NRC 2007-0018, Cycle 31 (U1C31) Core Operating Limits Report

Cycle 31 (U1C31) Core Operating Limits Report

ML071060339
Person / Time
Site:	Point Beach
Issue date:	04/16/2007
From:	Koehl D Nuclear Management Co
To:	Document Control Desk, NRC/NRR/ADRO
References
NRC 2007-0018
Download: ML071060339 (17)

Text

Committed to Nuclear Excellence Point Beach Nuclear Plant Operated by Nuclear Management Company, LLC April 16,2007 NRC 2007-0018 10 CFR 50.36 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington DC 20555 Point Beach Nuclear Plant Unit 1 Docket No. 50-266 License No. DPR-24 Unit 1 Cvcle 31 (U1C31) Core Oeeratina Limits Reeort In accordance with the requirements of Point Beach Nuclear Plant (PBNP)

Technical Specification 5.6.4, "Core Operating Limits Report (COLR),"

Nuclear Management Company, LLC (NMC), is submitting the Core Operating Limits Report for PBNP Unit 1 Cycle 31 (U1C31).

The PBNP U1C31 COLR was issued on April 13,2007.

This letter contains no new commitments and no revisions to existing commitments.

Dennis L. Koehl /

Site Vice-President, Point Beach Nuclear Plant Nuclear Management Company, LLC Enclosure cc: Administrator, Region Ill, USNRC Project Manager, Point Beach Nuclear Plant, USNRC Resident Inspector, Point Beach Nuclear Plant, USNRC PSCW 6610 Nuclear Road Two Rivers, Wisconsin 54241-9516 Telephone: 920.755.2321

ENCLOSURE CORE OPERATING LIMITS REPORT POINT BEACH NUCLEAR PLANT, UNIT 1 CYCLE 31 (UlC31)

TRM 2.1 CORE OPERATING LIMITS REPORT (COLR)

UNIT 1 CYCLE 3 1 REVISION 8

POINT BEACH NUCLEAR PLANT TRM2.1 U1 TECHNICAL REQUIREMENTS MANUAL Revision 8 April 13,2007 CORE OPERATlNG LIMITS REPORT (COLR) 1.0 CORE OPERATING LIMITS REPORT This Core Operating Limits Report (COLR) for Point Beach Nuclear Plant has been prepared in accordance with the requirements of Technical Specification (TS) 5.6.4.

A cross-reference between the COLR sections and the PBNP Technical Specifications affected by this report is given below:

COLR PBNP Description Section TS - 2.1 2.1.1 Reactor Core Safety Limits 2.2 3.1.1 Shutdown Margin 3.1.4 Rod Group Alignment Limits 3.1.5 Shutdown Bank Insertion Limits 3.1.6 Control Bank Insertion Limits 3.1.8 Physics Test Exceptions 3.1.3 Moderator Temperature Coefficient 3.1.5 Shutdown Bank Insertion Limit Control Bank Insertion Limits Nuclear Heat Flux Hot Channel Factor (FQ(Z))

Nuclear Enthalpy Rise Hot Channel Factor ( F ~ ~ H )

Axial Flux Difference (AFD)

Overtemperature AT Setpoint Overpower AT Setpoint RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits 2.12 Refueling Boron Concentration Figure 1 Reactor Core Safety Limits Curve Figure 2 Required Shutdown Margin Figure 3 Control Bank Insertion Limits I Figure 4 Hot Channel Factor Normalized Operating Envelope (K(Z)) for 0.422V+

Fuel I Figure 5 RAOC Summary of W(Z) with HFP AFD Band of -8/+9%

Figure 6 Flux Difference Operating Envelope Page 2 of 15

POINT BEACH NUCLEAR PLANT TRM2.1 U1 TECHNICAL REQUIREMENTS MANUAL Revision 8 April 13,2007 CORE OPERATING LIMITS REPORT (COLR)

I UNIT 1 CYCLE 31 2.0 OPERATING LIMITS The cycle-specific parameter limits for the specifications listed in Section I .0 are presented in the following subsections. These limits have been developed using the NRC approved methodologies specified in Technical Specification 5.6.4.

2.1 Reactor Core Safety Limits (TS 2.1.1)

The combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the limits specified in Figure 1.

Applicability: MODES 1 and 2 2.2.1 SDM shall be within the limits provided in Figure 2.

Applicability: MODES 1,2, and 3 2.2.2 SDM shall be 21 % AWk.

Applicability: MODES 4 and 5 2.3 Moderator Temperature Coefficient (TS 3.1.3) 2.3.1 The upper MTC limits shall be maintained within the limits.

2.3.2 The maximum upper MTC limits shall be:

1 5 pcmI0F for power levels 170% RTP 1 0 pcrnI0F for power levels >70% RTP Applicability: MODE 1 and MODE 2 with bff21 .O.

2.4 Shutdown Bank Insertion Limit (TS 3.1.5)

NOTE: This limit is not applicable while performing SR 3.1.4.2.

2.4.1 Each shutdown bank shall be fully withdrawn.

2.4.2 Fully withdrawn is defined as 2225 steps.

Applicability: MODES 1 and 2 Page 3 of 15

POINT BEACH NUCLEAR PLANT TRM 2.1 U1 TECHNICAL REQUIREMENTS MANUAL Revision 8 April 13,2007 CORE OPERATING LIMITS REPORT (COLR)

I UNIT 1 CYCLE 31 2.5 Control Bank Insertion Limits (TS 3.1.6)

NOTE: This limit is not applicable while performing SR 3.1.4.2.

The control banks shall be within the insertion, sequence and overlap limits specified in Figure 3.

Applicability: MODE 1 and MODE 2 with bff21.0 2.6 Nuclear Heat Flux Hot Channel Factor (Fq(Z)) (TS 3.2.1)

The Heat Flux Hot Channel Factor shall be within the following limits:

FQ(Z)5 CFQ

• K(Z) / P for P > 0.5 FQ(Z) I CFQ

• K(Z) 1 0.5 for P 1 0.5 Where P is the fraction of Rated Power at which the core is operating.

FQ(Z)is both:

Steady State FQ'(Z) = FQ(Z)

• 1.08 Transient F Q ~ ( Z=) F~"(Z)* W(Z)

CFo = 2.60 (422V+ Fuel)

K(Z) is the function in Figure 4.

W(Z) is the function in Figure 5.

Applicability: MODE 1 2.7 Nuclear ~ n t h a l p - r n The Nuclear Enthalpy Rise Hot Channel Factor shall be within the following limit:

2.7.1 F ~ A H4 . 7 7 x [ l + 0.3(1-P)] (422V+ Fuel) where: P is the fraction of Rated Power at which the core is operating.

Applicability: MODE 1 Page 4 of 15

POINT BEACH NUCLEAR PLANT TRM 2.1 U1 TECHNICAL REQUIREMENTS MANUAL Revision 8 April 13,2007 CORE OPERATING LIMITS REPORT (COLR)

) UNIT 1 CYCLE 3 1 2.8 Axial Flux Difference (AFD) (TS 3.2.3)

NOTE: The AFD shall be considered outside limits when two or more OPERABLE excore channels indicate AFD to be outside limits.

The indicated axial flux difference in % flux difference units, shall be maintained within the allowed operational space defined by Figure 6.

Applicability: MODE 1 with THERMAL POWER 250% RTP 2.9 Overtemperature AT Setpoint (TS 3.3.1, Table 3.3.1-1 note 1)

Overtemperature AT setpoint parameter values:

indicated AT at Rated Power, OF average temperature, OF 569.0°F 2235 psig 1.16 0.0149 0.00072 25 sec 3 sec 2 sec for Rosemont or equivalent RTD 0 sec for Sostman or equivalent RTD 2 sec for Rosemont or equivalent RTD 0 sec for Sostman or equivalent RTD f(A1) is an even function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests, where qt and qb are the percent power in the top and bottom halves of the core respectively, and qt + qb is total core power in percent of Rated Power, such that:

2.9.1 For qt - qb within -12, +5 percent, f(A1) = 0.

2.9.2 For each percent that the magnitude of qt - qb exceeds +5 percent, the AT trip setpoint shall be automatically reduced by an equivalent of 2.12 percent of Rated Power.

2.9.3 For each percent that the magnitude of qt - qb exceeds - 12 percent, the AT trip setpoint shall be automatically reduced by an equivalent of 2.0 percent of Rated Power.

Applicability: MODES 1 and 2 Page 5 of 15

POINT BEACH NUCLEAR PLANT TRM 2.1 U l TECHNICAL REQUIREMENTS MANUAL Revision 8 April 13, 2007 CORE OPERATING LIMITS REPORT (COLR)

I UNIT 1 CYCLE 3 1 2.10 Overpower AT Setpoint (TS 3.3.1, Table 3.3.1- 1 note 2)

Overpower AT setpoint parameter values:

indicated AT at Rated Power, OF average temperature, OF 569.0°F 1.10 of Rated Power 0.0262 for increasing T 0.0 for decreasing T 0.00103 for T 2 Tr 0.0 for T < T' 10 sec 2 sec for Rosemont or equivalent RTD 0 sec for Sostman or equivalent RTD 2 sec for Rosemont or equivalent RTD 0 sec for Sostman or equivalent RTD Applicability: MODES 1 and 2 2.1 1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits (TS 3.4.1) 2.1 1.1 T,, shall be 5574°F.

2.1 1.2 Pressurizer pressure shall be maintained 22205 psig during operation.

NOTE: Pressurizer pressure limit does not apply during:

1) THERMAL POWER ramp >5% RTP per minute; or

2) THERMAL POWER step >lo% RTP.

2.1 1.3 Reactor Coolant System raw measured Total Flow Rate shall be maintained 21 82,400 gpm.

Applicability: MODE 1 2.12 Refueling Boron Concentration (TS 3.9.1)

Boron concentrations of the Reactor Coolant System, the refueling canal, and the refueling cavity shall be maintained 22200 ppm.

Applicability: MODE 6

POINT BEACH NUCLEAR PLANT TRM 2.1 U1 TECHNICAL REQUIREMENTS MANUAL Revision 8 April 13,2007 CORE OPERATING LIMITS REPORT (COLR)

I UNIT 1 CYCLE 31 FIGURE 1 REACTOR CORE SAFETY LIMITS CURVE (Cores containing 422V + fuel)

Unaccepbble Operation Acceptable Operation Core Pawer (fraction of 1540 1WWt)

Page 7 of 15

POINT BEACH NUCLEAR PLANT TRM 2.1 U1 TECHNICAL REQUIREMENTS MANUAL Revision 8 April 13,2007 CORE OPERATING LIMITS REPORT (COLR)

I UNIT 1 CYCLE 3 1 FIGURE 3 CONTROL BANK INSERTION LIMITS NOTE: The "fully withdrawn" parking position range 2 225 steps can be used without violating this Figure.

POINT BEACH NUCLEAR PLANT TRM 2.1 U1 TECHNICAL REQUIREMENTS MANUAL Revision 8 April 13,2007 CORE OPERATING LIMITS REPORT (COLR)

I UNIT 1 CYCLE 3 1 FIGURE 4 HOT CHANNEL FACTOR NORMALIZED OPERATING ENVELOPE (K(Z))

FOR 0.422V+ FUEL 0 1 2 3 4 5 6 7 8 9 10 11 12 CORE HEIGHT (FT)

Page 10 of 15

POINT BEACH NUCLEAR PLANT TRM 2.1 U1 TECHNICAL REQUIREMENTS MANUAL Revision 8 April 13,2007 CORE OPERATING LIMITS REPORT (COLR)

I UNIT 1 CYCLE 3 1 FIGURE 5 RAOC Summary of W(Z) with HFP AFD Band of -8/+9%

(Top 15% and Bottom 12% Excluded)

Page 11 of 15

POINT BEACH NUCLEAR PLANT TRM 2.1 U l TECHNICAL REQUIREMENTS MANUAL Revision 8 April 13,2007 CORE OPERATING LIMITS REPORT (COLR)

I UNIT 1 CYCLE 3 1 FIGURE 5 (con't)

RAOC Summary of W(Z) with HFP AFD Band of -8/+9%

(Top 15% and Bottom 12% Excluded)

Page 12 of 15

POINT BEACH NUCLEAR PLANT TRM 2.1 U1 TECHNICAL REQUIREMENTS MANUAL Revision 8 April 13,2007 CORE OPERATING LIMITS REPORT (COLR)

( UNIT 1 CYCLE 3 1 FIGURE 6 FLUX DIFFERENCE OPERATING ENVELOPE

-35.00 -30.00 -25.00 -20.00 -15.00 -10.00 -5.00 0.00 5.00 10.00 15.00 20.00 25.00 30.00 35.00 Delta I (%)

NOTE: This figure represents the Relaxed Axial Offset Control (RAOC) band used in safety analyses, it may be administratively tightened depending on in-core flux map results. Refer to Figure 2 of ROD 1.2 for administrative limit.

Page 13 of 15

POINT BEACH NUCLEAR PLANT TRM 2.1 U1 TECHNICAL REQUIREMENTS MANUAL Revision 8 April 13, 2007 CORE OPERATING LIMITS REPORT (COLR)

I UNIT 1 CYCLE 31 TABLE 1 NRC APPROVED METHODOLOGIES FOR COLR PARAMETERS Surveillance Technical Specification,"

WCAP-14449-P-A, "Application of Best Estimate Large Break LOCA Methodology to Westinghouse PWRs with Upper Plenum Injection," Revision 1, October 1999 (cores containing 422V + fuel)

WCAP-10924-P-A, "Large Break LOCA Best Estimate Methodology, Volume 2: Application to Two-Loop PWRs Equipped with Upper Plenum Injection," and Addenda, December 1988 (cores not containing 422V + fuel)

WCAP-10924-P-A, "LBLOCA Best Estimate Methodology: Model Description and Validation: Model Revisions," Volume 1, Addendum 4, August 1990 (cores not containing 422V + fuel)

WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," August 1985 WCAP-10054-P-A, Addendum 2, Revision 1, "Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection into the Broken Loop and COSI Condensation Model," July 1997 Page 14 of 15

POINT BEACH NUCLEAR PLANT TRM 2.1 U1 TECHNICAL REQUIREMENTS MANUAL Revision 8 April 13,2007 CORE OPERATING LIMITS REPORT (COLR)

I UNIT 1 CYCLE 3 1 TABLE 1 NRC APPROVED METHODOLOGIES FOR COLR PARAMETERS of Constant Axial Offset Control FQ Surveillance Technical Specification,"

AT Trip Functions,"

Overtemperature AT Trip Functions,"

2.1 1 RCS Pressure, Temperature, and WCAP- 11397-P-A, "Revised Thermal Design Flow Departure from Nucleate Procedure," April 1989, for those events Boiling (DNB) Limits analyzed using RTDP WCAP-14787-P, Rev. 2, "Westinghouse Revised Thermal Design Procedure Instrument Uncertainty Methodology for Wisconsin Electric Power Company Point Beach Units 1

& 2 (Fuel Upgrade & Uprate to 1656 MWt-NSSS Power with Feedwater Venturis, or 1679 MWt-MSSS Power with LEFM on Feedwater Header)", October 2002.

WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985 for those events not utilizing RTDP 2.12 Refueling Boron Concentration WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985 Page 15 of 15