ML12132A087

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Cycle 23 Core Operating Limits Report
ML12132A087
Person / Time
Site: North Anna Dominion icon.png
Issue date: 05/04/2012
From: Huber T
Dominion Resources Services, Dominion
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
12-318
Download: ML12132A087 (24)
v • d • e


Text

Dominion Resources Services, Inc.

Innsbrook Technical Center 5000 Dominion Boulevard, 2SE, Glen Allen, VA 23060

  • JDomi n ioill May 4, 2012 U. S. Nuclear Regulatory Commission Serial No.12-318 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 1 CYCLE 23 CORE OPERATING LIMITS REPORT 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 23, Pattern UNK, 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:

Core Operating Limits Report for North Anna Unit 1 Cycle 23 Pattern UNK, Revision 1 Commitments made in this letter: None

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Serial No.12-318 Docket No. 50-338 Cycle 23 Pattern UNK 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 Ms. K. R. Cotton 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.12-318 Docket No. 50-338 ATTACHMENT CORE OPERATING LIMITS REPORT FOR NORTH ANNA UNIT 1 CYCLE 23 PATTERN UNK, REVISION I NORTH ANNA POWER STATION VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)

Serial No.12-318 Docket No. 50-338 N1C23 CORE OPERATING LIMITS REPORT INTRODUCTION The Core Operating Limits Report (COLR) for North Anna Unit 1 Cycle 23 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 (FNAH)

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.12-318 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.12-318 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.12-318 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 ZIRLO TM, July 2006.

Methodology for:

TS 2.1.1 - Reactor Core Safety Limits and TS 3.2.1 - Heat Flux Hot Channel Factor Page 4 of 21

Serial No.12-318 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 580 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.12-318 Docket No. 50-338 COLR Figure 2.1-1 NORTH ANNA REACTOR CORE SAFETY LIMITS 665 660 655 650 645 640 635 630 U-625 E 620 615 (A

610 fA 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.12-318 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/°F 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/kI 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 104 Ak/k/0 F [Note 2].

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

-4.7 x 104 Ak/k/0 F [Note 3].

SR 3.1.3.2 Verify MTC is within -5.0 x 104 Ak/k/OF (lower limit).

Note 2: If the MTC is more negative than -4.0.x 10-4 A*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 21

Serial No.12-318 Docket No. 50-338 3.1.5 Shutdown Bank Insertion Limits LCO 3.1.5 Each shutdown bank shall be withdrawn to at least 230 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 230 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 102 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.12-318

- Docket No. 50-338 COLR Figure 3.1-1

, North Anna 1 Cycle 23 Control Rod Bank Insertion Limits 230 0.5'49, 230' 220 210 200 1i 194/

190 C-BANK 180 3:170 0.

'V160 _ Fully w/d position = 230 in

  • ý150

'140 0130 120 0 0.0, 118 0110 r0100 90 80 70 60 50 40 30 20 10 0.048, 0 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.12-318 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 P>0.5 CFQ K(Z) 0.5 N(Z) for P_ 0.5 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 3.2.1 for this LCO requires the applicationof a cycle dependent non-equilibrium multiplier, N(Z), to the measured peakingfactor, FQM(Z), before comparing it to the 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 generatedfrom the actual EOC burnup distribution that can only be obtained after the shutdown of the previous cycle. The cycle-specific N(Z)function ispresented in COLR Table 3.2-1.

Page 10 of 21

Serial No.12-318 Docket No. 50-338 COLR Table 3.2-1 N1C23 Normal Operation N(Z)

NODE HEIGHT 0 to 1000 1000 to 3000 3000 to 5000 5000 to 7000 7000 to 9000 9000 to 11000 (FEET) MWD/MTU MWD/MTU MWD/MTU MWD/MTU MWD/MTU MWD/MTU 10 10.2 1.080 1.080 1.100 '1.108 1.120 1.131 11 10.0 1.079 1.080 1.105 1.116 1.119 1.129 12 9.8 1.077 1.082 1.111 1.124 1.123 1.132 13 9.6 1.076 1.087 1.120 1.133 1.132 1.140 14 9.4 1.074 1.090 1.124 1.137 1.136 1.144 15 9.2 1.078 1.095 1.128 1.142 1.143 1.151 16 9.0 1.088 1.104 1.138 1.155 1.160 1.166 17 8.8 1.100 1.113 1.147 1.166 1.177 1.181 18 8.6 1.105 1.116 1.151 1.171 1.183 1.186 19 8.4 1.108 1.117 1.151 1.173 1.185 1.187 20 8.2 1.113 1.118 1.153 1.178 1.190 1.193 21 8.0 1.115 1.118 1.153 1.181 1.192 1.196 22 7.8 1.116 1.119 1.153 1.182 1.193 1.197 23 7.6 1.115 1.117 1.150 1.179 1.190 1.196 24 7.4 1.112 1.114 1.145 1.174 1.187 1.195 25 7.2 1.111 1.112 1.141 1.169 1.185 1.196 26 7.0 1.110 1.110 1.138 1.165 1.183 1.196 27 6.8 1.107 1.107 1.135 1.160 1.180 1.196 28 6.6 1.103 1.103 1.130 1.153 1.176 1.195 29 6.4 1.101 1.097 1.122 1.141 1.171 1.191 30 6.2 1.101 1.092 1.115 1.129 1.164 1.184 31 6.0 1.101 1.087 1.113 1.126 1.162 1.182 32 5.8 1.102 1.087 1.113 1.126 1.157 1.173 33 5.6 1.098 1.088 1.103 1.116 1.142 1.154 34 5.4 1.096 1.094 1.100 1.110 1.129 1.138 35 5.2 1.095 1.101 1.107 1.111 1.123 1.133 36 5.0 1.102 1.112 1.118 1.118 1.124 1.133 37 4.8 1.114 1.123 1.126 1.126 1.128 1.131 38 4.6 1.127 1.133 1.134 1.134 1.132 1.131 39 4.4 1.136 1.141 1.141 1.139 1.134 1.133 40 4.2 1.147 1.148 1.148 1.142 1.134 1.134 41 4.0 1.158 1.155 1.155 1.144 1.134 1.135 42 3.8 1.170 1.165 1.164 1.148 1.132 1.130 43 3.6 1.180 1.176 1.175 1.154 1.131 1.127 44 3.4 1.187 1.184 1.181 1.156 1.136 1.127 45 3.2 1.194 1.192 1.186 1.158 1.146 1.135 46 3.0 1.206 1.200 1.191 1.158 1.154 1.144 47 2.8 1.220 1.210 1.199 1.162 1.162 1.154 48 2.6 1.233 1.221 1.210 1.170 1.165 1.158 49 2.4 1.249 1.235 1.223 1.181 1.174 1.165 50 2.2 1.270 1.253 1.233 1.194 1.191 1.179 51 2.0 1.285 1.265 1.243 1.204 1.204 1.189 52 1.8 1.289 1.269 1.253 1.208 1.205 1.190 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.12-318 Docket No. 50-338 COLR Table 3.2-1 (continued)

N1C23 Normal Operation N(Z)

NODE HEIGHT 11000 to 13000 13000 to 15000 15000 to 17000 17000 to 19000 19000 to EOR (FEET) MWD/MTU MWD/MTU MWD/MTU MWDIMTU MWD/MTU 10 10.2 1.144 1.145 1.127 1.127 1.077 11 10.0 1.142 1.143 1.127 1.127 1.085 12 9.8 1.140 1.141 1.125 1.124 1.097 13 9.6 1.141 1.140 1.124 1.124 1.114 14 9.4 1.143 1.136 1.118 1.124 1.123 15 9.2 1.151 1.136 1.121 1.133 1.134 16 9.0 1.166 1.145 1.142 1.156 1.158 17 8.8 1.181 1.158 1.169 .1.180 1.180 18 8.6 1.185 1.163 1.179 1.188 1.189 19 8.4 1.187 1.170 1.187 1.194 1.195 20 8.2 1.193 1.182 1.205 1.208 1.209 21 8.0 1.196 1.192 1.217 1.217 1.218 22 7.8 1.196 1.195 1.221 1.221 1.220 23 7.6 1.198 1.199 1.227 1.227 1.221 24 7.4 1.205 1.205 1.235 1.235 1.224 25 7.2 1.209 1.209 1.240 1.240 1.227 26 7.0 1.209 1.209 1.241 1.241 1.230 27 6.8 1.210 1.210 1.242 1.242 1.232 28 6.6 1.208 1.208 1.242 1.242 1.233 29 6.4 1.206 1.205 1.240 1.240 1.234 30 6.2 1.199 1.199 1.234 1.234 1.232 31 6.0 1.197 1.199 1.233 1.232 1.233 32 5.8 1.189 1.197 1.225 1.225 1.228 33 5.6 1.170 1.189 1.205 1.210 1.215 34 5.4 1.153 1.180 1.187 1.196 1.200 35 5.2 1.146 1.176 1.179 1.192 1.196 36 5.0 1.143 1.169 1.173 1.184 1.186 37 4.8 1.137 1.157 1.164 1.169 1.170 38 4.6 1.131 1.142 1.155 1.161 1.161 39 4.4 1.124 1.130 1.145 1.162 1.162 40 4.2 1.120 1.124 1.137 1.164 1.164 41 4.0 1.120 1.127 1.133 1.165 1.165 42 3.8 1.118 1.133 1.134 1.159 1.159 43 3.6 1.116 1.141 1.141 1.152 1.152 44 3.4 1.120 1.148 1.147 1.143 1.143 45 3.2 1.127 1.152 1.152 1.141 1.141 46 3.0 1.133 1.155 1.154 1.147 1.147 47 2.8 1.139 1.157 1.158 1.160 1.160 48 2.6 1.142 1.157 1.163 1.168 1.168 49 2.4 1.146 1.161 1.173 1.180 1.180 50 2.2 1.153 1.172 1.186 1.194 1.194 51 2.0 1.157 1.182 1.198 1.207 1.208 52 1.8 1.157 1.185 1.206 1.217 1.219 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.12-318 Docket No. 50-338 COLR Figure 3.2-1 K(Z) - Normalized FQ as a Function of Core Height 1.2 1.1

6, 1.0) 1.0 0.9 0.8 LLY0.7 N

-J

< 0.6 0

Z N,, 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.12-318 Docket No. 50-338 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (F NAH)

LCO 3.2.2 FN AH shall be within the limits specified below.

FN *_ 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.12-318 Docket No. 50-338 COLR Figure 3.2-2 North Anna 1 Cycle 23 Axial Flux Difference Limits 120 110 100 90 1.

80 70 I-60 0=

50 a.-

40 30 20 10 0

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

Page 15 of 21

Serial No.12-318 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_ ATAT o{ýK,-[T-K-K +77-+ - (P-p')-A (A,)}

I+K)(I(ts)[-P) where: AT is measured RCS AT, 'F AT0 is the indicated AT at RTP, 'F is the Laplace transform operator, secl T is the measured RCS average temperature, °F TP is the nominal Tavg at RTP, <586.8 IF P is the measured pressurizer pressure, psig pr is the nominal RCS operating pressure, > 2235 psig Kj _*1.2715 K 2 ->0.02174 /IF K3 > 0.001145 /psig r, z-2 = time constants utilized in the lead-lag controllerfor Tavg Tr ->23.75 sec T2 -<4.4 sec (1 + rIS)/(1 + ZYs) function generatedby the lead-lag controllerfor T.vg dynamic compensation fj(AI) > 0.0291{f- 13.0 - (qt - qb)} when (qt - qb) < -13.0% RTP 0 when -13.0% RTP _*(qt - qb)) < +7.0% RTP 0.0251 {(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.12-318 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 4 -K 5 [13S ]T-K6[T-T']-f 2 (AI)}

where: AT is measured RCS AT, °F.

AT0 is the indicated AT at RTP, °F.

s 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.

K4_ 1.0865 K5 > 0.0198 /°Ffor increasing Tavg K6 -Ž 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 Tmvg T3 > 9.5 see r3s/(1 + r3s) = function generated by the rate lag controllerfor Tag dynamic compensation f 2 (AI) = 0, for all Al.

Page 17 of 21

Serial No.12-318 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 IF; 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.12-318 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.12-318 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.12-318 Docket No. 50-338 NAPS TECHNICAL REQUIREMENTS MANUAL TRM 3.1 REACTIVITY CONTROL SYSTEMS TR 3.1.1 Boration Flow Paths - Operating Requir4ed Action D.2 Borate to a SHUTDOWN MARGIN > 1.77 % Ak/k at 200 IF, after xenon decay.

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