ML23249A259
| ML23249A259 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| 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) | |
Text
Do mi n ion En e rgy Se rv ices, In c.
5000 Do m in ion Bo u leva rd, Gle n A llen, VA 23 06 0 Domi nion En e rgy.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-N2C 30 Pattern UKR Revision 0
NORTH ANNA POWER STATION UNIT 2 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION ENERGY VIRGINIA)
N2C30 CORE OPERATING LIMITS REPORT
INTRODUCTION
The Core Operating Limits Report (COLR) for North Anna Unit 2Cycle30has 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 3.1.3Moderator Temperature Coefficient (MTC)
TS 3.1.4Rod 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 (FNL1 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)
In addition, a technical requirement (TR) in the NAPS Technical Requirements Manual (TRM) refers to the COLR:
TR 3.1.1Boration 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.
REFERENCES
1.VEP-FRD-42-A, Revision2, Minor Revision 2, Reload Nuclear Design Methodology, October2017.
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 ChannelFactor
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 6.VEP-NE-1-A, Revision 0, Minor Revision 3, Relaxed Power Distribution Control Methodology and Associated FQ Surveillance Technical Specifications,October2017.
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 2.0SAFETY LIMITS (SLs)
2.1SLs
2.1.1Reactor 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.
2.1.1.1The 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.
COLR Figure 2.1-1
NORTH ANNA REACTOR CORE SAFETY LIMITS 665
660
655 650 2400 psia
645
640 2250 psia 635
630 625 620 2000 psia
615 610 1860 psia
605
600 595
590 585
580
575 570 0102030405060708090100110120 Percent of RATED THERMAL POWER 3.1REACTIVITY CONTROL SYSTEMS
3.1.1 SHUTDOWN MARGIN (SDM)
LCO 3.1.1SDM shall be 1.77 % k/k.
3.1.3 Moderator Temperature Coefficient (MTC)
LCO3.1.3 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, and0.0 k/k/°Fwhen 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 10-4k/k/°F [Note 1].
The 60 ppm/ARO/RTP-MTC should be less negative than or equal to 4.7 x 10-4k/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 10-4k/k/°F.
3.1.4Rod Group Alignment Limits
Required Action A.1.1Verify SDM to be 1.77 % k/k.
Required Action B.1.1Verify SDM to be 1.77 % k/k.
Required Action D.1.1Verify SDM to be 1.77 % k/k.
3.1.5Shutdown Bank Insertion Limits LCO3.1.5Each shutdown bank shall be withdrawn to at least225steps.
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 3.1.5.1Verify each shutdown bank is withdrawn to at least 225 steps.
3.1.6Control Bank Insertion Limits LCO 3.1.6Control 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 3.1.6.1Verify estimated critical control bank position is within the insertion limits specified in COLR Figure 3.1-1.
SR 3.1.6.2Verify each control bank iswithin the insertionlimits specified in COLR Figure3.1-1.
SR 3.1.6.3Verify 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.bSDM is 1.77 % k/k.
SR 3.1.9.4Verify SDM to be 1.77 % k/k.
COLR Figure 3.1-1 North Anna 2Cycle 30 Control Rod Bank Insertion Limits Fully w/d position = 225 steps
230 220 0.524, 225
210 200 1.0, 194 190
180 C-BANK 170 160
150 140 130
120 0, 118 110 100 D-BANK 90 80 70
60 50 40
30 20 10 0 0.048, 0
00.10.20.30.40.50.60.70.80.91 Fraction of Rated Thermal Power 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:
( ) :5 p CFQ
- K Z ()for P >0.5
( ) :5 CFQ
- K Z ()for P :5 0.5
0.5 where
P = THERMAL POWER ; and POWERTHERMALRATED
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:
Q z Maximum Condition I (),
NZ F ()=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.
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) betweensurveillances. 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) versusburnup. A minimum value of 2% is used should any increase in steady-state or transient measured or predictedpeaking 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. Thedetermination and verification of the margin improvements along with the corresponding required reductions inthe Thermal PowerLimit and AFD Bands are performed on a cycle-specific basis.
COLR Table 3.2-1 N2C30 Penalty Factors for Flux Map Analysis Burnup Penalty (MWD/MTU) Factor %
0 - 499 2.0 500 -9992.5 1000 -EOC2.0 Notes:
- 1. PenaltyFactors are not required for initial power ascension flux maps.
- 2. All full power maps shall apply a PenaltyFactor 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 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (FN H)
LCO 3.2.2FN Hshall be within the limits specified below.
FN H 1.587{1 + 0.3(1 -P)}
P = THERMAL POWER where: POWERTHERMALRATED
SR 3.2.2.1Verify FN His within limits specified above.
3.2.3AXIAL 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.
COLR Figure 3.2-1 North Anna 2Cycle 30 Axial Flux Difference Limits
120
110
100 (-12, 100) (+6, 100)
90 Unacceptable Unacceptable Operation Operation 80
70 Acceptable Operation
60
50 (+20, 50)
(-27, 50)
40
30
20
10
0 20-100102030 Percent Flux Difference (Delta-I) 3.3INSTRUMENTATION
3.3.1Reactor 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.
1 2101 )()'('1s IfPPKTTsKKTT 13 2
where: Tis measured RCS T, °F T0 is the indicated T at RTP, °F s is the Laplace transform operator, sec-1 Tis the measured RCS average temperature, °F T is the nominal Tavgat RTP, 586.8 °F Pis the measured pressurizer pressure, psig P is the nominal RCS operating pressure, 2235 psig
K1 1.2715K2 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 Tavgdynamic 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 qtand qbare percent RTP in the upper and lower halves of the core, respectively, and qt + qbis the total THERMAL POWER in percent RTP.
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.
3sKKTT 540 IfTTKTs )(]'[1 26 3
where: Tis measured RCS T, °F.
T0 is the indicated T at RTP, °F.
s is the Laplace transform operator, sec-1.
Tis the measured RCS average temperature, °F.
T is the nominal Tavgat RTP, 586.8 °F.
K4 1.0865 K5 0.0198 /°Ffor increasing TavgK6 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.
3.4REACTOR COOLANT SYSTEM (RCS)
3.4.1RCS 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 3.4.1.1Verify pressurizer pressure is greater than or equal to 2205 psig.
SR 3.4.1.2Verify RCS average temperature is less than or equal to 591 °F.
SR 3.4.1.3Verify 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.
3.5EMERGENCY CORE COOLING SYSTEMS (ECCS)
3.5.6Boron Injection Tank (BIT)
Required Action B.2 Borate to a SDM 1.77 % k/k at 200 °F.
3.9REFUELING OPERATIONS
3.9.1Boron Concentration
LCO 3.9.1Boron concentrations of the Reactor Coolant System (RCS), the refueling canal, and the refueling cavity shall be maintained 2600 ppm.
SR 3.9.1.1Verify boron concentration is within the limit specified above.
NAPS TECHNICAL REQUIREMENTS MANUAL
TRM 3.1REACTIVITY CONTROL SYSTEMS
TR 3.1.1Boration Flow Paths -Operating Required Action D.2Borate to a SHUTDOWN MARGIN IV I> 1.77 % k/k at 200 °F, after xenon decay.