IR 05000445/1977001
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U. S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT
REGION IV
IE Inspection Report Nos. 50-445/77-01 Docket Nos. 50-445 50-446/77-01 50-446 Licensee: Texas Utilities Generating Company Category A2 Facility: Comanche Peak Steam Electric Station Units No. 1 and 2
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Location: Glen Rose, Texas Type of Licensee: Two W PWR's,1161 MW(e) each Type of Inspection: Routine, Unannounced Dates of Inspection: January 18-21 and February 2-4, 1977 Dates of Previous Inspection: December 9-10 and 14-17,1976 Principal Ins::ector:
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7/4 [77 m
R. C. Stawart, Reactor Inspecpr (Details I)
Date Acc:=panying Inspectors:
J. I. Tapia, Reactor Inspector (SSI Dam, Details I)
A. B. Rosenberg, Reactor Inspector (Details II)
Other Accompanying
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Personnel:
R. E. Hall, Chief, Engineering Support Section (2/4/77 only)
C. L. Heck, Engineering Aide (2/4/77 only)
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Resiewed By:
W. A. Cros5 man, Chief, Projects Section Date Q)
a512o20072 es1106 PDR FOIA GARDEB S-59 PDR
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76-09/III.B. Calibration of Argon Flow Meters The licensee previously determined that argon flow meters used for welding had not been included in the calibration program. There is no change on this matter since the last IE inspection. This matter will remain open until the flow meters are included in the calibration program and have been calibrated.
76-11/III. A.
Exoanded Metal Mesh Construction Joints During a prior inspection, the IE inspector observed expanded metal mesh left in a construction joint which was not bonded to the previous concrete placement. Gibbs & Hill Engineering has analy:ed the joints assuming no shear strength in the
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observed unbonded areas and have concluded that the construc-tion joints at placement 201-2805-004 are adequate to perform their design functions.
(Details II, paragraph 3.)
This matter is considered closed.
76-12/ III. A.
Drawing Control During a previous inspection, drawings were observed being used in the field marked "Not For Construction." Brown & Root Construction Procedure DCp-3 has been revised with Interim Change Notice No. 4, which discontinues use of the "Not For Construction" stamp and replaced it with an " Approved For Construction Except For Rev.
and Rev.
" stamp.
(Details II, paragraph 4.)
This matter is considered c~losed.
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76-10/III.2.
SSI Dam - Gradation Filter "A" Material As previously reportedl/ the licensee informed the Region IV staff by letter dated October 20, 1976, that gradation problems experienced with filter "A" materials were considered to be a Construction Deficiency (CDR) within the context of 10 CFR 50.55(e).
Under a cover letter dated December 10,1976, (TXX-2059) the licensee submitted a report outlining the details regarding the CDR; however, it was determined by the NRC Region IV staff that sufficient information relative to the deficiency and corrective action taken was not contained in the report. By letter dated January 17, 1977, the licensee provided supplemental information, further clarifying the deficiency and corractive action.
On the basis of previous IE follow-up inspectionsl/and the licensee's submittal of the formal report, this matter is censidered resolved.
1/ IE Insoection Reoort No. 50-445/446/76-12, dated 1/4/77-3-
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SUMMARY OF FINDINGS
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I.
Enforcement Actions A.
Items of Noncomoliance None B.
Deviations None
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II.
Licensee Action on Previously Identified Enforcement Matters A.
Items of Noncompliance None B.
Deviations None III.
New Unresolved Items 77-01/III. A.
Cold Weather Concrete Curing The licensee, during routine site surveillance. activities, identified concrete curing surface temperatures which were below specification limits. Tubsequent licensee investigation could not identify a documented pour plan in the placement records which might have concerned itself with the impending cold weather. This matter vill remain open pending resolution of site surveillance reports Nos. C-134-77 and C-135-77.
(Details II, paragraph 5.)
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IV.
Status of Previously Reported Unresolved Items
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76-09/III.A. Sucolier Control For Calibration Services The licensee had previously determined that purchase orders for calibration services for subcontractors should require QA controls. The licensee advised the IE Inspectors that procedures for the new calibration program would be avail-able for NRC review during the next inspection. This matter will remain open pending evaluation of the calibration program.
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76-03/III.B. Pioe Fabrication Shoo 0A/0C Program During a previous inspection 2,/, it was observed that the con-tractor's planned piping activities included on-site pipe-spool fabrication, 8" pipe sizes and under; however, a pipe fabrication shop QA/QC program was not fully developed as described in the
PSAR, paragraph 17.1.1.5.
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During the inspection period February 2-4, 1977, the IE Inspector
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i conducted a review of the recently develooed B&R construction procedure, PCP-5, " Pipe Fabrication Shop General," dated 12/14/76, and S&R QA/QC procedure QCP-3.4 " Inspection of Pipe Fabrication,"
(draft) dated 1/19/77. This matter is considered resolved.
(Cetails I, paragraph 6.)
V.
Design Chances None VI.
Unusual Occurrences None VII.
Other Sicnificant Items
None j
VIII.
Manacement Interview Site Meeting On January 21, 1977, at the conclusion of the inspection conducted January 18-21, a meeting was held udth the following
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licensee representatives in attendance:
Texas Utilities Generatine Comoany (TUGCO)
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D. N. Chapman, Manager, QA
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P. M. Milam, Site QA Supervisor C. L. Biggs, QA Engineer Texas Utilities Services, Inc. (TUSI)
L. F. Fikar, Vice President, Engineering and Construction E. G. Gibson, Project Engineer C. H. Gatchell, Resident Manager
,7 2/ IE Ins:ection Recort No. 50-445/446/76-03, dated 7/26-20/76-4-
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Gibbs & Hill (G&H)
F. L. McAllister, Mechanical Engineer i
Brown & Root, Inc. (B&R)
H. C. Dodd, Project Manager C. E. Bonin, Assistant Project Manager P. L. Bussolini, Project QA Manager W. E. Childress, Jr., Project Engineer R. N. Best, DDR Supervisor Freese & Nichols (F&N)
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J. M. Dodson, Resident QA Manager
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i Mason-Johnston Associates (MJA)
R. C. Mason, Project Manager During the meeting, the IE Inspectors reviewed the status of the five previously identified unresolved items as identi-fied in Sections III and IV of the surnary of this repart.
In diseassing the new unresolved item regarding the lack cf adequate cold weather concrete curing protection (Details II, paragraph 5.), the IE Inspectors expressed concern that this problem was allcwed to recur.
The licensee representatives indicated that the matter will
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be reviewed in depth and corrective measures will be initiated.
In discussing matters regarding the CDR relative to the SSI
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Dam filter "A" material, the IE Inspector stated that the final formal report has been reviewed by the IE Region IV staff and that there are no further questions regarding the matter.
Cn February 4,1977, at the conclusion of the inspection con-t ducted February 2-4, a meeting was held with the following l
licensee representatives:
Texas Utilities Generatino Comoany (TUGCO)
D. N. Chapman, QA Manager
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P. M. Milam, Jr., Site QA A. Vega, Senior Engineer, QA
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Texas Utility Services, Inc. (TUSI)
C. H. Gatchell, Resident Manager Gibbs & Hill (G&H)
J. J. Moorhead, Resident Engineer Brown & Root, Inc. (B&R)
H. C. Dodd, Project Manager C. E. Bonin, Assistant Project. Manager P. J. Karnoski, Nuclear QA Manager P. L. Bussolini, Project QA Manager D. L. Hansford, Senior-QC Engineer D. G. Horton, Houston Coordinator W. E. Childress, Jr., Chief Project Engineer Freese & Nichols (FLN)
R. A. Thompson III, Project Engineer J. M. Dodson, Resident QA Manager Mason-Johnston Associates (MJA)
R. C. Mason, Project Manager The IE Inspector stated that the scope of the inspection conducted during the period 2/2-4/77, was limited to a review
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of the B&R Pipe Fabrication Shop QA/QC program. The specific procedures reviewed were identified and discussed. The IE Inspector indicated that no discrepancies were identified during the inspection.
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50-?45/77-01
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50 105/77-01
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DETAILS I Principal Inspector:
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R. c. Stewart, Reactor Inspector # Projects Section s
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Accompanying Inspector:
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(Safe Shutdown Dam Only)
VJ ea h J. I. Tapia, Reactor [Ipspector tern, Engineering Support Wection Reviewed By:
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W. A. Crossman, cnief, Projects Section 1.
Persons Contacted Texas Utilities Generating Comoany (TUGCO)
D. N. Chapman, QA Manager P. M. Milam, Jr., Site QA Supervisor Brown & Root (B&R)
H. C. Dodd, Project Manager W. E. Childress, Jr., Chief Project Engineer P. L. Bussolini, Project QA Manager D. L. Hansford, Senior QC Engineer R. A. Peters, QC Inspector (CB&I)
S. Miller, QC Engineer (Civil)
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Mason-Johnston Associates (MJA)
R. C. Mason, Project Manager W. T. Cromeans, Resident Chief Technician R. D. Cody, Assistant Resident Technician
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Freese & Nichols (F&N)
W. L. McRath, Assistant SSI Dam QA Administrator Gibbs & Hill, Inc. (G&H)
J. V. Hawkins, Site QA Represeni:ative R. V. Fleck, Site QA Supervisor
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Chicago Bridoe & Iron Comcany (C3&I)
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E. H. Dildy, Project Supervisor M. Jeffers, Welding and QA Supervisor C. R. Gregory, Welding Supervisor K. Botkins, Welding Supervisor 2.
Scope of Insoection The scope of the inspection was limited to the observation of work activities relative to installation of the Unit No. 2 containment floor plate liner; follow-on review of the Safe Shutdown Impoundment (SSI) Dam work progress and QC activities; and a review of the B&R fabrication shop QA!QC program.
3.
Status of Project The licensee has reported that the overall plant progress is 25.2%
complete as of January 11, 1977.
(Unit No.1, 22.3%, and Unit No. 2, 2.9%.)
Installation of the Unit No. I containment liner is currently com-plated through the 14th ring. Preparations are complete for concrete placement of the third lift of the Unit No. I containment exterior wall.
Installation of the Unit No. 2 containment liner floor plate (ele-vation 805') is in progress and estimated to be 20 complete.
Fill placement is continuing of the SSI Dam; however, as with most construction activities, recent cold and inclement weather has delayed work progress.
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4.
Unit No. 2 Reactor Containment Floor Liner plate Installation a.
Observation of Work and Work Activities During this portion of the inspection, the IE Inspector observed handling, placing, welding and NDE activities involved in the installation of the Unit No. 2 containment liner floor plate at the 805' elevation. Observations were made of fit-up and
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first pass welding of seam welds Nos. 20, 23, 25, 41, 30A, and 33A.
In addition, examination of completed welds Nos. 4, 8, and 11 was also conducted.
It was observed that all welds were identified as to weld number, welder identification, and type of NDE test conducted with corresponding dates. Although final acceptance had not s
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been made on any of the completed welds at tne time of this inspection, in-process work and QC activities appeared to be consistent and in accordance with G&H specification No. 2323-SS-14 Revision 3, dated 11/17/75, and the applicable CB&I weld procedures. No discrepancies were identified during this part of the inspection.
b.
Review of Ouality Records In conjunction with observation of the in-process field welding activities,' the IE Inspector conducted a QA/QC documentation record review.
C3&I drawings, 74-2428U, Sheets R10, R11, R12 and R13, are being utilized to maintain record control of in-process in-spection activities. The IE Inspector conducted a correlation review of the field welds and NDE inspection activities ob-served in the field, against that which had been recorded on the control drawings.
In addition, the inspector reviewed current status of welder qualifications and vacuum box and magnetic particle inspector qualifications. No discrepancies were identified during this part of the inspection.
5.
SSI Dam Activities a.
'a'eather Conditions Due to inclement weather, recent progress on the Safe Shut-down Impoundment (SSI) Dam has been very limited. The cog-nizant Freese & Nichols QA personnel indicated that the Dam construction status remains at sixty-seven (67) percent com-plete. This value has not changed since mid-December.
Review of forty-one (41) Freese & Nichols Inspector's Daily Reports and respective inspection checklists covering the period from December 27, 1976, to January 16, 1977, indicated that icing conditions impeded the placement of fill material during the period.
b.
Field Laboratory Tests
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The inspector reviewed the followi.ng quality control tests that were performed during the period of limited work:
Test Number Test Description Test Date 1161 Relative density 01 -04-77 1162 Relative density 01-04-77 1116 Atterberg limits 12-18-76 I-3
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Test Number Test Descriotion Test Date
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1159 Atterberg limits 12-30-76
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1151 Sand Cone 12-28-76 1154 Sand Cone 12-30-76 1155 Sand Cone 12-30-76
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1168 Sand Cone 01-04-77
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1169-R1 Sand Cone 01-04-77 1176 Sand Cone 01-07-77
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1177-R1 Sand Cone 01-07-77 1180 Sand Cone 01-07-77
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i 1181-R1 Sand Cone 01-07-77 1186 Sand Cone'
01-08-77 1187 Sand Cone 01-08-77 1107 Sieve Analysis 12-18-76 1131 Steve Analysis 12-22-76 1163 Steve Analysis 01-04-77 1165 Steve Analysis 01-04-77 1160 Drive Cylinder 12-30-76 Results of these tests were in conformance with acceptance criteria established in Freese & Nichols Specification No.
FNSSI-1, Revision 11. No discrepancies were identified during this part of the inspection.
c.
Recair Work As a result of the freezing weather., the impervious material ramp at the north abutment at elevation 770.5 was damaged.
This ramp is the first step in the upward continuation of each layer, and as such, was exposed on three sides. The inspector witnessed rework being performed en the north abut-ment impervious core material and found it to be in accordance with established construction procedures. Scarification depth varied between six and eight inches which was well below the frost line. No discrepancies were identified during this part of the inspection.
6.
76-08/III.S. Pipe Fabrication Shop QA/0C Program During the period 2/2-4/77, the IE Inspector conducted an overall
. review of the pipe fabrication shop QA/QC program in conjunction with B&R construction procedure 35-1195-PCP-5, " Pipe Fabrication Shop General Operation," Rev. O, dated 12/14/76, and the'G&H Spaci-fication MS-438, " Shop Fabrication of Piping in the Field," Rev.1, dated 1/5/76.
It was observed b procedure contains an extensive,y the IE Insoector that the B&Rdetailed step by step proc involving 18 additional supcorting procedures. The IE Inspector did not identify any significant discrepancies during the overall review of the program as described.
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The IE Inspector also conducted a ceneral review of a draft copy of the B&R QA/QC Procedure CP-QCP-3.4, " Inspection of Pipe Fabri-cation," dated January 19, 1977.
It was noted by the inspector that an additional nine QA/QC procedures are referenced as supporting procedures, including the B&R ASME Code Manual.
In ' discussing the overall pipe fabrication shop program with the cognizant licensee representatives, the inspector was informed that the start of "Q" listed pipe fabrication is currently scheduled for March 15,1977, pending final development, review, approval and issuance of the applicable procedures. The IE Inspector informed the licensee that in view of the apparent program development, this matter is considered resolved; howev'er, additional IE inspections vill be conducted of pipe fabrication activities during subsequent routine on-site visits by the NRC staff personnel.
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DN 50-445/77-01 50-446/77-01 DETAILS II Accompanying Inspector:
d-e A. B. Rosenoerg, ReactordTnspector, Engineering Support Section
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Reviewed by:
- - - Pr R. E. hall, Chief. Engineering Support Section
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1.
Person Contacted
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a.
Texas utilities ~ Generating Co. (TUGCO)
P. M. Milam, Site QA Supervisor b.
Gibbs & Hill (G&H)
J. J. Moorhead, Resident Engineer R. V. Fleck, Site QA Supervisor J. V. Hawkins, Site QA Representative c.
Brown & Root (B&R)
E. F. Beacham, Calibration QA Supervisor S. F. Miller, QC Civil Engineer R. O. Taylor, Document Control System Coordinator
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d.
R. W. Hunt (RWH)
8. K. Kinkade, Site Supervisor
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2.
Scoce of Insoection The scope of this inspection included review of previously reported unresolved items and review of a licensee identified item of non-
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compliance concerning cold weather concrete curing.
3.
Excanded Metal Mesh Construction Joints (76-11/III. A.)
This matter was previously identified as an unresolved item in Inspection Report 76-11, dated November 24, 1976.
During that inspection, the IE inspector identified expanded metal mesh at the construction joints for concrete placement No. 201-2805-004 that was not bonded to the previous placements.
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a.
Engineering Evaluation During this insoection, the inspector reviewed the G&H letter, GTN-14703, to TUGCO, dated January 4, 1977. The letter stated that the shear capacity of the constructic'n joints was analy:ed using a ccmputer analytical program, assun.'ing zero shear strength in the areas of question. The letter concluded that "the construction joints at the Unit 2 containment mat pour 201-2805-004 are adequate to perform their design functions."
Additional information relating to the input for the analysis was requested by the IE inspector. A G&H representative obtained the information via telephone from the G&H New York offices. The amount of unbonded mesh used for the analysis was 200 sq. ft.,
approximately 130 sq. ft. in one face and 70 sq. ft. in the other face, in the central area of the joints. This data is in con-formance with inspector observations provided in previous inspection
reports.
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b.
Quality Control The B&R QC Concrete Placement Checklist Revised December 16, 1976, was reviewed. The revised checklist included three new items concerning expanded metal mesh used in construction joints:
. Total area of expanded metal mesh at construction joint
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adequately secured to prevent void gaps.
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. Where small, localized areas of unsecured expanded metal exist, they must be removed prior to concrete placement.
. Where a large area of unsecured expanded metal exists, it must be removed prior to concrete placement.
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The inspector was told that the guidance leading to the checklist revisicn was promulgated thru informal meetings between G&H, B&R i
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and Texas Utilities Services Inc. (TUSI).
A memorandum from TUSI to B&R, TUF-2504, dated January 11, 1977,
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was reviewed. The memo requested that steps be implemented to prevent recurrence of placing concrete against unbonded mesh and in locations where overlap of the mesh could restrict grout flow, i
Part of this request had previously been addressed by B&R.
This matter is considered closed based on the results of the engineering
evaluation and the steps taken to initiate preventive actions.
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4.
Drawing Control (76-12/III.A.)
This matter was previously identified as an unresolved item in Inspection Report 76-12, dated January 4,1977.
During that inspection, it was
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observed that drawings wero being used in the field which were marked
"Not For Construction." Further investigation revealed that document i
control procedures (DCP-3) allowed this condition.
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During "this inspection. Interium Change Notice (ICN) No. 4, dated January 4,1977, to DCP-3 was reviewed.
The ICN discontinues the use of the "Not For Construction" stamp and implements the use of a new stamo, " Approved For Construction EXCEPT For Rev.
& Rev.
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The ICN did not provide for changing the drawings already issued to the fiel d.
A B&R representative stated the "Not For Construction" stamp on drawings already issued would be crossed off and replaced with
" Approved by TUSI" stamps as the latest revisions are approved. The
" Approved For Construction EXCEPT for Rev.
& Rev.
" stamps will also be crossed off and replaced by " Approved by TDTI" stamps as the latest revisions are approved.
This matter is considered resolved.
5.
Cold Weatner Concrete Curing (77-01/III. A.)
While reviewing recent B&R Deficiency and Disposition Reports (DDR's),
the inspector identified DDR-C-460, dated January 13, 1977, which
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identified surface temperatures during concrete curing which were below specifications.for the auxiliary building and service water in-take structure areas. The inspector recognized this matter as a recurrence of a licensee identified problem of approximately one year ago. Further investigation revealed two TUGCO/G&H Site Surveillance
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Reports (SSR's,), dated January 4 and 6,1977, which also dealt with cold weather concrete curing. SSR No. C-134-77 dated January 4, 1977, i
indicated violation of ACI-301, " Specification for Structural Concrete for Buildings," Section 14.5, " Curing and Protection," and ACI-306,
" Recommended Practice for Cold Weather Concreting," Section 1.4.
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These codes are incorporated by reference in G&H 3pecification 2323-55-9,
" Concrete," and B&R procedures CCP-13. " Concrete Curing," Revision 0 and CP-QCP-2.4, " Concrete Inspection and Testing," dated February 2,1976.
The SSR reported surface temperatures below 320F on the curing concrete of placement No. 101-2808-001 on both December 30 and 31, 1976.
J Site surveillance Report No. C-135-77, dated January 6, 1977, concerned
I the records for placement No. 101-2808-001. This SSR reported that there was no placement plan in the B&R QA document vault to meet requirements of Regulatory Guide 1.55 and the project PSAR.
Site Surveillance Report C-068-76, dated January 7,1976, cited an identical condition, " frozen" concrete during curing of the Unit No.1 Safeguards
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Building Mat.
This SSR and attendant conditions precipitated two B&R DDR's, DDR-201 concerning the temperature of the concrete and DDR-211 concerning the low strength of field cured cylinders for that same placement.
The preventive action of DDR-211 states:
'The concrete dept. of B&R Construction has since been monitoring daily posted weather forecasts and using the information to plan the curing ooerations (insulated blankets, space heaters, burlap
& water, etc.) appropriate for expected conditions."
i This matter will remain unresolved pending resolution of SSR C-134-77 ar.d SSR-C-135-77 by the Licensee.
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A.26.d. (Tay or)
itRC Investigation Report 79-11 discussed an allegation that a small amount of concrete was phced on the Unit 1 containment building dome under very unusual circumstances on a rainy evening in January 1979. The subsequent investigation substantiated the allegation and it was found that concrete was placed without the 3equired
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inspection and testing of the concrete. A tiotice of Vio-lation was issued to the Applicants for failing to fully implement their QA program.
The Applicants respended to the findings in the above NRC Investigation Report 79-11 by letters dated June 12, 1979, (TXX 2998) and September 17, 1979, (TXX 3043).
In the June 12, 1979, letter, the Applicants stated that Texas Utilities Services, Inc. ("TUSI") engineering retained the services of an established materials and concrete consultant for
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the purpose of evaluating the inplace condition of that portion of the deme in question.
In addition, construc-tion concrete supervisory personnel were instructed to notify senior construction management prior to batching
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and placing concrete should a similar situation occur.
The Applicants stated in their letter of September 17, 1979, that the results of the investigation and evaluation by the consultant indicates the questioned concrete satisfies design re'quirements.
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The responses were accepted, subject to verification, by
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Region IV letters dated July 5,1979, and October--10, A'y
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1979, respectively.
A.26.e. (Taylor)
NRC Inspection Report 79-20 discussed an allega-tion,receivedinalettertotheCommission,thatfrash
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had been placed in concrete formwork, and that concrete had been placed over the trash to cover it up.
The
, building was identified as a containment building, but not which one.
It was alleged that the incident occurred during a drunken Christmas party in December 1978.
The investigkion revealed that the letter writer had heard the story frem the father of the individual who was sup-posedtohaveparticipatedinthepartytot.keextentof
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driving the " drunks" home after work.
The investigators interviewed the individual; identified by the alleger as the source of the story, who prc=ptly denied thkt the incident took place as alleged.
Further investigation
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determined that the alleged party had to have taken place
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on December 3,1978.
It was established that the source of allegation, although employed at the station, was, in fact, not at work that day.
The interview with other persens known to have been associated with the concrete placement involved ' denied that there was a party during the placement. No items of noncompliance or deviations were identified.
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%;,"I" C/VSE )Ldd~ M 3 M 73 18. Gibbs & Hill Specification 2323-55-30 " Structural Embedments," March 19, 1981.
19. Gibbs & Hill Report, " Evaluation of LOCA Temperature Effects on Pipe Supports,"
August 26, 1982.
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20. NPSI Report, " Load Transfomation Study on Richmond Insert & Tube Steel Assam-blies, " September,1982.
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22. PSE Guidelines,Section VI, " Richmond Inserts and Anchor Bolts Stress Allowables."
23. TUGC0 Procedure CP-HBM-0.1,. "Hilti Bolt Inspection Manual," Revision 31.
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24. Pol'ytechnic Institute of Brooklyn Test Reports for Richmond Screw Anchor Ccmpany..
25. PSE Report, " Richmond Inserts--Prepared for 1/17/83 meeting with NRC.
26. The evaluation / analysis (including any and all calculacions) for the worst-case analysis of an eleven-foot long member under LOCA conditions using available load-displacement data, perfomed by the SIT.
i
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27. The evaluation / analysis (including any and all calculations) for the shear cone
l analysis made by Applicant regarding the allowable Richmond anchor tension loads (page 19 of I&E 82-26/82-14).
-
Mj 28. Appendix B of the Merican Concrete Institute's (ACI) " Code Requirements for Nuclear Safety-Related Concrete Structures, "ACI 349-76.
~
3. Richmond Screw Anchor Company Bulletin No. 6 (for allowable loads).
-
30. All documentation (memoranda, letters, calculations, etc.) for the Applicant's statement (reported on Page.19 of I&E 82-26/82-14) that the manufacturer of Richmond Inserts has " indicated that a factor of safety of less than three
has on occasion been recomended in the concrete precast tilt-up industry."
~31.
All documentation (memoranda, notes, letters, calculations, etc.) done by SIT in reviewing the manufacturer's data published in reference 24 above.
5 32. ACI 349-80, " Code Requirements for Nuclear Safety-Related Concrete Structures."
>
33. All test modeling and test dat'a available in reference to the Richmond safety factor used by Applicants.
American Society for Testing and Matericis (ASTM), " Standard Test Methods for Strength of Anchors in Concrete and Masonry Elements," ASTM E488-76.
35. All documentation, user guides, code / meaning lists, etc. necessary to properly comprehend the load transformation behavior of a typical insert and tube steel configuration based on a finite element model analysis using the STARDYNE computer code.
(E.g., the STAR 0YNE computer code and how to
.
use it.)
h 62%
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.
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.
.
.
196. CMC 81948, Revision 3 (and all previous revisions) dated October 28,
.
1982 which was shown to SIT durin cant." (I &E 82/26-82/14, pa5e 54f." subsequent discussions with Appli-
/'
.
197. All docmnentation (notes, etc.) comprising the SIT's verification of the replacement of the damaged tube steel, discussed on pag'e 54 of the I&E report.
198. All documentation (notes, etc.) comprising the SIT's review of the design status of the pipe supports identified by Messrs. Walsh and Doyle.
(I & E 82/26-82/14, pages 54 and 55.)
199. All samples of pipe supcort designs reviewed by SIT in its evaluation of the implementation of the design review process--i.e., all pipe. support designs which had completed the design evaluation process.and had been marked
-
" vendor' certified." (!&E 82/26-82/14, page 55.)
-
200. Military Standard 1050-63, " Sampling Procedures and Tables for Inspection by Attributes."
201. All documentation (notes, memoranda, etc.) cresented and/or developed during the exit interview conducted February 8,1983 with the Applicant.
Ac 35. 202. The Prestressed Concrete Institute handbook. (See I&E 82/26-82/14, cage j__
20.)
.
.
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.
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.
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.
__
_
CIVIL / STRUCTURAL ALLEGATION REVIEW CATEGORIES Category Est. Mandays Allegation Package Assigned Schedule No.
Subject to Complete Nos.
Prepared to Open Close Remarks 1.
Inadequate materials AC-16, AC-19 AC-20, used in concrete AC-21, AC-27 2.
Concrete placements AC-22,'AC-23 F
.
3.
Poor weather conditions AC-24, AC-35 placement of concrete 4.
Concrete voids / cracked /
AC-25, AC-32, AC-34, crumbled AC-41, AC-28, AC-33 5.
Miscel, concrete AC-17, AC-18, AC-26, AC-29, AC-31, AC-36, AC-42, AC-43 6.
Rebar improperly AC-30, AC-37, AC-38, installed / drilled or AC-39, AC-40 omitted
.
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.
On September 6,1983,'M, a former B&R carpenter and carpenter's foreman at CPSES, was telephonically interviewed by NRC Investigator H. Brooks GRIFFIN.
gstated he worked at CPSES for about 2 years, but stated he did not remember ne exact time frame. @ aid he worked under N a general foreman.
qsaid he had testified at a ASLB hearing regarding what he believed was
" sloppy" work related to the pouring of concrete.
$tated he believed the work he performed at CPSES was done properly. M said that while he worked at CPSES, he was not intimidated by anyone nor did anybody attempt to intimidate him. 4 said he had adequate freedom to perform his job and did not recall anyone threatening or harassing him. W said his ole concern was covered in his testimony to the ASLB regarding " bad concrete sork."
6 technical concerns were forwarded to Region IV staff for evaluation.
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on a concrete pour and they just go right ahead and do it
anycay.
The boss signs off or somethine.
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'
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was at tne cracx of cawn they would nave their concrete
crew out the<e ready to start pouring concrete.
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~
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a
pours and included so much garbage and nard hats and lunch
.
ooxes and crash and stuf f that TUGCO wouldn't let them
.
power concrete until they inspected.
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Department had made their inspection, QC made tneir
.
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difference QC people, electrical, instrument, mechhnical
"
and everybody goes in there and they sign off their part
,
of it.
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forms would still be full of trash.
I have seen Brown and Root still waiting at 2
o'clocx in th'e afternoon and TEDCO wouldn't let enem pour
concrete cecause the forms are full of trash, the forms
are in the wrong place and they left stuff out of the
pours.
Some of tne inspectors that found tnis stuff,
especially the dirt and junk lixe that, well, tnat is good F"
'
TAYLOE ASSOCIATES S
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I CROSS REF./0R C0HPLEil0N ALLEGER-0AIE RECEIVED TA5E SOURCE TRACKING CATEGORY 1-7 ScalEDULE SOURCE ND.
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/10/79
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-
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CIVIL / STRUCTURAL ALLEGATION REVIEW CATEGORIES Cctegory Est. Handays Allegation Package Assigned Schedule No.
Subject to Complete Nos.
Prepared to Open Close Remarks j
1.
Inadequate materials AC-16, AC-19, AC-20,
'
used in concrete AC-21, AC-27
,
,
.
,
[
2.
Concrete placements AC-22, AC-23
.
.
?
3.
Poor weather conditions AC-24, AC-35
.
placement of concrete'
4.
Concrete volds/ cracked /
AC-25, AC-32, AC-34,
,;
crumbled AC-41, AC-28 AC-33 i
l S.
Miscel. concrete AC-17, AC-18, AC-26,
,
AC-29, AC-31, AC-36,
AC-42, AC-43
.
,
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'
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Rebar improperly AC-30, AC-37, AC-38, installed / drilled or AC-39, AC-40
- '-
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ALLEGATION REVIEW
.
CASE NUMBER 4-83-A-01
@
DATE OPENED 01/07/83 FACILITY NAME CPSES 50-445 50-446 SUBJECT Improper Concrete Pour &
Metal Shavings in Fuel Pool SOURCE OF ALLEGATION ex B&R Const. Fore-man NUMBER OF ALLEG.
ASSIGNED TO RPS A CROSS REF. NO.
ACTION SCHEDULED Closed FIRST/LAST NAME T. Westerman DATE ASSIGNED 01/07/83 REPORT NUMBER 1st: 0 2nd:
Lst: 83-03 FUNCTIONAL AREA FTS NUMBER 8-728-8100 OUE DATE 00/00/0000 ALLEGATION SUBSTANT 02/01 SORT CODE X
OATE CLOSED 03/28/83 ACTION OFFICE RIV DETAILS: Alleger states (1) rejected aggregate was used in the reactor building basemat and (2) he was prevented from cleaning metaT' shavings from lamp posts that were drilled. These shavings. he alleges, could enter when the head is remevec and cause core damage by blocking flow through fuel elements.
The front end loader operator was We have contacted him but he will not come to see us. Technical insp
. ion of item (2) indicated that the work on the lamp puts was not covered by the QA program. An NOV is being issued for this item. Allegation (1) was not substantiated.
""
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C0HANClfE Pi AK NUCLf AR l'Oh!R Pl ANT AtttdATT541 AN5/6E~1EvilYl5Afi5El iURNARy
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CROSS REF./OR CDHPLE110N ALLEGER-DAIE RECEIVfD IA5a SOURCE TRACKING CAlfGORY l-1 SCIIE Diff t 50tlRCE M
AlttGAll0N OR CONCERN ACil0N/5TATUS ANON CONFIO BN/DATE SY5ftH NO.
IIAO OPEN COHetEIE
,qtJHtNI PAGE AC-22
" Sloppy"wordpouring
04-006 3/7/84 A-23 concr ete ART P. 16
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AC-23 Concrete pour discrepencies
84-D06 3/7/84 A-27 ART Testimony, P. 31 AC-24 Concrete placed on Unit 1 Initial dispositten ASLB
linknown 3/82 Testimony, containment done in the applicant response Pre-filed ARI did 5/24/82, P. 39 rain ILr, did 6/12/79, testimony A.26.d. IR 79-11 9/17/79. RIV (Taylor)
acceptance Itr 7/5/79, 10/10/19 Subject te Verifica-tion. Documented IR 79-13 & IR 79-24 AC-25 18ellow places in concrete
84-006; 3/7/84 -
behind stelnless liner ART A-4, Testinony,
P. 65-68 AC-26 fquipment set on concrete
84-006; 3/7/84 -
before adequate curing ARI A-4, Testimony, P. 69-70
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AC-27 Reject material used in
84 006;.4/7/84 -
concrete pours ART A-4,' Testimony, s
P. 73 AC-28 Concrete " crumbled" N
q4-83-Olt 5/23/83 ART A-24, P 1, 2 i
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(7) A truck load of concrete was place in a containme u j
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building well* with a slump of 41" (4" maxiinum
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Concrete cylinder compression strength test results (8)
were falsified at the direction of general foyeman
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and laboratory manager.
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Concrete cylinder compression tests were run (9)
purposely faster than allowed.
(10) Rece tification of N inspector's were done
"open book" with answers given.
The investigation of the ten allegations appearing in the news media was completed on May 7,1979, and documented in
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NRC Investigation Report 79-07.
Eight of the allegations could not be substantiated, but a detailed investigation
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of all aspects of these allegations determined that even
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if the practices in question had occurred, there would be One no impact on the quality of' concrete production.
The other allegation remaining allegation was refuted.
was substantiated, but the practice involved had been previously identifi'ed and corrected.
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s The one substantiated allegation related to the calibra-tion of a pressure * gauge by an individual who Was not
properly cer ified.
The test, however, was found to have been properly performed by a competent person who
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l was not, however, certified.
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A.26.c. (Stewart)
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NRC Investigation Report 79-09 discussed allegations appearing in the Fort Worth Star Telegram on April 4, 5, 6, and 8, 1979, concerning concrete inspection i
and testing activities.
It was alleged that:
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(1)
Agg egate tests were falsified.
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(2)
Equipment used to test aggregate was unused.
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(3)
Shortcuts were taken on tests involving sizing of
agg.egate and moisture content.
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Truck drivers added excessive water to trucks in
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(5)
V Concrete for turbine building placement was rejected,
but placed anyway.
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(6)
Concrete for Unit 1 containment basemat was placed
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without testing.
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A'.47.a. (Taylor) NRC Inspection Report 78-11 discussed a finding
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that B&R Pipe Department foremen were not assuring the cleanliness of piping runs as recuired by the S&R
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procedures.
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A.47.b. (Taylor)
NRC Insoection Report 79-04 discussed an;-
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ebservation that equipment was not being maintained in i
accordance with established requirements.
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A.a7.c. (Taylor)
NRC Inspection Report 79-04 also discussed a
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finding that wiring within the main control boards was not
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separated as required by the Final Safety Analysis Report (FSAR) even though the components had been inspected and l
accepted.
A.47.d. (Taylor)
NRC Inspection Report 79-06 discussed a finding
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that inspections of safety class alectrical cable tray
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supports were not being done in accordance with FSAR requirements.
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A.47.e. (Taylor)
NRC inspection Report 79-28 discussed a finding that electrical equipment inspection instructions were not sufficiently cceplete to assure an acceptable final
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installation.
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CIVIL / STRUCTURAL ALLEGATION REVIEW CATEGORIES Category Est. Handays Allegation Package Assigned Schedule
No.
Subject to Complete Nos.
Prepared to Open Close Remarks 1.
Inadequate materials AC-16, AC-19, AC-20, used in concrete AC-21, AC-27 A
w 2.
Concrete placements AC-22, AC-23
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3.
Poor weather conditions AC-24, AC-35
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placement of concrete 4.
Concrete voids / cracked /
AC-25, AC-32, AC-34, crumbled AC-41, AC-28, AC-33 5.
Miscel. concrete AC-17, AC-18, AC-26, AC-29, AC-31, AC-36, AC-42, AG-43 6.
Rebar improperly AC-30, AC-37, AC-38, installed / drilled or AC-39, AC-40 omitted
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Jr.was land off tapciober, weeks fle@dtomakejudgmentsonpotegttal-slon because he does not trust the.was fired three separate tim ld 61s not quali. to the Nuclear Regulatory Commit e
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safety dffsetal for the company after his charges prompted an ip-
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, ponasble for the Comanche Peak vestigation by Brown & Root Con-! ' safety basards thet involve thecom.
lie said the improperty tourer struction lac. The lavestigation - dation of concrete walls.
federslagencyteconductpalmpar. construction jobs at two difteter.t bagecouid shnft under tiit we.g et of mar power plant dismissed t!st lavesugition...
nuclear plants after testifying at tatted tosubstantiate anyofh
" lie's not quallfsed to be tpfklag Opponents of. auclear energy regulatory hearjogs about faulty i eWtostractor.causarigtpepipe
.rges Thursday that procedura
@dgr contract to bus 1d the plant for- *lle's not a structuraleptinger Ite) have questioned the comenission's. unde charges. Browa 8s Itoot as un. * about things like that,'Wasid.
wtlons at the plant equid cost reactor to break.
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. ho:n of douars to repaft and. If Texas Utstatsek.
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impartiahty in hearinge before the Ro%t f rontendloaderoperater. told corrected, could mean the po. ~ A second lavestigation, condu.' *.* not knowlgdgeable la engine 4 former Brown &
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Juanita E;lha,presidentof theCgt6 ct. ' technology orla concrete." jrtag
<s. tog Appeal Spar 4 f., r. < afederal Atomic Safety and Licens.. aens Assnegation for Sound I'nergy
. Inet for a nuclear disaster.
r.
. edbyTexasUtahtte:GeneratingCo.l' c* harg M allegations were goads by a company investigators fast August M 'most significaat.. Last esoth,a former weldtagin.
said alleast a ' dozen nuclear plant that he was ordered to use the rA *
one of three sister goinpagles that luer forerase who charged that employeesbackedoutof tesutying forms Texas t.ltthues esinvolved two alleged safety ciel were conducting e " covers awe of the charges., substantiateg.,, ytolapons that.he sold could result spector at.the plant.:who charged jectedaggregateresteria.byasupo. N
,.s, beforetheNA becausetheyfeared that he was fired la reprisal f or tgstl-reprisals. <
T*-
rter. Investigatore talked to two
? to hsee safety hasards that he liewaver,
. la a meltdown of the nucless corelit mony be gave before the NRC. was n
.6 One o
. ether men who worked in the area i others brought tf their atten-TUGCO'squgtityassurancefuperv>. tions ta e swpfg U4tgment sagteda e Department of 49 bot adsetnistra-6 *
'a charges was who said they knew nothing of th,'
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that rejet was mised with the concrete that a time is dead. o 4..astregate material incident. Maupervisor et the
'as fortase. N ' safety basar4..
sor. sand the vlolations de not pose a Dec. !R.
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.s live law fudge..
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lie sand he did act tatt hischarges *.t N 14spector was poured to form the bast for the
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nuclear reactor. '
.. Please set Comanche rsa k
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refuensg. "The shavings can be driuedthroughtbe stal reinforce' dada'I bsve the suthortty lo Put Dtik ~
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A' report on thelavestigatsoa said
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Bests on the concrete faued to show.
fuel ce M could' knowledge (ms to have Mk towork.,But evenitt
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wethtoed did, I wouldn't let him blackinau 8 "I I" even f se to the control rods "Diu,by treproperty dettle eles..
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us!" he said.
said the concrete was ingham said in his statement.
'.who supposed
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d be * if themetal rticles fused tothe yttnesse such hgles tietag drill
&nd Wlost his job
reac'hed or om s
I. c removedreds. 6spossiblethat they becausethenumberofemployeesis a
urso y.
g. could prevent the rods fross betag, cooperated to th9 cotapany"e Aug-being reduced as work at the plant ust investagstloa. la na interview tha e es gat a
replaced Properfy sed touldcauses ursday.besaldh is tist d *
s8ows den
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done
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postsby hts superintendent.The su-m str b
w pe said,waala too*
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December and asked !! he could ' After being assigned t' work in i
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h""sa*id the shadngs not erstlegationb6 h othislobback.Hsul6 ene area of the shop for several
s
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,w retsata 6aside the hottow. involved the aueged practice of
threatened to - mate ' the weeks. he called had then I
driiksg bolce. through steefrein. t arges pubhc if he was Det put
- lampposta.could become dislodged forced conc e retaarung walls at back to worL
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W.ht. Itace, group vice pres 6 dent of the
. by underwater currents when the the plaat pecial authertaatnes Mbo said his job is coq-own 4. Root Power Divtsson.
- reactor bead as being removed for must be give' n before holes are cerned caly with safety, hand he sd. lle sand he was land
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June 8, 1979
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In Raply Refer To:
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RIV-l
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Docket No. 50-445/Rpt. 79-09 Y
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50-446/Rpt '79-09
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Texas Utilities Generating Company
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j ATTN: Mr. R. J. Gary, Executive'Vice
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President and General Manager -
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2001 Bryan Tower Dallas, Texas 75201
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centlemen:
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This refers to the investigation conducted by Mr.' R. C. Stewart and other met:bers
'i of our staff during the period April 5,1979, through May 7, '1979, of activities
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autborized by NRC Construction Permits No. CPPR-126 and 127 for the Comanche Peak facility, Units No. I and 2, concerning allega'tions by-former Co=anche Peak e=ployees which appeared in news articles of the Fort Worth Star-Telegram on April 4, 5, 6 and 8, 1979.
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The investigation and our fin. dings are discussed in the enclosed investigation report.
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Within the scope of tlie investigation, no it' ems of noncompliance were identified..
Even though no items of noncompliance with NRC requirements were detected during this invest'igation, it was not possible.to either substantiate or refute several of the allegations. We would appreciate an opportunity to discuss the results of the inoestigation in our office af ter you have had an opportunity to review the report. The reason for this discussion stems from our desire to obtain your
!'
reaction to the vide range of matters covered in this investigation.
i; i
In accordance s-ith Section 2.790,of the NRC's " Rules of Practice," Part 2, Title
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10, Code of Federal Regulations { a copy of this letter and the enclosed investi-l'
gation report will be placed in the NRC's Public Docu=ent Room. ~ If the report, l
contains any information that you believe to be proprietary, it is necessary
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that you submit a written' application to this office, within 20 days of the date of this letter, request.ing that such information be withheld fro = public disclosure. The application must include a full staterent of the reasons why
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it is claimed that the infor=ation is proprietary.
The application should be
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Texas Utilities Generating Company-2-June 8, 1979 prepared so that any proprietary information identified is contained in an enclosure to the application, since the application without the enclosure vill also be placed in the Public Document Room. If we do not hear from you in this regard within the specified period, the report will be placed in the Public Document Room.
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Should you have any questions concerning-this investigation, we vill be pleased to discuss them with you.
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Sincerely, o
l-W. C. Seidis, Chief i
Reactor Construction and
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Engineering Support Branch
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Enclosure:
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II Investigation Report No. 50-445/79-09 50-446/79-09 i.
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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 an'd 2 Inspection At: Comanche Peak Steam Electric Station (CPSES), Glen Rose, Texas Inspection Conducted: October 1982 through February 1983 Inspector:
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s
'R. G. Taylor, Senior Resident Inspector-Date Construction
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M/r,hf Approved:
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"T. FV W(sterman, Chief '
Date Reactor Project Section A Insoection Summary Insoection Conducted October 1982 Through February 1983 (Recort 50-445/83-03; 50-446/83-01)
Areas Insoected:
Routine and special inspection, announced by the Senior Resicent Inspector-Construction (SRIC) including facility tours, investigation of allegations, participation and assistance to the Construction Assessment Team Inspection, and other inspection related activities.
The inscection involved 253 inspector-hours by one NRC inspector.
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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Details
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1.
Persons Contacted
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Principal Licensee Personnel l
R. G. Tolson, Site Quality Assurance Supervisor D. N. Chapman, Quality Assurance Manager
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8. R. Clements, Vice-President, Nuclear
,
J. T. Merritt, Manager of Startup J. B. George, Vice President and Project General Manager i
Other Personnel
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G. R. Purdy, Project Quality Assurance Manager, Brown & Root (B&R)
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O. Frankum, Construction Project Manager, B&R
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The SRIC also interviewed other licensee and contractor personnel during
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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 j
that a small amount of concrete placed on the Unit I reactor building dome
!
during a rain storm withcut 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 t
assurance was judged to be required.
The licensee has now completed the
structural acceptance test of the Unit I reactor building with special attention directed to the repair area.
The test was successful and no
anomolies were identified in the repair area and therefore, it is judged i
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 that the segregation of safety and nonsafety wiring in the panels was not in accordance with Regulatory Guide 1.75 but we ; in essential compliance with the panel design displayed by FSAR Figure 8.3-15.
After discus-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.
.
(Close'd) 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 control of the number and extent of repairs
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made to weld joints in stainless steel pipe had become less well con-i 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
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j 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 i
being complied with and therefore, has no further questions.
3.
Action on Licensee Identified Desion/ Construction Deficiencies
1 (Closed) Over-Torquaing of Safety Relief Valves.
On September 10, 1982,
'
the Itcensee informed the SRIC that a potentially reportable condition I
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
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found that the ex essive 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
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of the examination of the valves by the licensee.
The examination did not I
reveal any significant damage had occurred to any of the ' valves that would have prevented the valves from lifting under pressure which would satisfy
the safety function.
Some of the valves may have leaked under operating
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conditions which would be undesireable but not a safety hazard.
The SRIC
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l had no further questions on this matter.
4.
A11ecations 8v Dennis K. Culton I
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on September 16, 1982, Mr. Dennis K. Culton made a Itaited public j
appearance before the Atomic Safety and Licensing Board hearing in the i
matter of TUGCO's application for an operating license for the CPSES.
His
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statement during the appearance appears in the hearing transcript at 5551 through 5555.
In addition, Mr. Culton furnished the Board with a written
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statement which appears in the record at 5556 through 5559.
Based upon a j
review of the record, NRC Region IV determined that there were two areas
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of interest that should be evaluated for their validity and effect of
' safety of construction.
The first area dealt with the potential misuse
)
l 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
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assigned to make the evaluation.
5.
A11ecation Relative to BRHL's
.
l Mr. Culton's concern in this area appears at Tr. 5552 through 5554 and 5557 through 5558.
His concerns can be summari:ed as follows:
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j (a) Based upon limited information, he was directed to generate isometric
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drawings giving support locations.
He states at 5557 that he did not j
feel qualified to do this work in the manner directed.
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j (b),The drawings that he and others in this group generated were released
to the field unapproved and were used by the craft labor personnel to locate and install supports.
j A BRHL is an isometric drawing made from a modified piping installation isocetric drawing to identify the supports on the pipe and to provide i
j locational information at an appropriate point in time.
The drawing
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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.
t
Discussions with various licensee personnel who are familiar with the
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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
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as-build verification program as required by NRC IE Sulletin 79-14.
The very early phase of the work appears to have started at about the same i
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
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understand the type of information that appears on individual support i
drawings.
These drawings, which carry a prefix SRH and an entirely j
different number scheme, provide, in addition to the design details of the supoort, information as to where the support was to have been installed.
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The plan type information is provided by a small square generally with four n9tations indicating building column lines.
Within the square, there
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is usually a dimensional figure in feet and inches from one or more of
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the column lines.
The support elevation information is furnished on the
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sain face of the drawing where the elevation of the pipe and the building
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structure, as appropriate, are shown.
The use of this system requires either a substantial degree of familiarity with the various buildings and
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their column line grids, or ready reference to a set of the architectural
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layout drawings which clearly snow the column line grids.
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i 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
,
i one remaining person still in the original group when Mr. Culton worked
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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 locational information.
The pipe
isometric was reproduced such that information relative to pipe instal-l 1ation was deleted.
This would include deletion of weld joint data and
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i the bill of materials.
The remaining information was the isometric line j
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 symool.
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
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dimensions that still remained on the drawing after modification.
The i
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I early BRHL drawings did not give any dimensional information on the
supports and the first issues w3re stamped " issued for hanger identifica-
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tion and accountability only" in the drawing approval block.
Subsequent
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revisions, apparently beginning in early 1981, were updated and began to i
show dimensions for support locations.
The final versions of the drawings l
provide verified support locations, at which time the individual support drawings are revised to delete the location information.
According to i
i the present supervisor of the central document control center, the BRHL
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drawings have never been routinely distributed to any of the possible j
user organizations such as the support installations crews.
The drawings i
were only available on an individual requisition hasis which would be
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stamped "For Information Only" when given to the rtquisitioner.
The BRHL drawings were originally developed and updated periodically to facilitate
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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
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be modified after The initial as-built verification effort.
This arises
i by reason of the deletion of the support location information on the individual support drawing which then makes the use of the BRHL vital
i in order to find the support in the facility.
This situation only arises
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on a limited basis and is treated on a case basis by the support instal-l 1ation group and the document control center.
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In regard to Mr. Cuiton's two major concerns in this area, the SRIC was able to locate a few of the early BRHL drawings which carry the initials i.
"DKC" in either the draftsman identification block or in the checker block.
A comparison of these drawings to those generated by other drafts-
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men indicate no significant differences.
The SRIC can only conclude that j
Mr. Culton was as competent as the other people in the group.
Given the Il 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 j
that the drawings were released unapproved for use by the construction
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forces has been shown to be incorrect in two different ways.
First, the original issues were provided to the document control center for filing
j with the note " issued for hanger identification and accountability only"
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on face of each drawing and were approved in the appropriate block on
'I the drawing face.
Secondly, the drawings, while on file in the document control center, were never subject to a routine distribution and were not readily availaole to the construction force who in fact had no need for them.
In addition, numerous observations by the SRIC of the support installation process has indicated that the support location information
i on the support drawing was used to install and to inspect the supports t
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 SRHL drawings is thus considered to be refuted.
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6.
A11ecation Relative to the Solicino of Electrical Cables
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At Tr. 5551 through 5552 and 5556 through 5557, Mr. Culton stated that he
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had observed that "Q" electrical cables had been spliced and that these
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splices were in the Unit 1 spread room.
Following Mr. Culton's appearance
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before the Atomic Safety and Licensing Board, Mr. Culton was interviewed in the NRC Region IV offices on November 8, 1982, in an attempt to obtain
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more information on the matter.
The interview was tape recorded 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
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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
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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:
The cables in question are 800 or more feet j
long.
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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 from the aisle.
i 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 Mr. Culton.
The basis of the analysis was as follows:
J The applicant has a computerized listing of all cables for the entire a.
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.
I 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, Of the 42 cables, only 5 were shown by the routing records to be c.
I terminated in a relay panel, more correctly called relay racks.
j 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 "80P 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, i
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The cable pulling records indicate that the two cables were orginally e.
identified as E0009231 and E0009240 which were pulled on January 14 and 15, 1980, respect, ly.
Based on his employment records, these dates coincide with h.. Culton's employment in the construction labor force as an electrician.
f.
Engineering changes subsequent to the pulls changed the designation of cable E0009231 to A0009231 and E0009240 to SP009240.
The change a
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:
The SRIC, with the assistance of two other NRC inspectors, traced a.
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 identi fi ed.
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:
Project Specification ES-100 " Electrical Erection" does not totally a.
prohibit the splicing of safety-related cables as indicated by
Mr. Culton.
The specification allows splicing to be done based upon j
the engineer's direction and this has been done by the use of engineered function 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 the site engineers.
One of two repair measures are applicable within i
the procedure.
One of the methods utilizes heat shrinkaole plastic tubing when the damaged area is not prohibitively far from the end of i
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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 these 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.
The SRIC believes that yet another consideration may well be relevant c.
to this satter.
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 position 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.)
Since 17 feet of each of the identified cables were not inspected by d.
the NRC during the course of this special inspection, it was not possible to conclude positively that tne allegation is either confirmed or refuted.
Notwithstanding, the inability to positively state that the allegation made by Mr. Culton is substantiated or 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.
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(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
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.l (3) A jacket repair made with tape would be very difficult 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 i
unusual event as a splice being made to safety-related cable i
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.
An article apoearing on pag he Fort Wort Star-Telegram dated stated that had made allegations wnicn were subsequently investigated by personnsl of B&R and later
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by personnel of TUGCO.
The article stated that was then charging that these investigations were a " cover-up" to hide safety h
rds at the Comanche Peak nuclear power plant.
The article stated that had been employed at the construction site as a foreman and j
was laid off weeks after he made the allegations.
The article alve attributes three technical type allegations directly to In summary, the technical allegations appearing in the article were:
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(a)
apparently stated when interviewed by the writer of the article that rejected aggregate was mixed with concrete that was
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subsequently poured to form the base for the nuclear reactor.
The gC,M article stated that a M was the B&R equipment operator who had apparent first hand knowledge of the matter.
The article also stated that could not be reached for comment.
(b) A second allegation, that the article stated was never previously investigated, involved the construction of u rwater lamps for the pools surrounding the reactor.
charged that he was prevented from cleaning out dril shavi 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 contention that holes had been improperly drill through concrete walls and the interior reinforcing steel.
The article attributes the information to anothe. party identified as
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The SRIC assigned to the Comanche Peak station obtained both the B&R and
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TUGC0 files pertaining to the investigations that were stated by the i
newpaper article to have occurred.
The B&R file was found to contain an
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undated and unsigned letter addressed to 6 6 of B&R. The letter is indicated in two different places to have been
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prepared by 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 6 by h on August 6, 1982.
The memorandum was dated August 13, 1982.
The undated letter to M contained eight violations that the writer stated he had observed or had knowledge of that had occurred during his
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period of employment at the Comanche Peak station.
Review of the letter addressed t indicated that only one of the eight violations I
correlated with t egations appearing in the newspaper article, this being the item outlinad 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 substantiaMd but the corrective measures were already Iipsa% case, by ' hat is assumed to beMs signature, taken.
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acknowledged his satisfaction with the B&R findings.
The aoove memorandum indicates that a number of other eople were interviewed by the B&R investigative group one of whom wa
apparently did not confi 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). M 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, M tated concerns about a possibly missed hold point during the welding of the fuel pool liner and that some welding had been done by an uncertified welder.
In an internal B&R memorandum dated August 17, 1982, the B&R investigators summarized
@ 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 j
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 allegations, the memorandum states that the allegations were investigated and found to be without basis but provides no other information.
Personnel of TUGC0 performed a separate investigation of 6 allegations (@e M letter) during g of 1982.
The results of 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
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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 investi ators documented a phone call from6 in which apparently made yet additional allegations.
One of these allegations regarded welding done by an uncertified welder on the turbine-generatcr pedestal (by implication).
also apparently further mentioned 6 who wts supposed to know about a sensor that had broken personally driven a front loader twt retur% was alleged to have off and was buried in th.e main dam.
Also, ned dry and lumpy cement that had been rejected to the bin, that this cement had been subsequently used in the reactor core, '.od that this was why the cracks happened.
The writer of the memorandum stated that he had encourge to take his concerns to the NRC.
in turn was reported as saying that he had intended on going to the newspapers and Congress instead.
It appears that 6 arried out his above stated intention in that the above referenced newspaper article has appeared and to the best of SRIC's knowledge, as 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 Mletter and in the telephone conversation with TUGC0 should not.
This decision was based on the premise that has had his earlier concerns satisfied except for those appearing in the article.
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Regarding the above summarized allegation (a), the SRIC established that W as 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 to contact @r by telephone at his new address, to no avail.
A registered lette, receipt requested, was then sent to@ requesting that he contact Region IV as soon as possible.
Receipt of the letter was acknowledged but as of this date, M as not contacted the region.
It appears that does not intend to assist the NRC in investigating allegations attrioused to him.
It should be noted that oniv the B&R investigative group has been able to establish contact with M 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 inside and had not been removed.
It was alta 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
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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 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
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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 have been included in the QA program.
This is considered to be a violation of Appendix B 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 significance of the chips is very small.
The uncontrolled (no QA) problem with 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 Under these circumstances, the NRC has determined that it can address the issue in a more satisfactory manner by d evaluating the second party's allegation rather than 8.
Posting of NRC Form 3 10 CFR 50 was revised by 47 FR 30452 to add 10 CFR 50.7 " Employee Protec-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 viewed by employees on their way to or from their place of work.
It has 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 in 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 office, sometime between Thanksgiving and Christmas and were posted on the 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 variou; federal statutes or regulations.
The posting of an additional form probably would not draw much attention from the average worker.
As of the time of
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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 Ihe $RIC 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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S. t:UCLEAR REGULATORY CC"MISS10t!
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0FFICE OF ItSPECTIO: A!!D ErlFORCE",Et!T
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o.EG10ti IV
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Report tio. 50 445/77-13; 50-445/77-13
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Cocket i:o. 50-445; 50-446 Category A2 Licensee:
Texas Utilities Generating Conpany 2001 Bryan Tower Dallas, Texas 75201 Facility f;are:
Comanche Peak, Uni ts 1 & 2 Inspection at:
Conanche Peak Site, Glen Rose, Texas Inspection conducted:
troverter 28 - December 2, 1077 O
Inspectors:
S p-e.
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R. C. StewartT RIat c.u inspector, PVojects Section Date
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(Paragraphs 1, 2, 3, 4, 9 & 10)
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W. G. Hubacek, Reactor Inspector, Projects Section Date (Paragraphs 5 & 6)
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R. A. Herr. ann, Reactor Inspector, Engineering Support Date Section (Paragraphs 7 & 8)
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["_L. D. Gilbert, Reactor Inspector, Engineering Support Date /
Section (Paragraphs 7 & 8)
Other Accompanying Perso.cnel:
R. E. Hall, Chief. Engineering Support Section (tiovember 30 and Decerter 2, 1977)
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Approved:
/Wh>'- ^
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/z/bo/7.7 W. A. Crossman, Cntef, Projects Section Date
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A R. E. Hall, Cnief, Engineering Support Secticn Date
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Irsoection 5.t.rary Ins:ection en November 28 - December 2.1977 (Recort No. 50-445/77-13;
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50-4-6/77-13)
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Areas Inscected:
Routine, unannounced inspection involving observation of work perforrance and record review of doce liner and fuel pool liner fabrication; follow-on review of safety related piping shop and field fabrication; observation of work performance and record review of the installation of the reactor coolant system component supports, review of the QA pr:gran implecenting procedures for electrical and instrument cables and terr.inations; and indeoendent reviews concerning construction c
deficiencies for which the licensee has submitted reports in accordance with 50.55(e).
The inspection involved one hundred thirty-nine inspector-hours on site by four MP.C inspectors.
Results: No items of nonccmpliance or deviations were identified.
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CETAILS 1.
Persons Contacted Printical Licensee Ecolevees
- J. B. George, TUS!, Nuclear Construction Manager
- 0. N. Cnapran, TUGCO, QA Manager
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- R. G. Tolson, TUGCO Site QA Supervisor
- J. T. Merritt, TUS!, Resident Manager
- C. L. Eiggs, TUGCO, QA lead Engineer R. V. Fleck, TUGC0/G&H, Site QA Supervisor J. V. Hawkins, TUGCO/GaH. Site QA Representative
'D. E. Ceviney, TUGCO, QA Technician Other Personnel
- H. O. Kirkland, B&R, Project General Manager H. C. Dodd, B&R, Project Manager
'U. D. Douglas, B&R, Assistant Project Manager
- J. J. Mcorhead, G&H, Resident Engineer The inspectors also interviewed other centractor employees during the course of the insoection. They included B&R field engineers, B&R QC inspectors and S&R construction personnel.
- denotes those present at the exit interview.
2.
Licensee Action en Previous Irsoection Findings (0 pen) ;oncompliance (50-445/77-10; 50-446/77-10): Failure to Remove Weld Surface Defect Prior to Final Acceptance. The licensee's written response, dated Neverter 17, 1977, di.d not reflect audits and/or sur-veillance activities being teclecented to prevent recurrence of this item. This matter remains cpen cending IE review of supplemental information to be provided by the licensee.
(0 pen) ':encocpliance (50-4a5/77-10; 50-446/77-10):
Failure to Provide Welding Procedures at the Location Where the Prescribed Activity is Pe rfo rme d.
The licensee's written response, dated Noveeber 17, 1977, did not reflect audits and/or surveillance activities being implemented to prevert recurrence of this item.
~his ratter remains open pending IE review of supplemental infor ation to be provided by the licensee.
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(Closed) Unresolved Item (50-445/77-11; 50 a46/77-ll):
Indication of an Uncontrolled Welding Casign Change.
Curing this inspection, the IE inspector reviewed B&R inter-office memo (TSV-0087), dated flovember 30, 1977, which documents the corrective actions initiated to resolve this matter. The inspector had no further questions regarding this item.
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3.
Potential Construction Deficiency - Vendor Sucolied Steel Erbeds
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On i;ovember 23, 1977, the licensee reported by telephone that the site construction staff discovered that "B" series Cadweld sleeves were welded to eight steel plate embedments in reversed orientation.
During this inspection, the IE inspector reviewed the current status of this discrepancy and found that the specific steel enbeds had not been embedded in concrete and corrective measures were initiated; however, due to insufficient information at this time, the question of similar conditions of reversed orientation of "B" series Cadwelds on previously installed embeds can not be answered until an on-going review and evaluation is completed. This matter remains unresolved.
Allecation of Poor Workmanship The licensee informed the NRC, Region IV office on flovember 23,1977, by telephone, of a call on fioverter 22, 1977, from an unidentified woman who was apparently concerned with the workmanship at the site regarding the
- 'g use of "rotofoam" as a temporary spacer being utilized in construction in maintaining the required air space between Category I seismic structures.
During this inspection, the IE inspector reviewed the subject allegation and found that contrary to the woman's belief, all temocrary "rotofoan" blocks have been removed from the subject areas. The E&R QA/QC inspection staff have initiated an inspection and documentation progran to assure that the required 1" gap between Category I seismic structures is being raintained in the as-built condition. This catter will remain coen pending IE review of the QA/QC inspection results.
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5.
Review of OA Manual Provisions for Electrical Construction Activities The inspector reviewed the Brown & Root QA manual to ascertain whether appropriate and adequate procedures were provided to assure that activities related to electrical cables and terminations and electrical components are controlled in accordance with ilRC requirements and licensee comitments.
The follcwing procedures and specifications were reviewed:
ACP-3, "itaterial Receiving Storage and Handling" QCP-1,1, "QC Receiving Inspection" QCP-1.2, "QC Surveillance of Storage, Warehousing and Control" QCP-1.6, "QC Surveillance of itechanical, Electrical and Instrumen-tation Equipment"
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QCI-1.5-ll, " Safety Relatec Mechanical and Electrical Equipment Storage Maintenance" GC:-i.1-11. " Receiving Inspection for TUSI/G&H Procured Safety palated Equipment"
!icp-10. " Storage and Storage Maintenance of Pechanical and
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Electrical Ecuienent" ECp-lC, " Cable Tray and Hangers"
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ECp-19 " Exposed Conduit and Hangers" Specification.':o. 2323-ES-lCO, " Electrical Erection Specification" Specification No. 2323-ES-19, " Cable Tray Specification" The inspector noted that several work and inspection procedures related to electrical construction activities are being developed and will be issued in the future.
These procedures will be reviewed during subsequent inspections.
No itets of noncocpliance or deviations were identified.
6.
Electrical Cable and Ecuioment Storace The inspector observed storage of electrical cable which was stored at the site.
P. eels of electrical cable were stored outdoors on a concrete pad. The inspector noted that several QC tags attached to cable reels were becc.ing faded froc exposure to weather and were difficult to read.
A licensee recresentative stated that new weather-resistant tags were being procured to replace the faded tags.
The inspector also cbserved storage of several items of electrical equip-cent which were located in warehouses.
These items included: three con-taineen spray pt=p totors, one component cooling water pump motor, two safety ir.jection pumps, and two motor operated valves.
The inspector reviewed receiving records for electrical cable and equip-r.ent =aintenance reccrds for one containment spray pump and two motor cperatec valves.
No itecs_ cf noncorpliance or deviations were identified.
7.
Safety F. elated Structures a.
Review of QA Imolecentine Procedures The inspector reviewed the program for the fabrication, erection, welding and inspection of the stainless steel liners for the refueling cavity, transfer canal, spent fuel storage and cask-5-
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loading pits to ascertain if the commitments stated in the PSAR and Gibbs & Hill (G&H) specification 2323-55-18. Rev. 2 were being implemented. The inspector' reviewed Crown & Root (B&R)
construction procedure 35-il95-CCP-38, " Stainless Steel Liner Erections," and B&R QA procedures CP-QCP-2.11. " Inspection of Stainless Steel Pool Liner Systems," and CP-QCI-2.ll-1, " Weld Inspection and Fit-Up of Stainless Steel Liners," to ascertain
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if the above stated requirements had been implemented.
Additional QA and work procedures in the areas of weld expendable material
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control, welder and weld procedure qualification, HDE and welding surveillance were reviewed to assess control of these activities.
No items of noncompliance or deviations were identified.
b.
Cbservation of Work Activities (1) Stainless Steel Liners The welding of fillet joints for the attachment of leak chase -
channels and of tacks for the attachment of backing bars for the butt weld seans for stainless steel liners was inspected.
Weld procedures and welders were found qualified in accordance with the requirements of the ASME B&PV Code,Section IX. The welding was performed in accordance with WPSs 99020 and 88023 and placed as specified by B&R drawing WRB-10559.
Work and inspection activities were performed as prescribed by the procedures discussed in the previous section.
No items of noncompliance or deviations were identified.
(2)
Reactor Coolant System Component Suoports A limited inspection of the Vertical Columns - C1 as shown and described on Westinghouse drawings 1457F29 and 1457F27 was perforned in the site storage yard. The inspector reviewed the PSAR and Westinghouse specification G-952628, Rey, 1, " Fabrication Requirements For the Reactor Coolant System Component Supports," and determined the vertical column fabrication requirements were ASME B&PV Code,Section III, Div.1, NF,1974 edition as a minimum. The inspector was unable to find any documentation in the preliminary data package and certificates of conformance or on the components.that the articles.were fabricated in accordance with AS"E III, NF and that volumetric inspection of the full penetration welds had been performed as prescribed i
by ASt'E III, LF, paragraph NF-5212. The licensee is obtaining
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the comolete data package for these items to determine if the items were fabricated and inspected as prescribed.
This item is considered unresolv:d.
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Safety Related pioino (Weldico)
i The inspector observed the welding in the pipe shop of weld #2, 4"-pipe to fitting, SF-1-1519.-3 per WPS CS023, Rev. 2.
The welders and welding
procedure were qualified in accordance with the ASME B&PV Code, Section i
IX. Wald technique, parameters, gases and expendable materials were as
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prescribed by the WPS.
Inspections were as prescribed by B&R QCP-3.4 as
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noted or. Held Data Card 00893.
The inspector revier:ed the radiographs of welds 2 and 3, 24"-CC-1-AS-12, cceponent cooling line. The radiography was performed in accordance with
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procedure CP-NCEP-101, " Radiographic Examination (Piping)," which c6mplies
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with the requirements of ASME S&PV Code, Sections III and V,197' 5:4 tion
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including Sucrer 1974 Addenda. The inspector reviewed twelvt cc 5 +;1 f
i radiographs and radiographs of repairs as required.
No itets of noncompliance or deviations were identified.
9.
Unresolved items Unresolved items are natters about which more information is required in order to ascertain whether they are acceptable items, items of noncompliance or deviations. The following item was disclosed during this inspection re-garding fabrication and inspection of reactor coolant system component supports:
Identifier Title Reference 77-13-1 Adequacy of the fabrication and Paragraph 7.b.(2)
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inspection of reactor coolant systec component supports
i 10.
Exit Interview The inspectors met with the licensee representatives (denoted in paragraph 1) at the conclusion of the inspection on December 2,1977. The inspectors summarized the purpose and the scope of the inspection and the findings.
The licensee representatives acknowledged the unresolved item (paragraph 7.b.(2)) concerning lack of documentation regarding the fabrication of the
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reactor coolant system component supports.
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U. S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT
REGION IV
Report No. 50-445/80-16; 50-446/80-16 Docket No. 50-445; 50-446 Category A2 Licensee: Texas Utilities Generating Company 2001 Bryan Tower Dallas, Texas 75201 Facility Name: Comanche Peak, Units 1 and 2 Inspection at: Comanche Peak Steam Electric Station, Glen Rose, Texas Inspection conducted: July 7-31, 1980
[/htffo Inspector:
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h. G. Taylor, Resid@t Reactor Inspector
'Datd v
Projects Section
[k/C Approved:
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. A. Crossman, ChMf, Projects Section Date Inspection Summary:
Inspection During July 1980 (Report No. 50-445/80-16; 50-446/80-16)
Areas Inspected: Routine, announced inspection by the Resident Reactor Inspec-tor (RRI) including general site tours; safety-related piping installation and welding; electrical cable installation; protection of major components; and follow up on allegations received. The inspection involved eighty-nine inspec-tor-hours by one NRC inspector.
Results: No items of noncompliance or deviations were identified.
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DETAILS 1.
Persons Contacted Principal Lienesee Employees
- J. B. George, TUSI, Project General Manager
- J. T. Merritt, TUSI, Construction and Engineering Manger
- D. N. Chapman, TUGCO, Quality Assurance Manager
- R. G. Tolson, TUGCO, Site Quality Assurance Supervisor Other Persons F. W. Gettler, Gibbs & Hill, Vice-President for Power Engineering E. Horowitz, Gibbs & Hill, Assistant Chief Mechanical Engineer R. L. Moller, Westinghouse, Site Manager The RRI also interviewed other licensee and Brown & Root employees during the inspection period including both craft labor and QA/QC personnel.
- Denotes those persons with whom the RRI held on-site management meetings during the inspection period.
2.
Site Tours The RRI toured the safety-related plant areas several times weekly during the inspection period to observe the general progress of construction and the practices involved. Eight of the tours were accomplished during por-tions of the second shift where the main construction activity involves the installation of electrical cables and the application of protective coatings.
No items of noncompliance or deviations were identified.
3.
Electrical Installation Activities The RRI made a number of observations of electrical cable pulling opera-tions during the inspection period. The RRI observed the activities of each of the seven cable pulling crews one or more times in order to ascertain whether they were working within the parameters of the site installation procedures and good practices. The RRI also observed the activities of the QA/QC personnri assigned to monitor the pulling activities to evaluate their knc.wledge of requirements and their dili-gence in assuring conformance. The RRI also reviewed the qualification and training files of several randomly selected QA/QC personnel assigned to various electrical installation activities for conformance to the gui-dance provided by Regulatory Guide 1.58 and ANSI N45.2.6, both of which are titled " Qualification of Nuclear Power Plant Inspection, Examination, and
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Testing Personnel." The RRI examined randomly selected segments of installed cable tray for freedom from burrs and other sharp surfaces, identification, and freedom from undesirable debris.
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No items of noncompliance or deviations were identified.
4.
Protection of Major Installed Equipment The RRI observed, during the general plant tours, that the Reactor Pressure Vessels in Unit No. I and 2 were covered and protected in the manner recom-mended by the supplier. The Unit 1 Reactor support structure (internals)
remain in their enclosures within the Reactor Containment Building while those for Unit 2 remain in protective outdoor enclosures. The RRI observed that pump electric prime movers and motors for safety-related valves have protective space heaters powered by temporary wiring or by permanent wiring as evidenced by a hand warm condition. The main control room and the Hot Shutdown Panel are adequately air conditioned to limit the temperature rise due to the recent extremely hot weather from adversely affecting the rela-tively delicate electronic solid state devices and instruments. Other electrical components such as switchgear and motor control centers which need protection from excess humidity until permanently energized are heated above ambient by means of electric lights within the enclosure in accordance with the supplier recommendations and normal industry practice.
No items of noncompliance or deviations were identified.
5.
Special Inspection Relative to Allegations Received The RRI performed a special inspection relative to a series of three allegations received via another NRC inspector assigned at another nuclear power station from an employee of that station who was a former employee of the general contractor at CPSES, Brown & Root. The alleger will here-after be referred to as Individual A for purposes of identification. The allegations as received by the RRI were:
a.
Anchor bolts used in the concrete walls of the safety-related build-ings were pulling out under load. No specific location was given.
b.
There are numerous concrete voids in the building walls that can be ik?
located by sounding the walls with a hammer and listening for a hollow
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sound.
I c.
The Safeguards I panel on the 790' level has loose bus bars and ground wire connections. He had told his supervisor of the problem, but was told to forget the matter.
The RRI reviewed the Brown & Root maintained personnel record jacket relative to Individual A and interviewed his former General Foreman rela-tive to Individual A's activities while employed at CPSES. The following facts were gleaned from the records and the interview:
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Individual A was employed by Brown & Root at CPSES as an electrician for a period of five weeks early in 1980. The records indicate Individual A was terminated due to insubordination.
b.
The interview with the General Foreman indicated that Individual A had worked most, if not all, of his period of employment on the task of installing grounding cable on cable trays with run-offs to equip-ment such as Safeguards Panels. The task was predominately performed on the 790' elevation area of the Unit 1 Safeguards Building. The General Foreman stated that Individual A had been a good reliable worker. The General Foreman also stated that he had terminated Individual A for insubordination after a Brown & Root Personnel Safety Inspector had found Individual A on a high scaffold without his safety belt in violation of published site safety standards. When admonished by the Safety Inspector, Individual A gave the Safety Inspector a "hard time".
The Safety Inspector then contacted Individual A's Fore-man who was also given a "hard time" by Individual A.
The foreman asked the General Foreman to terminate Individual A for failure to follow the site safety rules even when ordered to do so; i.e.,
insubordination.
The RRI's inspection relative to each of the allegations revealed the following facts and conclusions:
Allegation a It is true that the anchor bolts in use at CPSES will on occasion pull out when loaded which is the exact purpose of loading each bolt to a prescribed value, i.e., will the bolt pull out under load. This situation comes about because of the design of the bolt wherein a hole of a prescribed drill bit size is drilled into the concrete, the bolt with its pre-assembled wedges is driven in and then pulled back to set the wedges. On occasion, for one of the several reasons such as not holding the drill steady, the hole in the concrete actually is larger than the drill bit size. When this happens, the bolt may not develop full wedging action and will pull out at some force less than required depending on just how much the hole is oversized. The RRI could find no evidence that the anchor bolts are pulling out after the bolt has achieved its full load as indicated by being torqued to a specified value. When a bolt does pull out because of hole oversizing, the condition can be relatively easily corrected by
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installation of the next larger size bolt.
The allegation has no technical merit.
Allegation b The RRI identified two locations where a hollow sound could be obtained by tapping a wall with a hammer. The co'ndition was identified to the licensee, who found several more in the same general vicinity, all in the walls of the
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Safeguards Building I on the 790' elevation. Each of the five areas rauged from about I to 2 square feet. Each area was marked out and excavated down to the reinforcing steel or about two inches. No voids or other faulty condition was found in this stage of the investigation. The licensee then proceeded to excavate behind the reinforcing steel such that the excavations could be repaired in a sound manner. During several of the excavation, efforts, the RRI was present and observed nothing of an unusual nature.
The RRI also queried the craft personnel doing the work about what might have occurred when the RRI was not present and was informed that all of the excavations revealed nothing except sound concrete. The RRI again tapped the concrete af ter it was excavated to a depth of about four inches and determined that in each instance the hollow sound was no longer present.
The RRI can offer no explanation as to why the hollow sound exists in some localized areas but can assert that the sound is not indicative of actual hollowness or otherwise inferior concrete based on the observations i
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and examinations described above.
The allegation appears to have been based on an inference that a hollow sound indicates a hollow wall which has been shown to be an incorrect
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inference. The allegation is thus without merit.
Allegation c The RRI has inspected the Safeguards Panel referenced in the allegation, i.e., Motor Control Center IEB3-1, and found that the power bus and ground bus connections were tight as of the time of inspection. It appears that the allegrtion might have been representative of the panel's condition early in 1980 when the alleger was employed at CPSES. The panel in ques-tien, however, has been partially turned over to the licensee's test and startup organization for initial energization. The test and startup procedures provide that each power and ground bus bar connection be veri-fied to be tight prior to initial energizing of a panel. Based upon an interview of the cognizant test engineer and upon personal observation by the RRI, all bus connections were tight prior to initial power application to the panel. The purpose of grounding the bus is to assure
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the safety of operations and maintenance personnel in the event that a power
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connection were to become inadvertently grounded. Any loose power bus connections would be revealed during startup testing of the various devices powered from the Motor Control Center either by the device being non-opera-tive or by arcing in the loose connection.
The allegation has no technical merit as of this time.
6.
Pipe Hangers and Supports The RRI continued the inspection of the Class 2 and 3 pipe hangers and supports initiated in June 1980 (see Inspection Report No. 50-445/80-13;
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50-446/80-13) with particular interest being directed toward a design configuration of several supports that were not typical of that experienced during the inspection of other major piping installations. This configur-ation is one wherein the pipe is supported on an essentially line contact
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basis by a simple steel menber such as a box beam or a "v" beam rather than by a shoe or clamp which would have the effect of spreading the load
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over a larger area. Review of ASME,Section III Code, Subsection NB, NC and ND for Class 1, 2 and 3 systems respectively, and Subsection NF for supports, failed to reveal any definitive requirements although Subsection NF seemed to imply that such shoes or clamps should be utilized. The RRI contacted knowledgeable NRC personnel in the areas of stress analysis, fracture mechanics and pipe support design for technical advice on the subject.
The RRI was advised that the design configuration observed had become a relatively common feature in designing supports for piping systems which must be seismically sound. The line contact induced stresses would be reduced by a slight and unharmful deformation of the pipe and the support elements to a point well within the strength characteristics of each element.
The RRI had no further concern in this area.
7.
Safety-Related Piping System Installation The RRI made several observations of the handling practices relative to stainless steel safety-related piping, particularily in regard to that involving the Reactor Coolant piping in Unit 2 which was being adjusted for fit-up preparatory to welding to the Reactor Pressure Vessel and
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Reactor Coolant Pump in No. 3 loop. The actions observed were consistent with applicable site procedures and those of good industry practice.
On or about May 30, 1980, the RRI was informed by the licensee that p4r-sonnel representing the fabricator of the Unit 2 Reactor Pressure Vessel were on the site to conduct a crecial examination of the outlet nozzle safe-ends. The licensee stated that the fabricator had discovered that
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a comparable vessel still in the fabrication shop did not conform to the
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fabrication design drawing requirements in that the safe-end thickness on the outside diameter was less than specified. The reduced thickness was postulated to potentially cause cracking in the factory accomplished Inconel weld of the safe-end to the primary nozzle during the subsequent field weld of the safe-end to the Reactor Coolant pipe, particularily if the field weld stainless steel were to overlap onto the Inconel weld.
It was ascertained that one of the four outlet nozzles was substantially thinner than specified but not so thin that any special welding technique would be required. The vessel fabricator and the supplier (Westinghouse)
further established to the satisfaction of the licensee and the NRC that any cracking of the Inconel weld which might occur would be readily detectable during the normally conducted pre-service baseline inspection
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by means of liquid penetrant examination. Westinghouse was also able to establish that even if the Inconel weld did crack, the resulting crack
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would be such that the welded connection would not fail under either operating or seismic loading conditions or both in combination. The RRI has reviewed the licensee's file on this matter and has had discussions of the matter with cognizant NRC personnel and has no question on the correctness of the licensee's decision that the matter was not formally reportable in accordance with 10 CFR 50.55(e) as indicated in the licensee's letter dated June 26, 1980.
The RRI observed welder BBN during a portion of the welding of joints FW-5 and FW-6 as identified on Isometric drawing WP-1-RB-05 representing line No. 2-RC-1-035-2501R1 in the Reactor Coolant primary pressure boundry.
The welder was verified to be welding within the parameters of welding procedure No. 88021 using weld wire heat No. 746100. The RRI reviewed Weld Data Card and Weld Filler Metal Los in the possession of the welder and verified that assurate data had been recorded. The RRI subsequently verified that the welder, weld procedure and weld filler metal had each been properly qualified and certified in accordance with applicable portions of the ASME Code.
The RRI examined the following radiographs for completed safety Class 1 and 2 welds as denoted by the last number in the designation. The radio-graphs indicated that the welds conform to the requirements of ASME,Section III for the applicable class. The radiographs themselves met the requirements for radiograph quality as required by ASME Section V.
Weld Isometric Line
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W-13 & 31 BRP-RC-1-RB-15 6-RC-1-108-2501R1 FW-4 BRP-SI-1-RB-54 6-SI-1-328-2501R1 FW-2 BRP-SI-1-RB-54 6-SI-1-327-2501R1 FW-6 BRP-SL-1RB-40 10-SI-1-021-2501R1 W-14 BRP-SI-1RB-037 10-SI-1-021-2501R1 FW-2 BRP-CS-2-AB-079 4-CS-2-55-151R2
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W-16 BRP-CS-1-RB-01 2-CS-1-107-2501R1 W-2 BRP-CS-2-RB-037A 2-CS-2-103-2501R1
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W-14 BRP-CS-2-RB-22 3-CS-2-079-2501R1 W-2 & 3 BRP-RC-2-RB-044 6-RC-2-029-2501R1
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FW-7 BRP-CS-1-RB-028 2-CS-1-112-2501R1 W-9 BRP-SI-2-RB-52 6-SI-2-172-2501R1 W-7A BRP-SI-1-RB-16 6-SI-1-101-2501R1 W-12 BRP-CS-2-RB-013 2-CS-2-107-2501R1 FW-1 BRP-RC-1-RB-020 4-RC-1-075-2501R1 FW-11 BRP-RC-1-RB-08 2-RC-1-053-2501R1 The RRI verified that piping lines 12-SI-1-031-151R2, 8-SI-1-323-151R2 and 8-SI-1-324-151R2 as shown on flow diagram 2323-M1-0261 have been installed as required.
In conjunction with this verification, the RRI also examined the piping runs of lines 8-CS-1-567-151R2, 8-CS-1-063-151R2 and valves 1-LCV-112D, 1-LCV-112E and 1-8546 as shown on piping flow diagram 2323-M1-0255.
The RRI examined the training and qualification records of six of the QA/QC personnel assigned to the inspection of piping system fabrication, installation and examination. The personnel were randomly selected by the RRI and represent approximately 17% of the QA/QC force assigned to the activity. The training and qualification requirements were based on SNT-TC-1A as required by the ASME,Section III Code and supplemented by the Guidance of ANSI N45.2.6.
The records indicated an educational / experience
level and training background that meet these requirements.
No items of noncompliance or deviations were identified.
8.
Management Interviews The RRI met with one or more of the persons identified in paragraph 1 on July 7, 8, 9, 10,11, 14,15,16, 21, 25 and 29, 1980, to discuss inspection findings and the licensee's actions and positions.
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INSPECTION PLAN II Inspection Raport No.
50-445/80-16; 50-446/80-16 i
Licensee:
Texas Utilities Generating Company
'
Location:
Glen Rose. Texas
,
I Facility:
Comanche Peak, Units 1 & 2 Type of Licenses:
W, PWR, 1150 MWe
,
Type of Inspection: Resident, routine, announced j
Detes of Inspection: July 1980 Dates of Previous Inspection: June 1980 Inspectors:
RGTaylor SCOPE OF INSPECTION Resident Inspection Program for Construction hbYO-(*yg.' -
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U. S. f UCLEAR REGULATORY C0F111SSION OFFICE OF INSpECTI0fl AND EtiFORCEt'ENT i
REGION IV
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Report No. 50-445/78-01; 50-446/78-01
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Docket No. 50-445; 50-446 Category A2 j,
Licensee: Texas Utilities Generating Cortpany f
2001 Bryan Tower
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Dallas, Texas 75201 Facility f; ate: Comanche Peak, Units 1 & 2 Inspection at: Comanche Peak Site, Glen Rose, Texas Inspection conducted:
January 3-13, 1978 Inspectors:
MM
/
=- --
M ?.. C. Stewart, Reactor Inspector, Projects Section Date
(Paragraphs 1, 2, 3, 4, 5, 6, 7, 10, 11 & 12)
/h k
M. 3. Rosenberg, Reactor Inspector, Engineering Oste '
//
Support Section (January 9-13,1978)
(Paragraphs 8 & 9)
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Approved:
$/M
//.:Ko[78
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W. A. Crossman, Cnief, Projects Section Date
/
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_ /, o R. E. Hall, Cntef Engineering Support Section Ofte
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i i
Insoection Su. nary:
t Insoection en January 3-13, 1978 (Recort tio. 50-445/78-01; 50-446/78-01_)
Areas Insoec ed:
Routine, unannouncec inspection involving a follow-on review anc observation of work activities related to containnent building concrete placerent, er:ed steel support structures; polar crane assembly
,
and installation; and a review of a problem relating to "B" series Cadweld l
>
sleeves being inadvertently installed in reverse orientation.
In addition, i
reviews were conducted of a recent change to the project organization and
,
r.anagerent, and two previously identified items of noncompliance.
The j
inspection involved eighty-five inspector-hours on site by two flRC
inspectors.
Results: f:0 items of noncompliance or deviations were identified.
~
8512o20000 851106 hDE
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