ML20004B078
| ML20004B078 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 05/19/1981 |
| From: | Mills L TENNESSEE VALLEY AUTHORITY |
| To: | James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| References | |
| 10CFR-050.55E, 10CFR-50.55E, SQRD-50-328-81, NUDOCS 8105270199 | |
| Download: ML20004B078 (2) | |
Text
o e-
~.
I:5 :I I'
.~ - -
- a 400 Chestnut. Street Tower II May 19, 1981 SQRD-50-328/81-15 Mr. J P. O'Reilly, Director Office o nspection and Enforcement U.S. Nuclea Regulatory Coinniasion Region II -
ite 3100 101 Marietta reet Atlanta, Georgi 30303
Dear Mr. O'Reilly:
SEQUOYAH NUCLEAR PLANT UNIT 2 - REACTOR ESSEL FIELD ' ELD UIDICATIONS -
SQRD-50-328/81 THIRD IITTERIM REPORT The subject deficiency was initially reported to NRC-0IE Inspector M. Thomas at January 20, 1981 in accordance with 10.:FR 50.55(e) as NCR's 2511 and 2547. Interim reports were submitted on February 19 and April 6, 1981. We expect to submit our next report by May 22, 1981.
The enclosed thini interim report describes the current status of this NCR as discussed with your staff on May 15, 1981.
If you have any questions, please get in touch with D. L. Lambert at FTS 857-2581.
Very truly yours, TENNTESEE VALLEY AUTHORITY L. M. Mills, Manager Nuclear Regulation and Safety Enclosure
/
cc:
Mr. Victor Stello, Director (Enclosure)V Office of Inspection and Enforcement U.S. Nuclear Regulatory Commission Washington, DC 20555
{$f fr4 lb? s So/7 6 e,. @9.
d e ess-et 5
ea, :
i
- 2. u.5 @N, I}
l
.c a[.s G/
N' 8105270 N 3
e ENCLOSURE SEQUOYAH NUCLEAR PLANT UNIT 2 REACTOR VESSEL FIELD WELD INDICATIONS SQRD-50-328/81-15 10 CFR 50.55(e)
THIRD INTERIM REPORT Interim Progress Following het functional testing the reactor vessel nozzle weld inoications which were reduced to an acceptable size before hot functional tests were reexamined by liquid penetrant testing (PT). Also, additional metallography was performed. This inspection revealed that several of the indications which had previously been reduced to an acceptable size, meeting the requirements of ASME XI (Summer 1978 Addenda), had increased in size but still met ASME XI criteria. It is postulated that this increase could have been within the tolerance of the PT test, but this could not be proven. During this examination, several indications were also discovered that were not present during the prehot functional examination. At this time the decision was made to examine the remaining six nozzles. All nozzles had some indications that were not present before hot fbnctional tasting. The indications that have been discovered on the additional six nozzles are similar in characteristic to those discovered on nozzles 2RC-9 and 2RC-17. Approximately 75 percent of the indications are parallel to the direction of coolant flow and lie in a section circumferential1y around the outside diameter of the nozzles. The remaining 25 percent are either randomly oriented or rounded and lie in the same circumferential section.
The indications are located in the 309 stainless steel weld metal
" buttering" which was deposited on the end of the carbon steel nozzles to facilitate field welding.
Of the remaining unit 2 nozzles examined, no nozzle had cracking as extensive as that found on 2dC-17. All indications have been cleared from these nozzles. The maximum depth of grinding on 2RC-17 was 7/32 inch, six of the nozzles had less than 1/16 inch, and the eighth had less than 3/32 inch metal removal. Repair will consist of blending those nozzles which have sufficient wall thickness remaining to at least a 3 to 1 taper and l
weld repair of any nozzles which require additional material. Nozzle 2RC-17 has already been weld repaired.
TVA elected to inspect all eight unit 1 reactor vessel nozzles during the current shutdown. Of the eight nozzles inspected, all but two had indications which met ASPE Section III criteria. These two indications
~
are 1/8 inch in length and meet Section XI requirements. We do not see the same pattern of extensive axial indications as was seen on nozzle 2RC-17.
Westinghouse has researched their inservice inspection records and has found no previous history of this type of problem on other vessels of similar design, i.e., 309/308L buttered nozzles.
TVA has removed a small chip of metal from 2RC-17 and is performing a metallu gical investigation in conjunction with Westinghouse to obtain murther information regarding cause of this problem. Based on evidence from preliminary investigation to date, this condition appears to be stress-assisted corrosion cracking, although hot cracking has not been raled out.
We will submit a final report on t.a condition once our evaluation is completed.
Since the nozzles have now been thermally cycled and all unacceptable indications repaired, we conclude that saf* operation of unit 2 can proceed.
-