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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20217D2721999-10-12012 October 1999 Safety Evaluation Supporting Amends 248 & 239 to Licenses DPR-77 & DPR-79,respectively ML20217B3651999-10-0606 October 1999 Safety Evaluation Supporting Amends 247 & 238 to Licenses DPR-77 & DPR-79,respectively ML20212J6311999-10-0101 October 1999 SER Accepting Request for Relief from ASME Boiler & Pressure Vessel Code,Section Xi,Requirements for Certain Inservice Insp at Plant,Unit 1 ML20212F4761999-09-23023 September 1999 Safety Evaluation Supporting Amends 246 & 237 to Licenses DPR-77 & DPR-79,respectively ML20212F0831999-09-23023 September 1999 Safety Evaluation Granting Relief from Certain Weld Insp at Sequoyah Nuclear Plant,Units 1 & 2 Pursuant to 10CFR50.55a(a)(3)(ii) for Second 10-year ISI Interval ML20196J8521999-06-28028 June 1999 Safety Evaluation Authorizing Proposed Alternative to Use Iqis for Radiography Examinations as Provided for in ASME Section III,1992 Edition with 1993 Addenda,Pursuant to 10CFR50.55a(a)(3)(i) ML20206G3751999-05-0404 May 1999 Safety Evaluation Supporting Amends 244 & 235 to Licenses DPR-77 & DPR-79,respectively ML20205N0361999-04-12012 April 1999 Safety Evaluation Supporting Amend 234 to License DPR-79 ML20204E8211999-03-16016 March 1999 Safety Evaluation Supporting Amends 243 & 233 to Licenses DPR-77 & DPR-79,respectively ML20206U4331999-02-0909 February 1999 Safety Evaluation Supporting Amends 242 & 232 to Licenses DPR-77 & DPR-79,respectively ML20196B0231998-11-19019 November 1998 Safety Evaluation Supporting Amends 239 & 229 to Licenses DPR-77 & DPR-79,respectively ML20238F2961998-08-28028 August 1998 Safety Evaluation Supporting Amends 235 & 225 to Licenses DPR-77 & DPR-79,respectively ML20239A0631998-08-27027 August 1998 SER Accepting Licensee Response to GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves, for Sequoyah Nuclear Plant,Units 1 & 2 ML20236Y2091998-08-0707 August 1998 Safety Evaluation Accepting Relief Requests RP-03,RP-05, RP-07,RV-05 & RV-06 & Denying RV-07 & RV-08 ML20236S4081998-07-0101 July 1998 Safety Evaluation Supporting Amends 233 & 223 to Licenses DPR-77 & DPR-79,respectively ML20248L1961998-06-0808 June 1998 Safety Evaluation Supporting Amends 232 & 222 to Licenses DPR-77 & DPR-79,respectively ML20217K4471998-04-27027 April 1998 Safety Evaluation Supporting Requests for Relief 1-ISI-2 (Part 1),2-ISI-2 (Part 2),1-ISI-5,2-ISI-5,1-ISI-6,1-ISI-7, 2-ISI-7,ISPT-02,ISPT-04,ISPT-06,ISPT-07,ISPT-8,ISPT-01 & ISPT-05 ML20216E4701998-03-0909 March 1998 Safety Evaluation Approving Exemption from Updated FSAR Requirements of 10CFR50.71(e)(4) for Sequoyah Nuclear Plant,Units 1 & 2 ML20203K6811998-02-20020 February 1998 Safety Evaluation Supporting Amends 231 & 221 to Licenses DPR-77 & DPR-79,respectively ML20198P9171998-01-13013 January 1998 Safety Evaluation Supporting Amend 230 to License DPR-77 ML20217D7881997-09-29029 September 1997 Safety Evaluation Supporting Amends 229 & 220 to Licenses DPR-77 & DPR-79,respectively ML20211G2061997-09-23023 September 1997 Safety Evaluation Supporting Amends 228 & 219 to Licenses DPR-77 & DPR-79,respectively ML20210J5951997-08-12012 August 1997 Safety Evaluation Supporting Amends 227 & 218 to Licenses DPR-77 & DPR-79,respectively ML20140F0311997-06-10010 June 1997 Safety Evaluation Supporting Amends 224 & 215 to Licenses DPR-77 & DPR-79,respectively ML20138D2581997-04-28028 April 1997 Safety Evaluation Authorizing Licensee Proposed Alternative to Use 1989 Edition of ASME Boiler & Pressure Vessel Code, Section XI for Performance of Containment Repair & Replacement Activities Until 970909 ML20097D4191996-02-0707 February 1996 Safety Evaluation Supporting Amends 218 & 208 to Licenses NPF-77 & NPF-79,respectively ML20095F9891995-12-11011 December 1995 Safety Evaluation Supporting Amends 216 & 206 to Licenses DPR-77 & DPR-79,respectively ML20094N5341995-11-21021 November 1995 Safety Evaluation Supporting Amends 215 & 205 to Licenses DPR-77 & DPR-79,respectively ML20092H0811995-09-15015 September 1995 Safety Evaluation Supporting Amends 211 & 201 to Licenses DPR-77 & DPR-79,respectively ML20087L4671995-08-22022 August 1995 Safety Evaluation Supporting Amends 207 & 197 to Licenses DPR-77 & DPR-79,respectively ML20086C2861995-06-29029 June 1995 Safety Evaluation Supporting Amends 205 & 195 to Licenses DPR-77 & DPR-79,respectively ML20087J3291995-04-28028 April 1995 Safety Evaluation Supporting Amends 197 & 188 to Licenses DPR-77 & DPR-79,respectively ML20077L2421994-12-27027 December 1994 Safety Evaluation Supporting Amends 192 & 184 to Licenses DPR-77 & DPR-79,respectively ML20078H2911994-11-0909 November 1994 Safety Evaluation Supporting Amends 190 & 182 to Licenses DPR-77 & DPR-79,respectively ML20078B6421994-10-20020 October 1994 Safety Evaluation Supporting Amends 189 & 181 to Licenses DPR-77 & DPR-79,respectively ML20071K5331994-07-26026 July 1994 Safety Evaluation Supporting Amends 185 & 177 to Licenses DPR-77 & DPR-79,respectively ML20065D9661994-03-31031 March 1994 Safety Evaluation Supporting Amends 178 & 169 to Licenses DPR-77 & DPR-79,respectively ML20067B5891994-02-10010 February 1994 Safety Evaluation Supporting Amends 176 & 167 to Licenses DPR-77 & DPR-79,respectively ML20058G6021993-11-29029 November 1993 Safety Evaluation Supporting Amends 173 & 164 to Licenses DPR-77 & DPR-79,respectively ML20059L8301993-11-0909 November 1993 Safety Evaluation Supporting Amend 162 to License DPR-79 ML20057F8441993-10-14014 October 1993 SER Granting Relief Giving Due Consideration to Burden Upon Licensee That Could Result If Requirements Imposed on Facility ML20057D6351993-09-28028 September 1993 SER Granting Relief as Requested for Both ISPT-2 & ISPT-3 Per 10CFR50.55a(a)(3)(i) & 10CFR50.55a(g)(6)(i) ML20057D5321993-09-28028 September 1993 SER Granting Licensee 921117 Relief Requests ISPT-2 & ISPT-3 Re Inservice Pressure Test Program ML20056E0041993-08-0202 August 1993 Safety Evaluation Supporting Amends 169 & 159 to Licenses DPR-77 & DPR-79,respectively ML20128K0221993-02-11011 February 1993 SE Accepting Util Justification for Break Exclusion of Main Steam Lines in Valve Vaults Provisionally Until End of Refueling Outages ML20128E9161993-01-0606 January 1993 SE Approving Request for Relief from ASME Requirements Re First 10-yr Interval ISI Plan ML20125C9281992-12-0808 December 1992 Safety Evaluation Supporting Amends 165 & 155 to Licenses DPR-77 & DPR-79,respectively ML20114D4391992-08-31031 August 1992 Evaluation Supporting Relief Request from ASME Section XI Code Requirements for ten-yr Insp Intervals of Reactor Vessel ML20099C6981992-07-24024 July 1992 Safety Evaluation Supporting Amends 160 & 150 to Licenses DPR-77 & DPR-79,respectively ML20091C9041992-03-30030 March 1992 Safety Evaluation Supporting Amend 146 to License DPR-79 1999-09-23
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEAR05000327/LER-1999-002-03, :on 990916,EDG Started as Result of Cable Being Damaged During Installation of Thermo-Lag for Kaowool Upgrade Project.Caused by Inadequate pre-job Briefing. Involved Individuals Were Counseled on Event1999-10-15015 October 1999
- on 990916,EDG Started as Result of Cable Being Damaged During Installation of Thermo-Lag for Kaowool Upgrade Project.Caused by Inadequate pre-job Briefing. Involved Individuals Were Counseled on Event
ML20217D2721999-10-12012 October 1999 Safety Evaluation Supporting Amends 248 & 239 to Licenses DPR-77 & DPR-79,respectively ML20217B3651999-10-0606 October 1999 Safety Evaluation Supporting Amends 247 & 238 to Licenses DPR-77 & DPR-79,respectively ML20212J6311999-10-0101 October 1999 SER Accepting Request for Relief from ASME Boiler & Pressure Vessel Code,Section Xi,Requirements for Certain Inservice Insp at Plant,Unit 1 ML20217G3721999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Sequoyah Nuclear Plant.With ML20212F0831999-09-23023 September 1999 Safety Evaluation Granting Relief from Certain Weld Insp at Sequoyah Nuclear Plant,Units 1 & 2 Pursuant to 10CFR50.55a(a)(3)(ii) for Second 10-year ISI Interval ML20212F4761999-09-23023 September 1999 Safety Evaluation Supporting Amends 246 & 237 to Licenses DPR-77 & DPR-79,respectively ML20212C4761999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for Sequoyah Nuclear Plant.With ML20212A1841999-08-25025 August 1999 Errata Pages for Rev 0 of WCAP-15224, Analysis of Capsule Y from TVA Sequoyah Unit 1 Reactor Vessel Radiation Surveillance Program ML20210L4361999-08-0202 August 1999 Cycle 9 12-Month SG Insp Rept ML20210L4451999-07-31031 July 1999 Unit-2 Cycle 10 Voltage-Based Repair Criteria 90-Day Rept ML20216E3781999-07-31031 July 1999 Monthly Operating Repts for July 1999 for Sequoyah Nuclear Plant,Units 1 & 2.With ML20210G6631999-07-28028 July 1999 Cycle 9 90-Day ISI Summary Rept ML20196H8621999-06-30030 June 1999 NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data, June 1999 Rept ML20209H3831999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Sequoyah Nuclear Plant.With ML20211F9031999-06-30030 June 1999 Cycle 9 Refueling Outage ML20196J8521999-06-28028 June 1999 Safety Evaluation Authorizing Proposed Alternative to Use Iqis for Radiography Examinations as Provided for in ASME Section III,1992 Edition with 1993 Addenda,Pursuant to 10CFR50.55a(a)(3)(i) ML20195K2951999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Sequoyah Nuclear Plant,Units 1 & 2.With 05000327/LER-1998-003-01, :on 981109,Vital Inverter 1-IV Tripped.Caused by Failed Oscillator Board with Bad Solder Joint,Attributed to Mfg Defect.Replaced Component & Returned Inverter to Operable Status1999-05-27027 May 1999
- on 981109,Vital Inverter 1-IV Tripped.Caused by Failed Oscillator Board with Bad Solder Joint,Attributed to Mfg Defect.Replaced Component & Returned Inverter to Operable Status
05000327/LER-1999-001-04, :on 990415,exceedance of AOT Occurred Due to Failure of Centrifugal Charging Pump.Rotating Element Was Replaced,Testing Was Completed & Pump Was Returned to Svc1999-05-11011 May 1999
- on 990415,exceedance of AOT Occurred Due to Failure of Centrifugal Charging Pump.Rotating Element Was Replaced,Testing Was Completed & Pump Was Returned to Svc
ML20206Q8951999-05-0505 May 1999 Rev 0 to L36 990415 802, COLR for Sequoyah Unit 2 Cycle 10 ML20206G3751999-05-0404 May 1999 Safety Evaluation Supporting Amends 244 & 235 to Licenses DPR-77 & DPR-79,respectively ML20206R5031999-04-30030 April 1999 Monthly Operating Repts for April 1999 for Sequoyah Units 1 & 2.With ML20205N0361999-04-12012 April 1999 Safety Evaluation Supporting Amend 234 to License DPR-79 ML20205P9811999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Sequoyah Nuclear Plant,Units 1 & 2.With ML20204E8211999-03-16016 March 1999 Safety Evaluation Supporting Amends 243 & 233 to Licenses DPR-77 & DPR-79,respectively ML20204C3111999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for Sequoyah Nuclear Plant,Units 1 & 2.With ML20205B6631999-02-28028 February 1999 Underground Storage Tank (Ust) Permanent Closure Rept, Sequoyah Nuclear Plant Security Backup DG Ust Sys ML20203H7381999-02-18018 February 1999 Safety Evaluation of Topical Rept BAW-2328, Blended U Lead Test Assembly Design Rept. Rept Acceptable Subj to Listed Conditions ML20206U4331999-02-0909 February 1999 Safety Evaluation Supporting Amends 242 & 232 to Licenses DPR-77 & DPR-79,respectively ML20211A2021999-01-31031 January 1999 Non-proprietary TR WCAP-15129, Depth-Based SG Tube Repair Criteria for Axial PWSCC Dented TSP Intersections ML20198S7301998-12-31031 December 1998 Cycle 10 Voltage-Based Repair Criteria 90-Day Rept ML20199G3641998-12-31031 December 1998 Monthly Operating Repts for Dec 1998 for Sequoyah Nuclear Plant,Units 1 & 2.With 05000327/LER-1998-004-02, :on 981120,failure to Perform Surveillance within Required Time Interval,Was Determined.Caused by Leaking Vent Valve.Engineering Personnel Evaluated Alternative Methods for Performing Channel Check1998-12-21021 December 1998
- on 981120,failure to Perform Surveillance within Required Time Interval,Was Determined.Caused by Leaking Vent Valve.Engineering Personnel Evaluated Alternative Methods for Performing Channel Check
05000327/LER-1998-003-04, :on 981109,automatic Reactor Trip Occurred. Caused by Component in Bridge Circuit of Vital Inverter Failed.Inverter Bridge Circuit Replaced1998-12-0909 December 1998
- on 981109,automatic Reactor Trip Occurred. Caused by Component in Bridge Circuit of Vital Inverter Failed.Inverter Bridge Circuit Replaced
ML20197J5621998-12-0303 December 1998 Unit 1 Cycle 9 90-Day ISI Summary Rept ML20197K1161998-11-30030 November 1998 Monthly Operating Repts for Nov 1998 for Sequoyah Nuclear Plant,Units 1 & 2.With ML20196B0231998-11-19019 November 1998 Safety Evaluation Supporting Amends 239 & 229 to Licenses DPR-77 & DPR-79,respectively 05000328/LER-1998-002-05, :on 981016,turbine Trip Occurred Followed by Reactor Trip.Caused by Generator Lockout.Mod Being Evaluated to Physically Isolate Relays from Vibration of Transformers & Adding Two of Two Logic for Actuation of Sudden Pressure1998-11-10010 November 1998
- on 981016,turbine Trip Occurred Followed by Reactor Trip.Caused by Generator Lockout.Mod Being Evaluated to Physically Isolate Relays from Vibration of Transformers & Adding Two of Two Logic for Actuation of Sudden Pressure
ML20195F8061998-10-31031 October 1998 Monthly Operating Repts for Oct 1998 for Sequoyah Nuclear Plant.With ML20154H6091998-09-30030 September 1998 Monthly Operating Repts for Sept 1998 for Sequoyah Nuclear Plant,Units 1 & 2.With 05000328/LER-1998-001-05, :on 980827,turbine Trip Occurred Followed by Reactor Trip.Caused by Failure of Sudden Pressure Relay on B Phase Main Transformer.Control Room Operators Responded & Removed,Inspected & Replaced Failed Relays1998-09-28028 September 1998
- on 980827,turbine Trip Occurred Followed by Reactor Trip.Caused by Failure of Sudden Pressure Relay on B Phase Main Transformer.Control Room Operators Responded & Removed,Inspected & Replaced Failed Relays
ML20154H6251998-09-17017 September 1998 Rev 0 to Sequoyah Nuclear Plant Unit 1 Cycle 10 Colr ML20153B0881998-08-31031 August 1998 Monthly Operating Repts for Aug 1998 for Sequoyah Nuclear Plant.With ML20238F2961998-08-28028 August 1998 Safety Evaluation Supporting Amends 235 & 225 to Licenses DPR-77 & DPR-79,respectively ML20239A0631998-08-27027 August 1998 SER Accepting Licensee Response to GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves, for Sequoyah Nuclear Plant,Units 1 & 2 05000327/LER-1998-002-03, :on 980716,inadequate Surveillance Testing Was Discovered.Caused by Misinterpretation of ANSI Standard. Revised Appropriate Procedures to Provide Required Guidance1998-08-14014 August 1998
- on 980716,inadequate Surveillance Testing Was Discovered.Caused by Misinterpretation of ANSI Standard. Revised Appropriate Procedures to Provide Required Guidance
ML20236Y2091998-08-0707 August 1998 Safety Evaluation Accepting Relief Requests RP-03,RP-05, RP-07,RV-05 & RV-06 & Denying RV-07 & RV-08 ML20237B5221998-07-31031 July 1998 Monthly Operating Repts for July 1998 for Snp ML20237A4411998-07-31031 July 1998 Blended Uranium Lead Test Assembly Design Rept 1999-09-30
[Table view] |
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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555 0001 S
%4.....p SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 247 TO FACILITY OPERATING LICENSE NO. DPR-77 AND AMENDMENT NO. 238 TO FACILITY OPERATING LICENSE NO. DPR-79 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT. UNITS 1 AND 2
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DOCKET NOS. 50-327 AND 50-328
1.0 INTRODUCTION
By application dated February 26,1999, the Tennescee Valley Authority (TVA, the licensee) proposed amendments to the Techni' al Specifications (TS) for Sequoyah Nuclear Plant (SON) c Units 1 and 2. The requested changes would relocate TS 3/4.7.6, " Flood Protection Plan," from the SON TS to the SQN Technical Requirements Manual (TRM). In addition, the appropriate TS Bases sections and index pages would be revised to reflect this change.
2.0 BACKGROUND
SON's Flood Protection Plan is designed to minimize impact of floods above plant grade on safety-related facilities. Currently, the SON's TS provide the limiting conditions of operation (LCO) and surveillance requirements to verify the implementation of the Flood Protection Plan to minimize the consequences of floods. The TS also include an LCO for occurrence of a seismic event or recognizable seismic activity in the east Tennessee region. Procedures for predicting
' rainfall floods, arrangements to warn of upstream dam failure floods, and lead times available and types of action to be taken to meet related safety requirements for both sources of flooding are described therein.
TVA requests the proposed change W remove requirements associated with tha F! cod Protection from the SON TS on the basis that they do not meet the criteria in Title 10, Code of Federal Reaulations (10 CFR), Section 50.36. TVA's position is that the proposed change is consistent with the U.S. Nuclear Regulatory Commission's (NRC's) " Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors," published in the Federal Reaister on July 22,1993 (58 FR 39132). The NRC Final Policy Statement states that TS requiremt.nts that do not meet any of the screening criteria for retention may be proposed for removal from the TS and reloc sted to Ikansee-controlled documents, such as the Final Safety Analysis Report or TRM. TVNs proposed change would allow revisions to the Flood Protection Plan, in accordance with 10 CFR 50.59, without requiring a license amendment request and adds flexibility to processing necessary changes.
9910120218 991006 l
PDR ADOCK 05000327 1
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f 3.0. EVALUATION The Commission requires that LCOs meeting one or more of the four criteria stated in 10 CFR 50.5S(c)(2)(ii) must be included in the plant's TS. The NRC staff has reviewed TVA's proposed TS change and concludes that it is consistent with the guidance in the Commission's Final Policy Statement on Technical Specifications, as discussed above, and with NUREG-1431, Revision i, " Standard Technical Specifications, Westinghouse Plants," dated April 1995.
NUREG-1431 does not include Flood Protection Plan TS requirements because tnis plan does not meet the criteria in 10 CFR 50.36. In addition,'IVA evaluated SON's current flood protection TS requirements against the criteria of 10 CFR 50.36. The following discussions address the applicability of the 10 CFR 50.36 criteria to SON's TS for the Flood Protection Plan.
Criterion 1: Installedinstrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation Jf the reactor coolant pressure boundary.
SON's Flood Protection Plan is designed to minimize the impact of floods above plant grade on safety-related facilities. Procedures for predicting rainfall floods, arrangements to warn of upstream-dam-failure floods, and lead times available and types of action to be taken to meet related safety requirements for both sources of flooding are features of the plan. SON's Flood Protection Plan is not instaihd instrumentation that is used to detect and indicate, in the control room, a significant abnotmal degradation of the reactor coolant pressure boundary.
Accordingly, the SON Flood Protection Plan does not satisfy Criterion 1.
l Criterion 2: A process variable, design feature or operatir;g restriction that is an initial condition of a Design Basis Accident (DBA) or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
l SQN's Flood Protection Plan is not a process variable that is an initial condition of a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission j
product barrier. Therefore, the SON Flood Protection Plan does not satisfy Criterion 2.
l Criterion 3: A structure, system or component that is part of the primary success path and which l
l'inctions or actuates to mitigate a DBA or Transient :nat either assumes the failure of or l
presents a challenge to the integrity of a fission prodact barrier.
SON's Flood Protection Plan is not a structure, system or component that is part of the primary success path for accident mitigation. In addition, the Flood Protection Plan does not function or actuate to miligate a DBA or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Accordingly, the Flood Protection Phn does not satisfy Criterion 3.
1 l
Criterion 4: A structure, system or component, which operating experience orprobabilistic l
safety assessment has shown to be significant to public health and safety.
)
Operational experience and deterministic safety assessment evaluation as identified in the SON Generic Letter 88-20 response have not shown the SON Flood Protection Plan to be significant I
to the public health and safety. Therefore, the Flood Protection Plan does not satisfy Criterion 4.
I J
. The proposed relocation of the flood protection requirements to the SON TRM is acceptable based on the above discussions. Therefore, the staff finds the amendments, as proposed by TVA to be acceptable. The relocated requirements will be controlled in accordance with those established for the TRM. These requirements include appropriate administrative controls and reviews and a 10 CFR 50.50 evaluation which will ensure changes are not implemented that would reduce the functionality of or introduce an unreviewed safety question to SON's Flood Protection Plan.
4.0 STATE CONSULTATION
in accordance with the Commission's regulations, the Tennessee State official was notified of the proposed issuance of the amendments. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is nc, significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (64 FR 14286). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
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 l
Commission's regulations, and (3) the issuance of the amendments will not be inimical to the l
common defense and security or to the health and safety of the public.
I 1
l Principal Contributor: Richard J. Laufer, NRR
]
Dated: October 6, 1999 l
1 1
l
e o
Mr. J. A. Scalice -
Tennessee Valley Authority.
SEQUOYAH NUCLEAR PLANT cc:
Mr. Karl W. Singer, Senior Vice President Mr. Pedro Salas, Manager Nuclear Operations Licen. sing and Industry Affairs Tennessee Valley Authority Sequoyah Nuclear Plant 6A Lookout Place Tennessee Valley Authority 1101 Market Street P.O. Box 2000 Chattanooga, TN 37402-2801 Soddy Daisy, TN 37379 Mr. Jack A. Bailey Mi D. L. Koehl, Plant Manager i
Vice President Sequoyah Nuclear Plant Engineering & Technical Services Tennessee Valley Authority Tennessee Valley Authority P.O. Box 2000 6A Lookout Place Soddy Daisy, TN 37379 1101 Market Street Chattanooga, TN 37402-2801 Mr. Melvin C. Shannon Senior Resident inspector Mr. Masoud Bajestani Sequoyah Nuclear Plant Site Vice President U.S. Nuclear Regulatory Commission Sequoyah Nuclear Plant 2600 Igou Ferry Road Tennessee Valley Authority Soddy Daisy, TN 37379 P.O. Box 2000 Soddy Daisy, TN 37379 Mr. Michael H. Mobley. Director TN Dept. of Erivironment & Conservation General Counsel Division of Radiological Health Tennessee Valley Authority 3rd Floor, L and C Annex ET 10H 401 Church Street
- 400 West Summit Hill Drive Nashville, TN 37243-1532 Knoxville, TN 37902 County Executive Mr. N. C. Kazanas, General Manager Hamilton County Courthouse
-Nuclear Assurance Chattanooga, TN 37402-2801 Tennessee Valley Authority SM Lookout Place 1101 Market Street Chattanooga, TN 37402-2801
' Mr. Mark J. Burzynski, Manager l
Nuclear Licensing Tennessee Valley Authority I
4X Blue Ridge I
1101 Market Street Chattanooga, TN 37402-2801 b