ML20151T382

Insp Repts 50-445/85-07 & 50-446/85-05 on 850401-0621. Violation Noted:Failure to Promptly Correct Identified Problem w/RTE-delta Potential Transformer Tiltout Subassemblies

ML20151T382
Person / Time
Site:	Comanche Peak
Issue date:	01/28/1985
From:	Cummins J, Hunnicutt D, Norman D, Phillips H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20151T370	List: IR 05000401/2006021 ML20151T367 ML20151T374 ML20151T382
References
50-445-85-07, 50-445-85-7, 50-446-85-05, 50-446-85-5, NUDOCS 8602100251
Download: ML20151T382 (32)

See also: IR 05000401/2006021

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APPENDIX B ;

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U[S.JNUCLEARREGULATORYCOMISSION

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NRC. Inspection Report: 50-445/85-07- , Permit: CPPR-126

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/ 50-446/85-05 CPPR-127

Docket: 50-445; 50-446

~ . Applicant: Texas Utilities Electric Company (TVEC)

' Skyway. Tower

'400 North Olive Street

Lock Box 81

Dallas, Texas' 75201

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Facility.-Name: Comanche Peak Steam Electric Station (CPSES)

Units 1 and 2

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Inspection At: Glen Rose Texas

Inspection Conducted: April-1, 1985,.through June 21, 1985

! Inspectors:1 - *

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H. 5. Phillips, Senior Resident Date-

Reactor Inspector Construction

_(pars. 1,.2, 3, 8,-9, 10, 11, 15, 16,

'.17, 18, and 19).

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dp l Is 9x 2A m a d & / 6 /85

J. E. Cunstins, Senior Resident Reactor IIatd

') Inspector Construction (April 1 - May 10, 1985)

(pars. 1,;3, and 19)

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D. E. Norman, Reactor Inspector

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Date

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(pars. 1, 12, 13, 14,:and 19)

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D. M. H~unnicutt, Section Chief ITate

r t Reactor Projects Branch 2

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L(pars. 1; 4, 5, 6, 7, and 19)

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D. M. Hunnicutt, Section Chief,

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Reactor Project.Section B -

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Inspection Summary

Inspection Conducted April 1,-1985, through June 21,1985(Report 50-445/85-07)

, t- Areas Inspected:. Routine eannounced and unannounced inspections of Unit 1

7 which included plant tours and review of plant status, action on previous NRC

f nspection findings (violations / unresolved items), review of documentation for

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site dams, and review of 10 CFR Part 21 and 10 CFR Part 50.55(e) construction

. . ' deficiency status The inspection involved 77 inspector-hours onsite by four

NRC inspectors.

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., Results: Within the areas inspected, five violations were identified: fail-

ure to promptly correct an identified problem with RTE - Delta Potential .

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', Transformer Tiltout Subassemblies, paragraph 3.a.; consnercial non-shrink * grout x

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l . was used to grout the Unit I reactor coolant pump and steam generator supports

' ' in lieu.of ' Class "E" concrete, paragraph 3.b.; hydrogen recombiners out-of +

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specification voltage. recorded on quality release document but QC receipt

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inspector accepted, paragraph 3.c; failure to provide objective evidence to

1 show that central and truck mixer blades were inspected, paragraph 8; and '

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. failure to' issue.a deficiency report on cement scales that were out-of-calibra-

x tion, paragraph 9.c.

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-Inspection Conducted April 1, 1985, through June 21, 1985 (Report 446/85-05). ' (

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Areas' Inspected: Routine, announced and unannounced inspections of= Unit'2 ,

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which. included plant-tours and review of plant status, action on previous NRC '

inspection findings (violations / unresolved items), review of documentation for "

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site dams, review of documentation for voids behind the ~ stainless steel cavity

, liner;of reactor building, observation'of NDE on liner plates, inspection of

concrete batch plant,: review of calibration laboratory records for batch plant, , ,

review of concrete laboratory testing, in pection of level C and D storage, '

review of reactor pressure . vessel (RPV)'and piping records / completed work, and- ,

review of 10 CFR Part 21 and 10 CFR Part 50.55(e) construction deficiency-

a status, and review of violation and unresolved items status. The inspection

involved 335 inspector-hours onsite by four NRC inspectors.

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Results: Within the sixteen arev inspected three violations were identified:

failure to correct RTE-Delta transformer problem. paragraph 3.a; failure to

provide objective evidence to show that concrete central and truck mixer blades

were inspected, paragraph 8; and failure to issue a deficiency report on cement

scales that were out-of-calibration, paragraph 9c.

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DETAILS

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,, 1. Persons Contacted ,

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4I ' Applicant Personnel.

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M. McBay, Unit 2 Reactor Building' Manager ,

B.. Ward, General Superintendent, Civil .

' D. Chandler. QA/QC Civil Inspector

1 W. Cromeans QA/QC, .TUGC0 Laboratory / Civil Supervisor .

*fJ. Merritt, Assistant Project-General Manager

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• fP.;Halstead, Construction Site QA Manager '

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1. C. Welch,QASupervisorTUGC0(Construction)

'J. Walters, TUGC0 Mechanical Engineer

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K. Noman,' TUGC0 Hechanical Lngineer

J. Hite, B&R Materials Engineer

G. Purdy, BAR CPSES QA Manager

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• Denotes those present at May 10, 1985 exit interview.

1. Denotes those present at June 10, 1985 exit interview,

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The NRC inspectors also interviewed other applicant employees during this

inspection period.

2. Plant Status

Unit 1,

At the time of this inspection, construction of Unit I was 99 percent

complete. The fuel loading date for Unit 1 is pending the results of l

ongoing NRC reviews.

Unit 2

- At the time.of this inspection, construction of Unit 2 was approximately

74 percent complete. Fuel loading is scheduled for approximately 18

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months atter Unit 1 fuel loading. ,

3. Applicant Action on Previous NRC Inspection Findings s

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a. (Closed) Unresolved Item 445/8440-02: Potential Problem with .

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Potential Transformer Tiltout Subassemblies.

By letter dated June 15, 1983 Transamerica Delaval notified the

applicant of an RTE - Delta 10 CFR Part 21 report to the NRC

reporting a potential problem with the primary-disconnect clips of

the potential transformer tiltout assembly used in the emergency

diesel generator control panels at CPSES. ~ The Transamerica Delaval

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letter also provided instructions for correcting the problem.

However, the NRC inspector could not determine if the problem had

been corrected at CPSES and made this an unresolved iten. The

applicant determined that.the problem had not been corrected and

subsequently performed the recommended corrective action. The Unit I

corrective action work activities were documented on startup work

permits Z-2912 (train A) and Z-2914 (train B). The Unit 2 work

activities are being tracked as master data base (MDB) item 3003-31.

The failure to promptly correct this identified problem is an

apparent violation l445/8507-01; 446/8505-01). .

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, b. (Closed) Unresolved Item 445/8416-03: Commercial Grout Used in Lieu

of Class "E", Concrete

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The applicant determined that the use of nonshrink comercial grout

in lieu of the Class "E" concrete specified on drawing 2323-51-0550

was acceptable. Design Change Authorization 21179 was issued to

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drawing 2323-51-0550 accepting the use of the corrercial non-shrink

grout. However, the failure to grout with Class "E" concrete as

specified on the drawing at the time the work was' accomplished is an

apparent violation (445/8507-02).

c. (Closed) Unresolved Item 445/8416-04: Hydrogen Recombiners -

+ Out-of-Specification Voltage Recorded on Westinghouse Quality ,

Release Document

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Quality Release N-41424 was revised by Westinghouse changing the

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specified voltage from 10+-2V to 12+-2V which put the questionable

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voltage within specification linits. However, the failure of receipt

. . inspection to verify that the QRN-41424 was filled out accurately as

required by' Procedure QI-QAP7.2-8 is an apparent violation

.(445/8507-03).

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c d.. (0 pen 1 Unresolved Iten 445/8432-06; 446/8411-06; Lobbin Report

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, Described Site Surveillance Program Weaknesses

During this reporting period the NRC inspector reviewed the status of  ;

this open item several times and interviewed TUEC management and site '

surveillance personnel. The Lobbin report stated that the scope and

objectives of the site surveillance program were unclear, lacking

both purpose and direction.

There is no specific regulatory requirement to have a surveillance

program; however, TUEC committed to have a surveillance program and

1as established procedures to implement such a program as a part of

the 10 CFR Part 50, Appendix B, QA program. This extra effort is a

strength; however, the NRC inspector also observed, as did the Lobbin L

Report, that the surveillance progran lacks both purpose and

direction to be effective and complimentary to the audit and

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inspection programs. Since the TUEC audit group is not located on ,

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site, the TUEC surveillance program on site takes on added

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significance. , ,

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'This item was discussed with the TUEC site QC manager who described a

reorganized site surveillance function and changes that have

1 occurred. New procedures which describe this organization's duties

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and responsibilities are forthcoming.- ,

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TUEC has elected to defer responding'to the violations pertaining to

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the audit function in NRC Inspection Report 445/84-32; 446/84-11, but

rather to ha've the Comanche Peak Response Team (CPRT) respond to this

report and other QA-matters. .The surveillance issue is closely tied '

to the audit deficiencies in NRC Inspection Report No. 445/84-32;

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' 446/84-11. This item will remain open pending the review and imple-

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< mentation of the CPRT action plan. A special point of interest will

be how audits and surveillance work together to evaluate the control

-of all safety-related activities on site to assure quality,

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especially the overview of quality control effectiveness.

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4. Document Inspection of Site Dams ,

The NRC inspector reviewed documents describing the inspection activities

p(erformed

SSI) for im>oundingon thecooling

Squaw Creek

water for theDam (SCD)

two' units and the

at CPSES. Thesafe shutdown i

purpose of tie SCD is to impound a cooling lake for CPSES. A secondary

reservoir (SSI) is formed by a channel connecting the SCD impoundment to

the SSI.

Three documented ins'pections have been perfonned since 1980. The

inspections were:

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a. Relevant data for SCD is contained in Phase ! Inspection', National

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Dam Safety Program, Squaw Creek' Dam..Somervell County, Texas, Brazos

River Basin, inspection by Texas Department of Water Resources. Date

of Inspection

June 10, 1980.

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=b. Inspection on August 25, 1982,-bj registered professional engineers

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from Mason-Johnston & Associates, Inc., and Freese & Nichols, Inc..

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c. Inspection on September 19, 1984, by a registered professional

engineer from Mason-Johnston & Associates Inc.

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l The ins >ection activities consisted of visual inspections by inspection

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teams t1at included accompanying Texas Utilities Service, Inc. (TUSI),

!: andTexasUtilitiesGeneratingCompany(TUGCO) representatives.

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Photographs were taken as a part of the documentation. The data for the

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piezometer. observations and the data for the surface reference monuments

'were reviewed by applicant personnel and Mason-Johnston engineers.

No items of significance were observed or reported by these inspection

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teams. Slight erosion areas were observed and reported. A cracked area

on the service spillway upstream right bridge seat was observed by the

inspection teams and continued monitoring of this area was recommended by

Mason-Johnston and Associates. No signs of cracks, .;ettlements, or

horizontal movement at any location within the SCD or the SSI were

reported.

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The NRC inspector reviewed the applicant's records and the Mason-Johnston

inspection reports. These documents indicated that the SCD and SSI were

structurally stable and that the applicant was performing inspection

activities to maintain the structural integrity of these dams.

The state of Texas requires periodic inspections of these dams

(principally the SCD) due to inhabited dwellings downstream. The

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applicant has met these inspection requirements.

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No violations or deviations were identified.

5. Voids Behind the Stainless Steel CaviA1.iner in Unit 2 Reactor Buildina

In review of previous related TRT concerns, the NRC inspector reviewed

applicant records, including NCR C-82-01202; NCR C-1784, Rev.1; NCR

C-1784,

Rev.1; NCR Rev.C-1824,

2; NCRRev.

C-1766, Rev.1; NCR

2; Significant C 1791,

Deficiency Rev.1;Report

Analysis NCR C-1824,(SDAR)

- 26, dated December 12,1979; DCA-20856; and Gibbs and Hill Specification

2323-55-18. The review of records and documentation and discussions with

various applicant personnel indicated the following:

Structural concrete was placed in Unit 2 reactor building at

elevation B19 feet 6-3/4 inches to 846 feet 6 inches on June 21,

1979. This concrete was placed adjacent to the stainless steel liner

wall s. The concrete forms for this pour were not removed until

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October 1979 due to subsequent concrete placements for the walls to

elevation 860 feet 0 inches. When the forms were removed, honeycombs

and- voids were observed by appi tcant personnel . The applicant's

review of the extent of unconsolidated concrete resulted in the

issuance of SDAR-26 on December 12, 1979. Investigations were begun

and Meunow and Associates (It&A) of Charlotte, North Carolina, were

contracted to perform nondestructive testing on in-place concrete.

M&A performed these tests on a two foot grid pattern on the

compartment and liner sides of all four steam generator (SG)

compartment walls. The selectcd test locations did not include the

locations where the voids were later found to be located.

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In August 1982, preparations were made to pour the concrete annulus

around the reactor vessel. When the expanded metal formwork was

removed from the reactor side of the compartment walls, voids were

observed and NCR C-82-01202 was prepared. DCA 20856 was prepared as

a procedure to repair the void area. DCA 20856 indicated that the

voids were not extensive (a surface area of about 28 square feet by 8

inches maximum depth) and that the repair procedure assured that the

total extent of voids had been identified. One half (0.5) of a cubic

yard of concrete was used to complete the repairs as indicated on

grout pour card 261.

The applicant's review and evaluation of the gird pattern and a

comparison of SG compartments 2 and 3 to 1 and 4 indicated that voids

did not exist in SG compartments 2 and 3. The review of tv.st girds

extended down to elevation 834 feet, which is the floor elevation of

the liner. The liner walls of SG compartments 1 and 4 were not

tested at elevation 834 feet, but at elevation 836 feet which is

above the area of the identified voids. No testing was done on the

liner side of the area of the voids below elevation 836 feet. The

program also included removal of 2 inch x 2 inch plugs from the

stainless steel liner at locations where test indications raised

questions concerning the concrete. The inspections of the concrete

by applicant personnel after the plugs were removed confirmed that

there were no additional unconsolidated concrete areas (voids).

In accordance with OCA 20856, the applicant removed stainless steel

liner plates from three areas (one area about 1 foot by 1 1/2 feet

and two areas about 3 feet by 1 foot, excavated or chipped to sound

concrete, and cleaned the concrete surface area. One and one-quarter

inch (1 1/4) diameter probe holes and grout access holes were drilled

in the liner plates to determine the extent of and to assure full

definition of the void area. Air access holes were drilled in the

stainless steel liner plates to assure that grouting would be

accomplished in accordance with the procedure.

The procedure (DCA-20856) specifed that the grout was to be cured for

28 days or until the grout reached a compressive strength of 4000

psi. Repairs to the liner plates were specified in DCA-20856 and G5H

Procedure 2323-SS-18.

OCA-20856 required that under no circumstances was cutting of the

liner across weld seams, across embedded weld plates, or into leak

chase seal welds or drilling.through the liner at leak chase

channels, embeds, or weld seams permitted. Documentation review-

indicated that DCA-20856 was adhered to and that no cutting or

drilling occurred in prohibited locations.

No violations or deviations were identified.

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- 6. Nondestructive Testing Observations of Liner Plates in Fuel Transfer Canal

- The NRC inspector observed portions of non-Q liquid penetrant examinations

. (PT) being performed on liner plate welds following re-installation of the

. liner plates in the areas of the fuel transfer canal removed for

. inspection and repair of the concrete. The inspector. performed the PT on

the welds as required by the. repair package and the procedure

(QI-QP-11.18-1, " Liquid Penetrant Examination"). Scattered weld porosity

was. identified by the. inspection. 'The porosity was ground out'and a

repeat PT was performed. The final. inspection is scheduled to be

performed by QC inspection personnel. Thesliner plate areas to be

Linspected by PT were identified in DCA 20856.~

No violations or deviations were identified.

7. Cadweld Splice Observations and Records

a. -Calibration of Tensile Tester

The NRC inspector observed the calibration of the Tinus-Olson

Universal Testing Machine (Model Number 600-12 Identification Number

M&TE-784) on April 2 and May 7, 1985. The machine was calibrated'

just prior to performing tensile testing of cadweld splices and

subsequent to completion of tensile testing each day that tensile

testing was performed. The machine calibration.date for April 2

1985, prior to start of tensile testing was observed by the NRC

inspector and recorded as follows:

Nominal load Calibration Reading Error Error Remarks

-(1bs) (1bs) (1bs)  %

0 0 0 0 0 machine en

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4/2/85

100,000 99,750 +250 +0.25

.200,000- 199,600 +400 +0.2

300,00 299,450 +550 +0.18

350,000 350,300 -300 -0.08

400,000 401,200 -1200 -0.03

500,000 501,350 -1350 -0.27

600,000 602,450 -2450 -0.40

The NRC inspector reviewed calibration data for March 4, March 8

April 2, April 3. April 30, and May 7, 1985. All calibration data

met within the +/- 1% accuracy requirement specified by Calibration

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Procedure 35-1195-IEI-37, Revision 3, dated March 11, 1982. The

reference standards were identified as follows:

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Calibration Due Date ,

RS-75 ? 'BLH Electronics . January 27, 1987 -

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1 - y 'RS-75.3 .BLH. Electronics January 27,~1987

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b.- Observation of Cadweld Splice-Tensile ^ Testing

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! .-(1) LQ ualification Tensile Testing

• On April 2,1985, the NRC inspector observed the following

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' tensile testing of cadweld splices for cadwelder qualification:-

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.y EBD Q8, GBH Q1,' GBH Q2, GBV Q1, BFD Q4, BFD Q3,: BFH Q4, GAH Q1,

GAV Q1,-and GBV Q2..

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Each of the above qualification cadweld splices was tensile

. tested'to 400,000 pounds (100,000. psi) and met the requirements

stated in.the procedure..

b- (2) Production Tensile Testing

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c The NRC inspector observed the ~ tensile' tester calibrations and

,the following production cadweld splices tensile testing on

May 7. 1985: FXD 3P ,FYD 4P, FYO 8P, FRD 87P, and FUD 6P.

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' Each of the above production cadweld splices was tested to .'

400,000 pounds (100,000 psi)and met the requirements stated in 4

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, -the procedure.

(3) Installation of Production Cadweld Splices  ;

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The NRC inspector observed installation'of rebar. and cadweld

splices at frequent intervals (five or more observations per

week during the weeks of April 8 and 15; May 6,13, 20, and'27;

-and June 3,1985). The rebar installation for the Unit 2-

closure was performed in the area identified as elevation 805

feet to elevation 875 feet and azimuth 300. degrees to 335

degrees. .The' installation activities observed included rebar

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spacing, location of cadwelds, observation of ' selection and

removal for testing of cadweld splices for testing, and

determination-of location of rebars and cadwelds for the

as-buil t .drawi ngs. .

, - (4)'DocumentationReviewed

Ihe NRC inspector reviewed the folicwing documentation for the -

y . rebar placement and cadwelding for the Unit 2 containment

-(reactor building) closure area:

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Drawings DCAs NCRs

2323-S-0785, Rev.7 22616, Rev. 1 C85-200294

2323-S-0786, Rev.9 22728 C85-200339, Rev.1

2323-S1-500, Rev.5 22737 C85-200355, Rev.1

2323-S1-506, Rev.5 22836

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2323-S2-505, Rev.5 22878 (Sheets 1-7)

, 2323-S2-508, Rev.2 22772

2323-52-506, Rev.3

No violations or deviations were identified.

8. Concrete Batch Plant Inspection, Unit I and 2

The NRC inspector inspected the concrete production facilities for the

following specific characteristics for the following areas: (1) material

storage and handling of cement, aggregate, water and admixture, (2) batching

equipment scales, weighing systems, admixture dispenser, and recorders,

(3) central mixer (not applicable because it had been- dismantled),

(4), ticketing system, and (5) delivery system.

The current batching is a manual operation since almost all concrete has

been placed. The central mixer was dismantled and removed from site two

or three years ago when concrete placement was virtually completed.

Presently, the backup batch plant (which was a backup system for the

central mixer) is.in operation to complete the remaining concrete

placements. This batch plant is in good condition and complied with the

subject checklist except for one area.

The NRC inspector inspected the inside of one of three trucks used for

mixing concrete (that is, the batch plant dispenses the correct weight of

materials as required by the specific design mix numbers and the truck

then mixes the batch to be placed.) The blades inside the truck are

subject to wear and should _be checked at a reasonable frequency. The

Brown & Root (B&R) representative responsible for checking the blades in

accordance with B&R Procedure 35-1195-CCP-10, Revision 5, dated

December 4,1978, was asked _for evidence that the blades had been checked

for wear on a quarterly basis as required by procedures and it was found

that there was no record of such checks dating back to 1977 when they were

initially checked.

In the FSAR Volume V, Section 3.8.1.2.3, the applicant commits to

ACI 304-73. In- ACI 304, the maintenance of mixer blades is required.

Procedure CCP-10, paragraph 3.10 " Truck Mixing " is silent on blade wear

but Section 3.11 infers that the blades should be checked for both central

and truck mixing. The inspection of both central and truck mixing blades

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was not documented, although the B&R representative stated that the mixing

blades were periodically inspected and laboratory testing would have

probably indicated if there was a problem with the mixing blades.

Strength and uniformity tests have consistently been within the acceptable

range indicating that concrete production was acceptable even though

mixing blade inspection was not documented.

Otherwise, the condition of the inside of the truck was satisfactory as

the drum and charging / discharging were clean. The water gage and drum

counter were in good condition.

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This failure to follow procedures is a violation of 10 CFR 50, Appendix B,

' Criterion V. Subsequent to the identification of this violation, the

blades were checked for wear and blade wear was presently within allowable

limits (445/8507-04; 446/8505-02).

No other violations or deviations were identified.

9. Calibration Laboratory for Batch Plant Unit 1 and 2

The NRC inspector obtained batch plant scale numbers from tags which

indicated ~ that the scales had been calibrated and were within the

calibration frequency. Cement (MTE 779), Water (MTE 766), admixture scale

(MTE 764), and aggregate (MTE 780) were reviewed. The scales had been

periodically calibrated since the batch plant was activated. The records

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were adequate except as follows:

a. Scales MTE 766 records do not differentiate between the

required accuracy of the scale and the digital readout.

b. Scales MTE 779 and 780 records show various accuracy ranges for the

same scale; i.e., MTE 779 (SN749687) records the following: report

dated January 1976 gives 1%; report dated July 1976 gives 1% while

'the report dated October 1976 gives +/- 0.2%.

The calibration appeared to be proper, however, the above items are unre-

solved pending further review of the applicant's actions regarding the

correction of these records (445/8507-05; 446/8505-03).

c. Records for scales MTE 779 records contained B&R memo IM-1108 dated

July 16,1975, which described a nonconforming condition. This condi-

tion affected the water and cement scales causing a 24-48 pound deviation

(7,000 pound scale) during the calibration test. The memo stated that

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the condition was corrected and the scales were then calibrated; however,

no deficiency report was written as required by B&R Procedure

CP-QCP-1.3, " Tool and Equipment Calibration and Tool Control" dated

July 14,1975, and CP-QAP-15.1, " Field Control of Honconforming

Items," dated July 14, 1975. As a result there is no evidence that

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g7 , ' production was adversely affected.

This failure to assure that a nonconforming condition was evaluated

is a-violation of Criterion XV of 10 CFR Part 50, Appendix ~ B,

(44G/8507-06; 446/8505-04).

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, . 10. Concrete Laboratory Testing Units 1, and 2

TUGC0 Procedure QI-QP-11.1-1, Revision 6, was compared with ASME

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Section III, Division 2, Subsections 5222, 5223 and 5224 to assure that

>~ .eac -h ASTM testing requirement was incorporated into the procedure.

The NRC -inspector inspected the testing laboratory equipment and found'the.

test area and equipment were in good condition and each piece of equipment

. :was tagged with a calibration sticker which showed it to be within the .

required calibration frequency. Test personnel were knowledgeable of test

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requirements and equipment.

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The NRC inspector witnessed field tests performed by laboratory personnel

as follows:

Date Truck No. tiix No. Ticket No. Air Content (%) Slump'(in.) *

Temp ('F)

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6/3/85 RT-41 925 .64013 Req 8.2-10.3 NA 70 max

Mea 8.7-9.1 NA 57

6/3/85 RT-35 128. 64014 Req 5.0-7.0 5 max 70 max

lica 6.6 6.25* 57

• Initial slump was high; however, after additional truck rotations the

slump was found acceptable.

The following laboratory equipment was checked and found to be within

calibration: Forney Compression Tester, MTE 3031; Temperature Recorder

MTE 3013a ' nd 3014; Unit Volume Scale, lite 1053; Pressure Meters lite 3000B,

3002 and 3004; Sieves MTE 1286, 1239, 1272, 1274, 1136A, 1156, 1094, 1093,

1095,1178,1179,1300 and 1180; Aggregate scales, MET 1058 and 1067; and

2" grout mold MTE 1111.

The following test records for placement number 201-5805-034 were

reviewed: (1) concrete placement inspection, (2) concrete placement

summary and, (3) unit weight of fresh concrete.

No violations or deviations were identified.

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11. LInspection of Level C and D Storage Unit 1 and 2'

The NRC-inspector inspected all laydown areas where piping, electrical

conduit, cable, and structural reinforcing steel were stored. These.

materials were neatly stored outside on cribbing in well drained areas ,

.. which allowed air circulation and avoided trapping water. This met the

, Level "D" storage requirements of ANSI N45.2.2.

The electrical warehouse contained miscellaneous electrical hardware.

.This building was required to be fire resistant, weathertight, and well

ventilated in order to meet Level "C" storage requirements. This

warehouse was well kept and met all requirements except for a lock storage

area located upstairs at the rear of this building (electrical termination

tool room). Two minor problems were identified and the warehouse

personnel initiated action to correct them.

The first problem noted was that a box of nuclear grade cement was marked

" shelf life out of date" but it had no hold tag. The box was subsequently

tagged inaccordance with TUGCo nonconformance Procedure CP-QAP-16.1,

Revision 24 (Nonconformance Report (NCR) E85-200453) after being identified

by the NRC. During discussions with the warehouseman, the NRC determined

that engineering told the warehouseman to mark the material and lock it up,

but did not tell him to apply an NCR or hold tag. Also, the NRC inspector

noted a very small leak in the roof above the electrical termination tool

room.. This leak was in an area that did not expose hardware to moisture.

The roof is currently being repaired.

The millwright warehouse storage area was inspected; however, only a small

number of items or materials were stored in this area. The overall

storage conditions in this area met or exceeded Level "C" storage

requirements.

No violations or deviations were identified.

12. Reactor Pressure Vessel and Internals Installation - Unit 2

This inspection was performed by an NRC inspector to verify final

placement of the reactor pressure vessel (RPV) and internals by examining

the completed installation and inspection records.

a. Requirements for Placement of RPV

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Requirements for placement of the RPV to ensure proper fit-up of all

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other major NSSS equipment are in Westinghouse Nuclear Services

Division (WNSD) " Procedure for Setting of Major NSSS Corrponents",

l' Revision 2, dated February 13, 1979, and " General Reactor Vessel

- Setting Procedure" Revision 2, dated August 30, 1974. The NRC

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inspector reviewed the following drawings, which were referenced in

the RPV operation traveler, to verify implementation of WNSD

recommendations: ,,

o WNSD drawing 1210E59 " Standard - Loop Plant RV Support Hardware

Details and Assembly"

o WNSD drawing 1457F27 " Comanche Peak SES RCS Equipment Supports -

Reactor Yessel Supports"

o CE drawing 10773-171-004 " General' Arrangement Elevation"

o CE drawing 10773-171-005 " General Arrangement Plan"

Neither site prepared installation drawings nor specifications -(which

implemented the WNSD recommended procedures) were available and the

drawings examined did not show certain specific installation

criterion such as centering tolerances, levelness tolerances and

clearance between support brackets and support shoes.

The inspector considers this matter unitsolved. (446/8505-05)

b. Document Review

The NRC inspector reviewed B&R Construction and Operation Traveler

No. HE79-248-5500 which described the field. instructions for

installation of the Unit 2 RPV. Requirements recommended by WNSD

procedures were implemented in the traveler. Worksheets attached to

the traveler showed the RPV to be centered and leveled within the

established tolerances. Traveler operation 19. required verification

-of a 0.020 to 0.005 inch clearance between the support bracket and

support shoe, after, applying the shim plates.' Change 5 subsequently

changed the clearance to a 0.015 to .025 inch clearance. The

installation data reflected in attachment 3B of the' traveler ,

indicated an as-built clearance of 0.012 to 0.026 inch which~ exceeds

both the original and revised tolerances. 'This. condition was'

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accepted on the traveler based on Westinghouse concurrence, and there

was no documented engineering evaluation ensite justifying the final

?- tolerances. This matter is considered unresolved pending documentation

validating the final installation tolerances.- (446/8505-06)

The NRC inspector reviewed the following receiving records for the,

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RPV hardware and found them to be in order:

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o Report No.14322 for 54 each closure studs, closure nuts, and

closure washers

o Report No. 09507 for vessel S/N 11713, Closure Head 11713 and 26

0-Rings

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o Deviation notices and corrective action statements

The NRC inspector reviewed the following completed travelers for

internals installation and found them to be satisfactory:

o ME-84-4641-5500, " Assemble Upper Internals"

o ME-84-4503-4000, " Install and Adjust Roto Locks"

o ME-81-2145-5500, "Retorque UI Column Extension"

o RI-80-385-5500, " Transport and Install Lower Internals"

o ME-84-4617-5500, " Repair Lower Internals"

o ME-84-4640-5500, " Assemble Lower Internals"

c. Visual Inspection

At this time, visual inspection of the internals by the NRC inspector

was not possible, and inspection was limited on the vessel placement

to a walk-around beneath the vessel to inspect the azimuth markings

and for construction debris between the vessel and cavity. No

problems were identified in this area.

d. Records of QA Audits or Surveillances

The NRC inspector r 'uested TUGC0 QA audits or surveillances

performed by TUGC0 u the Unit 2 RPV installation. TUGC0 did not

make available any documentation of an audit or surveillance which

evaluated specified placement criteria, placement procedures,

hardware placement, or as-built records. This item is unresolved

pending a more comprehensive review of these activities

(446/8505-07).

No deviations were identified; however, two unresolved items were

identified and are described in the above paragraphs. (11.aandd)

13. Reactor- Yessel Misorientation

On February 20, 1979, the applicant reported to the NRC Resident Inspector

that a design error had resulted in the reactor support structures being

placed in the wrong position on the reactor support pedestal such that the

reactor would be out of position by 45 degrees. Initially Unit 2 was to

be a mirror image of Unit 1, however, a design change was initiated to

permit identical components for both units. The design change was

implemented for the reactor vessel, but not for the pedestal support

locations. The problem was not considered by the applicant to be

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. reportable under provisions of 10 CFR Part 50.55(e) since the error could

not have gone undetected.

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The deficiency was reported to the NRC Office of Inspection and Enforce-

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ment on February 22, 1979 and during a March 27, 1979 meeting in Dethesda,

Maryland, the applicant presented the proposed redesign and rework proce-

dures for relocating the pedestal supports. No unresolved safety concerns

with the repair were identified at the meeting.

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During this inspection the NRC inspector reviewed various documentation

relative to the misorientation problem, including design changes and the

construction traveler which implemented the repair.

The following documents were reviewed;

o NRC Inspection Reports 50-446/79-03; 50-446/79-07; 50-446/79-13

o TUSI Conference :lemo, dated March 1, 1979 H. C. Schmidt to S.

Burwell (NRC Licensing PM)

o .TUGC0 letter TXX-2980, dated April 30, 1979, to W. C. Seidle

o NRC letter to TUGC0 dated May 29, 1979

o DCA 3872, Revision 1, dated February 28, 1979, Subject: Rework of

Structure for Placement of the RPV Support Shoes

o DCA 4122, dated March 22, 1979, Subject: Replacement of Rebar for

RPV Supports

o Construction Traveler CE79-018-5505, dated Itarch 14, 1979, Subject:

Rework of Reactor No. 2 Cavity - New RPV Support Locations

o Grout Replacement Cards No. 007, 008, 009, 010, 014, and 015, various

dates, Subject: Replacement of Grout around Rebar for Repair.of RPV

Support Shoes

o Various Inspection Reports for Grout Properties and Application for

RPV Support Shoes

!- No violations or deviations were identified.

14. Reactor Coolant Pressure Boundary (RCPB) Systems

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The inspection was performed to verify: the applicants system for

preparing, reviewing, and maintaining records for the RCPB piping and

components; that selected records reflected compliance with NRC

requirements and SAR comitments for manufacture, test and installation of ,

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. items; and as-built hardware was a'dequately marked and~ traceable to

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records. The following items were randomly selected and inspected:

.o , a. Pressurizer Safety Valve - This item was . inspected to the comitment ,'

stated in FSAR, Table 5.2-1 which includes ASME Section III, 1971

Edition through Winter 1972 Addenda. Valve S/N N56964-00-007, which

is installed in the B position, was inspected. The following records

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were reviewed:

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o ~QA Receiving Inspection Report No. 21211

o Code Data Report Form NV-1 .

I o Valve Body Certified Material Test Reports -(CHTRs)

The valve was in place, however, installation had not been completed;

therefore, the hardware installation inspection consisted of

verifying that the item was. traceable to the records.

b. CVCS Spool Piece 301 - Requirements for this item are stated in ASME,

Section III,1974 Edition through Sumer 1974 Addenda, which is the

comitment from the FSAR, Table 5.2-1. The item was field fabricated

from bulk piping and purchased elbows and installed in the CVCS with

field welds number 1 and 6 (ref. BRP-CS-2-RB-076). The following

records were reviewed:

o- B&R Code Data Report

, o Field Weld Data Card

o .NDE Reports

o QA Receiving Reports for piping and elbows

o CMTRs

The installed spool piece was inspected for weld quality and to

verify that marking and traceability requirements had been met. The

item had been' marked with the spool piece number (3Q1).and the B&R

drawing number which provided traceability to the material

certifications.

c. Loop 3'RC Cold Leg - Requirements for this item are stated in ASME,-

Section. III,1974 Edition through Sumer 1974 Addenda, which is the

comitment from the FSAR, Table 5.2-1. This piping subassembly

'C consists of a 27.5 inch cast pipe with a 22 degree elbow on the

reactor end, a 10 inch 45 degree nozzle, a 3 inch nozzle,-and three 2

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1/2 inch thermowell installation bosses. 'The following records' were

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reviewed for the subassembly:

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y o- QA Receiving Inspection Report No. 12389

o Westinghouse Quality Release (QRN 47523)

o- ' Code Data Report Fonn NPP-1

0 27 1/2 inch line CMTR

o 3 inch nozzle CMTR

o Field Weld Data Cards

o NDE Reports

(1) Sandusky Foundry and Machine Company test report for the cold

leg pipe certifies that material meets requirements of ASME

Section~II, 1974 editions through winter 1975. Southwest

Fabrication and Welding Company code data report NPP-1 Form

certified that the _ cold leg subassembly met requirements of ASME ,

Section III, 1974 edition through winter 1975.

(2) The NRC inspector reviewed the procedures and hydro test-data

applicable to Unit 1 since Unit 2 hydro had not been completed.

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Requirements for the tests were presented in Procedures

CP-QAP-12.2, " Inspection Procedure and Acceptance Criteria for

ASME Pressure Testing" and CP-QAP-12.1, "ASME Section.III

Installation, Verification, and N-5 Certification."-Procedure

CP-QAP-12.1 requires that a data package to be used in the test,

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be prepared with the test boundary and the additional following

data shown:

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o Base metal . defects in which filler material has been added,

and the depth of the base metal defect exceeds 3/8 inch or

10% of the actual thickness, whichever is less.

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{[ -o . Untested. vendor performed piping circumferential welds.-

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o Approximate location and material identification and

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description for permanent pressure boundary attachment with

applicable support number referenced.

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J o ~ Weld history, which shall reflect weld removal and/or weld-

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The completed hydro data package (PT-5501) for Unit 1, loop 3

cold leg was reviewed for compliance with the above

requirements. Drawing No. BRP-RC-1-520-001 had been used to

annotate the test boundary. A handwritten statement on the

drawing indicated: "No major base. metal repairs could be ,

located" and "No hangers with weld. attachments could be

located." Welds performed by the pipe subassembly vendor.

including the 22 degree circumferential weld and the penetration

fittings had not been identified. The following items are

. unresolved pending further review to determine:

o If the ~ statement "no major base metal repairs" was based on

a visual inspection or on a review of vendor and. site

inspection and repair records.

o If the shop circumferential weld attaching the 22 degree

elbow to the pipe assembly was inspected during the test.

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o If welds for penetrations into pipe assembly were inspected

since Procedure CP-QAP-12.1 does not require identification of

such welds and they were not identified on the drawing.

The above issues will remain unresolved pending further -

-evaluation by the applicant (445/8507-07; 446/8505-09).

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d. Personnel Qualifications - Personnel who had performed selected tasks

were identified during inspection of installation records. Training

and experience records for the personnel were reviewed to verify that

employee qualifications and maintenance of- records were current and

, _ met requirements. Names or codes for.five welders and two NDE

examiners, who had performed tasks during installation of the items

being inspected, were identified and their qualification records

. reviewed. There were no questions in this area of the inspection.

No violations or deviations were identified.

15. Special Plant Tours (Unit I and Unit 2) -

.0n May 23, 1985, the NRC inspector conducted a tour of selected areas of

Unit 1 and Unit.2. The group consisted of one NRC inspector, two NRC

Technical Review Team (TRT) representatives, two allegers, and several

TUEC representatives. The TUEC representatives tagged each area where'a

deficiency was' alleged. .With the alleger's consent, a tape recorder was

also used.to note locations and describe any alleged deficiencies. The

allegers indicated that-they had identified all deficiencies during the

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tour and all.other deficiencies that they had knowledge. The NRC TRT is

analyzing this information and will decide what action, if any, should be

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. During this tour the NRC inspector independently identified a . questionable

. practice in that_the-top of the the pipe chase at the north end of room 88

l-> in Unit-1, safeguards building had two large sti.ckers which stated that

' areas' 'on the wall were reserved for pipe hangers GHH-SI-1-SB-038-006 and ,

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^R1(?)1-087-X11. These stickers were dated 1980. It was not evident-

whether hangers were missing or none were needed in these locations'"and -

ic w the reserve tags were not removed. TUEC representatives were unable to

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answer the question immediately. This item is unresolved pending further

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review during a. subsequent inspection. (445/8507-08).

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No violations or deviations were identified.

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16. Routine Plant Tours (Units 1 and 2)

..At various times during the inspection period NRC inspectors conducted

general tours of the reactor building, fuel building, safeguards building,

electrical and control building, and the turbine building. During the

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tours, the NRC inspector observed housekeeping practices, preventive

. maintenance on installed equipment, ongoing construction work, and

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discussed various subjects with personnel engaged in work activities. ^<

No violations or deviations were identified.

17. _ Review of Part 21 and 10 CFR 50.55(e) Construction Reports- Status

The NRC inspector reviewed all reports issued to date to assure that NRC

and TUEC status logs were complete and up to date. A total of 183 reports

have been submitted to date. This inspection period one Part 21 report on

Diesel Generator Oil Plugs and two 10 CFR 50.55(e) reports on the

Equipment Hatch' Cover and SA106 Piping (light wall) were submitted.

1h) viol 5tions or deviations were identified.

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18. Ex!t Interviews

The NRC inspectors met with members of the TUEC staff (denoted in

paragraph 1) on May 10 and June 10, 1985. ~ The scope and findings of the-

inspection were discussed. The applicant acknowledged the findings.

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