ML20136B657
| ML20136B657 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 08/31/1982 |
| From: | Madsen G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | Herr R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| Shared Package | |
| ML20136B196 | List:
|
| References | |
| FOIA-84-779 NUDOCS 8511200257 | |
| Download: ML20136B657 (2) | |
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MEMORANDUM FOR:
R. K. Herr, Director, Investigation THRU:
J. T. Collins, Regional Administrator J. E. Gagliardo, Director, Division of Resident, Reactor Project and Engineering Programs FROM:
G. L. Madsen, Chief, Reactor Project Branch 1
SUBJECT:
ALLEGED IMPROPER WELD PRACTICES AT CPSES, (DN:
50-445; 50-446) Q4-82-0005 A report of inquiry dated August 2,1982, (attached) requested an evaluation by the technical staff.
The results of our efforts are outlined in the attached letter from R. G. Taylor to G. L. Madsen, dated. August 15, 1982.
On the basis of the above, I support Taylor's recommendation that follow-up is not warranted.
. O :;ir.s! si: red by 1
L.L.I'A03CNMadsen, Chief G.G Reactor Project Branch 1
Attachment:
As stated cc: d. T. Collins J-F. Westerman G. L. Madsen 8511200G257 8510Q5 PDR FOIA GARDEB4-779 PDR RPB1 DRRP&EP RA-RIV GLMadsen JEGagliardo JTCollins 8/ /82 8/ /82 8/ /82 B -3
9.. ( 2 - ca r Alleoation No. A-82-57 (Dati Opened 10/ua)
ALLEGATION REVIEW
SUBJECT:
IRREGULARITIES IN RADIOGRAPHY FA:ILITY AND D3CKET N;. CP5ES A.EGER NO'E:
" John' ADDRESS:
TELE. N;-:
POSITION / TITLE:
SOURCE OF ALLEGATION:
From SRI DETAILS: Allecer has made several allecations reaarding radicaraphic operations a: tne CPSES.~ (1 ccrre:tfilmdens)ity;Automaticfilrprocissorspeedhasbeenadjustedtogetthe (2) 'T' holes in penetrameters have been reamed to ia ger size to give' impression of adequate sensitivity; and (3) Defective welds have Deen maske out wnen shooting adja:ent thin wall repairs.
A: tion Assigned To: Incuiry (Inv.)
Disposition:
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_- J April 25, 1983 TXX-3658 Mr. G. L. Madsen, Chief Reactor Project Branch 1 U. S. Nuclear Regulatory Comission Office of Inspection and Enforcement 611 Ryan Plaza Drive, Suite 1000 Docket Nos.: 50-445 Arlington, Texas 76012 50-446 COMANCHE PEAK STEAM ELECTRIC STATION RESPONSE TO NRC NOTICE OF VIOLATION INSPECTION REPORT N0. 83-03/01 FILE N0.:
10130
Dear Mr. Madsen:
We have reviewed your letter dated March 28, 1983 on the inspection conducted by your Senior Resident Reactor Inspector, Mr. R. G. Taylor, of activities authorized by NRC Construction Pemits CPPR-126 and CPPR-127 for Comanche Peak, Units 1 and 2.
We have responded to the finding listed in Appendix A of that letter.
To aid in the understanding of our response, we ' ave repeated the requirement and your finding followed by our cor.ective action. We feel the enclosed information to be responsive to the Inspector's finding.
If you have any questions, please advise.
N Very truly yours, RJG:aq IQ5a M'
Enclosures cc: NRC Region IV - (0 + 1 copy)
Director, Inspection & Enforcement (15 copies)
U. S. Nuclear Regulatory Commission Washington, DC 20555 B -H W
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TXX ~4658 April 25, 1983 APPENDIX A NOTICE OF VIOLATION Texas Utilities Generating Company Docket: 50-445/83-03 i
Comanche Peak Steam Electric Station 50-446/83-01 i
Permits: CPPR-126 i
CPPR-127 i
Based on the results of an NRC inspection conducted during the period of October 1982 through February 1983, and in accordance with the NRC Enforcement Policy (10 CFR Part 2, Appendix C), 47 FR 9987, dated March 9,1982, the following violation was identified:
Failure to Imolement a Ouality Assurance Program for the Fabrication ano Installation of Electrical Underwater Flooolight Pole Assemblies Criterion II of Appendix B to 10 CFR 50 requires that the applicant shall identify the structures, systems, and components to be covered by the quality assurance program and that the program shall provide control over activities affecting quality of the identified structures, systems and components.
FSAR Section 1A(b) commits the applicant to compliance with NRC Regulatory Guide 1.29 which in paragraphs 2 and 4 require the applicant to identify those structures, systems, and components whose continued function is not required (in a design basis accident) but whose failure could reduce the functioning of any plant feature identified in other paragraphs to an unacceptable level.
Contrary to the above, the Senior Resident Inspector-Construction has determined from investigation of allegations, observation of construction activities and review of design drawings that group of devices collectively identified as " Electrical Underwater Floodlighting Poles" (Drawing 2323-EL-0925-02) were not identified as required by Regulatory Guide 1.29 and were not included within the licensee's Quality Assurance Program.
Mechanical failure of the devices in a seismic event could damage fuel during reactor core installation activities or in the spent fuel storage pools, although the possibility of such mechanical failure of the pole assesly resulting in damaging
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fuel is very remote due to the design of upper and lower pole retention
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devices.
This is a Severity Level V Violation.
(Supplement II.D.)
CORRECTIVE STEPS WHICH HAVE BEEN TAKEN AND THE RESULTS ACHIEVED The drawing governing installation of the light poles has been changed via Design Change Authorization to reflect Seismic Category II requirements. This t
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TXX-3658 April 25, 1983 change causes the light poles to be inspected in accordance with the CPSES QA i.
Program. Any discrepancies discovered have been or will be documented and resolved in accordance with established procedures.
CORRECTIVE STEPS WHICH HAVE BEEN OR WILL BE TAKEN TO AVOID FURTHER VIOLATIONS Items designated "Non-Nuclear Safety Related" have been previously reviewed for
- i applicable Seismic Category II requirements and upgraded as needed.
Considering the scope of the evaluation and subsequent results, adequate confidence has been established that appropriate QA Program controls are in effect.
DATE OF FULL COMPLIANCE The required drawing change was accomplished on February 4,1983.
Inspection activities are on-going and will be completed and resolved consistent with the project schedule.
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k'd 14 1933 In Reply Refer To:
Dockets:
50-445/83-07 50-446/83-04 Texas Utilities Generating Company ATTN:
R. J. Gary, Executive Vice President & General Manager 2001 Bryan Tower Dallas, Texas 75201 Gentlemen:
This refers to the inspection conducted by Messrs. L. D. Gilbert and C. E.
Johnson of our staff during the period January 27-28 and February 1-2, 1983, of activities authorized by NRC Construction Permits CPPR-126 and CPPR-127 for the Comanche Peak facility, Units 1 and 2, and to the discussion of our findings with Mr. R. G. Tolson of your staff at the conclusion of the inspection.
Areas examined 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 inspectors.
During this inspection, it was found that certain of your activities were in violation of NRC requirements.. Consequently, you are required to respond to this violation, in writing, in accordance with the provisions of Section 2.201 of the NRC's " Rules of Practice," Part 2, Title 10, Code of Federal Regulations.
Your response should be based on the specifics contained in the Notice of Violation enclosed with this letter.
Two new unresolved items are identified in paragraph 5 of the enclosed report.
In accordance with 10 CFR 2.790(a), a copy of this letter and the enclosure will be placed in the NRC Public Document Room unless you notify this office, by telephone, witnin 10 days of the date of this letter, and submit written application to withhold information contained therein within 30 days of the date of this letter.
Such application must be consistent with the requirements of 2.790(b)(1).
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The response directed by this letter and the accompanying Notice is not subject to the clearance procedures of the Office of Management and Budget as required by the Paperwork Reduction Act of 1980, PL 96-511.
Should you have any questions concerning this inspection, we will be pleased to discuss them with you.
Sincerely, Cr:. I:". c c/
.. I. 6. 'lin Grrr..n 1
G. L. Madsen, Chief Reactor Project Branch 1
Enclosures:
1.
Appendix A - Notice of Violation 2.
Appendix B - NRC Inspection Reports 50-445/83-07 50-446/83-04 j
cc w/ encl:
Texas Utilities Generating Company ATTN:
H. C. Schmidt, Project Manager 2001 Bryan Tower Dallas, Texas 75201 bcc to DMB (IE01) bec distrib. by RIV:
Resident Inspector Section Chief RPB1 RPB2 TPB MIS SYSTEM l
RIV File RA C. Wisner M. Rothschild, ELD TEXAS STATE DEPT. OF HEALTH Juanita Ellis i
David Preister L. Gilbert E. Johnson G. Vissing, NRR
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APPENDIX A j
NOTICE OF VIOLATION Texas Utilities Generating Company Docket:
50-445/83-07 Comanche Peak Unit 1 Permit:
CPPR-126 Besed on the results of an NRC inspection conducted during the period Jtnuary 27-28 and February 1-2, 1983, and in accordance with the NRC Enforcement un1 Policy (10 CFR Part 2, Appendix C), 47 FR 9987, dated March 9, 1982, the following violation was identified.
Failure to Follow Procedure Criterion V of Appendix B to 10 CFR Part 50 requires that activities 4
affecting quality shall be prescribed by documented procedures of a type appropriate to the circumstances, and shall be accomplished in accordance with the procedures.
8rown & Root, Inc., Construction Procedure CP-CPM 9.10, Revision 8, paragraph 3.2.4, specifies that the fit-up gap for "T" fillet joints should not exceed 1/16"; however, if the gap is in excess of 1/16", but not exceeding 5/32", this condition will be considered acceptable provided the leg of the fillet weld is increased by the amount of separation in excess of 1/16".
Contrary to the above, on January 28, 1983, the NRC inspector determined that Pipe Support Mark No. SW-1-102-106-Y33K had been welded with a fit-up gap of 1/4" at one end of the "T" fillet joint which attached piece 7 to piece 4 The 1/4" fit-up gap exceeds the 5/32" maximum allowable fit-up gap requirement of Procedure CP-CMP 9.10.
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This is a Severity Level V Violation.
(Supplement II.E) (445/8307-01) o Pursuant to the provisions of 10 CFR 2.201, Texas Utilities Generating Company hi is hereby required to submit to this office, within 30 days of the date of this Notice, a written statement or explanation in reply, including:
(1) the corrective steps which have been taken and the results achieved; (2) corrective stsps which will be taken to avoid further violations; and (3) the date when full compliance will be achieved.
Consideration may be given to extending your i
response time for good cause shown.
1 Dated:
March 14, 1983 onn,-Avn-3 O
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APPENDIX B U. S. NUCLEAR REGULATORY COMMISSION REGION IV Reports: 50-445/83-07 50-446/83-04 Dockets: 50-445 50-446 Category A2 Licensee:
Texas Utilities Generating Company 2001 Bryan Tower Dallas, Texas 75201 Facility Name:
Comanche Peak, Unit 2 Inspection At:
Comanche Peak Steam Electric Station Inspection Conducted:
January 27-28 and February 1-2, 1983 k_
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's L. D. Gilbert, Reactor Inspector, Engineering Dite/
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E.~ Johnson, Reactor Inspector, Engineering Dtte /
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Section (paragraphs 1, 5, 6, and 7) t, Reviewed:
[2 he/ '3 T. F. Westerman, Chief, Reactor Project Section A Date Approved:
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Inspection Summary Inspection Conducted January 27-28 and February 1-2, 1983 (Report 50-445/83-07)
Areas Inspected:
Special, announced inspection to follow up on allegations on pipe and pipe support welds and protective coatings.
The inspection involved 48 inspector-hours by two NRC inspectors.
Results: Within the three areas inspected, one violation was identified (failure to follow procedure, paragraph 2.a).
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- Inspection Conducted January 27-28 and February 1-2, 1983 (Report 50-446/83-04)
Areas Inspected:
Special, unannounced inspection to follow up on allegations on pipe welding records.
The inspection involved 12 inspector-hours onsite by one NRC inspector.
Results:
In the area inspected, no violations or deviations were identified, i
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DETAILS 1.
Persons Contacted Principal Licensee Personnel
- R. G. Tolson, Site Quality Assurance (QA) Supervisor, TUGC0
- C. T. Brandt, Quality Assurance / Quality Control (QA/QC) Supervisor, TUGC0
- 0. N. Chapman, QA Manager, TUGC0 Other Personnel
- G. R. Purdy, Site QA Manager, Brown & Root (B&R)
W. E. Baker, Senior Project Welding Engineer, B&R E. Opelski, Site NDE Level III, B&R J. Patton, QC Mechanical Superintendent, B&R H. Williams, QC Superintendent, Non-ASME, B&R B. Wallace, Lead QC, Protective Coatings, B&R B. Dunham, Lead QC, Backfit Program, B&R The NRC inspectors also interviewed other licensee and contractor employees during the course of the inspection.
" Denotes those attending the exit interview.
2.
Followuo to Allegations on Pipe Support Welds in Unit 1.
A special inspection was performed to follow up on an allegation that three pipe support fillet welds were fit-up and welded with excessive gap.
The three pipe supports were identified as:
SW-1-012-010-A33R, SW-1-102-106-Y33K, and CC-1-087-004-A33A.
a.
A documentation review of Pipe Support Mark No. SW-1-102-106-Y33K provided the following information.
The support is a safety-related ASME Code Class 3 NF support for the service water piping system.
The weld attaching piece 7 to piece 4 is classified as a "T"
fillet joint that was welded by welder BLU in August 1981 using Welding Procedure WPS 11032 and Construction Procedure CP-CPM 9.10.
To permit inspection of the fit-up gap, a portion of the weld was removed from the fillet weld designated on Pipe Support SW-1-102-106-Y33K alleged to have an excessive gap.
The NRC inspector measured the gap as being 1/4".
The 1/4" gap exceeds the 5/32" maximum allowable fit-up gap requirement of Procedure CP-CPM 9.10 for "T"
fillet joints.
This is an apparent violation of 10 CFR 50, Appendix B, Criterion V, for failure to follow procedure.
The allegation, as it relates to Pipe Support SW-1-102-106-Y33K, was substantiated.
2,
. b.
A documentation review of Pipe Support Mark No. SW-1-012-010-A33R provided the following information.
The support is a safety-related ASME Code Class 3 NF support for the service water piping system.
The weld attaching piece 18 to piece 19 is classified as a "T" fillet joint.
The support was welded by welder BLU in November 1980 using Welding Procedure WPS 11032 and Construction Procedure CP-CPM 9.10.
To permit inspection of the fit-up gap, a portion of the weld was removed from the fillet weld designated on Pipe Support SW-1-012-010-A33R as having excessive gap.
The NRC inspector measured the gap as being 1/16".
The 1/16" gap meets the fit-up gap requirements of Procedure CP-CPM 9.10 for "T" fillet joints.
No violations or deviations were identified.
The allegation as it relates to Pipe Support SW-1-012-010-A33R was not substantiated.
c.
A documentation review of Pipe Support Mark No. CC-1-087-004-A33A provided the following information.
The support is a safety-related ASME Code Class 3 NF support for the component cooling water piping system.
The weld attaching piece 45 to piece 55 is classified as a partial penetration "T" joint.
The support was welded by welder BMG in February 1981 using Welding Procedure WPS 11032 and Construction Procedure CP-CPM 9.10.
In reviewing the joint design and fit-up requirements for the partial penetration "T" joint, the NRC inspector determined that the joint design requires a 3/8" bevel on the 1/2" thick material and a 0-3/16" fit-up gap which would give the appearance of having an excessive fit-up gap for a "T" fillet joint.
Also, it should be noted that the strength of the partial penetration "T" joint would not be affected by increasing the fit-up gap, where as, the strength of the "T" fillet joint could be affected by increasing the fit-up gap.
No violations or deviations were identified.
The allegation, as it relates to Pipe Support CC-1-087-004-A33A, was deemed to be without technical merit.
3.
Followup to Allegations on Pipe Welding in Unit 1 A special inspection was made to follow up on an allegation that a weld, identified as FW 1-B on Drawing CT-1-SB-017, contained unconsumed con-sumable insert material hanging down on the interior of the pipe and excessive radial shrinkage which could ca'use flow problems.
. After performing an inspection of the weld and reviewing the radiographs of the weld, the NRC inspector determined that the material hanging down approximately 1/16" was caused by melting through the weld root for an area 3/32" in diameter.
The melt-thru condition meets the ASME Code requirements for weld reinforcement and radiographic quality.
Radial shrinkage occurs in all pipe weldments, but its effects are predominate in large bore, thin wall stainless steel piping.
Weld FW 1-B does exhibit radial shrinkage which was accepted by QC on final visual inspection and adequate flow within the pipe will be demonstrated during the preoperational test of the containment spray system (Preoperational Test 1-CT-PT-48-01).
No violations or deviations were identified.
The allegation, although partially substantiated, is without technical merit.
4.
Followuo to Alleoations on Pipe Weldina in Unit 2 A special inspection was made to follow up on an allegation that the weld do.cumentation for Weld FW 1 on Drawing MS-2-RB-19 for the main steam system has been lost and reconstructed, but does not include the two repairs and, therefore, the weld may be defective.
The NRC inspector reviewed Nonconformance Report NCR M-3425, Revision 1, which stated that the weld data card package had been lost and disposi-tioned the weld to "USE AS IS" based on other information being available to reconstruct the weld data card sufficiently to satisfy the ASME Code requirements for weld documentation.
After reviewing the nonconformance report, reconstructed weld data card, and other available documentation which included weld rod issue records, weld material certification records, drawing requirements, piping material certification records, welder qualification records, procedure qualifica-tion reports, and radiographs and radiographic inspection reports for the weld; the NRC inspector determined that sufficient information was avail-able on the weld and two repairs to provide adequate assurance that the weld, as repaired, is of good quality and meets the ASME Code requirements for Class 2 welds.
No violations or deviations were identified.
The allegation that the weld data card was lost was substantiated, but is without technical merit.
5.
Protective Coatings The purpose of this inspection was to substantiate allegations concerning unsatisfactory coatings application in the Skimmer Pump Room and the Drain Tank Heat Exchanger in Reactor Building (RB) Unit 1 being accepted by QC
. as satisfactory.
There was also some concern of the use of manufacturer's recommendations when performing patch tests when using the Elcometer 106 Adhesion Tester.
The NRC inspector also reviewed the overall coatings program to determine if there were any other QA problems.
The inspection covered a review of procedures, records, observation of the areas in question; a review of the backfit, and ongoing in process program.
a.
Review of Coating Procedures The NRC inspector reviewed protective coatings procedures and specifications for the application and inspection of work in process and for the backfit program established.
The purpose of this review was to determine if the licensee had incorporated all necessary standards, codes, and manufacturer's recommendations into the QC procedures used in the field by QC inspectors.
During this review of procedures, the NRC inspector observed that Gibbs & Hill Specification 2323-AS-31, " Protective Coatings," stated in paragraph 9.2.2.e, in part, "An Elcometer 106 Adhesion Tester or equal shall be used for patch testing in accordance with the recom-mendations of the manufacturer of the instrument."
The NRC inspector then reviewed the manufacturer's recommendations.
Tne NRC inspector observed an instruction of the manufacturer's operating instruction that was not referenced in any of the QC procedures.
This instruction states, "When the adhesive curing time has elapsed, a cut is made around the dolly through the coating to the substrate (this is especially critical for electro plated coatings) with a sharp knife or special cutter obtainable from Elcometer Instruments Limited."
After a brief investigation and icterviews with QC personnel by the NRC inspector of this matter, it was found that no QC procedure referenced this instruction, nor were QC inspectors performing this function.
There was some doubt by the NRC inspector whether this procedure was necessary.
The NRC inspector informed the licensee of this finding.
The licensee immediately contacted the manufacturer to determine if this procedure was necessary.
Through a telephone con-versation, the manufacturer of this instrument stated that the procedure in question was not necessary, except for electro plated coatings, which is not the case at this site.
The NRC inspector has determined that this is an unresolved item until a letter from the manufacturer confirms this.
(445/8307-02)
The NRC inspector observed in Civil Construction Procedure (CCP)-30, paragraph 4.4.1.1, item 6, that cure time before topcoat shall be in accordance with Attachment 3, which contains graphs that specify hours of cure time with respect to temperature of Carbo Zinc II l
primer after application.
During a discussion with QC personnel, it
. 6 was found that QC inspectors check for ambient conditions before primer is applied and record time.
However, after primer is applied, the start of curing is not recorded on any inspection report (IR).
The NRC inspector has determined that if QC is to perform an accurate curing time in accordance with the graphs, they must record a starting time after application and an ending time for curing on the IR.
This will be considered an unresolved item.
(445/8307-03)
Procedures reviewed are listed below.
CP-CPM-13.1, " General Calibration Procedure" CP-QP-18.0, " Inspection Reports" CP-QP-2.1, " Training and Certification of Inspection Personnel" CCP-30, " Coating Steel Substrates Inside Reactor Buildings and Radiation Areas" QI-QP-2.1-4, " Qualification of Protective Coating Inspection Personnel" 2323-AS-31, " Protective Coatings" 35-1195-IEI-15, Revision 5, " Calibration of Mechanical Ory Film Thickness Gages" 35-1195-IEI-36, Revision 2, " Calibration of Coating Adhesion Tester" CCP-30A, " Coating Steel Substrates Inside Reactor Buildings and Radiation Areas" QI-QP-11.4-1, " Inspection of Steel Substrate Surface Preparation and Primer Application" QI-QP-11.4-5, " Inspection of Steel Substrate Primer Repair and Seal and Finish Coat Application and Repair" QI-QP-11.4-23, " Reinspection of Seal Coated and Finish Coated Steel Substrates For Which Documentation Is Missing or Discrepant" QI-QP-11.4-24, " Reinspection of Protective Coatings on Concrete Substrates for which Documentation is Missing or Discrepant" b.
Observations M&TE Calibration Shop The NRC inspector reviewed calibration procedures for the Mechanical Dry Film Thickness Gage (DFT) and the Elcometer 106 Adhesion Tester, as indicated in paragraph 5.a.
The inspector also observed the calibration of the Adhesion Tester by a technician.
The test was performed by the weight method, which is an acceptable test.
Calibra-tion records were up to date and the equipment was calibrated during the required intervals.
No discrepancies were noted.
.. Skimmer Pumo Room The NRC inspector and several QC inspectors visually inspected the Skimmer Pump Room and the R. C. Drain Tank Heat Exchanger Room in the reactor buildir.g at E1.808'-00".
This inspection was performed because of allegations concerned over unsatisfactory protective coatings being accepted as satisfactory.
The main concern was the floor in the heat exchanger room.
The NRC inspector and QC inspectors noticed some unsatisfactory areas on the floor,.however, some unsatis-factory spots had already been prepared for repair and the remainder had not because of the temporary stop work order.
During the course of the inspection, the NRC inspector located all irs related to coatings application in these areas.
It was found that proper documentation of unsatisfactory conditions were reported for these areas in question and work was being performed as corrective action.
There is no substantial evidence to this allegation in the areas of concern.
c.
Review of Records The NRC inspector reviewed many records of steel and concrete coatings applications and inspections.
Documentation was readily retrievable and legible.
There were no discrepancies observed.
irs reviewed are listed below.
IR#
NCR#
PCR 00879 C-82-0030 4
PCR 00874 PCR 00899 PCR 02274 PCR 00872 PCR 02246
- PC 48805
- PC 48760
- PC 48693
- PC 48721
- PC 48302
- PC 48670
- PC'48750
- PC 48190 l
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. *PC 48675
- PC 48747
- PC 48633
- Skimmer Pump Room /R. C. Heat Exchanger Room d.
Review of QC Personnel Qualifications The NRC inspector reviewed approximately 18 qualification records to determine that inspection personnel were qualified to inspect activ-ities in the area of protective coatings.
The inspectors were qualified in accordance with site procedures.
e.
Protective Coatings Program The overall coatings program consisting of the backfit program and ongoing in process program since report 81-15 shows substantial improvement in areas such as retrievable documentation and more specific inspection criteria for QC.
No violations or deviations were identified.
6.
Unresolved Items Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, violations, or deviations.'
An unresolved item related to testing of protective coatings 4
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is discussed in paragraph 5.a.
A second unresolved item, related to curing time for protective coatings, is discussed in paragraph 5.a.
7.
Exit Interview The NRC inspectors met with licensee representatives (denoted in paragraph 1) and R. G. Taylor (NRC resident reactor inspector) on January 27-28 and February 1-2, 1983, and summarized the scope and findings of the inspection.
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AUG 241993 In Reply Refer To:
i Docket: 50-445/83-24 50-446/83-15 i
1 Texas Utilities Generating Compary ATTN:
R. J. Gary, Executive Vice President & General Manager 2001 Bryan Tower Dallas Texas 75201 Gentlemen:
This refers to the inspection conducted by our Senior Resident Inspector, Construction, Mr. R. G. Taylor, during the period March through July 1983, of activities authorized by NRC Construction Permits CPPR-126 and CPPR-127 for Ccr.anche Peak, Units I and 2, and to the discussion of our findings with Mr. R. G. Tolson, and other members of your staff during the inspection.
Areas examined during the inspection included review, inspection, and evalua-tion of seve~ral allegations made to various NRC persons, including the Atomic Safety and Licensing Board in their proceedings regarding the operating license for Comanche Peak Steam Electric Station (CPSES).Within these areas, the I
inspection consisted of selective examination of procedures and representative records, interviews with personnel, and observations by the inspector. These i
findings are documented in the enclosed inspection report.
)'
During this inspection, it was found that certain of your activities were in violation with NRC requirements. You were notified of one such -violation by our letter of i%y 31, 1983, to which you have responded.
Details of the item enclosed with our MAy 31, 1983 letter are included in the enclosed inspection report.
One unresolved item is identified in paragraph 15 of the enclosed inspection report.
We have als%
o examined actions you have taken with regard to previously identified inspection findings.
The status of these items is identified in paragraph 2 of the enclosed report.
In accordance with 10 CFR 2.790(a), a copy of this letter and the enclosure will be placed in the NRC Public Document Room unless you notify this office, by telephone, within 10 days of the date of this letter, and submit written application to withhold infomation contained therein withir. 30 days of the date of this letter.
Such application must be consistent with the reouire-
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ments of 2.790(b)(1).
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Ccapany G g 41993 should you have any questions concerning this inspection, we will be pleased s
to discuss them with you.
1 Sincerely.
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c.L. MAcccN" 1
G. L. i'adsen, Chief
' Reactor Project Branch 1
Enclosure:
Appendix - fiRC Inspection Report 50-445/03-24 50-446/E3-15 cc w/encls:
Texas Utilities Generating Company ATTH: H. C. Schmidt, Project f:anager 2001 Bryan Tower Dallas, Texas 75201 Texas Utilities Generating Company ATTN: B. R. Clements, Vice President, Muclear 2001 Bryan Tcwer, Suite 1735 rellas, Texas 75201 bec to DM3 (IE01) bec distrib. by RIV:
RPB1 D. Kelley, SRI-Ops RPB2 R. Taylor, SRI-Cons y
i TPB SectionChief(RPS-A)
J. Collins, RA J. Gagliardo, DRRP&EP 1
C. Wisner, PA0 i
M. Rothschild, ELD
- i MIS SYSTEM RIV File 1
TEXAS STATE DEPT. OF HEALTH j
Juanita Ellis David Preister W.
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l APPENDIX U. S. NUCLEAR REGULATORY COMMISSION REGION IV NRC Inspection Report: 50-445/S3-24 4
50-446/83-15
-l
.j Docket: 50-445 Ca tegory:
A2 50-446 Licensee:
Texas Utilities Generating Company (TUGCO)
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2001 Bryan Tower i
Dallas, Texas, 75201 9
Facility Name:
Comanche Peak Steam Electric Station (CPSES), Units 1 and 2 3
Inspection At:
Comanche Peak, Units 1 and 2, Glen Rose, Texas Inspection Conducted: March through July 1983 bfX) vsw f-9// Ff83 Inspectors: 4 1
s R. G. Taylor, Senior Resident Inspector Date /
j Construction (SRIC)
Approyed:
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8// <//83 D. M. Hunnicutt, Chief Da te '
Reactor Project Section A Insoection Surrnary
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Insoection Conducted f'. arch throuah July 1983 (Recort 50-445/83-24 and 83-446/S3-15)
Areas Insoected: Special inspections, announced and unannounced, related to aliegations mace to various NRC persons including the Atomic Safety and i
Licensing Board in their procedings regarding the operating license for Comanche Peak Station.
The inspections involved 449 inspector-hours by one NRC inspector.
Results: The inspection confirmed the need to issue four violations initially icentified by the Construction Appraisal Team (CAT) (NRC Inspection Report 50-445/83-18; 50-446/S3-12).
These involved the areas of HVAC, Ecuipment i
Installation, Document Control, and Storage of Eouipment.
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2 Details 1.
Persons Contacted Principal Licensee Employees
- R. G. Tolson, Site QA' Supervisor '
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- C. T. Brandt, Non-ASME QC Supervisor
- J. R. Merritt, Engineering, Construction and Startup !% nager
- J. B. George, Project General Penger'
- D. N. Chapman, QA Manager
- B. R. Clements, Vice-President, Nuclear F
Brown & Root (B&R)
- G. R. Purdy, Project QA Manager
- D. Frankum, Construction Project Manager The SRIC also interviewed many other licensee, B&R, and subcontractor personnel during the course of the inspection.
- Denotes those persons who attended one or more management interviews with the SRIC.
2.
Licensee Action on Previous Insoection Findinas (Closed) Unresolved Item (50-445/82-22-02), " Analysis of Weld Discrepancies."
This unresolved item concerned a substantial number of identified defects in a large whip restraint essentially surrounding the mainsteam and feed water lines located several' feet outside of the ASt1E code boundry point.
The device was engineered by the licensee's A/E and manufactured by NPS Industries.
Due to the overall size of the structure, it has been nick-named " George Washington Bridge" by the site labor and quality forces.
The licensee had reported the finding of the defects as a potential 50.55(e) item to the SRIC on September 30, 1982, which was subsequently stated not reportable in a letter dated December 27, 1982.
An NRC inspector-followed up on the matter during a visit to the offices of the A/E, as documented
~in NRC Inspection Report 50-445/83-12.
This review pertained to all of the defects involved with the exception of two cracked welds that had not been analyzed at the time of the inspection.
The engineer has recently analyzed these two defects _and has determined that had they not 'been detected, the l
structure could have fulfilled it's function.
The SRIC has reviewed the-location of the cracks and their length in relation to the size of the welds and'the functional application of the structure.
Since the structure:has no continuous service application and.is essentially subject to a one-time loading, the cracks would not have the potential for further propagation.
Further, the cracks are at points-in the structure that would receive rela-tively low stresses in the one-time impact based on 'their small size in relation to the members being welded.
It appears that the cracks formed due-to the stresses developed during the tightening of high strength bolting in-e
3 the im.ediate vicinity of the welds during the site assembly of the structure.
Taken in conjunction with the earlier documented review of the engineers ~
calculations and the SRIC's review of these cracks, the SRIC has concluded that the engineer's overall analysis was adequate.and that deficiency (s) were not reportable under 50.55(e).
Both the licensee's initial report (CP-82-12) and the above identified unresolved item are considered closed.
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It shou 1d be noted for the record that this closure only applies to the reportability aspects under 50.55(e) and not to the correction of the defects.
The defects, including the cracks, h. ave been documented on a nonconformance report. The final disposition and closure of the NCR will be evaluated during future routine inspections.
3.
Review of Licensee Self-Evaluation (Usine INPO Criteria)
The SRIC has reviewed a report of the licensee's self-evaluation performed during October 1982 which was based on criteria that has been developed for i
the purpose by INPO.
The evaluation was performed in behalf of the licen-see by personnel in the employment of Sargent & Lundy, an architect-engineer firm with substantial nuclear power involvement.
A copy of the report was j
furnished to the NRC, and subsequently, to the Atomic Safety and Licensing Board in the matter of Comanche Peak Station operating license by letter l
dated May 2, 1983.
The purpose of the review by the SRIC was to determine if any of the 47 findings in the report were of a type and of sufficient significance to have been reported to the NRC as required by 10 CFR 50.55(e).
l The SRIC reviewed each of the 47 findings and the supporting documentation in the report pertaining to-each finding.
This review revealed that none
~ f the 47 items were based upon identified deficiencies in structures, o
i systems, or components nor were there any significant deficiencies in design, engineering, or testing that would constitute conditions reportable under 10 CFR 50.55(e).
4.
Car Wash In Containment l
During the limited appearance statement portion of the Atomic. Safety and Licensing Board hearing on P.ay 16, 1983, a person stated at transcript page 6152 that he understood that the containment looked something like a car wash. The person stated that it was his nderstanding that the situa-u tion developed at about the same time that there was a meeting at the D/FW Airport between the NRC and any interested parties to discuss NRC decen-tralization.
Tnat meeting took place on April 5,1983.
For the purposes of evaluating this allegation, the SRIC expanded the period of interest to include the 3 weeks ~ prior to the meeting.
During.this entire period, the Unit I reactor system was undergoing what is referred to as " Hot Func-l tional Testing".
Tnis particular test is an accurate simulation of the operation of the reactor system and its appurtenances but without a reactor l
core being in place.
The heat and pressure in the system is' generated by the reactor coolant pumps in conjunction with the chemical and volume con-l trol system charging ' pumps.
Tne test could readily be construed to be a pressure test but in fact is an operational test at pressure.
This-parti-l cular test extended overall for about 90 days beginning late in February r
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4 and continuing until late May. The SRIC monitored the test but was by no means continously in the containment.
The SRIC interviewed personnel in the licensee's startup test group, QC inspectors who had reason to be in the building and others to obtain a picture of the events that occurred in the Unit 1 Containment Building during the period of interest.
The SRIO also reviewed the licensee's control room logs for any indication of oper-ational prpblems indicative of a major leak in any of the fluid filled systems under test. The picture obta'ined was that there were several small 4
i leaks, generally at the gaskets between valve bodies and their bonnets.
In addition, there was a considerable amount of condensation dripping from the i
reactor coolant pump motor cooling coils.
This was caused by the cold water in the coils cond;nsing the humidity from the atmosphere within the building and was not indicative of a leak in the reactor coolant system.
The SRIO found from the control room logs that on March 29, a steam leak occurred during one phase of the test when a drain valve was partially open.
Perhaps this valve should have remained closed.
The room in which the valve was located was apparently filled with steam vapor which would have condensed out on the cooler walls as water. On March 30, the reactor vessel head vent valves were partially opened, which in turn would give some amount of steam blowoff into the reactor refueling cavity area and would rise up into the building until coo]ed and condensed out as water.
None of these events are typical of any major leak indicative of piping or piping component (such as a valve) failure. The type of small events described above are, within the experience of the SRIC, typical of what would be expected during such a test and is one of the reasons for performing the test.
5.
Desian of the HVAC System Succorts By letters, both dated March 11, 1983, Citizens Association for Sound Energy (CASE) notified the NRC's Offices of Inspection and Enforcement.and the Executive Legal Director of a concern that the HVAC system for Comanche Peak had not been properly supported, nor had it been properly considered in regard to seismic load conditions or its treatment as potential mis-s iles.
CASE specifically states that from their review of the FSAR, it appears that the licensee has not analyzed the HVAC supports for a seismic load condition.
Specific reference is made to Sheet 21 of Table 17A.
In addition, the personal observations of Messrs. Walsh and Doyle are relied upon to point out that there are no lateral supports on the HVAC systems within the containment.
CASE also states that all HVAC components and supports inside containment should be treated as missiles under Cri-terion 4 of the General Design Criteria for huclear Power Plants, 10 CFR 50, Appendix A.
Sheet 21 of Table 17A of the FSAR lists the containment ventilation sys-tems as being Seismic Category II. Apparently, it has been assumed by CASE that this category excludes seismic loading in the design.
This assumption is incorrect since the FSAR, Section 3.2.1.2 defines Seismic Category II as being those portions of systems or components whose m =
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5 continued function is not required but whose failure could reduce the func-tioning of any Seismic Category I system or component required to satisfy the requirements of C.I. A through C.1.Q of Regulatory Guide 1.29 to an unacceptable safety level or could result in incapacitating injury to occupants of the control room.
These systems are' designated Non-Nuclear Safety (NNS) Seismic Category II and are designed and constructed so that
- t. safe shutdown earthquake (SSE) will not cause such a failure.
' ~
CASE alsoitates that if the HVAC systems within the containment failed during a SSE, this would allow the temperature within the containment to rise quickly to unacceptable levels ~which could over time cause compon-ents and monitoring equipment to fail and which could also mean that it might be impossible for workers to enter the containment due to the heat.
Containment heat removal is required by Criterion 38 of the General Design Criteria for Nuclear Power Plants.
The system to remove heat from the reactor containment at Comanche Peak does not rely on the HVAC system but rather is composed of two separate containment spray recirculation trains each with 100 percent capacity.
Each train contains two separate pumps, one heat exhanger, and seven spray headers, and each system is fed from its individual electrical Class IE bus.
The containment heat removal system is designed to ensure that the failure of any single active compon-ent, assuming the availability of either onsite or offsite power exclusively, does not prevent the system from accomplishing its planned safety function.
CASE's concern with being able to enter the containment following certain design basis accidents is unfounded in that it is not a requirement.
In order to assess the adequacy of the design of HVAC supports, an inspec-tion was conducted at the home office of " Corporate Consulting & Develop-ment Company, LTD.," the support design consultant.
It was determined that all permanent HVAC supports are analyzed for seismic loading.
Two methods are utilized:
Zero Peak Accleration (IPA), or 1.5 Times the Peak-Accelera-tion When the Fundamental Frequency Falls Below 20 Hertz.
Of the latter
^
method of design, only about 6 out of 4000 supports have been designed that way.
A typical HVAC duct run is supported axially at every third support This may explain why Messrs. Walsh and Doyle may have felt that there were no lateral supports on the HVAC systems.
The NRC inspector reviewed the design of a typical HVAC duct run at elevation 852'-6" in the Auxiliary Building.
Supports were designed utilizing two computer programs entitled FEASA-2D and FEASA-3D.
The acronym stands for frame eigenvalue and stress analysis. The -2D version is used on the transverse supports and the -3D version is used on the axial supports.
The inclusion of equivalent weights from both up and downstream transverse supports and accesories such as vol-ume dampers and vane turns in the design of the axial supports was verified.
This inspection verified the adequacy of the siesmic design techniques being utilized for the design of HVAC supports at Comanche Peak.
The concerns expressed by CASE have been found to be without merit.
Persons contacted during the course of the inspection at Corporate Consulting J
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& Development Company, LTD. were:
J. Roland Yow, President & Chief Executive Officer Gary Hughes, Vice-President for Operations David Lindley, Principal Engineer Stephen Lehrman, Seismic Department Manager Daryl Hughes, Project Engineer 6.
Heatinc, Ventilation, and Air Conditionino System (HVAC)
During the CAT inspection (NRC Inspection Report 50-45/83-18;50-446/83-12),
the CAT inspectors noted that a significant portion of the welds on the ducting support structures were deficient in relation to the applicable welding code requirements. The dominate deficient condition noted was that the welds were significantly undersized.
Based upon this information the SRIC toured various areas of the facility with special emphasis on the ducting in the Unit 2 Containment Building since that was one of the more recent areas of installatien by the HVAC contractor.
In accordance with the design drawings, the bulk of the welds should have been fillet welds with hinch leg size. The SRIC noted by visual comparison to the hinch thick base metal that very few of the welds were of proper size.
The CAT inspectors also found cases where the bolting and gaskets between ducting sections were loose and/or missing.
The CAT inspectors also found that some support members were not within the dimensional ~ tolerances on the design drawings.
It was noted that the contractor's inspection records did not reveal these various facts, indicating ineffectual QC by the contractor.
- Further, e
a review of the licensee's audit program indicated that the licensee was unaware of these several problems in the fabrication, installation, and inspection of the HVAC systems.
Based upon the CAT inspectors' findings and his own observations, the SRIC recomended that a notice of violation be issued to the licensee pertaining collectively to these matters (Notice of Violation issued on May 31, 1983.
Reference 50-445/83-18 and 50-446/83-12, item 4).
7.
Installation of Major Items of Eouipment The CAT inspectors noted during their inspections of certain major items of equipment that there were several variables in how the equipment was fastened to the building equipment pads.
In some instances, tanks for example, CAT inspectors found that there were two nuts (double nuts) on the embedded bolts securing the equipment, other bolts had one nut, (single nut) and some had a combination of both single nuts and double nuts on one piece of equipment.
The CAT personnel also noted that certain heat exchangers had slotted holes in one of the mounting bases to allow for thermal expansion during operation. The holddown nuts appeared to be installed too tightly and may have prevented freedom of movement.
The SRIC obtained the design and installation drawings for two of the referenced heat exchangers identified in the CAT report.
Both were found to be horizontal Utube heat exchangers whose function is ncnsafety, but whose pressure boundary in the tubes is safety-related since the process fluid could be radioactive.
The SRIC found that the construction drawings for the mounting pedestals had a flat steel plate on one m
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pedestal that would be suitable for the type of mounting detail on these heat exchangers.
The SRIC then reviewed the installation travelers for each heat exchanger and found that these documents did not note or address the slotted details, the plate, or the fact the bolts should be left loose.
The SRIC would note that the i
vendor manual which provides the details does not provide information on how loose or tight the nuts should be nor how these nuts are to be locked at that looseness or some torque value.
The SRIC with j
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the assistance of site QC and craft labor had one of six nuts loosened on heat exchanger TCX-CSAHLD-01.
On all six of the studs i
involved, each had only one nut (single nut).
The one nut that was loosened had been very tight, as evidenced by the amount of 3
force required to break the nut loose.
On another heat exchanger of comparable design, it was found that each stud was double nuted j
and when the top nut was loosened, the second nut was approximately
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one flat (about 1/6 of a turn) from being fully tight.
This degree of looseness should allow sufficient freedom of movement.
During i
the' document review, the SRIC found that the engineer had specified that all rotating and vibrating equipment should be double nutted and that other equipment could be secured with only one nut.
No i
document could be located that established the identity of vibrating equipment nor were there any apparent provisions made to lock nuts where they must be deliberately left loose.
This was considered overall to be a violation of Criterion V of Appendix B to 10 CFR,50 (Notice of Violation was issued on May 31, 1983.
Reference:
Notice of Violaticn 50-445/83-18 and 50-446/83-12, item 1).
8.
Maintenance of Eauipment In Outdoor Storace Areas The CAT found that a considerable amount of equipment such as pipe support struts, elamps, and like items, normally stored outdoors, was not being properly maintained in accordance with procedure MCP-10,
" Storage and Storage Maintenance of Mechanical and Electrical Equipment", as evidenced by rusting bolts and adjustment screws on struts.
In addition, the strut bearings were' dirty from dust and i
the bearing load pins, in some instances, were rusted.
By a tour i
of the storage areas, the SRIC confirmed the CAT inspectors find-ings. The SRIC would also note that the INPO Self-Evaluation l
Report at page 111 describes essentially the sape finding.
This situation was detertnined to be a violation of Criterion XIII of l
Appendix B to 10 CFR 50 (Notice of Violation issued on May 31, 1983.
Reference:
Notice of Violation 50-445/83-18 and 50-446/83-12, i
item 2). The SRIC would note -for the record that there is little evidence that any items which indicated substantial deterioration from such storage conditions have in fact been installed in the i
nuclear power block.
It would appear that the various items involved have been cleaned and restored prior to installation such that they can perform the required function.
9.
Obsolete and/or Illeoible Drawines In The Field The CAT inspectors found a group of drawings in one particular area adjacent to.the control room that were found to be out.of date. by up to several issues and further, that some drawings in other areas were incomplete in the title and revision blocks.
The SRIC discussed
8 the finding with supervisory personnel of the licensee's central document control center who indicated that they had located the drawings identified by the CAT inspectors along with many more that were obsolete in other areas.
It was stated that distribution system for engineering drawings had become faulted by the simple volume and by the need for so many points of distribution and audit verification thereof.
Since problems are obviously still present, it was determined that the licensee had violated Criterion VI of Appendix B to 10 CFR 50 (Notice of Violation was issued on May 31, 1
1983.
Reference:
- Notice of Violation 50-445/83-18 and 50-446/83-12,
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item 3) and that substantial steps would be required to correct the problems.
10.
Allecations Relative To Imorocerly Suocorted Items In The Control Room The president of CASE in a letter dated March 11, 1983, addressed to Mr. Richard C. DeYoung, Director of the NRC Office of Inspection and Enforce-ment, indicated that CASE had received infomation from an unidentified sour; to the effect that:
a.
There is field run conduit above the control room supported only by wire.
b.
There is drywall (or sheet rock) that is supported by wire.
c.
There may be lights that are supported by wire.
The SRIC has examined the suspended ceiling and the area above the sus-pended ceiling in the control room area and has examined the pertinent engineering drawings depicting both in relation to these allegations with the following findings:
a.
There is a considerable amount of both safety-related and nonsafety related conduit in the area above the suspended ceiling.
The safety-related conduit is supported by Seismic Category I supports typical of those used in other areas of the facility.
The nonsafety-related conduits are generally supported by simpler and less substantial sup-ports that are typical of those that the SRIC has observed in large open factories and are not designed to seismic standards.
In each case examined, the non-seismic support was structurally paralleled with a small stainless steel cable that would assume the full weight of the conduit were the nomal support to fail in a seismic event.
b.
The drywall materials were found to be part of the suspended ceiling above the central part of the control room and to fom a part of the sloping wall area below the control room observation room.
These dry-wall materials have been securely fastened to a metal frame work (metal batten) which in turn is supported by conventional and non-seismic straps and wires to the concrete primary building.
The frame work is also attached to a system of stainless steel cables which in l
turn also attach to the primary structure such that if normal sup-ports fail during a seismic event, the weight of the framing and drywall will be assumed by the cabling thus preventing the materials frcm falling.
j 9
i c.
The lighting fixtures in the control room are suppcrted from an intermediate substructure of "unistrut" by light-weight conduit.
The substructure is likewise supported by the same type of conduit t
from the primary structure ceiling.
The conduit used appears to be the typical of that supporting the light fixtures in most offices with suspended ceilings.
Paralled with each conduit are two small stainless steel cables which would assume the load if the conduit or its-attachment were to fail.
In the case N
of the actual light fixtures, the cable is attached to the light 3
fixture at the edge of the reflector assembly.
The SRIC would note for the record that above described design
.l features appear to fully satisfy the intent of the licensee's commitment to -
comply with NRC Regulatory Guide 1.29. " Seismic Design Classification."
[
The licensee has used terminology in the classification system that is at variance with that of the regulatory guide but is explained and defined in Section 3.2 of the FSAR.
In essence, the licensee has defined all safety-related items that must remain fully functional during and after a seismic event as Seismic Category I.
Items not having a safety function but whose failure could damage components which have a safety function or cause injury to the occupants of the control room during an event are referred to as Seismic Category II.
In the case of the items involved in this allegation, all are Seismic Category II since their falling could cause injury to the control operators.
The cabling system described can be expected to prevent such a fall even though the normal supports could possibly fail.
The stainless steel cable used in this design feature, which at a short distance away looks much like bright galvanized common steel i
wire, is of relatively high strength.
As an example, the test strength of an 1/8-inch cable is in excess of 1760~ pounds. With four cables attached l
to a light fixture, two at each end, the total support capability of the cables is over 7000 pounds.
It is apparent that the designers have elected to use conventional suspended ceiling and light fixture support techniques in order to use. conventional and available materials and then provide a
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high strength backup support system in a seismic event.
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'No violations or deviations were identified during this special inspection effort.
11.
Placement anc Curinc of Concrete Durinc Freezina Weather J
During the limited public appearance portion-of. the Atomic Safety and Licensing Board (Board) hearing conducted on.May 15, 1983, there were two references to the placing of concrete.in freezing weather at the Comanche l
Peak Station which in turn lead to a question from the Board to the NRC staff as to whether there were any NRC personnel present with knowledge of the matter.
The two references are at 6106 and 6134 of the hearing i
transcript while the Board question is at 6109.
Also at 6109, an uni-I dentified voice responded to the Board that the matter had been reported in IE inspection reports.
Research of the NRC inspection reports revealed i
that there had been such 'a discussion in NRC Inspection Report 50-445/77-01 which was categorized as an' unresolved item pending the licensee's review and action on their finding of the problem.
The unresolved item was further discussed in NRC Inspection Report 50-445/77-04 with the closure of the item by an improvement in the QA procedures.
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i The SRIC has reviewed the matter, particularily with a view toward deter-mining whether the practices involved actually caused damage to the concrete involved. The primary focus of NRC Inspection Report 50-445/77-01 (Details II, paragragh 5) was directed toward two licensee " Site Surveillance Reports" which had been prepared approximately 2 weeks earlier than the inspection period covered by the inspection report.
The first of the licensee's reports (C-134-77} was directed specifically to findings by a licensee inspector that tfie-surface temperature of Concrete Placement 101-2808-001 some 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after the placement was completed were well below freezing in some locations.
The other licensee report (C-135-77) was directed toward records and was not considered in this review. The~5RIC obtained the necessary records i
to review the matter and found that placement 101-2808-001 had taken j
place on December. 30, 1976, being completed at approximately 6:00 p.m.
Later, the.same evening at approximately midnight, the licensee inspector found that some surface areas were chilled to as low as 200F.
The recorA reflect, however, that there was disagreement between the S&R inspectior personnel assigned to monitoring the curing of the placement and the licensee's inspector as to what the surface temperatures actually were.
The B&R personnel contended that the licensee inspector was actally mea-suring the air temperature rather than the temperature of the concrete.
f40 resolution of that disagreement was reflected in the records.
The SRIC interviewed the licensee ins'pector of record during the course of this review to gain a clearer understanding of the events which took place.
The licensee inspector stated during the interview that he was confident i
that his measurements were accurate and also stated that there was no phy-sical evidence that the concrete was frozen even though the surface temperatures were well below freezing.
The records also reflect that in 4
i order to resolve the issue, swiss hammer tests were run on the suspect areas after the concrete had fully cured.
These tests indicated that the suspect areas had attained strengths comparable to known properly cured areas, indicating that the concrete had not been damaged even though the M
possibility exists that it had been frozen for a period of time.
The i
records reflect that good concrete curing temperatures, i.e., above 400F Y
were established and maintained shortly after the licensee's inspector's J
observation.
For the record, the SRIC would note that Placement 101-2801-001 took place i
in the Unit 1 Reactor Building. The placement became the open area floor at the lowest full floor in the building.
This floor area, while suppor-ting some equipment, serves primarily as a walk area.
As such, it is fully topped with an architural concrete making the structural concrete no longer accessable.
tiRC Inspection Report 50-445/77-01 also discussed comparable events to that documented on Surveillance Report C-135-77.
One of these events was docu-mented by Surveillance Report C-063-76 on January 7,1976, and on B&R deficiency / disposition reports (now titled nonconformance reports).
These documents indicate that on January 7,1976, the surface temperature l
of Placement 105-2773-001, the foundation basemat for the Unit 1 Safecuards Building, were found frozen as evidenced by frozen wet burlap over certain areas that were not covered by insulating blankets.
The records also O
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Although the placement should not have been allowed to freeze in the time frame involved in accordance with the project speci-fication, the placement was accepted "use-as-is" on the premise that the
~
curing temperatures during the 7 days were conducive to a good cure and that after 7 days there would be little free water in the concrete to freeze even though the burlap was froze. This conclusion is considered valid by the SRIC based-on his review of publications of the American Concrete Institute and the Bureau of Reclamation.
Further, in responding to a separate finding that the field cure test cylinders made for the placement tested lower than allowed by the project specificatioris, swiss hammer tests were perfomed..
The swiss hammer tests indicated the concrete placement had full specified s trength.
Relative to the low reported strengths of the field cure cylin-ders, the SRIC would note that in his experience field cure cylinders will frequently test low under cold weather conditions.
The reason is that the cylinders' small mass generates little heat of hydration, thus making them either more vulnerable to freezing and/or curing much slower than normal due to their depressed temperat re.
The final events covered by NRC Inspection Report 50-445/77-01 included DDR-C-460 which in turn discussed low temperatures during the curing per-iod of threa separate placements that were made during the late December time period of 1976.
In each case, the records refle-t that the placements were accepted "use-as-is" since the least amount of cure time was 9 days, again with good conditions until the cold weather occurred.
The NRC inspector involved in NRC Inspection Report 50-445/77-04 which closed the unresolved issue has stated that he had visually inspected each of the placements discussed in NRC Inspection Report 50-445/77-01 for evidence of damaged concrete and found none.
NRC Inspection Report 50-445/77-04 did not reflect those inspections since the NRC inspector was aware that the concern was for prevention of repetition rather than any specific concern about the quality of the placements involved.
The SRIC would note for the record that there are no regulatory or industry prohibitions on placing concrete in cold weather conditions.
The American Concrete Institute and the Bureau of Reclamation both indicate that if the 0
fresh concrete is above 40 F at the time of place. lent, the chemical process of hydration will generate sufficient heat to prevent the concrete from freezing provided that precautions are taken to prevent heat loss.
In mass concrete applications, the greatest danger to the concrete is on the exposed surface areas, particularily at corners and other edges of the placement.
It would be exceedingly rare for the mass of the concrete to freeze and sustain damage.
These publications also indicate that even if frozen, the concrete will nomally cure to full design strengths if temperatures con-ducive to the hydration process are restored.
12.
Alleaations Relative To The As-Built Verification and Desian Verification Activities.
During April 1953, NRC personnel received allegations to the effect that
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the QA group performing as-built verifications were not measuring support member dimensions and therefore, the " Vendor Certified Drawings" of the supports would not be accurate. A second allegation from the same person indicated that the QA group charged with responsibility for verifying that E
design changes ~have been incorporated into the plant and that the inspection records for the installations accurately reflected that incorporation was
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being required with the use of a computer generated status document to l
make the verification of records.
The allegation was that the computer list-ing was faulty and therefore, the verification effort was equally faulted.
The -SRIC has examined each of these al' legations as to the factualness of 'the' allegation and as to whether the allegation has or will have an effect on the safety of the facility when operating.
In regard to the first allega-tion, the SRIC found that the allegation was and is factual.
The allegation, however, does not appear to have any significant impact on safety in that the as-built inspection was not developed to assure that the " Vendor Cer-i tified Drawing" was an accurate representation of the support in all aspects.
l The as-built program was established to assure only that the support loca-j tion on the supported pipe and the direction of support is accurate for the purposes of perfoming the final pipe stress analysis.
The responsibil-j ity for assuring that the support members and other characteristics of the j
individual support reflect the design drawing requirements reside ~in other j
QA groups associated with the fabrication and installation efforts.
To also perform these functions in the as-built verification inspection would be a redundant inspection that would not contribute significantly to the safety
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function of any given support.
l 1
Recarding the second allegation, the SRIC found that it too was factual but only at the specific time the allegation was made. When making the allega-tion, the alleger provided the NRC personnel with a reference to a QC f.
inspection report which he said would fully display his concern.
This 4
l report, identified as IR DCV-00421, was found to contain notation that the 4
j verification was based on a computer tabulation and that the report was M
j being completed at the direction of the inspector's supervisor.
The original i
report was dated April 4,1983. The pemanent file copy was found to have l
been marked " voided" by the originating inspector as of May 20, 1983, with i
a notation that the report had been superceded by IR DCV-00423.
This i
latter inspection report was examined by the SRIC and found to document l
essentially the same inspection effort by the same inspector but without j
any notation of having been based upon a computer tabulation and without j
notation of apparent protest of directions given by supervision.
The j
SRIC interviewed the QC inspector who prepared and signed all of the reports noted above in order to ascertain what had and is transpiring in s
j-the QC design verification program effort.
The inspector stated that the attempt to use the computer based data in the perfomance of the assigned I
task was in error from the beginning because of errors by persons genera-l ting the computer data. The interviewee stated that only the one verifica-tion effort had been done using the computer based data and that all prior and subsequent verifications have been done by the assigned inspectors l
directly and personally examining the existent quality records in compli-l ance with applicable QC procedures for the task.
He stated that the only i
13 l
procedural deviation was the one instance stated in the allegation. Dis-cussions between the group supervisor at the time the allegation was i
received and the SRIC indicated that he had attempted to use the computer tabulation to expedite the task on a trial basis by management direction and that he had caused the original inspection report to be filed as it was to give management a picture of the faults in the computerized data.
It i
thus appears that the design verification effort has been perfonned in q
accordance with procedures euept for the one-time pertubation that was
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subsequent correctly reaccomplished in accordance with approved proce-dures.
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No violation to NRC requirements were revealed during this special inspection effort.
13.
Imorocerly Certified Licuid penetrant Examination Materials 1
i The CASE informed the Atomic Safety and Licensing Board by a letter dated
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May 18, 1983, of a potential problem with the liquid penetrant materials in use at the Comanche Peak Station.
The letter stated that CASE had been made i
aware of the potential problem during a phone conversation with Charles A.
Atchison, who in turn learned of the " problem" from a Dallas area represen-j tative of the Magna-Flux Corporation, the orginal manufacturer of the material.,
The letter states that the problem surfaced only 7 to 10 days earlier.
Based on the date of the letter, it would seem that the problem arose between approximately May 8 to May 11, 1983.
i The situation bears close resemblance to the situation outlined beginning with NRC Inspection Report 50-445/82-18;50-446/82-09 based upon an inspection conducted during the period of September 7-10, 1982.- The NRC inspector noted that some certified test result documents had been altered by " pen and ink" i
4 changes not immediately explainable. The matter was considered unresolved i
at that time.
During a second inspection of the matter, conducted during i
November 1982 and documented in NRC Inspection Report 50-446/82-11, the inspector found that previous corrective actions were not adequate and fur-j 1
1 ther that the " pen and ink" changes sometimes didn't match the type of i
material being certified.
A Notice of Violation was issued as part of the i
inspection report on the matter. The licensee responded to the Notice of I
Violation by a letter dated December 21,1982, wherein he stated that a supplier had altered the certificates but that the original manufacturer had been able to furnish valid certificates and further, that all future i
purchases would be direct from the manufacturer rather from a " middle-man" i
supplier. The licensee also stated that specific receiving inspection pro-l cedures had been implemented to prevent repetition.
NRC Inspection Report i
1 50-445/83-10;50-446/83-05 documented verification that the licensee's actions
- i were acceptable and the matter was closed.
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It appears that the situation outlined in the CASE letter parallels the NRC findings in all details except for the dates which probably arcse l
as a result of misunderstood or incomplete communications between the
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i 14 Magna-Flux representative and Mr. Atchison and/or with CASE.
CASE also posed two questions on the matter as follows:
a.
Has an NCR been written on this problem?
Answer:
The above discussed inspection reports document a total of five NCR's that were issued.
b.
Has either TUGC0 or Texas Utilities or B&R notified the NRC of this probl em?
Answer:
The roles of reportability were effectively reversed in that the NRC identified the problem and notified the licensee.
A need for further NRC action on this matter has not been identified and tfie matter is considered closed.
14 Penetration Seals This special inspection was undertaken to ascertain the validity and sig-I '
nificance of allegations received initially by an NRC Headouarters Duty Officer on or about March 22, 1983, which were confirmed and added to during a telephone interview with the alleger on March 23, 1983, by the SRIC and a NRC inspector assigned to NRC Region I.
The allegations, as understood by the SRIC, were:
a.
The overlap seal for flexible boots should be 3 inches whereas 2 inches is being used by BISCO.
i b.
There maybe a problem with the strength of the fabric used in the flexible boots since the material supplier and BISCO are involved in a lawsuit.
c.
The aggregate used in a radiation seal may separate giving rise to improper personnel protection.
Since BISCO was and is on the Comanche Peak site installing seals, Region IV was selected for the purpose of this special inspection although the ccm-pany has involvement at several other nuclear power sites throughout the United States.
The SRIC obtained from the BISCO site manager all of 'the production and quality procedures applicable to the work at CPSES as well i
as some that are not.
The alleger specifically mentioned that the NRC should review Procedures QC-507, SP-504, SP-505, SP-505-1, and SP-505-2 in regard to the flexible boot overlap problem.
Each of the above procedures was in the books offered to the SRIC for review. A brief discussion fol-lows as to the contents of these procedures:
a.
QCP-507:
This procedure ccvers the final inspection of installed
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15 flexible boots.
The amount of overlap is not mentioned in the procedure, although the procedure does require that the seam be examined for evidence of poor sealing such as " fish-mouthing" which is taken to mean that the exposed edge of l
the overlap is puckered and not adhering to the base fabric.
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b.
SP-504:
This procedure provides instructions and a calculation sheet to initially cut the fabric into a shape that would subse-quently allow the fomation of a truncated cone.
The fomula on the calculation sheet requires that 1-inch be added at each edge of the fan shaped fabric which is evidently to pro-vide the overlap.
The base formula prior to adding the 1-inch provides a dimension just equal to the circumference of the pipe and/or sleeve to which the boot will be attached.
Thus, the 1-inch at each edge will provide for 2-inches of overlap, assuming that the pipe and sleeve are concentric.
If pipe and sleeve are not concentric, the resulting cone will be skewed and the seam overlap will be something other than 2-inches.
c.
SP-505:
This is a generic procedure for the installation.of flex-ible boots.
It was noted that the procedure requires that the adhesive for the overlap seam be spread over a 3-inch depth from the fabric edge prior to fitting up the fabric where it is to be installed.
Although not so stated, it appears that the 3-inch width of adhesive is to provide sufficient area of adhesive in the event the above men-tioned cone skewing occurs, d.
SP-505-1 and SP-505-2: These are additions to SP-505 having appli-cation when the boots are used as a simple pressure seal f
only and for when the boot is used as part of a fire pro-tection seal, respectively.
The SRIC interviewed the BISCO site manager as to whether the procedures had ever required a 3-inch overlap.
The site manager indicated that 3-inch seam had been used up to sometime in 1979 and that his homeoffice engin-eering had then changed the seal seam detail.
The SRIC reviewed the results of a pres ure differential test performed by BISCO in September 1979 which indicated that the fabric boot would withstand a differential pressure of 44 psig without sustaining damage.
The project specification (2323-MS-3SF) requires that the pressure seal maintain its integrity only up to 2 psig.
,j While the BISCO test data does not specifically state what the overlap seam width was on the test boot, it would strongly appear that the strength mar-gin is so high that even a reduction of 1/3 in the area of the overlap would have the effect of changing the safety factor from 22:1 to approximately 14: 1.
It is the SRIC's conclusion that while the allegation relative to the reduction in seam from 3 to 2 inches is correct, the reduction would have no significant effect on the performance of the boot in service at CPSES and that, therefore, the allegation has no technical merit.
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1 16 Regarding the matter of the possibility of some undefined problem with the boot fabric, the BISCO site manager stated thet his company has been engaged in a law suit with the supplier of the fabric but only in regard to the per-formance of the fabric in one application which is understood to involve the tearing of the fabric after being punctured.
It is understood that the puncturing has occurred when a gel type radiation seal hardens und.er radia-i i
tion. _Since the specific design. involved is not scheduled for use at CPSES, the allegation has no technical merit.
Regarding the matter of possible sep,aration of the radiation seal aggregate material from the carrier material, the SRIC can only conclude that t'e al-legation is potentially correct but without apparent merit.
The BISCO test reports indicate that the seals involved met the engineers specification.
The separation of the aggregate (powdered lead) from the carrier (a silicone material) would appear to be process sensitive in that if they are not well mixed, pockets of lead might forin with resulting pockets of silicone without sufficient lead.
Since the specification and the BISCO procedures require careful control and monitoring of the mixing process, the SRIC can only con-clude that these measures are effective in production operations as they were in preparation of the test samples.
- 15. Electrical Cable Solicina The SRIC became aware that the Comanche Peak project electrical engineer had authorized the splicing of safety-related and auxiliary electrical cables within several control panels during the inspection period.
Since the licensee has committed in FSAR Section 8.1 to ccmply with IEEE 420,
" Trial-Use Guide for' Class lE Control Switchboards for Nuclear Power Gener-ating Stations," which forbids splicing of wiring in such panels, the SRIC judged that the licensee was deviating from these commitments. The licen-see engineer indicated that he interpreted the IEEE standard to prohibit such splicing only between the cabinet terminal boards and the cabinet devices and did not prohibit such splicing in the field run cables attach-ing to the terminal boards.
The engineer stated that action had been initiated with the NRC Office of Nuclear Reactor Regulation to clarify the issue in the FSAR.
The SRIC confirmed that such action had been initiated by a telephone conversation with the NRR Licensing Program Manager for Comanche Peak.
Pending action by NRR, this matter will be considehd as an i
unresolved matter.
- 16. Unresolved Items Unresolved items are matters about which more infonnation is required in order to ascertain whether they are acceptable items, items of non-i compliance, or deviations.
i One such item, disclosed during the inspection, is discussed in paragraph 15 i
above. This item is identified as " Splicing of Electrical Cables in Cabinets."
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- 17. t'anacement Interviews The SRIC met with one or more of the persons identified in paragraph 1 of this report at frequent intervals during the inspection period to discuss the licensee's position and proposed, actions on a significant 1'
number of ' issues which occurred during the period.
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-g gg,D 2 01983 In Reply Refer To:
Occkets:
50-445/83-27 Texas Utilities Generating Company ATTN:
R. J. Gary, Executive Vice President & General Manager 2001 Bryan Tower Dallas, Texas 75201 Gentlemen:
This refers to the inspection conducted by Mr. R. C. Stewart of this office during the periods May 10-July 1, and September 9-22, 1983, of activities authorized by NRC Construction Permit CPPR-126 for the Comanche Peak, Unit 1.
Areas examined during the inspection included inspection of alleged improper construction practices expressed by Robert L. Messerly and an individual wno requested confidentiality.
Within these areas, the inspection consisted of selective examination of procedures and representative records, interviews with personnel, and observations by the inspector.
These findings are documented in the enclosed inspection report.
Within the scope of the inspection, no violations or deviations were identified.
In accordance with 10 CFR 2.790(a), a copy of this letter and the enclosure will be placed in the NRC Public Document Room unless you notify this office, by telephone, within 10 days of the date of this letter, and submit written application to withhold information contained therein within 30 days of the date of this letter.
Such application must be consistent with the requirements of 2.790(b)(1).
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i APPENDIX U. S. NUCLEAR REGULATORY COMMISSION REGION IV NRC Inspection Report:
50-445/83-27 Docket:
50-445 Construction Permit: CPPR-126 Licensee:
Texas Utilities Generating Company (TUGCO) 2001 Bryan Tower Dallas, Texas 75201 Facility Name:
Comanche Peak, Unit 1 Inspection At:
Comanche Peak, Unit 1 Glen Rose, Texas b
Inspection Conducted:
y 10-Julv 1, and September 9-22, 1983
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Inspector:
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M E"8& 85 T. C. Stewarr;--RItMor Inspettor Date Reactor Project Section A Approved:
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D. M. Hunnicutt, Chief Date Reactor Project Section A 4
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2 Inspection Summary Inspection Conducted May 10-July 1, and September 9-22, 1983 (Report 50-445/83-27)
Areas Inspected:
Special, unannounced inspection of alleged improper construc-tion practices expressed by Robert L. Messerly in an affidavit dated February 3, 1983, prepared for Citizens Association for Sound Energy (CASE) and in an interview conducted on April 14, 1983, by members of the NRC Office of Investigations Field Office, Region IV.
The inspection involved 120 inspector-hours onsite by one NRC inspector.
Additional information was received from an individual, who requested confidentiality, that a former B&R millwright had drilled holes through rebar without the required engineering approvals.
This supplemental inspection involved 10 inspector-hours onsite by one NRC inspector.
Results:
Of the seven allegations regarding improper construction practices expressed by Mr. Messerly, five were found to be unsubstantiated.
One allegation regarding improper documentation was found to be substantiated, however, the error was properly corrected by the licensee and appears to lack technical merit; and one allegation regarding the posting of NRC Form 3, could neither be refuted nor substantiated, however, it too appears to lack technical merit.
No violations or deviations were identified.
Results of Sucolemental Inspection l
The allegation that unauthorized cutting of rebar during installation of I
" trolley tracks" in the fuel handling building is considered to be l
unsubstantiated.
No violations or deviations were identified.
l
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3 Details A.
Persons Contacted Texas Utilities Services Incorporated (TUSI) Employees B. G. Scott, Quality Engineering Supervisor G. Tanley, General Superintendent C. R. Hooton, Lead Civil Engineer R. M. Kissinger, Project Civil Engineer C. Fleming, Field Engineer Brown & Root (B&R) Emoloyees W. Wright, Project Welding Engineer B. Hauser, Field Engineering Superintendent C. Osborn, Tool Crib Foreman The NRC inspector also contacted other licensee and contractor employees curing the course of the inspection.
Note:
Prior to this inspection, separate and independent investigative interviews were conducted by members of the Office of Investigation Field Office, Region IV (see attached Report A4-83-005, dated May 20, 1983).
B.
A11eced Imorocer Construction Practices The NRC inspector, through an interpretative review of Mr. R. L. Messerly's affidavit, dated February 3,1983, and his statements during his interview, April 14,1983, determined that there were seven specifically alleged matters that required a detailed inspection effort to assess their technical merit and/or their potential impact on safety-related systems, component, and structures.
The seven areas of NRC concern which Mr. Messerly alleged to have occurred are summarized as follows:
1.
That B&R employees drilled undocumented and unauthori:ed holes that cut through reinforcing steel and that such drilling and cutting was done at the direction of supervisors.
Mr. Messerly provided a copy of a personal diary which, he alleged, reflected undocumented and unauthorized drilling.
2.
That one of the main steam lines in Unit 1 was moved using the polar crane, thereby placing the section of pipe line in an unsafe stressed condition.
3.
That he had cut through concrete reinforcing steel as directed by work instructions that were not in accordance with t.a.e acproved a
method of documentation.
4 4.
That tubular hanger / support steel anchor bolt holes were enlarged with a burning torch which he said was unauthorized.
5.
That (Richmond) anchor bolts were not perpendicular to concrete surface and, therefore, unacceptable.
6.
That stainless steel pipe attachments were welded on piping without an inerting purge.
7.
That NRC Form 3, " Notice to Employees" was not posted on three main bulletin boards.
C.
Inspection Findings Allegation 1 1.
Discussion Mr. Messerly stated that during his assignment as foreman over the first crew responsible for drilling through concrete and reinforcing steel (rebar) during installation of cable tray and pipe hanger supports, ne was ordered by his supervisors to loan out drill bits and/or drill undocumented and unauthorized holes through rebar.
To further support his allegation, Mr. Messerly named B&R employees responsible for the alleged improprieties and those who could substantiate his allegations. 1/
In addition, Mr. Messerly provided the NRC staff a copy of his personal daily diary in which he logged drilling of holes for electric cable trays / hanger supports and rebar cutting details. He stated that this diary also identified noles he drilled, in or through, rebar and concrete without having documentation and author-ization.
2.
Chronological Findings 1978-1982 In order to determine the magnitude of implication ano the resulting findings of Mr. Messerly's allegations.
1/
See attacned " Assistance to Inspection Report," Report A4-83-005, dated May 20, 1983
5 The NRC inspector reconstructed, through the use of record archives and interviews with site personnel, the onsite construction activities and QA/QC program being implemented in the specific area of concern during the period 1978-1979.
3.
Rebar Cutting Capabilities The NRC inspector found from B&R purchases that during 1975 through 1982, the type of onsite equipment (drills) capable of cutting through rebar and available to craft personnel were restricted almost exclusively to the (water cooled) type diamond core drill bits (rebar eater) and associated drill motors, purchased from Drillco Equipment Company, Inc., (Drillco) Miami, Florida.
The Drillco water cooled diamond core drill bits purchased are hollow, tubular in shape, varying in sizes from 1/2" to 16" in diameter and from 2" to 14" in length.
The drilling end has a series of carbide rectangular shaped teeth impregnated with industrial diamond dust.
When worn, or dull, the bits can be reconditioned and reused.
The NRC inspector found that the initial core drilling requirements (1975 to 1978) were under the control of the concrete department.
Drilling was restricted to investigative type core drilling (identif-ing concrete honeycomo, voids or cold joints) in the base mats (NRC Inspection Report 445/446/76-04 dated April 20, 1976).
In late 1977, record archives contain copies of the original " Core Drilling Procedure," HCP-13, dated September 27, 1977, and issued for implementation April 21, 1978.
The procedure was developed for core drilling through walls and slabs for the purpose of installing pipe sleeves, conduits, instrumentation sleeves, etc.
Penetrations which were shown on drawings or included in design documents prior to concrete placement and inadvertently omitted, or penetrations which were added by the architect engineer (A/E) but for which the installa-tion information was not available to the field prior to concrete placement were covered by this procedure.
The procedure was applicable for all core drilling required in the plant.
Core drilling was assigned to the millwright department.
The procedure and its controlling document, " Core Drill Request Form," requires delineation of exact location, size and rebar location, and contains review and approval signoffs.
This procedure continues to be the principal core drilling procedure (Revision 3, dated December 2, 1981).
However, current policy (as determined by ths cognizant project civil engineer and reflected in documented records) is the assignment of core drilling of 2-1/2" ofameter and larger to the millwright department and 1/2" to 2" diameter core drilling to the steel faerication department drilling crew.
The NRC inspector also noted that " Core Drilling Request Forms" do not imply recar cutting; in fact, rebar cutting has for the most part, een avoidea where possible as stated by the project civil enginee" during ciscus-sions with engineering personnel. This fact was ooserved by the NRC L
6 inspector during his review of randomly selected " Core Drilling Request Forms" (1978 through 1982).
Construction records indicate that electrical cable tray, conduit hangers, and pipe hanger support installations were initially started in late 1978.
This coincides with the formation of the steel fabrica-tion department pipe hanger crew (s), special drilling crew (headed up by Mr. Messerly), and the requisition of the water cooled diamond core drills and motors by the steel fabrication department (of which Mr. Messerly was a member) on September 6,1978.
A record search indicated a Design Change / Design Deviation Authorization 2470, dated September 5,1978, authorizing rebar cutting for Cable Tray Support No. 597.
This was an initial rebar cut made on September 9,1978, and identified by Mr. Messerly in his personal handwritten diary (see paragraph 6).
The primary anchor and fasteners utilized at CPSES for the attachment of cable tray supports, conduit supports, pipe hanger supports, etc.,
to concrete surfaces are the "Hilti" drilled-in concrete expansion anchor and " Richmond" screw anchor.
The Richmond screw anchor is positioned prior to concrete placement, whereas the Hilti requires concrete drilling and placement at the time of component installation (a licensee representative stated, that based on purchase orders, over one million Milti bolts 1/2" to 1-1/4" in diameter, have been installed to date).
Drilled-in expansion bolts are bolts having expansion wedges so arranged that, when placed in a drilled hole and the nut tightened, the wedges are expanded and the bolt is securely anchored.
The most predominant means of drilling holes into concrete for expansion bolts is the use of Hilti power drills, using Hilti carbide masonry bits of the same nominal size as the bolt.
This form of drilling does not have the capability to drill through rebar.
In limited access areas where the Hilti power drills cannot be used, a flexible Drillco drive drill with drill press / vacuum base anc Orillco water cooled carbide / diamond bits are used.
This form of drilling has the capability of drilling through rebar and was restricted to the steel fabrication department special drilling crew (headed by Mr. Messerly from September 1978 through October 1979).
For these two methods of drilling, no authori:ation is requirec for Hilti bolt installations (other than an approved hanger support installation " traveler" with its accompanying location drawings). A design change authorization is only required if relocation is beyond the drawing tolerance limits, or if rebar is encountered anc requires cutting.
Construction quality programs of this nature rely heavily on each incividuals personnal integrity to adhere to prescribed procedure requirements.
i 7
i A research of purchase orders for 1978 through 1979 conducted by the i
NRC inspector, indicated that only seven (Drillco) power drives that facilitate water coolirg capability were purchased during that time frame.
Two were issued to the millwright department and five were issued to the steel fabrication department (under the control of i
Mr. Messerly).
Mr. Messerly requisitioned (from the B&R warehouse)
I three drill machines, with water cooling splash guards, and one flex l
shaft unit on September 6, 1978.
An additional flex shaft unit was requisitioned by Mr. Messerly on October 6,1978.
j In discussing the method of drilling with the Orfilco water coo' led diamond bits with cognizant site personnel, the NRC inspector was informed that when drilling with the diamond core bits, water cooling i
is mandatory.
The water provides two primary functions: it removes i
drilling debris (concrete / steel) as drilling progresses, otherwise the drill bit would bind; secondly and most important, without water cooling, the drill bit will readily " burn up," particularly when 4
attempting to cut through rebar steel.
In addition, a drilling foreman stated that, drilling equipment is heavy and bulky and drilling set-up time (mounting to walls or ceiling) generally takes i
half an hour to one hour.
When drilling, tha water cooling creates a i
concrete / water mist deluge requiring crew members (normally two) to wear rain type outer protective clothing.
i 4.
Diamond Core Drill Bit Control In verifying the purchase and control of the diamond core drill bits, 3
]
the NRC inspector reviewed 21 B&R purchase orders awarded to Orillco dating from January 13, 1978 through February 13, 1980.
i The NRC inspector found that of the total 21 purchase orders,10 requisitions were initiated by the steel fabrication department genersi superintendent, representing 293 core drill bit purchases, and 11 purchase orders were intiated by mi11 wright supervisory 3
personnel representing 122 core drill bit ourchases.
1 4
In reviewing the accompanying warehouse requisitions contained in
]
each of the purchase order files, the NRC inspector noted that in the i
t case of the steel fabrication department orders, all requisitions 1
i bore the signatures of Mr. Messerly or his oepartment personnel, Correspondingly all equipment ordered by the mi11 wrights was issued to and signed for by a cognizant mi11 wright forcinan.
The NRC inspector conducted an inspection at each of the respective department tool crib areas (millwrights and steel fabrication).
The j
mi11 wrights maintain a tool crib area enclosed by heavy gauge wire screen and a locked counter door access.
The ' tool crib attendant j
maintained a clip board type log specifically for the control of i
Drillco diamond core bits.
The log identified the individual, along i
with eneckout and return dates.
Entries in this log date back to l
October is, 1978.
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1
8 4
The steel fabrication department maintains a small separate building where the hanger installation crew controls the drilling equipment i
and bits.
The NRC inspector observed that the Drillco diamond core bits were separately stored in a large wooden cabinet with an accompany-ing combination lock.
The method of control over drills and bits was discussed with the cognizant foreman.
The foreman stated that he had been in charge of diamond core bits and the fabrication department drilling crew since April of 1982.
He stated that he did not cut any rebar without an approved " request for rebar cutting" form, which he further demonstrated by utilizing an inprocess form dated June 14, i
1983, No. 135.
The NRC inspector determined that this was in accordance
~
with the prescribed procedure, CC-P-47, " Request for Rebar Cutting,"
dated June 17, 1981.
In interviewing former supervisors, foremen, and members of diamond core drilling crews 1/, all interviewees stated that the present method of controlling diamond bits has been in effect since the initial purchase of Orillco bits; i.e., only cognizant supervisors, foremen, or drill crew members have access to the diamond bits (those interviewed included five former members of Mr. Messerly's drill crew).
s 5.
Procedure Reviews and Procedure Imolementation During the inspection, the NRC inspector reviewed B&R procedures and procedural implementation applicable to concrete core drilling and drilling requirements for Hilti bolt installations.
Included in the review were the original versions of issued procedures from archive files that were applicable during 1978 and 1979.
Applicable procedures reviewed included the following:
B&R Procedure 35-1195-CEI-20, " Installation of 'Hilti' Drilled-In Bolts," dated May 31, 1978; B&R Procedure 35-1195-CEI-20, " Installation of 'Hilti' Orilled-in Bolts," Revision 8, dated January 26, 1983; TUSI Procedure QI-QP-11.3-2, " Cable Tray and Conduit Hanger Inspection," dated June 3, 1978; B&R Procedure 35-1195-MCP-13, " Core Orilling," dated September 27, 1977; B&R Procedure 35-1195-MCP-13. " Core Orilling," Revision 1, dated April 21, 1978; TUSI Procedure CP-QP-11.2, " Surveillance and Inspection of Concreta Anchor Bolt Installation," dated Decameer 13, 1979; y
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B&R Procedure 35-1195-CCP-47, " Request for Rebar Cutting," dated
. June 17, 1981; TUSI Procedure QI-QP-11.2-1, " Concrete Anchor Bolt Installation,"
9 dated December 13, 1979; and Ll G&H Specification 2323-55-30, " Structural Embedments."
H is The principal construction procedure applicable for Hilti bolt installation was B&R Procedure 35-1195-CEI-20, originally issued May 31, 1978.
Section 3.2.1 states, " Expansion bolt holes shall not be drilled into concrete reinforcing steel unless approved by the Gibbs & Hill, resident engineer or his representative." This require-ment has been retained in all subsequent (eight) revisions to the
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procedure.
The statement is currently found in Section 3.1.2.1 of j
Revision 8 dated January 26, 1983.
i-l In discussing the method of " engineering approval" established in the l-period 1978-1979 with the cognizant project civil engineer, the NRC l'
inspector was informed that an " Interference Task Force" was established i
in September of 1978, composed of three TUSI project civil engineers v
l who coordinated any design changes or rebar cutting with the cognizant 4.
onsite, A/E Civil Design Engineer.
Where interference between the expansion bolt and reinforement was encountered, the bolt location j
was generally adjusted within the tolerances allowed by the design
{
drawings, otherwise a design change / design deviation authorization (DC/00A), design change authorization -(DCA), or a component modif-l ication change (CMC) was initiated and issued.
The various forms of j
design change documents have subsequently been reduced to the DCA and CMC forms of design change approval.
Where interference with reinforc-ing steel cannot be avoided and the cutting of rebar is required, the approval authorization is initiated by the A/E site project civil engineer who evaluates all requests for cutting rebar.
The criteria i.
for such evaluation is based on design parameters determined by the I
A/E. Final design approval for any rebar cutting remains the responsi-bility of the A/E's New York office.
The A/E site project civil engineer maintains a CMC OCA issuing log, for rebar cutting.
The earliest entry noted by the NRC Inspector is CMC 0188, dated October 3, 1978.
The information on the DCA or CMC; i
f.e., number of rebar cut, size and location is transferred to a j
separate set of building structural drawings especially establitned for showing "as-built" renar cutting entitled "rebar drawings cutting criteria."
In interviews with the cognizant A/E site project civil i
engineer assigned during 1978-1979 1/, the NRC investigators were
~
informed that although requests to cut rebar. came f rom a number of
?
different 81R craft personnel, he, almost always, gave the approving l
CMC to Mr. Messerly, since his crew did the rebar cutting.
He further stated that he had no knowledge of rebar cutting without l
i
,m 10 engineering approval.
The NRC inspector subsequently conducted a detailed review and documentation verification of the above proce-dures.
6.
Messerly's Diary (Loa)
During the interview on April 14, 1983, Mr. Messerly provided the NRC investigators with a copy of his personal diary log entitled, " Start of New Crew and New Operation Rebar Cutting Detail." The diary consists of 24 handwritten pages of columniation entries on standard 8-1/2" x 11" paper dating from September 7, 1978, through October 17, 1979.
Five columns delineating print numbers (cable tray / hanger support numbers); building location; rebar cut; day and date; and position (floor, wall, flex, DC/00A, OCA, or CMC number) were recorded by Mr. Messerly.
In addition, various notes regarding work activities are. interspersed thoughout the 24 pages.
During a detailed review of the diary, the NRC inspector observed that (barring errors due to legibility) Mr. Messerly recorded drilling a total of 2976 holes associated with 415 hanger / supports.
Of the 2976 holes drilled, 280 rebars were cut.
This means tnat rebar requiring cutting was encountered in less than 10% of the holes drilled.
All rebar cuts, as noted by Mr. Messerly, were identified by either a DC/00A, DCA, or a CMC.
A total of 84 such authorizations were identified.
Twenty-one of these rebar cuts were related to nonsafety-related buildings; therefore, the NRC inspector did not review these particular authorizations.
In addition, of the 2976 holes drilled, 247 were identified by Mr. Messerly as being in the turbine building.
Of the remaining 63 documents authorizing rebar cutting, the NRC inspector made a random selection of 32 authcrizations for a comparative verification against Mr. Messerly's diary.
The NRC inspector verified 132 rebar cuts identified in the 32 authoriza-tions.
In all cases, the location, size, ano number of rebar were identified on the DCA or CMC.
In acdition, all 132 cut rebars were traced to, and identified on, the specific building structural drawings, "rebar drawings cutting criteria," with the corresponding authorizing document number.
There was no rebar cutting, as identified by Mr. Messerly in his diary, that does not have a corresponding authorization number.
It was also observed by the NRC inspector, that a handwritten note in the diary (assumed to be written by Mr. Messerly) states
" Ordered to drill by (name withheld) - floor S.W. I."
Adjacent to tne date July 23, 1979, and Hanger / Support Number SW-2-035-004-JC3R.
Under tne rebar cutting column Mr. Messerly noted, "None ?".
Mr. Messerly also noted that eight holes were crillec.
During
11 an investigation of this particular support (SW-2-035-004-J03R) in the service water intake struct6re (S.W.I.), the NRC inspector found that the support was deleted on July 30, 1980.
The original bolt holes were subsequently grouted and concrete
~
surfaces painted.
It is assumed that, by indicating a question mark after his notation, Mr. Messerly was not a witness to the actual drilling of the specific holes drilled by his crew members, and since seven persons formerly associated with drilling operations have stated 1/ that they have no knowledge of unauthorized rebar cutting.
The NRC inspector did not pursue this matter further.
It was also observed by the NRC inspector that, during a verification review of the 32 OCA's and CMC's identified by the Mr. Messerly's diary, CMC 3307 identified 48 rebar cuts in the service water tunnei alone. This was also mentioned by Mr. Pesserly during his interview.
All 48 rebar cuts were traced to the design change authorization documents.
Although Mr. Messerly's diary consistently identified the percentage of rebar cut, the established G&H design criteria considers any reduction in individual bars a 100% loss of the bar.
The NRC inspector found no unauthorized rebar cutting identified by Mr. Messerly in his handwritten diary.
7.
Conculsion - Allecation 1 Mr. Messerly's alle;ation that B&R employees drilled undocumented and unauthorized holes that cut through reinforcing steel could not be
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substantiated for the following reasons:
i
,i Mr. Messerly's statements lack sufficient specificity as to who a.
he " loaned" the water cooled diamond drill bits to cut rebar, or who specifically ordered him to cut rebar when aad where.
Former supervisors deny ordering Mr. Messerly to " loan" out drills or cut unauthorized rebar, nor did any of the five former crew members support.this contention, b.
In the event an unauthorized person did use a water cooled diamond bit, i t is highly unlikely that cutting of rebar would be accomplished without the accompanying water cooling drive equipment, or if a orill bit was " loaned" for drilling concrete only, it is conceivable that drilling would be successful witnout ' water cooling, but not necessarilly resulting in defective workmanship.
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Although Mr. Messerly implied that his personal diary centained identification of unauthorized and undocument rebar cutting, unless shrouded by omission or misinformation, the NRC inspector could not identify a rebar cut that was not authorized by
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DC/DDA, DCA, or CMC.
I d.
Although the method of diamond bit accountability / control exhibits a weakness, the need for relying on individual personal integrity would not be diminished.
The inspection findings did not, nor do not, suggest indiscriminate cutting of rebar was done.
Documented records exhibit a purposeful avoidance of rebar interference.
Furthermore, the Messerly diary demon-strates that less than 10% of the recorded total holes drilled by his crew encounted rebar that required cutting.
There were no violations or deviations identified in this area of the inspection.
Allegation 2 1.
Discussion Mr. Messerly stated in his affidavit of February 3,1983, and in his interview on April 14, 1983, that he had witnessed the use of the Unit I reactor containment building polar crane by a pipefitter supervisor in relocating a main steam line in a manner that put undue tension on the pipe.
In addition, Mr. Messerly provided the names of persons involved with the movement of the steam line 1/.
2.
Conclusion - Allegation 2 Although B&R personnel named by Mr. Messerly contradicted his allegation 1/, the NRC inspector conducted an independent review of the onsite documented records regarding this matter.
It was observed by the NRC inspector that the specific 32-inch steam line mentioned by Mr. Messerly is, Loop 1, Line number MS-1-RB-001-1302-2, and the reactor building polar crane was utilized in a vertical lift to assist repositioning a section of this permanent piping.
The licensee has maintained a documented engineering record of the specific line movement.
The NRC inspector noted that the movement of the line was necessary in order that a large section of temcorary I
piping (attached to the steam generator feedwater nozzle and previously used for water flushing) be removed and to relocate the permanent section of the main steam line that had " sagged" due to the weight of the temporarly installed flushing pipe.
The record folder contains meeting notes (memorancum) which reflect discussions with Westinghouse (NSS Supplier) and the cognizant A/E representatives prior to the work activity, in adaition to establishing engineering limitations and acceptability.
The line was moved on January 15, 1982 under the
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13 supervision of the field mechanical engineering group, and was witnessed by an engineering representative who observed the installation and use of the dynamometer (to register crane lifting loads) throughout the operation.
The lift connections and applied forces were recorded
'e and retained in the file.
The lifting points were consistent with I
the hanger locations to simulate the permanent support system.
The as-built configuration was analyzed for stress and the acceptability r.
of the line confirmed.
In addition, the recent completion of the
" Reactor Hot Functional Test" did not reveal any undue stress conditions.
1 This allegation cannot be substantiated.
No violations or deviations were identified in this area of the inspection.
Allegation 3 1.
Discussion During Mr. Messerly's interview on April 14, 1983, Mr. Messerly (in referencing his personal diary) stated that he initially started drilling rebar based on the' instructions of three part memos, DC/DDAS, and subsequently the CMC.
Although Mr. Messerly did not allege that the CMC was an improper document, he did imply that the DC/DDA and the three-part memo were not the right documentation.
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2.
Conclusion - Allegation 3 l'i During the NRC inspector's review of Mr. Messerly's personal diary (paragraph 6), it was observed by the inspector that the first four holes (rebar cuts) he drilled on September 7 and 8,1978, for cable tray hangers 596, 642, and 643, Mr. Messerly made the notation E
"RFIC".
In researching the archive files, the NRC inspector found
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the original Request for Information or Clarification (RFIC) documents, N
Request Nos. EH-14 and EH-15, dated August 29, 1978.
Although the f;
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instructions authorizing rebar cutting contained in the RFIC were correct and authorized by the cognizant A/E design engineer, the RFIC
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document was-not the " approved" method of authorizing a design
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change.
The NRC inspector noted that this documentation error was corrected by CMC No. 00766 issued on Ottober 16, 1978.
The original document, the RFIC contained a note to this effect.
On September 9, 1978, Mr. Messerly's diary contains a reference to DC/DDA No. 2489 for two rebar cuts for hanger No. 597.
In researching this particular 1
DC/DDA, the inspector found that DC/DDA No. 2489 was not related to hanger No. 597.
The NRC inspector found that DC/DDA No. 2470 correctly a
identified the rebar cutting authorization.
The location and number of rebar cut was also traced to CMC No. 01146, cated September 20, 1978, and to the as-built building structrual drawings, "Rebar Drawings Cutting Criteria." This allegation by Mr. Messerly was -
substantiated; however, the original documentation error was identified a short time after its occurrence and immediately corrected and did not impact on plant safety.
No violations or deviations were icentified in this area of the inspection.
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14 Allegation 4 1.
Discussion During Mr. Messerly's interview on April 14, 1983, and as stated in his February 3, 1983 affidavit, Mr. Messerly indicated that anchor bolt holes in tubular steel hanger supports were enlarged with a burning torch in order to compensate for the angularity of the previously installed (Richmond) anchor bolts, rather than redrill the holes.
2.
Conclusion - Allegation 4 The results of the interviews of eight B&R employees, whose names were provided by Mr. Messerly and alleged to have knowledge concern-ing the improper use of cutting torches on hanger material, is contained in the attached " Assistance to Inspection Report." 1/ Two
' ndividuals stated that they recall an instance during a redesign i
modification of a hanger where it was discovered that holes had been enlarged by a burning torch, therefore, that portion of the hanger was scrapped.
During the onsite followup inspection concerning this matter, the NRC inspector discussed the use of cutting torches with the licensee's welding engineers and fabrication department engineers.
The NRC inspector was informed that the use of cutting torches is not prohibited, provided it is done in accordance with prescribed B&R procedures and/or ASME,Section III, Subsection 4211 (thermal cutting).
In the case of tubular hanger installations, the preferred method of correction for hole misalignment is to drill offset hole (s).
This has been done on many occasions via the design change CMC document.
The cognizant project engineer, responsible for approving and issuing CMC's for '
hanger modifications, stated that he knew of no CMC that involved authorization of hole enlargement or hole relocation on tubular hanger supports utilizing thermal cutting; however, thermal cutting has been permitted as necessary on other types of carbon steel supports, base plates, etc.
The NRC inspector conducted a walk-through of the containment building to examine accessible installed tubular hangers, specifically in the plant areas mentioned by Mr. Messerly during his interview.
The inspector-examined approximately-60 hangers at the 905' and 860' elevations in the containment building.
Although limited in visual accessibility to each 1" or 1-1/4" drilled hole in each section of the tubular hangers, the NRC inspector cio not find any hole that was enlarged by a cutting torch.
In addition, the NRC inspector discussed the subject of thermal cutting with the cognizant QC supervising inspector who was involved with inspections of tubular hanger installation curing 1980-1982.
Tne QC supervisor stated, that neither he nor any inspector discoverec
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Based on the lack of specificity by Mr. Messerly, the lack of corroborative testimony by Messerly's witnesses, interviews by the NRC inspector with cognizant site personnel, and the (limited) examinations of installed hangers, this allegation could not be substantiated.
There were no violations or deviations identified in this area of the inspection.
Allegation S 1.
Discussion During the interview on April 14, 1983, Mr. Messerly stated that Richmond Insert anchor bolts installed between elevations 905' and 860' in the reactor containment building have not been installed perpendicular to the concrete surfaces and, therefore, are unaccept-able.
In addition, Mr. Messerly stated, "... whatever angle it is, we would drill it at that angle so that it would ccme through the tube (i.e., tubular steel) and when it comes out the other side of the tube, it comes out as close to center as we could get it."
Mr. Messerly also stated, "Just go out there and pull any.
studdeo rod out of there, pull three of them and two of them is [ sic]
crooked."
2.
Conclusion - Allegation 5 During the NRC inspector's onsite follow up of this matter, the i
inspector found that the B&R Procedure CP-CPM 9.10, " Fabrication of 2
ASME-Related Component Supports," (original issue 12/28/78) is the primary construction installation procedure to be implemented and followed by the hanger installation crews.
The " General Fabrication and Installation Requirements," Section 3.3.1.2 " Installation Tolerances," states in part,
" Field Fit Tolerances "The tolerances discussed above shall be maintained for support fabrication activities.
However, if during the installation, the support won't fit, the members may be " field fit" provided the piping and elevation tolerances shown below have been maintained.
All other tolerances regarding axial location, alignment, and base plate attachments must be adhered to unless otherwise noted on the drawing."
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Where the surface of a high strength bolted part has a slope or more than 1:20, a beveled washer shall be used to compensate for the lack of parallelism."
During discussions with the cognizant design engineers concerning the specific installation requirements relative to the limiting perpen-dicular angle of the anchor bolts (Richmond Inserts), the NRC inspector was informed that the limiting perpendicular angle of anchor bolts (Richmond Inserts) to the concrete surface is, aside from the requirements of Section 3.3.2, is handled on a case-by-case basis.
No enlargement of the existing predrilled holes in the tubular steel is permitted without prior approval; however, numerous CMC's have been issued wherein offset holes have been authorized.
The approval is generally accompanied by the requirement that the l
large square bolt washer be welded in place using a 1/4" fillet on 2 i
sides. The cognizant engineer further stated that the requirement above only applies to safety-related supports (ASME III, Subsection NF, Classes 1, 2, and 3 component supports).
Enlargement of the predrilled holes in the tubular steel for nonsafety supports is permitted without prior engineering approval.
Since Mr. Messerly specifically referred to the 860' and 905' elevations in the reactor containment building in his testimony, it was assumed by the NRC inspector that his specific concern was in reference to the permitted angularity of the safety-related Richmond
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Insert anchor bolts.
Mr. Messerly was apparently of the opinion that the anchor bolt should be precisely perpendicular to the concrete surface, which appears to be a misunderstanding on his part of the installation specification.
Furthermore, Mr. Messerly's testimony reflected his awareness and knowledge of the procedural requirements, therefore, it must be assumed that Mr. Messerly did not ignore procedural requirements and did not inciscriminately enlarge pre-drilled tubular steel holes in safety-related supports.
- Further, that any offset or enlargement done by Mr. Messerly had prior engineer-ing approval as required.
As noted in Allegation 4, paragrapn 2, the NRC inspector conducted a limited visual examination of approximately 60 hanger supports at the 905' and 860' elevations in the containment building.
During the examination, the NRC inspector found no hole enlargements or anchor bolt angles (parallelism of bolt nut surface to washer surface) that appeared to violate the above installation specifications.
It is concluded by the NRC inspector that this specific allegation appears to be more of a design concern by Mr. Messerly, than an improper installation construction' practice having been implemented by him, i
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There were no violations or deviations identified in this area of the inspection.
Aller:ation 6 1.
Discussion During Mr. Messerly's interview on April 14, 1983, Mr. Messerly stated, "There was a welding foreman out there that done [ sic] a lot of welding illegally without documentation, such as lugs _on pipes without purges."
In addition, Mr. Messerly identified three individuals who would have knowledge of attachments (lugs) being welded on pipe without an inerting purge 1/, with specific reference to the 832' elevation in the reactor containment building.
2.
Conclusion - Alleoation 6 i
As noted in attachment 1/, two individuals identified by Mr. Messerly were interviewed concerning their alleged knowledge of lugs improperly welded on to stainless steel pipe without purging the pipe when required.
Both interviewees denied any knowledge of improper welding activities.
During this inspection, the NRC inspector conducted an onsite follow up review of this matter.
1 The licensee's pipe welding procedures had been established prior to the initial piping installation early in the construction phase.
The procedures and implementation activities had been inspected and documented on numerous occasions throughout that phase of construction by the NRC senior resident inspector and independently by NRC regional staff personnel.
Therefore, during this inspection, the NRC inspector limited the review to pipe welding purge requirement established by the licensee.
The NRC inspector observed that the primary welding procedures associated with safety-related piping are B&R CPM-6.9, Appendix D,
" Welding and Related Processes," and B&R Inspection Procedure QI-QAP-11.1-26, "ASME Pipe Fabrication and Installation Inspection."
Paragraph 3.5 of this procedure, states, in part,
" Purging snall be maintained for welding of attachments to stainless steel pioing naving a wall thickness of 1/4 inch or less for field welds only.
This may be waived on a case-cy-case basis by the PWE.and Engineering.
This waiver shall be documented on the applicable WDC."
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18 In discussing this matter with the cognizant project welding engineers, the NRC inspector was informed that when a welding purge is required for attachment welds, the requirement would be noted on the weld data card (WDC) and a " Hold Point" established for verification by a QC inspector.
However, in instances where the purge is waived, an interoffice memo waiving the purge is attached to the WDC.
The i
interoffice memo is controlled by a chronological numbering system and filed within the permanent record files.
It was further pointed out by the B&R welding engineers that the majority of stainless steel piping at the 832' elevation, have pipe wall thickness in excess of the limiting 1/4" wall, therefore, an inerting purge would not be required for weld of attachment lugs.
Based on the fact that prior NRC inspections have not identified a concern in this area, that Mr. Messerly's allegation lacks speci-ficity (i.e., safety-related piping, pipe line numbers, size, location, etc.), that the majority of stainless steel piping at the 832' elevation exceeds 1/4" wall thickness, and that persons named by Mr. Messerly did not support the allegation, this allegation was not substantiated.
There were no violations or deviations identified in this area of the inspection.
Allegation 7 1.
Discussion It was observed by the NRC inspector in Mr. Messerly's affidavit of February 3, 1983, and during his interview on April 14, 1983, he stated he did not remember seeing the posting of a copy of NRC Form 3, " Notice to Employees," on three main onsite bulletin boards.
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2.
Conclusion - Allegation 7 The Code of Federal Regulations, Part 50 (10 CFR 50), was revised by
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47 FR 30452 to add 10 CFR 50.7, " Employee Protection." The change was published July 14, 1982, and had an effective date of October 12, 1982.
An important element of the change is that of a require-ment to post NRC Form 3 at loca; ions where the form can be readily viewed by employees on their way to or from their place of work.
During a prior review of this matter by the NRC senior resident inspector (SRI) (see NRC Inspection Report 50-445/83-03; 50-446/83-01, dated March 28, 1983), the NRC Form 3 was observed by the SRI to be posted in early January 1983.
However, the precise date (between October through January) of the posting of NRC
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19 Form 3 could not be established.
B&R personnel records indicate that Mr. Messerly was terminated on December 6,1982.
The allegation cannot be refuted nor substantiated.
Furthermore, the matter lacks any technical merit relative to an impact on the
'e safety of t,'1e plant.
There were no violations or deviations identified in this area of the
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20 SUPPLEMENTAL INSPECTION September 9 - 22, 1983 i
1.
Ofscussion As noted in the attached assistance to NRC inspection report,
" Supplemental," dated September 7, 1983 2/, during the course of an unrelated investigation, information was received that a former B&R millwright had drilled holes through 'rebar without the required engineering authorization.
During the period September 9 - 22, 1983, the NRC inspector conducted an onsite follow up on this matter.
From the information provided by the interviewees, the NRC inspector identified the specific " Trolley Tracks" 2/, as the drum and spent filter handling equipment, liner transfer trolley process aisle rails, located on the 810'-6" floor level, in room 252, of the fuel handling building.
The system is currently in the preoperational testing phase; however, this system is not a safety-related system.
In reviewing the construction documentation records regarding the installation of the rail assemblies, the NRC inspector found that the rail base plates, rail clips, drilled Hilti anchor bolts, and rails were installed per drawing,
" Anchoring Details for Radwaste Solidification System," Figure 39, Sheet 5 of 5, and by direction of Design Change Authorization (OCA) 7041, Revisions 4, 8, and 9, dated October 22, 1980, October 28, 1982, and November 11, 1982, respectively.
It was observed by the NRC inspector that Orawing Figure 39, Sheet 5 of 5, contained the following pertinent notes, "2:
Expansion bolts and base plate may be moved in east west direction to avoid interference with rebar running in north-south direction. " and, "3:
For rebar running in east west direction, holes may be drilled through the uppermost #18 bar @ only one rail location and expansion bolts shall be installed through the hole (it is assumed that bar interference shall occur at any one rail only)."
2/ See attached assistance to inspection report " Supplemental," dated September 7, 1983, Report No. A4-83-005.
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In addition, Revision 8, of DCA 7041 directed the addition of extending l
the length of the rails from the original 24'-3" long to 27'-6" (3'-3" section added to east end); also, Revision 9 permitted the modification of Hilti bolts (shortening) to avoid cutting any additional
.q rebar.
1
<j The NRC inspector met with the superintendent of the millwright department and interviewod millwright craft personnel that were directly i
1 involved in installation of the rail assemblies.
During the interviews, the NRC inspector found that the rail assemblies were installed during two different time periods.
Although actual dates were not established, it appears that the initial 24'-3" rail sections were installed in late
+
1982 and the 3'-3" extension sections were installed early in 1983.
The individual interviewed on September 1, 1983 2/, stated that he was not aware of the 3'-3" extension of the rails; therefore, his reference to his work activities involved only the installation of the initial 24'-3" rail sections.
In addition, it has been established that, aside from the core drilling foreman, five millwrights and one millwright foreman were directly involved in the installation of the base plates and rail assemblies.
(Three of the millwrights and the millwright foreman were individuals also interviewed.)
2.
Inscection Findinos As a result of the onsite followup inspection, records review, and interviews with personnel, the inspection findings are as follows:
As stated by the millwright interviewed on September 1, 1983 2/, and a.
acknowledged by other millwrights, only the east-west, #18 rebar, t'
running parallel with the east-west rail, was drilled through to accommodate the 1/2" Hilti bolts which secure the rail base plates i
to the 810'-6" floor.
This rebar cutting was authorized per Note 3, Drawing Figure 39, Sheet 5 of 5, DCA 7041.
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The alleger stated that the 3'-3" extension rails were installed in y
accordance with the DCA 7041, and that rebar was drilled through for the south rail Hilti bolts by the steel fabrication department j.;
drilling crew and that no unauthorized rebar was cut during g
installation of the 3'-3" rail extension.
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The millwright foreman stated that during installation of the 24'-3"
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rail base plates, the steel fabrication department drilling crew foreman arrived with the "rebar eater" drilling equipment by s
i himself, therefore, he assigned one of the millwrights to assist the
' drilling crew foreman in drilling the holes in which rebar required being cut.
He further stated that only rebar that was authorized to be cut per the DCA was cut.
d.
During the inspection, two of the millwrights interviewed stated that north-south rebar was encountered during drilling Hilti bolt holes for base plates for the north rail and that since cutting of the particular rebar was not permitted by the OCA, the Hilti bolt was modified (shortened) as authorized by Revision 9 of DCA 7041.
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The NRC inspector had a TUGC0 licensee representative locate and verify the modification of the specific Hilti bolt.
The bolt was located at the west end of the north rail and further supports the millwright's contention that no unauthorized rebar was cut.
In discussing the use of the core drilling equipment with the craft e.
supervisory personnel, the NRC inspector was informed that there is
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no hard set policy as to who can or who cannot use the core drilling equipment as long as the equipment is used properly and the drilling being done is authorized and directed by craft foreman or supervisory personnel.
As with the millwright interviewed 9
September 1, 1983 2/, wherein he stated that when the core drilling s
foreman did not show up, he (the millwright) completed drilling the remaining (approximately 10) 1/2" diameter holes for the south rail base plate Hilti bolt 3.
f.
The NRC inspector found no evidence to support the allegation that j
unauthorized cutting of rebar was done during installation of the j
" Trolley Tracks" for the drum and spent filter handling equipment.
3.
Results The allegation that unauthorized cutting of rebar was done curing installation of the drum and spent filter handling equipment process aisle rails is considered to be unsubstantiated.
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION a
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OFFICE or INVESTIGATIONS FIELD OFFICE REGION l\\
e' 511 RYAN PLAZ A ORIVE SulTE 1000
- 4, ARLINGTON. TE A AS 76011 i
ASSISTANCE TO INSPECTION REPORT f g May 20, 1983 i
SUBJECT:
COMANCHE PEAK l
ALLEGED IMPROPER CONSTRUCTION PRACTICES 1
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REPORT NUMBER: A4-83-005 1
i
.1.
On February 3,1983, Robert L. MESSERLY provided an affidavit to the Citizens j
Association for Sound Energy (CASE), an intervenor that included three alle-gations regarding improper construction practices by Brown & Root personnel at the Comanche Peak Steam Electric Station.
MESSERLY alleged the following:
That Brown & Rcot employees drilled undocumented unauthorized holes through a.
rebar, and such cutting of rebar was done at the direction of supervisors.
b.
That the main steam line pipe in Unit I was moved using the polar crane, thereby placing the pipe under unsafe tension.
That a Brown & Root employee used a cutting torch on hanger material in c.
violation of procedure.
2.
On April 6, 1983 MESSERLY was contacted by the reporting investigator, and a meeting was arran,ged with MESSERLY for the following day.
MESSERLY contacted reperting investigator on April 7,1983, and requested the meeting be changed to April 8,1983.
3.
On April 8,1983, NRC OIFO Director R.K. HERR and the reporting investigator met MESSERLY at a restaurant in Fort Worth, Texas.
MESSERLY was acccmpanied by Ms. Juanita ELLIS, a CASE representative, and Ms. ELLIS' husband.
Ms. ELLIS wished to record the meeting; however, 01FO:RIV was not previously j
informed of her intended presence nor of her desire to record the interview.
OIFO did not have a recceder, and in accordance with 01's policy, the ceeting was rescheduled.
On April 10, 1983, arrangements were made to use a room at the U. S. Attorney's office, Fort Worth, Texas, and for a court j
reporter to transcribe the interview of MESSERLY.
i 4.
On April 14, 1983, MESSERLY was interviewed at the U.S. Attorney's office with j
Ms. ELLIS present.
MESSERLY's testimony was taken under cath, Attachment (1),
and Ms. ELLIS made her own personal recording of the interview.
In his testi-e many, MESSERLY expanded in detail on his original allegations.
MESSERLY named Brown & Root employees responsible for the alleged improprieties and those who could substantiate his allegations.
MESSERLY also identified numerous employees by title, and agreed to later provide the corresconding names when he was able to refresh his memory with his personal records located at his residence. MESSERLY also provided the NRC with a copy of a log.
MESSERLY
'e.v.plained that he r.aintained this log tc dccument the cutting of rebar at 1
Ccranche Peak.
(Note:
MESSERLY did net allege that ali the entries in the log documented urauthorized cuts thecugh rebar, bu. rather that some of the entries in the log may have teen for holes drilled through retar
.nat ma;. not have had the aporopriate acccmpanying authorizations.)
During this interview, MESSERLY made a fourth allegation regardinc instances of m, a r e et
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5.
On April 21, 1983, a copy of the recorded testimony was mailed to MESSERLY at his residence.
On April 27, 1983, MESSERLY was contacted by HERR, and acknowledged receipt of the transcript, but postponed giving the names of the Brown & Root employees he had identified by title in the transcript.
MESSERLY
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stated he had not as yet had an opportunity to read his entire testimony.
On April 29, 1983, MESSERLY was again contacted by HERR, but he again postponed providing the names, explaining he was very busy.
On May 1, 1983, the reporting investigator telephoned MESSERLY at his residence, and MESSERLY provided twelve, additional names of Brown & Root employees at Cemanche Peak he alleged had kncwledge of unauthorized cuts through rebar, 6.
On May 3, 1983, interviews were initiated at the Comanche Peak site addressing the four allegations.
MESSERLY identified 38 individuals allegedly responsible for, or having knowledge of, the allegations.
Review of employment records determined that eighteen individuals were no longer employed at Ccmanche Peak.
i 7.
Between May 3, 1983 and May 10, 1983, 19 Brown & Rcot employees and 1 Dravo Constructors Inc. employee (formerly employed by Gibbs & Hill) named by i
MESSERLY were interviewed, and signed, sworn statements were taken from 19 of them. One Brown and Root employee interviewed left on vacation before a signed, sworn statement was obtained from him, and his testimony was recorded in the form of a Results of Interview.
One Piping Design Serivces Inc. -
engineer was identified by the reporting investigator as responsible for the movement of the main steam line.
This engineer was interviewed, and executed a signed, sworn statement.
8.
Nine individuals alleged to have knowledge of improper, unauthorized cutting of rebar were interviewed and provided sworn statements.
These individuals denied having knowledge of rebar that was cut without proper authorization.
A 10th individual responsible for issuing the Component Modification Cards (CMC),
authorizing cuts through rebar, was interviewed and provided a signed, sworn 1
statement denying knowledge of any procedural viclations.
Testimony icentified instances where rebar was accidentally cut, but this testimony aisc establishedq that in these instances, CMC's were obtained after the cuts were reported to the enigneers.
There was no testimony received indicating that holes were drilled or rebar was cut without proper documentation, and no evidence was found to contradict the testimony of these individuals.
9.
Three Srewn & Root employees alleged to have knowledge concerning the use of the polar crane to move a portion of the main steam line in Unit I were inter-viewed and provided signed, sworn statements.
A Piping Design Services Inc.
engineer responsible for relocating the steam line, provided testimony of his evaluation and directicn of the relocation of the line.
The testimony taken from these four witnesses indicated that the relocation of the main steam line was done under the direction of engineers, and was acccmplished tc remove stress on the line and to return it to its designed lccaticn.
No testimony was recievec to indicate that the line was " cold sprung" or installed under stress.
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A4-83-005 Page Three 10.
Eight Brown & Root employees alleged to have knowledge concerning the improper use of cutting torches on hanger material were interviewed.
Two witnesses stated they remembered an instance during the redesign of a hanger in which a piece of tube steel was discovered to have had the bolt holes enlarged using a torch, which was a procedural violation.
The testimony of the two witnesses indicated that this hanger was scrapped because of the procedural violation, and was replaced with new material.
The other six had no knowledge of improper use of cutting torches or hangers.
11.
Two Brown & Root employees were interviewed concerning their alleged knowledge of lugs improperly welded onto stainless steel pipe without purging the pipe.
Both executed signed, sworn statements, and indicated that they did not know of any instances where welding was done on stainless steel pipe which required purging by procedure unless a " purge deletion" was received from the engineers.
12.
All of the employees mentioned by MESSERLY in his affidavit who were still employed or available for interview denied the allegations made by MESSERLY.
No evidence was uncovered during these inquiries which indicated deception on the part of the witnesses.
The witnesses ranged from pipe fitter helpers to Brown & Root superintendents.
A Piping Design Services Inc. engineer and the Dravo Constructors Inc. project manager also provided testimcny which centra-dicted the allegations.
13.
The signed, sworn statenents are maintained in 01F0:RIV.
No further inquiries are anticipated unless staff inspections identify additional pertinent information that would tend to substantiate the allegations or discredit the interviewees.
Attachments:
(1)
Testimcny of MESSERLY - dated 4-14-83 (2)
List of Interviewees (3)
List of terminated employees identified in Attachment (1)
REPORTED BY:
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.o H. 'Brtoks Griffin, !>fvestigator 01 Field Office, Recion IV APPROVED BY:
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[ p Ricnard K. herr, Director 01 Field Office, Region IV cc:
W. Ward, 0I:DF0 - w/ attachments W. Collins, RIV - w/ attachments E. Johnson, RIV - w/o attach: rents
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Ett P) AN PLAZ A opivt SU:Tl 100C
' ' '......C AR'INGTON TEAAS 76%'1*.
ASSISTANCE TO INSPECTION REPORT
" SUPPLEMENTAL" September 7, 1983 SU5 JECT:
COMANCHE PEAK STEAM ELECTRIC STATION:
ALLEGED IMPROPER CONSTRUCTION PRACTICES.
REPORT NUMEER:
A4-E3-005 1.
Durine the course of an unrelated investigation, information was received, from an indivicual wrio recuested conficentiality, that a former Ero.n & Root, Inc.,
millwright hac crilled holes thrcuch rebar without the recuirec engineering authorization.
2.
On September 1,1953, this millwright was interviewed and provided information wherein he stated he possibly drilled holes through.rebar in a concrete floor without a Component Modification Card (CMC) or a Design Change Authorization (DCA).
He explained that he drilled about 10 holes in January 1983 while installing 22 metal plates using a core drill.
He said these metal plates were used te secure the trolley tracks located ir the Fuei handiino Euildino as par: c' the Waste Menitor System.
He stated tha-he anc his crew used a' ccre drill b:rr:we: free the Core Crilling Crew.
Tr.e riib. -igtt saic that tne heles nace wi:n the cere crill were locatec cr tr.e scu:n.es: ccrner of ne trolley tracks.
he explained tha the ::ivecrints he usec astnor12ec tne cu ting of cne :,iece of reba or. sect h,le, ar.c he accec ina; it is his belief the hcies wire crillec prc;erly.
3.
Tne Resuits of Ir.terview with the former Erewn & Roct r.iliorigr. is maintainee ir 0; Field Office, Region IV.
Attact.er.: (1: - Fesults of Ir.terview.i r. tiliwright, datec Se::e. er 1, 1953.
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T. F. kesterrar., RIV (w/c attach er-)
ATTAC& TENT 2
l III OfffcTaI File ce77 DEC 211982 Dock 3ts:
50-445/82-29 50-446/82-15 Texas Utilities Generating Company ATTN: Mr. R. J. Gary, Executive Vice President & General Manager 2001 Bryan Tower Dallas, Texas 75201 Gentlemen:
This refers to the investigation conducted by Mr. R. K. Herr of the Office of Investigation, and Messrs. L. E. Martin and D. L. Kelley of our staff during the period August 4 to September 17, 1982, of activities authorized by NRC Construction Pennits CPPR-126 and CPPR-127 for the Comanche Peak i
facility, Units 1 and 2.
l Areas examined during the investigation and our findings are discussed in the enclosed investigation report.
Within the scope of this investigation, we found no instance where you failed to meet NRC requirements.
I In accordance with 10 CFR 2.790(a), a copy of this letter and the enclosure will be placed in the NRC Public Decisnent Room unless you notify this office, by telephone, within 10 days of the date of this letter and submit written cpplication to withhold information contained therein within 30 des of the date of this letter. Such application must be consistent with the require-ments of 2.790(b)(1).
Should you have any questions concerning this investigation, we will be pleased to discuss them with you.
Sincerely, l
'Drjgtnal Signed by g 3,gAoSEN
G. L. Madsen, Chief Reactor Project Branch 1 Enciosure:
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Appendix - NRC Investigation Report op' 50-445/82-29; 50-446/82-15 D Off.tcrar Fi.le ce~".
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My name is Arvill Dillingham, Jr., better known as "J.R. "
I live in Glen Rose, Texas.
I worked for Brown & Root f or approximately ten years, including about seven year. at the Conianche Fe4 nuclwr power plant.
During that time I was a boilemaker, and latei a builermaker General foreman for about three or four years. Wnile 1 was performing my duties as a general foreman, I saw a lot of things at the Comanche Peak plant which were not according to procedures and travelers, many of which could jeopardize the health and safety of the public.
I knew that something needed to be done before the fuel came in for the plant.
I thought it over and decided that I would go to Houston to the Brown
& Root main office and report these violations to the President of Brown & Root direct, Mr. Tnomas Feehan.
I went to Houston and met with Mr. Feehan.
I told him of some violations that were going on and told him that after ten years of experience, I thought some of those people might know better and show better craftsmanship and intelligence than they had used at the nuclear plant in Southport, North Carolina.
I told him about working at the North Carolina plant on some weld seams that were not supposed to lose more than an uunce of radiation per year (we were told). These seams, some of which were approximately 100 feet long, were in some instances leaking as much as 65 lbs. every three seconds, according to what was indicated when we turned the gauge of f and watched the gauge drop.
At first we tool the test channel off and started repairing the bad welds, which took several days. We were getting approximately 28 pin holes and indi-cations per foot. This was taking too much tinie to repair properly, apparently, so somebody in Erown E Root's organization canie up with the brainy idea of having us take the first foot or eighteen inches off, repair the weld, replace the eighteen inches of test channel, block it off, and just hydro the one-foot area.
QC assumed we had the whole 100 feet under pressure and bought it off as is.
In my opinion, that is lack of craftsmanship or experience or just downright sabotage.
Personally, I think some of these people should go to prison for this, and I told Mr. Feehan that.
I also gave Mr. Feehan a letter with some things I knew about and that other people had told ne about. He said he would send an investigation crew to investi-gate tne Comanche Peak charges, and he did.
I know of much more that I did not tell him or the investigative crew and have rot in t.e past reported to the local NRC office because I feel the NRC is not inteiestec ir. protecting the people's welfare and it will be coverea up.
I nave been told by CASE (Citizens Association for Sound Energy), the intervenor 'n the tomanche Peak operating license hearings, that I am required by law to report any problems which might affect the health and safety of the Dublic to the Nuclear Regulatory (omission.
I am therefore asking CASE to send this affidavit to the NM as my method of reporting it', in the hope that by doing so it will force tha NK to really look at the problems I've identified.
I would like to get this information to Congress or someone who's really interested in the safety and welfare of the people of Texas, 59' m de t sen ff Q pulv P J B -m
. Some of the ciuff that onus on, for instance:
My superintendent had us install some lignt poles inese pole 3 am ondopo or litt. ir..i h the stain-less steel liner around the reactor cavity.
They are used when unloading the reactor underwater so the people can see what they are doing.
They are nothing out stainless steel pipes with one end tapped off and holes drilled in them.
As we were drilling the holes, drill shaving; fell inside the pipe. We also used some cutting oil. My superintendent came out to the shop where we had the holes drilled and said "You don't have those poles in the hole yet?" And I said, "No, we're going to take a pencil grinder and deburr them, and take a steam hose and steam all the oil and shavings out."
He said, "That's bull.
Get those poles on down to the hole so the electricians can install the lights I
on the poles." By our not taking about 15 minutes a pole to clean them right, the poles are now installed in the proper location. They pose a serious safety problem. Wnen they're refueling, the shavings can be washed out of the pipe by the current when removing the reactor head underwater, and also, removing the old fuel cells causes a current.
The shavings can be washed inside the reactor, which can jam the fuel cells, could even fuse to the control rods and possibly cause a meltdown.
I feel that their doing that is lack of sense, lack of ex-perience, or sabotage. Maybe Brown & Root's got an explanation for it, but I'd like to hear it.
I also have information which indicates that during the early stages of
.l construction around the time when the reactor cavity was being poured, concrete aggregate material from a reject pile was used. My concern is that if the 700 ton reactor is sitting on rejected concrete, it could result in the weight shifting to the loop pipe, causing it to crack or shear off, which could result in a meltdown.
I am convinced tnat because I went tc the Brown & Root Houston office with my concerns about the safety of the Comanche Peak plant and also the one in North Carolina, I ac no longer employed with Brown & Root.
They already had my name pulled off the board as General Foreman when I got back from Houston before the investigation group ever got to the plant. My future was already oecided before 1 ever got back and before the investigation was ever started.
Later I was confined to one area of the shop for five weeks.
I called Mr.
Rice in Houston and asked him how long I was going to be confined and told him that if I was going to be confined, the people I had made the charges against should also be confined because they were still violating procedures.
I had called TUSCO before I called Houston and they removed me from standing ir. the shop and put me in a little tool room in the shop.
Mr. Rice said, "As far as I'm concernec. the investigation nas been over with and furthermore, you called TUGCO.
If I worked for Brown & Root. I'd call Brown & Root.
If I wo'rked for TUGCO, I'd call TUGCO.
If I worked for the federal government, I'd call the federal government. But you called TUGCO."
I said, "I tried to call you guys first, but you weren't there." He replied, "You think we're going to sit by this G.D. phone and wait for you to call?"
I said, "Well, maybe they're try-ing to discourage me here until 1 quit." He said, "Maybe you're finally getting the idea." I said, "As long as you guys can pay me General Foreman's wages, I'll sit in this little tool room forever." And he said, "We'll see about that."
Then the next Monday morning, I was given the choice of either working as a pipe journeyman, which would have greatly reduced my salary and relieved me of all my responsibility as a supervisor.
So they R0F'd me (laid off as part of reduction of force).
But they were increasing my department at the same time and after.1 left.
The information precedino was given tn CASE in the form of an affidavit on December 18,19d:.
Howew r. I did not want L4m to turn. It in in the hearings or to turn it over to the NRC or the utility.
I gave the information to some newspaper reporters, and an article ran in the FORT WORTH STAR-TELEGRAM on January 7,1983.
A copy of that article is attached.
After that interview, I was shot at and have been on the run ever since and have been in touch with CASE a few times by phone from different states.
One night when I came home, j
I found my cat; its head had been cut off smooth and its body was missing.
Since the article appeared in the paper, I have had a front-end problem with three different vehicles (one truck and two cars); they all appear to have the same problem -- the nuts were just about to fall off the tie-rod ends.
I've been scared to go back and sign up every six weeks for my unemployment because I'm scared someone may shoot me.
Some of the reasons I'm scared is because of the things I know about at Comanche Peak and another nuclear plant where !'ve worked, the South Port, fiorth Carolina, Brunswick Project liuclear Plant. As I mentioned before, there are weld seams around the Reactor Core and new spent fuel pools which we were told were not supposed to lose 1/2 ounce of contaminated liquid per year per seam.
i These seams were approximately 100 feet long;when we tested these seams, some i
of them were losing approximately 65 lbs. a minute Instead of repairing some of these seams, the gauge was blocked off and pressure was put on the gauge only.
When the inspector passed the weld seam, he thought the whole 100 feet was under i
pressure, not just a few inches.
Also, some of the stainless liner walls broke loose from embedded plates that are in concrete walls which some of these plates were improperly welded.
By these walls breaking loose they sprang out several s
inches from concrete wall; therefore, when refueling the reactors, the stainless steel liners were flooded with water. Of course, tne weicht of the water will push the liner walls back to the concrete.
Af ter the re eling process is over i
and water is drained out of the liner, the walls will sr ing back out, which l
could result in welds cracking or walls splitting. When I reported these viola-tions to Brown & Root's Vice President, he told me he was not that concerned about the gauges being blocked off but he was concerned about the walls breaking loose.
If I had told him of improper welding on these walls, I wonder if he would have been concerned at all? I feel these problems should be repaired.
Regarding Comanche Peak nuclear plant, there are safety violations such as torquing.
For instance, quality control is supposed to verify the torquing of piping support that should be torqued at 130 lbs.
The hanger is on a 20 foot ceiling with' a scaffold built to them. Quality control is on the floor; the torque wrench is sent down to get QC to verify the number and setting of the torque,
4 wrench and carried back up and placed on the nut before torquing.
QC hears a torque wrench click twice on each nut and buys off (approves) the hanger. What QC did not know was that the construction personnel-had a second torque wrench and also had a nut welded on the scaffold. The second torque wrench was set at a low torque poundage such as 3 lbs. and they clicked it twice. Therefore, the nut on the hanger was never torqued; only the nut on the scaffold was torqued.
There were also violations such as pipe supports around the pipe. For instance, 3/16" clearance is supposed to be maintained on each side and on top and the pipe is supposed to be resting gently on the bottom of the support.
For instance, a 2" pipe: a constructisn supervisor will climb on the pipe and get some of his creenen so when QC comes to inspect the support, the weight will push the pipe to the bottom.
In some cases, the pipe was binding so tight they would use a timber to jack the pipe down from the ceiling while OC bought off the pipe.
In some cases, when they can't get the right clearance on each side of the pipe, they take a grinder and grind between the pipe and tube steel, which in some cases results in a~ reduction of wall thicknbss of pipe.
I believe this could result in a rupture of the pipe.
Construction has also tried to straighten a pipe support by using a sledge hammer; this is done quite of ten. An employee told me that while hitting on the hanger he also hit the pipe and caved in the side of the 2" pipe 1/2 inch or more.
He reported it to his supervisor who said not to tell anyone and covered i' up with I.D.
tags.
Another incident is improper personnel designing and engineering pipe s upports.
For instance, one helper told me while he was employed at the plant he designed many pipe supports for engineers. One day he wondered if they were using his engineering and if they were then checking his work, so he decided 1
that he would design a hanger improperly and send it to engineering. The engineer i
passed it on to construction which built the hanger and it is presently installed i
impro perly. The helper said that he did not want to 90 to any NRC hearing but he would love to have a showing and he could show many things if he was allowed to take investigators and actually shoa them the supports in the plant. Other helpers have also been involved in making major decisions for which they are not qu'alified.
Another violation is a sensor in a dam was run over and broken by a bulldozer.
I understand tha'. these sensors are placed in the dam in a vertical position in order to tell Enether the dam moves or not. This sensor was not removed or re-paired.
It vas held up and dirt packed around it while being embedded in the dam.
The canstruction company, Brown & Root, lost a $3 million contract at Crystal River Foaer Company in Florida, by a dam breaking, I was told by one cf the Vice Presidents of Brown & Root. What concerns me is that if this dam breaks, they will lose more than a 53 millinn contract; it will endanger many li ves.
There is also a violation that concerns me regarding the use of rejected concrete material in the early stages of the plant when the reactor core was poured. A friend of mine told Brown & Root's Vice President's investigating crew that he was a front end loader operator at the concrete plant and one day a QC inspector told him that the concrete should be thrown away because it was hard and dried. The inspector walked away and qy friend started throwing it away and a supervisor told him to put it back in and use it and they did.
1 My friend also told the investigating crew of some type of sampling machine that tells whether there are good samples or bad samples in the concrete.
Il l
had.a wire run to it while QC watched the machine to verify the use of good samples.
2 Personnel would pull the wire to make it read good when it was not. My friend also told o' other people that know of these violations and as far as I know, i
Brown & Root did not contact any of these people, but talked with one of their supervisors and his brother that worked at the batch plant; they, of course, told them that they knew nothing of this incident and since the superintendent is deceased, they did not see any further investigation of this incident.
I am sure that the 4
NRC is aware of this statement, because it was in the FORT WORTH STAR TELEGRAM article (attached).
Undoubtedly, they are not concerned about the situation.
I have not been contacted and nei ther has qy friend.
J I
If indeed a 700 ton reactor is setting on rejected concrete, you have a very serious problem as tne reactor get'., hot and pegins to rpove around, the concrete can give putting stress on the reactor piping, which could cause it to shear or crack, wnich could even result in a meltdown. This could also be a problem in case of an earthquake.
I know of many, many more problems and violations than I can mmember right ncw. What is funny to me is the big deal everyone made of Russia's 300 lb.
nuclear satelite falling back into earth's atmosphere, when we have a possibility of a 700 ton reactor setting on rejected concrete and no one is concerned. If all nuclear power plants in the U. S. are built the way the ones that I have worked at am, we are in trouble. We'd better make friends with Russia so we will have somewhere to 90.
But speaking seriously, I think this should be investigated by someone with a little construction experience or common sense.
It has been drawn to my attention that I am not a civil or a mechanical engineer and that it is not up to me to decide whether these plants are safe or not. But I feel it does not take a civil or mechanical engineer.
Even a 6-year-old would know these violations should be corrected.
I have read the foregoing 5-page affidavit, which was prepared under my personal direction, and it is true and correct to the best of my knowledge and belief. The thoughts and words expressed therein are my own thoughts and words (with the exception of minor gramatical changes, either to correct spell-ing or to clarify what I rear.t, which did not cnange the intent of my thoughts).
m Or
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Date:
March 31,1983 STATE OF TEXAS On this, the 31st day of March,1983, personally appeared Arvill "J. R."
Dillingham, Jr., known to re to be the person whose name is subscribed to the foregoing instrument, and acknowledged to me that he executed the same for the purposes therein, expressed.
5 Subscribed and sworn before me on the 31st day of March,1983.
A 4
My Commission Expires: #d pr
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