ML22188A234
| ML22188A234 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 06/07/2022 |
| From: | Standley B Virginia Electric & Power Co (VEPCO) |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 22-072 | |
| Download: ML22188A234 (21) | |
Text
June 7, 2022 VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 United States Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555 Serial No.:
NRA/ENC:
22-072 RO Docket Nos.:
50-339 License Nos.: NPF-7 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION ENERGY VIRGINIA}
NORTH ANNA POWER STATION UNIT 2 CORE OPERA TING LIMITS REPORT NORTH ANNA 2, CYCLE 29, PATTERN BTN, REVISION 0 The North Anna Power Station Unit 2 Core Operating Limits Report (COLR) has been revised to reflect the reload design evaluations supporting Cycle 29. Therefore, pursuant to Technical Specification 5.6.5.d, attached is a copy of the COLR for North Anna Unit 2, Cycle 29, Pattern BTN, Revision 0.
If you have any questions or require additional information, please contact Ms. Erica N.
Combs at (804) 273-3386.
Sincerely, B. E. Standley, Director Nuclear Regulatory Affairs Dominion Energy Services, Inc. for Virginia Electric and Power Company
Attachment:
COLR-N2C29, Revision 0, Core Operating Limits Report, North Anna Unit 2, Cycle 29, Pattern BTN 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 NRC Senior Project Manager - North Anna Power Station U.S. Nuclear Regulatory Commission One White Flint North Mail Stop 08 B 1-A 11555 Rockville Pike Rockville, Maryland 20852-2738 NRC Senior Resident Inspector North Anna Power Station Serial No.: 22-072 Docket No.: 50-339 Page 2 of 2
ATTACHMENT COLR-N2C29, Revision 0 CORE OPERA TING LIMITS REPORT North Anna Unit 2, Cycle 29, Pattern BTN North Anna Power Station Unit 2 Virginia Electric and Power Company Dominion Energy Virginia Serial No.: 22-072 Docket No.: 50-339
N2C29 CORE OPERA TING LIMITS REPORT INTRODUCTION Serial No.: 22-072 Docket No.: 50-339 Attachment Page 1 of 18 The Core Operating Limits Report (COLR) for North Anna Unit 2 Cycle 29 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 AH)
Axial Flux Difference (AFD)
Reactor Trip System (R TS) 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 totheCOLR:
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.
COLR-N2C29, Revision 0 EVAL-ENG-RSE-N2C29, Revision 0, Attachment A
REFERENCES Serial No.: 22-072 Docket No.: 50-339 Attachment Page 2 of 18
- 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. Plant-specific adaptation of WCAP-16009-P-A, "Realistic Large Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM),"
as approved by NRC Safety Evaluation Report dated February 29, 2012.
Methodology for: TS 3.2.l - 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, "VANT AGE+ 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 COLR-N2C29, Revision 0 EVAL-ENG-RSE-N2C29, Revision 0, Attachment A
Serial No.: 22-072 Docket No.: 50-339 Attachment Page 3 of 18
- 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 3, "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 COLR-N2C29, Revision 0 EVAL-ENG-RSE-N2C29, Revision 0, Attachment A
Serial No.: 22-072 Docket No.: 50-339 Attachment Page 4 of 18 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.
COLR-N2C29, Revision 0 EVAL-ENG-RSE-N2C29, Revision 0, Attachment A
COLR Figure 2.1-1 Serial No.: 22-072 Docket No.: 50-339 Attachment Page 5 of 18 NORTH ANNA REACTOR CORE SAFETY LIMITS 665 660 655 650 645 640
'u:' 635
~
I!! 630
- I...
RI 625 GI a.
E 620
~
GI 615 11.Q RI...
GI 610
<(
"ii 605
~
GI > 600 595 590 585 580
.......... ~
psia
~"""---
,..........,~
psia
~~
~
r--........
-..... ~ I'\\
....... r--,....,..
2000 psia
--....... r\\.\\
~~
\\ ' I\\
I
---.. ~
\\ "
r-,......._
1860 osia
............. r---...._
~ '
...... ~ '\\
'\\.
.................... '\\
~'\\
\\
575 570 I
0 10 20 30 40 50 60 70 80 90 100 110 120 Percent of RATED THERMAL POWER COLR-N2C29, Revision 0 EVAL-ENG-RSE-N2C29, Revision 0, Attachment A
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.: 22-072 Docket No.: 50-339 Attachment Page 6 of 18 LCO 3.1.3 The MTC shall be maintained within the limits specified below. The upper limit ofMTC is +0.6 x 104 Ak/k/°F, when< 70% RTP, and 0.0 Ak/k/°F when~ 70%
RTP.
The BOC/ARO-MTC shall be S +0.6 x 104 Ak/k/°F (upper limit), when< 70%
RTP, and S 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 l].
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.
COLR-N2C29, Revision 0 EVAL-ENG-RSE-N2C29, Revision 0, Attachment A
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.
3.1.5 Shutdown Bank Insertion Limits LCO 3.1.5 Each shutdown bank shall be withdrawn to at least 230 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.
Serial No.: 22-072 Docket No.: 50-339 Attachment Page 7 of 18 SR 3.1.5.1 Verify each shutdown bank is withdrawn to at least 230 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 102 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 COLR Figure 3.1-1.
Verify each control bank is within the insertion limits specified in COLR Figure 3.1-1.
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.
COLR-N2C29, Revision 0 EVAL-ENG-RSE-N2C29, Revision 0, Attachment A
230 220 210 200 190 180 170 160
~ 150 3:
/
/
COLR Figure 3.1-1 North Anna 2 Cycle 29 Control Rod Bank Insertion Limits Fully w/d position = 230 steps
/
0.549,230
/
/
/
/
/
C-BANK I/
Serial No.: 22-072 Docket No.: 50-339 Attachment Page 8 of 18 1.0, 194
~
/
~v
/
,,V
/
g-140 t:
":. 130
/
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C 0 i 120 0
~ 110 0
~ 100
"'C 0
90 cc 80 70 60 50 40 30 20 10 0
~{:18
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o.048,o 0
0.1 0.2 0.3 0.4 0.5 0.6 0.7 Fraction of Rated Thermal Power COLR-N2C29, Revision 0 EVAL-ENG-RSE-N2C29, Revision 0, Attachment A 0.8 0.9 1
3.2 POWER DISTRIBUTION LIMJTS 3.2.1 Heat Flux Hot Channel Factor (F Q(Z))
Serial No.: 22-072 Docket No.: 50-339 Attachment Page 9 of 18 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:
for P > 0.5 for P :5 0.5 where:
THERMAL POWER p = RATED THERMAL POWER ; and K(Z) is provided in COLR Figure 3.2-1 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:
FJ(Z) = FS(Z)
- N(Z) where:
N Z = FQ(Z),Maximum Condition I
( )
FQ(Z),Equilibrium Condition I The discussion in the Bases Section B 3.2.l 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.
COLR-N2C29, Revision 0 EVAL-ENG-RSE-N2C29, Revision 0, Attachment A
The cycle-specific penalty factors are presented in COLR Table 3.2-1.
Serial No.: 22-072 Docket No.: 50-339 Attachment Page 10 of 18 Also discussed is the application of the appropriate factor to account for potential increases in F Q(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 jha mapping is invoked (7 EFP D ).
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.
COLR-N2C29, Revision 0 EVAL-ENG-RSE-N2C29, Revision 0, Attachment A
COLR Table 3.2-1 N2C29 Penalty Factors for Flux Map Analysis Burnup Penalty (MWD/MTU)
Factor%
0-499 2.0 500-999 4.0 1000-1999 2.5 2000-2999 2.0 3000-3999 2.0 4000 - 4999 2.0 5000-6999 2.0 7000-8999 2.0 9000-10999 2.0 11000-12999 2.0 13000-14999 2.0 15000 - 16999 2.0 17000 - 18999 2.0 19000 - EOC
2.0 Notes
Serial No.: 22-072 Docket No.: 50-339 Attachment Page 11 of 18
- 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 N2C29 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%andS 1%
S 98.0%
2: 0.5%
2: 1.0%
> 1% andS2%
S97.0%
2: 1.5%
2: 2.5%
>2%andS3%
S 95.0%
2: 1.5%
2: 3.0%
>3%
S50%
NIA NIA
- Axial Flux Difference Limits are provided in COLR Figure 3.2-2 COLR-N2C29, Revision 0 EVAL-ENG-RSE-N2C29, Revision 0, Attachment A
1.2 1.1 1.0 0.9 0.8 N a
- u. 0.7 C w N :J
<C 0.6
~
0 z
~ 0.5
~
0.4 0.3 0.2 0.1 0.0 COLR Figure 3.2-1 K(Z) - Normalized FQ as a Function of Core Height 6, 1.0)
I I
I I
I I
Serial No.: 22-072 Docket No.: 50-339 Attachment Page 12 of 18 I
I (12,.925)
I I
I 0
1 2
3 4
5 6
7 8
CORE HEIGHT (FT) 9 10 11 12 13 COLR-N2C29, Revision 0 EVAL-ENG-RSE-N2C29, Revision 0, Attachment A
3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (FN AH)
LCO 3.2.2 FNI\\H shall be within the limits specified below.
FNMI S 1.587{1 + 0.3(1 - P)}
THERMAL POWER where:
p = RATED THERMAL POWER SR 3.2.2.1 Verify FNI\\H is within limits specified above.
3.2.3 AXIAL FLUX DIFFERENCE (AFD)
Serial No.: 22-072 Docket No.: 50-339 Attachment Page 13 of 18 LCO 3.2.3 The AFD in % flux difference units shall be maintained within the limits specified in COLR Figure 3.2-2.
COLR-N2C29, Revision 0 EVAL-ENG-RSE-N2C29, Revision 0, Attachment A
120 110 COLR Figure 3.2-2 North Anna 2 Cycle 29 Axial Flux Difference Limits Serial No.: 22-072 Docket No.: 50-339 Attachment Page 14 of 18 100 90 r2* 101
\\6. 100)
Cl)
~
0 D.
"iii E
Cl)
.c I-Cl)..
IV 0::....
0 -
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~
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D.
80 70 60 50 40 30 20 10 0
-30 Unacceptable I Operation I
j I
/
J
(-27. 50)
-20 COLR-N2C29, Revision 0 Unacceptable
\\
Operation
\\
Acceptable Operation
' \\ \\
(+20, 50)
-10 0
10 20 Percent Flux Difference (Delta-I)
EVAL-ENG-RSE-N2C29, Revision 0, Attachment A 30
3.3 INSTRUMENTATION 3.3.1 Reactor Trip System (RTS) Instrumentation TS Table 3.3.1-1 Note 1: Overtemperature 8T Serial No.: 22-072 Docket No.: 50-339 Attachment Page 15 of 18 The Overtemperature 8T Function Allowable Value shall not exceed the following nominal trip setpoint by more than 2% of 8T span, with the numerical values of the parameters as specified below.
where: 8T 8To s
T T'
p P'
is measured RCS 8T, 0 P is the indicated 8T at RTP, 0P is the Laplace transform operator, sec-1 is the measured RCS average temperature, °F is the nominal Tavg at R TP, S 586.8 °F is the measured pressurizer pressure, psig is the nominal RCS operating pressure, ~ 223S psig K1::; 1.271S K2 ~ 0.02174 /°F K3 ~ 0.00114S /psig
't1, 't2 = time constants utilized in the lead-lag controller for Tavg
't1 ~ 23.7S sec
't2 S 4.4 sec (1 +t1s)/(l +'t2S) = function generated by the lead-lag controller for T avg dynamic compensation fi(Af) 2: 0.0291 {-13.0- (qt -qb)}
0 0.02S1 {(qt -qb)- 7.0}
when (qt-qb) <-13.0% RTP when -13.0% RTP::; (qt - qb)::; +7.0% RTP 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.
COLR-N2C29, Revision 0 EVAL-ENG-RSE-N2C29, Revision 0, Attachment A
TS Table 3.3.1-1 Note 2: Overpower 8T Serial No.: 22-072 Docket No.: 50-339 Attachment Page 16 of 18 The Overpower 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.
is measured RCS 8T, °F.
is the indicated 8T at RTP, °F.
where: 8T 8To 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.
l<4 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' TJ = time constant utilized in the rate lag controller for T avg
'!'3 ~ 9.5 sec TJs I (1 + 'Os) = function generated by the rate lag controller for T avg dynamic compensation fa(Af) = O, for all Al.
COLR-N2C29, Revision 0 EVAL-ENG-RSE-N2C29, Revision 0, Attachment A
3.4 REACTOR COOLANT SYSTEM (RCS)
Serial No.: 22-072 Docket No.: 50-339 Attachment Page 17 of 18 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.
SR3.4.l.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.
SR 3.9.1.1 Verify boron concentration is within the limit specified above.
COLR-N2C29, Revision 0 EVAL-ENG-RSE-N2C29, Revision 0, Attachment A
NAPS TECHNICAL REQUIREMENTS MANUAL TRM 3.1 TR 3.1.1 REACTIVITY CONTROL SYSTEMS Boration Flow Paths - Operating Serial No.: 22-072 Docket No.: 50-339 Attachment Page 18 of 18 Required Action D.2 Borate to a SHUTDOWN MARGIN~ 1.77 % Ak/k at 200 °F, after xenon decay.
COLR-N2C29, Revision 0 EVAL-ENG-RSE-N2C29, Revision 0, Attachment A