IR 05000348/1990031
| ML20058K210 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 11/23/1990 |
| From: | Blake J, Chou R, Newsome R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20058K207 | List: |
| References | |
| 50-348-90-31, 50-364-90-31, NUDOCS 9012170032 | |
| Download: ML20058K210 (21) | |
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UNITE 3 ST ATES
/pa rt;oq'o NUCLEAR REGULATORY COMMISslON
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[' n RE GION 11 lt #g 101 MARIETTA STREET, * t ATL ANTA, GEOROl A 30323
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Repcrt Hos.: 50-348/90-31 and 50-364/90-31 Licensee: Alabama Power Company Power Company 600 North 18th Street Birmingham, Al 35291-0400 Docket Nos.: 50-348 and 50-364 License Nos.: NPF-2 and NPF-8 l Facility Nane: Farley 1 and 2-Inspection Conducted: . ' October 29 - Noveniber 2',1990 1nspectors: 0A lA] ~7 ,mu .
// M - 90 Date Signed L') C .Newsome s O( '
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ho Date Signed Approved by: -
// 23ha Blake, Chief Date Signed a rials and Processes Section n ineering Branch Division of Reactor Safety SUMMARY Scope:
This routine, announceo inspection was conducted in the areas of Inservice inspection (ISI) including the eddy current examination of the Unit 2 Steam Generator (SG) tubing and snubber maintenance program. The inspection included a review of the Unit 2 ISI inspection plan. for this outage; reviews of examination procedures; observations of in-progress examinations; independent examination verifications; reviews of personnel qualifications; reviews of nondestructive examination (NDE) equipment calibration and material certification documentation; and, a review of completed examinetton data. Also, NRC previously opened items were addresse Results:
In the areas inspected, violations or deviations were not ic entifie This inspection indicated that ISI examinations and snubbea maintenance wcre being adequately controlle PDR ADOCK 05000340 i G PDR g
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Previous inspector followup item IFI 50-364/90-12-03 was followed up regarding - 1 repeated cracking of supports on-Main' Steam Line C. The inspector reviewed
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numerous technical analyses of the problem and concluded, that the failure ;
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mechanism and . fix has not yet- been- found for certain. This matter will' be :
reviewed further during a future inspectio ,
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REPORT DETAILS
, Persons Contacted Licensee Employees R. Badham, ISI Engineer i
- R. Berryhill, Systems Performance Manager
- S. Casey, Systems Performance Supervisor
- M. Dove, Snubber Maintenance Engineer S. Fulmer, Safety Audit and Engineering Review Supervisor
- H Garland, Mechanical Maintenance Superintendent
- D. Hartline, Systems Performance Engineering Supervisor
- R. Hill, Assistant General Manager, Operations D. Morey, General Manager, Operations
- S. Norman, Mechanical Maintenance Supervisor
- J. Sims, Project Engineer
- L. Stinson, Assistant General Manager, Support
- J. Thomas, Maintenance Manager Other licensee employees contacted during this inspection included craftsmen, engineers, security force members, technicians, and administrative personne Other Organizations J. Campbell, ISI Coordinator, Westinghouse Electric Corp. (W)
E. Conrad, Westinghouse Electric Technician L. McClain Southern Company Services, level III C. !isu, Civil Group Supervisor, Bechtel Power Corporation G. Lushbaugh, Assistant Project Engineer, Bechtel Power Corporation C. Vaz, Site Engineer, Bechtel Power Corporation-T. Damico, Senior Consultant, SMC O'Donnell, Incorporated T. Hazlett, Senior Engineer, SMC O'Donnell, Incorporated NRC Resident Inspectors j
- Maxwell, Senior Resident Inspector M. Morgan, Resident Inspector
- Attended exit interview Acronyms and Initialisms used throughout this report are listed in the last paragrap , .
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! Inservice Inspection
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The inspectors reviewed documents and records, conducted independent evaluations, and observed activities, as indicated below, to determine whether ISI was being conducted in accordance with applicable procedures, reaulatory 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,1983 edition with addenda through Summer 1983. Alabama Power personnel are primarily acting as coordinators for 151 contractor personnel. Westinghouse has the primary responsibility as the ISI contractor for conducting the ultrasonic (UT), liquid penetrant (PT), magnetic particle (MT), visual (VT), and primary eddy current (EC)
steam generator tubing data evaluation and collection while Conam Inspection is conducting a second evaluation of all the EC data. Alabama Power maintenance personnel conducted all snubber testing and examination Southern Company Services is performing overview functions for the licensee in all ISI NDE areas, ISI Program / Plan Review, Unit 2 (73051)
' The inspectors reviewed the inspection plan for this outage Unit No. 2 Inservice Inspection Interval-1 Period-3 Outage-2, to determine whather the program / plan had been approved by the licensee and to assure that procedures and plans had been established (written, reviewed, approved and issued) to control and accomplish the following applicable activities: organizational structure including qualifi-cations, training, responsibilities, and duties of personnel -
responsible for 151; audits including procedures, frequency, and qualification of personnel; general Quality Assurance requirements including examination reports, deviations from previously established program, material certifications, and identification of components to be covered; work and inspection procedures; control of processes including suitably controlled work conditions, special methods, and use of qualified personnel; corrective action; docunient control; l control of examination equipment; quality records including documenta-tion of indications and NDE findings, reviow of documentation, provisions to assure legibility and retrievability, and corrective .
action; scope of the inspection including description of areas to be ;
examined, examination category, method of inspection, extent of )
examinations, and justification for any exception; definition of j inspection interval and extent of examination; qualification. of NDE personnel; and, controls of generation, approval, custody, storage and naintenance of NDE record l The review of the 151 plan indicated that the plan was properly approvod and contained the necessary information.
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b. Review of NDE Procedures, Units 1 and 2 (73052) i (1) The inspectors reviewed the procedures listed below to determine whether these procedures were ' consistent with regulatory requirements and licensee commitment The procedures were also reviewed in the areds of procedure approval, requirements for !
qualification of NDE personnel, and compilation of required records; and, if applicable, division of responsibility between the licensee and contractor personnel if contractor personnel are involved in the ISI effor FNP-0-NDE-157.1(R2) Preservice and Inservice .
Inspection Documentation 1983 Code
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FNP-0-NDE-157.5(R3) Manual Ultrasonic Examination of Bolts Studs and Nuts 1983 Code
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FNP-0-NDE-157.7(R3) Manual Ultrasonic Examination of Welds In Vessels 1983 Code
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FNP-0-NDE-157.12(R2) Manual Ultrasonic Examination of Welds 1983 Code
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FNP-0-NDE-157.14(R4) Manual Ultrasonic Examination of Welds In Cast Stainless Steel Pipe 1983 Code
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FNP-0-NDE-157.17(R3) Manual Ultrasonic Examination Of Inner Radius Corners 1983 Code
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FNP-0-NDE-157.18(R4) Ultrasonic Examination of Studs and Bolts from the Bore Hole 1983 Code
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FNP-0-NDE-157.4(R2) Liquid Penetrant Examination 1983 Code
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FNP-0-NDE-157.11'(R3) Magnetic Particle Examinations 1983 Code i
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FNP-0-NDE-157.3(R2) Visual Examination VT-1 1983 Code
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FNP-0-NDE-157.16(R2) Visual Examination YT-3 1983 Code
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FNP-2-STP-610.2(R11) Accessible Snubbers Visual Inspection
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MRS2.4.2APC-6(RS) Digital Multi-Frequency Eddy Current. Inspection of Preservice and Inservice Heat Exchanger Tubing
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All procedures listed above have been reviewed during previous imC inspections. Only current revisions were reviewed during '
this inspectio (2) The inspectors reviend the Ultrasonic procedures to ascertain '
whether they had been reviewed and approved in accordance with the licensee's established QA procedures. The procedures were '
also reviewed for technical adequacy and conformance with ASME,Section V, Article 5 and other licensee commitments / requirements in the following areas: type of apparatus used; extent of coverage of weldment; calibration requirements; search units; beam angles; DAC curves; reference level for n.onitoring discontinuities; method for demonstrating penetration; limits for evaluating and recording indications; recording significant indications; and, acceptance limit (3) The inspectors reviewed the Liquid Penetrant procedure to ascertain whether it had been reviewed and approved in accordance with the licensee's established QA procedures. The procedure was also reviewed for technical adequacy and conformance with ASME,Section V, Article 6, and other licensee commitments / requirements in the following areas: specified method; penetrant material identification; penetrant materials analyzed for sulfur; penetrant materials analyzed for total halogens; surface temperature; acceptable pre-examination surface conditioning; method used for pre-examination surface cleaning; surface drying time prior to penetrant application; method of penetrant application; penetrant dwell time; method used for excess penetrant removal; surface drying prior to developer application, if applicable; type of developer; examination technique; evaluation techniques; and, procedure requalificatio (4) The inspectors reviewed the Magnetic Particle proceduro to ascertain whether it had been reviewed and approved in accordance with the licensee's established QA procedures. The procedure was reviewed for technical adequacy and for conformance with the ASME Code Section V Article 7, and other licensee commitments / require-ments in the following areas: examination methods; contrast of dry powder particle color with background; surface- temperature; suspension medium and surface temperature requirement for wet particles; viewing conditions; examination overlap and directions; pole or prod spacing;. current or lifting power (yoke);and,acceptancecriteri (5) The inspectors reviewed the Visual examination procedures to determine whether they contained sufficient instructions to assure that the following parameters were specified and; controlled within the limits permitted by the applicable code, standard, or any other specification requirement: method ' -
direct visual, remote visual or translucent visual; application'-
hydrostatic testing, fabrication procedure, visual examination of
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I welds, leak testing, etc.; how visual examination is to be performed; type of surface condition- available; method or implement used for surface preparation, if any; whether direct or i remote viewing is used; sequence of performing examination, when '
applicable; data to be tabulated, if any; acceptance criteria is specified and consistent with the applicable code section or controlling specification; and, report form completion, (6) The inspectors reviewed the Eddy Current procedure for technical content relative to: multichannel examination unit, multir.hannel-examination indication equipment is specified, examination sensitivity, method of examination, method of calibration and calibration sequence, and acceptance criteri All procedures reviewed appeared to' contain the necessary elements for conducting the specific examination, c. Observation of Work and Work Activities, Unit E (73753)
The inspectors observed work activities, reviewed certification t records of NDE equipment and materials, and- reviewed NDE personnel qualifications for personnel that had been utilized during the required ISI examinations during this outag The observations and reviews conducted by the inspectors are documented belo (1) The inspectors observed calibration activities'and the in-process l ultrasonic (UT) examinations being conducted on 3 Main Steam Header welds identified on drawing APR-2-2110 as welds 4-1, 4-3, and 4-3LS. These observations were compared with the applicable ,
procedures and the ASME B&PV Code in the following areas: !
availability of and compliance with approved NDE procedures; use of knowledgeable NDE personnel; use of NDE personnel qualified to the proper level; type of apparatus used; calibration j requirements; search units; beam angles; DAC curves; reference '
level for monitoring discontinuities; method of demonstrating isnetration; extent of weld / component examination coverage; iimits of evaluating and recording indications; recording significant indications; and, acceptance limit The inspectors conducted an independent ultrasonic verification !
examination, using W equipment, on portions of weld 4-1 l previously observed Eeing examined by the ultrasonic examiner This examination was conducted in order'to evaluate the technical adequacy of the ultrasonic- examination procedure being used by the licensee and to assess the validity of the information being
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reported by the ultrasonic examiner The verification ultrasonic examination conducted by the inspectors indicated that the procedure being used to conduct the 1 examinations is adequate and the verification examination results compared favorably with the information being reported by the !
ultrasonic examiner i l
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The following listed ultrasonic equipment and materials certification records were reviewed:
Ultrasonic Instruments Manufacturer /Model - Serial N Sonic /MK I 781304 Sonic /MK I 00890E a Sonic /MK I 07856E- )
Sonic /MK I 07855E
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Ultrasonic Couplant Batch Numbers 8874-and 8979 The inspectors reviewed spectrum analysis data-for the ultrasonic transducers listed below:
l Serial N Size Frequency J08327 .5" x 1" 2.25 MHz 00833T .5" x 1" 2.25 MHz l
Ultrasonic Calibration Blocks ALA-23,_ALA-24. and ALA-25 i (2) The inspectors observed the in-process liquid penetrant -(PT)
examinations of 4 RHR Return Split to the SIS pipe . welds identified on drawing APR-2-2317 as welds 4, 5, 6, and 29. -The observations were compared with the applicable procedure and the ASME B&PV Code in the following areas: specified method,_
penetrant materials identified; penetrant materials analyzed for halogens and sulfur; acceptable pre-exemination-' surface; surface temperature; surf ace orying time-prior to penetrant application; method of penetrant application; penetrant dwell time; method used for excess penetrant removal; surface . drying 1 prior to developing, it - applicable; type of developer; examination-
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technique; evaluation technique; and, reporting of ' examination result The inspectors re-evaluated welds 6 and 29 following the PT l examiners evaluation of the welds but prior to. the developer being removed from the weld surfaces. This. re-evaluation was conducted in order to determine if the evaluations performed by the PT_ examiners was in accordance-with the applicable ^ procedure i acceptance criteria and to determine if the examination.results were being reported as. required. The re-evaluations conducted by the NRC inspectors indicated that the proper evaluation was made by the PT examiners and that the examination results were beitig reported as require *
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halogen content of the material'was' within acceptable content' .
limit '
Materials Batch' Numbe .
Liquid Penetrant 88L88K: +
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Cleaner / Remover 90009P Developer 89H075S (3). The inspectors observed 1 in-process magnetic particle (MT)
examination of a Boron Injection Tank, upper nozzle to tank! weld identified on drawing APR-2-1230 and weld 3. The observation was compared with the l applicable' procedures and the Code- in the following, areas: examination methods; contrast of dry powder particle cCor with background;- surface temperature; suspension medium for wet particles, if. applicable; viewing conditions;
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examination overlap and directions; pole or prod spacing; curren or lifting power (yoke); and acceptance criteri ;
The examination appeared to be adequat The inspectors reviewed documentation indicating that a-10 pound lift test had been ,' performed on magnetic particle alternating current yoke MT-100. The certification record for the lift test plate that was used to conduct the. test, NSG-0084, was reviewed =
to confirir the weight of the test plat A review of the magnetic particle material certification records for batch numbers 86G028 and 87A003 indicated the particles met the applicable specifications requirement (4) Independent Verification of Visual Examinations The licensee's ISI program contains a total population of 454 pipe supports which are subject to VT-3 examination under requirements of ASME Section XI. Of the 454 supports,145 were scheduled for examination during this inspection interval and 40-were scheduled for examination during this refueling outage. Most other supports and snubbers required for this inspection interval had been completed since the last refueling outage. In addition, 1 100 percent of al.1 snubbers are-required to' be examined during each refueling outage per~ Technical Specification'4.7.9 a and b.- l The total snubber population is 696 and is : divided into two '
types, of which, 396 are mechanical ' snubbers and 300 are I hydraulic snubber A portion of. these snubbers .are also included in the 454 pipe support population and the requirements of ASME Section XI 'for VT-3 visual examination appl Therefore,
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in some instances,Jsome of the snubbers must be examined under ,
the dual requirements of ASME Section' XI and Technical Specification (a) The NRC inspectors conducted independent visual examinations [
of 17 pipe supports and snubbers selected at random.:These '
items included- 2 mechanical and 8 hydraulic snubbers, examined under Technical Specification requirements,land 7L pipe supports examined under ASME Section XI requirunents',
that are located -in the Auxiliary Buildin These examinations were conducted in order to evaluate the-adequacy of- the examination procedures, being used by. the
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licensee and to assess the validity of the information'being I reported by the examiner These verification examinations I generally agreed with the findings of the-visual examiners, ;
u During the review' of the inspection procedure and the verification examinations of snubbers, the-inspectors noted the following:
' The licensee did not perform an examination on the snubber ' structures (including gang supports), for gaps and tolerances ~ between the spherical bearings and washers or clevis at each end of snubbers.- I excessive gaps exist between the: spherical bearings and washers or clevis, the pins may be damaged due'to the wear from constant movements or impact from the sudden lock-up of snubbers, therefor inspection and established tolerances for the gaps are necessar The snubbers were examined between pin to pin, but an- ,
overall examination 'is not conducted to -include the supporting structures and foundations 1to: insure the snubber will perform its intended functions. Defects-or degradation of the structural components can have a safety impact on the systems operability and the -
ability of the snubber to functio _n properly. The defects or degradations could include crach of welds -
and members, deformation of members, ' loose nuts, corrosion, et Snubbers are required.to have established hot and cold settings, depending on the system condition, which are stated on the snubber's detail drawings ~. The licensee did record' the cold or hot settings in the inspection sheets. The cold or hot settings were compared to Table 1, Minimum and Maximum Piston Rod' Measurements, of inspection procedure No. FNP_-2-STP-610.2._ This table was established based on the two extreme cases of the e -.4
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minimum and maximum piston movement location The j licensee stated that the field inspection personnel checks the cold or hot settings to insure theyiare- l within the minimum and maximum ranges and the snubber engineer compares the settings against the cold or, hot l settings contained in the computer data _ base which stores the settings information based on the detaile drawing If the . settings are not acceptable, the
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'l snubber engineer will request the design engineerL to evaluate the settings. If the actual settings are not in. the proper positions the snubbers could be damaged-(bottom out or- top out) duringlthe next hot > or cold -
movement, therefore, the actual settings being compared: l to the ' designed ' settings are very important. :The j licensee's snubber engineer indicated that thel actual l settings- have been compared to the data base, but,' '
there are no' requirements in the procedure to -;
accomplish this and' there is no objective evidence '
indicating that the' settings were compare To insure that an adequate examination is accomplished,-the'
licensee agreed to add the above items to the inspection'
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procedures in the next revision. .
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(b) During the review of seven pipe support drawings which were used for the ISI walkdown verification, the inspectors'found ,
that three supports 2RHR-R72, 2RHR-583, and 2CVC-R228 had- i used Wej-It anchor. bolts in the: original construction and modifications. . . Design capacity problems for Wej-It anchor bolts were. first reported to NRC'several: years ago by
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Crystal River Nuclear Power Plan The Crystal River Plant tested the capaci_ty of Wej-It anchor-bolts at the site and --found that the tested capacity was-about 50 to 60 percent of the_ capacity stated ins the manufacturer's catalog. The apparent cause of the poor Wej-It anchor bolts performance was determined to be the low hardness value of the materials, such as aggregates, sand, etc.. used in~the bonding concrete. The Crystal River Plant revised all of the ~ support calculations containing the Wej-It anchor bolts, based. on. the tested capacity, and .
modified the supports, where required, in the fiel ,
't In addition, Turkey Point Nuclear Power Plant was found by the NRC to have used Wej-It anchor bolts 'on site and they were requested to test the design capacity of Wej-It anchor i bolts on sit The test results showed similar results to l those reported by= Crystal Rive Turkey Point also revised-all of the pipe support calculations, based on the test results, and modif.ied the supports as require .. . ,
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Therefore, based on the findings of,these tests, the design I capacity stated in the manufacture's catalog is not reliable - l due_to the strength differences of the materials contain in '
the construction concrete and _the test samples used in the catalo After-discussing the matter with the licensee's engineers ,
and the design supervisor of Bechtel. Power Corporation, the
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inspectors. understanding of this matter is'as -follcws: 1
- The Wej-It anchor bolts were used during the original- <
construction and for some modifications required for IE Bulletins 79-02 and 79-1 The use of Wej-It anchor bolts was discontinued at i Farley Nuclear Plant during the- support I modifications. for-IE Bulletin 79-14.' pipe
- The total number of pipe supports using Wej-It anchor bolts is unknow >
- The design capacity of the Wej-It anchor bolts used in the pipe support calculations was apparently based onL the manufacturer's catalo The licensee's engineers indicated that there might have been a: pullout capacity test accomplished during -
the anchor -bolt test ' program for IE Bulletin 79-02, however,. the licensee was not able to retrieve the 4 information during this inspectio Based on the results of the testing by Florida Power Corporation and Florida Power and Light, the NRC has concerns about the use of Wej-It anchor bolts using the design capacities specified in the manufacturer's catalo Pending a review of the number and location of Wej-It anchor 4 bolts installed at Farley; the results of any testing done '
to support the use of Wej-It bolts during IE Bulletin 79-02 -
work; and, the design capacities used in the support calculations, this item is identified as Inspector Followup Item (IFI) 50-348, 364/90-31-01, Possible Design Capacity Problem For Wej-It Anchor Bolts. This item will be reviewed during the Unit-1 refueling outage in the spring of 199 (5) Steam Generator (SG) Tubing Eddy Current (EC) Examination The inspectors observed the EC activities indicated belo The l observations were compared with the applicable procedures and the Code in the following areas: method for maximum sensitive is ~
applied; method of examination has been recorded; examination l equipment has been calibrated in accordance with -the applicable l
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performance reference; amplitude and phase angle Lhave . been calibrated with the proper calibration reference - and is recalibrated at predetermined frequency; required coverage of steam generator tubes occurs during the examination; acceptance:
criteria is specified: or referenced and is consistent with; the . :1 procedure or the ASME Code; and, results are consistent with thet acceptance criteri (a) Steam generator tube eddy' current data col _lection was -
being accomplished by W personnel. . In-process tube data ,
acquisition, including calibration confirmation and tube-location verifications, was observed for 30 SG tubes, some from each S ,
(b) In-process eddy current -data evaluatio including ,
calibration confirmation, was observed - for. 56. SG tube '
Primary data analysis, being- conducted -by W,;was observed for 46 SG tubes, 24 from SG-A and 22 from fG-C. . Secondar data analysis, being conducted by Conam, was~ observed.for'10 :
SG-A tube The inspectors co-evaluated -27- of' the SG Ltubesi during the observations of the primary and secondary anaiysts evaluations, 1 10 f rom SG-A and 17 from SG.. The-sample sf evaluations, some 1 having reportable indication! nd nme with no reported indications, was conducted in order'10 confirm the validity of-the repoited tubing condition. . The co-evaluation analysis .
conducted by the inspectors' agreed well with the _ reported result q Certification records for EC calibration standards MGT-005-90 and MGT-016-90 were reviewed for material type. correct fabrication, and artificial flaw location and siz .
(6) The inspectors reviewed 28 personnel qualification documentation records including EC personnel, Alabama' Power maintenance-
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personnel, and W visual examiners. These personnel qualifications were reviewed fn the following areas: employer's name; person certified; activity - qualified to . perform; current period- of certification; signature of employer's. designated representative; basis used for certification; and, annual visual acuity, color vision examination, and periodic recertification, d. Data Review and Evaluation, Unit 2(73755).
(1) . Records of completed examinati.ons for 7 UT, 47 PT, 1 MT, and ,
18 VT examinations were selected and reviewed to ascertain l whether: the methods (s), technique, and extent of the examination complied with the ISI plan and applicable NDE I procedures; findings were properly recorded and evaluated by l
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qualified personnel; programmatic. deviations were recorded as j required; personnel,- instruments, calibration blocks, and NDE !
materials (penetrants, couplants) were. designate (2) The inspectors reviewed the eddy current data analysis results and a sample of associated completed records _for'60 SG tubes, 20 tubes from each of three Steam Generators. The reviews were compared with the applicable procedures and the ASME B&PV Code in the following areas: the multichannel eddy current examination equipment has been identified; material permeability has been recorded; method of examination has been recorded; and, results- I are consistent-with acceptance criteri All cf the examination reports reviewed ' appeared to ' contain th I required examination information including disposition ,of indications, if an A random sample of current examination resultsL were compared with historical examination result No major- discrepancies were_ noted during the compariso ,
In the areas inspected, violations or deviations were not identifie . Licensee Action on Previously Opened Items-(92701)
(0 pen) IFI 50-364/90-12-03, Evalut tion .of Cause and Corrective Action on Failure of Steam Line Supports 2MS R84.and 2MS-R8 This matter concerned the weld anc ' member cracks- on supports' 2MS-R84 an MS-R85 for Main Steam Line C. Supoort 2MS-R84 has'a history of cracks' and repair The inspectors reviewed 1.he information provided and discussed this matter with the licensee's eng:neers and engineers from consulting- 3 companies - Bechtel Power Corporation and SMC O'Donnell Inc. Main Steam - '
Line C is longer than A and B and has five straight portions and four turns in a total of 110' and two vertical drops ofl 47' and 10' respectivel There are no axial restraints except at. each_ end 'of the line and two snubbers located along the line. Support 2MS-R84 is a vertical rigid' type support with sliding surfaces. The loads eventually are transferred to a ,
connection of 6 X 6 X 1/2 and 8-X 8 X 1/2' inch steel tubes. The recurrence of cracks were found at this connection'. Support 2MS-R85 consists of two snubbers which act as an axial restraint'whose pipe attachment _ location is- '
only l'- 31/2" from support 2MS-R84 and whose other 'end is connected to the structural steel of support 2MS-R84. 'This axial restraint (snubbers)
is supposed to resist huge quick movement loads induced by earthquake, pipe breaks, or steam hammer. A connection of support 2MS-R85 was repaired twice due to the cracks on welds and structural ' members. The. connection between the 6 x 6 and 8 x 8 inch tube steel for : support 2MS-R84 has been broken and repaired six time A history of this connection is shown
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Crack Crack Weld . Crack PCN N ..
Sequence Time Detail Locations Issued-l-
1 Oc /16" fillet welds _
- All around weh' 83-2-2450-1983 around intersection'of area at tube tubes 6"x6"'and 8"x8" 8"x8" .
2 Ja /8" fillet welds Tube 6"x6" 85-2-3077 1985- around' intersections of tube 6"x6",_ wing plates, s 3/4".end plates, and
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three tubes,8"x8"
- 3 Ap Reinfo'rced tube 6"x6"' Tube 6"x6" and- 86-2-3623 1986 wi'.h 1/2" cover plates 3/4" end plates- 1 at top and.two side around the' edges The iest was same as of horizontal abov _ wing plates
- 4 1/2" fillet welds
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Oc 'At tube 8"x8" 87-2-4274 1987 around the intersection. around and out-of tubes 6"x6" and side weld area 8"x8" -of tube 6"x6" 5 No Pin connection between 1" long hair-1988 tube 6"x6" and-two line crack in gusset plates (no gaps). weld and gusset which welded to tube plates 8"x8" with 1/4" fille weld Ma Same connection as crack './8" wide 89-2-5763 1989 above at the top of two gisset plates, tapered '
down esenly through-the weld to a hairline at the botto Se The same type of con- 11" long hair-1989 nection shown above ex- line cracks in cept 1/8" gaps at both weld and 3/4" sides between tube and gusset plates gusset plates in order to reduce vibration'-
effects
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I Crack Crack Weld Crack- PCN N Sequence Time Detail Locations Issued (cont'd) :
Ap Same connection as, 1/2" wide crack 90-2-6581*** 1 1990 shown above (widest) in weld J and'3/4" gusset plates from top ;
-to. bottom
Cracks were first' found at~ connections on supports 2MS-R85, R89, <
R90, and R98. Supports R89, R90, and RS8 are in Main Steam-Lines
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- Cracks on two connsctions of support 2MS-R85 were found two-separate time *** This Production Changt. Notice (PCN) will be implemented du' ring this refueling outage to use full penetration welds between two gusset plates and the reinforced cover plates on the 8"x8" tube, instead of fillet welds used previousl The licensee perform a walkdown inspection for all supports and a second walkdown inspection of the pipe and the supports at support' locations with the insulation removed at the support locations on Main Steam Line C. The licensee also performed inspections on nearby supports each time cracks were found at support 2MS-R84 The inspections;of the adjacent supports did not find any evidence of cracks, deformation, wearout, etc. on the adjacent supports or piping except those problems stated above. . Following the failures on suoports in lines - A and B in 1986, Bechtel Power-Corporation, an Architecture / Engineer firm, submitted twelve samples of failure materials from various supports to its laboratory, Bechtel
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National, Inc., for testin Bechtel National, Inc concluded that the suspected mode of failure was fatigue and the failure was not due to ductile overload based on the scanning electron microscopic examination on the fracture surface This conclusion was contained.in an interoffice memorandum, File No. TCW-076-01, dated-July'l', 190 On April 25, 1986, Southern Company Services (SCS), an Architecture and-Engineering firm, mainly for Georgia; Power. Company and Alabama Power
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Company, was-brought in by the licensee to' investigate the crack problems after the cracks in the various supports in lines A, B, and C were discovere SCS installed a monitoring system- consisting of thirteen strain gages to monitor movement' and six accelerometers to monitor vibration to detect vibration and movements on the previous failure
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locations at various supports and piping.in all main steam lines in Unit The monitoring system was installed on the repaired areas before the containment was closed on May 1,1986. Vibrations and stresses on the supports were to be monitored as the plant returned to normal full-load operating conditions in order to identify the occurrence and possible cause of fatigue and other stresses in the supports. - The system monitored l
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vibration and movement from May 12,1986 -(startup following the refueling-outage) to July 18,1986 (shutdown to resolve charging pump problems).
During the monitoring period..the power had reached 97 percent constant and came down to 55 percent quickly at one tim The test results and conclusion were summarized by SCS in a report- of'
" Measurement of Strain and Vibration on Main Steam Line Restraints for-Unit 2 Farley Nuclear Plant", dated July 1986.which wasJattached to SCS letter file: ENG 15-86-652, 86-2-3623, N11 for FNP:86-1183, dated August 5, 1986. The results shown .for- vibration were low frequencies, vibration increased with power, and the source of vibration came from the pipe due to its displacements being longer than ones in the supports. The reported results . indicated maximum stresses in steady strain 'due to movements in loading changes such as temperature variation, steam flow,- ,
etc. and were -5800 psi during heatup and 15570 psi during powerup. The results shown for maximum stress 1n. v n:ratory strain, due to vibration, was-about 3000 psi peak to peak of stress. SCS concluded that the stresses- '
caused and measured in steady 1or vibratory strains would not exceed _ an endurance limit of 29000 psi for the support material, A-36 steel, and
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should not cause fatigue-cracking. SCS gave two possible explanations for the apparent contradiction to the previous Bechtel reports (fatigue caused by vibration): the maximum stress occurred higher than the current testing; and, the stresses that occurred at the crack areas were m':ch higher than the stresses measured just a few inches away. SCS-also made a conclusion of two possibilities for cracks, such as:
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The initial stresses in the ' restraints are much larger than calculated. However, this case seems unlikel There are much higher stresses. developed at the crack locations than are developed at the strain: gages. If this is the case, then a redesign of the restraints-.to' eliminate the stress concentrations is necessar Cracks occurred again on the same connection of support 2MS-R84 following repair and after the installation of the monitoring system to detect the
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movement and vibratio This connection was modified six times by Bechtel using a trial and error method from a normal rigid connection to an extra rigid connection, . back to a normal rigid connection, and twice to a pin
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p I After the hairline cracks were found again in September 1989, the licensee contracted this crack problem to an independent consulting company, SMC l O'Donnell Inc. for a study and resolution. Two study reports performed by l
O'Donnell Inc. were reviewed by the inspectors, report number 2066-400-001-00, "Three Dimensional Finite Element Analysis of Support 2MS-R84/85 - Farley Nuclear Plant, Unit 2", dated March 1990, and . report '
number.2066.01-400-001-00, " Evaluation of the Failed Gusset Plates Removed from and Recommended Design Modifications to the Farley Nuclear Plant, Unit 2 Main Steam Line Support 2MS-R84", dated July 199 . ...
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1 The first O'Donnell report used a finite element method and: ANSYS computer program to include the whole support in the analysis: plus a submodel and i assumed the repeated ~ failures were caused by high cycle fatigue .due to j vibration which was based on Bechtel reports and the: examinations on the:
failure materials. The most probable source of the cycle _ loading is flow induced vibration from the main steam line. The input of displacements due to vibration, 3.9 mils (or 0.0039 inches) at' the lateral direction, i'
mils at the vertical direction, and 5.7 mils at the axial direction were the values shown in the SCS report stated previously which were based on-the vibratory monitoring ~ measurement. The model actually used an _1/4" ;
fillet weld connection with a gap between gusset plates and the reinforced- '
cover plates on 8 x 8 inch tube stee The maximum unintensified stress:
was 3071 psi and occurred in the. throat of the 1/4'? fillet weld, near the center of the gusset plates. The ASME Code, Appendix XIV states.that the I evaluation of fillet welds for cyclic loading shall. include a' fatNue '
strength reduction factor: of fou Therefore, the: maximum- altsrnating ,
stress intensity (intensified). at. the fillet weld was!3071 psi x 4 = 12284 psi. From the ASME Code design fatigue' curve, the- endurance limit for -
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carbon steel (faiged materials) with y ' ultimate' strength = 70000 psi is- l 12500 psi at-1x10 cycles. ~This.1x10 cycles.can be reached in less than ?
-one day assuming peak vibration amplitudes occur at an average-of 10 Hz which is based on the SCS repor The other research data indicated that, the fatigue curve can be decrgased and the fatigue strength can be dt. creased to 11400 psi at 1x10 cycles, which is obtained at about four months of full power operation. :In this = case, cracks in the weld.were obser ad in as little as.six months. Therefore, the 0.'Donnell Inc. report concluded that the high cyclic stresses generated? from' the measured vibrations are clearly capable of being the cause of.the observed fatigue failures at support 2MS-R84 O'Donnell Inc. recommended an inexpensive method of changing 'the 1/4" fillet weld into a full penetration weld'for this connection and downgraded the other two proposals of eliminating' thel source of the vibration and redesigning this support. Based on the assumption of anfull penetration weld connection, the above.'1/4" fillet weld model wasLrerun using an integrated connection (a coupled technique) due to .the full penetration weld instead of a gap connection-(an uncoupled technique) due to the fillet weld between the gusset plates and the reinforced cover plates on 8 x 8 inch steel tube. The results showed that the-maximumTalternating' stress intensity for the unintensified, full < penetration weld, still remained at the throat of the 1/4" fillet weld as 3060 psi which was' only 10 psi less I than the previous model using only a 1/4" fillet weld. The ASME Code does not suggest a value for the fatigue. strength reduction factor"for a full l penetration weld. .However, the major benefit of the full penetration weld i is the improvement in the intensified stress of 3060Lx 1.5 = 4590~ psi (a j stress reduction factor of'1.5 is a conservative assumption ) compared with- I the intensified stress of 12284 psi for the fillet weldg which exceeded the '
predicted fatigue strength of 10400 psi at '1x10.0 ' cycles which is the assumed vibration accumulated for 30 years in full operation. However, if the actual vibration amplitudes are approximately twice the- SCS values, a fatigue failure of the proposed full penetration weld could occu . _
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The second O'Donnell report evaluated thi failed gusset plates an recommendeo that the 1/4 fillet welds between the l/4": thick gusset plates
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and the 1/2" thick tube steel cover plates be changed to-single-bevel-groove welds (full penetration welds) while continuing the vibration testing system for a period of time to obtain new data and determine if th vibration is larger than_ the previously. recorded vibration dat The-licensee agreed with the recommendation and issued a PCN No. B-90-2-6581 to be implemented during this refueling outag The inspectors reviewed the stress calculations, analysis for main steam line C for the before = and after crack data of the welds and members of supports 2MS-R84 and 2MS-R85. The stress calculation for the after_ crack case was analyzed without supports. 2MS-R84 and .2MS-R85 being active. :The ordinary support calculations for supportsL2MS-R84 and R85 were reviewed to see if they had an adequate design. All of the support calculations ,
in line C for operability concern were also Lreviewed._. These support 1 calculations used the new loads from stress reanalysis without the presence of the supports 2MS-R84 and R85. The purpose of1 the operability concern
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evaluation was to demonstrate that the system is still safe:to operate t
without these two supports and that all actual . stresses on piping and cupports _ without those two supports are. acceptable when compared to the allowable stresses. All stress and. support _ calculations reviewed wera considered to be adequate and acceptabl The inspectors consider the arrangement of the axial restraint for_ the 110'
horizontal line between the two anchors (penetration and steam generator nozzle) to be strange and unusual and a possible' problem since no actual rigid restraint is in this line except two snubbers at support 2MS-R85 The snubbers, in theory, are assumed to resist dynamic loads such as steam hammer or earthquake due to their huge-loads capacity and quick actions, in reality, the steam hammer due to the pipe run length between valves or
, the steam flow may not act quickly enough to activate the snubbers and resist the loads but these conditions can still produce big loads and-damage the connectio The maximum anticipated steam hammer load is 28 1 kips in the axial direction acting on support 2MS-R85 which,-in turn, will l l be resisted by support 2MS-R84 which was not designed:to resist the axial l l
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load Therefore, a load of less than 28 kips acting._on support 2MS-R84 may create a problem at this connectio SCS concluded that there may be an unforseen huge load acting on the cracked support, 2MS-R84 i The O'Donnell report revealed that there was only a 10. psi difference in '
stress -between a model using a 1/4" fillet weld with a gap (uncoupled) and !
a model using a full penetration weld without a gap as an integrated 'l connection (coupled) plus a 1/4" fillet wel Both maximum stresses l l occurred at the 1/4" fillet weld.which means the full penetration weld may l not work as expecte Table 9.3.1 Al_lowable Stresses in Welds and Table ,
L 9.4 Fatigue Stress Provisions and Figures 9.4A, 9.4B, and 9.4C Design l i Stress Range Curves of ANSI /AWS Dl.1-88, Structural. Welding Code did not
, have a stress allowable reduction between fillet and full penetration welds for the fatigue stresses which were adopted from the Standard
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Specifications for Highway Bridges which deals with a lot of vibration Therefore, the inspectors have doubts about the cracks being' caused purely due to vibration without other forces being involve O'Donnell and-the licensee-selected the least expensive method to fix~this-problem. However, the licensee agreed that if this modification does not work and cracks occur-again, they willinot- try to modify this connection again, will rerun the stress calculation, will relocate this support,~and take any other steps necessary to solve the problem completel This -item remains open pending future examination for cracks during the-next refueling outage following the modificatio Exit Interview The inspection scope and results were summarized on November.2, 1990, with those persons indicated in paragraph 1. The inspectors described _the areas inspected and discussed in detail the inspection results. The item listed--
below was discussed with the licensee during. an NRC - exit interview conducted on November 9,1990. Although reviewed during this inspection,
. proprietary information is not contained in this report.; : Dissenting-comments were not received from the licensee.
l One new inspector followup item was identified. IFI 50-348, 364/90-31-01, Possible Design Capacity Problem For Wej-It Anchor Bolts, paragraph:2.(4) '
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i 5. Acronyms-and Initialisms '
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ANSI - American National Standard Institut ASME - American Society of Mechanical: Engineers-AWS - American Welding Society-B&PV - Boiler and Pressure Vessel DAC - Distance Amplitude Curve EC - Eddy Current Hz -
Hertz kips - Kilo Pounds ISI - Inservice Inspection <
MT - Magnetic particle MHz -
Megahertz NDE -
Nondestructive Examination N Number NRC - Nuclear Regulatory Commission PCN -
. Production Change Notice psi -
Pounds Per Square Inch PT - Liquid penetrant [
QA -
Quality Assurance R -
Revision RHR' - Residual Heat Removal SCS -
Southern Company Services' '
SG -
Steam Generator i SIS -
Safety Injection System UT -
Ultrasonic VT -
Visual W -
Westinghouse Electric Corporation-i j
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