IR 05000445/1981009
| ML20010A405 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 07/21/1981 |
| From: | Crossman W, Renee Taylor NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20010A402 | List: |
| References | |
| 50-445-81-09, 50-445-81-9, 50-446-81-09, 50-446-81-9, NUDOCS 8108110397 | |
| Download: ML20010A405 (8) | |
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U. S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT
REGION IV
Report No.
50-445/81-09; 50-446/81-09 e
Docket No.
50-445; 50-446 Category A2 Licensee:
Texas Utilities Generating Company 2001 Bryan Tower Dallas, Texas 75201 Facility Na;ae:
Comanche Peak, Units 1 and 2 Inspection at:
Comanche Peak Steam Electric Station Inspection conducted:
May and June 1981 7/2//8/
Inspector:
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R. G. Taylor, Residunt Reactor Inspector Date Reactor Project Section 3 7[2//8/
Approved:
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W. A. Crossman, Chief Date Reactor Project Section 3 Inspection Summary:
Inspection during May and June 1981 (Report 50-445/81-09; 50-446/81-09)
Areas Inspected:
Routine, announced inspection by the Resident Reactor Inspector (P.RI) including general site tours; followup on previously identified inspection findings; followup on licensee identified construction c'eficiencies; protection of major installed equipment; installation and welding of safety-related piping; and protection of electrical and instrument cables.
The inspection involved 154 inspector-hours by the RRI.
Results:
No violations or deviations were identified.
8108110397 810723 PDR ADOCK 05000445
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DETAILS 1.
Persons Contacted Prinicipal Licensee Employees
- D. N. Chapman, TUGCO, Quality Assurance "6uoger
- R. G. Tolson, TUGCO, Site Quality Assurance Supervisor
- J. R. Merritt, TUSI, Engineering and Construction Manager Other Persons J. V. Hawkins, Brown & Root, Project Quality Assurance Managar Th9 RRI also interviewed other licensee and Brown & Root employees during the inspection period.
- Denotes those persons who attended one or more management meetings with the RRI.
2.
Licensee Action On Previous Inspection Findings (Closed) Unresolved Item (50-445/80-27; 50-446/80-27) Qualification of Okonite Electrical Cable.
The RRI reviewed instrument chart recordings and test engineering log books which provide the necessary supportive detail that the Okonite Company had properly qualified the electrical cable supplied to this site.
The data reviewed was of a nature that would have been very difficult to reproduce as site specific since the tests were performed as type tests as allowed under the IEEE Standards.
The RRI had no further questions on this matter.
(Closed) Unresolved Item (50-445/80-20) Storage of Fuel Storage Racks.
The licensee covered the fuel storage racks almost immeaiately after the above inspection report was issued and has maintained the coverit.g to the current time.
The racks have also now been labeled as "Q" items and included within the QA surveillance of storage conditions.
An Operations Traveler has been issued, although not implemented, which requires that the lig',t rust in some areas of the racks be removed by wire brushing and
acetone prior to installation in the Fuel Building.
The cleaning is considered a routine operation of the installation sequence and will be observed at an appropriate time.
The RRI had no further questions on this matter.
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3.
Action On Licensee Identified Design / Construction Deficiencies (Closed) NSSS Design. Westinghouse reported to the NRC under 10 CFR 21 that it had discovered that it was possible to damage one of three high head charging pumps under a postulated single failure of the instrument system for controlling the level in the volume control tank. Westinghouse further stated that this would occur only if the plant operator did not intercede in the series of events by taking manual control. Westinghouse recommended that each plant review their operating procedures in conjunc-tion with their plant specific design as a preventative measure.
The plant operations staff has conducted such a review and has determined that the level instrumentation provided within the CPSES design has the necessary redundance to be essentially single failure proof; the failure of the instruments would be alarmed in the control room; and that exist-ing procedures provide the operator with all necessary instructions to properly respond to the alarm.
On this basis, the licensee has concluded that the matter is not reportable under 10 CFR 50.55(e).
The RRI had no further questions on this matter.
(Closed) Design of 6.9 KV Breaker.
On April 8, 1981, the licensee informed the RRI that a possible design problem with a 6.9 KV circuit breaker had been identified in preliminary startup testing.
By letten
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dated May 4, 1981, the licensee stated that the previously identified problem was not considered reportable under 10 CFR 50.55(e).
The RRI has reviewed data which indicates that breaker performed properly after the linkzge system was cleaned and lubricated.
The licensee also tested a group of comparable breakers in other circuits and all performed properly, which affirms that the original report of a possible design problem was incorrect.
The RRI had no further questions on this matter.
(Closed) Design of HVAC Ductwork and Fan Supports.
On March 24, 1981, the licensee informed the RRI that an audit of Bahnson Co. (HVAC Con-tractor) had discovered that the design drawings did not stipulate the embedded length of concrete anchor bolts used to secure HVAC struc-tures to the buildings.
On April 12, 1981, the licensee filed a formal report on the matter under the requirements of 10 CFR 50.55(e). The RRI has monitored the licensee's QA/QC program to verify the embedment lengths by ultrasonic measurements of these bolts and has further verified that the Bahnson procedures both for design and quality control have been clarified to prevent recurrence of this matter.
The RRI will follow the work of replacing anchor bolts identified as not having suf-
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ficient strength k r the application as a routine inspection item.
The RRI has no further questions on this matter at this time.
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(Closed) Design of the HVAC System Serving the Vital AC Converters and DC Battery Charger Rooms.
On March 30, 1981, the licensee informed the RRI that their A/E had discovered that the existing design for the HVAC-would not supply the necessary cooling to maintain the room spaces housing the vital AC converters and the DC chargers at temperatures not exceeding the temperatures to which the units had been qualified under certain postulated failures.
By letter dated April 24, 1981, the licensee filed a formal report under 10 CFR 50.55(e) indicating that a new adequately sized HVAC system would be designed and installed for these spaces.
The licensee further evaluated the overall design QA program to ascertain if the above described situation was the result of a program breakdown.
By letter dated June 23, 1981, the licensee stated that his examination indicated that the inappropriate design was the result of a slow evolution wherein equipment loads were increased with the resultant higher heat loads being imposed that finally exceeded those to which the equipment had been qualified.
The RRI will follow the installation of the redesigned HVAC systm for these spaces as a matter of routine inspectson and has no further quotions at this time.
(Closed) Vendor Manufacturing Errce.
Transamerica Delaval corp. notified the NRC by letter dated September 19, 1980, under the requirements of 10 CFR 21, that a possible error in the machining of certain linkage rods in the Emergency Diesel Generator engines furnished to CPSES for Units 1 and 2 had been discovered.
The linkage assemblies for three of the engines were inspected by the licensee and a representive of the vendor during December 1980 while the fourth engine was inspected during March 1981.
In each case the linkage coun+erbore was found to be one-sixteenth
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inch undersize in relation to the vendor's specified depth.
By letters dated December 30, 1980 and April 21, 1981, the vendor stated that undersized counterbore found in the four engines is acceptable and that no replacements are necessary.
Based on the vendor's evaluation of the site inspection results, the RRI has no further questions on the matter.
(Closed) Plasite 7122 Coating In Service Water System Piping.
On September 18, 1980, the licensee informed the RRI that a condition had been identified that might be reportable under 10 CFR 50.55(e) wherein the Plasite coating within the carbon steel portion of the service water system had been found to be coming loose on a flange surface.
This finding was construed to indicate that the lining might not be adhering in other areas of the pipe.
By letter dated October 29, 1980, the licensee informed the NRC that the matter was not considered reportable and data was on file to substantiate the position.
The RRI reviewed the data that the conclusion was, in part, bes M on the fact that the
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coating of the pipe was not considered to be safety-related although the pipe system was so considered.
FSAR Section 9.2 appeared to be taking credit in the safety ant. lysis for the coating although Table 17A failed to list the coating specifically as either safety or non-safety.
In Amendment 21 to the FSAR the licensee specifically stated that the Plasite coating was specified by the Engineer for economic reasons, i.e., prevent long term deterisration of the pipe, and that the presence of the coating was not considered from a safety standpoint.
The RRI had no further questions on this matter.
(Closed) Piping Minimum Wall.
The licensee informed the NRC on May 22, 1979, that a condition had been identified wherein the wall thickness of piping had in some instances been found to be less than the wall thickness allowed by the ASME Code.
On June 3, 1981, the licensee filed a formal report unde" 50.55(e) discussing the problem, the safety consequeaces, and the corrective actions being taken.
During the period from the initial to final notification, the RRI had been following the licensee's program for identifying necifically those piping areas whe.'e the wall thickness was below mini. a allowable thickness.
This process was time consuming due to the n s 7 ness of locations and was
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accomplished where possible bv physical measurement.
In other instances, the measurements were accomplished by calibrated ultrasonic thickness measurement devices. Where these measurements were found below the-Code allowed values, the specific values were submitted to the Engineer for review.
Generally, those pipe areas found to be more than.010 inches less than the minimum wall have been or are being repaired by welding.
Those pipe areas between the minimum wall and.010 inches less than minimum wall have generally been determined by the Engineer, using standard stress analysis techniques, as acceptable without repair.
By observatior: and review of documentation, the RRI is satisfied that the licensee has taken appropriate action in this matter.
The RRI will continue to follow the remaining repair work under the routine inspec-tion program relative to piping systems.
4.
Site Tours The RRI toured all of the safety-related plant areas at least twice during the inspection period.
Additional tours were made of those areas where construction activities are the most concentrated including the Unit 1 Reactor Building, the Unit 1 Safeguards Building, and the common Auxiliary Building.
One of the tours was made with an ACRS sub-committee during their site visit.
The tours were oriented to
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observe the status of construction and the contruction practices being utilized by the various crafts.
The RRI considered the general house-keeping in each area of potential effect on the quality of the workman-ship.
In addition, the RRI observed the overall activities of the licensee's QC force as indicated by their presence in the various work areas.
The RRI noted that the craft personnel were apparently knowledge-able of the requirements established by site procedures; the housekeeping was adequate and in some areas very good; and the QC force was in most instances available and observing assigned activities.
No violations or deviations were identified.
5.
Protection of Major Installed Equipment The RRI observed that the Unit 1 Reactor Vessel internals remain partially installed within the Vessel.
The Reactor Vessel Head with the control rod drive mechanisms and cooling shroud were also installed and adequately protected with a wood cover and with plastic.
The uninstalled portion of the Unit 1 Internals remained within the temporary wood enclosure as they have for a sustained period.
The Unit 2 Reactor Vessel remained covered with a substantial steel deck protecting the top flange surface and the inside of the vessel.
The RRI observed portions of the movement of the Unit 2 Reactor Internals from the outside temporary enclosure to their storage positions within the Unit 2 Reactor Building.
The physical handling during the movement process was carefully managed as was the final positioning on the storage fixtures.
The RRI observed the final efforts of enclosing and protecting the two portions of the internals. -These efforts evidenced good planning and proper implementa-tion.
The RRI also observed that electric motors for pumps and valves throughout both Units 1 and 2 had their space heaters energized as evidenced by their pilot lights and their hand warm condition.
No violations or deviations were identified.
6.
Safety-Related Piping Installation and Welding The RRI observed the handling practices involved in moving the Unit 2 Reactor cross-over pipe for loop 1 into the Reactor Building and down into the steam generator cavity.
The RRI also observed a pre-installation liquid penetrant examination of the weld-end preparation areas for the same doing.
The examinations were performed by Welding Engineering i
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technicians for information purposes since the supplier had already performed similiar examinations as required by the ASME Code. The examination revealed only a few Code acceptable anomalies, primarily tooling marks left from the weld-end machining operations.
The RRI
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observed the subsequent welding operations involved in the installation
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of the cross-over piping.
Two of the three welds in the loop were being welded simultaneously utilizing two semi-automatic welding machines, each with two welding heads.
The welders observed were AHX, BBI, BUF, and AMS.
Each of these operators were verified to have been properly qualified in accordance with ASME Section IX and were known by the RRI to be fully competent to weld with the machines as based en various observations during the past three years. The weld wire heat number recorded on the Weld Filler Metal Logs and marked on the wire drums by tne supplier was 464176.
The RRI subsequently verified that the heat was fully qualified in accordance with ASME Sections II and III.
The weld procedure being utilized was 99028 and each machine control panel was noted to be set in accordance with the parameters established by the procedure. The welds observed during the above operations were identified as FW-4 and FW-6 on drawing RC-2-520.
The RRI observed welder ABT while making weld FW-4-2A as identified on isometric drawing CT-1-SB-015.
The weld was being made under weld
procedure 88021 using weld rod identified as heat 463638.
The RRI
verified that the welder had been qualified r., reauired by ASME Section IX and that the weld rod had been qualif*ed as required by ASME Sections II and III.
The weld procedure is a commonly used one that had been previously verified to have been properly qualified by the RRI and other NRC inspectors.
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The RRI also reviewed radicgraphs for the below listed welds during the inspection period for compliance with the requirements of site procedures and ASME Sections III and V.
Weld Number Isometric Number Line Number
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FW-14-4A SI-1-RB-037 10-RC-1-021-2501R1 FW-14A SI-1-RB-021 3-SI-1-339-2501R1 FW-1-1A SI-1-RB-059 10-SL-1-180-2501R1 FW-5 and FW-4-1 SI-1-RB-021 3-SI-1-303-2501R1 W-17A SI-1-RB-021 3-SI-1-303-2501R1 FW-3 SI-1-RB-038B 6-SI-1-092-2501R1 FW-9 SI-1-RB-060 10-RC-1-055-2501R1 FW-1A and FW-6-1 CS-1-RB-028 2-CS-1-112-2501R1
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W-41 and W-13 C-2-RB-072 6-RC-2-046-2501R1 FW-5-3, FW-5-2, FW-78 CS-1-RB-028 2-CS-1-112-2501R1 W-12, FW-13, FW-11A CS-1-RB-026 3-CS-1-019-2501R1 FW-1 SI-1-RB-059 10-SI-1-180-2501R1 FW-3A VA-1-RB-005 Ventilation Duct FW-2 RC-1-RB-026 14-RC-1-135-2501R1 FW-38-2A RC-1-RB-015 3-RC-1-146-2501R1 FW-11 and FW-19 RC-2-520 Steam Gen. to Cross-over FW-5-1 CS-1-RB-028 2-CS-1-112-2501R1 FW-148 and FW-15A SI-1-RB-021 3-SI-1-339-2501R1 No violations or deviations were identified.
7.
Electrical and Instrument Installations The RRI toured selected areas where extensive work had been recently accomplished involving welding or the attachment of heavy components above or adjacent to cable trays to ascertain if the installed cable had been damaged and the extent that debris had been allowed to collect in the trays.
The RRI also observed on going activities of a comparab!
nature to ascertain if the cables were being adequately protected during routine activities.
The RRI'found no evidence of damage either from welding operations or from heavy falling objects.
The trays in most areas had accumulated some amount of small debris such as small pieces of wire and some rags and hand-wipes frequently used by the craft personnel.
This type of debris does not appear to be harmful at this stage of construction since relatively few of the circuits are energized.
The various crafts performing welding over or adjacent to the cable trays were found to be covering the tray with flame resistant blankets prior to welding or cutting to keep the hot slag from damaging the cable.
No violations or deviations were identified.
8.
Management Interviews The RRI met with one or more of the persons identified in paragraph 1 on May 5, 8, 13, 19, and June 4, 15, 18, and 30, 1981, to discuss inspection findings and the licensee's actions and positions.
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