ML13280A285

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Cycle 24 Core Operating Limits Report, Revision 1
ML13280A285
Person / Time
Site: North Anna Dominion icon.png
Issue date: 09/30/2013
From: Huber T
Dominion, Dominion Resources Services
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
13-550
Download: ML13280A285 (24)
v • d • e


Text

Dominion Resources Services, Inc.

lnnsbrook Technical Center 5000 Dominion Boulevard, 2SE, Glen Allen, VA 23060 IODominion September 30, 2013 U. S. Nuclear Regulatory Commission Serial No.13-550 Attention: Document Control Desk NLOS /ETS One White Flint North Docket No. 50-338 11555 Rockville Pike License No. NPF-4 Rockville, MD 20852-2738 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)

NORTH ANNA POWER STATION UNIT I CYCLE 24 CORE OPERATING LIMITS REPORT, REVISION 1 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 1.

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 1.

Commitments made in this letter: None U-

Serial No.13-550 Docket No. 50-338 Cycle 24 - Pattern BUS COLR 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.13-550 Docket No. 50-338 ATTACHMENT 1 CORE OPERATING LIMITS REPORT FOR NORTH ANNA UNIT 1 CYCLE 24 PATTERN BUS, REVISION 1 NORTH ANNA POWER STATION VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)

Serial No.13-550 Docket No. 50-338 N 1C24 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 NAH)

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 21

Serial No.13-550 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 21

Serial No.13-550 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

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 Page 3 of 21

Serial No.13-550 Docket No. 50-338 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 Specificationparameter.This is due to the transitionfrom AREVA fuel to Westinghouse fuel which requires the use of different vendor proprietary methodologies to verify the two fuel products meet the applicable regulatory limits.

Page 4 of 21

Serial No.13-550 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 21

Serial No.13-550 Docket No. 50-338 COLR Figure 2.1-1 NORTH ANNA REACTOR CORE SAFETY LIMITS 665 1 660 -

655 62400 psia 645 ______ __

640 2250psi 635 __

S630 __

S625 __ __ __ __

tw 615 E 620

.........l-610 605

> 600 .. . ................... ...

595 590 585 580 575 -

570 0 10 20 30 40 50 60 70 80 90 100 110 120 Percent of RATED THERMAL POWER Page 6 of 21

Serial No.13-550 Docket No. 50-338 3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 SHUTDOWN MARGIN (SDM)

LCO 3.1.1 SDMshallbe_>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/'F, when < 70% RTP, and 0.0 Ak/k/0 F when > 70%

RTP.

The BOC/ARO-MTC shall be _ +0.6 x 10-4 Ak/k/0 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/°F (lower limit).

The MTC surveillance limits are:

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

-4.0 x 10"4 Ak/k/°F [Note 2].

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

-4.7 x 10"4 Ak/k/°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 Ž 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 21

Serial No.13-550 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 21

Serial No.13-550 Docket No. 50-338 COLR Figure 3.1-1 North Anna 1 Cycle 24 Control Rod Bank Insertion Limits 230 220 210 200 190 180 170

'W160

=C- 150

._ 140 o 130 a.

C-5 120 0

I-II Q 110 o 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 21

Serial No.13-550 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)

F A4 (Z) <forP P>O.5 N(Z)

CFQ K(Z)

F*1 (Z) < for P.<0.5 0.5 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 FQM(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. 1for this LCO requiresthe applicationofa cycle dependent non-equilibrium multiplier,N(Z), to the CFQ limit. N(Z) accountsfor power distributiontransients encountered duringnormal operation.Asftunction N(Z) is dependent on the predictedequilibriumFQ(Z) and is sensitive to the axialpower distribution, it is typically generatedfrom the actualEOC burnup distributionthatcan only be obtainedafter the shutdown of the previous cycle. The cycle-specific N(Z) function is presented in COLR Table 3.2-1.

Page 10 of 21

Serial No.13-550 Docket No. 50-338 COLR Table 3.2-1 N1C24 Normal Operation N(Z)

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

Serial No.13-550 Docket No. 50-338 COLR Table 3.2-1 (continued)

N1C24 Normal Operation N(Z)

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

Serial No.13-550 Docket No. 50-338 COLR Figure 3.2-1 K(Z) - Normalized FQ as a Function of Core Height 1.2 1.1 + F 4 + 4 + F 4 F 4 +

6,11.0 1.0 0.9 (12-.925) 0.8 a

IL 0.7 0

LU N

-j 0.6 0

z Z 0.5 0.4 0.3 0.2 0.1 0.0 0 1 2 3 4 5 6 7 8 9 10 11 12 13 CORE HEIGHT (FT)

Page 13 of 21

Serial No.13-550 Docket No. 50-338 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (F NAH)

LCO 3.2.2 FN'A shall be within the limits specified below.

FNAH 1.587{1 + 0.3(1 - P)}

THERMAL POWER RATED 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 COLR Figure 3.2-2.

Page 14 of 21

Serial No.13-550 Docket No. 50-338 COLR Figure 3.2-2 T

North Anna 1 Cycle 24 Axial Flux Difference Limits 120 110 -

(-12, 100) (+6,100) 0 Unacceptable

,Dpera ion Un ccep able 90 - perat ion S80 --

M Ac cepta le Op eratio E 70 a,-

I--

" 60 0 50

(-27, 50) (+20, 50)

a. 40 30 20 10 0

-30 -20 -10 0 10 20 30 Percent Flux Difference (Delta-I)

Page 15 of 2l

Serial No.13-550 Docket No. 50-338 3.3 INSTRUMENTATION 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< ATo {K-K 2 *I- [T-T,]+K3(P-P')-f_ (AW)}

where: AT is measured RCS AT, OF AT0 is the indicated AT at RTP, OF is the Laplace transform operator, sec 1 T is the measured RCS average temperature, OF T' is the nominal Tavg at RTP, <586.8 OF P is the measured pressurizer pressure, psig pr is the nominal RCS operating pressure, _>2235 psig K, < 1.2715 K2 ->0.02174 /OF K 3 > 0.001145 /psig rl, r2 = time constants utilized in the lead-lag controllerfor Tavg T,1 Ž 23.75 sec T 2 -<4.4 sec (1 + l1s)/(] + r2s) = function generatedby the lead-lag controllerfor Tavg dynamic compensation fj(Al) > 0.0291 {- 13. 0 - (qt - qb)} when (qt - qb) < -13.0% RTP 0 when -13.0% RTP < (qt - qb) < +7.0% RTP 0.0251 f{(qt - qb,) - 7.0) 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 16 of 21

Serial No.13-550 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{K4 5 [ 1-3s 3K Z-gK 6 [T-zT]-f 2 (AI)}

+r 3 S where: AT is measured RCS AT, OF.

AT 0 is the indicated AT at RTP, OF.

is the Laplace transform operator, sec-.

T is the measured RCS average temperature, °F.

T' is the nominal Tavg at RTP, < 586.8 OF.

K4*< 1.0865 K5 ->0.0198 / 0 Ffor increasing Tavg K6 > 0.00162/OF when T > T' 0 /OF for decreasing Tavg 0 /OF when T < T' r3= time constant utilized in the rate lag controllerfor Tca,g

-13 9.5 sec r3s/(1 + r3s) = function generatedby the rate lag controllerfor T"g dynamic compensation f 2(AI) = 0, for all Al.

Page 17 of 21

Serial No.13-550 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 18 of 21

Serial No.13-550 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 19 of 21

Serial No.13-550 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.

Page 20 of 21

Serial No.13-550 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 at 200 OF, after xenon decay.

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