 - (Cit - qb)}

0 0.0251 {(qt - qb) - 7.0}

when (qt - qb) < -13.0% RTP when -13.0% RTP < (qt - qb) - +7.0% RTP when (qt-qb) > +7.0% RTP Where qt and qb are percent RTP in the upper and lower halves of the core, respectively, and qt + qb is the total THERMAL POWER in percent RTP.

Page 17 of 22

Serial No.13-643 Docket No. 50-338 TS Table 3.3.1-1 Note 2: Overpower AT The Overpower AT Function Allowable Value shall not exceed the following nominal trip setpoint by more than 2% of AT span, with the numerical values of the parameters as specified below.

AT<ATo {K 4a-K 5 [ V3S[]T-K 6[T-T']-f 2 (AI)}

where: AT ATo T

T' is measured RCS AT, °F.

is the indicated AT at RTP, OF.

is the Laplace transform operator, secl.

is the measured RCS average temperature, OF.

is the nominal Tavg at RTP, <586.8 OF.

K4 :5 1.0865 K5 >0.0198 /OF for increasing Tavg 0 /OF for decreasing Tavg K6 >0.00162 0 /OF

/OF when T > T' when T < T' r3= time constant utilized in the rate lag controller for Tag T3 Ž 9.5 sec r3s/(1 + r3s) = function generated by the rate lag controller for T*,g dynamic compensation f2(AI) = 0, for all Al.

Page 18 of 22

Serial No.13-643 Docket No. 50-338 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4,1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits LCO 3.4.1 RCS DNB parameters for pressurizer pressure, RCS average temperature, and RCS total flow rate shall be within the limits specified below:

a. Pressurizer pressure is greater than or equal to 2205 psig;
b. RCS average temperature is less than or equal to 591 OF; and
c. RCS total flow rate is greater than or equal to 295,000 gpm.

SR 3.4.1.1 Verify pressurizer pressure is greater than or equal to 2205 psig.

SR 3.4.1.2 Verify RCS average temperature is less than or equal to 591 OF.

SR 3.4.1.3 Verify RCS total flow rate is greater than or equal to 295,000 gpm.

SR 3.4.1.4 NOTE-------------------------

Not required to be performed until 30 days after > 90% RTP.

Verify by precision heat balance that RCS total flow rate is _> 295,000 gpm.

Page 19 of 22

Serial No.13-643 Docket No. 50-338 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.6 Boron Injection Tank (BIT)

Required Action B.2 Borate to a SDM _> 1.77 % Ak/k at 200 OF.

Page 20 of 22

Serial No.13-643 Docket No. 50-338 3.9 REFUELING OPERATIONS 3.9.1 Boron Concentration LCO 3.9.1 Boron concentrations of the Reactor Coolant System (RCS), the refueling canal, and the refueling cavity shall be maintained > 2600 ppm.

SR 3.9.1.1 Verify boron concentration is within the limit specified above.

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Serial No.13-643 Docket No. 50-338 NAPS TECHNICAL REQUIREMENTS MANUAL TRM 3.1 REACTIVITY CONTROL SYSTEMS TR 3.1.1 Boration Flow Paths - Operating Required Action D.2 Borate to a SHUTDOWN MARGIN >_ 1.77 % Ak/k at 200 IF, after xenon decay.

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