ML070650286
| ML070650286 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  (DPR-033, DPR-052, DPR-068) |
| Issue date: | 04/16/2007 |
| From: | Ellen Brown NRC/NRR/ADRO/DORL/LPLII-2 |
| To: | Swafford P Tennessee Valley Authority |
| Chernoff M, NRR/DORL, 415-4041 | |
| Shared Package | |
| ML070650326 | List: |
| References | |
| TAC MD3921, TAC MD3922, TAC MD3923 | |
| Download: ML070650286 (16) | |
Text
April 16, 2007 Mr. Preston D. Swafford Interim Chief Nuclear Officer Nuclear Support Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801
SUBJECT:
BROWNS FERRY NUCLEAR PLANT, UNITS 1, 2 AND 3 ISSUANCE OF AMENDMENTS REGARDING SCRAM TIME TESTING ACTIVITIES (TAC NOS. MD3921, MD3922, MD3923) (TS-456)
Dear Mr. Swafford:
The Commission has issued the enclosed Amendment Nos. 270, 299, and 258 to Renewed Facility Operating Licenses Nos. DPR-33, DPR-52 and DPR-68 for the Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3, respectively. These amendments are in response to your application dated December 21, 2006.
The amendments adopt the Technical Specifications Task Force (TSTF)-484, Revision 0, Use of Technical Specifications (TS) 3.10.1 for Scram Time Testing Activities, and revise TS Limiting Condition for Operation 3.10.1 and the associated TS Bases, to expand its scope to include provisions for temperature excursions greater than 212 degrees Fahrenheit as a consequence of inservice leak and hydrostatic testing, and as a consequence of scram time testing initiated in conjunction with an inservice leak or hydrostatic test, while considering operational conditions to be in Mode 4.
A copy of the Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
Sincerely,
/RA/
Eva A. Brown, Project Manager Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-259, 50-260, and 50-296
Enclosures:
- 1. Amendment No. 270 to DPR-33
- 2. Amendment No. 299 to DPR-52
- 3. Amendment No. 258 to DPR-68
- 4. Safety Evaluation cc w/enclosures: See next page
ML070650286 *No significant change from SE Input Memo NRR-058 OFFICE LPL2-2 LPL2-2/PM LPL2-2/LA NRR/ITSB/BC OGC LPL2-2/BC NAME MGutierrez EBrown BClayton TKobetz*
ACuratola TBoyce DATE 4/5/07 4/5/07 4/4/07 02/20/2007 4/14/07 4/16/07
TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-259 BROWNS FERRY NUCLEAR PLANT UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 270 Renewed License No. DPR-33 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated December 21, 2006, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-33 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 270, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 60 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
Thomas H. Boyce, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Operating License and Technical Specifications Date of Issuance: April 16, 2007
ATTACHMENT TO LICENSE AMENDMENT NO. 270 RENEWED FACILITY OPERATING LICENSE NO. DPR-33 DOCKET NO. 50-259 Replace Page 3 of Renewed Operating License DPR-33 with the attached Page 3.
Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
REMOVE INSERT 3.10-1 3.10-1
TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-260 BROWNS FERRY NUCLEAR PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 299 Renewed License No. DPR-52 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated December 21, 2006, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-52 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 299, 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 its date of issuance and shall be implemented within 60 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
Thomas H. Boyce, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Operating License and Technical Specifications Date of Issuance: April 16, 2007
ATTACHMENT TO LICENSE AMENDMENT NO. 299 RENEWED FACILITY OPERATING LICENSE NO. DPR-52 DOCKET NO. 50-260 Replace Page 3 of Renewed Operating License DPR-52 with the attached Page 3.
Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
REMOVE INSERT 3.10-1 3.10-1
TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-296 BROWNS FERRY NUCLEAR PLANT, UNIT 3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 258 Renewed License No. DPR-68 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated December 21, 2006, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-68 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 258, 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 its date of issuance and shall be implemented within 60 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
Thomas H. Boyce, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Operating License and Technical Specifications Date of Issuance: April 16, 2007
ATTACHMENT TO LICENSE AMENDMENT NO. 258 RENEWED FACILITY OPERATING LICENSE NO. DPR-68 DOCKET NO. 50-296 Replace Page 3 of Renewed Operating License DPR-68 with the attached Page 3.
Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
REMOVE INSERT 3.10-1 3.10-1
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 270 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-33 AMENDMENT NO. 299 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-52 AMENDMENT NO. 258 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-68 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT, UNITS 1, 2, AND 3 DOCKET NOS. 50-259, 50-260, AND 50-296
1.0 INTRODUCTION
By application dated December 21, 2006 (ADAMS Accession No. ML063610039), Tennessee Valley Authority (TVA, the licensee) requested changes to the Technical Specifications (TSs) for the Browns Ferry Nuclear Plant (BFN) Units 1, 2, and 3.
The proposed changes would revise Limiting Condition for Operation (LCO) 3.10.1, and the associated Bases, by expanding its scope to include provisions for temperature excursions greater than 212 degrees Fahrenheit (F) as a consequence of inservice leak and hydrostatic testing, and as a consequence of scram time testing initiated in conjunction with an inservice leak or hydrostatic test, while considering operational conditions to be in Mode 4.
These changes are based on Technical Specification Task Force (TSTF) Change Traveler TSTF-484, Revision 0, that has been approved generically for the boiling-water reactor (BWR)
Standard TSs, NUREG-1433, (BWR/4) and NUREG-1434 (BWR/6) by revising LCO 3.10.1, and the associated Bases, to expand its scope to include provisions for temperature excursions greater than 212 degrees F as a consequence of inservice leak and hydrostatic testing, and as a consequence of scram time testing initiated in conjunction with an inservice leak or hydrostatic test, while considering operational conditions to be in Mode 4. A notice announcing the availability of this proposed TS change using the consolidated line item improvement process was published in the Federal Register on November 27, 2006 (71 FR 68642).
2.0 REGULATORY EVALUATION
2.1 Inservice Leak and Hydrostatic Testing The Reactor Coolant System (RCS) serves as a pressure boundary and also serves to provide a flow path for the circulation of coolant past the fuel. In order to maintain RCS integrity,Section XI of the American Society of Mechanical Engineers (ASME) Pressure Vessel Code requires periodic hydrostatic and leakage testing. Hydrostatic tests are required to be performed once every 10 years, and leakage tests are required to be performed each refueling outage. Appendix G to Title 10, Code of Federal Regulations (10 CFR), Part 50, states that pressure tests and leak tests of the reactor vessel that are required by Section XI of the ASME Pressure Vessel Code must be completed before the core is critical.
NUREG-1433, General Electric Plants, BWR/4, Revision 3, Standard Technical Specifications (STSs) and NUREG-1434, General Electric Plants, BWR/6, Revision 3, STSs both currently contain LCO 3.10.1, Inservice Leak and Hydrostatic Testing Operation. LCO 3.10.1 was created to allow for hydrostatic and leakage testing to be conducted while in Mode 4 with average reactor coolant temperature greater than 212 degrees F, provided certain secondary containment LCOs are met.
TSTF-484, Revision 0, Use of TS 3.10.1 for Scram Time Testing Activities, modifies LCO 3.10.1 to allow a licensee to implement LCO 3.10.1, while hydrostatic and leakage testing is being conducted, should average reactor coolant temperature exceed 212 degrees F during testing. This modification does not alter current requirements for hydrostatic and leakage testing as required by Appendix G to 10 CFR Part 50.
2.2 Control Rod Scram Time Testing Control rods function to control reactor power level and to provide adequate excess negative reactivity to shut down the reactor from any normal operating or accident condition at any time during core life. The control rods are scrammed by using hydraulic pressure exerted by the control rod drive (CRD) system. The Browns Ferry Units were designed and constructed based on the proposed General Design Criteria (GDC) published by the Atomic Energy Commission in the Federal Register (32 FR 10213) on July 11, 1967 (draft GDC). TVA reviewed the differences between the draft GDC and final GDC. As discussed in the Nuclear Regulatory Commission (NRC, the Commission) Staff Requirements Memorandum for SECY-92-223, the Commission decided not to apply the final GDC to plants with construction permits issued prior to May 21, 1971. As the Browns Ferry units were licensed before the final GDC were formally adopted, these units were evaluated on a plant-specific basis, determined to be safe, and licensed by the Commission. The draft GDC applicable to these units are maintained in Appendix A to the Updated Final Safety Analyses Report. According to draft-GDC 6, the reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences. The scram reactivity used in design basis accidents and transient analyses is based on an assumed control rod scram time.
NUREG-1433, General Electric Plants, BWR/4, Revision 3, STSs and NUREG-1434, General Electric Plants, BWR/6, Revision 3, STSs both currently contain surveillance requirements (SRs) to conduct scram time testing when certain conditions are met in order to ensure that draft GDC-6 is satisfied. SR 3.1.4.1 requires scram time testing to be conducted following a shutdown greater than 120 days, while SR 3.1.4.4 requires scram time testing to be conducted following work on the CRD system or following fuel movement within the affected core cell.
Both SRs must be performed at reactor steam dome pressure greater than or equal to 800 pounds per square inch gauge (psig) and prior to exceeding 40-percent rated thermal power (RTP).
TSTF-484, Revision 0, Use of TS 3.10.1 for Scram Time Testing Activities, would modify LCO 3.10.1 to allow SR 3.1.4.1 and SR 3.1.4.4 to be conducted in Mode 4 with average reactor coolant temperature greater than 212 degrees F. Scram time testing would be performed in accordance with LCO 3.10.4, Single Control Rod Withdrawal - Cold Shutdown. This modification to LCO 3.10.1 does not alter the means of compliance with draft GDC-6.
3.0 TECHNICAL EVALUATION
The existing provisions of LCO 3.10.1 allow for hydrostatic and leakage testing to be conducted while in Mode 4 with average reactor coolant temperature greater than 212 degrees F, while imposing Mode 3 secondary containment requirements. Under the existing provision, LCO 3.10.1 would have to be implemented prior to hydrostatic and leakage testing. As a result, if LCO 3.10.1 was not implemented prior to hydrostatic and leakage testing, hydrostatic and leakage testing would have to be terminated if average reactor coolant temperature exceeded 212 degrees F during the testing. TSTF-484, Revision 0, Use of TS 3.10.1 for Scram Time Testing Activities, modifies LCO 3.10.1 to allow a licensee to implement LCO 3.10.1, while hydrostatic and leakage testing is being conducted, should average reactor coolant temperature exceed 212 degrees F during testing. As discussed in the TSTF Safety Evaluation (SE), the modification will allow completion of testing without the potential for interrupting the test in order to reduce reactor vessel pressure, cool the RCS, and restart the test below 212 degrees F. Since the current LCO 3.10.1 allows testing to be conducted while in Mode 4 with average reactor coolant temperature greater than 212 degrees F, the proposed change does not introduce any new operational conditions beyond those currently allowed.
SR 3.1.4.1 and SR 3.1.4.4 require that control rod scram time be tested at reactor steam dome pressure greater than or equal to 800 psig and before exceeding 40-percent RTP.
Performance of control rod scram time testing is typically scheduled concurrent with inservice leak or hydrostatic testing while the RCS is pressurized. Because of the number of control rods that must be tested, it is possible for the inservice leak or hydrostatic test to be completed prior to completing the scram time test. Under existing provisions, if scram time testing cannot be completed during the LCO 3.10.1 inservice leak or hydrostatic test, scram time testing must be suspended. Additionally, if LCO 3.10.1 is not implemented and average reactor coolant temperature exceeds 212 degrees F while performing the scram time test, scram time testing must also be suspended. In both situations, scram time testing is resumed during startup and is completed prior to exceeding 40-percent RTP. TSTF-484, Revision 0, Use of TS 3.10.1 for Scram Time Testing Activities, modifies LCO 3.10.1 to allow a licensee to complete scram time testing initiated during inservice leak or hydrostatic testing. As stated earlier (and as discussed in the TSTF SE), since the current LCO 3.10.1 allows testing to be conducted while in Mode 4 with average reactor coolant temperature greater than 212 degrees F, the proposed change does not introduce any new operational conditions beyond those currently allowed. Completion of scram time testing prior to reactor criticality and power operations results in a more conservative operating philosophy with attendant potential safety benefits.
It is acceptable to perform other testing concurrent with the inservice leak or hydrostatic test provided that this testing can be performed safely and does not interfere with the leak or hydrostatic test. However, it is not permissible to remain in TS 3.10.1 solely to complete such testing following the completion of inservice leak or hydrostatic testing and scram time testing.
Since the tests are performed with the reactor pressure vessel (RPV) nearly water solid, at low decay heat values, and near Mode 4 conditions, the stored energy in the reactor core will be very low. Small leaks from the RCS would be detected by inspections before a significant loss of inventory would occur. In addition, two low-pressure emergency core cooling systems (ECCSs) injection/spray subsystems are required to be operable in Mode 4 by TS 3.5.2, ECCS-Shutdown. In the event of a large RCS leak, the RPV would rapidly depressurize and allow operation of the low pressure ECCSs. The capability of the low pressure ECCSs would be adequate to maintain the fuel covered under the low decay heat conditions during these tests. Also, LCO 3.10.1 requires that secondary containment and standby gas treatment system be operable and capable of handling any airborne radioactivity or steam leaks that may occur during performance of testing.
The protection provided by the normally required Mode 4 applicable LCOs, in addition to the secondary containment requirements required to be met by LCO 3.10.1, minimizes potential consequences in the event of any postulated abnormal event during testing. In addition, the requested modification to LCO 3.10.1 does not create any new modes of operation or operating conditions that are not currently allowed. The NRC staff has determined that no factors specific to BFN change the applicability of the TSTF SE analysis with respect to the proposed changes.
Therefore, the NRC staff finds the proposed change acceptable.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Alabama State official was notified of the proposed issuance of the amendment. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding issued on February 13, 2007 (72 FR 6791). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
6.0 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.
7.0 REFERENCES
1.
NUREG-1433, General Electric Plants, BWR/4, Revision 3, Standard Technical Specifications (STS), August 31, 2003 2.
NUREG-1434, General Electric Plants, BWR/6, Revision 3, Standard Technical Specifications (STS), August 31, 2003 3.
Request for Additional Information (RAI) Regarding TSTF-484, April, 7, 2006, ADAMS Accession Number ML060970568 4.
Response to NRC Requests for Additional Information Regarding TSTF-484, June 5, 2006, ADAMS Accession Number ML061560523 5.
TSTF-484 Revision 0, Use of TS 3.10.1 for Scram Times Testing Activities, May 5, 2005, ADAMS Accession Number ML052930102 6.
TSTF Response to NRC Notice for Comment, September 20, 2006, ADAMS Accession Number ML062650171 Principal Contributor: Trent L. Wertz Date: April 16, 2007
Mr. Preston D. Swafford BROWNS FERRY NUCLEAR PLANT Tennessee Valley Authority cc:
Mr. Ashok S. Bhatnagar, Senior Vice President Nuclear Operations Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 Mr. Larry S. Bryant, Vice President Nuclear Engineering & Technical Services Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 Brian OGrady, Site Vice President Browns Ferry Nuclear Plant Tennessee Valley Authority P.O. Box 2000 Decatur, AL 35609 General Counsel Tennessee Valley Authority ET 11A 400 West Summit Hill Drive Knoxville, TN 37902 Mr. John C. Fornicola, General Manager Nuclear Assurance Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 Mr. Bruce Aukland, Plant Manager Browns Ferry Nuclear Plant Tennessee Valley Authority P.O. Box 2000 Decatur, AL 35609 Mr. Masoud Bajestani, Vice President Browns Ferry Unit 1 Restart Browns Ferry Nuclear Plant Tennessee Valley Authority P.O. Box 2000 Decatur, AL 35609 Mr. Robert G. Jones, General Manager Browns Ferry Site Operations Browns Ferry Nuclear Plant Tennessee Valley Authority P.O. Box 2000 Decatur, AL 35609 Mr. Larry S. Mellen Browns Ferry Unit 1 Project Engineer Division of Reactor Projects, Branch 6 U.S. Nuclear Regulatory Commission 61 Forsyth Street, SW.
Suite 23T85 Atlanta, GA 30303-8931 Ms. Beth A. Wetzel, Manager Corporate Nuclear Licensing and Industry Affairs Tennessee Valley Authority 4X Blue Ridge 1101 Market Street Chattanooga, TN 37402-2801 Mr. William D. Crouch, Manager Licensing and Industry Affairs Browns Ferry Nuclear Plant Tennessee Valley Authority P.O. Box 2000 Decatur, AL 35609 Senior Resident Inspector U.S. Nuclear Regulatory Commission Browns Ferry Nuclear Plant 10833 Shaw Road Athens, AL 35611-6970 State Health Officer Alabama Dept. of Public Health RSA Tower - Administration Suite 1552 P.O. Box 303017 Montgomery, AL 36130-3017 Chairman Limestone County Commission 310 West Washington Street Athens, AL 35611 Mr. Robert H. Bryan, Jr., General Manager Licensing and Industry Affairs Tennessee Valley Authority 4X Blue Ridge 1101 Market Street Chattanooga, TN 37402-2801