ML20210T243

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Partially Withheld Memo Forwarding Potentially Generic Issue Data Sheet Re Steam Leaks in Main Steam Bypass Lines Due to Fatigue Failure.Diagrams & Deviation/Event Rept Also Encl
ML20210T243
Person / Time
Site: Grand Gulf, 05000000
Issue date: 10/19/1984
From: Robert Lewis
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To: Bishop T, Denise R, Partlow J
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I), NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV), NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
Shared Package
ML20210T193 List:
References
FOIA-87-50 NUDOCS 8702180125
Download: ML20210T243 (7)
v • d • e


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MB RANDUM FOR:

James G. Partlow, Deputy Director, Division of Quality Ur 4 g/.

Assurance, Safeguards and Inspection Programs, IE Richard W. 5tarostecki, Director, Division of Reactor ere. pct f u 2 I'?MI17,, Director, Division of Reactor u-Projects, RIII -

Richard P. Denise, Director Division of Reactor safety and Projects, RIV Thomas W. Bishop Director, Division of Reactor Safety and Projects, RV FRC :

Richard C. Lewis, Director, Division of Reactor Projects SUE ECT:

PDTENTIALLY GENERIC PROBLEM AT GRAND GULF REGARDING THE MAIN STEAM BYPASS LINES.

The enclosed potentially generic issue data sheet concerning s' team leaks in l

mat steam bypass lines due to fatique failure is forwarded for information per TI 5D0/3.

i hichard C Lewis q

Enc >sures:

1.

Potentially Generic Issue Data Sheet No. RII:DRP:84-16 2.

Supporting Dats 3.

System / Hardware Diagrams a

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Heltenes, Jr., AEDD R. I Bangart, RIV 6

I CON',CT:

l R. I Carroll Jr.

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ENCLOSURE 1

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i Dg Sheet No.:RII:DRP-84-16 Appendix A TI 2500/3 1

4/1/80 p0TENTIALLY GENERIC ISSUE DATA $NEET Fa liity Grand Gulf Docket No(s).

50-416 Da e of Event 9/16/84 and 9/23/84 Inspection or other Report F0-416/84-38 '

1.

Brief Description of Issue (Not risquired if included in supporting data)

See Enclosure 2 2.

How Found (If appropriate)

Discovery of steam leaks in the turbine building 3.

Why considered Potentially Gener1c (i.e. - reference applicable criteria or give reason)

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These fatique failure induced torpass line steam leaks, have occurred at Grand Gulf and at a Bayernwerk facility (Grafenrhienfeld-KKG) in Germany.

Like Grand Gulf, who's turbine generator and associated hypass hardware were supplied by Allis-Chalmers, it is possible t. hat other nuclear power plants have a similar three line main steam bypass design.

V 4.

II R.I. Carroll P. R. Beets /D. N. Verre111 Region OrEginator.

Section Chief / Branch Chief l

5.

Dther Reeton Reportine That The Probles Has Also Been Identified By Them Region,

, Chief

, Reporting

. Docket No.

l 6.

Evaluat*:on by IE:HO Bu11ettei L/

Circular. / /

Information Notice /_,/

Other No further action required

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OCT 121984 ENQt.05URE. 2 5_! eortine Jata G:.nd Gulf unit 1, has empertenced two separate steam leaks in the "A" main steam b) : ass line (a non-safety related syn,5en) while tqypassing steam at the 155 to 2C%

pr 'er level. The first event (g/16) was due to e crack induced failure of the wi d which connected a 24-inch capped pipe stub to the "A" hypass line's pressure

b. ahdoun assembly (pBA); (See Enclosure 3).

The failure of this weld was es.rlbuted to earlier vibration of a. drain line that was connected to the pipe (The drain line on all three bypass lines' PBAs had been removed earlier p

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in the plant's life when preoperational testing shoued they were not required.)

Ti PBA. including a 1-foot section of 18-inch diameter P3 pipe and a 2%-inch d

noter drain nozzle upstream, as well as a short section of 24-inch P1 pipe di nstream, was furnished by Utility power Corporation. Inc. (previously known as l

The PBA was manufactured in Germany using German steals., i A' is-Chalmers).

overy action included replacing the damaged PBA and a portion of tho' 18-inch Rt ss valve.

stop/yA"PBAaswell st.edule 40 pipe located betueen the PBA and the "A" U'..rasonic and magnetic particle testing w"as perfomed on the new 0: on the existing ones in the "B" and C bypass lines.

Tl 9/23 occurrence resulted from a 7-inch crack, propagating in both directions l

m the new repair weld connecting the existing 18-inch pipe to that portion of ft inch pipe that was replaced during the 9/16 recovery period. Off-stte anal-It s confimed that both the 9/16 and 9/23 events were due to fatique failure; y:

welding deficiencies.

The 9/23 event has been attributed to pulsating / Y

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nt onste vibration in the 18-inch schedule 40. pipe when steam flow was passing rr tt ough the line.

It is also believed that this resonate vibration was the ur '.orlying cause of the 9/16 event.

. sequent modifications to all three hypass lines included replacing the 18-inch 5t.edule 40 piping with schedule 100 : piping.

(That portion of the PBAs con-st ning the Zh-inch capped pipe stub was also eliminated.) The two welds in each ta3. ass line which connects the 18-inch schedule 100 piping to the stop/ bypass b

vi ve and the remaining portion of the, PBA were stress relieved a'nd radiograph-11y tested.

These modifications were based on those made at a Bayernwerk l

itfa ility (Grafenrhienfeld-KKG) in Gemany, which experienced two similar occur-r4 ces in 1979.

No further problems have been encountered there since these m(.ifications were made.

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()eg No. 004 8 9 9 pg - Edison Deviation / Event Report Nembe, wss-,oi ,a,. t et ~I Identification and Description Part 1 4 A. Title 30" By Pass Line Field Weld 25 & 26 . Occurrence Date Time C. Safety Related D. 5ystem

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0 y,s B No y 13 in l-li 17 el 31 il a d d a a d A F. Description of occurrence or condition Nuclear Operations Welding Manual Sect. VII states that if a split backinn is to be used the split shall not exceed 3/32" contrary to this the split in backing for field weld 25 & 26 is approx 1/2" also during the welding of this field weld it was noted that weld layers were 7/8" to 1" wide, the max that can be carried is 5 times the diameter of electrod. 1/8" diameter electrodswere being used. G. Im mediate action . / Ff W H. References L init ~ t r Org./Section Date e Nuclear Operations Weldinz Manual / ve/In M M/7//Arar I //~h Yf Part 2 Repoltability A. 10CFR50 72 O ves E No Person Contacted n t Hr O 4 Hr. Date Time 10CFR$0 73 n Yes IX No LE R Number 10CFR21 O Yes [M No LE R Due Date

8. Detconined ByM ( M orarsec C atfL b A/o Cate f 8 9 7 Part 3 l:

~ l CARB Review l. A. Accept description of occurrence and immediate action (X Yes M No Accept reportability determination M Yes M No . 0$RO meeting required n Yes IMNo Meetinq No A) A Date D. Evaluation and Action t ssignment : F. Carb.N P Only M RadChe n M NQA R NT C NA O NOM n Maintenance M Technical f5L NE D N5 M NP M MNPM M Operations M Other E-aluator Section Evaluator Section E. Evaluation due date G. AR8 App ital Date h l //f/VfEl' // ~ / f* 8 T" Part 4 Evaluatfon l l A. Describe the root cause of the devsstion PE R. WNCf. DER FDtJ% Ally 'Td[ FJ T-t)b 19A/d 11 I6LOlhh Dln A/D T FCLL6h) 774E /ALSTJtLJd YsCN S BN ~7 d~ G /dELb PROCESS C n N T R n L_ S//EtT ffAP( 6 )

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8. Remarks

[ \\ (If reauired, attach a Corrective Action Sheet) Part 5 l Work Completion 7~" C. Approved By Org / Section Date I Part 6 l-NQA. Cornpletion Review tviewed By Section Da'e I l l

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O l Of 963 66331184C5 Detroit Edison Corrective Action Sheet Dr R ue Paca c8 Part 4 l Evaluation l Part 5 l l Completion l-C. Proposed Remedial Corrective Action:

1. 05RO meeting A. Remedial Action Taken:

_1st R w. m n n.sr_ rne D ves t% no .Ecs rotICT/et)$ inst J7Acjets)&-

2. Prior to declaring overable

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~ h Date/JdM C. Approved by Date E PicposedCorr/sve Actionto Prevent 1 dSRO meeting O Action Tab en to Prewent Recurrente' Recurrence: "Jg" r f~p,tt,f,w _ O yes D no idftbE*f_ '. AA/A 7~s r 7 s E 6 2 Prior to declaring operable . 1 NL/ckl'L*13 tD1 Y14 "TAf b 0 Yes E No ___E d A.,0 A /2f - 7 E 3 Ass 1nedto JN.<*r/2 0L TC1) /M YtJE' org f)u c. lit ftL.Th}/fL: nM hienkb6 Section VO D$ kQ 4 Read completion date l*]~0 71/ElJ1stryffN;5 . tr n td PC 5, 5 CARD aporoval M M%d __eF Date /2/p 7,(9 5 E. Com pleted by Date n n,,, F. Deterrnined by\\b [M Dave ff,-f]* M F. Approved by Da'e

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