RS-16-199, Supplement to Request for License Amendment Regarding Transition to Areva Fuel

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Supplement to Request for License Amendment Regarding Transition to Areva Fuel
ML16272A376
Person / Time
Site: Dresden, Quad Cities  Constellation icon.png
Issue date: 09/28/2016
From: Simpson P
Exelon Generation Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-16-199
Download: ML16272A376 (15)
v • d • e


Text

Exelon Generation 4300 Winfield Road Warrenville, IL 60555 www.exeloncorp.com RS-16-199 September 28, 2016 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Dresden Nuclear Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-19 and DPR-25 NRC Docket Nos. 50*237 and 50-249 Quad Cities Nuclear Power Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-29 and DPR-30 NRC Docket Nos. 50-254 and 50-265 10 CFR 50.90

Subject:

Supplement to Request for License Amendment Regarding Transition to AREVA Fuel

Reference:

Letter from P.R. Simpson {Exelon Generation Company, LLC (EGC)) to U.S. NRC, "Request for License Amendment Regarding Transition to AREVA Fuel," dated February 6, 2015 In the referenced letter, EGC requested amendments to Renewed Facility Operating License Nos. DPR-19 and DPR-25 for Dresden Nuclear Power Station (DNPS), Units 2 and 3, and Renewed Facility Operating License Nos. DPR-29 and DPR-30 for Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2. The proposed changes support the transition from Westinghouse SVEA-96 Optima2 (Optima2) fuel to AREVA, Inc. (AREVA) ATRIUM 10XM fuel at DNPS and QCNPS. Specifically, EGC proposed to revise Technical Specification (TS) 5.6.5, "Core Operating Limits Report {COLR)," Paragraph b, to delete methodologies that are no longer required and to add the AREVA analysis methodologies to the list of approved methods to be used in determining the core operating limits in the COLR. Also, in support of the planned transition to AREVA ATRIUM 10XM fuel, EGC proposed to revise DNPS and QCNPS TS 3.2.3, "Linear Heat Generation Rate (LHGR)1" and TS 3.7.7, "Main Turbine Bypass System."

During discussions with the NRC it became apparent that there is a need to clarify that the transition to the AREVA ATRIUM 1 OXM fuel will occur at DNPS and QCNPS in the order described in Table 1 below:

September 28, 2016 U.S. Nuclear Regulatory Commission Page 2 Table 1: Schedule for Implementation of AREVA ATRIUM 10XM Fuel at DNPS and QCNPS Station Unit Refueling Outage Outage Schedule Dresden 3

D3R24 Fall 2016 Quad Cities 1

Q1R24 Spring 2017 Dresden 2

D2R25 Fall 2017 Quad Cities 2

Q2R24 Spring 2018 Implementing Cycle 25 25 26 25 Once the changes proposed in the referenced letter are approved, EGC plans to implement Technical Specifications (TS) Pages 3.3.4.1-3, 5.6-3, and 5.6-4 as shown in Attachments 1 and 2 prior to entering into Mode 2 on the first plant startup following Refueling Outages D3R24 and 01 R24, for DNPS and QCNPS, respectively. All other TS pages will be implemented as proposed in Attachments 2 and 3 of the referenced letter prior to entering Mode 2 following Refueling Outages D3R24 and Q1R24.

Following Refueling Outages D2R25 and Q2R24. the interim TS changes described in Attachments 1 and 2 will be superseded with the TS pages proposed in Attachments 2 and 3 of the referenced letter.

In accordance with 10 CFR 50.91 (b), EGC is notifying the State of Illinois of these issues by transmitting a copy of this letter and its attachments to the designated State Official.

EGC has also reviewed the information supporting a finding of no significant hazards consideration, and the environmental consideration, that were previously provided to the NRC in of the referenced letter. The supplemental information provided in this submittal does not affect the bases for concluding that the proposed license amendments do not involve a significant hazards consideration. In addition, the additional information provided in this submittal does not affect the bases for concluding that neither an environmental impact statement nor an environmental assessment needs to be prepared in connection with the proposed amendments.

September 28, 2016 U.S. Nuclear Regulatory Commission Page 3 There are no regulatory commitments contained in this letter. Should you have any questions related to this letter, please contact Mr. Mitchel A Mathews at (630) 657-2819.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 28th day of September 2016.

Respectfully,

~R Patrick R. Simpson Manager - Licensing Attachments:

1. Markup of Dresden Nuclear Power Station, Units 2 and 3 Technical Specifications Pages Showing Proposed Changes to be Implemented During Interim Period Between Unit 3 Refueling Outage D3R24, and Unit 2 Refueling Outage D2R25
2. Markup of Quad Cities Nuclear Power Station, Units 1 and 2 Technical Specifications Pages Showing Proposed Changes to be Implemented During Interim Period Between Unit 1 Refueling Outage Q 1 R24, and Unit 2 Refueling Outage Q2R24 cc:

Regional Administrator-NRC Region Ill NRC Senior Resident Inspector - Dresden Nuclear Power Station NRC Senior Resident Inspector - Quad Cities Nuclear Power Station Illinois Emergency Management Agency - Division of Nuclear Safety

ATTACHMENT 1 Markup of Dresden Nuclear Power Station, Units 2 and 3 Technical Specifications Pages Showing Proposed Changes to be Implemented During Interim Period Between Unit 3 Refueling Outage D3R24, and Unit 2 Refueling Outage D2R25 3.3.4.1-3 5.6-3 5.6-4 DRESDEN INSERT A

ATWS-RPT Instrumentation 3.3.4.1 SURVEILLANCE REQUIREMENTS

NOTE-------------------------------------

When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> provided the associated Function maintains ATWS-RPT trip capability.

SR 3.3.4.1.1 SR 3.3.4.1.2 SR 3.3.4.1.3 SR 3.3.4.1.4 Dresden 2 and 3 SURVEILLANCE Perform CHANNEL CHECK.

Calibrate the trip units.

Perform CHANNEL FUNCTIONAL TEST.

Perform CHANNEL CALIBRATION. The Allowable Values shall be:

a.

Reactor Vessel Water Level-Low Low:

~ -54.15 inches with time delay set to~ 8.3 seconds and~ 9.7 seconds; and

b.

Reactor Vessel Steam Dome Pressure-High:

s 1241 ~*

~

I s 1241 psi g <Unit 2) I s 1198 psig (Unit 3).

3.3.4.1-3 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program (continued)

Amendment No. 237/230

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT CCOLR)

(continued)

3. The LHGR for Specification 3.2.3.
4. Control Rod Block Instrumentation Setpoint for the Rod Block Monitor-Upscale Function Allowable Value for Specification 3.3.2.1.
5. The OPRM setpoints for the trip function for SR 3.3.1.3.3
b.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following Dresden 2 and 3 documents :

..... (A_p_p-li-ca_b_le-to_U_n_i_t 2-0-n-ly.....,)

1.

Commonwealth Edison Company Topical Report NFSR-0091, "Benchmark of CASMO/MICROBURN BWR Nuclear Design Methods.

2.

NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel."

3.

NED0-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," August 1996.

4.

CENPD-300-P-A, "Reference Safety Report for Boiling Water Reactor Reload Fuel.

11

5.

WCAP-16081-P-A, "lOxlO SVEA Fuel Critical Power Experiments and CPR Correlation:

SVEA-96 Optima2."

6.

WCAP-15682-P-A, "Westinghouse BWR ECCS Evaluation Model:

Supplement 2 to Code Description, Qualification and Application."

7~

WCAP-16078-P-A, "Westinghouse BWR ECCS Evaluation Model:

Supplement 3 to Code Description, Qualification and Application to SVEA-96 Optima2 Fuel."

8.

WCAP-15836-P-A, "Fuel Rod Design Methods for Boiling Water Reactors - S_uppl ement 1."

(continued) 5.6-3 Amendment No. 234/227

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT CCOLR>

(continued)

9.

WCAP-15942-P-A, "Fuel Assembly Mechanical Design Methodology for Boiling Water Reactors, Supplement 1 to CENPD-287.

11

10.

CENPD-390-P-A, "The Advanced PHOENIX and POLCA Codes for Nuclear Design of Boiling Water Reactors."

11.

WCAP-16865-P-A, "Westinghouse BWR ECCS Evaluation Model Updates:

Supplement 4 to Code Description, Qualification llN_S_E_R_T_A ___

__.~

and Appl i ca ti on, 11 Revision 1, October 2011.

5.6.6 The COLR will contain the complete identification for each of the TS referenced topical reports used to prepare the COLR Ci.e., report number, title, revision, date, and any supplements).

c.

The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems CECCS) limits, nuclear limits such as SOM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d.

The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

Post Accident Monitoring CPAM> Instrumentation Report When a report is required by Condition B or F of LCD 3.3.3.1, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days.

The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

Dresden 2 and 3 5.6-4 Amendment No. 247/240

DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3 - INSERT A:

12. XN-NF-81-SBCP)(A) Revision 2 and Supplements 1 and 2, "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model," Exxon Nuclear Company, March 1984. (Applicable to Unit 3 Only)
13. ANF-89-98(P)(A) Revision 1 and Supplement 1, "Generic Mechanical Design Criteria for BWR Fuel Designs," Advanced Nuclear Fuels Corporation, May 1995. (Applicable to Unit 3 Only)
14. EMF-85-74(P) Revision O Supplement 1 (P)(A) and Supplement 2 (P)(A), "RODEX2A CBWR) Fuel Rod Thermal-Mechanical Evaluation Model," Siemens Power Corporation, February 1998.

(Applicable to Unit 3 Only)

15. BAW-10247PA Revision 0, "Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors," AREVA NP, February 2008.

(Applicable to Unit 3 Only)

16. XN-NF-80-19(P)(A) Volume 1 and Supplements 1 and 2, "Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis," Exxon Nuclear Company, March 1983.

(Applicable to Unit 3 Only)

17. XN-NF-80-19(P)(A) Volume 4 Revision 1, "Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads," Exxon Nuclear Company, June 1986.

(Applicable to Unit 3 Only)

18. XN-NF-80-19(P)(A) Volume 3 Revision 2, "Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description," Exxon Nuclear Company, January 1987.

(Applicable to Unit 3 Only)

19. EMF-2158(P)(A) Revision 0, "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASM0-4/MICROBURN-82," Siemens Power Corporation, October 1999.

(Applicable to Unit 3 Only)

20. EMF-2245(P)(A) Revision 0, "Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel,"

Siemens Power Corporation, August 2000. (Applicable to Unit 3 Only)

21. EMF-2209(P)(A) Revision 3, 11SPCB Critical Power Correlation, 11 AREVA NP, September 2009. (Applicable to Unit 3 Only)
22. ANP-10298P-A Revision 1, "ACE/ATRIUM lOXM Critical Power Correlation, 11 AREVA, March 2014. (Applicable to Unit 3 Only)
23. ANP-10307PA Revision 0, "AREVA MCPR Safety Limit Methodology for Boiling Water Reactors," AREVA NP, June 2011.

(Applicable to Unit 3 Only)

Page 1of2

DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3 - INSERT A:

24. XN-NF-84-105(P)(A) Volume 1 and Volume 1 Supplements 1 and 2, "XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis," Exxon Nuclear Company, February 1987.

(Applicable to Unit 3 Only)

25. ANF-913CP)(A) Volume 1 Revision 1 and Volume 1 Supplements 2, 3, and 4, "COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses," Advanced Nuclear Fuels Corporation, August 1990. (Applicable to Unit 3 Only)
26. EMF-2361(P)(A) Revision 0, "EXEM BWR-2000 ECCS Evaluation Model,"

Framatome ANP, May 2001. (Applicable to Unit 3 Only)

27. EMF-2292(P)(A) Revision 0, "ATRIUM'-10: Appendix K Spray Heat Transfer Coefficients," Siemens Power Corporation, September 2000.

(Applicable to Unit 3 Only)

28. ANF-1358(P)CA) Revision 3, "The Loss of Feedwater Heating Transient in Boiling Water Reactors," Framatome ANP, September 2005.

(Applicable to Unit 3 Only)

29. EMF-CC-074(P)(A) Volume 4 Revision O~ "BWR Stability Analysis:

Assessment of STAIF with Input from MICROBURN-82," Siemens Power Corporation, August 2000. (Applicable to Unit 3 Only)

Page 2of2

ATTACHMENT 2 Markup of Quad Cities Nuclear Power Station, Units 1 and 2 Technical Specifications Pages Showing Proposed Changes to be Implemented During Interim Period Between Unit 1 Refueling Outage Q1R24, and Unit 2 Refueling Outage Q2R24 3.3.4.1-3 5.6-3 5.6-4 QUAD CITIES INSERT A

ATWS-RPT Instrumentation 3.3.4.1 SURVEILLANCE REQUIREMENTS

NOTE-------------------------------------

When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> provided the associated Function maintains ATWS-RPT trip capability.

SR 3.3.4.1.1 SR 3.3.4.1.2 SR 3.3.4.1.3 SR 3.3.4.1.4 SR 3.3.4.1.5 SURVEILLANCE Perform CHANNEL CHECK.

Calibrate the trip units.

Perform CHANNEL FUNCTIONAL TEST.

Perform CHANNEL CALIBRATION. The Allowable Values shall be:

a.

Reactor Vessel Water Level-Low Low:

FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program

~ -56.3 inches with time delay set to

~ 7.2 seconds and~ 10.8 seconds; and In accordance with the Surveillance Frequency Control Program

b.

Reactor Vessel Steam Dome Pressure-High: < 1219 ~

Perform LOGIC SYSTEM FUNCTIONAL TEST including breaker actuation.

s 1195 psig (Unit 1)

--~ 1219 psig (Unit 2).

In accordance with the Surveillance Frequency Control Program Quad Cities 1 and 2 3.3.4.1-3 Amendment No. 248/243

5.6 Reporting Requirements Reporting Requirements 5.6 5.6.5 CORE OPERATING LIMITS REPORT CCOLR)

(continued)

3.

The LHGR for Specification 3.2.3.

4.

Control Rod Block Instrumentation Setpoint for the Rod Block Monitor-Upscale Function Allowable Value for Specification 3.3.2.1.

5. The OPRM setpoints for the trip function for SR 3.3.1.3.3.
b.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRG, specifically those described in the following documents:

1.

NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel."

2.

Commonwealth Edison Topical Report NFSR-0091, "Benchmark of CASMO/MICROBURN BWR Nuclear Design Methods." ~(Applicable to Unit 2 Only) f

3.

NED0-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," August 1996.

4.

CENPD-300-P-A, "Reference Safety Report for Boiling Water Reactor Reload Fuel."

5.

WCAP-16081-P-A, "lOxlO SVEA Fuel Critical Power Experiments and CPR Correlation:

SVEA-96 Optima2."

6.

WCAP-15682-P-A, "Westinghouse BWR ECCS Evaluation Model:

Supplement 2 to Code Description, Qualification and Application."

7.

WCAP-16078-P-A, "Westinghouse BWR ECCS Evaluation Model:

Supplement 3 to Code Description, Qualification and Application to SVEA-96 Optima2 Fuel."

(continued)

Quad Cities 1 and 2 5.6-3 Amendment No. 246/241

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT CCOLRl (continued)

8.

WCAP-15836-P-A, "Fuel Rod Design Methods for Boiling Water Reactors - Supplement l."

9.

WCAP-15942-P-A, 11 Fuel Assembly Mechanical Design Methodology for Boiling Water Reactors Supplement 1 to CENPD-287."

10.

CENPD-390-P-A, "The Advanced PHOENIX and POLCA Codes for Nuclear Design of Boiling Water Reactors."

11.

WCAP-16865-P-A, "Westinghouse BWR ECCS Evaluation Model Updates:

Supplement 4 to Code Description, Qualification and Application," Revision 1,

~

October 2011.

5.6.6 The COLR will contain the complete identification for each of the TS referenced topical reports used to prepare the COLR (i.e., report number, title, revision, date, and any supplements).

c.

The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems CECCS) limits, nuclear limits such as SOM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d.

The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

Post Accjdent Monjtorjng CPAM) Instrumentation Report When a report is required by Condition B or F of LCO 3.3.3.1, "Post Accident Monitoring CPAM) Instrumentation, 11 a report shall be submitted within the following 14 days.

The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

Quad Cities 1 and 2 5.6-4 Amendment No. 260/255

QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2 - INSERT A:

12. XN-NF-81-58(P)(A) Revision 2 and Supplements 1 and 2, "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model,

11 Exxon Nuclear Company, March 1984. (Applicable to Unit 1 Only)

13. ANF-89-98(P)(A) Revision 1 and Supplement 1, "Generic Mechanical Design Criteria for BWR Fuel Designs," Advanced Nuclear Fuels Corporation, May 1995. (Applicable to Unit 1 Only)
14. EMF-85-74CP) Revision 0 Supplement 1 {P)(A) and Supplement 2 (P)CA), "RODEX2A CBWR) Fuel Rod Thermal-Mechanical Evaluation Model,

11 Siemens Power Corporation, February 1998.

(Applicable to Unit 1 Only)

15. BAW-10247PA Revision 0, "Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors," AREVA NP, February 2008.

(Applicable to Unit 1 Only)

16. XN-NF-80-19{P)(A) Volume 1 and Supplements 1 and 2, "Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis," Exxon Nuclear Company, March 1983.

(Applicable to Unit 1 Only)

17. XN-NF-80-19{P)(A) Volume 4 Revision 1, "Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads, 11 Exxon Nuclear Company, June 1986.

(Applicable to Unit 1 Only)

18. XN-NF-80-19(P)(A) Volume 3 Revision 2, "Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description," Exxon Nuclear Company, January 1987.

(Applicable to Unit 1 Only)

19. EMF-2158(P)(A) Revision 0, "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASM0-4/MICROBURN-B2," Siemens Power Corporation, October 1999.

(Applicable to Unit 1 Only)

20. EMF-2245CP)(A) Revision 0, "Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel,"

Siemens Power Corporation, August 2000. (Applicable to Unit 1 Only)

21. EMF-2209CP){A) Revision 3, "SPCB Critical Power Correlation," AREVA NP, September 2009. (Applicable to Unit 1 Only)
22. ANP-10298P-A Revision 1, "ACE/ATRIUM lOXM Critical Power Correlation," AREVA, March 2014. (Applicable to Unit 1 Only)
23. ANP-10307PA Revision 0, "AREVA MCPR Safety Limit Methodology for Boiling Water Reactors," AREVA NP, June 2011.

(Applicable to Unit 1 Only)

Page 1of2

QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2 - INSERT A:

24. XN-NF-84-lOS(P)(A) Volume 1 and Volume 1 Supplements 1 and 2, "XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis," Exxon Nuclear Company, February 1987.

(Applicable to Unit 1 Only)

25. ANF-913CP>CA) Volume 1 Revision 1 and Volume 1 Supplements 2, 3, and 4, "COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses," Advanced Nuclear Fuels Corporation, August 1990. (Applicable to Unit 1 Only)
26. EMF-2361(P)(A) Revision 0, "EXEM BWR-2000 ECCS Evaluation Model,"

Framatome ANP, May 2001. (Applicable to Unit 1 Only)

27. EMF-2292(P)(A) Revision 0, "ATRIUM'-10: Appendix K Spray Heat Transfer Coefficients," Siemens Power Corporation, September 2000.

(Applicable to Unit 1 Only)

28. ANF-1358(P)(A) Revision 3, "The Loss of Feedwater Heating Transient in Boiling Water Reactors," Framatome ANP, September 2005.

(Applicable to Unit 1 Only)

29. EMF-CC-074(P)(A) Volume 4 Revision 0, "BWR Stability Analysis:

Assessment of STAIF with Input from MICROBURN-82," Siemens Power Corporation, August 2000. (Applicable to Unit 1 Only)

Page 2of2

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