IR 05000269/1981016
| ML15224A451 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 09/04/1981 |
| From: | Economos N, Herdt A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML15224A447 | List: |
| References | |
| 50-269-81-16, 50-270-81-16, 50-287-81-16, NUDOCS 8111050782 | |
| Download: ML15224A451 (15) | |
Text
eB REGU UNITED STATES
- CLEAR REGULATORY COMMISSIO
REGION II
101 MARIETTA ST., N.W., SUITE 3100 ATLANTA, GEORGIA 30303 Report Nos. 50-269/81-16, 50-270/81-16 and 50-287/81-16 Licensee:
Duke Power Company 422 South Church Street Charlotte, NC 28242 Facility Name:
Oconee Docket Nos. 50-269,50-270 and 50-287 License Nos. DPR-38, DPR-47 and DPR-55 Inspection at Ocon e site near Clemson, SC and Babcock & Wilcox Lynchburg Research e, Lyn r Virginia Inspector:
a?
N. Ec MDat4 Signed Approved by:
_
A. R. Herdt, Section Chief Da e Signed Engineering Inspection Branch Engineering and Technical Inspection Division SUMMARY Inspection on July 22 to August 13, 1981 Areas Inspected This routine, unannounced inspection involved 90 inspector-hours onsite and at B&W's Lynchburg Research Center (LRC) in the areas of Unit 1 ten year inservice inspection of the reactor pressure vessel and other Class 1/2 components; core barrel assembly thermal shield broken bolts; reactor coolant makeup system modifications.
Results Of the three areas inspected, two violations were identified in two areas: (Vio lation -
Failure to control welding electrodes and maintain appropriate environ mental conditions around the welding area -
Paragraph 5.b.; Violation -
Failure to maintain required cleanliness levels in the transfers canal area -
Paragraph 5.e.).
No deviations were identified.
. 8111050782 811027 PDR ADOCK 05000269 G
REPORT DETAILS 1. Persons Contacted Licensee Employees
- J. E. Smith, Station Manager
- J. M. Davis, Superintendent of Maintenance
- R. J. Brackett, Senior QA Engineer
- J. N. Pope, Superintendent Operations
- T. E. Cribbe, License Engineer
- G. Rothenburger, Support Engineer, Mechanical Maintenance J. Vignati, Construction Engineer W. W. Galcman, Mechanical Maintenance Support Engineer B. Milbsaps, Assistant Engineer C. B. Cheezem, ISI Engineer Other Organizations Babcock and Wilcox, Nulcear Power Generation Division (NPGD)
H. W. Stoppelmann, ISI Coordinator M. G. Hacker, Level III Examiner G. S. Clevinger, Manager, Nuclear Materials Technology, LRC R. V. Straub, Service Manager, Operating Plant Services Robert Michelski, Level II Examiner/ARIS Supervisor Oak Ridge National Laboratory J. H. Smith, NDT Engineer Consultant Sandia National Laboratories John H. Gieske, NDE Engineer NRC Resident Inspector
- W. Orders D. 0. Myers
- Attended exit interview 2.
Exit Interview The inspection scope and findings were summarized on July 24, 31 and August 13, 1981 with those persons indicated in paragraph 1 above.
The inspector described the areas inspected and discussed in detail the
inspection findings listed below.
No dissenting comments were received from the licensee.
a.
(Open)
Violation 269/81-16-01: Failure to Control welding electrodes and to maintain appropriate environmental conditions around the welding area, paragraph 5b.
b.
(Open)
Violation 269/81-16-02: Failure to maintain required cleanliness levels in the transfer canal area, paragraph 5.e.
c.
(Open) Inspection Followup Item 269/81-16-03: Requests for relief ISI requirements, paragraph 6a.
3.
Licensee Action on Previous Inspection Findings Not inspected.
4.
Unresolved Items Unresolved items were not identified during this inspection.
5.
Independent Inspection Effort a. Core Barrel Assembly Thermal Sheild Bolts Broken (LER 269/81-11)
On July 15, 1981, the licensee reported that while performing reactor vessel internals inspection with a remote video camera, as part of the 10 year inservice program, several loose parts (bolt sections) were discovered in the bottom of the Unit 1 vessel.
They were tentatively identified as parts missing from the internal vessel structures.
Further inspection revealed that the loose or missing parts consisted of four of 96 bolts and their locking clips used to secure the bottom ends of the thermal sheild. On June 26, 1981, Unit 1 was shutdown for this refueling and inspection outage.
On July 24, 1981 the inspector viewed the video tape showing the broken bolt section (head and shanks) resting on the bottom of the vessel and another showing the lower portion of the thermal shield where it attached to the lower grid forging showing the bolts in question. The machined and fractured surfaces of the broken bolts, as seen on the bottom of the reactor vessel, showed no discernible evidence of plas tically deformed metal.
Discussions held with the cognizant licensee personnel disclosed that approximately 80% of the aforementioned bolts appeared to be loose and were believed broken as they appeared to have backed out between 0.1 to 0.5 inches from the as installed position. Four of the 96 bolts had separated and were found to be missing along with three locking clips
used to secure the bolts.
In addition the licensee representative stated that one of the lateral restraint guide blocks was also missing.
At the time of this discussion, the reactor vessel guide lugs, core ledges and other parts of the vessel had been visually inspected with the aid of a remote control video camera. Other items inspected by the same method included the core barrel lower grid area, thermal shield bolting, core barrel bolting, upper thermal shield restraints, guide blocks, flow distributor, the upenders and canal.
The licensee repre sentative stated that no significant problems were identified.
This discussion also revealed that three broken bolt shanks and two bolt heads, retrieved from the vessel, had been forwarded to B&W's Lynchburg Research Center (LRC)
for a metallurgical examination/failure analysis in order to determine the cause of failure.
The licensee representative stated the bolts were made from material produced to ASTM Specification A453-65 Gr. 660 Condition A (A-286).
This is a high strength heat resisting alloy with a nominal composition of 15% Cr, 26%
Ni, 1.25% Mo and 2% Ti with an ultimate strength of approximately 130,000 psi.
On August 6, 1981, inspectors N. Economos, A. R. Herdt and J. Fair met with B&W Nuclear Material Technology personnel in Lynchburg, VA, to discuss the status and findings of the ongoing investigation and, the generic implications of this occurence on other B&W plants.
The B&W representative stated that in terms of a micro examination their findings showed that no evidence of necking, bottoming, gross extension or ductile tearing associated with overloading.
The material exhibited a uniform hardness in the range of 27 to 32 Rockwell "C"(RC).
A microscopic examination disclosed that a variation in grain size existed between the bolt head and the shank.
This is undergoing further evaluation. The inspectors noted that the bolt head may have been hot-upset formed which may account for the grain size variation.
Tensile tests, performed with a sample showing an activity of about 1R produced similar values as that from archive material.
Hardness surveys over the entire bolt cross-sectional surface was being con ducted at B&W's Alliance plant. Results of a SEM examination showed evidence of intergranular corrosion covering a large portion of the fracture surface area with associated small branch cracking. A lesser portion of the fracture surface exhibited evidence of transgranular cracking with striations indicating a low stress fatigue mechanism contributed to the failure.
Hence the failure mode was tentatively identified as corrosion fatigue associated with low stress levels.
In discussing the fabrication history and design changes which resulted in use of the these bolts, the inspectors confirmed that the thermal shield (TS) modification for Oconee Unit 1, was performed in the field whereas Oconee Units 2 & 3 TS's were modified at the B&W's shop.
In
response to questions regarding the type of lubricants (brand, compo sition etc.) used to drill and tap the holes in the TS and lower grid forging, B&W stated that this information was not available at this time but that it would be researched. The inspectors stated that this information could be useful and may help to identify the agressive species that may have initiated the intergranular corrosion attack on the bolts. In conclusion the B&W representative stated that RII would be kept abreast of their findings as more information became available.
In discussions held with the licensee representative, during the August 10-13, 1981 inspection, the inspector noted the following:
1. The licensee had removed bolt numbers W1O-2 from the TS and was making preparation for its shipment to LRC for analysis.
2. Removal of the remaining bolts was scheduled for the week of October 12, 1981, and installation of new bolts during the period of November 2 thru 22, 1981.
3. The material tentatively selected to replace the present bolts was X-750 inconel.
4. B&W was performing an information type ultrasonic examination on the TS upper restraint and lower section areas including bolts, guide blocks, guide lugs, flow distributor (FD)
and related bolts.
It is anticipated that a bolt will be removed from the FD and forwarded to LRC for analysis.
b.
Reactor Coolant Makeup System Modifications, Unit-1.
This work effort was a followup to earlier inspections on piping and welding activities associated with the safe shutdown facility (SSF).
The designated code governing welding and related pipe activities is ASME Section III (74 S75).
Pipe system modifications inside Unit 1 reactor building include the reactor and the auxiliary service water system.
At the time of the inspection, reactor coolant hangers were being installed but the tie-in to existing systems was not anticipated during this outage for lack of the required pump. Work on the auxiliary feedwater system (ASME Class 3, six inch (6") carbon pipe material) was in progress. It was antici pated that this system would be functional prior to start-up.
On July 30, 1981, ramdomly selected field welds were observed to ascertain whether fabrication practices, workmanship and weld appear ance was consistent with procedural and code requirements.
Welds selected were as follows:
Weld System Size Condition
- DOCIF 51-3-11
-
Makeup (51)
3" Schedule 40 Completed
"
-3-12
""
"
-3-13A
"
"
"
-3-14
"
"
"
-4-01
"
"
"
-4-03
"
"
"
DOCIF 51-6-17
"
1 1/2" Schedule 40 Tacked
" -7-11
"
"
Welding Out
- Radiographs for these welds were reviewed.
In addition to the above attributes, the inspector reviewed fabrication records, personnel qualifications and material quality records.
Within these areas the inspector noted that around the area where welding was performed, in the basement of the Unit-1 reactor building, the floor was covered with water. A number of welded pipe sections were laying on the floor and a significant number of partially consumed bare wire electrodes, stainless steel and carbon steel, electrodes of various lengths from approx imately six to twenty inches were observed scattered about. The inspector notified the project manager and corrective action was implemented imme diately. The inspector stated these findings were in violation with ASME Section III paragraph NC-4411 which states in part that the manufacturer is responsible for control of welding electrodes....
and that,...precautions shall be taken to minimize moisture absorption. This violation was identi fied as item 269/81-16-01,
"Failure to control welding electrodes and maintain appropriate environmental conditions around the welding area".
c. Auxiliary Feedwater Nozzle Thermal Sleeve Inspection This work effort was a followup to the thermal sleeve cracking problem discussed earlier in RE:II 50-269/80-03. Discussions with the cogni zant personnel disclosed that nozzle number "2" of 1A OTSG was removed and visually inspected for cracks. The licensee representative stated that no cracks were observed. In addition the inspector learned that B&W has not yet submitted a final fix for these components.
d. OTSG Upper Manway Stud Inspection Discussions with the licensee's cognizant engineer on cracked manway stud problems discussed covered in earlier RE:II report 50-269/80-03 disclosed that the studs from OTSG(s) A and B were removed, inspected and hardness tested. The inspection showed no evidence of cracking.
A uniform hardness range of 32 to 36 Rockwell C was recorded. This was in contrast to the hardness test results in Unit-3 where the studs exhibited nonuniform hardnesse of 26 to 38 Rockwell C, along with evidence of cracking.
e. Housekeeping The inspector performed a walk thru type inspection inside the reactor building to observe ongoing activities around the transfer canal, reactor vessel head storage area and mainsteam line hanger repair. On July 22, 1981, the inspector noted the following conditions existed around the walking and working area of the transfer canal.
1. Various types of trash and debris including a corroded 7 1/2 volt dry-cell battery; rubber overshoes; loose rope, wire, loose plastic material, tape and other sundry type litter/debris had been allowed to accumulate on the floor.
2. A substantial amount of debris was obsreved floating on top of the water in the transfer canal and, what appeared to be a rag was observed resting on the upper section of the planum.
These conditions were contrary to paragraph 6.3 of N.S.D. 3.11.4 "Clean liness Zone in Safety Related Areas" which applies to the Transfer Canal area and, states in part that...."all waste material shall be placed in approved receptacles, and removed from the area daily or more often as necessary....also that this area shall be kept free from dirt and debris".
Failure to follow procedural requirements was in violation with 10 CFR 50, Appendix B, Criterion V, as implemented by DPC Topical Report "Duke-i-A",
Section 17, paragraph 17.1.5.
This violation was identified as item 269/81-16-02; "Failure to Maintain Required Cleanliness Levels in the Transfer Canal Area".
Except as noted in paragraphs 5b.
and 5.e, no other violations or deviations were identified.
6. Ten Year Refueling Outage Inservice Inspection (ISI), (Unit-1)
a. Code Requirements This inservice inspection work effort coincides with the sixth outage of the first interval and has been identifed as the "TEN YEAR OUTAGE".
Components scheduled for inspection are listed in the Oconee-1 ten year ISI plan and in the ISI plan written for and approved by the licensee for this outage on July 29, 1981.
ASME Section XI (74S75)
as required be 10CFR 50.55(a), paragraph (g)
has been invoked as the governing Code for this activity.
Other controlling documents include the following:
1. Reactor Vessel, Pressurizer and Once Thru Steam Generators a)
b) Article 4. ASME Section V (77S78) Instruments and transducer performance checks; calibration block material requirements.
2. Once Thru Steam Generator Tubes Reg.
Guide 1.83 Rev.
1 ASME Section XI (77S78)
3.
Reactor Coolant Pump Flywheel Reg. Guide 1.14 Rev.1 Following are requests for relief from inservice inspection require ments submitted for review and approval to NRR (Licensing) on or about May 4, 1981.
1. Core flood Nozzle-to Safe-End; Relief from surface Safe-End to Pipe Welds.
examination.
2.
Reactor Coolant Pressure Use of Article 4 Boundary Section V (77S78) as a subsitute for Appendix 1 Section XI for the remaining one-third of the first 10 yr. interval.
3.
Reactor Coolant System Reduce volume of Piping Branch Connection examination to 1/3 T as per ASME Section XI (S78).
4.
Reactor Vessel, Pressurizer Delete all clad-patch Steam Generator Cladding surface inspections for remainder of first ten-year interval.
5. Class 2 Piping Welds Delete UT examination and substitute surface examination for pipe welds less than 0.25" nominal thickness.
6. 9" and 13" Thick Basic T Location of holes Calibration Block out of Code allowable tolerance.
Subsequent to the close of this inspection the inspector ascertained through discussions with cognizant headquarters personnel, that no action (review)
had been taken on the aforementioned items.
On August 24, 1981, the inspector informed the licensee by telephone that these requests for relief have been identified as an Inspector Followup Item (IFI) until such time that appropriate action can be taken on this matter. This item was identified as number 269/81-16-03 "Requests for relief from ISI requirements".
b.
Reactor Vessel Inspection (Unit-1).
Ultrasonic examination of reactor vessel welds as required by the aforementioned Code and recommended by Reg. Guide 1.150 was performed by B&W in accordance with procedures written to meet these require ments/recommendations. Compliance with Reg. Guide 1.150 recommendation was achieved through licensee and Commission staff discussions/meetings held on March 24, 1981, and again on July 22-26, 1981.
Procedures used to conduct the examination were as follows:
ISI-131 Rev. 6 Remote Ultrasonic examination.
Using the ARIS Device.
ISI-130 Rev. 16 Ultrasonic examination of vessel welds and nozzle inside radius sections.
In general the UT equipment included a six channel 600 Krautkramer pulse-echo ultrasonic flaw detector, with single element immersion search units (transducers) consisting of a 00, 450, 600 and two one-half inch 700 angle beam crystals.
Calibration blocks with a 2% notch in addition to the code required 3/8" diameter flat bottom holes were used to calibrate the equipment.
The 2% notch was to used set gain (Sensitivity) for calibration, whereas the 700 angle beam transducer was used to examine the near surface volume of 2" metal thickness.
Welds were scanned automatically fron both sides on the same surface, except that when reflectors greater than 20% were observed scanning was reverted to manual mode. In the case where reflectors required charac terization/mapping, the measurement increments (indexing)
were no greater than nine-tenths of the 70' crystal diameter or approxiamtely 0.3 inches. Scanning sensitivity was set at minimum of two times (6 db)
above primary reference level.
However, in most cases the gain was set at 7 db to 8 db above the primary reference level.
The inspection program provided for the examination of all reactor vessel welds with the exception of the Dutchman to Lower Head-Circle Seam and the Upper Nozzle Belt to Lower Nozzle Belt-Circle Seam.
In both instances only five percent (5%) of these welds were examined as required by the applicable code.
The inspector observed a part of the examination performed on the following welds:
Sequence Figure No.
Location and Type Number of Weld Comments
B1.1.3 Mid Shell to Lower Calibration and Shell, Circle-Seam mapping of Indica tions.
Reexamined weld area with indications to verify repeata bility.
B1.2.2 Lower Shell to Calibration and Dutchman, Circle-Seam part of examina tion.
--
Calibration for sequence # 12,
"Upper Nozzle Belt to Flange Circle Seam".
B.1.4.1 Outlet Nozzle to Calibration and Vessel from Vessel I.D.
part of examina tion.
The inspector observed the above activities to ascertain whether procedural requirements were being followed, personnel were throughly familiar with the inspection system, examination of weld metal and designated base metal was being accomplished, accurate and reproducible documentation of indications as required was being accomplished, initial and intermediate calibrations were being performed as per procedural requirements. In addition to the above, the inspector reviewed personnel qualifications, transducer certifications which included spectrum and acoustical analyses records, instrument calibra tion records and quality records of calibration blocks 40390, and 40392.
Preliminary inspection results of reactor vessel welds at the closing of the inspection on August 13, 1981, were as follows:
Location & Type of Weld Examination Summary Upper Shell to Mid Shell-Fig. B1.1.6; 00 scan, Circle Seam Six recordable laminar indications of acceptable size.
Lower Nozzle Belt to Upper Fig. B1.1.9; 0' scan, Shell-Circle Seam 49 recordable laminar indications, of acceptable size.
Limited scan because of nozzle belt transition.
Mid Shell to Lower Shell-Fig. B1.1.3; 00 scan, Circle Seam Nine (9) recordable laminar indications of acceptable size.
600 scan, three (3); 700 scan, Sixteen (16).
Two of the indications identified with the 700 transducer were confirmed with the 600 transducer. Their amplitudes varied between 0.1 to 0.75 inches while thru wall thickness was estimated to be between 0.1 to 0.3 inches. All except two of the 700 scan indications were in the base metal.
Two indications appeared to be nonfusion defects located near the fusion line, about 1 1/2 inches in from the root of the weld.
Although these indications were observed during the preservice inspec tion (baseline),
any meaningful correlation between the data was difficult-because of insufficient baseline information. Two 600 scan indications identified during baseline appear to be in the same location. Again, meaningful correlation could not be done for the reasons stated above.
Lower Shell-two Long Seams Fig. B1.1.1 No recordable indications.
Fig. 81.1.2. Twenty One, 00 scan recordable indica tions -
all acceptable laminar.
Mid Shell Two Long Seams Fig.B1.1.4 Twenty Five, (25) 00 scan recordable laminar indications-all acceptable.
Fig.B.1.5 No recordable indications.
Upper Shell-Two Long Seams Fig.B1.1.5 No recordable indications.
Fig. B1.1.8 No recordable indications.
Lower Shell to Dutchman -
Circle Seam Fig.B1.2.2 No recordable indications.
Dutchman to Lower Head -
Circle Seam Fig.B1.2.3 No recordable indications.
Upper Nozzle Belt to Fig.B1.#.1 No recordable Flange -Circle Seam indications.
c. Work Observation - Other ISI Activities In addition to the reactor vessel examination discussed earlier in this report, the inspector observed various methods of NDE examinations including ultrasonic inspection, radiography and magnetic particle inspection.
Components, type of examination and results are listed below Ultrasonic Examination A2 Section Drain Line Tee to Pipe B4.05.212 00, 450, 600 ISI-120 Rev. 13 Pipe to Tee B4.05.211 00, 450, 60'
"
Pipe to Elbow B4.05.210 00, 450, 600 n
Calibration Sheet # 3600058, 3600059, 3600061 Pressurizer Support Lugs B2.08.001 3.5" thickness 450 ISI-130 Rev. 16 B2.08.003
"
B2.08.013
"
"
Recordable type indications were identified and documented for evalua tion to determine reportability.
Calibration Shell #3600093 Radiographic Examination C2.01.278 Pipe to Elbow M.S.A 36" x 1.20" C2.01.277 Elbow to Valve M.S.A 12" x 0.60" C2.01.315 Pipe to Cap M.S.B 6" x 0.40" ISI-DOCI-OIA-01-1-X6 34" x 1.125" ISI-DOCIF-03A-04-35 6" x 0.432" ISI-DOCIF-03A-04-X6 6" x 0.432"
The radiographs for the above welds were reviewed and found to be consistent with applicable code requirements.
Magnetic Particle Inspection Feedwater Line
- Attachment Weld C2.05.069 24" diameter ISI-270 Rev. 9
"
"
C2.05.074 6" diameter
- Linear indications documented for disposition.
Within the areas inspected no violations or deviations were identified.
7.
Eddy Current Inspection (Unit 1)
The eddy-current (EC) examination of OTSG tubes was being conducted as per requirements delineated in paragraph 4.17 of the Oconee TS and procedure ISI 416 Rev. 3, "Multifrequency Eddy Current Examination of OTSG tubing in 177 steam generators". This procedure was written to comply with ASME Section XI (77S78),
and Reg. 1.83 July 1975.
Equipment used for examinations included the Zetec MIZ-12 multifrequency generator, Techtronic 5100 storage, display module insert Mo. MIZ-12 and a Zetec probe driver Model 2-D.
Inspection frequencies used for the examination included 200 khz, 400 khz and 600 KHz with a differential coil.
An examination to determine the extent of debris and sludge acummulation was performed using an absolute coil and a frequency of 35 khz. Following is a tabulation of the tubes examined and preliminary results:
OTSG A OTSG B Specific Interest 240 541 Lane Region 367 351 Random Selection 1334 (10%)
937 (6%)
Perephery 6347 Tubes with Obstructions
12 Sleeved None 6 sleeved
retainer added Tubes taken out of Service
37 Supplementary Sampling 926(6%)
Continue:
OTSG "A" OTSG "B" Total Number of Tubes Out of Service to Date
240 Within the areas inspected no violations or deviations were identified.
D-UKE,,
B'WEr?
z;QPANY 422 SOUTH CHURCH STREET, CHARLOTTE, N. C. 28242 WILLIAM 0. PARKER,JR.
VICE PRESIDENT TELEPHONE:AREA 704 STEAM PRODUCTION February 12, 1981 373-403 Mr. James P. O'Reilly, Director U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, Suite 3100 Atlanta, Georgia 30303 Re:
RII:DEM 50-269/80-37 50-270/80-33 50-287/80-31
Dear Sir:
.
With regard to Mr. W. B. Kenna's letter of February 3, 1981 which transmitted the subject inspection report, Duke Power Company does consider the information contained therein to be proprietary pursuant to 10CFR 2.790(d).
Very truly yours, William 0. Parker, Jr.
JLJ:pw