ML20198P772
| ML20198P772 | |
| Person / Time | |
|---|---|
| Site: | 05000000, Oyster Creek |
| Issue date: | 05/10/1983 |
| From: | Ostrowski R, Pinelli R GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| Shared Package | |
| ML20151H203 | List:
|
| References | |
| FOIA-86-26 IEB-82-03, IEB-82-3, PROC-830510, NUDOCS 8606060390 | |
| Download: ML20198P772 (29) | |
Text
.
f OYSTE11 CREEK RECIRCULATION SYSTEM PTPING INSPECTION PROCHAM Response to NRC I&E Bulletin No. 82-03, Rev.1 p
Topical Report No. 012 1
(Rev. 0)
Project No.: 328025 R. A.
Pinelli R. Ostrovski AUTil0RS PATE pnv 10, 1983 APPROVALS :
a
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84.-
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_LUtION !!AtiAGER
- ATE MSN3 vM DEPAllDZNf MAI;fGER DATE' f.!!O f% 3 (414[u (4A M o (ECTOR - QUALn( ASSURA
- CE DATE
_? td idw VICE PRESIDENT DATE TECl!NICAL FU!;CTI0t!S t
(SIGNIFICANT IMPACT REVIEW) 8606060390 860319 PDR FOIA PATTERSO86-26 PDR
%//t3 n.
,h
- t.S s tra c t The augmented inservice inspection program for I&E Bulletin 82-03 for Oyster Creek's Recirculation System has been performed.
Preliminary results of this ultrasonic inspection reveal that two welds contain indications similar to those of IGSCC and several welds possess geometric / fabrication indications.
Some of these indications cannot be compicLely characterized at this time. Follow-up internal examinations are required t o provide more complete inf ormation. These inspections are scheduled following reactor vessel drain and the disassembly of recirculation system suction and discharge valves.
4 9
l l
o.
0YSTER CREEK RECIRCULATION SYSTEM PIPING INSPECTION PROGRAM INTROD UCTION In October, 1982, the NRC issued I&E Bulletin No. 82-03 pertaining to "Stre ss Corrosion Cracking in Thick-Wall Large-Diameter, Stainless Steel, Recirculation System Piping at BWR Plants".
The bulletin concentrated on Nine Mile Point's recirculation system pipe cracking experiences.
With Oyster Creek being identified as an addressee of the bulletin and being most similar in design to.Nine Mile Point, GPUN developed and implemented an Inspection Program for ultrasonic examination of recire. system piping welds which would satisf y the new requirements.
Below is a supplemental response to Action Items 4, 2, and 3 of the bulletin.
METHODS 4.a.(1) & (2)
As part of the Inspection Program, CPUN and its NDE contractor, utilizing CPUN approved procedures and calibration standards, qualified inspection teams on cracked pipe sections from NMP at Battelle Columbus Latoratories in accordance with the Bulletin's requirements.
The initial inspection sample size, as shown in Figure 1 and reported in GPUN's 82-03 submittal dated December 1,1982, was 16 welds. The selection of the welds to be inspected was not a random sampling. Rather, the welds
s.
selected were those which would be more susceptible to IGSCC cracking, based on three conditions (1) high stress rule indices, (2) weld repairs performed during construction and (3) welds being similarly located to those at Nine Mile Point where cracks were initially detected.
Prior to the start of in spectio n, further information became available concerning weld repairs during construction.
Accordingly, additional welds were included in the inspection plan, increasing the size to 22 welds (Figure 2).
After prelimi-nary inspection re sults revealed two locations with possible ICSCC, the inspection plan was again increased to include all similarly located welds on the remaining recirculation loops (Figure 3), for a total of 31 welds in spected.
4.a.(3)
Previously re sponded to by Ref.1.
- 4. a. (4 )
Previously re sponsed to by Ref.1.
4.b. th rou gh 4.d Previously re sponded to by Ref.1.
INSPECTION RESULTS 2.
The original inspection results revealed that only two welds contain indications which may be intergranular stress corrosion cracking (IGSCC) and five welds with indications characterized as fabrication / geometric-type.
While further evaluation is needed on the two welds with possible IGSCC.
inspections were scheduled for nine additional welds.
These nine included all welds similar in location to the two above mentioned welds plus athose selected under the requirements of ASME Section XI Inservice Inspection.
I
-3_
l The additional inspections are complete.
The results of the additional inspections reveal th ree welds with indications which have been preliminarily characterized as fabrication or geometric conditions. Appendix 1 provides a detailed summary of the inspection results.
PLAN OF ACTION h
Bounding fracture mechanics analyses of the two possible IGSCC indica-tions reveal that both indications can be lef t as found (i.e., will not propa-gate beyond acceptable limits) for the next operating cycle without any form of weld repair.
Further characterization of the 2 welds indicative of IGSCC and five of i
the remaining eight welds with fab / geometry indications will require both radiography and internal, remote (visual and/or liquid penetrant) examina-
^
tions.
These examinations shall be performed after fuel of f-load, reactor i
ve esel drain a nd recire. valve disassembly.
I RECOMMENDATIONS
- 3. (Co n t ' d. )
Final disposition of the 10 welds with indications shall be made af ter a review of the additional internal examinations scheduled following reactor ve ssel drain.
However, possible courses of action being considered include weld over-lay repair for the two welds containing IGSCC-type indications -au induction heating stress improvement for welds repaired during construction and welds with high SRI's (greater than 1.4).
l
y s
s I
. REFERENCES 1) letter to Ronald C. Haynes from Peter B. Fiedler dated December 1, 1982, Re: Oyster Creek Nuclear Generating, Station Docket 50-219, I&E Bulletin 82-03.
2)
Letter to Ronald C. Haynes from Peter B. Fiedler dated February 25, 1983, Re: Oyster Creek Nuclear Generating Station, Docket 50-219, I&E Bulletin 82-03.
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- l. LOOPS "N,"B') "C"I "O" H AVE A TOTAL OF 15 WELDS EACH.
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////svm xA Entar-Othy Ma.Mandu)iA..
08 /Ases Date April 5, 1983 S PP( OMI:l-614 Subject Itectre. PtpInn Examinatton Results To T.
J.
Pat t e rson Location Oyutur Creek
.s The following information delineates the current status of exmuination results obtained from the initial " Quick Look" for ICSCC on the recirculdtion piping system.
The initial " Quick Look" sample was comprised of twenty two (22) welds (Attachment 1)..The sample size was later expended.during the examination to include four (4) scheduled ISI welds and five (5) welds which closed outthe inspection sets for areas where recordable indications were found.
These additions brought the total number of wclds,cxamined during the " Quick Look" to thirty one (31).
An a result of the examinations the following data was generated:
1.
'Nonty one (21) weld heat affected zones (llAZ) were found clear, that in, there were no recordable indications per the requirements of the Ultrasonic procedure used (6130-QAP-7209.08 Rev. 0).
2.
Ten (10) weld heat affected zones were documented as having recordable indications per the requirements of the U. T.
procedure.
Locations, evaluations, and repair status of these 10 welds are shown in Attachments 2, 3, 3A, and 4.
Two (2) weld (llAZ) are considered at this time to have indications characteristic of IGSCC (NG-D-11 and NG-D-5).
Planning for supplemental examinations is as follows.
Remote visuals are to be conducted on the tuo welds characterized au 1GSCC.
If renults of thouu examinations are not will have to be performed.conclusivo a remote surface examination (Dye Penetrant)
In addition, radiography is corruntly planned for fcur (4) welds which were evaluated as geometry / fabrication indications and the remaining wold (NG-ll-5) characturi. zed as ICSCC, Plans ace also undcrway to utilize an ultrasonic signal characterizer (ALN-4060) to be provided thru the EPRI Institute, as a means to further evaluate the recordable indications referenced in Attachment 3.
April 11.
The use of the 4060 is scheduled for the week of Except for the examinations utilizing the ALN-4000 all other supplemental exams will be scheduled after vessel drain, now scheduled for April 27.
A0000040
i' r y ',i t 7.
T. J.
PaLLersun A further update concerning these examinations will be-made us the supplemental examinations are completed and the welds dispositioned.
pC/d' sa R. Outrowski Supervisor, Corporate ISI RO/dg Attachments cc:
H. Allgaior w/ attachments J. Chardos w/ attachments R. DeMuth w/ attachments N. Kazunau w/uttachmentu R. Pinelli w/ attachments S. Pruitt w/uttachments C. Tracy w/ attachments G
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taenetseet s u.es todications Ite e sen Ex.usan.n a s qas httot' I
- e;-A-t at' vem Nu-A-14 @ cu"(fA -.* O w-a-ix T..
ves NC-C-12 fue v..
NC-O-11 ygg Yee 608 @ A.T.. Video W-A-3 Q IN;-M -7 (3 NG-N-2 b
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flC-E-24 ye, NC-A-6 NG-C-4 Yes NC-D-4 Yes yee flC-E-6 yee I
kEEI.3 total: $
Total 22 Tot.sle 1
@ taantuatten Team Numv'er
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e IWACTult RECIRCULAT10N l'It' LNG, OYSTER CREEP, 4
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NG-A-14 (A Loop)
NG-B-12 (B Loop)
NG-C-12 (C Loop) 4 g
NG-D-11 (D Loop) bS6 pgt4 #
___..____,_______m._.._,.
i_su.s. 4 s is mm t _nt..tsj. i a Ut _ t Nbl*LG i L OH II.H HLLU', WLitt NOTLD TO itAVE RtCORDAULE 1HDICAll0NS TO DATE l>LANNED
'LD ADDITIONAL ADDITIONAL FINAL
).
VARIABLE EVALUATION NDE NDE DISPOSITis i-B-12 SIDE-1 Ind.#200-Lack of thru-wall dimension discounts IGSCC SIDE-1 Ind.#201-Location, Lack of length (spot) and lack of thru-wall dimension discounts IGSCC SIDE-l Ind.#202-Location and CRT pattern amplitude preclude possibility of IGSCC SIDE-1 Ind.#203-Location discounts IGSCC'
-C-12 SIDE-1 Ind. #200-Fabrica tion /geane try condition SIDE-1 Ind.#201-Fabrication / geometry condition D-ll SIDE-1 Ind.#200-Characteristic of IGSCC 60, R.T.
Video 0
Ind.#201-Characteristic of IGSCC 600, R.T.
Video E-4 SIDE-1 Ind.#200-Fabrication, geometry condi tion Ind.#201-Fabrication, geometry condition C-22 SI DE-2 Ind.#200-Fabrication, geometry condition C-23 SIDE-2 Ind.#200-Fabrication, geometry R.T.
condition SIDE-1 Ind #201-Fabrication, geometry R.T.
condition SIDE-1 Ind.//202-Fabrication, geometry R.T.
condition B-5 SIDE-2 Ind.#200-Fabrication, geometry R.T., Video condition SIDE-2 Ind.#201-Fabrication, geometry R.T.,Vi deo condition SIDE-2 Ind.#202-Fabrication, geometry R.T., Video condition SIDF-2 Ind.#203-Characteristic of IGSCC R.T., Video SIDE-2 Ind.#204-Characteristic of IGSCC R.T., Video
- -4 SIDE-2 Ind.#200-Fabrication, geometry condition 1-4 SIDE-2 Ind #200-Fabrication, geometry R.T.
condition Ind.#201-Fabrication, geometry R.T.
condition
e PRELIMINARY RESULTS OF INSPECTION TLN WELDS WERE NOTED To llAVE RECORDAllLE INDICA i.ONS i
ILD TO DATE PLANNED ADDITIONAL ADDITIONAL FINAL D.
VARIABLE EVALUATION NDE NDE 0!SPOSI1.
%-A-14 SIDE-1 Ind.#200-Fabrication condition 600 R.T.
Ind.#201-Fabrication condi tion R.T.
.*e
WELD llISTORY (WELDS WITl! RECORDA!1LE U.T.
INDICAT10NS)
- DATE OF LOCATION NATURF h*0.
FAllR ICAT e oM op ggpafg gp g,, g NG-A-14 Pre-1969 3:00 1.F.
(3 spotu) repaired NG-11-12 6:00-12:00 I.P. - repaired 6:00-10:00 Gas Pockets / Porosity - repaired tiG-C-12 6:00 to 9:00 I.P. - repaired NC-D-11 5 of 6 on film I.P. and Slug Incluulons-rupulce.
NG-E-4 None NG-C-22 None - Elbow was removed during construction and rewe L NG-C-23 Nono - Elbow was removed,during construction.and rowe,.
NG-B-5 2-3 on film I.P'.
- repaired NG-C-4 None NG-D-4 None 4
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EXAMINATION TEAMS TEAM NO.
EXA!!INE11S/ LEVEL U.T.
1 McCaully, II/Licbold, II 2
Reichert, II/ Trotter I 3
McCaully, II/D. Koua, I 4
Deskiewicz, II/Valden, I 5
Rad 1 beck, II/D. Itusa, I 6
Deskiewicz, II/Floyd. I 7
lx:skiewicz, II/C. Itos a. I 8
Collins, II/Butkiewicz, I 9
Radlbeck, II/Floyd, I 10 Minyon, III/Valden, I 11 Liubuld, II/Valden, I 12 Reichert, II/Licbold, II 13 Collins, II/Minyon, III 14 Reichort lI/C. Rosa, i 3
15 Rad 1 beck, II/Valden, I 16 Minyon, III/McCaully, II 17 Reichert, II/Minyon III
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sI UfdlTED STATES i
NUCLEAR REGULATORY COMMISSION s
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.._,..e!E E._LETIN NO. 84-01:CF.ACKS IN SWR MARK I CONTAINMENT VENT m :- e r..
Re:
Oyster Creek Nuclear Generating Station On Februa ry 3,19E4, Secrgia Power Company reported a through-wall crack airost co eletely around the vent header within the contain ent torus of "aten brit 2.
Later tnat day IE Eulletin 64-01, " Cracks in Sciling Water
- eactor ark I Cor,tainment Vent Headers," was issued for action te the licensees of EVR facilities with Mark I containments that were in cold EFutdexn.
The bulletin recuired inspection for cracks in the containment vert header. Tne bulletin also suggested that the operating EWR plants wi-h vark I contai.n ents should review their plant data on differentiel i
cressure betv.een the wetwell' and drywell for anomalies that could de irdicatise of cracks.
By letters dated February 10 and Sectember 14, 1984, GPU Nuclear (the licensee) has responded to IE Sulletin 84-01.' The licenses stated in the letter dated February 10, 1984, that inspections perfor ed in response to the bulletin and a' eady dcne in conjunction with torus modificetion work in the Cycle 10 refueling outage showed no indication of cracking in the ver.: rin; header, main header, vacuur breaker line, or nitr, ogen purge Diring as experienced at Hatch 2.
Attached to this letter were copies of ths data sheets cf the inspections on the vent header and nitrogen Durge li,e.
In the ie::e* dated Se:tember 14, 19E4, the licensee provided su:rie ental information which gave the licensee's respcnses to the five re:cmmencatior.s in the General Electric (GE) Service Information Letter (SIL) Nc. 4C2 on the hatch Unit 2 Vent Header Cracks.
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~. e s t a -9 revievec the licensee': resconses to IE Evi*e:in '*-01 anc cer.cleted that no further action by the licer,see needs to be taker. but j
- ne Cc.r.issien reeded to concuct a follow-up inscection at Cyster Creek to verify the ccrpletion of the GE SIL No. 4C2 reco=endatier.s cemitted.
- o Oy -he licensee.
Pegien ! concucted Inspection 50-219/E5-01 e: Oyster C eer en IE Euiletin 82-01 and GE SIL Nc. 402. The ins:ection report da ed Er:F 14, IEEE closed cut this issue.
There'cre, this letter cieses cut the steff's acticns cr. this issue.
II 1
f, 3 Sincerely, g, v, s
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John ". Zwolinski, Chief j
Operating Reactors Brancb =5 Division of Licensing cc:
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20C36 Forked River, New Jersey 0E731 J.E. Liberran, Escuire Commissioner
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New Jersey Depa*trent of Energy
- E5 Avenue of the Anericas 101 Commerce Street
':sw Ycrk, New York 10036 Newark, New Jersey 07102 Eugene Fisher, Assistant Director
- ecional Administrator Division of Environmental Ouality Nucisar Regulatory Commission Departnent of Environmental Eecien I Office Protection 631 Park Avenue 380 Scotch Road King of Prussia, Pennsylvania 19406 Trenton, New Jersey 08628 e;3 Licensing Manager G:V Nuclear 100 Interpace Parkway l
Parsippany, New Jersey 07054 1
Decuty Atterr.ey General State of New Jersey Cepartment of Law and Public Safety 2f West State Street - CN 112 Trenten, New Jersey 08525 Mayor Lacey Township E1E West Lacey Road Ferked P.iver, New Jersey OS731 U.S. Environnental Protection Agency Pe; ion II 0#fice i
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Regicnal Radiation Recresentative 3
26 Federal Plaza New Ycrk, New York 10007 D. G. Holland Licensing Panager Oyster Creek Nuclear Generating Station j
i Post Office Box 388 Forked River, New Jersey 08731
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- 1882 MEMORANDUM FOR: Regional Administrators FROM:
William J. Dircks Executive Director for Operations
SUBJECT:
BWR PIPE CRACK PROBLEMS Recently there has been an increasing number of problems involving degra-dation and cracking in thick-wall large-diameter stainless steel piping at BWR plants.
As a result of the degradation identified in the recirculation system piping in the reactor coolant pressure boundary at Nine Mile Point Unit 1 the NRC issued IE Bulletin 82-03 entitled, " Stress Corrosion Cracking in Thick-Wall Large-Diameter Stainless Steel Recirculation Piping at BWR Plants." Inspection pursuant to this IE Bulletin has revealed instances-of piping degradation at Monticello and Hatch Unit 1.
Indications of piping degradation have also been identified at Hatch Unit 1 during inspections conducted pursuant to NUREG-0313, Revision 1 (Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping).
0(
NRR has generic responsibilities in this area, has chaired the Pipe Crack Study Groups, and has the lead responsibility for resolution of Generic Task A-42 (Pipe Cracks in Boiling Water Reactors).
Pursuant to that task, NRR published NUREG-0531 (Investigation and Evaluation of Stress-Corrosion Cracking in Piping of Light Water Reactors), several generic letters, and NUREG-0313.
NUREG-0313, Revision 1, published in July 1980, sets forth the NRC staff's revised acceptable methods to reduce the intergranular stress corrosion cracking susceptibility of BWR ASME Code Class 1, 2, and 3 pressure '
boundary piping and safe ends, and provides inservice inspection requirements.
The NRC needs a central point for dealing with these technical issues.
It is essential that the NRC response to the pipe crack problems be technically correct, regionally consistent, and integrated with ongoing studies and research. Accordingly, for any pipe crack problem that you judge requires substantive NRC review, I would like you to transfer responsibility for NRC action to NRR.
Because we have actions underway in this area, the following guidelines apply:
1.
Licensee responses to IEB 82-03 will be reviewed by the appropriate Region and the results forwarded to IE.
IE will evaluate and consolidate IEB 82-03 responses and provide the results to NRR.
NRR will employ this information in fonnulating plans for NRC action.
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2.
Substarl.tial plant-specific problems identified during inspections pursuant to NUREG-0313 will be forwarded to NRR.
3.
Substantial plant-specific problems with large-diameter safety related BWR piping will be forwarded to NRR.
4 NRR will evaluate and take appropriate licensing action regarding plant-specific repairs and modifications made as a result of pipe degradation identified pursuant to NUREG-0313 or IE Bulletin 82-03 inspections.
Transfer of responsibility to NRR shall be made by memorandun to the Director, Division of Licensing, NRR.
The memorandum should transmit as much information about the problem as you initially have available.
You will, of course, retain cognizance over any potential enforcement actions that may be related to pipe degradation problems (e.g., failure to report,,
management control weaknesses, violation of tech specs, etc.);
(Signed) William J.Dircks William J. Dircks Executive Director for Operations cc:
H. R. Denton R. C. DeYoung s
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TASK No.
6 63 DATE:
November 10 1
- S u. Mt., X.982 TASK INTERFACE AGREEMENT TAC f:
3LEM:
Evaluation of BWR Recirculation Piping ISI Inspections - IEB 82-03 LERD OFFICE: C IE C NRR C
REGION C
JOINT NOTIFICATION:
REFERENCES:
IE Bulletin 82-03: Stress Corrosion Cracking in Thick-Wall, Large Diameter, Stainless Steel, Recirculation System Piping at BWR Plants.
ACTION PLAN:
With respect to IEB 82-03 IE will accomplish:
IE:
1.
Coordinate with IPRI and the Regional Offices to assure that the demon-strations of the UT methods used by various ISI or.ganizations are witnessed by NRC.
2.
Evalutte, with the assistance of the Regional Offices, the performance capabilities and effectiveness of each ISI organization's UT methods.
3.
Issue final report on evaluation of ISI organiz,ation's performance tests required by the bulletin.
4 Assist in review of NRR Item-1, bel ow.
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7.2 Regions
9'-
1.
Assist IE in witnessing and evaluating ISI organization's performance tests required by the bulletin.
2.
Inspect ISI activities perfomed at affected plant sites to assure that these are consistent with the methods and procedures' used to demonstrate the effectiveness of the UT methods.
(CONTINUED ON NEXT PAGE)
NRR:
Designate Lead Project Manager to assign TACS and coordinate correspondence, meetings, and reports (ORB *2 R. Clark
). x27162
__ t OFF8CE COORDINATORS:
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V. Stello, ROGR J. Sniezek IE
Regional Admin.
R. DeYoung, IE D. Eisenhut, NRR LEAD Project Manager Taylor, IE C. Michelson, AEOD R. Vollmer, NRR R. Purple, NRR
. Grimes IE H. Denton, NRR G. Lainas, NRR R. Wessman, NRR E. Jordan, IE E. Case, NRR T. Novak, NRR W. Hazelton, NRR R. Baer, IE R. Mattson, NRR T. Ippolito, NRR W. Johnston, NRR IE J. Kramer, NRR F. Miraglia, NRR K. Wichman, NRR l
W. Mills ick, IE
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TASK No.: Ibn DATE: Novemb2r 10,1982 i
IE (continued):
3.
Inspect repair activities at plant sites as required.
NOTE:
If no evidence of piping degradation is identified at a particular facility, the decision regarding facility restart will be made by the Region.
NRR:
1.
Based upon results of IE report, review of licensee's inspection program, and IEB 82-03 responses, determine generic significance of degradation in recirculation piping and need for augmented ISI requirements. (MTEB) 2.
For plant-specific issues, assume lead responsibility for NRC review of pipe degradation problems, licensee submittals relating thereto, and licensing actions that may be required.
(Project Managers will coordinate with appropriate NRR Branches.
Plant specific TIAs will be develcped on a case basis.)
3.
Assist in evaluation of IE Item 2, above, c'
NOTE:
If evidence of piping degradation is identified at a particular
(
facility, the decision regarding facility restart will be r:ade by NRR.
APPROVED:
Facility PA-TAC Browns Ferry 2 1162 49044 t.1. M s h.k y dde% st/9 h it G Brunswick 1 1162 49045 Thomas T. Martin, Director Region I Dresden 2 1163 49046' Division of Engineering & Technical Programs Duane Arnold 1163 49047 Hatch 1
- 1162 49048 Millstone 1 1161 4904 9 (C3' A. Ha,lt b 'T.O.
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Monticello 1163 49050.
John Olshinski, Directob Region II Oyster Creek 1161 @9051 '
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C rles E. Norelius, Director Region III l
D'visi n f Engineering & Technical Programs l (hQYl v
u James Gagliardo, Director Region IV Di' vision of Resident Reactor Project &
Engineering Programs
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7-NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C.
20555 October 28, 1982 IE BULLETIN NO. 82-03, REVISION 1:
STRESS CORROSION CRACKING IN THICK-WALL, LARGE-DIAMETER, STAINLESS STEEL, RECIRCULATION SYSTEM PIPING AT BWR PLANTS Addressees:
Tiose licensees of operating boiling water reactors (BWR's) identified in Table 1 for action.
All other licensees and holders of construction permits (cps) for information only.
Purpose:
This bulletin is to notify all licensees and CP holders about a matter that may have a high degree of safety significance, and to require specific actions as set forth below for those licensees listed in Table 1.
Specifically, this matter involves the degradation in the recirculation system piping in the reactor coolant pressure boundary (RCPE) that was found at the Nine Mile Point Unit 1 Nuclear Generating s
Station.
This information was detscribed in considerable detail in Information Notice 82-39, dated September 21, 1982.
Action by the affected licensees identified in Table 1 is required to (1) provide a reasonable level of assurance that inspections which are currently being performed or scheduled are sufficient to detect cracking in BWR thick-wall recirculation piping welds
- and (2) to assist the NRC in determining the R1 generic significance of the piping degradation found at Nine Mile Point.
s The affected licensees are those owners whose plants *re currently in or scheduled to be in a refueling mode or extended outage through January 31, 1983.
This bulletin is provided to all other licensees and holders of construction permits for information only at this time.
Licensees not listed in Table 1 will be notified by January 15, 1983 as to the scope and extent of any required actions.
Description of Circumstances:
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During a primary system hydrotest in March 1982 at Nine Mile Point Unit I (NMP-1), leakage was visually detected at two of the ten l
l
- Large bore piping that is not designated as " Service-Sensitive" in R1 accordance with NUREG-0313, Rev. 1.
It should be noted that NUREG-0313, R1 Rev. I designates the recirculation riser lines as " Service-Sensitive."
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IES 42-03, Rev. 1 October 28, 1982 Page 2 of 5 Table 1 Plants Currently in or Scheduled to Be in a Refueling Mode or Extenced Outage Through January 31, 1983 LICENSEE PLANT Northern States Power Company Monticello Nuclear Generating Station Tennessee Valley Authority Browns Ferry Unit 2 Nuclear Generating Station Commonwealth Edison Company Quad Cities Unit 1 Nuclear Generating Station Dresden Unit 2 Nuclear Generating Station Northeast Utilities Millstone Unit 1 Nuclear Generating Station Georgia Power Ccmpany Hatch Unit 1 Nuclear Generating Station Carolina Po er & Light Company Brunswick Unit 1 Nuclear Ge'nerating w
8 Station
- Jersey Central Power & Light Company Oyster Creek Nuclear Generating Station Iowa Electric Light & Power Company Duane Arnold Nuclear Generating Station s
- To be performed during the November 1982 refueling outage, not the current outage.
IEB 82-03, rev. 1 October 28, 1982 Page 3 of 5 furnace-sensitized, recirculation system safe-ends.
Further visual inspection _ revealed three pinhole indications and a single -inch-long axial indication, all of which were located in the heat-affected Zone of the welds where the safe-end joined the pipe.
About nine months before the leak, these safe-ends were ultrasonically (UT) inspected; at that time, the inspection did not disclose any reportable indications.
Subsequent to the leak, the UT procedure was modified; UT examination of the two affected safe-ends and one other safe end confirmed the presence of indications of intermittent cracking around the pipe's inside diameter (ID).
Additional examinations revealed cracking in heat-affected Zones of recirculation pump discharge welds.
Dye penetrant examination confirmed these crack indications.
The UT examinations were extended to other welds in the five loops of the recirculation system.
The results of these examinations disclosed ID cracking in a large number of the welds examined.
Two boat samples removed from the area of the through-wall cracks in one safe-end were sent for evaluation--one to General Electric Co. and the other to Battelle Laboratories.
In addition, a boat sample from the crack region of the elbow weld was evaluated by Sylvester Associates, consultants to the licensee.
The results of these metallurgical evaluations concluded that the degradation resulted f rom intergranular stress corrosion cracking (IGSCC) in the sensitized region of the weld's heat-affected Zones.
Based on the fact that NMP-1 has furnace-sensitized safe-ends, the licensee deciced to replace all 10 recirculation system safe-ends without further
'i investigation beyond that described above.
Based on recirculation system findings, the licensee decided to also replace all recirculatioh system piping while the facility was shut down for safe-end replacement.
On September 16, 1982, a meeting was held between General Electric, BWR licensees, and NRC staff to review past IGSCC experiences and the general implications of NMP-1 IGSCC degradation in main recirculation piping welds.
The staff had the benefit of the metallurgical evaluation of the NMP-1 event and an update of the general IGSCC experiences relative to all operating BWR s plants.
On September 27, 1982, a meeting was held between BWR licensees and the NRC staf f to discuss the extent and results of examining welds in th? recirculation system for all BWR licensees with plants currently in or scheduled to be in a refueling mode or extended outage through January 31,.1983. As a result of this meeting, the NRC staff has determined that additional information is needed to assess the effectiveness of the UT methods employed or planned to be
~
used and to determine whether such piping should be designated " service-sensitive" in accordance with NUREG-0313, Rev. 1, issued by NRC letter dated February 26, 1981.
To provide a reasonable level of assurance that inspections which are currently being performed or scheduled are sufficient to detect cracking in thick-wall, recirculation system piping welds and to assist the NRC in further evaluating this issue, the affected licensees (identified in Table 1) are requested to take the following actions.
t IE8 82-03, Rev. 1 October 28, 1982 Page 4 of 5 Actions To Be Taken by Licensees of BWR Facilities Identified in Table 1:
1.
Before resuming power operations following the current refueling or extende'd outage, the licensee is to demonstrate the effectiveness of the detection capability of the ultrasonic methodology used or planned to be used to examine welds in recirculation system piping.
This demonstra-tion shall be made on representative service-induced cracked pipe samples.
Arrangements should be made to allow NRC to witness this demonstration.
This demonstration shall employ those procedures and standards, the same type of equipment (same transducer size, frequencies and calibration-standards), and representative UT personnel from the inservice inspection (ISI) organization utilized or to be utilized in the examinations at the plant site."
2.
Before resuming power operations following the current refueling or extended outage, the licensee is to provide a listing of results of recirculation system piping inspections.
3.
Before resuming power operations following the current refueling or extended outage, the licensee (if the inspections indicate the presence of cracks) is to describe the corrective actions taken and report these in accordance with the appropriate regulations.
4.
To assist NRC's further evaluation of this issue, the following shall be submitted by December 1, 1982:
a.
A description of the sampling plan used or to be used during this outage for UT examinations of recirculation system piping welds and the bases for the plan. The description should:
(1) Provide an isometric drawing of the recirculation system piping showing all the welds, and the number of welds and their loca-~ '
tion that have been examined or will be examined.
(2) Identify criteria for weld sample selection (e.g., stress rule index, carbon content, high stress location, and their values for each weld examined).
(3) [escribe piping material (s), including material type, diameter, tnd wall thickness.
1 (4) Estimate the occupational radiation exposure incurred or expected and briefly summarize measures taken to maintain individual and collective exposures as low as reasonably achievable.
- We understand that Electric Power Research Institute (EPRI) has arranged to have samples from the Nine Mile Point Unit I plant available for industry demonstrations of UT methodology.
The samples have been taken to Battelle Memorial Institute in Columbus, Ohio for characterization and subsequent use.
6 m,
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IEB 82-03, Rev. I October 28, 1982 Page 5 of 5
~
b.
A summary description of the UT procedures and calibration standards
- used or to be employed in the examination at the licensee's plant site.
This description should include the scanning sensitivity, the evaluation sensitivity and the recording criteria.
c.
A summary of the results of any previous inspection of the recircula-tion system piping welds which used the validated examination meth-odology as discussed in Action Item 1 above.
d.
An evaluation of the crack-detection capability of ultrasonic meth-odology used or planned to be used to examine recirculation system piping welds.
This evaluation should result from conducting the i
demonstration required in Action Item 1 above, and should include a comparison of the service-induced pipe crack sample to those welds actually examined in the licensee's plant in terms of pipe wall thickness and diameter, weld geometry, and materials.
5.
The written reports required by Items 2, 3, and 4'shall be submitted to the appropriate Regional Administrator under oath or affirmation under provisions of Section 182a, Atomic Energy Act of 1954, as amended.
The original copy of the cover letters and a copy of the reports shall be transmitted to the U.S. Nuclear Regulatory Commission, Document Control Desk, Washington, D.C. 20555 for reproduction and distribution.
~
This request for information does not require Office of Management and Budget approval since the number of plants asked to provide the information is limited j
to nine reactor plants.
Although no specific request or requirement is intended, the following informa-tion would help the NRC evaluate the cost of implementing this bulletin:
o Staff time to perform requested demonstration o
Staff time to prepare written responses If you have any questions regarding this matter, please contact the Regional Administrator of the NRC Regional Office or one of the technical contacts listed below.
daf Y
/ ~
hard C. DeYoung, Direct ice of Inspection and Enforcement Technical
Contact:
William J. Collins, IE 492-7275 Warren Hazelton, NRR 492-8075
Attachment:
{
List of Recently Issued IE Bulletins
_