ML20199A605

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Rev 0 to Procedure CP-QAP-14.1, Insp of Storage & Maint of Mechanical Equipment
ML20199A605
Person / Time
Site: Comanche Peak, 05000000
Issue date: 01/12/1982
From:
BROWN & ROOT, INC. (SUBS. OF HALLIBURTON CO.)
To:
Shared Package
ML17192A346 List:
References
FOIA-85-313, FOIA-85-59, FOIA-86-A-20 CP-QAP-14.1, NUDOCS 8606120529
Download: ML20199A605 (9)
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BROWN & ROOT, INC. ISSUE CPSES NLMBER REVISION DATE PAGE -

J08 35-1195 CP-QAP-14.1 0

}AN 231NE 1 of 8

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TITLE: ORIGINATOR: .

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( DATE INSPECTION OF STORAGE AND MAINTENANCE OF REVIEWED BY: /4v /- n .p1 MECHANICAL EQUIPMENT / OATE APPROVED BY: su A 1/l%f82

'5ite QA ManagC- DATE

1.0 REFERENCES

f 1-A CP-MCP-10 " Storage and Storage Maintenance of Mechanical and Electrical Equipment" ,

1-B G&H Project No. 2323-03049 Equipment List 1-C CP-QAP-16.1, " Control of Nonconforming Items" .

C 2.0 GENERAL e. m.U G 2.1 PURPOSE . R

, The purpose of this instruction is to describe methods l utilized by Quality Centrol (QC) to ensure that items are stored and required maintenance is being performed on mechanical equipment in accordance with Reference 1-A.

l 2.2 SCOPE This instruction is applicable to all maintenance of mechanical equipment required by Reference 1-B, from the time of arrival of the equipment on the job site until the equipment is turned over to the Owner.

I 2.3 RESPONSIBILITY l

The QC Manager shall be responsible for impl ion of this procedure.

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COPY eer 8606120529 860604 PDR FOIA GARDE 86-A-20 PDR

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-c l Issut BROWN & ROOT, INC. ' NUMBER REVISION CATE PAGE CMU J08 35-1195 CP-QAP-14.1 0 JAN12 Igg 2 of 8 3.0 PROCEDURE P

3.1 STORAGE SURVEILLANCE A surveillance of each storage location shall be conducted ,

at least once per month by the Mechanical Equipment QC Group to assure that mechanical equipment is being stored in accordance with Reference 1-A.

Results of this surveillance shall be documented on a Storage Surveillance Report Attachment 1.

Minor storage violations such as: -

a. Missing or improperly installed caps on piping;
b. Piping subassemblies or canponents not properly stored i

on dunnage;

c.
  • Stainless steel stored in contact with carbon steel;
d. Heaters which are not energized; and

,e. Poor housekeeping may be reported on a Field Deficiency Report (FDR) in i

accordance with Reference 1-C.

! Other storage violations shall be reported as nonconforming j in accordance with Reference 1-C. -

J l 3.2 EQUIPMENT MAINTENANCE l Equipment Maintenance Record (EMR) Cards (Reference 1-A) shall be prepared for all mechanical equipment contained in Reference 1-8 which requires maintenance.

3.2.1 Inspection of Maintenance of Mechanical Eauipment Scheduled maintenance- dates are established by the

, Millwright Superintendent in'ac ,r i t.R f er ce 1-A.

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ISSUE BROWN & ROOT, INC. NUMBER REVISION DATE PAGE CPSES J05 35-1195 CP-QAP-14.1 0 JAN 13 g 3 of 8

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l The Millwright Superintendent shall notify QC when

maintenance is to be perfomed. Except as noted in

{ Section 3.2.2, the Mechanical Equipment QC inspector shall witness all required maintenance. QC shall verify that:-

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a. Equipment nameplate agrees with the EMR;
b. The required maintenance and maintenance frequency are ,,

listed on the EMR and the EMR has been approved by -

Engineering;

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c. Required maintenance has been perfomed and description '

of the maintenance perfomed and date perfomed entered on the EMR;

! d. Any calibrated instruments used in maintenance have

current calibration sticker and have had M&TE numbers and calibration due date entered under "Date" on EMR; Upon completion of a through ~d above, the Mechanical l Equipment QC inspector shall initial and date the " Verified j by QC" block.

3.2.2 ' Wa'1ver of Inspections ,

i Based on the complexity of required maintenance and/or a l degree of confidence established from previous maintenance j intervals, the Quality Engineering Supervisor may waive I witnessing up to and including 2 consecutive maintenance

! intervals on a particular iten. If this is done, the l Quality Engineering Supervisor shall note on the EMR that

witnessing of the particular maintenance interval has been i waived, and sign and date this note. Additionally, the Mechanical QC Equipment inspector shall verify that required maintenance has been perfomed, and enter a "V" in the
" Verified by QC" block on the EMR and 'date i adjacent to the "V". p'

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j ISSUL BROWN & ROOT. I.1C. NUM8ER REVISION DATE PAGE i

CPSES J08 35-1195 CP-QAP-14.1 0 JAN 12 19R2 4 of a f

3.3 EQUIPMENTMAINTENANCELOG The QC Mechanical Equipment Lead Inspector shall maintain an Equipment Maintenance Log (EML). Attachment 2. for all mechanical equipment which have EMR's. The Equipment Maintenance Log shall contain as a minimum: ,

a. Equipment description including tag number and serial number, if applicable;
b. Arrival date on job site;
c. Storage location (obtained from the Equiment Record Card or from Material Control personnel ). In the , ,

event that the storage location changes, the location noted on Attachment 1 shall be changed by drawing a single line through the old location and the new location entered. The change shall be initialed and dated by the person making the change; and

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d. Pemanent installation location including building and elevation. -

E: In the event that equipment is to be " Stored in Place" imediately after arrival on site, the storage location on the EML shall be noted

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t " Stored in Place".

( The EML shall be completed concurrently with the EMR to l

provide for a means of verifying maintenance is being l

t perfomed as required, l

The date of maintenance and the inspector's initials shall i appear in the appropriate month block. Additionally, the

! inspector shall indicate whether maintenance was witna<"

i or verified. I WIO

'; 4 ' 'Kj' t,,I d,.Mj EXAMPLE 1: l ,

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[ This example indicates that maintend '-pir$med 'n i i

the fif th through the eighth of the ir > And was yttnessed i by QC. ' ,

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Issue j i BR0'dN & ROOT, INC. NUMBER REVISION DATE .

PAGE ,

CPSES I j J05 35-1195 CP-QAP-14.1 0 AN 12 BE 5 of 8

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EXAMPLE 2: 4 l

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This example indicates that weekly maintenance was perfomed on the sixth, seventeenth, twenty-fourth, and thirty-first T i of the month, and was witnessed by QC on the sixth and twenty-fourth, and that QC verified that maintenance was

! perfomed on the seventeenth and thirty-tirst by reviewing

the EMR. -

) If the entry was logged on the EML by an inspector who initialed the EMR, then that inspector's initials shall be

added to the EML and the entry initialed and dated by the
, inspector logging the entry, r

,i EXAMPLE:

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f l An X shall be entered in the month the equipment is turned over to the Owner.

The Mechanical Equipment QC Lead Inspector shall review the Equipment Maintenance Log at the end of each day, week or

! month, as requi red, to assure that all maintenance is i

being perfomed, witnessed and/or verified as required.

l s A new Maintenance Log shall be established prior to January -

L of each year and shall encompass all equipment within the L .

scope of this instruction, with the exception of equipment

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s for which the Owner has accepted maintenance responsibilit j

Completed EML's shall be transmitted go h t WON 9 b

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ISSUE BROWN & ROOT, INC. . NUMBER REVISION DATE PAGE CPSES J08 35-1195 CP-QAP-14.1 0 M 12 1N2 6 of 8 i

3.4 LACK 0F REQUIRED MAINTENANCE The following conditions shall be reported by the

Mechanical Equipment QC inspector as nonconfoming in accordance with Reference 1-C.
a. Maintenance is not perfomed as required by the EMR;
b. Scheduled daily or weekly maintenance is not perfomed  ;

on schedule;

c. Scheduled monthly maintenance is not perfomed within seven days of the scheduled date; or -
d. Damage to equipment.

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BRoldN & ROOT, INC.  !$3UE C7553 NWSG . REVISION DATE PAGE J08 35-1195 CP-QAP-14.1 0 M32 M 7 of 8 ATTACHMENT 1 STORAGE SURVEILt.ANCE REPORT p

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CPSEE .4mSER . REVISION CATE PAGE J08 35-1195

CP-Q#.P-14.1 0 M11 NEZ 8 of 8 I

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JOB NO. 35-1195 12-21-79 TO: Historical Procedures File

SUBJECT:

B&R QAP Manual Revisions Designation Number Quality Assurance Procedure designation number CP-QAP-14.1 is teminated for the QAP titled " Field Control of Inspectaon Status of Items and Material", Rev. 9 The QAP procedure " Field Control of Inspection Status of Items and Materials" new number designa- -

tion is CP-QAP-15.'1.

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< . In Reply Refer To:

Dockets: 50-445/83-18 50-446/83-12

,g Texas utilities Generating Company ATTN: R. J. Gary, Executive Vice l President & General Manager I 2001 Bryan Tower Dallas, Texas 75201 Gentlemen:

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$ Thank you for your letter of June 26, 1983, in response to our letter and

( Notice of Violation dated May 31, 1983. Based on our review of your June 28, i

1983, letter and to discussions concerning your reply between Mr. T. F. I Westerman of this office and Mr. R. F. Heishman of the Office of Inspection and Enforcement, we have no further questions at this time and will review your corrective actions during a future inspection.

Sincerely, mriginal SL::ned by:

G.1:. MAOSEN" G. L. Madsen, Chief Reactor Project Branch 1 cc:

Texas Utilities Generating Company ATTN: H. C. Schmidt, Project Manager .

2001 Bryan Tower Dallas, Texas 75201 I

Texas Utilities Generating Company ATTN: B. R. Clements, Vice President, Nuclear

? 2001 Bryan Tower, Suite 1735 Dallas, Texas 75201 bec to DMB (IE01) bec distrib. by RIV:

D. Kelley, SRI-Ops R.-Taylor, SRI-Cons '

Section Chief (RPS-A) RIV File TEXAS STATE DEPT. OF HEALTH Enforcement Assistant i RPB1 %

i GMadsen/dsm g 7/* /83 o,,~-w \s-

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.... . June 28, 1983 TXX-3996 . ..

Mr. G. L. Madsen, Chief Reactor Project Branch 1 U.S. Nuclear Regulatory Comission ggg Office of Inspection and Enforcement 611 Ryan Plaza Drive, Suite 1000 Docket Nos.: 50-445 Arlington, TX 76012 -

50-446 COMANCHE PEAK STEAM ELECTRIC STATION RESPONSE TO NRC NOTICES OF VIOLATION INSPECTION REPORT NO. 83-18/12 g FILE NO.: 10130

Dear Mr. Madsen:

We have reviewed your letter dated May 31, 1983 on the construction appraisal inspection conducted by the Office of Inspection and Enforcement and by Mr. R. G. Taylor regarding Comanche Peak, Units 1 and 2. he have responded to the findings listed in Appendix A of that letter.

To aid in the understanding of our response, we have repeated the requirements and your findings followed by our corrective actions. We feel the enclosed information to be responsive to the Inspector's findings. If you have any questions, please advise.

Very truly yours, RJG:1n Enclosures .

cc: NRC Region IV - (0 + 1)

Director, Inspection & Enforcement (15 copies)

U. S. Nuclear Regulatory Comission Washington, DC 20555

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TXX-3996 NOTICE OF VIOLATION i Texas Utilities Generating Company Dockets: 50 445/83-18 Comanche Peak Steam Electric Station 50-446/83-12 Permits: CPPR-125 CPPR-127 Based on the results of an NRC inspection conducted during the period of February 14 - March 3,1983, and in accordance with the NRC Enforcement Policy (10 CFR Part 2, Appendix C), 47 FR 9987, dated March 9,1982, the following violations were identified:

1. Failure to Provide Adeouate Procedures, Instructions, or Drawing for Installation of Major Items of Eouloment Criterion V of Appendix B to 10CFR 50 requires that activities affecting quality shall be prescribed by documented instructions, procedures, or i drawings of a type, appropriate to the circumstances.

Project Specification 2323-MS-101 requires in paragraph 4.7.2 that the manufacturers recommendations and instructions for the installation of equipment be complied with by the installer, and in paragraph 4.12.3 requires that all bolting attaching rotating or vibrating equipment to the building structure be provided vith locking or with double nuts.

Contrary to the above, the Senior Resident Inspector (SRI) Construction determined by review of the " Construction Operations Traveler" for the _

installation of Heat Exchanger TCX-CSAHLD-01 that the recommendations by Westinghouse, the supplier, that the nuts attaching the component to the building be left loose to allow for thennal expansion were neither referenced nor included in the instructions and further verified that the nuts were fully tightened as installed. The NRC Construction Assessment Team (CAT) noted in NRC Inspection Report 50-445/83-18; 50-446/83-12 that j various items of like equipment such as tanks sometimes had only one nut on the attachment bolts, sometimes similar tanks had double nuts, and some j had a combination of single and double nuts on the same component.

l This is a Severity Level V Violation. (Supplement II.0) (50-445/8324-01; l 50-446/8315-01) i CORRECTIVE STEPS WHICH HAVE SEEN TAKEN AND THE RESULTS ACHIEVED:

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a. The specification and the construction installation procedures for mechanical equioment installation have been revised to clarify
bolting requirements for safety-related mechanical ecuipment.
o. A reinspection crogram has been instituted to verify mounting details on all safety-related mechanical equipment.

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TXX-3996 Page 2 CORRECTIVE STEPS WHICH '.!!LL BE, OR HAVE BEEN TAYsEN TO PRECLUDE RECURRENCE:

a. Project specifications and the installation and inspection procedures have been revised to clarify installation requirements,
b. Engineering, construction, and QA/QC personnel have been indoctrinated in the revised documents.

DATE OF PJLL COMPLIANCE:

a. Corrective action on revisions to specifications, procedures, and instructions, and indoctrination in the revisions have been completed.

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b. The reinspection program will be completed as soon as possible and no ~

later than room completion.

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TXX-3990 -

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2. Failure to provide Adecuate Maintenance of Materials and Ecuioment in Outocor Warenouse-Areas Criterion XIII of Appendix B to 10 CFR 50 requires that measures shall be '

established to prevent damage or deterioration of stored equipment.

l Brown & Root (B&R) Procedure MCP-10 " Storage and Storage Maintenance of -

Mechanical and Electrical Equipment," in paragraph 3.19 requires that materials to be stored outdoors shall be prepared for storage by the application of caps, preservatives or other means. Paragraph 3.11 requires that permanent carbon steel bolts supplied on equipment be protected with a suitable preservative.

Contrary to the above, the CAT identified various piping suoport j

! components with dry and dirty bearings and with rusting bolts and pins. -

The SRI during 91 tour of the storage areas for pipe support components noted many instances where bolts and pins were rusted and a few instances where tne support components were rusted through the paint and/or where the paint had chipped off.

  • This is a Severity Level V Violation. (Supplement II.0) (50-445/8324-02; 50-446/8315-02)

CORRECTIVE STEPS WHICH HAVE BEEN TAKEN AND THE RESULTS ACHIEVED:

i a. Construction oersonnel have corrected the storage conditions to prevent further deterioration and a preservation program has been established to clean, preserve, and/or paint any items showing signs of dirt or corrosion.

I b. QC inspection instructions have been revised to specifically address the verification of the material condition of items prior to installation.

CORRECTIVE STEPS WHICH HAVE BEEN, OR WILL BE TAKEN TO PRECLUDE REPETITION:

l a. Outside storage areas will be monitored on a periodic basis by QA to i assure the adequacy of the storage and preservation programs.

DATE OF FULL COMPLIANCE:

l Corrective action is conplete.

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i TXX-3996 Page 4

3. Failure to Remove Obsolete Drawings From Construction Work Areas l Criterion VI of Appendix B to 10 CFR 50 requires that only current drawings be used to accomplish work. l FSAR Section 17, paragraph 17.1.6 states that changed documents will be distributed to. and used at all work locations.

B&R Procedure DCP-3, "CPSES Document Control Program" provides that all revised drawings will be distributed to users on a controlled basis and that each user is accountable for the currentness of his files.

Contrary to the above, a CAT inspector identified a group of drawings in the construction work area that contained approximately 23 percent j drawings which were not of. the proper revision. Further, the CAT -

inspector noted that some drawings were sufficiently illegible in the title and revision block areas of the drawing to prevent identification of the drawing and/or the revision level.

This is a Severity Level IV Violation.

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(Supplement II.0)

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(50-445-8324-03; 50 446/8315-03)

CORRECTIVE STEPS WHICH HAVE BEEN TAKEN AND THE RESULTS ACHIEVED:

a. The area in which the out-of-date drawings were located has been audited by Document Control Center personnel and all out-of-date or damaged drawings have been replaced with current revision.

CORRECTIVE STEPS WHICH WILL BE, OR HAVE BEEN TAKEN TO PRECLUDE REPETITION:

a. A different process for reproducing drawings from microfilm is now in use which uses a incre durable paper thereby prolonging the life of the ,

drawings.

b. The present document control system for distribution of controlled documents from DCC to field file custodians is being replaced by a

" Satellite" system, staffed and controlled by the central DCC. These satellites will be responsible for controlling, replacing and issuing controlled documents.

DATE OF FULL COMPLIANCE:

a. The out-of-date and illegible drawings have been corrected.
b. All satellites are scheduled to be operational on or before August 1, 1983.

TXX-3996 '

page 5

4. Failure to provide Adecuate Control of Ventilation System Fabrication Criterion VII of Appendix B to 10CFR50, requires that measures be established to assure that materials, equipMnt, and services purchased through or from a contractor or subcontractor conform to the procurement '

documents. These measures must include as appropriate, source evaluations and inspections, generation of objective evidence of quality by the contractor or subcontractor, and assessment of the contractor or subcontractor's control of quality by the applicant.

Bannson Service Company Contract. 35-1195-0526, which incorporates Gibbs &

Hill, Inc. Specification 2323-MS-85, Section 2.1.4.d requires that all fabrication processes conducted by Bahnson Service Company shall be performed in accordance with dimensioned detail drawings. j Contrary to then above, it has been established by NRC inspections that subcontractor Bahnson Service Company, supplied:

a. A substantial number of welds, approximately 40 percent, on supports for ventilation ducting that are smaller than specified.by the design drawi ngs,
b. There were instances in which the bolting at duct joints was loose or missing and that gaskets in duct joints was missing or improperly install ed.
c. Five of the nine supports inspected had dimensional deficiencies.
d. The Bahnson Service Company quality control records failed to reflect either of the above conditions.
e. The licensee's audits and assessments of the subcontractor's control of quality failed to identify the above conditions.

This_is a Severity 1.evel IV Violation. (Supplement II.D)

(50-445/8324-04; 50-446-8315-04) i l

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TXX-3996 l.

page 6 l

a. A substantial number of welds, approximately 40 percent, on supports for

. ventilation ducting that are smaller than~ specified by the design drawings.

CORRECTIVE STEPS WHICH HAVE BEEN TAKEN AND THE RESULTS ACHIEVED:

A task force was established to inspect a sample of hangers, including weld size and length, to define the extent of the concern.

On the basis of this study, a computer sort of the HVAC hangers was conducted by 'the HVAC support designer, Corporate Consulting and Development Co., Ltd. (CCL). This sort, assuming that all hangers had the worst case conditions identified by the task force,

identified 239 hangers that would have loads in excess of 50% of OBE .

i allowable stresses using combinations that include full SSE load. On!

the basis of the computer analysis, CCL has determined that all hangers meet the functional design requirements with?ut -

modifications.

! Of the aforementioned 239 hangers,177 accessible hangers were

! included in a reinspection of a statistical sample of 280 hangers i with the highest stress value. This reinspection of 280 hangers e results in a 95% confidence level that at least 95% of the hanger population have welds larger than worst case conditions. The results of this reinspection have been transmitted to CCL for evaluation.

l CORRECTIVE STEPS WHICH WILL BE, OR HAVE BEEN TAKEN TO PRECLUDE RECURRENCE:

Welding personnel have undergone requalification and all inspectors have been retrained. The training program for inspectors has been 3 improved to provide an increased emphasis on practical inspection .

aspects. This includes reindoctrination and training in the use of inspection hardware including fillet gauges. The obtained

, proficiency level has been assessed by both Bahnson QA and by

experienced welding inspection personnel as part of a TUGCo audit.

j The level of inspector proficiency is now considered adequate.

In addition, Bahnson has assigned a QA Manager to the CPSES site. A 4

new Level III QC. Inspector has been assigned to provide closer day-to-day supervision and direction to the inspection force.

In addition, Bahnson Corporate QA and TUGCo QA will conduct audits of CPSES on a more frequent basis. TUGCo QA will perform audits of

} Bahnson on a quarterly basis placing emphasis on welding activities and configuration control, until all corrective actions are adequately implemented.

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_. TXX-3996 Page 7 DATE OF FULL COMPLIANCE:  ;

The CCL analysis of the reinspection results is expected to be l complete by July 1,1983. Requalification, training and assignment l of new personnel have been completed.

b. There were instances in which the bolting at duct joints was loose or missing and that gaskets in duct joints was missing er improperly installed.

CORRECTIVE STEPS WHICH HAVE BEEN TAKEN AND THE RESULTS AC"IEVED:

This concern is attributed to a failure to restore HVAC systems after ~

startup air Salancing activities. A reinspection of HVAC duct under an upgraded inspection program, defined by duct inspection procedure QCI-CPSES-003, is in progress to assure all systems have been restored to design requirements. ,

CORRECTIVE STEPS WHICH WILL BE, OR HAVE BEEN TAKEN TO AVOID FURTHER VIOLATIONS:

Bahnson has issued DFP-TUSI-013 which requires inspection and signoff to close Startup Work Authorizations (SWA's) and Startup Work Permits (SWP's). Bahnson QA has been added to distribution for SWA's and .

SWP's.

DATE OF FULL COMPLIANCE:

The reinspection is expected to be complete July 1,1983. Bahnson Procedures QCI-CPSES-003 and DFP-TUSI-013 are issued and in effect.

c. Five of the nine supports inspected had dimensional deficiencies.
CORRECTIVE STEPS WHICH HAVE BEEN TAKEN AND THE RESULTS ACHIEVED

Bahnson Service Co. (BSC) Procedure QCI-CPSES-014, Revision 0, now requires QC inspection and verification of hanger as-built drawings, including hanger location, configuration, member size and length.

Indoctrination and training of inspectors to this procedure have been completed. This procedure was implemented on the reinspection of the statistical sample discussed in item a. above, and the results transmitted to CCL for analysis.

TXX-3996 i

Page 8 CORRECTIVE STEPS WHICH WILL BE, CR HAVE BEEN tar.EN TO PRECLUDE RECURRENCE:

'he BSC inspectors have been trained and indoctrinated to the requirements of Procedure QCI-CPSES-014. <

Both Bahnson Corporate QA and TUGCo QA will provide an increased level of QA audit activity and emphasize configuration control in their audits.

In addition, the Bahnson QA Engineer at CPSES nas been replaced by a QA Manager, and a Level III QA Inspector has also been assigned to the Bahnson site QA organization.

DATE OF FULL COMPLIANCE: -

The CCL analysis of the reinspection results is expected to be complete by July 1,1983. All other corrective tctions are complete.

'd. The Bahnson Service Canpany quality control records failed to reflect either of the above conditions.

CORRECTIVE STEPS WHICH HAVE BEEN TAKEN AND T}dE RESULTS ACHIEVE Drocedure QCI-CPSES-003 has been revised to require the use of a Ductwork Installation Checklist and a Discrepancy Report, together with BSC drawings to conduct ductwork inspections. This procedure now provides a checklist for documentation of reinspection and closecut of observed deficiencies. The procedure further requires the retention of these documents as QA records. The procedure was issued and reindoctrination and training of inspectors has been completed. This procedure is being used in the reinspection program discussed in Item (a.) above.

CORRECTIVE STEPS WHICH WILL BE, OR HAVE BEEN TAKEN TO PRECLUDE RECURRENCE:

BSC Procedure QCI-CPSES-003, Rev. I has been issued. The BSC 4 inspectors have been reindoctrinated and trained to the requirecents of the revised procedure.

BSC Corporate QA and TUGCo QA will audit this activity on an on-going basis to verify compliance with this procedure.

4 DATE OF FULL IMPLEMENTATION:

Corrective action is complete.

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TXX-3996 Page 9

e. The licensee's audits and assessments of the subcontractors control of quality failed to identify the above conditions.

CORRECTIVE STEPS WHICH HAVE BEEN TAKEN AND THE RESULTS ACHIEVED:

In addition to active participation of the aforementioned corrective i actions, TUGCo QA, Dallas has performed two audits of Bahnson activities. The audit teams have included qualified weld inspectors.

CORRECTIVE STEPS.WHICH WILL BE, OR HAVE BEEN TAKEN TO PRECLUDE REPETITION:

TUGCo QA will provide an increased level of QA audit activity. The audit teams will be supplemented with certified welding inspectors, as appropriate, to provide increased emphasis in hardware related .

activities. I TUGCo QA will use certified weld inspectors as appropriate on audits of contractors where TUGCo QA does not perform or supervise physical

, inspection of work activities.

DATE OF FULL COMPLIANCE:

Audits were conducted on Bahnson during March and April . The increased audit activity will continue on an ongoing basis to assure adequate implementation.

1

C&oq \ Op h?l 0) ]

(

AUG 241983 In Reply Refer To:

Docket: 50-445/83-24 50-446/83-15 N

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Texas Utilities Generating Company

/f

, ATTN: R. J. Gary Executive Vice 2

President & General Manager

2001 Bryan Tower y

! Dallas Texas 75201 3 f g /7 Gentlemen:

This refers to the inspection conducted by our Senior Resident Inspector, e Construction, Mr. R. G. Taylor, during the period March through July 1983, ~

.of activities authorized by NRC Construction Permits CPPR-126 and CPPR-127 fcr Cananche Peak, Units 1 and 2, and to the discussion of our findings with

, Mr. R. G. Tolson, and other members of your staff during the inspection.

Areas examined during the inspection included review, inspection, and esalua-tion of several allegations made to various NRC persons, including the Atomic Safety a.9d 1.icensing Board in their proceedings regarding the operating license r

for Comanche Peak Steam Electric Station (CPSES). Within these areas, the inspection consisted of selective examination of procedures and representative records, interviews with personnel, and observations by the inspector. These findings are documented in the enclosed inspection report.

During this inspection, it was found that certain of your activities were in violation with NRC requirements. You were 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 inspection report. -

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

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

In accordance with 10 CFR 2.790(a), a copy of this letter and the enclosure t

will be placed in the NRC Public Document Room unless you notify this offica, by telephone, within 10 days of the date of this letter, and submit written

' application to withhold information contained therein within 30 days of the date of this letter. Such application must be consistent with the require-j ments of 2.790(b)(1).

ae>b.~--- M' BIT official File cop 7

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.AUG 24 g Should you have any questions concerning this inspection, we will be pleased to discuss them with you.

Sincerely,

= original signed Wa

. c, L,. M ADCEN

G. L. Padsen, Chief Reactor Project Branch 1

Enclosure:

Appendix - NRC Inspection Report 50-445/83-24 50-446/63-15 ,

i ec w/encls:

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

  • Texas Utilities Generating Company T-ATTH: B. R. Clements Vice President, Nuclear 2001 Bryan Tower, Suite 1735 rellas, Texas 75201 .

bec to DMB (IE01) bec distrib. by RIV:

RPB1 D. Kelley, SRI-Ops RPB2- R. ~ Taylor, SRI-Cons ' .

TPB Section Chief (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

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APPENDIX I

U. S. NUCLEAR REGULATORY COMMISSION REGION IV NRC Inspection Report: 50-445/83-24 50-446/83-15 Docket: 50-445. Category: A2 50-446 Licensee: Texas Utilities Generating Company (TUGCO) 2001 Bryan Tower Dallas, Texas, 75201 i

Facility Name: Comanche Pe'ak Steam Electric Station (CPSES), Units 1 and 2 Inspection At: Comanche Peak, Units 1 and 2, Glen Rose, Texas Inspection Conducted: March through July 1983 i

, Inspectors: bhX} e:,W R. G. Taylor, Senior Resident Inspector 9//7/83 Otte /

l

! Construction (SRIC) -

l Approved: h)&,M 8//f/83 i D. M. Hunnicutt, Chief Dhte '

' Reactor Project Section A Inspection Sumary l -

Inspection Conducted March through July 1983 (Report 50-445/83-24 and 83-446/83-15) 1 l Areas Insoected: Special inspections, announced and unannounced, related to

! allegations made to various NRC persons including the Atomic Safety and i

Licensing Board in their procedings regarding the operating license for Comanche Peak Station. The inspections involved 449 inspector-hours by one NRC inspector.

t Results: The inspection confirmed the need to issue four violations initially

icentified 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 Princical Licensee Employees
  • R. G. Tolson, Site QA Supervisor
  • C. T. Brandt, Non-ASME QC Supervisor
  • J. R. Merritt, Engineering, Construction and Startup Manager
  • J. B. George, Project General Manger
  • D. N. Chapman, QA Mana'ger
  • B. R. Clements, Vice-President, Nuclear Brown & Root (B&R) .

, .e.

  • G. R. Purdy, Project QA Manager
  • D. Frankum, Construction Project Manager 1

The SRIC also interviewed many other licensee, B&R, and subcontractor j

personnel during the course of the inspection,. -

j i

  • Denotes those persons who attended one or more management interUiews with the SRIC.
2. Licensee Action on Previous inspection Findinos (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 restraint essentially surrounding the mainsteam and feed i water lines located several feet outside of the ASME code boundry point.

  • l 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- ,

named " George Washington Bridge" by the site labor and quality forces. The l licensee had reported the finding of the defects as a potential 50.55(e) ittm to the SRIC on September 30, 1982, which was subsequently stated not reportable in a letter dated December 27, 1982. An NRC inspector followed up on the matter during a visit to the offices of the A/E, as documented i in NRC Inspection Report 50-445/03-12. This review pertained to all of the '

defects involved with the exception of two cracked welds that had not been analyzed at the time of the inspection. The engineer has recently analyzed these two defects and has determined that had they not been detected, the 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 and the functional application of the structure. Since the structure has no continuous service application and is essentially subject to a one-time loading, the cracks would not have the potential for further propagation.

Further, the cracks are at points in the structure that would receive rela-tively low stresses in the one-time impact based on their small size in relation to the members being welded. It appears that the cracks formed due to the stresses developed during the tightening of high strength bolting in

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

l 3

the imediate vicinity of the welds during the site assembly of the structure.

Taken in conjunction with the earlier documented review of the engineers

- calculations and the SRIC's review of these cracks, the SRIC has concluded that the engineer's o'verall analysis was adequate and that deficiency (s) were not reportable under 50.55(e). Both the licensee's initial report (CP-82-12) and the above identified unresolved item are considered closed.

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 the defects.

i The defects, including the cracks, have been documented on a nonconfomance report. The final disposition and closure of the NCR will be evaluated during future routine inspections. I i 3. Review of Licensee Self-Evaluation (Usina INPO Criteria).

l The SRIC has reviewed a report of the licensee's self- evaluation perfomed I during October 1982*which was based on criteria that has been developed for the purpose by INPO. The evaluation was performed in behalf of the lican-l see by personnel in the employment of Sargent & Lundy, an architect-engineer

fim with substantial nuclear power involvement. P. copy of the report was l furnished to ,the NRC, and subsequently, to the Atomic Safety and
Licensing Board in the matter of Comanche Peak Station operating license by. letter l

! dated May 2, 1983. The purpose of the review by the SRIC was to~ determine l

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 10 CFR 50.55(e).

The SRIC reviewed each of the 47 findings and the supporting documentation-in the report pertaining to each finding. This review revealed that none of the 47 items were based upon identified deficiencies in structures, systems, or components nor were there any significant deficiencies in design, engineering,(or 10CFR50.55e). testing that would constitute conditions reportable und .

4. Car Wash In Containment During the limited appearance statement portion of the Atomic Safety and l

Licensin'g Board hearing on Pay 16, 1983, a person stated at transcript l

page 6152 that he understood that the containment looked something like a car wash. The person stated that it was nis understanding that the situa- '

tion developed at about the same time that there was a meeting at the D/FW Airport between the NRC and any interested parties to discuss NRC decen-tralization. That meeting took place on April 5,1983. For the purposes of evaluatir.g this allegation, the SRIC expanded the period of interest to include the 3 weeks prior to the meeting. During this entire period, the Unit I reactor system was undergoing what is referred to as " Hot Func-tional Testing". This particular test is an accurate simulation of the operation of the reactor system and its appurtenances but without a reactor core being in place. The heat and pressure in the system is generated by the reactor coolant pumps in conjunction with the chemical and volume con-trol system charging pumps. The test could readily be construed to be a pressure test but in fact is an operational test at pressure. This parti-cular test extended overall for about 90 days beginning late in February

4 4

and continuing until late Pay. 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 the building and others to obtain a picture of the events that occurred in the Unit 1 Containment Building during the period of interest. The SRIO 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 leaks, generally at the gaskets between valve bodies and their bonnets. In addition, there was a considerable amount of condensation dripping 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 l during one phase of the test when a drain valve was partially open. Perhaps y

! this valve should have remained closed. The room in which the valve was a

located was apparently filled with steam vapor which would have condensed out on the cooler walls as water. On March 30, the reactor vessel head vent valves were partially opened, which in turn wo.uld give some amount of steam blowoff into the reactor refueling cavity area and would rise up into the building until cooled and c.ondensed out as water. None of these events l 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, i 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 i 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 -

l 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 a

[ seismic loao condition. Specific reference is made to Sheet 21 of Table 17A.

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 systems within the containment. CASE also states that all HVAC components

, and supports inside containment should be treated as missiles under Cri-l terion 4 of the General Design Criteria for Nuclear Pcwer 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 assumed by f CASE that this category excludes seismic loading in the design. This

! assumption is incorrect since the FSAR, Section 3.2.1.2 defines Seismic l Category II as being those portions of systems or components whose w.w. ~ . --- ~ = - , - _ = ~

b 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 t unacceptable safety level or could result in incapacitating injury to occupants of the control room. These systems are designated Non-Nuclear Safety (NNS) Seismic Categor a safe shutdown earthquakeSSE) (y II and are cause will not designed suchand constructed so that 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-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 2.

rather is composed pf two separate containment spray recirculation trains e each with 100 percent capacity. Each train contains two separate pumps, I

s one heat exhanger, and seven spray headers, and each system is fed from its individual electrical Class IE bus. The containment heat removal system is designed to ensure that the failure of any single active compon-ent, assuning the availability of either onsite or offsite power), exclusively, does not prevent,the system from accomplishing its planned safety-function, j 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-j ment Company, l.TD. " the support design consultant. It was detemined that o

all permanent HVAC supports are analyzed for seismic loading. Two methods

[ are utilized: Zero Peak Accleration (ZPA), or 1.5 Times the Peak Accelera-

  • tion When the Fundamental Frequency Falls Below 20 Hertz. Of the latter method of design, only about 6 out of 4000 supports have been designed that h way. A typical HVAC duct run is suppcrted 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 r

design of a typical HVAC duct run at elevation 852'-6" in the Auxiliary p Building. Supports were designed utilizing two computer programs entitled

? FEASA-20 and FEASA-3D. The acronym stands for frame eigenvalue and stress i analysis. The .?D version is used on the transverse supports and the -30

[ version is usui on the axial supports. The inclusion of equivalent weights e 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.

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

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

[

[ Fersons contacted during the course of the inspection at Corporate Consulting y

b -

I L _ . _ _ _ - ~. ~..= _ _ .__ _ - - _ - -

f 6

& Development Company, LTD. were:

J. Roland Yow, President & Chief Executive Officer Gary Hughes , Vice-President for Operations David Lindley, Principal Engineer i Stephen Lehrman, Seismic Department Manager

Daryl Hughes, Project Engineer
6. Heatino. Ventilation, and Air Conditioning System (HVAC)

During the CAT inspection (NRC Inspection Report 50-45/83-18;50-446/83-12),

the CAT inspectors noted that a significant portion of the welds on the ducting support structures were deficient in relation to the applicable welding code requirements. The doniinate deficient condition noted was that the welds were i significantly undersized. Based upon this information the SRIC toured various 1 areas of the facility with special emphasis on the ducting in the i 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 hinch leg size. The SRIC noted by visual comparison to the hinch thick base metal that very few of the welds were of l

proper size. The CAT inspectors also found cases where the bolting and gaskets between ducting sections.were loose and/or missing. .

The CAT inspectors also found that some support members were not;.

~

within the dimensional tolerances on the design drawings. It was-noted that the contractor's inspection records did not reveal these various facts, indicating ineffectual QC by the contractor. Further, a review of the licensee s audit program indicated that the licensee was unaware of these several problems in the fabrication, installation,

~

l i and inspection of the HVAC systems. Based upon the CAT inspectors'

findings and his .own observations, the SRIC reconsnended that a notice of violation be issued to the licensee pertaining collectively
to these matters (Notice of Violation issued on May 31, 1983.

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

I 7. Installation of Ma.ior Items of Equipment

! The CAT inspectors noted during their inspections of certain major

! items of equipment that there were several variables in how the l equipment was fastened to the building equipment pads. In some l instances, tanks for example, CAT inspectors found that there were i two nuts (double nuts) on the embedded bolts securing the equipment,

! other bolts had one nut. (single nut) and some had a combination of both single nuts and double nuts on one piece of equipment. The CAT personnel also noted that certain heat exchangers had slotted

holes in one of the mounting bases to allow for thermal expansion during operation. The holddown nuts appeared to be installed too
tightly and may have prevented freedom of movement. The SRIC obtained the design and installation drawings for two of the referenced heat exchangers identified in the CAT report. Both were found to be horizontal Utube heat exchangers whose function is nonsafety, but whose pressure boundary in the tubes is safety-related since the process fluid could be radioactive. The SRIC found that the construction drawings for the mounting pedestals had a. flat steel plate en one w -- n m _. - - - . ~ -

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7 Sb7b pedestal that would be suitable for the type of mounting detail on these heat exchangers. The SRIC then reviewed the installation travelers for each heat exchanger and found that these documents did not note or address the slotted details, the plate, or the fact the bolts should be left loose. The SRIC would note that the vendor manual which provides the details does not provide information

on how loose or tight the nuts should be nor how these nuts are to l

be locked at that looseness or some torque value. The SRIC with 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 force required to break the nut loose. On another heat exchanger
of comparable design, it was found that each stud was double nuted i 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  ;

I of looseness should allov sufficient freedom of movement. During a

. .the document review', the SRIC found that the engineer had specified 4

that all rotating and vibrating equipment should be double nutted and that other equipment could be secured with only one nut. No t document could be lochted that established the identity of vibrating l equipment nor were there any apparent provisi.ons made to lock nuts

(

where they must be deliberately left loose. .This was considered - I

overall to be a violation of Criterion V 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 1).

8. Maintenance of Equipment In Outdoor Storage Areas The CAT found that a considerable amount of equipment such as p'ipe support struts, elamps, and like items, normally stored . outdoors,

> was not being properly maintained in accordance with procedure MCP-10, l "Stdrage and Storage Maintenance of Mechanical and Electrical l Equipment", as evidenced by rusting bolts and adjustment screws on ,

struts . In addition, the strut bearings were dirty from dust and

. c the bearing load pins, in some instances, were rusted. By a tour ~D "

j , of the storage areas, the SRIC confirmed the CAT inspectors find- p l ings. The SRIC would also note that the INPO Self-Evaluation %g. 96,%

Report at page 111 describes essentially the same finding. This c c

M situation was detennined to be a violation of Criterion XI:I of ' '

L (y{ Appendix B to 10 CFR 50 (Notice of Violation issued on May 31, 1983.

Reference:

Notic..e of vim 4.ti.go iQ-445/83-18 8 and 50-446/83-12, VwoCT 13*= 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 the nuclear power block. It would appear that the various items involved i have been cleaned and restored prior to installation such that they can perform the required function.

9. Obsolete 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 l _

were incomplete in the title and revision blocks. The SRIC discussed

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 f The president of CASE in a letter dated March 11, 1983, addressed to
Mr. Richard C. DeYoung, Director of the NRC Office of Inspection and Enforce-
ment, indicated that CASE had received infomation from an unidentified spuri j to the effect that: -

4

a. There is field run conduit above the control room supported only by wire.

l b. There is drywall (or sheet rock) that is supported by wire.

j  : ..

There may be lights that are supp'orted by wire.

^

c.

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 i

the following findings:

a. There is a considerable amount of both safety-related and nonsafety
related conduit in the area above the suspended ceiling. The safety-related conduit is suppo-ted by Seismic Category I supports typical i
of those used in other areas of the facility. The nonsafety-related conduits are generally supported by simpler and less substantial sup-
ports that are typical of those that the SRIC has observed in large l open factories and are not designed to seismic standards. In each case exvnined, the non-seismic support was structurally paralleled with a small stainless steel cable that would assume the full weight of the conduit were the nomal support to fail in a seismic event.
b. The drywall materials were found to be part of the suspended ceiling above the central part of the control room and to fom a part of the sloping wall area below the control room observation room. These dry-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. ,

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c. The lighting fixtures in the control room are supported from an I intermediate substructure of "unistrut" by light-weight conduit.

f The substructure is likewise supported by the sane type of conduit i from the primary structure ceiling. The conduit used appears b to be the typical of that supporting the light fixtures in most offices with suspended ceilings. Paralled with each conduit are

~

i two small stainless steel cables which would assume the load if the conduit or its attachment were to fail. In the case i of the actual light fixtures, the cable is attached to the light fixture at the edge of the reflector assembly.

[ 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."

f The licensee has used terminology in the classification system that is at

! variance with that of the regulatory guide but is explained and defined l in Section 3.2 of the FSAR. In essence, the licensee has defined all .

safety-related items that must remain fully functional during and after a I seismic event as Seismic Category I. Items not having a safety function bu't whose failure could damage components which have a safety function or cause injury to the occupants of the control room during an event are referred to as Seismic Category II. In the case of the items involved in I this. allegation, all are Seismic Category II since their falling could I cause injury to the control operators. The cabling system described can '

be expected to prevent such a fall even though 'the normal supports rould j possibly fail. The stainless steel cable used in this design feature, I which at a short distance away looks much like bright galvanized common steel wire, is of relatively high strength. As an example, the test strength of -

an 1/8-inch cable is in excess of 1760 pounds. With four cables attached i to a light fixture, two at each end, the total support capability of the
cables is over 7000. pounds. It is apparent that the designers have elected to use conventional suspended ceiling and light fixture support techniques in order to use conventional and available materials and then provide a high strength backup support system in a seismic event.

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

j 11. Placement and Curing of Concrete During Freezing Weather During the limited public appearance portion of the Atomic Safety and f Licensing Board (Board) hearing conducted on May 15, 1983, there were two i 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 j staff as to whether there were any NRC personnel present with knowledge p

of the matter. The two references are at 6106 and 6134 of the hearing Also at 6109, an uni-transcript while the Board question is at 6109.

dentified voice responded to the Board that the matter had been reported

,. in IE inspection reports. Research of the NRC inspection reports revealed i that there had been such a discussion in NRC Inspection Report 50-445/77-01 i which was categorized as an unresolved item pending the licensee's review

$ and action on their finding of the problem. The unresolved item was f further discussed in NRC Inspection Report 50-445/77-04 with the closure of the item by an improvement 'in the QA procedures.

L

10 The SRIC has reviewed the matter, particularily with a view toward deter-

! mining whether the practices involved actually caused damage to the concrete 8 involved. The primary focus of NRC Inspection Report 50-445/77-01 (Details II, paragragh 5) was directed toward two licensee " Site Surveillance Reports" which had.been prepared approximately 2 weeks earlier than the inspection period covered by the inspection report. The first of the licensee's reports (C-134-77) was directed specifically to findings by a licensee inspector 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.

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 licensee inspector

! found that some surface areas were chilled to as low as 200F. The records l reflect, however, that there was disagreement 'between the B&R inspection v 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 actally mea-l suring the air temperature rather than the temperature of the concrete. No resolution of that disagreement was reflected in the records. The SRIC

. interviewed the licensee inspector of record during the course of this

' review to gain a clearer understanding of the events which took place.

l 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 tnough the surface i

temperatures were well below freezing. The records also reflect that in l order to resolve the issue, swiss hamer tests were run on the suspect

areas after the concrete had fully cured. These tests indicated that the

! suspect areas had attained strengths comparable to known properly cured l areas, indicating that the concrete had not been damaged even though the i possibility exists that it had been frozen for a period of time. The .

I records reflect that good concrete curing temperatures, i.e., above 40cF were established and maintained shortly after the licensee's inspector's l observation.

For the record, the SRIC would note that Placement 101-2801-001 took place '

in the Unit 1 Reactor Building. The placement became the open area floor  ;

at the lowest full floor in the building. This floor area, while suppor .

l l ting some equipment, serves primarily as a walk area. As such, it is fully topped with an architural cuncrete making the structural concrete no longer

accessable.

l

, NRC Inspection Report 50-445/77-01 also discussed comparable events to that documented on Surveillance Report C-135-77. One of these events was docu-l

, 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 temperature

'of Pl acement.105-2773-001, the foundation basemat for the Unit 1 Safeguards Building, were found frozen as evidenced by frozen wet burlap over certain areas that were not covered by insulating blankets. The records also  ;

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11 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 1

after 7 days there would be little free water in the concrete to freeze even though the burlap was froze. This c.onclusion is considered valid by the L . .SRIC based on his review of publications of the American Concrete Institute and the Bureau of Reclamation. Further, in responding to a se;arate finding that the field cure test cylinders made for the placement testei lower than '

allowed by the project specifications, swiss hammer tests were prfonned.

, The swiss hamer tests indicated the concrete placement had full specified l

strength. Relative to the low reported strengths of the field cure cylin-g ders, the SRIC would note that in his experience field cure cylinden will frequently test low under cold weather conditions. The reason is that the .

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

l either more vulnerable to freezing and/or curing much slower than nonnel he j 'to their depressed temperature.

i

! The final events covered by NRC Inspection Report 50-445/77-01 included i 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 i time perioil of 1976. In each case, the recor'ds reflect that the placements l 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 clo' sed the unresolved issue has stated that he had visually inspected each of the placements discussed in NRC Inspection Report 50-445/77-01 for evidenc: of i 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 i concern was for prevention of repetition rather than any specific concern i about the quality of the placements involved.

' The SRIC would note for the record that there are no regulatory or industry prohibitions on placing concrete in cold weather conditions. The American Concrete Institute and the Bureau of Reclamatioa 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 j concrete applications, the greatest danger to the concrete is on the exposed surface areas, particularily at corners and other edges of the placement.

t It would be exceedingly rare for the mass of the concrete to freeze and l sustain damage. These publications also indicate that even if frozen, the l concrete will normally 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 i Activities.

During April 1983, NRC personnel received allegations to the effect that l

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the QA group performing 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 j design changes have been incorporated into the plant and that the inspection

' 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-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 allegation and as to whether the allegation has or will have an effect on 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, however, does not appear to have any significant, impact on safety in that the as-built inspection was not developed to assure that the " Vendor Ctr- ~

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

The as-built program was established to assure only that the support loca-tion on the supported pipe and the direction of support is accurate for the purposes of perfonning the final pipe stress analysis. The responsibi1~-

ity for assuring that the support members and other characteristics of the individual support reflect the design drawing. requirements reside in other QA groups associated with the fabrication and insta11ation' efforts? To also perform these functions in the as-built verification inspection would be a redundant inspection that would not contribute significantly to the safety 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-tion, the alleger provided the NRC personnel with a reference to a QC 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 being completed at the direction of the inspector's supervisor. The original report was dated April 4,1983. The permanent file copy was found to have been marked " voided" by the originating inspector as of May 20,1983, with a notation that the report had been superceded by IR DCV-00423. This latter inspection report was examined by the SRIC and found to document essentially the same inspection effort by the same inspector but without 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 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

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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 i and that he had caused the original inspection report to be filed as it was j to give management a picture of the faults in the computerized data. It

thus appears that the design verification effort has been perfonned in

! accordance with procedures except for the one-time pertubation that was l ' subsequent correctly reaccomplished in accordance with approved proce-l dures. -

No violation to NRC requirements were revealed during this special inspection effort.

l i

13. Improperly Certified Liquid Penetrant Examination Materials f.

The CASE informed the Atomic Safety and Licensing Board by a letter dated I

! May 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-tative of the Magna-Flux Corporation, the orginal manufacturer of the material.

The letter states that the problem surfaced only 7 to 10 days earlier. Based 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 situa' tion outlined beginning I' 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 no'ted i that some certified test result documents had been altered by " pen and ink" changes not immediately explainable. The matter was considered unresolved at that time. During a second inspection of the matter, conducted during NoYember 1982 and documented in NRC Inspection Report 50-446/82-11. the 4 inspector found that previous corrective actions were not adequate and fur-ther that the " pen ar.d ink" changes sometimes didn't match the type of

. material being certified. A Notics. of Violation was issued as part of the j inspection report on the matter. The licensee responded to the Motice of Violation by a letter dated December 21, 1982, wherein he stated that a i supplier had altered the certificates but that the original manufacturer 3 had been able to furnish valid certificates and further, that all future S purchases would be direct from the manufacturer rather from a " middle-man" supplier. The licensee also stated that specific receiving inspection pro-cedures had been implemented to prevent repetition. NRC Inspection Report 50-445/83-10;50-446/83-05 documented verification that the licensee'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 l

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

CASE also posed two questions on the matter as follows:

a. Has an NCR been written on this problem?

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

b. Has either TUGC0 or Texas Utilities or BaR 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. .

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

14. Penetration Seals 4

This special inspection was undertaken to asc'ertain the validity.and sig-

nificance of allegations received initially' by an NRC Headquarters Duty
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 understood by  :

the SRIC, were: '

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

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b. There maybe a problem with the strength of the fabric used in the flexible boots since the material supplier and BISCO are involved in r*

a lawsuit.

, c. The aggregate used in a radiation seal may separate giving rise to i improper personnel protection.

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

pany has involvement at several other nuclear power sites throughout the i United States. The SRIC obtained from the BISCO site manager all of the ~

production and quality procedures applicable to the work at CPSES as well 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  ;

regard to the flexible boot overlap problem. Each of the above procedures .

, 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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