ML20245H447

Insp Repts 50-413/89-14 & 50-414/89-14 on 890424-28. Violation Noted.Major Areas Inspected:Inservice Insp, Including Review of Automated Reactor Insp Sys Data from Unit 2 Hot Leg Nozzle inner-radius Indication Areas

ML20245H447
Person / Time
Site:	Catawba
Issue date:	05/25/1989
From:	Blake J, Coley J, Glasman M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20245H435	List: ML20245H432 ML20245H447
References
50-413-89-14, 50-414-89-14, NUDOCS 8906290488
Download: ML20245H447 (11)

See also: IR 05000413/1989014

Text

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NUCLEAR REGfjLATORY COMMISSION , [\\ REGION ll n

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ATLANTA, GEORGI A 30323 %....+,[ Report Nos.: 50-413/89-14 and 50-414/89-14 Licensee: Duke Power Company 422 Church Street Charlotte, NC 28242 ] 1 Docket Nos.: 50-413 and 50-414 ^ Facility Name: Catawba 1 and 2 i Inspection Conducted: April 24-28, 1989 Inspectors: n (_- 2 8Ci M. Glasman ~ Date Signed \\.Y 5'oY$ 9 A J. V Dat Signed ~ I Approved by: - . / 6 J. . B ak'e, Chief Date Signed Ma eri is and Processes Section E gir) ering Branch Division of Reactor Safety SUMMARY Scope This routine, unannounced inspection was conducted in the areas of inservice inspection and included a review of the Automated Reactor Insnection System (ARIS) data from the Unit 2 hot leg nozzle inner-radius areas in which indications were found. In addition, the inspectors followed the replacement of steam generator tube plugs identified as being susceptible to stress corrosion cracking; witnessed search for and removal of a loose part in a Unit 2 steam generator, and addressed previously opened NRC Unresolved and Inspector Followup Items. Also included in this report is a review of the Unit 1 End of Cycle 3 Inservice Inspection Report. Results Within the areas inspected, a major weakness in the area of protection of permanent plant equipment was identified, in that adequate procedural guidance in this area does not exist. (paragraph 3.e.) NRC inspectors witnessed ' personnel climbing and walking on equipment important to safety in the lower containment. In the area of inservice inspection, the licensee's progiam is adequate in . that personnel were knowledgeable and well-q'. lified to perform inspection their respective areas of certification, hot er, procedurt qualification documentation of examinations not fully performed to Code requirements was not available during the performance of the subject examinations, or for NRC review. i B906290480 890622 ~ PDR ADOCK 05000413 G PNU _ _ _ _ _ - - - - _ _ _ _ i

_ - _ - _ _ _ - _ _ _ - ____ - - _ . l '- . (; . . . , 2 I Within the areas inspected, a violation was identified: " Inadequate Procedural Guidance for Protection of Permanent Plant. Equipment" (paragraph 3.e.). One inspector followup item was identified involving the ARIS inspection of the reactor vessel het leg nozzles (paragraph 2.a.(2)). . , i - _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ . . _ . _ _ _ . _ _ . _ _ _ _ _

_ - _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ . F ' . l REPORT DETAILS l 1. Persons Contacted L Licensee Employees

• J. Barbour, Quality Assurance Director Operations

R. Giles, Site Inservice Inspection Coordinator

• R. Glover, Technical Support / Operations

R. Kaye. Maintenance Engineer Service Specialist

• V. King, Technical Support / Compliance
• P. LeRoy, Duke Power Corporation Regulatory Compliance
• T. Owens, Catawba Station Manager
• G. Robinson, Quality Assurance, Catawba

Babcock & Wilcox Employees G. Bryant, Eddy Current Task Leader M. Hacker, Supervisor, Ultrasonic Technology R. Patterson, Task Leader, Inservice Inspection NRC Resident Inspectors M. Lesser, Resident Inspector

• W. Orders, Senior Resident Inspector
• Attended Exit Interview

2. Inservice Inspection (ISI) Units 1 and 2 The inspectors reviewed documents and records, and observed activities as indicated below to determine whether ISI was being conducted in accordance with applicable procedures, regulatory requirements, and licensee commitments. The applicable code for ISI is the American Society of Mechanical Engineers Boiler and Pressure Vessel (ASME B&PV) Code, , Section XI,1980 edition with addenda through Winter 1981 (80 W' 81). Commercial operation commenced June 29, 1985, for Unit 1, and August 19, 1986, for Unit 2. Unit 1 has recently completed the third outage of the first 40 month period of the first ten year interval. At the time of this inspection, Unit 2 was in the second outage of the first 40 month period of the first ten year interval. Duke Power Company (DPC) nondestructive examination (NDE) personnel were performing the liquid penetrant (PT), magnetic particle (MT), radiography (RT), and visual examinations (VT) under the DPC Quality Assurance (QA) program. Babcock & Wilcox (B&W) and DPC personnel were conducting j ultrasonic (UT) examinations, under the B&W QA program. I I i ! ,

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. 2 4 a. Observation of ISI Work and Work Activities, Unit 2(73753) The inspectors observed examination activities, and reviewed NDE personnel qualification records for personnel that performed ISI examinations this outage. The observations and reviews are listed below. (1) Radiographic Examination (RT) The inspectors independently verified results of the RT examit,ation:: performed in the Units 1 and 2 outages. The inspectors also reviewed documentation associated with the respective examinations listed below to assure that the i ! examinations were consistent with Section XI. The inspectors reviewed the documentation and films to ensure that the following was evident and/or recorded: type of material; material and weld surface condition requirements; type of radiation source and its intensity; effectue source-to-film . distance; effective focal spot, or effective source size; film brand and type; number of films in cassette; minimum source-to-film distance; type and thickness of intensifying screens and filters; quality of radiographs; film density and contrast for single and composite viewing; use of densitometers for assuring compliance with film density requirements; system of radiograph identification; use of location markers; methods j of reducing and testing for backscatter; selection of l penetrameters, including penetrometer placement; number of penetrameters; shims under penetrameters; quality of film and its general condition; film storage adequate; proper evaluation of indications; proper density of film, and proper film viewing conditions. Unit 1 Radiographs Weld Identification Diameter x Wall Thk. System ISM 29-3 34" x 1.451" Main Steam ISM 29-7 32" x 1.459" Main Steam ISM 18-1 32" x 1.375" Main Steam Unit 2 Radiographs Weld Identification Diameter x Wall Thk. System 2SM 11-2 34" x 1.375" Main Steam 2SM 1A-I 34" x 1.375" Main Steam 2SM 11-3 34" x 1.375" Main Steam

_ __. _ _ . _ __ - , 9- ' . 3 l l Within the areas inspected, violations or deviations were not identified. (2) Evaluation of ARIS Data, Reactor Vessel Hot Leg Nozzle Inside . ' Radius During the week of April 17-21, 1989, Region II was notified that indications had been detected during the ultrasonic examinations of the inside radius of reactor vessel hot leg nozzles A, C, and D on Unit 2. On April 24, 1989, the ' inspectors arrived at the Catawba facility to review B&W's evaluation of the reported evaluation of the reported i ndi cations . Discussions with the Level III examiner revealed that B&W's preliminary evaluation of the indications indicated , that the ultrasonic reflectors appeared to be located totally i in the stainless steel clad. The Accusonex (B&W's i Automated Ultrasonic / Computer System) data further revealed that the indications were laying in a circumferential direction around the nozzles, in the same plane as the clad deposition, and spaced between each pass of the weld metal. The indications had initially been detected with a 70 degree

refracted longitudinal (RL) wave transducer which examines the i clad and base material for 1/2 inch below the clad. This transducer is excellent for near-surface detection, however, the long beam angle required for near-surface > detection tends to over size indications considerably. To j prevent distortion in indication size, B&W examined the J indications with focused 5 Megahertz (mHz) straight-beam transducer, and 40 and 60 degree angle beam RL transducers. The result of this ultrasonic enhancement revealed that the indications were not planar oriented like a ) crack, but had volume like an inclusion, and contained within i the clad weld material. The focused SmHz readily displayed the indications revealing width and no depth. The 40 degree and 60 degree RL transducers did not detect crack tips or facets to the indications. A test block with apparent inclusions and electro-discharge machined (EDM) notches in the clad and examination area also revealed similar type ultrasonic reflectors for visual / ultrasonic comparison. At the conclusion of the inspection, B&W had not completed their evaluation, since they intended to further enhance the focused 5 mHz data to further pin-point the exact location in the clad of the indications. However, the inspectors were confident that the preliminary eval M tions were sound, based on the data reviewed, and the DPC rLrporate QA supervisor agreed to pursue the following three concerns raised by the inspectors as a result of their examination of the evaluation activities: Clad welding fixture / process data should be determined by

contacting the fabricator prior to classification of the indications as a particular type of welding abnormality, i i

- _ _ _ _ _ _ . i ! . 4 The test block used for comparison should be , polished, etched, and examined with magnification on a l , side edge to determine the type of inclusion in the test block and whether the inclusions have any cracks running from them into the base material. l Demonstrate the sizing capabilities of the ARIS system (Accusonex) on real underclad cracks, using the same type of angle beam transducers used to evaluate the Unit 2 nozzle inner radius indications. B&W intends to perform this demonstration for the B&W owners group and Duke in several months at the Electric Power Research Institute in Charlotte, NC. The inspectors requested that the licensee notify Region II, Materials and Processes Section, when this demonstration is performed so that Region II can observe the capabilities of this system. The first two concerns will be addressed by the licensee prior restart. The third concern addresses system capabilities and limitations resulting from the inspectors' review of the system's display of the trailing tip on the EDM notches, and concerns that cracks would be even harder to detect. This item was reported to the licensee as Inspector Followup Item (IFI) 50-414/89-14-01, " Evaluation of ARIS Data." Within the areas examined, no violations or deviations were identified. (3) Visual Examination and Retrieval of Foreign Material in Steam Generator A, Unit 2 . 1 ' The inspectors observed B&W efforts to examine Steam Generator A Unit 2 (S/G 2A) for apparent loose material which had damaged three Alloy 600 tubes. This retrieval required the Licensee to bore an additional hole in S/G 2A near the level of the third support plate in order to expedite this retrieval effort. Several days of visual examination were required, and ' various fixtures for the remote video probe were tried before B&W located and retrieved the loose part. The inspectors observed B&W's visual examination to determine the condition of the Alloy 600 tubes in contact with the loose part. In addition, the inspectors observed the examinations to determine if upplicable drawings of the steam generator's internal structure and tube alignments were available; whether required tools and examination aids were available; if specific areas, locations, and extent of examinations were clearly defined, and if inspection results were recorded and reported in a prescribed manner. The loose part was retrieved on April 25, 1989, and consisted of a 3" x 5/8" x 1/4" tube wedge that had apparently been left in the steam generator since fabrication. - - _ _ _ ._. l

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L . , I 5 ' Within the areas examined, violations or deviations were not identified. (4) Observation of Steam Generator A Unit 2 Tube Plug Replacement l and Plug Repair In a B&W letter, from James H. Taylor of B&W Licensing Services to Dr. Thomas E. Murley, Director, NRR, dated September 6, 1988, B&W fridicated that certain steam generator (S/G) tube plugs fabricated from Inocnel 600, heat W-592-1, possessed a microstructure that may be susceptible to stress corrosion , I cracking (SCC). This letter identified the plants which had plugs installed that were manufactured from this heat of material. Catawba Unit 2 was identified as having six non-conforming plugs installed that were fabricated from Heat W592-1. The inspectors observed B&W replace these six plugs in S/G 2A with conforming material. In addition, three plugs were installed because of tube degradation caused by loose parts in S/G 2A as described above in paragraph 2.a.(3). The inspectors observed the plugging of the following tubes: Steam Generator A, Unit 2 Tube Number Location Replacement / Repair 49-54 Cold Leg W592-1 Replacement 49-39 Cold Leg W592-1 Replacement 24-67 Cold Leg

• W592-1 Replacement / Repair

24-68 Cold Leg Repair 24-69 Cold Leg Repair 15-77 Cold Leg W592-1 Replacement

• Denotes tube with non-conforming plug installed, and

degradation caused by foreign object on steam generator. The non-conforming W592-1 S/G tube plugs were only in S/G 2A, and have been in service since February 23, 1988, or the duration of one fuel cycle. The locations the non-conforming plugs occupied are shown in the table below: Steam Generator 2A Tube Number Location 15-77 Hot Leg 15-77 Cold Leg 24-67 Hot Leg 24-67 Cold Leg 49-39 Cold Leg 49-54 Cold Leg l _ _ _ ___ - - - - - - i

_ _ _ _ _ _ _ _ _ _ _ _ _ _ . . a i 6 At the time of the inspection, the licensee and B&W did not have any definite plans for laboratory evaluation or NDE to determine if the W592-1 plugs which were removed suffered i service induced degradation. Within the areas examined, violations or deviations were not- identified. b. Inservice Inspection Data Review and Evaluation, Unit 2 (73755) (1) Records of completed nuadestructive examinations were selected and reviewed to ascertain whether: the methods, technique, and extent of the examination complied with the ISI plan, and applicable NDE procedures; findings were properly recorded and evaluated by qualified personnel; programmatic deviations were recorded as required; personnel, instruments, calibration blocks, and NDE materials (penetrants, materials) were designated. The records selected for this review are listed below: j Item Number Weld / Component Method 802.011.002 2PZR-W8E UT Pressurizer B02.012.002 2PZR-W9D UT Pressurizer B09.011.044 2NC25-05 UT Reactor Coolant B09.011.312 2N161-05 UT Safety Injection B09.011.021 2NC13-W2 UT Reactor Coolant B05.010.005B 2RPV202-121ASE PT Reactor Vessel ' l Hot Leg Safe End l B05.040.002A 2PZR-W2SE PT Pressurizer Safe End C05.011.181 2ND13-7 PT Decay Heat Removal C05.011.300 2NI18-01 PT Safety Injection I

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_ _ _ _ _ _ _ , $- p .xp l ' %: <y , ,p'. . . ' :. + , IMAGE EVALUATION <^% < Y/o '\\ // [ $ff TEST TARGET (MT-3) / f.[6 44 k k '" " I.0 .: ll M i,l M M4 ' 1.25 1.4 1.6 __ 4_ . _ . _ . . . - 150mm > 4...._._... _ . - - - 6" - -> b'//4% ., ,Y ggi e, , p' ,, // w, ,ys ' h s{p 4c p _ . . -am,____, ,_.__m ,

y,, _ l[ ( 7 ' . - - I, - . " 7 , , d} , B09.021.011 2NC44-23 PT ' " Reactor Coolant C05.011.322 2NIf0-06 PT Sa' sty Injection C05.011.400 2SM25-01 MT Main Steam C05.012.300 2SV-02L MT Main Steam C05.021.054A 2CF-27-C MT Feedwater , B07.070.049 2NI-175 VT-1 Safety Injection B07.070.023 2ND-36B VT-1 Decay Heat Pemoval y F1.03.959 2-R-TE-0021 VT-3 Feedwater Turbine 2-R-KC-0414 F1.03.269 VT-3 Component Cooling Water (2) Review of Catawba Unit 1 Cycle 3 Inservice Report The Inspectors reviewed the DPC Inservice Summary Report for Catawba Unit 1, cycle 3 refueling outage. This outage was completed February 6,1989. The report was reviewed to determine whether the inspections performed were consistent with the requirements of Section XI of the ASME B&PV Code, applicable addenda, and licensee commitments. Specific areas examined for compliance were: minimum examinations completed; o limitations recorded; examination methods recorded, and indications were recorded, evaluated, and reported as required. Within the areas inspected, violations or deviations were not identified. 3.- Action on Previous Inspection Findings (92701) a. (0 pen) URI 50-413,414/89-08-01, "MT Demonstration" L In the performance of the MT examinations on reactor vessel studs and nuts, the licensee used less amperage than Section XI requires. The Code, however, allows this if it can be shown to the r _- _-.

_ - _ _- _ - _ - _ __ _ __ _- - - - _- --- . _ _ t ,. , . . . ' 8 satisfaction of the Authorized Inservice Nuclear Inspector (ANII) that the _ alternate method (s) are equivalent or superior to the Code-specified method. The inspectors found that a procedure < qualification document approved by the ANII was on file for the alternative MT exam.for the nuts, however, the procedure , ! qualification documentation for the studs ras not prepared at the time of the inspection. The inspectors found, however, that the ANII did, in fact, witness, and approve the alternate MT examination for the studs. Pending NRC review of the DPC procedure qualification documentation, this item will remalr. open. b. -(0 pen) IFI _50-413,414/89-08-02, "ISI Baseline Inspection Method" The Code requires that repair work be inspected using the same method that was used to detect the defect prior to performance of the required repair work. Two DPC NDE procedures (NDE 25, Revision 12, and NDE 35, Revision 10) did not incorporate this requirement, and at the time of the inspection, were being revised. Pending NRC review of the revised NDE procedures, this item will remain open. _ c. (Close)IFI 50-413,414/89-08-03, " Unavailable ISI Recoros" Spectrum analysis data for transducer numbers M18416, M18423, and data indicating hot leg wide range temperature instrument A0668 was in calibration were shown to the. inspectors. This item is closed, d. (0 pen) IFI 50-413,414/89-08-04, " Valve Identification" . A Valve tagged "27SV27A". was incorrectly identified as "2SV027" on a DPC drawing used by inspectors as reference in performing ISI. The subject valve is a S/G power operated relief block valve; the "A" indicating either A or B train power for the valve. Further review by the licensee indicated that additional drawings which included this valve required correction, however, the inspector found that the valve was correctly identified on the Unit 2 main control board. The inspectors also checked nine other A and B train valves for consistency between drawings (Flow Isometrics and Design Isometrics) . and tagging on the main control board and found no additional inconsistencies. Pending NRC review of the revised DPC drawings, this item will remain open. e. (Close) URI 50-413,414/89-08-05, " Protection of Installed Plant Equipment" During a tour of the Unit 2 Containment lower level, the inspectors confirmed that personnel were climbing and walking on safety-related equipment, including cable trays, small-bore piping, and pipe supports. At one point, the inspectors observed a number of personnel climbing down a long, vertically -run cable tray, scaling its rungs as if it were a ladder. The inspectors also found an . _ _ - - _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ __ j

.____ ___ __ __-_____ _ - . . 4 9 - example of a damaged line (WL400-12) during this tour at the 558' 270 degree location in the pipechase. The line, part of the liquid waste system, carries radwaste and was obviously struck with a heavy object, and badly dented at the point of impact. At the time of the inspection, the inspectors found that the licensee's procedure, CNS Station Directive 3.11.1, page 4, states, in part, that " Adequate working space shall be provided where possible utilizing proper work stages and platforms having accessibility by stairs or a ladders." Discussions with the licensee indicated, however, that this Station Directive was not intended to provide guidance to personnel regording protection of installed plant equipment during regular moveme nt through the containment. Based on these observations, this unresolved item will be closed, and elevated to a violation for eack of procedural coverage for protection of i permanent plant equipment. This matter will be identified as Violation 50-413,414/8914-02, " Inadequate Procedural Coverage for Protection of Permanent Plant Equipment." 4. Exit Interview The inspection scope and results were summarized on April 28, 1989, with those persons indicated in pa;agraph 1. The inspectors described the areas inspected and discussed in Jerail the inspection results listed below. Although reviewed during this inspection, proprietary information is not contained in this report. Dissenting comments were not received from the licensee. Item Number Description and Reference 50-414/89-14-01 IFI, " Evaluation of ARIS Data", Paragraph 2.a.(2) 50-413, 414/89-14-02 Violation, " Inadequate Procedural Coverage for Protection of Permanent Plant Equipment," Paragraph 3 Licensee management was informed that one IFI and one URI discussed in paragraph 3 were closed. -. . __ . - _ _ _ _ . __ ._ _ - - - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ - _ - _ - _ _ _ _ _ _ _ _ _ _ - - . - }}