ML23249A259

From kanterella
Revision as of 01:33, 29 September 2023 by StriderTol (talk | contribs) (StriderTol Bot insert)
(diff) ← Older revision | Latest revision (diff) | Newer revision → (diff)
Jump to navigation Jump to search
Core Operating Limits Report Cycle 30, Pattern Ukr, Revision 0
ML23249A259
Person / Time
Site: North Anna Dominion icon.png
Issue date: 09/06/2023
From: Standley B
Dominion Energy Services, Virginia Electric & Power Co (VEPCO)
To:
Office of Nuclear Reactor Regulation, Document Control Desk
References
23-236
Download: ML23249A259 (1)
v • d • e


Text

Domi nion En ergy Se rvices , Inc.

5000 Dom inion Bouleva rd, Glen All en, VA 23060 Domi nion En ergy .com September 6, 2023 U. S. Nuclear Regulatory Commission Serial No.: 23-236 Attention: Document Control Desk NRA/JHH: RO Washington, DC 20555 Docket No.: 50-339 License No.: NPF-7 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION ENERGY VIRGINIA)

NORTH ANNA POWER STATION UNIT 2 CORE OPERATING LIMITS REPORT NORTH ANNA UNIT 2, CYCLE 30, PATTERN UKR, 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 2, Cycle 30, Pattern UKR, Revision 0.

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

Sincerely, B. E. Standley, Director Nuclear Regulatory Affairs Dominion Energy Services, Inc. for Virginia Electric and Power Company

Attachment:

Core Operating Limits Report, COLR-N2C30 Pattern UKR, Revision 0 Commitments: None.

Serial No.: 23-236 Docket No.: 50-339 COLR-N2C30, R0 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. G. E. Miller NRC Senior Project Manager - North Anna Power Station U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 9E3 11555 Rockville Pike Rockville, Maryland 20852-2738 NRC Senior Resident Inspector North Anna Power Station

Serial No.: 23-236 Docket No.: 50-339 ATTACHMENT Core Operating Limits Report COLR-N2C30 Pattern UKR Revision 0 NORTH ANNA POWER STATION UNIT 2 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION ENERGY VIRGINIA)

Serial No.23-236 Docket No. 50-339 Attachment N2C30 CORE OPERATING LIMITS REPORT INTRODUCTION The Core Operating Limits Report (COLR) for North Anna Unit 2 Cycle 30 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 (FNL1H)

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 17

Serial No.23-236 Docket No. 50-339 Attachment 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 Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology), November 2016.

Methodology for: TS 3.2.1 - Heat Flux Hot Channel Factor

3. EMF-2328(P)(A), PWR Small Break LOCA Evaluation Model, S-RELAP5 Based, as supplemented by ANP-3467P, Revision 0, North Anna Fuel-Vendor Independent Small Break LOCA Analysis, 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, 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

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 Page 2 of 17

Serial No.23-236 Docket No. 50-339 Attachment

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 4, Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code, including Appendix C, Qualification of the Westinghouse WRB-2M CHF Correlation in the Dominion VIPRE-D Computer Code, August 2010 and Appendix D, Qualification of the ABB-NV and WLOP CHF Correlations in the Dominion VIPRE-D Computer Code, September 2014.

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-A, Optimized ZIRLO', July 2006.

Methodology for:

TS 2.1.1 - Reactor Core Safety Limits and TS 3.2.1 - Heat Flux Hot Channel Factor Page 3 of 17

Serial No.23-236 Docket No. 50-339 Attachment 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.

Page 4 of 17

Serial No.23-236 Docket No. 50-339 Attachment COLR Figure 2.1-1 NORTH ANNA REACTOR CORE SAFETY LIMITS 665 660 I J 655 II II J I ,I 650 I I 2400 psia 645 640 j

, I I I I /I I 2250 psia J

I ,I 635 ,

Vessel Average Temperature [°F]

630 J I J I/

I I I I / VV 625 I I

620 I 2000 psia I V

~

j 615 610 I I j I II 1860 psia J

I I/

/ ,V 605 600 I I I II l/' /

/

595 ' ~

II V 590

/

585 /

580 .v 575 570 0 10 20 30 40 50 60 70 80 90 100 110 120 Percent of RATED THERMAL POWER Page 5 of 17

Serial No.23-236 Docket No. 50-339 Attachment 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.

Page 6 of 17

Serial No.23-236 Docket No. 50-339 Attachment 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 225steps.

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 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 % 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 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 % k/k.

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

Page 7 of 17

Serial No.23-236 Docket No. 50-339 Attachment COLR Figure 3.1-1 North Anna 2 Cycle 30 Control Rod Bank Insertion Limits Fully w/d position = 225 steps 230

[' 0.524, 225 220 210 200

~

190 180

~

" 1.0, 194

' C-BANK 170 160 150 Rod Group Position,# step w/d 140 130

' '~

,, 120 110 I

0, 118 '

100 90

' '~ D-BANK 80 70 60 50 40 "'

30 20 10

"'~

0

~ 0.048, 0 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 Page 8 of 17

Serial No.23-236 Docket No. 50-339 Attachment 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 = 2.32 The Heat Flux Hot Channel Factor, FQ(Z), shall be limited by the following relationships:

CFQ

  • K (Z )

( ) :5 p for P > 0.5 CFQ

  • K (Z )

( ) :5 for P :5 0.5 0.5 THERMAL POWER where: P=  ; and RATED THERMAL POWER 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).

( )= ( ) (1.03) * (1.05)

The expression for FQT(Z) is:

FJ (z) = FU (z) *N (z) where:

F Q(z ),Maximum Condition I

( )=

NZ F Q (Z ),Equilibrium Condition I 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.

Page 9 of 17

Serial No.23-236 Docket No. 50-339 Attachment 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.

Page 10 of 17

Serial No.23-236 Docket No. 50-339 Attachment COLR Table 3.2-1 N2C30 Penalty Factors for Flux Map Analysis Burnup Penalty (MWD/MTU) Factor %

0 - 499 2.0 500 - 999 2.5 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 N2C30 Required Operating Space Reductions for FQT(Z) Exceeding its Limits Required FQT(Z) Required Negative AFD Band Positive AFD Band Margin THERMAL POWER Reduction from AFD Reduction from AFD Improvement Limit (% RTP) Limits* (% AFD) Limits* (% AFD)

>0% and ::; 1% < 98.0% - 0.5%

> - 1.0%

> 1% and ::; 2% < 97.0% - 1.5%

> - 2.5%

> 2% and::; 3% < 95.0% - 1.5%

> - 3.0%

> 3% - 50%

< N/A N/A

  • Axial Flux Difference Limits are provided in COLR Figure 3.2-1 Page 11 of 17

Serial No.23-236 Docket No. 50-339 Attachment 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)}

THERMAL POWER P=

where: RATED THERMAL POWER 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.

Page 12 of 17

Serial No.23-236 Docket No. 50-339 Attachment COLR Figure 3.2-1 North Anna 2 Cycle 30 Axial Flux Difference Limits 120 110 100 (-12, 100) (+6, 100) 90 - /

~

Unacceptable / Unacceptable Operation Operation 80 V

~/

~

Percent of Rated Thermal Power

~ ....

70 Acceptable Operation 60 V

"-.... /

50

(-27, 50) (+20, 50) 40 30 20 10 0

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

Page 13 of 17

Serial No.23-236 Docket No. 50-339 Attachment 3.3 INSTRUMENTATION 3.3.1 Reactor Trip System (RTS) Instrumentation TS Table 3.3.1-1 Note 1: Overtemperature T V V

The Overtemperature T Function Allowable Value shall not exceed the following V

nominal trip setpoint by more than 2% of T span, with the numerical values of the parameters as specified below.

V> V - 1 .i+ )

1 s T }

T0 K 1 K 2 ( .i+ ) - -

T T ' K 3 ( P P' ) f 1 ( I )

+[ - { V 1 2 sf V

where: T is measured RCS T, °F V VT0 V 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 5 P is the measured pressurizer pressure, psig P is the nominal RCS operating pressure, 2235 psig <

K1 5 1.2715 K2 < 0.02174 /°F K3 < 0.001145 /psig 1, 1, 1, 2 = time constants utilized in the lead-lag controller for Tavg 1, 23.75 sec 1, 4.4 sec 5 1 < 2 1, 1, (1+ 1s)/(1+ 2s) = function generated by the lead-lag controller for Tavg dynamic compensation f1( I) V < 0.0291{-13.0 - (qt - qb)} when (qt - qb) < 13.0% RTP 0 5 5 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 14 of 17

Serial No.23-236 Docket No. 50-339 Attachment TS Table 3.3.1-1 Note 2: Overpower T V The Overpower T Function Allowable Value shall not exceed the following nominal trip V

setpoint by more than 2% of T span, with the numerical values of the parameters as specified V

below.

s V> V T

_[] _}

T0 K 4 K 5 1

3 3 s

{ V T K 6 [T T ' ] f 2 ( I )

where: TV is measured RCS T, °F. V T0 V is the indicated T at RTP, °F.

V 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. 5 K4 5 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 5 I

,1 3= time constant utilized in the rate lag controller for Tavg

,1 3 < 9.5 sec 3s / (1+ 3s) =

,1 ,1 function generated by the rate lag controller for Tavg dynamic compensation V

f2( I) = 0, for all I. V Page 15 of 17

Serial No.23-236 Docket No. 50-339 Attachment 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.

Page 16 of 17

Serial No.23-236 Docket No. 50-339 Attachment 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 IV 1.77 % k/k I>

at 200 °F, after xenon decay.

Page 17 of 17

Retrieved from "https://kanterella.com/w/index.php?title=ML23249A259&oldid=3618357"