IR 05000327/1993015
| ML20044G245 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 05/24/1993 |
| From: | Blake J, Coley J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20044G238 | List: |
| References | |
| 50-327-93-15, 50-328-93-15, NUDOCS 9306020225 | |
| Download: ML20044G245 (8) | |
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UNITED STATES
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NUCLEAR REGULATORY COMMisslON s
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101 MARIETTA STREET.N.W.
~f ATLANTA, GEORGI A 30323
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9.....g Report Nos.:
50-327/93-15 and 50-328/93-15 Licensee:
Tennessee Valley Authority 6N38A Lookout Place
1101 Market Street Chattanooga, TN, 37402-2801 Docket Nos.: 50-327 and 50-328 License Nos.: DPR-77 and DPR-79 Facility Name:
Sequoyah I and 2
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Inspection Condu3ted April 26 - 30 and May 4 - 7, 1993
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Inspector:
O ey, Jr.
Date Signed goJ.
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Approved by:
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/Maferials and Processes Section
. Er Blake, Chief Date Signed Engineering Branch
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i Division of Reactor Safety SUMMARY Scope:
This routine announced inspection was conducted in the inservice inspection areas of: (1) Observation of work and work activities associated with the 10-year ultrasonic examination of the Unit I reactor vessel and (2) Ultrasonic examination of Unit I steam generator feedwater, nozzle-to-transition-piece welds and associated base materials.
Results:
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In the areas inspected, no violation or deviation was identified. During the two week inspection the inspector observed TVA and their vendors examine the Unit 1 Reactor Vessel and all four steam generator feedwater, nozzle-to-transition-piece welds.
Known cold cracks and reheat cracks in the reactor vessel nozzles were sized
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for growth. The vessel examinations were conducted by Southwest Research Institute (SwRI) utilizing their automated vessel PAR device and their
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enhanced data acquisition system (EDAS). The SwRI examiners were experienced technicians, their computer system was user friendly, and their transducer technology for under-clad crack detection and sizing was very good. TVA management and level III examiners maintained close surveillance of the SwRI process activities.
9306020225 930525 c
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The steam generator feedwater, nozzle-to-transition-piece welds were examined this outage by Asea-Brown-Boveria (ABB). These examinations identified areas of suspected cracking on the nozzles for all four loops. These nozzle transition pieces and the feedwater piping elbows, for each loop, had been replaced 11 months ago and the fact that cracks were presently thought to be detected required serious investigation. Therefore, the inspector observed as TVA's level III examiner conducted evaluations on each indication, for all four loops. The examiner used eight, different, enhanced, examination techniques, each requiring a separate calibration, to confirm whether cracks had actually been detected and to perform accurate sizing of the indications if they were determined to be cracks. The examiner's use of the enhanced detection and sizing techniques was excellent. The level III determined that the transition pieces in loops 1, 2, and 4 contained cracks that exceeded the acceptance criteria of the ASME Code.
Fracture analysis of the defects was performed for TVA by Westinghouse and the fittings were determined to be adequate for another cycle of operation. The inspector noted that TVA's management was very proactive in every area examined by the_ inspecto c
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REPORT DETAILS 1.
Persons Contacted Licensee Employees
- R. Bentley, Technical Specialist NDE - Level III Examiner
- R. Fenech, Vice President, Sequoyah
- T. Flippo, Site Quality Manager
- F. Leonard, Technical Specialist NDE - Level III Examiner
- R. Poole, Instrument Maintenance Manager
- J. Proffitt, Compliance Licensing Engineer
- B. Schofield, Licensing Manager
- M. Skarzinski, Technical Programs and Performance
- R. Thompson, Compliance Licensing Manager
- H. Turnbow, Manager, Inspection Services
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- J. Ward, Manager, Engineering and Modifications
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Contractor Personnel
- W. Jensen, SwRI, Project Engineer
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Other licensee and contractor employees contacted during this inspection included engineers, quality assurance and quality control personnel, l
technicians and administrative personnel.
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NRC Employees
- W. Holland, Senior Resider,t Inspector
- S. Shaeffer, Resident Inspector
- Attended exit interview
Acronyms and initialisms used throughout this report are listed in the i
last paragraph.
2.
Inservice Inspection - Observation of Work Activities (ISI) Unit 1 (73753)
The inspector reviewed documents and recorded data, related.to the observed ISI work activities indicated below, in order to determine t
whether ISI was being conducted in accordance with applicable
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procedures, regulatory requirements, and licensee commitments.
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applicable code for ISI is the American Society of Mechanical Engineers l
Boiler and Pressure Vessel (ASME B&PV). Code,Section XI, 1977 Edition, r
Summer 1978 Addenda, except that the extent of examination for pipe i
welds, categories B-J and C-F, is the 1974 Edition, Summer 1975, and NDE
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techniques are in accordance with the 1986 Edition of the' Code.
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The licensee performs their own ISI NDE, in accordance with TVA procedures, using TVA and contractor examiners. The Site Quality
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Organization (SQO), with the aid of level III examiners from the
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Corporate Inspection Services Organization (IS0), is responsible.for implementation of the ISI program. The corporate ISI/IST Programs
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Organization is responsible for issuing and revising the ISI program and necessary drawings; 150 is responsible for issuing and revising NDE procedures and scan plans. The reactor pressure vessel (RPV)
inspections were performed by Southwest Research Institute (SwRI), using SwRI procedures and examiners, under the direction of licensee NDE
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personnel.
For inspections other than the RPV, contract NDE examiners were furnished by ABB AMDATA.
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Observation of the 10 year automated ultrasonic (UT) examination of the Unit I reactor pressure vessel
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t Background
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By letter dated January 8,1993, TVA submitted a request to delete the requirement to perform supplemental inspections of the reactor vessel nozzles required by Technical Specification 4.4.10.b.
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These inspections are unique to Sequoyah and were added to the technical specification prior to initial startup in 1980 due to
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concerns related to underclad cracking in the reactor vessel
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nozzles.
Initial staff evaluation of the request found that insufficient
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data existed to justify deletion of the inspection requirement, i
and that TVA would be required to submit additional information before further action on the submittal was possible. Therefore, a demonstration was held at SwRI on March 23, 1993, which compared
the UT techniques used in 1980 to those planned for the 10-year l
vessel inspection in 1993.
In addition, in their letter dated April 1,1993, TVA submitted additional information, along with a description of the nozzle testing that would be performed during the Unit 1 Cycle 6 refueling outage, and a request for
confirmation that the scope and technique that would be used to perform the nozzle examinations would be acceptable.
- Based on the above communications, TVA committed to satisfy the Technical Specification requirements by performing the following:
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(1). An examination of the volume of all eight RPV nozzles as defined for the 10-year inservice inspection by the American
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Society.of Mechanical Engineers Boiler and Pressure Vessel (ASME B&PV) Code,Section XI.
(2). An examination of the clad-to-basemetal interface region of the nozzle as defined by ASME Section XI, Figure IWB-2500-7 (can coincide with item 1 above).
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(3). An examination of all reportable indications (reheat cracks,
cold cracks, inclusions, etc.) identified in the
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supplemental examination that was submitted to the NRC by
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letter dated February 28, 1980 that described the results of i
the original examination.
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NRC's letter dated A)ril 16, 1993, confirmed that the SwRI UT technique was satisf actory for inspection of the Unit I reactor
vessel nozzles for cold underclad cracking during the Cycle 6 refueling outage in 1993.
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On April 26, 1993 the inspector arrived at the Sequoyah facility to inspect TVA/SwRI work activities associated with the 10-year reactor pressure vessel examinations. The examinations started on April 27, 1993 and the reactor vessel nozzle examinations were the first examinations to be performed by SwRI. The automated examinations detected 40 of the 61 original indications which had
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been initially detected utilizing manual UT techniques.
By the conclusion of the inspector's two week inspection, EDAS data for three cold crack indications in nozzle N-11, which is the loop 1 inlet, and thirteen cold crack indications in N-12, which i
is the loop 2 inlet, had been evaluated by SwRI and the inspector.
The evaluation revealed that the indications had grown in length,
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and through-wall depth, from the recorded preservice length, and
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the projected preservice though-wall depth.
Even with the
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increase in through-wall measurements, none of the indications had.
approached the analyzed critical flaw depth.
In fact, none had gone into the base material beyond the clad heat effected zone of
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.300 inch, which is where the cracks were predicted to remain.
from the evaluation of the above data, the inspector concluded that the licensee should not seek removal of the examination
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requirements from the technical specifications this 10-year
inspection interval. They should use the data obtained this
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outage as a base line for the subsequent 10-year examinations so that data can be compared utilizing the same UT techniques. The
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licensee and NRR were notified of the inspector's recommendations.
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In addition to reviewing the EDAS evaluation data for underclad l
examinations of the nozzles, the inspector also reviewed the EDAS detection data for the inner-radius and underclad examinations on
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all eight reactor pressure vessel nozzler, observed the SwRI
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technicians acquire data; verified system calibrations; reviewed personnel and equipment certifications; and reviewed the SwRI Nuclear Quality Assurance Manual, the Project Quality Plan and Scan Plan, and the following examination procedures:
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i Procedure 10 to.
Title SEQ-AUT-15 R1 Automated Inside Surface Ultrasonic
Examination of Ferric Vessels
Greater Than 2.0 inches in Thickness SwRI-AUT2 R3 Automatic Inside Surface Ultrasonic l
Examination Indication Resolution and Sizing
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i SEQ-AUT14 R2 Automatic Inside Surface Ultrasonic j
Examination of Pressure Piping Welds
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Within the areas examined, no violations or deviations were
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identified.
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b.
Observation of UT examinations of Steam Generator (SG) feedwater l
piping, utilizing enhanced detection and sizing techniques.
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Upon arriving at the Sequoyah facility the inspector was notified,
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by the licensee, that cracking had just been reported on the loop
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1 SG nozzle-to-pipe transition piece. All Unit 1 feedwater nozzle-to-transition welds as well as the transition-to-elbow welds were in the process of being examined by TVA's vendor Asea-
Brown-Boveria (ABB). ABB had been trained and tested by TVA prior to the outage to be able to properly identify thermal fatigue cracking which would be the mechanism of failure for these welds.
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t The inspector immediately notified TVA's Inspection Services that he would like to observe the level III examiner conduct an-
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evaluation of the indication using the same equipment and techniques used to previously disposition the weld.
TVA complied with the inspector's request and the subsequent examinations
verified to the inspector that the indication had the proper
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signature to be classified as a thermal fatigue crack. Although,
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during the sizing examinations a crack tip could not be discerned l
because of the shallow depth of the indication. A crack tip signal would have been the most conclusive evidence the examiner
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could get for his evaluation.
After the re-examination of the indication, the inspector i
continued to investigated why the crack had occurred so soon.
i (All of the nozzle transition pieces and the elbows had been
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replaced 11 months earlier (April 1992) due to thermal fatigue cracking. Therefore, it did not seem probable that cracking would i
be detected so soon.) Previous UT reports, radiographic film
taken during the 1992 transition / elbow replacements, as well as
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the magnetic particle examination reports for the inside surface
examinations were reviewed by the inspector.
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During the review process, the inspector discovered that the SG Auxiliary Feedwater (AFW) transition pieces had already seen over 31 days of service while in operation, since the April 1992 replacement work. This is a significant length of time, in that
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it accounts for over 1/3 of the time that the fittings have before
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they see a fatigue factor of one.
In addition, when the plant was
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in the startup mode, feedwater to the SGs was being supplied i
automatically resulting in operation of the AFW in a " batch l
feeding" mode.
" Batch feeding" or " slugging" is the most
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detrimental mode of AFW operation for the fitting materials,
because of the thermal cycling that it causes.
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By April 29, ABB had completed all of the nozzle-to-transition and i
transition-to-elbow weld examinations. Crack-like indications had been identified in all four nozzle-to-transition welds, but the indications in loop 3 were evaluated as not being surface connected. The inspector had also accompanied the TVA level III i
examiner during evaluations of the nozzle welds for the other i
three loops. The examiner performed eight different examinations l
on each indication, each examination requiring a separate r
calibration. Cracks were verified in the counterbore region of
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the transition ring on loops 1, 2, and 4.
The flaws varied in
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size, with the largest having an accumulated length of 8 5/8" along the circumference, and a depth of 0.150" at the deepest point. This indication was on loop 2 and was the only indication which provided a crack tip signal.
Westinghouse performed analyses of the indications.for TVA and concluded that, based on the indication depth-and the predicted time that the plant will have the AFW in operation, Unit I could operate another cycle. After that cycle, (approximately 18 months) the nozzle-to-transition welds would be re-examined again, and repaired if necessary.
During the exit meeting the inspector and plant managemet discussed the need for establishing a limit on the use of AFW service to the steam generators.
In the event that some unusual operational circumstances cause the plant to approach this threshold limit, the unit should be brought down and the welds re-examined. TVA's senior management concurred that a threshold should be set.
Within the areas examined, no violations or deviations were identifie.-
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3.
Exit Interview i
i The inspection scope and results were summarized on May 7, 1993, with
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those persons indicated in paragraph I.
The inspector described the areas inspected and discussed in detail the inspection results.
Proprietary information is not contained in this report. Dissenting comments were not received from the licensee.
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Asea - Brown - Boveria
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American Society of Mechanical Engineers AUT
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Automatic
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Boiler'and Pressure Vessel EDAS
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Enhanced Data Acquisition System ETC
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Et cetera ID
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Inside Diameter ISI
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Inservice Inspection
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Inspection Services Organization
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Numbers
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NRC'
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Nuclear Regulatory Commission
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Nuclear Reactor Regulations
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Program and Remote
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Revision
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Reactor Pressure Vessel Sequoyah
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SEQ
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SQ0
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Site Quality Organization SwRI
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Southwest Research Institute
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Tennessee Valley Authority
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Ultrasonic testing
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