ML20134P378

Licensee Contractor and Vendor Inspection Status Report. Quarterly Report,April-June 1985. (White Book)

ML20134P378
Person / Time
Issue date:	08/31/1985
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE)
To:
References
NUREG-0040, NUREG-0040-V09-N02, NUREG-40, NUREG-40-V9-N2, NUDOCS 8509060260
Download: ML20134P378 (245)

Text

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NUREG-0040 l Vol. 9, No. 2 i

LICENSEE CONTRACTOR

, AND VENDOR INSPECTION

! STATUS REPORT  !

l l QUARTERLY REPORT l l APRIL 1985 - JUNE 1985 I

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Available from 1 Superintendent of Documents i

U.S. Government Printing Office Post Office Box 37082 Washington, D.C. 20013 7082 A year's subscription consists of 4 issues for this publication.

Single copies of this publication are available from National Technical

! Information Service, Springfield, VA 22161 i

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NUREG-0040 Vol. 9, No. 2 LICENSEE CONTRACTOR 4 AND VENDOR INSPECTION l

STATUS REPORT QUARTERLY REPORT APRll1985 JUNE 1985 oYe"fu$hI[$u* guy t985 Division of Quality Assurance, Vendor and Technical Training Center Programs Office of Inspection and Enforcement U.S. Nuclear Regulatory Commission i Washington, D.C. 20555 l

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CONTENTS PAGE

1. Preface................................ iii

2. Reporting Format (Sample).............. vii

3. Contractor with NRC Letters Confirming QA Program Implementation.............. ix

4. Sanple Letter.......................... xi

5. Inspector Reports...................... 1

6. Index of Inspection Reports. . . . . . . . . . . . 235

7. Table of Vendor Inspection Reports Rela ted to Reac tor Plants . . . . . . . . . . . . . . 239 I

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! PREFACE i

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! A fundamental premise of the Nuclear Regulatory Consnission's (NRC) nuclear j facility licensing and inspection program is that a licensee is responsible for j the proper construction and safe operation of nuclear power plants. The l total government-industry system for the inspection of nuclear facilities has

been designed to provide for multiple levels of inspection and verification.

! Licensees, contractors, and vendors each participate in a quality verification i process in accordance with requirements prescribed by, or consistent with,

{ NRC rules and regulations. The NRC inspects to determine whether its

requirements are being met by a licensee and his contractors, while the great bulk of the inspection activity is performed by the industry within the frame-l work of sequential ongoing quality verification programs.

j In implementing this multilayered approach, a licensee is responsible for i developing a detailed quality assurance (QA) plan as part of his license

{ application. This plan includes the QA programs of the lit.ensee's 1 contractors and vendors. The NRC reviews the licensee's and contractor's l QA plans to determine that implementation of the proposed QA program would be satisfactory and responsive to NRC regulations.

l Firms designing nuclear steam supply systems, architect engineering firms doing i design work on nuclear power plants, and certain selected vendors are currently

) inspected on a regular basis by the NRC. NRC inspectors, during periodic i inspections, ascertain through direct observation of selected activities j (including review of processes and selected hardware, discussions with j employees and selected record review) whether a licensee or contractor is satisfactorily implementing a QA program. If nonconformances with QA commitments are found, the inspected organization is requested to take appropriate corrective action and to institute preventive measures to preclude

recurrence.

! In addition to the QA program inspections, NRC also conducts reactive inspec-i tions of the licensee's contractors and vendors. These are special, limited l scope inspections to verify that organizations supplying safety-related l equipment or services to licensed facilities are exercising appropriate corrective / preventive measures when defects or conditions which could adversely affect the safe operation of such facilities are identified and that these

organizations are complying with the NRC requirements which govern the j evaluation and reporting of such conditions.

In the case of the principal licensee contractors, such as nuclear steam

! supply system designers and architect engineering finns, the NRC encourages j

i submittal of a description of corporate-wide QA programs for review and acceptance by the NRC. Upon acceptance by NRC, described QA programs provide

{ written bases for inspection on a generic basis, rather than with respect to specific consnitments made by a particular licensee. Once accepted by NRC, a corporate QA program of a licensee's contractor will be acceptable for all license applications that incorporate the program by reference in a Safety lii

Analysis Report (SAR). In such cases, a contractors's QA program will not be reviewed by the NRC as part of the licensing review process, provided that the incorporation in the SAR is without change or modification. However, new or revised regulations, Regulatory Guides, or Standard Review Plans affecting l QA program controls may be applied by the NRC to previously accepted QA programs.

The NRC Vendor Program Branch inspects the implementation of QA programs of nuclear steam supply system designers and architect engineering firms which have been submitted to and approved by the NRC in the form of Topical Reports l or Standardized Programs. Upon completion of inspections confirming satisfactory implementation of QA programs, NRC will issue a confirming letter to the nuclear steam system supplier or architect engineering firm.

Licensees and applicants that have referenced the NRC approved Topical Report, or Standardized Program, in SARs (or have adopted the total QA program described in the Topical Report or Standardized Program) may, at their option, use the confirming letter to fulfill their obligation under 10 CFR Part 50, Appendix B, J Criterion VII, that requires them to perform initial source evaluation eudits and subsequent periodic audits to verify QA program implementation. For additional details concerning the NRC letter, refer to " SAMPLE LETTER" included in this report.

Licensees or construction permit holders may choose not to make use of a contractor's NRC accepted program, or such an schepted program may not exist.

In such cases, the Vendor Program Branch inspections of nuclear steam supply system designers, architect engineering firms, or other licensee contractors, subtier contractors, or suppllers, will be based on programs developed to meet the commitments made by the licensee or construction permit holdcr. These inspections will not relieve the licensee or applicants from any inspection /

verification responsibilities required by Criterion VII.

The NRC currently is continuing their evaluation of proposed program for NRC acceptance of third-party (ASME) certification of Vendor QA programs. Should the proposed program be endorsed by NRC, it is anticipated that, subject to NRC audits of the third-party program, licensees and applicants would be able to use the ASME nuclear certification and inspection system to fulfill that part of their obligation under 10 CFR Part 50, Appendix B, Criterion VII, which required them to perform initial source evaluation / selection audits and subse-quent periodic audits to assess the QA program implementation.

A third party category of firms consists of organizations whose QA programs or manufacturing processes have not been reviewed and approved by NRC, or by a third party (such as ASME). This category of firms is subject to NRC inspection based on the safety significance and performance of products or services provided by such firms. Since such firms will not receive a third-party review of their QA programs, results of the direct NRC inspections may not be used to fulfill the licensee's obligations under Criterion VII.

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} The White Book contains information normally used to establish a " qualified I i -suppliers" list; however, the information contained in this document is not

adequate nor is it intended to stand by itself as a basis for qualification L of suppliers. ,

i l Correspondence with contractors and vendors relative to the inspection data 1

!  ! contained in the White Book is placed in the USNRC Public Document Room,  ;

located in Washington', D.C.

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i Coples uf.the White Bo'ok. may be obtained at a nominal cost by writing to

' the National-Technical Information Service, Springfield, Virginia 22161.

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ORGANIZATION: COMPANY, DIVISION CITY, STATE REPORT Docket / Year INSPECTION INSPECTION N0.: Sequence DATE(S): ON-SITE HOURS:

CORRESPONDENCE ADDRESS: Corporate Name SAMPLE PAGE Division (EXPLANATION OF FORMAT ATTN: Name/ Title AND TERMIN0 LOGY)

Address City / State / Zip Code l ORGANIZATIONAL CONTACT: Name/ Title TELEPHONE NUMBER: Telephone Number PRINCIPAL PRODUCT: Description of type of components, equipment, or services supplied.

NUCLEAR INDUSTRY ACTIVITY: Brief statement of scope of activity including percentage of organization effort, if applicable.

ASSIGNED INSPECTOR: Signature Name/VPB Section OTHER INSPECTOR (S): Name/VPB Section APPROVED BY: Signature Name/VPB Section INSPECTION BASES AND SCOPE:

A. BASES: Pertain to the inspection criteria that are applicable to the activity being inspection; i.e.,10 CFR Part 21, Appendix B to 10 CFR Part 50 and Safety Analysis Report or Topical Report comitments.

B. SCOPE: Sumarizes the specific QA program areas that were reviewed, and/or identifies plant systems, equipment or specific components that were inspected. For reactive (identified problem) inspections, the scope summarizes the problem that caused the inspection to be performed.

PLANT SITE APPLICABILITY: Lists docket numbers of licensed facilities for which equipment, services, or records were examined during the inspection.

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ORGA!IZATION: ORGANIZATION CITY,_ STATE I

REPORT INSPECTION NO.- RESULTS: PAGE 2 of 2 A. VIOLATIONS: Shown here are any inspection results determined to be in violation of Federal Regulations (such as 10 CFR Part 21) that are applicable to the organization being inspected.

B. NONCONFORMANCES: Shown here are any inspection results determined to be ,

in nonconformance with applicable commitments to NP.C requirements. In I addition to identifying the applicable NRC requirements, the specific industry codes and standards, company QA manual sections, or operating procedures which are used to implement these commitments may be referenced.

C. UNRESOLVED ITEMS: Shown here are inspection results about which more information is required in order to determine whether they are acceptable items or whether a violation or nonconformance mr.y exist. Such items will be resolved during subsequent inspections.

D. STATUS OF PREVIOUS INSPECTION FINDINGS: This section is used to identify the status of previously identified violations, items of nonconformance, and/or unresolved items until they are closed by appropriate action.

For all such items, and if closed, include a brief statement concerning action which closed the item. If this section is omitted, all previous inspection findings have been closed.

E.1 OTHER FINDINGS OR COMMENTS: This section is used to p'rovide significant ir. formation concerning the inspection areas identified under " Inspection Scope." Included are such items as mitigating circumstances concerning a violation or nonconformance, or statements concerning the limitations or depth _of inspection (sample size, type of review performed and special circumstances or concerns identified for possible followup). For reactive inspections, this section will be used to summarize the disposition or status of the condition or event which caused the inspection to be performed.

SAMPLE PAGE (EXPLANATION OF FORMAT AND TERMIN0 LOGY) viii f

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l CONTRACTOR WITH NRC LETTERS CONFIRMING QA PROGRAM IMPLEMENTATION l (See Next Page for Example of Confirming Letters)

CONTRACTOR TOPICAL REPORT REVISION DATE OF NRC LETTER Babcock & Wilcox BAW 10096A Revision 4 December 30, 1983 Stone & Webster SWSQAP 1-74A Revision D August 10, 1983 Westinghouse NTD WCAP-8370 Revision August 29, 1984 9A/5A 4

Bechtel - Gaithersburg BQ-TOP-1 Revision 3A November 2,1981 Bechtel - San Francisco BQ-TOP-1 Revision 3A June 12, 1981 Ebasco Services, Inc. ETR-1001 Revision 12 May 4, 1984 Combustion Engineering CENPD-210-A Revision 3 June 2, 1981 Gibbs & Hill, Inc. GIBSAR 17-A Amendment 8 February 27, 1985 United Engineers &

Constructors UEC-TR-001-3A Revision 6 September 16, 1962 General Electric Company NED0-11209-04A Revision 5 April 19, 1985 Sargent & Lundy Engineers SL-TR-1A Revision 6 April 14,1983 Bechtel - Los Angeles- BQ-TP-1 Revision 3A December 20, 1982 Gilbert / Commonwealth GAI-TR-106 Revision 3 May 24, 1984 Bechtel - Ann Arbor BQ-TP-1 Revision 2A May 7, 1981 l

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, k o UNITED STATES NUCLEAR REGULATORY COMMISSION g :y WASHINGTON, D C. 20555

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(ADDRESSEE)

Gentlemen: l l

A series of Nuclear Regulatory Comission (NRC) inspections have been conducted to review your implementation of the quality assurance program applicable to NRC applicants or licensees who have contracted for services from the 1 l (applicable corporate entity). These inspections consisted of selective '

examination of procedures and representative records, interview of personnel, and direct observation by the inspectors. As a result of these inspections, the NRC has concluded that the QA program described in Topical Report i is being implemented satisfactorily. Neither this conclusion nor the remainder of this letter applies to manufacturing activities or construction-related activities conducted at reactor sites.

Licensees and applicants that have referenced the above Topical Report in their 4 Safety Analysis Reports (or have adopted the total quality assurance program described in that Topical Report) may, at their option, use this letter to fulfill their obligation under 10 CFR Part 50, Appendix B, Criterion VII, that requires them to perform initial source evaluation / selection audits and subsequent periodic audits to assess the quality assurance program implementation.

The NRC expression of satisfaction with the implementation of your quality assurance program does not assure that a specific product or service offered by you to your customer is of acceptable quality, nor does it relieve the applicant or licensee from the general provision of Criterion VII which requires verification that purchased material, equipment, or services conform to the procurement documents. It is recognized that in some cases this assurance can -

be made by the applicant or licensee without audits or inspections at your facility.

. Continuing acceptability of implementation of your quality assurance program is contingent upon your maintaining a satisfactory level of program implemen-tation, certified through periodic NRC inspection, throughout all corporate organization units and nuclear projects encompassed by your program. Should your program implementation at any time be found unacceptable you will be notified by letter and requested to correct the deficiencies promptly. In the event you fail to correct the deficiencies promptly, or if the record of defi-ciencies is such as to indicate generally poor program implementation, you and the applicants and licensees who have referenced your quality assurance program j will be notified that the generic implementation of your program is no longer i

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(ADDRESSEE) (DATE) acceptable to the NRC. All of the audit / inspection requirements of Criterion VII, Appendix B,10 CFR Part 50, must then be implemented by the applicants or licensees. The NRC will reinstate its letter of acceptability 4

of implementation of your quality assurance program only after our inspectors have concluded, based on reinspection, that you have again demonstrated full compliance. 1 Except as noted above, the conclusions expressed in this letter will be effective for 3 years from the date of issue of the letter. At that time, program performance over the previous 3-year period will be evaluated and this letter reissued, if appropriate.

The results of our inspections are published quarterly in the Licensee Contractor and Vendor Inspection Status Report (NUREG 0040), which is made available to NRC facility applicants, licensees, contractors, and vendors as ,

well as to members of the public, by subscription.

Sincerely, i

Director Division of Quality Assurance, Vendor and Technical Training Center Programs Office of Inspection and Enforcement i

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I ORGANIZATION: THE ADVANCED PRODUCTS COMPANY NORTH HAVEN, CONNECTICUT l

REPORT INSPECTION INSPECTION N0.: 99900898/85-01 DATE(S): 1/28-2/1/85 ON-SITE HOURS: 28 CORRESPONDENCE ADDRESS: The Advanced Products Company ATTN: Mr. H. R. Sommer President 33 Defco Pork Road North Haven, Connecticut 06473 ORGANIZATIONAL CONTACT: Mr. J. J. Devine, QC Manager TELEPHONE NUMBER: (203)239-3341 PRINCIPAL PRODUCT: Metal-C-rings, 0-rings, spring-energized rings, and V-rings.

NUCLEAR INDUSTRY ACTIVITY: Approximately 3 percent of the 1984 sales.

ASSIGNED INSPECTOR: M%- 4 'f 85 J T. Conway, R6a ive Inspection Section (RIS) Date

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OTHER INSPECTOR (S): C. J. Czajkowski, Consultant

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APPROVED BY: . b J3!8(

d, E' W. Merktfoff, Chief, RIS, Vendor Program Branch Date INSPECTION BASES AND SCOPE:

A. BASES: 10 CFR Part 50, Appendix B and 10 CFR Part 21.

B. SCOPE: The purpose of this inspection was to conduct a programmatic evaluation of the implementation of the QA program in the areas of training / qualifications, procurement document control, control of purchased material / services, control of special processes, control of measuring and test equipment, audits (internal / external), and reporting of defects.

PLANT SITE APPLICABILITY: QA Data Packages for 0-rings: Oyster Creek (50-219),

Point Beach (50-266), Maine Yankee (50-309), Hatch (50-321), Prairie Island (50-282, 306), Salem (50-272), Kewaunee (50-305), Catawba (50-413), Arkansas Nuclear One (50-313), Beaver Valley (50-334), Connecticut Yankee (50-213),

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ORGANIZATION: THE ADVANCED PRODUCTS COMPANY NORTH HAVEN, CONNECTICUT REPORT INSPECTION NO.: 99930898/85-01 RESULTS: 3 AGE 2 of 8 PLANT SITE APPLICABILITY (continued): Wolf Creek (50-482), Summer (50-395),

Point Beach (50-266), Yankee Rowe (50-29), North Anna (50-338, 339),

Surry (50-280, 281), and Midland (50-329).

A. VIOLATIONS:

1. Contrary to Sections 21.6 and 21.21 of 10 CFR Part 21:

a. Copies of 10 CFR Part 21 and Section 206 of the Energy Reorganization Act were not posted.

b. Appropriate procedures to evaluate deviations or inform the licensee or purchaser of the deviation did not exist.

2. Contrary to Section 21.31 of 10 CFR Part 21, a review of 17 document-ation packages for nuclear 0-rings revealed that customer purchase orders (P0) to Advanced Products Company (APC) specified 10 CFR Part 21 as an applicable requirement, but APC P0s to tubing suppliers did not similarly specify that 10 CFR Part 21 would apply.

B. NONCONFORMANCES:

1. Contrary to Criterion V of Appendix B to 10 CFR Part 50, Section 1.1 of the Quality Control Manual (QCM), and subsection NCA-4134.5 of Section III of the Code, a review of procedures and specifications revealed the absence of instructions or procedures for the following activities:

a. Tube forming

b. Removal and rewelding of nonconforming welds

c. Comp,ession testing

d. Functioning of the material review board

2. Contrary to Criteria IV and V of Appendix B to 10 CFR Part 50 and Section 5.0 of Westinghouse document QCS-2 there was no documented i

evidence that APC identified QA requirements on P0s to suppliers of stainless steel and inconel tubing which were used as nuclear 0-rings.

3. Contrary to Criterion V of Apperdix B to 10 CFR Part 50 and Sections 3.1 and 6.2 of Addendum V of the QCM, there was no documented evidence that APC had audited and/or reviewed ar.d accepted the QA programs for the following subcontractors: J. Dirats and Company (chemical and 2

ORGANIZATION: THE ADVANCE 0 PRODUCTS COMPANY NORTH HAVEN, C0P.NECTICUT REPORT INSPECTION NO.: 99900898/85-01 RESULTS: ) AGE 3 of 8 physical testing), Reliable Cor Company (chemicals and testing)poration (silver supplier), and SELRE

4. Contrary to Criterion V of Appendix B to 10 CFR Part 50 and Section 9.2 of the QCM, it was noted that the last independent verification of silver plate thickness was accomplished on June 19, 1973.

5. Contrary to Criterion V of Appendix B to 10 CFR Part 50 and Section 6.0 of specification P1216, it was noted that:

a. Independent analyses of various plating solutions were not performed on the above schedule.

b. The make up control schedule was not adhered to for different solutions during the following time periods when two nuclear orders (0yster Creek in March 1980 and Point Beach 1 in August 1980) were inprocess:

ACID NEUTRALIZER from 8/13/79-5/14/81(21 months)

NICKEL STRIKE from 7/23/80-3/12/81 (P months)

SILVER STRIKE from 8/26/80-3/12/81 (7 months)

SILVER PLATE from 8/13/79-4/1/80 (7 months) frcm 8/26/80-5/12/81 (7 months)

6. Contrary to Criterion V of Appendix B to 10 CFR Part 50 and Sections 2.3.2 and 2.4.1 of the OCM, it was noted that:

a. The calibration control log only listed the date of next calibration, did not list the calibration inspector's name, and partially listed serial numbers,

b. The calibration certificate for the thermocouple wire purchased from Projects Inc. on P0 4197472 dated April 15, 1983, was not traceable to a National Bureau of Standards (NBS) test number.

7. Contrary to Criterion V of Appendix B to 10 CFR Part 50 and Sections 10 and 13.5 of ASTM A-213, a review of P0s for nuclear 0-rings and certified material test reports (CMTR) from Superior Tube and Tube Methods for stainless steel tubing fabricated to ASTM A-213 revealed that:

a. A product analysis was not performed on orders for Westinghouse (P0s 546-CPD-506987/-51332), Northern States Power (P0 A96569 MQ),

and Yankee Atomic Electric (P0 104313).

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ORGANIZATION: THE ADVANCED PRODUCTS COMPANY NORTH HAVEN, CONNECTICUT REPORT INSPECTION N0.: 99900898/85-01 RESULTS: PAGE 4 of 8

b. A hydrostatic test or nondestructive electric test was not performed on orders for Northern States Power (P0 A96569 MQ),

Yankee Atomic Electric (P0 104313), Westinghouse (P0 546-CPD-467588), and Wisconsin Electric (P0 B55724-S).

8. Contrary to Criterion V of Appendix B to 10 CFR Part 50; Section 1.3 of procedure P1057 and Sections 8.0, 9.6.1, and 9.7.3 of SNT-TC-1A, a review of records for nondestructive test personnel revealed the following:

a. Specific and practical examinations were missing for the two Level II personnel certified to perform dye penetrant and radiographic testing.

b. There was no statement that the two Level II personnel had satisfactorily completed training in accordance with procedure P1057.

c. There was no statement in procedure P1057 covering the duration of interrupted service requiring re-examination and recertification.

9. Contrary to Criterion V of Appendix B to 10 CFR Part 50 and Section 2 of procedure P1140, a review of data packages for nuclear 0-rings revealed that the Technical Check List was not signed or initialed by the QC Manager for P0s from Westinghouse (Nos. 506987, 49616, and 513332), Northern States Power (No. A96569MQ), Wisconsin Public Service (No. 59663), and Maine Yankee (No. 32594).

C. UNRESOLVED ITEMS:

None.

D. OTHER FINDINGS OR COMMENTS:

1. QA Data Packages - There were no reactive items or allegations pertaining to APC so the inspection consisted of record reviews and plant practice observations. Seventeen final QA data packages for nuclear 0-rings were reviewed for the following reactors:

Oyster Creek Beaver Valley 1 Hatch '

Point Beach 1 North Anna 1 & 2 Point Beach Midland Virgil Summer Prairie Island 1 & 2 Salem 1 & 2 Yankee Rowe Kewaunee Catawba 1 Surry 1 & 2 Connecticut Yankee Wolf Creek Maine Yankee Arkansas Nuclear 1

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ORGANIZAT10N: THE ADVANCED PRODUCTS COMPANY

?!0RTH HAVEN, CONNECTICUT REPORT INSPECTION N0.: 99900898/85-01 RESULTS: PAGE 5 of 8 A data package consisted of a customer P0, APC's P0 Technical Check-list, P0 tosupplier, CMTR or Certificate of Conformance (CC) from supplier, Receipt Inspection Report Inspection Traveler, Final Inspection Report, Material Certification, Radiographic Certification and APC's CC.

The following discrepancies were identified for several P0s:

1. Twenty-four P0s to suppliers were not approved by QA, and the P0 Technical Checklist was not signed off by the QC Manager (6) and by the President (3) (Ref. Nonconformance B.9).

2. P0 829436 (Northeast Utilities) dated January 16, 1984, for five sets of Inconel 718 rings - Station 5B " Final Inspection" was not signed off on the Inspection Traveler.

3. P0 32594 (Maine Yankee) dated June 24, 1982, for three sets of Inconel 718 rings - The Radiographic Certification was missing from the data package.

4. P0 A96569 fiQ (Northern States Power) dated January 6,1983, for 20 sets of stainless type 304 rings - The Radiographic Certifi-cation noted that one ring (S/N 1601) was rejected, rewelded, ard passed further radiographic examination. However, there was no documentation to show that a nonconformance report was generated or that procedures existed to remove and reweld the defective joint.

5. A comparison of information on CMTRs from Superior Tube and APC's material certification (MC) disclosed numerous transcribing differences as follows:

Reactor Mat. Code Item CMTR 3 f

Salem 1 A 2 - check analysis .063 .06 (C) check analysis 52.7 52.6 (Ni + Co)

Beaver Valley -

flatten test 3 2 Oyster Creek 3120 check analysis 5.28 5.38 (Cb + Ta) ,

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ORGANIZATION: THE ADVANCED PRODUCTS COMPANY NORTH HAVEN, CONNECTICUT REPORT INSPECTION NO.: 99900898/85-01 RESULTS: PAGE 6 of 8 Reactor Mat. Code Item CMTR MC Oyster Creek (continued) 3120 grain size 4 2 3329 grain size 4 2 1 3329 intergranular 4 2 l corrosion coupons 3330 check analysis 52.94 53.94 (Ni + Co) 3330 grain size 4 2 3330 flatten test 5 2 3330 intergranular 4 2 corrosion coupons Virgil Summer 3937 flatten test 1 4 Kewaunee 3827 flatten test 2 4 Connecticut Yankee 3892 flatten test 1 2 Yankee Rowe 3951 hydro test no yes liquid penetrant exam 10% 100%

2. Control of Purchased Material and Services - The inspector reviewed applicable sections of the QCM and four procedures relating to procurement and control of purcFased material and services.

Approximately 17 P0s to suppliers of stainless steel Type 304 and Inconel 718 tubing were reviewed to assure that applicable technical and quality requirements were included or referenced in P0s and that QA personnel reviewed and approved each P0 (ref. Nonconformance B.2).

It was noted that the tubing for the stainless steel nuclear 0-rings was ordered to ASTM A-213 which requires a chemical analysis (i.e.,

check analysis) be performed of either a billet or a tube from each heat and also a hydrostatic test or nondestructive electric test of each tube. A review of CMTRs from the tube suppliers indicated that a check analysis was not performed on three orders and testing was not performed on four orders for nuclear 0-rings, and there was no documented evidence as to why these requirements were waived (ref.

NonconformanceB.7).

External audit reports were reviewed from 1975 to the present to assure that material and services were purchased from qualified i vendors. APC performed periodic audits of their tubing suppliers -

Superior Tubing and Tube Methods, Inc. The audits reviewed appeared

! adequate, but it was noted that APC failed to perform audits of I

three suppliers of material aad services. The three suppliers were 6

ORGANIZATION: THE ADVANCED PRODUCTS COMPANY NORTH HAVEN, CONNECTICUT REPORT INSPECTION NO.: 99900898/85-01 RESULTS: PAGE 7 of 8 J. Dirats & Co. (chemical and physical testing), Reliable Corporation (silver supplier), and SELREX Company (chemicals erd testing) (ref.

Nonconformance 8.3),

3. Calibration of Measuring & Test Equipment (M&TE) - Records for M&TE and certifications for reference standards used by service vendors to calibrate M&TE were reviewed. An observation of MATE at various

- work stations was also performed to assure that M&TE are properly identified, controlled and calibrated at specified intervals.

The calibration status of M&TE was verified for the following items:

Radiographic Film Densitometer Furnaces (2)

Crystallab Model GS-20A Survey Meter (S/N 3375)

Calibration Blocks Various Flow Meters Western Arctronics Welding Machine Model 530 Chart Recorder Millivolt Potentiometer Thermocouple Wire The survey meter was last calibrated in August 1982, but this instrument is not included in the normal calibration program. It was noted that the calibration control log was not maintained in 1

accordance with the requirements of the QA program and that a calibration certificate for thermocouple wire was not traceable to NBS (ref. Nonconformance B.6).

4. Control of Special Processes - Applicable sections of the QCM, five nondestructive examination (NDE) procedures relating to liquid penetrant and radiographic examination, and three procedures for silver plating were reviewed to assure that special processes are controlled and accomplished by qualified personnel using qualified procedures. A review of the qualification records for NDE personnel and the procedure establishing APC's written policy on NDE did not fully meet the requirements of SNT-TC-1A (ref. Nonconformance B.8).

The silver plating solution control log was reviewed for various years from 1975 to the present. Various " gaps" were in evidence where the make up controls had not been instituted on the required frequency.

Discussions with appropriate personnel disclosed the fact that independent analyses of the solutions had not been accomplished regularly and that only once had the silver plating ever been tested t

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ORGANIZATION: THE ADVANCED PRODUCTS COMPANY NORTH HAVEN,. CONNECTICUT REPORT INSPECTION RESULTS: PAGE 8 of 8 N0.: 99900898/85-01 for thickness by independent verification (ref. Nonconformances B.4 and B.5).

Procedures - APC procedures used for the manufacture of nuclear j 5.

rings are delineated in a variety of APC specifications. Many of 1 these specifications have been tailored to a particular order or customer.

Approximately 17 procedures and specifications were reviewed, and the inspection and production areas were evaluated to assure that activities affecting quality are prescribed by an accomplished in accordance with documented procedures and specifications. It was noted that procedures did not exist for activities such as tube forming, removal and rewelding of nonconforming welds, compression testing, and the functioning of the material review board (ref.

Nonconformance B.1).

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ORGANIZATION: A&G ENGINEERING CO. II, INC.

ANAHEIM, CALIFORNIA REPORT INSPECTION INSPECTION NO.- 99901006/85-01 DATF(Si- a/90.5/9/nq nu_e t Tr unime . e, CORRESPONDENCE ADDRESS: A&G Engineering Co. II, Inc.

ATTN: Mr. Marvin Thomas President 4640 E. LaPalma Avenue Anaheim, California 92806 ORGANIZATIONAL CONTACT: Mr. John Thalasinos, QA Manager TFlFPHONF NIIMRFR- (71A) 770 Ainn PRINCIPAL PRODUCT: Ferrous & nonferrous bolting.

NUCLEAR INDUSTRY ACTIVITY: 10%

i ASSIGNED INSPECTOR: 7[ 'y 7 [k W// E /v/J/!AT N. J/ Miegel) Reactive Inspection Section (RIS) Date OTHERINSPECTOR(S): T. F. Burns, Consultant APPROVED BY: . .

28!05" E. W. Merschoff. Shfef.

s RIS. Vendor Prnaram Branch nato INSPECTION BASES AND SCOPE:

A. BASES: 10 CFR 50 Appendix B, 10 CFR 21.

i B. SCOPE: This inspection was performed as the result of an allegation made to the NRC concerning inadequate implementation of the A&G Engineering Co. II, Inc. (A&G) quality assurance program.

PLANT SITE APPLICABILITY: Not determined during this inspection.

9

ORGANIZATION: A&G ENGINEERING CO. II, INC. i ANAHEIM, CALIFORNIA REPORT INSPECTION NO.- 99901006/85-01 RESULTS: PAGE 2 of 8 A. VIOLATIONS:

None.

B. NONCONFORMANCES: l

1. Contrary to Criterion VII of 10 CFR 50 Appendix B and Section V, Paragraph II.C.1 of the A&G Engineering Co. 11. Inc. (A&G) Quality Assurance Manual (QAM) sufficient objective evidence was not i documented on vendor audit checklists to substantiate and/or verify the A&G auditor's findings.

2. Contrary to Criterion XII of 10 CFR 50 Appendix B, Section XX, I Paragraph 11.1.2 of the A&G QAM and A&G Quality Assurance Procedure (QAP) PR XX-A, Rev. A, paragraph VI:

a. the performance of the required " periodic checks" of measuring equipment could not be verified,

b. gauge 92A was incorrectly color coded.

4 C. UNRESOLVED ITEMS:

None.

D. OTHER FINDING! AND COMMENTS:

10 CFR Part 21 The procedures adopted by A&G to comply with 10 CFR Part 21 were reviewed and the inspector verified that the procedures were adequate. The posting of these procedures, 10 CFR Part 21, and Section 206 of the Energy Reorganization Act of 1974 was also verified. A&G has never made a Part 21 report, nor was the QA manager aware of any instance where A&G material was the subject of a Part 21 report issued by either an A&G customer or a utility.

There were no findings in this area of the inspection.

Documentation Review i Fourteen purchase order (PO) packages from eight different customers were i reviewed. The packages typically included: A&G's nuclear price quotation, certified material test reports, invoices, shipping manifests, A&G Certificetes of Compliance, the customer's incoming P0, A&G's P0 to a 1

1 10 1

ORGANIZATION: A&G ENGINEERING C0. II, INC.

ANAHEIM, CALIFORNIA REPORT INSPECTION NO.- 04401006/85-01 RESULTS: PAGE 3 of 8 vendor to procure material, A&G nuclear receiving inspection report, A&G nuclear receiving ticket, reports of physical testing results, and special customer requirements when applicable. The review was undertaken to ensure that A&G procures nuclear material and services from properly i qualified sources, correctly invokes federal and customer requirements on their vendor, maintains traceability of material, and properly implements the requirements of their quality assurance program. A review of the A&G Qualified Vendor List and selected vendor audits was also performed in conjunction with this P0 review. The vendor audits were 1

perfonned by personnel qualified in accordance with NQA-1 and using written checklists. However, the checklists were not always suited to the particular activity performed by the vendor, for example, calibration services or physical and chemical testing and analysis. Sufficient objective evidence to substantiate the audit findings (such as a list of procedures or P0s reviewed) was also generally lacking.

Nonconformance B.1 was identified in this area of the inspection.

It was not possible to review and evaluate A&G's procedure for upgrading material to ASME Section III Class 1 since all ASME Section III Class 1 material which A&G has supplied has been excluded from any additional NDE testing requirements due to its small size (i.e., less than one inch diameter).

Corrective Action Section XXIII, " Corrective Action," of the A&G QAM, the Corrective Action Request Log - Nuclear and 36 Corrective Action Reports (CARS) dated from January 20, 1982, through April 18, 1985, were reviewed.

All CARS examined had been properly completed and signed off by A&G personnel as required by the QAM. The appropriateness of the -

corrective action taken was not evaluated, i There were no findings in this area of the inspection.

Nonconformance Material Section XVII, " Nonconforming Material Control," of the A&G QAM, the Nonconforming Material Log, and 73 Nonconforming Material Reports (NMRs) dated from January 15, 1982 through April 4,1985, were reviewed. All NMRs reviewed had been completed and signed off by A&G personnel per the requirements of the QAM.

There were no findings in this area of the inspection.

11

ORGANIZATION: A&G ENGINEERING C0. II, INC.

ANAHEIM, CALIFORNIA REPORT INSPECTION MO.- 99901006/85-01 RESULTS: PAGE 4 of 8 ;

Internal AuditsSection XXII, " Audits" of the A&G QAM and internal audit reports for the years 1981, 1982, 1983, 1984, and 1985 were reviewed. All audits were performed within the prescribed twelve month time interval. However, 1 prior to 1984 the audit reports were generally inadequate and objective )

evidence to support the auditor's conclusions was not provided. In 1984 A&G hired a private firm to perform both the internal audits and management system audit. The private firm developed checklist questions and provided lists of the documents examined during the audits (for example: P0s, shop travellers,etc.).

There were no findings in this area of the inspection.

Plant Tour A tour was conducted of the A&G facilities to evaluate the adequacy of material control and storage activities. The facility was found to be clean and orderly. Raw materials were found to be in protective storage segregated by product form and size. Also, both raw and finished products were tagged or hard marked to identify the manufacturer / supplier, size, ASTM or AISI specification and quantity. One order being processed was identified as nuclear and was found to be segregated from commercial items, hard marked (die stamped) and further identified by a shop

traveler. The shop traveler identified the heat code customer, item, size, quantity, and each step in the manufacturing process.

There were no findings in this area of the inspection.

Calibration of Measuring and Test Equipment Section XX, " Control of Measuring and Testing Equipment," and calibration records (Gauge Control Records and Calibration Cards) dated from 1981 through April 1985 for nine different gauges were reviewed. A&G subcontracts to outside calibration services the task of calibrating their measuring and test equipment in accordance with a predetermined schedule. The gauges are color coded to identify the specific months when each must be calibrated. A " periodic check" as defined in Section XX is also required each time an instrument is used to ensure continued accuracy.

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ORGANIZATION: A&G ENGINEERING CO. II, INC.

ANAHEIM, CALIFORNIA REPORT INSPECTION Nn - 40401006/95-01 RESULTS: PAGE 5 of 8 Nine gauges were selected at random to verify that each was color l coded and calibrated correctly. One gauge, 92A, was found to be marked with an incorrect color code. It was also impossible to verify the performance of the " Periodic Checks." The check is to be performed by a quality control / quality assurance person who makes the pertinent entries on the Gauge Control Record. A review of the Gauge Control Records revealed that there were no entries indicating that the checks had been performed. Specific examples of such records are:

Gauge " Periodic Checks" not verified 6A 9/2, 8, 13, 15, 26 and 30 (1982) 1/22/83 to 6/1/83 6/21/83 to 12/1/83 12/1/83 to 6/1/84 23A 2/4/82 to 8/11/82 9/21/82 to 10/25/82 5/5/83 to 1/29/85 67A 4/14/82 (one day only) 5/3/82 to 6/21/82 i

7/12/82 to 9/23/82 3/18/82 to 9/2/83 9/12/83 to 6/16/84 Nonconformance B.2 was identified in this area of the inspection.

Training Section XXI, " Training of Personnel" of the A&G QAM was reviewed, and A&G training records were examined. Lesson plan content was examined for the period 1981 through 1985 and was found to be substantive.

Training was directed to those employees having duties and responsibil-ities affecting the quality of A&G products. Personnel who have received i

this training in,lude the president, QA manager, QA auditor, production foreman, purchasing agent, quality control inspectors, manufacturing manager, and document control clerk. Topics covered during training sessions included but were not limited to:

l a. examinations, tests, and reports l b. identification control of product

c. receipt inspection

d. audits j e. control of special processes 13

ORGANIZATION: A&G ENGINEERING CO. II, INC.

ANAHEIM, CALIFORNIA REPORT INSPECTION NO.- 99901006/85-01 RESULTS: PAGE 6 of 8

f. certification

g. QAM and subsequent revisions There were no findings in this area of the inspection.

Allegation An allegation was made to the NRC concerning inadequate implementation by A&G of their quality assurance program. This allegation was not substan-tiated in that, with the exception of several isolated instances (Nonconformances B.1 and B.2), the A&G quality assurance program is adequately implemented.

, E. PERSONS CONTACTED:

• J. Thalasinos, QA Manager, A&G

• M. Thomas, President, A&G

• W. Cook, Assistant QA, A&G F. DOCUMENTS EXAMINED:

1. QAM, Rev. C, 4/17/85, A&G Engineering Co. II, Inc. Quality Assurance Manual.

2. Documentation Package, 14 documentation packages for orders from 8 different customers packages typically included: - A&G Nuclear Price Quotation - CMTRs - Invoices - Shipping Manifests - A&G Certificates of Compliance - Customer's Incoming P0 - A&G P0 to Vendor to Procure Material - A&G Nuclear Receiving Inspection Report - Special Customer Requirements (when applicable) - A&G Nuclear Receiving Ticket - Report of Physical Testing Results.

3. Other, Approved Vendors List and Vendor Audit Checklists.

4. Other, 9 Shop Travelers.

5. Procedure, A&G Procedures for Compliance with 10 CFR Part 21.

6. Other, Exhibit XXXI, Rev. A, 5/7/82, Nonconforming Material Report Log.

7. QCD, 73 Nonconforming Material Reports dated 1/15/82 - 4/4/85.

l

• attended exit meeting 14

ORGANIZATION: A&G ENGINEERING C0. II, INC.

ANAHEIM, CALIFORNIA REPORT INSPECTION NO.- 99901006/85-01 RESULTS: PAGE 7 of 8

8. QCD, Exhibit XI, Rev. A, 5/7/82, Corrective Action Request Log -

Nuclear, j

9. QCD, 36 Corrective Action Reports dated 1/20/82 - 4/18/85.

10. Other, Internal Audit Reports - 1982, 1983, 1984, 1985.

11. Other, Auditor Qualification Records for two A&G Employees.

12. Internal Memo, 12 Semi-Annual Reports on the Status and Adequacy of the QA Program.

13. Other, Six reviews of ASME Code Addenda.

14. Lesson Plan, 1981-1985, Twenty Lesson Plans - Training of Personnel (for period 1981 thru 1985).

15. Purchase Order,12489-S,1985, Material " Upgrade" - 20 pcs. SA 540 Bolts, 8 pcs SA194 Nuts.

16. Purchase Order, H-58798, 2/1/85, Material " Upgrade" - 2000 pcs.

SA194 Gr. 6 Nuts.

17. Audit Report, 5/25/84, Audit Report of a material manufacturer.

18. Audit Report, 1983/1984, Audit Report of a material manufacturer, 12/6 & 7/83 and 11/29 & 30/84.

19. Audit Report, 5/18/84, Audit Report of a material manufacturer.

20. Audit Report, 6/15/84, A.:dit Report of a calibration service.

21. Audit Report, 3/11/85, Audit Report of a testing lab.

22. Purchase Order, 8E 104035,10/11/84, Cap Screws - Class 3, 1974 Edition 1, ASME III, SA 193 B7.

23. Purchase Order, 8E 204103, 10/23/84, Safety Related "Monel" B-164 Nuts.

24. Purchase Order, 8W 094124,10/2/84, Safety Related Bolts ASTM A193 B8M. ,

25. Purchase Order, 8K 084011, 8/29/84, Safety Related Bolts and Nuts SA 193 B8M, SA 194, 8.

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ORGANIZATION: A&G ENGINEERING CO. II, INC.

ANAHEIM, CALIFORNIA REPORT INSPECTION RESULTS: PAGE 8 of 8 NO.- 99901006/85-01

26. Purchase Order, 4R-67787, 7/12/84, ASTM A 193 Gr. B6 Main Steam Check Valve, Hex Head Bolts - Replacement Parts.

27. Purchase Order, 61057, 8/8/84, SA 193 Gr. B7 (10' length) SA 617 ,

l Visual, Bar Steel - All threaded inspection.

28. Audit Report, 4/30.84, Audit Report of a laboratory. '

29. Gauge Calibration Records, various documeat nos., 1982-1983 and 1984-1985, Nine Gauge Calibration Records 6A, 23A, 35A, 67A, 92A, 113A, 139A, 127A, 38.

j 16 l

ORGANIZATION: BABC0CK & WILC0X, A MCDERM0TT CO.

UTILITY POWER GENERATION DIVISION LYNCHBURG, VIRGINIA REPORT INSPECTION INSPECTION N0.: 99900400/85-01 DATE(S): 2/4-8/85 ON-SITE HOURS: 210 CORRESPONDENCE ADDRESS: Babcock & Wilcox, A McDermott Co.

Utility Power Generation Division ATTN: Mr. D. E. Guilbert, Vice President and General Manager Post Office Box 1260 l

Lynchburg, Virginia 24505 ORGANIZATIONAL CONTACT: Mr. T. Stevens, Nuclear QA Manager TELEPHONE NUMBER: (804) 385-3138 PRINCIPAL PRODUCT: Nuclear steam supply systems and nuclear cores.

NUCLEAR INDUSTRY ACTIVITY: The total effort committed to providing domestic nuclear steam systems and nuclear cores is approximately half of the Utility Power Generation Division. Principal activities include the design and procurement of these projects: Bellefonte, Units 1 and 2; and Washington Public Power Supply System, Unit 1, and providing engineering services under contracts and fuel reload contracts.

ASSIGNED INSP TOR: O bMO P. M. Sears, Special Projects Inspection Section (SPIS) Date OTHER INSPECTOR (S): W. Bannister (EG&G) J. Cozzual (EG&G)

W. Shier (BNL) R. Harris (EG&G)

B. Tollman (EG&G)

APPROVED BY: .

- MC / A.4 [oI 2thT

-Po(2. J. W. Craig, Chief, SPIS, W ndor Program Branch latd INSPECTION BASES AND SCOPE:

A. BASES: 10 CFR 50 Appendix B and Topical Report BAW-10096A.

B. SCOPE:

T7FTnspe(ct the implementation of B&W's QA program.1) Ascertain th PLANT SITE APPLICABILITY: Bellefonte (50-438, 50-439) and WNP (50-460) l i

17

ORGANIZATION: BABC0CK & WILC0X, A MCDERM0TT C0.

UTILITY POWER GENERATION DIVISION LYNCHBURG, VIRGINIA REPORT INSPECTION INSPECTION N0.: 99900400/85-01 DATE(S): 2/4-8/85 ON-SITE HOURS: 210 CORRESPONDENCE ADDRESS: Babcock & Wilcox, A McDermott Co.

Utility Power Generation Division ATTN: Mr. D. E. Guilbert, Vice President and General Manager Post Office Box 1260 Lynchburg, Virginia 24505 ORGANIZATIONAL CONTACT: Mr. T. Stevens, Nuclear QA Manager TELEPHONE NUMBER: (804) 385-3138 PRINCIPAL PRODUCT: Nuclear steam supply systems and nuclear cores.

NUCLEAR INDUSTRY ACTIVITY: The total effort committed to providing domestic nuclear steam systems and nuclear cores is approximately half of the Utility Power Generation Division. Principal activities include the design and procurement of these projects: Bellefonte, Units 1 and 2; and Washington Public Power Supply System, Unit 1, and providing engineering services under contracts and fuel reload contracts.

ASSIGNED INSP CTOR: O- SNO P. M. Sears, Special Projects Inspection Section (SPIS) Date OTHER INSPECTOR (S): W. Bannister (EG&G) J. Cozzual (EG&G)

W. Shier (BNL) R. Harris (EG&G)

B. Tollman (EG&G) l APPROVED BY: .

. MG / M

-902. J. W. Craig, Chief, SPIS, W ndor Program Branch h23.[Q3-latd INSPECTION BASES AND SCOPE:

A. BASES: 10 CFR 50 Appendix B and Topical Report BAW-10096A.

j B. SCOPE:

t T7) Tnspe(ct the implementation of B&W's QA program.1) Ascertain the l

PLANT SITE APPLICABILITY: Bellefonte (50-438, 50-439) andWNP(50-460) l 17

l

! ORGANIZATION: BABC0CK & WILC0X, A MCDERM0TT C0.

UTILITY POWER GENERATION DIVISION LYNCHBURG, VIRGINIA REPORT INSPECTION RESULTS: PAGE 2 of 12 NO.: 99900400/85-01

! A. VIOLATIONS:

None.

B. NONCONFORMANCES:

1. Contrary to Criterion XVI of 10 CFR 50 Appendix B, and B&W QA Topical Report BAW-10096A, Section 16, B&W does not have in place procedures requiring prompt evaluation of Potential Safety Concerns (PSC). For example, PSC 5-83, concerning the operability of the safety-related high pressure injection pumps during transient conditions, not included in the equipment specification, has not been evaluated in a timely manner such that corrective action, if necessary, could be implemented.

2. Contrary to Criterion VII of Appendix B to 10 CFR 50, B&W failed to ensure that adequate control measures over subcontractors were in place, resulting in certain components (Lambda Power Supplies) not being of adequate configuration or quality to assure (a) the function of the sub-supplier's product or (b) the proper output voltage charac-teristics for a certain range of input power and environmental conditions that are important to the B&W system function.

3. Contrary to the B&W Quality Assurance Program NPG-0402-01, the calculation of reactor protection system setpoints, as demonstrated in calculation number 32-1150653-00, Oconee 1, Cycle 9, RPS Set Point Calculations, do not contain information to demonstrate the accuracy of the noncertified setpoint calculations (personal computer computa-tions). Thus, the calculations are found to be in nonconformance with l

the above requirements regarding the use and verification of safety-related, noncertified computer calculations.

4. Contrary to Paragraph VII.I. of B&W Administrative Manual Procedure NPG-0902-06, Rev. 9 B&W did not perform sufficient testing of the T3 PIPE computer code to give reasonable assurance that the stress indices used in the ASME code Section III piping equations and, therefore, the results of these equations are correct.

C. UNRESOLVED ITEMS:

Potential Safety Concern PSC 17-83, "0vercooling Events at Low Reactor Power," identified a class of steam generator secondary side transients that, when initiated from low power levels, could produce a relatively high peak power without a reactor trip and thus, lead to violation of technical specifications. These transients, including excessive feed-18

ORGANIZATION: BABC0CK & WILC0X, A MCDERM0TT C0.

UTILITY POWER GENERATION DIilSION LYNCHBURG, VIRGINIA REPORT INSFECTION NO.: 99900400/85-01 RESULTS: PAGE 3 of 12 l

water addition during spurious turbine bypass or governor valve opening, are considered " moderate frequency events" by B&W. This PSC is applicable

. to all operating plants of the 177 fuel assembly (FA) design.

l This PSC was entered into the PSC file on June 30, 1983. An analysis performed in August 1983 and included in the PSC file, indicated that a slow overfeed transient to both steam generators "can produce unacceptable minimum (departure from nucleate boiling ratio) DNBR results" and thus violate the technical specification for the 177 FA plants.

B&W stated that additional analyses have indicated that the overfeed transient is not as severe as the original analysis indicated. Hcwever, this assessment is based on results obtained from a training simttlator and may have included reactor trip functions that are not safety grade and not normally included in safety analyses. B&W also stated that a mechanism to produce an overfeed transient of the type that could exceed technical specification limits has not been identified. However, no documentation of analyses supporting this statement was available.

The inspector stated that the analyses performed with the training simulator are considered "best estimate" calculations and not appropriate for the evaluation of technical specification limits. Furthermore, the use of a non-safety grade trip function is also inappropriate. It was also noted that this PSC has remained open for a considerable time

period (approximately 18 months) and dealt with a serious concern that could be considered reportable under 10 CFR 21. This item relates to Nonconformance B.l.

This item and the associated analyses will be subject to a detailed review during a future inspection.

D. STATUS OF PREVIOUS INSPECTION FINDINGS:

1. (Closed) Nonconformance (83-03): PSC 24-83 identified a concern that could have affected the technical specifications for all -plants with B&W Model 177 fuel assemblies. This PSC involved the analyses of the potential for fuel damage during accident situations.

B&W has completed the required analysis and the results indicate that, using a new methodology, the current operational and safety limits provide adequate margin to the fuel damage criteria and technical specification changes are not required. This PSC was closed in January 1983. The nonconformance is considered closed.

i I

19 i

ORGANIZATION: BABC0CK & WILC0X, A MCDERM0TT C0.

UTILITY POWER GENERATION DIVISION LYNCHBURG, VIRGINIA REPORT INSPECTION N0.: 99900400/85-01 RESULTS: PAGE 4 of 12

2. (0 pen) Nonconformance (83-03): This nonconformance concerned three computer code certification files for the CRAFT 2 code that were reviewed by the supervisor of the originator.

B&W QA procedure NPG 0403-11, " Technical Document Signatures," is being revised (Rev.12) to require that Unit Managers document that the reviewer was properly independent. The !! nit Manager will also be required to document the justification for reviewers who are not sufficiently independent. During this inspection a draft revised procedure was reviewed; however, it will not be issued until March 1, 1985. Thus, this item will be reviewed during a future inspection.

3. (0 pen) Nonconformance (84-03): This nonconformance was concerned with an uncertified computer code (the CORE code) that was used in a safety-related analysis.

The documentation required for the certification of the CORE computer code has been completed and is currently under review.

As part of the preventive actions associated with this nonconformance, B&W has prepared a list of uncertified computer codes used for safety-related calculations in the Engineering Department. In addition, QA Procedure NPG-0902-06, " Computer Program Development and Certification' has been revised to preclude the certification exemption for any computer code that performs safety-related calculations. This procedure is scheduled to be issued on March 1, 1985. This item will be reviewed during the next inspection.

4. (Closed) Nonconformance (84-03): Two modeling additions to the small break version of the CRAFT 2 code (i.e., non-equilibrium pressurizer and primary metal heat structure model) were not properly verified.

A review of the CRAFT 2 certification file indicated that the two models have been properly verified against hand calculations inde-pendent of the CRAFT 2 code and then tested for operability with the CRAFT 2 code structure. In addition, a training session was held for all computer code responsible engineers to emphasize the need for a clear description of test cases and verification records in code certification files. This nonconformance is considered closed.

5. (Closed) Nonconformance (84-03): This nonconformance was concerned with the incomplete explanation of the adequacy of agreement with experimental data for a test problem associated with a steam generator modification in the CRAFT 2 computer code.

20 i

1

ORGANIZATION: BABC0CK & WILCOX, A MCDERM0TT CO. )

UTILITY POWER GENERATION DIVISION i LYNCHBURG, VIRGINIA REPORT INSPECTION NO.: 99900400/85-01 RESULTS: PAGE 5 of 12 The certification file for CRAFT 2 computer code was reviewed and an adequate explanation of the comparison of CRAFT 2 calculation with Alliance Research Center (ARC) Test 13 for a once-through steam l generator,was observed. As part of the preventive 3ctions B&W has implemented conceving this nonconformance, a training session was held for all computer code responsible engineers to emphasize the need for clear identification and explanation of verification calculations.

This nonconformance is considered closed.

6. (0 pen) Nonconformance (84-03): This item concerned the description of computer code limitations in computer program manuals.

B&W QA procedure NPG 903-03, " Development and Control of Computer Program Manuals" is being revised to reflect changes in the computer code limitation description. This revision (Rev. 10) is scheduled to be issued on March 1, 1985 and this item will be reviewed during a future inspection.

E. OTHER FINDINGS OR COMMENTS:

1. Potential Safety Concerns B&W QA Procedure NPG-1707-01, " Processing Safety Concerns," requires that records be established that document any concern which has been discovered during design, analysis, fabrication, installation, testing, inspection, training and operations activities of a nuclear power plant and which has or may have safety implications.

These records are identified as PSCs prior to the completion of the evaluation of the need to report the item to the NRC per 10 CFR 21.

As part of this inspection, the PSC files were reviewed for the period of January 1983 through January 1985 and the findings are described below.

a. PSC 9-83, TAC 0 Code Error: This PSC concerns a computer code error that was identified in the TAC 02 computer code. The TAC 02 code is used to calculate the fuel rod internal pressure that is used in loss of coolant accident analyses. This item was entered into the PSC file in April 1983 and closed on September 20, 1983.

b. PSC 3-83, Effect of Non-condensibles During Small Break Loss of

+

CoolantAccident(SBLOCA): This PSC concerns the possibility that noncondensible gases generated in the primary system follow-ing SBLOCA could cause erroneous instrumentation readings and indicate a higher than actual degree of primary subcooling, thus l

I 21

ORGANIZATICN: BABC0CK & WILC0X, A MCDERM0TT C0.

UTILITY POWER GENERATION DIVISION LYNCHBURG, VIRGINIA REPORT INSPECTION N0.: 99900400/85-01 RESULTS: PAGE 6 of 12 belayingactuationofvarioussafetysystems(i.e.,highpressure injection and auxiliary feedwater). This PSC was more signifi-cant'for plants with the 205 fuel assembly (FA) design (Bellefonta I 1 and 2) than the 177 FA plants due to higher subcooled margin available in the 177 FA designs. This item was entered into PSC

- the file in January 1983 and closed in November 1983.

c. PSC 5-83, HPI Pump Thermal Shock / Thermal Stress: This PSC concerns a potential thermal stress problem with high pressure injection (HPI) and makeup (MU) water systems following activa-tion of the HPI pumps. The concern is applicable to all B&W plants. This item was entered into the PSC file in January 1984 and has not been closed.

d. PSC 1-84, Excessive OTSG Steam Velocities: This PSC concerns the potential for excessive steam exit velocities from the steam generator during transient conditions and the possibility of flow induced vibrations (FIV) resulting in steam generator tube ruptures. The item was entered into the PSC file in January 1984 and has not been closed.

e. PSC 17-83, Overcooling Events at Low Reactor Power: This PSC concerns certain secondary side steam generator transients (e.g., excessive feedwater addition and spurions turbine bypass or gover.nor valve opening) that can lead to high reactor power levels without reactor trip when initiated from low power levels.

This situation could result in exceeding departure from nucleate boiling ratio (DNBR) safety limits and thus, violate the technical specification for operating B&W plants. This PSC is applicable to all operating plants with the 177 FA design.

This item was ent'ered into the file on June 30, 1983 and has not been closed. A preliminary analysis was documented which indicated that a slow overfeed transient to both steam generators would not generate a reactor trip and possibly exceed DNBR safety limits. During this inspection, B&W stated that additional analyses have indicated that the overfeed transient was not as severe as originally indicated. However, the inspector stated

> that the documentation of the additional analyses was not suffi-cient to support this conclusion. For example, the additional analyses were performed using a training simulator that may have taken credit for a non-safety grade trip function. The training simulator is considered a "best estimate" analytical method and is not normally considered applicable safety analyses supporting i

22 l

l .

1 ORGANIZATION: BABC0CK & WILC0X, A MCDERM0TT C0.

UTILITY POWER GENERATION DIVISION l LYNCHBURG, VIRGINIA REPORT INSPECTION NO.: 99900400/85-01 RESULTS: PAGE 7 of 12 technical specification limits. Furthermore, non-safety grade protective functions are not included in safety analyses unless they affect the calculation in an adverse manner.

B&W tlso stated that a mechanism to produce an overfeed event that exceeded technical specification limits has not been identi-fied. However, documentation of analysis to support this statement was not available.

One nonconformance (see Section B.3 above) was identified in this part of the inspection.

f. PSC 18-83, Unanswered-Questions and Problems Identified as Part of a Design Review Board (DRB) Meeting on TVA-Bellefonte in August 1983: This PSC concerned questions raised during a DRB meeting held in August 1983. A PSC was received by B&W Licensing on September 9, 1983. In December of 1984, B&W determined that the FSC did not represent a reportable item based on the fact that all of the questions and problem areas were identified as part of the normal design process, which includes a DRB review

of the design. A major hardware redesign'and rework program was scheduled and implemented to resolve this PSC.

g. PSC 22-83, Potential Safety Concern Relating to Certification of Computer Prograns: This PSC concerns procurement of computer programs from vendors without certification for nuclear use, and subsequent use by B&W customers as commerical-grade items (i.e.,

not certified for safety-related application). During the period of January 1982 through October 1983, GPU Nuclear (GPU) made use of a number of coniputer programs in the B&W Ccmputer Sciences Library under the impression that these programs were certified for nuclear use when, in actuality, the programs were not so certified. This condition was discovered by GPU during an audit of B&W Computer Services on October 19, 1983.

B&W issued a preliminary report of safety concern on October 25, 1983. Subsequently, however, in January 1985, B&W concluded

, that the concern did not constitute a substantial safety hazard or reportable condition under 10 CFR 21. Corrective actions by B&W included adding.several of the programs.in question (ANSYS, GT STRUD, and ADL PIPE) to B&W's Quality Assured Library, thus qualifying their use for safety-related work. Also, error notice i ,

! have been sent to GPU so that they can determine the extent to which program errors may have affected the results of calculation  ;

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23

0 ORGANIZATION: BABC0CK & WILC0X, A MCDERM0TT C0.

- UTILITY POWER GENERATION DIVISION LYNCHBURG, VIRGINIA REPORT U!SPECTION NO.: 99900400/85-01 RESULTS: PAGE 8 of 12

, which were performed. This item was entered into the PSC file in October 1983 and closed in January 1985.

h. PSC 23-83, Lambda Power Supplies: This PSC concerns numerous failures the power supplies manufactured by Lambda Power Supplies which have been reported using B&W site Problem Reports (SRPs).

B&W's investigation of the Lambda Power Supplies problems revealed a generic quality problem of varying configuration and poor workmanship. B&W's failure to ensure that adequate controls and checks were placed on the subcontractor of procured materials resulted in Nonconformance B.2.

Hardware problems were identified in Octaber 1983 in site Problem Report SPR 13-16-337. System reliability, functional ability, and/or performance characteristics have been shown to be affected

'N* by the lack of capacitor "C26" or diode "CR6" or both. The fact that the Lambda Power Supplies components are a continuing prob-lem is evidenced by Bellefonte Unit 1 Site Problem Reports SPR-13-15-0683, dated November 9, 1984, and SPR 13-15-700, dated December 20, 1984. This probl.em is also evident in Bellefonte Unit 2 Site Problem Report SPR 13-16-33, Revision 5, dated October 26, 1984.

Lambda stated on November 4,1983, that the power supplies should have both the capacitor and the diode, to ensure proper operation.

However, specific problems being identified a year later include:

C26 capacitor missing; C26 capacitor polarity reversed; CR6 diode missing; corroded transformer.

B&W's emphasis has been primarily on hardware correction since the PSC was opened on November 11, 1983, A " front end" meeting was held on August 22, 1984, after timeliness was addressed by the NRC (NRC. Inspection Report 99900400/84-03). On October 26, 1984, B&W determined that this PSC was not reportable based on an analysis that-determined that the Bellefonte reactors will be shut down upon failure of the power supplies. However, these i

l shutdowns also will cause thermal transients which are not j desirable. A B&W Site Instruction i_s being sent to TVA i Bellefonte sites to correct Lambda power supplies that do not have the capacitor and the diode properly installed.

i. PSC 7-78 Nonconservative Analysis of Neutron Flux as it Relates to Bellefonte. Nuclear, Plant Units 1 and 2: On February 2, 1978,.

this PSC was issued concerning significant differences between 24

,a .

l ORGANIZATION: BABC0CK & WILC0X, A MCDERM0TT C0.

UTILITY POWER GENERATION DIVISION LYNCHBURG, VIRGINIA REPORT INSPECTION NO.: 99900400/85-01 RESULTS: PAGE 9 of 12 excore indicated power and heat balance observed during a power calibration measurement at one of the Oconee plants. Stemming from this concern, B&W identified (in October 1980) a potential I problem with assumed measurement errors used in determining reactor protection system (RPS) setpoints that may be noncon-servative under specific plant conditions for Bellefonte Nuclear Plant Units 1 and 2.

As of January 1985, B&W had completed their analysis program of this condition and have determined that no changes to plant hardware or setpoints are necessary, although initial condition normal operating limits on rod position and axial offset were revised. B&W has also determined that no further actions are required to prevent a reoccurrence of this problem at BLN.

Auditing of this concern included a review of the Summary Report for B&W's program to resolve the flux measurement error concern,

j. PSC 28-79 and PSC 40-80, Attached Piping: This PSC concerns an inconsistency between the assumptions relative to pipe breaks in the loss-of-coolant accident (LOCA) analysis and the structural analysis of certain connecting pipes in the affected or broken loop.

This item was reported to B&W by TVA on July 31, 1979. PSC 28-79 was written on August 6,1979 licensing evaluation was begun, and the final evaluation report was distributed on January 10, 1980.

This report states that the surge line must not fail during a LOCA event. B&W customers were notified on January 27, 1981, that B&W was not going to perfonn additional work on this problem B&W supplied displacement values for the pipes affected by a LOCA to the utilities for use in the analysis for other lines attached to these pipes.

Another PSC, (40-80) concerning large seismic displacements and normal operating vibration of in-core piping was written on June 6, 1980. All customers with 177 and 205 plants were notified on December 19, 1980, that B&W was not going to perform any additional work on this problem. However, B&W is still active in the solution of these two problems under funding from TVA. Nine interim reports have been submitted by TVA to the NRC Region II office. The in-core piping problem is still not resolved. B&W is in the process of analyzing the affects of a surge line break during a LOCA event.

1 S

25 1

ORGANIZATION: BABC0CK & WILC0X, A MCDERM0TT CO.

UTILITY POWER GENERATION DIVISION LYNCHBURG, VIRGINIA REPORT INSPECTION N0.: 99900400/85-01 RESULTS: PAGE 10 of 12

k. PSC 21-82, Water Hammer in External Auxiliary Feedwater (AFW)

Headers: This PSC concerns a postulated water hammer problem in AFW headers in B&W plants. Two scenarios were postulated which could produce water hammer conditions: The first is leakage in the lines or valves of the system might permit steam to replace the water held stagnant in the system when not in use. The second is boiling of the stagnant water as a result of heat transfer from the operating gencrators to the headers. After analyzing the problem, B&W concluded that it was not reportable under the requirements of 10 CFR 21 or 10 CFR 50.55(e).

2. T3 PIPE Computer Code:

T3 PIPE is a computer code that performs analyses in accordance with the ASME Nuclear Power Piping Code,Section III, NB-3600. Piping stresses are calculated beforehand using a finite element code and then used as input to T3 PIPE. A new version of T3 PIPE, Version 7.0 (the ninth version), is in the process of being certified. All earlier versions were removed from certification on January 11, 1985, because an error was found in Version 6.0. Five of the eight versions of the code were developed, partially or completely, as a result of errors found in the calculations for butt welded fittings and branch connections. Some of these errors caused the program to stop.

However, on three separate occasions the program did not stop and the execution gave erroneous results. Users are contacted after each error is identified to determine whether these options of the code had been used in safety-related analyses. The users are being contacted a third time to detennine whether the most recently identified errors involve options of the code which had been t. sed in safety-related analyses. The prior two errors did not afrect safety-related analyses.

This sequence of events indicates improper verification of these joint types from Version 1.0.

A second item involving T3 PIPE code verification was identified. In the 1981 winter addendum to the ASME Code,Section III, several stress indices were changed for various piping products. Verification was performed for Version 4.0, showing that the revised stress indices given by the T3 PIPE program was correct. It was stated that the uncnanged indices were visually checked and noted to be correct, but no record was made of this check. In addition, no general check of stress indices was made before Version 4.0 to show that the program was using the correct indices.

One nonconformance, discussed in Section B.4, resulted from this part of the inspection.

t 26

)

ORGANIZATION: BABC0CK & WILC0X, A MCDER*40TT C0.

UTILITY POWER GENERATION DIVISION LYNCHBURG, VIRGINIA REPORT INSPECTION N0.: 99900400/85-01 RESULTS: PAGE 11 of 12

3. NULIF Code Certification File Review: The NULIF code is a proprietary B&W code for calculating spectrum weighted cross-sections for input into the spatial neutronic codes.

The certification file was reviewed for compliance with the QA procedure NPG-0902-06. The latest version of the code was checked for proper documentation and verification. No nonconformances or violations were found in this part of the inspection.

4. PDQTHF Code Certification File keview:

The PDQTHF code is a proprietary B&W code for calculating the two-dimensional, neutron diffusion / depletion for reactor cores.

The certification file was reviewed for compliance with the QA procedure NPG-0902-06. The latest version of the code was checked for proper documentation and verification analysis. No noncon-formances or violations were found in this part of the inspection.

5. TACO 2.0 Code Certification File Review: The TAC 0 2.0 code is a fuel behavior code for steady state, burnup-dependent fuel para-meters (fuel tempe:ature profiles, fuel stored energy, fuel densification, fuel rod swelling, cladding creep and fission gas release, fuel rod internal pressure). The code is used in licensing 1

analysis for initializing LOCA transients.

The certification file was reviewed for compliance with the QA procedure NPG-0902-6. No nonconformances or violations were found in this part of the inspection.

6. Review of the Thermal-Hydraulic Analysis File for Oconee 1 Cycle 9 Reload: The file containing the calculation and transmittal docuTent-ation for the thermal-hydraulic analyses performed for the Cycle 9 reload of Oconee-1 was audited. Verification of the calculation results was reviewed as per B&W Administrative Manual Procedure NPG-0402-01, Rev. 17 No violations or nonco(Preparing nformancesand Processing were UPGD found in this partCalculations).

of the investigation.

7. REDBL5 Computer Code The REDBL5 code is being developed from the RELAP5/M001 computer code that was obtained from the Idaho National Engineering Laboratory (INEL) for use in various safety-related analyses. The code has been i

l 27

i l '0RGANIZATION: BABC0CK & WILC0X, A MCDERM0TT C0.

! UTILITY POWER GENERATION DIVISION LYNCHBURG, VIRGINIA

! REPORT INSPECTION N0.: 99900400/85-01 RESULTS: PAGE 12 of 12 conditionally certified and is available on the B&W computer for licensing calculations. The modification and verification of REDBL5 were reviewed during this inspection. No violations or nonconform-ances were found in this part of the investigation.

I 28 l

l

ORGANIZATION: BECHTEL POWER CORPORATION '

EASTERN POWER DIVISION GAITHERSBURG, MARYLAND REPORT INSPECTION INSPECTION NO.: 99900519/85-01 DATE(S): 1/7-11/85 ON-SITE HOURS: 117 CORRESPONCENCE ADDPESS: Bechtel Power Corporation Eastern Power Division ATTN: Mr. H. W. Wahl Vice President & General Manager

l. 15740 Shady Grove Road Gaithersburg, Maryland 20877-1454 ORGANIZATIONAL CONTACT: Mr. D. C. Kansal, Deputy Division QA Manager TELEPHONE NUMBER: (301) 258-3776 PRINCIPAL PRODUCT: Architect-Engineering Services NUCLEAR INDUSTRY ACTIVITY: The Bechtel Eastern Power Division, Gaithersburg, Maryland has approximately 2000 people employed on domestic nuclear projects.

N n \ \ \ cr n .

ASSIGNED INSPECTOR: b M /wIl 6[7(/ S'I Date 7{Section(SPIS)D. Milano, Special Proj$ts" Inspection OTHER INSPECTOR (S): R. L. Pettis, SPIS 0 P Gormley, Program Coordination Section APPROVED BY: I hk i

/j.W.Craig, Chief,SPIS,yendorProgramBranch Datd INSPECTION BASES AND SCOPE:

A. BASES: 10 CFR Part 21 and 10 CFR Part 50, Appendix B.

B. SCOPE: Evaluate the procurement control program for safety-related equipment for the Standardized Nuclear Unit Power Plant Systems (SNUPPS) projects. ,

l PLANT SITE APPLICABILITY: Callaway(UnionElectric)(50-481)andWolfCreek (Kansas Gas & Electric) (50-483).

29 l

ORGANIZATION: BECHTEL POWER CORPORATION EASTERN POWER DIVISION GAITHERSBURG. MARYLAND REPORT INSPECTION NO.: 99900519/85-01 RESULTS: PAGE 2 of 5

A. VIOLATIONS

None.

B. NONCONFORMANCES:

1. Contrary to Criterion V of 10 CFR 50, Appendix B, and Bechtel Engineering Department Project Instruction EDPI-4.58-01, design test results were approved for the Battery Chargers which did not meet specification requirements.

2. Contrary to Criterion V of 10 CFR 50, Appendix B, and Bechtel Engineering Department Procedure EDP-4.54, objective evidence could not be provided to substantiate the certification of the Professional Engineers authorized to certify design specifications.

3. Contrary to Bechtel EDPI-4.58-01, the Seismic Analysis Report

' prepared by Wachter Associates, in accordance with Technical i

Specification 10466-C-175 for the Spent Fuel Storage Racks, utilized a damping value of 5%, which was provided by Bechtel but contrary to the specification requirement of 4%.

C. UNRESOLVED ITEMS:

1. Appendix A, Part 4.0, of Bechtel Technical Specification,10466-C-175, Spent Fuel Storage Racks, requires the Seismic Analysis Report to contain a Certification of Compliance (C of C) attesting that all work was performed under the direction of a Registered Professional 1

Engineer, whose seal should be displayed on the form. This document

! was also listed on Bechtel Form 3210, " Engineering and Quality Verification Documentation," in the Specification as an item requiring Bechtel Engineering review and approval.

During the inspection, the above document could not be produced although the Seismic Report had already received Bechtel approval.

The resolution of this item will be pending the location and identi-i fication of the Certificate of Compliance by Bechtel.

, D. STATUS OF PREVIOUS INSPECTION FINDINGS:

1. (Closed) Nonconformance (84-01): Nine approved Engineering Department Project Instructions (EDPIs) did not have an issue date listed on the instruction.

The EDPIs were reviewed to verify incorporation of the issue date and found satisfactory.

30

ORGANIZATION: BECHTEL POWER CORPORATION EASTERN POWER DIVISION GATTHERSBURG, MARYLAND REPORT INSPECTION N0.: 99900519/85-01 RESULTS: PAGE 3 of 5

2. (Closed) Nonconformance (84-01): The design review process allowed power cable supplying Valcor soler.oid valves to be specified that was not suitable for the required application.

A design change has been issued by the SNUPPS project to replace the field wiring with qualified high temperature wiring. Based on the response provided in the Bechtel letter of June 29, 1984, the corrective and preventive actions are acceptable.

E. OTHER FINDINGS AND COMMENTS:

1. Procurement Process Control Selected documentation was reviewed for twelve (12) safety-related equipment purchase orders. This review evaluated the effect on the original design and regulatory commitments for changes to the purchase orders, Supplier Design Deviation Requests (SDDRs), Quality Supplier Deviation Reports (QSDRs), and nonconformances. In addition, Engineer-ing and Quality Verification documentation, such as design reports, test reports, seismic and environmental qualification reports, and calculations, were reviewed for compliance with the specification i

requirements. Finally, correspondence and other documentation contained in the Supplier Quality and Purchasing Department files were reviewed for problems and evidence of proper evaluation and correction, such as Rework Plans,

a. During the review of the purchase order documentation for the battery chargers, E-051, it was noted that these units experi-enced difficulty during initial operation which required the vendor to rework the units in the field. Because of this, the design change packages, called Rework Plans, and the subsequent field testing requirements were reviewed to determine that the units continued to comply with the design specification require-ments. Earlier documentation of the design and production testing performed by the vendor was also reviewed. In this area, it was noted that the strip charts associated with the alternat-ing and direct current transient voltage withstandability tests indicated that these tests were not fully conducted in accordance with National Electrical Manufacturers Association (NEMA)

Standard PV-5-1976 which was delineated in the specification.

Also, the battery size utilized for the ripple voltage measure-ment was smaller than stated by the NEMA Standard. The tests were reviewed by the Bechtel Supplier Quality Pepresentative and approved by the Design Engineer, but without any statements as to 31 i

ORGANIZATION: BECHTEL POWER CORPORATION EASTERN POWER DIVISION GAITHERSBURG, MARYLAND i REPORT INSPECTION NO.: 99900519/85-01 RESULTS: PAGE 4 of 5 the acceptability pf these deviations. The short-circuit test method described in Power Conversion Products Process Specifi-cation PS-77-8, " Production Test Procedure of SNUPPS Battery Chargers in accordance with NEMA PV-5-1976 and Technical Specifi- l cation 10466," dated March 16, 1977, did not correspond with the method of the NEMA Standard PV-5. However, a supplemental test record sheet provided with the final test data described the actual test conducted and results which were in compliance with the short-circuit test requirements of the NEMA standard,

b. The review of the purchase order for Spent Fuel Storage Racks, C-175, centered on the seismic analysis and mechanical reports, Bechtel document numbers C-175-0014 and 0020 respectively. From the seismic analysis report, several areas were considered to lack conformity to the technical specification, Appendix A, which outlines the criteria for the report preparation. For example, the value used for significant seismic event (SSE) damping departured from specification requirements. In the generation of the dynamic response spectrum input, Bechtel provided to Wachter Associates an SSE damping value of 5% vice the 4% stated in the specification.

Additicnally, Part 4.0 of Appendix A to the specification for the spent fuel storage racks, required that a Certification of Compliance, signed and ap) roved by a Registered Professional Engineer, be provided wit 1 the seismic analysis report. This documentation could not be provided during the inspection although the report was found to be approved by Bechtel,

c. The other purchase orders reviewed included: M-218A (Sub B),

j mechanical shock suppresors; M-067, spent fuel pool bridge crane; l

M-6278, Ruskin fire dampers; E-018, motor control center; J-301 t l

Rosemount pressure and differential pressure cells; and J-359 i containment hydrogen analyzers. The inspection in these areas did not result in any significant findings. Of note, however, was that the purchase order for the pressure and differential pressure cells was conducted without the assignment of a Supplier Quality Representative. It was Bechtel's determination that this instrumentation was of a standard design and that source inspec-tion was not necessary, although several models were specifically nuclear grade and required seismic and environmental qualification 1 In general, deviations from the specification were at times documented by various forms of correspondence, e.g., telexs, 32

ORGANIZATION: BECHTEL POWER CORPORATION EASTERN POWER DIVISION GAITHERSBURG. MARYLAND REPORT INSPECTION NO.: 99900519/85-01 RESULTS: PAGE 5 of 5 memos, letters, telegrams. Most of these changes from the specifications were of a nature that supporting analysis was not necessary (i.e., the substitution of a load limit switch for the motor overtorque protection required by the spent fuel pool bridge crane specification). More complex designs were processed by the formally established design change systems.

From this area of the inspection, two (2) nonconformances and one (1) unresolved item were identified.

2. Qualification of Personnel Performing ASME III Code Certification Activities During the review of the documentation for the purchase orders, Bechtel could not provide objective evidence to support compliance with the qualification requirements of Engineering Department Procedure EDP-4.54 for two Registered Professional Engineers engaged in certification activities for both (a) the owners design specifi-cation and (b) the N certification holders design report. However, both employees were certified to perform such activities in March 1984 by their Chief Engineer. In the case of the torsional restraint design specification, employee #390844 certified a report in November 1984 over 15 months after the issuance of EDP-4.54.

The basis for the issuance of this EDP stems from the summer 1980 addenda to ASME III which became effective in December 1980. This addenda revised NCA-3255 to include ANSI /ASME N 626.3-1979 as the basis for establishing qualifications and duties of Professional Engineers engaged in Divisions 1 and 2 certification activities.

Bechtel's EDP-4.54 was issued over three years after the June 30, 1980 release of the summer addenda.

i From this area of the inspection, one (1) nonconformance was l identified.

l 33

~

ORGANIZATION: BECHTEL POWER CORPORATION WESTERN POWER DIVISION / HOUSTON PROJECT OFFICE LOS ANGELES, CALIFORNIA

)

REPORT INSPECTION INSPECTION NO.: 99900521/85-01 DATE(S): 2/4-8/85 ON-SITE HOURS: 120 CORRESPONDENCE ADDRESS: Bechtel Power Corporation i Western Power Division / Houston Project Office i ATTN: Mr. L. G. Hinkleman

, Post Office Box 60650, Terminal Annex i

Los Angeles, California 90060 j j

ORGANIZATIONAL CONTACT: Mr. J. Gatewood 4 TELEPHONE NUMBER: (512) 972-3611 X4168 l PRINCIPAL PRODUCT: Architect-Engineering Services "

! NUCLEAR INDUSTRY ACTIVITY: The Houston Project Office is currently providing

] principal architect-engineering and construction management services for the j South Texas Project.  ;

4 i I

l 1 i ASSIGNED INSPECTOR: , w i

,!m ~/6[I5 7  !

j Dath  !

PJ(D.

i SPIS)Milano, Special Proje' cts Irppection Section  !

i OTHER INSPECTOR (S)

R. L. Cilimberg, SPIS T. L. Bridges, EG8G Idaho, Inc.

K. .

,ard, G&G daho Inc.

APPROVED BY: ( <. A '

Ncv i Ffp f 3

,JpnW.Craig,ChiefSPI'S}VendorProgramBranch ' Date I

INSPECTION BASES AND SCOPE:

A. BASES: 10 CFR Part 21 and 10 CFR Part 50, Appendix B.

B. SCOPE: Site procurement process control for material and services. I t

i PLANT SITE APPLICABILITY: South Texas Project Units 1&2(50-498,50-499).

4 35

ORGANIZATION: BECHTEL POWER CORPORATION WESTERN POWER DIVISION / HOUSTON PROJECT OFFICE LOS ANGELES, CALIFORNIA REPORT INSPECTION NO.: 99900521/85-01 RESULTS: PAGE 2 of 7 A. VIOLATIONS:

i None.

B. NONCONFORMANCES:

None. -

C. UNRESOLVED ITEMS:

1. (0 pen) 85-01: Questions concerning the qualification of the coating system on the steel containment liner of Unit I will be reviewed by the NRC. This item will remain open pending completion of this review.

D. STATUS OF PREVIOUS INSPECTION FINDINGS:

Previous findings were not covered during this inspection.

Previous inspection findings are discussed in Nonconformance B.1 and Unresolved Items C.1 thru C.4, of the previous inspection report.

84-01.

These items will be covered during a future inspection.

E. OTHER FINDINGS AND COMMENTS:

1. Site Procurement Process Control A review of the Bechtel field office procurement process for the purchase of material and services was conducted and focused upon the area of mechanical and electrical procurements. Purchase order packages for nine orders including any changes processed af ter contract award were reviewed. The aajority of the changes to the orders for these packages were the result of Specification l Change Notices (SCNs) and nonconformance reports (NCRs). With one exception, none of the purchase orders involved any source inspection. Thus, the acceptance of the material was done by receiving inspection along with, in some cases, receipt of Supplier Certificates of Conformance.

36

ORGANIZATION: BECHTEL POWER CORPORATION WESTERN POWER DIVISION / HOUSTON PROJECT OFFICE LOS ANGELES, CALIFORNIA REPORT INSPECTION NO.: 99900521/85-01 RESULTS: PAGE 3 of 7 Procedure change number NPC-27 to South Texas Project (STP)

Procedure WPP-QCI-4.0, " Receiving Inspection," Revision 8, dated January 24, 1985, made changes to indicate what Bechtel responsibility will be for the review of Westinghouse Electric Corporation documentation supplied with the shipment of ASME materials for which Bechtel has and maintains Code responsibility for the system design. In these cases, the procedure now states that "Bechtel QC shall perform a technical review of the Westinghouse documentation supplied with shipment...." A review of the purchase order for inspection and repair of high pressure safety injection, low pressure safety injection, and containment spray pumps (P.O. BF-2255 with Westinghouse / Pacific Pump),

identified that neither of the two Receiving Inspection Plans were performeo by Bechtel QC. While one of the plans had this QC inspection requirement deleted by a "N/A" (not applicable) in the signoff area, the other inspection, as described by the responsible Quality Control Inspector, only verified the physical receipt of the Westinghouse data that came with the shipment.

The documentation for Receiving Inspection Plan No. 1387 for P.O.

BF-2255 was reviewed, a Westinghouse Quality Release for Shipment was attached describing the information that the Westinghouse

source inspector reviewed prior to accepting the items. While other documentation stated that no repairs were necessary on the pumps, the Quality Release indicated that Welder Performance Qualification Records (PQRs) were reviewed. The NRC inspector was not able to determine the type of work that was done which would require welder certifications on this order.

The purchase order for machining services on the Reactor Vessel Lateral Restraints (BF-G689), while found to be adequate, contained a nonconformance report with a questionable resolution.

Nonconformance Report BM-00154 was prepared when it was determined that the precision straight edge used in verifying the surface flatness requirement had not been calibration checked. Rather than recheck the instrument, as the original disposition stated, the disposition was changed to accept the data as-is stating that the straight edge does not nred to be calibrated. Additionally, the order requirement for flatness within .001 inch could not be accurately measured by the system utilized. First, a piece of ,

.001 inch shim stock was checked with a 0 to 1 inch micrometer.

Then, the shim was used as a feeler gage with the straight edge.

This method of record measurement transfer leads to potential I

37

ORGANIZATION: BECHTEL POWER CORPORATION WESTERN POWER DIVISION / HOUSTON PROJECT OFFICE

, LOS ANGELES. CALIFORNIA REPORT INSPECTION NO.: 99900521/85-01 RESULTS: PAGE 4 of 7 inaccuracies that are greater than the actual measurement required.

The purchase order for a replacement valve positioner (BF-2828) was reviewed. When the NRC inspector questioned why a second j positioner was ordered af ter the first was received, it was stated that the person specifying the replacement did not completely record the correct model number from the damaged l unit. The supplier initially sent and the constructor installed an incorrect unit. The error was identified when the unit would not function properly when tested.

The purchase order for the repair of a damaged weld on a gate valve (BF-2687) was also reviewed and found to be adequate.

However, the Certified Material Test Reports (CMTRs) were not included with the Receiving Inspection Documentation nor was their location referenced. The CMTRs were later found in an equipment file.

2. Receiving Inspection of Home-office Procured Transmitters The receiving inspection criteria were reviewed for the pressure and differential pressure transmitters procured from Rosemount

(P.O. 14926-4332). This order was selected to be reviewed even though the purchase order was initiated by the Houston Project Office because these items do not receive source inspection prior to shipment. The Bechtel procurement policy identifies these instruments as standard, off-the-shelf, commodities and thus, do not warrant source inspection. Therefore, acceptance of the material is based on the receiving inspection and Certificates of Compliance.

The review of the Receipt Inspection Plans for this order indicated that Certified Material Test Reports (CMTRs) were to be received t

and evaluated per the technical specification requirement.

However, the specification requires that, if the vendor does not submit the required information that is required, the vendor is required to maintain and not destroy the data. Approximately half of the Receiving Inspection Plans gave no indication that the CMTRs did not arrive although this was stamped as approved.

Rosemount provided only a statement in a Certificate of Compliance that " Records are kept on file which identify the process wetted l surface materials used in the transmitter. The materials are in

accordance with applicabic material specifications; the chemical

~

38

ORGANIZATION: BECHTEL POWER CORPORATION WESTERN POWER DIVISION / HOUSTON PROJECT OFFICE LOS ANGEL,ES, CALIFORNIA REPORT INSPECTION N0.: 99900521/85-01 RESULTS: PAGE 5 of 7 and physical reports are on file." In order to determine whether l the materials used at the Project Site are correct, the Rosemount '

installation drawings must be reviewed. However, these are not I utilized during receiving inspection.

The same Certificate of Compliance is utilized for acceptance of the cleanliness inspection and hydrostatic test. The actual hydrostatic test pressure specification is not provided. But a statement that the test was conducted at 150% of maximum rated pressure or a minimum of 2000 PSI is included. The Certificate

' also indicates that the transmitter was tested for nuclear service. Rosemount Qualification Report D830040 was identified but did not contain a statement to indicate that the hydrostatic test results were acceptable.

In accordance with Section 10.2 of ANSI N45.2.13-1976, Quality Assurance Requirements for Control of Procurement of Items and Services for Nuclear Power Plants, the Purchaser should provide a means to verify the validity of Supplier certificates and the effectiveness of the certification system, such as during the performance of audits of the Supplier or independent inspection or test of the items. Such verifications should be conducted by the Purchaser at intervals commensurate with the Supplier's past quality performance. The validity of Certificates takes on an added importance when the acceptance of the item, such as the case of the transmitters, is based solely on this document.

The Bechtel Quality Assurance Program for Nuclear Power Plants, BQ-TOP-1, Rev. 3A, dated October 1980, which forms Section B of the STP Quality Assurance Program Description, Rev. 9, dated December 28, 1984, provides Bechtel's alternate interpretation to the ANSI N45.2.13. It states that the means of verification may include source witness / hold points, source audits, and document reviews; independent intpections at the time of material receipt; user tests on selected commodities, such as concrete components; and tests after installation on selected components and systems. All of these means verify whether or not a supplier has fulfilled procurement document requirements and whether or not a certification system is effective.

i l In the case of the Rosemount transmitter purchase order, independent inspections are not done upon receipt and the material properties are not independently tested. The absence of source inspection and the lack of source audits that verify 1

Certificate validity for these transmitters raise questions 39 i

ORGANIZATION: BECHTEL POWER CCRPORATION l WESTERN POWER DIVISION / HOUSTON PROJECT OFFICE

! LOS ANGELES, CALIFORNIA REPORT INSPECTION NO.: 99900521/85-01 RESULTS: PAGE 6 of 7 concerning the effectiveness of this system. It appears that the present Bechtel approach is: not question the validity unless a problem occurs which may relate to the area addressed by the Certificate. A further review of this area may be performed at a later date.

3. General Coments on Receiving Inspection The STP procedure WPP-QCI-2.2, Control of Inspection Planning, Rev. 6, dated June 20, 1984, requires that Receiving Inspection Plans (RIPS) be approved by the Project Management Coordinator, Project Quality Assurance Engineer, and the Project Quality Control Engineer. Generally, a preprinted " Receiving Inspection Plan for Permanent Plant Items" tform STP-010A) is utilized which contains the information listed in Section 5.2 of ANSI N45.2.2. This form is not formally approved for each use and is usually modified at the discretion of the QC inspector.

The QC Inspector fills out the form including the reference

> documents such as the technical specification and deletes those attributes that he considers not applicable. Although modified in cases, this form is considered by the Project to be preapproved since it is included as a " Typical form" in STP Proced. 7 WPP-QCI-4.0.

Section 1.7 of the Receiving Inspection Plan (RIP) for Permanent Plant Items states that the QC should sign and date the Material ReceivingRequest(MRR). When reviewing the MRRs, it was noted that the QC inspector signs a block on the MRR stating "QC Approved." When reviewing MRR B11827 and RIP 4043, on purchase order BF-2255 for the inspection and repair of selected pumps, the NRC inspector identified that the MRR was approved on August 2,1984 but the RIP was not completed indicating item acceptance until January 30, 1985.

4. Protective Coatings During the review of purchase ~ orders and specifications for protective coatings it was determined that a zinc rich primer and an epoxy top coat was applied to the steel containment liner of Units 1 and 2. The NRC inspectors reviewed the procurement and receiving inspection documents for these coatings. A material Certificate of Compliance was not provided by the coating supplier (Eastern Imperial Coatings) for the 27 RIPS reviewed. The coating supplier did provide the following statement on all but three RIPS:

l 40 l

ORGANIZATION: BECHTEL POWER CORPORATION WESTERN POWER DIVISION / HOUSTON PROJECT OFFICE LOS ANGELES, CALIFORNIA REPORT INSPECTION N0.: 99900521/85-01 RESULTS: PAGE 7 of 7

" Eastern Imperial Coatings Corporation certifies that the coating material supplied to Bechtel Energy Corporation (South Texas Project) on Purchase Order 1496-BF-1788, t essentially has the same required performance properties l and composition as the coating material on which approval was granted.

Discussions with the NRC Resident Inspector and Bechtel representatives revealed that the coatings on the Unit I containment liner were removed down to bare steel from all areas except those areas behind installed ducts, Bechtel Specification No. 5A819AQ1001.

Revision 2 dated July 18, 1984, requires that the Unit I containment liner plate be coated with two coats of epoxy and the Unit 2 containment liner plate retain the initial coating system (zinc primer with epoxy top coat).

The inspectors questioned the suitability of this coating system for Unit I since the coating will be routinely exposed to high humidity during outages when containment integrity is not maintained. The inspector noted that this two coat epoxy system may result in accelerated corrosion if the bare steel should become exposed to moisture through voids in the coating system. A potential concern noted by the inspectors was that corrosion may result in separation of the coating from the steel with subsequent entrapment in the containment sump.

At the exit meeting the inspectors stated that this item would be an unresolved item pending additional NRC review of the specification and acceptance tests for the two coat epoxy system used on the liner of Unit 1.

i 41

ORGANIZATION: BORG WARNER CORPORATION NUCLEAR VALVE DIVISION VAN NUYS, CALIFORNIA REPORT INSPECTION INSPECTION NO.: 99900289/85-01 DATE(S): 3/26-28/85 ON-SITE HOURS: 48 CORRESPONDENCE ADDRESS: Borg Warner Corporation Nuclear Valve Division ATTN: Ms. Norma Moore, Manager. Quality Engineering 7500 Tyrone Avenue. P. O. Box 2185

! Van Nuys, California 91409 ORGANIZATIONAL CONTACT: Norma Moore, Manager, Quality Engineering TELEPHONE NUMBER: (818) 781-4000 PRINCIPAL PRODUCT: Nuclear valves and hydraulic valve operators.

NUCLEAR INDUSTRY ACTIVITY: Approximately 30%.

. n #1 ASSIGNED INSPECTOR: (( [ s'/ir/as-Date 5:= k. E. Ollergeactive Inspection Section (RIS)

OTHERINSPECTOR(S): L. Vaughan, Program C ordination Section APPROVED BY: .

sdr[Pr E. W. Merschoff ief RIS, Vendor Program Branch Date INSPECTION BASES AND SCOPE:

A. BASES: 10 CFR Part 21 and Appendix B to 10 CFR Part 50.

B. SCOPE: The purpose of this inspection was to collect licensee and vendor interface information with regard to valve operation and main-tenance manuals and 10 CFR Part 21 notifications to customers.

Additionally, a followup was performed on a Borg Warner Nuclear Valve Division (BW-NVD) 10 CFR Part 21 report concerning Perry Station; a Bellefonte 50.55(e) report; a Comanche Peak 50.55(e) report and the I status of previous inspection findings.

PLANT SITE APPLICABILITY: As identified in this report.

l l

l 43 I

! ORGANIZATION: BORG WARNER CORPORATION NUCLEAR VALVE DIVISION VAN NUYS, CALIFORNIA REPORT INSPECTION l NO.: 99900289/85-01 RESULTS: PAGE 2 of 14 A. Subjects Inspected:

1. Collection of: (a) operation and maintenance manuals (OMM),

information for safety related valves (SRV), and/or valve operators (V0) shipped to various nuclear power plants (NPP) during the period of 1977 through 1984, and (b) identification of 10 CFR Part 21 notification letters to NPPs concerning potential generic deficiencies.

2. Potential gate valve guide problem at Perry NPP.

3. Over-torquing of valve motor operator studs at Bellefonte NPP.

4. Check valve malfunction due to broken tack welds at Comanche Peak NPP.

5. Status of previous inspection findings.

B. Supplementary Information:

Borg Warner, Nuclear Valve Division located in Van Nuys, California, manufactures nuclear valves and hydraulic valve operators for the nuclear industry. They hold ASME N and NPT type authorizations for this location.

C. Inspection Details:

1. The collection of information concerning OMMs furnished with SRV and V0 to various NPPs was accomplished by selecting SRVs and V0s for each plant site identified in the OMM log for the period of 1977 through 1984.

The information needed to identify NPP sites, OMMs and valves is shown in the following Table A.

I 44

ORGANIZATION: BORG WARNER CORPORATION NUCLEARVAL9EDIVISION VAN NUYS, CALIFORNIA REPORT INSPECTION NO.: 99900289/85-01 RESULTS: PAGE 3 of 14 l TABLE A BW-NVD OPERATION AND MAINTENANCE MANUALS l

Project OMM No. Item Description Tag and Other I.D. Nos.

Arkansas Nucl' ear 1063 2" C.S., M.O., Tag Nos. 2"-CBA-GB-MO-One - Unit 2 Y-Globe Valve CV2815-ANO-1, 2"-CBA-GB-MO-CV2816-AN0-1 Beaver Valley 2099 16"x10"x16", 900 LB., Tag Nos. 2 FWS-HYV 157A, Unit 1 C.S. FW. Iso. Valve B and C Bellefonte 2227 6" & 10", LP, CS & TVA Mark No.

S.S. Globe Valve 3 AWO412-ND-60 3 AWO416-NS-20 3 BWO456-KC-45 Byron /Braidwood 2023 16"x10"x16", 900 LB., Tag Nos. 1FW009A,-B,&C; Units 1 and 2 C.S. FW, Iso. Valve 2FW009A,-B,&C Catawba 2239 2", 1635 LB., 5 S Duke It. No. 9J-244 Gate Valve, Air List No. CN-1500-10 Op., Fail Close. Order No. A98528.

Comanche Peak 2337 4", 300 LB., S.S. Tag Nos. 1FP-591; Units 1 and 2 Gate Valve, M.0. IFP-592; 2FP-449, 2FP-450; Type No.

4G B1605 BHMO.

Ginna 1082 2"-Y Type, C.S. P.O. 73622-1302 Lift Check Valve Valve I.D. No. V-5 Valve Item V-5 Grand Gulf 1086 1/2 thru 2 inch Bechtel P.O. 9645-M Station S.S. & C.S., M.0. -250.2 Y-Type Globe Valves Limerick Station 1036 1-1/2 & 2 inch, ID No. 1-1/2-CCD-GB Units 1 and 2 Y-Type Globe Valve, ID No. 2-CCD-GB

., S.S.

l 45

f ORGANIZATION: BORG WARNER CORPORATION NUCLEAR VALVE DIVISION VAN NUYS, CALIFORNIA REPORT INSPECTION NO.: 99900289/85-01 RESULTS: PAGE 4 of 14 TABLE A (Continued)

Profect OMM No. Item Description Tad and Other I.D. Nos.

McGuire Station 2246 4", 1500 LB, C.S., Item 6H-102, Gate Valve, M.0. List No. MC-900-12 Order No. A98513 Monticello 2232 1 " & 2"; 1500 LB., Tag No. 2"-CCA-GB-AO-Station S.S. Globe Valve ZS-1 through -4; Air Op. Tag No. 1"-CCA-GB-AO-ZS-1 through -4.

Nine Mile Point 2187 2", 1500 LB., C.S. Tag Nos. 44.2-15; Unit 1 Globe Valve Air 44.2-16; 44.2-17, Op. 44.2-Spare; and 44.2-Spare Oconee Station 2078 4" & 8", LP, S.S. & Duke Item 2B-468; C.S. Gate Valves, and 5B-266; M.0. List Nos. CN-0150-12 and CN-0150-10; Order Nos. A98521 and A98528.

Palo Verde 2346 6", 1500 LB, S.S. Tag Nos. RC130 and Globe Valve, M.0. RC131.

PASNY 2207 1", 150 LB., C. & Unique No. WLD-MOV-505 S. Gate Valve, M.0.

Perry Units 2010 28", 900 LB, M.0. Tag Nos. RNN229 and I and 2 Gate Valve RNN268.

Pilgrim Station 2364 1", HP, Diaphragm Tag No. 1-CCC-YGB-S-Unit 1 Y-Type Globe, (R, $)

Valve, S. S.

Rancho Seco 2340 2", 1500 LB, Lift Tag Nos. V2167; V2168; Check, Valve, S. S. V2169 and V2556 46

ORGANIZATION: BORG WARNER CORPORATION NUCLEAR VALVE DIVISION VAN NUYS, CALIFORNIA REPORT INSPECTION NO.: 99900289/85-01 RESULTS: PAGE 5 of 14 TABLE A (Continued)

Project OMM No. Item Description Tag and Other I.D. Nos.

Seabrook 2027 18", 900 LB, C.S. Tag Nos. 1FWV30, -39, Units 1 and 2 FW, Iso. Valve -48, -57 and 2FWV30,

-39, -48, -57.

Shearon Harris 2026 16", 921 LB, C.S. Tag Nos. 2FWV275AB-1, Units 1-4 Gate, Valve 2FWV285-AB-1,-2 Hydraulic Operator 2FWV275-AB-2,-3 2FWV28S-AB-3,-4 2FWV275-AB-4 2FWV265-AB-1,-2,-3,-4.

SNUPP 2086 1", 1500 LB, C.S., Bechtel Tag No.

Callaway and Gate Valve, M.0.

Wolf Creek 1-CBC-GTE-Q-W-MA-K South Texas 1074 1-1/2" C.S. Gate Tag Nos. 01-1/GH22SS3 Project Valves -ABC; 01-1/2 GH25SS3 Units 1 and 2 -ABC; 01-1/2 GH25SS3A-C St. Lucie 1043 3", S.S., Gate Valve Item No. 9; Unit 2 Tag Nos. V-1441,-2,-3,

-4 Code No. 31-14-55-0622 Susquehanna 1042 2", S.S., H.O.,

Units 1 and 2 Item Nos. 29.1.1 and Y-Type Globe 29.2.1 Valve Equip. Nos.

2-CCA-GB-H0-FV-12603-P 2-CCA-GB-MO-FV-22603-P Vogtle Project 2372 Electro-Hydraulic Fisher Controls /

1C Operator Bechtel P.O. No.

S181972 WPPSS Unit 2 2345 3", 150 LB, C.S. Tag Nos. FOR-V-219, Gate Valve. Air -220, -221, -222, Op., Fall Close -394, -395

)

47

ORGANIZATION: BORG WARNER CORPORATION NUCLEAR VALVE DIVISION VAN NUYS, CALIFORNIA REPORT INSPECTION NO.: 99900289/85-01 RESULTS: PAGE 6 of 14

2. Collection of information also included customer notification letters sent by BW-NVD, concerning 10 CFR Part 21 reports in which the potential defects were generic to several NPPs. Review of seven l

10 CFR Part 21 reports identified four which were generic in nature.

These reports are identified in the following Table B.

TABLE B 10 CFR PART 21 CUSTOMER NOTIFICATIONS BY BW-NVD

! Report Customer / Site Valve 10 Deficiency Date 2/7/80 Duke Power / Tag Nos. 9J-239, Qty 6 Inadequate gate guides Catawba / and 9J-221, Qty 6 in valve under test McGuire at Duke Power. Needs rework.

Bechtel/ Tag No. ACA-GT-M0 Same Arkansas -2CV4698-1-PL, and Nuclear One 3ACA-GT-M0-2CV4697-2-PL Unit 2 TVA/Bellefonte Tag No. 3BW0456-K-30 Same Qty. 16.

i Combustion Tag Nos. SI604 and Same Engineering / S1609, Qty. 4.

Palo Verde 5/23/80 WSH/B0 ECON / Serial Numbers: Stem nuts or extensions GERI/WPPS-2 26210 thru 26216, may become loose l

22337 thru 22339, during operation of 22355, 26727, and valve. Needs stem l

35653 thru 35656 modification and staking.

^

1 i 48  !

l

ORGANIZATION: BORG WARNER CORPORATION NUCLEAR VALVE DIVISION VAN NUYS, CALIFORNIA REPORT INSPECTION i

NO.: 99900289/85-01 RESULTS: PAGE 7 of 14 TABLE B (Continued) i Report Date Customer / Site Valve ID Deficiency Combustion Unit #1 - Qty 20 Same Engineering / Unit #2 - Qty 17 Palo Verde Unit #3 - Qty 10 See BW-NVD letter dated 5/29/80 to Cleveland Electric Illuminating for serial numbers.

Bechtel/ Grand Units 1 and 2, Qty 93; Same Gulf (BPC) see BW-NVD letter dated 5/30/80 to BPC for Tag Nos.

7/21/80 Duke Power / Mk Item No.28-396 Possible " worst case" Catawba SNs 35557 and -8, tolerance buildup Mk Item No.28-393 condition in which SNs 31755 thru 31764. gate could disengage from guide when valve is in backseat position.

5/2/83 Gibbs & Hill / Total 53 affected Potential swing check Comanche Peak valves. See list of valve malfunction due Units 1 and 2 valves attached to to possible broken BW-NVD trip report tack weld securing dated 2/1/83. disc to operating arm.

Requires reinspection and rework.

Arkansas Total of 12 af fected Same Nuclear One valves. See BW-NVD (ANO) letter dated 5/2/83 to ANO for serial numbers.

49

! ORGANIZATION: BORG WARNER CORPORATION

! NUCLEAR VALVE DIVISION VAN NUYS, CALIFORNIA REPORT INSPECTION N0.: 99900289/85-01 RESULTS: PAGE 8 of 14 TABLE B (Continued)

Report J Date Customer / Site Valve ID Deficiency j Mill Power / Affected valves: Same Catawba / Catawba Qty 19 McGuire/ McGuire Qty 14 Oconee Oconee Qty 1.

See BW-NVD letter dated 5/2/83 to Mill Power for customer tag nos.

Pacific Gas Total of 7 affected Same

& Electric / valves. See BW-NVD Diablo Canyon letter dated 5/2/83 to PG&E for serial numbers.

Rochester Gas Total of 4 affected Same.

& Electric valves. See BW-NVD Corp./Ginna letter dated 5/2/83 to RG&E for serial numbers.

1 i

1 50 l

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ORGANIZATION: BORG WARNER CORPORATION NUCLEAR VALVE DIVISION VAN NUYS. CALIFORNIA REPORT INSPECTION N0.: 99900289/85-01 RESULTS: PAGE 9 of 14 l

3. Misalignment of Valve Body Guides and Inadequate Fillet Welds in 20 Inch Valves for Perry Units 1 and 2.

l a. Introduction On November 21, 1984, BW-NVD notified the NRC by means of a 10 CFR Part 21 report, of two deficiencies in 20 inch gate valves supplied to Perry Station, Units 1 and 2. The first deficiency involved ten valves wherein a misalignment occurred in the fabricated valve body guides which could cause the gate to stick. The second problem involved inadequate fillet welds used for welding in the previous mentioned guides in the bodies of certain valves which might be subject to a differential pressure in excess of 600 psi during cycling of the valve.

This second defect was determined to affect four of the previous ten valves furnished to the Perry Station.

b. Findings By review of records and interviews with BW-NVD personnel, the NRC inspectors verified that information supplied in the November 1984, 10 CFR Part 21 report was accurate and complete in regard to the identified deficiencies.

During the shop processing of ten 20 inch gate valve body castings, the cast-in guides were found to contain unacceptable porosity. To correct this condition, BW-NVD removed these cast guides and replaced them with welded-in fabricated guides. During this change the shop failed to use welding fixtures and this resulted in potential lineal distortion of the guides. Furthermore, engineering failed to include lineal dimensions for the welded-in guides on the fabrication drawing No. 80996. As a result, these guides were not properly inspected after installation.

The second problem concerned the fillet welds on the guides in four of the above valves. During the November 1984 site inspection of the valves by BW-NVD engineering personnel, they observed that the guide fillet welds in the valve flow path might be inadequate for certain of the 20 inch l

51 1

?

ORGANIZATION: BORG WARNER CORPORATION NUCLEAR VALVE DIVISION VAN NUYS, CALIFORNIA REPORT INSPECTION N0 : 99900289/85-01 RESULTS: PAGE 10 of 14 valves which would be subject to operating conditions where the differential pressure exceeds 600 psi.

This deficiency was verified by stress calculations, and BW-NVD recommended that the welds in the affected valves be built up. l Cleveland Electric Illuminating Company (CEI) determined that the inspection and rework of the . valve guides would be done at the Perry site by CEI. BW-NVO determined that the two problems were unique to the Perry order and had no generic implications.

4. Overtorqued Valve Operator Studs ac Bellefonte Station

a. Introduction In November 1983, the NRC was notified by a TVA/Bellefonte Station 50.55(e) report that motor operator studs on a 6 inch gate valve could not be torqued.to the 140-165 ft. Ib.

value specified on the BW-NVD Drawing No. 79760. As a result of this report, this item was inspected.

o

b. Findings By review of records and interviews with cognizant personnel,

, the NRC inspectors determined that the above condition was s'

due to a BW-NVD drawing error which specified the wrong torque values. At the request of TVA, BW-NVD engineering performed a review of all drawings related to TVA contracts.

i This review was documented on Customer Specification i Discussion sheets. This review verified that the error

, was isolated to BW-NVD drawing No. 79760. This drawing was

! > then revised to' provide the proper torque values of 60-70 l ft. 16. ,

5. Broken Tack Welds in Swing Check Valves at Comanche Peak Station l e.'  : Introduction e In February,1983, Comanche Peak Station notified the NRC by'a 50.55(e) report, of a malfunction of a swing check

, valve due to loose parts resulting from a broken tack weld.

i

. 52

ORGANIZATION: BORG WARNER CORPORATION NUCLEAR VALVE DIVISION VAN NUYS, CALIFORNIA REPORT INSPECTION NO.: 99900289/85-01 RESULTS: PAGE 11 of 14 This tack weld connects the disc stud to the operating arm.

As a result of this notification, this item was inspected l at BW-NVD.

b. Findings ,

By review of records and interviews with cognizant personnel, the following information was obtained. Prior to 1977, tack welds joining the valve disc stud to the operating arm were the standard design used by BW-NVD for the subject valve.

In 1977 the tack weld was changed to a fillet weld. When notified of the tack weld failure at Comanche Peak in January 1983, BW-NVD visited the site. During this visit the affected valves were inspected and a repair procedure was discussed. Since this problem had generic implications BW-NVD performed a review of all swing check valves which they had manufactured. The swing check valves manufactured up to the 1977 date of the design change plus six months were considered suspect by BW-NVD and were identified in the notification letters sent to five purchaser's of the valves, including Comanche Peak. In addition to the lists of valves, BW-NVD sent each customer an inspection and rework procedure which utilized fillet welds in place of the tack welds. BW-NVD personnel indicated that all of the affected customers had responded by telephone to the notification letters. In May 1983, BW-NVD submitted a 10 CFR Part 21 report to the NRC.

A copy of the report, notification letters and lists of affected valves are maintained for reference in the VPB Reactive File.

D. Persons Contacted P. L. Milinazzo, QA Manager, Borg Warner, NVD.

M. N. Rielly, Project Engineer, Borg Warner, NVD.

W. R. Wheaton, QA Specialist, Borg Warner, NVD.

E. Documents Reviewed

1. Drawing No. 80996, Rev. D, 5/31/79, Body Assembly, 20", 900 lb.,

Gate Valve, C.S. (Ref. NR-0007).

2. Drawing No. 81040, Rev. H, 3/12/85, Valve Assembly Gate, 20",

900 lb., C.S., w/ Gear Operator.

53

ORGANIZATION: BORG WARNER CORPORATION NUCLEAR VALVE DIVISION VAN NUYS, CALIFORNIA REPORT INSPECTION N0.: 99900289/85-01 RESULTS: PAGE 12 of 14

3. Drawing No. 79558, Rev. B, 8/6/79, Body Casting Commercial, 20", 900 lb. Gate Valve, C.S.

4. Parts List, No. PL-810-40, Rev. C, Valve Assembly, 20", 900 lb. ,

C.S., Gear Operated, Class 1.

5. Engineering Change Notice No. 81040, Rev. B, Valve Assembly, 20", 900 lb. , C.S. , Gear Operated, Class 1.

6. Engineering Change Notice No. 80996, Rev. B., 3/28/79, Body Assembly, 20", 900 lb. , Gate Valve, C.S.

7. TWX, 2/6/85; Cleveland Electric Illuminating (CEI) Inc. to BA-NVD, i.e., schedule for valve disassembly of 5 valves.

8. TWX, 2/7/85; BW-NVD response to CEI TWX dated 2/6/85.

9. Letter, 11/14/84; CEI to BW-NVD, notification of nonconforming conditions.

s

10. Letter, 12/14/84; BW-NVD to CEI, providing additional information/

investigation into nonconforming conditions.

11. Letter, 11/28/84; BW-NVD to CEI explaining guide rail problem and BW-NVD notification to NRC, i.e., NR0007.

12. QA Manual, Rev. 7, 12/4/84.

13. Letter, 2/26/85; BW-HVD to CEI, covering recommendations for reworking the tack welds, and justification for accepting same welds.

14. Letter, 11/19/84; BW-NVD to NRC, Notification of N.R. 0007 concerning gate valve, 20", 900 lb., body guide rail problem.

15. Letter, 4/6/83; BW-NVD to NRC, notification of N.R. 0006, concerning sensitivity and density variations in radiographs of valve welds.

16. Letter, 5/2/83; BW-NVD to NRC Region V, Interim notification report for N.R. 0005, concerning failure of tack welds in swing check valves.

17. Trip Report, 2/1/83; BW-NVD QA Manager's trip to Comanche Peak to investigate cause of tack weld failures.

54

ORGANIZATION: BORG WARNER CORPORATION NUCLEAR VALVE DIVISION VAN NUYS, CALIFORNIA REPORT INSPECTION NO.: 99900289/85-01 RESULTS: PAGE 13 of 14

18. Letter, 5/2/83; BW-NVD to Arkansas Nuclear One, notification of possible hardware problem of tack welds in swing check valves.

19. Letter, E/2/83; BW-NVD to Mill Power, notification of possible tack weld problem in swing check valves for Catawba, McGuire and Oconee.

20. Letter, 5/2/83, BW-NVD to Pacific Gas & Electric (PG&E), noti-fication of possible tack weld problem in swing check valves for Diablo Canyon.

21. Letter, 5/2/83; BW-NVD to Rochester Gas & Electric (RG&E),

notification of possible tack weld problem in swing check valves for Ginna.

22. Procedure, 5/2/83; BW-NVD Procedure for Reinspection and Rework

, of NVD Swing Check Valves.

23. Letter, 7/22/80; BW-NVD to NRC OIE, notification of N.R. 0003, concerning possible tolerance buildup condition on gate valve stems.

24. Letter, 7/22/80, BW-NVD to NRC Region V, notification of N.R. 0003 and including corrective action to be taken.

25. Letter, 5/30/80; BW-NVD to WPPS-2, notification of N.R. 0002 concerning loosening of stem nuts and extensions on Globe valves with SCM operators.

26. Letter, 5/30/80; BW-NVD to Bechtel Power / Grand Gulf, notification of N.R. 0002 concering loosening of stem nuts and extensions on globe valves with SCM operators.

27. Letter, 6/11/80; BW-NVD to WPPSS-2, recommended corrective action for loosened stem nuts.

28. Internal Memorandum; 6/10/80; BW-NVD, consisting of instructions for documenting field modifications on valve stems.

29. Letter, 2/12/80; BW-NVD notification to NRC concerning N.R. 0001, possible failure of gate valves to close due to inadequate guides.

30. Letter, 2/12/80; BW-NVD to Combustion Engineering /Palo Verde, notification of possible hardware problem due to inadequate guides in gate valves.

l l

l 55 L

ORGANIZATION: BORG WARNER CORPORATION NUCLEAR VALVE DIVISION VAN NUYS, CALIFORNIA REPORT INSPECTION N0.: 99900289/85-01 RESULTS: PAGE 14 of 14 l 31. Procedure P-77910, Rev. A, 10/7/80; BW -NVD Assembly Procedure to l Upgrade Four Valves in Line at C.E. (Palo Verde).

32. Procedure No. 11-8, Rev. O, 5/25/84, Reporting of Defects and l Noncompliances to the NRC. '

33. Drawing No. 79760; Valve Assembly, 6 Inch Gate Va.'ve,150 lbs.,

C.S., M.0.

34. Engineering Change Notice, 1/16/84; Change modifying Drawing No.

> 79760, Note 37 added.

35. Specification, 6/19/84; BW-NVD Customer Specification Discussion Sheets for TVA specifications.

36. Letter, 7/26/83; BW-NVD to NRC Region V, i.e., all affected valves rework completed, see NR-0004.

r I

i l l 56

ORGANIZATION: CARDINAL INDUSTRIAL PRODUCTS CORPORATION LAS VEGAS, NEVADA REPORT INSPECTION INSPECTION NO.: 99900840/85-01 DATE(S): 4/1-5/85 ON-SITE HOURS: 76 CORRESPONDENCE ADDRESS: Cardinal Industrial Products Corporation ATTN: Mr. D. Fielder President i

3873 W. Oquendo Road

! Las Vegas, Nevada 89118 ORGANIZATIONAL CONTACT: Mr. Michael J. Donovan, Chairman TELEPHONE NUMBER:

PRINCIPAL PRODUCT: Fasteners.

NUCLEAR INDUSTRY ACTIVITY: Approximately 75% of Cardinal Industrial Products Corporation's (CIPC) sales are to the nuclear industry.

ASSIGNED INSPECTOR: (f 2/ C 7 E. T. Baker, Reactive Inspection Section (RIS) Ohte OTHER INSPECTOR (S): T. F. Burns, Brookhaven National Laboratory

/

APPROVED BY: f E. W. Merschoff ief, RIS, Vendor Program Branch Date INSPECTION BASES AND SCOPE:

A. BASES: 10 CFR 21, 10 CFR 50 Appendix B, NCA-3800.

B. SCOPE: This inspection was made to review the implementation of CIPC's corrective action plan and determine the extent to which previously identified hardware deficiencies apply to material supplied to Diablo Canyon Nuclear Plant.

PLANT SITE APPLICABILITY: Diablo Canyon (50-275/323); River Bend (50-458);

Clinton Nuclear Power Station (50-461/462).

57 l

10 CFR 2.790 INFORftATION HAS BEEN DELETED

ORGANIZATION: CARDINAL INDUSTRIAL PRODUCTS CORPORATION LAS VEGAS, NEVADA REPORT INSPECTION NO_ 94400R40/85-01 RESULTS: PAGE 2 of 12 A. INSPECTION ISSUES:

1. Is CIPC effectively implementing the Action Plan reviewed and approved by the NRC?

2. Do documentation packages for material supplied to Diablo Canyon exhibit the same deficiencies (e.g., failure to perform NDE, purchases from unapproved sources, insufficient physical testing, loss of traceability) observed in documentation packages for material shipped to other plants?

B. INSPECTION FINDINGS:

The following findings were made during the inspection of the CIPC facility in Las Vegas, Nevada.

1. Followup item on corrective action resulting from Inspection Report Nos. 99900840/83-01 and 84-01.

a. Summary of Issue The CIPC " Action Plan" submitted to the NRC for review and approval, committed to performing revalidation surveys on all suppliers, reviewing all Certified Material Test Reports (CMTRs) for material exceeding 1" diameter, reverifying material chemical and mechanical properties on a sampling basis for material 1" diameter and less, formalizing agency agreements and commitments to pass CIPC's quality requirements down through the procurement chain, and hiring a monitor to supervise and control operations affecting material properties.

b. Inspection Findings (1) For the revalidation audits performed by a third party, Hartford Steam Boiler Inspection Company (HSBIC), the audit results were sent to CIPC and CIPC did not send the noncon-formances to the subtier supplier nor did CIPC request corrective action by the subtier supplier. In addition, CIPC failed to note that companies supplying calibration services to subtier suppliers had not been audited.

10 CFR 2.790 INFORMATION HAS BEEN DELETED 58

ORGANIZATION: CARDINAL INDUSTRIAL PRODUCTS CORPORATION LAS VEGAS, NEVADA REPORT INSPECTION NO.- 99900840/85-01 RESULTS: PAGE 3 of 12 (2) The checklist CIPC was using to review material certifi-cation packages was inadequate in that the CIPC review did not identify the following nonconformances; lack of or insufficient magnetic particle testing, insufficient number of tensile tests, and purchases from unapproved sources. l (3) In qualifying a new supplier, facility, CIPC failed to document the audit. ,

1 A nonconformance was issued on each of these three findings.

2. Investigation of allegations expressed in 2.206 petition from the Government Accountability Project concerning Diablo Canyon and CIPC.

a. Sumary of Issues The allegations investigated during the inspection at CIPC included the weldability of A-307B bolts welded as studs to the containment liner, the quality of the PG&E vendor approval system, and the extent to which previously identified problems, e.g., purchases from unapproved sources, failure to perform NDE and destructive tests, and loss of traceability applied to material supplied to Diablo Canyon.

b. Inspection Findings (1) The inspectors reviewed the certified material test reports for nine heats of A-307b bolts supplied to Diablo Canyon between February 1980 and October 1982. All heats were found to be readily weldable.

(2) The first audit of CIPC by PG&E was conducted in December 1984, after NRC Information Notice IN 84-52 was published.

Prior to that time, PG&E placed CIPC on their Approved Vendor List (AVL) based on audits performed by their subcontractors, Pullman Power Products and Bechtel. The inspection report issued by PG&E was reviewed, and while two problems with material supplied under one purchase were detected, PG&E auditors failed to detect insufficient magnetic particle testing and tensile tests on some addi-tional material in the same order. In addition, nonconform-ances involving material supplied on other PG&E orders went undetected (use of unapproved sources and loss of traceability).

10 CFR 2.790 INFORMATION HAS BEEN DELETED 59

ORGANIZATION: CARDINAL INDUSTRIAL PRODUCTS CORPORATION LAS VEGAS, NEVADA REPORT INSPECTION Nn - QQQOnRan/AR 01 RESULTS- PAGE 4 of 12 (3) In reviewing material supplied under seven PG&E purchase orders and 85 Pullman Power purchase orders, materials for two of the PG&E orders and 15 of the Pullman Power orders were found to have problems similar to those described in IN 84-52, loss of traceability, lack of or insufficient NDE, and insufficient tensile tests.

Since previous nonconformances already address these deficiencies and CIPC has initiated corrective action covering these discrepancies, and PG&E was made aware of the discrepancies during the exit interview, no new nonconformances were issued. CIPC committed to informing other affected licensees as required by their Part 21 procedures.

C. SUPPLEMENTARY INFORMATION:

1. Review of Corrective Action Plan Implementation

a. The inspectors reviewed four revalidation surveys performed by HSBIC for CIPC and two followup audits performed by CIPC. The HSBIC July 31, 1984, and August 1, 1984, of the the October 1984 followup audits of these facilities by CIPC, the HSBIC August 7, 1984, audit of and the HSBIC August 3, 1984, audit of were reviewed by the inspectors. Seven significant nonconformances were identified in these reports. However, contrary to CIPC's QAM, CIPC did not send the subcontractors copies of the nonconformances nor did they request that the subtier suppliers perform any corrective action.

During the review of CIPC's followup audit reports a note on one of the reports stated that had been audited and found acceptable and should be placed on CIPC's AVL. However, there was no completed checklist or audit l

report documenting the audit. Otherwise the followup audit reports appeared adequate.

b. The inspectors reviewed the checklist developed and used by CIPC in reviewing CMTRs for material exceeding 1" diameter.

The checklist did not require the reviewer to go back to the original P.O. to establish whether the customer initiated any supplemental requirement such as NDE, the checklist did not -

require the reviewer to consult the product specification to 10 CFR 2.790 INFORMATION HAS BEEN DELETED 60

ORGANIZATION: CARDINAL INDUSTRIAL PRODUCTS CORPORATION LAS VEGAS, NEVADA REPORT INSPECTION NO - 90900R40/85-01 RESULTS: PAGE 5 of 12 assure that the proper number of samples were examined by NDE, the checklist did not require the reviewer to consult the heat treat record or the product specification to assure that the proper number of tensile tests were performed, the checklist did not require the reviewer to establist that traceability 4 existed back to the material manufacturer's CMTR, nor did the '

checklist require the reviewer to establish that the material was purchased from an approved source.

The inspectors reviewed several P.O.s which had already been reviewed by CIPC personnel to determine if the checklist inadequacies affected the review. A discrepancy was discovered where the number of samples selected for magnetic particle examination did not comply with the AQL of 0.25 as specified in ASTM A490-81. Specifically, material was supplied to Baldwin Associates (Clinton Nuclear Power Station) on CI 31267, Cert.

No. 38789 for 11 - 7 x 3, A490-81 Type 1 bolts (Heat No. 110039) from an eight hundred piece production lot where only fifty pieces had been inspected. In order to fully comply with the standard, it would be necessary to inspect a minimum of eighty pieces. An additional discrepancy was found in regards to the number of tensile test specimens that were tested from each production lot. ASTM A490 requires (for the Production Lot Method as used by CIPC) that for lots of 801 to 8000 pieces, two tensile specimens be tested. The records examined indicated that only one tensile specimen had been tested from the lot of 816 pieces processed on 10/8/82.

CIPC agreed that the checklist was deficient and committed to reviewing all previously reviewed documentation packages for these deficiencies,

c. Under the material reverification testing program for material 1" diameter and less only a very small number of the material heat / product form combinations had been retested for chemical and mechanical properties at the time of the inspection. These were reviewed to determine that the proper sample size had been tested and that the test results were within specification. No nonconformances were noted.

d. The inspectors reviewed CIPC Standard Practice 8.004, " Agent Surveillance" which establishes how CIPC monitors their Trading Agents, as well as reviewed the standard language CIPC uses in formalizing their agreements with their Trading Agents.

Both documents appeared adequate.

10 CFR 2.790 INFORMATION HAS BEEN DELETED 61 l

l

ORGANIZATION: CARDINAL INDUSTRIAL PRODUCTS CORPORATION LAS VEGAS, NEVADA REPORT INSPECTION NO - 00900R40/R5-01 RESULTS: PAGE 6 of 12

2. Review Documentation for Material Supplied to Diablo Canyon

a. An evaluation of weldability was made on a selection of lots of ASTM A-307 fastener material supplied by CIPC for use at the Diablo Canyon nuclear facility. The selection was made from available records covering the period from 1980 to 1982.

The following heats of material were evaluated for the listed Purchase Orders (P.O.s).

Analysis P.O. Date Heat C Mn P S Si 8364 2/15/80 L24813 .17 .38 .021 .018 --

8791 9/4/80 322H380 .19 .42 .017 .045 --

9521 4/28/81 B850587 .16 .37 .030 .019 .01 9520 5/22/81 EA8222 .17 .44 .011 .008 .01 10906 8/9/82 KA8222 .17 .44 .011 .008 .01 5952356 9/7/82 369410 .05 .31 .01 .023 .25 5952356 9/7/82 363233 .04 .33- .015 .024 .230 5952356 9/7/82 367348 .05 .28 .011 .017 .210 5952356 10/26/82 212027 .23 .58 .024 .048 .150 All of the above materials can be classified as low carbon steels (carbon content of less than 0.25% and an alloy content of less than 1.50%). The above materials would be considered as generally weldable without special precautions.

It should be noted that ASTM A-307 does not impose maximum limita-tions on the carbon or manganese content of the material. This

, places considerable discretion in the hands of the manufacturer with the purchaser presumed to be aware of the effect carbon and

! manganese content has on the weldability of the material.

The data examined represents an approximate fifteen percent of the total orders made by Pullman Power products from CIPC for use at Diablo Canyon.

b. Because of time constraints the inspectors only reviewed the December 6-7, 1984, and January 3-4, 1985, PG&E audit reports.

Pullman Power Products audit reports were not reviewed.

10 CFR 2.790 INFORMATION HAS BEEN DELETED

.62

ORGANIZATION: CARDINAL INDUSTRIAL PRODUCTS CORPORATION LAS VEGAS, NEVADA REPORT INSPECTION NO.- 99900840/85-01 RESULTS: PAGE 7 of 12 The December 6-7, 1984, audit report indicated that PG&E auditors had reviewed P0 4R-66170 and identified that CIPC had not performed sufficient tensile tests on a lot of 11" x 61" A-490 bol ts. The auditors failed to identify that insufficient tensile tests were performed on a lot of li" x 51" A-490 bolts supplied under the same P0 or that insufficient magnetic particle testing was performed on both sizes of bolts.

In addition, on PG&E PO 577706 CIPC supplied ASTM A-194 Grade 4 1-7/8" Heavy Hex Nuts from an unapproved source and lost tracea-bility on the ASTM A-193 B-161-7/8" x 10" studs supplied under the same purchase order.

During the exit interview PG&E personnel stated that the intent of the December 6-7, 1984, audit was twofold; to review a sample of PG&E orders to ascertain whether or not previously identified problems affected material supplied to Diablo Canyon, and to review CIPC's new program to determine whether or not CIPC could be placed back on PG&E's AVL. PG&E was only reviewing a sample of the P0s because CIPC's corrective action program would cover the remainder of the material.

The January 3-4, 1985, audit was performed to review corrective actions required by PG&E as a prerequisite to restoring CIPC to PG&E's AVL. The result of the audit was to reinstate CIPC to the AVL on March 13, 1985.

c. The inspectors reviewed seven PG&E purchase orders and 85 Pullman purchase orders.

(1) On PG&E P0 4R-66170 for 340 11" x 51" and 340 11" x 61" A-490 bolts, CIPC failed to perform sufficient tensile and magnetic particle tests. For production heat treatment lots from 801 pieces to 1000 pieces, ASTM A-490 requires that two tensile tests be performed. A-490 also requires that the sample size for magnetic particle testing be based on an acceptable quality level (AQL) of 0.25. The heat treatment lot sizes for both sizes of bolts exceeded 801 pieces and the lot sizes that were magnetic particle tested were 1000 and 500, respectfully. A-490 requires that two tensile tests be performed for the heat treatment lots for both sizes of bolts and only one had been performed.

The production sample sizes for magnetic particle testing 10 CFR 2.790 INFORMATION HAS BEEN DELETED 63 l

ORGANIZATION: CARDINAL INDUSTRIAL PRODUCTS CORPORATION LAS VEGAS, NEVADA REPORT INSPECTION i NO.- 99900840/85-01 RESULTS- par # R nf 17

< should have been 80 bolts and 50 bolts, respectively, but

! CIPC only tested 32 bolts'and 20 bolts, respectively. In l

addition, one of the 32 bolts failed. A-490 then requires i 100% examination of all bolts in the lot. CIPC did not perform any additional testing as a result of the failure. ,

(2) On PG&E order 577706 for 48 ASTM A-193 B16 1-7/8" x 10" studs and 48 ASTM A-194 Grade 4 1-7/8" Heavy Hex Nuts, CIPC purchased the nuts from an unapproved source and lost traceability on the studs. The nuts were purchased from and were manufactured by in 1977. The first audit of was performed in August 1978 and of in November 1980. This is contrary to paragraph 8-3(b) of the January 23, 1982, revision of CIPC's QAM which is what CIPC certified the material to. According to the shop traveler for the studs, the material for the studs was cut into 10" lengths on 2/29/84, visually inspected on 3/1/84, charpy impact tests performed on 3/8/84, magnetic particle tested as random lengths on 3/13/84 and shipped 3/13/84. Traceability between the material cut to 10" lengths on 2/29/84 and the material magnetic particle tested as random lengths on 3/13/84 could not be established.

(3) A number of purchase orders to CIPC from Pullman Power Products for Diablo Canyon nuclear plant for various quantities of ASTM A-490 alloy steel bolts were issued during the period 1980 to 1982. An examination was made of the documents which covered the items supplied in fulfillment of some of these orders. It was discovered that several orders had been filled by shipping bolting material which had not been magnetic particle tested in accordance with the requirements of ASTM A-490 (para. 11).

These purchase orders are as follows:

Magnetic P.O. Pcs. CI No. Heat No. Date Particle 8528 25 19439 H1124L 5/8/80 No 8528 19 19439 8878948 5/8/80 No 8523 32 19406 Y31200 4/30/80 No 8523 10 19406-1 H1124L 5/8/80 No 8523 '2

. 19406 Y51054 4/30/80 No l

l l

10 CFR 2.790 INFORMATION HAS BEEN DELETED 64 I l i l

ORGANIZATION: CARDINAL INDUSTRIAL PRODUCTS CORPORATION LAS VEGAS, NEVADA REPORT INSPECTION NO.- 99900840/85-01 RESULTS: PAGE 9 of 12 Magnetic ,

P.O. Pcs. CI No. Heat No. Date Particle j l 8641 24 20166 8878948 6/9/80 No i

8695 10 20606 8067977 7/3/80 No 10331 150 25952 N32625 2/15/82 No 5952356 10 27541 L83395 9/28/82 No 5952356 16 27541 L21392 9/28/82 No 11124 10 27738 N12900 10/18/82 No Additional shipments of material were made during this period where the required magnetic particle tests had been performed. It appeared that CIPC personnel had simply overlooked the requirement. CIPC could offer no other explanation for this omission.

(4) During the review of orders placed by Pullman Power Products for the Diablo Canyon nuclear plant and the certified material test raports issued, certain discrepancies regard-ing the origin of this material were discovered. The term

" Truck Orders" refers to " Truck Load" shipments of bolt blanks or finished parts received from subtier suppliers.

This type of material and part procurement by CIPC occurred during the period 1979-1981. It is possible that this time frame could be expanded by further investigation. The inability to verify the origin and true identity of the material stems from the fact that all relevant data (heat number, chemical analysis and physical properties) had been transcribed on to certification sheets from Cleveland, Ohio. No original certified material test reports from the material manufacturer were available to confirm this data. For example, one order to Pullman Power Products (7177-7604, 4/6/79) for 4 pcs of 1-3/4"-8UN-2AX8 3/4 A490 Hex Head Machine Bolts was shown as Heat No.

43152 with complete chemical analysis and physical proper-ties including charpy impact results at a specified temperature. However, it was noted that the report from which this data was taken also listed forty three additional heats of material by grade, heat number, chemistry and

physical properties.

l l

(5) The substitution of various grades and/or types ;f bolting hardware is permitted by the various ASTM standards in those i cases where the purchaser has neglected to indicate his need l in the purchase requirements. This is a common omission 1

t 10 CFR 2.790 INFORMATION HAS BEEN DELETED 65

ORGANIZATION: CARDINAL INDUSTRIAL PRODUCTS CORPORATION LAS VEGAS, NEVADA REPORT INSPECTION NO.- 99900840/85-01 RESULTS: PAGE 10 of 12 on the part of the buyers of bolting hardware. Virtually all orders emanating from Pullman Power Products for Diablo Canyon did not specify the type of bolt covered by the standard until about the period beginning 1984.

Consequently, CIPC made substitutions at their facility, i ASTM A-325 permits the manufacturer to substitute Type 1 or Type 2 material when the buyer neglects to specify type.

A Type 3 is available within this standard but the manu-facturer is not permitted to make this substitution without the consent of the purchaser. Type 1 and 2 bolts are low carbon steels, although the carbon in Type 1 is specified as a minimum with no maximum, and Type 3 is a low alloy material .

During a review of documentation regarding orders placed for ASTM A325 by Pullman Power Products where Type 1 bolts had been specified, it was discovered that a low alloy material (AISI 4140 composition)-had been supplied.

This material did not fall into any of the classes of ASTM A-325 and, there was no evidence that the purchaser had agreed to the change. The specific orders examined were:

P.O. CI Item Quantity Heat No. Ship. Date

1) 7177-10331 25952 7/8 - 9x2) 20 2-0500 2/4/82

2) 7177-10331 25952 3/4 - 10x2-3/4 -

2-3200 2/4/82

3) 7177-7387 12195 1-3/4 - 8x1' 4" 2 6752 2/23/79 For these three different items the following material analyses were provided:

C Mn P S Si Ni Cv Mo Cu

1) .40 .79 .011 .025 .23 .12 .85 .166 .25

2) .42 .81 .019 .033 .24 .16 .94 .227 .25

3) .45 .97 .016 .019 .23 .16 1.02 .21 .12 The above material analysis is within the specified range for AISI 4140 - a low alloy material. The physical proper-ties (tensile, yield, elongation and reduction in area) for this material will differ considerably from low carbon steel even in the annealed condition. In some applications, it 10 CFR 2.790 INFORMATION HAS BEEN DELETED 66 i

ORGANIZATION: CARDINAL INDUSTRIAL PRODUCTS CORPORATION LAS VEGAS, NEVADA REPORT INSPECTION NO - 99000R40/85-01 RESULTS: PAGE 11 of 12 would be desirable from the end users point of view to know that the material supplied was of a much higher strength level. Also, there could be certain applications or environments where the end user should be aware of the substantial difference in chemical analysis and physical properties. PG&E was not informed of any substitutions and therefore could not review the application of the material.

i D. PERSONS CONTACTED:

CIPC M. J. Donovan, Chairman D. Fielder, President T. Graham N. Henderson, QA Manager L. Mydzowski, Admin. Manager PG&E D. Aaron R. F. Locke B. Norton S. Skidmore M. Tresler CYGNA D. Smedley NRC Region V R. T. Dodds E. DOCUMENTS EXAMINED:

Pacific Gas & Electric Documentation Packages P0 T521791 dated 2/11/83 P0 4R-66170 P0 577706 dated 1/30/84 P0 4R67175 dated 5/11/84 P0 593976 dated 3/11/84 P0 594705 P0 057152 10 CFR 2.790 INFORMATION HAS BEEN DELETED

67 l

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t

ORGANIZATION: CARDINAL INDUSTRIAL PRODUCTS CORPORATION LAS VEGAS, NEVADA REPORT INSPECTION NO.- 99900840/85-01 RESULTS: PAGE 12 of 12 PG&E Audit Report for Audits performed December 6-7, 1984 and January 3-4, 1985.

Stone and Webster Documentation Packages (River Bend)

P0 12210-28638 dated 4/20/84 P0 12210-27906 dated 3/12/84 CIPC Standard Practices 8.004 " Agent Surveillance" CIPC Followup Audits dated 12/14/84 Hartford Steam Boiler Inspection Company Audits of CIPC Subcontractors performed 7/31-8/1/85, 8/7/84, 8/3/84.

i 10 CFR 2.790 INFORMATION HAS BEEN DELETED 68 l

ORGANIZATION: COMBUSTION ENGINEERING POWER SYSTEMS GROUP WINDSOR, CONNECTICUT REPORT INSPECTION INSPECTION N0.: 99900401/85-01 DATE(S): 2/25-3/1/85 ON-SITE HOURS: 62 CORRESPONDENCE ADDRESS: Combustion Engineering, Inc.

Power Systems Group ATTN: Mr. C. W. Hoffman, Director Quality Assurance 1000 Prospect Hill Road

! Windsor, Connecticut 06095 ORGANIZATIONAL CONTACT: Mr. P. D. Ford, Supervisor, Group QA TELEPHONE NUMBER: (203) 285-9210 PRINCIPAL PRODUCT: Nuclear Steam Supply Systems.

NUCLEAR INDUSTRY ACTIVITY: The Power Systems Group, Combustion Engineering (CE),

had contracts for 16 domestic reactor units to date, of which five (5) are in the design and construction phase. In addition, they have modification / repair /

service contracts for 16 reactor units.

i ASSIGNED INSPECTOR: E.M W b R. P. McIntyre, Special Projects Inspection h26f86

'Dat6 Section (SPIS)

OTHERINSPECTOR(S): W. Shier, BNL APPROVED BY:

4 Io 26ffys fo( John W. Craig, Chief, SPIF, Vendor Program Branch ) ate INSPECTION BASES AND SCOPE:

A. BASES: 10 CFR Part 50, Appendix B and Topical Report CENPD-210-A.

B. SCOPE: The purpose of this inspection was to obtain and review selected Field Action Requests (FAR), Corrective Action Reports, and Availability Data Program Infobulletins for followup at CE designed plants.

l PLANT SITE APPLICABILITY: Multiple plant applicability including Palo Verde (50-528, 529, and 530).

69

7 ORGANIZATION: COMBUSTION ENGINEERING k POWER SYSTEMS GROUP f WINDSOR, CONNECTICUT f  !

REPORT INSPECTION I

p N0.: 99900401/85-01 RESULTS: PAGE 2 of 5 9 h

A. VIOLATIONS: '

None.

8. NONCONFORMANCES:

If None. [t p

C. UNRESOLVED ITEMS:

None.

(

D. STATUS OF PREVIOUS INSPECTION FINDINGS:

1. (0 pen) Nonconformance (84-02): No internal audits have been performed on the error reports pertaining to the CESEC computer code. I Not inspected during this inspection. i l'

2. (0 pen) Nonconformance (84-03): Computer code FATES 3A verification analysis (0000-TH-186) was found to have insufficient information k

[

concerning the test problems to evaluate the intent or adequacy of r the verification runs. j Not inspected during this inspection.

[

.i

3. (0 pen) Nonconformance (84-03): No verification calculations were available for the STRIKIN 11 computer code.

Not inspected during this inspection.

4. (0 pen) Nonconformance (84-03): The verification calculations performed for the CELOA and HCROSS computer codes were not independently reviewed.

Not inspected during this inspection.

5. (0 pen) Nonconformance (84-03): A modification implemented in the 78226 version of the CELDA computer code was not tested and verified.

Not inspected during this inspection.

70

l ORGANIZATION: COMBUSTION ENGINEERING POWER SYSTEMS GROUP WINDSOR, CONNECTICUT REPORT INSPECTION NO.: 99900401/85-01 RESULTS: PAGE 3 of 5 E. OTHER FINDINGS OR COMMENTS:

During this inspection, the Combustion Engineering (CE) Corrective Action Program, the Availability Data Program, and the Field Action Request 4

system were surveyed to identify significant safety related occurrences and concerns affecting CE operating plants and CE plants currently under construction. In addition, procedure API- U, for identification, evaluation, and disposition of potential ifety hazards as required by 10 CFR Part 21, was reviewed.

1. Corrective Action Program (CAP) - This program monitors the perform-ance of operating reactors to identify safety concerns with systems 4

and components supplied by CE. Sources of information for the CAP i

include Licensee Event Reports (LERs), CE Field Action Requests (FARs), and utility feedback through the Availability Data Program.

The CAP provides quarterly reports to CE Management as a mechanism for identifying safety related problems before they occur and for

initiating resolutions of identified safety concerns when necessary.

l 2. Availability Data Program (ADP) - This is a reporting system for

• performance and reliability concerns for both operating plants and plants under construction incorporating CE Nuclear Steam Supply d

Systems. Through this program CE receives feedback information on equipment problems which might affect the performance of systems

, at a nuclear power plant. The program provides an advisory service to utilities that have purchased CE nuclear plant equipment.

Input information for the program is obtained from Licensee Event Reports (LERs) and various other reports related to plant operations such as, monthly operating reports, plant outage reports, and main-tenance reports. CE requests the utilities to return this information to assist CE in developing availability design improvements for operating as well as future plants. The ADP issues quarterly reports to Utility hanagement summarizing recent problems and occurrences.

CE also issues a more detailed Availability Data Program Infobulletin to utilities who purchase CE nuclear plant equipment. Infobulletins discuss technical developments related to the application or operation of nuclear plant equipment supplied by CE. Acknowledgement of receipt of the ADP Infobulletin by the utilities is requested by CE. Followup action on these items is then considered to be the responsibility of the utilities.

1 il

ORGANIZATION: COMBUSTION ENGINEERING POWER SYSTEMS GROUP WINDSOR, CONNECTICUT REPORT INSPECTION )

N0.: 99900401/85-01 RESULTS: PAGE 4 of 5 1

3. Field Action Requests (FARs) - This system was established by CE for nuclear plants under construction to coordinate required field changes between CE Windsor Engineering personnel and CE staff at the construction site. FARs may be generated at Windsor or at the site.

In addition, concurrence for FAR closure is required by both Windsor and the construction site personnel.

This system provides a mechanism for tracking design changes, equip-ment malfunctions, and procedure modifications that are identified during construction. FARs associated with safety-related equipment l at the Palo Verde project were chosen as representative examples for review during this inspection.

l

4. API Reporting of Safety Hazards - This CE administrative manual l describes the procedures to be followed for reporting and evt.luating potential substantial safety hazards to conform with the requirements of 10 CFR Part 21. The system includes a four part report that includes: (1) a description of the potential safety hazard by the originator, (2) an evaluation by the immediate manager, (3) an evaluation by the applicable department director, and if both eval-uators concur that a substantial safety hazard might result, then (4) the Substantial Safety Hazard Report is forwarded to the Nuclear Power Systems Nuclear Safety Committee for further evaluation. If their evaluation also concludes that a substantial safety hazard exists.

then the NRC is notified in accordance with 10 CFR Part 21.

It was noted during this inspection that API-17 provides for traceability of the disposition of the potential safety hazard report at each stage except the first (i.e., from the originator to the immediate manager). The report does not enter a formal log until a transmittal is made between the manager and the department director.

5. Summary - During this inspection selected issues in Corrective Action Program Quarterly Reports, Availability Data Program Infobulletins, Field Action Requests for the Palo Verde Project, and API-17 Reports were reviewed. The information reviewed will be utilized during future NRC inspections of vendor / utility information interface to review utility receipt, evaluation and implementation of action, if 72

ORGANIZATIC N: COMBUSTION ENGINEERING POWER SYSTEMS GROUP WINDSOR, CONNECTICUT REPORT INSPECTION NO.: 99900401/85-01 RESULTS: PAGE 5 of 5 any, as determined by the utility to be appropriate. CE provided the NRC inspector with copies of Availability Data Program Infobulletins for 1982, 1983, aid 1984 (43 total).

i d

1 o

! 73

ORGANIZATION: CORPORATE CONSULTING & DEVELOPMENT COMPANY, LTD RESEARCH TRIANGLE PARK, NORTH CAROLINA REPORT INSPECTION INSPECTION NO.: 99900511/85-01 DATE(S): 5/6-10/85 ON-SITE HOURS: 67 CORRESPONDENCE ADDRESS: Corporate Consulting

& Development Company, Ltd.

ATTN: Dr. J. R. Yow President P.O. Box 12728 Research Triangle Park, N.C. 27709 ORGANIZATIONAL CONTACT: Mr. Carson Blanton, Jr. , QA Manager TELEPH0tlE NUMBER: (919) 362-8800

PRINCIPAL PRODUCT

Engineering, consulting, and testing services.

NUCLEAR INDUSTRY ACTIVITY: Corporate Consulting and Development Company, Ltd.

(CCL) provides engineering consulting and testing services to the nuclear industry for seismic analysis, testing, and nuclear environmental qualifications of equipment.

ASSIGNED INSPECTOR: fa1 & ffft,.i d YJMA/ffS l Rf N. Moi ~st, Equipsfent Qualification Inspection Date Section (EQIS) OTHER INSPECTOR (S): G. T. Hubbard, EQIS i I APPROVED BY: b -SAP G -1 MI U. Potapovs, Chief, EQIS', VPB Date I' INSPECTION BASES AND SCOPE: A. BASES: Appendix B to 10 CFR Part 50 and 10 CFR Part 21. B. SCOPE: This inspection consisted of : (1) a technical evaluation of equipment qualification (EQ) test activities for safety-related equip-ment, (2) witnessing of inprocess EQ testing; and (3) verification of implementation of the quality assurance (QA) program. PLANT SITE APPLICABILITY: Brunswick 1 & 2 (50-324 and 50-325) [ l l 75

ORGANIZATION: CORPORATE CONSULTING & DEVELOPMENT COMPANY, LTD RESEARCH TRIANGLE PARK, NORTH CAROLINA REPORT INSPECTION MO - 99900511/85-01 RESULTS: 3 AGE ? of 5 A. VIOLATIONS: None i B. NONCONFORMANCE:

1. Contrary to Criterion V of Appendix B to 10 CFR Part 50 and section 5 of the Quality Assurance Manual (QAM), no documented procedure existed for the control of mixing the chemical spray solution used during design basis event testing and monitoring the solution's PH.

C. UNRESOLVED ITEMS: None D. OTHER FINDINGS OP COMMENTS:

1. Background Carolina Power and Light Company (CP&L) contracted with CCL to conduct EQ testLine Energy (vibration Break aging),

(HELB) thermal aging, radiation, of limit, control seismic and pushbutton and High switches, terminal blocks, and solenoid valves. Test requirements were defined in three Patel Engineers Qualification Test Plans (QTPs), (PEl-TR-83-4-25 REVC-QTP for switches, PEl-TR-83-4-26 REV C-QTP for terminal blocks, PEl-TR-83-4-27 REV E - QTP for solenoid valves as required by CP&L purchase order (P0) to CCL. CCL wrote detailed test proce-dures for Qualification Testing based on the three Patel Engineers QTPs. These Qualification Tests were performed under CCL Job Number (JN) 1857.

2. Post HELB Inspection and Functional Tests for JN 1857:

The NRC inspectors observed and visually inspected switches and terminal blocks after the test chamber cover was removed and again after the removal of the specimens from the chamber. Visual inspection of a DC terminal block GE CR1510 showed a build-up . of foreign material between two terminal lugs. Following functional I testing (FT) the CP&L representative sent the terminal block to a lab for analysis of the foreign material. CCL personnel took photos of the terminal blocks to document their conditions. The inspectors witnessed post HELB FT (insulation resistance and continuity measure-ments) for the switches and terminal blocks. The NRC inspectors reviewed detailed Test Procedure (TP) 1857-1-1/REV 1/FT for switches, TP 1857-1-2/REV 1/FT knob and key switches, TP 1857-1-3/REV 1/FT pushbutton switches, TP 1857-1-4/Rev 1 FT indicating lights and 76

ORGANIZATION: CORPORATE CONSULTING & DEVELOPMENT COMPANY, LTD RESEARCH TRIANGLE PARK, NORTH CAROLINA REPORT INSPECTION un - coonng11/ng_n1 RFquiTs- PAGE 3 of 5 TP 1857-1-7/REV 3/FT terminal blocks. It was determined that the tests and inspection activities were being performed in accordance with the detailed test procedures and CCLs QAM Revision 1 dated 12-10-84. The NRC inspectors also evaluated the calibration status l and accuracy of the instrumentation used to perform the Post HELB FT by review of instrumentation calibration records.

3. Test Results of Post HELB FT for JN 1857 FT was started May 7, 1985, and was completed on May 9, 1985.

During the testing, four test anomalies were identified and were documented by Record of Anomaly (R0A) per CCL's QAM. The four ROAs are summarized as follows: (1) The knob on a Honeywell oil tight manual control switch would not turn clockwise or counterclockwise. (2) The pushbutton on a GE oil tight pushbutton switch could not be depressed. (3) A Honeywell limit position switch OP-AR did not meet the accep-tance criteria of the QTP for IR measurements. (4) A Honeywell limit position switch OPD-AR did not meet the accep-tance criteria of the QTP for continuity measurement. CCL and CP&L are currently reviewing and evaluating the four ROAs. The results of the evaluation will be reviewed in a future inspection.

4. Technical Evaluation The NRC inspectors performed an in-depth technical evaluation and review of previous testing conducted on three test programs for qualification of safety-related electrical equipment. The following table summarizes the test programs examined including equipment type and types of documents examined.

Test Program Equipment Type Documents Examined 1857 Solenoid valves; terminal QTP, TPs, P0's, blocks; switches; and receiving slips, indicating lights. test data, test monitor log sheets. 77

ORGANIZATION: CORPORATE CONSULTING & DEVELOPMENT COMPANY, LTD RESEARCH TRIANGLE PARK, NORTH CAROLINA REPORT INSPECTION NO.- 99900511/85-01 RESULTS: PAGE 4 of 5 Test Program Equipment Type Documents Examined 1859 Electro switches, indi- TR, calculation cating lights, fuseblock, file for thermal fuse, terminal block & aging, P0s, speci- l enclosure. ficatien test plan. 1914 Electre switches, indi- TR cating lights The NRC inspectors reviewed the EQ process prescribed in each test plan and reviewed test results, including the bases for accelerated thermal aging and radiation, and verified calculations. Each of the three EQ test plans and related engineering documents were examined for the following:

a. Adequate test instrumentation and their accuracies were described and used to meet the requirements of NUREG- <

0588/IEEE-STD-323/1974.

b. Equipment interfaces were addressed.

c. Test acceptance criteria were established as described in the test specification or in the design engineering documents, such as calculations and engineering letters to meet the requirements of NUREG-0588/IEEE-STD-323/1974.

d. Same equipment was used for all phases of testing and represented a standard production item.

l l e. Environmental conditions were established and described l (e.g., pressure and temperature profiles, and thermal l aging factors were consistent with those outlined in l the test specification or test plan).

f. Test results were adequately reduced and evaluated against established acceptance criteria described in customer test specifications or purchase orders and these requirements had been met.

g. All prerequisites for the given tests as outlined in the test specification had been met.

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ORGANIZATION: CORPORATE C0ilSULTING & DEVELOPMENT COMPANY, LTD RESEARCH TRIAL:GLE PARK, NORTH CAROLINA l 1 REPORT INSPECTION I NO.- 99900511/85-01 RESULTS: PAGE 5 of 5

h. Test equipment included a description of all material, parts, and subcomponents.

i. ROAs were properly documented.

No nonconformances were noted during this review.

5. Other Testing The NRC inspectors observed a test set up for Loss of Coolant Accident (LOCA) testing of conduit seals and anti-wicking splices for cables. During the observance of the test setup the NRC inspector determined, after interviews with CCL personnel, that CCL did not have a documented procedure for the control of mixing the chemical spray solution and monitoring the solution's PH.

(see nonconformance B.1) The QA Manager provided the NRC inspectors prior to leaving CCL a copy of a new data sheet called " Chemical Spray Data Sheet Form TF-114" which showed calculations of the mixing ratios, initial PH and Daily PH readings, a revised test monitor log which now verifies that Form TF-114 has been prepared, and a revised Accident Simulation testing checklist which now contains a check for preparing Form TF-114. I 1 79 i J

ORGANIZAT10N: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY PUSINESS OPERATIONS SAN JOSE, CALIFORNIA REPORT INSPECTION INSPECTION NO.: 99900403/84-04 DATES (S) 10/15-19/84 ON-SITE-HOURS 244 CORRESPONDENCE ADDRESS: General Electric Company Nuclear Energy Business Operations ATTN: W. H. Bruggeman, Vice President & General Manager 175 Curtner Avenue San Jose, California 95125 l ORGANIZATIONAL CONTACT: Mr. J. J. Fox, Senior Program Manager TELEPHONE NUMBER: (408) 925-6195 i PRINCIPAL PRODUCT: Nuclear steam system, services and fuel. NUCLEAR INDUSTRY ACTIVITY: General Electric Company (GE), Nuclear Energy ' ' Business Operations (NEB 0), has a work force of approximately 6,500 people assigned to domestic power plant activity. [. . ASSIGNED INSPECTOR: A Pg Sears, Special Projects Inspection Section 7NT

                                                                                  ~Date OTHER INSPECTOR (S):  0. Gormley, NRC           W. Shier, BNL R. McIntyre, NRC          G. Parker, EG8G D. pl,EG&G             E. Harris, EG8G R.7 ailson, EG&G APPROVED BY:                                    h Jo' n W.Craig, Chief, SPIS, VPB                          [)

Date INSPECTION BASES AND SCOPE: A. BASES: GE Topical Report No. NED0-11209-04A and 10 CFR Part 21. B. SCOPE:

1. Status of previous inspection findings.

2. Validation / verification of General Electric's (GE's) plant transient computer codes and structural computer codes used at GE. (continued next page) PLANT SITE APPLICABILITY: Multiple plant applicability (BWRs) 81

F u ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS SAN JOSE, CALIFORNIA REPORT INSPECTION RESULTS: PAGE 2 of 16 NO.: 99900403/84-04

3. GE's computer code error handling.

l

4. GE's Potentially Reportable Condition files.

5. Various reactive items as listed in this report.

A. VIOLATIONS None B. NONCONFORMANCES:

1. Contrary to GE Quality Assurance Topical Report NED0-11209, Rev. 4 Section 3.12, " Design Change Control," Engineering Operating Pro-cedures (E0P)40-3.00 " Engineering Computer Programs" (ECPs), does not require that Control Components (responsible engineers for ECP's) define other design documents affected by computer code changes or errors, or coordinate these changes 'with other responsible engineers whose documents are affected. Further, Section 4.1 of the same pro-cedure (EOP 40-3.00) does not reqtiire that the Contrcl Component interface with responsible engineers affected by a computer code error, and assess effects of computer code errors on designs, past and present.

2. Contrary to E0P 40-3.00, " Engineering Computer Programs," the Design Pecord File (DRF) for the CRNC-04 computer code (No. A00-01619) did not include all of the code testing specified in the Software System Specification.

3. Contrary to E0P 42-6.00, " Independent Design Verification," The ver-ification< of calculations described in GE Topical Report NEDE-25518 was not completed until a_f ter issuance of 'the report.

4. Contrary to E0P 42-10.00, " Design Records Files," the DRF for the PANACEA Core Design System (No. 670-0005) did not always identify the originator, reviewer, or date performed.

) 5. Contrary to Section 3.10 of the QA' topical report NED0-21109-04A, l application of tre SAP 4G07 code was not fully verified in the follow-ing areas:

a. Two options of the beam element (fixed end forces and shear deformetion analysis) and one option of the pipe element (the c

ASME code analysis) had no verification provided. l l l ! 82 t

ORGANIZATION: GENERAL ELECTRIC COMPANY i NUCLEAR ENERGY BUSINESS OPERATIONS ' SAN JOSE, CALIFORNIA REPORT INSPECTION N0.: 99900403/84-04 RESULTS: PAGE 3 of 16

b. One nodal point option (slaved degrees of freedom) and one option of the beam element (released degrees of freedom) had verification for the latest version only. However, an earlier version of the SAP 4G07 code (which is a Level 3 program), is still available for use on safety-related designs.

6. C.ontrary to E0P 42-6.00, the method from which analytical results were obtained in the SAP 4G07 computer program verification problems 4.1, 4.2, 5.1, 8.1, and 14 was not referenced, nor were any hand calculations included.

7. Contrary to E0P 40-3.00, " Engineering Computer Programs," users are reporting potential computer code errors verbally to the responsible engineer without the required documentation.

C. UNRESOLVED ITEMS

1. Potentially Reportable Condition File (PRC) 84-30, " Errors in the RVRIZO2 Engineering Computer Program," deals with errors that were identified in the RVRIZ02 code following its use in safety-related calculaticns for containment system design. GE has corrected these errors, implemented additional code modifications, and reevaluated the analyses. The results indicate that the previous calculations were conservative for the MARK I containment system. However, GE has supplied this code to utilities and architect angineers who may have used it for Mark I, II, or III containment system calculations.

Thus, these potential code users external to GE are being notified by GE of the code errors. In addition, a report in the PRC F+le . (dated October 8, 1984) indicated that the code has been designated as Level 2 (approved for design applications) since 1977 and may have been used in other safety-related analyses such as piping design. Thus, a review of all uses of RVRIZO2 is being performed by GE to insure that all safety-related code applications have been identified and evaluated. The modified version RVRIZO2 (designated as REFIX) has been the subject of a Design Review but has not been approved as a Level 2 code. The Design Team had several obser 6.'ons regarding further development and verification prior to C # ? approval and noted that

          " REFIX" is not qualified for desigr (ci t ions but only for " general design assessment." Thus, a review u L A es of the code should be performed to insure that it has not been used beyond the current level of qualification.

This item will be covered in a future inspection. 1 l l l 83 l

ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS SAN JOSE. CAU FORNI A_ , l ! REPORT INSPECTION N0.: 99900403/84-04 RESULTS: PAGE 4 of 16 D. STATUS OF PREVIOUS INSPECTION FINDINGS

1. (0 pen) Nonconformance (84-02): Contrary to Engineering OperM k.9 Procedure (E0P) 42-10.00, Section 4.2.d.4, concerning Caign Record Files (DRF's), the DRF's that supported the verifiption of computer calculations for SAFER 02 computer code (DRF's No. A00-01249, A00-1320, and E00-137) did not identify the reviewer and date when performed.

In addition, the calculations did not always identify the originator and date performed. The DRF's supporting the verification calculations for the SAFER 02 computer code (DRF's No. A00-01249, A00-1320, and E00-137) were not updated prior to this inspection. However, two actions have been taken by GE to prevent recurrence of this type of nonconformance: (1) the Manager, Core and Fuel Technology, has issued a letter to all engineers responsible for engineering computer programs reiterating the DRF requirements for verification calculations; and (2) Quality Assurance Newsletter (dated August 1984) has been issued to all engi-neers and managers that includes a "DRF Closecut Checklist" with reminders about signing and dating DRF entries. This item will be the subject of a future inspection at GE.

2. (Closed) Nonconformance (84-02): Contrary to E0P 40-3.00, Section 4.3.B.1., concerning generating and maintaining the DRF, the DRF for the computer code SAFER 02 did not contain a completed user's manual.

A review of the SAFER 02 DRF (No. A00-1716) indicated that a missing reference to the User's Manual (NEDE-30463) is now included in the DRF, thus satisfying the requirements of E0P 40-3,00.

3. (0 pen) Nonconformances (84-02): Contrary to E0P 42-1.00, Section 3.3.2, regarding design control, no documentation was available for the analyses described in GE topical reports NEDE 23785-1-P, Vol. II, and NEDE-24984. These topical reports were submitted to the Office of Nuclear Reactor Regulation for review.

This nonconformance concerns the need for DRF supporting topical re-ports that describe the analytical methods used in safety-related computer codes. In particular, no DRF's are available for two topical s reports related to computer codes: NEDE-23785-1-P (SAFER 02 Code) and NEDE-24984 (0DYN04 Code). GE has stated that these topicals are not design documents; however, since these codes are used in the design , process, documentation similar to that provided for design application l analyses should be provided. This item will be the subject of a future inspection at GE. 84 1

ORGANIZAT10N: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS SAN JOSE, CALIFORNIA _ REPORT INSPECTION N0.: 99900403/84=04 RESULTS: PAGE 5 of 16

4. (Closed) Nanconformance (84-02): Contrary to GE Topical Report NED0-11209, Section 3.12, concerning design change control, error reports l' affecting the ODYN computer code were not fnrmally distributed to all user groups. A responsible engineer had used ODYN for a safety-related calculation, but was not notified of an error that was discovered after completion of a licensing analysis.

This nonconformance is related to the procedure for transmitting com-puter code errors to all responsible engineers affected by the error. The particular code error involved in this nonconformance affected the use of the ODYNM04C Code. GE has stated that a responsible engineer was mistaken when he indicated to the NRC Inspector that he had applied the 0YDNM04C Ccde without knowledge of an error that existed in the code. A review of the DRF's associated with these analyses showed that the 0DYN02 Code had actually been used. In addition, it was stated that the ODYN04 Code was never used in this Component. The Manager of Core and Fuel Technology has issued a letter to NEB 0 managers emphasizing the importance of communicating error reports. However, the issue concerning the mechanism of error report trans-mittal is not resolved (See Section B.1 above).

5. (Closed) Nonconformance (84-02): Contrary to Quality Control In-struction 7.2.17, Revision 12, Paragraph 3.4.1, regarding information to be included in an audit report, the GE auditor did not include the required evaluation statement regarding the effectiveness of the quality assurance program elements which were audited in the Brown Boveri, Inc. quality assurance audit report dated 2/3/84.

The Inspector verified that the GE letter of 9/7/84 (referenced in GE audit response dated 9/24/84) was on file and had been sent to GE auditors. Two audits had been completed and filed between 9/7/84 and the time of inspection, and both audit files were reviewed by the inspector. One audit file contained a conclusion statement that con-sidered Brown Boveri, Inc. an acceptable supplier; the other contained a conclusion statement indicating that the QA program was being im-plemented in a satisfactory manner. Both phrases indicate that the supplier is satisfactorily implementing a QA program. This item is considered closed.

6. (Closed) Nonconformance (84-01): Contrary to Section 9 of GE Topical Report NED0-11209-04A, Revision 4, GE did not assure that pipe bending done at a vendor's facility was accomplished by qualified personnel, nor did GE assure that the pipe bending was accomplished using a quali-fied procedure for the pipe bend rate, heating, and annealing.

l 85

ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS SAN JOSE. CALTFORNIA REPORT INSPECTION NO.: 99900403/84-04 RESULTS: PAGE 6 of 16 GE now verifies by QA surveillance the following:

a. The vendor's procedure's are being implemented by which operators are being qualified by on-the-job training and records are being kept to confirm the qualifications.

b. The pipe bending process is now controlled as a special process.

c. The vendor's present QA program is acceptable.

Trip reports were presented to the NRC inspector verifying the afore-mentioned. This item is considered closed.

7. (Closed) Unresolved Item (84-01): GE's remedial actions concerning crack / indications in replacement recirculation piping shipped to the Hatch nuclear power plant were reviewed during the 99900403/84-01 inspection. GE reported that fourteen 12" risers were penetrant tested at Hatch by GE personnel after receipt and were determined to have indications. These risers had been tested using a die penetrant ex-amination and were passed at a GE subcontractor's facility.

GE's actions concerning the replacement recirculation piping were reviewed during this inspection and during GE's inspection of the subcontractor. There were no adverse findings and this item is con-sidered closed.

8. (Closed) Unresolved Item (84-01): Representative samples of preloaded (stiff) pipe clamp applications were selected for analysis by an NRC consultant to examine their effects on piping. The stresses induced in the pipe by the clamps were calculated. Those stresses included thermal, preload, and dynamic stresses in areas in the pipe under or

near the clamps. The object of the analysis was to determine if the total stresses are within ASME code allowables. Based upon this analysis the piping systems reviewed satisfy the ASME Code. This item is considered closed.

! E. OTHER FINDINGS OR COMMENTS

1. Thermohydraulic Computer Codes During this inspection, the development and verification of three GE thermohydraulic computer codes (REDYO6, PANAC06, and CRNC04) were reviewed. In addition, the Potentially Reportable Condition (PRC) file was surveyed for items pertaining to thermohydraulic code ap-plications. Throughout the inspection, GE Quality Assurance Topical Report NED0-11209 and the " Boiling Water Reactor Engineering Operating Procedures" (EOP) were reviewed and utilized. The findings and obser-vations are summarized in the following sections.

i 86 l

ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS SAN JOSE, CALIFORNIA REPORT INSPECTION N0.: 99900403/84-04 RESULTS: PAGE 7 of 16

a. REDY Computer Code: The REDY Code is used to simulate the transient behavior of a BWR system during various operational and accident conditions. During this inspection, the version of the code designated as REDYO6 was reviewed and the comments are de-scribed below.

i. The responsible engineer for REDYO6 stated that the code is used by approximately 20 engineers most of whom work in the same Component (Plant Transient Performance Engineering).

REDYO6 is also used in anticipated transients without scram (ATWS) analyses for new designs but it is not used in any reload applications. ii. REDYO6 is a consolidation of features included in several previous versions of the code. Some of the more significant models implemented in REDYO6 include: an eight node steam-line representation, an extended downcomer bulkwater model, a safety / relief valve model obtained from the SAFE Code, and a boron reactivity model. iii. It was stated that REDYO6 has been classified as a Level 2 code since 1980 and no errors have been reported. iv. The review of the DRF for REDYO6 (DRF N0. A00-593) indicated that the code verification relied on comparisons with calcu-lations obtained from other codes. Six calculations were performed with models representing operating reactors: four were compared with ODYN02 results, one compared with REDY05, and one compared with REDY03. In each case, the outputs were compared and the major differences in the results were explained.

v. The Engineering Computer Program Abstract for REDYO6 dis-cusses restrictions on the use of the code. In particular, for non-ATWS analyses, the code is restricted to simulating events that do not involve significant vessel pressure and power increases. For ATWS calculations, REDYO6 can only be used after comparison calculations are preformed with DDYN02.

The DRF supporting an application of REDYO6 in an ATWS analysis (DRF No. A13-00179) was reviewed and indicated that a comparison had been performed between ODYN02, REDYO3, REDYO6. The results indicated that REDYO6 calculated higher j (conservative) peak neutron power, heat flux, and steam dome pressure than the other two codes. l 87

ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS SAN JOSE. CALIFORNI A l REPORT INSPECTION i N0.: 99900403/84-04 RESULTS: PAGE 8 of 16 vi. GE Topical Report NEDE-25518 describing an application of i REDYO6 in an ATWS analysis for Perry Units 1 and 2 was l reviewed. The DRF supporting this analysis (No. A13-00141) l provided a reasonably complete description of the calcula-tions; however, in some cases, the independent review was dated after the issuance of the report, vii. The DRF supporting the basedeck used for the ATWS calcula-tions described in NEDE-25518 was reviewed. This DRF des-cribed the model input changes implemented for the analysis and identified the responsible engineer and the independent reviewer. There was one nonconformance (See Section B.3) identified during this part of the inspection,

b. PANACEA Computer Code: PANACEA is a three-dimensional coupled nuclear thermal-hydraulic computer code used for the steady state analysis of BWR core designs. The current version approved for safety-related calculations is designated as PANAC06/PANAC06C.

The code calculations are also used as input for safety-related analyses performed with other codes (e.g., ODYN). The results of the review of the verification and application of PANACEA are described below.

i. The responsible engineer for PANACEA stated that approxi-mately 100 people from four different Components use the code on either of two computer systems. The code is checked for consistency on both computers by comparing the results of a number of test cases. The large usage of the code was confirmed by reviewing a computer usage sumary for a pre-vious month that indicated that PANAC06 was accessed more than 1700 times.

ii. PANAC06 has been created from a previous version of the code and includes some modelling improvements, editing additions, and minor error corrections. A review of the DRF indicated that more than 150 test cases are run and compared with the results produced by a previous code version. Some of these test cases include comparison with actual plant data, j iii. It was stated that PANAC06 includes a new methodology iden-l tified as the "New Physics" models; this will be the subject of a separate verification program and is not currently used for safety-related calculations. 88

ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS SAN JOSE, CAITFORNIA REPORT INSPECTION N0.: 99900403/84-04 RESULTS: PAGE 9 of 16 i iv. The DRF's for two applications of PANAC06 (No. L11-00226 and No. L12-00608) were reviewed. Each contained a reasonably complete description of the analyses performed and were independently verified.

v. GE Nuclear Engineering Technical Procedure 384HA872, " Pre-paration of 3D Code For Initial Core Analysis" was reviewed.

This procedure providas methods and examples to be used in the prepare.tions of data for entry into the GEBALS PANACEA libraries. In addition, approved sources of the input parameters are identified. vi. The DRF supporting the current PANACEA data base was revi s.wed. The input data for all except one of the speci-fied plants was identified. It was stated that the remaining data was identical to several of the existing data bases. However, an independent verification was not indicated for all of the data base entries. There was one nonconformance (B.4) identified during this part of the inspection.

c. CRNC Computer Code: The CRNC program is used as interface be-tween the 3-D BWR core simulator code (PANACEA) and the 1-D core transientcode(0DYN). The code collapses various neutronic parameters to one dimension in a format appropriate for use in the 10 codes. CRNC-04 is the current version approved for Level 2 use. Comrents on the development and verification of the code are described below.

i. The DRF for CRNC-04 (N0 A00-01619) was reviewed. This pro-vided a complete description of the methodology implemented in the code and the verification program. However, the Software Requirements Specification (SRS) stated that CRNC-04 would be tested by comparing the code results with a hand calculation for a small core problem. The results of this test case were not available in the DRF and it was stated that the calculation was not performed.

89

ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS SAN JOSE. CALIFORNIA REPORT INSPECTION N0.: 99900403/84-04 RESULTS: PAGE 10 of 16 ' ii. The verification program for CRNC-04 was reviewed and it was determined that only comparison calculations with a previous version of the code were performed. It was stated that this provided verification of only the changes implemented in the new version and not the complete methodology implemented in the various code versions. However, a detailed review of the DRF's for earlier code versions showed that the ori-ginal code (CRNC-01) was verified against hand calculations for a simplified core test case. There was one nonconformance (B.2) identified during this part of the inspection.

d. RVR1Z02 Computer Code Errors: The Potentially Reportable Condi-tion Files (PRC) were reviewed to determine if only items related to computer code errors had been entered and evaluated. A recent PRC, designated as 84-30, was identified. This PRC deals with errors that were discovered in the RVRIZ02 computer code that is used by GE in containment system and piping design calculations.

RVRIZ02 has been approved for Level 2 applications (safety-related calculations) since 1977. The application identified in this PRC relates to the use of the code for sizing of safety relief valve discharge line vacuum breakers and the determination of timing requirements between safety relief valve actuations. Analyses, identified in the PRC file as being performed by GE, have been for the Mark I containment design. However, GE has released the code to utilities and architect engineers who may have used the code for other applications. It was stated in the PRC file that Mark I utilities have been notified of the errors; however, Mark II and III utilities have not yet been identified and notified. In addition, since the code had been designated Level 2 since 1977, assurance should be provided that it has not been used in other safety-related applications. The GE review of RVRIZ02 resulted in the correction of three code errors and the implementation of two modelling modifications. The reanalysis of the Mark I containment calculations with the modified code indicated that the previous results were conserva-l tive assuming that the GE application guidelines regarding the 1 90

ORGANIZATION: GENERAL ELECTPIC COMPANY  ; NUCLEAR ENERGY BUSINESS OPERATIONS SAN JOSE, CALIFORNIA REPORT INSPECTION NO.: 99900403/84-04 RESULTS: PAGE 11 of 16 method of analysis were followed. In addition, the revised code has performed some verification calculations that have been re-viewed by a design review team. However, the review team noted that the revised version of the code, designated as RZFIX, "is not qualified for design application but only for general design assessment." Thus, a review of the code applications should be performed to insure that it has not been applied beyond its current level of qualification. One Unresloved Item (See Sec-tion C.1 above) was identified during this part of the inspection.

2. Structural / Dynamic / Heat Transfer Computer Codes

a. SAP 4G07 Computer Code: The General Electric (GE) SAP 4G07 com-puter code is a large, general purpose, finite element code which is in wide use at GE. SAP 4G07 is basically the SAP 4 code written by the University of California, with options added by GE. GE had accepted responsibility for verification of the code.

It was noted that SAP 4G07 is a Level 2 program, meaning it is among the least restrictive programs used for design applications. A program can only become Level 2 through the design review pro-cess. Following an application to raise a code to Level 2 status, a presentation of the code capabilities is made to a design review team, which has the responsibility to accept or reject the appli-cation. The presentation includes verification problems demon-strating that the code gives results comparable to those from ) either experimental data or alternate solution techniques, as per E0P-40-3.00. The verification problems for SAP 4G07 were reviewed during this inspection. The verification status of SAP 4G07 was found in Table 1-5 of the SAP 4G07 user's manual. The NRC Inspector determined that all elements had been verified for static analysis and mode and ! frequency analysis. Most of the verification problems demonstrating the previous dynamic analysis techniques (i.e., l direct integration, modal superposition, and response spectrum analysis) involve the use.of the pipe element. Two options of the beam element (fixed end forces and shear deformation analysis) and one option of the pipe element (the ASME code analysis) had no verification provided. l l 91

ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS SAN JOSE. CALIFORNIA REPORT INSPECTION NO.: 99900403/84-04 RESULTS: ' AGE 12 of 16 Various combinations of element types and analysis types were included in the table, and their verification was implied by the verification of the separate components. However, no veri-fication problems were provided for the combinations. It was further noted that other element and analysis type combinations l were designated as nonverified and were not to be used. Further- ! more, each element contained options, such as type of loading l and special features such as the released degrees of freedom, ! which also required verification. However, one nodal point option (slaved degrees of freedom) and one option of the beam element (released degrees of freedom) had verification for the latest version of SAP 4G07 only, although earlier versions of the code were still available for use in safety-related designs. Two nonconformances (See Section B.5 & B.6 above) were identified during the inspection of the SAP 4G07 code.

b. ANSYS Computer Code: ANSYS is a large structural / heat transfer com-puter code that has been developed commercially by Swanson Analysis Systems Inc. and licensed by GE. ANSYS version 2, update 180 is the production version on line at GE for safety-related design. GE is currently in the process of putting a later version of ANSYS (4.lb) on line. At this time version (4.lb) is not accessible for produc-tion use.

The NRC inspector reviewed past errors and the procedures used in handling these errors. The responsible engineer for ANSYS as well as ANSYS users were interviewed. GE stated that they have never received any error notices from Swanson Inc., concerning Version 2, update 180. GE could produce no documentation to show that they have contacted Swanson Inc. to search out errors on ANSYS version 2, update 180. GE individually verifies each design application in which ANSYS has been used. The NRC inspector reviewed 5 recent DRF's which use ANSYS 03 in safety-related designs. ANSYS 03 is the GE classification for ANSYS ver-sion 2, update 180. The DRF's included verification by alternate calculations. Also reviewed were DRF 175-0013-1, ANSYS program material. This included a list of capabilities for which the veri-fication program applies as well as exclusions to certain capabilities which could yield potential errors. No Nonconformances or Violations were identified in this part of the inspection. l ! 92 1

ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS l SAN JOSE, CALIFORNIA l REPORT INSPECTION N0.: 99900403/84-04 RESULTS: PAGE 13 of 16 i

3. Engineerina Computer Program (ECP) Error Report Handling The NRC inspector performed an overall review of GE procedures for structural engineering computer program error report handling.

l E0P 40-3.00, Engineering Computer Programs, delineates procedures for the process of reporting computer code errors. ECP technical usage problems, including potential errors are documented and reported to the responsible engineer. The responsible engineer documents, for inclusion on the DRF, all identified errors in approved ECPs used for safety-related design applications. Required documentation includes a description of the error and the corrective action to be taken. This is submitted within 30 days after identification of the error, for control com-ponent approval. The control component has overall management responsibility of all ECPs. However, the procedure does not assign responsibility for identifying design documents associated with the application of the computer code or for coordinating code changes with other responsible engineers whose design documents may be affected. Further, there is no requirement in the procedure that the control component interface with responsible engineers affected by the computer code error, and assess the effects of computer code errors on designs, past and present. During this part of the inspection, the inspector determined that computer code users are reporting potential computer code errors

verbally to the responsible engineer without the required documenta-i tion.

Two Nonconformances (See Section B.1 and B.7) were identified in the part of the Inspection.

4. Review of the Potentially Reportable Conditions (PRC) Files

Both open and closed files were reviewed and comments are noted below

a. PRC-84-03: In January 1984, several Type CR2840 three-position keylock switches were found to exhibit a tendency for all of the required contacts to not close when an attempt was made to re-move the key while the switch was in the process of being actuated. Further investigation showed that only those relays with P022 as the last four digits had failed. The accumulated i tolerances in some of these switches would allow the action as described above.

l 93 l

t ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS SAN JOSE, CALIFORNIA REPORT INSPECTION N0.: 99900403/84-04 RESULTS: PAGE 14 of 16 i l Panel locations for the problem switches have been listed and GE l is continuing its review. During a future inspection this item j may be reviewed further.

b. PRC-84-45: During environmental testing, a GE CR2940 switch failed to operate af ter a radiation exposure of 37 megarads.

(The switch was approved for radiation exposures up to 20 megarads.) The exposure had hardened the cam material of the switch to the point that it fractured. Although the switch would fail safe, the possibikity of the leads shorting and thus degrading the safety IE bus suggested further investiga-tion was warranted. After further study of the problem it was concluded that failure of the switch under high radiation con-ditions would not affect the termination wire on the switch. Investigation of secondary problems due to failure of this switch is continuing at GE. During a future inspection this item may be reviewed further.

5. HPCS Diesel Generator and Motor Control Centers During a GE QA Systems audit at Powell Electrical Manufacturing Company October 21 and 22,1981, it was noted that Powell could not produce design and other documentation required in GE specification 21A9301BA Rev. O, Paragraph 4.7 which called for this equipment to be qualified in accordance with the requirements of IEEE 323. Because, in January 1984 this issue had still not been resolved to the satisfaction of GE

^ QA, a PRC file was opened. In order to resolve this matter, GE Safety and Licensing requested, from the cognizant engineer, both description of environments (radiation, pressure, temperature, humidity, spray, etc., as applicable) in which the HPCS Diesel Generator and related Motor Control Center must operate in order to perform their intended safety functions and a demonstration that the Diesel Generator has or has not been qualified to these environments. The required docu-mentation has not been assembled. This item will be covered-during a future inspection.

6. Implementation of 10 CFR Part 21 Of 35 submittals of Potentially Reportable Conditions in 1983 three were still being evaluated in October 1984. Six were reported to NRC, and the average time for a report to be filed was 9 1/2 months.

After a decision that a condition was reportable, GE complied with the requirement that NRC be notified within 48 hours. As a further example, of 54 Potentially Reportable Conditions identified in the 94 I l

ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS SAN JOSE, CALIFORNIA REPORT INSPECTION NO.: 99900403/84-04 RESULTS: PAGE 15 of 16 first 6 months of 1984 more than half were still being worked on in mid October. Likewise 15 of the 38 submitted in the first 6 months of 1984 were still being evaluated at the time of this inspection. No independent judgments of what constitutes a rea-sonable indication that a substantial safety hazard could have been created were made during this inspection. Potentially Peportable Condition files reviewed dealt with delivery of un-qualified components. This appeared to result from continuing to purchase those components from vendors which no longer operated QA or other programs necessary for qualification, or from use of sub-stitute commercial vendors without taking the necessary additional actions to qualify the components before they are shipped to the utility. This item will be further reviewed during a future inspection.

7. Potential Violation of the 1% Plastic Strain Limit on the Linear Heat Generation Rate (LHGR) for the Lead Test Fuel Assemblies An LER initiated by the Tennessee Valley Authority (TVA) pertain-ing to the Browns Ferry facility indicated that a reload calcu-lation yielded results in which the LHGR exceeded the 1% plastic strain limit.

GE received a request from TVA relative to an interpretation of the LNGR results, GE had provided earlier, in which the Rod Withdrawal Error (RWE) LHGR exceeded the 1% plastic strain limit. GE clarified what appeared to be a technical specification violation es outlined below.

a. The analysis for 1% plastic strain limit was not performed speci-fically for the lead test assembly but was conservatively included by subtracting 2.2% from the calculated limit,

b. The rod withdrawal error LHGR was calculated for the con-figuration which inherently includes a 2.2% power spiking penalty which enhances the conservativeness.

c. When reporting the RWE value of the LHGR and the 1% limit the RWE was further increased by 2.2% due to a lack of proper definition as to the use of the 2.2% limit. As a result, excessive conservatism on the RWE value and the 1% limit yielded an apparent violation in which the LHGR for the RWE condition exceeded the 1% limit.

{ 95 l

ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS SAN JOSE. CALIFORNIA REPORT INSPECTION NO.: 99900403/84-04 RESULTS: ' AGE 16 of 16

d. GE recalculated the 1% plastic strain limit for the lead test assemblies specifically and properly applied the 2.2% penalty to.the RWE value of the LHGR. The results were within the technical specification envelope. A violation did not occur and as a result of the activity, GE has started an effort to revise the reporting and documenting function for the 1% strain limit and the LHGR based on the RWE.

The above mentioned calculations were reviewed by the NRC inspector. No violations or nonconformances were identified during this part of the inspection.

8. Qualification of ASCO Scram Valves Used at Susquehanna I During a control rod trip actuation test at Susquehanna I on October 7,1984, four control rods failed to insert. The cause of the failure was determined to be the adherence of a valve disc to its seat in the scram pilot solenoid valve. During this inspection the inspectors obtained information on the use of this type of valve in other plants and on past experience and other factors which led to changes in disc material. Interviews with GE staff disclosed that two types of disc material had been used in ASCO scram pilot solenoid valves. The first type of material is used in valves located outside containment such as those et Susquehanna. The second type of material is used in valves located inside containment. The second type of disc material is an improved material included in all valve rebuild kits which meets current design criteria for use in harsh environments and is suitable for use inside or outside containment. The second type disc material was developed before GE had operating experience with the ASCO valves which used the first type of disc material. As a result of the improved material being included in the rebuild kits for ASCO valves, each time a valve is rebuilt the second type of disc material is installed. The qualification data for valves with I both types of disc material were reviewed. The temperature and radiation values used in the tests were found to exceed those listed in the Susquehanna I FSAR.

During a telephone conversation with GE staff after the inspection, it was disclosed that ester oil contamination in the instrument air system had been found at Susquehanna I. When a scram pilot solenoid valve was tested with similarly contaminated oil the disc adhered to the seat. GE had earlier advised all utilities of the need to keep the instrument air system free from oil and other contaminants. No violations or nonconformances were identified in this part of the inspection. 96

I ORGANIZATION: GULFALLOY, INC. HOUSTON, TEXAS REPORT INSPECTION INSPECTION l NO.: 99900343/85-01 DATE(S): 5/6-10/85 ON-SITE HOURS: 56 l CORRESPONDENCE ADDRESS: Gulfalloy, Inc. ATTN: Mr. P. R. Dalton President 4730 Darien, Post Office Box 52518 Houston, Texas 77052 l ORGANIZATIONAL CONTACT: Mr. G. W. Gross, Manager Quality Assurance TELEPHONE NUMBER: (713) 672-7451 PRINCIPAL PRODUCT: Pipe, fittings, and flanges NUCLEAR INDUSTRY ACTIVITY: Approximately 17 percent of the 1984 sales. ASSIGNED INSPECTOR: _ M ["4-8C

0. T.~Conway, Rea ive Inspection Section (RIS) Date V

OTHER INSPECTOR: E. Trottier, RIS I f APPROVED BY: . 28 C E.W.derschoff/ fief,RIS Date INSPECTION BASES AND SCOPE: A. BASES: 10 CFR Part 50, Appendix 50 and 10 CFR Part 21. B. SCOPE: The inspection was made as a result of the receipt of an allegation pertaining to purchasing commercial grade material and supplying it to a licensee as nuclear grade.

+

c i PLANT SITE APPLICABILITY: Allegation - St. Lucie-1 (50-335) and Turkey Point

; 3 & 4 (50-250/251) 97

ORGANIZATION: GULFALLOY, INC. HOUSTON, TEXAS REPORT INSPECTION NO.: 99900343/85-01 RESULTS: PAGE 2 of 12 A. Nonconformances:

1. Contrary to Criterion V of Appendix B to 10 CFR Part 50, Sections )

3.1 and 5.1 of the " Quality Assurance Program Materials Indenti- I fication and Verification Manual" (QAM), and subsection NCA-3861(b) of Section III of the ASME Code, a review of documentation packages for 20 nuclear orders from Florida Power & Light (FPL) and 2 nuclear orders from Texas Utilities Generating Company (TUGCo) indicated that material on 11 FPL orders and one TUGCo order was purchased by Gulfalloy from the following suppliers / manufacturers who were either surveyed / audited after the purchase order (P0) was placed or who were never surveyed / audited by Gulfalloy: Customer Order (date) Supplier / Manufacturer FP&L Delivery & Work Authorization 39541 (December 6, 1982) Benko Fitting & Flange (1) Cardinal Industrial Products (2) Tube-Line (1) 38624 (March 19, 1982) Ametek (1) Tube-Line (1) 39666 (January 3, 1983) Standard Fittings (1) Hackey (1) National Flange & Fittings Benko Fittings & Flange (1) 38625 (March 19, 1982) Tube-Line (1) 24503C (November 10, 1981) Gulf Supply (3) i 26398W (January 25, 1983) Cardinal Industries l Products (4) 286608 (September 3, 1981) Gulf Supply (3) 17461A (April 1, 1982) All States Steel (1) Gulf Supply (3) 17328 (February 5, 1982) All States Steel (1) 17387 (March 3, 1982) All States Steel (1) 98

ORGANIZATION: GULFALLOY, INC. HOUSTON, TEXAS REPORT INSPECTION NO.: 99900343/85-01 RESULTS: PAGE 3 of 12 TUGCo P0 CPF 1000 - S Copperweld Regal Tubing (4) (1) Never surveyed / audited by Gulfalloy (2) Surveyed / audited 6 months after P0 date (3) Last surveyed / audited in February 1978 (4) Sur~veyed/ audited 1 month af ter P0 date

2. Contrary to Criterion V of Appendix 8 to 10 CFR Part 50 and Section 3.3 of the QAM, there was no documented evidence of a Quality System Certificate (Material) for Nippon Kokan, Kawasaki Steel Corporation and Nippon Benkan who are ASME certificate holders. In 1982, these three manufacturers supplied pipe and fittings to Gulfalloy through Benko Fittings & Flange, a sister company of Gulfalloy. Subsequently, Gulfalloy sent a number of the items to the Turkey Point nuclear facility for orders on PWA 39541 dated December 6,1982 and DWA 39666 dated January 3, 1983.

3. Contrary to Criterion V of Appendix B to 10 CFR Part 50, Section 2.5 of the QAM and Subsection NCA-3967.4(c) of the ASME Code, a review of documentation packages for 20 "QL-1" P0s from FP&L revealed that Gulfalloy certified on Material Test Reports (MTR) i.

' or Certificate of Conformances (CC) that the requirements of SQADs 1001 and 1002 applied to the purchased items but the manu-facturer's Certified Material Test Report (CMTR) did not document this fact. The affected nuclear orders with improper certifi-tions are as follows: DWA/PO No. MTR/CC Date 34648 October 26, 1981 39641 February 3 and 17,1983 January 24, 1983 l 39541 January 26, 27, and 28, 1983 February 7 and 17, 1983 38617 A April 13 and 20, 1982 38624 B April 12 and 20, 1982 May 4, 1982 l l 93

i ORGANIZATION: GULFALLOY, INC. HOUSTON, TEXAS REPORT INSPECTION NO.: 99900343/85-01 RESULTS: PAGE 4 of 12 DWA/PO No. (con't) MTR/CC Date i 39666 March 3, 1983 May 4, 1983 January 31, 1983 38625 May 13, 1982 April 22, 1982 21834S August 11, 1982 219355 August 11, 1982 276395 January 5, 1984 17461 A April 20 and 23, 1982 17328 February 10, 1982 17387 April 20, 1982

4. Contrary to Criterion V of Appendix B to 10 CFR Part 50, Section 5.2.2 of the QAM, and Subsection NCA-3867.4(c) of the ASME Code, it was noted that Gulfalloy P0 22-27-22672 dated July 14, 1984 to Copperweld Regal Tubing (CRT) did not reference the code require-ments identified on TUGCo's P0; and Gulfalloy's MTR dated September 23, 1981 certified that the material met the requirements of "ASME Section III, 1974 Edition thru 1974 Winter Addenda, NB-2000 for Class 1 material", but CRT's CMTR dated September 3, 1981 did not document that the material met the code requirements.

l B. UNRESOLVED ITEMS: None C. STATUS OF PREVIOUS INSPECTION FINDINGS:

1. (Closed) Violation (84-01):

Current copies of 10 CFR Part 21 and Section 206 of the Energy Reorganization Act of 1974 were not posted in a conspicuous place. Appropriate procedures to evaluate deviations or inform the licensee or purchaser of the deviation did not exist. 100

ORGANIZATION: GULFALLOY, INC. HOUSTON, TEXAS REPORT INSPECTION NO.: 99900343/85-01 RESULTS: PAGE 5 of 12 Current copies of 10 CFR Part 21 and Section 206 of the Energy Reorganization Act of 1974 were found posted on the main office l bulletin board. Quality Assurance Procedure N-10, " Reporting i of Defects and Noncompliance", Revision 0 dated October 13, 1984 ( was also posted. A requirement to check for continued posting of a current copy of 10 CFR 21 and Section 206 has been added as l Item F, Section 2 of QA Procedure N-4, " Internal Audit."

2. (Closed) Violation 84-01:

P0s placed with Gulfalloy specified 10 CFR Part 21 as an applicable requirement for ASME Code Section III material. However, in ordering the material from manufacturers, Gulfalloy did not " pass on" 10 CFR Part 21 as a requirement. The inspector reviewed a random sample of 10 P0s for material to be supplied by Gulfalloy to nuclear power plants. The P0s were placed in late 1983 through 1984. In each case, Gulfalloy's P0 to their supplier stated that 10 CFR 21 applied to the material being ordered. Section 5.8 of Gulfalloy's internal audit procedure also has been revised to assure that 10 CFR 21 is imposed on Gulfalloy's suppliers. This is verified by a random sampling of applicable P0s. In addition, Gulfalloy notified all their suppliers (57) by mail on October 10, 1984, that 10 CFR 21 applies to orders that require implementation of their approved Quality Assurance Program. This letter required a positive response that the company was and continues to be in compliance with 10 CFR Part 21 for such nuclear-related orders. As of the date of the inspection, 40 replies had been received.

3. (Closed) Nonconformance (84-01):

A receipt inspection was not performed for 6, 6 in. SA-234 tees. The customer order file was reviewed to verify that a shipping inspection was performed prior to the parts leaving the Gulfalloy facility. Gulfalloy also committed to expand their normal sample size when reviewing PO files during the following scheduled internal audit (January, 1985); ensure the QA Manager reviews all nuclear-related P0s prior to close-out; and conduct training on the appli-cable section of the Gulfalloy QA Manual. The inspector verified that the number of P0 files reviewed during the internal audit conducted in January - February 1985, was 101

l ! ORGANIZATION: GULFALLOY, INC. HOUSTON, TEXAS REPORT INSPECTION NO.: 99900343/85-01 RESULTS: PAGE 6 of 12 doubled over the number reviewed in previous audits. The inspector also verified that a stamp with QA Manager's signature block is now affixed to all nuclear-related P0s, and that training in the requirements of QA Manual Section 5, Procurement, was conducted on July 26, 1984.

4. (Closed) Nonconformance (84-01):

Three P0s for nuclear-related material did not require a QA program certification statement (supplier's QA program approved by ASME or Gulfalloy). Gulfalloy reviewed the customer order file for each of these supp-liers. Two had ASME certificates in effect at the time the orders were placed; the third supplier's QA program had been previously audited and approved by Gulfalloy. Thus, although it was after the fact, each supplier was shown to be properly qualified to provide the subject material.

5. (Closed) Nonconformance (84-01):

A Nondestructive Examination (NDE) vendor performed NDE services for Gulfalloy in May 1983, without their written practice of training and performance being on file at Gulfalloy. Further, there was no evidence that Gulfalloy had approved the NDE vendor's written practice of providing NDE services. The inspector verified that Gulfalloy has since audited and approved their NDE supplier's program (Service Report Evaluation dated 9-14-84) and that their written practice (QAP 970900-75, Rev. 5 dated July 9, 1982) is presently on file at Gulfalloy.

6. (Closed) Nonconformance (84-01):

Gulfalloy obtained the services of a pipe coating supplier without surveying or auditing the supplier. Gulfalloy reconstructed the events surrounding the purchase of the pipe coating service and determined that even though the company shared yard space with Gulfalloy, it remained a subcontracted service. Gulfalloy is thus now fully aware of what constitutes a subcontracted service transaction and the proper administrative control thereof. All nuclear-related material and subcontracted services P0s are required to be reviewed and approved by the Gulfalloy Menager of Quality Assurance. 102

ORGANIZATION: GULFALLOY, INC. HOUSTON, TEXAS REPORT INSPECTION NO.- 99900343/85-01 RESULTS: PAGE 7 of 12

7. (0 pen) Nonconformance (84-01):

1 r Gulfalloy Power Sales / Purchasing personnel had not received QA I t Program indoctrination training between 1980 and 1984. l Gulfalloy committed to perform the required training in July and November,1984. The inspector noted that the training scheduled for July was accomplished on July 26, 1984. The training scheduled for November 1984 was not accomplished as of the date of this inspection.

8. (0 pen) Nonconformance (84-01):

The cause and corrective action taken to resolve deficiencies noted in audits conducted in January 1982 and January 1984 was not docu-mented. Also, reaudits were not performed in any of the deficient area.s noted in the January 1982 audit. Gulfalloy committed to document the cause of the deficiencies noted in the audits of January 1982 and January 1984. Also, Gulfalloy committed to reaudit the deficient areas. The inspector noted that the root cause of past audit deficiencies has not yet been determined. It was noted that the format of the Gulfalloy Nonconformance Report (ie, the vehicle for tracking and resolving such deficiencies) does not specifically identify "cause" as a topic of investigation and resolution. D. OTHER FINDINGS OR COMMENTS:

1. Allegation - In August 1984, the NRC Region II Office of Investi-gation was made aware of an allegation, which pertained to pur-chasing comercial grade material and supplying it to FP&L as nuclear grade.

The NRC inspector reviewed 20 procurement documentation packages for nuclear items ordered by FP&L and sent to the Turkey Point (15) and St Lucie (5) nuclear facilities. Documcntation packages consisted of written inquiries; FP&L P0s, specifications, and Special Quality Assurance Documents (SQAD); Gulfalloy P0s and Receiving Inspection Reports; manufacturer's CMTRs plus heat treat charts and nondestruc-tive examination (NDE) reports, where applicable; and Gulfalloy MTRs, CCs, and Delivery Tickets. 103 1

ORGANIZATION: GULFALLOY, INC. HOUSTON, TEXAS REPORT INSPECTION NO.: 99900343/85-01 RESULTS: PAGE 8 of 12 On March 6, 1981, FP&L issued a blanket P0 38497-39056C which authorized Gulfalloy to furnish nuclear safety related items for the Turkey Point nuclear facility. The P0 referenced SQADs 1001 and 1002. SQAD 1001 established minimum QA requirements to be satisfied by Gulfalloy, and SQAD 1002 addressed the requirements of P0s requiring compliance to 10 CFR Part 21. Specific orders were placed on Delivery and Work Authorizations (DWA) in accor-1 dance with the terms and conditions of the blanket P0. Each DWA noted that it was a " Nuclear Safety Related Order QL-1" and referenced SQAD 1001 and SQAD 1002. QL-1 orders for items for St. Lucie were placed on separate P0s which also referenced SQAD 1001 and 1002. The NRC inspector reviewed the following nuclear orders: No. DWA/PO (Date) Item Manufacturer 1 34648 (9-29-81) 3 pipe / fittings / Damascus Tubular flanges Products Taylor Forge WFI 2 39641 (1-17-83) 8 pipe / fittings Sandvik ' Custom Alloy Camco Fittings Alloy Stainless Products 1 t 3 39541 (12-6-82) 28 pipe / fittings / flanges / Quanex fasteners Nippon Kokan Kawasaki Steel Cardinal Indus-trial Products Capitol Manu-facturing Nippon Benkan WFI Tube-line 4 38617A (3-19-82) 9 pipe Quanex , U. S. Steel Sandvik 5 79935W (6-29-84) 1 - bar WFI 104 I

ORGANIZATION: GULFALLOY, INC. HOUSTON, TEXAS REPORT INSPECTION NO.: 99900343/85-01 RESULTS: PAGE 9 of 12 1 M. DWA/PO (Date) Item Manufacturer 6 386248 (3-19-82) 9 - fittings Ametex Tube-Line l Capitol Manu-facturing i 7 39666 (1-3-83) 10 pipe / Standard Fittings fittings / Hackey flanges National Flange & Fittings Nippon Benkan Nippon Kokan Quanex 8 38625 (3-19-82) 25 - fittings / WFI flanges Tube-Line Capitol Manu-facturing 9 21834S (7-26-82) 1 - caps Custom Alloy

,       10              219355 (8-2-82)           2 - fittings      Camco Fittings 11               22046S (8-18-82)          order was cancelled 12               24503C (11-10-81)         19    pipe /      Tube-Line fittings    WFI Tube Turns Standard Fittings 13              26112C (12-17-82)          QL-2 order for flexible hose 14              26398W (1-25-83)           2 - fasteners      Cardinal Indus-

! trial Products 15 399140 (11-3-81) 1 - fittings Camco Fittings L 16 286608 (9-3-81) 12 pipe / Sandvik fittings Camco Fittings flanges Taylor Forge ITT Grinnell Hackey 17 27639W (10-15-83) 1 - fasteners Texas Bolt I 105

                                                                                   # -# -_,w r-, - , ,

ORGANIZATION: GULFALLOY, INC. HOUSTON, TEXAS REPORT INSPECTION NO.: 99900343/85-01 RESULTS: PAGE 10 of 12 No. DWA/P0 (Date) Item Manufacturer 4 18 17461A (4-1-82) 15 - pipe / All States Steel shapes 19 17328 (2-5-82) 3 - plate All States Steel 20 17387 (3-3-82) 7 - plate /bar/ All States Steel shapes It was noted that items for purchase nos. 4 (partial), 5, 6, 9, 10, and 15 were ordered to Section III/ Class 2 of the ASME Code. Items on the remaining orders were to the following ASTM specifications: pipe (A312, 53 and 106); fittings (A182, 234, 403, and 105); flanges (A105 and 181); fasteners (A307, 193, and 194); and shapes (A36). The Gulfalloy P0s for orders to Section III/ Class 2 specified the ASME Code requirements to manufacturers. The majority of manufacturers were holders of a Quality System Certificate (Materials) or had been audited by Gulfalloy. However, P0s were placed with Tube-Line (nos. 3, 6, 8 and 12); Cardinal Industrial Products (nos. 3 and 4), Ametex (no. 6), Standard Fittings (nos. 7 and 12), Hackey (nos. 7 and 16) and National Flange

                                                                              & fittings (no. 7), and these manufacturers were never audited by Gulfalloy or were audited af ter the items were ordered (see Nonconformance A.1).             Items manufactured by Nippon Kokan (NK),

Kawasaki Steel (KS), and Nippon Benkan (NB) on order no. 3; Hackey, NB, and NK on order no. 7; ITT Grinnell and Hackey on order no. 16 were ordered thru Gulf Supply and/or Benko Fittings and Flange, two sister companies of Gulfalloy. Gulf Supply and Benko Fittings and Flange have never been audited or surveyed by Gulfalloy. NK, KS, and NB are ASME certificate holders, but Gulfalloy did not have a copy of a Quality System Certificate (Materials) for each company (see Nonconformance A.2). Although FP&L imposed CFR Part 21 requirements en all the QL-1 l orders, Gulfalloy failed to specify Part 21 requirements on their ! P0s to manufacturers for purchase nos. 1, 2, 3, 4, 6, 7, 8, 9, 12, 15, 16, 17, 18 and 19. Failure to pass on Part 21 requirements to vendors was identified during a previous inspection conducted on July 16-20, 1984 (see Violation A.2, report no. 99900343/84-01). With the exception of the CMTRs from Capitol Manufacturing (order ro. 3), WFI (no. 5), Custom Alloy (no. 9), Tube Turns (no. 12), and Camco Fittings (no.15), CMTRs from the other manufacturers on all the orders did not certify that 10 CFR Part 21 was required 106 l

ORGANIZATION: GULFALLOY, INC. HOUSTON, TEXAS l REPORT INSPECTION  ; NO - 99900343/85-01 RESULTS: .PAGE 11 of 12 during the fabrication of the items. However, for all orders Gulfalloy certified on applicable MTRs or CCs that the items were furnished under the requirements of SQAD 1002 (ie, Part 21) (see Nonconformance A.3). The improper certification of MTRS and CCs by Gulfalloy should have been detected by FP&L since Gulfalloy trans-mitted a copy of all the CMTRs to the licensee. Based on the results of examining the documents relating to 20 nuclear orders from FP&L from September 1981, through January 1983, it was noted that the items on order 1, 2, 3, 4 (partial), 7, 8, 12, 14, 16, 17, 18, 19, and 20 were purchased as commercial grade and supplied to FP&L as nuclear grade thus substantiating the allega-tion. There was no documented evidence that Gulfalloy upgraded any material in accordance with the requirements of Subsection NCA-3867.4(e) of Section III of the ASME Code.

2. Control of Purchase Material - The inspector reviewed approximately 52 procurement documentation packages pertainin to nuclear orders from 1975 thru 1985. The customers were TUGCo 4), Baltimore Gas

        & Electric (3), Stone & Webster (3), Ebasco (8), Bechtel (31) and Pacific Gas & Electric (3). The review was undartaken to assure that applicable regulatory, technical, and QA program requirements are included or referenced in procurement documents, and that nuclear was purchased from approved vendors.

TUGC0 - The 4 orders were from July 1981 through March 1982. With the exception of order CPF 1000-S (see Nonconformance A.4), infor-mation certified by Gulfalloy on their MTR or CC agreed with the technical data contained in the manufacturer's CMTR. However, it was noted that Gulfalloy's Pos did not specify the requirements of 10 CFR Part 21. Baltimore Gas & Electric (BGE) - The 3 orders from BGE from November 1979 thru December 1980 were for Section III/ Class 1 tubing. Gulfalloy P0s were to Teledyne Columbia (TC), who had been audited by Gulfalloy, for direct shipment to Calvert Cliffs nuclear facility. Gulfalloy MTRs agreed with TC CMTRS. Stone & Webster - All the items for the three purchases ir. May and July 1981 for Riverbend were ordered by Gulfalloy from certificate holders or vendors approved by Gulfalloy. 107 l

ORGANIZATION: GULFALLOY, INC. HOUSTON, TEXAS REPORT INSPECTION NO.- 99900343/85-01 RESULTS: PAGE 12 of 12 Ebasco - Six orders (October 1981 thru December 1981) were for items for Waterford and 2 orders (April and May 1975) were for St. Lucie No. 1. Gulfalloy P0s for 2 of the 6 orders for Waterford referenced Part 21. All the material was ordered from approved vendors. Bechtel - Thirty-one procurement packages for Palo Verde, Hope Creek and Arkansas Nuclear received between 1980 and 1984 were reviewed. In each case the material was intended for nuclear power plant application with Bechtel acting as agent for the utility. Upon review, it was found that all Gulfalloy suppliers (e.g., Sandvik, Quanex, Capitol Manufacturing, Hawley Forge) were either certificate holders or had been audited and approved by the Gulfalloy QA Department. No irregularities were found with the material requirements of the customers P0 and the material certified by Gulfalloy. Pacific Gas & Electric - The three orders for Diablo Canyon were placed from September 1984 thru March 1985. All of Gulfalloy's P0s were in order as well as their MTRs and CCs. 108

ORGANIZATION: INRYC0 INC. BEDFORD PAPK, ILLIN0IS REPORT INSPECTION INSPECTION i NO.: 99900731/85-01 DATE(S): 3/5-8/85 ON-SITE HOURS: 68 CORRESPONDENCE ADDRESS: INRYC0 Inc. l ATTN: Mr. D. W. Waitkus Supervisor, Quality Assurance 72005 Narragensett Avenue Bedford Park, Illinois 60638 1 ORGANIZATIONAL CONTACT: Mr. D. W. Waitkus, Supervisor, Quality Assurance TELEPHONE NUMBER: (312) 594-7300 PRINCIPAL PRODUCT: Post tensioning systems for reactor containments. NUCLEAR INDUSTRY ACTIVITY: Loss than 10%. Activities are limited to providing services for the surveillance of tendons. Inryco cu:Yently has an active contract for tendon surveillance at Farley, including rep.lacement of damaged anchor heads. ASSIGNED INSPECTOR: b K. Naidu, Reactive Inspection Section (RIS)

                                                                              ///21/d7 Cate OTHER INSPECTOR (S):    T. Burns, Brookhaven National Laboratory (BNL)

APPROVED BY: s[/4!Ps-E. W. Merschoff Chief, RIS, Vendor Program Branch Date INSPECTION BASES AND SCOPE: A. BASES: 10 CFR Part 21 and 10 CFR 50 Appendix B. B. SCOPE: This inspection was made as a result of failed tendons at the J. M. Farley Nuclear Power Station in Alabama, to verify the implementation of INRYCO's QA program relative to procurement document control (Criterion IV), control of purchased material,. equipment and services (Criterion VII), quality assurance records (Criterion XVII), and audits (Criterion XVIII). The inspection included visits to two of Inryco's suppliers of material or services. PLANT SITE APPLICABILITY: J.M. Farley Nuclear Station (50-364); Marble Hill Nuclear Station Unit 1(50-546); and Marble Hill Nuclear Station Unit 2 (50-547). 1 109 I

ORGANIZATION: INRYC0 INC. BEDFORD PARK, ILLIN0IS REPORT INSPECTION N0.: 99900731/85-01 RESULTS: PAGE 2 of 9 i A. INSPECTION ISSUES: During'an inspection of post-tensioning systems in January 1985 at the Unit 2 J. M. Farley Nuclear Power Station, tendon anchor head abnormalities were observed. INRYC0 supplied components for the post-tensioning systems. The objective of this inspection was to verify the implementation of INRYCO's QA program relative to the control of purchased material, equip-ment and services. B. INSPECTION FINDINGS:

1. Violations:

None.

2. Nonconformances:

Contrary to Criterion VII of 10 CFR 50 Appendix B and section 7 of the INRYC0 QA manual wnich requires the establishment of measures to assure that purchased material, equipment, and services, whether purchased directly or through contractors and subcontractors, conform to purchase documents. Specifically: (a) INRYC0 failed to assure that their purchase order to Machine Specialities for the procurement of tendon anchor heads

                   . reflected the requirements of Purchase Specification Y 2721, issued to INRYC0, by not including the relevant sections of the specification which required the vendors and subvendors to conform to 10 CFR Part 21 and 10 CFR 50 Appendix B.

(b) INRYC0's measures to assure that material supplied by Western Concrete' Structures (WCS) conform to purchase documents were inadequate in that INRYC0 did not take timely corrective action to bring WCS in compliance with the training requirements for WCS QC personnel. 1 (c) Measures established by INRYC0 to assure that the services supplied by Downey Heat Treating facilities (DOWNEY), subvendor to WCS, were inadequate in that neither audits nor surveillances were performed and dccumented which would indicate that Downey followed INRYC0's heat treatment procedure and that the furnaces conformed to MIL-Spec #6375F. 110 f 4 t

l ORGANIZATION: INRYC0 INC. BEDFORD PARK, ILLIN0IS REPORT INSPECTION NO.: 99900731/85-01 RESULTS: PAGE 3 of 9 C. OTHER FINDINGS AND COMMENTS:

1. Background Information:

The two unit J. M. Farley (Farley) Nuclear Power Station is operated by Alabama Power Corporation (APC). APC was preparing to perform the 5-year integrated leak rate test (ILRT) on Unit 2 during January 1985. While visually inspecting the tendons, one broken and one cracked field anchor head were observed on 1/25/85 and 1/30/85 respectively. One 2/26/85, APC0 reported an additional broken anchor head. Each anchor head retains 170 tensioned tendons. INRYC0 supplied the tendons, anchor heads, and other post-tensioning components for Farley Units 1 and 2. INRYC0 procured the tendon anchor heads from Western Concrete Structures (WCS). WCS fabricated the tendon anchor heads from steel manufactured by Republic Steel. The fabricated tendon anchor heads were heat treated by Downey Heat Treatment company to INRYC0's specification. The replacement tendon anchor heads at Farley were originally procured and supplied by INRYC0 for installa-tion at the Marble Hill Nuclear Power Plant. These anchor heads were not installed and were purchased as spares for the Fort St. Vrain Nuclear Power Plant. Documents reviewed indicate that these tendon anchor heads were fabricated by Machine Specialities and heat treated by FPM Heat Treating Company and supplied by INRYCO.

2. Review of Audits The NRC inspectors ascertained whether licensees or their representa-tives audited INRYC0 to evaluate the adequacy of the INRYC0 QA program and its implementation by selectively reviewir.g the following audits:

i

a. Preaward audit performed by Bechtel during 1976 for the supply l of post-tensioning components for J. M. Farley Nuclear Plant.

l b. Four audits performed by South Carolina Electric & Gas Company between 1981 and 1985. l c. Three audits performed by Commonwealth Edison Company between 1977 and 1984. j d. One audit performed by Tennessee Valley Authority during 1975.

e. One audit performed by Carolina Power and Light Company in 1984.

f. One audit performed by Baltimore Gas and Electric Company in 1984.

111

i ORGANIZAT' ION: INRYC0 INC. BEDFORD PARK, ILLIN0IS REPORT INSPECTION N0.: 99900731/85-01 RESULTS: PAGE 4 of 9 It was not readily discernable that any of the aoove audits scrutinized the INRYC0's implementation of 10 CFR 50 Appendix B Criterion IV and VII which relate to procurement document control and control of purchased material.

3. Review of INRYCO's Audits on Western Concrete Structures The inspectors reviewed the surveys, audits, and surveillances performed by INRYC0 on Western Concrete Structures (WCS) to assure that WCS had an acceptable QA program which was implemented at WCS and its subvendors. Records indicate that WCS supplied anchor heads to INRYC0 during the years 1975-1981. WCS subcontracted Downey to heat treat post tensioning components including anchor heads. INRYC0 qualified WCS in 1974 based on historical data. WCS was placed on INRYC0's approved vendors list (AVL) and continued to be on the AVL until 1981. INRYC0 made available the following audits.

a. INRYC0 audited WCS on 12/15/76 and identified the following adverse findings:

1. The INRYC0 auditor questioned the qualifications of a QC inspector. WCS stated that he was a graduate mechanical engineer with demonstrated ability. The INRYC0 auditor stated that this satisfied the immediate question but needed other documentary evidence to be presented during a subsequent audit.

2. The shop was not furnished with up-to-date INRYC0 drawings and procedures. It was determined that the quality of the components was not being affected because the later revisions only corrected typographical errors and did not influence the design.

b. INRYC0 performed a followup audit on 8/15/77 and identified the following findings:

1. Heat or furnace charts were being submitted as objective evidence that the INRYC0 procedures were being followed without evidence that the furnace thermocouples had been calibrated.

2. INRYC0 observed that while WCS used qualified procedures to accomplish special processes, the qualifications of personnel were questionable.

 , t )                                            112

ORGANIZATION: INRYC0 INC. BEDFORD PARK, ILLIN0IS REPORT INSPECTION N0.: 99900731/85-01 RESULTS: PAGE 5 of 9

3. Similarly, INRYC0 determined that while the equipment and procedures used in the special process complied with the applicable codes and I

standards, the qualifications of the personnel were questionable.

4. In the QA records section, the question "Do records include closely related data such as qualifications of personnel" was answered with an individual's name and a statement that "The inspector is now in training."

c. INRYC0 letter dated May 21, 1979, to WCS indicates the following:

Field anchor head (170 W 18) identified PC 048 was found with extensive cracking through the walls of 80% of the wire holes and were very visible to the naked eye. The anchor head was sent to an outside laboratory (Lab). The Lab was of opinion that the cracks were caused by the quenching operating during heat treatment. INRYC0 personnel observed that the inner surface of the wire holes of the shop and field anchor heads do not have the 0.500R completed. INRYC0 stated that they would complete the operation. INRYC0 personnel observed one anchor head where the holes had drifted together from top to bottom. This anchor head was scrapped. INRYC0 requested WCS to increase their inspection in the above areas to eliminate the shipment of defective or unfinished products.

d. INRYCO's telex to WCS dated 11/21/79 stated that one field head 170 W1B, INRYC0 heat code PY 136, heat #18381 supplied under a purchase order 21T 781-5 broke into two separate pieces, approximately 10 days after stressing. The failures noted j here and in c. above, though not related to the Farley tendon problem, indicate deficiencies in the control of heat treatment.

l e. INRYC0 ir.teroffice memorandum from D. Waitkus (QA Supervisor) l to G. Davis (Project Manager) dated August 18, 1980, recommended l the termination of Western Concrete Structures (WCS) as a source j of supply of anchor heads or any other services based on the following: 113

ORGANIZATION: INRYC0 INC. BEDFORD PARK, ILLIN0IS REPORT INSPECTION NO.: 99900731/85-01 RESULTS: PAGE 6 of 9

1. WCS did not maintain the requirements of their QA program including controls of their qubtier suppliers and documentation.

2. The WCS staff was reduced to a minimum and only one inspector was available to check their incoming and fabricated products. The person who acted as the Quality Control (QC) inspector spent less than 10% of the time on QC activities. Review and verification of paper work including heat treatment records was inadequate resulting in incomplete documentation packages.

3. WCS had various problems with two numerical controlled drills, which resulted in incorrect holes being drilled,

f. The NRC inspectors informed INRYC0 of the information furnished in subparagraphs a and b, INRYC0 identified an inadequacy in WCS training program in 1976 and this inadequacy continued in 1977 without being corrected; therefore, INRYCO's corrective actions to bring WCS in compliance with training requirements were not timely. INRYC0's purchase documents required WCS to implement a QA program meeting the requirements of 10 CFR 50 Appendix B.

Section VII of the INRYC0 QA program established measures to assure that vendors comply with requirements in the procurement documents. The inspectors informed INRYC0 that failure to take timely corrective action to bring WCS in compliance with procure-ment documents was in noncompliance with Criterion VII of 10 CFR 50 Appendix B and Section VII of the INRYC0 QA manual.

g. Relative to the heat treatment services provided by Downey Heat Treatment Company (DOWNEY), no objective evidence was provided to the NRC inspectors that either INRYC0 or WCS verified (through surveys by independent agencies) that the heat treat- 1 ment furnaces were in conformance with MIL SPEC 6875F. Records I

furnished to the NRC inspectors indicated that INRYC0 relied on the furnace heat treatment charts to verify compliance with their procedures without knowledge that the thermocouples in the furnaces were within calibration limits. The inspectors informed INRYC0 that inadequate measures were established to verify the Downey, the subvendor to WCS complied with the purchase order requirement, specifically, compliance to MIL SPEC 6875F and that this was another example contrary to 10 CFR 50 Appendix B Criterion VII. 114

ORGANIZATION: INRYC0 INC. BCDFORD PARK, ILLIN0IS REPORT INSPECTION NO.: 99900731/85-01 RESULTS: PAGE 7 of 9 One item of nonconformance with two examples was identified in the above area. l

4. Review of Management Audits Audits were performed by INRYC0 upper management to review the status and adequacy of the Concrete Systems Division. Audits performed by their customers and results of internal audits were used in the performance of the review. The results of the audits were reviewed by the Vice President as acknowledged by his signature on the document. This audit complies with the require-ment stated in paragraph 7 of Criterion 2 of the INRYC0 QA manual.

No noncompliances were identified in the above area.

5. Review of INRYCO's Purchase Orders The inspectors reviewed the purchase orders initiated by INRYC0 to vendors including vendors to sub-vendors relative to the procurement of post-tensioning components and services to comply with Purchase Specification Y2721 titled " Specification for Post-Tensioning Work (Sargent & Lundy, Agent for Public Service Company of Indiana Inc.,

Marble Hill Generating Stations Unit 1 and 2). Paragraph 115 of this specification states that the Quality Assurance criteria including 10 CFR Part 21 applies to both contractors and subcontractors. INRYC0 attaches their Form D-904 to their purchase order to convey requirements specified in Y2721. Form D-904 is a checklist containing numerous general and special requirements including the applicability of 10 CFR Part 21 and 10 CFR Part 50 Appendix B. The objective of this review was to ascertain whether INRYC0 conveyed the applicability of compliance to 10 CFR Part 21 and 10 CFR Part 50 Appendix B to their vendors and whether the vendors in turn forwarded these requirements to their subvendors,

a. The following purchase orders (P0) were reviewed.

(1) INRYC0 P0 45 ST 2742 dated 3/25/80 to Copper Weld Steel Company for the supply of A322 Grade 4140 low alloy steel tubing. (2) INRYC0 P0 45 ST 2749 dated 5/16/80 to Timken Company i for the supply of low alloy steel tubing. l 115 l I

l ORGANIZATION: INRYC0 INC. l BEDFORD PARK, ILLIN0IS l REPORT INSPECTION NO.: 99900731/85-01 RESULTS: PAGE 8 of 9 (3) INRYC0 P0 45 ST 2749 dated 11/14/80 to Timken Company for the supply of A519 Grade 4140 low alloy steel tubing. 1 (4) INRYC0 P0 45 ST 2757 dated 9/5/80 to Machine Specialities for the supply of various quantities of field tendon anchor heads and bushings. Fourteen separate specifica-tions/ procedures were attached which covered the complete fabrication and heat treatment of the fabricated components. (5) INRYC0 P0 41 MS 00015 dated 3/24/81 to Machine Specialities for various quantities of field tendon anchor heads and bushings. (6) INRYC0 P0 45 SY 2756 dated 4/5/80 for material for 170W1A (398 feet) and 170W15 (616 feet) to be saw cut. (7) Machine Specialities P0 9913 dated 12/3/80 to their subvendor FPM Heat Treating Company (FPMHTC) for the heat treatment of 700 bushings to the requirements of INRYC0 procedure PT 5.2.1, Revision 0. (8) Machine Specialities P0 9924 dated 3.2.81 to their subvendor FPM HTC for the heat treatment of 500 pieces 171W1B type field anchor heads to INRYC0 specification PT 5.2.1, Revision 0. (9) Machine Specialities P0 9931 dated 4/28/81 to their subvendor FPM HTC for the heat treatment of 1100 pieces of shop anchor heads to INRYC0 specification PT 5.2.1. (10) INRYC0 P0s 45 T 938-81 and 45 T 938-80 both dated 9/12/78 to Rode Welding Service Inc. to fabricate 324 pieces of type 170 W 50 B and 186 W 50 type trumplates. Eleven separate specifications / procedures were attached to the purchase order.

b. The results of the review indicated the following.

(1) For items a (2) and a (3), INRYC0 did not attach Form D-904 to the P0. The reason documented in a telecon dated 4/25/80 is that Timken Company refused to comply with 10 CFR Part 21 and 10 CFR Part 50 Appendix B requirements. 1 116

ORGANIZATION: INRYC0 INC. BEDFORD PARK, ILLIN0IS REPORT INSPECTION N0.: 99900731/85-01 RESULTS: PAGE 9 of 9 (2) For items (4), (5) and (6) in the above paragraph a, there was no objective evidence at INRYC0 that INRYC0 forwarded D-904 with the applicable requirements checked to Machine Specialities. Examination of the P0s received at Machine l Specialities indicated that Form D-904 was not received. (3) Consequently, for items (7), (8) and (9) Machine Specialities did not convey the applicability of 10 CFR Part 21 and 10 CFR 50 Appendix B to FPMHTC. (4) For item (10) there was no indication that INRYC0 forwarded D-904 to Rode Welding Services imposing the requirements of 10 CFR 50 Appendix B.

c. The inspectors informed INRYC0 that based on the above results, INRYC0 failed to assure that applicable regulatory requirements specified in Technical Specification Y2721, which are necessary to assure adequate quality, were included or referenced in the procurement documents to control the procurement of post-tension-ing equipment, material and services and that this was another example of noncompliance to Criterion IV to 10 CFR 50 Appendix B.

i One nonconformance was identified in this area. I D. PERSONS CONTACTED:

                                                                                     ^

1

1. INRYC0 Bedford Park, Illinois

             *W. Corson, Manager of Sales i             *H. F. Hendrickson, Project Manager l             *J. Heise, Manager, Contract Management l              C. E. Matykiewicz, Controller R. Kujawa, Sales Engineer

2. Machine Specialities, Berkley, Illinois E. W. Vehrs, President l

3. FPM Heat Treating Company Elk Grove Village, Illinois D. Guy, Quality Control Manager

• Denotes the individuals who attended the exit meeting at INRYC0 on 3/7/85 117 1

ORGANIZATION: ISOMEDIX (NEW JERSEY) INCORPORATED WHIPPANY, NEW JERSEY j i l REPORT INSPECTION INSPECTION NO.: 99900913/85-01 DATE(S): 4/8-11/85 ON-SITE HOURS: 44 l CORRESPONDENCE ADDRESS: Isomedix (New Jersey) Inc. ATTN: Mr. G. R. Dietz Executive Vice President and Secretary 11 Apollo Drive l Whippany, New Jersey 07981 ORGANIZATIONAL CONTACT: Mr. Steve Thompson, QA Manager TELEPHONE NUMBER: (201) 887-4700 PRINCIPAL PRODUCT: Gamma irradiation services. NUCLEAR INDUSTRY ACTIVITY: Five percent of Isomedix's business is irradiation of safety-related electrical equipment for environmental qualification testing.

ASSIGNED INSPECTOR

Le/o w 5/J20[/5' S. D. Alexander, Equip. Qual. Inspec. Sec. (EQIS) ' Dafte I

OTHER INSPECTOR (S): E. H. Richards, Sandia National Laboratory (SNL) , APPROVED BY: U. Potapovs, Cfiief, EQIS, Vendor Program Branch

                                                                            ////Date INSPECTION BASES AND SCOPE:

A. BASES: 10 CFR Part 21 and 10 CFR Part 50, Appendix B. B. SCOPE: The purpose of the inspection was to conduct a technical review and evaluation of irradiation performed on selected safety related components. PLANT SITE APPLICABILITY: Not identified. I 119

ORGANIZATION: ISOMEDIX (NEW JERSEY) INC. WHIPPANY, NEW JERSEY REPORT INSPECTION NO.- 99900913/85-01 RESULTS: PAGE 2 of 4 A. VIOLATIONS: None. B. NONCONFORMANCES:

1. Contrary to the requirements of Criterion V of Appendix B to 10 CFR Part 50 and the requirements of Paragraph 6.4 of the Isomedix Quality Assurance Manual (QAM), Receiving and Product Accountability Record (RPAR) No. 00766 and RPAR No. 00370 did not identify the test items by serial number, part number or other suitable means.

2. Contrary to the requirements of Criterion V of Appendix B to 10 CFR Part 50 and the requirements of paragraph 4.3 of Appendix C to the QAM, irradiation test reports dated 3/13/85 for customer 28325, 2/14/85 and 4/1/85 for customer 24200, 3/12/85 for customer 44330, and 3/4/85 for customer 45500 did not contain dose rate uniformity of the field information.

3. Contrary to the requirements of Criterion V of Appendix B to 10 CFR Part 50, the requirements of paragraph 6.10.3 and 6.11 of the QAM, paragraph B.2 of Appendix B to the QAM, paragraph 17.3a-f of Addendum I to the QAM, and the technical specifications of customer 44330 pur-chase order (P0) 54-7-FLC-40416, the dose rate required by the P0 was not achieved and there was no documentation in the test file identi-fying, explaining, justifying or authorizing the deviation.

C. UNRESOLVED ITEMS: None.

D. STATUS OF PREVIOUS INSPECTION FINDINGS

None. l E. OTHER FINDINGS OR COMMENTS:

1. A technical review and evaluation of EQ irradiation of Class 1E components was conducted on selected projects from one NSSS supplier, two manufacturers and two test laboratories. The test files were reviewed for technical completeness, accuracy, and consistency and for compliance with referenced specifications and regulatory requirements.

120 I l

i ORGANIZATION: ISOMEDIX (NEW JERSEY) INC. WHIPPANY, NEW JERSEY REPORT INSPECTION NO.- 99900913/85-01 RESULTS: PAGE 3 of 4 The following test files were reviewed: Control No. Component Status

a. 00084 3 Damper assemblies Completed

b. 00085 Large pump motor Completed

c. 00144 Pads, seals Completed (assoc.

with 07142)

d. 00309 Cables, trays, tape Completed

e. 00370 Valve seats, pads, discs Completed

f. 00424 Valve actuator (see 00636) Completed (aging dose) 9 00524 Motor controller Completed

h. 00578 Cable connectors Coinpleted

i. 00636 Valve actuator (see 00424) Completed (DBE dose)

j. 00766 Magnetic sensors Completed

k. 07142 Connector Completed

1. 07245 Cables In progress During review of the above test files, the nonconformances of section B were identified.

2. With respect to nonconformance B.2, in the cases in which the test items were small enough to see an insignificant flux gradient over the space occupied by the items, the dose rate could appropriately have been described as essentially uniform, but the test report makes no reference to uniformity. In cases where the raw data indicates non-uniform dose rates and their distribution over the field, these data were not included in the test reports cited.

3. With respect to nonconformance B.3, Isomedix stated that a customer

! representative had delivered the samples for irradiation prior to Isomedix's receipt of the P0. The customer representative provided Isomedix a single sheet excerpt from a test plan which he identified verbally as the test specification but which bore no identifying , information. The wording of this document was similar to that of l the subsequently received P0 with the exception that it provided for dose rates of less than or equal to 1 Mrad per hour. Isomedix further stated that the customer representative was aware of the parameters of the radiation delivered. l l 121

ORGANIZATION: ISOMEDIX (NEW JERSEY) INC. WHIPPANY, NEW JERSEY REPORT INSPECTION Nn - 99900913/85-01 RESULTS: PAGE 4 of 4

4. Test report letters reviewed contained the following statement:

           " Radiant heat from the source heated the samples somewhat, but the temperature did not exceed 100 degrees Fahrenheit as indicated by previous measurements of an oil solution in the same relative position."

Isomedix stated that this information was intended only as a rough indicator of the radiant heating effect in the irradiator and they do not claim it to be an accurate measure of sample temperature. The NRC inspector recommended clarification of the statement to reflect its intended significance and prevent it from being misconstrued. Isomedix stated that they were considering discontinuing use of the statement and that in practice, when a customer needs actual sample temperature data, Isomedix requires the customer to provide instrumen-tation and support personnel as necessary.

5. The irradiation data forms in use were not the same as described in related documentation and the NRC inspector noted numerous inconsistencies in how they were filled out. These forms also had no place to record deviations or anomalies as required by procedures.

Isomedix stated that the form was being redesigned, and that related procedural documentation was being revised to refer to the new form and to describe in detail the information required to be recorded.

6. Paragraph 7.2 of the QAM describes records relating to irradiation as " permanent." It requires QA records to be retained for 5 years and requires records relating to purchase and calibration [of dosimetry and test equipment] to be retained for 1 year beyond the l life of the equipment. This is inconsistent with paragraph 17.4 of Addendum I to the QAM which requires records to be retained for 7 years. Isomedix stated that the QAM was under revision and that this inconsistency would be rectified.

Isomedix's actions regarding concerns addressed in paragraphs E.2 through E.6 above will be reviewed in a future inspection. 122

ORGANIZATION: MORRISON-KNUDSEN COMPANY, INC. ROCKY MOUNT, NORTH CAROLINA REPORT INSPECTICN INSPECTION NO.- 99900702/85-01 DATE(Sh 5/14-16/R; nN RTTF MAMDR. 1A CORRESPONDENCE ADDRESS: Morrison-Knudsen Company, Inc. ATTN: Mr. W. Frank Jones Vice President and General Manager Post Office Box 1928, 101 Gelo Road Rocky Mount, North Carolina 27802 i ORGANIZATIONAL CONTACT: Mr. Harry W. Falter, Division Engineer TEtFPHONF N!NRFR- (410) 477 M?n PRINCIPAL PRODUCT: Emergency diesel driven power systems. NUCLEAR INDUSTRY ACTIVITY: Currently the company is involved in providing modifications and maintenance services to diesel engines installed in nuclear power plants and developing training and service instructions for personnel in the nuclear power plants. m I ASSIGNED INSPECTOR: (f -( K. R. Naidu, Reactive Inspection Section (RIS) 7/2/ff Date

                                                                                         ~

OTHER INSPECTOR (S): E. L. Burns, Consultant t

                                                  )

APPROVED BY: . 7!! E. W. Merschoff/ Ehief. RIS. Vendor Proaram Branch Date INSPECTION BASES AND SCOPE: A. BASES: 10 CFR Part 21 and 10 CFR 50 Appendix B. B. SCOPE: Reviewed corrective action taken on previously identified items, reviewed the status of Part 21 reports. i PLANT SITE APPLICABILITY: Browns Ferry 2/3, Davis Besse, Grand Gulf (HPCS) 1/2, ' St. Lucie 2, Sequoyah 1/2, Watts Bar 1/2, Susquehanna, Oconee, Three Mile Island, Trojan 1, Fitzpatrick, Rancho Seco. L l 123 l

ORGANIZATION: MORRISON-KNUDSEN COMPANY, INC. ROCKY MOUNT, NORTH CAROLINA 9 REPORT INSPECTION NO.- 99900702/85-01 RESULTS: PAGE 2 of 7 A. INSPECTION ISSUES: The Power Systems Division of Morrison-Knudsen Company, Inc. (MKC) packages emergency electric power systems with diesel engines manufactured by other companies. Two 10 CFR Part 21 items were reported on the diesel engines procured by MKC related to a lube oil modification problem and an electrical generator failure. B. BACKGROUND INFORMATION: The Power Systems Division of Morrison-Knudsen Company (MKC) is an engineering company which packages emergency power systems. MKC purchased diesel engines manufactured by the Electromotive Division (EMD) of General Motors, and Cooper Bessemer Company. MKC used electrical generators manufactured by Electric Products, Beloit Power Systems (currently known as Louis-Allis), Westinghouse, Electric Machining Manufacturing and General Electric. Excitation cubicles for the electrical generators were either supplied by the manufacturer of the electrical generator or by Basler Company. MKC has submitted bids for emergency power systems for Nuclear application with quotations from diesel engines manufactured by Motoren and Turbinen Union (Friedrichshafen, West Germany); Sulzer Brothers (Switzerland) and Grandimatori (Italy). MKC employes 100 persons at Rocky Mount. Currently, the company is involved in providing modifications and maintenance services to diesel engines installed in nuclear power plants and developing training and service instructions for personnel at nuclear power plants. C. FOLLOWUP ON PREVIOUSLY IDENTIFIED ITEMS:

1. (Closed) Violation (83-01): This item related to the corrective action taken to resolve excessive engine room ambient temperature presumed to be the result of under-designed heating, ventilation and cooling (HVAC) equipment. MKC developed a generic procedure for licensees to test the installed engines. It was realized that the ambient temperature is dependent on the velocity and quantity of air j moved over the diesel engine and will vary with the size of the room and location of the HVAC ducts'. Tennessee Valley Authority (TVA) conducted tests at the Watts Bar Nuclear Power Plant based on the MKC procedure. The results at Watts Bar indicate that the revised i heatload is conservative. The results of a test conducted at the Trojan Nuclear Power Plant indicates that the revised heatload was acceptable. Other nuclear plants notified of this problem by MKC have not yet responded with their results.

124 l

l ORGANIZATION: MORRISON-KNUDSEN COMPANY, INC. ROCKY MOUNT, NORTH CAROLINA REPORT INSPECTION j NO - 9990070?/85-01 RESULTS: PAGE 3 of 7 '

2. (Closed) Nonconformance 83-01: This item related to the omission of imposing the requirements of 10 CFR Part 21 and 10 CFR 50 Appendix B on Purchase Order 50206-6036/379 for the supply of level indicators I

and adapters. As indicated in the MKC letter dated October 17, 1983, to the NRC, the applicable MKC procedure (PCP 201) was revised. The revised procedure requires the originator to impose the 10 CFR Part 21 and 10 CFR 50 Appendix B requirements on safety related items. The current Revision 5 of PCP 201 dated April 25, 1984, contains these requirements in paragraphs 6.1.8, 6.1.9 and 6.1.10. The level indicators and the adapters referenced were not safety related.

3. (Closed) Violation Item A 84-01: This violation identified that measures adopted by MKC did not provide for review, evaluation, and escalation into a Part 21 system for all deviations. MKC developed procedure EP-205 titled " Reporting of Defects and Noncompliance."

This procedure provides for review, evaluation and escalation. Records indicate that all cognizant personnel were trained in the ise of this procedure.

4. (Closed) Violation Item B 84-01: The violation identified that MKC did not notify the NRC of the existence of a defect relative to the "A.C. Generator roller bearing" and the " Lube oil cooler leak" within the required time limit in section 21.21(b)(2) of 10 CFR Part 21.

The MKC Quality Assurance Instruction QAI 15-1 was revised and paragraph 6.3.2 states in part " Notification shall be made within 48 hours. If this notification is by means other than written communication, a written report shall be submitted to the appropriate office within five (5) days after the information is obtained."

5. (Closed) Nonconformance Item 1 84-01: This item identified lack of source evaluation records on four companies. Records were available to indicate that evaluations were performed on Starret on August 27, 1984; Anglo Repair Service on August 20, 1984; TRW-J.W. Williams Division on August 23, 1984; and Dresser Industries on September 29, 1984. The current MKC Approved Suppliers List reflects source evaluations.

6. (Closed) Nonconformance Item 2 84-01: This item identified inadequate implementation of precedure QCP-7 relative to the placement on and maintenance of the MKC Approved Suppliers List (ASL). MKC developed

, .QA Procedure QAP to control placement and on-going status of vendors I listed on the ASL. Corborundum and Machine Welding Supply were deleted from the ASL. Faster Supply and System Service Corporation l were audited on March 30, 1984 and March 31, 1984, respectively. 125

ORGANIZATION: MORRISON-KNUDSEN COMPANY, INC. ROCKY MOUNT, NORTH CAROLINA l REPORT INSPECTION NO.- 99900702/85-01 RESULTS: PAGE 4 of 7 i

7. (Closed) Nonconformance Item 3 84-01: This item related to failure l to follow procedures to include Quality Levels (QL) on Purchase Requisitions (prs) and Purchase Orders (P0s) and failure of QA to l review and approve P0s. As stated in MKC's response dated October 12, 1984, Change Orders were issued to prs and P0s to include the QLs.

8. [ Closed)NonconformanceItem484-01: This item related to failure to follow procedures relative to certification of NDE personnel. All currently employed inspectors were administered the appropriate examinations, including visual acuity. The certifications were examined and determined to be acceptable.

9. (Closed) Nonconformance Item 5 84-01: This item related to nonavail-ability of welder qualification records for several welders. The welder qualification recvds were examined and determined to be adequate. The records identified the welder, the Weld Procedure Specification (WPS) used and the month welded. The Level III inspector assures that the welders maintain their qualification by welding to each WPS within three months.

10. (Closed) Noncompliance Item 6 84-01: This item identified that a procedure was not developed to perform heatload tests on an installed diesel generator. MKC developed a procedure dated May 5, 1982, to determine the heat rejected by an EMD Model 645E4 diesel engine manufactured by General Motors. This test has been completed at Watts Bar and Trojan Stations.

D. INSPECTION FINDINGS AND OTHER COMMENTS:

1. Processing of NRC Information Notices i

t The Division Engineer distributes copies of NRC Information Notices (IN) to his staff alerting them of events reported relative to Diesel Engine problems experienced at nuclear plants so they can review these

events for applicability to their machines. Records indicate that NRC ins 85-08 (Corrosion Lining in Tanks), 85-27 (Defects Reporting),

and 85-25 (Consideration of Thermal Conditions in the Design and j Installation of Supports # tor Diesel Generator Exhaust Silencers) l were distributed to the engineering staff. 126

ORGANIZATION: MORRISON-KNUDSEN COMPANY, INC. ROCKY MOUNT, NORTH CAROLINA REPORT INSPECTION NO.- 99900702/85-01 RESULTS: PAGE 5 of 7 ,

2. Part 21 Report by Hub, Inc.

MKC evaluated a Part 21 report from Hub, Inc. to ascertain whether any SA-106 pipe manufactured by Phoenix Steel Corporation and supplied by Hub, Inc. was used in their installations. The Part 21 report stated that the SA-106 pipe may have a wall thickness problem due to the inspection method used by Phoenix. MKC determined that SA-106 pipe manufactured by Phoenix had not been used.

3. Westinghouse AC Generator Failure (MKC Serial # 0023)

MKC reported on April 19, 1985 that an AC generator bearing seal failed during a test at the Napot Point Nuclear Plant (Napot) in the Phillipines. The electrical generator was manufactured by Westinghouse at their Round Rock plant located in Texas. Records indicate that Westinghouse manufactured only four of these AC generators of which two were supplied to Napot in the Phillipines and two to Kori Nuclear Power Plant in South Korea. During a routine full load surveillance test performed at Napot, the generator bearing experienced a problem. Subsequent investigation revealed that the oil thrower lockwasher had been omitted from the assembly which, in time, allowed the outboard oil thrower / slinger to back off the thread such that it contacted the bearing cap and the thread forced the two pieces together. Extreme heat developed due to friction and the bearing lube oil was lost because the slinger no longer performed ! the lubrication function. Inspection revealed that the insulation had deteriorated and allowed the bearing to drop slightly causing contact with both inboard and outboard seals. MKC stated that prior to this incident, two generators were returned to Westinghouse Round Rock plant for slip ring replacement because the slip rings had rusted during storage. MKC stated that during the replacement of the slip rings, the oil thrower lockwasher was inadvertantly omitted during the reassembly. This failure is restricted to one of the two generators supplied to Napot. MKC notified the NRC that the Part 21 notification is being withdrawn because the failure is not reportable. l 127

ORGANIZATION: MORRISON-KNUDSEN COMPANY, INC. ROCKY MOUNT, NORTH CAROLINA REPORT INSPECTION NO - 9990070?/85-01 RESULTS- PAGE 6 of 7

4. Lube Oil Modification on 6 GPM Circulating System MKC (serial # 0020) reported a 10 CFR Part 21 relative to the low standby lube oil pressure alarm switch. MKC modified the lube oil systems in certain models of diesel engines manufactured by the Electro Motive Division (EMD) of General Motors. Specifically, the lube oil modification divides the standby lube oil flow into two channels. Channel one consists of a motor driven pump that draws lube oil from the sump and pumps through a filter to the engine turbocharger. The other draws lube oil from the sump and pumps to the engine scavenging pump piping system. The latter furnishes lube oil to the piping between the lube oil scavenging pump discharge and the strainer mounted on the engine. The oil flow in this channel is monitored by a pressure switch located in the piping between the motor driven pump discharge and the 30 psi spring loaded relief check valve discharging into the piping system. The spring loaded check valve allows the pressure to be monitored since the pressure in the scavenging piping system is too low to unseat the check valve. (The report did not apply to installations where a single motor driven pump is used to furnish the lube oil scavenging pump with oil.) The modification required the installation of a check valve on the discharge side of each pump to prevent one pump from pumping back through the other pump. The pressure alann switch is attached to a section of piping which has a check valve at each end. These check valves will close when the pumps are off and will trap pressure between them, which, in effect defeats the function of the alarm switch. The control switches which start and stop the motor driven pumps are not located in this piping section. Therefore, the function of the motor driven pumps is not affected. Corrective action recomended was to drill a 1/16" diameter hole in the disc of the 30 psi spring loaded relief check valve to allow the pressure to bleed into the scavenging pump piping. MKC reviewed the applicability of 10 CFR Part 21 and determined that the oil pressure switch will indicate a high pressure condition, but will not cause the diesel engine to fail to start or fail to perform its safety function. The MKC evaluation is considered acceptable.

5. High Capacity Turbo Charger Problems MKC (serial # 0021) informed the NRC of High Capacity Turbo Charger failures identified in diesel engines manufactured by Electro Motive

, Division (EMD) of General Motors. These failures were experienced in diesel engines used in freight locomotives. EMD's preliminary evaluation is that this problem is confined to locomotive operation. 128 1

ORGANIZATION: MORRISON-KNUDSEN COMPANY, INC. ROCKY MOUNT, NORTH CAROLINA REPORT INSPECTION NO.- 9990070?/85-01 RESULTS: PAGE 7 of 7 EMD has not forwarded to MKC a detailed report. EMD recommended a precautionary inspection program for high capacity planetary gear trains of 17.9:1 ratio turbo chargers. EMD informed NKC that they found grinding notches in high capacity style sun gears at or below the optimum radius at the root of the gear tooth. EMD contends that the gear tooth stress levels are altered as a function of gear tooth notch radius and unequal loading of the planet gears due to eccentricities of the gearshaft. To date no problems have been identified in standby emergency diesel generators. MKC is awaiting further information from EMD. E. The inspectors met with MKC personnel identified in paragraph F at the conclusion of the inspection and discussed the scope and results cf the inspection. F. Morrison-Knudsen Company, Inc.

       *H. W. Falter, Division Engineer
       *M. Vann Mitchel, Manager, Quality Assurance H. E. Loewe, Senior Quality Assurance Engineer D. S. Odom, Associate Engineer, Special Projects K. Lewis, Supervisor, Technical Services and Maintenance S. Cramton, Field Engineer l

• Denotes those individuals who attended the exit interview on 5/16/85.

129

ORGANIZATION: MULTI-AMP SERVICES COMPANY DALLAS, TEXAS REPORT INSPECTION INSPECTION NO.: 99900539/85-01 DATE(S): 4/9-12/85 ON-SITE HOURS: 40 l CORRESPONDENCE ADDRESS: Multi-Amp Services Company i ATTN: Mr. John J. Lapicola l Vice President-General Manager 4271 Bronze Way Dallas, Texas 75237 ORGANIZATIONAL CONTACT: Ms. Tienna Rogers, QA Manager TELEPHONE NUMBER: (214) 333-3201 PRINCIPAL PRODUCT: Technical Services ! NUCLEAR INDUSTRY ACTIVITY: Approximately 90 percent. l - ASSIGNED INSPECTOR: s/ir[rr r.=R.E.Oller,pctiveInspectionSection(RIS) Date 0THER INSPECTOR (S): J. Harper, RIS l APPROVED BY: sdr[ts-E. W. Merschoff p ief, RIS, Vendor Program Date l Branch INSPECTION BASES AND SCOPE: A. BASES: 10 CFR Part 21 and Appendix B to 10 CFR Part 50. l B. SCOPE: This inspection was made as a result of an allegation received by l the NRC Region IV office concerning deficiencies in the implemen-l tation of the Multi-Amp Services' Quality assurance program. I Additionally, the status of the previous inspection findings was reviewed. PLANT SITE APPLICABILITY: 50-483 131

i l ORGANIZATION: MULTI-AMP SERVICES COMPANY l

                  -DALLAS, TEXAS REPORT                               INSPECTION NO.:   99900539/85-01               RESULTS:                             PAGE 2 of 4 A. VIOLATIONS:

None B. NONCONFORMANCES: None C. UNRESOLVED ITEMS: None D. STATUS OF PREVIOUS FINDINGS:

1. (Closed) Violation (Report No. 83-01): Contrary to 10 CFR Part 21, Section 21.21, Multi-Amp Services failed to: (a) adopt appropriate procedures to provide for evaluating deviations and informing the purchaser of deviations, and (b) post a notice which described the regulations / procedures and states where the current copies may be examined.

During the current NRC inspection it was verified that Multi-Amp Services has adopted written procedure No. QAP 15-001, Revision 0, dated March 30, 1984, to provide the required reporting measures. Also, they have posted in a conspicuous location current copies of the documents required by 10 CFR Part 21.

2. (Closed) Nonconformance (Report 83-01): Contrary to Criterion V of Appendix B to 10 CFR 50, and procedure No. QAP 2-001, Multi-Amp Services had certified two personnel to ANSI Levels II and III with-out either person having had the required combination of education and experience or documented evidence of demonstratef qualifications or equivalency.

During this inspection it was verified that the Level II technician was not required by his duties to be certified and appropriate qualification records for the Level III technician are now available. E. OTHER FINDINGS OR COMMENTS:

1. Initial Management Meeting and Exit Interview: The Multi-Amp Services Company _ management representatives were informed of the allegation which was the reason for the inspection. The scope of the inspection and its required documentation were explained. During the exit meeting, the inspection findings were explained.

132 l

ORGANIZATION: MULTI-AMP SERVICES COMPANY DALLAS, TEXAS REPORT INSPECTION N0.: 99900539/85-01 RESULTS: PAGE 3 of 4

2. Allegation:

a.

Introduction:

On March 7, 1985, the NRC Region IV office received information by telephone with regard to potential deficiencies in the implementation of the Multi-Amp Services Company QA program in the area of audits. Multi-Amp Services furnishes technical, metrology and training services to numerous nuclear power plants (NPP). As a result of this allegation, an NRC inspection was performed on April 9-12, 1985, at the Multi-Amp Services' plant located in Dallas, Texas.

b. Findings: The NRC inspectors performed an independent assess-ment of the telephonic allegation by means of discussions with the alleger, and review of the Multi-Amp Services' QA manual, procedures, audit records and internal memoranda. The alleger had indicated the problem started in April 1984 and consisted of not being allowed by corporate management to perform audits of Multi-Amp Services site technician's certification records located at the Calloway NPP and six other sites, and of Alnor Instruments. Alnor is a Multi-Amp Services' subsupplier of calibration service for certain instruments used at Calloway.

Discussions established that the alleger knows of no safety related problems at the other six NPPs, since audits had been performed where scheduled. An audit of Multi-Amp Services' site personnel certification records at Calloway was scheduled for January 1985, but was not performed until March 13, 1985. During this audit, no problems were identified. With regards to Alnor Instruments, the NRC inspectors verified that the December 1983 and 1984 annual audits had not been performed. Alnor was subsequently audited by Multi-Amp services April 2,1985, and no problems were identified. Included in the allegation was a statement that three items of concern, referenced as allegations numbers 1, 2 and 10 in the NRC Inspection Report No. 99900539/83-01, were still true. These allegations had been determined by the NRC in November 1983, as not having been substantiated. These items included: (a) Multi-Amp's QA had inadequate organizational freedom and independence; (b) there was inadequate separation between QA and corporate management, and (c) QA was not effective. Review of the Multi-Amp Services' QA manual Section 1, Organiza-tion; the implementing procedure No. QAP-1, Revision B, and internal memoranda generated during 1984 and 1985, failed to identify any new objective evidence that would change the find-ings made by the NRC in 1983. 133

ORGANIZATION: MULTI-AMP SERVICES COMPANY DALLAS, TEXAS REPORT INSPECTION NO.: 95900539/85-01 RESULTS: PAGE 4 of 4 In summary, this inspection substantiated the allegation in part, as it applied to the Alnor Instruments. The safety related con-cerns relative to Calloway and six other NPPs were not substantiated. The three allegations referenced in the 1983 NRC report were not substantiated. In summary no safety related problems were identified. I ( l l 134

ORGANIZATION: NATIONAL TECHNICAL SERVICES HARTWOOD, VIRGINIA REPORT INSPECTION INSPECTION NO.: 99900914/85-01 DATE(S): 3/4-8/85 ON-SITE HOURS: 72 CORRESPORDENCE ADDRESS: National Technical Services ATTN: Mr. W. J. Ison Division Vice President State Route 748, Box 38 Hartwood, Virginia 22471 ORGANIZATIONAL CONTACT: Mr. W. C. Hartman, Quality Control Manager TELEPHONE NUMBER: (703) 752-5300 PRINCIPAL PRODUCT: Testing Laboratory. NUCLEAR INDUSTRY ACTIVITY: Approximately 15 percent of the National Technical , Systems (NTS) total business (dollar value) is a result of testing of equipment for the nuclear power industry.

                                      ,A ASSIGNE0 INSPECTOR:      bd \d (n. m o                                      f-14-8'I G. T. Hubbard, Equip. QVal.' Inspec. Section (EQIS)      Date OTHER INSPECTOR (S):   R. N. Moist, EQIS l

. APPROVED BY: M e at 2,-p\/2.-- El4-tf U. Potapovs, Chief, EQIf, Vendor Program Branch Date INSPECTION BASES AND SCOPE: A. BASES: Appendix B to 10 CFR Part 50. B. SCOPE: This inspection consisted of: (1) a technical evaluation of equipment qualification (EQ) test activities for safety-related equip-ment; (2) verification of implementation of corrective action (CA) on the nonconformances identified in NRC Inspection Report Nos. 99900914/83-01 and 84-01; and (3) verification of implementation of the quality assurance (QA) program. PLANT SITE APPLICABILITY: Not identified. i 135

ORGANIZATION: NATIONAL TECHNICAL SERVICES HARTWOOD, VIRGINIA REPORT INSPECTION NO.: 99900914/85-01 RESULTS: PAGE 2 of 6 A. VIOLATIONS: None. B. NONCONFORMANCES:

1. Contrary to Criterion V of Appendix B to 10 CFR Part 50 and paragraph 4.3.2.4 of the Quality Control Manual (QCM), the test technician had not initialed and dated corrections to recorded data on data sheets for Master Job Order (MJ0) 558-1572.

2. Contrary to Criterion V of Appendix B to 10 CFR Part 50 and paragraph 2.4.3.1 of the QCM, the test technician had not initialed the appropriate column of Job Traveler Forms (JTFs) for MJ0's 556-1720, 558-1686, and 557-1382 when it had been ascertained that a test had been conducted and completed in accordance with applicable specifi-cations.

3. Contrary to Criterion V of Appendix B to 10 CFR Part 50 and para-graphs 12.2.8 and 12.3.1 of the QCM, the audit report for the May 1984 corporate quality internal audit was not issued within thirty days of the conclusion of the audit, did not request a response date for corrective action, and there was no documented evidence to indicate that the required follow up had been performed by the Lead Auditor.

4. Contrary to Criterion V of Appendix B to 10 CFR Part 50 and paragraph 6.3.1.1.(7) of IEEE-STD 323/1974, NTS test plans or test procedures did not list test equipment requirements for MJO 558-1572 or list all test equipment accuracies for MJ0s 558-1572 and 557-1382.

C. UNRESOLVED ITEMS: None. D. STATUS OF PREVIOUS INSPECTION FINDINGS:

1. (Closed) Nonconformance (84-01): The current annual review of the l

QCM had not been performed. The NRC inspector verified that the

annual review and revision of the NTS QCM was completed as committed I in NTS's corrective action response.

2. (Closed) Nonconformance (84-01): The current yearly re-evaluation of test personnel had not been performed for four test personnel. The NRC inspector reviewed four supplements to personnel records and verified that the NTS re evaluation of the four test personnel was 136

ORGANIZATION: NATIONAL TECHNICAL SERVICES HARTWOOD, VIRGINIA l REPORT INSPECTION N0.: 99900914/85-01 RESULTS: PAGE 3 of 6 conducted and completed as committed in their response. Additionally, , the NRC inspector verified adequate CA on this nonconformance by ' reviewing the internal scheduling system set up by the Quality Control Department.

3. (Closed) Nonconformance (84-01): General data log sheets were found for MJO 557-1434 which the test technician had not signed or dated.

Tfie NRC inspector randomly selected general data log sheets from four MJO files to verify that data sheets are now being signed and dated by the test technician. The NRC inspector also reviewed the quality control audit report and associated checklist which was for an internal audit conducted on December 17-18, 1984, by the NTS, Hartwood, QC Manager. The checklist contained a characteristic relative to signing and dating general data log sheets and no deficiencies were identified in this area during the December 1984 audits.

4. (Closed) Nonconformance (84-01): Original test data sheets for seismic tests conducted under MJO 557-1434 were not filed in the " Job Package" and were not available for inspection by the NRC inspector. The NRC inspector verified that the original data sheets for seismic tests i

were now in the job package. The inspector also reviewed three additional job packages to verify that all test sheets are being kept in their associated MJO files.

5. (Closed) Nonconformance (84-01): NTS's test report reviews do not

always assure that the results of test programs are accurately reported. The NRC inspector verified that the appropriate corrections were anotated in Revision 2 of Nuclear Qualification Report (NQR) 557-14778. The inspector also reviewed four additional test reports and their associated MJO files'and determined that the reports are accurately reflecting results of test programs.

6. (Closed) Nonconformance (84-01): No enveloping profile or synopsis of tae acceleration of the total duration of the design basis event (DBE) and post-DBE from one year to approximately 30 days was provided in NQR 557-14778. The NRC inspector verified that the correction was anotated in Revision 2 of NQR 557-14778. The inspector's review of four additional MJ0s determined that test reports are adequately including all appropriate data and analysis necessary for complete understanding of conducted test programs.

137 t

ORGANIZATION: NATIONAL TECHNICAL SERVICES HARTWOOD, VIRGINIA REPORT INSPECTION , NO.: 99900914/85-01 RESULTS: PAGE 4 of 6

7. (Closed) Nonconformance (83-01): The audit report for the April 1982 corporate quality internal audit did not request a response date for CA and there was no documented evidence to indicate that the required follow-up had been performed. The NRC inspector verified that the CA forms which were generated by a previous Hartwood QC manager relative to the April 1982 corporate audit and which were not available for NRC review during the May 1984 NRC inspection, were permanently lost. Based on discussions with the new QC manager and review of available documentation and actions taken by the new manager, the NRC inspector determined that adequate CA had been accomplished relative to the April 1982 corporate audit.

D. OTHER FINDINGS OR COMMENTS:

1. QC Program Implementation Review - The NRC inspectors reviewed four MJO files containing documentation which supported testing efforts.

Documentation included purchase orders, receiving and shipping documents, test plans, test procedures, test reports, and test data. The inspectors' review of the above documentation was to verify continued implementation of the NTS QA program. During the above review and evaluation, the NRC inspectors identified the nonconform-ances described in B.1, B.2 and B.3. The NRC inspector also reviewed changes that were incorporated into Revision 3 of the QCM. The inspector determined that the changes were minor and did not change the NTS's QA program with respect to the requirements of Appendix B to 10 CFR Part 50.

2. Technical Evaluation - The NRC inspectors performed an in-depth technical evaluation and review of four MJO files for qualification testing of safety-related electrical equipment. The following table summarizes the MJO file numbers examined including equipment type and types of documents examined.

MJO Equipment Type Documents Examined 557-1382 Solenoid valves Test report, test plans, test procedures, test specification, purchase order. 558-1572 "MI" cable and Connector Test report, test plans,

Assemblies test procedures, test specification, purchase order.

l l 138

ORGANIZATION: NATIONAL TECHNICAL SERVICES HARTWOOD, VIRGINIA REPORT INSPECTION NO.: 99900914/85-01 RESULTS: 4 PAGE 5 of 6 I i MJO Equipment Type Documents Examined 558-1686 Insulated Motor Stator Test report, test plans, test procedures, test specification, purchase order. 558-1720 Solenoid Valve Test report, test plans, test procedures, test specification, purchase order. The NRC inspector reviewed the EQ process prescribed in each test plan and reviewed test results, including the bases for accelerated thermal aging and radiation and verified calculations. ', Each of the four EQ test plans and related engineering documents were examined for the following:

a. Adequate test instrumentation and their accuracies were described and used to meet the requirements of NUREG-0588/

IEEE-STD-323/1974.

b. Equipmer.t interfaces were addressed.

c. Test acceptance criteria were established as described in the test specification or in the design engineering documents, such as calculations and engineering letters to meet the nuclear regulatory requirements of NUREG-0588/IEEE-STD-323/1974.

d. Same equipment was used for all phases of testing and represented a standard production item.

e. Environmental conditions were established and dese-ibed (e.g., pressure and temperature profiles, and thermal aging factors were consistent with those outlined in the test specification or test plan). >

f. Test results were adequately reduced and evaluated ,

against established acceptance criteria described in customer test specifications or purchase orders and that these requirements had been met. i 139

                                                                                       ) > :

i ORGANIZATION: NATIONAL TECHNICAL SERVICES HARTWOOD, VIRGINIA REPORT INSPECTION NO.- 99900914/85-01 RESULTS: PAGE 6 of 6

g. All prerequisites for the given tests as outlined in the test specification had been met.

h. Test equipment included a description of all materials, parts, and subcomponents. l During the above evaluation and review, the NRC inspectors identified the nonconformance described in paragraph B.4.

1 9 I f I i 140 1

ORGANIZATION: NATIONAL TECHNICAL SYSTEMS ACTON, MASSACHUSETTS REPORT INSPECTION INSPECTION N0.: 99900912/85-01 DATE(S): 4/1-5/85 ON-SITE HOURS: 91 CORRESPONDENCE ADDRESS: National Technical Systems ATTN: Mr. G. Dowd General Manager 533 Main Street i Acton, Massachusetts 01720 ORGANIZATIONAL CONTACT: Mr. A. A. Dorr, Quality Assurance Manager TELEPHONE NUMBER: (617) 263-2933 PRINCIPAL PRODUCT: Environmental testing services. NUCLEAR INDUSTRY ACTIVITY: National Technical Services (NTS) provides qualifi-cation testing services for the military and both the conventional and nuclear ' power industry. Approximately 25 percent of the services are for the commer-cial nuclear power industry. ASSIGNED INSPECT 0 : hlM 6 ,w _ G. T. Hubbard, Equip. lual. Inspec. Section (EQIS) CG-f(C Date OTHER INSPECTOR (S): R. N. Moist, EQIS j E. R. Richards, Sandia National Laboratories APPROVED BY: W C VC U. Potapovs, Chief E0 S, Vsndor Program Branch Date INSPECTION BASES AND SCOPE: A. BASES: Appendix B to 10 CFR Part 50 and 10 CFR Part 21. 1 B. SCOPE: This inspection consisted of: (1) a technical evaluation of Equipment Qualification (EQ) test activities for safety related equip-ment, (2) verification of implementation of the quality assurance program (QAP), and (3) evaluation of test data relative to a 10 CFR Part 21 report issued by Anacon concerning the loss of adhesive properties of the epoxy used in manufacturing the probes for Anacon chlorine monitors. PLANT SITE APPLICABILITY: 50-312 and 412 (Beaver Valley 1 and 2) and 50-334 (Rancho Seco). 141

ORbANIZATION: NATIONAL TECHNICAL SYSTEMS ACTON, MASSACHUSETTS REPORT INSPECTION NO.- 99000912/85-01 RESULTS: PAGE 2 of 5 A. VIOLATIONS: None. B. NONCONFORMANCES:

1. Contrary to Criterion V of Appendix B to 10 CFR Part 50 and IEEE Standard 323-1974 paragraph 6.3.1.1(7), NTS test plans or test proce-dures did not list test equipment requirements or their accuracies.

2. Contrary to Criterion V of Appendix B to 10 CFR Part 50 and the paragraph entitled, " Test Reports" of NTS standard operating procedure, 50P 10/83, dated February 1, 1985, the individual who

      )                had approved revision 1 to qualification test report 17807-1 was not on NTS's list of individuals authorized to approve test reports.

C. UNRESOLVED ITEMS: t None. D. OTHER FINDINGS OR COMMENTS:

1. Technical Evaluation: The NRC inspector and Sandia consultant performed an in-depth technical evaluation and review of seven test programs for qualification testing of safety-related electrical equipment. The following table summarizes the test programs examined including equipment type and types of documents examined.

Program Number Equipment Type Documents Examined 18227 Components for Test procedure (TP), test Remote Shutdown specification, purchase l Panel order, log books, test data sheets 17244-82N-C Magnetrol Level Test reports (TRs), purchase 1 82N-E Controls, Micro orders (P0s) j- 82N-F Switches, 0-rings, 82N-F-1 Switches, phenolic 82N-8 terminal blocks and switchboard cable 142

ORGANIZATION: NATIONAL TECHNICAL SYSTEMS ACTON, MASSACHUSETTS REPORT INSPECTION NO.- 99900912/85-01 RESULTS: PAGE 3 of 5 i Program Number Equipment Type Documents Examined 18577-83N S0R pressure TR switch Test program 18227 had been completed, but the TR had not been written; therefore the TR will be reviewed during a future inspection. Supporting documentation for programs 17344-82N-C, -E,

            -F,  -F-1 and -8 and 18577-83N had been reviewed during previous inspections when the TRs had not been complete; therefore, only the TRs and P0s for these programs were reviewed and evaluated during this inspection.

The NRC inspectors and Sandia consultant reviewed and evaluated the EQ process prescribed in each test program. The test documentation reviewed for each of the seven programs was examined to verify that:

a. Adequate test instrumentation and its accurancy was described and used to meet the requirements of NUREG-0588.

b. Equipment interfaces were addressed.

c. Test acceptance criteria were established as described in the test specification or in the design engineering documents, such as calculations and engineering letters to meet the nuclear regulatory requirements of NUREG-0588/IEEE-STD-323-1974.

d. Same equipment was used for all phases of testing and represented a standard production item.

e. Environmental conditions were established and described (e.g.,

pressure and temperature profiles, and thermal aging factors were consistent with those outlined in the test specification or test plan).

f. Test resuits were adequately reduced and evaluated against established acceptance criteria described in customer test specifications or purchase orders.

g. All prerequisites for the given tests as outlined in the test specification had been met.

l l 143

ORGANIZATION: NATIONAL TECHNICAL SYSTEMS ACTON, MASSACHUSETTS REPORT INSPECTION NO.- 99900912/85-01 RESULTS: PAGE 4 of 5

h. Test equipment included a description of all materials, parts, and subcomponents.

During the above evaluation and review, the NRC inspector identified the nonconformance described in paragraph B.1.

2. Upcoming Testing: During the inspection, the NRC inspectors reviewed purchase orders and preliminary test procedures associated with three test programs scheduled to start in the next several months. These three programs (project numbers 19547, 17297-1-82N, and 18990) consist of the testing of flow transmitters, current transformers, and pressure transducers to simulated environmental test conditions up to and including loss-of-coolant accident (LOCA)/high energy line break (HELB) environments. The test efforts on the above programs will be reviewed and evaluated during future NRC inspections.

3. Followup on 10 CFR Part 21 Report: Anacon, Burlington, Massachusetts, filed a 10 CFR Part 21 report with NRC Headquarters concerning the loss of adhesive properties of the epoxy used in manufacturing the probes used in their chlorine monitors.

The NRC inspector reviewed and evaluated one TP, one TR, and the failure notification file relative to NTS's qualification testing of the Anacon chlorine monitors. The inspector determined that seven chlorine detector probes, along with other associated equip-ment that comprise the chlorine monitors, started the test program, which included functional tests, thermal aging, radiation aging, seismic testing, and abnormal environment simulation. All seven probes were exposed to thermal and radiation aging; however, only four probes were initially subjected to seismic testing. During this seismic testing of the four probes, the epoxy adhesive failures l were observed. Following the adhesive failures, which were attributed to the 85 C temperature for 543 hours used during thermai aging, the failed probes were taped together and seismic testing was completed. Following this, the tape was removed and three of the four probes were subjected to and successfully completed abnormal environment testing and final functional testing. Subsequent to the above, the three probes which were not initially subjected to seismic testing were modified with mechanical fasteners performing the functional requirement of the failed epoxy adhesive. These three mechanically fastened probes then successfully completed seismic testing. The mechanical fastener modification to the probes included the insertion of size 00, 1 inch, 18-8 stainless steel Type U drive screws at 90 intervals around the body of the probe into the 144

ORGANIZATION: NATIONAL TECHNICAL SYSTEMS ACTON, MASSACHUSETTS REPORT INSPECTION l NO - 94Q00917/85-01 RESULTS: PAGE 5 of 5 l top and bottom plates of the probes. In addition similar screws were inserted through the filament housing and the bottom of the probe housing. This modification is documented in Anacon drawing l No. 1700002, Revision D. The effect of this modification is that the structural functions previously performed by the epoxy adhesive is now performed by non-age sensitive mechanical fasteners. Since only test data was reviewed at NTS, review of information relative to Anacon's evaluation and notification to customers of the adhesive failure will be followed up during a future inspection at Anacon. During the abtve 10 CFR Part 21 evaluation, the NRC inspector identified the nonconformance described in paragraph B.2. The inspector further recommended to the QA manager that NTS should consider clarifying their procedures on the issuance of documents; such as, test reports and procedures, so that it is clear as to how revisions to these type documents are generated, reviewed, and approved. i i l i i 145 9

ORGANIZATION: NES MANUFACTURING GREENSBORD, NORTH CAROLINA REPORT INSPECTION 6/10-15/85 INSPECTION N0.: 99901018/85-01 DATE(S): 6/24-28/85 ON-SITE HOURS: 146 CORRESPONDENCE ADDRESS: Nuclear Energy Services Manufacturing ATTN: Mr. Frank Sugar i General Manager 101 Swing Road Greensboro, North Carolina 27409 ORGANIZATIONAL CONTACT: Mr. Lon Ludwig, QA Manager TELEPHONE NUMBER: (919) 852-3400 PRINCIPAL PRODUCT: Pressure Vessels. NUCLEAR INDUSTRY ACTIVITY: Fuel storage racks and shipping and storage canisters. ASSIGNED INSPECTOR:

• dn+ i R. L. Cilimberg, Spfcial Projects Inspection Date Section (SPIS)

OTHER INSPECTOR (S): -J. Conway, Reactive Inspection Section (RIS) i

0. Gormley, Program Coordination Section (PCS)

                            . He m EG G APPROVED BY:             1 7 26/fr$

gphnW.Craig, Chief,\SfIS,VPB 'Date l INSPECTION BASES AND SCOPE: l A. BASES: 10 CFR Part 21 and 10 CFR Part 50, Appendix B. l B. SCOPE: Verify the implementation of the NES QA program during the fabrication of defueling canisters for TMI-2. These QA requirements were specified in Bechtel purchase order TC-016172, Rev. O dated December 4,1984, and Bechtel Technical Specification 15737-2-M-101A(Q), Rev. I dated November 30, 1984 PLANT SITE APPLICABILITY: TMI-2(50-320) 147

ORGANIZATION: NES MANUFACTURING GREENSBOR0, NORTH CAROLINA REPORT INSPECTION NO - QQQO101R/RE 01 RFSt!! TS- PAGE 2 of 18 A. VIOLATIONS: Contrary to Section 21.31 of 10 CFR Part 21, a review of 50 purchase orders (P0), 35 for materials and 15 for services, pertaining to the Three Mile Island Unit 2 (TMI) defueling canisters indicated that while 10 CFR Part 21 was imposed upon NES by Bechtel (ref. P0 TC-016172 dated December 4, 1984), NES P0s to 23 vendors (11 material - P0s 4010, 4009, 3998, 4008, 4012, 4011 4292A, 4292, 4356, 4293, and 4291; 12 service - P0s 4297, 4664, 4657, 4642, 4639, 4632, 4608, 4607, 4337, 4681, 4467, and 4359) did not specify that 10 CFR Part 21 requirements wculd apply. B. NONCONFORMANCES:

1. Contrary to Criterion V of Appendix B to 10 CFR Part 50, Sections 2.1.2 and 5.5.2 of Bechtel Specification 15737-2-M-101A, Section 4.3 of NES Procedure N-10, and Section 3.2.10 of NES Procedure Q-12, receipt inspection was not performed on the following items purchased by Bechtel and shipped directly to the NES facility in Greensboro, North Carolina: neutron poison shrouds, lower and upper closure heads, bulkheads, filter bundles, recombiner catalyst, and DE0X0-D catalyst.

2. Contrary to Criterion V of Appendix B to 10 CFR Part 50, Section 3.2.1 of NES Procedure Q-13, and Section 4.4 of NES Procedure N-15:

a) approximately 10 pieces of 14" diameter nonconforming pipe were not segregated from acceptable pipe in a storage area, and the i nonconforming pipe was not marked with a red tag, and b) nonconforming poison tubes were segregated from acceptable tubes on carts in a storage area without tagging the nonconforming items. l 3. Contrary to Criterion V of Appendix B to 10 CFR Part 50 and welding procedure WPS-001, voltage was not maintained within the specification l limits (10-18 volts) at the lower head welding station for the filter I canister subassembly on five different occasions when a calibrated voltmeter on welding machine S/N 12RT-73449 was reading 8 volts.

4. Contrary to Criterion V of Appendix B to 10 CFR Part 50 and Section 3.4.3 of NES Procedure No. MC-03 a review of 50 P0s, 35 for materials and 15 for services, and the Qualified Source List (QSL) indicated that orders were placed with 18 vendors who were not on the QSL: 13 material vendors and 5 service vendors.

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ORGANIZATION: NES MANUFACTURING GREENSBOR0, NORTH CAROLINA REPORT INSPECTION NO.- 99901018/85-01 RESULTS: PAGE 3 of 18

5. Contrary to Section 6.5 of Bechtel Specification 15737-2-M-101A, items such as boral shrouds, stainless steel, and canisters were observed in outside storage areas without protection from corrosion l and damage.

6. Contrary to Criterion V of Appendix B to 10 CFR Part 50 and Section 4.1.4 of NES Procedure N-7, a review of 50 P0s indicated that orders were placed with 22 material vendors and 10 service vendors, but an audit was not performed on 18 of these 32 vendors (12 material A-Jay Metal Supply, Cambridge Wire Cloth, Southern Spring and Stamping, Automotive Fasteners, Charlotte Valve and Fittings, Dixie Bearing, ENSCO, B&W-Advanced Ceramics, Engineered Plastics, B&B Hose and Rubber, Air Products and Advanced Products and 6 service vendors Wallace Manufacturing, Custom Industries, SAC Tool and Die Shop, K&C Machine, Machinex, and Conam Inspection).

7. Contrary to Criterion V of Appendix B to 10 CFR Part 50, Section 7.2 of Bechtel Specification 15737-2-M-101A, Section 5 of ANSI N45.2, and Section 4.7 of Procedure N-4 of the NES QAM, a review of 50 P0s, thirty-five (35) for material and fifteen (15) for services related to the defueling canister fabrication program, revealed that none of the P0s required the vendor (i.e.,

contractor / subcontractor) to have a QA program consistent with ANSI N45.2 or Appendix B of 10 CFR Part 50.

8. Contrary to Criterion V of Appendix B to 10 CFR Part 50, Section 4.3.2 of Bechtel Technical Specification 15737-2-M-101A, and Section 11.1.1 of Bechtel Specification 15737-G-300, unapproved welding procedure WPS-002 "GMAW-Short Circuiting Transfer" was used on joint No. 4 of the filter canister subassembly for Traveler S/N 4104

9. Contrary to Section 4.3.1.5 of Bechtel Specification 15737-2-M-101A, recombiner elements installed in a number of lower heads were exposed to dirt and grinding particles on the floor of the shop next to the head closure welding operation.

10. Contrary to Criterion V of Appendix B to 10 CFR Part 50 and Sections 3.3, 4.5, 4.7, and 4.9 of NES Procedure MC-04, material was not properly identified, stored, or accounted for as evidenced by:

a) Tags or other identifying means were not properly used such that the heat numbers and other identification were not maintained. l i l 149 l 1

i ORGANIZATION: NES MANUFACTURING GREENSBOR0, NORTH CAROLINA REPORT INSPECTION I NO.- 99901018/85-01 RESULTS: PAGE 4 of 18 b) Integrity of " lots" identified by single tags or other means of identification was not properly maintained in that similar material from different heats was mixed together. c) Heat numbers were obliterated in the manufacturing

               - processes without the installation and maintenance of effective compensating identification measures.

d) Quantities of materials stated on tags and travelers did not match the quantities in the given lots.

11. Contrary to Criterion V of Appendix B to 10 CFR Part 50 and Section 6.1 of Bechtel Specification 13587-G-400, the tape currently in use in the fabrication area was not certified as meeting the limits for halogens and sulfur.

12. Contrary to Criterion V of Appendix B to 10 CFR Part 50 and Section 4.5.2 of NES Procedure Q-10:

a) One Work Order Change Notice issued a drawing revision for use without specifying that the obsolete drawing should be recalled. b) Eight obsolete drawings were found in the Master Work Order drawing file without the required " obsolete" stamp.

13. Contrary to Section 3.1 of NES Procedure N-8, identification numbers etched on the outside diameters of some poison tubes did not relate to applicable documents while other tubes did not contain etched identification.

14. Contrary to Criterion V of Appendix B to 10 CFR Part 50, Sections 5.5 and 5.7 of NES Procedure N-7, Section 4.3 of NES Procedure N-10,-and Section 3.2.6 of NES Procedure Q-12, there was inadequate document-ation to show that receipt inspection was performed on certain items and drawings could not be located to determine whether or not a dimensional check was performed by NES on items machined or formed by vendors (P0s 4297, 4657, 4639, 4607, and 4467); SQHLs were missing for two material vendors (P0s 4404, 4356), and one service vendor (P0 4467).

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ORGANIZATION: NES MANUFACTURING GREENSB0RO, NORTH CAROLINA I REPORT INSPECTION NO.- 99901018/85-01 RESULTS: PAGE 5 of 18

15. Contrary to Criterion V of Appendix B to 10 CFR Part 50, Section 5.5 of NES Procedure N-7, Section 4.1 of NES Procedure N-10, and Section 4.1 of NES Procedure N-17, a review of Supplier Quality History Logs I (SQHL) for receipt inspection activities of material and services l purchased by NES indicated that none of the SQHLs described the type of observation, identified the inspector, or documented the results of the inspection.

16. Contrary to Criterion V of Appendix B to 10 CFR Part 50, Section 3.2.9 of NES Procedure Q-12, and Section 5.2 of NES Procedure N-17, a review of QA records indicated that CMTRs and/or CCs from material

             .uppliers/ manufacturers were missing for items on P0s 4302, 4298, 4639, 5103, and 4467.

17. Contrary to Section 3.3 of NES Procedure N-5, a note was attached to Traveler SN 004096 which had not been approved by the NES QA Department and which provided instructions for identifying Poison Tube Assemblies with poison lot and tray numbers.

18. Contrary to Criterion I of Appendix B to 10 CFR Part 50, the NES QAM, Revision 0 dated April 1984 does not in all cases contain the current information with respect to the authority and duties of persons performing activities affecting quality.

The QAM does not reflect the current organization at NES.

19. Ccntrary to Criterion V of Appendix B to 10 UR Part 50 and Section 3.1 of NES Procedure N-6, " Document Control," of the NES Quality Assurance Manual (QAM), it was noted that NES Policy / Procedure Q-10,

            " Documental Control," dated March 30, 1984 did not contain measures to assure that current procedures are retained in manuals as evidenced by a review of the "Bechtel Canister Program Procedures - 84091" manual. A review of copies of this manual in two fabrication areas revealed that a copy of Procedure No. 15737-2-M101A-00031-02,
            " Packaging and Shipping," dated January 2, 1985 was missing; and superseded copies of weld procedure WPS-001 "GTAW" dated December 15, 1981, and February 7, 1985 were in both manuals.

20. Contrary to Criterion V of Appendix B to 10 CFR Part 50, Section 5.1.1 of Bechtel Specification 15737-2-M-101A, Section 5.6 of NES Procedure N-2. Section 9.5 of SNT-TC-1A, and Section 3.2.5.4 of NES Procedure Q-11, there was no documented evidence that NES had copics of the written practices of Conam Inspection (CI) or Pittsburgh Testing Laboratory (PTL) for all phases of certification of NDE personnel, or that NES hao approved the two written practices.  !

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ORGANIZATION: NES MANUFACTURING GREENSBOR0, NORTH CAROLINA REPORT INSPECTION NO.- 99901018/85-01 REstILTS: PAGE 6 of 18

21. Contrary to Criterion V of Appendix B to 10 CFR Part 50 and Section 3.2.5.4 of NES Procedure Q-11, there was no documented evidence that NES had approved the Level III status of a PTL employee and a Conam employee who had cercified two RT-Level II examiners in April and July 1985. One Level II performed RT on weldments on four occasions from April thru June 1985, and the other performed RT on weldments on seven occasions in May 1985.

22. Contrary to Criterion V of Appendix B to 10 CFR Part 50 and Sections 2.2 and 4.5.1 of NES Procedure Q-04, it was noted that.the QA Manager and four QC inspectors had passed a written examination for certification, but the certification form had not been completed.

C. OPEN ITEMS: None. D. OTHER FINDINGS AND COMMENTS:

1. Review of Radiographs Radiographic films of the longitudinal welds which required repair were reviewed. These welds are in the 304 L stainless steel pipe supplied by Armco Steel, Wildwood, Florida to be used for canisters.

The North Carolina state authorized inspector (AI) had previously reviewed the radiographs and written "LOF" (lack of fusion) in pencil on several reader sheets and film envelopes. The NRC inspectors questioned whether the Al had rejected the welds. The Al stated that the copy of the reader sheet was written for his information and that he had not rejected any of the Armco radiographs. Review of the radiographs on pipe identified by S/N 28PI indicated a weld repair area between radiographic stations 12 and 13. The weld repair was exhibited on the film as a lighter area than the adjacent original automatic fusion weld. Because the repair weld was not identified on the film or the reader sheet, a visual inspection was made by the NRC' inspectors (of the weld on the pipe which had been fabricated into a fuel canister) to determine if the weld had been repaired. The visual inspection indicated that no weld repair had been made in the logitudinal seam that was exhibited on the radiograph. However, there was a small 2 inch repair weld that ran diagonally across the seam weld near station 13 that was not exhibited i 152

ORGANIZATION: NES MANUFACTURING GREENSB0RO, NORTH CAROLINA REPORT INSPECTION NO.- 99901018/b5-01 RESULTS: pAGE 7 of 18 by the radiograph. The absence of a radiograph that was represen-tative of the weld seam between stations 12 and 13 is in nonconfor-mance with Subsection UW-51(2) of Section VIII, Part UW (lethal) of , the ASME Code. Visual inspection of the weld surface on the inside l of the pipe was not possible because the canister was lined with concrete. All of the Armco radiographs of longitudinal welds which had been repaired were reviewed to determine if the above finding was an isolated case. Approximately 13 weld repair areas that were noted on the radiographs were compared to the welds on the pipes and found to correlate, based upon this review the above finding was determined to be an isolated case. Radiographs of lower head welds were not reviewed because the AI rejected the welds of the lower heads due to stepping (distortion) between the head and the shell resulting from welding, i After the NRC inspection June 10-15, 1985, Bechtel advised the NRC inspector by telephone conversation on June 18, 1985, that the weld repair between stations 12 and 13 had been made by fusion welding at Armco on the inside surface of the pipe, Bechtel stated that this was the reason why the NRC inspectors did not observe the repair on . the outside surface. The NRC inspectors concluded that if the repair was made by fusion welding on the inside surface, then the defects were not removed by mechanical means or thermal gouging which would have required the use of filler metal. The latter methods of weld repair are required by the ASME Code for pressure vessels. The method of weld repair used by Armco was rot in accordance with Subsection UW-38 of Section VIII, Part UW (lethal) of the ASME Code and the canister cannot be code stamped for use as a pressure vessel without additional review of the Armco method of weld repair.

2. NES Organization The NES Quality Assurance Manual (QAM) exhibits an organizational structure that was in effect during April 1984. The QAM has not been revised to reflect the reorganization in April 1985 (See i

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ORGANIZATION: NES MANUFACTURING GREENSBOR0, NORTH CAROLINA REPORT INSPECTION NO.- 99901018/85-01 RESULTS: PAGE 8 of 18 Nonconformance B. 18). The purchase order for the canisters was awarded on November 30, 1984, and there have been three QA Managers since December 1984, indicating significant changes to the NES organization without a corresponding update of the QAM.

3. Poison Tube Assembly During the setup of two inch diameter tubes for loading of B,C pellets, the NRC inspectors observed that some tubes were identified with etched numbers while some were not, and one tube exhibited a gouge on the outside surface which was filled with rust. When questioned, the NES " expediter" produced a note to explain how the tubes should be identified with prison lot and tray numbers. The note was not properly approved (See Nonconformance B.17). Documents were not available to explain identification of the tubes (See Nonconformance B.13). When the NRC inspector identified the nonconformances to the NES machine shop foreman, he stopped work on the tubes. The gouged tube was placed on a cart without properly tagging the tube (See Nonconformance B.2).

4. Recombiner Contamination The NRC inspectors observed that recombiner elements had been installed in a number of lower heads and partially covered with plastic. The lower heads were stacked against the wall next to the automatic welder used to weld the lower heads to the shells. Large quantities of grinding particles and dirt were observed on the plastic and the plastic was not sealed to prevent the debris from sliding through the folds of the plastic and contaminating the recombiners (see Nonconformance B.9).

5. Corrosion and Damage of Stored Material

During the tour with the NES General Manager and others, the NRC

! inspectors observed that canisters and in process materials / components l were stored outside which exposed this material to the weather, dirt, i and other potential debris and damage. This was discussed with the l NES General Manager and prior to the end of the inspection, end caps , were installed on pipes and canisters, parts were placed on wooden l skids and boxes containing boral shrouds were repaired (See Nonconformance B.5).

6. Procurement Document Control 50 P0s, 35 material vendors and 15 service vendors, were reviewed to determine whether or not applicable regulatory, technical, and 154

ORGANIZATION: NES MANUFACTURING l GREENSBOR0, NORTH CAROLINA REPORT INSPECTION NO.- 99901018/85-01 RESULTS: PAGE 9 of 18 QA program requirements were included or referenced in procurement documents. Bechtel had imposed the requirements of 10 CFR Part 21

              " Reporting of Defects and Noncompliance" and the QA Program l

' Requirements in ANSI N45.2 upon NES in Technical Specification 15737-2-M-101A for fabrication of defueling canisters. The NRC inspectors found that NES had failed to pass the requirements of 10 CFR Part 21 to 23 of 35 vendors, 11 material vendors (P0s 4010, 4009, 3998, 4008, 4012, 4011, 4292A, 4292, 4356, 4293, and 4291) and 12 service vendors (P0s 4297, 4664, 4657, 4642, 4639, 4632, 4608, 4607, 4337, 4681, 4467, and 4359) (See Violation A). In addition, NES failed to pass on QA program requirements to 22 material vendors and 10 service vendors for the 50 P0s reviewed (See Nonconformance B.7). QA personnel had initialed all of these P0s with the exception of P0 3948 dated November 16, 1984 to Carolina Steel and P0 4337 dated January 25, 1985 to K&C Machine.

7. Control of Special Processes The NRC inspectors reviewed applicable portions of the NES QAM and three NES procedures to determine whether special processes were being conducted by qualified personnel using qualified procedures and equipment. A review of 20 travelers relating to welding (9-filter canister subassemblies and 11-fuel canister subassemblies) revealed that all individual operations were properly initialed or stamped and dated. In addition, the hold points for witnessing by the Bechtel site inspector were signed or initialed and dated. The traveler package also contained a weld map which showed a sketch of the particular joint and the weld procedure to be used. A " Welder and NDE Record Sheet" identified the welder and procedure used as well as the NDE examiner. An attached " Liquid Penetrant Inspection Report" also identified the PT procedure as wel.1 as the examiner.

For Traveler S/Ns 4111, 4110, 4113, and 4112, the inspector noted that the PT Inspection Report identified Procedure QIP-PT-V as containing the acceptance standards, but a review determined that QIP-PT-V, " Penetrant Inspection" does not contain acceptance standards. For Traveler S/N 4104, it was noted on the " Welder and NDE Record Sheet" that welder No. 20 welded joint No. 4 using WPS-002 "GMAW-Short Circuiting Transfer". WPS-C02 is not an approved welding procedure for the defueling canister program (See Nonconformance B.8). l 155 i i

ORGANIZATION: NES MANUFACTURING GREENSB0RO, NORTH CAROLINA REPORT INSPECTION NO.- 99901018/85-01 RFSULTS: PAGE 10 of 18 l While inspecting the plant fabrication area and witnessing various l ongoing activities, the NRC inspectors observed, on five different occasions, that a lower voltage (8 volts) rather than the value specified (10-18 volts) in welding procedure WPS-001 was being used for the gas tungsten arc welding of the lower head for the filter canister subassembly (See Honconformance B.3). The inspectors also noted that two manuals containing procedures utilized by the individuals in the shop area were not current (See Nonconformance B.19). Qualification records for 10 welders who had worked on the defueling canisters (job No. 89071) were reviewed. With the exception of welder No. 20, nine welders were qualified to weld using procedure WPS-001 "GTAW" and/or WPS-004 "GMAW." The qualifications were signed off by a welding engineer (consultant) and the QA Manager. A Qualification Maintenance Log, updated every three months by the QA Department, documented that each welder and welding operator had welded using a process to maintain their qualification in accordance with Section IX of the ASME Code. A review of three Procedure Qualification Records (PQR) for WPS-001 and one PQR for WPS-004 and four test reports indicated that all testing had been performed as required in accordance with Section IX of the ASME Code by Law Engineering Testing Company. The disposition of welding filler metal appeared to be in compliance with procedure Q-07 " Welding Rods, Electrode and Filler Metal Control."

8. Training / Qualifications The NRC inspectors reviewed applicable sections of the QAM, one procedure, and training records from 1983 to the present for 20 employees (1-manufacturing, 1-materials control, 5-engineering, 1-plant manager, 1-expediting, 11-QA/QC) to determine whether personnel performing and verifying activities affecting quality had received the necessary training and qualifications.

Qualification Records of four personnel who conducted audits of vendorr, in the capacity of a lead auditor were evaluated. The audits were performed from October 1979 thru May 1985. The review of these records indicated that all of the auditors were qualified and met the requirements of Procedure Q-02 " Qualification of Audit Personnel" which describes the requirements for the qualification of personnel conducting audits both within and outside NES. l l 156 i

I l ORGANIZATION: NES MANUFACTURING GREENSBOR0, NORTH CAROLINA REPORT INSPECTION NO.- 99901018/85-01 RESULTS: PAGE 11 of 18 Training was given in various disciplines of the QAM and quality procedures. QC inspectors performing inspections, examinations and tests were required to pass a written examinaticn and were l certified as Level I, II, or III. An annual evaluation is also performed. In the case of the QA Manager and four QC inspectors, a certification form had not been completed even though it was apparent that they had passed the examination (See Nonconformance B.22).

9. Control of Purchased Material and Services The inspector reviewed the data packages for the material purchased by Bechtel and shipped directly to the NES facility in Greensboro, North Carolina. Bechtel purchased items included the neutron poison shrouds from Brooks & Perkins, pipe from Armco, filter bundles from Pall Trinity Micro Corporation, upper and lower heads and bulkheads from Jessop Steel and Hackney Iron Works, and catalyst material from Atomic Energy of Canada and Engelhard Corporation. The data packages, which consisted of certified material test reports, (CMTRs), certifi-cate of compliance, (CC) NDE reports, and hydrostatic test reports, where applicable, were furnished by Bechtel to NES. Bechtel performed source surveillance audits for all the items with the exception of the recombiner catalyst and the DE0X0-0 catalyst.

Although NES verified the number of items received, receipt inspection was only performed on the 14 inch diameter pipe fabricated by Armco. The required receipt inspection was not performed by NES on items purchased by Bechtel and shipped directly to NES. Specifically, receipt inspection was not performed by NES on items such as neutron poison shrouds, lower and upper closure heads, bulkheads, filter bundles, recombiner catalyst, and DE0X0-0 catalyst. (See Nonconformance B.1). Activities be performed as part of receipt inspection and identified on a Detail Traveler included (1) a dimensional check for pipe (0D, ID, straightness, and length), (2) visual examination for cracks and damage, and (3) verification of documents (e.g., CMTR, test reports). The three activities were to be verified by the Bechtel site representative. Six shipments of pipe were made to NES in November and December 1984. A review of five travelers indicated that 208 pieces of pipe had undergone a dimensional check, but there was no indication of the serial number (S/N) for the measuring equipment, and none of the travelers was initialed / stamped by either the NES inspector or the Bechtel site representative. Also, a traveler for I 157

ORGANIZATION: NES MANUFACTURING l GREENSBOR0, NORTH CAROLINA REPORT INSPECTION i NO.- 00nn1018/85-01 RESULTS: PAGE 12 of 18 the remaining 53 pieces of pipe could not be located. After these NRC findings were discussed, NES initiated Nonconformance Reports ' 179 (in process pipe) and 180 (raw pipe) on June 22 and June 23, 1985, respectively, noting that the pipe was not fully receipt inspected. As of June 28, 1985, NES was awaiting Bechtel's formal notification regarding the disposition of the nonconformances. For NES purchased material and services, the NRC inspectors reviewed applicable sections of the QAM, two procedures, the Qualified Source List (QSL) and external audits of vendors to determine whether material and services were purchased from qualified vendors. A review of 50 P0s indicated that orders were placed with 22 material vendors and 10 service vendors (5-machining, 1-forming, 1-calibration, and 3-testing) from November 1984 through April 1985. Four of the material vendors were ASME certificate holders. NES had audited three vendors - J. T. Ryerson in October and November 1984, Piedmont Hub in October 1979, and Carolina Steel in October 1982. Four service vendors were audited - Pittsburgh Testing and American GFM Corporation in January and May 1985 and both Gage Laboratory (GL) and Law Engineering & Test in May 1983. Thirteen of the material vendors and five service vendors had not been audited by NES, and they were not on the NES QSL (See Nonconformances B.4 and B.6). During a review of CMTRs and CCs from material suppliers and manufacturers, the NRC inspectors determined that these documents were missing on five P0s (See Nonconformance B.16). The NRC inspectors reviewed drawings and SQHLs for 11 items (P0s 4297, 4280, 393G, 4664, 4657, 4639, 4607, 4404, 4356, 5103, and 4467) to determine whether receipt inspections had been performed. Six items were machined by outside vendors, but drawings to assure that a dimensional check was performed by NES were missing on five of the items. SQHLs were missing for three other items (See Nonconformance B.14). The SQHL is the document used to establish that a receipt inspection has been performed, identify the P0, quantity received, acceptance and the type of inspection performed (i.e., dimensional, cleanliness, P0 requirements). It does not identify the inspector, list S/Ns of measuring equipment, or describe the type of observation and the specific results (See Nonconformance B.15). In addition NES inspections were conducted without a checklist.

10. Control of Measuring and Test Equipment (M&TE)

The NRC inspector reviewed applicable sections of the QAM, one procedure and calibration records to determine whether M&TE was properly identified, controlled and calibrated at specified 158

ORGANIZATION: NES MANUFACTURING GREENSBOR0, NORTH CAROLINA REPORT INSPECTION NO.* 99901018/85-01 RESULTS: PAGE 13 of 18 intervals. A review of the Tool Check Log indicated that issuance of M&TE from the gage crib is controlled. Fifteen Gage Maintenance Cards for inspection equipment (6-micrometers calipers, and thread measuring wires); instruments (5- pressure gages, voltmeters, and ammeters); and reference standards (4-gage block set, dead weight tester, and panel meter) were reviewed. The cards identified S/Ns of the instruments, calibration date, and calibration frequency. -Information contained on calibration stickers on each item was in agreement with the applicable gage card. Internal calibration activities at NES are handled by one individual. Gage Laboratory (GL) performed calibration services on several items: 20 piece weight set, precision level, pyrometer, gage block set, panel meter, hardness tester, and dead weight tester. A review of 10 applicable certifications from GL indicated that all standards utilized were traceable to the National Bureau of Standards. The most recent audit performed by NES of GL was in May 1983.

11. Pipe In addition to the two NES nonconformance reports for the 14 inch diameter pipe (See Section 0.9), another nonconformance report was generated by NES when a crack was detected on pipe No.

103P2. The crack was in the weld between stations 11 and 12, and subsequent PT examination of the pipe's OD and ID revealed that the crack on the OD surface was ;ot a through wall defect.

This pipe was one of 10 pieces of pipe that were identified as not being stored in a segregated area by the NRC inspector during the June 10-14, 1985 inspection (See Nonconformance B.2). As of June 28, 1985, the status of the 261 pieces of pipe is as follows: 140-in storage, 8-rejected for dimensional noncompliance, 107-in process (29-filter /49-fuel /29-knock out), and 6-sent to Joseph Oat Company.

12. Nondestructive Examination (NDE)

The NRC inspector reviewed the qualification and certification records of NDE personnel from NES (9), PTL (2), and CI (1) to determine whether the individuals performing ultrasonic (UT), radiographic (RT) and/or liquid penetrant (PT) testing were certified to SNT-TC-1A. 159 _ J

i ORGANIZATION: NES MANUFACTURING GREENSB0RO, NORTH CAROLINA REPORT INSPECTION NO.- 99901018/85-01 RESULTS: PAGF 14 of 18 The NES QA Manager was certified to a Level III for visual examination (VT), RT, PT, magnetic particle examination (MT) by the General Manager, but the certificatica was not dated. The QA Manager had successfully passed written examinations in July l'980 for PT, RT, & MT and in March 1985 for VT. With the exception of the QC Supervisor whose certification was missing, the NDE records for the remaining seven individuals appeared to be in order. An April 1985 memo from NES's QA Manager designated R. Dovicsak from PTL .ss NES's Level III examiner in MT, UT, RT and PT. PTL had certifiel Dovicsak in May 1984 to a Level III, but copies of his tests were not available. An eye examination dated November 1983 was in the file, and a eye examination report dated January 1985 was added to the file during the NRC inspection. The NRC inspectors reviewed NES's written practice for all phases of certifying NDE personnel, which appears to be consistent with SNT-TC-1A. However, NES did not have copies of PTL's or CI's written practices nor had they approved the two written practices (See Nonconfortrance B.20). The NES procedures for VT and PT were approved by Bechtel in December 1984 and February 1985, respectively. NES approved PTL's procedures for UT and RT in January 1985 with a stipulation that Bechtel approve them prior to use. Although Bechtel approved both procedures on June 12, 1985, Dovicsak of PTL performed UT of weldments on May 10 and 14,1985 and June 3, 1985. A. Morrision, a Level 11 from PTL, performed RT at NES on April 17 and 22, 1985 and June 6 and 8, 1985. In addition, J. Miller, a Level II from CI, performed RT on weldments using PTL's procedure on seven occasions in May 1985. A. Morrison was certified by B. Bruce and J. Miller was certified by D. Fister, but there was no documented evidence that NES had approved the Level III status of Bruce or Fister, therefore, the Level II .itatus of Morrison and Miller was, in effect, not valid at the time they performed the tests (See Nonconformance B.21).

13. Drawing Control The NPC inspectors reviewed the procedures and functioning of the drawing control process. They determined that the Master Work Order folder contained obsolete drawings related to the canister program which were not stamped " obsolete" as required (See Nonconformance i

160

ORGANIZATION: NES MANUFACTURING GREENSB0RO, NORTH CAROLINA REPORT INSPECTION NO.- 99901018/85-01 RESULTS: PAGE 15 of 18 l B.12). Further, one drawing was issued for use without specifying that the obsolete drawing should be recalled. However, inspection of the Engineering Control Clerk's drawing file revealed that the l obsolete revision had been removed. The inspector also noted that drawing No. 1154027 Rev. 3 had been released for fabrication by Bechtel without the required signatures in the revision block by Babcock and Wilcox who made the changes.

14. Material Control The NRC inspectors reviewed the procedures and the functioning of the material control system and determined that, while the system relies on the use of green or yellow " accepted" tags or label markings to maintain identification and status of materials, parts and corponents, the actual use of such tags or labels was limited both in scope of application and in effectiveness. In addition when a single traveler or tag was used to indicate the status of a " lot" (i.e., group) of material, parts or components, the controls provided on the envelope of the " lot" were insufficient to prevent mixing of material between lots with the resulting loss of hea+ numbers and other identity, and loss of any indication of status which was on the traveler or tag. Finally, there was no evidence of adjustments of traveler quantities when material was removed from a " lot". This resulted in mismatches between quantities of material in given " lots" and the corresponding quantities stated on the travelers or tags when such travelers or tags were present.

(See Monconformance B.10). Examples of the failure to adequately control material, parts or components include:

1. Fifty-two eight inch and 56 fourteen inch canister skirts were in separate piles on two pallets each with a single acceptance tag. Heat number identities were not available. Neither was there effective envelope control to prevent addition or subtraction of material from the piles.

2. Fourteen knockout canister intermediate "A" support plates were stored in a box. No' status tag or other identification was available. Machining had taken place which obliterated or removed sections of heat numbers.

3. Twenty knock out canister intermediate support plates were stored in and around two fiber drums and a cardboard box.

The fiber drums also contained rain water. There were no tags or other accompanying identification or indication of status. The lids of the drums had heat numbers written on l 161 ! 1 l

ORGANIZATION: NES MANUU.CTURING GREENSB0R0, NORTH CAROLINA REPORT INSPECTION Nn - QQQn101R/RR_n1 RF91tI TS - PAGF 16 of 18 them but had become separated from the drums. Heat numoers which had been written on the plates were obliterated in whole or in part by previous machining operations. Where parts of heat numbers were still legible, the inspector determined that the contents of the two drums had been mixed. 4.- Fourteen partially machined knockout canister upper heads were stored on a pallet on a gravel surface trapping spattered gravel and rain water in blind holes. There was no indication present as to the status of these components.

5. A pallet outside the tool room door, containing 47 lower canister head assemblies, was indicated by the welder as the place where he had been instructed to obtain lower heads for assembly to canisters. There was no indication of acceptance status or other information such as a traveler on or around the pallet, although the welder said there had been a tag there at one time.

6. Next to the fit up and weld station there were three pallets each with a a tag identifying the pallet load as a " lot" of 30 canister lower head assemblies. One pallet had 34, the second had 27 and the third had 29 assemblies.

7. Four boxes of knockout canister entrance tube components were in the assembly area in various stages of welding.

The tag in the box of 11 unassembled bent tubes noted a quantity of 140 pieces. The other boxes were unmarked and without tags and contained tubes tack-welded into double and triple bends.

15. Control of Materials Used in Contact with Canisters and Sub Components Because of the potential for l' ngo term corrosion and malfunctioning of the catalytic recombiner catalysts, the inspector made a review of the materials used in the fabrication and testing process. CMTRs and other appropriate documentation were reviewed for marking materials, the water system, cleaning solutions, liquid penetrant.

l 162

ORGANIZATION: NES MANUFACTURING GREENSBOR0, NORTH CAROLINA REPORT INSPECTION NO.- 4990101R/85-01 RESULTS: PAGE 17 of 18 etc. No adequate documentation or certification could be shown for the " tuck" type tape used to fasten travelers to the canisters and subassemblies, and for other purposes in the canister program. The QA Engineer produced a certificate of compliance dated 1984 for 24 rolls of tape and stated that the tape was kept in a locked cabinet in the QA room. The cabinet contained no tape. Neither was the more-than-6-months-old certificate of compliance for 24 rolls of tape consistent with the usage of tape currently in evidence. A person using tape stated that he obtained it from the tool room. The QA Engineer later stated to the inspector they had run out of certified tape and had submitted a nonconformance report to engineering for evaluation. This nonconformance report was not reviewed during the inspection. In addition, there were no CCs for adhesive-backed accept / reject labels which were called for in the material identification and control procedure (See Nonconformance B.11).

16. Canister Cleanliness After observing the practices in the shop and having conversations with workers, the inspectors reviewed the procedures for storage and cleaning. No requirements for storage of material during fabrication were identified. The shop practices were found to be in compliance with the approved cleaning procedure. However, the approved procedure for cleaning of material and parts, in process, and for internals of the canister assemblies does not meet usual standards for components in contact with primary coolant or for stainless steel components where extended lifetime is desired. For example, material in process is stored outside - some in a gravel surfaced area - or inside where other operations such as weld preparation and other grinding take place. There is no requirement for cleaning anything except weld areas and the external surfaces of the completed canisters (i.e., a requirement for removing grinding chips and dust, adhesive residue or other foreign material on and in internal subassemblies or other components installed inside the canisters does not exist). Similarly, there is no requirement for cleaning the internal surfaces of the canisters prior to installation of the internals. Based upon observations and discussions with NES personnel, the inspectors determined that internal surfaces were not being cleaned prior to assembly. In addition, the approved NES/Selamco cleaning procedure 84091 CP, applicable to weld areas and exterior canister surfaces, allows for 15 square inches of rust in any one foot of surface to which cleaning requirements apply.

l ! 163

ORGANIZATION: NES MANUFACTURING GREENSBOR0, NORTH CAROLINA REPORT INSPECTION l NO.- 99901018/85-01 RESill.TS: PAGE 18 of 18 l l

17. Installation of Catalytic Recombiner and Verification Although no activities related to the installation of the catalytic recombiner were being conducted at the time of the inspection, the inspectors reviewed the procedures for installation and inspection of the recombiner and determined the status of recombiner installation. All bottom head assemblies which include recombiner installation had been completed, inspected, accepted, and stored in folded plastic in the fabrication area. Because of unrelated fabrication problems only four of the upper head assemblies had been completed.

The procedure for filling the cavities with recombiner appears to lack internal consistency. Although it calls for obtaining premixed, premeasured, and presumably carefully controlled packages of recombiner from the store room and then putting the contents of the packages into the cavities, it also calls for adding additional recombiner from an unidentified source to fill the cavity if additional material should be required. The inspector noted the Bechtel inspection rate of 1 in 5 head assemblies for verification of recombiner installation. A future inspection will evaluate the controls applied to the activities of receipt, mixing, packaging, issue, and installation of recombiners. 164

ORGANIZATION: NUCLEAR PARTS ASSOCIATES (NUPA) ST. FRANCISVILLE, LOUISIANA REPORT INSPECTION INSPECTION NO.: 99901010/85-01 DATE(5): 4/24/85 ON-SITE HOURS: 08 CORRESPONDENCE ADDRESS: Nuclear Parts Associates (NUPA) P.O. Box 1399 St. Francisville, Louisiana 70775 ATTN: Mr. James I. Miller I Site Operations Manager ORGANIZATIONAL CONTACT: Mr. John J. Zullo - QA Manager TELEPHONE NUMBER: (504) 635-4875 I PRINCIPAL PRODUCT: River Bend Station Spare Parts Procurement and Warehousing Services NUCLEAR INDUSTRY ACTIVITY: Approximately 99% on ASSIGNED INSPECTOR: _ J

                                 /                                                 (/5 95 '

J. Petrosino, Reactive Inspection Section (RISV 4 ate OTHER INSPECTOR (S): L. D. Vaughan, Program Coordination Section (PCS) APPROVED BY: [ s' E. W. Merschof'f . let, RIS Date INSPECTION BASES AND SCOPE: A. BASES: 10 CFR Part 21 B. SCOPE: 1. Determine what percentage of spare parts are supplied by manufacturers located outside of the United States.

2. Review approved vendor qualification methods.

3. Review dedication process of commercial grade items.

PLANT SITE APPLICABILITY: River Bend Station (50-458) l 165

\ ORGANIZATION: NUCLEAR PARTS ASSOCIATES (NUPA) ST. FRANCISVILLE, LOUISIANA l l REPORT INSPECTION i NO. 99901010/85-01 RESULTS: PAGE 2 of 4 A. Inspection Issues:

1. Determine NUPA's involvement with components supplied by foreign vendors.

2. Review the adequacy of NUPA's qualified vendor audit / survey process.

3. Revlew NUPA's safety related component dedication process.

B. Inspection Findings:

1. NUPA has less than one percent involvement with foreign vendors.

2. No deficiencies were noted in NUPA's vendor qualification process.

3. Dedication and upgrading of commercial grade to safety related is not within the scope of NUPA's contractual requirements.

C. Supplemental Information: The NRC inspection was conducted at the Gulf States Utilities Company (GSU) corporate offices, Beaumont, Texas, and at the Nuclear Parts Associates offices, St. Francisville, Louisiana. NUPA is a recently formed joint business venture between General Electric Company, San Jose, California, and Stone and Webster, Cherry Hill, New Jersey. NUPA has been contracted by GSU for River Bend Station spare parts procurement and warehousing services. NUPA and GSU are currently discussing the possibilities of collective spare part pooling with other nuclear plant licensees. However, at this time NUPA's only client is GSU at the River Bend Station. GSU has two (2) agreements concerning their purchases of material, com- , ponents, and services. One agreement is with the State of Louisiana and the other is with Cajun Electric Company, (a River Bend Station stockholder). Cajun Electric Company's agreement requires, in part, that GSU will buy American products / services unless they cannot be adequately obtained domestically.  ; I l l l 166 1

i ORGANIZATION: NUCLEAR PARTS ASSOCIATES (NUPA) ST. FRANCISVILLE, LOUISIANA REPORT INSPECTION NO. 99901010/85-01 RESULTS: PAGE 3 of 4

  !       The State of Louisiana agreement requires in part, that GSU purchase from Louisiana vendors, unless they cannot be adequately obtained from l

Louisiana sources. Minimal foreign vendor activity was observed or is anticipated, as evidenced by the presence of only two foreign vendors on the GSU/NUPA qualified supplier list (QSL). D. Background Information:

1. A review of GSU's qualified supplier list from 1982 to present and NUPA's QSL was performed.

Only two vendors which were located outside of the U.S. were identified. Both vendors are located in Canada. Velan Engineering, (supplies valves and accessories) and Guildline Precision (supplies instruments and calibration services).

2. The NUPA QA procedures for quality auditing does not provide for any difference between the audit requirements of domestic or foreign

.f suppliers.

Both GSU and NUPA audit requirements appear adequate fer overall program evaluations. NUPA requires that audit plans be "...deve-loped from applicable procedures, specifications, instructions, codes, and regulatory requirements". Discussions with GSU and NUPA personnel indicated that the same degree of audit effort is performed at domestic and foreign vendors.

3. Document reviews and discussions with cognizant NUPA and GSU personnel, concerning dedication of commercial grade components for nuclear use, were performed.

Currently, NUPA does not plan to be involved in any safety related component dedication. GSU will continue to control all commercial grade components which are to be dedicated as safety related. E. Comments:

1. The NUPA organization at the River Bend Station is not in full operation at this time. Contractual agreements between GSU and i NUPA, for the spare part procurement and warehousing, were not 167

, ORGANIZATION: NUCLEAR PARTS ASSOCIATES (NUPA) l ST. FRANCISVILLE, LOUISIANA i REPORT INSPECTION NO. 99901010/85-01 RESULTS: PAGE 4 of 4 finalized until earlier this year. Therefore, implementation of the NUPA program could not be fully evaluated. F. Persons Contacted: Tom Crouse GSU Manager - QA Brandon Bryon GSU Director - Task Projects Lyle P. Gerac GSU Director - Contracts John H. Curless GSU Manager - Project Controls James E. Booker GSU Manager - Engineering & Licensing Lester R. Sutton GSU-RBS Q.A. Engineer Garland Mahan CSU-RBS Q.A. Engineer Jim Miller NUPA Manager - Operations John Zullo NUPA Q.A. Manager R. E. Hebert GSU-RBS Senior Purchasing Agent ' G. E. Kelley GSU-RBS Supervisor - Purchasing Materials , 1 ) l l 168

ORGANIZATION: NUTHERM INTERNATIONAL INCORPORATED (NI) MOUNT VERNON, ILLIN0IS REPORT INSPECTION INSPECTION N0.: 99900779/85-01 DATE(S): 4/29/85 ON-SITE HOURS: 6 CORRESPONDENCE ADDRESS: Nutherm International Incorporated ATTN: Mr. James S. Hanner Executive Vice President j 501 South Eleventh Street i Mount Vernon, Illinois 62864

ORGANIZATIONAL CONTACT: Mr. Ronald J. Heifner, _QA Manager TELEPHONE NUMBER: (618) 244-6000 PRINCIPAL PRODUCT: Electrical Control Components ard Systems NUCLEAR INDUSTRY ACTIVITY: 99% of total work is nuclear related, which includes military, commercial power, and research work. Current in house job orders are Beaver Valley 2 (NI-1323), Nine Mile Point (NI-1344), and Knolls Atomic (NI-1372/1385).

I 1 1 M ASSIGNED INSPECTOR: [ - 8 !85 J jl Petrosino, Reactive Inspection Section (RIS')/ Date OTHER INSPECTOR (S): APPROVED BY: - 6!7 E. W. Merschofff)def, RIS, Vendor Program Branch Date i INSPECTION BASES AND SCOPE: A. BASES: 10 CFR Part 21. B. SCOPE: 1. Obtain vendor-licensee corresponderce regarding Elma Engineeririg Incorporated (EEI) ferroresonant power supplies (FPS). ! 2. Verify where each eel-FPS was shipped and the total ! number of units at each location. PLANT SITE APPLICABILITY: . Pilgrim (50-293); Browns Ferry (50-259/260/296); Clinton(50-461). l l 169

Y ORGANIZATION: NUTHERM INTERNATIONAL INCORPORATED MOUNT VERNON, ILLIN0IS REPORT INSPECTION NO 99900779/85-01 RESULTS: PAGE 2 of 3 A. Inspection Issues:

1. Verification of correspondence regarding Nutherm International (NI) )

customers who were shipped safety-related EEI ferroresonant power j supplies regarding potential problems with these power supplies.

2. Obtain information regarding the total number of EEI - ferroresonant power supply units shipped.

B. Inspection Findings:

1. Three (3) commercial nuclear plants were supplied EEI - FPS units by NI. Correspondence was transmitted by NI to each customer who pur-chased the EE! units regarding the EEI power supply problems.

Additionally, each customer acknowledged receipt of these NI notifications.

2. A total of fifty one (51) safety-related EEI power supplies were purchased by NI. Of that total, thirty two (32) units were shipped to Browns Ferry station, four (4) units went to Clinton station, and eight (8) units went to Pilgrim station. NI retained seven (7) units for equipment qualification testing.

C. Supplemental Information: Recently, Elma Engineering Incorporated Ferroresonant power supply units were identified as having two potentially generic problems. NRC-IE Information Notice Number 83-04 discussed the loss of output of EEI's - FPS due to overheating and subsequent capacitor failure. EEI reported to the NRC that their testing concluded that improper vertical mounting of FPS units could result in capacitor overheating and sub-sequent degraded operation. Additionally, FPS units utilized in a parallel load sharing application will have a greater heat output and therefore will increase the FPS unit enclosure temperature. Recom-mendations of EEI were to-perfonn analyses for load sharing, adequate cooling and efficiency at low loads. The second potentially generic problem was identified in a 10 CFR part 21 notification letter submitted by NI. Deficiencies identified were found by removing a mounting plate which enclosed the electrical wire connection area. Inadequate electrical wire soldering connections and a leaking oil filled capacitor were observed by NI. Routine receipt inspections by NI had not included removing the mounting plate to examine electrical wire connections. This was the subject of a nonconformance issued to Nutherm in Inspection Report 99900779/84-01. 170

t G' 'j ORGANIZATION: NUTHERM INTERNATIONAL INCORPORATED j MOUNT VERNON, ILLIN0IS u l REPORT INSPECTION j NO.:99900779/85-01 RESULTS: PAGE 3 of 3 D. Persons Contacted: l i James S. Hanner - Executive Vice President - NI Bill E. Ellis - Quality Assurance Engineer .- NI - i Ronald J. Heifner - Quality Assurance Manager - NI A 4 i i l i

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ORGANIZATION: POWER CONVERSION PRODUCTS INC. CRYSTAL LAKE, ILLIN0IS REPORT INSPECTION INSPECTION N0.: 99900741/85-01 DATE(S): 5/22-23/85 ON-SITE HOURS: 8 CORRESPONDENCE ADDRESS: Power Conversion Products Inc. Forty-two East Street P.O. Box 380 Crystal Lake, Illinois 60014 l ORGANIZATIONAL CONTACT: C. F. Seyer, Vice President TELEPHONE NUMBER: (815) 459-9100 i PRINCIPAL PRODUCT: Battery Chargers NUCLEAR INDUSTRY ACTIVITY: Current Nuclear activity is approximately 5%. l l' l' ASSIGNED INSPECTOR: Vk6.J(. [. v K, R. Naidu, Reactive Inspection Section (RIS) 7//[6~ Date OTHER INSPECTOR (S): APPROVED BY:

                             '                                                  7       rr E. W. Merschoff Chief, RIS                                Date INSPECTION BASES AND SCOPE:

A. BASES: 10 CFR Part 21 and 10 CFR 50 Appendix B. B. Review the corrective action taken on identified 10 CFR Part 21 items, implementation of the QA program in selected areas and observation of work in progress for refurbishing battery chargers for the South Texas l Project. PLANT SITE APPLICABILITY: LaSalle County Stations 1 & 2 (50-373, 50-374) l South Texas Project 1 (50-498). 173

ORGANIZATION: POWER CONVERSION PRODUCTS INC. CRYSTAL LAKE, ILLIN0IS REPORT INSPECTION NO - 99900741/85-01 RESULTS: PAGE 2 of 5 A. Inspection Issues

1. Illinois Power Company, the owner of Clinton Power Station, notified the NRC on October 19, 1984, of a potential defect in the wire-connector crimping in the battery chargers manufactured by Power Conversion Products (PCP).

2. PCP notified the NRC on May 13, 1982, of a potential defect based on information received from LaSalle County nuclear power station.

Specifically, the 200 ampere fuses in the battery chargers type 3S-130-200 were failing due to overheating.

3. Brown and Root (B&R), the then Architect Engineer for the South Texas Project, notified the NRC on October 7, 1980, of a potential defect in the battery chargers supplied by PCP. Specifically, B&R stated that some components in the battery chargers were different than those specified.

B. Background Information PCP manufactures battery chargers for safety-related applications at Nuclear Power Plants. Battery chargers are normally delivered to the nuclear power plant construction sites and are stored in warehouses for extended periods before they are installed and placed in service. , As such, they may not be removed from the packing crates and inspected  ! until just before installation, and since battery chargers contain capacitors which have limited shelf life, these capacitors may have to be replaced at this time. C. Action Taken on 10 CFR Part 21 Reports l

1. 10 CFR Part 21 Report on Crimping l

! Illinois Power Company, the owner of Clinton Nuclear Power Station, i reported on October 19, 1984, a potential defect relative to improper crimping in PCP battery chargers. The report stated that during start-up tests on three battery chargers, 27 out of 31 vendor crimped  ; lugs failed. On November 28, 1984, PCP dispatched their Assistant - Field Service Manager (AFSM) to investigate the potential defect. The AFSM field report dated November 30, 1984, indicated that only one j wire separated from the lug during pull tests. l When heat is applied to the heat shrinkable wire markers they tend to obscure the indentations left by the crimping tool, which may lead to an erroneous conclusion that the lugs have not been properly crimped. 174

ORGANIZATION: POWER CONVERSION PRODUCTS INC. CRYSTAL LAKE, ILLIN0IS REPORT INSPECTION NO.- 99900741/85-01 RESULTS: PAGE 3 of 5 The NRC inspector interviewed the ASFM and he confirmed that he pulled all the wires and only one wire separated from the lug. These battery chargers were shipped to Clinton Site in 1979. Illinois Power reported that all suspected wires were recrimped reinspected and determined the connections to be acceptable. l 2. 10 CFR Part 21 Report on Fuse Failures The 200 ampere fuse located in the center phase of the BC model 3S-130-200 was failing due to inadequate heat dissipation. Documents indicate that PCP notified all the owners of BC model 35-130-200 of the problem and recommended two solutions. One was to relocate the fuses to increase the efficiency of heat dissipation, the other was to replace the existing 200 ampere fuses with 300 ampere fuses. All the owners opted to replace the 200 ampere fuses with 300 ampere fuses. Subsequent to the replacement, no failures were reported. PCP has since modified the design and mounts the fuses horizontally l to increase heat dissipation.

3. 10 CFR Part 21 Report on Material Used l

l Brown & Root (B&R) reported that PCP used components inside the l battery chargers which were manufactured outside the United States and therefore did not meet their design specification 3E239ES047. PCP t stated that these commercial grade components were manufactured out-side the U.S. for parent companies such as General Electric. B&R's letter dated November 9,1981, retracted the 10 CFR Part 21 report on the basis that it was not a generic problem. To resolve this matter, on October 25 and 26, 1983, the PCP QA Manager visited the South Texas Project site to inspect the battery chargers. He identified that the shelf life of the DC electolytic filters (5 years) expired, that some indicating instruments were damaged and observed some minor workmanship problems. In the interim, the Architect Engineering responsibilities had been transferred to Bechtel Energy Corporation i (BEC) from B&R. BEC revised specification 3E239ES0047 to require a DC breaker with auxiliary contacts to remotely indicate a tripped l condition of the DC breaker and a 24 hour timer. Arrangements were made to ship the twelve battery chargers for modifications to PCP. The remaining four which are currently in operation will be returned for modification to PCP on receipt of the modified units. i i l

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175 l

ie ORGANIZATION: POWER CONVERSION PRODUCTS INC. CRYSTAL LAKE, ILLIN0IS REPORT INSPECTION NO - 99900741/85-01 RESULTS: PAGE 4 of 5 D. Inspection Findings and Other Comments

1. Shop Tour The NRC inspector toured the shop. Work was in progress to refur-bish the battery chargers for the South Texas Project. The inspector physically verified the integrity of the crimping of the control wires by exerting a nominal pulling pressure and observed no failures.

Calibration records were reviewed and determined to be acceptable. Calibration certificates indicate that standards traceable to the National Bureau of Standards were used.

2. Review of QC Inspector Qualification Records Review of the PCP QC inspector qualification records indicate that QC inspectors were qualified to established procedures. QA inspec-tors were normally selected from craftsmen who worked on the assembly floor for a specific period and given additional training. Two individuals were interviewed and determined that they were know-ledgeable in the circuits and trouble-shooting techniques. Pecords indicate that training sessions were given to individuals on various occasions on different subjects. Visual acuity test results, health records and qualification records were complete.

3. Review Trouble-Call Records The inspector reviewed the " trouble call" logs in which the calls received from customers are documented. The substance of the inquiry and the recommended suggestion is recorded. Review of the records for the past one year limited to nuclear power plants indicated that the calls were few and that problems were not con-centrated in failure in any particular operation mode.

4. Review of Service Call Records l

l The inspector reviewed the record of trips made to nuclear power plants to service installed battery chargers. The review indicated no consistent failures in any given area of battery charger operation. 176

ORGANIZATION: POWER CONVERSION PRODUCTS INC. CRYSTAL LAKE, ILLIN0IS REPORT INSPECTION NO.- 99900741/85-01 RESULTS: PAGE 5 of 5 l l E. Exit Interview The inspector met with representatives of PCP mentioned in paragraph F and discussed the scope and the results of the inspection, i F. Persons Contacted l *C. F. Seyer, Executive Vice President

       *M. F. Behr, Quality Assurance Manage i        M. J. Grant, Quality Control Manage D. L. Hunt, Assistant Field Service Manager

• Denotes those who attended the exit interview.

l l l l l l l l l l 177

ORGANIZATION: RUSKIN MANUFACTURING COMPANY GRANDVIEW, MISSOURI REPORT INSPECTION INSPECTION NO.: 99900716/85-01 DATE(S): 2/11-14/85 ON-SITE HOURS: 52 i l CORRESPONDENCE ADDRESS: Ruskin Manufacturing Company ATTN: Mr. Thomas D. Hill President i Box 129 3900 Dr. Greaves Road Grandview, Missouri 64030 ( ORGANIZATIONAL CONTACT: TELEPHONE NUMBER: Mr. Richard Yarges, QA Manager

(816)761-7476 PRINCIPAL PRODUCT: Air handling components j NUCLEAR INDUSTRY ACTIVITY: Approximately 6%

l

                              >0 ASSIGNED INSPECTOR:                                                               SM.1/gf G7 Petrosino, Reactive Inspection Section (RIS) ' Date OTHER INSPECTOR (S):      . L. Burns, Brookhaven National Laboratory APPROVED BY:       .    -

a E. W. Merschofy)2flief, RIS, VPB Date INSPECTION BASES AND SCOPE: A. BASES: 10 CFR Part 21 B. SCOPE: 1. Review circumstances concerning Fire Dampers failing to close.

2. Obtain information regarding Electro-Thermal Link (ETL) pro-blems.

3. Obtain information for NRC evaluation of HVAC component generic issues and functionability.

l l l 179

ORGANIZATION: RUSKIN MANUFACTURING COMPANY GRANDVIEW, MISSOURI REPORT INSPECTION NO.- 99900716/85-01 RESULTS: PAGE 2 of 8 l l Plant Site Applicability: Arkansas One (50-313), Beaver Valley 1&2 (50-334/412), Bellefonte 1&2 (50-433/439), Braidwood 1&2 (50-456/457), Browns Ferry 1,2&3 (50-259/260/296), Byron 1&2 (50-454/455), Callaway 1 (50-483), Catawba 1&2 (50-413/414), Clinton 1 (50-461), Comanche Peak 1&2 (50-445/446), D.C. Cook 1&2 (50-315/316), Davis Besse (50-346), Diablo Canyon 1&2 (50-275/323), Duane Arnold (50-331), Farley 1&2 (50-348/364), Hope Creek (50-354), EI Hatch 1&2 (50-321/366), Indian Point 3 (50-286), LaSalle 1&2 (50-373/374), Maine Yankee (50-309), McGuire 1&2 (50-369/370), Millstone 3 (50-423), Monticello (50-263), Oconee 1,2 & 3 (50-269/270/287), < Palo Verde 1,2, & 3 (50-528/529/530), Perry 1 (50-440), Quad Cities 1&2 (50-254/265), River Bend (50-458), Sequoyah 1&2 (50-327/328), South Texas 1&2 (50-498/499), Susquehanna 1&2 (50-387/388), Turkey Point 3&4 (50-250/251), Watts Bar 1&2 (50-390/391), Waterford 3 (50-382), WPPSS 1&2 (50-397/460), Wo1f Creek (50-482), and Zion 1&2 (50-295/304), (63 Total Plants). 1 180

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ORGANIZATION: RUSKIN MANUFACTURING COMPANY GRANDVIEW, MISSOURI REPORT INSPECTION NO.: 99900716/85-01 RESULTS: PAGE 3 of 8 A. Inspection Issues:

1. Failure of interlocking blade fire dampers to close under normal

, flow conditions. 1

2. Failure of Electro-Thermal Links (ETL) to function as designed, resulting in loss of fire damper closure capability.

B. Inspection Findings:

1. Ruskin Manufacturing Company (RMC) has adequately notified their affected purchasers in regard to fire dampers failing to close under normal duct flow. Additionally RMC recommended testing of the affected fire dampers to verify proper operation.

2. RMC documents were not in evidence to assure that the Palo Verde Gen-erating Station fire damper modifications were subject to complete design control measures, equal to those applied to the original design, prior to the actual field modifications being performed.

3. Installation problems with the ETL, and its associated conduit were identified at two facilities.

C. 5ucpitmental Information:

1. The NRC was notified by Ruskin Manufacturing Company, Grandview, Missouri by letter, on November 6, 1984, of an equipment deficiency with their curtain type interlocking blade fire dampers (IBFD).

The IBFDs are utilized to block air flow in Heating, Ventilating, Air Conditioning (HVAC) systems during a fire. RMC manufactures IBFCs at their Parsons, Kansas facility. The affect-ed Model numbers are NIBD23, IBD23, and IBD21. The Waldinger Corporation (TWC), Buckeye, Arizona, notified RMC of the deficiency by letter, dated August 3, 1984. Specifically, TWC iden-tified seven (7) dampers that failed to cycle during an Arizona Public Service system startup test. RMC in their November 6, 1984 letter, states in part, "...the test methods originally utilized to test dampers under flow may not.. depict actual installed conditions...Ruskin recommends that fire dampers 181 1

ORGANIZATION: RUSKIN MANUFACTURING COMPANY GRANDVIEW, MISSOURI REPORT INSPECTION NO.: 99900716/85-01 RESULTS: PAGE 4 of 8 supplied with closure springs, which require closure under air flow conditions be tested to veri fy proper operation. . .", and ". . . is currently researching... methods to modify fire dampers to ensure closure under air flow...". l 2. RMC personnel indicated that at least three design modifications were

planned or implemented at the Palo Verde Generating Station due to damper operability problems.

The design modifications include; (a) Modified blade catch assembly (b) Modified bottom blade bracket & spring attachment, and (c) Increased tension and size of negator spring. Within this area the NRC inspectors reviewed documents and testing records to assure that the fire damper design changes were subject to design contral measures equal with those applied originally. Specifically, prc4edure no. 112084PV, for fire damper closure modi-fications, and two Ruskin 1985 fire damper testing reports were reviewed and provided incomplete assurance that complete design control measures had been implemented. However, Ruskin submitted a RMC analysis report, dated 2/25/85 to the inspector several weeks after the inspection. The analysis states that RMC verifies the adequacy of the horizontal spring bracket assembly. Complete design control measures appear to be in place with the addi-tion of this analysis.

3. Two additional deficiencies were discussed in a review of Ruskin trip reports for their 2/85 Watts Bar, and 8/84 Palo Verde problem evalua-tion / inspection visits. These trips were made due to initial notifi-cation to RMC of fire damper closure problems. Both trips were per-formed to obtain information for use in RMC's problem evaluations.

The RMC Palo Verde evaluation resulted in the November 6, 1984, 10 CFR Part 21 report, and the Watts Bar problem resulted in a TVA noncon-formance report, number W-210P, and subseqtent licensee 10 CFR 50.55(e) report. 182 l

5 ORGANIZATION: RUSKIN MANUFACTURING COMPANY GRANDVIEW, MISSOURI REPORT INSPECTION NO.: 99900716/85-01 RESULTS: PAGE 5 of 8

a. A total of nine IBFDs with ETLs at Watts Bar and Palo Verde have experienced closure problems during startup and functional testing, l due to electrical conduit interference. The electrical conduit causing the interference is typically attached to the ETL. After the ETL melts and separates, the conduit may block or spring against the damper blades,

b. RMC's Watts Bar 2/85 trip report also noted two dampers that had not closed due to "S" hook interferences with the IBF0 blade pack-ages where the blades were caught in the "S" hook loops.

                   "S" hooks are utilized to hold the ETL or Thermal Links across the fire damper blades, until " heat" melts the link and releases the damper. RMC uses "S" hooks on all IBF0 applications. RMC recom-mended to TVA that installation personnel be instructed on the proper orientation of "S"    hooks at Watts Bar to prevent blade package interferences.

4. During a fire detection system test on November 20, 1984, TVA/ Watts Bar nuclear station identified 14 out of 47 Electro-Thermal Links (ETL) that ".. . failed to function as designed. . ." resulting in a loss of fire dem er closure capability.

However, these failures do not appear to be the result of a failure of the ETL's to function as designed. As discussed in item 3.a above, other factors caused the fire damper closure failures. TVA has concluded that inadequate tension on the ETL, and other workmanship problems caused the failures. TVA has documented this root cause on their Watts Bar nonconformance report numbers W-210P and W-220P. The lack of adequate tension applied to the ETL to cause link separa-tion, and electrical conduit interferences with the blade package may be generic issues.

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ORGANIZATION: RUSKIN MANUFACTURING COMPANY GRANDVIEW, MISSOURI REPORT INSPECTION N0.: 99900716/85-01 RESULTS: PAGE 6 of 8 D. Evaluation and Nuclear Application Information:

1. RMC manufactures nuclear related components at Parsons, Kansas and Minden, Louisana, (Louvers are manufactured at Minden and all other j components at Parsons).

I 2. RMC. components which may be utilized for nuclear applications are;

a. Sluce Gate Dampers

b. Louver Assemblies

< c. Backdraft Dampers

d. Isolation Dampers

e. Control Dampers

f. Volume Dampers

g. Balancing Dampers

3. The highest failure rate of the curtain type fire dampers, in regard to closure under air flow, are those which are installed horizontally. The Waldinger Corp. letter, dated August 24, 1984 to RMC, states in part, IBFD cycling problem was "... judged to be isolated to the horizontal dampers. ..", as concluded by RMC.

4. IBFD closure spring (negator spring) tensions and sizes are being increased in most RMC applications, to aid closures.

5. Remote damper operators (actuators) are mainly utilized on shutter I type dampers by RMC. Some tynical. actuator manufacturers used by RMC are: Bettis, Limitorque, ITT, Paul-Monroe, and Johnson. RMC is cur-rently completing the last of the Bettis actuator Elastomer-Lubricant

             " change outs". The " change out" was due to a previous Part 21 report which identified an incompatability between the elastomer and lubricant on Bettis actuators.

6. Standards for manufacturing are Underwriters Laboratory Standards UL 555 and UL 10B.

Testing Standards utilized are Air Moving and Conditioning' Association (AMCA) Standard AMCA 500. Wy le Laboratories performed seismic qualification testing for RMC's IBFOs on 3/23/79 & 7/22/80 per IEEE-344 requirments.

7. RMC stated that ETL installation instructions are provided to all customers. Instruction sheets are usually inserted between the fire damper and ETL prior to component shipping.

184 l l

ORGANIZATION: RUSKIN MAMFACTURING COMPANY GRANDVIEW, MISSOURI REPORT INSPECTION NO.: 99900716/85-01 RESULTS: PAGE 7 of 8 I

8. Installation instructions do not address RMC's IBFD orientation to air flow requirements.

9. The Ruskin NIBD23 Technical Data Sheet, states in part, "... ratings are based on AMCA standard 500 using test set up apparatus figure 5.3 (Damper installed with duct upstream & downstream). ..". The original test results were actually based on figure 5.5 of AMCA-5000, which does not have duct installed downstream of damper. Due to this discrepancy, RMC has recommended that testing to verify fire damper closure under air flow be performed. I

10. TVA/ Watts Bar has cuncluded that one or both of the following root causes prevented the closure of 14 out of 47 fire dampers, as indi-cated on Attachment A to TVA nonconformance report number W-210-P/

W-220-P.

a. Concealed damage to the ETL power leads or resistive bridge which occurred during handling, shipping, and/or installation.

b. Improper installation which resulted in inadequate tension on the ETL. This could have been caused by improper positioning of S-hooks or from external forces exerted by the flexible conduit connecting the ETLs to the power source.

E. Comments / Observations:

1. The RMC trip report to TVA/ Watts Bar dated 2/5/85 states in part, "...

majority of dampers in question are used for compartmentalization of the control room and are equipped with side seals and ETLs, and are referred to as Smoke Dampers...The side seals, which had been noted as previously holding up the blades, caused no problems during this test... Damper XFD-76 had jammed...the conduit may have held up the leading edge of... blade... Damper XFD-99... the conduit had fallen into the bottom of the frame, preventing complete closure... Damper XFD-74... had section held open by a backwards "S" hook...".

2. South Texas Project purchase order specification required RMC's IBFDs to operate at 3,550 FPM velocity and maintain a 9 inch Water gage pressure. This P.O. revision was a licensee to Region IV correc-tive action commitment, which was completed.

3. RMC Iateroffice Memo from Ruskins Palo Verde Project Manager to Ruskins Design Engineer, dated 12/13/84 states, "As you know, due to the recent testing of fire dampers under air flow for Palo Verde, modifications 185

ORGANIZATION: RUSKIN MANUFACTURING COMPANY GRANDVIEW, MISSOURI REPORT INSPECTION NO.: 99900716/85-01 RESULTS: PAGE 8 of 8 were made to the horizontal blade lock and spring attachment to blade. Please verify these changes will not adversely affect the seismic quali-fication". The return memo from the Design Engineer, dated 12/12/85 states, "The changes recently made to the horizontal dampers (blade lock & blade spring attachment) will not adversely affect the seismic qualifications." However, no design analysis, calculations or testing reports to verify the validity of the above RMC statement could be produced. After the conclusion of the NRC inspection, RMC approved an analysis titled,

              " Horizontal Spring Bracket Analysis", dated 2/25/85.

G. Persons Contacted: Richard J. Yarges OA Manager Ruskin Manufacturing Cosslett W. Moore Project Manager Ruskin Manufacturing Robert Van Becelaere V.P. Engineering Ruskin Manufacturing Connie S. Hubbard QA Engineer Ruskin Manufacturing Tim Arnold Sr. Project Manager Ruskin Manufacturing Robert Strait

• Project Engineer The Waldinger Corp.

Donald Wheeler

• QA Supt. The Waldinger Corp.

Johns Ciminski* Project Manager The Waldinger Corp. S.P. Weise* Engineering TVA Herb Wilhite* Engineering TVA Jim Worthy

• CDR-Engineering TVA
• Telephone contact only l

l 186

ORGANIZATION: STEWART & STEVENSON SERVICES, INC. HOUSTON, TEXAS REPORT INSPECTION INSPECTION N0.: 99900760/85-01 DATE(S): 2/25-26/85 ON-SITE HOURS: 22 CORRESPONDENCE ADDRESS: Stewart & Stevenson Services, Inc. ATTN: Mr. T. Michael Andrews Vice President Engineering / Manufacturing Post Office Box 1637  ! l Houston, Texas 77251-1637 ORGANIZATIONAL CONTACT: Stephen W. Bowman, Nuclear Projects Manager TELEPHONE NUMBER: (713) 923-2161 PRINCIPAL PRODUCT: Engineering services and diesel generators. NUCLEAR INDUSTRY ACTIVITY: Less than 1%. ASSIGNED INSPECTOR: 77. 47 %/d 5/#/Af N. # Mid@e6 Reactive Inspection Section (RIS) ~ Date OTHER INSPECTOR (S): T. F. Burns, Consultant, BNL APPROVED BY: T[/4!ff E. W. Merschoff p ief, RIS, Vendor Program Branch Date INSPECTION BASES AND SCOPE: A. BASES: 10 CFR Part 21. B. SCOPE: Determine how Stewart & Stevenson Services, Inc. (S&S) notifies customers of pertinent design modifications, maintenance instructicns, and changes to general operating procedures; and obtain a customer list to verify that customers receive the information from S&S and properly implement the recommended actions. PLANT SITE APPLICABILITY: Arkansas Nuclear 1(50-313), Braidwood 1 and 2 (50-456, 457), Byron 1 and 2 (50-454, 455), Clinton 1 and 2 (50-461, 462), LaSalle 1 and 2 (50-373, 374), Nine Mile Point 2 (50-410), Perry 1 and 2 (50-440,441), River Bend 1 and 2 (50-458, 459), San Onofre 2 and 3 187 l l 1

ORGANIZATION: STEWART & STEVENSON SERVICES, INC. HOUSTON, TEXAS - REPORT INSPECTION N0.: 99900760/85-01 RESULTS: PAGE 2 of 4 PLANT SITE APPLICABILITY (continued): (50-361, 362), WPPSS (50-397), St. Lucie 1 (50-335), Dresden (50-10), Prairie Island 1 and 2 (50-282, 306). A. INSPECTION OBJECTIVES:

1. Determine how Stewart and Stevenson Services, Inc. (S&S) notifies customers of pertinent design modifications, maintenance instructions, and changes to general operating procedures.

2. Obtain a customer list to verify that customers receive the informa-tion from S&S and properly implement the recommended action.

B. INSPECTION FINDINGS:

1. Diesel generator units assembled by S&S for commercial nuclear use are powered by engines manufactured by either General Motors Electro-Motive Division (EMD) or Detroit Diesel.

a. EMD engines - EMD issues " Maintenance Instructions" (mis) and

                   " Power Pointers" to their diesel engine customers. The mis and Power Pointers cover recommended maintenance procedures and design changes for EMD engines. S&S receives the mis and Power Pointers from EMD and in turn distributes them to their own customers having diesel generator units powered by EMD engines. The S&S Service Department maintains a master list of all S&S customers. The list is revised as needed to reflect address changes, and customers remain on the list until S&S receives notice that the unit has been sold. S&S does not have a written procedure covering how these notifications are processed, nor do they have an official means to confirm receipt of the literature by their customers and/or implementation of the recomendations. EMD does not generally send maintenance l                   information directly to end users of their engines unless EMD supplied the engine directly, or the end user specifically requested EMD to forward the information. EMD does not have a system by which end users can register engines with them.

Proper dispersal of information from EMD to the end user is dependent upon S&S maintaining accurate and up to date records. EMD does not require that the action recommended by the mis or Power Pointers be performed, nor does EMD state that engine operation and/or reliability will be jeopardized if the mis and Power Pointers are not followed. For this reason, mis and Power Pointers should be evaluated by knowledgeable personnel (i.e., a 188

ORGANIZATION: STEWART & STEVENSON SERVICES, INC. HOUSTON, TEXAS REPORT INSPECTION N0.: 99900760/85-01 RESULTS: PAGE 3 of 4 mairitenance engineer) to determine the applicability of the instructions to a particular engine.

b. Detroit Diesel engines - Unlike EMD, Detroit Diesel has a system by which end users of their engines register the units directly with them. S&S is an area distributor for Detroit Diesel. When S&S receives information pertaining to Detroit Diesel engines, S&S forwards the information to all Detroit Diesel engine owners that are located within the area served by S&S, regardless of whether or not the unit was purchased from S&S. If S&S has supplied a Detroit Diesel unit to a customer outside of their region, S&S does not send them the information. S&S relies on their customers properly registering the unit with Detroit Diesel, and the distributor that serves the customer's particular area co rectly passing on information, to ensure that proper notification of design changes and maintenance information is made. S&S has sold twelve Detroit Diesel units for nuclear use. Nine of the twelve are in service, three are not. S&S advises their customers to maintain a rapport with the local Detroit Diesel distributor to be sure they receive all pertinent information.

C. BACKGROUND INFORMATION: S&S takes a basic diesel engine and tailors it with subsystems to meet i each customer's particular requirements. For a typical nuclear order, S&S will generally design the unit, purchase the diesel engine and all other compor.ents needed, (such as heat exchangers, generators, etc.) and assemDie the components per the unit's design. S&S neither designs nor manufactures components, instead they assemble or " package" diesel generator units. S&S currently will not accept new purchase orders which reference e'ither 10 CFR 50 Appendix B or 10 CFR Part 21. However, they will certify that a part is the same as one supplied earlier when an Appendix B quality program was in place. S&S maintains a Part 21 program to meet their responsibilities under old purchase orders. D. 10 CFR PART 21: S&S " Procedure for Reporting of Defects and Noncompliance in Accordance with 10 CFR Part 21" dated February 8, 1983, was reviewed. The inspector verified that the procedures were adequate and in compliance with Part

21. The inspector also discussed with the S&S Nuclear Projects Manager 189

ORGANIZATION: STEWART & STEVENSON SERVICES, INC. HOUSTON, TEXAS REPORT INSPECTION NO.: 99900760/85-01 RESULTS: PAGE 4 of 4 three potentially reportable defects found with S&S units. Corrective action and/or deficiency reports covering these incidents were not available for review. The inspector was informed verbally that the evaluations of the defects by S&S determined that they were not reportable because they were isolated incidents, i.e., unique to a particular plant and unit. E. PERSONS CONTACTED: T. Michael Andrews, Vice President, Engineering / Manufacturing, Stewart & Stevenson.

 !

• Stephen W. Bowman, Nuclear Projects Manager, Stewart & Stevenson.

Jim Sutherland, QA Manager, Stewart & Stevenson John Fine, Warranty Administrator, Stewart & Stevenson David Whisenhunt, Branch Service Manager, Stewart & Stevenson

• Mick Lay, Stewart & Stevenson F. DOCUMENTS EXAMINED:

1. Index, dated 1984, GM-EMD Maintenance Instruction (MI) Index

2. QAM, dated 1/85, S&S QAM (Draft)

3. PRO, dated 2/8/83, Procedure for Reporting of Defects and Noncompliance in Accordance with 10 CFR 21.

l 4. REP, document no. 70-84, dated 11/1/84, INP0 Significant Event Report. l l 5. LTR, dated 1/21/85, S&S Letter from S. Bowman to R. F. Pariani of GE.

6. MI, 23 GM-EMD mis.

7. PP, 9 GM-EMD Power Pointers.

• Attended exit meeting 190

ORGANIZATION: TELEDYNE ENGINEERING SERVICES WALTHAM, MASSACHUSETTS I REPORT INSPECTION INSPECTION I NO.: 99900513/85-01 DATE(S): 1/7-11/85 ON-SITE HOURS: 62 i CORRESPONDENCE ADDRESS: Teledyne Engineering Services ATTN: Mr. F. C. Bailey l i President 130 Second Avenue Waltham, Massachusetts 02254 l ORGANIZATIONAL CONTACT: Mr. A. E. Johnson, QA Manager l TELEPHONE NUMBER: (617) 890-3350 PRINCIPAL PRODUCT: Engineering and consulting services NUCLEAR INDUSTRY ACTIVITY: Approximately 90% of Teledyne Engineering Services (TES) staff at Waltham, Massachusetts, facility is involved in nuclear activities. i ASSIGNED INSPECTOR: h k 4DC fif klef[95 R.P.McIntyre,'SpecialPropctsInspectionSect. Date OTHER INSPECTOR (S): M. Russell (EG&G) l APPROVED BY: 4ohn Craig, Chief, SpecM1 Project Inspection Section

                                                                             ' Date i

INSPECTION BASES AND SCOPE: A. BASES: 10 CFR Part 50, Appendix 8 and 10 CFR Part 21. ! 8. SCOPE: The purpose of this inspection was to review the following items: (1) program verification of Teledyne Engineering Services (TES) owned computer code THRSAP, (2) computer program error handling procedures, and (3) pipe support design calculations. PLANT SITE APPLICABILITY: Watts Bar (50-390, 50-391) 191

ORGANIZATION: TELEDYNE ENGINEERING SERVICES WALTHAM, MASSACHUSETTS l REPORT INSPECTION NO.: 99900513/85-01 RESULTS: PAGE 2 of 4 A. VIOLATIONS: None t B. NONCONFORMANCES:

1. Contrary to Section 5.1 of Project QA Program for TES project 6235C and Section 6.0 of Impell procedure WBNP-001 Rev 0., Watts Bar pipe support calculations 62-2CVC-R168 and 62-2CVC-R253 did not include documented justification for the omission of the effects of forces and moments due to pipe movements on the sizing of welded plate attach-ments, nor was there any documentation of qualification by engineering judgement.

2. Contrary to Section 3.1 and 3.7 of the TES Quality Assurance Manual, a vertical friction force was incorrectly applied at Node 9 of the GTSTRUDL model for Watts Bar Pipe Support calculation 62-2CVC-R42, TES project 6235C. Additionally, there was no documentation of this error with respect to structural adequacy of the support.

C. UNRESOLVED ITEMS None D. OTHER FINDINGS OR COMMENTS

1. Computer Program Verification: The development and verification of the computer program THRSAP, which is used by TES in the design of safety related items was reviewed during this inspection. Technical Engineering Procedures TEP-1-005, Application Computer Program Deve-lopment, was reviewed and utilized throughout the inspection of TMRSAP.

The computer code TMRSAP, which was developed internally by TES, is used for static and dynamic analysis of linear piping systems. It employs a finite element solution technique with a library con-sisting of curved and straight pipe elements, and a boundry element for simulation of pipe restraints. TMRSAP provides capability for analysis of such static loading as deadweight, thermal, and pressure elongation loadings. Capabilities for dynamic analysis include res-ponse spectrum and time history (both modal and direct) analysis. Solution methods include Gaussian elimination for static solutions, i and determinant search or subspace iteration for the modal dynamic solutions. Direct integration is performed with the Wilson-0-method. 192

ORGANIZATION: TELEDYNE ENGINEERING SERVICES WALTHAM, MASSACHUSETTS REPORT INSPECTION NO.: 99900313/85-01 RESULTS: PAGE 3 of 4 TES verified TMRSAP by a comparison of the output of 22 verification problem solutions with either the results of hand calculations or the output of other computer codes (STARDYNE, EPIPE, ANSYS, AND ADLPIPE). During this inspection all verification problems were reviewed. Although the verification of this code was done according to a general design control procedure (Section 3.0 of the TES Quality Assurance Manual), it was found to meet the requirements of the latest TES pro-cedure controlling computer code verification (TEP-1-005), with one exception. The exception was that the source code listing and compu-tation outputs were not included in the verification manual. However, the computation output includes a source code listing that was clearly identified in the verification manual and was readily available at the TES office. No violations or nonconformances were identified during this part of the inspection.

2. Computer Program Error Handling Procedures:

The procedures for handling computer program errors are covered in Section 14.0, Error Reporting, TEP-1-005 Rev 1. TEP-1-005 was revised 12/19/84 to include error reporting procedures as well as a Computer Program Error Notification and Disposition Report form, which is completed for all reported computer program errors. If it is deter-mined that a program error can impact current or previous analyses, then a Computer Program Error Project Disposition Report form must be completed. An investigation by TES project is initiated to search for and correct any current or previous analyses affected by the error. Since the procedures for error reporting were recently put into effect, the NRC inspector was unable to track a code error through the new system. TES has received a large number of computer code error reports from computer service bureaus, Control Data Corporation and Urited Information Services. TES is in the procest. of performing an investi-gation of the errors for the computer program ANSYS which have the poten-tial to impact past analyses. The search and disposition of these errors on past and present TES projects may be reviewed during a future NRC inspection. No violations or nonconformances were identified during this part of the inspection.

3. Pipe Support Design Calculations:

TES is a sub-contractor to Impell and provides services for the design and analysis of certain pipe supports for the Watts Bar Nuclear Power 193 I

ORGANIZATION: TELEDYNE ENGINEERING SERVICES WALTHAM, MASSACHUSETTS l l REPORT INSPECTION N0.: 99900513/85-01 RESULTS: PAGE 4 of 4 Plant. Impell has the overall contract with the Tennessee Valley l Authority (TVA). TES will provide support analysis packages and support drawings to Impell for the systems within their scope of the contract. During this inspection a sample of pipe Watts Bar Pipe support drawings and design calculations for the Chemical Volume and Control System (CVCS) were chosen for a detailed review. The review included verifi-cation of compliance to TVA Pipe Support Design Procedures, Impell Engineering Procedures for Pipe Support Designs, TES Project 6235C QA program requirements and applicable design codes and criteria. The inspectors reviewed the correctness of the numerical calculations related to support member stress, member deflections, weld design, base plate analysis, and concrete anchor bolt analysis. Additionally, computerized input and output for computer programs such as GTSTRUDL and BASEPLATE II was reviewed. The inspector also examined the appro-priateness of references and applicability of assumptions used in these programs. Within this area of the i.m pection, two nonconformances were identified (see Section B, items 1 and 2 above). These items are discussed below. Impell Engineering Procedure for P1pe Support Design (WBNP-001) expli-citly states that when engineering juQement is used as a means of qualification for a portion of a calculation, it should be documented and justified in the calculation package. When reviewing pipe support calculations 62-2CVC-R168 and 62-2CVC-R253, TES project 6235C, the NRC inspector found instances where engineering judgement was being uti-lized but not documented. These calculations did not include forces and moments resulting from piping movement when sizing the welds for the attachment of rigid sway struts to a bolted base plate and an embedded plate. This was confirmed during interviews with the project engineer on the Watts Bar Project. A vertical friction force was incorrectly applied at Node 9 of the GTSTRUDL model for a welded frame in the calculation of pipe support 62-2CVC-R42. It should have been applied at Node 5 of the model. Since there was already a horizontal load at Node 5, the error re-sulted in lower stress for the member incorporating Node 5. However, the loading conditions on the frame were low and the total ratio of calculated stress to allowable stress was well below the ASME code limit of 1.0 for combined stresses. Therefore, this input error will not affect the structural adequacy of this support. 194

ORGANIZATION: WESTERN CONCRETE STRUCTURES, INC. GARDENA, CALIFORNIA REPORT INSPECTION INSPECTION NO.: 99901003/85-01 DATE(S): 3/25-29/85 ON-SITE HOURS: 54 CORRESPONDENCE ADDRESS: Western Concrete Structures, Inc. ATTN: Mr. A. H. Stubbs President 19113 S. Hamilton Avenue Gardena, California 90248 ORGANIZATIONAL CONTACT: Mr. A. H. Stubbs TELEPHONE NUMBER: (213) 321-1571 PRINCIPAL PRODUCT: Post-tensioning components and installation services. NUCLEAR INDUSTRY ACTIVITY: Western Concrete Structures has supplied and installed post-tensioning systems for Nuclear Power Plants. Current work is limited to non-nuclear jobs. ASSIGNED INSPECTOR: Y( #c . t s  % ~ ' ' i' K. R. Naidu, Reactive Inspection Section (RIS) Date OTHER INSPECTOR (S): T. F. Burns, Brookhaven National Laboratory , APPROVED BY: . r[1!fS-E.W.Merschofffhief,RIS,VendorProgramBranch Date INSPECTION BASES AND SCOPE: A. BASES: 10 CFR Part 21 and 10 CFR Part 50 Appendix B. B. SCOPE: This was a followup inspection relative to the failed tendon anchor heads at J. M. Farley Nuclear Station. The failed tendon anchor heads were fabricated at Western Concrete Structures and heat treated by their subcontractor Downey Heat Treatment Company. The inspection included a review of materials and services for the post-tensioning components supplied to Palo Verde Nuclear Power Station. No items of noncompliance were identified. PLANT SITE APPLICABILITY: J. M. Farley Nuclear Station (50-364) and Palo Verde Nuclear Station (50-528). i 195

ORGANIZATION: WESTERN CONCRETE STRUCTURES, INC. GARDENA, CALIFORNIA REPORT INSPECTION RESULTS: PAGE 2 of 12 NO.: 99901003/85-01 A. INSPECTION FINDINGS: Tendon anchor head abnormalities were observed at the J. M. Farley Nuclear Power Station (JMFNPS). INRYC0 (located in Chicagu) supplied these anchor heads which were manufactured by Western Concrete Structures (WCS) to JMNFPS. WCS fabricated and installed post-tensioning systems at the Palo Verde Nuclear Generating Stations (PVNG). The inspection objective was to evaluate the WCS QA Program implementation relative to the procurement of material and services as a vendor to INRYC0 for JMFNPS and as a supplier to PVNG. B. INSPECTION FINDINGS:

1. Violations:

i None.

2. Nonconformances:

None. C. OTHER FINDINGS AND COMMENTS:

1. Background Information:

The two unit J. M. Farley Nuclear Power Station is operated by Alabama Power Corporation (APC). APC observed two broken field anchor heads and one cracked field anchor head during January and February of 1985. Each of these anchor heads retains 170 0.25 inch

!             diameter tensioned tendons. The tendon anchor heads were fabricated by Western Concrete Structures (WCS) frcm steel supplied by INRYCO.

The inspectors visited Downey Heat Treating Company where these tendon anchor heads were heat treated and selectively reviewed the heat treat-ment furnace charts and furnace calibration records. During this inspection, the NRC inspectors reviewed documents such as furnace charts which were not made available to them during their visit to INRYC0 during the March 4-8, 1985 visit. WCS supplied post-tensioning components to the PVNG Units 1, 2, and 3 which are owned by Arizona Public Service Company. Bechtel Corporation (Bechtel) is the architect engineer and constructor. WCS recently completed the installation of the post-tensioning systems for the three units. The NRC inspectors reviewed the WCS procurement ' documents, and field installation checklists, and audits performed i on WCS. 196 l l

ORGANIZATION: WESTERN CONCRETE STRUCTURES, INC. GARDENA, CALIFORNIA REPORT INSPECTION NO.: 99901003/85-01 RESULTS: PAGE 3 of 12

2. Review of Purchase Orders:

The inspectors reviewed the following purchase orders issued by INRYC0 to WCS to ascertain the adequacy of the technical and , quality assurance requirements: l

a. INRYC0 P0 to WCS 21T505-5 dated 5/22/73 for the supply of 360 sets of post-tensioning anchors for the J. M. Farley Unit 2 Nuclear Station. The following instructions specified the components.

(1) Supplementary instructions (SI) for certification of 170W1B for field applied anchor head, revision dated 9/1/71 with drawing attached. (2) SI for certification of 170W1A for shop applied anchor head, revision dated 9/1/71 with drawing attached. (3) SI for certification of 170W15 field applied bushing revision dated 9/1/71 with drawing attached. 1

b. The sis mentioned in the above paragraph contained the following information:

(1) Paragraph 3.a.1 states " Components to be batched by heat codes and sent to the heat treater. Vendor will receive certificate from the heat treater stating that 50% of each batch has been checked for conformance to range." (2) Paragraph 3.a.2 states " Vendor to give entire batch 100% inspection in three places 120* apart and report average Rockwell on 'C' scale for each head on certificate of inspection." (3) Paragraph 6b states "INRYC0 to retain all original certification documents." During the NRC inspection at INRYCO, INRYC0 personnel informed the inspectors that certification documents were not available and that they (INRYCO) requested WCS to provide them the infor-mation. The current inspection at WCS was to verify whether these records were available. WCS informed the inspectors that 197 l l

                                                                                     ]

ORGANIZATION: WESTERN CONCRETE STRUCTURE 4 INC. GARDENA, CALIFORNIA REPORT INSPECTION NO.: 99901003/85-01 RESULTS: PAGE 4 of 12 they shipped to INRYC0 all the original documents as required by the sis.

c. INRYC0 PO to WCS 21T 455-53 dated 5/22/73 for the supply of 360 i sets of post-tensioning anchors for J. M. Farley Unit 1 Nuclear l Station.

d. INRYC0 P0 to WCS 2T 505-395 dated 9/3/76 for the supply of addi-tional quantities of field anchor heads, shop anchor heads and field bushings.

The following supplementary instructions (SI) were referenced: (1) SI for the fabrication of type 170W1A shop anchor heads dated 11/20/73 with drawing dated 6/18/76. (2) SI for the fabrication of type 170W1B field anchor heads dated 11/20/73 with drawing dated 6/18/76. (3) SI for the fabrication of type 170W15 field applied bushing with drawing, both dated 6/18/76.

3. Review of Contractor Quality Requirements The subcontractor quality class Q Programs are addressed in Appendix l 011 to Bechtel Specification 13-CM-158 for " Furnishing and Installing Containment Building Post-Tensioning Systems at Palo Verde Nuclear Generating Stations 1, 2, 3." The Appendix D 11 requires the sub-contractor, in this case WCS, to conform to the quality assurance requirements stated in 10 CFR 50 Appendix B. Western Concrete Structures developed and maintained a list of Approved Vendors (AVL) to use as their contractors.

a. The inspectors selected Varco International (VARCO) who supplied Heat Treating Services to WCS to ascertain whether WCS followed the requirements in their QA manual to approve VARCO.

198

i ORGANIZATION: WESTERN CONCRETE STRUCTURES, INC. GARDENA, CALIFORNIA REPORT INSPECTION N0.: 99901003/85-01 RESULTS: PAGE 5 of 12 A WCS vendor checklist indicates that WCS inspected VARC0 on November 3,1978, and verified selected attributes of the VARC0 i , , QC manual. Based on the results, WCS determined that VARC0 was capable of meeting the WCS quality and technical requirements and placed VARC0 on their approved vendors list. The qualification process complies with WCS Quality Procedure No. 7.

b. The inspectors reviewed WCS QC checklists which were used to verify the quality of the shop anchor heads (SAH). The fabrication of the shop heads is done in three stages. WCS drills into the shop head with machines specially designed for this purpose using gun drills. Then the SAHs are sent to a machine shop - M&R Machines, Los Angeles, to thread them. Finally they are heat treated. WCS stated that they conduct inprocess inspections in addition to final inspections at every stage. They stated that all parts return to WCS before going to another subcontractor. The Certifi. cates of Inspection - Machine Components - identifies the QC inspector who verified the following attributes for i

shop anchor heads for compliance to WCS procedure QCP 1.2.B. (1) Sampled 25% of the internal and external diameter (2) Sampled 10% for drilling diameter and drift (3) 100% check on hardness testing - average of 3 values (4) 100% of external threads "go" tests (5) 25% of external threads "go/no go" tests. No items of noncompliance were identified in the above area.

4. Review of Quality Assurance Records The inspectors reviewed the qual.ity assurance records relative to the base material. The assurance of quality base material is of fundamental importance to the successful performance of the anchor heads (both shop and field installed) and the bushings. This is due to the continuous application of tension on a complex part.

The following certified material test reports were examined to l 199 l E

ORGANIZATION: WESTERN CONCRETE STRUCTURES, INC. GARDENA, CALIFORNIA REPORT INSPECTION t RESULTS: PAGE 6 of 12 NO : 99901003/85-01 f 1 determine if the specification requirements for base material were fulfilled: Order Date Supplier Quantity Heat No. Specification Cert. Date Jorgenson 1 Bar 17660 ASTM A 322-76 6/12/78 4/15/77 Gr 4142 , Jorgenson 2 Bars 18201 ASTM A 322 6/2/78 4/5/78 Gr 4142 5/10/78 Jorgenson 4 Bars- 21072 " 6/2/78 Jorgenson 5 Bars 18274 6/2/78 5/11/78 This examination revealed that the material supplied was in conformance with the specification (ASTM A 322, Grade 4142) requirements. Chemical analysis, product form, grain size and manufacturing practice (vacuum degassed) were verified on the applicable test reports. No discrepant items were noted on these reports. A review of the WCS manufacturing sequence was made to verify that heat identification was maintained. This processing sequence of base material consists of the following steps:

a. Receipt inspection

b. Color code bar stock,

c. Saw cut,

d. Stamping of Serial Number,

e. Drilling,

f. Machine Outside Diameter,

g. Heat treat,

h. Finish-Inspect,

i. Assemble-Ship.

The exanination of color code indices, certified material test reports, and assigned serial numbers revealed that material to component traceability had been maintained. No items of noncompliance were identified.

5. Corrective Action Taken By WCS l

WCS informed Bechtel that one of the tendon anchor heads fabricated l by them and supplied to INRYC0 for use in the Byron Nuclear Power Station failed in 1979, 10 days after tensioning. Bechtel l l 200 i

           .?               ,
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ORGANIZATION: WESTERN CONCRETE STRUCTURES, INC. GARDENA, CALIFORNIA REPORT INSPECTION NO.: 99901003/85-01 RESULTS: PAGE 7 of 12 i l reviewed this report and determined that it was not applicable for Palo Verde and hence not reportable. One of the considerations was that the Rockwell Hardness (RC) for the tendon anchor heads specified for Palo Verde was 38 2 versus 40 2 for the Byron Station. l Additionally, WCS conducted additional load tests to confirm ! the adequacy of the design. The results indicated that the l fabrication of the post-tensioning components for Palo Verde were j adequate to meet the design requirements. i No items of noncompliance were identified.

6. Review of WCS Installation Records The inspectors selectively reviewed the inspection records for the installation of the post-tensioning system at Palo_ Verde Unit 3 to l ascertain the adequacy of the WCS QC inspections. The following records were reviewed:

l l a. Certificates of Inspection - Tendon Installation - identify l the following attributes for compliance to procedure PTP 6.8: (1) Tendon number l (2) Location or Buttress number l (3) Tendon identification number j (4) Tendon length (5) Tendon length checked , (6) Wire condition (7) Water removed from sheath (8) Corrosion protection applied (9) Shop head threads protected (10) Shop head orientation correct.

b. Certificates of Inspection - Installation of Field Anchor Head and Field Button Heading - identify the following attributes for compliance to WCS procedure PTP 7.8:

(1) Tendon number (2) Location or Buttress number (3) Field anchor head oriented correctly (4) Tendon pushed back (5) Tendon temporary protection adequate (6) Machine qualified l (7) Button head eccentricity and head (8) Size verification (9) Inspection for splits. 201 i

ORGANIZATION: WESTERN CONCRETE STRUCTURES, INC. GARDENA, CALIFORNIA REPORT INSPECTION N0.: 99901003/85-01 RESULTS: PAGE 8 of 12 l c. Certificate of Inspection - Tendon Stressing and Installation j of Shims and Grease Caps - identify the following attributes for j compliance to WCS procedure PTP-8: (1) Tendon number (2) Location (3) Calibration of stressing equipment (4) Gage (5) Verification of tendon bearing plate (6) Tendon stressing record complete (7) Grease caps installed (8) Shim stacks installed.

d. Certificates of Inspection - Tendon Greasing - identify the following attributes for compliance to WCS procedure PTP-9:

(1) Tendon number (2) Location or buttress number (3) Filler material certification on hand (Viscon rust 2090 P4) (4) Filler temperature at tank (5) Filler temperature at cap (6) Sheathing purged at drain and vent (7) Caps leak tight at pumping end, far end, and at jumper connection (8) Pipe plug inserted at Y (9) Grease lines blown out. No items of noncompliance were identified in the above area.

7. Shop Tour The inspectors conducted a shop tour of the machining area at WCS.

All work currently in progress was of a commercial _ nature, and these activities were on a reduced scale. The facilities for drilling holes in the anchor heads were examined and found to be state-of-the-art numerically controlled gun drilling machines. The work areas observed were found to be clean and orderly. No items of noncompliance were identified in the above area.

8. Audits The inspectors reviewed the audits performed by independent sources

, and Bechtel on WCS. 202

ORGANIZATION: WESTERN CONCRETE STRUCTURES, INC. GARDENA, CALIFORNIA REPORT INSPECTION N0.: 99901003/85-01 RESULTS: PAGE 9 of 12

a. Independent Audits A review of the independent audits performed on WCS by outside sources was conducted to evaluate their level of compliance with the quality assurance program during the period when work was being processed for the Palo Verde Nuclear Generating Station.

The reports covering periods 1978, 1979, 1981, 1982, and 1983 were reviewed and the following was noted: Year Location Dates Number of Findings 1978 Gardena, CA February 2 & 3 8 1979 Gardena, CA October 3 & 4 7 1981 Gardena, CA October 13 & 14 4 1982 Gardena, CA December 13-16 12 1983 Palo Verde Site December 13 & 14 11 These audits were predominately of a programmatic nature. Corrective action was taken on all findings except for the period of 1981. WCS could not locate the records which detailed the corrective action taken during that period.

b. Purchaser Audits The inspectors reviewed the audits performed by Bechtel Corporation (Post Tensioning System - P.O. 10407-13-CM-158) for the periods 1978 through 1981 for the Palo Verde Nuclear Station.

No deficiencies were identified as a result of the audits of 1979, 1980 and 1981. Two "nonconformances" were identified in the 1978 audit which were program rather than hardware related. No items of noncompliance were identified in the above area.

9. Review of Heat Treating Facilities l There were two heat treating facilities that were involved in the l

performance of hardening and tempering of the anchor heads (shop l and field). The inspectors visited each facility to evaluate their l compliance to MIL-H-6875F (specification which establishes the heat l treatingprocessandequipmentcontrols). 203 1

ORGANIZATION: WESTERN CONCRETE STRUCTURES, INC. GARDENA, CALIFORNIA REPORT INSPECTION N0.: 99901003/85-01 RESULTS: PAGE 10 of 12

a. Varco Heat Treating Company 774 North Eckhoff Street Orange, California 92668 This facility had recently been sold by Varco International of which it was the heat treating subsidiary. While a subsidiary of Varco International, it was customary practice to solicit outside work to fill available heat treating capacity.

Purchase orders for heat treatment which were issued to Varco by WCS were as follows: P.O. Date Part Quantity Material 3715 11/20/78 Bushing 700 ASTM A-519 Gr. 4140 3813 1/9/79 Field Head 720 ASTM A-322 Gr. 4142 2723-R1 12/5/79 Shop Head 720 ASTM A-322 Gr. 4142 These items were to be heat treated to a Rockwell "C" hardness of 36/40 "per MIL-H-6875F." Selected instrument and furnace calibration records required by MIL-H-6875 were reviewed for the periods 1979,1980, and l'81  ; with the following results: Temperature Controlling Instrumentation Date Furnaces Not Tested 4 7/16/79 9D, 14D, 150, 16D 8/15/79 All were tested. 9/18/79 9D, 140, 150, 16D 1/9/80 9D 2/13/80 9D, 14D, 150 3/17/80 28, 2AD, 80, 90, 140, ISD, 160 4/22-28/80 14D, 150, 16D 6/26/80 90, 14D, ISD 7/26/80 All were tested. 8/27/80 9D, 140, 150, 16D, 2AD 10/10/80 2AD 11/19/80 28, 2AD 12/12/80 28, 2AD l 1/16-29/81 28, 2AD l 7/1/81 All were tested. 204

ORGANIZATION: WESTERN CONCRETE STRUCTURES, INC. I GARDENA, CALIFORNIA j REPORT INSPECTION N0.: 99901003/85-01 RESULTS: PAGE 11 of 12 Furnace Temperature Uniformity Surveys Furnace No. 2B (Used for Anchor Head Hardening) Tested on 2/27/85, 11/27/84, and 8/3/84 - no prior records were available. Tempering Furnaces (Used for Anchor Heads) 8D 9D 14D 15D 16D 2/5/85 12/3/84 1/14/85 1/10/85 6/18/82 7/26/84 4/12/84 9/23/82 6/18/82 3/12/82 1/11/84 8/30/82 3/26/82 10/7/81 9/14/83 6/2/82 1/11/80 2/27/81 6/9/83 3/4/82 12/12/79 5/14/80 3/3/83 8/28/81 3/22/79 6/21/82 3/18/81 3/29/82 2/24/81 9/22/81 12/7/80 9/15/81 2/28/80 Prior records were not available. An effort was made by the Varco Quality Control Manager to locate prior test results of furnace instrument calibration and temperature uniformity surveys, but this effort was unsuccessful. The Quality Control Manager had been on temporary layoff for over a year and had only recently been recalled to work. He stated that the records were complete as of two years ago but had been misplaced or lost during his absence. Also, it was during this time frame that the heat treating subsidiary was sold by Varco International. In these two areas, the records (or absence of them) indicates a failure to completely comply with the imposed requirement of MIL-H-6875F, governing the heat treating equipment,

b. Downey Heat Treating Company 9629 Nance Street Downey, California 90241 l

l l 205 I

ORGANIZATION: WESTERN CONCRETE STRUCTURES, INC. GARDENA, CALIFORNIA REPORT INSPECTION NO.: 99901003/85-01 RESULTS: PAGE 12 of 12 This facility is an independent heat treating organization of relatively small size. Only limited work was performed by this organization for WCS. Purchase orders for heat treatment issued to Downey by WCS were as follows: P.O. 3644 10/3/78 20 Bushings 2940 6/3/76 2 Field Heads 2933 - 4 Bushings 2933 - 3 Field Heads 2933 - 3 Shop Heads The items for heat treating on Purchase Orders 2933 and 2940 were identified as prototypes. The requirements of MIL-H-6875F were also imposed on this heat treating organization by WCS. However, due to limited time at this facility, the insnectors examined only the temperature uniformity survey results for the equipment which was used to perform the hardening of the post tensioning hardware. These records covered the period of 1976 to 1978 and were found to be in compliance with the requirements of MIL-H-6875F. For this period, the temperature uniformity surveys were performed within the required time frame and temperature variations were found to be within specification requirements. No items of noncompliance were identified in the above area. D. PERSONS CONTACTED:

a. Western Concrete Structures (WCS)

A. H. Stubbs, President l b. Varco Heat Treating Company (VARCO) R. Durham, Quality Control Manager, 774 N. Eckhoff Street, Orange, California 92668.

c. Downey Heat Treating Company (Downey)

J. R. Patterson, Vice President and Sales Manager, P. Cresswell, Manager, 9629 Nance Street, Downey, California 90241. 206

ORGANIZATION: WESTINGHOUSE ELECTRIC CORPORATION , NUCLEAR FUEL DIVISION PITTSBURGH, PENNSYLVANIA REPORT INSPECTION 12/3-7/84 INSPECTION N0.: 99900005/84-01 DATE(S): 12/17-18/84 3N-SITE HOURS: 49 CORRESPONDENCE ADDRESS: Westinghouse Electric Corporation Nuclear Fuel Division ATTN: Dr. Richard Slember General Manager Post Office Box 355 Pittsburgh, Pennsylvania 15230 ORGANIZATIONAL CONTACT: Mr. R. Cost, Manager Operations Product Assurance TELEPHONE NUMBER: (412) 374-2359 PRINCIPAL PRODUCT: Nuclear fuel assemblies. NUCLEAR INDUSTRY ACTIVITY: Nuclear fuel supplier for Westinghouse (W) designed Cores.

                                     . a ASSIGNED INSPECTOR:    R.                hd                               <     k R. L. Cilimberg, Spegial Projects Inspection       Date Section (SPIS)

OTHER INSPECTOR (S): APPROVED BY: t I /St$ Craig, Chief, SPIS, Ven q Program Branch 'Datd i INSPECTION BASES AND SCOPE: i A. BASES: 10 CFR Part 50, Appendix B. B. SCOPE: Manufacturing and special process control including fuel pellet fabrication, fuel rod loading, bundle assembly, and follow-up on previous inspection findings. PLANT SITE APPLICABILITY: Ginna 1 (50-244), Turkey Point 4 (50-251), Point Beach I (50-266), Surry 2 (50-281), Point Beach 2 (50-301), Cook 1 (50-315), Millstone 2 (50-336), Trojan (50-344),Farley 1(50-348), Farley 2(50-364), McGuire 1 (50-369). 207

ORGANIZATION: WESTINGHOUSE ELECTRIC CORPORATION NUCLEAR FUEL DIVISION PITTSBURGH. PENNSYLVANIA r REPORT INSPECTION N0.: 99900005/84-01 RESULTS: PAGE 2 of 4 A. VIOLATIONS: None. 1 B. NONCONFORMANCES: None. C. UNRESOLVED ITEMS: None. D. STATUS OF PREVIOUS INSPECTION FINDINGS:

1. (Closed) Nonconformance (Item A, 82-01): A strap inspection procedure, QCI 933025 was not retained as required by Quality Control Instruction (QCI)-000147, Control and Distribution of QCIs.

All W files reviewed during this inspection were found to be complete and in compliance with QCI-000147 requirements. QCI-000147 was revised in October 15, 1982, to include an instruction to area supervisors that they must verify that inspectors being qualified to perform a particular activity have signed cover sheets for all pertinent QCIs.

2. (Closed) Nonconformance (Item B, 82-01): Operating procedure (0P)-

715604 was not in the assigned book in work area 61 (Spider Rework Station) as required by Manufacturing Operating Procedure (M0P)-14. OP-715604 was placed in the assigned book, supervisors were reinstruc-ted, and all work areas reviewed during this inspection were found to be in compliance with M0P-14. E. OTHER FINDINGS OR COMMENTS: The inspector reviewed a number of items which are described below and are considered closed.

1. Manufacturing and Special Process Control - Control of the manufac-turing of fuel assemblies and special processes was verified and the performance of functions by operators and inspectors was observed to meet the requirements of written procedures and criteria.

2. Posting of 10 CFR Part 21: The September 12, 1983, edition of 10 CFR Part 21 was posted rather than the May 9,1984, edition. W personnel stated that they believed that the May 9, 1984, edition did not contain any substantive revisions and therefore, the current edition was not posted. The inspector discussed with W personnel the require-208 1

ORGANIZATION: WESTINGHOUSE ELECTRIC CORPORATIOW NUCLEAR FUEL DIVISION PITTSBURGH, PENNSYLVANIA REPORT INSPECTION N0.: 99900005/E1-01 RESULTS: PAGE 3 of 4 ment that current copies of 10 CFR Part 21 be posted (10 CFR 21.6). The May 9, 1984, edition of 10 CFR Part 21 was posted prior to the end , of the inspection. l

3. Damaged Fuel Cladding at Farley Unit No.1: During the fuel off-loading from cycle 4 at Farley 1 on January 29, 1983, damaged fuel was observed. Subsequent analysis determined that the damage was caused by water jetting which resulted from a 30 psig pressure differential from a faulty flow path. The problem was corrected by the conversion from a downflow system to an upflow system during February 1984.

4. Broken Hold Down Springs at McGuire Unit No. 1: Inspection of the top fuel assemblies at McGuire 1 on March 11, 1983, revealed that a number of hold down springs on burnable poison assemblies were broken. All of tne upper head injection springs have been replaced with a new design which eliminates the vibration which was resulting in fatigue failure of the broken hold down springs. Discussions with W personnel indicate that breakage of the newly designed springs has not occurred.

5. Broken Hold Down Springs and End Post Wear at Millstone Unit No. 2:

Inspection of top fuel assemblies at Millstone 2 during June 1983, l discovered broken hold down springs and galled end posts. Evaluation of the broken springs indicates that cross flow resulted in fatigue failure of one spring per assembly in 15 assemblies because the material exhibited unsatisfactory fatigue properties. The fatigue strength of the new design has been increased by a factor of ten through improved processing and larger springs. The galled end posts were caused by fretting wear between the posts and stop pins. The problem was identified as improper clearance between stop pins and the end posts. The new design features increased clearances to minimize fretting.

6. Clad Defects in One Fuel Assembly at Point Beach Unit No. 2: Higher than normal levels of I-131 in the primary coolant of Point Beach 2 during the Spring 1983 refueling outage prompted an inspection which

found through-wall penetrations of the cladding of several fuel rods.

, The holes in the cladding were caused by fretting wear in a gap between the cladding and the bottom spring clips. W concluded that the gap was the result of misalignment of the rod as it was inserted in the fuel bundle through a guide plate fixture. The assembly procedure has been modified by W to prevent misalignment of rods so the recurrence of this problem is not likely. The modified procedure was reviewed

during this inspection.

209

ORGANIZATION: WESTINGHOUSE ELECTRIC CORPORATION NUCLEAR FUEL DIVISION PITTSBURGH, PENNSYLVANIA REPORT INSPECTION N0.: 99900005/84-01 RESULTS: PAGE 4 of 4

7. Fuel Pellets with Improper Diameter and Enrichment at Sequoyah Unit 2:

Nineteen fuel rods containing pellets of an improper diameter and enrichment were loaded into fuel assembly P39 and shipped to Sequoyah 2. The problem was discovered by a QC engineer on August 24, 1983, while inspecting records and P39 was not loaded into the reactor. The problem occurred because gamma scan instructions permitted the processing of a faulty rod lot because the defect coding was not entered into the computerized data system to prevent further processing. The gamma scan instructions have been revised and the proper defect coding entered into the data system to ensure that defective rods are rejected at the time of inspection to prevent further processing. Physical barriers were installed between pellet production lines to prevent mixing enrichments. The above corrective actions were reviewed during this inspection.

8. Clad Defects in One Fuel Pin at Salem Unit No.1: Inspection of a fuel rod at Salem 1 on November 21, 1982, found two holes in the cladding. W evaluation determined that the holes were caused by hydriding. This fuel was manufactured in 1976 when the control on hydrogen was not sufficient to prevent hydriding. The hydrogen specification was revised in 1981 to minimize the chance of hydride failures and reviewed during this inspection.

l l 210 l

ORGANIZATION: WESTINGHOUSE ELECTRIC CORPORATION l

 ;                    PITTSBURGH, PENNSYLVANIA REPORT                          INSPECTION                                          INSPECTION N0.: 99900404/85-01             DATE(S): 3/4-6/85                                   ON-SITE HOURS: 51 CORRESPONDENCE ADDRESS: Westinghouse Electric Corporation                                                                  -

Nuclear Technology Division .. ATTN: Mr. J. L. Gallagher, General Manager i Post Office Box 355 ' " Pittsburgh, Pennsylvania 51230 ORGANIZATIONAL CONTACT: Mr. P. T. McManus, Quality Assurance e TELEPHONE NUMBER: (412) 825-7988 ,

                                                                                                                         '[

PRINCIPAL PRODUCT: Nuclear Steam Supply Systems NUCLEAR INDUSTRY ACTIVITY: Westinghouse provides NSSS components, other i safety and non-safety related components, and services. l n O f ASSIGNED INSPECTOR: .) rt fo[/ir[P5' PQ D. Milano, Specialtoj@:ts Inspection Bate j 5ection (SPIS) OTHERINSPECTOR(S): J. W. Craig, SPIS W. Pq Haass, Program Coordination Section APPROVED BY: i G[(d5 CphnW.Craig, Chief,yIS,VendorProgramBranch Date INSPECTION BASES AND SCOPE: A. BASES: 10 CFR Part 21 and 10 CFR Part 50. B. SCOPE: The purpose of the inspection was to followup on previous inspec-tion findings and to review the Westinghouse system for providing informa-tien to their customers pertaining to installation and operation of Westinghouse supplied equipment. PLANT SITE APPLICABILITY: Multiple: Westinghouse NSSS facilities. 211

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ORGANIZATION: WESTINGHOUSE ELECTRIC CORPORATION PITTSBURGH, PENNSYLVANIA REPORT INSPECTION N0.: 99900404/85-01 RESULTS: PAGE 2 of 5 A. Violations and Nonconformances None B. Status of Previous Inspection Findings

1. (Closed) Nonconformance (A/84-03): No documentation was available to determine the effect of an error identified in Problem Report No. 1511 on the design activity accomplished utilizing the WECAN computer code.

The Westinghouse Plant Engineering Division " Evaluation of WECAN Problem Report 1511," dated 12/14/84, was reviewed. The evaluation indicated that there was no impact on safety-related design since the feature with the error was not utilized in piping modeling or equipment analysis. However, WECAN users and affected manage-ment were informed of the requirements for evaluation and documen-tation of error reports.

2. (0 pen)Nonconformance(B/84-03): No procedures were in place to review and document the effect of computer program and system errors on design, past and ongoing.

The procedure for verification of computer programs is being revised to address the review and documentation of the effect of computer code errors. However, the procedure had not been issued at the time of the NRC inspection.

3. (0 pen) Nonconformance (C/84-03): Field Change, Control No. CAE-9455, authorizing design changes to the Instrumentation and Control Pro-tection System, was approved and constructed without Quality Assur-ance and Nuclear Safety Department approvals.

Westinghouse is reviewing the Commonwealth Edison Projects active Change Controls to ensure that any field changes which did not have Quality Assurance or Nuclear Safety approval were properly evaluated. The item remains open until the reviews are completed.

4. (Closed) Nonconformance (D/84-03): The computer program ANSYS, procured from SWANSON, remained available for use on safety-related designs after Westinghouse was notified that the program contained I errors. The errors were of a nature that erroneous results could be obtained. The Advance Energy Systems Division also continued to use and maintain ANSYS for its own safety-related design activities

( 212

ORGANIZATION: WESTINGHOUSE ELECTRIC CORPORATION PITTSBURGH, PENNSYLVANIA REPORT INSPECTION N0.: 99900404/85-01 RESULTS: PAGE 3 of 5 after receipt of the error notices. When requested, no documentation could be produced to justify the acceptability of the version of ANSYS that contained the errors. Westinghou'se Plant Engineering Division (PED) utilized ANSYS on three occasions during this period, none of which was for safety-related applications. On February 22, 1985, PED disapproved the use of ANSYS for further work in the division. The Advance Energy Systems Division did not utilize the program after this date due to the termination of the Clinch River Breeder Reactor Program in 1983. Further inves-tigation for other potential users is ongoing.

5. (Closed) Nonconformance (E/84-03): An error identified and cor-rected in the computer program NOTRUMP verification calculation note SEC-RFFA-1381-C0 (Generic CE Plant N0PRUMP Small Break Spectrum) was not similarly corrected in the affected verification calculation note SEC-RFFA-1381-C0 (CE NOTRUMP Input Deck).

Westinghouse Safeguards Engineering and Development (SE&D) calcula-tion note system allows revisions to calculation notes by adding the next higher "Cx" suffix number. This allows for revisions of notes that have been archived on microfilm. In this case, while note SEC-RFFA-1381-Cl was added to correct a portion of the original note, its title was sufficiently different to be misinterpreted as a i different note. SE&D Instruction and Guidance NS-SED-IG-3 has been clarified so that calculation notes that are not related by specific title will have different calculation note numbers.

6. (Closed) Nonconformance (F/84-03): The computer program NOTRUMP verification calculation notes SEC-RFFA-1381-C0 and SEC-RFFA-1381-C1 l contain reviewer coments requiring resolution. However, the com-ments did not include the documented resolution and the verifier's signature.

The Safeguards Engineering and Development Instruction and Guidance Material (NS-SED-IG-3) Preparation of Calculation Notes has been modified to require that " reviewer comment resolution must be docu-mented in to calculation note." In the above cases and in the past, the design verifier's final signature was used as evidence that all comments had been resolved.

7. (Closed) Nonconformance (G/84-03): The set of problems identified in the computer program WECAN User Manual, Table 5-2, Volume II, did not support the verification of gap element, STIF 77, which demonstrates i

213

ORGANIZATION: WESTINGHOUSE ELECTRIC CORPORATION PITTSBURGH, PENNSYLVANIA REPORT INSPECTION N0.: 99900404/85-01 RESULTS: PAGE 4 of 5 the capability of the element to simulate the opening and closing of a gap between structural components. This item of nonconformance was corrected during the course of the inspection. By means of WECAN User's Memo 85-04, dated February 27, 1985, all users were notified of the verification completion of the STIFF 77 element.

8. Since the implementation of the procedures, modification and evalu-ation of the 10 CFR 21 reporting system are still in progress, the remaining open items from NRC Inspection Report 99900404/84-02 were not reviewed.

C. Inspection Report Summary An internal Westinghouse system was initially developed in the late 1950's to provide information relating to installation, operation, and maintenance of equipment. This system was comprised of Data letters which were sent to Westinghouse service personnel. This system was used in special cases as a means of providing information to Westinghouse customers. In 1977, Westinghouse instituted a system of Technical Bulletins to supple-ment the system of Data Letters. While the Technical Bulletins contained information similar in nature to Data Letters, Technical Bulletins were utilized if the information affected the written guidance contained in instruction books rather than providing supplemental information to Westinghouse service personnel. Technical Bulletins and Data Letters were initially used as a means to transmit information within Westinghouse. However, Technical Bulletins were sometimes sent to the Westinghouse customers. Based upon discussions with Westinghouse personnel and absent written guidance describing the purpose of Data Letters or Technical Bulletins, the specific use of either document is somewhat unclear. However, Data Letters have not been issued since 1982. Each Division within Westinghouse is required to determine when a Technical Bulletin should be issued. Technical input is then provided to the Nuclear Service Integration Division (NSID) which is responsible for the prepara-tion, review, approval, and issuance of Technical Bulletins. NSID prepares the Bulletins with concurrence from the responsible design organizations. l l Technical Bulletins can be and are used to provide supplemental and revised l information for instruction (Installation, Operat.on, Maintenance Manuals) I manuals. The affected documents are not revised at a later date to incorporate this information. 214

ORGANIZATION: WESTINGHOUSE ELECTRIC CORPORATION PITTSBURGH, PENNSYLVANIA REPORT INSPECTION NO.: 99900404/85-01 RESULTS: PAGE 5 of 5 After the Salem ATWS event in 1982, Westinghouse implemented a more formal l control mechanism for tracking their customers' receipt of safety-related Technical Bulletins. A computerized tracking system is used to track return receipts for those Technical Bulletins that affect safety-related equipment (basic component), as defined by Westinghouse. These Technical  ; Bulletins contain a form which Westinghouse requests that their NSSS customers to return acknowledgir g receipt of the Bulletin. During the inspection, selected Technical Bulletins were reviewed. The receipt, evaluation and implementation of actions, if any, that are determined by utilities to be appropriate, will be reviewed during future NRC inspections. 1 215

ORGANIZATION: WESTINGHOUSE ELECTRIC CORPORATION NUCLEAR TECHNOLOGY DIVISION FOREST HILLS, PENNSYLVANIA REPORT INSPECTION INSPECTION NO.: 999M900/85-01 DATE(S): 3/18-22/85 ON-SITE HOURS: 116 CORRESPONDENCE ADDRESS: Westinghouse Electric Corporation Nuclear Technology Division l l ATTN: Mr. J. L. Gallagher, General Manager Post Office Box 355 Pittsburgh, Pennsylvania 15230-0355 ORGANIZATIONAL CONTACT: Mr. P. T. McManus, QA Manager, NTD , TELEPHONE NUMBER: (412) 825-7988 PRINCIPAL PRODUCT: Functional and environmental testing of nuclear power plant equipment. NUCLEAR INDUSTRY ACTIVITY: Westinghouse Nuclear Technology Division (W-NTD) Forest Hills test laboratory performs developmental, verification and quali-fication testing of both nuclear and non-nuclear power plant components. Loss-of-coolant accident (LOCA)/ thermal aging equipment qualification testing of nuclear power plant safety-related equipment comprises approximately 10 % of the facility's work. ASSIGNED INSPECT 0 : JE. APE C-9 -8f G. T. Hubbard, Equip. Qual. Inspec. Section (EQIS) Date OTHER INSPECTOR (S): S. D. Alexander, EQIS E. H. Richards, Sandia National Laboratories (SNL) M. Jacobus, g f . APPROVED BY: MO hNetA%Lo f #-kT 1 U. Potapovs, Chief, EQIS p Vdndor Program Branch Date INSPECTION BASES AND SCOPE: A. BASES: 10 CFR Part 50, Appendix B,10CFR Part 21,10 CFR Part 50.55(e), and Topical Report WCAP-8370, Revision 10A. B. SCOPE: This inspection consisted of: (1) technical review and eval-uation of certain environmental qualification test plans prepared by Westinghouse Nuclear Services Integration Division (W-NSID) and non-Westinghouse sponsors and (2) a followup inspection as a result of reports under 10 CFR Part 21 and 10 CFR Part 50.55(e) concerning a ootantial oroblem with the Westin< house 7300 orocess protection system. PLANT SITi APPLICABILITY (DOCKET NO./ NAPE) 413,414/ Catawba-1 & 2 50-338,339/ North Anna-1 & 2 395/ Virgil C. Summer 423/Hillstone-3 348,364/Farley-1 & 2 400/ Harris-1 & 2 424,425/Vogtle-1 & 2 369,370/McGuire-1 & 2 412/ Beaver Valley-2 443,444/Seabrook-1 & 2 217

ORGANIZATION: WESTINGHOUSE ELECTRIC CORPORATION NUCLEAR TECHNOLOGY DIVISION FOREST HILLS, PENNSYLVANIA ! REPORT INSPECTION l NO.: 99900900/85-01 RESULTS: PAGE 2 of 7 l PLANT SITE APPLICABILITY (DOCKET NO./NAME): (continued) l 50-445,446/ Comanche Peak-1 & 2 456,457/Braidwood-l & 2 483/Callaway-1 454,455/ Byron-1 & 2 482/ Wolf Creek-1 498,499/ South Texas-1 & 2 A. VIOLATIONS: None. B. NONCONFORMANCES: None. C. UNRESOLVED ITEM: The NRC inspector's review of the test report for Forest Hills test program 83-0296 identified that W-NTD did not receive from its radiation services vendor all the specific data required by W-NTD purchase order (P0) 546-MZY-505117-SN, dated April 19, 1984. From discussions with W-NTD personnel, the inspector determined that the vendor had been audited in 1982 and approved as a W-NTD vendor. This approval was based in part on the data the vendor committed to provide in fulfilling P0s issued to him, even though the commitment was not in full compliance with W-NTD's " Quality Procurement Specification for Radiation Services," TR-QCP-82-005. W-NTD had issued PO 546-MZY-505117-SN after the date of the above audit and had included specification TR-QCP-82-005, with no exceptions to its requirements, as a contractual provision of the P0. This action was inconsistent with the basis for approval of the vendor. Review of the test documentation revealed that W-NTD subsequently accepted a level of compliance with the specification below that called for in the purchase order. D. STATUS OF PREVIOUS INSPECTION FINDINGS: None. E. OTHER FINDINGS OR COMMENTS:

1. Technical Evaluatinn of Qualification Test Programs: NRC inspectors and Sandia consultants evaluated ten EQ test specifications to deter-mine whether they met the approved methodology of WCAP 8587, Revision 6A, Methodology For Qualifying Westinghouse WRD Supplied NSSS Safety-Related Electrical Equipment, and regulatory requirements.

The EQ process prescribed in each test plan, bases for accelerated j thermal aging and irradiation, calculations, assumptions, acceptance limits , and test results including any deviations or anomalies were reviewed. l 218

ORGANIZATION: WESTINGHOUSE ELECTRIC CORPORATION NUCLEAR TECHNOLOGY DIVISION FOREST HILLS, PENNSYLVANIA

,   REPORT                                  INSPECTION i

NO.: 99900900/85-01 RESULTS: PAGE 3 of 7

a. The EQ test plans and related' engineering documents for the tdst programs listed below were examined to verify that the following items were considered, as applicable. Pertinent i

l inspector comments and observations are included. (1) Test equipment included a description of all materials, parts, and subcomponents. (2) Equipment interfaces were addressed. (3) Same specimens underwent all phases of testing and were representative of in use or to-be-installed plant equipment. (4) Test acceptance criteria were established as described in the test specification or in the design engineering documents to meet regulatory requirements. (5) All prerequisites for the given tests as outlined in the test specification had been met. (6) Environmental conditions were established and described (e.g., pressure and temperature profiles, and thermal aging factors were consistent with those outlined in the test specification and those prescribed in WCAP d587). (7) Adequate test instrumentation was described and used to meet requirements. (8) Test results were accurately reported. No nonconformances were identified in this examination, however, the review of applicable procedures revealed the need for clari-fications described in paragraph E.3. b. The examination of documentation described above included the following test programs: (1) Seismic test program 84-0342 for thermocouple (T/C) Bulkhead Connector and Mineral Insulated (MI) cable: The test items had completed functional testing following radiation aging and were awaiting the start of seismic testing at the time of the inspection. 219 l l

ORGANIZATION: WESTINGHOUSE ELECTRIC CORPORATION NUCLEAR TECHNOLOGY DIVISION FOREST HILLS, PENNSYLVANIA REPORT INSPECTION NO.: 99900900/85-01 RESULTS: PAGE 4 of 7 (2) Test program 85-0358 for Resistance Temperature Detector (RTO) and T/C Splice and Hardline Cable: The test items were undergoing functional checkout following radiation exposure at the time of the inspection. LOCA/high energy line break (HELB) testing will follow this functional checkout. (3) Test program 82-0199 for post-design basis event (DBE), high-range, radiation monitoring equipment (including ion chamber, cables and connectors): Samples had failed in November, 1984 and were retested in February, 1985. At the time of the inspection, the cause of failures had not been determined and the test report had not been completed. (4) Test program 84-0343 for T/C connector / adapter (part of a post-accident, in-cora, temperature monitoring system): This program was in the post-DBE simulation phase at the time of the inspection. (5) Test program 84-0344 for Integrated Head Package (IHP) connectors (part of post DBE monitoring system): The connectors were undergoing radiation aging at the time of the inspection. (6) Test program 82-0193 on an air control valve: Results of testing completed at Forest Hills and test anomalir.s/ deviations had been documented and reported. (7) Test program 84-0331 for a T/C cold reference junction box: Testing had been completed, but the final test l report was in review and not available for examination r during the inspection. (8) Test program 84-0183 for an electronic pressure trans-mitter: Test parameters were achieved but one of the samples suffered a failure of an instrument lead seal fitting. This was repaired and the test completed. , All results and anomalies were reported.  ! (9) Test program 82-0182 for electronic differential pressure transmitters consisted of LOCA/ post-0BE testing: The  ; testing has been completed and results and anomalies i documented and reported to the sponsor.  ! 220

ORGANIZATION: WESTINGHOUSE ELECTRIC CORPORATION NUCLEAR TECHNOLOGY DIVISION FOREST HILLS, PENNSYLVANIA REPORT INSPECTION NO.: 99900900/85-01 RESULTS: PAGE 5 of 7 (10) Test program 83-0292 is a generic type test sequence to qualify a large pump motor (LMP) (emergency injection coolant charging pump) for the HELB environment: The i sample is plant specific but will also be considered representative of similar LMP's in a range of sizes with  ; similar HELB susceptability (due to location). The I specimen and the test facility were being prepared for I the HELB test at the time of the inspection.

2. Followup of 10 CFR 50.55(e)/10 CFR Part 21 Reports by Westinghouse:

A potential problem in the Westinghouse 7300 process protection system (PPS) was reported under 10 CFR Part 21 for operating plants and under 10 CFR Part 50.55(e) for plants under construction. These items were inspected during NRC inspection 99900240/83-01, at Westinghouse-Industry Electronics Division and foilowed up during subsequent inspections 99900900/83-03 and 84-01 at W-NSID. During the 84-01 inspection, the NRC determined that this item could be closed out pending verification of issuance of a change control order in accordance with design control manual procedure NTD-DPP-5A dated July 24, 1981. During this inspection, the NRC inspector reviewed W-NSID Change Control Order No. 6258, dated March 6,1985, which requires the input test relay on the temperature channel test cards of the 7300 PPS to be bypassed. Verification of the issuance of this change order was the only outstanding item from the previous inspection; therefore, this is considered closed.

3. During the inspection, the NRC inspector recommended that certain procedures be clarified to ensure that practices prescribe by sub-tier QA procedures are consistent with the intent of the Westinghouse Water Reactors Divisions (WRD) Quality Assurance Plan (Topical Report)

WCAP 8370 (Rev. 10A). The WRD positions on and commitments to various regulatory guides and associated industry standards are given in table

,            17-1 of WCAP 8370. Items for which clarification was recommended are as follows:

a. Under paragraph 4 (QA Review of Procurement Documents) of the above table, the WRD position on the provisions of section 3.3 of ANSI Standard N45.2.13-1976 corresponding to the subsection entitled " Contract Award" allows either of the following two methods for processing procurement documents:

(1) QA reviews purchase requisitions. Purchase orders are then reviewed by QA after issue by Purchasing. 221

ORGANIZATION: WESTINGHOUSE ELECTRIC CORPORATION NUCLEAR TECHNOLOGY DIVISION FOREST HILLS, PENNSYLVANIA REPORT INSPECTION NO.: 99900900/85-01 RESULTS: PAGE 6 of 7 I l (2) QA reviews purchase requisitions prior to contract award and i purchase orders are reviewed " concurrently" with issue. l l W-NTD operating policy / procedure WRD-0PR-4.0 (Rev. 3), " Pro-curement Document Control," of the NTD/S0D Design Control Manual (WCAP 9565) (a sub-tier document to WCAP 8370), under the heading " RESPONSIBILITIES" and the subheading " PREPARATION AND REVIEW 0F PROCUREMENTS AND CHANGES THERETO," requires

                 " Product Assurance" (PA) to review purchase requisitions for quality requirements and to review purchase orders "at the time of issue."

Although WRD-0PR-4.0 is more restrictive in prescribing the second of the two methods allowed by WCAP 8370, W-NTD interprets WRD-0PR-4.0 to allow the first method. The NRC inspector recom-mended clarifying WRD-0PR-4.0 to reflect this interpretation since the practice and intent are to follow the guidance of the first method given in WCAP 8370,

b. Paragraph 5 of Table 17-1 of WCAP 8370, under the WRD position corresponding to the subsection on nonconformances of ANSI Stan-dard N45.2.13-1976, requires suppliers to submit deviations from WRD technical procurement requirements for approval, but it does not require the supplier to submit formal nonconformance reports on deviations from their own, WRD approved, manufacturing or pro-cess procedures.

Paragraph 5 ("Nonconformances"), under " CONTENTS OF THE PROCURE MENT DOCUMENTS", of procedure WRD-0PR-4.0 of WCAP 9565 requires the procurement documents to include WRD requirements for repor-ting and dispositioning nonconformances including (1) (when ap-plicable) a system for identification, documentation and evalu-ation of discrepancies and for alerting the suppliers cognizant management of the need for corrective action and (2) deviation notices on nonconforming materials. Whereas WCAP 8370 requires deviation notices in cases of deviation (or proposed deviation) from technical procurement specifications but does not require formal reports when suppliers deviate from their own procedures, the words of WRD-0PR-4.0 do not make that distinction and imply that reports of all nonconformances are required. Additionally, this section of WRD-0PR-4.0 is interpreted by W-NTD, PA and pur-chasing personnel interviewed to exclude purchase of irradiation and seismic testing services for equipment qualification. l l 222 l

ORGANIZATION: WESTINGHOUSE ELECTRIC CORPORATION NUCLEAR TECHNOLOGY DIVISION FOREST HILLS, PENNSYLVANIA REPORT INSPECTION NO.: 99900900/85-01 RESULTS: PAGE 7 of 7 i The NRC inspector therefore recommended clarification of this part of WRD-0PR-4.0 (1)to resolve the inconsistency with WCAP 8370 and provide for clearly transmitting the intended noncon-formance reporting requirements of WCAP 8370 to vendors and (2) to state clearly the situations to which it is applicable and not applicable

c. " Quality Procurement Specification for Radiation Services",

TR-QCP-82-005 is invoked by P0s for EQ testing and contains a prohibition only against certain specific changes to a compo-nent or its radiation dose without W-NTD approval, but incon-sistent with WCAP 8370 and WCAP 9565, it has no provision for , reporting nonconformances with other technical procurement specifications. The NRC inspector reccomended clarification of TR-QCP-82-005 to be consistent with WCAP 8370 in requiring prior permission and/or deviation notices for all deviations from technical procurement specifications. 223 l L

1 ORGANIZATION: Wyle Laboratories Scientific Services and Systems Group Huntsville, Alabama REPORT INSPECTION INSPECTION N0.: 99900902/85-01 DATE(S): 6/10-14/85 ON-SITE HOURS: 59 ' CORRESPONDENCE ADDRESS: Wyle Laboratories Scientific Services and Systems Group ! ATTN: Mr. W. W. Holbrook, General Manager Eastern Test and Engineering Operations

7800 Governors Drive Huntsville, Alabama 35807 ORGANIZATIONAL CONTACT

Mr. E. W. Smith, Director, Contracts and Purchasing TELEPHONE NUMBER: (205) 837-4411 PRINCIPAL PRODUCT: Research, engineering, and test operations i NUCLEAR INDUSTRY ACTIVITY: Wyle Laboratories; Huntsville, Alabama, provides a variety of nuclear services to the industry which includes environmental

 ! and seismic qualification testing of safety-related equipment, refurbishment i  and recertification of valves, valve and component flow testing, mechanical 4

and hydraulic snubber testing, decontamination, anr! repair. ASSIGNED INSPECTOR: 1/Al a/A M //, M N k If .7/Ji>L//f/6 A. N. Mcfist, Equiparent Qualification Inspection Date Section (EQIS) OTHER INSPECTOR (S): E. H. Richards, Sandia National Laboratories , APPROVE 0 BY: M ~1- 29 -15" U. Potapovs, Sectioh Ch'ief EQIS Date INSPECTION BASES AND SCOPE: A. BASES: 10 CFR Part 21 and 10 CFR Part 50, Appendix B. B. SCOPE: This inspection consisted of: (1) a technical evaluation of

!        equipment qualification (EQ) test activities for safety-related equipment; (2) witnessing of inprocess EQ testing; and (3) verifica-tion of implementation of the quality assurance (QA) program.

PLANT SITE APPLICABILITY: La Salle County 1 & 2 (50-373, 50-374); Dresden 1, 2, 3 (50-10, 50-237, 50-249); Quad Cities 1 & 2 (50-254, 50-265); Nine Mile Point 2 (50-410). 4 l i 225 i

ORGANIZATION: Wyle Laboratories, Scientific Services and Systems Group l Huntsville, Alabama l REPORT INSPECTION NO.- 99900902/85-01 RESULTS: PAGE 2 of 4 , A. Violations l None. B. Nonconformances None. C. Unresolved Items None. D. Other Findings or Coments

1. Technical Evaluation The NRC inspector and Sandia consultant performed an in-depth technical evaluation and review of four test programs for quali-fication of safety related electrical equipment. The following table summarizes the test programs examined, including equipment type and types of documents examined.

Test Program Equipment Type Documents Examined Plant 17605-1 SRV Solenoid Pilot Test Report (TR), La Salle Valve Qualification Plan (QP), County Purchase Order (P0), Data Sheets (DS) and Letter 17604-1 Level Switches TR, QP, P0, DS and La Salle Letter County 45916-03 Instrumentation TR, QP, P0, Letter, Dresden, and Power Cables Specification and dss Quad Cities 17590-1 Area Radiation TR, QP, Specification, La Salle , Monitor Sensor and and Letter County  : Converter l The NRC inspector and Sandia consultant reviewed the qualification , prescribed in each QP and reviewed test results, including the basis I l for accelerated thermal aging and radiation, and verified ) calculations, l 226

ORGANIZATION: Wyle Laboratories, Scientific Services and Systems Group Huntsville, Alabama REPORT INSPECTION NO - 99900902/85-01 RESULTS: PAGE 3 of 4 Each of the four QPs and related engineering documents were examined to verify the following:

a. Adequate test instrumentation and their accuracies were described and used to meet the requirements of IEEE-STD-323/1974.

b. Equipment interfaces were addressed.

, c. Test acceptance criteria were established as described in the test specification or in the design engineering documents, such as calculations and engineering letters to meet the requirements of IEEE-STD-323/1974.

d. Same equipment was used for all phases of testing and represented a standard production item.

e. Environmental conditions were established and described (e.g., pressure and temperature profiles, and thermal aging factors were consistent with those outlined in the test specification or test plan).

f. Test results were adequately reduced and evaluated against established acceptance criteria described in customer test specifications or purchase crders.

g. All prerequisites for the given tests as outlined in the test specification had been met.

h. Test equipment included a description of all materials, parts, and subcomponents,

i. Notices of Anomalies were properly documented.

j. Appropriate margins were applied.

The EQ documentation package for the level switch Model 751 manu-factured by Magnetrol identified a test anomaly during post thermal functional testing where actuation of the switch mechanisms of both specimens would not cause sufficient travel of the micro-switch plungers to change the state of the contacts. Testing of two spare microswitch assemblies also had the same results. Because of this anomaly, a maintenance interval was incorporated into the test program. Malfunctions of the level switch during the accident test were also observed. Notices of anomalies were documented by Wyle Labs. The effect of these anomalies j 227

ORGANIZATION: Wyle Laboratories, Scientific Services and Systems Group Huntsville, Alabama REPORT INSPECTION NO.- 99900902/85-01 RESULTS: PAGE 4 of 4 on the qualification of the level switches are +o be determined by Comonwealth Edison Company (CECO) and/or Sargent & Lundy (S&L).

2. Observation of Testing Activities The NRC inspector and Sandia consultant observed thermal aging testing of a Target Rock Solenoid Valve 6 X 10 Model 7567F being performed for General Electric Company (GE). The NRC inspector and Sandia consultant determined after review of test data, in-process environmental chamber surveillance records, test equipment calibration labels and chart recorder records that testing was performed in accordance with QP 47736-00 Rev. C dated 5/25/85.

It was determined by Wyle Laboratories that the leakage rate exceeded the acceptance criterion for the valve operator during the post thermal functional test. The leaks were located at two plugs on the operator. Disposition of a Notice of Anomaly by the GE representative directed Wyle to repair one plug using nuclear grade sealant which caused the leak rate to be brought back within the acceptance criterion.

3. Followup on 10 CFR Part 21 Report Crosby Valve and Guage Company filed a 10 CFR Part 21 report with NRC Headquarters concerning a latch roller bearing failure of a Crosby main steam line isolation valve similar to that used at Nine Mile Point 2.

This failure occurred during seismic testing at Wyle Laboratories which was being conducted for Gulf and Western. The NRC inspector reviewed and evaluated one TR and verified that two notices of anomalies were documented for the above failure. The NRC inspector determined that the information contained in the notices of anomalies supported the information contained in the Part 21 report. Gulf and Western / Stone and Webster requested that the specimen be returned to Gulf and Western facility in the "as failed" condition for their further evaluation. 228

ORGANIZATION: WYLE LABORATORIES NORCO, CALIFORNIA ~ REPORT INSPECTION INSPECTION NO.: 99900905/85-01 DATE(S): 5/13-17/85 ON-SITE HOURS: 99

CORRESPONDENCE ADDRESS

Wyle Laboratories

( Western Operations ATTN: Mr. R. C. Sadlier General Manager 1841 Hillside Drive Norco, California 91760 ORGANIZATIONAL CONTACT: Mr. Larry Houston, QA Manager TELEPHONE NUMBER: (714) 737-0871 PRINCIPAL PRODUCT: Equipment testing and engineering services NUCLEAR INDUSTRY ACTIVITY: The Wyle Laboratories, Western Operations facility provides engineering and test services to military, utility and commercial nuclear power organizations. Approximately fif teen percent of the facility's capability is currently committed to environmental qualification testing of safety-related equipment for the commercial nuclear power industry.

 /.SSIGNED INSPECTOR:                                                           y[dd f/.)bd                                      d-/9-f6
                                                                            @. N. M6ist, Equipment Qualification Inspection        Date Section (EQIS)

OTHER INSPECTOR (S): G. T. Hubbard, EQIS E. H. Richards, SNL V.J. Dana,SMg APPROVED BY: 4b d; JA (2-d f3 U. Potapovs, Chief, EQIS, NPB (-l~1f.I Date INSPECTION BASES AND SCOPE: , A. BASES: Appendix B to 10 CFR Part 50 and 10 CFR Part 21 B. SCOPE: This inspection consisted of: (1) a technical evaluation of Equipment Qualification (EQ) test activities for safety-related equipment; (2) witnessing EQ testing; and (3) verification of implementation of the quality assurance (QA) program. PLANT SITE APPLICABILITY: Not identified ' 229

ORGANIZATION: WYLE LABORATORIES NORCO, CALIFORNIA REPORT INSPECTION NO.: S9900905/85-01 RESULTS: PAGE 2 of 5 A. VIOLATIONS: None B. NONCONFORMANCES:

1. Co,ntrary to Criterion V of Appendix,8 to 10 CFR Part 50 and para-graph 6.1 of Standing Practice Procedure 518-3-B dated March 15, 1982, Test Plans 566-1674 Revision A dated May 1, 1984.and 566-1674-1 Revision A dated May 1,1984, had no documented objective evidence of approval by the same organization that performed the original review and approval.

2. Contrary to Criterion V of Appendix B to 10 CFR Part 50 and Sec-tion 11, paragraph 11.10 of W'ley Laboratories Quality Assurance Manual Revision E dated March 1, 1985, Test Reports 58883-1 dated November 20, 1984 and 58883 dated August 24, 1984, documented instances where deviations to the prescribed requirements occurred during testing but Notices of Deviation were not written.

3. Contrary to Criterion V of Appendix B of 10 CFR Part 50 and Section 5 paragraph 5.0 of Wyle Laboratories Quality Assurance Manual Revision E dated March 1, 1985, there was no documented procedure for the control of mixing the chemical spray solution used during design basis event testing and monitoring the solution's PH.

C. UNRESOLVED ITEMS: None D. OTHER FINDINGS OR COMMENTS:

1. Technical Evaluation - The NRC Inspectors and Sandia Consultants performed an indepth technical evaluation and review of eight programs for qualification testing of safety-related electrical equipment. The following table summarizes the programs examined including equipment type and types of documents examined. l Program Equipment Type Documents Examined 26443 Thermocouples Test Report (TR),

Qualification Plan (QP), l Data Sheets i 230

ORGANIZATION: WYLE LABORATORIES NORCO, CALIFORNIA REPORT INSPECTION NO.: 99900905/85-01 RESULTS: PAGE 3 of 5 53273 Pneumatic Hydraulic Oper. TR, Purchase Order (P0), i Test Procedure (TP), Data Sheets 53309 Electro-Hydraulic Actuator Job File (JF) 58681 Electrical Penetration TR, JF, TP, P0 58801 Operator Assembly & Servo TPs TR, Specification,

   ,                                                            Data Sheets, P0s 58883                       Cable, Conduit Seals          TP, P0, TF, TRs 58939                       Limitorque Fast Acting        QP, TR, Specification Actuator 58941                       AC Motor Actuator             TR The NRC inspectors and Sandia Consultants reviewed the EQ process pre-scribed in each test plan and reviewed test results, including the bases for accelerated thermal aging and radiation and verified calculations.

Each of the eight EQ test plans and related engineering documents were examined for the following:

a. Adequate test instrumentation and their accuracies described and used to meet NUREG-0588/IEEE-STD-323/1984.

b. Equipment interfaces addressed.

c. Test acceptance criteria established as described in the test specification or in the design engineering documents, such as I calculations and engineering letters to meet NUREG-0588/IEEE-

STD-323/1974.

d. Same equipment used for all phases of testing and represented a l standard production item.

l e. Environmental conditions established and described (e.g., pressure and temperature profiles, and thermal aging factors l were consistent with those outlined in the test specification l or test plan). l 231

ORGANIZATION: WYLE LABORATORIES NORC0, CALIFORNIA REPORT INSPEC~i10N NO. 99900905/85-01 RESULTS: PAGE 4 of 5

f. Test results adequately reduced and evaluated against established acceptance criteria described in custoner test specifications or purchase orders and demonstrated that these requirements had been met.

g. All prerequisites for the given tests as outlined in the test specification met.

h. Test equipment description included a description of all materials, parts, and subcomponents,

i. Notices of Deviations properly documented.

During the above evaluation and review, the NRC inspectors and Sandia Consultants identified the Nonconformances described in paragraph B.1, B.2 and B.3 and as follows:

a. The NRC inspector review of the summary section of TR 58883-1 dated March 20,198a, showed that test specimen No. I started to leak at the beginning of Loss of Coolant Accident (LOCA) testing. The conduit seal was plugged off and left in the LOCA tank for the remainder of the test. No Notice of Deviation was written. (See Nonconformance B.2)

b. The NRC inspector review of paragraph 5.1.4. of TR 58883 dated August 24, 1984, showed that after the specimens were mounted in the LOCA chamber and the chamber was sealed, a controlled steam /

superheated steam blowdown of the LOCA chamber was performed for six hours and forty-five minutes. The test was interrupted at three hours and fifteen minutes due to the loss of superheated steam. During the down time (approximately nine hours) the LOCA chamber was maintained at 200F and ambient pressure. The super-l l heater was recharged and testing continued. No Notice of Devia-tion was written. (See nonconformance B.2)

c. The NRC inspectors' and Sandia Consultants' review and evaluation of test program 58939 being conducted for General Electric deter- ,

mined that two motors failed during LOCA testing of the Limitorque i SB-3-150 motor actuator. Wyle and their customers are currently reviewing and evaluating the failures. The NRC Inspector will follow-up on these failures during a future inspection at GE. 232

ORGANIZATION: WYLE LABORATORIES NORCO, CALIFORNIA REPORT INSPECTION NO.: 99900905/85-01 RESULTS: PAGE 5 of 5 The NRC Inspectors and Sandia Consultants also observed a portion of the continuing test, inprocess functional test during LOCA, of a third motor of the Limitorque SB-3-150 Motor Actuator. One opening and one closing of the Motor Actuator was conducted during l i the observation. 233 J

INDEX FACILITY REPORT N0. PAGE The Advanced Products Company, North Haven, Connecticut 99900898/85-01 1 A&G Engineering Co., II, Inc., Anaheim, California 99901006/85-01 9 Babcock & Wilegx, A McDermott Co., Utility Power Generation Division, Lynchburg, Virginia 99900400/85-01 17 Bechtel Power Corporation, Eastern Power Division, , Gaithersburg, Maryland 99900519/85-01 29 Bechtel Power Corporation, Western Power Division / Houston Project Office, Los Angeles, California 99900521/85-01 35 Borg Warner Corporation, Nuclear Valve Division,

Van Nuys, California 99900289/85-01 43 Cardinal Industrial Products Corp.,

Las Vegas, Nevada 99900840/85-01 57 Combustion Engineering, Power Systems Group, Windsor, Connecticut 99900401/85-01 69 Corporate Consulting & Development Co., Ltd. Research Triangle Park, North Carolina 99900511/85-01 75 General Electric Company, Nuclear Energy Business Operations, San Jose, California 99900403/84-04 81 Gulfalloy, Inc., Houston, Texas 99900343/85-01 97 l Inryco, Inc., Bedford Park, Illinois 99900731/85-01 109 Isomedix(NewJersey)Inc., Whippany, New Jersey -99900913/85-01 119 Morrison-Knudsen Company, Inc. Rocky Mount, North Carolina 99900702/85-01 123 235

INDEX.(continued) FACILITY REPORT N0. PAGE 1 l Multi-Amp Services Co., j Dallas, Texas 99900539/85-01 131 National Technical Services,  ! Hartwood, Virginia 99900914/85-01 135 l National Technical Systems, Acton, Massachusetts 99900912/85-01 141 NES Manufacturing, Greensboro, North Carolina 99901018/85-01 147 f Nuclear _ Parts Associates, St. Francisville, Louisiana 99901010/85-01 165 Nutherm International Inc., Mount Vernon, Illinois 99900779/85-01 169 Power Conversion Products, Inc., Crystal Lake, Illinois 99900741/85-01 173 Ruskin Manufacturing Co., Grandview, Missouri 99900716/35-01 179 Stewart & Stevenson Services, Inc., Houston, Texas 99900760/85-01 187 Teledyne Engineering Services, Waltham, Massachusetts 99900513/85-01 191 Western Concrete Structures, Inc., Gardena, California 99901003/85-01 195 Westinghouse Electric Corporation, Nuclear Fuel Division Pittsburgh, Pennsylvania 99900005/84-01 207 Westinghouse Electric Corporation, Nuclear Technology Division, Pittsburgh, Pennsylvania 99900404/85-01 21. 1 Westinghouse Electric Corporation, Nuclear Technology Division, Forest Hills, Pennsylvania 99900900/85-01' 217 236 i

    .      _       _ _ _ , . - ~ . .        .. _ _ _ _ __   __    _      . _ _ . - _ . .             .  - _ _ __

l l- INDEX(continued) I i-T j FACILITY ~ REPORT NO. PAGE Wyle Laboratories, Scientific Services and Systems Group,

i. Huntsville, Alabama 99900902/85-01 225 i

l Wyle Laboratories, Western Operations, Norco, California 99900905/85-01 229 I ? 1 1 i j l 4 l 4 i + l '. I r i i 4 237

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e s a t e s' a . 2 h 2 %

                                                                                        >   2 2

Mvanced P% ducts Company X X X A8di Engineering Co. . . II. Inc.(1) Babcock & Wilcox Bechtel Power Corp. (519/85-01) Bechtel Power Corp. (521/85-01) X 8org Warner Corp. X X X X X Cardinal Ind. Prod. Corp. I Combustion Engineering (3) Corp. Consult. and Develop. Co., Ltd. General Electric Co. (2) Gulfalloy, Inc. X X

 ; Inryco, Inc.
                                                                                                      ~--

Isomedix Inc. (1) Morrison-Knudsen Co. , Inc. X X X X X ( Multi-Amp Services Co. National Technical Services (1) National Technical Systems j NE5 Manufacturing X Nuclear Parts Assoc. X Nuthern International Inc. Power Conversion Prod. Inc. I Ruskin Manufacturing Co. X X X X X X X Stewart & Stevenson Services. Inc. X X X Teledyne Engineering Services X Western Concrete Structures. Inc. Westinghouse Electric Corp. (5/84-01) X X X Westinghouse Electric Cc o. (404/85-01) (3) westinghouse Electric Corp. (900/85-01) X X X X Wyle Laboratories (902/85-01) Wyle Laooratories (905/85-01) (1) (1) Not identified (2)' Multiple BWRs (3) Multiple PWRs 242

i 1 l I i

                                              )

VENDOR INSPECTION REPORTS RELATED TO . I REACTOR PLANTS T &

                                   =

2 m i f  : 1 5 VENDORS g j 2 ;q Advanced Products Company X X A&G Engineering Co.,

  *II. Inc.(1)

Baocock & wilcox X Bechtel Power Corp. (519/85-01) X Bechtel Power Corp. {521/85-01) Borg Warner Corp. X X Cardinal Ind. Prod. Corp. Combustion Engineering (3) Corp. Consult. and Oevelop. Co.. Ltd. General Electric Co. (2) Gulfalloy, Inc. Inryco, Inc. Isomedia Inc. (1) Morrison-Knudsen Co. , Inc. M.;lti-Amp Services Co. 5 hational Technical Services (1) National Technical Systems NES Manufacturing huclear Parts Assoc. huthern International Inc. Power Conversion Prod. Inc. Ruskin Mar facturing Co. X X X Stewart ,. Stevenson Serv ice . Inc. X Teledyne Engineering Services Western Concrete Structures. Inc. Westinghouse Electric Corp. (5/84-01) Westinpouse Electric Corp. (404/85-01) (3) Westingnouse Electric Corp. (900/85-01) X Wyle Laboratories (902/85-01) Wyle Laboratories (905/85-01) (1) (1) Not identified (2) Multiple BWRs (3) Multiple PWRs i 243

g, EOR == u Nucus.a uruuTOR, C-e ,0 .. u ,Ou Nu.. ,A . .Vr,oc. v.,u...,,,, L"o*" % BIBUOGRAPHIC DATA SHEET NUREG-0040 St E INSTRUCTIONS ON Ts*E river $t Vol. 9, No ~ 2

2. TITLE AND busTITLE 3. LE AVE BLANE j Licensee Contrac or and Vendor Inspection Status Report 1 Quarterly Report April 1985 thru June 1985 -

                                                                                                                            / DATE REPORT COMPLEf f o
                                                                                                                        .p T R                         T AR l
     . Auf OR,,,                                                                                                   July                               1985
                                                                                                                                  . OAT. R ,ORY ssuno
                                                                                                                       /

' j oONT,, v.AR l Afgust 1985 7 PE,*FORU6NG ORGA412AT60N NAME AND M AILS ADORES 5 fsacswerle C.ed S PR CTsf ASK. WORE unsef NUMSER Division of Quality Assur ce, Vendor, and Technical Training Center Programs . Oa GaANT Nu-aia Office of Inspection and En cement U.S. Nuclear Regulatory Commi ion Washinoton. D.C. 20555 2 10 SPONSORING ORGANt2ATION FeAWE ANO MAeLING AOOREES 4eer le Cases lie Yvet OF REPORT Same as 7 above. Quarterly b PERIOD COv tRED fiac4asrae mmsJ

                                                                                                ~

April 1985 thru June 1985

    ,, su,,u..N Y AR v NOT .
    ..       T R .C , _ _ ,

This periodical covers the results of 1 ctions performed by the NRC's Vendor Program Branch that have been dis ibuted to the inspected organizations during the period from Apr 1985 through June 1985. Also included in this issue are the results certain inspections performed prior to April 1985 that were not incl ed i previous issues of NUREG-0040. I A A

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