ML022940478

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Revision to Unit 3 Cycle 17 Core Operating Limits Report
ML022940478
Person / Time
Site: Dresden Constellation icon.png
Issue date: 10/03/2002
From: Hovey R
Exelon Generation Co, Exelon Nuclear
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
RHLTR: #02-0070
Download: ML022940478 (22)
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Text

Exel ns Exelon Generation Dresden Generating Station wwwexeloncorp com NEucle a 6500 North Dresden Road Morris, IL60450-9765 Tel 815-942-2920 October 3, 2002 RHLTR: #02-0070 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington D.C. 20555-0001 Dresden Nuclear Power Station, Unit 3 Facility Operating License No. DPR-25 NRC Docket No. 50-249

Subject:

Revision to Unit 3 Cycle 17 Core Operating Limits Report The purpose of this letter is to transmit the revision to the Core Operating Limits Report (COLR) in accordance with Technical Specification (US) Section 5.6.5, "CORE OPERATING LIMITS REPORT (COLR)." The analytical methods used to determine the operating limits have been previously approved by the NRC. The COLR is enclosed as an attachment to this letter.

This revision includes allowance for a final feedwater temperature reduction of 120 degrees F, instead of the current 100 degrees F allowed reduction. On June 6, 2002, Framatome ANP Richland Inc., submitted analyses to Exelon documenting the acceptability of reducing final feedwater temperature by 120 degrees F prior to the implementation of Extended Power Uprate (EPU). There were no changes to the basis for developing the limits as a result of this revision.

Should you have any questions concerning this letter, please contact Mr. J. Hansen at (815) 416-2800.

Respectfully, R.J. Hovey Site Vice President Dresden Nuclear Power Station

Attachment:

Core Operating Limits Report, Dresden Station Unit 3 Cycle 17, dated August 2002 cc: Regional Administrator - NRC Region III NRC Senior Resident Inspector- Dresden Nuclear Power Station

Core Operating Limits Report Dresden Station Unit 3 Cycle 17 August 2002 I

ISSUANCE OF CHANGES

SUMMARY

Affected Section Affected Pages Summary of Changes Date All All Original Issue Cycle 17 10/00 5.2 5-1, 5-2, 5-3 Section 5.2 was revised to be consistent 3/01 with the format used in the D2 C1 7 COLR.

6.0 6-1, 6-2 Added references CTS 6.9.A.6.b.1 6.9.A.6.b.13 (ITS 5.6.5.b.1-5.6.5.b.13),

including the reference revision number and date 1.0 1-2 Added Allowable Value to the title heading of table 1.2-1 to be consistent with ITS.

5 iii, 5-2, 5-4 Revised OLMCPR to reflect analysis 7/01 results in response to Part 21 with notification event # 37874 All ii, 1-1, 1-2, 2-1, Removal of historical Current Technical 2-3, 3-1, 4-1, Specification References, identification 5-1, 5-3, 6-1 of new Technical Specification references, and re-titling to reflect the new Technical Specifications 5 iii, 5-2, 5-4 Insert reference to allow 120°F Final 8/02 Feedwater Temperature Reduction, updated Figure 5.2-2.

Dresden Unit 3 Cycle 17 i August 2002 1

TABLE OF CONTENTS Page REFERENCES ............................................................................................................. iii LIST O F FIG U R ES ......................................................................................................... iv LIST O F TABLES ........................................................................................................... v 1.0 ROD BLOCK MONITOR (3.3.2.1, Table 3.3.2.1-1) .................................................. 1-1 1.1 TECHNICAL SPECIFICATION REFERENCE ................................................. 1-1 1.2 D ESC R IPTIO N.................................................................................................. 1-1 2.0 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (3.2.1, 3.4.1) ............... 2-1 2.1 TECHNICAL SPECIFICATION REFERENCE ................................................. 2-1 2.2 D ESC R IPTIO N .................................................................................................. 2-1 2.3 MAPLHGR LIMIT MULTIPLIERS ..................................................................... 2-1 3.0 LINEAR HEAT GENERATION RATE (3.2.3) ........................................................... 3-1 3.1 TECHNICAL SPECIFICATION REFERENCE ................................................. 3-1 3.2 DESC R IPTIO N .................................................................................................. 3-1 4.0 FUEL DESIGN LIMIT RATIO - CENTERLINE MELT (FDLRC) (3.2.4) ................... 4-1 4.1 TECHNICAL SPECIFICATION REFERENCE ................................................. 4-1 4.2 DESC R IPTIO N.................................................................................................. 4-1 5.0 MINIMUM CRITICAL POWER RATIO (3.2.2, 3.4.1, 3.7.7) ...................................... 5-1 5.1 TECHNICAL SPECIFICATION REFERENCE ................................................. 5-1 5.2 D ESC R IPTIO N.................................................................................................. 5-1 6.0 METHODOLOGY (5.6.5.b) ........................................................................................ 6-1 Dresden Unit 3 Cycle 17 ii August 2002

REFERENCES

1. Commonwealth Edison Company Docket No. 50-249, Dresden Nuclear Power Station, Unit 3, Facility Operating License DPR-25.
2. Letter, D. M. Crutchfield (NRC) to All Power Reactor Licensees and Applicants, Generic Letter 88-16, Concerning the Removal of Cycle-Specific Parameter Limits from Technical Specifications.
3. Dresden LOCA-ECCS Analysis MAPLHGR Limits for ATRIUM-9B and 9x9-2 Fuel, EMF-98-007(P), January 1998, NFS NWIT No. 9800072 Seq. 00.
4. DGOO-001047, Dresden Unit 3 Cycle 17 Plant Transient Analysis, Siemens Document EMF-2406, Revision 0, July 2000.
5. DGOO-001 046, Dresden Unit 3 Cycle 17 Reload Analysis, Siemens Document EMF-2421, Revision 0, July 2000.
6. DGOO-000907, Dresden Unit 3 Cycle 17 Neutronics Licensing Report (NLR), July 28, 2000, TODI No. NFM0000086 Seq. 00.
7. Dresden Units 2 and 3 Generic Coastdown Analysis with ATRIUM-9B, EMF-92-149 (P) and EMF-92-149(P) Supplement 1, Revision 1, September 1996, NFS NDIT No. 960137 Seq. 00.
8. Letter, David Garber (SPC) to Dr. R. J. Chin, "Dresden Operation with Final Feedwater Temperature Reduction," DEG:00:176, July 24, 2000.
9. GE DRF C51-00217-01, Instrument Setpoint Calculation Nuclear Instrumentation Rod Block Monitor, December 1999.
10. Letter, David Garber (SPC) to Dr. R. J. Chin, "Dresden Unit 3 Cycle 17 Evaluation of Fuel Thermal Conductivity (Non-Proprietary Version for Exelon)," DEG:01:079, May 14,2001.
11. Letter, DEG:01:082R, "Clarification of Results for Dresden Unit 3 Cycle 17 Fuel Thermal Conductivity Evaluation - Non-Proprietary Version for Exelon," June 14, 2001.
12. Letter, NFM-MW:02-0271, "Approval of FANP Analysis of D3C1 7 Feedwater Temperature with FHOOS for 120°F Reduction," July 24, 2002.

Dresden Unit 3 Cycle 17 iii August 2002 1

LIST OF FIGURES Figure Title I Description Page 2.2-1 MAPLHGR Limit vs. Planar Average Exposure 2-2 3.2-1 Steady State LHGR (SLHGR) Limit vs. Planar Average 3-2 Exposure 4.2-1 Transient LHGR (TLHGR) Limit vs. Planar Average Exposure 4-2 5.2-1 Operating Limit MCPR for Manual Flow Control 5-3 5.2-2 Main Turbine Bypass Valve OLMCPR Adders 5-4 Dresden Unit 3 Cycle 17 iv August 2002 1

LIST OF TABLES Table Title I Description Page 1.2-1 Control Rod Withdrawal Block Instrumentation Setpoints 1-2 2.3-1 Equipment Out Of Service MAPLHGR Limit Multipliers 2-3 5.2-1 Operating Limit MCPR 5-2 Dresden Unit 3 Cycle 17 V August 2002 1

1.0 ROD BLOCK MONITOR 1.1 Technical Specification Reference 3.3.2.1 - Control Rod Block Instrumentation Table 3.3.2.1 Control Rod Block Instrumentation 1.2 Description The Rod Block Monitor Upscale Instrumentation Setpoints are determined from the relationships shown in Table 1.2-1.

Dresden Unit 3 Cycle 17 1-1 August 2002 1

TABLE 1.2-1 CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION SETPOINTS TRIP FUNCTION: ALLOWABLE VALUE Rod Block Monitor Upscale (Flow Bias)

Dual Loop Operation Less than or equal to (0.65 Wd plus 55)*

Single Loop Operation Less than or equal to (0.65 Wd plus 51)*

  • Wd - percent of drive flow required to produce a rated core flow of 98 MIb/hr.

Dresden Unit 3 Cycle 17 1-2 August 2002 1

2.0 AVERAGE PLANAR LINEAR HEAT GENERATION RATE 2.1 Technical Specification References 3.2.1 - AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) 3.4.1 - Recirculation Loops Operating 2.2 Description The Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limit versus Planar Average Exposure for each fuel type is determined from Figure 2.2-1.

2.3 MAPLHGR Multipliers The appropriate multiplicative factor, during power operation with equipment out of service, to apply to the base MAPLHGR limits specified in Section 2.2 is shown in Table 2.3-1.

Dresden Unit 3 Cycle 17 2-1 August 2002 1

FIGURE 2.2-1 MAPLHGR LIMIT VS PLANAR AVERAGE EXPOSURE 14 12 S10 8

6 0 10 20 30 40 50 60 70 Planar Average Exposure (GWDIMTU)

Planar Average MAPLHGR MAPLHGR Limit (kW/ft)

Exposure Limit (kWIft) ATRIUM-9B (offset &

(GWDIMTU) 9x9-2 non-offset) 0 12.5 13.5 5 12.5 13.5 15 12.5 13.5 20 11.9 13.5 55 7.7 9.3 60 8.7 61.1 8.6 Dresden Unit 3 Cycle 17 2-2 August 2002 1

TABLE 2.3-1 EQUIPMENT OUT OF SERVICE MAPLHGR LIMIT MULTIPLIERS Technical Specifications Scenario Multiplicative Multiplicative Factor, Factor, 9x9-2 ATRIUM-9B (offset & non offset) 3.2.1 AVERAGE PLANAR Single Loop 0.90 0.90 LINEAR HEAT Operation GENERATION (SLO)

RATE (APLHGR) 3.4.1 Recirculation Loops Operating Dresden Unit 3 Cycle 17 2-3 August 2002 1

3.0 LINEAR HEAT GENERATION RATE 3.1 Technical Specification Reference 3.2.3 - LINEAR HEAT GENERATION RATE (LHGR) 3.2 Description The Steady State LHGR (SLHGR) limit versus Planar Average Exposure for each fuel type is determined from Figure 3.2-1.

Dresden Unit 3 Cycle 17 3-1 August 2002 1

FIGURE 3.2-1 STEADY STATE LHGR (SLHGR) LIMIT VS. PLANAR AVERAGE EXPOSURE 6

10 0 10 20 30 40 50 60 70 PNw-ArAw E-xomi.r(GMfJ)

Planar Average SLHGR Limit (kW/ft)

Exposure ATRIUM-9B (offset &

(GWD/MTU) non-offset) 0.0 14.4 150 144 61.1 832 Planar Avorage Exposure SLHGR Limit (kWlft)

(GWDIMTU) 9x9-2 0.0 14.5 5.0 14.5 252 108 480 7.2 550 7.2 Dresden Unit 3 Cycle 17 3-2 August 2002 1

4.0 FUEL DESIGN LIMIT RATIO - CENTERLINE MELT (FDLRC) 4.1 Technical Specification Reference 3.2.4 - Average Power Range Monitor (APRM) Gain and Setpoint 4.2 Description The Transient LHGR (TLHGR) limit versus Planar Average Exposure for each fuel type is determined from Figure 4.2-1.

Dresden Unit 3 Cycle 17 4-1 August 2002 1

FIGURE 4.2-1 TRANSIENT LHGR (TLHGR) LIMIT VS. PLANAR AVERAGE EXPOSURE 3=

E 0 10 20 30 40 50 60 70 Ptw Av-we Expore (OmVRAT9 Planar Average TLHGR Umit (kWlft)

Exposure ATRIUM-9B (offset &

(GWDIMTU) non-offset) 0.0 19A 150 19A 61.1 112 Planar Average Exposure TLHGR Limit (kWIft)

(GWDIMTU) 9x9-2 0.0 19.2 254 16.9 432 10.8 480 10 550 10 Dresden Unit 3 Cycle 17 4-2 August 2002 1

5.0 MINIMUM CRITICAL POWER RATIO 5.1 Technical Specification References 3.2.2 - MINIMUM CRITICAL POWER RATIO (MCPR) 3.4.1 - Recirculation Loops Operating 3.7.7 - The Main Turbine Bypass System 5.2 Description

a. The Operating Limit MCPRs for D3C17 are listed in Table 5.2-1 for 9x9-2 and ATRIUM-9B. The OLMCPRs calculated for D3C17 are based on Technical Specification CRD Scram Insertion Speeds (3.1.4). When necessary for slower than normal bypass valve opening times or for operation with inoperable turbine bypass valves 1 , apply the appropriate Operating Limit MCPR adder provided in Figure 5.2-2.
b. During Manual Flow Control, the Operating Limit MCPR for each fuel type at reduced core flow conditions can be determined from (i) or (ii),

whichever is greater:

i. Figure 5.2-1 using the appropriate flow rate, or ii. The Operating Limit MCPR determined from Table 5.2-1 as appropriate and supplemented by Figure 5.2-2 as appropriate.
c. Automatic Flow Control is not supported for D3C17.
d. Core Flow must be maintained _ 108% of rated.1
e. During operation at core average exposure > 30,837 MWd/MTU, power is limited to the lesser of 100% CTP or the following:
i. Apply the appropriate limits for no CTP overshoot as described in section 5.2.b and monitor and maintain CTP as follows:

CTP (% rated)-<100-10*(currentcoreaverageexposure (MWD/MTU)-30,837 (MWD/MTU)10 CTP  % raed)

(crren cor aveage1000) ii. Apply the appropriate limits for 15% CTP overshoot as described in section 5.2.b and monitor and maintain CTP as follows:

CTP (% rated)< 100-10"(current core average exposure1000(MWD/MTU)-32,337 (MWD/MTU)J

f. The following conditions are supported without penalty:
  • 40% TIP channels unavailable e 50% LPRMs unavailable
  • 4 Safety Valves OOS' 0 1 Relief Valve OOS'
  • 2500 EFPH LPRM Calibration interval (2000 EFPH + 25% Grace) 1 Ensure the unit's licensing basis permits operation in this condition prior to crediting this flexibility.

Dresden Unit 3 Cycle 17 5-1 August 2002 1

TABLE 5.2-1 OPERATING LIMIT MCPR OLMCPR for Core Average Exposure < 30,837 MWd/MTU OR OLMCPR for Core Average Exposure > 30,837 MWd/MTU with no CTP Overshoot Operating Scenario 9x9-2 Fuel ATRIUM-9B1' Operating Limit Operating Limit MCPR MCPR Two Loop 2

Operation and CTP maintained per 1.44 1.44 5.2.e.i Single2Loop Operation and CTP maintained per 1.45 1.45 5.2.e.i OLMCPR for Core Average Exposure > 30,837 MWd/MTU with 15% CTP Overshoot Operating Scenario 9x9-2 Fuel ATRIUM-9B Operating Limit Operating Limit MCPR MCPR Two Loop3 Operation and CTP maintained per 1.47 1.47 5.2.e.ii Single 3Loop Operation and CTP maintained per 1.48 1.48 5.2.e.ii I Includes both offset and non-offset ATRIUM-9B fuel designs.

2 Includes up to 120 degrees F reduced feedwater temperature from normal. For core average I

exposure >30,837 MWd/MTU, the 15% CTP Overshoot limits must be applied if feedwater temperature reduction causes CTP to exceed the limits of 5.2.e.i.

3 Includes up to 120 degrees F reduced feedwater temperature from normal. I Dresden Unit 3 Cycle 17 5-2 August 2002 1

FIGURE 5.2-1 OPERATING LIMIT MCPR FOR MANUAL FLOW CONTROL 3.00 2.50 A

0.

It 2.00 E

1.50 1.00 0 20 40 60 80 100 120 Reactor Core Flow (% Rated, 98 Mlblhr) 110% Maximum Flow ( B3.2.2 ASA & TRM TSR 2.1.a.1)

Reactor Operating Limit MCPR - Flow Core Flow ATRIUM-9B (offset & non-offset)

(% Rated) and 9x9-2 110 1.10 100 1.20 30 1.99 0 2.50 Dresden Unit 3 Cycle 17 5-3 August 2002 I

FIGURE 5.2-2 MAIN TURBINE BYPASS VALVE OLMCPR ADDERS ATRIUM-9B (offset & non-offset) & 9x9-2 007 006 005 004 0.03 0

0.02 0 01 0.00 0 200 400 600 800 1000 1200 1400 Bypass Valve Delay Time Relative to Time of TSV Full Closure (msec)

Bypass Valve (BPV) OLMCPR Adder1 ' 2 (ACPR)

Delay Time (msec) ATRIUM-9B (offset and non-offset) & 9x9-2 0<t<50 0.00 50< t < 89 0.00 89 < t_< 215 0.01 215 <t< 1010 0.02 1010 < t< 1075 0.03 1075 <t< 1149 0.04 1149 <t< 1262 0.05 t > 1262 0.06 Two or more BPV 0.06 inoperable 2

' Includes the effects of one BPV inoperable with no OLMCPR adjustment required.

2 Ensure the unit's licensing basis permits operation with one or more BPV inoperable prior to utilizing the associated penalty.

Dresden Unit 3 Cycle 17 5-4 August 2002 1

6.0 METHODOLOGY The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in the latest approved revision or supplement of the topical reports describing the methodology. These methodologies are listed in TS section 5.6.5.b, the complete identification for each of the TS referenced topical reports used to prepare the COLR are listed below.

1) ANF-1125 (P)(A) and Supplements I and 2, "ANFB Critical Power Correlation."

Advanced Nuclear Fuels Corporation, April 1990.

2) ANF-524 (P)(A), Revision 2 and Supplements I and 2, "ANF Critical Power Methodology for Boiling Water Reactors." Advanced Nuclear Fuels Corporation, November 1990.
3) XN-NF-79-71 (P)(A) Revision 2, "Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors." Exxon Nuclear Company, November 1981.
4) XN-NF-80-19 (P)(A) Volume 1 Supplement 3, Supplement 3 Appendix F, and Supplement 4, "Advanced Nuclear Fuels Methodology for Boiling Water Reactors: Benchmark Results for the CASMO-3G/MICROBURN-B Calculation Methodology." Advanced Nuclear Fuels Corporation, November 1990.
5) XN-NF-85-67 (P)(A) Revision 1, "Generic Mechanical Design for Exxon Nuclear Jet Pump Boiling Water Reactors Reload Fuel." Exxon Nuclear Company, September 1986.
6) ANF-913 (P)(A) Volume I Revision I and Volume I Supplements 1, 2, 3, and 4, "uCONTRANSA2: A Computer Program for Boiling Water Reactor Transient Analysis." Advanced Nuclear Fuels Corporation, August 1990.
7) XN-NF-82-06 (P)(A), Qualification of Exxon Nuclear Fuel for Extended Bumup Supplement I Extended Bumup Qualifications of ENC 9x9 BWR Fuel, Supplement 1, Revision 2, Advanced Nuclear Fuels Corporation, May 1998.
8) ANF-89-14 (P)(A), Advanced Nuclear Fuels Corporation Generic Mechanical Design for Advance Nuclear Fuels Corporation 9x9-IX and 9x9-9X BWR Reload Fuel, Revision 1 and Supplements 1 and 2, Advanced Nuclear Fuels Corporation, October 1991.
9) ANF-89-98 (P)(A) Generic Mechanical Design Criteria for BWR Fuel Designs, Revision 1 and Revision 1 Supplement 1, Advanced Nuclear Fuels Corporation, May 1995.
10) ANF-91-048 (P)(A), Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors EXEM BWR Evaluation Model, Advanced Nuclear Fuels Corporation, January 1993.
11) Commonwealth Edison Company Topical Report NFSR-0091, "Benchmark of CASMO/MICROBURN BWR Nuclear Design Methods." Revision 0 and Supplements on Neutronics Licensing Analyses (Supplement 1) and LaSalle County Unit 2 Benchmarking (Supplement 2), December 1991, March 1992, and May 1992, respectively.

Dresden Unit 3 Cycle 17 6-1 August 2002 1

12) ANF-1 125 (P)(A) Supplement 1 Appendix E Rev 0, ANFB Critical Power Correlation Determination of ATRIUM-9B Additive Constant Uncertainties, Siemens Power Corporation, September 1998.
13) EMF-85-74 (P) Revision 0, Supplement I (P)(A) and Supplement 2 (P)(A),

RODEX2A (BWR) Fuel Rod Thermal Mechanical Evaluation Model, Siemens Power Corporation, February 1998.

Dresden Unit 3 Cycle 17 6-2 August 2002 1

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