ML19332C729

Forwards Info Re Status of Implementing USI Requirements for Facility for Which Technical Resolution Achieved Per Generic Ltr 89-21

ML19332C729
Person / Time
Site:	Sequoyah
Issue date:	11/21/1989
From:	Shell R TENNESSEE VALLEY AUTHORITY
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
REF-GTECI-MI, REF-GTECI-SC, TASK-***, TASK-OR GL-89-21, NUDOCS 8911280479
Download: ML19332C729 (7)

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a TENNESSEE VALLEY AUTHORITY CH ATT ANOOGA, TENNESSEE 37401 SN 157B Lookout Place NOV 211989 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Hashington, D.C. 20555 Gentlemen:

In the Matter of ) Docket Nos. 50-327 Tennessee Valley Authority ) 50-328 SEQUOYAH NUCLEAR PLANT (SQN) - NRC GENERIC LETTER 89-21, REQUEST FOR INFORMATION CONCERNING STATUS OF IMPLEMENTATION Of UNRESOLVED SAFETY ISSUE (USI) REQUIREMENTS As requested in the subject generic letter, Enclosures 1 and 2 provide the

-status of the USI requirements for-SQN Units 1 and 2, respectively. For items in the enclosures that are not complete, the date is either a date that is a previous TVA commitment or a projected date for completion. Theprojected dates are not % sidered as new commitment dates.

Please direct questions concerning this issue to Kathy S. Whitaker at (615) 843-7748.

Very truly yours, TENNESSEE VALLEY AUTHORITY Y'

eManager, Nuclear Licensing and Regulatory Affairs Enclosures cc (Enclosures):

Ms. S. C. Black, Assistant Director for Projects TVA Projects Division U.S. Nuclear Regulatory Commission One White Flint, North

/ 11555 Rockville Pike Rockville, Maryland 20852 8 Mr. B. A.' Hilson, Assistant Director

.,jga.

0 for Inspection Programs

-o TVA Projects Division

$$ U.S. Nuclear Regulatory Commission g Region II ho 101 Marietta Street, NH, Suite 2900 Atlanta, Georgia 30323 /

oc 'Ok N NRC Resident Inspector b'd E8

1. 0 a-Sequoyah Nuclear Plant 2600 Igou Ferry _ Road f Soddy Daisy,. Tennessee 37379 An Equal Opportunity Employer

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e Enclosure 1 -

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'SEOUOYAH NUCLEAR ptANT - 1MIT'l Status of'unresalved safet, issues for Which A final Technical Resolution has been__Achievsd J

-USI/MPA Number Title status /Date~ Remarks A-1 Water Hasuner MC TVA completed preoperational testing of the steam generator feedring design. -

x A-2/- Asyssnetric Blowdown C-7/89 MRC approved TVA's request t'o apply TA D 19 . toads on Reactor Primary Teak-before-break technology to SQN Coolant Systems primary loop piping.

A-3 Westinghouse Steam ' C-12/87 Procedures have been implemented to Generator Tube Integrity address full feagth steast generator tube.

inspections. 72-month inspection intervals. and a compre % sive condenser program.

A-4 CE Steam Generator Tube NA - Not applicable to Westinghouse reactors.

Integrity A-5 B&W Steam Generator NA Not applicable to Westinghouse reactors.

Tube Integrity T.-6 Mark I Containment NA .Not applicable to pressurized water Short-Term program reactors.

A-7/ Mark I Long-Tem NA Not applicable to pressurized esater D-O' Program reactors.

l A-8 Mark II Containment NA Not applicable to pressurized water Pool Dynamic Loads reactors.

C - Complete NC - No Changes Necessary NA - Not Applicable

. -f - Incomplete

. E - Evaluating Actions Required -

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USI/MPA' ,. .

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Status /Date Remarks

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N_umbe r Title A-9' Anticipated Transients _ I-6/90' Actions.will-be completed during the Without Scram Cycle 4 refueling outage.

A-10/ BWR Feedwater Nozzle NA :Not applicable to pressurized water MPA B-25 Cracking reactors.

A-II Reactor Vessel Material C-2/80 TVA has used low leakage core designs for Toughness the last two cycles to reduce neutron flux at the reactor vessel.

A 12 Fracture Toughrtess of NC The resciution of this issue contained Steam Generator and no backfit requirements.

Reactor Coolant Pump Supports A-17 Systems Interactions NC SQN's IPE will address internal flooding.

51D-1.3.1 outlines the NER program, and AI-18.18 addresses telephone notification to NRC and preparation of LERs.

A-24/ Qualification of Class IE C-12/88 Equipment qualifications were completed MPA B-60 Safety-Related Equipment during the extended outage.

A-26/ Reactor Vessel Pressure I-3/91 Revised setpoints to be provided in MPA B-04 Transient Protection accordance with NRC Generic letter (GL) 88-11 by 3/91.

A-31 Residual' Heat Removal NC Original residual heat removal system Shutdown Requirements design was evaluated and found acceptable by NRC.

A-36/ Control of Heavy Loads I-7/92 Initial inspections'of welds and lifting C-10 Near Spent Fuel devices are still outstanding.

C-15 A-19 Deterinination of SRV NA Not applicable to pressurized water Pool Dynamic loads rear. tors.

and Pressure Transients b

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Number Title Status /Date Remarks' _

A-40' Seismic Design Criteria - C-2/88 SON calculations for flexible vertical trnks have been revised to address NRC concerns about tant flexibility.

A-42/ ' Pipe Cracks in Boiling NA Not applicable to pressurized water HPA B-05 Water Reactors reactors.

A-43 Containment Emergency C-9/87 Westinghcuse study and Norris Laboratories Sump Performance evaluation indicated acceptable sump performance.

AM Station Blackout I-6/92 TVA completion is dependent on issuance of SER by NRC.

A 45 Shutdown Decay Heat NC Vulnerability to loss of decay heat Removal Requirements removal is to be addressed by SQM's IPE.

A-46 Seismic Qualification I-6/90 Replacement Items Project is scheduled for of Equipment in f ull ' implementation by Cycle 4.

Operating Plants A-47 Safety Implication E-3/90 NRC GL 89-19 is under review.

of Control Systems A-48 Hydrogen Control Measures C-II/84 Redundant trains of ignitors have been and Effects of Hydrogen installed. Future analytical work i?.

Burns on Safety Equipment planned, and the schedule is tied to the completion of analysis work by Duke Power Company. Reference TVA letters dated 2/2/87 and 1/29/68 and NRC letter dated 3/31/87.

A-49 Pressurized Thermal Shock I-3/91 - The remaining action is to submit revised reactor pressure vessel heatop and cooldown curves in accordance with NRO GL 88-11.

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SE000YAH NUCLEAR PLANT - LMIT 2 : e r. > ~ , ,

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~ Status of' Unresolved Safety Issues'for Which . . . .

.-A Final Technical Resolution has been Achievgd... -

'USI/MPA.

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' Number Title' Status /Date Remarks A-1 - Water Hammer ~ NC . l TVA completed preoperational testing of .R.

. the steam generator feedring design. ,

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A-2/  : Asynenetric Blowdown .

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C-7/89 '- NRC approved TVA's request to apply..

leak-before-break technology to SQN -

MPA D-10 .

Loads on Reactor Primary -

Coolant Systems ' primary loop piping.'

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A-3 Westinghouse Steam C-12/871 ' Procedures have been implemented to-address ~ f ull-length steam generator tube - ;T Generator Tube Integrity:

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- inspections,'72-month inspection intervals, and a comprehensive' condenser program. ~_

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A-4 CE Steam Generator Tube NA Not applicable to Westinghouse reactors.

Integrity A-5 B&W Steam Generator NA Not applicable to Westingnouse reactors. ,'

Tube Integrity A-6 Mark I Containment NA' Not applicable to pressurized water Short-Term Program reactors.

A-7/ Ma-k I long-Term. NA' ' Not applicable to pressurized water s D-01 Program reactors. -

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A-8 Mark II Containment MA Not applicable to pressurized water Pool Dynamic loads reactors.

C - Complete ,

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NC - No Changes Necessary NA - Mot Applicable i.- Incomplete ,

C - Evaluating Actions Required 1

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'. Number ' Title' .Stattis/Da te " Remarkt x .;

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- A-9 Anticipated Transients - :I-12/90 < Actions ~willL-be completed during the. -

- Cycle 4 refueling outage.

. Without' Scram: , q

^ '

'A-10/?  : BWR Feedwater Nozzle ' NA:: Not4 applicable. to pressurized water '

MPA B-25 Cracking " reactors. -

, 4:

Reactor Vessel Material - . TVA has used low leakage core designs for-.

A-11 NC -

Toughness the:last-two cycles to reduce' neutron flum

-at" the reactor vessel.

A '2 Fracture Toughness of NC The resolution of this issue contained :

Steam Generator and no'backfit requirements.

Reactor Coolant Pump , _

Supports *...

A-17 Systems Interactions NC SQM's IPE will' address internal flooding, STD-1.3.1 outlines the NER program, and AI-18.18 addresses telephone notification to NRC and preparation of LERs.

A-24/ Qualification of Class lE C-2/88 Equipment qualifications were completed ,

MPA B-60 Safety-Related Equipment ' during the extended' outage.

A-26/ Reactor Vessel Pressure I-3/91-~ Revised setpoints to be provided in MPA B-04 Transient Protection accordance with NRC GL 88-11 by 3/91.

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A-31 Residual Heat Removal NC . Original residual heat' removal system Shutdown Requirements design was evaluated and found acceptable by NRC.

A-36/ Control of Heavy Loads I-6/91 ~ Initial inspections of welds and lifting-C-10, Near Spent fuel devices are still outstanding.

C-15 A-39 Determinaticn of SRV NA Not applicable to pressurized water Pool Dynamic tcads .

reactors.

and Pressure Transients O

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-+. s jA - Seis.mic Design Criteria : -C-2/88f ...SQN ca}culations.for' flexible vertical. ,

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tanks have -been revised to ' address NRC ' '

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. concerns abcutstank flexibility. -

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A-42/ Pipe Cracks in Boiling .'

NA; 1Not' applicable.to pressurized water; s MPA B-05 Water Reacto~rs' reactors.: ,;

A-43 . Containment Emergency 'C-9/87) Westinghouse' study and NoPris' Laboratories

' Sump Perfor9ance. .. evaluation. indicated acceptable. sump'

. performance.

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.A-44 Station Bl&ckout I-6/92' TVA ' completion is dependent on' issuance of

SER by NRC. -

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,. r A-45 Shutdown Decay Heat. NC' 1 Vulnerability to loss of decay heat Removal Requirements removal _ is to be addressed by SQN's IPt:. '

5 A-46 Seismic Qualification. 1-12/90 Replacement.' Items Project is scheduled for. '

full implementation oy.. Cycle 4.

~

of Equipment in- j.

Operating Plants.

A-47 Safety Ieplication E-3/90 NRC GL 89-19 is under review. -

of Control Systems A-48 Hydrogen Control Measures C-II/84 - Redondant trains of. ignitors have been <

and Etfects of Hydrogen installed. Future analytical work is' ' ,

Burns on Safety Equipment -planned. and the schedule-is tied to the -

f

' completion of. analysis work by . Duke Power :

'Coepany.' Reference TVA letters dated -

2/2/87 and 1/29/88 and NRC letter dated

'3/31/87.

A-49 Pressurized Thermal Shock I-3/9'1 'The remaining action is to submit revised

• reactor pressure vessel heatup and ' '

Tcooldown curves in accordance with NRC GL 88-11. ;e.

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