ML20024E419

Affidavit of Lv Witt Re IE Investigation Repts 50-445/83-03 & 50-446/83-01

ML20024E419
Person / Time
Site:	Comanche Peak
Issue date:	06/27/1983
From:	Witt L Citizens Association for Sound Energy
To:
Shared Package
ML20024E412	List: ML20024E411 ML20024E413 ML20024E415 ML20024E416 ML20024E417 ML20024E418 ML20024E419
References
NUDOCS 8308100392
Download: ML20024E419 (23)

Text

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, o s-AFFIDAVIT OF LARRY V. WITT Q: Please state your name and address for the record.

A: My name is Larry V. Witt. I live at P. O. Box 278, Glen Ecse, Texas 76043.

Q: Have you reviewed a copy of Investigation Report 50-445/83-03, 50-446/83-01, which was prepared by the Nuclear Regulatory Commission (NRC) regarding allegations made by Arvil Dillingham, Jr. ("J . R. ")?

A: Yes, I have.

Q: We call your attention to page 9 of Appendix B of that Report, item 7 (a) , which states:

"Mr. Dillingham apparently stated when intereviewed by the writer of the article that rejected aggregate was mixed with concrete that was subsequently poured to form the base for the nuclear reactor. The article stated that a Larry Witt was the B&R equipment operator who had apparent first hand knowledge of'the matter. The article also stated that Mr. Witt could not be reached for comment."

Are you familiar with the article mentioned in the report?

A: Yes, I read the article in the paper.

Q: Are you the Larry Witt mentioned in that article and in the NRC investigation report?

A: Yes, I am.

Q: Please read the portion of the investigation report beginning on page 10, about the middle of the page, the sentence which begins "The above memorandum indicates that a number of other people were interviewed by the B&R investigative group, one of whom was Mr. Witt" through the end of that paragraph.

Please comment on that porti.sn of that paragraph.

A: The material was actually rejected gravel and perhaps rejected sand (I am not certain about the sand). Regarding the scales, the wires were pulled leading to the scale head for the ice bin (instead of leaning on the wires as stated in the report).

Q: The report states that the "B&R memorandum indicates that the investigation relative to use of rejected aggregate was apparently par-tially substantiated but of no concern in that the aggregated pile, rather than actually being unacceptable, simply had not been tested prior to

~

I use as required." Do you agree with that statement?

8308100392 830803 l PDR ADOCK 05000445 G PDR

e A: No. I assume they are referring to a time or two when we took material from the outside set of bins going into the concrete pour without being checked. But what I was referring to was material that had been taken out of the bins because it was rejected and failed to meet the specifications. We dumped it out on the ground with a front end loader and later fed part of it back into pours. And on some of them, they were Q pours (safety-related pcurs). It was done so they could continue.

Q: Did you personally operate a front end loader, as stated on page 11 in the first full paragraph of the report?

A: Yes, but the material was not dry and lumpy cement that had been rejected. This makes it sound like the material had already been poured and hardened, then broken up and reused again. This is not what I was referring to.

Q: Is there other information you are aware of regarding the concrete porrs and the concrete batch plant?

A: Yes. There were times when I was told to take a front-end loader and pull back the material from the mouth of the feed hopper so that a sample could not be taken. This was done because it was feared many times that as the bins got low in material, segregation problems within the gravel could cause material to be rejected. Therefore, many times when the bins were getting low, I moved the material. This can be checked by reviewing the final daily sample reports of Hunt Lab. You'd find that many times there was no material left in the bins at the end of a pour. It's pretty hard to plan a concrete pour in such a way that you only got just exactly enough sand and gravel to make a pour. The odds are very much against your taking a pile of sand and a pile of gravel and saying that you're going to use just that amount of sand and gravel for a pour of concrete. It can happen, but the records will show that it happened far more often than the mathematical odds would bear out.

Q: Why did you do these things?

A: We did it under orders. My orders usually came from Bob Morris.

I was told to do it when they were getting low on materials so they could continue with the concrete pour.

a a s Q: Do you know where the concrete which contained this rejected material was used?

A: No. It could have been used anywhere. But I do know that a lot of times it was for Q pours.

Q: Please read the portion of the report on page 11, last paragraph, beginning with the first sentence "Regarding the above summarized allega-tion (a) . . . " through the sentence "It should be noted that only the B&R investigative group has been able to establish contact with Mr. Witt; all others have apparently failed."

Have you ever been contacted by the SRIC (Senior Resident Inspector-Construction, R. G. Taylor)?

A: No.

Q: Is it true that you are no longer an employee at CPSES?

A: Yes.

Q: Is it true that you have relocated from the Glen Rose, Texas, area to another state?

A: No. We do have a field office in Hobart, Oklahoma. But I still live in Glen Rose, Texas, and my business is still listed as Witt Energy Resources, Inc., in Glen Rose, Texas. Many of the people at the plant with Brown & Root, including Doug Frankum, B&R Project Manager, know me and know that I am still in Glen Rose.

Q: To your knowledge, have any attempts been made to contact you by phone at the of fice in Hobart, Oklahoma, or in Glen Rose, Texas?

A: Not to my knowledge, and I believe someone would have passed this information along to me.

Q: Did you ever sign for a registered letter, receipt requested, from the NRC Region IV office?

A: No, I did not.

Q: To your knowledge, did anyone affiliated with your business either in Hobart, Oklahoma, or in Glen Rose, Texas, sign for such a registered letter?

A: Not to my knowledge, and as I stated before, I believe someone would have passed this information along to me. No one I have asked about it knew anything about such a registered letter.

Q: How did you first become aware of the NRC investigation report about this matter?

.- .- . - . . . - . . - ~ . . . --.

A: J. R. Dillingham brought me a copy of it which he stuted he had received from Juanita Ellis, with CASE.

Q: And this is the first knowledge you had that the NRC wanted to contact you or wanted you to contact them?

A: Yes.

Q: Is it your intention to deliberately withhold information from the NRC?

A: No.

Q: Is it your intention to refuse to cooperate with the NRC or to assist the NRC in investigating allegations attributed to you?

A: No. I have authorized Juanita Ellis, with CASE, the intervenor in the operating license hearings for Comanche Peak, to get this affidavit into the hands of NRC investigators by the means she deems best so that they can proceed with their investigation.

Q: Do you have any knowledge or idea why the NRC inspector would have stated in the investigation report:

"Regarding the above summarized allegation (a) , the SRIC established that Mr. Witt was no loner an employee at CPSES and further established that he had relocated from the Glen Rose, Texas, area to another state. NRC Region IV personnel made several attempts to contact Mr. Witt by telephone at his new address, to no avail. A registered letter, receipt requested, was then sent to Mr. Witt requesting that he contact Region IV as soon as possible. Receipt of the letter was acknowledged but as of this date, Mr..Witt has not contacted the region. It appears that Mr. Witt does not intend to assist the NRC in in-vestigating allegations attributed to him. It should be noted that only the B&R investigative group has been able to establish contact with Mr. Witt; all others have apparently failed."

A: I don't think they want to talk to me.

Q: Obviously, CASE was able to contact you, weren't we?

A: Yes. Witt Energy Resources, Inc., is listed in the phone book in Glen Rose. And any of the Witts listed in the Glen Rose phone book would know how to get in touch with me. Also, Tommy Gosdin with TUGCO, the Sheriff, the County Judge, the Commissioners' Court, and many others could have told them how to contact me.

I have read the foregoing 4 -page affidavit, which was prepared under my personal direction, and it is true and correct to the best of ray knowledge and belief.

The foregoing affidavit was prepared under my personal direction, and the thoughts and words expressed therein are my own . thoughts and words (with the exception of minor gramatical changes, either to correct spelling or to clarify what I meant, which did not change the. intent of my thoughts).

Whem questions were posed, they were posed by CASE.

c , L lam y V. Witt Date: crim o 27. 1983 STATE OF TEXAS On this, the 23rd day of June ,1983, personally appeared Larry V. Witt , known to me 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.

Subscribed and sworn before me on the 23rd day of June , 1983.

l MK%

Notary Public in and/or the State of Texas l

My Conmission Expires:D 6 /%

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UNITED SVATES -

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NUCLEAR REGULATORY COMMISSION REGION IV

$11 RYAN PLAZA DRIVE. sVITE 1000

1. ""\ ,, - ,8

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• ARLINGTON, tex As 76011 .

In Reply Refer To: .

MAR 2 81983 Dockets: 50-445/83-03 50-446/83-01 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 under the Resident Inspection Program by our Senior Resident Reactor Inspector, Mr. R. G. Taylor, during the period of October 1982 through February 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 you and other members 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 e.

selective examination of procedures and represeitative records, interviews with personnel, and observations by the inspector.  :

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.

In accordance with 10 CFR 2.790(a), a copy of this letter and the enclosures will be placed in the NRC Public Document Room unless you notify this of fice, 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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Texas Utilities Generating 2 MAR 2 8 583 Company Should you have any questions concerning this inspection, we will be pleased to discuss them with you.

Sincerely, 1

/

G. L. Madsen, Chief Reactor Project Branch 1

Enclosures:

1. Appendix A - Notice of Violation

2.

• Appendix B - NRC Inspection Report 50-445/83-03

, 50-446/83-01 cc w/encis:

Texas Utilities Generating Company ATTN: H. C. Schmidt, Project Manager 2001 Bryan Tower Dallas, Texas 75201

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APPENDIX A '

NOTICE OF VIOLATION Texas Utilities Generating Company Docket: 50-445/83-03 Comanche Peak Steam Electric Station 50-446/83-01 Permits: CPPR-126 .

CPPR-127 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 Implement a Quality Assurance Program for the Fabrication and Installation of Electrical Underwater Floodlight 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 stified structures, systems, and components'. FSAR Section IA(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 assembly resulting in damaging fuel is very remote due to the design of upper and lower pole retention davices.

This is a Severity Level V Violation. (Supplement II.D.)

Pursuant to the provisions of 10 CFR 2.201, Texas Utilities Generating Company is hereby required to submit to this office, within 30 days of the date of this Notice, a written statement or explahation in reply, including: (1) the corrective steps which have been tak'en and the results achieved; (2) corrective steps 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 response time for good cause shown.

Dated: MAR 2 81983

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APPENDIX B O. S. NUCLEAR REGULATORY COMMISSION REGION IV Report: 50-445/83-03 50-446/83-01 Dockets: 50-445; 50-446 Category: A2 Licensee: Texas Utilities Generating Company (TUGCO) 2001 Bryan . Tower Dallas, Texas 75201 Facility Name: Comanche Peak, Units 1 and 2 Inspection At: Comanche Peak Steam Electric Station (CPSES), Glen Rose, Texas Inspection Conducted: October 1982 through February 1983 Inspector: x [

' R. G. Taylor, Senior Resident Inspector-

/ /N .5 Date Construction Approved:

'T. P/ W(sterman, Chief v I/6h2 Date Reactor Project Section A Inspection Summary Inspection Conducted October 1982 Through February 1983 (Report 50-445/83-03:

50-446/83-01)

Areas Insoected: Routine and special inspection, announced by the Senior Resident Inspector-Construction (SRIC) including facility tours, investigation of allegations, participation and assistance to the Construction Assessment Team Inspection, and other inspection related activities. The inspection involved 263 inspector-hours by one NRC ins'pector.

Results: Within the areas inspected, one violation was identified (failure to implement a QA program for fabrication and installation of underwater lighting poles.) '

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Detail's .

1. Persons Contacted .

Principal Licensee Personnel R. G. Tolson,' Site Quality Assurance Supervisor D. N. Chapman, Quality: Assurance Manager B. R. Clements, Vice-President, Nuclear J. T. Merritt, Manager of Startup J. B. George, Vice President and Project General Manager Other Personnel G. R. Purdy, Project Quality Assurance Manager, Brown & Root (B&R)

O. Frankum, Construction Project Manager, B&R The SRIC also interviewed other licensee and contractor personnel during the inspection period.

2. Licensee Action on Previous Inspection Findings (Closed) Unresolved Item (50-445/79-24) Quality of Unit 1 Reactor Building Dome Concrete. This item related to the need for additional assurance that a small amount of concrete placed on the Unit 1 reactor building doce i during a rain storm without appropriate controls by quality control was adequate. An earlier evaluation of the in situ concrete by a proprietary testing program had indicated that the material was acceptable. The testing program was found to be unauditable and therefore, some additional

! assurance was judged to be required. The licensee has now completed the structural acceptance test of the Unit 1 reactor building with special i attention directed to the repair area. The test was successful and no

] anomolies were identified in the repair area and therefore,'it is judged that the concrete is'of adequate quality.

~(Closed) Unresolved Item (50-445/80-20; 50-446/80-20) Design of the AC Instrument Distribution Panels. This item involved a finding th'at the segregation of safety and nonsafety wiring in the panels was nut in acccrdance with Regulatory Guide 1.75 but was in essential compliance with the panel design displayed by FSAR Figure 8.3-15. After discus-t sions between the SRIC, NRR personnel and the licensee's electrical engineering group, a method of correcting the-matter was developed.

FSAR Figure 8.3-15 was revised by Amendment 27 to reflect the method of correction. The SRIC has examined the implementation of the change in two of the four panels involved and had no further questions.

(Closed) Unresolved Item (50-445/81-14; 50-446/81-14) Control of Stainless Weld Repairs. This item involved an observation that a previously well controlled program for the con' trol of the number and extent of repairs

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3 made to weld joints in stainless steel pipe had become less well con-trolled due to personnel changes. The licensee revised Construction Procedure CP-CPM-6.90 to reflect proper identification controls for craft and QC actions effective in January 1982. The engineering controls were promulgated effective with the issuance of Procedure CP-EP-10.0 in. March -

1982. The SRIC has not. observed any instance where the procedures are not being complied with and therefore, has no further questions.

3. Action on Licensee Identified Design / Construction Deficiencies (Closed) Over-Torqueing of Safety Relief Valves. On September 10, 1982, the licensee informed the SRIC that a potentially reportable condition under the purview of 10 CFR 50.55(e) had been identified. It was reported that the main steam safety relief valves had been over seated by excessive torqueing to stop leaks during the main steam hydrostatic test. . It was found that the excessive tightening had damaged the valve seats in some instances and that some of'the valves appeared to have the valve stems bent out of tolerance.' 'By letter dated November 10, 1982, the licensee informed the NRC that after review, the matter was not considered formally reportable under the regulation. The SRIC has reviewed the documentation of the examination of the valves by the licensee ~. The examination did not reveal any significant damage had occurred to any of the valves that would have prevented the valves from lif ting under pressure which would satisfy the safety function. Some of the valves may have leaked under operating conditions which would be undesireable but not a safety hazard. The SRIC had no further questions on this matter.

4. Allegations By Dennis K. Culton On September 16, 1982, Mr. Dennis K. Culton made a limited public appearance before the Atomic Safety and Licensing Board hearing in the matter of TUGCO's application for an operating license for the CPSES. His statement during the appearance appears in the hearing transcript at 5551 through 5555. In addition, Mr. Culton furnished the Board with a written statement which appears in the record at 5556 through 5559. Based upon a review of the record, NRC Region IV determined that there were two areas of interest that should be evaluated for their validity and effect of

' safety of construction. The first area dealt wita the potential misuse of a group of drawings referred to as BRHL while the second dealt with the alleged splicing of safety-related or "Q" electrical cables. The SRIC was assigned to make the evaluation.

5. Allegation Relative to BRHL's Mr. Culton's concern in this area appears at Tr. 5552 through 5554 and 5557 through 555.8. His concerns can be summarized as follows:

(a) Based upon limited information, he was directed to generate isometric drawings giving support locations. He states at 5557 that he did not feel qualified to do this work in the manner directed.

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(b) The drawings that he and others in this group generated were released to the field unapproved and were used by the craf t labor personnel to locate and install supports _.

A BRHL is an isometric drawing made from a modified piping installation -

isometric drawing to identify the supports on the pipe and to provide locational information at an appropriate point in time. The drawing series has no unique title with the BRHL appearing only before the drawing number to distinguish it from the parent pipe isometric which carries the same number except for its unique prefix, BRP.

Discussions with various licensee personnel who are familiar with the history of the development of the BRHL's indicate that the need to gen-erate the drawings became apparent when planning was initiated for the as-build verification program as requ' ired by NRC IE Bulletin 79-14. The very early phase of the work appears to have started at about the same time that Mr. Culton was assigned to the drafting department and it is understood that he and others were hired and/or recruited from the field labor forces specifically for the effort.

As an aid to further understanding this matter, it is also necessary to understand the type of information that appears on individual support drawings. These drawings, which carry a prefix BRH and an entirely different number scheme, provide, in addition to the design details of the support, information as to where the support was to have been installed.

The plan type information is provided by a small square generally with four notations indicating building column 1ines. Within the square, theFe is usually a dimensional figure in feet and inches from one or more of the column lines. The support elevation information is furnished on the main face of the drawing where the elevation of the pipe and the building structure, as appropriate, are shown. The use of this system requires either a substantial degree of familiarity with the various buildings and their column.line grids, or ready reference to a set of the architectural layout drawings which clearly show the column line grids.

It can no longer be established just exactly what information was given to Mr. Culton for his use in generating the drawings. An interview with the one remaining person still in the original group when Mr. Culton worked there, indicated that the package contained a piping isometric and the individual support drawings along with any of their outstanding change documentation (CMC) that changed the~1ocational information. The pipe isometric was reproduced such that information relative to pipe instal-lation was deleted. This would include deletion of weld joint data and the bill of materials. The remaining information was the isometric line detail, location data (again in the form of column lines and elevation) and reference to connecting isometrics. The modified isometrics were then annotated with a symbol that was to depict the approximate location of supports and a support number was assigned to each symbol. The location of a given support appears to have been estimated from the support drawing using the building column lines, elevations, and the piping isometric dimensions that-still remained on the drawing after modification. The

5 early BRHL drawings did not give any dimensional information on the supports and the first issues were stamped"" issued for hanger identifica-tion and accountability only" in the drawing approval block. Subsequent revisions, apparently beginning in early 1981, were updated and began to show dimensions'for support. locations. The final versions of the drawings -

provide verified support locations, at which time the individual support drawings are revised to delete the location information. According to the present supervisor of the central document control center, the BRHL drawings have never been routinely distributed to any of the possible user organizations such as the support installations crews. The drawings were only available on an individual requisition basis which would be stamped "For Information Only" when given to the requisitioner. The BRHL drawings were originally developed and updated periodically to facilitate the final as-built stress analysis. The only use of the BRHL by other than the stress . analysis' groups presently occurs when a support has to be modified after the initial as-built verification effort. This arises by reason of the deletion of the support location information on the individual support drawing which then makes the use of the BRHL vital in order to find the support in the facility. This situation only arises on a limited basis and is treated on a case basis by the support instal-lation group and the document control center.

In regard to Mr. Culton's two major concerns in this area, the SRIC was able to locate a few of the early BRHL drawings which carry the initials "DKC" in either the draftsman identification block or in the checker block. A comparison of these drawings to those generated by other drafts-men indicate no significant differences. The SRIC can only conclude that Mr. Culton was as competent as the other people in the group. Given the non-use of the drawings at the time they were originally developed, this -

level of competency appears to have been adequate. Mr. Culton's statement that the drawings were released unapproved for use by the construction forces has been shown to be incorrect in two different ways. First, the original issues were provided to the document control center for filing with the note " issued for hanger identification and accountability only" on face of each drawing and were approved in the appropriate block on the drawing face. Secondly, the drawings, while on file in the document control center, were never tabject to a routine distribution and were not readily available to a1e construction force who in fact had no need for them. In addition, numerous observations by the SRIC of the support installation proce'ss has indicated that the support location information on the support drawing.was used to install and to inspect the supports and that any use of the BRHL for this purpose was so limited in frequency of occurrence that it was never detected. Mr. Culton's allegations regard-ing BRHL drawings is thus considered to be refuted.

6. Allegation Relative to the Splicinq of Electrical Cables At Tr. 5551 through 5552 and 5556 through 5557, Mr. Culton stated that he had observed that "Q" electrical cables had been spliced and that these splices were in the Unit 1 spread room. Following Mr. Culton's appearance

6 1

i before the Atomic Safety and Licensing Board, Mr. Culton was interviewed l in the NRC Region IV offices on November 8, 1982, in an attempt to obtain more information on the matter. The interview was tape recor'ded by a representative of the intervenor CASE in the proceedings. At Tr. 5552, Mr. Culton stated that he observed the splicing to have occurred two times .

and further that there were other instances for which he had some papers.

During the interview, Mr. Culton also indicated that he had other drawings available to him that would pin point the matter and promise to make them available to the NRC, or alternatively he would provide a sketch that would provide more detail. For the record, Mr. Culton has not yet made available to the'NRC any of the documents to which he has alluded.

Based on the information in the hearing record and in a transcript of the interview referred to above, the SRIC, initiated an investigation that attempted to determine what cables may have been' involved when Mr. Culton made his observation. The following key statements were utilized in attempting to isolate the involved cables from the other estimated 6,000 "Q" cables in the Unit 1 spread room:

a. At Tr. 5552 and 5556: Th6 cables in question are 800 or more feet long.

b. At Tr. 5556: Two cables.were observed to have been spliced.

c. At Tr. 5557: The cables were going to a relay panel.

d. Interview Record, Page 3: The relay panel was the third one in froe the aisle.

Using the above statements, the SRIC was able to narrow the number of possibilities down to two cables, presumably the same two.as observed by i Mr. Culton. The basis of the analysis was as follows:

a. The applicant has a computerized listing of all cables for the entire facility. By arrangement with tne computer operators, the SRIC was able to obtain a selected sort of the cables based on the "Q" identification and those in excess of 800 feet.

b. The list was reviewed by the SRIC to eliminate'those cables that were not routed to equipment in either the cable spread room or the control-room. A total of 42 cables were then involved.

c. Of the 42 cables, only 5 were shown by the routing records to be terminated in a relay panel, more correctly called relay racks.  ;

d. Of the seven relay racks, only one is the third one from an aisle and also has "Q" cables terminated in it, this being a cabinet identified as the " BOP Auxilary Relay Rack 1" with Tag Number CP1-ECPRCR-03. Of the five cables terminated in relay panels, only two are terminated in this panel.

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e. The cable pulling rec'ords indicate that the two cables were orginally identified as E0009231 and E0009240 which were pulled on January 14 and 15, 1980, respectively. Based on his employment records, these dates coincide with Mr. Culton's employment in the construction labor force as an electrician. .

f. Engineering changes subsequent to the pulls changed the designat. ion of cable E0009231 to A0009231 and E0009240 to SP009240. The change from "E" to "A" signifies that a previously identified safety function had been downgraded to nonsafety with,the cable still routed with safety grade cables. The change from "E" to "SP" indicates that the electrical circuit involved has been deemed to be no longer required and the cable has become a spare.

Specific findings relative to cables A0009231 and SP009240 are as follows:

a. The SRIC, with the. assistance of two other NRC inspectors, traced cable A0009231 through the spread room cable tray system from the point at which the cable entered the room until it left the tray to pass through a conduit into the relay rack. The only portion of the cable not examined was the approximately 17' of cable in the conduit.

Of the estimated 50-60' of cable in the tray, there were no anomalies identified.

b. The SRIC found that cable SP009240 had been removed from the tray system on or about November 23, 1982, in response to NCR E-82-01210 which stated that certain tray sections were overfilled. The  : ,

engineering solution was to remove several cables that been spared by other design changes. The removal was through the tray system but left the cable in the conduit entrance to the relay rack, again approximately 17'. The SRIC located the removed portion of the cable in a storage yard and visually examined the entire 400' with no anomalies identified.

General findings and considerations:

a. Project Specification ES-100 " Electrical Erection" does not totally prohibit the splicing of safety-related cab 1'se as indicated by Mr. Culton. The specification allows splicing to be done based upon the engineer's direction and this has been done by the use of engineered junction boxes. It should be noted the industry standards (IEEE) do not prohibit field run splices provided they are properly qualified.

, b. There have been a number of instances where the cable jackets have been repaired when the jackets were damaged either in the process of manufacture or during installation. These repairs have been accom-

) plished under a standard repair procedure, EEI-13, when directed by i

the site engineers. One of two repair measures are applicable within the procedure. One of the methods utilizes heat shrinkable plastic tubing when the damaged area is not prohibitively far from the end of 3

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the cable and,thus allows the tubing to slip down'the cable. The other procedure which was much more generally used early in 1980 involved the use of a fire resistant tape wrapped in half-laps over the damaged area. . The former procedure produces a very neat slim appearance while the latter procedure is relatively bulkly and might -

well appear to be a splice. A number of both types of those repairs were identified during.'the examination of the specific cables discussed above and during an earlier more extensive examination of several tray runs in the spread room. All of these anomalies were judged by the NRC inspector to be jacket repairs.

c. The SRIC believes that yet another consideration may well be relevant to this matter. The consideration involves a much earlier allegation that cables had been repaired in an unauthorized manner. The allega-tion was received by the SRIC sometime during February 1980.from an electrician assigned to the electrical cable pulling crew that had pulled the cable then in question and the two cables identified with Mr. Culton's allegation. All three cables were pulled during early to mid-January 1980. The SRIC's recollection.of the person was that he was a journeyman electrician and assisted the foreman in the detail '

supervision of the crew of about 16 men,-all of whom were classified as helpers except for the foreman and the journeyman. Since the electrician was sufficiently concerned to report a cable jacket repair involving the use of Scotch 33 tape rather than the approved tape, it seems to follow that he would have also reported an actual cable splice for which there is no approved repair. Given the electrician's,positi,on with the pulling crew, it also seems unlikely that he would not have been aware of an. error of a magnitude that would have caused such splices to be made. (For more information about the 1980 allegation and the results of the subsequent investigation, see NRC Inspection Report 50-445/80-08; 50-446/80-08.)

d. Since 17 feet of each of the identified cables were not inspected by I the NRC during the course of this special inspection, it was not l possible to conclude positively that the allegation is either l confirmed or refuted. Notwitnstanding, the inability to positively i state that the allegation made by Mr. Culton is substantiated or l refuted, the SRIC believes that no further action is warranted based on the following cumulative information as follows:

(1) Cable jacket repairs utilizing wrapping with a rubber tape were not and are not unusual.

(2) There has been no identified reason why the splices should have been necessary. The rubber-like jackets on the cable are relati.vely easy to cut with even a dull edge but the wire insulation material and the wire itself are relatively hard to cut.

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l (3) ~A jacket repair made with tape would 'be very difficuit for the inexperienced person to. distinguish from an actual splice. The splice generally would have a somewhat bulkier shape and would probably be somewhat lumpy rather than smooth.

(4)^ The probability that the SRIC would have learned of such an unusual event as a splice being made to safety-related cable

'through contacts that had been established in the electrical crew involved. -

(5) Removal of the cable would probably cause damage to the nearly

.30 cables in each of the conduits.

(6) Neither of the two cables now have a safety-related function and there are no requirements that prohibit the splicing of nonsafety cables.

! 7. Allegations by Arvil Dillingham, Jr.

An article appearing on page 13A of the Fort Worth Star-Telegram dated January 7,1983, stated that Arvil Dillingham, Jr. had made allegations which were subsequently investigated first by personnel of B&R and later by personnel of TUGCO. The article stated that Mr. Dillingham was then*

charging that these investigations were a " cover-up" cto hide safety hazards at the Comanche Peak nuclear power plant. The article stated that Mr. Dillingham had been employed a.t the construction site as a foreman and was laid off weeks after he made the allegations. The article also  ?

attributes three technical type allegations directly to Mr. Dillingham.

j In summary, the technical allegations appearing in the article were: -

(a) Mr. Dillingham apparently stated when interviewed by the writer of the article that rejected aggregate was mixed with concrete that was subsequently poured to form the base for the nuclear reactor. The article stated that a Larry Witt was the B&R equipment operator who had apparent first hand knowledge of the matter. The article also stated that Mr. Witt could not be reached for comment.

(b) A second allegation, that the article stated was never previously investigated, involved the construction of underwater 1 amps for the pools surrounding the reactor. Mr. Dillingham charged that he was prevented from cleaning out drill shavings from the lampposts and that these shaving could be washed into the reactor during refueling and could jam the fuel cells and could even fuse to the control rods.

(c) The third allegation dealt with a contentioi. that holes had been improperly . drill through concrete walls and the interior reinforcing-steel. The article attributes the information to another party identified as Danny Grisso.

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.The SRIC assigned to the Comanche Peak station obtained both the B&R and TUGC0 files pertaining to the investigations that were stated by the newpaper article to have occurred. The B&R file was found to contain an undated and unsigned letter addressed to Mr. Thomas Feehan, President of

.B&R. The letter is indicated in two different places to have been prepared by Arvil Dillingham, Jr. The letter is stated in a memorandum addressed to a group vice president of B&R from a vice president of the B&R division to have been hand delivered to Mr. Feehan by Mr. Dillingham on August 6, 1982. The memorandum was dated August 13, 1982. The undated letter to Mr. Feehan contained eight, violations that the writer stated he had observed or had knowledge of that had occurred during his period of employment at the Comanche Peak station. Review of the letter addressed to Mr. Feehan indicated that only one of the eight violations correlated with the allegations appearing in the newspaper article, this being the item outlined in (c) above pertaining to the drilling of holes in the concrete walls. The B&R memorandum of August 13, 1982, which is a report of the internal B&R investigation of the eight violations, indicates that seven of the allegations were found to be either without a basis or were not substantiated. The remaining item was considered in effect to have been substantiated but the' corrective measures were already taken. In_each case, by what is assumed to be Mr. Dillingham's signature, Mr. Dillingham acknowledged his satisfaction with the B&R findings. The above memorandum indicates that a number of other pecple were interviewed by the B&R investigative group, one of whom was Mr. Witt. Mr. Witt apparently did not confirm Mr. Dillingham's allegations but made additional allegations related to his experiences during his past employment at CPSES. One of these allegations appears to be substantially-the same as that appearing in the summarization of the news article as (a). Mr. Witt also charged that some personnel biased the operation of the-concrete batch plant scales by leaning on the wires connecting'the scales to the sensors. Additionally, Mr. Witt stated concerns about a possibly missed hold point during the welding of the fuel pool liner and that some welding had been done by ac_ uncertified welder. In an internal

B&R memorandum dated August 17, 1982, the B&R investigators summarized 1

Mr. Witt's concerns and their findings relative to the concerns. The B&R memorandum indicates that the investigation relative to use of rejected aggregate was apparently partially substantiated but of no concern in that

the aggregated pile, rather than actually being unacceptable, simply had not been tested prior to use as required. The matter was documented on Deficiency and Disposition Report C-446 dated December 9, 1976, which appears as attachment A to the memorandum. In the matter of the missed

! hold point for the fuel pool liner weld, attachment B to the memorandum documents that no hold point was missed. Regarding the two remaining Witt I

allbgations, the memorandum states that the allegations were investigated j and found to be without basis but provides no other information.

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Personnel of TUGC0 performed a separate investigation of Mr. Dillingham's

( allegations (the Feehan letter) during August of 1982. The results of

l. that investigation were furnished to TUGC0 management by memorandum dated

! September 2, 1982. This investigation found that two of eight items were

substantiated with one of these being the same item that was substantiated l

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11 by the B&R investigation. Both of the substantiated allegations were found by the investigators to have been adequately documented and that corrective measures had been taken or were in progress.

In a separate memorandum dated December 10, 1982, one of the TUGC0 investigators documented a phone call from Mr. Dillingham in which Mr. Dillingham apparently made yet additional allegations. One of these allegations regarded welding done by an uncertified welder on the turbine-generator pedestal (by implication). Mr. Dillingham also apparently further mentioned Mr. Witt who was supposed to know about a sensor that had bro, ken of f and was buried 'in the main dam. Also, Mr. Witt was alleged to have personally driven a front loader that returned dry and lumpy cement that had been rejected to the bin, that this cement had been subsequently used in the reactor core, and that this was why the cracks happened. The writer of the memorandum stated that he had encourged Mr. Dillingham to take his concerns to the NRC. Mr. Dillingham in turn was reported as saying that he had intended on going to the newspapers and Congress instead.

It appears that Mr. Dillingham carried out his above stated intention in that the above referenced newspaper article has appeared and to the best of SRIC's knowledge, Mr. Dillingham has made no contact with any component of the NRC. NRC Region IV determined that the allegations in the news article should be investigated but that those made in the.Feehan letter and in the telephone conversation with TUGC0 should not. This decision was based on the premise that Mr. Dillingham has had his earlier concerns satisfied except for those appearing in the article. '

Regarding the above summarized allegation (a), the SRIC established that Mr. Witt was no longer an employee at CPSES and further established that he had relocated from the Glen Rose, Texas, area to another state. NRC Region IV personnel made several attempts t'o contact Mr. Witt by telephone at his new address, to no avail. A registered letter, receipt requested, was then sent to Mr. Witt requesting that he contact Region IV as soon as l possible. Receipt of the letter was acknowledged but as of this date, i

Mr. Witt has not contacted the region. It appears that Mr. Witt does not intend to assist the NRC in investigating allegations attributed to him.

It should be noted that only the B&R investigative group has been able to establish contact with Mr. Witt; all others have apparently failed.

Regarding summarized allegation (c), the SRIC, with assistance of another

! Region IV inspector, was able to establish that the underwater lighting standards were fabricated in such a manner as to leave drilling chips insida and had not been removed. It was also established that the lighting standards were fabricated completely outside the licensee's QA program which included various welding operations. There are no records of inspection or of the welders involved or of the weld procedures utilized. Review of the design drawings do not reflect that the A/E considered the lighting standards to be within the QA scope, yet l

should the standards physically fail during the seismic event, fuel could be damaged. Given the possibility of failure, the standards should have been classified as Seismic Category II (licensee's FSAR definition for A

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e 12 components which have.no safety function but must not fail in a seismic event since such failure could jeopardize the functioning of a safety-related component) and should have been included 'in the QA program.

This is considered ~to be a violation of Appendix 8 to 10 CFR 50.

Regarding the premise that the drilling. chips inside the standards could be swept into the reactor during refueling and cause an accident, the SRIC found six of the standards are normally located with their bottoms just about the floor level of the refueling pool,and that the chips that might have worked their way out of the bottom of the standard could have carried into the reactor at the conclusion of.the refueling process. The size.of the chips that could work their way out through the 1/2" holes are not of a size .that could be expected to plug a water channel through the reactor core and create a hot spot. Further, the idea that the chip could fuse to the control rods is equally remote in that far higher temperature would be required in the core to achieve such fusion than actually will exist there, the differential being 600' to 800'F. Thus,' the safety signific'ance of the chips is very small. The uncontrolled (no QA) problem wit,h the -

standards is relatively more important since workmanship on the devices has not been established. Regarding sumamrized allegation (c), the allegation has been the subject of another allegation by a person who appears to have substantially more direct knowledge of the matter than indicated by Mr. Dillingham. Under these circumstances, the NRC has determined that 'it can address the issue in a more satisfactory manner by investigating and evaluating the second party's allegation rather than Mr. Dillingham's.

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8. Posting of NRC Form 3 ,-

10 CFR 50 was revised by 47 FR 30452 to add 10 CFR 50.7 " Employee Protec-i tion." The change was published July 14, 1982, and had an effective date

of October 12, 1982. An important element of the change was that of a requirement to post NRC Form 3 at locations where the form can be readily l viewed by employees on their way to or from their place of work. It has l been alleged that the licensee did not post the form. The SRIC learned of the allegation during early January 1983 and found.that the form was posted throughout the main construction administration building and on a bulletin board where most of craft labor force can readily see it, particularily when departing from the construction area. The SRIC has been informed by licensee employed personnel that they received and posted the forms 1.n the administration building about the first of 1983. A senior B&R manager indicated that the forms were received, he believed from B&R's Houston l office, sometime between Thanksgiving and Christmas and were posted on the l craft labor bulletin board near the " brass alley" well before the first of the year. It is thus clear that the forms were not posted on the specified effective date of the change to 10 CFR 50 as alleged. It is much less clear as to when the forms were actually posted nor is it clear that most people would even have been aware of the posting. The " brass alley" bulletin board is a large board, perhaps 4' by 6' in size with many postings. The majority of the postings are required under various l federal' statutes or regulations. The posting of an additional form probably j would not draw much attention from the average worker. As of the time of I

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13 inspection by the SRIC, the licensee and his principal site contractor were found to have the form posted and to be in compliance with the regulation.

9. Management Interviews .

The SRIC held management interviews with one or more of the persons identified in paragraph 1 on a nearly daily basis throughout the inspec-tion period to discuss NRC findings developed during various special inspections and investigations. The discussions also included the licensee's positions on the NRC findings.

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