IR 05000445/1981016
| ML20033D284 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 11/16/1981 |
| From: | Crossman W, Renee Taylor NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20033D280 | List: |
| References | |
| 50-445-81-16, 50-446-81-16, NUDOCS 8112070488 | |
| Download: ML20033D284 (7) | |
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APPENDIX
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U. S. NUCLEAR REGULATORY COMMISSION
REGION IV
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Report:
50-445/81-16; 50-446/81-16 Dockets:
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50-445; 50-446 Category:
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Licensee:
Texas Utilities Generating Company 2001 Bryan Tower
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Dallas, Texas 75201 jfh'-
Facility Name: Comanche Peak, Units 1 and 2
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Inspection At:
Comanche Peak Steam Electric Station Inspection Conducted: October 1981
'4 Inspectors:
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. G. Tay
,~ Resident Reactor Inspector Dste/
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N//42W Approved By:
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- W. A. Cross
, Chief, Reactor Projects Section 3 Ddte'
Inspection Summary:
Inspection conducted during October 1981 (50-445/81-16; 50-446/81-16)
Areas Inspected:
Routine, announced inspection by the Resident Reactor Inspector (RRI) including general site tours, follow up on previously identified inspection findings; inspection of structure steel access platforms and building attachments; installation of safety-related piping systems; storage of major installed equip-ment; and installation of the liner _in the Unit 1 construction opening.
The inspection involved 71 inspector-hours by the RRI.
Results:
No violations or deviations were identified.
8112070488 811120 PDR ADOCK 05000445
.O PDR
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DETAILS 1.
Persoi;s Contacted Principal Licensee Employees
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- B. R. Clements, TUGCO, Vice President-Nuclear
- D. N. Chapman, TUGCO, Quality Assurance Manager
- 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 Manager The 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 (50-445/81-02; 50-446/81-02):
Engineered Pepth of
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Concrete Anchor Bolts.
This item concern-the depth of anchor bolt embedment where a grout was placed between the hanger baseplate and a wall or ceiling to which it was attached without case basis engineering approval.
The licensee determined t!at a total of 77 hangers had been grouted at the time of the inspection. With the exception of 20, all of the hangers have been reanalyzed by engineering based on actual bolt embedments.
Where the analysis indicated less than satisfactory embedment, necessary corrections have been made.
The 20 hangers not yet analyzed have been posted in a " tickler file" for review at a later date when all of the information relative to the hanger is more defined.
The licensee has revised Quality Control Procedure QI-QAP-ll.1-28 to require the quality control inspectors to measure bolt protrusion above the base concrete surface from which the embedment may be derived.
This information will be provided to engineering for their analysis of the "as-built" hanger. The RRI had no further questions on this matter and it.is considered closed.
3.
Site Tours The RRI toured each safety-related work area at least once during the inspec-
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tion period to observe the status of construction and construction practices being utilized by the various craft disciplines in each area.
The RRI noted that ths housekeeping in each area was about normal for the status and type of construction in each area. Those areas with a low-level of activity
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-3-were generally very neat and relatively clean, while those areas with heavy, sustained activity tended to be more cluttered with tools, some debris with cable and hoses used in welding operations in some profusion.
No violations or deviations were identified.
4.
Fire Protection Piping System Installation
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The RRI conducted a brief investigation relative to allegations received from another NRC Resident Inspector stationed at another site.
The allegations involved poor quality welding and a lack of well implemented quality control at a manufacturing facility of the contractor, Grinnell Fire Protections Systems, Inc.
The particular facility involved in the allegation was ident-ified as was the fact that the facility was primarily supplying prefabricated pipe spools to the other site.
The RRI was able to establish that the particular facility had not supplied any prefabricated spools to the Comanche Peak site with the exception of two standard commercial inline filters of welded construction. These two components are standard commercial, and
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widely used, filter elements that are under the control of the Underwriters Laboratories and the Factory Mutual Insurance Company.
The RRI found that the vast bulk of the piping runs are of threaded type construction rather than welded.
The weld type of pipe installation.is confined to six-inch and larger pipe sizes and have all been field fabricated on site under the Grinnell Fire Protection Quality Assurance Plan.
Based on observations over the past year and one-half nf the Grinnell installation activities, and upon a tour of the Grinnell pipe storage yard, the RRI concluded that the allegation as received from the other sitt had no connection to t;,y Comanche Peak site and the RRI had no further questions.
No violations or deviations were identified.
5.
Installation of Steel Structural Access Platforms The RRI observed, during various plant tours conducted during the past several months, that a number of fairly large platforms for operational maintenance eccess were being installed in various plant areas, including the Reactor duilding and Safeguards Building.
Certain of the platfoms were noted to be of a size and location such that if they were to fail in a seismic event,
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they migV ' ell damage nearby safety-related equipment and thus reduce the ability,t che equipment to respond as required to a design basis accident.
The platforms appeared to fit the intent of paragraph 2 of NRC Regulatory Guide 1.29 which would require them to be seismically designed and installed under appropriate quality controls.
FSAR Chapter 3.2 generally discusses the concept and lists several equipment classifications included in the licensee's definition of Seismic Category 2.
The platforms in question, however, are not listed.
Reference to FSAR Chapter 17A, a listing of various plant components and structures with appropriate safety or non-safety designations and/or seismic and non-seismic designations, also failed to identify the platforrs.
The RRI focused in on four platforms that had been
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-4-installed in the Steam Generator compartments in the Unit 1 Reactor Build-l ing whose apparent purpose was to provide worker access to the Steam Generator manways during future maintenance.
These platforms are of a size and in a l
location where if they were to fail structurally, substantial damage might occur to nearby safety-related components.
The RRI located the platforms on l
Gibbs & Hill Drawing 2323-S1-0555.
The RRI noted that the drawing contained the notation "NNS" next to " plates and shapes" under the outline of plat-
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form and again opposite a note relative to bolting materials.
The drawing also referred the reader to 2323-51-0500 for general notes and references.
Note 1 of this drawing lists a number of applicable project specifications, among which are SS-16A for non-seismic structural steel, 55-168 for Category 1 structural steel, and SS-17 for miscellaneous steel.
The RRI was not able L
to clearly discern which of the specifications the engineer deemed to be most applicable but after reading each of the documents, concluded that the most applicable specification was SS-17 for miscellaneous steel since it contains
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l both QA requirements where the drawings indicate that the steel is Category 1, and provisions for not performing QA were so noted ir. toe drawings.
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further compound the confusion, the top referenced drawing 2323-51-0500 states "*(asterisk) designates those structures which are not class 1."
l Since no asterisk was applied to the drawing 2J23-S1-0555, the RRI concluded the engineer meant that the structural platforms are " Class 1."
The term i
" Class 1," however, is usually used to denote the safety class of a component
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such as a pump, valve or pipe as in the ANSI system, while the term." Category" l
is applied to the seismic classification.
The RRI interviewed cognizant QA/QC l
personnel as to whether these platforms had been inspected during installation l
and was subsequently informed they had not been based on the "NNS" notation on drawing 2323-S1-0555. The RRI also was not able to ascertain whether the (
structural shapes involved were purchased and received with " Certified
Material Test Reports" as committed to in Chrter 3.8 of the FSAR which
states that all structural miscellaneous steel in the Reactor Building will
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be so certified.
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During and after discussions of this matter between the RRI, licensee engineer-ing, and quality assurance supervisory personnel, the RRI was provided with correspondence which indicated that there was a growing awareness that there might be a specification / quality assurance problem in this area but, the
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dates of the correspondance also indicated that no definitive action had been evolved in over six months until apparently prompted by the RRI's interest at which time the engineer asked QA to consider performing an inspec-tion of as yet to be determined stru"tural steel assemblies.
The RRI also became aware, during the course of reviewing the correspondence, that yet another area of concern was structural steel asserblies which are supporting stairways.
In some locations, these structures are in the immediate proximity of safety-related equipment and have the same damage potential as the structures discussed above.
This matter will be considered as an unresolved item until such time as one of the following actions has taken place:
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The Office of Nuclear Reactor Regulation (NRR) accepts an FSAR amend-ment defining the status of the platforms and stairways as not falling within the licensee's definition of Category II seismic structures (NRC Category I), or; b.
The licensee institutes a quality assurance inspection program meeting the requirements of Appendix 8 and other commitments of the FSAR.
6.
Design and Installation of Building Supplemental Steel Structures For Supporting Pipe The RRI noted during tours of the facility that a number of large steel structures had been installed in portions of Unit 2, particularly in the yard piping tunnel, that were notably different than the overall installation in the comparable area on Unit 1.
The RRI researched the reference files of construction design drawings attempting to locate these structures, but was unable to do so.
Continued probing revealed that the necessary installation drawings had been prepared and released for construction by the licensee's Comanche Peak Project Engineering group and were installed for the most part in a short time span in early 1980.
Further probing
turned up a set of QC generated inspection records that proved appropriate
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to the installation.
The design drawings revealed that the structures had been designed under the American Institute for Steel. Construction (AISC)
Structural Building Code to serve as supports for ASME and non-ASME pip-ing runs through the tunnel and certain building corridors.
In general, there will be ASME Section III, Subsection NF pipe supports between the new building structure and the actual pipe runs, although in a few instances this may not be the case.
Utilizing the design drawings and the QC inspection records, the RRI selectively verified that the structural frames were fabricated from the materials selected by the designers and welded in a quality manner utilizing weld sizes specified by the designers.
The inspection records indicated that the welds had been performed by a group of several welders, all of whom were qualified to perform the welds under
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both the ASME Code and the AISC/AWS Codes.
Discussions between the RRI and the current manager of the site Pipe
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Support Design Group (a component of the Comanche Peak Project Engineering Group) revealed that his group is currently reviewing the design to assure that structures are capable of supporting all necessary loads, including those seismically generated.
No violations or deviations were identified.
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Safety-Related Pipe Installation and Welding During this inspection period, the RRI did not observe any pipe joint.
. welding since such work has become very random as to location and relatively limited in quantity as the piping runs are near completion in the Unit l'
area and only limited work has been initiated in Unit 2.
The RRI did
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-6-cxamine the radiographs for the welds identified below for conformance to ASME Section III for weld quality and ASME Section V relative to the quality of the radiographs:
Weld No.
Isometric No.
Line No.
FW-20, 21, 22, 28, 29 & 30 RC-2-520-001 Unit 2 Reactor Loop FW-16 SI-1-RB-020 3-SI-1-020-2501R1 FW-1 & FW-22 SI-1-RB-014 2-SI-1-059-250lR1 FW-2A & FW-2-1 SI-1-RB-014 2-SI-1-059-250lR1 FW-1-1 SI-1-RB-014 2-SI-1-059-2501R1 FW-7A & FW-8 RC-1-RB-28A 6-RC-1-100-2501R1 FW-3-4A RC-1-RB-016 4-RC-1-091-2501R1 FW-Il RC-1-RB-05 6-RC-1-008-2501R1 FW-12 SI-1-RB-033 5-SI-1-092-2501R1 FW-25 RC-1-RB-032 3-RC-1-159-2501RI.
FW-6 SI-1-RB-056 6-SI-1-056-250lR1 8.
Reinstallation of the Unit 1 Containment Liner Personnel of Chicago Bridge & Iron Co. returned to the site after an absence of several months to reinstall that portion of the Unit 1 Reactor Building containment liner where the building construction access opening had been provided.
The liner material utilized in the work was that that had been removed and stored when the liner was initially erected.
The RRI observed two welders welding the lower plates which had been satisfactorily fit-up and pinned in place.
The RRI interviewed one of the welders sufficiently to obtain his name, the weld procedure he was utilizing, and the.identifi-cation of the weld rod involved.
The RRI reviewed the weld procedure, WPS-E-8010 74-2427/8, and found that it had been properly qualified and further, that it allowed the use of downward progression weld for the cover pass that the RRI had observed.
The welder had been initially qualified for the weld process in 1970 and had maintained his qualification contin-uously as noted on the records..The weld rod, identified as JJJ-046, was-documented as Heat Number 20415,. Lot G620B21AD, and furnished with proper Certified Material Test Report by the Chemetron Company.
The RRI
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-7-was informed that the welders, weld procedures, and the weld filler metals were the same as was utilized in the original construction of both containment liners and that in all probability would be used to also finish out the Unit 2 liner construction opening.
No violations or deviations were identified.
9.
Protection of Major Installed Equipment The RRI observed that the reactor vessels and internals for both units remain well protected.
Electric motors for pumps and valves were noted to be hand-warm in relation to surrout. ding metals and thus protected from moisture. The major equipment in Unit 1, for the most part, has tha component heaters energized from the design power sources rather by temporary power connections.
In Unit 2, most of the equipment is still warmed by temporary connections to internal heaters or by arrangement of large electric lights within the enclosures.
Equipment requiring physical protection from nearby construction activities were noted to be adequately covered with heavy reinforced plastic or wooden covers.
No violations or deviations were identified.
10.
Unresolved-Items Unresolved items are matters about whici: more information is required in
' order to ascertain whether they are acceptable items, violations, or deviations.
One such item was disclosed during the inspection and is discussed in paragrpah 5.
11.
Management Interviews The RRI met with one ir more of the persons identified in paragrpah 1 on October 2, 7, 19, 22, 23, 28 and 29, 1981, to discuss inspection findings and the licensee's actions and positions.
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