ML20198K568

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Rev 1 to Spec 2323-GS-900, General Painting Requirements for Mechanical Equipment. Supporting Info Encl
ML20198K568
Person / Time
Site: Comanche Peak  Luminant icon.png
Issue date: 08/27/1974
From:
GIBBS & HILL, INC. (SUBS. OF DRAVO CORP.), TEXAS UTILITIES SERVICES, INC.
To:
Shared Package
ML20197J316 List: ... further results
References
FOIA-85-59 2323-GS-900, NUDOCS 8606040051
Download: ML20198K568 (29)
v • d • e


Text

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I TEEAS UTILITIES SERVICES INC.

AGENT FOR TEEAS UTILITIES GENERATING ' COMPANY ACTING FOR DALLAS POWER S LIGHT COMPANY TEXAS ELECTRIC SERVICE COMPANY.

TEEAS POWER S LIGHT COMPANY t

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i COMANCHE PEAK STEAM ELECTRIC STATION NUCLEAR UNITS NOS. 1S2 I

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GENERAL PAINTING REQUIREMENTS FOR MECHANICAL EQUIPMENT ~

REVISION - 1 1 .

l SPECIFICATION 2323-GS-900 .

l AUGUST 27, 1974

!l l!

l GIBBS S HILL, INC.

EIGINEERS, DESIGNERS S CONSTRUCTORS NEW YORK, N.Y.

8606040051 860527 PDR FOIA CARDE85-59 PDR '

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Specification 2323-GS-900 August 27, 1974, Rev. 1 Page i f TABLE OF .'ONTENTS i

j SECTION TITLE PAGE 1

4 l 1.0 SCOPE 1-(

l 2.0 SURFACES NOT REQUIRING PRIME OR FINISH COATINGS 1 i

i 2.1 -

INSULATED SURFACES- 1 If 2.2 UNDERSIDE OF FABRICATED STEEL EQUIPMENT AND BASES 1

2.3 MACHINED SURFACES 1 -

l 2.4 SURFACES ADJACENT TO FIELD WELDS 1 3

~

3.0 WELDING CLEAN UP 2 3.1 EXTERIOR FERROUS SURFACES 2 3.2 INTERIOR FERROUS SURFACES 2 4.0 COATING OF FERROUS SURFACES 3 4.1 GENERAL 3 e

4.2 CRITICAL EQUIPMENT 3 Rev.1 4.2.1 EXTERIOR SURFACES 3 .

I 4.2.2 INTERIOR SURFACES 3 l 4.2.3 FINISH COATINGS 3 l

4.3 NONCRITICAL EQUIPMENT -

- TEMPERATURES GREATER THAN 250 F 4 i

4.3.1 EXTERIOR SURFACES 4 l

i .

4.3.2 INTERIOR SURFACES 4

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

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Gibbs & Hill, Inc.

Specification 2323-GS-900 August 27, 1974, Rev. 1 Page ii

. SECTION TITLE PAGE 4.3.3 FINIS!! COATING 4 j 4.4 NONCRITICAL EQUIW1ENT -

TEMPERATURES LESS THAN 250 F 4
e

's .4.4.1 ' EXTERIOR SURFACES 5 1

'i 4.4.2 INTERIOR SURFACES 5 4.4.3 FINISH COATING 5

. Rev.1

-i 5.0 NONFERROUS SURFACES AND

! STAINLESS STEEL 5 j! .

6.0 MARKING 5 7.0 SUBMITTAL FOR APPROVAL 6 l

8.0 ACCEPTABLE COATING MANUFACTURERS 6 l

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Specification 2323-GS-900 August 27, 1974, Rev. 1 Page 1 -

3 GENERAL, PAINTING RFQUIREMENTS FOR MECHANICAL EQUIPMENT

,i

'l 1.0 SCOPE I These general requirements shall apply to all mechanical

{ equipment, except as specified in the Technical Specification.

i

. 2. 0 SURFACES NOT REQUIRING FRIME OR FINISH COATINGS 2.1 INSULATED SURFACES Surfaces to be insulated shall be prepared and coated in accordance with Paragraph 4.0 of this Specification, except as speci.fied in the Technical Specification. -

2. 2 UNDERSIDE OF FABRICATED STEEL EQUIPMENT AND BASES The underside of fabricated steel equipment and bases which are

.I not accessible except by disassembly or moving shall be protected by Gaco Neoprene Asphalt Na-62 as manufactured by Gaco Western, Inc.

2.3 MACHINED SURFACES Machined surfaces (exposed parts or trim not requiring coating) shall be protected with at least 4 mils (.004 inches) dry film thickness of MIL-C-16173, Grade I, Preservative (or a similar .

product approved by the Engineer) .

2. 4 SURFACES ADJACENT TO FIELD WELDS Surfaces within two (2) inches of field welds or which will be embedded in concrete shall not be painted.

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Specification 2323-GS-900

' ' August 27, 1974, Pev. 1 f

Page 2 i

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3.0 WELDING CLEAN-UP i .

j 3.1 EXTERIOR FERROUS SURFACES .

I Before surface preparation is initiated:

! a. All welding scale, spatter and excess metal shall be removed.

i i b. In the case of stainless steel, the following shall apply:

1. Only iron-free silica sand, alumina oxide grinding

wheels and/or stainless steel wire brushes shall be used. -

?

2. The following non-halogenated organic

- solvents are approved for cleaning and degreasing.

- Xylene Rev.1

' - Toluene 3 - All other. solvents are subject to the Engineer's l approval prior to use.

3.2 INTERIOR FERROUS SURFACES

's

! Before surface preparation is initiated
'
a. All welding scale, spatter and excess metal shall be removed.
b. Edge of weld bead chall be blended into parent metal.

c.

In the case of stainless steel, the following shall apply:

1. Only iron-free silica sand, alumina oxide grinding 1

.i wheels and/or stainless steel wire brushes shall be used.

I

2. Rev.1 l

The following non-halogenated organic solvents are 4

t approved for cleaning and degreasing.

i -

Xylene '

i Toluene 4

All other solvents are subject to the Engineer's

} approval prior to use

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Specification 2323-GS-900

  • August 27, 1974, Rev. 1 Page 3 4.0 COATING OF FERROUS SURFACES t

i 4.1 ~ GENERAL t

Prime coating of equipment will be required only, unless j

specified otherwise in the Technical Specification.

I

.[ 4.2 CRITICAL EQUIPMENT critical equipment is defined as that equipment that will be located in the~ containment and/or will be exposed to ionizing radiation and contamination by radioactive nuclides. All surface preparation. and prime coating will satisfy the requirements of j latest revisions to ANSI N101.2-1972. *

! 4.2.1 EXTERIOR SURFACES I -

a.

All exterior surfaces shall be prepared in accordance with 4

coating manufacturers instruction and be in accordance with i

., " Steel Structures Painting Council Surface Preparation Specifications". Minimum requirement for surface preparation shall be " Steel Structures Painting Council Surface l Preparatlon Specification No. 6, Commercial Blast Cleaning".

b. The specified shop prime shall be in accordance with MIL-P-38336 (self curing, zine-filled inorganic coating eg. , Carbo j sinc. 11), and shall be applied in strict accordance with coating manufacturers instructions. Coating shall be t

suitable for a continuous service temperature of 750 F.

c. The Seller shall submit all detaila of surface preparation and selected coating for appr. oval by Purchaser prior to start of surface coating.
l. 4.2.2 INTERIOR SURFACES The interior surface preparation and coating, if required, shall be specified in the Technical Specification.

, 4.2.3 FINISH COATINGS Finish coatings shall be applied by others.

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Specification 2323-GS-900 August 27, 1974, Rev. 1 Page 4 ,

4. 3 NONCRITICAL EQUIPMENT - TEMPERATURES GREATER THAN 250 F.

Noncritical equipment - temperatures greater than 250 F are j defined as equipment that will be located in an area where minimum possibility exists for exposure to radiation and

! contamination by radioactive nuclides and subject to temperatures above 250 F and not exceeding temperatures of 750 F.

! 4.3.1 EXTERIOR SURFACES

a. All exterior surfaces shall be prepared in accordance with coating manufacturers instruction and be in accordance with j " Steel . StructuresMinimumPainting Council Surface Preparation Specifications". requirement for surface preparation shall be " Steel Structures Painting Council Surface
Cleaning".

Preparation Specification No. 6, Cou.mercial Blast

b. The specified shop prime shall be in accordance with MIL-P-38336 (self curing, zinc-filled inorganic coatings) and shall be applied in, strict accordance with coating manufacturers instructions. Coating shall be suitable for a continuous service temperature of 750 F.
c. The Seller shall submit all details of surface preparation and selected coating for approval by Purchaser prior to start i of surface coating.

! 4.3.2 INTERIOR SURFACES t The interior surface preparation and coating, if required, shall l be specified in the Technical Specification.

4.3.3 FINISH COATING i

Finish coating shall be applied by others.

4. 4 NONCRITICAL EQUIPMENT - TEMPERATURES LESS THAN 250 F Noncritical equipment - temperatures less than 250 F are defined j as equipment that will be located in an area where minimum

! possibility exists for exposure to radiation and contamination by l

radioactive nuclides and subject to temperatures less than 250 F.

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, /My;ggc0 FIGURE 1. Page 1 of 1 COMANCHE PEAK STEAM ELECTRIC STATION DESIGN CHANGE AUTHORIZATION l.

(MEX) (WILL NOT) BE INCORPORATED IN DESIGN DOCUMENTS DCA NO. 12,626 XX l 1. SAFETY RELATED 00CIMENT: YES NO 2.. ORIGINATOR: CPPE XX' ORIGINAL DESIGNER ,,,-

} 3. DESCRIPTION:

2323-MS-023 2 _

l A. APPLICABLE SPEC /0WG/000 Welt 736J823 Sheet 2 REY.

I l

t 8. DETAILS i~

PROBLEM: 3/4" piping for hutwell conductivity probe from stainless drio trays in condenser as installed is carbon steel oice. Pinina is corroded with l rust. Referenced pipino was furnished with condenser by Westinahouse-I t

SOLUTION: Reolace carbon steel oinina with stainte<< steel niaina en nininn -

specification catecory 152G.

h i

JUSTIFICATION: Stainless steel oinino will not be subinet to internal enerneinn i

4. SUPPORTING 00CUMENTATION:

l TOCR #818

{ ( p M p p m =,7a gga p mp,,

) wunwanna yal );t.]

.- 5. APPROVAL SIGNATURES: :GS:jb March 11, 1982 A. ORIGINATOR: 4D a / J e DATE //!/k I

B. DESIGN REPRESENTATIVE: -

DATE h b I 6. VENDOR TRANSMITTAL REQUIREDi YES X NO

7. STANDARD DISTRIBUTION:

OCA FORM 11-80 ARMS (Original) 1) P ECEIVE^

Quality Engineering 1)  !

TS for Orig. Design. .

ex ,' i l Westinghouse-Site ,

' ' ~ ~~~

j g fi ECEIVE3  !

m _ _

-s _ _ __ _ _. , _ _ _ _ ,_.. _

C bgog i1 \ ; W 9 7; **  %

TEXAS UTILITIES SERVICES INC. A sons e.esv 4.s nsw,;> . DAI.1,M D;XA76 MJH 4

TCH-2931 December 1, 1976 B & R DCC DIST.!l ,

JOB NO. 351195 i

Mr. R.' E. Hersperger E C E 1 V E sn Gibbs & Hill, Inc. f. ,y Dv0 ' /

393 Seventh Avenue New York, New York 10001

['~:) CEC - 8 13I3

[ , ~,, 3 l.[; ;. CPU. ".M /[/

d LL E C E I V E Wl r.,

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' COMANCHE PEAK STEAM ELECTRIC STATION W#

TRANSMITTAL OF SIGNED PURCHASE ORDER CP- 0023 SUPPLEMENT NO. 5 - MAIN CONDENSERS fpjy j/k PILE NO. 08023 g1 ,g Dr A%

TCH

/

i

Dear Mr. Hersperger:

d) /%

Enclosed please find one (1) copy of the signed Purchase Order CP- 0023 Supplement No. 5 for the above mentioned equipment.

If there are any questions or comments, please contact this office.

Very truly yours, i

O I H. C. Schmidt Project Manager-Nuclear Plants

BCS,baskf 1 Enclosure cc
L. A. Ashley Ic.1A H. C. Dodd 2c,2A

.i C. R. Catchell Ic ,2A D. N. Chapman Ic J. C. Kuykendall !c,1A .

~

F01A-85-59 N9%

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- O F FIC E M E M O P. A N D U M To ._, Glen the. Texas Subject Condenser Tube Sheet Leakage During initial filling perations of condenser after turnover from construction, leakage was detected going from the condenser side to the tube sheet pressurization

, annulus. The following paragraphs describe the re-rolling, the leak test results, the redesign of the tube sheet pressurization system and one area of concern.

f 1. The B condenser west waterbox lower bundle was selected as the first bundle

- to test methods. The bundle was' tested with air and water through the tube sheet pressurization system. The bundle leaked 40 psi of air in 15 seconds.

Water could not be added fast enough to even pressurize the annulus. In both I

cases, all leaks were into the condenser, no leakage was detected from the annulus to the waterbox side. The entire tube bundle was re-rolled. Water test of the annulus revealed eight tubes still leaking. These tubes were rolled one additional time a,nd leakage stopped. Final leakage on this bundle was'.021 gallons per hour.

2. After completion of the first bundle, Westinghouse, TUSI Engineering and Startup agreed to only re-roll tubes that were leaking. Using this method, we eli-minated all visible leakage into the condenser steam space except for B con-denser west discharge upper bundle. Af ter approximately six tubes were removed because leakage could not be stopped, it was determined that the tube sheet -

holes were tapered. Startup, TUSI and Westinghouse then agreed to modify the existing roller to roll a tapered hole (roller taper is .004"/in) which stopped all visible leakage.

Final Condenser Annulus Leakage:

1

' Condenser A Inlet West Top .020 gal /hr f Center .017 gal /hr i Bottom .007 gal /hr Condenser A Inlet East Top .009 gal /hr Center .008 gal /hr

{; Botcom .006 gal /hr 5 Condenser A Discharge West Top .020 gal /hr Center .026 gal /hr Bottom .240 gal /hr y Cendenser A Discharge East Top .074 gal /hr Center .055 gal /hr q

Bottom .035 gal /hr TOTAL .517 gal /hr

) '

F0%855 ~

4 I-PM3

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Su-81134 Page 2 Condenser B Inlat West Top .015 gal /hr Center .023 gal /hr i

Bottom .020 gal /hr 5 Condenser B Inlet East Top .047 gal /hr Center .010 gal /hr Bottom .015 gal /hr l Condenser B Discharge West Top .042 gal /hr

  • Center .140 gal /hr

! Bottom .021 gal /hr Condenser B Discharge East Top .024 gal /hr Center .030 gal /hr.

Bottom .032 gal /hr l . TOTAL .419 gal /hr'

3. Testing of the tube sheet pressurization systems revealed a maximum makeup
rate of less than 30 gallons per minute. Westinghouse advised that leakage at normal operating temperature and pressure could be as high as 100 spa. The original system was designed on site, and TUSI Engineering agreed to redesign the system,to attain 100 gpm makeup rate. TUCCO Operations suggested an alter-

' nate feed to the tube sheet pressurization system be incorporated along with the normal feed from the discharge side of the condensate pumps. TUSI Engineer-

  • ing agreed to provide the alternate supply from deserated domineralized water in their redesign.

I 4. One problem incurred during tube re-rolling was annulus blockage. Five tub's e I

had to be cut out and their annulus' enlarged to connect their annulus to adjacent tube annulus to allow proper flow. Annulus blockage was'noted in these five tubes only because they were pressurization system inlet tubes.

i Also, two of the holes drilled from the edge of the tube sheet to the annulus 1 were found not to connect to the annulus. There is still concern about block-age in the annulus in the center of a bundle and no practical way to test it at this time. ,

r

,\ . ,

Dick Camp

}

a ,.

f REC / B/jb a Distribution: J. C. Kuykendall .

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GNANOIE PEAK STEAM EIET. RIC STATION ALLEGED D@ ROPER CG4STRUCTI(N PRACTICES INVOLVING UNIT 1 MS LMIT 2 MAIN GNDENSERS The main condensers and circulating water systen are not required to i shut down the reactor and, as such, are not safety related. Therefore, a g 10CFR50 Appendix B Quality Assurance Program was not invoked during con- i struction activities. However, in accordance with the same sound engineering practices followed during other generating station construction activities, 1UG00 Engineering personnel, TUG 00 Operations Maintenance personnel, Welding Engineering personnel and vendor personnel monitored the activities involved in the construction of the condensers; e.g., Welding Engineering monitored

)

the field welding on the condenser shells; 1VG00 Operations Maintenance monitored the connection of the condensers to the low pressure section of i .

the turbine; 1UG00 Engineering and Operations Maintenance monitored the tube installation and tube rolling proegss; and the vendor monitored all of the erec-t t

tion activities.

) 1he condenser design provides for pressurization, with condensate, of f the tube sheets to preclude any lake water leakage into the stema side fmn j the waterbox side. This design wasyffectedwsing the integral groove. The

'i j groove was machined perpendicular,.to the longitudinal axis of the tube hole in the tube sheet and at the center of the,, thickness of the tube sheet. The i l

diameter of the groove was such to allow for overlap between adjacent grooves,
thus providing a flow path for the condensate. .#

During initial filling opera-tions, leakage from the waterbox was observed going into the tube sheet l

pressurization ann 61us. The tubes were re-rolled and leak tested. No l 1eakage was observed going fmn the waterbox into the annulus. Upon pressuriza-tion of the annulus with condensate, leaka,ge rates of 0.5 gal./hr. and 0.4

' gal./hr were observed going from annulus to the stesa side of condensers A Yb

r Page 2.

S and B, respectively.

>; 1he main condensers have been tested in accordance with Acceptance Test 27-1, which specified a vactam on the steam side and lake water on ,

the tube side. 1he Waers passed this Acceptance Test. During Hot

]

Punctional Testing of the stem supply, operating conditions in the

.t condenser were achieved insofar as a vacuta existed on the stem side .

1 with stema going to the stem side; tube sheet pressurizatioh was in ll ,

place and lake water was ptmped through the.-tube side. .The steam side

~!; '. -

...: .4. . . . . . .p. .

, chemistry was monitored' and no evidence of leakage was noted. r ' c.'.'

~

Dp.< , . .,. ;. . > ;. - .

4, eg,. ,,.

ji ..in order to preclude unsatisfactory' water chemistry on the _ W ry

I side of the steam generator, two measures have been taken. 1) The tube
1

,j sheet pressurization systan, as described above, will preclude lake water -

i

)i from leaking into the steam side of the condenser by injecting condensate into the annulus around each, tube in both tube sheets. Since the condensate e , ,

. J ' ..c .W!an.t.. . , ,

' -in the annulus area is at s' higher pressure than the lake water, any leak

.v:q.J. ;..

will be from the tube sheet pressurization systen to the lake side. 2) A

' condensate polishing systan'is., employed to provide a' continuous contaminant

?z renoval process, thus maint=ining satisfactq;ry stean generator secondary

- side water chenistry. The polishing'systen treats 100% of the condensate and is located between the condensate pinp and the feedwater pinnp on the downstrean side of the condenser.

A e i

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, (4 kO* J bM 33 $ h FOLDER NO. 14 Draft No. Date COMANCHE PEAK ALLEGATION WORK PACKAGE i.

't MECHAN'ICAL & PIPING Category 14 - Miscellaneous Piping l

Allocation Numbers: AP-6. AP-14. AP-15, AP-17 l8 Statement of Allecation: a) A pipe fitter dropped a crowbar "down in a pipe

'I ic the reactor core"; b) penetration seals for flexible boots were the wrong

! size, not the proper strength, and the filler material that absorbs radiation l may, separate out; and c) nonconforming material was used to " butter" piping 360 around the pipe to achieve the proper wall thickness.

1

'; Reference Documents:

'i

j See source documents marked on attached pages from allegation list.

.}

'i Source of Allegation: Anonymous allegers and an individual who testified at 4 the hearinas -

  • 1 Date Received: 3/22/83; 3/7/84 i
The above information prepared by H. S. Phillips 6/12/84 Date Name Group Leader Name Date Assigned ll Team Members

!' Date Assigned 1

Date Assigned Date Assigned i

i

,i I Date Assigned l F0lA-85-59 n-aw a - .- - -- - - . __- . -

- . - . _ . . . . ~._ .

,, ,, Cg%oS N ; P M u 'l ' 4a m 9 3

$ l6 EeShe.,3. N o \ 4-COMANCHE PEAK OPEN ISSUE ACTION PLAN o

Task:

Ref. No.: AP-6, AP-14, AP-15, AP-17 I

Characterization: a) A pipe fitter dropped a crowbar "down in a pipe in the

reactor core"; b) penetration seals for flexible boots were the wrong size, j not the proper strength, and the filler material that absorbs radiation may, separate out; and c) nonconforming material was used to " butter" piping 360 j around the pipe to achieve the proper wall thickness.

Initial Assessment of Stanificance: The alleged problems relate to hardware that is deficient or extraneous objects that could fail to perform its intended

function or be damaged and not be able to perform its function.

Source: Anonymous allegers and an individual who testified at the hearings.

Approach to Resolution:

i

! 1. Review background material referenced in the allegation list.

2. Identify specific problems alleged by reviewing NCR and deficiency logs.

Discuss specific hardware deficiencies with QA, construction, and/or engineering personnel to gather data.

{ 3. Review NCRs, DCNs or other related documents pertaining to these

! deficiencies. Also review correspondence files.

4. If possible, examine the alleged deficient equipment or material.
5. Refer any examples of wrongdoing to the TRT manager. .
6. Evaluate allegations for generic / safety implications.
7. Report on results of review / evaluation of allegations.

Related Open Issues

!l *

1. Using system codes, pull open items, previous inspection findings, etc.,

from the tracking system open item list. (Region IV identify and add to this work package.)

i 2. Review activities necessary to close or partially close related items, either based cn inspection conducted above or reasonable additional inspection while the inspector is familiar with the areas.

! 3. While performing physical inspections above, examine surrounding systems, 1* components, and structures for any apparent defect or indicator of faulty workmanship.

4. If workmen are still in the area.of a physical inspection, interview them for any knowledge of other potential deficiencies.
5. Complete portion of IE Module on QA/QC nonconformances or access control /

cleanliness of reactor coolant piping.

t

=!

F01A-85-59 y =

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I b Status:

Review Lead: M . D . (_ H u

,j Support:

1 Estimated Resources: 6 m bM4 S i

j Estimated Completion
bwn gg, A%

t CLOSURE: ,

, Reviewed by:

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IAsa 50ueCE tantillsG CAf tGeef I F SCNE tutt 50ustE s_e AttIGAf tese te Centlass AClloss/5 fATUS Asmus CosW IS e_m/DAE SYSIfM me. [f_ne erls Cougrt[ gec_seent eact AP-9 Stelatens piplag heated to 8 54-006 3/7/94 A-28 este pfplag. AGI festlemay. P.20.Pl.22 )

M,.i. A sie- ,i,e leien.e. for leitlei die. i i. ,9 ii the Walt B turtlas fell peeltlee IG 79-88 ASI of f of a tract and struck a vallreed tract. It i es .

taben back to e storage j eroe and hidden.

AP-Il Beecter vessel stoport 8 94-0061/7/04 A-7 blects and several naastes AAI test lemay. P. 4-9 g eneleed *crects sad calPs* t durlag lestellettee. (tiere J

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TEXAS UTILITIES GENERATING COMPANY

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, subms COMANCHE PEAX S E2M I'.I- 1*: 5 4 ::N D'NETRATION SEAli 23*:C::CSI A:c:0VI 9EF: SP-11 2ei. 1*

' SP-5C4 Rev. .

', SP-5CS lev. I SP-805-1 Rev. 2 l We ave ev'ewed the referenced ::rocedures submitted for approval, the results af tha: review are indicated bel:w:

SP-110 Aoproved. (Paragraph 1.5 TL'GC3 not US!)

SP-504 Approved.

5P-505 1. The orevious revision did not mention the 4000 upper Ifmit for pipes, as does Para. 3.1 now. Are we in compliance with this criteria as currently stated.

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2. Paragraph 6.2.1 mentions SP-505-A, we don't have a #[/$ (&ty/' N@)

copy of that procedure.

U Paragraph 6.3.-1 mentions a 350 upper limit, the

. prior revision did not. Are we in compliance with ygg 3.

l this criteria as currently stated.

l 4 Paragraph 6.5 we do not use the collars, we have done SC calculations to show scattering is not a problem.

SP-805-1

5. Section 6.5.6.

Approved.

Not used at CPSES. 8b .

, SD-110 is approved with comment, SP-505 will require further clarification and/or revision prior to acceptance. SP-504 and 805-1 are approved as is.

Should you have any questions please contact Kim Anger, s '

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, M/WWI ROB
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FOSG Supervisor cc: ARMS l C. G. Cr.eamer t

F. L. Powers

,' J. C. Youngbloc V. C. English I

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4 .. Log No. TXX-4219

( Ahe yo 19j bW 47) M f' File No. 10110

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TEXAS UTILITIES GENERATING COMI'ANY

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Mr. R. L. Bangart, Director ,

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. F Region IV Comanche Peak Task Force l D -- a Nuclear Regulatory Commission Region IV i 611 Ryan Plaza Drive, Suite 1000

, Arlington, TX 76011

Subject:

Construction of CPSES Condensers Dockets: 50-445 & 50-446

Dear Mr. Bangart:

This refers to your letter of June 19, 1984, concerning alleged improper construction practices involving the Comanche Peak Steam Electric Station main condensers. .

The a: sin conden.ers and circulating water system are not required to shut down the reactor and, as such, are not safety-related. This equipment is not listed Therefore, in a 10CFR50 Appendix 17A of the FSAR (list of Quality Asstirance items).

' Appendix B Quality Assurance Program was not invoked during construction activi-ties.

However, in accordance with the same sound engineering and construction practices followed by TUCCo during other generating constructions activities, TUCCo engineering personnel TUCCo Operations maintenance personnel, velding engineering

' personnel and vendor personnel monitored the activities involved in the construc-tion of the condensers; e.g., velding engineering monitored the field welding on the condenser shells; TUCCo Operations maintenance monitored the connection of the '

condensers to the low pressure section of the turbine; TUCCo engineering and, Operations maintenance monitored the tube installation and tube rolling process; and the vendor monitored all of the erection activities.

f, The condenser design provides for pressurization, with condensate, of the tube i; sheets to preclude any lake water leakage into the steam side from the waterbox

(; side. This design was effected using the integral groove. The groove was machined perpendicular to the longitudinal axis of the tube hole in the tube sheet and atThe

! the center of the thickness of the tube sheet.

' to allow for overlap between adjacent grooves, thus providing a flow path for the condensate. During initial _ filling operations, leakage _fgom.the va_terbox_wAE.

The tubes were re-

, observed gqing into the tube sheet pressuriYation annulus.from thg warexbox into lha.

ro'1Ted and leak tested. No leakage was ob~se6ted'eafne Earnblum -'Daoutessurjaa3$.oJLof the annMus_vith condensate _,.. leakage rates_3of _0 5_

Ja,1,:/.,h,,,r,. and 0.,4, gal,.fhr h were observed going from annulus to the steam side of z ~ * - - - --

y ondensers A,and B res ect_1 YlOA. Servenum ur sma s.ros.orse,. sacrno . com.ar L ..

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sy4.s(No.TXX-4219 Page 2 of 2 The main condensers have been tested in accordance with Acceptance Test 77 which specified a vacuus on the_ steam side _and_ lake _ water on_the. tube. aide. The_ son-

! densers passed this Acceptance Test. Durina Hoe Fustional Testing of the steam l lsupply,operatingcondjionsinthecondenserwereachieved. During thisJime , , l tube sb==* -ressurlzation was in service. The steam side chemistry was' ~ monitored ~ ~ " ~

and no evidence of.leakad Fas~.noted.-- ---

i  ;

I

! In order to preclude unsatisfactory water chemistry on the secondary side of the steam generator, two measures have been taken.

1) The tube sheet pressurisation system, as described above, will preclude lake water from leaking into the steam side of the condenser by injecting condensate into the annulus around each tube in both tube sheets. Since the condensate in the annulus area is at a higher pressure than the lake water, any leak will be from the tube sheet pressuri:ation system to the lake side. 2) A condensate polishing system is employed to provide a con-l tinuous contaminant removal process, thus maintaining satisfactory steam generator i secondary side water chemistry. The polishing system treats 100% of the condensate-

.I and is located between the condensate pump and the feedwater pump on the downstream I side of the condenser. There is a condenser sampling system which allows early detection of lake. vater in ,the condensate side of the condenser. Additionally,* the

, condensate system has salinity and conductivity cells to detect leakage into the condensate system. Alarms from these detectors warn the operators of any leakage.

i We appreciate your bringing this matter to our attention. ,

Yours truly, I

BRC;ah i ec: Messrs. T. A. Ippolito B. J. Youngblood ,

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$.f;.ts AUG 241983

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In Reply Refer To:

Docket: 50-445/83-24 R / //

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d 0 /f Texas Utilities Generating Company L ATTN: R. J. Gary. Executive Vice L President & General Manager j 2001 Bryan Tower j g /7 j Dallas. Texas 75201 d Gentlemen:

O L This mfers to the inspection conducted by our Senior Resident Inspector, .

, construction, Mr. R. G. Taylor, during the period March through July 1983, cf activities authorized by NRC Construction Permits CPPR-126 and CPPR-127

l for Canande Peak, Uhits 1 and 2, and to the discussion of our findings with Mr. R. G. Tolson, and other mes6ers of your staff during the inspection.

Areas examined during the inspection included review, inspection, and evalua-tion of several allegations made to various NRC persons, including the Atomic -

D Safety and Licensing Board in their proceedings regarding the operating license ll for Comande Peak Steam Electric Station (CPSES). Within these areas, the L inspection consisted of selective examination of procedures and represen ative a

records, interviews with personnel, and observations by the inspector. These

(; findings are documented in the enclosed inspection report.

h During this inspection, it was found that certain of your activities were in violation with NRC requirements. You wem notified of one such violation by our letter of May 31, 1983, to which you have responded. Details of the item enclosed with our May 31, 1983 letter are included in the enclosed j

inspection report.

L L One unnsolved item is identified in paragraph 15 of the enclosed inspection report.

We have also examined actions you have taken with mgard to previously identified inspection findings. The status of these items is identified in paragraph 2 of the enclosed report.

In accordance with 10 CFR 2.790(a), a copy of this letter and the enclosure l' will be placed in the NRC Public Document Room unless you notify this office.

by telephone, within 10 days of the date of this letter, and submit written application to withhold information contained therein within 30 days of the date of this letter. Such appilcation must be consistent with the mquim-ments of 2.790(b)(1).

BIT 01Ticial File Copy

.- . . . _ . . . . . . . _ . . - . _ . _ ..__y_.

RPB1/h,,

. 9.M,; 7090 N3 fl L

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"*.R,Tay,1,o,q, lac,,DHunni,' cutt ,, ,Gf tadsen, ,,,,,,, ,,,,,,, ,,, , ,,,,,,

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Texas Utilities Generating 2 i Ccmpany AUG g4ggg j Should you have any questions concerning this inspection, we will be pleased t.

to discuss them with you.

Sincerely, "Ortsinal s!gned W Q. L. M Ao0!!N" G. t.. t'.adsen, Chief Reactor Project Branch 1 t

Enclosure:

Appendix - fiRC Inspection Report 50-445/03-24 50-446/63-15 .

cc w/encls:

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

' Texas Utilities Generating Company ATTH: B. R. Clements, Vice President, !!uclear 2001 Bryan Tower, Suite 1735 fallas, Texas 75201 bec to DMB (IE01) bec distrib. by RIV:

RPB1 D. Kelley, SRI-Ops RPS2 R. Taylor, SRI-Cons TPS SectionChief(RPS-A)

J. Collins, RA J. Gagliardo, DRRP&EP C. Wisner, PA0 M. Rothschild, ELD MIS SYSTEM RIV File TEXAS STATE DEPT. OF HEALTH Juanita Ellis David Preister i

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  • APPENDIX t

U. S. NUCLEAR REGULATORY C0!HISSION l REGION IV

[

NRC Inspection Report: 50-445/83-24 i_

50-446/83-15 Category: A2 Docket: 50-445 50-446 e

Licensee: Texas Utilities Generating Company (TUGCO) lj 2001 Bryan Tower i Dallas, Texas, 75201 Facility Name: Comanche Peak Steam Electric Station (CPSES), Units 1 and 2 i

Inspection At: Comanche Peak, Units 1 and 2 Glen Rose Texas Inspection Conducted: March through July 1983 Inspectors: b fX } _n f 9//7/83 Ot te /

R. G. Taylor,(Senior Construction SRIC) Resident Inspector Approved: . h) nof 8//d7/83 D. M. Hunnicutt, Chief Dhte /

Reactor Project Section A 1

Inspection Summary Inspection Conducted March through July 1983 (Report 50-445/83-24 and 83-446/83-15)

Areas Inspected: Special inspections, announced and unannounced, related to allegations made to various NRC persons including the Atomic Safety and Licensing Board in their procedings regarding the operating license for Comanche Peak Station. The inspections involved 449 inspector-hours by one NRC inspector.

l Results: _ The inspection confirmed the need to issue four violations initially identified by the Construction Appraisal Team (CAT) (NRC Inspection Report 50-445/83-18; 50-446/83-12). These involved the areas of HVAC, Equipment Installation, Document Control, and Storage of Equipment.

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1. Persons Contacted _

Princioal Licensee Employees

! *R. G. Tolson, Site QA Supervisor

! *C. T. Brandt, Non-ASME QC Supervisor

! *J. R. Merritt, Engineering, Construction and Startup Manager i *J. B. George, Project General Manger 4 *D. N. Chapman, QA Manager

  • B. R. Clements, Vice-President, Nuclear i

Brown & Root (B&R) .-

  • G. R. Purdy, Project QA Manager
  • D. Frankum,. Construction Project Manager The SRIC also interviewed many other licensee, B&R, and subcontractor personnel during the course of the inspection.

-ll

  • Denotes those persons who attended one or more management interviews with '

the SRIC.

2. Licensee Action on Previous Inspection Findinos l

(Closed) Unresolved Item (50-445/82-22-02), " Analysis of Weld Discrepancies."

This unresolved item concerned a substantial number of identified defects in a large whip r'estraint essentially surrounding the mainsteam and feed

' water lines located several feet outside of the ASME code boundry point.

The device was engineered by the licensee's A/E and manufactured by NPS Industries. Due to the overall size of the structure, it has been nick- The i

named " George Washington Bridge" by the site labor *and quality forces.

l' licensee had reported the finding of the defects as a potential 50.55(e)

l. item to the SRIC on September 30, 1982, which was subsequently stated not i

reportable in a letter dated December 27, 1982. An NRC inspector followed t; up on the matter during a visit to the offices This of the A/E, review as documented pertained to all of the in NRC Inspection Report 50-445/83-12.

D defects involved with the exception of two cracked welds that had not been

!, The engineer has recently analyzed analyzed at the time of the inspection.these two defects and has det d: ' structure could have fulfilled it's function.

The SRIC has reviewed the location of the cracks and their length in relation to the size of the welds 1

and the functional application of the structure. Since the structure has no l

continuous service application and is essentially subject to a one-time

! loading, the cracks would not have the potential for furth tively low stresses in the one-time impact based that It appears on their small size the cracks in due formed l relation to the members being welded.

to the stresses developed during the tightening of high strength bolting in t

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the immediate vicinity of the welds during the site assembly of the structu Taken in conjunction.with the earlier documented review of the engineers calculations and the SRIC's review of these cracks, the SRIC has con that the engineer's o'verallBoth analysis was adequate the licensee's and initial report that deficiency (CP-82-12)

)

i not reportable under 50.55(e).and the above identified unresolved i It should be noted for the record that this closure only applies to the reportability aspects under 50.55(e) and not to the correction of th The defects, including the cracks, have been documented on a nonconfom 1

report.

The final disposition and closure of the NCR will be evaluated l during future routine inspections.

3.

Review of Licensee Self-Evaluation (Using INPO Criterial .

The SRIC has reviewed a report of the licensee The evaluation was performed in behalf of the lican-the purpose by INPO.

see by personnel in the employment of ASargent copy of the & report Lundy, wasan archite fim with substantial nuclear power involvement.

furnished to ,the NRC, and subsequently, to the Atomic Safety and Lic

Board in the matter of Comanche Peak Station operating license by lette -

dated May 2, 1983.

The purpose of the review by the SRIC was to detemine if any of the 47 findings in the report were of a type and of sufficient ,

significance to have been reported to the NRC as required by The SRIC reviewed each of the 47 findings and the supporting docum This review revealed that none in the report pertaining to each finding.

of the 47 items were based upon identified deficiencies in structures, systems, or components nor were there any significant deficien engineering,(or 10 CFR 50.55 e).

testing that would constitute condi

4. Car Wash In Containment _

During the limited appearance 16,.1983, statement a person stated portion of the Atomic Sa at transcript Licensing Board hearing on t'.ay page 6152 The that person he understood stated that it that his was theunderstanding containmentthat looked the some situa-car wash.

tion developed at about the same time that there was a meeting at Airport between the NRC and any interested parties For theto discuss NRC purposes

! That meeting took place on April 5,1983.

tralization.

of evaluating this allegation, the SRIC expanded During this entirethe period of intere period, include the 3 weeks prior to the meeting.

the Unit I reactor system was undergoing what is referred to as " Ho tional Testing". This particular test is an accurate simulation of the operation of the reactor The heat and tystem andin its pressure the appurtenances system is generated by but witho core being in place.

'the reactor coolant pumpsThe in conjunction test could readilywithbe the chemical construed to and be a volume c trol system charging pumps. This parti-pressure test but in fact is an operational test at pressur m . _._ _~- _ ,. - ._ _ - . _ ~ _ m-_- &.._. ._

. . _ = ___. _ . _ . - . . . _ . . . _. . . _ .

1 I' 4 l

and continuing until late May. The SRIC monitored the test but was by no means continously in.the containment. The SRIC interviewed personnel in

! the licensee's startup test group, QC inspectors who had reason to be in

y the building and others to obtain a picture of the events that occurred in The SRIO the Unit 1 Containment Building during the period of interest.

also reviewed the licensee's control room logs for any indication of oper-ational problems indicative of a major leak in any of the fluid filled systems under test. The picture obtained was that there were several small In leaks, generally at the gaskets between valve bodies and their bonnets.

addition, there was a considerable amount of condensation drippino from the reactor coolant pump motor cooling coils. This was caused by the cold water in the coils condensing the humidity from the atmosphere within the building and was not indicative of a leak in the reactor coolant system. The SRIO found from the control room logs that on March 29, a steam leak occurred Perhaps during one phase of the test when a drain valve was partially open. .-

this valve should have remained closed. The room in which the valve was located was apparently filled with steam vapor which would have condensed On March 30, the reactor vessel head out on the cooler walls as water.

vent valves were partially opened, which in turn would give some amount of steam blowoff into the reactor refueling cavity area and would rise up into a

the building until cooled and c.ondensed out as water. None of these events i are typical of any major leak indicative of piping or piping component '

(such as a valve) failure. The type of small events described above are, within the experience of the SRIC, typical of what would be expected during  ;

such a test and is one of the reasons for perfonning the test.

5. Design of the HVAC System Supports _

By letters, both . dated March 11, 1983, Citizens Association for Sound

' Energy (CASE) notified the NRC's Offices of Inspection and Enforcement and the Executive Legal Director of a concern that the HVAC system for Comanche Peak had not been properly supported, nor had it been properly considered in regard to seismic load conditions or its treatment as potential mis-siles. CASE specifically states that from their review of the FSAR, it appears that the licensee has not analyzed the HVAC supports for seismic load condition.

In addition, the personal observations of Messrs. Walsh and Doyle are relied upon to point out that there are no lateral supports on the HVAC

, CASE also states that all HVAC components systems within the containment.and supports inside containment shou terion 4 of the General Design Criteria for Nuclear Power Plants, 10 CFR 50, Appendix A.

Sheet 21 of Table 17A of the FSAR lists the containment ventilation sys-tems as being Seismic Category II. Apparently, it has been assumedThis by CASE that this category excludes seismic loading in the design.

assumption is incorrect since the FSAR, Section 3.2.1.2 defines Seismic Category II as being those portions of systems or components whose

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. . 5 continued function is not required but whose failure could reduce the func-tioning of any Seismic Category I system or component required to satisfy the requirements .of C.I. A through C.1.Q of Regulatory Guide 1,29 to an unacceptable safety level or could result in incapacitating injury to 1

occupants of the control roor. These systems are designated Non-Nuclear .

Safety (NNS) Seismic Category II and are designed and constructed so that a safe shutdown earthquake (SSE) will not cause such a failure.

CASE also states that if the HVAC systems within the containment failed during a SSE, this would allow the temperature within the containment to rise quickly to unacceptable levels which could over time cause compon-J ents and monitoring equipment to fail and which could also mean that it might be impossible for workers to enter the containment due to the heat.

Containment heat removal is required by Criterion 38 of the General Design ,

Criteria for Nuclear Power Plants. The system to remove heat from the reactor containment at Comanche Peak does not rely on the HVAC system but ,.

l rather is composed of two separate containment spray recirculation trains each with 100. percent capacity. Each train contains two separate pumps, one heat exhanger, and seven spray headers, and each system is fed from its individual electrical Class IE bus. The containment heat removal i system is designed to ensure that the failure of any single active compon-

-i ent, assuning the availability of either onsite or offsite power exclusively, -

does not prevent the system from accomplishing its planned safety function.

CASE's concern with being able to enter the containment following certain design basis accidents is unfounded in that it is not a requirement.

In order to assess the adequacy of the design of HVAC supports, an inspec-tion was conducted at the home office of " Corporate Consulting & Develop-1 ment Company, LTD. ." the support design consultant. It was detemined that Two methods all permanent HVAC supports are analyzed for seismic loading.

are utilized: Zero Peak Accleration (IPA), or 1.5 Times theOf Peak Accelera-tion When the Fundamental Frequency Falls Below 20 Hertz. the latter 1

method of design, only about 6 out of 4000 supports have been designed that way. A typical HVAC duct run is supported axially at every third support This may explain why Messrs. Walsh and Doyle may have felt that there were no lateral supports on the HVAC systems. The NRC inspector reviewed the design of a typical HVAC duct run at elevation 852'-6" in the Auxiliary i Building. Supports were designed utilizing two computer programs ue andentitled stress FEASA-20 and FEASA-30. The acronym stands for frame eigenva 1 analysis. The -20 version is used on theThe transverse inclusionsupports and the of equivalent -3D weights version is used on the axial supports.

i from both up and downstream transverse supports and accesories such as vol- '

uma dampers and vane turns in the design of the axial supports was verified.

This inspection verified the adequacy of the siasmic design techniques being utilized for the design of HVAC supports at Comanche Peak.

The concerns expressed by CASE have been found to be without merit.

Persons contacted during the course of the inspection at Corporate Consulting

-. . . - . . . . ~-~ . - _ - _ - - - -_n - . - - -

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6 ,

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l' & Development Company, LTD. were:

0 j J. Roland Yow President & Chief Executive Officer H Gary Hughes , Vice-President for Operations 9 Davidgindley, Principal Engineer StepRei Lehrman, Seismic Department Manager F "

Daryl Hughes, Project Engineer

6. Heatino, Ventilation. and Air Conditionino System (HVACl 50-45/83-18;50-446/83-12),

During the CAT inspection (NRC Inspection Reportthe CAT inspec ll l

i support structums were deficient in relation to the applicable welding code

! requirements. The dominate Based deficient upon this condition information noted thewas SRICthat the various toured welds were

!i significantly undersized. areas of the facility with special emphasis on the d H Unit 2 Containment Building since that was one of the more recent ..

areas of installation by the HVAC contractor. In accordance with i

the design drawings, the bulk of the welds should have been fillet

? welds with inch leg size. The SRIC noted by visual comparison to i the inch thickThe base metal that very few of the welds were of CAT inspectors also found cases where the bolting proper size.

b and gaskets between ducting sections were loose and/or missing.

d The CAT inspectors also found that some support members were It not was j

within the dimensional tolerances on the design drawings.

noted that the contractor's inspection records did not reveal Further, these i

various facts, indicatiny ineffectual QC by the contractor.a re was unaware of these several problems in the fabrication, installation, Based upon the CAT inspectors' and inspection of the HVAC systems.

findings and his .own observations, the SRIC recen1 mended that a

!!. notice of violation be issued to the licensee pertaining collectively ir to these matters (Notice of Violation issued on May 31, 1983.

Reference 50-445/83-18 and 50-446/83-12, item 4).

f}l

7. Installation of Major Items of Equipment I

The CAT inspectors noted during their inspections of certain major items of equipment that there were'several variables in how the In some 9 equipment was fastened to the building equipment pads.

i;t tanks for example, CAT inspectors found that there were instances two nuts (double nuts) on the embedded bolts securing the equipment, j other bolts had one nut (single nut) and some had a combination

'. The of both single nuts and double nuts on one piece of equipment.

CAT personnel also noted that certain heat exchanger during operation. The holddown nuts appeared to be installed too

l. Tht SRIC tightly and may have prevented freedom of movement.
obtained'the design and installation drawings Both for two of the referenced

! were found to l heat exchangers identified in the CAT report.be horizontal Utube l

l but whose pressure boundary in the tubes is safety-related since theTh process fluid could be radioactive.

drawings for the mounting pedestals had a flat steel plate on one

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s pedestal that would be suitable for the type of mounting detail i

on these heat exchangers. The SRIC then reviewed the installation travelers for each heat exchanger and found that these documents

,t did not note or address the slotted details, the plate, or the fact J. the bolts should be left loose. The SRIC would note that the vendor manual which provides the details does not provide infomation on how loose or tight the nuts should be nor how these nuts are to

' be locked at that looseness or some torque value. The SRIC with 1

the assistance of site QC and craft labor had one of six nuts loosened on heat exchanger TCX-CSAHLD-01. On all six of the studs involved, each had only one nut (single nut). The one nut that

.: . was loosened had been very tight, as evidenced by the amount of 1 force required to break the nut loose. (h another heat exchanger of comparable design, it was found that each stud was double nuted and when the top nut was loosened, the second nut was approximately one flat (about 1/6 of a turn) from being fully tight. This degree of looseness should a11ov sufficient freedom of movement. During  :-

the document review, the SRIC found that the engineer had specified that all rotating and vibrating equipment should be double nutted and that other equipment could be secured with only one nut. No document could be located that established the identity of vibrating equipment nor were there any apparent provisions made to lock nuts l J where they must be deliberately left loose. This was considered -

overall to be a violation of Criterion V of Appendix 8 to 10 CFR 50 (Notice of Violation was issued on May 31, 1983.

Reference:

Notice of Violation 50-445/83-18 and 50-446/83-12, item 1).

8. Maintenance of Eautoment In Outdoor Storace Areas _

1 The CAT found that a considerable amount of equipment such as pipe i support struts, clamps, and like items, normally stored outdoors, l

was not being properly maintained in accordance with procedure MCP-10,

" Storage and Storage Maintenance of Mechanical and Electrical Equipment", as evidenced by rusting bolts and adjustment screws on struts. In addition, the strut bearings were dirty from dust and i

the bearing load pins, in some instances, were rusted. By a tour of the storage areas, the SRIC confirmed the CAT inspectors find-

. ings. The SRIC would also note that the INPO Self-Evaluation 1 Report at page 111 describes essentially the same finding. This situation was detemined to be a violation of Criterion XIII of Appendix 8 to 10 CFR 50 (Notice of Violation issued on May 31, 1983.

Reference:

Notice of Violation 50-445/83-18 and 50-446/83-12, i tem 2) . The SRIC would note for the record that there is little evidence that any items which indicated substantial deterioration from such storage conditions have in fact been installed in tne nuclear power block. It would appear that the various items involved have been cleaned and restored prior to installation such that they can perform the required function.

9. 6bsolete and/or Illegible Drawings In The Field The CAT inspectors found a group of drawings in one particular area adjacent to the control room that were found to be out of date by up to several issues and further, that some drawings in other areas were incomplete in the title and revision blocks. The SRIC discussed
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, 8 the finding with supervisory personnel of the licensee's central document control center who indicated that they had located the drawings identified by the CAT inspectors along with many more that were obsolete in other areas. It was stated that distribution system for engineering drawings had become faulted by the simple

volume and by the need for so many points of distribution and audit verification thereof. Since problems are obviously still present, -

it was determined that the licensee had violated Criterion VI of Appendix B to 10 CFR 50 (Notice of Violation was issued on May 31, 1983.

Reference:

Notice of Violation 50-445/83-18 and 50-446/83-12,

'

  • item 3) and that substantial steps would be required to correct the

' problems.

. 10. Allegations Relative To Improperly Supported items In The Control Room _

i The president of CASE in a letter dated March 11, 1983, addressed to i

Mr. Richard C. DeYoung. Director of the NRC Office of Inspection and Enforce-ment, indicated that CASE had received infonnation from an unidentified source to the effect that:

a. There is field run conduit above the control room supported only by wire.
b. There is drywall (or sheet rock) that is supported by wire.

i

c. There may be lights that are supported by wire.

The SRIC has examined the suspended ceiling and the area above the sus-pended ceiling in the control room area and has examined the pertinent engineering drawings depicting both in relation to these allegations with

l the followLng findings

l

a. There is a considerable amount of both safety-related and nonsafety The safety-fi

,related conduit in the area above the suspended ceiling.relate The nonsafety-related of those used in other areas of the facility.

conduits are generally supported by simpler and less substantial sup-ports that are typical of those that the SRIC has observed In in each large open factories and are not designed to seismic standards.

case examined, the non-seismic support was structurally paralleled with a small stainless steel cable that would assume the full weight of the conduit were the nonnel support to fail in a seismic event.

a

b. The drywall materials were found to be part of the suspended ceiling

' above the central part of the control room and to fonn a part of the These dry-sloping wall area below the control room observation room.

wall materials have been securely fastened to a metal frame work (metal batten) which in turn is supported by conventional and non-seismic straps and wires to the concrete primary building. The frame work is also attached to a system of stainless steel cables which in turn also attach to the primary structure such that if normal sup-ports fail during a seismic event, the weight of the framing and drywall will be assumed by the cabling thus preventing the materials from falling. ,

9 1 .

c. The lighting fixtures in the control room are supported from an

' intermediate substructure of "unistrut" by light-weight conduit.

The substructure is likewise supported by the same type of conduit The conduit used appears l from the primary structure ceiling.

to be the typical'of that supporting the light fixtures in most l offices with suspended ceilin s. Paralled with each conduit are d

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two small stainless steel cab es which would assume the load '

1 In the case if the conduit or its attachment were to fail.

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of the actual light fixtures, the cable is attached to the light '

fixture at the edge of the reflector assembly.

f A The SRIC would note for the record that above described design features appear to fully satisfy the intent of the licensee's commitment to comply with NRC Regulatory Guide 1.29. " Seismic Design Classification."

' The licensee has used terminology in the classification system that is at variance with that of the regulatory guide but is explained and defined -

In essence, the licensee has defined all in Section 3.2 of the FSAR. 2 safety-related items that must remain Items fully notfunctional having a safetyduringfunction and after a seismic event as Seismic Category I.

I but whose failure could damage components which have a safety function l or cause injury to the occupants of the control room during an event are referred to as Seismic Category II.

j this. allegation, all are Setsmic Category II since their falling couldThe  : c cause injury to the control operators. -

- be expected to prevent such a fall even though the normal supports could possibly fail. The stainless steel cable used in this design feature, which at a short distance away looks As much like bright an example, the testgalvanized strength of-comon stee wire, is of relatively high strength. With four cables attached an 1/8-inch cable is in excess of 1760 pounds.

, to a light fixture, two at each end, the total support capability of th cables s over 7000. pounds.to use conventional suspended ceiling and light fixtur in order to use conventional and available materials and then provide a

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' high strength backup support system in a seismic event.

No violations or deviations were identified during this special inspection effort.

l 11.

Placement and Curina of Concrete Durino Freezing Weather _

j 1: During the limited public appearance portion 15,of1983, the Atomic there were Safetytwo and G' Licensing Board (Board) hearing conducted on May references to the placing of concrete in freezing weather at the Comanche

, Peak Station which in turn lead to a question from the Board to the NRC l

staff as to whether there were any NRC personnel present with knowledge

! The two references are at 6106 and Also 6134 of the hearing at 6109, an uni-of the matter.

! transcript while the Board question is at 6109.

l I dentified voice responded to the Board that the matter had bee j

in IE inspection reports. 50 445/77-01 that there had been such a discussion in NRC Inspection Report rL which was categorized as an unresolvedThe item pending the licensee's review unresolved item was ll and action on their finding of the problem.50-445/77-04 with the closure of further discussed in NRC Inspection Report j

the item by an improvement'in the QA procedures.

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10 The SRIC has reviewed the matter, particularily with a view toward deter-mining whether the practices involved actually caused damage to the concrete 50 l- involved. The primary focus of NRC Inspection Report i

paragragh 5) was directed toward two licensee " Site Surveillance Reports" which had been prepared approximately 2 weeks earlier than the inspection i The first of the licensee's reports l

period covered by the inspection report.

(C-134-77) was directed specifically to findings by a licensee inspector j

that the surface temperature of Concrete Placement 101-2808-001 some 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after the placement was completed were well below freezing in some locations.

i The other licensee report (C-135-77) was directed toward records and was not considered in this review. The SRIC obtained the necessary records

' to review the matter and found that placement 101-2808-001 had taken place on December 30, 1976, being completed at approximately 6:00 p.m.

Later, the same evening at approximately midnight, the licenseeThe inspector records found that some surface areas were chilled to as low as 200F.

reflect, however, that there was disagreement between the S&R inspection "

personnel assigned to monitoring the curing of the placement and the licensee's inspector as to what the surface temperatures actually were.

The B&R personnel contended that the licensee inspector was acta 11y mea- No suring the air temperature rather than the temperature of the concrete.

The SRIC resolution of that disagreement was reflected in the records.

interviewed the licensee inspector of record during the course of this review to gain a clearer understanding of the events which took place. ,

~The licensee inspector stated during the interview that he was confident that his measurements were accurate and also stated that there was no phy-sical evidence that the concrete was frozen even though the surface Tne records also reflect that in temperatures were well below freezing. order to resolve the issue, s areas after the concrete had fully cured. These tests indicated that the I suspect areas had attained strengths comparable to known properly cured areas, indicating that the concrete had not been damaged even though The the

, possibility exists that it had been frozen for a period of time i

were established and maintained shortly after the licensee's inspector's observation, 101-2801-001 took place For the record, the SRIC would note that PlacementThe placement became f in the Unit 1 Reactor Building. This floor area, while supper-at the lowest full floor in the building. As such, it is fully I ting some equipment, serves primarily as a walk area.

topped with an archttural concrete making the structural concrete no longer accessable.

NRC Inspection Report 50-445/77-01 also discussed comparable events to tha documented on Surveillance Report C-135-77. One of these events was docu-mented by Surveillance Report C-068-76 on January 7,1976, and on B&R deficiency / disposition reports (now titled nonconformance reports).

,These documents indicate that on January 7,1976, the surface te of Placement 105-2773-001, Building, were found frozen as evidenced by fro:en wet burlap over certain areas that were not covered by insulating blankets. The records also m -

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reveal that the reported finding took place almost 7 days after the place-ment of the concrete. Although the placement should not have been allowed

- to freeze in the time frame involved in accordance with the project speci-fication, the placement was accepted "use-as-is" on the premise that the curing temperatures during the 7 days were conducive to a good cure and that after 7 days there would be little free water in the concrete to freeze even i though the burlap was froze. This conclusion is considered valid by the l 5

. SRIC based on his review of publications of the American Concrete Institute

! and the Bureau of Reclamation. Further, in responding to a separate finding i that the field cure test cylinders made for the placement tested lower than 1 allowed by the project specifications, swiss hammer tests were perfomed.

The swiss hanner tests indicated the concrete placement had full specified s trength. Relative to the low reported strengths of the field cure cylin-

, ders, the SRIC would note that in his experience field cure cylinders will

. frequently test low under cold weather conditions. The reason is that the .

cylinders' small mass generates little heat of hydration, thus making them _

either more vulnerable to freezing and/or curing much slower than normal due ~

to their depressed temperature.

The final events covered by HRC Inspection Report 50-445/77-01 included DDR-C-460 which in turn discussed low temperatures during the curing per-

, iod of three separate placements that were made during the late December

, time period of 1976. In each case, the records reflect that the placements .

were accepted "use-as-is" since the least amount of cure time was 9 days, again with good conditions until the cold weather occurred.

The NRC inspector involved in NRC Inspection Report 50-445/77-04 which closed the unresolved issue has stated that he had visually inspected each of the placements discussed in NRC Inspection Report 50-445/77-01 for evidence of damaged concrete and found none. NRC Inspection Report 50-445/77-04 did not reflect those inspections since the NRC inspector was aware that the

concern was for prevention of repetition rather than any specific concern about the quality of the placements involved.

l The SRIC would note for the record that there are no regulatory or industry prohibitions on placing concrete in cold weather conditions. The American i

Concrete Institute and the Bureau of Reclamation both indicate that if the fresh concrete is above 400F at the time of placement, the chemical process of hydration will generate sufficient heat to prevent the concrete from freezing provided that precautions are taken to prevent heat loss. In mass

concrete applications, the greatest danger to the concrete is en the exposed

, surface areas, particularily at corners and other edges of the placement.

j It would be exceedingly rare for the mass of the concrete to freeze and

, sustain damage. These publications also indicate that even if frozen, the concrete will nomally cure to full design strengths if temperatures con-ducive to the hydration process are restored.

12. Allegations Relative To The As-Built Verification and Design Verification Activities.

During April 1983, NRC personnel received allegations to the effect that u - __. __. _-

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12 t- . . l l the QA group perfoming as-built verifications were not measuring support member dimensions and therefore, the " Vendor Certified Drawings" of the supports would not be accurate. A second allegation from the same person indicated that the QA group charged with responsibility for verifying that design changes have been incorporated into the plant and that the inspection lj} records for the installations accurately reflected that incorporation was

being required with the use of a computer generated status document to make the verification of records. The allegation was that the computer list-i} ing was faulty and therefore, the verification effort was equally faulted.

! The SRIC has examined each of these allegations as to the factualness of the i allegation and as to whether the allegation has or will have an effect on j jl the safety of the facility when operating. In regard to the first allega-

,- tion, the SRIC found that the allegation was and is factual. The allegation, l however, does not appear to have any significant impact on safety in that ,

j the as-built inspection was not developed to assure that the " Vendor Cer- ~

tified Drawing" was an accurate representation of the support in all aspects.

i The as-built program was established to assure only that the support loca-I tion on the . supported pipe and the direction of support is accurate for

! the purposes of perfoming the final pipe stress analysis. The responsibil-1 ity for assuring that the support members and other characteristics of the i- individual support reflect the design drawing requirements reside in other

! QA groups associated with the fabrication and installation, efforts. To also ,

i perfom these functions in the as-built verification inspection would be a

!, mdundant inspection that would not contribute significantly to the safety j, function of any given support.

Regarding the second allegation, the SRIC found that it too was factual but

only at the specifi.c time the allegation was made. When making the allega-l1 tion, the alleger provided the NRC personnel with a reference to a QC j'

inspection report which he said would fully display his concern. This report, identified as IR DCV-00421, was found to contain notation that the verification was based on a computer tabulation and that the report was

!a being completed at the direction of the inspector's supervisor. The original j report was dated April 4,1983. The pemanent file copy was found to have 1

been marked " voided" by the originating inspector as of May 20,1983, with j a notation that the report had been superceded by IR DCV-00423. This j latter inspection report was examined by the SRIC and found to document '

b essentially the same inspection effort by the same inspector but without ll any notation of having been based upon a computer tabulation and without notation of apparent protest of directions given by supervision. The SRIC interviewed the QC inspector who prepared and signed all of the

reports noted above in order to ascertain what had and is transpiring in l the QC design verification program effort. The inspector stated that the attempt to use the computer based data in the perfomance of the assigned task was in error from the beginning because of errors by persons genera-ting the computer data. The interviewee stated that only the one verifica-tion effort had been done using the computer based data and that all prior and subsequent verifications have been done by the assigned inspectors directly and personally examining the existent quality records in compli-ance with applicable QC procedures for the task. He stated that the only m; 2 -.. r . me_ n_ -_ _
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13 i . .

procedural deviation was the one instance stated in the allegation. Dis-cussions between the group supervisor at the time the allegation was

. received and the SRIC indicated that he had attempted to use the computer i .- tabulation to expedite the task on a trial basis by management direction t and that he had caused the original inspection report to be filed as it was i to give management a picture of the faults in the computerized data. It l thus appears that the design verification effort has been performed in I accordance with procedures except for the one-time pertubation that was

' subsequent correctly reaccomplished in accordance with approved proce-dures.

No violation to NRC requirements were revealed during this special

! inspection effort.

13. Improperly Certified Liquid Penetrant Examination Materials The CASE informed the Atomic Safety and Licensing Board by a letter dated --

M4y 18, 1983, of a potential problem with the liquid penetrant materials in use at the Comanche Peak Station. The letter stated that CASE had been made aware of the potential problem during a phone conversation with Charles A.

Atchison, who in turn learned of the " problem" from a Dallas area represen-g tative of the Magna-Flux Corporation, the o ginal manufacturer of the material.

The letter states that the problem surfac'.d only 7 to 10 days earlier. Based e on the date of the letter, it would seem that the problem arose between -

approximately May 8 to May 11,1983.

The situation bears close resemblance to the situation outlined beginning with NRC Inspection Report 50-445/82-18;50-446/82-09 based upon an inspection conducted during the period of September 7-10, 1982. The NRC inspector noted

that some certified test result documents had been altered by " pen and ink" j changes not immediately explainable. The matter was considered unresolved at that time. During a second inspection of the matter, conducted during

> November 1982 and documented in NRC Inspection Report 50-446/82-11, the j inspector found that previous corrective actions were not adequate and fur-ther that the " pen and ink" changes sometimes didn't match the type of

. material being certified. A Notice of Violation was issued as part of the inspection report on the matter. The licensee responded to the Notice of

. Violation by a letter dated December 21, 1982, wherein he stated that a l supplier had altered the certificates but that the original manufacturer

.: . had been able to furnish valid certificates and further, that all future

!! purchases would be direct from the manufacturer rather from a " middle-man" y supplier. The licensee also stated that specific receiving inspection pro-l' cedures had been implemented to prevent repetition. NRC Inspection Report i 50-445/83-10;50-446/83-05 documented verification that the Itcensee's actions were acceptable and the matter was closed.

It appears that the situation outlined in the CASE letter parallels the

! NRC findings in all details except for the dates which probably arose

! .as a result of misunderstood or incomplete communications between the I'

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Magna-Flux representative and Mr. Atchison and/or with CASE.

CASE also posed two, questions on the matter as follows:

l_

a. Has an NCR been written on this problem?

1 Answer: The above discussed inspection reports document a total of l five NCR's that were issued.

1 4

b. Has either TUGC0 or Texas Utilities or B&R notified the NRC of this problem?

! Answer: The roles of reportability were effectively reversed in that the NRC identified the problem and notified the licensee. _

A need for further NRC action on this matter has not been identified and the matter is. considered closed.

14. Penetration Seals t

4 This special inspection was undertaken to ascertain the validity and sig- -

nificance of allegations received initially by an NRC Headquarters Outy Officer on or about March 22, 1983, which were confirmed and added to during a telephone interview with the alleger on March 23, 1983, by the SRIC and a NRC inspector assigned to NRC Region I. The allegations, as understcod by the SRIC, were: s

a. The overlap seal for flexible boots should be 3 inches whereas 2 inches 1 is being used by BISCO.

s

b. There maybe a problem with the strength of the fabric used in the t

j flexible boots since the material supplier and BISCO are involved in l

a lawsuit.

c. The aggregate used in a radiation seal may separate giving rise to

. improper personnel protectioni i

l; Since BISCO was and is on the Comanche Peak site installing seals, Region IV was selected for the purpose of this special inspection although the com-

[.l pany has involvement at several other nuclear power sites throu hout the

  • United States. The SRIC obtained from the BISCO site manager a 1 of the i

I production and quality procedures appitcable~f6 the^ work at CPSES as well l as some that are not. The alleger specifically mentioned that the NRC should review Procedures QC-507, SP-504, SP-505, SP-505-1, and SP-505-2 in L Each of the above procedures regard to the flexible boot overlap problem.

i was in the books offered to the SRIC for review. A brief discussion fol-lows as to the contents of these procedures:

a. QCP-507: This procedure covers the final inspection of installed

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f flexible boots. The amount of overlap is not mentioned in

. the procedure, although the procedure does require that the seam be examined for evidence of poor sealing such as " fish-mouthing" which is taken to mean that the exposed edge of the overlap is puckered and not adhering to the base fabric.

b. SP-504: This procedure provides instructions and a calculation sheet to initially cut the fabric into a shape that would subse-quently allow the fonnation of a truncated cone. The fomula on the calculation sheet requires that 1-inch be added at each edge of the fan shaped fabric which is evidently to pro-
l. vide the overlap. The base fomula prior to adding the 1-inch provides a dimension just equal to the circumference of the pipe and/or sleeve to which the boot will be attached. ,

Thus, the 1-inch at each edge will provide for 2-inches ~

of overlap, assuming that the pipe and sleeve are concentric.

If pipe. and sleeve are not concentric, the resulting cone

. will be skewed and the seam overlap will be something other than 2-inches.

c. SP-505: This is a generic procedure for the installation of flex-

! ible boots. It was noted that the procedure requires that ,

the adhesive for the overlap seam be spread over a 3-inch depth from the fabric edge prior to fitting up the fabric where it is to be installed. Although not so stated, it appears that the 3-inch width of adhesive is to provide sufficient area of adhesive in the event the above men-tioned cone skewing occurs.

d. SP-505-1 and SP-505-2: These are additions to SP-505 having appli-cation when the boots are used as a simple pressure seal only and for when the boot is used as part of a fire pro-tection seal, respectively.

The SRIC interviewed the BISCO site manager as to whether the procedures had ever required a 3-inch overlap. The site manager indicated that 3-inch seam had been used up to sometime'in 1979 and that his homeoffice engin-eering had then changed the seal seam detail. The SRIC reviewed the results of a pressure differential test perfomed by BISCO in September 1979 which indicated that the fabric boot would withstand a differential pressure of 44 psig without sustaining damage. The project specification (2323-MS-38F) requires that the pressure seal maintain its integrity only up to 2 psig.

While the BISCO test data does not specifically state what the overlap seam width was on the test boot, it would strongly appear that the strength mar-gin is so high that even a reduction of 1/3 in the area of the overlap would have the effect of changing the safety factor from 22:1 to approximately 14:1.

It is the SRIC's conclusion that while the allegation relative to the reduction in seam from 3 to 2 inches is correct, the reduction would have no significant effect on the perfomance of the boot in service at CPSES and that, therefore, the allegation has no technical merit.

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16 Regarding the matter of the possibility of some undefined problem with the boot fabric, the BISCO site manager stated that his company has been engaged in a law suit with the supplier of the fabric but only in regard to the per-4 il formance of the fabric in one application which is understood to involve the

'j

tearing of the fabric after being punctured. It is understood that the puncturing has occurred when a gel type radiation seal hardens under radia-tion. Since the specific design involved is not scheduled for use at CPSES, the allegation has no technical merit.

4 Regarding the matter of possible separation of the radiation seal aggregate i

material from the carrier material, the SRIC can only conclude that the al-i legation is potentially correct but without apparent merit. The BISCO test l' reports indicate that the seals involved met the engineers specification.

! The separation of the aggregate (powdered lead) from the carrier (a silicone y material) would appear to be process sensitive in that if they are not well mixed, pockets of lead might form with resulting pockets of silicone without -- ,

sufficient lead. Since the specification and the BISCO procedures require  !

careful control and monitoring of the mixing process, the SRIC can only con-clude that these measures are effective in production operations as they were in preparation of the test samples. p

15. Electrical Cable Splicing The SRIC became aware that the Comanche Peak project electrical engineer

- had authorized the splicing of safety-related and auxiliary electrical cables within several control panels during the inspection period. Since

!' the licensee has committed in FSAR Section 8.1 to ccmply with IEEE 420,

" Trial-Use Guide for Class lE Control Switchboards for Nuclear Power Gener-

' ating Stations," .which forbids splicing of wiring in such panels, the SRIC

judged that the licensee was deviating from these commitments. The licen-
see. engineer indicated that he interpreted the IEEE standard to prohibit

- such splicing only between the cabinet terminal boards and the cabinet

,j devices and did not prohibit such splicing in the field run cables attach-ing to the terminal boards. , The engineer stated that action had been initiated with the NRC Offica of Nuclear Reactor Regulation to clarify the issue in the FSAR. The SRIC 2onfirmed that such action had been initiated by a telephone conversation with the NRR Licensing Program Manager for Comanche Peak. Pending action by NRR, this. matter will be considered as an unresolved matter.

H'

16. Unresolved Items Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, items of ncn-compliance, or deviations.

One such item, disclosed during the inspection, is discussed in paragraph 15 above. .This item is identified as " Splicing of Electrical Cables in Cabinets." (8324-01)

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17. Management Interviews ,

- The SRIC met with one or more of the persons identified in paragraph 1 of this report at' frequent intervals during the inspection period to '

i discuss the licensee's position and proposed actions on a significant number of issues which occurred during the period.

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