ML20197B043

Licensee Contractor and Vendor Inspection Status Report. Quarterly Report,April-June 1997.(White Book)

ML20197B043
Person / Time
Issue date:	11/30/1997
From:	Office of Nuclear Reactor Regulation
To:
References
NUREG-0040, NUREG-0040-V21-N02, NUREG-40, NUREG-40-V21-N2, NUDOCS 9712230277
Download: ML20197B043 (98)

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NUREG-0040 Vol. 21, No. 2 Licensee Contractor and Vendor Inspection Status Report  :

Quarterly Report April - June 1997 U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation

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AVAILABILITY NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in NRC publications will be available, from one of the following sources:

1. The NRC Public Document Room 2120 L Street, NW., Lower Level Washington, DC 20555-0001

2. The Superintendent of Documents. U.S. Government Printing Office, P. O. Box 37082 Washington, DC 20402-9328

3. The National Technical ir. formation Service, Springfield, VA 221CI-0002 Although the listing that follows represents the malnrity of documents cited in NRC publica.

tions, it is not intended to be exhaustive.

Referenced documents available for inspection and copying for a fee from the NRC Public Document Room include NRC correspondence and internal NRC memorandal NRC bulletins, circulars, information notices, inspection and investigation notices licensee event reports; vendor reporta and ccrrespondence: Commis0bn papers; and applicant and licensee docu-ments and correspondence.

The following documents in the NUREO series are available for purchase from the Government Printing Office: formal NRC staff and contractor reports NRC sponsored conference pro-ceedings, international agreement reports, grantee reports, and NRC booklets and bro-chures Also available are regulatory guides, NRC regulations in the Code of Federal Aeguis-tlons, and Nuclear Regulatory Commission issuances.

Documents available from the National Technical Information Service include NUREG series reports and technical reports prepared by other Federal agencies and reports prepared by the Atomic Energy Commission, forerunner agency to the Nuclear Regulatory Commission.

Documents available from public and special technical libraries include all open literature items, such as books, journal articles, and transactions. Federal Register notices, Federal and State legislation, and congressional reports can usually be obtained from these libraries.

Documents such as theses, dissertations, foreign reports and trans,iations, and non-NRC con-ference proceedings are availab'e for purchase from the organization sponsoring the publica-tion cited.

Single copies of NRC draft reports are available free, to the extent of supply, upon written request to the Office of Administration. Distribution and Mall Services Section, U.S. Nuclear Regulatory Commission, Washington, DC 20555-000t.

Copies of industry codes and standards used in a substantive manner in the NRC regulatory process are maintained at the NRC Library, Two White Flint North,11545 Rockville Pike, Rock-ville, MD 20852-2738, for use by the public. Codes and standards are usually copyrighted and may be purchased from the originating organization or, if they are American National Standards, from the American National Standards institute,1430 Broadway, New York, NY 10010-3308.

A year's subscription of this report consists of four ovarterly issues.

NUREG-0040 Vol. 21, No. 2 Licensee Contractor and Wndor Inspection Status Report .

Quarterly Report April- Jiine 1997 Manmeript Canpleted: October 1997 Date -!'ublidicd: November 1997 Division or Reactor Controls and lluman Factors Omce of Nuclear Reactor Regulation U.S. Nuclear Regulatory Conunission Washington, DC 20555 0001 qv' *'ev f

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- NUREG-0040, Vol. 21.No. 2 has been reproduced from the best available copy.

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ABSTRACT l

i This periodical covers the results of inspections performed between April 1997 and June 1997 -

by the NRC's Special Inspection Branch, Vendor Inspection Section, that have been distributed to the inspected organizations, i i

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f CONTENTS PAGE Abstr a et . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ........................................................................iil I n t r o d u cti o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1 i n spe ction R e port s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

ABB Service Incorporated (99901281/97-01) .... .................. 2 Regional Service Center Fenton, MO C&D Charter Power Systems, Inc. (99901305/97 01) ........ ......... ... 7 Conshohocken, PA Dubose National Energy Services, Inc. (99900861/97-01) ..... ... ...... ..... 21 Clinton, NC General Electric Nuclear Energy (99900003/97 01) ...... ........ .... . 42 Wilmington, NC SynTech Products Corporation (99901313/97-01) ........ .......... . 58 Toledo, Ohio Nuclear Advanced Technology Division (99900404/97-01) ...... ... ..... .. .. 74 Westin9 house Electric Corporation Select Generic Correspondence on the Adequacy of Vendor ..... .. .... .......... ... ..... 88 Audits and the Quality of Vendor Products V

INTRODUCTION A fundamental premise of the U. S. Nuclear Regulatory Commission (NRC) licensing and inspection program is that licensees are responsible for the proper construction and safe and efficient operation of their nuclear power plants. The Federal government and nuclear industry have established a system for thJ inspection of commercial nuclear facilities to provide for multiple levels of inspection and verification. Each licensee, contractor, and vendor participates in a quality verification process in compliance with requirements prescribed by the NRC's rules and r?gulations (Title 10 of the Code of Federa/ Regulaflons). The N9C does inspections to oversee the commercial nuclear industry 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 quality verification programs.

The licensee is responsitile for developing and maintaining a detailed quality assurance (QA) plan with implementing procedures pursuant to 10 CFR Part 50. Through a system of planned and periodic audits and inspections, the licensee is responsible for ensuring that suppliers, contractors and vendors also have suitable and appropriate quality programs that meet L'tC 1

requirements, guides, codes, and standards.

The NRC reviews and inspeats nuclear steam system suppliers (NSSSs), architect engineering (AE) firms, suppliers of products and services, independent testirq laboratories performing equipment qualification tests, and holders of NRC construction permits and operating licenses in vendor related areas. These inspections are done to ensure that the root causes of reported vendor related problems are determined and appropriate corrective actions are developed. The inspections also review vendors to verify conformance with applicable NRC and industry quality requirements, to verify oversight of their vendors, and coordination between licensees and vendors.

The NRC does inspections to verify the quality and suitability of vendor products, licensee-vendor interface, environmental quahfication of equipment, and review of equipment problems found during operation and their corrective action. When nonconformances with NRC requirements and regulations are found, the inspected organization is required to take appropriate corrective action and to institute preventive measures to preclude recurrence.

When generic implications are found, NRC ensures that affected licensees are informed through vendor reporting or by NRC generic corresponder.ce such as information notices and bulletins.

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This quarterfy report contains copies of all vendor inspection reports issued during the calendar quarter for which it is published. Each vendor inspection report lists the nuclear facilities inspected. This information will also alert affected regional offices to any significant prob!em areas that may require special attention. This report also lists selected bulletins, generic letters, and information notices, and include copies of other pertinent correspondence involving vendor issues.

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• ) NUCLEAR REGULATO9Y COMMIS810N WASHINGTON, D.C. 30406 4 001

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% ..... / June 4, 1997 Mr. D. P. Polizzi Manager, Regional Service <

ABB Service Inc.

Regional Service Center 400 Blitmore, Suite 520 fenton, MO 63026

SUBJECT:

NRC INSPECTION REPORT NO. 99901281/97-01

Dear Mr. Po11zzt:

This letter refers to the inspection conducted by Mr. K. R. Naidu and Mr. V.

L. Beaston of this office on May 15, 1997. The inspection included a review of activities conducted at your facility in fenton, Missouri. The enclosed report presents the results of that inspection.

The insp:ctors evaluated the program that JisB Service Inc. (ABB) established to implement the provisions of Part 21 of fitle 10 of the Code of Federal Reaulations (10 CFR Part 21) for reporting defects and noncompliances, and Appendix B to 10 CFR Part 50. The inspectors reviewed the docu.nentation on selected metal-clad K-line circuit breakers that Illinois Power Company's Clinton Power Station sent to your Fenton shop for refurbishing. Within these areas, the inspection consisted of an examination of procedures and representative records, interviews with personnel, and oMervations by the inspectors. The ins)ectors identified one unresolved item relative to the availability of brea(er specific drawings to verify the correct contact configuration and propose to obtain additional information on this item during a future inspection at your Houston Service Center.

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

Si cerely

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\ g Robert M. Ga o, r1 Special Inspection Branch Division of Inspection and Support Programs Docket No.: 99901321

Enclosure:

Inspection Report 99901321/97-01 ,

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INSPECTION REPORT-U.S. NUCLEAR REGULATORY COMMISSION c OFFICE OF NUCLEAR REACTOR REGULATION '

DIVISION OF INSPECTION AND SUPPORT PROGRAMS .

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ORGANIZATION: ABB Service Inc. l Fenton, Missouri r REPORT NO.: 99901321/97-01 ORGANIZATIONAL Mr. D. P. Polizzi, Manager, Regional Service CONTACT: 314 343 0232 -

ABB Service Inc. services switchgekr

' NUCLEAR INDUSTRY manufactured by I.T.E.

ACTIVITY:  :

INSPECTION CONDUCTED:

May 15, 1997  :

Kamalakar R. Naidu, NRR INSPECTORS:

Virgil L. Beaston, NRR l

APPROVED BY:

Gregory C. Cwalina, Chief Vendor inspection Section Special Inspection Branch Division of Inspection and Support Programs Office of Nuclear Reactor Regulation

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i 1 teFECTION $UMMARY AB8 Service Inc., located in Fenton, Missouri, performs switchgear overhauling '

services for commercial and nuclear customers. Recently it performed services on 480 vcit metal-clad circuit breakers installed at Illinois Power Company's (IPC) Clinton Power Station (Clinton). The circuit breakers were originally  :

designed and manufactured by I.T,E imperial Company (ITE). The ownership subsequently changed and the company became known as I.T.E.- Gould, Gould- f Brown Poveri, Brown Boveri Electric and finally ASEA Brown Boveri (ABB). ABB manufactures low-voltage metal-clad circuit breakers, and medium-voltage circuit breakers at Florence, South Carolina. ABB at Sanford, Florida, i assembles complete switchgear installations. ABB has established service centers at several locations in the U.S. The service center at Fenton currently services the circuit breakers installed in Clinton, Recently, ABB i decided to service its nuclear customers from four locations, Columbia, -

Maryland; Cleveland, Ohio; Charlotte, North Carolina; and Houston, Texas. In the future, breakers which need service will be shipped to ABB Services, Houstone Texas. '

The inspectors reviewed the documentation on selected metal-clad K-iine circuit breakers that IPC's Clinton sent for refurbishment.

The inspection bases were:

• Appendix B, " Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," to Part 50 of Title 10 of the Code of Federal Reaulations (10 CFR Part 50) 10 CFR Part 21. " Reporting of Defects and Noncompliance" During this inspection, the NRC found no instance where ABB failed to meet NRC requirements. However, the inspectors identified one unresolved item regarding the availability of breaker specific drawings (Section 3.1).

2 STATUS OF PREVIOUS INSPECTION FINDINGS This is the first inspection at ABB Service Inc., Fenton, Missouri.

3 INSPECTION DETAILS 3.1 Breaker Refurbishment

a. SLDDR On May 15, 1997, the inspectors reviewed the records documenting the refurbishment performed on Clinton breakers and reviewed the refurbishment activities with the personnel who performed the work at the Fenton shop.

The inspectors also reviewed the purchase order (PO) that IPC issued to ABB for the refurbishing services including selected revisions to the PG to determine if the work preformed on selected Clinton breakers met the quality requiren,ents.

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b. Observatic,ns and Fin (ing1 IPC issued PO 554794 to ABB Sanford, Florida, to rework / refurbish a circuit breaker with serial number Sl?52A-9-02616 with 55-13 type solid state trip device in compliance with ABB Procedure MS 3.1,1.9-20, " Maintenance and Surveillance, Low Voltage Switchgear Equipment." In subsequent revisions to the P0, Clinton added additional breakers to be refurbished.

IPC attached to the P0 a copy of its " Supplier Quality Assurance Program Requirements," QAP-407-02F02, Revision 2, which invoked 10 CFR Part 50 Appendix B, and 10 CFR Part 21. IPC required ABB to obtain its approval for all nonconformance reports initiated by ABB during the refurbishment where ABB would recommend "use as is" or " repair." IPC required ABB to issue a Certificate of Rework or Repair certifying that the item was reworked using original equipment parts by a qualified technician, and that the item will function in accordance with the original design specification.

Revision 02 to the P0, dated June 23, 1995, required the replacement of all EQ sensitive (organic) items with less than 40-year qualified life. As a minimum, Agastat time delay relays, Gould auxiliary relays (J13PA4312), and control and power Shawmut type fuses were to be replaced.

Revision 09 to the P0, dated November 29, 1995, required ABB to test the breakers for "as-found" conriitions. ABB was also to rework / refurbish and lubricate the breakers, and test the trip devices to verify the time / current curve.

Revision 12 to the P0, dated March 12, 1996, added the refurbishment of breaker Serial Number (S/N) 51752E-9-02616. ABB was to (1) reconfigure the breaker and update documentation as appropriate, (2) replace the electrically operated mechanism with a manually operated mechanism, (3) replace the lower molding and CT assembly with 600-A variety, (4) change the solid state trip device to 600-A (SS 14), (5) add an alarm switch, and other minor accessories as required to conform with Drawing 708 665, Revision 0.

The records on work performed on breakers with S/Ns 51752AB-109-201054, 517528-9-11456, 51752E-9-02616, 51752B-9-01456, 51752F-ll-02616, and 51752F 01616, indicate that ABB personnel had operated the Clinton breakers, disconnected the breaker mechanisms, and inspected the main roller and carrier and other major pins and moving components in the breakers. Records indicated that the operating mechanisms were dirty and had to be disassembled, cleaned, lubricated and reassembled. On receipt of the breakers from ABB, Clinton personnel performed receipt inspections and documented adverse findings (discussed below) in Condition Reports (CRs). The inspectors discussed the adverse findings in these CRs with the ABB personnel and reviewed the refurbishment records.

During receipt inspections, Clinton identified a deficiency related to incorrect wiring of leads 11 and 12 on the alarm switch in four breakers.

Clinton corrected this error by switching the leads, and informed ABB during a meeting at the Clinton plant. The inspectors interviewed the ABB personnel 2

who worked 09 the breakers, and determined that the technician who wired the switch had made an error. ABB personnel-stated that they could not verify the correct wiring of the alarm switch because Clinton did not send them the drawings for the individual breakers. At the time of the inspection, ABB personnel stated that this condition had not been documented in a noncon-formance report because the technician who made the error had been away working on a job in Maine, and returned to the shop only the previous week.

ABB personnel sutsequently documented this in a nonconformance report.

Clinton also identified a problem related to the configuration of the 8-point auxiliary switch which is mounted on the "C" phase side below the Power Shield unit.- The switch is connected through a right-hand link assembly to the jack shaft of the breaker. The contacts in the auxiliary switch change state when the breaker operates. Clinton personnel had documented that, when the breakers were receipt inspected, the contact configuration did not meet the breaker drawing requirement. ABB personnel stated that different breakers have different contact configurations. A 90-degree rotation of the operating lever will change the contact configuration. ABB personnel informed the inspectors that Clinton did not provide them with the drawings for the individual breakers. Therfore, ABB could not verify the correct contact configuration. Further, ABB did not identify the need for the drawings to perform safety-related refurbishments.

c. Conclusion Initially, Clinton did not provide individual breaker drawings to ABB, and ABB did not request them. The inspectors considered this to be an indication of inadequate Licensee - Vendor interface. Adequate technical information exchange (e.g., circuit breaker drawings) between personnel at Clinton and ABB could have reduced the number of problems that were identified in the condition reports. Since the service center at Fenton will no longer refurbish safety-related breakers, this concern is being identified as an unresolved item and will be followed up at a future inspection of the ABB Service Center at Houston, where ABB proposes to refurbish the Clinton breakers in future. (Unresolved item 99901321/97-01-01)

4. PERSONS CONTACTED ABB Service Company D. P. Polizzi, Manager, R(,ional Service R. Clostermann, Service Engineer J. O. Webb, Director of Quality ITEMS OPENED, CLOSED, AND DISCUSSED Onened 99901321/97-01-01 URI unavailability of breaker drawings 3

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s**%4 y *~ J- t UNITED STATES

_g . f} NUCLEAR REGULATCRY COMMISSION

'*- I WASHINGTON, o.C. 30666-0001 f

• ..,* May 16, 1997 Dr. Leslie S. Holden-Vice President Technology C&D Charter Power Systems, Inc.

Washington & Cherry-Streets Conshohocken, PA 19428

SUBJECT:

-NRC INSPECTION REPORT 99901305/97-01

Dear Dr. Holden:

OnJanuary9,-1997,thpU.S.NuclearRegulatoryCommission(NRC)completedan inspection at the C&D* Charter Power Systems, Inc. (C&D) facilities at Leola, Pennsylvania, regarding cylindrical lead-acid battery cells, "round cells," that C&D manufactures and supplies to nuclear power plant facilities as a contractual manufacturing representative for Lucent Technologies, Incorporated, (Lucent). The round cells are designed, qualified and contractually controlled by Lucent, and are used in Class IE electrical safety-related applications at five nuclear power plant facilities. The enclosed report presents the results of the inspection.

The NRC team determined that C&D has manufactured and controlled the nuclear power plant customer orders under its commercial manufacturing controls, in association with Lucent's specifications. C&D has established and implemented an International Organization for Standardization (150)-9001, " Quality Systems-Model for Quality Assurance in Design / Development, Production, Installation and Servicing" quality assurance program at its Leola facility through its corporate office in Blue Bell, Pennsylvania. Although the team assessed portions of your quality assurance program implementation, your activities were not assessed to Part 50 of Title 10 of the Code of Federal Reaulations (10 CFR Part 50), Appendix B quality assurance requirements; therefore, no response to this letter is required.

However, the team determined that Lucent contractually accepted 10 CFR Part 50, Appendix B quality requirements from the Duke Power Company but did not pass those requirements on to C&D. Therefore, the team reviewed certain portions of C&D's quality activities to Appendix B quality requirements to assess Lucent's compliance with regulatory requirements. The results of that review are discussed separately in NRC's in:.pection report of Lucent Technologies, Report 99901309/96-01.

1 C&D is a registered trademark of C&D Charter Power Systems, Inc.

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Dr. L. S. Holden In accordance with 10 CFR 2.790 of the NP.C " Rules of Practice," a copy of this letter and-enclosures will be placed in the NRC's Public Document Room (PDR).

-Should you have any questicas concerning the issues discussed in this letter, we will be pleased to discuss them with you.

Sincerely, hl/F 9(

Robert M.~Gallo, Chief Special Inspection Branch Division of Inspection and Support Programs Office of Nuclear Reactor Regulation Docket No.: 99901305

Enclosure:

Inspection Report 99901305/97-01 cc: Mr. T.J. Kinder , Director Quality Assurance C&D Charter Power Systems, Inc. <

Washington & Cherry Streets Conshohocken, PA -19428 Mr. Greg Stoermer, Plant Manager C&D Powercom 82 East Main Street Leola,-PA. 17540-1940 I

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Enclosure U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR RETCTOR REGULATION Report No.: 99901305/97-0)

Organization: C&D*' Charter Power Systems, ' Incorporated 1400 Union Meeting Road Blue Bell, Pennsylvania 19422-0858 F

~ Contact Address: Terry J. Kindan, Director of Quality Assurance Washington & Cherry Streets Conshohocken, PA 19428 (610) 825-2150 ext. 245 i

facility C&D PowerCom Inspected: 82 East Main Street Leola, Pennsylvania  ;

Nuclear Industry Manufacturers of Lucent Technologjes cylindrical Activity at lead-acid battery cells, Lineager 2000 Round Cell Facility: batteries (round cells), that are supplied to nuclear power plant facilities for Class IE electrical H safety-related applications.

Inspection Date: January 6-0, 1997 Inspector:: Joseph J. Petrosino, Team Leader  ;

Saba N..Saba, EELB:DE:NRR Stephen D. Alexander, VIS:PSIB: DISP Approved by: Gregory C. Cwalina, Chief Vendor Inspection Section Special Inspection Branch Division of Inspection and Support Programs Office of' Nuclear Reactor Regulation k

' C&D is a registered-trademark of C&D Charter Power Systems, Inc. 1

'2 . A division of C&D Charter Power Systems, Inc.

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. Registered-trademark of Lucent technologies, Incorporated (formerly

known as American Telephone.and Telegraph Company [AT&T)).

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1 INSPECTION

SUMMARY

During this inspection, the NRC inspectors reviewed the implementation of selected portions of C&D Charter Power Systems, Incorporated (C&D) quality assurance (QA) program, reviewed activities associated with the manufacture of round cells, reviewed pertinent design, engineering and manufacturing documents, reviewed the Lucent Technologies, incorporated (Lucent) manufacturer requirements regarding Lucent's control of C&D's Leola manufacturing activities, and compared th(. Lucent requirements with actual manufacturing procedures and practices. The team was aware that several nuclear plant round cell. batteries experienced premature loss of capacity.

The inspecticn bases were:

• Appendix B, " Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," to Part 50 of Title 10 of the Code of Federal Reaulations (10 CFR Pa-t 50)

• 10 CFR Part 21, " Reporting of Defects and Noncompliance"

• The American National Standard Institt.te/The American Society of Mechanical Engineers (ANSI /ASME) StandaH N45.2-1977, " Quality Assurance Program Requirements for Nuclear Facilities" (ANSI N45.2-1977), as endorsed by-U. S. Nuclear Regulatory Commission Regulatory Guide (RegGuide) 1.28, " Quality Assurance Program Requirements (Design and Construction)," Revision 2, February 1979 (RegGuide 1.28)

Although C&D's quality program for its manufacture of round cells was found to be generally well established and implemented, the team found some weaknesses that are discussed in this report. Those weaknesses are also discussed in NRC Inspection Report 99901309/96-01, Lucent Technologies-Dallas, Texa:.

2 STATUS OF PREVIOUS INSPECTION FINDINGS This was the first NRC inspection of C&D's facilities at Leola, Pennsylvania.

3 INSPECTION FINDINGS AND OTHER COMMENTS 3.1 Ouality Assurance Proaram and Reauirements

a. ScoDe The inspection team (team) reviewed the establishment and implementation of selected portions of the-C&D quality program, which was documented in the C&D Quality System Program Manual (QSPM), Issue P, dated June 1995.

b. Observations (nd findinos The QSPM stated that its purpose was to establish the basic operating policies and procedures to be employed by C&D, and to meet applicable requirements of International Organization for Standardization (150)-9001, " Quality Systems-Model for Quality Assurance in 2

-_ Design / Development, Production, Installation and Servicing,"

ANSI [N)45.2-1977 and other imposed specifications. The team noted that C&D's Leola-facility had been certified to 150-9001.

The team was apprised by C&D that it assures that its components conform to the specified requirements through implementation of C&D's quality system program that is delineated in the QSPM. The QSPM's foreword explains that C&D's quality system program. is implemented through the-QSPM, departmental . standard operating procedures, work instructions, drawings, bills of materials, material specifications, process specifications, and quality and test procedures. The team reviewed applicable 1_icensee and Lucent procurement documents and did not find any instances where AT&T/ Lucent had imposed any unique nuclear quality assurance requirements upon C&D.

The team noted that the C&D facilities that are under C&D's quality system' program are specified in the QSPM, specifically: (1) Corporate Headquarters-Blue Bell, PA-Control and maintenance of design and qualification aspects of all C&D safety-related components; (2) Attica, IN-Class IE station batteries; (3) Dunlap, TN-Class lE station battery chargers; (4) Conshohocken, PA-seismically qualified, Class lE station battery racks; (5) Leola, PA-Class -lE station batteries, round cell; (6) Conyers. GA-commercial products; (7) Huguenot, NY-commercial products; and (8) Ratelco, Incorporated, Seattle, WA-electronics, including battery charger assemblies used in safety-related applications of C&D battery chargers.

c. Conclusions The team determined that C&D-Leola manufactured the round cells according to its commercial QA program. No nuclear qualification, manufacturing or testing requirements were imposed onto C&D by Lucent for the manufacture of the round cells.

3.2 10 CFR Part'21 Proaram

a. Scope The inspectors evaluated the procedure adopted by C&D to implement the requirements of 10 CFR Part 21 by reviewing Standard Policy and Procedure Number A-14-4, " Reporting of Defects and Nonconformances in accordance with Federal Regulation 10 CFR 21 (US Nucle Regulatory Commission)," dated March 18, 1997.

b. Observations and findinas Although C&D is_not responsible for the requirements of 10 CFR Part 21 at its Leola facility, the team determined that C&D had implemented the provisions of Procedure A-14-4 into its Leola facility round cell 3

o>eration for nuclear power plant orders. Additionally, the team noted tiat documents required by Part 21 had been posted at various locations at the Leola facility.

During a previous inspection, in September 1996, at C&D's Attica, Indiana, facility, the team performed a review of Procedure A-14-4. The review determined that some weaknasses existed in the procedure and C&D committed to modify the procedure to address the NRC concerns within 90 days of the date of Inspection Report 99901304/96-01. (March 17, 1997).

Procedure A-14-4 had not yet been revised and provided to the C&D-leola facility, c Conclusions The team did not identify any concerns in this area.

3.3 Imolementation and Control Of Ouality Proaram

a. Scope During the inspection at the C&D-Ltola facility, the team observed manufacturing activities and reviewed associated documents to assess the adequacy of the manufacture of the rnund cells,

b. Observations and Findinas

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Process Controls The team observed numerous manufacturing process control activities at C&D's facility and reviewed associated documents.

The team noted that many of the C&D manufacturing work stations had key sections of the applicable process control work instructions posted for use of the area personnel. The team also noted that a Lucent required resident inspector is stationed at the C&D facility. The resident inspector is typically on the premises several days a week. The resident inspector witnesses testing activities of round cell battery strings, and performs in-process manufacturing verifications and random inspection of incoming material. No anomalies were noted by the team in the manufacturing process control activities that were observed.

Charaina & Conditionina The team performed a comparison of the C&D procedures, practices and documentation used during the process of initial charging and conditioning of round cell strings with the charging and conditioning requirements (design bases) of Lucent's design and manufacturing specification KS-20472.and associated documents. The team reviewed the charging records for string 345-96 for McGuire Station and strings 223-92 and 224-92 for. Oyster Creek.

C&D conducts all of its manufacturing activities (including charging and conditioning) in the same manner, in accordance with its documented ISO-9001 quality system in conjunction with Lucent's design-and specification documents. However, the team identified several instances 4

in which C&D's procedures, practices and documentation for the charging and. conditioning process departed from the requirements of-the Lucent ,

design specification. The following are examples of these departures:

• KS-20472, Section 5.25, Step (c), prescribed the initial formation charge'for the string. The step included a cautionary note that '

stated that the cell temperature should not be allowed to exceed a certain value. Although C&D stated that the ten.peratures are monitored, wey were not recorded, nor was there a provision in the C&D record forms to do so.

• KS-20472, Section 5.25, Step (d), prescribed the first conditioning discharge following the initial formation charge, per Table 11 for-List IS round cells, to M : certain amperage for a certain duration. This discharge removed a specific amount of ampere (ah) from each cell of the string. The step included the warning "No individual cell should be allowed to go below (specified] volts."

KS-20472, Section 5.25, Step (e) required that, following the Discharge, the string be recharged to a specific percentage of the ampere-hours removed during the discharge for a minimum time duration. This charge then should have _ restored a minimum and a maximum value.

However, tne overall charging record sheet, called " Procedure Sheet 1," was pre-printed with the annotation, "[specified number) amps for icertain number of] hours" for this charge. This charge would put in more ampere-hours than allowed by the specification.

Although C&D produced a 1986 memorandum from Lucent (then AT&T/ Bell Laboratory) authorizing-this departure from KS-20472, the specification had not been updated,_in over four years, to reflect this mndification to the prescribed charging and conditioning (C&C) process. The charge records reflected that the string was, in fact, recharged per Procedure Sheet 1 instead of KS-20472.

• KS-20472, Section 5.25, Step (f), prescribed a "First Taper Discharge," per Table 111 for List IS round cells, to be for a specific periods at certain current values to remove a specific number of ampere-hours. The first post-taper recharge per Step (g) was supposed to ensure that a certain percentage of the ampere-hours were removed in the previous discharge. However, Procedure Sheet I was pre-printed with the annotation, "(specified value) amps for (certain number of] hours" followed by "(specified value of] amps for (certain number of] hours," more than allowed by the specification. Although C&D produced ancther 1986 memorandum from Lucent (then AT&T/ Bell- Laboratory) authorizing this departure from KS-20472, the specification did not reflect this four-year old design basis change to the prescribed charging and conditioning (C&C) process. -The charge records reflected that the string was recharged per Procedure Sheet 1 instead of KS-20472.

5

• KS-20472,.Section 5.25, Steps (h) and (j), prescribed the required sequential "[specified value)-Amp Discharges." The steps stated

" Discharge at the Table IV rate until the first cell reaches

[specified value) volts." However, C&D 01-20-0900.2, " Charging" (Section 6.3.16, " Water Discharges"), did not provide for normal discharge termination upon reaching a prescribed voltage (on the first cell to reach it). Instead, Step 6.3.16.3 called for termination "at the. normal discharge end time." The note above step 6.3.16.3 stated: "If any cell has a voltage of [certain number of) volts or less, terminate the discharge and notify your supervisor."

Therefore, the C&D procedure called for discharging until an end

-time that is not defined as the termination point for this particular discharge in KS-20472. Also, the C&D procedure prohibited allowing any cell to go below a certain voltage, but did not require discharging until reaching a specified voltage on the first cell that reaches this voltage as the language of the Lucent specification states.

In view of the explicit language in Step (h) of the specification (as distinguished from languaga explicitly prohibiting discharging below a specified voltage elsewhcra in the specification), the language in the C&D 01 was contrarv *n the requirements of specification KS-20472. As a result, the charge records showed that both specified discharges went only to a certain voltage on the lowest cells. Although the practice of discharging until the first cell reached the specified voltage (instead of the prescribed voltage) had been authorized in 1986 on a limited, experimental basis, C&D could not provide documentation authorizing this departure from KS-20472 for general production use. This resulted in the string not being discharged in this step to the extent required by KS-20472.

KS-20472, Section 5.25, Step (1), prescribed the recharges for the two sequential "[specified value]-Amp Discharges" of Steps (h) and (j) respectively. The recharges were supposed to be a specified percentage of the ampere-hours removed in the previous discharge.

Since both discharges continued for several hours, they were I

discharged to a certain value. Therefore, according to the KS-20472 formula, a specified minimum of the ampere-hours should have been recharged. However, Procedure Sheet I was pre-printed with the annotation, "[specified value of) amps for [certain number of) hours," less than required by the KS-20472 formula using the abbreviated discharges. This was clso much less than intended by KS-20472 because the recharge amount, even if calculated and performed in accordance with the KS-20472 formula, would have been based on a discharge to a nominal voltage, less than the amount of discharge specified by KS-20472 in Steps (h) and (j).

l

• KS-20472, Section 5.25, Step (k), prescribed a "Second Taper," or l "[specified value)-amp Taper" discharge per Table V for List IS I

round cells, to be conducted in several steps at specific values.

Step (1) required discharge to "[a specified value of] volts per 6

l l

14

cell average," or for a specified ampere-hour value with the caution that no cell should be allowed to go below a certain voltage.

However the discharge record (Form RS533) showed that no cell went below the minimum allowed voltage during the phase and none reached the specified voltage during any part of the second taper discharge.

Although the minimum total ampere-hours were discharged, no cell achieved the lowest voltage specified by KS-20472.

• The final capacity testing specified by KS-20472 following cha.rging and conditioning was modified for Strings 223-92 and 224-92 by a procedure required by Lucent's customer, GPU Nuclear, licensee for Oyster Creek. GPU Nuclear had contracted National Technical Systems (NTS), Inc., of Acton, Massachusetts, to perform the dedication of the round cell strings for Oyster Creek purchased from Lucent as commercial grade items. N15 Procedure 60145-93N, Pevision 2, was used as the basis for dedication of the strings. The NTS dedication procedure modified the standard capacity test discharge rate following the specified float period prescribed in KS-20472.

Instead, the NTS procedure called for a high-current discharge, recharge, and a several hour discharge, intended to better simulate design basis service conditions at Oyster r. reek. The test record contained a notice of anomaly indicating that only 95-percent capacity had been demonstrated by the discharge tests, but that this was acceptable. The explanation of the reduced capacity was the interruption of the standard " conditioning" sequence to perform the NTS dedication tests. However, this language did not accurately characterize the circumstances in that the conditioning phase of the KS-20472 process, including the final several-day float charge, was, in fact, completed. Only the final capacity test discharge, not considered part of conaitioning, was omitted in favor of the NTS multiple hour discharges.

• In reviewing the test and cell selection records for strings 223-92 and 224-92, the inspector noted that one test for 223-92 was terminated several minutes early. A notice of anomaly (NOA) also noted the early stop, but had the wrong time. In addition to the records having a number of other anomalies, such as not all forms being fully or correctly filled out and some erroneous operation ending times recorded. For example, one recharge for a specific discharge was listed as staring at a certain value on 11/21/92, but ending at the same value on 11/22/92, a longer period than specified on Procedure Sheet 1; yet, the second discharge was recorded as starting earlier on 11/22, and the inspector believed this entry to likely be the correct one. The charge end time entry of the first entry on 11/22 was apparently in error, but there was no indication that the record had been reviewed and the error identified by C&D QA, engineering or management or representatives of Lucent or GPU Nuclear.

7 The_ inspector also noted that the cell in position 43 in string 223-92-(cell serial no.- 78471); finished the test with a low-out-of-specification voltage,-yet this cell was selected for shipment by GPU Nuclear according to the selection record. Conversely, although the cell in positicn 64 had a low voltage, but met the acceptance criterion, it was rejected. The record indicated no rationale for tha apparently inappropriate cell selection, nor_ could C&D offer any explanation.

• The inspector also reviewed the charging and conditioning records-for round cell string 345-96, produced for Lucent's customer, Duke Power for Class IE service at McGuire Station. The-same process was used for this string as for_ the Oyster Creek strings discussed above except that the standard final capacity test discharge'was conducted. Therefore, this string underwent charging and c0nditicning with the same deviations from KS-20472 as occurred with the Oyster Creek strings, only some of which, as above, had ever been authorized by Lucent memorandum, and still without formal revision to KS-20472 (Issue 8 still in effect).

Chemical Analysis of Ecoxy Mixture The team reviewed the-applicable section of KS-20472 regarding preparation and mixture verification of epoxy used to s al the round cell jars around the terminal posts. Also reviewed were the Lucent epoxy titration procedure referenced by KS-20472 that was prescribed on AT&T Drawing L193472, Issue 5, and the C&D epoxy titration procedure, BT-4, Revision 0, dated March 7, 1995.

The procedures specify the quantitatfve analysis by titration of percent resin in hardener (done on each new mixed batch) and of the equivalent weight of hardener (done on each lot of hardener and each shift), the results of which are used in the percent-resin-in-hardener analysis calculation.

Following this review, the te m observed a routine set of epoxy titration analyses, performed ay C&D's usual analyst in the C&D laboratory. The analyses observed were the determination of percent hardener-in resin, and the determination of the equivalent weight of hardener. The team identified two instances in which C&D epoxy titration procedure BT-4 deviated from AT&T Drawing L193472, specifically:

L193472 called for titrating into the. sample in a 125-ml Erlenmeyer flask; whereas BT-4 prescribed a 250-m1 flask, and

• L193472 required adding a specific number of drops of Brom Creosol green indicator; whereas, BT-4 specified a range of drops that could be added.

The team also determined that the size of the flask can affect the efficiency of mixing of the sample (by means of a magnetic stirrer) with the indicator, and more importantly, with the drops of titrant (sulfuric acid). The amount of indicator added can affect the visibility of the endpoint. Other deficiencies in C&D procedure BT-4 noted were:

8

l-:

f

• -The C&D procedure inappropriately defined _the titration endpoint as -

ia color change from blue-to vellow:- however, the analyst correctly- ,

described the color change as being from blue to creen..

• - The acceptance criteria for-percent-hardener-in-resin and for hardener equivalent weight given on L193472 were not-in the C&D

procedure fornuse by the analyst, although the analyst was familiar

-with them. ,

t

• - -No. sample vortex characteristics were specified to ensure uniform mixing in the flask.

The team observed two instances in which the analyst deviated from C&D-

procedure BT-4:

• Instead of the 250-m1 Erlenmeyer sample flask specified by BT-4, the analyst used a 125-m1 flask as_ specified by the Lucent drawing; although the analyst was unaware of the drawing requirement. He explained, correctly, that for the specified 50-m1 sample size, the smaller flask provided the depth of liquid necessary to achieve an adequate vortex (using the specified magnetit stirrer) for proper mixing of_the indicator and the titrant in the sample.

• -The C&D analyst stated that he typically adds more drops of indicator than that specified in BT-4, nearly twice as much indicator as specified in drawing L193472.

In addition, the analyst did not employ the standard titration technique when near the expected endpoint of so-called " drop splitting" or " drop transfer" using a glass rod to take each drop allowed to form at the burette nozzle as the titrant level is lowered by the smallest burette graduation and placing the drop in the sample. Instead, tht analyst simply slowed the expected near-endpoint rate of titration to about I drop'per second,-which may not allow for complete mixing and reaction of the titrant with the sample and indicator so as to show the endpoint when it first occurs before the next drop falls into the sample. This practice, with a less experienced analyst, and also using the 250-m1 flask if BT-4 were followed, could cause the actual endpoint to be missed, thus degrading the sensitivity of the analysis or the accuracy of the results.

Audits and Surveillance of C&D-leola by AT&T/Lqgini A review of the AT&T Quality Management & Engineering Department (QM&E) audits and surveys indicated that AT&T/ Lucent performed on-site audits of C&D to observe their capabilities and performance as required by Lucent's qual _ity manual and technical specifications and drawings. The team noted that--_the QM&E personnel identified numerous quality problems over

.a several year: period. - However, contrary to .the 10 CFR Part 50, Appendix B,: quality assurance requirements which were imposed upon-AT&T/ Lucent, the team determined that Lucent did not take prompt; action to address the effectiveness of C&D's control of . product quality.- That

.is, although Lucent identified numerous exa:::ples _of conditions adverse 9

_ _ _ _ _ _ _ _ _ _ _ _ \

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

to-quality over a three year period-at the C&D-Leola facility, Lucent _

did not promptly correct those conditions adverse to quality identified or assure that C&D addressed the ineffectiveness of its program' i impi m ntation consistent with the importance and complexity of the~-

round call: product.- it was also noted that the quality and reliability

-of_the-product that was-shipped to nuclear power. plants did not appear- -

to-be-~ addressed in the: documents reviewed by the team.

Consequently, the team determined that Lucent did not assure that the manufacturing services'that it contracted from C&D, for fabricating its round cells for at least the McGuire station, consistentiy conformed to the procurement document specifications and requirements during the charging and conditioning manufacturing activities.

' For example, a November 2,1994, AT&T Bell Laboratories letter to the-C&D Leola facility quality control (QC) Manager and associated documents discussed the findings of a September 21, 1994, mini quality audit of the round cell manufacturing and quality systems. AT&T's correspondence indicated that 15 separate areas were investigated during a period when C&D was in the process of increasing production rates of its round

-cells. The correspondence indicated that 11 of the 15 areas audited did not conform to AT&T's specification recnirements. The correspondence also indicated-that in light of a recent problem encountered with 1992 series ISH cells sold to Arizona Public Service (Palo Verde), "it would behoove C&D to spend additional time and effort on training new employees and increasing inspection" at those operations where these new employees are located. The conclusion reached by.the AT&T auditors was that the overall round cell-quality was not up to usual standards. The letter to C&D itemized specific areas of concern and requested C&D's corrective action plan.

A December 19, 1994, AT&T Bell Laboratories Trip report documenting a December 14, 1994, follow-up mini quality audit stated that C&D had not compiled or composed the required corrective action plan, and the overall manufacturing process qualit) was found to be lower than C&D's usual standards. The trip report stated that although the observed deficiencies were not critical, they indicated "a general relaxation in standards "

The December 1994 report stated that on December 14, the inspection of round cell manufacturing process quality indicated that the manufacturing quality standards were continuing to trend downward. The -

letter indicated that the observation had been made over a period of

time, was not specific to any one area, and that-C&D's emphasis on quality and the commitment-to quality have diminished over the last six months.--The staff engineer rcuommended that AT&T and C&D meet to address:both Companies' concerns and to " refocus on the critical importance of' quality to this product's continued success."

Three months.later on March 15, 1995, another mini quality audit was conducted and identified findings in 18 areas at the C&D-leola facility.

. The overall conclusion reached was that quality is poor." The AT&T 10

J staff performing the audit stated that work instructions need to be reviewed with both the operators and C&D quality control inspectors and the importance of producing good-quality parts needs to be reinforced.

c. Conclusions -

No anomalies were noted by the team in the manufacturing-process control activities that were observed during the performance of this inspection.

However, in the charging and conditioning process area, the. team concluded that:

• C&D charging and conditioning procedures and practices deviated from KS-20472 and only some of the departures were authorized by Lucent which had not updated its specification in several years to reflect the different practices

• C&D procedures were not always followed and records were not always-fully accurate, nor apparently carefully scrutinized by Lucent, the ultimate user or its representatives, and

• the 64 cells selected by GPU Nuclear from each of the 72-cell strings produced and tested, did not all meet the acceptance criteria following the final test discharge.

In the chemical analysis of epoxy mixture area, the team concluded that the anomalies from the AT&T/ Lucent drawing as well as the quantitative analysis practices employed could affect the accuracy and repeatability of the epoxy titration analyses for percent-resin-in-hardener and equivalent weight of hardener.

Additionally, although there is a wide range in the_ acceptance criteria for-the epoxy titration analyses, the team was given the understanding that neither Lucent nor its customers had identified any of the anomalies observed by the team in the titration process. Therefore, the effects of these anomalies had not been evaluated by Lucent or their customers to determine their significance.

The team concluded that the AT&T/ Lucent auditors identified negative quality aspects that would be indicative of an ineffectively established or implemented quality program. However, the team noted that neither Lucent nor C&D took appropriate measures to determine the cause of the problem and implement corrective action.

Additionally, neither the product or process appeared to be appropriately questioned regarding quality and reliability of- the product. These issues are also discussed in NRC's inspection results of Lucent's Dallas, Texas facility, Inspection Report 99901309/96-01-02.

11

1 13 . 4 Entrance & Exit Meetinac  ;

In the entrance meeting on January-6, 1997, the NRC Team Leader l discussed the scope of_the inspection, outlined the areas to be  ;

inspected, and established-interfaces _ with C&D management. In the exit meetir.:'on January 9,1997, the inspectors discussed their findings and-t Concer.1s . ,

t 3.5 PERSONS-CONTACTED C&D Charter Power Systems. Inc.

G. Stoermer, Plant Manager-Leola '

T. Kinden. Director, Quality Assurance ,

D. Johnson, Q.C. Manager (L;ent lechnoloaies-Bell Laboratories. Murray Hill. New Jersey Staff H.C. Weeks. Technical Staff J.B. Baldasty, Technical Staff 12

pA #8cg I#t UNITED STATES

[ NUCLEAR REGULATORY COMMISS'ON

's .

WASNINGTON D.C. 30MH001

1. 's, April-29, 1997 Mr. James Dailey, Manager, Quality Assurance DuBose National Energy Services, Inc.

900 Industrial Drive -

Clinton, NC 28328 )

l

Subject:

NRC INSPECTION REPORT 99900861/97 01 AND NOTICE OF NONCONFORMANCE l

Dear Mr.' Dalley:

On January 24,199'/, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection of DuBose National Energy Services, Inc. The enclosed report presents the results of that inspection.

During this inspection, the NRC inspectors found that the implementation of your quality assurance pror, rem failed to meet certain NRC requirements imposed on DuBose by customer,s. Specifically, the inspection identified instances where DuBose accepted material from qualified suppliers without receiving all documentation required by the applicable ASME Code or your purchase ordsr provisions. In'severalinstancet DuBose also supplied material under your ASME Quality Systems Certificate without providing all documentation required by the applicable ASME NCA 3800 quality assurance program. Additionally, the inspection identified tnat DuBose supplier audits did not consistently identify objective evidence to support the audit conclusions, and the audit conclusions were not always consistent with the approved vendor list scope restrictions or the purchate order conditions.

These noncanformances are cited in the enclosed Notice of Nonconforma' ce (NON), and the circumstances surrounding them a e described in detall in the c nclosed report. You are requested to respond to the nonconformances and should follow the instructions specified in the enclosed NON when preparing your response.

_ . _ . _ _ . _ . _ _ = . .._ _.

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 NRC's Public Document Room.

Sincerely, >

, .6 W. t, Robert M. Gall,., C sef Special Inspection Branch Division of Inspection and Support Programs Office of Nuclear Reacior Regulation -

Docket No.: 99900861

Enclosures:

1. Notice of Nonconformance

2. Inspection Report 9990861/97-01 9

22-

NOTICE OF NONCONFORMANCE DuBose National Energy Services,inc. Docket No.: 99900861 Clinton NC Based on the results of an inspection conducted on January 21 through 24,1997;-

it appears that certain of your activities were not conducted in accordance with .

NRC requirements.

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

Paragraph NCA-3861(b) states that the Material Organization shall transmit all certifications received from other Material Organizations or approved suppliers to the purchaser at time of shipment.

Paragraph NCA 3862.1(b) of Section lli of the ASME Code states that "When required chemical analyses (including mill heat analysis), heat treatment, tests, examinations, or repairs are subcontracted, the approved supplier's certification for the operations performed shall be furnished as an identified attachment to the Certified Material Test Report."

The following examples demonstrate failure to comply with the above requirements anc constitute Nonconformance 99900861/97 0101

1. Contrary to the above, DuBose Certified Material Test Report (CMTR) for material supplied to San Onofre Generating Station (SONGS),

under Purchase Order (PO) 6N236016, certified that this material was supplied in accordance with their ASME Quality Systems Certificate (OSC), but did not reference or include the mill heat analysis as an identified attachment.

2. Contrary to the above, DuBose CMTRs for material supplied to Union Electric Co. under PO 094006, dated June 24,1996, and to Carolina Power & Light Co. (CPL), under PO BM1717 CJ, dated November 1, 1996, certified that the material was supplied in accordance with their OSC, but did not reference or include certifications from approved suppliers as identified attachments to their CMTRs. ,

-23

f B. Criterion Vil, " Control of Purchased Material, Equipment, and Services," of Appendix 8 to 10 CFR Part 50, requires, in part, that purchased material coriforms to procurement documents.

Paragraph 10.3.1.4 of DuBese qu6;ity systems program (OSP) requires, in part that GA will review the vendor documentation against requirements of the PO, material specification, and, when required, the ASME Code. When unacceptable docuniontation is received, the meterial will be segregated until proper documentation can be obtained.

Paragraph 2.3.4.1 of DuBose OSP re auires, in part, that POs shall require certification that the material was fv nished under a program accepted by DuBose.

The following examples demonstrate failure to comply with the above requirements and constitute Nonconformance 99900861/97-0102

1. Contrary to the above, DuBose accepted material supplied under PO 2284 63, dated February 5,1993, without the starting material supplier being identified on the CMTR as required by this PO.

2. Contrary to the above, DuBose accepted material under POs 2284 63 and 1546 63, dated February 5,1993, and August 24,1992, respectively, without documente'd evidence that all requirements of the material specification (macroetch test) had been completed.

3. Contrary to the above, DuBose accepted material under POs 11645-66 and 10612-62, dated December 9,1996, and December 12, 1996, respectively, without certification that the material was furnished under a quality program accepted by DuBose.

C. Criterion IV, " Procurement Document Control," of Appendix B to 10 CFR Part 50 requires, in part, that requirements necessary to assure adequate quality are included in the documents for procurement of material.

Paragraph 2.3.4.1 of the DuBose QSP requires, in part, that POs shall require certification that the material was furnished under a program accepted by DuBose.

2

1 e

h Contrary to the above, DuBose issued POs 11495-61, and 1'0649I 68, dated; November 11,1996,' and November.15,1996, respectively, without _

l Including a requirement that the material must be furnished under a program

accepted by DuBose.- These POs had been reviewed and approved by DuBose QA on November 12,-1996, and November 18,1996, respectively.

(Nonconformance 999008610103)

.D.  : Criterion Vil, " Control of Purchased Material, Equipment, and Services," of Appendix B to 10 CFR Part 50 requires', in part, that the effectiveness of the control of quality by cor. tractors and subcontractors shall b.e assessed at

~

intervals consisten't with the importance, complexity, and riuantity of the j product Paragraph 7.4.1 of DuBose QSP states, in part, that evaluation (of vendor) is parformed to adequately evaluate the quality and technical capability of the vendor.

Paragraph 2.4.3 of DaBose QSP states, in part, that the approved vendor's list (AVL) shall be compiled on the basis of a site evaluation of the vendor's quality program and Paragraph 2.4.2 states that the AVL shall show the scope of activity approved with any limitations. ,

Paragraph 2.3.2 states that quality related material and services shall only be purchased from vendors appearing on the AVL.

The following examples demonstrate failure to comply with the above requirements end constituta Nonconformance 99900861/97 01-04.

1. Centrary to the above, DuBose audit of Colonial Machine Company on April 11,1996, did not document that all areas of the AVL approved scope of activity had been evaluated and determined to be adequately '

controlled. Specifically, the audit documentation,did not describe what was reviewed in the areas of rc0E, heat treatment, upgrading of unqualified source material, and control of suppliers of source material and weld repair activities were not addressed.

2.' Contrary to the above, DuBose audit of Capitol Manufacturing Company on February 13,1996, did not document that all_ areas of the AVL approved scope of activity had been evaluated and were adequately controlled. Soecifically,-the audit documentation did not u

(

(- .

l-3 i

-- . , - . - =__ _ --  ; - ~ _ _ - - .

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

5 describe what was reviewed in the areas of material testing and NDE, and did not address heat treatment and the. upgrading of unqualified.

- source material.1

~ 3. - Contrary to the above,' on November 11,1996,-'and on November 15,  ;

L 1996, DuBose issued POs 11495 61 and 10649 68, respectively, to Marmon/ Keystone Corporation for material that was not included in -

the product scope of the' AVL.

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, Special inspection Branch, Division of Inspection and Support

- Programs, 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 Nonconformance" and should 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. l Dated at Rockville, Maryland

. this day of 1997-Enclosure 1 j

9 E

4 4

i i

~

, s--r-- ,- ---+v.w . w e ,- -, - -. .-- ew-~.- - - -ww -m!-- y

t U.S. NUCLEAR REGULATORY COMMISSION

' OFFICE OF NUCLEAR REACTOR REGULATION t

Report No: 99900861/97 01 Organization: DuBose National Energy Services, Inc.

Contact:

James Dalley, Quality Assurance Manaper (910) 590 2151 Nuclear Industry Activity: Supplier of threaded fasteners, bars, tubular products, forgings, plates, flanges, fittings, and other items used primarily for nuclear applications.

Dates: January 21 through 24,1997 Inspectors: Uldis Potapovs, Senior Reactor Engineer Steven M. Matthews, Quality Assurance Engineer Approved by: Gregory C, Cwalina, Chief Vendor inspection Section Special Inspection Branch Division of Inspection and Support Programs Enclosure 2

1 INSPECTION

SUMMARY

During this inspection, the inspectors reviewed the implementation of selected portions of DuBose National Energy Services, Inc. (DuBose) quality af.surance (QA) program for supplying material under the American Society of Mechanical Engineers (ASME) Boller and Pressure Vessel Code (Code) requirements, and for providing dedicated commercial grade material. The inspection was focused on the review of DuBose activities related to the supply of nonconforming material to Southern California Edison for use in the San Onofre Generating Station (SONGS).

The inspection bases were:-

e Appendix B, " Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," to Part 50 of Title 10 of the Code of Federal Reoulations (10 CFR Part 50).

e 10 CFR Part 21, " Reporting of Defects and Noncompliance" e ASME Code, Section Ill, Subarticle NCA 3800.

During this inspection, four instances where DuBose failed to conform to NRC requirements imposed upon them by NRC licerisees were Identified.

These nonconformances are discussed in Sections 3.2 and 3.3 of this report.

2 STATUS OF PREVIOUS INSPECTION FINDINGS This was the first NRC inspection of DuBose at its present location and under the present management. The last NRC inspection of DuBose was conducted on May 21 through 25,1984, at Roseboro, NC. The status of findings from that inspection were not reviewed.

3. INSPECTION FINDINGS AND OTHER COMMENTS 3.1 Descriotion of Facilities and Activities DuBose has been accredited by the ASME as a Material Organization (MO),

supplying a wide range of metallic materials, including bars, threaded fasteners, castings, forgings, plates, fittings, flanges, tubular products, structural shapes, and weld filler material. According to DuBose management, approximately 70 80% of sales volume consists of safety related material supplied to nuclear utilities. Of that amount, approximately 50% is sold under their ASME Quality Systems Certificate as complying 2

with the requirements of NCA 3800. The company has an on site laboratory facility capable of performing spectrochemical analyses, tensile testing, hardness testing, and hydrostatic testkig. No manufacturing noerations, other than subdividing of material, are performed.

3.2 Follow un on Nonconformina material sucolled to SONGS

a. Backaround On May 10,1996, DuBose issued a Certificate of Conformance/ Compliance /

CMTR (CMTR) to SONGS for forty six 3/4-10 ASME SA 194, Grade 7, heavy hex nuts, supplied as item 2 of SONGS purchase order (PO) 6N236016. SONGS receipt inspection of these nuts resulted in a nonconformance report (NCR) for failure to meet System 22 dimensional inspection requirements, as specified in the PO. lt was also identified that chemical analysis of ten nuts showed that, while they complied with the specification limits, the nuts apparently came from two different heats of material (all 46 nuts were certified as coming from a sirule heat). SONGS performed a 10 CFR PART 21 reportability evaluation and determined that this event was not reportable.

Review of documentation obtained from SONGS indicated that the material was shipped from T&T Enterprises, Corona California, a supplier qualified by DuBose. DuBose provided the CMTR for this material, referencing their ASME QSC nu,mber and certifying compliance with the PO and ASME Code requirements. Attached to the DuBose CMTR was a CMTR from T&T Enterprises, certifying the performance of dimensionalinspection (System 22), visualinspection, and macroetch (ASTM E-381), in accordance with their DuBose approved QA program. Also furnished with the material was a CMTR from A&G Engineering II, Inc. (A&G), dated February 8,1993 and issued to DuBose. The CMTR certified that the material was provided under A&G's OSC and contained the results of chemical and mechanical tests, apparently transcribed from Korea Bolt Ind Co., LTD., inspection Certificate, dated October 10, 1 ~491, attached to the A&G CMTR. The certifications did not include a CMTR from the producing steel mill, Lnd there was no indication of who produced the steel.

On September 12,1996, NRC contacted DuBose and requested information concerning the apparent loss of traceabi;ity for this material. DuBose responded on September 19,1996, by stating that, based on their review and additional chemical analyses, they determined that there was no conclusive evidence to indicate that the nuts were from mixed heats of material. This conclusion was based primarily on the fact that the heat code was forged into each nut by the manufacturer and that the chemical 3-

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

. overcheck results (15 samples) were within)cceptable variance limits of  ;

ASTM- A 29. -

i

bl Ohnervations and Findings  !

_. 1 Documentation available at DuBose for the material discussed above did not contain a CMTR from the producing mill. Also, other than the macrootch -

that was apparently. performed by T&T Enterprises, as discussed above, on a finished nut, there was no indication that macroetch testing was performed as a part of the steel producer's quality procedures, as required by the -

applicable material specification,"ASME SA 194 A macroetch test on'a -

= forged nut, performed after the completion of manufacturing is not-i considered an acceptable method of achieving compliance with the material specification requirement. With respect to the missing CMTR from the steel-

_ producer, the inspector noted that, for material supplied under DuBose's '

OSC,~ paragraph NCA 3862.1(b) of the ASME Code requires that certifications of approved suppliers be furnished as identined attachments to the QSC holder's CMTR. In this case, the steel mill was considered an approved supplier, and the mill CMTR, or heat analysis, should have been available, and should have been provided to the recipient of the nuts. ,

With respect to DuBose's comment, as discussed above, that forging of the heat code into each nut during the manufacturing process assures heat traceability, the inspector noted that the forged in heat markings could not be relied upon so provide full heat traceability for this situation, since a material mixup may have occurred at the producing mill or at an intermediate supplier to the nut manufacturer.

With respect to DuBose statement that, since the results of their chemical-analyses of-15 nuts were within acceptable product analysis limits'of ASTM A 29, there was no conclusive evidence that the nuts were from mixed heats, the inspector noted the applicable variance limits in A-29 were not i

intended to verify heat homogeneity, and that multiple analyses of a single heat of material would be expected to show less variability than indicated in the tables referenced in A 29..

Review of the nonconforming thread issue identified during SONGS receiving

!- inspection showed that DuBose issued Non Conformance Report (NCR) No.

1039 on June 18,1996, as a result of communication from SONGS. Based on the NCR, a Corrective Action Report was issued to T&T Enterprises on 1

- June 19,'1996,- describing the unsatisfactory condition, and requesting a response by July 19,1996.- On September 6.- 1996, T&T Enterprises provided a response by stating that the internal threads were gaged with GO NOGO plug gages and GO-NOGO cylinder gages which satisfied the ,

~ ..= .. - - . - . - --- - - - -- =-

J functional requirements of System 22. They identified the cause of the nonconfortnance as unavailability of the cone and vee type internal variable gages to appraise the values of the internal pi'ch diameters over the lull length of the thread. The described correct:ve action was to assure the proper equipment !s used to make these measurements on future orders.

DuBose accepted the proposed corrective action on September 9,1996.

The inspector also reviewed several procurement docurnents related to DuBose's acquisition of material from A&G, as summarized below:

e The nuts oiscussed above were purchased as a part of DuBose PO 2284-63, dated February 5,1993 for approximately 85,000 ASME SA-194 nuts in sizes ranging from 5/8 inch to 1 1/2 inches. The nuts were to be provided under A&G's OSC. The inspector noted that PO contained an attached Form D 2. Paragraph 6. of this form stated that "The starting material manufacturer is to be identified on the CMTR produced by your company." Paragraph 7. of the same form stated that the use of foreign starting materialis not permitted without DuBose's approval and that, when approvalis given, backup certification is to be provided. The PO 'was amended on February 8,1993, to allow foreign starting material. All otner cone'itions were to remain the same. There was no further explanation concerning " backup certification." The inspector's review of documentation supplied by A&G did not identify any reference to " backup certification," and did not show any evidence that the starting material manufacturer had been identified as specified in the atta'chment to the DuBose PO for this material, e On August 24,1992, DuBose issued PO 1546-63 to A&G for 20 tons of threaded rod in accordance with ASME SA-193, Grade B7 in 12 foot lengths, and diameters ranging from 1/2 to 1 inch, to be supplied under A&G's QSC. The A&G CMTR for this material contained apparently transcribed results of the required chemical and mechanical tests and certified the material as being provided in accordance with their QSC.

The CMTR did not include certification from the steel mill or include any certifications by approved suppliers. Also, there was no indication that a macroetch test, as required by the material specification, was performed during the steelmaking process,

c. Conclusions As discussed above, the inspectors determined that, for SONGS PO N236016, DuBose did not provide all of the documentation required by the ASME Code provisions governing the application of their QSC when certifying Code material. Specifically, contrary to the requirements of ASME Code, SC lli, paragraph NCA 3862.1(b), DuBose did not include the producing mill heat analysis as an identified attachment to their CMTR. This item was identified as an example of Nonconformance 99900861/97-01-01.

The inspectors also determined that, contrary to Paragraph 10.3.1.4 of their QA Manual, A&G certifications for material supplied to fill DuBose POs 2284-63 and 1546 63 were either not reviewed by DuBose QA, or the review was not effective. Specifically, this material was accepted without identification of " starting material manufacturer" on these certifications, as required by Form D 2, attached to the DuBose POs. Additionally, the A&G certifications did not indicate that all of the testing, required by the material specification (macroetch), had been performed These items were identified as examples of Nonconformance 99S00861/97 01-02.

With respect to the mixed heat issue, based on review of the results of DuBose chemical analyses of 15 nuts, it appears likely that more than one heat of material was represented in the sample. However, as discussed above, since none of the samples tested by DuBose or SONGS failed to meet the requiremerits of the material specification, co-mingling of heats, if it did occur, appe*,rs to have been limited to the same grade of material.

Additionally, a July 15,1996, memorandum from the DuBose QA Manager to its Prwoent Indicates that this material was being moved to the DuBose commercialinventory. The dimensional nonconformance of these nuts apparently resulted from the subcontractor's failure to perform all of the required measurements, and was dispositioned through the DuBose corrective action process.

3.3 QA Prooram imolamentation 3.3.1 Prooram Descriotion The DuBose quality assurance program is described in their Quality System Program (OSP) and in Quality Control Procedures (OCPs). Revision 5 of the QSP, dated January 10,1996, was in effect at the time of the NRC inspection. The QSP was written to comply with the ASME Code, Section lli requirements applicable to Material Organizations, and also indicates compliance with 10 CFR 50, Appendix B, MIL l 45208, MIL Q 9858, NOA 1, ANSI 45.2, And ISO 9002.

Only selected areas of the QA program and its implementation were reviewed during this inspection.

2

l 3.3.2 Procurement Control t

a. inanaction Scoon  !

The inspectors reviewed DuBose approved vendors list (AVL) dated January 21,1997, and selected several suppliers from the AVL to evaluate DuBose capability to quellfy vendors for placement on the AVL.

b. Dhaarvations and Findinns b.1 Approved Vendors According to Section 2.4, " Approved Vendors List," cf DuBose OSP, vendors are placed on th6 AVL either on the basis of a site evaluation of the vendor's OA program in accordance with Section 7.4, " Vendor Surveys and Audits," of the OSP, or on the basis of verification that the vendor has an ASME issued OSC or an N type Certificate of Authorization. Vendors that are placed on the AVL based on a DuBose site evaluation are qualified to one of two levels: For material traceability only, or as a DuBose quallfled MO with OA program meeting the requirements of Subarticle NCA 3800 of Section lll of the ASME Code.

To evaluate DuBose capability to qualify vendors for placement on the AVL as an approved supplier, the inspectors reviewed the following audits:

(1) Marmon/Khystone Corporation, Charlotte, North Carolina The last DuBose audit of Marmon/ Keystone was performed on October 13,1994. The scope of the audit identified Marmon/ Keystone as a supplier of ferious and non ferrous, seamless, and welded without filler metal, tubular products. Based on a review of the audit report, the inspectors determincd that DuBose had adequately evaluated Marmon/hystone, Charlotte, North Carolina, as a supplier of source mateoal with controlled traceability.

However, the AVL's approved scope for purchases from Marmon/ Keystone was, "MS carbon welded tubular products in accordance with QA manual,1st Edition, Revision 0, dated January 27, 1988." Therefore, in this instance, the AVL approved scope for purchases was inconsistent with the scope of DuBose's audit of Marmon/ Keystone since it did not include nonferrous or seamless tubular prcducts, 7

I

(2) Alloy Rods incorporated The last DuBose audit of Alloy Rods Inc., was performed on April 15, 1994. The scope of the audit was, *MO - ferrous and nonferrous bare i and covered wire.* On the basis of their review of the audit, the inspectors determined that DuBose had adequately evaluated Alloy Rods Inc. as a quellfled MO.

However, the AVL's approved scope for purchases from Alloy Rods Inc.

was, "MO - ferrous and nonferrous braring material, bare and coverti  ;

flux cored electrodes, and bare wire." Therefore, the AVL's approved l scope for purchases was not consistent with, and was expanded beyond the scope of the DuBose audit of Alloy Rods Inc. Specifically, <

the audit scope did not include braring material or bare and covered flux cored electrodes. This condition would permit DuBose buyers to procure material that was not included in the audited, and therefore, qualified capability of Alloy Rods Inc.

5 (3) Colonial Machine Company,Inc., Pleasantville, Pennsylvania The last DuBose audit of Colonial Machine Company was performed on April 11,1996. The scope of audit was: " MO - ferrous and nonferrous fittings, flanges and pipe hangers - machining and procening of DuBose supplied material." However, the Vendor Quality Assurance. Survey / Audit Cover Sheet (Form D 46), also listed additional items including the following:

o bars e plates e scam less fittings e forgings e flanges e seamless tubing product '

e NDE (nondestructive examination)

• HT (heat treatment)

• weld repair o material testing e upgrading unqualified source material e approval & control of suppliers of source material e approval & control of suppliers of services l

. . v..--.. . , 4-, ._--,i 3 --- .m,,.,.- ,._, -y-. .- ,. --. -

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

t On the basis of their evaluation of DuBose April 11,1996, audit of Colonial Machine Company, the inspectors determined that the audit report did not contain objective evidence that the audit had adequetely "

evaluated the following areas:

e for NDE, HT, upgrading unqualified source material, and approval & coritrol of suppliers of source material, the DuBose auditor did not document objective evidence of what was audited e weld repair was not addressed e for approval & control of suppliers of services, the DuBose auditor did not document objective evidence that supplier audits were evaluated.

Additionally, the AVL's approved scope for purchases from Colonial Machine Company was, "MO - ferrous and nonferrous fittings, flanges and pipe hangers - machining and processing of DuBose supplied material- welding in accordance with ASME B31.1 Code." Therefore, the AVL's approved scope for purchases was not consistent with and was expanded beyond the scope of DuBose's audit of Colonial Machine Company Specifically, welding in ac::ordance with ASME B 31.1 Code was not included in the audit scope. This condition would permit DuBose buyers to procure material that was not included in the audited, and therefore, qualified capability of Colonial Machine Company.

(4) Capitol Ma'nufacturing Company The last DuBose audit of Capitol Manufacturing Company was performed on Fet tuary 13,1996. The scope of audit was: "MO -

manufacturing ' arbon and stainless f.ttings." However, DuBose Form '

D 46, also listed additional items including:

e seamless fittings

• NDE

• HT e material testing e upgrading unqualified source material e approval & control of suppliers of source material e approval & control of suppliers of services

. 9 35-

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

. On the basis of its evaluation of DuBose February 13,1996, audit of Capitol Manufacturing Company, the inspectors determined that the audit failed to adequately document evaluation of the following areas:

o for NDE and material testing, DuBose did not document objective evidence of what was evaluated e HT was.not addressed e for upgrading unqualified source material, DuBose did not document a review of upgrading activities. '

b.2 Procurement Document Control To evaluate DuBose capability to control procurement documents, the inspectors reviewed the following purchase orders (POs):

(1) PO 11495 61 DuBose issued PO 11495 61, dated November 11,1996, to Marmon/ Keystone Corporation, Charlotte, North Carolina, for 20 feet (ft.), random length,8 inch nominal diameter, schedule 160, A 106 Grade B, seamless pipe. The PO was reviewed and approved by DuBose QA on November 12,1996.

However, the PO failed to impose on Marmon/ Keystone the requirement that materjal must be controlled in accordance with the quality control system and procedures audited by DuBose. Marmon/ Keystone also failed to provide DuBose certification that the material was furnished under a quality program accepted by DuBose.

Additionally, the AVL approved scope for purchases from Marmon/ Keystone was, *MS carbon welded tubular products in accordance with QA manual,1st Edition, Revision 0, dated January 27, 1988.* Since this PO procured seamless pipe, the PO did not comply with the limitations of the DuBose AVL.

(2) PO 10649 68 DuBose issued PO 10649 68, dated November 15,1996, to Marmon/ Keystone Corporation, Charlotte, North Carolina, for 60 ft.,

5/8 inch outside diameter (OD) x .083 wall x 20 ft., A 269 type: 304 stainless steel tubing. The PO was reviewed and approved by DuBose OA on November 18,1996.

'f 36- -

I

However, the PO failed to impose on Marmon/ Keystone the requirement that material must be controlled in accordance with the quality control system and procedures audited by DuBose. Marmon/ Keystone also failed to provide DuBose certification that the material was furnished under a quality program accepted by DuBose.

Additionally, the AVL approved scope for purchases from Marmon/ Keystone was, "MS - carbon welded tubular products in accordance with OA manual,1st Edition, Revision 0, dated January 27, 1988." Since this PO procured stainless steel tubing, the PO did not comply with the limitations of DuBose AVL.

(3) PO 11645 66 DuBose issued PO 11645 66, dated December 9,1996, to TAD USA, Inc., for one piece of 2-inch, schedule 80 x 10 ft., SA 335, P22, seamless pipe. The PO was reviewed and approved by DuBose QA on December 11,1996.

Even though DuBose imposed on TAD USA the requirement that material furnished must be controlled in accordance with the quality control system and procedures audited by DuBose, TAD USA failed to provide DuBose certification that the material was furnished under a quality program accepted by DuBose.

(4) PO 10612562 DuBose issued PO 10612 62, dated December 12,1996, to Castle, A.M. & Company for one piece of 2 3/4-inch thick x 35 inch wide x 36-inch long, SA 516 Grade 70, plate.

Even though DuBose imposed on Castle, A.M. & Company the requirement that material furnished must be controlled in accordance with the quality control system and procedures audited by DuBose, Castle, A.M. & Company failed to provide DuBose certification that the material was furnished under a quality program accepted by DuBose,

c. _ Conclusions The inspectors determined that the DuBose material traceability audits of suppliers appeared to be adequate for their intended purpose.

1 However, the inspectors determined that the audits of DuBose qualified {

NCA 3800 suppliers (MOs) were lacking in detail end programmatic in their i conduct and/or documentation. The inspr ctors identified two instances ]

(Colonial Machine Company, Inc., and Capitol Manufacturing Company) where DuBose audits of its vendors failed to (a) adequately assess the l effectiveness of the control of quality by its vendors and (b) compile its AVL, for all vendors, on the basis of an adequate site evaluation of the vendors quality program for the AVL's approved scope for purchases. These j audits often lacked objective evidence that upgrading and dedication activities were adequately verified or that the implementation of these activities was evaluated. These findings were identified as examples of Nonconformance 99900861/97 0104.

Additionally, in certain instances identified by the inspectors and described above, DuBose AVL approved scope for purchases was expanded beyond the scope of DuBose's audit and therefore permitted DuBose buyers to procure material that was not included in the audited and qualified capability of the supplier.

In other instances identified by the inspectors and described above, DuBose procured material form vendors on its AVL that was not included in the AVL approved product scope (POs 11495 61 and 10649 68). These instances were identified as examples of Nonconformance 99900861/97 0104.

The inspectors, identified two instances (POs 11645 66 and 10612 62) where DuBose accepted material from its vendor without certification that the material was furnished under a quality program accepted by DuBose.

These items were identified as examples of Nonconformance 99900861/97-01 02.

On the basis of its review of the POs desciibed above, the inspectors identified two instance (POs 11495 61 and 10649 68) where Dubose failed to require its vendors to certify that the material was furnished under a quality program accepted by DuBose. These items were identified as examples of Nonconformance 99900861/97 01 03, 3.3.3 Review of Recent Material Sales To NRC Licensees

a. Scoon The inspectors reviewed selected documentation packages for safety related material recently supplied to NRC licensees under the DuBose's QSP to verify the implementation of the applicable QA program requirements and compliance with 10 CFR 50, Appendix B, provisions.

1 l

-3 8-l

i l

b. Observations ond_Findinos Document packages for mater!al supplied to Appendix B requirements, as well as material certified under DuBose OSC, were reviewed. Significant observations and findings are summarized below.  :

e Union Electric PO 094006, dated June 24,1996, item (2) for twenty  :

ASME SA 564 Grade 630 studs (3/410,51/2in long), with 1100 F i aging treatment, and 302 BHN minimum hardness.

This material was certified under DuBose OSC as complying with the  !

applicable Code and PO requirements. Appended to DuBose CMTR was a certification from Walker Bolt, a DuBose qualified supplier. Attached to the Walker Bolt certification was certification from Ugine Savols ,

(French company) containing chemical and tensile test results, satisfactory macroetch statement and statement that the material was provided in solution treated condition. No quality program statement was attached to this certification, Indicating that Ugine Savois was not i en approved supplier and that the material was upgraded by Walker Bolt. However, no test reports, indicating that upgrading of this material was performed were attached to the Walker Bolt certification.

There was also nn indication of what organization performed the required aging treatment on this material, or the actual time and temperature of the heat treatment.

The inspector noted that Paragraph 3.3.10 of the DuBose QSM section on utilization of unqualified source material states that "... the lab test report shall be attached to the DuBose Certificate of Conformance."

Paragraph 3.3.11 states that the Certificate will contain a statement

" upgraded per NCA 3855.5."

Before the completion of this inspection, Walker Bolt telefaxed an amended certification to DuBose indicating that upgrading tests were performed at their facility and including a certification of the heat treatment.

o Carolina Power & Light Co. PO 8M1717 CJ, dated November 1,1996 ,

for thirty four 18,21/2 inch long ASME SA 193, Grade B7 Cap screws.

This material was certified under DuBose OSC, as complying with the applicable Code and PO requirements. Appended to the DuBose CMTR was a certificate from T&T Enterprises (T&T), a DuBose qualified supplier, containing chemical analyses, results of tensile and hardness tests, description of heat treatment, visualinspection, and a statement i

13-l l

Indicating that a macrostch test had been performed with satisfactory results. The T&T certification also contained a quality program statement that all activities had been performed in accordance with their '

DuBose approved QA program. A Republic Steel Co. CMTR for this I heat of material was attached. The Republic Steel certificallon was sent 4 to Fry Steel Co., a distributor, Indicating that T&T obtained this material as unqualified source material and upgraded it.

The inspector noted that, since the starting material was in the form of 1 inch bar, cold or hot heading operations, as well as heat treatment would have been involved in the manufacturing process, requiring the use of approved suppIlers. However, no certifications for these operations were in the data package, e Public Service Electric & Gas Co. PO P2 09029931850 0000 for four 10 gage,4 by 8 foot steel sheets.10 CFR 50, Appendix E OA requirements and 10 CFR Part 21 were invoked by the PO.

DuBose certified this material as complying with their QSP and the PO requirements. DuBose obtained the material from NUCOR Steel Co., a commercial supplier, and dedicated it under their program. The dedication consisted of performing complete chemical analysis on each of the four sheets, and tensile and bend tests on one sheet. The laboratory report, including the test results and the NUCOR Steel Co.

material certification were attached to the Dubose CMTR. The chemical analyses o'f samples from the four sheets showed close agreement to the NUCOR Steel Co. CMTR, supporting the bases for verifying mechanical properties of only one sample.

c. Conclusions Failure to include certifications from approved suppliers as identified attachments to DuBose CMTRs issued to Union Electric (PO 8M1717 CJ) and Carolina Power and Light (PO 8M1717 CJ), were identified as examples of Nonconformance 99900861/97 0101, 3.4 Entrance and Ex[t Meetinos in the entrance meeting on January 21,1997, the NRC inspectors discussed the scope of the inspection, outlined the areas to be inspected, and established interfaces with DuBose management. In the exit meeting, on January 24,1997, the inspectors discussed their findings and concerns.

14

I PARTIAL LIST OF PERSONS CONTACTED Carl M. Rogers, President & CEO John Hottel, Vice President '

James Dalley, Manager, Quality Assurance Allen Jones, Manager, Sales & Marketing Laurie Meyer Dickson, Certification engineer ITEMS OPENED, CLOSED, AND DISCUSSED Onened [

99900861/97 01 01 NON Failure to provide required documentation 99900861/97 01 02 NON Acceptance of material with incomplete documentation 99900861/97 01 03 NON Inadequate procurement Documents 99900861/97 01 04 NON Inadequate supplier audits f

15- ,

-41 r

poq q ,

,/ W. k.]

s*

UNITED STATES NUCLEAR RECULATCRY CSMMimaN

' wuwmot ow, o.c. somme t  :

\, ./ June 11, 1997 Mr. Craig P. Kipp Plant Manager General Electric Nuclear Energy P.O. Box 780 Wilmington, NC 28402-0780

SUBJECT:

NONDROPRIETARY VERSION OF NRC INSPECTION REPORT 99900003/97-01 AND NOTICE OF NLNCONFORMANCE

Dear Mr. Kipp:

This letter transmits the nonproprietary version of the U.S. Nuclear Rcgulatory Commission's (NRC's) inspection report at the General Electric (GE)

Nuclear Energy facility. Our letter to you dated May 20, 1997, transmitted the original (proprietary) version of the report. On the basis of our discussions and review of the information in your letter (RJR-97-076) of June 6, 1997, attached affidavit, and attached Proprietary Information Summary Sheet, we have concluded that the specific items identified in your letter could be regarded as proprietary and, as such, were removed from the inspection report. In tie revised nonproprietary (public) version of the report, we have deleted or revised the text.

Your response to either this letter or our letter dated May 20, 1997, and their enclosures are not subject to the clearance procedures of the Office of Management and 8udget, as required by the Paperwork Reduction Act of 1980, Pubite Law No. 96-51).

In accordance with Section 2.790(a) of the NRC ' Rules of Practice," of Title 10 of the Code of Federal Reaulations, a copy of this letter and its enclosures will be placed in the NRC Public Document Room. Should you have any questions concerning the matter, please contact Anil S. Gautam of my staff at 301/415-2988.

S cerel f Rb er . Ilo, Chief Special Inspection Branch Division of Inspection and Support Programs Office of Nuclear Reactor Regulation Docket No. 99900003

Enclosures:

NRC letter to GE, May 20, 1997

aeg ye t* UNITED STATES NUCLEAR REGULATOAY COMMISSION

• WASHINGTON D.C. 3046H001 May 20, 1997 Mr. Craig P. Kipp Plant Manager General Electric Nuclear Energy P.O. Box 780 Wilmington, NC 28402-0780 SUBJEC1: NRC INSPECTION REPORT 99900003/97-01 AND NOTICE OF NONCONf0RMANCE

Dear Mr. Kipp:

On March 14, 1997, the U.S. Nuclear Regulatory Commission (NRC) ccmpleted an inspection at the General Electric (GE) Nuclear Energy facility. The enclosed report presents the results of tha', inspection.

The inspection was conducted to assess GE's design interface with licensees that implemented the average power range monitor-rod block monitor-technical specification (referred to as ARTS) modificatien, that allowed licensees to bypass the rod block monitor (RBM) for a rod withdrawal error (RWE) event if the core had sufficient minimum critical power ratio margin. The inspectors assessed RBM operability and related technical specifications actions with respect to protecting the fuel cladding in the boiling water reactor cores of ARTS plants for an RWE event. The inspectors also assessed GE's monitoring of the effectiveness of its design control program.

During the inspection, the inspectors determined that GE dio not adequately inform ARTS licensees of the need to consider the fuel cladding plastic strain limits and the assoc.iated mechanical overpower (MOP) limits in addition to the minimum critical power ratio limits when considering RBM operability for potential RWE events. This inadequate interface between GE and ARTS licensees contributed to (1) the failure of licensees to ensure through their technical specifications that the RBM was operable to protect- fuel cladding at applicable plants, and (2) occasions during certain operational cycles of the Fermi, Hatch, Brunswick, and Duane Arnold nuclear plants when, based on GE's RWE analyses, the fuel cladding had exceeded its M0P limits and had the potential of exceeding its plastic strain limits.

The inspectors determined that GE modified certain design parameters useu in the RWE analysis without procedures or other documented basis, and did not assess the overall effectiveness of GE's design control program after Detroit Edison Company (Deco) identified deficiencies in design adequacy. Based on the above, the inspectors concluded that the implementation of GE's quality assurance program did not meet certain NRC requirements imposed on GE by its customers.

l l

t -

L C. Kipp l I

These issues are cited in the enclosed Notice of Nonconformance (NON), and the  !

circumstances surrounding them are described in detail in the enclosed report. '

You are requested to respond to the nonconformances and should follow the instructions specified in the enclosed NON when preparing your response.

In accordance with 10 CFR 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.

f Sincerely, ,

orisin.t eis t,y ,

Robert M. Gallo, Chief Special Inspection Branch Division of Inspection and Support Programs Office of Nucloar Reactor Regulation ,-

Docket No. 99900003

Enclosures:

1. Notice of Nonconformance

2. Inspection Report 99900003/97-01 ,

-44

N0i!CE OF NONC0dFORMANCE General Electric Nuclear Energy Docket No.: 99900003 Wils;ington, NC On the basis of an inspection by the staff of the U.S. Nuclear Regulatory Comission (NRC) from March 10 through 14, 1997, it appears that the following activities were not conducted in accordance with NRC requirements:

1 Criterion 111 of Appendix B to 10 CFR Part 50, ' Design Control,'

requires, in part, that measures shall be established for the identification and control of design interfaces and for coordination among participating design organizations. These measures shall include the establishment of procedures among participating design organizations for the review, approval, release, distribution, and revision of documents involving design interfaces.

Paragraph 3.3 of Section 3. " Design Control," General Electric (GE)

Nuclear Energy Quality Assurance Manual NEDO-11209, dated March 31, 1989, requires, in part, that GE design documents be furnished to the customer to provide for interface compatibility review and coordination by owner design organizations.

Contrary to the above requirements, GE did not adequately inform licensee design organizations implementing the average power range monitor-rod block monitor-technical specification (ARTS) modification of the need to consider the 1 percent fuel plastic strain limits and the associated mechanical overpower IMOP) limits in addition to the minimum critical power ratio limits when evaluating rod block monitor (RBM) ope. ability for a rod withdrawal error (RWE) event. GE's supplemental reload licensing reports for ARTS plants did not adequately address requirements for RBM operability with regard to the M0P limits. This inadequate interface between GE and ARTS licensees contributed to (1) the failure of licensees to ensure through their plant technical specifications that the RBM was operable to protect fuel cladding at applicable plants, and (2) occasions during Fermi Cycles 4, 5 and 6, Hatch Unit 1 Cycles 16 and 17, Hatch Unit 2 Cycles 13 and 14, Brunswick Cycle 10, and Duane Arnold Cycle 14 when based on GE's RWE analyses the fuel cladding had exceeded its MOP limits and had the potential of exceeding its plastic strain limits. (99900003/97-01-01) 2 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 documented instructions and procedures of a type appropriate to the circumstances, and shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished.

GE Nuclear Energy Quality Assurance Manual NEDO-ll209, Section 5,

" Instructions, Procedures, and Drawings," dated March 31, 1989, requires, in part, that documented instructions, procedures, and drawings be utilized to communicate quality requirements throughout all phases of design.

Enclosure 1

-4 5-

l Contrary to the above-requirements, GE (1) modified peaking factors when the MOP limits were exceeded in the RWE analyses, (2) applied alternate rod patterns in addition to normal rod patterns in the RWE analyses, and i (3) revised the theoretical density values used in the peak cladding l temperature (PCT) analysis to reduce the calculated PCT, without ,

documented instructions or procedures. (99900003/97-01-02) ,

3 Criterion XVI!! : Appendix B to 10 CFR Part 50, ' Audits,' requires, in  !

Part, that per'.ute audits shall be carried out to determine the {

effectiveness of the quality assurance )rogram. Followup action, '

including reaudit of deficient areas, siall be taken where indicated.

GE Nuclear Energy Quality Assurance Manual NEDO-ll209, Section 18, )

• Audits,' dated March 31, 1989, requires, in part, that the audit '

program provide for followup action, including any necessary reaudit of ,

deficient areas. -

Contrary to the above requirements, Detroit Edison Company (Deco) audited GE in 1992 and 1993 and observed several deficiencies regarding  !

design control, including GE's failure to inform DECO (and other ARTS ,

licensees) that the MOP limits would be exceeded if the RBM was not operable during an RWE event. GE took corrective actions for specific deficiencies but did not conduct followup action, including reaudit of the design control area, to determine the effectiveness of the program.

(99900003/97-01-03)

Please send a written statement or explanation to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555-000), with a copy to the Chief. Special Inspection Branch, Division of Inspection and Support Programs, Office of Nuclear Reactor Regulation, within 30 days of the date of the letter transmitting this Notice of Nonconformance. Your reply should be clearly marked as a " Reply to a Notice of Nonconformance" and should -

contain for the nonconformances (1) a description of steps that have been or will be taken to correct these items, (2) a description of stups that have been or will be taken to prevent recurrence of these items, and (3 the dates your corrective actions and preventive measures were or will be com)pleted.

Dated at Rockville, Maryland i this 20th day of May 1997 i

1

_n .- -.A,

I U.S. NUCLEAR REGULATORY COMMISSION l Off!CE OF NUCLEAR REACTOR REGULATION l P

i i

Report No: 99900003/97-01 P Organization: General Electric Nuclear Energy Wilmington, North Carolina

Contact:

Ralph J. Reda, Manager Fuels & Facility Licensing 910/675 5000 Nuclear Industry Nuclear fuel assemblies and related reload core Activity: designs, core e.omponents, safety analyses, and services Dates: March 10-14, 1997 Inspectors: Anil S. Gautam, Team Leader, NRR Laurence E. Phillips, NRR Kombiz Salehi, NRR Edward D. Kendrick, NRR Geoffrey R. Golub, NRR John F. Carew, Brookhaven National Laboratory Carl E. Beyer, Pacific Northwest National Laboratory Approved by: Gregory C. Cwalina, Chief Vendor inspection Section Special Inspection Branch Division of Inspection and Support Programs Enclosure 2 r -. _

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! !NSPECTION

SUMMARY

During this inspection, the NRC inspectors reviewed activities associated with General Electric (GE) Nuclear Energy's design interface with licensees that implemented the average power range monitor-rod block monitor-technical specifications (hereafter referred to as ARTS) modification. The ARTS 1 i

modification allowed licaisees to bypass the rod block monitor (RBM) for a rod withdrawal error (RWE) above 30 percent power (hereafter referred to as RWE i event) if the core had sufficient minimum critical power ratio (MCPR) margin.

The inspectors focused on evaluation and implementation of RBM operability to prevent exceeding fuel cladding 1 sercent plastic strain limits (hereafter referred to as strain limits) in tae boiling water reactor fuel cores of the The inspectors examined technical ARTS specifications _ plants (TSduring)an actions toRWE ensureevent.

RBM operability during an RWE event to keep fuel cladding from exceeding the strain limits and associated MOP limits.

The inspectors also ..sessed GE's monitoring of the effectiveness of its design control program.

The inspection bases were as follows:

• Appendix B " Quality Assurance Criteria for Nuclear Power Plants and fuel Reprocessing Plants," to Part 50 of Title 10 of the f.g.de of Federal Reaulations (10 CFR Part 50).

• 10 CFR Part 21, " Reporting of Defects and Noncompliance."

• GE Nuclear Energy Quality Assurance Program Description NEDO-ll209-04A, Revision 8, dated March 31, 1989, and associated implementing procedures.

During this inspection, the inspectors noted three instances in which GE failed to conform to NRC requirements imposed upon it by NRC licensees. These nonconformances are discussed in Section 3.1 and 3.3 of this report.

2 STATUS OF PREY!0US INSPECT!0N FINDINGS This inspection was limited in sc ue to address GE's design interface with licensees. Previous inspect i findings were not evaluated.

3 INSPECTION FINDINGS AND OTHER COMMENTS 3.1 Desian Interface with ARTS Licensees

a. Inspection Scope The inspectors assessed the adequacy of GE's design interface with ARTS licensees regarding RBM operability during an RWE tvent. The inspectors examined GE's RWE analyses and supplemental reload licensing reports (SRLRs) and GE's interaction with licensees regarding the assur?. ions and results of its analyses.

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b. Observations and Findinos The RWE event assumes erroneous withdrawal of the highest worth rod from full in to full out at any time during the operating cycle. To protect against this event and keep the fuel from reaching a 1 percent plastic strain during an RWE event, the RBM monitors control rod movements and initiates a rod block during a RWE in which the local power exceeds a  :

predetermined setpoint. The 11ercent plastic strain of the cladding is ,

conservatively defined as the t1reshold below which fuel damage due to fuel clad overstraining is not expected to occur.

GE uses conservative mechanical overpower (MOP) limits to assure compliance with strain limits. GE started using MOP limits in 1987 after a study showed that the fuel assemblies in the error cell in newer fuel designs could potentially exceed the strain limits during an RWE event. >

Using the MOP limits in the RWE analysis added a margin of protection against exceeding the strain limits. GE determined that if fuel assemblies did not exceed the MOP limits during an RWE event, they would not exceed the strain limits. If the RWE analysis indicated that the MOP limits would be exceeded, licensees could take credit for RBM operability, perform an exact site-specific analysis to ascertain whether the strain limits would be exceeded without RBM operability, or reconfigure the core design to avoid exceeding the MOP limits during an RWE event.

The MCPR limits were imposed to avoid fuel damage due to severe overheating of the fuel cladding. The ARTS modification allowed licensees to bypass the RBM for an RWE event if the core had sufficient MCPR margin. GE determined that for certain ARTS plants during certain fuel cycles the RBM needed to be operable to prevent exceeding the MOP limits during an RWE event, even if the MCPR margin was within the design parameters established by the ARTS modification.

The inspectors reviewed activities associated with GE's design interface with ARTS licensees regarding RBM operability during a potential RWE event at ARTS plants. The inspectors assessed GE's interaction with Detroit Edison Company (DECO) for fermi Atomic Nuclear Plant Unit 2 (Fermi); with Soutiern Nuclear Operating Company (SNOC) for Edwin 1.

Hatch Nuclear Plant, Units 1 and 2 (Hatch); with Carolina Power & Light for Brunswick Steam Electric Plant, Units 1 and 2 Company (Brunswick);(CP&L)ith w Philadelphia Electric Company for Limerick Generating Station, Units 1 and 2 (Limerick); with Nebraska Public Power District for Cooper Nuclear Station (Cooper); with IES Utilities (lES) for Duane Arnold Energy Center (Duane Arnold); and with Boston Edison Company for Pilgrim Nuclear Power Station, Unit 1 (Pilgrim). The inspectors' review of factors pertaining to RBH operability at ARTS plants is summarized >

below:

(1) FERMI Cycle 3 commenced June 1, 1991. Cycle 4 commenced November 1, 1992 and concluded with the turbine trip event in December 1993, in April 1992 Deco audited GE's activities regarding fuel design control 3

l for fermi Cycle 4 (DECO Report 92-018, dated May 6, 1992). Deco determined that based on GE's RWE analysis there was a potential for the fuel cladding exceeding its MOP limits if the RBM was not operable and control rod patterns deviated from GE's recommended normal operation cycle patterns. GE had not informed Deco that the MOP limits at Fermi could be exceeded if the RBM was bypassed, and that the fermi TS 3/4.1.4.3 (in accordance with the ARTS modification) allowed bypassing of the RBM. DECc recommended that GC analyze both Cycle 3 and 4 cores te either affirm or refute the validity of the RWE analysis with respect to MOP limits in intermediate control rod patterns, and whether the Fermi TS 3/4.1.4.3 " contained adequate operational restraints."

GE analyzed Fermi Cycles 3 and 4 and informed DECO that for Cycle 3 the MOP limits were not exceeded. GE also informed DECO that, for Cycle 4 the MOP limits for GE-Il fuel would be exceeded but that the strainlimitswouldnotbeexceededbecauseofasubstantial conservative margin between the MOP limits and the strain limits. GE concluded that for Cycle 4 the RBM did not need to be operable, and that the TS 3/4.1.4.3 RBM operational requirements were adequate.

DECO audited t>E on December 7 through 10, 1933 (DECO Report QA 1006, dated January 10, 1994), and again determined that GE had not adequately communicated design changes to Deco. GE had not informed DECO that fermi's TS basis for RBM operability required consideration of the M0P limits. The inspectors determined that the Fermi reload analysis for Cycle 4 provided in the SRLR 23A7075 Revision 0, dated April 1992, did not address RBM operational requirements for not

$xceeding the MOP limits. In November 1994, as part of its response t SFCo's audit observation, GE issued letters to inform all ARTS licensees that they should consider the strain limits and associated MOP limits when evaluating RBM operability for the ARTS modification.

For Cycle 5, GE informed DECO that the HOP limits could be exceeded.

in preparation for Cycle 5. GE's RWE analyses, dated November 11, 1993, and June 22, 1994, indicated that the RBM needed to be operable to prevent the M0P limits from being exceeded. Cycle 5 commenced on Deccmber 17, 1994. In May 1995, Deco requested GE to further evaluate the need for PBM operability. On August 22, 1995, during Cycle 5, GE completed an RWE analysis using the actual core loading and control rod patterns and concluded that the MOP limits were not exceeded and that the RBM did not need to be operable, for Cycle 6, GE informed Deco that the RBM was required to be operable. Based on the PANACEA analyses, GE informed DFCo that the HOP limits for fuel GE-Il would be exceeded within the error cell during an RWE event if the RBH was not operable. The SklR for Cycle 6, issued in November 1996, stated: "At least one RDM channel must be operable when moving rods in order to protect for mechanical overpower l limits." The inspectors considered the SRLR to be inadequate in that l it did not reference the 30% power level for RBM operability and i

addressed operability of only one RBM channel.

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During the inspection, GC informed the inssectors that the strain limits were not exceeded at Fermi during tie above operational cycles.

However, the inspectors concluded'that GE did not adeauntely inform DECO of the need to consider the strain limits and the associated MOP limits in addition to the MCPR limits when evaluating RBM operability i for an RWE event. DECO actions regarding implementation of I appropriate TS revisions for Cycles 5 and 6 are addressed in section  ;

3.2 b of this report.

(2) HATCH i For Hatch Unit 1 Cycles 16 and 17, and Unit 2 Cycles 13 and 14, GE  ;

infomed SNOC that the MOP limits would be exceeded during an RWE f event if the RBM was not operable. A GE memorandum tc SNOC, dated ,

October 31, 1994, stated that the 'l percent plastic strain is met, if one channel of the RBM remains operable." In addition, on October 31, 1904, GE informed SNOC by letter that they should consider the strain limits and associated MOP limits when evaluating RBM operability for  :

the ARTS modification. GE's SRLR for Hatch 1 Cycle 16 and 17, and .

Hatch

• Cycles 13 and 14 informed Hatch of necessary protective .

i measures but failed to clearly address the conditions under which the l RBM should be operable. During the inspection, GE informed the .

inspectors that based on its analyses the MOP limits had been exceeded l for Hatch Unit 1 Cycles 16 and 17, Unit 2 Cycle 13, and Cycle 14. The strain limits were not exceeded during these cycles. SNOC's actions regarding implementation of appropriate TS revisions are addressed in section 3.2 ) of this report.

(3) PILGRIM, LIMERICK, COOPER Based on the SRLRs for these plants, the RBM did not need to be operable for Pilgrim Cycle 12, Limerick Unit 1 Cycles 6 and 7 Limerick Unit 2 Cycle 4 and 5, and Cooper Cycle 18. The inspectors determined that the MOP limits were not exceeded during these cycles, in November 1994, GE issued letters to these licensees indicating that they should consider the strain limits and associated MOP limits when evaluating RBM operability for the ARTS modification.  ;

(4) BRUNSWICK AND DUANE ARNOLD in January 1995, GE conducted an RWE analyses for the Brunswick Cycle 10 and Duane Arnold Cycle 14 and determined that the MOP limits would be exceeded if the licensees bypassed the RBMs. Based on the ,

inspectors review of GE's desigt, documents, GE did not inform these  ;

licensees that the MOP limits would be exceeded. Based on GE's RWE analysis, the MOP limits were exceeded at Brunswick. On April 1, 1997, CP&L issued a 50.72 notification (Event No. 32058) regarding RBM cperability.

In March 1997 (before the NRC inspection), GE reviewed reload design  :'

documents to dLtermine which ARTS plant cycles resulted in exceeding the strain limits. Based on its review of prior RWE analyses, GE i

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notified CP&L and IES by letter dated March 11, 1997, that the  ;

trunswick Cycle-10 and Duane Arnold Cycle 14 GE-10 fuel had exceeded the MOP limits. GE did not inform the inspectors during the inspection that the MOF limits were ex;eeded for these two plants.  ;

Subsequent to the inspection, GE performed a cycle-specific analysis ,

for both plants. In a letter dated April 25, 1997, GE informed the .

NRC that.the strain limits were not exceeded at Brunswick and Duane Arnold. On May 9, 1997, the inspectors contacted GE staff by telephone to ask why licensees were not informed in January 1995, and  ;

why-the inspectors were not informed during the March 1997 inspection i that based on GE'r, RWE analyses, Brunswick and Duane Arnold had .

exceeded the MOP limits. GE statec that it did not inform the  !

licensees or the NRC because GE's " generic" RWE analysh had indicated that the strain limits would not be exceeded. CP&L and IES's actions  :

regarding implementation of appropriate TS revisions are addressed in section 3.2 b of this report.

Based on the above review, the inspectors determined that GE did not adequately inform licensee design organizations implementing the ARTS modification of the need to consider the strain limits and the associated MOP limits in addition to tiie MCPR limits when evaluating RBM operability for an RWE event. GE's SRLRs for ARTS plants did not adequately address reautrements for RBM operability with regard to the MOP limits. This i inadequate interface between GE and ARTS licensees centributed to (1) the failure of licensees to ensure through their TS that the RBM was operable to protect fuel cladding at applicable plants, and (2) occasions during '

Fermi Cycles 4 and 5, Hatch Unit 1 Cycles 16 and 17, Hatch Unit 2 Cycles  !

13 and 14, Brunswick Cycle 10, and Duane Arnold Cycle 14 when based on GE's RWE analyses the fuel cladding had exceeded it MOP limits and had the potential of exceeding its strain limits. This constitutes Nonconformance 99900003/97-01-01.

In its March 19, 1997, letter to the ukC, GE stated that its '

' communications to licensees in the SRLR have not been adequate.' GE also stated that, in > art, it.(1) had sent letters to all customers '

apprising them that tie RBM is required to ensure that the MOP lisaits were not exceeded (2) had revised technical design procedure (TOP) 0035 to clarify required utility communications, and (3) planned to review the SRLR format and revise the GESTAR 11 standard format, if necessary, and make further revisions to TDP-0035 if necessary to be consistent with the final SRLR format. - -

c. Conclusions i The strain limits were not exceedad at the ARTS plants, GE's design-interface with licensees-was weak, as noteo in %e nonconformance herein. ,

GE did not adequately inform ARTS licensees of sne conditions for RBM

. operability, contributing to the ARTS licensees' failure to implement RBM protection-for fuel during potential RWE events. SRLRs did not clearly identify conditions for RBM operability. Based on GE's RWE analyses, the 6

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fuel cladding had exceeded its MOP limits and had the potential of exceeding its strain limits at Fermi, Hatch, Brunswick, and Duane Arnold.

3.2 TECHNICAL SPECIFICATIONS (TS) ACTIONS

a. Insoection Scoce The inspectors examined TS actions taken by GE and ARTS licensees to ensure RBM operability during an RWE event to protect exceeding the strain limits and associated MOP limits.

b. Observations and Findinos DECO audited GE in 1992 and 1993 and observed GE's failure to inform DECO (and other ARTS licensees) that the MOP limits would be exceeded if the In November 1994, GE informed RBM was not operable during an RWE event.

all ARTS licensees of the need to consider RBM operability to not exceed strain limits.

During Cycle 5, GE informed Deco that the RBM must be operable. In Section 10 of the Cycle 5 SRLR, dated December 1993, GE stated that "at least ont channel of the RBM must be operable when moving rods in order to meet the MOP limits for an RWE event." GE's memorandum to Deco, dated June 22, 1994, regarding the Cycle 5 analysis stated: "GE still recommends that at least one RBM channel be operable when moving rods."

However, DECO did not revise the Fermi TS; rather, in May 1995 Deco requested GE to further evaluate the need for RBM operability. On August 22, 1995, GE concluded that the MOP limits were not exceeded during Cycle 5.

In preparation for Cycle 6, DECO addressed RBM operability by placing a st tement in the core opesating limits report (COLR). The NRC resident inspectors became aware of DECv's use of the COLR in this manner and on November 22, 1996, the NRr inform.cd DLCo by telephone that the addition of this statement to the COLR without a corresponding request to amend the TS would not comply with the requirements of 10 CFR 50.36. The NRC permitted Deco in start up using administrative controls after the licens9e submitted the TS amendmest request on December 2, 1996, to requ're RBM operability ebove 30 percent power. The TS amendment (No.

110 to facility operating license No. NPF-43) was issued on May 15, 1997.

For Hatch Unit 1 Cycles 16 and 17, and Unit 2 Cycles 13 and 14, GE inforined SNOC that the MOP limits would be exceeded during an RWE event if the RBH was not operable. A GE memorandum to SNOC dated October 31, 1994 stated that the "I percent plastic strain is met, if one channel of the RL4K remains operaole." SNOC did not revise the Hatch TS.

I In January 1995, GE's RWE analysis indicated that the Brunswick Cycle 10 and Dua% Arnold Cycle 14 GE-10 fuel would exceed the HOP limits. Based on doc' nonts provided to the inspectors, GE did not inform licensees of the potential for exceeding the MOP limits, in r: arch 1997, GE informed l 7 63

I CP&L and IES that based on its RWE analyses the fuel at Brunswick and Duane Arnold had exceeded its MOP limits. CP&L and IES did not revise the Brunswick and Duane Arnold TS.

The inspottors asked GE why they did not explicitly inform ARTS licensees I to revise their TS to ensure RBM operability for an RWE event. On May 1, 1997, during a telephone conversation with the inspector, GE agreed to inforn licensees of the need to examine or consider revisions to their TS, as applicable, with respect to RBM operability. l

c. Conclusions feral and Hatch apparently knew that the RBM must be operable for a potential RWE but did not revise their TS to require RBM operability for an RWE event. Other ARTS licensees, including Brunswick and Duane Arnold, were informed by GE in November 1994 of the need to consider RBM operability to not exceed strain limits but did not revise their TS.

GE's failure to explicitly inform licensees to revise their TS to ensure RBM operability during an RWE is considered a weakness.

3.3 DESIGN CONTROL

a. Insoection Scope The inspectors reviewed the adequacy of GE's RWE analyses and GE's assessment of its design control program,

b. Observations and findinos The inspectors determined that GE modified peaking factors when the MOP limits were exceeded in the RWE analyses, and applied alternate rod patterns in addition to normal rod patterns in the RWE analyses without documented instructions or procedures. The inspectors also determined that GE changed design parameters for its peak cladding temperature (PCT) analysis by revising the theoretical fuel pellet density values without procedures for tracking the removal of conservatisms in the FCT analysis.

GE's use of the incorrect fuel pellet densification value had resulteo in a lower >eak cladding temperature. When GE corrected the densification value, t1e Loss of Coolant Accident analysis of the BWR-6 plant indicated an unacceptable increase of the PCT. CE revised the theoretical density values to reduce the PCT without issuing any documented instructions or procedures for using the revised density values. This constitutes Nonconformance 99900003/97-01-02.

Subsequent to the inspection, in its March 19, 1997, letter _to the NRC, GE stated that, as part of its corrective actions, it (1) revised TDP 0035 te improve its communication with licensees on the need for RBM operability, (2) implemented an explicit requirement for MOPS calculations for all fuel types, (3) plans to review TDPs for other areas of improvement, e.g., added guidance on local peaking factor adjustments, and (4) plans to document conservatisms used in its analyses.

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Deco audited GE in 1992 and 1993 and observed several deficiencies regarding design control, including GE's f ailure to inform Deco (and other ARTS licensees) that the MOP limits would be exce-ded if the RBM was not operable during an RWE event. The inspectors determined that GE took corrective actions for sp(:ific deficiencies but did not conduct followup action, including reaudit of the design control area, to determine the effectiveness of the program. This constitutes Nonconformance 99900003/97-01-03. During the inspection, GE stated that it planned to perform an internal audit to address the effectiveness of its design program, in its March 19, 1997 letter to the NRC, GE stated '.

that its corrective actions included (1) strengthening its requirements for annual adequacy reviews of TOPS; planning to perform adequacy reviews during the second quarter of 1997, and (2) perisrming a QA followup audit of a revised adequacy review process rheduled for late 1997.

The inspectors observed that GE's generic analysis of HOP limits for RWE events at ARTS plants may not be valid for cores containing GE 8x8 fuel or GE-10. -11. -12, and -13 fuel not included in the 1987 assestment of generic HOP limits. GE informed the inspectors that in March 1997 it began conducting cycle-specific RWE analyses for all plants and all fuel types. No immediate safety concerns were identified.

The inspectors reviewed GE's PANACEA analysis and observed that GE evaluated U0, rods rather than Gadolinium (Gd) rods in an error cell for an RWE event. The generic GESTR-H strain analysis indicated that the Gd rods were more limiting than the U0, rods. Consideration of the UC rodt but n9t the Gd rods in the error cell for the PANACEA analyses may ,not identify the limiting rod for the strain limits. The inspectors also observed that the PANACEA analyses considered planned rod patterns for certain cycles but did not include possible alternate rod patterns. Use of only planned rod patterns for certain analyses may not identify the most limiting operation within the error cell for the analyzed cycle because alternate rod patterns may be used. The inspectors determined that exclusion of Gd rods and alternate rod patterns was a weakness in GE's PANACEA analyses.

The inspectors assessed the initial conditions assumed in cycle-specific accident analyses. One example reviewed involved a Fermi turbine trip event with nominal reactor steam dome pressure assumed for the start of the transient. The inspectors exenined whether the safety analysis was bounding and the MCPR limit was sufficient if a transient would be initiated from a higher pressure within the bounds of the TS operating limits. In response to questions by the inspectors, GE performed a sensitivity study to evaluate the effect on the thermal margin of pressure variations of plus or minus 20 psi, including cases with GE-9 and GE.13 fuel. The results indicated a pressure sensitivity of a small delta critical power ratio (CPR)/ initial critical power ratio (ICPR) over this pressure range. Higher initial pressure resulted in a lower delta CPR/ICPR. In reviewing safety analysis methods, the NRC examines assumptions vogarding the initial conditions of the analyses and whether operating limits derived from the analyses provide bounding protection for all operating conditions permitted by the plant TS. Since the l

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thermal margin is dependent on several operating parameters that do not vary independent of pressure, thermal margin sensitivity to pressure reduction is not necessarily indicative of the thermal margin sensitivity to overall off-nominal plant operating conditions for large pressure >

increments. No concerns were identified.

The inspectors observed that in late february 1997, Tenaessee Valley  ;

Authority determined that the Browns Ferry Nuclear Plant Emergency Operating Procedures (EOPs) revision for GE 9x9 fuel had not considered all the input parameters affected by the new fuel design. The input parameters are specific to fuel type and would be used by the licensee to calculate plant specific curves or limits to maintain Emergency Procedure Guidelines (EPGs) specified action levels. Input parameters for E0P calculations are contained in Appendix C of NEDO-31331 Revision 4 .

issued March 1987. At the time of issuance of revision 4. GE 9x9 fuei  !

was not in use by any licensee. Upon incorporation of 9x9 fuel, licensees .pparently did not recognaze that certain parameters identified

• as " generic data" in EPG Appendix C were affected by fuel type. The parameters were four steam-cooling-related input parameters and two -

shutdown boron weight-related input parameters, in response to the event, GE issued Services Information Letter (SIL) 529, Supplement 1, on March 14, 1997, informing BWR customers of the fuel type-specific nature of the input parrneters and recommending that potentially affected plants reevaluate their EPG calculations if the fuel design has changed since 4 the Appendix C calculations. The Sil also provided a list of calculations which may be impacted by revised input parameters.

c. Conclusions GE did not have documented instructions or procedures to control modifications in RWE analyses when MOP limits were exceeded. GE's generic MOP limits evaluations nec:! to be replaced with cycle-specific analyses. The PANACEA fuel strain analysis was weak.

3.4 Entrance and Exit Meetinos in the entrance meeting on March 10, 1997, the NRC inspector; discussed the scope of the inspection, outlined the areas to be inspected, and established interactions with GE management, in the exit meeting on March 14, 1997, the inspectors discussed their findings and observations.

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4 PARTIAL !!ST OF PERSONNEL CONTACTED General Electric Nuclear Enerov ,

Chris Monetta, Manager, Nuclear Quality Assurance Glen Watford, Manager, Nuclear Fuel Engineering James Klapproth, Manager, Product Defirittion i Steve Congdon, Manager, Nuclear Technology Chuck Papandrea, Mar ager, SC Quality Jerry Potts, Manager, fuel Performance Kevin Theriault, Manager, Fuel QA Anne Sullivan, Manager, Fuel Engineer, QA Faul Sick, Manager, Nuclear Quality Assurance Elwood Mobley, Manager, Quality Assurance Rendy Hannelt, Manager, Nuclear Fuel Projects Ralph Reda, Manager, fuel & Facility Licensing Caroline Smith, Sr. Engineer, Fuel & Facility Licensing John Embley, Jim Rash, Sr. EngProgineer, Fuel & Facility Licensingam Manager, fuel & Facility Li Robert R:nd, Principal Engineer, Nuclear Fuel Fran Bolger, Principal Engineer, Nuclear fuel <

Barry R. Fischer Technical Program Engineer, huclear Fuel ITEMS OPENED, CLOSED, AND DISCUSSED Ooened 99900003/97-01-01 Para 3.1 b NON inadequate design interface 99900003/97-01-02 Para 3.3 b NON inadequate procedures 99900003/97-01-03 Para 3.3 b NON i_nadequate assessment of design effectiveness ACRONYMS USED

-ARTS Average Power Range Monitor-Rod Block Monitor-Technical Specificatiens GE General Electric RBM Rod Block Monitor RWE Rod Withdrawal Error RWE Event Rod Withdrawal Error above 30 percent power MCPR Mirimum Critical Power Ratio TS Technical Specifications SRLR Sup)1emental Reload Licensing Report MOP Mecianical Over Power

-1DP Technical Design Procedure COLR Core Operating Limits Report PCT Peak Cladding Temperature Gd Gadolinium U0 Uranium Dioxide CPk Critical Power Ratio ICPR Initial Critical Power Ratio EP0s Emergency Operating Procedures EPGs Emergency Procedure Guidelines SIL Services Information Letter 11 47-

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• k UNITED STATES

• NUCLEAR REGULATORY COMMISSION g wAsumotow, 0.c. asse+ soot

%,**e** April 22, .997 Mr. John M. Leslie, President SynTech Products Corporation 520 East Woodruff Avenue Toledo, Ohio 43624

SUBJECT:

NRC INSPECTION REPORT 99901313/97-01

Dear Mr. Leslie:

On February 4-6, 1997, the U.S. Nuclear Regulatory Commission (NRC) conducted an inspection at your Woodruff Avenue facility in Toledo, Ohio. The enclosed report presents the results of that inspection.

The NRC inspection team evaluated the commercial quality program that SynTech established and implemented to control the manufacture and supply of vartous cleaning and maintenance chemical products. Although SynTech supplies several of its products to NRC-licensed nuclear power plants, the inspectors determined that those products are not basic components as defined in Section 21.3 of Part 21 of Title 10 of the Code of Federal Reaulations (10 CFR Part

21) because, for facilities licensed under 10 CFR Part 50, the products are not designed or manufactured under a quality assurance program meeting the provisions of 10 CFR Part 50, Appendix B, nor are they treated as basic components for procuremeat purposes by the NRC licensees to whom you supply them.

The inspection consisted of an examination of selected procedures, purchase orders, test reports, certificates of compliance and other associated records, interviews with personnel and observations by the inspectors. No violation of NRC requirements was identified. Also, during this inspection, the inspectors reviewed the circumstances surrounding SynTech's submission to Consumers Power Company (Palisades) of what appeared to be a Detroit Edison Company chemical test report, but which reported data of indeterminate origin and cited a sample lot for which Detroit Edison or its subsidiary, Utility Technical Services, had no record of testing. Prior to the inspection, the information available to the inspectors suggested a fraudulent SynTech report. However, your explanation of the circumstances surrounding the creation of the report i

in its unusual form was plausible, if not conclusive, and was supported by other information obtained by the inspectors. Although you could not offer a satisfactory explanation for the report's not being effectively screened through management review before issue, the incident appeared to have been isolated. The inspectors could not conclusively determine that there was any deliberate intent to mislead the recipient of the report and found no evidence of a pattern of similar activity.

58-i __ _ . ,

I i

J. Leslie  !

In accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of  :

this letter and its enclosures will be p1* ' in the NRC's Public Document i Room (PDR).

Sincerely, 1.

Robert M. Gallo, Chief Special Inspection Branch '

Division of Inspection and Support Programs Office of Nuclear Reactor Regulation Docket No. 99901313

Enclosures:

1. Inspection Report 99901313/97-01 l r

i t

U.S. NUCLEAR REGULATORY COMMISSION-  ;

OFFICE OF NUCLEAR REACTOR REGULATION k

Report No.: 99901313/97-01 ,

Organization: SynTech Products Corporation 520 East Woodruff Avenue 1' Toledo, Ohio 43624

Contact:

D. Moore, Quality Assurance Manager (419) ?41-1215 ,

l Nuclear Industry Manufactures and supplies industrial cleaners, Activity: solvents, and other maintenance products such as penetrating oils, cutting fluid, gasket / paint strippers, etc., in bulk quantities or in aerosol cans.  :

Dates: February 4-6, 1997 Inspectors: Stephen D. Alexander, Reactor Engineer (Team Leader) i Kamalakar R. Naidu, Senior Reactor Engineer Approved by: Gregory C. Cwalina, Chief Vendor Inspection Section Special Inspection Branch Division of Inspection and Support Programs -

Office of Nuclear Reactor Regulation j I

Enclosure 1 .

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1 INSPECTION SUMARY During this inspection of SynTech Products Corporetion, the NRC inspect >-

reviewed selected procyrement documents, test certificates and other documents, examined products and storage facilities, and interviewed Synicch staff regarding the supply of industrial cleaning and maintenance products to NRC-licensed nuclear power plants. The inspection bases were Part 21 of Title 10 of the Code of Federal Reaulations (10 CFR Part 21) and Appendix B to 10 CFR Part 50.

Based on a SynTech purchase order (PO) to a licensee-operated testing laboratory for analysis of a SynTer.h product that invoked 10 CFR Part 50, Appendix B, and 10 CFR Part 21, Sy 'ech n.ay have supplied some of its products as basic components as defined in -, CFR 21.3. However, the inspectors subsequently determined that SynTech products are not designed (developed) or manufactured under a 10 CFR Part 50, Appendix B, quality assurance (QA) program and synTech has not supplied, nor does it currently supply its products as basic components. Although many of SynTech's nuclear plant customers required certificates of conformance (CoC) to their PO requirements and a certified chemical contaminant analysis test report traceable to the batches or lots of the products supplied, the inspectors found no instances in which NRC licensees imposed nuclear-unique design or QA requirements on SynTech or invoked 10 CFR Part 21 in their procurement dm:9ments.

On the basis of a review of SynToch's commercial quality controls and review of independent analyses of SynTech products, the inspectors concluded that the contaminant content of SynTech Froducts is variable enough to require batch-or lot-traceable certified chemical contaminant analysis reports. Such reports would confirm that levels of undesirable substances such as halides, sulfur and heavy metals are within acceptable limits befnre using SynTech products on safety-related equipment or allowing them to come in contact with safety-related systems.

The inspectors also reviewed the circumstances surrounding SynTech's 1995 submission of what appeared to be a Detroit Edison (DE) Company chemical analysis report, but which reported data of indeterminate origin and cited a sample lot for which DE or its subsidiary, Utility Technical Services (UTS),

had no record of testing. Prior to the inspection, the information available  ;

to the inspectors suggested a fraudulent SynTech repott. However, SynTech's explanation of the circumstances surrounding the crea; ion of the report in its unusual form was plausible, if not con:lusive, and was supported by other information obtained by the inspectors. Although SynTech could not offer a satisfactory explanation for the report's not being effectively screened through management review before issue, the incident appeared to have been isolated. The inspectors could not conclusively determine that there was any deliberate intent to mislead the recipient of the report and found no evidence of a pattern of similar activity.

2 STATUS OF PREVIOUS INSPECTION FINDINGS This was the first NRC Inspection of SynTech.

2 3 INSFECTION FINDINGS AND WHER COMMENTS 3.1 Commercial Quality Controls / Chemical Contaminant Control

a. Scone In order to evaluate SynTech commercial quality controls, particularly regarding control of chemical contaminants in SynTech products sold to NRC-

-licensed facilities, Go inspectors toured the SynTech facility, interviewed persunnel, examined rew materials and finished products in various forms end in various stages of prepe ation. Also examined were products awaiting shipment either to catowrs or :qbcontractor manufacturing or packaging l facilities. The inspem ors also reviewed numerous chemical analysis reports  !

on the contaminant con'.ent of SynTech products as well as correspondence with the materials engineer at General Electric Nuclear Energy (GE NE) regarding the eligibility of Sy1 Tech products for GE NE's " nuclear grade" designation,

b. Findinas and Observations SynTech manufactures several industrial cleaning and maintenance chemical products including the following which are used in nuclear power plants.

SynTech supplies these products in aerosol cans, in bulk containers, or bot 5:

Easy Tap: Cutting and tapping fluid (aerosol and squeeze bottles)

Parts Kleen: Solvent and degreaser (aerosol and bulk containers)

N-Grade: Gasket and paint stripper (aerosol)

NOC: (Non-organic Cleaner) All purpose cleaner Rust Buster !!: Penetrating oil Titanic C: All purpose floor cleaner (bulk)

Touch It Up: All purpose cleaner (aerosol and bulk)

SynTech manufactures (blends) a batch of a product supplied in bulk from raw materials or receives finished product in bulk form from one of its suppliers or contract manufacturers. It then sends a sample of that numbered batch to a subcontractor laboratory, currently Philadelphia Electric Con:pany's (PECO Energy's) Wayne, Pennsylvania, laboratory, for chemical analysis to determine if the levels of total halogens / halides, sulfur, and other undesirable elements, mostly heavy metals (e.g., As, Bi, Sn, Pb, Hg, etc.), are within specified acceptance levels. For products in aerosol cans, the finished products are sent in bulk form to a SynTech subcontractor for packaging in the aerosol cans. Samples from numbered lots of aerosol cans (which are supposed to be filled from one batch of bulk product) are also sent to PECO's laboratory for analysis. SynTech has samples of aerosol can lots analyzed after repackaging because SynTech has found that the packaging process or cans themselves may introduce some of the contaminants of concern.

Batches or lots of products routinely sold to nuclear plants as well as other customers that are awaiting results of chemical analysis are stored in a locked cage designated the " Commercial Quality" area with QA hold tags or 3

labels affixed. If the results of the chemical analysis indicate that the levels of the contaminants of- concern are within acceptable limits,- the batches of. chemicals are released for shipment to nuclear power plants. If the levels are above the acceptance limits, the products are removed from the

'Commerctal Quality Area" and stored in general storage for sale to non-nuclear facilities.

As discussed-in correspondence with and from GE NE's materials engineer, SynTech established the acceptance limits to maintain a " nuclear grade" rating in accordance with GE I:E Haterial Specification D50VP12 and Appendix A of NEDE-31295P, GE's 'BA: Operator's Manual for Materials and Processes," June, 1986. However, individual licensees sometimes specify their own limits, either explicitly or by reference to a material standard or specification (e.g., those of the American Society for Testing and Materials (ASTM)). For nuclear customers, SynTech stated that it checks that the batch / lot analysis report indicates the customer!s specific limits are met before shipping the product unless specifica11y' requested otherwise by the customer. Review of numerous lot analysis re) orts on file at SynTech indicated that the amounts of contaminants or undesiraale substances such as halides, sulfur, and heavy metals-(e.g., mercury, bismuth, arsenic, lead, tin, etc.) are somewhat variable, sometime:,-being marginal or even exceeding GE NE nuclear grade. or customer-specific limits.

- SynTech brought two such instances to the inspectors' attention. In one case, SynTech had discovered that tin was leaching out of aerosol cans into the product.- This-was corrected by SynTech requiring their aerosol can packaging subcontractor henceforth to use cans with a non-metallic coating inside.

Subsequent analysis reports indicated that this measure corrected the problem and did not introduce additional prot,lems such as adding halides. in the other instance, the NRC licensee, Philadelphia Electric Company (Pero) requested that SynTech ship a batch of product for which the analysis results had not yet been obtained because it was PEco's laboratory that was currently i- performing the contaminant analyses for SynTech. As Murphy's Law would have it, sure enough, this particular batch of the product contained excessive sulfur and was of course captured and sent back to SynTech. Shipping records reviewed by the inspectors confirmed this and also indicated that, as stated by SynTech, it routinely provides copies of the batch / lot chemical analysis test reports to nuclear customers whether or not they are required.

In the Commercial' Quality Area, the inspectors noted information on the tags affixed to 55-gallon drums containing Titanic C, NOC, and Parts Kleen, and a tag affixed to a box containing Part Kleen aerosol cans, and reviewed the documents associated with their manufacture. The inspectors observed that the access to this storage area was controlled with a lock and key.

One of the products, Titanic C floor cleaner, was manufactured on February 28, 1996, and stored in a 55-gallon drum. A tag attached to this drum identified the product as belonging'to Lot 095, and Batch 1089. The batch manufacturing sheet for Lot 095, dated February 28,- 1996, indicated the various ingredients that were used to manuf acture this lot of Titanic C. The SynTech QA manager stated that all this material was stored in an area designated for comercial quality products. . The inspectors observed some of the ingredients used in the 4

_._m_-__

manufacture of Titanic C stored in their original bags in the Commercial Quality Area. SynTech issued P0 1520) on February 4, 1997, to PECO Energy, its current independent testing laboratory, to analyze a sample of Lot 095 for contaminant content including halogens, sulfur and heavy metals.

The tag attached to a 55-gallon drum of NOC (non-organic cleaner) was marked Lot NOC 01, Batch 10994. The tag indicated that a sample of the contents of the drum had been tested. SynTech received certification on October 3, 1996, that the levels of halogens, sulfur and heavy metals were within the acceptable limits.

A 55-gallon drum of Lot 0408942 of Parts Kleen, blended for SynTech by Chem Central in Toledo had a hold tag attached to it indicating that it was awaiting the results of chemical analysis of a sample of it that had been sent to one of SynTech's contract laboratories for chemical analysis. The laboratory report, dated August 23, 1994, indicated the content of halogens, sulfur and heavy metals were within the usual prescribed limits, but the hold tag we.s left on so that the report would be checked against specific customer requirements prior to shipment.

A hold tag was attached to a pallet of 90 cases of Parts Kleen aerosol cans, Lot 6m30, in the commercial quality area packaged for SynTech by American Jetway of Detroit, Michigan. The hold tag indicated that SynTech had sent one can of Lot No. 6m30 to the PECO Energy laboratory for chemical analysis.

The inspectors reviewed the documents on a shipment of I dozen aerosol cans of Touch It Up for Niagara Mohawk Power Company's Nine Mile Point site. The documents included a SynTech CoC that certified that the contents were from tot #4K31 which conformed to Specification 15586/94-004215, and the halides, sulphur and haavy metals were within acceptable limits. SynTech's certification was based on a certified laboratory report providing the chemical contaminant analysis results.

c. Cnnelusions SynTech has implemented commercial quality controls to provide reasonable assurance that its products shipped to NRC-licensed facilities typically contain less than the maximum acceptable levels of chemical contaminants of concern. SynTech had established measures to routinely control the release of products to nuclear plant customers pending receipt of certified chemical contaminant analysis reports from SynTech's contract laboratory. According to SynTech, this had become their standard practice, regardless of whether such reports were required by a particular customer. No further concerns were identified in this area.

3.2 Review of Procurement Documents,

n. Econg The inspectors reviewed SynTech customer procurement document files to determine if NRC-licensees had impnsed ti"; requirements of 10 CFR 50, Appendix B, or 10 CFR Part 21 on SynTech, i.e., if licensees had treated 5

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SynTech products as basic components. In addition, the inspectors examined selected files of SynTech procurement documents to determine if SynTech had passed these requirements to its suppliers of materials and services.

b. Observation and Findinas Prior to the inspection, the inspectors had obtained a copy of SynTecn P0 No.

10022, dated January 28, 1993, and a letter accompanying a sample of Lot A050-303 of " Touch-It-Up" to a subsidiary of Detroit Edison (DE), Utility Technical Services (UTS), which stated that " Appendix B and Part 21 apply,10 CFR 50"

[ sic). This suggested that SynTech may have supplied some of its products to nuclear power plants as basic components as defined in 10 CFR 21.3. According to SynTech and documents on file at SynTech, when this PO was issued, SynTech had been contracting DE's Technical and Engineering Services laboratory for several years to perform the contaminant analyses on its products being shipped to nuclear plants. However, the DE lab reports, notably in the form of letters on DE letterhead, had covered multiple products and multiple samples. The inspectors found three such reports in SynTech's files, dated March 26, 1990, December 6, 1990, and January 15, 1992, in addition, there had historically been significant time delays in obtaining results from DE.

SynTech stated that at about the same time that DE's Technical and Engineering Services laboratory was put under the new DE subsidiary organization, UTS, SynTech was advised by the QA manager at UTS that if SynTech invoked 10 CFR Part 50, Appendix B, and Part 21 in the P0, then UTS would be able to perform the analyses expeditiously in addition to the added benefits of a certified analysis report for one lot of one product at a time. Although in translating these unfamiliar requirements, SynTech cited them incorrectly, the inspectors observed that subsequently, UTS test reports were for one lot of one product each and contained the statement "the work was performed in accordance with the Technical and Engineering Services quality assurance program." The inspectors identified no other instances in which SynTech had invoked 10 CFR Part 21 or imposed 10 CFR Part 50, Appendix B, on any of its suppliers or subcontractors.

The inspectors reviewed P0s issued to SynTech by several NRC licensees including Boston Edison Company, Centerior Services Company, Niagara Mohawk Company, Detroit Edison Company, Consumers Power Company, and Commonwealth Edison Company between 1992 and 1995 for the supply of variros quantities of different SynTech products. Although many of SynTech's nuclear plant customers required CoCs to their P0 requirements or a certified chemical contaminant analysis test report, traceable to the batches or lots of the products supplied, the inspectors found no instances in which NRC-licensees had imposed nuclear-unique QA requirements on SynTech or had invoked 10 CFR Part 21 in their procurement documents to SynTech.

In addition, the inspectors determined through information obtained from some licensees that they typically procure SynTech products as commercial grade material, although some handle them under their augmented quality programs in order to maintain certain controls, such as requiring analyses for contaminant exclusion, when they are used in maintenance of safety-related systems and equipment.

6

Finally, during discussions with SynTech personnel and review of procedures and records, the inspectors also determined that SynTech had established a comercial quality program, but did not have a 10 CFR Part 50, Appendix B, quality assurance program. SynTech was not familiar with the requirements of to CFR Part 50, Appendix B, and 10 CFR Part 21.

c. Conclusion The inspectors concluded that SynTech's commercial quality program does not, nor was it intended to meet Appendix B QA requirements. Therefore, the inspectors concluded that SynTech products have not and are not being designed or manufactured under a 10 CFR Part 50, Appendix B, quality assurante program.

Further, SynTech products have no nuclear-unique design requirements, they are used in non-nuclear appliiations and they may be purchased on the basis of SynTech's published product descriptions. Therefore, SynTech products meet the definitions of commercial grade items in 10 CFR 21.3 for both 10 CFR Part 50 licensees and non-Part 50 licensees, and are not procured or supplied as '

basic components.

-3.3 Ouestionable SynTech Analysis Report and CoC

a. ScoDe In preparation for this inspection, the inspector reviewed documentation provided by UTS and Consumers Power Company (CPCo) at Palisades relevant to the circumstances surrounding SynTech's submission to Consumers Power Company (CPCo), Palisades' licensee, of what appeared to be a DE chemical analysis report, dated May 15, 1995, that was not issued by DE. Although SynTech did not have a copy of this report in its files at the time of the NRC inspection, the inspectors had reviewed a copy along with other SynTech-related documen-tation obtained prior to this inspection. During this inspection, SynTech brought this issue to the attention of the inspectors before being asked about it specifically. The inspectors reviewed relevant documentation on file at SynTech and interviewed SynTech personnel on this issue.

b. Observat!ans and Findinal b.1 Report Anomalies The report exhibited several anomalies, most notably that, although it appeared to have been printed on DE letterhead or stationery, unlike the genuine DE reports that SynTech had on file, this report contained results of analysis of only one lot of only one product, specifically Lot 761-055 of SynTech N-Grade Gasket Remover and Paint Stripper. The report was in the form of a letter that was typed on what initially appeared to be DE letterhead, and contained a signature block indicating the name of the analyst who had signed the previous DE reports, but his signature itself was not present on the report in question. In addition, the report pavided data for a sample lot for which UTS stated it had no record of testing.

7 b.2 The Palisades Response Upon receipt of the product, Palisades conducted its own analyses on a sample from Lot 761-055, as is their routine practice, but obtained unacceptable results, most notably excessive chlorides. This prompted Palisades Procurement Engineering to obtained SynTech's CoC and the attached analysis report for comparison. Although the report was on what appeared to be DE letterhead, the procurement engineer involved stated that she did not notice that the report was not in the usual format because normally only receiving sees them and they didn't identify that as an anomaly. Nevertheless, due to the significently differing results, Palisades procurement engineering called the named analyst at DE/UTS in order to resolve the differences. She also telefaxed to the analyst a copy of the report.

The Palisades procurement engineer further stated that she later contacted the QA Manager at Syniech who reportedly told her that the report was a mistake by a new person. The SynTech QA manager also reportedly requested that Palisades send back all the product and the original Coc. Palisades informed the inspectors that they did this except for the one lot 761-055 can from which the CPCo sample was taken.

With respect to 10 CFR Part 21 evaluation, the Palisades procurement engineer explained to the NRC inspector that UTS had told her that they would make any reports required to the NRC; therefore CPCo did not evaluate the deviation under Part 21 procedures (a) because of this assurance from DE/UTS, (b) because the products were not considered basic components, and (c) because CPCo had never used this batch in Palisades. Previous batches used, she stated, would have been verified satisfactory or rejected.

)

b.3 The UTS Reply According to an August 24, 1995, handwritten mene to Palisades from the DE/UTS ,

analyst named in the report, written in response to Palisades' call and fax, i DE did not issue that report. The memo also stated that the report was not in the standard format of current UTS reports in that it did not have a report i number or supervisory signatures, and if " nuclear related," was missing a traveller number and QA review signatures. However, the inspectors noted that the memo did not mention the fact that DE had provided similar reports (although covering multiple products and lots) on its letterhead until 1993.

1 l

b.4 UTS Correspondence with SynTech In an August 29, 1995, letter to SynTech in response to the apparent unauthorized use of DE letterhead, and after a reported telephone conversation between a UTS QA representative and a SynTech sales representative, UTS demanded that SynTech stop using DE stationery, inform UTS of every instance in which SynTech had published product analysis results on DE st Cionery, and inform all its affected customers that DE did not perform the analyses.

Again, there was no mention of the earlier reports DE had provided SynTech on DE letterhead before the establishment of UTS.

8 47-

b.5 SynTech Reply to UTS The inspectors also reviewed a copy of a SynTech letter, also dated August 29, 1995, from the SynTech QA director, addressed to the UTS QA representative, in reply to the UTS letter of the same date, that denied any possession or use of DE stationery. However, this letter also did not clearly answer the UTS questions. Its explanation of the circumstances surrounding the issuance of the report in question was vague and still did not explain how SynTech would have had legitimate possession of DE letterhead, not blank stationery, b.6 UTS Contact with NRC in February 1996, in response to the inspector's inquiries to DE and UTS, a UTS QA representative informed the inspector that a SynTech sales representative had stated to him (in the telephone conversatisi mentioned above prior to issuance of the UTS August 29, IM5, letter) that SynTech had l

' routinely reported results of chemical analyses of its products to licensees '

and others on Detroit Edison letterhead." This statement was construed by UTS as an admission by SynTech. In addition, to compound the misunderstanding, during this reported conversation, the SynTech sales representative allegedly could not recall, when asked by UTL, how SynTech came into possession of DE stationery, b7 NRC Review at SynTech When the NRC inspectors first learned of this issue, the inspectors were not aware of the DE letterhead reports on file at SynTech and the UTS QA representative did not mention them. However, in light of the DE reports and other documents reviewed by the inspectors at SynTech, the telephone conversation statement attributed by UTS to SynTech's sales representative (confirmed during the inspection), was consistent with the circumstances.

During this inspection, the SynTech sales represer,tative stated that he had admitted at the time of the reported telephone conversation that he could not offer a satisfactory explanation for the issuance of the analysis report in question (a) on what appeared to be DE letterhead during a period when UTS had taken over analysis of SynTech products, and (b) reported results on a lot of N-grade that neither UTS, nor apparently DE, had a record of analyzing. He stated that his remark about routinely issuing reports on DE letterhead was made in an effort to m plain to the UTS QA representative how SynTech came to have possession of DE letterhead. However he failed to make it clear that the letterhead to which he was referring was what the DE reports had been printed on, not blank DE stationery. SynTech stated during this inspection that it never had blank DE stationery. Nevertheless, the unclear explanation provided by the SynTech Sales representative, interpreted by UTS as an admission by SynTech, served to confirm UTS's suspicion that SynTech was routinely sending out fraudulent reports on DE letterhead (presumably using blank DE stationery), thus prompting UTS's August 29, 1995, letter. The inspectors found no evidence to suggest that SynTech had done more than send a single report on what appeared to be DE letterhead, and apparently only to Palisades on the one occasion in question.

9

- .=. - _ .

-As mentioned above, the inspecters found three reports on DE letterhead that had been sent to SynTech by DE, during the period before the formation of UTS, when DE's Technical and Engineering Services group had been.the name of the t

organization whose laboratory was performing contaminant analyses of SynTech

! products. The reports, dated in 1990 and 1992, were all signed by the analyst whose name, but not his signature, appeared on the report in question. The inspectors noted that none of the analysis reports from DE on file at SynTech.

! contained results of analysis of N-grade, and the UTS report on file at SynTech that did contain N-grade analvsis results was for lot 761-193, not lol 761-055 as in the SynTech report to P'alisades.

Review of associated records during this inspection confirmed that SynTech had in fact routinely reported its product test results to its customers simply by fonvarding copies of the DE reports which included results for the products and lots being shipped, among others. UTS had also stated that it had never given SynTech permission to use its stationery, letterhead, name or logo, but the inspectors found no evidence that either DE or UTS had ever prohibited SynTech from forwarding DE or UTS analysis results to SynTech customers on the original reports. UTS also confirmed the inspectors' findings at SynTech that, although the UTS laboratory had performed a number of analyses for SynTech on its products over the last two years, the lot listed on the apparently fraudulent report was never analyzed by UTS.

b.8 SynTech Internal Investigation and Explanation Presumably in anticipation of questions about this issue by the inspectors, SynTech disclosed the issue of the questionable report to the inspectors and offered their explanation of the circumstances. SynTech stated that in response to the August 29, 1995, letter from UTS, it had conducted an internal investigation to establish the facts and circumstances surrounding the issuance of the report in question. On the basis of the results of this investigation, SynTech developed the following explanation: At the time the report in question was sent to Palisades, May 1995, SynTech was using UTS to perform its product contaminant analysis for nuclear customers. However, on this particular occasion, in response to an order from Palisades for some of the N-Grade Gasket Remover and Paint Stripper (hereinafter referred to as "N-Grade"), SynTech discovered that it did not have a UTS analysis report on file for the lot of N-grade it had in stock, Lot 761-055. SynTech stated that upon informing Palisades of this fact, Palisades agreed to accept so-called

" typical" results for this order, because they would be testing a sample of the received product themselves anyway. Therefore, according to SynTech, instructions were given to a clerk (who was reportedly new at the time and who is no longer with the company) to prepare a report for Palisades of typical contaminant analysis results for N-Grade lot 761-055. SynTech explained (and produced examples from their files in support of the explanation) that after sending out DE reports with multiple product and lot results on them, it had become their practice for a while to scan into their computer the letter reports from DE. The reports could then be conveniently edited to produce customized reports containing only the data for the product and lot being shipped under the report and the accompanying CoC. The rationale was that this practice would result in fewer, if any, transcription errors, less 10 confusion, and admittedly, would not disclose unnecessarily the few results not pertinent to the shipment being certified that did not meet contaminant specifications.

During this inspection, the inspectors examined some of these edited reports and found that they were printed out on plain paper and in a font not

_ resembling the font of the original DE letters, but some were left partially in the form of letters from one of SynTech's contractor laboratories addressed either to the SynTech QA Manager or to the SynTech sales representative who '

was involved, depending on the vintage. The inspectors pointed out to SynTech that while it would have been appropriate to quote the results of the product and lot of interest from the test report and/or to refer to the report or even to attach a copy, it was not appropriate to edit the original report to l isolate the desired set of results, yet leave it partially ie the form of a '

letter report to SynTech which would result in a document that was fundamentally misleading as to its origin, if not the content. SynTech agreed, but stated that they had discontinued this practice entirely when their investigation to determine how the report in question could have been produced revealed that the scanning and editing practice was a significant contributing factor.

-SynTech stated that as nearly as they were able to determine, the clerk who had been told to produce a report for the Palisades N-grade order in question citing typical results for this product had called up an old DE report that l

had been scanned into the computer to obtain the results of analysis of a

! previous lot of the same product. !iowever, instead of using the report (not

! scann?d) provide by UTS in 1992 of Lot 761-193 of N-Grade that SynTech had on l file, the clerk reportedly deleted the data for products not being shipped, l leaving only the N-grade data and then edited the date of the report and the

! lot number to read May 15, 1995 and Lot 761-055 that was being shipped.

! SynTech stated that they were not able to find the original report from which the "typicals" presumably were taken so the origin or correctness of the data

could not be verified. Then, SynTech explained, and as is apparent when l examining the report, for an unknown reason, the clerk did not edit out the DE logo that appeared in the upper left hand corner of the page. The logo consisted of printed words in a stylized version of a bold block letter font and it had been scanned as part of the document by the optical character recognition (OCR) software being used by SynTech as part of the text of the letter instead of as a graphic, in a font as close to the one scanned as the software could reproduce. Upon close examination, the inspectors were able to see that in fact, what had appeared to UTS and CPCo to be DE letterhead, was a computer reproduced facsimile of the logc that had reportedly been on the stationery on which the original report (similar to ones seen by the inspectors in SynTech's files) had been written. Close examination of the letters revealed that the reproduced font differed slightly from the actual DE logo in the thickness, spacing and proportions of the letters, the alignment of the words, and in particular, in the form of the letter "t" as it appeared in the word " Detroit" in the name and address of DE in the logo. In the actual DE logo, the stylized "t" is not fully crossed. In the custom DE font, the cross of the "t" extends from the vertical part only to the right, not to the left as on a conventional "t " This feature, once noticed was evident on all the actual DE reports on file. Whereas, the scanned document, exhibited a 11

simliar font, but, as previously stated with slightly differently formed and aligned letters and with normal t's on which the cross piece extended fully from the left to the right sides. This information tended to support SynTech's explanation of how the report in question had been produced. Also, SynTech assured the inspectors that although it no longer had the scanning software used so that could not demonstrate the effect, it was named and described as readily available so that the NRC could test SynTech's explanation.

Nevertheless, SynTech was not able to offer a satisfactory explanation of why the scanned and incorrectly edited report was sent out in that form and not at least edited to the extent that others had been. Even the error of a new clerk would not explain why su>ervisors had not caught the error. SynTech could only postulate that in tieir haste to fill the order and deliver the paperwork quickly in response to customer pressure, they had failed to adequately review all the submitted documentation, including the accompanying CoC letter which also referenced the lot number instead of stating that the accompanying results were supposed to be typical,

c. Conclusion

Prior to the inspection, the information available to the inspectors, as discussed above, suggested a fraudulent SynTech report. However, SynTech's explanation of the circumstances surrounding the creation of the report in its unusual form, was plausible, if not conclusive, and was supported by other information obtained by the inspectors. Although SynTech could not offer a satisafactory explanation for the report's not being effectively screened through management review prior to issue, the incident appeared to have been isolated. The inspectors could not conclusively determine that there was any deliberate intent to mislead the recipient of the report and found no evidence of a pattern of similar activity.

3.4 procurement Status and Chemical Analysis Results

a. Scope To determine the validity of SynTech reported contaminant content of Lot No.

761-055 (as well as other products), and the consistency of contaminant content in SynTech products, the inspectors interviewed analysts at two of SynTech's former contract laboratories, both NRC licensee subsidiaries, had the results evaluated by the NRC Chemical Engineering and Materials Branch, reviewed numerous analysis reports on SynTech products and reviewed the nuclear-grade evaluation of SynTech products by GE NE Materials Engineering,

b. Observations and Findinas During a telephone conversation on July 15, 1996, between the inspector and a representative of Consumers Power (CPCo), Palisades Procurement Engineering, the Palisades procurement engineer stated that Syntech Products are treated l

under CPCo's consumables program as Augmented Quality items because SynTech is l

not a qualified / approved supplier. CPCo had, for a number of years, purchased

! 12 l

l l

the product in question ("ikGrade Gasket Remover and taint Stripper -

Certified Non-Chlorinated, Clinging Gel") among other SynTech products as non-safety-rtlated. However, because the products are used on safety-related plant components, CPCo said it always independently tested a sample of each lot / batch in its own Trail Street Labs (owned by CPCo) to verify the product meets CPCo's safety-related or Category A requirements. For consumables used for cleaning, etc., even though they are not allowed to remain in safety-related systems, CPCo conservatively prefers not to allow any undesirable substances (e.g., fluorides, chlorides, sulfides, heavy metals, etc.) to come in contact with safety-related metals. Therefore, a sample of each received lot is routinely sent to the CPCo lab by the Palisades receiving department for analysis. In September of 1995 LPCo's Trail St. Labs obtained results reported as " total chloride'[ sic) of 10,492 ppm on the sample of lot 761-055.

This was significantly higher than the 120 ppm "Cl" [ sic] results reported on the May 15, 1995 SynTech report in question. In accordance with CPCo i procedures, the issue was then referred to Procurement Engineering for i resolution.  !

i On August 6, 1996, the inspector talked to a chemist at CPCo's Trail Street Labs, about the analysis report provided by SynTech to Palisades of Lot No.

761-055 of its "N-Grade" solvent / cleaner and gasket remover. Discussed were the differences between her lab's analysis of the product, and the SynTech reported results, which differed widely as discussed previously. She confirmed the inspector's concern that the SynTech results of heavy metal analysis may be questionable because many of the heavy metals (e.g., As, Bi, Pb, Sn, At, Hg, etc.) were reported by SynTech on the order of <1 to <5 ppm, whereas, CPCo's analyses of these same elements were reported typically as <50 or <100 ppm because of the minimum sensitivity of the analyses. At the time, the inspector was concerned that unless SynTech (or its contract laboratories) had more sensitive instruments and processes for analyzing for these elements, the SynTech-reported results may have been below what is typically achievable, and would therefore, be questionable.

During this inspection, however, the inspectors found comparable results on another lot of the same product in a test report by UTS. Also, the report was reviewed by the NRC's Chemical Engineering and Materials Braach who stated that the SynTech-reported results were not below the sensitivity of the more sophisticcted techniques in common use in the industry.

During subsequent conversations with the analyst at UTS who had signed the old DE reports to SynTech and also the analyst who had performed the analyses for UTS, they indicated that results of analyses could vary widely depending on analytical method used if only minimums are being reported due to minimum sensitivity or detectab',e qualities. Also, for halides, sulfur, it is necessary to know whether the analysis was for total amounts or leachable amounts. If being analyzed for " total" content, chlorinated or fluorinated compounds will release large quantities of halides when certain analytical methods are employed.

The second conversation the inspector had with the CPCo chemist at the Trail Striet Labs confirmed that her analysis was for total chlorides, yielding 10,492 ppm, indicating tha+ the product contained chlorinated compound (s). On 13 August 29, 1996, the Palisades procurement engineer faxed the NRC copies of previous reports of analyses of SynTech products done by the CPCo lab for themselves and said they would locate the records of analyses they used to do for SynTech several years ago, just as UTS had done until recently. The inspectors found some of these old-(pre DE) CPCo reports on file at SynTech,

c. Conclusion ,

On 1he basis of the information from the UTS and CPCo laboratories, the NRC Chalcal Engineering and Materials Branch review, the GE NE nuclear-grade evaluation, and the analysis reports reviewed, the inspectors concluded that although N-Grade, Lot 761-055, contained excessive chloride, even for total (vice leachable) content, the results of the other substances analyzed for were consistent with the other analyses of N-Grade. The inspectors further concluded that the SynTech reported results, although their specific origin was indeterminate, appeared to be typical of N-Grade and were consistent with the capabilities of the analytical methods used.

However, on the basis of review of many SynTech product analysis reports and review of GE NE Materials Engineering recommendations in their nuclear-grade evaluation, the inspectors concluded that the contaminant content of SynTech products is variable enough such that " typical" analyses are not useful.

Therefore,- the practice of most of SynTech's nuclear customers to require batch-traceable or lot-traceable certified chemical contaminant analysis reports is prudent. Such reports would ensure that levels of undesirable substances such as halides, sulfur and heavy metals are within acceptable limits.

4. PERSONS CONTACTED

• +J. Leslie, President

+M. Bramson, Executive Vice President

• +0. Moore, Quality Assurance Manager J. Broderick, Sales T. Mosier, Assistant QC Coordinator

+ Attended the entrance meeting on February 4, 1997.

• Attended the exit meeting on February 6, 1997.

14 c8 44 y+ .*, UNITED STATES j

t NUCLEAR REGULATORY COMMISSION WASHINGTON. Df. 20666 0001

\*****/ thy 2,1997 Mr. Nicholas J. Liparulo, Manager Nuclear Safety and Regulatory Activities Nuclear and Advance ' Schnology Division Westinghouse Electric Lorporation P O. Box 355 Pittsburgh, Pennsylvania 15230 1

SUBJECT:

NRC INSPECTION NO. 9500: "J4/97 01 i

Dear Mr. Liparulo:

On April 17,1997, the U. S. Nuclear Regulatory Commission (NRC) completed an incpection of AP600 design control quality assurance activities at the Westinghouse Energy Canter in Monroeville, Pennsylvanic The enclosed report presents the results of that inspection.

The purpose of the irisuection was to evaluate the Westinghouse technical and quality oversight of Empresa Nacional de Ingenieria y Tecnologia, S.A. (INITEC) design activities in general, and to assess Westinghouse's evaluation and disposition of the audit findings identified in the August 24,1994 NRC " Summary of Audit of the AP600 Structural Desigri," report conceming NRC-identified errors in INITEC calculations.

During this inspection, the NRC determined that the implementation of the Westinghouse quality assurance program for AP600 design certification activities failed to meet certain NRC requirements. Specifically, the team identified nonconformances with program implementation with respect to your: (1) failure to initiate appropriate root cause determinat'on and corrective actions for find ngs described in the August 24, i994, NRC Audit Summary report, and (2) failure to adequately evaluate and assess INITEC's performance during your annual quality assurance review of AP600 suppliers, and to conduct an evaluation of INITEC's February 15,1995, response to the NRC audit findings as requested in Westinghouse's August 3,1994, letter, during your triennial audit. During the required triennial supplier audit of INITEC performed in February 1995, Westinghouse's audit plan and scope did not include the calculation audit findings identified in the August 1994 NRC Audit Summary report, and any associated INITEC corrective action activities, as issues that needed to be addressed Ic verify the effectiveness of INITEC's quality assurance program implementation.

Based on the nonconformances identified at ove, the NRC is concerned that these quality assurance deficiencies may have introduced a level of uncertainty on the acceptability of design deliverables provided by AP600 technical cooperation agreement participants. Of particular concern to the NRC, is Westinghouse's failure to recognize and appropriately address a condition adverse to quality, requiring a root cause evaluation and determination and appropriate corrective actions, even when such a condition was iduntifiod by an NRC audit and resulted in re design of the AP600 foundation basemat.

Mr. Nicholas J. Liparulo 2-Accordingly, the NRC requests that Westinghouse: (1) determine and evaluate the impact of these nonconformances on completed or related design deliverables and/or activities performed by all AP600 technical cooperation agreement participants; (2) identify the steps that it has taken, or intends to take, to demonstrate that other design deliverables provided by AP600 technical cooperation agreement participants do in fact achieve the level of integrity in design verification and quality assurance necessary to satisfy the design certification provisions of 10 CFR Part 52: and (3) provide a list of all AP600 technical cooperation agreement participants, including a description of their AP600 program work scope and involvement. The NRC is identifying these concerns as an unresolved item. Please provide the above requested information within 30 days of receipt of this letter.

The responses requested by this letter and the enclosed Notice of Nonconforri . ace 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.i90 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 thern with you.

Sincerely,

_ (

Robert M. Gallo, Chief Specialinspect on Branch Division of Inspection and Support Programs Office of Nuclear Reactor Regulation Docket No.52-003

Enclosures:

1. Notice of Nonconformance

2. Inspection Report No. 99900404/97-01 cc w/encis: See Next Page m

Mr. Nic6olas J. Lipard., Docket No.': 52 003

'Nestinghouse Electric Corporation - AP600 cc:

Mr. 8.A. McIntyre - . Mr. Ronald Simard,. Director Advanced Plant Safety & Licen:ang Advanced Reactor Programs i

- Westinghouse Electric Corporation Nuclear Energy Institute -  !

Energy Systems Business Unit 1776 Eye Street, N.W.

! P.O. Box 355 Suite 300 Pittsburgh, PA 15230 - Washington, DC 20006 3706 Ms. Cindy L. Haag- Ms. Lynn Connor l

- Advanced Plant Safety & Licensing Doc Search associates Westinghouse Electric Corporation Post Office box 34 Energy Systems Business Unit Cabin John MD,20818 Box 355 Pittsburgh, PA 15230 -

Mr. James E Quinn, Projects Manager

. - Mr. M.D. Beaumont LMR and SBWR Programs Nuclear and Advanced Technology Division GE Nuclear Energy Westinghouse Electric Corporation '175 Curtner Avenue, M/C 165 One Mnntrose Metro San Jose, CA 95125 11921 Rockvilla Pike Suite 350- -

Rockvillo, MD 20852 Mr. Robert H, Bucholz

,; GE Nuclear Energy Mr. Sterling Franks 175 Curtner Avenue, M/C 781 U.S. Department of Energy San Jose, CA 95125 NE 50

_ 19901 Germantown Road Barton 2, Cowan, Esq. ,

Germantown, MD 20874 Eckert Seamans Cherin & Mellott 600 Grant Street 42nd Floor .

Mr. S.M. Modro Pittsburgh, PA 15219

. Nuclear Systems Analysis Technologies Lockheed Idaho Technologies Company Post Office Box 1625 - Mr. Ed Rodwell, Manager

-idaho Falls, ID 83415 PWR Design Certification Electric Power Research Institute-

! Mr. Frank A. Ross 3412 Hillview Avenue

~

U.S. Department of Energy, NE 42 Palo Alto. CA 94303 Office of LWR Safety and Technology 19901 Germantown Road Mr. Charles Thompson, Nuclear Engineer Germantown, MD 20874 AP600 Certification U.S. Department of Energy ,

NE-50 i Washington, DC 20585

l NOTICE OF NONCONFORMANCE Westinghouse Electric Corporation Docket Nos.: 52 003 99900404 Pittsburgh, Pennsylvania Based on the results of a Nuclear Regulatory Commission (NRC) inspection conducted on April 17,1997, of activities supporting Westinghouse Electric Corporation's AP600 design certification,it appears that certain activities were not conducted in accordance with NRC requirements.

A. Criterion XVI, " Corrective Action," of Appendix B to 10 CFR Part 50, requires, in part, that measures be established to assure that conditions adverse to quality, such as f ailures, malfunction *, deficiencies, deviati ons, defective material ar.d 6:;uipment, and nonconformances are promptly identified and corrected.

WCAP 12600, "AP600 Quality Assurance Program Plan," Revision 2, dated December 15,1993. Section 16, " Corrective Action," states, in part, " Application of WCAP 8370 to AP600 activities includes the following:

• The root cause of significant conditions adverse to quality are determined and documented, and the impact of such conditions on completed or related items and activities is evaluated,"

WCAP 8370, Quality Assurance Plan (OA Topical Report), Revision 12A, dated April 1992, Section 16. " Corrective A:: tion," states, in part, in:

Section 1 t> " General," that conditions adverse to quality such as f ailures, malfunctio . 'nconformances, and out of cortrol processes (including failure to follow proceove shall be identified. These adverse conditions are also analyzed, documented, and corrected commensurate with their importance to safety.

Section 16.1, " Corrective Ac an," that personnel performing activities in accordance with this plan identify conditions adverse to quality and suggest, recommend or provide solutions to the conditions as appropriate. For significant conditions adverse to quality, the causes are determined and documented and the impact of such conditions on items and services is evaluated for significant trends and reported to the appropriate level of management.

Section 1S.2, " Follow up," that for corrective action resulting from reports (e.g.,

nonconformance reports, audit reports, computer software error reports, NRC inspection reports, customer audit reports, etc.) quality assurance participates in verifying that appropriate corrective action is documented and implemented.

Enclosure 1

Contrary to the above, Westinghouse:

-1,- Old not identify, analyze, document, a.id correct conditions adverse to

-quality as required by the AP600 Quality Assurance program. During a July 1994 NRC structural audit of the nuclear island foundation mat, errors were -

identified in calculations performed by INITEC that resulted in significant re design of the AP600 foundation basemat. Th6 findings described in.the August e,1994, NRC ' Summary of Audit of the AP600 Structural-Design," report were not identified as a condition adverse to quality requiring or receiving quality assurance participation in verifying that appropriate l corrective action is documented ar.d imple: inted. l 2, Did not adequately determine and document the root cause of INITEC's basemat :alculation erro's nor evaluate tha impact of such a condition i adverse to quality on completed or related INITEC AP600 design daliverables and activities, (99900404/97 01-01)

B. Criterion Vil, " Control of Purchased Material, Equipment and Services." of Appendix B to 10 CFR Ptrt 50, requires, in part, that measures shall be established to assure that purchased material, equipment, and services, whether purchased directly or througn contractors or subcontra:: tors conform to the procurement documents.

WCAP 12600, "AP600 Quality Assurance Program Plan," Revision 2 dated December.15,1993, Section 7 " Control of Purchased items and Services," states, in part, "AppHcation of WCAP 8370 to AP600 activities includes, but is not limited to, tne following:

o The initial qualification and suosequent performance evaluation of suppliers to which technicalcooperation apreaments (ernphasis added) apply is performed in the same manner as for suppliers of purchased items and services.

• The performance of aach supplier is evaluated on an annual basis, commensurate with the comp'exity and importance to safety of items or services provided. The evaluation is documented and includcs evidence, based on direct observation of work performed by the supplier, that the supplier's quality assurance program is continuing to operate successfully."

WCAP 8370, Quality Assurance Plan (OA Topical Repcrt), Revision 12A, dated April 1992, Section 7. " Control of Purchased items and Services," states, in part, in:

Section 7.3, " Supplier Performance Evaluation," that "A formal evaluation of suppliers is performed each year to determine if additional actions such as audits are required during the upcoming year. This evaluation includes a review of some or all of the following: prior quality program audits, supplier surveillance activities,

...results of audits from other sources (customers, ASME, NRC, crc.)if available temphasis sdded), ...and the suppliers's respcnsiveness and cooperatin in resolving 2-

goality questions or problems. As a result of this evaluation, supoliers requiring a +

complete quality program reaudit are identified...Regardless of the results of the evaluation. suppliers are reaudited every three years."

Contrary to the above, Westinghouse did not provide appropriate OA oversight of design activities performed by INITEC. Afte the basemat calculation errors were identified by the NRC in July 1994, Westinghouse did not evaluate or assess the ,

impact of the errors on other work performed by INITEC. Specificaily: ,

i 1, evestinghouse failed to adequately evaluate or atasess INITEC's ennual performancei as required by WCAP 8370, Part B,-Section 7.3, "Supplie, Performar,ce Evaluation," for a supplier of AP600 design deliverables that had been the subject of an adverse NRC audit finding.

2- In its February 1995 triennial audit of INITEC, Westinghouse f ailed to conduct an evalention of INITEC's response to Westinghouse's Aug.3t 3, .

1994, letter, and any associated corrective actions taken. The letter to INITEC described the basemat design caiculation issues identified by the NRC during the July 1994 structurol Jesign audit. (99900404/97 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

• he Chief, S peciat inspection Branch, Division of I.ispection und Support Programs, 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 Nonconformance" and shouH include for each nor.conformance: (1) a description of ti..

steps that were or will be taken to correct these items; (2) a description of the steps that have or will be taken to prevent recurrence; and (3) the dates your corrective actions aid preventative measures were or will be completed.

Dated at Rockville, Mayr land This_M day of ncay ,1997 l

I 3-

. 79 h

L .. .- , - -

U. S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION 4

Report _no.: 99900404/97 01_

i Organization: ' Westinghouse Electric Corporation .

Nuclear and Advan:.ed Technology Division .

Pittsburgh, Pennsylvania 15230

)

Contact:

h.r. Nicholas J. Liparuto.- Manager l Nuclear Safety and Regulatory Activities Nuclear ' Industry - . Nuclear steam supply system riesign, components and

. Activity: services

{

4 Date: April 17,1997 2

?

Inspectors: Richard P McInty'e, PStB/ DISP -i Juan D. Peralta, HOMB/DRCH Goutam Bagchi, ECGB/DE ,

Diane Jackson, PDST/DRPM i

- Approved: Gregory C. Cwalina, Chief Vendor inspection Section

Specialinspection Branch Division of inspection and Support Programs r

t a

Enclosure 2 n ,,n .- - - . >,.N- ,,-m,--, s .~c - ~.. .s; , L.w.,..-.. . .

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

r 1' LINSPECTION

SUMMARY

The purpose of the inspection was.to evaluate the Wastinghouse sechnical and quality-overtight of Empresa Nacional de Ingenieria y Twenologia.' S.A. (INITEC)-design activities in-generst, and to assess Westinghouse's thoroughness in its evaluation and disposition of the audit findings identified in the August 24,1904, NRC " Summary of Audit of the AP600 Structural Design" concer. ling NRC staff-luentified errors in INITEC calculations, as described in Section ,.,...

The inspection basea were:

I

• - Appendix B, " Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," to Part 50 vi ime 10 of the Code of Federal Reaulations l

-(10 CFR~ Part 50).

• AP600 SSAR, Revision 11, Section 17.3, " Quality Assurarece" 1

• - WCAP 12600, Revision 2, dated December 1993, " AP600 Quality Assurance Program Plan" During this inspection, two instances where Westinghouse failed to conform to NRC requirements were identified. Also, one unresolved item was identified.

~1.1 Nonconformance

• Nonconformance 99900404/97-01 01 was identified and is discussed in Section 3.3 of this report.

• Nonconformance 99900404/97-01 02 wcs identified and is discuj sed in Section 3.4 of this. report.

._1. 2 Unres Jved item

• Unresolved item 9?900404/97 0103 was identified anef is discussed in Section 3.5 of inis repcrt,

- 2 . STATUS OF PREVIOUS INSPECTION FINDINGS j

No previous inspection findings were reviewed during tids insp?ction.

9 i

2-

3 INSPECTION FINDINGS AND 01HER COMMENTS 3.1 AP600 Quality As'urance Proaram Chaptar 17 of the AP600 standard safet) a talysis report (SSAR) describes the West;nghouse Electric Corporation quality assurance (OA) pro 0 ram for the design phase of the AP6;. .Advancod Light Water Reactor (ALWR) Plant Program.

AP600 SSAP, Revision 11, Section 17.3, " Quality Assurance " states that activities performed prior to March 31,1996, were performed in accordence with the quality assurance plan described in Westinghouse topical report WC AP-8370, " Energy System Business Unit Power Gene,a.,on Business Unit, Quality Assurance Plan,"

Revision 12a, dated April 1992. WCAP 8370 applied to all Westinghouse activities affecting sality of items and services supplied to nuclear power plants . nd establishes Wettinghouse's corapliance with the provisions of Appendix B to 10 CFR 50.

WCAP 12600, "AP600 Quality Assurance Program Plan," dated December 1993, a peoject specific OA plan, was developed by Westinghouse to enhance WCAP-8370 in specific areas and to establish additional commitments needed to support the AP600 Design Certification and First Of A Kind (FOAKE) programs. WCAP-12600 establishes the responsioility of the Advance Technology Business Area of the Energy Systems Business Unit fo AP600 Design Certification and FOAKE programs and for control of the technicalinterface between Westinghouse and engineering groups and suppliers providing engineering services under such programs.

Additionally, WCAP-12601, "AP600 Program Operating Procedures," Revision 13, dated July 8,1994, was developed by Westinghouse to establish requirements and responsibikties for developing, approving, implementing, revising, and maintaining operating procedures to meet the OA and administrative requirements of the AP600 progra n. WCAP 12601 includes an "Al 600 Program Procedure Matrix," currently Revision 15, dated April 4,1995, which identifies the correlation between the Westinghouse commitments to the OA requirements of (1) ANSI /ASME NOA-1,

! " Quality Assurance Program Requirements for Nuclear Facilities," 1983 Edition (as endorsed by Regulatory Guide 1.28, Revision 3) and (2) ANSl/ASME NOA 1 1989 Edition through NOA 1b-1991 Addenda, and the corresponding implementing l guidance embodied in WCAP 9565, " Nuclear and Advanced Technology Division (NATO) Quality Assurance Program" currently Revision 34, dated May 2,1994, and l in V' CAP 12601 WCAP-9565 governed the implementation of all NATO activities related to areas within the scope of WCAP-8370.

3.1.1 AP600 Desion Certification and FOAKE Proarams Control of Pyrehased items and Services WCAP 12600, Section 7 " Control of Purchased items and Services," describes the application of WCAP 8370, Part 8 Section 7, provisions to AP600 activities and clarifies that the initial qualification and subsequent performance evaluation of 1

l l

suppliers to which technicalcooperation agreements apply (emphasis added)is j

performed in the same manner as for suppliers of pi rehased items and services, WCAP 8370, Part B, Section 7.3, " Supplier Performance Evaluation," requires, in part, "A formal evaluation of suppliers is performed each year to determine if add.tional actions such as audits are e ,uired daring the upcoming year. This evaluation includes a review of some or a'l of the following: prior quality program audits, supplier surveillance activities, nature and frequency of hardware discrepancies, results of audits from other sources (customers, ASME, NRC, etc.)

(emphasis added). ..." This sect;on also specifies that as a result of this evaluation, suppliers requiring a complete quality program reaudit are identified, 3.1,2 AP600 Particioation Prooram Between INITEC and Westinahouse Electric CRD2Ull!Da On March 5,1992, Westinghouse and INITEC signed an agreement under which INITEC would perform engineering, rnanagement and execution of certain structural and piping tasks associated with the APCOO Program. The following tasks were identified as INITEC's work scope: (1) Analp,a and Detail Design of Pressurizer Safety and Relief Valve (PSAR\/) Module, (2) Structurat Steel Framing (SSAR Section 3.8.4), (3) Floor Slabs (c,SAR Section 3.8.4), (4) Nuclear Island Basemat (SSAR Section 3.8.5), (5) Nuclear Island chear Walls (SSAR Section 3.8.4), (6) Analysis and Detail Design of Piping, and (7) Structural Steel Framing (SSAR Section 18.4).

Additionally, the agreement specified that design inputs, methods, required formats and other design information for structural tasks would be provided by Bechtel Power Corporation (Bechtel) and allowed for technical correspondence and interfaces for the3e tasks to take place directly between INITEC and Bechtel.

In Section 3.0, " Quality Assurance," the agreement specified that INITEC shall establish, implement, and maintain a quality assurance program meeting the requirements of ANSI /ASME NOA-1. "Qualits Assurance Program Requirements for Nuclear Facilities" (1986 edition). The agreement also specified that a Quality Assurance Program Plan (OAPP) would be prepared by INITEC and be subject to review and approval by Westir.,nouse prior to performing any quality-related work.

3.2 Nuclear Island Basemat Desion Bac':around On July 11 through 14,1994, the NRC performed an audit of the structural design of the AP600 at the Bechtel offices in San Francisco, California. The results of this audit were documented in a letter to Westinghouse dated August 24,1994,

" Summary of Audit of the AP600 Structural Design."

One of the key issues identified during the audit were errors found by the audit team in design calculations performed by INITEC (Calculation No.1010 CCC-001, Rev. A).

The NRC audit team identified the following deficiencies: (1) errors in shear and flexural rebar assessment, in the use of punching shear formula, and in the use of finite element dimension; (2) no consideration of accident pressure loads, and loads 4

83-

frnm con,tructh,.. equence, and (J) out of phase overturning moment from shield and containment buildings, in its August 2,1994, letter to the NRC, Westmghouse acknowledged its co.nmitment to: (1) perform an independent review of the basemat design calculations, (2) verify the :dequacy a ""TFC's in house post process computer programs used for the foundation mat oesign. (3) perform simplified analyses as appropriate to confirm the existing design results, and (4) provide the results r'f this independent review to the NRC, (This issue was identified by the NRC in the AP600 DSER as DSER Opeu Item 3.8.5 21), i 1

3.3 Review of Westinohouse Corrective Actions for July 1994 NRC Structural Audit Findinos 1

a. Scoce The inspectors reviewed Westinghouse's evaluation and disposition of the audit findings identified in the August 24,1994, NRC " Summary of Audit of the AP600 Structurai Design" report concerning NRC-identified errors in INITEC calculations.

b. Observations and findinos As indicated above, in its August 2,1994. Mter to the staff regarding errors identified in INITEC post process computer programs used for the AP600 foundation mat design, Westinghouse made a commitment to conduct an independent review of such calculations, in a letter to INITEC (FOK/IN10141) dated August 3,1994, Westinghouse forwarded: (1) comments presented by an NRC consultant at the July

, 1994 meeting, (?l additionalinterpretation and comments by Westinghouse technical staff, and (31 a copy of the Aagust 2,1994, letter to NRC Westinghouse also l stated that a Bechtel employee had been selected to perform the independent review and the scope of basemat review was attached to the letter (Attachment 4).

The inspectors reviewed the independent review report, " Independent Review of AP600 Nuclear Island Basemat Design," dated November 11,1994, performed by Bechtel to address the basemat calculation issues identified by the staff. The inspectors noted that ii Section 6.0, " Summary and Recommendations," of the report, Bechtel made several observations and recommendations related to uncertainties in the INITEC calculation that required further evaluation or study.

The inspectors reviewed an INITEC letter to Westinghouse (INI/FOK0175), dated February 15,1995,in which INITEC provided its response to address the root causes of the quality issue relative to the Nuclear Island Basemat Calculation,

identified by NRC in its July 1994 structural design audit, as well as the measures taken by INITEC in order to avoid the occurrence of similar situations during the performance of present and future structural analysis. The inspectors noted that INITEC's response had been formulated almost contemporaneously with the triennial audit being conducted by Westinghousa on February 20 through 22,1995, at 4

1NiTEC's f acilities. Yet Westinghouse's triennial audit report (OLAllN10007) does not provide any evidence that INITEC had identif.ed this issue as a condition adverse to i quality requiring root cause determination or corrective actions in accordance with INITEC's Westinghouse approved QAPP fsee Section 3.4.a.).

During the inspection, the inspectors inquired as to whether Westinghouse had l initiated any roct cause determinations or ar"rective actions, as required by WCAP 12600J 3ection 16, ' Corrective Action," since this issue was first identmed by the NRC in July / August 1994 or whether Westinghouse had formally accepted or rejected INITEC's proposed corrective actions identified in INITEC's February 1995 letter. We"inghouse representatives responded that, as of April 17,1997, this issue had not been identified as a condition adverse to quality requiring a root cause determination or corrective actions in accordance with WCAP 12600 nor had Westinghouse 1ntmally responded to INITEC's February 1995 letter.

c. Conclusions -

Based on the above, the inspectors concluded that Westinghouse: (1) failed to identify or recognize the NRC identified basemat design deficiencies as a condition adverse to quality requinng a root cause duermination or corrective actions in accordance with WCAP 12600, Section 16, and (2) f ailed to determine and evaluate the impact of such design deficiencies on completed or related INITEC AP600 desigr deliverables and activitit This issue was identified as Nonconformance 99900404/97 01 01.

3.4 Review of Westinahouse Oversicht of INITEC AP600 Desian Activities

a. Scone The inspectors reviewed Westinghouse technical and quality oversight of INITEC design activities in general, including the documents desenbing the contractual agreements between Westinghouse, INITEC and Bechtel for AP600 program design activities.

b. Observations and findinas Westinghouse appropriately specified all technical and quality assurance requirements consistent with WCAP 12600. As required by tha agreement signed on March 5,1992, INITEC provided a QAPP for Westinghouse's review and approval.

As described in Section 3.3 above, in September 1994, Westinghouse contracted the services of Bechtel for an independent review of INITEC's analysis and design methodology for the nuclear island basemat. The results of Bechtel's independent review were documented and made available to Westinghouse in a report dated November 11,1994.

4

'Although Westinghouse had identified the submitta! of this independent review report as a commitment in its August 2,1994, response to the NRC, the inspectors

were informed that a submittal of the report on the AP600 docket was still outstanding. No explanation was given as to why this report had not yet been made svailable for NRC review.

The inspectors learned that on August 3.15s94, Westinghouse had sent a letter to INITEC (FOK/lNIO181) requesting that INITEC provide its response to the NRC audit findings (see Section 3.3.a.). Howaver, it appeared that Westinghouse did not l consider the re.iutts of the NRC structural audit findings and concerns in its January  !

1995 annual review of INITEC's performance. j Additionally, the inspectors reviewed Westinghouse's Audit Report OLA/INIOOC7, j dated March 20,1995, that documented a triennial audit conducted on February 20 through 22,1E95, at INITEC's f acilities in Madrid, Spain. Upon reviewing the report, the inspectors found that: (1) an evaluatic.i of INITEC's response to Wr-tir ghouse's August 3,1994, letter had not been included in the audit scope, and (2) the audit did not identify any evidence to suggest that INITEC had initiated any internal root cause analysis, and evaluation. Further, th # inspectors determined that no corrective actions had been formally identified by INITEC's OA organization to determine the cause, and document t' a impact of the design deficiencies identified in Westinghouse's August 3,1994, letter, on INITEC's AP600 design deliverables.

Finally, the inspectors questioned Westinghouse's conclusion, in its March 22,1995, letter to INITEC, " Westinghouse AP600 Audit WES 95 211," that "lNITEC Design Quality Assurance Program and implementing procedures meet the applicable NOA 1 requirements for INITEC's AP600 work," in light of the fact that the audit report identified 10 findings of programma'.ic deficiencies in INITEC's implementation of its l OA DP, i

c. - Conclusiom Based on the above, the inspectors concluded that Westinghouse's (1) failed to o adequately evaluate or assess INITEC's performance, as required by WCAP 12600, L for a supplier of AP600 d-aign deliverables that had been the subject of an adverse NRC audit _ finding, and (2) failed to conduct an evaluation of INITEC's response to Westinghouse's August 3.1994, letter, and any associated corrective actions taken.

This issue was identified as Nonconformance 99900404/97 01 02, 3.5 Westinohouse Oversicht of AP600 Desian Activities t

L Based on the nonconformances identified above, the NRC is enncerned that these quality assurance deficiencies may have introduced a level of uncertainty on the acceptability of design deliverables provided by AP600 technical cooperation agreement participants. Of particular concern to the NRC, is Westinghouse's failure to recognize and appropriately address a condition adverse to quality, requiring a root cause evaluation and determination and appropriate corrective actions, even when 7

l

5 such a condition was identified by an NRC audit and resulted in re design of the  !

AP600 foundation basemat.

Westinghouse's f ailure to address this design and quality assurance program deficiency in & timely manner has raised the issue of whether this is an isolated case and that other design deliverab%s provided be AP600 technical cooperation

.egreement participants do in fact possess the level of integrity in design verification and quality assurance necessary to satisfy the design certification provisions of 10 CFR Part 52. This issue was loentified as Unresolved item 99900404/97-0103. l 3.6 Entrance and Exit Meetinas i

Since this was a one day inspection, the entrance and exit meetings were both held on April 17,1997, in the entrance meeting the NRC inspectors discussed the scope of the inspection and outlined the areas to be inspected. In the exit meeting the inspectors discussed the inspection findings and unresolved item, 4 PERSONNEL CONTACTED

, Wasiinnhouse Electric Corocration David Alsing, AP600 Quality Systems Manager Ken Kloes, Projects Quality Assurance Engineer Robert Tupper, Advanced Technology,' Project Engineer Richard Orr, AP600 Structures Advisory Engineer Narenda Prasad, Engineering Technology Department, Fellow Engineer Donald Lindgren, AP600 Licensing ITEMS OPENED, CLOSED, AND DISCUSSED ,

Onened

{: 99900404/97 01 01 - NON- inadequate corrective action 99900404/97-01 02 NON inadequate quality and technical oversight of INITEC 99900404/97 01-03 - URI acceptability of AP600 design deliverables 4

3 i

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Selected Generic Correspondence on the Adequacy of Vendor Audits and the Quahty of V.endor Products identifier lille information Notice 97-15 Reporting of Errors and Changes in Large-Break Loss-of-Coolant Accident Evaluation Models of Fuel Vendor 3 ;nd Compliance with 10CFR50.46(a)(3)

Information Notice 97-32 Defective Worm Shaft Clutch Gears in Limitorque Motor-Operated Valve Actuators NRc posv sW U & NuCLLAR REGULAIDRY COMMISSION 1. RE PORT NUMDE3 rr sa :see,,,e ,, N=c, u, yet so,,, ne, ,

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