ML25281A193

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Core Operating Limits Report, Cycle 32, Pattern GOP, Revision 0
ML25281A193
Person / Time
Site: North Anna Dominion icon.png
Issue date: 10/08/2025
From: Harrell J
Virginia Electric & Power Co (VEPCO)
To:
Office of Nuclear Reactor Regulation, Document Control Desk
References
25-217
Download: ML25281A193 (1)
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Text

VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 October 8, 2025 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Serial No.:

25-217 NRA/JHH:

R1 Docket No.:

50-338 License No.:

NPF-4 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION ENERGY VIRGINIA)

NORTH ANNA POWER STATION UNIT 1 CORE OPERA TING LIMITS REPORT NORTH ANNA UNIT 1. CYCLE 32, PATTERN GOP, REVISION 0 Pursuant to North Anna Power Station Units 1 and 2 Technical Specification 5.6.5.d, attached is a copy of the Core Operating Limits Report for North Anna Unit 1, Cycle 32, Pattern GOP, Revision 0.

If you have any questions or require additional information, please contact Julie Hough at (804) 273-3586.

Sincerely, arrell Director, Nuclear Analysis & Fuel and Regulatory Assurance Dominion Energy Services, Inc. for Virginia Electric and Power Company

Attachment:

Core Operating Limits Report, COLR-N1C32, Pattern GOP, Revision 0 Commitments: None.

Serial No.: 25-217 Docket No.: 50-338 Page 2 of 2 cc: U. S. Nuclear Regulatory Commission, Region II Mr. G. E. Miller NRC Senior Project Manager - North Anna Power Station NRC Senior Resident Inspector North Anna Power Station

Serial No.: 25-217 Docket No.: 50-338 NORTH ANNA POWER STATION UNIT 1 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION ENERGY VIRGINIA)

ATTACHMENT Core Operating Limits Report COLR-N1C32 Pattern GOP Revision 0

COLR-N1C32, Revision 0 EVAL-ENG-RSE-N1C32, Revision 0, Attachment A Page 1 of 17 N1C32 CORE OPERATING LIMITS REPORT INTRODUCTION The Core Operating Limits Report (COLR) for North Anna Unit 1 Cycle 32 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 (FN H)

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.

Serial No.25-217 Docket No. 50-338 Attachment

COLR-N1C32, Revision 0 EVAL-ENG-RSE-N1C32, Revision 0, Attachment A Page 2 of 17 REFERENCES

1. VEP-FRD-42-A, Revision 2, Minor Revision 2, Reload Nuclear Design Methodology, October 2017.

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. WCAP-16996-P-A, Revision 1, Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology), November 2016.

Methodology for: TS 3.2.1 Heat Flux Hot Channel Factor

3. EMF-

-RELAP5 Based, as supplemented by ANP-3467P

-Vendor Independent Small Break as approved by NRC Safety Evaluation Report dated March 19, 2021.

Methodology for: TS 3.2.1 Heat Flux Hot Channel Factor

4. WCAP-12610-P-A April 1995.

Methodology for:

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

5. VEP-NE-2-A, Revision 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 Serial No.25-217 Docket No. 50-338 Attachment "Realistic LOCA Evaluation Methodology Applied to the Full 2328(P)(A), "PWR Small Break LOCA Evaluation Model, S

, Revision 0, "North Anna Fuel LOCA Analysis,"

, "VANTAGE+ FUEL ASSEMBLY -REFERENCE CORE REPORT,"

COLR-N1C32, Revision 0 EVAL-ENG-RSE-N1C32, Revision 0, Attachment A Page 3 of 17

6. VEP-NE-1-A, Revision 0, Minor Revision 3, Relaxed Power Distribution Control Methodology and Associated FQ Surveillance Technical Specifications, October 2017.

Methodology for:

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

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

Methodology for:

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

8. 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

9. DOM-NAF-2-P-A, Revision 0, Minor Revision 5, Reactor Core Thermal-Hydraulics Using the VIPRE-D 2M CHF Correlation in the Dominion VIPRE-and Appendix

-NV and WLOP CHF Correlations in the Dominion VIPRE-D Methodology for:

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

10. WCAP-12610-P-A and CENPD-404-P-A, Addendum 1-Methodology for:

TS 2.1.1 Reactor Core Safety Limits and TS 3.2.1 Heat Flux Hot Channel Factor Serial No.25-217 Docket No. 50-338 Attachment Computer Code," including Appendix C, "Qualification of the Westinghouse WRB D Computer Code," August 2010 D, "Qualification of the ABB Computer Code," September 2014.

A, "Optimized ZIRLO' " July 2006.

COLR-N1C32, Revision 0 EVAL-ENG-RSE-N1C32, Revision 0, Attachment A Page 4 of 17 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.

Serial No.25-217 Docket No. 50-338 Attachment

COLR-N1C32, Revision 0 EVAL-ENG-RSE-N1C32, Revision 0, Attachment A Page 5 of 17 COLR Figure 2.1-1 2400 psia 2250 psia 2000 psia 1860 psia 570 575 580 585 590 595 600 605 610 615 620 625 630 635 640 645 650 655 660 665 0

10 20 30 40 50 60 70 80 90 100 110 120 Percent of RATED THERMAL POWER NORTH ANNA REACTOR CORE SAFETY LIMITS Serial No.25-217 Docket No. 50-338 Attachment Vessel Average Temperature [°F]

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COLR-N1C32, Revision 0 EVAL-ENG-RSE-N1C32, Revision 0, Attachment A Page 6 of 17 3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 SHUTDOWN MARGIN (SDM)

LCO 3.1.1 SDM shall be 1.77 % k/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 k/k/°F, when < 70% RTP, and 0.0 k/k/°F when 70%

RTP.

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

RTP, and 0.0 k/k/°F when 70% RTP.

The EOC/ARO/RTP-MTC shall be less negative than 5.0 x 10-4 k/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 k/k/°F [Note 1].

The 60 ppm/ARO/RTP-MTC should be less negative than or equal to 4.7 x 10-4 k/k/°F [Note 2].

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

Note 1: If the MTC is more negative than 4.0 x 10-4 k/k/°F, SR 3.1.3.2 shall be repeated once per 14 EFPD during the remainder of the fuel cycle.

Note 2: 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 k/k/°F.

Serial No.25-217 Docket No. 50-338 Attachment A

A A

COLR-N1C32, Revision 0 EVAL-ENG-RSE-N1C32, Revision 0, Attachment A Page 7 of 17 3.1.4 Rod Group Alignment Limits Required Action A.1.1 Verify SDM to be 1.77 % k/k.

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

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

3.1.5 Shutdown Bank Insertion Limits LCO 3.1.5 Each shutdown bank shall be withdrawn to at least 225 steps. The cycle specific fully withdrawn position shall be 229 steps.

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

Required Action B.1 Verify SDM to be 1.77 % k/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 A and B shall be withdrawn to at least 225 steps. Control banks C and D 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. The cycle specific fully withdrawn position shall be 229 steps and the overlap limit during withdrawal shall be 101 steps.

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

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

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

SR 3.1.6.1 Verify estimated critical control bank position is within the insertion limits specified in LCO 3.1.6 above.

SR 3.1.6.2 Verify each control bank is within the insertion limits specified in LCO 3.1.6 above.

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 % k/k.

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

Serial No.25-217 Docket No. 50-338 Attachment IV IV I>

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COLR-N1C32, Revision 0 EVAL-ENG-RSE-N1C32, Revision 0, Attachment A Page 8 of 17 COLR Figure 3.1-1 North Anna 1 Cycle 32 Control Rod Bank Insertion Limits Fully w/d position = All Rods Out = 229 steps 0, 118 0.524, 225 1.0, 225 0.048, 0 1.0, 194 0

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

0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

Fraction of Rated Thermal Power C-BANK D-BANK Serial No.25-217 Docket No. 50-338 Attachment Rod Group Position,# step w/d I

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COLR-N1C32, Revision 0 EVAL-ENG-RSE-N1C32, Revision 0, Attachment A Page 9 of 17 3.2 POWER DISTRIBUTION LIMITS 3.2.1 Heat Flux Hot Channel Factor (FQ(Z))

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

CFQ = 2.5 The Heat Flux Hot Channel Factor, FQ(Z), shall be limited by the following relationships:

where:

POWER THERMAL RATED POWER THERMAL P

and K(Z) = 1.0 for all core heights, z FQE(Z) is an excellent approximation for FQ(Z) when the reactor is at the steady-state power.

FQ(Z) from the incore flux map results is increased by 1.03 for fuel manufacturing tolerances and 1.05 for measurement uncertainty to obtain FQE(Z).

The expression for FQT(Z) is:

where:

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 steady state FQE(Z). N(Z) values are calculated for each flux map using analytically derived FQ(Z) values (scaled by relative power), consistent with the methodology described in VEP-NE-1. N(Z) accounts for power distribution transients encountered during normal operation.

Serial No.25-217 Docket No. 50-338 Attachment for P > 0.5 for P $; 0.5 FJ(Z) = F!(Z)

  • N(Z)

N(Z) = FQ(Z),MaximumConditionI FQ(Z),Equilibrium Condition I

COLR-N1C32, Revision 0 EVAL-ENG-RSE-N1C32, Revision 0, Attachment A Page 10 of 17 The cycle-specific penalty factors are presented in COLR Table 3.2-1.

Also discussed is the application of the appropriate factor to account for potential increases in FQ(Z) between surveillances. This factor is determined on a cycle specific basis and is dependent on the predicted increases in steady-state and transient FQ(Z)/K(Z) versus burnup. A minimum value of 2% is used should any increase in steady-state or transient measured or predicted peaking factor be determined unless frequent flux mapping is invoked (7 EFPD).

The required operating space reductions are included in COLR Table 3.2-2.

Should FQT(Z) exceed its limits the normal operating space should be reduced to gain peaking factor margins. The determination and verification of the margin improvements along with the corresponding required reductions in the Thermal Power Limit and AFD Bands are performed on a cycle-specific basis.

Serial No.25-217 Docket No. 50-338 Attachment

COLR-N1C32, Revision 0 EVAL-ENG-RSE-N1C32, Revision 0, Attachment A Page 11 of 17 COLR Table 3.2-1 N1C32 Penalty Factors for Flux Map Analysis Burnup (MWD/MTU)

Penalty Factor %

0-499 2.0%

500-999 3.0%

1000-EOC 2.0%

Notes:

1. Penalty Factors are not required for initial power ascension flux maps.
2. All full power maps shall apply a Penalty Factor unless frequent flux mapping is invoked

( 7 EFPD).

COLR Table 3.2-2 N1C32 Required Operating Space Reductions for FQT(Z) Exceeding its Limits Required FQT(Z)

Margin Improvement Required THERMAL POWER Limit (% RTP)

Negative AFD Band Reduction from AFD Limits* (% AFD)

Positive AFD Band Reduction from AFD Limits* (% AFD)

> 0% and

< 100.0%

> 1.0%

> 0.0%

< 100.0%

> 2.0%

> 0.0%

< 100.0%

> 3.0%

> 0.0%

> 3%

< 50.0%

N/A N/A

  • Axial Flux Difference Limits are provided in COLR Figure 3.2-1 Serial No.25-217 Docket No. 50-338 Attachment
S 1%

> 1 % and ::S 2%

> 2% and ::S 3%

COLR-N1C32, Revision 0 EVAL-ENG-RSE-N1C32, Revision 0, Attachment A Page 12 of 17 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (FN H)

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

FN H 1.587{1 + 0.3(1 P)}

where:

POWER THERMAL RATED POWER THERMAL P

SR 3.2.2.1 Verify FN H 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 3.2-1.

Serial No.25-217 Docket No. 50-338 Attachment

COLR-N1C32, Revision 0 EVAL-ENG-RSE-N1C32, Revision 0, Attachment A Page 13 of 17 COLR Figure 3.2-1 North Anna 1 Cycle 32 Axial Flux Difference Limits 0

10 20 30 40 50 60 70 80 90 100 110 120

-30

-20

-10 0

10 20 30 Percent Flux Difference (Delta-I)

(+6, 100)

(-12, 100)

Acceptable Operation Unacceptable Operation Unacceptable Operation

(-27, 50)

(+20, 50)

Serial No.25-217 Docket No. 50-338 Attachment Percent of Rated Thermal Power

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COLR-N1C32, Revision 0 EVAL-ENG-RSE-N1C32, Revision 0, Attachment A Page 14 of 17 3.3 INSTRUMENTATION 3.3.1 Reactor Trip System (RTS) Instrumentation TS Table 3.3.1-1 Note 1: Overtemperature T The Overtemperature T Function Allowable Value shall not exceed the following nominal trip setpoint by more than 2% of T span, with the numerical values of the parameters as specified below.

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P K

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s K

K T

T 2

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1 where: T is measured RCS T, °F T0 is the indicated T 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 P

is the measured pressurizer pressure, psig P

is the nominal RCS operating pressure, 2235 psig K1 1.2715 K2 0.02174 /°F K3 0.001145 /psig 1, 2 = time constants utilized in the lead-lag controller for Tavg 1

23.75 sec 2 4.4 sec (1+ 1s)/(1+ 2s) = function generated by the lead-lag controller for Tavg dynamic compensation f1( I) 0.0291{-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.

Serial No.25-217 Docket No. 50-338 Attachment

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COLR-N1C32, Revision 0 EVAL-ENG-RSE-N1C32, Revision 0, Attachment A Page 15 of 17 TS Table 3.3.1-1 Note 2: Overpower T The Overpower T Function Allowable Value shall not exceed the following nominal trip setpoint by more than 2% of T span, with the numerical values of the parameters as specified below.

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K K

T T

where: T is measured RCS T, °F.

T0 is the indicated T 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.

K4 1.0865 K5 0.0198 /°F for increasing Tavg K6 0.00162 /°F when T > T 0 /°F for decreasing Tavg 0 /°F when T T 3 = time constant utilized in the rate lag controller for Tavg 3

9.5 sec 3s / (1+ 3s) = function generated by the rate lag controller for Tavg dynamic compensation f2( I) = 0, for all I.

Serial No.25-217 Docket No. 50-338 Attachment I >

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COLR-N1C32, Revision 0 EVAL-ENG-RSE-N1C32, Revision 0, Attachment A Page 16 of 17 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 °F; 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 °F.

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.

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

Required Action B.2 Borate to a SDM 1.77 % k/k at 200 °F.

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.

Serial No.25-217 Docket No. 50-338 Attachment IV IV I>

IV IV

COLR-N1C32, Revision 0 EVAL-ENG-RSE-N1C32, Revision 0, Attachment A Page 17 of 17 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 % k/k at 200 °F, after xenon decay.

Serial No.25-217 Docket No. 50-338 Attachment IV I>

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