ML14113A286
| ML14113A286 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 07/31/2014 |
| From: | Farideh Saba Plant Licensing Branch II |
| To: | James Shea Tennessee Valley Authority |
| Saba F DORL/LPL2-2 301-415-1447 | |
| References | |
| TAC MF0877, TAC MF0878, TAC MF0879 | |
| Download: ML14113A286 (70) | |
Text
OFFICIAL USE ONLY PROPRIETARY INFORMATION UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. Joseph W. Shea Vice President, Nuclear Licensing Tennessee Valley Authority 1101 Market Street, LP 3D-C Chattanooga, TN 37 402-2801 July 31, 2014
SUBJECT:
BROWNS FERRY NUCLEAR PLANT, UNITS 1, 2, AND 3-ISSUANCE OF AMENDMENTS REGARDING TECHNICAL SPECIFICATION (TS) CHANGE TS-478 ADDITION OF ANALYTICAL METHODOLOGIES TOTS 5.6.5 AND REVISION OF TS 2.1.1.2 FOR UNIT 2 (TAC NOS. MF0877, MF0878 AND MF0879)
Dear Mr. Shea:
The Commission has issued the enclosed Amendment Nos. 285, 311, and 270 to Renewed Facility Operating Licenses Nos. DPR-33, DPR-52, and DPR-68 for the Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3, respectively. These amendments are in response to the Tennessee Valley Authority's (TVA's) application dated February 28, 2013, as supplemented by letters dated September 30, 2013, and May 16, 2014.
The amendments add three additional AREVA analysis methodologies to the list of approved methods to be used in determining core operating limits in the Core Operating Limits Report. In addition, the amendments implement a change to the Safety Limit Minimum Critical Power Ratio value for BFN Unit 2. The changes support a planned transition to AREVA ATRIUM 10XM (XM) fuel design. TV A intends to transition BFN Unit 2 to the XM fuel design starting with Cycle 19 in the spring of 2015, Unit 3 in the spring of 2016, and Unit 1 in the fall of 2016.
The Nuclear Regulatory Commission (NRC, Commission) staff has completed its review of the information provided by the licensee. The NRC staff's safety evaluation (SE) is enclosed. The NRC staff has determined that its documented SE (Enclosure 4) contains proprietary information pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Section 2.390, "Public inspections, exemptions, requests for withholding." Accordingly, the NRC staff has prepared a redacted, nonproprietary version (Enclosure 5). However, the NRC will delay placing the nonproprietary SE in the public document room for a period of 1 0 working days from the date of this letter to provide TVA with the opportunity to comment on any proprietary aspects. If you believe that any information in Enclosure 5 is proprietary, please identify such information line-by-line and define the basis pursuant to the criteria of 10 CFR 2.390. After 10 working days, the nonproprietary SE will be made publicly available.
Document transmitted herewith contains Sensitive Unclassified Non-Safeguard Information in its Enclosure 4. When Separated from Enclosure 4. this document is decontrolled.
OFFICIAL USE ONLY PROPRIETARY INFORMATION
OFFICIAL USE ONLY PROPRIETARY INFORMATION To facilitate the NRC staff's review of your comments, please provide a marked-up copy of theSE showing proposed changes and provide a summary table of the proposed changes.
A Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
Docket Nos. 50-259, 50-260, and 50-296
Enclosures:
Sincerely, Farideh E. Saba, Senior Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
- 1. Amendment No. 285 to Renewed License No. DPR-33
- 2. Amendment No. 311 to Renewed License No. DPR-52
- 3. Amendment No. 270 to Renewed License No. DPR-68
- 4. Safety Evaluation* (Proprietary Information)
- 5. Safety Evaluation (Non-Proprietary Information) cc with Enclosures 1, 2, 3, 4, and 5: Addressee only cc with Enclosures 1, 2, 3, and 5: Distribution via Listserv (10 days after issuance of the amendments to the licensee.)
OFFICIAL USE ONLV PROPRIETARY INFORMATION
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-259 BROWNS FERRY NUCLEAR PLANT UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 285 Renewed License No. DPR-33
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated February 28, 2013, as supplemented by letters dated September 30, 2013, and May 16, 2014, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
- 2.
Accordingly, the license is amended by changes to the Operating License and Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-33 is hereby amended as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 285, are hereby incorporated in the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3.
The renewed license is also amended by adding License Condition (17) to Section 2.C of the Renewed Facility Operating License No. DPR-33 and reformatting page 7 of the license as indicated in the attachment to this amendment.
(17)
The fuel channel bow standard deviation component of the channel bow model uncertainty used by ANP-10307PA, "AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, Revision 0" (i.e., TS 5.6.5.b.11) to determine the Safety Limit Minimum Critical Power Ratio shall be increased by the ratio of channel fluence gradient to the nearest channel fluence gradient bound of the channel measurement database, when applied to channels with fluence gradients outside the bounds of the measurement database from which the model uncertainty is determined. This license condition will be effective upon the implementation of Amendment No. 285.
- 4.
This license amendment is effective as of its date of issuance and shall be implemented during the Unit 1 refueling outage in fall of 2016.
Attachment:
FOR T/E NUCLEAR REGULATORY COMMISSION
- lifo--
Lisa M. Regner, Acting Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Changes to the Renewed Operating License and Technical Specifications Date of Issuance: July 31, 2014
ATTACHMENT TO LICENSE AMENDMENT NO. 285 RENEWED FACILITY OPERATING LICENSE NO. DPR-33 DOCKET NO. 50-259 Replace the following pages of the Renewed Operating License DPR-33 with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
REMOVE 3
6 7
INSERT 3
6 7*
Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
REMOVE 5.0-24b 5.0-24c INSERT 5.0-24b 5.0-24c
- No changes to this page, only text from page 6 is moved to page 7.
3-(3)
Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use at any time any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4)
Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form for sample analysis or equipment and instrument calibration or associated with radioactive apparatus or components; (5)
Pursuant to the Act and 10 CFR Parts 30 and 70, to possess but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C.
This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter 1:
BFN-UNIT 1 Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1)
Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 3458 megawatts thermal.
(2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 285, are hereby incorporated in the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.
For Surveillance Requirements (SRs) that are new in Amendment 234 to Facility Operating License DPR-33, the first performance is due at the end of the first surveillance interval that begins at implementation of the Amendment 234. For SRs that existed prior to Amendment 234, including SRs with modified acceptance criteria and SRs whose frequency of performance is being extended, the first performance is due at the end of the first surveillance interval that begins on the date the surveillance was last performed prior to implementation of Amendment 234.
Renewed License No. DPR-33 Amendment No. 285 Following Implementation:
(a) The first performance of SR 3.7.4.4, in accordance with TS 5.5.13.c.(i), shall be within a specific frequency of 6 years, plus the 18-month allowance of SR 3.0.2, as measured from November 10, 2003, the date of the most recent successful tracer gas test.
(b) The first performance of the periodic assessment of the Control Room Envelope (CRE) Habitability, Technical Specification 5.5.13.c.(ii), shall be within 9 months following the initial implementation of the TS Change. The next performance of the periodic assessment will be in a period specified by the CRE Program. That is 3 years from the last successful performance of the Technical Specification 5.5.13.c.(ii) tracer gas test.
(c) The first performance of the periodic measurement of CRE pressure, TS 5.5.13.d, shall be within 24 months, plus 180 days allowed by SR 3.0.2 as measured from the date of the most recent successful pressure measurement test.
(17) The fuel channel bow standard deviation component of the channel bow model uncertainty used by ANP-10307PA, "AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, Revision 0," (i.e., TS 5.6.5.b.11) to determine the Safety Limit Minimum Critical Power Ratio shall be increased by the ratio of channel fluence gradient to the nearest channel fluence gradient bound of the channel measurement database, when applied to channels with fluence gradients outside the bounds of the measurement database from which the model uncertainty is determined. This license condition will be effective upon the implementation of Amendment No. 285.
D.
The UFSAR supplement, as revised, submitted pursuant to 10 CFR 54.21 (d), shall be included in the next scheduled update to the UFSAR required by 10 CFR 50.71(e)(4) following the issuance of this renewed operating license. Until that update is complete, TVA may make changes to the programs and activities described in the supplement without prior Commission approval, provided that TVA evaluates such changes pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.
E.
The UFSAR supplement, as revised, describes certain future activities to be completed prior to the period of extended operation. TVA shall complete these activities no later than December 20, 2013, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.
F.
All capsules in the reactor vessel that are removed and tested must meet the test procedures and reporting requirements of the most recent NRC-approved version of the Boiling Water Reactor Vessels and Internals Project (BWRVIP) Integrated Surveillance Program (ISP) appropriate for the configuration of the specimens in the capsule. Any changes to the BWRVIP ISP capsule withdrawal schedule, including spare capsules, must be approved by the NRC prior to implementation. All capsules placed in storage BFN-UNIT 1 Renewed License No. DPR-33 Amendment No. 285 must be maintained for future insertion. Any changes to storage requirements must be approved by the NRC, as required by 10 CFR Part 50, Appendix H.
G.
(1)
During the power up rate power ascension test program and prior to exceeding 30 days of plant operation above a nominal 3293 megawatts thermal power level
( 1 00-percent OL TP) or within 30 days of satisfactory completion of steam dryer monitoring and testing that is necessary for achieving 1 05-percent OL TP (whichever is longer), with plant conditions stabilized at 1 05-percent OL TP, TVA shall trip a condensate booster pump, a condensate pump, and a main feedwater pump on an individual basis (i.e., one at a time). Following each pump trip, TVA shall confirm that plant response to the transient is as expected in accordance with previously established acceptance criteria. Evaluation of the test results for each test shall be completed and all discrepancies resolved in accordance with corrective action program requirements and the provisions of the power ascension test program.
(2)
Deleted.
H.
The licensee must complete the thirteen (13) Unit 1 restart commitments that are discussed in Appendix F of the license renewal application, dated December 31, 2003, as supplemented by letters dated January 31, 2005, March 2, and April 21, 2006.
Completion of these activities must be met prior to power operation of Unit 1.
I.
This renewed license is effective as of the date of issuance and shall expire midnight on December 20, 2033.
Attachments:
FOR THE NUCLEAR REGULATORY COMMISSION Original Signed By J. E. Dyer J. E. Dyer, Director Office of Nuclear Reactor Regulation
- 1. Unit 1 -Technical Specifications-Appendices A and B Date of Issuance: May 4, 2006 BFN-UNIT 1 Renewed License No. DPR-33 Amendment No. 285
Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT CCOLR) (continued)
BFN-UNIT 1
- 10. XN-NF-84-1 05(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.
- 11. ANP-10307PA Revision 0, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, AREVA NP, June 2011.
- 12. ANF-913(P)(A) Volume 1 Revision 1 and Volume 1 Supplements 2, 3 and 4, COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses, Advance~ Nuclear Fuels Corporation, August 1990.
- 13. ANF-1358(P)(A) Revision 3, The Loss of Feedwater Heating Transient in Boiling Water Reactors, Framatome ANP, September 2005.
- 14. EMF-2209(P)(A) Revision 3, SPCB Critical Power Correlation, AREVA NP, September 2009.
- 15. EMF-2245(P)(A) Revision 0, Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel, Siemens Power Corporation, August 2000.
- 16. EMF-2361 (P)(A) Revision 0, EXEM BWR-2000 ECCS Evaluation Model, Framatome ANP Inc., May 2001 as supplemented by the site-specific approval in NRC safety evaluation dated April27, 2012.
- 17. EMF-2292(P)(A) Revision 0, ATRIUM'-10: Appendix K Spray Heat Transfer Coefficients, Siemens Power Corporation, September 2000.
5.0-24b (continued)
Amendment No.~.~.~.
23-1-, 285
Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (COLRl (continued)
BFN-UNIT 1
- 18. EMF-CC-07 4(P)(A), Volume 4, Revision 0, BWR Stability Analysis:
Assessment of STAIF with Input from MICROBURN-B2, Siemens Power Corporation, August 2000.
- 19. BAW-1 0255(P)(A), Revision 2, Cycle-Specific DIVOM Methodology Using the RAMONA5-FA Code, AREVA NP, May 2008.
20.BAW-10247PA Revision 0, Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors, AREVA NP, February 2008.
21.ANP-10298PA Revision 0, ACE/ATRIUM 10XM Critical Power Correlation, AREVA NP, March 2010.
22.ANP-3140(P) Revision 0, Browns Ferry Units 1, 2, and 3 Improved K-factor Model for ACE/ATRIUM 10XM Critical Power Correlation, AREVA NP, August 2012.
5.0-24c (continued)
Amendment No. ~. 239, ~.
~.285
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-260 BROWNS FERRY NUCLEAR PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 311 Renewed License No. DPR-52
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated February 28, 2013, as supplemented by letters dated September 30, 2013, and May 16, 2014, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
- 2.
Accordingly, the license is amended by changes to the Operating License and Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-52 is hereby amended as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 311, are hereby incorporated in the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3.
The renewed license is also amended by adding License Condition (17) to Section 2.C of the Renewed Facility Operating License No. DPR-52 and reformatting page 7 of the license as indicated in the attachment to this amendment.
(17)
The fuel channel bow standard deviation component of the channel bow model uncertainty used by ANP-1 0307PA, "AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, Revision 0" (i.e., TS 5.6.5.b.1 0) to determine the Safety Limit Minimum Critical Power Ratio shall be increased by the ratio of channel fluence gradient to the nearest channel fluence gradient bound of the channel measurement database, when applied to channels with fluence gradients outside the bounds of the measurement database from which the model uncertainty is determined. This license condition will be effective upon the implementation of Amendment No. 311.
- 4.
This license amendment is effective as of its date of issuance and shall be implemented during the Unit 2 refueling outage in spring of 2015.
Attachment:
F(O:;jUCLEAR REGULA TORY COMMISSION
- z£W-~-__/
Lisa M. Regner, Acting Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Changes to the Renewed Operating License and Technical Specifications Date of Issuance: July 31, 2014
ATTACHMENT TO LICENSE AMENDMENT NO. 311 RENEWED FACILITY OPERATING LICENSE NO. DPR-52 DOCKET NO. 50-260 Replace the following pages of the Renewed Operating License DPR-52 with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
REMOVE 3
6 7
INSERT 3
6 7*
Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
REMOVE 2.0-1 5.0-24a 5.0-24b INSERT 2.0-1 5.0-24a 5.0-24b
- No changes to this page, only text from page 6 moved to page 7.
3-sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4)
Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form for sample analysis or equipment and instrument calibration or associated with radioactive apparatus or components; (5)
Pursuant to the Act and 10 CFR Parts 30 and 70, to possess but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C.
This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter 1:
Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1)
Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 3458 megawatts thermal.
(2)
Technical Specifications
- 3)
BFN-UNIT 2 The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 311, are hereby incorporated in the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.
For Surveillance Requirements (SRs) that are new in Amendment 253 to Facility Operating License DPR-52, the first performance is due at the end of the first surveillance interval that begins at implementation of Amendment 253. For SRs that existed prior to Amendment 253, including SRs with modified acceptance criteria and SRs whose frequency of performance is being extended, the first performance is due at the end of the first surveillance interval that begins on the date the surveillance was last performed prior to implementation of Amendment 253.
The licensee is authorized to relocate certain requirements included in Appendix A and the former Appendix B to licensee-controlled documents.
Implementation of this amendment shall include the relocation of these requirements to the appropriate documents, as described in the licensee's Amendment No. 311 Renewed License No. DPR-52 Following implementation:
(a) The first performance of SR 3.7.4.4, in accordance with TS 5.5.13.c.(i),
shall be within a specific frequency of 6 years, plus the 18-month allowance of SR 3.0.2, as measured from November 10, 2003, the date of the most recent successful tracer gas test.
(b) The first performance of the periodic assessment of the Control Room Envelope (CRE) Habitability, Technical Specification 5.5.13.c.(ii), shall be within 9 months following the initial implementation of the TS change. The next performance of the periodic assessment will be in a period specified by the CRE Program. That is 3 years from the last successful performance of the Technical Specification 5.5.13.c.(ii) tracer gas test.
(c) The first performance of the periodic measurement of CRE pressure, TS 5.5.13.d, shall be with 24 months, plus 180 days allowed by SR 3.0.2 as measured from the date of the most recent successful pressure measurement test.
( 17)
The fuel channel bow standard deviation component of the channel bow model uncertainty used by ANP-1 0307PA, "AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, Revision 0," (i.e., TS 5.6.5.b.1 0) to determine the Safety Limit Minimum Critical Power Ratio shall be increased by the ratio of channel fluence gradient to the nearest channel fluence gradient bound of the channel measurement database, when applied to channels with fluence gradients outside the bounds of the measurement database from which the model uncertainty is determined. This license condition will be effective upon the implementation of Amendment No. 311.
D.
The UFSAR supplement, as revised, submitted pursuant to 10 CFR 54.21 (d),
shall be included in the next scheduled update to the UFSAR required by 10 CFR 50.71 (e)(4) following the issuance of this renewed operating license.
Until that update is complete, TVA may make changes to the programs and activities described in the supplement without prior Commission approval, provided that TVA evaluates such changes pursuant to the criteria set forth in 1 0 CFR 50.59 and otherwise complies with the requirements in that section.
E.
The UFSAR supplement, as revised, describes certain future activities to be completed prior to the period of extended operation. TVA shall complete these activities no later than June 28, 2014, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.
F.
All capsules in the reactor vessel that are removed and tested must meet the test procedures and reporting requirements of the most recent NRC-approved version of the Boiling Water Reactor Vessels and Internals Project (BWRVIP) Integrated Surveillance Program (ISP) appropriate for the configuration of the specimens in the BFN-UNIT 2 Renewed License No. DPR-52 Amendment No. 311 capsule. Any changes to the BWRVIP ISP capsule withdrawal schedule, including spare capsules, must be approved by the NRC prior to implementation. All capsules placed in storage must be maintained for future insertion. Any changes to storage requirements must be approved by the NRC, as required by 10 CFR Part 50, Appendix H.
G.
This renewed license is effective as of the date of issuance and shall expire midnight on June 28, 2034.
Attachments:
FOR THE NUCLEAR REGULATORY COMMISSION Original Signed By J. E. Dyer J. E. Dyer, Director Office of Nuclear Reactor Regulation
- 1. Unit 2-Technical Specifications-Appendices A and B Date of Issuance: May 4, 2006 BFN-UNIT 2 Amendment No. 311 Renewed License No. DPR-52
SLs 2.0 2.0 SAFETY LIMITS (Sls) 2.1 Sls 2.1.1 Reactor Core Sls 2.1.1.1 With the reactor steam dome pressure< 785 psig or core flow
< 1 0% rated core flow:
THERMAL POWER shall be:::; 25% RTP.
2.1.1.2 With the reactor steam dome pressure ~ 785 psig and core flow
~ 1 0% rated core flow:
MCPR shall be ~ 1.06 for two recirculation loop operation or~ 1.08 for single loop operation.
2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.
2.1.2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall be :::; 1325 psig.
2.2 SL Violations With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:
2.2.1 Restore compliance with all Sls; and 2.2.2 Insert all insertable control rods.
BFN-UNIT 2 2.0-1 Amendment No.~.~. ~.
~. 311
5.6 Reporting Requirements (continued)
Reporting Requirements 5.6
- 4. ANF-89-98(P)(A) Revision 1 and Supplement 1, Generic Mechanical Design Criteria for BWR Fuel Designs, Advanced Nuclear Fuels Corporation, May 1995.
- 5. 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.
- 6. 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.
- 7. EMF-2158(P)(A) Revision 0, Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASM0-4/MICROBURNB2, Siemens Power Corporation, October 1999.
- 8. 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.
- 9. 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.
10.ANP-10307PA Revision 0, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, A REV A NP, June 2011.
11.ANF-913(P)(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.
12.ANF-1358(P)(A) Revision 3, The Loss of Feedwater Heating Transient in Boiling Water Reactors, Framatome ANP, September 2005.
- 13. EMF-2209(P)(A) Revision 3, SPCB Critical Power Correlation, AREVA NP, September 2009.
(continued)
BFN-UNIT2 5.0-24a Amendment No. 287, ~. 311
5.6 Reporting Requirements (continued)
Reporting Requirements 5.6
- 14. EMF-2245(P)(A) Revision 0, Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel, Siemens Power Corporation, August 2000.
- 15. EMF-2361 (P)(A) Revision 0, EXEM BWR-2000 ECCS Evaluation Model, Framatome ANP Inc., May 2001 as supplemented by the site-specific approval in NRC safety evaluation dated February 15, 2013.
16.EMF-2292(P)(A) Revision 0, ATRIUM'-10: Appendix K Spray Heat Transfer Coefficients, Siemens Power Corporation, September 2000.
- 17. EMF-CC-07 4(P)(A), Volume 4, Revision 0, BWR Stability Analysis: Assessment of STAIF with Input from MICROBURN-82, Siemens Power Corporation, August 2000.
- 18. BAW-1 0255(P)(A), Revision 2, Cycle-Specific DIVOM Methodology Using the RAMONA5-FA Code, AREVA NP, May 2008.
19.BAW-10247PA Revision 0, Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors, AREVA NP, February 2008.
20.ANP-10298PA Revision 0, ACE/ATRIUM 10XM Critical Power Correlation, AREVA NP, March 2010.
21.ANP-3140(P) Revision 0, Browns Ferry Units 1, 2, and 31mproved K-factor Model for ACE/ATRIUM 10XM Critical Power Correlation, AREVA NP, August 2012.
(continued)
BFN-UNIT 2 5.0-24b Amendment No. ~.
dQQ, 311
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-296 BROWNS FERRY NUCLEAR PLANT, UNIT 3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 270 Renewed License No. DPR-68
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated February 28, 2013, as supplemented by letters dated September 30, 2013, and May 16, 2014, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
- 2.
Accordingly, the license is amended by changes to the Operating license and Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-68 is hereby amended as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 270, are hereby incorporated in the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3.
The renewed license is also amended by adding License Condition (13) to Section 2.C of the Renewed Facility Operating License No. DPR-68 and reformatting page 7 of the license as indicated in the attachment to this amendment.
(13)
The fuel channel bow standard deviation component of the channel bow model uncertainty used by ANP-10307PA, "AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, Revision 0" (i.e., TS 5.6.5.b.1 0) to determine the Safety Limit Minimum Critical Power Ratio shall be increased by the ratio of channel fluence gradient to the nearest channel fluence gradient bound of the channel measurement database, when applied to channels with fluence gradients outside the bounds of the measurement database from which the model uncertainty is determined. This license condition will be effective upon the implementation of Amendment No. 270.
- 4.
This license amendment is effective as of its date of issuance and shall be implemented during the Unit 3 refueling outage in spring of 2016.
Attachment:
NUCLEAR REGULATORY COMMISSION Lisa M. Regner, Acting Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Changes to the Renewed Operating License and Technical Specifications Date of Issuance: July 31, 2014
ATTACHMENT TO LICENSE AMENDMENT NO. 270 RENEWED FACILITY OPERATING LICENSE NO. DPR-68 DOCKET NO. 50-296 Replace the following pages of the Renewed Operating License DPR-68 with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
REMOVE 3
6 7
INSERT 3
6 7*
Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
REMOVE 5.0-24a 5.0-24b INSERT 5.0-24a 5.0-24b 5.0-24c
- No changes to this page, only text from page 6 moved to page 7.
3-(3)
Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use at any time any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4)
Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form for sample analysis or equipment and instrument calibration or associated with radioactive apparatus or components; (5)
Pursuant to the Act and 10 CFR Parts 30 and 70, to possess but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C.
This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter 1:
BFN-UNIT 3 Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1)
Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 3458 megawatts thermal.
(2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 270, are hereby incorporated in the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.
For Surveillance Requirements (SRs) that are new in Amendment 212 to Facility Operating License DPR-68, the first performance is due at the end of the first surveillance interval that begins at implementation of the Amendment 212. For SRs that existed prior to Amendment 212, including SRs with modified acceptance criteria and SRs whose frequency of performance is being extended, the first performance is due at the end of the first surveillance interval that begins on the date the surveillance was last performed prior to implementation of Amendment 212.
Amendment No. 270 Renewed License No. DPR-68 (3) Following Implementation:
(a)
The first performance of SR 3.7.4.4, in accordance with TS 5.5.13.c.(i), shall be within a specific frequency of 6 years, plus the 18-month allowance of SR 3.0.2, as measured from November 10, 2003, the date of the most recent successful tracer gas test.
(b)
The first performance of the periodic assessment of the Control Room Envelope (CRE) Habitability, Technical Specification 5.5.13.c.(ii), shall be within 9 months following the initial implementation of the TS Change. The next performance of the periodic assessment will be in a period specified by the CRE Program. That is 3 years from the last successful performance of the Technical Specification 5.5.13.c.(ii) tracer gas test.
(c)
The first performance of the periodic measurement of CRE pressure, TS 5.5.13.d, shall be within 24 months, plus 180 days allowed by SR 3.0.2 as measured from the date of the most recent successful pressure measurement test.
(d)
For License Amendment 268,the licensee shall implement changes to BFN, Unit 3 TSs 5.6.5 and 3.3.1.1 within 60 days of approval. The remaining BFN, Unit 3, changes will be implemented upon completion of required supporting modification work and prior to entering Mode 3 (i.e., Hot Shutdown) from the spring 2014 refueling outage.
(13)
The fuel channel bow standard deviation component of the channel bow model uncertainty used by ANP-10307PA, "AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, Revision 0," (i.e., TS 5.6.5.b.1 0) to determine the Safety Limit Minimum Critical Power Ratio shall be increased by the ratio of channel fluence gradient to the nearest channel fluence gradient bound of the channel measurement database, when applied to channels with fluence gradients outside the bounds of the measurement database from which the model uncertainty is determined. This license condition will be effective upon the implementation of Amendment No. 270.
D.
The UFSAR supplement, as revised, submitted pursuant to 10 CFR 54.21 (d),
shall be included in the next scheduled update to the UFSAR required by 10 CFR 50.71 (e)(4) following the issuance of this renewed operating license.
Until that update is complete, TVA may make changes to the programs and activities described in the supplement without prior Commission approval, provided that TVA evaluates such changes pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.
E.
The UFSAR supplement, as revised, describes certain future activities to be completed prior to the period of extended operation. TVA shall complete these activities no later than July 2, 2016, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.
BFN-UNIT 3 Renewed License No. DPR-68 Amendment No. 270 F.
All capsules in the reactor vessel that are removed and tested must meet the test procedures and reporting requirements of the most recent NRC-approved version of the Boiling Water Reactor Vessels and Internals Project (BWRVIP) Integrated Surveillance Program (ISP) appropriate for the configuration of the specimens in the capsule. Any changes to the BWRVIP ISP capsule withdrawal schedule, including spare capsules, must be approved by the NRC prior to implementation. All capsules placed in storage must be maintained for future insertion. Any changes to storage requirements must be approved by the NRC, as required by 10 CFR Part 50, Appendix H.
G.
This renewed license is effective as of the date of issuance and shall expire midnight on July 2, 2036.
Attachments:
FOR THE NUCLEAR REGULATORY COMMISSION Original Signed By J. E. Dyer J. E. Dyer, Director Office of Nuclear Reactor Regulation
- 1. Unit 3 - Technica! Specifications - Appendices A and B Date of Issuance: May 4, 2006 BFN-UNIT 3 Renewed License No. DPR-68 Amendment No. 270
Reporting Requirements 5.6 5.6 Reporting Requirements (continued)
BFN-UNIT 3
- 4. ANF-89-98(P)(A) Revision 1 and Supplement 1, Generic Mechanical Design Criteria for BWR Fuel Designs, Advanced Nuclear Fuels Corporation, May 1995.
- 5. 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.
- 6. 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.
- 7. EMF-2158(P)(A) Revision 0, Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASM0-4/MICROBURNB2, Siemens Power Corporation, October 1999.
- 8. 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.
- 9. XN-NF-84-1 05(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.
10.ANP-10307PA Revision 0, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, AREVA NP, June 2011.
- 11. ANF-913(P)(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.
5.0-24a (continued)
Amendment No.~. ~. ~.
~.~.270
Reporting Requirements 5.6 5.6 Reporting Requirements (continued)
BFN-UNIT 3 12.ANF-1358(P)(A) Revision 3, The Loss of Feedwater Heating Transient in Boiling Water Reactors, Framatome ANP, September 2005.
- 13. EMF-2209(P)(A) Revision 3, SPCB Critical Power Correlation, AREVA NP, September 2009.
- 14. EMF-2245(P)(A) Revision 0, Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel, Siemens Power Corporation, August 2000.
- 15. EMF-2361(P)(A) Revision 0, EXEM BWR-2000 ECCS Evaluation Model, Framatome ANP Inc., May 2001 as supplemented by the site-specific approval in NRC safety evaluation dated February 15, 2013 16.EMF-2292(P)(A) Revision 0, ATRIUM'-10: Appendix K Spray Heat Transfer Coefficients, Siemens Power Corporation, September 2000.
- 17. EMF-CC-07 4(P)(A), Volume 4, Revision 0, BWR Stability Analysis:
Assessment of STAIF with Input from MICROBURN-82, Siemens Power Corporation, August 2000.
18.BAW-10255(P)(A), Revision 2, Cycle-Specific DIVOM Methodology Using the RAMONA5-FA Code, AREVA NP, May 2008.
- 19. BAW-10247PA Revision 0, Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors, AREVA NP, February 2008.
20.ANP-10298PA Revision 0, ACE/ATRIUM 10XM Critical Power Correlation, AREVA NP, March 2010.
5.0-24b (continued)
Amendment No. ~. ~.
~.~.~.270
Reporting Requirements 5.6 5.6 Reporting Requirements (continued)
BFN-UNIT3 21.ANP-3140(P) Revision 0, Browns Ferry Units 1, 2, and 3 Improved K-factor Model for ACE/ATRIUM 10XM Critical Power Correlation, AREVA NP, August 2012.
5.0-24c (continued)
Amendment No. ~. ~.
~. 2e9, 268,270
OFFICIAl USE ONlY PROPRIETARY INFORMATION UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 285 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-33, AMENDMENT NO. 311 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-52, AND AMENDMENT NO. 270 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-68 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT, UNITS 1, 2, AND 3 DOCKET NOS. 50-259, 50-260, AND 50-296
1.0 INTRODUCTION
By letter dated February 28, 2013 (Reference 1 ), as supplemented by letters dated September 30, 2013 (Reference 2), and May 16, 2014 (Reference 37), Tennessee Valley Authority (TVA, the licensee) requested an amendment to the Technical Specifications (TSs) for Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3. The license amendment request (LAR) adds three additional AREVA analysis methodologies to the list of approved methods to be used in determining core operating limits in the core operating limits report (COLR). In addition, the amendment request implements a change to the safety limit minimum critical power ratio (SLMCPR) value for BFN Unit 2. The changes support a planned transition to AREVA ATRIUM 1 OXM (XM) fuel design. TVA intends to transition BFN Unit 2 to the XM fuel design starting with Cycle 19 in the spring of 2015, Unit 3 in the spring of 2016, and Unit 1 in the fall of 2016.
The Nuclear Regulatory Commission (NRC) staff has reviewed and evaluated the following reports and analyses on fuel mechanical design, thermal-hydraulic design, fuel cycle design, reload safety analysis, thermal conductivity degradation, fuel rod thermal mechanical evaluation, equilibrium fuel cycle design, loss-of-coolant accident (LOCA) analysis, and reload safety analyses, which were submitted as attachments to the LAR.
ANP-3150P Revision 0, "Mechanical Design Report for Browns Ferry ATRIUM 10XM Fuel Assemblies," AREVA NP, Inc., October 2012.
ANP-3082P, Revision 1, "Browns Ferry Thermal-Hydraulic Design Report for ATRIUM 10XM Fuel Assemblies," AREVA NP Inc., August 2012.
ANP-3145P, Revision 0, "Browns Ferry Unit 2 Cycle 19 LAR Fuel Cycle Design,"
AREVA NP Inc., August 2012.
OFFICIAl USE ONlY PROPRIETARY INFORMATION
OffiCIAl USE ONlY PROPRIETARY INFORMATION
- ANP-3167P, Revision 0, "Browns Ferry Unit 2 Cycle 19 Reload Analysis,"
AREVA NP Inc., November 2012.
ANP-3152P, Revision 0, "Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel," AREVA NP Inc., October 2012.
ANP-3153P, Revision 0, "Browns Ferry Units 1, 2, and 3 LOCA-ECCS [Emergency Core Cooling System] Analysis MAPLHGR [Maximum Average Planar Linear Heat Generation Rate] Limits for ATRIUM 1 OXM Fuel, AREVA NP Inc., October 2012.
ANP-3170P, Revision 0, "Evaluation of Fuel Conductivity Degradation for ATRIUM 10XM Fuel for Browns Ferry Units 1, 2, and 3," AREVA NP Inc., November 2012.
ANP-3159P, Revision 0, "ATRIUM 10XM Fuel Rod Thermal-Mechanical Evaluation for Browns Ferry Unit 2 Cycle 19 Reload BFN2-19," AREVA NP Inc., October 2012.
ANP-3148P, Revision 0, "Browns Ferry ATRIUM 10XM Equilibrium Cycle Design Summary," AREVA NP Inc., August 2012.
51-9191258-001, Revision 1, "Browns Ferry Unit 2 Cycle 19 MCPR [Minimum Critical Power Ratio] Safety Limit Analysis with SAFLIM3D Methodology," October 2012.
ANP-3140P, Revision 0, "Browns Ferry Units 1, 2 and 3, Improved K-factor Model for ACE/ATRIUM 10XM Critical Power Correlation," AREVA NP Inc., August 2012.
The supplements dated September 30, 2013, and May 16, 2014, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register on August 13,2013 (78 FR 49301).
2.0 REGULATORY EVALUATION
TVA intends to transition BFN Unit 2 to the XM fuel design starting with Cycle 19 in the spring of 2015, Unit 3 in the spring of 2016 and Unit 1 in fall of 2016. The proposed LAR will add three additional AREVA analysis methodologies to the list of approved methods to be used in determining core operating limits in the COLR. The additional methodologies are:
RODEX4 for fuel rod thermal mechanical analyses (Reference 3)
ACE correlation for critical power monitoring of XM fuel (References 4 and 5)
SAFLIM-3D for SLMCPR analyses (Reference 22)
The proposed addition of the above methodologies to all three BFN units is to support a planned fuel transition to ATRIUM 1 OXM fuel design for non-extended power uprate power conditions (i.e.,
1 05-percent original licensed power level only).
In addition, a change to the Unit 2 SLMCPR TS values is requested to reflect a conservative reduction from the current values by using one of the new methodologies (Reference 22).
OffiCIAl USE ONlY PROPRIETARY INFORMATION
OFFICIAl USE ONlY PROPRIETARY INFORMATION The XM fuel design was developed using the thermal mechanical design bases and limits outlined in Reference 7, the compliance with which ensures the fuel design meets the requirements for fuel system damage, fuel failure, and fuel coolability criteria identified in Standard Review Plan (SAP, NUREG-0800) (Reference 8).
The SAP Section 4.2, "Fuel System design," Section 4.3, "Nuclear Design," and Section 4.4, "Thermal and Hydraulic Design," provide guidance for the review of fuel rod cladding materials, the fuel system, the design of the fuel assemblies and control systems, and thermal and hydraulic design of the core. In addition, the SAP provides guidance for compliance with the applicable General Design Criteria (GDC) specified in Appendix A to Title 10 of the Code of Federal Regulations (1 0 CFR) Part 50. Specifically, as discussed in SAP Section 4.2, the fuel system safety review provides assurance that:
The fuel system is not damaged as a result of normal operation and anticipated operational occurrences (AOOs),
Fuel system damage is never so severe as to prevent control rod insertion when it is
- required, The number of fuel rod failures is not underestimated for postulated accidents, and Coolability is always maintained.
The NRC staff, upon review of the LAR to evaluate the applicability of AREVA methodology to the BFN units' TSs and changes in the SLMCPR, confirms that the use of the methodology is within the NRC-approved ranges of its applicability, and verifies that the results of the analyses are in compliance with the applicable requirements of the following GDC:
GDC-1 0, "Reactor design," requires the reactor design (reactor core and associated reactor coolant system (RCS), control, and protection systems) assures that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of AOOs.
GDC-12, "Suppression of reactor power oscillations," requires the reactor core and associated coolant, control, and protection systems to be designed to assure that power oscillations which can result in conditions exceeding specified acceptable fuel design limits are not possible, or can be reliably and readily detected and suppressed.
GDC-15, "Reactor coolant system design," requires the RCS and associated auxiliary, control, and protection systems to be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including AOOs.
GDC-20, "Protection system functions," requires the protection system to be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of AOOs and (2) to sense accident conditions and to initiate the operation of systems and components important to safety.
OFFICIAl USE ONlY PROPRIETARY INFORMATION
OffiCIAl USE ONlY PROPRIETARY INfORMATION
- GDC-25, "Protection system requirements for reactivity control malfunctions," requires the protection system to be designed to assure that specified acceptable fuel design limits are not exceeded for any single malfunction of the reactivity control systems, such as accidental withdrawal (not ejection or dropout) of control rods.
GDC-26, "Reactivity control system redundancy and capability," requires two independent reactivity control systems of different design principles be provided. One of the systems shall use control rods, preferably including a positive means for inserting the rods, and shall be capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including anticipated operational occurrences, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded. The second reactivity control system shall be capable of reliably controlling the rate of reactivity changes resulting from planned, normal power changes (including xenon burnout) to assure acceptable fuel design limits are not exceeded. One of the systems shall be capable of holding the reactor core subcritical under cold conditions.
GDC-27, "Combined reactivity control system capability," requires the reactivity control systems to be designed to have a combined capability, in conjunction with poison addition by the ECCS, of reliably controlling reactivity changes to assure that under postulated accident conditions and with appropriate margin for stuck rods the capability to cool the core is maintained.
GDC-28, "Reactivity limits," requires the reactivity control systems to be designed with appropriate limits on the potential amount and rate of reactivity increase to assure that the effects of postulated reactivity accidents can neither (1) result in damage to the reactor coolant pressure boundary greater than limited local yielding, nor (2) sufficiently disturb the core, its support structures or other reactor pressure vessel internals to impair significantly the capability to cool the core. These postulated reactivity accidents shall include consideration of rod ejection (unless prevented by positive means), rod dropout, steam line rupture, changes in reactor coolant temperature and pressure, and cold water addition.
GDC-35, "Emergency core cooling," requires a system that provides abundant emergency core cooling to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented, and (2) clad metal-water reaction is limited to negligible amounts.
3.0 TECHNICAL EVALUATION
1 3.1 Mechanical Design of XM Fuel for BFN Units of Reference 1 provides the mechanical design details of the XM fuel design for the BFN Units 1, 2, and 3. The XM fuel design comprises of a 1 Ox1 0 array of fuel rods with a square internal water channel that displaces a 3x3 array of rods, with 79 full length rods (FLRs), and 12 part length fuel rods (PLFRs). The active length of the PLFR is approximately one half length of the FLRs. The use of the PLFRs is expected to improve the fuel utilization in the high void upper 1 The information in (( )) contains proprietary information.
OffiCIAl USE ONlY PROPRIETARY INfORMATION
OFFICIAl USE ONlY PROPRIETARY INFORMATION region of the bundle, enhance the shutdown margin, improve stability, and pressure drop performance. Relative to ATRIUM-1 0, the PLFRs are shorter, and their radial placement within the assembly is different. Other changes in XM fuel relative to ATRIUM-10 fuel include the following:
((
))
The XM fuel assembly consists of a lower tie plate (L TP), 91 fuel rods, 9 spacer grids, a central water channel with ((
)), and miscellaneous assembly hardware. The structural connection between the L TP and upper tie plate (UTP) is provided by the central water channel.
((
))
((
))
OFFICIAl USE ONlY PROPRIETARY INFORMATION
OffiCIAl USE ONlY PROPRIETARY INfORMATION Fuel Rods
((
))
Fuel Channel and Components
((
))
3.1.1 Fuel Design Evaluation The objectives of the fuel design are that (i) the fuel system does not fail as a result of normal operation and AOOs, (ii) fuel system damage is never so severe as to prevent control rod insertion when it is required, (iii) the number of fuel rod failures is not underestimated for postulated accidents, (iv) fuel coolability is always maintained, (v) the mechanical design of the fuel assemblies shall be compatible with co-resident fuel and the reactor core internals, and (vi) fuel assemblies shall be designed to withstand the loads from handling and shipping. The first four objectives are addressed in SRP Section 4.2 and the latter two are to assure the structural integrity of the fuel and the compatibility with the existing reload fuel (co-resident fuel). This fuel design evaluation contains only fuel structural analyses where the fuel rod evaluation is documented in Attachment 21 of Reference 1 and will be discussed in Section 3.2.2 of this safety evaluation (SE).
Stress. Strain, or Loading Limits on Assembly Components American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) code (Reference 9) is used as a guide to establish acceptable stress, deformation, and load limits for standard assembly components. These limits are applied to the design and evaluation of the UTP, LTP, spacer grids, springs, and load chain components, as necessary and applicable. The fuel assembly structural component criteria under faulted conditions are based on Appendix F of the ASME B&PV Code Section Ill with some criteria derived from component tests. Outside of faulted conditions, most structural components are under the most limiting loading conditions during fuel handling.
In response to the NRC staff's request for additional information (RAI) SNPB RAI-9, AREVA has provided a summary of their stress evaluations performed to confirm the design margin and to OffiCIAl USE ONlY PROPRIETARY INfORMATION
OFFICIAl USE ONlY PROPRIETARY INFORMATION establish a baseline for adding accident loads for the determination of loading limits on fuel assembly components (Enclosure 1, Reference 2). To evaluate the stresses under normal operating conditions, ((
)). The maximum normal operation ((
)) for BFN Unit 2 Cycle 19 is then compared against the limit to ensure that adequate margin is maintained.
((
))
The NRC staff determined that the stress analysis performed has shown sufficient design margin under normal operating and AOO conditions.
Fatigue and Fretting Wear
((
)) NRC staff agrees that the lack of significant wear at the spacer cell locations relaxed to EOL conditions provides further assurance that no significant fretting will occur at higher exposure levels.
Rod Bow Differential expansion between the fuel rods and cage structure, and lateral thermal and flux gradients can lead to lateral creep bow of the rods in the spans between spacer grids. This lateral creep bow alters the pitch between the rods and may affect the peaking and local heat transfer.
The criterion for fuel rod bowing is ((
))
OFFICIAl USE ONlY PROPRIETARY INFORMATION
OFFICIAl USE ONlY PROPRIETARY INFORMATION Axial Irradiation Growth Fuel assembly components such as fuel channel must maintain clearances and engagements throughout their design life. There are three specific growth calculations for XM fuel design:
(1) minimum fuel rod clearance between L TP and UTP, (2) Minimum engagement of the fuel channel with the L TP seal spring, and (3) external interfaces (e.g., channel fastener springs).
((
)) Assembly growth is dictated by the water channel growth. ((
))
Assembly Liftoff The acceptance criterion in the NRC-approved topical report (Reference 7) is that the fuel assembly shall not levitate under normal operating, AOO or faulted conditions. Under postulated accident conditions, the fuel shall not become disengaged from the fuel support. These criteria assure control blade insertion is not impaired. For normal operating conditions, the calculated net axial force acting on the assembly due to addition of the loads from gravity, hydraulic resistance from coolant flow, difference in fluid flow entrance and exit momentum, and buoyancy will be in the downward direction, indicating no assembly liftoff. The net force calculation is performed at maximum hot channel conditions because the greater two-phase flow loses produce a higher uplift force. Mixed core conditions for assembly lift-off are considered on a cycle-specific basis, as determined by the plant and other fuel types. ((
))
3.1.2 Structural Deformations Evaluations performed for the fuel under combined seismic/LOCA loadings by using approved methodologies include mechanical fracturing of the fuel rod cladding, assembly structural integrity, and fuel assembly liftoff (References 11, 12, and 13) restricting fuel uplift and limiting fuel channel deformation under accident conditions permit insertion of the control blades.
AREVA indicates that the testing and analyses have shown the dynamic response of the XM fuel design to be very similar to the ATRIUM-10 since AREVA intends to deliver XM fuel assemblies with the Advanced Fuel Channel (AFC) design. The dynamic responses of the XM fuel and the ATRIUM-10 are very similar to other Boiling Water Reactor (BWR) fuel designs that have the same basic channel configuration and weight. This includes the previously analyzed GNF fuel at BFN. The structural response of XM to combined seismic/LOCA loadings ((
)) (Reference 2). ((
)) The Table listed for SNPB RAI-1 0 in Reference 2 lists the stiffness (area moment of Inertia), mass and natural frequency for the BFN Seismic analysis, ATRIUM-1 0 with 100/75 AFC, XM fuel with 100/75 AFC.
The results show that the existing accident analysis remains applicable to the XM fuel design and the channeled fuel assembly will experience safe shutdown earthquake acceleration with an acceptable value.
OffiCIAl USE ONlY PROPRIETARY INFORMATION
OFFICIAl USE ONlY PROPRIETARY INFORMATION NRC staff has found that the ATRIUM-10 fuel assembly component analysis remains applicable to XM fuel design based on a detailed comparison between the fuel assembly components. The allowable stress or load limits for the XM fuel were updated to new limits based on testing of XM fuel components and are listed in Table 3-1, Criterion 3.4.4 of Attachment 6 of Reference 1.
In response to NRC staff SNPB RAI-10(b), the licensee provided details of how the acceptance criteria discussed in SRP Section 4.2, Appendix A,Section IV is satisfied for LOCA for
( 1) prevention of fuel rod fragmentation and (2) not to exceed 10 CFR 50.46 temperature and oxidation limits. A full test campaign for channeled and unchanneled XM fuel assembly was used to support the transition to the XM fuel at BFN units in regards to the structural response of the fuel as described above. The licensee stated that these tests were performed in accordance with the NRC-approved methodology described in References 12 and 13. These tests included fuel assembly axial load test, spacer grid lateral impact strength test, tie plate strength tests, debris filter efficiency test, fuel assembly fretting test, fuel assembly static lateral deflection test, fuel assembly lateral vibration test and fuel assembly impact tests. A summary of test results is provided in Attachment 6 of Reference 1.
In summary, the NRC staff has reviewed the evaluation of the structural design of the assembly and fuel channel and found that the fuel assembly and channel meet all mechanical compatibility requirements for use in BFN Units 1, 2, and 3 and that the compatibility extends to both co-resident fuel and the reactor core internals.
3.2 XM Fuel Rod Thermal-Mechanical Evaluation This section presents the results of NRC staff review of fuel rod thermal-mechanical (T-M) analyses for the XM fuel that will be inserted for operation in BFN units. T -M analyses were performed using the approved codes and methodology (References 3 and 14). The cladding external oxidation limit was reduced when RODEX4 code was first implemented (Reference 14).
The RODEX4 fuel rod T-M analysis code and methodology is used to analyze the fuel rod for fuel centerline temperature, cladding strain, rod internal pressure (RIP), cladding collapse, cladding fatigue and external oxidation.
3.2.1 Fuel Material (Isotopic) Composition In 1997, TVA agreed to take several tons of highly enriched uranium to be converted to blended low enriched uranium (BLEU) for use as fuel in BFN units. AREVA and TVA had an agreement for AREVA to provide this BLEU fuel in ATRIUM-10 BWR fuel assemblies for BFN units. The primary difference between BLEU and commercial grade uranium (CGU) is the concentration of U234 and U236. BLEU material has a higher concentration of these isotopes when compared to the maximum allowed values for enriched CGU defined by American Society for Testing and Materials (ASTM) document ASTM C966-1 0. Chemically, there is no difference between BLEU and CGU. Within the fuel manufacturing process, the U234 and U236 isotopes are inseparable from the original BLEU feed stock.
OFFICIAl USE ONlY PROPRIETARY INFORMATION
OffiCIAl USE ONlY PROPRIETARY INFORMATION Both CGU and BLEU material are subject to the same maximum U235 enrichment of 4.95 percent.
The table below provides a CGU versus BLEU comparison of U234 and U236 concentrations.
Enriched CGU from ASTM C966-1 0 Typical BLEU Isotope Max allowable Equivalent Weight%
Concentration Weight%
u~::s4 1.1 OE+04 11g/g U235 0.0546 wt% u 0.09 wt% U u~::so 250 llg/g u 0.025 wt% u 1.60 wt% u The small changes in isotopic impurities of the BLEU fuel do not significantly affect the physical properties of the fuel. Isotopes of uranium (e.g., U234, U235, U236, and U238) have the same electronic structure. They also occupy the same space. Consequently, the substitution of a U234 or U236 for a U238 (or U235) atom in the lattice does not constitute a point defect and does not change the local electronic configuration. The fuel thermal conductivity is therefore independent of the U234 and U236 content as it is also independent of the amount of U235* Other T-M properties such as thermal expansion, heat capacity, enthalpy, Young's modulus, Poisson's ratio, creep, melting temperature, and emissivity are also not affected.
The primary difference in neutronic characteristics of BLEU relative to CGU fuel is decreased reactivity due to the higher concentration of U236* U236 has a neutron poisoning effect. The combination of BLEU and CGU rods has the net effect in reducing the reactivity approximately equivalent to a 0.3-percent reduction in U235 enrichment. This means that the enrichment of a BLEU assembly would need be approximately that much higher to provide the same amount of energy production. Since both CGU and BLEU assembly enrichments are limited by the same enrichment requirements, the impact is seen only in a larger required batch fraction for BLEU versus an equivalent CGU reload (Reference 2).
The lattice depletion code, CASM0-4 and 3D core simulator code, MICROBURN-B2 are used to track the isotopes of the BLEU fuel to account for their concentrations. By including the U234 and U236 concentrations explicitly in the fuel design and licensing process, the reload batch size is increased by modifying lattice enrichment and gadolinium loading.
3.2.2 Fuel Rod Design Evaluation An XM fuel rod is slightly larger in diameter than the ATRIUM-10 fuel rod. ((
)) The PLFRs are shorter in length than the length of PLFRs in ATRIUM-10. ((
))
Table 2-1 of Attachment 21 of Reference 1 provides the main parameters for the fuel rod and components. Tables 3-1 through 3-4 provide results from the fuel rod design analyses along with the design criteria for each of the items. The fuel rod analyses such as those for fuel centerline temperature and cladding strain cover normal operating conditions and AOOs.
OffiCIAl USE ONlY PROPRIETARY INFORMATION
OFFICIAl USE ONlY PROPRIETARY INFORMATION Internal Hydriding The absorption of hydrogen by the cladding can result in cladding failure due to reduced ductility and formation of hydride platelets. This is prevented by careful moisture control during fuel fabrication and reduces the potential for hydrogen absorption on the inside of the cladding.
Cladding Collapse Creep collapse of the cladding and the subsequent potential for fuel failure is avoided in the design by limiting the gap formation due to fuel densification subsequent to pellet-clad contact.
Creep collapse of the clad is evaluated using RODEX4 (Reference 3}. RODEX4 uses a statistical method and the code gives best-estimate results for nominal inputs. The maximum gap formation is calculated such that the expected fraction of fuel rods below the maximum value is 99.9 percent with a 95-percent confidence level.
Overheating of fuel pellets In order to avoid fuel failure from overheating of the fuel pellet, the centerline temperature of the fuel pellets must remain below the melting point during normal operation and AOOs. The melting point is adjusted for gadolinia content in the fuel. AREVA establishes a linear heat generation rate (LHGR) to protect against fuel centerline melting during steady-state operation and during AOOs.
Fuel centerline temperature is evaluated using the RODEX4 code (Reference 3) for both normal operating conditions and AOOs. RODEX4 fuel model considers the fuel column divided in to axial and radial regions, gap region, cladding, gas plena and the fill gas and released fission gases.
The operational conditions are controlled by the ((
)) The heat conduction for the clad and the fuel is calculated with a general variable mesh to accommodate steep temperature gradients.
Mechanical processes include ((
))
((
))
((
OFFICIAl USE ONlY PROPRIETARY INFORMATION
OffiCIAl USE ONlY PROPRIETARY INfORMATION )) (Reference 19).
((
))
((
))
Stress and Strain limits Cladding strain caused by transient-induced deformations of the cladding is calculated using the RODEX4 code and methodology as described in Reference 3. The calculated strain is reported to be less than 1 percent.
Cladding stresses are calculated using solid mechanics elasticity solutions and finite element methods. Stresses are calculated for the primary and secondary loadings. Primary loadings consist of uniform hoop, axial and radial stresses (membrane) and differential pressure and ovality and flow-induced vibration (bending). Secondary loadings consist of differential thermal expansion steady-state pellet cladding interaction (PCI) (membrane) and restraint against mechanical and thermal bow steady-state PCI spacer contact stresses (bending). The stresses are found to be less than the design limits prescribed by ASME B&PV Code Section Ill.
Fuel densification and swelling Fuel densification and swelling are limited by the design criteria for fuel temperature, cladding strain, cladding collapse, and rod internal pressure criteria.
Fatigue
((
)) A maximum value that encompasses OffiCIAl USE ONlY PROPRIETARY INfORMATION
OFFICIAl USE ONlY PROPRIETARY INFORMATION 99.9 percent of the fuel rods with a 95-percent confidence is determined. The maximum cumulative usage factor for the cladding remains below the design criterion.
Oxidation. Hydriding, and Crud Buildup The RODEX4 calculation of cladding external oxidation includes an enhancement factor that is derived from poolside measurement data to obtain a fit of the expected oxide thickness. An uncertainty on the model enhancement factor also is determined from the data. RODEX4 analysis implicitly includes the thermal effect from normal levels of crud. Specific analyses are performed for higher than normal crud deposition. An abnormal level of crud is defined by a formation that increases the calculated fuel average temperature by 25 degrees Celsius above the design basis calculation. The corrosion model also takes into consideration the effect of the higher thermal resistance from the crud on the corrosion rate.
TVA instituted a healthy fuel inspection program in the mid-2000s, as part of establishing a baseline for fuel performance at BFN. The scope of this program typically involves the inspection of one highest-exposure one-time-burnt assembly, a high-exposure twice-burnt assembly and a high-exposure thrice-burnt assembly from each batch following completion of an operating cycle.
The inspection typically includes a peripheral examination of the bundle with the channel removed, to assess general performance and ensure no abnormal physical distortion is present. A limited number of fuel rods are removed, washed to remove the crud, and measurements of liftoff are taken on each of the removed fuel rod, along with profilometry measurements and eddy current testing for flaws. The licensee stated that the scope of this program has been expanded with the implementation of On Line Noble Chemistry (OLNC) (Reference 2) beginning with BFN Unit 3 Cycle 15.
Specifically, the OLNC program involves bulk analysis of selected fuel deposit particulate samples for elements and metallic distribution and crystal grain size and evaluation of the OD and inner diameter of the buildup (deposit) for porosity and density, the distribution and size of boiling chimneys, and the elemental and metal oxide distribution within any given flake. TVA has reported that using the data obtained from OLNC program AREVA has benchmarked the crud and corrosion risk assessment tool for BFN Unit 3 (Reference 2).
A safety evaluation report (SER) restriction imposed on RODEX4 required that the calculations account for an expected, design basis crud thickness and it may be based on plant-specific history. As part of the RAI responses to the RODEX4 topical report, it was stated that the existing corrosion model includes a design basis level of crud. In the case of the BFN, plant-specific liftoff data are available from the fuel surveillance program described above. Fuel assemblies with the highest end-of-cycle exposures were selected for measurement at the BFN units with the intent to obtain liftoff values that conservatively represent the reload batches.
During the first reload application of RODEX4, the initial approved limit of corrosion was challenged by NRC staff because of a concern about the effect of spallation on the cladding integrity. To avoid the issue of spallation, the limit was reduced to ((
)). The ((
))
limit was established from a review of historical liftoff measurement data on AREVA BWR fuel. As a result, the licensee made a regulatory commitment to reduce the limit on predicted cladding peak oxide thickness, as defined by RODEX4 to ((
)) (see Section 5.0 of this SE).
In response to an NRC staff's RAI, the licensee reported (Reference 2) that the maximum calculated corrosion that represented the 99.9-percent quantile with a 95-percent confidence OFFICIAl USE ONlY PROPRIETARY INFORMATION
OffiCIAl USE ONlY PROPRIETARY INFORMATION level was found to be ((
)). This means that, it is expected that ((
((
))
)) The maximum calculated corrosion limit is lower than the limit established by the NRC for AREVA BWR fuels.
Therefore, the NRC staff has found this calculated result to be acceptable.
In response to an RAI from NRC staff regarding the regulatory commitment made to reduce the limit on the calculated oxide to the value indicated in Reference 3 of ANP-3159P, the licensee has made a commitment related to complying with the reduced oxide limit being included in the BWR specific requirements appendix of the TVA procedure that governs the overall process of core design and nuclear analysis as part of the implementation of this license amendment (Enclosure 2 of Reference 2). TVA stated that the oxide thickness will be included in the BFN procedures as an item to be verified prior to each reload cycle where the RODEX4 code is applied.
The licensee has assured the staff that this procedural requirement will include a cross reference back to the commitment tracking number as an NRC commitment. TVA confirmed that the requirement to meet this calculated oxide limit will be included in the Reload Requirements Specification for each reload design.
Rod internal pressure Fuel rod internal pressure is calculated using the RODEX4 code and methodology. The maximum rod pressure is calculated under steady-state conditions and transients. Rod internal pressure is limited to ((
)) above the rated system pressure.
In summary, the NRC staff has reviewed the application of approved code and methodologies in the fuel rod T-M analyses of XM fuel design that will be inserted for operation in BFN units. The staff has determined that the fuel design criteria as set forth by the applicable regulations and Section 4.2 of SRP have been satisfied for the safe operation of the BFN units.
3.3 Thermal Hydraulic Design of XM Fuel Assemblies for BFN This section describes the staff review results of BFN thermal-hydraulic analyses presented to the NRC to demonstrate the hydraulic compatibility of the XM fuel with coresident fuel.
TVA is proposing to transition to XM fuel design in BFN Unit 2 starting with Cycle 19 (spring 2015),
Unit 3 in spring 2016, followed by Unit 1 in fall of 2016. The LAR is mainly for the lead cycle application, which is Unit 2 Cycle 19. The Unit 2 transition design is representative of an XM fuel transition for the other two units, since for all three units the balance of the core will be standard ATRIUM-10 fuel when the first XM fuel reload is inserted. BFN Unit 1 currently has a mix of ATRIUM-10 and GE14 fuel, all of the GE14 fuel will be discharged in the fall of 2016 outage when XM will be introduced.
Thermal-hydraulic analyses per Reference 7 are performed to verify that design criteria are satisfied and to help establish thermal operating limits with acceptable margins of safety during normal reactor operation and AOOs. Due to reactor and cycle operating differences many of the analyses supporting these thermal-hydraulic operating limits are performed on a plant and cycle-specific basis and are documented in plant and cycle-specific reports (Reference 1 ).
Table 3.1 of Attachment 8 of Reference 1 lists the applicable thermal-hydraulic design criteria.
Analyses and results are presented in Tables 3.5 to 3.10 for hydraulic compatibility, thermal margin performance, fuel centerline temperature, rod bow, bypass flow, stability, LOCA analysis, OffiCIAl USE ONlY PROPRIETARY INFORMATION
OFFICIAl USE ONlY PROPRIETARY INFORMATION control rod drop accident analysis, ASME overpressurization analysis and seismic/LOCA liftoff.
Sections below summarize the results from selected design criteria and analyses results.
3.3.1 Hydraulic Characterization Basic dimension parameters for XM, ATRIUM-10 and GE14 fuel designs are summarized in Table 3.2 of Attachment 8 of the BFN submittal. Compared to ATRIUM 10, the XM fuel has a
((
)). Table 3.3 of Attachment 8 provides loss coefficients that include modifications to the test data reduction process. These modifications account for the ((
))
The transition core at BFN consists of ATRIUM-10 with both Standard FUELGUARD (SFG) and Improved FUELGUARD (IFG). The ATRIUM 10A with SFG has ((
)) blades and (( )) grid rods, with blades assembled in slots and grid rods inserted and brazed together. The IFG is similar to SFG except for ((
)) half interstitial strips that run parallel to the grid rods and are utilized to increase filter efficiency. Figures 4-1, 4-2, 4-3, and 4-4 of Reference 2 (ANP-3248P) provide detailed diagrams of SFG and IFG, top, bottom, and side views.
The impact on pressure drop between the SFG and IFG LTP is best shown in the full core evaluations shown in Tables 3.9 and 3.10 of ANP-3082P (Attachment 8 of Reference 1 ). The Table 3.9 provides core pressure drop, core bypass flow fraction for transition cores ranging from full core GE14, full core ATRIUM-10 SFG, transition core loading 1 through transition core loading 4 (Table 3.4 of ANP-3082), full core ATRIUM-10 IFG and full core XM for BFN 1 00-percent current licensed thermal power (CL TP) and 1 00-percent flow conditions and for radial peaking factors 1.00 and 1.50, the corresponding critical power ratios (CPRs) and assembly flow. Table 3.10 lists the same thermal hydraulic results as in Table 3.9, except at 62-percent CL TP and 37.3-percent core flow. The differences between core pressure drops at rated conditions (1 00-percent CL TP and 1 00-percent flow) and at off rated conditions (62-percent CL TP and 37.3-percent flow) for the two cases of L TPs are not significant. The results given in Tables 3.9 and 3.10 of ANP-3082 indicate that the critical power performance of the SFG is not significantly affected by the introduction of the IFG L TP design. The NRC staff accepts that the introduction of IFG for ATRIUM-1 0 fuel design has not significantly affected the hydraulic characterization for loss coefficients and pressure drops for the BFN mixed cores during the transition to XM fuel design.
3.3.2 Hydraulic Compatibility The thermal-hydraulic compatibility analyses were performed in accordance with the AREVA thermal hydraulic methodology for BWRs (Reference 15). The X COBRA code predicts steady-state thermal hydraulic performance of the fuel assemblies of BWR cores at various operating conditions and power distributions. Thermal-hydraulic compatibility analysis evaluates the relative thermal performance of the XM, ATRIUM-10, and GE14 fuel designs that are in BFN one time or other. Analyses have been performed for full core GE14, full core ATRIUM-1 0 SFG, full core ATRIUM-1 0 IFG, and full core XM configurations. Analyses for mixed core configurations were also performed to demonstrate the thermal-hydraulic compatibility for resident and co-resident fuel designs.
OFFICIAl USE ONlY PROPRIETARY INFORMATION
OFFICIAl USE ONlY PROPRIETARY INFORMATION Hydraulic compatibility analysis models each of the fuel assembly channels in the core such that the pressure drop across the channels is the same. Table 3.4 of ANP-3082 lists a summary of all inputs in to the hydraulic compatibility analysis. The inputs are for rated (1 00-percent power/1 00-percent flow) and off-rated (62-percent power/37.3-percent flow) conditions of power and core flow. The addition of this off-rated statepoint is to demonstrate hydraulic compatibility is maintained for both rated and off-rated conditions. Analysis at two state points (at rated and off-rated conditions) was performed in the compatibility analyses to adequately support operation in the power/flow map. The selection of the off-rated state point considered the following operational conditions: ((
)) For BFN, this is represented as 62-percent power/37.3-percent flow.
The evaluations are made for all transition core loadings 1 through 4 and with bottom-, middle-,
and top-peaked axial power distributions as presented in Table 3.4 of ANP-3082P. Results of the hydraulic compatibility analyses are presented in Tables 3.5 through 3.10 and Figures 3.2 through 3.9 of ANP-3082 for bottom-peaked power distribution. Results for the middle-peaked and top-peaked axial power distributions show similar trends.
Transition Core Loading 1 is a core consisting of approximately one third XM fuel and one-third ATRIUM-10 IFG fuel with the remainder GE14 fuel. For rated and off-rated conditions the flow to the maximum power XM and ATRIUM-1 0 IFG assemblies are within ((
)) of the flow to the GE14 assembly at the same power level as illustrated in Figure 3.2 and the results provided in Tables 3.5 and 3.9 of ANP-3082. Based on the reported changes in pressure drop and assembly flow caused by the transition from GE14 fuel, the NRC staff considers that both the XM and ATRIUM-10 IFG designs are hydraulically compatible with the GE14 design since the thermal-hydraulic design criteria are satisfied.
Transition Core Loading 2 (Table 3.4) is a transition core consisting of approximately one-third XM fuel and one-third ATRIUM 10 IFG fuel with the remainder ATRIUM-10 SFG Fuel and is considered a representative transition core from a full core of ATRIUM-10 SFG fuel to a full core of XM fuel including ATRIUM-10 IFG fuel. The core average results and the differences between the fuel designs for both rated and off-rated statepoints are within the range considered hydraulically compatible as shown in Table 3.6 of ANP-3082.
Transition Core Loading 3 (Table 3.4) is a transition core consisting of approximately one-third XM fuel with the remainder ATRIUM-1 0 IFG fuel and is a representative transition core from a full core ATRIUM-1 0 IFG fuel to a full core of XM fuel. Based on the reported results for changes in pressure drop and assembly flow caused by the transition from ATRIUM-1 0 IFG to XM fuel, the XM design is considered hydraulically compatible with the ATRIUM-10 IFG design since the thermal-hydraulic design criteria are satisfied.
Transition Core Loading 4 (Table 3.4) is a transition core consisting of approximately two-thirds XM fuel with the remainder ATRIUM-10 IFG fuel and is a representative transition core for a second reload of XM fuel at BFN. Based on the reported results on changes in pressure drop and assembly flow caused by the second reload of XM at BFN, the XM design is considered OFFICIAl USE ONlY PROPRIETARY INFORMATION
OFFICIAl USE ONlY PROPRIETARY INFORMATION hydraulically compatible with the co-resident fuel (ATRIUM-10 IFG) since the thermal-hydraulic design criteria are satisfied.
In summary, the NRC staff has determined that hydraulic compatibility analyses for the transition cores at BFN units has provided assurance that the resident and co-resident fuel designs satisfy the thermal-hydraulic design criteria for mixed cores.
3.3.3 Thermal Margin Performance Thermal margin analyses were performed using the thermal hydraulic methodology and the XCOBRA code. The calculation of fuel assembly CPR (thermal margin performance) is established by means of an empirical correlation based on results of boiling transition test programs. The details of the CPR calculations will be discussed in separate sections of this SE (Sections 3.6 and 3.7). The CPR methodology for XM fuel is the approach used by AREVA to determine the margin to thermal limits for BWRs and the methodology is described in Reference 5 and Attachment 27 (ANP-3140P) of Reference 1. For the ATRIUM-10 SFG, ATRIUM-1 0 IFG and GE14, the CPR values are calculated using the SPCB critical power correlation (Reference 16).
Assembly design features are incorporated in the CPR calculation through the K-factor term in the ACE correlation and the F-effective term for the SPCB correlation. The K-factors and F-effectives are based on the local power peaking for the nuclear design and on additive constants determined in accordance with approved procedures.
Table 3.5 through Table 3.8 of ANP-3082P lists representative CPRs of the XM fuel, ATRIUM-10 SFG, ATRIUM-10 IFG and GE14 fuel designs. Tables 3.9 and 3.10 of ANP-3082P show similar comparisons of CPR and assembly flow for the various core configurations. The NRC staff has determined that the introduction of XM will not cause adverse impact on thermal margin for the co-resident fuel.
3.3.4 Rod Bow Differential expansion between the fuel rods and cage structure, and lateral thermal and flux gradients can lead to lateral creep bow of the rods in the spans between spacer grids and this lateral creep bow alters the pitch between the rods and may affect the peaking and local heat transfer. The design criteria related to rod bow is that ((
)) Less rod bow is expected for the XM compared to ATRIUM 10 due to a larger diameter fuel rod and a reduced distance between most spacer grids (Attachment 6 of Reference 1 ).
At higher exposures a CPR penalty is determined as a function of exposure and fractional rod closure. AREVA's BWR rod bow CPR penalty is based on rod closure using an open literature data (Reference 17) and it is concluded that thermal margins were not substantially reduced for closures up to 30 percent. To assure that the model is conservative, a CHF test was performed on an ATRIUM-10 bundle in which two rods were welded together. As indicated in the SNPB RAI 5-2 of Reference 2, the penalty factor was over-predicted by a factor of 2. AREVA used the NRC-approved correlation described in Reference 18, Supplement 1. The gap closure correlations reported in AREVA's response to staff's RAI (Reference 2) assume a linear functionality of fuel burnup and initial rod-to-rod spacing. Once the gap closure is calculated at a given burnup level, the corresponding CPR penalty is determined from the Figure SNPB RAI 5-1 OFFICIAl USE ONlY PROPRIETARY INFORMATION
OFFICIAl USE ONlY PROPRIETARY INFORMATION of Reference 2. From the correlation in Reference 18, it is evident that the primary factors impacting rod bow are ((
11 ((
11 3.3.5 Bypass Flow Total core bypass flow is defined as leakage flow through the L TP flow holes, channel seal, core support plate, and L TP-fuel support interface. Tables 3.9 and 3.10 of ANP-3082P provides results for the core bypass flow fraction for the rated and off-rated conditions of power and core flow during the transition from a full GE14 fuel core to a full XM core. Differences in bypass flow fractions between other transition core combinations of AREVA fuel and G E 14 are either equal to or less than the full core GE14 fuel to a full core XM fuel results. The staff is assured that adequate bypass flow will be available with the introduction of the XM fuel design. Therefore, fuel design guidance in SRP as supported by applicable regulations are met.
In summary, the NRC staff has reviewed all thermal-hydraulic analyses and results to demonstrate that the XM fuel design is hydraulically compatible with ATRIUM-1 0 and GE14 fuel in the BFN Units. The staff has determined that the generic thermal-hydraulic design criteria as approved by the NRC in the topical report, ANF-89-98PA Revision 1 and Supplement 1 (Reference 7) have been used in the analyses. The NRC staff concluded that though the XM, ATRIUM-10 (SFG and IFG), and the GE14 fuel assemblies are geometrically different, they are hydraulically compatible.
3.4 BFN XM Equilibrium Cycle Design 3 to Reference 1 (ANP-3148P) summarizes the fuel cycle design and fuel management calculations for an XM equilibrium cycle for BFN. These analyses have been performed using the CASM0-4, a lattice depletion code, for generation of nuclear cross sections data, and MICROBURN-82, the 3-dimensional core simulator code, for pin power reconstruction for thermal margin analysis (References 19, 5, and 20).
The BFN, Units equilibrium cycle, arbitrarily named Cycle 25 has 280 fresh assemblies with enrichment slightly over 4 weight percent (w/o) was determined to meet the energy requirements for the BFN/TVA. Appendix B, Figures 8.1 through 8.3 of ANP-3148P, provide Cycle 25 fresh reload fuel design axial enrichment and gadolinia distributions. The loading pattern maintains quarter core symmetry within a scatter load fuel management scheme. Appendix A of ANP-3148P shows acceptable power peaking and associated margins to limits for projected Cycle 25 operation. The specific core location of the fresh assemblies in Cycle 25 is provided in Appendix C of ANP-3148P.
Table A.1 of Appendix A of ANP-3140P (Reference 4) provides control rod patterns and operating parameters including such as calculated kef!, core power, inlet subcooling, core minimum CPR, core maximum LHGR and core maximum average planar LHGR (APLHGR) for each incremental burn up step during the depletion of fuel. Table A.2 provides thermal margin calculation results for core limiting CPR, fraction of limiting CPR, core limiting LHGR, fraction of limiting LHGR, core limiting and fraction of APLHGR for depletion steps. Table 3.2 of ANP-3148P presents hot operating target ke11 values at various cycle exposures and the ke11 and margin to limits from the design depletion analysis are presented graphically in Figures 2.1 and 2.2 of ANP-3148P.
OFFICIAl USE ONlY PROPRIETARY INFORMATION
OFFICIAl USE ONlY PROPRIETARY INFORMATION The equilibrium cycle design calculations in Reference 4 have demonstrated adequate hot excess reactivity, standby liquid control shutdown margin and cold shutdown margin throughout the cycle as illustrated in Tables 3.5 and 3.6 of ANP-3148. The shutdown margin is in conformance with the TS value limit of cycle R value plus 0.38-percent delta (~)k/k at beginning of cycle (80C).
The licensee has shown how the hot excess reactivity and shutdown margin are maintained per TS values during the transition cycles and during the equilibrium cycle of operation of the 8FN units. The design and licensing process requires that cycle exposure dependent hot and cold critical eigenvalues be selected for the design cycle of interest. Once the design eigenvalue bases of a cycle are established, and the core is designed and licensed, the site is provided with data to support the testing used to demonstrate compliance with the reactivity-related TSs. The hot eigenvalue as a function of cycle exposure is provided to the site via a Cycle Management Report, and the surveillance procedure for reactivity anomaly for the unit of interest is updated to include the cycle specific data from this report. The difference between the calculated eigenvalue for the current core state point conditions and the design eigenvalue at the cycle exposure where the test is being performed is computed, to verify the difference is within the plus or minus 1-percent ~klk tolerance required by the TSs.
The cold shutdown margin test is performed at 80C. The test determines the kelt of the core in the state point condition where criticality was achieved, and then calculates the difference between this kelt and the analytically determined strong rod out kelt. The test data provided to the site also includes the R value, which is the calculated difference between the 80C and cycle minimum calculated shutdown margin values. The 80C demonstrated shutdown margin value is reduced by the R value prior to comparing to the TS criteria of 0.38-percent ~klk. The cold eigenvalues for cycle exposures past 80C are chosen to be conservative, because the amount of cold critical benchmark data past 80C is more limited in comparison to the available hot data. In addition to the selection of a conservative cold eigenvalue assumption, the core is designed to a minimum target shutdown margin of 1-percent ~k/k, which allows for additional uncertainty in the design calculations. The combination of the two conservatisms provides a high degree of confidence that the TS shutdown margin is maintained.
The NRC staff has determined that the cycle design calculations and the projected control rod patterns for the equilibrium cycle are developed to be consistent with a conservative margin to thermal limits.
3.5 8FN Unit 2 Cycle 19 Fuel Cycle Design 0 to Reference 1 (AN P-3145P) summarizes the fuel cycle design and fuel management calculations for an XM Cycle 19 for 8FN that is expected to start operation in spring 2015. These analyses have been performed using the CASM0-41attice depletion code for generation of nuclear cross sections data and local peaking factors and MICR08URN-82, the 3-dimensional core simulator code for pin power reconstruction for thermal margin analysis (References 19, 5, and 20). Design results for the Cycle 19 core loading include the projected control rod patterns and evaluations of thermal and reactivity margin. The Cycle 19 results are based on Cycle 18 core operational history.
The 8FN Unit 2 Cycle 19 has 272 fresh assemblies with enrichment slightly below 4 w/o was determined to meet the energy requirements for the 8FN. Appendix 8, Figures 8.1 through 8.3 of OFFICIAl USE ONlY PROPRIETARY INFORMATION
OFFICIAl USE ONlY PROPRIETARY INFORMATION ANP-3145P, provide Cycle 19 fresh reload fuel design axial enrichment and gadolinia distributions for BFN Unit 2. The loading pattern maintains full core symmetry within a scatter load fuel management scheme. Appendix A of ANP-3145P shows acceptable power peaking and associated margins to limits for projected Cycle 19 operation. The specific core location of the fresh assemblies in Cycle 19 is provided in Appendix C of ANP-3148P. Key results are summarized in Table 2.1 of ANP-3145P.
Table A.1 of Appendix A of ANP-3145P provides control rod patterns and operating parameters including such as calculated keff, core power, inlet subcooling, core minimum CPR, core maximum LHGR and core MAPLHGR for each incremental burnup step during the depletion of fuel.
Table A.2 provides thermal margin calculation results for core limiting CPR, fraction of limiting CPR, core limiting LHGR, fraction of limiting LHGR, core limiting and fraction of APLHGR for depletion steps. Table 3.2 presents hot operating target keffvalues at various cycle exposures and the keff and margin to limits from the design depletion analysis are presented graphically in Figures 2.1 and 2.2 of ANP-3145P.
Nuclear design analyses are used for nuclear fuel assembly design and core design. The core design analysis demonstrates operating margins for minimum critical power ratio (MCPR),
MAPLHGR, and LHGR. An LHGR limit and an allowable set of LHGR power and flow-dependent multipliers, that reduce the LHGR limit as a function of power and flow, are established for each fuel design. The LHGR limit is a steady state operating fuel design limit (FDL), and the multipliers are determined to protect the fuel against power transients (Reference 3). The FDL is established to ensure that the T-M criteria are met using power histories from expected operation but generated assuming an AOO occurs during the life of the assembly. The RODEX4 licensing topical report (Reference 3) details the approved methodology used to calculate the FDL including transient evaluations. The results of the cycle-specific licensing analyses are also used to establish power and/or flow dependent multipliers to the LHGR limits where needed. These power and flow dependent multipliers are identified as LHGRFACp and LHGRFACt. respectively.
The application of these multipliers ensures that the thermal mechanical criteria are met throughout the operating domain for both steady-state operation and potential licensing transients.
The steady-state LHGRs for XM and ATRIUM-1 0 fuel designs are listed in Table 8.4 of Reference
- 21. The LHGRFACp and LHGRFACp multipliers for the reference cycle are listed in Tables 8.5 and 8.6, respectively, of Reference 21. The sets of LHGR multipliers are established such that they limit the transients, as necessary, and such that the analyses satisfy the transient criteria for cladding strain (1.0-percent strain limit) and fuel overheating (fuel melt limit) (Reference 2).
Reference 2, RAI-12(b) provides a detailed summary of the LHGR multiplier limit analyses.
AOO events are divided into two basic categories-slow transients and fast transients. Both sets of analyses are performed using the RODEX4 T-M methodology. A slow transient is defined as one that can be analyzed using a steady-state solution. The slow or steady-state transient LHGR design limit for RODEX4 is expressed in terms of an allowable overpower ratio for a transient.
This ratio is defined as the maximum tolerable rod nodal power calculated during a transient divided by the steady-state power level just prior to the transient. This ratio is determined such that the transient design criteria are satisfied.
The NRC staff has determined that the TVA's Cycle 19 design calculations have demonstrated adequate hot excess reactivity, SLC shutdown margin and cold shutdown margin throughout the cycle as illustrated in Tables 3.5 and 3.6 of ANP-3145P. The shutdown margin is in conformance with the TS value limit of cycle R value plus 0.38-percent flklk at BOC.
OFFICIAl USE ONlY PROPRIETARY INFORMATION
OFFICIAl USE ONlY PROPRIETARY INFORMATION The licensee has shown how the hot excess reactivity and shutdown margin are maintained per TS values throughout the operating cycle as explained in Section. 3.4 of this SE.
The NRC staff has determined that the cycle design calculations and the projected control rod patterns for the Cycle 19 of BFN Unit 2 are developed to be consistent with a conservative margin to thermal limits.
3.6 BFN Units Improved K-Factor Model for ACE/XM Critical Power Correlation This section describes how the ACE/XM critical power correlation (CPC) (Reference 5) as revised by Attachment 27 of Reference 1 (ANP-3140P (Reference 4)) is used in the licensing analyses for BFN Units 1, 2, and 3. Reference 5 presents the approved ACE/XM CPC for XM fuel design. A deficiency was identified in the calculation of the K-factor within the ACE/ATRIUM-10XM CPR correlations (Reference 4). These deficiencies were shown to have influenced the predicted results in a non-conservative manner for this CPR correlation, for fuel assemblies with a downskew axial power shape. Since the K-factor was integrated over the entire heated length of the assembly, it was possible for the local peaking factors in the upper lattices to contribute significantly to the K-factor used, even when dryout occurs much lower in the bundle.
Reference 4 presents a revision to the ACE critical power correlation for XM fuel. The Reference 4 correlation is very similar to the Reference 5 critical power correlation with a couple of exceptions. The K-factor methodology was modified in response to deficiencies found in the axial averaging process. Also, the additive constants were revised as a result of the change to the K-factor model. Reference 4 is a plant specific application of the original ACE/XM critical power correlation as modified for application at BFN Units 1, 2, and 3 for the proposed introduction of XM fuel design.
The model equation for the ACE correlation is given in Equation 3.1 of ANP-3140P and Reference 5. In the original formulation (see ANP-10298PA, Revision 0), the local rod K-factor was calculated as an axial average, by integrating over the heated length of the rod. This approach was intended as a means of capturing the total effect of individual rod power on critical heat flux behavior in the assembly, including the effect of axial variation in individual rod power distribution. However, the integration introduces a subtle bias into the local rod K-factor values, with the implicit nonphysical assumption that downstream conditions as well as upstream conditions affect local dryout behavior. The correction for this deficiency, as documented in Reference 4 was to reformulate the dryout power model to calculate the local radial peaking term as a function of axial location, rather than integrating to obtain an axial average value.
Equation 3.1 of Reference 4 is solved for the margin to dryout at each location and the power is adjusted until the node with minimum margin is at dryout. This means that the margin at each axial node is based on the integration of the equation up to the node (i.e., the solution at the limiting node is independent of the conditions in the node above the limiting node (dryout node)).
Use of the local conditions rather than the axial average results in a definition of the local rod K-factor that more accurately captures the effect of axial power distribution on dryout behavior.
The critical power behavior of the individual fuel rods within the fuel bundle is influenced by the spacers and the bundle geometry. Additive constants are factors that distinguish the critical power performance of each rod and they are position dependent. They are considered as a flow/enthalpy redistribution characteristic for a given bundle and spacer design. All other components of the model were unchanged, but the axial resolution of the model was increased to more accurately capture the shape of the axial power distribution for each rod in the assembly.
OFFICIAl USE ONlY PROPRIETARY INFORMATION
OFFICIAl USE ONlY PROPRIETARY INFORMATION With these revisions implemented, the additive constants were re-derived, using the same procedure documented originally in ANP-10298PA, Revision 0. All other empirical coefficients of the ACE/ATRIUM-1 OXM CPR correlation were unchanged in this procedure, which generated a new set of additive constants for use with this correlation in applications to reactor analyses.
Estimation of initial additive constants for full length and part-length rods as well as the iteration scheme for the determination of final additive constants are discussed in detail in Sections 3.3.1 through 3.3.4 of Reference 4.
The NRC staff finds that the overall effect of the changes documented in the improved K-factor methodology introduces physically realistic modeling of the local subchannel hydrodynamics that influence dryout behavior in a fuel rod array. The staff finds that the proposed corrections are an acceptable improvement to the dryout modeling approach used in the ACE/ATRIUM-10XM CPR correlation that will be used in the thermal margin calculations at BFN units.
3.7 MCPR Safety Limit Analysis with SAFLIM3D Methodology 3.7.1 Purpose and Methodology NRC approved AREVA methodology for the determination of the SLMCPR is presented in Reference 22. The SLMCPR methodology was updated to incorporate full implementation of the ACE critical power correlation (References 4, 5, and 23), a realistic fuel channel bow model (Reference 3), and expanded coupling with the MICROBURN-82 core simulator (Reference 19).
This section provides SLMCPR results for BFN Unit 2 Cycle 19 (BFN2-19) using the Reference 22 methodology to support a change in the list of approved methodologies in the TSs and also a change in the TS SLMCPR values for two-loop operation (TLO) and single-loop operation (SLO).
The deficiency in the original formulation of ACE/XM correlation (Reference 5) as described in Section 3.6 of this SE does not have a significant impact on the representative BFN Unit 2 Cycle 19 SLMCPR values. However, the implementation of the modified K-factor and additive constants that constitute the ACE correlation as described in Section 3.6 and in Reference 4 rectifies the deficiency. In addition, the NRC staff has approved a supplement to Reference 5, which describes the details of the improved methodology (Reference 23).
The SLMCPR is defined as the minimum value of the CPR, which ensures that at least 99.9 percent of the fuel rods in the core are expected to avoid boiling transition (BT) during normal operation or an AOO. The SLMCPR methodology uses a statistical approach that employs a Monte Carlo process where the input parameters are perturbed. The SLMCPR analysis is performed with a power distribution that conservatively represents expected reactor operating states that could both exist at the operating limit MCPR (OLMCPR) and produce an MCPR equal to the SLMCPR during an AOO. The conservative approach used in the analysis through uncertainties creates near limiting MCPR assemblies that are more likely to produce rods in BT.
The AREVA SLMCPR methodology includes the effects of channel bow on the critical power performance.
3.7.2 Analysis and Results The BFN Unit 2 Cycle 19 design supports licensed rated power of 3,458 MWt and operation to licensing end of cycle (EOC) cycle exposure of approximately 16,941 MWd/MTU. The design includes extensions for final feedwater temperature reduction (FFTR) and coastdown. The core is made up of XM and ATRIUM-10 fuel. Analyses were performed ((
)) for the BFN power/flow map OFFICIAl USE ONlY PROPRIETARY INFORMATION
OffiCIAl USE ONlY PROPRIETARY INfORMATION for MELLLA operation as shown in Figure 2 of Reference 6. The ACE/XM critical power correlation (References 4, 5, and 23) is used for the XM fuel while the SPCB critical power correlation (Reference 16) is used for the ATRI UM-1 0. The fuel-and plant-related uncertainties used in the BFN2-19 SLMCPR analysis are presented in Table 1 of Reference 6. The radial and nodal power uncertainties used in the analysis include the effects of up to 40 percent of the traversing incore probe channels out-of-service (OOS), up to 50 percent of the low-power range monitors (LPRMs), and a 2500 effective full-power hour LPRM calibration interval.
Table 3.1 below presents a comparison of the percentage of rods in BT calculated for the lowest supportable and submitted TLO and SLO SLMCPRs. The calculated rods in transition are equal to or worse when implementing the updated ACE/XM correlation.
Table 3.1 Safety Limit MCPR and Percent of Rods in Boiling Transition ACE/XM Critical Power NRC-Approved ACE/XM Loop Configuration SLMCPR Correlation with Nodal K-factor method Critical Power Correlation Percent Rods in Boiling Transition TLO 1.04 0.0834 0.0820 1.06 0.0417 0.0360 SLO 1.05 0.0921 0.0719 1.08 0.0331 0.0331 The NRC staff has reviewed the results from the TLO and SLO SLMCPR analyses that incorporated realistic channel bow model, expanded coupling with MICROBURN-B2 through the pin-power distributions. The staff finds the application of methodology and the results acceptable for BFN Unit 2 operation.
3.8 Emergency Core Cooling System Performance ECCS is designed to mitigate postulated LOCAs caused by rupture in primary system coolant piping. The ECCS performance under all LOCA conditions and the evaluation model must satisfy 1 0 CFR 50.46 and 10 CFR Part 50 Appendix K. The ECCS performance analysis consist of (1) LOCA break spectrum analysis to identify the parameters that result in the highest calculated peak cladding temperature (PCT) during a postulated LOCA (Reference 24) and (2) to identify the maximum average planar linear heat generation rate (MAPLHGR) limit versus exposure for XM fuel and to demonstrate that the MAPLHGR limits is adequate to ensure that the LOCA-ECCS criteria in 10 CFR 50.46 are satisfied for operation at or below the limit (Reference 25).
3.8.1 BFN Units LOCA Break Spectrum Analysis for XM Fuel Introduction For a BWR, a LOCA may occur over a wide spectrum of break locations and sizes. Responses to the break vary significantly over the break spectrum. The largest possible break is a double-ended rupture of a recirculation pipe; however, this is not necessarily the most severe OffiCIAl USE ONlY PROPRIETARY INfORMATION
OFFICIAl USE ONlY PROPRIETARY INFORMATION challenge to the ECCS. In addition to break location dependence, different break sizes in the same pipe produce quite different event responses, and the largest break area is not necessarily the most severe challenge to the event acceptance criteria. Because of significant variations in responses over a break spectrum, an analysis covering the full range of break sizes and locations is performed to identify the limiting break characteristics (Reference 24).
Methodology and LOCA Description Regardless of the initiating break characteristics, the LOCA event is separated into three phases, the blowdown phase, the refill phase, and the reflood phase. During the blowdown phase of a LOCA, there is a net loss of coolant inventory, an increase in fuel cladding temperature due to core flow degradation, and the core may become fully or partially uncovered depending on the break size. During the refill phase of a LOCA, the ECCS is functioning and there is a net increase of coolant inventory due to the activation of the core sprays that provide core cooling. The low pressure and high pressure coolant injection systems supply coolant to refill the lower portion of the reactor vessel. During the reflood phase, when the coolant inventory has increased, the cooling is provided above the mixture level by entrained reflood liquid. The ECCS must be designed such that the plant response to a LOCA meets the acceptance criteria specified in 10 CFR 50.46(b):
The calculated maximum fuel element cladding temperature shall not exceed 2200 °F.
The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation.
The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.
Calculated changes in core geometry shall be such that the core remains amenable to cooling.
After any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core.
A MAPLHGR limit is established for the XM fuel type to ensure that these criteria are met (Reference 25).
The evaluation model used for the BFN units LOCA analysis is the NRC-approved EXEM BWR-2000 LOCA analysis methodology described in Reference 26. The EXEM BWR-2000 employs three major computer codes, RELAX, HUXY, and RODEX2, to evaluate the system and fuel response during all phases of a LOCA. RELAX is used to calculate the system and hot channel response during the blowdown, refill and reflood phases of the LOCA. The HUXY code is used to perform heatup calculations for the entire LOCA, and calculates the PCT and local clad oxidation at the axial plane of interest. RODEX2 is used to determine fuel parameters (such as stored energy) for input to the other LOCA codes (References 27 and 28). A complete analysis for a given break size starts with the specification of fuel parameters using RODEX2. RODEX2 is then used to determine the initial stored energy for both the blowdown analysis (RELAX hot OFFICIAl USE ONlY PROPRIETARY INFORMATION
OFFICIAl USE ONlY PROPRIETARY INFORMATION channel} and the heatup analysis (HUXY). This is accomplished by ensuring that the initial stored energy in RELAX and HUXY is the same or higher than that calculated by RODEX2 for the power, exposure, and fuel design being considered.
RODEX2 code is used to calculate fuel mechanical parameters for use in the HUXY computer code that potentially impact the clad ballooning and rupture models. Though clad ballooning has a small impact on PCT, and metal water reaction (MWR), clad rupture can have a significant impact on PCT, depending on event timing. Sensitive studies have shown that the BWR LOCA analyses are insensitive to initial stored energy. After the initial phase of a LOCA, the heat transfer coefficient at the cladding surface is degraded due to the loss of coolant (low flow and high quality). As a result, the heat transfer from the fuel is primarily controlled by the surface heat flux, and the temperature profile across the pellet is very flat. When compared to the rod surface thermal resistance, the pellet thermal conductivity is not a significant portion of the fuel rod total thermal resistance. Therefore, LOCA calculations are not sensitive to the U02 thermal conductivity used in RELAX and HUXY.
To demonstrate that LOCA calculations were not sensitive to U02 thermal conductivity, assessments were performed for multiple BWRs. Assessments of the potential impact of exposure degradation of U02 thermal conductivity on the fuel mechanical parameters were made using the RODEX4 computer code. The RODEX4 code explicitly incorporates the impact of U02 thermal conductivity degradation with exposure. The assessment based on the RODEX4 computer code was used to make adjustments to input for RODEX2 to reflect the impact of the U02 thermal conductivity degradation with exposure. Based on the assessment describe above, the potential impact of U02 thermal conductivity degradation with exposure on calculated PCT and MWR for the BFN XM fuel is not significant relative to existing margins to limits (Reference 29).
The LOCA break spectrum analysis is performed for a full core of XM fuel. Table 3.2 provides a summary of reactor initial conditions used in the break spectrum analysis.
Table 3.2 Initial Conditions for Break Spectrum Analysis and Heatup Analysis (Reference 24, Table 4.1)
Parameter Values for Values for ((
((
11 Reactor power (% of rated) 102 102 Total core flow(% of rated)
((
11 Reactor Power (MWth) 3527 3527 Total core flow (Mib/hr)
((
11
((
11 Steam flow rate (Mib/hr) 14.50 14.50 Steam dome pressure (psia) 1054 1054 Core inlet enthalpy (Btu/hr)
((
11 XM hot assembly MAPLHGR 13.0 13.0 (kW/ft)
((
11 ECCS fluid temperature (°F) 120 120 OFFICIAl USE ONlY PROPRIETARY INFORMATION
))
OFFICIAl USE ONlY PROPRIETARY INFORMATION Parameter Values for Values for ((
))
((
))
Axial power shape Mid and top peaked Mid and top peaked (Figure 4.7)*
(Figure 4.8)*
- Reference 24 The LOCA analyses identify the limiting break location, break type, break size, and ECCS single failure. Potential break locations are separated into two groups: recirculation line breaks and non-recirculation line breaks. The single failures and available ECCS for each failure assumed in the break spectrum analysis are summarized in Table 5.1 of Reference 24 for recirculation line breaks and in Table 5.2 of Reference 24 for non-recirculation line breaks.
Large recirculation line break analyses are performed for breaks in both the discharge and suction side of the recirculation pump. Two break types (geometries) are considered for the recirculation line break. The two types are the double-ended guillotine (DEG) break and the split break. For a DEG break, the piping is assumed to be completely severed resulting in two independent flow paths to the containment. A split type break is assumed to be a longitudinal opening or hole in the piping that results in a single break flow path to the containment. The break spectrum analyses in the intermediate and small break region consider break sizes between 1.0 square feet (ft2) and 0.05 ft2* Break sizes and single failures are analyzed for both suction and discharge recirculation line breaks.
In addition to the recirculation line breaks, breaks in other reactor coolant system piping are also considered in the LOCA break spectrum analysis. The non-recirculation line breaks considered in this analysis are (1) main steam line break inside the containment, (2) feedwater line breaks, (3) high-pressure coolant injection line breaks, (4) low pressure core spray line breaks, (5) low-pressure coolant injection line breaks, (6) reactor core isolation cooling line breaks, (7) reactor water cleanup line breaks, and (8) instrument line breaks.
Limiting Break Analysis Results for Recirculation Breaks The licensee has performed LOCA analyses for breaks in both of these locations with consideration for both DEG and split break geometries. The break sizes considered included DEG breaks with discharge coefficients from 1.0 to 0.4 and split breaks with areas ranging between the full pipe area to 0.05 ft2. The analyses demonstrate that the limiting (highest PCT) recirculation line break is the 0.20 ft2 split break in the pump discharge piping for single failure of battery (DC) power, board A (SF-BATTIBA) and a top-peaked axial power shape. The limiting PCT is 1909°F, maximum local cladding oxidation is 1.2-percent, and the maximum planar average MWR is 0.69-percent (Reference 24).
The limiting PCT for single failure of battery (DC) power, board B (SF-BATTIBB) is bound by the limiting PCT for SF-BATTIBA. For SF-BATTIBB, the highest PCT occurred for a recirculation split line break of 0.5 ft2 in the pump discharge piping and a top-peaked axial power shape. The PCT is 1658°F, maximum local cladding oxidation is 0.36-percent, and the maximum planar average MWR is 0.18-percent.
OFFICIAl USE ONlY PROPRIETARY INFORMATION
OFFICIAl USE ONlY PROPRIETARY INFORMATION The maximum PCT calculated for a recirculation line break occurs in the pump discharge piping.
The maximum PCT calculated for a recirculation line break occurs for a 0.20 fe split break for SF-BATTIBA.
Limiting Break Analysis Results for Non-Recirculation Breaks The licensee has performed LOCA analyses for breaks in the feedwater and low pressure core spray (LPCS) lines. Breaks in other non-recirculation lines are less limiting than for recirculation breaks since they do not necessarily result in the most severe challenge to event acceptance criteria. The analysis indicates that the limiting non-recirculation line break is the 0.4 fe DEG break in the LPCS line with SF-BATTIBA and a top-peaked axial power shape. The PCT for the limiting ECCS line break is 1511 oF and maximum local cladding oxidation is 0.15-percent. The feedwater line break was confirmed to be nonlimiting. Also the LPCS line break was found to be less limiting.
The key event times for the limiting non-recirculation break are provided in Table 7.1. Table 7.2 presents PCT results for the non-recirculation line breaks. The 1511 oF maximum PCT for non-recirculation line breaks is lower than the maximum PCT for recirculation line breaks of 1909°F. Therefore, non-recirculation line breaks are nonlimiting (References 24 and 25).
Summary The LOCA break spectrum analysis results (i.e., the limiting break characteristics) presented in this section are applicable for a full core of XM fuel as well as for XM fuel in transition cores. Since the thermal-hydraulic characteristics of the XM fuel and coresident fuel assemblies are similar, the overall core response during a LOCA will not be significantly different for transition cores of XM fuel and coresident fuel.
The limiting LOCA break characteristics are:
Location:
Type/Size:
Single failure:
Axial power shape:
Initial state:
recirculation discharge pipe split/0.20 fe battery (DC) power, board A top-peaked
((
))
The break spectrum analysis was performed using the NRC-approved AREVA EXEM BWR-2000 LOCA methodology. All SER restrictions and ranges of applicability for the EXEM BWR-2000 methodology were reviewed prior to final documentation of the LOCA analysis to ensure compliance with NRC requirements and methodology limitations. The NRC staff finds the LOCA limiting break spectrum analysis methodology and results acceptable.
3.8.2 BFN Units LOCA-ECCS Analysis MAPLHGR Limits for XM Fuel Introduction This section documents the LOCA-ECCS analyses for BFN Units 1, 2, and 3 to specify the MAPLHGR limit versus exposure for XM fuel and to demonstrate that the MAPLHGR limit is adequate to ensure that the LOCA-ECCS criteria in 10 CFR 50.46 are satisfied for operation at or OFFICIAl USE ONlY PROPRIETARY INFORMATION
OFFICIAl USE ONlY PROPRIETARY INFORMATION below the limit. The analysis also establishes the licensing basis PCT and corresponding local cladding oxidation from the MWR for XM fuel used at BFN Units 1, 2, and 3.
The break spectrum analysis is documented in Section 3.8.1 of this SE. The break spectrum analysis included the determination of break size, type, location, axial power shape and ECCS single failure. The limiting LOCA characteristics are listed in the summary Section 3.8.1 of this SE.
The LOCA break spectrum analysis documented in Reference 24 and in Section 3.8.1 is based on a generic XM fuel neutronics design at beginning of life conditions. The heatup analyses were performed for limiting neutronics design and fluid conditions from the limiting LOCA break spectrum analysis.
LOCA Description and Analysis Regardless of the initiating break characteristics, the event response is separated into three phases: the blowdown phase, the refill phase, and the reflood phase. The relative duration of each phase is strongly dependent upon the break size and location. The last two phases are often combined and will be discussed together in this report. During the blowdown phase of a LOCA, there is a net loss-of-coolant inventory, an increase in fuel cladding temperature due to core flow degradation, and for the larger breaks, the core becomes fully or partially uncovered.
There is a rapid decrease in pressure during the blowdown phase.
In the refill phase of a LOCA, the ECCS is functioning and there is a net increase of coolant inventory. During this phase the core sprays provide core cooling and, along with low-pressure and high-pressure coolant injection, supply liquid to refill the lower portion of the reactor vessel. In general, the core heat transfer to the coolant is less than the fuel decay heat rate and the fuel cladding temperature continues to increase during the refill phase.
In the reflood phase, the coolant inventory has increased to the point where the mixture level reenters the core region. During the core reflood phase, cooling is provided above the mixture level by entrained reflood liquid and below the mixture level by pool boiling. Sufficient coolant eventually reaches the core hot node and the fuel cladding temperature decreases.
The Code of Federal Regulations prescribes specific acceptance criteria (1 0 CFR 50.46) for ECCS in the event of a LOCA as well as specific requirements and acceptable features for Evaluation Models (1 0 CFR Part 50 Appendix K). These acceptance criteria are listed in Section 3.8.1 of this SE.
The LOCA evaluation model based on EXEM BWR-2000 (Reference 26) methodology employs three major computer codes to evaluate the system and fuel response during all phases of a LOCA. These are the RELAX, HUXY, and RODEX2 computer codes. RELAX is used to calculate the system and hot channel response during the blowdown, refill, and reflood phases of the LOCA. The HUXY code is used to perform heatup calculations for the entire LOCA, and calculates the PCT and local clad oxidation at the axial plane of interest. RODEX2 is used to determine fuel parameters (such as stored energy) for input to the other LOCA codes. The principal results of a HUXY heatup analysis are the PCT and the-percent local oxidation of the fuel cladding, often called the percent maximum local metal water reactor.
((
OFFICIAl USE ONlY PROPRIETARY INFORMATION
OffiCIAl USE ONlY PROPRIETARY INFORMATION ))
MAPLHGR Analysis Results An exposure-dependent MAPLHGR limit for ATRIUM 10XM fuel is obtained by performing HUXY heatup analyses using results from the limiting LOCA analysis case identified in Reference 24 and Section 3.8.1 of this SE. The response of the reactor system is shown in Figures 5.1 through 5.18 of Reference 25. In the MAPLHGR analysis, the XM fuel rod stored energy is set to be bounding at all exposures and the RELAX hot channel peak power node is modeled at the highest MAPLHGR, which is 102 percent of 13.0 kW/ft. for the XM fuel. Table 5.2 of Reference 25 lists the results from XM MAPLHGR analysis results. The HUXY analysis is performed at 5 gigawatt-days per metric ton of uranium (GWd/MTU) exposure intervals for exposures between 0 and 65 GWd/MTU and an ending exposure of 67 GWd/MTU.
The XM limiting PCT is 1903°F at the 0.0 GWd/MTU exposure. The corresponding maximum local cladding oxidation at the PCT limiting exposure is 1.16 percent. Analysis results show the core average metal-water reaction is less than 1.0-percent total hydrogen generated (Reference 25).
Summary The MAPLHGR limit was determined by applying the EXEM BWR-2000 Evaluation Model for the analysis of the limiting LOCA event. The fuel assembly hydraulic response during a LOCA is primarily dependent on assembly geometry, hydraulic characteristics, and initial power. The characteristics for a typical XM neutronic design are used in the break spectrum analysis fuel heatup analyses to determine the fluid conditions that occur in a fuel assembly during the limiting LOCA. Thermal-hydraulic characteristics of XM and coresident fuel assemblies are similar. The limiting break characteristics, and ultimately the calculated PCT, are not significantly different during the transition to a full core of XM fuel. Therefore, the XM MAPLHGR limits documented in this SE are applicable for a full core of ATRIUM IOXM as well as transition cores.
The licensee performed the break spectrum analysis and LOCA analysis using the NRC-approved AREVA EXEM BWR-2000 LOCA methodology. All SER restrictions and ranges of applicability for the EXEM BWR-2000 methodology were reviewed prior to final documentation of the LOCA analysis to ensure compliance with NRC requirements and methodology limitations.
The NRC staff finds the LOCA limiting analysis methodology and results acceptable.
3.9 BFN Units Reload Safety Analyses Introduction Reload licensing analyses in support of BFN's fuel transition were performed for potentially limiting events using NRC-approved methodologies and computer codes (Reference 21 ).
Results of the analyses are used to establish the TSs/COLR limits and ensure design and licensing criteria are met. Design and safety analyses are based on both operational assumptions and plant parameters provided by the licensee. The results of the reload licensing OffiCIAl USE ONlY PROPRIETARY INFORMATION
OFFICIAl USE ONlY PROPRIETARY INFORMATION analysis support operation for the power/flow map presented in Figure 1.1 of Reference 21 and also support operation with the equipment out-of-service (EOOS) scenarios presented in Table 1.1 of Reference 21.
The core consists of a total of 764 fuel assemblies, including 272 fresh XM fuel assemblies and 492 irradiated ATRIUM-1 0 assemblies. Licensing analyses support the core design presented in 0 to Reference 1.
3.9.1 Disposition of Events The licensee performed a disposition of events analysis to identify the limiting events that must be analyzed to support operation at the BFN with the introduction of XM fuel. Events and analyses identified as potentially limiting are either evaluated generically for the introduction of AREVA fuel or on a cycle-specific basis. The disposition process involved identification of the licensing basis of the plant and fuel related system design criteria compliance for regulatory and safe operation of the plant.
A disposition event summary is presented in Table 2.1 of Reference 21 and the summary presents a list of events and analyses, the corresponding final safety analysis report (FSAR) section, the disposition status, and comments. The disposition is categorized as (1) FSAR analysis, (2) generic analysis, (3) plant specific analysis, and (4) cycle specific analysis. The NRC staff has reviewed the Table 2.1 and has accepted the validity of the disposition analysis, and verified that NRC-approved methodology and code are used in the analyses.
3.9.2 Core Hydrodynamic Stability The licensee has implemented the BWR Owners Group Long Term Stability Solution Option Ill (Oscillation Power Range Monitor (OPRM)) and has performed reload validation in accordance with the methodology in Reference 30. The stability based Operating Limit MCPR (OLMCPR) is provided for two conditions as a function of OPRM amplitude setpoint in Table 4.3 of Reference
- 21. The two conditions evaluated are for a postulated oscillation at 45-percent core flow steady state operation and following a two recirculation pump trip (RPT) from the limiting full power operation state point. Power-and flow-dependent limits provide adequate protection against violation of the SLMCPR for postulated reactor instability as long as the operating limit is greater than or equal to the specified value for the selected OPRM setpoint. Setpoints supporting EOOS operating conditions are provided in Table 4.3 of Reference 21.
AREVA performed delta-over-initial CPR versus oscillation magnitude (DIVOM) calculations to obtain relative change in CPR as a function of the calculated hot channel oscillation magnitude (HCOM). These calculation were performed using the NRC approved methodology presented in Reference 31. RAMONAS-FA employs a coupled neutronic-thermal-hydraulic 3-dimensional transient model for the purpose of determining the relationship between the relative change in flCPR and the HCOM on a plant specific basis. The stability-based OLMCPRs were calculated using the most limiting calculated change in relative flCPR for a given oscillation magnitude.
In cases where the OPRM system is declared inoperable, backup stability protection (BSP) is provided consistent with Reference 32. BSP curves have been evaluated using an approved methodology (Reference 32) to determine endpoints meeting decay ratio criteria for the BSP Base Minimal Region I (scram region) and Base Minimal Region II (controlled entry region).
Stability boundaries based on these endpoints can then be determined using the generic shape OFFICIAl USE ONlY PROPRIETARY INFORMATION
OFFICIAL USE ONLY PROPRIETARY INFORMATION generating function from Reference 32. Endpoints for the SSP regions provided in Table 4.4 of Reference 21 have global decay ratios less than or equal 0.85, and regional and channel decay ratios less than or equal 0.80.
Based on the information provided by the licensee and discussed above, the NRC staff finds that the stability analysis and evaluation performed in support of the LAR provides reasonable assurance that the proposed transition to XM fuel and methods will not adversely impact BFN ability to satisfy G DC 1 0 and 12.
3.9.3 Anticipated Operational Occurrences This section summarizes the analyses performed to determine the power-and flow dependent MCPR operating limits for base case operation. COTRANSA2 (Reference 33), XCOBRA-T (Reference 34), XCOBRA (Reference 35), and CASM0-4/MICROBURN-82 (Reference 19) are the major codes used in the thermal limits analyses as described in the AREVA THERMEX methodology report (Reference 35) and neutronics methodology report (Reference 19).
COTRANSA2 is a system transient simulation code, which includes an axial 1-dimensional neutronics model that captures the effects of axial power shifts associated with the system transients. XCOBRA-T is a transient thermal-hydraulics code used in the analysis of thermal margins for the limiting fuel assembly. X COBRA is used in steady-state analyses. The ACE/XM critical power correlation (References 5, 4, and 23) is used to evaluate the thermal margin for the XM fuel.
At BFN, direct scram on turbine stop valve (TSV) position and turbine control valve (TCV) fast closure are bypassed at power levels less than 30-percent of rated (Pbypass). Scram will occur when the high pressure or high neutron flux scram setpoint is reached.
Reductions in feedwater temperature of less than 10 oF from the nominal feedwater temperature and variation of plus or minus 1 0 psi in dome pressure are considered base case operation, not an EOOS condition. Analyses were performed to determine the limiting conditions in the allowable ranges. FFTR is used to extend rated power operation by decreasing the feedwater temperature. The amount of feedwater temperature reduction is a function of power with the maximum decrease of 65 oF (55 oF plus 10 oF bias) at rated power. Analyses were performed to support combined FFTR/Coastdown operation to a core average exposure of 34,147.6 MWd/MTU. The analyses were performed with the limiting feedwater and dome pressure conditions in the allowable ranges.
Load Rejection No Bypass (LRNB)
The load rejection causes a fast closure of the turbine control valves. The resulting compression wave in the steam lines into the vessel creates a rapid pressurization. Pressurization causes a decrease in voids and a rapid increase in power. The turbine control valve closure causes a reactor scram. LRNB analyses were performed for a range of power/flow conditions to support generation of the thermal limits. Base case limiting LRNB transient analysis results used to generate the near end of cycle (NEOC) and end of cycle licensing basis (EOCLB) operating limits, for both TS scram speed (TSSS) and nominal scram speed (NSS) insertion times, are shown in Table 5.3 of Reference 21. Responses of various reactor and plant parameters during the LRNB event initiated at 1 00-percent of rated power and 1 05-percent of rated core flow with TSSS insertion times are shown in Figures 5.1-5.3 of Reference 21.
OFFICIAL USE ONLY PROPRIETARY INFORMATION
OffiCIAl USE ONlY PROPRIETARY INfORMATION Turbine Trip No Bypass (TTNB)
The turbine trip causes a closure of the turbine stop valves. The resulting compression wave travels through the steam lines into the vessel and creates a rapid pressurization. The increase in pressure causes a decrease in core voids, which in turn causes a rapid increase in power. The closure of the turbine stop valves (TSV) also causes a reactor scram. The excursion of the core power due to the void collapse is terminated primarily by the reactor scram and revoiding of the core. In addition to closing the TSV, a signal is also sent to close the TCV in fast mode. The consequences of a fast closure of the TCV are very similar to those resulting from a TSV closure.
Analyses were performed demonstrating that the TTNB event is equivalent to or bound by the LRNB event; therefore, the thermal limits established for the LRNB will also protect against the TTNB event (Reference 21 ).
Feedwater Controller Failure (FWCF)
The increase in feedwater flow due to a failure of the feedwater control system results in an increase in the water level and a decrease in the coolant temperature at the core inlet. The increase in core inlet subcooling causes an increase in core power. The water level continues to rise and eventually reaches the high water level trip point. The high water level trip causes the turbine stop valves to close in order to prevent damage to the turbine from excessive liquid inventory in the steam line. The valve closures create a compression wave that travels to the core causing a void collapse and subsequent rapid power excursion. The closure of the turbine stop valves also initiates a reactor scram and an RPT. In addition to the turbine stop valve closure, the turbine control valves also close in the fast closure mode. Because of the partially closed initial position of the control valves, they will typically close faster than the stop valves and control the pressurization portion of the event. However, TCV closure characteristics are nonlinear so that the resulting core pressurization and b.CPR results may not always bound those of the slower TSV closure at rated power (steam flow increases above rated before fast TCV closure). FWCF analyses were performed for a range of power/flow conditions to support generation of the thermal limits. Table 5.5 of Reference 21 presents the base case limiting FWCF transient analysis results used to generate the NEOC and EOCLB operating limits for both TSSS and NSS insertion times.
Loss of Feedwater Heating (LFWH)
The LFWH event analysis supports an assumed 100° F decrease in the feedwater temperature.
The result is an increase in core inlet subcooling, which reduces voids, thereby increasing core power and shifting the axial power distribution toward the bottom of the core. The axial power shift and increase in core power causes the voids to build up in the bottom of the core, acting as negative feedback to the increased subcooling effect. The negative feedback moderates the core power increase. The increase in core thermal power event does not result in a corresponding increase in steam flow because some of the added power is used to overcome the increase in inlet subcooling. The increase in steam flow is accommodated by the pressure control system via the TCVs or the turbine bypass valves, so no pressurization occurs. The licensee performed a cycle-specific analysis in accordance with Reference 36 methodology to determine the change in MCPR for the event. The results are presented in Table 5.6 of Reference 21.
OffiCIAl USE ONlY PROPRIETARY INfORMATION
OFFICIAl USE ONlY PROPRIETARY INFORMATION Control Rod Withdrawal Error (CRWE)
The CRWE transient is an inadvertent reactor operator initiated withdrawal of a control rod. This withdrawal increases local power and core thermal power, lowering the core MCPR. The CRWE is typically terminated by control rod blocks initiated by the rod block monitor (RBM). The CRWE event was analyzed assuming no xenon and allowing credible instrumentation OOS in the RBM system. The analysis further assumes that the plant could be operating in either an A orB sequence control rod pattern. Analysis results indicate standard filtered RBM setpoint reductions are supported. The licensee's analyses demonstrated that the 1-percent strain and centerline melt criteria are met for both XM and ATRIUM-1 0 fuel, for the LHGR limits and their associated multipliers presented in Section 8.2 of Reference 21.
Slow Flow Runup Analysis The slow flow excursion event assumes recirculation flow control system failure such that core flow increases slowly to the maximum flow physically attainable by the equipment (107-percent of rated core flow). An uncontrolled increase in flow creates the potential for a significant increase in core power and heat flux. A conservatively steep flow runup path was used in the analysis.
Analyses were performed to support operation in all the EOOS scenarios. XCOBRA is used to calculate the change in critical power ratio during a two-loop flow run up to the maximum flow rate.
The MCPRt limit is set so an increase in core power, resulting from the maximum increase in core flow, assures the TLO safety limit MCPR is not violated. The licensee performed calculations over a range of initial flow rates to determine the corresponding MCPR values causing the limiting assembly to be at the safety limit MCPR for the high flow condition at the end of the flow excursion.
The licensee performed flow runup analyses with CASM0-4/MICROBURN-82 to determine flow-dependent LHGR multipliers (LHGRFAC1) for XM and ATRIUM-1 0 fuel. Analysis results are presented in Table 5.9 of Reference 21. MCPRt limits providing the required protection are presented in Table 8.3. MCPRt limits are applicable for all exposures (Reference 21 ).
Equipment OOS (EOOS) Scenarios The following Table lists of EOOS scenarios that were supported for the BFN units operation.
Turbine bypass valves OOS (TBVOOS)
Power load unbalance OOS (PLUOOS)
Combined TBVOOS and FHOOS Combined TBVOOS and PLUOOS Combined FHOOS and PLUOOS Combined TBVOOS, FHOOS, and PLUOOS Single-loop operation (SLO)
Each of the above EOOS scenarios include EOC-RPT-OOS.
The effect of operation with TBVOOS is a reduction in the system pressure relief capacity, which makes the pressurization events more severe. While the base case LRNB and TTNB events are analyzed assuming the turbine bypass valves OOS, operation with TBVOOS has an adverse OFFICIAl USE ONlY PROPRIETARY INFORMATION
OffiCIAl USE ONlY PROPRIETARY INFORMATION effect on the FWCF event. Analyses of the FWCF event with TBVOOS were performed to establish the TBVOOS operating limits.
The FHOOS scenario assumes a feedwater temperature reduction of 65°F (55°F plus 1 oaF bias) at rated power and steam flow. The effect of reduced feedwater temperature is an increase in core inlet subcooling, changing axial power shape and core void fraction. Additionally, steam flow for a given power level decreases because more power is required to increase coolant enthalpy to saturated conditions. FWCF events with FHOOS conditions are generally worse due to a larger change in inlet subcooling and core power prior to the pressurization phase of the event.
The PLU device in normal operation is assumed to not function below 50-percent power.
PLUOOS is assumed to mean the PLU device does not function for any power level, and does not initiate fast TCV closure. Analyses were performed for LRNB events assuming PLUOOS.
The licensee performed FWCF analyses with both TBVOOS and FHOOS. Operating limits for this combined EOOS scenario were established using these FWCF results and results previously discussed.
Limits were established to support operation with both TBVOOS and PLUOOS. No additional analyses are required to construct MCPRP operating limits for TBVOOS and PLUOOS since TBVOOS and PLUOOS are independent EOOS conditions (TBVOOS only impacts FWCF events; PLUOOS only impacts LRNB events).
The licensee performed LRNB analyses with both FHOOS and PLUOOS. Operating limits for this combined EOOS scenario were established using these LRNB results and results previously discussed.
Limits were established to support operation with TBVOOS, FHOOS, and PLUOOS. No additional analyses are required to construct MCPRp operating limits for TBVOOS, FHOOS, and PLUOOS since TBVOOS and PLUOOS are independent EOOS conditions (TBVOOS only impacts FWCF events; PLUOOS only impacts LRNB events).
The NRC staff has reviewed the analyses that support the operation with the EOOS presented in Reference 21 and finds them acceptable.
3.1 0 Special Analyses 3.1 0.1 ASME Overpressurization Analysis The licensee performed a maximum overpressurization analysis to demonstrate compliance with the ASME B&PV Code. The analysis shows that the safety/relief valves have sufficient capacity and performance to prevent the reactor vessel pressure from reaching the safety limit of 110 percent of the design pressure.
The licensee performed Main Steam Isolation Valve (MSIV) closure, TSV closure, and TCV closure (without bypass) analyses with the AREVA plant simulator code COTRANSA2 (Reference 33) for 1 02-percent power and both 81-percent and 1 05-percent flow at the highest cycle exposure. The assumptions that were made for analysis are (1) the most critical active component was assumed to fail, (2) to support one main steam relief valve OOS (lowest MSRV),
OffiCIAl USE ONlY PROPRIETARY INFORMATION
OFFICIAl USE ONlY PROPRIETARY INFORMATION {3) using TSSS insertion times, (4) the initial dome pressure set to maximum, and (5) the anticipated transient without scram (A lWS)-RPT assumed.
Results of all the overpressurization analyses are listed in Table 7.1 and Figures 7.1 through 7.4 of Reference 21. The maximum pressure of 1336 pounds per square inch gauge (psig) occurs in the lower plenum. The maximum dome pressure for the same event is 1301 psig. Results demonstrate the maximum vessel pressure limit of 1375 psig and dome pressure limit of 1325 psig are not exceeded for any analyses.
The NRC staff has reviewed these analyses for maximum overpressurization and MSIV, TSV, and TCV closure presented in Reference 21 and finds it acceptable, since the results do not exceed the limited values.
3.1 0.2 A lWS Overpressurization Analysis This AlWS overpressurization analysis was performed to demonstrate that the peak vessel pressure for the limiting A lWS event is less than the ASME Service Level C limit of 120 percent of the design pressure (1500 psig). Overpressurization analyses were performed at 1 00-percent power at both 81-percent and 1 05-percent flow over the cycle exposure range for both the MSIV closure event and the pressure regulator failure open (PRFO) events. The PRFO event assumes a step decrease in pressure demand such that the pressure control system opens the turbine control and turbine bypass valves. The assumptions made in the analysis are listed in Section 7.2 of Reference 21.
Analyses results are presented in Table 7.2 of Reference 21. The response of various reactor plant parameters during the limiting PRFO event are shown in Figures 7.5-7.8 (Reference 21 ).
The maximum lower plenum pressure is 1399 psig and the maximum dome pressure is 1379 psig.
The results demonstrate that the AlWS maximum vessel pressure limit of 1500 psig is not exceeded.
The NRC staff has reviewed the AlWS overpressurization analysis presented in Reference 21 and finds it acceptable, since the results do not exceed the limited values.
3.11 Summary and Conclusion The NRC staff has reviewed the LAR (Reference 1 ), in conjunction with the supplemental information (References 2), the responses to the staff's requests for additional information to evaluate the acceptability of the BFN transition to XM fuel with AREVA fuel performance assessment, safety analysis and core design methodologies. Based on its review, the NRC staff has determined that the licensee provided adequate technical basis to support the proposed TSs changes. Specifically, the NRC staff finds the licensee has demonstrated that (1) BFN complies with the staff limitations and conditions imposed for application of the topical reports, (2) AREVA codes and methods are applicable for BFN, {3) the BFN specific safety analysis results based on the AREVA methodology meet the applicable licensing criteria, and (4) the proposed TSs changes are acceptable.
3.12 Impact of Using XM Fuel on the BFN Design Basis Dose Consequence Analyses The NRC staff reviewed the impact of using XM fuel on the licensee's design basis dose consequence analyses to ensure that the use of XM fuel will not result in an increase in the dose OFFICIAl USE ONlY PROPRIETARY INFORMATION
OFFICIAl USE ONlY PROPRIETARY INFORMATION consequences and that the resulting calculated doses will remain within the design criteria specified in 10 CFR 50.67, "Accident source term," and the accident specific design criteria outlined in Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors." Specifically, the NRC staff reviewed the impact of the use of XM fuel on all events currently analyzed in the BFN Updated Final Safety Analysis Report (UFSAR) that could have the potential for significant dose consequences, namely; the control rod drop accident, the LOCA, the refueling accident and the main steam line break.
For the control rod drop accident, the licensee's evaluation has determined that the number of XM fuel rods calculated to fail is well below the 850 rods assumed to fail in the UFSAR analysis. Since the power level is unchanged by this LAR and the number of XM rods calculated to fail is lower, the inventory released during this accident and the dose consequences will be bounded by the UFSAR analysis.
The LOCA analysis performed to show compliance with the design criteria of 10 CFR 50.67 uses a deterministic source term as defined in RG 1.183. Since the percentages of core inventory released from the fuel is specified in guidance, the amount of activity assumed to be released is only a function of power level. Since the power level is unchanged by this LAR, the inventory released and the dose consequences will be unchanged from that stated in the UFSAR analysis.
For the refueling accident, the licensee has determined that the number of XM fuel rods assumed to fail exceeds the number of rods assumed to fail in the current licensing basis (CLB) UFSAR evaluation. As described in the LAR, the XM fuel contains a higher number of fuel rods than the fuel analyzed in the CLB dose analysis. As a result, the activity produced during power operations is distributed over a significantly larger number of XM fuel rods. Therefore, the amount of activity per XM fuel rod is significantly less than for the fuel rods analyzed in the CLB analysis.
Although a larger number of XM fuel rods are assumed to fail, the lower inventory per rod more than makes up the difference resulting in a lower calculated release. The NRC staff has confirmed the licensee's determination that the amount of activity assumed to be released and the resulting dose consequences for a refueling accident with XM fuel will be bounded by the CLB analysis in the UFSAR.
The licensee's evaluation of the main steam line break accident has shown that fuel damage is not expected to occur. The radionuclide inventory for this accident is based on TS coolant activity limits and is therefore unaffected by fuel design characteristics. The licensee concluded that the use of XM fuel will have no impact on the main steam line break dose consequence analysis.
The NRC staff evaluated the impact of the using XM fuel design fuel on the licensee's design basis dose consequence analyses and has determined that the use of XM fuel will not result in an increase in the dose consequences and that the resulting calculated doses will remain within the design criteria specified in 10 CFR 50.67, and the accident specific design criteria outlined in RG 1.183. Therefore, the NRC staff finds that the use of XM fuel at BFN is acceptable from a dose consequence perspective.
4.0 LICENSE CONDITION During review of Brunswick Steam Electric Plant (BSEP) XM fuel transition LAR, the NRC staff determined that the predictive model for channel bow was validated against empirical data that was not bounding by BSEP's expected performance. To resolve this issue, the licensee for BSEP OFFICIAl USE ONlY PROPRIETARY INFORMATION
OffiCIAl USE ONlY PROPRIETARY INFORMATION agreed to increase the channel bow uncertainty in the SLMCPR calculation for the most severely deflected fuel channels. In view of the excessive channel bow that occurred at BSEP, the licensee proposed a license condition for BSEP Units 1 and 2 in connection with the use of AREVA channel bow model outside the range of the channel bow measurement database from which its uncertainty was quantified (
Reference:
Letter, BSEP 13-0002, from Michael J.
Annacone (Duke Energy) to NRC, "Supplement to License Amendment Request for Addition of Analytical Methodology Topical Report to Technical Specification 5.6.5, CORE OPERATING LIMITS REPORT (COLR), and Revision to Technical Specification 2.1.1.2 Minimum Critical Power Ratio Safety Limit," January 22, 2013).
TVA, in its response (Reference 2} to the NRC staff SNPB RAI-6(b) stated that, given that the AREVA channel bow model used for the BSEP will also be applied to the BFN, it is possible that the calculated channel fluence gradients could exceed the bounds of the channel bow database for a limited number of channels in a given cycle. Therefore, the license condition provided for the BSEP units is also applicable to the BFN units. The TVA's proposed license conditions for BFN Units 1, 2, and 3, which are included in Enclosure 4 of Reference 2, are listed below:
BFN Unit 1 License Condition The fuel channel bow standard deviation component of the channel bow model uncertainty used by ANP-10307PA, "AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, Revision 0" (i.e., TS 5.6.5.b.11}, to determine the SLMCPR shall be increased by the ratio of channel fluence gradient to the nearest channel fluence gradient bound of the channel measurement database, when applied to channels with fluence gradients outside the bounds of the measurement database from which the model uncertainty is determined.
This license condition will be effective upon the implementation of Amendment No. 285.
BFN Unit 2 License Condition The fuel channel bow standard deviation component of the channel bow model uncertainty used by ANP-10307PA, "AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, Revision 0" (i.e., TS 5.6.5.b.1 0), to determine the SLMCPR shall be increased by the ratio of channel fluence gradient to the nearest channel fluence gradient bound of the channel measurement database, when applied to channels with fluence gradients outside the bounds of the measurement database from which the model uncertainty is determined.
This license condition will be effective upon the implementation of Amendment No. 311.
BFN Unit 3 License Condition The fuel channel bow standard deviation component of the channel bow model uncertainty used by ANP-10307PA, "AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, Revision 0," (i.e., TS 5.6.5.b.1 0) to determine the SLMCPR shall be increased by the ratio of channel fluence gradient to the nearest channel fluence gradient bound of the channel measurement database, when applied to channels with fluence gradients outside the bounds of the measurement database from which the model uncertainty is determined.
This license condition will be effective upon the implementation of Amendment No. 270.
OFFICIAl USE ONlY PROPRIETARY INFORMATION
OFFICIAl USE ONlY PROPRIETARY INFORM.UION 5.0 REGULATORY COMMITMENTS During the first reload application of RODEX4 the initial approved limit of cladding external oxidation was challenged by NRC staff because of a concern about the effect of spallation on the cladding integrity. To avoid the issue of spallation, the limit was reduced to ((
)). The
((
)) limit was established from a review of historical liftoff measurement data on AREVA BWR fuel. As a result, the licensee made a regulatory commitment to reduce the oxidation limit (see Section 3.2.2 of this SE). The licensee, in Attachment 1 of the enclosure to the LAR dated February 28, 2013, provided the following regulatory commitment to be completed upon implementation of the amendments.
Commitment Completion Date When using AREVA Topical Report BAW-10247PA, Upon implementation of "Realistic Thermal Mechanical Fuel Rod Methodology for Units 1, 2, and 3 license Boiling Water Reactors," Revision 0, February 2008, to amendments authorizing the determine core operating limits, the fuel cladding peak oxide incorporation of AREVA thickness calculated by RODEX4 will be limited to less than Topical Report BAW-10247PA the proprietary value defined in Section 3.2.7 of AREVA into Technical Specification report ANP-3159P, Revision 0, dated October 2012.
5.6.5.b.
Therefore, the NRC staff finds the above regulatory commitment acceptable, since it assures that the fuel cladding peak oxide thickness will be limited the values specific to BFN Units 1, 2 and 3.
6.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Alabama State official was notified of the proposed issuance of the amendment. The State official had no comments.
7.0 ENVIRONMENTAL CONSIDERATION
The amendments change a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding {78 FR 49301 ). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
OFFICIAl USE ONlY PROPRIETARY INFORMATION
OFFICIAl USE ONlY PROPRIETARY INFORMATION
8.0 CONCLUSION
The NRC has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the NRC's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
9.0 REFERENCES
- 1. Letter from J. W. Shea, TVA to NRC, "Technical Specification Change TS-478-Addition of Analytical Methodologies to TS 5.6.5 for Browns Ferry 1, 2, and 3, and Revision of TS 2.1.1.2 for Browns Ferry Unit 2 in Support of ATRIUM 10XM Fuel Use at Browns Ferry," dated February 28, 2013, (Agencywide Documents Access and Management System (ADAMS)
Accession No. ML13070A307).
- 2. Letter from J. W. Shea, TVA to NRC, "Response to Request for Additional Information on Technical Specification Change TS-478," dated September 30, 2013 (ADAMS Accession No. ML13276A063). (Enclosure 1, ANP-3248P, Revision 1, "AREVA NP Responses for Browns Ferry ATRIUM 10XM Fuel Transition," AREVA NP, September 2013; Enclosure 2, "TVA Response to Request for Additional Information," September 2013.)
- 3. BAW-10247PA Revision 0, "Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors," AREVA NP, February 2008.
- 4. ANP-3140P Revision 0, "Browns Ferry Units 1, 2, and 3 Improved K-factor Model for ACE/ATRIUM 10XM Critical Power Correlation, AREVA NP Inc., August 2012.
- 5. ANP-10298PA Revision 0, "ACE/ATRIUM 10XM Critical Power Correlation," AREVA NP Inc.,
March 2010.
- 6. Engineering Information Record 51-9191258-001, "Browns Ferry Unit 2 Cycle 19 MCPR Safety Limit Analysis with SAFLIM3D Methodology, AREVA NP Inc., October 2012.
- 7. ANF-89-98PA Revision 1 and Supplement 1, "Generic Mechanical Design Criteria for BWR Designs," Advanced Nuclear Fuels Corporation, May 1995.
- 8. NUREG-0800, "Standard Review Plan for Review of Safety Analysis Reports for Nuclear Power Plants," NRC, March 2007.
- 9. ASME Boiler and Pressure Vessel Code, Section Ill, Division 1, American Society of Mechanical Engineers.
- 10. W. J. O'Donnell and B. F. Langer, "Fatigue Design Basis for Zircaloy Components," Nuclear Science and Engineering, Volume 20, January 1964.
- 11. EMF-93-177PA Revision 1, "Mechanical Design for BWR Fuel Channels," Framatome ANP Inc., August 2005.
OFFICIAl USE ONlY PROPRIETARY INFORMATION
OFFICIAl USE ONlY PROPRIETARY INFORMATION 12. XN-NF-81-51 PA, "LOCA-Seismic Structural Response of an Exxon Nuclear Company BWR Jet Pump Fuel Assembly," Exxon Nuclear Company, May 1986.
- 13. XN-NF-84-97PA, "LOCA - Seismic Structural Response of an ENC 9x9 BWR Jet Pump Fuel Assembly," Exxon Nuclear Company, August 1986.
- 14. Letter from Farideh E. Saba, NRC, to Michael J. Annacone, CP&L, "Brunswick Steam Electric Plant, Units 1 and 2 -Issuance of Amendments Regarding Addition of Analytical Methodology Topical Report to Technical Specification 5.6.5)," Agencywide Documents Access and Management System (ADAMS) Accession No. ML11101A043, dated AprilS, 2011.
- 15. XN-NF-80-19PA Volume 4 Revision 1, "Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads," Exxon Nuclear Company, June 1986.
- 16. EMF-2209PA Revision 3, "SPCB Critical Power Correlation," AREVA, September 2009.
- 17. EMF-95-52P, "Fuel Design Evaluation for Siemens Power Corporation ATRIUM-10 BWR Reload Fuel," Siemens Power Corporation, December 1968.
- 18. XN-NF-75-32PA Supplements 1 through 4, "Computational Procedure for Evaluating Fuel Rod Bowing," Exxon Nuclear Company, October 1983.
- 19. EMF-2158PA Revision 0, "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASM0-4/MICROBURN-B2," Siemens Power Corporation, October 1999.
- 20. ANP-2637 (Attachment 16 to Reference 1 ), Revision 4, "Boiling Water Reactor Licensing Methodology Compendium," AREVA NP Inc., November 2012.
- 21. ANP-3167P (Attachment 12 of Reference 1 ), Revision 0, "Browns Ferry Unit 2 Cycle 19 Reload Analysis," AREVA NP, Inc., November 2012.
- 22. ANP-1 0307PA, Revision 0, "AREVA MCPR Safety Limit Methodology for Boiling Water Reactors," AREVA NP, June 2011.
- 23. ANP-10298PA, Supplement 1P, Revision 0, "Improved K-factor Model for ACE/ATRIUM 10XM Critical Power Correlation," AREVA NP, December 2011. Final SEdated March 31, 2014 (ADAMS Accession No. ML14072A353), as acknowledged in SE Revision 1 of Reference 5 will be issued as the approved method by incorporating Reference 23 into Reference 5. Revision 1 of ANP-10298P will be issued within 3 months of the Reference 23 SE date of March 31, 2014.
- 24. ANP-3152P, Revision 0 (Attachment 14 to Reference 1), "Browns Ferry Units 1, 2 and 3 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel, AREVA NP, Inc., October 2012.
- 25. ANP-3153P, Revision 0, "Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limits for ATRIUM 10XM Fuel, AREVA NP Inc., October 2012.
OFFICIAl USE ONlY PROPRIETARY INFORMATION
OFFICIAl USE ONlY PROPRIETARY INFORMATION 26. EMF-2361 PA, Revision 0, "EXEM BWR-2000 ECCS Evaluation Model," Framatome ANP, May 2001.
- 27. XN-CC-33A, Revision 1, "HUXY: A Generalized Multi rod Heatup Code with 10 CFR 50 Appendix K Heatup Option User's Manual," Exxon Nuclear Company, November 1975.
- 28. XN-NF-81-58PA, Revision 2 and Supplements 1 and 2, "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model," Exxon Nuclear Company, March 1984.
- 29. ANP-3170P, Revision 0 Attachment 19,20 to Reference 1), "Evaluation of Fuel Conductivity Degradation for ATRIUM 10XM Fuel for Browns Ferry Units 1, 2 and 3," AREVA NP Inc.,
November 2012
- 30. NED0-32465-A, Licensing Topical Report, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology and Reload Applications," GE Nuclear Energy, August 1996.
- 31. BAW-10255PA, Revision 2, "Cycle Specific DIVOM Methodology Using the RAMONA5-FA Code," AREVA NP, May 2008.
- 32. OG02-0119-260, "Backup Stability Protection (SSP) for Inoperable Option Ill Solution,"
GE Nuclear Energy, July 17, 2002.
- 33. ANF-913PA, Volume 1, Revision 1; and Volume 1 Supplements 2, 3, and 4, "COTRNSA2: A Computer Program for Boiling Water Reactor Transient Analyses," Advanced Nuclear Fuels Corporation, August 1990.
- 34. XN-NF-84-105PA, 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.
- 35. XN-NF-80-19PA, Volume 3, Revision 2, "Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description," Exxon Nuclear Company, January 1987.
- 36. ANF-1358PA, Revision 3, "The Loss of Feedwater Heating Transient in Boiling Water Reactors," Framatome ANP, September 2005.
- 37. Letter from J. W. Shea, TVA to NRC, "Technical Specification Change TS-478-Addition of Analytical Methodologies toTS 5.6.5 for Browns Ferry 1, 2, and 3, and Revision of TS 2.1.1.2 for Browns Ferry Unit 2 in Support of ATRIUM 10XM Fuel Use at Browns Ferry," dated May 16, 2014 (ADAMS Accession No. ML14139A180}.
Principal Contributors: Mathew Panicker Mohammed Razzaque Date: July 31, 2014 OFFICIAl USE ONlY PROPRIETARY INFORMATION
ML14108A334 Non-Proprietary LTR ML14113A286
- By a memorandum
- a 11 syan ema OFFICE LPL2-2/PM LPL2-2/LA SNPB/BC(A)*
SRXB/BC*
ARCB/BC NAME FSaba BCiayton JDean CJackson US hoop DATE 07/10/14 07/10/14 04/08/14 04/08/14 05/30/14 OFFICE STSB/BC OGC-NLO**
LPL2-2/BC(A)
LPL2-2/PM NAME A Elliott CKanatas LRegner FSaba DATE 06/02114 07/09/14 07/31/14 07/31/14