ML101060511

Cycle 21 Core Operating Limits Report

ML101060511
Person / Time
Site:	North Anna
Issue date:	04/13/2010
From:	Funderburk C Dominion Resources Services, Virginia Electric & Power Co (VEPCO)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
10-229
Download: ML101060511 (23)

Text

Dominion Resources Services, Inc.

5000 Dominion Boulevard, Glen Allen, VA 23060 LJDomi nion Web Address: www.dom.com April 13, 2010 U. S. Nuclear Regulatory Commission Serial No.10-229 Attention: Document Control Desk NLOS /ETS One White Flint North Docket No. 50-339 11555 Rockville Pike License No. NPF-7 Rockville, MD 20852-2738 VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNIT 2 CYCLE 21 CORE OPERATING LIMITS REPORT Pursuant to North Anna Technical Specification 5.6.5.d, attached is a copy of the Virginia Electric and Power Company Core Operating Limits Report for North Anna Unit 2 Cycle 21 Pattern FRY.

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

Sincerely,

< C. L. Funderburk Director - Nuclear Licensing and Operations Support Dominion Resources Services, Inc.

for Virginia Electric and Power Company

Attachment:

CORE OPERATING LIMITS REPORT, North Anna 2 Cycle 21 Pattern FRY Commitments made in this letter: None AUc

Serial No.10-229 Docket No. 50-339 COLR Cycle 21 Page 2 of 2 cc: U.S. Nuclear Regulatory Commission Region II Marquis One Tower 245 Peachtree Center Avenue, 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 State Health Commissioner Virginia Department of Health James Madison Building - 7 th floor 109 Governor Street Suite 730 Richmond, Virginia 23219 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.10-229 Docket No. 50-339 ATTACHMENT CORE OPERATING LIMITS REPORT FOR NORTH ANNA UNIT 2 CYCLE 21 PATTERN FRY NORTH ANNA POWER STATION UNIT 2 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)

N2C21 CORE OPERATING LIMITS REPORT INTRODUCTION The Core Operating Limits Report (COLR) for North Anna Unit 2 Cycle 21 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 Test 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 20

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 3.1.5 - Shutdown Bank Insertion Limit, TS 3.1.4 - Rod Group Alignment Limits, TS 3.1.6 -

Control Bank Insertion Limits, TS 3.1.9 - Physics Test 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. VEP-NE-2-A, Rev. 0, 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)

3. VEP-NE 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)

4. 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)

5. 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 Test 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)

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

(Methodology for TS 2.1.1 - Reactor Core Safety Limits, TS 3.2.1 - Heat Flux Hot Channel Factor) r

7. 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)

8. 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 2 of 20

9. 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)

10. DOM-NAF-2, Rev. 0.1-A, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code," including Appendix A, "Qualification of the F-ANP BWU CHF Correlations in the VIPRE-D Computer Code," July 2009.

(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 3 of 20

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 limitý 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 < 5173°F, decreasing by 65°F per 10,000 MWD/MTU of burnup.

Page 4 of 20

COLR Figure 2.1-1 NORTH ANNA REACTOR CORE SAFETY I LIMITS 665 660 655 650 645 640 635 LA.

630 625 E 620 tw 615 610 (A

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 5 of 20

3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 SHUTDOWN MARGIN (SDM)

LCO 3.1.1 SDMshallbeŽ>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 104 Ak/'k/F (upper limit), when < 70%

RTP, and *0.0 Ak/k/°F when Ž_70% RTP.

The EOC/ARO/RTP-MTC shall be less negative than -5.0 x 104 Ak/k/°F (lower limit).

The MTC surveillance limits are:

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

-4.0 x 104 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 104 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' 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 6 of 20

3.1.5 Shutdown Bank Insertion Limits LCO 3.1.5 Each shutdown bank shall be withdrawn to at least 226 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 226 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 98 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 % Akk.

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.

SR3.1.6.3 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 7 of 20

COLR Figure 3.1-1 North Anna 2 Cycle 21 Control Rod Bank Insertion Limits 230 220 210 200 190 180 170 160 0.

a, 150 C- 140 0

0. 130 0.

120 0

"cc O~

-o 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 8 of 20

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)

FP (Z) < for P>O.5 P N(Z)

CFQ K(Z)

F* (Z) . N 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 3.2.1 for this LCO requires the application of a cycle dependent non-equilibrium multiplier, N(Z), to the measured peaking factor, 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 equilibriumFQ(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 is presentedin COLR Table 3.2-1.

Page 9 of 20

COLR Table 3.2-1 N2C21 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 MWDIMTU MWD/MTU 10 10.2 1.093 1.119 1.130 1.131 1.132 1.151 11 10.0 1.092 1.117 1.128 1.130 1.131 1.149 12 9.8 1.094 1.115 1.126 1.129 1.129 1.145 13 9.6 1.099 1.114 1.125 1.129 1.129 1.143 14 9.4 1.101 1.111 1.124 1.127 1.128 1.141 15 9.2 1.105 1.112 1.129 1.129 1.129 1.142 16 9.0 1.119 1.122 1.143 1.143 1.136 1.142 17 8.8 1.133 1.133 1.158 1.158 1.145 1.145 18 8.6 1.138 1.137 1.163 1.163 1.149 1.148 19 8.4 1.139 1.139 1.164 1.164 1.152 1.152 20 8.2 1.141 1.141 1.167 1.167 1.157 1.157 21 8.0 1.141 1.141 1.168 1.168 1.161 1.161 22 7.8 1.140 1.140 1.168 1.168 1.163 1.162 23 7.6 1.139 1.139 1.170 1.169 1.166 1.164 24 7.4 1.137 1.137 1.172 1.172 1.171 1.167 25 7.2 1.136 1.136 1.174 1.174 1.173 1.169 26 7.0 1.137 1.137 1.173 1.173 1.173 1.171 27 6.8 1.138 1.138 1.173 1.174 1.174 1.172 28 6.6 1.138 1.138 1.172 1.173 1.173 1.172 29 6.4 1.136 1.136 1.168 1.170 1.171 1.170 30 6.2 1.132 1.132 1.161 1.164 1.166 1.166 31 6.0 1.131 1.130 1.159 1.163 1.166 1.165 32 5.8 1.126 1.126 1.151 1.157 1.161 1.162 33 5.6 1.113 1.113 1.132 1.139 1.145 1.156 34 5.4 1.102 1.106 1.116 1.125 1.132 1.150 35 5.2 1.097 1.110 1.113 1.123 1.127 1.147 36 5.0 1.098 1.118 1.117 1.127 1.129 1.143 37 -4.8 1.102 1.123 1.122 1.128 1.133 1.137 38 4.6 1.107 1.128 1.127 1.130 1.136 1.135 39 4.4 1.112 1.130 1.130 1.131 1.138 1.137 40 4.2 1.118 1.133 1.133 1.129 1.137 1.136 41 4.0 1.126 1.136 1.136 1.127 1.134 1.134 42 3.8 1.136 1.137 1.138 1.126 1.130 1.130 43 3.6 1.145 1.138 1.139 1.129 1.128 1.128 44 3.4 1.152 1.141 1.137 1.130 1.128 1.129 45 3.2 1.158 1.148 1.137 1.134 1.133 1.134 46 3.0 1.164 1.157 1.141 1.139 1.140 1.142 47' 2.8 1.173 1.169 1.149 1.149 1.148 1.151 48 2.6 1.183 1.180 1.154 1.155 1.152 1.152 49 2.4 1.196 1.194 1.165 1.165 1.159 1.156 50 2.2 1.213 1.212 1.181 1.182 1.173 1.170 51 2.0 1.224 1.224 1.193 1.193 1.182 1.179 52 1.8 1.227 1.228 1.195 1.195 1.184 1.180 These decks are generated for normal operation flux maps that are typically taken at full power ARO.

Additional N(z) decks may be generated for the specific plant conditions at the time of the flux map, if necessary, consistent with the methodology described in the RPDC topical (Reference 3). EOR is defined as Hot Full Power End of Reactivity.

Page 10 of 20

COLR Table 3.2-1 (continued)

N2C21 Normal Operation N(Z)

NODE HEIGHT 11000 to 13000 13000 to 15000 15000 to 17000 17000 to EOR (FEET) MWD/MTU MWDIMTU MWD/MTU MWD/MTU 10 10.2 1.151 1.119 1.119 1.119 11 10.0 1.149 1.117 1.118 1.118 12 9.8 1.145 1.115 1.117 1.116 13 9.6 1.143 1.115 1.116 1.116 14 9.4, 1.141 1.110 1.112 1.110 15 9.2 1.142 1.113 1.114 1.113 16 9.0 1.140 1.126 1.122 1.134 17 8.8 1.143 1.143 1.135 1.159 18 8.6 1.148 1.150 1.140 1.166 19 8.4 1.158 1.158 1.149 1.172 20 8.2 1.171 1.172 1.165 1.187 21 8.0 1.181 1.181 1.176 1.197 22 7.8 1.184 1.184 1.180 1.200 23 7.6 1.188 1.189 1.186 1.203 24 7.4 1.194 '1.197 1.196 1.209 25 7.2 1.197 1.202 1.203 1.214 26 7.0 1.197 1.204 1.204 1.216 27 6.8 1.197 1.206 1.206 1.218 28 6.6 1.195 1.206 1.206 1.220 29 6.4 1.190 1.207 1.207 1.222 30 6.2 1.182 1.204 1.204 1.221 31 6.0 1.177' 1.205 1.205 1.224' 32 5.8 1.170 1.199 1.199 1.218, 33 5.6 1.159 1.183 1.183 1.205 34 5.4 1.149 1.168 1.168 1.191 35 5.2 1.146 1.161 1.161 1.188 36 .5.0 1.145 1.156 1.156 1.180 37 :4.8 1.142 1.151 1.151 1.165 38 4.6 1.139 1.149 1.149 1.150 39 4.4 1.136 1.149 1.149 1.138 40 4.2 1.132 1.145 1.145 1.140 41 4.0 1.129 1.139 1.139 1.152 42 3.8 1.127 1.139 1.139 1.160 43 3.6 1.129 1.143 1.143 1.165 44 3.4 1.130 1.144 1.144 1.168 45 3.2 1.134 1.146 1.146 1.169 46 3.0 1.142 1.146 1.148 1.168 47 2.8 1.151 1.148 1.152 1.167 48 2.6 1.151 1.149 1.153 1.164 49 2.4 1.157 1.157 1.162 1.171 50 2.2 1.172 1.174 1.180 1.190 51 2.0 1.183 1.187 1.194 1.207 52 1.8 1.185 1.190 1.198 1.214 These decks are generated for normal operation flux maps that are typically taken at full power ARO.

Additional N(z) decks may be generated for the specific plant conditions at the time of the flux map, if necessary, consistent with the methodology described in the RPDC topical (Reference 3). EOR is defined as Hot Full Power End of Reactivity..

Page 11 of 20

COLR Figure 3.2-1 K(Z) - Normalized FQ as a Function of Core Height 1.2 1.1 1.0 0.9 0.8 LR0.7 N

-j

< 0.6 0

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 12 of 20

3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (F NAH)

LCO 3.2.2 FNAH 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 13 of 20

COLR Figure 3.2-2 North Anna 2 Cycle 21 Axial Flux Difference Limits 120 110

(-12, 100) (+6,100) 100 Unaccep able pera ion Un accep able

___ ____tin pera ion 90 80 0

Ac cepta le Op ratio 70 60 0

50

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

40 30 20 10 0

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

Page 14 of 20

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.

S l+ l2S AT:* ATj KIK2(1+/-vTs)[T-T}-+K,(P-P')-f J (A!)l 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.

P is the measured pressurizer pressure, psig.

P' is the nominal RCS operating pressure, >_2235 psig.

Ki < 1.2715 K 2 > 0.02172 /-F K3 > 0.001144 /psig Tl, z2 = time constants utilized in the lead-lag controllerfor Tavg ti1 --23.75 sec T2 -<4.4 sec (I + "ris)/(]+ r2s) = function generated by the lead-lag controllerfor Tavg dynamic compensation fl(AI) _ 0.0165{-35 - (qt- qb)} when (qt - qb) < -35% RTP 0 when -35% RTP < (qt - qb) -<+3% RTP 0.0198{(qt - qb) - 3} when (qt - qb) > +3% RTP Where q, 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 15 of 20

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 L[ ]T-K 13S 6 [T-T' ]-f 2 (AJ)}

where: AT is measured RCS AT, 'F.

AT 0 is the indicated AT at RTP, 'F.

s is the Laplace transform operator, sec-1 .

T is the measured RCS average temperature, 'F.

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

K 4 < 1.0865 K 5 > 0.0197 /-F for increasing Tavg K6 > 0.00162 /F when T > T' 0 /1F for decreasing Tavg 0 /F when T _ T' z3 time constant utilized in the rate lag controllerfor Tavg T3 - 9.5 sec r3s/(1 + r3s) = function generated by the rate lag controllerfor Tavg dynamic compensation f 2(AI) = 0, for all Al.

Page 16 of 20

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 IF.

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 Ž_90% RTP.

Verify by precision heat balance that RCS total flow rate is Ž_295,000 gpm.

Page 17 of 20

3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.6 Boron Injection Tank (BIT)

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

Page 18 of 20

3.9 REFUELING OPERATIONS 3.9.1 Boron Concentration I

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 19 of 20

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.

Page 20 of 20