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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 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 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 ML20198C0211998-12-16016 December 1998 Safety Evaluation Supporting Amends 241 & 231 to Licenses DPR-77 & DPR-79,respectively ML20196C4091998-11-19019 November 1998 Safety Evaluation Supporting Amends 238 & 228 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 ML20195G3271998-11-17017 November 1998 Safety Evaluation Supporting Amends 237 & 227 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 ML20151L7981997-07-14014 July 1997 Safety Evaluation Supporting Amends 226 & 217 to Licenses DPR-77 & DPR-79,respectively ML20148S0581997-07-0101 July 1997 Safety Evaluation Supporting Amends 225 & 216 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 ML20137Y8911997-04-21021 April 1997 Safety Evaluation Supporting Amends 223 & 214 to Licenses DPR-77 & DPR-79,respectively ML20101M5761996-04-0303 April 1996 Safety Evaluation Supporting Amend 211 to License DPR-79 ML20100N5711996-03-0404 March 1996 Safety Evaluation Supporting Amends 220 & 210 to Licenses DPR-77 & DPR-79,respectively 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 ML20094D2451995-10-30030 October 1995 Safety Evaluation Supporting Amend 204 to License DPR-79 ML20093E1191995-10-11011 October 1995 Safety Evaluation Supporting Amend 214 to License DPR-77 ML20093B9471995-10-0404 October 1995 Safety Evaluation Supporting Amends 213 & 203 to Licenses DPR-77 & DPR-79,respectively ML20092N0711995-10-0202 October 1995 Safety Evaluation Supporting Amends 212 & 202 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 ML20092G7071995-09-13013 September 1995 Safety Evaluation Supporting Amends 210 & 200 to Licenses DPR-77 & DPR-79 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 ML20085G5811995-06-13013 June 1995 Safety Evaluation Supporting Amends 203 & 193 to Licenses DPR-77 & DPR-79,respectively ML20085G5091995-06-13013 June 1995 Safety Evaluation Supporting Amends 202 & 192 to Licenses DPR-77 & DPR-79,respectively ML20083L1931995-05-10010 May 1995 Safety Evaluation Supporting Amends 198 & 189 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 ML20082E4291995-04-0404 April 1995 Safety Evaluation Supporting Amends 196 & 187 to Licenses DPR-77 & DPR-79,respectively ML20078M8441995-02-0909 February 1995 Safety Evaluation Supporting Amends 195 & 186 to Licenses DPR-77 & DPR-79,respectively 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 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 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 ML20216E3781999-07-31031 July 1999 Monthly Operating Repts for July 1999 for Sequoyah Nuclear Plant,Units 1 & 2.With ML20210L4451999-07-31031 July 1999 Unit-2 Cycle 10 Voltage-Based Repair Criteria 90-Day Rept 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 ML20199G3641998-12-31031 December 1998 Monthly Operating Repts for Dec 1998 for Sequoyah Nuclear Plant,Units 1 & 2.With ML20198S7301998-12-31031 December 1998 Cycle 10 Voltage-Based Repair Criteria 90-Day Rept 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
ML20198C0211998-12-16016 December 1998 Safety Evaluation Supporting Amends 241 & 231 to Licenses DPR-77 & DPR-79,respectively 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 ML20196C4091998-11-19019 November 1998 Safety Evaluation Supporting Amends 238 & 228 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 ML20195G3271998-11-17017 November 1998 Safety Evaluation Supporting Amends 237 & 227 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
1999-09-30
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UNITED STATES s.
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SAFETY EVALUATION BY THE OFFICE OF SPECIAL PROJECTS EMPLOYEE CONCERN ELEMENT REPORT EN 23801 CONDUIT OVERFILLS AND CABLE DAMAGE TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 I.
SUBJECT Category:
Engineering (20000)
Subcategory:
Raceway Overfills and Cable Pulling (23000)
Element:
Conduit Overfills and Cable Damage (23G01)
The basis for Element Report EN 23801,'Rev. 3, dated May 12, 1987 was the generic applicability determination resulting from Watts Bar Nuclear Plant (WBN) Employee Concerns Employee Concern:
The following conduit overfills and cable damage concerns are identified as follows:
IN-85-432-001 IN-85-036-001 IN-86-310-001 IN-85-313-001 IN-85-506-001 IN-85-622-001 IN-85-685-001 IN-85-743-008 IN-86-034-001 IN-86-226-003 IN-85-642-001 IN-85-856-003 IN-86-028-002 IN-86-262-001 IN-85-832-001 IN-86-312-001 IN-85-734-001 IN-85-367-001 IN-86-262-004 IN-86-254-009 IN-86-206-001 OW-85-007-003 II.
SUMMARY
OF ISSUE The overfilling of conduits may cause cable damage during installation, over-heating of cables, and is not in accordance with the National Electric Code (NEC).
III. EVALUATION TVA reviewed documents of employee concerns, NRC investigative interviews, FSAR commitments, engineering and construction procedures for conduit overfill problems and interviewed personnel associated with scheduling and installation of the conduit raceway system.
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2-The TVA review indicated that the FSAR and the Design Criteria are.in agree-ment. Conduits shall not have a cable fill where the cross section area of the cable exceeds 40% of the cross section area of the inside of the conduit.
However, the TVA Electrical Design Standard anc' the NEC allow 53% for conduits having one cable and 31% for conduits having two cables.
Cables were manually routed in conduits by designers and documented in the computer cable schedule.
The designer performing the cable reuting was responsible for determining the total cross section area (CSA) of the cables in the conduit.
The evaluation revealed that accurate conduit fill information is not readily available and, therefore, compliance with FSAR comitment for conduit fill is not verifiable through QA documentation.
After September 1986 a procedure was issued to require a checker to verify the manual cable routing and check CSA prior to releasing the cable pull slips for cable installation.
- Further,
.I additional concerns of CSA were raised because the cable diameters used by the designers were not frnm an approved QA list, therefore, causing overfill and overheating. The employee concerns for WBN cited some specific locations, but the implied generic concerns for.SQN are general in nature.
The SQN concern addresses overall problems related to conduit overfill and cable damage with a specific concern related to conduit fill which exceeded that established in the Electrical Design Standard, SCR'SQNEEB 8529 R0.
The concerns associated with conduit overfill discussed in the TVA report are as follows:
i Cable ampacity may not be valid because the cable diameters.used by the~
designers were not from an approved QA list.
These cable diameters used, if less than actual, cculd cause conduit overfill and thus overheating.
Cable supports nay not be adequate, because cable weight were not used from a QA list.
The cable fill criteria in the FSAR and Design Criteria are not in agreement with neither the Electrical Design Standard nor the National Electrical Code for one and two cables in a conduit.
Cable damage cculd haue occurred because the ranufacturers recomrend side wall pressure may have been exceeded from excessive bends, pullbys, and jaming.
TVA determined that the cable OD differences would have no effect on the cable ampacities and darating.
Cable ampacities in conduit are a function of the-number of cables in a conduit and not the physical conduit fill.
TVA cetermined that there was no program to verify adequacy of corduit supports for overfill conduits.
TVA has retained the services of United Engineers and Construction (UE&C) to conduct a full systematic analysis of the SON cable and conduit scheduling program.
UE&C will identify any necessary corrective-
-actions required to establish the accuracy of the conduit and cable _ schedules.
The UE&C effort will include a review of the practices and procedures utilized for routing, installing, and abandening cables in conduit during SQN's design, construction, and modification phases up to present.
All corrective actions
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s required to resolve problems identified by the UE&C review will be evaluated per restart criteria.
The review will also determine conduit support adequacy.
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Those items meeting the criteria will be corrected before Unit 2 restart.
1 The remaining items will be completed as part of a long-term program after restart. The details of staff evaluation with regard to the structural / support adequacy is addressed elsewhere (CEB-16 Calculation Review Inspection Open item).
The staff concurs with this effort to determine potential root causes for discrepancies and to correct any identified problem areas resulting from these root causes.
Electric utilities are exempt from the requirements of the National Electrical Code (NEC) for those facilities which are used for electrical power generation.
However, the licensee, TVA, will revise both the FSAR and the Design Criteria to agree with the Electrical Design Standard and, NEC concerning cenduit fill requirements.
The staff concurs that TVA sFeuld follow the industry standards concerning allowable conduit fill.
Cable damage associated with pullbys and janning is addressed in Employee Concern Element Report C010900-SON and TVA's cable test program submitted for staff review on July 31, 1987.
Other Employee Concerns for conduit design and installation are addressed it: Element Report C0 19201-SQN.
IV.
CONCLUSION The NRC staff concludes that the licensee's investigatien of the concerns were adequate and their resolution of the concerns described in Elenent Report EN 23801-SQN, Revision 3, is acceptable except for identification of cable damage.
However, the staff concludes that the cable test program has adequately addressed the cable damage ccncern for purposes of restart.
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