IR 05000424/1987047
| ML20237G901 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 08/25/1987 |
| From: | Blake J, Hallstrom G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20237G889 | List: |
| References | |
| 50-424-87-47, 50-425-87-34, NUDOCS 8709030023 | |
| Download: ML20237G901 (16) | |
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o UNITED STATES g
NUCLEAR REGULATORY COMMISSION o-
$
E REGION ll 101 MARIETTA ST., N.W. SUITE 3100 o
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ATLANTA, GEORGI A 30303 g*****
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Report Nos.:
50-424/87-47 and 50-425/87-34 Licensee: Georgia Power Company P. O. Box 4545 Atlanta, GA 30302.
Docket Nos.:
50-424 and 50-425 License Nos.:
NPF-61 and CPPR-109 Facility Name:
Vogtle 1 and 2 Inspection Conducte - August 3-6, 1987 Inspector:
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Date sig ed j
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8//) U Approved by:
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J.J./Take, Chief Da'te Signed E gi 6ering Branch iv sion of Reactor Safety j
SUMMARY Scope:
Ti11 s routine, unannounced inspection was in the areas of Inspector l
Followup Items (IFIs) (Units 1 & 2), Presservice Inspection. Summary Report (Unit 1) Housekeeping and Materials Control (Unit-2), Criteria for Arbitrary Intermediate Pipe Breaks on High Energy lines (Unit 2) and Welding and Non-welding Activities Associated With Reactor Vessel Internals (Unit 2).
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Results:
No violations or deviations were identified.
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8709030023B{$$$424 PDR ADOCK ppn O
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REPORT DETAILS i
1.
Persons Contacted Licensee Employees j
- P. D. Rice, Vice President and Vogtle Project Director R. H. Pinson, Vice. President, Construction
- C. W. Hayes, Project Quality Assurance (QA) Mi. nager
- E. D. Groover, QA Site Manager, Construction j
- G. A. McCarley, Project Compliance Coordinator
- D. M. Fiquett, Unit II Field Construction Manager
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J. E. Sanders, Assistant Project Construction Manager, Unit II
- L. B. Glenn, Manager of Quality Control
- W. R. Bainbridge, Engineering Supervisor, Electrical Field Operations
(EFO)
- C C. Matson, Cable Pulling Supervfsor, EF0
- R.
Hollands, Electrical Construction Supervisor
- A. W, Harrelson, Electrical Discipline Manager W. C. Gabbard, Senior Regulatory Specialist, Operations J. A. Caudill, Senior Plant Engineer, Operations Other licensee employeer contacted included construction craftsmen, engineers, technicians, mechanics, and office personnel.
Other Organizations
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- W. C. Ramsey, Southern Company Services (SCS), Project Engineering Manager
- C. D. Markham, Westinghouse, Project Manager D. D. Wieland, Westinghouse, Site Manager B. Reed, Westinghouse, Project Welding Engineer T. E. Richardson, Bechtel Power Corporation (BPC), Project Engineering Manager D. Capito, BPC, Plant Design Group Supervisor J. Hawley, BPC, Project Engineer - Plant Design Mechanical
- D. E. Strohman, BPC, Project Quality Assurance Engineer NRC Resident Inspectors H. Livermore, Senior Resident Inspector (Construction)
J. Rogge, Senior Resident Inspector (Operations)
R. Schepens, Resident Inspector (Operations)
C. Burger, Resident Inspector (Operations)
- Attended exit interview
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'2.
Exit Interview i
The inspection scope and findings were summarized on August 6, 1987, with
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those persons indicated in paragraph 1 above. The inspector described the areas inspected and discussed in detail the inspection. findings.
No dissenting comments were received from the licensee.
The licensee did identify as proprietary some of the materials provided to and reviewed by the inspector during this inspection; however, details from those materials are not included in this report.
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3.
Licensee Action on Previous Enforcement Matters Licensee Actions on previous enforcement matters within the scope of.this
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inspection were not complete.
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Unresolved Itenis Unresolved items were not identified during this inspection.
5.
Independent Inspection. Effort Housekeeping (548348), Material Identification and Control (429028), and
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Material Control (429408)-
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The inspector conducted a general inspection on Unit 2 containment to -
I observe activities such as housekeeping, material identification and i
control; material control, and storage.-
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Within the areas examined, no violations or deviations were identified.
6.
Reactor h ssel Internals - QA, Work Observation and Records Review (50051, 50053 and 50055) (Unit 2)
The inspector examined nonwelding activities associated with Unit 2 Reactor Vessel internals to determine whether implementing procedures were appropriate and adequate to assure that'these activities are controlled and performed according to NRC requirements and SAR commitments and whether quality records conformed with established procedures and
. reflected work accomplishment consistent with requirements.
The i
applicable code for Unit 2 Reactor Vessel internals'is the ASME Boiler and Pressure Vessel Code,Section III, 1977 edition with ' addenda through i
.Winter 77.
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a.
QA Review (50051)
The below listed procedures were reviewed to determine whether procedures are appropriate and adequate to ensure control of the following specific activities.
Receipt Inspection and Handling
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Receipt inspections verify that the reactor vessel,
associated components, and internals are undamaged and are in conformance with specifications, including any special protection requirements.
Handling and storage activities are in accordance with
established procedures, i
Record-keeping requirements are met.
- Installation records reflect the actual as-installed conditions.
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RPV and Internals Installation Inspection Inspection activitics cover pertinent installation
activities (such as vessel and internals placement, leveling, and final adjustment) to assure that applicable specifications and work procedures are accomplished as specified.
Inspection activities meet established procedures including
record-keeping requirements and qualifications of inspection
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personnel responsible for RPV and internals installation i
inspection, j
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Post-Installation Activities Vessel and internals protection procedures and internal
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cleanliness preservation procedures are established prior to need.
Procedures established above are adhered to.
- Document No.
Title ES-4028-Vogtle-2-7, Rev. A Upper and Lower Internals Storage Stand Assembly ES-4028-Vogtle-2-17, Rev. A Reactor Vessel Internals Assembly ES-63, Rev. C Procurement and Material Control a
ES-67, Rev. C Cleanliness Requirements and Control
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ES-100-2, Rev E Liquid Penetrant Examination ES-110, Rev. A Stop Work Procedure
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ES-116-1, Rev. L Qualification and Certification on Non-destructive examination
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personnel.
Written Practice for Corporate and all field sites ES-116-2, Rev. I Qualification and Certification of <!nspection Personnel to ANSI N45.2.6, Written Practice for Corporate and all field sites
ES-116-3, Rev. C Qualification and Certification of Auditing Personnel l
ES-126, Rev, C Quality A'ssurance Records and Retention ES-142, Rev G Deviations and Corrective Actions ES-145, Rev. D i'
Change Notices and Change Requests l
ES-146, Rev. E Document Control I
ES-174, Rev. C Process Control b.
Observation of Work (50053)
The below listed activities associated with the Reactor vessel internals storage, protection. preservation, handling, installation and inspection were observed to determine whether requirements of applicable specifications, work and inspectiop (QC) procedures are being met in the following areas:
Protection of Stored Vessel Internals Internals storage
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Protection / protective coverings
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Storage supports Installation Techniques
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Lifting and handling are consiatent with established requirements and precautions.
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Lifting equipment is as specified and required testing has been completed prior to lifting.
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Placement is being (or hr been) accomplished in i
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conformance with requir..ents.
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Installation is being (or has been) accomplished in conformance with requirements, including special requirements and precautions.
The inspector observed upper and lower internals storage / protective activities and rigging / lifting installation techniques associated with adjustment of lower internals storage stand guide brackets to
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the required gap in conformance to process control sheet 2-185-54.
During this inspection the inspector observed gaouged areas on the bottom machined face of the lower internals core plate at approximately 41" from the 240 support lug.
Subsequent investigation by cognizant licensee personnel determined that the damage had not been previously identified and DR-NI-270 was issued to accomplish necessary rework (surface blending) and inspection (dye penetrant examination).
The inspector observed completion of rework and inspection in conformance to DR-NI-270 and noted that no significant damage had occurred.
Subsequent discussions with licensee personnel and review of Westinghouse programmatic controle (as clarified by GAEW-87-1701) established that several additional visual inspections intenced to identify any damage detrimental to the internals design and operation were scceduled.
The inspector concluded that programn stic controls were sufficient to resolve any potential NRC
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Records Review (50055)
The below listed quality records associated with the Unit 2 lower internals and lower internals starage stand assembly were examined to ensure that the records reflected work accomplishment consistent with requirements on the following areas:
Receipt inspection and material certification
Storage and installation inspe.:tions Nonconformance/ Deviation Records Qualification Records of Craft, QA and Inspection (QC) personnel
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Process Control Sheets (PCS)
Issue No.
Date Scope 2-185-2 2/1/85 Temporary Assembly of Lower Internals Storage Stand 2-185-9 6/8/85 Installation of Lowr e Internals Storage Stand at Permanent Location
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2-185-10 6/8/85 Installation of Lower Internals Storage Stand in the Temporary J
Positions i
2-185-54 4/26/87 Installation of the Lower Internals i
Storage Stand into the Permanent i
Position 2-105-12 3/26/86 Installation of the Cruciform Instrumentation, Butt Instrumentation Offset Instrumentation and Secondary-Support Columns, Upper and Lower Tie Plates, Energy Absorbers and Secondary Core Support Base Plate 2-105-12-1 4/16/86 Machining of Offset Instrument Guide Shim 2-105-12-2 4/16/86 Assembly and Weld Energy Absorber Guide Post 2-105-12-3 2/12/87 Rework Energy Absorbers to DR-NI-238 2-105-12-4 2/13/87 Rework Lower Internals Per Interim Disposition of DR-NI 241 2-105-12-5 2/24/87 Rework Lower Internals Per Final Disposition of DR-NI-241 2-105-12-6 3/3/87 Rework Lower Internals Per Disposition
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of DR-NI-244 2-105-12-7 3/4/87 Rework Lower Internals Per Disposition of DR-NI-245 l
2-105-12-8 3/9/87 Rework Lower Internals Per Disposition of DR-NI-244, Rev. 1
2-105-12-13 4/10/87 Obtain Dimensions for Machining the Clevis Inserts and Energy Absorbers l
2-105-12-15 4/10/87 Machining of the Guide Posts
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Nonconformance/ Deviation Reports (NCRs/DRs)
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Closure No.
Date i
DR-NI-221 2/5/87 DR-NI-238 2/12/87 DR-NI-241 3/13/87 DR-NI-242 2/27/87 i
DR-NI-244 3/11/87 DR-NI-245 3/11/87
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DR-NI-270 8/5/87
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Field Change Requests (FCRs)
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Date
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l M-FCRB-16920 4/8/87 l
M-FCRB-16480 1/2/87 i
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Receiving Inspection Reports (RIRs)
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Inspection
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Date(s)
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2021 5/13/86 & 2/4/87
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2C24 2/2/87 & 2/16/87 2C18 3/11/86 2C13 5/22/86 2C14 4/2/87 2C19 5/15/87
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CMTRs/C0Cs Class / Type Size Heat / Lot No.
ER308 Weld Rod 3/32" x 36" 05394 ER308 Weld Rod 0.035" J09513 SA36 Plate 1/2" Thick 85386 SA36 Plate 1-1/4" Thick 85326 SA36 Plate 3/16" Thick D8891/39F NDE Materials Liquid Penetrant Type SKL-HF/S 84H027 Cleaner / Remover Type SKC-S 86B051 Developer Type SKD-NF/2P9B 84J005
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l Qualification Records
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Welder Stencil No.
Qualification WPs N-66 80.2.3 l
N-80 80.2.3 QC Inspector ASNT Qualifications GRG PT-III, RT-III, UT-III, MT-III, VT-II Within the areas inspected, no violations or deviations were identified.
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Reactor Vessel Internals - Welding (Unit 2) (55050)
The inspector examined welding activities pertaining to field welding of Unit 2 reactor vessel internals to determine whether applicable code and procedure requirements were being met. The applicable code for Unit 2 reactor vessel internals is the ASME Boiler and Pressure Vessel Code,Section III, 1977 Edition with Addenda through Winter 77.
a.
Visual Examination of Welds The below listed welds were examined relative to the following:
location, length, size and shape; weld surface finish and appearance; transitions between different wall thickness; weld reinforcement --
height and appearance; joint configurations on permanent attachments and structural supports; removal of temporary attachments, arc strikes and weld spatter: finish grinding or machining of weld surface, surface finish and absence of wall thinning; surface defects, cracks, laps, lack of penetration, lack of fusion, porosity, slag, oxide film and undercut exceeding prescribed limits.
(1) Cruciform Instrumentation Welds Do./ ell Pin to Nut and nut locking cap to cruciform (4 welds) at lower core plate and secondary core plate (2 elevations)
Core Location Cruciform ID No.
Weld No.
B-3 76615 AAW B3 B-6 76566 AAW B6 B-8 76568 AAW B8 C-7 76572 AAW C7 C-8 76602 AAW C8 E-11 76548 AAW E11 F-14 76565 AAW F14 G-12 76593 AAW G12
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H-11 76595 AAW H11 H-13 76603 AAW H13 i
H-15 76607 AAW H15 l
M-7 76592 AAW M7 i
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N-6 76599 AAW N6 N-8 76604 AAW N8 R-6 76606 AAW R6 R-8 76608 AAW R8 (2) Energ3 Absorber to Base Plate Welds Core Location Weld No.
H-10 MMW 0
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K-8 MMW 90 H-6 MMW 180 F-8 MMW 270 (3) The following inspector qualification status records and "QA/QC j
Inspector Qualification / Certification" records were reviewed l
relative to inspection of the weld joints listed above.
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Inspector ASNT Certification GRG PT-III, RT-III, UT-III,
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MT-III, VT-II
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(a) The following liquid penetrant materials certification records relative to inspection of the weld joints listed above were reviewed to ascertain if the sulfur and halogen content of the materials was within acceptable content limits.
Materials Type Batch / Lot No.
Liquid Penet ant SKL-HF/S 84H027 Cleaner / Remover SKC-S 85J058 Developer SKO-NF/2P9B 84J005 b.
Welding Procedure Specification and Quality Assurance Procedures i
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(1) Welding Procedure Specification (WPS) applicable to the weld joints listed in paragraphs 7.a.(1) and 7.a.(2) were selected for review and comparison with the ASME code as follows:
WPS Process PQR 8.2.3, Rev. D
- GTWA (manual)
205, 216
- GTAW - Gas Tungsten Arc Welding l
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The above WPSs and their supporting Procedure Qualification Records (PQRs) were reviewed to ascertain.whether essential, supplementary and/or nonessential variables, including thermal treatment, were consistent. with Code requirements; whether the WPS were properly qualified and their supporting PQRs were accurate and retrievable; whether all mechanical tests had been
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S performed and.the results met the minimum requirements; whether-the PQRs had been reviewed and certified. by appropriate personnel and 'whether any. revisions and/or changes to nonessential variables were noted.
WPSs are qualified in
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accordance with ASME section IX, 'the latest edition and addenda at the time of qualification.
(2) The below listed Nuclear'Insta11ation Services Company (NISCO)
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procedures were reviewed in addition to those reported in paragraph 6.a. to ascertain whether the reactor vessel internals welding program had been approved, by the licensee ' and: whether adequate plans and procedures had been established to assure
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that welding would be controlled. and accomplished consistent
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with commitments and regulatory requirements.
Document Number Title ES-300, Rev. F General Welding Procedure ES-100-5, Rev. E Visual Inspection of Welds l
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ES-56, Rev. F Welding Filler Material Control Procedure (Safety-Related)-
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Welder Qualification The inspector reviewed the NISCO's program for; qualification of i
welders and welding operators for compliance with QA procedures and i'
ASME Code requirements.
The following welder qualification status records and " Records of l
Performance Qualification Test" were reviewed relative to the weld joints listed in paragraphs 7.a.(1) and 7.a.(2),
l Weld Symbol WPS
'N-66 80.2.3 N-80 80.2.3
- Welder qualification tests were conducted on a full scale mockup of productinn weld l i
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Welding Filler Material Control The inspector reviewed the NISCO's program for control of welding materials to determine whether materials were being purchased, accepted, stored, and handled in accordance with QA procedures and applicable code requirements.
The following specific areas were j
examined:
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Purchasing, receiving, storing, distribution, and handling procedures, material identification j
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The following welding. material applicable to the weld joints listed in paragraphs 7.a.(1) and 7.a.(2) were selected for review of purchasing ar.d receiving records for conformance with applicable procedures and code requirements:
Type Size Heat / Control No.
ER308L 0.035" J09513 ER308L 3/32" X 36" 05394 Within the areas examined, no violations or deviations were identified, l
8.
Previously Identified Inspector Followup Items a.
(Closed) Inspector Followup Item (424/86-18-07), Structural Sizing Calculations Review - Followup on the review of two calculations from member sizing along with transfer of data to drawings.
j This item was identified by I&E during the examination of Readiness Review Module 8, Structural Steel.
The inspector had been informed during a previous inspection (Inspector Report Nos. 424/86-93 ana 425/86-43) that additional GPC responses had resolved I&E concerns.
However, this item was left open pending final confirmation that all concerns were resolved.
The inspector reviewed additional information during this inspection and determined that all concerns had been resolved.
This item is considered closed.
b.
(Closed) Inspector Followup Item (425/87-10-01): Revision of Bechtel Specification X3AR01 Section 68-Raceway Systems Thi s item concerned need for improvement of upper-tier Bechtel Specification X3AR01 Section E8.
During earlier inspections, the inspector had noted sufficient outstanding FCRs/0CNs to cause undue complexity in the field and a lack of conformance to PRM requirements for revision of specifications / drawings (Section C-26). This problem had been previously identified by a licensee audit (CAR No. VS-87-209)
and completion of corrective action was not-complete.
This item was opened pending NRC review of CAR corrective actions and later
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revisions of Section E8. The majority of outstanding.FCRs/CSNCs were the result of scheduled implementation " lessons learned" from Unit 1.
The inspector was informed by cognizant' licensee personnel that -
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" lessons learned" were now implemented.
The inspector was.further
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informed of planned revisions to PRM - Section C-25 which would q
reduce the number of allowed FCRs/DCNs before requiring that a i
specification be revised.
The PRM revisions will include clarif1 cations intended to. improve:
scheduling efficiency in accomplishing revisions.
The inspector i
reviewed the l ate'st revision of Section E8 (Revision 26 issued
June 10, 1987) and the current out' standing FCRs/DCNs (FCR Nos. ~ 20, 104 & 8147; DCN Nos. 574 & 575) and concluded that NRC concerns were resolved.
This item is considered closed.
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Preservice Inspection Summary Report (Unit 1)
On June 25, 1987 GPC transmitted to NRC.a Preservice-Inspection Report (SL-2702) in lieu of the Form' NIS-1 Report specified by the ASME Code,.
Section. XI.
The submittal was based on code interpretation XI-1-83-15.
NRC had previously agreed that a report in lieu of, but consistent with, an NIS-1 be submitted.
The report included a summary of preservice inspection activities and noted that supporting information was available at the plant site for review.
The inspector reviewed previous inservice inspection activities by Region II (Report Nos. 424/85-07, 85-18, 85-25, 85-46,86-101, 86-129) and selected a random sample of items (UNR 86-129-01, IFI 86-46-01, UNR 85-02-02) which were closed based on. proposed changes to later revision (deviations) of the specifications involved.
Current revisions of the specifications were reviewed to verify that anticipated changes had been incorporated. The inspector also reviewed supporting information for randomly selected details in the summary report as follows.
Mechanized ultrasonic examinations of reactor vessel interior by.
Combustion Engineering.
100*s eddy current examination of steam generator tubing by Zetec, l
Inc.
i Augmented inspections (100% volumetric) of large diameter main steam J
and feedwater system welds between the containment penetration' and the first rigid restraint'outside containment.
Class 1 & Class 2 component plan and program documents ISI-P-001 and ISI-P-002.
Use of a dual-element refracted-longitudinal wave 1.0 MHZ transducer i
and an NRC approved technique for the ultrasonic examination ~of cast l
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stainless steel reactor coolant loop piping.
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Eval'uation and either resolution and acceptance or repair of components with reportable indications.
No discrepancies were identified during the above inspection and the inspector. concluded that an. independent verification supported.the accuracy of the Preservice Inspection Summary Report as transmitted to NRC. The inspector informed cognizant licensee personnel that Region II would recommend approval of~the report as transmitted.
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Criteria for Arbitrary Intermediate Pipe Breaks on High Energy Lines Inspector Followup Item 424/86-39-02, 425/86-19-02 concerned proper application of a minimum distance criteria (50 Criteria) from welded attachments for the postulation of intermediate pipe breaks on high-energy ASME Code class piping. The inspector's previous inspections in this area had identified need for field examinations to assure the as-built support locations in order to conduct the engineering analyses involved.
The inspector had previously noted that support location tolerances for small diameter piping amounted to as much as three piping diameters.
Therefore, a 50 minimum distance criteria (or some alternate criteria requiring greater precision) could not be assured without' field examinations.
Agreements reached between GPC and the NRC Office of Nuclear Reactor Regulation (letters, dated August 4, 1986 -. log GN.- 1022 and September 30, 1986 - Log GN-1090), regarding arbitrary intermediate pipe break criteria-includes a minimum distance less than 50 for all high energy class 2 and 3 systems other than main steam and main feedwater systems.
The proposed criteria (34T) requires greater precision regarding pipe support locations than does 50.
The inspector had previously noted that the method of implementation will be the use of as-built piping support and piping isometric drawings to determine the distance between the edge of the welded attachment and the edge of the break location. The inspector further noted that PPP procedure X-24 (As-built Piping Systems and Related j
Components) did not require that exact distances be recorded unless support locations were outside the tolerances allowed by PPP Support Installation Procedure IX-50. Therefore, "as-built'! drawings ' reviewed to date for imposing the subject criteria were suspect unless exact field measurements had been recorded.
This item was previously closed (Report No. 424/87-05) based -on information provided by Piant Systems Support Group (PSSG) memo No. 34 -
Group No.1, dated October 20, 1986.
PSSG No. 34 provides. detailed instructions to BPC engineers on the evaluation of welded attachments in j
the vicinity of high stress locations; i.e. application of the approved i
arbitrary intermediate pipe break criteria. PSSG No. 34 provides several
levels of screening criteria to ensure conservatism in establishing the j
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i distance from high stress point (potential arbitrary intermediate break)
to nearest welded attachment. High stress points were to be selected for review (screened) versus the arbitrary intermediate break criteria as follows:
All high stress points at 100 from centerline of welded attachment All high stress points at SD (or 3/RT) from edge of welded attachment using worst case installation tolerance Field measurement by BPC engineers Cognizant Licensee personnel informed the inspector, during this inspection, that " lessons learned" from Unit 1 were being applied to Unit 2 in that PPP Procedure X-24 had been revised and was able to provide dimensional information accurate enough to apply AIP8 criteria. A review of the latest revision of X-24 (dated 2/13/87) determined that tolerances involved (i 1") would not provide accuracy sufficient to apply 3 /RT l
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Cognizant licensee personnel provided proposed changes to Design Criteria 1018 and PSSG memo 34 to clarify that either field measurements to the accuracy required or worst case application of tolerances would apply.
The inspector informed cognizant licensee personnel that these changes would be sufficient to prevent further NRC I
concern.
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s USNRC-DS F181 SEP -2 A 9 57
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