ML022770566

Amendments 278 & 269 to Licenses DPR-77 & DPR-79, Respectively. Changing Technical Specifications to Allow Irradiation of Up to 2256 Tritium Producing Burnable Absorber Rods (Tpbars)

ML022770566
Person / Time
Site:	Sequoyah
Issue date:	09/30/2002
From:	Hernan R NRC/NRR/DLPM/LPD2
To:	Scalice J Tennessee Valley Authority
Shared Package
ML022790007	List: ML022770566
References
-nr, TAC MB2972, TAC MB2973, TSC-00-06
Download: ML022770566 (77)

Text

September 30, 2002 Mr. J. A. Scalice Chief Nuclear Officer and Executive Vice President Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801

SUBJECT:

SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENTS REGARDING TECHNICAL SPECIFICATION CHANGE NO. 00-06 (TAC NOS. MB2972 AND MB2973) (TSC 00-06)

Dear Mr. Scalice:

The Commission has issued the enclosed Amendment No. 278 to Facility Operating License No. DPR-77 and Amendment No. 269 to Facility Operating License No. DPR-79 for the SEQUOYAH Nuclear Plant, Units 1 and 2 (SQN 1 & 2), respectively. These amendments are in response to your application dated September 21, 2001, as supplemented by letters dated June 11, July 19, August 9 and 30, and September 5 and 12, 2002. The application requested changes to the SQN 1 & 2 plant Technical Specifications (TSs) to allow for irradiation of up to 2256 Tritium Producing Burnable Absorber Rods (TPBARs). This request was in response to an agreement between the Tennessee Valley Authority and the U.S. Department of Energy.

The proposed license amendments involve (1) revising the measurement range for the source range monitors (TS Table 3.3-9), (2) increasing the required boron concentration for both the cold leg accumulators (TS 3/4.5.1) and the refueling water storage tank (TS 3/4.5.5),

(3) deleting the boron concentration and the spent fuel storage requirements for the cask pit pool (TS Section 3/4.7.14), (4) adding a limit on the number of TPBARs that can be irradiated (TS Section 5.3), and (5) providing storage requirements for spent fuel assemblies that contain TPBARs during irradiation (TS Section 5.6). A TPBAR consolidation activity was also added to the licensing basis. This submittal also provided revisions to the TS Bases in Section 3/4.6.4 associated with combustible gas control and revisions to the Bases for TS Sections 3/4.7.13 and 3/4.7.14.

The staffs Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions next biweekly Federal Register notice.

Sincerely,

/RA/

Ronald W. Hernan, Senior Project Manager, Section 2 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-327 and 50-328

Enclosures:

1. Amendment No. 278 to License No. DPR-77

2. Amendment No. 269 to License No. DPR-79

3. Safety Evaluation cc w/enclosures: See next page

ML022770566 *See previous concurrence OFFICE PDII-2/PM PDII-2/LA Technical Editor OGC PDII-2 NAME RHernan BClayton PKleene* SHom* HBerkow*

DATE 9/30/02 9/30/02 9/25/02 9/27/02 9/30/02 Distribution:

PUBLIC PDII-2 Reading H. Berkow A. Howe R. Hernan (Hard Copy)

B. Clayton (Hard Copy)

M. Padovan (Hard Copy)

P. Fredrickson, RII R. Dennig ACRS OGC J. Wermiel F. Akstulewicz S. Jones J. Hannon S. Weerakkody M. Reinhart S. La Vie M. Kowal A. Attard R. Jasinski G. Hill (4 Hard Copies)

TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-327 SEQUOYAH NUCLEAR PLANT, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 278 License No. DPR-77

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Tennessee Valley Authority (the licensee) dated September 21, 2001, as supplemented by submittals dated June 11, July 19, August 9 and 30, and September 5 and 12, 2002, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions 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 Commissions 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 Commissions regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-77 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 278, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of its issuance, and shall be implemented prior to starting up from the outage where TVA inserts tritium-producing burnable absorber rods in the core, provided there have been no changes to the facility that materially change the bases for approval of the amendment.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Herbert N. Berkow, Director Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: September 30, 2002

ATTACHMENT TO LICENSE AMENDMENT NO. 278 FACILITY OPERATING LICENSE NO. DPR-77 DOCKET NO. 50-327 Replace the following pages of the Appendix A Technical Specifications with the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.

REMOVE INSERT Index Page IX Index Page IX Index Page XIV Index Page XIV Index Page XVI Index Page XVI 3/4 3-51 3/4 3-51 3/4 5-1 3/4 5-1 3/4 5-11 3/4 5-11 3/4 7-43 3/4 7-43 5-4 5-4 5-5 5-5 5-5a 5-5a 5-5b 5-5b 5-5c 5-5c 5-5d 5-5d 5-5e 5-5e 5-5f 5-5f 5-5g 5-5g 5-5h 5-5h 5-5i 5-5i 5-5j 5-5j

-- 5-5k

-- 5-5l

-- 5-5m

-- 5-5n

-- 5-5o B3/4 6-4 B3/4 6-4 B3/4 7-9 B3/4 7-9 B3/4 7-10 B3/4 7-10 B3/4 7-11 B3/4 7-11 B3/4 7-12 B3/4 7-12 B3/4 7-13 B3/4 7-13 B3/4 7-14 --

B3/4 7-15 --

TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-328 SEQUOYAH NUCLEAR PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 269 License No. DPR-79

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Tennessee Valley Authority (the licensee) dated September 21, 2001, as supplemented by letters dated June 11, July 19, August 9 and 30, and September 5 and 12, 2002, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions 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 Commissions 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 Commissions regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-79 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 269, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of its issuance, and shall be implemented prior to starting up from the outage where TVA inserts tritium-producing burnable absorber rods in the core, provided there have been no changes to the facility that materially change the bases for approval of the amendment.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Herbert N. Berkow, Director Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: September 30, 2002

ATTACHMENT TO LICENSE AMENDMENT NO. 269 FACILITY OPERATING LICENSE NO. DPR-79 DOCKET NO. 50-328 Replace the following pages of the Appendix A Technical Specifications with the attached pages. The revised pages are identified by amendment number and contain a vertical line(s) indicating the area of change.

REMOVE INSERT Index Page IX Index Page IX Index Page XIV Index Page XIV Index Page XVI Index Page XVI 3/4 3-52 3/4 3-52 3/4 5-1 3/4 5-1 3/4 5-11 3/4 5-11 3/4 7-54 3/4 7-54 5-4 5-4 5-5 5-5 5-5a 5-5a 5-5b 5-5b 5-5c 5-5c 5-5d 5-5d 5-5e 5-5e 5-5f 5-5f 5-5g 5-5g 5-5h 5-5h 5-5i 5-5i 5-5j 5-5j

-- 5-5k

-- 5-5l

-- 5-5m

-- 5-5n

-- 5-5o B3/4 6-4 B3/4 6-4 B3/4 7-9 B3/4 7-9 B3/4 7-10 B3/4 7-10 B3/4 7-11 B3/4 7-11 B3/4 7-12 B3/4 7-12 B3/4 7-13 B3/4 7-13 B3/4 7-14 --

B3/4 7-15 --

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Table of Contents Abbreviations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii

1.0 INTRODUCTION

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1.1 Background . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1.1.1 DOE Strategy To Produce Tritium . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1.1.2 Producing Tritium in Commercial Light-Water Reactors (CLWRs) . . . . . . . . . . . . . . 2 1.1.3 Public Meetings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1.1.4 TVAs Amendment Requests for the SQN 1 & 2 Licenses . . . . . . . . . . . . . . . . . . . . 4 2.0 EVALUATION OF NUREG -1672 INTERFACE ITEMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 2.1 Interface Issue 1 - Handling of Tritium-Producing Burnable Absorber Rods (TPBARs) . . . 5 2.1.1 Removing Irradiated TPBARs From Fuel Assemblies and Consolidating the Rods . . 5 2.1.2 Preparing TPBARs for Shipping . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 2.1.3 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 2.2 Interface Issue 2 - Procurement and Fabrication Issues . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 2.2.1 Tritium Production Program (TPP) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 2.2.2 Tritium Production Program Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 2.2.3 Quality Requirements (Direct Suppliers) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 2.2.4 Pacific Northwest National Laboratorys (PNNLs) Quality Assurance (QA) Program 8 2.2.5 WesDynes Quality Assurance Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 2.2.6 Quality Requirements (Material/Service Subcontracts) . . . . . . . . . . . . . . . . . . . . . . . 9 2.2.7 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 2.3 Interface Issue 3 - Compliance With Departure From Nucleate Boiling (DNB) Criterion . . 10 2.3.1 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 2.4 Interface Issue 4 - Reactor Vessel Integrity Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 2.4.1 Licensee Analyses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 2.4.1.1 Neutron Fluence . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 2.4.1.2 Surveillance Capsule Withdrawal Schedule . . . . . . . . . . . . . . . . . . . . . . . . . . 13 2.4.1.3 Heatup and Cooldown Pressure-Temperature (P-T) Curves . . . . . . . . . . . . . . 13 2.4.1.4 Pressurized Thermal Shock (PTS) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 2.4.1.5 Emergency Response Guideline (ERG) Limits . . . . . . . . . . . . . . . . . . . . . . . . 13 2.4.1.6 Upper Shelf Energy (USE) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 2.4.2 Staff Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 2.4.3 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 2.5 Interface Issue 5 - Control Room Habitability Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 2.5.1 Specific Design Basis Accidents (DBAs) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 2.6 Interface Issue 6 - Specific Assessment of Hydrogen Source and Timing of Recombiner Operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 2.6.1 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 2.7 Interface Issue 7 - Light-Load Handling System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 2.7.1 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 2.8 Interface Issue 8 - Station Service Water System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 2.8.1 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 2.9 Interface Issue 9 - Ultimate Heat Sink (UHS) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 2.10 Interface Issue 10 - New and Spent Fuel Storage . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 2.10.1 Criticality Calculations Associated with Assemblies Containing TPBARs . . . . . . . . 25 i

2.10.2 Methodology . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 2.10.3 Normal Conditions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 2.10.4 Abnormal or Accident Conditions and Soluble Boron Requirements . . . . . . . . . . . 26 2.10.5 Criticality Analyses Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 2.10.6 Heat Load Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27 2.10.7 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27 2.11 Interface Issue 11 - Spent Fuel Pool Cooling and Cleanup System (SFPCCS) . . . . . . . 27 2.11.1 The Capability To Cool the Spent Fuel Pool . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27 2.11.2 Increase in SFP Heat Load . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28 2.11.3 Effect of SFP Boiling . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 2.11.4 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 2.12 Interface Issue 12 - Component Cooling Water System (CCS) . . . . . . . . . . . . . . . . . . . 29 2.12.1 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 30 2.13 Interface Issue 13 - Demineralized Water Makeup System (DWMS) . . . . . . . . . . . . . . . 30 2.13.1 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 30 2.14 Interface Issue 14 - Liquid Waste Management System . . . . . . . . . . . . . . . . . . . . . . . . 30 2.14.1 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 2.15 Interface Issue 15 - Process and Effluent Radiological Monitoring and Sampling System 32 2.15.1 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 2.16 Interface Issue 16 - Use of LOCTA_JR Code for Loss-of-Coolant Accident Analyses . . 33 2.17 Interface Issue 17 - Anticipated Transient Without Scram (ATWS) Analysis . . . . . . . . . 33 3.0 EVALUATION OF TECHNICAL SPECIFICATION CHANGES . . . . . . . . . . . . . . . . . . . . . . . . 33 3.1 Technical Specifications Changes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33 3.1.1 TS Table 3.3 Remote Shutdown Monitoring Instrumentation - Revised Backup Source Range Monitor Measurement Range . . . . . . . . . . . . . . . . . . . . . . . . . . . 33 3.1.2 TS 3/4.5.1 - Cold Leg Injection Accumulators - Boron Concentration Increase TS 3/4.5.5 - Refuel Water Storage Tank - Boron Concentration Increase . . . . . 34 3.1.3 TS 3/4.7.14 and Bases - Cask Pit Pool Minimum Boron Concentration -

Deletion of Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35 3.1.4 TS 5.3.1 - Design Features/Reactor Core/Fuel Assemblies . . . . . . . . . . . . . . . . . . . 36 3.1.5 TS 5.6 and the Basis for TS 3/4 7-13 SFP Storage Requirements . . . . . . . . . . . . . 36 3.1.6 TPBAR Consolidation Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 36 3.1.7 Technical Specification Bases 3/4 6.4 - Hydrogen Recombiners . . . . . . . . . . . . . . . 36 3.2 Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 36

4.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION

DETERMINATION . . . . . . . . . . . . 36 5.0 COMMENTS ON PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41

6.0 STATE CONSULTATION

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 42

7.0 ENVIRONMENTAL CONSIDERATION

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 42

8.0 CONCLUSION

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 42 APPENDIX AFLOW OF QA PROGRAM REQUIREMENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . A-1 APPENDIX BDETAILS OF NRC STANDARD REVIEW PLAN (SRP) REVIEW . . . . . . . . . . . . B-1 ii

Abbreviations ABGTS Auxiliary Building Gas Treatment System ANS American Nuclear Society ANSI American National Standards Institute ARC alternate repair criteria ART adjusted reference temperature ASL acceptable suppliers list ASME American Society of Mechanical Engineers ATWS anticipated transient without scram BPRA burnable poison rod assembly CCS component cooling system CFR Code of Federal Regulations CLA cold leg accumulator CLWR commercial light-water reactor COLR Core Operating Limits Report CRE control room envelope DAC derived air concentration DBA design-basis accident DNB departure from nucleate boiling DNBR departure from nucleate boiling ratio DOE U.S. Department of Energy DOPC doppler-only power coefficient DWMS demineralized water makeup system ECCS emergency core cooling system EFPD effective full-power day EFPY effective full-power year EGTS emergency gas treatment system EOL end of life (or end of license)

EPA U.S. Environmental Protection Agency ERCW essential raw cooling water ERG emergency response guideline FHA fuel-handling accident GDC General Design Criterion/Criteria HLSO hot leg switchover HTC heat transfer coefficient HZP hot zero power IFBA integral fuel burnable absorber Keff effective multiplication factor LBLOCA large break loss-of-coolant accident LOCA loss-of-coolant accident LPZ low population zone LTA lead test assembly MSLB main steamline break iii

NEI Nuclear Energy Institute NQAP nuclear quality assurance program NRC U.S. Nuclear Regulatory Commission ODCM off-site dose calculation manual PCT peak cladding temperature PNNL Pacific Northwest National Laboratory PTLR pressure/temperature limits report PTS pressurized thermal shock QA quality assurance RAI request(s) for additional information RCCA rod cluster control assembly RCS reactor coolant system RG regulatory guide RHR residual heat removal RWST refueling water storage tank SBLOCA small break loss-of-coolant accident SFP spent fuel pool SFPCS spent fuel pool cooling system SFPCCS spent fuel pool cooling and cleanup system SGTR steam generator tube rupture SLB steamline break SQN 1 & 2 SEQUOYAH Nuclear Plant, Units 1 and 2 SR Surveillance Requirement SRP Standard Review Plan TCF TPBAR [tritium-producing burnable absorber rod] consolidation fixture TDCT tritiated drain collector tank TEDE total effective dose equivalent TPBAR tritium-producing burnable absorber rod TPC tritium production core TPP tritium production program TS technical specification TVA Tennessee Valley Authority UFSAR updated final safety analysis report UHS ultimate heat sink USE upper-shelf energy WABA wet annular burnable absorbers WBN 1 Watts Bar Nuclear Plant, Unit 1 WGDT waste gas decay tanks

/Q atmospheric dispersion factor iv

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 278 TO FACILITY OPERATING LICENSE NO. DPR-77 AND AMENDMENT NO. 269 TO FACILITY OPERATING LICENSE NO. DPR-79 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328

1.0 INTRODUCTION

By application dated September 21, 2001, as supplemented by letters dated June 11, July 19, August 9 and 30, and September 5 and 12, 2002, the Tennessee Valley Authority (TVA, or the licensee) proposed to the U.S. Nuclear Regulatory Commission (NRC) amendments to the Technical Specifications (TSs) for the TVAs SEQUOYAH Nuclear Plant, Units 1 and 2 (SQN 1 & 2). The requested changes to the SQN 1 & 2 TSs would allow irradiation of up to 2256 tritium-producing burnable absorber rods (TPBARs). This change would allow SQN 1 & 2 to insert TPBARs into the reactor cores to support the U.S. Department of Energy (DOE) in maintaining the nations tritium inventory for the strategic nuclear weapons program.

The proposed license amendments involve (1) revising the measurement range for the source range monitors (TS Table 3.3-9), (2) increasing the required boron concentration for both the cold leg accumulators (CLAs) (TS 3/4.5.1) and the refueling water storage tank (TS 3/4.5.5),

(3) deleting the boron concentration and the spent fuel storage requirements for the cask pit pool (TS Section 3/4.7.14), (4) adding a limit on the number of TPBARs that can be irradiated (TS Section 5.3), and (5) providing storage requirements for spent fuel assemblies that contain irradiated TPBARs (TS Section 5.6). A TPBAR consolidation activity was also added to the licensing basis. The TVA submittal also provided revisions to the TS Bases in Section 3/4.6.4 associated with combustible gas control, and revisions to the Bases Sections 3/4.7.13 and 3/4.7.14.

These changes would allow SQN 1 & 2 to produce tritium during the course of normal commercial operation of the units. The tritium would later be extracted by DOE to meet its needs. The supplemental letters provided clarifying information (responses to NRC requests for additional information (RAIs)) that did not change the proposed no significant hazards consideration determination, nor did they expand the application beyond the scope of the amendment request notice published in the Federal Register on December 17, 2001 (66 FR 65000).

1.1 Background

Tritium, an essential material in U.S. nuclear weapons, is an isotope of hydrogen that decays at a rate of about 5 percent per year (half of it decays in about 12 years). Thus, tritium must be routinely replenished in nuclear weapons. The United States (U.S.) has not produced tritium since 1988, when the DOE shut down its last production facility at the Savannah River site in South Carolina. The U.S. is currently recycling tritium from dismantled U.S. nuclear weapons to ENCLOSURE

2 meet its immediate tritium needs. Resuming tritium production is essential to maintain the U.S.

nuclear weapons stockpile, according to the DOE.

1.1.1 DOE Strategy To Produce Tritium The DOE is responsible for reestablishing the capability to produce tritium gas by the end of 2005, in accordance with a Presidential directive. During the period 1995 to 1998, DOE considered two options for producing tritium. One option was to use a new linear accelerator, which does not require NRC approval. The second option was to use a commercial nuclear power plant to produce tritium, which would require NRC review and approval. On May 22, 1996, the Secretary of Energy and the Chairman of the NRC signed a joint Memorandum of Understanding that establishes the basis for NRC review and consultation regarding DOEs use of commercial reactors for producing tritium.

The DOE has developed a technology for producing tritium using lithium, rather than boron, in burnable absorber rods installed in commercial pressurized-water reactors. Neutron irradiation of the lithium burnable absorber rods in the reactor cores will convert the lithium to tritium. The rods will then be removed from the fuel assemblies and shipped to the Savannah River site, where the DOE will extract the tritium.

1.1.2 Producing Tritium in Commercial Light-Water Reactors (CLWRs)

In the first phase of the tritium program at CLWRs, the NRC evaluated the DOEs proposal to irradiate a limited number of fuel assemblies (lead test assemblies or LTAs) containing TPBARs in a commercial nuclear reactor. An NRC report, NUREG-1607, A Safety Evaluation Report Related to the Department of Energys Proposal for the Irradiation of Lead Test Assemblies Containing Tritium-Producing Burnable Absorber Rods in Commercial Light-Water Reactors, May 1977, documents the results of the NRCs review. The DOE contracted with TVA to conduct a one-time confirmatory test at the Watts Bar Nuclear Plant, Unit 1 (WBN 1), near Spring City, Tennessee. In April 1997, TVA applied for an amendment to the WBN 1 operating license to put 32 TPBARs in the WBN 1 reactor core during one fuel cycle. The NRC issued a license amendment to TVA in September 1997 authorizing the test irradiation of the TPBARs.

TVA loaded the TPBARs into the WBN 1 core during the units first refueling outage and irradiation began the next month. TVA removed the 32 irradiated TPBARs from the reactor during the WBN 1 spring 1999 outage. The DOE shipped the rods offsite and subjected them to nondestructive examination and destructive postirradiation examination to confirm the TPBAR design methodology and performance. The WBN 1 LTA program authenticated the capability and efficacy of producing tritium in a commercial light-water reactor. The TPBARs performed as expected incore during the irradiation, as shown by the monitoring performed during the 17-month irradiation and the subsequent examinations.

The Secretary of Energy announced on December 22, 1998, that he had chosen the CLWR technology as the primary means for producing tritium and would retain the accelerator option as a backup. On May 6, 1999, the Secretary of Energy selected TVAs WBN 1 and SQN 1 & 2 nuclear power plants in Tennessee as the facilities which will irradiate the TPBARs for DOE.

CLWRs are designed and constructed to produce electrical power. A CLWR nuclear reactor core normally has standard burnable poison rod assemblies (BPRAs) inserted into many of the fuel assemblies. The BPRAs absorb neutrons to help control power distribution in the reactor, ensuring an even distribution of heat generation and extending the time between refueling outages. The DOE developed an alternative type of BPRA, called a TPBAR. A TPBAR functions in a similar manner to a BPRA, but the TPBAR absorbs neutrons using lithium aluminate instead of boron. There is no uranium or plutonium in a TPBAR. The licensee will

3 install TPBARs in fuel assemblies where standard BPRAs are normally placed. A TPBAR assembly contains up to 24 TPBARs.

Tritium is produced when the neutrons strike the lithium material in a TPBAR. A solid zirconium material in the TPBAR (called a getter) captures neutrons as they are produced. The tritium remains trapped in the getter material. To extract the tritium the DOE will later heat the TPBARs in a vacuum at much higher temperatures than normally occur during the operation of a light-water reactor. DOE will extract the tritium in its Tritium Extraction Facility at its Savannah River Site in South Carolina (currently under construction).

The second phase of the tritium program focuses on using TVAs WBN 1 and SQN 1 & 2 nuclear plants to produce tritium. The NRC has reviewed the DOEs safety assessments submitted in its Tritium Production Core Topical Report, NPD-98-181, dated July 30, 1998, as revised on February 10, 1999. The DOEs topical report assessed how inserting up to 3300 TPBARs into the reactor core of a typical large nuclear power plant (a standard or reference plant) would affect the core. NRC reviewed the topical report and DOEs responses to the staffs RAIs against the guidance in NRC NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants. NRC's review is documented in NUREG-1672, Safety Evaluation Report Related to the Department of Energys Topical Report on the Tritium Production Core, May 1999. Section 5 of NUREG-1672, Summary and Conclusions, states that the DOE satisfactorily addressed many technical issues in its topical report, as documented in the NUREG. However, the NUREG identified 17 issues that an NRC licensee would need to address before operating with TPBARs in the cores because the parameters of the reference plant might not bound the actual design of an individual plant.

Thus, the staff concluded that a licensee wanting to irradiate TPBARs must first apply to the NRC for an amendment to the individual facility operating license to do this. NUREG-1672 further indicated that the licensees application must address the 17 plant-specific interface issues before operating with TPBARs to produce tritium. As mentioned above, TVA submitted a license amendment request for tritium production to the NRC for SQN 1 & 2 on September 21, 2001. The current schedule for loading the TPBARs into the SQN 1 & 2 reactor cores (up to 2256 TPBARS per core) is fall 2003. Irradiation of TPBARs in the WBN 1 reactor is expected to begin earlier in 2003. A license amendment was issued to TVA for WBN 1 for tritium production on September 23, 2002.

1.1.3 Public Meetings Over the past 5 years, the NRC, DOE, and TVA have held several public meetings to give the public an opportunity to comment on the technical issues of the confirmatory test and to inform the public of NRC activities. An initial meeting was held at NRC Headquarters in Rockville, Maryland, in February 1997. Another public meeting was held near TVA's WBN 1 in Tennessee in August 1997, prior to loading LTA TPBARs into the reactor core as part of the concept verification process. In addition, technical meetings that were open to the public were held on March 23 and August 24, 2000, and August 20, 2001. The NRC staff discussed the status of technical issues and proposed schedules for licensing activities at these meetings.

The NRC also held a meeting on October 2, 2001 with citizens of the eastern Tennessee area in Evensville, Tennessee to discuss the DOE/TVA program to produce tritium at the WBN 1 and SQN 1 & 2 nuclear plants and to review the NRCs plans regarding the license amendment applications and to receive public comments. The public comments were reviewed by the NRC staff and responded to in the NRC meeting summary following that meeting. The NRC plans to hold another public meeting in Rhea County, Tennessee, on the evening of October 30, 2002, before any TPBARs are loaded at either TVA site. The NRC will explain the results of its environmental assessments and safety evaluations of TVAs license amendment requests for WBN 1 and SQN 1 & 2.

4 1.1.4 TVAs Amendment Requests for the SQN 1 & 2 Licenses TVAs amendment request of September 21, 2001, proposes inserting up to 2256 TPBARs into each SQN 1 & 2 reactor. TVA would irradiate the TPBARs for one fuel cycle (about 18 months). TVA would then remove the irradiated TPBARs, consolidate them into shipping containers and casks, and the DOE would then take possession and ship the TPBARs to its Tritium Extraction Facility at its Savannah River site in South Carolina. Current plans are to have TVA insert new TPBARs in the SQN 1 & 2 reactors and repeat the process for the life of the plants, as required by the DOE.

There are 17 NUREG-1672 interface issues and a number of TS and other license changes that the NRC was required to review and approve before TVA can operate with TPBARs in the SQN 1 & 2 reactor cores. NUREG-1672 listed the following interface issues that needed to be addressed prior to issuance of license amendments for SQN 1 & 2:

(1) handling of TPBARs (2) procurement and fabrication (3) compliance with departure from nucleate boiling (DNB) criterion (4) reactor vessel integrity analysis (5) control room habitability systems (6) specific assessment of hydrogen source and timing of recombiner operation (7) light load handling system (8) station service water system (9) ultimate heat sink (UHS)

(10) new and spent fuel storage (11) spent fuel pool cooling and cleanup system (SFPCCS)

(12) component cooling water system (13) demineralized water makeup system (DWMS)

(14) liquid waste management system (15) process and effluent radiological monitoring and sampling system (16) use of LOCTA_JR code for loss-of-coolant accident (LOCA) analyses (17) anticipated transient without scram (ATWS) analysis Each of these interface issues is discussed separately below in this safety evaluation.

The staffs Safety Evaluation concludes that TVAs amendment request for SQN 1 & 2 is acceptable.

5 2.0 EVALUATION OF NUREG -1672 INTERFACE ITEMS The NRCs evaluation of each of the 17 NUREG-1672 tritium production program (TPP) interface issues appears below.

2.1 Interface Issue 1 - Handling of Tritium-Producing Burnable Absorber Rods (TPBARs)

This issue deals with the following three activities as described in NUREG-1672:

• removing irradiated TPBARs from fuel assemblies

• moving irradiated TPBARs to the cask pit

• consolidating and preparing TPBARs for shipping 2.1.1 Removing Irradiated TPBARs From Fuel Assemblies and Consolidating the Rods As described by TVA in Enclosures 1 and 4 of its application to the NRC dated September 21, 2001, TPBARs will be shipped to the SQN 1 & 2 site with new fuel assemblies and, after inspection, placed into the spent fuel pool (SFP). TVA routinely reuses fuel assemblies that were irradiated for only one fuel cycle (once-burned fuel assemblies), which is normal for commercial nuclear power plants. Fuel assemblies are normally inserted for three cycles (in different locations in the cores). For once-burned fuel assemblies containing TPBARs, plant operators will remove the TPBAR assemblies from the fuel assemblies and temporarily store the TPBAR assemblies in old spent fuel assemblies or TPBAR-assembly-holding fixtures in the SFP.

The TPBAR and the BPRA are nearly identical in design and weight (58 and 65 lbs respectively), which will permit the use of the BPRA tool to transfer TPBARs to selected fuel assemblies. The fuel assemblies with the TPBARs will then be transferred to the reactor vessel and placed in the cores using the same methods, procedures, and equipment as for non-TPBAR fuel. Following irradiation in the reactor, the spent fuel assemblies containing the TPBARs will be returned to the SFP and stored in the spent fuel racks. The licensee states that TPBAR consolidation will begin approximately 30 days after refueling is complete. When consolidation begins does not depend on any safety issues (e.g., decay heat), but rather on scheduling. The 30-day estimate corresponds to when the licensee expects to be finished with all outage- related activities and can then begin the consolidation process.

For the TPBAR consolidation process, the licensee will use a specially-designed TPBAR consolidation fixture (TCF), which will be installed in the cask loading pit for the consolidation process. During the consolidation process, the TPBARs will be withdrawn from their storage location and transferred from the SFP to the TCF using the TPBAR-assembly-handling-tool suspended from the SFP bridge crane. A specially designed release tool will be used to detach each TPBAR from the baseplate. Once released, the TPBAR slides into a specially designed canister, which can store 300 TPBARs. Prior to operation at the SQN site, the licensee will perform both dry and wet testing to ensure proper operation of the TPBAR consolidation process. Loaded canisters will be returned to the spent fuel racks until they are removed from the site. The spent fuel bridge crane is used to handle the canisters. A review of this operation is provided in Section 2.7.

The licensee evaluated the consequences of a mishandled, or dropped, fuel assembly containing an assembly of 24 TPBARs. The consequences of a postulated drop of a TPBAR assembly are bounded by the fuel-handling accident (FHA). The licensee evaluated the results of an accident both in containment and in the fuel-handling area of the Auxiliary Building. The licensee stated that the accident occurring in containment resulted in the largest offsite doses, but the calculated consequences remain well within the limits of Title 10 of the Code of Federal Regulations (10 CFR), Part 100.

6 2.1.2 Preparing TPBARs for Shipping The transportation and consolidation of TPBARs will involve the movement of heavy loads using the auxiliary building crane. The loaded consolidation canisters will be transferred from the SFP to a specially designed cask that will be placed in the cask-loading pit. Once loaded, the cask will be removed from the cask-loading pit and prepared for transportation. These activities involve heavy load lifts. The initial installation of the TCF in the cask pit and the transportation of the TCF to the cask laydown area to allow for cask handling and loading also involve heavy load lifts.

The cask-loading pit and cask laydown area are separated from the spent fuel storage racks by a wall. This physical separation protects the stored fuel during movement of the shipping cask or the TCF for some potential heavy-load-handling events. Due to the close proximity of stored fuel during movement of the TCF, the licensee will ensure that all lifts will be controlled by site procedures. The lifting devices and the interfacing lift points for the TPBAR heavy loads are in compliance with the increased safety factors in Sections 5.1.1 and 5.1.6 of NUREG-0612, Control of Heavy Loads at Nuclear Power Plants, and Sections 1 through 6 of American National Standards Institute (ANSI)/American Society of Mechanical Engineers (ASME) N14.6, Standard for Special Lifting Devices for Shipping Container Weighing 10,000 Pounds (4500 kg) or More for Nuclear Materials. The licensee further commits to maintain all the controls on movement of heavy loads committed to in its response to Generic Letter 81-07, Control of Heavy Loads (i.e., crane interlocks to prevent transport of heavy loads over the new fuel pool and SFPs, definition of safe load paths, crane inspection, and operator training).

However, the staff concluded that the movement of the TPBAR shipping cask would involve additional risks not present in the TCF movement because a large number of TPBARs could be damaged in a load-handling accident. To address this concern, TVA stated in its letter dated July 19, 2002, it is in the process of converting the main hoist of the 125-ton capacity Auxiliary Building crane to the Ederer X-Sam system (TVA expects to complete the conversion early in 2003, well in advance of when cask handling will be required), and that this crane will be used to lift the TPBAR shipping casks. This hoist has been accepted as part of overhead cranes found to satisfy the criteria of NUREG-0554, Single Failure Proof Cranes for Nuclear Power Plants. When these conditions are satisfied, the criteria of NUREG-0612 will be satisfied and cask movement will be acceptable.

2.1.3 Summary Based upon this evaluation, the staff concludes that the equipment and administrative controls that TVA states will be applied to handling TPBAR assemblies provide reasonable assurance of safety. The physical separation of the TCF from the stored fuel and the proposed personnel training, equipment inspections, and procedural controls provide adequate defense-in-depth to assure an extremely small probability of a load drop during TCF-handling operations that would damage spent fuel, which satisfies NUREG-0612 guidelines. The movement of the TPBAR shipping cask will be acceptable with the provision that the SQN Auxiliary Building crane satisfies the criteria of NUREG-0554. The design of the TPBAR assemblies allows the use of existing equipment and procedures to safely handle burnable poison rod assemblies and fuel assemblies. In the unlikely event of an accident associated with TPBAR assembly handling, TVA adequately demonstrated that applicable regulatory offsite and onsite dose limits will not be exceeded. Therefore, the existing light-load-handling system satisfies Section 9.1.4 of the Standard Review Plan (SRP), NUREG-0800, for TPBAR assembly handling, and the proposed TPBAR-assembly-handling process is acceptable.

7 2.2 Interface Issue 2 - Procurement and Fabrication Issues This section documents the staffs review of the adequacy of TPBAR procurement and fabrication processes with respect to TVA compliance with the requirements of 10 CFR Part 50, Appendix B, and 10 CFR Part 21. The staff also reviewed the quality assurance (QA) relationships among the contracting parties, focusing on the TPBAR procurement and fabrication activities of Pacific Northwest National Laboratory (PNNL) and WesDyne International LLC (WesDyne). Appendix A shows the overall flow of QA program requirements for DOEs tritium production program.

2.2.1 Tritium Production Program (TPP)

The DOE manages the TPP, including major procurements. The DOE buys TPBAR design, fabrication, irradiation, and transportation services to deliver irradiated TPBARs to its Tritium Extraction Facility. Major DOE suppliers include PNNL, WesDyne, and TVA.

The DOE/TVA Interagency Agreement of January 1, 2000, provides a means for imposing TVA requirements directly on DOEs TPBAR suppliers. DOE procurement documents for safety-related material, items, and services are reviewed by TVA for acceptance. The DOE/TVA Interagency Agreement defines TPBARs as safety-related, basic components that shall be supplied in accordance with a QA program that complies with 10 CFR Part 50, Appendix B.

The DOE/TVA Interagency Agreement further requires that TPBARs and services must conform to the guidance of NRC Regulatory Guide (RG) 1.28, Rev. 3, Quality Assurance Program Requirements (Design and Construction).

The DOE/TVA Interagency Agreement passes TVAs Tritium Production Program Requirements (TVA-TPPR-99-01, Rev. 1, dated October 6, 2000) through to TPBAR design, analysis, and fabrication suppliers. The DOE/TVA Interagency Agreement also requires that suppliers be included on TVAs Acceptable Suppliers List (ASL). The ASL identifies vendors whose QA programs conform with 10 CFR Part 50, Appendix B. Those vendors are acceptable as suppliers of items and services consistent with the requirements of TVA's QA program.

2.2.2 Tritium Production Program Requirements Activities associated with TPBAR design, procurement of materials and services, fabrication, and delivery are performed in accordance with TVA's NRC-approved nuclear QA program (NQAP), TVA-NQA-PLN89-A, dated July 25, 2000. TVAs NQAP includes provisions for TVA to specify the applicable QA requirements for items or services supplied by others. In accordance with the DOE/TVA Interagency Agreement, TVA has elected to qualify TPBAR suppliers as though they were direct suppliers to TVA. DOEs direct suppliers (PNNL and WesDyne) have submitted their QA programs to TVA and have been placed on TVA's ASL in accordance with the NQAP. Suppliers are maintained on the ASL through annual evaluations and triennial audits.

TVA requires that all TPBAR safety-related materials, items, and services comply with the technical, functional, and quality requirements of TVA-TPPR-99-01. This document requires that TPBARs be designed, fabricated, and delivered in accordance with the methods of the basic and supplementary requirements of ASME NQA-1-1994, and comply with the regulatory positions of RG 1.28, Rev. 3. The staff has reviewed the DOE/TVA Interagency Agreement and TVA-TPPR-99-01, and finds that they establish an effective method for controlling the TPBAR procurement and fabrication process in accordance with applicable NRC regulatory procurement and QA requirements.

8 2.2.3 Quality Requirements (Direct Suppliers)

DOE selected the following two suppliers of TPBARs:

• PNNL, located in Richland, Washington, is a DOE-owned site operated by the Battelle Memorial Institute.

• WesDyne, a wholly owned subsidiary of Westinghouse Electric Company LLC, also supplies TPBAR fabrication and procurement services. The TPBAR Fabrication Facility is located at the Westinghouse Fuel Fabrication Plant in Columbia, South Carolina.

2.2.4 Pacific Northwest National Laboratorys (PNNLs) Quality Assurance (QA) Program PNNLs scope of work includes design and fabrication process improvements associated with tritium production and material/subcomponent procurements to support TPBAR irradiation under a full-scale production program.

TVA has audited PNNL's QA program applicable to this work. The TPBAR design interface agreement between TVA and PNNL is documented in TVA-TPPR-99-02, Rev. 0, dated November 30, 1999. This agreement identifies the controls associated with PNNL obtaining and using technical information from TVA and the TVA fuel vendor in TPBAR design activities.

The NRC reviewed TVAs agreement and determined it provides the necessary detail to satisfy 10 CFR Part 50, Appendix B, Criterion 3, which mandates establishing measures to identify and control design interfaces and to coordinate among participating design organizations. PNNL has issued a similar document (TTQP-1-021) for satisfying design interface requirements.

PNNL developed and qualified the TPBAR design and fabrication processes, fabricated and delivered the TPBARs for use as LTAs, obtained LTA irradiation services from TVA, and performed LTA postirradiation examinations. TVA irradiated the LTAs in the WBN 1 reactor core during Cycle 2, as authorized under License Amendment No. 8, dated September 15, 1997. Monitoring during the 17-month irradiation in WBN 1 and subsequent postirradiation examinations showed the LTAs performed as expected.

The scope of the staff's review included implementing procedures and activities for controlling design, procurement, fabrication, assembly, and handling of the LTAs. As part of its review of this license amendment, the staff reviewed the QA program implemented at the PNNL facility and found the program to be acceptable. PNNL's quality controls were effective in assuring compliance with the requirements of 10 CFR Part 21 and 10 CFR Part 50, Appendix B.

2.2.5 WesDynes Quality Assurance Program WesDynes TPBAR Fabrication Facility performs the following activities:

• receives materials and subcomponents purchased by PNNL

• procures some items (such as end plugs and spring clips)

• fabricates, processes, and assembles TPBARs

• delivers certified TPBARs to TVA or TVA's nuclear fuel manufacturer(s) for use in SQN.

9 WesDyne has subcontracted the TPBAR assembly process to Westinghouse Nuclear Fuels in Columbia, South Carolina, and has developed an interface agreement with Westinghouse Nuclear Services for support services.

WesDynes TPBAR Fabrication Project Quality Plan (DOE Contract DE-AC02-00DP00229, dated July 31, 2001) commits to comply with 10 CFR Part 50, Appendix B. WesDyne implements the NRC-approved Westinghouse Quality Management System to comply with these requirements (NRC letter to Westinghouse Electric Company LLC, Changes to Quality Management System Document, Rev. 4, dated December 7, 2000).

The QA requirements controlling all WesDyne activities related to TPBAR fabrication are defined under the TPBAR Fabrication Project Quality Plan. DOE established and imposed QA program and procurement requirements through the quality requirements of TVA-TPPR-99-01 (described above). As previously mentioned, TVA placed WesDyne on its ASL and has audited WesDyne's QA program.

2.2.6 Quality Requirements (Material/Service Subcontracts)

Quality oversight (such as program reviews, source surveillances, and audits) of material, services, and subcomponent suppliers is the responsibility of PNNL or WesDyne, with a TVA observer participating periodically. Suppliers currently producing parts or providing services to be used in the TPBARs have established and implemented QA programs that meet the requirements of 10 CFR Part 50, Appendix B. These suppliers are on PNNL's Qualified Supplier List. Some suppliers are still establishing manufacturing processes and are not currently producing parts or providing services for producing TPBARs. These suppliers have been placed on PNNL's Qualified Supplier List as approved with restrictions until the programs have been established and verified through PNNL oversight activities.

For items procured by WesDyne, WesDyne requires subcontractors and suppliers to implement QA programs that meet 10 CFR Part 50, Appendix B, requirements.

2.2.7 Summary TVAs Tritium Production Program Requirements document (TVA-TPPR 99-01) describes requirements which are applied to the procurement and fabrication of TPBARs to be irradiated in the WBN 1 and SQN 1 & 2 reactor cores. This document provides the technical, functional, and quality requirements associated with the design, analysis, materials, fabrication, and delivery of TPBARs.

The DOE/TVA Interagency Agreement identifies TPBARs as safety-related, basic components.

As such, TPBARs and related services furnished under the agreement must be supplied in accordance with a QA program that complies with 10 CFR Part 50, Appendix B, in accordance with RG 1.28, Rev. 3. In addition, the reporting requirements of 10 CFR Part 21 are imposed on suppliers of TPBARs and related services.

DOE selected PNNL and WesDyne to be suppliers of TPBARs. The activities of these suppliers are conducted in accordance with QA programs that the NRC has determined to be acceptable in that they conform to the requirements of 10 CFR Part 50, Appendix B. TVA has audited the supplier QA programs for compliance and maintains them on its ASL. In accordance with contractual requirements imposed by the direct TPBAR suppliers, their subcontractors and suppliers of components and services have implemented QA programs that meet the requirements of 10 CFR Part 50, Appendix B, and the reporting requirements of 10 CFR Part 21.

10 TVA has reviewed the QA programs of DOEs direct suppliers and qualified the suppliers as having QA programs that conform to the requirements of Appendix B to 10 CFR Part 50.

Through its procurement process, DOE has imposed 10 CFR Part 21 requirements on all TPBAR suppliers. Accordingly, the NRC staff concludes that the overall TPP structure provides for effective control of all supplier activities in compliance with applicable regulatory QA and procurement requirements.

2.3 Interface Issue 3 - Compliance With Departure From Nucleate Boiling (DNB) Criterion NUREG-1672, Section 2.4.4 states that DOEs analyses regarding the incorporation of the TPBARs in the reference plant showed that the bypass flow will remain within its design limit of 8.4 percent, and that the DNB criterion will continue to be met with no feature of the TPBAR component affecting the coolability of the cores. The staff agrees with this assessment.

However, continued compliance with the DNB criterion must be evaluated for the operating conditions of a particular plant. The staff has identified this as an interface item that must be addressed by a licensee referencing the TPC topical report in its plant-specific application for authorization to irradiate TPBARs.

To address this interface item, TVA provided detailed information in Sections 1.5.3, 2.4.3 and 2.4.4 of its amendment application dated September 21, 2001, and in its response to RAI question 1 dated July 19, 2002. Framatome-ANP applied standard reload design analysis methods (BAW-10163P-A, dated June 1989) to evaluate the SQN 1 & 2 TPC designs. The analysis involved the application of DNB-based maximum allowable peaking limits, generated using the LYNXT code (BAW-10156-A, dated August 1993). The results of the analysis showed that the 7.5-percent total core bypass fraction that has been traditionally applied to the SQN licensing analyses will continue to apply to the TPBAR cores and that the presence of TPBARs in the reload core design does not challenge the DNB criterion. The DNB peaking margins preserved for the SQN 1 & 2 TPCs are similar to those preserved in recent SQN 1 & 2 standard reload core designs. The effects of local power spikes that result from axial gaps between TPBARs were accounted for in the power distribution analysis. Continued compliance with the DNB criterion is ensured through evaluations performed for each SQN core reload design.

The licensee also provided justification for continued application of its current DNB correlations.

In its July 19, 2002, RAI response, the licensee stated that the applicability of the BWCMV-A and BWU critical heat flux correlations is not impacted by the presence of TPBARs. The fuel geometry of the Mark-BW17 fuel is compatible with the TPBAR assembly and the correlations continue to be applied within their respective ranges of applicability.

2.3.1 Summary Framatome-ANP used standard analytical methods consistent with NRC-approved Topical Report BAW-10163P-A to evaluate thermal-hydraulic conditions with TPBARs in the cores to ensure that an adequate safety margin exists in the thermal-hydraulic design.

Framatome-ANPs analysis show that operating with TPBARs will not have adverse effects on the thermal and hydraulic design or performance of the SQN 1 & 2 cores. Framatome-ANP demonstrated that DNB will not occur on the most limiting fuel rod or TPBAR with at least a 95-percent probability at a 95-percent confidence level as specified in SRP Section 4.2. The thermal-hydraulic design basis of the TPBARs will also meet the applicable acceptance criteria outlined in Section 4.4 of the NRCs SRP with no feature of the TPBAR challenging the cooling capacity of the cores. Based on the results of the analysis supplied by TVA, the NRC staff concludes that the TPBARs will not have any adverse effects on compliance with the DNB criterion.

11 2.4 Interface Issue 4 - Reactor Vessel Integrity Analysis NUREG-1672, Section 2.5.3 states that the TPC topical report identifies the applicable regulations and describes methods for demonstrating compliance with Appendix G and H to 10 CFR Part 50 and with 10 CFR 50.61. In the TPC topical report, DOE concludes, and the staff agrees, that the reference plants pressure-temperature (P-T) limits report (PTLR) and Updated Final Safety Analysis Report (UFSAR) would need to be updated to reflect the change to the pressurized thermal shock value and include the updated P-T limit curves for the applicable effective full-power years (EFPYs). In addition, because the reactor vessel integrity analyses are dependent upon the plant-specific material properties and neutron fluence, the staff concluded that a licensee participating in DOEs program for the CLWR production of tritium must provide a discussion of the material properties for its reactor vessel(s) and perform analyses to demonstrate the reactor vessels will meet the requirements of 10 CFR 50.61 and Appendices G and H to 10 CFR Part 50. The staff has identified this as an interface item that must be addressed by a licensee referencing the TPC topical report in its plant-specific application for authorization to irradiate TPBARs.

To address this interface item, the licensee provided detailed information in Section 1.5.4 of its submittal dated September 21, 2001, and in its response dated July 19, 2002, to question 25 in the staffs initial RAI. The licensee performs several analyses to determine the impact of neutron irradiation on the SQN 1 & 2 reactor vessels. These analyses include a surveillance capsule withdrawal schedule, heatup and cooldown pressure-temperature limit curves, pressurized thermal shock calculations, and upper shelf energy evaluations. All of these analyses and evaluations can be affected by changes in the neutron fluences and operating temperatures and pressures.

To satisfy the tritium production requirements, a larger number of feed batch assemblies are to be provided in each refueling, necessitating that once burned assemblies are placed on the periphery of the cores. This loading will result in higher core leakage compared to typical low leakage core loading patterns. Higher neutron leakage (neutron source) will increase the vessel fluence. To mitigate the potential impact this would have on the vessel fluence, the SQN TPC designs will employ suppression methods to reduce the power production in peripheral core locations. These methods will include loading fuel assemblies with higher burnups or loading BPRAs containing B4C-Al2O3 in key peripheral core locations.

The acceptability of the resulting fluence values as well as the methodology used in the evaluation of the fluence is part of this review. Calculated and best estimate fluence values were determined for the SQN Units 1 & 2 reactor vessels. Fluence values were projected to operating times of 20, 32, and 48 EFPYs, assuming cycle 11 and beyond are operated with tritium producing cores and at an uprated reactor power level of 3455 Mwt, which was approved by license amendments for SQN 1 & 2 on April 30, 2002. The method used to estimate the projected value of pressure vessel fluence involves 2-dimensional neutron transport calculations by a 3-dimensional synthesis technique as recommended in NRC RG 1.190. The values were calculated using a synthesis of R-theta, R-Z, and R calculations in accordance with the RG. The calculated fluence values are recommended to be used for the projection of vessel material properties (up to 48 EFPYs). Best estimate fluence values were also determined and are based on the calculated values with a plant-specific bias factor applied.

For the SQN Units, eight surveillance capsules have been analyzed and the average ratio of measurement to calculation for these capsules is 1.076 for fluence (E>1.0 MeV). This value falls well within the +/- 20 percent tolerance specified in RG 1.190.

Based on this analysis, the licensee determined that the maximum vessel exposure point has a lower fluence with the tritium production core fluence projections than for the previous projections made for the SQN 1.3 percent power uprate program. This occurs because of the vessel fluence reduction methods the licensee will employ in SQN TPC designs.

12 2.4.1 Licensee Analyses TVAs application addressed this issue. Appendices G and H to 10 CFR Part 50 and 10 CFR 50.61, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock (PTS), pertain to this issue. Appendix H to 10 CFR Part 50 specifies material surveillance program requirements to monitor changes in the fracture toughness properties of the reactor vessel that occur as a result of exposure of these materials to neutron irradiation. The requirements of Appendix G to 10 CFR Part 50 and the PTS Rule ensure that the material property changes due to neutron irradiation (termed neutron embrittlement) are sufficiently low that the plant is operated only under conditions with sufficient safety margins against fracture of the reactor vessel.

Appendix G to 10 CFR Part 50 states that the Charpy V-notch upper-shelf energy (USE) level for the reactor vessel must exceed 68 joules (50 ft-lb) throughout the service life. If it is anticipated that a vessel may fall below 68 joules (50 ft-lb) before the end of the plant license, continued plant operation can occur if the NRC approves the licensees analysis demonstrating margins of safety against fracture equivalent to those required by Appendix G of the ASME Code.

The PTS Rule establishes screening criteria that define a limiting level of neutron embrittlement beyond which plant operation cannot continue without further plant-specific evaluation. The PTS screening criteria are given in terms of reference temperature, RTPTS.

TVAs September 21, 2001, submittal addressed the following considerations that apply to Interface Issue 4:

• surveillance capsule withdrawal schedule

• heatup and cooldown pressure-temperature curves

• PTS

• emergency response guideline (ERG) limits

• USE Neutron embrittlement of the reactor vessel material affects reactor vessel integrity. Neutron fluence is the main parameter used to characterize the extent of embrittlement expected in the future. The level of neutron embrittlement depends on neutron fluence, with higher fluence equating to greater embrittlement, assuming that all other parameters remain constant.

Therefore, change in neutron fluence is a key consideration in evaluating the impact of the TPC in the reactor vessel integrity analysis.

2.4.1.1 Neutron Fluence For the TPC, TVA will insert new fuel assemblies containing TPBARs into the reactor during each refueling outage. During that evolution, TVA will likely put once-burned assemblies on the periphery of the cores. This loading results in higher core leakage than low-leakage loading configurations. Higher neutron leakage (neutron source) will increase the vessel fluence. This section reviews the acceptability of the new fluence values and the methodology TVA used to evaluate fluence.

The method TVA used to estimate the projected value of pressure vessel fluence (to the end of the license) is based on the discrete ordinates approach with a one- and two-dimensional

13 synthesis (following RG 1.190). The loading pattern was based on a uniform distribution of new assemblies throughout the cores, which resulted in a more conservative fluence evaluation than to low-leakage patterns which were subsequently implemented at SQN 1 & 2. Comparing the relative values of the source distribution of the design basis and the equilibrium tritium production cycle indicates that the design basis is conservative by a large margin. Therefore, the design basis neutron fluence value used to determine the end-of-license (EOL) neutron embrittlement for the reactor vessel is conservative relative to the actual estimated EOL value.

The presence of the TPBARs does not alter the spectrum of the neutron leakage source because lithium-6 is a thermal neutron absorber. Therefore, although the TPBARs suppress the local power, they do not affect the neutron leakage spectrum. This is one of the reasons why the plant-specific source estimate is conservative with respect to the design value. The presence of the TPBARs is exploited to lower the peak vessel fluence because the bars will be inserted in key peripheral locations to reduce the power in specific assemblies important to peak vessel fluence.

2.4.1.2 Surveillance Capsule Withdrawal Schedule TVA developed a withdrawal schedule to periodically remove the surveillance capsules from the reactor vessel in order to effectively monitor the condition of the reactor vessel materials under actual operating conditions. TVA determined that the fluence projections for the SQN 1 & 2 TPC do not exceed the fluence projections for the 1.3-percent power uprate for SQN 1 & 2.

The staff concurs with this conclusion.

2.4.1.3 Heatup and Cooldown Pressure-Temperature (P-T) Curves The heatup and cooldown P-T curves for SQN 1 & 2 were reviewed by comparing the fluence projections used in the calculation of the adjusted reference temperature (ART) for all the beltline materials in the reactor vessel for the uprated power conditions to the fluence based on the tritium production design conditions. The revised fluence projections did not exceed the fluence projections used in developing ART values for the uprated power conditions.

Therefore, the P-T curves remain valid.

2.4.1.4 Pressurized Thermal Shock (PTS)

TVA stated that the RTPTS values for the uprated power conditions do not exceed the screening criteria of the PTS Rule. Furthermore, TVA stated that the fluence projections for the TPC design conditions do not exceed the fluences used in developing the RTPTS values for the power uprate. Therefore, SQN will remain within the screening criteria of the PTS rule.

2.4.1.5 Emergency Response Guideline (ERG) Limits TVA developed the ERG limits using the design basis EOL neutron fluence projection. Since this fluence value bounds the estimated EOL neutron fluence projection, the ERG limits remain valid.

2.4.1.6 Upper Shelf Energy (USE)

TVA stated that, based on 1.3-percent uprated conditions, all beltline materials are expected to have USEs greater than 50 ft-lb through EOL (32 EFPYs) as required by 10 CFR Part 50, Appendix G. The EOL (32 EFPYs) USE was predicted using the EOL 1/4T fluence projection.

Furthermore, TVA stated that the fluence projections at tritium production core design conditions do not exceed the fluence projections for the uprated power conditions without the TPBARs and that the current predicted USE values for SQN 1 & 2 remain valid.

14 The staff reviewed TVAs reasoning and concurred with its conclusions regarding USE.

2.4.2 Staff Evaluation To satisfy the tritium production requirements, a larger number of feed batch assemblies are to be inserted in the cores during each refueling, necessitating that once-burned assemblies be placed in the periphery of the cores. This loading will result in higher core leakage than typical low-leakage core loading patterns. Higher neutron leakage (neutron source) will increase the vessel fluence, and the licensee has employed designs intended to reduce the power production in peripheral core locations. The staff reviewed the acceptability of the resulting fluence values and the methodology the licensee used to evaluate the fluence.

Calculated and best-estimate fluence values were determined for the SQN 1 & 2 reactor vessels. Fluence values were projected to operating times of 20, 32, and 48 EFPYs, assuming Cycle 11 and subsequent operation with a tritium-producing core and at an uprated reactor power level of 3455 MWt. The method used to estimate the projected values of pressure vessel fluence involved calculating two-dimensional neutron transport by a three-dimensional synthesis technique (i.e., a synthesis of R-theta, R-Z, and R calculations) as recommended in RG 1.190 for the projection of vessel material properties (up to 48 EFPYs). The licensee also projected best-estimate fluence values by applying a plant-specific bias factor to the calculated values. For SQN 1 & 2, eight surveillance capsules were analyzed. The average ratio of measurement to calculation for these capsules was 1.076 for fluence (E>1.0 MeV). This value falls well within the +/- 20-percent tolerance specified in RG 1.190.

2.4.3 Summary Based on this analysis, the licensee determined that the maximum vessel exposure point has a lower fluence with the tritium production core fluence projections than for the previous projections made for the SQN 1 & 2 1.3-percent power uprate. This is because the use of TPBARs results in a reduction in neutron leakage from core locations where the peak vessel fluence occurs.

The staff has reviewed the licensees analysis and has found it to be credible. The NRC staff, therefore, finds that the reactor vessel integrity interface item has been satisfactorily addressed.

2.5 Interface Issue 5 - Control Room Habitability Systems This section addresses the impact of TPBARs on design basis accident (DBA) radiological consequences, specifically control room habitability as addressed in TVAs submittals of September 21, 2001, and June 11, July 19, August 9 and 30, and September 5 and 12, 2002.

DOEs Tritium Production Core Topical Report, NDP-98-153, concluded that the control room doses for a LOCA would exceed the limits of General Design Criterion (GDC) 19, Control Room. In the topical report, DOE stated that this situation arises due to the assumed high leakage rate from emergency core cooling systems (ECCSs) outside of the containment.

Accordingly, TVA analyzed the doses for all DBAs at SQN, although the interface issue specifically referred to the ECCS leakage contribution to the LOCA doses. TVA described the analyses in Section 2.15.6, Radiological Consequences of Accidents, of Enclosure 4 to its letter of August 20, 2001. Section 2.15.6 addresses the potential radiological impact of operation with the maximum number of TPBARs installed for various design basis accidents.

The NRC staffs review of TVAs analyses is provided below.

15 2.5.1 Specific Design Basis Accidents (DBAs)

This section provides the staffs evaluation of the DBA analysis results reported in the amendment submittal. The staff evaluated each analysis input and assumption against analysis descriptions in the SQN UFSAR, regulatory guidance, and staff experience in performing similar reviews. The staff performed independent analyses to confirm the conservatism of the TVA analyses. However, the findings of this safety evaluation are based on the descriptions of the TVA analyses and other supporting information docketed by TVA.

TVA considered the impact of TPBAR operation on the previously analyzed DBAs. The DBAs considered included:

Loss of Offsite Power (LOOP)

Waste Gas Decay Tank Failure (WGDTF)

Loss of Coolant Accident (LOCA)

Main Steam Line Break (MSLB)

Steam Generator Tube Rupture (SGTR)

Fuel Handling Accident (FHA)

Rod Ejection Accident (REA)

TVA determined that the impacts of the TPBARs on the REA were bounded by the consequences for the LOCA. For the remaining analyses, TVA has stated that analysis inputs and assumptions will not change for the TPC with a few exceptions. Each DBA will be discussed individually. Analysis changes common to most DBAs included the following:

Initial source terms were revised, taking into account TPC parameters.

ARCON96 code control room atmospheric dispersion coefficient (/Q) values were determined and used for many of the analyses. ( is the Greek symbol chi)

TVA determined the total effective dose equivalent (TEDE) for each accident. These results supplement, but do not supplant, the whole body and thyroid dose results currently in the SQN licensing basis.

TVA used dose conversion factors from Federal Guidance Report (FGR) 11 and 12.

For accidents other than the FHA and LOCA, TVA has assumed that all of the tritium gas in two failed TPBARs (2.4E-4 Ci) would be released to the reactor coolant.

TVA did not report control room doses for accidents other than the LOCA and MSLB. In response to staff questions, TVA stated that they concurred that GDC-19 applied to the spectrum of accident scenarios in the SQN UFSAR, but that the scenarios other than the LOCA and MSLB had not been analyzed against the GDC-19 criteria for main control room habitability. Although the results of the LOCA analysis will often be bounding for the other accident sequences, there can be plant-specific differences that may cause another accident scenario to be more limiting. For example, facilities using alternative repair criteria for steam generators may be limited by the results of the MSLB. TVA made a commitment in the August 30, 2002, letter to conduct control room habitability analyses for the UFSAR accident

16 scenarios and provide the results including a tabulation of analyses inputs and assumptions to the NRC by December 2002.

In order to support a staff determination that SQN will be in compliance with GDC-19 for all accident scenarios prior to completing the additional analyses, TVA docketed an assessment of their compliance by comparison to the similarly designed WBN 1, which has analyzed all scenarios for control room doses. In this comparison, TVA considered differences and similarities in the core power, core inventories, means of control room isolation actuation, location of release points relative to control room intakes, control room volume and ventilation system flow rates, atmospheric dispersion (/Q) values, and the control room dose results of the LOCA and MSLB analyses which have been performed for both SQN and WBN. TVA concluded that, based on the similarities of the data and the fact that all calculated radiological consequences were well within acceptance criteria, there was a high level of assurance that TVA meets the requirements of GDC-19 for all postulated events. The staff concurs in this evaluation subject to the commitment of TVA to conduct the additional analyses by December 2002 to confirm this conclusion. In performing its review of this amendment request, the staff performed analyses of the control room doses. Based upon its own analyses, the staff has reasonable assurance that the additional analyses to be performed by TVA will confirm their conclusion that GDC-19 will be met.

2.5.1.1 Loss of Offsite Power (LOOP)

In the event of a loss of offsite AC power, emergency diesel generators will start and power vital loads. The main condenser circulating water pumps are not vital loads and will be unavailable.

Without circulating water, the main condenser will not be available to receive and condense steam generated during the plant cooldown. As such, this steam is released to the environment by steam safety valves and atmospheric dump valves. This steam may be contaminated due to leakage of reactor coolant into the steam generators via small tube leaks (i.e., primary-to-secondary leakage). It is assumed that the cool down, and the release, will continue for 8 hours0.333 days <br />0.0476 weeks <br />0.011 months <br />. The release is based on the transfer of reactor coolant containing radionuclides corresponding to the maximum allowed by technical specifications at a rate equivalent to a primary-to-secondary leak rate of 3.7 gpm per steam generator. Pre-accident and coincident iodine spike cases were also considered. The release also includes the radionuclides initially in the steam generators at the event onset. The reactor coolant activity is assumed to include all of the tritium gas in two TPBARs.

In evaluating the coincident iodine spiking case, TVA assumed 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for the spike duration.

TVA assumes that the coincident iodine spike continues until such time as the RCS specific activity increases to 60 times the equilibrium value. This was determined to be 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. This assumption is not consistent staff practice which assumes a fixed 8-hour duration. TVA assumed a 2-hour spike duration in 1995 for the steam generator alternative repair criteria (ARC) amendment. In a 1998 amendment request for ARC, TVA assumed 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> since the iodine release rate assumption had increased, making the reduced duration possible. Both of those amendments were approved by the staff. The staff performed analyses to determine the sensitivity of the final result to the value of this assumption. The staff found that the low population zone (LPZ) and control room doses would be increased if 8 hours0.333 days <br />0.0476 weeks <br />0.011 months <br /> were assumed, but that the results would still be within the acceptance criteria for offsite and control room dose.

The staff has determined that the 1-hour assumption is acceptable in the context of the current amendment. The staff bases this determination on (1) the results of the sensitivity analyses, (2) the fact that accident consequences of the increased tritium concentration associated with TPC operation are independent of iodine spiking, and (3) the TPC operation does not affect the core inventory or release transport of iodine. This assumption was also used in the MSLB and SGTR analyses and this determination applies to those events as well.

17 As such, the staff accepts TVAs 1-hour spike duration assumption in the context of the current proceeding. The staff also notes that the accident consequences of the increased tritium concentration in a TPC is independent of this assumption. TVA also used this assumption for the MSLB and SGTR and it also applies to the analyses of those events As discussed above in Section 2.2, TVA has concluded that there is reasonable assurance that the criteria of GDC-19 will be met for this accident. The staffs own analyses of this event are consistent with TVAs conclusion. The staff finds TVAs analysis to be acceptable. Table 1 to this safety evaluation tabulates the analysis parameters.

2.5.1.2 Waste Gas Decay Tank (WGDTF) Failure Waste gas decay tanks (WGDT) contain radioactive gases generated during plant operation until the gases have decayed sufficiently to allow release to the environment. This DBA postulates the failure of a waste gas decay tank that releases its contents prior to adequate decay. TVA has assumed that the tank is filled with the highest concentration of each radionuclide from all sources into the WGDT. TVA assumes that the reactor coolant contains the tritium released from two TPBARs, and that 10 percent of this tritium exists as a gas which is transported to the WGDT. The contents of the tank are released non mechanistically to the environment over a period of 2 hours0.0833 days <br />0.0119 weeks <br />0.00274 months <br />.

As discussed above in Section 2.2, TVA has concluded that there is reasonable assurance that the criteria of GDC-19 will be met for this accident. The staffs own analyses of this event are consistent with TVAs conclusion. The staff finds TVAs analysis to be acceptable. Table 1 to this safety evaluation tabulates the analysis parameters.

2.5.1.3 Loss of Coolant Accident (LOCA)

A LOCA is a failure of the reactor coolant system that results in the loss of reactor coolant and, if not mitigated, fuel damage possibly including a core melt. The containment building holds up the majority of the radioactivity released from the core. Evaluation of the effectiveness of plant safety features, such as ECCS, has shown that core melt is unlikely. The objective of this DBA is to evaluate the ability of the plant design to mitigate the release of radionuclides to the environment in the unlikely event that ECCS is not effective. Two release pathways are considered: (1) leakage of containment atmosphere, and (2) leakage from systems that recirculate contaminated sump water outside of containment (e.g., certain ECCS). TVA updated the previous LOCA analysis to reflect the use of TPBARs. TVA assumes that all of the core inventory of tritium (2.68 E-7 Ci) is released to the containment atmosphere as tritiated water. TVA assumes that the tritium is not removed by the emergency gas treatment system (EGTS), the auxiliary building gas treatment system (ABGTS), or by the ice condenser. For the ECCS leakage path, TVA assumes that all of the core inventory of tritium and 50 percent of the core iodine inventory is in the sump water. The control room /Q values were recalculated for use in this analysis using ARCON96. Other analysis inputs and assumptions are unchanged.

The TVA analyses show the control room doses to be within GDC-19 criteria. The staff finds TVAs analysis to be acceptable. Table 1 to this safety evaluation tabulates the analysis parameters.

2.5.1.4 Main Steam Line Break (MSLB)

This DBA postulates an unisolable failure in one of the four main steam lines at a location outside of containment, resulting in the release of steam from the affected steam line. Since a loss of offsite power is assumed to occur, the main condenser is not available as a heat sink and the unaffected steam generators are used to cool down the plant by dumping steam to the environment. The released steam may be contaminated due to leakage of reactor coolant into

18 the steam generators via small tube leaks (i.e., primary-to-secondary leakage). TVA assumes that there is 3.7 gpm primary-to-secondary leakage into the faulted steam generator, and 150 gpd primary-to-secondary leakage into the unaffected steam generators. Two release paths are considered: (1) via the faulted steam line, and (2) via the nonaffected steam generators used for the plant cooldown. The reactor coolant activity is assumed to include the entire tritium inventory in two TPBARs. Preaccident and coincident iodine spike cases were considered (see the discussion on iodine spike duration in Section 2.2.1 above.). The release also includes the radionuclides initially in the steam generators at the event onset. It is assumed that the cool down, and the release, will continue for 8 hours0.333 days <br />0.0476 weeks <br />0.011 months <br />. ARCON96 /Q values were used for control room dose calculations. Other analysis inputs and assumptions are unchanged.

In evaluating the coincident iodine spiking case, TVA assumed 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for the spike duration.

TVA assumes that the coincident iodine spike continues until such time as the RCS specific activity increases to 60 times the equilibrium value. This assumption is not consistent staff practice which assumes a fixed 8-hour duration. TVA assumed a 2-hour spike duration in 1995 for the steam generator ARC amendment. In a 1998 amendment request for ARC, TVA assumed 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> since the iodine release rate assumption had increased, making the reduced duration possible. Both of those amendments were approved by the staff. The results of the staffs confirmatory calculations using an 8-hour duration were within acceptance criteria for offsite and control room doses.

The TVA analyses show the control room doses to be within GDC-19 criteria for both spike cases. The staff finds TVAs analysis to be acceptable.

2.5.1.5 Steam Generator Tube Rupture (SGTR)

This DBA postulates a rupture in a tube in one of the four steam generators resulting in the transfer of reactor coolant water to the ruptured steam generator. Since a loss of offsite power is assumed to occur, the main condenser is not available as a heat sink and the unaffected steam generators are used to cool down the plant by dumping steam to the environment. The released steam may be contaminated due to leakage of reactor coolant into the steam generators. In addition to the reactor coolant transferred through the ruptured tube, TVA assumes that there is 150 gpd primary-to-secondary leakage into the unaffected steam generators. Two release paths are considered: (1) via the ruptured steam generator, and (2) via the nonaffected steam generators used for the plant cooldown. The reactor coolant activity is assumed to include the entire tritium inventory in two TPBARs. Pre-accident and coincident iodine spike cases were considered. (See the discussion on iodine spike duration in Section 2.2.1 above.) The release also includes the radionuclides initially in the steam generators at the event onset. It is assumed that the cooldown, and the release, will continue for 8 hours0.333 days <br />0.0476 weeks <br />0.011 months <br />. Other analysis inputs and assumptions are unchanged.

As discussed above, TVA has concluded that there is reasonable assurance that the criteria of GDC-19 will be met for this accident. The staffs own analyses of this event are consistent with TVAs conclusion. The staff finds TVAs analysis to be acceptable.

2.5.1.6 Fuel Handling Accident (FHA)

This DBA postulates the drop of an irradiated fuel assembly during refueling operations. All of the fuel rods in the assembly, including the 24 TPBARs, are assumed to rupture, releasing the radionuclides within the fuel clad gap to the fuel pool or reactor cavity water. Two cases are considered: (1) an FHA within the containment, and (2) an FHA in the auxiliary building spent fuel pool area. The analysis assumes that the release from the containment is isolated in 10 minutes and that 100 percent of the activity is released in 2 hours0.0833 days <br />0.0119 weeks <br />0.00274 months <br /> for an event in the auxiliary building.

19 As discussed above in Section 2.2, TVA has concluded that there is reasonable assurance that the criteria of GDC-19 will be met for this accident. The staffs own analyses of this event are consistent with TVAs conclusion. The staff finds TVAs analysis to be acceptable.

2.5.1.7 Rod Ejection Accident (REA)

TVA stated that the radiological consequences of this event are bounded by the evaluation of the DBA LOCA. The SQN licensing basis does not include a radiological consequence analysis specific to the REA. This approach was accepted as a part of the original licensing of SQN.

The staff, based on its experience in reviews at similar plants, accepts the TVA conclusion.

2.5.2 Control Room Unfiltered Inleakage TVA assumed an unfiltered inleakage into the control room envelope (CRE) of 51 cfm. The staff requested TV to provide additional information to substantiate this assumption. In response to this request, TVA described unfiltered inleakage testing performed in August 1980.

This test assessed the leakage from two pressurized ducts that pass through the CRE at 7.7 cfm and 21.3 cfm. TVA includes 19 cfm for pneumatic components and 3 cfm for door activities to arrive at the assumed 51 cfm. Since 1980, TVA has increased the rate of filtered pressurization from 200 to 1000 cfm and made modifications to ductwork. TVA also removed the functions of the normal control building pressurization fan that accounted for 21.3 cfm of the unfiltered inleakage measured in 1980. TVA has performed walkdowns of the remaining pressurized duct that accounted for 7.7 cfm of the unfiltered inleakage measured in 1980 to confirm that the configuration hasnt been changed and that the ductwork and sealing are in excellent condition. TVA also performs a periodic pressurization test that demonstrates that a positive pressure can be maintained in the CRE. TVA states that these considerations support its view that 51 cfm is an appropriate value for unfiltered inleakage.

Integrated testing at several U.S. power reactors has shown leakage exceeding that assumed in control room habitability analyses. The testing experience also indicated that the typical pressurization surveillance test may not be reliable in identifying sources of unfiltered inleakage.

The NRC has conducted several public meetings with its stakeholders on this issue since 1998 and recently published a series of proposed draft regulatory guides and a proposed generic communication on control room habitability (67 FR 31385). The intent of the final generic communication will be to formally alert licensees of the staffs findings related to inleakage testing and to request licensees to submit information that demonstrates that the facility CRE complies with current licensing and design basis, and applicable regulatory requirements.

While TVA will be subject to the generic communication when it is issued, the staff has determined that there is adequate assurance that the radiation doses to control room personnel will not impede response actions necessary to protect the public in the interim. The staff bases this finding on (1) the evaluations and integrity testing already performed by TVA, (2) the available margin between doses postulated assuming 51 cfm and the applicable habitability criteria, and (3) the limited impacts that tritium doses have on the previously analyzed doses.

As such, the TVA unfiltered inleakage assumption of 51 cfm is acceptable for the current proceeding. However, this approval does not exempt TVA from future regulatory actions that may become applicable due to the final generic communication.

2.5.3 Control Room Atmospheric Dispersion Coefficient (/Q) Values TVA calculated new /Q estimates for the plant vent and the plant stack for both units for the normal and emergency control room intakes. These values were used for the reanalysis of the LOCA and MSLB. All of the values were determined as ground level releases using meteorological data collected for the years 1976 through 1993, using the NRC-sponsored

20 ARCON96 computer code (NUREG/CR-6331, Rev.1, Atmospheric Relative Concentrations in Building Wakes). The meteorological data were obtained from the SQN meteorological tower.

TVA performed a sensitivity analysis of the value of the building cross-sectional area used in wake corrections and reported the most limiting result.

The staff performed a series of statistical checks on the meteorological data to evaluate its suitability. This evaluation compared stability class, wind speed, and wind direction distributions observed in each year between 1987 through 1993 with the other years in that range and with corresponding distributions provided in the UFSAR. The data showed a large degree of consistency, confirming the accuracy of the meteorological data sets input to ARCON96. The staff compared the ARCON96 code inputs used by TVA to the site release point and intake configuration and found the inputs to be acceptable. Since the license used a NRC-sponsored computer code and provided its code printouts for review, the staff only ran one confirmatory case. The staff finds the /Q values listed in Table 1 to be acceptable.

2.5.4 Use of Total Effective Dose Equivalent (TEDE)

The DBA analyses in the SQN licensing basis determined the whole body doses and the thyroid doses due to inhalation of radioiodine were consistent with Part 100 and Part 50 Appendix A, GDC-19. In addition, TVA also reported, for informational purposes values for the TEDE. TVA calculated these TEDE results to appropriately account for the radiological consequences of the increased tritium in the TPC core. TVA has stated that they do not consider the TEDE results to be the licensing basis for SQN. The existing whole body and thyroid dose calculation methods remain the licensing bases for SQN.

The staff agrees that tritium is not adequately addressed by the existing whole body and thyroid dose criteria since (1) the decay emission energy of tritium is insufficient to penetrate the skin and contribute significantly to the whole body dose, and (2) the thyroid dose is explicitly limited to inhalation of radioiodine. The use of TEDE as a dose quantity addresses these limitations.

The staff considered whether it would be appropriate to request that TVA incorporate the TEDE results into the SQN licensing bases along with corresponding dose criteria. However, the TVA analyses show that the impact of TPC operation on postulated TEDE doses is small, thereby indicating that use of the existing dose results, expressed in whole body and thyroid dose quantities, continue to provide bounding estimates of the radiological consequences of the accident, concurrent with TPC operations. TVA should perform TEDE analyses when analyzing offsite and control room accident doses as long as TPC operations continue at SQN to ensure that this conclusion remains valid.

2.5.5 Conclusion The staff reviewed the assumptions, inputs, and methods used by TVA to assess the radiological impacts of the proposed TPC operation at SQN. In doing this review, the staff relied upon information placed on the docket by TVA and staff experience in doing similar reviews. The staff finds that TVA used analysis methods and assumptions that are adequately conservative and consistent with regulatory guidance, where applicable. The staff accepted the TEDE acceptance criteria proposed by TVA and compared the doses estimated by TVA to the proposed criteria. The staff also compared the whole body and thyroid doses estimated by TVA against the corresponding criteria.

The staff finds, with reasonable assurance, that the licensees estimates of the EAB, LPZ, and control room doses will continue to comply with Part 50, Appendix A, GDC-19. Therefore, TPC operation at SQN is acceptable with regard to NUREG-1672, Interface Item No. 5.

21 2.6 Interface Issue 6 - Specific Assessment of Hydrogen Source and Timing of Recombiner Operation This issue pertains to post-LOCA hydrogen generation inside containment. The concern is that an uncontrolled hydrogen-oxygen recombination inside containment could cause containment pressure to rise to unacceptable levels and damage the containment.

A reactor loaded with TPBARs can impact post-LOCA hydrogen generation inside containment by adding tritium and hydrogen to the non-TPBAR hydrogen inventory. For plants operating with conventional cores (without TPBARs) four sources of hydrogen need to be considered following a LOCA:

• metal-water reaction with the fuel cladding

• corrosion of materials in contact with spray and sump solutions

• radiolysis in the sump and core solutions

• hydrogen in the reactor coolant system (RCS) inventory prior to the accident When operating with TPBARs in the core, two additional sources of post-LOCA hydrogen need to be considered: the metal-water reactions with the zirconium components in TPBARs, and the tritium and hydrogen in the TPBARs prior to the accident. The increase in tritium production from radiolysis, which is a function of decay energy of the fission products, would be negligible for a TPC, since the burnup is not considerably different than for a conventional core operating with 18-month fuel cycles.

Criteria for combustible gas control are provided in relevant requirements of 10 CFR Sections 50.44 and 50.46, as well as GDC -5, GDC-41, GDC-42, and GDC-43. The analysis should demonstrate that a single system train is capable of maintaining the combustible gas concentrations at levels that preclude uncontrolled hydrogen-oxygen recombination.

Specifically, RG 1.7 states that the lower flammability limit for hydrogen of 4-volume percent should not be exceeded in the containment. The licensee states that, for SQN, with a total containment free volume of 1,230,000 cubic feet, a volume of 4-volume percent equates to approximately 49,200 standard cubic feet (scf) of hydrogen.

The licensees analysis identified three sources of TPBAR tritium and hydrogen inventories:

(1) zirconium-water reactions with the TPBAR getter material, which is composed of nickel-plated-Zircaloy (NPZ), (2) tritium gas released from the TPBARs, and (3) hydrogen released from 3He(n,p)T reactions inside the TPBARs. For the getter material in the TPBARs, the maximum amount of zirconium for a full TPC was considered to be oxidized. This also included the internal surface of the 12-inch burst node length of the TPBAR, since it could be postulated that TPBAR cladding could be breached following a large break LOCA (LBLOCA). Only the metal located in the active core region would be subjected to the high temperatures (in excess of 1800 EF) necessary for the zirconium-water reaction to occur. A total of 1122 scf of hydrogen was determined by the licensee to be generated from the TPBARs by the zirconium-water reaction. The licensee calculated the tritium gas contained within the TPBARs to be 357 scf, conservatively using the design limit of 1.2 grams per rod at the end of the fuel cycle for 204 TPBARs. The third source of hydrogen associated with the TPBARs is generated from the 3

He(n,p)T reaction within the rods. At the end of the fuel cycle, this reaction was found to be capable of generating up to an additional 16 scf of hydrogen. Thus, the total additional hydrogen contributed by the TPBARs for release to the containment following an LBLOCA was calculated to be 1495 scf.

22 The previous analysis for a conventional (non-TPBAR) core indicated that for an LBLOCA, with no recombiners started, the hydrogen volume within containment 1 day after event initiation would be 35,403 scf, which is equivalent to a containment hydrogen concentration of 2.88 percent. With an additional 1495 scf of hydrogen from a TPC, the containment hydrogen concentration would reach 3.0 percent at 1 day following event initiation. Figure 6.2.5-1 of the SQN UFSAR, Hydrogen Concentration in Containment after a LOCA, shows that placing one recombiner in service at 24 hours1 days <br />0.143 weeks <br />0.0329 months <br /> following event initiation limits the peak containment hydrogen concentration to approximately 3 percent. Therefore, the additional 1495 scf of total hydrogen production resulting from operation of the TPC would allow maintenance of an adequate margin between the peak containment hydrogen concentration and the lower flammability limit of 4 percent. Having up to 24 hours1 days <br />0.143 weeks <br />0.0329 months <br /> to place a recombiner train in service to maintain the containment hydrogen concentration below 4 percent is adequate in satisfying the guidance of RG 1.7.

2.6.1 Summary Based on the above evaluation, the staff concludes that, for operation of a TPC at SQN, the TPBARs will not be a significant contributor to the post-LOCA hydrogen inventory and will not have a significant impact on the total hydrogen concentration within the containment when compared to the values associated with the non-TPBAR core. The maximum containment hydrogen concentration can be maintained at less than the lower flammability limit of 4-volume percent with one recombiner train started at approximately 24 hours1 days <br />0.143 weeks <br />0.0329 months <br /> after an LBLOCA.

2.7 Interface Issue 7 - Light-Load Handling System This section deals with SQN 1 & 2's light-load handling systems (e.g., the SFP bridge crane and associated canister lifting tool) for handling TPBARs and loaded consolidation canisters.

This issue needs to be evaluated because the TPBARs weigh slightly more than the standard BPRAs that the system normally handles (65 versus 58 lb). Section 9.1.4 of the SRP provides guidance applicable to light-load handling related to refueling. The review criteria in this SRP section ensure that the light-load handling system design is reliable and that the consequences of potential light-load drops are bounded by the analyzed consequences of the design basis FHA. Light-load handling (handling loads less than the weight of one fuel assembly and its associated handling tool) encompasses TPBAR handling and consolidation because both the weight of a fuel assembly with TPBARs and the weight of a loaded consolidation canister are less than the weight of a fuel assembly with a rod cluster control assembly.

Although NUREG-0612 was written for heavy-load handling, the licensee has incorporated many of the guidelines of NUREG-0612 in its canister-handling program to demonstrate an extremely small potential for unrestrained drops of loaded canisters rather than evaluate the potential consequences of unrestrained drops of a loaded canister. In its letter dated July 19, 2002, TVA described the specific measures applicable to consolidation canister handling and its degree of conformance with NUREG-0612 guidelines.

During the consolidation process, the irradiated TPBAR assembly will be transported from its location in the SFP to the consolidation fixture located in the cask-loading pit. The consolidation process consists of releasing the TPBAR rods from the baseplate and placing these rods in a specially designed canister. Each canister is designed to accommodate a maximum of 300 TPBARs. The loaded canister is then transported to the designated SFP cell location using a canister-handling tool suspended from the SFP bridge crane.

The SFP bridge crane will be used to handle TPBAR assemblies and consolidation canisters within the pool. The weight of a loaded canister submerged in water is less than 700 lb, and a loaded canister weighs less than 1000 lb in air. Since the SFP bridge has a rated load capacity of 4000 lb, the bridge has a large structural safety factor when handling the consolidation

23 canister. An additional safety measure is provided by administrative controls requiring the use of a safety lanyard on the canister-handling tool. This lanyard limits the canister descent in the fuel pool, prevents canister tipping, and is sized to stop the canister from the maximum hook speed of 40 ft per minute. A PNNL analysis determined that no TPBAR cladding failures are expected to occur during an accidental impact with a rigid structure at that hook speed.

Therefore, the TPBARs are adequately protected from damage during routine handling.

Although lacking many features specified in NUREG-0612 for single-failure-proof cranes, the SFP bridge crane and the associated canister-lifting tool provide a reliable means of handling the TPBAR consolidation canisters. The large safety factor resulting from the low weight of the consolidation canister relative to the rated load of the SFP bridge crane assures that critical load supporting components (e.g., hoist brakes, load-block attachment point, wire rope, reeving system, and hoist drum bearings) have an extremely low probability of failure. In addition, the SFP bridge crane has some features common to single-failure-proof cranes (e.g., interlocks to prevent simultaneous motion of the bridge and hoist, redundant and diverse upper-limit switches, and a load-monitoring device). The canister-lifting tool is designed in accordance with the requirements of American National Standards Institute (ANSI) N14.6, Standard for Special Lifting Devices for Shipping Container Weighing 10,000 Pounds (4500 kg) or More for Nuclear Materials, such that it will also have increased safety factors. The tool will be tested and inspected to ensure safe operation, and the tool has an air-actuated, fail-closed safety latch to ensure the tool hook will not inadvertently disengage from the canister lifting bail.

Fuel will also be moved inside the SFP. To minimize the potential for damage to stored consolidation canisters from potential FHAs, the licensee stated that administrative controls or design features will be implemented to preclude such an occurrence. Potential administrative controls include placement of canisters in the outside row near one corner of the fuel pool, which is well away from any planned fuel movement path.

The licensee states that the light-load handling system complies with the intent of NUREG-0612 for consolidation canister handling. Since each consolidation canister holds up to 300 TPBARs, and since the proposed maximum core inventory of TPBARs is 2256 TPBARs, the staff expects the total number of consolidation canister handling evolutions per operating cycle to be small.

2.7.1 Summary Considering the expected number of consolidation canister handling evolutions and the features identified to reduce the potential of a load drop, the staff concludes that the potential for a drop of a consolidation canister is extremely small. The administrative controls applied to fuel handling ensure that the potential for damage to a consolidation canister from an FHA is also extremely small. This outcome satisfies the guidelines of Section 5.1 of NUREG-0612.

Therefore, the staff finds that the equipment and administrative control measures proposed for handling of the consolidation canisters are acceptable.

2.8 Interface Issue 8 - Station Service Water System The effect of TPBARs on a plant station service water system and the extent of the effect on available margins in the system are plant-specific. Therefore, TVA performed a specific analysis for the effect of TPBARs on the SQN 1 & 2 station service water system. In the staffs safety evaluation contained in NUREG-1672, the staff identified the need to quantitatively evaluate the effect of increased SFP heat load resulting from placement of irradiated TPBARs in the pool on affected systems, including the service water system. The design basis function of the station service water system, which is called the essential raw cooling water (ERCW) system at SQN 1 & 2, includes providing a cooling loop for heat removal from the component cooling system (CCS). The ERCW system supplies water from the ultimate heat sink (UHS) to cool the CCS. The CCS intermediate cooling loop, in turn, provides a heat sink to the SFP

24 cooling system (SFPCS) and the residual heat removal (RHR) system. The licensees evaluation of this interface item and documentation of the NRC staffs review of this evaluation follow.

The licensees analysis showed the TPC impact on core heat loads was approximately 0.5 MW, or 1.7 million British Thermal Units per hour (MBTU/hr) based on conservative, full pool SFP conditions. The estimated heat load included both the decay heat generated by freshly discharged fuel assemblies during a refueling outage and the additional residual decay heat from the increased discharge rate of fuel assemblies into the pool.

The licensee evaluated the additional heat transfer available through utilization of reduced SFP heat exchanger fouling factors and lower CCS temperature. This analysis showed that the SFP cooling heat rejection to the SFPCS could increase by approximately 9.7 MBTU/hr without exceeding SFP design temperature limits, with 1.7 MBTU/hr attributable to the TPC and 8 MBTU/hr available to allow an earlier start of core offloads. However, the latter portion of the decay heat (i.e., the 8 MBTU/hr) did not represent additional heat load on the CCS or ERCW, because it would be decay heat shifted as a result of the earlier start of core offload from the RHR system to the SFPCS, both of which were cooled by CCS and ERCW.

In its application, the licensee indicated that the net increase of 1.7 MBTU/hr, of decay heat due to TPC was well within the design basis limiting heat load imposed on the ERWS. The licensee analysis showed that the increase of 1.7 MBTU/hr of decay heat due to TPC produced an insignificant increase (less than 0.1 oF) in the ERCW temperature leaving the plant site.

2.8.1 Summary Based upon the review of the licensees analysis for the increased decay heat due to TPC and the proposed change of the offload time, a comparison of the ERCW design capacity and the increased heat load to ERCW, and the resultant temperature increase of less than 0.1 oF, the staff concludes that the ERCW system has adequate cooling capacity and margin to perform its safety and nonsafety functions with the additional heat loads imposed by TPC activities and that tritium production activities will not have an adverse impact on the ERCW heat removal capabilities.

2.9 Interface Issue 9 - Ultimate Heat Sink (UHS)

The Tennessee River is the UHS for SQN 1 & 2. The purpose of the UHS is to provide a source of cooling water for decay heat removal. The SQN 1 & 2 CCS removes heat from the SFP cooling and RHR systems and transfers it to the ERCW system. Heat in the ERCW system is then transferred to the UHS. During plant cooldown, additional heat from irradiated TPBARs in the SFP must be transferred to the UHS. NUREG-1672 stated that the UHS needs to be analyzed on a plant-specific basis because plants have various UHS designs.

In NUREG-1672, the staff identified the need to quantitatively evaluate the effect of increased SFP heat load resulting from placement of irradiated TPBARs in the pool on affected systems, including the UHS. The design basis function of the UHS is to provide an uninterrupted source of cooling water for decay heat removal.

As discussed in Section 2.4 of TVAs submittal, the net increase in decay heat associated with operation of a TPC was approximately 1.7 MBTU/hr. The additional increase in the decay heat load to the SFPCS was decay heat shifted from RHR system to the SFPCS as a result of the earlier start of core offload and does not represent a net increase in CCS heat load on the UHS.

Because the Tennessee River flow is large relative to the ERCW flow, the increased heat rejected to the river has no effect on inlet ERCW temperature. The licensees analysis showed that the increase of 1.7 MBTU/hr in decay heat due to operation of a TPC produced an

25 insignificant increase (less than 0.1 oF) in the ERCW temperature leaving the plant site.

Therefore, the additional heat loads imposed by the TPC operation will not have an adverse impact on the UHS heat removal capabilities.

2.10 Interface Issue 10 - New and Spent Fuel Storage This issue involves maintaining new and spent fuel assemblies in a safe, subcritical, coolable array during all credible storage conditions. NUREG-1672 said that applicants for TPC license amendments would need to analyze this issue because fuel rack analysis is plant-specific.

The July 19, 2002, supplemental information submittal (response to question 24 to the staffs June 6, 2002, RAI) transmitted the criticality analyses conducted by Holtec International supporting the criticality analysis associated with SFP storage of assemblies containing TPBARs (Holtec Report No. HI-2012629). A previous analysis performed by Holtec International for allowing credit for boron in the SFP contained the analysis of spent fuel storage containing no burnable absorber rods. The more recent TPBAR analysis from Holtec International also addressed the presence of other types of burnable absorbers, such as wet annular burnable absorbers (WABAs), integral burnable absorbers (IFBAs) rods, gadolinia rods, BPRAs, and burnup. The Holtec report contains the criticality analyses accounting for the presence of the TPBARs and the burnable absorber rods in fuel assemblies, while maintaining Keff less than or equal to 0.95 under normal and abnormal conditions. The TPBARs are removed from the once-burned (or in some cases, twice-burned) fuel assemblies shortly after the assemblies are placed in the SFP.

2.10.1 Criticality Calculations Associated with Assemblies Containing TPBARs NRC SRP Sections 9.1.1, 9.1.2, and NRC regulations such as GDC-62, provide guidelines applicable to the storage of new and spent fuel, respectively. Conformance with the SRP ensures that the design and arrangement of the new and spent fuel storage racks will maintain a subcritical array during all storage conditions. At SQN 1 & 2, new fuel assemblies (including assemblies with TPBARs, WABAs, and IFBAs) may be stored in the new fuel storage vault prior to insertion into the reactor core. New fuel assemblies containing TPBARs will have a lower reactivity than unpoisoned fresh fuel assemblies. The existing criticality analysis for the new fuel storage configuration will remain valid for new fuel assemblies with TPBARs because of their lower reactivity. The current criticality calculation did not credit any discrete or integral poisons in the fuel. This licensees analysis, namely, Holtec International Report HI-2012629, supplements the current SQN SFP analysis by including an additional analysis accounting for the presence of the TPBARs and other burnable poisons.

The current licensed SFP storage capacity at SQN is 2091 spent fuel assemblies. There are currently 244 spent fuel assemblies in the SFP. As discussed in the licensees submittals, the licensee has requested deletion of current TS Sections 3/4.7.14 and 5.6 that address spent fuel storage provisions for the cask pit pool. Similar to the new fuel, the spent fuel racks have been analyzed for fuel assemblies up to 235U 5.0-weight percent. For conservatism, the spent fuel was assumed to operate with TPBARs or WABAs, which were removed before the fuel was put into the spent fuel racks. Credit was taken for fuel burnup and IFBAs where appropriate.

2.10.2 Methodology The Holtec International analysis used the most reactive design and the most reactive temperature to set the storage requirements. The analyses included means to account for the bias and uncertainty associated with the benchmarking of the methodology, a bias for the underprediction of reactivity due to boron particle self-shielding, and the uncertainty due to mechanical tolerances from the manufacturing process. The licensee also included additional uncertainties related to irradiated fuel as described in the burnup credit methodology discussed

26 in the Holtec report. The licensee calculated these uncertainties at the 95/95 probability-confidence level, using procedures described in the regulatory guidance of an NRC memorandum dated August 19, 1998.

2.10.3 Normal Conditions The licensee performed depletion calculations using the CASMO-4 computer code to evaluate the reactivity of the fuel assemblies currently in use or anticipated for storage in the SQN spent fuel racks. Criticality calculations were performed using two different assembly designs: the Westinghouse 17x17 fuel assembly and the Framatome Cogima Fuel (FCF) Alliance fuel.

Burnable poison loadings (including TPBARs) were explicitly included in the modeling.

SFP calculations are usually performed with the highest reactivity worth fuel design. For this analysis, Holtec International had determined that the Westinghouse 17x17 fuel design is more bounding than the FCF fuel design. CASMO-4 calculations were carried out to account for uncertainties associated with densities, enrichments, mechanical tolerances, and TPBAR loadings.

2.10.4 Abnormal or Accident Conditions and Soluble Boron Requirements Although the NRC permits credit for the soluble boron (poison) normally present in the SFP water, most abnormal or accident conditions will meet the limiting reactivity effective neutron multiplication factor Keff less than or equal to 0.95, even in the absence of soluble poison. The licensee analyzed all postulated accidents (dropped fuel assembly, water temperature and density effects, eccentric positioning of a fuel assembly within the rack, abnormal placement and misplacement of fresh fuel assembly, etc.) for this amendment request. The licensee analyzed its spent fuel storage racks by taking into account boron credit in accordance with the methodology described in the previously approved Holtec report. This methodology ensures that Keff is less than or equal to 0.95 as recommended in ANSI/American Nuclear Society (ANS) 57-1983 and NRC guidance, such as 10 CFR Part 50.68.

The analysis contained in the Holtec report of this submittal, shows that the most serious postulated condition is the misplacement of a fresh fuel assembly without Gadolinia rods in the location of a water cell in Region 3 of the SFP. In this case the misplaced assembly has the potential for exceeding the limiting reactivity, Keff less than or equal to 0.95, but is always less than 1.0, should there be a concurrent and independent accident where all the boron had been lost in the SFP. To reduce Keff to less than or equal to 0.95, the addition of 700 ppm of soluble boron is required, bringing Keff in the storage rack to below or equal to the regulatory requirement of 0.95, including bias and uncertainties.

2.10.5 Criticality Analyses Results The licensee evaluated four different storage configurations (SFP Regions 1a, 1b, 2, and 3) of fresh and spent fuel assemblies in the SQN 1 & 2 SFP analyses. The analyses showed that all four storage configurations meet the NRC regulatory requirement of Keff equal to or less than 0.95. The results of all four configurations were tabulated and provided in Holtec International report HI-2012629 (submitted with the July 19, 2002, TVA response to the NRCs RAI). The analysis results provided in the July 2002 submittal were found to be comparable to those of the previous analysis without TPBARs. No additional restrictions on the interfaces between regions for fuel types using TPBARs and regions using fuel without TPBARs were found necessary. All critical analyses inherently include axial distribution effects where appropriate. A summary of the conditions and conclusions is provided in the Holtec International report of the July 2002 submittal.

27 2.10.6 Heat Load Analysis The licensee evaluated the heat production from a fully loaded canister and its potential effect on the spent fuel racks. A TPBAR rod will produce approximately about 3 watts of heat 30 days after reactor shutdown. Assuming a fully loaded canister contains a maximum of 300 TPBAR rods, this translates into a maximum heat load of 900 watts per canister. This heat load is small, considering that adequate circulation is provided through the open-topped canister and the drainage/cooling holes located on the sides and bottom of the canisters. Therefore, the staff concludes that this configuration will provide adequate natural circulation cooling to preclude adverse effects on the fuel storage racks.

2.10.7 Summary The NRC staff reviewed and evaluated the licensees analysis pertaining to the inclusion of TPBARs and other burnable absorbers into its spent fuel analyses, and has determined that the analyses were conducted in accordance with the requirements of GDC-62 and other regulatory documents and, thus, are acceptable. The analysis for the abnormal and accident conditions is also acceptable since it satisfies the requirement that Keff be no more than 0.95, as prescribed in the above regulatory requirements.

The licensee proposed to modify Section TS 5.6.1.1 to accommodate new provisions that address the storage of TPBARs. Clarifying information regarding the storage of TPBARs in all core regions has been included. This revision will also delete current TS Section 5.6.1.1.d for each unit, which addresses spent fuel storage provisions for the cask pit pool. Because these revisions will not impact the safety of the SFP operation, the staff finds these requests acceptable.

2.11 Interface Issue 11 - Spent Fuel Pool Cooling and Cleanup System (SFPCCS)

The SFPCCS cools the spent fuel during all operating and accident conditions. NUREG-1672 documented the staffs review of the effect of TPBARs on this system for a typical Westinghouse plant. However, TVA was required to perform an analysis for SQN 1 & 2 to determine SFP temperatures with TPBARs installed because the capacity of the SFP and associated cooling system design varies among plants. Specifically, in the safety evaluation documented in NUREG-1672, the staff identified the need to quantitatively evaluate the effect of increased SFP heat load resulting from placement of irradiated TPBARs in the pool on affected systems, including the SFPCS. As a part of its submittal for a license amendment to implement the TPC at SQN, the licensee proposed a change in the SFP cooling analysis methodology that would increase the maximum allowable SFP decay heat load by taking credit for lower CCS water temperatures and lower SFP heat exchanger fouling factors. The licensee proposed this change so that it would have the capability to offload the core during outages as early as 100 hours4.167 days <br />0.595 weeks <br />0.137 months <br /> after shutdown and compensate for the projected increase in SFP decay heat from tritium production activities.

2.11.1 The Capability To Cool the Spent Fuel Pool The SFPCS at SQN 1 & 2 is designed to remove the decay heat generated by stored spent fuel assemblies. The system consists of two seismic Category I cooling trains, each equipped with one heat exchanger and one pump. A third pump is provided to serve as a backup to the pump in either train. Heat is removed from the SFP heat exchangers by the CCS.

The licensees current analysis established 45.3 MBTU/hr as the bounding heat load for planned refueling outages and 183 oF as the maximum bulk water temperature in the SFP with only one train of SFPCS operating. The current approach is to perform outage-specific decay heat analyses for each outage to determine the acceptable point in time that the core offloading

28 activities may commence without exceeding the design basis maximum allowable heat load (45.3 MBTU/hr). This method ensures that the maximum SFP water temperature will not exceed 183 oF.

In its amendment request, the licensee described a change in the SFP cooling analysis methodology that would increase the maximum allowable SFP decay heat load up to a maximum of 55 MBTU/hr, an increase of 9.7 MBTU/hr, by taking credit for lower CCS water temperatures and lower SFP heat exchanger fouling factors. The licensees current analysis utilized design fouling factors of 0.0005 hr-ft2-oF/BTU for both the tube and the shell side fouling, and design maximum CCS temperatures of 95 oF for the cooling water on the shell side of the SFP heat exchangers. However, based on more than 20 years of data from SQN, the licensee stated that the actual fouling had been found to be considerably less than design without cleaning. It was also noted that the CCS maximum design temperature of 95 oF was very conservative relative to the actual amount of heat being rejected to the CCS. The design basis for the CCS included significantly higher decay heat loads based on RHR system heat loads shortly after shutdown. However, by the time the core is completely offloaded, the decreasing decay heat from the fuel has been transferred to the SFPCS and the RHR heat load is essentially zero. By increasing the flow of the essential raw cooling water (ERCW) to the CCS heat exchanger, the CCS temperature can be decreased to values less than 95 oF.

The proposed alternative method used the additional heat removal capacity that could be obtained from the lower fouling factors and CCS temperatures, but utilized the same basic methodology, equations, and data (other than fouling factors and CCS temperatures) as the current analysis. The alternative methodology varied both the SFP heat exchanger fouling and SFP heat exchanger coolant (CCS) temperature to calculate thermal balances on the SFP.

The results showed that a single train of the SFPCS could remove 55 MBTU/hr of decay heat from the SFP under the following conditions: (1) an SFPCS heat exchanger shell and tube side fouling factor of 0.0001; (2) a CCS temperature of 80 oF, and (3) an SFP temperature of 183 oF.

Under otherwise identical conditions, operation of both trains of the SFPCS could remove the same decay heat with the SFP at 144 oF.

The staff concludes that the proposed alternative methodology for calculating the maximum SFPCS heat removal is acceptable since it utilizes the same basic methodology, equations, and data as the current analysis, and thus is essentially equivalent to the current method and maintains the currently established maximum temperature of the SFP water. The proposed alternative methodology incorporates the use of actual, rather than conservative, values for SFPCS heat exchanger fouling factors and CCS temperatures.

2.11.2 Increase in SFP Heat Load Based on the alternative analyses, the licensee proposed to increase the maximum allowable decay heat in the SQN SFP from 45.3 MBTU/hr to a range between 45.3 MBTU/hr and 55 MBTU/hr. Of the additional heat removal capability, up to 8 MBTU/hr would be used to allow SQN to begin its full core offload operation as early as 100 hours4.167 days <br />0.595 weeks <br />0.137 months <br /> after shutdown, and the remaining 1.7 MBTU/hr would provide an allowance for increased decay heat from the tritium production activities. The licensee stated that exceeding the lower design value of 45.3 MBTU/hr would only be permitted with actual fouling of the SFPCS heat exchanger, and by taking credit for actual CCS temperature. Since the analyses (current and alternative) are primarily an overall system heat balance, the source of the decay heat is not important. The SQN SFPCS can accommodate tritium production operations by adjusting the beginning of the core offloading until the design allowable heat load can accommodate core and residual decay heat. Therefore, the staff concludes that the SFPCS can accommodate the additional decay heat load imposed by commencing the core offload as early as 100 hours4.167 days <br />0.595 weeks <br />0.137 months <br /> and compensate for the projected increase in SFP decay heat from tritium production activities.

29 In its supplement dated July 19, 2002, TVA stated that administrative procedures require defense-in-depth for the decay heat removal function performed by the SFPCS. Defense-in-depth is provided by operating both SFPCS trains during planned refueling offloads and supporting their operation with cooling water supplied from the operating unit. The provision of a backup pump capable of operation in either train provides additional assurance that two trains of SFP cooling will be available. For the increased heat load, the existing cooling system satisfies the requirements of GDC-61 of 10 CFR Part 50, Appendix A, with respect to provision of a residual heat removal capability having reliability that reflects the importance to safety of decay heat and other residual heat removal.

2.11.3 Effect of SFP Boiling The licensee also provided the results of analyses to evaluate the minimum time to SFP boiling and the maximum rate of boiloff in the unlikely event that there is a complete and sustained loss of cooling. The analyses assumed the maximum decay heat load of 55 MBTU/hr with one cooling train in operation and employed conservative, worst-case assumptions. The minimum time to SFP boiling decreased from 2.64 hours2.667 days <br />0.381 weeks <br />0.0877 months <br /> for the current method to 1.14 hours0.583 days <br />0.0833 weeks <br />0.0192 months <br /> for the alternative analysis, and the maximum boiloff rate increased from 103 gpm to 118.2 gpm. The SQN UFSAR identified multiple sources of makeup water that can be aligned to the SFP in less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. These include the refueling water storage tank (RWST) via the refueling purification pumps, the demineralized water system, and the fire protection system. All piping, valves, and pumps from the RWST to the SFP cooling loop are seismically qualified. The licensees analysis of loss of forced cooling found that available makeup water sources are capable of maintaining an adequate cooling water inventory. The staff concludes that, in the unlikely event that there is a complete loss of forced cooling, cooling the SFP at SQN by adding makeup water conforms with the guidance described in the SRP Section 9.1.3, and operation of the TPC does not adversely affect the ability to maintain an adequate coolant inventory in the SFP under accident conditions.

2.11.4 Summary The staff concludes that the increase in SFP heat load is acceptable based on SFP temperature remaining within the current licensing basis values, provision of an SFP cooling system having reliability consistent with the importance of decay heat removal at high heat loads, and provision of a makeup system capable of maintaining SFP water level under accident conditions. These capabilities are consistent with the cooling requirements of 10 CFR Part 50, Appendix A, GDC-61.

2.12 Interface Issue 12 - Component Cooling Water System (CCS)

As mentioned in Interface Issue 9, the SQN 1 & 2 CCS removes heat from the SFP cooling and residual heat removal systems and transfers it to the ERCW system. NUREG-1672 said that irradiated TPBARs place little additional heat load on the SFP cooling system. However, the additional heat load from full-core offloads associated with the TPC could increase the demands on the CCS. Thus, TVA was required to analyze the effect of full-core offloads on the CCS at WBN 1.

The design basis function of the CCS includes providing an intermediate cooling loop for heat removal from several safety-related heat exchangers and several non-safety-related components. Two of the highest heat loads placed on the CCS are the SFPCS and the RHR system. These two heat removal systems are the primary means for cooling the plant and removing residual decay heat during the later stages of plant cooldown and during outages.

The licensees analysis showed the TPC impact on core heat load was approximately 0.5 MW, or 1.7 MBTU/hr, based on conservative, full SFP conditions. The estimated heat load included

30 both the decay heat generated by freshly discharged fuel assemblies during a refueling outage and the additional residual decay heat from the earlier discharge of fuel assemblies into the pool. The decay heat load increase resulting from operation of the TPC is less than 2 percent of the total design heat load on the CCS. The remaining increase in decay heat is merely a transfer of heat load from the RHR system to the SFPCS.

The effect of these increased decay heat loads on the CCS has been evaluated in conjunction with the analysis of the SFPCS. The increased heat loads are allowed only when lower CCS temperatures can support the increased heat loads.

2.12.1 Summary Based upon the review of the licensees analysis for the increased decay heat load, the staff finds that the CCS has adequate cooling capacity and margin to perform its safety and nonsafety functions with the additional heat loads imposed by tritium production activities and that tritium production activities will not have an adverse impact on the CCS heat removal capabilities.

2.13 Interface Issue 13 - Demineralized Water Makeup System (DWMS)

The DWMS of a typical Westinghouse plant is nonsafety-related and supplies high-purity makeup water to other plant systems. Having TPBARs in the core will increase tritium levels in the RCS. TVA may have to use the DWMS to replace some of the water in the RCS to dilute the tritium. NUREG-1672 concluded that a licensee must analyze the plant-specific capability of the DWMS since this system differs from plant to plant. The NRCs review of TVAs analysis for SQN 1 & 2 follows.

The licensee performed an evaluation of the impact on the DWMS from the increased amount of tritium expected to be released into the RCS. To maintain tritium levels within the RCS at current levels, additional feed and bleed operations may be needed, which would require additional demineralized water as a makeup supply.

The licensee determined that the plants current feed and bleed operations are adequate to handle the potential maximum tritium levels of 9 µCi/g with the TPBARs. There is no NRC regulatory limit on RCS tritium level. The licensee performed an evaluation which determined that the existing demineralized water makeup system has a capacity to produce demineralized makeup water at a nominal rate of 175 gpm. Additionally, there are 500,000 gallons of demineralized water stored in tanks. The licensees evaluation concluded that the demineralized water makeup system capacity is adequate for plant operation with the TPBARs.

2.13.1 Summary Based on the above-stated tank capacities and assumptions, the staff concludes that the SQN 1 & 2 licensee has demonstrated that sufficient storage and water makeup capacity will be available to adequately meet any increased feed and bleed demands from normal and abnormal TPC events. The worst-case need for makeup water (i.e., 150,000 gallons in the event of two failed TPBARs) is far less than the capability of the DWMS to provide water. The staff agrees with the licensees evaluation that it has adequate demineralized water storage and production capacity to meet the expected increase in feed and bleed requirements during routine operation as well as abnormal operation (a two-TPBAR failure event).

2.14 Interface Issue 14 - Liquid Waste Management System This issue concerns the impact of increased tritium levels in the RCS from TPBARs on the liquid waste management system. Larger volumes of primary coolant need to be processed

31 and discharged through the liquid waste management system to remove tritium buildup in the RCS. In NUREG-1672, the NRC staff said that a licensee needs to analyze plant-specific effluent concentrations, dose limits, dilution flow rates, and the tritium monitoring programs for a plant operating with TPBARs. NRCs review of these SQN 1 & 2-specific analyses continues below.

Primary coolant discharge volumes have the potential to rise with TPC operations if increased feed and bleed is required. As stated in Interface Issue 13, operation with a TPC may increase RCS tritium levels to 9.0 µCi/gm, due to normal reactor tritium production plus tritium permeation from the TPBARs. TVA must assure the adequacy of SQN 1 & 2s storage capacity for potential increased discharge volumes. To accommodate this liquid waste resulting from RCS feed and bleed operations, SQN has two shared chemical and volume control system holdup tanks, each with a capacity of 126,000 gallons. Other tanks that may be used to hold and process liquid waste for release to the environment include a liquid waste monitor tank with a capacity of 21,600 gallons, the tritiated drain collector tank, which has a capacity of 24,700 gallons, and a cask decon collector tank, which has a capacity of 15,000 gallons. Plant conditions would dictate the length of temporary storage time required, which could vary from several days to several weeks. These tanks process liquid from the two holdup tanks and dilute it with cooling tower blowdown flow before releasing that liquid to the environment.

The effective operation of the liquid waste management systems is dependent on the amount of liquid waste that is sent to the systems. The licensee performed an evaluation of the major sources of liquid waste to the systems and determined that for normal TPBAR operation the normal RCS feed and bleed operation for boron control will be maintained throughout the plants operating cycle. Therefore, anticipated coolant discharges with the TPC will be comparable to the current plant discharges.

For its evaluation, the licensee collected site-specific data during the first WBN 1 operating cycle with the LTAs to estimate the impact on station radiological conditions from the expected increased amounts of tritium. The RCS maximum recorded tritium levels were approximately 2.5 µCi/g with a cycle RCS tritium mean value of approximately 1.0 µCi/g. To assess the potential impact on plant workers from the tritium, the licensee calculated the peak tritium derived air concentration (DAC) fractions that could be expected inside the containment building. The DACs are regulatory radiation values, contained in 10 CFR Part 20, for airborne radioactive material and are used to assess the dose to a radiation worker. The licensee calculated that with no modifications to the current boron control feed and bleed methodologies, the design basis RCS maximum tritium values will be approximately 9 µCi/g with a cycle mean of approximately 3.6 µCi/g. These values translate into an estimated containment peak tritium DAC fraction of approximately 0.6 and an average containment tritium DAC fraction of about 0.3. The licensee estimated that the containment average tritium DAC fraction equates to an effective dose rate of about 0.7 mrem/hr. The addition of this dose rate will not have a significant impact on the licensees ability to control worker radiation doses using the controls and practices in the existing Radiation Protection Program.

Population doses from radioactive liquid and gaseous effluent releases associated with both TPC normal and abnormal operation (failure of two TPBARs under irradiation and the associated inventory of tritium that is assumed to be released to the primary coolant) will remain below applicable Offsite Dose Calculation Manual (ODCM) limits, and tritium release concentrations will remain within 10 CFR Part 20 and ODCM limits.

In addition, the licensee reviewed its existing radioactivity monitoring programs for outdoor liquid storage tanks and verified that a postulated accident involving the uncontrolled release of the tanks contents will not exceed the 10 CFR Part 20 limits at the nearest potable water supply or the nearest surface water supply.

32 2.14.1 Summary The staff has reviewed TVAs analyses regarding this interface issue and finds no errors or inappropriate assumptions. The liquid waste management systems meet NRC SRP guidance and the TPP will not significantly change the ability of these systems to properly dispose of radioactive effluents. Waste liquid storage tank capacity far exceeds the volumes expected to result from the TPP, even under abnormal conditions. Based on the above, the staff agrees with the licensees evaluations and finds that the licensees assessment of the potential radiological impact on plant workers, members of the public, and the environment from operation with the TPC is acceptable to ensure that compliance with all regulatory dose limits will be maintained. This capability satisfies the guidance of Section 11.2 of the NRC SRP and is acceptable.

2.15 Interface Issue 15 - Process and Effluent Radiological Monitoring and Sampling System NUREG-1672 said that the NRC staff must review the details of laboratory instrumentation and sampling frequencies and locations for process and effluent monitoring at plants with TPBARs.

The acceptance requirements for the process and effluent radiological monitoring and sampling system are based on meeting the following regulations:

• 10 CFR 20.106, as it relates to radioactivity in effluents to unrestricted areas

• GDC-60, as it relates to waste management design and the control of releases of radioactive materials to the environment

• GDC-63 and GDC-64, as they relate to the radioactive waste system design for monitoring radiation levels and leakage The licensee evaluated its current process and effluent radiological monitoring instrumentation and sampling system capability and determined that modifications would be required for a TPC.

The exhaust will be continuously sampled during the period of release instead of the previous grab samples. The licensee determined that its current techniques for tritium air sampling, liquid monitoring, and liquid scintillation counting are adequate and will not be changed.

Other procedural modifications include enhanced RCS tritium sampling and SFP liquid sampling. Currently, for a non-TPC, weekly sampling of RCS tritium concentration is performed. For TPC operations, RCS tritium concentration sampling will be performed three times a week for concentrations below 9 µCi/gm and daily for concentrations between 9 µCi/gm and 15 µCi/gm, and for concentrations above 15 µCi/gm mitigative actions will be implemented, which will involve expanded tritium monitoring. While irradiated TPBARs are stored in the SFP, liquid tritium sampling will be collected on a weekly frequency. In addition, during movement of TPBARs, liquid sampling of the SFP will be performed daily. Sampling of the waste gas decay tanks for tritium will also be performed prior to any planned release when TPBAR irradiation is performed.

2.15.1 Summary The staff finds that the licensees change to continuous tritium air sampling for the Auxiliary and Shield buildings is adequate to ensure that tritium emissions are adequately monitored, assessed, and reported in accordance with license conditions and regulatory requirements.

Effluent releases from TPC operation for both normal and abnormal (two TPBAR failures) conditions are not expected to result in exceeding 10 CFR Part 20 or ODCM limits.

33 2.16 Interface Issue 16 - Use of LOCTA_JR Code for Loss-of-Coolant Accident Analyses The LOCTA_JR computer code is used to evaluate the thermal response of TPBARs to a LOCA. This code performs one-dimensional radial heat conduction calculations for a fuel rod.

NUREG-1672 states that the NRC staff had not yet approved this code for licensing applications. The purpose of Interface Issue 16 is to document the staffs subsequent approval of this code.

On June 23, 2000, TVA submitted Westinghouse Topical Report WCAP-15409, Description of the Westinghouse LOCTA_JR 1-D Heat Conduction Code for LOCA Analysis of Fuel Rods, to the NRC for review and approval. The NRC staff completed its review of WCAP-15409 and concluded in a January 17, 2001, letter to TVA (Accession No. ML010170152) that the report was acceptable for referencing in licensing analyses.

2.17 Interface Issue 17 - Anticipated Transient Without Scram (ATWS) Analysis An ATWS is an anticipated operational occurrence requiring a reactor scram (automatic insertion of control rods) that does not occur due to a common-mode fault in the reactor protection system. NUREG-1672 states that the TPP applicant must submit a plant-specific ATWS analysis to the NRC. The NRC reviewed the ATWS analysis for SQN 1 & 2, described in TVAs letter of September 29, 2000. The NRCs letter of March 16, 2001 (Accession No. ML010750049), concluded that using TPBAR cores in SQN 1 & 2 conformed to the ATWS rule as specified in 10 CFR 50.62(c)(1) for Westinghouse plants and was, therefore, acceptable.

3.0 EVALUATION OF TECHNICAL SPECIFICATION CHANGES This section documents the NRC staffs evaluation of the licensees proposed TS changes for SQN 1 & 2 listed in the Introduction section of this Safety Evaluation. Each proposed TS change and the basis for the staffs approval are addressed in this section.

3.1 Technical Specifications Changes 3.1.1 TS Table 3.3 Remote Shutdown Monitoring Instrumentation - Revised Backup Source Range Monitor Measurement Range This proposed change will revise the measurement range of the backup source range monitor (SRM). The current range is 1 to 106 counts per second and the proposed range is 0.1 to 105 counts per second. The current measurement range for the backup SRM provides an acceptable range of values for the current fuel loading configurations and the typical boration levels of the RCS. With the higher levels of boron concentrations that will be utilized with the TPCs, the availability of neutrons to be detected by the backup source range monitor will be reduced. This proposed change improves the ability to monitor neutron activity for verification of shutdown conditions, which is the primary function of this monitor.

The backup SRM provides an indication of core criticality conditions in the auxiliary control room. This monitor would be used if the main control room was required to be evacuated and shutdown conditions needed to be monitored from a remote location. This monitor is used for indication of the core shutdown conditions and does not include the trip functions associated with the main control room monitors that support plant startup functions. Based on this, the staff agrees that lowering the backup SRM measurement range by one decade will ensure adequate neutron monitoring indication under the proposed TPBAR conditions and will not adversely impact nuclear safety. The staff finds this change to be acceptable.

34 3.1.2 TS 3/4.5.1 - Cold Leg Injection Accumulators - Boron Concentration Increase TS 3/4.5.5 - Refuel Water Storage Tank - Boron Concentration Increase The post-LOCA long-term core cooling analysis for SQN 1 & 2 requires maintaining a subcritical boron concentration following a LOCA after all boration sources are injected and mixed in the containment sump, without taking credit for any rod cluster control assembly (RCCA) insertion. These boration sources include the CLAs, the RWST, and the melted ice from the ice condenser containment.

The proposed use of TPBARs at SQN introduces an additional neutron poison into the reactor core. When large amounts of excess neutron poison are added to a core, such as with TPBARs, there is competition for neutrons from all the poisons and the negative worth of each poison (including the RCS boron) decreases. Following a LOCA, the positive reactivity insertion due to the negative moderator coefficient that occurs during the cooldown from hot full power to cold conditions must be entirely overcome by RCS boron. Because the RCS boron will be worth less with a TPBAR core, a higher concentration is needed to maintain subcriticality. The ice (at approximately 2000 parts per million [ppm]) is a dilution source which has to be overcome by the CLAs and the RWST concentrations in order to prevent criticality. Therefore, the licensee is proposing to increase the CLA boron concentration from the present range of 2400 to 2700 ppm to a range of 3500 to 3800 ppm, and the RWST boron concentration from the present range of 2500 to 2700 ppm to a range of 3600 to 3800 ppm. To verify that these increased boron concentration ranges are adequate to maintain subcriticality following a LOCA, Framatome-ANP performed analyses using the NRC-approved core simulator, NEMO. This analysis conservatively assumed failures of TPBARs and various adverse reactivity conditions.

The licensee evaluated the impacts of the proposed increase in CLA and RWST boron concentration ranges on the LOCA analyses. For both the LBLOCA and small break (SBLOCA), the licensee stated that these analyses do not explicitly model the boron concentration levels present in the CLAs or RWST. However, although not modeled in the analyses, any additional boron injected due to the increased concentration levels would increase the margin by which the core is maintained subcritical. The PCT and clad oxidation are not functions of the boron concentration. Therefore, the increased levels of CLA and RWST boron concentration will not adversely impact the results of the LBLOCA and SBLOCA.

With respect to post-LOCA long-term core cooling requirements, the licensee provided a summary of the hot leg switchover (HLSO) time evaluation model it used to establish that boric acid will not precipitate in the long term following certain loss-of-coolant accidents. The model is consistent with the traditional 1975 model used by licensees operating Westinghouse-designed nuclear steam supply systems. The licensee additionally provided a discussion of conservatisms and nonconservatisms associated with the model. Predicted times available for initiation of hot leg injection included the following:

Case Case Description HLSO Time, hours 1 Traditional analysis with no allowance for boric acid saturation concentration uncertainty 7.25 2 Traditional analysis with allowance for boric acid saturation concentration uncertainty 5.59 3 Case 1 with Appendix K decay heat generation rate assumption 5.35 4 Case 2 with Appendix K decay heat generation rate assumption 4.15

35 Based on this information, the licensee stated that the SQN emergency operating procedures will be revised to require initiation of hot leg ECCS [emergency core cooling system]

recirculation 3 hours0.125 days <br />0.0179 weeks <br />0.00411 months <br /> following a large break LOCA for the tritium production core rather than 5.5 hours0.208 days <br />0.0298 weeks <br />0.00685 months <br />. The 3-hour switchover time requirement does not increase operator burden during LOCA mitigation and recovery and will provide an added measure of conservatism with respect to the tritium production core long-term cooling analysis.

The licensee also evaluated the impacts of the increased CLA and RWST boron concentrations on non-LOCA transients. The CLAs do not inject for any of the SQN non-LOCA transients and therefore the higher CLA boron concentration will have no impact on any of the UFSAR Chapter 15 non-LOCA transients. The following non-LOCA accidents model the RWST boron concentration:

SLB at hot zero power feedwater line break spurious operation of the safety injection system at power mass and energy releases steam generator tube rupture containment mass and energy releases The licensee concluded that the results of these accidents are not impacted by the increased boron concentrations in the RWST. The feedwater line break, SGTR and containment mass and energy releases conservatively do not credit the CLA or RWST boron concentrations and are therefore not impacted. For the SLB at hot zero power (HZP), dryout of the broken steam generator and a subsequent reduction in RCS cooling ends the core power excursion prior to the introduction of boron into the RCS. The SLB mass and energy release evaluation relies on control rods for shutdown margin and assumes a minimum boron concentration. The spurious operation of the SI system at power analysis is postulated to maximize the insertion of negative reactivity and assumes a maximum boron concentration. The SQN UFSAR analysis of record assumes a boron concentration of 20,000 ppm, which conservatively bounds the proposed RWST boron concentration of 3800 ppm.

The licensee also addressed the issue of boron precipitation from solution at the proposed increased concentrations and concluded that this is not credible at the minimum RWST temperature of 60 EF and minimum CLA temperature of 70 EF. The minimum acceptable temperature associated with the proposed RWST and CLA boron concentrations is near the freezing point (32 EF).

Based on this discussion, the NRC staff concludes that the increased levels of boron concentration in the RWST and CLAs are acceptable.

3.1.3 TS3/4.7.14 and Bases - Cask Pit Pool Minimum Boron Concentration - Deletion of Requirements The licensee is proposing to delete this TS because it plans on using the cask pit pool for TPBAR consolidation activities. SQN has never stored spent fuel in the cask pit and has no plans to do so in the future. This TS only addresses the potential for storage of spent fuel in the cask pit pool, and eliminating this TS will not have any adverse impact because the storage function was never utilized and a specific boron concentration is not required. This TS requires a minimum boron concentration of 2000 ppm, only if spent fuel is stored in the cask pit pool. In accordance with the TS requirements for the SFP, the licensee will continue to maintain proper boron concentration for the storage of spent fuel in the SFP to control inadvertent criticality events. If TVA chooses to utilize the cask pit pool for spent fuel storage in the future, the appropriate analysis, along with a license amendment request, will be required.

36 The NRC staff approves the removal of this TS for SQN 1 & 2.

3.1.4 TS 5.3.1 - Design Features/Reactor Core/Fuel Assemblies The licensee is requesting a change to TS Section 5.0, Design Features, to allow the insertion of a maximum of 2256 TPBARs into the SQN reactor core for irradiation purposes. The specific number of TPBARs to be irradiated during a given cycle will be identified in the COLR, but will always be less than or equal to 2256 TPBARs. To support the proposed change, the licensee performed extensive analysis, testing, and evaluation of the TPBARs as reported previously in the DOE TPC topical report. Additionally, Framatome-ANP performed SQN-specific evaluations of the various impacts of the TPBARs on the SQN reactor cores. The Framatome-ANP evaluation topical report includes the confirmatory checks recommended by the DOE TPC topical report, and plant-specific evaluations requested by the NRC staff in the TPC SER (NUREG-1672).

3.1.5 TS 5.6 and the Basis for TS 3/4 7-13 SFP Storage Requirements See Section 2.10 3.1.6 TPBAR Consolidation Activities See Section 2.1 3.1.7 Technical Specification Bases 3/4 6.4 - Hydrogen Recombiners See Section 2.6 3.2 Conclusions The staff has completed its review of the proposed SQN TS changes associated with irradiation of TPBARs in SQN 1 & 2. Based on the discussions in this Safety Evaluation, the staff has determined that for the proposed TS changes, all applicable acceptance criteria will be satisfied and, therefore, the proposed TS changes are acceptable. The associated TS changes include revising the measurement range for the source range monitors in TS Table 3.3-9, increasing the required boron concentration for both the CLAs (TS 3/4.5.1) and the refueling water storage tank (TS 3/4.5.5), deleting the boron concentration and spent fuel storage requirements and associated Bases for the cask pit pool in TS Section 3/4.7.14 and Section 5.6, adding a limit on the number of TPBARs that can be irradiated in TS Section 5.3, providing storage requirements for spent fuel assemblies that contained TPBARs during irradiation in TS Section 5.6 and the Bases for TS Section 3/4.7.13, and revising Combustible Gas Control - Hydrogen Generation Sources in TS Bases 3/4.6.4. Additional background information considered by the staff in its findings is documented in Appendix B to this safety evaluation, which provides information regarding the nexus between the licensees submittal (and supplements) and the NRC SRP.

4.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION

DETERMINATION The Commissions regulations in 10 CFR 50.92 state that the Commission may make a final determination that a license amendment involves no significant hazards consideration if the operation of the facility, in accordance with the amendment, would not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

As required by 10 CFR 50.91(a), the licensee provided its analysis of the issue of no significant hazards consideration in its September 21, 2001, amendment request. The staff reviewed the

37 licensees analysis and, based on its review, it appeared that the three standards of 10 CFR 50.92(c) were satisfied. Therefore, the NRC staff proposed to determine that the amendment request involves no significant hazards consideration, and published its proposed determination in the Federal Register for public comment on December 17, 2001 (66 FR 65000).

The staff has completed its evaluation of the licensees proposed amendment as discussed in Sections 2.0 and 3.0 above. Based on its evaluation, the staff has determined that the proposed amendment does not significantly increase the probability or consequences of an accident previously evaluated; does not create the possibility of a new or different kind of accident from any accident previously evaluated; or does not involve a significant reduction in a margin of safety. The evaluation below, in relation to the three standards of 10 CFR 50.92, supports the staffs final no significant hazards consideration determination.

First StandardThe amendment does not significantly increase the probability or consequences of an accident previously evaluated.

TVA and the NRC staff considered the impact of operating SQN 1 & 2 with TPBARs on the following previously analyzed DBAs:

• loss of offsite power

• waste gas decay tank failure

• loss-of-coolant accident

• main steamline break

• steam generator tube rupture

• fuel-handling accident

• failure of small lines carrying primary coolant outside containment

• rod ejection accident Operating SQN 1 & 2 with TPBARs does not significantly increase the probability that any of the above accidents can occur. The proposed changes do not influence the probabilities of the above accidents. A nuclear reactor core usually has standard BPRAs inserted into many of the fuel assemblies. A TPBAR is simply another type of BPRA. A TPBAR functions in a very similar manner to a BPRA, and there is no uranium or plutonium in the TPBARs. SQN 1 & 2 plant operators would install these TPBARs in fuel assemblies where standard BPRAs are normally placed.

Regarding the FHA, TVA will use the same methods, procedures, and equipment to handle TPBARs as it does to safely handle fuel assemblies that do not contain TPBARs. TPBARs are nearly identical in design to BPRAs, and weigh about the same (58 versus 65 lbs, respectively).

The staff evaluated TVAs provisions for handling TPBARs and concluded that the equipment and administrative controls that TVA will apply to handling TPBAR assemblies and transport casks provide reasonable assurance of safety. The combination of the following provides adequate defense-in-depth to assure an extremely small probability of a load drop during TCF and cask-handling operations and satisfies NUREG-0612 guidelines:

• using the auxiliary building crane for these manipulations

• proposed personnel training

• equipment inspections

• procedural controls Plant operators now handle fuel assemblies containing rod cluster control assemblies during refueling manipulations. A fuel assembly with up to 24 TPBARs and a loaded consolidation canister each weigh less than a fuel assembly with a rod cluster control assembly. Considering the expected number of consolidation canister handling evolutions and the features identified to reduce the potential of a load drop, the staff concluded that the potential for dropping a

38 consolidation canister is extremely small. The administrative controls applied to fuel handling ensure that the potential for damage to a consolidation canister from a fuel-handling accident is also extremely small. This outcome satisfies the guidelines of Section 5.1 of NUREG-0612.

Accordingly, handling a fuel assembly with TPBARs or a loaded consolidation canister does not increase the probability of an FHA.

Increasing boron concentration in the RCS accumulators and the RWST does not affect any initiating event for the DBAs. Thus, operating with TPBARs does not significantly increase the probability that any of the above accidents can occur.

TVA and the staff analyzed the consequences of plant operation with the maximum number of TPBARs installed for all of the above DBAs. In Section 2.5.1 of this safety evaluation, the staff determined that most DBA analysis inputs and assumptions did not change for the SQN 1 & 2 core fully loaded with TPBARs. The most changes occurred to the FHA. The staff determined that these changes resulted in suitably conservative offsite and control room doses. Section 2.5.4 of this safety evaluation explains that TVAs DBA analyses determined the whole-body dose and the thyroid dose due to inhalation of radioiodine, are consistent with 10 CFR Part 100 and GDC-19. The staff concluded that TVAs estimates of the exclusion area boundary, low-population zone, and control room doses will continue to comply with 10 CFR Part 100 and GDC-19 with TPBARs in the reactor core.

The staff evaluated the consequences of postulated FHA accidents on reactivity in Section 2.10 of this safety evaluation. TVA analyzed all postulated accidents (dropped fuel assembly, water temperature and density effects, eccentric positioning of a fuel assembly within the rack, abnormal and misplacement of fresh fuel assembly, etc.) for this amendment request. TVAs analysis showed that misplacing a fresh fuel assembly in a water cell is the most serious postulated condition. The analysis showed that adding 55 ppm of soluble boron would bring keff in the storage rack to less than or equal to 0.95. The staff determined that the analyses were conducted in accordance with the requirements of GDC-62 and other regulatory documents for preventing criticality in fuel storage and handling, and thus are acceptable. The spent fuel analysis shows that the design and operations of the SQN 1 & 2 plant will maintain keff less than or equal to 0.95 under all postulated accidents for the abnormal and accident conditions. Thus, the consequences of previously evaluated FHAs involving reactivity are not significantly increased.

As discussed in Section 3.2.1 of this safety evaluation, the change in the boron concentration in the RCS accumulators and the RWST increases the amount of boron in the sump during a LOCA. These increases in boron concentrations decrease the time to reach boron precipitation limits in the reactor core, thus requiring a shorter switchover time. TVA changed the relevant Bases to revise the switchover time for containment sump to hot leg recirculation from 9 hours0.375 days <br />0.0536 weeks <br />0.0123 months <br /> to 3 hours0.125 days <br />0.0179 weeks <br />0.00411 months <br /> to prevent any potential boron precipitation in the core. However, the time being shortened does not change the switchover function, and the evolution is controlled by emergency operating procedures. Based on the above, the staff concludes that the amendment does not significantly increase the consequences of an accident previously evaluated.

Second StandardThe amendment does not create the possibility of a new or different kind of accident from any previously analyzed.

As noted in various sections of this safety evaluation, the staff evaluated the proposed changes in accordance with the applicable NRC regulations, NUREGs, RGs, SRP sections, and industry codes and standards. No unproven techniques or methodologies were used in the analyses and design of the TPBARs. Section 1.1.2 of this safety evaluation explains how TVA irradiated a limited number of fuel assemblies containing test TPBARs (LTAs) in the WBN 1 core during

39 one fuel cycle. TVA irradiated the 32 TPBARs and removed them from the reactor. DOE then shipped the rods offsite and examined them to confirm TPBAR design and analytical modeling assumptions. The TPBARs performed as expected during the irradiation, as shown by the monitoring performed during the 17-month irradiation and the subsequent examinations.

The staff evaluated the effect of TPBARs on reactor core cooling in Section 2.3 of this safety evaluation. The analysis method Westinghouse used to determine the acceptability of operating with TPBARs in the reactor core is consistent with the current standard Westinghouse methods for inserting new components into Westinghouse cores and DOEs NRC-approved Topical Report NDP-98-153, Tritium Production Core (TPC). Westinghouses analyses demonstrated continued integrity of the core in the presence of TPBARs. Fuel integrity was assured by demonstrating that DNB will not occur on the most limiting fuel rod with at least a 95-percent probability at a 95-percent confidence level. The results of the DNB analyses showed that the departure from nucleate boiling ratio design basis will continue to be met. The thermal-hydraulic design of the TPBARs meets the requirements of Section 3.6.2 of the NRCs SRP and is below the operating temperature analyzed in the structural analysis of Section 3.4 of that document. Thus, the analyses conducted by Westinghouse show that operating with TPBARs will not have adverse effects on the thermal and hydraulic design or performance of the SQN 1 & 2 core or create the possibility of a new or different kind of accident.

Dropping a loaded consolidation canister or a fuel assembly containing a TPBAR assembly is bounded by the fuel-handling accident analysis. Accordingly, handling a loaded consolidation canister or a fuel assembly with TPBARs does not create the possibility of a new or different kind of accident.

As discussed above and in Section 3.2.1 of this safety evaluation, the change in the boron concentration in the RCS accumulators and the RWST increases the amount of boron in the sump during a LOCA. These increases in boron concentrations decrease the time to reach boron precipitation limits in the reactor core, thus requiring a shorter switchover time. TVA revised the switchover time for containment sump to hot leg recirculation from 9 hours0.375 days <br />0.0536 weeks <br />0.0123 months <br /> to 3 hours0.125 days <br />0.0179 weeks <br />0.00411 months <br /> to prevent any potential boron precipitation in the core. However, the time being shortened does not change the switchover function, and the evolution is controlled by emergency operating procedures. Thus, this does not create the possibility of a new or different kind of accident. Based on the above, this amendment does not create the possibility of a new or different kind of accident from any previously analyzed.

Third StandardThe amendment does not involve a significant reduction in the margin of safety.

In evaluating a potential significant reduction in the margin of safety, the staff evaluated the following safety issues that could be impacted by operating SQN 1 & 2 with TPBARs:

• handling of TPBARs

• procurement and fabrication of TPBARs

• compliance with departure from nucleate boiling criterion

• reactor vessel integrity analysis

• control room habitability systems

• specific assessment of hydrogen source and timing of recombiner operation

• light-load handling system

• station service water system

• ultimate heat sink

• new and spent fuel storage

• spent fuel pool cooling and cleanup system

40

• component cooling water system

• demineralized water makeup system

• liquid waste management system

• process and effluent radiological monitoring and sampling system

• use of LOCTA_JR code for loss-of-coolant accident analyses

• anticipated transient without scram analysis The NRC staff concluded in this safety evaluation that TVA satisfactorily addressed the above issues, and all proposed changes met applicable NRC regulations, NUREGs, RGs, SRP sections, and industry codes and standards. NRC staff determined the following to conclude that the amendment does not involve a significant reduction in the margin of safety:

• The design of the TPBAR assemblies allows for the use of existing equipment and procedures that safely handle BPRAs and fuel assemblies. The existing light-load handling system satisfies Section 9.1.4 of the NRCs SRP (NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants) for TPBAR assembly handling, and the proposed TPBAR assembly handling process is acceptable.

• The overall TPP structure provides for effective control of all supplier activities in compliance with applicable regulatory QA and procurement requirements. TVA has reviewed the QA programs of DOEs direct suppliers and they qualified as having QA programs that conform to the requirements of Appendix B to 10 CFR Part 50. Through its procurement process, DOE has imposed Part 21 requirements on all TPBAR suppliers.

• Operating with TPBARs will not have adverse effects on the thermal and hydraulic design or performance of the SQN 1 & 2 cores. The analyses also showed that the bypass flow and the DNB criterion will continue to be met with no feature of the TPBAR challenging the cooling capacity of the core.

• Material property changes due to neutron irradiation (termed neutron embrittlement) are sufficiently low that the plant is operated only under conditions with sufficient safety margins against fracture of the reactor vessel.

• The exclusion area boundary, low-population zone, and control room doses will continue to comply with 10 CFR Part 100 and GDC-19.

• TPBARs will not be a significant contributor to the post-LOCA hydrogen inventory.

Operating with TPBARs will not significantly impact the total hydrogen concentration within the containment when compared to the values associated with the non-TPBAR core. The maximum containment hydrogen concentration reaches 3.78-volume percent 2 days following event initiation. This is less than the lower flammability limit of 4-volume percent.

If one recombiner train is started at 24 hours1 days <br />0.143 weeks <br />0.0329 months <br /> after event initiation for the TPC case (when containment hydrogen concentration is 3.19-volume percent), the peak containment hydrogen concentration is limited to 3.56-volume percent for up to 6 days.

• TVA will use the same methods, procedures, and equipment to handle TPBARs as it does to safely handle fuel assemblies that do not contain TPBARs, as discussed under the First Standard segment above. The staff evaluated TVAs provisions for handling TPBARs and concluded that the equipment and administrative controls that TVA will apply to handling TPBAR assemblies and transport casks provide reasonable assurance of safety. A combination of provisions provides adequate defense-in-depth to assure an extremely small probability of a load drop during TCF- and cask-handling operations and satisfies NUREG-0612 guidelines. Considering the expected number of consolidation canister handling evolutions and the features identified to reduce the potential of a load drop, the staff

41 concluded that the potential for dropping a consolidation canister is extremely small.

• The SQN 1 & 2 essential raw cooling water system has adequate cooling capacity and margin to perform its safety and nonsafety functions with the additional heat loads imposed by TPC activities. TPP activities will not have an adverse impact on the ERCW heat removal capabilities.

• The ultimate heat sink has adequate cooling capacity and margin to perform its safety and nonsafety functions with the additional heat loads imposed by TPC activities. TPP activities will not have an adverse impact on the UHS heat removal capabilities.

• Analysis of the SFP containing TPBARs showed that adding 55 ppm of soluble boron would bring keff in the storage rack to less than or equal to 0.95. The staff determined that the analyses were conducted in accordance with the requirements of GDC-62 and other regulatory documents for preventing criticality in fuel storage and handling, and thus are acceptable.

• The SFPCCS had adequate capacity and cooling margin to perform its safety and non-safety functions with the additional heat loads imposed by TPP activities.

• The CCS has adequate cooling capacity and margin to perform its safety and non-safety functions with the additional heat loads imposed by tritium production activities. TPP activities will not have an adverse impact on the CCS heat removal capabilities.

• SQN 1 & 2 has sufficient storage and water makeup capacity to adequately meet any increased feed and bleed demands from normal and abnormal TPC events.

• The process and effluent radiological monitoring and sampling systems at SQN 1 & 2, with the described enhancements, are sufficient to adequately meet regulatory requirements for TPC operations. Effluent releases from TPC operation for both normal and abnormal (two TPBAR failures) conditions will not result in exceeding 10 CFR Part 20 or ODCM limits.

• Westinghouse Topical Report WCAP-15409 is acceptable for referencing in licensing analyses

• Using TPBAR cores in SQN 1 & 2 conforms to the ATWS rule as specified in 10 CFR 50.62(c)(1) for Westinghouse plants.

• Changing the switchover time for containment sump to hot leg recirculation from 9 hours0.375 days <br />0.0536 weeks <br />0.0123 months <br /> to 3 hours0.125 days <br />0.0179 weeks <br />0.00411 months <br /> is controlled by emergency operating procedures and does not change the switchover function.

Thus, the proposed amendment does not involve a significant reduction in the margin of safety.

On the basis of the above evaluation, the NRC has made a final determination that the proposed amendment does not involve a significant hazards consideration.

5.0 COMMENTS ON PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION The NRC received written comments by a letter dated January 16, 2002, from Dr. Kenneth D.

Bergeron regarding the staff's December 17, 2001, proposed no significant hazards consideration determination (66 FR 65000). The Blue Ridge Environmental Defense League also submitted written comments on January 16, 2002, which incorporated Mr. Bergeron's comments by reference. The NRCs letter to Dr. Bergeron from Dr. Brian Sheron of September 6, 2002, addressed the above comments that directly pertain to TVAs license

42 amendment request. This letter (Accession No. ML022410310) is accessible electronically from the Agencywide Documents Access and Management System (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html.

6.0 STATE CONSULTATION

On September 30, 2002, in accordance with the Commissions regulations, the Tennessee State Official, Ms. Debra Schults, was notified of the proposed issuance of the amendment.

The State official had no comments.

7.0 ENVIRONMENTAL CONSIDERATION

Pursuant to 10 CFR 51.21, 51.32, and 51.35, an Environmental Assessment and Finding of No Significant Impact has previously been prepared and published in the Federal Register on September 23, 2002 (67 FR 59581). Accordingly, based on the environmental assessment, the Commission has determined that the issuance of this amendment will not have a significant effect on the quality of the human environment.

8.0 CONCLUSION

The Commission 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 Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Date: September 30, 2002 Principal Contributors:

A. Attard M. Kowal S. Jones S. La Vie T. McLellan R. Hernan

A-1 APPENDIX AFLOW OF QA PROGRAM REQUIREMENTS TVA/DOE Interagency Agreement TVA Tritium Program Requirements (TVA-TPPR-99-01)

APPENDIX BDETAILS OF NRC STANDARD REVIEW PLAN (SRP) REVIEW The licensee evaluated the impacts of TPBARs on all aspects described in the NRCs NUREG-0800, Standard Review Plan. The licensee submitted these evaluations in the Framatome-ANP topical report referenced in the staffs safety evaluation and provided by TVA in its supplemental information. This evaluation is necessary because SQN utilizes Mark-BW fuel assemblies. The reference plant evaluated in the NRC's Safety Evaluation Report, issued as NUREG-1672, Safety Evaluation Report Related to the Department of Energys Topical Report on the Tritium Production Core, May 1999, assumed a Westinghouse plant using Westinghouse-designed fuel. Additionally, SQN has incorporated design changes to the tritium producing burnable absorber rods (TPBARs) design previously evaluated in the staff Safety Evaluation Report (NUREG-1672). This section is organized according to applicable sections of the SRP.

SRP Section 4.2 - Fuel Design Evaluation SRP Section 4.2 provides the basis for the staffs requirements regarding fuel system design.

The objectives of the fuel system review are to provide assurance that:

(a) the fuel system is not damaged as a result of normal operation and anticipated operational occurrences, (b) fuel system damage is never so severe as to prevent control rod insertion when it is required, (c) the number of fuel rod failures is not underestimated for postulated accidents, and (d) coolability is always maintained.

The TPC topical report identified that the TPBAR design shall meet 12 functional requirements.

These functional requirements are consistent with the SRP Section 4.2 fuel design criteria that include structural integrity, thermal and mechanical performance, and nuclear characteristics.

In a response to a staff question, the licensee illustrated how these functional requirements will be met by the TPBAR design.

The licensee stated that an analysis was performed for each functional requirement to ensure that the TPBAR design can be irradiated to 550 effective full-power days (EFPD). The analyses included structural integrity, swelling and rupture, vibrational fatigue, creep collapse, fretting wear, corrosion, dimensional integrity, and coolant temperature. The results showed that the TPBAR design conservatively met all the design limits. The staff reviewed these analyses and agreed with the licensee assessment. Based on the licensees conservative analyses and results, the staff concludes that the TPC TPBAR design is acceptable for 550 EFPD exposure in SQN.

SRP Section 4.3 - Nuclear Design The acceptance criteria for SRP Section 4.3, assure that the requirements of GDC Nos. 10, 11, 12, 13, 20, 25, 26, 27 and 28 are satisfied. These criteria ensure that acceptable fuel design limits are specified (GDC-10), a negative prompt feedback coefficient is specified (GDC-11) and power oscillation is addressed in (GDC-12). A control and monitoring system is required (GDC-13) which automatically initiates a rapid reactivity insertion to prevent exceeding fuel design limits in normal operation or anticipated transients (GDC-20). The control system is required to be designed so that a single malfunction or a single operator error will cause no violation of fuel limits (GDC-25). A reactor coolant boration system is provided which is capable of bringing the reactor to cold shutdown conditions (GDC-26) and the control system is required to control reactivity changes during accident conditions when combined with the engineered safety features (GDC-27). Reactivity accident conditions are required to be limited so that no damage to the RCS boundary occurs (GDC-28).

To evaluate the impacts of the TPBARs on the nuclear reactor core design, the licensee developed and analyzed first and equilibrium cycle core designs using NRC approved methodology. The approved methods are described in the SQN UFSAR. The SQN topical report describes the nuclear design methodology, design bases, core design descriptions, core power distribution and reactivity coefficient evaluations, and effects of extended shutdowns for the representative SQN TPC designs.

In the SQN TPCs, the TPBARs function in a manner that is similar to conventional burnable absorbers. While the depletion behavior of the TPBARs is different than that of conventional burnable absorbers, this does not necessarily lead to significant differences in core physics behavior. Through TPC design and analyses, the licensee determined that the behavior of core designs with respect to power distributions, reactivity coefficients, and other core physics parameters is comparable to that of current SQN core designs. TPCs are designed to meet the same safety analysis criteria as standard reload cores. The licensee determined that the only new safety analysis criteria added for the TPBAR cycles is the minimum and maximum tritium production per rod. These criteria, including uncertainties, are checked by the core designer each cycle to ensure that the maximum and minimum tritium production remains within the TPBAR design limits of 1.2 and 0.15 grams, respectively.

Based on the SQN TPC reload design analyses, some features of the SQN TPBAR and nuclear design are different from those evaluated in the NRC staff Safety Evaluation Report on the TPC. The licensee identified the following significant differences between the SQN TPC and the design described in the TPC topical report and NRC Safety Evaluation Report:

(a) A feed batch of 96 Mark-BW fuel assemblies was used instead of 193 and 140 VANTAGE+TM fuel assemblies.

(b) Two 6Li concentrations are used instead of one; concentrations slightly higher (0.032 gm/in) and lower (0.029 gm/in) than that in the generic TPC topical report analysis (0.030 gm/in) were used.

(c) A singular, longer 6Li poison column length of 132 inches, centered with respect to the fuel stack was used. The TPC Topical Report analysis used 127.5 and 128.5 inch lengths, and the WBN 1 LTAs used a 142 inch length.

(d) Gadolinia (Gd2O3) was used as integral burnable absorber instead of IFBA (ZrB2);

fuel enrichment was slightly reduced in the fuel pellets that contain gadolinia.

(e) Burnable Poison Rod Assemblies (BPRAs) containing B4C-Al2O3 pellets were used on the periphery for fluence control in the equilibrium fuel cycle instead of TPBARs.

(f) As few as 12 tritium producing burnable absorber rods (TPBARs) in a single cluster were used in the transition cycle whereas no fewer than 20 per cluster were used in the TPC topical report analysis.

(g) No fuel rod enrichment zone loading was employed except for fuel rods containing gadolinia.

The licensee addressed each of these design differences. Each of these design parameters was incorporated into the CASMO-3/NEMO core simulator model that was used to evaluate the SRP Section 4.3 acceptance criteria. Because these items are explicitly modeled, the effects of each of these items is considered in each phase of the reload safety evaluation. Additionally, the licensee modified the cross section libraries and cross section generation process in the CASMO-3 and NEMO computer codes to include isotopes important to TPBAR cores (tritium, helium and lithium isotopes).

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The SQN TPC design calculations and analyses demonstrated that all key safety analysis parameters fall within the ranges and limits normally assumed for SQN reload analysis, with four exceptions. These exceptions include shutdown margin (resulting in the relocation of four RCCAs), Doppler Only Power Coefficient (DOPC) at zero power, HZP ejected rod worth at beginning of cycle (BOC) (which may affect rod insertion limits), and post-LOCA recriticality (which affects the RWST and CLA boron concentration). The post-LOCA recriticality issue is discussed in Section 3.2 of this Safety Evaluation. The other exceptions are discussed here.

In order to ensure that the SQN shutdown margin (SDM) requirement of 1.6 percent k/k is satisfied while not compromising the amount of tritium produced, the licensee determined that four RCCAs would need to be relocated. Four RCCAs that were previously located in peripheral core locations were moved inward to existing spare control rod drive mechanism locations. The RCCA relocation was implemented under the 10 CFR 50.59 process for SQN Units 1 and 2 during the Cycle 11 refueling outages. The licensees 10 CFR 50.59 evaluation concluded that a license amendment is not required, and that the RCCA relocation does not have any negative impacts on any of the UFSAR Chapter 15 transient analyses.

The licensee identified that the most negative Doppler-Only Power Coefficient (DOPC) at HZP falls outside the acceptable range for this key safety analysis parameter. At zero power and end of cycle, flux redistribution causes the DOPC to be more negative than the limit. The licensee provided a discussion which justifies why this is acceptable and noted that this issue is not specific to TPBAR core designs. The discussion provided by the licensee explains that any accidents starting at HZP are conservatively analyzed assuming a least negative DOPC, because a more negative value will result in a lower final power level.

The licensee identified that evaluations for the HZP ejected rod event for the first transition fuel cycle failed to meet the BOC ejected rod worth requirement. Satisfactory results were obtained by increasing the HZP rod insertion limit specified in the Core Operating Limits Report (COLR) for the first transition core by eight steps. NRC approved methods are utilized for all parameters included in a COLR. The licensee will evaluate the need to make rod insertion limit modifications during each refueling cycle as part of the reload safety evaluation.

To account for the effects of flux peaking caused by axial gaps between absorber pellets in a pellet stack or between pellets in adjacent pencils, the licensee applied conservative augmentation factors in the power distribution analysis. The gap peaking augmentation factors were applied as a function of fuel enrichment, 6Li concentration, TPBAR pattern type, location of the fuel rod within the assembly, and elevation along the fuel rod. These factors were bounding for all burnup values.

For an extended shutdown near end of life, the buildup of 3He through tritium decay can have a significant impact on core reactivity. The TPC topical report showed that the 3He buildup after a 6-month shutdown could reduce the critical boron concentration at HFP by about 80 ppm upon startup. For SQN, the expected reactivity effects of 3He buildup will be smaller than that in the TPC topical report design because of the smaller number of TPBARs to be used and the harder neutron spectrum in the fuel lattice. To address this concern, the licensee will provide guidance for use by reactor operators and engineers to identify and assess parameters that would indicate that a formal evaluation of an extended shutdown should be conducted. This evaluation would ensure that the core reactivity and power distribution limits specified in the TSs and COLR remain valid, or be revised (as appropriate), prior to resumption of power operation.

The NRC staff has reviewed the licensees submittals associated with SRP Section 4.3, Nuclear Design, and based on the above discussion, concludes that TPC designs can be developed for SQN that achieve typical cycle energy goals, generate tritium for DOE, and continue to meet all safety requirements for nuclear design.

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SRP Section 4.4 - Thermal and Hydraulic Design The acceptance criteria for the thermal and hydraulic design of the reactor core and the RCS assure that the requirements of 10 CFR 50, Appendix A, GDC-10 and 12 are satisfied. The licensee performed a thermal-hydraulic evaluation utilizing the following design criteria to demonstrate acceptable operation with TPBARs.

(a) The mechanical integrity of the thimble tube is maintained during the life of the fuel with the presence of the TPBAR by demonstrating adequate cooling of the thimble tube to preclude excessive component temperatures and corrosion.

(b) The core will remain protected from departure from nucleate boiling (DNB) by assurance that there will be at least a 95 percent probability at a 95 percent confidence level that the hot fuel rod in the core does not experience a departure from nucleate boiling or transition condition during normal operation or anticipated operational occurrence.

(c) The core departure from nucleate boiling ratio (DNBR) predictions account for the localized rod power influence associated with the positioning of TPBARs within the thimble tubes.

(d) Centerline fuel melting will not be permitted for normal operation or anticipated operational occurrences.

To demonstrate that these acceptance criteria are satisfied with a TPC at SQN, the licensee evaluated the impact of the TPBARs using approved methodologies for licensing the Mark-BW17 fuel design at the SQN units. The LYNXT thermal-hydraulic code was used to predict the local coolant and surface temperature conditions within the thimble tubes and surrounding subchannels. The BWCMV-A CHF correlation was applied in analyzing the impact of localized fuel rod power perturbations associated with TPBARs. The BWU CHF correlation was used to predict the DNBR for the steamline break analysis due to its better performance at low pressure conditions. All other DNB analyses utilized the BWCMV-A correlation. The TACO and GDTACO fuel performance codes were used to quantify the impact of TPBAR fuel cycle design steady-state peaking changes on centerline fuel melt limits as compared to non-TPBAR fuel cycles for UO2 and gadolinia fuel rods, respectively.

The licensee also used the LYNXT code to evaluate the impact of peaking spikes associated with the axial gaps between the TPBARs on local DNBR. Predicted allowable peaks from the LYNXT analysis were used to develop allowable peaking reduction factors that account for the effect of the local power spikes on DNB. The licensee assessed these maximum allowable peaking limit reduction factors, which ranged in value from approximately 0.3 percent to 0.4 percent, as part of the core power distribution analysis.

The results of these analyses show that no bulk boiling will occur in the thimble tube, thereby precluding excessive thimble tube temperatures that could jeopardize the tube integrity. The core bypass flow rate through a thimble tube occupied by a TPBAR is comparable to a tube occupied by a thimble plug and has little impact on the overall core bypass flow rate. The licensee will continue to verify that the cycle-specific core bypass flow rate remains bounded by that assumed in the DNB analyses. The magnitude of the augmentation factors attributed to the axial peaking spikes formed by the axial gaps between the TPBARs is generally small and will be applied to fuel rod peaking margin calculations during the SQN reload safety evaluation process. The licensees evaluations of the TPBAR transition and equilibrium fuel cycle designs shows acceptable DNBR performance for both steady-state and transient conditions. The centerline fuel melt limits previously established for SQN reloads was shown to be applicable for SQN cores containing TPBARs.

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The licensee utilized NRC-approved codes and methods to demonstrate that acceptable core thermal-hydraulic conditions are predicted for the operation of TPBARs in future SQN TPCs.

These criteria include fuel rod integrity, thimble tube integrity, maximum core bypass flow rates, and DNB. Explicit checks of the these criteria are included in the cycle-specific reload safety evaluation performed for each SQN reload core design.

The NRC staff has reviewed the licensees submittals related to SRP Section 4.4 and based on the above discussion, concludes that core thermal-hydraulic conditions will be acceptable for future SQN TPCs.

SRP Section 3.9.1 - Nuclear Steam Supply System (NSSS) Design Transients Evaluation The NSSS design transients are used as input for the component fatigue stress analyses of the various RCS components. They describe the thermal and hydraulic (i.e., pressure, temperature, and flow) variations that occur during various normally expected plant maneuvers and during unanticipated transients. The TPC topical report stated that a TPC could impact these parameters such that a modification of the design transients may be necessary. In the SQN topical report (Table 4 -1), the licensee states that the pertinent operating parameters (NSSS power, RCS flow, RCS temperatures, RCS pressure, steam temperature, feedwater temperature, and steam flow) for the TPC are unchanged from those previously evaluated.

Therefore, for a SQN TPC, the existing NSSS design transient curves remain valid.

SRP Section 4.6 - RCCA Rod Drop Time Evaluation The TPC topical report stated that a TPC could slightly increase the core pressure drop, and thus slightly increase the calculated RCCA rod drop time. To address this possibility, the licensee performed an analysis for the TPC design conditions. The analysis showed that the TPC has no effect on the RCCA drop times relative to the uprated SQN core design. This is consistent with the conclusions of the TPC topical report. Additionally, RCCA rod drop times will continue to satisfy the TS requirement.

SRP Section 5.2 - Cold Overpressure Mitigation System (COMS) Evaluation The TPC topical report stated that a TPC could impact the operation of the COMS system if pertinent operating parameters (Reactor Vessel Appendix G limit, heat and mass input design bases, Pressurizer PORV setpoint or capacity, RHR relief valve setpoint or capacity) are changed. In the SQN topical report (Table 4 - 1), the licensee states that the pertinent operating parameters for a SQN TPC are unchanged from those previously evaluated.

Therefore, the existing COMS analyses and setpoints remain applicable for the SQN tritium program.

Standard Review Plan Section 5.4.7 - Residual Heat Removal (RHR) System Evaluation The TPC topical report recommended that a plant-specific evaluation of the net effect of a TPC on the RHR system cooling capability be performed. In general, decay heat from the RCS will be reduced with TPBARs due to the reduced average maximum burnup in the core. This reduction in decay heat will be offset by an increased heat load on the SFP heat exchangers.

The licensees analysis determined that the increase in heat load imposed on the SFP as a result of TPC is less than 0.5 MWt, which includes both the increase in reactor core decay energy and higher decay energy in the SFP from the cumulative affect of multiple TPC core discharges over the time to full capacity of the SFP. This small increase in heat load is insignificant relative to the overall conservative cooldown assumptions utilized in the Westinghouse cooldown methodology, and the increase represents a small fraction of the total RHR heat load at maximum decay heat (approximately 38 MWt). Based on this analysis, the B-5

licensee determined that the existing RHR cooldown times are not affected by a TPC core.

SRP Section 15 - Accident Analysis Non-LOCA Transients As part of the reload analyses for the first transition and equilibrium TPBAR cycles at SQN, the licensee determined the key safety analysis parameters and compared them to the parameters used in the current applicable safety analysis for SQN. The licensees evaluation of these key safety analysis parameters showed that:

a. No changes have been identified in the nominal plant operating conditions (power, coolant temperature, pressure and flow rate) assumed in the plant safety analysis in order to accommodate the TPBARs. Therefore, the existing safety analysis calculations for SQN are not affected by any changes in plant parameters as a result of the TPBARs.

b. No changes to the reactor core thermal hydraulic characteristics or power peaking factors, which could affect the core thermal limits (DNBR and overpower), have been identified as a result of the use of TPBARs. Therefore, the plant thermal limit protection setpoints do not change as a result of the TPBARs.

c. The nuclear design and fuel rod design calculations performed for the TPBAR reload core design have identified no safety analysis parameters outside of the bounds of the current applicable reload safety analysis parameters. Therefore, no change to the existing licensing-basis safety analysis is required as a result of the TPBAR core design at SQN.

The licensee provided a table which compares the limit values of the key safety analysis parameters to the cycle-specific values for a first transition SQN TPC cycle, an equilibrium SQN TPC cycle and a recent SQN cycle without TPBARs. The comparison of these three designs shows that all fall within the limits and ranges of the parameters assumed in the safety analysis with the exception of the most negative DOPC at HZP, which is discussed in the SRP Section 4.3 section of this safety evaluation.

The licensees confirmatory checks of the key safety analysis parameters demonstrated that the SQN TPC has not changed any of the bounding values assumed in the safety analyses.

Therefore, the SQN safety analysis for the non-LOCA events is unaffected by the TPBAR core design, and all of the applicable acceptance criteria continue to be met.

LOCA Evaluations The licensee performed an evaluation to determine the response of the TPBARs to the design basis LBLOCA and SBLOCA. The LOCTA_JR code was used to calculate the TPBAR temperature during a LOCA for the SQN TPC. LOCTA_JR uses as boundary conditions the cladding temperature of the surrounding fuel rods and the core steam and entrained liquid convective heat transfer coefficients and temperatures. The boundary conditions are taken from Appendix K LOCA analyses of record for SQN Units 1 and 2. Because of uncertainties in the application of the LOCA hot rod heat transfer coefficient (HTC) to the guide thimble, two cases were run for both LBLOCA and SBLOCA. The first case is considered to be a reasonable approach, while the purpose of the second case was to maximize the thimble temperature throughout the transient to quantify an upper bound temperature. For the LBLOCA, the peak guide thimble temperature obtained was 2127 EF with a corresponding TPBAR temperature of 2109 EF. The SBLOCA limiting thimble temperature was determined to B-6

be 1040 EF with a corresponding peak TPBAR temperature of 1034 EF. These temperatures remain below the 10 CFR 50.46 criteria of 2200 EF. This temperature is acceptable for the thimble. The TPBAR pressure boundary is made of 316 stainless steel, which is not prone to the metal/water reaction like zirconium at these temperatures and is therefore not affected.

Appendix K modeling of the downcomer region and downcomer boiling have recently been shown to substantially impact peak cladding temperature (PCT) and oxidation following a LOCA, especially for ice condenser containments. As discussed in the previous paragraph, the licensee used the worst case (Appendix K) LOCA analysis PCT response to predict post-LOCA TPBAR thermal response. The downcomer boiling issues can have an adverse effect on the TPBAR temperatures. Framatome-ANP, the fuel vendor for SQN was notified of this concern by NRC letter dated June 3, 2001 and is working with the staff to resolve this modeling issue.

In response to an RAI on the impacts of downcomer boiling, the licensee demonstrated that the current SQN Appendix K model contains sufficient conservatism to offset the expected impacts of downcomer boiling, and has provided a schedule for reanalysis.

Additionally, the licensee has provided a statement that SQN and its vendor have ongoing processes which assure that LOCA analysis input values for peak cladding temperature-sensitive parameters bound the as-operated plant values for those parameters. In a telephone call with SQN on September 9, 2002, the licensee clarified their response to question 17 of the NRC RAI dated September 4, 2002. The licensee stated that they did not intend to provide an all-inclusive list of such processes, but rather, provided a sampling to illustrate the types of processes they have in place. The licensee stated that processes and procedures in addition to those included in their RAI response are used to ensure LOCA analysis input values are appropriate.

The licensees evaluation of the impacts of TPBARs on the LBLOCA and SBLOCA analyses considered the key core design parameters related to LOCAs. The licensee determined that current and future key parameters for both the LBLOCA and SBLOCA are met for a TPC. The only change found to be necessary as a result of the TPBARs is to increase the RWST and CLA boron concentration levels to ensure post-LOCA subcriticality. This change was evaluated in detail in Section 3.2 of the staffs safety evaluation.

Based on the above discussion, the staff finds that the use of TPBARs at SQN is acceptable with respect to LOCA analyses.

TPBAR Evaluation Production TPBAR Design The TPC topical report listed plant-specific analyses required for TPBARs. The plant- specific analyses were intended to verify that the TPBARs were compatible with the existing core design. The analyses included core design conditions, nuclear design, thermal-hydraulic design, and mechanical performance.

The core design conditions included verification of compliance with requirements of production, power peaking, cycle length, thimble flow, power density, and generic reactor conditions. The licensee stated that the goal of the TPC design was to maximize the amount of tritium production within the design capabilities of the TPBAR while assuring that all safety criteria are met. The licensee analyses verified that the core design conditions including power peaking, cycle length, power density, and thimble flow were consistent with the current licensing basis which was validated for each cycle. The staff reviewed the analyses and agrees with the licensees assessment.

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The nuclear design included verification of compliance with, and conservatism of input with, limits of power distribution and power peaking. The licensee stated that a cycle-specific power distribution analysis will be performed to ensure that the reload core design is conservative with respect to the power distribution limits. The licensee also confirmed that acceptable peaking margins to the core safety limits will be maintained through the reload cores. Based on the licensees commitment, the staff considers that the licensees approach is acceptable.

For thermal-hydraulic design, the licensee is required to verify conservatism of conditions for thimble flow, pressure drop, and margin to boiling. The licensee confirmed that a conservative analysis using the LYNXT code reflected adequate margin to boiling within the thimble tubes.

In addition, the thimble flow and pressure drop in the thimble tubes were conservatively analyzed to maintain adequate flow and pressure. The staff reviewed the analysis and agrees with the licensee assessment.

For mechanical performance, the licensee is required to verify compliance and conservatism of conditions for (1) tritium and helium production, pressure, and cladding stress, (2) getter loading, and (3) tritium release. For each cycle, the licensee performed an analysis of maximum production of tritium and helium, and maximum pressure loading on cladding. The result showed that the cladding stress due to pressure loading was below the stress limit. The licensee also analyzed and confirmed that the TPBAR getter loading and tritium release were conservatively predicted and within the limits. The staff reviewed the analyses and agrees with the licensee assessment.

Based on the licensee response to the plant-specific analyses, the staff concludes that the licensee has adequately address the plant-specific requirements, and thus the TPC TPBAR design is acceptable for SQN.

Thermal-Hydraulic Evaluation of TPBARs The licensee evaluated the effects of the reactor core thermal-hydraulic conditions on the function and integrity of the TPBARs. For this evaluation, Framatome-ANP analytical tools and methods were applied to evaluate bypass flow through the assembly guide thimble tubes, guide tube boiling and TPBAR component temperatures. The results of these analyses show that the total core bypass flow remains below the 7.5 percent design value currently in place for SQN, there will be no bulk boiling in the guide thimble tubes, there will be no surface boiling from the core component rod within the dashpot region of the guide thimble tubes and the maximum TPBAR component temperatures remain below their respective melting temperatures during Condition I or Condition II and III events.

TS 5.6 and TS 3/4.7.13 Bases - Design Features/Fuel Storage and Spent Fuel Pool Minimum Boron Concentration - Revised Storage Requirements for Fuel Assemblies Containing TPBARs.

Technical Evaluation The Holtec International topical report, submitted to the NRC in a TVA RAI response, analyzed four different configurations for the storage of new and spent fuel at SQN.

1. Region 1a. Checkerboard storage of fresh and spent fuel assemblies (one fresh and three spent fuel assemblies in a 2X2 array). The fresh fuel assembly is assumed to contain no Gadolinia rods (Configuration 1).

2. Region 1b. Checkerboard storage of fresh and spent fuel assemblies (1 fresh and 3 spent fuel assemblies in a 2X2 array). The fresh fuel assembly is assumed to contain 4 or 8 2.0-weight percent Gadolinia rods with a 4.95 weight percent 235U carrier enrichment (Configuration 2).

3. Region 2. Storage of spent fuel face adjacent to each other with no other restriction except burnup (Configuration 3).

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4. Checkerboard storage of 2 fresh fuel assemblies and 2 empty cells in a 2X2 array. The fresh fuel assemblies are assumed to contain no Gadolinia rods (Configuration 4).

Methodology The Holtec Corporation analyzed the SQN spent fuel storage racks and documented the analyses in Holtec report HI-2012629. Even though the NRC never formally generically approved the methodology described in the Holtec report, it is recognized as the industry standard, and is reviewed and evaluated on a plant-specific basis. The methodology described in the Holtec report ensures that Keff remains less than or equal to 0.95 as recommended in ANSI/ANS-57-1983 and NRC guidance, such as CFR part 50.68. The methodology also takes partial credit for soluble boron in the SFP criticality analyses.

The Holtec International analysis used the most reactive design and the most reactive temperature to set the storage requirements. The analyses included means to account for the bias and uncertainty associated with the bench-marking of the methodology, a bias for the under-prediction of reactivity due to boron particle self-shielding, and the uncertainty due to mechanical tolerances from the manufacturing process. The licensee also included additional uncertainties related to irradiated fuel as described in the burnup credit methodology discussed in the Holtec report. The licensee determined these uncertainties at the 95/95 probability/confidence level.

The licensee performed the criticality analysis of the reactivity effects of fuel storage at CQN using the three-dimensional MCNP4a and MCNP4b as well as the KENO-Va Monte-Carlo computer codes. MCNP4b is MCNP4a incorporating fission product cross-sections. MCNP is a three-dimensional multi-group Monte-Carlo code developed at Los Alamos National Laboratory.

CASMO-4 is a two-dimensional multi-group transport theory code used for depletion analyses.

Since MCNP is not a depletion code, depletion calculations were performed using the CASMO-4 code (with explicit description of the fission product nuclide concentration) and compared to the criticality results of MCNP. All these codes are industry standard codes that were validated through bench-marking to relevant critical experiments, and are acceptable for licensing applications.

Normal Conditions The licensee performed depletion calculations using CASMO-4 to evaluate the reactivity of the fuel assemblies currently in use or anticipated for storage in the SQN spent fuel racks.

Criticality calculations performed using two different assembly designs; the Westinghouse 17x17 fuel assembly and the Framatome Cogima Fuel (FCF) Alliance fuel. Burnable poison loadings (including TPBARs) were explicitly included in the modeling.

Spent fuel pool calculations are usually performed with the highest reactivity worth fuel design.

For this analysis, Holtec International had determined that the Westinghouse 17X17 fuel design is more bounding than the FCF fuel design.

CASMO-4 calculations were carried out to account for uncertainties associated with densities, enrichments, mechanical tolerances, and TPBARs loadings. Since CASMO-4 does not include Li-6, typically used in the TPBARs, an equivalent boron was used to simulate the absorption in Li-6. Since this approximation could introduce some uncertainty, Holtec performed a sensitivity analysis was performed using KENO-5a by increasing the boron concentration in the simulated TPBARs by 25 percent. The result of the sensitivity study showed that the effect on the residual reactivity was small (0.0009 delta k) at 20 GWD/MTU, and (0.0030 delta k) at 60 GWD/MTU. In the above stated analyses for all four stated configurations, no soluble boron was required to achieve a keff less than 1.0. To reduce the keff to the regulatory requirement of less than or equal to 0.95, the addition of 500 ppm of soluble boron was required for Regions 1a, 1b, and 2. The keff for Region 3 is 0.95 without the addition of soluble boron.

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Abnormal or Accident Conditions and Soluble Boron Requirements Although the NRC permits credit for the soluble boron (poison) normally present in the SFP water, most abnormal or accident conditions will meet the limiting reactivity (Keff less than or equal to 0.95) even in the absence of soluble poison. The licensee analyzed all postulated accidents (i.e., dropped fuel assembly, water temperature and density effects, eccentric positioning of a fuel assemble within the rack, abnormal and misplacement of fresh fuel assembly, etc) for this amendment request. The licensee analyzed its spent fuel storage racks by taking into account boron credit in accordance with the methodology described in the Holtec report. This methodology ensures that Keff less than or equal to 0.95 as recommended in ANSI/ANS-57-1983 and as required in NRC regulations, such as 10 CFR 50.68.

The analysis contained in the Holtec report of this submittal, shows that the most serious postulated condition is the misplacement of a fresh fuel assembly without Gadolinia rods in the location of a water cell in Region 3. In this case the misplaced assembly has the potential for exceeding the limiting reactivity, Keff less than or equal to 0.95, but is always less than 1.0, should there be a concurrent and independent accident where all the boron had been lost in the SPF. To reduce Keff to less than or equal to 0.95, the addition of 700 ppm of soluble boron is required, bringing Keff in the storage rack to below or equal to the regulatory requirement of 0.95, including bias and uncertainties.

The licensees analyses included the assumptions that the moderator was pure water at a temperature of 4 °C and a density of 1.0 gm/cc. The analyses also included treatment for uncertainties due to tolerances in fuel enrichment and density, storage cell inner diameter, storage cell pitch, stainless steel thickness, assembly position, calculation uncertainty and axial burnup. The licensee also appropriately determined the uncertainties at the 95/95 probability/confidence level and included a methodology bias (determined from benchmark calculations) as well as a reactivity bias to account for the effect of the normal range of spent fuel pool water temperatures.

The NRC staff reviewed and evaluated the licensees analysis pertaining to the inclusion of TPBARs and other burnable absorbers into its spent fuel analyses, and have determined that the analyses were conducted in accordance with the requirements of GDC-62 and other regulatory documents and thus are acceptable. The analysis for the Abnormal and Accident Conditions is also acceptable since it satisfies the requirement of Keff of 0.95 as prescribed per the above regulatory requirements.

The staff finds the proposed SQN license amendment request acceptable in its safety evaluation and that it meets the requirements of GDC- 62 for the prevention of criticality in fuel storage and handling. The staff concluded that the licensee conducted the necessary analysis in accordance with NRC Guidelines and the ANSI standards, and the analysis shows that the design and operation of the SQN plant will maintain the maximum Keff within the acceptance criteria under all postulated accidents.

Bases 3/4.6.4 - Combustible Gas Control - Hydrogen Generation Sources The licensee proposes to revise the Bases section for TS 3/4.6.4 to add a fourth hydrogen generation source for the combustible gas control discussions. As a result of the SQN tritium program, possible sources of combustible gases now include the hydrogen and tritium gas inventories inside the TPBARs.

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