ML20114C600

Licensee Contractor and Vendor Inspection Status Report. Quarterly Report,April - June 1992.(White Book)

ML20114C600
Person / Time
Issue date:	07/31/1992
From:	Office of Nuclear Reactor Regulation
To:
References
NUREG-0040, NUREG-0040-V16-N02, NUREG-40, NUREG-40-V16-N2, NUDOCS 9209020331
Download: ML20114C600 (129)

Text

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.NUREG-0040 Vol.16, No. 2 Licensee Contrac:or and Vendor Ins;pection S':atus Repor:

Quarterly Report April-June 1992 n

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

Single copies of this publication i

are available from National Technical information Service, Springfield, VA 22161 i

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N UllEG-0040 Vol.16, No. 2 i

9h6lSime4SA Licensee Contractor and Vendor Insaection Status Report Quarterly lleport April-June 1992 Manusript Completed: July IW2 Date l'ubbshed: July IW2 Division ofitenctor inspection and Safeguards Olilee of Nuclear Itenctor Ittgulation U.S. Nuclear Hegulatory Conunission Wasliington, DC 20555 8

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ADSTRACT This periodical covers the results of inc7ectionc performed by the NRC's Vendor Inspection Branch that have been distributed to the incpected organization during the period from April 1992 through June 1992.

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l TABLi' OF CollTEllTS PAGE 111 Abstract...................................................

vii Prefaco....................................................

Index......................................................

ix 1

Inspection Reports Selected Bulletins and Information Noticos Concerning Adequacy of Vendor Audits and Quality of Vendor Products 114 115 Correspondence Related To Vendor Issues V

l PREFACE A fundamental premise of the Nuclear Regulatory Commission (NRC) i licensing and inspection program is that licensees are responsible for the proper construction and safe and efficient 1

-operation of their nuclear power plants.

The total government-industry system for the inspection of commercial nuclear facilities has been designed to provide for multiple levels of inspection.and_ verification.

Licensees, contractors, and vendors i

cach participate in a quality verification process in compliance 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 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 ~1s 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 vnd 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 standards.

The Vendor Inspection Branch (VIB) reviews and inspects nuclear steam system suppliers (NSsss), architect'ongineering (AE) firms, suppliers of products and services, independent testing-laboratories performing equipment qualification tests, and holders of HRC 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.

Tho'VIB inspection 1 emphasis is-placed on the-quality and suitability of vendor products, licenser-vendor interface, 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 _in required to take appropriate i

corrective action and to institute: preventive measures to preclude _ recurrence.

When generic implications are identified, NRC assures that affected licensees are informed through vendor reporting or:by NRC generic correspondence such as information

notices and bulletins.

This-periodical (White Book) is published quarterly and contains copies of all vendor inspection reports issued during the calendar _ quarter for which it is. published.

Each vendor vil-

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

In addition, the affected Regional Offices

.are notified _of any significant problem areas that may require 3

special attention.

The White Book also contains a-list of selected bulletins and

'l Information notices involving vendor issues.

Copies of other pertinent correspondence involving vendor issues are also included in this White Book issue.

Correspondence with contractors and vendors relative to inspection data contained in the Phite~ Book is placed in the USNRC Public bocument Room, located in Washington, D.C.

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INDEX FACILITY REPORT NUMBEB PAGE

.ABB Power T&D Company Inc.

99901247/92-01 1

Allentown, Pennsylvania ITT Barton Fluid Technology Corp.

99900113/92-01 15 City of Industry,- California r

Power Systems Energy Services, Inc.

99901241/92-01 25

-Windsor, Connecticut Nebraska-Public Power District 05000298/92-201 37 Cooper Nuclear Station Columbus, Nebraska s

Westinghouse-Electric Corporation 99900404/92-03 73 Nuclear and Advanced Technology Division Pittsburgh, Pennsylvania Westinghouse Eloctric Company 99900104/92-01 82 Nuclear Components Division Pensacola, Florida' Yuasa..Exide. Incorporated 99901246/92-01 105

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3' UNITED STATES 3

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NUCLEAR REGULATORY COMMISSION

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WASHINGTON. D C, EUA o

r Docket No. 99901247 April 30, 1992 Mr. Gary Campbell, General Manager Protective Relay Division ABB Power T&D Company Inc.

7036 Snowdrift Road, Suite 2 Allentown, Pennsylvania 18106 Dear Mr. Campbellt

SUBJECT:

NOTICE OF NONCONFORMANCE (NRC INSPECTION REPORT 99901247/92-01)

This letter addresses the inspection of your facility at Allen-town, Pennsylvania, conducted by Mr.

R.

C. Wilscn of this office on April 6-9, 1992, and the discussion of his findings with mem-bers of your staff on April 9, 1992.

The purpose of the inspec-tion was to determine if safety-related electrical relays have been supplied by ABB Allentown in accordance with nuclear utility specifications.

The inspector reviewed your implementation of your quality assurance program for the design, manufacture, and testing of Class 1E relays for safety-related applications using commercial grade parts.

' Areas examined during the NRC inspection and our findings are discussed in the enclosed report.

This inspection consisted of an examination of procedures and records, interviews with personnel, and observations by the inspector.

The inspection identified that-the implementation of your QA program failed to meet certain U.S.

Nuclear Regulatory Commission (NRC) require-ments.

Specifically, on numerous occasions your procedures did not contain adequate instructions for activities affecting qual-ity because the procedures were outdated or were inapplicable.

In some cases appropriate procedures-did not exist.

The inade-

'quately controlled activities included receipt inspection of parts, final assembly, and (until February 1992) calibration and testing of Class 1E relays.

In addition, you did not perform independent reviews.of numerous activities affecting quality.

Specifically, from November 1988 until January-1991 you performed no independent engineering re-view of design changes; and until February 1992 you performed no independent check of calibration and testing of Class 1E relays.

Protective relays supplied by your facility are used extensively in commercial nuclear power plants licensed by the NRC.

Failures of these relays could significantly impact plant safety.

You have identified to us a plan to identify and correct deficiencies in your program, including underlying causes.

I trust that as 1

l Mr. Gary Campbell you carry out that plan, you will bear in mind your responsibil-ities under 10 CFR Part 21 to report any safety-related devia-tions that you discover that could affect previously-shipped relays, as well as current production.

After you have completed your upgraded program to meet the quality assurance requirements of Appendix B to 10 CFR Part 50, we will perform a followup inspection of your facility.

The specific findings and references to the pertinent require-ments for the above nonconformances are identified in the enclosed Notice of Nonconformance.

The response requested by the enclosed Notice is not subject to the clearance procedures of the Office of Management and Budget as required by the Paperwork Reduction Act of 1980, Public Law No.96-511.

In accordance with 10 CFR Part 2.790 of the NRC's " Rules of Practice," a copy of this letter and its enclosures will be placed in the NRC's Public Document Room.

Sincerely,

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, ir Leif 5.,fforrholm, Chief Vendor Inspection Branch

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Division of Reactor Inspection and Safeguards Office of Nuclear Reactor Regulation

Enclosures:

1.

Notice of Nonconformance 2.

Inspection Report 99901247/92-01 2

ENCLOSURE 1 NOTICE OF NONCONFORMANCE ABB Power T&D Company Inc.

Docket No.:

99901247/92-01 Allentown, Pennsylvania During a U.S.-Nuclear Regulatory Coumission (NRC) inspection con-ducted at the Protective Relay Division of ABB Power T&D Company Inc. in Allentown, Pennsylvania on April 6-9, 1992, the NRC l

inspector determined that certain activities were not conducted in-accordance with NRC requirements that were contractually imposed on ABB Allentown by purchase orders from NRC licensees.

The NRC has classified-these Atoms as nonconformances to the requirements of Title 10 of the Code of Federn) Reaulations, Part 50 (10 CFR Part 50), Appendix B.

A.-

Criterion V of Appendix B to 10 CFR Part 50, " Instructions, Procedures, and Drawings," requires in part that activities affecting quality shall be prescribed by, and accomplished in accordance with, documented instructions, procedures, or drawings.

Section 5 of ABB Allentown's Quality Assurance Manual Revision 5, dated Tobruary 3, 1992, states that activities affecting quality are defined by documented instructions,

-proceduree or drawings, and that quantitative and quali-

-tative criteria are used to determine satisfactory work performance and quality compliance.

Contrary to the above, on-numerous occasions ABB Allentown's procedures did not exist, were outdated, or were inappli-cable.

Specific examples include the following-1.

Quality Assurance Procedure No. TP-3001 Revision 1 l

dated December 5, 1978, did not specify the trip LED l

test of a type 51-Y relay, and the proper procedure for the relay tested was not at. hand end was stated to exist only in draft form.

2.

No-procedure or drawing existed for final assembly of a type 51-Y relay (92-01-02).

B.

Critorion'X of Appendix B to 10 CFR Part 50, " Inspection,"

requires in.part-that a program for' inspection of activities-affecting quality shall.be established and executed; such

. inspection shall be performed by individuals other than-those who performed the work being inspected.

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l Section 10 of ABB A11entown's Quality Assurance Manual Revision 6, dated February 3, 1992, states that where applicable and specified, inspections or tests are performed in accordance with the requirements of the appropriate specifications.

Purchase orders for Class 1E relays typically.specified the apn14.cability of Appendix B to 10 CPR Part 50, and ABB Allentown typically so certified.

Contrary to the above, on numerous occasions ABB Allentown did not perform independent reviews of activities affecting quality.

Specific examples include the following.

1.

From November 1988 until January 1991, ABB Allentown conducted no independent review of design changes.

2.

Until February 1992, ABB Allentown performed no independent check of calibration and testing of Class 1E relays (92-01-02).

Please provide a written statement or explanation to the U.

S.

Nuclear Regulatory Commission, ATTN:

Document Control Desk, Washington, D.C.

20555 with a copy to the Chief, Vendor Inspec-tion Branch, division of Reactor Inspection and Safeguards, Office of Nuclear Reactor Regulation, within 30 days of the date of the letter transmitting this Notice of Nonconformance.

This reply should be-clearly marked as a " Reply to a Notice of Noncon-formance" and.shuuld include for each nonconformance:

(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.

Dated at.3ockville, M ryland this

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day of //

1992.

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ORGANIZATION:

PROTECTIVE RELAY DIVISION ABB POWER T&D COMPANY INC.

ALLENTOWN, PENNSYLVANIA REPORT NO.:

99901247/92-01 CORRESPONDENCE Mr. Gary Campbell, General Manager ADDRESW Protective Relay Division ABB Power T&D Company Inc.

Allentown, Pennsylvania ?.8106 ORGANIZATIONAL liarry A. Hinch, Quality Assurance Manager CONTACT:

215/393-7333 NUCLEAR INDUSTRY.

Manufacture and supply of Class 1E protective ACTIVITY:

relays for safety-related applicat. ions INSPECTION.

April 6-9, 1992 CONDUCTED:

SIGNED:

M 4-O.92._.

Richard C.

Wilson, Senior Engineer D6te Reactive Inspection Section No. 2 Vendor Inspection Branch (VIB)

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APPROVED:

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'YL 9A Jef frey/B. Jacobson, Acting Chief Date Reactiv CInspection Section No. 2 Vendor Inspection Branch INSPECTTON BASES:

10 CFR Part 21 and 10 CFR Part 50, Appendix B INSPECTION SCOPE:

To review the design, manufacturing,.and testing activities conducted under ABB

- Allent own's quality assurance program.

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-PLANT SITE.

Numerous.

APPLICABILITY:

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1 INSPECTION

SUMMARY

1.1 NoncqnfrrJn.cnged 1.1.1 Contrary to Criterion V,

" Instructions, Procedures, and Drawings," of 10 CFR Part 50, Appendix B, ABB Allentown's procedures for controlling numerous activities atfccting quality were nonexistent or outdated (Nonconformance 92-01-01, seo Section 3.13 of this report).

1.1.2 Contrary to Criterion X,

" Inspection," of 10 CFR Part 50, i

Appendix B, on numerous occasions ABB Allentown performed no independent review of activities affecting quality.

(14 3ncon-formance 92-01-02, see Section 3.13 of this report).

2 STATUS OF PREVIOUS INSPECTION FIllDINGS There were no previous llRC inspections of this facility.

3 INSPECTION FINDINGS AND OTilER COMMENTS 3.1 Entranco a_ntilgit MeetiDas

.Tn the entrance meeting on April 6,

1992, the NRC inspector dis-cussed the scope of the inspection, outlined areas of concern, and established interfaces with ABB Allentown's management and staff.

In the cxit meeting on April 9, 1992, the inspector dis-cussed the findings and concerns with ABB Allentown's management and staff.

3.2 lugpgetion Scong L

The ABB Power T&D Company Inc., Protective Relay Division, Allen-town, Pennsylvania location (ABB Allentown) manufactures several types of protective relays, including ground fault, overcurrent, voltage, and direct current.

The location occupies 34,000 square feet and has about 90 employees.

Annual sales are about $17 million, of which 3 to 6 percent is for commercial nuclear power plants.

Partial customer lists provided by ABB Allentown show 47 commercial nuclear plant sites for which Class 1E relays were purchased from ABB Allentown, either directly by the licensee or through a third party.

Thc headquarters of ABB Power T&D Company Inc. is located in Blue Bell, Pennsylvania.

The Relay Division headquarters is in Coral Springs, Florida, and the division has a third manufacturing facility in Puerto Rico.

Allentown's relays are often sold to customers by the ABB Power Distribution Company of Sanford, Flo-rida or other corporate divisions which incorporate the relays in higher level switchgear assemblies such as motor control centers.

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The solid state relays produced at ABB Allentown were developed in thn carly 1960s by the I-T-E (for inverse time element)

Circuit Breaker Company.

Because of mergers, reorganizations, acquisitions, and joint ventures, the relays have been supplied

-undex several corporate names.

These include I.T.E.

Imperial Corporation; Gould Incorporated; Gould-Brown Boveri; Brown Boveri Electric, Inc.;-BBC Brown Boveri, Inc.;-ASEA Brown Boveri, Inc.;

Westinghouse ABB Power T&D Company; and ABB Power T&D Company._

Facilities were moved from Philadelphia to Horsham, Pennsylvania in 1974 and to Allentown in 1984.

The senior development engineer stated that the only significant design change made since seismic qualification testing was performed in the early 1980s was a packaging modification to facilitate testing by Customers.

The NRC inspector reviewed the procurement, rocciving, quali-fication, manufacturing, testing, and inspection activities conducted under ABB Allentown's quality assurance program pursuant to the requirements of Appendix B to 10 CFR Part 50.

The NRC inspector also reviewed four recent audits by customers-of ABB Allentown, and ABB's progress in resolving the findings of

'those audits.

The NRC inspector also reviewed ABB Allentown's program for conforming to the requirements of 10 CFR Part 21.

3.3 AllditD of ABB Allentown The NRC' inspector reviewed four recent audits of ABB Allentown:

(1) ABB Power Distribution Company (Sanford, Florida), October 22-23, 1991; (2) Quality Systems Incorporated (QSI) for three licensees, November 11-12, 1991; (3) Westinghouse Electric Corporation,-Nuclear Systems Division, February 12-14, 1992; and (4) Tennessee Valley Authority (TVA), February 26-27, 1992.

'After receiving ABB Allentown's response to their audit findings, QSI notified the NRC of_their concern that ABB Allentown may have been shipping commercial grade relays identified as safety-grade.

l These four audits had a total of 46 findings, some multiple-part and with some overlap.

About half of the findings were identi-l fled in the ABB Sanford audit.. At the time of the.NRC inspec-tion, the status of these audits was as'follows:

ABB'Allentown had responded to all findings of the ABB Sanford audit, and a followup. audit was expected later in April 1992.

ABB Allentown

'had responded to_the QSI audit,.QSI had conducted two followup visits, and-fiveitindings remained open.

ABB Allentown's re-l

'sponse to the-Westinghouse audit was:in preparation, and expected to be:one-weekflate. -ABB-Allentown had-not yet received the TVA audit report.

No~ work had boon-performed on componente for.

Westinghouse at the time of their inspection; the other three auditors all placed holds on shipments of safety-related orders.

Several of the audit findings. seriously questioned the implemen-tation of~ABB Allentown'a Appendix B QA program.

No independent 3

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review of design changes was performed from November 1988 to l

January 1993; during this time the same person served as design j

engineer, engineering manager, and QA manager.

No QA review was performed on seven purchase orders received from March 1991 to March 1992.

Auditors found numerous instances of procedures not conforming to actual practice.

Instances were found of improper distribution of controlled documents, use of unapproved proco-dures, activities affecting quality performed without written procedures, a superseded drawing revision in the shop, inadequate I

record storage, and out-nf-calibration test equipment.

Suppliers of calibration services were not audited; when ABB Allentown performed an audit in response to the finding, the ABB auditor was unqualified (in this case ABB qualified the auditor and repeated the audit).

ABB Allentown had initiated corrective action for most of these findings.

Independent review of all release forms for Class 1E equipment had been completed to identify those requiring further review. Most of the specific details addressed in the findings had been corrected, and several responses stated that procedures were out of dato and would be corrected.

3.4-EyAluation of-Supoliers Section 4, Procurement Document Control System,- of ABB Allen-l town's QA Manual states that all vendors and their products are l

cvaluated by one of the specified methods.

Since the suppliers y

for all parts and materials were commercial grade, the NRC l

inspector assessed the supplier evaluations as dedication I

activities.

(Suppliers of calibration services are addressed separately below.)

ABB Allentown's acceptance of suppliers was based on a combina-l tion of product evaluation and vendor / product history.

Product evaluations were conducted by engineering.

To assess the evaluations performed, the NRC inspector selected several types of parts used in Class 1E relays. A documented evaluation was found for only one of the six cases, an alternate supplier of a transistor.

The evaluation consisted of reviewing published catalog data and testing five samples on a curve tracer.

This evaluation verified that the nominal performance of the alternate supplier's transistor satisfied ABB Allentown's design require-ments, which were based on the published performance data for the original supplier.

The development engineer stated that this evaluation was typical, and that suppliers were not visited as part of engineering evaluations.

No evidence was found of QA audite of parts suppliers.

With respect-to suppliers of_ calibration services, the NRC

. inspector-noted chat two of the customer audits found that ABB Allentown had not audited such supp.'.iers.

In response to these findings, the ABB Allentown QA manager audited one of the i

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same data sheet and an-unnumbered procedure titled Overcurrent Relay Test Procedure dated November 19, 1991, the tester then verified one value each for the time curve and instantaneous trip checks performou during second stage calibration and testing.

i The NRC inspector noted that there was no procedure or drawing for the final assembly operation.

Each of the two line testers review.ponsible for enecking her own work, with no. independent was res Although tha second stage and final testing operations were performed with different sets of tast equipment, there was no requirement against the same person performing both tests.

Thus, the same'purson could perform calibration adjustments, final assembly operations, and final testing with no ir. dependent checking by anyone else.

No additional testing of Class 1E relays was-performed until February, 1992.

The NRC inspector was particularly concerned about the lack of procedural control on final assembly and the absence of inde-pendent checks on the calibration and testing because the relays were built from commercial-grade parts (sco section 3.5 of this report).

The additional QA conformance testing initiated in February 1992 provides a basis for safety-grade dedication of commercial grade relava by testing and experience.

Relays shipped before February 1992 did not roccive the benefit of the

-improved testing..

Section 3.12 of this report identifies ABB Allentown's intention to implement a corrective action program.

Because of the large number of identified discrepancies and the imminent corrective f

action program,-the NRC inspector did not attempt to assess the overall acceptability of ABB Alluntown's corrective actions ~at the time of the_inspectien.

However, the simple accomplishment of seyeral performance measurements on the completed relay, by an

~1ndapandent person using different test equipment, is a major improvement in upgrading the-rolays toward safety-grade status.

3.7 Desian Control The NRC_ inspector reviewed ABB's in-progress activity to review design changes that had not received an independent design review from November 1988 to January 1991.- ABB Allentown had initiated L

this activity as a result of customer _ audit findings.

Two former engineering managers are performing the review in accordance with-

Engineering Standard Practice 502,-which specifies criteria for determining whether a change requiresEfurther engineering eval-untion.7 ABB"Allentown had completed review of the Class 1E j

release forms,,and other documents such as aging stress =calcu-lations were being reviewed.- The development engineer stated

'that the last-significant-design change occurred in 1986-1987, when-the Circuit Shield'rxlay design was changed to a test case configuration;.that design was then seismically tested in 1988.

Many of the-drawing revisions involved corporate name changes and 7

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other non-technical matters.

No class 1E relay will be shipped until the review of related design changes is completed.

The NRC inspector concluded that the design change review appeared to be adequate, pending verification of its adequacy after completion.

3.8 Oualification Rxpqr_ta_p 1 Certifications The NRC inspector reviewed qualification documentation for a type 27N relay shipped to the Waterford 3 Nuclear plant under ABB Sales Ordcr No. 27755 dated November 19, 1991.

Seismic testing, stress analysis, and certifications provided by ABB Allentown appeared to be generically consistent.

The NRC inspector concluded that the generic base for supplying safety-grade. relays _ appeared to be acceptable, provided that the quality of the specific relay shipped and the traceability of its design to the qualification tests andianalyses performed from l

1981-to 1783-are-shown to be adequate.

3.9 S2qrcoation of Reiqqtod Parts The-NRC inspector observed storage of rejected material in the receiving area.

The material was segregated only in that it was placed on the last of a series of open shelf assemblies (the other shelves did not contain rejected material).

The segregated rrca was not posted with signs, and no consistent tagging system a

was evident.

Different cartons had.various combinations of pink, white, blue, and yellow paper attached.

Tha~ incoming inspector stated that the form previously marked to identify rejected-materials was no longer in use, and r,s new procedure had been i

issued.

He stated that he had improvised one tagging system, and his partner another.

The NRC inspector concluded that procedural control of rejected material was not in place.

ABB Allentown personnel indicated that immediate ccurective action would be initiated.

3.10 Other QA MaDMAL.Cpncerna The ABB QA manager stated that as a result of the inspection, he l

would make some clarific*+'ons in the QA Manual. Section 4.1 implies that vendor evalu cions are' performed consistent with the

-requirements'of Appendix B to 10 CFR Part 50,: which is inappro-priate because neither the vendors nor the evaluations are at that level.

Section 10_ states that when modifications or repairs are performed, affected characteristics are retested.

Complete retesting was11ntended, and the manual will ha revised to so state.

Section 12.7-states that if test equipment is consis-tently found-to be out of calibration, it it tagged and removed-froml service..The development engineer told the NRC inspector 8

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u that this statement was written when analog test equipment subject to drift was ueed (the manual was originally issued in 1981), and that out-of-calibration test equipment in no longer Common.

3.11 10 CFR Part 21 Activities The NRC inspector reviewed ABB Allentown's procedures for con-forming to the requirements of 10 CFR Part 21, which provided l

instructions and requirements for identification, control, and documentation of nonconformances.

Appropriate notification was posted.

The only part 21 notification made by ABB Allentown recently was dated October 34, 1991.

It ocncerned type 59 and related relays that were manufactured from 1975 to 1982.

The notification to the NRC identified the basis for the concern, recommended re-placement of all affected relays with catalog number 211E1175 and specific named printed circuit board numbers, and listed affected nuclear-customers.

As followup action, ABB Allentown telephoned the NRC ar.d then directly notified the af fected commercial nuclear power plants in October 1991.

The inspector concluded that ABB Allentown was adequately addressing the notification requirements of 10 CFR part 21, and that the corrective actions taken for the only notification provided to date were satisfactory.

The inspector had no concerns in this area.

3.12 Cgirective-Actions During the inspection, ABB Allentown personnel indicated that they would prepare a corrective action plan that would ensure their capability to supply Class 1E relays.

On April 15, 1992, ABB Allentown faxed to the NRC-a "Contrcl Document to Improve Implementation of Appendix B GA Progra!."

This document describes a corrective action plan, and contains appropriate approval signatures including the general manager's.

The NRC i

inspector particularly noted that the plan includes activities to identify program'"clements which satisfy Appendix B or follow other methods," and,"to review results of recent audits to identify underlying causes."

The control document also. outlines an interim method of producing safety-related products until the improved program is implemented.

3.13 Epnclusions The NRC inspector concluded that ABB Allentown's program for supplying Class 1E relays failed to meet the requirements of

. Appendix B of 10'CFR part 50.

Many of the deficiencies were identified by customers and.were already being addressed by ABB

'Allentown.

The NRC inspector found additional deficiencies that 9

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1 are included in the scope of ABB Allentown's corrective action plan.

None of the identified deficiencies necessarily affected the hardware product, but the c,tmulative effect of the programmatic and documentation deficiencies is significant.

Most of the deficiencies identified during this inspection can be grouped into failure to comply with two criteria of Appendix B to 10 CFR Part 50.

Criterion V,

" Instructions, Procedures, and Drawings" requires in part that activities affecting quality shall be prescribed by and accomplished in accordance with documented instructions, procedures, or drawings.

The numerous instances of ABB Allentown's f ailure tr procedurally control activities affecting quality constitute Nonconformance 92-01-01.

Criterion X,

" Inspection," requires in part that activities

~

affecting quality shall be inspected by individuals other than tnose who performed the work beir.g inspected.

The significant examples of ABB Allentown's failure to independently review activities affecting quality constituto Nonconformance 92-01-02.

A future NRC inspection will review ABB Allentown's implementa-tion of its improved QA program.

4 PERSONNEL CONTACTED AM:

L.

R. Volkel, Manager Total Quality H.

A.

Hinch, Quality Assurance Manager J.

Martin, Manager, Operations H.

Parke, Controller V.

L. Morris, Manager, Purchasing C.

Tomlinson, Manager, Marketing R.

M. Garrett, Manager, Engineering R.

Conrad, Qualif. and Sr. Development Engineer C.

A.

Berger, Line Tester C.

Schildt, Line Tester G.

E.

butr, Incoming Inspector C.

Downs, Product Manager, Marketing W.

H. Wallace III, Manager, Total Quality, ABB Relay Division, Coral Springs, Florida HRC:

A.

S.

Gautam, Acting Section Chief, VIB/DIR/NRR Attended the exit meeting on April 9, 1992 10 14 l

1

OS 40%

' 7.

UNITED STATES

.I NUCLEAR REGULATORY COMMISSION "o

I WAshtNutoN, o c. 20zs s.,...../

Mey 4, 1992 Docket No. 99900113/92-01 Mr. Roger F. Murphy General Manager ITl Barton Fluid Technology Corport' ion 900 South Turnbull Canyon Road City of Industry, California 91749-1082

SUBJECT:

NRC INSPEC(10N REPORT NO. 99900113/92-01 This letter addresses the inspection of your facility at City of Industry, California, conducted by Mr. K. R. haidu and Mr. R. K. Frahm, Jr. of this office on April 14-16, 1992, and the discussions of their findings with you and your staff at the conclusion of the inspection <

The purpose of the inspection was to review the implementation of the ITT Cart.m (Bartun) quality assurance program in selected areas during the n>anufacture of transmitters and related products.

The inspactors also reviewed Barton's corrective actions for three 10 CFR Part 21 reports and for a recent problem you identified regarding a transmitter supplied to the Big Rock Point nuclear power station.

Arcu examined during the NRC inspection tnd our findings are discussed in tne enclosed report.

This inspection consisted of an examination of precedures and representative records, interviews with personnel, and observations by the inspectors.

No violations, nonconformances, or unresolved items were identified during this inspection.

In accordance with 10 CFR Part 2.790 of the NRC's " Rules of Practice," a copy of this letter and its enclosure will be placed in the NRC's Public Document Room.

Sincerely, 7

i

[

J,r

' d A j' O y

b J

i Leif J. Norrh'olm, Chief Vendor inspection Branch Division of Reactor Inspection and Safeguards Office of Nuclear Reactor Regulation

Enclosure:

Inspection Report No. 99900113/92-01 15

OIEANIZATIQ1:

ITT Barton Fluid 'lbchrology Corinration City of Intetry, California RUCRP 10:

99900113/92-01 CDimESIOt3DD4CE Mr. G. R. Welt, Director ADDRESS:

Qaality AstAuarce ITT Barton Fluid Tbchnology Corporation U

900 South 'hirnball Canyon Road City of Irdustry, Calliornia 91749-1882 3

OIGAllIZATIOtE Mr. Gerald R. Welt, Qaality Assurance Itunger 001TTAC1' NUCLEAR DIDUSTRY Precsure switches, pressure transmitters, ACTIVITY:

analog trip system ard valve actuators IllSPirrIOtt April 14-16, 1992 00tIDUCTID:

q f

/ )'

b: c um g/q hz E5malakar R. Nakhl, 'Ibam Leader

'Date spccial Projects section Verdor Inspection Branch (VIB) 4 Roruld F.. Prahm, Jr. VID M7 5 Y APPIOVID BY:

_Gregorf q7Cwalira, Odef Dite Special Projects Section Verdor 1rupection Branch INSPECT 104 BASIS:

10 CFR Part 21 ard 10 CFR Part 50, Apperdix B INSPIrrIOtf SCOPE:

Paview corrective action taken on previously reported 10 CFR Part 21 itm3; observe the irplenentation of the quality assuranco progran durirg the fabrication ard testing of pressure transmitters ard pressure switdrs.

PLAlfr SITE APPLICABILITY:

Numerous pressurizai Water reactors ard boiling water reactors.

16

i 1

-INSPECTION

SUMMARY

i The inspectors identified no violations, noncanformances or unresolved items during this inspection.

l l

2 STATUS OF PREVIOUS INSPECTION FINDINGS There were no outstanding findings from previous inspections, s

3 INSPECTION FINDINGS AND OTHt.R COMMENTS 3.1 Entrance and Exit Meetinas During the entrance meeting on April 14, 1992, the inspectors informed the ITT Barton (Bartor.j staff of the scope of the inspection, outlined areas of concern, and determined which persons on the Barton staff to contact during the: inspection.

During the exit: meeting on April 16, 1992, the NRL inspectors summarized the results of the inspection for Barton management.

3.2 Review of 10 CFR Pa-t 21 Proar_aJD The inspectors reviewed Barton's Part 21 program to determine compliance with the requirements of Part 21 of Title 10 of the Code of Federal Regulations, (10 CTR Part 21) for reporting of defects and noncompliance. The inspectors reviewed Standard Practico GE 12-3, "NRC Regulations to 10 CFR Part 21," of March 1 1987, and QU 25-1, " Product safety Program," of January 15, 1992.

The inspectors also observed the pxting of 10 CFR Part 21 Section 206 of the Energy Reorganization Act of 1974, and Barton's Standard Practice GE 12-3.

The inspectors concluded that Barton had not incorporated the July 31, 1991, revision to-10 CFR Part 21 into its program.. This revision was published in the-Federal Reaister and became effective on October 29, 1991. Barton stated

.that.-it knew that Part-21 had been changed and-ordered the latest revision to Title 10 of the Code of federal Regulations.

Barton received the 1991 Jrevision to Title.10 e i

'$ does not include the July 1991 change.

The inspectors provided Bai with the July 1991 revision which it_immediately posted and Barton has begun to revise its Part 21 procedure to refle.t the latest revision.

The inspectors noted-that Barton did try to obtain the

' latest revision to Part.21 and corporate practice h_ad been changed -to reflect the revision.

(.

L The' inspectors were also concerned that procedure GE 12-3 may be misinterpreted by Barton employees or consultants.

Section D, entitled H

" Reporting,".of this procedure includes the statement that "any Barton L

employee who is: aware of a de.f1qi in a Barton product to which 10 CFR Part 21 applies, shall immediately and personally assure that the General Manager and Director:of Quality Assurance are notif ei d of the defect." This statement could be interpreted as requiring the Barton employees to perf orm an. initial evaluation of a deviation te determine-if it could result in a substantial:

o p

-safety hazard.

Thus, this statement may dissuade the employee from reporting i

~17

..-,_ _.m.

. =.

l l

possible pioblems to management.

Barton employees should only be responsible for reporting any deviations or problems to management, which is responsible for determining the safety significance o' the possible defect.

The inspectors discussed these concerns with the Barton management during the inspection.

Barton has committed to revising its Part 21 procedure to address the NRC's concerns.

3.3 _R_eyiew of Barton's 10lFR Parl_jlLReportl 11T Barton had recently submitted five Part 2] reports to the NRC.

Three were reports discussing potential safety hazard issues, and the ther two were interim followup reports.

The three issues are discussed below.

The first Issue concerned the possibility of misunderstanding information in the marketing literature on the performance of Model 763 transmitters. Model 763 is a zero-suppressed transmitter which was redesigned in 1984 to correct a series of reported defects.

Since 1984, both the original and the new (Model 763A) designs were available to the nuclear industry.

The product specifications did not clearly compare the long term otift between the two transmitters.

The literature states that the long term drift for the Model 763 is 5.0 percent of the maximum span each year and for the Model 763A is 1.0 aercent of the maximum span per year.

These reported drift errors apply to suppressed range applications only, both the 763 and the 763A have a 1.0 perccat drif t for zero-range applications.

Barton has committed to reviewing and revising all applicable marketing literature and issuing a general information letter to all users by June 29, 1992, which will include the performance specification differences.

The second issue concerned a possible defect in the Model 764 and 765 differential pressure transmitters and the Model 763 and 763A static pressure transmitters. The lead wire insulation at the point of exit from the transmitter housings had become damaged and could lead to short circuiting the output of the transmitter.

After invrstigating this issue, Barton determined that the tool used to insert the harness assembly into the housing 9as poorly designed and could cut through the insulation to the conductor.

Barton improved the tool design as evidenced by draw ng number 0764-1174-B, " Seal Connector Tool Ass'y,"

Revision 05 of January 9, 1992.

Barton also inproved the harness assembly design by adding protective sleeves to the lead wires at the point of entry as evidenced by drawing number 0764-1221-B, " Connector Assy," Revision 04 of November 4, 1991. On November 7, 1991, Barton revised its process sheet for this connector assembly to instruct the assembler to add the sleeving to correct work in process. Barton sent copies of the notification and recommended instructions to all customers, advising them to modify the affected transmitters or return them to Barton for modification.

The third issue concerned an unrelated failure in the Model 763 and 763A static pressure transmitters. The Kansas Gas and Electric Company, (the licensee for the Wolf Creek Generating Station) reported a short circuit between a connector pin and an adjacent mounting screw.

Barton's common 18

practice was to clip the connector pin during the manufacturing process.

However, this requirement was not clearly documented, and apparently was not followed consistently.

To correct the problem, Barton engineers rcoesigned the connector assembly on which the pin is mounted to move the location of the pin further from the

. adjacent mounting screw as documented-on drawing 0763-1028B, " Strain Gage Connector Assembly," Revision 004 of April 2, 1992.

Barton sent copies of the notification and recommended instructions to all customers, advising them i

to. inspect the clearanco between the connector pin and screw head and modify the affected transmitters if the clearance is not at least 0.01 inch.

On November 21, 1991, Barton revised its process sheet for this connector assembly to instruct the assembler to cut the appropriate terminal pin.

Barton eliminated its existing stock of transmitters to ensure that all new 1

transmitters have the correct clearance.

3.4 Review of a Problem Exogrienced at _the Bia Rock PoinLHuclear Pqwer Elant The Big Rock Point nuclear plant uses two Barton 764/351 transmitters to monitor the primary containment water level during normal and accident conditions.

Under loss-of-coolant accident conditions, these transmitters provide containment water level indication, from which the operator transfers core spray to recirculation mode.

On March 23, 1992, the Consumers Power Company, the licensee for the Big Rock

. Point nuclear plant (Big Rock Point) issued a licensee event report stating that Barton had notified-it that Barton had returned one of its Model 764/351 transmitters with the sensing lines filled with water instead of silicone.

The licensee had-installed the transmitter from January 12, 1990, to June'28, 1991.

The Barton Model 764/351 transmitter consists of a Model 764 transmitter connected to a Model 351 remote sensor by means of.a capillary tube.

Before June 1982,.Barton had used distilled water as the medium in the capillary tubes to transmit the pressure variations from the sensor to the

transmitter. After being alerted in_ June 1982 to the possibi_lity that distilled water may flath to steam under high-temperature accident conditions Barton began using DC 702 type silicone gel instead of distilled water in the capillaryTubes of level indicating transmitters installed inside the ontainment.

c l:

The NRC inspectors reviewed the documentation available at Barton on the

~

764/351 transtaitters including the purchase orders (P0s) Consumers Power had issued to~.Barton.to procure and repair Model 764/351 transmitters, Barton's

" Register," which is the internal-tracking-sheet that specifies the material, andcinternal instructions to either manufacture-or repair the transmitters.

The. inspectors determined that the licensee purchased five Barton Model

-764/351 transmitters to moni_ tor the containment water level.

Two (serial numbers _1643 and 3164) of--these were purchased with distilled water as the i

fill fluid in the capillary tubes.

At Censumers Power's request, Barton replaced the fluid.with DC 70? type silicone fluid.

From_1984 to 1991, operators at Big Rock Point determined that the transmitters responded slowly

~ to level variations,- and operators experienced difficulty in calibrating the

19

.- - ~

transmitters at the lower end of the range.

The operators cut the capillary tubes and returned only the transmitters to Barton with purchase orders to repair and recalibrate the transmitters according to the original purchase order.

Consumers Power's purchase orders to Barton were adequately identified with a suffix "Q" to indicate that the equipment being serviced was safety-related, and required compliance with its quality requirements.

The purchase orders included statements that 10 CfR Part 21 was applicable.

The purchase orders frequently included the statement " repair and return."

Except in two instances, Barton personnel were able to research the original Pos and the subsequent changes and fill the capillary tubes with DC 702 type silicone fluid even though the technical requirements in 9e purchase order were not

specific, in the other two instances, Barton fulfilled the PO requirements by repairing the transmitters in accordance with the original requirement, i.e.

distilled water in the capillary tube. To preclude reoccurrence, Barton informed the inspectors that it had taken corrective action to revise its procedure for handling repair orders.

Barton will process future repair orders from nuclear power plants through quality assurance representatives who will review the requirements of both the original purchase and subsequent revisions.

The Barton-representative must also contact the customer and obtain verbal confirmation of the problem and the precise repair requirement.

The inspectors reviewed the propu:,ed Barton action and determined it responsive.

l 3.5 Problems Experienced With_Barton level Transmitters The Barton staff inforned the inspectors that other plants had experienced problems with level inoicating transmitters installed inside the containment.

These level indicators consist of Barton Model 764 transmitters and Model 350 series sensors (also referred to collectively as Model SO48).

The sensor transmit's the pressure to the transmitter through a capillary tube filled with a flui_d (DC 702 type silicone). The ends of the capillary tubes are welded to the transmitter and the sensor.

The problems in level transmitters with remote sensing devices appear as a gradual upward shift in the transmitter output without a corresponding shift in the measured water level-and appear as a difficulty in calibrating the transmitter in the lower range. The problems are caused by gas (oxygen and nitrogen) entrapped in the capillary tube.

One of the characteristics of DC 702 type silicone fluid is its affinity to absorb oxygen and nitrogen.

Barton has-not been abl.e to conclusively determine the manner in which the gases accumulat.e in the level system; but postulates two possible ways.

If the gases are not completely removed from fill fluid during filling operations, any entrained gases within the fluid will migrate to the highest point in the housing and cause the transmitter to respond slowly.

Air can also seep into-the transmitter center block and housing over time because 6f the compounded effects _of the sensing leg being under negative pressure and the affinity of the-DC 702 type siliconc.for absorbing oxygen and nitrogen.

These problems appear to depend on the specific fill procedure and the installation configuration at the plant.

The time to accumulate air appears to depend on the vertical separation between the transmitter and the sensor. 20

- _ =. - -. - - -

i Barton recommends the following actions to avoid these problems in level indicsting systems:

Minimize the height differential between the transmitter and the lower j

sensor.

Minimize the length of the capillary tubing between the transmitter and I

the sensor with predominantly vertical runs and as few bends as

possible, Check and vent the level system or refill the system as necessary during l

refueling outages.

Barton informed the inspectors that its service personnel have successfully bled entrapped air from level transmitters and had re-calibrated and restored them to operability at some plants.

Barton has prepared an inuustry informatis.1 notice on this subject for general distribution.

The inspectors reviewed the industry information_ notice and determined that it contained the necessary details for licensees to take adequate corrective action to preclude problems.

3.6 Review of Mrton's Control of Materials Dyrina Manufactur_ing

't The inspectors reviewed the list of-materials used in assembling a Model

~

M 288A pressure switch which was being supplied to the Laguna Verde Nuclear Power Plant in Mexico and selected various components to examine.

The inspectors verified the control-of these materials from the quality assurance documents and determined the following:

Barton purchase orders to the vendors specified the apprcpriate part number which is traceable to a drawing, the applicable quality assurance crittria, and_ military standard specifications-(Mil-STDs) where applicable.

Vendors furnished-certified material test reports (CMIRs) with details of the physical-and chemical properties of the metallic products supplied.-.The CMTRs provided heat numbers to establish traceability of the material;_ as appropriate, the CMTRs included additional details on L

topics such as elongation tests, tensile strengths before and after age L

hardening, and heat treatsent.- Barton independently tested coupons from l

all the materials supplied' to verify _the validity of the CMTRs.

u Parker 0-Rings, Kentucky, supplied-ethylene propylene terpolymer (EPT)

' type 0-rings of various sizes and.provided certificates of compliance (C0C) for the 0-rings.

Information-on the C0C included designation of the material, its cure -date, shelf life, and compliance to the L

applicable MIL-STD and American Society for Testing Materials (ASTM) tests. ;

21

e Parker Packing, Utah, supplied Fluro Elastomer (Viton) +vpe backup rings of various sizes and provided COC with information on the designation of the material, its cure date, shelf life, and compliance to the app!Icable MIL-STD and ASTM tests.

Barton purchased cable from two vendors. The cable includes insulation certified to DuPont Tefzel 280 which has been proven to be radiation-resistant during environmental qualification tests.

Barton purchases most of the components as commercial grade items and dedicates them during receipt inspections using attributes specified by engineering and quality assurance personnel. The inspectors reviewed the attributes to dedicate various items procured as commercial grade and determined them acceptable.

Barton supplemented these attributes by selectively sending samples of the procured items to either internal or outside laboratories for independent tests to verify the validity of the certification.

For example, Barton performed swell tests on 0-rings and sent copper-nickel bars to-an independent laboratory for analyses.

The inspectors observed that Barton's receiving inspectors had completed rejection reports when they observed components that did not meet the established acceptance criteria and fc warded the reports to the engineering department for evaluation and disposition. Components determined to be unacceptable by engineering were returned to the vendor.

The inspectors observed the following:

Rojection Report (RR) 37251 of March 11, 1992, in which Barton noted that the' tensile strengths of several samples of beryllium copper tubes (for Bourdon Tubes) did not meet the specified values before being age hardened. However, the tensile strengths after age hardening were acceptable. A Material Review Board (MRB) consisting of representatives from various departments including engineering _and quality assurance

. determined to acce,nt the tubes as-is because the tensile strengths of

.the tubes'after age hardening were acceptable.

l Barton completed RRs 35948 and 35973 on March 7 and 13, 1991, l

l respectively, to document that ITW type snap-action switches did not meet the specified acceptance criteria. The MRB determined to return the switches to the vendor.

Receipt inspectors observed that they received 10-terminal Kulka type terminal blocks and verified that these terminal blocks met the dimensions specified by the manufacturer.

However, the receipt inspectors observed that the terminal blocks did not meet the Barton drawing 0038-0033T for the part because the dimensions shown were for a 5-terminal block.

On January 28, 1992, Barton completed Document Change Request 24645 to revise the dimensions.

The inspectors determined that Barton's control of commercially-procured items was adequate and-that Barton had taken precaution to prevent the receipt of unacceptable material.

_7 s

22

3.7 Review of Barton's Audits of its Vendors The inspectors-reviewed the audits performed by Barton of the Parker Packing Company, Salt Lake City, Utah, and of Parker 0-Ring.

Barton's quality L

assurance (QA) representatives performed the audits using checklists reflecting appropriato criteria of hIl-1-45208 and HIL-STD-45662.

The

' auditors docomented unacceptable observations as identified findings; QA performed followup audits to verify that the findings were corrected.

The inspectors determined that Barton is exercising adequate controls on its vendors.

3.8 Review of Quality Assutanst Audits Performed on Barton The inspectors reviewed the most recent Nuclear Utilities' Procurement inspection Committee (NUPIC) Jcint Audit report of January 31, 1991, which was conducted by the Baltimore Gas and Electric Company (BG&E).

The audit results sufficiently indicated that' Barton's Quality Assurance (QA) Hanual meets the requirements of Appendix B to.10 CFR Part 50 and that it has adequately implemented its QA program. The audit produced no new findings, and closed the. remaining open items frors the previous. audit.

The inspectors 'also reviewed the most recent internal management audi report of December 17,(1990.- 'While reviewing-this-audit report, the inspectors noted four discrepancies that required a response from Barton.-Barton responded to Lthe findings outlining actions to correct the present problems and preclude repeating them in the future.

On January 8,1991, NUPIC accepted Barton's response and concluded that its QA program was deemed to be properly implemented and-in ~accordance with its latest QA manual.

The inspectore determined that external.and internal audits were conducted and that actions had been taken to correct adverse findings._

g L

ll i

ol'g 23

o 4-PERSONNEL CONYACTED D. Bell, Electronics Supervisor J. P._ Doyan, General Manager, Barton Sales J. Dwyer, Manager, Quality Assurance Programs

+

+

M. Henderson, Manager of Materials T. Hnitedge, Quality Products Manager J. Incontri, Product Marketing Manager

+ '*

R. Krechmery, Director of Engineering

+

• J. Meyer, Manager, Nuclear Engineering

+

R. F. Murphy, General Manager

+

• D. Norman, Quality Datumentatior Enecialist

~

+

G. Ricci, Manager, Marketing 'jsta:-

J. Schwable, inspect r T. Tran, 'achni-ian

+

G. R. Welt, u;.ector of Quality Assurance M. W. Williams, Manager of Manufacturing

+

~

4 Attended Entrance Meeting on April 14, 1992 Attended Exit Meeting on April 16, 1992

_g_

24 l

ps nog Ps UNITED s. ATES 0"

yo

.i NUCLEAR REGULATORY COMMISSION 3

y I

WASHINGTON, D.C. W6 o,s,...../

Docket No. 99901241 Mr. Calvin J. Creekncre, Mm39er Quality Assurance Power Systens Energy Services, Inc.

995 Day Hill Road, Suite 100 Windsor, Conrr<:ticut 06095

Dear Mr. Creekmre:

SUELTDCT: NRC INSPIITION REFOR1' NC. 99901241/91-01

'Ihis letter addresses the inspection at your facility in Wirdsor, Connecticut conducted by Messrs. Walter P. Raass, Anthony W. Markley, aM Daniel R. Carter of this office on December 10-13, 1991, ard the M':<'Mian of their findirgs with you and other nembers of the Power Systers and Energy Services, Inc.

(PSES) staff at the conclusion of the inspection.

'Ihe primary purpose of the inspection was to review PSES's process for providirq qualified decontamination and radiation protaction technicians to nuclear plant sites to aug: rent the licensee's staff during planned outages.

'Ihe min areas of concern incitdod the determination ard evaluation of personnel qualifications, ard the translation of these qualifications to the licensee for evaluation.

'Ib accanplish this effort, the inspectors held discussions with various ccgnizant members of the PSES staff, and examirwxi procedures, program descriptions, ard representative records.

'1he areas discussed and examined during the NRC inspection, ard our firdings, are presented in the enclosed report.

It was concluded that the PSES process for identifying, evaluatirg, ard supplyirg qualified decontamination ard radiation protection personnel adequately accmplishes its objective. It was notcd that the PSES activities c e conducted without invokirg the provisions of 10 CFR Part 50, Apperdix B and 10 CFR Part 21 since confomance to these regulations is not required by the li nsecs' purchase orders.

As a further outccme of this inspection, the followirq recomnondations for inproving the overall process of supplying contract radiation protection tectinicians to nuclear facilities were identified:

While it was concluded that PSES has well established ucthods and procedures for providirg contract perram1, hwnting this internal p vm would assist in the training of staff personnel, assure better consistency in perforwince, and facilitate understardity of the process.

1 25

Mr. Calvin Creekncre While scre positive initial steps have been taken to formlize the trainirn of the contract radiation protection tectudcian work force includirq licensee progranc and a cxxpilation of radiation protection technolcgy by an irdastry firm, there is a definite need for a nore fomalized ard stimod training sta;dard by irdustry to provide greater consistercy of personnel knowledge ard qualifications. (bntinued efforts in teaming arrangments with licensees to develop ard inplenent trainirg supans for contract radiation protection techniciari are crcouragcd.

Development by the irdustry of consistent, detailed qualification stardards for radiation protection technicians, similar to thoce develcped for quality control personnel, would be beneficial to licensees, vendors, ard contract radiation protection personnel.

'Ihe qualification "andards should ircltde a rulti-level certification program.

Since the functicns performed by raalation protection technicians are considered to be safety related because they may affect the health ard safety of the public as well as the nuclear plant staff, it is important that a clear and well structmmd distribution of responsibility be established with raytrd to assuriry that the tenporary radiation protection technicians supplied are properly qualified for their assigned tasks. 'Ihe division of responsibility for perfomirg functions includirg conparirg personnel qualitications to applicable irdustry standards, verifyiry the accuracy of resume information, trainity of personnel in discipline technology ard in plant procedures, testing of personnel to verify retained krmledge, and evaluating the acceptability of past performance should be clearly established between licensee and contractor.

In accordance with '10 CFR 2.790 of the Cornission's regulations, a copy of this letter ard the enclosed inspation report will be placed in the NRC's Public Docunnnt Roam.

26

Mr. Calvin Crockmore

~3-Should ycu have any questions concerniry this inspection, w are available to diraiss them with ycu.

Sirrerely,.

\\d,) t

(

N j 'Lc; /

Leif J. I 1m, Chief Verdor Inspection Brant Division of Reactor Irspection ard Safegthutis Office of Nuclear Reactor Regulation Dv::locure:

Inspection Report No. 99901241/91-01 27

Enclosure OFGANIZATION:

Power Systems Energy Services, Inc.

995 Day Hill Road, Suite 100 Wirdsor, Connecticut 06095 REFORT NO. : =

99901241/91-01 001TTACT:

Calvin J. Creckmore, Manager - Quality Assurance Telephone: 203-285-9860 NUCLEAR INDUSTRY Provide decontamination ard radiation ACTIVITY:.

protection technicians on a comercial-grade basis to augment licensecs' plant staffs during planncd outages.

INSPECTION CONDUCTED:

LWv= Mr 10-13, 1991

INSPECHORS:

66

[

/ 5jht.

no,..

W. P. Haass, Team Inader

. Nte '

Special ProjectsSection Vendor Inspection Brar &

A. W. Ibrkley, Inspector, Region III Division of Radiation Safety ard Safeguards D. R. Carter, Health Ihysicist Radiation Protection Branch, 1ER APPROVED BY::

)

W

' b b 2.

~

' Special Projecics Sect} ion G. C. Cwalina, 011ef Date Vendor Inspection Branch

. INSPECTION BASES:

10.CFR Part 50, Appendix B 10 CFR Part 21 ANSI N18.1-1971 (Reg. Guide 1.8)

ANSI /ANS 3.1-1978-Regulatory Guide 1.33 INSPECTION SCOPE:

' Announced inspection to examirv he verdor's process for the evaluation, selection, wx1 supply of decontamination and radiation protection techr'icians to augment nuclear plants staffs durire planned outages.

PLANT SITE APPLICABIIJTY:

Numerous 28

I 1-

'INSPIrrION '

SUMMARY

1.'1 yjolations No violations of Imc regulations were identified durirg this inspection.

1.2 Nonconformnces No nomonformnces were identificd during this inspection.

1.3 Unresolved Items

. No unresolved items were identified durirg this inspection.

2 INSPirrION FINDUCS AND OHIER CDtEDTTS 2.1.

Status -of' Previous Insr>oction Findims

' Ibis was the initial inspection oorducted by NRC of this verdor.

3.

INSPIUPION FDIDI1GS AND ODIER CDMDTTS 3.1 lY.rance ard Exit Meetims An entrance-meetiry was held on Dece @ J 10, 1991 in which the NRC inspectors discussed the purp x.e and scope of the planned inspection with the Quality Assuranm Manager and other members of the ABB Power Systems Energy Services, Inc. -(PSES) mnagement staff. During the exit

'meetiry held on tw y= N r 13, 1991, the NRC inspectors summarized the

- areas reviewed and the -conclusions drawn during the -inspection.

Ih.uuuerdations and concerns were also identified for generic

impmvement in the ' pro ss.

3.2-Bachrourd

~

The NRC has' W concerned tr.at the overall quality and performance of contract radiation protection technicians (RPTs) ucilized at licensee facilities may be in decline over the past few years. The basis for i~

this concern is'information gathered through inspection effortt.at I

licensee. facilities, information prtwided directly to the NRC, and informtior: obtained at professional meetings with irdustry personnel.

As a result, ~ the NRC decidal that it was appropriate to inspect a vendor active in the field of providing radiation protcction (RP) personnel on a contractual basis to licensees. PSES was selected as a representative supplier of these services. _ :'Ihere were no previously identified concerns or other regulatory issues involved in the selection of this-Einn. 29

l l

PSES is crqagcd in providirq technical personnel primrily to the nuclear irdustry. Personnel are provided in the functional areas of ergirmerirg, trainirn, quality assurance, quality control, decontamination, ard radiation protection. The service usually consists of providiry temporary perreel to augment the licensees' plant staffs durity refueliry ard mintenance cutages at the nuclear utility sites.

These services alco include providirg personnel to support specific tasks ard projects e5ther directly to nuclear utilities or in teamirg arr igements with other ABB sulnidiaries such as ABB Ccabustion Ergineerity ard ABB 1mpell Corporation.

PSES was incorporated in 1987 ard does an annual business volum in the

$40-50 million rarge. Of the total business volume, about 90% ir.volves the nuclear irdustry consisting of 50-60 clients. The company has about 33 full time employees, but its peak employment level during the Sprirg ard Pall nuclear plant outages reaches as high as 1000 contract e1ployees. PSES has access to about 11,000 RPTs in their databank with varyirq levels of experience aid qualifications rengiry frcm janitors to consultirg engineers to draw on to fill contractual needs.

Approximately 4000 of the RP1b are qualified at the senior level of expertise. Although PSES does have doctrrented programs ard procedures that meet the NRC requirements urder 10 CFR S0, Appendix B, ard 10 CFR 21, 2e work perforned for nuclear plant licensees may not conform to these regulations since the limnsee purchase orders (PO) generally do not specify their implementation by the contractor.

3.3 Proces for Supolvim Atrtmented Staff The inspectcsrs reviewcd the process for supplyirg qualified perrmel to licensees, ard interviewed the cognizant vendor management personnel.

PSES has well established mthods for identifying and evaluatirq prospective contract personnel. Although the vendor's internal process was not documented, management personnel providcd an adequately detailed overview of the process ard responded to all of the inspectors' questions pertainiry to the process.

Upon receipt of a purchase order (PO) frca the licensee, PSES cmmnces a review of its files, database ard other sources of information to identify potential-candidates for satisfying the PO requirements.

A list.of cardidates for the respective positions, as required, of RP supervisor, senior RI71', junior RPT, ard decontamination technician is generated. PSES contacts these irdividuals to elicit their interest in the particular job assignment with regard to location, tim frarae for work, and the rwture of the duties.

If the irdividual is interested in the position, PSES will then proceed to evaluate the individual for suitability to the position requirements.

Evaluation of the individual's qualifications by PSES consists of obtaining the inctividual's resume from the individual, comparity it to the licensee's program requirements ard specific project needs as identified in the PO, and transcribiry it onto the PSES letterhead. 30

Program requirencnts norm 11y include, but are mt limited to, metirg Ancrican National: Standards Institute (ANSI) experience and trainiry requirements (as crdorsed by the licensee's nuclet.r plant 7bchnical-Specifications) and specific limnsee administrative requiremnts.

Specific project needs my include mrtain types of work' experience such

.as prior assignrents at boility or pressurized water reactors, steam generatc,r repair and replacement, recirculation pipiry replacement, refuelirg, ard control rod drive mechanism mintenance. 'Ihe individual resumes determined to be acceptable are forwarded to the li nsee for furt.her evaluation as nemory.

hxordirg to PSES, if the icensee is not familiar with the prqxrmd irdividual, verification of the individuals' qualifications will be conducted through professioml and/or personal refereme checks. 71m licensee then notifies PSES of the personnel selected frm the proposed group. ' PSES will then contact the selected individuals to mke arrargements for mobilization, security screeniry for unescorted access, and staffirq.

Again, acrordire to PSES, upon arrival at the licensee's facility, tlm accepted individuals are placed in the licensee's trainirg programs.

These include general employee training, fitness for duty training, specialty training, ard trainirg in plant arrangement, policies ard procedures.1Trainirg classes are frequently given over a period of a few days to a week. During this time, acconfirg to PSES, the individuals are tested to dotamine their. knowledge of radiation protection technology, plant procedures, and industry practices. PSES indicated that almost all licensees have established testire methods ard on-the-job training pivgrams for m ntract RPI's that provide the final assurance that the contract RPIs are properly qualified for the safety-related functions to be performed. PSES, however, was not able to provide any further infomation regarding the quality _or adeq1acy of these licensee training and testing programs.

~ Based on the above review,'the inspectors-concluded that the supplying of contract RFfs' by PSES to nuclear plant licensees is perfomed in accordance with licensee PO requirements with subsequent verification',

trainina ard testing activities, according to PSES, conducted by the licensee to provide the rvwmry assurance that the contract personnel are pmperly qualified. These licensee activities are necessary since the service perfomed by the contract RPTs affects public health ard safety ard is therefore categorized as safety-related. 31

o 3.4' Licensee Purchase Ordez Based on the neod for augmented resources to support operation, mainterance ard refueliry of the nuclear plant, rcquests for quote (REQ) are issuM by the licensee. *1he RFQ defines the commercial, administrative, and regulatory requirements that the vendor r.ust address in the responding propocal. The PJQa generally specify tle level of qualification of personnel required rather than the number of personnel, the latter varyirq durirg the course of the work. None of the REQs reviewod required that the veMor certify that the personnel proposed for the scope of work are qualifiod.

In general, the REQs identified the jJdustry standards to be uscd to determine the level of qualification and that an appropriate quality assurance program be utilized. The inspectors did not identify any REQs that included a specification for conpliance with 10 CF'R 21.

While the need for. a quality assurance prcgram was irdicated, none of the RFQs reviewed specified a need for the quality assurance program to meet the requirements of -20 CFR 50, Apperdix B.

'1he vendor respords to the RFQ with a proposal that addresses or takes exceptions-to the RFQ. The prowl specifies the verdor maragement involvencnt, specific ccrnmercial and administrative requirenants to be met,- identifles the exceptions as appropriate, ailresses the qualification requirements of the contract RPlt, aM provides the resumes of :the vendor nanagenent and technical personnel who will parform the work.

Once the licensee has ' selected the desired veMor, a purchase order is written. The purchase order frcquently references the requirements of the RFQ ard the resporrilrq propr.al. The FO also specifies the applicable regulatory r@mts.

Bascd on the inspectors' review of several RFQs, proposals, and PCs, it was concludcd that the PSES commitment for supplyirg contract RPIs is consistent with the limasce's direction.

3.5 Contro111rn Ileauircrents The inspectors reviewed the NRC regulations, regulatory guidance, licensee 'Ibchnical Specifications, ard the ANSI qualification standards that were determined to be applicable to the PSES activities urder review during this inspection. The followirg describes the pertinent requirements that are applicable to the utilization of contract RFIb at licensed facilities.

10 CFR Part 21:

In past correspordence (Information Notice 85-52,, dated July 10, 1985), the NRC has clearly indicated that the potential. exposure of the public, which includes licensee personnel and contractor employees, to excessive 1cvels of radiation ard/or radioactive materials is cornidered a substantial safety hazard to which Part 21 is applicable. -

32

me supplyirs of contract RPPs to licensees is tantanount to providing basic cmponents for use in safety-related applications.

me utilization of an irdividual who lacks the proper qualifications in a safety-related activity is amltgcus to the use of a defective basic ocoponent. m erefore, the use of an imdequately qualified RPr to perforn safety-related activities could create a sutstantial safety hazard thereby subjectirg the use of contract RPIb to the provisions of Part 21.

Regulatory Guide 1.33: his guide, entitled " Quality Assurance Program Requirer,ents (Operations)," endorses ANSI N18.7-1976/ANS 3.2 (subject to additioml staff positions) as acceptable quality assurance criteria meeting 10 CFR 50, Apperdix B for the operatiom1 phase of a nuclear facility. Apperdix A, item 7 (a-h) to that guide lists several safety-related activities, inchdiry procedures for the control ~of radioactivity ard radiation, which are subject to the provisions of the guide. 'Iherefore, radiation protection is a safety-related activity urder the control of the Apperdix B operational quality assurance program.

Technical Specifications: All licensees have statenents in their Technical Specifications that require personnel at the licensee's facility to meet minimum qualification levels as defined in tne ANSI standards. In addition, most licensees have either an en3crsenent or statements that incorporate the intent of Regulatory Guide 1.33 in their 'Ibchnical Specifications.

Therefore, the-licensee's radiation protection program is subject to the quality assurance requirements of 10 CFR 50, ~4perdix B.

ANSI N18.1 1971 ard ANSI /ANS 3.1-1978: 'Ihese standards are commonly used to specify the minimum qualifications for the various levels of RFIt at nuclear plants.

For full performance technicians (senior technicians), ANSI N18.1-1971 requires that technicians have two years of workire experience.

In addition to this work experience, the technician should have one year of related technical trainirg. Also for the full performan technician (senior technician), ANSI /ANS 3.1-1978 requires that the technician have three years of experience, one year of which -

should be related technical training.

Since the stardards do not alcolutely rcq ire related technical training, as such, limrmes-b have not required that their vendors provide personnel who possess l

applic@le related technical trainirg.

Rather, the focus has been on.the accumulation of experience to meet the minimum qualification requirements.

As diwW with PSES mmgement, each licensee frequently interprets the ANSI stardard differently. One licensee may give credit for a given L

type of experience, such as decontamination activities or Navy nuclear E

experience, while another licensee does not.

Some licensees will give l

credit for certain types of formal education while other licensees will l

-not.

l' l L

33 1

l I

PSES managemnt indicatcd that a significant experditure of rescuroes is inattrred in its ef forts to provide personnel that meet the iniividual licensee's requiremnts.

PSES mmge:nont inlicated that detailed, prescriptive qualification requiremnts for RPTs, similar to quality control personnel, would be beneficial to the irdustry. :

such were the case, then all licensees would ham essentially the sare interpretation of the qualification requirerents and the verdor's coct of doiry basiness would te lowertd significantly PSES mnagemnt also irdicatcd that if detailed, prescriptive qualification standards were adopted by the irdustry, the ccrpetitive environment would beacre rore levelized thereby forcing unscrupulous verdors to ctrpete fairly by supplyirs only qualified p e nnel. Accordity to PSES, the prescnt broadly integreted qualification rcquiremnts allow mrgi.nd oc even unqualified personnel resunes to be suhnitted to licenses for employmnt consideration.

In summary, licensees are not obligatiry the suppliers of contract RPrs in their Pos to meet all of the provisions of the above described requiremnts. The licensee perfonts a portion of the qualification activity, moct likely consisting of acilitional trainire, testirg ard verification as descrited previously (see Section 3.3), to assure that the contact RPTs are properly qualified to corduct safety-related activities. The activities corducted by the licensee nust be perforred prior to the RPT's initial performance of duties et the licensee's plant.

3.6 Oualification Reviews of Proposed RP Technicians The inspectors reviewud several sele'.;ted PSES personnel files containing both the resume subnitted by the prospective contract RPT ard the correspondirg work experience resume transcribed by the verdor under a PSES letterhead for transmittal to the licensee.

In general, the vast majority of the information was transcribed uncharged. However, sevaral instar; es appeared to contain some minor enhancements that my have resultee in the irdividual beiro qualified at the senior rather than the orior RPr level. Another exagle was noted in which the individual's irdicated experience with steam generator mintenance ard refueliry was supplemntcd in the transcription by experience in reactor coalant purp minternnce. Other examples were identificd in which the individual's resu:te irdicated experience with reactor coclant purp ard steam generator minternme ard the transcribed version includal refuelire experience. '1hree specific examples of the inspectors' review irdicating questionable review ard quality control are described belcu.

The irdividual's subaitted resum irdicated a decontamimtion worker backgrourd with laborer experience.

Approxirrately 1-1/2 years later, the individual subnitted an updated resume to PSES which not only added recent experierce but also significantly upgradad the Inture and characterizatior. of a previcus job experience at Connonwealth Edison's Byron Nuclear Station.

~7-34

-.-n 7he revision indicated that a significant portion of the previous decx>ntamimtion worker experience (approximately 4300 hours0.0498 days <br />1.194 hours <br />0.00711 weeks <br />0.00164 months <br />) was now categorized as junior health physics technician experience.

PSG transcribed this informtion onto the cxmpany letterhead resume without performing any quality control checks or verification of this experience. This individual was subsequently proposed aM accepted as a senior (ANSI qualified) RPT to cammonwealth Edison's laSalle County Nuclear Station in Docmber 1989.- Without credit given fcr this upgrade in experience, the iMividual would not have been qualified to perform as a senior RPT. During this inspection, this individual was accepted for a position with another licensee.

A scooM individual was identified for which an extraon11mry amount of-credit was allowed for decontamination worker experience. The individual was supcad and acceptal as a senior (ANSI qualified) RPT to Ocrucnwealth D31 son's laSalle County Nuclear Station in November 1989. With a more normal ne m em nt of experience, this inlividual would have been approxim tely 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> short of meeting the ANSI requirement for this position.

A third individual was proposed and acx:epted as a senior (ANSI qualified) RPT to Northeast Utilities' Millstone Nuclear Plant in thrch 1991. This individual's resume contained a job entry for the period of August 1984 through April 1988 (7,500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br />) titled

-" Health Physics Coordimtor/ Advanced Radiation Worker.'? The resume entry described a qualification program rather than the individual's position and associated duties during his enployment by Arkansas Power and Light (AP&L). The individual's perscnnel file contained what appeared to be an extended self-monitoring program.

Discussion by the inspector with PSES management indicated that AP&L had experienced significant loss of licensee RP personnel at one time and had carpensated by training other perronnel such as auxiliary and equipment operators to perform l

radiological sur mys.

Since the previous job entry delineated dutir.s as an auxiliary operator, 'it would appear that this job entry was a continuaticn of the previous entry..Since no information becam available during the inspection regan11ng the proportion of time spent and the mture of the dutjes performed partaininy to RP for this job entry, the inspector was unable to-determine % tether the individual would have been qualified for this position. PSES transcribed this infomation onto the ocupary letterhead resume without performing any quality. control check or verification of this experience.

As.a result.of the inspectors' review of individual resume information, inspectors' it was concluded that, in general, accurate transmittal of information received from the RPT applicant was made to the licensee.

_g_

35

~w-

+

ikuever, instances were noted in which minor enhancement of the irdividual's experience entries was alparently done; in sca:e instances, the reverse was roted in that the length of work experience claimed by the prospective RPf "as rcducal by PSES. Consistent with the practices utilizcd by PSES, no quality control checks and verifications of the irdividual's restu:n entries were performed. 'Ihe latter, accordity to PSES, would be the responsibility of the licensee.

3.7 Licenser Eya3Mion ard Selection of Atrmentcd RP Staff Verification of the validity of the pros M ive employee's resume informtion is often a necessary step in assurity the qualifications of technicians proposed for site staff augnentation. As noted above, the PSES recruiter obtains resuno infornation directly fram the technician and determines qualifications bascd on tnat input. Other than as incidental to a backgrourd investigation performed for security pirposes, no indeperdent verification of the technician's industry experience is perfomed by PSES.

PSES did irdicate that such verification activities are urdertaken by the licenree as a final check on the IJdividual's qualifications prior to caployment. According to PSES, the licensee's verification activities include contact with previous enployers, probably on a spot-check basis, to confirm periods of work, sixcific tasks performed, ard suitability for rehiring. In addition, the licensee usually administers an examiraticn as a further check on the applicant's qualifications.

The licensee's verification activities rerve to supplement the PSES activities ard provide the necessary assurance of the RPT's qualifications to perform the assigned tasks.

4.

PSES PEIG3NNEL OJtTTALTfD The folloaing PSES personrel were contacted during the course of the inspection:

+

• J. R. Wyvill President

+

• R. S. Bell Vice President and General Omrsel

+

• C. J. Creekmore Manager, Quality Services 4

• R. G. Ne@lm Marager, Support Services

+

• S. M.'Wargo Marnger, Human Relations W. J. OTalfant Manager, Dqineerity and Training Services

+ Attended the entrance meeting on December 10, 1991

• Attended the exit nretiry on December 13, 1991

_9 1

36

-~

.... -. - ~~

. ~.. -...

- -.~.-..-.

- [g no k w.

z,

$3P UNITED STATES i

. ' y( oE NUCLEAR REGULATORY COMMISSION -

- WASHIP'1 TON, D C. 20666

~

y*****/

-June 9, 1992 L:

Docket No. 50-298 o,

Hr. Gay R. Horn-Nuclear Power Group Panager Nebraska-Public Iw er District

' P. O. : Box -499 -

Columbus, Nebraska 68602-0499_

C

SUBJECT:

'INSPDNIGi OF BIE PROC 11RDM AND C0lHERCIAL GRADE DEDICA2701 PROGRAMS AT MIE CD3PER NUCLEAR STATION, (REPORP 110. 50-298/92-201)

'Ihisiletter trarsmits the report of the inspection corducted February 24 through 28,.1992, at Nebraska Public Power District's (NPPD's) _ Cooper Nuclear 2 Station (CNS) ' by R i Pettis, S. Alexander, W. Gleaves,. and B. Rogers of the Nuclear _ Regulatory.Comntission's (NRC's) Vendor Irsoecticri Branch afd L; Ellershaw +nd R. Evans of NRC Regitn IV..

The inspect. ion was related to activitjes at the plant sitt authotizcd by NRC license DPR-46.- At the

conclusion of the inspecticn, we diwW our findirns with you and the members of your staff identified in Section 5 of the emlosed inspection report.

. Ihe' inspection was conducted to review.the implementation of the HPPD programs

~

for the procurement and dedication of commercial grade;itess (CGIs) used.in

.. safety-related = applications lat CNS. 'Ihe results of the inspection irdicate that CIS failed to properly dedicate certain Q3Is procured for use in safety-related applications.o Cbnsequently, CGIs of indetermirate quality were iirstalled or accepted forfinstsllation in plant safety systems. 'Ihe specific deficiencies ' contributing to this condition included failure to identify __

safety. functions of.the CGI,Efailure to identify' critical characteristics-irelating-to the specific safety functions of the CGI, failure to adequately i verify the critical characteristier that were identified when usirq the cmmercial gradWsnecification approach,,and the-. failure to identity ard verify methcds for the seismic qualification of most CEIs.-

Ihe most significant area requiring' increased attention was NPPD's extersive reliance on the use of-the' essential-cmmercial_ grade -(EDG) apprcach, which -

' relies on the perfomance of broad-based, programmatic surveys, ard post-installation testing, fcr product acceptance of suppliers who do re maintain or commit to having a 10 GR Part 50 Appendix B quality Essuranac program.

'Ihis approach does:not provide a sufficient basis lfor verifying' that the -

supplier's activities centrol the required critical characteristics..It should be nottu that audits conducted in accordance with Criterion XVIII,

>" Audits,"'of Appendix B to_10 CFR Part 50 and the guidance provided in NRC Regulatory Guide 1.144 and American National =Standarxis Institute (ANSI)

N45.2.12-1977, can be used as'an alternative to Electric Power Research

Institute ' (EPRI) Method 2, "Ctrimercial Grade SurwrL " provided that the atxiit 37 a

.~

Mr. Guy Horn confirms, by direct observation, that procedures are in place for controlling the required critical characteristics identifico by the purchaser. QJS prccMures did not recognize the need to perform crranercial grade specific surveys until Revision 17 of CJS Quality Assurance Instruction (QAI-16),

" Supplier Approval," May 17, 1991, addressed ocrmercial grW surveys ard required the use of the Nuclear Drocurement Issues Omnittee ccxanercial grade survey checklist. However there was no technical evaluation required as part of the D33 process and therefore, no nahanism for formally determinirg critical characteristics to be uscd in such surveys.

Furthernere, it was noted that only a _ few ccrenercial grade cpecific surveys have been utilized by CiS to date. A review of these surveys identified sane implementation deficiencies.

I Additionally, W--installation testirg pformed urder the E ap2 was ger'erally not sufficient to verify the full rarge of design corditions, including seis:nic, necessary for the item to perfom its safety functions.

Utilization of this approach does not ensure ccrpliance with 10 CFR Part 50 Appendix B.

We recognize that NPPD identified certain program ard iraplenentation deficiercies itself, as evidenced by NPPD's canitm2nts to strengthen the prcgram by July 1,1992 (Appendix), ard consider this self-identification a positive actiot. However, the najority of NPPD's program improvements appeared to be initiated at about the tilte of NRC's announcement of its inspection on January 15, 1992; our peinspection site visit in 1sw Anuary; and NPPD's retaining a consultant to review its prcgram before our inspection cornmenced. All licensees are ommitted to reviewing and improving their programs in accordance with the Nuclear Maragenent and Resources Cbuncil's first initiative on the dedication of CGIs by January 1,1990. 'Ihis initiative stated that licensee programs should meet the intent of the

. guidance provided in the EPRT Final Report NP-5652, " Guideline for the Utilization of Caramercial Grade Items in Nuclear Safety Related Applications (NCIG-07)," by Jartuary 1, 1990.

The inspection firdiajs presented to your regesentatives duma the exit meetity at OJS on February 28, 1992, ard discussed in this letter and in the l

eaclosed report, are considered deficiencies in your procurenent ard commercial grade dedication activities, ard will be referred to the NRC Region IV office for any apprcpriate enforcement action.

i In accordance with 10 CFR 2.790(a), a copy of this letter and the enclosure will te placed in the NRC Public Document Room.

l 38

- _ _ - ~

Mr. Guy Horn,

1 Should you have any questions ecrernirg this incpection, we will be pleased to discuss them with you. 'Ihank you for your cooperation in this inspection.

Sinmrely, M

, Director Bruce A.

Division of Reactor Projects III, IV Office of No:: lear Reactor Rogulation, V Enc 1ccure:

Inspection Report 50-298/92-201 cc: see next page f

39

l l Mr. - Guy R. Ibrn (baper Nuclear Station Nuclear Power Grct'p M1naga.r oc:

Mr. G. D. Watson, General Ctunsel Nebraska Public Power District P. O. Box 499 Cblumbus, Nelraska 68602-0499 Cooper r%10" Station

{

ATIN: Mr. John M. Meacham Division }hruger of Nuclear Operations P. O. Dox 98 Brcunville, Nebraska 68321 Randolph Wood, Director Nebraska Departmnt of Divirorrnental Control-P. O. Box 98922 Lincoln, Nebraska 68509-8922 Mr. Iarry Bohlken, cuirmn Nenuha ccunty Board of Ctanissioners Dennha County Courthouse 1824 N Street Auburn, Nebraska 68305 Senior Resident Inspector U.S. Nuclear Rogulatory Ctrnission P. C. Box 218 Brcunvil)e, Nebraska 68321 Regional Nininistrator, Regicn IV U.S. Nuclear Rcqulatory Otrrdssion 611 Ryar. Plaza Drive, Suite 1000 Arlington, 'Ibxas 76011 Mr. Harold Borchert, Director Division of Radiological }:ealth Nebraska Department of fleslth 301 Cer.tennial Mall, So th P. O. Box 95007 Lincoln, Nebraska 68509-5007 l

l l

39 a

l

!1 U.S. NUCLEAR R EULNIG W OW KISSIQi OFFICE OF NUCLEAR RENCICR RHAHATIQi DIVISIO{ OF REAC'IQt INSPECTIO4 AND SAFEKDRDS Report No.:

50-298/92-201 Docket No.:

50-298 License No.:

DPR-46 Lioerr,ee:

Nebraska Public Power District P. O. Box 499 Columhm, NekrarJca 63602-0499 Facility Nam:

Cooper Nuclear Station Inspection at:

Nanaha County, Nebnska Inspcction Otnketed: February 24 through 2_8, 1992

[

\\ r, 6

-b Robert L. Pettis, Jr., P.E.,

Dath

'li3am TMr Venkr Inspection Brard (VIB)

Division of Ra% tor Inspectico ard Safeguartis Office of Nuclear Reacter Rogulation Other Inspectors:

S. Alexander, EQ ard Test Egiwar, VID W. Gleaves, Mechanical Engineer, VIB B. Rogers, Reactor Ergirhaer, VIB L. Ellershaw, he+rw-Egineer, RIV R. Evans, Real&nt InEqutxxr, PIV

(

'A f[L7 7 L.-

p g:

Imif J. I 1m, C11ef Date '

verdor cri Brarxt

-Divisicn of Reactor Inspectica ard Safeguartis Office of Nuclear Rasctor Regulation 1

J

(

40

1 i

i TAB M OF Ca nDfIS Pane D3DJrIVE SONARY....................................................

i 1 DTIPODUCTIQi.......................................................

1 2 CDHDtCIAL GRADE DEDICATIQi ITOGRAM REVIW.........................

2 2.1 Procedures Revisa.............................................

2 2.1.1 Essential-Ortnercia1 Grade Method.....................

4 2.1.2 Whtion Method for Ctxnmercial Grade Itens pit: cured as thential...............................

5 2.2 Ctrraerr.ial Grade Su@ lier Surveys.............................

6 2.2.1 'Ihird Party Surveys....................................

3 2.2.2 Source Verifications...................................

8 2.3 Parts M m e.n1f ication..........................................

9 4

2.4 Trending of Suppliers.........................................

11 2.5 Detecticn of Fraudulent }hterials.............................

12 3 DEDICATION PAQGGE PIV1EH..........................................

12 3.1 Items Puranased as Essential-Ctzmnrcial Grade.................

13 3.2 Itans Purchared as Ncnessential...............................

18 3.3 Lubricants and Fluids Used in Essential ard Essential-D2 A m11 cations..................................................

2L 4 PPOCURD4ENT AND DEDICATIQi 'IRADUlG................................

22 5 DGT MEETDG......................

23 APPDOIX - NPPD PFOCURDIDir PROGRAM D4HANCEMD"IS PROVIDED AT 'nIE EXIT MEETDC...................................................

A-1 41

EXILVrIVE SWMARY FYcra February 24 through February 28, 1992, representatives of the Nuclear Rcqulatory Ctanission's (NRC's) Vendor Inspection Branch and the Region IV office inspectM Nebraska Public Power District's (NPPD's) ac'ivities related to the procurc2cnt and dedication of cxrrnercial grade items (CGIs) used in safety-related applications at the Cooper Nuclear Station (01S). %e

~inspecticn team reviewed NPPD's procurement and dedication prcgram to assess the licensee's compliance with the quality assurance (QA) requirements of AppcMix B to Part 50-of Title 1LC of the Code of Federal Regulations

_(10 CFR Part 50 AppcMix _ B).

On August 24,1990, the NRC staff forwarded to the Commission SECY-90-304,

. JU4 ARC Initiative.s on Procurcrent," in which the staff reported the status of the Nuclear Management and Resources Council's (IU4 ARC's) initiatives on general procurement practices.

Procurement initiatives as described in IU4 ARC 90-13, " Nuclear Procurement Program Irprovements," dated October 1990, ccanitted licensees to assess their procurement programs and take specific actim to strengthen inadequate picpun.- W e initiative on the dedication of CGIs, which was supposed to be accomplished by January 1, 1990, stated that i

licensee programs should meet the intent of the guidance provided in the Electric Power Research Institute (EPRI) Fiml Report NP-5652, " Guideline for the Utilization of Catmercial Grade Items in Nuclear Safety Related Applications (NC1G-07)," dated June 1988. W e staff also stated in SECY-90-304 that it would corduct awents at selected sites to review the licensees' implementation of improved procurement and commercial grade L

dedication programs, assess improvements made in the areas covered by the IMARC initiatives, and report the results of those a--nts to the Commission. From February through July 1991, the NRC's Vendor Inspection Branch conducted eight a w - nts of selected licensees to determine the current status _of-activities to improve the procurement program related to industry initiatives and NRC requirements. On September 16, 1991, the NRC staff _ forwartled to the Ctanission SECY-91-291, " Status.of NRC's Procurement Assessments and Resumption of Programmatic Inspection Activity," in which the staff reported on the results of its-awennnts aM noted that it was resuming inspection and enforcement activities.

NRC conducted this inspection at 01S, the second since completing the eight i

earlier assessments, to review NPPD's procurement and dedication programs and their implementation since January 1, 1990, the effective date of the I M ARC initiative on dedication of CGIs. Se inspection focused on a review of procedures and representative records (incltrlirg approximately 30 procurement and dedication packages for mechanical and electrical items); interviews with

NPPD _ staff (including NPPD senior management and WS site personnel); and obcervations by members of the inspection team. 'Ihe inspection team also met with NPPD's management to discuss relevant aspects of commercial grade dedication and to identify areas requiring additional information. Se

-inspection team's findirgs were discussed with NPPD's representatives and senior mnagement at the exit. meeting held on February 28, 1992. We inspection team identified two deficiencies, with multiple examples, summarized below, l

1 42

)

Deficiency 92-201-01

'Ihe incpection team identified numetuis examples in which NPPC either installcd CGIs in safety-related plant applicat. tons or had identifiod them as available for installation in safety-related appiications at 01S without adcquate review

_e cuitability of application of these mterials, narts,

^

equipment, and proocsses that were essential to the safety-relatal functions of structures, systems and components. NPFD failed to adequately determine the suitability of application of CGIs which resulted in the use or warehousing of CGIs of irdeterminate quality, as irdicated in the following examples:

(1)

Items Purchaned as FasentialhJcial Grade (FD3)_

'Ibe following are examples of C31s furchased urder the DOG item classification that were improperly dodicated by 01S. This approach to dedication is delineated in cis Plant Services Procedure 1.13, " Utilization of Essential-Commrcial Grade Items in Safety-Related Applications," Revision 0, April 18, 1990, which acknowlei es the limitations of the suppliers' QA program ard also 3

recognizes the neal for NPIO to take responsibility for 10 CFR Part 21 report-ing sin these suppliers do not have 10 CFR Part 50 Apperdix B QA programs.

'Ihe following purchase orders (Pos) were ismod for items that were purchascd as EDG ard either installed or mde available for installation in essential (10 CFR Part 50, Appendix B [ safety-related]) applications at OIS.

a.

PO 312069: Autmatic Switch Comp 1ny (ASCO) " Red Hat" solenoid-operated valves (SOVs) were purchasal frun the John Day Company of Omaha, Nebraska, on January 24, 1990. 01S relied on a 10 CFR Part 50, Apperdix B atdit of ASCO which focused on AS00's dedication of parts used for its nuclear, environmentally qualified (EC) FP-1 series SOVs.

" Red Hat" SOVs are manufactured urder ASCD's commercial program which does not meet Appendix B raluirements. NPED Mmorardum QAD 9100010 placed this PO on hold; three other similar Pos were also identified and placed on hold, b.

PO 342961: Sixteen Belleville spring washers to be installed in the high pressure coolant injection pump were purchased from Georgia Power Company's (GPC's) Plant Hatch on January 17, 1991. Plant Hatch was placed on 01S's suppliers list (SL) as a distributor for Dresser-Pand (the manufacturer) bascd on a telephone interview and a review of an NRC assessment report issued in May 1991. Dresser-Pand appeared on the 01S SL as a supplier of E03 iters. The package reviewed did not contain documentation to support the traceability of the 100 washers purchased as nonsafety-related by GPC, nor did it contain any certification to establish the traceability of the 16 washers supplied to 01S. During the inspeculoa, 01S perforrcd an operability evaluation to determine suitability of the installal washers. As a result, 01S dcungraded the washers to nonessential (nonsafety-related) on the basis of information received from Dresser-Rand.

-ii-l 43

p c.

M 311091: Wire wound resistors to be installed in the control room's IMIS augmentation system were purchased frca Dale Electronics on January 8, 1990. OfS ao pted a January 16, 1990, test report that rh'mmited the nranthly testim of resistors using samplim criteria established under Military Specification MIIrR-26E; hcuever, 01S relied on a cmmercial grade survey that was performd nine nonths after the PO was placed (October 25-26, 1990).

d.

PO 311792: 'Ihirty-eight 150-pourd gaskets purchased from F]exitallic Gasket Cmpany, Pennsauken, New Jersey on January 17, 1990.

Flexitallic appeared on the 04S SL as an EXI; supplier based on a broad-based progranmatic audit which was not critical characteristic specific i

to the otrimercial ituts parchased. Other fos reviewed that contained similar deficiencies were Pos 311015 ard 310601.

e.

PO 312109: One hundred half-inch-diameter Nelson studs were purchased from T W -Nelson, Welding Division, on January 18, 1990. 'Ihe PO required-the supplier to provide certified material test reports (OtrRs) and maintain ard apply the supplier's QA program which ncets the applicabls portions of 10 CFR Part 50 Appendix B ard American National Stardards Institute (ANSI) N45.2.

Such requirements are in violation of Section 8.2.1.2(a) of Q1S Procedure 1.13 for FIII purchases.

(Another example of this is PO 314872 for Viton o-rings purchased from the Parker-Hannifin Corporation.) T W-Nelson was qualified as an approved supplier based on a broad-based prograntiatic atulit perfoned by NPPD on May 29, 1989, which utilized a Coordinated Agency Supplier Evaluation-Nuclear Section checklist, f.

PO 329366: Six service water pump shaft couplirqs were purchased from W/IP International (formerly Borg-Wamer Industrial Products) on

. January 31, 1991. 'Ihe PO required the couplings to be manufactured frm American Society for 'Ibsting ard Materials (ASH 4) A-479, Type 410 material and a certificate of confontance to be supplied verifying that the itenu cupplied meet all EO requirements. A review of the package fourd that at least two rJiaft couplirgs had been installed in service water pump 1C on Octnhw 19, 1991. After post-mairtenance testing, the pump was declared operable on November 5, 1991. W/IP was approved only as an essential supplier based on a Nuclear Procurement Issues Committee (NUPIC) joint utility audit performcd in August 1989 in which W/IP's 10 CFR Part 50, Apperrlix B QA program was reviewed and approved. However, the subject PO was placed as ECG. During the inspection, the NRC inspection team questioned the suitability of the installed couplings since there was no documentation available to support their comercial quality. On February 27, 1991, 01S received documentation from W/IP statirq that although the couplings were ordered commercial grade, W/IP processed the order urder its nuclear QA program ard accepted the reportirg responsibilities of 10 CFR Part 21.

W/IP also supplied 01S with 04 irs from Earle M. Jorgensen (the material manufacturer) that verified the specified material.

g.

Other Pos: Several Pos that were reviewed related to various emergency diesel generator (EDG) replacement parts purchased from Cooper Energy

-iii-44 l

Services (CFS) ard installed in both EDGs. Each PO rcquested that CES provide a certificate of conforwince statirg that the iters were equal to or better than those origimlly supplied to 01S (1964) and to irrxse CES's 10 CFR Part 50, Appendix B QA program including ANSI 1145.2, despite the fact that these were D33 Pos. This practice violated Section 8.2.1.2(a) of CIS Procedure 1.13 which prohibits invoking unique nuclear rcquirements on a comnnrcial supplier. GS's dedication primrily relicd upon acceptance of CES's certifications combincd with a stardard visual receipt inspection ard post-installation tests (PITS) nontally required by plant technical specifications. CES stated that all fos accepted af ter July 1,1991, would not contain the " equal to or better than" statencnt Lut would contain a reviscd statement that the items are considered 031s and as such, CES makes rc clairs to form, fit, or function. IOs reviewod that contained these deficiencies were IOs 329844, 326792, 312074, 322798, ard 336439.

(2) RQDqssential Items Purchased Urder the Conmercial Grade SrecificAt; ion

.(03S) Classification

'1he following are exanples of 03Is purchased urder the CGS classification that were improperly dedicated by CNS. This classification is delineated in W S Engineering Procedure 3.22 ard requires a technical evaluation that identifies the item's safety function, critical characteristics, and acceptance methcds.

01S staff has performed approximately 40 CGS dedications since January 1,

1990, a.

Dedication Package 90-031 (IO 177637): CR 2940U310, a circuit breaker enclosure rackdcwn interlock switch, was putrhased from the Ceneral Electric Supply Company on November 21, 1980. The switch was installed on titrcia 26, 1990, in a Class 1E, 4160-Vac circuit breaker associated with the core spray pump. Concerns identified incltded listing "near infinity" as the acceptance criterion to verify open contact resistance, ard a check to verify the temiral-to-grourd resistance using a PIT. However, documentation did not exist to support that the resistance-to-grourd test had ever been perfonted.

b.

Dedication Package 90-032: Six hirge pins 'or several 18-inch tilt disc check valves were purchased from Anchor-Darling (A/D) Valve ocupany on June 16, 1989. A review of the package irdicated that four of the six pins ordered were installed in several safety-related reactor feedwater check valves that act as containment isolation valves. The technical evaluation identified safety function, environmental ard design criteria, ard such critical characteristics as outside diameter end material. Deficiencies noted durity the team's review included no docunentation to support the identification of the mterial, failure to address design differences between the safety and nonsafety-related check valves that these pins were to be generically used in, and failure to provide a basis for the purchare of nonessential items from an essential (10 CFR Part 50 Appendix B) supplier.

During the latter part of the inspection, NPPD perfornyxl an operability review for the installed pins. As a result, A/D committed

-iv-45

to providing NPPD mtcrial certifications by early Much (1992) for the six pins,.despite the fact that they were ordered as nonessential.

(3)

Lubricants s1Euids Purchased as Nonessential ard U. sed in Essential and yssential-ED Arvilcatipmt

'Ihe following are examples of lubricants, oils and greases purchased as nonessential but used in essential applications, includirq EQ equipment, for which 01S had no procedures in place for determining their suitability in a safety-related molication, Additionally, no analysis exi:;ted to document and ensure similarity to the lubricants testod as documented in the EQ test report

-or to establish traceability back to the original equipment manufacturer (OEM).

a.

PO.326028:. DAG 156 lubricant, procured from Acheson Colloids Company

.on November 15,-1990,- was used on safety-related_ main steam isolation valve _ (MSIV)' stems, guide rcxis, and. internal threads. 'Ihe team's review identified no documentation to support the campat bi ility of this

- material with the Versilube lubricant used on the MSIV pneumatic actuator o-rirns or with the elastomers used in the MSIV's SOVs supplied by ASCO, which could be exposed to this material-as air is exhausted through the SOVs durirs the MSIV closing. cycle.

b.

PO 346760f, -Mobil UI'E 797 oil, procured from the Allied Oil and Supply Ctxnpany on January _.17,_1992, was.used in various safety-related EQ applications such as the core spray pump motor bearing. A review of the General Electrjc (GE) drawing for the core spray pump and motor

'showed that GE specified the bearing's minimm viscosity.to be 45 Saybolt Universal-Seconds (sds), whereas the Mobil product data sheet specified 44' SUS. -Ihere was no documented resolution of or i

-justification for the discrepancy.. Other deficiencies identified 4

' included: no traceability to the OEM,_ critical' characteristic of

' environmental qualification not wrified, _ and no similarity to the EQ

sample _or traceability to the original EQ test report. Other Pos reviewed which exhibited the sam types of _ deficiencies included

- PO 250546 -(Chevron SRI No. 2 grease used -in IQ electric motors), and POs 343117 and 315910_(Mobilgrease 28 used:in Limitorque actuator limit switch gear boxes).

Lp

' Deficiency 92-201-02 The inspection team identified several generic weaknesses in the procurement i

program and in implementation that contrihited'to the specific examples of

~{

h

- deficient' 03I dedication. described in Deficiency 92-201-01, j

.y i

ll L'Ihe mast significant weakness concerned the use of the EOG approach to l

Tdedicatina 03Is for safety-related applications. Under this approach, there is-no requiremnt in CNS Procedure 1.13 that a technical evaluation be performed to identify the item's safety functions arxl/or failure modes from

- which critical characteristics could be identified, but rather utilizes a-1 l-stardard, routine receipt inspectica and post-installation tests usually L

-v-i 46

m _.

required uMer mst plant technical specifications, as the mms by which the

. item was accepted for nuclear safety-related service.

Such PITS usually cannot adcquately verify critical characteristics r-mry to-verify the full rarse of design corditions, includire seismic, even_ if such characteristics

-were required to be identified. m is approach relied predominantly on

.qualifyirs the supplier usirq a broad-bascd, programmtic survey, instead of

. performirs a well-focused, critical characteristic-specific survey of the commercial supplier's p m gram controls in place to control selected critical characteristics.- Finally, this approach does not ensure conpliance with the requirements of 10 CPR Part 50 Apperdix B, since the use of a broad-based audit / survey does not verify the ability of the supplier's program to control those critical characteristics necessary for the item to (erform its safety functions.

Another weakness in the CIS dedication process was the failure, in some instances, to identify safety function, cril.ical characteristics, and related acceptance methods in the technical evaluation when using the 03S approach.

Such parameters are required urder GS Engineerirg Procedure 3.22.

Jeneric weaknesses within the dedication process included the failure to ve~ify that the original seismic qualification for replacencnt electriccl and mechanical items was still valid.

If 01S identifies no charges to configuration (form, fit, function, and materials), then it is assumed that the item is identical ard, therefore, that the original seismic qualification has been maintained. As mentioned previously, 01S relied on broad-based pmgrammtic audits / surveys in lieu of a well-focused commercial grade survey, to verify that the supplier has the nc m e ry controls in place to handle charges made in the design, the manufacturity process, ard materials.

Also, important characteristics for greases, lubricants, and oils used in safety-related and environmental qualification applications are not required to be

.identifled and verified per current licensee procedures.

Because such items are classified as nonessential (nonsafety-related) they are not required _to be inspected upon receipt or dedicated in order-to be used in safety-related applications at 01S. - Additionally, traceability: to the Om and similarity to the original environmental qualification test report are not required, witich raised concerns over the suitability of application of these materials.

Another generic weakness concerned specifyi19 PITS.as part of the' verification for critical characteristics without ensurits that.the PIT actually verified the identified critical characteristics. Most of these PITS are routine tests l

used to verify that the item functions normally. W e team also identified several examples in which_ unique nuclear requirements were imposed on suppliers furnishirg items under 01S's E03 procurement classification without

.specifying in the procurement documents that 10 CFR Part 21 applied. This' practice violated Section 8.2.1.2(a). of 01S Procedure 1.13.

In response to the NRC inspection team's identification of these program and implementation deficiencies, the NPPD/01S staff committed during the.

inspection to placing on hold all material associated with approvim tely 212 purchases made since January 1, 1990.

GS will use such material only if the supplier is rcqualified usirg the NUPIC Commercial Grade Items Survey-Gecklist, or if the item is formally dedicated urder 01S Ergineerirg Procedure 3.22, "Conmercial Grade Specification." NPPD also corttitted to

-vi-47

roquality all Ba3 suppliers before January 1, 1993, using the NUPIC checklist.-

NPPD's procurem.nt program enhanoaments emntitted to during the inspection are founti in the appendix to the report.

l

[

t I-

-vil-48

f 1

INITODUCTIG1 During this inspection, the Nuclear Pequlatory Ccanission (NRC) inspection team (team) frcxn the Vendor Inspection Branch (VIB) of the Division of Reactor Inspection and Safeguan3s of the Office of Nuclear Reactor Regulation reviewed the Nebraska Public Power District (NPPD) program ard its implementation for the procurement of commercial grade items (CGIs) used in safety-related applications in the Cooper Nuclear Station (01S). The team also reviewed the NPPD program and its implomntation at CIS for determination or verification of suitability of those CGIs for their intended or approved safety-releced applications, a process referred to as " dedication."

Part 21 of Title 10 of the Code of Fcderal Regulations (10 CFR Part 21) defines dedication as the point at which an item or service becomes a " basic ccuponent," which it defines essentially as items (or servims) with safety-related functions. However,10 CFR Part 21 also defines CGIs (Section 21.3(a)(4)(a-1)), as distirguished from items procured as basic cmponents. The regulation then allows the procurement of items that are to becone basic camponents, but that meet its definition of CGIs, without invoking 10 CFR Part 21 in the procurement documents, hhen 03Is are procured for safety-related service, their procurement and dedication constitute activities affecting quality ard, therefore, these activities must be controlled in accordance with the requirerents of Appendix B, " Quality Assurance Paquiremnts for Nuclear Power Plants," to 10 CFR Part 50.

In particular, Criterion III, " Design Control," ard Criterion VII, " Control of Pun:hased Material, Muipmnt, and Services," of 10 CFR Part 50, Appendix B are rest pertinent to procurement aM dedication of 03Is; therefore, the NPPD program governirq these activities aM the implementation of that program at 04S were reviewed for ccepliance with these (primarily) aM other applicable Apperdix B criteria, as well as with the rcquirements of 10 CFR Part 21.

Acklitionally, the NRC has provided further guidance ard interpretation to amplify and clarify the requirements of Appendix B as they pertain to the procurement and dedication of 031s in NRC Generic Letter (GL) 89-02, "Actionn

'o Improve the Detection of Counterfeit and Fraudulently thrketed Products,"

dated March 21,1989, ard GL 91-05, " Licensee Commrcial-Grade Procurement and Dedication Programs," dated April 9, 1991. Therefore the NPPD CGI procurerent and dedication program ard its implementation were also evaluatcd for incorporation of and consistency with the guidance and NRC staff positions promulgated in these GLs.

Finally, with respect to procurement in general, including procurement and dedication of CGIs, NPPD has ccanitted to various iMustry stardards and other publications (as erdorsed or coMitionally erdorsed by NRC regulatory guides (BCs), NURII;s (NRC documents), ard GIs), as stated in the "NPPD QA Program for Operations" policy document referenced in Appendix D of the NPPD Updated Safety Analysis Report (USAR) for 015, and as expressed for the industry by the Nuclear Management ard Resources Council (NUMARC) in the "NUMARC Initiative on the Dedication of CGIs" (adopted by NIEMIC in May 1989). 49

In Inrticular, IMO, like other nuclear utilities, was acrimitttd to octablich a program for prccurencnt and dcdication of 031s consistent with Electric Ituer Research Institute (DRI) Pqort 142-5652, " Guideline for the Utilization I

of Constercici Grade Items in Nuclear Safety Related Applications (11CIG-07),"

on or tofore Jaratary 1,1990.

h nowptance mthods describcd in HP-5652 were conditic*ully erdorsed b/ the IEC in GL 89-02 and the NRC staff positions on several dodication issues were later clarified in GL 91-05.

Therefore, the team ascessed the dcgroo to which the IE4V CGI prccurenent and dodication program in effect sinou January 1990, and its implementation, were consistent with the portirent industry commitrents, r

2 OlHDtCIAL GRADE DEDICATIOli PROGRAli REVID4

~

2.1 PrcmhtreEtEgy1W

'1ho NPIV program for procurement erd dedication of oils for safety-relatcd applications at clS is described and prescribed in a hierarchy of procedural documentation boginnirg at the 12PD cotrrate level (Columbus General Office, er "030," Nuclear lYuer Croup, or "NP; with IEG Directive 3.13, " Nuclear Procurencat." LEG 3.13 incorportted the NPPD general guidance for 01S procurenent activitics. The team reviewed the currently effective revision of NPG 3.13, Revision 3, cbted February 20, 1990, ard mado the follosirq b

observations:

Urder Section IV, " Responsibilities," the procedure charycd thc, QA Manager, 01S, and the QA Manager, 030 with the responsibility for revicuirg procurencat documents for safety-relatcd materials ard services and for evaluatire suppliers of items for safety-related applications. Paragraph V.K, under Section V, " Requirements," directed that all procurements of essential mterials and services ard mterials requiring cquipent qualification were to be ruvicued ty tha "QA Division" and Paragratt V.L required that all essential procurerents be from suppliers evaluated ard approved by the QA group.

Although it was this procedure in which IETO codified its colicy that the requirements of 10 CFR Part 50 lucrdix B, and the intent of American National Standards Institute (ANSI) Starristd H45.7. ;3-1974, " Quality Assurance Requirements for Control of Procurement of Items and Services for Nuclear lWor Plants," shall be met with respect to procurerent, this prondure did not delinoats any forml policy with respect to primary res;tnsibility for preparire procurement docummts with the etdant technical evaluations and acceptance planniry. Neither did it assign responsibility explicitly for activities relatiry to prccurement ard dedication of 03Is or evaluation of suppliers of 031s. Although a so-called " task force" for 031 prccurerent and doditation had been established, there was no IUT0/OJO-level policy statement that formily recognizM ccencrcial grade procurencnt and dodication ard established a group or assigncd an existing group within the NFPD/Cis staff to be cognizant of CGI procurement ard dodication activities. '1his apparent lack of corporate recognition in official prcccdural docunentation and lack of commitment was reflected in the deficiencies observed by the NRC team in the quality ard inplcmentation of the 03I program and practices at 04S ard in the ur.3atisfactory prcgress mde by CIS in inplementing industry commitments (i.e., !M4 ARC initiatives). 50

'lhe prircipal irplerentin; procedures pertinent to procurcrent ard dcxlication of CGIs were containod in the 01S Cierations tbnual ard covend activities ircltrlity safety classification, surplier evaluation, dcdication, ani receipt ingoction. Volum 1 of the mnual, " Plant Services Prccedures" (PSPs),

contairKd the general prccurcrent quichnoe in PSP 1.4, " General Procurcrent Program." Roooipt inspection activities for all tyres of mterial were cpverncd by PSP 1.5, "Wa.cchotse Receivirg." Procurerent ard dedication of CGIs were covered in PSP 1.13, " Utilization of Essential Ctencrcial Grade Itcro in STfety-Relatecl Applications," ard Erginacrirg Proccdure (EP) 3.22, "Ctxmercial Grade Spccification." System /c x w nt safety classification was corducttd urder QiS EP 3.13, "14uipremt Saf ety Classification," of which the current revluion, Revision 7, datcd June 21, 1990, was reviewcd.

14uipnent ard ccr.porents, that is, itens with unique identifiers a tag numbers called componer.t identification codes (CICs), classified as essential in accordance with EP 3.13, would be put on the Q-lict. aiS EP 3.24, "Part Safety Classifi-cation," ses desigrwxi to docurent the prccess of justifyirg classification of prts of a component differently (i.e., as nonessential) than the prent equipnent, co@olyant or systm; Rcvision 0, October 3,1988, the current revision of EP 3.24, was reviewcd.

14ulgent aid components classifio:1 as noressential in accordaMn with EP 3.13 would not be put on the Q-list.

Reclassificaticqdcurgradiry of components and subocanponents ard parts of essential systcmo.aM ccraponents to noressential status was to be docurentcd on a Safety Classification Osecklist (Attachnent A to EP 3.24).

EP 3.24 assigrKd the CIS sy-@m or design ergineer the responsibility for determinity the safety chtsa of _' i item to be procurcd, ard hence, whether the itcan was to te procurcd urder QA program controls.

'Ihe review of '31ese proccdures is discusstd in greater detail later in this report, but it is appropriate to note here that EP 3.24 establichcd sono important furxhnental concepts in 1988 that appearcd to have toen laryoly supplanted by the current (primrily EERI) dcdicntion rhiloccphy, nanely, that (1) consideration was to be given to pocsible failure modes in determination of p1rt safety function (Paragraph II.B.1.c); (2) the plant applications of prts should not be classi;ied noressential merely because the supplier cannot cmply with 10 CITt Part 21, or is not on the Approvcd Suppliers List (A3L);

(3) the procurcment safety classification is khat is determined by the safety classification, the supplier's QA program /ASL status, aM khether the part must be dedicattd (Paragraph II.B.4); ard (4) critical characteristics are thcce prorcrties or attributes of a part that are essential to the safet/

functica of the prent cxxTonent (Paragraph II.C.3).

While these concepts, if adhered to ard adequately implemented, would have significantly strerythencd the O!S OGI procurcment aM dcdication prcyram, the team notcd that scre were conti vened by other procedures ard prevalent practices, and that when a part was classificd as essential, the prcccdure itself ncgatcd the benefits to the 03I dedication process of the rigorous tcchnical evaluation that would presumbly be corductcd in accotxhnce with its i@lcmentiry provisions (1) by the statement that no classificz.cion oer the proctdure wan requircd khen it was certain that a part was essential (Paragraph II.B.5) aM (2) by the lack of rcquirenents to feed any saf ety functions ard failure modes if thus identified into the dedication process. l 51

___m._ _ - _ -

l'or OlS, there were two rain safety classifications.

Safety-relattd aid envirorwentally qualifini (IV) applicatiors were desigmtal "escential" ard "cenential-IQ" (a cutclass of essential) respa tively.

!Jonmfety-relatcd plant applications were desigmttd "ronesnential," nid anticipatcd transient without neram, radwarte, fire protection, etc., alplications were Intdlcd as renessential.

Ilusever, llPIO choce to use cimi]hr tem to distirguich anong its three tyyrs of procuremnts. 'Ihese categories were (1) " essential" or E-type procuremnts for safety-relatrxl application, prucurcd f rom a curplier who ostensibly mnuiactures/ supplies the item urder a 10 OV Part 50, Apirrdix B program aid accepts 10 QR Part 21 reportiry responsibilities, (2) "escential-ecamercial grade," (IIG) or C-type procurcrents f or safety-relatol'nsential application, Lut procurable as a on (presumbly metiry the 10 cts.'1.3 (a) (4) 's-1) tests) from a Suppliers List (sL)-listed c-type nupplier with an approval roomnercial QA" program, but who does rot accept 10 CHt Part 21, ard (3) "nonecsential" or N-type procurcrents in khich the it,m is not ncessarily intended for an essential application.

Accordirn to c4S procurement cratf, normlly only when necded replacenent

}nrts for essential applications, or new pirts for essential ncdifications were not available from approved E-type surpliers, would cuch parts then ir procurcd as CGIs -(ard presumably only if they not the 10 Cnt Part 21 OGI def n tion) by one of the two a ternato rethcds, llowever, the team actcd that ii l

in practice the apparent prefererce was to attenot first to obtain like-for-like replaecments, rcgardless of the type of procurcrent rcquircd.

t Also noted was the overshcIntiry mjority use of C-type CH procurments (numberity nere than 200) durity-the period of interest, which were sulstantially sinpler than 11-type prcxxtremnts (numtcrity about 20), with their nore traditional (ard much nore detailcd) dcdication rcquirenents.

2.1.13pcggial-Commemial Grade tvdicatio0RM1od Procurements of CGIS as D33 were rcquircd to be conductcd in acco1Tbnce with PSP 1.'13.

She team revicwd Favision 0 of PSP 1.13, dattd April 18, 1990, khich, although not in effect as early as January 1,1990, was the currently effective revision at the tino of this inspection ard was effective for most of the procurements of interest. PSP 1.13 was brief, describiry a four step process for dodication of OGls pitcured as D33.

s

- '1he first step!was-qualifyiry certain suppliers who have no-called ccamcrcial QA programs that either were at one tJm bascd on 10 GT1 Part 50, Aprendix B L and the associated stardards, or whoce G. program resembles /noets the intent of Arperdix B for 01S purpcoes *ut none of shem will accept the reportirn responsibilities of 10 CFR Part

. -Nevertheless, 01S effectively treats these suppliers as -if they were

.ly approved Apperdix B surpliers once they -

have had a satisfactory " commercial atdit" by ois or a third lurty audit in

' accottlaroe with QA Instruction (QAI)-16, " Supplier Approval." '1hese suppliers are then placed on the SL in Section C, ard although restrictions are suppoccd to be placcd on their socie of cupply, in practice their listire in the SL is treated largely as bhtnket authority to purchase any JNns in their prcduct line. Urder PSP 1.13, there is no technical evaltati. process, so no safety

-[g

• 52

l l

functiors or critical characteristics are identificd ard, until about April of 1991, the atdits of them ragpliero consinted primrily of a brcad-inr4d, prcgransnatic audit.

'Ibe scmd stqp n the IIU process involved issuiry the f0.

'Ihe procurenent docunents weru to to revlwod per PSP 1.4 and the revlw was to incitdo a determimtion that the item was, in f at.t a (nI, that is, that it ret the tests in 10 G P Part 21.

Paragraph 8.2.1.2. (r.) stated that the procurcrent docunents should inpr.e the suppliers' IIPfD-approved QA prcgram, tut that the docurents thould mko no refererce to 10 GP Part 50, Apperdi.x B or N4SI fl45.2, or 10 GP Part 21.

As discusscd later in this report, the team noted in revlwirq several IIC procurcments that the los violatcd this requirerent by requirirn a supplier to luve a QA prcgram reeting 10 GR Part 50, A[pendix B, ard NISI N45.2.

'Ihis had the effect of irpirg design

~

ruluircments unique to imC-licemed facilities ard, hence, the 10 was orderity Malc components, not CGIs.

'Ibe third step involved performiry a receipt inspection in accordance with PSP 1.5.

However, despite the heavy reliarce on surplier controls lor quality, the proocdures did rot adoquately address capture ard prcper re f ew of docunents to establich traceability to the origiml cqui rent mnutacturer 1

(oui).

Finally, the fourth step involvcd termal pre-installation testiry as my te required by PSP 1.5, or post mintenance (installation) testirg urder the administrative controls of !binterance Procedure (MP) 7.0.1, " Work Item Trackirg. " Several nther proc (dures were referenccd for testirg associatcd with plant design changes. 110 wever, reviw of numerous C-type dcdications irdicated that the routine testirg usually consisted of simple cparational checks uider nominal conditions md was not alwayr adcquate to ensure performnce ol' safety function, or no failures detrinental to safety, urder all denign conditions. 'Ihe team roted that the vast mjority of CUI procuti nents since January 1,1990 (approximately 212) were IIE per PSP 1.13.

~

conscquently, in c.ffcet, 04S. ras relyirg almost entirely on the C-type suppliers to control the critical characteristics of the items (though not identificd) without actually verifyiry such controls through a proper comnercial grade survey that was item ard critical characteristic-specific.

2.1.2 Dedjcationythcd For CGls Prccured As "Ncngegrttinl" In contrast to these C-type procurements, those (Els needed for essential applications, but that either had already been procurcd as nonessential (no essential use initially identificd), or that ware not available from a C-type supplier, were to be upgradcd/ dedicated (in the more traditional sense) in accordance with EP 3.22, "Ccuercial Grade Specification." 'Ihe team reviewed Revision 3 of EP 3.22, dated Dccomber 28, 1989, which was in effect as of January 1,1990, as well as Revision 4, datcd January 23, 1992, ard the new Revision 5, datcd February 20, 1992, which was currently effective. 'Ihe N-type procurements ard dcdications reviawed were evaltatcd against the revision of EP 3.22 in cffoct at the tine they were prepared. 53

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1he revicu of & 3.22 irdicatcd that the procedure was generally consistent with the pIUvisions of mu llP-5652; altlwugh, the guidance on the principles ard considerations in the process of obtainity critical characteristics, or noro correctly, derivity them frm safety functions ard other essential application suitability requirencnts (e.g., safety affectiry failuro ncdes) was reager.

Instead, cxanples of critical characteristics were given. With recpcct to the definition of critical diaracteristics, the team nottd that the term was definod (in the definition section) as it is in WRI t@-5652, that is, those attributes that provido reasomble assurance that the item receivcd is the item specificd. licuever, the procedural section (8.0) did contain a goad workiry definition of the term that was consistent with the 10C position as expressed in GL 91-05.

Shroo other concem s were identifiod with the IF 3.22 dodication proccas, as written. Urder the description of acocptance nethod 2 (sano as mu nothcd 2), rcgardity camercial grado surveys, there was to procedural reference to other guidance on surveys, that is, QAI-16, ard also the neod to-verify distributors' controls, as applicablo, as stated in GL 89-02, was not addrecsod. Although 013 had a distributor surveillance proceduro, OA Guidelire 3.15, it was not referenced or otherwise ticd in to the dodication process prcgranetically. Firally, the proccdure did not adcquately address capture and review of docunentation to establish traceability of the CGIs as received to their GIMs, which would be recessary for establishily the validity ard applicability c1 verdor controls ard/or verdor-supplied informtion/ documentation, to the extent they are relied upon to suITort dodication (ard/or qualification).

Accordity to the procedure, the QA group was to review escential, LU), ard nonessential (that were to boome essential) procurenent docunents for technical ard quality rcquirements ard the responsibic orgineers were to prepare the CGI acceptance plans (APs) which were to be used to document tha OA crd technical rcquirements for procuriJg ard accepting CGIs ard services as well as the critical characteristics technical evaluation ard pertinent special instructions.

7hc lack of adcquate guidance on derivation of critical characteristics from safety functions, failuro mode informtion, or other essential safety-related application suitability rcquirements, ard the lack of a r(quirement to verify all critical characteristics once they were properly identified, lead to numerous exanples (fcerd by the team during this inspection of CGIs) that were inadcquately dedicated. The irndcquata dodications of these CGIs, which are discussed in detail later in tl.is report, scue of which had been installed, constituted a failure by NPPD to perform ard document an adoquate review for suitability of application and, in scen cases, adcquate design verification

-(seismic /DQ), for items interded-for safety E.ervice, contrary to the requirements of Criterion III of 10 CFR Part 50 Appendix B.. 1he inadequate dodications also. constituted a tailure to verify that the items received net the specifications for their safety-relattd applications contrary to the-requirements of Criterion VII of 10 CFR Part 50 Apperdix B.

2.2 Connercial Grade SurxdigLEurveys NPm Quality Assurance Division QA Instruction No.16 (QAI-16), entiticd

" Supplier Approval," providcd general requircnents for evaluatirg suppliers l 54 V

ard mintemrx:e of the als SL.

'Ibe team reviewcd the currently effective revision of QAI-16, Revision 17, dattd Iby 17, 1991. 'Ihis revision had ircorporattd, amorg other thi ys, nw provisions for cxntercial grade supplier surveys (also temcd "cxntercial aMits" at Q4S), "ocamercial surveillance" or "ocramrcial ocurce surveillance" (04S terro for source verifications), ard directtd the une of the lluclear Procurenent Issues Cantnittee (11UPIC) audit or cancercial grade curvey procedures ard ct.ccklists for atdits or eczmercial surveys by 11WD or IJUPIC joint or neder audits or canticrcial surveys, also allcuirg equivalent dmcklists as approvcd try the QA Supplier Supervisor. 'Ihe

@ Supplier Supervisor at 030 was assigned the responsibility by QAI-16 1or 1rplementirg of the cupplier approval program.

Although the !@lc survey prcmdures were nentioncd (urder Paragraph 3.2.1.6) along with prescribirg the use of the IJUPIC chocklists, it was not clear that the IJUPIC procedures were beiry consistently or uniformly follcwed. Revics of several liUPIC joint ard nember surveys recently conducted (Coltoc-IIPit, September 25-26, 1991; Parker-llannifin-GSU, January 29-30, 1991; ard Woodward Governor-GSU, July 16-18, 1991) revealcd some inconsistency and variability in the implcrentation of the checklists in terno of iten ard critical duracteristic specificity aM level of docunented objective evidence. lior e significantly, there was no docunentxd guidance on hcv or frcin whom to obtain the critical characteristics for the subject 031s, which are necessary in order to cxniuct the r;urvey properly ard with which to fill in the survey plan form in the NUPIC dmcklist.

In response to this concern, ifPPD QA staf f explaincd that normily the QA supplier auditor and the 04S responsible cryineer discussed this informily aM agrc(d upon a set of critical duracteristics that would be listed on the checklist. llevertheless, the team notal that the use of QAI-16 for contrercial grade surveys (ard hence the use of the liUPIC chocklist) was prescribod only by PSP 1.13 for 00G procureecnts and was not ncntioncd urder EPRI liethod 2 in EP 3.22 for nonensential dedications. Ilowever, PSP 1.13 had no 03I technical evaluation requirenents; that is, no rcquirements for identification of safety functions ard from them, derivation of critical characteristics to be used in a survey. 'Iherefore, even if QAI-16 had directed that safety functions ard critical characteristics be formily establishcd thrcugh a dccunented technical evaluation, there was no mechanism in place to do this.

'Ibe team also identified the followiry canocr s.

Section 3.3, " Supplier Reevaluation," of QAI-16 required that surved be ccMuctcd e* least evet'y three years. Also, an antral update was rcquircd that anounted to a brief review (not necessarily requirirg a visit) of durges to the supplier's QA program since the last atdit. Although there was also a provision for audits for cause, the team was concerncd that this my rot be adcquate coverage deperdity on several factors, includirg, but not nccessarily limited to (1) the complexity of the 03I(s) in question, (2) the frequency ard size of purdunes, (3) the critical characteristics to to verified by survey and the extent to which thoce are relied upon to support dedication, (4) the strength of the surplier's controls on design, mterials, mnufacturirq prccesses, ard subcuppliers of parts ard services, ard (5) the streryth of the supplier's ccrunitment/ obligation to either not mke charges in certain prcducts, or at least to inform the customer of any changes rade. 55

2.2.12hird Party Survevn Soction 4.5 of QAI-16 generally addressal audits ard surveys by third parties, such as NUPIC joint or nomber-corducted atdits ard surveys. Paragraph 4.5.1 i

requirtd that third party surveys be mraluated in accordance with itPID requirtments. _ licuever, this instruction did rot limit the time (or circumtances) preocdiry the interdad procuremnt for khich such a survey cculd be considercd valid. Althcu3h specifyiry that third-party surveys wem to be evaluatcd in acconlance with LIPID requiremnts, this procedure did not contain any other guidaroe or acceptaroe criteria for evaluatirg such surveys.

Upcn reviewirn these requirements, the team fourd the evaluation criteria t:s be general in mture and largely slanted tarard broad-bascd prcgramnatic QA atdits. 'lhe proccdure did tot specify survey applicability to the sam (or similar) items bairg prcoured by itPIV, ard there were no rcquiremnts (1) for the third-party survey to have verificd that the supplier had docuented _ and

- of foctively implerented ocumercial quality controls, (2) timt the specific critical characteristics selected by IWID for verification by survey were, in fact, verifird ard documental in the third-party survey, and (3) that both distributor ard mnufacturer controls were veriflod where applicable.

'Ibe NUPIC process provides for canvassing the mcenbership to otsupile a generic set of critical characteristics. 'Ihe survey then neal only verify generically that the supplier has controls for the critical characteristics (within liPPD's programmtic lintits on thooe selectcd for verification an discusscd above) associated with the CGIs in the nupplier's product line (presumbly only thcco of interest to NPPD).

licwever, although this provision my theoretically ensure that the supplier has controls for a given critical characteristic for sorno CGI it can prcduce, it does rot nocmsarily ensure that particular critical characteristic is contro11cd for the CGI being procured ard dodicattd by llPPD. lience, it does not ensure that every critical characteristic (selectcd for verification by survey) of each CGI to be dodicatcJ by NPPD Will be controlled. 'Ihc team was concernod that such a survey might verify that e

the supplier controls a given critical characteristic, but not newrily for the Cul of interest. 'Ihe team also noted that the procedure did rot provide for surveys of distributors as well as manufacturers, where applicable, as discusscd in GL 09-02.

Although, QA Guideline 3.15, " Distributor Surveillanca," had been written, it had rot been formally tied into the process.

In order to as-tho offuctiveness of the iqplementation of NPPD's nammercial grade survey program in support of dodication, the team also reviewcd a number of cxxqpleted survey reports associated with same _of the irdiviuual dedication packages reviewcd.

Any such surveys thus evaluated are discussed later in this report in conjunction with the discussion of the associated dcdication.-

2.2.2 Source Verifications NPPD's CGI dodication procedures providcd for acceptance of CGIs through source verifications (EPRI Methcd 3). Accordingly, the team revicvod the NPPD proccdure governing this method, QAP m 0.01, "fource Inspections." "he team reviewcd the currently effective revision of N10.01, Revision 18, da ed 56

i I

l t

Septeniber 9,1991. She team fourd tlat this prcxxdure providcd accey

'e guidance for the performroe of sauron verifications aM did specify verification of critical characte11stics. The only weakness identiflod w,a that the detailed instmetions for the inspection report, Attachnent 2, callcd for a "rnrrative sumary of inspection activities," lut did not specifically requiro tint the pirticular critical characteristics be listed ard that their method of control ard verification and recults be documentcd to provide docunentcd objective evidence of those critical chracteristics.

2.3 Egte Clannificat190

.Ihe team reviewo:3 the proocss used at 01S to determine the safety classification of irdividual component parts ard determined it was adacpate.

1ho process offercd the tanic guidance rcquired for classifyirn parts.

Irdustry-roccgnized documents were used as references in the review of the spire parts safety classification program. 1he mothcdology uscd for classifyiry parts ard sub scamblies of essential camponents was described in OfS EP 3.24, Revision O.

The proceduro providad a mthcd of documentirq the basis.of coch part classification usity the Safoty Classification diecklist (Attachnent A to the procedure). The responsible system ergineer was rcquircd to use the checklist, which consisted of a series of questions to be miswered mechani'al, electrical, yes or no._1here were four categories of questions:

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instnmentation ard controls, and structural reviews. The procedure also gave instructions to ensure that the spue parts inventory list for equip:ent was revised if a part was actually reclassiflod.

'Ibe NRC inspection team determined that lack of docunentation was the primry procedural weakness. Only part number, mnuf acturer/ supplier, the applicable camponent identification code number, basis for evaluation results, ard references had to be doctmented.. A section for notes was also providcd.

She procedure did not require the followiry specific itams to be documentcd:

(1) safety function of the parent camponent, (2) technical evaluations, l

includity the failure modes ard effects analysis,- (3) safety function modes, either active or passive, of the part and parent ccuponent, and (4) the nano of the part. The procedure instnacted the person usirg the checklist to completa the review and aralysis, ht didn't actually require that the review i

l be docunented. - Docunentation was missiry in soveral ocupleted chocklists that were reviewed.

NPPD had performed 26 reclassification reviews as of the date of the inspection. The team reviewod five of NPFD's reclassification reviews for technical adequacy anel adherence to adn.inistrative requirements.

Of the five checklists reviewod, all were technically adcq2 ate. The basis for reclassification was marginal in several cases. several-basis statements were actually sunmary statements, and were tot subatantiated with dcctmnntcd results. There were no irdications that certain attributes important to the part, such as part material or seismic requirenents, had even been considered.

One possible weakness was observed during the checklist reviews.

Dawrgrading a part from safety-related to nonsafety-related could lead to uncontrolled

-changes in part m terial. The formal-review process c,f the checklists was inconsistent because the " reviewed by" signature blank was missire from the

_9 57-

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i checklists. Of the five chccklistr s viewed, one was missiry the nnamry form 1 review. We five safety el fication checklists (Attachment A to EP 3.24) reviewod durity the inspr i are diroW in the psragraphs that follcu.

• Q1ggMist 0-90-01 Valve IRdy-to-Bonnet Gasket 1ho safety classification M the body-to-bonnet gas.ket of a primry l

containment isolation valve was domyraded in cafety classification because the gasket was not required for mintainirg the pressure bourdary. Win was the only checklist exanple of mterial properties leirg considered ard uocumented in the basis ocction.

• Q1cgk11st 0-91-03 Barton Differential Pressure Switch Gasket

%e gasket function was to provide a seal to ensure moisture, fumes, ard dust did not enter the iMicator through the faceplate bezel. Win gasket was downgraded on the bac a of envirornintal qualification data that irdicated the gasket did.ot have to ptuvide a leak-tight seal from steam intrusion. Win chocklist applied to approxistely 30 instruments. We most severe service rcquirement for all applications rhould have been notcd and considered in the analysis. 'ahe instruments were Da to withstard the effects of exposure to high radiation. W e effect of sut exposure on the gasket was not documented in the basis sections.

• Q1gg}: list 0-91-06. liigh-Prescure Coolant Injection (IIPCI) lubricatirg Oil Filter Element We lube o!! filter is a passive component that filters out contaminants jn the IIPCI Lurbine lubricatirn ard control oil systems. W e justification for reclassifyirg this element was bascd on ccuponent design.

Internal bypass valves would allcw flow to bypass a filter clogged with debris. We amlysis only considered one failure node (blockage). Other failure modes, such as crosion, corrosion, and loss of material properties, were apparently not considered.

• Checklist 0-91-10 Air IIandling Unit Cooliry Coil Gasket 2

We purpose of the part was to mintain pressure integrity (a leak-tight i

scal)- to prevent loss of cooliry water fluid. W e basis statement for the reclassification was: "a leakirn gasket on ooil cover will not preclude coil frcu performing its safety function, i.e., cooling of air." We' team considered this a summary statement ard not an adcquate insis for reclassifying the gasket. W is statement was not. supported with documented analyals. No failure modes were listed; these could have-included crackirg, embritticurent, loss of seal,- loss of resiliency, or loss of material properties. We references cited were margimlly acceptable because only the vendor mnual was listed. We verdor mnual offered very little informtion about the gasket.. Checklist Q-91-10 was considered only mrginally technically adequate. 58

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• Cleghlist O-91-11, Breaker Ctotrol Relay

%c put was a 125-Vdc control relay for the power supply breakers for the station air crrpressors.

Failure of the p1rt was detennincd to inve no safety effect; hoever, a loes of station air crrpressors would inhibit plant rwovery follwing an accident condition. Of the five checklists reviwod, 0)ocklist 0-91-11 documented the technical evaltation best. We perton who prcpired it exceedcd the mininum levels of docunentation establithed by the procedure.

We reviw of G 3.24 identifiod several other areas in which procedares needed to be improved.

For exanple, the definition of critical characteristics listed in Section 2.12.9 of & 3.22, Revision 5, was different frcxn the definition of critical characteristic in W 3.24, Step II.C.3.

We wonling of the definition in W 3.24 agrood with NRC GL 91-05 recomicidations; the definition in & 3.22 did not. An NPPD representative stated that the incorrect won 11ng of the definition of critical characteristics in D 3.22 was an NPPD oversight, and that the step rcquired revision. Also, the ?.ttachment A structural reviw section was missirg a critical yes/no question about the put's effect on fuel novenent. Of the checklists reviewcd, this question was not applicabib to the function of the puts or parent camponents. Finally, the procedure did not clearly require a written basis if all the criteria questions in a subocction were answered no; therefore, the basis sections of 8

several chocklists were simply left blar.k.

We methcd to classify generic pirts for such items as gaskets, o-rirgo, and valve packityJ was not proceduralizcd at 01S. Generic parts have not bcen classificd to chte, but are evaluated cc a care-by-case basis.

If generic ptrts are to be classificd, prcmdures have to be revised.

In conclusion, the ord result (reclassification of parts to ronsafety-related) appeared acceptable. Domacntation of the basis for the reclassification was narginal in several cases. Also, there was no indication that certain attritutes ivortant to the part had been considercd in the arnlysis.

2.4 Trendim of AtIgliern

%e failures of 03Is installed in safety-related applications at 01S are trended according to existirg plant procedures. Wese failures are tracked using the nonconfornunce report (NCR) process per 03S Procedure 0.5.1, Revision 6, "Nonconfornance and Corrective Action," which establishes measures to ensure corditions adverse to plant safety ard reliability are promptly identified and corrected. NCRs are written for equiprnt failures, malfunctions, deficiencle.7, deviations, defective materials, and other sintilar nonconformirg corditions. NCR dispositions normally in mudcd root-cause evaluations, a reviw of pmvious !!CRs and equipent history to identify repetitive occurrences or adverse trends, and corrective actions to prevent u.CurrenCo.

Plant procedures do not require that Ioceipt inspection failures te tracked.

01S PSP 1.5, Revision 12, Step 8.2.7, stated, " Vendor shipped materials identificd as nonconforming or defective in accordance with purchase order 59

criteria or cpocifications chall be relocatcd to a designatcd warehouse

'RPJII'r' area. An fiCR, for 10 CFR Part 21 reportability, is not rcquircd to identify the situation or the return of an item to the veMor." Hoseser, the roccipt inspectors are informlly questioned duricg the annual supplier update evaluation. Probl es encountered with the supplier or the determination of any nonaanpliances nottd would then be discussed vd docuxnted. 031 acceptance test failures were also not required by the procedure to be tracked.

Stcp 8.3.1 of EP 3.22, Revision 5, stated, "In the event that the copnt fails the specified inspection and/or test, the perforter will notify the responsible cryineer for dicposition." 'Ihe procedure dic' not discuss ti.a disposition procem further.

In conclusion, failures of installed U;Is at CIS are trendcd accordirg to existiro plant procedures. Receipt inspecticns ard acceptance test failures

~

for 03Is not install.cd in the plant were not formlly trendcd. Trendire of all 031 failures before parts are installed would give NPPD valuable data which should be includcd in NPPD's annual update of its supplier survey.

These data could be uscd to flag suppliecs of cxamponents that fail at a higher than norml rate ard could be used to feed informtion back into the 03I dedication process.

2.5 potection of t'raudulent tDterials Fraudulent mtcrials at OlS were dettettd primrily by the physical receipt inspectors; these inspecten were qualificd in accxardance with Training Program Description (TPD) 0315, " Physical Rooelpt Inspector," Revision 0, February 11, 1992. 'lhe licensee had documentcd that all receipt inspectors had completed I.esson File No. SKLD33-02-01, " Physical Receipt Inspection,"

Revision 1, ley 6,1991, a requircment of TPD 0515.Section IV of the lesson file, " Identification of Fraudulent !bteriale," discussed NRC GL 89-02, NRC Information Notices 89-70 and 91

1, general information on nonconforming molded case circuit breakers (MCCBs), ard specific information on General Electric, Westirghouse, Siemens-Gould-ITE, and Equare D MCCBs.

Physical receipt irgoction was performed in accordance with PSP 1.5, Revision 12, October 10, 1991. Paragraph 8.2.1.13 of PSP 1.5 required inspection of items such as valves, circuit breakers, ard mtor control centers, for fraudulent or substandard materials or materials that had been tampercd with.

FiniinJs were required to be documented on an attachment to the receipt inspection report. 'Ihe NRC inspection team discussed the detection of fraudulent materials with a receipL inspector who demonstrated familiarity with the mthods.

It was noted that the area in which the receipt inspectors performed their work had a latxje posting of informtion and methcds related to detection of fraudulent m terials. The inspection team mncludcd that personnel responsible for detectiry fraudulent materials were adequately trained and had properly implementtd the appropriate tuthods.

3 RDICATION PAGEE P1VIDJ

'Ib facilitate the NRC review of irdividual dadications, NPPD prepared (at the NRC's request) a number of dedication record review files, compiled from 60

diverse records, but cadt pertainirg to one dcdication, as selected by the team frcn its review of the 01S dodication file lis,ts. The review pad. ages were organized by discipline into electrical ard instmnentation, mocnanical, ard mterials (irclinirg lubricants).

In addition, NPPD providad the associated cumercial atdit or camercial grade survey reports in separate files. '1he team reviewed the available rccords for the selected dedications, includity 100, involoes, receivity reports, receipt inspcction reports, dedication acceptance plans / records, mintemnce work requests (!Ms), and qualification reports.

3.1 Item PurcMned as Esrq1t;1aL-Qomercial Grado (IE) 01S PSP 1.13 was the docunent establichod to control the activitics anscciated with the procurement ard use of 031s for safety-related applications. 1he procedure required that D33 items be procured from a emrce whose approved commercial quality prcgram had been invoked durirg the mnufacture of these ituis ard prohibits reference of unique nuclear requjrements in the procurement doctments, to Appendix B to 10 Clm Part 50, ANSI N45.2, or 10 Cm Part 21.

Ilowever, the atdits perforined were generally prcgrammatic in nature ard did not verify the styplier's ability to cx>ntrol the critical characteristics. Additionally, critical characteristics ard safety functions were not required to be identified uider this approach. A doctmentcd receipt inspection is performed upon receipt, and itema subsequently issued for maintemnce purposes were to receive either pre-installation tests or post-maintenance tests, while items issued for design change or equipment specificatic>n charges were to receive acceptance testirg. The procedure stated that dedication occurrcd at the time the item was placed in cervice followirg acceptance. The followiry exarples are items that were purchascd as ECG ard either installed or made available for installation in essential (10 CFR Part 50, AppeJdix B) applications at CIS without beitg adcquately reviewed for suitability.

a.

10 change authorization 342961 ordered 16 Belleville sprirq Warhers fram Georgia power Company (GPC), Plant Hatch, on January 17, 1991.

The 10 to GPC did not require certification that items surplied by Dresser-Rard to GPC were the same items GPU supplied to NPPD. The package contained a copy of the 10 frcu GPC to Dresser-Rand for 100 washers, certification from Dresser-Rand to GIC certifyire part nicber ard that it was supplied to 01S as a 031. Plant Hatch was placed on 01S's SL as a distributor for Dresser-Raid on the basis of a phone interview on October 30, 1991, ard an NRC Essessment of Hatch dated

.May 3, 1991, that was critical of GPC's extensive reliance on broad-based, programmatic audits to qualify commercial grade suppliers.

Dresser-Rard appeared on the 01S SL as an ECE supplier. Durity the inspection, the NRC team found that the washers had been installcd in the HPCI turbine. 01S performed an operability evaluation to deternine suitability and inforned the team that the spring washers muld ba dowrgraded to rt;nsafety-related. Dresser-Rard participated with NPPD in this decision, b.

10 311091, January 8, 1990, to Dale Electronics, purchased various vire-wound resistors mnufactured to Military Specification MIL R-26E l

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l ard installed in the PKIS atg.rentaticn }hase Il system located in the control rocn. We PO rerpircd Ihle Electronics to provide hst Report No. 26080, January 16, 1990, which docurented the testirg of a nonthly sanple of these resistors. We test report covercd resistax.e, tolerance, tharval shcck, short-tire overicnd, resistance tenperature coefficient, dielectric withstard veitsgo, insulation resistarce, high-tenperature exposure, roisture resistance, ard Icw-tenperature storage.

01S cpalified tule Electronics as an IIn supplier on the basis of a ocmmercial grade surwy (SA 90-48) perfomad nine months af tu thc 10 was placxd (October 25-26, 1990). W e survey was progranratic and referenood Mile-I-45208, MIIrs'ID-45622, MIIr6Te105, and MIIrQ-9858.

We Imc inspection team's review of the ccamercial grade survey fourd that, although not rcquired by ots procedures, the survey addressed critical characteristics such as resistarco, tolerance, power rating, dielectric strength, ard seismic qualification of the resistors. A technical evaluation had not been performed to identify safety function or critical characteristics.

c.

TO 311709 was for thirty-eight 150-pourd gaskets that were procured on January 17, 1990, frcn the Flexitallic Gasket Carpany, Pennsauken, New Jersey. Flexitallic (Pennsauken) appeared on the Q1S SL as an DOG suppl.4.r. NPPD's Quality Assurance Follow-up Checklist, Audit SAB8-31, was a fol w2p to Atdit S87-36 shich identified prob 1 cms with Flexitall, supplier list and procedure revision control. W e audits were prograntatic in nature and did not verify thu supplier's ability to control specific critical characteristics. Several other Flexitallic orders reviewed which containcd similar deficiercies were IOs 311015 ard 310601.

d.

TO 312069, January.4,1990, was for Autcatatic Switch Company (ASOD)

HB8320A90 " Rod Ibt," solenoid operated valves (SOVs), procured frca the John Day Caquny of Omha. We file contained an ASCD drcp-chip packing slip, February 14, 1990, ard a receipt inspection report (RIR),

i February 27, 1990. We fornut and usage practice of the RIR left the applicability of attributes up to the receipt inspector instead of beirg pre-approved. Consequently all testing blocks, ircluding post i

installation _ tests, were marked ret applicable. Upon ingairy by the team, CIS determined that the SOV frcn this PO was in the warehouse and available for installation in essential applicauions (e.g., SW-SOV-SPV857). %e dedication of these OGIs was inadcqtate in that it was based on an Appeniix B audit of ASCO that focused on ASCO's dodication of parts for its nuclear, D2, Appendix B QA s ufram-manufactured, catalog NP-1 line of SOVs. However, the ASCO " Pad Hat" SOV is not unnufactured to this program tut to their etnmercial quality program ard cuntrols. 'Ihe atdit was not a prger ocumercial grade survey, since it was not item specific, ard the critical characteristics were not verified for_the itens purchased. Note that 10 312069 was'not listed as beirq put on hold for this reason in NPFO Mercrandum Q?M 100010, Janutry 7,1991, as were three other siJttilar IOs (327193, 315840, ard 312427). 61 a

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e.

IOs 315840 and 327193, also issued to Jchn thy Ompany, for ASCO SOVs ard rebuild kits tad the sano problems as 10 312069, but iters fran this 10 were captured ard put on QA hold accordity to NPID QAD Manorardum QAD9100010, dated Jarnlary 7,1991, listiry ASOC SOVs ard rebuild kits fran 100 372193, 315840, ard 312427. Note that this uso did not list 10 312009, khich was apparently overlockcd. CIS raported that the item was in the warehouse, and was available for issue for such essential application as SW-sov-SIve57, f.

10 312109, January 18, 1990, was issucd to SW-Nelson, Weldirg Division for 100 Nelson stuis, 1/2" x 2", P/N 101-017-315. The lo spccified that the supplier was to provide ocrtified mterial test reports (Q41Rs) attestin) to omplDince wita Ancrican Weldity Society (AWS)

Oode D1.1-1985. The PO, contrary to pro:xxtural rcquirenents, also required the supplier to mintain and apply a prcgram that was in accordance with those applicable portions of 10 cm Part 50, Appendix B ard ANSI N45.2, thus invokirn unique nuclear rcquircrents in a camnercial grade purchase. '1his practice violated Q1S proccdure 1.13, Section 8.2.1.2(a). 'IW-Nelsen supplied the studs ard a prcduct certification dated Fehwiry 8,1990, which attested to compliance with AWS, AS'1H, Apperdix B, hrd MISI N45.2, without TW-Nelson ocarnittirri to inplementing an Apperdix B QA program. The ce"tification also proir. dea the mtcrial grade, heat rnInber, chemistn analysis, aid physical prcperties. Receipt inspection consisted of verification of quantity, olnious physical damge, ard review of TW-Nelson's product certification. There was no evidence that any special tests had been performed.

'Ihe 04S SL shcsod that TW had been approvcd as a supplier of IIG products only. The triennial evaluation (NPPD Audit FA89-27) was perforrmd on May 29, 1989, usity a CASE (Coordinated Agency Supplier Evaluation)-Nuclear Section cha:klist. A review of the complettd checklist indicated that the evaluation was programnntic ard did not address the cupplier's ability to control specific critical characteristics. Annual supplier evaluations were perfouned on i

November 8,1990, ard April 29, 1991.

g.

PO 311631, January 16, 1990, was issued to Cocaha Valve ard Fittirq Co.

for 14 valve bonr.ets (3 half-inch and 11 threo-eights-inch Union Bonnets) ard certifications. 'Ihc PO specified the mnufacturer as k

Whitey Ompany, and indicatcd the tort numlnrs for each of the two sizes of valve bonnets. The Po, contrary to procedural raluirements, specifled that the supplier was to mintnin and apply a program that was in compliance with the applicable portions of 10 CFR Part 50 Appendix B *M ANSI N45.2.

7*c PO also stated that all require:rnts were to to

'ttcd ard i paced on any marufacturer or subtier suppliers 11..

in the nenufacture of the ccrnponents.

The 14 valve lxannets were received ard : eceipt inspected on January 29, 1990. 'Ihe RIR showed that the technical data we-e reviewcd ard approved on February 8, 1990; that data consistcd of two certificates of ccepliance fran Whitey Company, dated January 19, 199C. 61 b

l 1he oc.rtificates rhcwcd the apprispriate quantities ard prt numbers, ard stated that the Type 316 stainless steel usM to mnuf acture the bonnets had Lun purctused ard certified as beirq in acconlance with unterial specificadoru ASD4 A-479 aid A-262, ard that the tonnets had been testod aM packagd in accordance with WS-22 ard WS-23, recroctively. We procurencnt package also containod warehouse issue and return tickets which showed that at least seven of the valve bonnets had bmn installed in cafety-related applications, usiry INR 90-1179. The lER also chcued that all work was ocrpleted and the equipmnt was doclared ready for service on April 30, 1990.

We 01S SL chowcd thLt Waitry Campny had been approved for the procurement of EOG itena oi:1y. The approval bocaro effective on October 30, 1989, and was besod on a triennial evaluation (liPIO QA Atnit SA89-20 perforced lia) 23-24, 1989, usiM a prcgranratic CASE-fluclear S(ction checklist.

Ihe procurement package did not cor hain any arinual supplier evaluations of hhitey Ccxtpany.

It should be rotcd that program implezentation effcctiveness was not reviewcd during the inspection.

h.

10 329'66, Jantury 31, 1991, through TO change authorization E, J

Juif 22, 1991, was issued to IM/IP Intermtiornl, Incorporatr_d, for six rhaft couplim s P/11 7002355, Drawirq No. IF-6921, made from AS1M A-479 Type 410 Class 2 noterial. % e E0 stipulated that all work was to be performed at the Verron, California, facility and that the supplier nust inpoco the qtnlity prcgram that had been previously approved by NPPD. A certificate of annformnoe was also required to be cut 2nitted.

We team roted that IM/IP's Ptmp Division was listcd on the 01S SL as an essential supplier. S crefore, procedurally, an DOG PO should not lave been insucd. We couplings were recei.vod, rcceipt inspected, ard approved on July 22, 1991. The receipt inspcction consistcd of identifying the coupliins, verifyirg their characteristics ard quantity, and approviry the supplier's documentation. We certificate of conformance, revised on July 22, 1991, attested that the parts met all of the requirements of the f0.

In acklition, the certificate provided such informtion as part ntmber, drawing number, and material type, eM stated that the parts were producM under FM/IP's QA Program Rinual (2rd edition, Revision 2), July 18, 1990.

The procurement package contained INR 90-4017, kbich rhowed that at least two shaft couplirgs had been installed in service water pump 1C on October 19, 1991, aM that subsequent post-mintenance testing had been perforwd. We equirnent was declared ready for service on November 5, 1991. At the tire the Po was issued, IM/IP had been apptoved as an essential supplier based on a NUPIC audit (AG89-018) corducted betwoon August 14 ard 18,196 -

%e audit was a joint nuclear utility QA atdit in whic Union Electric Canpany lud the lead.

The atdit verifiod tlut IM/IP's QA program was based on 10 CFR Part 50 Appendix B, ANSI N45.2, ANSI /ASME !QA-1 (1986), ard 10 CFR Part 21.

%e NRC inspection team's general review of this audit shcwed it was nore carprehensive ard performnce-bascd than the others that it had reviewtd ainilarly. 62

'Ihe team expresscd canonin rcqardire the afp'trent lad of traceability that would provide assurarm tlat the counllins tud been it.inufactund urder the m progren that flUPIC had approvtx1. Further, there vere no OUPs frm the origiml mt&lal mnufacturer to confirn the actual mterial uncd in the mnufacture. As a result of telephone ocrunnication between flPID and IM/IP,1 axes were received on February 27, ibl, attestily to the use of the m prcgram that !ad teen approved as mcotiry the requireents of Apperdix 8 to 10 QR Part 50 ard that the mterial had lxen procenrAd in accordarce with tk rrquircr:ents of 10 CFR Part 21.

Acklitiornlly, it was attestrd that the mterial was surpliert to IM/IP urrler an approvcd @ program f rom the uterial supplier, thrle M. Jorgensen. A fax of a OUR, February 27, 1990, provided the informtion raxled f or establishirg mtarial traceability.

i.

IO 329844, was for four 2-inch elbcm, purchancd frcin Cocpar EnenJy Servioos (CES) on February 1, 1991, stich wra installed on emrgency diesel generator (EIXi) 130. 1 durity the rcantly ccrpletcd ref uelire outage under }MR 91-1892. 'Iha 10 rcquested a certificate of conformnce that the parts were egaal to or better than those origimlly surplied (1964) and Jarxced CES's m program as previously approved by NPID. The hisir. of flacity CES on the SL consistcd of a prcyramp.ic-type m atdit which did not verify CES's ability to control cpecific critical characteristics.

j.

IO 326792, was for nine Viton fuel oil filter gaskets, purchascd fram CES on Ikwember 11, 1990, which were installed on IDG IJo.1 durity the recently otoplcted refusi.ity cutage urder MWR 91-1658.

Dodication conciatix! only of performiry stardard past mintenance ard inservice p

leak tests in accordarce with CIS's proondures.

h k.

10 312074, was for 12 gaskets, Inrchased f rom CES on January 18, 1990, which were insta11cd on intercooler pipiry for Em tio. 2.

015 relied upon a certificata of conformnco ctatiry the parts are cqual to or better than those origitally turplied; however, CES classifled the parts as roncritical. The lo invokcd such unique nuclear requirements as 10 C1m 50 Apperdix B ard At1S1 li45.2, which were not appropriate for these Em items. O1S perfortwxl n stafriant inservice leak test Ixtr CIS Procedure 7.0.8.1 ard accepted a certificate of conformnce from CES l

which was based on a brcad-based programmatic QA audit perforwd by IIPID in January 1988. Receipt inspection, per the EE rethcd, only consistcd of a visual inspection ard no critical characteristics were required to ba identified or verified.

Impusirq unique nuclear rcquirenents violated Section 8.T.1.2(a) of 01S Prccedure 1.13.

Other 100 reviewtxl which had the same deficiencies were Pos 322798 ard 336439.

A surveillance of CCS vas perf0rmed by GiS in Septertber 1991 (SS91-57) to evaluate current NPFU fos with regan1 to the newly revised tethcd of certification to CES's procedures only. The report statcd that af ter July 1,1991, CES will no lorger certify that the items supplied are 63

(qual to or Mtter than thcce origimlly supplied, tut will only certify that the itenn are considered as 0319 ard as such, CES mkes no clairo to form, fit, or function.

01S reviesed approximtely 10 f03 dated betwwn Januar/ 1, ard July 1,1991, ard oo"cluded that no problem exhtcd, airce CES's m program, applied to these orders at the tine, was.vnsistent with current industrf procurement practice. As a result, scrc parts procurcd throgh these 103 have been installcd in the plant and othern are precently available in the warehouse.

3.2 ltrm_Lmpingd as lionerwathl I>:anples follcu of items that were purchastd as nonoscential and were either instalicd or made available for installation in essential 00 CIR Part 50, Apperdix B) ar.plications at OJS without the performance of an adequate review for suitability. QJS EP 3.22, "Cancrcial Grade Specif: ation," was the document used to control the dcdication activities assoc.ated with these iteca. The prcxxdure requirod a 03I dedication package to be prepared that ccr.sisttd of a 031 technical evaluation ard an acceptan plan (AP).

a.

Dedication Package 90-031, 10 177637, datad November 21, 1990, was for a General Elcctric (GE) G29400310 circuit breaker enclosure rackdcun interlock switch, that was procured as a nonessential item from Cencral Electric Surply ampany, Omha, Nebraska, dodicatcd unior AP 90-031, ard installed urder !M 90-1617, dated Mrch 26, 1990, in C. Tass 1E, 4160-Vac circuit breaker EE-CB-4160G asscciattd with the core spray pmp. The dealcation package was placcd on hold for engineering review on January 27, 1992, as a result of deficiencies identified in NPPD interml QA Audit SG90-1400Ir24 (for example, seismic qualification not addressed in the technical evaluation perfonTd per EP 3.22). 1he O1S response to the audit firdiry stated tnat seicmic qualification was covered by the like-for-like determination based on sane part number and visual camparison. In response to the lac concern about the adequacy of the G S response on this issue, 01S prepared a menorandum to the 90-031 file containirq a scismic qualification justification statenent. The bases for the critical characteristics were not clear frcan the technical evaluation. 'Ihis had also been identificd in the internal QA audit, although the charactcristics listed app 2ared technically sound. The team also identiflod scme additional concerns.

'1he tochnical evaluation list of critical characteristics included opcn and closcd contact resistance (>10 mogcb and <1 chm, respectively) separately and then listed them again as evidence -

satisfactory operation, int with "nnar infinity" listed as the acceptance criterion i

for open contact resistance. 'Ibe technical evaluation also listed switch terninal-to-frame (lasulation) resistance as a critical charac-teristic. 7b^ AP ca)?ed for bench testirq accordirgly, except that for the operational test, the acceptance criterion for open contact resis-tance was >1 magohm (not >10 magcfr or near infinity). Uhc AP, appropriately, also callcd for checkiry the termiml-to-grourd resistance as a post-installation test.

Hcuever, the quality control testirg checklist attached to 14R 90-1617, urder which the switch was installed, stated that no test was rcquirtx1, tud the post-mintenance testing checksheet listcd only " verify proper operation." Hence, there 1 l

l 64

was no doctmented objective evidence that the insulation resistance-to-grourd test was ever prformod. The inspection team noted that this was not one of the intermi atdit firdirgs.

b.

- Dedication Package 90-032 consisttd of six hirge pins for Anchor-Darlirn (A/D) tilt disc check valves. The IO, dated June _16, 1989, spucificd pins with an outside diameter of 1.992 inches in the btmhirn area totween steps for an 18-irrh check valve. The 10 was nonsafety-relattd ard A/D appeared on t'e Ois SL as an approved Appendix B supplier, qualified by a NUPIC Apperdix B audit. The 10 also specified that the valves were to be mde to A/D Part No. 764-3D-5 and Drawing Noo. 764-3 (Revision DS) ard 920-3.

The 03I technical evaluation revealed that four of the six pins were to t

be used in safety-related reactor fecxtaater check valves RF-CV-13CV, 14cV,15CV, ard 16CV, which act as isolation valves ard are locatcd inboard and outi;oard of containment on lines A ard B.

The technical evaluation included the items' end use, camponent saf ety function and envirorment, design criteria with mnufacturer's dercription and design code, critical characteristics (outside diameter afd mterial), ard acceptance criteria (part number, outsido diameter, ard local leak rate test). Docunented in the dodication package were outside diamter measurements for two of the four safety-related valves. The remining two pins were to be used in nonsafety-related foodwater check valves (RF-CV-10CV aid 11CV).

Another review IG'ntified A/D Drawirq No. 920-3, which specified a pin diameter of 1.975 inches ard A276 Type 410 stainless steel pins for U-CV-13CV,14CV,15CV and 16CV check valves. A/D Drawing No. 764-3D-5 specified a pin diancter of 1.992 inches ard A582 Type 416 stainless steel for the nonsafety-related check valves. A previous discussion between an A/D design ergineer ard a 01S enployee allowed the substitution of t' e 1.992-inch A582 Type 416 pins (designed for the nonsafety-related valves) for the 1.975-inch A276 Type 410 pins (designed for the safety-1 4 ted valves).

The pins were dedicated by measuremnt of the outside diameter, verification of part nunber, ard an ASME Section XI 1MA 521(e),1980, visual inspection of valve bonnet leakage. This was perforned under M4R 90-0525, Test 90-111, on May 5,1990,_ for the 15CV valvo, and under MiR 90-0524, Test 90-110, on April _ 21, 1990, for the 16CV valve. A deficiency noted durity the review was the aboence of material t

' verification or documentation addressity seismic qualification of the entire aremhly, given the fact that the newly. installed hinge pin was not identical to the pin replaced.

In addition, the original PO was for nonessential items despite A/D's classification as an essential supplier on the OJS SL.

Since the pins werc _ installed in four safety-related feedurater check valves without. proper dedication, NPPD performed an operability evaluation durirg the latter part of the inspection. As a result, A/D ocumtitted to providing materici certifications for the six h?.nge pins, although they had been originally ordered as nonsafety-relattd, by early March 1992. 65

3.3 1&dgante ard Fluid?_UmLJnEpentJal_mr1E9r/intial-IE14plicaltione

'lhe team selected several lubricant ard fluid procuresents for review Irm those listtd for essential and essential-ID alplicatiors on the CIS lbster lubricant List, dated February 9,1991.

Certain fluids (e.g., diesel fuel, snubber hydraulic fluid, ard min steam isolation valve (MSIV) stem ard guide lubricants) were belig procurcd as essential or IIU, but other lubricants for 01S had teen, ard were tchy procured as mnesser' tal, ard without dcdica. ion urder IF 3.22.

Although Cis had a program for routine, in-service lui.ricant samplirn (e.g., for breakdwn or excessive bearirg wear), to acceptarce testirn was required or was beirq rerforud. 'Ihc follcuirg exanples are items that were purhastd as nonessential ard either were used in safety-related (essential) equipent at CIS (includirg equipr:ent urder the IQ rule), or were mde availab'le for use in such equipent, without beirg reviewd for I

suitability.

a.

10 326028, llovember 15, 1990, for DAG 156 lubricant, was procured as nonossential from Acheson Colloids conpany, Yansas City, Missouri, for i

use on essential MSIV stem (urder MWR 90-3914), guido rods, ard internal threads. The file contairxd no invoice or packiry slip ard j

the warehouse " pick" ticket in the file referenced l'O 287238 instead of 326023.

(Despite an irquiry, OJS never explairxd this discrepancy durity the inspection.) 'Ihere was no receipt inspection or testing record. Scite of this type of lubricant w;s used on MSIVs urder MWR 90-3914. There was to evaluation evident of the campatibility of this mterial with the Vers 11ube uscd on the MSIV pneumtic actuator o-rirns, or with the o-rirys themselves, er with clastcmers uscd in the MSIVs' ASCD SOVs which cxxild be exposed to this mterial as air from the actuatirg cylirders was exhausted through the SOVs durire the MSIV closiro cycle.

b.

10 ~ 34 6760,- January 17, 1992, was for Mobil UTE 797 oil, procured as nonessential from Allied oil & Supp'ly ocmpany. 'Ihis oil is uscd in such essential-ID applications as the core spray pump motor bearing.

'Ihera was to dccunnnted technical evaluation or acceptance plan, so no Gafety functions or critical characteristics were identified; nor was l

there any acceptance sanplirg/testiry required or perforned. General Electric Nuclear Energy (GHE) Drawiry 11o. 234C735CX, Revision 8, for e

the 01S core spray pmp ard mtor, stated that the minimum viscosity for the lubricant at 210 dcgrees Fahrenheit ('F) was to be 45 Saybolt Universal Seconds.(SUS); whercar the Mobil prcduct data sheet stated that the vicccuity for DTF 797 o. nt 210*F was 44 SUS. '1he discrepancy was neither resolved or docunented, nor was it justified.

In response to this con rn, NPPD contactcd GENE ard prcduced a record of'tolophone conversation with a CDE representative on February 27, 1992, that did not specifically resolve the discrepancy, but reportedly s,tated that UTE 797 oil (anong other lubricants) was acceptlle, other deficiencies included: no receipt-inspcction, no traceability to the ODi, critical characteristic of envitannental qualification not verified, ud no similarity to ID sarple or traceability to rp test report.

I I

l.

66

~.-. -.-

.m _.

Ailitionally, the 11RC ingoction team had conocrns associatcd with the environmental qualification itself thich imludad:

(1) MpligAbility to 01S's comercial-ctred? Mobl1 DTE 797 oil of Ibbil's IUTfDI Emineerp_ImpgrtJD-01-101. Reyip, ion 1 March 2h 1987, "IhvirprEIltal Oualification of Ibbil Oils ard Greams"...

Other ocumercial-grade Mobil oils and greases in casential-DQ use i

at Q15 includits Ibbilgrease 28 (D2 appilcation: the Limitorque POV actuator limit switch gear case). Accordirg to the Mobil product data sheet on "Mobilrad"-series nuclear-grade lubricants, cantnercial-grade DrE 797 oil (Mobil Product lio. 60011-4), ard

.others (!bbilgreare 28, Product !!o. 52062-6) worn used as D2 test sanples, but Report no-01-101, intended to apply to Ibbilrd-ocries lubricants (specific CIS exampics: Mobilrad oil 797, Product lio. 60006-4 ard Mobilrad SHC 28 greare, Prcduct tio. 53060-0), reportedly were of the same composition as the CGIs, but only Mobilrad-series lubricants were claimed to be supportcd lif Ibbil's 10 CFR Part 50 Apperdix B QA/IO CTR Part 21 prcgrams; the inplication clearly beirn guaranteed consistent similarity to the testod camples.

- (2) Rplig3hility of sipulation jn reoort to OJS arolicatlOD conditi9ng, that is, similarity of tested to instalkd corditions:

'Ihe test fixture bearirq loadirq was-40 to 60 pourds, based on producing a 45'C oil terperature riru through the bearisg; whereas, according to the core spray pump motor drawing, vertical bearirg static loadiry (continuous dcwn thrust) was 2525 pounds, l

but there was no documented resolution of the difference. 'Ihe 01S lubrication engineer contended that bearirgs of the typa in question are generally designed for a 45*C rjse, but he did not address the possible effects of specific loadiry, _or any' synergism between heat and nochanical Wrking that might be accounted for by temperature rise alone.

L (3)

Problems witb the rerort itself: - Ibbil Report MD-01-101, Revision 1, March 24, 1987, Apperdix D, " Calculation Package to Establish 'Ibmperature Agiry Times for Mobil' Oils ard Greases," was prepared for Mobil by IvUITui Engineers, San Jose, California; agirg and loss-of-coolant accident (BOCA)- by Wyle Laboratories-llorco, California; sanple irradiation agiry by Radiation 3

Sterilizers; and design-insis accident (DBA) IrcA dose sample 7

irradiation by ISOMEDIX, I:rorporated; sanple physical and chemical amlysis by an irdeperdent laboratory; ard infrared spect.roscopy scans by Mobil itself. IUrD01's Arrhenius calcu-lations-for accelerated thernal aging were ambiguous.. 'Ihey contained a questionable determirntion of sample activation energies based on re-called " life" data -(but with no erd-of-life

- point or cordition defined) using only two data points, for exanple, for UTE 797 oil:

410 " minutes" (sic] at 150*C and 385

" minutes"-[ sic] at 140*C. _However, the Institute _of Electrical ard Electronics Drgineers (IEEE) Stardard 101-1972, " Guide for

~21-67-

~

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

. -. ~ - -. - - -

Statistical Arulysis of 1hernal Life 7bst tuta," referenced for Arrhenius nocelerated agirg/ activation energy calculations in IDI Stanhrd 323-1974, "lDI Stan1Trd for Qualifyirq Class if 11;uirrent for Nuclear pWer Gercrating Stations," (accordire to which rethods the lubricants were suppocod to be qtulified) rcquires a mininn of three data points.

In acklition, it was not clear that thn agiry tires, for exanple, 272 unspecifiod units of tin at 150*C (for DrE 797 oil) for cquivalent degradation to 36

.i raths at 95'C, were apprcpriate.

If the agirg tire was expresscd in hours, which mkos core sense, it would then have been extrapolats_d, that is, not tourdcd 17/ activation enenJy life data, liwever, if it was oc,rrectly exprenscd in minutes, which would te ocanparable to the life data at similar temperatures, then it did not meet the minian agirq tire rcquirenont,100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />, of IDI 323-74.

In response to these questions, GIS agrecd to obtain clarification fram ifUTIDI, ard others, as apprcpriato.

1hree additiorn1 types of lubricants used, cocorTlirq to the CIS Master lubricant List, in essential-IQ applications were procurcd or receivcd since January 1,1990, on various ronesseatial IOs, includirrJ: TO 250546 (01cvron SRT 11o. 2 grease, use in.lQ electric rotors), 10 345259 (Exxon IJebula D2-1, pcssible use in EQ equirrent, e.g., Limitongue actuators), IOs 343117 ard 315910 (!bbilgrease 28, use: ID-Limitorque actuator limit switch gear tox).

'Ihis informtion was obtaincd ftm a 01S cpare ynrt informtion retrieval system printout.

'I*iene nonessential lubricants were prccured under the same controls as' the 11 Jricants Separately revicWcd alOve ard would be expected to exhibit the same types of dodication ard/or environnental qualificatica.

In response to these concerns, some of which were origirully rairead durire the NRC'r, preinspection visit, 01S preparcd a pooition paper on procurerent of lubricants for essential (ard essential-IQ) applications. The team reviewed the position paper ard fourd that although it committed CIS to initiate a program of-randca acceptance sanpliry ard arnlysis (by a Mobil laboratot,y, as an "enhancerent" to provide " greater initial assurance that what we ordercd is khat we receivod"), the acceptability of lubricants was still lanjely stated to be bascd on reliarro on the various lubricant IQ reports. The questions of documented, verifiable traceability to the ODi ard ODi test reports and consistert similarity (i.e., batch homogeneit~ ard traceability) to IQ-testcd sanples were not fully acklressed.

4 IECURD(ENf NJD DEDICNTIOtt TRAINI!G l

'1he inspection team reviewed CIS's training activities in support of the pacess of dcdication of 03Is uscd in safety-related applications performed

- after January 1,- 1990.- In Ray 1990,- QiS's ibchnical Staf f Training (TST) group Ixtrformed a job survey ard task analysis which identifiod the required campor.mts needed for performing such specific taaks as commercial grado dedication.. The reviewers fourri that knowlcdge in a variety of areas was rcquired,. inc'. lire cystern ard cmponents, codes and stantirds, updatcd safety analyr report (USAR) ard system safety functions, ard determination of ccanponent

, iga characteristics.

4 4 4

68

l j

Dyirymrirn Departrent Instruction 89-04, " System Dyineer Prcgram,"

Ibvision 1, Febntary 26, 1990, rcquired system cryincers to be certificd in accordance with 1TU 502, "Todanical Staff," IMvision 8, April 23, 1991, whidi requircd crxpletion of courses identified in the tank aralysis as necensar'y knwlcxkje required to support the dodication proocss. Ccurse NE004-01-01,

" Codes, Stardards, ard Classifications," Revision 1, llovcrber 21, 1990, discussed the Code of Federal Regulations, Imc ngulatory guides, and the M2E Ccde. Course MH009-01-01, "USAR Overviw," Pcvision 0,14 arch 22,1991, discussed the USMt ard the safety classification of the plant's systems.

Course ME011-01-01, "Testiro Overvicv," Revinion 0, January 24, 1992, had becn given on February 19, 1992, to a class that incitdod six system ergineera. We cource covered the major asgets of the dcdication process, such as identifyirn safety function, determinirg critical daracteristics, ard colectity verification nothods. %c cource was not a requirement in the systcmo cryincer's trainirg curriculum at the tine of the IRC inspction, although it was indicated that TPD 502 would be revised to irclude thin course.

In addition, IUr was in the process of developiry a not of ergineering workbooks, each of which would be unique to a particular system, ard indicated that trainity requircments would be reviscd to require the system ergineer to canplete the portions of the ergineer workbcok nquired for each system that the ergirv.er nealed to kncw ainit. The inspection team revicwed an ergineer wrkbook proposed for the core spray system ard deteratined that, once the workbook was ccoplete, it would familiarize students with parametern and operatire characteristics of the system and would support the technical aspects of comncrcial grado dcdication. After revicvity 01S's tank analysis information, trainiry rcquirenents, ard various cources that were nquired ard were beirq planned, the inspection team conclud<xt that ist had developcd an adequate technical framework to support the process of the dodication of 031s litt had failed to require any trainiry that specifically addressed the process itself. %c course epocific to the dedication process, "Testire overview,"

had been given to approximately 25 percent of the system cryincers in Februat'y 1992.

Dyineers were not yet required to ccuplete the ocurce before performity dedication-related activities.

5 FXF1 MDTIl1G on Febnlary 28, 1951, the inspection team coMucted an exit meetirq with members of tne NITO staff ard nonagerent at tla 04S site. Durire the exit meetirq the tem cunnarized the inspection findirns ard circrvations. We folicwing irdividuals were present:

~23-j 69

lhiGTdA.IM!211c IWr Digtriqi G. llorn, !Mclear Ibrir Group Ramger J. Meadnm, Division IMclear Operations Mimger G. fknith, IMclear Lloenairg and Safety Ramger M. I:stes, CPI Tank Force Inader J. Larcon, @ Supplier Supervisor D. Robinson, @ Mamger M. Sponoer, lhgineerity Pru3 rams Supervisor R. Gardner, Senior Operations Mamger S. Peterson, Senior Kanager of 7bchnical Support Services M. Doan, Licensing Supervinor L. Brag, Rcquiatory coupliance Spccialist

11. liitdi, Plant Services Rimger J. Flaherty, cis Dgineering Manager D. Overbock, Purdiasirq and Lcerials Supervisor V. Wolstenholm, Division Rmager - @

B. 7bline, Technical Staff Trainity Instructor R. Giboon, Audit and Procuru:ent @ Sugervisor R. Wilbur, Division Rmager, Nuclear Engineerity and construction D. W11tman, Division Manager, Nuclear Support R. Wenzl, lHD Sito Dyineerirg Masoger J. Dutton, Trainirn R1 nager R. Uhri, ibch/Cpo Supervinor Ligglear Pmulatarv Conmission L. Norrbolm, Olief, VIB U. Potaptws, Section Q11cf, VIB I. Batncs, Section Chief, RIV R. Pettis, Team Inader, VIB B. Rogers, Reatt.er Dyineer, VIB W. Gleaves, Mechanical Dyineer, VIB S. Alexarder, D2 ard Test Dgineer, VID L. Ellerchaw, Ibactor Inspector, RIV R. Evans, Resident inspector, RIV W Walker, Resident Inspector, 01S Other Ormnizations B. Bradley, Senior Project Rsnager, NIEARC T. Spink, Riterials Rinagement Services, TD ERA J. Grace, SRAB Member A. Itubil, SRAB Administrator H. Green, SRAB Cutside Meaber L. Payne, Mamger, supplier Quality, Wolf Creck N. Iloadley, Kanager, R1uip:ent Engineerire, Wolf Creek 70 1

ApFENDIX NEBRASKA PUBl.IC POWER DISTRICT PROCUREMENT PROGRAM ENilANCEMENTS The purpose of this paper is to provide a status of the enhancerrents to the District's Procurement Program which have been implemented and to idcntify those which are to be implemented by July 1, In92.

A Procurerr nt Initiative Task Force was created in October 1990 to evaluate the NRC's eight assessment inspections, the NUMARC Initiatives, and to upgrade CNS's Procurement Program accordingly. The NUMARC Comprehensive Procurement Initiative has been analyzed and an action plan developed.

In addition, an industry recognized expert was utilized to review the current CNS Procurement

Program, llis recommendations have been factored into this action plan.

A number of actions have been t aken as a result of the t,bove activities.

They are as follows:

1.

The eight NRC Procurement Assessment inspections have been evaluated, the findings categorized and summarized, and an action plan developed.

2.

Procedure 3.22

" Commercial Grade Specification," was revised to address several program improvertent.s.

3.

A "llold" statement has been placed in each approved dedication package, pending review to ensure compliance with current procedural requirements.

4 A position paper has been generated on the classification and use 4

of lubricanta which will form the basis for producing a dedication package.

S.

Special trftining has been conducted with System Engineers on the latest revision t o CNS Procedure 3.22, " Commercial Grade Specification."

The following actions are planned to be implemented by July 1, 1992:

5 1.

Establish procedural requirements to provide formal documentation af critical characteristics as applied to Essential. Commercial Grade (ECG) procurement.

2.

Formalize the Engineering Programs Department independent review of dedication packages and ECC technical evaluations.

3.

Irtprove testing and inspection capabilities.

4.

Review at ? revise procurement procedures (e.g 3.22, 3.24, 1.13, QAI-16) as appropriate A-1 71

.._._.-._m...__

5.

Enhance Quality Assurance supplier audits.

-d.

Implement testing for lubricants along with a dedication package p.E decide to purchase under a 10CFR50, Appendix B Program, i

In addit!on, the NRC Inspection of February 24, 1992, identified the Commercial Surveys of Essential Commercial Grade Suppliers from January 1, 1990, to May 1991 to be broad-based, programmatic, Appendix B type audits and

- not product / critical characteristic commercial surveys. The following actions will address this concern:

1.

A

• HOLD" will be placed on all items in the Varehouse purchased as ECG since January 1, 1990, until such time as either the survey is re performed using the NUPIC Commercial Survey Checklist, or the item is formally dedicated.

2.

Focused, commercial surveys of ECG suppliers will be completed prior to January 1, 1993, using the NUPIC Commercial Survey i

Checklist.

All of the above items have been reviewed and approved for implementation by

- Nuclear Power Group Management.

.A sY C. Mi hael Estes Comprehensive Procurement Initir.tive Task Force 1.eader A-2 1

-72

I

/p* *8 cp %

UNITED $T ATES f,",,, q '#h NUCLE AR REGULATORY COMMISSION

<-t W ASHING T ON, D. C. 70565 o

%' e.,, e */

April 9, 1992 Docket Ib. 999004G4 Mr. Nidolas J. Lipuulo, livager ibclear Safety ard IWJulatory Activitien Dnirocrirg 'Ibdmology lbclear ard Advarced Tochrology Division Enertyy Systrina Dusfress Unit Westingtwuse Electric 03rporaticn Itst of fioe Dax 355 PittchnTJh, Ibnrmylvania 15230-0355

Dear Mr. Lipatulo:

SUa7DCT:

DD~RADCD IUN%NCE CDtIDITIQ1 OT ' DIE IPOCESS ITorDCTIQi SYSIDi Di

%"rDGQ5E PRESSURIZED 1&TER IU7CIOR PLA!GS (Imc DEFECTIQi REIU C No. 99900404/92-01)

'Ihis letter refers to the inspwtien acrductal by Mr. Stef en Alexander of f

this offlee cn February 5 and 6,1992. 'Ihc ircyction ircluded a revis of activities authorized for your !bclear ard Advanced Technology Divisicn (WRTD) facility in Manteeville, Ibnnsylvania. At the octrlusicn of the inspection, the fin 11ngs were dirmeM with those renbers of your staff identified in the encicsod report. Subacquent to the inspecticn, there were telephcne dirmee-icns cn April 7,1992, between rir..*.lexarder and Mr. J.S.

Srinivasan ard Ms. I. Girgerich of ycur staf f riguuiry the inspction.

Areas examined durity the inspecticn are dMwn1 in the report. The 100 inspcctor assec> sed the adequacy of R&TD's evaluation anl reporting of a degradcd performroe canditicn of the Westirghause plum protection systm (PPG) cucrterperature/ differential torperature reactor trip functicn, ard the distribution of relevant technical inforwtion by MM to affected NRC licensees. Within these areas, the inspectico consisted of celective enhtion of procedures and representative records and interviews with personnel.

'Ihe iropr rottd that WRTD was helpful in prwidL-mformaticn to supplement NRC gercric ccr1munications to licensocs for the areas inspected.

IL%wer, a findirg was identificd with rogard to the generic safety analysis perfomad by RETD to evaluate the ' graded perforrance conditicn of the PPS.

'Ihe results of the Ra'ID generic arJ ysis were significantly differunt frm the ocoalunions of an analysis performed by Duke Pcuer Orpany for the Catawba Nuclear Ib cr Plant and were rxx.-conse2Vative.

It could rut be detemired durire this inspection whether the RWu analysis was gercrically conservative ani adequately accounts for potential worst case 73

Mr. Nicholas J. Lipuulo, MR7D ocn11tiarc of other accident se/erity factors, mx:h as rui rAxkuiry tMt may exist at spccific plants. It therefore realm unrtsolved stxMer the deviation oculd croote a subatantial safety Mr.ard, crnld ocntrilute to excocdirn a rafety lir. tit, or, by causirg violatico of 11oerce technical spccificaticn setpoint limits could acrctitute a failure to cxrply, as defirrd in 10 CFR Part 21.

Because this firdirg cxnid rot be satisfactorily resolvcd within the scqe of this irspcction, it is identi'!cd in the rtport as an unresolved its perdirg further zwiw bf the IEC.

Purctant to the pru/isions of Section 2.790 of 10 CTR Part 2, the IEC's " Pules

~

of Practioc," a cxp/ of this letter ard its crclocure will te placed in the 1&c's lublic Docurent Rocan.

Shculd you Mvc any questicrs corcernirg this irspecticn, we will be pleascd to discuss the:n with you.

Sircerely, y v.

he ow y leif.. !_

e lJn, 0 11 e f Vendor Ircpecticn Branch Divisicn of Reactor Irspectico ard Safogturds Office of liuclear Reactor Regulaticn Delocure:

lEC Ircpcction Report 99903404/92-01 oc:

w/etcl:

N. P. Maeller, Kvager, control Systers Mdysis 74

)

l

Delosure DGPErrIQi madd' U.S.1RX:Ir>R REDUIAIGN CONISSIQi OTTICE OF 10CIEAR IT/CIUt REGJ1ATIQi DIVLSIQi OF REACIUt DEPDCTIQi AND SAFIIUARDS Verdor Name:

Westirghcuse Nuclear ard Advanood Techrralogy Divisico (WWID)

Enertyf Systes 111siness Unit Westinghouse Electric Corporaticn Aiiress:

Post Office B:* 355 Pittch:rgh, Pennsylvania 15230-0355 Report No.:

99900404/92-01 In W ' Activity:

Design, ergineering, ard other technical servloes for Westiryhause pressurizcd water rector plants Inspecticn Ihtes:

February 5 ard 6,1992 Inspecto1:

Stephem D. Alexander, Equi mnnt Qualification l

ard Test Dgineer Prepartd Dy:

m S.D. 'Alexardai, Reactive Inspection

/Date Secticn 2 (RIS-2)

Verdor I "pcction Iturth Approved Rf:

}

9

.2 VArill S.

uyhm, Acting 011ef, RIS-2

' D6te i

Vendor ion Brand) l Divisico of Reactor Inspccticn ard Safeguards Irspecticn Bases:

10 GR Part 21 ard 10 GR Part 50, Apperriix B Inspectico Scope:

'Ib assess the evaluaticn ard trporting by MWID of the degraded performrce conditico of the Westinghouse process protection system (PPS) overteTerature/

differential tenperaturu (OIUT) reactor trip functicn, and the distribution by MRTD of technical informtion on the scalirg of the PPS to licensees of Westinghouse presstirizcd water reactor plants.

Plants Affected:

Pnu rized water reactors with Westiryhouse Type 7100 arn #,300 ard Foxboro process protecticn equi; rent 75

,w 1.0 DEPDCIIret MWARY 1.1 Vipirations

~

ikm.

g

[:;%4

1. 2 ' 13cnconformnoq2

}kre.

g 3j 1_

k, 1.3 Unresolved It m h

'Ibe Nx; inspector mviewed a safety analysis performd by the Westinghouse h

Nuclear Mu.rcoi 'Ibchnology Divisim (MRTD) of the degraded performnoe of the overterperature/ differential taperature (OIUT) rwr: tor trip functicn of a.-

tha Westirghause 'B(pe 7300 process proteWm systm (TS). R WID stated that the degraded pertonrarce restatoi frr-scalirg of the FPS, vy st due to the diversity of reactor RRTD's gercric safety aralys'

. per'ormance of the 01Ur trip protective furrticns available

.e would not create a substantial

.y nazard, as defimd by 10 CFR P. art 21.

-liowever, Duke Power Cmpany (Da..e) had ccncluded from their core themal-hydraulic analysis that, ft.r the Catawba leimr Iwcr Plant (Catasta), an urcontrolled M red withiranl frun 10 vcuJa pcwer with OIUr disabled

. auld result in a departure frtu nuc:m".e,uD *Jg (Ils) cxrditicn in tha hat channel. Duke postulated potential oft' tits. radioactivity releases in Utis mC '9emer, these wleases were less than 10 percent of 10 CIR Part 100 y: e 11.es.

'Ibe inspector had no inxxiiate safety concerns because Duke and ot'er affected lf NJsets had taken corrective action, or had been alerted to the problem throtgh R9r: Inferration Notice 91+52 ard a Westirghause Nuclear Services.

Divisica (RED) technical bulletin. tkuever, the irspecte* noted that there were significant differerrxct in the results of the arnlyses perforrtd by Duke ard WWID. 'Ibe adequcy of RMID's 10 CFR Part 21 evaluation could not be determined without a detailud cxrpariron of these differences. 'Iherefore this issue could not be resolved durirg dus inspection. Fendits further NRC myiew, this issue is considered an unresolved item (99900404/92 01-01, see sectico 3.2.1 of this report).

2.0 SIAIUS OF TELVIOUS DEPECTICH FDCDES Previous innection firriirgs did not pertain to the scop s this inspecticn

- aM were nct adckeM.

3.0 DEPECTIOl' FD0 DES AND CmER CrNMDTTS

3. 7.

Entrance ard FRit Meetjagg Durity the entrance metirg en February 5,1992, the NRC inspector mt with the cognizant staff ard rarngement of WWID's Cbntrol Systas Analysis Group.

'Ibe inspector din'wmi the basis ard scope of the inspecticn, highlighted 2

76

- re-m c

,e.

e

areas of crrcern, ard catablichcd workiry inten' aces. At the cxrclusion of the inspection on February 6,1992, the incpector sunrarizM the firdirg ard corcems, ard MUGD's nvwgccr:nt ard staff ackreuloiged this informticn.

Sectico 4.0 of this report identifles the personrel who atten%d these rectirgs.

3.2 Wstinghouse Type 7300 PPS, M Reactor Trip Functico orrmenwealth Edison carpar.y's Dyron and

.dwood Nuclear Pcwer Plants, ard Duke roe.r Ctrpany's (Duke's) }tGuire ard Catawba }Aclear Power Plants, reocntly reported a saturatico probim wath the e reactor trip furction of their Mrtirghause Type 7300 PPS. S e problem involved the K reactor trip set point, used by the coircidence logic of the reactor protectN system (RPS), not beiry reduced (to the extent requirvd) by rising reactor coolant average ter:perature (hvgj over the full rarge of Tavg (530*F to 630*F). his had resulted frm an irconect distribution of arplifier gain among the cirulit cards of the Tavg charrel of the OIUr functicn during the scalirg pr e e.

Se IEC was notified of the problem by licensee event report (ID) 369/91-09 frm kbGuire ard ITR 413/91-09 frm Catawba, both dstod Atgust 22, 1991, cont,equently, the IRC issued Informtico Notice (IN) 91-52, "Ncnconservative Errors in overterperature Delta 'Itrporaulm (OIUr) Setpoint caused by Improper Gain Settirgs," dated ALyust 29, 1991, to alert '.icensees to this prtblem.

MED r,ubcoquently MM technical bulletin }GD-TO-91-0840 on rs % r 13, 1991, to ackiress this problem. Wis bulletin (1) exparded the scope of the problem to ircitde Westinghouse Type 7100 ard Ftoni:cro PPS equipoent, (2) acMressed the transient effect accountirn for the rate of incrmse of Tavg, and (3) provided a recotrended ccalirq nethod without hardware scxilfications that would ensure that the OIUr setpoint would reach its design mininum value before the hvg chanrel saturated under the worst case transient ocnditicns.

We MGD technical bulletin scaling gain distribution plu involved raisirq the gain of the OIUr sumirq amplifier (the last card in the cinnnel) as far as r - ary above unity. Bis rade it p::csible to diselbute 1cuer gains over the other canis in the channel and still achieve We overall gain rtquired by fuel calculation parameters.

During scalirg, the process variables nottd in the plant's "Precauticos, Limitations, aid Setpoints" Manual (PIS) are converted into electrical paraneters ard used to set the equipnen Ibwever, if emive gain is Itmpod cn the Tavg lead / lay rxdule (lead card in the Tavg channel), the output signal of the nodule will reacn its raxirm value before a risiry Tavg reaches 630*F. - We lead / lag redule also introduces a rate corponent into the sigml.

%e z.ambinatica of the ircornrt gain distribution ard the rate ccrpcnent msults in the output of the Tavg lead / lag rodule rmchirq its maxinum value at an even loser Tavg. Berefore, the rate et increase of Tavg results in the OIUT setroint getti q no further reduction frm the rising Tavg. During this inspection, the irq ector reviewed WWID's evaluation and reportirg of the dcgraded perforrance condition of the Westiryhouse plant protecticn system OIUr reactor trip function, ard the distribution by MIA'ID of technical infomation on the scaling of the PPS to affectM NRC licensees.

3 77

3.2.1 ' safety Amlysis of 01Ur Reactor Trip runction 7tw impoccor tuviewed the mmrnry of a safety amlysis, EP-TEA 'IA-II-91-436, "overtorperaturu Delta-T Rescalirg ' transient Amlysis," dated }kneeber 20, 1991, (kNeloped by tM Transient Analysis II Secticn of the Nuclear Safety Aralysis Group of the Dyinocrity Tochrology Dcpartment of M' AID. The ircpoctor also ocntactM ccgnizant Dha cagineers for details of an cmlysis Duke cryineers had perfo:ved for the degradaticn of the GIUr trip function.

Both the Wa1D aid Duke ulyses conside T the unomtrolled bank rod withdrawal (UH% ovent to M the worst t i against which the UIUr trip provided the first and principal protectico. The inspector roted tmt the results of the RRID ard DLike amlyses differtd Fignificantly.

Duke had stated that, for Ottawba the departure frun rucleate boilirq (um) l therm 1 limit wuld be expoorkd in tlw design hot chnnel with scrae postulated corv damage ard effsite re'nases. This conclusico was based on the precent i

ocn e life, with a dograded CIUr functim, ard with a tmRW at 10-percent ruector pcuer. '7ho inspector discussed this with Dake perscrinal by teleptone while at 14RTD aM confirwd tMt tM resultJrg ckparture trun rucloato boillrg ratio (DER) was calculated by Duke to go belcw 1.0, implying that DG would cocur.

In contrast, the Westinghouse ergineers who perforred the MRID safety j.

l amlysis stated tMt althcugh their calculated DE went belcw the design I

mrgin valuo (ncaniml) of 1.3, it did not, in their amlysis, go belcw 1.0.

(

Thereform, accordirq to the MRID amlysis, while tM syste suffered a I

reducticn of safety margin te the thermal limit, the Da limit would rot be actually exceled. This ccrclusion was satal to N resed cn the diversity of protective f.incticos available, ircitdire the b',gh flux trip, the high pressurizer pressure trip, high prussurizer level trip, ard the overpawer-differential r eperature (OPDF) trip. The Duke analysis also took into ocnsideration the contributians of tM other applicable tuactor protactive functions, but had concluded that the protective functions did not take cffact coan erngh in the postulated event to prevent e=uiirg the De 1

thermal limit.

The inspector interviewod the MRTD erginocrirg personnel ute performed the analysis to tuview assunptions and conservatisms that might Locount for the differify conclusions with the Duke analysis. In addition, the inspector obtained clarificaticns frun the cognizant Duke cryinocru cm the calculated DR at Catawba for a tmRW event urder the postulated coniiticns. Tho

. inspector-noted that Duke had considered the lcu-settirg high-flux trip, ncaninally at 25-pcIrent reactor power, to be disabicd at 10-percent paer.

Duke had also considermd the severity factor of ocotrol rod shadcwirg that was characteristic of the current rod positionirg scheme for the Catawba core at the present time in core life. The MR7D cafety analysis sunmary indicated that MWID had also considered a rarge of initial power levels, ard wtwa WR7D crgineering personnel were interviewed, they statal that 10-perrent power had been incitdod. licwever, b. vas rot clear frun interviews with the transient l

analysts if MR1D had taken into account the fact that the lcu-settirg high-flux trip was disabled at 10-perrent power. Subcoquent to the inspection, MR1D reported that it did not tW crulit for the lcu-setting high-flux t-ip 4

l 78

l in tM generic tefety amlysis. MRTD also stated that it had perforrrd tM analysis with the aid of tM "IDPnWG" cxrtuter code rodel for com therml-hydraulic analyses, but explaired that the code watGd not have separately, or directly, auoounted for specific cevere Ird shackwirg, usirn insted, scrae namimi values of amicgaus or related parameters. Althatyh tk lack of low settirg high-flux trip was reportedly crnsidertd, it annins unresolved whether the MRTD analysis was generically conservative ard shether the nethodolcgy a&qmtcly accounts for potential wrst case canditicrw of other severity factors, such as red shadcuirg.

The inspx: tor had no LMate aafety crrrems because DAe and other affected licensces had taken corrective actico, or rad been alerted to the prtble thrtugh Imc Inforetion Notice 91-52 and a Westirghause 1Aclear Services Divisicn (RGD) technical bulletin. Nevertheless, ee inspector otncitdad tMt the differerces beween the Insults of the Dub ard MR7D amlyces could not be roccnciled without a detailed occpariscn of tM calculatican.

kndirg further imC review, this issue is cmsidertd part of Unresolval Its No.

99900404/92-01-01.

3.2.2 MRTD 10 GR Part 21 Evaluation of OIUT Reactor Trip Nnctico 7he inspetor ^e-ed the adequacy of Ma1D's evaluaticn of the degradaticn of the OIUT trip furstion as a deviaticx1 frm technical procurement spocifications. MRID's 10 GR Part 21 evaluaticn was perforrxd by the Instrunent and Contrul Syh Licensirq (ICSL) organizaticn.

The inspector fcund several interml rmas regA* the detemimtien of the status of all potentially affected plants, the MED 7bchnical Dalletin, IGC Informtion Notice 91-52, and other backgrourd dociraents supportirg ICSL's corrlusim that the potential defect (deviaticn] or failure to cxmply could

+

rot create a substantial safety hazard. The bases of their conclusicn was the safety amlysis (see section 3.2.1) perforrod by MRID's Nuclear Safety Analysis Gruup. The in Wsr cbserved that MRID had evaluated the deviatian and determined that it was not a defect because it would not create a substential safety hazard. Ibwever, its contributicn to cyrwairq a safety limit was not explicitly addressed. The inspector also roted that affected licensees or purchasers had baen informed (via the MED technical bulletin) in order that they rey evaluate the deviatico ard take appropriate corrective actim. Ikwver, tne irqxrtor had questiens cn the nau]uacy and conservatim of the amlysis and could rot reach a ccnclusim cn the adegaacy of MRID's 10 GR Part 21 ccrpliarre.

The inspector further rxt.ed that reducticn of the I2ER belcu the ncninal design value, kttich constitutes a rutuction of safety margin, my not per se

. violate a specific safety limit, as defined in technical specifications (part p

of the IOC plant operating license); ard even the severity of the Catawba s&mrio ray not reach the level of creation of a substantial safety hazard as defined in 10 GR Fart 21. On this prtnise, the deviation my rot meet the definition of a defect, reportable urder 10 GR Part 21.

However, the degradation of the performnoe of the PPS as a result of premature OIUr saturation cb.s cause the violation of license technical specification setpoint limits ard ther7 fore, could be detemined upcn further review to 5

4 79

ocnstitute a failure to ocuply, as definte in 10 CFR Part 21, with a " rule, rcquiation order, or limnse of the Ommission relatiry to a sutstantial safety hazard." On the insis of such a deternimtion, the corditico would be cxxsidertd reportable to the }RC urder 10 CFB Part 21.

Peniirq further NRC review of the safety amlyses, this issue is desigmted an unresolved itan, 99900404/92-01-01.

3.2.3 Distrilution of Scalirq Infornation by MRID The inspector asserad the ade:pacy of the activit.cs of MRID and other Westirghause organizations, as aguqa-iate, with regard to the distrilution ard availability of technical inforration and <h'entation relatirq to the scalirg of Westirghause-supplied PPS aquiptent.

Westirghcuse PPS scalirg rarmals werm provarod by MRID for specific plants to aid licensoes in scalire their PPS equipaent. MRTD stated that the scalirg nanuals wem considertd "opticnal" tutnical th'entation ard were provided to those plants that sp3cifically requested them as part of the procurment of the Nuclear Steam Supply System (NSSS) software package. MRID stated that it was stardard MR7D practice to have custcrors mark up a ocimuter printcut software standard that listed all available technical documentation. MRID produced a sa nple software stardard frcra the Westirghcuse software schedulirg systen. that did list the scalirn manual and identified it as the stardard vehicle by which custczners irdicated what cptioral technical ch'~ntation they wished to ru tduse. The marked-up stardard would then be transcribed by MRID onto a software order package to be premi as part of the customer purctase order for the plant. M&1D also prcduced proccdures irdicating that the scalig nanual would have been ircluded in the software stardards offered to all custcrers of the PPS.

MRTD also unintaintd that the scalirg techniques provided in its scaling

' tanuals were well kncun to techniciars experlerced in the set up ard operation of amlog process control equignent.

It was therefore expccted that scrne utilities wculd choosa ret to rurtime the scalirs nanual if they Irv mi or could athervise obtaim

.cquirtd realirq expertise. This pcniticn was supportxd tyf eviderre that rac utilities apprently managed to perform proper scaling without navirg scalirg annualc.

In contrast, others scaled the equipmnt incorrectly despite havirg the scalirg manual.

.7he inspector'noted that the plant specific pr,a ss variables used for scaling were ackiressed in the " Precautions, Limitations, and Setpoints" Manuals, ard that these nanun.ls were provided to all custwo as part of the standard NSSS scope of-supply. MR2D also stated that all its custcriers were offered fonnal training on scaliry techniques ard that MRID was in the se of providing assistance to affected licensees to help tt m correct the degraded condition of the OIUT trip function.

The: inspector concluded that there was sufficient evidence that MRID had been providity ard/or makirg available scaling manuals to their customrs. No further concerns were identified.

6 80

4.0 PIRSCROEL CCREACTID Westirrtyxtge Electric Gn u.n ation

• R. B. Miller, Fellow Dgirer, Nuclear Safety ard Regulatory Activitics

• N..P. Haeller, Manager, Control Systems Analysis

• J. S. Srinivasan, Pri.Tipal Dgirar, Ocntrol Systes Aralysis M. P. Osborno, Harage't, Naclear Safety Analysis I. Girgerich, Dv3 rwr, Nuclear Safety Analysis 1

Noclear Pandatory Ornissign A. Gautam, Actirg Sectico Chief, VIB, NRR a

w 9

Denotes those attenling de exit interview on February 6,1992, at the canclusion of the inspection.

7 81

pa nov J'

UNITED STATES n

E

> i NUCLEAR REGULATORY COMMISSION k

/

WASHINGTON D.C. 20556

%,,,,,8 May 21, 1992 Docket No. 99900104 Mr. B. R. Snelstoys Plant lumger Westirghause Electric ompany Nuclear orponents Division 8301 Scenic Hicdway Pensacola, Florida 32514 Dear Mr. Smelstoys SUT17DCT:,

NCfTICE OF VIOUJICt1 AND NCt1CQEORMANCE (imC INSPECTION RDORP

10. 99900104/92-01)

'Ihis letter addresses the inspection of your facility at Pensacola, Florida corducted by Messrs. L.L. Cupbell, W.C. Glem is, S.M. Matthew, ard U. Potapyr. on April 6 through 10, 1992, a 6 the M m'Fsions of their findings with yco a.d other acmbers of your staff at the conclusion of the inspection.

'Ihe inspection was conducted to assesc Westirghause Electric Cc@tny Nuclear Oxponents Division, Pensacola Plant's (WPP's) cmpliance with the U. S.

Nuclear Regulatory n=imion (IEC) requirements i-M in Virginia Electric Power Ctnpany's purchase order for three replacement steam generator

- suhwMlies for Unit 1 of the North Anna Power Station. 'Ihe performance-bascd inspection was conducted to evaluate WPP's quality program ard its inplementation in selected areas such as (1) design pr m rm and interfaces, (2) purchased material and services, (3) special rem, and (4) inspection and test.

Areas examined during the NRC inspection and our findings are dise'W in the inspection report (Erclosure 3).

'Ihis inspection consisted of an examination of procedures anci representative records, interviews with personnel, ard observations by the inspectors.

'Ihe inspection identified that scne of your activities appeared to be in violation of NRC requirements,. as specified in the enclosed Notice of Violation (Enclosure 1).

'Ibe violation identified that neither WPP's nor the corporate Westirghouse proccdures addressed the requirements of Section 21.21,

" Notification," of Title 10 of the Code of Federal Recrulations (10 CFR) as revised ard effective on Octr+w 29, 1991 (e.g. the 60 day period for evaluating potential defects and failures or filing an interim report was not addressed). _Also, the irspection team identified the failure of WPP to invoke the requirements of Section 21.31, " Procurement Docume its," of 10 CIR on three safety-related purchase orders issued to surpliers of basic ccuponents.

82

Mr.. B. R. Snelstays -In acconlarm with Section _III of Appendix C, "Gercral Statemnt of Policy aM Procedure for imC Enfc:urnent Actions," of 10 CFR Part 2 (1990), these firdirns are desigrated violations 92-01-01 aM 92-01-02 as described in the Notice of Violation (Enclocure 1) and have been classified as Severity Invel IV violations.

You are rcquirtd to respond to this letter and should follow the instructions specifled in the exclosed Notice of Violation when prc. paring your response.

In-your response, you should document the specific actions taken aM any additional actions you plan.to prevent recurrence. After reviewirg your response to this Notice, incitrling your propcced cuwetive actions and the

-results of future inspections, the NRC will determine whether further NRC enforecrent action is rumwry to ensure ccepliance with Imc regulatory requirenents.

In addition, durirg the inspection it was found that the inplementation of your quality assurarre (QA) program failed to rect certain NRC requirements.

We inspection identified instances in khict) WPP failed to follow the req.11renents of its Nuclear Quality Assurance Program }hnual for the control

-of reasuring aM test equipnent. It was also'determired that WPP failed to adequately verify that the supplier of the flow limiting insert was effectively inplementirg its QA pn. gram.

Please provide us within 30 days frcxn the date of this. letter a written statem.nt in accordance with the instructions specified in the enclosed Notice of Nonconfonnance.. We will consider extendirg the response tire if you can show good cause for us to do so.

We responses requested by this_ letter and the enclosed Notices are not subject to the clearance p.Warcs of the Office,of Rmagcrent and Budget as required by the Papelvork FMuction Act of 1980, Public Law No.96-511.

In accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of this letter ard its enclosures will be placed in the NRC Public Document Room.

If you have any questions concernirg this _ inspect. ion, we will be pleased to -

discass them with you.

Sincerely,

[

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Imif J Norrholm, Chief Verdor Inspection Branch Division of Reactor Inspection and Safeguards Office of Nuclear Reactor Regulation

Enclosures:

-1.

Notice of Violation 2.

Notice of-Nonconfonance 3.

Inspection Report _99900104/92-01 83

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DK2DSURE 1

)KTTICE OF VIO1ATICt1 Westinghouse Electric Ccepany Docket No.: 99900104/92-01

Nuclear m w ts Division Pensacolt

- D2 ring an inspectim conducted April' 6 through 10, 1992, at the Westinghouta Electric Otmpany's (li's)- Nuclear Otmponents Division (WPP) facility located in Pensacola, Florida, the U.S. Nuclear Regulatory Otmnbion (NRC) inspection j

team identified violations of IF requirements. In accordance with the Section III.of Appendix C, " General Statement of Policy and Procedure for NRC D1foronment Acticos," of. Title.10 of the Cbde of Federal Reculations, Part 2 (10 CFR Part 2) (1990),l^.he violaticos are ' listed belcw:-

-A.-

- Section'21.21,- " Notification of Failure to Ocmply or Dcistence of a Defect and its Evaluation," of 10 CFR requires, in part, that each corporation subject to the regulations adopt. am vr iate pccslures for either evaluating deviations and failures to comply, or informiry the licensee or purchaser of the deviation or failure to occply.

Contrary to. the above requimnents, WPP had not implemented adequate -

procedures to addmss the evaluation of deviations and failures to oceply. On December 10, 1991, WPP was rotified by Westirghouse Product

~ Assurance transmittal PA-91-1766 of substantive revisions to 10 CFR Part 21"and provided a.new poster that-addressed the mvisions 'and satisfied-the posting requirements of-10 CER 21.6.

However, WPP had not revised its walures, required by 10 CFR 21.21, to address the r

substantive revisions to 10 CFR Part 21 that hem _ma effective on-Octe h 29, 1991.(99900104/92-01-01).

p

'Ihe NRC has classified this violation as Severity Imvel IV (Supplement VII).

B.

Section 21.31, '" Procurement th'mts," of 10 CFR Part 21, requires, in part,: that each corporation shall assure that each procurement docunent-

_.for a basic ocmponent specifies, when applicable, that the provisions of 10 CFR Part 21 apply..

1 Contrary to_ the above, the followirg three purchase orders (tos) are exarples in which WPP failed to-invoke the reportirg requirements of 10

~

' CFR Part :21 in safety-related IOs.to_ subvendors of basic ccmponents-or materialsi L (99000104/92-01-02).

PO-PE 72056-M-CD,' issued to Wisconsin Centrifugal, Inc. for safety-related flow limiting inserts for Unit 1 of the North Anna L

Power Station-(NAPS-1). -

l-E 84 o

PO 7075590, issued to Piping and Bpipwnt, In::. for the calitratico of an Ashcroft pressure gatge used by WPP for calibrating measuring ard test equipTnt (1%TE) to be used durirg the hydroctatic testing of replacemnt stea:n generator surarsenblies (RSGS) for laps.

PO PE-21955-B,JB-0-000, issued to MDS Calinraticn laboratory for the calibration of 1%TE used during the mnufacture of RSGS for 1RPS.

The lac has classified this violation as a Severity luvel IV violation (Suppleront VII).

Pursuant to the provisions of 10 Cm 2.201, WPP is hereb/ reqaired to subnit a written statamnt or explanation to the U.S. }A> clear Regulatory Ccrrnission, ATIN: temnt Control Ded, Washington, D.C. 20555, with a o:py to the 011ef, Verdor Inspection Branch, Division of Reactcr InsW ion ard Safoguards, office of Nuclear Peactor Rcqulath, withir.

30 days of the date of the. letter transmitting this Notim of Violation.

This reply should be clearly m rked as o " Reply to Notice of Violation" ard should ircitde for each violation (1) the reason for the violation, or if contested, the basis for disputirg d>e violation, (2) the corrective steps that have been taken ard the results addeved, (2) the corrective steps that will be tak n to avoid further violations, and (4) the date when full empliance will be achieved. Nhere good cause is shown, the lac will consider extendirg the response tim.

\\

Dam! :.c Pockville, Marylard this ~L day of M6A

, 1992.

/ 85

l DiCLDSURE 2 IUTICE OF IniCCti1VRMAliCE Westirghoase F.loctric Ocqnny Docket No:

99900104 Nuclear Otrponents Division Persacola Plant During an inopoction conducted at the Westinghouse F.loctric Oc y ny Nuclear Otrponents Division, Persacola Plant (WFP) facility in Pensacola, Norida on April 6 through 10, 1992, the inspecticn team from the U.S. Nuclear Pegulatory Commission (NRC) deternincd that certain activities were not cmducted in accordance with NRC requirements, which were contractually inpcsod cn WPP bf purchase orders frcn NRC licensees. The NRC has classified these items, as set forth-belcu, as nonconformnoes to the requirtments of Title 10 of the Cbde of Fed _gralJmulatiorn, Part 50 (10 CFR 50), Appendix B, inpocod on WFP by contract and the supplemental reqairements of NRC licerscas.

A.

Criterion V, "Imtruction, Procedures, and Dravirgs," of Apperdix B to 10 CFR Part 50 and Sectico 5.0, " Instructions, Procatares, and Drawirgs,' of the WPP Nclear Quality Assurance Pecgram Partaal, Revision 21, dated Och 9,1991, rcquire in part, that activities affcctirq quality be acocuplished in accordance with instructions or proccdures.

Section 12.0, " Control of Measuring ard hstirg Riuip:ent," of the WPP Nuclear QA Program mnual rcquires, in part, that calibration services be subcontracted to an approved gauge / calibration laboratory gaalifled in acconisnce with Section 7.0, "ocotrol of Purchased Items ard Services," of the QP Program Manual.

7 Section 4.0, "Procuremult Docunent Coritrol," of the NPP Nuclear QA Program Manual requires, in part, that purdase orders (ios) for qsslity related items and services be reviewed by Quality Assurance Engineering to assure applicable design, Code, ard contractual quality requirenents

're ircitrbd, and core purchascd from approved suppliers for the scope of work.

Section 7.0, of the WPP Nuclear QA Program m nual requires, in part, that WFP receivirg inspection verify that the certificate of conformance (CoC) received for a safety-related item or service irdicates that the item or service was produced under a quality prcgram reviewed ard approved by WN.

Contrary to the above the follcuing Pos for safety-related calibration services were not reviewed bf Qualitv Assurance; did not invoke the requirements of Appendix B to 10 CFR Part 50 or the reportirg requirements of 10 CFR Part 21; were procured from suppliers not apprcved for the assigned qualit1-related work scope; ard the CoC received was not reviewed by WPP receiviry inspection and did not irdicate the QA nanual under which the calibration service was protaced.

86 l

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l l

1)

PO 705590, isoxd to Piping and B;[uiFent Irc. for the calibration of an Arhraft pressure gatge used to calibrate measurirg and test ocpipent (l%TE) uscd during the mnufacture of replacemnt steam generator snhmmblies (RSGS) for Unit 1 of the }kirth Anna PJwer Station (lRPS).

2) 10 PE2D55-B-JB-O-000, issued to MDS Calibraticn Tahnratcr y for the calibration of a tcog neter used to calibrate PGTE used durity the manufacture of RSGS for IMPS (99900104/92-01-03).

B.

Critericn V, "Instructicos, ProoMures, and Drawings,' of Apperdix B to 10 CFR Part 50 require, in part, that activities affectirg quality be acocruplished in ecoordance with instructions, proocdures, or drawilgs.

Section 5.0, "In=ructicns, Proccdures, and Drawings," of the WPP Nuclear Quality Assurance Program Manual ruquires, in part, that personnel involved in Quality Assurance (m) supam activities perfom activities related to quality in accordance with procedures.

Cantrary to the above,-WPP failed to follow the instnx:tions on the Quality releas1 (QR) Fom and to docu ent a deviation, Documnt Subsittal Form (DSF) 3388, cn CR 168665 (9P900104/92-01-04).

C.

Criterion II, " Quality Assurance Program," of Apperdix B to 10 CIR Part 50 requires, in part, that activities affecting quality be

'acccuplinho3 in accordance with a quality assurance program which shall be documented by written policies, procedures, ard instructions.

Criterion VII, "Omtrol of Purchased Material, Bpigent, and Services,"

of Appendix B to 10 CFR Part 50 rrquires, ir part, that reasures be established to assure that Ivrchased material conforms to procurement documents and that the effectiveness of the control of quality by the suppliers of mterial shall be assessed by the purchaser at intervals consistent with ^.he inportance of the prcduct.

Contrary to the above, during the procurement of materials for the RSGS, the WPP m program did not require the perfonunce of any activities to verify that suppliers, accredited by ASME, were effectively inplementirg their m progrr s.

As a result of this deficiency, WPP did ret perform any r-dits, source inspections, or m terial overchecks of the supplier of '..e flow limiting inserts which WPP riu; plied to NAPS 99900104/92-01-05).

Please provide a written stntenent or explemtion to the U.S.1) lear Regulatory Oxnission, ATIN: Document Control Desk, Wa.Airgta D.C. 20555, with a copy to the Chief, Vudor Inspection Branch, Division a

'eactor inspection and Safoguards, Office of Nuclear Peactor EcgulatL... within 30 days of the date of the letter transmitting this Notice of Nonconformnoe.

1 87

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This reply should be clearly marked as'a " Reply _ to a Notice of Ncoconforuance"

-and should'incitx$e for each nanoanformance:

(1) a description of steps that have been 'or will bo.taken to h4=ct these = itans:

(2) a h iption of steps that have been or kill be taken to. prevent recurrtence; and. (3) the dates your corrective acticra-and prwentive maamures were or will be crmpleted.

b DateditRockville, Maryland-

.this % L

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1992.

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GGTI2ATICit:

Westinhouse Electric Otrpany Nuclear Ocqxnents Divisico Pensaco!s, Florida REFORT 10. t 99900104/92-01 CIEGDilttIDDiCE AIIEEC5:

Mr. B. R. Smelstafs, Plant }bmger Westirrfeuse Electric Ocrpany Nuclear Cdq;onents Division 8301 Secrtic Mj tray g

Pens. cola, Florida 32514 GWII2ALTtRL CatTDCT:

Mr. Jose M. &rtinez, P1xduct Assurance M1mger NUCIIAR DiDUSIRY ACTIVITY:

}hnufactures nuclear cxrpcnents incltdirrj replacenunt steam generator subassenblies (RSGS) for the North Anna Pcuer Station INSPECTICri CCUCUCTED:

April 6 thrurp 10, 1992 h PARID BY:

"Y 6 //8/ 94 L.L. C:fmpbell, 'Ibam Irader Date Reactive Inspect:. ion Section No.1 Verdor Inspection Branch (VIB)

OIRER INSPDCIORS:

W. C. Gleaves, VIB S. M. Matthewz., VIB U. Potaper Sr-VI%

F i

APPROVID:

dA d

G 20-92 Uldis Potapovs, Chief Date Reactive Inspection Sdction No. 1 Vendor Inspection Branch IliSPECTICri BASES:

1C GR Part 21 ard Part 50, Appeniix B INSPIETICU BASES:

'Ib assess the Westirghouse Electric Cmpany, Nuclear Ctr.ponents Division, Persacola Plant (WPP) quality assurance (CA) program ard its irplenentation in selected areas sudt as the control of (1) the design process ard interfaces, (2) purchasc:1. raterial ard services, (3) special processes, and (4) inspection and test.

PIRC SITE APPLICABILITY:

North Anm Power Station, Urtit 1, ard other pressurized water reactors. 89

l l

DISPECTICri SOf%RY

- 1.1 Violations Cbntrary to the rapirecents of 10 CJR Part 21, We:?.irghcwe Electric Corporation (W) Huclear Ctrprents Division, Pensacola Plant's (UPP's)

Proocdure IQ-02-OO7, Revision J, d3ted }%y 23,1991 ard the u.u ruate Westirgh ase Electric ChTnny's (H's) Prccedure OPR-19, Revision 3, dated Novenber 21, 1988, failcd to inplemnt the requirencnts of the October 29, 1991, revision of Faction 21.21(a) of Title 10 of the Code of fpderal PmulatioIn (10 CIR) Part 21 (Violation 99900104/92-01-01, see Section 3.2 of this report).

In addition, contrary to the rtquire:r_nts of Section 21.31 of 10 CFR Part 21 three purchase ottiers (FO's) to suppliers of basic ccrpone.h failed to invoke C

the reqairements of 10 CFR Part 21 (Violation 99900104/92-01-02, see Sections 3.4.3 ard 3.8 of this report).

1.2 Norconf.ormnots Contrary to Sections 4.0, 5.0, 7.0, and 12.0 of the WPP Nuclear QA Marual, two p

Pos for safety-related calibration se2 Vices were rot reviewod by WPP Quality Assurance; did not irrxAe the rcquireronts of Appendix 3 to 10 CFR Part 50; were procured frcra suppliers rot approved fx the assigned quality-relatal work scope; and the certificate of conformnce (CDC) received frcxn the supplier was rot reviewed by WPP receivirg inspection aid did not irdicate thc QA mnual under which the calibration scivioe was performed (Nonconformrce 99900104/92-01-03, see Section 3.8 of this report).

Contrary to Criterion V of AppeJXiiX B to 10 CFR Part 50 ard Section 5.0 of the WPP Nuclear QA Prcgram Marnal, WPP failed to docunent a deviation, Docunent Sulnittal Form (DSF)- 3388, on Quality Pelease (QR) 168665 (Norconformance 99900104/92-01-04, see Section 3.4.1 of this report).

Contrary to Criteria II and VII of Apperxiix B to 10 CFR Part 50, durirg the procurenent of mterials for the VinJinia Electric Pcwr Ctrpany (VEPCD) replacencot steam generator mhmblics (RSGS), the WPP Nuclear QA Program Manual did rot require the performnoe of any activity to verify that suppliers, accrcdited by the Arctican Society of Mechanical Dgineers (ASME),

were effectively inplerentirq their QA program (Nonconformnce 99900104/92-01-05, see Section 3.4.3 of this report).

2 STATUS OF PREVIOUS INSPFETIOti FD1DDIGS There were no open findirgs frcra the previcus inspection report. 90

3 DISPIITIGI FINDDGS AND 01EER O} fens 3.1 1htrance and Ddt Meetim Dur319 the entrance meeting en April 6,1992, the U.S. fMclear ikxjulatory Comission (NRC) inspectors dimme.03 the scope of the insycticn, outlinnd areas of concern, Nd established interfaces with WPP.ramgement and staf f.

Durity the exit utd.irg un April 10, 1992, the NRC inspectors dLuissed their f 3Jdirns ard concerns with WPP'a maragement and staff.

3.2 Impecticn for Omqliaroo with 10 C'!V Part 21 he NRC inspectors reviewed WPP's inplementatico of ard ccepliance with the requirenants of 10 GR Part 21 for VEMD's procurement of RSGSs ard cther equipment supplied in accordance with the VFPCD Specificaticn NAP-0001, "Specificatice for Repairtd Steam Generators, North Anrn Power Station, Unit 1," Revis>.' 3, dated September 6, 1990. During this review, the NRC insW., it. anti Jiui two iru.tnes in aich WPP's activjties were in violation of NRC /t.quirend m e violations are di m m wl separately as follows.

3.2.1 Irplennntation of 10 GR Part 21 Procedures he NRC inspectors reviewed WPP's Procedure FQ-02--OO7, " Identification ard Reportiry of Substantial Safety Hazards, Significant Deficiencies, ard Unreviewed Safety Questions," Revision 3, dated May 23, 1991.

Procedure PQ-02-007 provided the WPP Safety Review Cammittee (SRC) with the responsibility for evaltating deviations and failures to ccrply ard reportiry the results of its evaluation to the corporace Westirghause Electric Ompany Encryy Systems Business Unit (ESBU) SRC. S e ESBU/SRC had responsibility to evaluate tre results of WPP ERC's evaluation ard detemine 1,.hether the deviation or failure to comply is reportable. WPP, hcw ver, failed to revise Procedure PQ-02-007 to address the revisiurs to 10 GR Part 21 that became effective on October 29, 1991.

S e NRC inspectors determined that the H ISBU/SRC's procedures for evaluating deviations ard failures to emply were contairyxl in Operatire Procedure OPR-19, " Identification ard Reportirg of Substantial Safety Hazards, Significant Deficiencies, and Caresolvtd Safety Questions," Revision 3, dated November 21, 1988, of Section III, "Related Information," of the E I N 9550,

" Energy Systems Business Unit Quality Assurance Program for Basic Ompsrents,"

Revision 21, dated August 15, 1991. W e NRC in g;octors reviewod OFR-19 ard deterednod that E also failed to revise OPR-19 to address the Ot h.r 20, 1991, revision of 10 GR Part 21.

E, through WPP's staff, rade a verbal comitnent to the hHC inspectors that E woul-1 review ard revise OPR-19 and WCAP-9550, as applicable, b/ the end of May, 1992. WPP committed to review its Procedure PQ-02-007 for any inpact the revised OPR-19 or WCAP-9550 may have (Violation 99900104/92-01-01). I 91

3. '. 2 Incorporation of 10 CFR Part 21 in Procurment remnts i

201RC inntors reviewed several IOs for basic ocrponentA and fcurd three TOs f or tanic ocquents in which WPP failed to invoke the requiroments of 10 CFR Part 21 as requirtd by Section 21.31, " Procurement rM_w'nts," of 10 CTR plolation 99900104/92-01-02, see Sections 3.4.3 ard 3.8 of this f

report).

3.3 Wsicin Contrgl

%e imC inspectors evaluatM design activities performod by WFP f a cmpliance with Crite ia III, " Design Cmtrol," of Apperdix B to 10 CFR PaI: 60, and with the requi2cments o'

.ttachment 3, Specification NAP-0001, to VEPOD's purchase order (IV) BNr273343, with Change On3er No. 3, dated Febmiry 20, 1990.

c itet of the design hw'nts rcquired by Specification be-0001 such as the ASP 2 code,Section III, stress ard design reports, the therml-hydraulic amlysis report, and seismic qualification summary for the RSCS were not finalized, but were in various stages of capletion. Design inputs for the ccq. uter program generati19 the required amlysh were reviewed as well as preliminuy design documents. We follcuire design activities performed by WPP wera reviewed by the NRC inspectors.

3.3.1 Steam Generator Stress Report Design Inputs-Nozzle Ioads

%c inrpoctors reviewd Volume 3 of hWEP-9106, "ASME Section III Stress Report For the Primuv Nozzle Amlysis," Revision 0 (Draf t), with a tantative issue date of July, 1992. We NRC inspectors determined that the Seismic Design Base Earthquake Fx design inprt value was correctly listed and the resultirg design output value on Page 5-9 of 7tchnical Record Book, 'IR 0520, appeared to be carrect based on a hard calculation performod durire the ingxx: tion by WPP.

We insph also reviewd Volume 11, " Minor Shell Taps Analysis," of bMEP-9106 and determincd that the 1.0 ir4h auxiliary nozzle load design inputs

- for the steam gercrator drain had been correctly identified. A)though these values differed fram the loads listed in VEP00's Specification NAP-0001, WPP presented the inspectors correspondence fim VEPCD alorg with a ccpy of Revision 4 to Table 2.14, " Steam Generator Insign Icads, Auxiliary Nozzles,"

of NAP-0001 that reflected the design inputs used by WPP.

3.3.2 Tube Suppart Plates he NRC inspectors reviewd data used for the f]cu-irduced ard turbulence-induced vibration amlysis. Section 2.8.6, " Tube Supports," of VEPCO's Specification NAP-0001 re@ ires that the diametrical gap between tubes and U-bend supports shall average 0.005 inches or less.

Exhibit B-IMP of WPP Process Specification 80307 RN, " Steam Gernrator Anti-Vibration Bar Assembly ard Inspection Specification 51F-1dvance Dasign," Revision E, controls the prem" of determinity the diametrical gap. Approximtely one out of six column gaps ard every outside gap was measured ard ascraged. A total of 1006, 1032, ard 1133 gaps were measured for Steam Generator 11970, 11972, and 11971, respo tively. Average gaps of 0.00231, 0.00241, and 0.00253 inches were

-p 92 l

oltaired for Steam Generator 11976, 11072, and 11971, respectively. Se }&C fnspectors determined that these average diametrical gaps not the respirements 00 Specification IRP-0001.

3.3.3 Stran Nozzle Flcw Limitirq Davice no 15C inspectors reviewtd ard diso.Issed with WPP personnel tN design and the performroe requirenenta fur the steam nuzzle flcu limitirg device. 'Ihe IEC inspectors also reviewed the nozzle design for ocx:pliance to VEPCD Specification NAP-0001. Of specific interest to the NRC inspectors were performance requirements listtd in Table 11, pages 17,18 ard 19 of 1RP-0001 shich are derivcd frcra the North Arra Power Station (!W) Updated Fiml Safety Analyses Report (UFSAR) Tables T6B-12, T6B-21, ard T6B-22, respectively.

During the course of the inspectico the NRC inspectors reviewed VEPCD Specification NAP-0001, Section 2.8.12, "Immries aid 43are Parts,"

Table 1.1, ard the "}hin Steam Line Break Curve 13," on pages 137 ard 150 of WPP Specification 409A99, Revision 2, datcd November 30, 1991, and a hard caledatibn performed by WPP engineering perrennel nq or about April 8,1992, titled " North Anna RSG Steam Line Break Initial Water level at Top 'Ibbe Support Plate llot Stardby Initial Condition." 'Ihe NRC inspectors ruted that the afo crentioned hard calculatico vas rot final ard was subject to additic.al review and approval. The NRC inspectors determined from the calculation that the stardardized design of the steam nozzle flw limitiry device was adequate to reek the requirements of VEPCO Specification 1&P-0001.

Hcuever, the NRC inspector; ntfd that the' steam rozzle ficw limitirg device's f1w limitiry performn appearcd to be approxirately an order of magnitido hicther than what wa-11uired by VEPCO Specification IRP-0001, Table 1.1.

'Ihe NRC inspectors were ' wormed durity the inspection that a recting between VE'00, kTP, and VEPa)'s design agent was scheduled at a later date to discuss these performnce characteristics.

3.3.4 Ocrputer Sof tvare

'Ihe NRC inspectors reviewed tNo computer programs to detennine whether or not kTP's programs roccived an irdeperdent review and validation. Orputer program kTrAN, " General Purpose Finite Element Analyses Program,"

Version 0-12-10B, datcd June 29, 1991, used in various analyses such as the steam generator stress report, was reviewed by the NIC inspectors. WPP independently validated the WDCAN prcgram, ard performs a yearly validation of the p.w, by comparing program results to approxima*ely 400 problem sets which includes hand calculations ard results frczn tests and o+her pt grams.

If the WDCAN program is revised, WPP validates the revision by checking the revision against a sutset of approximately 90 of these solved problems.

khTF-8648, "Cbnputer Program FIU/IB: Flcw-Irduccd Vibration of a Multi-Span Stnictural Member," Revision 1, dated January,1990, was also reviewcd.

Methods used to irdependently validate WNEP-8648 included the use of alternate calculations. 1 l

93

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Prirary Side lif rectatic hst Proordure d

3 3.5 2e IUtC inspectors revicved WPP prcoadare DG'5956, "It/droctatic Nst of Imr Steam Comratar Primry Side," Revision 2, datId Septaber 17,1991, ard l.

-informtical copies of proculares QIP 16. 3388 ard QIP No. 3389, both titled, i-

"AIP Izuer Stea:n Gercrator Primry Itydrotest," both datal June 1,1988, and i

ccrpared them to rtquirumnts of the 1996 Dlition of the ASME Ocde, i

Section III, Divisicri I, Sulxectical NB, which WPP was craltractually obligated to met.- 'Ihe 101C insp,ctors determined that the hydrostatic test pressure, hold tine, ard test taperature cpecifiod in 06"S956 met applicable requiremnts of the 1986 ElitMK) of the AfME III Code ard VEPCD SpecifJcation NAP-0001. The lutC inspectors rottd that Specification NAP-OOO1 specifled the L

primry side hydrotest pressure to be 3107 pcig at 70*F shile D#P-5956 allcued i

taqct Mrds for trzperatare and pres. care. Idiiticmily, the NRC aspectors I

reviewed ard discunr4d with WPP personnel the Q7. vendnr file includirg the supplier evaluation form, dated Jamary 29, 1992, for Dowell SchlumbcIger, Inc. (DS), the subcontractor perfomirg the hydrostatic test. S e IURC inspnctors noted that this service was a nuc r sWcy related service and that 10 CFR Part 21 was -invoked in the draft 10 tr DS.

7he IURC irspectors were inforrrd that this service would be perfcInd urder WPP's QA program ard that WPP personnel *.mld oversee all subcontractor operations. W e 101C inspecto:s deterrtincd that WPP had correctly cranslatcd hydroccatic test requiremnts frm the VriCO specification into their prccalures.

3.4 Control oLMaterigleiParts, ard Conrcrnat;s l

l 2e IUlc inspectors reviewal the procurement ard receipt irqwction of mterial I

for several mjor items of the RSGS to determine whether or rd. WPP h3d established ard inplemental adequate masures to control mterial, parts, and crrponents. me area reviewd we described separately as follcus.

3.4.1 Primary Nozzle Safe-Dd Fo girgs he NRC inspectors reviewed the documntation for the pi b ary side nozzle l

safe-end forgims' purdlar4d by WPP from Irrape Forge, Inc. (Lonape) iri l_

accordance with the requirnents of WPP Specification A336c01, Revision B, and l

ASME Code,Section II, Material Specification SA-336, Class F 316 IN, with special requiremnts. _ Icmpe was qtulified by hTP to supply' material in l~

accordance with the rcquiremnts of MME Section III, -NCA-3800.

Review of the Ienapo mterial certification for serial number (S/N) 494G-1A ard S/N 494G-1B forgings ard archive camples shcued the material grain size to u

- bc reportrd in the rarge of 3.0 to _3.5 khile DP rpecification 336C01 reg.lires the material grain size to be 4.0 or greata ' he grain size is finer as the number irmues). 3his mterial had teen accepted bf WFP on Quality Release l

(QR) 168665 as meetirq the specification rcquirenents with to deviations

~

l:

notrd. Review of additional docunentation irdicated that tempe sutrtitted a j_

. Docatment Subnittal-Fom (DSF) 3388 requestiry a deviation frun the grain size l'

requirement urd that this request had been approvcd by WPP.

l

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'Iho ' certification block of m 168665 contains the statemnt, "We certify that

. for_ the equipnent ard material released, all contractual requiranents have been w t._ Approvcd deviations, if any, are listed above." Since the grain size deviaticn was rot listed cn OR 168665, it appears that the @ eminmr signirg this certification was rot aware of the DSF ard failed to observe that the mterial did rot meet s;rcification. 'Ihe deficiency in certification of QR 168665 was identified as an exarple of ncnconfomance (See Noncanformance 99900104/92-01-04)..

It was noted _that a.h DSF had been approvcd on the safe-erd mterial allowing comer content to be reported for infomation caly, h DSF indicated that the actual copper content of the noterial was 0.20 percent uttile specificatico 336C01 limited copper to 0.15 percent. A review of WPP m terial sp'cificaticn 336C01, Pcvision B, Pa mver failed to identify any restrictions on the copper content of this mterial ard the basis for the suhnittal or for the approval of this DSF could rot S identified durim this insIxx: tion.

3.4.2 Steam Generator Food Ring

'Ihe NRC inspectors reviewed the documentation fer the steam generator feed ring purchased by WFP frca Tioga Pipe Supply, Inc. (Ticga) in accordance with the reirements., of WPP Specification A106C04, Revision D, ard ASME Code,Section II, M1terial Specification SA-106, Grode B, with special requirements.

Tioga is an A9E accredited waterial supplier. Tioga pJrdused the material frca Dalmine (foreign sumlier) ad upgradcd it to /eME requirements in acconian e with ASME Code, Sectico III, NCA 3867.4(g). Tioga also performd or_ contracted for special tests to meet WPP Specification A106C04.

Review of the WPP atdit file for Tioga iniicated that before aid during this procurenent cycle Ticna had been placed and mintained on the qualified supplier' list on the basis of their ASME Quality Systems Certificate (QSC) ard an annual performance review by WPP.. 'Ibere was n. docunentation to irdicate that Westirghouse had performed any atdits to evaluate the implementation of Ticga's % program in connecticn with this procurement. Sursoquently the NRC inspector was shown a repcrt of an atdit of Tioga performd on June 6,1990, by the WestiJghouse Electro Mechanical Division, Cheswicx, Pa. However, there was no evidence that the resalts of this atx11t had been considered in connection with this procuturent. - It was noted that the last entry in the

-_ Tioga atdit file (February 21, 1992) stated, "Atdit req 11 rod durirg life of che ASME Certificate if supplier is used."

3.4.3 Flow Limiting Inserts

'Ihe NRC inspectors reviend the documentation for the f *.cw limiting inserts purtMsed by WPP frcn Wisconsin Centrifugal, Inc. L the reqaircrents of WPP Specification M494C01, Revision C, and the 1986 Fdition of the ASME Oode, Sectior. III, 9*wstion NB, Division 1, Class 2.

Wisconsin Centriftgal, Inc.

is an AME accredited Material Marraf acturer.

WPP PO PE 72069-M-GD, dated April 29, 1991, for this, mterial requirai the seller to maintain a quality assurawe prcgram consistent with the applicable 95

o portions of 10_ CFR Part 50, Appendix B and particns of the ASHE Code referencxxl in this onler. me purchase order, however, failed to invoke the

'r@cnonts of 10 CFR Part 21 as required by tMt regulaticn.

Fr11ure to invoke 10 CFR Part 21 in a 10 for a ocrponent is cited as a violation (See Violation 99900104/92-01-02).

Review of the Westisghouse audit file for Wisconsin Centrifugal, Inc.

indicat d that the vendor had been placed and mintairvxl cr the qualified supplier list based on their ASME QSC %cre was no document:ation to irdicate that Westirghause had perfomed any quality assurance program implementation audits of.this supplier. Section 3.5 of this report providcu additicml disalssicn on WP's controls for manufacturers and suppliers of Ctda material, i

me failure to perform program implemntation audits is identiflod as a j

nonconformance (Sac Nonconfomance 99900104/92-01-05).

3.4.4 h bes WPP issued purchase order (PO) PE-70165-H-SA, dated Dmrde 11, 1989, to AP Sandvik Steel of Sandvik, Sweden, for the steam generator tutes and the tu k bending. We tube material was ordered to (1) the material specification of ASME SB-163, Nickel-C3uuninu-Iron Tubing, Alloy 690 (UNS N06690), for seamless condenser _ard heat excharger tubirg; (2) the requirements for Class I caponents in accordance with ASME (bde,Section III, ' Division 1, and Code Case N-20; (3) Apperriix B to 10 CFR Part 50; ard, (4) the rupcrtiry requirements of 10 CFR Part 21.

Le NRC inspectors reviewed the tube material data packages with the WP metallurgist, who was also involved in WPP's supplier qualification of Sandvik Steel and sairce inspection durirg the tubing manufacturing. Sandvik Steel held ASME QSC-457 as a material manufacturer. Wrough photanicrographs taken frcan a sar:ple of tube mterial from each heat, lot, and themal treatuent, WPP verifled that the grain bourdary precipitaticn of chronium carbides met the visual stardards for corrosion resistance. 2e data packages reviewod L

contained the test reports required for ASME Cbde,Section III, Class 1, ard met WPP's mterial' requirements.

3.4.5 Tube Plates l.

ji WPP issusd PO PC-69905-M-CD, dated November 15,19C9, to Kobe Steel Ltd.,

Takasaco Plant, for the steam generators' tube plates (tubesheets). Se tube L

phte material was ordered to _(1) the mterial specification of ASME SA-508, l:

Class 3a for quenchcd ard tempered vacuum-treatal steel forging; (2) the ne! rements for Class 1, cx:mponents in accordance with ASME Coda, Section d

L III, Division ?, -(3) Apperdix B to 10 CFR Part 50;- (4) ard the reportirg requirenents of 10 CFR Part 21.

We 1C4C inspectors reviewed the tube plate material data packages. Kobe Steel Ltd., hkasago Plant,- held ASME QSC-215 as a aterial manufacturer ard WPP subcontracted source surveillance to Inteco Japan Limited. W e data cackages L

revievcd _containod the test reports required for ASME Code,Section III,_

- Class 1,. ard met WPP's material requirements. 96

1.-

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3.4.6 Barrel Forgirgs WFP issucd PO PE-69967-M-CD, dated Novcenter 22, 1989, to Japan Castirg ard Fomirg Corp. (JCFC), %tata Steel Works, for the barrel forgirgs. Two seamlens inrrel forgirgs were joimd together by a circumferential weld to fom the pressure-bourdary shall section of the steam generator's secordary side. S e RSGSs for VEPOO vere the first units WPP had manufactured without longitudiral weld-scans in the pressure-bourdary shell section.

20 seamless barrel forgirg material was ordered to (1) the mterial specification of ASME SA-508, Class 3a for quenched ard tapered vacuum-treated steel forgirg; (2) the requirements for Class 1, cmpcnents in accorrhnce with ASME Code,Section III, Divisicn 1; (3) AppeJdix B to 10 CFR Part 50; (4) ard the rerui.irn requirements of 10 CFR Part 21.

ne NRC inspectors reviewed the barrel forgim material data packages. JCFr held ASME QSC-215 as a material manufacturer and WPP sub;.ehacted source surveillance to Inteco Japan Limited. De data packages reviewed contained

~ the test reports required for ASME Chie,Section III, class 1, ard met WPP's nuterial requirementr..

3.4.7 Transition Cone Forgirq WPP issued IV PE-69906-M-GD, dated November 22, 1989, to Kobe Steel Ltd.,

Takasago Plant, for the steam gacrators' transition cone forgings. he transition cone forgirg mtcrial was ordered to (1) the mterial specification of JGE SA-508, Class '3 for querched ard tempered vacuum-treated steel forgirg; (2). the requirements for Class 1, cmponents in accordance with ASME Ocde,Section III, Divis.icn 1; (3) A;pendix 9 to 10 CFR Part 50; (4) ard the

-reportirg requirements of 10 CFR Part 21.

We IRC insp:ctors reviewed the transition cone forgirg material data packages. Kobe Steel Ltd., Takasago Plant, held ASME QSC-215 as a material manufacturer ard WPP subuv.irected source surveillarce to Inteco Japan Limited. Se data packages reviewd contained the test reports required for

~ ASME Code, Sectico III, Class 1, and met WPP's material requirements.

3.4.8 Channel Head Forgirgs WPP issued PO PE-70164-M-GD, dated [*mNr 8,1989, to the Japan Steel Works, Ltd., Maroran Plant for the steam generators' channel head forgig s.

Japan Steel Works, Ltd., Muroran Plant, held ASME Certificate of Authorization,'

N-2725, as a cmponent manufacturer and WPP subcontracted sorv;e surveillance to -Intaco Japan Limited.--- 2e channel head forgig material was ordered to (1) the material specification of ASME SA-508, Class 3 for quenched ard tempered vacuum-treated steel'forgirg; (2)' the -requirenents for Class 1, ccrponents in acnordance with ASME Ctde,Section III, Division 1; (3) Apperdix B to 10 CFR Part 50; (4) ard the reportirg rcquirements of 10 CFR Part 21.

' he NRC inspectors reviewcd the channel head forging material data packages and determincd that the data packages aviewed contained the test reports and 97

the b2 data rrport requirtd for AmE Code,Section III, Class 1, and met h?P's mterial requircrents.

3.4.9 Anti-Vibration Bars (AVBs) kPP issued FO PE-71347-M-CD, dattd October 11, 198), to Crtrible Specialty tbtals f or the anti-vibration tar coil raterial. The anti-vibratico bar ooil mterial was ordertd to (1) the mterial specification of AME SA-479, Type 405 (LNS S40500) stain 1 css steel hot-rollcd round bar; (2) the requirenents for Class 1, corponents in accordaroe with AME Ocde,Section III, Division 1; (3) Arpendix B to 10 CFR Part 50; (4) ard the reportirg requirc=cnts of 10 CD1 Part 21.

kTP issued FO PE-71348-M-G dated July 16, 1990, to Telodyne Pittsburgh 7tol Steel (Telodyne), tbnaco, FA for cmversicn of the coil raterial to the fiml shape, size, vacuun-anneal, and cut straight lergth. Durirg Telcdyne's mnuf acturirg process, hTP cLiscovered that Telcdyne had not been properly qualificd as a 10 CFR Part 50, Appen11x B supplier and that Teledyne had not prcperly identifiod ccrplettd AVBs. WPP iraued a Marrafacturing Ibid For Corrective Action (!OiCA), MHCA No. 048, dated June 21, 1991, perdirg inplementation of hlodyne's corrective actions correrning noterial traceability and quality prcgram. Telodyne developed Coality Assurance Plan, TP-QP-002, Revision 1, datcd August 12, 1991, that accordirg to WFP addressed its correrns. VDOD closed its Supplier Narconformrce Report No. 236 on August 27, 1991, after assurirg that the AVBs supplicd by hledyne could be used as-in ard that the hledyne issue represents a unique corcern, not a WPP prcgrammatic ca.wrn.

The imC inspectors' review of hledyne's mnufacturing process observed that the final vacuum-anneal (three hourt minimum soak tire at 1450'F-1500'F then sltu cooled at 50*F per hour to 1100*F follcutd by still-air cooliry to rocn temperature) was perforW while the AVB mterial was in the c 311 form.

Hardross values were obtained to determire the mterial's sof tness.

(Since the AVBs are in contact with the generator's tubes, hantra values for the AVBs, similar to the hardness values of the tubes, is desirable to minirize ttion degradaticn.) Itwever, the hardness values wre not taken after the AVBs wro lightly cold workrd durity the straightenirg process. The 1mc insp>ctors observed that WPP had not perforrai a prcduction process cpilification of Telodyne's mnufacturirg procera to ensure that the corpletcd cut length AVBs have the corrcct hardness values, h?P issued 10 PE ~'2029-N-GD, clated January 28, 1991, to SL Modern Hard Cunne to fana, cut ard grird, perform the finial inspcction, and clean ard package the AVDs for shlinant to hTP.

3.5 thrnfactumq_and_J2roliers g{ ASME Code Patedal 3.5.1 The IEC inspcctors reviewed a sample of hTP's audit / surveys uscd to qualify mterial suppliers ard raterial ranufacturers to the NCA-3800 requirements of AME Code, Scction III, 7he IRC inspectors observed that the reparts of these audit / surveys failed to adcquately document the recessary objective eviderce to support the qualification of the vendor, kTP staff 98

irdicated that its audit / survey chocklist was beirg revisM to a n-- Wte, prcper documentation of the cbjective eviderce evaluated during the audit / survey.

3.5.2 The NRC inspectors determined throtgh interviews with WFP's staff that WPP had rot made it a quality practice to verify that ASME-accredited suppliers of certain A9E Ctde, Sectico III, iters are effectively.

impicmen~c.irg their quality assurance program, hTP's positicn was determined

- to be cxxitrary to the NRC's expectations described in Infor7tation Notice (D1)

No. 86-21, "RD00GGTIQi OF AMDUCAN SOCIETY OF MEDWIICAL DGIhTJRS ACLREDITATIQi PIOGRAM FOR N bTAMP IDIDIRS," datM Mirch 31, 1986; includirg Supplement 1, dated Dw=har 4,1986, and Supplement 2, dated April 16, 1991.

WPP's staff acknowle& ed this contradiction and advised the team that its J

Procedure PQ-01-005, " Supplier Auditing / Surveillance Prcgram," Revision 7, dated April 3, 1992, had been revised to address this concern.

The }mC inspectors reviewod paragraph 6.9.1 of Procai2re PQ-01-005 and determined that WPP had addrescM this ocncern by ocmittirq to audit A9E-accredited suppliers a mininum of cree durirg the life of the three year period of'ASME's accreditation certificate. %e team determined that, even though the otanitwet of paragraph 6.9.1 was an improvement over WPP's earlier position, WPP should review Supplenuit 2 of IN 86-21 ard consider acticns to address the NRC's staff positicos stated in Regulatory Guide (BG) 1.144, Revision 1, "Atriitirg of Quality Assurance Prograns-for Naclear Power Plants,"

and PG 1.28, Revisicn 3, " Quality Assuran Program Requirements (Design aM Constructicn)."

3.6 Cbntrol of Special Processes i

The NRC inspectors reviewed the special processes usM by WPP to mrrafacture the RSGSr. %c results of the review of these special processes ard the

' inspectors' conclusions are described in the followirg sections.

3.6.1 Welding Durirg the corduct of the inspecticn WPP was weldjrg the channel head safe-ems ard the channel head to the tube plate. The NRC inspectors focused their inspection on the welding of the stainless steel safe-end to the channel head primary inlet nozzle, which had becn tuttered with inaannel weld filler 1 mterial.

3.6.1.1 Weldirq Procedure Specification ard Welder Quali_fication The weldirg precedure specification (WPS) used for weldirq the channel head safe-erds was WPS 4533-1, Revision 00, dated July 18, 1990, a gas tungsten arc-hot wire (GTAW-HW) machine weldirg process. The p uculare qualification recorti (PQR) supportirg the qualification of WPS 4533-1 was PQR 8506B, dated February 22, 1990.

Bcth the WPS ard supportirg PQR appeard to be adequate for the weldirg of the channel safe-ends.

The NRC inspectors reviewed the qualifications of one of the weldiry

- operators, Stanp No. 5421, who was weldirg the channel head safe-erd. The 99

qualificaticr1 of the weldint, operator appeared to be adeqate for weldirg the cinnnel head safe-crd.

3.6.1.2 Weld Tiller Material

%e NRC inspectors reviwod the docu~entation for the weld filler mterial, Heat Code IQO3, used to weld the ciannel head safe-erds, he NRC -impectors reviewed WPP PO No. PD2081, dated April 1,1991, issucd to Sandvik Stoel Inc.

for 2000 pcunds of 0.045 irch diarcter DMICr-3 weld wire cri 25 pourd spools.

Other dowrentaticri reviwod irclixkd chemical analysis, certifled mterial test reports, rowivirg inspection reports, weld filler mt^ rial issue tickets, and shcp travellers for veldirg the channel head safe-cnis using the DMICr-3 weld wire. Se Pos and dcmentation packages reviewed oppearcd to

~

be adcquate and to meet the regairenents for Class 1 mterial as spa:ified in SubGection NS-2000 of the AShE Cbde, Secticri III,1996 Dlition, ard VEPCO Specification NAP-0001. It is noted that WPP substituted their Heai:. Code, IQO3, f r the Sandvik Heat Cbde, SNX7321D, for the DMICr-3 weld wire.

3.6.2 Postweld Heat Treatrent We NRC inspecurs reviewed VEPCO's Specification NAP-0001 and determined that ocItain requirements, such as the requirerent to solution anreal the stainless steel safe-crds if subjected to a terpcrature of 800*F or above, were rcquired to be considered durity the conduct of Postweld Heat Treatrent (Mrr) of the channel head to tube plate weldment. Se NRC IILp.h reviewed WPP's FWift prcccdure, flip-5524, " Heat Treatment Procolare Includirg Preheat, Interpass, Gougirg, Hydrogen Bakirg ard Post Weld Heat Treatment," Revision 33, dated January 23, 1992, ard Drawiro No. IDSK 422277E, " North Anru Peplacerent steam Generator Madel 51F Onnnel Head P.W.E T. hernoccuple Lccations," initial issue, approved January,1992, in order to determine if hTP had adequately addrerz.ed the requirerrents of Specification NAP-0001. %e NRC inspectors also reviewed the chcp route rheet for performing the Nfr and WPP Prccess Stocification 830311rr, '*1bbed hbeplate Weldment PWiff Requirements,"

Pevision A, dated January 9,1931.

Follcuiry a review of these dxurents, the NRC IILc.pectors ret with WPP personnel to di-how the IWfr of the channel head to tube plate weldment would be performed. his discussion resulted in the NRC inspectors retirg that presently WPP pmwlures arn drawirgs for performing IWfr of the chanrcl head to tune plate weldment have to provisions to stcp IMfr activities if the therroocuple terperature adjacent to the stainless steel safe-end aporoaches 800*F.

3.6.3 Other Prmw Reviewed (Nordestructive Denination, Cleanliness, Material Identification, ard Wold Filler Material Control)

We NRC inspcctors reviewed three radiographs (Film Nos. 28-30) for the C seam (A to B Barrel) weld on the No.11972 steam generator ard deternined that the radiographs had been correctly interpreted. Se use of soapstore are acetone for mrkiry and cleanirg the safe-erd welds was also reviewed and determincd to be acceptable. We NRC inspectors reviewod the 011Tt for the B Barrel of j

the No.11970 steam gercrator ard verifiod that the raterial heat nunber on the Kobe Steel Ltd. mterial certification, W91074-1, was the same heat number mrked on the B Barrel. Se NRC impectors also observed the issuance of weld 100

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l filler naterial, taperature checks of wld filler mterial holdirg ovens, ard the color ervHrg ard storage of weld filler mterial. %e wld filler sterial control practices observed ameared to be adoquate.

3.7 lbnconformnces ard Cbrrective Actigog he imC inspectors reviewcd WPP's implere sation of Sectiers 15.0, "Ocotrol of Nonconformirg Items," ard 16.0, "Urrective Acticos," of WPP's }bclear QA Program }hnual. We impleentation review for renconformiry items consiAed of reviewing a canple :>f WPP's Material RevicM Reports (NEs), tuterial Rejection Notias (MRNs), and Phterial Disposition Reports (MDRs) issued by WPP and relatirg to VEPCO's RSGS. The implementation review for corrective actions consisted of reviewirg a scrple of WPP's Manufacturirq Hold for Cbrrective Actions (MHCA). Sectico 3.4.9 of this report pmvides an exanple of WPP's MICA.

%e IRC inspectors fcuM WPP's implenentation ard control of ronconformrces and corrtstive actions ccrplicd with its ibclear QA Program tenaal ard de::anstrated an adequate ard active interface with the assigned Authcrized ibclear Inspector for ASME Code,Section III activities.

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3.8 Cdntrol of Meamrenent ard 7bst Etuirmmt The imC inspectors conducted an inspection of WPP's calibration lab with empnasis in the area of reasurirq ard test equipaent (M&TE) that was beirg used for the construc'clon of the NAPS-1 RSGSs. The imC inspectors review d calibration records, WPP audits, and PDs issucd for the follcuiry calibration services.

3.8.1 Calibrations Performed By MDS Calibration L*boratory The imC inspectors reviewcd the calibration receds for the IPO4978-GIAW welder controller, the LPO2559 ard LPD1339 GrAW weldirg mehines, and the HA4925 and HA322 strip durt recorders. The imC inspectors also reviewed the calibration records for the follcuirg mster M&TE: L520 calibration block thster set, IPO 4357 Aarobe amneter, IPO 4586 Columbia Electric torg arreter, and LPO 4948 Wahl ohmneter. W e PRC inspectors noted that all of the aforementioncd masters were calibrated under open PO PE-21955-B-JB-0-00, dated July 14, 1989, to }W ^alibration Laboratory (MDS). The NRC irqxrtors noted that the PO did not > ' >oke the requiremnts of 10 r 3, Part 21 ce Appendix B to 10 CFR Part 50.

D"N review of the calibration records revealcd that contraty to Sectiona 4.0, "Procurerent Document Control," ard 7.0, " Control of Purchased Items ard Servims," of the WPP QA Prcgram lunual, the PO to MDS was not reviewed by QA engineerity to asmre proper contractual quality requirenents were included, ard CoCs supplied with the equiprnt did not irdicata that the servloe was perfomed under a quality prcgram reviewed and approved by WPP. The NRC inspectors rcted that the MDS CoC for LPO 4586 listed the as found cordition of the Columbia Electric tong reter as "OUT OF 70LUV.NCE" ard therefore the roter should have been processed in acxx>rdance with the WPP lbclear QA Program !bnual, Section 12.d.

The team also reviwed the audit report of HDS performed by Gilbert Ccrimoruealth Ergineers ard Consultants, dated June 28, 1989. The audit verified that MDS was qualified 101

to NOA-1 for calibration cnly and was based cn HDS's Supplier Quality Prcgram nmual, S.O.P.88-008, dated Noveraber 4,1988. 7he team ruted that contrary to requirments in the WPP Nuclear QA Progra= K'umal Section 12.0, " Quality Assurance Records," WPP did not have a copy of the approved QA Innual for MDS.

Ackiitionally, the amiit of HDS identified under Aniit ceilist criterion Number 19.0, "Ihe supplier shall be familiar and cmply with Federal Regulation 10 CPR 21, or equivalent," however a hanchtritten entry cn the checklist identifial this as being N/A, with the reason stated, "Does not supply any nuclear, safety related parts or services."

Failure to invoke 10 CFR Part 21, in the IOs for a basic ocrpanent is cited as a violation (See Violation 92-01-02) ard the failure to process the Pos as a sdety-related procurement is cited as a raconfonnnce (See Noncon-formance 99900104/92-01-03).

3.8.2 Calibrations Perfonted By Pipirg and Dpi;nent, Inc.

he team also reviewed the calibration records for the Inster gauge that was to be used bf WPP to calibrate the RSGS primry side hydrostatic test gauges.

The team reviewed Purchase Agrce:cnt (PA) No. 705590, Revision 00, dated October 9,1991, to Pipirg ard Brailrnent, Im. for one Ashcroft pressure puge having a range of 0-80000 psi. Cbntrary to requirements of Appendix B to 10 CFR Part 50,10 CFR Part 21, S e,tions 4.0, 7.0, ard 12.0 of the WPP Nuclear QA Program Emual and Section G.2 of WPP Specification 409A99, Revision 2, the team noted several discrepancies. The supplier evaluation form datal August 9,1991, listed Piping and Dyipment, Inc. as an approved ccreercial grade supplier for fittirgs, pipirq and valves, however WPP obtained a calibrated gauge frcan them for use in safety-related activities. The Ashcroft master gauge was manufactured and calibrated at Dresser Industries Instru ent Division, which did not appear on WPP's qualified suppliers list (QSL) when this PO sas issued.

It is also noted that Piping and Dpl mnt, Inc. ws not

~i an approved distrilutor for Dresser Irdustries Instnment Division, anc.

'at Dresser Industries was not approved by WPP to provide calibration services for the gauge. WPP PA No. 705590 did not invoke the rcquirements of Apperdix B to 10 CFR Part 50 or 10 CFR Part 21 to Pipiry ard Dyiprent, Inc.

Failure to invoke 10 CFR Part 21 in the 10 for a basic cxrponent is cited as a Violation (See Violation 99900104/92-01-02) ard the failure to process the FO as a safety-related procurement is cited as a nonconformnce (See Nonconformarce 99900104/92-01--03). 102

4 PERS0tNEL 02fIACTED WESTDGrESE FHUBUC O'RPORATI(tUNUCLFAR CUf0tTDTTS DIVISICt1

+*

B. Smelstoys, Plant Kvager

+*

J. }hrtirez, Prcduct Assurance Marnger

+

D. Albert, Humn Resources Karager

+

S. Anderson, Purdasirg Manager

+*

J. Allen, Qaality Assurarce Dyincer T. Allen, Dyineerirg }hnager C. Alvarez, Area lunager W. Arrold, Weldig Dgireer G. Bieberbach, Design Ergineerity }hnager R. Bolling, Quality Centrol Inspector R. Burt, Works Engireerirg 'Ibchrtician M. Carpenter, Weldirg Dyineer K. Qun, Design Dgineer

+

C. Fillirgin, Quality Assurarce Dgineer E. Fitzpatrick, Design Dyincer

+

R. Frisbey, Quality Control Mimger D. Furth, Weldirg Dgineer D. Harrod, Metalurgist M. Helbar, Senior Project Dyineer J. Hobbs, ihnufacturirq Dyineer B. Hood, Design Dyineer R. Johnson, ngineerity Analysis Dgineer

+

D. Koko, Ccrmercial Products Quality Assurance Marager P. Largford, Design Engineer

+

L. Lauoie, Production Supervisor

+*

O thchado, Nuclear Projects Mm3ger J. Phtthews, Weldirg 'Ibchnician S. McCrea, Weld Crib Attendant

+*

K. Marritt, Procurement and Records Systems QA }hrnger

+* -

W. Middlebrooks, Dyineering Analysis nim;er

+*

J. }brtara, 'Ibchnical Services Manager

+*

K. Olmstead, Quality Assurance Dyineer

+*

M. Palner, Materials Manager

+*

W. Rosenberger, Oaality Assurance Marager

+*

W. Sanders, ocumunication ard Trainiry }hnager C. Schraishuhn, Manafacturirg }hnager R. Schreider, Weldirg/Special Process turnger G. Sntith, Design Engineer J. Visaria, Engineerirq Analysis Dgineer M. Weatherly, Quality Assurance Dginecr/NDE Invel III

+*

J. Yourg, Quality Assurance Dyineer Atterded the entrance recting on April 6,1992

+

Attended the exit meeting on April 10, 1992 103

=

VITGINIA MCTRIC R%'IR CrHPMN -

+.

G. Clark, Cur.uate Qaality Assurance l'arager

+

L. Carter, Quality Assurarca Dwineer J. Orlando, Quality Assurarce Dyineer L. Spain, Senior Staff Dyineer

+

+

- FACIORY }RTIUAL INSURN1CE COIPMW

+*

W. Jones, Authorized 1Anclear Inspector l

l l

NUCLEAR REGUIATORY COMISSICN

+

L. Norrhobu, Chief, Verdor Inspection Branch, IER

+*

U. Potapovs, Chief, Reactive Section No. 1, VID l

+*

L. Carphell, Reactor Engineer, VIB

+*

W. Gleaves, Mechanical Dgineering Intern, VIB

.+

• S. Matthews, Quality Assurance Specialist, VIB Attended the entrance meetirq on 3pril 6,1992

+

- Attended the socit moeting on April 10, 1992 104

po Mcoq$*$

l l

• y'_

UNiiTD STATES n

E NUCLEAR REGULATORY COMMISSION

[

WASHINGTON. D.C. 20666

\\

/

JUN l'.1992 Docket No. 99901246 Mr. Mike Ehlerman, President Yuasa Exide Incorporated 645 Penn Street P.O. Box 14145 Reading, Pennsylvania 19601

SUBJECT:

EVALUATION AND REPORTING 0F BATTERY FAILURES BY YUASA EX1DE INCORPORATED (NRC INSPECTION REPORT 99901246/92-01)

We are forwarding the report of a Nuclear Regulatory Commission (NRC) inspection performed from April 6 through 8, 1992, at Yuasa Exide Incorporated (Exide), Reading, Pennsylvania, involving activities authorized by 10 CFR Fart 21. The inspection was conducted by Mr. Randolph N. Moist of the Vendor inspection Branch (VIB) of-the Office of Nuclear Reactor Regulation (NRR), and other NRC representatives. An exit meeting was held on April 8, 1992, and again by telephone on May 5,1992, during which we discussed our findings with Mr.1. Baeringer and other members of your staff.

Areas examined during the inspection are discussed in the enclosed copy of our inspection report.

The inspection team a:,sessed Exide's evaluation and reporting of failures recorded during battery qualification testing and

~

service conditions, and reviewed the implementation of Exide's battery qualification program.

The inspection also consisted of a selective review of relevant procedures, representative records, and interviews with engineering and technical support staff.

The performance of Exide in the assessed areas was found unsatisfactory.

The team determined that (1) Ex Ne failed to notify the NRC of the cracking of terminal post radial gland seals affecting Class lE Exide Type GC batteries at the Fort Calhoun Nuclear Station, and (2) Exide failed to notify the NRC or licensees that failures during qualification testing of Type GC and GN batteries had reduced the qualified life of the batteries from 20 years to 4 years.

In addition, Exide's field modifications to repair the cracking of terminal post seals were inadequate.

As a result, failures of Type GC batteries occurred at Fort Calhoun.

It is our understanding that Exide took corrective action by implementing design changes and field modifications to the Type GC and GN batteries; however, Exide could not demonstrate through documentation whether all batteries with defective post seals had been repaired in nuclear plants.

Exide also did not take measures to provide a unique identification or record 105

l Mr. Mike Ehlerman for the modified batteries so as to provide customers traceability to defectiva batteries.

Specific findings and references to the pertinent requirements are-identified in tha enclosed Notice of Violation and Notice of Nonconformance.

Pursuant to the provi: inns of 10 CFR Part 2.201, you are required to respond to this letter and should toilcw the instructions specified in the enclosed Notice of Violation when preparing your response.

in your response, you should docuinent the specific actions taken and any additional actions you plan

{

to prevent recurrence. After reviewing your response to thi: Notice of Violation, including your proposed corrective actions and the results of future inspections, the NRC will determine whether further NRC enforcement action is necesaary to ensure compliance with regulatory requirements.

You are also requested to provide a written statement in accordance iith the instructions specified for the enclosed Notice of Nonconformance.

The responses requested by this letter for the enclosed Notice of Violation and Notice of Nonconformance are not subject to the clearance procedures of the Office of Management and Budget as required by the Paperwork Reduction Act of 1980, Public Law No.96-511.

In accordance with 10 CFR Part 2.790 of the NRC's " Rules of Practice," a copy of this letter and its enclosures 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,

,[

4 Leif J. Norrholm, Chief Vendor Inspection Branch Division of Reactor Inspection and Safeguards Office of Nuclear Reactor Regulation

Enclosures:

1, Notice of Violation 2.

Notice of Nonconformance 3.

Inspection Report 99901246/92-01 106

ENCLOSURE 1 NOTICE OF VIOLATION Yuasa Exide Incorporated Docket No. 99901246 Reading, Pennsylvania ~19601 Report No. 92-01 During the Nuclear Regulatory Commission (NRC) inspection conducted from April 6 through 8,1992, at Yuasa Exide Incorporated (Exide), Reading, Pennsylvania, violations of NRC requirements were identified.

In accordance with the " General Statement of Policy and Procedure for NRC Er arrement

' Actions," 10 CFR Part 2, Appendix C (1992), the violations are listed below:

.A.

Section 21.21 (a) of 10 CFR Part 21, " Notification of failure to comply or existence of a defect," requires each individual, corporation, 1

partnership or other entity subject to this regulation, u 7 aluate deviations or inform the licensee or purchaser of the dev c. on, and to inform an appropriate director or responsible officer of licensee or purchaser if a basic component supplied by such facility contains a

-defect that could create a substantial safety hazard.

Section 21.21 (b) requires the director or responsible officer of a facility subject to these regulations to notify the commission when he obtains information reasonably indicating a defect sffecting the operation of a licensed facility.

1.

Contrary to the above, in 1982, Exide failed to notify the NRC of the cracking of terminal post radial gland seals on Exide Type GW batteries-installed at the Southern New England Telephone Company, even though identical seals existed on Class 1E Exide Type GC batteries installed at Fort Calhoun.

In 1991 these batteries failed as a result of the seal cracking (92-01-01).

This is a Severity Level H violation (Supplement VII).

2.

Contrary to the above, in 1979, Exide failed to notify the NRC or licensees thatLfailures of Type GC and GN Exide batteries during the-1919 qualification testing had resulted in a qualified life of 4 years for these batteries.

Instead, Fort Calhoun and other licensees were provided certificates of conformance indicating a-qualified life-of 20 years-for these batteries (92-01-02).

This is a: Severity Level H violation (Supplement VII).

Pursuant to the provisions of 10 CFR Part 2.201, you are required to submit a written statement or explanation-to the U.S. Nuclear Regulatory Commission,

' ATTN: Document Control Desk, Washington, D.C. ?0555 with a copy to the Chief, Vendor Inspection Branch, Division of Reactor Inspection and Safeguards, Office of Nuclear Reactor Regulation, within 30 days of your receipt of this Notice of Violation.

This reply should be clearly marked as a " Reply to a Notice-of. Violation" and should include the following for each of the violations:

107-

1.

The reasons-for the violation, or, if contested, the basis for disputing the violation, 2._

The corrective steps that have, or will be taken, and the results

.cchieved, 3, -

The corrective steps that have,-or will be taken, to avoid further violation, and 4.

.The date when full compliance will be achieved.

Wherefgood cause.is shown, the staff will consider extending the response time.

Dated at Rockville, Maryland

. this 17 day of Ivat -1992.

. 108

~-

~

T ENCLOSURE 2 NOTICE OF NONCONFORMANCE Yuasa Exide Incorporated Docket No.:

9990124E Reading, Pennsylvania 19601 Report No.:

92-01 During a Nuclear Regulatory Commission (NRC) inspection conducted from April 6 through 8, 1992, at Yuasa Exide Incorporated (Exide), Reading, Pennsylvania, l

the inspection. team determined that certain activities were not conducted in accordance with NRC requirements.

These requirements were contractually imposed by purchase orders from licensees to Exide.

The NRC has classified these-items as a Nonconformance to the requirements of-Title 10 of the Code of Federal Requiations, Part 50 (10 CFR Part 50), Appendix B.

A.

Criterion VIII, " Identification and Control of Components," of Appendix B to 10 CFR Part 50 states, in part, that " measures shall assure that tne identification of the item is maintained by part number, serial nu6er or other appropriate means, either on the item or on records traceable to the item, as required throughout fabrication, l

erectica, instai h tion, and use of the item.

These identification and control measures shall be designed to prevent the use of incorrect or defective materials, parts and components."

Contrary to the above, Exide implemented design and field modifications l

.to its Type GN and GC batteries but did not provide a unique number or other appropriate means to provide traceability to these modifications.

As a result, Exide could not provide traceability to the kind of modifications made by Exide to the batteries, and documentation to confirm if any batteries with defective post seals were currently installed in nuclear plants (92-01-03).

' Please provide a written statement _or explanation to the United States Nuclear

- Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555 cith-a copy to the Chief, Vendor Inspection Branch, Division of Reactor Inspection and Safeguards, Office of Nuclear Reactor Regulation, within-30 days of the date of the letter transmitting this Notice of Nonconformance.

This reply sbnuld be clearly marked as a " Reply-to a Notice of Nonconformance"

- and should it.,

le for each nonconformance:

(1) a description of steps that have been:or k 1 be taken to correct these_ items; (2) a description of steps that~kne been or will be taken to prevent recurrence; and (3) the dates your corrective actions and prevantive measures were or will be completed.

. Dated at-Rockville, Maryland

- this p day of 6 4 1992. 108a

ENCLOSURE 3 ORGANIZATION:

Yuasa Exide Incorporated (Exide)

Reading, Pennsylvania REPORT NO.:

99901246/92-01 CORRESPONDENCE Mr. Mike Ehlerman, President ADDRESS:

Yuasa Exide Incorporated 645 Penn Street P.O. Box 14145 Reading, Pennsylvania 19601 ORGANIZATIONAL Mr. Rich Berjer, Director for Qua'ity CONTACT:

(215) 371-0400 NUCLEAR INDUSTRY Class IE Station Batt'eies ACTIVITY:

INSPECTION April 6 through 8, 1992 CONDUCTED:

SIGNED:

d,vfMM 27 ///041[

6/7[#/2-andolph' N. Moist, Team Leader Date Reactive Inspection Section No. 2 Vendor Inspection Branch (VIB)

OTHER INSPECTOR:

Frederick H. Burrows, NRR Electrical Systems Br;nch 5/

~

APPROVED:

M h

C-

'Anil S. Gautam, Acting Chief Dat~e Reactive Inspection Section No. 2 Vendor Inspection Branch INSPECTION BASES:

10 CFR Part 21 and 10 CFR Part 50, Appendix B INSPECTION SCOPE:

Assess the adequacy of (1) Exide's evaluation and reporting of Type GC and GN battery failures during qualification testing and during service' conditions, and (2) the implementation of Exide's battery qualification program.

PLANT SITE APPLICABILITY:

Numerous 109

. -. - ~ ~. _ -

-1.0 INSPECTION

SUMMARY

11.1' Violations 1.1.l~ Contrary to Section 21.21 (a), " Notification of failure to comply or existence-of a de.ect," of 10 CFR Part 21, in 1982, Exide failed to notify the NRC of the cracking of terminal post radial gland seals on Exide Type GW batteries _ installed at the Southern New England Telephone Company, even though identical seals. existed on Class lE Exide Type GC batteries purchased by Omaha Public Power District (0 PPD) for the Fort Calhoun Nuclear Station.

In 1991 these batteries. failed as a result of the seal cracking (Violation 92-01-01, see-section 3.2 of this repart).

Contrary to Section 21.21 (b), " Notification of failure to comply or existence of a-defect and.its evaluation," of 10 CFR Part 21, in 1979, Exide failed to notify the NRC or licensees, including OPPD, that failures of Type GC and r,N l

Exide batteries had occurred during the 1979 qualification testing conducted by Exide at the Wyle Laboratories (Wyle), and that these failures had resulted in a qualified life of only 4 years for these batteries.

Exide provided OPPD and other licensees certificates of conformance (CoCs) to specifications invoked by. licensee purchase orders, even though these specifications required

-a qualified life of 20 years (Violation 92-01-02, see section 3.3 of this report).

l

-l.2 Nonconformance 1.2.1 Contrary to Criterion VIII, " Identification and Control of Materials, Parts and Components," of Appendix B to 10 CFR Part 50, Exide implemented design and field modifications to its Type GN and GC batteries out did not provide ~a unique number or other appropriate means to provide traceability to

~

these changes.

Exide could also not provide records -to match the type of modified battery with the' appropriate nuclear plant. As a result, Exide could not demonstrate-traceability to the design changes or field modifications made by Exide to the Type GN and GC batteries, and could not confirm through documentation if any batteries with defective post seals were currently installed in nuclear plants (Nonconformance 92-01-03, see section 3.4 of this report).

2.0 STATUS OF PREVIOUS INSPECTION FINDINGS No previous inspection findings exist.

l 3.0: INSPECTION FINDINGSLAND 0THER COMMENTS

(

3.1 -Entrance and Exit-Meetinas

-During the' entrance meeting on April 6, 1992, the NRC inspectors discussed t" scope _of: inspection,. outlined areas of concern and established interfaces with Exide management and staff. At the conclusion of the inspection on April 8, 1992,-and again by telephone on May 5, 1992, the inspectors summarized their

_z_

110

~

findings and concerns, and Exide management and staff acknowledged this information, 3.2 Eailure of Battery Post Seal On October 7, 1991, OPPD reported cracking on the face of Class lE Exide Type GC-23 battery jars installed at the fort Calhoun nuclear station (Fort Calhoun), causing leakage of electrolyte to the floor.

The radial gland terminal post seal cracked in two cells in the plant, leading to cracking of the positive terminal seal nut, and the subsequent cracking of the jar face.

These failures occurrcd despite Exide performing a field modification to these batteries in 1983 to preclude such cracking.

Exide stated that the root cause of the failure was an inadequate design of the cell terminal post seal, out that this defect had not been detected during qualification testing or analysis.

As a result, f ailures of Type GC batteries occurred at Fort Calhoun and resulted in a substantial safety hazard.

The team reviewed the history of Type GC and GN batteries to determine if there were any past failures or reporting of defects by the manufacturer The team noted that in 1982, the Southern New England Telephone Company informed Exide of the cracking of terminal post seals on a Type GW battery purchased from Exide.

Exide described Type GW and GC to be commercial designations for the Type GN (nuclear grade) batteries and that Type GC, GW, and GN batteries had identical terminal post radial gland seals.

Exide stated that, in response to the cracking of Type GW seals, they implemented corrective action by performing field modifications to post seals of Type GN batteries at nuclear plants, including Fort Calhoun.

Exide also stated that, on April 19, 1983, pursuant to 10 CFR Part 21, Exide notified the NRC of the cracking of the Type GN battery seals.

However, the team noted that Exide had made no notification of the relevance of this cracking to the Type GC batteries, even though the Type GC batteries installed at Fort Calhoun had terminal post seals identical to the seals on the Type GW and GN batteries.

The team concluded that Exide had failed to notify the NRC of a defect concerning these batteries despite having been notified of this defect by a customer.

This is considered a Violation of 10 CFR Part 21 (92-01-02).

3.3 Failures of Batteries Durina 0ualification Testina lhe Type GC batteries were originally purchased by OPPD as Class lE from the Exide Industrial Battery Division via Purchase Order (PO) 44643 dated September 20, 1979 (Yuasa and Exide Industrial Battery Division form.d a joint venture in July 1991, to form Yuasa Exide Incorporated).

Purchase order 44643 imposed the requirements of 10 CFR Part 21,10 CFR Part 50, Appendix B, and OPPD Specification MR NO. FC-79-03, " Specification for Replacement Storage Batteries for Fort Calhoun Unit #1," Revision 1, dated August 24, 1979, on Exide. The batteries were to be qualified to Class lE requirements for 20 years of service in accordance with the licensee procurement spe:ifications.

In 1979, Exide conducted qualification testing of Type GC and GN batteries at the Wyle Laboratories, as documented in Wyle qualification report number 44478-1, dated April 16, 1980.

The report identi?ied the test specimens to be 111

l Type GN and Type GC batteries.

Failures occurred during the thermal aging of the specimens, and were documented by Wyle as test anomalies.

One anomaly identified the discovery of cracks on the heat seal covers during the fourth week of the test.

Two additional anomalies identified a 50-percent increase in current due to internal shorting of cells.

The shorted cells were removed from the test.

Exide elected to terminate the thermal aging of the batteries af ter 45 days of testing because of the failures and to establish a qualified life of 4 years for the Type GC and GN battery specipens.

Exide stated that it decided to establish a shorter qualified life because several licensees needed immediate delivery of the Type GN batteries and could not wait for the 2 year qualification test program required to establish a qualified life of 20 years.

However, there was no evidence that Exide had notified licensees of the 4 year qualified life of the batteries.

For example, Exide provided a CoC to OPPD for the Type GC battery at Fort Calhoun. The CoC stated that the batteries met the requirements of OPPD specification MR-FC-79-03. This specification required the qualified life of the Type GC batteries to be 20 years.

Exide stated that they sold OPPD a

{

commercial grade Type GC battery.

However, based on the OPPD purchase order and documentation provided to the team, the Type GC batteries purchased by OPPD were Class lE and were supposed to be qualified for 20 years of service.

The team concluded that Exide failed to notify the NRC and licensees, including OPPD, of a defect or deviation that reduced the qualified life of the Type GC and GN batteries from 20 to 4 years, and that OPPD and other licensees had been misinformed through Cots that these batteries had a qualified life of 20 years.

This is considered a Violation of 10 CFR Part 21 (92-01-01),

3.4 Lack of Battery Identification _and Records The team r,ated that Exide had not assigned any unique identifiers to identify

~

the specific design differences and modifications made by Exide to the Type GC and GN batteries.

During 1979 through 1983, Exide attempted to qualify the Type GC and GN batteries to Class lE standards.

As noted in Section 3.3 of this report, failures occurred during the qualification testing. Subsequent to these failures, Exide implemented a new manufacturing process for all GC and GN batteries. However, no unique identifier was provided for traceability to the new modificd batteries.

In 1982, a design defect was discovered involving the cracking of the battery post seals, as noted in Section 3.2 of this report.

To correct this problem.

Exide developed three different field modifications (Level I, II, and III) to repair batteries installed in nuclear plants.

Exide stated that they performed the level of modification requested by the licensees for batteries supplied to nuclear plants. However, no unique number was provided for traceability to the modified batteries for the work performed by Exide.

In.

112

l some cases, the battery would have to be disassembled to identify the type of modification.

In March 1983, Exide redesigned the post seal to eliminate the defect (and need for field modification), and began producing a new design of Type GC and GN batteries.

However, no unique number was provided for traceability to the new batteries.

The team also noted a lack of traceability to the type of batteries tested in the qualification documentation (section 3.3 of this report).

For example, when the team reviewed Wyle qualification report number 44478-1, dated April 16, 1980, they noted that battery test specimens were referred to as Type GC and GN interchangeably.

In one section of the report, the test specimens were identified as Type GN-13 (Part Number 84913) and Type GN-23 (Part Number 84918) batteries; however, the report also identified the same specimens as Type GC-13 and Type GC-23.

The team could not determire if the qualification reports supported the Type GC (connercial grade) or the Type GN (nuclear grade) battery.

Based on their review of Exide qualification reports, licensee procurement specifications, and related documentation, the team concluded that no unique identifiers had been assigned to reflect specific design differences and modifications 1 ie to these batteries.

Exide also could not provide records to match the type of modified or redesigned battery with the respective nuclear plant or licensee who purchased the battery.

As a result, Exide could not demonstrate to the team if any Exide Type GC or GN batteries with defect've post seals currently existed in nuclear plants.

Exide's failure to provide identification and control of materials, parts, and components is considered a Nonconformance to 10 CFR 50 Appendix B (92-01-03).

4.0 PERSONNEL CONTACTED Yuasa Exide incorocrated

+*

I. Baeringer, Vice President of Engineering

+*

R. Bender, Director for Quality

+*

C. Carr, Quality Assurance Manager

+*

M. Patel, Manager of Stationary Products

+*

C. Reichart, Contract Administrator

+

W. Mettin, Senior Stationary Product Engineer Nuclear Reaulatory Commission

/

Gautam, Acting Section Chief, VIB, NRR, NRC

• Attended the exit meeting

+ Attended the entrance meeting 113

Selected Bulletins and Information Notices Concerning Adequacy of Vendor Audits and Quality:.of Vendor Products ISSUED.

ILT1*g]

. 1.

Information Notice 92-27 Thermally Induced Accelerated Aging and Failure of ITE/Gould A.C.

Relays Used in Safety-Related Applications F

- 2.

Information Notice 92-31 Electrical Connection Problem in Johnson Yokogawa Corporation'YS-80 Programmable Indicating Controllers 3 '.

Information Notice 92-43

. Defective Molded Phenolic Armature Carriers Found on i

Elmwood Contactors 4.

Information Notice 92-44 Problems With Westinghouse DS-206 and DSL-206 Type Circuit. Breakers 5.

Information. Notice 92-45 Incorrect Relay Used in Emergency Diesel Generator Output. Breaker Control Circuitry l-i d

2 h

114

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

?

- CORRESPONDENCE RELATED TO VENDOR ISSUES l

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C UNITED STATES 1

jij NUCLEAR REGULATORY COMMISSION g

o

WASHINGTON, D.C. 206E6 J-JUN 2 61992 Docket No.

-99900100 Mr. P. G. McQuillan, Manager

- Nuclear /Special Projects Limitorque' Corporation 5114 Wooda11~ Road P. O.

Box;11318 Lynchburg,-Virginia 24506-1318

SUBJECT:

APPLICABILITY OF 10 CFR PART 21

- I am responding to_your: letter, dated; June 15, 1992, which

- requested.NRC's position relative-to the applicability of 10 CFR

- Part 21 to equipment that will perform a safety related function in a nuclear - facility not based'in the United States.

With respect to' equipment installed only in a foreign facility, the requirements of--10 CFR Part 21 would have no applicability.

. Only:the parallel _ requirements of the host country, if any, woul<*.

be applicable..

However,-in those cases in which the same or similar equipment is i

-installed both in foreign facilities and'Un ted States.

ifacilities, discovery of a-deviation in the foreign-installation would require-the evaluation and reporting requirements-of 10 CFR

- Part 21 to;be-performed.- Once it is determin'ed that a defect or' failure to.-comply actually exists in particular equipment, even if installed;atia foreign facility, it-isistill incumbent-.on the

- vendor-to-provide the appropriate 10 CFR.Part 21 notifications to

-domestic licensees and theiNuclear_' Regulatory Commission, if that same defect could exist.in dor *stic nuclear plants.

If you have_further questions,-please contact me.

Sincerely,

,4 1/

6 j.

Leif

._N rrholm,sChief

' Vendor Inspection Branch

= Division ~of Reactor Inspection.

and Safeguards Office of Nuclear Reactor Regulation e

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Vol. 16, No. 2

i. ina AND 50Dilid Licensee Co'1 tractor and Vendor inspection Status Report 3

D AT E R( POHT PUBt ISHE D Quarterly Report l

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" Aa April-June 1992 July 1992 4 FIN OR GR ANT NUMetH f4 AU T HQR(S) 6 THE Of HLPORT Quarterly

1. VL H lOD COV t H L D isnenus,ee Demb April-June 1992 8i FORMS ANIZ AT lON - NAMt ANO AUDRt SS for NRC. provue Onunen. Otr w or noyean. U $ Naor nevererer 'onaseven, ense moonee e.stuess or runteervor, orpessse a

Division of Reactor Inspection and Safeguards Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555

9. SPONSORING ORGANIZ ATION - N AME AND ADDHi SS Its Nac. e,s=

seaw as et=>.,~. sa coas,ec tur, piernse 4ee o<veinsa. Or<.c, e, s,,=,n. v s eucaw n,uee<ory com,== wen.

cant answer enresaJ Same as above

10. SUPPLEMENTARY NOTES

11. CAST R AC T (Aw. ara w arss>

This periodical covers the results of inspections performed by the NRC's Vendor inspection Branch that have been distributed to the inspected l

organizations during the period from April through June 1992, I

12. CC L Y WORDS/DE $CH!P TOR S (ta4 =o<ds or paceses ran war esme renereners m #acenas rae vroper.!

12 ava L Ato.a t y r 'itut Nt -

Unlim" vendor inspection

,. 34,Jribiul o~

lihotlegel Uncl as si fiec'

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Unclass'#ied 15 NUMOLH Of PAGts Ili PHILt NHC FONu 33 O a191

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- THIB: DOCUMENT WAS PRINTED USING RECYCLED Pl.PER w.. :.. -

snctat rouamctass nArt UNITED STATES rostace Awo rrts rao NUCLEAR REGULATORY COMMISSION usunc WASHINGTON, D.C. 20555-0001 NRMtT NO. G-07 OFFICIAL BUSINESS PENALTY FOR PRIVATE USE. $300 l

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