ML24088A269
| ML24088A269 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 03/27/2024 |
| From: | Standley B Virginia Electric & Power Co (VEPCO) |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| 24-126 | |
| Download: ML24088A269 (1) | |
Text
V IRGINIA E LECT RIC AND P OWER C O MPANY R IC HMON D, V IRGINIA 2326 1 March 27, 2024 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Serial No.:
24-126 NRA/JHH:
RO Docket No.:
50-338 License No.:
NPF-4 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION ENERGY VIRGINIA)
NORTH ANNA POWER STATION UNIT 1 CORE OPERATING LIMITS REPORT NORTH ANNA UNIT 1, CYCLE 31, PATTERN SOS. 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 31, Pattern SOS, 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-N 1 C31 Pattern SOS, Revision 0 Commitments: None.
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 NRG 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.: 24-126 Docket No.: 50-338 Page 2 of 2
ATTACHMENT Core Operating Limits Report COLR-N1C31 Pattern SOS Revision 0 NORTH ANNA POWER STATION UNIT 1 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION ENERGY VIRGINIA)
Serial No.: 24-126 Docket No.: 50-338
NI C31 CORE OPERA TING LIMITS REPORT INTRODUCTION Serial No.24-126 Attachment Page 1 of 17 The Core Operating Limits Report (COLR) for North Anna Unit 1 Cycle 31 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 TS 3.1.1 TS 3.1.3 TS 3.1.4 TS 3.1.5 TS 3.1.6 TS 3.1.9 TS 3.2.1 TS 3.2.2 TS 3.2.3 TS 3.3.1 TS 3.4.1 TS 3.5.6 TS 3.9.1 Reactor Core Safety Limits Shutdown Margin (SDM)
Moderator Temperature Coefficient (MTC)
Rod Group Alignment Limits Shutdown Bank Insertion Limit Control Bank Insertion Limits PHYSICS TESTS Exceptions - Mode 2 Heat Flux Hot Channel Factor Nuclear Enthalpy Rise Hot Channel Factor (FN,m)
Axial Flux Difference (AFD)
Reactor Trip System (RTS) Instrumentation RCS Pressure, Temperature, and Flow DNB Limits Boron Injection Tank (BIT)
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.
REFERENCES Serial No.24-126 Attachment Page 2 of 17
- 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
Serial No.24-126 Attachment 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 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
Serial No.24-126 Attachment 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.
665 660 655 650 645 640
.... 635 LA.
I!...
f 630 IQ 625 CII a.
E 620
{!!.
CII 615 b.O IQ...
CII 610
<(
li 605 Ill Ill
~ 600 595 590 585 580 575 570
~
r-,,....__
COLR Figure 2.1-1 Serial No.24-126 Attachment Page 5 of 17 NORTH ANNA REACTOR CORE SAFETY LIMITS r---..............
-.......... ~
psia
....... --""'~
-......... ~
psia
--~.....
I
'\\
r-,.....,__
~""'
\\
2000
~
psia
............. \\. \\
--~
\\, \\
\\ \\.
1860 psia
!lo,..
'\\
\\.
...................... \\..
\\ '\\
\\
0 10 20 30 40 so 60 70 80 90 100 110 120 Percent of RATED THERMAL POWER
3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 SHUTDOWN MARGIN (SDM)
LCO 3.1.1 SDM shall be~ 1.77 % Ak/k.
3.1.3 Moderator Temperature Coefficient (MTC)
Serial No.24-126 Attachment Page 6 of 17 LCO 3.1.3 The MTC shall be maintained within the limits specified below. The upper limit ofMTC 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~ 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/°F [Note 1].
The 60 ppm/ARO/RTP-MTC should be less negative than or equal to
-4.7 x 104 Ak/k/°F [Note 2].
SR 3.1.3.2 Verify MTC is within -5.0 x 10-4 Ak/k/°F (lower limit).
Note 1: 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 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 104 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.l.1 Verify SDM to be~ 1.77 % Ak/k.
3.1.5 Shutdown Bank Insertion Limits Serial No.24-126 Attachment Page 7 of 17 LCO 3.1.5 Each shutdown bank shall be withdrawn to at least 225 steps. The cycle specific fully withdrawn position shall be 227 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 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 227 steps and the overlap limit during withdrawal shall be 99 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.l Verify SDM to be~ 1.77 % Ak/k.
SR 3.1.6.1 SR 3.1.6.2 SR 3.1.6.3 Verify estimated critical control bank position is within the insertion limits specified in LCO 3.1.6 above.
Verify each control bank is within the insertion limits specified in LCO 3.1.6 above.
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 SOM is~ 1. 77 % Ak/k.
SR 3.1.9.4 Verify SOM to be~ 1.77 % Ak/k.
230 220 210 200 190 180 170 160
-c 150 3
- a. 140 41 ti 130 r: 120 0
E 110 Ill 0
o.. 100 Q.
90 0...
C, 80
'tJ 0
a::
70 60 so 40 30 20 10 0
/
COLR Figure 3.1-1 North Anna 1 Cycle 31 Control Rod Bank Insertion Limits Full w/d osition = All Rods Out = 227 ste s 0.524, 225,
/
V
/
I/
/ : -BANK
/
/
~v
/
/
~/
Serial No.24-126 Attachment Page 8 of 17 I
1.0,2 25 I
I 1.0, 194
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l/fi-BAI ~K
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V
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o.048,o 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 Fraction of Rated Thermal Power 1
3.2 POWER DISTRIBUTION LIMITS 3.2.1 Heat Flux Hot Channel Factor (Fo(Z))
Serial No.24-126 Attachment Page 9 of 17 LCO 3.2.1 Fo(Z), as approximated by FoE(Z) and FoT(Z), shall be within the limits specified below.
CFQ=2.5 The Heat Flux Hot Channel Factor, Fo(Z), shall be limited by the following relationships:
where:
CFQ
- K(Z)
FQ(Z) ::;
p for P > 0.5 for P::;; 0.5 THERMAL POWER d
p
*an
= RATED THERMAL POWER '
K(Z) = 1. 0 for all core heights, z FoE(Z) is an excellent approximation for FQ(Z) when the reactor is at the steady-state power.
Fo(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 FoT(Z) is:
FJ(Z) = FJ(Z)
- N(Z) where:
N Z = FQ(Z),MaximumConditionl
( )
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.
Serial No.24-126 Attachment Page 10 of 17 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 F Q(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 F QT (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.
Notes:
COLR Table 3.2-1 N1C31 Penalty Factors for Flux Map Analysis Burnup Penalty (MWD/MTU)
Factor%
0-499 2.0 500-1999 3.0 2000-EOC 2.0 Serial No.24-126 Attachment Page 11 of 17
- 1. Penalty Factors are not required for initial power ascensionfl,ux maps.
- 2. All full power maps shall apply a Penalty Factor unless frequent flux mapping is invoked
(~7 EFPD).
COLR Table 3.2-2 N1C31 Required Operating Space Reductions for FaT(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) s; 1%
s; 100%
~ 1.0%
~0.0%
> 1%ands;2%
s; 100%
~2.0%
~0.0%
>2%ands;3%
s; 100%
~3.0%
~0.0%
>3%
s; 20%
NIA NIA
- Axial Flux Difference Limits are provided in COLR Figure 3.2-1
3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (FNiiH)
LCO 3.2.2 FN LiH shall be within the limits specified below.
FNMI S 1.587 {l + 0.3(1 - P)}
THERMAL POWER where:
p = RATED THERMAL POWER SR 3.2.2.1 Verify FNiiH is within limits specified above.
3.2.3 AXIAL FLUX DIFFERENCE (AFD)
Serial No.24-126 Attachment Page 12 of 17 LCO 3.2.3 The AFD in% flux difference units shall be maintained within the limits specified in COLR Figure 3.2-1.
120 110 100 90 80 Q)
~ 70 0 a.
io E 60 Q)
.c "C
50 Q) -"'
a::....
0 40 C
Q)
~
Q)
- a.
30 20 10 0
-30 COLR Figure 3.2-1 North Anna 1 Cycle 31 Axial Flux Difference Limits Serial No.24-126 Attachment Page 13 of 17 r2* 101
\\6, 100)
Unacceptable /
Unacceptable Operation
\\
Operation I
\\
)
Acceptable Operation I
' \\
/1/
\\
(-27, 50)
(+20, 50)
-20
-10 0
10 20 30 Percent Flux Difference (Delta-I)
3.3 INSTRUMENTATION 3.3.1 Reactor Trip System (RTS) Instrumentation TS Table 3.3.1-1 Note 1: Overtemperature 8T Serial No.24-126 Attachment Page 14 of 17 The Overtemperature 8T Function Allowable Value shall not exceed the following nominal trip setpoint by more than 2% of 8 T span, with the numerical values of the parameters as specified below.
where: 8T 8To s
T T'
p P'
is measured RCS 8T, °F is the indicated 8 T at R TP, 0P is the Laplace transform operator, sec-1 is the measured RCS average temperature, 0P is the nominal T avg at R TP, ~ 586.8 °F is the measured pressurizer pressure, psig is the nominal RCS operating pressure, ~ 2235 psig K1 s 1.2715 KJ ~ 0.001145 /psig
't1, 't2 = time constants utilized in the lead-lag controller for Tavg
't1 ~ 23.75 sec
't2 ~ 4.4 sec (1 +'t1S)/(l +'t2S) = Junction generated by the lead-lag controller for Tavg dynamic compensation fi(Af) 2: 0.0291 {-13.0- (qt - qb)}
0 0.0251 {(qt-qb)-7.0}
when (qt-qb) < - 13.0% RTP when - 13.0% RTP s (qt-qb) s +7.0% RTP when (qt-qb) > +7.0% RTP Where qt and qb are percent R TP in the upper and lower halves of the core, respectively, and qt+ qb is the total THERMAL POWER in percent RTP.
TS Table 3.3.1-1 Note 2: Overpower ~T Serial No.24-126 Attachment Page 15 of 17 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.
is measured RCS ~T, °F.
is the indicated ~ T at R TP, °F.
where: ~T
~To s
is the Laplace transform operator, sec-1.
T T'
is the measured RCS average temperature, °F.
is the nominal Tavg at RTP, S 586.8 °F.
K4 S 1.0865 Ks ~ 0.0198 /°F for increasing Tavg K6 ~ 0.00162 /°F when T > T' 0 /°F for decreasing T avg 0 /°F when T ~ T'
, 3 = time constant utilized in the rate lag controller for T avg I'J ~ 9.5 sec I']s I (1 + I"Js) = function generated by the rate lag controller for T avg dynamic compensation fa(Af) = 0, for all AI.
3.4 REACTOR COOLANT SYSTEM (RCS)
Serial No.24-126 Attachment Page 16 of 17 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 SR 3.4.1.2 SR 3.4.1.3 SR 3.4.1.4 Verify pressurizer pressure is greater than or equal to 2205 psig.
Verify RCS average temperature is less than or equal to 591 °F.
Verify RCS total flow rate is greater than or equal to 295,000 gpm.
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 % Ak/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.
SR3.9.l.1 Verify boron concentration is within the limit specified above.
NAPS TECHNICAL REQUIREMENTS MANUAL TRM 3.1 REACTIVITY CONTROL SYSTEMS TR 3.1.1 Boration Flow Paths-Operating Serial No.24-126 Attachment Page 17 of 17 Required Action D.2 Borate to a SHUTDOWN MARGIN~ 1.77 % Ak/k at 200 °F, after xenon decay.