ML21099A156

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Core Operating Limits Report Cycle 29, Pattern Cvd, Revision 0
ML21099A156
Person / Time
Site: North Anna Dominion icon.png
Issue date: 04/08/2021
From: Standley B
Virginia Electric & Power Co (VEPCO)
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
21-091
Download: ML21099A156 (21)
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VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 April 8, 2021 United States Nuclear Regulatory Commission Serial No.: 21-091 Attention: Document Control Desk NRA/MLW: R1 Washington, D.C. 20555 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 CYCLE 29, PATTERN CVD, REVISION 0 Pursuant to North Anna Technical Specification 5.6.5.d, attached is a copy of the Core Operating Limits Report (COLR) for North Anna Unit 1 Cycle 29, Pattern CVD, Revision

0. The COLR was revised to reflect the reload design evaluations supporting Cycle 29.

If you have any questions or require additional information, please contact Mr. Michael Whitlock at (804) 273-3123.

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

Attachment:

COLR-N1 G29, Revision 0, Core Operating Limits Report, North Anna Unit 1 Cycle 29, Pattern CVD Commitment Summary: There are no new commitments contained in this letter.

Serial No.: 21-091 Docket No.: 50-338 COLR N1 C29 Pattern CVD, Rev. 0 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. Edward Miller NRG Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North Mail Stop 08 B1-A 11555 Rockville Pike Rockville, Maryland 20852-2738 NRG Senior Resident Inspector North Anna Power Station

Serial No.: 21-091 Docket No.: 50-338 Page 1 of 19 ATTACHMENT COLR-N1 C29, Revision 0 CORE OPERATING LIMITS REPORT North Anna Unit 1 Cycle 29 Pattern CVD North Anna Power Station Unit 1 Virginia Electric and Power Company

Serial No.: 21-091 Docket No.: 50-338 Attachment, page 2 of 19 N1C29 CORE OPERATING LIMITS REPORT INTRODUCTION The Core Operating Limits Report (COLR) for North Anna Unit 1 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 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 (FNAH)

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 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-N1C29, Revision 0 EVAL-ENG-RSE-N1C29, Revision 0, Attachment A

Serial No.: 21-091 Docket No.: 50-338 Attachment, page 3 of 19 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.l 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. WCAP-10054-P-A, "Westinghouse Small Break:ECCS Evaluation Model Using the NOTRUMP Code," August 1985.

Methodology for: TS 3.2.1-Heat Flux Hot Channel Factor

4. WCAP-10079-P-A, "NOTRUMP, A Nodal Transient Small Break and General Network Code,"

August 1985.

Methodology for: TS 3.2.1- Heat Flux Hot Channel Factor

5. WCAP-12610-P-A, "VANTAGE+ FUELASSEMBLY REFERENCE CORE REPORT,"

April 1995.

Met'1.odology for:

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

6. 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-N1 C29, Revision O EVAL-ENG-RSE-N1 C29, Revision 0, Attachment A

Serial No.: 21-091 Docket No.: 50-338 Attachment, page 4 of 19

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

Methodology for:

TS 3.2.1- Heat Flux Hot Channel Factor and TS 3 .2.3 - Axial Flux Difference

8. WCAP-8745-P-A, "Design Bases for the Thermal Overpower AT and Thermal Overtemperature AT Trip Functions," September 1986.

Methodology for:

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

9. 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

10. 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

11. 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-N1C29, Revision 0 EVAL-ENG-RSE-N1 C29, Revision 0, Attachment A

Serial No.: 21-091 Docket No.: 50-338 Attachment, page 5 of 19 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-N1C29, Revision O EVAL-ENG-RSE-N1C29, Revision 0, Attachment A

Serial No.: 21-091 Docket No.: 50-338 Attachment, page 6 of 19 COLR Figure 2.1-1 NORTH ANNA REACTOR CORE SAFETY LIMITS 665 660

~~

655 650

~

,......... psia 645

~ ......... ~

640

-~ ...........

.... ~ o psia ~"'-..... ....

635

'u:' ~ ........

~

~ 630 r---....... \

....:::sra ....

... 625

-....... ......... ....... ~

~ \

(II a.

E 620

............. 2000 psia

~ ....... \'

~ ~~

(II bl)

...racu 615 610 r--..... 1860 osia ............... \ \'

I\,

~

1ii V,

V,

~ 600 605

............. r--......

~ \

595 590

" ... ' \ -... '-.

\.

585

\\

580 \

575 570 0 10 20 30 40 50 60 70 80 90 100 110 120 Percent of RATED THERMAL POWER COLR-N1C29, Revision O EVAL-ENG-RSE-N 1C29, Revision 0, Attachment A

Serial No.: 21-091 Docket No.: 50-338 Attachment, page 7 of 19 3.1 REACTMTY 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)

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~ +0.6 x 10-4 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 104 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 s 60 ppm is less negative than-4.7 x 104 Ak/k/°F.

COLR-N1 C29, Revision 0 EVAL-ENG-RSE-N1 C29, Revision 0, Attachment A

Serial No.: 21-091 Docket No.: 50-338 Attachment, page 8 of 19 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 fusertion 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 2:: 1.77 % .Ak/k.

SR3.l.5.1 Verify each shutdown bank is withdrawn to at least 230 steps.

3.1.6 Control Bank fusertion Limits LCO 3.1.6 Control banks shall be limited in physical insertion as shown in COLRFigure 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 2:: 1.77 % Ak/k.

Required Action B.1.1 Verify SDM to be 2:: 1.77 % Ak/k.

Required Action C.1 Verify SDM to be;;:: 1.77 % .Ak/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 2:: 1.77 % Ak/k.

SR 3.1.9.4 Verify SDM to be;;:: 1.77 % .Ak/k.

COLR-N1C29, Revision O EVAL-ENG-RSE-N1C29, Revision 0, Attachment A

Serial No.: 21-091 Docket No.: 50-338 Attachment, page 9 of 19 COLR Figure 3.1-1 North Anna 1 Cycle 29 Control Rod Bank Insertion Limits Fully w/d position= 230 steps 230 / 0.549,230 220 210 /

200

/

190

/ 1.0, 194 180 / /

/

170 / C-BANK 160  !/ /

.'.!:..150 / ,,v

==

51' 140 / /

t;

=Ii: 130 / ,J /

c 0

~ 120 ~,G:18

/

0

~110

,J/

e 100 I.!)

/

"CJ Vo-BAr K

~ 90 80 /

70 V

60

/

50 V

40 /

30

/

20

/

10 /

/o.048,0 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 COLR-N1 C29, Revision 0 EVAL-ENG-RSE-N1C29, Revision 0, Attachment A

Serial No.: 21-091 Docket No.: 50-338 Attachment, page 1O of 19 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:

for P > 0.5 for P :s; 0.5 THERMAL POWER where: p= RATED THERMAL POWER ; autl 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:

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.

COLR-N1C29, Revision 0 EVAL-ENG-RSE-N1C29, Revision 0, Attachment A

Serial No.: 21-091 Docket No.: 50-338 Attachment, page 11 of 19 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 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 1'Q(Z)IK(Z) versus burnup. A minimum value of2% is used should any increase in steady-state or transient measured or predictedpealdng 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 pealdng 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-N1C29, Revision 0 EVAL-ENG-RSE-N1 C29, Revision 0, Attachment A

Serial No.: 21-091 Docket No.: 50-338 Attachment, page 12 of 19 COLR Table 3.2-1 N1C29 Penalty Factors for Flux Map Analysis Burnup Penalty (MWD/MTU) Factor%

0-499 2.0 500-999 5.5 1000-1999 4.5 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:

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

($7EFPD).

COLR Table 3.2-2 N1C29 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:S 1% :S 98.0% 2: 0.5% 2: 1.5%

> 1%and:S2% :S 96.0% 2: 1.0% 2:2.5%

> 2%~nd:S3% :S 94.0% 2: 1.0% 2: 3.0%

>3% :S 50% NIA NIA

  • Axial Flux Difference Limits are provided in COLR Figure 3.2-2 COLR-N1 C29, Revision 0 EVAL-ENG-RSE-N1 C29, Revision 0, Attachment A

Serial No.: 21-091 Docket No.: 50-338 Attachment, page 13 of 19 COLR Figure 3.2-1 K(Z)

  • Normalized FQ as a Function of Core Height 1.2 I

i 1.1 6, 1.0) 1.0 I

--+--- r--- r---

I r--- r-----..

0.9 (12 . .925)

I 0.8 i N

ci LL 0.7

~

j 0.6 0 I z  :  !

~ 0.5 0.4 0.3 0.2 I 0.1 0.0 +

0 1 2 3 4 5 6 7 8 9 10 11 12 13 CORE HEIGHT (FT)

GOLR-N1G29, Revision O EVAL-ENG-RSE-N1G29, Revision 0, Attachment A

Serial No.: 21-091 Docket No.: 50-338 Attachment, page 14 of 19 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (FN,:\H)

LCO 3.2.2 FNAH shall be within the limits specified below.

FNAH ~ 1.587{1 + 0.3(1-P)}

THERMAL POWER p

where: RATED THERMAL POWER SR 3 .2.2.1 Verify FNAH 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 3.2-2.

COLR-N1 C29, Revision O EVAL-ENG-RSE-N1C29, Revision 0, Attachment A

Serial No.: 21-091 Docket No.: 50-338 Attachment, page 15 of 19 COLR Figure 3.2-2 North Anna 1 Cycle 29 Axial Flux Difference Limits 120 110 100 (-12, 100) \6, 100) 90 I Unacceptable / \ Unacceptable 80 Operation

~ Operation Cl)

)

V \

~ 70 Acceptable Operation ll.

1ij E 60 7 \

/1 \

Cl)

..c:

,,I-Cl) 50 1u (-2 7 , 50) (+20, 50) 0:::

'o C:

40 Cl) 0 Cl) ll. 30 20 10 0

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

COLR-N1C29, Revision 0 EVAL-ENG-RSE-N1C29, Revision 0, Attachment A

Serial No.: 21-091 Docket No.: 50-338 Attachment, page 16 of 19 3.3 INSTRUMENTATION 3.3.1 Reactor Trip System (RTS) Instrumentation TS Table 3.3.1-1 Note 1: Overtemperature AT The Overtemperature AT Function Allowable Value shall not exceed the following nominal trip setpoint by more than 2% of AT span, with the numerical values of the parameters as specified below.

where: AT is measured RCS AT, °F ATo is the indicated AT 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 s 1.2715 K2 ~ 0.02174 /°F K3 ~ 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 +1:2s) = function generated by the lead-lag controller for Tavg dynamic compensation fi(AI) ~ 0.0291 {-13.0- (qt-qb)} when (qt qb) <-13.0% RTP 0 when-13.0% RTP s (qt qb) s +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.

COLR-N 1C29, Revision 0 EVAL-ENG-RSE-N1 C29, Revision 0, Attachment A

Serial No.: 21-091 Docket No.: 50-338 Attachment, page 17 of 19 TS Table 3.3.1-1 Note 2: Overpower AT The Overpower AT Function Allowable Value shall not exceed the following nominal trip setpoint by more than 2% of AT span, with the numerical values of the parameters as specified below.

where: AT is measured RCS AT, °F.

ATo is the indicated AT 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 s 1.0865 Ks ~ 0.0198 /°F for increasing Tavg K6 ~ 0.00162 /°F when T > T' 0 /°F for decreasing Tavg 0 /°F when T s T'

'{3 = time constant utilized in the rate lag controller for Tavg

'C3 ~ 9.5 sec 1:3s I (1 + ns) function generated by the rate lag controller for Tavg dynamic compensation f2(Af) = O, for all AI.

COLR-N1 C29, Revision O EVAL-ENG-RSE-N1C29, Revision 0, Attachment A

Serial No.: 21-091 Docket No.: 50-338 Attachment, page 18 of 19 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.

SR3.4.1.4 ------------------------------NOTE--------------------------------------------

Not required to be performed until 30 days after 2 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 2 2600 ppm.

SR 3.9.1.1 Verify boron concentration is within the limit specified above.

COLR-N1C29, Revision 0 EVAL-ENG-RSE-N1C29, Revision 0, Attachment A

Serial No.: 21-091 Docket No.: 50-338 Attachment, page 19 of 19 NAPS TECHNICAL REQUIREMENTS MANUAL TRM 3 .1 REACTIVITY CONTROL SYSTEMS TR 3 .1.1 Borati on Flow Paths Operating Required Action D.2 Borate to a SHUTDOWN MARGIN~ 1.77 % Ak/k at 200 °F, after xenon decay.

COLR-N1C29, Revision O EVAL-ENG-RSE-N1C29, Revision 0, Attachment A

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