ML14014A100

North Anna, Unit 1, Cycle 24 Core Operating Limits Report, Revision 2

ML14014A100
Person / Time
Site:	North Anna
Issue date:	01/07/2014
From:	Huber T R Dominion, Dominion Resources Services, Virginia Electric & Power Co (VEPCO)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
13-643
Download: ML14014A100 (25)

Text

j"Dominion Dominion Resources

Services, Inc.Innsbrook Technical Center5000 Dominion Boulevard, 2SE, Glen Allen, VA 23060January 7, 2014U. S. Nuclear Regulatory Commission Attention:

Document Control DeskOne White Flint North11555 Rockville PikeRockville, MD 20852-2738 Serial No.NLOS /ETSDocket No.License No.13-64350-338NPF-4VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)

NORTH ANNA POWER STATION UNIT ICYCLE 24 CORE OPERATING LIMITS REPORT. REVISION 2Pursuant to North Anna Technical Specification 5.6.5.d, attached is a copy of theDominion Core Operating Limits Report for North Anna Unit 1 Cycle 24 Pattern BUS,Revision 2.If you have any questions regarding this submittal, please contact Mr. Thomas Shaub at(804) 273-2763.

Sincerely, T. R. Huber, DirectorNuclear Licensing and Operations SupportDominion Resources

Services, Inc.for Virginia Electric and Power Company

Attachment:

1. Core Operating Limits Report for North Anna Unit 1 Cycle 24 -Pattern BUS,Revision 2.Commitments made in this letter: None Serial No. 13-643Docket No. 50-338COLR, North Anna 1 Cycle 24, BUS R2Page 2 of 2cc: U.S. Nuclear Regulatory Commission

-Region IIMarquis One Tower245 Peachtree Center Ave., NE, Suite 1200Atlanta, Georgia 30303-1257 Mr. J. E. Reasor, Jr.Old Dominion Electric Cooperative Innsbrook Corporate Center4201 Dominion Blvd.Suite 300Glen Allen, Virginia 23060NRC Senior Resident Inspector North Anna Power StationDr. V. Sreenivas NRC Project ManagerU. S. Nuclear Regulatory Commission One White Flint NorthMail Stop 08 G-9A11555 Rockville PikeRockville, Maryland 20852-2738 Serial No. 13-643Docket No. 50-338ATTACHMENT ICORE OPERATING LIMITS REPORT FOR NORTH ANNA UNIT 1CYCLE 24 PATTERN BUS, REVISION 2NORTH ANNA POWER STATIONVIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)

Serial No. 13-643Docket No. 50-338N 1 C24 CORE OPERATING LIMITS REPORTINTRODUCTION The Core Operating Limits Report (COLR) for North Anna Unit 1 Cycle 24 has been prepared inaccordance with North Anna Technical Specification 5.6.5. The technical specifications affected bythis report are listed below:TS 2.1.1 Reactor Core Safety LimitsTS 3.1.1 Shutdown Margin (SDM)TS 3.1.3 Moderator Temperature Coefficient (MTC)TS 3.1.4 Rod Group Alignment LimitsTS 3.1.5 Shutdown Bank Insertion LimitTS 3.1.6 Control Bank Insertion LimitsTS 3.1.9 PHYSICS TESTS Exceptions

-Mode 2TS 3.2.1 Heat Flux Hot Channel FactorTS 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (F NA)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 LimitsTS 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 bythe 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 22 Serial No. 13-643Docket No. 50-338REFERENCES

1. VEP-FRD-42, Rev. 2.1-A, "Reload Nuclear Design Methodology,"

August 2003.Methodology for:TS 3.1.1 -Shutdown Margin,TS 3.1.3 -Moderator Temperature Coefficient, TS.3.1.4

-Rod Group Alignment LimitsTS 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 FactorTS 3.5.6 -Boron Injection Tank (BIT) andTS 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 3.2.1 -Heat Flux Hot Channel Factor3. WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model Using theNOTRUMP Code," August 1985.Methodology for: TS 3.2.1 -Heat Flux Hot Channel Factor4. WCAP-10079-P-A, "NOTRUMP, A Nodal Transient Small Break and General NetworkCode," August 1985.Methodology for: TS 3.2.1 -Heat Flux Hot Channel Factor5. WCAP-12610-P-A, "VANTAGE+

FUEL ASSEMBLY

-REFERENCE CORE REPORT,"April 1995.Methodology for:TS 2.1.1 -Reactor Core Safety LimitsTS 3.2.1 -Heat Flux Hot Channel Factor6. VEP-NE-2, Rev. 0-A, Statistical DNBR Evaluation Methodology, June 1987.Methodology for:TS 3.2.2 -Nuclear Enthalpy Rise Hot Channel Factor andTS 3.4.1 -RCS Pressure, Temperature and Flow DNB LimitsPage 2 of 22 Serial No. 13-643Docket No. 50-3387. VEP-NE-1, Rev. 0.1-A, Relaxed Power Distribution Control Methodology and Associated FQSurveillance Technical Specifications, August 2003.Methodology for:TS 3.2.1 -Heat Flux Hot Channel Factor andTS 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 andTS 3.3.1 -Reactor Trip System Instrumentation

9. WCAP-14483-A, Generic Methodology for Expanded Core Operating Limits Report, January1999.Methodology for:TS 2.1.1 -Reactor Core Safety Limits,TS 3.1.1 -Shutdown Margin,TS 3.1.4 -Rod Group Alignment LimitsTS 3.1.9 -Physics Tests Exceptions

-Mode 2TS 3.3.1 -Reactor Trip System Instrumentation, TS 3.4.1 -RCS Pressure, Temperature, and Flow DNB LimitsTS 3.5.6 -Boron Injection Tank (BIT) andTS 3.9.1 -Boron Concentration

10. BAW-10227P-A, Rev. 0, "Evaluation of Advanced Cladding and Structural Material (M5) inPWR Reactor Fuel," February 2000.Methodology for:TS 2.1.1 -Reactor Core Safety Limits andTS 3.2.1 -Heat Flux Hot Channel Factor11. EMF-2103 (P) (A), Rev. 0, "Realistic Large Break LOCA Methodology for Pressurized Water Reactors,"

April 2003.Methodology for: TS 3.2.1 -Heat Flux Hot Channel Factor12. EMF-96-029 (P) (A), Rev. 0, "Reactor Analysis System for PWRs," January 1997.Methodology for: TS 3.2.1 -Heat Flux Hot Channel FactorPage 3 of 22 Serial No. 13-643Docket No. 50-33813. BAW- 10168P-A, Rev. 3, "RSG LOCA -BWNT Loss-of-Coolant Accident Evaluation Model for Recirculating Steam Generator Plants,"

December 1996. Volume II only(SBLOCA models).Methodology for: TS 3.2.1 -Heat Flux Hot Channel Factor14. DOM-NAF-2, Rev. 0.2- P-A, "Reactor Core Thermal-Hydraulics Using the VIPRE-DComputer Code," including Appendix A, "Qualification of the F-ANP BWU CHF Correlations in the Dominion VIPRE-D Computer Code," and Appendix C, "Qualification of theWestinghouse WRB-2M CHF Correlation in the Dominion VIPRE-D Computer Code,"August 2010.Methodology for:TS 3.2.2 -Nuclear Enthalpy Rise Hot Channel Factor andTS 3.4.1 -RCS Pressure, Temperature and Flow DNB Limits15. WCAP- 12610-P-A and CENPD-404.-P-A, Addendum 1 -A, "Optimized ZIRLOTM,

July 2006.Methodology for:TS 2.1.1 -Reactor Core Safety Limits andTS 3.2.1 -Heat Flux Hot Channel FactorNote: In some instances, the North Anna COLR lists multiple methodologies that are used toverify a single Technical Specification parameter.

This is due to the transition from AREVAfuel to Westinghouse fuel which requires the use of different vendor proprietary methodologies to verify the two fuiel products meet the applicable regulatory limits.Page 4 of 22 Serial No. 13-643Docket No. 50-3382.0 SAFETY LIMITS (SLs)2.1 SLs2.1.1 Reactor Core SLsIn MODES 1 and 2, the combination of THERMAL POWER, Reactor CoolantSystem (RCS) highest loop average temperature, and pressurizer pressure shallnot exceed the limits specified in COLR Figure 2.1-1; and the following SLsshall 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 DNBcorrelations and methodologies specified in the References Section.2.1.1.2 The peak fuel centerline temperature shall be maintained

< 5080'F,decreasing by 58'F per 10,000 MWD/MTU of burnup, for Westinghouse fuel and < 5173°F, decreasing by 65'F per 10,000 MWD/MTU of burnup,for AREVA fuel.Page 5 of 22 Serial No. 13-643Docket No. 50-338COLR Figure 2.1-1NORTH ANNA REACTOR CORE SAFETY LIMITS4-Eba(Uw,6656606556506456406356306256206156106056005955905855805755700 10 20 30 40 50 60 70 80Percent of RATED THERMAL POWER90 100 110 120Page 6 of 22 Serial No. 13-643Docket No. 50-3383.1 REACTIVITY CONTROL SYSTEMS3.1.1 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 limitof MTC is +0.6 x 10-4 Ak/k/IF, when < 70% RTP, and 0.0 Ak/k/OF 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/0F when > 70% RTP.The EOC/ARO/RTP-MTC shall be less negative than -5.0 x 10-4 Ak/k/OF (lowerlimit).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/0F [Note 2].The 60 ppm/ARO/RTP-MTC should be less negative than or equal to-4.7 x 104 AkWk/°F [Note 3].SR 3.1.3.2 Verify MTC is within -5.0 x 10-4 Ak'k/°F (lower limit).Note 2: If the MTC is more negative than -4.0 x 10-4 Ak/k/°F, SR 3.1.3.2shall be repeated once per 14 EFPD during the remainder of the fuel cycle.Note 3: SR 3.1.3.2 need not be repeated if the MTC measured at theequivalent of equilibrium RTP-ARO boron concentration of _ 60 ppm isless negative than -4.7 x 10-4 Ak/k/]F.3.1.4 Rod Group Alignment LimitsRequired 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.Page 7 of 22 Serial No. 13-643Docket No. 50-3383.1.5 Shutdown Bank Insertion LimitsLCO 3.1.5 Each shutdown bank shall be withdrawn to at least 225 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 LimitsLCO 3.1.6 Control banks shall be limited in physical insertion as shown in COLR Figure3.1-1. Sequence of withdrawal shall be A, B, C and D, in that order; and theoverlap limit during withdrawal shall be 97 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. 1 Verify SDM to be > 1.77 % Ak/k.SR 3.1.6.1 Verify estimated critical control bank position is within the insertion limitsspecified in COLR Figure 3.1-1.SR 3.1.6.2 Verify each control bank is within the insertion limits specified in COLRFigure 3.1-1.SR 3.1.6.3 Verify each control bank not fully withdrawn from the core is within thesequence and overlap limits specified in LCO 3.1.6 above.3.1.9 PHYSICS TESTS Exceptions

-MODE 2LCO 3.1.9.b SDM is _ 1.77 % Ak/k.SR 3.1.9.4 Verify SDM to be __ 1.77 % Ak/k.Page 8 of 22 Serial No. 13-643Docket No. 50-338COLR Figure 3.1-1North Anna 1 Cycle 24Control Rod Bank Insertion Limits0.1,4.In00W-0.0~23022021020019018017016015014013012011010090807060504030201000.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0Fraction of Rated Thermal PowerPage 9 of 22 Serial No. 13-643Docket No. 50-3383.2 POWER DISTRIBUTION LIMITS3.2.1 Heat Flux Hot Channel Factor (FQ(Z))LCO 3.2.1 FQ(Z), as approximated by FQM(Z), shall be within the limits specified below.CFQ = 2.32The Measured Heat Flux Hot Channel Factor, FQM(Z), shall be limited by the following relationships:

CFQ K(Z)P N(Z)) for e>0.5P N(Z)CFQ K(Z)F0 (Z) -for P0.5Q0.5 N( Z)THERMAL POWERwhere: P = RATED THERMAL POWER ; andK(Z) is provided in COLR Figure 3.2-1N(Z) is a cycle-specific non-equilibrium multiplier on FQ (Z) to account for powerdistribution transients during normal operation, provided in COLR Table 3.2-1.The discussion in the Bases Section B 3.2.1 for this LCO requires the application of a cycledependent non-equilibrium multiplier, N(Z), to the CFQ limit. N(Z) accounts for power distribution transients encountered during normal operation.

As function N(Z) is dependent on the predicted equilibrium FQ(Z) and is sensitive to the axial power distribution, it is typically generated from theactual EOC burnup distribution that can only be obtained after the shutdown of the previous cycle.The cycle-specific N(Z) function is presented in COLR Table 3.2-1.Page 10 of 22 Serial No. 13-643Docket No. 50-338COLR Table 3.2-1N1C24 Normal Operation N(Z)NODE HEIGHT(FEET)10 10.211 10.012 9.813 9.614 9.415 9.216 9.017 8.818 8.619 8.420 8.221 8.022 7.823 7.624 7.425 7.226 7.027 6.828 6.629 6.430 6.231 6.032 5.833 5.634 5.435 5.236 5.037 4.838 4.639 4.440 4.241 4.042 3.843 3.644 3.445 3.246 3.047 2.848 2.649 2.450 2.251 2.052 1.80 to 1000MWD/MTU1.1281.1281.1331.1401.1431.1441.1501.1551.1581.1591.1621.1621.1621.1601.1571.1521.1471.1451.1431.1341.1231.1181.1131.1001.0921.0921.0961.0991.1011.1021.1021.1041.1131.1271.1371.1461.1571.1701.1821.1931.2031.2131.2221000 to 3000MWD/MTU1.1391.1471.1551.1611.1661.1681.1731.1761.1771.1751.1731.1691.1641.1601.1571.1521.1471.1451.1431.1351.1231.1161.1081.0911.0791.0771.0801.0821.0861.0911.0971.1041.1121.1211.1281.1351.1441.1541.1631.1721.1801.1871.1953000 to 5000MWDIMTU1.1431.1481.1541.1611.1651.1681.1731.1761.1761.1761.1761.1751.1751.1741.1711.1671.1631.1611.1581.1501.1371.1311.1211.0981.0811.0771.0791.0791.0821.0871.0911.0961.1041.1141.1221.1291.1351.1421.1531.1641.1741.1831.1925000 to 7000MWD/MTU1.1441.1481.1521.1561.1561.1581.1701.1811.1861.1851.1861.1851.1851.1801.1731.1671.1631.1621.1621.1581.1521.1491.1421.1261.1121.1071.1091.1111.1141.1151.1151.1151.1131.1151.1241.1371.1471.1551.1581.1661.1831.1951.1977000 to 9000MWD/MTU1.1211.1281.1371.1481.1531.1571.1701.1811.1861.1851.1851.1841.1831.1841.1871.1891.1891.1901.1881.1851.1781.1751.1671.1481.1311.1221.1201.1211.1231.1211.1201.1201.1161.1151.1231.1371.1471.1551.1591.1671.1831.1951.197These decks are generated for normal operation flux maps that are typically taken at full powerARO. Additional N(z) decks may be generated, if necessary, consistent with the methodology described in the RPDC topical (Reference 7). EOR is defined as Hot Full Power End ofReactivity.

Page 11 of 22 Serial No. 13-643Docket No. 50-338COLR Table 3.2-1 (continued)

N1C24 Normal Operation N(Z)NODE HEIGHT(FEET)10 10.211 10.012 9.813 9.614 9.415 9.216 9.017 8.818 8.619 8.420 8.221 8.022 7.823 7.624 7.425 7.226 7.027 6.828 6.629 6.430 6.231 6.032 5.833 5.634 5.435 5.236 5.037 4.838 4.639 4.440 4.241 4.042 3.843 3.644 3.445 3.246 3.047 2.848 2.649 2.450 2.251 2.052 1.89000 to 11000MWDIMTU1.1191.1181.1231.1331.1381.1431.1561.1681.1731.1751.1801.1821.1821.1831.1871.1891.1891.1901.1881.1851.1781.1751.1671.1471.1321.1301.1311.1301.1271.1231.1191.1191.1181.1221.1301.1441.1561.1681.1761.1851.1991.2081.20911000 to 13000MWD/MTU1.1201.1181.1211.1271.1281.1321.1441.1581.1641.1701.1821.1891.1921.1931.1971.1981.1981.1971.1951.1961.1971.1981.1941.1841.1731.1681.1591.1451.1311.1221.1231.1331.1421.1501.1561.1611.1631.1681.1741.1851.1991.2081.20913000 to 15000MWD/MTU1.1121.1111.1081.1091.1111.1201.1391.1591.1651.1701.1821.1891.1921.1931.1971.1981.1981.1971.1951.1961.1971.1981.1941.1841.1731.1681.1631.1551.1451.1341.1301.1341.1411.1501.1561.1611.1651.1671.1691.1701.1701.1711.17315000 to 17000MWD/MTU1.1191.1191.1181.1171.1121.1131.1231.1381.1461.1571.1751.1881.1931.2001.2111.2171.2191.2211.2221.2221.2191.2181.2111.1961.1811.1741.1691.1611.1541.1481.1441.1401.1341.1281.1221.1221.1301.1451.1531.1631.1751.1861.19417000 to EORMWDIMTU1.1191.1191.1181.1171.1121.1131.1231.1391.1481.1601.1791.1941.2001.2071.2181.2261.2281.2311.2311.2321.2291.2291.2231.2091.1941.1881.1821.1721.1641.1591.1541.1481.1351.1241.1221.1331.1451.1601.1701.1811.1961.2091.220These decks are generated for normal operation flux maps that are typically taken at full powerARO. Additional N(z) decks may be generated, if necessary, consistent with the methodology described in the RPDC topical (Reference 7). EOR is defined as Hot Full Power End ofReactivity.

Page 12 of 22 Serial No. 13-643Docket No. 50-3381.21.11.00.90.8N0 0.7M.'N--I40.60Z-0.5N0.40.30.20.10.0COLR Figure 3.2-1K(Z) -Normalized FQ as a Function of Core Height6,1.0(12 .925)0 1 2 3 4 5 6 7 8 9 10 11 12 13CORE HEIGHT (FT)Page 13 of 22 Serial No. 13-643Docket No. 50-3383.2.2 Nuclear Enthalpy Rise Hot Channel Factor (FNAH)LCO 3.2.2 FNAH shall be within the limits specified below.FNA _< 1.587(1 + 0.3(1 -P)}THERMAL POWERwhere: P RA TED THERMAL POWERSR 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 the applicable COLR Figure (3.2-2-A or 3.2-2-B).

Page 14 of 22 Serial No. 13-643Docket No. 50-338COLR Figure 3.2-2-AApplicable Burnup: BOC* to 5000 MWD/MTUNorth Anna 1 Cycle 24Axial Flux Difference Limits1201101000C-ra,09080706050403020100-30 10 0 1020Percent Flux Difference (Delta-I) 30*Figure 3.2.-2-A was implemented at a core bumup of approximately 2000 MWD/MTU.Page 15 of 22 Serial No. 13-643Docket No. 50-338COLR Figure 3.2-2-BApplicable Burnup: 5000 MWD/MTU to EOCNorth Anna 1 Cycle 24Axial Flux Difference Limits1201101000E-09080706050403020100(-12,100)

(+6,10))Unacceptabl/

eUnateiont Unacceptable Operation Acc ptable Operation

(-27,50)

(+20,50)-30-20-100102030Percent Flux Difference (Delta-I)

Page 16 of 22 Serial No. 13-643Docket No. 50-3383.3.1 Reactor Trip System (RTS) Instrumentation TS Table 3.3.1-1 Note 1: Overtemperature ATThe 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 theparameters as specified below.AT<~C +~ {K K)[TT,]+K3 (P-p,)_fl(A)}

where: ATAT0STT'PP1is measured RCS AT, OFis the indicated AT at RTP, OFis the Laplace transform

operator, sec1is the measured RCS average temperature, OFis the nominal Tavg at RTP, < 586.8 OFis the measured pressurizer

pressure, psigis the nominal RCS operating

pressure,

_> 2235 psigKi < 1.2715K2>- 0.02174 /OFK3 -> 0.001145

/psigr/, r2 = time constants utilized in the lead-lag controller for TavgTi ! 23.75 sec T2 -4.4 sec(1 + rls)/(J + r2s) = function generated by the lead-lag controller for Tag dynamiccompensation f1 (AI) > 0.0291 {- 13.0 -(Cit -qb)}00.0251 {(qt -qb) -7.0}when (qt -qb) < -13.0% RTPwhen -13.0% RTP < (qt -qb) -+7.0% RTPwhen (qt- qb) > +7.0% RTPWhere 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 17 of 22 Serial No. 13-643Docket No. 50-338TS Table 3.3.1-1 Note 2: Overpower ATThe Overpower AT Function Allowable Value shall not exceed the following nominaltrip setpoint by more than 2% of AT span, with the numerical values of the parameters asspecified below.AT<ATo {K4a-K5[ V3S[]T-K6[T-T']-f2(AI)}where: ATAToTT'is measured RCS AT, °F.is the indicated AT at RTP, OF.is the Laplace transform

operator, secl.is the measured RCS average temperature, OF.is the nominal Tavg at RTP, <586.8 OF.K4:5 1.0865K5 >0.0198 /OF for increasing Tavg0 /OF for decreasing TavgK6 >0.001620 /OF/OF when T > T'when T < T'r3= time constant utilized in the rate lag controller for TagT3 9.5 secr3s/(1 + r3s) = function generated by the rate lag controller for dynamiccompensation f2(AI) = 0, for all Al.Page 18 of 22 Serial No. 13-643Docket No. 50-3383.4 REACTOR COOLANT SYSTEM (RCS)3.4,1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) LimitsLCO 3.4.1 RCS DNB parameters for pressurizer

pressure, RCS average temperature, andRCS 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 OF; andc. 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 OF.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 _> 295,000gpm.Page 19 of 22 Serial No. 13-643Docket No. 50-3383.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)3.5.6 Boron Injection Tank (BIT)Required Action B.2Borate to a SDM _> 1.77 % Ak/k at 200 OF.Page 20 of 22 Serial No. 13-643Docket No. 50-3383.9 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 21 of 22 Serial No. 13-643Docket No. 50-338NAPS TECHNICAL REQUIREMENTS MANUALTRM 3.1 REACTIVITY CONTROL SYSTEMSTR 3.1.1 Boration Flow Paths -Operating Required Action D.2 Borate to a SHUTDOWN MARGIN >_ 1.77 % Ak/k at 200 IF,after xenon decay.Page 22 of 22