ML20053A848
| ML20053A848 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 05/24/1982 |
| From: | Spencer S TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC) |
| To: | |
| Shared Package | |
| ML20053A827 | List: |
| References | |
| NUDOCS 8205270349 | |
| Download: ML20053A848 (96) | |
Text
&
O UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION O, r. 4? U.,' U-N' BEFORE THE ATOMIC SAFETY AND LICENSING BOARD L E'
In the Matter of
)
)
"EXAS UTILITIES GENERATING
)
Docket Nos. 50-445 and et al.
)
50-446 COMPANY, -
)
(Comanche Peak Steam Electric
)
(Application for Station, Units 1 and 2)
)
Operating Licenses)
TESTIMONY OF SUSAN L.
SPENCER REGARDING DISPOSITION OF NRC I&E REPORTS Ol.
Please state your name, residence and educational and professional qualifications.
A1.
My name is Susan L.
Spencer.
I reside in Dallas, Texas.
A statement of my educational and professional qualifications is attached hereto as Attachment 1.
Q2.
What it your current position?
A2.
I am employed by Texas Utilities Generating Company,
("TUGCO") in the position of Quality Assurance Auditor.
In this position, I am responsible for monitoring and assuring resolution of issues raised by the NRC in Inspection & Enforcement Reports ("I&E Reports") for the Comanche Peak Steam Electric Station.
03.
What is the purpose of your testimony?
A3.
The purpose of my testimony is to demonstrate that all matters raised in NRC I&E Reports (including unresolved items and Notices of Deviation and Notices of Violation) 8205270 M 5
3 J
_2_
cited by Intervenor CASE as pertinent to Contention 5 have been resolved and such resolution confirmed by the NRC Staff, or that Applicants have taken appropriate corrective action that is subject to verification by the NRC Office of Inspection and Enforcement.
In addi-tion, I will address four specific topics raised in I&E Reports in order to detail the measures taken to resolve the issues, and the status of acceptance by the NRC of those measures.
Q4.
What procedures are followed by the NRC in performing inspections at Comanche Peak?
A4.
The NRC retains a Resident Reactor Inspector ("RRI")
on site at the Comanche Peak facility.
This individual and other Region IV staff personnel perform, on a rou-tine basis, inspections at the plant of the activities authorized by the construction permits for Comanche Peak.
Inspections generally consist of an examination of procedures and representative records, interviews with licensee construction personnel and management and observation of ongoing activities and completed work at the plant.
QS.
Does the Inspector prepare a report-of each inspection?
AS.
Yes.
Following each inspection, the inspector pre-pares a report which details the findings during the inspection.
These findings may include instances which require further examination or information to determine I
I
. acceptability, but which do not constitute existing variations from NRC requirements.
These are identified in the report as " unresolved items."
Where apparent variations from NRC requirements (including apparent variations from procedures or instructions established to assure compliance with NRC requirements) are dis-covered, the Inspector will recommend and the NRC Office of Inspection and Enforcement will transmit to the licensee either a Notice of Violation (involving either a violation, infraction or deficiency) or a Notice of Deviation, depending upon the severity of the variation.
These notices are generally transmitted to the Applicants before the I&E Reports are issued and thus are also referenced in the I&E Reports.
06.
What measures do the Applicants take to respond to NRC findings in I&E Reports?
A6.
To resolve issues identified as " unresolved items,"
Applicants generally meet with the Inspector to pre-sent the results of the Applicants' investigations of the matter. Additional documentation is provided, as The RRI will review the information pre-necessary.
sented and recommend further action such as closing the item, seeking more information or issuing a Notice of Deviation or Violation.
When an ite.n is closed out, a notation to that effect generally is made in the first I&E Report issued after such action is taken.
. 1 l
If a Notice of Deviation or Violation is issued, the Applicants transmit a formal written reply within a specified period of time responding to each issue raised.
The Applicants provide the information necessary to respond to the discrepancy as requested.
The Applicants' response describes any corrective steps already taken and the results achieved, corrective steps to be taken to avoid further noncompliance and the date of full compliance.
When the NRC is satisfied that appropriate action has been taken by the Appli-cants, the Notice is closed out in a future inspection by the NRC, and a notation to that effect is made in a subsequent I&E Report.
07.
How often does the NRC perform inspections?
A7.
Since 1973 the NRC has performed routine inspections, at least monthly, of preparations for and actual con-struction activities at Comanche Peak, as documented by I&E Reports.
During this period over 150 NRC I&E Reports have been issued.
Our records reflect that over 8000 NRC inspector hours have been spent on site conducting the inspections and investigations which are the subject of those reports.
Significant addi-tional NRC resources have been spent off site on inspection activity as well.
The results of these inspections and Applicants' response to each, when
~ required, are a matter of public record.
In every case, the findings of the NRC Inspectors were reviewed by Appli-cants and actions were taken to correct any matters raised therein.
08.
What NRC I&E Reports are the subject of this testimony?
A8.
I have examined CASE's Answers to Applicants' First, Third and Fifth Sets of Interrogatories, dated September 3, 1980 (Supplemented December 1, 1980),
March 16, 1982 and April 20, 1982, respectively.
I have also examined Answers to the NRC Staff's First and Fourth Sets of Interrogatories, dated February 17, 1981 (Supplemented April 6, 1981) and March 15, 1982, respectively.
All NRC I&E Reports which are cited in those answers are included within the scop'e of my testimony.
In addition, each I&E Report cited in ACORN's Offer of Proof, served August 29, 1980, is within the scope of my testimony. Finally, I do not l
l address I&E Report 80-09 regarding groundwater with-drawal rates (which has not been formally closed-out),
because it does not involve matters of quality assurance.
09.
What is the status of the matters raised in those NRC I&E Reports?
A9.
All but two issues raised in I&E Reports which are cited by CASE as pertinent to Contention 5 have been resolved and that resolution verified by the NRC Staff.
Such verification, in all but one instance, has bs+n by a formal close-out in a subsequent I&E Report.
i
- 010. What was the one instance for which formal close-out has not been effected?
A10. Applicants have completed work pursuant to a commit-ment for resolution of an unresolved item raised in I&E Report 80-20, involving the spacing of, and circuit breakers for, safety and non-safety cables in the AC Instrument Distribution Panels.
I&E Report 80-20 is attached hereto as Attachment 2.
Applicants' commit-mitment for resolution of the item is set forth in FSAR s8.3.2.1 Paragraph 7.c (Applicants' Exhibit 3).
The commitment was accepted by the NRC Staff in SER Supple-ment No.
1, 58.4.4,
- p. 8-1 (Staff Exhibit 2), although a formal close-out has not been made in a subsequent I&E Report.
Oll. What are the two unresolved issues?
All. The unresolved issues involve certain procedures for the inspection of coatings raised in I&E Report 81-15 (attached hereto as Attachment 3) and a concrete pour on the Unit 1 dome raised in I&E Report 79-11 (attached hereto as ).
012. From a QA perspective, what are the facts regarding the concrete pour on the Unit 1 dome?
A12. As described in I&E Report 79-11, on March 30, 1979, the NRC Region IV office received a telephone call from an individual who identified himself as a former CPSES
. employee. The individual alleged that during a concrete pour on the Unit 1 dome in January, 1979, a rain washed away part of newly poured concrete, and that the affected area was repaired with grout.
Subsequent investigation by the NRC RRI indicated that the inci-dent apparently occurred on January 18, 1979.
It was determined that on that day a concrete pour on the Unit 1 dome was begun under good weather conditions.
The weather subsequently deteriorated to the point that heavy rain stopped work at about 7:30 p.m.
The pour area was covered and the incoming shift was instructed to clean the area so the pour could resume the next day.
The RRI subsequently identified time sheets of individuals which indicated they had been " placing concrete" during the later shift.
Q13. Who discovered the facts concerning the employee's allegations?
A13. On April 17, 1979, the RRI was asked whether Brown
& Root could check with its personnel regarding the allegations.
It was informed that it could do so because the RRI had completed his on-site investiga-tion.
The next day, the Applicants reported to the RRI that they had identified the craft General Foreman l
l
. who was involved.
As described in I&E Report 79-11, it was determined that concrete placement had occurred that evening to replace a small amount of concrete which had washed away in the rain.
The General Foreman had himself mixed the concrete (one-half yard) in accordance with design mix data for the dome concrete, although no QA personnel were present.
Consequently, a Notice of Violation was issued for the incident.
Q14 What is the status of that Notice of Violation?
A14. In I&E Report 79-24/23, the NRC Office of Inspection and Enforcement formally closed the Infraction cited in I&E Report 79-11 regarding implementation of the Quality Assurance program.
A copy of I&E Report 79-24/23 is attached hereto as Attachment 5.
As stated in that Report, Applicants advised the Region IV Office of Inspection & Enforcement that reviews by consultants and Gibbs & Hill had been completed and the in-place concrete was found satisfactory.
In addition, Applicants informed Region IV that construction supervisory per-sonnel have been informed that should a similar situation occur, no additional concrete shall be batched or placed without prior notification to senior construction manage-ment.
. 015. Do any items remain open regarding this matter?
A15. The structural integrity of the concrete used in this pour was cited as an unresolved item in I&E Report 79-24/23.
This item will be closed out following the Structural Integrity Tests to be performed pureuant to 10 C.F.R. Part 50, Appendix J, which test the leak-tight integrity of the entire primary reactor containments of both Units 1 and 2, as described in SER % 2.8.1 at p.
2-18 (Staff Exhibit 1).
016. What are the facts regarding the inspection of coatings?
A16. In I&E Report 81-15 (Attachment 3 hereto), the NRC raised certain matters concerning documentation of coating applications for miscellaneous steel, cable tray supports or pipe supports inside Units 1 and 2 Containment Buildings.
Also, the records reviewed by the RRI for the Unit 2 Containment Steel Liner revealed incomplete checklists without recorded visual inspections and Dry Film Thickness readings.
Q17. What actions has TUGCO QA taken with regard to the matters raised in I&E Report 81-15 concerning inspec-l tion of coatings?
l i
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. 1 A17. The discrepancies cited in I&E Report 81-15 have been identified as nonconforming conditions in accordance with established QA procedures.
In response, reinspect-ions of the subject coatings using both scratch and l
adhesion tests to evaluate the condition of the applied coatings are being conducted.
A complete review of existing records and reinspection of affected areas is being conducted. Any discrepancies are being identified and corrected in accordance with approved procedures.
018. What measures have been taken to prevent recurrence l
l of this problem?
A18. To prevent recurrence of this matter, Application (Construction) Procedures were revised and reissued to clearly indicate pot life at all temperatures within the applicable range for application of the coating system. In addition, Inspection (Quality) Procedures /
Instructions were revised to clarify applicable require-ments and were reissued. Also, an identification system providing traceability of inspection documentation from blasting through installation and final coating for l
miscellaneous steel and supports was established.
Formal close-out of this item will occur upon verifi-cation by the NRC Office of Inspection & Enforcement of satisfactory completion of the actions described I
above.
l L
~ Q19. Are you familiar with OA aspects of the matter involving honeycombing in the Unit 2 steam generator compartment?
A19. Yes, I am.
Q20. What are the facts regarding that matter?
A20. In October 1979, routine Quality Control inspections identified and documented areas in the concrete place-ment of the Unit 2 steam generator compartment walls where exposed concrete contained honeycombed condi-tions.
Following engineering review of the condition, repair work was authorized to begin in early December, 1979.
During repairs, Engineering and Senior Quality Control personnel recommended that the integrity of inaccessible portions of the placement be investigated further.
It was determined on the basis of that recommendation that further investigation was required.
Upon completion of further investigation and evaluation based on all data available, it was concluded by engin-eering personnel that the inaccessible portions of the e
placement, excluding the honeycombed portions already identified, met or exceeded design requirements or l
contained no hidden internal defects which would be detrimental to the safety or utility of the structure.
021. Did the NRC Inspector inspect the repair work?
A21. As discussed in I&E Report 80-08 (a copy of which is attached hereto as Attachment 6), a subsequent examina-tion by the NRC RRI of the repair work on the honeycombing 1
. was conducted in March, 1980.
The RRI found that work was being accomplished in accordance with detailed instructions generated at the site and the recommenda-tions set forth in " applicable portions of the U.S.
Bureau of Reclamation ' Concrete Manual', a recognized authoritative publication on concrete work."
The NRC Staff conducted extensive reviews in April and May, 1980 and concluded that no items of noncompliance or deviations existed.
Those conclusions are contained in I&E Report 80-11, a copy of which is attached hereto as Attachment 7.
022. What involvement did TUGCO/TUSI QA have in the repair work?
A22. In accordance with pertinent GA procedures, TUGCO/TUSI CA personnel verified that all repair work was con-ducted in accordance with appropriate specifications and procedures.
023. Are you familiar with the QA aspects of rework per-l formed on the Unit 2 reactor vessel support structure?
A23. Yes, I am.
1 024. What are the facts involving that matter?
I A24. On February 20, 1979, Applicants reported to the NRC t
l RRI that an error had been discovered in the design l
l of the Unit 2 reactor vessel support structure.
Appli-l j
cants reported that the reactor vessel support shoes, the ventilation duct work, and the surrounding i
, reinforcing steel had been rotated forty-five degrees from correct positions through a design error.
See I&E Report 79-03, a copy of which is attached hereto as Attachment 8.
Q25. How did the design error occur?
A25. The design error resulted from the fact that evolving design criteria were not coordinated adequately.
The evolving design included changes to the reactor i
vessel supports to reflect the revised design cf the Unit 2 reactor vessel to be identical to the Unit 1 vessel (rather than a mirror image), with associated changes to the primary coolant systems.
As a result, the mounting pads for the Unif. 2 vessel were misoriented by about 45 degrees.
l 026. Was the rework performed to the satisfaction of the NRC?
A26. Yes, it was.
Applicants and NRC representatives held a 1
l meeting in March 1979 to discuss the repair procedures I
for relocating the vessel support pads.
The meeting focused on differences between the original design of the reactor vessel support pedestal and Ehe repaired l
pedestal.
The NRC. concluded t at "no' unresolved safety l
concerns associated with the' repair design for the Unit 2 pedestal were' identified at the. meeting."
See 9
f
. NRC Meeting Summary (May 15, 1979), a copy of which is attached hereto as Attachment 9.
Q27. What was the involvement of TUGCO/TUSI QA regarding this matter?
A27. Prior to commencement of construction activities, TUGCO QA reviewed the scheduled activities to assure establish-ment of required hold points.
During and upon comple-tion of rework activities, TUGCO QA conducted necessary inspections to assure completion in accordance with applicable requirements.
Q28. In sum, what is the status of matters raised in NRC I&E Reports cited by Intervenor CASE?
A28. As demonstrated above, all matters raised in NRC I&E Reports cited by Intervenor CASE as pertinent to Contention 5 either have been resolved to the satisfaction of'the NRC Staff or have been the sub-
~
ject of corrective action by Applicants which is subject to verification by the NRC Office of Inspection and Enforcement prior to formal close-out of the item by the NRC.
ATTACHMENT 1 s
SUSAN L.
SPENCER STATEMENT OF EDUCATICNAL AND PROFESSICNAL-QUALIFICATIONS POSITION:
Quality Assurance Auditor FORMAL EDUCATICN:
Texas Tech University, 21974-75 Texas A&M University, 1976-78 SEA-Management OTHER:
Certified.in auditing nuclear quality assurance progqams by Stat-A-Matrix Institute - 1979.
Eighty hours l ok ' formal training in Radiographic' Testing'- Westinghouse Electric Corporation - 1979.'
Forcy hours of training in Ultrasonic Exar'ination - Rockwell In.ernational - 19 79.
EXPERIENCE :
~
i 1979 to Present Texas Utilities Generating Company, Dallas, Texas,; Quality Assurance Auditor.
Responsible for codrdinating responses to Nuclear Regulatory Cor. mission Inspection and Enforcement Reports.
Perform audits as necessary,,to support QA Division activities.
s
- . I PROFESSICNAL:
American Society for Quality Control, Member x
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e ATTACM'ENT 2 e arco UNITED STATES
-[.s NUCLEAR REGULATORY COMMISSION o
fh" REGloN IV
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c 611 RY AN PLA:A o RIVE, SUITE 1000 g
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Oc:ober 2',
1980 1
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In Reply Refer To:
v -._ ;.I.~~g RIV Docket No. 50-445/Rpt. 80-20 0C7 "3 50-446/ apt. 80-20 b_d_
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Texas Utilities Generating Cc=pany ATTN:
Mr. R. J. Gary, Executive Vice OCT 221980 President and General Manager 2001 3ryan Tower R.J. GA?l Dallas, Texas 75201 Gentle =en:
This refers to the inspection conducted by our Resident Reactor Inspector, Mr. R. G. Taylor, during Septe=ber 1980, of activities authorized by NRC Construction Per=its No. C??R-126 and 127 for the Co=anche Peak facility, Units No. I and 2, and to the discussion of our findings with Mr. R. G.
Tolson and other = embers of your staff at the conclusion of the inspectic'n.
Areas exa=ined during the inspection and our findings are discussed in the enclosed inspection report.
'a'ithin these areas, the inspection censisted of selec:ive exa=inatien of procedures and representative records, interviews with personnel, and observations by the inspector.
During the inspectica, it was found tha: certain activities under your license appear to be in ncncc=pliance with Appendix 3 to 10 CFR 50 of the NRC Regulation, " Quality Assurance Criteria for Nuclear ?cwer Plants." The Notice of Violation for :he ite= of nonco=pliance reported in the enclosed inspectica report was forwarded to you by our letter dated Septe=ber 24, 1980; therefore, this le::er does not require further response regarding this =a::er.
During the inspection, it was found 1. hat certain of your activities appeared to deviate f c= cc _it=ents in the FSAR.
This ite= and refe' ences to the specific co =1:=ents are identified in the enclosed Notice of Deviation.
In your reply, please include your ec==ents concerning this ite=, a description of any steps that have been or will be taken to correct it,
.1 descriptien of any steps that have been or will be taken to prevent recurrence, and the date all corrective actions or preventive =easures were or will be ce=pleted.
Two new unresolved ite=s are identified in paragraphs 4 and 7 of the enclosed report.
i ATTACOiENT 2 i
Texas Utili:ies Generating Cc=pany 2
Oc:cber 21, 1980 In accordance w1:h See icn 2.790 of the NRC's " Rules of Prae:1:e," Par: 2, Title 10, Code of Federal Regulations, a copy of :his le::er and the enciesed inspection report will be placed in the NRC's Public Docu=en: Roc =.
If the report cen:ains any infer:ation that you believe :o be preprie:ary,1: is necessary that you sub=1: a written applica:icn to this office, wi:hin 20 days of the date of this letter, requesting that such infor:a:1cn be withheld f c=
public disclosure.
The applicatica =ust include a full state =ent of :he reasons why it is clai=ed tha: the infor=ation is proprietary.
The application -should be prepared so that any proprietary infor=atica identified is contained in an enclosure to the application, since the applica:icn withou: the enclosure will also be placed in the Public Docu=ent Roc =.
If we do not hear frc= you in this regard within the specified period, the report will be placed in the Public Docu=en: Room.
Should you have any questions concerning this inspectica, we will be pleased i
to discuss the= with you.
Sincerely, W.C.Sedlb, Chief Reactor Conhjruction and Engineering Supper: 3 ranch
Enclosures:
1.
Appendix 3, Notice of Deviation 2.
IE Inspec:fon Report No. 50-445/80-20 50-446/80-20 cc: w/ enclosures Texas Utilities Genera:Ing Cc=pany ATTN:
Mr. H. C. Sch=idt, Project Manager 2001 3ryan Tcuer Dallas, Texas 75201 V, C.*
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50-445/80-20 30-446/80-20 Appendix 3 NOTICI 0F DEV'_A'"ICN Based on the resul:s of :he NRC inspection condue:ed during Septe:ber 1980, 1: appears tha: certain of your activities devia:e fro co-d :=ents =ade in your Final Sade:y Analysis Report (?SAR) as indicated below:
Incorrect Design of Pressurizer Serav Control 7alve ? icing FSAR, Section 5.1, Figure 5.1-1, Note 2 states, " Spray pipe sloped to provide va:er seal between pressurizer and spray valvas.
Valves :o be a =ini=um of 10 ft. below pressurizer :op."
Contrary to the above:
The Resident Reac:or Inspec:or observed that the partially installed pressuriser spray control valves were above the top of the pressurizer.
Subsequent review of the engineered design as displayed on iso =e ric drawings RC-1-R3-016, 017 and 018 established tha: valves 1-pVC-4553 and C are placed 1 foo: 4 inches and 2 fee: 11 inches, respec:1vely above the top of the pressuri ar rather chan a: leas: 10 fee: below.
This is a deviation.
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ATTACFJENT 2 U. S. NUCEIAR REGUI.ATORY COSISSION OFFICI 0F INS?ECTICN AND ENIORCE. tit REGION IV Report No. 50-445/80-20; 50-446/80-20 Docket No. 50-445; 50-446 Category A2 Eicensee:
Texas Utilities Generating Company 20013ryan Tower Dallas, Texas 75201 Facility Name:
Comanche Peak, Units 1 and 2 Inspection at:
Comanche Peak Steam Electric Station, Glen Rose, Texas Inspection conducted:
September 1980 Inspec or:
d;()
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- l7sfft R.6.'TaylorfResiden Reactor Inspector, Projects Section Dfte/
f Approved:
W
/#[24[fc W. A. Crossman, Chief, Proj ects Section Date Inspection Su==ary:
Inspection during Secte=ber 1980 (Recor No. 50-445/80-20; 50-446/80-20)
Areas Inspected:
Routine, announced inspection by the Resident Reactor Inspector (RRI), with limited support by two Regional II inspectors, in-cluding general site tours; piping, electrical and instrumentation, installa-tion activities; protection of installed and uninstalled equipmen and pipe supports and restraints.
The inspection involved ninety-two inspector hours by the RRI and four hours by two Regional II inspectors.
i l
Results:
No items of noncompliance were identified in five of the areas.
one item of noncompliance and a deviation (infraction - unsuitable weld surface as required by magnetic particle test procedures paragraph 6; deviation - incorrect design of pressuriner spray sys:em control valve piping paragraph 5) were identified in two other areas.
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ATTACHMENT 2 DETAILS 1.
Persons Contacted Princiral Licensee E=clovees
- R. 3. Clements, TUGCO, Vice-President, Nuclear Operations
- D. N. Chapman, TUGCO, Quality Assurance Manager
- R. G. Tolson, TUGCO, Site Quality Assurance Supervisor Other Persons J. V. Hawkins, Brown & Root, Project Quality Assurance Manager D. C. Frankum, Brown & Root, Construction Project Mangager The RRI also intervtewed other l'icensee and Brown & Root employees during the inspection period including both craft labor and QA/QC personnel.
- Denotes those persons with whom the RRI held on-site management meetings during the inspection period.
2.
Site Tours The RRI toured the safety-related plant areas several times weekly during the inspection period to observe the general progress of construction and the practices involved.
Four of the tours were accomplished during portions of the second shift where the main activity continues to be installation of electrical cables and the application of protective coatings.
1 No items of noncompliance or deviations were identified.
3.
Protection of Major Installed Ecuirment The RRI observed that the Reactor Vessel Internals (core support structure) were partially installed in the Unit I vessel as part of the final alignment machining operation.
The internals and the vessel were well covered and protected from any construction j
debris.
The Unit 2 Reactor Vessel continued to be protected during periods when personnel had to enter the vessel for coolant pipe welding and weld inspection.
The Unit 2 internals remain in their storage enclosures near a facility warehouse.
The RRI observed that randomly selected electric prime movers for pumps have their space heaters energized as have the motors associated with motor operated valves.
The safety-related switchgear and motor control centers either have their space heaters energined or have large light bulbs installed to keep them dry.
The control room panels and boards were maintained l
in an air-conditioned environment during the period to preclude heat damage to the electronic components.
No items of noncompliance or deviations were identified.
l ATTACIDiENT 2 4
Storage of Uninstalled Co=conents The RRI toured the warehouses and outdoor laydown areas during the inspection period.
All co=ponents observed, including a substantial at=ber of prefabricated pipe spools and cable reels, appeared to be well cared for in accordance with nor=al industry practice and the recommendations of the various suppliers of major components, with one exception.
The RRI noted in touring an outdoor storage area what appeared to be spent fuel storage racks identified by an attached sign as "NON-Q" (not within the scope of the licensee's quality assurance program).
Also the lay-down positions of some of the rack modules were such that water might collect in some areas in the event of rain.
The RRI reviewed the files of the on-site vendor coordination and found that the racks had been furnished by Westinghouse as Seismic I components as indicated by the FSAR Section 17.2 and that there was a Westinghouse quality release document on file.
The RRI initiated a discussion with licensee QA personnel and was infor=ed that they were aware of the problem and that a Nonconformance Report had been initiated but not yet finalined.
This situation will be considered to be an unresolved item pending completion of the Nonconformance Report and whatever actions are taken to rectify the matter.
5.
Safetv-Related Picing Installation and Welding I
The RRI made several general observations of the handling practices l
for piping components during the inspection period both in the on-site fabrication shop and within the main plant buildings.
The practices were found to be consistent with the construction practices outlined in Construction Procedure 35-1195-CPM 6.9 which in turn are i
consistent with the Project Specification MS-100 and good industry practice.
The RRI observed, during a portion of one of the tours, that the installation and welding of a portion of the pressuriner spray piping had evolved to a point where the final layout configuration of this subsystem of the reactor coolant system was evident.
The RRI obtained the pertinent isometric drawings for the subsystem and verified that the piping was installed essentially as shown.
The design layout, however, was found not to match the commitments of Section 5.1 of the FSAR in that Note 2 of Figure 5.1-1 had not been achieved.
This note requires that pressuriner spray control valves be located a minimum of ten feet below the top of the pressur-iner while the isometric drawings indicate that the two valves are actually located two feet eleven inches and one foot four inches, l
respectively above the top of the pressuriner.
This is considered to be a deviation to the FSAR commitments.
3
ATTACIU1ENT 2 i
The RRI observed the following welds being =ade during the inspection period:
Weld No.
Isometric Filler Ht.
Welder (s)
Procedure Process FW-4 AF-2-53-003 87401 2PK 11021 GTAW i
W-7-12 SW-1-SB-003 762550 SWP 88025 GTAW FW-5 RC-2-520-001 434788 ARN-ERS-AFP 99028 GTAW (Machine)
FW-1 RH-1-R3-001 746100 AEN 88021 GTAW The RRI verified that the weld filler metals being utilized in the above welds were certified by the suppliers via Certified Material i
Test Reports as meeting the applicable requirements of ASMI Section II.
The welders and weld procedures noted were verified to have been properly qualified in accordance with ASME Section IX.
The RRI also examined the radiographs of the following welds for compliance to the requirements of ASME Section III for weld quality l
and Section V for the quality of the radiographs themselves.
a 4
Weld No.
Isometric Line No.
1
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W-16 RC-2-RS-073 4-RC-2-091-2501R1 FW-16A CS-1-RB-028 2-CS 112-1501R1 W-4 CS-2-RE-042 3-CS-2-235-2501R1 FW-14 SI-1-RB-017 6-SI-1-04 6-2501R1 l
FW-7A CS-1-RS-023 2-CS-1-112-2501R1 FW-7 SI-1-RS-037 10-SI-1-179-2501R1 V-13 SI-2-R3-042 2-SI-2-036-2501R1 FW-11A RC-1-RS-006 6-RC-1-070-2501R1 W-19 FW-2-SB-029 6-FW-2-097-1303 l
Except as noted above, no items of noncompliance or deviations were identified.
6.
Piping System Sucports and Restraints l
The RRI initiated an inspection of a group of some one hundred piping system moment restraints that had been fabricated by Chicago Bridge & Iron Co. under Brown & Root (acting as a purchasing agent for the licensee) subcontract 35-1195-0585 and shipped to the site complete, except for sandblasting and painting.
The inspection was initiated as a result of an investigation of allegations received by Region IV as documented in Inspection Report 50-445/80-22; 50-446/80-20.
The allegation essentially stated that some of the weld surfaces of the fabrications were ~ such that a meaningful magnetic particle inspection within the requirements of ASME Section III, subsection NF could not have been done as certified to by the subcontractor.
ATTACEMDIT 2 i
A secondary consideration contained within the allegation, although not specifically stated, was tha: the Authorized Nuclear Inspector assigned to the facility was derelict in that the fabrication had been stamped and certified as having me: the Code requirements.
The RRI, with the assistance of two NRC inspectors specialized in veld inspections and nondestructive exa=ination techniques, performed a magnetic particle and visual examination of select:f portions of the welds of four of the fabrications using the Chicage Bridge &
Iron procedures.
Several of the selected areas examined revealed indications that could only be resolved by surface removal of the welds, thus substantiating the specific allegation.
The secondary allegation referred to above was effectisely refuted when the RRI found that the magnetic particle inspection was required by the Architect / Engineer as a supplemental addition beyond the require-ments of the ASME Code.
Subsection NA of the Code specifically excludes the Authorired Nuclear Inspector from taking responsibility for requirements imposed by a designer that are in excess of the Code requirements.
The Code requirement was for a visual examination only of these components and each of the welds examined by the NRC personnel appeared to satisfy the Code stated acceptance criteria.
The licensee was informed of the NRC findings infor= ally on September 19, 1980, and formally by a Region IV letter dated September 24, 1980.
The licensee placed a QC hold on all of the components and'has initiated a reinspection program involving an identified one hundred five units received under the order.
The RRI also beca=e aware of another substantial group of large fabrications used as pipe whip restraints that were fabricated by Chicago 3 ridge & Iron under another subcontract and has initiated an inspection of these that is not complete.
7.
Electrical Installacion Activites The RRI made a number of observations of electrical cable pulling operations during the period.
The RRI observed the activities of each of the six cable pulling crews one or more times during the period in order to ascertain whether they were working within the parameters of the site installation procedures and good practices.
j The RRI alo observed the activities of the QA/QC' personnel assigned to monitor the pulling crews.
The RRI found both groups (craft and QA/QC) to work consistently within their respective procedures, EEI-7 and QE-QP-11.3-26.
The RRI also inspected randomly selected cable tray segments in Unit 1 Safeguards area for freedom from cable damaging burrs and excessive debris.
The selected segments were found to be in a satisfactory condition for cable pulling.
The RRI examined the wiring within the 11S VAC Distribution Panels shown on FSAR Figure 3.3-15 and identified as IPC1 thru IPC4 It was found that from two to four consafety cables entered each panel 5
o
^ - - ^C e"'
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4 J
l along with several safety-rela ed cables.
Withi: the panel, the safety and =onsafety cables were tightly tied together which appears a
i to be contrary to commitments of the ISAR to maintain a six inch space between safety and consafety cables withis panels.
Reference to the above FSAR Figure revealed that safety and sensafety cables were shown and tha: the consafe y cable was to be landed on circuit breaker-1 asse= bled a=ong the safety-related breakers which would make the sii j
inch spacing =early i=possible regardless of cable routing techniques.
The same Figure also shows the consafety cabling being powered off a i
safety bus with no other seans than a circuit breaker of an over-curren type to serve as isolation.
This situation appears to violate the recommendations of Regulatory Guide 1.75.
It appears tha: the nonsafety cables should be designated as associated safety-related cables and i
given segregated routing through the cable raceway system.
A Since the issue is clearly shown on the referenced FSAR Figure and since the electrical / instrument wiring scheme is still under active review by NRR, this matter will be considered to be an unresolved item rather than a deviation subject to either a FSAR revision or specific approval by NRR of the as-built design.
8.
Instrument Installation Activities i
The RRI observed during the inspection period that various instruments are being placed on their supports throughout the Unit 1 and Cocoon plant areas.
The instruments observed were noted to be covered with wooden boxes wired in place for the purposes of physical p:ctection.
j Where the instrument impulse tubing has not been connected to the instru=ent, the ports have been capped to protect from the entrance of debris, all considered to be good industry practices.
1 t
i The RRI continued to examine the engineer's drawing in the electrical and instrument areas to establish whether all safety-related devices
}
and circuits have been identified and to also develop the II inspection plan.
The RRI noted the Engineer's Instrument Tabulation List does not appear to address those safety-related devices associated with the heating and ventilation controls as being within the QA/QC scope even though they are connected with safety-related cabling to safety-1 related control circuits.
The RRI discussed this matter with cogniza=:
I engineering personnel and was informed that the instrument list was under active review and would be corrected in a timely ma==er.
No items of concompliance or deviations were identified.
-{
9.
Unresolved Items 1
Unresolved items are matters about which more infor=ation is required in order to ascertain whether they are acceptable items, items of noncompliance, or deviations.
Unresolved items disclosed during the
)
inspection are discussed in paragraphs 4 and 7 and will, in future inspection reports, be referred to as:
4 1
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ATTACIDENT 2 50-445/30-20; 50-446/30-20 Spen: Fuel Storage Racks a.
b.
50-445/30-20; 50-446/30-20 Design of the AC Instrunen:
Distribucion Panels 10.
. Manage =en: IntervieVs The RRI net with one or more of the persons identified in paragrach 1 0n September 2, 3, 5, 12, 15, 19, and 24, 1980, to discuss inspection
,.indings and the licensee's actions and positions, i
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w ere ATTACFJ.ENT 3
. [;
o UNITED STATES
[',~gj NUCLEAR REGULATORY COMMISSION REGION IV 0, ?M
[
611 RYAN PLAZA ORIVE, SulTE 10C0
%.'.b' AR LINGTON, TEXAS 76011
.s November 6, 1981 In Reply Refer To:
RIV Dockets : 50-445/Rpt. 81-15 50-446/Rpt. 81-15 T v c n.
Dcr
._. L uv-e Texas Utilities Generating Company ATTN: Mr. R. J. Gary, Executive Vice WW 9 192i President and General Manager
~
2001 Bryan Tower v..
, m Dallas, Texas 75201
' ". ". T D. :..a Gentlemen:
This refers to the inspection ccnducted by Mr. C. E. Johnson of our staff during the pericds October 13-16 and 19-23, 1981, of activities authorized by NRC Construction Permits CPPR-126 and CFPR-127 for the Comanche Peak facility, Units 1 and 2, and to the discussien of our findings with Mr.
R. G. Tolson and other memcers of ycur staff at the conclusion of the inspection.
Areas examined during the inspection and our findings are discussed in the enclosed inspecticn report. Within these areas, the inspection censisted of selective examination of procedures and representative records, interviews
,ith persennel, and cbservatiens by the inspector.
w During the inspecticn, it was found that certain activities under ycur license appear to be in violation of Appendix 3 to 10 CFR Part 50 of the NRC Regulation, " Quality Assurance Criteria for Nuclear Pcwer Plants." The Notice of Holation for the violation reported in the enclosed inspecticn i
report was fonvarded to ycu by cur letter, dated October 23, 1981; therefore, this letter dces not require further response regarding this matter.
One new unresolved item is discussed in paragraph 3.a of the enclosed report.
In accordance with 10 CFR 2.790 of the Ccmmission's regulations, a ccpy of this letter and the enclosed inspection report will be placed in the NRC's Public Document Roem.
If this report contains any information that you believe to be exempt from disclosure under 10 CFR 9.5(a)(4), it is necessary j
that you (a) notify this office by telephone within 10 days frem the date of this letter of your intention to file a request for withholding; and (b) sub-mit within 25 days from the date of this letter a written application to this office to withhold such information.
If your receipt of this letter has been delayed such that less than seven days are available for your review, please notify this office prcmptly s.o that a new due date may be established.
Received NOV 9 1981 R. J. GARY t
I ATTACH.'eiT 3 l
Taxas Utilit;!es Generating Company Ncvember 5, 1981 Consistent with Section 2.790(b)(1), any such application must be acccmpanied by an affidavit executed by the cwner of the information which identifies the document or part scught to be withheld, and which centains a full statement of the reascns en the basis which it is claimed that the informaticn shculd be withheld frca public disclosure.
This secticn further requires the statement to address with specificity the censiderations listed in 10 CFR 2.790(b)(4).
The information sought to be withhald shall be incorporated as far as possible into a separate part of the affidavit.
If we do not hear from you in this regard within the specified periods noted above, the report will be placed in the Public Document Room.
i Should you have any questfcils concerning this' inspection, we will be pleased to discuss them with you.
Sincerely,
,-k W
9 W. C. Se1 Chief Engineeri'. Inspection Branch
Enclosure:
Acpendix - IE Inspection Report 50-445/81-15 50-446/81-15 cc: w/ enclosure Texas Utilities Generating Ccmpany ATTN: Mr. H. C. Schmidt, Project Manager 2001 Bryan Tcwer Dallas, Texas 75201 U.C '.
OdLL
ATTACHMENT 3
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APPENDIX U.S. NUCLEAR REGULATORY CCMMISSION OFFICE OF INSFECTION AND ENFORCEMENT REGION IV Report:
50-445/81-15; 50- 4 6/81-15 Occkets:
50-445; 50-446 Category A2 Licensee:
Texas Utilities Generating Cc=pany 2001 Bryan Tower Dallas,-Texas 75201 Facility Name:
Comanche Peak, Units 1 and 2 Inspection at:
Comanche Peak Site, Glen Rose, Texas-Inspection Conducted: October 13-16 and 19-23, 1981 f
i Inspector: /- ' f /-
/
e uh-/N Date C. i. Jonnsen, Reactor inspec or, Engineering anc Materials $3ction other Accomcanying R. E. Hall, Acting Chief, Engineering and Materials Section Personnel:
(October 23, 1981)
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Approvad: /
Date gt R. i. nai Acting Calef, ingineering anc Material. Section r
i Insoection Summary 50-345/31-15:
Insoection Conducted October 13-16 and 19-23, 1981 (Recort No.
50-446/81-15)
Routine, unannounced inspection of construction activities Areas Insoected:
incluaing a site tour and observation of work and review of records for safety-related pipe su ports, protective coatings, and restraint systems in The inspection involved sixty-five inspector-hours by one NRC Units 1 and 2.
inspector. In the three areas inspected, no violations or deviations were Results:
One violation was identified in the area of icentified in two areas.
protective coatings (violation - failure to follow procedures for the inspection of coatings - paragraph 3.b).
ATTACEIENT 3
UETAILS 1.
Persons Contacted Princical Licensee Personnel
- 0. N. Chapman, Texas Utilities Generating Ccmpany (TUGCO), Quality l
Assurance Manager "R. G. Tolson, TUGCO, Site Quality Assurance Supervisor S. C. Scott, TUGCO, Quality Engineering a
Other Personnel "J. R. Merritt, Texas Utilities Services, Inc., (TUSI), Engineering and Construction Manager J. V. Hawkins, Brown & Root (S&R), Project Quality Assurance Manager H. Williams, S&R, QC Superintendent, Non-ASME J,. Ragan, S&R, Site QC Manager The NRC inscector also contacted other licensee and Brown & Root employees during the inspection perioc including botn craft labor and QA/QC personnel.
- Denotes presence at the exit interview conducted October 23, 1981.
2.
Site Tour The NRC inspector toured the site several times during the insoection to observe construction activities in progress and to inscect hcusekeecing.
The areas toured included the Units 1 and 2 Reactor Containment Suildings, Safeguards Building, Auxiliary Building, paint shep, and paint wareneuse.
There were no violations or deviations identified.
3.
Protective Coatino l
a.
Concrete Coatincs Acclication 4
The NRC inspector reviewed six records of protective coating application on concrete in Reactor Containment Building No. 2.
The NRC inspector identified some procedural inconsistencies l
including those listed below:
(1) no inspector initials or signature i
(2) incorrect pot life specified or indicated i
(3) dates missing (4) induction time signed N/A when there should have been a minimum time listed j
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a ATTAC. HINT 3
)
In fcur of the six records cbserved, the finisn c:at had been applied; hcwever, checklists documenting the preceding acclica:icns had neither been c mpleted nor finally accepted.
The majority cf the inspecticn dccumentatien had not had final ac:sp;ance.
The NRC inspect:r informed the licensee representative of the discrepancies fcund. The licensee representative then informed the NRC inspector that they were aware of the problecs. After a brief discussion, the licensee representative shcwed the NRC inspector an audit report performed during September 14-18, 1981, by TUC',0 QA pertaining to this matter. The audit reocrt was titled, " Protective Coatings," QA Audit File: TCP-24 This audit identified the same deficiencies as did the NRC inscector. The sc:pe of the audi:
included receiving, storage and handling, certifica:icns, measuring and testing-equipment, application, and inspecticn dccumentaticn regarding protective coatings en c:ncrete. This audit recuested a response to the deficiencies and concerns identified by Novemcer 6, 1981.
j This matter will be listed as an unresolved item until a satisfact:ry reply to the QA Audit, TCP-21 has been documented in a timely manner and reviewed by an NRC inspect:r en a subsecuent inspecticn.
b.
Steel Coatines Acolicatien The NRC inspector also reviewed records en protective coating a:pli-cation en steel.
During the review cf these records, tne NRC inspector noticed that the most current recorcs for miscellanecus steel and suppcrts dated fr:m late Septemcer 1979 and back.
The NRC inspect:r rsviewed cne package which contained several rec:rds' j
of miscellaneou: embedments and decr frames in Reac:cr Building No. 2.
This package cer.tained d:cuments that were ne: preperly filled cut or were incemclete.
For example, dry film thickness (DFT) of primer after application was not recorded and initialed by Quality Centrol inspectors en the che'cklist prior to a:plicaticn of the seal ::a:
as required. The NRC inspector also noticed incorrec po life require-ments en the protective coating material identifica:icn and mixing checklist. The licensee representative was informed of the defi-ciencies noted by the NRC inspect:r.
The NRC inspector inquired abcut records of inspecticn c:nducted since September 1979.
The licensee representative inf rmed the NRC inspector that frem late September 1979, there had been no checklists maintained as required by the precedures for miscella-neous steel, cable tray supports, and pipe suppcrts noted in paragraph 3.f belcw. The licensee representative presented a leg back that Quality Centrol was maintaining en cable tray i
supports. The licensee recresentative also informed the NRC inspector that they have maintained a log bcck en miscellaneous steel, pipe hangers, and supports.
After reviewing the icg bock, the NRC inspector determined that this did not appear to satisfy i
- - - _.,,.,,,,. -.,,.. _,,., _ _,,., - -., - ~ - -
n
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ATTACsiENT 3 the recerts recuirements of the pr cecures.
The NRC inspector also reviewed pr0:ec-ive c:a:ings records of Unit 2 Reac: r Building steel liner plate, fr:m September 1950 to Oc::ber 1950.
In reviewing these rec rds, the NRC ins ector noticed that documentaticn en the " Steel Substra:e Primer.acclica:icn Checklist" nad nc been c:: letely filled out pricr to seal c:a: applications as required by precedure.
This appears to be in violaticn of precedural requirements and 10 CFR 50, Appendix B, Criterien V.
c.
Paint Shco The NRC inspec:cr toured the paint shco and reviewed coating receres in process by a Quality Centrol inscector.
All rec:rds reviewed at the paint shop apfear to c:nform to Prececures QI-QP-ll.4-1, QI-QP-11.4-2, and QI-QP-11.4-3 as listed in paragrach 3.f.
The NRC inspector discussed procedural requirements w1:n the Quality Cen rel inspecter in charge of the paint shcp.
The Quality Centrol inspect:r demenstrated an a: parent level cf kncwledge sufficient Oc inspect protective coating en steel.
The NRC inspect:r also discussed precedural requirements and qualiff-ca:icns for a:plicaters with a paint fereman.
The paint foreman also a:: eared to have the level of kncwledge required for protective coating applica:1cns.
During the : ur, the NRC inspector c: served scme sandblasting taking place; hcwever, there was no c:ating a: plica icn going en at :ne time en safety-related items.
There were no violatiens er deviatiens identified.
i d.
Pain: Warehcuse Sterace The NRC inspec cr t ured the "0" paint wareneuse and randomly inspec ed for cpened centainers and expired dates for shelf life of c:ntainers.
There are two warehcuses, ene a ncn "Q" and the other a "Q'" warehcuse.
The NRC inspector verified the readings cf the temperature recorder which varied between 70 degrees and 75 degrees.
The temperature was well within the specification.
The temperature recorder calibratien das
- was current.
The NRC inspector reviewed recorcs of tne tem:er-ature charts and the surveillance checklist for st: rage and c:ntrol of ceatings. The temperature charts ccvered frem January 1,1981, to August 31, 1981, and the temperature was maintained wf:hin the sceci-fied range.
The checklists were filled cut and signed.
The enly i
discrepancy notec en the checklist was in the block " shelf life exoired", which had been checked in the "yes" blocks. After inquiring into this matter, it was found to be in errer since the shelf life had not been exceeded. This was an isolated case which was immediately corrected.
There were no violatiens er deviations identified.
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ATTAC.TEr 3 j
Cualification Records (Coatincs) e.
The NRC inspector reviewed qualification recorcs for six Quality i
Control inspectors acc ten painters.
The Quality Control inspectors and painters selected for review were indicated on tne recorcs of six safety-related items wnica had been paintec anc inspected.
The NRC inspector traced the qualifications of eaca Qualitiy Control inspector and painter on the records during that time period to verify each individual's qualifications.
All Quality Control inspectors and painters were qualified during the period of applications.
There were no violations or deviations identified.
f.
Review of Coatinc Procedures 1
The NRC inspector reviewed the coating procedures listed below.
All procedures reviewed appeared to include sufficient instructions and appropriate quantitative and qualitative acceptance critaria for determining that important activities would have been satisfactorily accomplished.
Procedures reviewed:
Steel CP-QP-il.4, Revision 4, " Inspection of Protective Coatings" QI-QP-11.4-1, Revision 3, " Inspection of Steel Substrate Surface Preparation" QI-QP-ll.4-2, Revision 4, " Inspection of Steel Substrate Protec-tive Coating Mixing Operations" QI-QP-11.4-3, Revision 5, " Inspection of Steel Substrate Prime Coat Applications" QI-QP-il.4-5, Revision 4, " Inspection of Steel Substrate Finish Coat Application" i
l QI-QP-11.4-17, Revision 1, " Storage and Handling of Protective Coatings" Concrete QI-QP-ll.4-10. Revision 1, " Inspection of Concrete Substrate Surface Preparation" QI-QP-ll.4-12, Revision 1, " Inspection of Concrete Substrate Surface Application" i
1 I
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ATTACE'4ENT 3
)
a QI-QP-11.4-14, Revision 1, " Inspection of Concrete Substrate Finisn Coat Applicatien" There were no violations or deviations icentified.
4.
Safetv-Related ? ice Succort and Restraints a.
Mcment Limitina Cccconents The NRC inspector teured Unit 1 Reacter Containment Building and randemly selected three mcment restraints listad belcw for inspection:
(1)
S-1-0538 Safety Injection System (incomplete)
(2) 5-1-0538 Safety Injection System (ccmplete)
(3) S-1-0538 Chemical and Volume Control System (c mplete)
The NRC inspector visually inspected the completed mcment restraints in accordance with referenced design drawings.
Welds, attachments, and location all appeared to be in compliance with design crawings.
The NRC inspector reviewed the traveler package contents on work being performed in Unit 1 Reactor Containment Building, on tne Safety Injection vertical line restraints in each of the four steam All documentation a:peared to be in the generator compartments.
All changes and modifications had traveler package as required.
been approved.
The mcment restraints reviewed were all faoricated by an off-site The constructor erected, aligned, and welded as required vendor.
by the cesign drawings.
The records of the two completed restraints indicated that verifi-cation of elevation, location, plumb, and levelness had been verified by Quality Control.
All welds were in accordance with design drawings and verified by Quality Control.
Comoletion of all operations were verified, signed, and initialed by Quality Control.
The NRC inspector verified welder qualifications from the completed travelers and also for work in process in the field.
All welders were fcund to be qualified by precedure for the work being performed.
b.
pice Succorts The The NRC inspector reviewed the records of five pipe supports.
The records indicated the type and classification of the suoports.
records confirmed that the specifications and installation precedures Weld The required scope of QA/QC inspections was met.
had been met.
identification and location corresponded to respective weld card, i
drawing, and work crder.
Welders were qualified for the welding procedures used.
1
y ATTACFS. INT 3 Records reviewed were for the pipe se:: orts listed below:
CT-1-029-018-C325 - (C:ntainment Spray System)
CT-1-014-421-C52R - (Containment Spray System)
CT-1-031-007-C92R - (Containment Spray System)
C5-1-074-047-542R - (Chemical and Volume Control System)
CS-1-074-041-542R - (Chemical and Volume Control Systam)
There were no viciations or deviations identified.
5.
Unresolved Items Unresolved items are matters a: cut which more information is required in order to ascertain wnether they are acceptacle items, violations, or deviations.
An unresolved item disclosed during this inspection is dis-cussed in paragrach 3.a.
5.
Exit Interview The NRC inspector met with the licensee recresentatives (denoted in paragrapn 1) and R. G. Taylor, NRC Resident Reactor Inspector, on October 23, 1981, and summari:ed the purpose, scope, and findings of the inspection.
t s
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ATTACFJ.ENT 4
/.g anoi%q uNitro STATES RECEIVED NuctrAa RscVLATORY COMMISSION 2 bg,i c'
ancieNiv ggy1o;g7c ge g
611 RYAN PLAZA DRIVE SUITE 1000 o
ARLINGTCN, TEXAS 76011 e
f NUO:. EAR CIV.
May 14,1979 m
In Reply Refer To:
W RIV Cocket No. 50-445/Rpt. 79-11
~
50-446/Rpt. 79-11 4
Texas Utilities Generating Company t
ATTN:
Mr. R. J. Gary, Executive Vice President and General Manager 2001 Bryan Tower Dallas, Texas 75201 Gentlemen :
This refers to the investigation conducted by Mr. R. G. Taylor and other members of our staff on April 2-3 and April 13-23, 1979, cf. activiti es authorized by NRC Construction Per:11ts No. CPPR-125 and 127 for the 1-Comanche Peak facility, Units No.1 and 2, concerning allegations by a former Comanche Peak employee.
The investigation and our findings are discussed in the enclosed investigation report.
During the investigation, it was found that certain activities under I
your license appear to be in noncompliance with Appendix 3 to 10 CFR 50 of the NRC Regulations, " Quality Assurance Criteria for Nuclear m.
Power Plants." The item of noncompliance and references to the per-tinent requirements are identified in the enclosed Notice of Violation.
This notice is sent to you pursuant to the provisions of Section 2.201 of the NRC's " Rules of Practice," Part 2, Title 10, Code of Federal Regulations.
Section 2.201 requires you to submit to this office, within 30 days of your receipt of this notice, a written statement or explanation in reply including: (1) corrective steps which have been taken by you, and the results achieved; (2) corrective steps which will be taken to avoid further noncompliance; and (3) the date when full compliance will be y
achieved.
d In accordance with Section 2.790 of the NRC's " Rules of Practice," Part 2, Title 10, code of Federal Regulations, a copy of this letter and the d
enclosed investigation report will be placed in the NRC's Public Document Room.
If the report contains any information that you believe to be
ATTACIDE:IT 4 Texas Utilities Generating Company gay p, )c79.
q oroprietary, it is necessary that you submit a written application to D'I+]
this office, within 20 days of the date of this letter, requesting that 2.->
such infonnation be withheld frcm public disclosure.
The application i
must include a full statement of the reasons why it is claimed that the
- ]a information is proprietary.
The application should be prepared so that any proprietary information identified is contained in an enclosure to the application, since the application without the enclosure will also be placed in the Pubite Document Room.
If we do not hear from you in this regard within the specified period, the report will be placed in the Public Document Room.
' Should you have any questions concerning this investigation, we will be pleased to discuss them with you.
Mg Sincerely, J
n Q. m/L
~
W. C. Seid
, Chief
~ Reactor Coirstruction and Engineering Support Branch
~
r Enclosures :
- { ;
1.
Appendix A, Notice of Violation 2.
IE Investigation Report No. 50-445/79-11 50-446/79-11 hj cc: w/ enclosures E
Texas Utilities Generating Company ATTN: Mr. H. C. Schmidt, Project Manager i
2001 Bryan Tower i
Dallas, Texas 75201 I
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ATTACHMznT 4
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50-445/79-11 50-446/79-11 Appendix A NOTICE OF VIOLATION grii Based on the results of the NRC investigation conducted during the periods
~
April 2-3 and April 13-23, 1979, it appears that certain of your activities were not conducted in full compliance with the conditions of your NRC
,_ e.
Construction Permit No. CPPR-126 as indicated below:
Failure to Imolement the Ouality Assurance Procram For Civil I
Construction 10 CFR 50, Appendix B, Criterion II requires that a quality assurance I
~
program be established and implemented for the construction of the I
structures important to safety of the nuclear plant.
The Texas Utilities Generating Company Comanche Peak Steam Elactric Station Quality Assurance Plan affirms the intention to fulfill this require-
"' if ment.
The CPSES " Civil Inspection Manual" provides a body of r4 inspection and testing procedures required to implement the Quality
)gj Assurance Plan.
3 Contrary to the above:
On January 18, 1979, personnel of the civil construction labor force placed an undetermined amount of concrete of an unknown quality on the dome of the Unit 1 containment without the knowledge of your Quality Assurance organization and without
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benefit of required inspections and testing of the concrete.
This is an infraction.
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ATTACEME:IT 4
'd
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U. S. NUCLEAR REGULATORY CCMMISS,10 L OFFICE OF INSPECTION AND ENFORCEMENT REGION IV
.L Report No. 50-445/79-11; 50-446/79-11 Docket No. 50-445; 50-446 Category A2 Licensee: Texas Utilities Generating Ccmpany 1
2001 Bryan Tower Callas, Texas 75201 Facility Name:
Comanche Peak, Units 1 & 2 l
Investigation at:
Comanche Peak Steam Electric Station, Glen Rose, Texas
_I
~3 Investigation Conducted: April 2-3 and April 13-23, 1979
[5 8[/4[7f Inspectors:
2n-
. G. Taylor, Resicent Reactor Inspector, Projects Date Section obf j
- 0. P. Tomlinson, Reactor Inspector, Engineering Date/
Support Section (April 13, 1979, Interview )
f okf
[A. B. Beach, Reactor Inspector, Engineering Date /
b' Support Section (April 23, 1979, Inter /iew)
M Approved:
w_ --
/d[7f C.t W. A. Crossman, Chief, Projects Section Date
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R. Ee' Hall, Chief, Engineering Support Section Date /
ATTAC:9ENT 4 Investigation Suntary:
Investication on Acril 2-3 and Acril 13-23, 1979 (Recort 50 445/79-11; 50-446/79-11)
! p.gf}
Areas Investigated:
Special investigation of allegations received indi-l cating that concrete-had been placed on the Unit 1 dome during a rainstorm 2,=.
in January 1979, without QC or documentation; that pipe with sandblasted-off markings was being used in Unit 1; that steam system pipe was damaged
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by a handling accident and covered up; and that welders were not being properly qualified.
The investigation involved thirty-six inspector-hours y
by the Resident Reactor Inspector and three inspector-hours by two Region I
IV based inspectors.
Resul ts : The allegation relative to the concrete placement was confirmed l
(nonccmpliance - failure to implement the QA program - infraction).
No I
items of noncompliance or deviations were identified relative to the I
balance of the allegations.
3
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INTRODUCTION Comanche Peak Steam Electric Station (CPSES), Units 1 and 2, are under construction in Somerville County, Texas, near the town of Glen Rose, Texas.
w Texas Utilities Generating Company is the Construction Permit holder with Brown and Root, Inc., as the constructor and Gibbs and Hill, Inc., as the Architect / Engineer.
j REASON FOR INVESTIGATION I
The Region IV Reactor Construction and Engineering Support Branch office received a telephone call from a fomer CPSES employee who reported several allegations indicating a potential breakdown in the CPSES Quality Assurance program.
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SUMMARY
OF FACTS J
.3 7-On March 30, 1979, the Region IV Reactor Construction and Engineering Support Branch received a telephone call frem a party who identified himself as a fomer CPSES employee.
The call was taken by an on-duty Reactor Inspector in the Projects Section who in turn provided the information to the assigned Resident Reactor Inspector at CPSES on i
April 2,1979. The allegations, as received on March 30, 1979, were:
1.
During a concrete pour on the Unit l containment dome in January 1979, a rain occurred which washed away part of the concrete.
The affected area was repaired by the use of grout.
Workers involved were requested to " keep it quiet." Two workers, who j
are still at the site, have knowledge of this occurrence.
I 2.
The identity of a lot of "Q" and "non-Q" pipe (6" or less) being used for Unit 1 has been lost due to obliteration of heat numbers I
by sandblasting and loss of identifying tags. Workers are guessing l
as to the proper identification of the pipe.
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3.
A steam pipe intended for the Unit 1 turbine fell 'off of a truck and struck a railroad track.
It was taken back to a storage area and hidden.
4.
Third class helpers are being qualified in less than three months Ll and are being used for safety related welding on Unit 1.
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On April 13, 1979, the Pesident Reactor Inspector assigned to CPSES and accompanied by another Region IV inspector interviewed the alleger in an effort to obtain additional information on the allegations.
The additional infomation is summarized as follows:
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The concrete used for the. repair was not grout as originally indi-cated but was known to contain gravel.
The concrete came frem the batch plant where it was mixed on the ground and carried in a bucket j
to a tower crane at the Unit 1 Containment Building and hoisted to j
the deme area.
The work was accomplished sometime during the middle of the second shift, possibly around 10:00 to 10:30 p.m. (January 1979, no day specified).
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i 2.
The pipe in question was not prefabricated pipe but rather bulk pipe joints.
Sometimes, the pi occurrence not identified)pe'is sandblasted on the outside (rate of which removes all of the heat marking used for traceabil_ity.
3.
The steam pipe was being moved during the second shift from the "Dodd's Spur" storage area to the plant area when it was dropped f,- s off the truck.
A couple of the large " cherry-picker" type cranes were dispatched to the indicent to pick up the pipe and place it back on the truck.
The crew with the truck decided instead to put the pipe back into the storage area and leave it there for another shift to pick up and perhaps be blamed for dama5ing the pipe.
The alleger did not know if the pipe had actually suffered any damage.
He was aware the pipe in question was "non-Q" but
' 7' expressed a concern that if the craft could get away with a cover-Q up on "non-Q," they probably are also doing it on the "Q" pipe and other equimpment.
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The alleger indicated he was concerned with what must be incompetent Mi welders working on "Q" welds, since they could not have very much i-experience and still only be considered third class labor.
j CONCi.USIONS 1
Research of various records and interviews with both craft labor and E_.
i Brown & Root QC personnel produced the following conclusions:
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1.
The allegation relative to the concrete placement on the deme of t
Unit 1 is essentially correct and is evidence of a breakdown in the licensee's Quality Assurance program.
The incident will be i
considered an item of noncompliance.
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2.
The allegation relating to the loss of pipe tFiceability markings could not be confirmed.
The Resident Reactor Inspector's finding was that on occasion the sandblasting, with attendant loss of readily i
visible markings, probably does occur through human error, but that there are other means which will re-establish the identity of the i
pipe without guessing on the part of the craft labor force.
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3.
The piping in the "Codd's Spur" storage area is for the turbine portion of the plant and is not safety related from a nulcear
'n951 standpoint and is therefore not within the jurisdiction of the NRC inspection program.
The more generalized concern of cover-up of improper handling practices is not consistent with the obser-vations of the Resident Reactor Inspector and other NRC inspectors made during the course of routine inspections.
The allegation cannot be verified or refuted at this time, but should subsequent observations verify that the alleged situation is occurring,
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appropriate action will be taken.
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Welders are qualified in accordance with the provisions of the t
r; ASME Boiler and Pressure Vessel Code,Section IX, " Welding and Brazing Qualifications," as required by NRC regulations and the ij?!
licensee's commitments as contained in the Safety Analysis Recort e
submitted to obtain a Construction Permit.
The labor classifica-tion, and therefore the pay, of the welders is not an element of the ASME Code welder qualification program, enly the ability of the person being tested to weld on a specified weld coupon.
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, DETAILS
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1.
Persons Contacted
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Non-Licensee or Contractor Dersons
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The a?ieger is a former employee of Brown & Root (the site general
' contractor).
The person identified himself as a former equi;; ment Fj operator and foreman of equipment operators.
'PrincioalLicenseeEmoloveks
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Construction Manager, Texas Utilities Generating Co.
. 'e Supervisor cf Product Assurance, Texas Utilities Generating Co./
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Gibbs:& Nfil
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Brown & Root, Inc.
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Project General Manager
- c Construction Project Manager F
General Foreman, Building Department
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Superintendent, Building Decartment Quality Control Inspector, Civil 5-2.
Prel'iminary Investication Aoril 2-3 1979
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Allegation 1: The Resident Reactor Inspector (RRI) initiated ls a preliminary investigation of the allegation as soon as received.
The RREwas aware that a number of concrete place-ments had been necessiry!to' ccmplete the dcme area of Unit 1
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JJ62 and that a substantis) portion of these placements occurred in January.1979.
Schedule completion data indicated that five of the total of thir cen doce placements occurred in January 1979.
Rainfall data for January was then obtained from the licensee's meteorology unit which' indicated rain had fallen r.
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on Janaury 15,1979 (with the rainfall totalizer reset to zero) and again in the period between January 15 and 22,1979,
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when the totalizer was again zerced.
The data suggested that M
placement 101-8805-013, the final. placement on the doce, was the most likely candidate since 2.72 inches of rain had occurred l
about the placement date cf January 18, 1979.
The RRI then i..
examined the QC inspection records for the placement which stated,
- " Pour stopped at 8:00 p.m.1/18/79 due to incle.ent weather.
Pour was topped out all but to' a 20' radius which was cleaned up and finished 1/19/79."
i The RRI then interviewed the QC inspector of. record for the
.p'atement and was informed that the placement had started
- . under good weather conditions on January 18, but that the e
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} ATTACEME: T 4
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weather subsequently developed into a li$it mist and dri::le which did not interfere with the placement.
By late evening, the weather deteriorated further and became a full rainstorm with thunder and lightning.
By 7:30 p.m. or so it was decided that the placement would have to be stopped for reasons of A
personnel safety.
The placement area was covered to keep the rain off the fresh concrete and the second shift was instructed g
to water blast and clean up the area so the placement could be 4
resumed the following day.
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Allegations 2, 3 & 4: No attempt was made to perform a pre-i liminary investigation of these allegations since the infor-i mation was too vague.
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Licensee / Contractor Recort of A11ecations
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I During the course of the above preliminary investigation, personnel of the licensee's management and QA organizations approached the
-- 1 RRI and stated that they too had received an allegation relative to the dome placement.
It was stated that licensee management had received a telephone call on or about March 19, 1979, on the subject
,,j and that licensee management had visited the alleger at his home on j
March 20, 1979, to ascertain the facts of the allegation.
The alleger then was invited to visit the site and discuss the allegation, which the alleger is reported to have done on March 25, 1979.
On the basis of these interviews, the licensee's product Assurance personnel under-took an investigation which concluded that the allegation had no merit.-
~ 4. Interview with Allecer by NRC Personnel r! The Region IV office made several attempts to establish contact with the alleger during the period following March 30, 1979, when the Q; allegation was received, through April 12, 1979, when the interview p date and location were established. The RRI and another NRC inspector met with the alleger and a friend on April 13, 1979. The alleger provided the following 'infarination about himself: a. He had been employed by Brown & Root at CpSES for 2-1/2 to 3 years and had quit in mid-March because he was dissatisfied with how the night shift equipment operators ~were being .i dispatched and supervised. b. He had been an equipment operator, primarily on cherry-pickers, g and also a foreman for equipment operators at an earlier time. 2.,. 4 i 1 2 l..
ATTACEMINT 4 He stated that he had made the~ allegations to licensee management c. and Brown & Root management earlier but had not been at all sat-isfied with the answers he had received to his allegations. The alleger provided the following additional information relative to , n..) each of the allegations: ^= J J Allegation 1: The incident occurred well after the time that ~ the placement had been stopped. He could not be sure of the time but thought it was probably 10:00 to 10:30 p.m. when scme i equipment was dispatched to the concrete batch plant to bring . ~ dcwn a bucket of concrete to Unit I and thought it strange. The concrete was taken to the dcme by a tower crane. He was sure i that the concrete,was not batched by the batch plant and certainly } was not delivered by the usual concrete mix truck. Allegation 2: The alleger made it clear that he was not referring to completed pipe spools but rather to bulk pipe. The cherry-i picker operators routinely move the pipe frem one location to another on the site and that the pipe involved was bulk pipe or joints. He stated that the pipe was sometimes sandblasted in such a way as to obliterate the heat number markings or tags and that he was pretty sure that there was a lot of unidentified pipe in ~* the safety systems in Unit 1. This sandblasting sometimes happened to various steel forms used to make supports. Allegation 3: The alleger described being dispatched with his equipment out to "Dodd's Spur" to pick up a length of pipe that had fallen off a truck after being loaded. The pipe had fallen a on the spur railroad track. The RRI was not familiar with the q term "Dodd's Spur." The alleger stated it was the area were the 1 turbine components are stored. When he (the alleger) arrived at 3 the site of the incident, he was told not to reload the pipe on the truck but to take it back into the storage area and put it down. The pipe crew indicated to him that they hoped that a day shift crew would come for the pipe and would probably be blamed for any damage that might have occurred to pipe when it fell. He stated that he did not know if the pipe had been damaged. He i stated that he knew it was "non-Q" pipe but thought the NRC should l be aware that such things were going on at the site. 5. Final Investication - Acril 16-23, 1979 a. Allegation 1. The RRI obtained the craft labor time sheets for both shifts for January 18 and 19,1979. Review of the time ~ sheets for the day shift on January 18 indicated that a portion of that shift worked on placement 101-8805-013. The records indicated that the day shift was tenninated at approximately _a-A
ATTAC'OE:IT i d 8:30 p.m. relative to the placement as Ire the personnel at the concrete batch plant. The batch plant has no second shift operators. The RRI found that a large number of people, well in excess Of fifty, had then worked on the placement during a substantial portion of the "=lC second shift. One crew of twelve people was shown by the time sheets ~ to have been placing concrete, a notation not consistent with the 4 fact that the batch plant was closed during tha shift. The RRI then ,1 utilized the time sheets to develope a list of persons to be inter-I viewed in connection with the incident with special concentration on the persons listed on the time sheet indicating " placing concrete 101-8805-013." The B&R personnel office records indicated that eight of the ten names included in this specific crew had been ~ l terminated at various times since January 18; the records did not suggest that any action was being taken to get. rid of possible confinnatory personnel. Late on April 17, 1979, two of the senior B&R construction manage-fmg ment personnel very informally asked the RRI how the investigation y of the allegations was coming along'. The RRI responded that the j on-site phase appeared to be complete and that NRC personnel would undertake the effort to locate and interview selected personnel '~T irrediately since it appeared that the allegation might be well ~ founded. They asked the RRI if they could check with their people down to the General Foreman level as to the incident the night of ' January 18. The RRI indicated that such an inquiry on their part would probably not interfere with any future investigative action by the NRC. On April 13, 1979, the licensee's Product Assurance Supervisor informed the RRI that he had information which indicated that the incident had occurred and that the craft General Foreman was the y person responsible. On April 23, 1979, the RRI, accompanied by another NRC Inspector, interviewed the General Foreman and his ir=ediate supervisor, the night shift B&R Building Department Superintendent. These men related that on the night of January 18 the weather.seemed to worsen i and got to the point where the rain was so heavy that the people could hardly see. The freshly placed concrete developed into a problem when the plastic cover could not take the rainfall water K- ! load. Some of concrete began to sag back down the dome slope and - d one sma;l area actually washed out and fell to the ground below. These men related that they and their entire crew of up to about one hundred-fifty worked on into the night trying to save a very bad situation. The sagged concrete was worked back into position and the crew protected it in any way they could to allow it to take a set. _g. I
ATTACEMDT 4 ( t ... I The General Foreman went to the batch plNt, got it open and operated the plant himself to make enough material to patch the washec out area. He stated that he found the design mix data used for the concrete on the deme and calculated the necessary C-= E weight of ingredients to prepare a half a cubic yard of concrete. The required data was put into the control system for the back-up ory batch plant, dropped into a skiff, and carried over to the w quarter yard concrete mixer at the site test laboratory. It was mixed in two batches and placed into a skiff and carried to the 9 dome where most of the half yard was used as a patch in the washed om area. Both the General Foreman and his Superintendent were aware that i there were no Quality Control personnel around to observe any of these actions since they had all gone home when the weather got j really bad. Both men related to the RRI a picture of almost panic proportions in which the presence or absence of Quality Control simply did not matter; they were going to save a concrete place-4 ment from what they considered a disasterous situation, regardless. They indicated that while the night shift Assistant Construction 4 project Manager was generally aware of the situation on the dome i 1 that night, he probably was unaware of the fact that Quality Control personnel were not there or of the batching of the concrete under the conditions indicated. In response to a question from the General Foreman as to "what _._.l happens now" the RRI stated that the NRC had no choice but to issue a Notice of Violation to the licensee since it had become very clear that the licensee's Quality Assurance program had broken down for the entire evening of January 18, 1979, and that a substantial amount of concrete on the deme was of an unknown quality. ._ r b. Allegation 2. The RRI visited the paint shop sandblasting area during the course of the final investigation to ascertain if 'this allegation could reasonably happen. The RRI interviewed a foreman of painters who is also in charge of the sandblasting activity and was told that three main categories of piping material routinely are sandblasted. These are: 7 (1) Completed carbon steel spool pieces which are blasted on f the outside prior to painting. The identity of these pieces is on an attached stainless steel band on which the identifying is encoded by stamping. Should the band M come off, the spool piece identity can be re-established by the pipe fabrication shop since each spool is unique i and is fully described by isometric drawings.. w eg.msy.e-.e + = ag e.e t =
~ i ( g ') ATTACF.MIr ; (2) Carbon steel cut lengths, but otherwise in an unfabricated condition, are sent to sandblasting to have the insi& cleaned prior to further fabrication. The outside, which usually carries the heat marking in paint is supposed to be untouched. . -- g (3) Bulk carbon steel pipe materials used for making equipment e stands and supports is blasted and painted prior to fabrica-n tion. D The material is used for such items as instrument supports. D-The RRI found a number of examples of each of the above categories as well as steel shapes in the sandblast area. During the tour of the area, the RRI did not find any matvial that could not ba identified except that in category three. The RRI interviewed one of the sandblasting personnel and came to the conclusion that the person might make an occasional mistake on category 2 material since he seemed confused when asked what he was going to do with a number of pieces ready for him to work on. It appeared that he might well blast the outside of a pipe when he should blast the ..y inside. ~ Subsequent discussions with the paint shop foreman and with a 1 n. Brown and Root Quality Control inspector in the pipe fabrication shop revealed that all cut, but unfabricated material, is trans-ferred to the paint shop by memo which details the size, schedule and length of the cut section and the pipe spool isometric drawing involved. Should the outside of the pipe be inadvertently blasted, the piece can be reidentified relatively easy by measuring its size, j schedule and length. The isometric drawing used to make the cut length is annotated with the pipe heat number prior to the cutting ~ ! operation and verified by QC. It appeared most unlikely to the RRI that two otherwise identical pieces but with different heat I numbers would be inadvertently blasted within the same time period. The RRI concluded that the allegers remark that " workers are l guessing on the identity of pipe" might be true, but that there was an adequate cross-check system built into the quality assurance program to preclude untraceable pipe from being installed in the safety related systems. All of the steel shapes used in safety related supports for pipe ".d and cable tray that have been examined by the RRI and other NRC O inspectors have been sufficiently marked to establish their origin. J These materials are.also subject to a system of quality control verifications at various stages of fabrication sufficient to make "'I it very unlikely that any improperly identified or unidentified material is used and installed. . i.
-) ATTACEMENT 4 i / c. Allegation 3: Based on the interview with the alleger, no further action was taken to investigate the specifics of the allegation since the pipe in question was clearly not safety related and therefore not within the jurisdiction of the NRC inspection pro-gram. The more general concern that the pipe. handling incident was a possible indicator of the general attitude of the craft personnel, particularly the riggers and pipefitters, appeared to be unfounded. The RRI has cbserved durin geyj the past nine months (since August 1978) g many plant tours over that the material hand-t ling activities of the craft personnel have been acccmplished under well controlled conditions in so far as they relate to safety related equipment and materials. An allegation of possible j cover-up of improper actions by the craft personnel in behalf of I other craft personnel is almost impossible to either confirm or completely refute. d. Allegation 4: No further investigation was made into the charge that third class welders are being used to perform safety related 2 pip'ing system welds on the basis that the welders are all qualified under a program prescribed by the ASME Boiler and Pressure Vessel Code Section IX, " Welding and Bra:ing Qualifi:ation." The applica-tion of the Section IX program has been reviewed a number.of times ~ by the RRI and other NRC inspectors since it was implemented at CPSES. The implementation has been found to be consistent with the requirements. These requirements, however, do not address themselves to the experience or inexperience of the person seeking qualification as a welder, but rather to whether he can accomplish a weld in'one or more of the Code prescribed positions that will pass the test criteria imposed by the Code. The terminology " third ~ sq: class," as it applies to the labor force, relates primarily to the )i pay category in which a person is hired and previous experience is a factor in this determination.
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~p. ATTAC*4'dr;; 5. 4 UNITED STATES /pa *8 c,e, o NUCLEAR REGULATORY COMMISSION ' 7., R EGICN IV .,- Q M j C 611 RY AN PLAZA oRIVE,5UITE 1000 g r. e f.hyd 2 ARUNCTCN, TEXAS 76012 %*4,,,. 1,J Nove=ber 27, 1979 NOV2 31979 e R. J. GAR ie In Reply Refer To: RIV i Docket No. 50-445/Rpt. 79-24 50-446/Rpt. 79-23 RECElyED Texas Utilities Generating Cc=pany ATTN: Mr. R. J. Gary, P.xecutive Vice President and General Manager MC0 (:. 2001 Bryan Tower Dallas, Texas 75201 M LJ: Gentle =en: This refers to the inspection conducted by our Resident Inspector, Mr. R. G. Taylor, during the period of October 1979 of activities authorized by NRC Construction Per=1ts No. CPPR-126 and 127 for the Cc=anche Peak facility, Units No. 1 and 2, and to the discussion of our findings with Mr. R. G. Tolson and other = embers of your staff during the inspection period. Areas exa=ined during the inspection and our findings are discussed in the enclosed inspection report. Within these areas, the inspection consisted of selective examination of procedures and representative records, interriews with personnel, and observations by the inspector. Within the scope of the inspection, no ite=s of nonco=pliance were identified. We have also exa=ined actions you have taken in regard to a previously identi-fied finding. The status of this ite: is identified in paragraph 2 of the enclosed report. Two new unresolved ite=s are identified in paragraphs 2 and 7. In accordance with Section 2.790 of the NRC's " Rules of Practice," Part 2, Title 10, Code of Federal Regulations, a copy of this letter and the enclosed inspection report will be placed in the NRC's Public Docu=ent Roo=. If the report contains any infor=ation that you believe to be proprietary, it is necessary that you sub=it a written application to this office, within 20 days of the date of this letter, requesting that such infor=ation be withheld fro = public disclosure. The application =ust include a full statement of the reasons why it is clai=ed that the infor=ation is proprietary. The application should be prepared so that any proprietary infor=ation identified is contained in an enclosure to the application, since the application without the enclosure
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_ =. ATTACHMENT 5 Texas Utilities Generating Conpany Neve=ber 27, 1979 will also be placed in the Public Decu=ent Roc =. If we do not hear frc= you in this regard within the specified peried, the report will be placed in the Public Document Roc =. Should you have any questions concerning this inspection, we will be pleased to discuss the= with you. Sincerely, \\ / fJ. C. Seidle, Chief Reactor Coh truction and Engineering Support Branch J
Enclosure:
IE Inspection Report No. 50-445/79-24 1 40-446/79-23 cc: w/ enclosure Texas Utilities Generating Cenpany ATTN: Mr. H. C. Sch=idt, Project Manager 2001. Bryan Tower Dallas, Texas 75201 D.W OL / l ps: \\ tD.?.N<. W o l I l P I i 9 i ,,---,---~_---,---.-,.--.--.,e ,..,-w-s, y w - --+ e
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ATTACHF.ENT 5 U. S. NUCI.EA*1 PIGUI.ATORY CO:D!ISSION 0FFICE CF INSPECTION AND ENFORCEMENT REGION IV Report No. 50-445/79-2'; 50-446/79-23 Docket No. 50-445; 50-446 Category A2 Licensee: Texas Utilities Generating Company 2001 Bryan Tower Dallas, Texas 75201 Facility Name: Comanche Peak, Units 1 & 2 Inspection at: Comanche Peak Steam Electric Station, Glen Rose, Texas kaspectionconducted: October 1979 [ ./ h ~L.-.., M, / i AL /N/D Inspector: L-s z. R. G'. Taylor, Resident Reactor Inspector, Projects Date' Section /
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\\ Approved: h.. L2 1 'L. - vu - A n. 8.,W /.f W. A. Crossmam/ Chief, Projects Section Date Inspection Summarv: i Inspection durinz October 1979 (Report No. 50-445/79-24: 50-446/79-23) j Areas Inscected: Routine inspection by the Resident Reactor Inspector (RRI) of construction progress and practices; concrete construction activities; piping system installation and welding; storage and maintenance of equipment; construction fire protection; electrical cable installation; and followup i on previous inspection findings. The inspection involved sixty-nine inspector-hours by one NRC in:pector. Results: No items of noncompliance or deviations were identified. l 4 I f
ATTACHMENT 5 DETAILS 1. Persons Contacted Principal Licensee E:ployees
- R. G. Tolson, TUGCO, Site QA Supervisor
- J. B. George, TUSI, Project General Manager
- J. R. Merritt, TUSI, Construction and Engineering Manager
- D. N. Chapman, TUGCO, Quality Assurance Manager Brown and Root Employees
- U. D. Douglas, Construction Project Manager J. P. Clarke, Project QA Manager i
J. V. Hawkins, QC Supervisor The RRI also interviewed other licensee and Brown and Root employees during the inspection period.
- Denotes those persons with whom the RRI held on-site management
=eetings. 2. Action en Previous Inspection Findings 1 (Closed) Infraction (50-445/79-11): Failure to Implement the Quality Assurance Program for Civil Construction. The licensee notified RIV by n letter dated September 17, 1979, that all contemplated actions by his consultants and the Architect / Engineer have been completed and that the in place concrete was found satisfactory. The substantiating data to support this contention were reviewed by the RRI and personnel of the RIV Engineering Support Section. The II inspectors found that the stated positions were essentially based upon an examination of the in situ concrete by a sonic technique. This technique was developed, used and is interpreted by only one person in the industry and, as such, is not verifiable by any other party. Pending some additional and verifiable assurance that the in situ concrete has the necessary qualities required by the design, this matter will be considered as an unresolved item. (Closed) Unresolved Item (50-445/79-13): Potential Deficiency j i Regarding Design of Pipe Supports. The licensee notified RIV by letter dated September 8, 1979, that this matter had been investigated and deemed to be not reportable within the context of 50.55(e). Supporting data reviewed by the RRI and discussions with cognizant site personnel substantiated this determination. The RRI had no further questions on this matter. 2
= - - ATTAC'91ENT 5 a 3. Site Tours The RRI toured the safety-related plant areas several times during the inspection period to observe the progress of construction and the general practices involved. Three of these tours were conducted during portions of the construction labor second shift which continues to be relatively s=all and substantially devoted to electrical installation activities. During several of these tours, the RRI observed a general deterioration in plant area housekeeping. This matter was brought to the attention of licensee manage =ent who responded i= mediately. The construction force was directed to cleanup and recove the accumulated construction debris which was prc=ptly done. No items of noncompliance or deviations were identified. 4. Concrete Construction Activities The RRI observed a portion of the concrete placement activities for the Unit 2 dome. This placement, identified as 201-3805-013, was the final placement in the Unit 2 containment shell exclusive of the construction opening. The RRI observed the preparation of the concrete at the batch plant and the condition of the cement and aggregate storage activities. The RRI also observed the transportation of the concrete to the placement area via trucks and two y.i:d buckets including performance of required tests for slu=p, temperature and air content of the fresh =aterial. On October 24, 1979, at approxi=stely 11:15 a.m., the RRI received a es11 on the plant area telephone system. The caller, who refused to identify himself, stated that he and several other persons, also unidentified, had overheard the Brown and Root QC inspector say, "I didn't inspect this placement, but since the trucks are here go ahead." The caller said that the pour was in progress inside Reactor Building 2. The RRI went immediately to the placement area, which was a portion of an interior wall, and discussed the accusation with the QC inspector of record. The QC inspector premptly and emphatically denied having made the statement and star 2d positively that he had inspected the placement area. The RRI asked that all personnel associated with the activity be made available for an interview. A subsequent and longer interview with the QC inspector of record indicated that he had inspected the area on October 23, 1979, and was satisfied, except for cleanup, an element which was satisfactorily verified between 6:00 a.m. and 7:00 a.m. on October 24, 1979. The inspector did relate that he found one s=all area of the placement that had to be fully inspected just prior to initial delivery of concrete which was held up for a few minutes. This occurred, the I 3
i ATTACEMDIT 5 inspector said, because he had misconstrued the exact place =ent boundries. The inspector indicated that he had discussed the locali:ed lack of inspection with a craft general foreman in charge when he discovered his error and was immediately informed as to what the boundry really was. The QC inspector reiterated that the entire placement area had been properly inspected prior to initiation of concreting. A second B&R QC inspector who had assisted the inspector of record between 6:00 a.m. and 7:00 a.m. could shed no light on the quality of the inspection on October 23, 1979, but stated that the placement area was clean and ready prior to place =ent. He also indicated that he was aware of the inspector of record's problem with the small uninspected area but had no reason to raise a question sinra he had observed that the area had finally been inspected. The RRI subsequently intervir.wed sc=e seventy-four persons of the labor force who might have possibly overheard the alleged conversation or might have some knowledge of the quality of the placement. With two exceptions, no one admitted to being a party to or overhearing the alleged conversation / One of two exceptions was the previously referenced general foreman who recalled the conversation with the inspector of record about the small uninspected area and the short ensuing delay, but could not recall the exact words used. The other exception was a carpenter crew fore =an who said that he overheard a portion of a conversation between the general fore =an and the inspector. The foreman stated that to the best of his recollection the inspector said, "I didn't inspect that, but I'll get on it," and indicated that the inspector was pointing to an area of the placement. The foreman was aware that the placement was held up shortly for QC to finish inspecting the area. The various general foremen and foremen actively involved in the placement activity and cleanup process stated that they had observed and assisted the inspector of record on October 23 and October 24, 1979, and had no question as to his thoroughness. A few workers substantiated this review. Most of the workers indicated that they were not in a position to have had any specific knowledge relative to the quality of the inspections. Based upon the results of the interviews and upon the l'apsed time between when the conversation had to have taken place; ie., approximately 7:00 a.m. and the receipt of the phone call (11:15 a.m.), the RRI i can only conclude that the call was a hoax. The purpose of the hoax could i not be identified. No items of noncompliance or deviation w.ere identified. I 5. Piping Svstems Installation and Welding The RRI observed the general handling and installation of Reactor Coolant Pressure Boundry and other safety-related piping system ec=ponents during the inspection period. These activities were accomplished in accordance 4 l l -.n---. r-
ATTACHMENT 5 i 1 j with good industry practices. The RRI examined the following weld joint l radiographs for conformance to the requirements of ASME Section III: Joint Number Isometric Drawing Line Numbeg FW-19 BRP-RC-1-520-1 Reactor Main Loop i FW-14 BRP-RC-1-520-1 Reactor Main Loop FW-20 BRP-RC-1-520-1 Reactor Main Locp l FW-21 BRP-RC-1-520-1 Reactor Main Loop FW-22 BRP-RC-1-520-1 Reactor Main Loop FW-29 BRP-RC-1-520-1 Reactor Main Loep i FW-14-1 BRP-SI-1-RB-21 3-SI-1-339-2501R1 1 i FW-1 BRP-SI-1-RB-053 6-SI-1-329-2501R1 l FW-2 BRP-RH-1-RB-002 12-RH-1-002-25d1R1 i W-4 BRP-RC-1-R3-05 6-RC-1-008-2501R1 1 W-2 BRP-RC-1-RB-028B 6-RC-1-096-2501R1 W-3 BRP-RC-1-RB-028B 6-RC-1-096-2501R1 l W-4 BRP-RC-1-RS-02SB 6-RC-1-096-250131 l W-6 BRP-RC-1-R3-028B 6-RC-1-096-2501R1 I W-7 BRP-RC-1-RB-0283 6-RC-1-096-2501R1 W-14 BRP-RC-1-RS-0283 6-RC-1-096-2501R1 W-16 BRP-RC-1-RB-028B 6-RC-1-096-2501R1 W-18 BRP-RC-1-RB-028B 6-RC-1-096-2501R1 W-20 RRP-RC-1-RS-02SB 6-RC-1-096-2501R1 l I W-21 BRP-RC-1-R3-0283 6-RC-1-096-2501R1 l W-2 BRP-RC-1-RB-028A 6-RC-1-108-2501R1 W-3 BRP-RC-1-RB-028A 6-RC-1-108-2501R1 W-5 BRP-RC-1-RB-028A 6-RC-1-108-2501R1 5 6 i.,.
ATTACHMENT 5 W-6 3RP-RC-1-R3-028A 6-RC 108-2501R1 W-7 BRP-RC-1-R3-02SA 6-RC-1-108-2501R1 W-9 BRP-RC-1-R3-02SA 6-RC-1-108-2501R1 W-10 BRP-RC-1-R3-028A 6-RC-1-108-2501R1 W-17 BRP-RC-1-RB-028A 6-RC-1-103-2501R1 W-10 BRP-SI-1-RS-017 6-SI-102-2501R1 W-9 3RP-SI-1-R3-017 6-SI-102-2501R1 W-12 BRP-51-1-R3-017 6-SI-102-2501R1 FW-2 BRP-SI-1-RS-053 6-SI-1-330-2501R1 FW-6 BRP-CT-2-RB-09 16-CT-2-014-301R2 The six weld joints noted above as being in the Reactor Main Coolant Loop are the last of thirty-two field welded connections in the Unit 1 Main Loop piping. No ite=s of noncompliance or deviations were identified. 6. In-place Storage and Maintenance of Safetv-Related Components The RRI randomly selected several mechanical and electrical components during the period to observe the storage and maintenance practices being employed. Among these components were safety-related =otor operated valves, main control boards, switchgear cabinets, heat exchangers, reactor pressure vessels in both units and the Unit one Reactor Vessel internals. Each of the components observed were protected by adequate covering and were being maintained in a manner commensurate with supplier instructions and/or good industry practice. No items of noncomoliance or deviations were identified. 8 7. Construction Fire Protection The RRI verified that an adequate number of portable fire extinguishers displaying a properly charged condition were present in areas where welding and/or flame cutting operations were observed. The RRI observed on one occasion that welding operations of a structural nature were being carried on above a cable tray containing safety-related electri-cal cable that was unprotected from the weld spatter. Although none of spatter fell on the cable during the sustained period of observation, this was judged to be more of a fortunate accident than a deliberate action. The RRI ascertained that at present there is no coordinated inter-craft method of controlling such velding operations. The RRI discussed the 6 i
n.. c.C2 9. ... 5 i matter with licensee construction and Quality Assurance management, both indicated an awareness of the potentirA prr lem and stated k that a control method would be developed. This matter will be considered an unresolved item pending an opportunity to review and observe i=plementation of such controls. 8. Electrical Cable Installation During this period, the RRI observed the installation of a three conductor, nu=ber 6 AVG safety train A cable. The cable which runs from motor control center II31 to the Channel static inverter was approximately 410 ft. in length going through various segments of cable tray and conduit runs. The RRI verified that the cable utilined was of the type specified and verified, on a selective basis, that the cable was being routed as shown on the engineer furnished cable pull card. The RRI observed a portion of the cable through conduit pulling operation for consistency with project pro-cedures. The ERI interviewed and observed the activities of the QC inspector assigned to the activity. The QC inspector appeared to be knowledgeable of the requirements and diligent in his work effort. As a result of a licensee manage =ent audit and review of the cable installation program, the licensee determined that it was desirable to stop all safety-related cable pulling activities to allow time for an in depth review of project specifications, construction procedures and quality control procedures along with a review of appropriate personnel qualifications. The review was initiat-d in the latter part.cf the period and will probably last two to four weeks according to the licensee provided information. No items of nonecapliance or deviations were identified. 9. Unresolved Ite=s Unresolved items are catters about which more information is required in order to ascertain whether they are acceptable ite=s, items of nonce =pliance or deviations. Two such items are discussed in this report. The applicable paragraph and item title reference are as follows: i Paragraph 2: Unit 1 Contaic=ent Dcme Concrete Paragraph 7: Protection of Installed Electrical Cable 10. Management Interviews The RRI met with one or more of the persons identified in paragraph I on October 9,11,15, and 19,1979, to discuss various inspection findings and to discuss licensee actions and positions. 7
>. 6 /s >= rac \\c,j e UNITED STATES ATTACH.'-!ENT 6 y4. NUCLEAR REGULATORY COMMISSION ~ REGloN IV
- the t'
) c "*b " E $11 RYAN Pt.AZA oRIVE SulTF.1000 ARLINGTON, TEXAS 76012 .[ April 18, 1980 In Reply Refer To: ?:i Docket No. 50-445/Rpt. 80-08 50-446/Rpt. 80-08 Texas Utill:1es Generating Company N:TN: Mr. R. J. Gary, Executive Vice President and General Manager 2001 3ryan Tower Dallas, Texas 75201 Gentlemen: This refers to the inspection conducted by our Resident Inspector, Mr. F.. G. Taylor, during March 1980, of activities authoriced by NRC Construe:1on Per=its No. CPPR-126 and 127 for the Comanche ?eak facility, Units No. 1 and 2, and to the' discussion of our findings with Mr. R. G. Tolson and other me=bers of your staff at the conclusion of the inspection. Areas ern+ed during the inspection and our findings are discussed in the enclosed inspection report. Within these areas', the inspec: ion consisted of selective en-"=: ion of procedures and representative records, inte: views with perso:sel, and observations by the inspector. During the inspection, it was found that certain activities under your license appear :o in noncompliance with Par: 50, Title 10, of the Code of Federal Regu-lations. One apparent item of noncompliance, which is discussed in paragraph 5 of the enclosed inspection report, was forwarded to you by our letter dated l April 2, 1980. A second apparent itr-of noncomoliance and references to the pertinent requirements is identified in the enclosed no: ice of Violation and in paragraph 6 of the enclosed inspection report. ,l This notice is sent to you to pursuan: to the provisions of Section 2.201 of the NRC's " Rules of Practice," Part 2, Title 10, Code of Federal Regulations. Section 2.201 requires you to submi: to this office, v1 W 30 days of your receipt of this notice, a written statement of explanation in reply including: (1) cc:- rective steps which will be taken by you, and the results achieved; (2) corrective steps which will be taken to avoid further noncompliance; and (3) the date when full compliance will be achieved. I e I 9 ~
ATTACFMENT 6 i \\ Texas U:ili:ies Generating Cc=pany April 13, 1980
- 4e have also exa ined the actions you have taken with regard :c previously identified inspection findings.
The s:a:us of these 1:e=s is identified in paragraph 2 of the enclosed report. Two new unresolved ite=s are identified in paragraph 6 of the enclosed reporr. In accordance with Section 2.790 of de NRC's "parles of Practice," Part 2, Title 10, C,de of Federal Regulaticus, a copy of this letter and the enclosed inspection reper vill be placed in the NRC's Public Docu=en: Roc =. ~f the report contains any infcr=acic: tha: ycu believe to be proprietary, it is necessary that you submit a written application to this office, within 20 days of the date of this letter, requesting that such infor=ation be withheld frc= public disclosure. The application =ust include a full statement of the reascus why it is claimed that the infor=ation is proprietary. The applica:1cn should be prepared so that any proprietary infor=ation iden:1fied is contained in an enclosure to the applica:1cn, since the application without the enclosure vill also be placed in the Public Document Roc =. If we do not hear frc= you in this regard within the specified period, the reper: vill be placed in the Public Docuren: Roc =. Should you have any cuestions cenceW g this inspection, we will be pleased to discuss them with you. i Sincerely, "J. C. Seidle, Chief Reacter Ccustrue: ice and Engineering Supper: 3 ranch i
Enclosures:
1. Appendix A, Notice of Violatien l 2. II Inspection Report No. 50-445/80-08 { 50/446/80-08 cc: w/ enclosures Texas Utilities Generating Cc=pany AT"N : Mr. H. C. Sch=idt, Project Pmzager 2001 Bryan Tower Dallas, Texas 75201 l l i i l i g
ATTACFJENT 6 Docke: No. 30-445/Rpt. 80-08 50-446/Rpt. 80-08 Appendix A NOTICE 07 VIOLATION 3ased on :he resul:s of an NRC inspee:1on conducted during March 1980, 1: appears that certain of your activities were not condue:ed in full co=pliance vi:h the conditions of your NRC Construction Per=1ts No. C??R-126 and 127 as indica:ed below: Failure to Follow Procedures for Recorting and Reeair of Da=azed Electrical )I Cable 10 CFR 50, Appendix 3, Criterion 7 states in part, "Ac:ivities affec:ing quali:y shall be prescribed by docu=ented instrue:1ons, procedures or drawings, of a type appropriate to the circu= stances and shall be acco=- plished in accordance with these instructions, procedures, or drawings." Brown and Roo: Procedure 35-1195-EEI-13 " Repair of Electrical Cable Jacket," requires that da= aged cable shall be reported :o Electrical Engineering by the craft or by QC via a Field Deficiency Repor: and requires that da= aged jackets on control type cable be repaired using Okonite No. 35 Jacketing Tape in accordance with Okonite Drawing No. D-5715 when the da= age involves penetra: ion through the jacket. Contrary to :he above: l The Residen: Reactor Inspector (RRI) learned of a Saf ety Train A control cable :ha: had been da= aged during pulling of the cable through :he buried condui: sys:e= frc= the prinary power plan: building to the Service 'a'ater Intake Building. The cable was pulled back on or about March 28, 1980, and i: vas verified by the RRI that the da= age had been sustained, the jacket was penetrated, and :he repair was =ade with standard vinyl electri-cal tape rather than Okonite No. 35 and not in the =e: hod required by Okonite Drawing No. D-5715. The original da= age and initial repair were not reported via a Field Deficiency Report. This is an infraction. l l t i l r r - ~ - - - - - - -~ ' ^ ^ ~ ' "^
A2 TACE'4E2iT 6 (, ' U.S. NUCI. EAR REGULATORY CC." MISSION OFFICE OF INSPECTION AND ENFORCE. tit REGION IV i Report No. 50-445/80-08; 50-446/80-08 Docket No. 50-445; 50-446 Category A2 Licensee: Texas Utilities Generating Company 2001 Bryan Tower Dallas, Texas 75201 Facility Name: Comanche Peak, Units I and 2 Inspection at: Comanche Peak Steam Electric Station, Glen Rose, Texas Inspection conducted: March 1980 Inspector: &iSV 9'//S/?0 g R. G. Taylor, Resident Reactor Inspector ~ Date Projects Section Y/4/80 Approved: W. A. Crossman, Chief, Projects Section Date Insoection Summarv: Inspection During March 1980 (Recort 50-445/80-08: 50-446/80-08) Areas Inspected: Routine, announced inspection by the Resident Reactor Inspector (RRI) including follow up to previous inspection findings; general site tours; safety related piping installations; concrete placement activities; electrical installation activities; and protection,of major components. The inspection involved ainety-seven inspector-hours by one NRC inspector. Results: Of the six areas inspected, no items of noncompliance were identi-fied in four areas; two items of noncompliance were identified in two areas (infraction - failure to follow procedures for repair of damaged cable - paragraph 6; deficiency - failure to report a significant construction deficiency paragraph 5). ---n-nn-
/ A;TACP. GEST 6 } a DETAILS SECTICN 1. Persons Contacted Princioal Licensee E=clovees
- J. 3. George, TUSI, Project General Manager
- J. T. Merritt, TUSI, Construction and Engineering Manager
- D. N. Chapman, TUGCO, Quality Assurance Manager
- R. G. Tolson, TUGCO, Site Quality Assurance Supervisor
- J. C. Kuykendall, TUGCO, Plant Superintendent The RRI also interviewed other licensee and Brown & Root e=ployees during the inspection period including several randomly selected craft and Quality Control personnel that were interviewed in a semi-for=al, private atmosphere.
- Denotes those persons with whom the RRI held on-site =anagement meetings during the inspection period.
2. Action on Previous Inscection Findings (Closed) Unresolved Item (50-445/79-07; 50-446/79-07): ITT-Grinnell Piping Assemblies Containing Bends. The licensee informed the RRI that a complete records search had been accomplished at ITT-Grinnell which identified all pipe spool components that had been hot bent in the fabrication process and had provided the site with documentation to support that each had been solution annealed after the hot bending process. The RRI selected at random a total of eight spools, three installed and five in storage, that by either of the wall thickness or the sine had probably been hot bent. The necessary documentation was filed with the spool fabrication records attesting to the items having been solution annealed by various subcon-tractors to ITT-Grinnell. The RRI had no further questions on this matter. (Closed) Unresolved Item (50-445/79-08; 50-446/79-08): Pressuriner Safety Valve Spring Material. The licensee informed the RRI that the Crosby Valve drawing clearly indicated the material composition of the primary valve closure spring and that Westinghouse had approved the drawing and certified it to the owner. The RRI reviewed drawing DC-C-56964, Revision C and the Westinghouse letter WPT-877 and had no further questions on this matter. (Closed) Unresolved Item (50-445/79-16; 50-446/79-16): Design of Service Water System Cooling for Diesel Generators. The licensee notified RIV, by a letter dated October 5,1979, that a potential design problem with valves and controls at the Service Water System and Auxiliary Teedwater Syste= interface had been analy=ed and found not to be a problem. The letter further indicated that data to support this contention was on file in the offices of the Architect / Engineer (A&E), Gibbs & Hill, Inc. 2
~ ATTACEMENT 6 s, The RRI referred the matter to the Vendor Inspection 3 ranch for review during a routine inspection of that A/E. This review was acco=plished by VI3 personnel who informed the RRI of their concurrence with the A/E's analysis. For further infor=ation see Inspection Report 99900524/80-01. (Closed) Unresolved Item (50-445/80-01; 50-446/80-01): Engineer's Review of Test Reports. The licensee obtained a clarification statement from his A/E relative to the meaning of the sta=p appearing on various test report documents indicating " Approved For Arrangement Only." The A/E has stated that the review is for complete compliance to related contract requirements placed on the vendor and that the test report indicates that the equipment satisfied contractual requirements. The RRI had no further questions on this matter. (Closed) Unresolved Item (50-445/80-03; 50-446/80-03): Materials For Service Water Valve Discs. The licensee informed RIV, by letter dated Mard 25, 1980, that they have reviewed the vendor reported lack of ASME Code required heat treatment of aluminum-bron=e discs after welding has been performed and are reasonably sure that no safety problem exists. However, the licensee is returning the valves to the vendor for the Code required treatment prior to fuel loading or at some other time that will not interfere with early phases of pre-operational testing. This is a necessary step to prevent abrogation of the ASME certification for the valves. The RRI had no further questions on this matter but will follow it as an element of routine inspection.- (Closed) Infraction (50-445/80-03; 50-446/80-03): Failure to Follow Pro-cedures for Cable Pulling. The licensee notified RIV, by letter dated March 5, 1980, that the specific cables in question had been pulled back and visually examined and electrically tested with no damage being evident as a result of the failure to follow proper procedures. The licensee also committed to having fore =an level supervision is attendance during all cable pulling activities in the future and to use cable lubrication when prescribed by approved site procedures. The RRI has verified through several inspections that each of the corrective actions has been implemented. The RRI is also satisfied, based upon document reviews and interviews with appropriate personnel, that the cables did not suffer damage. The RRI had no further questions on this matter. (Closed) Unresolved Item (50-445/79-23; 50-446/79-22): Component Installation Activities. This unresolved item was written to express a possibility that neither Operational Travelers nor Engineering Instructions would be utilized for the setting and alignment of mechanical equipment and, therefore, that the work might be improperly accomplished and/or not inspected to a proper set of instructions. The RRI has no evidence gained during various inspections over the past twenty months that would indicate that this possibility is also a probability. The evidence, as based on these inspections, is that the 3
ATTACHMENT 6 s necessary instructions for setting and aligning equip =ent have been issued, have been followed, and have been verified by the site QA/QC organisation. i The RR1 has no further questions on this =atter. 3. Site Tours The RRI toured the safety-related plan areas several times weekly during the inspection period to observe the progress of construction and the general practices involved. Three of the tours were conducted during portions of the construction second shift, with a primary e=phasis on electrical cable pulling activities. No items of noncompliance or deviations were identified during the tours. 4. Safetv-Related Picine Installations and Welding The RRI made several observations of the general handling and installation practices of safety-related piping components including spool pieces and valves with a primary concern for those fabricated from stainless steel. These observations included operations being carried on within the plant primary buildings, the pipe fabrication shop and finally, how the finished, uninstalled co=ponents are stored. The RRI observed the work of one rela-tively new welder that had not been previously observed. The welder observed was 3FZ accomplishing the post-fitup tack welds on FW-2 of iso-metric CT-2-R3-005 in line 4-CT-2-097-301R2. The weld rod being used as obtained from the rod flag-tag and the weld filler metal log was from Sandvik heat 463638. The weld procedure being e= ployed was 88021. The RRI subsequently verified that the welder, weld material, weld procedure and the adjacent components being welded all were consistent with the requirements of the ASMI Code at that point in time. The RRI also observed the activities of the QA/QC person present and found him diligent and apparently knowledgeable of require =ents. The RRI also observed the activities of a QC person performing a liquid penetrant examination of two socket weld joints of a single spool in the containment spray system being finalised in the fabrication shop. The liquid penetrant examination was being carefully accomplished in a =anner consistent with Brown & Root Procedure CP-NDEP-300, ASMI Code requirements and good practice. In addition to the above, the RRI also examined the following weld radio-graphs which were found to be consistent with the requirements of ASME, Section III as to weld quality and ASME, Section V for radiograph quality: 4
ATTACH'4ENT 6 Weld Isometric Line Identificatice W-11 BRP-SI-1-R3-27 1.5-SI-1-057-2501R1 W-4 3RP-SI-2-R3-48 1.5-SI-2-023-250121 W-6 3RP-SI-2-RS-008 3-SI-2-003-2501R1 W-4 3RP-SI-2-R3-008 3-SI-2-339-2501R1 W-3 3RP-SI-2-R3-008 3-SI-2-339-2501R1 W-5 BRP-RC-2-R3-22 1.5-RC-2-079-2501R1 W-12 BRP-CS-1-R3-029 2-CS-1-112-2501RI W-7 3RP-RC-1-R3-016 2-RC-1-132-2501R1 W-24 3RP-RC-1-RB-032 1-RC-1-159-2501R1 W-6 BRP-RC-1-RB-016 2-RC-1-132-2501R1 W-24 BRP-RC-1-R3-033 1-RC-1-159-2501R1 W-15 3RP-SI-1-RS-022 1.5-RC-1.079-2501R1 W-8 3RP-SI-2-R3-008 1.5-SI-2-027-2501R1 FW-2 3RP-SI-1-R3-014 2-SI-1-059-2501R1 F's-8 BRP-CS-2-RB-021 2-CS-2-112-2501R1 FW-7 3RP-CS-2-RB-021 2-CS-2-112-2501R1 FW-9 BRP-CS-2-R3-021 2-CS-2-112-2501R1 W-20 3RP-SI-2-RB-060 6-SI-2-092-2501R1 FW-17 BRP-SI-1-R3-038 10-RC-1-092-2501R1 W-10 3RP-RC-1-R3-008 3-RC-1-052-2501R1 W-18 3RP-RC-1-R3-008 3-RC-1-052-2501R1 FW-9 BRP-FW-1-SB-019 18-FW-1-26-2002-2 FW-12 BRP-FW-1-SB-017 18-FW-1-34-2003-2 ~ 5
ATTACH'4ENT 6 i During an interview with a pipefitter welder, the RRI was infor=ed that sometimes the pipefitter fore =an is able to convince the Quality Control personnel to accept an out-of tolerance fitup at a weld jed-* The welder indicated that this see=ed to occur =ainly when the joint was in a pipe size where the inside of the joint could be ground after welding. The RRI obtained enough infor=ation fro: the welder to pinpoint one weld where this had occurred, but by the time the RRI could get to the joint to make an examination, access to the inside of the pipe had become blocked by additional installations. The only recourse under this circumstance was then to rely on exanination of the final weld radiographs and an ultrasonic measurement of the pipe wall thickness in the near vicinity of the veld and through the weld. The radiographs indicate a Code acceptable weld which does indicate evidence that some amount of internal grinding was accomplished prior to the radiograph having been taken. The ultrasonic wall thickness measure =ents show that adequate wall thickness was maintained even though the weld area was ground on both the interior and exterior surfaces. The RRI also interviewed the QC inspector who performed the fitup inspection. The inspector related that the joint in question was marginal in fitup, but that a conse=able ring had been used and that the fitup could not be more than a very few thousandths of an inch over specification. The RRI inquired as to what the inspector and the welders considered to be the requirements and was informed that 3/32nds of an inch was the maximum offse: allowable. The Code, however, allows as much as 3/16ths inch offset in the wall thickness involved. The Code further allows the fairing (grinding) of the interior of such a joint to provide a smooth transition across the weld. It appears that the tighter tolerance used by Brown & Root primarily comes from the verbage of the " General Piping Procedure" CPM-6.9 and further that the welders much prefer the better fitup since it is easier for them to achieve a satisfactory veld. The RRI has concluded that the weld joint in question is satisfactory; i.e., it meets all applicable requirements of the ASME Code and that the infor=ation received from the welder was largely based on a misunderstand-ing of the requirements. No items of noncompliance or deviations were identified. 5. Concrete Placement Activities The RRI examined the area preparation and the special formwork being installed preparatory to repairing the " honeycomb" condition in the interior walls of the Unit 2 Containment Building as discussed in Inspection Report 50-446/80-01. The work is being accomplished in accordance with a site generated set of detailed instructions while the RRI's basis for inspection is both these instructions and the applicable portions of the U.S. Bureau of Reclamation " Concrete Manual", a recognized authoritative publication on concrete work. The work to date appears to be progressing in accordance with the site instruction and the recommendations of the referenced publication. 6
( ) ATTACEMENT 6 w-The RRI did note : hat a period of nearly three =en:hs had passed between the ine :ha: the RRI was firs: no:1fied of the "heneyec=b" situa:icn and the ini:1ation of significant repair efforts. Based on the work perfer=ed by a consul: ant :o :he licensee (see Inspec: ion Repor: No. 50-446/c0-01); :he observed trips :o the si:e by represenza:ives cf :he A/I; and the tine span before repairs were started; it appeared :ha: an extensive engineering review had occurred ei:her for :he purpose of deter =ining the =e: hod of repair or :o develope a basis for possibly needing to =ake :he repair at all for other than cos=e:1c reasons; i.e., no: tha: the structural soundness of the walls was not affec:ed sudficiently
- o have a safe:y i= pac:.
10 CFR 50.55(e) in sub-paragraph (1) (iii) indicates that a deficiency is reportable if an extensive engineering evaluation is required to s1= ply de:er=ine the safety significance of such a deficiency as appears to be the situation. Since the licensee did not file an interi or final writ:en repor: to the RIV office wi:hin thirty days following i=nediate notification of the inciden: to the RRI on Dece=ber 13, 1979, the licensee was found to be in nonec=pliance with paragraph (3) of 10 CFR 50.55(e). A for=al No: ice of Violation was sen: to the licensee on April 2,1980.
- During an interview with a craft person, the person related a concern :ha:
some of the concrete in :se ceiling over a corridor in the facili:y Cc==en Fuel Building was no: wra : it is supposed to be. He stated that se=e:i=e over a year ago, he was drilling holes in the concrete :o inse- "'J-d" bolts frc= which he was going to suspend a pipe hanger. He said tha: the dust which ca=e fro = the drilling was nearly coal black ra:her than nearly white which is usually encoun:ered. He had asked his fore =an :o check with sc=eone on the apparen: proble=. He was subsequently told to go ahead and drill the holes; : hat nothing was wrong, but was given no explana: ion. The RRI de:er=ined frc= the person's description of the loca: ion of the inciden: and by reference to :he civ'il/ structural design drawings tha: the concre:e in the particular ceiling was designated :o be " heavy weight" concrete to provide added personnel shielding frc= radiation since the ceiling is also the support floor slab for the refueling canal which cen-nects the spen: fuel pools to the contain=en:s and the corridor below is a pri=ar7 passageway. The job specifications require :ha: a =agne:ite ore be used as both :he fine and coarse aggregate in the concrete to achieve the =uch higher density. With the aid of a licensee e=ployee, the RRI located a block of heavy weigh: concrete that had been originally cas as a test weight for hoisting equip =en: and drilled a hole in the block wi:h a standard "E11:i" bol: carbide drill. The resulting drilling dus was nearly coal black just as described by the person. Although 1: has not j yet been done, the RRI intends to locate the person expressing the concern and infor= hi= that his concern is needless while thanking h1= for relating his concerns to the RRI. Except as noted above, no itens of nonce =pliance or deviations were iden:1-fied of a technical safety nature in this area of the inspection. N 1. l I i l l
) ATTAC'91E::T 6 .e i \\_ ' 6'. Electrical Svste= Installation Activities The RRI =ade an on-going series of observations. of the labor and QC activities as they relate to electrical cable installation and ter=ination. The RRI observed a number of v:rious sizes of contrcl and low voltage power cables being installed in both of the pri=ary safety trains. The RRI found that the labor force is carefully handling the cables and is lubri-cating them thoroughly when pulling through conduits already containing cables. Random checks in the cable tray system indicates that the trays are properly installed and adequately clean and further that there is presently no evidence of inter =izing of either of the trains with each other or of either with nonsafety cable. No attempt was made to trace any given cable through its nuting system since this is more efficiently and effectively done by other means than visual. The RRI randomly selected.two, wire lug crt= ping tools observed in use and examined them for apparent wear or evidence of careless use. The tools appeared to be in good condition. The tools (CT-1224 and CT-1323) were also used as a vehicle for examining the crimping tool control system. The tools are checked each three months by the site calibration facility using vendor recommended procedures and certified go-no go gauges in accordance with precedures 111-98 and~IE.T-103, respectively. The RRI examined, on a random basis, the actual ter=inations in various cabinets and observed that the lugs were correctly crimped onto the wires and would not pull loose with application of reasonable force. The RRI selected a cable type observed being installed as a vehicle for examining the qualification of the cable as required by the cocnit-ments contained in the FSAR, Sectiot 8. The cable selected was W-847 from reel W-847-2 and is a 12 conductor unit made up of nu=ber 12 AWG individual wires and including their insulation with filler plus a jacket. The cable was supplied by Rockbestos in accordance with Project Specification IS-133.1. The Project Specificatica and the referenced FSAR section both require that the cable be qualified in accorcance with 1777-383-74, "Tirr Standard For Type Test of Class II Ilectric Cables, Field Splices and Connections for Nuclear Generating Plants." The RRI obtained documentation indicating that Rockbestos had performed the i stipulated type and production tests required by the specification with the exception that there was no clear evidence that the three separate type tests of cables, as required by paragraph 2.5.4.3 of I7 7-383, had been accomplished nor was there evidence in the report that individual conductor tests had been performed as required by paragraph 2.5.6. The report did contain an attachment indicating that such tests had been performed and were sucesssful. s This matter will be considered an unresolved item pending receipt of specific test data to substantiate the vendor statement 8
,n y n n..sc =._y r_ y. n. g .f 4 i During an in:erview, :he ER: was infor=ed that a safe:y-rela:ed control cable had been da: aged while pulling 1: into :he buried condui: sys:e= running fro: :he =ain plan: buildings :o the Service Water intake Building. j The interviewee related :ha: early in January 1980, a 3 or 7 condue:or, orange (Safe:y Train A) control cable had slipped between a pully wheel and :he pulley fra=e during an interrup:icn in pull and tha: when :he pull was resu=ed, :he jacke: of the cable had been cu: open. The pully had been installed in the firs =anhole outside of the =ain buildings to aid the elec:ricians in =aking :he nearly nine:y degree direction change :hrough the =anhole. The person further related tha: the elec:rical crew fore- =an in charge had instructed his people :o tape up the da= aged area with standard electrical tape (Sco:ch 33) and continue :he pull which placed the da= aged area so=ewhere in the buried condui: that is nearly five hundred fee: long. Neither :he fore =an or any of his crew had apparently seen fi: :o report the inciden: nor was QC apparen:ly aware of it. The i RRI discussed the =a::er with the site electrical engineering personnel who indicated tha: the standard elec:rical tape used in the repair was probably not adequate as a jacket repair considering the location and the ti=e tha: cable would have perfor=ed a safe:y fune: ion; i.e., for:7 i years. The engineering personnel deter =ined : hat only so=e eigh: 5 and/ i or 7 condue:or, orange cables were likely to be involved since :he balance of the cable in the particular condui: vas ei:her =uch larger or s= aller and the person interviewed had been very specific in his rela: ion of the The RRI verified that each of the identified cables had been pulled event. wi:hin tha ti=e fra=e of :he related event by review of :he elec:rical cable pull cards. The licensee elected to deter =ina:e the cables and draw the: back to the =anhole. The RRI exa=ined the single cable found to have :he da= age and found tha: a cut of about one inch long had occurred tha: pene:ra:ed the cable jacket, but also found : hat no da= age had occurred to the individual wire insulation within the cable. Based on the RRI's knowledge of the characteristics of the wire insulation =aterial, cross-linked polyethylene (IL?I), 1: is very doub:ful :ha: the fune:1ocabili:7 of the cable would have even been i= paired and, therefore, the cut jacket has no direct i= pac: on safety. The i=p11 cations of :he incident do, however, have a potential i= pact on saf e y in : hat it is j indicative of a breakdown in the Construe:1on Qualiry Assurance Progra= as evidenced by the fact that an electrician fore =an took 1: upon hi=self to deter =ine the need for the type of repair that was to be =ade to a da= aged cable rather than reporting the =atter through proper channels and allowing engineering to =ake the decision. The inciden: is a violation of the inten: of Appendix 3, 10 C7R 50, Criterion V in that cable was not repaired in accordance with the applicable procedure. i For the record, the RRI would note that at the ine of the incident, specific instructions had been issued addressing the area of cable da= age or repair after the da= age had occurred. The procedure provided for reporting da= age to engineering and also provided for the use of a self-vulcanizing rubber tape to =ake jacket repairs ra:her than esi=g Scotch 33. 9
-J 1 ? +. 7 4 y n..n' C."'EN. "> 6 ( { la regard' to the aforenen:ioned cable'.repad: procedure (III-13), :he REI's ' review,l along with discussions with app:cpria:e Quali:7 Ingineering personnel ~ indicate a lack of clarity in.1:s recuire=en:s. Thefprocedure curren:ly allows :he replace =en: of wire insula: ion =sterial in a =ul:iconductor cabl'e'wi:h the self-vulcani:ing : ape. There is currently no evidence avail-able ~which would show tha: :he : ape has :he sane or be: er fla=e retardance characteristics as the fac:ory applied IL?E insula:1on. The verbage utilized in the procedure also essentially requires discussion with the procedure writers in order to achieve an understanding of what was in: ended by the writers. This =a::er vill be considered to be unresolved pending clarifica: ion. s 7. Protectice of Major Safe:v-Related Cc=cenents ~ The RRI verified tha: he reac:or vessels in both units are adequa:ely protected to preven: likely da= age and/or con:a=ina ion. The Uni: 1 reactor vessel head is well covered and pro:ected in its lay-down area. The Uni: 1. reactor vessel core-support co=ponents (internals) re=ain in their_ enclosed lay-down areas wi:hin :he refueling pool area. No ite=s of nonce =pliance or' deviations were identified. 8. Unresolvedkhe=s ~ Unresolved itp=s are =a::ers about which =cre Lafor=a:icn is required in order to ascertain whether they are acceptable ite=s, 'ite=s of nonec=pli-ance or devi2-ions. Two such ite=s are discussed in paragraph 6 and will j be referenced;in the future as: Clarifica: ion of Rockbestos Elec:rical Cable Qualifica:1cn a. j b. Clarificacion of Elec:rical Cable Repair Procedures 9. Marsge=en: In:erviews The RRI =e: with one or = ore of the persons identified in paragraph 1 on i March 3, 4. -18,19 and 29,1980, to discuss various inspection findings and to discuss 1,1censee actions and positions. t \\ t 7 I I 1.0
s ~ UNITED STATES ATTACEMINT 7 [p a'cw%, NUCLEAR REGULATORY COMMISSION OV -f' REGloNIV n l I ' I /2, c 611 RY AN PLAZA oRIVE, SUITE 1000 I
- 7. k d..,/
ARLINGTcN, TEX AS 76012 m, / s%.,. > e f June 17, 1980 In Reply Refer To: RIV j Docket No. 50-445/Rpt. 80-11 i 50-446/Rpt. 80-REC"qi g i 20 Texas Utilities Generating Co=pany bSO ATTN: Mr. R. J. Gary, Executive Vice Tyggg 0", Presid it and General Manager j 2001 Bryan Tow-g S 1 Dallas, Texas 01 Gentle =en: This refers to the inspection conducted by our Resident Reactor Inspector, Mr. R. G. Taylor, during April and May 1980 of activities authorized by NRC Construction Per=its No. CPPR-126 and 127 for the Co=anche Peak Facility, Units No. 1 and 2, and to the discussion of our findings with'Mr. R. G. Tolson and other =e=bers of your staff at the conclusion of the inspection. Areas exa=ined during the inspection and our findings are discussed in the enclosed inspection report. Within these areas, the inspection consisted of selective examination of procedures and representative records, interviews with personnel, and observations by the inspector. During the inspection, it was found that certain activities under your license appear to be in nonco=pliance with Part 50, Title 10, of the Code af Federal-Regulations. An apparent item of nonco=pliance, which is discussed in para-graph 7 of the enclosed inspection report, was forwarded to you by our letter dated April 9, 1980. You responded to our letter by your letter of May 5, 1980, and our Resident Reactor Inspector has verifed i=plementation of your co=mitment as discussed in paragraph 2 of the enclosed report. No further response is required. We have also examined the actions you have taken with regard to previously 4 l identified inspection findings. The status of these ite=s is identified in paragraph 2 of the enclosed report. I i { l e f +* v m. ...-..,r-,.,-~ ,4 -,----w.~
= - + ATTACE.LMIT 7 e Texas Utilities Generating 2 June 17, 1980 Conpany In accordance with Section 2.790 of the !?RC's " Rules of Practice," Part 2, Title 10, Code of Federal Regulations, a copy of this letter and the enclosed inspection report will be placed in the !!RC's Public Document Room. If the report contains any infor=ation that you believe to be proprietary, it is necessary that you submit a written application to this office, within 20 days of the date of this letter, requesting that such infornation be withheld from public disclosure. The application must include a full stata=ent of the reasons why it is clai=ed that the infor=ation is proprietarf. The application should be prepared so that any proprietary infornation identified is contained in an enclosure to the application, since the application without the enclosure will also be placed in the Public Docu=ent Room. If we do not hear from you in this regard within the specified period, the report will be placed in the Public Docunent Room. Should you have any questions concerning this inspection, we will be pleased to discuss them with you. Sincerely, / / ffU W. C. Seidle, Chief Reactor CSrAtruction and Engineering Support Branch
Enclosure:
IE Inspection Report To. 50-445/S0-11 50-446/80-11 cc: w/ enclosure Texas Utilities Generating Cocpany ATTN: Mr. H. C. Sch=idt, Project Manager 2001 3rfan Tower Dallas, Texas 75201 i YC. : 0. W V -m
ATTACEMINT 7 U. S. NUCIEAR REGUI.4 TORY CCMMISSION OFFICE OF INSPECTION AND ENFORCEMENT REGION IV Report No. 50-445/30-11; 50-446/80-11 Docket No. 50-445; 50-446 Category A2 Licensee: Texas Utilities Generating Ccspany 2001 3ryan Tower Dallas, Texas 75201 Facility.Name: Cocacche Peak, Units 1 and 2 Inspection at: Comanche Peak Steam Electric Station, Glen Rose, Texas Inspection conducted: April and May 1980 G/6/ga Inspector: e-R. G. Taylor, Resident Reactor Inspector Date Projects Section d[//[Jo Approved: h== W. A. Crossman, Chief, Projects Section Date 1 Inspection Su= mary: i Inscection Durine Acril and May 1980 (Report 50-445/80-11: 50-446/80-11) Areas Inscected: Routine, announced inspection by the Resident Reactor Inspector (RRI) including follow up to previous inspection findings; general i site tours; safety-related pipe and equip =ent installations; concrete repair activities; electrical installation activities; and protection of major cosponents. The inspection involved one hundred fifty-five inspector-hours by one NRC inspector. Results: Of the seven areas inspected, no items of noncompliance were identified in six areas. One item of noncompliance was identified in one area (infraction - failure to follow piping installation procedures - paragraphs 2 .and 7). t l
ATTACHMENT 7 DETAILS 1. Persons Contacted Principal Licensee E=cloyees
- J. B. George, TUSI, Project General Manager
- J. T. Merritt, TUSI, Construction and Engineering Manager
- D. N. Chapman, TUGCO, Quality Assurance Manager
- R. G. Tolson, TUGCO, Site Quality Assurance Supervisor The RRI also interviewed other licensee and Brown & Root e=ployees during the inspection period including both craft labor and QA/QC personnel.
- Denotes those persons with whom the RRI held on-site =anagement meetings during the inspection period.
2. Action on Previous Inspection Findings (Closed) Infraction (50-445/79-18): Failure to Control Inspection Stamps. As noted in paragraph 2 of Inspection Report 50-445/79-27; 50-446/79-26, the use of nt=bered inspection sta=ps has been discontinued and the imple-menting procedure cancelled. The licensee's Site. Surveillance Group interviewed all QC personnel to whom such stamps had been issued and who had failed to return them when the cancellation took place to ascertain the reason for the nonreturn and appreximately when the stamp was lost or misplaced. Personnel no longer in QC by reason of termination were not interviewed. The dates of loss and/or termination were then used as the basis for an extensive QA records search to determine if the missing stamps had been i= properly used. The records search failed to reveal any such i= proper use and the licensee concluded that the loss of the stamps was attributed to personnel carelessness rather than any overt act. The RRI had no further questions on this matter. (Closed) Unresolved Item (50-445/80-01; 50-446/80-01): Class 1 to Class 2 Transition Chifices. The licensee has issued Component Modification Cards 33001 and 33002 which require the installation of the required transition crifices in the manner orginally called for in the design drawings at a location approximately six inches from the improperly sized orifices. The improperly si:ed orifices will be plugged and seal welded. TheRRIhadnofurtherquestionsonthisbatterbutwillfollowthe implementation of the above Component Modification Cards during routine inspections. (Closed) Infraction (50-445/80-01; 50-446/80-01): Failure to Provide Instructions and Procedures Appropriate to the Circumstances. The licensee informed RIV, by letter dated February 19, 1980, that their analysis of 2 - ~
ATTACEMENT 7 the as-built =ounting of the battery chargers indicated tha: the =cunting provided adequate strength to satisfy seismic requirements. The licensee also stated that engineering procedures were being revised to require an Architect / Engineer review of equipment ounting details in addition to that already required by the equipment vendor. The RRI has verified that the procedure has been revised and imple=ented. The RRI had no further questions on this =stter. (Closed) Deficiency (50-445/80-08; 50-446/80-08): Failure to Report a Significant Construction Deficiency. The licensee infor:ed RIV, by letter dated April 21, 1980, that a review of the reporting require =ents of 10 CFR 50.55(e) had been accomplished and that a meeting in the RIV office, as docu=ented by Inspection Report 50-445/80-12; 50-446/80-12, had rendered further clarification of the require =ents. The licensee stated that necessary instructions had been given to appropriate personnel in the =atter. The RRI has interviewed these personnel and is satisfied that they now understand and will imple=ent the require =ents fully. For further infor=ation relative to the " honeycomb" condition referred to in the original finding, see paragraph 4 of this report. This item is considered closed. (Closed) Infraction (50-445/80-11; 50-446/80-11): Failure to Follow Piping Installation Procedures. This infraction, which is discussed in paragraph 7 of this report, was forwarded by RIV letter dated April 9,1980. The licensee reported to RIV, by letter dated May 5,1980, that an analysis of the reported situation showed that no excessive strain had been placed on the pump no::le involved. The RRI reviewed these calculations with the NSSS supplier and was satisfied that no damage had been incurred. The licensee also cot =itted to additional inspection for like ite=s which was acco=plished and the results documented. Other situations were found of a like nature and fortunately no harm to equipment was involved. The licensee stated that piping installation procedures have been revised to =ake it clear to the craft labor force that piping copnections to equipment are not to be made until the piping is supported properly with hangers rather than by si=ple cribbing. l The RRI observed, during tours of the facility, that the revised procedures had been i=plemented and had no further questions. (Closed) Infraction (50-445/80-08; 50-446/80-08): Failure to Follow Procedu.res 4 for Reporting and Repair of Damaged Elect:ical Cable. The licensee in-for=ed RIV, by letter dated May 14, 1380, that a new cable would be pulled through the buried bus duct to replace the damaged cable. In addition, new cables were also pulled to replace several other cables in the duct that were damaged in the search for the cable originally reported. The licensee also stated, in the referenced letter, that Management / Supervisory i Seminars had been held to emphasize the need to follow all project procedures. The RRI reviewed docu=entation indicating that eighty-two persons, including 3
ATTACH. MENT 7 electrical depart =ent superintendents, general fore =en, and foremen, attended one of two such seminars. Interviews with two electrical crew fore =en indicate that they are aware of the procedural require- =ents. The RRI had no further questions. 3. Site Tours The RRI toured the safety-related plant areas several times weekly during the inspection period to observe the general progress of construction of the practices involved. Five of the tours were accocplished during portions of the second shift. Since the principal effort of the second shift is the installation of electrical cables, primary e=phasis was placed on this activity. No items of noncompliance or deviations were identified. 4. Concrete Recair Activities The RRI observed substantial portions of the activities involved in the removal of the defective concrete in the " honeycomb" areas of, the Unit 2 Reactor Contain=ent Building internal walls as discussed in Inspection Reports 50-445/80-01; 50-446/30-01 and 50-445/80-08; 50-446/30-08. The RRI examined a nu=ber of the cavities after re= oval of the " honeycomb,' after application of concrete bonding agents, and again after the repair for= work was in place for the concrete pour back. In one area, the sleeve through the wall for the reactor coolant pipe had to be partially re=oved to gain access to the defective concrete. The RRI observed portions of the weld repair to the sleeve to re-establish its orginal configuration. The welding was acco=plished in accordance with the engineer's instructions by qualified welders utilining qualified weld procedures. As of the end of the inspection period, the entire repair effort was essentailly complete and appeared to have been done in a sound manner in accordance with recogni:ed concrete repair practices. The licensee officially infor=ed RIV of the above matter as required by 10 CFR 50.55(e) in a letter dated April 21, 1980. The report outlines the engineering evaluations perfor=ed, the safety impact had the defects gone unrecognized and/or unrepaired, and the repair =ethods to be utilized. No items of noncompliance or deviations were identified. 5. Major Cemoonent Installation Activities During the inspection period, the RRI observed the efforts involved in installing the last two steam generators and the first two Reactor Coolant pumps in the Unit 2 Reactor Contain=ent Building. The RRI observed the 4
ATTAC:0!E:IT 7 initial preparation of the steam generators for hoisting into the building, the actual hoisting and :ove=ent, and finally the setting and align =ent of the units on their support colu=ns. Each step was observed to be in accor-dance with Operation Travelers RISO-369-3400 and ME30-2005-5500 governing the work of the riggers and millwrights, respectively. The RRI also reviewed the steps indicated by the two Operation Travelers with the NSSS supplier representatives on site and verified that the steps utilired were in consonance with the supplier's written reco==endations. The RRI reviewed data developed by the site field engineers (surveyors) which showed that the generators are well within the established vertical-require-ments of the vendor and that each of the four support colu=ns are carrying approximately equal load. In regard to the Reactor Coolant pu=p instal-lation, the RRI observed the work involved in setting the pumps on their col"-"s and establishing the pu=ps into sa. essentially level position. The RRI also observed the preliminary installation of two of the Reactor Coolant pipe legs through the sleeves leading to the Reactor Pressure Vessel. These pipe sections were carefully handled and placed into posi-tion in accordance with good practice. No items of noncompliance or deviations were identified. 6. Re2ctor Coolant Pressure Boundary Pipine Installation The RRI made limited observations of piping component handling in the Reactor Coolant Pressure 3oundary area during the period. The RRI observed two welds in process as follows: Weld Number: W-3A IW-20 Isometric: RC-1-R3-026 SI-1-R3-037 Line Identification: 14-RC-1-135-2501R1 10-RC-1-021-2501R1 Welder Identification: AWT and 3MK 3AG Weld Procedure: 99025 (Machine GTAW) 88025 (Manual GTAW) Filler Metal Identification: 463870 762550 Subsequent to the observation of welding, the RRI verified that the welders, weld procedures and weld filler metals were each properly qualified in accordance with the ASMI Code, Section III or IX as appropriate. In addition, the RRI also examined the radiog'raphs taken of the welder quali-fication test coupons for welders BAG, 3LU, AXC, SPA and AED. These i l l 5 t
e ATTACH'4DT 7 e radiographs, which are an examination alternative of ASME, Section IX (the other alternative is prescribed bend tests), indicated a sensitivity technically acceptable per Section V of the Code. The RRI discussed the radiographs with the supervising radiographer w'co stated that the fuz iness of the radiographs was caused by energy scatter from the source (Iridium 192). Since the radiographs =et all technical requirements of the Code, he felt there was no problem. The RRI agreed that the Code had been technically satisfied, but at a marginal or minimum level and the radio-graphs could be substantially improved by a better technique. The RRI will pursue this mat er during future inspections. The above discussed radiographs indicated that each welder had accomplished a weld or velds that satisfied the Code requirements and were, thus, fully qualified to perform production welding. The RRI also examined radiographs of the following reactor coolant boundary (Class 1) welds: Weld Identification Isometric Line Number W-6 SI-2-R3-042 2-SI-2-086-2501R1 FW-12 SI-1-R3-21 3-SI-1-033-2501R1 W-14 SI-2-R3-042 2-SI-2-086-1501R1 W-14 SI-1-R3-020 1.5-SI-1-020-2501R1 W-12 SI-2-RB-042 2-SI-2-086-2501R1 W-6 CS-1-R3-0313 2-CS-1-105-2501R1 FW-1-1 RC-1-R3-15 3-RC-1-111-2501R1 i FW-10-2 RC-1-RS-15 3-RC-1-111-2501R1 FW-38-1 RC-1-RS-15 3-RC-1-146-2501R1 W-10 CS-1-R3-0313 1.5-CS-1-249-2501R1 W-8 CS-1-R3-0313 1.5 -CS-1-105-2501R1 W-9 CS-1-R3-0313 1.5-CS-1-105-2501R1 W-7 CS-1-RS-0313 1.5-CS-1-105-2501R1 2 W-2 CS-2-RB-074 2-CS-2-112-2501R1 W-5 CS-1-RB-0313 2-CS-1-105-2501R1 6
s ATTACEMENT 7 W-3 CS-1-RB-023 2-CS-1-112-2501R1 W-18 RC-1-R3-15 3-RC-1-111-2501R1 FW-42 RC-1-R3-15 3-RC-1-146-250121 FW-6 RC-1-R3-08 3-RC-1-052-2501R1 W-6 RC-1-RB-06 6-RC-1-70-2501R1 FW-3 RC-1-R3-017 4-RC-1-075-2501R1 FW-38-2 RC-1-RB-05 3-RC-1-146-250121 FW-2 RC-1-RS-017 4-RC-1-075-250lR1 W-5 SI-2-R3-042 2-RC-2-086-2501R1 W-35 SI-1-R3-015 2-SI-1-086-2501R1 FW-11 SI-1-RB-021 3-SI-1-033-2501R1 FW-1 RC-1-RB-06 12-RC-1-069-2501R1 FW-5A RH-1-RS-02 12-RE-1-022-2501R1 No items of noncompliance or deviations were identified. 7,. Other Safetv-Related Piping Installation Activities The RRI observed welder AHI during a period when the welder was working on joint FW-3 as identified on isometric CT-1-RB-17 in line 10-CT-1-027-301R2. The welder was working to Weld Procedure 88021 using filler metal Heat Number 463638. The qualification of the procedure and this heat of l filler metal have been verified several ti=es during previous inspection. Review of the welder qualification records for AHI indicate that he has been properly qualified in accordance with ASME, Section IX. The RRI also examined the licensee actions in regard to i=ple=entation of his commitment to radiograph and repair those field welds in the Safety Class 3 Component Cooling Water and Auxiliary Feedwater Systems that do not require radiographs under the Code. (For more information regarding this cocmitment, see Inspection Reports 50-445/79-12 and 50-445/79-17.) The personnel canaging the, program indicated that approximately 56% of the 1842 welds involved have, to date, been radio-graphed and that about 37% of those requiring repair have been repaired. l The RRI randomly selected the following radiographs for review: 1 I l 7 i l i
ATTAC'CiENT 7 Weld Identification Isemetric Line Nu=be-FW-10 AF-1-SB-23 4-AF-1-102-152-3 FW-15 AF-1-ID-05 3-AF-1-8 6-152-3 FW-13 CC-1-R3-042 3-CC-I-232-152-3 FW-10A CC-1-RB-58A 3-CC-1-234-152-3 FW-30 CC-1-RS-58A 3-CC-1-234-152-3 FW-1 CC-2-AB-045 3-CC-2-118-152-3 FW-22-R1 AF-1-SB-10 6-AF-1-33-15 2-3 FW-28-R1 AF-1-SB-15 4-AF-1-102-152-3 FW-3-R1 AF-1-SB-72 3-AF-1-72-152-3 FW-24-R1 CC-1-RS-041 3-CC-1-232-152-3 The RRI made nu=erous observations of the general pipe and component handling operations in both Units 1 and 2 during the inspection period and found that good practices were being followed as outlined in the General Piping Procedure CPM-6.9. In one instance howesar, the RRI observed a situation that was of concern in that possible major safety component damage might have occurred which could easily have gone undetected. The RRI found that a pipe asse=bly, consisting of several feet of six inch diameter pipe, was being entirely suspended by attach-ment to the suction co:nle flange of the Unit 2 Train A Safety Injection pump TCX-SIAPSI-01. Further investigation developed that the pipe assembly would place a torque load on the no:nle of between 1500 and 2000 foot pounds. The RRI found that CPM 6.9 did not provide instruc-tions on this matter to the labor force, although the project Mechanical Erection Specification (MS-100) specifically prohibited such practices. The RRI notified the licensee of the situation which was in turn followed up with a Notice of Violation dated April 9,1980. The licensee responded to the initial notification by having the other installed pu=ps and valves in Unit 2 checked for like situations. A very limited number of other comparable situations were identified during this inspection. The RRI identified situation and others identified by the licensee were detailed on Nonconformance Reports which vere submitted to the component 8
5 ATTACHENT 7 vendor, Westinghouse, for analysis of possible damage to the components. The analysis indicated that no dacage was likely to have occurred due to the static loading on the nor:les, although had the pipe been of a heavier schedule or longer in length, such damage would have occurred. The West-inghouse analysis was reviewed by the RRI who had no question of its accuracy. The licensee's investigation of the circumstances surrounding the incident indicated that the pipefitters had the pipe supported by te=porary wooden blocks or jack stands when they left the work area. These workers were subsequently reassigned to other work and did not return to the area. In the meanti=e, it appears that a group of painters were assigned to paint the floors in the area and removed the te=porary shoring under the piping leaving it suspended from the no:cles. The labor force was notified that this practice must cease and the licensee also revised CPM 6.9 to provide specific instructions in the matter. All of these actions were consu= mated during the period covered by this report, and as noted in paragraph 2, this item of noncompliance is con-sidered to have been satisfactorily closed. Except as noted above, no ite=s of noncompliance or deviation were identified. 8. Electrical Installation Activities The RRI =ade a number of observations of electrical cable installations during the inspection period. The primary inspection effort was directed toward observing the activities of the various cable pulling crews and toward this end at least five crews were checked. During most of the period there were seven active crews working safety-related cable. Each of the crews observed appeared to be knowledgeable of the prescribed methods of pulling cable and of the limitations i= posed by site pro-cedures and good practice. The RRI also examined most of the Main Control Room cabinets and the termination cabinets in the Cable Spread Room of Unit I relative to the quality of the workmanship displayed in termination of the cables. No instances were found in which the termination was less than satisfactory as evidenced by the application of the correct size wire lug that was properly cri= ped and tightly installed on the terminal boards. The RRI also examined a nu=ber of terminations for correct connection on the te=sinals as indicated on the electrical design drawings with no errors being detected. This effort was pri=arily directed toward the main 6.9 KV switchgear in Safety Train A. No items of noncompliance or deviations were identified. 9. Protection of Major Safetv-Related Ec_uiement During the course of general plant tours, the RRI noted that the major plant components continue to be well cared for as evidenced by space 9 i
ATTACIC4ENT 7 y l i l heaters being energined and where appropriate, because of on going work, the equipcent is adequately covered The Unit I and 2 Reactor Vessels were noted to be well protected even though extensive civil construction work was in progress in the ir:=ediate vicinity. The Unit 1 Reactor Vessel internals were noted to be in their enclosures and apparently adequately protected as was the Unit 1 Vessel head with the installed Control Rod Drive Mechanisms. No items of nonco.pliance or deviations were identified. 10. Management Interriews The RRI met with one or more of the persons identified in paragraph 1 on April 2, 3, 9, 10, 15, IS, and May 13 and 29, 1980, to discuss inspection findings and to discuss licensee actions and positions. f 6 I I l i I i j i 1 l 10 I
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ATTACEMENT 8 tg UNITED STATES '\\ E
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,0g NUCLEAR REGULATORY COMMISSION 3. %. - 2 e REGION IV p..,"3 .w. 3 611 RYAN PLAZA ORIVE. SUITE 1000 l'h i -- !U! O
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.[ AR LINGTON, TEXAS 76011 c:<fr~: & rtm E3!Tu; STATO March 14, 1979 In Reply Refer To: RIV - ~~l Docket No. 50-445/Rpt. 79-03 50-446/Rpt. 79-03 '\\ L J+ Texas Utilities Generating Company ATTN: Mr. R. J. Gary, Executive Vice President and General Manager 2001; Bryan Tower Dallas, Texas 75201 Gentl emen: This refers to the inspection activities perfor ed by our Resident r. Q Inspector, Mr. R. G. Taylor, during the period February 2-28,1979, of activities authorized by NRC Construction Permit Nos. CPPR-125 and 127 for the Comanche Peak facility, Units No.1 and 2, and to the discussion of our findings with Mr. R. G. Tolson and other members of your staff during the course of the inspection. Areas examined during the inspection and our findings are discussed in the enclosed inspection report. Within these areas, the inspection con- ~l sisted of selective examina, tion of procedures and representative records, interviews with personnel, and observations by the inspector. During the inspection, it was found that certain activities under your license appear to be in noncompliance with Appendix B to 10 CFR 50 of the l NRC Regulations, " Quality Assurance Criteria for Nuclear Power Plants." The Notice of Violation for the item of noncompliance reported in paragraph 9 of the enclosed inspection report was forqarded to you by our letter, i dated February 20, 1979; therefore, this letter does not require further j response regarding this matter. In accordance with Section 2.790 of the NRC's " Rules of Practice," Part 2, Title 10, Code of Federal Regulations, a copy of this letter and the enclosed inspection report will be placed in the NRC's Public Document !,,,1 Room. If the report contains any information that you believe to be proprietary, it is necessary that you submit a written application to
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ATTACH 24ENT 3 U ,) Texas Utilities Generating Company March 14, 1979 this of fice, within 20 days of the date of this letter, requesting that it~ such information be withheld frem public disclosure. The application { must include a full statement of the reasons why it is claimed that the ,i information is proprietary. The application should be prepared so that any proprietary inferr.ation identified is contained in an enclosure to the application, since the application without the enclosure will also M be placed in the Public Document Roem. If we do not hear from you in this regard within the specified period, the report will be placed in I the Public Occument Room. Should you have any questions concerning this inspection, we will be pleased to discuss them with you. Sincerely, _d W.C.Sei'ij, Chief d Reactor Cbnstruction and Engineering Support Branch
Enclosure:
l IE Inspection Report flo. 50-445/79-03 50-446/79-03 cc: w/ enclosure Xj,-. Texas Utilities Generating Company ATTil: Mr. H. C. Schmidt, Project Manager 2001 Bryan Tower Dallas, Texas 75201 ~ l l '. _ j J ~ l.. $1 }(Q lt4 1
-- 8 O U. S. NUCLEAR REGULATORY CCMMISSION OFFICE OF INSPECTION AND ENFORCEMENT REGION IV ,y + 0-l Report No. 50-445/79-03; 50-446/79-03 J l Docket No. 50-445; 50-446 Category A2 1j Licensee: Texas Utilities Generating Company 2001 Bryan Tower .._1 callas, Texas 75201 Facility Name: Comanche Peak, Units.1 & 2 Inspection at: Comanche Peak Steam Electric Station, Glen Rose, Texas Inspection condu;ted: February 2-23, 1979, and meeting at TUGC0 Corporate Office on February 2,1979 Inspector: (( .3//3/79 . G. Taylor, Resioent Reactor Inspector Date Projects Section Other Accompanying -- l Personnel at Corporate Meeting on February 2,1979: W. C. Seidle, Chief, Reactor Construction and Engineering Support Branch '-4 R. E. Hall, Chief, Engineering Support Section -. - 1 /3/79 Approved: e' j W. A. Crossman, Chief, Projects Section Date i i Inscection Summary: lp Inscection on February 2-28,1979 (Recort No. 50-445/79-03; 50-446/79-03) i Areas Inscected: Routine inspection by the IE Resident Reactor Inspector (RRI) of safety related construction activities including installation and j welding of reactor coolant and other piping systems; structural building l activities; construction fire protection; and follow up on licensee identi- !] fled items. The inspection involved seventy-five inspector-hours by one NRC. inspector. Resul ts : One item of noncompliance (infraction - failure to follow concrete placement procedures - paragraph 9) was identified in one of the ten con-struction areas inspected. I 1 ~...
1 ATTACEENT 8 h ) DETAILS 1. Persons contacted _3-- 'l ' Princioal Licensee Noloyees
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- J. B. George, TUSI, Project General Manager
- R. G. Tolson, TUGCO, Site QA Supervisor 8
- J. V. Hawkins, TUGC0/G&H Product Assurance Supervisor
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- 0. N. Chapman, TUGC0 Quality Assurance Manager C. C. Buffkin, TUGCO/G&H QA Welding Specialist I
- D. E. Deviney, TUGC0 QA Technician
- R. J. Gary, TUGCO, Executive Vice President & General Manager
- L. F. Fikar, TUSI, Executive Vice President & General Manager Other Personnel q
. M
- H. O. Kirkland, Project General Manager, Brown & Root (B&R)
W
- U. D. Douglas, Construction Project Manager, B&R
- J. H. Magner, Construction Project Engineer, B&R Tne Resident Reactor Inspector (RRI) also interviewed a number of other site construction and quality control personnel during the course of the inspection.
- denotes those ir..'ividuals attending the management interviews.
] 2. Licensee Construction Ceficiency Recorts (10 CFR 50.55(e)) i j The licensee has determined that two of three potentially reportable 'j construction deficiencies discussed in IE Inspection Report 50-445/79-01 ; 50-446/79-01 are not formally reportable as 50.55(e) items. The status of the third item has not yet been determined. l a. The ' licensee notified RIV by letter, dated February 6,1979, that the problem of leakage through a designed expansion joint would 1 l not meet the criteria of 50.55(e) and was, therefore, not con-sidered : eportable. The RRI reviewed the referenced supporting iM documentation which essentially indicated that while leakage ~ Ct would occur, the worst case level of contamination in the leakage ? would be below the limits imposed by 10 CFR 20 Appendix B, Table 1.
- d T.4e RRI discussed the licensee's analysis with knowledgeable NRC j
personnel who concurred. This item is considered closed.
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ATTACIDEN"' 3' L.) b. The licensee notified RIV by letter, dated February 20, 1979, that the matter of the reversed reactor coolant loop elbows was not reportable. The RRI reviewed the referenced docu=en-tation which concludes that the reversal is readily corrected E by addition of weld metal at the connecting joints and that the initial report was premature since appropriate subsequent m' QA ard construction actions had not yet taken place. The evaluation by the licensee concluded that the item was not reportable within the context of 50.55(e). The RRI had no M further questions and the item is considered closed. 3. Mislocated Reactor Vessel Succort Structure l The licensee reported to the RRI on February 20, 1979, that a major error had been detected in the design of the Unit 2 reactor vessel support structure. It had been determined that the reactor vessel support shoes, their ventilation duct work, and the surrcunding a reinforcing steel had been rotated forty-five degrees from the -] correct positions through a design error. As a result of the rota-tion, the reactor vessel would not match the vessel support feet ~i nor would the piping system to the other reactor loop ccmponents fit. The licensee concluded that since the vessel could not be installed and connected this precluded the deficiency from being reported as a 50.55(e) item, but was necessarily of concern because significant design and construction changes would be required to accommodate the vessel and its connecting piping. As of the date of this report, it has been detemined that the existing structure can be modified to reorient the vessel correctly to the balance of the system. The exact details of the modification, however, are not known at this time. The RRI will follow the modification effort in detail until it has been concluded. ~ 4. Plant Tours l The RRI toured one or more plant areas several times weekly during i the reporting period to observe the progress of construction of i safety related structures and the installation of equipment. Two i of the tours were conducted during second shift hours of work. The RRI observed that housekeeping practices have improved during the period, apparently through the ccmbined efforts of the construction My~ ' and quality control forces. The RRI has been infomally advised that a general housekeeping program is now operative, although undocumpted. The work force and work effort continues as previously indicated.- .},g No items of noncompliance or deviations were identified. ) i jfIE Inspection Report No. 50-445/79-01; 50-446/79-01 3-
ATTAC9ENT 8 O ) 5. Construction Fire Protection The RRI observed, during various plant tours, that installed equipment is adequately protected from incidental damage frrm probable heat sources such as area welding operations. protection is primarily by 25=j covering of equipment with plywood and fire retardant plastic and by l availability of properly charged portable fire extinguishers. Inter-views with randomly selected craft personnel indicated an awareness of where the fire extinguishers were available. i The RRI has no further questions regarding fire orotection at this j stage of construction. 6. Reactor Vessel and Internal Protection The RRI observed that the Unit i vessel internals remained stored in site warehouse facilities and continue to be protected from probable contaminants. The RRI observed that the prefabricated steel reactor vessel cover in Unit I was removed on or about February 27, and replaced with a wood beam and plastic cover. All vessel no::les are now partially or completely welded to piping. Access control and personnel logging for work on or in the vessel continues in effect. i The RRI reviewed plans and procedures related to receipt, off-load'ing and temporary storage of the Unit 2 reactor vessel scheduled to arrive early in March 1979. Records of tests and maintenance indicate that the handling equipment proposed has met all requirements and that planned storage is consistent with the supplier's (Westinghouse) j recontendations - No items c f noncompliance or deviations were identified. 7. Reactor Coolant System Installation and Weldina The RRI observed the welding of Control Rod Drive Mechanisms (CRCM) to the reactor vessel head CRCM no::les during this period. The total construction effort involved seventy-eight assembly and welding i operations with associated nondestructive and hydrostatic testing i over a period of approximately seven days. The welds involved are nonstrength seal welds referred to as canopy' welds and falling under J the provisions of ASME, Section III, Article NB-4360. The welding observed was accomplished using an automatic gas tungsten machine with an especially designed welding head powered by an " Astro-A.c" control and supply unit. The consumable insert, the only weld filler a .-r.-+ '( -4
} ATTACE'4ENT 8 O .) ^ metal involved, was tacked in place prior to CRCM assembly by welder AFX while the machine was operated by welder ABT during the time of observation. The machine welding was being performed in accordance with Weld Procedure Specification (WPS) 99029, Revision 0 as evidenced by the control unit settings and the strip chart recorder attached. =K Review of WPS 99029 indicated that the procedure and three welders, including AFX and ABT, had been qualified as required by the above referenced article of ASME, Section III. Each of the welds involved ); was authorized and documented by individual Weld Data Cards. The RRI reviewed seven of the cards and found them to have been ccmpleted ( consistent with project procedures. j The RRI also observed a portion of the welding operations on FW-23, .j which joins the cold leg reactor pipe to the coolant pump in loop 3. t The welders observed were BBI, AXB and AZF using a Dimetrics auto-matic TIG machine in accordance with WPS 99028, Revision 1. The weld filler metal was of Heat No. 434788 as evidenced by Weld Material Requisitions A175373 and A175374 and vendor identification on the wire supply reel. The machine control panel settings were verified as being consistent with WPS 99028 requiraments for the h.g observed weld pass; i.e., number seven. The RRI reviewed the _~ qualification documents relative to the above identified persons and found them to be in accordance with ASME, Section IX requirements. The RRI reviewed acceptance level radiographs for Reactor Coolant Loop welds FW-10, FW-26 and P4-31 during the report period. These welds join piping to the loop 2 steam generator hot leg; the loop 4 steam generator hot leg, and loop 4 pump to cold leg, respectively. The radiographs indicated acceptable weld quality in accordance with ASME, Section III and~ displayed sensitivity as required by ASME, Section V. J No items of nonccmpliance or deviations were identified. I 8. Other Safety Related Installation and Weldina i The RRI observed several piping installation and welding operations during the period with emphasis on stainless steel components. It was observed that in all instances the pipe ecmoonents, including j i valves, were handled in accordance with project procedures to prevent (M carbon steel contamination via the use of nonmetalic slings, use of .q marked "for stainless only" grinders, brushes and files. l Two welds were selected for in-depth observation and document review. i I The RRI observed welder ARP working on repairing FW-5 as shown on I i isometric BRP-CT-1-SB-09-0 in Containment Spray line 16"-CT-1-SB-( ) 014-301 R2. The welder was utilizing WPS 88023, Revision 4, a manual .--=,_;
~ ~ ATTACFjENT 8 O 3 TIG process. Filler metal was ER308, Heat tio. 463730. Review of documentation indicated that repair involved removal of five zones of root lack of fusion totaling approximately 3.3 inches in about 50 inches of original weld. 5 The RRI observed welder AGP working on joint PA-2B identified on =" isometric BRP-RH-1-SB-02-0 in line 12-RH-1-003-60lR2, a Residual m Heat Removal system line. This weld was the second full replacement weld for an original weld. WPS 88021, Revision 0 was being utilized a with filler metal Heat tio. 463730. I j The RRI subsequently verified that welders ARP and AGP were both qualified to the welding processes involved in accordance with ASME, Section IX. Documentation for the off-site fabricated spool pieces being joined by the above welds indicated that the items met l Project Specific'. tion MS-43A and ASME, Section III requirements. The RRI also reviewed radiographs of the following safety system ~ welds: _A y Weld Isemetric Line { Pd-12A BRP-SI-1-SB-14 4-SI-1-300-150lP2 FW-13A BRP-SI-1-SB-14-5 4-SI-1-045-150lP2 P4-5 BRP-CT-1-SB-05 16-CT-004-151 R2 2.lf The RRI observed the conduct of liquid penetrant examination of '~ PA-3, isemetric BRP-CS-1-SB-01-0 in line 8-CS-1-063-151R2. All [d of the identified nondestructive examination efforts indicated
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acceptable weld joints with examination techniques consistent with ASME Code requirements. -f j tio items of nonccmpliance or deviations were identified. ? l 9. Concrete Construction activities i The RRI selected Placement tio. 201-5805-019 in the Unit 2 containment as typical of concrete construction activities for the period. The s concrete batch plant was again batching out ten cubic yard loads of 'w e Design Mix 133 concrete from the automatic stationary mixer utilizing ) calibrated scales. The site test laboratory personnel were observed i. at both concrete pumping stations taking required field tests of slump, air content and temperature at the correct frequency and utilizing _j project proceduralized techniques. The RRI verified that the placement j fonnwork was tight and had adequate clearance from the reinforcing steel l
.~..__ _ _ __ _ _ _ ATTACEME:IT 8 p* ') .w to provide specified concrete cover; and that the lower preceding placement was clean and damp. It was observed that the placement crews were consolidating the freshly deposited concrete in a thorough but not excessive manner. {.- Af ter a change frem one deposit chute to another, the RRI observed that the falling concrete frem the new chute was aimed such that a portion of the stream impinged on a crcss piece (shear tie) rein- .)j forcing steel element which shredded the stream and caused the coarse aggregate to segregate frcm the stream and fly around the placement area. The RRI pointed out the situation to the Brcwn & Root QC inspec-l tar who ordered the crew to stop and add additional length to the chute ,j (elephant trunk) to get the concrete down through the rebar without direct impingement. The RRI verified that the other placement crew nad I ~ also added the elephant-trunk extenders after a short shutdown. Since neither the Brown & Root crew nor the attending QC inspector had taken any irm:ediate corrective action without being prcmpted by the RRI, it appeared that an undocumented nonconformance to project specifications =-M 2 and industry standards had occurred and might have been allowed to M continue. N ~~ - ACI-301-72, paragraph 8.3 as invoked by Project Specification 2323-SS-9, 3 'y " Concrete," prohibits any placement procedure which will cause segre-gation. Project QC procedure QI-QP-ll.0-3 also references SS-9 as criteria with an inspection checkpoint for segregation. The RRI con-sidered this to be in nonccmpliance with 10 CFR 50, Appendix B and so informed the licensee. 10. Management Meetings c. The RRI met with licensee site representatives on February 14, 15, .F 22 and 23, 1979, to discuss inspection findings and the licensee's j. plans for modification to the Unit 2 containment for the problem identified in paragraph 3. The RRI also accompanied the RIV Reactor Construction Branch Chief and Projects and Engineering Support Section Chiefs during a meeting on February 2,1979, at the licen-see's corporate headquarters to discuss safety related piping system i i welding problems. The RIV representatives indicated that welding i I reject rates with the acccmpanying repair cycles had created concern l l since a repaired weld is frequently not as satisfactory as an original 5 weld due to metallurgical effects in the heat affected zone adjacent is to the weld. The licensee stated that the site construction manage-N+j ment had developed a program endorsed by licensee management that would improve the situation as follows: a. Improve the welder training and qualification program beyond that required by the ASME Code such that qualification testing would be more representative of field production conditions. >veeg
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ATTAC'912NT S 5 '. .3 .) b. Tighten the administrative production controls such that immediate identification of inadequately trained welders could be achieved; these to be retrained as necessary. c. Provide highly qualified welding technicians to the field to 7: p work with the welders in a training and-advisory capacity. i"' The licensee management acknowledged their concern for the reject M rate and expressed hope that the above program would substantially reduce the rate in the near future. The licensee representatives __ s emphasized that welds accepted had met all requirements of ASME i Code and FSAR commitments; a condition acknowledged by the RIV t representatives. .. g i w m i i
f\\ ATTACain 7 g a Mcg'o ga / UNITED STATES !\\ c ^,% NUCLEAR REGULATORY COMMISSION
- q WASHINGTON, D. C. 20555 IVED I
Ja / M 15 G hfAY g g ygp. Nuc:.& DOCKET NOS. 50-445 AND 50-446 APPLICANT: TEXAS UTILITIES GENERATING COMPANY FACILITY: COMANCHE PEAK STEAM ELECTRIC STATICN, UNITS 1 & 2
SUBJECT:
SUMMARY
OF MARCH 2'7, 1979 MEETING ON REPAIR OF REACTOR VESSEL SUPPORT PEDESTAL FOR COMANCHE PEAK, UNIT 2 Summary A meeting was held with representatives of the Texas Utilities Generating Company on March 27, 1979 in Bethesda, Maryland. The purpose of the meeting was to dis-cuss the repair procedures for relocating the misoriented vessel support pads on Comanche Peak Steam Electric Station, Unit 2. The discussion was directed primarily at how the design of the repaired pedestal will differ from the original design of the reactor vessel support pedestal. The applicant des-cribed the repair design for the load carrying members as structurally equiva-lent and physically similar to those on the original design on Unit 1. A meeting attendance list is enclosed. Backcround The misorientation of the reactor vessel support structure at Comanche Peak, Unit 2 was described in our " Preliminary Notification of Event or Unusual Occur-rence" PNO-79-028, issued by the Office of Inspection and Enforcement on February 22, 1979. The applicant reported that the reactor vessel support pede-stal was being constructed such that its mounting pads would not mate with the support pads on the reactor vessel. The architect-engineer designed many features of the Unit 2 containment building, including the reactor vessel pede-stal, as a mirror image of Unit 1. However, the nuclear steam system supplier did not design the primary coolant systems mirror image for this two unit station. The reactor vessel supplied for Unit 2 is a duplicate of the reactor vessel installed in Unit 1. As a result, the mounting pads on the Unit 2 pede-stal are misoriented around the reactor vessel's vertical axis by about 45 degrees. The applicant elected to modify the top of the pedestal to install the mounting pads in the prop'er orientations. Constructicn had progressed to elevation 819 feet, several feet below the support pad, when the error was dis-covered. RECEIVED f.iAY 30 m - IUC @ CAc1 i
ATTACHP.ENT 9 l' . EY 151979 feetinc Details - The original design for the Unit 2 pedestal anticipated that the vessel would be supported on four pads which are approximately equally spaced around the periphery of the vessel. The revised design shifts the location of each pad 45 degrees about the vertical axis of the reactor vessel. The reactor vessel support pedestal is a reinforced concrete structure which by design includes a greater density of reinforcing steel beneath the four vessel support pads. Shifting the support pads 45 degrees therefore requires the installation of additional reinforcing steel under the new pad locations. The applicant was strengthening the pad locations by drilling 64 holes in the face of the pede-stal under each new pad location and grouting in #11 rebar. The holes are spaced on about 10 inch centers and extend almost five feet into the pedestal. Each new pad position also will have about 12 vertical #9 rebar installed in a similar fashion. The applicant states that the steel installed beneath the new pad position is equivalent to the initial design for this service, and that the new pads are designed to support the same loads as the pads on Unit 1. The applicant advised that the modification of the Unit 2 pedestal required that the ventilation ducts to the reactor vessel supports be extended to the new support locations. This did not cause any change in the ventilation systems performance, and was of no safety significance. The applicant considers this repair of the Unit 2 reactor vessel support pede-stal a field design change, having no safety impact. The applicant further does net consider the repaired pedestal to be structurally different from the Unit i design and therefore does not plan to make separate structural load analyses for each unit. No unresolved safety concerns associated with the repair design for the Unit 2 pedestal were identified at the meeting. yq pikmr. fu L>..a e Sgottswood B. Burwell Light Water Reactors Branch No. 2 ( Division of Project Management Enclosure : 1 Attendance List ces w/ enclosure: See next page i
( ATTACF.'4ENT 9 Texas Utilities Generating Company ccs: Nicholas S. Reynolds, Esq. Debevoise & Libernan 1200 Seventeenth Street ' Washington, D.C. 20036 Spencer C. Relyea, Esq. Worsham, Forsythe & Sampels 2001 Bryan Tower Dallas, Texas 75201 Mr. Homer C. Schmidt i Prcject Manager - Nuclear Plants Texas Utilities Generating Company i 2001 Bryan Tower Dallas, Texas 75201 Mr. H. R. Rock Gibbs and hill, Inc. 393 Seventh Avenue New York, New York 10001
- 1 Mr. A. T. Parker Westinghouse Electric Corporation P. O. 50x 355 Pittsburgh, Pennsylvania 15230 Mr. R. J. Gary Executive Vice President and General Manager Texas Utilities Generating Company 2001 Bryan Tower Dallas, Texas 75201 i
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~ ~'- ATTACE E::: 9 Encicsure PJLY 15 59 ATTENDANC E_. LIST COMANCHE PEAK STEAM ELECTRIC STATION MARCE 27, 1979 NRC - STAFF S. S. Burwell ~F. Rinaldi R. E. Shebaker . TEXAS UTILITIES GENERATING r.0MDANY H. C. Schtridt GIBBS & HILL H. R. Rock-E. G. Gibson E. L. Berkar Cherim Zion 8 l l 1 I J ) om-4 ,...c.m., - - r-- _.-r--, -,ne---m-e- ---+3 e e- ,-ser=-r -rwe--~--e*"N*' ' " " " ' ' -T'"' - " Y ' " - ^ * ' ~ ' " - " " ' - - - - * ' " " '}}