ML20138J246

Licensee Contractor and Vendor Inspection Status Report. Quarterly Report,July-September 1996.(White Book)

ML20138J246
Person / Time
Issue date:	01/31/1997
From:	Office of Nuclear Reactor Regulation
To:
References
NUREG-0040, NUREG-0040-V20-N03, NUREG-40, NUREG-40-V20-N3, NUDOCS 9702060133
Download: ML20138J246 (147)

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! NUREG-0040 i Vol. 20, No. 3 i

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! Licensee Contractor j and Vendor Inspection l Status Report i

l Quarterly Report

] July-September 1996 j

i l U.S. Nuclear Regulatory Commission i

Office of Nuclear Reactor Regulation I

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1 AVAILABILITY NOTICE 1 Availability cf 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, Washiagton, 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 Information Service, Springfield, VA 22161-0002 Although the listing that follows represents the majority of documents cited in NRC publica-tions, it is not intended to be exhaustive.

Referenced docurnents available for inspection and copying for a fee from the NRC Public Document Room include NRC correspondence and internal NRG memoranda: NRC bulletins, circulars, information notices, inspection and investigation notices; licensee event reports; vendor reports and correspondence; Commission papers; and applicant and licensee docu-ments and correspondence.

The following documents in the NUREG 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 FederalRegula-tions, 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 translations, and non-NRC con-ference proceedings are available 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 Mail Services Section, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.

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-l 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 i Standards, from the American National Standards Institute,1430 Broadway, New York, NY 10018-3308.

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A year's subscription of this report consists of four quarterly issues. j J

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NUREG-0040 Vol. 20, No. 3 Licensee Contractor and Vendor Inspection Status Report Quarterly Report i July- September 1996 Manuscript Completed: January 1997 i Date Published: January 1997 l Division ofInspection and Support Programs Omcc of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 f  %,,,

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ABSTRACT This periodical covers the results of inspections performed by the NRC's Special Inspection Branch, Vendor Inspection Section, that have been distributed to the inspected organizations during the period from July 1996 through September 1996.

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l CONTENTS E8EE 1 Abstract ................................................................. iii ,

Introduction ............................................................. vii Inspection Reports ....................................................... 1 ABB-Electro Mechanics (99901297/96-01) ......... 2 New Britain, CT l Aerofin Corporation (99901302/96-01) ......... 13 Lynchburg, VA Arkwright mutual Insurance Company (99901296/96-01) ......... 24 Waltham, MA Commercial Union Insurance Company (99900632/96-01) ......... 44 Boston, MA Framatome Technologies, Inc. (99901300/96-01) ......... 57 Lynchburg, VA General Electric Nuclear Energy (99900003/96-01) ......... 72 Nuclear Energy Production Wilmington, NC Nuclear Logistics, Inc. (99901298/96-01) ........ 120 Fort Worth, TX Select Generic Correspondence on the Adequacy of Vendor ................. 138 Audits and the Quality of Vendor Products

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INTRODUCTION ,

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j A fundamental premise of-the U. S. Nuclear Regulatory Commission (NRC) l licensing and inspection program is that licensees are responsible for the

proper construction and safe and efficient operation of their nuclear power i j pl ants. The Federal government and nuclear industry have established a system l

for the inspection of comercial nuclear facilities to provide for multiple levels of inspection and verification. Each licensee, contractor, and vendor i participates in a quality verification process in compliance with requirements  :

3 prescribed by the NRC's rules and regulations (Title 10 of the Code of Federal j Regulations). The NRC does inspections to oversee the commercial nuclear l industry to determine whether its requirements are being met by licensees and i j their contractors, while the major inspection effort is performed by the industry within the framework of quality verification programs. )

l The licensee is responsible 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 NRC requirements, guides, codes, and standards.

The Vendor Inspection Section (VIS) of the Special Inspection Branch reviews and inspects nuclear steam system suppliers (NSSSs), architect engineering (AE) firms, suppliers of products and services, independent testing laboratories performing equipment qualification tests, and holders of NRC 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 I coordination between licensees and vendors.

The VIS does inspections to verify the quality and suitability of vendor products, licensee-vendor interface, environmental qualification of equipment, and review of equipment problems found during operation and their corrective action. When nonconformances with NRC requirements and regulations are found, the inspected organization 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 correspondence such as information notices and bulletins.

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This quarterly report contains copies of all vendor inspection reports issued I 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 problem areas that may l require special attention. Appendices list selected bulletins, generic letters, and information notices, and include copies of other pertinent correspondence involving vendor issues.

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INSPECTION REPORTS  !

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pweg k UNITED STATES O E NUCLEAR REGULATORY COMMISSION U

f WASHINGTON, D.C. 20555-0001 p'

August 22. 1996 Mr. Don F. Pedritti, President & General Manager ABB-Electro Mechanics 150 John Downey Drive New Britain, CT 06050-0750

SUBJECT:

NRC INSPECTION REPORT 99901297/96-01

Dear Mr. Pedritti:

I On June 20, 1996, the U.S. Nuclear RNulatory Commission (NRC) completed an inspection at your ABB-Electro Mechanics (ABB-EM) facility. The enclosed report presents the results of that inspection.

The inspection was conducted to ascertain whether licensees effectively monitored your control of quality for safety-related instrumentation and control systems and associated spare a,d replacement parts purchased by l licensees for nuclear power plants. The inspector assessed specific attributes and implementation of your licensees' monitoring of these areas. quality control program and the Two licensees purchased ABB-EM items from your facility and performed effective audit and surveillance activities, including identifying problems concerning your dedication of commercial-grade components. Both licensees removed ABB-EM from their list of approved suppliers. Twelve licensees that procured your items through your parent company ABB-Combustion Engineering (ABB-CE) relied on ABB-CE audits of your audits of your quality assurance program,even programthough rather thanitems procured performing were direct assembled at your facility. The Nuclear Procurement Issues Committee (NUPIC) also relied on ABB-CE audits rather than perform direct audits of your program. By relying upon the ABB-CE :"dits, the 12 licensees and NUPIC were unaware of the problems identified with your commercial arade dedication program.

During this inspection, the NRC inspector determined that the implementation of your quality assurance program failed to meet certain NRC requirements imposed on you by your customers. Specifically, you failed to take corrective action to prevent recurrence of deficiencies related to dedication of commercial-grade components for safety elated functiora, as required by 10 CFR Part 50, Appendix B.

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D. Pedritti This nonconformance is cited in the enclosed Notice of Nonconformance (NON),

and the circumstances surrounding it are described in detail in the enclosed report. Please respond to the nonconformance and 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 anti its enclosures will be placed in the NRC's Public Document Room (PDR).

Sincerely, 8

9 Ilobert M. Gallo, Chief Special Inspection Branch Division of Inspection and Support Programs Office of Nuclear Reactor Regulation Docket No. 99901297

Enclosures:

1. Notice of Nonconformance

2. Inspection Report 99901297/96-01 cc: See next page 3

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Michael P. Lilley, Manager, Quality Assurance  !

Rochester Gas & Electric Corporation 89 East Avenue Rochester, NY 14649 1

Richard H. Fassler, Supervisor, Quality Assurance Niagara Mohawk Power Corporation l Nine Mile Point Nuclear Station P.O. Box 63 l Lycoming, NY 13093 )

Ronald Casavant, Technical Specialist Washington Public Power Supply System P.O. Box 968 t Richland, WA 99352

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l Daniel Geneva, '

Supervisor / Vendor Audits Baltimore Gas & Electric Company l

39 West Lexington St. I 19th Floor Baltimore, MD 21202 Michael Dobrzensky, Engineer Pacific Gas & Electric Company P.O. Box 770000, Mail Code A10D San Francisco, CA 94177 i

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NOTICE OF NONCONFORMANCE ABB-Electro Mechanics (ABB-EM) Docket No.: 99901297 New Britain, Connecticut On the basis of an inspection conducted on June 17 through 20, 1996, it appears that the following activity was act conducted in accordance with NRC requirements:

• Criterion XVI of Appendix 8 to 10 CFR Part 50, " Corrective Action,"

requires, in part, that deficiencies and nonconformances are promptly identified and corrected. In the case of significant conditions adverse to quality, the measures shall assure that the cause of the condition is determined and corrective action is taken to preclude repetition.

Section 5.8.6 of ABB-EH's Nuclear Safety Related Designation Program for Commercial-grade Items, " Critical Characteristics," Revision 6, requires, in part, that critical design characteristics shall be defined for those items that have been determined ta have a safety-related function.

Contrary to the above, ABB-EM's dedication activities failed to establish critical characteristics for certain commercial-grade components (e.g., j test jacks, diodes, and other PC Board components) procured for safety-  !

related functions. ABB-EM failed to take corrective action to preclude repetiticn of such deficiencies. (99901297/96-01-01)

Please send a written statement or explanation to the.U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555-0001, 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 l 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 nonconformance: (1) a description of steps that have been or will be taken to correct these items; (2) a description of steps that have been or will be taken to prevent recurrence; and (3) the dates your corrective actions and preventive measures were or will be completed.

Dated at Rockville, Maryland this 22nd day of August 1996 Enclosure 1 5

U.S. NUCLEAR REGULATORY COMMISSION OFFILE vr NUCLEAR REACTOR REGULATION l

Report No: 99901297/96-01 l l Organization: ABB-Electro Mechanics New Britain, CT 06050-0750 1

Contact:

Peter Ferwerda, Quality Assurance Manager 860/826-4105 )

Nuclear Industry Instrumentation and control systems and Activity: associated spare and replacement parts ,

1 Dates: June 17 - 20, 1996 I

Inspector: Anil S. Gautam, Senior Engineer I Approved by: Gregory C. Cwalina, Chief Vendor Inspection Section.

Special Inspection Branch '

Division of Inspection and Support Programs l

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Enclosure 2 6

I INSPECTION

SUMMARY

During this inspection, the NRC inspector reviewed the implementation of selected portions of the ABU-Electro Mechanies (ABB-EM) quality assurance (QA) orogram, and reviewed activitits associated with licensee monitoring of these areas.

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."

• NRC Regulatory Guide 1.144, " Auditing of Quality Assurance Programs for Nuclear Power Plants."

• ABB-EM's Quality Assurance Manual (QAM) 200, Revision 7, dated September 19, 1994, and associated implementing procedures.

The inspector noted one instance in which ABB-EM failed to conform to NRC requirements imposed upon them by NRC licensees. This nonconformance is discussed in Section 3.1 of this report.

2 STATUS OF PREVIOUS INSPECTION FINDINGS This was the first NRC inspection of ABB-EM.

3 INSPECTION FINDINGS AND OTHER COMMENTS 3.1 Quality Assurance Proaram

a. Insoection Scooe i The inspector examined ABB-EM's quality assurance program, policy, implementing procedures and management directives. The inspector assessed ABB-EM's conformance to procurement documents, evaluation and corrective actions in response to licensee audit findings, commercial-grade item dedication, Part 21 evaluations, monitoring of subvendors, responses to pertinent audits, and self-assessment of performance.

b. Observations and Findinas The inspector assessed ABB-EM's commercial-grade dedication process.

ABB-EM purchased commercial-grade items from subvendors and qualified them to Appendix B requirements. Section 5.8.6 of ABB-EM's Nuclear Safety Related Designation Program for Commercial-grade Items, " Critical Characteristics," Revision 6, requiret. critical design characteristics to be defined for those items determined to have a safety-related function. ABB-EM's dedication activities failed to establish critical characteristics for certain commercial-grade components (e.g., test 2

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Jacks, diodes, and other PC Board components) procured for safety-related functions for Rochester Gas & Electric (RG&E). P.G&E identified deficiencies in this area ar' ABB-EM corrected them before the items were released to RG&E. However, ABB-EM failed to take corrective action to preclude repetition of such deficiencies, as required by Criterion XVI, " Corrective Action," of Appendix B to 10 CFR Part 50. This constitutes Nonconformance 99901297/96-01-01.

The inspector assessed AB8-EM's implementation of licensee purchase j orders (P0s), including evaluating failures of purchased items. The l inspector observed that compones,t failures were documented in ABB-EM's data bases. The QA manager stated that ABB-EM was planning to implement

" Root Cause Analysis Procedure and Training" for staff to enhance efforts toward trending failures and prevent recurrence.

The inspector assessed ABB-EM's organization and responsibilities. The QA program was comprised of the QA manager, QA engineering manager, quality control (QC) supervisor, and three QC '.nspectors. The QA manager reported to ABB-EM's president and general manager. QC inspectors had the authority to stop production of a nonconforming item until the nonconforming conditions were corrected. AB8-EN's senior management supported its QA program, assigned responsibilities, and remained involved in the implementation of the process.

The inspector observed that the QA manager could supersede a QC inspector's finding and sign off approval to continue the work. The QA manager stated that such instan as occurred to resolve minor discrepancies but he planned to make it clear to staff on when and how management may supersede QC findings. To better document such decisions in the future, ABB plans to remove the "use as is" block from its problem disposition reports (nonconformance reports) and require l engineering evaluation of any superseded inspection findings.

The inspector examined certificates of conformance (C0Cs) for items purchased by RG&E and Niagara Mohawk Power's (NMP's) P0s NQ-14931 and 02326, respectively. The C0Cs attested that purchased items were processed in accordance with the QA Manual (QAM) 200, specifications, requirements, and drawings of the P0s. The C0C for P0 NQ-14931 (to purchase auxiliary relay rack assembly component boards) was validated by ABB-EM Test Record 11914-1, Revision A.

The inspector observed that there was no formal ABB-EM guidance about which nonconformances should bc transmitteo w e iicensee. For example, licensees were not routinely informed of nonconformances dispositioned as "use as is" or " repair," unless licensee purchase orders required this notification.

The inspector questioned if ABB-EM had found any design errors and failures during the past 3 years and if it had istued any Part 21 reports. ABB-EM reported one design deficiency in accordance with 10 CFR Part 21 regarding a jumper wire on a printed circuit.

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l ABB-EM conducted audits of its Appendix B subvendors. These audits i

addressed appropriate criteria. The inspector observed that the reports, in general, did not indicate ABB-EH's conclusions on the adequacy or acceptability of its subvendor's implementation of Appendix l B criteria. The QA manager stated that ABB plans to revise its audit i

checklist to determine whether the program audited complied with i Appendix B. This revision is expected to be completed by September 30, 1996.

l The inspector observed that QAM 200 was based on the policies and

! criteria of 10 CFR Part 50, Appendix B but did not specify whether the QA program was in compliance with Appendix 8. The QA manager agreed to l issue a revision to 0AM 200 stating that the manual was in compliance with Appendix B criteria.

The inspector assessed ABB-EM's internal audits conducted in 1995. QA routinely conducted internal audits for Appendix B criteria to evaluate ABB-EM's quality performance. The audits, in part, assessed the results l j of source surveillances, receipt inspections, nonconformance reports, I and corrective actions. The inspector observed that no unsatisfactcry l conditions were found.

c. Conclusions The inspector concluded that, in general, the QA manual, work instructions, and procedures were satisfactory, except for the nonconformance described.

3.2 Review of Licensee Monitorina of ABB-EM

a. Inspection Scope The inspector assessed licensee monitoring of ABB-EM's control of quality for safety-related items purchased by licensees. The inspector reviewed licensee audits /surveillances, any restrictions imposed by licensees on ABB-EM concerning the manufacture of items purchased by licensees, and ABB-EM's corrective actions in response to licensee. audit findings.

b. Observations and findinas RG&F and NMP purchased safety-related instrumentation systems and components from ABB-EM during 1993-1996.

RG&E audited ABB-EM on June 22-24, 1993. The audit comprised of i monitoring, witnessing, and observing activities. The inspector l

reviewed RG&E's audit report and determined that RG&E's audit was l effective and had been performed in accordance with proper criteria, l written procedures, and checkli: .s. The RG&E audit team identified six l findings concerning ABB-EM's compliance to P0 requirements, design control, material control, corrective action pertinent to PDRs, and procurement requirements imposed on subvendors. RG&E found four of 4

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ABB-EM's responses to the findings unacceptable. Further ABB-EM responses were evaluated and accepted by RG&E in December 1993 and in January 1995; however, RG&E was not satisfied with ABB-EM's commercial dedication (upgrading) of PC board components. RG&E placed a restriction on ABB-EM requiring ABB-EM to submit its commercial-grade item dedication plan to RG&E before shipping components or repaired assemblies incorporating any commercial-grade dedicated item. In its March 27, 1996, letter to ABB-EM, RG&E addressed requirements of its P0 NQ-14931-A-JW and stated that ABB-EM had not adequately addressed critical characteristics of certain components of PC boards. RG&E plac.ed restrictions on this PO urm, it issued a written confirmation of acceptance of ABB-EM's commercial dedication plan. In July 1996, RG&E removed ABB-EM from its list of approved suppliers because of concerns regarding commercial dedication. l The inspector asked NMP's QA staff (by telephone) whether NMP had monitored ABB-EM's control of quality and whether it had evaluated RG&E's findings. A member of NMP's QA staff gave the inspector a copy of NMP's checklist for evaluating RG&E's audit and stated.that NMP's monitoring was based on RG&E's audit, that the audit scope encompassed NMP's purchases, and that NMP placed six restrictions on ABB including  ;

review and approval of dedication plans prior to start of work. NMP removed ABB-EM from its list of approved suppliers in July 1996.

The following 12 licensees purchased ABB-EM instrumentation systems and components through ABB-EM's parent company ABB-Combustion Engineering (ABB-CE): Southern California Edison, Consolidated Edison, Arizona Public Service, Baltimore Gas and Electric, Florida Power and Light, Omaha Public Power District, Louisiana Power & Light, Commonwealth Edison, Northeast Utilities, Public Service Electric and Gas, Houston Light and Power, and Consumers Power. The inspector cbserved ti.at nor.e of these licensees audited ABB-EM even though all hardware and repair parts were assembled at the ABB-EM facility. The Nuclear Procurement Issues Committee (NUPIC) audited ABB-CE in October 1995. The inspector asked the NUPIC audit team leader tstaff members of Washington Public Power Supply System) why ABB-EM was excluded from the NUPIC audit. The team leader stated that NUPIC did not audit vendor subcontractors.

The inspector observed that the aforementioned 12 licensees and NUPIC relied on ABB-CE audits of ABB-EM rather than performing direct audits of ABB-EM's program. However, there was no documented evidence indicating that ABB-CE evaluated problems identified by RGL. and NMP concerning ABB-EM's failure to establish characteristics of certain 7 commercial-grade components (see enclosed Notice of Nonconformance) and whether these problems impacted items purchased from ABB-EM. "

Apparently, the 12 licensees and NUPic did not adequately monitor ABB-EM's program since ABB-CE did not identify any audit findings, including problems concerning commercial-grade dedication. The inspector contacted NUPIC members (staff membt s of Pacific Gas & Electric and Baltimore Gas & Electric) but could not determine whether ABB-EM would be included during the next NUPlc audit of ABB-CE, planned for late 1997.

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l Wisconsin Electric Power Company sent a questionnaire to ABB-EM in March 1993 to evaluate any changes to ABB-EM's product line, quality program, procedures, facilities, and personnel. The inspector observed that the questionnaire did not request information c,n any recent problems with P0 procurement specifications, subvendors, or other quality involvement.

The ABB-EM QA manager stated that Florida Power and Light, Southern California Edison, Consumers Power, and RG&E visited ABB-EM to witness the testing of purchased items. The inspector observed that there was no documentation available relevant to such activities, including whether the licensees approved resuits of the test surveyed. The QA manager stated that ABB-EM did not issue or ask licensees to provide any document regarding licensee surveillance and considered the customer's shipping release to indicate the customer's approval,

c. Conclusions In general, licensee monitoring of ABB-EM's quality control program for safety-related instrumentation systems and components purchased from ABB-EM was not always satisfactory. Two licensees that purchased items directly from ABB-EM did perform effective audit and surveillance activities. The concerns identified on the audit activities resulted in ABB-EM being removed as an approved supplier by both licensees.

However, 12 licensees and NUP!C relied upon ABB-CE's audits of ABB-EM which did not identify the concerns regarding commercial grade dedication.

I 3.3 Entrance and Exit Meetinas In the entrance meeting on June 17, 1996, the NRC. inspector discussed the scope of the inspection, outlined the areas to be inspected, and established interfaces with ABB-EM management. In the exit meeting on June 20, 1996, and by telephone on July 26, 1996, the inspector l

discussed his findings and observations. i l l I

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i PARTIAL LIST OF PERSONS CONTACTED ABB-Electro Mechanics

! Don Pedretti, President, General Manager

! John Davit, Director of Operations j Peter Ferwerda, Quality Assurance Manager Vanessa Boulier, Material Services Manager

William Hadovski, Engineering Manager 4

Alex Oja, Contract Manager Edward Rollins, Plant Manager i Ray Majka, Quality Control Supervisor

. William Wayland, Quality Assurance Engineering Supervisor i Rochester Gas & Electric (contacted by telephone) 1 Michael P. Lilley, Manager, Quality Assurance Niaaara Mohawk Power (contacted by telephone) l

? Richard H. Fassler, Supervisor, Quality Assurance Washinaton Public Power Supply System (contacted by telephone)

Ronald Casavant, Technical Specialist (1995 NUPIC team leader)

Baltimore Gas & Electric (contacted by teleohone)

Daniel Geneva, Supervisor / Vendor Audits (1997 NVPIC team leader)

Pacific Gas & Electric (contacted by telechone)

Michael Dobrzensky, Engineer ITEMS OPENED, CLOSED, AND DISCUSSED l

Opened 99901297/96-01-01 NON failure to take corrective action l

Closed l

None Discussed ,

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• • "% J l ,  % UNITED STATES

! E, E NUCLEAR REGULATORY COMMISSION I

U f WASHINGTON, D.C. 20555-0001

%,,,,,# September 3, 1996 _

Mr. Brian C. Elliott, Manager  !

Aerofin Corporation 4621 Murray Place Lyr.chburg, VA 24502 l

SUBJECT:

NRC INSPECTION REPORT 99901302.'96-01 ,

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Dear Mr. Elliott:

On July 19, 1996, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Aerofin Corporation facility. The enclosed report presents the results of that inspectio .

The inspection assessed specific attributes and implementation of your quality assurance program to ascertain whether it met NRC requirements, and whether licensees effectively monitored your control of quality for safety-related finned heat exchangers, piping subassemblies, and component support purchased by licensees for nuclear power plants.

During this inspection, the NRC inspector did not identify any instances in which your quality assurance program failed to meet NRC requirements for the areas inspected. Your corrective actions .nd monitoring of subvendors were also found to be adequate. Therefore, no response to this letter is required. l

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In general, licensee monitoring of your quality assurance program for safety-related heat exchangers was not always satisfactory. Five licensees did perform effective audit and surveillance activities. However, on the basis of licensee documents, two licensees did not address details of licensees' monitoring of third-party audits of Aerofin and three licensees did not address Aerofin's lack of dedication of commercial-grade items for safety-related heat exchangers. Details of these concerns are addressed in the report.

In accordance with 10 CFR 2.700 of the NRC's " Rules of Practice," a copy of this letter and its enclosures will be placed in the NRC's Public Document Room.

Sincerely,

/ Robert'M. allo, Chief

/ Special Inspection Branch Division of Inspection and Support Programs Office of Nuclear Reactor Regulation Docket No. 99901302

Enclosure:

Inspection Report 99901302/96-01 cc: See next page.

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f cc:

. Michael P. Lilley, Manager, Quality Assurance Rochester Gas & Electric Corporation 89 East Avenue Rochester, NY 14649 James Johns ,

Quality Service Supervisor ,

Duquesne Light Company P.O. Box 4, Mail Code ISI Shippingport, PA 15077 ,

William Poteat Sr. Procurement' Engineer l Carolina Power & Light Company ENP, Zone 17-A P.O. Box 165 State Road 1134 New Hill, NC 27562-165  :

William Jewell-Project Engineer  !

Consumers Power Company l 27780 Blue Star Memorial Highway i Covert, MI 49043 Samson E. Blue QA Specialist Duke Power Company Mail Code EC09G P.O. Box 1006 Charlotte, NC 28201-1006 Jerry Bragg Group Leader Detroit Edison Company 6400 North Dixie Highway Newport, MI 48166  !

William Moody ,

-Senior Engineer '

Southern Nuclear Operating Company i P.O.. Box 1295 i

. Eirmingham, AL 35201-1295 Lowell Arnold Senior Quality Specialist Northern States Power Company 414 Nicollet Mall Minneapolis, MN 55401-1927 14

U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION Report No: 99901302/96-01 Organization: Aerofin Corporation Lynchburg, VA 24502

Contact:

Sub DasGupta, Quality Assurance Director (804) 528-6282 Nuclear Industry Finned heat exchangers, piping subassemblies, Activity: and component support Dates: July 18-19, 1996 i I

l Inspector: Anil S. Gautam, Senior Engineer 1 1

Approved by: Gregory C. Cwalina, Section Chief Vendor Inspection Section Special Inspection Branch Division of Inspection and Support Programs l

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l Enclosure 15

l 1 INSPECTION SUMARY During this inspection, the NRC inspector review 2d activities associated with licensee monitoring of Aerofin Corporation's entrol of quality and implementation of selected portions of its quality assurance (QA) program.  !

The inspection bases were as follows: l

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

• 10 CFR Part 21, " Reporting of Defects and Noncompliance." i Aerofin's Nuclear Quality Assurance Manual (NQAM) and Nuclear Quality i Procedure Sheet (NQPS) for the manufacture of finned heat exchangers, j piping subassemblies, and component support to the American Society of Mechanical Engineers (ASME) Code Section III, Division 1, Classes 2 and 3, "N" and "NPT," Revision 7, dated February 20, 1996.

For the areas inspected, the inspector did not identify any instance in which Aerofin's QA program or practice did not con *orm to NRC requirements.

However, the inspector did identify weaknesses in some licensee's monitoring of Aerofin's QA program. These items are discussed in Section 3.1.

2 STATUS OF PREVIOUS INSPECTION FINDINGS l This was the first NRC inspection of Aerofin. i 3 INSPECTION FINDINGS aND OTHER COMENTS 3.1 Review of Licensee Monitorina of Aerofin

a. Insnection Scope The inspector assessed licensee monitoring of Aerofin's control of quality for safety-related items purchased by licensees. The inspector reviewed licensee audits ard surveillances of Aerofin, any restrictions imposed by licensees on Aerofin concerning the manufacture of items purchased by licensees, Aerofin's corrective actions in response to licensee observations and findings, and Aerofin's monitorir.g of '

subvendors.

b. Observations and Findinas Eight licensees purchased safety-related finned tube heat exchangers from Aerofin during 1993-1996. The licensees included Rochester Gas &

Electric Corporation (RG&E), Northern States Power Company (NSP),

Detroit Edison Company (DE), Carolina Power & Light Company (CP&L),

Consumers Power Company (CP), Duke Power Company (DP), Southern Nuclear Operating Company (SN), and Duquesne Light Compar.y (DL). The inspector evaluated licensee audit reports, surveillances, and disposition of 2

16

findings. The inspector determined that monitoring, in general, was performed in accordance with proper criteria, written procedures, and J checklists. The Nuclear Utilities Procurement Issues Comittee (NUPIC) did not audit Aerofin.

RG&E audited Aerofin on March 2-5, 1993. The audit comprised monitoring, witnessing, and observing activities, such as inspections, examinations, and performance tests. Three observations were made: (1) j Aerofin did not address consideration for Part 21 reportability on its

! nonconformance report (NCR) form, (2) Aerofin's NQPS 1 required an annual audit of subvendors while its NQAM allowed triennial audits, and l

I (3) Aerofin did not transfer purchase order (PO) requirements to internal Aerofin documents to ensure proper implementation of requirements. RG&E accepted Aerofin's responses to its observations.

As a QA hold point, RG&E required its QA inspector to witness final hydrostatic / pneumatic testing and review final documentation before items were shipped to RG&E. RG&E's audit report indicated that Aerofin did not have a commercial-grade dedication program and expected licensees to identify safety-related items to ensure " proper procurement by Aerofin."

NSP audited Aerofin on April 5-6, 1995. The audit assessed Aerofin's  :

implementation of its QA program to the requirements of Appendix 8, as applicable to NSP's P0 #PG 9266SQ. Two findings were identified concerning Aerofin's QA of software Aerofin utilized ir, its design process and Aerofin's inadequate control of revisions to its NQPS procedures. NSP noted that Aerofin did not perform commercial-grade dedication activities; rather, it performed " upgrade" activities in accordance with ASME Code requirements. NSP also noted that Aerofin's NQPS 23, " Commercial Grade Parts," and its bill of materials pertinent to NSP's P0 identified which items were being supplied as commercial-grade. NSP accepted Aerofin's responses to its findings.

The inspector contacted seven licensees by telephone and assessed l documents that evidenced licensee monitoring of Aerofin's control of qual ity.

In a letter dated March 4,1996, to Aerofin, DE indicated that it had performed a thorough desk evaluation of both Aerofin's NQAM and NSP's audit and was concerned that Aerofin did not have a commercial-grade dedication program for non-ASMF Code parts essential to the function of cooling coil assemblies. In its letter, DE asked Aerofin to supplement its NQPS 23 with instructions to identify come s.41-grade non-ASME Code J items that could affect safety functions. Aerofin disagreed about the need for a new procedure because it did not dedicate commercial-grade items. Aerofin told DE that ASME Code materials having a pressure-retaining or structural function integral to the safety function of the coil were procured as certified materials from qualified vendors.

Aerofin also told DE that if any ASME Code materials were purchased as commercial-grade, Aerofin qualif.ad them as cerufied materials. DE l

accepted Aerofin's responses.

l 3

l 17

CP's documents indicated that CP evaluated RG&E's audit and that the audit encompassed CP's quality requirements. CP's checklist indicated that CP had reviewed Aerofin's commercial-grade dedication activities.

CP performed a detailed surveillance in July 1995 to verify whether Aerofin had manufactured the cooling coils in accordance with CP&L's P0 G0109437. Five deficiencies were identified: (1) certified material test reports referenced incorrect ASME Code years, (2) a drawing referenced incorrect material, (3) lack of a procedure for Aerofin's tube rolling process, (4) certain header plate bolts were not being torqued, and (5) a cooling coil support plate was damaged at threaded cor.nections . Deficiencies were attributed primarily to the lack of experience of certain personnel on the shop floor. CP accepted Aerofin's responses to its findings.

SN's documents indicated that SN had evaluated NSP's audit, reviewed whether NSP's audit satisfied the requirements of SN, and recommended retaining Aerofin on SN's list of approved suppliers. Upon the basis of the documents provided, SN did not perform any surveillance of Aerofin.

Desk evaluations were performed annually 1994-1996 for pertinent Aerofin noncompliances, NRC informat W , Part 21 reports, and changes to Aerofin's product line, quality program, procedures, facilities, and personnel. SN's checklist inoicated that Aerofin did not dedicate commercial-grade items.

CP&L's documents indicated that CP&L evaluated RG&E's audit based on activities described in CP&L's QA Procedure VEQ-008 Revision 2, and that CP&L determined that RG&E's audit encompassed CP&L's quality requirements. Documents did not include or refer to details of the licensee's review of the RG&E audit, nor did th;y address Aerofin's lack of commercial-grade dedication. CP&L performed a surveillance on March 31, 1994, to witness a hydrostatic test, inspect welds, review calibration records and personnel certifications, and examined Aerofin test reports and certificates of conformance (C0Cs). CP&L's P0 7LE323AA was completed in April 1994. Aerofin was dropped from CP&L's list of approved suppliers for lack of active P0s.

DP's documents indicated that DP had evaluated NSP's audit and recommended retaining Aerofin on DP's list of approved suppliers.

Documents did not include any details of DP's review of the NSP audit or indicate whether DP had determined that the audit encompassed DP's quality requirements. DP stated that its evaluation of the NSP audit wa:. based on its Procedure NPP-400, Revision 1, " Evaluating Suppliers

d the Approved Supplier List," but the documents provided by the licensee to the inspector did not refer to this procedure. In addition, the documents did not address Aerofin's lack of commercial-grade dedication. DP performed two surveillances in June 1993 and witnessed hydrostatic tests, reviewed personnel certifications, and examined Aerofin test ree-ts. DP's P0 E33328-K5 was completed in July 1994.

DL's documents w .ed that DL had evaluated NSP's audit and reviewed whether the NSP k satisfied the requirements of DL. DL performed a surveillance of ' - .n in July 1996 and issued a detailed inspection 4

18

report. Inspection attributes included workmanship, QA program, purchases from subvendors, welding, brazing, testing, and documentation. i DL's inspection checklist did not address Aerofin's lack of commercial-grade dedication. I The Aerofin QA director stated that Aerofin (purposely) did not dedicate commercial-grade items and that licensees needed to review Aerofin's NQPS 23 and bill of materials to ensure that commercial-grade items ,

being supplied by Aerofin did not adversely affect plant safety  !

functions. For example, the header plate gasket was supplied-as  ;

commercial-grade even though the gasket facilitated mating of pressure- 1 I retaining surfaces. The inspectar asked she QA director whether the l l

heat exchanger was qualified to perform in the environmental conditions identified in DE's procurement design specifications 3071-542 and whether degradation of.the header plate gasket due to pressure, temperature, or radiation could affect the pressure-retaining function j

of the ASME Code header plate. Aerofin provided a brief analysis addressing DE's specified pressure and temperature conditions and

concluded that the environmental conditions would not affect the

! operation of the heat exchanger for these parameters. The analysis did not address potential degradation due to radiation- The QA director .

believed that leakage would be insignificant and stated that "no licensee had ever complained about gasket leakage" but could not confirm the amount of potential leakage. He also stated that Aerofin l

recommended that licensees replace the gasket every time the header plate was removed but could not provide a formal document indicating that licensees were informed of this recommendation.

l The inspector contacted five licensees regarding dedication of the header gasket. DE stated that the heat exchanger header plate gasket performed a safety-related function and it dedicated the gasket in March 1996 for its plant environmental conditions. RG&E stated that it purchased a non-gasketed header plate heat exchanger (which used welds l instead of the gasket). Three other licensees stated that the gasket l was non-safety-related and that it would not affect the pressure-

retaining function of the header plate. The inspector observed that DE performed material upgrades verification for non-ASME Code safety-related parts. However, there was insufficient evidence available during this inspection for the inspector to determine whether other l licensees performed pertinent dedication of all the commercial-grade items supplied by Aerofin.

The Aerofin QA director stated that no stop-work order (fm a severe i nonconformance) was ever imposed by licensees on Aerofin.

c. Conclusions In general, licensee monitoring of Aerofin's quality assurance program for safety-related heat exchangers was not always satisfactory. Five licensees did perform effective audit and surveillance activities.

however, on the basis of licensee dxuments, two licensees did not address details of licensees' monitoring of third-party audits of Aerofin and three licensees did not address Aerofin's lack of I

dedication of commercial-grade items for safety-related heat exchangers.

! 5 19

3.2 Quality A surance Proaram I

a. Inspection Scoce 1

I The inspector examined Aerofin's QA program, policy, implementing procedures, and management directives. The inspector assessed Aerofin's l

conformance to procurement documents, evaluation and corrective actions l

in response to licensees' audit findings, commercial-grade item dedication, Part 21 evaluations, mcnitoring of subvendors, responses to l pertinent NRC audits, information notices and bulletins, and self-assess 4..ent of performance.

b. Observations and Findinos The inspector reviewed Aerofin's manufacturing activities, including witnessing liydrostatic testing of a heat exchanger, and installation of cooling fins on heat exchanger tubes. The inspector examined markings and shop documents of safety-related header plates being manufactured in the shop, and segregation of safety-related materials in holding areas.

The QA director stated that Aerofin " upgraded" materials to meet ASME Code requirements under the provision of NCA-3800 but did not dedicate commercial-grade items (see Section 3.1.b of this report). The inspector asked the QA director whether Aerofin informed licensees that non-ASME Code parts (e.g., the header plate gasket) were supplied as commercial-grade for safety-related heat exchangers. The QA director stated that Aerofin's NQPS 23 identified the commercial-grade items used in heat exchangers, that licensees were provided a detailed bill of materials identifying the gasket as commercial-grade, and that Aerofin started production only after the licensee approved the bill of materials, j

The inspector assessed Aerofin's organization and responsibilities. The QA staff comprised the QA director, the manager, and three inspectors.

The QA director and manager reported to the Manager of Nuclear and Code Products and had direct access to the President. The QA director stated that inspectors had the authority to stop production of a nonconforming item until the nonconforming conditions were corrected. Aerofin's senior management supported its QA program and remained involved in the implementation of the process. For example, senior management funded and sponsored the creation of an ISO 9001 quality system program to further trai staff in improving quality in their areas of responsibility.

The inspector observed that no guidance existed for instances in which the QA manager could supersede a QA inspector's finding and approve continuation of the work. The QA director stated that management could only resolve QA findings through written disposition of the nonconformance report (NCR). The inspector interviewed one QA inspector and four.d no problems reg:Pding the employee's motivation to identify nonconformances.

l 6

20

The inspector reviewed selected Aerofin NCRs to determine whether adequate actions were taken to correct defects or weaknesses found by Aerofin quality inspections or licensee audits. The inspector reviewed NCR 74512 regarding 13 findings identified during an internal audit and j determined that disposition of the NCR was adequate. ,

l The inspector assessed Aerofin's measures for resolving pertinent I component defects or failures of item purchased by Aerofin from  !

subver. dors . Aerofin's NQAM Section 16 required the QA manager to l identify significant and recurring deficiencies and the need for corrective action, and to determine the cause and course of action to  !

preclude recurrence of the deficiencies. The inspector asked if trending of failures was performed by the QA department to evaluate root causes so as to prevent recurrence of the failures. The QA director stated that trending was not currently necessary because Aerofin had a low volume of nuclear work and its products were standardized, had minor failures (e.g., shippiig damage or normal wear and tear) and few NCRs.

He stated that NCRs were scanned on the computer screen for repetitions I before a new nonconformance was recorded. The inspector observed that nonconformances were documented in Aerofin's database. The QA director identified one instance in which one of four subvendors had a tendency to furnish undersized material and stated that this subvendor had been removed from Aerofin's list of approved suppliers.

The inspector examined C0Cs for items purchased by DL. The C0Cs attested that purcha:,ed items were proc ' sed in accordance with the NQAM, material specifications as de/ined in the ASME Code, and requirements and drawings of the P0. For example, the C0C for P0 D145184 for cooling coils manufactured and tested in accordance with the 1974 ASME Code Section III, Class 3, Subsection ND and NF, 1974 Winter Addendo, was validated by Aerofin Test Records 960322-TR-001 and 960322-TR-002, dated June 16, 1996. j The inspector asked if Aerofin had found any design errors or failures i during the past 3 years that required the issuance of Part 21 reports. l Aerofin reported one subvendor material control deficiency in accordance l with 10 CFR Part 21 regarding deficient bolting supplied by Consolidated Power Supply for a PO from CP.

The inspector asked whether nonconformances were transmitted to the licensee, for example, whether the licensee was routinely informed of .

nonconform nces dispositioned as "use as is" or " repair." The QA director stated that NCRs pertinent to contract documents listed in NQAM Section 17, Figure 17-1, were transmitted to the licensee.

The inspector observed that Aerofin conducted effective audits of its subvendors for implementation of 10 CFR Part 50, Appendix B, criteria and 10 CFR Part 21. Audit findings were documented on detailed checklists and based on proper criteria. The auditor was adequately trained and qualified. The inspector observed that the audit reports, in general, indicated acceptability of its subvendor's implementation of Appendix B criteria, 10 CFR Part 21, and ASME Code Section III.

7 21

Subvondors were tracked on a qualified approved vendors list. The inspector observed that the list indicated audits were "not required" for several subvendors. The QA manager stated that these audits were not required by the ASME Code, however, they were conducted for compliance to Appendix B. i The inspector assessed Aerofin's internal audit conducted in May 1996 by 4 an independent contractor. The purpose of the audit was to assess the I adequacy and implementation of At.rofin's QA program. The audit was, in general, performed in accordance with proper criteria, written procedures, and checklists. The scope of the audit ns detailed and assessed the results of Aerofin's subvendor surveillance, receipt .

inspections, NCRs, and corrective actions. Thirteen findings were identified, including problems with staff training schedules and content, P0 information released to Aerofin subvendors, information '

concerning Aerofin's approved vendor lit ',, information recorded on route cards (work packages), identification of materials in Aerofin's hold area, test results, the QA check list, the calibration status of instrumentation, the calibration tolerance of instrumentation, corrective action requests fcr <.udit findings, and closure of audit findings before verification of corrective action. Aerofin issued an NCR to address all findings and dispositioned the findings satisfactorily.

c. {pnclusions The inspector concluded that, in general, the QA manual, work l instructions, and procedures were adequate.

3.3 Entrance and Exit Meetinas In the entrance meeting on July 18, 1996, the NRC inspector discussed the scope of the inspection, outlined the areas to be inspected, and established interactions with Aerofin management. In the exit meeting on July 19, 1996, the inspector discussed his observations. I 1

j l

l l

i 8

22

l l

l f PARTIAL LIST OF PERSONS CONTACTED l

Aerofin

, Dave L. Corell, Presider.t ...d Chief Operating Officer Brian C. Elliott, Manager, Nuclear and Code Products t Sub DasGupta, Director, Quality Barry DeHart, Quality Assurance Manager Tony L. Hawks, Lead Auditor, Quality Assurance l

Licensees (contacted by telephone) l l

James Johns, Quality Service Supervisor, DL William Poteat, Sr. Procurement Engineer, CP&L ,

William Jewell, Project Engineer, CP <

Samson Blue, QA Specialist, DP  ;

l Jerry Bragg, Group Leader, DE William Moody, Senior Engineer, SN Michael Lilley, QA manager, RG&E ITEMS OPENED, CLOSED, AND DISCUSSED Opened None.

Closed None.

l 1

l 9 i

l l

23

@ *'c

. , - ?g UNITED STATES

[

j g

NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001

,o

.,,.. August 23, 1996 i

l Mr. Klaus Ullmann, Senior Vice President

& Chief Engineering Officer Mutual Boiler Division '

Arkwright Mutual Insurance Compay 225 Wyman Street P.O. Box 9198 Waltham, MA 02254-9198 I

'UBJECT: NRC INSPECTION REPORT 99901296/96-01 AND NOTICES OF VIOLATION AND i

NONCONFORMANCE

Dear Mr. Ullmann:

2 On June 18, 1996, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection of Arkwright Mutual Insurance Company in Waltham, Massachusetts.

The enclosed report presents the results of that inspection.

During this inspection, the NRC inspectors found that certain of your activities appeared to be in violation of NRC requirements. Specifically, the l 3 NRC inspectors determined that your procedures for implementing the provisions i

! of Part 21 of Title 10 of the Code of Federal Reaulations, " Reporting of Defects and Noncompliance" (10 CFR Part 21) did not adequately define conditions that need to be evaluated for potential reportability or provide for the evaluation of deviations which may result from Arkwright Mutual Insurance Company's activities at nuclear facilities or other covered entities. The violation is of concern because your 10 CFR Part 21 implementing procedures appear to limit the conditions subject to evaluation and, therefore, do not provide adequate assurance that all defects or failures to comply would be reported as required by this regulation.

This violation is cited in the enclosed Notice of Violation (NOV), and the circumstances surrounding the violation are described in detail in the enclosed report. Please note that you are required to respond to this letter and should fc!10w the instructions specified in the enclosed NOV when preparing your response. The NRC will use your responte, in part, to determine whether further enforcement action is necessary to ensure compliance with regulatory requirements.

In addition, the NRC inspectors found that the implementation of your quality assurance program failed to meet certain NRC and industry requirements imposed on you by your customers. Specifically, your quality assurance program did not comply with the applicable requirements of 10 CFR Part 50, Appendix B with ,

respect to organizational independence and separation from cost and scheduling i responsibilities of personnel designated to perform quality functions. The  ;

inspectors concluded that, in at least one instance, this program deficiency I caused quality and potential safety issues to be unduly influenced by cost and j schedule considerations. Nonconformances with the requirements of 10 CFR Part i 24

K. Ullman  !

, 50, Appendix B were also identified in the anas of training and l indoctrination of personnel and document control and distribution. The inspectors also identified instances where the implementation of your quality assurance program failed to fully comply with American Society of Mechanical Engineers (ASME) Code requirements that are applicable to your activities i

! under the scope of your ASME Certificate of Accreditation.

These nonconformances are cited in the enclosed Notice of Nonconformance (NON), and the circumstances surrounding them are described in detail in t:-  !

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 l Room.

i Sincerely, QN M M'C Robert M. Gallo, Chief Special Inspection Branch Division of Inspection and Support Programs Office of Nuclear Reactor Regulation Docket No. 99901296

Enclosures:

1. Notice of Violation

2. Notice of Nonconformance

3. ; spection Report 99901295/96-01 l

l i

I l

25

l NOTICE OF VIOLATION i

i Arkwright Mutual Insurance Company Docket No.: 99901296 Waltham, MA.

I i

During an NRC inspection conducted on June 17 and 18, 1996, a violation of NRC requirements was identified. In accordance with the " General Statement of Policy and Procedure for NRC Enforcement Actions," NUREG-1600, the violation is listed below.

Paragraph 21.21, " Notification of failure to comply or existence of a defect l and its evaluation," of Title 10 of the Code of Federal Reaulations (10 CFR),

Part 21 requires, in part, that each corporation subject to the regulations adopt appropriate procedures to ensure the evaluation and proper reporting of l deviations and failures to comply.  !

Contrary to the above, Section 1.5, " Nuclear Regulatory Commission (NRC)  :

Document" of Arkwright Mutual Insurance Company's (Arkwright) Quality Assurance Manual, Third Edftion, Revision 2, dated October 18, 1995, did not:

(1) adequately define conditions that need to be evaluated for 10 CFR Part 21 notification (deviations) or (2) provide for the evaluation of deviations l which may result from Arkwright activities at nuclear facilities or other ,

covered entities. (99901296/96-01-01) i This is a Severity Level IV violation (Supplement VII). I Pursuant to the provisions of 10 CFR 2.201, Arkwright is hereby required to submit 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 Violation. This reply should be l clearly marked as a " Reply to a Notice of Violation" and should include for l each violation: (1) the reason for the violation, or, if contested, the basis for disputing the violation, (2) the corrective steps that have been taken and ,

the results achieved, (3) the corrective steps that will be taken to avoid  !

further violations, and (4) the date when full compliance will be achieved. l Your response may reference or include previous docketed correspondence, if .

the correspondence adequately addressas the required response. Where good cause is shown, consideration will be given to extending the response time.

1 Dated at Rockville, Maryland this 23rd day of August, 1996 Enclosure 1 26

NOTICE OF NONCONFORMANCE 1

Arkwright Mutual Insurance Company Docket No.: 99901296 l Waltham, MA ,

Based on the results of an inspection conducted on June 17 and 18, 1996, it appears that certain of your activities were not conducted in accordance with -

NRC requirements. (

A. Criterion I, " Organization," of Appendix B to Part 50 of Title 10 of the Code of Federal Reaulations (10 CFR), requires, in part, that the persons and organizations performing quality assurance (QA) functions be provided authority and organizational freedom including sufficient j independence from cost and schedule when opposed to safety >

considerations. I Contrary to the above, the Arkwright Technical Services. Inc., Assistant ,

Vice President, who'was charged with establishing and maintaining i Arkwright Mutual Insurance Company's (Arkwright's) QA program, was also  ;

responsible for cost and scheduling considerations related to providing (

inspection services. The inspectors determined that, in at least one instance, the Assistant Vice Preitdent's lack of independence from cost and schedule adversely influenced decisions with potential safety significance. (Nonconformance 99901296/96-01-02)  :

! B. Criterion I, "Organi ation," of Appendix B to 10 CFR Part 50, requires, j

in part, that the authority and duties of persons and organizations performing activities affecting the safety-related functions of structures, systems, and components (including personnel performing QA  ;

functions) be clearly established and delineated in writing. j Contrary to the above, Arkwright's " Quality Assurance Manual for ]

Authorized Inspection Agency Inspection Services," Third Edition,  ;

Revision 2, October 18, 1995 (QA manual) did not clearly delineate _the  !

authority and duties of persons performing QA functions that, in part, l l ensure inspection activities comply with the requirements of American f l Society of Mechanical Engineers (ASME) standard N626, " Qualifications and Duties for Authorized Nuclear Inspection Agencies and Personnel,"

l and are commensurate with its ASME authorized inspection agency cer .ficates of accreditation.

In addition, the QA program failed to establish the authority and duties j of Arkwright's Mutual Boiler Division, as it appeared in the ASME l authorized inspection agency certificates of accreditation.

(Nonconformance 99901296/96-01-03) i C. Criterion V, " Instructions, Procedures, and Drawings," of Appendix B to 10 CFR Part 50, requires, in part, that activities affecting quality be prescribed by documented instructions or procedures and be accomplished in accordance with these instructions.

! Enclosure 2 27

Paragraph NCA-5220, Categories of Inspector's Duties," of Subsection  !

NCA of ASME Code Section III, requires, in part, that the authorized '

nuclear inspector perform duties as required in ASME N626b-1992.  ;

Paragraph NCA-5290, " Certification of Data Reports and Construction }

Reports," of Subsection NCA of ASME Code Section III requires, in part, ,

that the appropriate manufacturer's data reports be certified by the l authorized nuclear inspector after being satisfied that all requirements i of ASME Code Section III have been met. l Section 0-3, "The Authorized Nuc'e:.? Inspector," Subsection 0-3.2,  !

" Duties," Paragraph 0-3.2.18 of ASME N626b-1992, requires, in part, that l the authorized nuclear inspector keep a bound diary of activities and i inspections made. 'The information to be recorded shall include a  ;

description of the~ item inspected, the type of observation made, the j i

requirements that prompted the activity, and the results of inspection.  ;

Contrary to the above, Arkwright failed to establish measures in QAP.07,  !

"ASME Code Inspection Activity - Authorized Inspector (A!/ authorized nuclear inspector)," Revision 1, June 1996, e

that the authorized nuclear inspector document the inspection  :

results, the type of observations made, and the requirements that i prompted the activity  ;

that QA monitoring reports issued to manufacturers during the course  ;

of inspection activities be satisfactorily addressed thus ensuring that all requirements of the ASME Code have been met before the  ;

authorized nuclear inspector certified the ASME manufacturer's data '

report (Nonconformance 99901296/96-01-04)

D. Criterion II, " Quality Assurance P ogram" of 10 CFR Part 50, Appendix B states,-in part, "The program shall provide for indoctrination and training of personnel performing activities affecting quality as necessary to assure that suitable proficiency is achieved and j maintained. l Criterion XVII, " Quality Assurance Records" of.10 CFR Part 50, Appendix  ;

B states, in part, " Sufficient records shall be maintained to furnish '

evidence of activities affecting quality. The records shall include at  :

least the following: Operating logs ... and material analyses. The  !

records shall also include closely-relateJ data such as qu ifications of personnel, procedures, and equipment." -

Section 8.0, " Indoctrination and Training," of the Arkwright QA manual states, in part, in Paragraph 8.3 that all indoct.ination and training activities shall be documented and that all documentation is required to be maintained on file.

Contrary to the above, Arkwright tri..ning records did not include previous Factory Mutual Engineering Association (FMEA) ASME training records for those employees who had transferred from FMEA employment to 2

28

l Arkwright employment, when Arkwright assumed responsibility for the inspection sites at which they were assigned. (99901296/96-01-05) t E. Criterion VI, " Document Control" of 10 CFR Part 50, Appendix B states, in part, that measures shall be established to control the issuance of '

documents which prescribe activities affecting quality, and that these measures shall assure that documents, including changes, are approved and distributed to, and used at, the location where the prescribed activity is performed.

Sectior. 10, "The Quality Assurance 'anual," of the Arkwright. QAM, states, in Paragraph 10.3.1 that "the QA manual shall be implemented by all personnel involved with nuclear inspection activities," and also states, in part, that the Managing Supervisor of Inspection Services shall maintain a list of recipients of controlled manuals.

Contrary to the above, the Arkwright QA Manual, that had been approved and issued on October 18, 1995, was not distributed to inspection personnel for implementation until June 1996. Also, the list of the Distribution of Controlled Copies of the Arkwright Quality Assurance Manual was not up to date and did not indicate that the QAM had ever been sent out to inspection personnel. (99901296/96-01-06)

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 Nor.conformance" 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.

Dated at Rockville, Maryland this 23rd day of August, 1996 ]

1 l

l 3

29

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

Report No: 99901296/96-01 Organization: Arkwright Mutual insurance Company 225 Wyman Street 3

P.O. Box 9198 Waltham, MA 02254-9198 .

l l

Contact:

Cont.d M. D'Esopo, President Arkwright Technical Services, Inc. i Nuclear Industry i Activity: ASME accredited authorized inspection agency Dates: June 17 - 18, 1996 Inspectors: Uldis Potapovs, Senior Reactor Engineer Richard McIntyre, Senior Reactor Engineer Steven Matthews, QA Specialist )

Approved by: Gregory C. Cwalina, Section Chief Vendor Inspection Section Special Inspection Branch  ;

Division of Inspection and Support Programs Office of Nuclear Reactor Regulation  :

1 1

4 Enclosure 3 30

i 1

1 INSPEC(ION

SUMMARY

During this inspection, the NRC inspectors reviewed the implementation of j selected portions of the Arkwright Mutual Insurance Company's (Arkwright's) l quality assurance (QA) program and reviewed activities associated with the implementation of Arkwright's inspection services contract with Amer Industrial Technologies, Inc. (Amer).

l The inspection bases vere:

l l

• Appendix B, " Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessi19 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" American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code)

• ASME N626-1990, " Qualifications and Duties of Authorized Nuclear Inspection Agencies and Personnel."

• Arkwright's QA Manual for Authorized Inspection Agency Inspection Services, Third Edition, Revision 2, dated October 18, 1995.

During this inspection, a violation of NRC requirements was ident',fied and is discussed in Section 3.1 of this report.

The inspectors also identified five instances where Arkwright failed to conform to NRC requirements imposed upon them by NRC licensees or to ASME Code requirements imposed by the ASME as a condition for accreditation as an authorized inspection agency. These nonconformances are discussed in Sections 3.2.b, 3.2.c, 3.3.b, 3.4, and 3.5 of this report.

2 STATUS OF PREVIOUS INSPECTION FINDINGS This was the first NRC inspection of Arkwright 3 INSPECTION FINDINGS AND OTHER COMMENTS 3.1 10 CFR Part 21 Proaram The inspectors reviewed Arkwright's procedures for implementing the provisions of 10 CFR Part 21 for reporting defects and noncompliance:

QA Manual, Section 1.5 and Quality Assurance Procedure (QAP).11, which was revised in June, 1996. Section 15 of the QA manual requires all personnel who become aware of an adverse condition to bring it to the '

attention of management at the location and, if the management is not aware or fails to acknowledge that an adverse condition exists, to

, inform their (Arkwright) supervisor. It further states that the Assistant Vice President of Arkwright Techc.ical Services will discuss l

l 2 l

31

i the adverse condition with the President of Arkwright Technical Services (Vice President of Arkwright) who will decide on the course of any further action by Arkwright. QAP.ll states that the' President, ,

Arkwright Technical Services (Vice President Arkwright) shall be responsible for determining whether any defect or failure to comply ,

meets the reporting i,..;uirements of 10 CFR Part 21 and, if so, to report  !

to the NRC.

The inspectors noted that the procedure, contained in Section 1.5 of the QA Manual, used terminology (adverse condition) which is not defined in the n.anual or in 10 CFR Part 21. 1.' CFR Part 21 requires entities subject to this regulation to adopt procedures to evaluate deviations and failures to comply potentially u sociated with a substantial safety .

hazard and to assure that a director or responsible officer subject to the regulation is promptly notified if this evaluation determines that t the condition (s) constitute a failure to comply or a defect. These terms are explicitly defined in the regulation. The inspectors i determined that failure to adequately define conditions which are subject to evaluation under this regulation could contribute to potential non-reporting of defects or failures to comply.

The inspectors also noted that Arkwright's 10 CFR Part 21 procedure did not require the evaluation of deviations or failures to comply when these conditions are the direct result of services that Arkwright is providing to the manufacturers of basic components (ASME certificate holders). For example, the procedure did not address situations where the authorized nuclear inspector's inspection activities are found to be in noncompliance with the technical requirements of the purchase documents after the performance of the contracted services. An example would be an after-the-fact identification that an authorized nuclear inspector providing services to a ASME certificate holder or a nuclear plant lacked the required qualifications. L 1

Failure to adopt appropriate procedures to evaluate deviations or failures to comply as required by 10 CFR Part 21 constitutes Violation 99901296/96-01-01.

3.2 Ouality Assurance Proaram

a. Insoection Scone Arkwright's QA program is described in their QA Manual for Authorized Inspection Agency Inspection Services and several QAP's which provide additional guidance for program implementation. The manual attests compliance with the applicable Sections of the ASME Code and ASME QAl-1,

" Qualifications for Authorized Inspection." Although the quality program requirements of 10 CFR Part 50, Appendix B and applicable cr'teria of American National Standards Institute (ANSI) standard'N45.2

.d associated daughter standards h we been invoked by several utility procurement documents, Arkwright's QA Manual did not address or indicate compliance with these documents. The QA program review was focused on Arkwright's activities related to inspection services agreement with 3

32

l l

Amer that cas administered through Factory Mutual Engineering I Association (Factory Mutual), partially owned by Arkwright.

}

! Arkwright has been accredited by ASME :s an authorized inspection agency in accordance with ASME N626, " Qualifications and Duties for Authorized l

Nuclear Inspection Agencies and Personnel," 1990 edition and addenda through N626b-1992. Arkwright held two ASME authorized inspection agency certificates of accreditation. The first allowed Arkwright to provide inspection services directly as an authorized inspection agency  !

l and read as follows:

I l Arkwright Mutual Insurance Company Mutual Boiler Division l 225 Wyman Street l

Waltham, Massachusetts 02254-9198 j

The second certificate allowed Arkwright to provide inspection services by doing business as Factory Mutual and reads as follows:

Arkwright Mutual Insurance Company Mutual Boiler Division dba Factory Mutual Engineering Association 225 Wyman Street Waltham, Massachusetts 02254-9198 l

As required by Section 0-1, "The Authorized Inspection Agency," of ASME N626b-1992, Arkwright established and implemented an internal program to ensure that the authorized nuclear inspector supervisor and authorized nuclear inspector perform their duties in accordance with ASME N626 and that levels of inspection activity are commensurate with the scope of the ASME authorized inspection agency certificates of accreditation.

The team was told by Arkwright that the program was documented in Arkwright's QA manual and in QAPs used to ensure adequate levels of inspection activities by Arkwright field personnel. The team was also told by Arkwright that the QA manual described Arkwright's organization and its involvement with Factory Mutual in sufficient detail to address both ASME authorized inspection agency certificates of accreditation. ,

l

b. Observations and Findinas Oraanization 1 The Arkwright QA manual contained the following statements regarding delegation of the authority and responsibility for the QA program:

= Statement by the Arkwright President delegating the authority

' and responsibility for the QA program to the Arkwright Senior Vice President and Chief Engineering Officer.

• Statement by the Arkwright Senior Vice Pre.,ident and Chief l

Engineering Officer delegating the authority and responsibility for the QA program to the President, Arkwright 4

33

Technical Services, Inc., who is also an Arkwright Vice President. (Arkwright Technical Services is a wholly owned subsidiary of Arkwright.)

Statement by the Arkwright Technical Services President delegating the authority and responsibility for the QA program to the Arkwright Technical Services Assistant Vice President.

Additionally, during an interview, the Arkwright Technical Services Assistant Vice President told the team that he had delegated the authority and responsibility foi the QA program to the Managing Supervisor - Inspection Services, who is an Arkwright employee.

However, the team found that the QA program failed to describe the Arkwright Technical Services Assistant Vice President'_s delegation of the authority and responsibility for the QA program to the Arkwright Managing Supervisor - Inspection Services.

The team also found that the QA program failed to describe, establish, or delineate the responsibilities and duties of the Mutual Boiler Division of Arkwright, as it appears in both ASME authorized inspection agency certificates of accreditation. Moreover, the QA program failed to describe Arkwright's interfaces with both Arkwright Technical Services and Factory Mutual and their respective authority and duties for the functions of attaining quality objectives and quality assurance.

i for instance, Section 2.1.3 of the QA manual prescribed, in part, that l

Arkwright delegates to Arkwrigh'. Technical Services the responsibility for administering contracted inspection services and for implementing l the QA program. The same section stated that inspection services were i performed by_ Arkwright or Factory Mutual and that Factory Mutual's QA

~

manual was a supplement to Arkwright's QA manual. However, as required by Criterion I, " Organization," of Appendix B to 10 CFR Part 50 and ASME l N626b-1992, the QA program failed to describe how Arkwright' retains responsibility for either the Arkwright Technical Services or Factory Mutual QA programs or how it verifies that activities affecting safety-related functions (e.g., authorized nuclear inspector third-party {

oversight of manufacturers supplying safety-related components to NRC

licensees) are correctly performed. These findings constitute 3 Nonconformance 99901296/96-01-03.

Inspector's Duties Section 6, " Inspection," of the QA manual requireo that the inspection activities prescribed in QAP.07 be documented in the authorized nuclear inspector's bound diary. These documentation requirements were imposed on Arkwright through its ASME accreditation as an authorized inspection agency in accordance with ASME N626-1990 and Paragraph NCA-5220,

" Categories of Inspector's Duties," of ASME Code Section III.

To assess the adequacy of Arkwrignt's inspection activities, the team reviewed Arkwright QAP.07, "ASME Code Inspection Activity - Authorized 4

Inspector (AI/ANI)," Revision 1, June 1996. From its assessment, the I

5 34 e F-m'

-t-.M y--,e--+--+++" -----M 'T,--F emt-Es =f -- & - + - , ar' -7 ~p wu - r- s-- m s tr -

team determined that Subsection P, " Inspector's Diary," of Section IV,

[

" Inspection Activities," of QAP.07 failed to require that the authorized nuclear inspector document the inspection results, the type of observations made, and the requirements that prompted the activity as required by ASME N626-1990.

Additionally, from its evaluation of Arkwright's program requirements for authorized nuclear inspectors, the team noted that the authorized l

nuclear inspectors issued QA monitoring reports to manufacturers during i the course of inspection activities. However, the team's evaluation of QAP.07 showed that it failed to equire that all monitoring reports be satisfactorily addressed thus ensuring that all requirements of the ASME Code have been met before the authorized nuclear inspector certified the ASME manufacturer's data report. Therefore, open deficiencies may exist at the time the authorized nuclear inspector certifies the manufacturer's data report signifying that all ASME Code requirements had been met. The team also found that QAP.07 failed to establish measures to control the issuance, distribution, and use of monitoring l reports..

Arkwright's failure to establish measures (1) that the authorized nuclear inspector document the inspection results, the type of observations made, and the requirements that prompted the activity and (2) that all deficiencies identified on monitoring reports were satisfactorily addressed before the authorized nuclear inspector certifies the manufacturer's data report constitutes Nonconformance 99901296/96-01-04,

c. Q;nclusions The team concluded that the cascoding delegations described above and in Arkwright's QA manual resulted in the ASME-accredited authorized inspection agency (Arkwright) delegating the authority and responsibility for the QA program to Arkwright Technical Services, who eventually delegated it back to a staff-level Arkwright employee.

Arkwright's QA manual placed the responsibility for implementation of the QA program on Arkwright Technical Services, a wholly owned subsidiary, and charged an Arkwright Technical Services Assistant Vice President with the authority and responsibility for establishing and maintaining Arkwright's QA program. However, on the basis of interviews of Arkwright and Arkwright Technical Services staff, the team determined that Arkwright's QA program was established and maintained by the

....g:ng Supervisor - Inspection Services, an Arkwright employee.

Although authority and responsibility for the QA program were delegated through several layers of Arkwright and Arkwright Technical Services management to an Arkwright staff-level position, the QA program failed to describe how Arkwright retained responsibility for the QA program and how it verified that activities n'fecting inspection activities and safety-related functions were correctly performed. This chain of 6

l 35

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

delegated authority appeared to the team to obscure Arkwright's responsibility for the QA program as the ASME-accredited authorized inspection agency.

The team was also concerned that neither ASME authorized inspection agency certificates of accreditation allowed Arkwright to do business as Arkwright Technical Services, its wholly owned subsidiary.

Arkwright's failure to clearly establish the responsibility and duties of organization: performing activities affecting the safety-related functions of nuclear components i.. a.eir QA manual was cited as Nonconforniance 99901296/96-01-03.

The team also concluded that Arkwright's requirements with regard to  :

entries in a bound diary by the authorized nuclear inspectors did not '

meet the requirements of Criterion V, " Instructions, Procedures, and Drawings" of Appendix B to 10 CFR Part 50, ASME Code paragraph NCA-5220, and ASME N626b-1992. Failure to establish measures that meet the applicable r'quirements and duties for authorized nuclear inspectors constitutes Nonconformance 99901296/96-01-04.

3.3 Arkwricht Involvement with Factory Mutual I

a. Insoection Scoce During the NRC inspection of Amer, from January 29 through February 2, 1996 (NRC Inspection Report 99901292/96-01, March 21, 1996), the team evaluated four jobs comprising nuclear components shipped to NRC licensees. All four jobs had been inspected by authorized nuclear inspectors employed by Factory Mutual, and for each job an ASME manufacturer's data report had been signed by Amer and certified by the authorized nuclear inspector indicating that Amer had complied with all applicable requirements of ASME Code Section III. Although manufacturer's data reports for all four jobs stated that Amer had met the ASME Code requirements, the NRC inspection of Amer showed that none of the jobs evaluated completely met all applicable requirements of the ASME Code. On the basis of the NRC findings and other concerns raised i

during its inspection of Amer, the NRC conducted a followup inspection of Factory Mutual's inspection services activities.

,* During the followup inspection at Factory Mutual's district office in Bala-Cynwyd, Pennsylvania, and its home office in Norwood, l Massachusetts, from April 22 through 26, 1996 (NRC Inspection Report l

99901296/96-01, May 28, 1996), the team found that the authorized nuclear inspector inspection services contract for Amer was actually issued by Arkwright. (The ASME authorized inspection agency certificate of accreditation that allows Factory Mutual to provide authorized nuclear inspector inspection services is described in Section 3.2 of this report.)

During its inspection of Factory Mutual, the team determined that since the early part of 1993, factory Mutual's authorized nuclear inspectors 7

I 36 i

assigned to Amer, the Factory Mutual district office in Bala-Cynwyd,

, Pennsylvania, and the Factory Mutual home office in Norwood, Massachusetts, had repeatedly requested that Arkwright cancel the inspection services contract with Amer. These requests were largely based on Factory Mutual's assessments of Amer's technical performance, 1 which showed that Amer consistently failed to comply with ASME Code requirements. For instance, in its " Annual Shop Activity Report" for 1993 and 1994 (December 10, 1993, and December 21, 1994, respectively),

Factory Mutual recommended that the inspection services contract with Amer be canceled because Amer had routinely failed to comply with ASME

, Code rc auirements and several authL ized nuclear inspector supervisor audits of Amer had shown that Amer's activities were not in compliance with its own quality procedures. Additionally, Factory Mutual stated that continuing its inspection relationship with Amer would result in jeopardizing Factory Mutual's overall credibility and good reputation as an authorized inspection agency as well as the commissions of its authorized nuclear inspector supervisor and authorized nuclear inspectors.

However, despite several requests by Factory Mutual to cancel the Amer inspection services contract, A/kwright's management did not terminate this contract until September 1995. Arkwright's failure to take this action sooner prompted the tram to question Arkwright's involvement with Factory Mutual and its role in controlling the authorized nuclear inspector inspection services provided by Factory Mutual to 4

manufacturers of nuclear components.

The Arkwright Vice President (President of Arkwright Technical i Services), who was present during the Factory Mutual inspection exit meeting, stated that he had joined Arkwright Technical Services in January, 1995 and, therefore, had little information about the administration of the inspection services contract with Amer. He also told the team that he had no experience with the ASME Code before i joining Arkwright Technical Services and that those responsibilities were delegated to the Arkwright Technical Services Assistant Vice President.

Since the inspection of Factory Mutual did not produce conclusive answers to NRC concerns regarding the involvement of Arkwright's management in Factory Mutual's inspection services, the team examined these issues further during this followup inspection of Arkwright.

b. Observations and Findinas The authorized nuclear inspector's responsibilities, as described by the ASHE Code, are to provide third-party oversight of an ASME certificate  ;

holder (manufacturer) assumed to be conscientiously trying to comply 1 with the requirements of the ASME Code. However, the Factory Mutual files reviewed during the NRC inspection of Factory Mutual showed that the normal assumptions and principles of third-party oversight did not 1 apply at Amer. The documentation and interviews with the authorized i

nuclear inspector supervisor and authorized nuclear inspectors showed 8

37

l l

that from 1993 through 1995, the authorized nuclear inspectors had to l exercise extreme diligence (over and above that routinely expected of 1 third-party oversight) in the performance of their inspections and verifications because Amer's failure to comply with the ASME Code requirements was pervasive in Amer's fabrication activities.

After reviewing Factory Mutual's re .ts, records, and authorized i nuclear inspector bound diaries and interviewing the authorized nuclear I inspectors, the team determined that the authorized nuclear inspectors had very little confidence in Amer's compliance with ASME Code requirements. Because Amer's failure to comply with these requirements was so prevalent, normal authorized nuclear inspector oversight activities were not adequate to establish reasonable assurance that Amer had complied with all the applicable requirements of the ASME Code for the nuclear components supplied to NRC licensees.

Factory Mutual's manag. ment reiterated to it: authorized nuclear inspectors not to certify Amer's manufacturer's data reports when they knew that the ASME Code requirements were not being met. The team recognized that the authorized nuclear inspectors are not expected to verify every detail of ine manufacturing process and, under the existing conditions, the team determined that it was not reasonable for the authorized nuclear inspectors to assume that all details that were not verified actually met the ASME Code. The NRC findings during its previous inspection of Amer (described in NRC Inspection Report 99901292/96-01) support the team's conclusion that the authorized nuclear inspectors did not verify every detail of Amer's manufacturing process and that the third party oversight process.could not be adequately implemented under the existing conditions.

The inspectors determined that, because of the conditions that existed at Amer from 1993 through 1995, the Factory Mutual authorized nuclear inspectors performing inspections at that facility were in a difficult position since the third-party oversight inspections th' 7erformed were compromised by Amer's numerous failures to comply with E E Code requirements to the extent that NRC licensees could not rely on Amer's manufacturer's data reports as evidence that nuclear components supplied by Amer fully met all the applicable requirements of the ASME Code.

To evaluate Arkwright's management of the Amer inspection services contract, the team reviewed Arkvright's records and audit reports and held seve .! interviews. The team reviewed Arkwright's 1993 and 1994 l audits of Factory Mutual performed by the Arkwright Technical Services Assirtant Vice President, who used the same audit checklirt both in 1993 and in 1994. Although Arkwright's file contained memoranda documenting Factory Ma.ual's concerns about the conditions at Amer that preceded its audits of Factory Mutual, the auditor's (Arkwright Technical Services Assistant Vice President's) 1993 and 1994 audits of factory Mutual failed to ' tentify Factory Mutual's ,rrowing concerns about the conditions at Amer or to show whether those conditions were receiving timely attention.

9 38

The teaa noted that one audit checklist question was whether the Factory Mutual home office. authorized nuclear inspector supervisor told the Factory Mutual. Vice President and Manager-of Boiler and Machinery 7

Engineering when unsatisfactory conditions noted during audits of

! manufacturers were not receiving timely attention. The auditor's responses on the audit checklist were as follows:

• For the 1993 audit of Factory Mutual, conducted October 5, 1993, the auditor answered "Yas" to the above question and noted, "however no situations to verify."

• For the 1994 audit of Factory Mutual, conducted November 18, 1994,.the auditor answered "N/A" to the above question.

The team also reviewed the notes from Arkwright's quality program management assessment meetings dated October 13, 1993, and December 16,

! 1994, and found that neither of the management assessments noted Factory Mutual's growing concerns about the conditions at Amer. The management assessments were conducted by the Arkwright Technical Services Assistant Vice President and the Arkwright Senior Vice President and Chief Engineering Officer.

l Failure to note the Amer issue in either the Arkwright's audits of Factory Mutual or management assessments discussed above appeared to l result primarily from the lack of independence that existed; that is,

! the Arkwright Technical Services Assistant Vice President managed the

! Amer it.spection services contract with the concurrence of the Arkwright l Senior Vice President and Chief Engineering Officer. Therefore, both l individuals were responsible for the Arkwright decision not to honor Factory Mutual's request to cancel the contract with Amer.

During this inspection of Arkwright, the team interviewed the Arkwright Technical-Services Assistant Vice President, who also reported having little experience with or knowledge of the ASME Code, and reviewed Arkwright's contract file including a response to factory Mutual's

! request to cancel the inspection services contract with Amer because of growing concerns about factory Mutual's capability to verify code compliance as discussed in Section 3.3.a. above. The Arkwright  ;

Technical Services Assistant Vice President responded to Factory Mutual's concerns in a_ memorandum dated August 15, 1994, by stating that Arkwright did not believe that available information justified canceling

the inspection services contract with Amer. The Assistant Vice President's response suggests inadequate understanding of the ASME Code third-party inspection process and the basic elements of this process which provide the assurance that certified manufacturer's data reports can be relied upon as evidence that nuclear componente comply with the applicable ASME Code requirements.

After reviewing the Arkwright and Arkwright Technical Services reports and record files and interviewing th: President and Assistant Vice President of Arkwright Technical Services, the team determined that the decision of the Arkwright and Arkwright Technical Services management to  !

! 10 l 4

l 39

continue the inspection services contract with Amer was based on Arkwr yht's position that Amer owed it several payments for inspection services it had previously provided and that it would pursue collection of the debt and maintain the contract with Amer until the revenue deficit was decreased (memorandum dated October 22, 1993, from the Arkwright Technical h vices Assistant Vice President to the contract file).

Section 1.2, " General Responsibilities," of Arkwright's QA manual requires, in part, that should differences of opinion between parties of this company arise that cannot be r: solved, the Arkwright Technical Services Assistant Vice President will resolve the matter in accordance with ASME Code requirements. However, after being advised by Factory Mutual in 1993 that maintaining the inspection services contract with Amer could compromise ASME Code compliance, the Arkwright Technical Services Assistant Vice President and Arkwright's Senior Vice President and Chief Engineering Officer apparently made a decision to keep this contract in effect. Inadequate separation of QA functions from cost and schedule responsibilities constitutes Nonconformance 99901296/96-01-02.

Arkwright's position remained unchanged until the Arkwright Technical Services Assistant Vice President notified Amer in a letter dated July 24, 1995, that Arkwright was issuing Amer a 60-day notice of cancellation.

c. Conclusions The team determined the Arkwright Technical Services Assistant Vice President, with the concurrence of Arkwright's Senior Vice President and Chief Engineering Officer, decided to keep in effect the inspection services contract with Amer in order to decrease its revenue deficit with Amer (memorandum dated October 22, 1993, from the Arkwright Technical Services Assistant Vice President to the contract file). This decision was made contrary to the technical information supplied by-Factory Mutual in numerous memoranda and letters reviewed by the team that continuing its inspection relationship with Amer could undermine the basic elements of the third-party inspection process, from 1993 through 1995, Arkwright Technical Services Assistant Vice

. resident and Arkwright's Senior Vice President and Chief Engineering Officer rejected Factory Mutual's request to cancel the Amer inspection services contract because its third-party-oversight of Amer was compromised by Amer's routine failure to comply with ASME Code requirements (as determined by the NRC findings documented in Inspection Report 99901292/96-01, March 21, 1996) to the extent that NRC licensees could not rely on Amer's manufacturer's data reports, signed by Factory Mutual authorized nuclear inspectors, as evidence that nuclear components supplied by Amer fully met all the applicable requirements of the ASME Code.

The team concluded that, contrary to the requirements of Criterion I,

" Organization," of Appendix B to 10 CFR Part 50, Arkwright's QA program 11 40

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

l  !

\

I failed t'o. provide adequate separation of QA and cost / schedule functions

necessary to ensure that compliance and safety decisions are not unduly i influenced by cost.or schedule considerations (Nonconformance 1

99 % 296/96-01-02)'

i l

l Additionally, the team concluded that, as accredited by ASME (i.e., i Arkwright doing business as Factory Mutual), Factory Mutual should have 1 the authority and duty described in Section 1.2, " General

. Responsibilities," of Arkwright's QA manual to resolve any differences j of opinion between parties.

3.4 Personnel Qualification and Training l f

1 The inspectors reviewed the latest revision of QA Manual Section 8,

" Indoctrination and Training," which describes the general requirements j for the indoctrination and training of inspection personnel and the more

' specific training responsibilities for the Managing Supervisor and the

Field Supervisor.

! The inspectors reviewed the indoctrination and training records for

authorized nuclear inspectors and authorized nuclear inspector i supervisor who work for Arkwright, including individuals who had previously worked for Factory Mutual prior to Arkwright assuming 4 authorized nuclear inspector and authorized nuclear inspector supervisor

duties at several Factory Mutual sites in October 1995. The current

! Arkwright training records did not include records of previous Factory j Mutual training received for those employees who had transferred from 4

Factory Mutual employment to Arkwright employment when Arkwright assumed responsibility for the inspection sites at which they were assigned.

Also, the inspector training records did not include all documentation j

. of the ASME/ National Board training and qualification for authorized nuclear inspectors and authorized nuclear inspector supervisor as required by ASME QAl-1 and the Arkwright quality program. The

inspectors also identified certain inspector training records that could l not be located in the Arkwright files and were told by Arkwright that j these training records were located at the Managing Supervisor of i Inspection Services Georgia residence, since it also serves as his field j office. Nonconformance 99901296/96-01-05 was identified during this j portion of the inspection.

The inspectors determined that, currently, Arkwright does not maintain a training matrix of job functions versus training requirements for

. authorized nuclear inspectors and authorized nuclear inspector supervisor and that, as of the inspection, no formal ongoing annual training had been conducted. The inspectors concluded that training records of Arkwright inspection personnel had not been reviewed for i completeness and appropriate actions are needed by Arkwright to assure compliance with ASME and QA Manual requirements.

I y-j 12 i

I 41

3.5 Document Contrql The inspectors reviewed the latest revision of QA Manual Section 10, "The Quality Assurance Manual," which describes the process for QA Manual review and approval, QA Manual transmittal, and QA Manual implementation and revision.

While attempting to verify the distribution of the controlled copies of the QA Manual to all the authorized nuclear inspectors and authorized nuclear inspector supervisor, the inspectors learned that the QA Manual, ,

that had been approved and issued on October 18, 1995, was not i distributed to inspection personnel until June of 1996. The inspectors also determined that the list of the Distribution of Controlled Copies of the Arkwright Quality Assurance Manual was not up to date and did not indicate that the QA Manual had ever been sent out to the authorized nuclear inspectors and authorized nuclear inspector supervisor. No entry existed in either the update sent or the date returned column. 1 The inspector verified that the QA Manual had been distributed to inspection personnel in June 1996, eight months after original approval, but since the distribution list had no record of date returned by the  !

manual holder, the inspectors could not verify manual implementation.

Arkwright stated that the QAPs, that were issued to inspectors in l September 1995, contained similar guidance on QA program implementation I as in the QAM and that is the reason for not distributing the QA Manual until June 1996. Nonconformance 99901296/96-01-06 was identified during this portion of the inspection.

3.6 Entrance and Exit Meetinas In the entrance meeting on June 17, 1996, the NRC inspectors discussed the scope of the inspection, outlined the areas to be inspected, and established interfaces with Arkwright management. In the exit meeting on June 18, 1996, the inspectors discussed their findings and concerns.

i l

l 13 42

PARTIAL LIST OF PERSONS CONTACTED Conrad M. D'Esopo, President Arkwright Technical Services, Inc. and Vice President, Arkwright Mutual Insurance Company Peter Hoefler, Assistant Vice President, Arkwright Technical Services, Inc.

Timothy B. Rhodes, Managing Supervisor - Inspection Services ITEl." OPENED 99901296/96-01-01 VIO inadequate Part 21 procedure 99901296/96-01-02 NON inadequate QA independence 99901296/96-01-03 NON inadequate description of organization 99901296/96-01-04 NON inadequate procedure for ASME Code inspection activities 99901296/96-01-05 NON inadequate training management 99901296/96/01-06 NON inadequate control of documents i

i 14 l

43

p f %e p 4 UNITED STATES

& E NUCLEAR REGULATORY COMMISSION f WASHINGTON, D.C. 20556 0001

\...../ September 4. 1996 Mr. Edgar Whittle, Manager ASME Codes and Standards l Commercial Union Insurance Company One Beacon Street Boston, MA 02108-3106

SUBJECT:

NRC INSPECTION REPORT 99900632/96-01 AND NOTICE OF NONCONFORMANCE

Dear Mr. Whittle:

l On June 21, 1996, the U.S. Nuclear Regulatory Commission (NRC) completed an I inspection of Commercial Union Insurance Company's Contract Inspection Services activities at Commercial Union's Boston offices. The enclosed report presents the results of that inspection.

During this inspection, the NRC inspectors identified several instances where ,

the implementation of your quality assurance program failed to fully comply I with American Society of Mechanical Engineers (ASME) Code requirements that are applicable to your activities under the scope of your ASME Certificate of i Accreditation. Specifically, the authorized nuclear inspector diaries did not always include the required information, the use of monitoring reports was not adequately controlled, and quality assurance and shop inspection manuals did j not include or reference all procedures required for your quality program '

implementation.

These nonconformances 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.

Sincerely,

/

Rober

%A Gallo, Chief Special Inspection Branch Division if Inspectior .nd Support Programs Office of Nuclear Reactor Regulation Docket No. 99900532 1

Enclosures:

1. Notice of Nonconformance '

2. Inspection Report No. 99900632/96-01 44

NOTICE OF NONCONFORMANCE Commercial Union Insurance Company Docket No.: 99900632 Boston, MA On the basis of the results of an inspection conducted from June 19 through 21, 1996, it appears that certain of your activities were not performed in accordance with The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (ASME Code) requirements tnat are applicable to your activities under the scope of your certificate of accreditation.

A. Paragraph NCA-5220, " Categories of Inspector's Duties," of Subsection NCA, " General Requirements for Division 1 and Division 2," of Section III, " Rules for Construction of Nuclear Power Plant Components," of the ASME Code requires, in part, that the authorized nuclear inspector (ANI) perform the duties required in ASME N626-1990, " Qualifications and Duties for Authorized Nuclear Inspec' ion Agencies and Personnel," and addenda through N626b-1992.

Section 0-3, "The Authorized Nuclear Inspector," Subsection 0-3.2,

" Duties," Paragraph 0-3.2.18 of ASME N626b-1992 requires, in part, that the ANI keep a bound diary of activities and inspections made. The information to be recorded include a description of the item inspected, the type of observation made, the requirements that prompted the activity, and the results of tne inspection.

Subsection 4.2, " Duties cf ANI," of the Commercial Union Insurance Company (Commercial Union) " Quality Assurance Program for Implementation and Control of Inspection Activity in Accordance With ASME Section III, Division 1, and QAI-1," Revision 3, December 19, 1995 (QA manual),

requires, in part, that the information recorded in the bound diary include the requirements that prompted the activity and the results of the inspection.

Contrary to the above' requirements, the ANI diary entries for the inspection activities performed at Amer Industrial Technologies (Amer) from October 13, 1995, through June 17, 1996, did not, for the most part, include the inspection results, the type of observations made, and the requirements that prompted the activity.

(Nonconformance 99900632/96-01-01)

B. Paragraph NCA-5290, " Certification of Data Reports and Construction Reports," of Subsection NCA of Section III of the ASME Code required, in part, that the ANI certify the appropriate data reports after satisfying him or herself that all requirements of Section III of the ASME Code had been met.

Enclosure 1 45

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

Paragraph 4.2.17 of Subsection 4.2 of the Commercial Union's QA manual requires, in part, that the ANI include details of QA monitoring reports (MRs) in the bound dairy, in addition to filling out the appropriate forms.

Subsection 3.4, "QA/QC Monitoring Procedure," of the Commercial Union's

" Contract Inspection Services Shop Inspection Manual," Revision 2, December 1, 1995 (shop inspection manual), requires, in part, that each time a monitoring function is performed, it be recorded in the bound dia'y with the ANI's findings.

Subsection 7.1, "QA Monitoring Reports," of the shop inspection manual requires, in part, that the ANI describe the discrepancies on the MR I form and that before the MR is distributed, the manufacturer describe on

the form how the discrepancies will be resolved and give the completion i

date for the resolution.

Contrary to the above requirements, the ANI issued two MRs (November 1, 1995, and January 1, 1996) to Amer that were not noted in the bound diary and that had been distributed without documenting Amer's resolution plan for the discrepancies or the completion date.

Additionally, the QA manual and the shop inspection manual did not establish measures that ensure all job-specific MRs are closed before the ANI certifies the ASME manufacturer's data report and authorizes application of the ASME mark.

(Nonconformance 99900632/96-01-02)

C. Section 0-1, "The Authorized Inspection Agency," Subsection 0-1.2,

" Duties," Paragraph 0-1.2.5 of ASME N626b-1992, requires, in part, that 1 the agency shall establish and implement an internal program which shall l provide assurance that those of its employees holding the positions of authorized nuclear inspector supervisor (ANIS) and ANI perform work in accordance with the requirements of Part N626.0 of this Standard. This program shall be documented by written policies, procedures, or instructions and shall be carried out throughout the life of any agreement covering ASME Code Section III, Division 1 work, in accordance with these policies, procedures, or instructions.

Contrary to the above, the inspectors identified the following examples where Commercial Union failed to describe, referen<,e, or include I activities and documents that are part of the Quu tty Assur nce Program i implementation within the QA manual or the Shop Inspection Manual for l conducting these quality activities.

1. The QA Manual and the Shop Inspection Manual c'o not properly 1 identify, adequately reference, or include the checklist that is l used by the ANIS for the semiannual Section III nuclear shop audits '

and the semiannual audits of ANI duties.

2 l 46 l

4

2. The Home Office Technical Manager's annual audit of ANIS activities (which is usually delegated to Zone Manager) is conducted using the

" Zone Office Nuclear Audit Checklist" which is also not referenced or included in the QA Manual or Shop Inspection Manual.

3. Commercial Union is performing internal audits of the Home Office Technical Managers activities and overall QA program implementation.

The 1995 " Internal Audit of the ASME QA program," was performed by a zone technical specialist using the " Nuclear Audit Checklist - Home Office Acti"ities." The curren' QA manual does not describe or include the internal audit process currently being implemented and therefore, there are no documented QA program requirements (such as procedures) on how to accomplish this quality activity.

(Nonconformance 99900632/96-01-03)

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 the corrective actions and preventive measures were or will be completed.

I Dated at Rockville, Maryland this 4th day of September, 1996.

3 47

1

. U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION Report No: 99900632/96-01

~

Organization: Commercial Union Insurance Companies One Beacon Street Boston, MA 02108-3106 4

Contact:

Edgar Whittle, Manager ASME Codes and Standards Nuclear Industry Activity: Authorized Nuclear Inspection Agency Dates: June 19 - 21, 1996 I

Inspectors

Uldis Potapovs, Senior Reactor Engineer  !

Richard McIntyre, Senior Reactor Engineer

! Steven Matthews, QA Specialist I

Approved by: Gregory C. Cwalina, Chief Vendor Inspection Section Special Inspection Branch Division of Inspection and Support Programs i

j e

l l

Enclosure 2 48

l' INSPECTION

SUMMARY

During this inspection, the NRC inspectors reviewed the implementation of selrcted portions of the Comercial Union Insurance Company's (Commercial Uni >n's) quality assurance (QA) program and reviewed activities associated with the implementatinn of Commercial Union's inspection services contract wit h Amer Industrial Technologies, Inc. (fer) l The inspection bases were:

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

• American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code)

• ASME N626-1990, " Qualifications and Duties of Authorized Nuclear Inspection Agencies and Personnel."

During this inspection three instances were identified where Commercial Union failed to conform to ASME Code requirements imposed by the ASME as a condition for accreditation as an Authorized Inspection Agency (AIA). These t

l nonconformances are discussed in Sections 3.3 and 3.5 of this report.

t 2 STATUS OF PREVIOUS INSPECTION FINDINGS This was the first NRC insoection of Commercial Union's activities performed under an ASME Certificate of Accreditation.

l  !

3 INSPECTION FINDINGS AND OTHER COMMENTS 3.1 10 CFR Part 21 Proaram The inspectors reviewed Comercial Union's procedures for implementing the provisions of 10 CFR Part 21 for reporting defects and noncompliance:

l l

Contract Inspection Services Shop Inspection Manual, Revision 2, dated l

December 1, 1995, Section 3.10, " Reporting Defects & Noncompliance with 10CFR21" and Sections 2.1.7 and 3.2.8 of Commercial Union's QA manual, Revision 3, dated December 19, 1995.

l l

l The respective sections of Commercial Union's QA manual direct all authorized nuclear inspector supervisors (ANISs) to provide their staff with specific written instructions for contacting them when questionable circumstances which may be reportable under 10 CFR Part 21 are. identified. Such circumstances are then investigated by the Supervisors and reported in writing to the Technical l

Manager. The Technical Manager, with input from the ANIS assesses the nature and severity of the nonconforming condition, determines if 10 CFR Part 21 is applicable, and provides his conclusion in writing to the Responsible Officer The with a recommendation whether the NRC should or should not be notified.

final decision on notification is made by '.he Responsible Officer (Commercial 1

Union Vice President) based on his review of the data and recommendation in l

l the report. Review of records and recent correspondence during this In inspection indicated that this guidance was being effectively implemented.

i 2

5 49

at least one instance the ANIS identified a potentially reportable condition to the Technical Manager who evaluated the condition and determined that it was not reportable. Review of information on file verified adequate basis for this decision.

While the inspectors determined that the above procedures adequately addressed the evaluation and reporting requirements of 10 CFR Part 21, it was noted that the implementing instructions were focused on nonconforming conditions identified at manufacturing facilities and did not explicitly address the need to also evaluate the potential reportability of deviations which may result from identified nonconformances with Commercial Union's own QA program requirements. This was identified as a programmatic weakness. The Technical Manager committed to making appropriate clarifications to the existing procedures.

3.2 Ouality Assurance Fr itag Comercial Union has been accredited by ASME as an AIA in accordance with ASME N626, " Qualifications and Duties for isthorized Nuclear Inspection Agencies and Personnel," 1990 edition and addenda through N626b-1992. Their ASME certificate of accreditation read as follows:

Commercial Union Insurance Company Contract Inspection Services One Beacon Street Boston, Massachusetts 02108 As required by Section 0-1, "The Authorized Inspection Agency," of ASME N626b- <

1992, Comercial Union established and implemented an internal program to ensure that the ANISs and the authorized nuclear inspectors (ANIS) perform  ;

their duties in accordance with ASME N626 and that levels of inspection '

activity are commensurate with the scope of the ASME AIA certificate of accreditation. The program was documented in Commercial Union's " Quality Assurance Program for Implementation and Control of Inspection Activity in Accordance With ASME Section III, Division 1 and QAl-1," Revision 3, December 19,1995 (QA manual), and in "Comercial Union Insurance Company Contract inspection Services Shop Inspection Manual," Revision 2, December 1, 1995 (shop inspection manual).

The scope of Commercial Union's activities, as described in their QA manuti, is to provide the services of authorized inspection to manufacturers and contractors, at fabricating facilities and field sites, who hold ASME Section III, Division 1, certificates of authorization. The QA manual further states that Commercial Union does not intend to engage in ASME See. tion III, Division 2 or ASME Section XI inspection activities. Since Commercial Union does not have any contracts with NRC licensees, the QA manual does not address the requirements of 10 CFR Part 50, Appendix 5. The QA program, as reviewed and accepted by ASMt, appeared to adequately address the applicable requirements.

3 50

3.3 Commercial Union's Inspection Services l a. Insoection Scope During the NRC inspection of Amer, from January 29 through February 2, 1996 (NRC Inspection Report 99901292/96-01, March 21, 1996), the team

evaluated four jobs comprising nuclear components shipped to NRC

! licensees. All four jobs had been inspected by ANIS employed by Factory Mutual Engineering Association (Factory Mutual), and for each job an ASME manufacturar's data report (MDP) had been signed by Amer and certified by the Factory Mutual ANI indicating that Amer had complied l with all applicable requirements of the ASME Code,Section III.

i Although MDRs for all four jobs stated that Amer had met the ASME Code requirements, the NRC inspection of Amer showed that none of the jobs evaluated completely met all applicable requirements of the ASME Code.

On the basis of its fir. dings and other concerns raised during its inspection of Amer, the NRC conducted a followup inspection of Factory Mutual (NRC Inspection Report 99900603/96-01, May 28, 1996).

During the followup insrcction of Factory Mutual the team found that, since the early part of 1993, the Factory Mutual Anis assigned to Amer, the Factory Mutual district office, and the Factory Mutual home office

, had repeatedly requested that the inspection services contract with Amer

! be cancelled. These requests were largely based on several Factory 1 Mutual assessments of Amer's technical performance which showed that i Amer consistently failed to comply with ASME Code requirements. i However, despite several requests to cancel the Amer inspection services contract, the AIA's management (Arkwright Mutual Insurance Company) did not terminate this contract until September 1995 (see NRC Inspection Report 99901296/96-01, August 23, 1996).

In October 1995, Commercial Union began providing ANI inspection I services to Amer.

The scope of this NRC inspection was focused on reviewing the effectiveness of Commercial Union's activities to implement their inspection services contract with Amer,

b. Observations and Findinas The ANI's responsibilities, as described by the ASME Code, are to provide third-party oversight of an ASME certificate holder (manufacturer) assumed to be conscientiously trying to comply with the requirements of the ASME Code. However, the Factory Mutual files reviewed during the NRC inspection of Factory Mutual showed that the normal assumptions and principles of third-party oversight did not apply  !

1 I

at Amer. The documentation and interviews with the ANIS and ANIS showed that from 1993 through September 1995, the ANIS had to exercise extreme diligence (over and beyond that rout.nely expected of third-party l oversight) in the performance of their inspections and verifications l l because Amer's failure to comply with the ASME Code requirements was pervasive in Amer's fabrication activities.

4 i

l 51

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

I Factory Mutual's management reiterated to its ANIS not to certify Amer's ,

MORs when they knew that the ASME Code requirements were not being met. ]

The team recognized that the ANIS are not expected to verify every i detail of the manufacturing process and, under the existing conditions, '

the team duermined it was not reasonable for the ANIS to assume that all details that were not verified actually met the ASME Code. The NRC findings during its inspection of Amer support the team's conclusion I that the AN!s could neither verify every detail of Amer's manufacturing i process nor assume that those details not verified met the ASME Code.'  !

To evaluate Commercial Union's management of the Amer inspection services contract that began in October 1995, the team reviewed i Commercial Union's records and audit reports and held several 1 interviews. The team evaluated the Commercial Union's ANIS' semi-annual (

audit of Amer's QA program and found that the ANIS had made '

recommendations to correct certain nonconforming areas and to strengthen Amer's QA program. The team found the ANIS' audit to be thorough and well documented.

t The team reviewed the bound diary of the Commercial Union's ANIS for the '

inspections conducted at Amer's facilities from October 13, 1995, through June 17, 1996. This review showed that the ANIS had performed several inspection activities from pre-fabrication through final testing and MDR certification. The team determined that the amount of inspection activity documented by the AN!s was extensive and that it resulted in significant revisions to Amer's process control system.

These revisions were necessary to properly document inspection and verification activities by both the Amer QA/ quality control (QC) personnel and the Commercial Union ANIS.

However, the individual entries in the bound diary lacked sufficient l documentation to substantiate the various and numerous ASME Code '

complianca issues identified by the ANIS regarding Amer's performance. '

Specifically, the diary was, for the most part, annotated in short, cryptic notes that were not always clear about the specific inspections or verifications performed. The team's evaluation of the bound diary entries and its review of these entries with the ANIS during an i interview showed that the entries failed to document the inspection '

results, the type of observations made, and the requirements that prompted the activity.

These documentation requirements were specifiori hv Subsection 4.2, j

" Duties of ANI," of the Commercial Union QA manual. These requirements were also imposed on Commercial Union through its ASME accreditation as an AIA in accordance with ASME N626b-1992 and Paragraph NCA-5220,

" Categories of Inspector's Duties," of Subsection FCA,Section III, of the ASME Code.

Failure to demonstrate conformE'ce with the apt licable requirements of ASME N626b-1992 and Subsection 4.2 of the Commercial Union QA manual was identified as Nonconformance 99900632/96-01-01.

5 52

l Additionally, during its evaluation of the bound diary, the team noted .

that the ANIS had issued two QA monitoring reports (MRs) while I performing inspection activities at Amer. The first MR, dated i November 1, 1995, addressed Amer's failure to (1) record the welder's identification on the route sheets for job N391, (2) record material '

traceability numbers on the route sheets for job N391, and (3) include a design change order for Revision 7 of the drawings for jnb N391. ,

The second MR, dated January 10, 1996, addressed Amer's failure to recertify certain nondestructi"e testing personnel during the past 3-  :

year period and to calibrate a humidity and tempera Are measurement instrument.

The team found that neither MR was described in the bound diary; no entry in the bound diary existed for November 1, 1995, the date of the first MR; and the manufacturer had not described on the MR form how the '

discrepancies would be resolved and had not given the completion date for the resolution.

l As provided for in Subsection 4.2, " Duties of ANI," of the Commercial Union QA manual, the ANIS sh H d complete and record the MR details in the bound dairy. Subsection 3.4, "QA/QC Monitoring Procedure," of the Commercial Union shop inspection manual required that each monitoring  :

function be recorded in the bound diary, including the ANI's findings. )

Subsection 7.1, "QA Monitoring Reports," of the shop inspection manual required that the ANI describe the discrepancies on the MR form and that, before the MR was distributed, the manufacturer describe on the form how the discrepancies would be resolved and give the completion date for the resolution The team also found an Amer QA issue form, dated March 29, 1996, that the ANI had prepared to address an issue pertaining to the delta ferrite testing cf weld wire. However, the ANI had not noted the issuance of j this QA issue form in the bound diary.

I l

The team's evaluation of the QA manual and the shop inspection manual showed that Commercial Union had failed to require that all MRs be satisfactorily closed before the ANI certifies the MDR and authorizes the application of the ASME mark. Therefore, open deficiencies may exist at the time the ANI certifies the MDR.

In summary, the team found that the ANIS did not note the MRs in the bound diary as required and that Commercial Union failed to establish measures that all deficiencies identified on MRs were satisfactorily closed before the ANI certified the MDRs. These findings constitute Nonconformance 99900632/96-01-02.

c. Conclusions The team concluded from its review of the ANIS' bound dairy and

, interviews of the ANIS that Commercial Union ANIS and ANIS were exercising expected diligence in the performance of their inspection and 6

53

l l Verif' cation activities at Amer's facility. The team reiterated its l concern that NRC licensees should be able to rely on Amer's MDRs as evidence that nuclear components supplied by Amer fully met all the applicable requirements of the ASME Code.

l 3.4 Personnel Indoctrination. Trainino and Oualification The inspectors reviewed the various sections of the Commercial Union QA manual which describe the indoctrination and training requirements for ANIS and ANISs and the accompanying requirements for records and documentation. The ,

following Sections of the QA program were reviewed:

• Section 1.0, "AIA Qualifications"

• Section 3.0, " ANIS Qualifications"

• Section 4.0, "ANI Qualifications"

• Section 5.0, " Indoctrination, ANI"

• Section 6.0, " Training, General"

• Section 7.0, " Training, ANI"

• Section 8.0 " ANIS Indoctrination and Training,"

These procedures describe general as well as specific requirements for the indoctrination, training and qualifications of inspection personnel at Commercial Union. Training takes place in several different forms such as self-study, on-the-job training, informal question and answer sessions ,

formal classroom training with written quizzes, and appropriate National Board qualification training. It was also noted that Commercial Union employees receive training on 10 CFR Part 21. Indoctrination and training documentation is maintained at each zone office and the Boston home office and consist of documents such as personnel training files, formal training logs, attendance sheets, training matrix, and certificates or letters of training completion.

The training records reviewed by the inspectors included the appropriate documentation for the qualification and training required for ANIS per ASME QAI-1 and the Comercial Union quality program.

The inspectors also reviewed the training files and qualification records for 3 ANIS and 2 ANISs. The training files cor.tained all pertinent training record documents required by the Comercial Union QA program and appropriate ASME QAl-1 training and qualification records. The inspectors also verified l that each ANI and ANIS training file included the Comercial Union Qualification and Certification Forms signed by the Technical Manager.

l The inspectors verified on a sample basis that the Home Office Technical Manager had reviewed and initialed the training records files for ANIS and ANISs on an annual basis (for the last three years) as required by Comercial

Union QA Program Section 6.0 requirements.

l l

3.5 ASME Audits and Inspector's Performance The inspetors reviewed the various sectons of the Comercial Union QA manual and Section 5.0, " Surveys and Audits," of the Comercial Union Shop Inspection Manual, which describe the audit and survey requirements for the Home Office Technical Manager, the ANIS, and the ANI, and accompanying requirements for 7

l s4

i l

l audit report documentation. The following Sections of the QA program were reviewed:

• Section 2.0, "AIA Duties and Responsibilities"

• Section 3.2, " Duties of ANIS"

• Section 4.2, " Duties of ANI" The Commercial Union nuclear survey and audits program is written to meet the l

l requirements of ASME Section III, Division 1 and ASME N626.0/QAl-1. This included audit methods and documentatior audit reports of nuclear Section III l

work, authorized inspector performance audits and the audit schedule for each.

I The inspectors reviewed the current Authorized Nuclear Inspector Roster that listed ASME Section III shops, including information such as the ANIS and the l

ANI assigr.ad, and nuclear audit schedule. The ASME Section III shop audits are performed to review for applicable ASME Code compliance and for the implementation of the various sections of the shop's QA manual utilizing standard Commercial Union audit checklists. The nuclear shop and ANI

! performance audits are both performed semiannually. The inspectors reviewed the results of recent ANIS nuclear shop audits conducted at three ASME Section III Code shops; Amer Industrial Technology, Henry Vogt Machine Company, and Anchor Darling Valve Company. The inspectors also reviewed the semiannual inspector's performance audits conducted by the ANIS during his nuclear shop audits at the above mentioned Section III shops. The audit reports included the appropriate documentation required by Commercial Union procedures l including Nuclear Audit Deficiency Reports when applicable.

l l During the above review, the inspectors identified that QA manual Section 3.2 requires the audits of ANI performance be conducted using the checklist identified in Section 5.5, " Nuclear Audit," of the Shop Inspection Manual. l 1

The current Revision 0 of Section 5.5, dated September 1, 1992, specifies use of an attached checklist in the performance of nuclear audits. However, no such checklist was attached or included in Section 7.0, " Forms," of the Shop Inspection Manual. The inspectors also determined that the QA manual and the Shop Inspection Manual does not properly identify, adequately reference, or l

include the checklist that is used by the ANIS for the Section III nuclear shop audits and the ANI performance audit. The Home Office Technical Manager's audit of ANIS activities (which is usually delegated to Zone l Manager) is conducted using the " Zone Office Nuclear Audit Checklist" which is also not referenced or included in the QA program or Shop Inspection Manual.

The inspectors a.so determined that Comtrercial Union is performing internal audits of the Home Office Technical Manager's activities and overall QA program implementation. The 1995 " Internal Audit of the ASME QA program," was performed by a Zone Technical Specialist using the " Nuclear Audit Checklist -

Home Office Activities." The current QA program does not address or describe this internal audit process being implemented and therefore, there are no documented QA program requirements for accomplishing this quality activity.

i The failure to adequately describe, reference, or include activities and documents that are part of the QA program implementation within the QA program 8

55 l

or the Shop Inspection Manual is identified as Nonconformance  !

99900632/96-01-03.

3.6 Entrance and Exit Meetinos ,

In the entrance meeting on June 19, 1996, the NRC inspectors discussed the  ;

scope of the inspection, outlined areas to be inspected, and established e interfaces with Commercial Union's management. In the exit meeting on June 4 21, 1996, the inspectors discussed their findings and concerns. ,

3 LIST OF PERSONS CONTACTED Edgar Whittle, Technical Manager Steve Voorhees, Authorized duclear inspector Supervisor ,

i ITEMS OPENED 99900632/96-01-01 NON Incomplete ANI diary entries l 99900632/96-01-02 NON Inadequate procedures for ASME Code .

inspection activities  !

99900632/96-01-03 NON Inadequate control of quality program [

doct.ments i l

l I

l i

I 9

56

p Hog ye 74 UNITED STATES l "

• l NUCLEAR REGULATORY COMMISSION l Ea 4E WASHINGTON D 4 F555-00m l

%,e,,,,/ September 25, 1996 L

Mr. Lyle H. Bohn, Executive Vice President i Integrateo Nuclear Services Framatome Technologies, Inc.

i 10935 Old Forest Road -

Lynchburg, VA 24506-0935

SUBJECT:

NRC INSPECTION REPORT 99901300/96-01 AND NOTICE OF NONCONFORMANCE

Dear Mr. Bohn:

~

1 l

On July 12, 1996, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Framatome Technologies, Incorporated (FTI), Integrated Nuclear i Services facility. The enclosed report presents the results of that inspection.

The inspection team determined that the FTI eddy current testing program that l is managed by your Integrated Nuclear Services organization was generally well l implemented and controlled, with the exception of a few concerns that are

! discussed herein. During the conduct of this inspection, the NRC inspectors found that the FTI staff were experienced and knowledgeable regarding their respective areas, and the FTI personnel who were interviewed reflected a l strong sense of ownership and pride in their different job functions.

l One concern, as discussed in Section 3.5.4 of the enclosea report, relates to l a 1994 Nuclear Industry Assessment Committee (NIAC) Supplier Audit Summary Report of Zetec, Incorporated (Zetec), that FTI used as part of its basis for procuring eddy current (EC) computer software (EddynetM) . The team determined that the report did not present adequate narrative to determine whether the effectiveness of the supplier's implementation of its software quality program was adequately evaluated. Subsequent to the FTI inspection, the team performed an inspection at Zetec (NRC Inspection Report 99901037/96-

01) in July 1996 and identified some concerns with Zetec's software program l controls. Thesa soncerns contrasted with the conclusion discussed in the NIAC l Report. Consequently, the team concluded that NIAC's Report did not provide l an accurate and complete picture of the implementation of Zetec's Eddynet l software program.

The inspectors determined that the implementation of one portion of your quality assurance program failed to meet certain NRC requirements imposed on you by your customers. Specifically, onc of your activities relative to the quality of safety-related activities was not delineated in written instructions or procedures as required by 10 CFR Part 50, Appendix B. This nonconformance is cited in the enclosed Notice of Nonconformance (NON), and Eddynet is a registered tradamerk of Zetec, Inc.

I, 57 1

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

L.H. Bohn the circumstances surrounding it is described in detail in the enclosed report. You are requested to respond to the nonconformance and to 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 olaced in the NRC's Public Document Room.

Sincerely, i

ORIGINAL SIGNED BY G. CWALINA FDR:

Robert M. Gallo, Chief .

Special Inspection Branch j Division of Inspection and Support Programs '

Office af Nuclear Reactor Regulation Docket No. 99901300/96-01

Enclosures:

1. Notice of Nonconformance

2. Inspection Report 99901300/96-01 cc: Emily Mayhew, Acting Quality Assurance Director .

Framatome Technologies, Incorporated '

155 Mill Ridge Road I Lynchburg, VA 24502-4341 i

I d

58

l-i NOTICE OF NONCONFORMANCE E

< Framatome Technologies, Inc. Docket No.: 99901300

integrated Nuclear Services Lynchburg,. Virginia t

~

Based on the results of an inspection conducted July 9 through 12, 1996, it appears that certain of your activities were not conducted in accordance with i the requirements of the U. S. Nuclear Regulatory Commission.

2 Criterion V, " Instructions, Procedures, and Drawings," of Appendix B to

10 CFR Part 50 requires

• that activities affecting quality be prescribed by j documented instructions, procedures, or drawings 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.

i Section 5, " Instructions, Procedures & Drawings," of Framatome Technologies, Incorporated's, (FTI's) Quality Assurance Program manual (QAPM), Revision 2,

January 1, 1996, requires that measures be established and documented to i

assure that activities affecting the quality of items are established in

! Instructions, procedures, or drawings and are prepared, reviewed, approved, i and distributed before the activity is begun.

Contrary to the above, FTI did not establish procedures to control and ensure

. repeatable results regarding the development of safety-related eddy current computer data screening (CDS) techniques used by data analysts.

(99901300/96-01-01)

Please provide a written statement or explanation to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555-0001, 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 the following for each nonconformance: (1).a description of steps that have beer or will be taken to correct these items; (2) a description of steps 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 completed.

Dated at Rockville, Maryland, this 25th day of September, 1996 59 Enclosure 1 )

i

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

Report Ho: 99901300/96-01 I

Organization: Framatome Technologies, Incorporated Integrated Nuclear Services Lynchburg, Virginia

Contact:

Emily Mayhew, Acting Quality Assurance Director (804) 832-3331

)

i Nuclear Industry Supplier of eddy current nondestructive Activity: examination (NDE) related training, nuclear power plant steam generator tubing eddy current examination and analysis and NDE personnel services.

Dates: July 9-12, 1996 j Inspectars: Joseph J. Pet rosino, Team Leader  ;

John K. Ganiere, Electrical Engineer i Michael J. Morgan, Reacter Engineer l Pnillip J. Rush, Materials Engineer Michael E. Waterman, I&C Engineer i

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

l Enclosure 2 60

l 1 INSPECTION

SUMMARY

l During this inspection, the NRC inspectors reviewed the implementation of selected portions of the quality assurance (QA) program which is being implemented by the Integrated Nuclear Services Department of Framatome

! Technologies Incorporated (FTI) and activities associated with FTI's nuclear power plant steam generator (S/G) tube eddy current (EC) testing servicas.

The inspection bases were as follows:

l

• Appendix B, " Quality Assurance Criteria for Nuclear Power Plants and l

Fuel Reprocessing Plants," to Part 50 of Title 10 of the Code of Federal l ReaulatioD1 (10 CFR Part 50) a 10 CFR Part 21, " Reporting of Defects and Noncompliance" During this inspection, The NRC inspection team (team) identified three

instances where FTI failed to conform to NRC requirements imposed upon them by i NRC licensees which are discussed herein.

2 STATUS OF PREVIOUS INSPECTION FINDINGS

This was the first inspection of EC nondestructive examination services that l is managed by FTI's Integrated Nuclear Services f acility, formerly operated by l Babcock & Wilcox, Incorporated.

3 INSPECTION FINDINGS AND OTHER COMMENTS l

3.1 Ouality Assurance Proaram The NRC inspectors selectively reviewed FTI's Quality Assurance Program Manual 1 (QAPM), Revision 2, dated January 1, 1996, and its implementation for safety- i related activities concerning NRC licensee nuclear power plant S/G tube l inspections using nondestructive EC testing methodology. The team also l reviewed selected departmental work instructions for the Service and EC Departments.

The team found that FTI's QAPH encompasses its safety-related products and l services supplied to NRC licensees under NRC and other nuclear industry l requirements, American Society of Mechanical Engineers (ASME) Boiler and Pressure Code items and its non-safety related products and services. In its review, the team found that FTl was satisfactorily implementing its QAPM in the EC testing area, with the exception of some areas in which FTI did not have adequate instructions for controlling safety-related activities using written procedures or instructions. The team found that FTI was generally performing the required activities in these areas and had been typically performing additional activities which were not required procedurally, but enhanced or improved the particular process. That is, the team found that FTI had not established procedures to control thae additional activities (further discussion in Section 3.2). As a result of its review, the team concluded that the implementation of FTI's QA program as it related to EC testing was appropriate, except as discussed later in the report. j 2

\

l 61

3.2 Data Analysis Procedures and Acouisition Technioues l

The team reviewed the vendor's processes and controls for developing site-specific EC data analysis guidelines. Requirements for analysis guidelines l are specified in FTI's " Admin.strative Procedure for Control of Inservice j

Inspection Procedures and Procedure Qualifications," 151-1, Revision 8, dated  ;

I July 8, 1991. The team noted that additional requirements may be imposed on  ;

the vendor by a utility.

The team evaluated the vendor's processes are controls on the development of l site-specific procedures through discussions with FTI Nondestructive Examination (NDE) staff and also by reviewing previous F11 NDE inspections.

Analysis guidelines undergo numerous changes before inspections based on input from the utility and other vendors involved in the inspection. ISI-1,

! Revision 8, required a requalification of an inspection procedure if essential  ;

l variables are altered. By a review of a sample population of site-specific '

examination guidelines developed by FTI, the team verified that changes to the i guidelines were clearly documented. Overall, FTI's process for developing  !

site-specific guidelines and the controls on changes to these procedures -

appear to be satisfactory. A Level III Examiner stated that FTI generally l proposes the use of industry-qualified inspection techniques (Appendix H of the Electric Power Research Institute [EPRl] Guidelines), when possible. The i team verified this information through a review of acquisition technique j sheets incorporated into plant-specific S/G inspection guidelines.

The team concluded that FTI appeared to maintain good procedural controls for j the development of S/G examination guidelines and that changes to the  ;

l guidelines were satisfactorily incorporated in accordance with the vendor's procedures.

l 3.2.1 Control of Data Analyst Activities The team reviewed the vendor's measures to control data analyst activities related to S/G tube examinations. Industry guidance related to S/G inspections is described in EPRI NP-6201, "PWR Steam Generator Examination Guidelines," which made recommendations on analyst activities. The team verified that the vendor's program appeared to be generall. developed in l l accordance with current intentry practice.

However, the team found that FTl did not have any overall procedures, I

, instructions or written guidance for controliing EC data analyst activities. )

l Rather, the vendor relied on the requirements imposed by the different '

utilities for the particular S/G tube inspection. Utilities generally require l all data analysts to pass a site-specific examination before they can begin i

analyzing data for an inspection. In the absence of any formal procedures, j such examinations are typically administered, at the FTI facility, as close to

the actual inspection as possible, but the examination could occur while the l analysts are still performing work for other inspections. As a result, analyst activities between the time of the examination and the inspection could dirtinish the analyst's knowledge of the site-specific details emphasized i

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in the examination. Industry guidelines state that these examinations should be administered just before the start of the plant outage. FTI's program l

allowed this guidance to become more subjective than the industry guidance. l

\

FL ther, because of the lack of procedures for the control of analyst '

l activities, the team believed it to be possible for both primary and secondary

! data analysis activities to2 have occurred at the same location. This would l not be desirable during an inspection and analysis of acquired EC test data. l t FTI staff stated that they typically split their activities between offices in Lynchburg, Virginia and Benecia, California gRockridgej; however, no specific requirements appeared to be imposed by FTI to ensure that the two parties would remain in separate locations for the activity. The team concluded that FTI had not established adequate written guidance nor specificity in this area  ;

and it was considered as a weakness.

l Although FTI had not established formal procedures governing the control of

, data analysts, it appeared to have implemented effective practices for l managing its analysts, and the team found that good communication existed  !

l between the NDE staff and its supervision. Consequently, the team concluded that this area appears to be effectively implemented. The benefit of having established FTI procedures to ensure continued effectiveness was discussed with FTI staff.

3.2.2 Computer-Assisted Data Analysis The team reviewed the vendor's controls for computer data screening (CDS) techniques used in the analysis of EC inspection data to assess whether FTI adequately managed activities ir.volving the use of CDS. The team found that l FTI is often contracted to supply EC data analysis using a CDS algorithm.

These computer algorithms use specific criteria which was inputted by an FTI data analyst before a particular utility customer S/G tube EC inspection to l

identify expected EC indications. The individual responsibility is significant for the data analyst because they must develop these criteria based on expected modes of tube degradation for each different utility S/G tube EC examination. The team identified that FTI did not have a written procedure for controlling or establishing any specific process that is to be

, used by its data analysts for developing these criteria. As a result, the l screening criteria developed by one analyst may be significantly different from that of another analyst.

To establish the parameters for identifying indications in the EC data, a data l

analyst must independently develop the screening logic. Before FTI starts its EC inspection at a particular nuclear power plant site, CDS is typically subjected to a site-specific examination, which it must pass in order to be used in an inspection. The team's review identified that when developing CDS criteria, analysts will typically address known degradation modes. However, 2

, It is desirable to have the primary and secondary analyses for the same j acquisition EC data to be located in different locations and performed by different analysts.

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because of the different experience levels of the analysts, a variance in the effectiveness of the developed criteria can occur. This can allow types of degradation previously undetected by a particular analyst to be missed by the CDS algorithm. Therefore, some type of parameters, or range, need to be delineated to ensure that the widest spectrum of degradation modes will be identified. This aspect was not addressed in FTI's program which was used in the CDS criteria development.

The team also recognizes that FTI's data analysts all appear to have a much l wider kmwledge of degi adation mechani..as and may adapt to any new degradation mechani,ms that appear during inspections. The team observed that although FTI analysts have not been working with written procedures or instructions l regarding written guidance and instruction for the development of these l criteria, FTI was only using experienced analysts for developing CDS criteria.

The team concluded that the lack of such procedures could lead to differing

! performance capabilities for CDS techniques under actual inspection conditions

! and as a result, the potential exists that the CDS program may fail to identify defective tubes containing new or different modes of degradation ,

l during the actual inspections. The team identified the lack of written I guidance to govern the development of CDS criteria as Nonconformance 99901300/96-01-01.

3.3 Eddy Current Test Personnel Oualification/ Certification i The team reviewed FTI's Nuclear Servic n Group Procedure FTI 54-151-24, l

" Written Practice for Personnel Qualification in Eddy Current Examination," '

Revision 20, to understand FTI's procedures and process for FTI's eddy current  !

l NDE personnel qualification and certification. l The team found that trainees must follow the directions of certified I personnel, and they are not allowed to collect or interpret test data unless permitted to do so by certified supervisors. Trainees are not allowed to ,

independently conduct tests or interpret, evaluate, or report on any test )

results. Trainees are typically upgraded to an NDE Level I only after completion of a recommended 40-hour American Society for Nondestructive Testing (ASNT) training program and after accumulation of a minimum of 175 hours0.00203 days <br />0.0486 hours <br />2.893519e-4 weeks <br />6.65875e-5 months <br /> of " hands-on" experience.

FTI's Level I personnel perform specified setups, calibrations, and tests in accordance with written instructions, i.nd unlike trainees, they are allowed to record data. Level I personnel are authorized to r.. ..., test instructions only under the guidance of higher certified personnel, and Level I personnel  ;

are not allewed to independently evaluate or accept results.

FTI's Level I personnel are typically upgraded to Level 11 after completion of  :

a recommended ASNT training program and a minimum of 1,750 hours0.00868 days <br />0.208 hours <br />0.00124 weeks <br />2.85375e-4 months <br /> (approximately 9 months) of hands-on enerience. They perform setups and calibrations and interpret applicable codes, specificau ons, and standards.

They also note whether all data acquisition equipment functions properly and l whether such equipment requires replacement. They prepare written l

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instructions, are familiar with the scope end limitations of the testing method and provide on-the-job training (0JT) for all Level 1 and trainee personnel.

FTI's Level liA personnel perform the same activities as Level 11 personnel and they undargo documented training in the analysis of nonferromagnetic tubing EC testing data. As data analysts, they also accept or reject testing data and interpret, evaluate, report, and disposition the data. FTI's Level 11 and IIA personnel are upgraded to Level III only after completion of a recommendcd ASNT training program. l by establish examination techniques and procedures, analyze data, and interpret codes / standards and procedures / specifications and evaluate examination results based upon recognized codes and standards. Level III personnel specify examination methodologies, techniques, and procedures and they have experience in other forms of NDE.

Overall, the team concluded that this area appeared to be well controlled and straightforward, with only one exception where the team believed more attention to detail could be added. The team noted that FTI's delineation of activities which could satisfy the reouired 1,750 hours0.00868 days <br />0.208 hours <br />0.00124 weeks <br />2.85375e-4 months <br /> of hands-on experience for the site-specific and S/G specitn. training appeared weak and vague.

However, FTI's routine methodology of practice and the performance of supplemental S/G inspection training was strong and appeared to compensate for the team's concern. For example, the team found that supplemental training, such as documented on-site training and 0JT, appeared to adequately address different site requirements and unique S/G tube characteristics regarding variety and the accumulation of documented ASNT-recommended experience hours.

In summary, the team concluded that FTI's overall training appeared to be solid and well established, and the team found that FTI's EC personnel have an adequate basis for their qualification and certification, as measured by FTI's NDE Procedure 54-ISI-24. The team also concluded that this procedure appeared to be acceptable when compared to that of EPRI NP-6201, Appendix G. The procedure also appears consistent with the guidance in ASNT Recommended Practice SNT-TC-1A, 1984, the requirements of ASME Section XI, 1989, and the requirements of 10 CFR Part 50, Appendix B, Criterion IX.

3.4 Vendor Self-Assessments and Third-Party Audits 3.4.1 Sel f- Assessments In t' .

i . of internal audits and self-assessments, the team reviewed the methodology that FT1 had adopted to implement the provisions of Appendix B to 10 CFR Part 50 and other industry guidance.

The team found that FTI's internal auditing was accomplished using Procedure 1719-21, " Quality Assurance Audits of 'nternal Activities," Revision 10, May This procedure requires in-hourn auditing of EC functions, that is, 1996.

safety-related or ASME Code Sections I.1 and XI, activities, on at least a once-per-calendar-year basis. In the past two years, internal QA audits of l FTI's NDE/EC testing group were performed on two diffet<ent occasions, as discussed in Framatome/BWNT Report 94-08, August 1994, and 95-09, July 1995.

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A review of these reports showed that FTI generally took appropriate  !

corrective action to respond to identified problems and to preclude their "

recurrence-For example, one of four remote data acquisition units (RDAUs) identified in a l Crystal River EC examination report was misidentified by FTI's EC technician. ,

An an immediate corrective action, a memorandum that stressed the importance '

of entering correct serial numbers in data sumuaries was distributed to all EC '

technicians. To preclude similar problems, FTI now requires two-party ,

verificatian of all entries, and this var;ficath.n is stressed in current FTI activity control documents. The team also noted that supplemental training  ;

for EC Level IIA and Level III analysts was to be performed on at least an  ;

annual basis. For some of the analysts, no evidence of this required training  ;

was found for 1994. As an immediate corrective action in December 1995, FTI routinely scheduled supplemental training for every December-January time '

period. To preclude similar problems, FTI scheduled its subsequent training j within 12-to-15 month intervals, as recomended in EPRI NP-6201, Appendix G. j i

The team found that all issues were satisfactorily dispositioned. The team  ;

concluded that FTI's self-assessment program was adequate, and the team noted ,

that FTI's QA/ internal audits were acceptable and appeared to meet the -

guidance in EPRI NP-6201, Section 2.  ;

i 3.4.2 Third-Party Audits The NRC inspection team reviewed selected third-party audits of the FTI t organization to assess the scope of the audits and to review any identified weaknesses. The inspection team's review of audit reports of FTI's program found that some third-party assessments of FTI's EC testing program functions are periodically performed by the Nuclear Procurement Issues Committee (NUPIC) and nuclear power plant utilities.

The team reviewed two recent NUPIC audits of FTI's NDE group that were made in July 1994 and June 1996. During its review of these audits, the team noted that the audit results were generally satisfactorily, with a few exceptions.

For example, HUPIC noted some weakness in the areas of " starting the clock" for Level III recertification and the calibration of one RDAU, which was used on a Davis-Besse S/G inspection. The team also reviewed a limited-scope audit performed by Baltimore Gas & Electric Company (BG&E) Report in May 1995. The BG&E audit report stated that the implementing procedures and activities of FTI, (formerly BWNT), were appropriately perfornied and complied = :h FTI's QAPM. The BG&E report also indicated that FTI's QA Program was adequately implemented and effective in the areas evaluated. The team's review concluded that FTI adequately resolved any findings and performed appropriate corrective action. No concerns were identified in this area.

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3.5 Eddy Current Software j 3.5.1 Receipt The team assessed FTI's process relating to the receipt of the Eddynet software3 . It appeared that the supplier, Zetec, continues to update their Eddynet 95 software versions after customer receipt in response to customer requests and to improve man-machine interface characteristics of the software. ,

As a result, new versions, version updates, and prototypes have been received  !

by FTI. The tam found tnat Zetec typical./ supplied FTI with new versions of l its software on optical disks and FTI returned old versions to Zetec upon i receipt of new versions. The team found that the optical disk typically  ;

l contained " read me" files which described the changes or additions that were l made in the new version. Upon receiving the optical disk from Zetec, FTI stated that it performs a check-out of the acquisition and analysis software  ;

and this process was reviewed in detail by the team.  ;

FTI documents receipt of new Zetec software versions in its quality control inspection reports (QCIRs). Each QCIR is assigned a unique QCIR number, which is maintained in the FTI quality assurance system. Interviews with FTI ,

personnel revealed that there is no governing procedure for performing this  !

function. The staff reviewed two QCIRs: QCIR 96-00339 (Document 1237552A-00) and QCIR 96-00340 (Document 1222845A-2), both dated April 1996, under Calvert Cliffs (BG&E) Contract Number 1010441. QCIR 96-00339 documented completion of check out activities for Eddynet acquisition software, Zetec Eddynet 95  !

Software Version 2.0, Patch E95-2.1. QCIR 96-00340 documented completion of Eddynet 95 System Check-out Procedure, Zetec Eddynet 95 Software Version 2.0, Patch E95-2.1. The team saw that the Eddynet analysis program that FTI used for its check out utilized a set of QA-controlled test data, which had been verified to produce known analysis results. The procedure provided l unambiguous guidance for setting up the tests and test data and for protecting l the original set of test data. The procedure required a verification that the l software analysis results agreed with the baseline results. Additionally, the l functional testing verified the correct transmission of data between a work l station and a server. These functions were verified by the test personnel and l FTI management. i The inspectors found that although FTI's NDE manager expected his staff to  :

validate new or upgraded Eddynet software using an attribute sheet which l contained special data and outlined confirmatory analyses as described above, I l FTi had not establ4 ..ed any procedure to require the validation process to be  :

completed in a consistent, repeatable manner. The team concluded that this i process appeared to be satisfactory in practice, except FTI had not required 3

FTI's Approved Supplier List showec' that Zetec is an approved Appendix B supplier for NDE equip ent and as a source of NDE personnel for on-site EC testing but does not list Eddynet. Appendix B to 10 CFR Part 50 is 4

not imposed by FTI for procurement of Eddynet nor will Zetec accept Appendix B j for Eddynet. The Eddynet 95 software is a commercially available product.

8 67 l ,

the process to be completed by instruction or procedure. As a result, the team concluded that there was no assurance that future versions or version updates will be validated using the existing validation process.

The team was concerned that FT! had not established written controls for the Zetec software receipt inspection activities considering the importance of the i S/G tube eddy current test activities utili.ing this software. The team notes  ;

that this concern would have been identified as a nonconformance had Appendix B been imposed by FTI; however, Zetec does not accept Appendix B to 10 CFR Part 50 for its Eddynet. Consequently, the receipt of Eddynet would be as a commercial item by FTI.

3.5.2 New Version Releases FTI indicated that in addition to its internal check out of new Eddynet software versions, Zetec val.Jates all new version releases, version update releases, and prototype releases of the Eddynet software product.

Specifically, FTI referenced three Zetec work instructions that have been developed for the software validation of new versions, version updates, and prototypes. The inspectors reviewed the following three Zetec work instructions: (1) New Version Product Disk Validation Work Instruction, dated June 29, 1994; (2) Version Update Disk Validation, dated September 19, 1994; and (3) Prototype Disk Validation, dated September 1994. The team's discussions with FTI staff indicated that Zetec implemented these work instructions to ensure that new version releases of the Eddynet software would perform correctly by conducting check outs and tests of new software products.

The inspectors review of a sample of these version check out procedures l indicated that the procedures appear comprehensive and demonstrates that Zetec ,

performs software validation before releasing Eddynet software products. '

I 3.5.3 Software Chance Reouests t

The inspectors reviewed the process for reporting and resolving Eddynet software deficiencies. FTI indicated that there is no formal mechanism to internally report and track software problems or recommended changes.

However, FTI indicated that Zetec developed a Software Change Request (SCR) form for customers to report apparent problems or to suggest recommended changes and enhancements to tne Eddynet software. The inspectors reviewed the SCRs completed by FTI.

The inspectors ;, sed a question about how FT! tracked the disposition of each SCR. Specifically, upon reviewing the SCRs, the inspectors noted that the SCRs do not have unique identifiers to facilitate tracking and closure. In addition, FTI did not have any information on file as to the status of the SCR. Fil indicated that Zetec assigns an SCR number to each SCR when it receives the SCR from the customer. Zetec classifies each SCR as an error or a request for change and makes the final determination as to how to process the SCR. Typicay, enhancements or corre -tions to the Eddynet software are listed in an Eddynet bulletin that is sent to customers by Zetec when new versions of the software are distributed. in addition, tracking disposition of the SCRs is performed by the Eddynet Users Group. This group meets semiannually to assign priorities to SCRs from all customers. The team i

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recommended to FTl that it consider establishing a formal meUanism to .

internally report and track software problems or recommended changes that it ,

l identifies or is transmitted to Zetec.

3.5.4 Audits cf Zetec's Software Develorment Process l The team reviewed the FTl supplier audit records of Zetec, focusing on whether l FTI's suppiier audit documentation reflected a critique of Zetec's Eddynet software development process that would provide an adequate confidence level i for FTl to cor,'.inue procuring Zetec's softcare. The team found that FTI's i Approved Supplier List showed that Zetec is an approved Appendix B supplier l

l for NDE equipment and as a source of NDE personnel for on-site EC testing.

I FT! staff stated that it does not impose Appendix B to 10 CFR Part 50 for i procurement of Eddynet 95 software from Zetec because Zetec will not accept I the Appendix 8 requirement for Eddynet because Zetec's Eddynet software is a commercially available product. ,

The team assessed FT!'s basis for procurement and use of the software. FTI l provided the team with a copy of its Administrative Procedure No. 1719-22,

" Quality Assurance Audits of FTl Suppliers," May 1, 1996, and two audit reports of Zetec for 1991 and 1994, which were conducted by others. 1 Administrative Procedure 1719-22 contained requirements for evaluating and auditing quality assurance programs of FTI suppliers. The 1994 audit of Zetec was performed by the Nuclear Industry Assessment Committee (NIAC), and the  :

lead NIAC member was Asea Brown Boveri-Combustion Engineering (ABB-CE). The stated purposa of the inspection was to evaluate the effectiveness of implementation and the adequacy of the Zetec QA program to the applicable l criteria of 10 CFR Part 50, Appendix B. Software quality assurance processes i were included in the scope of the audit and were found in the NIAC audit checklist.

As a result of the audit, ABB-CE provided FTI with a Supplier Audit Summary  ;

Report 94-013, dated April 22, 1994, as well as the NIAC Audit Checklist.  ;

ABB-CE concluded that Zetec was effectively implementing the Zetec QA program l in accordance with Zetec's QAM, Z-QA, Revision 12, and related procedures.  !'

, The supplier audit report stated that Zetec is approved to supply quality class 1 (safety-related) NDE personnel, equipment, services, calibration of l equipment, manufacture of EC test standards and EC testing related computer l software (Eddynet). By memorandum dated August 8, 1994, the FTI Manager for l QA audits documented review and acceptance of the ABB-CE audit. Specifically, ,

the memorandum approved NDE personnel and NDE equipment, services, equioment '

l calibration and manufacture of EC test standards and EC testing related

! computer software.

The team reviewed the ABB-CE audit report and observed that no deficiencies were noted in the software area checklist and that all audit inspection attributes that were in the software area checklist were marked

" satisfactory." It appeared to the team that the audit report lacked l sufficient detail to make an assessment of the adequacy of Zetec's l establishment and implementation of its softwarc controls. Instead, the audit I report and checklist only indicated that, Zetec had a software development j procedure, Zetec staff performed a validation for version 24, and that the 10 f

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NIAC auditors reviewed three Zetec SCNs and a Zetec " bug" list. Therefore, the team did not believe that narrative or adequate objective evidence existed in the checklist or audit report package for one to reach the conclusion stated in the NIAC Audit Report. For example, the software area checklist section made the followir.g '.atement:

PDP-1, Revision 9 Software Product Development Procedure, is Zetec's procedure for design and development of EC analysis and I acquisition software; Final validation of Eddynet Version 24 was performed in January 1994; Reviewed Software Change Notice (SCN) numbers 404, 407 and 372. SCNs were in accordance with PDP-1, l Revision 9; Reviewed Eddynet Version 24 known bug list for l dispositions."

l The inspectors also reviewed the December 1991 audit of Zetec performed by FTI, which focused on the qualification of Zetec as a supplier of personnel for EC testing and analysis and calibration services. Software development and quality assurance processes were not included in the scope of this audit.

Therefore, the 1991 audit of Zetec was not applicable to its software product.

Subsequently, the NRC inspection team conducted an inspection at the Zetec facility on July 15 through 18, 1996, and did not reach the same conclusions as did the NIAC audit report regarding software. The NRC inspectors at Zetec identified some concerns in Zetec's software program. For example, Zetec did not establish a documented software requirement specification for its Eddynet software, nor had it performed a formal verification of its software to confirm that a given phase of the software development life cycle fulfilled the requirements imposed by the previous phase. Although software verification attributes were included in the NIAC checklist for verification, the audit assessment checklist did not identify that Zetec's software verification was not performed by Zetec. Based on the above information, the team concluded that FTI's NIAC audit report supporting the adequacy of Zetec's Eddynet did not appear to provide adequate specificity for a FTI's confidence level as a stand-alone document.

On the basis of the above discussion and circumstance, the team concluded that adequate objective evidence was not presented in the NIAC 1994 Audit Report on Zetec to conclude that Zetec's control of its software process was effectively implemented. FTI indicated to the team that the next Zetec evaluation is scheduled for December 1996, and the next audit is scheduled for March 1997.

Although some concern is noted by the team for this area, no nonconformances were identified.

3.5.5 Review of Software Purchase Orders  !

The staff reviewed the following four purchase orders (P0s) for materials and/or services from Zetec: P0 32640, South Texas Project, October 6, 1995, P0 38231, Crystal River, April 26, 1996, P0 32533, Salem, May 28, 1996, and P0 42800, South lexas Project-Unit 1, June 12, 1996. Although Zetec was audited

s an Appendix B supplier and listed in the FTl approved supplier list, FTI indicated that the Eddynet software is procured as non-safety grade software 11 70

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1 and is not dedicated as a basic component. This was substantiated by the team by review of the above-listed P0s, which were placed as orders for non-safety-l grade equipment. l In its review of the FTl Supplemental Checklist for Software Development,Section IV, " Procurement," (Figure 5), the team found that Zetec EC equipment and calibration standards were purchased on April 10, 1996, under P0 41215 and P0 34868; the item descriptions are MIZ-30-4, VH-4606, P-006. The supplier evaluation audit was performed in March 1994 and the acceptance documents were 56496 and 55929. No concerns were identified in this area.

3.6 Entrance and Exit Meetinas In the entrance meeting on July 9, 1996, the NRC team discussed the scope of l the inspection, outlined the areas to be inspected, and established FTI staff l contacts for interaction. In the exit meeting on July 12, 1996, the team l discussed its findings and concerns.

PERSONS CONTACTED Lyle Bohn Executive VP, Nuclear Services Charles England Vice President, Outage Services Dick Gill Director, Quality Assurance Emily Mayhew Manager, Quality Assurance Audits Todd Richards Manager, NDE Technology Nicholas Simile Supervisor, Quality Control Scott Wilson Manager, S/G Service Products 1

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p3 Kf C ye t UNITED STATES s* g NUCLEAR REGULATORY COMMISSION f WASHINGTON, D.C. 20SS4001

,,,,,# September 25, 1996 Mr. Craig P. Kipp Plant Manager General Electric Nuclear Energy Nuclear Energy Production P.O. Box 780, Mail Code A20 Wilmington, NC 28402-0780  :

1 SU8 JECT: NONPROPRIETARY VERSION OF NRC INSPECTION REPORT NO. 99900003/96-01 l

Dear Mr. Kipp:

This letter transmits the nonproprietary version of the U.S. Nuclear Regulatory Commission's (NRC's) Inspection Report 99900003/96-01. Our letter to you dated September 10, 1996, transmitted the original (proprietary)  ;

version of the report. On the basis of our discussions and review of the information in your September 19, 1996, letter (RJR-96-107), and its enclosure (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 the revised nonproprietary 1 (public) version of the report, we have briefly summarized the deleted or  ;

revised text. I Your response to either this letter or our letter dated September 10, 1996, and their enclosures are not subject to the clearance procedures of the Office {

of Management and Budget, as required by the Paperwork Reduction Act of 1980, i Public Law No.96-511.

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 this matter, pleased contact Steven M. Matthews of my staff at (301) 415-3191.

Sincerely, O

Ro rt M. Gallo, Chief Special Inspection Branch Division of Inspection and Support Programs Office of Nuclear Reactor Regulation Docket No.: 99900003

Enclosure:

NRC letter to GE, September 10, 1996 i

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g p

G k 2 UNITED STATES NUCLEAR REGULATORY COMMISSION f WASHINSTON, D.C. 20655-CM1

...../ September 10, 1996 Mr. Craig P. Kipp Plant Manager General Electric Nuclear Energy Nuclear Energy Production P.O. Box 780, Mail Code A20 Wilmington, NC 28402-0780

SUBJECT:

NRC INSPECTION REPORI N0. 99900003/96-01

Dear Mr. Kipp:

On May 10, 1996, the U.S. Nuclaar Regulatory Commission (NRC) completed an inspection of the General Eier tric Company (GE) activities at the Nuclear Energy Production facilities in Wilmington, North Carolina. This letter transmits the report of that inspection.

During this inspection, the Ni'C inspection team did not identify anj instances where your evaluation of the safety limit minimum critical power rat!o (SLMCPR) errors or the notificat Mn of NRC licensees failed to meet NRC requirements in Part 21 of Title 10 of the Code of Federal Reaulations (10 CFR). However, the team concluded that GE could have taken actions to evaluate the SLMCPR problem earlier than its " discovery date" of March 28, 1996, when GE initiated its potentially reportable condition evaluation. The team determined that in August 1995, GE determined that it was possible to have an R-factor distribution that would yield a higher number of fuel rods that would approach boiling transition than the bounding values used in its generic SLMCPR analyses. This information was contained in an internal GE document titled " Discussion of Deviation from Design Procedure," August 1995, based on the SLMCPR analysis for the Kernkraftwerk Krommel plant.

Pursuant to an April 17, 1996, meeting with NRC staff, CE provided a list of 13 NRC licensee plants that were most likely to be impacted by the SLMCPR errors. Letter notifications that included plant specific status were transmitted to all GE fueled boiling-water reactors (BWRs) by letters dated April 18, 1996, although some licensees were . orally notified a few days earlier. Thus, the team found that from the time the potential design defect l

was identitled (March 28,1996), approximately one month elapsed before GE notified licensees of needed interim corrective actions. GE's failure to evaluate its potential SLMCPR problem when indications of the problem first appeared in August 1995 and to timely notify licensees of the need to take interim corrective actions is considered a weakness in GE's responsiveness to the SLMCPR issue.

' In its May 24, 1996, letter, "10 CFR Part 21, Reportable Condition, Safety Limit MCPR Evaluations," GE informed NRC of the reportable condition finding

'or its SLMCPR evaluations. The 10 CFR Part 21 notification identified 11 NRC licensed plants that have SLMCPR limits that are above their currently licensed technical specification SLMCPR values. GE's reanalyses of 29 domestic GE fueled BWRs was reportedly complete on May 29, 1996, over two months from the official " discovery date" and approximately nine months after the problem was initially documented by GE in August 1995.

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C. Kipp The team concluded that GE's response to the SLMCPR plant specific evaluations was inadequate to provide clear and accurate corrective action information to NRC licensees on a schedule appropriate for the safety issues involved. The team also concluded that the clarity and effectiveness of GE's notifications could have been enhanced if GE had placed more attention to timely completion of the plant specific evaluations and to reporting and explaining the plant specific results rather than deemphasizing the impact of the SLMCPR error on affected plants. .

During this inspection the NRC team determined that the implementation of your quality assurance program failed to meet certain NRC requirements imposed on you by licensees and NRC approved methodology. The most significant  ;

nonconformance was the team's finding that GE failed to recalculate or reconfirm the applicability of the generically determined SLMCPR to new fuel [

bundle designs as required by Appendix B to 10 CFR Part 50, and Amendment 22 of the " General Electric Standard Application for Reload (GESTAR) II" topical report documented in NEDE-240ll-P-A, " General Electric Standard Application For Reactor Fuel" (approved by the NRC on July 23,1990). GE's failure to recalculate or reconfirm the applicability of the generically determined SLMCPR to new fuel bundle designs resulted in licensees operating 11 reactor i cores with incorrect and nonconservative technical specification safety limit l MCPR values. )

In addition, the staff notes that certain licensees failed to verify that GE complied with the NRC approved methodology required by the plant technical ,

specifications. l The team found that your recent generic SLMCPR evaluations did not comply with the NRC approved search procedure that assures a conservative power distribution representative of the limiting permissible rod configurations.

Specifically, the input power distribution assumptions employed by GE for ,

recent generic analyses did not conform to the inputs that were implicit to  :

the NRC approved methods in " General Electric BWR Thermal Analysis Basis (GETA8): Data, Correlation and Design Application," NEDE-10958-PA, January 1977, and referenced in GESTAR. In addition, the cycle-specific evaluations were based on planned operating conditions during the cycle and did not bound all rod configurations permitted by the plant technical specifications and ,

operating controls. The team also found that the R-factor definition was changed for Gell fuel; and the R-factor data used in the Gell SLMCPR analysu was not adequately verified and was not bounding for many plant specific Gell fuel designs.

These nonconformances 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 i should follow the instructions specified in the enclosed NON when preparing  !

your response. ]

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C. Kipp In accordance with Section 2.790(a) of 10 CFR, a copy of this letter and its enclosure will be placed in the NRC Public vocument Room and made available to the public unless you notify this office by telephone within 10 days of the date of this letter and submit a written application to withhold the information contained therein. Such application must be consistent with the requirements of 10 CFR 2.790(b)(1). Your response to this letter and its enclosure is not subject to the clearance procedures of the Office of Management and Budget, as required by the Paperwork Reduction Act of 1980, Public Law No.96-511.

l Should you have any questions concerning this inspection, we will be pleased ,

j to discuss them with you. Thank you for your cooperation during this process. f

! Sincerely, l

iRo Mert M. Gallo, Ciief l Special Inspection Branch i Division of Inspection l

and Support Programs ,

Office of Nuclear Reactor Regulation  !

Docket No.: 99900003 ,

Enclosures:

1. Notice of Nonconformance

2. Inspection Report 99900003/96-01 l

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

l l NOTICE OF NONCONFORMANCE t

General Electric Nuclear Energy. Docket No.: 99900003 Nuclear Energy Production Wilmington, NC i

On the basis of the results of an inspection conducted from May 6 through 10, i 1996, it appears that certain of your activities were not performed in  ;

accordance with requirements of the Nuclear Regulatory Commission (NRC).

Criterion III, " Design Control," of Appendix _B to Part 50 of Title 10 of the Code of Federal Reaulations (10 CFR), requires, in part, that design control measures shall be applied to reactor physics, stress, thermal, hydraulic, and L accident analyses, and shall provide for verifying or checking the adequacy of l the design. I Section 3.8, " Design Verification," of "GE Nuclear Energy, Quality Assurance Program Description," NED0-ll209-04A, Revision 8, Class 1 (approved by the NRC- l l on March 31, 1989, as meeting the requirements of Appendix B to 10 CFR ,

Part 50) requires, in part, that design verification is a process for independent review of designs against design requirements to confirm that the designer's methods and conclusion are consistent with requirements and that the resulting design is adequate for its specified purpose.

Section 1.1.5.A of Amendment 22 of the " General Electric Standard Application for Reload (GESTAR).II" topical report documented in NEDE-24011-P-A, " General Electric Standard Application For Reactor Fuel" (approved by the NRC on July 23,1990) requires, in part, that the safety limit minimum critical power l

ratio (SLMCPR) _shall be recalculated _ following the steps in 1.1.5.B or reconfirmed when a new fuel design or new critical power correlation is introduced. Section 1.1.5.8 describes the reference core conditions and input assumptions to be used when performing the SLMCPR calculations. Section 1.1.7.C describes the criteria for establishing new critical power correlation and refers to NRC approved " General Electric BWR Thermal Analysis Basis (GETAB): Data, Correlation and Design Application," NEDE-10958-A, January 1977, to determine coefficients in the correlation.

Appendix IV-4, "Effect Of Power Distribution On Statistical Rod Boiling Transition Analysis," of GETAB requires, in part, the rod patterns chosen are l l those which maximize the assembly powers in an annular zone of maximum radius.

L A. Contrary to the above, GE failed to recalculate or reconfirm the applicability of the generically determined SLMCPR to new fuel bundle designs. GE's failure to recalculate or reconfirm the applicability of the generically determined SLMCPR to new fuel bundle designs as required ,

by the NRC approved methods in GESTAR resulted in eight licensee reactor l cores loaded with Gell fuel and three reactor cores loaded with GE9 fuel l operating with incorrect and nonconservative SLMCPR technical  !

specification limits. =

(Nonconformance 99900003/96-01-01) f I

I Enclosure 1 76

s

i. .

B. -Contrary to the above, GE's input assumptions used in recent generic l SLMCPR analysis did not provide for a large annular peak power region ,

placing a large fraction of fuel bundles near the limit a; eauired by  ;

i the NRC approved methods in GETAB.

(Nonconformance 99900003/96-01-02)

C. Contrary to the above, GE's R-factor data used in the Gell SLMCPR analysis was not adequately verified since the R-factors were not bounding and the design verification of the R-factor data used in the ,

Gell SLMCPR analysis failed to identify this deficiency.

(Nonconformance 99900003/96-01-03)

Please provide a written statement or explanation to the U.S. Nuclear I Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555, with a copy to the Chief, Special Inspection Branch, Division of Inspection i 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 t

! 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 the corrective actions and preventive measures were or will be completed.

l l

i Dated at Rockville, Maryland l this 10 day of September, 1996 l l

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REVISED NONPROPRIETARY VERSION U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION Report No.: 99900003/96-01 Organization: General Electric Company (GE)

General Electric Nuclear Energy (GE-NE)

Nuclear Energy Production (NEP)

Wilmington, North Carolina

Contact:

Ralph J. Reda, Manager Fuels and Facility Licensing Nuclear Industry Activity: GE provides boiling-water reactor (BWR) reload core designs, safety analysis, and licensing, fuel assemblies, fuel-related core components, and fuel-related inspection services to the U.S. nuclear industry.

Dates: May 6 - 10, 1996 Inspectors: Steven M. Matthews, DISP /PSIB Carl E. Beyer, Pacific Northwest National Laboratory Dr. John F. Carew, Brookhaven National Laboratory Dr. Tai L. Huang, DSSA/SRXB Donald A. Lampe, Par & meter, Inc.

Kamalakar R. Naidu, DISP /PSIB Laurence E. Phillips, Chief, DSSA/SRXBA Kombiz Salehi, RGN-III/DRS Dr. Shih-Liang Wu, DSSA/SRXB Approved by: Gregory C. Cwalina, Chief Vendor Inspection Section Special Inspection Branch Division of Inspection and Support Programs Enclosure 2 78

I 1 INSPECTION

SUMMARY

From May 6 through 10, 1996, the U.S. Nuclear Regulatory Commission (NRC) conducted a performance-based inspection of the General Electric (GE) Nuclear Energy (GE-NE), activities at the Nuclear Energy Production (NEP) facilities in Wilmington, North Carolina. This inspection focused on technically i directed observations and evaluations of GE activities related to (1) under prediction of the safety limit minimum critical power ratio (SLMCPR) for several licensee plants and (2) out-of-specification low density fuel pellets (LDPs) that were loaded into fuel rods and shipped to licensees in fuel I

assemblies.

The inspection bases were:

=

General Design Criterion (GDC) 10, " Reactor Design," and GDC 12, "3uppression of Reactor Power Oscillations," of Appendix A, " General Design Criteria for Nuclear Power Plants," to Part 50, " Licensing of Production and Utilization Facilities," of Title 10 of the Code of Federal Reaulations (10 CFR).

• Appendix 8, " Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," to 10 CFil Part 50.

• 10 CFR Part 21, " Notification of Failure to Comply or Existence of a Defect."

• Section 4.2, " Fuel System Design," of NRC NUREG-0800, " Standard Review Plan" (SRP), Revision 2, dated July 1981, and its Appendix A,

" Evaluation of Fuel Assembly Structural Response to Externally Applied Forces," Revision 0. l l

• "GE Nuclear Energy, Quality Assurance Program Description," NE00-Il209- l 04A, Revision 8, Class 1 (approved by the NRC on March 31, 1989, as l meeting the requirements of Appendix B to 10 CFR Part 50), hereafter I referred to as the "QA topical report." ,

• Amendment 22 of the " General Electric Standard Application for Reload (GESTAR) II," topical report documented in NEDE-24011-P-A, " General Electric Standard Application For Reactor Fuel," (approved by the NRC on July 23,1990).

• " General Electric BWR Thermal Analysis Basis (GETAB): Data, Correlation and Design Application," NEDE-10958-PA, January 1977 (approved by the NRC and referenced in GESTAR).

During this inspection, three instances were identified where GE failed to conform to NRC requirements and approved methodology. These nonconformances are discussed in Section 3.1 of this report.

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During this inspection, the team noted weaknesses and observations concerning GE activities that affect quality. Neither the weaknesses nor the observations described in this report require any specific action or written response.

2 STATUS OF PREVIOUS INSPECTION FINDINGS The open findings from the previous NRC inspection of GE (conducted from August 14 through September 1, 1995, and documented in Inspection Report 99900003/95-01, March 5, 1996) were not reviewed during this inspection.

However, concerns associated with the determination of the SLMCPR that were identified during the previous inspection were reviewed during this inspection.

3 INSPECTION FINDINGS AND OTHER COMMENTS This inspection included an evaluation of the adequacy of GE processes and activities for determining the safety limit minimum critical power ratio (SLMCPR) which ensures that 99.9% of the fuel rods avoid boiling transition and out-of-specification low density Nel pellets (LDPs). The emphasis of the team's evaluation was on the fuel development and engineering, corrective actions, the requirements of 10 CFR Part 21, and GE licensee notifications related to the SLMCPR and LDP issues.

3.1 Safety Limit Minimum Critical Power Ratio (SLMCPR)

a. Inspection Scope On March 16, 1995, GE met with NRC staff and stated that they would submit a request to replace the current generic SLMCPR analysis with a cycle-specific analysis. During the presentation GE stated that  ;

(1) conservative generic analysis restricts benefits of more efficient designs, (2) actual designs show higher bundle peaking factors permitted by increased thermal margin, (3) cycle-specific evaluations can reflect more realistic radial power assumptions, and (4) its cycle-specific SLMCPR methodology would be submitted to NRC in 1995.

During the NRC inspection of NEP activities conducted from August 14 through September 1, 1995 (NRC Inspection Report 99900003/95-01, March 5, 1996) the team followed up on GE's March 16, 1995, presentation t the NRC staff by evaluating GE's methodology for determining the blMCPR.

That inspection identified several concerns associated with the determination of SLMCPR. Those concerns related to areas where significant changes had been made to the methodology used to determine the SLMCPR, which were considered to be outside the NRC approved generic method described in GESTAR. For instance, the revision to the R-factor methodology used in determining the SLMCPR for the Gell, GE12, and GE13 ,

fuel designs was of special concern since (1) the SLMCPR calculated with this new methodology increased from 1.07 for Gell fuel to 1.09 for GE13 fuel even though an additional spacer had been added in the GE13 design for critical power ration (CPR) improvement and (2) the revised methodology had not been submitted for NRC review.

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During this inspection, the team questionzd GE staff regarding its  :

response to the concerns raised by the previous NRC inspection team. i The team learned through interviews and discussions with GE staff that, aside from submitting the revised fuel rod R-factor model, NEDC-32505-P, ,

"R-Factor Calculation Method for Gell, GE12, and GE13 fuel," GE had  ;

taken no action with regard to the concerns raised by the previous NRC ,

I team.

• On March 27, 1996, GE advised NRC by telephone that its SLMCPR error '

issue was under review.as a potential reportable concern (PRC). GE stated that the generic Gell SLMCPR of 1.07 was not " bounding" for a  !

licensee operating in a 2-year cycle with a large batch fraction (37%)

(

of fresh Gell fuel. GE attributed the SLMCPR error to an analytical >

procedure deficiency and claimed to be evaluating all plants as rapidly  ;

as possible to assure that the plant / cycle-specific SLMCPR value was t bounded by the generic SLMCPR used as the licensing basis.

I i

However, during its April 17, 1956, meeting with the NRC staff (several ,

licensees were also present at the meeting), GE stated that the l erroneous safety limit MCPR values were due to nonconservatism in the  ;

l l

generic analysis methods and may not be restricted to Gell fuel and l large batch fractions.

The generic analysis for each fuel type (e.g., Gell, GE12, or GE13) is based on the analysis of a reference equilibrium core of the fuel type at the cycle exposure corresponding to the most limiting exposure state point. The reference core is designed with flat fuel rod bundle power distribution and core radial power distribution such that the generic SLMCPR value determined for the reference core was expected to bound all l

I reload core applications. GE discovered that this was not valid and concluded that cycle-specific analyses of each reload core was necessary l to confirm that the generic SLMCPR value t smained bounding for each GE fueled BWR licensee. GE indicated that they had performed preliminary cycle-specific analyses and based on the results, had informed licensees and recommended corrective actions for 13 plants most likely to-be impacted by the cycle-specific analyses.

f Therefore, the primary focus of this inspection was the nonconservative L under prediction of the Gell SLMCPR. In Section 3.1.b.4 of this report j the team traced the under prediction of SLMCPR to the miscalculation of l l the R-factors used in the engineering computer program GESAM for SLMCPR analysis. l l

b. Observations and Findines I b.1 Cycle-Specific Safety Limit MCPR Methodology As part of the evaluation of the cycle-specific SLMCPR methodology, the o

team reviewed the recent application of this method to the Edwin I.

Hatch unit 1 Cycle 17 SLMCPR reevaluation. The Hatch unit 1 Cycle 17 licensed safety limit MCPR is 1.07 and the initial GE screening evaluation based on the GESAM DIST analytical method also indicated a i

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- value of 1.07, at the limiting peak-hot-excess (PHE) reactivity  !

statepoint. The detailed cycle-specific GESAM Monte Carlo calculation, i which was completed and verified on May 3,1996, resulted in a SLMCPR of I 1.06. The Hatch unit 1 Cycle 17 calculation, together with all the  ;

SLMCPR reevaluations, was documented in DRF Jll-0286. The new cycle-  !

specific methodology included an extensive design verification checklist for both the PANACEA design basis statepoint evaluation and the GESAM-OLCPOW, Monte Carlo analysis. The Hatch unit 1 Cycle 17 SLMCPR analysis l was performed for the beginning-of-cycle (BOC), PHE reactivity, and end-l of-cycle (EOC) statepoints. ' The analysis used the approved set of plant L instrumentation and monitoring uncertainties. i

i The principal differences between GE's implementation of the NRC approved generic SLMCPR method (described in Amendment 22 of GESTAR) and the cycle-specific method, as described in the design record files (DRFs) are the determination of the core loading and the design basis I reactor statepoint. In the generic methodology, a conservative bounding )

core is established by selecting a large, high power densi;f,_  :

' equilibrium core with a large fuel batch fraction. In the cycle- )

specific approach the actual core loading (e.g., Hatch unit 1 Cycle 17) i is assumed, including the cycle-specific local fuel bundle exposures and  ;

fuel rod R-factors. In the generic method, the fuel bundle design (enrichments, spacers, etc.) and bundle exposure's are selected to give the flattest (most conservative) R-factor distribution.

The SLMCPR is very sensitive to the number of fuel bundles close (to i within ~0.10 AMCPR) to the MCPR limiting bundle, and the determination I of the core design basis CPR distribution is a critical step in the SLMCPR analysis. The GE procedure (Y1003C04, "GETAB Safety-Limit,"

Revision 2) used to implement the generic method recommends (but does not require) that the search for the design basis statepoint be performed by adjusting the rod pattern to maximize the power in an annular zone', under xenon-free conditions in order to maximize the control rod inventory and, consequently, the radius of the high powered annular zone. The team concluded that, by only recommending the method to search for the design M is statepoint instead of requiring the method to be used, GE's search method may not in all instances conform the NRC approved methods described in GETAB. For instance, through discussions with the GE staff, the team learned that the GE12 design basis power distribution included a centrally located cylindrical peak power region rather than an annular peak power zone.

j The NRC approved methodology described in GETAB requires that the searen for the design basis statepoint be performed by adjusting the rod pattern to maximize the power in an annular zone. Therefore, the team concluded that the input assumptions used in recent generic SLMCPR analysis of GE12 (and possibly other fuel designs) did not provide for a large annular peak pou r region placing a large fraction of fuel bundles near the limit.

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l t

The cycle-specific method makes no recommendation concerning the radial l

shape of the high powered region, however, all calculations-to-date (including Hatch unit 1 Cycle 17) have used a centrally located l cylindrical region. In addition, the cycle-specific calculations have l used equilibrium xenon conditions rather than the xenon-free conditions recommended in the generic method.

In both methods, the design basis statepoint is determined by performing a series of PANACEA (core analysis) calculations to obtain a bounding CPR distribution. In the cycle-specific method, the search is continued l

until the limiting bundle is at the operating limit MCPR and 10% of the fuel bundles are within 0.20 EPR of the limiting bundle. If this is not possible, after an unspecified number of trials, the MCPR distribution is renormalized so that the peak bundle is at the operating limit (if necessary) and the current CPR distribution is taken to be bounding. In the case of Hatch unit 1 Cycle 17, the 10% search criteria was satisfied for the BOC and E0C statepoints, however, for the PHE reactivity statepoint only 6.4% of the fuel bundles were brought to within 0.20 E PR of the limiting bundle and, therefore, the search criteria was not satisfied. The team concluded that this less conservative PHE statepoint was responsible, in part, for the reduced PHE SLMCPR of 1.03, relative to the 1.06 SLMCPR obtained for the B0C and EOC statepoints.

In recent applications of the generic method, bundle CPR distributions which are more conservative than the di '.ributions resulting from the cycle-specific 10% search criteria have been used. For example, the GE9 D-lattice and C-lattice bounding CPR distributions included ~20% and l

~15% respectively, of the fuel bundles within 0.20 EPR of the limiting bundle. The Gell SLMCPR analysis included ~20% of the fuel bundles ,

within 0.20 EPR of the limiting bundle. I 1

It is important to compare these recent SLMCPR statepoint distributions l with the design basis bundle power distribution provided in GETAB. l Bundle power distributions are compared since bundle CPR distributions were not provided in GETAB. The GETAB statepoint was conservatively flat and included ~28% of the fuel bundles within 10% of the peak powered bundle. The GE9 design basis statepoint included ~8% of the fuel bundles within 10% of the limiting bundle and the Gell statepoint included ~4% of the bundles within 10% of the limiting bundle. The team concluded that these comparisons show that the core statepoina used by GE in the recent SLHCPR calculations are not as conservative as that shown in the NRC approved GETAB statepoint methodology. Therefore, GE's failure to search for the design basis sta Npoint by maximizing the power in an annular zone constitutes Nonconformance 99900003/96-01-02.

In both the cycle-specific and generic methods, the search for the '

bounding CPR distribution is performed by varying the control rod pattern. In the cycle-specific method only the positions of the inserted rod sequence group (e.g., A2) are varied. However, in the generic method the search is not restricted to a specific rod group and resuits in a more conservative SLMCPR value.

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Based on discussions with the GE staff and the review of the DRFs, the team found that there were certain features of the generic analysis selection of the bounding plant, core conditions, fuel bundles and control rod pattern that, when taken together, provide a high level of confidence that the SLMCPR is bounding when applied to a specific plant.

Additionally, the team identified concerns with several aspects of the cycle-specific method that may result in a non-conservative value for the SLMCPR and are, therefore, considered by tne team to be weaknesses in the cycle-specific method. These weaknesses are summarized in the following:

(1) Design Basis CPR Distribution i The cycle-specific SLMCPR calculations only require 10% of the i fuel bundles to be within 0.20 EPR of the limiting bundle. The l generic method, characterized by an annular region of limiting '

bundles, maintains ~20% of the fuel bundles within 0.20 EPR of I the limiting bundle. In addition, the recent design basis '

statepoints appear to be less conservative than the GETAB statepoint. Since the SLMCPR is very sensitive to the number of fuel bundles close to the limiting bundle, these differences are expected to have a significant effect on the SLMCPR.

(2) Determination of the Limiting Control Rod Pattern The cycle-specific method determines the limiting control rod

pattern by varying only the inserted control rod sequence. In the i generic method, no constraint is placed on the selection of the control rods and a more conservative SLMCPR results.

(3) Termination of the Search for Maximum Safety Limit MCPR l

I

The procedure for terminating the search for the maximum SLMCPR in j the cycle-specific method allows the search to be terminated

, without satisfying the intended search criteria. Consequently, i

this procedure does not ensure that a bounding SLMCPR is determined. The team noted that in the case the search procedure does not ensure a bounding SLMCPR, a conservative addition to the SLMCPR may be required.

(4) Xenon-Free Design Basis CPR Distribution in order to maximize the control rod inventory and explind the f search for the maximum SLMCPR, the generic method assumas the 1 design basis statepoint is xenon-free. The cycle-specif N search for the bounding SLMCPR is more limited in that equilibr i xenon i

conditions are assumed.

84

The teaa considered these issues to be a weakness in the GE's cycle-l specific SLMCPR procedures to ensure that a bounding SLMCPR is determined, as required by the NRC approved methodology described in

! GESTAR.

b.2 Underestimate of Gell Safety Limit NCPR l The SLMCPR is determined by (1) the GEXL, plant instrumentation and core

! monitoring uncertainties, (2) the design basis bundle-wise CPR distribution, and (3) the local bundle R-factor distribution. The DRFs for the generic GE9 and Gell SLMCPR analyses, and the more recent cycle-specific SLMCPR evaluation for Hatch unit 1 Cycle 17 were reviewed to determine the cause of the underestimate of the Gell SLMCPR. The DRF l

documentation indicated that the instrumentation and monitoring uncertainties used in both the Gell and GE9 SLMCPR analyses were identical, and that the GEXL uncertainty was slightly higher for Gell  :

(3.1 vs. 3.0%). The uncertainties used were consistent with NRC approved values.

The design basis bw 'le-wise CPR distributions for the GE9, Gell and Hatch unit 1 Cycle 1/ iLMCPR analyses were also compared. The Gell CPR distribution was comparable or flatter (more conservative) than the GE9 and Hatch unit 1 Cycle 17 distributions, with more than 20% of the fuel bundles within 0.20 oCPR of the limiting bundle.

Both the Gell and Hatch unit 1 Cycle 17, GE12 fuel designs include part length rods and the R-factors are calculated with the new R-factor model described in NEDC-32505-P [ procedure no. corrected by staff), described in Section 3.1.b.3 of this report. A comparison of the Gell and Hatch unit 1 Cycle 17 GE12 pin-wise R-factor distributions indicated that the Gell R-factor distribution was highly peaked and nonconservative compared to the relatively flat Hatch unit 1 Cycle 17 R-factor distributions. The pin-wise Gell D-lattice R-factors ranged from 0.62 to 1.01, while the range of the typical Hatch unit 1 Cycle 17 GE12 D-lattice R-factors only varied from 1.025 to 0.89.

In view of the agreement and/or conservatism of the uncertainties and the CPR distribution, the team concluded that the use of the nonconservatively peaked R-factors was responsible for the under prediction of the Gell SLMCPR. In addition, the team noted that the large variation observed in the Gell R-factor distribution (0.62 to 1.01) typically only occurs at B0C and that B0C fuel bundles may have i been used in the Gell SLMCPR analysis. In order to understand the source of the incorrect R-factors (e.g., use of B0C fuel bundles, changes in the definition / application of the R-factor model) and i completely define the impact, a detailed description of the determination of the Gell R-factors is required. The team considered the lack of a detailed description of the source of the incorrect R-factors and GE's failure to perform a root-cause analysis to be a weakness in GE's response to this issue; as raised by the previous NRC inspection team and documented in Inspection Report 99900003/95-01.

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The bundle R-factors for the Gell fuel design were calculated by NEP's Nuclear Fuel Design Group and transmitted by letter to the group performing the SLMCPR analysis. While the transmittal stated that the fuel bundles were especially designed to provide the flattest R-factor distributions, no verification of the R-factor data was indicated. The Gell DRF included a design verification checklist which confirmed that bounding R-factors were used in the SLMCPR determination.

Since the R-factors were actually not bounding, the team concluded that the design verification of the R-factor data used in the Gell SLMCPR analysis was not adequate and the design verification of the R-factor data used in the Gell SLMCPR analysis failed to identify this deficiency. Therefore, GE's failure to adequately verify the R-factor l data constitutes Nonconformance 99900003/96-01-03.

b.3 Revised R-Factor Model The Gell fuel design included part length rods and required the I development of the new calculational model and procedure for determining I the bondie R-factors. As noted in S3ction 3.1.a of this report, this '

new R-factor definition and application were of concern to the previous NRC inspection team and resulted in a request for a detailed description of the new R-factor methodology. In response to the previous NRC inspection (Inspection Report 99900003/95-01, March 5, 1996) GE provided NRC a description of the revised fuel rod R-factor model (NEDC-32505-P, "R-Factor Calculation Method for Gell, GE12, and GE13 Fuel"). This model includes several new features including (1) a change in the i definition of the R-factor model in which the effects of adjacent rods i are based on the axially integrated rod power, rather than the local axially-dependent rod power, (2) the introduction of " correction factors" in the definition of the R-factor for the Gell and GE13 fuel designs, and (3) a change in the weighting factors for adjacent fuel l rods.

On the basis of its review of NEDC-32505-P and discussions which the GE staff, the team concluded that these changes are not simply changes in GEXL input but rather constitute basic changes in the GEXL/SLMCPR methodology that require NRC review. Also on the basis of its review of this method and the Gell R-factors, it appeared to the team that the new R-factor methodology was either not applied or applied incorrectly in the determination of the Gell SLMCPR.

b.4 Nonconservative Safety Limit MCPR Because of the complex and sensitive dependence of the SLMCPR on the bundle-wise CPR and the rod-wise R-factor distributions, a definitive observation of a nonconservative SLMCPR generally requires a complete engineering computer program GESAM/0LCPOW Monte Carlo reevaluation.

GESTAR requires that the SLMCPR be reevaluated for new fuel bundle designs which involve changes to the enrichment or spacer designs.

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However, the calculations involved are extremely long-running and GE typically determines the SLMCPR for each fuel type (e.g., Gell) but does not reevaluate the SLMCPR for subsequent cycle-specific core / bundle designs. Consequently, the nonconservative SLMCPR was not identified until a complete SLMCPR reevaluation was performed during the 1

application of the cycle-specific methodology to a mixed-core reload.

7 During the initial GE application of the cycle-specific methodology to Grand Gulf Cycle 9, the cycle-specific SLMCPR was calculated to be greater than the generic maximum SLMCPR. Since the Grand Gulf [ plant reference corrected by staff] core included both GE and Siemens Power Corporation fuel bundles, the cycle-specific / generic SLMCPR comparisons were inconclusive. In order to completely understand this difference, i GE immediately performed a cycle-specific analysis for River Bend Cycle

7. This analysis also resulted in cycle-specific values that exceed the i generic Gell SLMCPR, and show that the generic analysis was not bounding l as assumed.

The GE staff stated that this analysis was completed at the end of i February 1996, and that this was GE's first indication that the Gell l SLMCPR analysis was not bounding. However, GE's preliminary safety concern (PSC) 9608, to review the potential nonconservatis.m in the SLMCPR to determine if a potential reportable concern (PRC) evaluation should be initiated was not started until March 14, 1996. (This issue is discussed further in Section 3.3 of this report.) l l

The team concluded that in addition to the fuel design SLMCPR l reevaluation, another opportunity existed for GE to identify the nonconservative SLMCPR. In certain cases, the relative magnitudes of the SLMCPR may be determined by ccy aring the input distributions. For l example, if the R-factor distribution for a given bundle is always larger (i.e., the R-factor for each rod is higher), then the SLMCPR for this bundle design will be larger. Recognizing that the R-factors used in the Gell SLMCPR analysis were not bounding when compared to many other fuel bundles, a comparison of available R-factor distributions would have indicated that the Gell SLMCPR was nonconservative. The team concluded that this comparison was apparently not made by GE because the Gell R-factors were assumed to be bounding.  ;

The SLMCPR/GESAM program for the cycle-specific SLMCPR analyses is a recently developed program based on the OLCPOW program referenced in GESTAR. In response to the team's request, during this inspection, to produce benchmark results showing the validation of GESAM versus OLCPOW,  ;

GE submitted a letter to NRC staff after the inspection, RJR-96-065, June 12, 1996, attesting that the two codes are equivalent. During the PRC (10 CFR Part 21) evaluations, GE relied on procedure restrictions (a check list) and independent design verification for quality assurance of the new methods. The letter indicates that the restrictions and design verification requirements will be reduced after the ongoing qualification of GESAM in accordance with the QA program.

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The pr:cedure restrictions and design verification requirements alluded to by GE were reviewed during this inspection and appeared to be reasonable for the interim. However, the team identified some calculation procedure weaknesses that are of concern. In addition, QA of the SLMCPR methods for GESAM calculations had not been performed in accor:iance with the NRC approved QA topical and the team did not find any e.'idence that the GESAM calculation results had been validated.

b.5 Quality Assurance As a result of its review of the generic and cycle-specific determinations of the SLMCPR, as documented in the DRFs and supporting procedures, the team identified several concerns associated with the implementation of the GE QA topical. These concerns are (1) the calculation / data verification performed for the Gell SLMCPR analysh (2) the Gell and subsequent reload design reviews, and (3) the generic method used to perform the SLMCPR analysis.

In addition, the team concluded that both the Gell SLMCPR design review and the subsequent cycle-specif:: reload core design reviews failed to recognize the need to reconfirm that the Gell SLMCPR R-factors remain bounding. Also, recognizing that the Gell and GE13 design basis core power shapes are similar, the increase of the SLMCPR from 1.07 for Gell i to 1.09 for GE13 should have shown GE the need to review both the Gell and GE13 R-factors.

1 The generic method used to determine the SLMCPR is described in the GE I procedure Y1003C04, "GETAB Safety Limit," Revision 2. This procedure was written in 1977 and does not correspond to present GE practice. For example, the generic procedure recommends that the design basis statepoint should include an annular region of CPR limiting bundles under no-xenon conditions, while the current practice is to calculate the design basis statepoint for a centrally located cylindrical region of CPR limiting bundles under equilibrium xenon conditions. While this procedure was identified for review and updating on March 19, 1996, the team concluded that the adequacy of the periodic review and updating of these SLMCPR procedures was a weakness.

According to GE QA staff interviewed by the team, no deviation from a procedure involving safety related analysis was permitted except to implement a deviation by means of a design review. This step was followed by GE and was documented in an April 15, 1996, memo, DRF-J11-02861, "SLMCPR Calculation Process Design Review Report." The memo documents the approach that will be used for performing cycle-specific analyses to determine the SLMCPR. It identifies those areas in the current procedure Y1003C04 from which a deviation will be made. The team concluded that GE conformed to their QA procedures with respect to initiating a deviation from an existing safety related procedure.

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The team reviewed the method of verifying the cycle-specific analyses contained in DRF-Jll-02861. This reviewed showed that the verifier limited the verification to a check of the checklist prepared by the Responsible Engineer for the analyses. Although no restrictions are l r placed on the verifiers activities, the team's review of the i calculational files showed that the independent verifier did not deviate l from the scope of the verification or the method of verification specified by the Responsible Engineer. The team concluded that this lack of independent activity by the verifier was a weakness in GE's verification process.

In the generic method, the SLMCPR is determined by a Monte Carlo ,

calculation performed using the OLCPOW option of the GESAM program. The l GESAM/0LCPOW code is not a Level-2 program and, therefore, requires independent verification when applied in the SLMCPR analysis. This ,

verification is performed using the DIST analytical method. I l

Based on a review of the GESAM/0LCPOW and DIST calculations and discussions with the GE staff, the team concluded that the DIST method is not sufficiently accurate to provide the required independent verification of the SLMCPR calculation. The team, therefore, concluded that the verification of the GESAM/0LCPOW program was not adequate.

b.6 Mixed Core Analyses In the case of a mixed core including both GE and Siemens fuel bundles, a new GEXL correlation is determined for the Siemens fuel. The new correlation is based on CPR data calculated (by the licensee) using the Siemens CPR correlation rather than actual measured data. The calculated data is fit by using the GEXL correlation and adjusting the definition of the R-factor to obtain optimum agreement. The uncertainty in the new GEXL correlation, for application to Siemens fuel, is determined by a statistical (sum-of-squares) combination of (1) a conservative estimate of the Siemens CPR correlation uncertainty and (2) the uncertainty in the fit of the new GEXL correlation to the calculated data. This statistical combination of the uncertainties assumes that the fitting error of the Siemens correlation to the experimental data and the fitting error of the new GEXL correlation to the Siemens correlation are indeper. dent. However, the team found that GE had not documented a justification for this assumption nor was one l provided to the team. l 1

The team concluded that mixed cores had no direct impact on the input '

nonconservatisms previously discussed or on the failure to reconfirm the acceptability of fuel type design changes. However, the team did note that GE identified an additional error involving a failure to properly count the number of fuel pins that represent 0.1% of the core when mixed fuel bundles with different fuel rod arrays coexist. l 89

bc7 3-D MONICORE System The SLMCPR provides the critical power margin required to account for the uncertainty in the on-site core surveillance predictions and the uncertainty in the GEXL correlation. The SLMCPR is determined by a Monte Carlo uncertainty-propagation simulation in which the core surveillance instrumentation and input data are varied randomly, based on their expected uncertainties. The core surveillance at BWR plants is

. generally performed with the advanced 3-D MONICORE system, although one plant is still using the older P-1 core performance algorithm. While there are substantial differences between these two core surveillance methods, the SLMCPR calculation is currently performed using the older P-1 method.

In response to the inspection team's concerns raised during the previous August 1995 inspection, GE stated that the determination of the SLMCPR is not affected by these 3-D MONICORE/P-1 differences, and the current SLMCPR calculation is applicable to both plants using the 3-D MONIC0RE or the P-1 core surveillance method. However, in view of the several recent concerns associated with the SLMCPR, the team concluded that the inconsistency between the actual 3-D MONICORE plant monitoring and the assumed P-1 monitoring may result in a nonconservative determination of the SLMCPR. The team, therefore, considered this to be a weakness in the GE method for determining the safety limit MCPR.

b.8 Safety Limit MCPR Error Prior to this inspection of GE, GE attributed the erroneous safety limits MCPR for several plants to large batch fractions of Gell fuel in core reloads designed for 24 month operating cycles. The generic analyses for each fuel type is based on the analysis of a reference equilibrium core comprised only of the fuel type being evaluated.

GESTAR approved SLMCPR methods require that the reference core be designed with flat core radial power distribution and flat rod bundle power distribution so that a maximum number of fuel bundles / rods can be placed on limits with a selected control blade pattern. However the team could not confirm that this procedure was followed by GE.

GE expected that the SLHCPR determined for the reference core by analyses at the r.ost limiting exposure state point would be conservative with respect to any actual application. However, GE continued to change the rod powers and enrichment and R-factor distributions of the approved fuel type without further analyses and without confirming that the generic value of SLMCPR remained bounding. Cores consisting of mixed 8x8 and 9x9 fuel oundles (fewer rods in the core) and designed for flatter power distributions also contributed to cycle-specific non-conservative changes in the SLMCPR (i.e., lower number of failed rods needed to reach the 0.1% allowable limit).

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GESTAR and similar Amendment 22 compliance evaluation reports for other GE fuel types clearly identify SLMCPR sensitivity to fuel bundle design parameters affecting the bundle R-factor distribution and the core radial power distribution. The team's evaluation of the Gell reference design used for $!.MCPR determinations showed that the bundle R-factor distributions were much more peaked than typical reference designs and resulted in a non-conservative generic SLMCPR determination. Subsequent i design changes to fuel enrichment level and distribution produced

! flatter bundle R-factor Jistributions to degrade the SLMCPR in subsequent cycle-specific designs..

l Section 1.1.5. A of Amendment 22 of GESTAR (NEDE-24011-P-A), states that l the safety limit MCPR shall be recalculated following the steps in i 1.1.5.8 or reconfirmed when a new fuel design or new critical power correlation is introduced. Section 1.1.5.B describes the reference core l conditions and input assumptions to be used when performing the safety limit MCPR calculations. Section 1.1.7.C describes the criteria for establishing new critical power correlation and refers to GETAB to

determine coefficients in the correlation.

However, GE failed to recalculate or reconfirm the applicability of the generically determined SLMCPR to these new fuel bundle designs as i required by NRC approved methodology in GESTAR. For instance, GE9

, fueled cores were designed with core radial power distributions flatter than that of the reference core without confirming the applicability of l l the generic SLMCPR. Therefore, GE's failure to recalculate or reconfirm i the applicability of the generically determined SLMCPR to these new fuel l bundle designs as required by NRC approved methods in GESTAR constitutes l Nonconformance 99900003/96-01-01.

! c. Conclusions l The team found that the defect in GE's generic SLMCPR determination related to the selection of a reference fuel / core design with input i

condition power distribution that did not satisfy the objective of placing the maximum number cf fuel pins near peak power as required by the NRC approved GETAB statistical analysis procedure. Specifically, the team concluded that the input assumptions used in recent generic SLMCPR analysis did not provide for a large annular peak power region l placing a large fraction of fuel bundles near the limit as required by l the NRC approved methods in GETAB. (See Nonconformance 99900003/96 l 02 described in Section 3.1.b.1 of this report.)

i It appears that the cycle-specific methodology envisioned by GE is a means of eliminating or reducing conservatism employed in the generic analyses in order to compensate for the more limiting SLMCPR values in recent fuel designs. It turns out that GE current methods did not

preclude cycle-specific analyses, and in fact, the methods actually I required that the core and fuel designs be checked prior to the start of the cycle to be sure that previous generic fuel type SLMCPR analyses 1

l 91 t

were bounding from the standpoint of core power distribution of fuel bundle design R-factors. In addition, GE's failure to comply with analysis procedures referenced in plant technical specifications appears to place these licensees in nonconformance with their technical specifications.

The team identified the following concerns with GE's cycle-specific SLMCPR method that may result in a nonconservative SLMCPR value:

(1) the cycle-specific SLMCPR calculations only require 10% of the fuel bundles to be within 0.20 ACPR of the limiting bundle versus the generic method, characterized by an annular region of limiting bundles, that maintains ~20% of the fuel bundles within 0.20 ACPR of the limiting bundle; (2) the cycle-specific method determines the limiting control rod pattern by varying only the inserted control rod sequence versus the generic method where no constraint is placed on the selection of the control rods, therefore, a more conservative SLMCPR results, (3) the procedure for terminating the search for the maxis,w. 5LMCPR in the cycle-specific method allows the search to be trrminated without satisfying the intended search criteria; and (4) in order to maximize the control rod inventory and expand the search for the maximum SLMCPR, the generic method assumes the design basis statepoint is xenon-free where as the cycle-specific search for the bounding SLMCPR is more limited in that equilibrium xenon conditions are assumed.

The team concluded that GE's changes in the R-factor calculation method are not simply changes in GEXL input but rather constitute basic changes in the GEXL/SLMCPR methodology that require NRC review. The team also concluded that the new R-factor methodology was either not applied or applied incorrectly in the determination of the Gell SLMCPR. The team concluded that the R-factor data used in the Gell SLMCPR analysis was not adequately verified since the R-factors were actually not bounding and the design verification of the R-factor data used in the Gell SLMCPR analysis failed to identify this deficiency. (See Nonconformance 99900003/96-01-03 described in Section 3.1.b.2 of this report.)

GE assumed the generic methodology applied to new core designs without checking the applicability of the present methodology to the new core designs. At the time the supplemental reload licensing report is prepared for a particular plant and cycle, GE failed to confirm that the generic bounding SLMCPR analysis did indeed bound the curreni. loading pattern and the current bundle designs. Instead, GE assumed that the bounding analysis was valid. Specifically, GE's failure to recalculate or reconfirm the applicability of the generically determined SLMCPR to new fuel bundle designs was evident in the DRF for the supplemental reload design report. For instance, in the DRF for River Bend Cycle 7 the SLMCPR limits were listed by fuel assembly type but the DRF did not contain a statement that the bundle specific designs had been checked to confirm that the product line bounding analysis was valid. Nor was any comment made as to whether the bounding analysis was valid for the planned core loading pattern.

i l

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GE's failure to recalculate or reconfirm the applicability of the generically determined SLMCPR to these new fuel bundle designs as

! required by the NRC approved methods in GESTAR resulted in eight licensee reactor cores loaded with Gell fuel and three reactor cores loaded with GE9 fuel having SLMCPR limits higher than the licensed values. (See Nonconformance 99900003/96-01-01 described in Section 3.1.b.8 of this report.)

3.2 Low Density Pellets (LDPs)  ;

a. Inspection Scope In February 1996, GE employees found that the as-fabricated fuel pellet density of certain pellets was out-of-specification. These out-of-specification fuel pellets were LDPs and some had been loaded into fuel rods and shipped to various reactor sites. GE's initial assessment determined that the LDP problem was due to incomplete blending of a lubricant called " mixing lubricant" [ revised pursuant to 10 CFR 2.790 -

document described a specific trade name) with UO3 powder. " Mixing lubricant" [ revised pursuant to 10 CFR 2.790 - document described a  ;

specific trade name) is a powder used as a lubricant during the pellet i pressing operation.

Since the " mixing lubricant" (revised pursuant to 10 CFR 2.790 -

document described a specific trade name] process was first introduced l in late 1993, GE traced the production history and determined that the i

LDP problem could cover the period of late 1993 to February 1996. .

Several NRC licensees plants including some foreign reactors were  !

involved. Some of the involved fuel was either recalled or corrected on ,

site. However, some involved fual had been loaded into reactors for  !

irradiation. In addition, GE indicated that several fuel designs were  !

also involved, however, no gadolinia pellets were involved in this LDP problem because gadolinia pellets do not use the " mixing lubricant"  !

[ revised pursuant to 10 CFR 2.790 - document described a specific trade name] process.

Since the LDPs could affect the fuel performance and overall analyses, the team evaluated GE's safety assessment of LDP problem and the '

production of uranium fuel pellets from powder formation to pellet control and distribution for fuel rod loading. .

b. Observations and Findinas b.1 Characterization  :

l GE's initial characterization of the low density pellet (LDP) problem  ;

was summarized a letter to NRC, "GE Fuel Pellet Density Report," April 15, 1996. As a result of the team's assessment of the initial characterization, GE perfor.ned an additional characterization that was provided in a letter to NRC, " Low Density Fuel Pellets," June 25, 1996.  ;

The purpose of GE's characterization of the LDP problem were two fold:

[

i 93

e to define the physical characteristics of the LDPs such that these characteristics can be included in conservative bounding analyses that demonstrate that the thermal-mechanical specified acceptable fuel design limits (SAFDLs) are not exceeded with the introduction of the LDPs in operating reactors, and to identify the root cause of the problem.

1he former will be discussed in this section while the root cause analysis will be discussed later in this report. The major purpose of this effort was to identify those characteristics of the LDPs that are different from the standard fabricated pellet population and that may impact the thermal-mechanical analyses. The major characteristics that can impact the thermal-mechanical analyses are pellet density and densification. This section will be subdivided into GE's characterization of the pellet density and densification behavior of the LDPs. The first part of both subsections will discuss GE's initial characterization effort and the background for the team's conclusion that additional characterization was necessary. The second part of both subsections will discuss the additional characterization performed by GE and the team's evaluation of this additional characterization.

b.l.1 Pellet Density One of the problems in characterizing the pellet density of the LDPs is that the low density pellets showed up only sporadically in the fabrication campaigns since the introduction of a new lubricant in late 1993 for pressing the powder into pellets. GE's preliminary estimate is that less than 0.3% of the ~400,000 rods produced between late 1993 and January of 1996 are affected (out of 179,376 scanning traces of individual rods examined, 515 rod scanning traces showed the signature of LDPs) and that less than 15% of the pellets within an affected rod are actually below the specification for pellet density and, therefore, are characterized as LDPs.

1 GE attempted to reproduce the LDP problem by taking a can of UO 2 powder with the new lubricant and pressing and sintering the powder into pellets without mixing the powder. The pellets frem

, this fabrication effort were a little low in density from the nominal density but were within GE specifications. GE had to add a significantly larger amount of lubricant than normally used before they could reproduce the LDPs. The team reviewed this effort and concluded that another parameter must be influencing the occurrence of the LDFs in addition to insufficient powder mixing. This additional parameter may be powder activity which is  ;

a measure of how easily the powder can be sintered. For example, a high activity powder sinters very quickly while a low activity l powder sinters more slowly. In the case of LDPs, a high activity l 94

l pswder will sinter before the lubricant can escape as a gas and

! the gas is trapped in the pellet creating the LDPs. With this

! scenario two factors must be present to create the LDPs:

l (1) insufficient powder mixing and (2) high activity powder.

In order to characterize the pellet density of LDPs, GE initially used only one tray of recently fabricated fuel pellets that exhibited pellets with low density characteristics (some pellets in the tray showed slight bulging at the ends). It shnuld be noted that there are approximately 2000 pellets in a tray and a tray makes 5 to 6 fuel rods. GE measured the pellet density distribution of this tray of suspect pellets using a gamma l densitometer to obtain a distribution of pellet densities in this l tray. GE also used the gamma densitometer to measure the density ;

distribution of one recently fabricated tray without low density pellets. The tray with low density pellets had approximately 13% l of the pellet population below the GE specification for pellet l density. l There is a problem with the gamma densitometer measurement of

pellet density in that the method has a significant bias in

! measured density (measured density is biased low) at low pellet densities and also the accuracy of the measurement is unknown at the low densities. Because of the suspected bias in the gamma densitometer measurement at low densities, GE has taken three low

, density pellets (less than 90% theoretical density (TD) based on i the gamma densitometer measurements) at different densities and i

performed a more accurate measurement of pellet density using the vacuum impregnation / water immersion measurement technique (this technique will be referred to as the vacuum impregnation

' measurement). The vacuum impregnation measurement technique is considered to be one of the more accurate techniques for measuring fuel densities to within typically 0.2% TD. These three vacuum l impregnation measurements were then plotted against the gamma densitometer measurements for these same pellets and GE used this plot to estimate the true densities of all the gamma densitometa-l measurements made on the LDPs from the one tray. Based on these measurements GE initially determined that the lowest gamma densitometer measurement of density from the tray examined was l approximately 87% TD and the true density to be approximately 91%

TD based on the calibration to the three vacuum impregnation measurements. Based on these initial measurements GE has estimated the effect of the local pellet density on SAFDLs using a density of 91% TD as noted below in Section 3.2.b.2 of this report.

The team questioned GE on whether pellt.t density measurements using the gamma densitometer and/or the more accurate vacuum impregnation technique were performed on pellets from other trays or fuel rods with suspected low density pellets. GE stated that there were a few LDPs from other sources but the large majority of the LDPs examined to date were from the one tray. The team noted l

l 95 l

that the low density pellet distribution is based on the ,

exaoination of only one tray out of over 30,000 to 40,000 trays i fabricated during the time of the suspected LDP problem and .

I several hundred of these trays may have LDPs. Therefore, it is very difficult to determine if the one tray examined to date and its associated density distribution is representative of the trays with LDPs fabricated since late 1993.

Also, the team's examination of the three LDPs that were measured using the more accurate vacuum impregnation measurement of densitj indicates that the gamma densitometer measurement may have an  ;

uncertainty of 1% TD or greater but the accuracy of this method could not be determined for LDPs based on only the three GE measurements.

The team concluded that GE did not have an adequate estimate of the low density pellet distribution to be able to bound the lowest individual pellet densities fabricated since late 1993.

Therefore, the team requested that GE examine other LDPs from other fuel rod / trays fabr:cated during the time period since late 1993 and measure the densities using both the gamma densitometer and the vacuum impregnation techniques on at least 6 additional LDPs (the required number of vacuum impregnation measurements is much lower because they are only used to estimate the accuracy of the gamma densitometer measurement).

As a result of the team's request, GE examined the densities of additional LDPs from five other trays to (1) determine the accuracy of the gamma densitometer measurements by performing additional vacuum impregnation tests on the LDPs, and (2) verify that a density of 91% TD is a reasonable lower bound on the LDPs.

GE's further characterization of the LDPs is summarized in a letter to the NRC, " Low Density Fuel Pellets," June 25, 1996. The results of this further characterization demonstrates that the 2 sigma uncertainty in the gamma densitometer measurement at a density of 87 to 88% TD is approximately 1.3% TD. GE's further c;aracterization consisted of selecting 8 LDPs from 5 different trays. The purpose of GE selecting the LDPs was to find those pellets in the trays with the lowest densities and all selected pellets had to have actual densities below 94% TD. From the 40 LDPs selected, only one pellet had a density as low as the l west density found in their initial examination of all the pellets from one tray.

The team concluded that GE's further characterization of the LDP problem was adequate and demonstrated that GE's assumption of 91%

TD provides a reasonable lower bound of density for the LDPs for use in thermal-mechanical analyses.

96

f i b.1.2 Pellet Densification The team questioned GE on whether the pellet densification l

characteristics of the low density pellets had been examined in the initial GE characterization because, in general, lower density pellets have larger pellet densification than higher density i

pellets. In addition, larger pellet densification can adversely affect the fuel melting and LOCA analyses. The team also questioned GE on what values of densification were used as input to the GE LDP analyses of the SAFDLs. GE stated that they had performed pellet densification tests on the LDPs and the measured densification values for these pellets were within GE specifications for pellet densification. GE also stated that the densification values for standard pellets were used for the thermal-mechanical analyses of the LDP problem.

The team examined the initial densification data and found that only 8 measurements were made on the LDPs. The mean densification for these 8 LDPs was twice the amount of the mean densification of the GE standard density pellets.

The team also examined the GE specifications on pellet densification which states, "Mean Pellet Densification - The upper 95% confidence limit on the mean density increase for any pellet type in a project shall not exceed (GE densification limit is proprietary)" and for " Individual Pellet Densification - The upper 95/95 tolerance limit on the individual density increase in a project shall not exceed (GE densification limit is proprietary)."

The team was concerned that the upper 95% confidence limit on the mean density increase had been exceeded by the low density pellets contrary to the GE conclusions. Therefore, the team examined the GE definition of " pellet type" which is stated as "a population of pellets obtained from a single U02 powder source that has been processed under the same range of operational conditions and which exhibits a consistent and predictable densification behavior."

The team concluded that the LDPs were of a different pellet type ,

I than the standard density GE pellets because the LDPs were processed under different or non-standard conditions (i.e.,

insufficient mixing of the powder) and their densification behavior was inconsistent with the standard density pellets (i.e., ,

twice the densification of the standard pellets) which is a l statistically significant variation of the mean compared to the standard density pellets of this project. If the LDPs are considered to be of a different pellet type then the mean densification of the LDP type exceeds the upper 95% confidence limit specified by GE on mean density increase for a pellet type within a project. GE disagreed with the teams interpretat~on of the LDPs being a different pellet type because both the LDPs and standard density pellets were made during the same fabrication campaign for a given project.

l 4

97 ,

The team was less concerned with whether the densification of the LDPs exceeded the GE specifications because the LDPs already exceed GE specifications on density. The team was more concerned about whether the LDPs have been sufficiently characterized and the assumptions used in the LDP analyses sufficiently bound the LDP characteristics to provide assurance that the LDPs do not exceed thermal-mechanical SAFDLs.

The team concluded that therc is most likely a significant difference between the densification characteristics of the LDP's i and the standard density GE pellet but that there has not been adequate characterization of the LDPs population by GE to determine the densification characteristics of the LDPs.  !

The team requested that further densification tests were needed for the LDPs in order to determine appropriately bounding values l for input to the thermal-mechanical analyses. GE agreed to obtain '

further densification data on the LDPs by taking additional density and densification data on LDPs from several fuel rods or trays with LDPs.

As noted above, GE has selected 8 LDPs each from 5 different trays for a total of 40 additional LDPs. GE performed densification tests on each of these 40 LDPs and found the mean densification to be approximately 2.5 times the densification of the standard GE ,

pellets and the upper 95 percentile densification was also greater I than observed for standard GE pellets.

The team concluded that GE's further characterization of the densification properties of LDPs was adequate for use in evaluating the impact of LDPs on thermal-mechanical performance, b.2 Thermal-Mechanical Analyses GE's initial thermal-mechanical analyses presented in a letter to NRC, "GE Fuel Pellet Density Report," April 15, 1996, made some assumptions about the pellet density and densification properties of the LDPs based on their characterization of the LDP problem. These assumptions were used by GE to perform thermal-mechanical analyses to demonstrate that the LDP problem does not exceed their SAFDLs for normal operation, anticipated operational occurrences (A00s), and accidents.

l l

GE identified six analyses and SAFDLs that may be influenced by changes in the fuel pellet density and these are: (1) fuel melting, (2) fuel  !

rod internal pressure, (3) cladding strain, (4) thermal-hydraulic I performance, (5) rod drop accident, and (6) loss of coolant accidents l (LOCAs). These analyses are either dependent on local fuel properties l (local axial location in the rod) or rod-average fuel properties. This was important for modeling the GE LDP problem because it is a localized problem in a fuel rod where less than 15% of the pellets in an affected l rod are LDPs. Those analyses that are dependent on local fuel l properties ar e fuel melting, cladding strain, and LOCA. Those analyses 98

that are d:p:nd:nt en avarage fuel propertics are fuel rod internal pressures, thermal-hydraulic performance, and rod drop accident. As a result of the team's evaluation of the analyses in GE's April 15, 1996, letter to NRC, GE has revised the three analyses that are dependent on local LDP properties based on further GE characterization of the densification properties of the LDPs and these revised analyses were summarized in GE's June 25, 1996, letter to NRC.

For those analyses that are dependent on local fuel properties GE assumed that the minimum pellet density observed in their initial ,

characterization of the LDP was the density of the peak power pellet,  ;

and the average pellet density of the rod is assumed to be the lowest  ;

permitted by GE fuel weight tolerances. For those analyses that are dependent on rod average fuel properties the average pellet density is assumed to be the lowest permitted by GE fuel weight tolerances. GE correspondingly assumed a lower local linear heat generation rate (LHGR) and rod-average LHGR due to the lower fissile content resulting from the I lower local and average fuel densities. The remaining input, other than i local and rod-average pellet density and local LHGR, for these analyses i remained the same as those previously approved in GESTR-Mechanical I (NEDE-240ll-P) and GESTR-LOCA (NEDE-23785-1-P-A).

The team concluded that the GE assumption of average fuel rod density l being the lowest permitted by the fuel weight specifications is l conservative. However, as noted above in the Section 3.2.b.1 of this report, the team concluded that GE did not initially characterize the LDPs adequately to conclude that the lowest density observed and assumed in the local fuel analyses of the peak power pellet bounded the population of LDPs fabricated by GE. The characterization was '

inadequate because GE only selected one small population of LDPs to characterize the problem rather than sampling from the broader population of LDPs fabric'ated. As a result, GE performed additional characterization of the LDPs that demonstrated that the 91% TD assumed by GE is a raasonable estimate of the lower bound density of the LDPs.

The team also questioned GE on the assumptions used for pellet densification of the LDPs at the peak power pellet location because, in general, lower fuel densities result in higher pellet densification. GE stated that for the initial analysis of the LDP problem GE used the same densification as for their standard fuel pellets. As noted in the Section 3.2.b.1 of this report, the team examined the small amount of GE densification data on the LDPs and found the nominal densification to be twice that of GE standard pellets. The increased pellet densification of the LDPs is of concern because this can increase local fuel temperatures. The team concluded that GE initial characterization efforts had not adequately characterized the pellet densification of the LDPs. The team requested further characterization of the densification properties of the LDPs and that these properties be included in those thermal-mechanical analyses that are dependent on local fuel properties, such as fuel melting, cladding strain and LOCA analyses, in order to demnstrate that the respective SAFDLs are not exceeded.

99 i

i

As previously stated, GE performed the additional characterization of the LDPs including the densification properties of the LDPs. These properties were used by GE in their revised analyses of fuel melting, cladding strain, and LOCA, as presented in GE's letter to NRC dated June l

25, 1996.

b.3 GE Analyses GE issued an initial report in its letter to NRC dated April 15, 1996, that summarized analyses that assessed the impact of the LDP problem on the following six fuel performance SAFDLs: (1) fuel melting, (2) fuel rod internal pressure, (3) cladding strain, (4) thermal-hydraulic performance, (5) rod drop accident, and (6) LOCA identified in Section 4.2 of the NRC SRP (NUREG-0800). GE performed revised analyses of fuel melting, cladding strain, and LOCA as summarized in its June 25, 1996, letter to the NRC. GE concluded that the other SAFDLs identified in Section 4.2 of the NRC SRP such as cladding collapse, cladding fretting, axial growth, corrosion /hydriding, rod bowing, and seismic loadings are not affected by the LDP problem. The team concurs that only the identified six SAFDLs are affected by the LDP problem.

It should be noted that cladding collapse might be expected to be impacted by the presence of LDPs because pellet densification was one of the contributing factors for the observed cladding collapse in PWR rods in the early 1970s. However, the GE cladding collapse analysis is not dependent on pellet densification because GE conservatively assumes the axial gap in the fuel column is of sufficient size to allow the cladding to collapse into the gap, i.e., the fuelTherefore, pellet column offers no support an increase in pellet to the cladding to prevent collapse.

densification and subsequent axial fuel column gap size does not affect the cladding collapse calculation.

A summary of the initial GE analyses of the identified six SAFDLs from its April 15, and June 25, 1996 letters to NRC plus the team's evaluation of both analyses is provided in the following.

(1) Fuel Melting GE presented a summary of the initial results in its April 15, 1996, letter to NRC of their GESTR-Mechanical analyses for the GE88/GE9B/GE10/ Gell /GE13 fuel rod designs to evaluate the The GE potential for fuel centerline melting during an A00 event.

analyses used standard design and conservative licensing assumptions with the exceptions that (1) the fuel average density was assumed to be the lowest possible permitted by fuel weight tolerances on the rod, and (2) the peak power pellet was assumed to have the minimum pellet density found by GE in the characterization of the LDP problem. The lower fuel densities assumed in this analysis result in lower fuel thermal conductivity and higher fuel temperatures and finion gas release (FGR) in the 100

fuel rod on average and also higher temperatures in the peak power pellet and greater chance for fuel melting. The GE analyses l concluded that fuel melting will not occur during the bounding A00 l at the worst time in the fuel lifetime.

{

The team reviewed the analyses presented in GE's April 15, 1996, I letter to NRC and concluded that the input of pellet density and l

densification on the local hyh power pellet may not bound the  ;

LDPs because the LDPs had not been adequately characterized to I assure that the input was bounding. For example, the GE input for pellet densification of the high power pellet was assumed to be equal to a standard pellet. The few available GE densification l data on the LDPs demonstrated that they had twice the i densification on average as the GE standard pellet. An increase in densification of the peak pellet will increase the fuel temperatures of this peak pellet. Therefore, the team could not conclude that the GE fuel melting analyses demonstrated that fuel melting would not occur for fuel rods with the LDP problem.

l As a result GE has performed revised analyses (GE's June 25, 1996, j letter to NRC) using actual measured densification properties of l the LDPs and verified that the LDP density input of 91% TD is t

l indeed bounding based on further characterization. These revised

! analyses demonstrate that fuel melting will not occur for LDPs for the bounding A00 at the worst time in the fuel lifetime. The team concluded that GE's revised analyses was acceptable. j (2) Fuel Rod Internal Pressure GE has performed GESTR-Mechanical analyses for the GE88/GE9B/GE10/ Gell /GE13 fuel rod designs to evaluate the potential for fuel rod pressure to exceed the rod internal ,

pressure SAFDL. The GE analyses assume standard design and '

l conservative licensing assumptions with the exception that the

! fuel rod average density was assumed to be the lowest permitted by fuel rod weight tolerances. Fuel rod FGR is an integrated effect that takes place over the whole fuel length and, therefore, the GE assumption of lowest average pellet density based on the lowest fuel weight tolerance is conservative. GE also looked at the j sensitivity of having lower density and standard density zones at i

different locations within the fuel rod and obtained nearly identical'results as those with the assumption of the same average smeared density over the whole fuel length. The GE analyses demonstrated that fuel rods with LDPs will not exceed the rod internal pressure design ratio of 1.0 with at least 95% confidence l for a rod operating at the bounding envelope of power versus l exposure.

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l The team reviewed these analyses and concluded that even though the GE analysis results are approaching the rod pressure limit, there is adequate conservatism in the rod internal pressure analyses. Therefore, the GE LDPs meet the SAFDL for rod internal pressures.

(3) Cladding Strain GE presented the initial results in its April 15, 1996, letter to NRC of the GESTR-Mechanical analyses for the GE88/GE98/GE10/ Gell /GE13 fuel designs that evaluate the effect of LDPs on pellet / cladding interaction (PCI) and resulting cladding strain during an A00 event. This GE analysis used standard design and conservative licensing assumptions with the exceptions that, as with the fuel melting analyses, (1) the fuel average density is assumed to be the lowest possible permitted by fuel weight tolerances on the rod, and (2) the peak power pellet is assumed to have the minimum pellet density found by GE in the characterization of the LDP problem. The lower fuel densities assumed in this analysis results in lower fuel thermal conductivity and higher FGR, that results in higher cladding strains.

GE also examined the possibility of the LDPs having different fuel relocation than the standard pellets. If the LDPs experienced more cracking than the standard pellets it could be hypothesized that the LDPs could possibly result in more PCI than resulting from the standard pellets. As a result GE tried to induce thermal cracking in the LDPs by heating the LDPs and rapidly cooling them.

At fuel pellet temperatures typical of normal operation GE experienced difficulty in cracking the LDPs, however, pellet fracturing was observed when the temperatures were increased. GE also performed similar thermal cracking tests on GE standard density pellets and compared the cracking results of the sta.;dard and low density pellets. The results showed no difference in the cracking behavior between the standard and low density pellets.

From this GE concluded that pellet cracking behavior of the LDPs was similar to the standard pellets and relocation should be similar for these two pellet types. Therefore, the GE analyses assumed that the LDP fuel relocation was the same as the GE standard pellets.

The initial GE analyses demonstrated that cladding strain in fuel rods with LDPs remained below the 1.0% strain limit as required by Section 4.2 of the SRP for the bounding A00s at the worst time in the fuel lifetime.

The NRC team reviewed the initial GE cladding strain analyses and concluded that the input of pellet density and densification on the local high power pellet may not bound the LDPs because the LDPs had not been adequately characterized to assure that the input was bounding. In addition, the GE analyses showed that they 102

L were very close~to the 1.0% strain, limit.and changes to either

! pellet density or densification could change the results of the analysis. Therefore, the team concluded that the GE fuel melting analyses did not demonstrate that cladding strainLwould not exceed the 1.0% strain SAFDL for fuel rods with.the LDP problem.

- As a result,.GE performed revised. analyses.of cladding strain (June 25, 1996, letter to NRC) using. actual measured densification properties of the LDPs and verified that the LDP density input of 91% TD is indeed bounding based on further characterization.

These revised analyses demonstrate that cladding strain will not exceed the 1.0% strain SAFDL.for fuel rods with LDPs for the bounding A00 at the worst time in the fuel lifetime. The team concluded that GE's revised analyses was acceptable.

(4) Thermal-Hydraulic Performance  :

GE did not perform specific analyses of the impact of the LDPs on thermal-hydraulic performance but offered an argument in its April ,

15, 1996, letter to NRC as-to why the LDPs will not adversely affect this SAFDL, that argument is-summarized below. - Thermal-hydraulic performance is a function of the thermal time constant of the fuel rod and can be divided into two types of analyses (1) 4 a core wide analysis, and (2) a hot channel a M ysis. A decrease 1 in pellet density and an increase in pellet densification compared  !

to GE standard density pellets will increase the thermal time constant of the fuel rod.

For the core wide thermal hydraulic response, an increase in the thermal time constant decreases the thermal performance margin because it increases the magnitude of a rapid depressurization-type A00. However, because of the small number of fuel pellets )

and even smaller number of fuel rods affected by the LDP problem in a core, the decrease in core wide transient performance will most likely be imperceptible.

For the hot channel analysis, the increased thermal time constant delays the heat transfer to the coolant during an A00 and, therefore, increases the hot channel thermal performance margin.

However, the actual. magnitude.of this increased margin is also probably 'small (although the effect will be. larger than _for. the core) due to the small number of fuel pellets and rods with the

- LDP problem that would exist within an individual: assembly.

The team found GE's arguments acceptable for wny the LDP problem l has little impact on thermal performance.  ;

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(5). Control RSd Drop Accident GE did not perform specific analyses of the impact of the LDPs on the control rod drop accident but has offered an argument in lis April 15, 1996, letter to NRC as to why the LDPs will not-adversely affect this accident. The control rod drop accident is initiated from reduced power conditions (either cold or hot standby conditions) due, in part, to significanU.y higher control '

blade reactivity worth at these conditions. The reactivity insertion in the surrounding fuel-assemblies is then determined assuming adiabatic heat transfer conditions. The resultant fuel enthalpy' values are then compared to a SAFDL of 280 cal /gm that assures core coolability. The margin of acceptability for this l SAFDL is very large and independent of the thermal time constant because adiabatic conditions are assumed.

For the radiological consequences of this accident, the failure threshold of the fuel rods is a function of two limits (1) a-reactivity limit of 170 cal /gm and (2) a thermal performance limit of boiling transition. These two limits are used to determine the number of failed fuel rods. The reactivity of the fuel rod is calculated assuming adiabatic conditions and is not-impacted by the thermal time constant. Therefore, the 170 cal /gm limit is not effected by the LDP problem. As discussed in thermal hydraulic performance above, the thermal performance margin to boiling transition increases with an increase in the thermal time constant for the LDPs. Therefore, the radiological consequences of the control rod drop accident will most likely result in a small improvement with the presence of LDPs.

The team found GE's arguments acceptable for why the LDP problem has either no, or improved performance, with respect to the control rod drop accident.

(6) Loss-of-Coolant Accident GE performed specific analyses of the impact of the LDPs on fuel stored energy for their different fuel designs and argued that the results of LOCA sensitivity analyses to fuel stored energy (NEDE-30996P-A) demonstrate that the increased stored energy due to the  !

LDPs will not adversely affect LOCA peak cladding temperatures I (PCTs). A decrease in pellet density and increase in pellet densification due to local LDPs increase local fuel temperatures and, therefore, fuel stored energy at the onset of the LOCA.

However, for all GE BWRs (with exception of BWR/2s), the maximum PCTs occur during blowdown rather than at the initial loss of primary coolant stage of the accident. As a result, GE sensitivity analyses have demonstrated that the increase in stored energy of the peak power pellet due to the minimum observed local l

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density of the LDPs is more than offset by the decrease in decay heat caused by the lower fissile content of the LDPs. Therefore, GE states that no increase in PCTs-is expected for LOCA in GE plants due to the LDP problem.

For BWR/2s, the LOCA performance is more sensitive to the number and timing of fuel rod failures due to fuel rod internal pressures and initial PCTs, e.g., a reduction in the allowable maximum average planar linear heat generation rate (MAPLHGR) would be necessary if additional fuel rod failures resulted due to l increased rod pressures and higher PCTs during the initial heatup l stage of the LOCA. GE assumed for the BWR/2 analysis that the fuel rod average density.was equal to the lowest density allowed  !

by the GE fuel rod tolerance on fuel rod weight. This assumption maximizes fuel rod internal pressures but did not maximize the l initial PCTs.during heatup which is dependent on local pellet density and densification. The GE result of the BWR/2 LOCA analysis for the LDP problem is that the increase in. fuel rod pressure is not sufficient to significantly alter the onset of fuel' rod perforations and, therefore, does not alter the results l of the LOCA.

The team reviewed the GE analyses of stored energy in its April 15, 1996, letter to NRC for all GE plants and fuel designs with respect to LOCA and concluded that the input of pellet density and l densification on the local high power pellet may not bound the LDPs because the LDPs.have not been adequately characterized to assure that the input was bounding and, therefore, the calculated stored energy may not be bounding. >

As a result, GE performed revised stored energy analyses using actual measured densification properties of the LDPs as input, and summarized the results in its June 25, 1996 letter to NRC. These revised results shaw that stored energy increases for LDPs but the increase for all GE plants is still not sufficient to significantly increase LOCA PCTs based on the sensitivity analyses ,

in NEDE-30996P-A.- For all plants except BWR/2s the stored energy increase from the revised analyses is still offset by the reduced decay heat. For BWR/2s, the increase in rod pressures from the revised analyses still does not significantly al'er the onset of

' fuel rod perforations. GE was questioned if the increased stored energy from the LDPs in BWR/2s would increase PCTs and, therefore, alter the onset of fuel rod perforations. GE responded that the sensitivity analyses in NEDE-30996P-A demonstrate that the increased stored energy from LDPs has little impact on initial PCTs and, therefore, will not alter the onset of perforations.

The team found GE's revised stored energy analyses in the June 25, ,

1996, letter to NRC acceptable.

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b.4 Powd;r 81cnding The team observed GE's uranium powder process through and including the blending of uranium powder. Blending is the key to forming a uniform mixture of the powder. The team examined the process qualification record when " mixing lubricant" [ revised pursuant to 10 CFR 2.790 -  ;

document described a specific trade name) was first added to the process. Properly blended and mixed with the uranium oxide powder, the

" mixing lubricant" (revised pursuant to 10 CFR 2.790 - document described a specific trade name) enhu ced lubrication of the pellet press machine resulting in fuel pellets with a rre uniform density.

One of the acceptance criteria for the qualificition batch was a specified time (minutes) for the blending operation (mixture of " mixing lubricant" Irevised pursuant to 10 CFR 2.790 - document described a .

specific trade name) and uranium powder). The team concluded that the

" mixing lubricant" (revised pursuant to 10 CFR 2.790 - document described a specific trade name) blending process had been adequately qualified.

The team observed " mixing lubricant" [ revised pursuant to 10 CFR 2.790 -

document described a specific trade name] being weighed on a scale and determined that during the weighing process, the scale's pointer fluctuated to the extent that it was difficult to read the actual weight. GE informed the team that because the problem was verbally identified, rather than in a written format, it did not take corrective actions. The team concluded that for this instance, GE's actions lacked the proper attention to adequately perform problem identification and j resolution for a parameter that directly related to the quality of pellets. The team considered this a weakness in the pellet process.

The team found that the procedures for powder production did not specify the tumbling time to meet the qualification batch value. Further, there were no checklists or quality hold points to verify blending. The team considered this a procedural weakness. However, the team noted that the procedures and checklists were being revised. In addition, GE was adding computer controls to the blending process, b.5 Pelleting i

The team, during discussions with plant personnel, determined that operators of the pellet presses document their comments on a press production log which includes the following information: the date of operation, the shift, identification of the operator, the boat (pan used to hold pellets during the sintering process) card number, the identification of the can with fuel powder, the tare and gross weights of the boat, the net weight, the green density, the number of pellets pressed per minute, the press number, and the enrichment (in percent). ,

The team reviewed the press production logs generated during the l November 1995 to February 1996 period and found the following comments repeated:

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o nine instances of fluffy powder o seven instances of pellet density out of range e six instances of lumpy powder l

• five instances of bad powder

• four instances of weak powder

• three instances of not tumbling the cans to blend the acrawax with uranium oxide powder a two instances of pellets not holding together e one instance of the powder being like wet flour.

Additionally, on December 22, 1995, an operator recorded that powder can AUG 2106 had not been tumbled. On January 9, 1996, the press operator logged that cans AUG 2183 and 2184 were not tumbled resulting in chipped pellets. On January 13, 1996, a fuel pellet loader observed that the i

l pellets did not hold together because their ends were bloated (convex).

l

! On the basis of its review of the press operators log, the team noted several precursors that may have resulted in LDPs.

b.6 Sampling and Testing The team reviewed the adequacy of the methodology to verify pellet integrity on the basis of testing pellet samples for specific l characteristics during the manufacturing process. In F-S-179, " Quality

) Notice," Revision 4, October 16, 1995, GE establined a rationale for '

powder / pellet sampling. The quality notice prescribed the various  :

characteristics that are to be verified, the frequency at which they are to be verified, and the rationale for sampling. The team found that powder and pellets are sampled at the following stages of production.

1 (1) Preblended Powder The enrichment of Ammonia Diuranate (ADU) in the preblended powder is monitored by taking a sample from (deleted pursuant to 10 CFR 2.790 - document described a specific value] cans of powder to calculate the accountability factors and to obtain the dat' for blend feed. ,

(2) Blended Powder The estimation of impurities in the blend is monitored by taking a l sample [ deleted pursuant to 10 CFR 2.790 - document described a  !

l specific value] blend and determining the acceptability of the l l population and assurance of moderation control.  !

1 (3) U0, Pellets Pellet density in U02 pellets is monitored by testing [ deleted pursuant to 10 CFR 2.790 - document described a specific value] to determine if the product / process is performing to the applicable specification.

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[ deleted pursuant to 10 CFR 2.750 - document described a specific value) pellets were selected from each pellet boat after it completed the sintering stage, for initial density testing.

Before the LDP problem, (deleted pursuant to 10 CFR 2.790 -

document described a specific value) pellets were selected from the bottom of the boat and [ deleted pursuant to 10 CFR 2.790 -

document described a specif9c value) pellets from the top of the boat.

As part of its corrective actions to address the LDP problem, GE now selects [ deleted pursuant to 10 CFR 2.790 - document described i a specific value) pellets from the bottom of the boat, and one l pellet from the top of the boat.

GE was not able to provide the inspection team with a statistical bases l for the adequacy of test sample selection. The lack of a statistical basis of the pellet sampling plan was considered by the team to be a weakness in GE's pellet processing.

b.7 Pellet Integrity The team reviewed the various tests used by GE to determine the following characteristics of pellets which are measured during the manufacturing process: surface roughness, green density, sintered ,

density, dimensions, ^,hermal stability. The team identified no adverse '

findings in this area.

b.8 Quality Control The team reviewed the following documents used by QC inspectors to verify the integrity of the manufacturing process at various stages of production.

Quality Control Inspector Instructions (QCII) 2.2.1.1, " Acceptance and Release of Line Material In-House & Ship Powder Blends,"

Revision 9, dated February 14, 1996 QCII 3.2.2, " Pellet Sampling and Release," Revision 32, dated November 6, 1995 QCII 6.2.5.1, "(Gd, U)0, Pellet Sampling and Release," Revision 23, dated November 6, 1995

• Quality Control Operator Requirements, 2.1.1.46, " Sampling requirements on Blends," Revision 3, dated January 30, 1986 The team observed that QC implemented the above QCIIs and maintained statistics of the results to demonstrate to that there were no adverse trends. However, the team observed that there were no requirements for QC to independently examine work areas where problems are likely to occur and document the findings.

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For instance, QC did not attempt to evaluate the problems experienced by the fuel pellet pressers and the fuel pellet loaders which were ,

documented in the operator logs. The team concluded that if QC had brought the problems documented in the operator logs to the attention of the appropriate production engineers, they may have evaluated them to determine whether they were contributors to the LDP problem; if so, they may have been identified before LDPs were shipped to licensees. The team also observed that in some instances where problems were l identified, QC relied on statistical sampling to characterize or bound the problem without documenting an adequate basis to ensure the problem was bounded. The team considered these observations as weaknesses in i GE's pelleting process.

The team observed a single pellet from another fuel tray mixed with natural uranium pellet on a pellet tray prepared for rod loading. The pellet dimensions were different from the remaining pellets on the tray.

When the team disclosed this observation during the inspection, GE investigated this discrepancy and determined that this pellet was from a tray that was used to manufacture fuel over six months before. GE hypothesized that this pellet was dislodged and got mixed with the i pellets of the other tray. l The pellet had not been loaded into a fuel rod. GE stated that even though the operator had missed detecting the oversized pellet during the visual check, the scanning device, which monitored the dimensions of the pellets, would have rejected the pellet because it was oversized, and thus prevented its entry into a fuel rod. However, the team was concerned about the inadequate controls at the pellet and tray activities, which made such pellet mixup possible, and GE's belated i detection of the incorrect size pellet.

b.9 Corrective Actions Taken The team reviewed actions taken to correct the problem identified in Corrective Action Request (CAR) 0039. The CAR was issued on March 19, 1996, to identify that a specific tray contained pellets that exceeded the specification requirements relative to density and voids. According to GE the largest contributor to the LDP problem was the possibility that the containers were not tumbled. The corrective action recommended in CAR 0039 was to sensitize the operators involved in the importance of the process of tumbling the fuel powder and " mixing lubricant" (revised pursuant to 10 CFR 2.790 - document described a specific trade name]

mix. The team reviewed training documents which listed thc Temporary Operating Instruction (TOI) and the signatures of persons trained. The review indicated that individuals were trained on TOI B-3483 issued on March 1, 1996, " Gathering Information Related to Bloated Pellets," as acknowledged by the signatures of the trainees. The team identified no adverse findings in this area.

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b.10 Root Cause Analysis The team evaluated GE's root cause analysis. It showed that on February 22, 1996, GE first discovered the LDPs when an operator noticed two adjacent fuel pellets with slightly convex end faces. 'The team concluded that the analysis failed to consider the press operators previous comments on the quality of the blend, and investigate their affect on the problem.

On January 9,1996, for instance, the press operator on shift Z ,

documented that cans identified as AUG 2183 and 2184 had not been  ;

tumbled. Four days later on January 13, 1996, a fuel rod loading station operator attempted to pick up a group of pellets by squeezing )

its ends. The group of pellets did not hold together and fell apart. )

The operator observed that the pellets did not hold together because i their ends were convex (same indication was observed later on February 22, 1996, and determined by GE-to be characteristic of LDPs). A "Q Hold" tag was issued for the tray because it contained pellets with i raised end surfaces. A nonconformance was written to identify that the i end surfaces of the pellets were raised and not concave. The pellet I dimensions were verified to be acceptable, and the nonconformance was closed without examining the densitim of_the pellets. The nonconformance dispositioned the pe'.iets as "use as-is."

In another instance, on December 22, 1995, a pellet press operator documented that a can identified as AUG 2106 had not been tumbled. The

! team concluded that if GE had investigated these comments it may have determined whether they were precursors that contributed to the LDPs.

GE's root cause team did not evaluate the pellet press operator's complaints to determine if the pellets produced from those mixtures had low densities. Therefore, the team concluded that the GE's root cause  !

analysis was not adequate in identifying when the LDPs first occurred or i evaluating all contributing factors.

c. Conclusions The team evaluated the GE's safety analyses and corrective action for j l LDPs. Based on the team's evaluation and the additional >

characterization information, the team concluded that (1) there were no '

additional significant safety concerns, (2) the consequences of postulated accidents were not underestimated, (3) the control rod insertability was not affected, and, therefore (4) the LDPs problem was i satisfactorily resolved.

l The team's evaluation showed certain weaknesses in the GE's pelleting l process. Principal contributing factors to these weaknesses included apparent deficiencies in development of procedures, managing problem identification and corrective actions, and QA involvement.

Specifically, the team found weaknesses in (1) determining the weight of the mixture because the pointer on the blending mix weighing mamine fluctuates, (2) the lack of a statistical bases for selecting the quantity of samples, (3) the adequacy of corrective action taken when

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cans of powder had not been tumbled and GE did not examine the density  ;

of the pellets manufactured from those cans, (4) the instructions to )

i QA/QC to resolve verbal complaints, and (5) the root cause analysis did  ;

not identifying when the first LDPs occurred or evaluate all '

contributing factors.

3.3 10 CFR Part 21 Procram I

l a. Inspection Scopa  ;

I l In an August 1995 document titled " Discussion of Deviation from Design l Procedure" (observations made during the SLMCPR analysis for the Kernkraftwerk Krommel plant) GE determined that within a fuel type for a l particular fuel design it was possible to have an R-factor distribution  ;

i that would yield a higher number of fuel rods that would approach l l boiling transition than the bounding values used in its generic SLMCPR i

! analyses.  !

l However, as described in Section 3.1.b.4 of this report, GE claimed that its first indication that the Gell SLMCPR analysis was not bounding l occurred at the er.d of February 1996 as the result of completing a cycle-specific analyses for River Bend Cycle 7. The River Bend Cycle 7 analysis also resulted in cycle-specific SLMCPR values that exceed the generic Gell SLMCPR and show that GE's generic analysis was not bounding as assumed by GE.

Also in August 1995, an NRC inspection team raised several concerns j related to areas where GE had made significant changes to the  !

i methodology used to determine the SLMCPR. With regard to the previous  !

team's concerns, during this inspection the team was told by GE staff that no action had been taken to address those concerns.

Therefore, the team reviewed the chronology of GE's SLMCPR, as well the LDP issue, to evaluate GE's responsiveness to these issues and evaluated GE's notices sent to GE fueled BWR licensees to evaluate the accuracy and effectiveness of the notices,

b. Observations and Findinas b.1 Safety Limit MCPR A GE internal letter dated March 15, 1996, initiated potential safety concern (PSC) 9608 to review the safety limit MCPR issue. A follow-up GE internal letter requested that the PSC be evaluated to determine if it represents a potentially reportable condition (PRC) and that results be documented in a design record file (DRF). Corrective actions were to be identified and justification for PSC closure was to be provided within 10 working days or it would automatically become a PRC (i.e., an evaluation under 10 CFR Part 21 requirements would be necessary).

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On March 27, 1996, the NRC was advised, by a telephone call from GE, that the SLMCPR concern was under internal review as a PRC. The NRC was told that the generic Gell SLMCPR of 1.07 was not bounding for a plant (River Bend) operating in a 2-year cycle with a large batch fraction (37%) of the fresh Gell fuel. GE claimed that the error was attributed to an analytical procedure deficiency and stated that it was in the process of evaluating all plants as rapidly as possible to assure that the plant-specific SLMCPR was bounded by the generic SLMCPR used as the licensing basis. (The telephone call is documented in a March 29, 1996,  ;

letter from GE Fuel and facility Licensing Manager to NRC.) l On March 28, 1996, GE issued PRC 9604 to perform the 10 CFR Part 21 reportability evaluation of the SLMCPR issue.

On April 17, 1996, GE met with the NRC staff (several licensees were also represented at the meeting) for an update on the status of its SLMCPR review. At this meeting, the NRC staff was informed that the erroneous safety limits were due to non-conservatism in the generic '

analysis methods and may not be restricted to Gell fuel and large batch ,

fractions. GE concluded that ycle-specific analyses of each reload core was required to confirm that the generic SLMCPR value remained bounding. GE indicated that they had performed preliminary cycle-specific analyses and based on the results, had recommended corrective actions to 13 GE fueled BWR plants most likely to be affected by the  :

cycle-specific analyses.  !

During this meeting, the NRC staff agreed that administrative control of  ;

the operating limit MCPR, as recommended by GE, was an appropriate i response to the safety issue until correct safety limits could be verified and technical specifications and core operating limits reports (COLRs) could be revised to reflect the correct SLMCPR values. GE also i proposed long term actions to recover the lost thermal margin, but the NRC staff indicated these actions were relatively low priority.

In its April 22, 1996, letter, RJR-96-044, to NRC, GE documented its perspective of the April 17, 1996, meeting and transmitted the proprietary version of material presented at the meeting. Also included >

was a list of the 13 plants that had been identified, based on preliminary analyses of 22 plants, for priority attention in the reevaluation of the SLMCPR for current operating cycles. The letter listed GE's short-term action items as follow:

a perform SLMCPR evaluation of all plants with GE fuel as soon as possible a inform affected plants immediately of higher SLMCPR analysis results even if the value is preliminary

• licensees to immediately establish administrative controls to conform to a higher value of SLMCPR (even when unverified) and to implement changes in the process computer data upon receipt of a t verified higher SLMCPR value j

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• licensees to inform the NRC when administrative controls are implemented During this inspection, the team reviewed GE's customer correspondence on the SLMCPR issue. GE sent its customers a white paper entitled,

" Safety Limit MCPR Calculation," informing them of the SLMCPR error and of the April 17, 1996, meeting with NRC staff. These notifications, dated April 2, 3, and 4, 1996, differed only slightly, and no plant specific recommendations for action were provided. The customers were advised that no specific actions were believed to be necessary until plant specific evaluations are completed. At that time, GE believed the problem was limited to plants having large reload batch fractions of Gell fuel to support longer operating cycles.

The team concluded that these notifications did not address the recommended corrective actions that GE had claimed to have sent to 13 plants during the April 17, 1996, meeting with NRC staff.

On April 18, 1996, GE letters to 29 domestic plants transmitted the plant specific calculation results and recommendations for administrative actions that GE had claimed to have sent during the April 17, 1996, meeting with NRC staff. Of the 29 plants notified, 16 plants were told that they were at low risk for a non-conservative erroneous SLMCPR value. A summary of the team's findings from its review of GE's April 18, 1996, letter to selected plants follows:

a Clinton, classified as low risk, even though the E0C preliminary evaluation resulted in 1.082 value versus an existing SLMCPR value of 1.07. However, it was later determined that a verified calculation of 1.07 at E0C had been completed prior to the April 18, 1996, notification.

The status of the Clinton plant was further confused by the information provided in GE's May 24, 1996, letter to NRC, "10 CFR Part 21, Reportable Condition, Safety Limit MCPR Evaluations,"

which indicated that Clinton had an unverified SLMCPR value of 1.08 versus a technical specification value of 1.07. The NRC staff learned later during a telephone conversation with GE that verification of the Clinton MCPR was completed on May 29, 1996, and the existing technical specification value of 1.07 is correct.

However, the Clinton licensee did not provide 10 CFR 50.72(b) notification of any administrative control action while the unverified evaluation indicated that the technical spe ification was in error.

• Hatch unit 2, classified as low risk, was later informed (May 2, 1996, letter) that the calculated SLMCPR value was 1.08 versus a licensed value of 1.06 for its GE9 fuel and that administrative controls to protect the revised safety limit should be imposed.

On May 3,1996, the 1.08 value was verified and the Hatch unit 2 project manager was verbally notified. A May 8, 1996, letter from GE confirmed the verbal notification.

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Monticello, classified as low risk, was later informed (May 8, 1996, letter) that the verified SLMCPR value is 1.08 versus a licensed value of 1.07 for its Gell fuel and that administrative controls should be imposed to protect the revised safety limit.

Nine Mile Point unit 1, classified as low risk, was later determined to have a verified SLMCPR value of 1.10 for the Gell fuel in this BWR2 reactor.

Hope Creek was notified that the preliminary unverified value of SLMCPR was 1.09 (based on E0C calculations) versus a licensed value of 1.07 for its GE9 fuel. The licensee had already provided NRC with an April 16, 1996, event report of action to impose <

administrative controls for a 1.10 SLMCPR based on GE notification (believed to be verbal) on that date. Verified calculations for B0C and PHE were completed on May 7, 1996, and the maximum SLMCPR verified value was 1.08 at BOC. A May 8, 1996, letter to Hope Creek indicated that the final verified value was 1.08, even though a final calculation for the E0C had not been completed.

The team questioned the omission of the E0C calculation (which had been limiting in the preliminary assessment); GE personnel agreed that E0C should be evaluated but explained a logic which could support its omission. Further review of the Hope Creek DRF by the team showed that the E0C results could not be achieved oecause the limiting bundle could not be placed on the operating limit. At that time, additional calculations were planned at an exposure i life between PHE and E0C with sufficient rod worth remaining to I challenge the 80C maximum safety limit result.

Cooper, classified as low risk, was later notifud (April 25, 1996, letter) of an unverified SLMCPR value of 147 (tnc licensed value was 1.06), and then was reported to have a verified value of 1.05 (no impact) in GE's May 24, 1996, 10 CFR Part 21 notification letter to NRC.

In its May 24, 1996, letter "10 CFR Part 21, Reportable Condition, Safety' Limit MCPR Evaluations," GE informed NRC of the reportable condition of its SLMCPR evaluations. The 10 CFR Part 21 notification identified 12 plants with new SLMCPR limits above the currently licensed values. In addition to Hatch 2, Monticello, and Nine Mile Point 1 (original low risk plants), 9 of the 13 plants identified as high risk in the initial evaluation were found to have inadequate SLMCPR limits and license changes are needed. High risk plants that were found to have acceptable SLMCPR limits were Peach Bottom 2, Brunswick 1, Fermi 2, and Duane Arnold.

However, verified analyses had not.been completed for 6 plants reported in GE's May 24, 1996, letter, and as previously discussed, the existing SLMCPR was later (May 29,1996) verified for 1 of these plants (Cooper) that was affected by GE's preliminary analyses.

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Thus, from the time the potential design defect was identified (March 15,1096), approximately one month elapsed until notifications of needed interim corrective actions were issued by GE. The validated analyses were not completed until near the end of May, over two months after the PSC evaluation began. In the interim, one plant (Nine Mile Point 1) was not notified to take corrective action until May 21, 1996, even though the SLMCPR was 1.10 compared to a technical specification value of 1.07.

The final results of the GE's 10 CFR Part 21 evaluation showed nonconservative values of SLMCPR for 11 operating cores, 3 with GE9 fuel and 8 with Gell fuel. As described in Section 3.1.b.8 of this report, the team concluded that design defects resulting in nonconservative values of SLMCPR in 11 currently operating cores was due to GE's failure to recalculate or reconfirm the applicability of the generic SLMCPR as required by the NRC approved methods in GESTAR. In addition, GE's l failure to comply with analysis procedures referenced in plant technical l specifications appears to place these licensees in nonconformance with l

their technical specifications.

b.2 Low Density Pellets During its review of PSC 9606, " Low Density Fuel Pellets," April 4, 1996, and GE letter RJR 96-041 to NRC, "GE Fuel Pellet Density Report,"

April 15, 1996, the team raised questions concerning the adequacy of GE's characterization information used as an input to the fuel performance evaluation. The team also identified certain other weaknesses with GE's root cause evaluation. (See Section 3.2.b.1 and 3.2.b.10 of this report.)

Pursuant to the issues raised by the team, GE agreed to reopen this issue as a PRC to address the team's questions. In its letter RJR 96-070 to NRC, " Low Density Fuel Pellets," June 25, 1996, GE responded to the characterization questions raised by the team and assessed the impact to the original fuel performance evaluation or conclusions. i GE concluded that the revised characterization information was l consistent with its earlier information used as an input to the earlier fuel performance evaluation. ,

c. Conclusions c.1 Safety Limit MCPR According to GE, the SLMCPR problem was discovered about March 15, 1996,  ;

even though in August 1995, GE realized that it was possible to have an R-factor distribution that would yield a higher number of fuel rods that would approach boiling transition than the bounding values used in the generic analyses. Again in August 1995 [ year corrected by staff), an NRC inspection team raised several concerns related to changes to the methodology used to determine the SLMCPR, however, GE staff told the team that no action had been taken to address those concerns.

1 115

On the basis of its evaluation of the information that GE had in August 1995, the team concluded that GE could have taken actions to evaluate the SLMCPR problem earlier than its " discovery date" of March 28, 1996, when GE initiated its PRC evaluation.

The team found that from the time the potential design defect was identified (March 28,1996), approximately one month elapsed before GE notified licensees of needed interim corrective actions. GE's failure to evaluate its potential SLMCPR problem when indications of the problem first appeared in August 1995 and to timely notify licensees of the need to take interim corrective actions is considered a weakness in GE's responsiveness to the SLMCPR issue.

The team concluded that the results of detailed final analyses were sometimes inconsistent with results of the preliminary scoping analyses and that GE's understanding of the issue changed during the licensee notification process. Generally, only one sentence of a two page letter provided information on the cycle-specific core analyses, and for some plants, the status changed significantly with little explanation.

Consequently, sequential correspondence was sometimes conflicting and potentially confusing to licensee.

With regard to the GE assessment of the impact of the design error on affected plants, the team concluded that the notification correspondence was, in some instances, inaccurate and misleading in the characterization of NRC positions. For example, it was asserted that NRC had somehow relieved licensees of the need to evaluate impact of the SLMCPR error on previous operating cycles. Clearly, GE can not attribute an interpretation of the regulations to NRC without supporting NRC documentation. In another example, the notification correspondence from GE strongly implied that NRC is receptive to near term recovery of lost thermal margin by review and credit for claimed conservatism associated with power distribution, instrumentation, and direct gamma heating. In fact, the NRC indicated that submittals of this nature would receive low priority attention and was non-committal on the likelihood for approval.

The team concluded that the GE staffing level applied to the SLMCPR plant specific evaluations was inadequate to provide clear and accurate corrective information to its customers on a schedule appropriate for the safety issues involved. The team also concluded that the clarity and effectiveness of GE's notifications could have been enhanced if GE had placed less attention to deemphasizing the impact of the error on affected plants and more attention to timely completion of the plant evaluations and to reporting and explaining the plant specific results.

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

c.2 Low Density Pellets l Based on its evaluation of GE's additional characterization information for the low density pellet event, the team concluded that (1) there were t no additional significant safety concerns, (2) the consequences of

• postulated accidents were not underestimated, (3) the control rod insertability was not affected, and, therefore (4) the LDPs problem was satisfactorily resolved.

3.4 Entrance and Exit Meetinos In the entrance meeting on May 6, 1996, the NRC team met with members of GE management and staff, and discussed the scope of the inspcction. The team also reviewed its responsibilities for handling proprietary information, as well as those of GE. In addition the team established

contact persons within the management and staff of the applicable GE organizations.

l l During its exit meeting with GE management and staff, on May 10, 1996, l

the team discussed its findings and concerns, as well as GE's weaknesses, l

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{

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PARTIAL LIST OF PERSONS CONTACTED Armijo, J.S. General Manager, Nuclear Fuel Congdon, S.P. Manager, Design Process Improvement Currier, J.W. Manager, Customer Service Embley, J.L. Licensing Program Manager, Fuels & Facility Licensing Hauser, T.M. Manager, GENE, Quality Assurance Kipp, C.P. General Manager, Production Marlowe, M.0. Manager, Fuel Materials Programs, Nuclear Fuel McCaughey, D.A. Manager, Fuel Quality Potts, G.A. Manager, Operating Fuel Performance / Support Reda, R.T. Manager, Information Managerent Systems '

Sependa, W.J. Manager, Chemical Product Line '

Serell, D.C. Technical Program Manager, Nuclear Fuel Americas Sick, P.W. Manager, Nuclear Quality Assurance Smith, C.W. Sr. Licensing Engineer, Fuels & Facility Licensing Williams, R.D. Fuel Project Manager, Nuclear Fuel Americas ITEMS OPENED AND DISCUSSED Opened 99900003/96-01-01 NON Failure to recalculate or reconfirm generic SLMCPR ,

99900003/96-01-02 NON Failure to use NRC approved methods in GETAB '

99900003/96-01-03 NON Failure to verify R-factors Discussed

= GE letter MFN 074-96 to NRC, "10 CFR Part 21, Reportable Condition, l Safety Limit MCPR Evaluation," May 24, 1996 1

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! ACRONYMS USED l

[ A00 Anticipated Operational Occurrence l B0C Beginning Of Cycle l BWR Boiling-Water Reactor l

CAR Corrective Action Request CFR Code of Federal Regulation COLR Core Operating Limits Report CPR Critical Power Ratio A Delta (Differential) l DRF Design Record File l E0C End Of Cycle i FGR Fission Gas Release GDC General Design Criterion GE General Electric Company GE-NE General Electric Nuclear Energy l GESTAR General Electric Standard Application for Reload GESTR General Electric Stress and Thermal Analysis of fuel Rods GETAB General Electric BWR Thermal Analysis Basis LDPs Low Density Fuel Pellets LHGR Linear Heat Generation Rate LOCA Loss-of-Coolant Accident MAPLHGR Maximum Average Planar Linear Heat Generation Rate l MCPR Minimum Critical Power Ratio NEP Nuclear Energy Production NRC U.S. Nuclear Regulatory Commission PCI Pellet / Cladding Interaction PCT Peak Cladding Temperature ,

l PHE Peak Hot Excess  !

PRC Potentially Reportable Condition PSC Potential Safety Concern

QA Quality Assurance  ;

l QC Quality Control l l QCII Quality Control Inspector Instructions l

! SAFDLS Specified Acceptable Fuel Design Limits SLMCPR Safety Limit Minimum Critical Power Ratio i SRLR Supplemental Reload Licensing Report SRP Standard Review Plan

TD Theoretical Density l TOI Temporary Operating Instruction i 3D Three-Dimensional l

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pS MOg

% UNITED STATES

1. 3 NUCLEAR REGULATORY COMMISSION N WASHINGTON, D.C. 2055H21 f

August 28, 1996 Mr. Aron Seiken, President Nuclear Logistics, Inc.

l 7461 Airport Freeway Fort Worth, TX 76118

SUBJECT:

NRC INSPECTION REPORT 99901298/96-01 (AND NOTICE OF NONCONFORMANCE)

Dear Mr. Seiken:

l On July 11, 1996, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Fort Worth facility. The enclosed report presents the results of that inspection.

The NRC inspection team evaluated the program that Nuclear Logistics, Inc.

(NLI), established and executed to implement the provisions of Part 21 of ,

Title 10 of the Code of Federal Reaulations (10 CFR Part 21), reviewed selected portions of the NLI quality assurance program based on 10 CFR Part 50, Appendix B, that were used to control the dedication of commercial-grade purchased components for safety-related (Class IE) service in NRC-licensed nuclear power plants and that were used to control procedures used by your subcontractor, National Switchgear Services, for refurbishing switchgear used in safety-related applications. Within these areas, the inspection consisted of an examination of procedures and representative records, interviews with personnel and observations by the inspectors. '

The NRC inspectors determined that the implementation of your quality assurance program did not meet certain NRC requirements imposed on you by your customers. Specifically, you failed to establish adequate measures for the identification and control of low-voltage circuit breakers and failed to establish adequate measures to assure that purchased components conform to procurement documents as required by 10 CFR Part 50, Appendix B. These nonconformances are cited in the enclosed Notice of Nonconformance (NON) and 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.

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(

A. Seiken l In accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of l

this letter and its enclosures will be placed in the NRC's Public Document l Room (PDR).

Sincerely, r

1 )

Robert M. Gallo, Chief Special Inspection Branch I Division of Inspection and Support Programs Office of Nuclear Reactor Regulation Docket No. 99901298

Enclosures:

1. Notice of Nonconformance

2. Inspection Report 99901298/96-01 l

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NOTICE OF NONCONFORMANCE Nuclear Logistics Inc. Docket No.: 99901298 Fort Worth, Texas Report No.: 96-01 Based on the results of an inspection conducted on July 8 through 11, 1996, it appears that certain of your activities were not conducted in accordance with

NRC requirements.

1

1. Criterion VII of Appendix B tu 10 CFR Part 50, " Control of Purchased Material, Equipment and Services," requires, in part, that measures be established to assure that purchased material, equipment, and services, )

whether purchased directly or through contractors and subcontractors, i conform to the procurement documents.

Contrary to the above, verification plans for dedication of motor control center components for safety-related applications in the Vermont Yankee and Crystal River plants, VP 059-009 and VP 023-061 respectively, did not specify appropriate acceptance criteria, nor did associated test '

records contain adequate data, to verify that certain critical characteristics were met and that the associated components would perform all of their safety functions (99901298/96-01-01).

2. Criterion VIII of Appendix B to 10 CFR Part 50, " Identification and Control of Materials, Parts, and Components," requires, in part, that measures be established for the identification and control of materials, ,

parts, and components, including partially fabricated assemblies. These '

measures shall assure that identification of the item is maintained by heat number, part number, serial number, or other appropriate means, either on the item or on records traceable to the item, as required throughout the fabrication, erection, installation, and use of the item.

These identification and control measures shall be designed to prevent the use of incorrect or defective material, parts, and components.

NLI Quality Assurance Procedure NLI-QUAL-09, " Identification, Handling, Storage, and Shipping of Material," Revision 8, dated July 1996, section 2.1 states that it is NLI's responsibility to maintain traceability for items which are procured commercial grade and dedicated for safety-related use.

Contrary to the above, NLI had not maintained adequate quality assurance records to establish traceability to General Electric for a General Electric AKR-50 circuit breaker sold as new, safety-related to Entergy Operations, Incorporated (Entergy), for use at River Bend, to fill Entergy P0 95-G-72336 (99901298/96-01-02).

Please provide a written statement or explanation to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555, with a copy to the Chief, 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 l l

. Enclosure 1 I

122 1

include for each nInconformance: (1) a description of steps that have been or been or will be taken to correct these items; (2) a description of steps that have been taken or will be taken to prevent recurrence; and (3) the dates of j your corrective actions and preventive measures were or will be completed.

I l

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Dated at Rockville, Maryland this 28th day of August,1996.

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1 U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION Report No.: l 99901298/96-01 I

Organization: Nuclear Logistics, Inc. ,

7461 Airport Freeway l Fort Worth, Texas 76118 l

Contact:

Archie C. Ball, Quality Assurance Manager (817) 284-0077 i

i Nuclear Industry Servicing and refurbishing low- and medium-voltage  !

Activity: switchgear, third party dedication of commercial-grade ,

procured components. I l

Dates: July 8-11, 1996 i

4 Inspectors: Kamalakar R. Naidu, Senior Reactor Engineer Stephen D. Alexander, Reactor Engineer Billy Rogers, Reactor Engineer l

1 b

Approved by: Gregory C. Cwalina, Chief l Vendor Inspection Section Special Inspection Branch  ;

Division of Inspection and Support Programs '

Office of Nuclear Reactor Regulation Enclosure 2 2

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1 INSPECTION

SUMMARY

During this inspection, the NRC inspectors reviewed the implementation of selected portions of the Nuclear Logistics, Inc. (NLI), quality assurance (QA) program, and reviewed activities associated with the dedication of commercial-

grade procured items. The inspectors also reviewed procedures for refurbishing 4.16-kV General Electric (GE) circuit breakers.

The inspection bases were:

Appendix 8, " Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," to Part 50 of Title 10 of the Code of Federal l Reaulations (10 CFR Part 50, Appendix B) l 10 CFR Part 21, " Reporting of Defects and Noncompliance" A

3.1 minor violation of this report.of 10 CFR Part 21 was identified and is discussed in Section Two instances where NLI failed to conform to NRC QA requirements imposed upon l

them by NRC licensees were identified. These nonconformances are discussed in Sections 3.4 and 3.6 of this report.

l 2 STATUS OF PREVIOUS INSPECTION FINDINGS This was the first NRC inspection of NLI.

l 3 INSPECTION FINDINGS AND OTHER COMMENTS 3.1 10 CFR Part 21 Proaram The inspector evaluated the procedures adopted by NLI to implement the requirements of 10 CFR Part 21 by reviewing QA Procedures NLI-QUAL-06,

" Deviation Reporting," Revision 3, April 1994 and NLI-QUAL-08, i

"Nonconformances and 10 CFR 21 Reporting," Revision 1, April 1992. In NLI-QUAL-06, NLI's definition of " deviation" was different from s 21.3. This procedure does not require a review to determine if the deviation or a similar condition, not previously detected, affects any completed basic components, or ones that have been shipped. In NLI-QUAL-08, NLI used the term "nonconformances" to represent what Part 21 defines as deviations (failures to comply were not addressed). However, the procedure stated that " Material which does not meet the technical, quality or documentation requirements of the client P.0., but meets the plant design basis," was not considered as l nonconforming (and hence not requiring Part 21 evaluation). This statement is l does not conform to 621.21(a)(1) which requires that deviations be evaluated .

l to determine if they could create a substantial safety hazard or lead to  !

exceeding a technical specification safety limit, i.e., to determine if they !

are defects, because 621.3 defines deviation as a departure from technical l requirements included in a procurement document. l 1

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Section 6.1 of the NLI QA Manual, Revisicn 1, July 1991, appeared to be the principal procedure adopted pursuant to Part 21. The procedure used the terms

" potential defects" and "noncompliances" without defining them. It also uses the term "nonconformance" but in a sense not consistent with NLI-QUAL-C8. The procedores did not make it clear that deviations and failures to comply as defined in $21.3 must be evaluated per 21.21(a)(1) and just what deviations and failures to comply are. A preliminary evaluation is supposed to determine whether a safety evaluation is required, but does not ask the appropriate questions, such as, (1) Is the affected item a basic component?, (2) Is the problem a deviation (i.e., departure from technical procurement specification-excluded by NLI-QUAL-08) or does it fail to comply with the Atomic Energy Act of 1954, as amended, or any rule, regulation, order, or a Itcense of the NRC7, and (3) Has the item (or any similarly affected item) been shipped, such that the problem cannot be corrected?

The inspectors informed NLI that its 10 CFR Part 21 procedures were inconsistent with Part 21 and among themselves. The inspectors did not identify any instances in which NLI failed to comply with Part 21. In accordance with NRC enforcement policy as described in NUREG 1600, the Part 21 procedural deficiencies described in the paragraphs above are considered a minor violation of Part 21 and although tney are discussed here in the report, a Notice of Violation will not be issued. NLI agreed to revise the procedures to comply and conform to Part 21 requirements.

3.2 Deviation Reports The inspectors evaluated the implementation of NLI Quality Procedure NLI-QUAL-06, " Deviation Reporting," Revision 3, dated April 1994 The inspectors observed RR-042-0441-1 and -2 tags attached to 3-horse power (HP) motors manufactured by Baldor Company. During receipt inspection, a NLI technician noticed a discrepancy in the speed of motors specified in NLI purchase order 2344, Revision 1, which stated 1200 revolutions per minute (rpm), and the speed on the motor nameplate which stated 1140 rpm. NLI engineers had not formally evaluated the DR, although they pointed out that 1140 rpm was a typical slip speed for a 1200 rpm synchronous speed.

i The inspectors observed DR-023122-2, -4 and -5 tags attached to molded case Westinghouse type circuit breakers and reviewed documentation associated with them. NLI had not yet dispositioned these DRs. In general, NLI adequately implemented the procedure by initiating DRs and tagging items which deviate from the P0 requirements.

However, as discussed in paragraph 3.1, there appeared to be a problem with NLI's evaluations because NLI's deviations are typically referred to as nonconformances by most vendors and licensees, being conditions in which safety-related (basic) components fail to meet some specifications or other requirements, not necessarily those from procurement documents, and actions are taken to correct the discrepant condition (s) or to demonstrate that the material is acceptable. NLI has agreed to correct the inconsistency between definitions when they revise the procedures to conform to 10 CFR Part 21.

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3.3 Nonconformances Reports The inspectors evaluated NLI's implementation of its Procedure NLI-QUAL-08, Revision 1, "Nonconformances and 10 CFR Part 21 Reporting," dated April 14, 1992, by reviewing the evaluations and corrective actions taken to preclude recurrence in NLI nonconformance reports (NCRs).

The inspectors reviewed NCR-20 and Revision 1 to NCR-20, dated May 1, 1996, which were initiated after Salem personnel observed that lock washers installed in the 4.16-kV, GE, Magne-Blast circuit breaker arcchute assemblies were broken. NLI evaluated this problem and determined that contrary to the instructions in its Procedure NLI-TECH-P101, fasteners, including lock washers, were not replaced with dedicated hardware. The lockwashers had not been removed from the arcchute holddown assemblies before the assemblies were zinc plated. Also, three pawls were found to be chipped on the end; the pawls i engage the teeth of the crank wheel. The root cause was determined to be failure to follow procedures. Corrective action taken to prevent recurrence was to provide additional training for the NLI/NSS personnel in the requirements for replacing fasteners.

NCR-24 was initiated on June 4, 1996, to document that approximately eleven l hardened parts in a AM-4.16-250-9H GE Magne-Blast circuit breaker supplied to l the River Bend Station were zinc plated and that they were susceptible to hydrogen embrittlement. NLI recommended that River Bend return the breaker so l that it could replace the affected parts. NLI identified that the root cause of the problem was the lack of understanding of the impact of plating hardened parts. The zinc plating process is discussed further in paragraph 3.5.

l NCR-25 was initiated on June 14, 1996, to document that the hardened parts in i

low-voltage circuit breakers had been zinc plated and identified the 4 recipients. NLI evaluated the DR and determined that it did not zinc plate  !

any parts in one of the circuit breakers. NLI recommended the recipients to return the other breakers for replacement of those parts.

l As discussed in Paragraph 3.1 of this report, the definitions in Quality Procedure NLI-QUAL-08, Revision 1, are inconsistent with the definitions in  ;

t 10 CFR Part 21, and does not complement other NLI procedures. The NLI QA l Manager agreed to correct this when he revises the procedure.

3.4 Commercial Grade Dedication

( a. Inspection Scoce The inspectors reviewed selected portions and verified the implementation of the NLI Quality Assurance (QA) Manual, the current implementing procedures and instructions for the QA, Engineering, Procurement, and Service Departments, and several Shop Test Procedures related to NLI's dedication of commercial-grade items.

4 4

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b. Observations and Findinas The inspector reviewed the dedication activities (in progress) for several 480-Vac reversing motor starters, Project 059-009, for Vermont Yankee. Reversing motor starters consist of two sets of solenoid-operated line contactors, overload relays, auxiliary switches, and control contacts for automatic or remote operation.

The main line (480-Vac) contactors are connected in different phase sequences to effect phase (and hence motor) rotation reversal. The overload relays are thermal-inverse-time devices that act to open the main contactors by deenergizing their operating coils upon sensing sustained current overloads. The main contact operating solenoids and other control components typically operate at 120 Vac or 125 Vdc (as do the ones being dedicated). Reversing starters of the type and size being dedicated by NLI are typically used to provide control and sustained overload protection for Limitorque (or other) motorized valve actuators. Also installed in each motor control center (MCC) cubicle or " bucket" with the starters are molded-case circuit breakers, typically, as in this case, of a type with an instantaneous, magnetic-only overload trip function. Called

" motor circuit protectors" (MCPS) by Westinghouse, they are for protection of the circuits from faults or short circuits by tripping

" instantaneously" (meaning no intentional time delay) while allowing for momentary high current surges (such as starting current) without tripping.

NLI Purchase Order Documentation File (P0DF) 059-009 for the Westinghouse and I-T-E type MCC buckets being dedicated for Vermont Yankee included verification plans (VPs) and test data sheets for the individual MCC bucket components including the MCPS, starters, I and the overload relays and heater elements used in the overload I

relays. Attachment E te the P00F was VP-059-009-5, Revision 1, April 2, 1996, for Westinghouse 600-Vac-rated, 2- and 3-pole HMCPs (MCPS with higher interrupting capacity).

The inspector noted that the specified voltage for the dielectric withstand test was the customary twice rated voltage plus 1000 volts (totalling 2200 Vac in this case). However according to the test record, only a 500-Vdc megger check of insulation resistance was performed and documented. NL1 explained that Vermont Yankee had asked that the megger check be done instead of the usual dielectric withstand test. However, the inspectors found no documented evidence that Vermont Yankee requested NLI to perform the 500-Vdc megger test in lieu of the standard test.

In addition, the instructions in the VP for mfggering were as follows: "With breaker open, megger across each pole, with breaker closed, megger each pole line side to load side." The inspectors determined that these instructions were incorrect for the following reasons: (1) meggering or measuring insulation resistance with a "Megger" (registered trademark of Biddle Instruments, Inc.) "across" a pole implies measuring line to load side which was the second 5

128

step, (2) if the breaker is closed, meggering from line to load is useless, as the resistance (as measured by a megger) will be zero, (3) there were no instructions for meggering from phase to phase l

(line to line) or from each phase (pole) to ground as is normally l required by MCCB manufacturers' instructions and industry standards such as National Electrical Manufacturers Association (NEMA)

Standard AB 4-1991. AB 4-1991 also recommends meggering 600-Vac rated MCCBs at 1000 Vdc instead of at the 500 Vdc reportedly l requested by Vermont Yankee (which is much less than the peak l voltage that the MCCB insulation may have to withstand in service).

NLI explained that this VP had been developed using Vermont Yankee's standard MCCB test procedure, VYOPF 5210.04, (OP 5210, Revision 6) which, the inspector noted, also lacked the phase-to-ground l measurement.

The inspectors reviewed NLI's Standard Verification Plan (SVP)-1, l Revision 2, January 1996, and test data taken for a Westinghouse type reversing starter, and determined that it lacked insulation resistance data for one of the two sets of main line contactors.

Further, the data collected was not useable because it did not identify if it was for the forward or the reverse contactor.

VP 059-009-0T,, for MCC bucket final assembly testing (after component interconnection) specified meggering power (480-Vac) circuits to ground, power to control circuits, and control circuits to ground. Since both sets of contactors can not be closed at the same time and without energizing the contactors or manually closing  !

them, it would require a minimum of six data points for each j contactor to measure the power circuits to ground: phases A, B, and C, line side to ground, and phases A, B, and C, load side to ground.

However, the inspectors found only one value recorded. When the inspectors enquired about this discrepancy, the technicians who performed the testing on the starters and completed buckets stated i

that they had meggered both sides of both sets of contactors.

After finding that they all read the same high values, they recorded only one set of data. The QA manager agreed with the inspectors that the QA documentation was inadequate because it did not provide objective quality evidence that the critical characteristic of insulation resistance was acceptable. The QA manager informed the project engineer the as-documented test results represented incomplete test data and as such could not be used in the dedication.

VP 059-009-6, Revision 1, April 17, 1996, prescribed the testing for Westinghouse Type A series overload relays with FH-type heaters.

During initial testing of some of the overload relay heater l elements, NLI used Westinghouse rated full-load and trip time data l corresponding to the heater size for the installed heater. However,

! when Vermont Yankee later provided plant motor full-load amps and valve stroke time data for each bucket designated with its plant ID, the data indicated that some of the overload relays may not have had the correct heater installed. For example, the size FH29 element 6

129

was installed in the overload relay in bucket No. ' 98-4C. According to the 300-percent current / time characteristics for 3-phase relays with two or three H or FH type heaters in Table 3 of VY OP 5210, Rev. 6,: Attachment 2 to the VP, the expected / rated full-load amperes for two size FHP9 heaters were 3.61, and 3.12 amperes for three heaters. However, according to Attachment 1, the plant motor full-load and stroke time data, bucket 98-4C would see 4.0 amperes. This means. that, exposed to greater than their nominal rated full. load, FH29 heaters may cause the overload relay to begin to time out instead of holding in for the required' time. This condition should be detected or verified acceptable by the full-load hold-in test specified by the VP, which calls for the relays to hold in at motor full-load amps for up to three times the associated MOV stroke time. 1 NLI agreed to verify that all heaters would be checked against plant parameters (including those that were initially tested using only nominal 300-percent overload data), and confirm that the correct heaters were installed for the intended plant load.

Finally, the inspector's review of.VP 023-061-1, SVP-22, for shunt trip devices for circuit breakers for C&D Charter Power and Florida Power Corporation's Crystal River Nuclear Plant, revealed that the shunt trips were tested at only nominal rated voltage (125 Vdc).

This test does not verify that the shunt trips will perform their safety function under degraded voltage conditions (possibly as low as 90 Vdc), or that they meet the manufacturer's specification for minimum trip voltage (typically about 70 percent of rated).

Although a less likely occurrence, being called upon to trip without damage at the highest expected 125-Vdc vital bus voltage (possibly as high as 145 Vdc during an equalizer battery charge) would also not be verified when only nominal voltage is applied.

c. Conclusions

, The inspectors found that NLI's dedication procedures failed to I

specify appropriate technical acceptance criteria and test records failed to demonstrate that acceptance criteria had been met to verify that certain commercial grade components would perform their safety functions and that.they met the requirements of the procurement documents. The inspectors _ informed NLI that this is contrary to Criterion VII, " Control of Purchased Material, j Equipment, and Services," of Appendix B to 10 CFR Part 50 and j identified it as Nonconformance 99901298/96-01-01, 3.5 Refurbishina 4.16-kV Circuit Breakers.

a. SCoDe l The inspectors reviewed Procedure NLI-TECH-P101, " Remanufacture of General Electric (GE) 4.16-kV Magne Blast Breakers," Revision 6, dated June 1996, and its implementation and reviewed the adequacy of <

[ NLI's corrective action for adverse findings documented by Salem i

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! Generating Station's-(Salem) personnel in a trip report dated March I i 16, 1996. '

i i b. Observations and Findinas )

Salem contracted NLI to refurbish GE 4.16-kV Magne-Blast type i circuit breakers. NLI subcontracted the refurbishing activities to j

National Switchgear Services. (NSS) in Lewisville, Texas. NSS performed the refurbishing work implementing NLI's quality assurance  :

program and procedures. :NLI Procedure NLI-TECH-P101 was developed

in 1995 and revised frequently to include additional. inspection i requirements'and comments from Salem's personnel. Paragraph 2.0 of j this procedure lists the applicable GE Service Advice Letters (SALs)

and Service Information Letters (SILs). The inspectors observed  ;

, that the procedure did not list five GE SALs issued by the GE '

i Specialty Breaker Plant (SBP), the original manufacturer of GE i' circuit breakers. NLI personnel informed the inspectors that they >

3 would take adequate corrective action to revise Procedure NLI-TECH- )

! P101 to add the five GE SALs. i l

l1 On March 16, 1996, four Salem technical representatives inspected three refurbished 4.16 kV breakers and documented seven adverse findings in a trip report. They identified 1) significant misalignment in the area of the prop and prop pins, 2) a weld crack on the frame of a breaker which was repaired, 3) broken buffer

1. blocks in the primary element moveable contacts on one breaker, 4) 2 the release latch pawl rotation was impeded by the protrusion of an

auxiliary switch mounting bolt, 5) the alignment of the primary bushings could not be verified, 6) improper alignment of the bushing 2

rods at the pivot points between phases, 7) the braided wire 4

connecting bolts in the stationary contacts for the primary stabs

~

were found to be loose, and 8) generally poor workmanship relative to the proper placement of prop springs, beveled spring brackets and lubrication. The Salem personnel did not approve the shipment of these three breakers. The inspectors reviewed NLI's action taken to correct these findings and determined that NLI took adequate corrective actions by revising Procedure NLI-Tech-P101 to detect and correct problems with alignment, broken buffer blocks, and general workmanship, and borrowed the design for a template from Salem to correctly align primary bushings.

In May 1996, during receipt inspection of refurbished breakers returned from NSS, Salem technicians observed broken lock washers in the arcchute holddown assembly. After investigation, Salem -

personnel determined that NSS, to enhance the appearance of some metal parts, had zinc plated the arcchute hold down assembly. 1 Before zinc plating the assembly, NSS did not mask or remove the lock washers. During the zinc plating process, hydrogen may have diffued into the lockwashers which had been made from hardened steel and caused hydrogen embrittlement. During the preparation of Procedure NLI-TECH-P101, NSS had not considered the adverse effects of zinc plating hardened steel components and overlooked remedial i

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

L measures, such as baking the parts after zinc plating to expel the l hydrogen. Salem had reviewed and approved this procedure and its subsequent revisions. After the problem was identified, provisions in Section 10 which required all steel parts to be zinc plated were deleted from the procedure. NLI generated Nonconformances to identify zinc-plated hardened steel parts in other switchgear and these are discussed in paragraph 3.3. This matter is not being identified as a nonconformance because technicians at Salem identified this problem, and NLI had implemented corrective actions prior to the NRC inspection.

c. Conclusion The LLI/NSS is taking appropriate actions to correct weaknesses exhibited in Procedure NLI-TECH-P101.

( 3.6 Traceability of GE AKR Circuit Breakers

a. Inspection Scope The inspectors reviewed NLI Quality Procedure NLI-QUAL-09,

" Identification, Handling, Storage, and Shipping of Material,"

Revision 8, dated July 1996, which established the requirements for the identification, handling, storage, and shipping of safety-

related items for NRC licensees. The inspectors verified implementation of the requirements by reviewing NLI activities associated with the supply of 6 safety related GE AKR breakers to Entergy Operations, Inc. (Entergy). The breakers were purchased as commercial-grade items by NLI and upgraded to safety-related through NLI's commercial grade dedication program.

b. Observations and Findinas l During an inspection of the work in progress in NLI's facility the NRC inspectors observed a General Electric (GE) AKR circuit breaker with two nameplates. One was the original GE nameplate without the GE serial number. The second was a National Switchgear Service (NSS) nameplate with a NLI-assigned serial number. NSS is NLl's subvendor which refurbishes low- and medium-voltage circuit breakers. NLI informed the NRC inspectors that it had originally supplied the GE AKR-50 circuit breaker, as safety-related, to l Entergy, for use at its River Bend station, and that the breaker was l currently at NLI for additional work. NLI stated that they had removed the GE serial number and assigned an NLI serial number to i

track the circuit breaker. NLI had generated documentation to i maintain complete traceability of the circuit breaker through the NLI serial number to GE. In addition, NLI indicated that they had l

sold five additional GE AKR-30 circuit breakers to Entergy, also for i use at River Bend, as new, safety-related, with two nameplates - one NSS nameplate with the NLI-assigned serial numbers and one GE

nameplate deleting the GE serial number. NLI indicated that these six circuit breakers were the only circuit breakers that NLI had 9

132 l

sold to a customer after removing the manufacturer's serial numbers from the nameplates.

l NLI-QUAL-09, states in Section 2.0, " Identification and Control of Material," that it is NLI's responsibility to maintain traceability for items which are procured commercial grade and dedicated for safety-related uses.

To verify that NLI had maintained traceability for the six GE AKR circuit breakers, the inspectors reviewed P00F-052-013, and.

associated test documentation. The documents indicate that NLI purchased six GE AKR circuit breakers (five GE AKR-30s and one GE

' AKR-50) from National Switchgear Systems, Inc. (NSS) who had obtained them through Powell Industries, Inc. (Powell), a distributor. NSS, under NLI's QA provisions, had performed all the work and testing, associated with both customer P0 requirements and dedication, on the GE AKR circuit breakers supplied to River Bend.

Upon receipt of the circuit breakers, NSS installed National Switchgear nameplates with NLI-assigned serial numbers on them. NSS removed the serial numbers inscribed on the GE nameplates. The NLI certificates of conformance provided to Entergy listed the NLI serial numbers in lieu of the GE serial numbers.

l For the five GE AKR-30 circuit breakers, NLI had all the P0s on file, including the summary sheets originally packed with the i

! circuit breakers, which listed the GE serial numbers of the circuit breakers received and the Powell P0 number; the NSS test data sheets which listed the GE serial numbers, the NSS Job Number which related l

to the NSS P0, and the associated test data; all GE nameplate information; and the NLI certificates of conformance listing the NLI serial numbers.

The inspectors reviewed the associated purchase dour >ntation for the single GE AKR-50 circuit breaker supplied to River Bend.

l Documents indicated that with P0 95-G-72336, Entergy purchased the l

single GE AKR-50 circuit breaker from NLI as safety-related; NLI had purchased the circuit breaker from NSS as commercial grade with P0 1618; NSS had purchased the circuit breaker from the distributor as commercial grade with P0 7958; and the distributor had purchased the circuit breaker from GE as commercial grade with its P0104233.

For the single, GE AKR-50 circuit breaker, NLI had the P0s from Entergy to NLI, NLI to NSS, and NSS to the distributor on file, however, NLI did not have the P0 from the distributor to GE on file.

NLI was only able to produce the P0 after obtaining it from the distributor, following completion of the inspection. NLI had the summary sheet packed with the circuit breaker, which listed the GE serial number of the circuit breaker received and the distributor PO number; all GE nameplate information; anr, the NLI certificate of conformance listing the NLI serial number. NLI also had on file the NSS test data sheet which listed the NSS Job Number which related to the NSS P0, and that associated test data, however the sheet did not list the GE serial number but instead listed the NLI serial number.

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The inspsctors determined that NLI actions were not adequate in identifying the circuit breaker with its associated test data sheets. However, based upon the additional information provided by the P0s, summary sheets, and recorded nameplate information, and the fact that there was only one GE AKR-50 circuit breaker on the order, the inspectors concluded that the circuit breaker test sheet was applicable to the GE serial number on the associated documentation and therefore did not constitute a safety concern.

c. Conclusion The inspectors concluded that NLI had on file adequate information to demonstrate traceability for the five GE AKR-30 circuit breakers, GE the manufacturer.

The inspectors concluded that without the distributor's P0 to GE, and without the GE serial number on the applicable test data sheet, NLI inadequately documented traceability of the GE AKR-50 circuit breaker to GE, the circuit breaker manufacturer. The inspectors informed NLI management that tha inadequate documentation of traceability was contrary to Criterion VIII, " Identification and Control of Materials, Parts, and Components," of 10 CFR Part 50, Appendix B, and identified it as Nonconformance 99901298/96-01-02.

3.7 Control of Material

a. Inspection Scope The NRC inspectors reviewed and verified implementation of NLI's Quality Procedure NLI-QUAL-09, " Identification, Handling, Storage, and Shipping of Material," Revision 8, dated July 1996, which established the requirements for the identification, handling,

, storage, and shipping of safety-related items for NRC licensees.

b. Observations and Findinas Except for the GE breakers discussed above, NLI's preferred method of material identification was use of the manufacturer's identifier.

However, since many items do not have a unique identifier applied by the manufacturer, NLI assigned an NLI serial number or part number (for material or small components). NLI's identification number consisted of the PODF number followed by a component type designator (such as "BR" for circuit he d er) and a sequential number [(P00F#)-

(component type)-(sequential number)]. PODFs were either associated with a particular licensee purchase order number or a generic NLI PODF number which was used for inventoried items such as hardware.

NLI identified material as part of the material receiving process.

NLI subsequently controlled material by use of NLI serial numbers, material control tags, and material travellers, and by maintaining material in a segregated material storage room.

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The NRC inspectors Cbserved the application of the NLI l

identification system in the NLI nuclear material storage room, a locked, controlled room which contained received material and components in various stages of preparation for use. Specific areas l in the room included incoming items which had not been receipt inspected, items which had been receipt inspected but on which the dedication process had not begun, items which had been receipt inspected and were currently in the dedication or qualification i process, items which had been verified to meet any dedication or l qualification requirements and were ready for use, items on hold

! (deficient material), and test specimens from completed projects.

NLI had items in each of the described areas.

NLI maintained numerous bins containing various fasteners which had been dedicated for safety-related use; these items were identified with generic NLI PODFs used to control inventory. NLI had a smaller inventory of material staged for use in particular jobs. The staged  ;

items had an NLI identifier using a licensee P0DF as opposed to the  !

l generic NLI PODF used for the inventoried items. In addition, items l purchased for licensee jobs but not used (excess material) were identified by the original licensee PODF. The inspectors examined items stored in the various areas, numerous individual items contained in larger cartons, and numerous inventoried fastener bins and did not find any items had not been properly marked and located.

The inspectors observed that the material storage room was properly segregated and access to it was adequately controlled by means of the locked single entrance and that the material storage room was

very well organized with all items clearly labeled. Discussion with l the NLI personnel indicated that the material remained in the room until use without any casual removal and that no material left the premises. In addition, NLI stated that a dedicated truck (a contracted truck which goes directly between pickup and delivery with no stops and no other customer loads on board) had been used to move controlled equipment and material since the establishment of NLI.

c. Conclusion

The inspectors concluded that with the exception of the '

nonconformances identified in paragraphs 3.4 and 3.6, NLI had

! established adequate measures to identify and control material and components and was effectively implementing the measures.

3.8 Confiauration fontrol

a. Inspection Scope The NRC inspectors reviewed NLI's QA requirements and their implementation for technical document preparation and data collection for safety-related activities. Section 5.0, 12 l

135

l

" Configuration Control," of the NLI Quality Assurance Manual, Revision 1, dat(d July 1991, established the NLI requirements for configuration control, which acompassed data collection and documentation preparation for safety related activities.

b. Observations and Findinos NLI developed a Project Performance Plan for all safety-related projects which included Quality Plans, Technical Plans, Interface Plans, Training Logs, and Document Logs. The Project Performance Plan was considered a safety-related document and was accordingly maintained as a quality assurance record in accordance with NLI QA requirements. The Technical Plan defined the methods used to accomplish the project activities including the technical approach, assumptions, limitations, and regulatory requirements, and was the only document to provide technical guidance to project personnel. -

Document Logs were used to control project input documents (drawing, specifications, and calculations) and output documents (NLI prepared .

documents including reports, commercial grade surveys, test plans, and inspection plans). The Quality Plan defined the QA requirements associated with the project. The Interface Plan provided guidance on communication between NLI and the customer. The Training Logs documented project specific training.

NLI safety-related engineering documentation included Technical Reports, Technical Calculations, and Design Specifications.

Technical Reports included operation, testing, and maintenance procedures, design basis documents, safety evaluations, equipment descriptions, and testing and inspection plans and reports.

Technical calculations were directed by the Project Engineer who was responsible for assuring that project personnel had appropriate qualifications. NLI assigned an NLI tracking number to each technical calculation and performed independent verification of all technical calculations. Design specifications and drawings were developed by personnel selected by the Project engineer and also subject to independent verification. NLI's preferred method of independent verification was a design review performed in accordance with written criteria including inputs, assumptions, methodology, calculations, computer runs, results, and conclusions. In addition, NLI typically obtained customer approval of test documents including criteria and specification.

The NRC inspectors reviewed test documentation supporting several projects recently accomplished by NLI. The review included P0DF-03719 for Siemens HE3-B100 circuit breakers ordered by Florida Power and Light Company with P0 00009415, P0DF-052-019 for refurbishment of a GE AM-4.16-250-9H circuit breaker ordered by Entergy with P0 95-J-73166, and P0DF-052-013 for GE AKR-30 circuit breakers ordered by Entergy with P0 95-G-72336.

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c. Conclusions The inspectors concluded that with the exception of the nonconformances identified in paragraphs 3.4 and 3.6, NLI adequately implemented the established procedures and that the test documentation reviewed was well organized, clearly documented, and complete, with the appropriate approvals.

4 ENTRANCE AND EXIT MEETINGS In the entrance meeting on July 8, 1996, the NRC inspectors discussed the scope of the inspection, outlined the areas to be inspected, and established interfaces with NLI management. In the exit meeting on July 11, 1996, the inspectors discussed their findings and concerns.

PERSONS CONTACTED A. Seiken, President A. Bell, QA Manager M. Estrada, Engineering Manager 14 137

Selected Generic Correspondence on the Adequacy of Vendor Audits and the Quality of Vendor Products Identifier Title Information Notice 96-40 Deficiencies in Material Dedication and Procurement Practices and in Audits of Vendors Information Notice 96-43 Failure of General Electric Magne-Blast Circuit Breakers Information Notice 96-44 Failure of Reactor Trip Breaker from Cracking of Phenolic Material in Secondary Contact Assembly Information Notice 96-46 Zinc Plating of Hardened Metal Parts and Removal of Protective Coatings in Refurbished Circuit Breakers Information Notice 96-50 Problems with Levering-In Devices in Westinghouse Circuit Breakers Contact Assembly I

i l

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I U.S. NUCLE AR REGUL ATORY COMMIS$lON 1. REPORT NUMBER NRC FOAM 335 umbers $n nc 1102.

sm. 22e BIBUOGRAPHIC DATA SHEET NUREG-0040 isa ameructions on ene narwi yn3, 79, yn, 3 l 2. TITLE AND SUBilTLE Licensee Contractor and Vendor Inspection Status Report 3. DATE REPORT PUBLISHED Quarterly Report == r a j vtaa July - Septermber January 1997

4. FIN OR GRANT NUMBER

b. AUTHOR (S) 6. TYPE OF REPORT Quarterly

7. PE RI00 COV E R ED unctus,w peresp July - September 8, PE R F 0RMlNG ORGAN lZAT SON - N AME AND ADDR ESS 499 NRC. pmunk Dwasion. O!!oce or Region, U.S Nuckar Regulatory Commouron. emt maihne aMrou;19 eontractor. omvrer none omt methne addowul Division of Inspection and Support Programs Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 9 SPONSOR;NG oRGANIZATloN - N AM E AND ADDRESS ut NRc. tvoe *1;nme es abon"; ot contractor pmvide NRC Orvisoon. Onu:e or Region. us. Nucker Ragusatory Commsuen.

and meeting eddrent Same as above

10. SUPPLEMENTARY NOTES

11. ABSTRACT (200 were or kai i This periodical covers the results of inspections performed by the NRC's Special l Inspection Branch, Vendor Inspection Section, that have been distributed to the I inspected organizations during the period from July through September 1996. (

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13. AVAILAN,ui v st AT EMENT

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Printed on recycled paper Federal Recycling Program

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