IR 05000244/1995011
| ML17263B074 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 06/05/1995 |
| From: | Modes M, Patnaik P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17263B073 | List: |
| References | |
| 50-244-95-11, NUDOCS 9506140124 | |
| Download: ML17263B074 (11) | |
Text
Qa DOCKET/REPORT NO.
LICENSE NO.
LICENSEE:
FACILITY NAME:
INSPECTION AT:
INSPECTION OATES:
U.S.
NUCLEAR REGULATORY COMMISSION
REGION I
50-244/95-11 OPR-18 Rochester Gas and Electric Corporation 89 East Avenue Rochester, New York 14649 Ginna Nuclear Power Plant Ontario, New York May 1-3, 1995 INSPECTOR:
. PM=~
ra as atnai
,
eactor nspector Materials Section
'on o
Reactor Safe 5. 22-19
~ate APPROVED BY:
>c ae o es, ie Materials Section Oivision oF Reactor Safety ate 9 o ADQCK 05000244 50bi40i24 950b05 PDR PDR
EXECUTIVE SUMMARY This inspection covered reviews of the inservice inspection program, the eddy current examination program for steam generator tubes and an assessment of licensee's evaluation of defective sleeve welds found in "8" steam generator during eddy current examination.
The inservice inspection program during the spring 1995 outage complied with the 1986 edition of the ASME code,Section XI.
For thin wall Class II piping,
.
the licensee has exceeded the requirements of the code in selection of components for examination.
The steam generator (SG) eddy current examination also exceeded the regulatory requirements and was oriented to identify degraded and defective tubes and sleeves with the mechanism causing tube degradation.
During the inspection, the licensee discovered two unwelded sleeves and seven defective welds in the upper sleeve welds of "B" steam generator.
These weld defects and lack of weld in the sleeves were attributed to failure of detection by ultrasonic examination for installation acceptance conducted by an inexperienced technician.
The defective sleeves were repaired by plugging.
The licensee's significant hazard evaluation appropriately addressed the, concerns associated with the lack of adequate upper welds.
The incomplete tube sleeve weld and subsequent failure of the ultrasonic examination violate a number of criteria, including Ginna UFSAR paragraph 5.4.2.7.2, (Rev.
10, 12/93)
and
CFR Part 50, Appendix B Criteria IX,
"Control of Special Processes."
Because Rochester Gas and Electric identified the incomplete weld, it could not be reasonably expected to have been prevented by Ginna corrective action for a previous violation (because there weren't any); it was corrected in a reasonable time; and it was not willful; (10 CFR Part 2, Appendix C, B(2)) this is a Noncited Level IV Violatio DETAILS 1.0 INSERVICE INSPECTION (ISI) (73753)
1.1 Scope The conduct of inservice inspection using ultrasonic, magnetic particle, and liquid penetrant examination ensures integrity of the pressure boundary.
During this inspection, reviews of the ten-year ISI plan, the scope of work for the outage, a sampling of inservice inspection data and the steam generator tube examination data were conducted.
1.2 Findings Rochester Gas and Electric Corporation (RG&E), the licensee for Ginna Station, conducted an inservice inspection during the spring 1995 refueling outage of the unit, in accordance with the American Society of Hechanical Engineers (ASNE) code,Section XI, 1986 edition.
The unit was in the third outage of the second period for the third inspection interval.
The inspector reviewed selected portions of the ten-year ISI plan for class 1, 2, and 3 components and the relief requests to the NRC pertaining to the inservice inspection.
The inspector found that the licensee's basis for the relief requests was justified.
Within the scope of the review of the ten-year plan, the inspector did not find any discrepancy.
The scope of inservice inspection work for the outage was reviewed, The components listed in the work scope were identifiable from the ten-year plan.
The inspector noted that the licensee had included welds in thin wall piping belonging to the containment spray, the residual heat removal, and the high pressure safety injection systems, which were beyond the requirements of the applicable ASHE code,Section XI.
This would ultimately result in enhanced safety and incr eased availability oF the plant.
The examination requirements for components in the outage plan, and the percentage of welds examined by the end of the second period, complied with the requirements of the 1986 edition of the ASNE code,Section XI.
The non-destructive examination (NDE) data on the following welds were reviewe SYSTEM WELD EXAMINATIO IDENTIFICATION N METHOD RESULTS REMARK Reactor PL-FW-XII-LR Coolant System (Longitudinal RCS Loo A
seam UT No recordable indication Relief request to be generated for cast steel RCS-Loop A Pump-to-Pipe PL-FW-VI UT & PT No recordable indication No examination on pump side due to pump confi uration RCS - Loop B
PL-FW-XI-LR
,
UT & PT (Longitudinal seam No recordable indication Relief request to be generated for cast steel RCS - Loop 8 Elbow-to-Pump 4-inch Pressurizer Relief Line Reducer-to-Pum Seal water to Reactor Coolant Pump A, Pipe-to-Elbow 2-inch charging line to Loop B, cold leg.
Pipe-to-valve and Tee-to-reducer 2-inch alternate charging to Loop A, cold leg.
Elbow-to-cou lin 2-inch alternate charging to Loop B, hot leg.
Integral attachment.
PL-FW-XV
22 and 21C 30 and
CVU-48 UT& PT UT & PT PT PT PT PT No recordable indication No recordable indication No recordable indication No recordable indication No recordable indication No recordable indication Six manufacturing indications were removed by surface conditionin None Limited examination due to obstruction from a guide, but the coverage exceeds
ercent None None None
The above data conformed to the evaluation criteria set forth in the non-destructive examination procedures.
In the inspector's assessment, the licensee's data evaluation was properly done.
The written practices on qualification and certification of non-destructive examination personnel belonging to the contractor and the licensee were reviewed.
These documents were found to be in compliance with the American Society of Non-Destructive Testing (ASNT) Recommended Practice No. SNT-TC-IA, 1980 and 1984 edition, as applicable.
The certifications of all personnel who performed non-destructive examinations during the conduct of inservice inspection were available in the file, and the personnel were appropriately certified.
2.0 EDDY CURRENT EXAMINATION OF STEAN GENERATOR TUBES The initial eddy current examination of the "A" and "B" steam generators was performed utilizing a standard bobbin coil technique with data acquisition being performed with the EDDYNET Acquisition system.
The frequencies selected were 400, 200, 100, and 25 kHz.
The examination was performed primarily from the first tube support through the tubesheet region to examine open crevice, roll transition regions, and the area between the top of the inlet tubesheet to the first tube support plate.
Twenty-percent of all open tubes were also examined throughout the full length of tubing.
Additionally, eddy current examinations of the "A" and
"B" steam generators were performed utilizing the Zetec 3-coil motorized rotating pancake coil (HRPC) probe to examine the roll transition region, and selected crevices.
The frequencies used for these examinations were 400, 300, 100, and 25 kHz.
All row one and row two U-bends were also examined with HRPC.
All sleeve examinations (except for the B&W brazed and old style explosive sleeves)
were performed using the Zetec "plus-point" probe.
The B&W brazed sleeves and the old style explosive sleeves were examined using a saturation bobbin technique, due to the magnetic permeability of the sleeve material.
Prior to examination of the steam generators, an inspection program was established for the inlet and outlet sides of both the "A" and
"B" steam generators.
The inlet or hot leg, examination program plan was generated to provide the examination of 100% of each open unsleeved steam generator tube from the tube end through the first tube support plate, along with 20% of these tubes being selected and examined for their full length (20% random sample as recommended in the Electric Power Research Institute (EPRI)
guidelines) with the bobbin coil.
In addition, 20% of each type of sleeve was examined and the remaining tube examined full length.
All row I and row 2 U-bend regions were examined with the HRPC between the 06H and 06C supports from the cold 'leg sid As a result of examination, the following repairs were performed to each steam generator.
Number of tubes plugged Number of tubes sleeved
~A" SG
75
~G" S
G
91 Total Repairs Im HU The results of the examination indicated that intergranular attack (IGA) in the secondary side and primary water stress corrosion cracking (PWSCC)
continue to be active within the tubesheet crevice region on the inlet side of each steam generator.
The "A" steam generator had 88 tubes that were found to have tubesheet crevice indications (56 PWSCC L 32 IGA).
The "B" steam generator had 106 tubesheet crevice indications (26 PWSCC
80 IGA).
Additionally, 14 sleeved tubes were plugged (3 obstructed, 2 with PWSCC,
blowhole, 2 unwelded and 6 unfused),
accounting for a total of 120 repairs in the "B" steam generator.
2. 1 Steam Generator Sleeve Defects During eddy current examination of 515 ABB-CE sleeves (out of a total of 1,370 sleeves)
using a plus-point probe, 11 sleeves exhibited indications in the inside surface.
These eleven indications were comprised of one in "A" steam generator and ten in "B" steam generator.
Subsequent visual examination of sleeve upper welds indicated eight welds with minor weld defects considered to be acceptable, two blow hole indications and one sleeve with no weld in "B" steam generator.
The sleeve without any upper weld was a curved sleeve installed in 1990.
The lack of weld was attributed to equipment problem in installing curved sleeves which had even. occurred in laboratory testing.
However, ultrasonic testing for acceptance of installation did not identify this defect.
Subsequently, all data on plus-point probe for upper welds were reviewed and visual examination of curved sleeves that were not examined by plus-point probe was conducted.
One additional sleeve without an upper weld was found in "8" steam generator.
The licensee determined that both sleeves with missing upper welds were ultrasonically examined by the same technician in 1990, who had little experience in ultrasonic examination of sleeves.
One hundred and thirteen welds were identified that had been examined by this technician.
During this outage, all 113 welds were examined by either plus-point or remote visual with no negative results.
As a precaution, an experienced and qualified Level III in this particular UT, randomly tested 20K of the population and found one more sleeve with lack of fusion in the upper weld.
The sample was expanded to 100'.
An additional six sleeves with inadequate fusion were discovered for a total of nine.
All nine sleeves with lack oF fusion or missing welds in steam generator
"B" were repaired by plugging the tube.
CE attributed the problem to one UT technician used in 1990 and stated that for an inexperienced NDE technician, it was possible to mistake the expansion transition area as a wel The inspector reviewed the following welding records pertaining to sleeve welds.
~
Procedure for welding the 7/8-inch steam generator tube sleeve, upper and lower joints.
~
Preproduction sample preparation for steam generator tube sleeves.
Procedure for the ultrasonic examination of steam generator tube to sleeve upper welds.
~
Field activity logs during welding of sleeves.
~
Certifications of NDE technicians.
The review of procedures did not identify any procedural deficiency in the sleeving activities and the review of various logs even indicated that the activities were properly controlled.
The NDE technicians were appropriately certified.
The licensee evaluated the potential safety hazards associated with plant operation with the above-mentioned sleeves with inadequate or missing upper welds.
The evaluation addressed the following concerns associated with the lack of adequate upper welds.
~
Structural integrity of the sleeve with a missing weld.
~
Structural integrity of the parent tube with a missing weld.
~
Leak integrity of the sleeve/weld upper joint with a missing weld.
The licensee's evaluation concluded that the deficient sleeves did not result in a loss of structural integrity of the welded CE sleeves, and since the consequences of a complete severance are bounded by the existing design basis steam generator (SG) tube rupture accident, a substantial safety hazard did not exist due to the following:
There was no major defect in the design, construction, or installation of the CE welded sleeve, which would have resulted in a structural failure of the installed sleeve.
Additionally, the lack of adequate fusion during the installation process does not prevent the installed sleeved from functioning as a leak limiting sleeve.
There was no major degradation of essential safety equipment since the CE welded sleeves would maintain structural integrity.
Additionally, the parent tube would remain constrained by the tubesheet and the installed sleeve.
Therefore, reactor coolant system (RCS) pressure boundary integrity has been maintaine ~
Since the CE welded sleeves were determined to be capable of acting as leak limiting sleeves with integrated leakage well below the plant technical specification limit of 0. I gpm for normal plant operation, no significant increase in radiation exposure or release to the general public would have occurred.
The resulting primary to secondary leakage is well below the bounding leakage For the existing Ginna SG tube rupture accident.
The incomplete tube sleeve weld and subsequent failure of the ultrasonic examination violate "a number of criteria, including Ginna UFSAR paragraph 5.4.2.7.2, (Rev.
10, 12/93)
and
CFR Part 50, Appendix B Criteria IX,
"Control of Special Processes."
Because Rochester Gas and Electric identified the incomplete weld, it could not be reasonably expected to have been prevented by Ginna corrective action for a previous violation (because there weren't any); it was corrected in a reasonable time; and it was not willful; (10 CFR Part 2, Appendix C, B(2)) this is a Noncited Level IV Violation.
The inspector determined that the licensee appropriately addressed safety concerns and the inspector concurred with the above conclusions of the licensee's evaluations.
3.0 NANAGENENT OVERSIGHT/SELF ASSESSMENT The quality assurance department performed surveillances and audits during the past year in the following areas:
Visual and liquid penetrant examinations Wet fluorescent magnetic particle examination Ultrasonic weld examinations Eddy current examination of steam generator tubing VT-3 visual examination of a mechanical snubber Radiographic examination ISI/IST repair/replacement program Snubber inspection program Containment fan cooler service water isolation valve hydro Radiography of service water pipe welds The surveillances and audits provided a thorough self-assessment and oversight of the non-destructive examination program at site.
4.0 EXIT MEETING The findings of the inspection were presented to and discussed with members of the licensee's management at the exit meeting on Hay 3, 1995.
The licensee concurred with the findings of the inspection and did not voice any objections.
A list of attendees of the exit meeting is appended to this report as Attachment I,
ATTACHMENT I Rochester Gas and Electric Cor oration M. Saporito J.
Widay F. Klepacki P.
Lewis R. Marchionda Manager, Technical Performance
& Field Inspection Plant Manager ISI Engineer LIS Senior Engineering Assistant, NDE Plant Superintendent U.S. Nuclear Re ulator Commission P. Patnaik T. Moslak P. Drysdale Reactor Engineer Sr. Resident Inspector Sr. Reactor Engineer
C