ML20198K581
| ML20198K581 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 04/20/1982 |
| From: | Miller T TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC) |
| To: | |
| Shared Package | |
| ML20197J316 | List:
|
| References | |
| FOIA-85-59 NUDOCS 8606040056 | |
| Download: ML20198K581 (51) | |
Text
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C A b g o q l 3 A P '7 't k b o CPSES 4
/ Q PREREQUISITE TEST INSTRUCTION
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NUMBER OF PAGES TYPE OF 00CUMENT 1.
^- 4 Manufacturing Record Sheet (MRS) 2.
f Weld Data Card (WDC) Weld No(s). Flu '//
t 3.
f Weld Filler Material Log (WFML) 4.
e-A Material Identification Log (MIL) i I
5.
Non-Destructive Examination Report (NDER) 6.
Inspection Report (IR) 7.
Nonconformance Report (NCR)
F 8.
Vendor Documentation Jf I RMATIOly _~
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9.
Repair Process Sheet (RPS) 'le 5
l 10.
Operation Traveler (OT) 11.
Drawing (IncludingCMC)
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12.
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Material Requisition (MR) -
PPR/
13.
f Miscellaneous (Describe Below)
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The contents of this package as listed above have been reviewed per the requirements of CP-QAP-18.2 and are acceptable.
QC Superintendent
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Total Number of ages in Package ANI
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PLANT SYSTEM COMPONENT T AG/ SPIN /tO EN T NO I OR AWING / SPECIFIC A TION NO.
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3 Ccd%og G Mh 'A 31 Ts0-82495 TEXAS UTILITIES GENERATING COMPANY OFFICE MEMORANDUM Distribution June 23, 1982
., To Gen Rose. Texas Interim Change to XCP-ME-4 g
t System Cleanliness Verification FOR INFO ONLy' This interim change supersedes the changes issued under TNO-82319.
m l
The acceptance criteria for cleanliness level 1 and 2 effluent flush water requires i
the pH to be between 6.0 and 8.0.
The flushing medium pH as described in Table II, note 1, can be relaxed to between 5.8 and 8.0.
i Upon review of Westinghouse specification.292722 revision 9, dated April 27, 1979 I
and letter TBX-M-756; the pH requirements for cleanliness level 1 and 2 acceptance I
criteria and flushing medium criteria (Table II) for water grades A and B can be re-laxed to 5.5 to 8.0 due to CO absorption, subject to the following condition:
2 1.
The water in the storage tank from which the flush water is drawn must have a maximum conductivity of 1.0 micromhos/cm.
A note to this.effect will be added to the acceptance criteria of the procedure for cleanliness levels 1 and 2 at the next procedure revision. A copy of the revised Table II of XCP-ME-4 is attached.
Z
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J. C. K kendall JCK/ GDS /TPM/jb Distribution:
I J. C. Kuykendall, IL, 1A M. G. Eidson, IL, lA C. M. ?uffer, IL, 1A R. E. Camp, IL, IA S. M. Franks, IL. 1A A. D. Quam, IL, lA R. L. Moller, IL, IA E. L. Gastinel, IL, lA G. F. Riggio, IL, lA R. G. Tolson, IL, IA M. S. Harris, IL, IA D. F. Rohrer, IL, IA R.
Franks, IL, 1A K. E. Hemmila, IL, 1A R. B. Russom, IL, lA D. E. Deviney, IL, 1A J. G. Hennessy, IL, IA J. C. Sanders, IL, 1A W. E. Stone, IL, IA T. E. Hodge, IL, 1A G. C. Sandlin, IL, IA R. G. Taylor, IL, 1A C. R. Horne, IL, 1A S. L. Siebenaler, IL, 1A C. E. Scott, IL, IA T. L..Hutson,,1L, 1A G. D. Smith, IL, 1A M. J. Riggs, IL, lA A. S. Jamar, IL, 1A M. G. Smith, IL, 1A T. P. Miller, IL, 1A R. G. Johansen, IL, 1A F. R. Stough, IL, IA G. L. Kunkle, IL, IA B. W. Kaulfus, IL, IA I. M. Thomson, IL, 1A M. J. Michalka, IL, IA J. G. Kilpatrick, IL, IA J. A. VanGulik, IL, 1A Shif t Supervisor's Office, H. A. Lancaster, IL, IA
.W. J. Weis, IL, 1A (Control Room), IL, IA A. C. Lilly, IL, IA J. E. Wooten, IL, 1A D. B. Allen, IL, 1A D. A. London, IL, 2A J. C. Zimmerman, IL, 1A C. D. Beach, IL, 1A D. L. McKibbin, IL, 1A ARMS S. J. Berra, IL, IA D. L. Marston, IL, IA P. N. Boozer, IL, IA J. L. Martin, IL, IA D. R. Bowles, IL, 1A D. E. Mayer, IL, 1A
(
D. E. Bradley, IL, 1A J. K. Newton, IL, IA J.
Cardoza, IL, IA P. E. Olson, IL, 1A H. J. Cheatlcam, IL, 1A M. D. Phalen, IL, IA M. S. Clark, IL, 1A L. A. Porter, IL, 1A J. H. Collins, IL, 1A J. C. Prevo, IL, 1A
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TNO-82682 /
TEXAS UTILITIES GENERATING COMPANY OFFICE MEMORANDUM C
!I l
TO Distribution Glen Rose, Texas August 17, 1982 i
l i
SUMECT Interim Change to XCP-ME-4 Revision 5 7
i Change paragraph 4.1.7 to read as follows:
EOR INFO
.ONLY
.l 4.1.7 Drawing - applicable drawing marked to show flush flow' paths ano
.j boundaries.
1 Note "For Information Only" drawings
-I may be used.
1 4.1.7.1 Verify that all DCA's outstanding against the drawings are listed on the drawings. DCA's shall be used during testing
(
,j as follows:
A.
The DCA's that affect the component system to be tested may be drawn-in on the af fected drawing using RED for addi-tions and GREEN for deletions and highlighted during
(
testing, or Note The design change number (DCA, CMC, etc.) shall be indicated beside changes that are drawn j
in on drawing.
B.
The DCA's may be highlighted during testing and attached to the drawing.
Change made to address concern identified in DSR-82-004.
t
//
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C.
uykendall JCK/CDS/HAL/jb Distribution:
(See attached page) l FI og_RO
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TNo-82682 Page 2 f
EOR INFO ONLY f
Distribution:
J. C. Kuykendall, IL,1 A S. M. Franks, IL, IA G. F. Riggio, IL, 1A R. E. Camp, IL, IA E. L. Castinel, IL, IA D. F. Rohrer, IL, IA I
R. L. Moller, IL, 1A M. S. Harris, IL, IA R. B. Russom, IL, IA
'l R. G. Tolson, IL, IA K. E. Hemmila, IL, 1A J. C. Sanders, IL, 1A R.
Franks, IL, 1A J. G. Hennessy, IL, 1A G. C. Sandlin, IL, IA D. E. Deviney, IL, 1A T. E. Hodge, IL, IA S. L. Siebenaler, IL, 1A W. E. Stone, IL, IA C. R. Horne, IL, 1A G. D. Smith, IL, 1A f
R. G. Taylor, IL,1A T. L. Hutson, IL, 1A M. G. Smith, IL, 1A C. E. Scott, IL, IA A. S. Jamar, IL, 1A F. R. Stough, IL, IA jl M. J. Riggs, IL,1A R. G. Johansen, IL, IA
- 1. M. Thomson, IL, 1A T. P. Miller, IL, IA B. W. Kaulfus, IL, 1A J. A. Van Gulik, IL, 1A j
C. H. Welch, IL, IA J. G. Kilpatrick, IL, IA D. J. Weis, IL, IA
'j M. J. Michalka, IL, 1A H. A. Lancaster, IL, 1A J. E. Wooten, IL, 1A
. stf t Supervisor's Of fice, A. C. Lilly, IL, IA J. C. Zimmerman, IL, 1A (Control Room), IL, I A D. A. London, IL, 2A ARMS D. B. Allen,11, IA D. L. McKibbin, IL, 1A C. D. Beach,' IL, IA D. L. Marston, IL, 1A S. J. Barra, IL, IA D. E. Mayer. IL, 1A P. N. Boozer, IL, IA J. K. Newton, IL, IA D. E. Bradley, IL, IA P. E. Olson, IL, IA J.
Cardoza, IL, 1A M. D. Phalen, IL, 1A l
(
%. J. Cheatheam, IL, 1A L. A. Porter, IL, 1A M. S. Clark, IL, IA J. C. Prevo, IL, 1A J. H. Collins, IL, 1A C. M. Puffer, IL, 1A M. G. Eidson, IL, 1A A. D. Quam, IL, 1A I have received and inserted the above material in Manual No.
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M-83-00154S Rev. 1 UNIT STRUCTURE / SYSTEM ITEM / COMPONENT TAG /lO NUMBER LOCATION OR ELEVATION RIR NO.
1&X 4901 Light Pole P P1, P2, P3' Fuel Building Assemble
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E f
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REv REPORTEo BY:
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'r MAR 2 81933 In Reply Refer To:
0 :kets:
50-445/83-C' 50-446/83-01 d
Texas Utilities Generating Company ATTN:
R. J. Gary, Executive Vice President & General Manager i
2001 Bryan Tower Dallas, Texas 75201 r
Gentlemen:
This refers to the inspection conducted under the Re I
~
of October 1982 through February 1983 of activities authorized by NRC Construction Permits CPPR-126 and CPPR-127 for the Comanche Peak facility Units I 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 discusse enclosed inspection report. selective examination of procedures and represen
{
with personnel, and observations b( the inspector.
a During this inspection, it was found that certain of your activities wereCon in violation of NRC requirements.
i 2.201 to this violation, in writing, in accordance with the provi
- I Your response should be based on the specifics contained in the Notice of j'
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 office, i
by telephone, within 10 days of the date of this letter, I,
j Such application must be consistent with the date of this letter.
requirements of 2.790(b)(1).
m
'i M
DRRP&ER/RIV SRI
,. h RPS-RPB1 D/ES H GMads'en EJohnson JGagl'iWdo 3/./8S]
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TWesterman 3/ '/83 3/ /53 3/ /83 3//(./83
~~~ F0lA-85-59#
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2 MAR 2 81983 Texas Jtilities Generattr4 Cem:any Shoule you have any ouestions concerning this inspection, we will be pleased 1
to discuss tnem with you.
Sincerely, "Ori-inal Signed be
- c. L. MAOSEN"
.G. L. Madsen, Chief Reactor Project Branch 1 i
Enclosures:
j-1.
Appendix A - Notice of Violation
.-:f 50-445/83-03 2.
Appendix B - NRC Inspection Report 50-446/83-01 1
cc w/encls:
Texas Utilities Generating Company ATTN:
H. C. Schmidt, Project Manager 2001 Bryan Tower Dallas, Texas 75201 i
bec to DMB (IE01)
.)
bec distrib. by RIV: Resident Inspector RPB2 4
Section Chief l
TPB MIS SYSTEM E. Johnson RIV File RIV Reading File Br. Reading File C. Wisner RA M. Rothschild, ELD TEXAS STATE DEPT. OF HEALTH Juanita Ellis David Preister q'
J 4
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APP.ENDIX A NOTICE OF VIOLATION 50-445/83-03 1,
Docket:
50-446/83-01 Texas Utilities Generating Company j
Comanc!e Peak Steam E1'ectric Station Permits: CPPR-126 CPPR-127 i
i i d of i
Based on the results of an NRC inspection conducted during the per o NRC Enforcement October 1982 through February 1983, and in accordance with the 1982, the Policy (10 CFR Part 2. Appendix C), 47 FR 9987, dated March 9, following violation was identified:
i d
Failure to Implement a Quality Assurance Procram for the Fabricat o lies _
Installation of Electrical Underwater Flood 11oht Pole Asse
.4 i
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licant shall Criterion II.of Appendix 8 to 10 CFR 50 requires tha the
]7 ide control over quality assurance program and that the program shall prov t ms, and activities affecting quality of the identified structures, sys e 7
l the NRC Regulatory Guide 1.29 which in' paragraphs 2 and 4 require components.
ts whose applicant to identify those structures, systems, and componen
) but whose continued function is not required (in a design bas ified in
]
]
other paragraphs to an unacceptable level.
i has Contrary to the above, the Senior Resident Inspe t
tion activities and review of design drawings that group of devices i
Poles" q
collectively identified as " Electrical Underwater Floodlight ngwe l t ry
'i ce (Drawing 2323-EL-0925-02) 1 Guide 1.29 and were not included within the ifcensee's Quality Assur Mechanical failure of the devices in a seismic event t
damage fuel during reactor core installation activities or in the spen Program.
l l failure fuel storage pools, although the possibility of such mechanica t due to of the pole assembly resulting in damaging fuel is very remo e the design of upper and lower pole retention devices.
Violation.
(Supplement II.D.)
4 This is a Severity Level V ting Company Pursuant to the provisions of 10 CFR 2.201, Texas h date of li including:
(1) the this Notice, a written statement or explanation in reply,d (2) corrective corrective steps which have been taken and the results achieve ;(3) the d steps which will be taken to avoid further violations; andConsideration may full compliance will be achieved.
your response' time for good cause shown.
Eg' f& O
')
MAR 2 g 1993 Dated:
ty m
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Y, M:es
- 3
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APPENDIX E U. S. NUCLEAR REGULATORY COMMISSION REGION IV Report: 50-445/23-03 50-446/53-01 Category: A2 Dockets: 50-445; 50-446 Texas Utilities Generating Company-(TUGCO) i l
Licensee:
2001 Bryan Tower Dallas, Texas 75201 Comanche Peak, Units 1 and 2 Facility Name:
Comanche Peak Steam Electric Station (CPSES), Glen Rose, Texas Inspection At:
October 1982 through February 1983, M.
Inspection Conducted:
Date In'spector:
/ R. G. Taylor, Senior" Resident Inspector-s l
Construction
/6h2 s'
Date erw Approved:
'T. FY W(sterman, Chief '
Reactor Project Section A Insoection Summary 50-445/83-03; Insoection Conducted October 1982 Through February 1983 (Recort 50-446/83-011 Routine and special inspection, announced by the Senior Resicent Inspector-Construction (SRIC) including f acility tours, investiga l'
Areas Insoectedi t
of allegations, participation and assistance to the Construction Assessmen The inspection Team Inspection, and other inspection related activities.
involved 263 inspector-hours by one NRC inspector.
t
,l Within the areas inspected, one violation was identified (failure to implement a QA program for fabrication and installation of underwat ii Results:
l; Ij poles.)
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Details i
t Persons Contacted 1
1.
Principal Licens_ee Personnel l
R. G. Tolson, Site Quality Assurance Supervisor
- 0. N. Chapman, Quality Assurance Manager j
- 8. R. Clements, Vice-President, Nuclear J. T. Merritt, Manager of StartupJ. B. George, Vice President an
(
3 i
Other Personnel
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G. R. Purdy., Project Quality Assurance Manager, Brown & Root (
"{
1 D. Frankum, Construction Project Manager, B&R l during i
The SRIC also interviewed other licensee and contractor personne the inspection period.
},
Licensee Action on Previous Inspection Findings _
Building 2.
Quality of Unit 1 Reactor (Closed) Unresolved Item (50-445/79-24)This item rel t
building dome
'l that a small amount of concrete placed on the Unit 1 reac or Dome Concrete.
f i
trol was during a rain storm without appropriate controls by qua table. The testing program had indicated that the material wa adequate.
e additional j
assurance was judged to be required. structural acceptanc i l attention directed to the repair area.anomolies were identified it is judged l
that the concrete is of adequate quality.
Des gn of the AC
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i (Closed) Unresolved Item (50-445/80-20; 50-446/80-20) h Instrument Distribution Panels., This item involved a finding t t in segregation of safety and nonsafety wiring in the panels was no i l compliance
,1 accordance with Regulatory Guide 1.75 but was in essent a
- 15. After discus-with.the panel design displayed by FSAR Figure 8 l
trical developed.
I engineering group, a method of correcting the matter was d of l
FSAR Figure 8.3-15 was revised by Amendment 2 j
of the four panels involved and had no further questions.
I correction.
Control of Stainless (Closed) Unresolved Item (50-445/81-14; 50-446/81-14)T l
f repairs controlled program for the control of the nurnber and extent'o l
Weld Repairs.
- J n-
_-n
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31 made to weld joints in stainless steel pipe had become less well con-
=i The licensee revised Construction ll trolled due to personnel changes. Procedure CP-CPM-6.9D to reflect prop The engineering controls were and QC actions effective in January 1982.
promulgated effective with the issuance of Procedure CP being complied with and therefore, has no further questions, 1982.
j.
f Action on Licensee Identified Design / Construction Deficiencies if 3.
On September 10, 1982, (Closed) Over-Torqueing of Safety Relief Valves.the licensee i 1
It was reported under the purview of 10 CFR 50.55(e) had been identified i
- j It was torquaing to stop leaks during the main steam hydrostati
.4 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 y
informed the NRC that after review, the matter was not considered formally The SRIC has reviewed the documentation reportable under the regulation.
The examination did not
,}
of the examination of the valves by the licensee.
l reveal any significant damage had occurred to any of the valves that would it have prevented the valves from lifting under pressure which would satisfy Some of the valves may have leaked under operating l
The SRIC the safety function.
conditions which would be undesireable but not a safety hazard.
>j had no further questions on this matter.
H Allegations By Dennis K. Culton 4:
Mr. Dennis K. Culton made a limited public On September 16, 1982, matter of TUGCO's application for an operating licens His
,j' 5551 statement during the appearance appears in the hearing tran lf i
through 5555.
Based upon a statement which appears in the record at 5556 through 5559.
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 The first area dealt with the potential misuse
>I safety of construction.
of a group of drawings referred to as BRHL while the second dealt with th The SRIC was alleged splicing of safety-related or "Q" electrical cables.
assigned to make the evaluation.
Allegation Relative to BRHL's 5.
Mr. Culton's concern in this area appears at Tr. 5552 through 5554 and His concerns can be summarized as follows:
i:
5557 through 5558.
Based upon limited information, he was directed to generate isometric
'He states at 5557 that he did not (a) drawings giving support locations.
feel qualified to do this work in the manner directed.
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g 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 (b) locate and' instal) 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 The drawing locational information at an appropriate point in time.
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 The as-build verification program as required by NRC IE Bulletin 79-14.
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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
_3 understood.that he and others were hired and/or recruited from th
~
l 1 abor forces specifica11y for the effort.
As an aid to further understanding this matter, it is also necess'ary to t
understand the type of information that appears on individual support These drawings, which carry a prefix BRH and an entirely f
different number scheme, provide, in addition to the design details.of the drawings.
?
The plan type information is provided by a ses11 squ I
Within the square, there four notations indicating building column lines.
is usually a dimensional figure in feet and inches from one or more of The support elevation information is furnished on the the column lines.
main face of the drawing where the elevation of the pipe and the building The use of this system requires structure, as appropriate, are shown.
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 Ma column line grids.
It can no longer be establiNo %M exactly what information was given to An interview with the Mr. Culton for his use in gn< lg, the drawings.
one remaining person still in the Eginal group when Mr. Culton worked there, indicated that the package contained a piping isometric and the i
individual support drawings along with any of their outstanding change fl The pipe documentation (CMC) that changed the locational information.
'4 isometric was reproduced such that information relative to pipe instal-j This would include deletion of weld joint data and 1
lation was deleted.
The remaining information was the isometric line i
the bill of materials. detail, location data (again in the form of column lines an The modified, isometrics were then and* reference to connecting isometrics.
i annotated with a symbol that was to depict the approximate location of The location supports and a support number was assigned to each symbol.
of a given support appears to have been estimated from the support drawin using the building column lines, elevations, and the piping isometric The dimensions that still remained on the drawing after modification.
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5 early BRHL drawings did not give any dimensional in Subsequent-tion and account' ability only" in the drawing approval show dimensions for support locations.
provide verified support locations, at which time the individual support According to drawings are revised to delete the location information.
O L
the present supervisor of the central document control center, the BRH j
drawings have never been routinely distributed to any of the possible j
The drawings user organizations such as the support installations crews.
U ld be were only available on an individual requisition basis which wou J
The BRHL stamped "For Information Only" when given to the requ
.I d
The only use of the BRHL by other j
the final as-built stress analysis.
than the stress analysis groups presently occurs when a support has to D
This arises ze be modified after the initial as-built verification effort.
by reason of the deletion of the support location infor 2r in order to find the support in the facility.
l:
on a limited basis and is treated on a case basis by the support instal-ir 1ation 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 t block.
The SRIC can only conclude that men indicate no significant differences.
Given the Mr. Culton was as competent as the other people in the g j
i Mr. Culton's statement 1evel of competency appears to have been adequate.
l that the drawings were released unapproved for use by the construction l
First, the forces has been shown to be incorrect in two different ways.
original issues were provided to the document control center for filing j
l" with the note " issued for hanger identification and accountability on y on face of each drawing and were approved in the apprcpri
- 1 the drawing face.
control center,-were never subject to a routine distribution and were not readily available to the construction force who in f 1
installation process has indicated that the support location information for them.
on the support drawing was used to install and to inspect the supports and that any use of the BRHL for this purpose was s Mr. Culton's allegations regar -
ll d
1; ing BRHL drawings is thus considered to be refuted.
If Allegation Relative to the Splicinq of Electrical Cables 6.
At Tr.' 5551 through 5552.and 5556 through 5557, Mr. Culton stated that had observed that "Q" electrical cables had been spliced and that these f
splices were in the Unit 1 spread room.
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before the Atomic Safety and Licensing Board, Mr. C i
The interview was tape recorded by a j
At Tr. 5552, more information on the matter.
representative of the intervenor CASE in the proceedings.
i Mr. Culton stated that he observed the splicing to have occurred two t me and further that there were other instances for which he had some papers.
During the interview, Mr. Culton also indicated that he had other draw h
available to him that would pin-point the matter and promise to make t em available to the NRC, or alternatively he would provide a sketch t j
would provide more detail.available to the NRC any of the documents t Based on the information in the hearing record and in a transcript of th interview referred to above, the SRIC initiated an investigation that Culton attempted to determine what cables may have been involved 4
6,000 attempting to isolate the involved cables from the other estimated{
made his observation.
9 "Q" cables in the Unit 1 spread room:
The cables in question are 800 or more feet At Tr. 5552 and 5556:
a.
I long.
Two cables were observed to have been spliced.
b.
At Tr. 5556:
j The cables were going to a relay panel.
At Tr. 5557:
c.
j The relay panel was the third one in from Interview Record, Page 3:
d.
the aisle.
Using the above statements, the SRIC was able to narrow the num d by possibilities down to two cables, presumably the same two as observe
]
The basis of the analysis was as follows:
4 Mr. Culton.
The applicant has a computerized listing of all cables for the ent By arrangement with tne computer operators, the SRIC was a.
able to obtain a selected sort of the cables based on th facility.
]
identification and those in excess of 800 feet.
The list was reviewed by the SRIC to eliminate those cables that we U;
not routed to equipment in either the cable spread room or the b.
[l A total of 42 cables were then involved.
p control room.
Of the 42 cables, only 5 were shown by the routing records to be k
terminated in a relay panel, more correctly called relay rac s.
c.
d Of the seven relay racks, only one is the third one from an aisle a ified also has "Q" cables terminated in it, this being a cabinet ident d.
R 03. 'Of as the " BOP Auxilary Relay Rack 1" with Tag i
ted i
in this panel.
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7 The cable pulling records indicate that the two cables were orginally which were pulled on January la P
j identified as.E0009231 and E0009240 Based on his employment records, these e.
j and 15, 1980, respectively.
dates coincide with Mr. Culton's employment in the construction
'l labor force as an electrician, j
Engineering changes subsequent to the pulls changed the designatio The change of cable E0009231 to A0009231 and E0009240 to SP009240.
f.
from "E" to "A" signifies that a previously identified safety
.l function had been downgraded to nonsafety with the cable s to "SP" indicates that l
with safety grade cables.
the electrical circuit involved has been deemed to be no longer required and the cable has become a spare.
are as follows:
l*
- Specific findings relative to cables A0009231 and SP009240 The SRI.C, with the assistance of two other NRC insp i
I a.
point at which the cable entered the room until it left the tray toThe G
cable A0009231 j
pass through a conduit into the relay rack.
_ l cable not examined was the approximately 17' 1
identified.
had been removed from the tray The SRIC found that cable SP009240 in response to NCR E-82-01210 b.
23, 1982, The system on or about November which stated that certain tray sections were overfilled.
engineering solution was to remove several cables that been spa
^
The removal was through the tray system but 1 eft the cable in the conduit entrance to the relay rack, a other design changes.
7'l approximately 17'.in a storage yard and visually examined the entire 4
'+
anomalies identified.
General findings and considerations:
<l Project Specification ES-100 " Electrical Erection" does not tota prohibit the splicing of safety-related cables as indica a.
f the engineer's direction and this has been done by the use oIt sh Mr. Culton.
engineered junction boxes.(IEEE) do not prohibit field run splic 9
i,!
qualified.
There have been a number of instances whe f
b.
manufacture or during installation.
d by p11shed under a standard repair procedure, EEI-13, whe i
the site engineers.One of the methods utili::es heat shrinkable plastic d of tubing when the damaged area is not prohibitively far from the en the procedure.
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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~1atter procedure is relatively bulkly and might it well appear to be a splice.
A number of both types of these repairs were identified during the examination of the specific cables
,j-discussed above and during an earlier more extensive examination of ll several tray runs in the spread room.
All of these anomalies were judged by the NRC inspector to be jacket repairs.
i The SRIC believes that yet another consideration may well be relevant c.
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 z?
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
]f 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 il 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 position with the pulling crew, it also seems unlikely that he would not have been aware of an error of a magnitude that I
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; 5.0-446/80-08.)
1 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 possible to conclude positively that the allegation is either confirmed or refuted.
Notwithstanding, the inability to positively 4
state that the allegation made by Mr. Culton is substantiated or refuted, the SRIC believes th't no further action is warranted based a
on the following cumulative information as follows:
(1) Cable jacket repairs utilizing wrapping with a rubber tape were not and are not unusu,al.
(2) There has been no identified. reason why the splices should have been necessary.
The rubber-like jackets on the cable are relatively 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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(3) A jacket repair made with tape would be very difficult for the The inexperienced person to distinguish from an actual splice.
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 (4) 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 j
30 cables in each of the conduits.
l I
(6) Neither of the two cables now have a safety-related function and i
f there are no requirements that prohibit the splicing of f(
j nonsafety cables.
1 A11ecations by Arvil 01111naham. Jr.
7.
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 I,
charging that these investigations were a " cover-up" to hide safety J
hazards at the Comanche Peak nuclear power plant.
The article stated that Mr. Dillingham had been employed at the construction site as a foreman and j
was laid off weeks after he made the allegations.
The article also attributes three technical type allegations directly to Mr. Oillingham.
I In summary, the technical a'11egations 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 The subsequently poured to form the base for the nuclear reactor.
i article stated that a Larry Witt was the B&R equipment operator j
The article who had apparent first hand knowledge of the matter.
i also stated that Mr. Witt could not be reached for comment.
l (b) A second allegation, that th'e article stated was never previously investigated, involved the construction of underwater lamps 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.
The third allegation dealt with a contention that holes had been (c) 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 i
TUGC0 files pertaining to the investigations that were stated by the The B&R file was found to contain an newpaper article'to have occurred. undated and unsigned letter address
.i The letter is indicated in two different places to have been The letter is stated in a memorandum B&R.
l prepared by Arvil Dillingham, Jr. addressed to a group vice president of Feehan by Mr. Dillingham B&R division to have been hand delivered to Mr.
The memorandum was dated August 13, 1982.
The lj' on August 6, 1982. undated letter to Mr. Feehan contained eight violations that th I
stated he had observed or had knowledge of ?. hat had occurred during his Review of the letter j
period of employment at the Comanche Peak station. addressed correlated with the allegations appearing in the newspaper article, this i
being the item outlined in (c) above pertaining to the drilling of holes
!~
in the concrete walls.
The 8&R memorandum of August 13, 1982, which is a report of the internal 8&R investigation of the eight violations.
5-indicates that seven of the allegations were found to be either without a
~
- I The remaining item was considered in basis or were not substantiated.
effect to have been substantiated but the corrective measures were already l
In each case, by what is assumed to be Mr. Dillingham's signature, l
The Mr. Dillingham acknowledged his satisfaction with the B&R findings.
taken.
i d
I,
.above memorandum indicates that a number of other people were interv ewe Mr. Witt J
by the S&R investigative group, one of whom was Mr. Witt.
apparently did not confirm Mr. Dillingham's allegations but made additional allegations related to his experiences during his pastOne employment at CPSES.
the same as that appearing in the summarization of the n
'{
the concrete batch plant scales by leaning on the wires connecting the (a).
Additionally. Mr. Witt stated concerns about a possibly missed hcid point during the welding of the fuel pool liner and scales to the sensors.
In an internal that some welding had been done by an uncertified welder.the B&R investigators s 17, 1082, The B&R B&R memorandum dated August Mr. Witt's concerns and their fin'. lings relative to the concer
'i aggregate was apparently partially substantiated but of no concern in that the aggregated pile, rather than actually being unacceptable, simply had The matter was documented on not been tested prior to use as required.
Deficiency and Disposition Report C-446 dated December 9,1976, whichIn
[
appears as attachment A to the memorandum. hold point i
documents that no hold point was missed.
allegations, the memorandum states that the allegations were investigate 4
and found to be without basis but provides no other information.
j Personnel of TUGC0 performed a separate investigation of Mr. Dillingham's The results of allegations (the Feehan letter) during August of 1982.that i d
This investigation found that two of eight items were September 2, 1982. substantiated with one of these being the same item th 4 <
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1 Both of the substantiated allegations were by the B&R investigation.found by the investigators to have been adeouat corrective measu'res had been taken or were in progress.
10, 1982, one of the TUGC0 In a separate memorandum dated Decemberinvestigators docum One of these Mr. Dillingham apparently made yet additional allegations.
i allegations regarded welding done by an uncertified welder on the turb generator pedestal (by implication). mentioned Mr. Witt who d broken off and was buried in the main dam.
ht personally driven a front loader that returned dry and lumpy cement l
had been rejected to the bin, that this cement had been subsequent y d
The writer of the memorandum stated that he had encourg illingham to T
Mr. Dillingham in turn was reported as take his concerns to the NRC.saying that he had intended on goi ll
_~
ll instead.
f It appears that Mr. Dillingham carried out his above stated intention h b t of that the above referenced newspaper article has ap f
NRC Region IV determined that the allegations in the news t
j article should be investigated but that those made in the Feehan let er the NRC.
This decision and in.the telephone conversation with TUGC0 should not. lier concerns was based on the premise that Mr. Dillingham has had his ear lf satisfied except for those appearing in the article.
Regarding tne above summarized allegation (a), the SRIC establis Mr. Witt was no longer an employee at CPSES and further established C
he had relocated from the Glen Rose, Texas,. area to another state. N i
Region I'/ personnel made several attempts to contact Mr. Witt by f
A registered letter, receipt requested, at his new address, to no avail.
i i
IV as soon as was then sent to Mr. Witt requesting that he contact Reg on Iteceipt of the letter was acknowledged but as of this date,It f
possible.
Mr. Witt has not contacted the region.
intend te assist the NRC in investigating allegations attributed to him It should be noted that only the B&R investigative group has been a
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l failed.
j to establish contact with Mr., Wi t; all others have apparent y Regarding summarized allegation (c, the SRIC, with assistance o I
Region IV inspector, was able to establish that the underwater lig i
standards were fabricated in such a manner as to leave drilling ch ps It was also established that the inside and had not been removed. lighting standards were fab i
There are no QA program which included various welding operations.
records of inspection or of the welders involved or of the we e
(g ye t-procedures utilized.the A/E considered the lighting standards to be within th fuel could should tne standards physically fail during the seismic eve i
been classified as Seismic Category II (licensee's FSAR definition fo be damaged.
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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 hase been included in the QA program.
This is considered to be a violation of Appendix B to 10 CFR 50.
j.
[
Regarding the premise that the drilling chips inside the standards could be swept into the reactor during refueling and cause an accident, the SRIC j
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 The size of into the reactor at the conclusion of the refueling process.
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 l
Further, the idea that the chip could fuse to core and create a hot spot.
the control rods is equally remote in that far higher temperature would be i
required in the core to achieve such fusion than actually will exist i
there, the differential being 600* to 800*F.
Thus, the safety significance 3(
of the chips is very small. The uncontrolled (no QA) problem with the 17 l
- c standards is relatively more important since workmanship on the devices
~
Regarding sumamrized allegation (c), the has not been established.
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 j',
-J investigating and evaluating the second party's allegation rather than Mr. Dillingham's.
8.
Postino of NRC Form 3 10 CFR 50 was revised by 47 FR 30452 to add 10 CFR 50.7 " Employee Protec-l tion."
The change was published July 14, 1982, and had an" effective date An important element of the change was that of a of October 12, 1982.
requirement to post NRC Form 3 at locations where the form can be readily 5
It has viewed by employees on their way to or from their place of work.
The SRIC learned of been alleged that the licensee did not post the form.
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 f.orce can readily see it, particularily ai The SRIC has been informed by when departing from the construction area.
licensee employed personnel that they received and posted the forms in the A senior B&R manager administration building about the first of 1983.
{
indicated th'at the forms were received, he believed from B&R's Houston office, sometime between Thanksgiving and Christmas and were posted on the
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craft labor bulletin board near the " brass alley" well before the first of It is thus clear that the forms were not posted on the the year.
It is specified effective date of the change to 10 CFR 50 as alleged.
much less clear as to when the forms were actually posted nor is it clear The " brass that most people would even have been aware of the posting.in size with alley" bulletin board is a large board, perhaps 4' by G' The majority of the postings are required under various many postings.
The posting of an additional form probably federal statutes or regulations.
As of the time of would not draw much attention from the average worker.
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tractor inspection by the SRIC, the licensee and his principal s te con in compliance with the were found to have,the form posted and to be regulation.
Management Interviews 9.
The SRIC held management interviews with one or more of the person inspec-identified in paragraph 1 on a nearly daily basis throughout the i
s special tion period to discuss NRC findings developed during var ouThe discussio inspections and investigations.
licensee's positions on the NRC findings.
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-ysc.eteoas QA og yea racw of ator us se Jamansere ser eas e Assurance Prooram for the Fabrication and Failure to Imolement a Ouality Pole Assemoiies Installation of Electrical Unoerwater Floociiont h ll Criterion II of Aopendix B to 10 CFR 50 requires that the applicant s a identify the structures, systems, and components to be covered by the
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trol over quality assurance program and that the program shall prov de con d
si activities affecting quality of the identified structures, system NRC Regulatory Guide 1.29 which in paragraphs 2 and 4 require the
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h se app'licant to identify those structures, systems, and comoonents w o continued function is not required (in a design basis accident) but whose failure could reduce the functioning of any plant feature identified in s.
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- other paragraphs to an unacceptable level.
et !
Contrary to the above, the Senior Res'ident Inspecto
- n tion -
2 activities and rev'iew of design drawings that group of devices i-
' collectively identified as " Electrical Underwater Floodlighting ul
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(Drawing 2323-EL-0925-02)
Guide'1.29 and were not included within the licensee's Quality Assurance
- *l Mechanical failure of the devices in a seismic event could t
damage fuel during reactor core installation activities
- i. '
Program.
a f,]l to of the pole assembly resulting in damaging fuel is very remote due
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_ the design of upper and lower pole retention devices.
j,l (Supplement II.D.)
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