ML20058F925

Licensee Contractor and Vendor Inspection Status Report. Quarterly Report,April-June 1990.(White Book)

ML20058F925
Person / Time
Issue date:	10/31/1990
From:	Office of Nuclear Reactor Regulation
To:
References
NUREG-0040, NUREG-0040-V14-N02, NUREG-40, NUREG-40-V14-N2, NUDOCS 9011090222
Download: ML20058F925 (125)

Text

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i NUREG-0040 Vol.14, No. 2 4.

Licensee Contractor and Vendor Inspection Status Report Quarterly Report i

April-June 1990 U.S. Nuclear Regulatory Commission i

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Available from

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

Single copies of this publication are available from National Technical Information Service, Springfield, VA 22161 i

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I NUREG-0040 Vol.14, No. 2 4

i Licensee Contractor and Vendor Inspection Status Report Quarterly Report April-June 1990 Manuscript Completed: July 1990 Date Published: October 1990 t

Division of Reactor Inspection and Safeguards Omce of Nuclear Reactor Regulation U.S. Nuclear Regulatmy Commission Washington, DC 20555 l

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ABSTRAC.T This periMicui covers the results of inspections perfornud by the NRC's Vendor Inspection Branch that have been distributed to the inspected organization durits the period fiua April 1990 thrcugh June 1990.

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TABLE OF COtmNIS 1

PI G Abstract..........................................................

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Preface...........................................................

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Reporting Format..................................................

ix Index................................:............................

xi Inspection Reports................................................

1 Selected Bulletins and Information ibtices concernirg Adequacy of i

Verdor Audits and Quality of Verdor Products......................

109 Verdor L's[xece lons Related to Reactor Sites.......................

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PREFACE 7

A fundamental. premise of the Nuclear Regulatory Commission's (NRC) licensing and inspection program is that licensees are responsible for the proper con-struction and safe and efficient operation of their nuclear power plants. The total government-industry systeni for the inspection of comercial nuclear facilities has been designed to provide for multiple levels of inspection and verification. Licensees, contractors, and vendors each participate in a quality verification process in comp 11ence with requirements prescribed by the

- NRC's rules and regulations (Title 10 Code of Federal Regulations). The NRC performs an overview of the commercial nuclear industry by inspection to r

determine whether its requirements are being met by licensees and their contractors while the major inspection effort is performed by the industry within the framework of ongoing quality verification programs.

The licensee is responsible for developing and maintaining a detailed quality assurance (QA) plan with implementing procedures pursuant to 10 CFR 50.

Through a system of planned and periodic audits and inspections, the licensee is responsible for assuring that suppliers, contractors and vendors also have suitable and appropriate quality programs that meet NRC requirements, guides, codes and stenoards.

The Vendor Inspection Branch (VIB) reviews and inspects nuclear steam system suppliers (NSSSs), architect engineering (AE) firms, suppliers of products and-services, independent testing laboratories performing equipment qualification tests, and holders of NRC licenses (construction permit holders and operating licenses) in vendor-related areas. These inspections-are performed to assure that the root causes of reported vendor-related problems are determined and appropriate corrective actions are developed. The inspections-also review the

vendors' conformance with applicable NRC and industry quality requirements, the adequacy of-licensees' oversight of their vendors, and that adequate interfaces exist between licensees-and vendors.

The VIB inspection emphasis is placed on the quality and suitability of vendor products, licensee-vendor interf ace, environmental qualification of equipment, and review of equipment problems found during operation and their corrective action. When nonconformances with NRC requirements and regulations are found, the inspected ~ organization is required to take appropriate corrective action and to institute preventive measures to preclude recurrence. When generic

-implications are identified NRC assurer,that affected licensees are informed through vendor reporting or by NRC generic correspondence such as information notices and bulletins.

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The White Book is published quarterly and contains co)ies of all vendor inspec-tion. reports issued during the calendar quarter for wiich it is published.

Each. vendor inspection report lists the nuclear facilities to which the results are applicable thereby informing licensees and vendors of potential problems.

e-In' addition, the affected Regional Offices are notified of any significant c

problem areas that rey require.special attention.

ihe White Boo'k also contains a list of selected bulletins and information notices involving vendor issues.

Correspondence with contractors and vendors relative to inspection data contained in the White Book is placed in the USNRC Public Document Room, locatedinWashington,lD.C.

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REPORTING FORf%T ORGANIZATION:

COMPANY, DIVISION CITY, STATE i

REPORT.

INSPECTION INSPECTION NO.: Docket / Year / Sequence DATE:

ON-SITE HOURS:

CORRESPONDENCE ADDPESS:

Corporate Name Division ATTN: Name/ Title 1

Address City, State Zip Code i

ORGANIZATIONAL CONTACT:

Name/ Title j

TELEPHONE NUMBER:

Telephone Number NUCLEAR INDUSTRY ACTIVITY:

Description of type of components, equipment, or services supplied.

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ASSIGNED INSPECTOR Mame/ Vendor Program Branch Section M

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OTHERINSPECTOR(S):

Name/ Vendor Program Branch Section l

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APPROVED BY:

Name/ Chief - Section/ Vendor Program Branch M

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t INSPECTION BASES AND SCOPE:

A.

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

B.

SCOPE:

Summarizes the specific areas that were reviewed, and/or identi-Ties plant systems, equipment or specific components that were inspected.

For reactive (identified prcblem) inspections, the scope sunmarizes the problem that caused the inspection to be performed.

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PLANT SITE APPLICABILITY:

List plant name and docket rumbers of licensed facilities for which equipment, services, or records were examined during the inspection.

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r ORGANIZATION: ORGANIZATION CITY, STATE l

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 I

applicable to the organization being inspected.

.ONCONFORMANCES:

Shown here are any inspection results determined to B.

N be in nonconformance with applicable commitments to NRC requirements.

In 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

• 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, i

and/or unresolved items until they are closed by appropriate action, i

'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.

INSPECTION FINDINGS AND OTHER COMMENTS:

This section is used to provide significant information 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).

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.For' reactive inspections, this section will be used to summarize the disposition or status of the condition of event which caused th,e L.

inspection to be perfomed.

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PERSONS CONTACTED:

Typed, Name, Title

• present;during extt meeting l

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SAMPLE PAGE (EXPLANATIONOFFORMATANDTERMINOLOGY)

INDE FACILITY RDCRT NLMBER PAGE ABB 0;stustion Ergineerity Nuclear Power 99900401/90-01 1

Wirdaar, Connecticut Autunatic Valve Corportion 99901178/90-01 16 Novi, Michigan Cooper-Beasuner Recipm2,atirg 99900317/89-01 38 Grove City, Pengwylvania DIVESCO, Incorporated 99901117/90-01 48 Jackson Mississippi ham Services, incorporated 99900505/90-01 SB New York, New York Power Distrilution Technology 99901128/88-01 68 Rockville, Marylard Square D coupany 99900367/88-01 77 Peru, Ildiarm The Rmkh stos Coupany 99900277/90-01 83 How ILyon, Counticut 4

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UNITED STATES

'g NUCLEAR REGULATORY COMMISSION ti, W ASHING TON. D. C. 20Lb6 8

May 21, 1990 Docket No.: 99900401/90-01 i

i Mr. A. E. Scherer, Vice President i

Nuclear Quality.

ABB Combustion Engineering Nuclear Power 1000 Prospect Hill Road Windsor, Connecticut 06095 t

Dear Mr. Scherer:

This letter addresses the inspection of your facility at Windsor, Connecticut ponducted by Messrs. R. C. Wilson and R. N. Moist of this office on March 12 through 15, 1990, and the discussions of their findings with you and members of your staff at the conclusion of the inspection.

The inspection was conducted to review your control of procurement activities in selected areas. Areas examined during the NRC inspection and our findings are discussed in the enclosed report. This inspection consisted of an exami-nation of procedures and representative records, interviews with personnel, and observations by the inspectors.

Implementation reviews were conducted pr'marily for three suppliers of in-core instrumentation: Whittaker Corporation, Reuter-Stokes Canada, and Imaging and Sensing Technology Corporation.

This and previous' inspections found that the implementation of your quality assurance (QA) program failed to meet certain NRC requirements.

Discrepancies in your procurement of safety-related thermocouple cable assemblies, including electrical connectors, constitute a nonconformance in your implementation of 10 CFR Part $0, Appendix B, which was contractually imposed on you through purchase order contracts with NRC licensees. The enclosed Notice of Nonconformance and inspection report provide details.

The inspectors determined that you have already taken adequate corrective action and preventive measures to address the identified nonconformance and, therefore, no response to the Notice of Nonconformance is required. A second area, related to-audits of holders of certificates from the American Society l

ofMechanicalEngineers(ASME),isconsideredanunresolveditemandwillbe addressed in a future NRC inspection.

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46 Mr. A. E. Scherer.

6 Should you have are questions concerning this inspection, we will be pleased to discuss them with you.

Sincerely,

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1. E. William Erach, Chief Vendor Inspection Branch V

Division of Reactor Inspection and Safeguards Office of Nuclear Reactor Regulation

Enclosures:

1. ' Notice of Nonconformance 2.

Inspection Report 99900401/90-01 t

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NOTICE OF NONCONFORMANCE Based on the results of an NRC inspection coaducted on March 12 through 15, 1990, it appears that certain of your activities were not conducted in accordance with NRC requirements which were imposed on you through purchase order contracts with NRC licensees, 1.

Criterion Y of Appendix B to 10 CFR Part 50 states:

" Activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, and or i

drawings. -Instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished."

l Section 6.5.1 of C-E Quality Specification 00000-WQC-11.1, Revision D, states in part:

" Activities affecting quality, including inspections, tests and special processes, shall be controlled and performed in accordance with written procedures, instructions or drawings. These shall be submitted to, reviewed, and accepted by C-E prior to the start of fabrication."

l Contrary to the above, in numerous purchase orders with the Electronic Resources Division of Whittaker Corporation from 1984 to 1987, a silicone a

fluid treatment for which no written procedure existed was applied to certain safety-related electrical connectors.

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Criterion VII of Appendix B to 10 CFR Part 50, states in part:

" Measures shall be established to assure that purchased material, equipment, and a

services, whether_ purchased directly.or through contractors and subcontractors,-conform to the procurement-documents. These measures for source evaluation and shall include provisions, as appropriate, furnished by'the contractor or selection, objective evidence of quality i

subcontractor, inspection at the contractor or subcontractor source, and j

. examination of products upon delivery."

Sections 4.1.1 and 4.1.1.1 of C-E Quality Specification 0000-WQC-1.1.1 Revision D, state:

"The suppliers shall employ a quality program complying with the contract requirements. The quality program shall be delineated in writing. The program shall, as a minimum: Assure adequate Quality throughout all areas of contract performance by the supplier."

Contrary to the above,19 safety-related electrical conne:: tors were shipped in 1987 on Purchase Order 9770018-71186 by the Electroni::

Resources Division of Whittaker Corporation with the written approval of C-E QA and Engineering, prior to submittal of any connector drawing for C-E review and approval.

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'ORGANIZA110N: ABB-COMBUSTION ENGINEERING NUCLEAR POWER WINDSOR CONNECTICUT norum IN5PECTION INSPECTION NO.: 99900401/90-01 DATE: March 12-15, 1990 ON-SITE HOURS: 60 CORRESPONDENCE ADDRESS: Mr. A. E. Scherer, Vice President Nuclear Quality ABB Combustion Engineering Nuclear Power 1000 Propect Hill Road Windsor, Cennecticut 06095 ORGANIZATIONAL CONTACT:

Steven A. Toelle, Manager, Operating Reactor Licensing TELEPHONE HUMBER:

(213)285-5213 NUCLEAR INDUSTRY ACTIVITY:

ABBCombustionEngineering(C-E)hadNSSScontracts for 16 domestic reactors, and has support service contracts for aparoximately 40 rett'ers worldwide. C-E has support contracts to previde nuclear leplacement parts which is also a significant business area.

ASSIGNED INSPECTOR:

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R. C. Wilson, Senior Reactor Engineer, Reactive InspectionSectionNo.2,(RIS-2), Vendor Inspection Branch OTHER INSPECTOR (S):

R. N; Moist, QA Specialist, RIS-2, Vendor Inspection Branch i

APPROVED BY:

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f-to-y o U. Potapovs, Chief, RIS-4, Vendor Inspection Branch Date INSPECTION BASES AND SCOPE:

A.

BASES:-

10 CFR Part 21 and 50 B.

SCOPE:

Review documentation and interview personnel relating to C-E's control of procurement activities.

Programmatic review was supplemented byl vendor-specific reviews primarily covering purchase orders (P0s) with the following suppliers of.in-core instrumentation: Whittaker Corporation Electronic Resources Division; Reuter-Stokes Canada Division of Furst Manufacturing Corporation; and Imaging and Sensing Technology Corporation.

PLANT SITE APPLICABILITY:

All C-E plants and other sites with C-E services.

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ORGANIZATION: ABB-COMBUSTION ENGINEERING NUCLEAR POWER

-WIN 050R, CONNECTICUT REPORT INSPECTION

,NO.:

99900401/90-01 RESULTS:

PAGE 2 of 12 A.

VIOLATIONS:

None B.

NONCONFORMANCE:

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The following examples demonstrate inadequate implementation of 10 CFR Part 50, Appendix B criteria in applyin9 C-E Quality Specif-i ication.0000-WQC-11.1, Revision D, to procurements of safety-related thermocouple cable assemblies from the Electronic Resources Division of Whittaker Corporation.

(a) Contrary to Criterion V of Appendix B and Section 6.5.1 of i

the C-E specification, which require that activities affecting quality must be controlled by written procedures, a silicone fluid treatment for which no written procedure existed was applied to certain electrical connectors between 1984 and 1987.

(90-01-01)

J (b) Contrary to Criterion Vil of Appendix B and Section 4.1.1.1 of the C-E specification, which require a vendor program that assures quality throughout all areas of contract performance, 19 connectors were shipped in 1987 prior to submittal of any connector drawing for C-E review and approval.

(90-01-02)

C.

UNRESOLVED ITEMS:

1.-

After review of C-E's QA manual and procedures and discussion with C-E. personnel, the inspectors were not able to conclude whether C-E requirements for audits of ASME certificate holders were consis-tent with Regulatory Guide 1.28, Revision 3 and Information Notice 86-21.

(90-01-03)

D.

. STATUS OF PREVIOUS INSPECTION FINDINGS:

1.

Unresolved item 99900401/89-01-01, control of procurement activities, was reviewed.

Inspector review of C-E practices in this area resulted in the identification of Nonconformances 90-01-01 and'02. The inspector's review determined corrective actions: already taken by C-E to be adequate. Therefore, the previous unresolved item and the identified nonconformances were closed during this inspection.

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ORGANIZATION: ABB-COMBUSTION ENGINEERING NUCLEAR POWER WINDSOR, CONNECTICUT i

l REPORT INSPECTION NO.: 99900401/90-01 RESULTS:

PAGE 3 of 12 E.

INSPECTION FINDINGS AND OTHER COMMENTS 1.

Procurement Process a.

Procurement Document Control (Non-Spare parts)

The inspectors reviewed System 4, Revision 0 of the Nuclear Power Businesses (NPB) Division Nuclear Quality Assurance Manual (QAM) titled " Procurement Document Control" (PDC) and NPB Quality Assurance Procedure (QAP) 4.1, Revision 0 titled i

" Preparation, Review and Approval of Purchase Orders."

System 4 describes the controls for the preparation and i

issuance of procurement documents to assure that applicable recuirements are addressed, and QAP 4.1 defines the pro-ceture for document control to assure that appropriate technical and quality requirements are specified.

C-E uses three procurement categories during the procurement process. Category 1 pertains to safety-related items and services, Category 2 to commercial grade items which will be upgraJed for safety-related applications, and Category 3 to nonsefety-related items / services.

The major focus of this inspection was on procurement Category 1.

For Category 1 procurements, a purchase requisition (PR) is

. written. The PR which becomes a procurement package, includesthefollowing: statement of the scope of work to be performed, technical requirements, quality class and QA program,;10 CFR Part 21 requirements, documentation require-ments, nonconformance, proprietary information, andirelease l

status. When required, an independent technical reviewer reviews the procurement package to verify that the order includes appropriate technical requirements. Af ter resolu-tion of any comments the independent reviewer annotates the PR, which completes the design quality requirements of the document. Then a Nuclear Quality System (NQS) representa-tive performs a quality verification of the document.. The-NQS representative adds information to the document such as:

e the supplier's QA manual and revision level approved for use, methods to be used to accept the item or service, i

whether or not.the item will be drop shipped, receiving inspection at C-E, and witness hold points or integratd i

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ORGANIZATION: ABB-COMBUSTION ENGINEERING NUCLEAR POWER WINDSOR, CONNECTICUT REPORT INSPECTION NO.: 99900401/90-01 RESULTS:

PAGE 4 of 12 manufacturing quality plan (IMQP) if required for in-process curveillance. After completion of this review and resolution of any comments the NQS representative annotates the PR, which then signifies that the quality requirements have been completed for the PR.

The project manager then reviews the PR for administrative compliance with customer requirements and authorization for procurement. After all the required reviews, the purchasing agent issues a P0 to the approved supplier. Documents used to control the technical interface are the Technical Chan Request (TCR),DeviationfromContractRequirements(DCR)geand theRequestforApprovalandReview(RAR).

Changes to the procurement documents are controlled by requisition supplements which are subject to the same degree of control as used in preparation of original documents. The inspectors reviewed several procurement packages issued to Whittaker Corporation, Electric Resources Division; Reuter-Stokes Canada; and Imaging and Sensing Technology Corporation to verify implementation of the procurement program. The inspectors verified that C-E customer requirements were passed on in the P0's to C-E suppliers, b.

Procurement Document Control (Spare Parts)

The inspectors reviewed System 4'.0; Rev.1 of Nuclear Spare Parts (NSP) Quality Assurance Program Description (QAPD) titled " Procurement Document Control." System 4.0 describes procurement document preparation, review and change control to assure that appropriate technical and quality rotuire-t ments are specified. The procurement process is essentially the same as for non-spare parts except for a few minor changes. The spare parts program uses an Order Processing andManagementInformationSystem(OPHIS),whichserveras the basis for electronic P0 preparation and processing.

OPHISalsoprovidesasecuremethodforprep(NSPDB)I aring nuclear spare parts orderr using the NSP Data Base and for obtaining requireu approvals prior to release of f nal procurement documents. The following guidance is used when determining the appropriate methods for parts acceptance:

(a) documentation reviews, for items of simple design involving standard materials, processes and tests, or when the part is being supplied by the original equipment I

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ORGANIZATION: ABB-COMBUSTION ENGINEERING NUCLEAR POWER WINDSOR, CONNECTICUT REPORT INSPECTION NO.: 99900401/90-01 RESULTS:

PAGE 5 of 12 manufacturer and history indicates satisfactory performance; (b)sourceverification;or(c)dropshipment/ receiving inspection, only when the item is relatively 5imple or standard in design and quality characteristics can be checked after receipt. The NSPDB establishes Category 1 technical and quality requirements for warehouse stocking of items, since customer requirements are not known. The inspectors reviewed five C-E customer P0's(Northeast Utilities, Florida Power and Light, Baltimore Gas and Electric, Arizona Power and Light, and Alabama Power Company) and verified that the technical and quality requirements were included in C-E supplier PO's.

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Approved Suppliers List The inspectors reviewed NPB QAP 7.2, Revision 0 titled "ApprovedSuppliersList"(ASL). QAP 7.2 defines the method used by C-E to presare and maintain a list of companies and organizations whic1 have been formally approved for supply of materials, items and services to NPB. The Manager, Supplier Control has principal responsibility for imple-mentation of QAP 7.2.

The inspectors reviewed C-E's basis for addition, ren. oval and extensions of suppliers as part of the maintenance of the ASL. Suppliers are placed and main-tained on C-L's ASL if they hold a current ASME certificate for-the items procured, if C-E's customer directs them to use aspecificsupplier(notapplicabletoASMEsuppliers),orif they have been evaluated and approved by C-E Quality Opera-tions. During this inspection the inspectors reviewed several P0's and verified that the suppliers were on the C-E ASL.

d.

Control of Purchased Items and Services The inspectors reviewed the procedures ' listed below:

1 NPB, NQAM, System 7. Revision 0 Control of Purchased i

Items and Services NPB, NQAM, System 18 Revision 0 Audits NSP, QAPD, System 7. Revision 1 Control of Purchased items and Services l

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ORGANIZATION: ABB-COMBUSTION ENGINEERING NUCLEAR POWER WINDSOR, CONNECTICUT REPORT INSPECTION NO : 99900401/90-01 RESULTS:

PAGE 6 of 12 NPB, QAP 7.1, Revision 0 Supplier Evaluation and Approval NPB, QAP 7.3, Revision 0 Control of Suppliers Document or Review (RAR)

Process NPB, QAP 7.4, Revision 0 Control of Supplier Nonconformances Deviation From Contract Require-ments (DCR) Process NPB, QAP 7.6, Revision 1 Source Surveillance NPB, QAP 18.2, Revision 2 Supplier Audits The inspectors reviewed C-E's procurement and materials QA program for the evaluation and control of supplier quality performance for safety-related materials and services.

C-E's Quality Operations group has the responsibility for evaluating suppliers to ensure that they have the f acilities, QA program, personnel and equipment to supply safety-related items and services.

Suppliers of safety-related items or services are required to implement a QA program in accord-ance with 10 CFR Part 50, Appendix B, as applicable to the scope of work performed.

C-E ensures the capability of suppliers of safety-related items by in-facility pre-award surveys, ASME certificates, QA program audits, historical records, annual evaluations, and triennial implementation

audits, il One area of concern was the way C-E utilized ASME surveys of certificate holders.

NRC Information Notice (IN) 86-21

" Recognition of ASME Program for N stamp holders," dated March 21, 1986, recognizes that if ASME has surveyed the supplier and issued a certificate of authorization of appropriate scope, then the supplier may be placed on an ASL.

However, this recognition applies only to the prograngtic aspects of the ASME accreditation program and C-E is still responsible for ensuring that the supplier is effectively implementing its approved QA program.

C-E's interpretation appears to be inconsistent with IN 86-21 based on conversation with C-E personnel, but their practice of conducting source evaluations appears to be 9

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ORGANIZATION: ABB-COMBUSTION ENGINEERING NUCLEAR POWER WINDSOR, CONNECTICUT REPORT INSFECTION NO.: 99900401/90-01 RESULTS:

PAGE 7 of 12 appropriate. C-E contends that Regulatory Guide 1.28 Revision 3, dated August 1985, paragra>h 3.2 does not require C-E to audit ASME certificate 1olders on a triennial basis as long as ASME is auditing the supplier. Howen r, the NRC inspectors stated that the programmatic audits being performed by ASME can only be used as supporting documentation to keep the supplier on the C-E ASL. There still are require-ments for the implementation survey or audit, and for subsequent implementation or source audits for ASME suppliers to be consistent with IN 86-21. C-E further stated that C-E performs a source surveillance inspection when a P0 is awarded to an ASME supplier.

C-E internal procedures a) pear

+o be contradictory in that NPB NQAH System 4, Paragrap1 1.1.2b states that " suppliers who have ASME Certificates of Authorization or Quality Systems Certificates (materials) may be used without survey or audit by NPB or their subcon-tractors for the items contained in the scope of their certificates," while NPB NQAM System 7, Paragraph 2.3.1 l

relating to supplier pre-award evaluations states that "ASME Section III material manufacturers / suppliers shall be evaluated and qualified based upon appropriate requirements in ASME code Article NCA-3800. The evaluations shall include a review of the supplier's QA program manual, including supporting procedures as appropriate, and a survey to verify implementation at the location of work." Based upon the discussions and the apparent inconsistencies in procedures, this. concern is designated unresolved item 90-01-03, Section C of this report.

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The inspectors selected the ASME suppliers listed below for review of source surveillance inspection reports.

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Cajon Company Copes Vulcan Target Rock Ranor Incorporation Pacific Pumps / Dresser Only two of the five suppliers had any activity (Target Rock &

CopesVulcan). The inspectors reviewed two source surveillance reports for inspections conducted by C-E at Target Rock during the 1984 - 1985 timeframe and one C-E source surveillance inspection report for Copes Vulcan during 1978. The surveillances consisted of testing and 10 r,-

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i ORGANIZATION: ABB-COMBUSTION ENGINEERING NUCLEAR POWER WINDSOR, CONNECTICUT

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REPORT INSPECTION

~NO.:

9990W01/90-01 RESULTS:

PAGE 8 of 12 visual inspection of the products.

No anomalies were noted during this review, p

The inspectors reviewed in detail documents for two safety-related suppliers, Whittaker Corporation and Reuter-Stokes E

Canada. The Whittaker Corporation audit reports and annual k ',

evaluations were reviewed for the 1976-1990 timeframe and 4

Reuter-Stokes Canada for the 1971-1990 timeframe. The inspectors verified that initial supplier implementation 4

audits, annual reviews, and; subsequent audits were per-formed, as applicable; that documented objective evidence was identified in the audit reports and checklists to verify that an implementation audit was performed; and that supplier surveillances were performed when required. The i

anomalies identified'during these~ audits were generally

-i programmatic and did not affect the product. However, a

~ major.1988' performance-based audit'of Whittaker identified l

. product-related anomalies. The inspectors' reviewed Whittaker's responses to the 1988 audit. A supplier

. evaluation summary, dated February 7, 1990, was reviewed in which C-E accepted the revised Whittaker Corporation-QA Manual.- This evaluation also placed Whittaker on C-E's ASL as, conditionally approved for su components and' assemblies (e.g.,pply of QC-1 electrical I

mineral insulated cable, connectors, cable. assemblies, heated junction thermocouple 4

probes). C-E indicated that Whittaker's status may be-upgraded to unconditional upon satisfactory completion of QA program. implementation ve "ication as determined by future-C-E audits.: It was estimaced that Whittaker Corporation i

will complete sufficient work under their revised QA pran i by May 1990 to ollow C-E to schedule implementation aud 3s.-

During annual: evaluations, C-E reviews the supplier's Quality Assurance Manual, past purchase orders, receiving 1

~ inspections reports, DCR's, Field' Action. Reports, Corrective :

. Action Reports, supplicr's file for any restrictions NRC 1

1H's and Bulletins, past audit /

outstanding QA finding. reports) pre-award surveys'(for y

, and past source sur-veillance inspection reports.

C-E's audit process appeared to be satisfactory.

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4 0RGANIZATION: ABB-COMBUSTION ENGINEERING NUCLEAR POWER 4

4 W NDSOR, CONNECTICUT l

1 REPORT INSPECTION NO.: 99900401/90-01 RESULTS:

PAGE 9 of 12 I

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Dedication of Commercial Grade Equipment j

C-E has engineering and QA groups devoted to the growing-C spare parts business. C-E briefly described-the spare parts procurement program for QC Category 2 parts -- commercial L

grade parts intended to be dedicated for safety-related use by the licensee customer. C-E provides a " designation" process intended-to facilitate future dedication'for the specific end-use application. Critical characteristics are A

included in the program, often functional in nature.

2.

< Previous Inspection Follow-up

'j j

Two NRC inspection' reports addressed concerns related to C-E t

7 Pm procurements from the Electronic Resources Division of Whittaker i

k M Corporation:- 99900401/89-01 covering a May 22-25, 1989

inspecH on of C-E, and 99901164/89-01-covering a June-20-22 n

-1989. i_.pection of Whittaker. Unresolved Item 99900401/89-01-01 covering C-2's. control of procurement activities was identified-1 m0 as a result of those inspections. The two specific concerns are sumarized in Section E.3 of the Whittaker inspection report; i

e they involve a: silicone treatment process that was used without i

(5 procedural control, and 19 connectors that were shipped prior to

-drawing review and approval by C-E.

The inspectors examined

'1

. additional re o rds and interviewed personnel at C-E concerning i

these areas r, part of the inspection.of C-E's procurement controls.--

For the connectors procured in 1987 on PO 9770018-71186,-the P0 originally specified an exivting Whittaker connector design by part number, and specified a modified rear sleeve..A new Whittaker: drawing was generated, and a P0 supplement was initiated specifying the part number covered by the new drawing.

The next day,-prior.to submittal of the.new drawing, C-E approved l

shipment'of 19. connectors. When the new drawing was submitted, C-E noted that the method.of ground connection was different than originally specified and requested 'a change.

]

The new drawing was then revised, and five connectors were built s

E to the approval revision and shipped. Thus 19 connectors were built to the unapproved drawing with a ground contact crimp i

sleeve, and 5 were built to the approved revision without the F

-f t

~

g-12 e

Y ORGANIZATION: ABB-COMBUSTION ENGINEERING iiUCLEAR POWER WINDSOR; CONNECTICUT INSPECTION

- REPORT RESULTS:-

PAGE 10 of 12 NO.: 99900401/90-01 sleeve. The discrepancy was discovered by a third party to whom the connectors were sent for additional assembly work, prior to delivery to the user. -The 5 connectors were then modified or replaced so that all 24 were alike. The inspectors iso noted that the 19 connectors were inspected and shipped shortly after Whittaker's facilities were relocated. This event appears to be

-an isolated incident of breakdown of C-E's procurement control system;-no similar occurrence was found by the inspectors. The e

event, however, exemplifies deficiencies discovered in C-E's May 1988 audit of Whittaker with respect to inadequate procedural controls.

The silicone. treatment process concern is-addressed in detail in

.the-Whittaker:insp'ection report, 99901164/89-01. Briefly, the treatnent was applied to connector _ headers with low insulation

-resistance.

It was employed on connectors built for C-E from 1994 to 1987, and was considered proprietary. However, no written procedure was prepared until April 1537. Report 99901164/89-01 addressed the impact of the silicone treatment processontheenvironmentalqualification(EQ)ofthecon-nectors,-_and concluded that the possibility of an EQ concern is emote.

(referenceNonconformance 90-01-01 and 02) a.:

Environmental Qualification During the present inspection the NRL u.sgtt?rs incvsed on whether the cumulative effect'of multiple QA deficiencie's 4

jeopardizes the EQ of cable and connector assemblies built since the EQ type test specimens were made in 1984; that is, are the 1984 type test specimens truly representative of subsequent production equipment?

In November 1984, C-E conducted a technical and QA audit of Whittaker's design and process operations for newly-developed heated junction thermocouple (HJTC) assemblies.

The. technical audit report covers areas very similar to '

findings in the May 1988 C-E audit, which are also con-sistent with the 1989 NRC inspection of Whittaker and C-E.

The deficiencies are primarily related to' documents rather than to hardware.. Taking into account also the numerous other C-E audits and surveillances reviewed by the NRC inspectors,' hardware concerns. appear.to have been of three types:

small errors in assembly overall lengths, reversed I

l sif

l% '

1

0RGANIZATION: ABB-COMBUSTION ENGINEERING NUCLEAR POWER WINDSOR -CONNECTICUT c

REPORT INSPECTION NO.: 99900401/90-01 RESULTS:

PAGE 11 cf 12 connector pin assignments, and~the connector ground sleeve concern addressed above. The existence of these concerns 4

and the absence of others reflect a consistent pattern of

. production personnel building a product with high basic

[

Lintegrity but with defects in application-specific L

' characteristics typical of inadaquate production controls.

Product quality over the time span in question appears to o

derive from the skill of the workers more than from an ongoing control system. Even for the types of defects i

noted,-only the connector ground sleeve concern was undetected, causing shipment of discrepant hardware. The relatively simple nature of the product (almost no moving parts)..the long. history of aerospace use of the proprietary i

~

product. and the care exercised by C-E_in qualifying the

. nuclear. products (especially the new HJTC's) all con-

-tribute to the conclusion that environmental qualification

remains-acc6ptable.

b.

Nonconformance 1

Unresolved item 99900401/89-01-01 is now clas'sd ha Noncon-formance_99900401/90-01-01 and 02 described in Section B.1

~

of this report. :The noted examples were corrected in 1987 59 -

and 1988. Significant actions ~to prevent recurrence are cited

in the 1989.NRC inspection: reports. 'During this inspection the

'NRC. inspectors-reviewed C-E's. Suppliers Evaluation Summary and i

Supplier Evaluation Checklist for Revision L of-Whittaker's QA manual. Based on these evaluations on February 7, 1990, C-E accepted the revised QA manual and designated Whittaker as a conditionally approved: supplier.

Implementation audits are

-expected'in mid-1990.

In view of the actions taken by C-E and Whittaker, the nonconformances are considered closed during this

. inspection.

F.-

PERSONS CONTRACTED:

• 1+ A. E.- Scherer, Director, Nuclear Licensing

• + S. A. Toelle, Manager, Operating Reactor Licensing

• + J. M. Betancourt, Senior Consultant, Licensing

• + R.;J. Fitzgerald,; Director, Quality Operations g

• + D. L. Levan,-Manager,1 Supplier Evaluation.

J R. W. Driscoil, Senior Engineer, Quality.0perations

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_-___l___11_______

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w ORGANIZAT:.'ON:'-ABB-COMBUSTION ENGINEERING NUCLEAR POWER WINDCOR,: CONNECTICUT w

REPORT ~

a INSPECTION

,7..;

NO. :' : 99900401/90-01 oESULTS:

PAGE 12 of 12' j

x C

R. F. Ryan, Senior Enginee'r, Quality Operations-

. M. W. Stewart, Supervisor, Spare Parts QA~

'S. L. Mara, Senior Engineer, Quality Operations-A. J. Kostenko,l Senior Engineer, Quality Operations i,

+ J. P.. Pasquenza, Manager,. Quality Programs

?

C. B. -Fowler, Supervisor, Quality Programs.

J.. M. Burger, Manager, Reactor Mechanical Systems

.M. J. Linden, Senior-Engineer, Reactor Mechanical Systems'

• +10. E. McConnell, Manager, Nuclear Spare Parts

)

il T.S. Bernard,? Supervisor,TechnicalSupport(SpareParts).

P.LA. Hellandbrand Engineer, Spare Parts C. Keefer, Engineer, Spare Parts

• + M. J.,Sponza.: Engineer, Spare Parts l

1 i

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• Attended. entrance meeting on March.12, 1990

+ Attended exit meeting on March 15, 1990

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NUCLEAR REGULATORY COMMISSION

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WASHINGTON, D. C. 20$56 June 13,1990

+,

Docket No ' 99901178/90 m 13 Mi Mr. T. Hutchins, President

?

Automatic Valve Corporation 41144 Vicenti Court lL Novi, Michigan 48050-9435 gu 4

Dear Mr. Hutchins:

'This letter addresses; the inspection of your facility at Novi, Michigan, conducted by Messrs K. R. Naidu, H. L. Ornstein and T. J. Carter, of this office.on January 29, 1990 through February 1, 1990, and the discussions 4

of their findings with'you and Mr. D. Swinton at the conclusion of the-inspection.

I E

LThe purpose:of the inspection was to review the implementation of your quality

~j assurance program in selected areas and observe activities related to the i

manuf acture, assembly, and_ testing of pneumatic control assemblies intended for'use in= safety-related actuators to operate' main steam isniation-

' valves installed =in nuclear power plants.

Areas. examined during the NRC inspection and our findings are-discussed in the enclosed report. This' inspection consisted of an_ examination of procedures 4

and' representative records; interviews with personnel,_and observations by the S

l inspectors.c During this inspection,- it was observed that your quality _ assurance (QA), program

.and its implementation _ failed.to meet certain NRC~ requirements.

Specifically, the inspectors -determined that'your QA program was inadequate..in the areas des-cribing the authority and duties of persons performing safety-related activities; j!

measures to convey quality: requirements contained in customer purchase orders to subvendors; and' measures to verity the adequacy and' validity of documentary evi-dence related to purchased material. Furthermore,.the inspectors determined that.

your' procedure to implement 10 CFR Part 21 was inadequate to-evaluate and document evaluations of potentially reportable items. 'The: specific findings and references q

to the pertinent requirements are identified in the enclosures to this letter.

i RTbe enclosed Notice.of. Violation is sent to you pursuant to the provisions of W

TSection 206 of the Energy Reorganization Act of 1974.~ You are required to submit to:this office within 30 days of this letter a written statement

'o containing: L(1) a description of steps-that have been or will be taken to correct _these items; (2)=a description of steps that have or will be taken to prevent recurrance; and (3)~ the dates your corrective actions and preventive measures were or wills be completed. We will consider extending the response o

time if you can'show us good cause to do so.

n; 16

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x 4Mr.TT.'HutchinsL '

1 You are also requested to submit a written statement for each item which appears 1

D in the enclosed-Notice of Nonconformance.

e The' responses requested by this letter are not subject to the clearance f^:

.-procedures:of the Office of Management and Budget as required by the Paperwork i

Reduction Act of 1980, PL 96-511.

accordan'ce with 10 CFR 2.790 of the Comission's. regulations, a copy of

nis letter and +.he enclosed inspection report will be placed in the NRC's

~

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'Public' Document Room.

k

5hould'you

ha've any: questions concerning this inspection, we will be pleased M.

- to discuss them'with you.

q, hy Sincerely, W

i

. William Brach, (@hief-

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c Vendor Inspection Branch Divisbn of Reactor Inspection and Safeguards-Office of Nuclear Reactor Regulation

Enclosures:

1.

Appendix AlNotice of Violation 2.

Appendix B-Notice of Nonconforman:e 3.

Appendix C-Inspection Report 99901178/90-01~

4 I

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Autunatic Valve Corporation-a:

Novi, Michigan

~

APPENDIX A l

.O i

11 NOTICE OF VIOLATION As a result of. the inspection conducted on January 29 through February 1,

.1990, and in accorcance with Section 206~of the Energy Reorganization Act of 1974 and its implementing regulation 10 CFR Part 21,_ the following violation 1

was identified and categorized:in accorcance with the NRC Enforcement Policy (10CFP.Part2.-AppendixC),49_FR8583(March 8,1984):

Section 21.21 of.10 CFR Part 21, dated-August 21, 1987, states, in part: "Each ins 11 vie:1, corporation,' partnership or other entity subject to the regulations in this part shall adopt appropriate procedures.to:

(1) Provide for:

(1) Evaluating deviations or-(ii) informing the licensee or purchaser of 1

.the.deviationlin order that the licensee or purchaser may cause the deviation a

1 to be evaluated unless the deviation has been corrected; and...Any advice-related totthe defect or failure to comply.about the facility, activity, or basic component.that has been, is being, or will be given to purchaser or licensees."

Section 21.51 of 10 CFR Part'21 states in part:. "Each Licensee of a facility 1

ior activity subject to the regulations in this part shall maintain such records in ccnnection with the. licensed facility or activity as may be required to assure compliance with the regulation'in this part."

Contrary to the:above, the Automatic Valve' Corporation (AVC) did not establish a procedure to evaluate, document, and report to.their customers at least three

~

deviations-involving >1ubricants, design changes,-and use of U-cup seals in the neumatic control assenblies that had been supplied to nuclear power plants.

p(90-01-01)

This is a' Severity Level IV Violatiot ' Supplement VII)..

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18

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3 i

JAuiosiaticValveCorporation 3

i Novi, Michigan 4

APPENDIX B 4

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. NOTICE OF NONCONFORMANCE r

m p.7 Il Based on the results of an NRC inspection conducted on January 29 through

(

February

,1990at_theAutomatic.ValveCorporation(AVC),itappearsthat certain o your activities were not conducted in accordance with NRC

(

i requirements which were ' imposed on you through purchase order contracts with b'

.NRC-licensees.

Criterion' ll of Appendix B-to 10- CFR Part 50 states, in part, "... establish...a quality assurance program which complies with t'it requirements of_this' appendix.

,"" lThe program shall be documented by written polteies, procedures, or instructions s

• A

=and shall be carried out...in accordance with those policies, procedures, or Linstructions..;The quality; assurance program shall provide control over activities affecting;the quality of the id -+1f Md structures, systems, and

. ; components, to'an extent consistent with

i v rtance to safety."

l-Thi AVC quality assurance manual-(QAM), Revn ev., 8, dated July 13, 1989, is inadequate in the following areas:

!1. 1 Criterion I

" Organization," of Appendix B to 10 CFR Part 50, requires, in part, that the authority and duties of persons performing activities affecting safety-related: functions ~of components be-clearly established and delineated in writing.

Contrary to the above, the AVC'QAM does not reflect the present -

~

iorganization and does not clearly describe the authority and duties of persons currently. performing safety-related activities. -

(90-01-02)'

2.-

Criterion IV, " Procurement Document Control," of Appendix B-to

'10:CFR Part-50, requires.that measures be established-to assure that applicable regulatory requirements, design bases, and other

~

.requira.ments which are necessary to assure adequate quality are suitably included or referenced in the documents for procurement of

- material, equipsent, and services, whether purchased by the

• rnlicant or its contractors or subcontractors.

0

.rary to the above,'the AVC-QAM doeslnot require that quality s

.6surance requirenents received with customer purchase orders be impcsed,in turn, on their subvendors for the supply of materials such as steel, elastomers, and gaskets.

(90-01-03) 4 i(y 19

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L 3'. - Criterion Vil, " Control of Purchased. Material -Equipment, and l,

Services," of Appendix B to 10 CFR Part 50, requires that measures be L

established to assure that-purchased material, equipment and services 4

' conform to the procurement' documents. These measures shall includa 1 objective evidence of quality furnished.

,' g..

s.

Contrary to the above, AVC accepted cortified material test

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reports and certificates of. conformances sup(plied by) vendors without ~

. verifying the validity of such documents.

90-01-04 4

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' ORGANIZATION:: AUTOMATIC' VALVE CORPORATION:

-NOVI; MICHIGAW j

s I

REPORT:

. INSPECTION INSPECTION NO.:. 99901178/90-01J DATE: 1/29-2/1/90 l0N-SITEHOURS: 8 14 CORRESPONDENCE ADDRESS:/ T. Hutchins, President X

Automatic Valve Corporation 41144 Vincenti Court y

Novi, Michigan 48050-0435

,k, i

ORGANIZATIONAL CONTACT:. T. Mutchins i

TELEPHONE NUK3ER:

313-474-6700 L,,

T HilCLEAR INDUSTRY ACTIVITY: Manuft :tures pneumatic control assemblies for safety-related valve actuators.

}.

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if 4

J ASSIGNED 4 INSPECTOR:

[M'

$/3 /90 sfy X. R. Naidu, Reactive Inspection,Section No.1 Date.

o NR (RIS-1)

R, OTHERINSPECTOR(S):

H. L. Ornstein, Reactor Operations Analysis Branch, AE00 L

T.p. Carter, Events Assessment-Branch, DOEA l

he lTw APPROVED'BYf:'

ar.

2

(/'/ /d

.G. {walina, Chief,:Special. Inspection Section, Date a-

.Vendo: Inspection-Branch o

' y.

INSPECTION BASES AND SCOPE:

i m

A.

BASES: 110-CFR Part 21 andi10-CER 50 Appendix B.

- I o

T B.o SCOPE:~. Review the' implementation of the quality assurance 1 program in L-m

' selected' areas,= actions' taken on previously reported failures at the

.Susquehanna and Hope Creek nuclear power plants /and: observe activities

' associated with the manufacture of pneumatic control assemblies.

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,e PLANT SITEJAPPLICABILITY: All plants which utilize valve actuators with Li, pneumaticicontrol.' assemblies manufactured by Automatic Valve

'm Corporation.

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g hi 10RGANIZATION::iAUTOMATIC VALVE CORPORATION t

i NOVI, MICHIGAN:

t m

a 1 REPORT; INSPECTION 1 99901178/90 RESULTS:

PAGE 2 of 17

-j fNO.:

7A.

VIOLATIONS:

{

Contrary to Sections 21.21 and 21.51 of 10 CFR Part 21, Automatic' l

o "'

. Valve Corporation (AVC) had neither developed nor implemented an T-adequate' procedure to evaluate, document, and report deviations and t-17, defects detected in the pneumatic control assemblies'(PCAs) manu-J h'

WP ifactured and: supplied by AVC to nuclear power plants for use in M

' safety-related1 applications. Examples of such deviations not o

U

-: properly evaluated by AVC include lubricants, U-cup' seals, and design j

gg changes.

(Violation 90-01-01)

[

t.. :-Nonconformance: -

B w

JContrary to-Criterion 11 of -10 CFR 50, Appendix B, AVC did not l%

t establish and implement an adequate quality assurance program as_

$9 documented in their. quality essurance manual (QAM), Revision B, dated c

k E

g N uly'13, 1989,Las evidenced by the following:

V' 1._

Contrary to Criterion 11of 10 CFR 50, Appendix B, the authority "l

and duties of persons performing activities related to quality 1~'

g in-the manufacture of safety-related PCAs were not clearly g

-established and delineated in writing.

(Nonconformance

[

90-01-02) g 2.

Contrary to Criterion'IV'of 10 CFR 50, Appendix B, measures were notiestablished by AVC to include the applicable: quality requirements and 10 CFR Part 21 reporting: requirements listed in

=

' safety-related purchase orders-(P0s) received by:them in their=

P0siissued to their subtier suppliers.

(Nonconformance 90-01-03) um E

.were not established to verify that documentary evidence, test-3.

Contrarys to Criterion VII of 10 CFR 50, Appendix B,' measures-r

' reports, certificates of conformance,-and' certified material test reports received with purchased material met the P0 requirements. (Nonconformance 90-01-04);

l..

J

-C.

Unresolved item:

AVCistated that the existing QAM-is being revised to incorporate-d 1.

• g' provisions for. signatures and dates in assigned spaces in drawings to: identify;the individuals who drafted and verified r

the drawing instead'of.having:the Computer Assisted Design F

s L

machine print the initials of the' individuals and the~' dates when.

[

the draf ting land' verification activities were performed. This a

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-cl M-ORGANIZATION:' AUTOMATIC VALVE CORPORATION-

NOVI,. MICHIGAN-i REPORT!

INSPECTION

NO.:l:99901178/90-01.

-RESULTS:

PAGE 3 of 17 o

matter was identified as an unresolved item in the Ralph

,i

'A.H111er'(99901170/89-01) inspection report.

Pending review of t M revised QAM, this matter is considered an unresolved item.

(UnresolvedItem 90-01-05).

Or,.

D.

DETAILS:-

, l'. -. Purpose, u

- This? inspection was conducted to determine the adequacy of the f

implementation:of:the AVC quality assurance (QA) )togram

' e relative: to the manufacture of-PCAs. PCAs are sunssemblies of pneumatic / hydraulic 1 valve actuators and other types of ' valve gip ^

actuators.1 The PCA, as a subassembly of a main steam is'olation 4

valve;(MSIV) operator, directs compressed air to the top.or

y' bottom of..a piston inside the. pneumatic cylinder of the HSIV sperator, to close or open the MSIV.. Actuation of the pneumatic

cylindernis initiated by the motion of solenoid ~ operated valves (S0V).

,y('

n.

The following events; which prompted this inspection were

reviewed.

a.o On October 11, 1989, a faulty S0V in an AVC PCA caused one MSIV.to inadvertantly close during power operation at the LPeach Bottom nuclear power station. The faulty S0V was

repla.ced., Disassembly-and-examination of the faulty S0V indicated'that-a spring had been inserted upside down, s

.b. '

On: February 24,.1987, one MSIV failed to close:at the Hope Creek nuclear power station due to a faulty.50V. General 1

4-

.?_

-Electric Nuclear: Energy (GENE). San Jose,-California, the

supplier-of?the MSIV operator. examined:the failed S0V.

LGENE,'in a letter (GP-87-032), dated May 7,1987, informed i

Hope Creek that "the cause of failure appears to have been the, accidental inclusion of a foreign material, probably a-u loose' seal,:in:the solenoid cavity between the plunger and q

the upperlorifice."

c.

-On~ July 5, 1986, an outboard MSIV-failed a quarterly exer-cise test,at the Susquehanna Unit l' nuclear power plant.

s The failure was caused by the use of Houghton H-620 type lubricant which corroded the unanodized aluminum housing of theLPCA.

4 j -

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L ORGANIZATION: AUTOMATIC. VALVE CORPORATION 1

(

NOVI, MICHIGAN pf, JREPORT -

RESULTS:

PAGE 4 of 17 INSPECTION

(.},

LN04,,99901178/90 '

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On December 10, 1986, several MSIVs failed a monthly func-MD tional closure test at the Susquehanna Unit 2 nuclear power plant. It was reported that the failure was caused by the -

use of Houghton H-620 type lubricant which corroded the f,

unanodized' aluminum housing of the PCA.

f+

UJ.

2.

Background Information on AVC

~

AVC manufactures' general purpose, industrial pneumatic and vacuum

+

.' service valves.';TheLPCAs manufactured by AVC for nuclear-power V

- plants are for safety-related applications 'and are less than 7~ percent of the total:sa'les volume; Documents-at AVC indicate v,Bt,

that. initially ~AVC, received P0s to supply PCAs exclusivel the Nuclear Energy Division of General Electric Company (y fromGENE),.

j y

Sai Jose, California. However, for the last 5 to 6 years, AVC

-ale % Accepted safety-related P0s directly from licensees.

[yr N pping! documents indicate that AVC ships PCAs to either the

-acilitywherethevalveactuatorsaremanufactured(for y

< installation and test as en integral actuator unit), or

)

directly.to:the' nuclear power. plant. AVC also furnishes spare

?

H

, parts for PCAs' supplied by ' them.

3.

-Review of Implementation of 10 CFR'Part 21 Requirements

'The inspectors reviewed ?the AVC's program for' implementing t'.

10' CFR Part ' 21'.(Procedure D372-071289), and observed that copies of 10 CFR Dart 21,:Section 223(b) of the Atomic Energy Act, and-l

@M Section-2,, of the' Energy Reorganization.Act were posted.

The

inspectors determined that Procedure D372-071289'was inadequate to fully implement Sections 21.21'and.21.51;of 10 CFR Part 21, i -

cdatedl August 21,-1987. AVC did not have an adequate procedure

.for evaluating' deviations and maintaining.such records.

The inspectors. identified three areas related to the use of suitable lubricants, de' sign changes requiring the supply of' appropriate-tools,!and the use of U-cup seals in'3-and 4-way valves, which required MC's evaluation underJPart 211-These items are discussed fa the following paragraphs. (Violation 90-01-01).

3

. a.-

Review ofoProblems With Lubricants The inspectors examined the problems related to the use of.

Houghton H-620 and Super-0-Lube type lubricants in PCAs

,i installed in safety-related applications as discussed below:

i i

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ORGANIZATION:- AUTOMATIC VALVE CORPORATION H.

NOVI,-MICHIGAN-1-

y "m

I 0 REPORT RESULTS:

PAGE 5 of 17 INSPECTION

^

' NO.: 999011/8/90-01 y

SE (1): Pennsylvania Power and Light Company'(PP&L), the licensee

for the Susquehannna nuclear power plants Units 1 and 2, 4

.in a letter, dated December 30, 1985, to AVC, requested AVC's 1_

position regarding lubrication of 0-rings in the PCAs for

^4 MSIV actuators for Susqurbsnna.

PP&L's letter noted

)

6/

reservations expressed by AVC regarding the use of Houghton J:

H-620.

Their letter also noted that AVC-had reservations with a Wyle= qualification test re uit because' Wyle. encountered problems with the use of Super 0-lube during the tests.

_ m* $

(2)- In response to'the above, AVC, in a letter dated January 29, E

1986,Lto PP&L, statad thattsince 1972'they had not encoun-

tered any failures.with Super 0-Lube other then the one 7

observed during the~Wyle qualification. test. Wyle Nuclear Environmental Qualification Test-Report 17514-1, dated

-March 14,.1985, documents the results of the qualification

- test performed' on-AVC' PCAsi to meet:IEEE 323-1974, IEEE

'344-1975 and.IEEE 382-1980 for Tennessee Valley Authority

--(TVA), Detroit: Edison. Company (Fermi'2),PennsylvaniaPower and Light Company -(Limerick), and the Coninonwealth Edison Company :(Dresden and Quad Cities). The Wyle Test Report noted that Super-0-lube wasinot qualified for use in AVC PCAs because Super-0-lube' solidified during the tests.

The Wyle report attributed AVC valve qualification n

ifailures to " excessive amounts" of Super-0-lube applied The Parker Company.(Parker) port,-page xviii,; item 15).

tothevalves(Wyletestre

, the " manufacturer" of Super-0-lube, notified Wyle that Super-0-lube starts to solidify at 1.0 megarad.

Parker'also, notified Detroit Edisoni(Fermi. 2) in.a _ letter dated October 25, 1985, that Super-0-lube starts to-solidify, but at 1.9 megarads. The*AVC: letter also noted that during the above'Wyle Ltest Super-0-lube had solidified, not-only in AVC-supplied S0Vs, but.also in the 2, 3-and 4-way valves ~that are installedLin'PCAs. The 3-and 4-way valves did remain' operable;during the tests even though

'they contained Super-0-lube which was -degraded.- As a

" result'of the failure of the S0Vs with Super-0-lube, and subsequent successful testing of:Houghton H-620 Wyle recommended-the use-of Houghton H-620 lubricant.

The' records reviewed did not indicate that Wyle verified the compatibility of'Houghton H-620' lubricant with y

L 7

25

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10RGAHlZAT10N:L AUTOMATIC VALVE CORPORATION-

=NOVI, MICHIGAN M

REPORT RESULTS:

PAGE 6~of 17 i

INSPECTION NO.:-.99901178/90-01 E

!unanodized aluminum alloy components used in AVC PCAs.

AVC also noted that GENE had qualified Super-0-lube for N,'

nuclear service.

In conclusion, this letter stated that AVC's position on lubricants was to provide either 4

Super-0-lube, as specified by GENE, or Houghton H-620, as recommended by the Wyle Test Report, m

(3).OnMay28,'1986, AVC issued a Lubrication Specification (Procedure-D328-052886,. Revision 0) for Houghton H-620-type lubricant which stated that H-620 was rated " Noncom-1 patible" with unanodized aluminum by E. F. Houghton and Company, the manufacturer of H-620; and that H-620 was t

"to be used ONLY when specified by the customer."

N s

(4). On.May 29, 1986, AVC issued a Lubrication Specifications (Parker " Super-0-lube"), Procedure D329-052986, Revision 0 to indicate that drawings and manuals which refer to lubrication document D329-052986 are to have all specified parts covered.

with a light coat of " Super-0-lube."

On June-20, 1986 AVC revised their lubricant specifi-1

'(5)~~ cations for " nuclear power valves" (0129-052986, Revision 1).

It stated that unless specifically stated on the customer's purchase order, all valves for: nuclear power ~ service would be: lubricated with Super-0-lube; that the only acceptable substitute was Houghton'H-620; and that "It is the customer's-responsibility to use the correct lubricant, per their-plant specifications for.all valve maintenance." More importantly, AVC's lubrication. specification had the same caution note as the aforementioned May 28, 1986',

lubrication document which noted that Houghton-H-620 1

was rated noncompatible with unanodized aluminum by j

the lubricant manufacturer and that it was~"to be used ONLY when specified by the customer." This permission to use Houghton H-620 as an alternative: lubricant was removed in Revision 2 to D329-052986, dated December

)

i

-24, 1986.

(6) 'AVC certified for PP&L P0 6-07802-1 on June 30, 1986, that AVC manufactured eight assemblies each of 2, 3 n

and 4-way valves lubricated with Houghton H-620. This information also appeared on AVC Invoice' Number 74374, i

dated June 30, 1986. AVC lubricated the valves in the-26 TU

-}s 3

l0RGANIZATION:~~AUTOMATICVALVECORPORATION N0VI', MICHIGAN q.h 4

INSPECTION x

-REPORT <

~

RESULTS:

PAGE 7 of 17

. NO.1: 99901178/90-01.

i k SusquehannaPCAswithHoughtonH-620(June 30,1986),

C* 4 subsequent to their knowing of.the lubricant's noncom-patibility with their valves (May 28,1986). The use of Houghton H-620 was specified in the licensee's s

a purchase order.

4

.(7) On December 10, 1986, after installing the above-procured R

PCAs on MSIVs, PP&L experienced common-mode failures of the AVC PCAs for'all 16 MSIVs at Susquehanna Unit 2.

The Houghton;H-620. lubricant corroded the unanodized aluminum V'"

manifolds-in the PCAs. The-common mode failures of the AVC PCAs did not occur durina a design basis accident and it did not result in the co.romise of the public health and safety. The failures were reported to INP0.

Subsequently, on February 3, 1987, AVC provided PP&L.

with a cost effective method for repairing or replacing.

o the MSIV PCAs which had "corrosien problems caused by

.Houghton H-620 lubricant."

(8)

In.a letter ' dated: December 19, 1986, with copies to Commonwealth Edison and PP&L, AVC notified the NRC

. Region Ill office of the corrosion problems experienced at the Susquehanna without expressly mentioning 10~CFR Fart _21, and stated that PCAs for Commonwealth Edison Company's Quad Cities' plants were also lubricated with

.the same H-620' lubricant.

c

'(9)f0nFebruary 10,1987'AVCnotifiedDetroitEdison(Fermi 2) and TVA (Browns Ferry) of the problems which had been o

-experienced with Houghton H-620 lubricant (pitting and P

+

= corrosion) at Susquehanna.

The inspectors determined that contrary to 10 CFR Part 21.21(a)(1), AVC supplied PCAs to PP&L lubricated with Houghton-H-620 without informing the licensee of the existence of a deviation. As described in Section A above, the inspectors determined that AVC did not have adequate procedures for the' review, documentation and' reporting required under 10 CFR Fort 21. AVC did not have records to indicate.that AVC:

(1) adequately evaluated the advice of E. F. Houghton and impany regarding the corrosive properties of Hougi, ton H-620

' lubricant, (2) warned PP&L of the same, and (3) re-frained from using the lubricant'in PCAs furnished to PP&L. This matter is addressed in Violation 90-01-01.

4 27 e

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0RGANIZATION: AUTOMATIC VALVE CORPORATION.

.l g y, ;

NOVI, MICHIGAN N

,m REPORT-INSPECTION NO.:; 99901178/90-01 RESULTS:-

PAGE 8 of 17 g

,u

'b.

Review of Improper Tool Use-AVC changed the $0V= design during the period 1984 through

^

1985. The revised design. changed the base of the solenoid

6 guide _ tube which screws into a threaded cavity. This design it

change removed a hexagonal head nut from the SOV tube collar. 1The change required the use of a. spanner. wrench f

4

[

with two, nipples to tigh. ten or loos'en the-50V, tube in the cavity;during assembly or disassembly of the SOV. AVC incorporated-changes.into the instructions, but only I

provided the revised instructions and the spanner wrench to e

customers who ordered complete PCAs. AVC did not provide the revised instruction and the correct tool to customers who ordered spare SOVs or 50V kits. The use of pliers or c

vise grips _as prescribed in-AVC's previous instructions could lead to-damage and' malfunction of the new S0Vs. A S0V which failed at the Hope Creek nuclear power plant on 1

February:24', 1987, also sustained damage which appeared to

.have been caused by the use of. vise grips or pliers. The NRC i

-issued Information Notice 90-11, " Maintenance Deficiency Associated with Solenoid-0perated Valves" on February 28, 1990,-to alert all nuclear utilities of this problem. The i

inspectors determined that AVC did not evaluate the results of; their own inspection report, which documented-the observed idamage to the SOV, in sufficient' depth'to ascertain that the i

' root cause of the damage.could have been;from using vise grips or pliers-as described in pravious AVC; instructions.

Other utilities.using the "new" S0Vs without proper tools or

the. observation and stated that lthey will establish provisions:in'the revision to the QAM.for I

individuals-to sign-and date in the assigned spaces.to. denote

.t when'and by whom the activities were performed. This matter is q

= considered an unresolved. item.- (Unresolved-' Item 90-01-05)

'?

'7.

Procurement Document Controli Section 4 of. th'.QAM-references Procedures D119-81674 and 1

E e

D118-81670 entitled:". Customer Purchase Order" and " Supplier-Purchase Orders," respectively._ The former procedure is H

' intended to ensure that customer P0s are-_ processed in an

]

p g

accurate.: consistent, and timely manner; it neither addresses.

acknowledgenient of the quality and regulatory requirements nor

)

m-3 requires the1 translation of these requirements into AVC P0s to 4

subtier-. vendors..'The latter procedure.is intended to ensure s

that parts. procured are inspected, identified, handled, and storediin an accurate and consistent manner. The procedure neither requires AVC to impose the quality.and regulatory a

y I,

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' ORGANIZATION: AUTOMATIC VALVE CORPORATION 9-NOVI, MICHIGAN REPORT INSPECTION

- NO.:i.90901178/90-01 RESULTS:

PACE 11 of 17 irequirements specified in P0s to AVC in'AVC P0s issued to their

-vendors nor des vibes controls in place to otherwise establish necessary product quality. The inspectors reviewed several AVC

.P0s to their vendors and determined that AVC did not establish

'y and implement the requirements of 10 CFR 50, Appendix B, y" >

Criterion IV to' translate the quality and regulatory requirements received in various purchase orders issued to AVC

into AVC P0s to subtier suppliers.

(Nonconformance 90-01-03)

This matter is also discussed in paragraph 9 of this report.

8.

Control of Purchased Material, Equipment and Services-W The current-AVC QAM references Procedures D118-81674 Revision.K, dated December 28, 1989, to address the control of purchased 1

e material, equipment and services. The stated purpose of this procedure,is'_to ensure.that parts are-inspected,. identified, handled, and stored in an accurate, consistent and efficient w

manner. To ascertain if adequate measure'., were established to assure that purchased material, equipment and services, whether purchased directly or through contractors and subcontractors 3

confo m to procurement docunents, the inspectors reviewed the i

certifications received in response to AVC P0s to procure components and services. The inspectors determined that AVC had not audited:the> manufacturer or otherwise verified the accuracy of certifications or assured the traceability of purchased material for assembly in safety-related PCAs in the following-cases:

.a.

AVC-procured electrically operated SOVs from Snap-tite Incorporated (Snap-tite), Erie,' Pennsylvania. Snap-tite manufactures SOVs to meet AVC design drawings. AVC P0s to-Snap-tite do not require Snap-tite to implement a-quality assurance program during the manufacture of the SOVs to ensure consisten::y and reliability of-perform nce to assure that they meet the quality requirements of.lEEC-323 and:

IEEE-344. Until the time of this inspection, AVC'had not audited the manufacturing facilities of Snap-tite. Documents-reviewed by the inspectors indicate that Snap-tite purchases the electrical coils for SOVs from two-subtier suppliers - Five Star and Quality Coil. After the coil is ir.serted 10 the SOV and_ sealed with _ potting compound, the identity of the coil supplier cannot be determined.

For example, during the Erainent Qualification.of PCAs in 1984, Wyle supplied s

31 a

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0RGANIZATION

L AUTOMATIC VALVE CORPORATION NOVI, MICHIGAN REPORTT t

' INSPECT 10f NO.:- 99901178/90-01' RESULTS:

PAGE 12 of 17 Snap-tite with three failed S0Y coil assemblies with a 1 request to provide information on the manufacturer of the coils.

In response to=the request, Snap-tite, in a letter e

' dated November 8,1984, informed Wyle that they "cannot be certain as.to which of our two major coil suppliers made these' coils." Receipt inspections at AVC' verify that resis-tances^of the 50V coils received:are within the specified such as opera-

-acceptance limits; other characteristics,(which power plants bility of the coils at: degraded voltages 1

are' susceptible to) and insulation resistance are not verified.

Each SOV assembly, as illustrated in AVC drawings 6910-010 and 6910-020.for the 120-volt AC and the 125-volt DC S0Vs respectivelv, consists of components specified below:

(1)< One nut made of cold rolled steel and zinc plated.

(2) ' One 125-volt DC (or 120-volt AC) coil ud housing assembly. The housing was to be manufactured from Standard American Equivalent 1010 cold rolled steel with black oakite-paint. The S0V coil is wound with special magnet' wire conforming to American National Standards ~ Institute Class M35. The-SOV coil leads are to'be 22 inches-long with silicone insulation and-rated for 150-degree Centigrrie temperature, Norcast-type compound was to be used to pot the coil a

- in the housing.

(3)' One plunger guide made from 430 F-and 304-Type stainless steel.

(4) One compensating plunger and spring assembly consisting of a plunger manufactured from 430 F-Type stainless steel with a Viton ring and a 302-Type stainless steel return spring.

(5)'OneViton0-ring..

AVC could not demonstrate that the components in the 50Vs' conformed to the above specifications and are

therefore qualified to meet IEEE 323'and 344 standards.

AVC does Jnot perform an operability test.

Such a' test

^

may have' identified whether the-S0V supplied to Peach Bottem (paragraph la) was assembled correctly or with the spring upside down.

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LORGANIZATION: AUTOMATIC VALVE CORPORATION

NOV1,-MICHIGAN-7 REPORT' INSPECT 10N LNO.: 99901178/90 RESULTS:

PAGE 13 of 17 b.:

AVCprocuresViton(elastomer)0-ringsfrom-ZatkoffCompany (a~ distributor),FarmingtonHills, Michigan. AVC P0s to Zatkoff require Viton 0-rings to be manufactured from DuPont's E60C or 3M's' Fluorel 2174 base polymer, with cure dates not to exceed 12 months from the date of shipment.

- Zatkoff provides a Certificate of Conformance (CoC) that the Viton'0-rings are manufactured from either DuPont's E60C or 3M's Fluorel number 2174 and specifies cure dates of the 0-rings.-The inspectors determined that AVC had not.

at the: time of the inspection, independently verified the accuracy of any CoC. The inspectors reviewed a typical m

certificate provided by Parker Hannifin Corporation, 0-ring Division, Lexington, Kentucky, the manufacture of 0-rings to Zatkoff. A certificate, dated April 25, 1985, stated-that Parker can-certify that their compounds V-747-75 (black) and V884-75 (brown) are each made of either DuPont's E60C or

' 3H's Fluorel 2174' base polymer and both meet MIL-R-83248, Type l', Class.l.- The inspectors informed AVC that if a generic deficiency is identified in the Viton 0-rings, it wouldinot be'possible to' trace it back to the manufacturer

in the absence of the identity of the polymer base from

'which the'0-rings were manufactured.-

c.-

Similarly AVC procured'" poppet seals";with CoCs from the n

= Minnesota Rubber' Company.

The-inspectors reviewed a CoC-from Minnesota' Rubber, dated May 20,-1986.-- The CoC stated that 550 poppet seals,~Part Number 6717007A, were made with a 514AD type-compound'from either DuPont's E60C or: 3M's.

J<

Fluorel 2174 base polymer and that the cure date wast in the-2nd., quarter of 1986.

d.-

' Copper and Brass Sales, a distributor, provided~a certificate, dated April'2, 1986, to AVC for P0 43696

+

providing'the tensile strength values and chemical composition for 2024-T351' type aluminum plate and certified that~the aluminum supplied conformed to Federal

~

S)ecification QQ-A250/4E. The certificate did not indicate tie _mP 1'which provided the aluminum or include the test s;

-results.

Thefinspectors informed AVC that the above' examples. indicate instances where AVC 'did not establish measures to verify that the purchased material met the procurement documents specifi-cations.1 This matter is a nonconformance to 10 CFR Part 50,

Appendix B, Criterion Vll. '(Nonconformance 90-01-04) 33

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_0RGANIZATION: AUTOMATIC VALVE-CORPORATION I

L NOVI, MICHIGAN =

I:j{;

CREPORT INSPECTION

['

NO.: 99901178/90-01 RESULTS:

PAGE 14 of 17 y

i 9.

Review of Purchase Orders (P0s) Issued by AVC i_

The inspectors reviewed the following P0s issued by AVC for components to be assembled in safety-related PCAs and determined that AVC did not convey 10 CFR Part 21 requirements to subtier s

vendors :

to Snap-(tite a.

AVC issued P0 No. 49378, dated August 17, 1989, Incorporated, Erie, Pennsylvanic, for the supply of 30 125 VoltDC)S0VsmanufacturedtoAVCdrawing 6910-020. AVC

,also issued P0 No. 49667 dated December 12, 1989 to Snap-tite for the supply of 120 volts ac 60 Hz SOVs manufactured to meet AVC drawing 5910-010.

Both P0s required a CoC certifying that the S0Vs met the P0 requirements. Review of the above AVC P0s indicates ?. hat AVC did not convey the 10 CFR Part 21 requirements to their subtier vendors. AVC does not audit Snap-tite to establish or verify the quality and acceptability of the SOVs. The 4

. inspectors could not determine from the documents available 2

at AVC.if Snap-tite implements.a quality. assurance program during the manufacture of S0Vs. The inspectors determined W

that Snap-tite purchases solenoid coils from two subtier

. suppliers. After the solenoid coil is potted.in the 50V assembly, Snap-tite cannot establish the traceability of the coil _to the-subtier coil _ supplier.

b.

- AVC issued P0 No. 48853, dated -June-22,.1989, to S&C Tool &

Manufacturing Company Telegraph Road, Detroit, Michigan' t

for the supply of pistons manufactured to meet AVC Drawing 5140-005 and 74 bushings to. meet AVC Drawing 2195..AVC

. purchased the steel from a distributor with certified material test reports (CMTRs) and supplied the steel'to the above company: to manufacture the pistons and bushings. The P0 required the manufacturer to use the raw material-(provided to them by AVC) to manufacture the items to the drawings (supplied to them by AVC) specified in the P0.

c.;

AVC. issued'P0 No. 48935, dated August 1,1987, to L. F. Dale

' Precision Machining company (LFDPM), Farmington, Michigan,

> to supply'four manifolds to meet AVC Drawing 5140-010 and E

one manifold to meet AVC Drawing 5158-010. - AVC provided the aluminum' which was purchased by AVC with CMTRs from a distributor. The P0 required LFDPM to use the raw material supplied to them by AVC to manufacture the manifolds.

W p

34

.iq 1

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ORGANIZATION: AUTOMATIC VALVE CORPORATION NOVI, MICHIGAN REPORT INSPECTION NO.: 99901178/90-01 RESULTS:

PAGE 15 of 17 d.

AVC issued P0 No. 48989, dated July 28, 1989, to Assured Manufacturing, Madison Heights, Michigan to supply 27 manifolds to AVC Drawing C6930-010. AVC provided the raw material purchsed by AVC with CMTRs from a distributor.

The P0 required Assured Manufacturing to machine the manifolds from the raw material supplied to them, e.

AVC issued PO No. 49300, dated October 30, 1983, to R.

Zatkoff, a distributor, to supply Viton 0-rings to meet AVC part No-0914-008A. The PO equired a CoC from Zatkoff certifying that Viton 0-rings were either made from Dupont E600 or 3M's Fluorel 2174 base polymer with cure dates not to' exceed 12 months from the date of shipment.

The inspectors informed AVC that AVC did not establish adequate measures to convey the quality requireme. ts and 10 CFR Part 21 evaluation and reporting requirements imposed on AVC in various P0s issued to them by GENE and other licensees to their subtier vendors in the above examples.

The inspectors informed AVC staff that failure to include applicable regulatory and quality requirements in the AVC P0s is in nonconformance with 10 CFR 50, Appendix B, Criterion IV.

(Nonconformance 90-01-03)

10. Document Control AVC utilizes Procedure D151-3575, Revision J, dated December 29, 1989, to ensure that additions, changes, and deletions to documents are done accurately, consistently and efficiently. The inspectors reviewed the implementation of this procedure and determined that AVC adequately controlled the revisions to arocedures and drawings.

However, the inspectors identified a probl e inherent with the methodology adopted by AVC in allocating prt numbers (drawings numbers) to sub-components. During a review of procuretrent documents related to GENE and the Coninonweal?.h Edison Company (Ceco), the inspectors observed that the identification number of a component was being changed.

The inspectors obtained copies of the AVC Nuclear Parts List (from the Master Number Pook) and the procedures list. The inspectors determined that AVC maintains an up-to-date list of all parts and drawings / procedures used to control the components manufactired in-house and the parts-provided by subtier vendors. D C issues a new number to a component when the component is modified slightly, instead of issuing a revision to the original part number.

For example, a letter dated February 4,1985, to GENE stated that AVC originally 3

35

ORGANIZATION: AUTOMATIC VALVE CORPORATION NOVI, MICHIGAN REPORT INSPECTION NO.: 99901178/90-01 RESULTS:

PAGE 16 of 17 identified a 4-way valve as C5140-4H; kits to rework the valve were identified as K-C5140-4H.

Subsequently, the kit and its identification number were changed; the kit number became i

number (from C5140-4H) y, AVC changed the valve identification K-5497-1H.

Consequentl to C5497-1H to match the identification of the old valve with parts from the new kit. Because the new kit (withmodifiedcomponents)isnotqualified,thenew(reworked) valve will also be uno.ualified.

However, since the old valve was qualified, AVC was assuming the (reworkad) volve was also qualified and so was the kit.

The inspectors observed another instance c' en uncertain qualifi-cation process for a S0V supplied by AVC. A letter, dated December 29, 1988, from AVC to Ceco, indicates that S0Vs with nonreplaceable poppet seals were replaced with S0Vs having replaceable poppet seals. A letter, dated August 16, 1987, from AVC to GENE indicates that a PCA with the original nonreplaceable poppet seals was qualified by GENE for u:,e on valves manufactured by the Target Rock Corporation (the manufacturer of main steam relief valves) and installed at CEC 0 plants during the 1975 time frame. AVC stated that Wyle Laboratories qualified replace-able poppet seals during the qualification of a 3-way valve.

Thus, the original valve with the replaceable poppet seals was considered by AVC to still be qualified, and qualified on the basis of the Wyle qualifications.

11. Review of Audits lhe inspectors reviewed the following audits of AVC performed by licensees and GENE:

e.

Pennsylvania Power & Light (PP&L) performed an audit (Audit Report 88-075) on February 16, 1989, of AVC.

PP&L determined that AVC's program, as currently implemented, does net meet the requirements of 10 CFR 50, Appendix B or ANSI N45.2-1977.

PP&L identified one audit finding (88-0751) which essentially states that AVC's quality program does not properly dedicate commercial-grade items per 10 CFR Part 21 for use as a basic component in nuclear power plants.

In response to this finding, AVC established a dedicated area to store components for assembly in safety-related PCAs.

The NRC inspection team explained to AVC the intent of commercial-grade dedication. AVC acknowledged to the NRC inspectors that the action taken in response to the above PP&L finding was inadequate to address commercial-grade dedication and that additional measures would be established.

36

ORGANIZATION: AUTOMATIC VALVE CORPORATION NOVI, MICHIGAN REPORT INSPECTION NO.: 99901178/90-01 RESULTS:

PAGE 17 of 17 b.

Iowa Electric Light and Power Company, the licensee for the Duane Arnold nuclear power plant, conducted an audit on May 11, 1989 and identified four audit findir,,,and two observations.

The findings related to:

(1) The authority and duties of individuals performing quality related activities were neither discussed nor defined in the QA manual.

(2) AVC does not have a program in place to audit or evaluate their subtier vendors.

(3) AVC did not provide a formal training program for individuals assigned to QA/QC activities.

(4) AVC has neither an internal audit program nor a lead auditor to perform audits.

The audit observations also identified CoCs and QA/QC procedures not being microfilmed and the lack of a formal corrective action program.

AVC responded to the audit in a letter dated July 17, 1989,

(

and outlined the corrective action taken to resolve the l

audit findings.

c.

GPU Nuclear Corporation, the licensee for the Oyster Creek nuclear power plant performed an audit on December 10, 1987 and identified one adverse finding relating to lack of objective evidence that thread ring and plain plug gages were calibrated.

AVC responded to the audit finding on January 8,1989, stating that AVC revised Procedure D-308-1885 to provide provisions to record calibration dates for all major measuring equipment and that they had sent the measuring gages to a calibration agency.

E.

. EXIT INTERVIEW' The inspectors met with the individuals listed in Section F and discussed the scope t.ad findings of the inspection.

F.

PERSONS CONTACTED:

• T. Hutchins, President L

• D. Swinton, Manager, Engineerino L. Stone, Purchasing Agent R. Belcher, Assembly Builder B. Robson, Assembly Builder

• Denotes that the individuals were present at the exit interview.

37

pf UNITED STATES y

g NUCLE AR REGULATORY COMMISSION wassmoTON. O C. 20555 j

g...*/

4rn 4,1so Docket No. 99900317/89-01 Mr. W. H. A. Lambert, Manager Quality Assurance and Nuclear Operations Cooper-Bessemer Reciprocating Lincoln Avenue Grove City, Pennsylvania 16127

~

Dear Mr. Lambert:

This letter refers to the inspection conducted by Messrs. Walter P. Haass, Harvey M. Wescott, Edward B. Tomlinson, Robert A. Hermann and Carl H. Berlinger of this office on December 11-13, 1989 at your facilities in Grove City, Pennsylvanit. and to the discussions of our findings with yourself and other 1

p personnel in your company at the conclusion of the inspection.

The purpose of this inspection was to review the root cause analysis of the t

connecting rod failure that occurred in i, diesel generator at Unit 2 of the South Texas Project on November 28, 1989.

This inspection also included a review of the procurement process for connecting rod forgings, connecting rod design, manuf acturing documentation, and quality controls applicable to the design and manufactu-ing of connecting rods.

The results of the inspection indicate that the cause of the connecting rod failure is attributable to the stress raisers at the sharp corners of the 3/8 inch diameter center lube oil hole prouuted by the drill bit that partially penetrated the articulated rod pin bore during the drilling process. A Material Review Request (MRR) was written identifying the manufacturing error, but only the effect on the lubricating oil flow path was considered in the disposition of the MRR - not the structural aspects.

The scope of this problem with regard to generic implications was not fully defined at that time. At the time of manufacture, no specific control was established to assure that the,i/8 inch center lube oil hole was drilled to its proper depth as shown on the design drawing. You committed, therefore, to the performance of a finite element analysis to determine the structural acceptability of the connecting rods in the field with varied hole depths. Your company has since completed the finite element analysis and it is presently under NRC staff review.

It was noted that there is no record of the depth and number of overbored holes since these were not identified by MRRs; only those instances in which the holes penetrated the articulated pin bore were recorded on MRRs. These instances are the failed connecting rod at South Texas and two spare connecting rods - one located at South Texas and the other at Nine Mile Point.

It is our understanding h

that you consider the latter two to be acceptable since they were drilled completely through and the hole edges were radiused. We further noted that Cooper-Bessemer had issued a Part 21 report dated December 8,1989 to the NRC with copies to the Model KSV Users Group describing the incident, indicating that further analysis of the overbored holes is underway, and committing to advise users of the generic impact of the analysis.

i 38 w

Cooper-Bessemer Reciprocating Grove City, Pennsylvania NOTICE OF NONCONFORMANCES Based on the results of an NRC inspection conducted on December 11-13, 1989 at the Cooper-Bessemer Reciprocating Facility, Grove City, Pennsylvania, it appears tnat certain of your activities were not conducted in accordance with NRC requirements.

1.

Criterion X " Inspections" of Appendix B to 10 CFR Part 50 states in part "A program for inspection of activities affecting quality shall be established and executed by or for the organization performing this activity to verify conformance with the documented instructions, procedures and drawings for accomplishing the activity.... Examinations, measurements... of... products processed shall be performed for each wo-k operation where necessary to assure quality."

Contrary to the above, acceptance criteria for the 3/8 inch diameter center lube oil hole drilled into the connecting rod to provide lubrication at the crankshaft journal did not include verification of its depth to that shown on the engineering drawing.

(89-01-01) 2.

Criterion XV " Nonconforming Materials, Parts, or Components" of Appendix B to 10 CFR Part 50 states in part " Measures shall be established to control materials, parts, or components which do not conform to requirements in order to prevent their inadvertent use or installation.... Nonconforming items shall be reviewed and accepted, rejected, repaired or reworked in accordance with documented procedures."

Cooper-Bessemer quality procedure, QCP-10-6, " Quality Program Requirements for Review and Disposition of Nonconforming Material," paragraph 6.2 states that "Nonconformances shall be evaluated and dispositioned jointly by Quality Engineering and Engineering Design."

Contrary to the above, several Material Review Requests (MRR's) were dispositioned without the proper review by the affected engineering disciplines to assure that the necessary engineering considerations were included in the final disposition decision. Examples include MRR 07505, dated September 17, 1986, and MRR 08296, dated January 21, 1987.

l -

(89-01-02) l 39

L Mr. W. H. A. Lambert l During this inspection it was found that the implementation of your QA program l

failed to meet certain NRC requirements in that certain machining operations were not being verified, and the disposition of nonconformances was not being i

adequately reviewed. The specific findings and references to the pertinent requirements are identified in the enclosures to this letter.

Please provide us within 30 days from the date of this letter a written statement containing:

(1) a description of steps that have been or will be taken to correct these items; (2) a description of steps that have been or will be taken to prevent recurrence; and (3) the dates your corrective actions and preventive measures were or will be completed. Consideration may be given to extending your response time for good cause shown.

The responses requested by this letter are not subject to the clearance procedures of the Office of Management and Budjet as required by the Paperwork Reduction Act of 1980, PL 96-511.

In accordance with 10 CFR 2.790 of the Comission's regulations, a copy of this letter and the enclosed inspection report will be placed in the NRC's Public Document Room.

Should you have any questions concerning this inspection, we are available to discuss them with you.

Sincerely.

OriginalSIgnedbT E. William Brach, Chief Vendor Inspection Branch Division of Reactor Inspection and Safeguards Office of Nuclear Reactor Regulation

Enclosures:

1.

Notice of Nonconformances 2., Inspection Report No. 99900317/89-01 L

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)

40

ORGANIZATION: COOPER-BESSEMER RECIPROCATING GROVE CITY, PENNSYLVANIA xtrunt Ih5FECT10N INSPECTION NO.: 99900317/89-01 DATE: December 11-13, 1989 OH-SITE HOURS: 83 CORRESPONDENCE ADDRESS: Cooper-Bessemer Reciprocating Mr. W. H. A. Lambert, Manager Quality Assurance and Nuclear Operations Lincoln Avenue Grove City, Pennsylvania 16127 ORGANIZATIONAL CONTACT: Mr. W. H. A. Lambert, Manager of Quality Assurance TELEPHONE NUMBER:

412-458-8000 X-3395 NUCLEAR INDUSTRY ACTIVITY: Cooper-Bessemer Reciprocating is a manufacturer of diesel engines used in n: clear plants. The vendor also provides spare and replacement parts and services for these engines. The nuclear-related activity of the vendor comprises approximately 10 percent of its current business volume.

>}

11 m

ASSIGNED INSPECTOR:

/M/tNh [ /dtw+s -

3/37/N W. P. Haass, Acting Chief, Special Projects Section' Date OTHERINSPECTOR(S):

H. M. Wescott, SPS R. A. Hermann, Section Chief, EMCB E. B. Tomlinson, PD4 C. H. Berlinger, Branch Chief, GCB Irac[h, Chief,vendorInspectionBranchM' APPROVED BY:

6 FdM a

INSPECTION BASES AND SCOPE:

A.

BASES:

10 CFR Part 50, Appendix B and 10 CFR Part 21 B.

SCOPE:

Review the root cause analysis of the connecting rod failure at l

the South Texas facility including the related design, procurement, manufac-turing, and quality control processes. Also, determine the generic implica-t tions of the failure to assess its affect on other nuclear plants.

1 L

PLANT SITE APPLICABILITY: Nuclear plants that utilize the Model KSV diesel engine includir.g Braidwood 1 & 2. Byron 1 & 2. Cooper, Nine Mile Point 2 i

Palo Verde 1, 2 & 3, South Texas 1 & 2, Susquehanna 1 & 2, Waterford 3, and Zion 1 & 2.

41

ORGAfilIATION: COOPER-BESSEMER RECIPROCATING GROVE CITY, PENNSYLVANIA kEPORT INSPECTION NO.: 99900317/89-01 RESULTS:

PAGE 2 of 7 A.

V10LAT10NS:

None.

B.

NONCONFORMANCES:

Contrary to the requiremer.ts of Criterion X of 10 CFR Part 50, Appendix B which specifies the need "to verify conformances with the documented...

drawings for accomplishing the activity," acceptance criteria for the 3/8 inch diameter center lube oil hole did not include verification of its depth to that shown on the engineering drawing.

(89-01-01)

Contrary to the requirements of Criterion XV of 10 CFR Part 50, Appencix B which specifies that " nonconforming items shall be reviewed a!.c accepted, rejected, repaired, or reworked in accordance with documented procedures" and the Cooper-Bessemer procedure QCP-10-6, " Quality Program Requirements for Review and Disposition of Nonconforming Material," Paragraph 6.2 which states that "the nonconformance shall be evaluated and dispositioned jointly by Quality Engineering and Engineering Design " several Material Review Requests (liRR) were dispositioned without the proper review by the affected engineering disciplines to assure that the necessary engineering considerations were included in the final disposition decision. Examples included MRR 08296, dated Janeery 21,1987, (no engineering analysis was performed to justify acceptance of the oversized connecting rod

" asis"),andMRR07505,datedSeptember 17, 1986, (no engineering analysis was performed or review conduct 4d by the structural engineering discipline to determine the effect of st*ess raisers due to the sharp edges from the overbored hole in the disposition of the nonconforming connecting rod). (89-01-02)

C.

UNRESOLVED ITEMS:

The generic implications of the connecting rod failure at the South Texas facility have not been identified yet. The vendor comitted to perform a finite element analysis to determine the structural acceptability of the varied drilled depths of the 3/8" center lube oil hole. As of this date, the vendor has completed the required-analysis and it is presently under staff review. Since no specific control was established to assure that this hole was drilled to its proper depth as.shown on the design drawing, there is no record of the number of connecting rodswithoverboredholesandtheirdepth(noMRRswererequiredtobe written). While the vendor considers those holes that completely penetrated the pin bore and whose edges were radiused to be acceptable, it was re uested that all overbored holes be considered in the analysis.

(89-01-03 z

42

ORGANIZATION: COOPER-BESSEMER RECIPROCATING GROYE CITY, PENNSYLVANIA REPORT INSPECTION NO.: 99900317/89-01 RESULTS:

PAGE 3 of 7 b.

OTHER FINDINGS AND COMMENTS:

===1.

Background===

Approximately 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> into a 24-hour surveillance run of the No. 22 emergency diesel generator at the South Texas Project on November 28, 1989, a catastrophic failure of the number 4 master connecting rod occurred. A total of about 700 hours0.0081 days <br />0.194 hours <br />0.00116 weeks <br />2.6635e-4 months <br /> running time had been accumulated on the engine. The failure was attributed to high cycle fatigue of the section between the articulated pin bore and the crankpin bore due to overboring of a lube oil hole. A detailed metallurgical analysis of the fracture surface of the rod was performed by the licensee (Houston i

Light and Power) with support by the vendor.

Initiation of the fatigue failure occurred at the sharp edges created when the point of the

. drill bit partially broke through the bore surface for the articulated rod pin. At the time of manufacture, this hole was overbored, the error was identified in a Material Review Request (MRR), and the cor-rective action adopted was " acceptance as is" with the addition of screwed plug to assure the proper lubrication oil flow path. No

_ structural engineering considerations were apparently addressed in 2

the disposition of this nonconformance.

2.

Root Cause Analysis The design of the master connecting rod in the KSV-20-T diesel engine includes four oil holes to provide lubrication to the articulated rod pin and bushing. One hole is drilled through the area between the crankpin bore and the articulated pin bore and parallel to the axis of the crankshaft. This is referred to as the transverse or lateral oil hole. There are also two holes that are drilled at right angles to the transverse hole and through the center of the transverse hole from the crankpin bore into the articulated pin bore. The purpose of these holes is to direct lube oil flow from the transverse hole to the articulated pin bushing. These holes are located near the ends of the articulated pin and coincide with what is called the " balls".

Oil flow from the transverse oil hole into the articulated pin bail area is achieved by plugging both ends of the transverse hole, and by sealing the crankpin bore side of the bail area oil holes with the back of the crankpin bearing shell.

A fourth hole is drilled from the center of the crankpin bore at right angles to and into the center of the transverse hole. This hole communicates with the lubricating oil supply coming from the crankshaft and passing into the articulated rod for piston pin lubri-cation and piston crown cooling. The purpose of this hole is to pro-vide a flow path for lube oil into the transverse hole and I

E 43

ORGAh12AT10N: COOPER-BESSEMER RECIPROCATING GROVE CITY, PENNSYLVANIA REPORT INSPECTION NO.: 99900317/89-01 RESULTS:

PAGE 4 of 7 subsequently to the oil holes in the bail areas. The master rod design calls for this hole to extend only to the transverse hole, and not beyond, g-The master conr.ecting rod that failed at the South Texas plant had this fourth hole drilled beyond the transverse hole to a depth where the point of the drill bit penetrated the surface of the articulated pin bore. This partial penetration resulted in the formation of sharp edges at the surface of the articulated pin bore; the sharp edges acted as stress raisers that contributed to fatigue crack ini-tiation and growth and ultimately failure due to overload. Metallo-graphic analysis of the fracture surfaces of the master connecting j

rod clearly indicates that a crack initiated at the sharp edges of the center oil hole and propagated in an outward and downward direc-tion along a plane that intersects the centerlines of both the trans-verse and center oil holes. The crack then progressed in high cycle fatigue from the articulated pin bore surface downward to the trans.

verse hole and outward to the end holes. The remainder of the sec-tion between the crankpin bore and the articulated pin bore failed in a ductile tearing mode as a consequence of overloading the remaining uncracked section.

During the vendor's evaluation of the nonconforming condition, at the time of manufacture, no consideration was given to rounding on radi-using the sharp edges of the overbored hole at the articulated ein bore surface.

In fact, the shoulder left by the partially penetrated drill bit was utilized as a bearing point for the installation of the threaded plug to assure proper lube oil flow.

(Nonconformance 89-01-01) 3.

Scope of the Overb fing Problem The licensee, the vendor, and the NRC inspectors are in agreement regarding crack initiation at the sharp edges and shoulder of the over-bored center hole, and crack propagation due to high cycle fatigue.

What retaains in question is whether or not operational stresses could result in the initiation and propagation of a fatigue crack in a severely overbored hole that did not penetrate the surface.

It is known that there are numerous master connecting rods presently in nuclear service with center holes overbored to varying depths. No precise record exists to identify which rods contain overbored holes or their depth because overboring of the center hole was not consid-ered an inspection criterion at the time of manufacture. Only pene-tration to the articulated pin bore was considered important since it affected the lube oil flow path. Consequently, it is possible that center oil holes may have been drilled to just short of the point of penetration which would create a condition similar to the condition i

44

ORGANIZATION: COOPER BESSEMER RECIPROCATING GROVE CITY, PENNSYLVANIA REPORT INSPECTION NO.: 99900317/89-01 RESULTS:

PAGE 5 of 7 of the f ailed rod at South Texas. The vendor and an independent con-tractor are currently performing a three-dime.nsional finite element analysis to ascertain if any overbored condition could result in excessive localized stresses that would be sufficient to initiate a li fatigue crack. If the analysis predicts that cracks could initiate in Li an overbored oil hole, then the maximum hole depth that could be tol-erated would have to be determined, and all master rods in all the KSV engines would have to be inspected against the maximum hole depth criterion.

(Unresolveditem 8901-03)

In 1986, the vendor changed its inspection requirements to include

(

the depth of the center oil hole as an acceptance criterion.

4.

Connecting Rod Manufacturing Problems The inspectors reviewed the entire process for the manufacturing of r

connecting rods from the procurement of forgings from a supplier to the machining of the rods to final dinensions.

In particular, approximately 20 MRRs were reviewed to determine the type and magnitude of problems encountered during the manufacturing process T

and their disposition. Problems that were addressed in the HRRs included:

' oversizing of holes

• overboring of holes

• holes drilled at an angle

• excessive connecting rod length

• oversizing of lube oil channels

' mislocation of lube oil grooves and channels

• machining marks It was noted that no evidence was available to demonstrate that dis-position of the MRRs was supported by a documented engineering analy-sis, it was further noted that the appropriate engineering disci-plines were not always involved in the final disposition. As an example, one MRR involved a connecting rod of excessive length which was accepted "as is".

The rod length fell outside acceptable limits specified on the engineering drawing, but nevertheless was accepted.

No engineering analysis to determine the affect on compression ratio, for example, was performed to support the disposition decision that was made. As another example, for the connecting rod that failed at South Texas, the MRR that addressed the overbored lube oil hole was dispositioned without consideration of the stress aspects by a struc-tural engineer. Only the effect of the manufacturing error on the lubricating oil flow path was considered resulting in the insertion 45

ORGANIZATION: COOPER-BESSEMER RECIPROCATING GROVE CITY, PENNSYLVANIA REPORT INSPECTION NO.: 99900317/89-01 RESULTS:

PAGE 6 of 7 of a plug to assure design distribution of oil flow to the articula-ting rod pin bore. The need to radius the sharp edges to eliminate stress raisers was not considered.

(Nonconformance 89-01-02) 5.

Quality control Considerations The inspection indicated that the 3/8 inch center lube oil hole was apparently drilled to varying depths beyond the transverse hole -

contrary to the engineering drawing which called for a depth only to the transverse hole. The inspectors concluded that inadequate qual-ity control was exercised by not properly verifying the adherence of the manufacturing process to the engineering requirements. The inspectors found it unusual that such a variation in hole depth was not identified in MRRs.

It was also indicated that appropriate engineering discipline review of the manufacturing drawings should be performed to assure that all critical dimensions are identified for inspection and/or as hold points on the Quality Control Inspection Plan (QCIP).

6.

Part 21 Report issued by the Vendor Cooper-Bessemer issued a Part 21 report, dated December 8, 1989, to the NRC with copies to all uset5 of the Model KSV-20-T including the KSV Users Group. The report described the connecting rod failure j

event at the South Texas Project, the resulting damage that occurred, and provided the results of the root cause analysis.

It further indicated that a technical evaluation (finite element analysis) of the acceptability of other connecting rods with holes drilled to varying depths would be undertaken and users would be advised of the results at its completion, i

46

ORGANIZATION: COOPER-BESSEMER RECIPROCATING GROVE CITY, PENNSYLVANIA REPORT INSPECTION NO.: 99900317/89-01 RESULTS:

PAGE 7 of 7 E.

PERSONS CONTACTED:

The following Cooper-Bessemer personnel were contacted during this inspection:

f+F. Bruce Stolba, Vice President and General Manager John M. Horne, Manager, Analytical Engineering and Compressor Design John R. Schneider, Manager Manufacturing Engineering i+B. R. Sedelmyer, Director Manufacturing f+D. T. Blizzaro, D'. rector, Engineering R. J. Brager, Manager, Engine Design and Development f+ Archie E. Bice, 'Mnager, Nuclear Support M. A. Schleigh,.

.pervisor, Design Engineering William Ferguson, Manager, Materials Engineering f+W.' H. A. Lambsrt, Manager Quality Assurance and Nuclear Operations Ted Miller, Quality Assurance Technician N < d lowing NRC personnel participated in the inspection:

f+ Walter P. Haass, VIB Team Leader l

f+Harvey M. Wescott, VIB f + Robert A. Hermann, Section Chief, MTEB f+ Edward B. Tomlinson, PDIY

1. Carl H. Berlinger, Branch Chief, GCB

~

1. indicates attendance at the entrance meeting

+ indicates attendance at the exit meeting 47

of o

UNITE D ST ATES g

8 NUCLE AR REGULATORY COMMISSION 3s o

h f

W ASHING TON. D. C. 20555

• %,,,,,/

May 25, 1990 Docket No. 99901117/90-01 Mr. W. L. Strickland, Presidert DIVESCO, Incorporated A Subsidiary of Westland International Corporation 5000 U.S. Highway 80 East Jackson, Mississippi 39208 l

Dear Mr. Strickland This letter addresses the inspection of your f acility at Jacksor. Mississippi, conducted by Messrs.

S.

Alexander and R.

Moist of this office on April 2 through 6,1990, and the discussions of their findings with you and certain members of your staff at the conclusion of the inspection.

The purpose of the inspection was to review your corrective actions with regard to nonconformances from the previous NRC inspection and commitments made in your response letter to the previous inspection report, to evaluate the implementation of your quality assurance program in selected areus, and to followup on an allegation.

This inspection consisted of an examination of procedures and representative records, interviews with personnel, and observations by the inspectors as discussed in the enclosed report.

The inspectors found that your corrective actions from the previous nonconfor-mance and the implementation of your QA program were acceptable.

However, certain potential weaknesses were noted relating to procedures, receipt inspec-tions, and material traceability and storage that are discussed in detail in the enclosed report.

The allegation was not substantiated on the basis of our findings at DIVESCO.

Please advise this office of the results of your inquiries with Borg-Warner to resolve the marking irregularities identified on two of their stainless steel valves in your warehouse as discussed in section E.4 of the report.

In accordance with 10 CFR 2.790 of the Commission's regulations, a copy of this letter and the enclosed inspection report will be placed in the NRC's Public Document Room.

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

Sincerely, 4N

'E. William, ach, Chief Vendor inspection Branch Division of Reactor Inspection and Safeguards Office of Nuclear Reactor Regulation

Enclosure:

Appendix A-Inspection Report 99901117/90-01 48

ORGANIZATION: DIVESCO INCORPORATED JACKSON, MISSISSIPPI REPORT INSPECTION INSPECTION NO.: 99901117/90-01 DATES: April 2-6, 1990 ON-SITE HOURS! B4 CORRESPONDENCE ADDRESS: DIVESCO, Incorpora'ted A Subsidiary of Westland International Corp.

ATTN: Mr. W. L. Strickland, President 5000 U.S. Highway 80 East Jackson, Mississippi 39208 ORGANIZATIONAL CONTACT: Ms. Susan Kay Fisher, Vice President and QA Manager TELEPHONE NUMBER:

(601)932-1934 N'JCLEAR INDUSTRY ACTIVITY:

DIVESCO, Inc., (formerly NSSS, Inc.) has supplied nuclear plant components and equipment as safety-related items and as commercial grade items to various nuclear utilities, and currently supplies commercial grade items, handled and stored under a 10 CFR Part 50 Appendix B QA program.

DIVESCO obtains its stock primarily from surplus material from cancelled nuclear plants.

ASSIGNED INSPECTOR:

24 Av 5/2.c/fd

5. D. Alexanderf Reactive Inspection Section No. 2

'D a t'e (RIS-2)

OTHERINSPECTOR(S):

R. N. Moist, RIS-2 A

APPROVED BY:

52Sie U. Potapovs. Chief, RI5-2, Vendpr Ihspection Branch Date INSPECTION BASES AND SCOPE:

A.

BASES:

10 CFR Part 21 and 10 CFR Part 50, Appendix B B.

SCOPE:

Follow-up on corrective actions for the nonconformance from the previous inspection, evaluation of quality assurance (QA) program implementation in selected areas, and follow-up on an alle,gation.

L PLANT SITE APPLICABILITY:

Haddam Neck (50-213), Nine Mile Point 2 (50-410),

Hope Creek 1 and 2 (50-354/355), Beaver Valley 2 (50-412), Davis aesse(50-346),

[

River Bend (50-458), Clinton (50-461), Peach Bottom 1 and 2 (50-277/278),

Duane. Arnold (50-331),WNP-2(50-397), and Quad Cities 1 and 2 (50-254/265) 49

ORGANIZATION: DIVESCO INCORPORATED JACKSON, MISSISSIPPI 1

REPORT INSPECTION NO.: 99901117/90-01 RESULTS:

PAGE 2 of 9 A.

VIOLATIONS:

None f

B.

NONCONFORMANCES:

None C.

UNRESOLVED ITEMS:

None D.

STATUS OF PREVIOUS INSPECTION FINDINGS:

1.

(Closed) Nonconformance 99901117/88-01, Item B.1:

Contrary to Criterion IV of Appendix B to 10 CFR Part 50, DIVESCO, Inc.

(NSSS, Inc. at the time) had not established measures to assure that applicable regulatory requirements, design bases and other requirements necessary to assure adequate quality were suitably included or referenced in procurement documents for material later supplied to nuclear utilities as safety-related.

During this inspection, the NRC inspectors reviewed the correc-tive actions taken by DIVESCO that were committed to in their response letter to inspection report 99901117/89-01 and found them to be acceptable.

E.

OTHER FINbfNGS AND COMMENTS 1.

The inspectors noted that DIVESCO's purchase and resale of American Society of Mechanical Engineers (ASME) Code Section 111 and XI material without being an ASME N Stamp holder or Quality Systems Certificate (QSC) holder as well as the purchase of such material from DIVESCO by nuclear utilities may not be consistent with the requirements of the ASME code.

DIVESCO contended that because they make no representations as to quality or suitability of the material, but merely establish traceability and pass on the manufacturers certifications and store and maintain the material while in their possession in accordance with a 10 CFR Part 50 Appendix B quality assurance (QA) program, that they meet the intent of the ASME code.

This issue will be subject to further review by the NRC.

2.

During this inspection, the NRC inspectors reviewed tiie circum-stances surrounding alleged sales by DIVESCO of safety-related 50

ORGANIZATION: DIVESCO INCORPORATED JACKSON, MISSISSIPPI REPORT INSPECTION NO.:

99901117/90-01 RESULTS:

PAGE 3 of 9 valves [GE boiling water reactor (BWR) master parts list (MPL) numbers F42-F002 cod F004]

and refueling bridge winches (F11-E017) obtained as surplus from the Tennessee Valley Authority's (TVA's) cancelled Hartsville plant and other sources with inappropriate QA documents.

This allegation was not substantiated on the basis of information obtained at DIVESCO.

The NRC inspectors verified that DIVESCO had obtained QA documentation directly from the valve manuf ac-turer, Anchor Darling, for the eight 24-inch, 75-pound stainless steel gate valves DIVESCO had bought at the Hartsville site.

Records indicated that the appropriate documentation package was provided to 'ai Power along with its corresponding valve.

The inspectors reviewed the other cacumentation packages obtained from Anchor Darling in conjunction with visual examination of the seven remaining valves.

No discrepancies were noted.

With respect to the alleged sale of a Programmable and Remote (PAR) Systems refueling bridge winch to Commonwealth Edison Company's (CECO's) LaSalle County Station, records obtained at DIVESCO indicated that such equipment would have been part of a consignment to DIVESCO of material from Public Service -of Oklahoma's (PS0's) cancelled Black Fox plant.

According to DIVESCO, two such winches were held by them, but they were not l

sold.

Instead, DIVESCO stated that both winches were among the equipment returned to the consignor.

It was alleged that another PAR winch that was not part of the Black Fox consignment may have been sold by DIVESCO with QA documentation from the winches that had been returned to the consignor.

DIVESCO agreed that no QA documentation was included with the two winches they say they returned to the consignor, but maintained that the winches had no identifiable documentation with them when received by DIVESCO.

Although the disposition of both PAR winches could not be verified at DIVESCO, NRC review of DIVESCO sales records showed three sales of parts to LaSalle, but none of these were PAR winches.

In the review by the NRC inspectors of other equipment sales by DIVESCO to nuclear utilitier, no other sales of PAR winches were identified and no evidence of falsi-fication or improper substitution of QA documentation was identified.

Subsequent to the inspection, the NRC inspector confirmed DIVESCO's LaSalle sales information with the LaSalle procurement staff.

l

.1

ORGANIZATION:

DIVESCO INCORPORATED JACKSON, MISSISSIPPI REPORT INSPECTION NO.: 99901117/90 RESULTS:

PAGE 4 of 9 3.

QA Program Review In the course of evaluating selected portions of OlVESCO's QA program, the following QA policy documents were reviewed:

QAP 1.0 Organization QAP 8.0 identification and Control of Materials, Parts, and Components QAP 13.0 Handling, Storage, and Shipping QAP 15.0 Nonconforming Materials, Parts & Components QAP 17.0 Quality Assurance Records The following QA implementing procedures were also reviewed:

QAPR 4.0 Receipt Inspection QAPR 5.0 Monitoring of Material Control and Storage QAPR 6.0 Equipment Packaging and Shipping QAPR 7.0 Equipment Maintenance QAPR 9.0 Storage and Control of Nuclear Components and Assemblies QAPR 10.0 Part 21 Evaluation and Hotification On the basis of the inspectors review, the above policies and implementing procedures appeared adequate for their purposes with the exception of QAPR 7.0 discussed below.

The QA program review also included DIVESCO's Quality Assurance Policy Manual (QAPM), Revision 3, dated October 31, 1989 and Quality Assurance Procedures Manual (QAPRM), Revision 4, dated October 31, 1989.

The changes to both QAPM and QAPRM since the last revision clarified and defined in more detail the responsi-bilities and implementation of the QA program.

The following l

concerns were identified by the inspectors during this review:

• The Statement of Policy (50P) of the QAPM was ambiguous with respect to acceptance of safety-related orders, but supplying commercial grade material only.

• Criterion IV, " Procurement Document Control," and Criterion Vil, " Control of Purchased Material, Equipment, and Services,"

of Appendix B to 10 CFR Part 50 had been erroneously considered inapplicable to DIVESCO's scope of operations and were removed from DIVESCO's QAPM as a result of a licensee audit finding and recommendation.

• A deviation from the prescribed storage maintenance procedure was authorized by an internal memorandum, but had not been incorporated in a QAPRM.

52

ORGANIZATION:

DIVESCO INCORPORATr.D JACKSON, MISSISSIPPI REPORT INSPECTION NO.: 99901117/90-01 RESULTS:

PAGE 5 of 9 The 50P of the QAPM indicated that purchase orders (P0s) from nuclear utilities imposing 10 CFR Part 21 and 10 CFR Part 50 Appendix B would be accepted by DIVESCO only for equipment for which (1) documentation was available from the original manufac-turer, or (2) recertifications were obtained from the manufac-turer or (3) certifications were obtained from a testing facility apprcved by the utility, in another paragraph of the SOP, DIVESCO stated that because they are neither the manufacturer nor the original purchaser of nuclear components they cannot certify that the components are " safety-related." The inspectors pointed out that acce ping orders which imposed 10 CFR Part 21 and 10 CFR Part 50 Appendix B in fact constituted their acceptance of safety-related orders.

Accordingly, DIVESCO agreed to clarify the 50P.

Criterion IV of 10 CFR Part 50 to Appendix B " Procurement Document Control" and Criterion VII

" Control of Purchased Material, Equipment, and Services" were deleted from QAPM and QAPRM.

Some procurements made by DIVESCO are from suppliers other than those selling surplus material from cancelled plante j

on an as-is basis.

One such procurement was reviewed during '.'

inspection in which DIVESCO's customer invoked 10 CFR Part 21 e imposed 10 CFR Part 50 Appendix B QA requirements in its P0 for material DIVESCO obtained from another supplier, it appeared that this procurement was handled properly.

DIVESCO also sends some components back to original manufacturers for recertification.

Therefore, the inspectors pointed out to DIVESCO that since nuclear-specific requirements for material intended for safety-related applications imposed on them by their customers must be passed on to these suppliers of safety-related services (who must also have the appropriate, audited, and approved QA and 10 CFR Part 21 programs), these activities should be addressed in their formal QA program, even though DIVESCO is only a warehousing activity without the capability to perform complete dedication.

In addition, DIVESCO stated their plans to expand their business routinely to include such orders from other vendors who would drop-ship the material (some of which could be safety-related) directly to the utilities, and possibly to act as a distributor for at least one manufacturer. On the basis of this information, the inspectors concluded that these two criteria needed to be reincorporated in the QAPM and QAPRM to cover DIVESCO's expanded scope of operations to the extent of their activities affecting quality.

53

ORGANIZATION: DIVESCO INCORPORATED JACKSON, MISSISSIPPI REPORT INSPECTION NO.: 99901117/90-01 RESULTS:

PAGE 6 of 9 The deviation from maintenance procedurer was identified during review of Public Service Electric and Gas Company's (PSE&G's)

Nonconformance Report (NCR) number Q2-284899 which had been written by a PSE&G inspector during a source surveillance at DIVESCO.

The NCR indicated that two Anchor Darling gate valves being bought for the Hope Creek station had surface rust inside the nozzle outlet and inlet due to the valves' not being stored with end caps and dessicant.

The NCR also indicated that the ZEP-Preserve rust inhibitor used by DIVESCO was not approved by PSE&G for nuclear applications.

Attachment B to DIVESCO QAPR 7.0, Revision 2, dated October 31, 1989, " Equipment Maintenance,"

under " Listing of Maintenance Activities for Manual Inspection,"

shows inspection attributes which include checking components for covers or seals, installing dessicant and humidity indicators and inspecting valve stems and packing areas for corrosion.

DIVESCO showed the inspectors an internal memo, dated March 9,1988, which authorized the deviation from DIVESCO's procedures for maintenance and storage of nuclear valves.

The memo stated that the preservation measures for short-term storage recommended by valve manufacturers were not adequate to meet the long-term storage requirements of a high-humidity area.

DIVESCO argued that it was their considerable experience that leaving the valve in an unsealed condition enhanced the air flow and allowed for evaporation.

The memo that authorized deviation from QAPR 7.0.

was not intended to be a one-time deviation; therefore, DIVESCO l

i agreed that QAPR 7.0 should be consistent with DIVESCO's policy L

and practice for storage and maintenance as reflected in the DIVESCO memo.

However, the inspectors pointed out that in view of the rust found by the PSE&G inspector (which DIVESCO contended was only light " surface" rust), DIVESCO should reevaluate the effectiveness of their practice and consider the possibility that either better inhibitors, fresh and/or different dessicants and tighter sealing might be more effective and also be consistent I

with manufacturers' recommendations.

4.

QA Program Implementation Review The inspectors reviewed NCRs 14, 18, 35, 50, and 79 selected at random from DIVESCO's NCR log.

These NCRs were written during receipt inspections.

Three of the five NCRs were still open.

The inspector reviewed the disposition of all five NCRs and verified that the dispositions were adequate.

Final disposition of the three open NCRs is to be completed prior to shipment.

DVESCO stated that, in most cases, open NCR deficiencies are verbally communicated to their customers when they place an order.

1 54

I ORGANIZATION:

DIVESCO INCORPORATED JACKSON, MISSISSIPPI REPORT INSPECTION N0.: 99901117/90-01 RESULTS:

PAGE 7 of 9 l

The inspectors reviewed DIVESCO invoices 1380A,13808,1325, and 1512 and the associated QA records including material receiving i

reports (MRRs). audit surveys, shipping records and maintenance records.

The records reviewed were found to be in accordance with DIVESCO policies and procedures.

The inspectors toured both DIVESCO warehouses noting that house-keeping, storage, material and location identification on items selected to check were satisf actory.

Two Borg-Warner 12-inch, 300-lb, stainless steel gate valves, serial numbers 62944 and 2

62943, were examined and the inspectors noted several marking irregularities.

The date of manufacture on the nameplates of l

both valves. appeared to have been been changed from "1980" to j

l read "1985."

Also the "6" on the marking "PN0 82565" appeared to j

have been struck-over on the stainless steel body of one of the 1

l the valves.

And finally, the "3" on the P0 number, 9773107, on the nameplate of valve serial number 62943 appeared to be changed from a 5.

However, none of these discrepancies had been identi-fied on the MRR for these valves.

DIVESCO agreed to define receipt inspection and documentation requirements more clearly in the procedures and on forms, and to pursue satisfactory resolu-tion of the identified marking discrepancies with Borg-Warner.

i 5.

In verifying corrective actions by DIVESCO since the last NRC l

inspection, the inspectors reviewed the records of all sales of safety-related material identified by DIVESCO through September 1989 that were not already reviewed by the NRC as follows:

Purchap2 Purchase Order Invoice Plant Baldwin Assoc C53996 B11225-2 Clinton 1111nois Power X32969 1035 Clinton Illinois Power X34974 1061 Clinton Illinois Power X36014 1064 Clinton Illinois Power X39725 1103 Clinton Illinois Power

/39726 1111 Clinton Illinois Power 5C2247 1196 Clinton Illinois Power 516?48 1316 Clinton Illinois Power 510832 1325 Clinton WPPSS 09495^

1333 WNP-2 Baldwin Assoc C54327 B12265-1 Clinton Iowa Electric 535633 1215 Dwayne Arnold Illinois Power X39134 1096 Clinton PSE&G P2-284898 1380A&B Hope Creek PECO BW602574 1426 Peach Bottom g

ORGANIZATION: DIVESCO INCORPORATED JACKSON, MISSISSIPPI REPORT INSPECTION N0.: 99901117/90-01 RESULTS:

PAGE 8 of 9 The inspectors reviewed two letters that DIVESCO sent to their customers, dated August 12, 1988 and September 27, 1989 in which they addressed the problems identified by the NRC during the previous inspection. The files contained certified mail receipts for the mailings to the customers listed above and those identi-fied in the previous NRC inspection report.

The letters were intended to clarify DIVESCO's position that their certificates of conformance and compliance (C0Cs) were and are provided only as supporting documentation and are not to be used by the utility as the sole basis for dedicating the strictly commercial grade components supplied by DIVESCO.

It was noted during the review of the associated orders that DIVESCO had procured some piece-parts.from vendors who supplied documentation to the utility directly or in other instances, DIVESCO sent components to a vendor for recertification and shipping directly to the utility.

The inspectors also verified that DIVESCO P0s imposed the same requirements on the vendor as DIVESCO's customer requirements.

In addition, the inspectors reviewed the only new safety-related P0 that imposed 10 CFR Part 21 and Appendix B since September 1989.

P0 330860, dated March 24, 1990 from Commonwealth Edison for Quad Cities ordered a 12-inch, 300-psi Anchor Darling gate valve.

DIVESCO purchased the valve from Price Supply Company (PSC) who in turn shipped the valve to DIVESCO for final inspection and shipping to Quad Cities, it was noted that DIVESCO performed a source survey on PSC durb; 1988 and allowed PSC to use applicable procedures from DIVESCO's QAPM and QAPRM for the purposes of this sale.

In order to evaluate DIVESCO's practices in handling commercial grade orders (for material which could be used in safety-related applications), the inspectors also reviewed three utility P0s that did not impose Part 21 or Appendix B for material obtained by uiVESCO from non-surplus vendors as follows: (a) a 3-inch, 600-psi Valtek control valve with air operator supplied by Control Valve Services to Clinton Nuclear plant under Illinois Power PO 526592, dated February 15,1990, (b) a 10-inch V-Ball (rotary shaft) valve, Fisher Controls model V-100, shipped directly from Control Valve Services to Clinton Nuclear plant under Illinois Power P0 525506, Revision 1, dated January 23, 1990 (these valves were rebuilt), and (c) Agastat timing relays, j

model E7012AC002, 120VAC, (not manufactured between June 1981 and January 1982) shipped directly from PM sales to Fermi-2 under Detroit Edison P0 NR-22174, dated October 10, 1989.

Consistent with DIVESCO's stated policy, their processing of these orders included sufficient documentation to maintain traceability from 56

ORGANIZATION: DIVESCO INCORPORATED JACKSON, MISSISSIPPI REPORT INSPECTION NO.: 99901117/90-01 RESULTS:

PAGE 9 of 9 at least DIVESCO's source of supply to the end user in order to facilitate establishment of traceability to the manufacturer (or refurbisher) should later dedication be undertaken. No deficien-cies with respect, to DIVESCO'c handling of these orders were noted.

F.

PERSONS CONTACTED:

W. Strickland, President S. Fisher, Vice Presi'cnt and Quality Assurance Manager T. Westbrook, Administrative Manager W. Boggan, Quality Assurance Inspector and Warehouse Manager 57

'o UNITED STATES

'f NUCLE AR REGULATORY COMMISSION E

nAssiwotow, o. c. rosss

• • ...../

moa Docket No. 99900505/90-01 Mr. Charles R. Healy, Director Que'ity Assurance Ebasco Services Incorporated Two World Trade Center New York, New York 10048

Dear Mr. Healy:

This letter addresses the inspection of your office located in New York City conducted by Messrs. Robert L. Pettis, Jr., and Michael Snodderly of this office, and Michael Villaran of Brookhaven National Laboratory during the period May 7-9, 1990, and the discussion of their findings with Mr. William Stuckey and other members of your staff at the conclusion of the inspection.

The purpose of the inspection was to follow-up on a previous open item identified in NRC Inspection Repo:t No. 99900505/89-01 which related to Ebasco's basis for placing several suppliers on the Waterford 3 approved suppliers list.

The specific areas examined during the inspection and our findings are discussed in the enclosed report.

in addition to the review of procedures and records, the

-inspection consisted of interviews with Ebasco personnel and observations by the inspector.

Within the scope of this inspection, there were no instances found where Ebasco failed to meet NRC requirements. Should you have any questions concerning this inspection, we will be pleased to discuss them with you.

Sincerely, OM\\/A L E. William Brach, Chief Vendor Inspection Branch Division of Reactor Inspection and Safeguards Nuclear Reactor Regulation j

- Inspection Report No. 99900505/90-01 cc: Mr. N. S. Carns, Plant Manager Louisiana Power and Light 317 Barone Street l

New Orleans, Louisiana 70160 58

i i

I ORGANIZATION: EBASCO SERVICES INCORPORATED NEW YORK. NEW YORK REPORT INSPECTION INSPECTION N0.: 99900505/90-01 DATE: May 7-9, 1990 ON-SITE HOURS: 38 CORRESPONDENCE ADDRESS: Mr. Charles R. He'aly, Director Quality Assurance Ebasco Services Incorporated Two World Trade Center New York, New York 10048 ORGANIZATIONAL CONTACT: Mr. William Stuckey, Manager, QA Engineering TELEPHONE NUMBER:

(212)839-3306 NUCLEAR INDUSTRY ACTIVITY: Design, procurement, and quality assurance activities for several nuclear projects.

C s F.

O ASSIGNED INSPECTOR:

R. L. Pettis, Cr., Reactive Inspection Section

'Date No.1,(RIS-1)VendorInspectionBranch(VIB)

OTHER INSPECTOR (S):

M. Snodderly, RIS-1, VIB

(

M. Villaran, Consultant APPROVED BY:

h k

dw Zb[To C.A.VanDenburgh, Chief,RIS-1,GIB Date INSPECTION BASES AND SCOPE:

A.

BASES:

10 CFR Part 50, Appendix B, Ebasco Topical Report ETR-1001 liicTear Quality Assurance Program Manual," and 10 CFR Part 21.

B.

SCOPE:

Follow-up-inspection to review unresolved item (89-01-03) from RRI Tnspection Report No. 99900505/89-01.

PLANT SITE APPLICABILITY:

Waterford3SteamElectricStation(50-382) 59

p m.<..

p a

ORGANIZATION:

EBASCO SERVICES INCORPORATED NEW YORK, NEW YORK REPORT INSPECTION NO.: 99900505/90-01 RESULTS:

PAGE 2 of 9 A.

VIOLATION:

No violations were identified.

B.

NONCONFORMANCE:

Ne nonconformances were identified.

C.

UNRESOLVED ITEM:

No unresolved items were identified.

D.

STATUS OF PREVIOUS INSPECTION FINDINGS:

Background

The inspection was performed as a follow-up to an earlier inspection (NRC Inspection Report No. 99900505/89-01) performed at the Waterford 3 Steam Electric Station (SES) which was prompted by(DOL) by a former a complaint of discrimination filed with the Department of Labor Ebasco employee and referred to the NRC by 00L. The complaint alleged that Ebasco violated 10 CFR Part 50, Appendix B, and its internal proce-dures in the evaluation and approval of certain material and component suppliers for the Waterford 3 SES.

Specifically, the alleger claimed that certain suppliers were found unsatisfactory for supplying safety-related components and materials in the 1981-1983 time frame. Due to the potential impact this would have on the construction schedule,

-Ebasco sent other auditors (unqualified) to evaluate these suppliers who subsequently found them to be satisfactory.

To determine the validity of these allegations, an NRC inspection was performed at Ebasco's New York City office in August 1988 to review the pertinent records and related procedures.

Immediately prior to the inspection, the NRC. inspectors met with the alleger to clearly identify the specific records and areas of concern. The results of the inspection were documented in NRC Inspection Report No. 99900505/88-01, dated September 28, 1988. An additional inspection was performed at the Waterford 3 SES to review procurement documents generated by Ebasco during this period in order to determine the safety-related significance l

associated with the alleger's concerns.

Those documents were not located L

in the New York office during the August inspection.

The results of the' inspection were documented in NRC Inspection Report No. 99900505/89-01, I

dated March 31, 1989.

This inspection was performed to follow-up on previous inspection findings and to review the basis for Ebasco's responses to NRC related to these findings. The status of these items l

1s as follows:

60

ORGANIZATION:

EBASCO SERVICES INCORPORATED NEW YORK, NEW YORK REPORT INSPECTION N0.: 99900505/90-01 RESULTS:

PAGE 3 of 9 1.

(Closed)Nonconformance 89-01-02 This nonconformance concerned Ebasco's placement of 35 safety-related urchase orders (P0s) with Guarantee Instrument Systems of Missouri GISMO) without GISMO being listed on Ebasco's appioved suppliers list ASL). Since this inspection, Ebasco has responded to these concerns in correspondence dated 'ay 29, 1989; June 9, 1989; August 11, 1989; A

October 20, 1989; and January 17, 1990. As a result, this issue was closed by the NRC in a letter to Ebasco dated April 6, 1990.

The basis of such closure was reviewed during the inspection and is presented below.

The 35 P0s classified as safety-related involved the procurement of electrical interface material in accordance with Ebasco Specification LOV 1564.249R, " Fabricated Boxes, Covers, and Associated Hardware for Electrical Systems - Interface Material - Seismic Category I."

Electrical interface material consists of fabricated junction boxes, covers, conduct, and associated hardware which is non-electrical in nature, but is generally recogn) zed as being necessary to provide support for Class IE circuits.

Ebasco's position was that the specification and the associated junction box detail drawings, LOU 1564-B-353, allowed consnercial-grade fabrication g

.of the junction boxes which were then seismically analyzed and designed cy i

Ebasco and installed by Ebasco's contractor, Fischbach and Moore.

This position appeared to be consistent with the fact that quality assurance (QA) Specification 1209 QCB was prepared and made a part of the specifi-cation rather than the use of Ebasco's standard QA Specification 860-80 l

which invoked the requirements of 10 CFR Part 50, Apperdix B, and 10 CFR

.?

Part 21.

Ebasco stated that the P0's were never intended to comply with 10 CFR Part 50, Appendix B, Quality Assurance requirements. They-were designated safety-related because the f*ont page of each P0 only allowed the initiator the choice of nonsafety-related or safety-related. Since Appendix I, Revision 2, of Ebasco's nuclear quality assurance program manual defined the term safety-related as an item whose application was either Class IE, seismic Category I, or Safety Class 1, 2, or 3, the safety-related block on the P0 was marked.

Ebasco maintained that rigid l

steel galvanized (RSG) conduit, fittings, and boxes were intended to form a non-Ciass IE, seismically supported system.

Ebasco's classification of RSG boxes and fittings during the period 1977-1979 was discussed in a March 1, 1977, internal memo which classified l

l 61

l\\

ORGANIZATION: EBASCO SERVICES INCORPORATED NEW YORK, '1W YORK i

REPORT INSPECTION NO.: 99900505/90-01 RESULTS:

PAGE 4 of 9

(

%:4..et %

l electrical interface material as Class 1E, seismic Category I which would s

have required the supplier to maintain a full 10 CFR 50, Appendix B, QA program. A March 15, 1977, Ebasco memo discussed the fact that this material should be classified as non-Class 1E, non-seismic Category I, since it was non-electrical in nature and was primarily used for mechanical protection of cables. Under this classification, the material should be purchased to industry standards and a modified QA program per NRC Regulatory Guide (RG) 1.29 paragraph C.4.

It was further stated that the conduit and box supports should be treated as a seis i M eoory I system which would require Ebasco to perform a seismic analysis.

!his memo also pointed out that Waterford 3 was only committed to Institute of Electrical Engineers (IEEE) Standards 308-71 and 323-71 which did not address electrical interface material.

On May 18, 1977, Ebasco issued Nuclear Licensing Policy Statement (NLPS) 3.2-2 which revised the classification of electtical interface material to non-Class 1E, seismic Category 1.

This was based on the fact that these components are cecessary to provide support for the proper functioning of Class 1E circuits and their failure could degrade such circuits to an unacceptable level. On January 12, 1979, Ebasco issued NLPS 3.2-2, Revision 1, which introduced the term " interface" material and reclassified such material as seismic Catergory 11 since the material interfaces with Class IE electrical components and structural failure resulting from safe shutdown earthquake (SSE) conditions could adversely affect Class 1E systen performance.

Seismic Category 11 requires a modified 10 CFR 50, Appendix B, QA program as stated in NRC RG 1.29.

The NRC inspectors reviewed seismic calculations performed by Eba, o which confirmed the structural design adequacy of the RSG junction. Lox I

and related conduit system.

Ebasco's approach was conservative since the calculations assumed a minimum material yield stress of 25,000 pounds per square inch for commercial-grade carbon steel and used seismic l

design accelerations of-1.5 times peak accelerations.

In addition, a l

finite-element analysis to verify stresses within the junction box was L

performed to verify the anticipated seismic loading.

Ebasco concluded l

that deformation of the junction box during SSE conditions would not I

degrade the electrical performance of the system.

Ebasco used this engineering assessment to demonstrate the operability of of GISMO on Ebasco's ASL during the period (1977-1980)pport the placement the system since documentation was not available to su in which the P0s were issued. The inspectors noted that Ebasco audited GISM0 in July i

1977; however, documentation did not exist to support the placement of 1

GISMO on the ASL. On June 26, 1980, Ebasco re-audited GISMO and found them to be in compliance with Waterford 3 Project Specification LOU 1564.249R, and the applicable portions of 10 CFR Part 50, Appendix B.

62

ORGANIZATION: EBASCO SERYlCES INCORPORATED NEW YORK, NEW YORK 1

REPORT INSPECTION NO.: 99900505/90-01 RESULTS:

PAGE 5 of 9 limited to some that essentially required a complete 10 CFR Part 50, Appendix B, QA program depending on the nature of the item being supplied (e.g., galvanized rigid steel conduit, fittings, couplings, and related electricalinterfacematerial). This methodology is discussed in the In conclusion, although documentation approving GISMO for inclusion on the Waterford 3 SES ASL in 1977 could not be paduced by Ebasco, the 1980 audit (and sub xquent 1983 audit) of GISMO indicates the supplier was on a 3 year periodic audit cycle (in accordance with Ebasco Procedure No. QAP-9, Revision 1, dated September 29, 1978, paragraph 7.1.3),and gives reasonable assurance that the required corrective actions, arising from the 1977 supplier audit, had been implemented at some time prior to 1980.

In addition, the seismic adequacy, design, and component classifi-cation of the system appear to be in compliance with the regulatory posi-tion of NRC RG 1.29.

As a result, this item is closed.

(Closed) Unresolved item 89-01-03 During NRC Inspection No. 99900505/89-01, the inspectors were unable to review supplier evaluations performed by Ebasco's New York QA personnel for suppliers previously rejected by Ebasco but later approved and placed on Ebasco's ASL. As a result, during this inspection the NRC inspectors

)erformed a review of the documentation associated with eight Ebasco-performed supplier evaluations from early 1983. The evaluation audits for all eight suppliers were found to be adequate, thus supporting Ebasco's placement of tiese suppliers on the June 1, 1983, version of the Waterford 3 ASL.

According to Paragraph 7.1.3 of Ebasco Procedure No. QAP-9, Revision 1, dated September 29, 1978, " Quality Assurance Vendor Evaluations,"

qualified suppliers were placed on the ASL by an Ebasco review of their QA program.

An initial or pre-award audit was then conducted for those first-time suppliers placed on the ASL, with a follow-up audit performed on a cycle of 1 to 3 years, as determined by the Ebasco Vendor Evaluation Group Leader.

Suppliers included on the ASL, according to 3

Paragraph 10.1.1., were those whose documented quality )rogram and facility were considered satisfactory in accordance witi Paragraph

/

7.7.1 (Ebasco Procedure No. QAP-9) and therefore, authorized to receive a nuclear safety-related P0. At the end of the supplier's periodic audit date, the supplier must be re-audited or dropped from the ASL.

Project QA requirements were ;overned by Etasco Specification 860-80, "QA Requirements for Suppliers of Safety ". elated Equipment and Services,"

and 860-72, " Quality Control Requirements for Suppliers of Equipment and Services." Not all suppliers on the Ebasco Wataford 3 SES ASL were

{

63

i ORGANIZATION:

EBASCC SERVICES INCORPORATED G YORK, NEW YORK REPORT INSPECTION NO.: 99900505/90 01 RESULTS:

PAGE 6 of 9

~

reqai ed to implement or maintain a full 10 CFR Part 50, Appendix B, QA program.

In some cases, the quality requirements applicable to an order were defined in a procurement specification that was referenced in the basic PO to a supplier. The range of supplier quality programs required by these procurement specifications varied from some that were very Introduction to 10 CFR Part 50, Appendix B, which states that the per-tinent requirements of this appendix apply to all activities affecting the safety-related functions of those structures, systems and com-ponents.

In addition, Criterion !!, " Quality Assurance Programs,"

states that the program shall provide control to an extent consistent with their importance to safety.

A letter from Mr. M. Brooks (Ebasco-Waterford 3 site) to Mr. R. Williams, (Ebasco, New York), dated March 8,1983, indicated that several " key" sup-pliers of safety-related electrical interface material, soon due fo-re-audits, had been rejected by Ebasco's New York QA Department due to the suppliers inability to comply with the 10 CFR Part 50, Appendix B, QA requirements. The suppliers referenced were: Crouse Hinds Company; GISM0; Wheatland Tube Company; 0. Z. Gedney Company; Picoma Industries; Conduit Pipe Products; Electromark, and Appleton Electric. A review of the Ebasro Waterford 3 SES ASL, dated December 15, 1983, indicater. that all eight suppliers appeared on the ASL. The basis for placing these suppliers on the ASL was supplier audits performed by Ebasco New York auditors during the period March-May 1983.

As a result, these suppliers were deemed satisfactory by Ebasco, and were placed on the ASL despite earlier rejection by the Ebasco, New York QA Department.

The NRC inspector's review of these supplier audits is presented below, a.

Appleton Electric Company.

Ebasco performed a supplier audit of the Appleton Electric Company of Chicago, Illinois, on April 27-29, 1983, to deter-mine compliance of their QA arogram to the applicable portions of 10 CFR Part 50, Appendix 3, and Waterford 3 project Specifi-cation LOU 1564.249A ' Revision 4.

The audit was conducted by Mr. H. Bullen for the manufacture and supply of conduit boxes and fittings per the specification requirements.

Appleton's QA program, facility and manuf acturing operations were eraluated using Ebasco's standard nuclear QA checklist.

t l

The results of the audit were satisfactory and no corrective l

actions were required. As a result, Appleton Electric Company was considered a qualified supplier of electrical conduit

) oxes l

and fittings to the Waterford 3 SES. This verifies the basis

~

for retention of Appleton Electric Company on the Ebasco Water-ford 3 SES ASL, dated June 17, 1983.

64

l ORGANIZATION:

EBASCO SERVICES INCORPORATED NEW YORK, NEW YORK REPORT INSPECTION NO.: 99900505/90-01 RESULTS:

PAGE 7 of 9 b.

Crouse Hinds Company.

Ebasco performeo a supplier audit of the Crouse Hinds Company of Syracuse, New York, on May 4,1983, to determine compliance of their QA program to the applicable portions of 10 CFR Part 50, Appendix B, and Waterford 3 project Specification LOU 1564.249A, Revision 4.

The periodic audit was conducted by Mr. P. Ryan for the manufacture and supply of conduit boxes and fittings.

Crouse Hinds' QA program was evaluated using Ebasco QA checklist form 1306-14/-181," Supplier Quality Program Evaluation-Nuclear and/or Seismic 1," and the facili-ties and manufacturing operations were evaluated using Ebasco check-list Form 1306-2/2-78 " Supplier Manufacturing / Service Facility Quality Assurance Audit-Part 1 Nuclear and/or Seismic !."

The results were conditionally satisfactory with minor corrective actions required.

Ebasco approved the corrective actions for the facility in their letter to Crouse Hinds dated September 12, 1983, and for the Amarillo, Texas, facility on December 29, 1983, at which time Ebasco upgraded Crouse Hinds to a qualified supplier of elec-trical conduit boxes and fittings. These audits verify the basis for retention of Crouse Hinds Company on the Waterford 3 SES ASL, dated June 17, 1983.

c.

Guarantee Instrument Systems of Missouri (GISMO).

Ebasco performed a supplier audit of GISMO on May 3-5, 1983, to determine compliance of the GISMO QA program to the applicable portions of 10 CfR Part 50, Appendix B, and Waterford 3 project Specification LOU 1564.249R.

The audit was conducted by Mr. H. Bullen for the manufacture and supply of fabricated boxes, covers and associated electrical interface material.

GISM0's QA rogram was evaluated using Ebasco checklist form 1306-14/-181, p' Supplier Quality Program Evaluation-Nuclear and/or Seismic 1,"and thefacilitiesmanufacturingoperationswereevaluatedusingEbasco QA checklist form 1306-2/2-78 ' Supplier Manuf acturing/ Service Facility Quality Assurance Audit-Part i Nuclear and/or Seismic 1."

The results of the audit were conditionally satisfactory with several corrective actions required.

Ebasco approved the corrective actions in their letter dated June 23, 1983, and upgraded GISMO to a satisfactory supplier.

These audits verify the basis for retention of GISMO on the Waterford 3 SES ASL, dated June 17, 1983.

65

ORGANIZATION: EBASCO SERVICES INCORPORATED NEW YORK, NEW YORK REPORT INSPECTION NO.: 99900505/90-01 RESULTS:

PAGE 8 of 9 d.

Picoma Industries. Incorporated Ebasco performed a supplier audit of Picoma Industries, Incorporated, of Martins Ferry, Ohio, on March 22, 1983, to determine compliance of the Picoma Industries QA program with the applicable portions of 10 CFR Part 50, Appendix B, and Waterford 3 project Specification LOV 1564.270 Revision 6.

The audit was conducted by Mr. H. Bullen forthemanufactureandsupplyofelectricalinterfacematerialand accessories. Picoma's QA program was evaluated using standard Ebasco checklists.

The results of the audit were conditionally satisf actory with several corrective actions required.

Picoma implemented the corrective actions which were approved by Ebasco per their letter dated April 18, 1983. This verifies the basis for retention of Picoma Industries, incorporated, on the Ebasco Waterford 3 SES ASL, dated June 17, 1983.

e.

Wheatland Tube Company Ebasco audited the Wheatland Tube Company on March 29-30, 1983, to the requirements of Ebasco Specification 270. The results were conditionally satisfactory and corrective actions were required in the areas of internal audits, calibration procedures and management review of nonconformance reports.

On April 18, 1983, Whetland Tube Company forwarded their corrective actions to Ebasco who reclassi-fied them as satisf actory on July 8,1983. This verifies the basis for retention of Wheatland Tube Company on the Ebasco Waterford 3 SES ASL, dated June 17, 1983.

f.

Electromark Company Ebasco audited the Electromark Company on May 3,1983, to the requirements of Ebasco Specification 2490.

The results of the audit were satisfactory and Electromark was retained on the ASL to provide material in accordance with the Waterford 3 project speci-fications. This verifies the basis for retention of Electromark Company on the Ebasco Waterford 3 SES ASL, dated June 17, 1983.

g.

Conduit Pipe Products Company Conduit Pipe Products Company was audited on January 7-8, 1982, and the results were unsatisfactory.

The audit report stated that their present QA program did not address the necessary criteria of 10 CFR Part 50, Appendix B.

In a March 10, 1982, letter to Ebasco, Conduit Pipe Products Company said that their QA program met the 66

ORGANIZATION: EDASCO SERVICES INCORPORATED NEW YORK, NEW YORK REPORT INSPECTION NO.: 99900505/90-01 RESULTS:

PAGE 9 of 9 requirements of Ebasco Specification 270 and not 10 CFR Part 50 Appendix B.

Conduit Pipe Products was re-audited by Mr. H. Bullen on March 23-24, 1983, to one specification requirements and was found conditionally satisfactory. Corrective action was required in the areas of vendor evaluations, training, QA review, and internal audits. On April 20, 1983, Conduit Pipe Products forwarded their corrective actions to Ebasco who reclassified them as satisfactory on July 1,1983.

This verifies the basis for retention of Conduit Pipe Products on the Ebasco Waterford 3 SES ASL, dated June 17, 1983.

h.

O. Z. Gedney Company Ebasco audited the 0. Z. Gedney Company (Gedney) of Brooklyn, New York, on May 11, 1983, to the requirements of Ebasco Specification 249A.

The results were conditionally satisfactory.

Corrective action was required in the area of overdue calibration of equipment.

On May 31, 1983, O. Z. Gedney forwarded their corrective actions to Ebasco who reclassified them as satisfactory on July 1,1983.

Gedney's Birch Hill Foundry in Shoemakersv111e, Pennsylvania, was also audited by Ebasco on June 1, 1983, to the requirements of Ebasco Specification 249A and 249R.

The results were conditionally satisfactory. Corrective action was required in the area of segregation of material.

On June 9, 1983, O. Z. Gedney forwarded their corrective actions to Ebasco who reclassified them as satisfactory on July 1,1983.

This verifies the retention of 0. Z.

Gedney on the Ebasco Waterford 3 SES ASL, dated June 17, 1983.

As a result of the above, this item is closed.

E.

OTHER INSPECTION FINDINGS AND COMMENTS:

None were identified during the inspection.

F.

PERSONS CONTACTED EBASCO Services. Inc.

W. L. Stuckey, Manager, Quality Assurance Engineering L. A. Stinson, Quality Assurance Manager - Atlanta Area l

P. E. Grossman, Project Engineer I. 8ari, Supervising Civil Engineer 67

,[pD CIG k S

UNITED ST ATEs y

n NUCLE AR REGULATORY COMMISSION wassiNot oN, D. C,20665 k.....

May 14. 1990 Docket No. 99901128/88-01 Mr. R. S. Walker, President Power Distribution Technology 15215 Shady Grove Road Rockville, Maryland 20850

Dear Mr. Walker:

This letter addresses the inspection of your f acilities at Newport News, Virginia, conducted by Messrs. J. Jacobson and T. Silko of this office on October 17 through 19, 1988, and the discussions of their findings with Mr. R. Shematek, and other members of your staff at the conclusion of the inspection.

The purpose of the inspection was to review Power Distribution Technology activities in the area of circuit breaker refurbishment. Re1 case of this report was delayed during NRC's ongoing review of nonconforming and substandard vendor products.

Areas examined during the inspection and our findings are discussed in the enclosed report. This inspection consisted of an examination of procedures and representative records, interviews with personnel, and observations by the inspectors.

During this inspection, it was found that the implementation of your quality assurance program failed to meet certain NRC requirements, as imposed upon you in licensee safety-related procurement documents, relating to adequacy of procedures and test control. The specific findings and references to the pertinent requirements are identified in the enclosures to this letter.

Please provide us, within 30 days from the date of this letter, a written statement containing:

(1) a description of steps that have been or will be taken to correct the cited deficiencies; (2) a description of steps that have been or will be taken to prevent recurrence; and (3) the dates your corrective actions and preventive measures were or will be completed.

We will consider extending the response time if you can show good cause for us to do so.

68

Mr. R. S. Walker x.y 14, 1990 The responses requested by this letter are not subject to the clearance proce-dures of the Office of Management and Budget as required by the Paperwork Reduction Act of 1980, PL 96-511.

In accordance with 10 CFR 2.790 of the comission's regulations, a copy of this letter and the enclosed inspection report will be placed in the NRC's Public Document Room.

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

Sincerely, ORIGINAL SIGNED BY:

ULDIS POTAPOVS FOR E. William Brach, Chief l

Vendor Inspection Branch Division of Reactor Inspection and Safeguards l

Office of Nuclear Reactor Regulation

Enclosures:

1.

Appendix A-Notice of Nonconformance 2.

Appendix B-Inspection Report 99901128/88-01 l

i 69

Power Distribution Technology Docket / Report No. 99901128/88 01 APPENDIX A NOTICE OF NONCONFORMANCE Based on the results of an NRC inspection cvMucted Octobe-17 through 19, 1988, it appears that certain of your act!vities were not conducted in accordance with NRC requirements which were imposed on you by ourchase order contracts with NRC licensees.

Nonconformances with these requ >ements are as follows:

A.

Criterion V of Appendix B to 10 CFR Part 50, " Instructions, Procedur_es, and Drawings," states: " Activities affecting quality shall be prescribed i

by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings.

Instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfac-torily accomplished."

Contrary to the above, Power Distribution Technology procedure PDT-QA-07,

" Document Control and Distribution," was not appropriate to the circum-stances in that it did not require that current revisions to vendor technical manuals and product bulletins be obtained and used for work performed on safety-related circuit breakers. (88-01-01)

B.

Criterion XI of Appendix B to 10 CFR Part 50, " Test Control," states, in part: 'A test program shall be established to assure that all testing requireit to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accor-dance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents...."

Contrar,i to the above, Power Distribution Technology test procedure PDT-0P-04 did not incorporate acceptance limits contained in applicable design documents in that the trip time acceptance limits listed in the procedure were taken from a combination of sample curves in the vendor technical manual generated for a particular set of pickup and delay settings that may be (and, in the cited instance, were) different from those of breakers under test.

l l

70

l i

ORGANIZATION:

P)WER DISTRIBUTION TECHNOLOGY NfWPORT NEWS VIRGINIA REPORT INSPECTION INSPECTION NO.:

99901128/88-01 JATE:

10/17-19/08 ON-SITE HOURS:

16 CORRESPONDENCE ADDRESS:

Mr. R. S. Walker, President Power Distribution Technology 15215 Shady Grove Road Suite 202 Rockville, Maryland 20850 ORGANIZATIONAL CONTACT:

Mr. R. Shomatek (Newport News, Virginia facility)

TELEPHONE NUMBER:

(804) 873-1048 NUCLEAR INDUSTRY ACTIVITY:

Pruvides electrical power distribution main-tenance and refurbishment services to the U.S. Navy and the utility industry.

i n

0 ASSIGNED INSPECTOR:

(/%[ M

/ M/

J. B.,fac IS) pecial Projects Inspection

,S Date Selgi OTHERINSPECTOR(S):

Thomas B. Silko, SPIS l

L b

1. V'-

1 F!$$

APPROVED BY:

t E Potapovs Section Ch ef, SPIS Date INSPECTION DASES AND SCOPE:

A.

BASES:

Appendix B to 10 CFR Part 50 and 10 CFR Part 21 8.

SCOPE:

Inspection was conducted to review Power Distribution Technology activities in the area of circuit breaker refurbishment.

PLANT SITE APPLICABILITY:

Surry 1/2 (50-280/281), Indian Point Unit 3 (50-286).

71

ORGANIZATION: POWER DISTRIBUTION TECHNOLOGY NEWPORT NEWS, VIRGINIA REPORT INSPECTION NO.:

99901128/88-01 RESULTS:

PAGE 2 of 6 A.

VIOLATION:

None B.

NONCONFORMANCE:

1.

Contrary to Criterion V of Appendix B to 10 CFR Part 50, Power Distribution Technology (PDT) procedure PDT-QA-07 " Document Control and Distribution" does not require that current revisions of vendor technical manuals and product bulletins be obtained and used for work performed on safety-related circuit breakers

~

(88-01-01).

2.

Contrary to Criterion XI of Appendix B to 10 CFR Part 50 the acceptance criteria listed for circuit breaker trip time testing performed in accordance with PDT procedure PDT-OP-04 were incorrect.

The acceptance times listed were taken from a combination of sample vendor curves generated for a particular pickup and delay setting different from the tested breakers.

(88-01-02).

C.

UNRES0l.VED ITEMS:

None D.

STATUS OF PREVIOUS INSPECTION FINDINGS None. This was the first inspection of this facility.

E.

INSPECTION FINDINGS AND OTHER COMMENTS 1.

Introduction PDT of Newport News, Virginia is a subsidiary of Grove Engineering, Inc. of Rockville, Maryland.

PDT was established in 1984 to provide electrical power, distribution, maintenance and refurbishment services to the U.S. Navy and the utility industry.

Approximately 6 percent of PDT's business in 1987 was on safety-related breaker work activities.

No molded case circuit breakers (CB's) have been refurbished for nuclear utilities although insulated case breakers have been refurbished.

All electrical testing on CB's is performed on site.

x l

l 72

ORGANIZATION:

POWER DISTRIBUTION TECHNOLOGY NEWPORT NEWS, VIRGINIA REPORT INSPECTION NO.: 99901128/88-01 RESULTS:

PAGE 3 of 6 2.

Personnel Training The inspector reviewed PDT's QA procedure PDT-QA-02, Rev. 1

" Quality Assurance Introduction and Training." The inspector verified via review of QA test and attendance records that the annual Introduction and Training meeting to cover all aspects of the QA program was being held.

The inspector verified by dis-cussion with shop personnel that QA project briefings were being held to define each individual's QA responsibilities prior to the initiation of a new project. Training of PDT personnel is performed via the on-the-job training method.

The inspector noted that documentation is maintained on the breakers worked as part of employee on-the-job training to demonstrate qualificatien.

No deficiencies were identified.

3.

Technical Manual Control A review was conducted of PDT procedure PDT-QA-07 " Document Control and Distribution."

It was noted that this procedure did not include provisions for assuring that the latest revisions of technical manuals are obtained and used when overhauling CB's.

Additionally, no provisions existed for receiving or evaluating vendor product bulletins that apply to the particular CB's being overhauled.

Nonconformance 88-01-01 is cited in Section B of this report as a result of these deficiencies.

4.

Control of Measuring and Test Equipment (M&TE)

The inspector reviewed PDT's QA procedure PDT-QA-12, Rev. 2

" Control of Measuring ano Test Equipment," which applies to all M&TE used in the overhaul of nuclear safety-related or electrical apparatus built to military specifications.

The inspector reviewed the Test Equipment History Card, Proof of Calibration of tools. M&TE usage logs, and the test equipment itself.

No deficiencies were identified.

5.

Plating of Breaker Parts Plating of breaker parts is performed by Industrial Plating Corporation (IPC) of Tabbs, Virginia.

PDT's audit of IPC states that the cadmium plating is performed in accordance with Federal Specification QQ-P-416-C and the silver plating is performed in accordance with Federal Specification QQ-S-365C.

The silver plating is performed on contacts and the cadmium plating is 73

i ORGANIZATION:

POWER DISTRIBUTION TECHNOLOGY NEWPORT NEWS, VIRGINIA REPORT INSPECTION NO.: 99901128/88-01 RESULTS:

PAGE 4 of 6 performed on other breaker pieces to prevent corrosion.

The review of PDT's audit of IPC identified no deficiencies.

The inspector reviewed the Certificate of Conformance provided from IPC to PDT on the plating of parts used for the Surry Nuclear Plant safety-related breakers which were in the shop at the time of inspection.

No deficiencies were identified.

6.

Procurement Review As part of the inspection a review was conducted of PDT's procedure relative to the procurement of parts used in refur-bishment operations.

PDT-QA-03, " Purchase Requisition / Order" requires that all specialty parts be ordered from the Original Equipment Manufacturer (OEM) or from a supplier acceptable to the particular customer involved.

Specialty parts were defined as those parts specific to a particular breaker such as contacts, coils, mechanism, etc.

Nonspecialty parts would be generic parts such as barstock, wire, terminal blocks, etc.

All safety-related nonspecialty parts are required to be orde ed from vendors on the PDT approved vendors list in accordance with PDT-QA-10 " Source Evaluation." A review was done of four random purchases of safety-related spare parts.

Purchase order 5743-0-S was for aa arc chute assembly to Brown Boveri, Inc.

The P/0 invoked 10 CFR 50 Appendix B and 10 CFR Part 21.

Purchase order 704421 for arcing contacts was also for Brown Boveri. Inc. and similarly invoked Appendix B and Part 21.

Purchase order 5781-2-5 for moving arcing contacts and a power mechanism was to Westinghouse NSID and invoked Appendix B and 10 CFR Part 21.

No deficiencies were noted in this area, however, the PDT termi-nology of specialty, nonspecialty, and safety-related parts was found to be somewhat confusing.

7.

Approved Vendor List A review was conducted of the PDT approved vendor list.

It was noted that companies are put on the PDT approved vendors list based on a combination of several factors including review of their QA manual, past performance, or an audit. After being put on the approved vendors list the companies are required to be audited every three years.

The current PDT approved vendors list included eight companies.

No deficiencies were noted in this area.

74

ORGANIZATION:

POWER DISTRIBUTION TECHNOLOGY NEWPORT NEWS, VIRGINIA REPORT INSPECTION NO.:

99901128/88-01 RESULTS:

PAGE 5 of 6 8.

Analysis of Commercial Grade Parts The inspector reviewed three completed analyses to support the use of commercial grade parts in safety-related CB's. - The first component was an airpot overcurrent trip switch.

The switch activates an auxiliary relay whose contacts provide indication that the breaker tripped on overcurrent and also prevents reciosing of the breaker. The second component was a wire cover for a Westinghouse D5-416 1600 amp circuit breaker. The purpose of the wire cover is to protect the wiring from mechanical damage during removal and installation of the circuit breaker.

The third component was an operating link upper pin used to connect the contact operating link to the contact operating arm.

All commercial grade analyses were reviewed for adequacy and com-pleteness.

No inadequacies were identified.

j 9.

Review of Ccmpleted Work Packages A review was conducted of three completed PDT work packages.

Job number 223 was for refurbishment / overhaul of several ITE K-600 metal clad CB's for the Surry Nuclear Plant. All replace-ment parts used in this job were traced to safety-related pro-curements from vendors on PDT's approved vendors list.

It was noted that breaker deficiencies encountered during the overhaul i

are documented in the work package by the shop mechanics; however, no documented record of an evaluation to solve each deficiency existed.

The inspector was shown a draft procedure that included a new form for docena.. ding evaluations and PDT committed to issuing the draft procedure by November 11, 1988.

Job number 145-4 was for the refurbishment / overhaul of several Westinghouse DS-416 type metal clad CB's for the Indian Point Nuclear Plant.

All spare parts were procured from vendors on PDT's approved vendor list.

No deficiencies were noted with this work package.

Job number 224 was for refurbishment / overhaul of several Westinghouse 05-416 metal clad breakers for the Indian Point Nuclear Plant.

It was noted that during the calibration check for the electronic overloads of these breakers, the wrong acceptance criteria were used.

For these particular overloads variable pickup and delay times are achieved by the setting of two potentiometers en the overload face. Only sample (two) trip curves are provided in the technical manual for these breakers.

Each of these two curves corresponds to a particular set of pickup 75

ORGANIZATION: POWER DISTRIBUTION TECHNOLOGY NEWPORT NEWS, VIRGINIA l

l REPORT INSPECTION NO.: 99901128/88-01 RESULTS:

PAGE 6 of 6

\\

and delay settings.

For settings other th6n those shown on the sample curves, a new curve would need to be constructed.

For this particular job, the pickup and delay settings were different from those indicated on the sample curves; yet the sample curves were used for deriving the acceptance criteria. Even though the wrong acceptance criteria were used, the recorded trip times appeared to be within the limits of the applicable curve for this particular application.

Additionally, it was noted that the acceptance criteria were derived by the shop personnel without any engineering or QA input or review. Nonconformance 88-01-02 is cited in section B of this report as a result of PDT's failure to properly test these particular overloads and to develop proper acceptance criteria.

F.

PERSONNEL CONTACTED

• J. Canning, PDT, Production Manager

• T. J. Kielar, PDT Quality Assurance Manager f

• M. Denny, PDT Quality Assurance Representative

• R. G. Shematek, PDT, Director J. Danahy, Vice Presider.t of PDT R. S. Walker, VP of Grove Engineering and President of PDT l

[

• Attended exit meeting l

76

l Docket No.: 99900367/88 01 MAY ! 81990 Mr. L. West Quality Assurance Manager Square D Company 252 North Tippecanoe Peru, Indiana 46970

Dear Mr. West:

This letter addresses the inspection of your facility at Peru, Indiana conducted by Messrs. J. B. Jacobson and T. B. $11ko of this of fice on November 2-3, 1988, and the discussions of their findings with you and other members of your staff at the conclusion of the inspection.

The inspection was conducted to review activities relative to your supply of molded.:ase circuit breakers to the nuclear industry.

Release of this report was delayed during NRC's ongoing review of nonconforming and substandard vendor products.

Areas examined during the NRC inspection and our findings are discussed in the enclosed report.

This inspection consisted of an examination of procedures l

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

Within the scope of this inspection, we found no instance where you failed to meet the NRC requirements.

However, questions concerning your design change resiew activities, and activities relative to testing of safety-related breakers at the Cedar Rapids manufacturing plant are identified as unresolved items in the enclosed report.

Should you have any questions concerning this inspection, we will be pleas 2d to discuss them with you.

Sincerely, WJMT,'

OdQ E. William Brich, Chief Vendor inspection Branch Division of Rtactor Inspection and Safeguards Office of Nuclear Reactor Regulation

Enclosure:

Appendix A-Inspection Report:

99900367/88-01 77

-ORGANIZATION:

SQUARE D COMPANY PERV, INDIANA REPORT INSPECTION NSPECTION NO.:

99900367/88-01 DATE: 11/2-3/88 S SITE HOURS: 24 CORRESPONDENCE ADDRESS: Mr. L. West Quality Assurance Manager Square D Company 252 North Tippecanoe Peru, Indiana 46970 ORGANIZATIONAL CONTACT:

Lyle Hawkins TELEPHONE NUMBER:

317-472-3382 NUCLEAR INDUSTRY ACTIVITY:

Square D supplies molded case circuit breakers (CBs) and other electrical components used in safety-related Motor Control Centers (MCC's).

!/

/ /

//147 ASSIGNED INSPECTOR:

J. B. Jaco)is=onl,~5pecial' Inspection Section (SPI 5)

Date l'fl<' '

OTHERINSPECTOR(S):

T. B. Silko SPIS APPROVED BY:

/20 AtA2A p _ ~

l Ib Ef[

V. Potapovs. Section Chidf, SP I S te e

INSPECTION BASES AND SCOPE:

l.

A.

BASES: Appendix B to 10 CFR Part 50, 10 CFR Part 21.

B.

SCOPE: This inspection was conducted to review activities relative to Square D's supply of molded case CBs to the nuclear industry.

PLANT SITE APPLICABILITY:

All nuclear plants using Square D molded case circuit' breakers.

78

ORGANIZATION:

SOUARE D COMPANY PERU, INDIANA REPORT INSPECTION NO.:

99900367/88 01 RESULTS:

PAGE 2 of S A.

VIOLATION:

None.

B.

NONCONFORMANCE:

None C.

STATUS OF PREVIOUS INSPECTION FINDINGS:

Not reviewed during this inspection.

D; UNRESOLVED ITEMS:

1.

Square D's use of Underwriter Laboratories Inc. (UL) parameters rather than trip curve parameters for developing acceptance criteria for overcurrent testing and Square D's failure to notify customers of changes made to inoividual trip curves are identi-fied as unresolveo items pending further review.

2.

Square D's apparent lack of documentation supporting design change reviews performed since January 19, 1987 will remain an unresolved item pending review of pertinent records at the l

applicable manufacturing facilities.

E.

OTHER FINDINGS AND COMMENTS:

1.

Review of Dedication Activities All Square D supplied safety-related breakers are manufactured under a 10 CFR Part 50 App'endix B program. The majority of the breakers are manufactured at Square D's Cedar Rapids, Iowa plant which provides them to Peru, Indiana facility with a Certificate of Conformance (C of C).

The Square D Peru, Indiana facility then reviews the paperwork and issues a C of C to the customer.

A review was conducted of several specific purchases in order to evaluate the Square D certification process.

Purchaseorder(PO)6A068028 from Southorn California Edison for three FHL3600, 3 amp, molded case CBs was reviewed. The P0 invoked Appendix B. Part 21, and Bechtel Specification S023-302-04 which included specific seismic criteria.

Square D Peru in turn ordered these breakers from the Cedar Rapids manufacturing plant under interplant number 12-79249, also invoking Part 21.

79 l

ORGANIZATION:

50VARE D COMPANY PERU, INDIANA REPORT INSPECTION NO.: 99900367/88-01 RESULTS:

PAGE 3 of 5 Cedar Rapids supplied a C of C for these breakers stating that they had been tested and inspected and found to be in accordance with the applicable engineering records. Attached to the C of C was test data for a contact pressure check and magnetic calibration check. No thermal check was performed on these breakers as they are not equipped with a separate thermal element.

No deficitncies were noted.

PO RQ-88-01-25851 from Sacramento Municipal Utility District for one 255 amp LAB 36255 molded case breaker'was also reviewed. The P0 invoked Appendix B. Part 21, and required a C of C to Square D Test Report 8998-10.09-L45. The breaker was obtained from Cedar Rapids under interplant order number 0439985.

A C of C was supplied by Cedar Rapids along with test data for a 135 percent and 200 percent thermal trip test plus a contact pressure and a magnetic calibration check.

No deficiencies were noted.

Test data for one FAB 24030AC molded case breaker received from Cedar Rapids under interplant number 32-76886 was also reviewed.

It was noted that the acceptance criteria given for the COO percent trip test performed on this breaker was incor*ect.

The data sheet listed an acceptable time as being "less than 45 minutes." This is not in agreement with either the published trip curves or UL 489 which requires tripping in less than 2 minutes at 200 percent of the breaker's current rating. The actual measured trip time for this breaker at 200 percent was however less than the required 2 minutes.

This discrepancy was said to be a clerical error.

l One potentially generic concern was discovered during review of the test data. The concern stems from the fact that the acceptance criteria delineated for the 135 percent and 200 percent overcurrent trip tests are derived from UL 489 requirements which may or may not envelop the times one would derive from the published trip curves for the individual breakers.

For example the acceptance criteria for the 200 )ercent overcurrent check performed on an FAL 36080A2, 80 amp bren cer is listed as 8 minutes maximum. This matches the maximum value give in Table 15.1 of UL 489, but is higher than the 5 minutes maximum one would read from the published manufacturer's curves which utilities typically use for application and coordination studies.

l Additionally, it was noted that for any given type of CB the trip curves may have significantly changed over the years.

Utilities ordering replacement breakers may assume the new breaker meets the E

80

1 1

ORGANIZATION: SQUARE D COMPANY PERU, INDIAtiA REPORT INSPECTION NO : 99900367/88-01 RESULTS:

PAGE 4 of 5 original trip curves upon which the plant electrical protection scheme was developed.

If the new breakers do not meet the original trip curves, the plant's electrical protection system could be compromised.

Square D's use of UL versus trip

- ve paraneters and Square D's failure to notify customers of the us to trip curves will be left as c

an unresolved item pending further review, Control of Safety Related Breakers All safety-related breakers sold by Square D are processed directly through the Peru facility.

Square D's policy is to provide a C of C for breakers sold through Peru, but not to provide C of C's to Square D distributors.

This policy precludes a distributor from purchasing a refurbished breaker and selling it to a nuclear facility with a C of C.

An additional action taken by Square D to prevent the use of non-safety-related breakers in safety-related applications was to add an "A2" suffix to the number for safety-related equipment. Square D will only sell "A2" equipment to nuclear facilities for safety-related applications and will not sell this type of equipment to distributors.

Therefore, if a nuclear utility contacts a distri-butor for an "A2" part, the distributor will have to direct the utility to the Peru facility.

Upon receipt of an "A2" order, Peru instructs the manufacturing l

facility, located in Cedar Rapids, Iowa, to manufacture the breakers according to a separate set of procedures implementing 10 CFR 50 Appendix B, quality assurance requirements.

By doing so, Square D negates the need'for a dedication process. The inspector i

noted that Square D does not have a buy back program and therefore, all safety-related breakers sold must be specifically ordered.

The inspectors review of Square D's program for the handling of safety-related breakers did not identify a likely path by which refurbished breakers could be sold as safety-related.

No deficien-cies were identified.

Review of Audits The inspector reviewed two audits of Square D Peru, conducted by Sacramento Municipal Utility District (Rancho Seco) and Southern California Edison (SONGS 1, 2 end 3). The audits appeared complete and corrective actions taken by Square D to audit deficiencies were acceptable.

Nu deficiencies were identified.'

i 61

ORGANIZATION:

SQUARE D COMPANY PERU, INDIANA i

REPORT INSPECTION NO.: 99900367/88-01 RESULTS:

PAGE 5 of 5 l

Design Change Review A review was conducted of Square D's design change review program as necessary to establish similarity between current replacement products and those originally seismically tested.

During the inspection it was discovered that the design change reviews were not being performed at L

Peru, Indiana, but at the individual manufacturing locations. Prior to January 19, 1987 the reviews were documented on a " Request For Engineering Change" and/or " Cost Estimate" form. The form included a block for indicating the change would not affect nuclear 1E certi-fications. Subsequent to January 19, 1987 this form was no longer used, however the design chinge review was reportedly continued.

Square D's apparent lack of do;umentation supporting design change reviews performed since January 19, 1987 will remain an unresolved item pending review of pertinent records at the applicable manufacturing facilities.

^

F.

EXIT MEETING:

l Upon completion of the inspection an exit meeting was held.

The following people were in attendance:

NAMES TITLE

=

L. West Quality Assurance Manager Square D L. Hawkins Quality Assurance Engineer, Square D J. B. Jacobson Lead Inspector NRC T. B. Silko inspector, NRC 82 J

d'jo meeg4 UNITED STATES l

[

g NUCLE AR REGULATORY COMMISSION 5

-l W ASHING TON, D. C. k1666

\\, "*'/

APR 11 1%]

Docket No. 99900277/90-01 Mr. 5. Bond, President and General Manager The Rockbestos Company A Member of the Marmon Group Post Office Box Drawer 1102 New Haven, Connecticut 06504

Dear Mr. Bond:

This letter addresses the inspection of your facilities at New Haven and East Granby, Cone;cticut, conducted by Messrs. R. Moist and S. Alexander of this office ott January 30 through February 2,1990, and the discu.sions of their findings with Mr. George Littichales and other members of youa staff at the conclusion of the inspection.

The purpose of the inspection was to review the circur.a s and your actions related to certain 10 CFR Part 21 notifications and to it sw-up on nonconfor-mances, open items, and unresolved items from previoys inspect'ons.

Areas examined during the NRC inspection and our findings are dist.utsed in the enclosed report. This inspection consisted of an examination of procedures and representative records, interviews with perronnel, and observations by the inspectors.

During this inspection, it was found that the implementation of your QA program failed to meet certain NRC requirements, including failure to follow procedures relating to design control, corrective action and test control.

The specific findings and references to the pertinent requirements are identified in the enclosures to this letter.

Please provide us, within 30 days from the date of this letter, a written statement containing:

(1) a description of steps that have been or will be taken to' correct the cited deficiencies; (2) a description of steps that have been or will be taken to prevent recurrence; and (3) the dates your corrective actions and preventive r:easures were or will be completed.

We will consider extending the response time if you can show good cause for us to do so.

83

l Mr. 5. Bond 2

The responses requested by this letter are not subject to the clearance proce.

2 dures of the Office of Management and Budget as required by the Paperwork Reduction Act of 1980. PL 96 511.

In accordance with 10 CFR 2.790 of the Comission's regulations, a copy of this letter and the enclosed inspection -

will be placed in the NRC's l

Public Document Room.

i L

Should you have any questions concerning this inspection, we will be ple'ased l

to discuss them with you.

Sincerely, FAgTn'ahlgned by E. Ylilliam Br::h E. William Brach, Chief Vendor Inspection Branch Division of Reactor Inspection and Safeguards Office of Nuclear Reactor Regulation

Enclosures:

1.

Appendix A-Notice of Nonconformance 2.

Appendix B-Inspection Report 99900277/90 01 l

l l

84

ORGANIZATION:

THE ROCKBESTOS COMPANY NEW HAVEN AND EAST GRANBY, CONNECTICUT REPORT INSPECTION INSPECTION NO.: 99900277/90-01 DATES: Jan. 30 - Feb. 2. 1990 ON-SITE HOURS: 30 CORRESPONDENCE ADDRESS:

The Rockbestos Company A Menber of the Marmon Group Mr. S. Bond, President and General Manager Post Office Box Drawer 1102 New Haven, Connecticut 06504 ORGANIZATIONAL CONTACT: Mr. George Littlehales, QA Manager TELEPHONE NUMBER:

(203) 772-2250 NUCLEAR INDUSTRY ACTIVITY:

Rockbestos conducts thermal aging, loss-of-coolant accident (LOCA) simulations, and post-LOCA sample evaluation testing for environ-mental qualification (EQ) as well as flame testing for their nuclear adverse service Class 1E (safety-related) cables in the engineering laboratory at their New Haven, Connecticut factory. Nuclear service cable is manufactured at both the New Haven and East Granby, Connecticut, plants, with the majority of it being made at East Granby.

Approximately 40 percent of Rockbestos' business is from nuclear utilities.

ASSIGNED INSPECTOR: /defMM 8 82 d f J/'IY//O

/t. N. Mo'Ist,' Reagtive Inspection Section No. ;

~Date (RIS-2)

OTHERINSPECTOR(S):

S. Alexander, RIS-2 M. Jacob s, Sandia National ! aboratories (SNL)

I i

APPROVED BY:

% flee 4 2-cfo l

U. Potapovs, Chief, R15-2, Vendor Inspection Branch Date INSPECTION BASES AND SCOPE:

A.

BASES:

10 CFR Part 21 and 10 CFR Part 50, Appendix B B.-

SCOPE:

Follow-up on 10 CFR Part 21 notification, follow-up on nonconfor-mances, open items and unresolved items from previous inspections, evaluation of quality assurance (QA) program implementation in selected areas, and follow-up on an allegation.

PLANT SITE APPLICABILITY:

San Onofre 1, 2 and 3 (50-206/361/362); Haddam Neck (50-213); Nine Mile Point 1 and 2 (50-220/410); Dresden2and3(50-237/249);

Millstone-1, 2 and 3 (50-245/336/423); Turkey Point 3 and 4 (50-250/251);

(continued on next page) 85

.~

ORGANIZATION: THE ROCKEESTOS COMPANY NEW HAVEN AND EAST GRANBY, CONNECTICVT REPORT INSPECTION NO.: 99900277/90-01 RESULTS:

PAGE 2 of 24 PLANT SITE APPLICABILITY:

(continued) Palisades (50-255); Monticello (50-263); Quad Cities 2 (50-265); Point Beach I and 2 (59-266/301); Peach Bottom 2 and 3 (50-277/278); Prairie Island 1 and 2 (50-282/306); Indian Pilg(rim (50-293); Zion 1 and 2 (50-295/304);50-309); Arkansas Nucle Kewaunee Point 3 (50-286);

(50-305); Maine Yankee Calvert Cliffs 1 and 2 (50-317/318); Fitzpatrick (50-333); St. Lucie 1 and 2 (50-335/389); McGuire 1 (50-369); LaSalle 1 and 2 (50-373/374); Medical CollegeofHanover(50-377); Catawba 1and2(50413/414); and WNP-3 (50-508).

A.

VIOLATIONS:

None B.

NONCONFORMANCES:

1.

Contrary to the requirements of Criterion III of Appendix B to 10 CFR Part 50, " Design Control." and Section 1 of Rockbestos Procedure S-3, " Design Control Systems," the implementation of the Rockbestos program for approval and release of design data was inadequate in that Rockbestos provided cable ampacities and ampacity calculations (nuclear plant design data) that were not formally reviewed and approved, and which were prepared using a computer code that was not formally reviewed and approved for Firezone R cable sold to the Haddam Neck Plant.(99900277/90-01-01) 2.

Contrary to the requirements of Criterion V of Appendix B to 10 CFR Part 50,

" Instructions, Procedures and Drawings,"

Rockbe.cos test procedure RTP-0050 and TDB test data sheets specified by RTP-0050 were not appropriate to the circumstances because RTP-0050 had been revised improperly and TDBs had not been revised to reflect current practice prescribed in RTP-0050.

(99900277/90-01-02) 3.

Contrary to the requirements of Criterion XI of Appendix B to 10 CFR Part 50, " Test Control," Rockbestos Qualification Report QR-6802, Revision 1, stated that certain post-test insulation resistance values were passing when they were (1) taken with an instrument incapable of valid measurements of the required magnitude, (2) (in some cases not shown to meet the acceptancein other cases act criterion and 3) values.

(99900277/90-01-03) i 86

ORGANIZATION: THE ROCKBESTOS COMPANY NEK HAVEN AND EAST GRANBY, CONNECTICUT REPORT INSPECTION NO.: 99900277/90-01 RESULTS:

PAGE 3 of 24 4.

Contrary to the requirements of Criterion XVI of Appendix B to 10 CFR Part 50, " Corrective Action," Rockbes'.os measures estab.

11shed to correct certain conditions adverte to quality that were identified as nonconformances during previous NRC inspec-tions were not fully effective in that some of the same types of nonconformances were repeated and identified in this inspec-tion.

(99900277/90-01-04)

C.

UNRESOLVED ITEMS:

1.

Control of Off-Standards Reports During the conduct of this inspection, the inspectors identified weaknesses in the system used at the Rockbestos East Granby, Connecticut, plant to maintain control and accountability of Off-Standard Reports.

This issue is discussed in detail in section E.8 of this report.

(99900277/90-01-05) 2.

Activation Energy for Silicone Rubber The inspectors reviewed preliminary data from the new aging and elongation testing program in progress at Rockbestos to redeter-mine the activation energy for their KS-500 formulation of silicone rubber.

The data obtained thus far indicate that the activation energy may turn out to be lower than expected which could impact the qualified life of installed safety-related i

cable. Therefore, this issue, discussed in section E.2.a of this 1

report, remains unresolved pending completion of the new test l

series and recalculation of the KS-500 activiation energy.

(99900277/90-01-06)

D.

STATUS OF PREVIOUS INSPECTION FINDINGS:

1.

(Closed) Nonconformance 84 02, item B.1:

(East Granby Plant)

Contrary to Criterion V of Appendix B to 10 CFR Part 50 end Sections 4 and 5 of the Rockbestos Company's Quality Piacedure Q-10A, dated April 4,

1984, traceability could not be demon-strated for a reel of single conductor insulated wire, (Shop Order 90205-01).

The reels of cable which utilized the above mentioned wire for inclusion as one conductor of a two conductor cable are:

4A-536/D34964, 4A-845/034964B, and 4A-846/034964A.

i 87 l

)

i i

i ORGANIZATION: THE ROCKBESTOS COMPANY NEW HAVEN AND EAST GRANBY, CONNECTICUT REPORT INSPECTION NO.: 99900277/90 01 RESULTS:

PAGE 4 of 24 The NRC inspectors reviewed records, notes and the Rockbestos response to the USNRC concerning this nonconformance.

All reels of completed cable which incorporated the suspect single conduc-tor were identified with nonconformance (OSR) tags and were later scrapped.

The inspectors reviewed Rockbestos Quality Procedure Q-10A, Revision 6, dated July 7,1989, which established the method of traceability for cabling.

The inspectors verified traceability of cabling by reviewing six purchase orders and associated supporting data such as Certified Test Reports, Test Data A&B (TDA and TDB) sheets, shop order numbers and reel numbus.

2.

(Closed) Nonconformance 84-02, item B.2:

(East Granby Plant)

Contrary to Criterion V of Appendix B to 10 CFR Part 50 and Section 13 of the Rockbestos Company's Quality Manual for tool and gage control, six timer units used for rework processes and one unit used for hot modulus testing had not been calibrated as required.

The inspectors reviewed Rockbestos Quality Manual, Section 13, dated March 18, 1988, titled " Tool and Gage Control." Four timer units from the insulation rework area were selected by the inspector to verify implementation of the QA manual.

All units had current calibration stickers on the face of the units fcr the j

curing temperature, footage counter and curing time measurements, i

Unit 812 did not have a calibration sticker for the curing time i

measurement; however, a red sticker was on the unit which displayed do not use until unit is calibrated.

The NRC inspectors also reviewed the calibration records of all four units to verify i. hat information on the calibration sticker was correct.

l 3.

(Closed) Nonconformance 85-03, item B.1:

(East Grandby Plant) 4 i

Contrary to the commitment contained in the Rockbestos corrective action response letter, dated November 23, 1982, with respect to paragraph E.1 in the NRC Inspection Report ho. 99900277/82-02, the procedure for the calibration of irradiation units had not been followed.

This is evidenced by the lack of calibration stickers on the high voltage resistor string and the direct current tachometer digital

readout, as required by the

" Procedure for Calibration of Irradiation Units."

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PAGE 5 of 24 The NRC inspectors reviewed two high voltage resistor strings and verified thet each had a current calibration sticker attached.

The calibration records were also reviewed to verify that the information on the calibration stickers was correct.

4.

(r,losed) Nonconformance 83-03, Item B.2:

(East Granby Plant)

Contrary to Criterion V of Appendix B to 10 CFR Part 50 and paragraph 6 of Part C of Quality Procedure Q-19, dated May 15, 1982, the Chief Technician or desi not reviewed all completed Test Data Sheets "B" (TDBs) gnee had as evidenced by the lack of signature or initial and date, i

The NRC inspectors selected five current TDAs and verified that the test results were evaluated by the QA Analyst or designee to assure the test requirements had been met.

The NRC insectors also reviewed five current TDBs and verified that the Chief i

Technician or designee had reviewed the data sheets as evidenced by initials.

5.

(Closed) Nonconformance 83-03, item B.3:

(East Granby Plant)

Contrary to Criterion Y of Appendix B to 10 CFR Part 50 and paragraph 4 and its subparagraphs b and c of Quality Procedure Q-6A, dated January 12, 1983, the Quality Analyst Supervisor or designee had not revised TDB G323AK, dated March 24, 1983, for Shop Order 81396-03, Reel No. 54093, which was completed June 29, 1983.

The lack of revision is evidenced by an elongation requirement change (lined through) without identification of the responsible person or time.

The NRC inspectors reviewed Quality Procedure Q-6, Revision 12, dated December 19, 1988, titled " Procedure for Control of TDAs and TDBs."

Paragraph 4.1 of Q-6 states in part "for Nuclear Orders, any change to requirements must be reflected in a new revision of the TDB.

Also, it is permissible for the laboratory to write in results of additional tests not part of the TDB or to mark certain tests N/A where necessary."

The NRC inspectors verified. implementation of the above by selecting five purchase orders and associated TDBs.

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PAGE 6 of 24 6.

(Closed) Nonconformance 83-03, item B.4:

(East Granby Plant)

Contrary to Criterion V of Appendix B to 10 CFR Part 50 and paragraph 111 and its subparagraphs 1 and 2 of Quality Procedure Q-9, dated June 7,1982, Quality Control Instruction No.114A, dated, August 10, 1982, did not contain identification of the individuals preparing and reviewing and/or approving the instruction.

The NRC inspectors reviewed CUnlity Procedure Q-9, Revision 9 dated, January 20, 1989, titled " Document Control Procedure."

This procedure lists different classes of documents and who is responsible for approving these documents.

Also, the procedure requires the identification of the individuals preparing, reviewing and approving the docu.aents via initials or signature.

Several documents were reviewed to verify the implementation of this procedure at both the East Granby plant and the New Haven plant during the course of this inspection.

7.

(Closed) Nonconformance 83-03, Item B.5:

(East Granby Plant)

Contrary to Criterion V of Appendix B to 10 CFR Part 50 and Ebasco Services Incorporated's letter dated, April 29, 1982, referencing the D60-01 cable insulation restoration procedure, records were not available which would substantiate that the l

required thermocouple monitoring had been performed.

1 The NRC inspectors reviewed an internal memorandum dated, March 5, 1984, subject " Supporting Data D60-01 Test," laboratory Data Sheet dated, March 2,1984, and a circular temperature chart dated, March 1, 1984. This was a special test that was performed to show objective evidence that heating approximately 10,000 feet of 2/0 16 D60-01 in an oven for 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> at 225'F would bring cable to 225'F.

l Temperature was monitored at top, middle, and bottom of reel at 32 and 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> with a Technoterm 5501 Pyrometer calibrated March 11, 1983.

The test data verified that the inner layers on the cable reel reached 225' uniformly throughout the reel.

8.

(Closed)Nonconformance83-04,ItemB.1:

(NewHaven)

Contrary to the' requirements of Criterion XVII of Appendix B to 10 CFR Part 50 and Section 8.3 of IEEE-323-1974, Rockbestos had not maintained sufficient records to furnish evidence that activities affecting quality were performed.

Specifically, 90

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PAGE 7 of 24 Rockbestos did not have the supporting data for test report F-C3798 entitled, " Qualification Tests of Electrical Cables Under Simulated Reactor Containment Service Conditions Including Loss-of-Coolant-Accident While Electrically Energized," performed by the Franklin Research Center (FRC) for Rockbestos.

During the NRC inspection of January 7-10, 1985, the NRC inspectors reviewed the latest revisions of quality assurance documents, procedures, inspection plans, etc., for compliance with regulatory requirements and verified by cbservation of test i

and QA activities and review of associated records that controls were being implemented except as noted in the nonconformance identified during that inspection.

Subsequently, these nonconformances were closed out during an NRC inspection conducted on April 22-24, May 22-24 and September 11-12, 1985.

During this inspection, the NRC inspectors reviewed the latest i

revisions of selected QA documents, procedures, inspection plans, qualification test reports and associated records and verified that controls were being implemented.

I 9.

(Closed)Nonconformance83-04,ItemB.2:

(NewHaven)

Contrary to the requirements of Criterion XI, Appendix B to 10 CFR Part - 50, Rockbestos did not evaluate or document the evaluation of 4 cable failures which occurred during the F-C3798 qualification test and 11 cable failures that occurred during the i

F-C3859-2 qualification test.

Rockbestos concluded in both of the above instances that it is -

i virtually impossible to ensure that damage to the ceble samples can he eliminated at the point where they pass througt the flange penetrations.

Additionally, both the expansion characteristics of the insulation-and the thermal characteristics of the jacket material exacerbate the penetration problem. During the requali-fication program, the NRC inspectors verified that Rockbcstos -

evaluates and documents' cable failures.

Also, several follow-up items in this report confirm that Rockbestos is evaluating and documenting cable failures.

10.

(Closed). Nonconformance 83-02, Item B.1, Example (2) (New Haven)

Contrary to the. requirement of Criterion II of Appendix B to 10 CFR Part 50, Rockbestos failed to e:,tablish and implement an adequate QA program for its safety-related equipment qualifi-cation (EQ) testing effort.

This is evidenced by the number of i

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ORGANIZATION: THE ROCKBESTOS COMPANY NEW HAVEN AND EAST GRANBY, CONNECTICUT REPORT INSPECTION NO.: 99900277/90-01 RESULTS:

PAGE 8 of 24 generic deficiencies identified during this inspection.

Examples are:

(1) numerous calibration system deficiencies; (2) use of inadequate test instrumentation; (3) the failure of engineering to develop, review, and approve test plans; (4) the failure of engineering. to specify and describe test requirements; (5) the i

failure of engineering to identify and evaluate test nonconfor-mances, variations, and deviations during testing and to document the same in test reports; (6) the failure of eng)ineering to perform adequate evaluations of test results; (7 technical inconsistencies between raw test data and final EQ test reports; and (8) the failure of the QA and control organization to audit and monitor EQ testing.

The instrument in question was a LOCA chemical sp) ray flowmeter as stated in NRC Inspection Report 85-01 0.7.C. (2.

The maximum error recorded during before-use calibration of the flowmeter was 6.3 percent, exceeding the 5 percent allowed by paragraph 13.2.5 of t1e Rockbestos QAM.

A staff memorandum, dated December 11

-1984, indicated that the flowmeter accuracy specification should be changed to 120 percent.

However, procedures had not been revised accordingly.

A review of Section 13.2.6 of the Rockbestos QAM during this inspection indicated it now required that equipment for the Qualification and Test Department "shall not have an error greater than plus or minus 5 percent except ~as specifically excepted by the Technical Manual (TM) or applicable calibration procedures."

Section 4.2.1(b) of the Rockbestos TM states "the percent deviation from the standard for spray flowmeters -shall not exceed 20 percent."

11.

(Closed) Unresolved Item 83-04, Item C-Rockbestos had used single conductor or single twisted pair test results to establish qualification for a vastly different cable, such as an 80-:;onductor (80/C) cable.

This was a questionable use of similarity 'analy(sis.

The NRC inspector observed that I

requalification program RP) specimens now included 7/C and 2/C completed cable section samples as representatives of multicon-ductor (M/C) types.

Results of requalification. testing indicated satisf actory performance of both the jacketed, multiconductor cable and unjacketed single conductors of the Firewall III and Firewall SIS types of XLPE insulated wire respectively.

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PAGE 9 of 24 12.

(0 pen)UnresolvedItem83-03,ItemC The NRC inspector observed that the "011 Aged" section of work-sheet EGTL No. 4 indicated that calculations had been indicated complete by being stamped and dated.

However, data had not been entered in the " Average" blocks of the table.

Rockbestos stated that the spaces were not completed because additional data was required prior to performing the calculation.

Apparently, there were no requirements that addressed stamping / dating of in-process test calculations.

This item was not addressed during this inspection.

13.

(Closed) Unresolved Item 82-02, Item C. 2 (East Granby Plant)

It was not apparent that brazed connection of conductors had been measured, or a procedure existed for accomplishing the task. The validity of retests was questionable for accelerated water i.

absorption and flammability properties.

During NRC inspection

~83-03, the NRC inspector deferred action on this item until staff technical evaluation was complete.

During this inspection, the NRC inspectors reviewed Rockbestos' brazing procedure, dated March 1,1985.

The procedure delineated the steps to be per-formed during this operation and the inspection characteristics to be inspected.

No brazing of conductors in the stranding area was being performed during this visit to verify implementation.

However, the procedure appeared to be adequate to perform this operation.

(410 sed) Follow-upItem81-01 This item involved the evaluation of the original qualification testing of -the RSS-6-100 series coaxial cable.

In May 1981, General Atomics Company (now Sorrento Electronics Division of GA Technologies) submitted a 10 CFR Part 21 report on the failure of Rockbestos RSS-6-104 coaxial cable to maintain sufficient insula-tion resistance during' main-steam-line-break (MSLB) and oven tests.

Rockbestos responded to this failure by (a) redesigning the construction of the coaxial and triaxial cable series; (b) performing qualification type tests on the " newer" second and third generation. coaxial and triaxial cables; (c) recognizing that the redesign causes an additional technical concern for the larger diameter members of the RSS-6-100 series c3bles (for l

example RSS-6-109); and (d) performing specialized qualification L

tests for customers on the RSS-6-109 cables to alleviate qualifi-cation concerns.

This item was examined extensively during the j

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ORGANIZATION: THE ROCKBESTOS COMPANY NEW HAVEN AND EAST GRANBY, CONNECTICUT REPORT INSPECTION NO.: 99900277/90 RESULTS:

PAGE 10 of 24 inspection of June 6-10, 1983, and a number of questions concerning whether qualification had been demonstrated by the Rockbestos effort were raised.

These concerns were outlined in inspection report 83-01 and remained open pending the NRC evalua-tion of the Rockbestos requalification test program being conducted in response to inspection reports 83-01, 83-02, and 83-04.

The results of the Rockbestos requalification program on coaxial, twinaxial and triaxial cable of the RSS-6-100 series were documented in Rockbestos qualification report QR-6802.

QR-6802 indicates that the samples carried nominal / rated voltage and current successfully for 28 days of loss-of-coolant-accident (LOCA) and post-LOCA simulation exposure intended to follow a IEEE 323-1974 LOCA profile.

The published values of insulation resistance during and after LOCA exposure, for these cables, while not necessarily unacceptable for many qualified applica-

- tions, _ may still be too low with respect to the values needed for continuous satisfactory performance of the GA high range radiation monitor under some design basis accident conditions as first discussed in the GA 10 CFR Part 21 report.

During this inspection, the NRC inspectors reviewed QR-6802, raised technical concerns described in 1section E of this report, and identified the nonconformance cited in paragraph B.3 of this report.

15.

(Closed) Follow-upItemIR 99900277/86-01 paragraph E.4 The test failures of the triaxial cable sample C2 and one conduc-tor of the twinaxial cable sample D1 were further investigated by Rockbestos after the 86-01 NRC inspection.

The test report 4

documents the failures and the failure analysis.

The failure analysis concluded that extrusion of the cables' through LOCA chamber penetrations caused-severe stress on the cables and probable mechanical damage.

The information in the test report was considered sufficient for utilities to determine the suitability of this cable for their plant applications.

16.

(Closed) Follow-up Item IR 99900277/8E-02 paragraph - E.1.e.(4)

L L

The analysis of chemically cross-linked polyethylene-insulated Firewall SIS. sample failures had concluoed that failures were L

attributable to handling damage which ju',tified further testing.

According to records, Rockbestos repaired the handling-damaged sections of faulted samples and reperformed the voltage withstand test with satisfactory results.

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PAGE 11 of 24 17.

(Closed) Follow-up Item IR 99900277/85-02, Paragraph E.2.b The NRC inspectors had observed an incorrectly set power supply voltage prior to a LOCA test.

Rockbestos' normal pretest checks had not detected the problem.

The inspectors recommended that a pretest check procedure be instituted to prevent such problems.

The NRC inspector interviewed the test technician during this inspection and found that Rockbestos has made a practice of tagging each loading circuit with sample numbers and voltage and current requirements to preclude similar problems.

18.

(Closed) Follow-upItemIR 99900277/85-02 Paragraph E.4.b During inspection 85-02, the NRC team questioned the Rockbestos analysis showing that LOCA test conditions envelope a 2-minute main steam line break to 381*F based on a thermal lag heat transfer analysis.

The analysis has not been modified by Rockbestos.

They do not include the analysis in test reports when the actual test data is provided. Rockbestos stated that if a customer requests analysis for a higher temperature than was i

tested Rockbestos provides information that includes Arrhenius equivalent degradation analysis, justification of peak tempera-ture based on short circuit rating and a report with information with which the customer can calculate insulation resistance at a given temperature above the LOCA test temperature.

i 19.

(Closed) Follow-upItem,IR 99900277/85-01 Paragraph E.11-During a previous inspection, the NRC inspectors identified some elements on the Rockestos personnel QA/QC qualification record sheets of two test personnel that were marked with arrows proceeding from elements above in the block for instructors' initials and date of completion and were not inficated as completed as required by quality procedure _(QP) Q-25B, Revision "0", dated April 26, 1983. During this inspection, the in u ctor reviewed Rockbestos QP Q-25B, Revision 2 dated March 20, 1986.

Paragraph 2 of Q-25B now states, in part: "If each item is not separately initialed or signed, applicability of initials / signa-

_l tures shall-be clearly shown by -suitable brackets, arrows or similar means."

20.

(Closed) Follow-upItem,IR 99900277/85-01 Paragraph E.12 During a previous inspection, Rockbestos could not produce the record of the survey reportedly performed by Rockbestos on the calibration laboratory they were using at that time.

During this 95

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PAGE 12 of 24 inspection, the NRC inspectors reviewed a Rockbestos survey that had been completed on September 20, 1988 on their current cali-bration laboratory, EIL Instrument, Incorporated.

No anomalies were identified during this review.

E.

OTHER FINDINGS OR COMMENTS 1.

Rockbestos 10 CFR Part 21 Notification on Shielded Cable A 10 CFR Part 21 report was submitted to the NRC by Rockbestos on November 6,

1989 concerning a Rockbestos electrical cable 1

supplied to Consumers Power Company for use in the Palisades Nuclear Plant.

The-Consumers Power purchase order required the

.I special purpose, 600V, 90'C, shielded, 2/C, #12 AWG control cable i

to have a #34 AWG tinned copper braided shield, but it was supplied with a #34 AWG stainless steel braided shield instead.

The Rockbestos manufacturing product specification (MPS) for this order (revision A), dated April 20, 1989, prepared by the specification engineer and approved by the supervisor of specifi-cation engineering, specified the shield material as Type 302 stainless steel.

To prevent recurrence, Rockbestos has now developed a checklist which is used to check the MPS against the requirements of the shop order, Rockbestos Order Form 441, which is supposed to include all purchase order requirements.

The NRC inspector reviewed the checklist and determined that it would be adequate to verify that the correct technical requirements are translated into the MPS. A Rockbestos internal memorandum, dated November 6, 1989, stated that Consumers Power had performed a calculation with which they determined that a problem would not exist with installed circuits 200 feet or less in length.

Consumers Power reportedly believed that all installed circuits were less than 200 ft and that no cable would have to be removed.

2.

Notifications to the NRC regarding Rockbestos Silicone Rubber insulated cables The Rockbestos Company notified the NRC on October 2,1989 of a concern relating to activation energy. and ampacity values for Rockbestos silicone rubber (SR) insulated cable.

EcoTech/ RAM-Q Industries notified the NRC on October 9, 1989 of their concerns relating to findings from their source verification and perfor-mance-based audit of Rockbestos which was performed for Northeast Utilities (NU).

On January 2,1990, NU notified the NRC of a related concern regarding Rockbestos "Firezone-R,"

silicone rubber (SR)-insulated cable.

All the notifications pertained to 96

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PAGE 13 of 24 1

the same technical concerns, SR activation energy and Firezone-R

- ampacity, and related QA concerns. Specifically:

a.

Silicone Rubber Activation Energy issue:

l Rockbestos conducted an accelerated thermal aging program on their KS-500 formulation of SR in the mid 1970's.

The elongation test data generated from that program was used to develop time-temperature curves at 200'C,190*C and 180'C.

h The original data supporting the plots of the curves is no longer available, because it was apparently lost or inadver-tently discarded.

During the audit of Rockbestos mentioned above, EcoTech/

RAM-Q reviewed the data that Rockbestos had taken from the I

curves to calculate the SR activation energy and found problems with (1) the basis for the curves, (2) the basis for selection of the end point line at 50% retained elonga-

-tion, and (3) with the proximity of the-points actually chosen for the regression analysis to the actual intersec-tion of the time-temperature curves with the percent-retained elongation lines.

The activation energy value calculated by Rockbestos at 50%

,etained elongation was 1.805 ev.

However, in addition to the data supporting the Arrhenius plots not being available, the intersection points of the 190'C and 180'C curves with the selected 50% retained elongation line were not bounded by actual data points used to plot the curves, but were on sectors of those curves that were extrapolated from plotted elongation test results.

Furthermore, the points Rockbestos selected for the regression analysis did not exactly coincide with the curve-line intersections, but were skewed, resulting 'in an erroneously higher calculated activation energy than the actual intersections would have yielded.

According to the Ecotech audit report, their own calcula-tions performed for comparison, yielded a value of 1.718 ey at 50% retained elongation.

The lower value of activation energy reduces the qualified life of 125 C-rated SR from 95.06 as originally calculated, to 69.98 years at 125'C.

EcoTech/ RAM-Q then calculated activation energy at 55%

retained elongation in order to obtain intersection points on all three curves bounded by actual data points and

+

obtained a value of 1.622 ev.

Using this lower activation 97

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PAGE 14 of 24 energy, the Arrhenius formula yields a qualified life for SR further reduced to 49.77 years for the rated ambient temperature.

During this inspection the NRC inspectors performed regres-sion analysis using data.from the same plot.

At 55%

retained elongation, the NRC inspectors calculated the SR activation energy at 1.624 ev.

The result at 60% retained elongation was 1.611 ev,1.753 ev at 65 % and 2.092 ey at 70% retained elongation.

These figures give a reasonable generit, activation energy for the KS-500 compound in the range of 1.61-1.62 ev.

Using this activation energy valvo gives an acceptable qualified life for Firewall SR but the resultant qualified life for Firezone R is only abost 30 years - because of the limited amunt of accelerated thermal aging done on this type of cab ic.

Nevertheless, the amount of actual aging degradation that would be suffered by SR insulation in the Firezone R at 90*C in an actual installation is insignificant relative to the thermal capability of SR.

Less than 5% reduction in the elongation of SR would be expected in 40 years at 90'C (90'C because Firezone is arbitrarily derated to 90'C for it to be compatible with other cables with which it could share-raceways). Accordingly, it would be appropriate to select a higher retention of elongation figure for calculating the activation energy for the Firezone R, thus yielding values higher than 1.62 ev, and hence obtaining longer qualified life.

Calculations using the existing plots indicate that it is

'likely that there can be sufficient demonstrated qualified life for most plant applications of both the Firewall SR and the Firezone R if the data from which they are constructed are presumed to be correct.

However, in order to confirm the validity of_ the original curves-used to calculate activation energy for thermal a environmental qualification (EQ)ging purposes. in support of of Roc 6estos Firewall SR-and Firezone R cables insulated with the KS-500 compound, Rockbestos is conducting another aging program of the KS-500 formulation SR compound which is scheduled to be completed by mid-April of 1990.

Some initial data from the retests in progress ' appears to indicate a fairly significant deviation from previous data.

Pending completion of this aging and 98

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RESULTS:

PAGE 15 of 24 testing program, and further review during a future NRC inspection, this issue is designated as Unresolved Item 99900277/90-01-06, b.

Firezone R Ampacity issue:

Rockbestos had a contract with NU to provide them with some

1. 4/0 AWG Firezone R cable in accordance with Rockbestos sales specification RSS-5-144.

RSS-5-144 provided ampacity values for commonly used cable sizes but not for #4/0 AWG, Upon request for this information by NU, Rockbestos calcu-l lated a value for NU using a computer code developed in-house and based on the standard methodology, and provided this value to NU. During the Ecotech audit mentioned above, it was determined that ampacity calculation had not been reviewed and approved in accordance with design and document control QA procedures.

During this inspection, Rockbestos pointed out that standard ampacity tables such as those promulgated in the National Electric Code (NEC) can be used for most cables.

However, due to its special application and design, Firezone R cable does not conform to the model used to calculate the nominal i

values in the NEC. Among other substantial differences from conventional cable design to enhance fire environment oper-ability, Firezone R uses a combination copper and nickel conductor.

Therefore, this is considered special purpose cable and ampacity needs to be specially calculated.

In order to determine the accuracy of the Rockbestos figures and evaluate their calculation method, the NRC reviewed a manual calculation of the ampacity for 4/0 Firezone R cable using the standard formula, but with the parameters and values unique to Firezone R.

The manual calculation gave the same answer as the Rockbestos computer code that was the basis for the ampacity figure provided to NV.

The computer code had also been used to calculate the ampacities listed in RSS-5-144 for the other available sizes. The calculation 1

input and methodology were also reviewed and no deficiencies were noted.

Thus, the ampacity calculations appear to be correct _ based on agreement between the computer code and the manual calculation results.

Although no technical deficiencies were identified with either the ' methodology - or the results of the ampacity calculations, nonconformances to QA requirements were 99

' ORGANIZATION: THE ROCKBESTOS COMPANY NEW HAVEN AND EAST GRANBY, CONNECTICUT REPORT INSPECTION NO : 99900277/90-01 RESULTS:

PAGE 16 of 24 identified.

Rockbestos did not consider ampacity to be a cable design parameter, although it is a function of cable design.

However, vendor-provided ampacity values based on cable design and standard installation factors are used as nuclear design information by plant engineers in conjunction with plant-specific installation-related factors that determine the final installed ampacity and are a cable sizing consideration.

Therefore, ampacity must be treated as other safety-related design information and the calcula-tion of ampacity is an activity affecting quality, requiring QA controls.

The failure to properly and formally document, review and document, review approve the computer code, and to formally (by someone other and approve the individual calculations than the person directly responsible) constituted a

nonconformance with respect to Rockbestos QA procedures on design and document control and Criterion 111 of Appendix B to 10 CFR Part 50.

This is designated as Nonconformance 99900277/90-01-01.

I 3.

Quality Assurance Program The NRC inspection team reviewed certain aspects of the implemen-tation of Rockbestos' QA program by evaluating corrective actions taken by them on previously-identified NRC findings at the New Haven and East Granby plants and also the findings identified by EcoTech/ RAM-Q during their audit for NU.

Also, the inspectors reviewed current quality procedures,_ qualification test reports, qualification test plans, training records, audit reports, test data sheets, calibration records, engineering calculations, and certificates of conformance.

The inspectors ' found that the portions of the QA program reviewed were ~ being implemented in

-accordance with Appendix B-to 10 CFR Part 50 with the exceptions noted.

The types of findings identified in Section B of this report such as test control, design control, and failure to follow procedures are repetitive of findings identified in previous NRC inspections, showing lack of adequate corrective action and are designated Nonconformance 99900277/90-01-04.

4.

Review of Test Report for Rockbestos Silicone Rubber KS-550 The NRC reviewed qualification report QR-8802 for' silicone rubber KS-550 insulation and supporting data.

Standard Rockbestos proce-dures and test parameters were used in the test.

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PAGE 17 of 24 Thermal aging consisted of 1400 hr. at 180*C to give a qualified life of 40 yr, at 125'C.

The testing used 1-conductor, #14 AWG cables, one aged and one unaged.

The test plan called for.a random selection from each of the two aged and two thermally-unaged specimens.

Following irradiation of the 4 samples, one of the thermally-unaged specimens failed electrical tests. The only indication of mechanical damage from the visual observations was small indentations at 3,

9, and 12 feet along the sample.

Rockbestos believed that the failure was caused by (visually) undetectele mechanical. damage in transportation from Isomedix, the irradiation facility.

During the LOCA test, the #14 AWG cable was loaded to only 10A.

Although this is acceptable, it is possible that this reduced loading could create an unnecessary ampacity limitation for certain plant applications.

No findings were identified in this review.

5.

Unauthorized Repairs to Cable Armor NRC Region III referred to the Vendor Inspection Branch a report to them from Northern States Power Company (NSPC) regarding the licensee's discovery of damaged jacket material on some Rockbestos cable.

NSPC purchased Rockbestos 3/C #10 Firewall III cable with aluminum interlocked armor from the Rockbestos Company under Purchase Order NSP HIAWO2607, Rockbestos Shop Order 72908 for use in Prairie Island Nuclear Generating Plant.

NSPC discovered while installing the cable that the armor had separated and that the jacket material was damaged; however, the

-conductor insulator was not damaged. Ccring this inspection, the inspectors asked Rockbestos why the jacket material was damaged.

Rockbestos stated that an unauthorized repair was made (TIG welding instead of soldering).

Not following the repair procedure for interlocked aluminum armor resulted in overheating the jacket, which caused jacket embrittlement and separation.

Rockbestos determined that the condition of the

filler, flame-retardant polyethylene, indicated that the small area under the armor repair was exposed to a temperature of 160*C to 165"C.

This temperature degraded the jacket material (Hypolon), but not the crosslinked polyethylene (XLPE) insulation.

Rockbestos determined ' that the insulation was not damaged based on the following:

101

1 ORGANIZATION: THE ROCKBESTOS COMPANY NEW HAVEN AND EAST GRANBY, CONNECTICUT REPORT INSPECTION NO : 99900277/90-01 RESULTS:

PAGE 18 of 24 a.

The XLPE insulation meets the requirements of ICEA S-66-524 which requires survival and unaffected operation after exposure to 250*C short circuit conditions, b.

The insulation had been qualified to the double transient of IEEE 323-74 for combined PWR and BWR (340'F-171'C) profile after being fully-aged and without a protective jacket.

l c.

Heat aging studies have confirmed thet this insulation will retain 85% of its original elongation properties after seven 3

days exposure to 158'C.

d.

Rockbestos confirmed that Firewall III meets the vertical tray flame test without the added protection of the jacket.

As part of corrective action, the Rock 5estos QA Manager sent a memorandum to his manufacturing manager and general foreman l

prohibiting repairs on aluminum interlocked armor.

The only other shipment of this cable was to Virginia Power.

A review of off-standard reports (OSR's) showed that no repairs were conducted by Rockbestos on that order.

6.

Review of Rockbestos Qualification Report QR-6802 j

The inspectors reviewed Rockbestos environmental qualification report QR-6802, Revision 1, dated July 2,1987, for environmental qualification of Rockbestos coaxial, twinaxial, and triaxial cable for adverse nuclear Class 1E service.

Paragraph 12.1.B of Data Section 2 of the report gives the acceptance criteria for insulation resistance (IR) measured per test plan.TP-5803 in accordance with ICEA standard S-19-81.

for IR, expressed in terms of gegohms (10}he acceptance criterion ohms) per 1000 feet of cable length, was 100,000 (10 meggms for types "A,"

"D,"

and "E" cable samples and 1,000,000 (10 ) megohms' for types "B"

and "C"

samples.

The-table of post-loss-of-coolant-accident (LOCA) sample evaluation IR values following Section 12.2 of the report shows all results to be passing. The post-LOCA insulation resis-tance values were listed as " greater than" 360,000 megohms (corrected for 1000 feet) for type "B"

and "C"

samples and 4,500,000 megohms (uncorrected as measured) for type "D" samples.

In order to determine the validity of these readings, the inspec-tors reviewed the relevant procedures, tech manuals laboratory 102

ORGANIZATION: THE ROCKBESTOS COMPANY l

NEW HAVEN AND EAST GRANBY, CONNECTICUT REPORT INSPECTION NO.: 99900277/90-01 RESULTS:

PAGE 19 of 24 data sheets, interviewed the Rockbestos laboratory test equipment calibration technician and reviewed his calibration records for Genrad 1864 Megohmeter Bridge, Serial Number 2311, indicated in QR-6802 as having been the IR testor used for the listed IR measurements.

The test plan for this test. TP-5802, called for use of this model instrument, set for 500 volts per ICEA S-19-81.

The Rockbestos general procedure for measuring IR on coaxial, twinaxial, and triaxial cables, RTP-0050-3, dated August 9, 1978, revision dated January 1,1986, was in effect at the time the post-LOCA irs were taken, but it prescribes use of the Kiethly 610C electrometer and a separate. 240-amp power supply or

" current set" in order to obtain valid IR readings of the high magnituderequiredonshort(inthiscase,lgfeet)gestsamples of coaxial type cables to assure that the 10 and 10 -megohm per 1000 feet acceptance criteria are met.

Review of the technical manual for the Kiethly 6100 electrometer indicated that its gable range, used as an ameter, extends down to as -low as 10~

amperes.

Using the current set voltage setting of 100 volts specified by the type "B" cable-specific laboratory test data sheets, or "TDBs" referenced in RTP-0050 (suchasTDBG338ZforspecificationnumberRSS-6-100/LEcoaxig cable), the minimum current value would equate to an IR of 10 megohms.

However, review of the technical manual for the Genrad 1864 Megohmeter Bridge used revealed ghat, at 500 volts, its useful range.is not greater than 20 X 10 megohms (20 terohms) at the greatest recommended scale and needle deflection. Therefore, when-full scale readings of greater than 20 terohms were obtained during IR measurement of the 18-foot B and C samples, these were corrected for 1000- feet by dividing 20 terohms by a factor of 1000/18 (approx 55.56) yielding a figure listed in the QR-6802 data table as " greater than" 360,000 megohms per 1000 feet. The indication in the table that these were passing values is erroneous in that a maximum reading of " greater than 360,000 megohms" does not demonstrate that the acceptance criterion for the B and C samples of 1,000,000~meghoms is met.

The value of measured IR for the 18-foot D samples was listed in the report: as 4,500,000 megohms, which, when corrected for 1000 feet by dividing by a factor of 1000/18 yields a value of 81,000 megohms. The 4,500,000 megohm result was therefore characterized erroneously as " pass" in the table, when, in fact, the measured value corrected for 1000 feet is less than the specified accep-tance criterion of-100,000 megohms.per 1000 feet for D samples.

t 103

ORGANIZATION: THE ROCKBESTOS COMPANY HEW HAVEN AND EAST GRANBY, CONNECTICUT REPORT INSPECTION L

NO.:

99900277/90-01 RESULTS:

PAGE 20 of 24 L

in the interview of the Rockbestos laboratory test equipment calibration technician and review of the calibration procedures for the Genrad 1864 Megohmeter Bridge, it was learned that the procedure specifies that the instrument be calibrated such that the maximum error does not exceed 10% of the full scale reading.

The calibration records for Genr.ad 1864 Serial Number 2311 used for the QR-6802 IR measurements confirmed that this and other 3

Genrad 1864s are not calibrated above about 5 X 10 megohms.

This range is about.four orders of magnitude below the minimum rangeofaboutpX10 megohms needed to meet the lower IR speci-fication of 10 megohms-per-1000-feet with short sample (18-FT)

IR measurements'.

The calibration technician explained that this was the maximum range at which these instruments would have less-than 10% error; so that it was their practice not to attempt to calibrate them at any higher ranges and determine what the error was.

This meant, therefore, that Genrad 1864 Serial Number 2311 was uncalibrated for the range of IR values for which it was being used during post-LOCA sample evaluation for TP-5802 as published in QR-6802.

Thus, the values listed were (1) taken by an instrument that was not only uncalibrated in the range of interest, but was incapable of providing. a valid reading of the required magnitude, (2) not demonstrated for samples B and C to have passed as claimed, and (3) were below specification for the D samples.

This failure to I

adequately evaluate the test results to assure that requirements are met constitutes a nonconformance with respect to Criterion XI L

-of Appendix B to 10 CFR Part 50 and is designated as Nonconfor-mance 99900277/90-01-03.

7.

Review of IR Measurement Practices The QR-6802 report also contained, but made no reference to, certified test reports (CTRs) of tests done by production personnel on cable from the reels from which the.TP-5802 samples were taken using routine or standard IR measurement practices.

These reports indicated that satisfactory irs were obtained, e.g.,

on a Cfor type B samples showing 1.53 X 10{R, dated February 27, 1986, megohms per 1000 feet.

Therefore, the inspectors reviewed the current procedures and those in effect at the time of the testing in question, including the general procedures-and individual test data sheets or "TDBs" referenced in the procedures.

The Rockbestos general procedure for measuring IR on coaxial, twinaxial, and triaxial cables, RTP-0050, dated August 9,1978, revision dated September 24, 1986 (current revision), prescribes 104

ORGANIZATION: THE ROCKBESTOS COMPANY NEW HAVEN AND EAST GRANBY, CONNECTICUT

-REPORT INSPECTION NO.: 99900277/90-01 RESULTS:

PAGE 21 of 24 the use of a Genrad 1864 Megohmeter Bridge set for the voltage specified in the "TDB" (cc51e-specific laboratory test data sheet type "B").

This revision of the procedure also still specifies the use of a separate 240-amp power supply which was used in accordance with previous revisions with the Kiethly 610 electro-meter, but which is inconsistent with the use of the Genrad 1864 which is AC powered and contains its own internal test voltage /

current power supply.

The procedure then requires conversion of the measurement to units of megohms per 1000 feet of cable and reflects the current prac',1ce of measuring an entire cable reel at once by giving 2000 feet as the cable length measured in the sample calculation.

Thus, the procedure was improper 1v revised l

because the reference to the separate power supply should have been deleted when use r,f the Genrad 1864 was instituted.

The laboratory test data sheets referred to in RTP-0050, such as TDB G338Z, Revision E, dated December 17, 1874, for RSS-6-100/LE coaxial cable, however, specify a length of 20 feet for a specimen to be used for the test and specify a current set (power supply) voltage of 100 volts.

The sample length, current set, and voltage specified'are not used with a Genrad 1864, but would have been used with a Kiethly 610C electrometer consistent with previous revisions of RTP-0050 (such as the revision dated January 6,1986 in effect at the time of the testing in question) and requiring a separate power supply.

While it appeared that the IR measurements given in the CTRs were taken in accordance with effective procedures, the practice since the later revision which established use of the Genrad 1864 on L

whole cable reels has not been in accordance with contradictory L

and out of date TDBs referenced.

It should be noted that the l

post-LOCA IR values listed in QR-6802, while not all meeting the stated acceptance criteria, could be considered acceptable for some qualified applications, but since they were taken in an uncalibrated and in some instances, invalid range. of the instru-l-

ment, the error associated with those values is indeterminate.

The deficiencies noted with respect to revision of RTP-0050 and lack of revision to associated TDBs constitute nonconformances with respect to the requirements of Rockbestos quality procedures.

and Criterion V 10 CFR Part 50 Appendix 8 in that they represent lack of procedures appropriate to the circumstances for activi-ties-affecting quality and are designated as Nonconformance 99900277/90-01-02.

105

i ORGANIZATION:

THE ROCKBESTnF COMPANY NEW HAVEN AND EAST GRANBY, CONNECTICUT REPORT INSPECTION NO.:

99900277/90-01 RESULTS:

PAGE 22 of 24 8.

Followup on Allegation of Shipment of Damaged. Unrepaired Cable The NRC received an allegation that a 1,200-foot reel of 37-con-ductor (37/C), nuclear safety grade cable with damaged insulation on one conductor had been shipped from Rockbestos' East Granby plant to a nuclear utility without the damaged insulation being repaired.

The cable in question was identified as 37/C Firewall III, with radiation cross-linked polyethylene (RXLPE}

insulation.

The reported damage was a 1-inch section of insula-tion stripped from a conductor with blue insulation with a black stripe about 400 feet from the end of the 1,200-foot reel, and was believed to have been caused by a burr on the flange of the metal feed. reel which snagged the insulation of the conductor as it was being unwound from this reel into the jacket extruder machine to be jacketed with the other conductors.

The incident allegedly occurred near Labor Day,1988, and the alleger believed that the cable had been sent to the Philadelphia Electric Company plants) (based on PECO's being the only customer-known by the (PECO) licensee for the Peach Bottom and Limerick nuclear alleger to 'have bought such unusual cable in the past.

An OSR reportedly was initiated on the damaged insulation, but the dis-position of the OSR was unknown, and the alleger stated that he believed that the cable had been shipped without repair.

From the East Granby plant log of all shop orders for 1988, the NRC inspector identified all the 1988 shop orders for PECO.

The inspectors then searched the files of all shop orders (all custo-mers) for 1988 and identified only one for 37/C cable for PECO and - only one other one for 37/C cable.

Review of those shop orders revealed that the one for 37/C cable for PEC0 had been shipped in April of 1988, and the other one was shipped late in September,1988, but was for a' commercial electrical supplier.

The "0SR Report, the computer printout of the description and status of presumably all OSRs generated in a given year, was then searched for all OSRs involving PEC0, damaged insulation, 37/C cable, and the time frame of interest.

The inspector identified only one OSR for 37/C cable that was issued during the time frame i

of interest, but the customer was the commercial electrical supplier mentioned above. The problem was with insulation'on one conductor, but it was a striping discrepancy, not damage, and the color code marking was different than that alleged.

Review of related documentation, including the shop order, the OSR sheet for the shop order, and the OSR file copy itself, revealed that there indeed had been a 1,200-foot reel which was 106

ORGANIZATION: THE ROCKBESTOS COMPANY NEW HAVEN AND EAST GRANBY, CONNECTICUT REPORT INSPECTION NO.: 99900277/90-01 RESULTS:

PAGE 23 of 24 one of three on this order, but the OSR was on one of the other two reels and had been satisfactorily dispositioned.

In addition, although the shop order was for 37/C cable, it listed the cable as "Pyrotrol 210," the Rockbestos trade name for their non-environmentally qualified commercial grade version of the same formulation of RXLPE-insulated cable, and not nuclear safety grade Firewall 111 as the alleger had described it.

A search of all OSR file copies written during the time frame of interest was conducted, but none were found that either' matched or resembled the allegers description of the problem.

Suring this activity, the ~QC supervisor described the details of the routine OSR processing practice, disclosing the facts that led i

the inspectors to identify a lack of adequate OSR accountability.

L Rockbestos documents and tracks the disposition of the various i

conditions adverse to quality that may occur during cable L

production, including damage, nonconformances, etc., using a I

system of three-part tags or chits called Off-Standard Reports (OSRs).

According to Rockbestos' description of-this process, when a condition that requires some resolution is identified, the 4

cognizant supervisor is notified by the identifying worker or quality control (QC) inspector and then the QC inspector is notified, if not already informed, and fills out an OSR, attaches y

one part to the affected item, as applicable,- and turns in the copies to the' QC office for disposition by the QC supervisor.

The QA manager's assistant at the East Granby Plant records the OSRs that come into the office, enters the information into the OSR tracking computer data base and files the copies for review.

L

' mines the corrective action and causes it to be taken, then and-disposition by the QC supervisor.

The QC supervisor deter-l determines final disposition which is noted on the file OSR copy and the status is entered in the data base.

When corrective action, such as repair, is completed, the OSR tag is removed from the affected item, noted as completed and returned to QC, there the status is noted by the QC supervisor, entered in the tracking data base, and the closed-out OSR is filed.

The weaknesses identified in the practice of controllin the generation, tracking and disposition of OSRs were that' (1)g 0SRs are not necessarily logged and/or copies retained by the identi-fying worker.and/or QC inspector and (2) although the OSR forms have printed serial numbers, no record is kept of which pads or packs of OSRs, by numsrical series, are issued to various plant j

personnel; thus accountability for each form, by serial number, 107

j ORGANIZATION: THE ROCKBESTOS COMPANY NEW 'VEN AND EAST GRANBY, CONNECTICUT REPORT-INSPECTION

- NO. : ~ 99900277/90-01 RESULTS:

PAGE 24 of 24 is not maintained. This condition creates the potential for OSRs to be lost, damaged, or improperly cleared without an auditable accounting system to detect such occurrances.

In response to these concerns, the Rockbestos QA Manager agreed to investigate the matter and take appropriate corrective action.

The allegation follow-up indicated that if there had been damaged insulation shipped unrepaired, and if the OSR had been lost or even improperly removed, the cable was not shipped to PECO or any other nuclear utilities.

However, this allegation, while not substantiated, did serve to reveal the deficiencies in the Rockbestos OSR system which could potentially allow such problems to go undetected with nuclear safety-related material.

Therefore, this issue involving control of OSRs remains unresolved pending completion of corrective. action by Rockbestos and further review during a future NRC inspection, and is designated Unresolved item 99900277/90-01-05.

F.

PERSONS CONTACTED:

New Haven

• • H. Littlehales, QA Manager

• • G. Klein, Vice President of Engineering

• • B.-Gehm,.Jr., Manager, Electrical & Production Engineering M. Mennone, Test Technician J. Morgane111, Qualification and Test Administrator East Granby

• • G. Littlehales, QA Manager

' C. O' Conner, Chief Technician (Laboratory) )

B. Jendza, QC Supervisor (East Granby plant l

D. Schuler, Manager of Radiation Services 1

l M. Bliss, QA Analyst l

1 l-108

Selected Bulletins anx1 Infornation Ibtices Concerning Adequacy of Verxlor Audits and Quality of Vendor Products ISSUED 1.

Information Notice No. 89-70 Possible Irdication of Misrepresented Suppleont la Vendor Parts 2.

Infonution Notice No. 90-37:

Sheared Pinion Gear to Shaft Keys in Limitorque Motor Actuators 3.

Potential Failure of G.E. Magna-Blast Circuit Breakers aid AK Circuit Breakers

4. -

Information ibtice No. 90-43:

Mochcatical Interference with Thernul Trip Function in G.E.

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1. REPORT NUM6i R

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a.Tif LE AND SUBitTLg Vol. 14, No. 2 k.

Licensee Contractor and Vendor Inspection Status Report 3.

DATE REPORT PUBLtSHED

't-g woNin vtan Quarterly Report October 1990 April - June 1990 oln OR GR ANT NuMsE R

6. AUTHOR (S)

6. TYPE OF REPORT Quarterly

7. PL R100 COV E R E D isackseve pace.1 i

April - June 1990

8. PE R FORMING, omsre.AN12 ATION - N AME AND ADOR ESS (if NRC, pne Dwoma. Orta er Aseven, us Nwarar meeusere,y commason, sad meehas safrm ## cenerece ORG anne e+d meema aA I

Division of Reactor Inspection and Safeguards Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.

20555

9. SPONSORING ORG ANIZATION - NAME AND ADORESS ts# Nac type %me m ebove". sf coarramr.preekse Nac oween, otra er seven, ut Nuceer neouserery comedesma, sad madame seremJ i

Same as 8, above

10. SUPPLEMENT A8tY NOTES

11. A85T R ACT (200 wenn or mai 1

.This periodical covers the results of inspections perfornied by the NRC's Vendor Inspection Branch that have been distributed to the inspected organizations during the-period from April-1990 through June 1990.

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12. KE Y wCRoS/OEScRiPTORS re d..e,

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14. 34 GURI T y CLASSIF IC ATION Vendor Inspection tra,,see, Unclassified tra neoens Unclassified

16. NUMBER OF PAGES

16. PRICE

- NRC FORu 3M (2491 l

i UNITED STATES l

seec= nwwcuss urs

- NUCLEAR REriULAT4RY COMMISSl*N

• • ntg[85 Pao 9

WASHINGTON, D.C. 20566 j

OFFICIAL BU$lNESS PENALTY FOR PRIVATE USE,4300 t

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